ML20082J662

Regulatory and Technical Reports (Abstract Index Journal). Annual Compilation for 1994

ML20082J662
Person / Time
Issue date:	03/31/1995
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V19-N04, NUREG-304, NUREG-304-V19-N4, NUDOCS 9504180345
Download: ML20082J662 (133)

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i NUREG-0304 Vol.19, No. 4 1

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l Regulatory and Technical Reports

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l AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications i

Most documents cited in NRC publications will be available from one of the following sources l

1.

The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC l

20555-0001 2.

The Superintendent of Documents, U.S. Government Printing Office, P. O. Box 37082, Washington, DC 20402-9328 3.

The National Technical Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica-l tions, it is not intended to be exhaustive, t

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda: NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports;.

vendor reports and correspondence: Commission papers; and applicant and licensee docu-ments and correspondence.

The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, international agreement reports, grantee reports, and NRC booklets and bro-l chures. Also available are regulatory guides NRC regulations in the Code of Federal Regula-l tions, and Nuclear Regulatory Commission Issuances.

l Documents ava.ilable from the National Technical Information Service include NUREG-series l

reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions. Federal Reg / ster notices. Federal l

and State legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, fnreign reports and translations, and non-NRC con-ference proceedings are available for purchase from the organization sponsoring the publica-tion cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library Two White Flint North.11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308.

l NUREG-0304 Vol.19, No. 4 k

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l Regulatory and Technical Reports j

(Abstract Index Journal) i j

Annual Compilation for 1994 l

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i Date Published: March 1995 e

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i Regulatory Publications Branch Division of Freedom ofInformation and Publications Services j

Office of Administration i

U.S. Nuclear Regulatory Commission j

Washington, DC 20555-0001

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i CONTENTS Preface.......................................................................

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'Index Tab Main Citations and Abstracts........................................................... 1 l

e Staff Reports o Conference Proceedings

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e - Contractor Reports s Grant Reports e international Agreement Reports Secondary Report Number index...................................................... 2 Personal Author index................................................................. 3 Subject index.........................

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NRC Originating Organization index (Staff Reports)....................................... 5 NRC Originating Organization index (International Agreements),.....,.........,........... 6 NRC Contract Sponsor index (Contractor Reports) 7 Contractor index.....................................................................

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Intemational Organization index........................................................ g Licensed Facility index............................................................... - 10 1

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PREFACE This compilation consists of bibliographic 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 appreciated. Please send them to:

Technical Publications Section Publications Branch Division of Freedom of Information l

and Publications Services T-6 E7 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 -0001 l

TL9 main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUlEG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indexes:

Secondary Report Number index Personal Authorindex Subject index NRC Originating Organization index (Staff Reports)

NRC Originating Organization index (International Agreements)

NRC Contract Sponsor index (Contractor Reports)

Contractor index International Organization Index Licensed Facility index A detailed explanation of the entries precedes each index.

The bibliographic elements of the main citations are the following:

Staff Report NUREG-0808 MARK ll 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 Document Control 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.

j 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 pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use).

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Contractor Repu,t NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REAC-TORS-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 number, (2) report title, (3) report authors, (4) organ!zational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control Sys-tem accession number, (8) the report nun her of the originating organization (if given), (9) the microfiche ad-dress (for NRC intemal use).

Grant Report NUREG/GR-0013: APPLICATIONS OF A NEW MAGNETIC MONITORING TECHNIQUE TO IN SITU EVALJA-TION OF FATlQUE DAMAGE IN FERROUS COMPONENTS. JILES, D.C.; BINER, S.B.; GOVINDARAJU, M.; et t

al. lowa State Univ., Ames, IA.. lune 1994 41 pp. 9407250286. 80328:195.

Where the entries are(1) report nt mber, (2) sc,nort title (3) report authors, (4) organizational unit of authors or i

publisher, (5) date report was published, (6) numbc of pc ges in the report, (7) the NRC Document Control Sys-tem accession number, (8) the report number of the originating organization (if given), (9) the microfiche ad-dress (for NRC intemal use).

International Agreement Report NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA.

NEUMANN, U. Kraftweek 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 organdbo (if given), and (9) the microfiche address (for NRC Internal use).

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

ADD

- addendum APP

- appendix DRFT - draft ERR

- errata I

N - number R - revision l

S - supplement j

V - volume l

Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the NationalTechnical 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:

Superintendent of Documents U.S. Government 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) 512-2249 or (202) 512-2171. Non-U.S. customers must make payment in advance either by Intemational Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian sank, payable to the Superintendent of Documents.

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NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report.

Co:% dor-prTared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established codes such as ORNLJNUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship or the work being reported.

I in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference pro-ceedings NUREO/GR is used for NRC grant reports, and NUREG/lA is used for intemational agreement reports.

All these report codes are controlled and assigned by the staff of the Technical Publications Section of the NRC Division of Freedom of Information and Publications Services.

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io Main Citations and Abstracts

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The report listings,in this compilation are arranged by report number, where NUREG-XXXX is 3:

an NRC staff nated report, NUREG/CP-XXXX is an NRC-sponsored conference report, 2

i NUREG/CR-is an NRC contractor-preparedreport, and NUREG/lA-XXXX is an inter-

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i national agreement re: port. The bibliographic informaton (see Preface for details) is followed by a brief abstract of t1is report.

i NUMG 0000 Vie: LICENSED OPERATING REACTORS STATUS '

tion, that have boon diserttnsted to the irimpacead orgeruzetions

SUMMARY

REPORT.Dete As Of December 31, 1993.(Gray during the period from July through September 1994.

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Book I) HARTFIELD.R.A. Ofece of informaton Resources Man- '

1 agement (Post 800205). March 1994. 360pp. 9404110367.

NUMG 0000 Vie NOS: REPORT TO NSS ON AMOR-j 7eg1g.014 MAL OCCURRENCES. July-September 1993.

• Office for Analy -

The Nuciew Reguidory Commission's annual summary of E-sis & Evaluation of Operemonal Dete, Director March 1994.

j coneed nuclew power reactor date le bened primedy on #w 31pp. 9406060338. 79182 316.

l report of operating data submitted by Econeses tw each unit for Secton 208 of the Energy Reorganiraton Act of 1974 identi.

the month of Decembw har=== that report contains date for nu en abnormel occurmnce as an unscheduled incident or the month of December, the year to date (in this case calender emnt that the Nuclear Reguistory Cc wnesion dmermines 2 be l

year 1993) and cumuleeve dele, usueNy from the date of com-segnWicent from the standpont of potc heeNh or safety and re-i mercial operaton. The date is not independently vented, but quires a quarterly report of such wents to be made to Con- -

vanous computer checke we made. The report is dMded into grees. This report covers the perm from July 1 through Sep.

}

two sectons. The Arst contains summary higNights and the tomber 30, 1993. This report dismases two abnormel occur-second contems date on each individual unit in commercial op.

'ences et 'NRC-Hoensed facittles. W irwohed a mecNcal weton. Secton 1 cepecNy and availabihty factors are simple sodium iodide meedministraton and one lnvolved a 1981 fatal 3

enthmenc avere0es Secton 2 items in the cumulatwo column redletion exposure of a redlagrapher. One induolnel redlo-are generally as reponed by the beenees and notes as to the grapher overeuposure event and four medical smeedministre-j use of weighted averages and stening dates other then com-tions that were reported by the 4_

States are eleo dio-i rneresel operemon are provided.

cussed, bened on information provided by the A0reement States as of November 1,1993. The report also contams infor-00UREG 0040 V17 N04: LICENSEE CONTRACTOR AND motion updeeng four p t reported abnormal occurrences VENDOR INSPECTION STATUS REPORT. Quarterly at NRC-Ilconsed faceitos and three reported by the Agreement i

j, Report, October-December 1993.(White Book)

• Division of Re-States, and includes informaton on two other events of interest.

i actor tr=f*ri & Liconese Performance (Post 921004). Febru.

Appendix D has been added to this report which includes i

ary 1994. 80pp. 9403150242. 78501288.

events submitted by.V_ua States that are likely to be cat-j This perlocacei covers the results of inopochone portormed by egonzed as obnormal occurrences. For these events, insuffi :

the NRC's Vendor inspecton Branch that have been chstnbuted cient informaton was aweilable as of November 1.1993, to j

to the inspected orgemzations during the period from October poeltrwely identfy them as obnormal occurrences.

1 December a93.

l NUREG0080 Vis esse: REPORT TO CONGRESS ON ABNOR-NUREGco40 V18 N01: LICENSEE CONTRACTOR AND MAL OCCURRENCES. October-December 1993.

• Othoe for

(

~ VENDOR INSPECTION STATUS-REPORT. Quenerly Analyse & Evolustion of Operatonal Dese, Director. Aprd 1994.

t Report. January-March 1994.(White Book)

• Drvoson of Reactor 43pp. 9406310231. 79689:191.

I inspecton & Licensee Performance (Post 921004). April 1994.

Secton 208 of the Energy Reorgentzetion Act of 1974 identi- '

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250pp. 9405310289. 79587:032.

fios an abnormel occurrence as an unscheduled incident or This perweral covers the resurts of inspectont performed by event that the Nuclear Regulatory Commlesion determmes to be 3

1 the NRC's Vendor inspecton Branch that have been distnbuted sagraficent from the standpoint of pubile health or safety and re-j to the inspected organizations during the period from January quires a quarterly report of such events to be made to Con-f through March 1994.

grees. This report covers the period from October 1 through De-i

]

cember 31, 1993. This report discusses six abnormal occur-j NUREG 0040 V18 N02: LICENSEE CONTRACTOR AND rences at NRC-hcensed facudies. Five invo6ved medical bre-VENDOR INSPECTION STATUS FIEPORT. Quarterly chytherapy nusedmmistratons, and one involved an overeupo-4 Report.Apni-June 1994.(White Book)

• Division of Reactor in-sure to a nunang infant. Seven abnormal occurrences that were specton & Licensee Performance (Post 921004). October 1994.

reported by the Agreement States are eleo diarumaad bened on i

183pp. 9411290139. 81878:220.

information provided by the 41 J States as of February 8

This penodical covers the results of ir.soectons performed by 28,1994. Of these events, three swolved brachytherapy mined-l the NRC's Vendor inspecton Branch that have been distributed rnirustrations, one involved a teletherapy meedmmistraton, one 1

to the inspected organizatons during the period forH through involved a theft of radioactwo motorial during trenoport and im-j June 1994.

proper d=pr=al, and two involved lost sources j

NUREG 0040 V18 0003: LICENSEE CONTRACTOR AND NUREG 0080 V17 Net: REPORT TO CONGRESS ON ABNOR.

VENDOR INSPECTION STATUS REPORT. Quarterly MAL OCCURRENCES.Jenuary44erch 1994.

• Office for Analy-i Report.JulySeptember 1994.(White Book)

• Office of Nuclear ais & Evaluation of Operemonal Dete, Director, August 1994.

2 Reactor Regulaton, Director (Post 870411). November 1994.

33pp. 9400000134. 80817:073.

3 123pp. 9412300175. 82158:001.

Secton 208 of the Energy Reorgeruzeton Act of 1974 identi-This portodical covers the results of inspections performed by fios an abnormal occurrence (AO) as an unscheduled incident 4

t the NRC's 81parsal inspection Branch, Vendor inspection Sec-or event that the Nuclear Regulatory Commiseson determines to 1

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2 Main Citations and Abstracts be sigrufcant from the standpoint of pubic health or safety and NUREG-0304 V19 N02: REGULATORY AND TECHNICAL RE-requires a quarterfy report of such events to be made to Con-PORTS (ABSTRACT INDEX JOURNAL). Compilation For gress. This report provides a desenption of those events that Second Quarter 1994, April-June.

• Division of Freedom of Infor-have been determined to be abnormal occurrences during the mation & Publications Sstvices (Post 940714). September 1994.

period of January 1 through March 31, 1994 This report ad-46pp.9410130043.8126&OO1.

dresses seven AOs at NRC-licensed facilities. One involved in.

See NUREG-0304,V18 NW sbstract.

operable main steam isolation valves at a toiling water reactor, four involved medical brachytherapy misadninistration, one in.

NUREG 0325 R17: UA N90W.A3 REGULATORY COMMISSION ORGANIZATION

(" iR1s AND FUNCTIONAL volved a meJical teletherapy misadministra' ion, and or,e in.

volved four bst reference sources. One AO that was reported STATEMENTS.Octd;er J,10h Ofc of Personnel (Post tl70413). October iv94. tP2pp. 9411160041. 81758:001.

by an Agreement State is also discussed; the hformation is cur-Fonctional statements and ciganization charts for the U.S.

rent as of Apnl 25,1994. This event involved 1 therapeute ra.

Nuclear Regulatory Conxnission c ffices, divisions, and branches Gushininscoutical msadministration. The repart also contains updates on neven abnormal occurrences pwiously reported by are pmsented NRC hcensres and one abnormal occurrence previously report-NUREG 0383 V01 R17: DIRECTORY OF CERTIFICATES OF ed by an Ayeement State bcensee. For the period January 1 to COMPLIANCE FOR RADIOACTIVE MATERIALS March 31,1994, no new "Other Events of interest" were repat-PACKAGES. Report Of NRC Approved Padages.

• Divuton of ed but an update to a therapeuts misaderustration previously industnal & Medcal Nuclear &!ety (Post 870729). Oc:ober reported as an "Other Event of Interest" is included 1994. 513pp. 9411290132. 81f,75:001.

Ms 6metwy contabs a RW of E WW Padays NUREG-0090 V17 N02: REPORT TO CONGRESS ON ABNOR.

Wolume 4, w cates of Nnphance Wh 2h aM a MAL OCCURRENCES.Aph 1994.

• Office for Analysis &

Report of NRC Approved Quahty Assurance Programs for Ra-Evaluttion of Operational Data Director. October 1994. 41pp.

deoactive Materials Packages (Volume 3). The purpose of this 2130086.82006 M directory is to make available a convenient source of informa-Section 208 of the Energy Reorganization Act of 1974 6denti-tion on Quahty Assurance Programs and Packagings which fies an abnormal occurrence (AO) as an unscheduled incident have been approved by the U.S. Nuclear Regulatory Commis-cr event that the Nuclear Regulatory Commssion determnes to sion. Shipments of radioactive material utikring these packag-be sagraficant from the standpoint of public health or safety and ings must be in accordance with the provisions of 49 CFR requires a quarterty report of such events to be made to Con-t 173.471 and 10 CFR Part 71, as appleable. In satisfying the gress. This report provides a descnption of those events that requirements of Section 71.12, it is the responsibihty of the li-have been determined to be abnormal occurrences during the censees to insure themselves that they have a copy of the cur-penod of Apnl 1 through June 30,1994. This report addresses rent approval and conduct their transportation actrvities in ac-seven abnormal occurrences (AOs) at NRC-Icensed facihties.

cordance with an NRC approved quality assurance program.

Five involved modical brachytherapy mesadministrations, one in-volved a medcal teletherapy misedministration, and one in.

NUREG-0383 V02 R17: DIRECTORY OF CERTIFICATES OF volved a medcal sodium iodub rrusadministration. Four AOs COMPLIANCE FOR RADIOACTIVE MATERIAL S were reported by the Agreement States as of August 3,1994.

PACKAGES.Certifcates Of Comphance.

• Division of Industria.

Two invoived medcal beachyttvary n ;c1 ministrations, one in.

& Medcal Nuclear Safety (Post 870729). October 1994. 583pp.

volved a radation burn received by an irdustnal radiographer, 9411290134. 81877:001, and one involsed a lost well logging source. The report also See NUREG-0383,V01 R17 abstract.

contains updates of seven AOs prewously reported by NRC h-NUREG-0383 V03 R14: DIRECTORY OF CERTIFICATES OF censees and frve AOs previously reported by Agreement State COMPLIANCE FOR RADIOACTIVE MATERIALS hcensees. Three "Other Events of Interest" are also reported.

PACKAGES. Report Of NRC Approved Quality Assurance Pro-One involved a dehberate coverup of an error in a diagnoste grams For Radioactrve Materials Packages.

• Division of Indus-radiopharmaceutcal adtrurustration at an NRC Econsee, one in-trial & Medical Nuclear Safety (Post 870729). October 1994.

volved an Order Suspending License and Demand for informa-31 8187 N tion at an NRC Icensee, and one involved an overexposure of See NUREG-0383,V01,R17 abstract.

an industrial rascy.Wi at an Agreement State licensee.

NMEG-M M N0h M CAL MRT Hm NUREG-0304 V18 N04: REGULATORY AND TECHNICAL RE.

b n &

lys Sta 704

'ay 94 Opp PORTS (ABSTRACT INDEX JOURNAL). Annual Compilation For 1993.

• Drvision of Freedom of information & Pubications 9406200334. 79831:298.

Services (890206 940714). March 1994.125pp. 9405050332.

This report provdes industry with procedures for submrtting 79208:147.

topical reports, guidance on how the U.S. Nuclear Regulatory This journal includes all formal reports in the NUREG series Commission (NRC) will process and respond to topeal report prepared by the NRC staff and contractors, proceedings of con-submissions, and an accounting of all topical reports currently forences and workshops, grants, and international agreement being reviewed by the NRC staff. This report will be published reports. The entries in this compilation are indexed for accesh semiannually.

by title and abstract, secondary report number, personal autror, subject, NRC organization for staff and intemational agree-NUREG-0430 V13: LICENSED FUEL FACILITY STATUS ments, contractor, intemationa; orgaruzation, and licensed facih-REPORT. inventory Difference Data. July 1,1992 June 30, ty.

1993.(Gray Book 11) JOY,0.R. Office of Nuclear Matenal Safety

& Safeguards. February 1994.19pp. 9403150260. 78501:269.

NUREG 0304 V19 N01: REGULATORY AND TECHNICAL RE.

NRC is committed to the penode pubhcation of imensed fuel PORTS (ABSTRACT INDEX JOURNAL). Compilation For First facility inventory difference data, following agency review of the Quarter 1994 January-March.

• Dmsaon of Freedom of informa-information and completion of any related NRC investigat6ons.

tson & Pubications Services (890206-940714). June 1994. 48pp.

Information in this report includes inventory 6fference data for 9407250316. 80328:300.

actrve fuel fabrication facihties possessing more than ono effec.

See NUREG-0304,V18,N04 abstract.

tive kilogram of special nuclear matenat.

Main Citations and Abstracts 3

NUREG-0496 S01 DFC: FINAL ENVIRONMENTAL STATEMENT NUREG-0540 V16 N02: TITLE LIST OF DOCUMENTS MADE RELATED TO THE OPERATION OF WATTS BAR NUCLEAR PUBLICLY AVAILABLE. February 1 28,1994.

• Division of Free-PUNT UNITS 1 AND 2. Draft Report For Comment. Docket Nos.

dom of Informahon & Pubhcations Services (89020&940714).

50-390 And 50-391.(Tennessee Valley Authonty)

• Associate Di-April 1994. 304pp. 9405040142. 79106:001.

rector for Advanced Reactors & License Renewal (ADAR) (Post See NUREG-0540,V15,N11 abstract.

941001). November 1994.185pp. 9412130097. 82007:001.

The Final Environmental Statement (FES) issued in 1978 rep.

NUREG-0540 V16 NO3: TITLE LIST OF DOCUMENTS MADE resents the Nuclear Regulatory Commission's (NRC's) previous PUBLICLY AVAILABLE. March 1-31, 1994.

• Division of Free-environmental review relateri to the operation of Watts Bar Nu-dom of Informaton & Publications Services (890206-940714).

clear Plant (WBN). The purpose of ( a NRC review is to dis.

May 1994. 400pp. 9405310285. 79586:001.

cuss the effects of observed changes in environment and t,,

See NUREG-0540,V15 N11 abstract.

evaluate the changes in environmental impacts that have &

NUREG-0540 V16 N04: TITLE LIST OF DOCUMENTS MADE curred as a result of chang % in the WBN Plant design and pro-PUBUCLY AVAILABLE.Apnl 1-30, 1994.

• Division of Freedom posed methods of operatonb since the last envirorv.iental l

review. A full scope of em,ronmenbi topics he been evaluat-of Informaton & Publications Services (890206-940714). June

)

er', incicding regional dem> graphy, land and water u% meteor.

1994. 325pp. 9406290304. 80013:136.

See NUREG-0540,V15.N11 abstract.

Oogy, terrestnal and aqur. tic ecology, radioicWJ and non-radi-ologictJ impacts on humes and the r.nvironment, socioeco-NUREG-0540 V16 N05: TITLE LIST OF DOCUMENTS MADE nomic impacts, and environm.tal ).cce. The staff concluded PUBLICLY AVAILABLE.May 1 31, 1994.

• Division of Freedom that there are no significant changes in the environmental im-of Informaton & Publicabons Services (Post 940714). July 1994.

p. cts since the NRC 1978 FES OL from changes in plant 325pp. 9407250168. 80327:001.

design, proposed methods of operaton, or changes in the envi-See NUREG-0540,V15,N11 abstract.

ronment. The applicant's preoperational and operational monb toring programs were reviewed and found to be appropriate for NUREG-0540 V16 N06: TITLE LIST OF DOCUMENTS MADE establishing basehne condrbons and ongohg assessments of PUBLICLY AVAILABLE. June 1 30, 1994.

• Division of Freedom environrnental impacts. The staff also conducted an analysis of of Informabon & Publicatons Services (Post 940714). October plant operation with severe accident mitigaton design attema.

1994. 331pp. 9412090047, 81986:001.

tives (SAMDAs) and concluded that none of the SAMDAs, See NUREG-0540,V15,N11 abstract.

beyond the three procedural changes that the applicant commit-NUREG-0540 V16 N07: TITLE LIST OF DOCUMENTS MADE ted to implement, would be cost-beneficial for further mitigat ng enronmWal impacts.

PUBLICLY AVAILABLE. July 1-31, 1994

• Division of Freedom of Information & Publications Services (Post 940714). Novem-NUREG-0525 V02 R02: SAFEGUARDS

SUMMARY

EVENT LIST ber 1994. 320pp. 9412070108. 81958:288.

(SSEL).Janumry 1,

1990 Through December 31, 1993.

See NUREG-0540,V15,N11 abstract.

(24 Operations Branch. July 1994.

^

9 8

47 NUREG-0540 V16 N06: TITLE LIST OF DOCUMENTS MADE The Safeguards Summary Event List provides brief summa-PUELICLY AVAlLABLE. August 1-31, 1994.

• Division of Free-ries of hundreds of safeguards-related events involving nuclear Jom of Information & Publications Services (Post 940714). No.

material or facihties regulated by the U.S. Nuclear Regulatc,y vember M 3Mpp. N2070105. 8457:260.

Commissicn. Events are desenbed under the categories: Bomb See NUREG-0540,V15,N11 abstract.

related, intrusion, Missing / Allegedly Stolen, Transportabordrelat-NUREG-0540 V16 N09: TITLE LIST OF DOCUMENTS MADE ed Tamper ng/Vandahsm, Arson, Firearms-related, Radiological PUBLICLY AVAILABLE. September 1 30, 1994.

• Division of Sabotage, Norvradclogical Sabotage, and Miscellaneous. Be-Freedom of information & Publications Services (Post 940714).

cause of the pubhc interest, the Miscellarwous category also in-November 1994. 322pp. 9412300185, 82159:321.

ciudes events reported involving source r.terial, byproduct ma-See NUREG-0540,V15,N11 abstract.

terial, and natural uranium, which arc exenpt from safeguards requirements. Informaton in the event dr.i.,,,.Lis was ob.

NUREG-05a0 V16 N10: TITLE LIST OF DOCUMENTS MADE tained from offcial NRC sources.

PUBLICLY AVAIMBLE. October 1-31, 1994.

• Division of Free-NUREG-0540 V15 N11: TITLE LIST OF DOCUMENTS MADE dom of information & Pubhcations Services (Post 940714). De-cember 1994. 321pp. 9501180150. 82346:001.

PUBLICLY AVAILABLE. November 1 30, 1993.

• Drvision of See NUREG-0540,V15,N11 abstract.

Freedom of information & Pubhcatons Services (890206-940714). January 1994. 251pp. 9402150231. 78121:163.

NUREG-0711: HUMAN FACTORS ENGINEERING PROGRAM j

This dowment is a monthly pubhcahon contairung descrip-REVIEW MODEL O'HAHA.J.M.; HIGGINS.J.C.; STUBLER,W.F.;

tions of informaton recetved and generated by the U.S. Nuclear et al. Brookhaven Natiorel Laboratory. July 1994. 97pp.

Regulatory Commession (NRC). This information includes (1) 9408220036.80637:001, docketed matenal associated with crvihan nuclear power plants The staff of the Nuclear 8s Watory Commission is performing and other uses of radioactive materials, and (2) nondocketed nuclear power plant desirp.Sftsfcation reviews based on a material received and generated by NRC pertinent to its role as design process plan that describes the human factors engineer-a regulatory agency. The following indexes are included: Per.

ing (HFE) program elements that are necessary and sufficient to sonal Author, Corporate Source, Report Number, and Cross develop an acceptable detailed design specifcation and an ac-Reference of Enclosures to Pnncipal Documents.

ceptsSle implemented design. There are two principal reasons NUREG 0540 V15 N12: TITLE LIST OF DOCUMENTS MADE

• "#8 PUBLICLY AVAILABLE. December 1 31, 1993

• Divison of 0""*" *

Freedom of Informaton & Publications Sewes (890206 940714). February 1994. 371pp. 9404010178. 78719:033-

"*D

• See NUREG-0540,V15,N11 abstract.

factors issues arise early in the design process, review of the design process actnrities and results is important to the evalua-NUREG-0540 V16 N01: TITLE LIST OF DOCUMENTS MADE tion of an overal' design. However, current regulatons and guid-PUBLICLY AVAILABLE. January 1 31, 1994.

• Drvision of Free-ance documenti do not address the criteria for design process dom of information & Pubhcanns Services (890206-940714).

review. Therefo e, the HFE Program Rea:w *,.",W tHFE PRM)

March 1994. 288pp. 9404010174. 78718:109 was developer. as a basis for im,6,n y design eerbfcaton re-See NUREG-0540,V15,N11 abshact.

views that incude desagn process evaluations n well as review

m -

l i

4-

Main Citations and Alistracts of 9

e final deelgn. A cordral tenet of the HFE PRM is that the NUREG4700 V30 NOS: NUCLEAR REGULATORY COMMISSION l

HFE espects of the plant should be devaaapad designed, and ISSUANCES FOR MARCH 1994.Peges 91186.

• Division of I

evalueled on the beels of a structured top <fown system analysis Freedom. of IrWormation & Pid*anMs Services (800208-using eccepted HFE pnnoiples. The HFE PRM consists of ten 940714). May 1994.105pp. 9406200339. 79831:001.

component elements. Each element is divided into four sec-See NUREG 0750,V37 ebetract.1 tions: Background, Objective. Applicent Submittels, and Review Criteris. This report describes the development of the HFE PRM NUREG4700 V30 N04: NUCLEAR REGULATORY COMMISSION and gives e dotated doectlpton of each HFE review element ISSUANCES FOR APRIL 1994. Pages 187-247.

• Division of Freedom of information & Pubiloellons Services (000206-

'l NUREG4700 - VS7: NUCLEAR REGUuTORY > COMMISSION 940714). June 1994. Sepp. 9407010320. 00030:294.

ISSUANCES.Opimons And Decisions Of The Nucteer Regule-

- See NUREG4750,V37 ebetract.

. tory Commenion With Selected Orders.Jenuary June 1993. ? D6-i vision of Freedom of information & Pad *anMs Services NUREG4700 V30 NOS: NUCLEAR REGULATORY COMMISSION i

(890306440714). June 1994. 572pp. 9409010177. 80723:001.

ISSUANCES FOR MAY 1994.PeGoe 249 284.

• Diviolon of.

t Legal leeuences of the Commission, the Atomic Selety and Li -

Freedom of Information & Pubacomons Sen4ces (Poet 940714).

censing Board Penel, the Administrouve Law Judges, and NRC July 1994. 42pp. 9407250180. 80327:324 i

Program Offices are presented-See NUREG4750,V37 abstrect.

NUREG4700 V30: NUCLEAR REGULATORY. COMMISSION NUREG4700 V30 N00: NUCLEAR REGULATORY COMMISSION ISSUANCES.Opiruons And Deceens Of The Nucteer Regule-ISSUANCES FOR JUNE 1994.Pages 206-300.

• Division of tory Commission With Selected Orders. July-December 1993.

• Freedom of informenon & Pubilcagons Services (Poet 940714).

Division of Freedom of Informellon & Publicellons Services August 1994,112pp. 9400250025. 80000001.

t (890206-940714). June 1994. 430pp. 9400010179. 80724:208.

See NUREG4750,V37 abstract.

See NUREG4750,V37 abstract.

NUREG4700 V40101: INDEXES TO NUCLEAR REGULATORY:

NUMEG4700 V30101: INDEXES TO NUCLEAR REGULATORY Cnuusanar1N ISSUANCES. July-September 1994.

• Division of COMMISSION ISSUANCES. July *,i+ 1993.

• Dmeson of Freedom of information & Pubhcetions Services (Post 940714).~

Freedom of information & Pubaceuons Services (890206-m 1994. 33pp. 9501130244. 82314 293. -

I 940714). January 1994. 41pp. 9403140325. 78464:291.

" Sm NUREG 0750,V30,101 ebetract.

Dignets and indones for issuances of the Commiseson, the Atomic Safety and Lloenomg Board Penet, the Admimetreeve IsuREG4700 V401001: NUCLEAR REGULATORY COMMISSION Law Judges, the Directors' Decisions, and the Daniels of Pou-ISSUANCES FOR JULY 1994.Pegos 1-41,

• Division of Free-hans for Rulemaking are presorded, dom of informellon & Pid*=e%s Services (Poet 940714). Sep-1 tomber 1994. 40pp. M30023. 8126227.

NUREG4700 V30102: INDEXES TO NUCLEAR HEGULATORY

' See NUREG 0750,V37 abstrect COMMISSION ISSUANCES. July-Dooomber 1993.

• Dmeion of Freedom of informaton & Pimme%s Services (890206 NUREG4700 V40 NO2: NUCLEAR REGULATORY COMMISSION 940714). Apre 1994. 80pp. 9405040123. 79103.096.

ISSUANCES FOR AUGUST 1994. Pages43-132.

• Division of See WREG4750,V38,2 h Freedom of informellon & Publicellons Services (Poet 940714).

NUREG4700 V30 NOS: NUCLEAR REGULATORY COMMISSION October 1994. Sepp. 9411000019. 81647:133.

3 ISSUANCES FOR NOVEMBER 1993. Pe9es 187-288.

• Division See NUREG4750,V37 ebelrect.

of Freedom of Information & Ph*a*We Sen4ces (890206-NUREG4700 V40 9003: NUCLEAR REGULATORY COMMISSION 940714). February 1994.100pp. 9405240153. 79484:001.

ISSUANCES FOR SEPTEMBER 1994. Pages 133-145.

• Dwi-See NUREG.0750,V37 atstract elon of Freedom of informabon & Piedin=hs Sendoes (Post i

' 00UREG4700 V30 N06: NUCLEAR REGULATORY COMMISSION 940714). October 1994. 20pp. 9411100001. 81755:323.

ISSUANCES FOR DECEMBER 1993. Pe9es 289-391.

• Division See NUREG4750,V37 abstract.

of Freedom of informaton & Publicatons Services (890206-940714). February 1994.100pp. 9403140316. 76477:148.

NUREG4700 V40 0004: NUCLEAR REGULATORY COMMISSION See NUREG-0750,V37 ebstract.

. ISSUANCES FOR OCTOBER 1994. Pages 147-167.

• Division of. Freedom of informaton & PubliceNons Services (Post e

NUREG4700 V30101: INDEXES TO NUCLEAR REGULATORY 940714). November 1994. 27pp. 9412140189. 82032:272.

COMMISSION ISSUANCES.Jenuary -March 1994.

• Duelon of See NUREG4750,V37 abstract -

Freedom of Informaton & Publicatons Services (Post 940714).

f July 1994. 41pp. 9408150182. 80537:291, NUREG4037 V13 0004: NFIC TLD DIRECT RADIATION MONI.

See NUREG4750,V38,101 ebetract TORING NETWORK.Progrees Report. October December 1993.

UCK Y 1 (Post 820201). March 1994.

NUREG4700 V30 let: INDEXES TO NUCLEM REGULATORY COMMISSION ISSUANCES. January June 1994.

• Division of TNs W provides #m M and 6 of h NRC Thor.

Freedom of trWormation & Pubecatons Ser does rPoet 940714).-

(TLD) N Radleton W September 1994. Sepp. 9410130009, 81288:182.

Network. It presents the redletion levels meneured in the vicinity See NUREG 0750,V38,101 ebstrect.

of NRC licensed facMllies throughout the country for the fourth 00UREG4700 V30 0101: NUCLEAR REGULATORY COMMISSION quer1er of 1993.

- ISSUANCES FOR JANUARY 1994. Pages 1-45.

• Divieson of

~ NRC TLD DIRECT RADIATION MONI.

00UREG4037 V14 0001:

Freedom of informaton & Pimmaians Services. (890206-940714). Aprd 1994. 52pp. 9405040127. 79103;157.

TORING NETWORK.Progrees Report January-March,1994.

See NUREG4750,V37 abstract.

STRUCKMEYER,R. Region 1 (Poet 820201). June 1994. 231pp.

9407250293. 00329:001.

91U5W44700 V301003: NUCLEAR REGULATORY COMMISSION This report provides the status and results of the NRC Thor-ISSUANCES FOR FEBRUARY 1994. Pe9es 47 90.

• Deveion of rnoluminsecent Dommeter (TLD) Direct Radleton Monitoring Freedom ' of information & Pubilcanons Services (890206-Network. It presents the redemon levels meneured in the vicinity c

940714). Apre 1994. 50pp. 9405040131. 79103.200.

of NRC hoonced facHlhos throughout the country for the first :

See NUREG-0750,V37 ebetract quarter of 1994.

)

i h:

l -

,a

~

Main Citations and Abstracts 5

NUREG-0837 V14 N02: NRC TLD DIRECT RADIATION MONI-action, or is considering action, and all petitions for rulemaking TORING NETWORK. Progress Report. April-June 1994, which have been received by the Commisson and are pending STRUCKMEYER,R. Region 1 (Post 820201). August 1994.

disposition by the Commission. The Regulatory Agenda is ap-231pp. 9409260062. 81042:112.

dated and issued each quarter.

This report provides the status and results of the NRC Ther-moluminescent Dosimeter (TLD) Direct Radiaton Monitonng NUREG-0936 V13 N01: NRC REGULATORY AGENDA,Ouarterty Network. It presents the radiation levels measured in the viciruty Report, January-March 1994.

• Divison of Freedom of informa-of NRC licensed facilities throughout the country for the second ton & Pubicatons Services (890206-940714). June 1994. 58pp.

i quarter of 1994.

9407250298. 80328:235.

See NUREG4936,V12,N04 abstract.

I NUREG-0837 V14 NO3: NRC TLD DIRECT RADIATION MONI-TORING NETWORK. Progress Report. July-September 1994.

NUREG-0936 V13 N02: NRC REGULATORY AGENDA.Ouarterfy STRUCKMEYER,R. Regen 1 (Post 820201). December 1994.

Report April-Juno 1994.

• Division of Freedom of Information &

231pp. 9501180140. 82346:322.

Pubicatons Services (Post 940714). August 1994. 68pp.

This report provides the status and results of the NRC Ther-9410060320, 81223:167, See NUREG-0936,V12,N04 abstract.

moluminescent Dosimeter (TLD) Direct Radiaton Monitoring Network. It presents the radiation levels measured in the vicinity NUREG-0940 V12 NO3: ENFORCEMENT ACTIONS: SIGNIF1-of NRC licensed facilities throughout the country for the third CANT ACTIONS RESOLVED.Ouarterty Progress Report. July-quarter of 1994, September 1993.

• Ofc of Enforcement (Post 870413). Decem-Th@is compilaton summanzes segnifcant enforcement acto

3. 6% 9402220162. 78MM2.

NUREG-0847 S13: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WATTS BAR NUCLEAR PLANT UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Tenney that have been resolved dunng one quarterty penod (July - Sep-Valley Authonty) TAM P.S. Dnnsion of Reactor Projects - 1/11 th A aM inces @s of N. Wees, aM (Post 870411). April 1994.154pp. 9405310281,79588:001 Orders sent by the Nuclear Regulatory Comrnesson to licensees Supplement No.13 to the Safety Evaluaton Report for the with respect to these enforcement actions. It cs anticipated that n nate in Ws @ caw W W @ hminaW to appleation fded by the Tennessee Valley Authonty for license to operate Watts Bar Nuclear Plant Units 1 and 2, Docket Nos.

managers and employees engaged in activities licensed by the 50-390 and 50-391, located in Rhea County, Tennessee, has

, so at ades can W tahn to We safey W aM been prepared by the Offee of Nuclear Reactor Regulaton of ing future violatons similar to those desenbed in this publica-the Nuclear Regulatory Comrrnsson. The purpose of this sup-piement is to update the Safety Evaluation with (1) additonal in-NUREG-0940 V12 N04: ENFORCEMENT ACTIONS: SIGNIFi-formation submrtted by the applicant since Supplement No.12 CANT ACTIONS RESOLVED.Ouarterty Progress was issued, and (2) matters that the staff had under review Report, October-December 1993.

• Ofc of Enforcement (Post when Supplement No.12 was issued-870413). March 1994. 387pp. 9404010220. 78721:266.

NUREG-0847 S14: SAFETY EVALUATION REPORT RELATED This compilabon summarizes signif' cant enforcement actions TO THE OPERATION OF WATTS BAR NUCLEAR that have been resolved during one quarterly penod (October -

PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Ten.

December 1993) and includes copies of letters, Notices, and nessee Valley Authonty) TAM,P.S. Division of Reactor Projects Orders sent by the Nuclear Regulatory Commission to licensees 1/11 (DRPE) Post 941001). December 1994.132pp.9412300174.

with respect to these enforcement actons. It is anticipated that 82158:127, the informaton in this publication will be widely disseminated to Supplement No.14 to the Safety Evaluation Report for the managers and employees engaged in activities licensed by the appication filed by the Tennessee Valley Authonty for license to NRC, so that actions can be taken to improve safety by avoed-operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos.

ing future violations similar to those described in this publica-50-390 and 50-391, located in Rhea County, Tennessee, has

tion, been prepared by the Office of Nuclear Reactor Regulaton of the Nuclear Regulatory Commission. The purpose of this sup-NUREG-0940 V13N01P01: ENFORCEMENT ACTIONS: SIGNIFI-piement is to update the Safety Evaluaton with (1) additonal in-CANT ACTIONS RESOLVED REACTOR LICENSEES.Ouarterly formaton submitted by the appleant since Supplement No.13 Progress Report, January-March 1994.

• Ofc of Enforcement was issued, and (2) matters that the staff had under review (Post 870413). June 1994. 210pp. 9407260127. 80345:099.

when Supplement No.13 was issued.

This compilation summanzes significant enforcement actions that have been resolved dunng one quarterly period (January -

NUREG-0933 S17: A PRIORITIZATION OF GENERIC SAFETY March 1994) and includes copies of letters, Notees, and Orders ISSUES. EMRIT R. Division of Safety Issue Resoluton (880717 sent by the Nuclear Regulatory Commisson to reactor licens-941217). September 1994. 307pp. 9411290141. 81879:039.

ees with respect to these enforcement actions. It is antcipated The report presents the safety prionty ranking for genenc that the information in this publicaton will be widely disseminat.

safety issues related to nuclear power plants. The purpose of ed to managers and employees engaged in activities licensed these rankings is to assist in the timely and effeient allocation by the NRC, so that actons can be taken to improve safety by of NRC resources for the resoluton of those safety issues that avoiding future violatons similar to those described in this pubh-have a signifcant potential for reducing nsk. The safety poonty cation.

rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of nsk signifcance estimates, the rato of NUREG-0940 V13N01P02: ENFORCEMENT ACTIONS: SIGNIF1-nsk to costs and other impacts estimated to result if resoluton CANT ACTIONS RESOLVED MEDICAL LICENSEES.Ouarterly of the safety issues were implemented, and the consideration of Progress Report, January-March 1994.

• Ofc of Enforcement uncertainties and other quantitative or qualitative factors. To the (Post 870413). June 1994.145pp. 9407260176. 80346:001, extent practcal, estimates are quantitatrve.

See NUREG-0940,V13,N01 P01 abstract.

NUREG-0936 V12 N04: NRC REGULATORY AGENDA.Ouarterly NUREG-0940 V13N01P03: ENFORCEMENT ACTIONS: SIGNIFI-Report, October-December 1993.

• Division of Freedom of Infor-CANT ACTIONS RESOLVED INDUSTRIAL matson & Pubhcations Services (890206-940714). February LICENSEES.Ouarterty Progress ReportJanuary-March 1994.

• 1994.130pp. 9403150227. 78501:135.

Ofc of Enforcement (Post 870413). June 1994. 118pp.

The NRC Regulatory Agenda is a compilaton of all rules on 9407260183. 80346:146.

whch the NRC has recently completed acton, or has proposed See NUREG-0940,V13 N01,P01 abstract.

6 Main Citations and Abstracts NUREG-0940 V13N02P01: ENFORCEMENT ACTIONS: SIGNIFl.

improve safety by avoiding future violatons similar to those de-CANT ACTIONS RESOLVED REACTOR LICENSEES.Ouarterly scnbod in this pubhcaton.

Progress Report,Apnt-June 1994.

• Ofc of Enforcement (Post 870413). August 1994. 200pp. 9409270322. 81046:108.

NUREG-1021 R07 S01: OPERATOR LICENSING EXAMINER This compdaten summanzes signifcant enforcement actons STANDARDS

• Dmson of Reactor Controls & Human Factors that have been resolved dunng one quarter 1y pered (Apnl.

(Post 921004). June 1994.120pp. 9407070304. 80112:284.

Jee 1994) and includes copies of letters, Notees, and Orders The Operator Licensing Examiner Standards provide poicy sent by the Nuclear Regulatory Commessen to reactor licens-and guidance to NRC examiners and estabhsh the procedures ees with respect to these enforcement actons. It is antcipated and practices for examarung hcensees and appicants for reactor that the informaton in this pubhcaton wdl be widely disseminat-operator and senor reactor operator licenses at power reactor ed to managers and employees engaged in actmties Icensed facdities pursuant to Part 55 of Title 10 of the Code of Federal by the NRC, so that actons can be taken to improve safety by Regulations (10 CFR 55). The Examiner Standards are intended tvoiding future violatons similar to those desenbed in this pubh-to assist NRC examiners and facdity bcensees to better under-cation.

stand the initial and requahfcaton examination processes and to ensure the equitable and consistent administraton of exami-NUREG-0940 V13N02P02: ENFORCEMENT ACTIONS: SIGNIFI-natens to all appleants. These standards are not a substitute CANT ACTIONS RESOLVED MEDICAL LICENSEES Quarterly for the operator iconsang regulations and are subject to revision Progress Report.Aprildune 1994.

• Ofc of Enforcement (Post or other internal operator Icensing polcy changes. Revision 7 870413). August 1994. 354pp. 9410130039. 81263:001.

was pubbshed in January 1993 and became effectwo in August See NUREG-0940,V13,N02,P01 abstract.

1993. Supplement 1 is being issued primanly to implement ad-ministratwo changes to the requahficaten examinaten program NUREG-0940 V13N02P03: ENFORCEMENT ACTIONS: SIGNIFl-resu ng m e anwndment to 10 UR 55 mat ehaM N CANT ACTIONS RESOLVED INDUSTRIAL went b m kM paw m pass an MNm LICENSEES.Ouarterty Progress Report,Apol-kne 1994.

• Ofc ed mquaMeaton samaton as a Mm W Icense w of Enforcement (Post 870413). August 1994. 177pp.

newal. The supplement does not substantially alter either the 9410070244.81232:001.

initial or requahMon examaton processes and wm Mcome See NUREG 0940,V13 N02,P01 abstract.

effectue 30 days after its pubhcaton is notced in the Federal NUREG-0940 V13NO3P01: ENFORCEMENT ACTIONS: SIGNIFl.

Register. The corporate notifcaton letters issued after the et-CANT ACTIONS RESOLVED, REACTOR LICENSEES Quarterty factwe date wdl provide facihty licensees with at least 90 days Progress Report, July-September 1994.

• Ofc of Enforcement notice that the examinatons will be adrrunestered in accordance (Post 870413). December 1994. 315pp. 9501180207.

with the revised procedures.

82344:001.

NUREG-1022 R01 DR FC: EVENT REPORTING GUIDELINES This compdaten summanzes signifcant enforcement actons 10CFR50.72 AND $0.73.Second Draft For Comment.

that have been resolved dunng one quarterty perod (July - Sep.

tember 1994) and includes copies of letters, Notees, and ALLISON.D.P.; HARPER,M.R.; ISRAEL S.; et al. Offee for Anal-Orders sent by the Nuclear Regulatory Commesson to reactor ysis & Evaluaton of Operatonal Data, Director. February 1994.

hcensees with respect to these enforcement actons. It is antci.

182pp. 9402220125. 78192.041.

Revison 1 to NUREG-1022 clanfies the immediate notifica-pated that the information in this pubication wdl be widely dis.

seminated to managers and employees engaged in actuities b-ton requirements of Title 10 of the Code of Federal Regula-censed by the NRC, so that actons can be taken to emprove tions, Part 50, Section 50.72 (10 CFR 50.72), and the 30-day safety by avoiding future voistons similar to those desenbed in written licensee event report (LER) requirements of 10 CFR this pubhcaton.

50 73 for nuclear power plants. This revision was initiated to im-prove the reporting guidehnes related to 10 CFR 50.72 and NUREG-0940 V13N03P02: ENFORCEMENT ACTIONS: SIGNIFl-50.73 and to consohdate these guidehnes into a single refer.

CANT ACTIONS RESOLVED. MEDICAL LICENSEES.Ouarterty ence document. A first draft of this document was notced for Progress Report. July-September 1994.

• Ofc of Enforcement pubhc comment in the Federal Register on October 7,1991 (56 (Post 870413). December 1994. 316pp. 9501180211.

FR 50598). This document updates and supersedes NUREG-82343:001.

1022 and its Supplements 1 and 2 (pubbshed in September This compilaton summanzes signifcant enforcement actons 1983, February 1984, and September 1985, respectuety), it that have been resolved dunng one quarterty penod (July - Sep-does not change the reporting requirements of 10 CFR 50.72 tember 1994) and includes copies of letters, Notees, and and 50.73.

Orders sent by the Nuclear Regulatory Commesseon to medcal iconsees with respect to these enforcement actons. It ta antci.

NUREG-1100 V10: BUDGET ESTIMATES Fiscal Year 1995.

• De-pated that the informaton in this pubicaton wdl be widely dis-vision of Budget & Analysis (Post 890205). February 1994.

seminated to managers and employees engaged in actuities h-217pp. 9402220139. 78191:075.

censed by the NRC, so that actons can be taken to improve This report contains the fiscal year budget justificaton to Con-safety by avoiding future violatons simdar to those desenbod in gress. The budget provides estimates for salanes and expenses this pubhcaton.

and for the Offee of the Inspector General for fisca: year 1995.

NUREG-0940 V13NO3P03: ENFORCEMENT ACTIONS: SIGNIF1-NUREG-1125 V15: A COMPILATION OF REPORTS OF THE AD.

CANT ACTIONS RESOLVED MATERIAL LICENSEES (NON.

VISORY COMMITTEE ON REACTOR SAFEGUARDS.1993 MEDICAL)Quarterty Progress Report, July-September 1994.

• Annual.

• ACRS - Advisory Committee on Reactor Safeguards.

Ofc of Errorcement (Post 870413). December 1994. 328pp.

April 1994.156pp. 9405040139. 79104 001.

95011802i5. 82342 001.

This compilation contains 47 ACRS reports submitted to the This Nmpdaten summanzes signiftant enforcement actons Commission, Executrve Director for Operatons, or to the Office t% ha e been resolved dunnp y,e cuarterty penod (July - Sep-of Nuclear Regulatory Research, dunng calendar year 1993. It tembe. 1994) and includes copes of letters, Notees, and also includes a report to the Congress on the NRC Safety Re-Ordens sent by the Ntr,' ear %.htory Commisson to matenal search Program. All reports have been made avadable to the hc.msees (r,cin-medical, vtt' respect to these enforcement ac-pubic thrnugh the NRC Pubic Document Room and the U.S. Li-tons. It is anicipated th at :he information m this pubication wdt brary of Congress. The reports are categortied by the frost ap-proonate genere sub ect area and by chronological order within be widely disseminated tvi managers and employees engaged in t

actuities hcensed by the DRC, so that actions can be taken to subject area.

t l

Main Citations and Abstracts 7

NUREG 1145 V10: U.S. NUCLEAR REGULATORY COMMISSION NUREG-1266 V08: NRC SAFETY RESEARCH IN SUPPORT OF 1993 ANNUAL REPORT.

• Offee of Administrabon, Director REGULATION - FY 1993.

• Office of Nuclear Regulatory Re-(Post 940714). September 1994. 306pp. 9410130036.

search (860720 941217). June 1994. 100pp. 9407070294.

81262:001.

80113:209.

This report covers the major activities, events, decisions, and planning that took place dunng Fiscal Year 1993 wilfun the U.S.

TNs report, the ninth in a senes of annual reports, was pre.

Nuclear Regulatory Commission (NRC) or involving the NRC.

pared in response to congressional inquines conceming how nuclear regulatory research is used. It summarizes the accom-NUREG-1200 R03: STANDARD REVIEW PLAN FOR THE

• I REVIEW OF A LICENSE APPLICATION FOR A LOW-LEVEL FY 1993. A special emphasis on accomplishments in nuclear RADIOACTIVE WASTE DISPOSAL FACILITY.

• Division of power plant aging research reflects recogntbon that a number of Low-Level Waste Management & Decommissioning (870413-plants are entenng the final portion of their original 40-year op-940402). April 1994. 582pp. 9405100233. 79279:287.

ersting hcenses and that, in addition to current aging effects a The Standard Review Plan (SRP) is prepared for the guid-focus on safety considerations for license renewal becomes ance of staff revsewers in the Offee of Nuclear Material Safe y timely. The pnmary purpose of performing regulatory research is and Safeguards in performing safety reviews of applicahons to to develop and provide the Commission and its staff with the construct and operate a low-level waste disposal facility. The technical bases for regulatory decisions on the safe operation pnncipal purpose of the SRP is to assure the quahty and uni-of heensed nuclear reactors and facihties, to find unknown or formity of staff reviews and to present a well-defined base from unexpected safety problems, and to develop data and related which to evaluate proposed changes in the scope and require-information for the purpose of revising the Commission's rules, ments of reviews. It is also a purpose of the SRP to make infor.

regulatory guides, or other guidance.

i mahon about regulatory matters widely available and to improve communcation and understanding of the staff's review process NUREG-1272 V08 N01: OFFICE FOR ANALYSIS AND EVALUA.

by interested rnembers of the pubic and the nuclear industry.

TlON OF OPERATIONAL DATA.1993 Annual Report - Power NUREG-1200 consists of 11 chapters containing approximately Reactas.

• Offee for Ana9 sis & Evaluabon of Operational 60 individual SRP sechons. Each section idenbfies the disci-Data. Director. December 1994. 261pp. 9501180134.

82348:001.

l phnes appropriate to perform the review, the matters that are reviewed, the basis for review, how the review is performed' This annual report of the U.S. Nuclear Regulatory Commis-and the conclussons that are sought. The SRP will be revised si n's Office for Analysis and Evaluation of Operational Data from time to bme to reflect changes in regulatory policy and to (AEOD) describes actrvities conducted dunng 1993. The report update techncal information in the text.

is ublished in two parts. NUREG-1272, Vol. 8, No.1, covers

)

ower reactors and presents an overview of the operating expe.

NUREG-1214 R13: HISTORICAL DATA

SUMMARY

OF THE SYS.

rience of the nuclear power industry from the NRC perspecbve, l

TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE.

inewg cannets about the trends of some key perfwmance ALLENSPACH.F. Dnnsson of Reactor inspechon & Licensee measures. The reput also includes the pnncipal findings and Performance (Post 921004). March 1994. 94pp. 9404110359.

issues idenbfied in AEOD studies over the past year and sum-78651:200.

marizes informabon from such sources as licensee event re-The Histoncal Data Summary of the Systemate Assessment Ports, diagnoste evaluations, and reports to the NRC's Oper-of Licensee Performance (SALP) is produced penodca ly by the ations Center. NUREG-1272 Vol. 8, No. 2 covers nuclear mate-U.S. Nuclear Regulatory Commission. TNs summary provides rials and presents a review of the events and concerns during the results of the assessment for each facihty by NRC region 1993 associated with the use of heensed material in nonreactor and is further divided into the following sect #ons: Secton 1 pre.

apphcabons, such as personnel overexposures and medical sents the most recent SALP report rabogs for facahties in oper, misadministrations. Both reports also contain a discussion of i

ation. Section 2 presents a chronological listing of all SALP the incident investigabon Team program and summarize both report ratings for each operahng facihty since February 1,1989.

the incident Investigation Team and Augmented Inspection Team reports. Each volume contains a kst of the AEOD reports NUREG-1242 V03 PT01: NRC REVIEW OF ELECTRIC POWER issued from 1980 through 1993.

RESEARCH INSTITUTE'S ADVANCED LIGHT WATER REAC-TOR UTILITY REQUIREMENTS DOCUMENTS.Passrve Plant NUREG-1275 V10: OPERATING EXPERIENCE FEEDBACK Designs Chapter 1. Project Number 669.

• Associate Director for REPORT RELIABILITY OF SAFETY-RELATED STEAM TUR-Advanced Reactors & License Renewal (Post 910918). August BINE-DRIVEN STANDBY PUMPS. Commencal Power Reactors.

1994. 450pp. 9410060313. 81222:001.

BOARDMAN,J.R. Division of Safety Programs (Post 870413).

l The staff of the U.S. Nuclear Regulatory Commission has pre.

October 1994. 91pp. 94)1280281. 81851:135.

pared Volume 3 (Parts 1 and 2) of a safety evaluabon report This report documents a detailed analysis of failure initiators, (SER), "NRC Revew of Electric Power Research institute's Ad-cases and dessgn features for steam turbine assembhes (tur-vanced Light Water Reactor Utihty Requirements Passive Plant bines dh thwr related components, such as govemors and Designs" to document the results of its review of *e Electnc valves) whsch are used as drivers for standby pumps in the aux-Power Research institute's " Advanced Light Water Reactui Lmi.

ihary feedwater systems of U.S. commercial pressurized water i

ity Requirements Document." TNs SER grves the results of the reactor plants, and in the high pressure coolant injection and re-staff's review of Volume lli of the Requirements Document for actor core isolabon coohng systems of U.S. commercial boiling passive plant designs, which consists of 13 chapters and con-water reactor plants. These standby pumps provide a redundant tains utility design requirements for nuclear power plants for source of water to remove reactor core heat as specified in indi-whch passrve features will be used in their design (approxi-vidual plant safety analysis reports. The penod of revew for tNs matefy 600 megawatts-electnc) report was from January 1974 through December 1990 for h-censee event reports (LERs) and January 1985 through Decem-NUREG 1242 V03 PT02: NRC REVIEW OF ELECTRIC POWER ber 1990 for Nuclear Plant Reliabihty Data System (NPRDS) fail-RESEARCH INSTITUTE'S ADVANCED LIGHT WATER REAC-ure data. TNs study confirmed the continuing validity of conclu-TOR UTILITY REQUIREMENTS DOCUMENT. Passive Plant sions of earlier studes by the U.S. Nuclear Regulatory Commis.

Designs. Chapters 213. Project Number 669.

• Associate Direc-sion and by the U.S. nuclear industry that the most significant tor for Advanced Reactors & Lcense Renewal (Post 910918).

factors in failures of turbine-driven standby pumps have been August 1994. 642pp. 9410070247. 81230:001.

the failures of the turbine-dnvers and their controls. Inadequate See NUREG-1242,V03.P01 abstract.

maintenance and the use of inappropriate vendor techncal in-1

8 Main Citations and Abstracts formaton were identified as signifcant factors which caused re-generating capacity and average capacity factor for operating curnng failures.

U.S. commercial nuclear power reactors is obtained from momW operahng mpods mat am eM hW to N E NUREG-1307 R04:

REPORT ON WASTE BURIAL by the Icensee. This information is reviewed by the NRC for CHARGES.Escalabon Of Decommissioning Waste Disposal mnystency ony and no WM vaMah ah Mca-Costs At Low Level Waste Burial Facihties.

• Dmsion of Regula-bon is pehmed tory Apphcahons (870413-941217). June 1994. 54pp.

9408250008.80658.306.

NUREG 1368: PREAPPLICATION SAFETY EVALUATION 1

One of the requirements placed upon nuclear power reactor REPORT FOR THE POWER REACTOR INNOVATIVE SMALL licensees by the U.S, Nuclear Regulatory Commission (NRC) is MODULE (PRISM) LIQUID-METAL REACTOR. Final Report.

for the licensees to penodcally adjust the estmate of the cost DONOGHUE.J.E.: DONOHEW,J.N.; GOLUB,G.R.; et al. Asso6 of decw.nm,w.ing their plants, in dollars of the current year, ate Drector for Advanced Reactors & License Renewal (Post as part of the process to provide reasonable assurance that 910918). February 1994. 400pp. 9404010163. 78716:001.

adequate funds for decommissioning will be available when TNs preappication safety evaluation report (PSER) presents needed. This report, which is scheduled to be revised periodi-the results of the preapphcaton design review for the Power cally, contains the development of a formula for escalating de-Reactor innovatrve Small Module (PRISM) liquid-metal (sodium)-

commissioning cost eshmates that is acceptable to the NRC, cooled reactor, Nuclear Regulatory Commission (NRC) Project and contains values for the escalation of radioactive waste No. 674. The PRISM conceptual design was submitted by the bunal costs, by site and by year. The teensees may use the for-U.S. Department of Energy in accordance with the NRC's mula, the coefficients, and the bunal escalation from this report

" Statement of Polcy for the Regulabon of Advanced Nuclear in their escalabon analyses, or they may use an escalation rate Power Plants" (51 Federal Register 24643). This policy provides at least equal to the escalation approach presented herein.

for the early Commission review and interaction with designers and licensees. The PRISM reactor design is a small, modular, NUREG-1323 R00: LICENSEE APPLICATION REVIEW PLAN pool-type, liquid-metal (sodium)-cooled reactor. The standard FOR A GEOLOGIC REPOSI7ORY FOR SPENT NUCLEAR plant design consists of three identcal power blocks with a total FUEL AND HIGH-LEVEL RADIOACTIVE WASTE.

• Dmsion of electrical wtput rating of 1395 MWe. Each power block com-Waste Management (NMSS 940403). September 1994. 564pp.

prises thme reactw modules, each we a monnal rahng of 471 9411290144.81880-001.

MWt. Each module is located in its own below-grade allo and is In developing the MRP at tNs time, there are numerous ben.

connected to its own intermediate heat transport system and efits to be reahzed by the NRC staff. Frst, the LARP will help structure and coordinate the staff's efforts to develop a well-in-SMam generaW system. De macWs W,hze a mew @

fuel, a tomary alloy of U-Pu-Zr. The design mcludes passrve re-tegrated review capability supported by the staff's independent actor shutdown and passive decay heat removal features. The technical assessment capabihty and the results of NRC-spon-PSER is the NRC's preliminary evaluation of the safety features sored research. Second, identifying technical issues most im-in the PRISM design, including the projected research and de-portant to geologe repository performance will help focus and velopment programs required to support the design and the pro-prioritize the pre-icense appication reviews, independent tech-ncal assessment capabihties, and research needed by the staff.

posed testing needs. Because the NRC review was based on a Third, the regulatory and techncal bases for the staff's pre-li-conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provid-cense applicaton reviews and guidance will be improved due to the focus on regulatory and techncal issues. The LARP is dmd.

ed some guidance on appleable licensing critena, assessed the ed into three parts. Part A contains the License Apphcation adequacy of the preapphcant's research and development pro-Review Strategy, which gives general guidance to the staff in grams, and concluded that no obvims impediments to heensing conducting its Icense apphcation rertiews. Part B contains eight the PRISM design had been identified.

Individual review plans, to be used by the staff to review general NUREG-1415 V06 N02: OFFICE OF THE INSPECTOR informabon in the license appicabon. Part C contains 89 individ-GENERAL. Semiannual Report, October 1,1993 - March 31, ual review plans, distnbuted among ten chapters, wNch the 1994. NORTON,L; BARCHl.T.; HUBER,D. Ofree of the inspec-staff will use to review the Safety Analysis Report, the pnncipal tor General (Post 890417). April 1994, 46pp. 9409200338.

part of the hcense apphcation in which DOE provides the infor-80954:277.

mabon needed to demonstrate comphance with the techncal The inspoctor General is required by the IG Act of 1978, as requirements of 10 CFR Part 60. The organizabon of the individ-amended, to prepare a semiannual report to Congress whch ual review plans in Parts B and C is consistent with the organi-summarizes program actmties. The 6-month reportog period zabon of the hcense apphcaton as specified in the draft regula-ends March 31 and September 30th. The IG's report is submit-tory guide " Format and Content for the License Applicabon for tod to the Chairman of the NRC not later than April 30 ed Oc-the Waste Repository" Finally, each individual review plan has tober 31, respectively. The Chairrnan comments on the IG's a standard format consisbng of the following six sechons: (1) report and prepares his own, as required by t.ie Act, and sub-Applicable [10 CFR Part 60] Regulatory Requirements; (2) mata both reports to Congress no later than November 30 and Review Strategy; (3) Review Procedures and Acceptance Crite-May 31, respectively.

ria; (4) implementation; (5) Example Evaluation Findings; and (6)

References, NUREG 1415 V07 N01: OFFICE OF THE INSPECTOR GENERALSemiannual Report. April 1 -September 30, 1994.

NUREG-1350 V06: NUCLEAR REGULATORY COMMISSION IN-DONOVAN,D.; MULLEY,G.; BODENSTEINER.J.; et al. Offee of FORMATION DIGEST.1994 Editen. STADLER,L Dmson of the inspector General (Post 890417). October 1994. 50pp.

Budget & Analysis (Post 890205). March 1994. 128pp.

9412300183. 82159:204.

9406000130, 79618:009.

See NUREG-1415,V08,N02 abstract.

The Nuclear Regulatory Commission Informat#on Digest (dgest) provides a summary of informaton about the U.S. No-NUREG-1416: OPERATIONAL EXPERIENCE AND MAINTE-clear Regulatory Commisson (NRC), NRC's regulatory responsi-NANCE PROGRAMS OF TRANSAMERICA DELAVAL, INC.,

bilities, the actmbes NRC heenses, and general enformaton on DIESEL GENERATORS. RAJAN.J. Division of Engineering (Post domeste and worldwide nuclear energy. The digest, pubhshed 921004). May 1994. 42pp. 9405310249. 79589:001.

annually, is a compilation of nuclear-and NRC-related data and in response to concerns prompted by a crankshaft failure in a is desgned to provide a Quck reference to ma)or facts about Transamenca Delaval, Inc., (TDI) emergency diesel generator the agency and the industry it regulates. In general, the data (EDG) at Shoreham in August 1983,11 (now 8) U.S. nuclear rever 1975 through 1993, with exceptons noted. Informaton on utthty owners formed a TDI Diesel Generator Owners' Group to

Main Citations and Abstracts 9

address the related operational and regulatory issues. The tech-personnel. This report is to provide a comprehensive descripton necal resoluton involved implementaten of programs focused of the implementaten and venfcation status of TMI Acton Plan j

on the dessgn review of a large set of important engine compo-Requirements, USIs, GSis, and other MPAs that have haei; re-nents to ensure their adequacy from a manufacturing and oper-solved and involve implementaten of an acton or actons by li-ational standpoint. The most entical periode maintenance /sur-censees. This repor1 makes the informaton availabh to other veillance actons for certain components, such as connecting interested parties, including the public. An additonal prpose of rods, crankshafts, cylinder blocks, cylinder heads, piston skirts, this report is to serve as a follow-on to NUREG 0933, "A Priori-and turbochargers were incorporated as license conditions. On tization of Generic Safety issues," which tracks safety issues up the basis of its review of the substantial operational experience until requirements are approved for imposition at licensed plants of the TDI engines accumulated since 1985 and the inspecton or until the NRC issues a request for acton by licensees.

results of the EDG components, the NRC staff has concluded that there is adequate justifcation in removing the component.

NUREG-1460 RO1: GUIDE TO NRC REPORTING AND RECORD-based Icense conditons, and the TDI EDGs can be regulated in KEEPING REQUIREMENTS. Compiled From Requirements in a manner similar to other EDGs withan the nuclear industry with.

Title 10 Of The U.S. Code Of Federal Regulatons As Codified out the special requirements imposed in 1985.

On December 31,1993. COLLINS,M.; SHELTON.B. Office of in-formaton Resources Management (Post 890205). July 1994.

NUREG-1426 V02: COMPILATION OF REPORTS FROM RE-254pp. 9408180211, 80625:001, SEARCH SUPPORTED BY THE MATERIALS ENGINEERING This compilaton includes in the first two sectons the report-BRANCH DIVISION OF ENGINEERING.19911993. HISER,A.L.

Division of Engineenng (870413-941217). June 1994. 39pp.

ing and recordkeepmg requirements applicable to U.S. Nuclear 9408030136. 80425:196-Regulatory Commission (NRC) Icensees and appicants and to members of the pubic. It includes those requirements codified Since 1965, the Matenals Engineenng Branch, Division of E*

gineenng, of the Nuclear Regulatory Commission's Offee of Nu-in Title 10 of the Code of Federal Regulatons, Chapter 1, on December 31,1993. It also includes, in a separate secten, any clear Regulatory Research, and its predecessors dating back to the Atome Energy Commesson (AEC), has sponsored research of those requirements that were superseded or discontinued be-programs conceming the integnty of the pnmary system pres-tween January 1992 and December 1993 Finally, the appendix sure boundary of light water reactors. The components of com Ists mailing and delivery addresses for NRC Headquarters and cern in these research programs have included the reactor Regional Offices mentoned in the compilaton.

pressure vessel (RPV), steam generators, and the piping. These NUREG-1462 V01: FINAL SAFETY EVALUATION REPORT RE.

research programs have covered a broad range of topes, in-LATED TO THE CERTIFICATION OF THE SYSTEM 80+

ciuding fracture mechancs analysis and expenmental work for DESIGN. Chapters 1 14. Docket No.52-002. (Asea Brown RPV and piping appbcatons, inspecton method development Boveri-Combuston Engineering)

• Associate Director for Ad-and quahfcaton, and evaluaton of irradiaton effects to RPV vanced Reactors & License Renewal (Post 910918). August steels. This report provides as complete a listing as practical of 1994. 583pp. 9409200306. 80953:001, formal techncal reports submitted to the NRC by the armstiga-This final eefety evolustion report (FSEM dooumonto the tors working on these research programs. This listing includes techncal review of the System 80+ standard design by the topical, final and progress reports, and is segmented by tope area. In many cases a report will cover several topcs (such as U.S. Nuclear Regulatory Commiseen (NRC) seeW. The appica ton for the System 80+ design was adwnitted by Combuston in the tase of progress reports of multi-faceted programs), but Engmeenng, Inc., now Asee Brown Boveri-Comtwetion Engi-is listed ander only one topc. Therefore, in searching for reports on a specfc topic, other related topc areas should be checked neenng (ABB-CE) as an appleetion for desten approval and also. The pred:,w volume to this report covers the penod 1965 subsequent design certification pursuant to 10 CFR l$2.45, 1990-System 80+ is a presourtred water reactor with a rated power of 3914 megewette thermal (MWt) and a design power of 3992 NUREG-1435 S03: STATUS OF SAFETY ISSUES AT LICENSED MWt at whch mWto are analyred. Many features of the POWER PLANTS.TMI Acton Plan Requirements Unresolved System 80+ are osmilar to thoes of ABB-CE's System 80 Safety issues. Generic Safety issues.Other Muttsplant Action dessgn from which it evolved Unique features of the System issues.

• Program Management, Polcy Development & Analysis 80 + doesgn include: a large spherical, steel containment; an in-Staff (Post 870411). December 1993. 165pp. 9402220173.

contamment refuelmg water storage tank; a reactor cavity flood-78197:187.

ing system, hydrogen ignrlors, and a safety depreneurtretion As part of ongong U.S. Nuclear Regulatory Commission system for severe accident mitgetort a combuston gas turtune (NRC) efforts to ensure the quality and accountabetity of safety for an anernate ac source; and an adver,oed dlgitally beood issue informaton, a program was established whereby an control room. On the besse of its ovatustom and independent j

annual NUREG report would be published on the status of li.

analyses, the NRC staff concludes that AEB-CE's application censee implementaten and NRC venfcaton of safety issues in for design certlicaton meets tr's requremots of Subpert B of major NRC requirement areas. This information was compiled 10 CFR Part 52 that are applicable and technscally relevant to

)

and reported in three NUREG volumes. Volume 1, published in the System 80+ standard desigt.

March 1991, addressed the status of Three Mile Island (TMI)

NUREG 1462 V02: FINAL SAFETY EVALUATION REPORT RE-

{

Action Plan Requirements. Volume 2, published in May 1991*

LATED TO THE CERTIFICATION OF THE SYSTEM 80+

addressed the status of unresolved safety issues (USls).

l Volume 3, pubhshed in June 1991, addressed the implementa-DESIGN. Chapters 15-22 And Appendees. Docket No. 52-tion and venfcation status of generic safety issues (GSis). Sup-002.(Asea Brown Boveri-Combuston Engineenng)

• Associate a

piement 1, published in December 1991 combined these vol-Director for Advanced Reactors & License Renewaf (Post 910918). August 1994. 557pp. 9409200314. 80956:001, umes into a single report and provided updated informaton as See NUREG-1462,V01 abstract.

of September 30,1991. Supplement 2, published in December 1992, provided updated informaton on TMI, USI, and GSI NUREG-1470 V03: FINANCIAL STATEMENT FOR FISCAL YEAR issues and included status of all other Muttiplant Actions 1993.

• Office of the Controller (Post 890205). August 1994.

(MPAs). This annual NUREG report provides updated informa-110pp. 9409090026. 80816:001.

ton on TMI, USI, and GSI and other MPAs as of September 30, The Chief Financial Offcers Act of 1990 requires the NRC 1993. The data contained in these NUREG reports are a prod-Chief Financial Offcer to prepare and submit an annual financial uct of the NRC's Safety issues Management System (SIMS) da-statement to the Drector of the Off~ e of Management and c

tabase, whch is maintained by the Protect Management Staff in Budget (OMB). The OMB has replaced the requirement for the the Ottee of Nuclear Reactor Regulation and by NRC regonal CFO's Annual Report with the annual f:nanciaf statement. The l

1 i

1 l

10 Main Citations and Abstracts annual financial statement was previously included in the Chief MENT FOR THE CONSTRUCTION AND OPERATION OF CLAl-Financial Offcer's Annual Report. This report is the third annual BORNE ENRICHMENT CENTER, HOMER, LOUISlANA. Docket report for the NRC and includes an overview of the NRC, the No. 70-3070. Louisiana Energy Services, LP. Environmental audited pnncipal financial statements and audit reports for fiscal Impact Statement. ZEITOUN,A. Science Applications Interna-l year 1993, and supplemental financial and management infor.

tional Corp. (formerly Science Applications, Inc.). August 1994.

l mation.

390pp. 9409070096. 80790:001.

This Final Environmental impact Statement (FEIS) was pre-NUREG-1471: CONCEPT OF OPERATIONS WITH ORGANIZA.

pawd by the Nuclear Regulatory Commission in accordance TION CHARTS.NRC incident Response.

HIMES,J.;

with %RC regulation 10 CFR Part 51, which implements the Na-i JACKSON,K.; LOPRESTI.F. Divison of Operational Assessment tionn r nvironmental Poicy Act (NEPA), to assess the potential I

(Post 870413). February 1994. 69pp. 9403140209. 78456:175.

envirmmental impacts of the construction and operation of a The U.S. Nuclear Regulatory Commission (NRC) regulates nu-proposed Daseous centnfuge enrichment facility to be built in Claiborne Parish, LA. The proposed facility will have a produc-clear power plants and certain other crvihan nuclear facilities i

ton capacity of about 866 tonnes annuaHy of up to 5 percent and materials to protect the public health and safety and to pre-cNd M, usN a prwen centn@ the incW e

serve environmental quahty. While the foremost objecbve of e assessnwnt am conse, M anal opam aM regulaton is to prevent accidents, the NRC is also prepared to help its licensees and State and local 90vernments mit'9sie the tual decontaminaten and decommissioning of the site. In order consequences of any that might occur. This document de-to help assure that releases from the operation of the facihty scribes the NRC concept, purposes, and organization for per-and potential impacts on the pubhc are as low as reasonably forming essential funcbons dunng a Federal response to a achievable, an environmental monitoring program was devel-j severe reactor accident, with an emphasis on State and Federal oped to detect significant changes in the background levels of coordinabon, uranium around the site. Other issues addressed include the purpose and need for the facihty, the attematrves to the pro-NUREG-1475: APPLYING STATISTICS. LURIE.D. Office of the posed acton, the aste selection process, environmental justice, Controller (Post 890205). MOORE,R.H. Bonneville Power Ad-and tails dispositen. The NRC concludes that the facility can be ministrabon. February 1994. 594pp. 9405310242. 79584:001-constructed and operated with small and acceptable impacts on Applying Stabstes is both a reference book and a textbook the pubhc and the environment. The FEIS supports licensing.

on statistcal methods. Although the majonty of the book's ex-tmples are taken from the field of nuclear regulaton, the statis-NUREG-1484 V02: FINAL ENVIRONMENTAL IMPACT STATE-teal pnnciples that they embody are apphcable across a wide MENT FOR THE CONSTRUCTION AND OPERATION OF CLAl-spectrum of scientifc, techncal, and hfe problems. Many of BORNE ENRICHMENT CENTER HOMER, LOUISlANA. Docket these more widely based problems are addressed directly in the No. 70-3070. Louisiana Energy Servces, L.P. Comments And t;xt and illustrated with appropnate data. The 21 chapters are Responses. ZEITOUN A. Science Appications Intematonal l

interspersed with "For discussen:" sectons to promote the Corp. (formerty Se,sence Appleatons, Inc.). August 1994.480pp.

9409070125. 80792:001.

I reader's insights and intuitive understanding of the appropriate Volume 2 contains the comments rece#ved on the Draft EIS techniques. Each chapter is introduced by a hst of items cov-and the msponses to those comments.

l ered in the chapter and is concluded by a hst of concepts to be remembered. A special secten entitled "Afterword" lists areas NUREG-1486: FINAL SAFETY EVALUATION REPORT TO Li-of stabstes that are not covered in the book; it grves a brief de-CENSE THE CONSTRUCTION AND OPERATION OF A FACILi-senption of each of these areas along with one or two applea-TY TO RECEIVE, STORE AND DISPOSE OF 11E(2) BYPROD-ble and rocccnn,ociGmi references _ Applying Stabstes also pro-UCT MATERIAL NEAR CLIVE. UTAH. Docket No. 40-8989.(En-vides an extensive set of statistcal tables. Each table is ex-virocare of Utah.inc.) ABU-EID.R.; BRUMMETT E.; BYKOSKl.L plained and its use illustrated by at least one example; thus, the et al. Drvison of Low-Level Waste Management & Decommis-j tables can be used independently of the rest of the material, sioning (870413-940402). January 1994.116pp. 9402220120.

The book's notatonal conventions, as they apply to the eight 78191:291, common stabstcal distributions that support the various analy-The Final Safety Evaluaton Report (FSER) summanzes the ses, are displayed on.the inside front and back covers and the U.S. Nuclear Regulatory Commisson (NRC) staff's revimv of flyleaves.

Envirocare of Utah, Inc.'s (Envirocare's) apphcaton for a hcense to receive, store, and dispose of uratwum and thorium byproduct l

NUREG-1478: NON-POWER REACTOR OPERATOR LICENSING material (as defined in Secten 11e.(2) of the Atomic Energy Act l

EXAMINER STANDARDS.

• Divison of Reactor Controls &

of 1954, as amended) at a site near Clrve, Utah. Envirocare pro-Human Factors (Post 921004). June 1994. 200pp. 9408030122.

poses to dispose of high-volume, low-activity Section 11e.(2) by-80424:001.

product material in separate earthen disposal cells on a site The Non-Power Reactor Operator Licensing Examiner Stand _

where the apphcant currentty disposes of naturally occurring ra-ards provide pohey and guidance to NRC examiners and estab-deactrve material (NORM), low-level waste, and mixed waste under hcense by the Utah Department of Environmental Quality.

l Hsh the procedures and practces for examining licensees and The NRC staff review of the December 23,1991 hcense apphca-f apphcants for NRC operator icenses pursuant to Part 55 of ton, as revised by page changes dated July 2 and August Title 10 of the Code of Federal Regutabons (10 CFR Part 55).

10,M2, AM 5, 7, and 10, 23, and May 3, 6, 7, H, and 20 They are intended to assist NRC examiners and facility licens-1993, has identifed open issues in geotechnical engineering, ees to better treo,&u4 the examinaton process and to wa s u es e on, ra na nua nama! assuo ensure the equitable and consistent administration of examina-ance, and radelogical safety. The NRC will not issue a icense bons to all applcants. These standards are not a substitute for for the proposed acton until Envirocare adequately resolves the operator hcensing regulatons and are sub}ect to revision or these open issues.

l other internal operator examinaten hcensing polcy changes. As appropriate, these standards will be revised periodscally to ac-NUREG-1488: REVISED LIVERMORE SEISMIC HAZARD ESTI-commodate comments and reflect new informaton or experi.

MATES FOR SIXTY NINE NUCLEAR POWER PLANT SITES ence.

EAST OF THE ROCKY MOUNTAINS. Final Report. SOBEL,P.

Division of Engineenng (Post 921004). April 1994. 108pp.

j NUREG-1484 V01: FINAL ENVIRONMENTAL IMPACT STATE-9405310204. 79589.235.

i l

Main Citations and Abstracts 11 The draft versson of th#s report presented updated Lawrence NRC staff concludes that the applicant's descriptons, specifca.

Livermore National Laboratory (LLNL) probabiliste seisme bons, and analysos provide an adequate basis for safety review hazard analysis estimates for 69 nuclear power plant sites in of facility operatons and that constructen and operation of the the regon of the United States east of the Rocky Mountains.

facility does not pose an undue nsk to public health and safety.

LLNL performed a re-eleitation of seismcity and ground motion experts to improve their estimates of uncertainty in seismcity NU9EG 1492 DFC: REGULATORY ANALYSIS ON CRITERIA parameters and ground moton models. Us#ng these revised FOR THE RELEASE OF PATIENTS ADMINISTERED RADIO-inputs. LLNL updated the seismic hazard estimates documented ACTIVE MATERIAL. Draft Report For Comment. SCHNEIDER,S.:

in NUREG/CR-5250 (1989). These updated hazard estimates MCGUIRE,S.A. Dwison of Regulatory Applications (870413-will be used in future NRC actions. The draft was issued for 941217). BEHLING,U.H.; et al. S. Cohen & Associates, Inc. May public comment in October 1993. By the end of the pubic com-1994. 78pp. 9406290327. 80014:210.

ment penod, February 28,1994, comments had been received The Nuclear Regulatory Commisson (NRC) has received two from two nuclear industry comparues. The comrnents from pebtions to amend its regulations in 10 CFR Parts 20 and 35 as these companies neither contested nor suggested amendments they apply to doses received by members of the public exposed to the techncal data conveyed in the report. Rather, they both to patients released from a hospital after they have been admin-suggest changes in the Individual Plant Extemal Event Examina-istered radioactue material. While the two petitons are not tion (IPEEE) program scope. This report is not the forum for dis-identical, they both request that the NRC establish a dose limit cussion of the IPEEE program. Possible modifcation to the of 5 mdlisseverts (0.5 rem) per year for indwiduals exposed to scope of the iPEEE will be examined in its own setting. There-patients who have been administered radioactue matenals. This fore, there are no technical differences between the draft report and this final report. Any informatm as to modifications to the Regulatory Analysis evaluates three attematwes. Alteratwe 1 is IPEEE program will be provided to the public via an NRC gener-for the NRC to amend its patent release enteria in 10 CFR al communicaton.

35.75 to use the more stnngent dose limit of 1 mdhsievert (0.1 rem) per year in 10 CFR 20.1301(a) for its patient release crite-NUREG 1449: A REVIEW OF NRC STAFF USES OF PROBABI-ria. Alternatue 2 is for the NRC to continue using the existing LISTIC RISK ASSESSMENT.

• Ofc of the Executive Drector for patient release enteria in 10 CFR 35.75 of 1,110 megabecquer-t Operations. March 1994. 274pp. 9405040223. 79107:001.

els (30 millicuries) of actwity or a dose rate at one meter from The NRC staff uses probabiliste nsk assessment (PRA) and the patient of 0.05 milhsievert (5 milhrems) per hour. Alternatue nsk management as important elements of its Icensing and reg-3 is for the NRC to amend the pahent release enteria in 10 CFR ulatory processes. In October 1991, the NRC's Executwe Drec-35.75 to specify a dose limit of 5 millisieverts (0.5 rem) for pa-tor for Operations established the PRA Working Group to ad-tient release. The evaluation demoristrates that, except for a dress concerns identrhed by the Advisory Committee on Reac-few diagnoste procedures using iodine-131, diagnoste proce-tor Safeguards with respect to unevenness and inconsistency in dures are unaffected by the choice of alternatwe. Only some the staff's current uses of PRA. After surveying current staff therapeute administratons of radcactive material could be af.

i uses of PRA and idenbfying needed improvements, the Working fected by the choce of altematue. The evaluation indcates that Group defined a set of base pnnciples for staff PRA use and Alternatue 1 would cause a prohibituely large increase in the identified three areas for improvements: guidance development.

national health care cost from retaining pabents in a hospital training enhancements, and PRA methods development. For longer and would cause signifcant personal and psychological each area of irnprovement, the Working Group took certain ac-costs to pabents and their famihes. The choce of Alternatives 2 tons and recommended additional work. The Working Group recouniia n.6d integrating its work with other recent PRA-related or 3 would affect only thyroid cancer pabents treated with iodine-131. For those pabents, Alternatwe 3 would result in less actwitnes the staff completed and improving staff interactons with PRA users in the nuclear industry. The Working Group took hospitalization than Alternatue 2. Altematwe 3 has a potential two key actons by developing general guidance for two uses of decrease in national health care cost of $30,000,000 per year PRA within the NRC (that is, screening or priontzing reactor but would increase the potential collectue dose from released safety issues and analyzing such issues in detail) and develop-therapy pabents by about 2,700 person-rem per year, mainly to ing guidance on basic terms and methods important to the family members. Alterative 3 would also have personal and psy-staff's uses of PRA.

chologcal benefits for the patients and their fam6hes.

NUREG-1491: SAFETY EVALUATION REPORT FOR THE CLAl-NUREG-1494: ST/JF TECHNICAL POSITION ON CONSIDER-BORNE ENRICHMENT CENTER, HOMER. LOUISIANA. Docket ATION OF FAULT DISPLACEMENT HAZARDS IN GEOLOGIC No. 70 3070, Louisiana Energy Services,L.P.

• Dvision of Fuel REPOSITORY DESIGN. MCCONNELL,K.l.; LEE.M.P. Dvision of Cycle Safety & Safeguards (Post 930207). January 1994.

Waste Management (NMSS 940403). September 1994. 37pp.

300pp. 9402150293. 78120:017, 9411080076. 81649-261.

This report documents the U.S. Nuclear Regulatory Commis-Nuclear Regulatory Commission regulabons for the disposal sion (NRC) staff safety review and evaluation of the Louisiana of spent nuclear fuel and high-level radioactwe waste in a geo-Energy Servces, LP. (LES, the apphcant) appicaton for a li-logic repository recognize that fault displacement is a potentially cense to possess and use byproducts, source, and special nu-adverse condition (10 CFR 60.122(c)(11) and 60.122(c)(20)).

clear matenal and to ennch natural uranium to a maximum of However, they do not prohibit designing the geologic repository fue percent U-235 by the gas centnfuge process. The plant, to against the effects of such a potentialty adverse conditon. This be known as the Claibome Ennchment Center (CEC), would be Staff Techncal Possten recognaes the acceptability of design-i constructed near the town of Homer in Claibome Pansh, Louisi-ing the geologc repository to take into account the attendant ana. At full producton in a gwen year, the plant will recewe ap-effects (e g., displacement) of faults of regulatory concem and proximately 4,700 tonnes of feed UF(6), and produce 870 expresses the staff's views on what is needed from the U.S.

tonnes of low-enriched UF(6), and 3,830 tonnes of depleted Department of Energy if it chooses to locate structures, sys-UF(6) tails. Facility constructon, operation, and decommission-tems, and components important to safety or important to waste j

ing are expected to last five, thirty, and seven years, respectue-isolation in areas that contain faults of regulatory concem.

fy. The objectue of the review is to evaluate the potenbal ad-verse impacts of operabon of the facihty on worker and pubhc NUREG-1495: OVERALL REVIEW STRATEGY FOR THE NUCLE-health and safety under both normal operabng and accident AR REGULATORY COMMISSION'S HIGH-LEVEL WASTE RE-conditons. The review also considers the management organi.

POSITORY PROGRAM. JOHNSON,R.L. Dwision of Waste Man-zaton, administratwe programs, and financial quahfcations pro-agement (NMSS 940403). November 1994. 39pp. 9412070075.

I vided to assure safe design and operaton of the facility. The 81956.200.

I

12 Main Citations and Abstracts The Overall Review Strategy gives general guidance to the On July 6, 1993, the Nuclear Regulatory Commission's Nuclear Regulatory Commisson staff for conducting it's hcense (NRC's) Executwe Duector for Operatons estabhshed a review apphcaton and pre-license apphcation reviews. These reviews team to reassess the NRC's program for protecting allegers tre in support of the Commission's construction autnoruation against retaliation. The team evaluated the current system, and decision for a geologic repository for the disposal of high-level sohcited comments from varcus NRC offices, other Federal radioactwo waste. Objectues and strategies are defined that agencies, former allegers, and the pubic. TNs report is subject focus the staff's reviews on determining compliance with re-to agency review. The report summanzes current processes and querements of 10 CFR Part 60. These strategies define how the gives an overview of current problems. It discusses: (1) ways in staff pnontizes its reviews on those key techncal uncertainties which licensees can promote a quahty-conscious work environ-considered to be most important to repository performance.

ment, an wNch all employees feel free to raise concerns without Strategies also give guidance for developing, in an integrated fear of retahation; (2) ways to improve the NRC's overall han-way, the License Appicaton Review Plan together with support-dling of allegatons; (3) the NRC's involvement in the Depart-ing performance assessments, analyses, and research, ment of Labor process; (4) related NRC enforcement practices; and (5) methods other than investigaton and enforcement that NUREG-1496 V1 DFC: GENERIC ENVIRONMENTAL IMPACT may be useful in treating allegatons of potential or actual dis-i STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOG-crimination. Recommendations are gwen in each area.

ICAL CRITERIA FOR DECOMMISSIONING OF NRC-LICENSED NUCLEAR FACILITIES. Main Report. Draft Report For Comment.

NUREG-1500: WORKING DRAFT REGULATORY GUIDE ON RE-

• Dmson of Regulatory Appleatons (870413-941217). August LEASE CRITERIA FOR DECOMMISSIONING. NRC STAFF'S 1994.170pp. 9408300003. 80700:001.

DRAFT FOR COMMENT. DAILY,M C.;

HUFFERT,A.M.;

l The action being considered in tNs Draft Generic Environ-CARDILE,F.; et al. Dwision of Regulatory Appications (870413-mental Impact Statement (GEIS) is an amendment to the Nucle-941217). August 1994.168pp. 9410130081. 81264:171.

ar Regulatory Commisson's (NRC) regulatons in 10 CFR Part The Nuclear Regulatory Commission's (NRC) regulations in 20 to include radiological cntena for decommissioning of lands 10 CFR Part 20 are being amended to include radiological ente-and structures at nuclear facilities. Under the National Environ-ria for decommissoning of lands and structures at nuclear facih-mental Pohey Act (NEPA), all Federal agencies must consider ties.10 CFR Part 20, Subpart E estabhshes cntena for the re-tne effect of ther actons on the environment. To fulfill NRC's mediation of contaminated sites or facihte that will allow their responsibihties under NEPA, the Commission es prepanng this release for future use with or with-out restrictons. The enteria GEIS whch analyzes afternatue courses of action and the include a Total Effectwe Dose Equivalent (TEDE) hmit of 15 costs and impacts associated with those alternatwes. In prepar-mrem / year (0.15 mSv/y) that should not be exceeded by an av-ing the GEIS, the following approach was takert (1) a listing erage indmdual among those who could potentially recewe the was developed of regulatory alternatues for establisNng radio-greatest exposure from any residual actrvity within a facility or logcal cntena for decommessening; (2) for each alternatwe, a on a site. The entena also requre a licensee to reduce any re.

detailed analysis and companson of incremental rnpacts, both sedual radioactmty to as-low as-reasonably-achievable (ALARA) radiological and nonradologcal, to workers, members of the levels. TNs staff draft guide desenbes acceptable procedures pubhc, and the envronment, and costs, was porformed, and (3) for determining the prodcted dose level (PDL) from any residual based on the analysis of impacts and costs, prehmanary recom-radioactmty at the site. It desenbes the basic features of the mendations were provided. Recommendatons contained in the calculational models and the associated default assumptions GEIS include those related to the definiton of decommissioning, and parameter vakses the NRC staff would find acceptable in the scope of rulemaking, the radiological entena, restnetons on calculating PDLs. Appendees A, B, and C provide numercal use, citizen partespation, use of the GEIS in site specife cases, values that can be used to estimate the dose from residual ra-and minimizat>on of contamination.

deactivity remaining at a site. Since 10 CFR Part 20, Subpart E introduces several new concepts, definitens and discussons NUREG-1496 V2 DFC: GENERIC ENVIRONMENTAL IMPACT are Wed in a relaton positen cmcepts sede of the STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOG-guide to assist heensees in understandog some of the philoso-ICAL CRITERIA FOR DECOMMISSIONING OF NRC-LICENSED phy undertying the rule.

NUCLEAR FACILITIES. Appendices Draft Report For Comment.

• Divison of Regulatory Appications (870413-941217). August NUREG-1501 DRFT: BACKGROUND AS A RESIDUAL RADIOAC-1994. 600pp. 9408300006. 80698:001.

TIVITY CRITERION FOR DECOMMISSIONING. Appendix A To See NUREG-1496,V01,DRF FC abstract-The Draft Genenc Environmental Impact Statement in Support NUREG-1497: INTERIM LICENSING CRITERIA FOR PHYSICAL Of Rulemaking On Radologeal Cntena For Decommissioning PROTECTION OF CERTAIN STORAGE OF SPENT FUEL.

Of NRC... HUFFERT,A M.; MECK,R.A. Dmsson of Regulatory DWYER.P.A. Dwison of Fuel Cycle Safety & Safeguards (Post Appleations (870413-941217). MILLER.K M.

Energy, Dept.of, 930207). November 1994. 20pp. 9412300178. 82159:182.

Environmental Measurements Laboratory. August 1994.100pp.

82 00 TNs document presents intenm entena to be used in the physical protection Icensing of certain spent fuel storaoe instal-draft U.S. Nuclear Regulatmy Commesson (NRC) document en-latons. Installations that will be reviewed under tNs entena are titled, "Genere Envronmental Impact Statement in Support of those that store power reactor spent fuel at decommissimed Rulemaking on Radological Critena for Decommissioning of l

power reactor sites; independent spent fuel storage instal!ations located outside of the owner controlled area of operating nucle-NRC-Leensed Nuclear Facilities." Because of the great interest in this report by members of the pubic, citizen and envronmen-ar power reactors; monitored retnevable storage installatons owned by the Department of Energy, designed and constructed tal aganizatons, academicians, heensees, and regulatus, the NRC staff is pubhsNng this report separately, so that it can be specifically for the storage of spent fuel; the proposed geologic readily available to a dwerse audience. This report was created repository operatons area; or permanently shutdown power re-to assist both the NRC staff and interested members of the actors still holding a Part 50 leense. This cntena in this docu-ment does not apply to the storage of spent fuel within the pubhc in evaluating background radiaton (background) as a de-conwmssoning cnwon, by serymg as a pnmer on background l

owner-controlled area of operating nuclear power reactors.

and providing information on the existing apphcations of back-l NUREG-1499: REASSESSMENT OF THE NRC'S PROGRAM ground in regulatory entena and standards. This report also dis-i FOR PROTECTING ALLEGERS AGAINST RETALIATION.

cusses some of the methods available to measure and distrn-l Ofc of the Executwe Drector for Operatons. January 1994.

guish between the very low radiaton levels associated with 200pp. 9402 t 50379. 78123.349.

background and man-made sources of radiaton. Two approach-i l

l

Main Citations and Abstracts 13 i

es are conssdered for applying background as a decommission-mobilize uranium found in the rock. Uranium would then be re-ing cnterion; these are the use of background dose rates and moved from the solution using ion exchange technology in proc-background radionuclide concentrations. This report concludes essing plants located in three areas. A central plant would pro-that the temporal and spatial variability of background produces vide drying and packaging equipment for the entire project. This a wide range of doses to United States residents, which pre-DEIS was prepared by an interagency review group, including vents the applicaton of background dose rates as a decommis-the Nuclear Regulatory Commiss#on, the Bureau of Irdian Af-sooning cntenon. Instead, this report recommends that local fairs, and the Bureau of Land Management. After weighing the background radionuchde concentrations serve as a benchmark environmental, technical, and other benefits of the Hydro Re-for decommissioning criteria, while taking into account the con-sources mining project against environmental and other costs, cept of reducing residual radcactivity to a level as low as is rea-the reviewing agencies concluded that the appropnate action is sonably achievable.

to issue the requested license and leases authorizing the appli-NUREG-1502: ASSESSMENT OF DATABASES AND MODELING cant to proceed with the project as discussed in this DEIS. The DEIS desenbes and evaluates (1) the purpose of and need for CAPABILITIES FOR THE CANDU 3 DESIGN. CARLSON,D.E/l the poposed acton, waluated under the Natonal EnWronme MEYER,R.O. Division of Systerns Research (880717 941217)

July 1994.107pp. 9408220041. 80637:099.

tal Policy Act, and the agencies' implement #ng regulatons, (2)

As part of the research program associated with the prelimi-ahnahes conshed, (3) asbng emmW Mons, and (4) environmental consequences of the proposed action nary review of the CANDU 3 design, the NRC Office of Nuclear Regulatory Research (RES) has completed an assessment of and proposed mitigating measures.

databases and modeling capabilities that might be needed to NUREG-1511: REACTOR PRESSURE VESSEL STATUS support the CANDU 3 design. To ensure full coverage of the REPORT. STROSNIDER.J.; WICHMAN.K.; ELLIOT,B.; et al. Di-design, a detasled assessment methodology was developed by vision of Engineenng (Post 921004). December 1994.174pp.

the RES staff and was implemented with help from research 9412300173. 82158:251.

projects at three natonal laboratones. This report entegrates The Nuclear Regulatory Commission (NRC) issued Generic and summanzes the database and modeling assessments, in-Letter 92-01 (GL 92-01) to obtain informaton needed to assess ciuding major contnbutons from these laboratones, compliance with requirements and commitments regarding reac-NUREG-1503 V01: FINAL SAFETY EVALUATION REPORT RE.

tor vessel integnty in view of certain concems raised in the NRC LATED TO THE CERTIFICATION OF THE ADVANCED BOIL-staff's review of reactor vessel integrity for the Yankee Nuclear ING WATER REACTOR DESIGN. Docket No.52-001.(General Power Station. This report gives a brief descripton of the reac.

Electric Nuclear Energy)

• Associate Director for Advanced Re-tor pressure vessel (RPV). followed by a discussion of the radi-actors & License Renewal (Post 910918). July 1994. 876pp.

ation embnttlement of RPV beltline matenals and the two indi-9408260011. 80681:001, cators for measunng embnttlement, the end-of-license (EOL)

This Safety Evaluation Report (SER) documents the technical reference temperature and the EOL upper shelf energy. It also rem of the U.S. Advanced Boiling Water Reactor (ABWR) summanzes the GL 92-01 effort and presents, for all 37 boiling standard design by the U.S. Nuclear Regulatory Commisson water reactor plants and 74 pressunzed water reactor plants in (NRC) staff. The applicaton for the ABWR design was submit-the United States, the current status of compliance with regula-ted by GE Nuclear Energy. The NRC staff concludes tnat, sub-tory requirements related to ensuring RPV integrity. The staff ject to satisfactory resoluton of the confirmatory items identified has evaluated the matenal data needed to predsct neutron em-in Secton 1.8 of this SER, GE's application for design certifica-bnttlement of the reactor vessel beltline materials. These data tion meets the requirements of Subpart B of 10 CFR Part 52 will be stored in a computer database entitled the reactor vessel that are applicable and technically relevant to the U.S. ABWR integrity database (RVID). This database will be updated annual-standard desagn.

ly to reflect the changes made by the licensees in future submit-als a e

sa assess N issws e NUREG-1503 V02: FINAL SAFETY EVALUATION REPORT RE-LATED TO THE CERTIFICATION OF THE ADVANCED BOIL-ING WATER REACTOR DESIGN. Appendices. Docket No. 52-NUREG/CP-0127: PROCEEDINGS OF THE CSNI SPECIALISTS 001.(General Electnc Nuclear Energy)

• Associate Director for MEETING ON FUEL-COOLANT INTERACTIONS.

• Organiza-l Advanced Reactors & License Renewal (Post 910918). July tion for Economic Cooperatior' & Development. March 1994.

1994. 209pp. 9408250023. 80659:098.

377pp.9404110363. NEA/CSNI/R(93)8. 78818:001.

l See NUREG 1503,V01 abstract-A specialists meeting on fuel-coolant interactions was held in NUREG-1504: REVIEW CRITERIA FOR THE PHYSICAL FITNESS Santa Barbara, CA from January 5-7, 1993. The meeting was TRAINING REQUIREMENTS IN 10 CFR PART 73. BROWN,C.

sponsored by the United States Nuclear Regulatory Commisson Division of Fuel Cycle Safety & Safeguards (Post 930207). Sep-in collaboration with the Committee on the Safety of Nuclear in-tember 1994.18pp. 9410130302. 81276:081.

stallation (CSNI) of the OECD Nuclear Energy Agency (NEA)

This document provides review cntena that will be used in re.

and the University of Califomia at Santa Barbara. The objectives viewing and approving revised physical secunty plans submitted of the meeting are to cross. fertilize on-going work, provide op-by licensees which are required to meet the physical fitness re.

portunities for mutual check points, seek to focus the technical querements in 10 CFR Part 73.

issues on matters of practical significance and re-evaluate both the objectives as well as path of future research.

NUREG-1508: DRAFT ENVIRONMENTAL IMPACT STATEMENT TO CONSTRUCT AND OPERATE THE CROWNPOINT SOLU-NUREG/CP-0133 V01: PROCEEDINGS OF THE TWENTY FIRST TION MINING PROJECT,CROWNPOINT, NEW MEXICO. Docket WATER REACTOR SAFETY INFORMATION MEETING. Plenary No. 40-8968.(Hydro Resources,Inc.)

• Drvison of Waste Man-Session; Advanced Reactor Research; Advanced Control agement (NMSS 940403). October 1994.154pp. 9411160064.

System Technology; Advanced instrumentaten & Control Hard-81767:008.

ware; Human Factors.... MONTELEONE,S. Brookhaven Natonal i

This Draft Environmental Impact Statement (DEIS) addresses Laboratory. April 1994. 600pp. 9405100230, 79217:001.

issuing a combined source and byproduct matenal license and This three-volume report contains 90 papers out of the 102 minerals operating leases for Federal and Indian lands to Hydro that were presented at the Twenty First Water Reactor Safety Resources, Inc. This acton would authorize the company to information Meeting held at the Bethesda Mamott Hotel, Be-l conduct in situ leach uranium mining in McKinley County, New thesda, Maryland, dunng the week of October 25-27,1993. The Mexico. Such mining would involve drilling wells to the ore papers are printed in the order of their presentation in each ses-bodies, then recirculating ground water with added oxygen to son and describe progress and results of programs in nuclear

14 Main Citations and Abstracts safety research conducted in this country and abroad. Foreign NUREG/CP-0137 V01: PROCEEDINGS OF THE THIRD NRC/

partcipation in the meeting included papers presented by re-ASME SYMPOSIUM ON VALVE AND PUMP TESTING. Held At searchers from France, Germany, Japan, Russia, Switzerland.

The Hyatt Regency Hotel, Washington,DC, July 18-21, Taiwan, and United Kirgdom. The titles of the papers and the 1994. Session 1 A Session 2C.

• EG&G Idaho, Inc. July 1994.

names of the authors have been updated and may differ from 570pp. 9408250007, EGG-2742. 80657:001.

those that appeared in the final program of the meeting.

The 1994 Symposium on Valve and Pump Testing, jointly NUREG/CP 0133 V02: PROCEEDINGS OF THE TWENTY-FIRST sponsored by the Board of Nuclear Codes and Standards of the WATER REACTOR SAFETY INFORMATION MEETING. Severe Amencan SocW of Mechadcal Engine and W h ma Accident Research. MONTELEONE,S. Brookhaven National Regulatory Comnwsson, provides a forum for the discussion o Laboratory. Ap il 1994. 632pp. 9405100236. 79220.001.

mnent programs and athods W msee MsW aM rne See NUREC,'CP-0133,V01 abstract ~

operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improvo NUREG/CP 01tJ V03: PROCEEDINGS OF THE TWENTY-FIRST that testing in order to help ensure the reliable performance of WATER REACTOR SAFETY INFORMATION MEETING.Pnmary pumps and valves. The parteipation of industry representatives, System Integnty; Aging Research, Products & Appleations; regulators, and consultants results in tho discussion of a broad Structural & Seismic Engineenng. Seismology & Geology.

spectrum of ideas and perspectrves regarding the improvement MONTELEONE S. Brookhaven Natonal Laboratory. April 1994.

of inservice testing of pumps and valves at nuclear power 454pp. 9405100242. 79218.236 planta See NUREG/CP-0133,V01 abstract.

NUREG/CP-0135: WORKSHOP ON ENVIRONMENTAL QUALIFl-NUREG/CP-0137 V02: PROCEEDINGS OF THE THIRD NRC/

ASME SYMPOSIUM ON VALVE AND PUMP TESTING Held At CATION OF ELECTRIC EQUIPMENT.

LOFARO,R.;

De Hyan Regercy Hotel Washington M M 18 21, GUNTHER,W.; VILLARAN,M.; et al Brookhaven National Labo-ratory. May 1994. 309pp. 9406060111. BNL-NUREG-52409, 1994.Sesson 3A -Session 48., EG&G Idaho, Inc. July 1994.

79619:001 387pp. 9408030114. EGG-2742. 80424:197.

Questioris concoming the Environmental Quahficaton (EO) of See NUREG/CP.0137,V01 abstract.

electrical equipment used in commercial nuclear power plants NUREG/CP-0138: PROCEEDINGS OF WORKSHOP 1 IN AD-have recently become the subject of significant interest to the VANCED TOPICS IN RISK AND RELIABILITY ANALYSIS.Model U S. Nuclear Regulatory Comnwssion (NRC). Initial questions centered on whether comphance wrlh the EO requirements for Uncertainty:

its Characterization And Quantification.

older plants were adequate to support plant operaton beyond MOSLEH,A.; SMIDTS,C.; et al. Maryland, Univ. of. College Park, MD. SIU,N. EGaG Idaho, Inc. October 1994. 250pp.

40 years. After subsequent investgaton, the NRC Staff con-9411220239 81806.025 cluded that questions related to the differences in EO require ments between older and newer plants constitute a potential This report contains the proceedings of an internatonal work-generic issue whsch should be evaluated for backfit, independ-shop: Model Uncertainty: Its Charactenzation and Quantification, held October 20-22, 1993 in Annapolis, Maryland, USA. It in-ent of license renewal actrwties. EO testing of electric cables was performed by Sandia National Laboratories (SNL) under ciudes: i) a brief introduction to the tope of model uncertainty contract to the NRC in support of Icense renewal activities. Re-and to the workshop; li) contributed papers from 17 experts suits showed that some of the environmentally quahfied cables from a variety of organizatons (govemment agencies, research either failed or exhibited marginal insulaton resistance after a laboratories, universities, and pnvate industry), disciphnes (phys-simulated plant hfe of 20 years dunng accident simulation. This ics, engenng, stabsfics, law, medcme, managemed s&nce),

indicated that the EO process for some electnc cables may be and areas of interest (nuclear' reactor safety, waste repository non-conservative. These results raised questions regarding the performance, fire cafet/, human hea!!h, environmental nsk, deci-EO process including the bases for conclusons about the quali-sion analyses, and uncertainty analysis); and iii) summanes of fied hfe of components based upon artifcial aging prior to test-discussions of three working groups formed dunng the work-ing~

shop. The pnmary findings of the workshop as documented in these proceedings are as follows. First, a formal probabilistm NUREG/CP-0136: PROCEEDINGS OF THE DIGITAL SYSTEMS framework for dealing with model uncertainty exists; some fur.

RELIABILITY AND NUCLEAR SAFETY WORKSHOP. September ther development is needed to make it more practical. Socond, 1314,1993,Rockvdle Crowne Plaza Hotel,Rockvdle, Maryland.

less formal quantitahve approaches also exist and have been WALLACE.D.R.; CUTHILL.B.B.; IPPOLITO.L.M.; et al. Natonal apphed in some problems. Third, coping strategies whch don't Institute of Standards & Technology (formerfy National Bureau require direct quantification of modst uncertainty are available of Standa. March 1994. 373pp. 9405050310. NIST SP 500-216.

for immediate usage. Fourth, quantitative procedures that " aver-79163.001.

age out" irreconcilable differences between models and model-The Urvted States Nuclear Regulatory Commission (NRC), in ers do not provide suffeient informaton regarding predictive un-cooperation with the National Institute of Standards and Tech-certainty to nsk managers. A pnontzed research agenda for de-nology conducted the Digital Systems Reliabihty and Nuclear veloping a standardized approach for treating model uncertainty Safety Workshop on September 13-14,1993, in Rockville, Mary-is presented.

land. The workshop provided a forum for the exchange of infor-maton among experts within the nuclear industry, experts from NUREG/CP-0139: TRANSACTIONS OF THE TWENTY SECOND other industnes, regulators and academia. The informaton pre-WATER REACTOR SAFETY INFORMATION MEETING.

sented at this workshop provided in-depth exposure of the NRC MONTELEONE,S. Brookhaven National Laboratory. October staff and the nuclear industry to digital systems design safety 1994.141pp. 9411040088. 81647:001.

issues and also provided feedback to the NRC from outside ex.

This report contains summanes of papers on reactor safety perts regarding identified safety issues, proposed regulatory po-research to be presented at the 22nd Water Reactor Safety in-sstens, and intended research associated with the use of digital formation Meeting at the Bethesda Mamott Hotel, Bethesda, systems in nuclear power plants. Techncal presentatens pro-Maryland, October 24 26,1994. The summanes bnefly describe vided insights on areas where current software engineenng the programs and results of nuclear safety research sponsored practces may be inadequate for safety cntcal systems, on po-by the Office of Nuclear Regulatory Research, U.S. NRC. Sum-tential solutons for development issues, and on methods for re-maries of invited papers concerning nuclear safety issues from ducing nsk in safety cntical systems. Thts report contains an U.S. government lateratones, the electre utihties, the nuclear analysis of results of the workshop, the papers presented, panet industry, and from foreign govemments and industry are also in-presentatens, and summanes of discussons at this workshop.

cluded. The summanes have been compiled in one report to a

l l

Main Citations and Abstracts 15 1

l provide a basis for meaniVd rtscussioa and information ex.

NUREG/CR 3950 V09: FUEL PERFORMANCE REPORT FOR change dunng the course of the meeting and are given in the 1991. PAINTER,C L.; ALVis,J.M.; BEYER,C E.; et al. Battelle order of their presentaten in each session.

Memorial institute, Pacife Nethwest Laboratory. AuGJst 1994.

)

123pp. 9409200326. PNL-5210. 69963:162.

NUREG/CR-2850 V12: DOSE COMMITMENTS DUE TO RADIO-This annual report, the fourteentil in a senes, provides a bnef ACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES descripton of fuel performance dinng 1991 in commercial nu-IN 1990. BAKER.D.A. Battelle Memonal Institute, Pacific North.

c,uar power plants and an indcaton of trends. Onef summaries west Laboratory. November 1994.194pp. 9412120234. PNL.

c' mi operating expenenc?, fuel design changes, fuel surveil-

)

4221. 81997:001, laice programs, high-burnup expenence, problem areas, and j

Population and individual radiation dose ccmmitments have items of general signifbance are provided. References to more

]

been estimated from reported radionuchde releases from com.

detailed enformaton and re:ated U.S. Nuclear Regulatory Com-j mercial power reactors operating dunng 1990. Fifty-year dose misson evaluations are included.

commitments for a one-year exposure from both hquid and at' NUREG/CR-4219 V10 N1: HEAVY SECTION STEEL TECHNOL-mosphenc releases were calculated for four populaton groups (infant, child, teen-ager and adult) residing between 2 and 80 OGY PROGRAM. Semiannual Progress Report For October 1992 March 1993. FENNELL,W.E. Oak Ridge Natonal Labora-km from each of 72 reactor sites. This report tabulates the re-tory. September 1994.149pp. 9410120224. ORNL/TM-9593.

l suits of these calculatons, showing the dose commitments for 81254:001.

both water and airborne pathways for each age group and The Heavy-Secton Steel Technology (HSST) Program is con-organ. Also included for each of the sites is an estimate of indi-ducted for the Nuclear Regulatory Commisson by Oak Ridge vidual doses which are compared with 10 CFR Part 50, Apperb National Laboratory (ORNL). The program focus is on the de-dix 1 design obectives. The total collective dose commitments l

velopment and vahdation of technology for the assessment of (from both hquid and airbome pathways) for each site ranged fracture-prevention margins in commercial nuclear reactor pres-from a high of 15 person-rem to a low of 0.002 person-rem for sure vessels. The HSST Program is organized in 12 tasks: (1) the sites with plants in operaton and producing power canng program management, (2) fracture methodology and analysis, l

the year. The anthmetic mean was 1.1 person-rem. The total (3) material charactenzaton and properties, (4) special techncal l

populaton dose for all sites was estimated at 78 persorvrem for assistance (5) fracture analysss computer programs, (6) cleav.

j the 130 millon people considered at nsk. The individual dose age-crack initiation, (7) cladding evaluatons, (8) pressurized-l commitments estimated for all sites were below the Appendix 1 thermal shock technology, (9) analysis methods vahdaten, (10) design objectives.

fracture evaluatons, (11) warm prestressing, and (12) biaxial 1

loading effects. The program tasks have been structured to

%fREG/CR-2907 V12: RADIOACTIVE MATERIALS RELEASED place emphasis on the resoluton fracture issues with near-term FROM NUCLEAR POWER PLANTS. Annual Noort 1991.

hcensing signifcance. Resources to execute the research tasks TICHLER J.; DOTY,K.: CONGEMI,J. Brookhaven mal Labo-are drawn from ORNL with subcontract support from universi-ratory. May 1994. 360pp. 9406210302. BNL-4G-51581.

ties and other research laboratones. Close contact is main-i 79830 001.

tained with the sister Heavy.Secton Steel Irradiation Program at i

Releases of radioactive matenals in airborne a x1 houid ef-ORNL and with related research programs both in the United fluents from commercial hght water reactors dunng 1991 have States and abroad. This report provioes an overview of pnncipal been compiled and reported Data on solid waste shipments as developments in each of the 12 program tasks from October well as selected coerating information have been included. This 1992 to March 1993.

report supplements earher annual reports issued by the forme; Atome Energy Commisson and the Nuclear Regulatory Com.

NUREG/CR-4409 V05:[mTA BASE ON DOSE REDUCTION RE.

mission. The 1991 release data are summanzed in tabular form.

SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

Data covenng specifc radonuchdes are summarized.

KHAN,T.A.; YU C.K. Brookhaven National Laboratory. May 1994.193pp.9406200329. BNL-NUREG-51934. 79831:100.

I NUREG/CR-3145 V10: GEOPHYSICAL INVESTtGATIONS OF This is the fifth volume in a senes of reports that provide in-THE WESTERN OHIO. INDIANA REGION Final Report, October formaton on dose reduction research and health physcs tech-1986-September 1992. RUFF,L; LAFORGE,Ra THORSON.R.;

nologv W nuclear power plants. The information is taken from et al. Mehegan, Uruv. of, Ann Arbor, MI. Fetwa v 1994.115pp.

two M several databases maintained by Brookhaven Natonal 9403140217.78456.242.

Laboratory's ALARA Center for the U.S. Nuclear Regulatory Earthquake activity in the Wester, Ohio-India,1 regon has CommissKn. The research secten of the report covers dose re-been monitored with a seismogrgin network cons sting of nine ducton p'oiects that are in the expenmental or developmental statens located in west-centr :a Ohc and four stat, ins located phase. l' includes topes such as ste2m generator degradaten, in Indiana. Six local and regional earthquakes have been re-decontaminaton, robotes, improvements in reactor materials, corded from October 1990 to September 1992 with maprutudes and inepecton techniques. The secten on health physics tech-l ranging from 0.6 to 5.0. A total of 36 local and regional earth-nology discusses domdumn ems mat are m plaw m in quakes have been recorded in the past 6-year penod (October N p cess d W WemenW at dar pm #anh A 1986 to September 1992). Overall a total of 78 local and region-total of M5 mw w @ated polects am WM M poM a as am a a ar Mm al earthquakes have been recorded since the network went into es nals e access 2 a fax mW Wh w operaton in 1977. There was a peak in seismicity in 1986, irk ACEFAX system or a computer with a modem and the proper ciuding the July 12,1986 St. Marys* event (mb=4 5), followed communicatons software through our ACE system. Detailed de-by an anomalously low level of seismicity for about 2 years. The senptons of how to access all of our databases electronically most unusual feature of the seismcity in the past year is the are in the appendices of the report.

occurrence of three earthquakes in Indiana. The locatons of the felt earthquakes are scattered across central Indiana; an area NUREG/CR-4513 R01: ESTIMATION OF FRACTURE TOUGH-the had been aseismc. Analysis of amval time data accumulat-NESS OF CAST STAINLESS STEELS DURING THERMAL ed o er the past 14 years shows that the Anna regon crustal AGING IN LWR SYSTEMS. CHOPRA,0.K. Argonne Natonal

}

structure is " slower" than the average mid-continent crustal Laboratory. August 1994. 83pp. 9409230297. ANL 93/22.

structure. This imphes that the proposed Keewenawan nft in the 81010:229.

Anna repon has a different structure than that of other This report presents a revison of the procedure and correla-Keewenawan nfts in the rnsd-contsnent.

tons presented earlier in NUREG/CR-4513, ANL-90/42 (June i

I

16 Main Citations and Abstracts 1991) for predicting the change in mechanical properties of cast cracked pipe evaluations involved (1) material charactenzation stainless steel components due to thermal aging dunng serwce of B&W C-Mn-Mo submerged arc weld metal, and (2) 3D finite.

in light water reactors at 280-330 degrees C (535-625 degrees element mesh refinement study. The toughness of the bimetallic F). The correlations presented in this report are based on an weld fusion line was evaluated and showed unusual fracture be-esponded databene and have been opamazed with mechanical-havior based on the results of the Charpy tests. The dynamic P'eportr data on cast stamises stooie aged up to =58,000 h at strain aging J R tests confirmed the screening cntenon devel-i 2 ell 350 doyees C (554 633 degrees F). The correlations for oped earlier in the program. The results from this program to I

estimating the change in tensile stress, including the Ramberg/

date necessitated several additional efforts. These were initiat-Osgood parameters for strain hardening, are also desenbed.

ed and have been reported here. Presentation of the results The fracture toughness J-R curve, tensile stress, and Charpy*

from this program to the ASME Section XI Pipe Flaw Evaluation impact energy of aged cast stainless steels are estimated from Working Group is also summarized here.

known matenal information. Mechanical properties of a specific cast stainless steel are estimated from the extent and kinetics NUREG/CR-4639 V5R4P2: NUCLEAR COMPtJTERIZED Li-of thermal embnttlement. Embnttlement of cast stainless steels BRARY FOR ASSESSING REACTOR RELIABILITY is characterized in terms of room-temperature Charpy-impact (NUCLARR). Volume 5: Data Manual.Part 2: Human Error Proba-onergg The extent or degree of thermal embnttlement at "satu-bihty (HEP) Data. REECE W.J.; GILBERT.B.G.; RICHARDS.R.E.

ration, i.e., the minimum impact energy that can be achieved EG&G Idaho, Inc. September 1994. 453pp. 9410130115. EGG-for a material after long-term aging, is determined from the 24 81265M chemical composition of the steel. Charpy-impact energy as a i

This data manual contains a hard copy of the,nformation in function of time and temperature of reactor service is estimated the Nuclear Computerized Library for Assessing Reactor Reis-from the kinetics of thermal embnttlement, which are also deter, mined from the chemical composition. The initial impact energy abihty (NUCLARR) Version 3.5 database, which is sponsored by of the unaged steel is regured for these estimations.

the U.S. Nuclear Regulatory Commission. NUCLARR was de-signed as a tool for nck analysis. Many of the nuclear reactors NUREG/CR-4551 V01 R1: EVALUATION OF SEVERE ACCI-in the U.S. and several outside the U.S. are represented in the a

se.

s M Man m TAINME T RCE T M CON EOUENCE AND RIS pr bability estimates for workers at the plants and hardware fail.

INTEGRATION ANALYSES. GORHAM.E.D.; BREEDING.R.J.

ure data for nuclear reactor equipment. Aggregations of these Sandia National Laboratones. HELTON.J.C.; et al Arizona State data yield valuable reliability estimates for probabilistic risk as-Univ., Tempe, AZ, December 1993. 279pp. 9402220215.

sessments and human reliabiiity analyses. The data manual is SAND 861309. 78198:001.

organized to pd mamal searches of me informahon d me NUREG-1150 examines the risk to the pubhc from five nucle-ar power plants. The NU9EG-1150 plant studies are Level lli computerized version is not available, probabilistic nsk assessments (PRAs) and, as such, they consist NUREG/CR-4639 V5R4P3: NUCLEAR COMPUTERIZED L1-of four analysis components: accident frequency analysis, acci-BRARY FOR ASSESSING REACTOR RELIABILITY dent progisssion analysis, and source term analysis, and conse-(NUCLARR). Volume 5: Data Manual.Part 3: Hardware Compo-quence analysis. This volume summarizes the methods utilized nent Failure Data. REECE,W.J.; GILBERT,0.G.; RICHARDS R.E.

in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG.

EG&G idaho, Inc. September 1994.186pp. 9410130099. EGG-2458.81267:001.

1150 analysis approach is i,6:4,o on the following ideas: (1) general and relatwnty fast-runnsng models for the individual See NUREG/CR-4639,V05,R04,P02 abstract.

analysis components, (3) use of Monte Carlo techniques togeth-er with an efficient sampling procedure to propagate uncertain.

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACK-ties (4) use of expert panels to develop distnbutions for impor.

ING IN LIGHT WATER REACTORS. Semiannual Report,Apnl tant phei,ve,v;vg cal issues, and (5) automation of the overall 1993 - September 1993. CHOPRA.O.K.; CHUNG,H.M.;

analysis. Many features of the new analysis procedures were KARLSEN,T.; et at Argonne National Laboratory. June 1994.

adopted to facilitate a comprehensive treatment of uncertainty 70pp. 9407130212. ANL-94/16. 80189:133.

in the complete nsk analysis. Uncertainties in the accident fre-This report summanzes work performed by Argonne National quency, accident progression and source term analyses were Laboratory on fatigue and environmentally assisted cracking included in the overall uncertainty assessment. The uncertain-(EAC) in light water reacto s (LWRs) dunng the six months from ties in the consequence analysis were not included in this as-Apnl 1993 to September 1993. EAC and fatigue of piping, pres-sessment. A large effort was devoted to the development of sure vessels, and core components in LWRs are important con-procedures for obtaining expert opiruon and the execution of the cems as extended reactor lifetimes are envisaged. Topics that development of procedures for obtaining expert opinion and the have been investigated include (a) fatigue of low-alloy steel execution of these procedures to quantify parameters and phe-used in piping, steam generators, and reactor pressure vessels, nomena for which there is large uncertainty and divergent opin-(b) EAC of cast stainless steels (SSs). and (c) radiation-induced ions in the reactor safety community.

segregation and irradiation-assisted stress corrosion cracking of NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPING AND Type 304 SS after accumulation of relatrvely high fluence. Fa-PIPING WELDS.Serniannual Report, October 1992 - March tigue tests were conducted on medium-sulfur content A106-Gr 1993. WILKOWSKI,G.M.; BRUST,F.; FRANCINI,R.; et al. Bat.

B piping and A533-Gr 8 pressure vessel steels in simulated telle Memonal institute, Columbus Laboratories. March 1994.

PWR water and in air. Additional crack growth data were ob-112pp. 9404040024. BMI-2173. 78736:256.

tained on fracture-mechanics specimens of cast austenitic SSs This is the sixth semiannual report of the U.S. Nuclear Regu.

in the as-received and thermany aged condetsons in simulated latory Co.. es.c6's 4 year research program "Short Cracks in boiling-water reactor (BWR) water at 289 degrees C. The data Piping and Piping Welds" which began in March 1990. The ob-were compared with predictions based on crack growth correla-jective is to venfy and improve fracture analyses for circumfer-tions for wrought austenitic SS in oxygenated water developed entially cracked nuclear piping with cracks sizes typically found at ANL and rates in d: Irem Section XI of the ASME Code. Mi-during in-service flaw evaluations. Progress it the through-wall-croctwmical and microstructual changes in high-and commer-cracked pipe efforts involved (1) venfication of deformation cial-punty Type 304 SS specwnens from control-blade absorber plasticity under av,ip upvruunal loading, (2) evaluation of the tubes and a control-blade sheath from operating BWPs were effect of weld metal strength on various J-estimation schemes, studied by Auger electron spectroscopy and scanning electron and (3) development of new GE/EPRI fuiw::tions. Surface-n-ascmcc,p,

Main Citations and Abstracts 17 NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE idation of embnttfement prediction models by researchers. The CORE DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Main Power Reactor Embnttlement Data Base (PR-EDB) is such a Report And Appendix A.

COX,D.F.;

CLETCHER.J.W.;

comprehensive collection of data for U.S. commercial nuclear COPINGER.D.A.; et al. Oak Ridge Natonal Laboratory. Decem-reactors. The current version of the PR-EDB contains the ber 1993.150pp. 9402220223. ORNL/NOAC.232. 78198.280.

Charpy test data that were irradiated in 252 capsules of 96 re-Twenty-seven operational events with conditonal probabilities actors and consists of 207 data points for heat-affected. zone of subsequent severe core damage of 1.0 x 10E-06 or higher (HAZ) matenals (98 different HAZ), 227 data points for weld occurnng at enmmercial hght-water reactors dunng 1992 are matenals (105 different welds) 524 data points for base materi-considered to be precursors to potential core damage. These als (136 different base matenals), including 297 plate data are desenbed along with associated signifcance estimates, cat-egonzation, and subsequent analyses. The report discusses (1) points (85 different plates),119 forging data points (31) different the general rationale for this study, (2) the selection and docu-forging), and 108 correlation monitor materials data points (3 mentation of events as precursors, (3) the estimaton and use of different plates). The data files are given in dBASE format and conditonal probabilities of subsequent severe core damage to can be accessed with any personal computer using the DOS rank precursor events, and (4) the plant models used in the operating system. " User. friendly" utility programs are used to anahsis process.

retrieve and select specific data, manipulate data, dsplay data to the screen or pnnter, and to fit and plot Charpy impact data.

NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE The results of several studies investigated are presented in Ap-CORE DAMAGE ACCIDENTS:

1992 A

STATUS pendix D.

REPORT. Appendices B, C. D, E, F. And G. COX,D.F.;

CLETCHER,J.W.; COPINGER D.A.; et al. Oak Ridge Natonal NUREG/CR-4833: LARGE AREA SELF-POWERED GAMMA RAY Laboratory. December 1993. 688pp. 9402220241. ORNL/

DETECTOR. Phase il Development Of A Source Position Moni.

NOAC-232. 78194:299.

tor For Use On Industrial Radographic Units. LEVERT,F.E.

See NUREG/CR-4674,V17 abstract.

K.E.M.P. Corp. January 1994. 47pp. 9402150301. 78122:053.

NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE The purpose of this research was to develop a large area CORE DAMAGE ACCIDENTS: 1993 A STATUS REPORT. Main self. powered gamma detector (LASPGD) capable of detecting Report And Appendices A D.

VANDEN HEUVEL,L; the movement of sealed radiation sources into and out of indus-CLETCHER.J.W. COPINGER D.A.; et al. Oak Ridge National trial radiographic units and to construct a prototype source posi-Laboratory. September 1994. 200pp. 9411220300. ORNL/

tion monitor (SPM) for these units utilizing the LASPGD. Proto-NOAC-232. 81806.272.

type isotropic and directional LASPGDs, with solid and inert gas Sixteen operational events that affected sixteen commercial dielectncs, were developed and extensively tested using cali-light-water reactors during 1993 and that are considered to be brated gamma sources (i.e., Cs-137 and Co-60). The sensstrvi-precursors to potenbal severe core damage are desenbed. All ties of the isotropic detectors, with inert gas dielectncs, were these events had conditonal probabilities of subsequent severe found to be approximately a factor of ten greater than those were identified by first computer. screening the 1993 heensee tive self-powered detectors were found to exhibit a forward-to-core damage greator than or equal to 1.0 x 10(6). These events measured for the solid dielectnc LASPGDs. Directionally sensi-event reports from commercial hght-water reactors to identify back hemispherical sensitivity rato of approximately 2 to 1. In-those that could potentm!!y be precursors. Candidate precursors dustrial radographic units containing tr-192 sources with differ.

l were then selected and evaluated in a process similar to that ent activities were used to test the performance on the SPM.

l used in previous assessments. Selected events underwent engi.

The SPM, which utihzed a gas dielectnc LASPGD, performed as I

neering evaluaton that identified, analyzed, and documented designed. That is, the current generated in the LASPGD was the precursors. Other events dessgnated by the Nuclear Regula-tory Commesson (NRC) also underwent a similar evaluaton. Fe-converted to a voltage, amplified and used to control the on/off nally, documented precursors were submitted for review by h-state of an incandescent lamp. The incandescent lamp, which censees and NRC headquarters and regonal offices to ensure functons as the source /out warning indicator, flashes at a rate the plant dessgn and its response to the precursor were correct-of one flash per second when the source is in use (i e. out of its shield).

ly charactenzed. This study is a continuaton of earher work, which evaluated 1969-1981 and 19841992 events. The report NUREG/CR-4838: MICROCOMPUTER APPLICATIONS OF, AND discusses the general ratonale for this study, the selecton and MOOLFICATIONS TO. THE MODULAR FAULT TREES.

documentation of events as precursors, and the estimation of ZIMMERMAN,T.L.; GRAVES,N.L Energy, Inc. PAYNE,A.C.; et conditional probabdrbes of subsequent severe core damage for al Sandia Natonal Laboratones. October 1994. 300pp.

eents. This document rs bound in two volumes: Volume 19 9412070078. SAND 88-1887. 81956.241.

comns tne main report and Appendices A-D; Volume 20 con-tan Appendices E and F.

The LaSane Probabihstic Risk Assessment was the first major apphcation of the modular loge fault trees after the IREP pro.

NURLG/CR-4674 V20: PRECURSORS TO POTENTIAL SEVERE gram. In the process of performing the analysis, many errors CODE DAMAGE ACCIDENTS:

1993 A

STATUS were discovered in the fault tree modules that led to diffculties REPORT.Apps9dences E And F.

VANDEN HEUVEL,L; in combining the modules to form the final system fault trees.

CLETCHER,J.W.; COPINGER.D.A.; et at Oak Ridge National These errors are corrected in the revised modules hsted in this Laboratory. September 1994. 400pp. 9411220304. ORNL/

report. In additon, the applicaton of the modules in terms of NOAC-232. 81807:120.

editing them and forming them into the system fault trees was See NUREG/CR-4674,V19 abstract.

ineffcient. Two programs were wntten to help alleviate this NUREG/CR-4816 R02: PR-EDB: POWER REACTOR EMBRIT-Problem: (1) MODEDIT - This program allows an operator to re.

TLEMENT DATA BASE. VERSION 2.

Program Description, trieve a file for editing, edit the file for the plant specific apphca-STALLMANN.F.W.; WANG.J.A.; KAM F.B.K.; et al. Oak Ridge tion, perform some general error checking while the file is being National Laboratory. January 1994.122pp. 9402150283. ORNL/

modified, and store the file for later use, and (2) INDEX - This TM-10328. 78121:034.

program checks that the modules that are supposed to form investigatons of regulatory issues such as vessel integnty one fault tree all knk up appropriately before the files are loaded over plant hfe, vessel failure, and suffciency of current codes, onto the mainframe computer. Lastly, the modu 6es were not de-Standard Review Plans (SRPs) and Guides for heense renewal signed for relay type loge common in BWR designs but for solid can be greatly expedited by the use of a well-designed comput-state type logic. Some additonal modules were defined for enzed data base. Also, such a data base is essential for the val-modehng relay logic.

18 Main Citations and Abstracts NUREG/CR-4918 V07: CONTROL OF WATER INFILTRATION of evaluating sampling performance was to determine the effec-INTO NEAR SURFACE LLW DISPOSAL UNITS. Progress tiveness of the sampling plan for detecting and plugging defec-Report On Field Expenments At A Humid Region twe tubes. A summary of key results from the eddy current reli-Site,Beltsville. Maryland. SCHULZ,R.K. Cahfomia, Univ. of, Los abihty studies is presented. The analytcal and Monte Carlo sim-Angeles, CA. RIDKY,R.W. Maryland, Urw. of, College Park, MD.

ulation analyses are discussed along with a synopsis of key re-O'DONNELLE. Waste Management Branch (Post 941217). De-suits and conclusions.

cember 1994. 34pp. 9501180233. 82341:180.

The project oblective is to assess means for controlling waste NUREG/CR-5229 V06: FIELD LYSIMETER INVESTIGATIONS:

infiltration through waste disposal unit covers in humid regions.

LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM Expenmental work is being performed in large scale lysimeters FOR FISCAL YEAR 1993. Annual Report. MCCONNELL.J.W.;

(70'x45'x10') at Beltsville MD and results of the assessment ROGERS,R.D.; JASTROW,J.D.; et al. EGaG Idaho, Inc. May are apphcable to disposal of LLW, uranium mill tailings, hazard-1994. 63pp. 9406070051. EGG-2577. 79684:208.

ous waste, and sanitary landfills. Three concepts are under in.

The Field Lysimeter investigabons: Low-Level Waste Data veshgabon: (1) resistwo layer barrier, (2) conductue layer bar-Base Development Program, funded by the U.S. Nuclear Regu-rier, and boengineenng water management. The resistue layer latory Commisson, is (a) studying the degradaten effects in bamer consists of compacted earth (clay). The conductue layer EPICOR-il organc ion-exchange resins caused by radiaton, (b) barrier is a special case of the capillary bamer and it requires a examsnsng the adequacy of test procedures riw,w,eJwd in flow layer (e.g. fine sandy loam) over a capillary break. As long the Branch Technical Position on Waste Form to meet the re-as unsaturated conditions are maintained water is conducted by quirements of 10 CFR 61 using solidtfied EPICOR-il reams, (c) the flow layer to below the waste. This bamer is most effcient obtaining performance informaton on solidified EPICOR il ion-at low flow rates and is thus best placed below a resistive layer exchange resins in a disposal environment, and (d) determining bamer. Such a combinabon of the resistue layer over the con-the condition of EPICOR-il liners. Results of the eighth year of ductwe layer bamer promises to be highty effectue provided data acquisition from the field testing are presented and dis-there is no appreciable subsidence. Bioengineenng water man-cussed. During the continuing field testing, both Portland type i-agement is a surface cover that is designed to accommodato 11 cement and Dow vinyt ester-styrene we forms are being suossdence. It consists of impermeable panels whch enhance tested in lysimeter arrays located at Argonne Natonal Laborato-run-off and limet infiltraton. Vegetaton is planted in narrow ry-East in lilinois and at Oak Ridge Natonal Laboratory. The openings between panels to transpire water from below the study is designed to provide continuous data on nuchde release panels. This system has successfully dewatered two lysimeters and movement, as well as envwonmental conditions, over a 20-thus demonstrating that this procedure could be used for reme-year period.

dial acton (" drying out") existing water logged disposal sites at low cost NUREG/CR-5247 VD1 R2: RASCAL VERSION 2.1 USER'S GUIDE. SJOREEN.A.L Oak Ridge Natonal Laboratory.

NUREG/CR-5128 R01: EVALUATION AND REFINEMENT OF ATHEY,G.F. Athey Consulting. RAMSDELL J.V.; et al. Battelle LEAK-RATE ESTIMATION MODELS. PAUL D.D.; AHMAD.J.;

Memorial Institute, Pacife Northwest Laboratory. December SCOTT P.M.; et al. Battelle Memonal Institute, Columbus Lab-1994.196pp. 9501180194. ORNL-6820. 82345:001.

oratones. June 1994. 97pp. 9407260077. BMI-2164. 80345:001.

Leak-rate estimaton models are important elements in devel-The Radiological Assessment System for Consequence Anal-oping a leak-before-break methodology in piping integnty and ysis, Verson 2.1 (RASCAL 2.1) has been developed for use safety analyses. Existing thermal-hydraulc and crack opensng-dunng response to radological emergencies. The system sup-area models used m current leak-rate estimatons have been in-piemenis assessments based on plant conditons and quck es-timates cased on hand-calculabonal methods. The model is de-corporated into a sangle computer code for leak rate eshmaton.

The code is called SOUIRT, which stands for Seepage Ouantifi-signed to provide a companson to EPA Protectus Acton Guid-ance and thresholds for acute health effects. The model was cation of Upsets in Reactor Tubes. The SOUIRT program has been vahdated by comparing its thermal-hydrauhc predctions developed to allow consideraton of the dominant aspects of with the hrruted expenmental data that have been pubhshed on source term, transport, dose, and consequences. Revisions to two-phase flow through shts and cracks, and by companng its RASCAL 2.0 have required the release of this new verson of I

crack-opening-area predctons with data from the Degraded the system. Thrree new source-term calculatons have been Piping Program. In ad6 ton, leak-rate expenments were con-added to ST-DOSE ir, HASCAL 2.1. They are (1) source term ducted to obtain vahdaten data for a circumferential fatque based on the reactor containment morutor reading (2) a source crack m a carbon steel pipe girth weld.

term for a spent fuel pool accident, and (3) an esotopic concen-tration source term. Field Measurements to Dose (FM-DOSE) i NUREG/CR-5161 V02: EVALUATION OF SAMPLING PLANS calculabons have been modfied to include consideration of the FOR INSERVICE INSPECTION OF STEAM GENERATOR effect of delay for re-entry on first-year and second-year dose.

TUBES. Comprehensive Analytcal And Monte Carlo Simulation to incorporate a variable resuspenson rate, and to compute a Results For Several Samphng Plans.

KURTZ R.J.;

factor used to estimate first-year dose from R/hr measurements HEASLER,P.G.; BAIRD.D.B. Battelle Memonal institute, Pacific on the ground. Also, RASCAL 2.1 supports the savmg of cases Northwest Laboratory. February 1994. 72pp. 9404040034. PNL-for later display or modifcabon and the use of a mouse for user 6462. 78736:057.

input.

l This report summanzes the results of three prevous studies l

to evaluate and compare the effectiveness of samphng plans NUREG/CR-5247 V02 R2: RASCAL VERSION 2.1 WORKBOOK.

for steam generator tube inspections. An analytcal evaluaton ATHEY,G.F. Athey Consutting. SJOREEN,A L, Oak Ridge Na-and Monte Carlo simulaton techniques were the methods used tonal Laboratory. MCKENNA,T.J. Incident Response Branch.

to evaluate samphng plan performance. To test the perform-December 1994.148pp. 9501180200. 82345:197, ance of canddate sampling plans under a vanety of con 6tions.

The Radiological Assessment System for Consequence Anal-ranges of inspection system reliability were considered along ysis, Version 2.1 (RASC.Al 2.1) was developed for use by the with different dstnbutons of tube degradaten. Results from the NRC personnel who rrespond to radological emergencies. This eddy current rehabihty studes performed with the retired-from-workbook complements the RASCAL 2.1 User's Guide service Surry 2A steam generator were utskred to guide the se-(NUREG/CR-5247. Vol.1, Rev. 2). The workbook contams ex-lecten of appropriate probabshty of detection and flaw sizing ercises designed to famihanze the user with the computer.

models for use m the analysis. Different dstnbutons of tube based tools of RASCAL through hands-on problem solving. The degradaten were selected to span the range of conditons that workbook contains four maior sectons. The first RASCAL famil-might exist in operabog steam generators. The pnncipal means iarization exercise to acquaint the user with the operabon of the l

l I

i

Main Citations and Abstracts 19 forms, menus, on!Me help, and documentation. The latter three cause of the extensive amount of available information, the 1

sections contain exercises in using the three tools of RASCAL report was divided into two parts. Part i discusses cyclic internal Version 2.1: DECAY, FM-DOSE, and ST-DOSE. A discussion pressure loading and Part 11 discusses moment loadings Mr the section desenbing how the tools could be used to solve the branch and run. The cyclic pressure loading fatigue parameta problems follows each set of exercises.

are mostly based on leakage, whereas, if the parameters were NUREG/CR-5314 V05: INSIGHTS FOR AGING MANAGEMENT based on crack initiation, different and possibly higher valves OF MAJOR LWR COMPONENTS METAL CONTAINMENTS.

would be developed. The fatigue evaluation procedure, which SHAH,V.N.; SINHA,U.P. Idaho National Engineering Laboratory, attempts to relate fatigue strength of pipsng components to SMITH,S.K. Ogden Environmental & Energy Services (Formerly strain-controlled, pokshed bar, and fatigue data appears to be Multiple Dynamics Corp). March 1994.104pp. 9404060267.

inaccurate on the conservative side for high amplitude cycles EGG-2562. 78760:001, and inaccurate on the unconservatve side for low amplitude This report evaluates the available technical information and cycles.

field expenence related to management of aging damage to hght water reactor metal containments. A genenc aging man.

NUREG/CR-5403: PREDICTING THE PRESSURE CalVEN FLOW agement approach is suggested for the effective and compre.

OF GASES THROUGH MICRO-CAPILLARIES AND MICRO-hensive aging management of metal containments to ensure ORIFICES. ANDERSON.B.L; CARLSON.R.W.; FISCHER,LE.

their safe operation. The major concern is corrosion of the em.

Lawrence Livermore National Laboratory. November 1994.

bedded porton of the containment vessel and detection of this 22pp.9411220229. UCRL-ID-118245. 81806:001.

damage. The electromagnebc acoustc transducer and half-cell Expentnentally measured gas flow rates obtained from the lit-potential measurement are potential techniques to detect corro-erature were compared to the predictions obtained using the sion damage in the embedded portion of the contanment constitutive flow equations given in ANSI N14.5, the "American vessel. Other corrosion-related concerns include inspecton of Natonal Standard for Radcactive Materials-Leakage Tests on corrosion damage on the inaccessible side of BWR Mark I and Packages for Shipment," to determine whether the equations 11 containment vessels and corrosion of the BWR Mark I torus apply to the predctions of gas flow rates from leaking contain-and emergency core cooling system peping that penetrates the ment vessels used to transport radioactive rnatenals. The equa-torus, and transgranular stress corrosion cracking of the pene-tions were applied to both (1) the data set according to the rec-traton bellows. Fabgue-related concerns include reducten in ommendations given in ANSI N14.5, and (2) globally to the the fabgue hfe (a) of a vessel caused by roughness of the cor-complete data set. It was found that For flow rates 51 roded vessel surface and (b) of bellows because of any physcal etm a cm(3)/s, the predctions obtained with the continuum and damage. Maintenance of surface coatings and sealant at thE molecular flow equation provided good agreement with the ex-metal-concrete interface is the best protecton against corrosion of the vesset perimental values and the choked flow equaton resulted in over.predicion of the measured values. For flow rates > 1 NUREG/CR-5344 R01: REPLACEMENT ENERGY COST ANALY.

atm em(3)/s, the molecular and conhnuum flow equation over-SIS PACKAGE (RECAP): USER'S GUIDE. VANKUlKEN,J.C.;

predicted the measured flow rates and the predctions obtained i

WILLING,D.L Argonne National Laboratory. July 1994. 54pp.

with the choked flow equation agreed well with the experimental 9407250267. A.NL/EES-TM-364. 80328.141.

values. Since leakage rates from packages used to transport ra-A mscrocomputer program called the Replacement Energy deactive matenals are almost always

<1 atm cm(3)/s, it is j

Cost Analysis Package (RECAP) has been developed to assist suggested that the continuum and molecular flow equaton be j

the U.S. Nuclear Regulatory Commisson (NRC) in determining used for gas flow rate predictions related to these apphcations.

the replacement energy costs associated with short-term shut-downs or derabngs of one or more nuclear reactors. The calcu-NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DE-i I

latons are based on the seasonal, unit-specific cost estimates GRADED SHEAR WALL STIFFNESSES ON SEISMIC PLANT

)

for 1993-19% previously pubhshed in NRC Report NUREG/CR-RISK AND SEISMIC DESIGN LOADS. KLAMERUS,E.W.;

4012, Vol. 3 (1992), for alt 112 U.S. reactors. Because the BOHN.M.P. Sandia National Laboratones. JOHNSTON,J.J.; et i

RECAP program is menu-dnven, the user can define specife al. EOE Engineering Consultants (formerly EOE Engineering, case studies in terms of such parameters as the units to be in.

Inc.). February 1994. 588pp. 9404010183. SAND 93-0234.

cluded, the length and timing of the shutdown or derating 78720:042.

penod, the urut capacity factors, and the reference year for re-Test results sponsored by the USNRC have shown that rein-porting cost results. In additon to simuttaneous shutdown forced shear wall (Seismic Category I) structures exhibit stiff-cases, more complicated situatons, such as overlapping shut-nesses and natural frequencies which are smaller than those en penods or shutdowns that occur in different years, can be calculated in the design process. The USNRC has sponsored examn'ed through the use of a present-worth calculaton opton.

Sandia Natonal Labs to perform an evaluaton of the effects of NUREG/CR-5359: REVIEW OF ELASTIC STRESS AND FA-the reduced frequencies on several existing seismic PRAs in TIGUE-TO-FAILURE DATA FOR 6 RANCH CONNECTIONS order to determ,ne the seismic risk implications inherent in AND TEES IN RELATION TO ASOE DESIGN CRITERIA FOR these test results. This report presents the results for the re-NUCLEAR POWER PIPING SYSTEMS. RODABAUGH.E.C.;

evaluaton of the seismic risk for three nuclear power plants: the l

MOORE,S.E.; GWALTNEY,R.C. Oa6 Ridge Natonal Laboratory.

Peach Bottom Atomic Power Staten, the Zion Nuclear Power May 1994.124pp. 9406060115. ORNL/TM 11152. 79618 200.

Plant, and Arkansas Nuclear One - Unit 1 (ANO-1). Increases in This is the third in a series of reports on the state-of-the art core damage frequencies for seismic initiated events at Peach design guidance for piping system branch connections and tees Bottom were 25 to 30 percent (depending on whether LLNL or provided by Secton ill of the ASME Boiler and Pressure Vessel EPRI hazard curves were used). At the ANO-1 site, the corre-Code. The other reports covered pnmary or limit. loads and sponding increases in plant nsk were 10 percent (for each set nozzle flexibility. The pnncipal objective of this report, as with of hazard curves). Finally, at Zion, there was essentially no the others, was to identify needed improvements in the design change in the computed core damage frequency when the re-methods and entena of the Code based on the evaluaton of the duction in shear wall stiffness was included. In additon, an eval-available information. This report does not propose changes in uaton of determanistic " design-like" structural dynamic calcula-i the design procedure of the Code. This report discusses the tons with and without the shear stiffness reductons was made.

evaluaton of stresses in branch connectons and tees, correla-Deterministic loads calculated for these two cases typically in-ton of these stresses with fatigue failures, and the code rules creased on the order of 10 to 20 percent for the affected struc-for protection against fatigue failure in design appleations. Be-tures.

20 Main Citations and Abstracts NUREQ/CR-5535 V06:

RELAPS/ MOD 3 CODE and power reactors, nuclear medicine, and other industries that MANUAL. Validation Of Numencal Techniques In RELAPS/

either process or use nuclear materials.

MOD 3. SHIEH,A.S. EG&G Idaho, Inc. RANSOM,V.H. Purdue Univ., West Lafayette, IN. KRISHNAMURTHY Pacific-Nuclear NUREG/CR 5591 V02 N1: HEAVY-SECTION STEEL IRRADIA-Co. October 1994.154pp. 9411160091. EGG-2705. 81759:138.

TION PROGRAM. Semiannual Progrest. Report For October 1990 - March 1991. CORWIN,W.R. Oak Rdge National Labora-The RELAP code has been developed for best-estimate tran-sient simulation of light-water reactor coolant systems during tory. July 1994. 36pp. 9408180191. ORNL/TM-11568.

large and small break loss-of-coolant accidents and as well as 80613:001.

operational transients. The code models the coupled behavior The primary goal of the Heavy-Section Steel Irradiation Pro-of the reactor coolant system and the core dunng a severe ac-gram is to provide a thorough, quantitative assessment of the cident transient and models large-and small-break loss-of-effects of neutron irradiation on the material behavior, and in coolant accidents and operational transents, such as anticipat.

particular the fracture toughness properties, of typical pressure ed transient without scram, loss of offsite power, loss of feed-vessel steels as they relate to light water reactor presyure-water, and loss of flow. A genenc rnodeling approach is used vessel integrity. Effects of specimen size, material chemestry, r

that permits as much of a paricular system to be modeled as product form and microstructure, irradiation fluence, flux, tem-necessary. Control system and secondary system components perature and spectrum, and post-irradiation anneahng are being are included to permit modehng of plant controls, turbines, con-examined on a wide range of fracture properties. Analyses of densers, and secondary feedwater conditioning systems.

precleavage stable ductile tearing of high-copper welds 72W l

RELAPS/ MOO 3 code documentation is divided into five vol-and 73W demonstrated that the size effects observed in the umes Volume I provdes modehng theory and associated nu-transition region are not due to substantial differences in ductile i

merical schemes; Volume 11 contains detailed instruct:ons for teanng behavior. Drop. weight tests, Charpy impact tests, and code appication and input data preparation; Volume ill provides chemical analyses of the Mdiand reactor low upper-shelf welds the results of developmental assessment cases that demon-were completed showing large vanations in bulk copper con-strate and venfy the models used in the code; Volume IV pre-tent, transition temperature, and upper-shelf energy. Atom probe sents a detailed discussion of MLAPS models and correlations; field ion mcroscopy analyses revealed no evidence of fine Volume V contains guidelines tl,P We evolved over the past copper precipitates or clusters in the unirradiated Midland wclds several years through use of the RELAPS code; and Volume Vi but a substantial depletion in the copper concentration in the contains desenptions of numencet modehng of two-phase flow matnx.

used in RELAp 5 and discussions on statxhty, accuracy, and convergence of the numerical techniques in RELAPS.

NUREG/CR-5591 V02 N2: HEAVY-SECTION STEEL IRRADfA-TION PROGRAM. Semiannual Progress Report For April-Sep-NUREG/CR-5535 V07-RELAPS/ MOO 3 CODE temba 199L CMENE Oak Ridge Nahal Laha%

MANUALSummanes And ' Reviews Of Independent Code As-Octoba 1994.

37pp.

M 040083.

@NW-W8.

sessment Reports. SLOAN.S.M.; SCHULTZ.R.R.; WILSON,G.E.

81 EG ho, Inc. June 1994.120pp. 9407260037. EGG-2596.

The nkary goal of the Heavy-Section Steel irradiation Pro-gram is to provide a thorough, quantitatrve assessment of the Summanes of RELAP5/ MOO 3 code assessments, a listing of effects of neutron irradiation on the material behavior, and in the assessment matrix, and a chronology of the vanous ver-partcular the hache toughness prop @s, of Wcal presswe sions of the code are grven. Results from these code assess-vessel steels as they relate to light-water reactor pressure-ments have been used to formulate a compilation of some of vessel integnty. Effects of specimen size, material chemistry, the strengths and weaknesses of the code. These results are product brm and microstructure, irradiation fluence, flux, tem-documented in the report. Volume 7 was designed to be updat-perature and spectrum, and post-irradiation annealing are being ed periodically and to include the results of the latest code as-examaned on a wide range of fracture properties. The HSSI Pro-sessments as they become available. Consequently the user of gram is arranged into 10 tasks: (1) program management, (2)

Volume 7 should check that the latest revision is available.

K(lc) curve shift in high-copper welds, (3) K(la) curve shift in NUREG/CR-5569 R01: HEALTH PHYSICS POSITIONS DATA high-copper welds, (4) irradiation effects on cladding. (5) K(ic)

BASE. KERR,G.D.; BORGES.T.; STAFFORD.R.S.; et al. Oak and K(la) curve shifts in low upper-shelf welds, (6) irradiation ef.

Ridge National Laboratory. February 1994. 300pp. 9403140176-fects in a commercial low upper-shelf weld. (7) microstructural ORNL/TM 12067. 78453:001-analysis of irradiation effects, (8) in-service aged material eval-The Health Physics Positions (HPPOS) Data Base of the Nu-uations, (9) correlation monitor matenals, and (10) special tech-clear Regulatory Commission (NRC) is a collection of NRC staff nical assistance. This reiport provides an overview of the actrvi-positions on a wide range of topics involving radiation protection ties within each of these tasks from April to September 1991.

(health phys.cs). It consists of 328 documents in the form of let-NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSED ters, memoranda, and excerpts from technical reports. The DECAY HEAT GUIDE USING SAS2H/ORIGEN-S DATA.

HPPOS Data Base was developed by the NRC Headquarters HERMANN.O.W.; PARKS.C.V.; RENIER.J.P. Oak Ridge National and Regional Offices to help ensure uniformity in inspections, Laboratory. September 1994.130pp. 9411080038. ORNL-6698.

enforcement, and licensing actions. Staff members of the Oak 8 N 231.

Ridge National Laboratory (ORNL) have assisted the NRC staff Major revisions are proposed to the current U.S. Nuclear Reg-in summanzing the documents during the preparation of this ulatory Commission decay heat rate guide entitled " Regulatory NUREG report. These summaries are also being made available Guide 3.54, Spent Fuel Heat Generation en an Independent as a " stand alone" software package for IBM and IBM-compati-Spent Fuel Storage Installation, using a new data base pro-ble personal computers. The software package for this report is SAS2H anaW sequence of the SCAM ca!!ed HPPOS Version 2.0. A variety of indexing schemes were system. The data base for the proposed guide revision has used to increase the usefulness of the NUREG report and its

• • • O I'#

associated software. The software package and the summanes 9' *

• in the report are written in the context of "new" 10 CFR Part 20 actor (PWR) and boiling-water reactor (BWR) spent fuel assem-(H20.1001 - 20.2401). The purpose of this NUREG report is to bhes. Using generic PWR and BWR assembly models, calcula-allow interested individuals to familiartze themselves with the tions were performed with each model for six different bumups contents of the HPPOS Data Base and with the basis of many at each of three separate specific powers to produce heat rates NRC decisions and regulations. The HPPOS summaries and at 20 cooling times in the range of 1 to 110 y. The proposed -

original documents are intended to serve as a source of infor-procedure specifies proper interpolaton formutae for the tabu-mation for radiation protection programs at nuclear research

Main Citations and Abstracts 21 lated heat genersbon rates. Adjustment formulae for the inter.

This report summarizes the data from the semiannual reports 1

polated values are provided to account for dfferences in initial on Fitness for Duty programs subrnetted to the NRC by utihties (235)U enrichment and changes in the specife power of a cycle for two reportmg penods: January 1 through June 30,1993, and from the average value. Finally, safety factor formulae were de-July 1 through December 31,1993. During 1993, licensees re-rived as a function of burnup, cooling twne, and type of reactor.

ported that they had conducted 242,966 tests for the presence The proposed guide revision was designed to be easier to use.

of illegal drugs and alcohol. Of these tests, 1,512 (.62%) were Also, the complete data base and guide procedure is incorpo.

confirmed positrve. Positive test results vaned by category of i

rated into an interactrve code called LWRARC which can be ex-test and category of worker. The malonty of positive test results ecuted on a personal computer The report shows adequate (952) were obtained through pre-access training. Of tests con-cv m4,ie00s of heat rates computed by SAS2H/ORIGEN-S and ducted on workers having access to the protected area, there i

measurements for 10 BWR and 10 PWR fuel assembbes. The were 341 positive tests from random testmg and 163 positive average differences of the computed rnmus the measured heat tests from for cause testmg. Follow-up testng of workers who rates of fuel assembhes were -07 2 2.6% for the BWR and 1.5 had previously tested positive resulted in 56 positive tests. For-21.3% for the PWR In addtion, a detailed analysis of the pro-cause testing resulted in the highest percentage of poestive posed procedure indcated the method and equatens to be tests: about 22 percent of for-cause tests were positive. This vahd.

compares with a positrve test rate of 1.04 percent of pre-access j

NUREG/CR-5640 V01: THE IMPACT OF ENVIRONMENTAL tests and.23 percent of random tests. Positive test rates also CONDITIONS ON HUMAN PERFORMANCE. A Handbook Of varied by category of worker. When all types of tests are com-bined (pre-access, random, for-cause, and follow-up testing),

1 Environmental Exposures. ECHEVERRIA.D.; BARNES V.;

BITTNER,A.; et al Battelle Human Affairs Research Centers.

short-term contractor personnel had the highest positive test September 1994.110pp. 9411080032. 81650-001.

rate at.97 percent. Licensee employees and long-term contrac-j A comprehensive review of the techncal hterature was co*

tors had lower combined posttrve test rates (.25% and.21%, re-I ducted regardng the impact of environmental condtons on Spectively). Of the substances tested, marijuana was responsi-human performance applicable to nuclear power plant workers.

ble for the highest percentage of positive test results (49.56%),

1 The environmental constions considered were vibraten, noise, followed by cocaine (23.41%) and alcohol (22.65%).

heat, cold, and hght. Research staff identified potential human NUREG/CR-5412: MANAGING AGING IN NUCLEAR POWER performance defeits (e.g., decreased dextenty, impaired visson, PLANTS. Insights From NRC Maintenance Team 16,Ki;vn Re-heanng loss, memory defciency) along a continuum of increas-ports. FRESCO,A.; SUBUDHI,M.; GUNTHER,W.; et al. Brookha-ing occupational exposure, ranging from exposures that result in ven National Laboratory. December 1993. 204pp. 9402220158.

no defest to exposures that resulted in signifcant performance BNL-NUREG-52309. 78196:259.

problems. Specife defeits were included in the report if there A plant's maintenance program is the p-incipal vehicle was sound scientifc evidence that environmental exposure re-through which age-related degradation is managed. From 1988 suited in those performance defeats. The levels associated with to 1991, the NRC evaluated the maintenance program of every each defeit were then compared to the protect >on afforded by nuclear power plant in the United States. Forty-four out of a existmg occupational exposure standards. Volume i is a hand-total of 67 of the reports issued on these ir> depth team inspec-book for use by NRC mspectors to help them determine the tions were reviewed for insights into the strengths and weak-impact of specific environmental conditions on hcensee person-nesses of the programs as related to the need to understand nel performance. It discusses the units used to measure each and manage the effects of aging on r 4 lear plant systems, I

condition, dscusses the effects of the condetson on task per-structures, and cumpvnents. Relevant information was extract-

[

formance, presents an example of the assessment of each con-ed from these inspection reports and sorted into several cate-dition in a nuclear power plant, and discusses potential methods gones, including Specific Aging insights, Preventive Mainte-j for reducmg the effects of exposure to the condtiort nance, Predictive Maintenance and Condition Monitoring, Post NUREG/CR-5600 V02: THE IMPACT OF ENVIRONMENTAL Maintenance Testing, Failure Trending, Root Cause Analysis CONDITIONS ON HUMAN PERFORMANCE. A Cntcal Review and Usage of Probabihste Risk, Assessment in the Maintenance Of The Literature. ECHEVERRIA.D.; BARNES,V.; BITTNER A.;

ocess. SWe examples of mspecton and Wonng tech 4

et al. Battelle Human Affairs Research Centers. September nws smessW used by utshs to dotect degradaten due 1994. 248pp. 9411080026. 81650:109.

to agmg have been ideN he mfonnaton also was sorted 1

See NUREG/CR-5680,V01 abstract.

accadng to systems and wmnus, incbdv Auxnary i

Feedwater, Main Feedwater, High Pressure Intection for both NUREG/CR-5726: REVIEW OF THE OtABLO CANYON PROB BWRs and PWT;s, Sennce Water, instmment Air, and Emergen.

ABillSTIC RISK ASSESSMENT.

BOZOKI.G.E.;

cy Diesel Generator Air Start Systerns, cnd emergency diesel FITZPATRICK,R. Brookhaven Natonal Laboratory. BOHN,M.P.;

generators, electrical components such as vwitchgear, breakers, et al. Sanda National Laboratones. August 1994. 500pp.

relays, and motor control centers, motor o:ierated valves and 9409260069. BNL-NUREG-52288. 81028:001.

check valves. This informaton was compare d to insights gained i

This report details the review of the Diablo Canyon Probabilis-from the Nuclear Plant Aging Research (FPAR) Program. At-tc Risk Assessment (DCPRA). The study was performed under tributes of plant maintenance programs whe e the NRC inspec-contract from the Probabihste Risk Analysis Branch, Offee of tors felt that improvement was needed to property address the Nuclear Reactor Research, USNRC by Brookhaven National aging issue also are discussed.

Laboratory. The DCPRA is a full scope Level I effort and al-though the review touched on all aspects of the PRA, the mter.

NUREG/CR-5830: AUXILIARY FEEDWATER SYSTEM RISK.

nel events and seisme events received the vast majonty of the BASED INSPECTION GUIDE FOR THE MCGUIRE NUCLEAR review effort. The report includes a number of independent sys, POWER PLANT.

BUMGARDNER.J.D.;

LLOYD,R.C.;

tems analyses, sensitivity studies, importance analyses as well MOFFITT,N.E.; et al. Battelle Memorial Institute, Pacific North-as conclusions on the adequacy of the DCPRA for use in the west Labmatory. May 199( 36pp. 9406200324. PNb7784.

Diablo Canyon Long Term Seesr7ic Program.

na sponsored by the U.S. Nuclear Regulatory Com.

NUREG/CR-5754 V04: FITNESS FOR DUTY IN THE NUCLEAR mission (NRC), Pacifc Northwest Laboratory has developed and POWER INDUSTRY. Annual Surnmary Of Program Performance applied a methodology for denving plant-specifc risk-based in-Reports CY 1993. WESTRA,C,; FORSLUND,C.; FIELD,l.; et al.

specton guidance for the auxiliary feedwater (AFW) system at Battelle Memorial institute, Pactfe Northwest Laboratory. August pressurized water reactors that have not undergone probabiliste 1994.101pp. 9409230281, ML 9985. 81006:184.

risk assessment (PRA). This methodology uses existmg PRA re-

22 Wlit Citations and Abstracts suits and plant operahng experience information. Existing PRA-quahfcation issues. The study also identified optcal fiber sys-based inspecton guidance informaton recently developed for tems as technologies that are relatively new to the nuclear the NRC for vanous plants was used to identify genenc compo-power plant environment, and examined the failure modes and nent failure modes. This informabon was then combined with age-related degradaten mechanisms associated with fiber optic plant specifc and industry-wide component informaton and fail-cables and components. The data were then used to propose a ure data to identify failure modes and failure mechanisms for methodology for identifying circumstances in which accelerated the AFW system at the selected plants. McGuire was selected aging should be used in an equipment qualifcation program for as one of a senes of plants for study. The product of this effort "new" I&C technologies. An analysis of the Icensee event is a pnontized hstng of AFW failures whch have occurred at report database over a 10-y period (19821991) performed the plant and at other PWRs. This listing is intended for use by under this study showed that the fracton of EMI/RFl related NRC inspectors in the preparaton of inspecton plans address-protechon system events is signifcant compared to traditionally ing AFW nsk-important components at the McGuire plant.

recogrijzed environmental stressors such as elevated tempera-Wra pr em is Weh to be won mee sigWant b N NUREG/CR 5850: ANALYSIS OF LONG-TERM STATION safety systems due to the increased use of macroprocessor.

BLACKOUT WITHOUT AUTOMATIC DEPRESSURIZATION AT based echnology and software. Thus, it appears that while PEACH BOTTOM USING MELCOR (VERSION 1.8). MADNI,l.K.

sa% bysWns s AMs wd have to W quam to N sam Brookhaven National Laboratory. May 1994.

127pp.

envkmment as current LWRs, EMI/RFl emissions and suscepti-9406210270. BNL-NUREG-52319. 79837:001.

bility critena and guidehnes specifc to the nuclear power plant This report documents the results from MELCOR calculatons wmment SM M co@d WM WE Wh of the Long-Term Staton Blackout Accident Sequence, with fail-ments are addressed in a companion document, NUREG/CR-ure to depressunze the reactor vessel, at the Peach Bottom echage and RaMmecy Wem in (BWR Mark 1) plant, and presents compansons with Source Safety Cntcal l&C System.

Term Code Package calculatons of the same sequence. STCP has calculated the transient out to 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after core unco-NUREG/CR-5908 V01: ADVANCED HUMAN SYSTEM INTER-very. Most of the MELCOR calculatons presented have been FACE DESIGN REVIEW GUIDELINE. General Evaluaton ccmed out to between 15 and 16.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after core uncovery.

Model. Techncal Development, And Gudehne Desenpton.

The results include the release of source terms to the environ-O'HARA,J.M. Brookhaven Natonal Laboratory. Jufy 1994.

ment. The results of several sensibvity calculatons with 140pp.9408250026. BNL-NUREG 52333. 80660:115.

MELCOR are also presented, whch explore the impact of vary-Advanced control rooms wdl use advanced human-system ing user-enput modehng and timestep control parameters on the interface (HSI) technologies that may have signifcant impica-accident progression and release of source terms to the envi-tons for plant safety in that they will affect the operator's over-ronment. Most of the calculatons documented here were per-all role in the system, the method of informaton presentation, formed in FY1990 using MELCOR Version 1.8BC. However, the and the ways in which operators interact with the system. The appendices also document the results of more recent calcula-U.S. Nuclear Regulatory Commission (NRC) reviews the HSI as-tons ormed in FY1991 using MELCOR versions 1.8CZ and pects of control rooms to ensure that they are designed to good human factors engineenng principles and that operator perform-NUREG/CR-5861: CRACK-SPEED RELATIONS INFERRED ance and rehabdity are appropnately supported to protect puble FROM LARGE SINGLE-EDGE-NOTCHED SPECIMENS OF A health and safety The pnncipal guidance available to the NRC, 533 B STEEL SCHWARTZ,CW. Oak Rdge Nabonal Laborato-however, was developed more than ten years ago, well before ry. July 1994. 32pp. 9408180237. ORNLSUB79-77789.

these technological changes. Accordingly, the human factors 80595:312.

guidance needs to be updated to serve as the basis for NRC A relatonship between instantaneous crack-tip velocity, a dy-review of these advanced designs. The purpose of this project name stress-intensity factor,K(I), and temperature T for A 533 B was to develop a general approach to advanced HSI review and steel is estimated using dynaraic crack position vs bme data the human factors gudelines to support NRC safety reviews of measured in a senes of very large-scale crack-arrest tests. The advanced systems. This two-volume report provdes the results corresponding dyname stress intensity vs time history and the of the project. Volume 1 desenbes the development of the Ad-dynamic-arrest toughness for each test are obtained from gen-vanced HSI Design Review Guidehne (DRG) including (1) its erat on-mode elastodyname analyses based on cubc polynomi-theoretical and techncal foundation (2) a general model for the al fits to the discrete crack-positen data poents. Apphcation review of advanced HSis, (3) gudehne development in both mode elastodyname anatytcal predcitons based on the pro-hard-copy and computer-based versons, and (4) the tests and posed a-K(1)-T relaton are within 7% of expenmentally meas-evaluatens performed to develop and vahdate the DRG.

ured arrested crack lengths and within 50% of measured arrest Volume 1 also includes a discusson of the gaps in available bmes. These predctons represent signifcant improvements guidance and a methodology for addressing them. Volume 2 over resutts obtained using previous prehminary eshmates of provides the guidehne to be used for advanced HSI review and the a-K(1)-T relabon for A 533 8 steel. The influence of nonlin-the procedures for their use.

NUREG/CR-5908 V02: ADVANCED HUMAN. SYSTEM INTER-NUREG/CR-5904: FUNCTIONAL ISSUES AND ENVIRONMEN-FACE DESIGN REVIEW GUIDELINE. Evaluaton Procedures TAL QUALIFICATION OF DIGITAL PROTECTION SYSTEMS And Guidehnes For Human Factors Engineenng Reviews.

OF ADVANCED LIGHT WATER NUCLEAR REACTORS-O'HARA.J M ; BROWN.W.S. Brookhaven Natonal Laboratory.

KORSAH.K.; CLARK,R.L.; WOOD R.T. Oak Rdge Natonal Lab-BAKER,C.C.; et al. Cartow internatonal, Inc. July 1994. 277pp.

oratory. Apol 1994. 87pp. 9405310256. ORNL/TM-12164 9408250004. BNL-NUREG-52333. 80656:001.

79588:251.

See NUREG/CR-5908,V01 abstract.

A study of signifcant "new" technologies proposed for use in safety-relcied instrumentabon and controls (l&C) systems of ad-NUREG/CR-5919: REPOSITORY OPERATIONAL CRITERIA vanced hght-water reactors (ALWRs) was performed as part of COMPARATIVE ANALYSIS.

HAGEMAN,J.P.;

the Quahfcaton of Advanced instrumentaten and Control Sys.

CHOWDHURY,A.H. Center for Nuclear Waste Regulatory Analy-tems protect conducted for the Office of Nuclear Regulatory Re-ses. June 1994. 69pp. 9406290301. CNWRA 92 007.

search of the U.S Nuclear Regulatory Ccas,s.On (NRC). Tem-80013:065.

plates showing dagttal protection systems oi some ALWR de-The objective of the " Repository Operatonal Cntena (ROC) signs and the c%ct of expected envirorvnental stressors on Feasibihty Studies" (or ROC task) was to conduct comprehen-sys?em components were developed to Lustrate functional and sive and integrated analyses of repository design, constructon,

Main Citations and Abstracts 23 and operabons entena in 10 CFR Part 60 regulations consider.

nents in each system that could be subject to aging were ac-ing the interfaces among the components of the regulabons and counteJ for in the model to simulate the time-dependent effects impacts of any potenbal changes to those regulations. The ROC of ar,ng degradation, assuming no provisions are made to prop-task addresses regulatory entena and uncertainbes related to erty nnanage it. System unavailability as a function of increasing the preclosure aspects of the geologic repository. Those parts component failure rates was then calculated.

of 10 CFR Part 60 that require routine guidance or minor changes to the rule were addressed in Hageman and Chowd-NUREG/CR-5941: TECHNICAL BASIS FOR EVALUATING ELEC-hury,1992. The ROC task shows a possible need for further TROMAGNETIC AND RADIO-FREQUENCY INTERFERENCE regulatory clanty, by major changes to the rule, related to the IN SAFETY-RELATED l&C SYSTEMS.

EWING,P.D.;

design bases and siting of a geologic repository operabons area KORSAH K. Oak Ridge Natonal Laboratory. April 1994. 44pp.

and radologeal emergency planning in order to assure de.

9405310198. ORNL/TM-12221. 79590:001.

l fense-in-depth. The analyses, presented in this report, resulted This report discusses the development of the techncal basis l

in the development and refinement of regulatory concepts and for the control of upets and malfunctions in safety-related in-their supporting ratonale for recommendatons for potenbal strumentation and contel (l&C) systems caused by electromag-t major changes to 10 CFR Part 60 regulations.

netic and radio-frequency interference (EMI/RFI) and power NUREG/CR-5935:

SUMMARY

OF WORK COMPLETED UNDER surges. The research was performed at the Oak Ridge Natonal THE ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUALIFl-Laboratory (ORNL) and was sponsored by the U.S. NRC Offee CATION RESEARCH PROGRAM (EDOP). STEELE R f Nuclear Regulatory Research (RES). The motivation for re-BRAMWELLD.L.; WATKINS.J.C.: et al. EG&G Idaho, Inc. Fety..

search stems from the safety-related issues that need to be ad-ruary 1994. 79pp. 9403140229. EGG-2686. 78477:071.

dressed with the apphcation of advanced I&C systems to nucle-l Thes report documents the results of the main projects under.

ar power plants. Development of the technical basis centered l

taken under the Environmental and Dynamic Equipment Quahfi-arwnd establishing good engineering practces to ensure that i

cation Research Program (EDOP) sponsored by the U.S. Nucle-suHcient levels of electromagnetic compabbikty (EMC) are Regulatory Commission (NRC) under FIN A6322. Lasting maintained between the nuclear power plant's electronic and l

at from fiscal year 1983 to 1987, the program dealt with environ-electromechanical systems known to be the source (s) of EMI/

mental and dynamic (ircluding seismic) equipment quahficaton RFl and power surges. First, good EMC design and installaton j

issues for mechanical and electromechanical components and practices need to be established to control the impact of inter-i systems used in nuclear power plants. The research results ferm smrces on nearby circuits and systems. These EMC have since been used by both the NRC and industry.The pro-good practces include circuit layouts, terminatens, filtering, gram included seven major research projects that addressed grounding, bonding, shielding, and adequate physical separa-l the following issues: (a) containment purge and vent valves per.

tion. Second, an EMI/RFI test and evaluation program needs to forming under design basis loss of coolant accident loads, (b) be established to outhne the tests to be performed, the associ-containment piping penetrations and isolation valves performing ated test methods to be followed, and carefully formulated ac-under seismc loadings and design basis and severe accident ceptance entena based on the intended environment to ensure containment wall displacements. (c) shaft seals for primary cool-that the circuit or system under test meets the recommended ant pumps performing under staton blackout conditons, (d) guidehnes. Third, a program needs to be developed to perform electncal cabinet internals responding to in-structure generated confirmatory tests and evaluate the surge withstand capabihty motion (ratthng), and (e) in situ piping and valves responding to (SWC) and of l&C equipment connected to or installed in the seismic loadings. Another project investigating whether certain vicinity of power circuits within the nuclear power plant. By fol-containment isolation valves will close under design basis con.

lowing these three steps, the desegn and operabihty of safety-ditons was also started under ttws program. This report includes related l&C systems against EMI/RFI and power surges can be eight main sectons, each of which provides a bnef desenption evaluated, acceptance enteria can be developed, and appropri-of one of the projects, a summary of the findings, and an over-ate regulatory guidance can be provided view of the apphcation of the results. A bibhography lists the joumal arteles, papers, and repo'ts that document the research.

NUREG/CR-5960 STEAM EXPLOSIONS: FUNDAMENTALS AND ENGERGETIC BEHAVIOR. THEOFANOUS.T.G.; YUEN,W.W.;

NUREG/CR 5939: THE EFFECTS OF AGE ON NUCLEAR ANGELINI,S.; et al. California, Urvv. of. Santa Barbara, CA. Jan-POWER PLANT CONTAINMENT COOLING SYSTEMS.

uary 1994. 250pp. 9403140221. 78454:001.

LOFARO,R.; SUBUDHI,M.; TRAVIS R.; et al. Brookhaven Na-This report presents the results of a multifaceted research tional Laboratory. Apnf 1994. 160pp. 9405310236. BNL-NUREG-52345. 79589:042.

effort in the field of steam explosions. The scope ranges from the fundamentals to assessing the energetics in appications rel-A stuoy was performed to assess the ef:scts of aging on the evant to Severe Accidents in Light Water Reactors. The consid-performance and availability of containment cooling systems in eration of fundamentals is built around two key ideas: the water U.S. commercial nuclear power plants. This study is part of the depleton phenomenon dunng premixing and the mcrointerac-Nuclear Plant Agtng Research (NPAR) program sponsored by tons, including fragmentaten kinetes, dunng propagaten. The the U.S. Nuclear Regulatory Commission. The objectrves of this apphcation to reactor conditions includes consideraton of in.

program are to provide an understanding of the aging process vessel steam explosons in PWRs and ex-vessel explosions in and how it affects plant safety so that it can be property mark all five containment designs in current plants (in the USA). The aged. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identifi-report is structured in three parts, deahng with premixing, propa-gaton, and energetes, respectively.

cation and evaluation of degradaten caused by age. The ef-fects of age were charactenzed for the containtnent cooling NUREG/CR-5963: CONTINUOUS AE CRACK MONITORING OF system by reviewing and analyzing failure data from national da-A DISSIMILAR METAL WELDMENT AT LIMERICK UNIT 1.

tabases, as well as plant-specific data. The predominant failure HUTTON,P.H.; FRIESEL,M.A.; DAWSON.J.F. Battelle Memorial causes and aging mechanisms were identified, along with the Institute, Pacifc Northwest Laboratory. December 1993.100pp.

components that failed most frequently. Current inspecten, sur-9402220168. PNL-8844. 78197:095.

veillance, and monitonng practces were also examined. A con.

Acouste emisson (AE) technology for continuous surveillance tainment coohng system unavailabihty analysis was performed of a reactor component (s) to detect crack initiaton and/or crack to exarrune the potential effects of aging by increasing failure growth has been developed at Pacife Northwest Laboratory rates for selected components. A commonly found containment (PNL), operated by Battelle Memonal Institute, under support spray system dessgn and a commonly found fan cooler system from the U.S. Nuclear Regulatory Commession Offee of Nuclear design were modeled. Parametnc failure rates for those Compo.

Regulatory Research (U.S. NRC-RES). The technology was vali-

24 Main Citations and Abstracts dated off-reactor in several major tests, but it had not been vali-NUREG/CR-5973 R01: CODES AND STANDARDS AND OTHER dated by monitonng crack growth on an operating reactor GUIDANCE CITED IN REGULATORY DOCUMENTS.

system. A flaw indicaton was identified during normal inservice ANKRUM,A R ; NICKOLAUS.J.R.; VINTHER,R.W.; et al. Battelle inspecten of piping at Philadelphia Electric Cc,mpany (PECO)

Memonal Institute, Pacife Northwest Laboratory. August 1994.

Limenck Unit I reactor dunng the 1989 refueling outage. Eval-398pp. 9409230271. PNL-8462. 81009:001.

uaton of the flaw indcaten showed that it could remain in As part of the U.S. Nuclear Regulatory Commission (NRC) place dunng the subsequent fuel cycle without compromising Standard Review Plan Update and Development Program, Pa-safety. The existence of this flaw indcaten offered a long cafc Northwest Laboratory developed a hsting of industry con-sought opportunity to validate AE surveillance to detect and sensus codes and standards and other government and indus-evaluate crack growth dunng reactor operaton. Through the co-try guidance referred to in regulatory documents. In addition to operaton and support of PECO and the U.S. NRC-RES, AE in-updating previous information, Revision 1 adds citations from strumentation was installed by BNW and PECO under PECO the NRC Inspection Manual and the improved Standard Techni-Mod. No. 043-002 to monitor the flaw indication dunng two cal Specifcatons. This hsting identifies the version of the code complete fuel cycles. This report discusses the results obtained or standard cited in the regulatory document, and the current from the AE monitoring over the pered May 1989 to March version of the code or standard. It also providos a summary charactenzaten of the nature of the citaton. This listing was de-1992 (both fuel cycles).

veloped from electrone searches of the Code of Federal Regu-NUREG/CR-5965: MODELING FIELD SCALE UNSATURATED lations and the NRC's Bulletins, Informaton Notees, Circulars.

FLOW AND TRANSPORT PROCESSES. GELHAR,LW.;

Genenc Letters, Policy Statements, Regulatory Guides, and the CELIA,M.A.; MCLAUGHLIN D. Massachusetts Institute of Tech-Standard Review Plan (NUREG-0800).

i NUREG/CR-5985: EVALUATION OF COMPUTER-BASED UL.

f

nology, mbndge, MA. August 1994. 82pp. 9409230273.

TRASONIC INSERVICE INSPECTION SYSTEMS. HARRIS,R.V.;

A stochastic theory desenbing unsaturated flow and contami-ANGEL.LJ.; DOCTOR,S.R.; et al. Battelle Memonal Institute, nation transport in naturally hemerogeneous soils has been en-Pacife Northwest Laboratory. March 1994.126pp.9404040063.

hanced by adopting a more realistic charactenzaten of soit vari-PNL-8919. 78737.001.

ability. The enhanced theory is used to predict field-scale effec-This report presents the pnnciples, practces, terminology, tive properties and vanances of tenson and moisture content.

and technology of computer-based ultrasonc testing for inserv-Appicatons illustrate the ernportant effects of small-scale het-ice inspection (UT/ISI) of nuclear power plants, with extensive erogeneity on large-scale anisotropy and hysteresis and demon-use of drawings, diagrams, and UT images. The presentaten is strate the feassbshly of simulating two-dimensional flow systems techncal but assumes limited specifc knowledge of ultrasoncs at time and space scales of interest in radoactive waste dispos-or computers. The report is divded into 9 sectons covenng al investgatons. Numencal algonthms for predcting field scale conventional UT, computer-based UT, and evaluaton methodot-unsaturated flow and contaminant transport have been im-ogy. Convenbonal UT topes include coordinate axes, scanning, proved by requinng them to respect fundamental physical pnnci-instrument operaton, RF and vdeo sgnals, and A, B, and C-pies such as mass conservaton. These algonthms are able to scans. Computer-based topcs mclude samphng, dgrtizaton, provide reahsts simulatons of systems with very dry initial con-signal analysis, image presentaten, SAFT, ultrasonic hologra-dibons and hgh degrees of heterogeneity. Numencal simulation phy, transducer arrays, and data interpretaten. An evaluaton of the simultaneous movement of water and air in unsaturated methodology for computer-based UT/ISI systems is presented, soils has demonstrated the importance of air pathways for con-including quesbons, detailed procedures, and test block de-taminant transport. The stochaste flow and transport theory has signs. Bnef evaluatons of several computer-based UT/ISI sys-been used to develop a systemate approach to performance tems are given; supplementary volumes will provide detailed assessment and site charactenzaton. Predcton uncertaintes evaluatons of selected systems.

have been quanbfed by considenng the role of both natural het-erogenesty and measurement error. Hypothesis-testing tech-NUREG/CR-5990 THE EFFECTS OF SOLAR-GEOMAGNETI-neques have been used to determine whether modol predctons CALLY INDUCED CURRENTS ON ELECTRICAL SYSTEMS IN are consistent with observed data.

NUCLEAR POWER STATIONS. SUBUDHI,M. Brookhaven Na-tonal Laboratory. CARROLL.D P. Flonda Univ. of, Gainesville, NUREG/CR-5967: DEVELOPMENT AND APPLICATION OF DEG-FL. KASTURI,S. MOS, Inc. January 1994.196pp. 9402150350.

RADATION MODELING TO DEFINE MAINTENANCE PRAC-BNL-NUREG 52359 78123.019.

TICES. STOCK.D.; SAMANTA.P. Brookhaven National Labora-This report presents the results of a study to evaluate the po-tory. VESELY,W. Science Apphcatons International Corp. (for.

tential effects of geomagnetcally induced currents (GICs) merly Science Appicatons, Inc.).

June 1994. 63pp.

caused by the solar disturbances on the in-plant electncal distrL 9406290294. BNL-NUREG-52353. 80013 001, bution system and equipment m nuclear power stabons. The This report presents the development and appication of com-plant specifc electrcal distnbuton system for a typcal nuclear ponent degradaten modeling to analyze degradaten effects on plant is modeled using the ElectroMagnetc Transient Program rehability and to identify aspects of maintenance practces that (EMTP). The computer model simulates onhne equipment and trutgate degradaten and aging effects. Using continuous time loads from the station transformer in the switchyard of the Markov approaches, a component degradation rnodel is dis.

power staton to the safety-buses at 120 volts to whch all elec-cussed that includes informaton about degradaten and mainte-tronc devices are connected for plant monitonng. The analybcal nance. The component model commonly used in probabikste model of the plant's electrical distnbution system is studed to nsk assessments is a simple case of this general model. The identify the transient effects caused by the half-cycle saturabon parameters used m the general model have engineenng inter.

of the station transformers due to GlC. This study provides re-pretabons and can be estimated using data and engineenng ex-suits of the voltage harmoncs levels that have been noted at penence. The generation of equations for specsfic models, the varous electrcal buses insde the plant. The emergency circuits soluton of these equatons, and a methodology for estimating appear to be more susceptible to hgh harmonics due to the the needed parameters are all discussed. Appicatons in this normally Ight load conditions. In additon to steady-state analy-report show how these models can be used to quantitatively sis, this model was further analyzed simulating vanous plant assess the benefits that are expected from maintaining a com.

transient conditions (e g., loss of load or large motor start-up) ponent, the effects of different maintenance effeiencies, the occumng dunng GlC events. Detail models of the plant's pro-ments of different maintenance polcies, and the interaction of tective relaying system employed in bus transfer application surveillance test mtervals with mamtenance practces.

were included m this model to study the effects of the harmonic

Main Citations and E a ts 25 distorion of the voltage input. Potential harmonic effects on the through the instrument tunnel into the subcompartment struc-uninterruptable power system (UPS) are qualitabvely discussed tures and the upper dome of the simulated reactor containment as well.

building. The results of the IET expenments are given in this NUREG/CR-5994: EMERGENCY DIESEL GENERATOR: MAIN.

TENANCE AND FAILURE UNAVAILABILITY, AND THEIR RISK NUREG/CR-6051: EFFECTS OF TENSILE LOADING ON UPPER IMPACTE. SAMANTA,P.; KIM,1.; URYASEV.S.; et al. Brookha-SHELF FRACTURE TOUGHNESS. JOYCE,J.A.; LINK,R.E. U.S.

ven Naboial Laboratory. November 1994. 210pp. 9412130030.

Naval Academy, Annapolis, MD. March 1994. 129pp.

BNL NUREG-52363. 82006:001.

9405050401, 79165:001.

Ernergency Diesei Generators (EDGs) provide orwsite emer.

The elaste-plaste expenments described in this document in-gency alternabng current (ac) electnc power for a nuclear plant vestigated whether upper-shelf initiation toughness and J resist.

in tre event that all off-site power sources are lost. Existing reg-ance curves are dependent on T stress and O. Upper-shelf J ulations estabhsh requirements for desagrung and testing of resistance curves (J(Ic)) and teanng resistance (T(mat)) results these on-site power sources to reduce to an acceptable level were obtained for a range of applied constraint. The J-O and J-the probability of losing all ac power sources. Operating expen-T stress loci were developed and compared with the expecta-ence with EDGs has rarsed questens about their testing and tions of the O'Dowd and Shsh and the Betegon and Hancock maintenance to achieve the EDG rehabehty levels and the total analyses. Consbaint was vaned by changing the crack length EDG unavailab&ty expenenced (fracton of time EDG is out-of-and by changing the loadsng mode from bending to predomi-service due to teshng, maintenance, and failures). In this report, nantly tensile. The conclusion is that J(Ic) is apparently not de-recent operating expenence is used to assess EDG unavailabil-pendent on T-stress or O but the matenal teanng resistance is, ity due to testing, maintenance, and failures dunng reactor with the tearing modulus increasing as constraint increases.

power operation and dunng plant shutdown. Recent data show an improvement in EDG rehabibty, but an increase in EDG un-NUREG/CR-6053: COMPARISON OF MACCS USERS CALCULA-availability due to maintenance, a segrwficant portion of which is TIONS FOR THE INTERNATIONAL COMPARISON EXERCISE due to routinety scheduled maintenances. Probabilistic safety ON PROBABILISTIC ACCIDENT CONSEQUENCE ASSESS-assessments (PSAs) of selected nuclear power plants are used MENT CODES. NEYMOTIN,L Brookhaven National Laboratory.

to assess the nsk impact of EDG unavailability due to mainte-April 1994.

184pp.

9405310158.

BNL-NUREG.52380.

nance and failure dureng power operaton, and dunng different 79592:001.

stages of plant shutdown. The results of these nsk analyses Over the past several years, the OECD/NEA and CEC spon-suggest qualitative insights for scheduling EDG maintenance sored an internatonal program intercompanng a group of six that will have minimal impact on nsk of operabng nuclear power probabilistic consequence assessment (PCA) codes designed to plants-simulate health and economic consequences of radioactrve re-NUREG/CR-6042: PERSPECTIVES ON REACTOR SAFETY.

leases into atmosphens of radoactive maMnals foHowing sem HASKIN,F.E. New Mexico, Uruv. of, Albuquerque, NM.

accidents at nuclear power plants (NPPs): ARANO (Finland),

CAMP,A.L Sandia National Laboratones. March 1994. 589pp.

CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS 9404110335. SAND 93-0971. 78820:001.

(USA), and OSCAAR (Japan). In parallel with this effort, two The U.S. Nuclear Regulatory Comfrusson (NRC) maintains a separaM groups perfonned sinular calculations using the technical trairung center at Chattanooga, Tennessee to provide MACCS and COSYMA codes. Results produced in the MACCS appropnate trairung to both new and exponenced NRC employ, Users Group (Greece, Italy, Spain, and USA) calculatons and ees. This document desenbes a one-week course in reactor their comparison are contained in the present report. Version safety concepts. The course consists of frve modules: (1) histor.

1.5.11,1 of the MACCS code was used for the calculations.

ical perspectrve; (2) acesdent sequences; (3) accident progres.

Good agreement between the results produced in the four par.

sicn in the reactor vessel, (4) containment charactensbcs and ticipabng calculations has been reached, with the excepton of desegn bases; and (5) source terms and offsste consequences.

the results related to the ingeston pathway dose predactions.

The course text is accomparued by shdes and videos dunng the The main reason for the scatter in those particular results is at-actual presentation of the course.

tnbuted to the lack of a straightforward implementation of the specifications for agneuttural production and counter measures

• • UREG/CR-6044: EXPERIMENTS TO INVESTIGATE DIRECT critena provided for the exercise. A sigruficantly smaller scatter

'. ONTAINMENT HEATING PHENOMENA WITH SCALED in predictions of other consequences was successfully ex.

IdDELS OF THE ZION NUCLEAR POWER PLANT IN THE plained by differences in meteorological files and weather sam-S RTSEY TEST FACILITY. ALLEN M D.;

PILCH,M.M.;

pling, grids, rain distance intervals, dispersion model optons, B ANCHAT,T.K.; et al. Sandia National Laboratones. May 1994, and populaton distnbutons.

Ei op. 9406210288. SAND 93-1049. 79836:001.

as Surtsey Facihty at Sandia National Laboratories (SNL)is NUREG/CR 6063: INTRAVAL PHASE 11 MODEL TESTING AT usec 'o perforrP scaled expenments that simulate hypothetical THE LAS CRUCES TRENCH SITE. HILLS.R.G. New Mexico high-pi 7 weit ejection (HPME) accidents in a nuclear State Univ., Las Cruces, NM. WIERENGA,P.J. Arizona, Univ. of, por a ptert (NPP). These expenments are designed to investi.

Tucson, AZ. LUIS,S.; et al. Massachusetts insttute of Technolo-gate the effect of specific phenomena associated with direct gy, Cambridge, MA. January 1994. 143pp. 9402150366.

cor.tainment heating (DCH) ort the containment load, such as 78123:209.

the effect of physical scale, prototypic subwmpadment struc.

Several field experiments have been performed by scienbsts j

turet, water in the cavity, and hydrogen generation and com.

from the University of Anzona and New Mexico State University bush sn. In the Integral Effects Test (IET) series,1:10 knear at the Las Cruces Trench Site to provide data to test determs scale nodels of the Zon NPP structures were constructed in nistic and stochastic models for water flow and solute transport.

)

the Sui'sey vessel. The RPV was modeled with a steel pressure These expenments were performed in collaboration with INTRA.

vessel ttat had a Mp6-d: bottom head, which had a 4 cm VAL, an internatonal effort toward vahdation of geosphere i

hole in the Lottom head that simulated the final ablated hole models for the transport of radionuchdes. During Phase I of IN.

that would be formed by ejecten of an instrument guide tube in TRAVAL, quahtative comparisons between expenmenta: data a severe NPP accident. Iron / alumina / chromium thermite was and model predictons were made using contour plots of water used to simulate molten conum that would accumulate on the contents and solute concentrations. Detailed quantitative com-bottom head of an actual RPV. The chemically reactive mett sb pansons were not made. To provide data for more ngorous i

mulant was ejected by high-pressure steam from the RPV model teshng, a third Las Cruces Trench expenment was de-model into the scaled reactor cavity. Debns was then entrained signed by scienbsts from the University of Anzona and New i

i i

l Main Citations and Abstracts l

26 Mexico State Urwersity. Modelers from the Center for Nuclear groups. In the four new scenarios, consistency of the initial con-l Waste Regulatory Analysis, Massachusetts Institute of Technol-dations has been implemented by using insights from systems-i ogy, New Mexico State University, Pacific Northwest Laboratory, level codes. SCDAP/REMPS was used to analyze three short.

and the University of Texas provided predictons of water flow term station blackout cases with different lead rates. In all three and tritium transport to New Mexico State University for analy-cases, the hot leg or surge lirw failed well before the lower sis. The corresponding rnodels assumed soil charactenzations head and thus the primary system depressurized to a point ranging from uruform to deterministically heterogeneous to sto-where DCH was no longer considered a threat. However, these chastc. TNs report presents detailed quantitative comparisons calculations were continued to lower head failure in order to to field data.

gain insights that were useful in establisNn0 the initial and boundary cmdtons. De most M insWs are mat N M MUREG/CR-6074 V01: SEALED SOURCE AND DEVICE DESIGN pream is low at vessel beach nwtak Mages M N m SAFETY TESTING. Technical Report On The Findings Of Task regon do not men and retxate into me W W, aM

1. October 1991 April 1993. DORNES.E.; CUTSHALL,W.;

meltmg of uppa p6enum steel is correlated with hot leg failure.

GONZALEZ,M. Southwest Research Institute. April 1994.

THE SCDAP/REMP output was used as input to CONTAIN to 102pp. 9405040118. 04-4448-Y1T1. 79103:003 assess tM contamnwnt conns at vessd Wa& W m This report covers the Task 1 activities for the Sealed Source tainment-side conditons predicted by CONTAIN are similar to and Devee Design Safety Test Program. The NRC and Agree-mose onginah speified in NUREG/CR-6075.

ment States evaluate radiabon safety informaton and register i

the designs of products containing radioactive matenals. SwRI NUREG/CR-6076: TR.EDB: TEST REACTOR EMBRITTLEMENT has been contracted to perform an examination of the heensing DATA BASE,VEPSION 1.

STALLMANN,F.W.; WANG.J.A.;

information submitted by vendors. The intent of this investiga-KAM.F.B.K. Oak Ridge Natonal Laboratory. January 1994.

tion is to venfy vendor testing and analysis and to determme 119pp. 9403140311. ORNL/TM-12415. 78516:001.

where further testing is warranted to ensure the radiabon safety The Test Reactor Embnttlement Data Base (TR-EDB) is a of the sealed source and device designs. Results of interviews collection of results from irradiations in materials test reactors. It with users are presented and recommendations are made for complements the Power Reactor Embrittlement Data Base (PR-trsts to be developed dunng Task 2.

EDB), whose data are restricted to the results from the anelysis NUREG/CR-6075: THE PROBABILITY OF CONTAINMENT FAIL.

of surveillance capsules in comtnercial power reactors. The ra-URE BY DIRECT CONTAINMENT HEATING IN ZION.

tionale beNnd this restriction was the assumpton that the re-PILCH,M.M.

Sandia Natonal Laboratones.

YAN.H.;

suits of test reactor expenments may not be applicable to THEOFANOUS.T.G. Cahfomia Urw. of, Santa Barbara, CA. De.

power reactors and could, therefore, be challenged if such data cember 1994. 396pp. 9501190286. SAND 931535, 82359:001.

were included. For this very reason the embnttlement prede-TNs report is the fast step in the resoluten of the Drect Con.

tons in the Reg Guide 1.99, Rev. 2 were based exclusively on tainment Heating (DCH) issue for the Zion Nuclear Power Plant power reactor data. However, test reactor expenments are able using the Risk Onented Accident Anatysis Methodology to cover a much wider range of materials and irradiaton condi-(ROAAM). This report includes the definiton of a probabilists tions that are needed to explore more fully a vanety of models framework that decomposes the DCH problem into three proba.

for predicion of irradiation embnttlement. These data are also bility density functons that reflect the most uncertain initial con.

needed for the study of effects of annealing for hfe extension of ditions (UO(2) mass, zrcorwum oxidation fractiort and steel reactor pressure vessels that are diffeult to obtain from surveil.

mass). Uncertainties in the initial conditsons are significant, but lance capsule results. The current data collecton of the TR.

I our quantificaton approach is based on establishing reasonable EDB contains pnmanly Charpy test data, which are accompa-bounds that are not unnecessanly conservative. To tNs end, we nied in most cases by tensile tests for the same irradiaton con-also make use of the ROAAM ideas of enveloping scenanos dations. Information is available for 1,230 different irradiated and " splintering." Two causal relatons (CRS) are used in this sets, 797 of which are from base material (plates and forgings),

framework: CRI is a model that calculates the peak pressure in 378 from welds, and 55 from heat-affected-rone materials. The the containment as a function of the initial conditions, and CR2 chemistnes of the investigated materials span also a farfy wide is a model that retums the fluency of containment failure as a range, partcularly in the content of copper and nickel, whch are function of pressure within the containment. Uncertainty in CR1 considered the rnost important contnbutors to embrittlement is accounted for by the use of two independently developed sensitivity. Complete chemistry informaton is available for 1,095 l

phenomenologcal models, the Convecton Limited Containment of the 1,230 samples (after discarding the HAZ information).

l Heating (CLCH) model and the Two-Cell Equihbrium (TCE)

THe arcNtecture of the TR EDB is fully compatible with that of l

model, and by probabekstically distnbuting the key parameter in the PR-EDB so that the data from both databases can be easily both, which is the ratio of the melt entrainment time to the merged if dested. The data files are given in dBASE format and system blowdown time constant. The two phenomenolog' cal can be accessed with any personal computer using the DOS models have been compared with an extensive database includ.

operating system. " User fnendly" utshty programs have been ing recent integral simulations at two different physical scales.

wntten to investigate the radiaton embnttlement using tNs data The containment load distnbutons do not intersect the contain.

base. The utility programs are used to retrieve and select spe-ment strength (fragibty) curve in any significant way, resulting in cific data, manipulate data, display data to the screen or printer, containment failure probabihties less than 10( 3) for all scenar.

and to fit and plot Charpy impact data.

conssder Sensstivity analyses did not show any areas of NUREQ/CR-6077: DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST VI-6.

OSBORNE.M.F.;

NUREG/CR-6075 S01: THE PROBABluTY OF CONTAINMENT LORENZ,R.A.; TRAVIS,J.R.; et al. Oak thdge National Laborato-FAILURE BY DIRECT CONTAINMENT HEATING IN ZION.

ry.

March 1994. 67pp. 9404080092. ORNL/TM 12416.

PILCH,M.M.; ALLEN.M.D. Sandia Natonal Laboratones.

78794:197, KNUDSON,D.L.; et al. Idaho Nabonal Engineenng Laboratory.

Test VI-6 was the sixth test in the VI senes conducted in the December 1994.

450pp.

9501180334. SAND 93-1535.

vertical furnace. The fuel specimen was a 15.2-cm-long section 82340:001.

of a fuel rod from the BR3 reactor in Belgium. The fuel had ex.

Supplement 1 of NUREG/CR-6075 bnngs to closure the DCH perienced a burnup of ~42 mwd /kg, with inert gas release issue for the Zion plant. It includes the documentation of the during irradiaton of ~2%. The fuel specimen was heated in an l

peer review process for NUREG/CR-6075, the assessments of induction furnace at 2300 K for 60 men, initially in hydrogen, four new splinter scenanos defined in working group meetings, then in a steam atmosphere. The released fission products i

and modehng enhancements recommended by the working were collected in three sequentially operated collecten trains I

Main Citations and Abstracts 27 designed to facihtate sampling and analysis. The fission product cluded three or more team members with appropriate experbse inventones in the fuel were measured directly by gamma ray in rad 6abon oncology, medical physics, nuclear medicine tech-spectrometry, where possible, and were calculated by nology, nsk analysis, and human factors. The investigatons fo-ORIGEN2. Integral releases were 75% for (85)Kr, 67% for cused on causes of the event, consequences, mitigabng ac-(129)l, 64% for (125)Sb, 80% for both (134)Cs and (137)CS, tions, and corrective actons. The investigation produced seven 14% for (154)Eu,63% for Te,32% for Ba,13% for Mo, and major findings: 1) many misadministratons occurred pnmarily 5.8% for Sr. Of the totals released from the fuel,43% of the because procedures did not exist or because existing proce-Cs,32% of the Sb, and 98% of the Eu were deposited in the dures that were not sufficiently detailed, comprehensive, specif-outlet end of the furnace. Dunng the heatup in hydrogen, the ic, or clearty wntten; 2) although tfe NRC's quahty management Zircaloy cladding melted, ran down, and reacted with some of (QM) rule can prevent many misadministrabons, most hcensees the UO(2) and fission products, especially Te and Sb. The total in this study had not effectively implemented their OM pro.

mass released from the furnace to the collecton system, includ-grams; 3) the lack of substantial, direct involvement by radiaton ing fission products, fuel, and structural materials, was 0.57 g, safety officers and authorized users was often a direct cause of almost equally divided between thermal gradient tubes and fil.

misadministraton; 4) a change in routine or the advent of a ters. The release behaviars for the most volatile elements, Kr unique conditon often predisposed misadministration; 5) hard-and Cs, were in good agreement with the ORNL Diffusion ware failures, though rare, had severe consequences, particular-Model.

ly when operating procedures, staff training, or other factors NUREG/CR-6006: SELECTED FAULT TESTING OF ELECTRON.

wwe not wd kn>nM 6) kenms' care ams IC ISOLATION DEWCES USED IN NUCLEAR POWER PLANT wweo n narrow en fows; 7) N kenes lacM syswnato n

s fu deMng aM rndgaW a Msamskaton om OPERATION. VILLARAN,M.; HILLMAN.K.; TAYLOR,J.; et al.

an m occmd l

Brookhaven National Laboratory. May 1994.

126pp.

)

9405310260. BNL-NUREG 52385. 79588:141.

NUREG/CR-6092: RISK ASSESSMENT FOR THE INTENTIONAL Electronic isolation devices are used in nuclear power plants DEPRESSURIZATION STRATEGY IN PWRS. DINGMAN,S.E.

i to provide electncal separaton between safety and non-safety Sandia Natonal Laboratories. March 1994. 94pp. 9405040222.

circuits and systems. Major fault testing in an earher program in-SAND 93-1737. 79108:001.

dicated that some energy may pass through an isolaton device An accident managoment strategy has been proposed in when a fault at the maximum credible potential is apphed in the which the reactor coolant system is intentonally depressurized transverse mode to its output terminals. During subsequent field dunng an accident. The aim is to reduce the containment pres-quahficaten testing of isolators, concems were raised that the surizatson that would result from high pressure ejection of worst case fault, that is, the maximum credible fault (MCF), may molten debns at vessel breach. Probabihshc risk assessment not occur with a fault at the maximum credible potential, but (PRA) methods were used to evaluate tNs strategy for the Surry rather at some lower potential. The present test program inves-nuclear power plant. Sensitivity studies were conducted using tigates whether problems can anse when fault levels up to the MCF potential are applied to the output terminals of an ;=olator-event trees that were developed for the NUREG-1150 study. It was found that depressunzation (intentonal or unintentional)

The fault energy passed through an isolated device dunng a had minimal impact on the containment failure probability at fault was measured to determine whether the levels are great vessel breach for Surry because the containment loads as-enough to potentially damage or degrade performance of equip-sessed for NUREG-1150 were not a great threat to the contain-4 ment on the input (Class 1E) side of the isolator.

ment survivability. An updated evaluation of the impact of inten-NUREG/Ch-6067: THE EFFECTS OF AGING ON BOILING tional depressurizatton on the probability of having a high pres-WATER REACTOR CORE ISOLATION COOLING SYSTEMS.

sure melt ejecten was then made that reflected analyses that LEE BS. Broukhaven National Laboratory October 1994.

have been performed since NUREG 1150 was completed. The 218pp.9411160020. BNL-NUREG-52390. 81756:083.

updated evaluaton confirmed the sensitivity study conclusions A study was performed to assess the effects of aging on the that intentonal depressurization has minimal impact on the Reactor Core isolaton Cooling (RCIC) system in commercial probabsty of a high pressure melt election. The updated evalua-Bosng Water Reactors (BWRs). TNs study is part of the Nucle-toon did show a shght benefit from depressurizaten because de-1 at Plant Aging Research (NPAR) program sponsored by the pressunzation delayed core melting, which led to a hsgher prob-U.S. Nuclear Regulatory Commisson. The objectives of this pro-abihty of recovenng emergency core coolant injecten, thereby gram are to provide an understanding of the aging process and arresting the core damage.

how it affects plant safety so that it can be property managed.

TNs is one of a number of studies performed under the NPAR NUREG/CR-6093: AN ANALYSIS OF OPERATIONAL EXPERI.

program wNch provide a technical basis for the idenbficaton ENCE DURING LOW POWER AND SHUTDOWN AND A PLAN FOR ADDRESSING HUM AN RELIABILITY ASSESSMENT and evaluation of degradaten caused by age. The failure data from natonal databases, as well as plant specific data were re-ISSUES. BARRIERE,M.; LUCKAS.W. Brookhaven National Lab-i viewed and anatyzed to understand the effects of agrng on the oratory. WHITEHEAD,D %; et al. Sandia National Laboratories.

June 1994.

200ry 9408030127.

BNL-NUREG-52388.

RCIC system. TNs analysis identifned important components 8043010' that should receive the highest prionty in terms of aging man-agement. The aging charactenzation provided informaton on Recent nuclear power plant events (e.g. Chernobyl, Diablo the effects of aging on component failure frequency, failure Canyon, and Vogtie) and U.S. Nuclear Regulatory Commisson modes, and failure causes. Current inspection, surveillance, and (NRC) reports (e.g. NUREG-1449) have led to concerns regard-monitonng practices were also reviewed.

ing human reliabsty dunng low power and shutdown (LP&S) conditons and hmrtations of human reliabihty analysis (HRA)

NUREG/CR-8088:

SUMMARY

OF 1991 1992 MISADMINISTRA.

methodologies in adequately representing the LP&S environ.

TlON EVENT INVESTIGATIONS. OSTROM,L.T.; LEAHY,T.J.;

ment. As a result of these concerns, the NRC initiated two par-NOVAK.S.D. EG&G Idaho, Inc. March 1994.61pp 9404040060.

allel research projects to assess the influence of LP&S conde EGG-2707. 78736:193.

tions on human reliatxhty through an analysis of operational ex-Investigation teams composed of representatives of the Idaho penence at pressunzed water reactors (PWRs) and boihng Naborial Engineenng Laboratory (INEL), the U.S. Nuclear Regu-water reactors (BWRs). These research projects, performed by latory Commisson (NRC), and subcontractors investigated and Brookhaven National Laboratory for PWRs, and Sandia National analyzed seven misadministraton events selected by the NRC Laboratones for BWRs, identified unique aspects of human per-concerning medica. radioisotopes. Each team was led by an formance during LP&S conditions and provided a program plan INEL member and, depending on the nature of the event, in-for research and development necessary to improve existing i

28 Main Citations and Abstracts HRA methodologies. This report documents the results of the (44GROUPNDFS). This cross-section set has been tested analysis of LP&S operahng expenence and descnbes the im-against its parent 238-group line structure hbrary proved HRA program plan.

(238GROUPNDF5) using a set of 33 benchmark problems in order to demonstrate that the crAlapsed set was an acceptable NUREG/CR4094: CALCULATIONS IN SUPPORT OF A POTEN-representation of 238GROUFolDF5. Vahdation of the library TIAL DEFINITION OF LARGE RELEASE. HANSON.A.L; within the SCALE system wn based on a compenson of caecu-j DAVIS,R.E.: MUBAYl,V. Brookhaven National Laboratory. May lated values of k(eff) with 'nat of 93 experiments. The expon-1994.16Spo.9406200323. BNL NUREG-52387. 79830:037, ments pnmarily consisted of vanous configurations of light-The Nuclear Regulatory Commission has stated a hierarchy water reactor type fuel ref resentative of transportaticn and stor-of safety goals with the qualitative safety goals as LevelI of the age conditions. Additiont,i expenments were included to allow hierarchy, backed up by the quantitabve health objectives as companson with results obtained in earher valdabon of the Level fl and the contemplated large release guideline as Level 27GROUPNDF4 hbran,, Results show that the broad 44-group Ill. The large release guideline has been stated in qualitative structure is an acceptable representation of its parent 238-terms as a magnitude of release of the core inventory whose group hbrary for thermal as well as hard fast spectrum systems.

frequency thould not exceed 10(-6) per reactor year. However Accurate broad-group analyses of intermedate spectmm sys-the Commission did not provide a quantitative specification of a tems will require either a more detailed group structure in this targe release. One rh.w.w.wdstion was made that a large re-energy range or a more appropriate collapsing spectrum. Fur-lease be defined as having the potential for causing an off-site ther, validation calculations indicate that the 44-group hbrary is earty fatality within one mde of the site to the average individual.

an accurate tool in the predictio,1 of criticality for arrays of light-This report focuses on an examination of releases which have a water-reactor-type fuel assemblies as would be encountered in potenbal to lead to one prompt fatality. The basic informabon required to set up the calculations was denved from the simpli.

fresh or spent fuel transportation or storage environments, fied source Wms which were obtained from approximations of NUREG/CR-6103: PRIORITIZATION OF REACTOR CONTROL the NUREG-10 source terms. Since the calculation of conse-COMPONENTS SUSCEPTIBLE TO FIRE DAMAGE AS A CON-quences is. affecteo y a large number of assumphons, a genor-SEQUENCE OF AGING. LOWRY,W.; VIRGIL,R. Science & En-ic site witt' a populatis riensity and meteorology was specified.

gineering Associates, Inc. NOWLEN,S. Sandia National Labora-At this si'a, various enoney responses (including no re.

tories. January 1994. 45pp. 9402150309. SAND 93 7107.

sponse) w are assumed based on information derived from earip 78122:333.

er studies For each of the emergency response assumphons, a set of.:alculations were performed with the simphfied source The Fire Vulnerability of Aged Electncal Components Test terms These included adjustments to the source terms, such as Program is to identify and assess issues of plant aging that tho timing of the release, the core inventory, and the release could lead to an increase in nuclear power plant risk because of fractsons of different radionuclides, to have a potential to lead to fires. Historical component data and prior analgos are used to one mean prompt fatality in each case. Each of the source pnontire a list of components with respect to aging arH fire vul-terms, so defined, has the potential to be a candidate for a nerability and the consequences of their failure on plant safety large release, systems. The component list emphasizes safety system control components, but excludes cables, large equipment, and devices NUREG/CR4095: AGING. LOSS-OF-COOLANT ACCIDENT wnpassed in me EWpmd Quakahm (EQ Fogram h (LOCA), AND HIGH POTENTIAL TESTING OF DAMAGED test Narn selected componets Med in a u% sunmy CABLES. vigil,R.A. Science & Ergineering Associates, Inc.

and heloped test and ke cedibms nessary to maxN JACOBUS,M.J. Sandia National Lat' oratories. Apnl 1994. 74pp.

mess of me test program Re damage cmsidw.

ee 9405310297. SAND 93-1803. 79585:231.

abons were hmited to purely thermal effects.

Expenments were conducted to assess the effects of Ngh potental testing of cables and to assess the survivabihty of NUREG/CR4104: SHEAR WALL ULTIMATE DRIFT LIMITS.

aged and damaged cables under loss-of-coolant accident DUFFEY,T.A.

Consulting Engineer.

GOLDMAN,A/

(LOCA) conditions. High potential testng at 240 Vdc/md on un.

FARRAR.CA Los Alamos National Laboratory. Apra 1994$

damaged cables suggested that no damage was incurred on 116pp. 9405310177. LA-12649-MS. 79590:111*

the selected virgin cables. Dunng aging and LOCA testin0, Oko-Onh hmns fu Mnfaced-cmcrete shear waHs am inveshgated nite ethylene propylene rubber (EPR) cables with a bonded by reviewing the open hterature for appropriate experimental jacket en,ced unexpected failures. The failures appear to data. Dnft values at ultimate load are determined for walls with be primarily related to the level of thermal aging and the pres-aspect rabos ranging up to a maximum of 3.53 and undergoing ence of a bonded jacket that ages more rapidly than the insula-d'fferent types of lateral loading (cyclic static, monotonic static, tion. For Brand Rex cross linked polyolefin (XLPO) cables, the and dynamic). Based on the geometry of actual nuclear power results suggest that 7 mits of insulaton remaining should give plant structures exclusive of containments and concerns regard-the cCJes a high probabdity of surviving accident exposure foi-ing their response during seismic (i.e,cychc) loading, data are

%ving aging. The voltage necessary to detect when 7 mits of obtained from pertnent references for which the wall aspect inalation remain on unaged Brand Rex cables is approximatety ratio is less than or equal to approximately 1, and for which 35 k%1c. This voltage level would almost certainly be unaccept.

teshng is cyche in nature (typicah displacement controMed). In able to a utility for use as a damage assessment tool. However, particular, lateral deflections at ultimate load, and at points in addebonal tests indicated that a 35 kVdc voltage application the sonening rege beyond ultimate for which the load has would not damage virgin Brand Rex cables when tested in dropped to 90,80,70,60, and 50 percent of its ultimate value, water. Although two Rockbestos silicone rubber cables failed are obtained and converted to drift infamation. The stahshcal during the accident test, no correlation between failures and nature of the data is also investigated. These data are shown to level of damage was apparent.

be lognormally distritmted, and an analysis of variance is per.

NUREG/CR4102: VALIDATION OF THE SCALE BROAD formed. The use of these stabstes to estimate Probabihty of STRUCTURE 44-GROUP ENDF/B-Y CROSS-SECTION Ll-Failure for a shear wall structure is Hlustrated.

BRARY FOR USE IN CRITICALITY SAFETY ANALYSES.

DEHART,M.D.; BOWMAN,S.M. Oak Ridge National Laboratory.

NUREG/CR4105: HUMAN FACTORS ENGINEERING GUID-September 1994. 147pp. 9411080072. ORNL/TM-12460.

ANCE FOR THE REVIEW OF ADVANCED ALARM SYSTCMS.

81649:116.

O'HARA,J.M.; BROWN.WA; H!GGINS J.C. Brookhaven Nabon-This report documents the vahdation of the recently devel-al t eboratusy. September 1994. 124pp. 9411000068 BNL-oped 44-group ENDF/B-V based cross-sectinn hbrary NUREG.52391. 81649:001,

Main Citations and Abstracts 29 This report provides guidance to support the review of the NUREG/CR4114 V03: PERFORMANCE ASSESSMENT OF A human factors aspects of advanced alarm system designs in HYPOTHETICAL LOW-LEVEL WASTE FACILITY. Groundwater nuclear power plants. The report is organized into three major Flow And Transport Simulation. TALBOTT,M.E.; GELHAR.L.W.

sections. The first section desenbes the methodology and ente.

Massachusetts institute of Technology, Cambridge, MA. May ria that were used to develop the design review guidelines. Also 1994. 82pp. 9406200318. 79830:203.

included is a description of the scope, organization, and format Stochastic subsurface hydrologc theory is apphed to data for of the guidelines. The second section provides a systematic a hypothetical low-level waste site to demonstrate the features review procedure in which important characteristics of the alarm of the hydraule parameter estimation process, as developed by system are identified, desenbed, and evaluated. The third sec-Gelhar and others. Effectrve values of hydraulic conductivity, tion provides the detailed review guidelines. The review guide-macrodspersivh, and macrodspersivity enhancement are es4 rnated from the data n hs mannet A two.&nenemal seMab hnes are organized according to important characteristics of the ed now and transpwt We-elenet computer e is uM M alarm system includng: alarm definition; alarm processing and reduction; alarm priontizabon and availabihty; display; control; model the sete. Four efferent isotope inputs and two types of input configurations contnbute to an evaluation of model sensi-automated, dynarnic, and modifiable characterishes; reliability, tivities. These sensitivities of the mean concentrations and the test, maintenance, and failure indication; alarm response proce-uncertainties around the mean are explored using an analytical dures; and control-desplay integration and layout.

model as an example. Results indcate that the spatial hetero.

genesty of isotope sorption, through its contnbution to longitudi-NUfdG/CR4107:

SUMMARY

OF MELCOR 1.8.2 CALCULA~

nel dispersnnty enhancement, has a large effect on the magni-TIONS FOR THREE LOCA SEQUENCES (AG,S2D & S3D) AT tude of concentration predictions, especially for isotopes with THE SURRY PLANT, KMETYK,LN. Sandia National Laborato-short half-hves in compenson to their retarded mean travel ries. SMITH.L Geo-Centers, Inc. March 1994. 200pp.

times. This observation indicates the need for accurate site data 9404080073. SAND 93-2042. 78794:001-measurements that -@,661 the parameter estimation proc-Activities involving regulatory implementabon of updated ess. A comparison of simphfied analyt cal screening models with source term informahon were pursued. These activities include the numerical model predchons shows that the analytical the identificabon of the source term, the ident fication of the models tend to undereshmate concentration levels at low times, chemical form of iodne in the source term, and the timing of potentially as a result of oversimphfication of the flow field.

the source term's entrance into containment. These activities Future models could address aspects that are neglected in this are intended to support a more realistic source term for hcens-report, such as three-dmonsionality or unsaturated flow and ing nuclear power plants than the current TID-14844 source transport.

term and current hcenssng assumptions. MELCOR calculations NUREG/CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR were performed to support the technical basis for the updated source term. This report presents the results from three HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-PHIRE) VERSION 5.0. Technical Reference Manual.

MELCOR calculations of nuclear power plant accident se-RUSSELL,K.D.; ATWOOD,C.L; GALYEAN.W.J.; et al EG&G quences and presents comparisons with Source Term Code Idaho, Inc. July 1994. 131pp. 9408120223. EGG-2716.

Package (STCP) calculabons for the same sequences. The 80517:001.

three low-pressure sequences were analyzed to identify the ma-The Systems Analysis Programs for Hands-on integrated Reli-tetials which enter containment (source terms) and are available ability Evaluabons (SAPHIRE) refers to a set of several micro-for release to the environment, and to obtain bming of se-computer programs that were developed to create and analyze quence events. The source terms include fission products and probabilistic risk assessments (PRAs), primanly for nuclear other matenals such as those generated by core-concrete inter-power plants. This volume provides information on the princ6-actions. All three calculations, for both MELCOR and STCP, pies used in the construction and operabon of Version 5.0 of analyzed the Surry plant, a pressurized water reactor (PWR) the Integrated Reliabihty and Risk Analysis System (IRRAS) and with a subatmospheric containment design.

the System Analysis and Risk Assessment (SARA) system. It summarizes the fundamental mathemabcal concepts of sets i

NUREG/CR4112 DRF FC: IMPACT OF REDUCED DOSE LIMITS and logic, fault trees, and probabehty This volume then de.

ON NRC LICENSED ACTIVITIES. Major issues in The imple-scribes the algonthms that these programs use to construct a mentation Of ICRP/NCRP Dose Limit Recommendations. Draft fault tree and to obtain the minimal cut sets. It gives the formu-Report For Comment. MEINHOLD.C.B. Brookhaven National las used to obtain the probabihty of the top event from the mini-Laboratory. January 1994. 75pp, 9402220146. BNL-NUREG-mal cut sets, and the formulas for probabihties that are appro-52394. 78191:001, priate under various assumphons concorrung repairability and This report summarizes information required to estimate, at mission tsme, it defines the measures of basic event importance least in quaktative terms, the potential impacts of reducing oc.

that these programs can calculate. This volume gives an over-cupational dose limits below those given in 10 CFR 20 (Re-view of uncertainty analysis using sample Monte Carlo sampling vised). The data from a questionnaire developed for this project or Lahn Hypercube samphng, and states the algonthms used by and data from existing surveys were used to estimate the these programs to generate random-basse event probabilities I

impact of three dose hmit options; 10 mSv yr( 1),20 mSv yr(-1),

from various dstnbubons. Further references are given, and a and a combination of an annual hmit of 50 mSv yr( 1) coupled detailed example of the reductson and quanbfication of a simple with a cumulative hmit in rem equal to age in years. The overall fault tree is provided in an appendix.

conclusions of the study are: (1) Although 10 mSv yr(1) is a NUREG/CR4116 V02: SYSTEMS ANALYSIS PROGRAMS FOR reasonable limit for many licensees, such a limit could be ex*

HANDS-ON INTEGRATED RELLABILITY EVALUATIONS (SA-traordinanly dfficult and potentialty destructive to some. (2)

PHIRE) VERSION 5.0. Integrated Rehatulity And Risk Analysis Twenty mSv yr(-1) as a hmst is posseble for some of the latter System (IRRAS) Reference Manual.

RUSSELL,K.D.;

groups, but for others it wenid prove 4thcult. (3) Fifty mSv yr(-1)

KVARFORDT,K.J.: SKINNER.N.L; et al. EGSG Idaho, Inc. July and ago in 10's of mSv would appear acceptable both in terms 1994. 422pp. 9408150043. EGG-2716. 80542:032.

of the related blotime risk of cancer and severe genetsc effects The Systems Analysis Programs for Hands-on integrated Reli-to the most highly exposed and in terms of practicality of oper-abihty Evaluabons (SAPHIRE) refers to a set of several micro-ation. This

-x-Mz"y in some segments of the industry is computer programs that were developed to create and analyze based on the adophon of a " grandfather clause" for those ex-probabilishc nsk assessments (PRAs), pnmenty for nuclear coedng the cumulabve limst.

power plants. The integrated Rehatxhty and Risk Analysis

30 Main Citations and Abstracts System (IRRAS) is a state-of the-art, microcomputer-based NUREG/CR4116 V05: SYSTEMS ANALYSIS PROGRAMS FOR probabilistic rick assessment (PRA) model development and HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-analyss tool to address key nuclear plant safety issues. IRRAS PHIRE) VERSION 5.0. Systems Analysis And Risk Assessment is an integrated software tool that gives the user the ability to (SARA) Tutorial Manual. SATTISON.M B.; RUSSELL,K.D.;

create and analyze fault trees and accident sequences using a SKINNER,N.L. EG8G Idaho, Inc. July 1994.

110pp.

microcomputer. This program provides functons that range from 9408150072. EGG-2716. 80544:001, graphical fault tree constructen to cut set generaton and quan.

The Systems Analyss Programs for Hands-on integrated Reh-tification to report generation. Version 5.0 of IRRAS provides ability Evaluatons (SAPHIRE) refers to a set of several micro-the same capabilitios as earher versons and adds the ability to computer programs that were developed to create and analyze 1

perform location transformations, seismic analysis, and provides probabilistic nsk assessments (PRAs), pnmanly for nuclear power plants. This volume is the tutonal manual for the Systems enhancements to the user interface as well as improved algo-ahss aM M Assesse @@ Wm km M a nthm performance. Additonally, verson 5 0 contains new alpha-of a family', daw sysMm M M anaW h @ be ocen#

numenc fault tree and event tree editors, and a powerful set of

[i.e., a power plant, a manufactunng facili*y, any macro based rule editors. These editors are used for event tree facdtty on which a probabilistic nsk assessment (PRA) meget be rules, recovery rules, and end state partitioning performed) A senes of lessons is provided that guides the iser through some basic steps common to most analyses perfor.ned NUREG/CR 6116 V03: SYSTEMS ANALYSIS PROGRAMS FOR with SARA. The example problems presented in the lessons HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-build on one another, and in combinaton, lead the user tarough PHIRE) VERSION 5 0 Integrated Reliability And Risk Analysis all aspects of SARA sensitivity analyss capabilities.

System (IRRAS)

Tutonal Manual.

VANHORN,R.L.;

RUSSELL,K.D.; SKINNER.N.L EG&G Idaho, Inc. July 1994.

NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR 174pp. 9408150049. EGG-2716. 80544:106.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-The Systems Analysis Programs for Hands-on integrated Reli-PHIRE) VERSION 5.0. Fault Tree, Event Tree, And Piping & In-ablity Evaluatons (SAPHIRE) refers to a set of several micro-strumentaten Diagram (FEP) Editors Reference Manual.

computer programs that were developed to create and analyze MCKAY,M.K.; SKINNER,N L.; WOOD,S.T. EG&G Idaho, leic.

prababilistic nsk assessments (PRAs), pnmanly for nuclear July 1994.147pp. 9408150081. EGG 2716. 80537:144.

power plants. This volume is the tutonal manual for the integrat.

The Systems Analysis Programs for HandsSn integrated Reli-ed Reliability and Risk Analysis System (IRRAS) Version 5.0, a ability Evaluatens (SAPHIRE) refers to a set of several micro-state-of-the-art, microcomputer-based probabiliste nsk assess, computer programs that were developed to create and analyze ment (PRA) model development and analysis tool to address probabilist c nsk assessments (PRAs), pnmanly for nuclear key nuclear plant safety issues. IRRAS is an integrated software power plants. The Fault Tree, Event Tree, and Piping & Instru-tool that gives the user the ability to create and analyze fault mentation Diagram (FEP) editors allow the user to graphically trees and accident sequences using a microcomputer. A senes build and edit fault trees, event trees, and piping & instrumenta-of lessons is provided that guides the user through basec steps tion diagrams (PalDs). The software is designed to enable the independent use of the graphical-based editors found in the in-common to most analyses performed with IRRAS. The tutonal is WaW Ma@ aM M Assessment System WW M divided into two major sectons: basic and additonal features.

s canmM d h mak Ws Faun % M h, The basic section contains lessons that lead the student and Piping & Instrumentaten Diagram) and a utdity module. This through development of a very simple problem in IRRAS, high-reference manual provides a screen-by-screen guide of the lighting the program's most basic features. The additional fea-entire FEP System.

tures section contains lessons that expand on basc analysis features of IRRAS 5 0.

NUREG/CR-6116 V06: SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-NUREG/CR-6116 V04: SYSTEMS ANALYSIS PROGRAMS FOR PHIRE) VERSION 5.0.Models And Results Database (MAR D)

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-Reference Manual. RUSSELL,K.D.; SKINNER,N.L. EG&G Idaho, PHIRE) VERSION 5 0. Systems Analysis And Risk Assessment Inc. July 1994.134pp. 9408150094. EGG-2716. 80544280.

(SARA) Reference Manual. RUSSELL,K.D.; KVARFORDT,K.J.;

The Systems Analysis Programs for Hands-on integrated Reli-SKINNER.N L.; et al. EG8G Idaho, Inc. July 1994. 291pp.

ability Evalustons (SAPHIRE) refers to a set of several mero-9408150057. EGG-2716 80541:105.

computer programs that were developed to create and analyze The Systems Analysis Programs for Hands-on integrated Reli.

probabiliste nsk assessments (PRAs), pnmanly for nuclear ability Evaluatons (SAPHIRE) refers to a set of several mero-power plants. The pnmary function of MAR-D is to create a data computer programs that were developed to create and analyze repository for completed PRAs and Individual Plant Examina-probabilistic nsk assessments (PRAs), pnmanly for nuclear tions (IPEs) by providing input, converson, and output capabell-power plants. This volume is the reference manual for the Sys-ties for data used by IRRAS, SARA. SETS, and FRANTIC soft.

toms Analysis and Risk Assessment (SARA) System Verson ware. As probabiliste nsk assessments and individual plant ex.

amenatons are submitted to the NRC for review, MAR D can be 5.0, a mecrocomputer-based system used to analyze the safety used to const h nodels and msuus kan h s% fa use issues of a " family" p e., a power plant, a manufactunng facility, with IRRAS and SARA. Then, these data can be easily ac-any facility on which a probabilistic nsk assessment (PRA) might be performed). The SARA database contains PRA data pnmanly for the dominant accident sequences of a famdy and desenptive view of the functions available within MAR.D and step-by step information about the famdy including event trees, fault trees, operating instructions.

and system model diagrams. To simulate changes to f amily sys-tems, SARA users change the failure rates of initiating and NUREG/CR-6120: CONTROLLED FIELD STUDY FOR VALIDA-base events and/or modify the structure of the cut sets that TION OF VADOSE ZONE TRANSPORT MODELS.

make up the event trees, fault trees, and systems. The user WlERENGA.P.J 4 WARRICK,A.W.; et al Arizona, Univ. of, then evaluates the effects of these changes through the recal-Tucson, AZ. HILLS.R.G. New Mexico State Univ., Las Cruces, culation of the resultant accident sequence probabilities and im-NM. August 1994 27pp. 9409060223. 80785-001.

portance measures The results are displayed in tables and Predcton of radionuclide migration through soil and ground-graphs that may be ponted for reports.

water requires models which have been tested under a vanety of conditions. Unfortunately, many of the existing models have

- - J

Main C!tations and Abstracts 31 not been tested in the field, partty because such testing requires ing egulatons and pmch in the eani #980s were made in accurate and representative data. This report provides the resW. set to the accid ent at the Three Via Island Unit 2 nuclear design of a large scale field experiment representatrve, in terms power pant in 1979. These char.ges in.fuded widespread issu-of surface area and depth of vadose zone, of an actual disposal ance of new operatrr and licensing rc4rements and the estab-area. Expenments are proposed which will yield documented lishment of naborW training centers. A priod of relative stability data, of sufficient scale, to allow testmg of a variety of models followed these c1ango Changes in the %r half of the 1980s including effective media stochaste models and deterministe focused on continuir.g improvements arJ additics to training models. Details of the methodology and procedures to be used cumcula and methods, most notably incimed reliance on sim-in the expenment are presented.

ulator training.

hUREG/CR-6121: COMPONENT EVALUATION FOR INTERSYS-NUREG/CR OM6: COGNITIVE SKILL TRAINING FOR NUCLEAR TEM LOSS-OF-COOLANT ACCIDENTS IN ADVANCED LIGHT POWER PLANT OPERATIONAL DECISION-MAKING.

WATER REACTORS. WARE.A.G. EGAG Idaho, Inc. July 1994.

MUMAW,R.d; SMTZLER.D.; ROTH,E.M.; et al. Westinghouse 113pp. 9408250010. EGG-2717. 80658:177.

Electnc Corp. Ju,'M 1994. 200pp. 9407010300. 80031:240.

Usmg the methodology outlined in NUREG/CR 5603 this Training for operPor and other technical positons in the com-report evaluates (on a probabilists basis) design rules for com-mercial nuclear powor industry traditonally has focused on mas-ponents m ALWRs that could be subjected to intersystem loss-tery of the formal r,ocedures used to control plant systems and of-coolant accidents (ISLOCAs). The methodology is intended processes. However, decision-making tasks required of nuclear for piping elements, flange connections, orkline pumps and power plant operators involve cognatue skills (e.g., situation as-valves, and heat exchangers. The NRC has directed that the sessment, p'anning). Cognitive skills are needed in situations design rules be evaluated for BWR pressures of 7.04 MPa where formal procedures may not exist or may not be as pre-(1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 senptive, as is the case in severe accident management (SAM).

degrees C (350 degrees F), and has established a goal of 90%

The Westinghouse research team investigated the potential probabihty that system rupture will not occur during an ISLOCA cognitive demands of SAM on the control room operators and event. The results of the calculations in this report show that Technical Support Center staff wfo would be most involved in components designed for a pressure of 0.4 of the reactor cool-the selection and execution of severe accident control actions.

ant system operating pressure will satisfy the NRC survival goal A model of decision making, organized around six general cog-n most cases. Spectic recommendations for component native processes, was developed to identify the types of cogni-strengths for BWR and PWR appications are made in the tive skills that may be needed for effective performance. Also, 1

report. A peer review panel of natonally recognized experts was twelve SAM scenarios were developed to reveal specific deci-selected to rewew and entioue the insbal results of this program.

sion-making difficulties. Following the identifcaton of relevant NUREG/CR-6122: STAFFING I 'ECISION PROCESSES AND cognitive skills,19 approaches for training individual and team ISSUES. Case Studies Of Seven U.S. Nuclear Power Plants.

cognitive skills were identified. A review of these approaches re-MELBER.B.; ROUSSEL,A.; BAXER,K.; et al. Battelle Human Af-suited in the identificabon of general charactenstes that are im-fairs Research Centers. March 1994. 53pp. 9404040039. PNL.

portant to effective training of cognitive skills.

8428. 78736:001.

The objective of this repoit is to identify how decisions are NUREG/CR-6127: THE EFFECTS OF STRESS ON NUCLEAR made regarding staffing leveis and positions for U.S. nuclear POWER PLANT OPERATIONAL DECISION MAKING AND power plants. In this report, a 'ramework is provided for under.

TRAINING APPROACHES TO REDUCE STRESS EFFECTS.

standing the major forces dnvirs sta'fing and the impications of MUMAW.R.J. Westinghouse Electnc Corp. August 1994. 43pp.

staffing docesons for plant safety The focus of this report is on 9408250017. 80656:278.

dnving forces that have led to changes in staffing levels and to Operational personnel may be exposed to significantly levels the estabhshment of new positions between the mid 1980s and of stress during unexpocted changes in plant state and pit.nt j

the earty 1990s. Processes used at utilibes and nuclear power emergencies. The decison making that idenbfies operational ac-J plants to make and implement these staffing decisons are also tions, which is strongfy determined by procedures, may be af-discussed in the report. While general trenos affecting the plant fected by stress, and performance may be impaired. This report as a whole are presented, the major emphasis of this report is analyzes potential effects of stress in nuclear power plant (NPP) on staffing changes and practces in the operatons department, settings, especially in the context of severe accident snage-including the operations shift crew. The findings in this report ment (SAM). First, potential sources of stress in the f<r'sOing are based on intennews conducted at seven nuclear power are identified. This analysts is followed by a revor of the wayc 1

plants and thett parent utilities. A discusson of the key findings in whch stress is likely to affect performance, with an emphasis is followed by a summary of the impications of staffing issues on performance of cognitive skills that are linked to operational for plant safety.

decision making Finalty, potenbal training approaches for reduc-ing or eliminating stress affects are identified. Several training NUREG/CR-6123: AN INTERNATIONAL COMPARISON OF approaches have the potential to eliminate or mitigate stress ef-COMMERCIAL NUCLEAR POWER PLANT STAFFING REGU-fects on cognitive skill performance. First, the use of simulated LATIONS AND PRACTICE.1980-1990. MELBER,B.; HAUTH,J.;

events for training can reduce the novelty and uncertainty that TERRILL E.; et al. Battelle Human Affairs Research Centers.

can lead to stress and performance impairments. Second, train-March 1994. 70pp. 9404040052. PNL-8564. 78736:123.

ing to make cognitrve processing more efficient and less reliant A review of nuclear power plant staffing regulatory and 6ndus-on attenton and memory resources can offset the reductons in try practces in Canada, France, Germany, Japan, Sweden, and these resources that occur under stressful condibons. Third, the United Kingdom is presented in this report. Internatonal training that targets crew communicatons skills can reduce the trends in staffing regulatory approaches, industry practces, and likelihood that communicatons will fait under stress.

issues of concern that have potential relevance for the contin-ued development of nuclear power plant staffing polcy in the NUREG/CR-6t28: PIPING BENCHMARK PROBLEMS FOR THE United States are highhghted. The 1980s were marked by signif.

ABB/CE SYSTEM 80+ STANDARDIZED PLANT. BEZLER.P.;

icant growth in nuclear power operations internabonally; howev-DEGRASSI,D.; BRAVERMAN.J.; et al. Brookhaven Natonal er, growth of nuclear power is not expe:ted to continue in the Laboratory. July 1994. 234pp. 9408150117. BNL-NUREG-1990s except in France and Japan. A continuum of regulatory 52396 80543.090.

approaches was identified, ranging from presenptive regulations To satisfy the need for venfcation of the computer programs apphed to all licensees to staffing requirements agreed to in the and modeling techniques that will be used to perform the final plant operating Icenses. Most of the changes observed in staff-piping analysis for the ABB/ Combuston Engineenng System

32 Main Citations and Abstracts 80+ Standardized Plant, three benchmark problems were de-This report documents user instructions for several simplified veloped. The problems are representative piping systems sub-subroutines and dnver programs that can be used to estimate q

jected to representative dynamic loads with solubons developed various aspects of the long-term performance of cement-based using the methods being proposed for analysis for the System barners used in low-level radioactive waste disposal facshties.

80+ standard design. It will be required that the combined li-The subroutines are prepared in a modular fasNon to allow cense heensees demonstrate that their solutions to these prob-flexibility for a variety of applicabons. Three levels of codes are tems are in agreement with the benchmark problem set.

provided; the individual subroutines, enteractive dnvers for each NUREG/CR-6132: BIAXIAL LOADING AND SHALLOW-FLAW EF-of the subroutines, and an interachve main dnver, CEMENT, FECTS ON CRACK TIP CONSTRAINT AND FRACTURE that calls each of the individual drivers. The indivkiual subrou-TOUGHNESS. BASS,B.R.; BRYSON,J.W.; THEISS,T.J.; et al.

tines for the different models may be taken independently and Oak Ridge National Laboratory. January 1994. 74pp.

used in larger programs, or the driver modules can be used to 9403140303. ORNL/TM-12498. 78477:001.

execute the subroubnos separately or as a part of the main A program to develop and evaluate fracture methodologies dnver routine. A brief program descriphon is included and user-for the assessment of crack-tip constraint effects on fracture interface instructions for the individual subroutines are docu-toughness of reactor pressure vessel (RPV) steels has been ini-mented in the main report. These are intended to be used when bated in the Heavy-Secton Steel Technology (HSST) Program.

the subroutines are used as subroutines in a larger computer Crack bp constraint is an issue that significantly impacts fracture code. User instructons for the dnvers and example interactive mechanics technologies employed in safety assessment proce-screens for the main dnvers are provided in Appendix A. Exam-dures for commercially heensed nuclear RPVs. The focus of pies showing the use of the individual driver routines to execute studies desenbod herein is on the evaluation of two stressed-the different subroutines and test data are included in Appendix based methodolog,es for quantifying crack-tip constraint (i.e., J-B. Programmer notes are provided in Appendix C and the con-1 O theory and a micromechanical scaling model based on enbcal figuration control system is discussed in Appendix D.

stressed volumes) through applicatons to expenmental and fractographe data. Data were ublized from single-edge notch NUREG/CR-6139: CRACK-ARREST TESTS ON TWO IRRADIAT-bend (SENB) specimens and HSST-developed cruciform beam ED HIGH-COPPER WELDS. Phase ll: Results Of Duplex-Type specimens that were tested in HSST shallow-crack ard biaxial Specimens. ISKANDER,S.K.; CORWIN.W.R.; NANSTAD,R.K.

testing programs. Results from applicatons indicate hat both Oak Ridge Natonal Laboratory. March 1994. 138pp.

the J-O methodology and the micromechanical scaibg model 9403140206. ORNL/TM-12513. 78476:176.

can be used successfully to interpret expenmental dah fr~.

the shallow-and deep crack SENB specimen tests. When ap The objective of the Heavy-Section Steel Irradiation Program plied to the urwaxially and biaxially loaded cruciform specimens, Sixth trradiation Senes is to determine the effect of neutron irra-the two methodologies showed some promising features, but diation on the shift and shape of the lower-bound curve to also raised several questions conceming the interpretation of crack-arrest toughness data. Two submerged-arc welds with constraint conditions in the specimen based on near.bp stress copper contents of 023 and 0.31 wt % were commercially fab-fields. Fractographic data taken from the fracture surfaces of ricated in 200 mm-thicn plate. Crack-arrest specimens fabricat-the SENB and cruciform specimens are used to assess the rei, ed from these weids were irradiated at a nominal temperature evance of stress-based fracture characterizatons to conditons of 288 degrees C to an average fluence of 1.9 x 10(19) neu-at cleavage initiaton sites. Unresolved issues identified from trons/cm(2)(> 1 MeV). This is the second report giving the re-these analyses require resolution as part of a validaten process suits of the tests on irradiated du, plex-type crack arrest s,peci-for biaxial loading apphcatens. This report is designated as mens. Charpy V-notch specimens irradiated in the same cap-HSST Report No.142.

sules as the crack arrest specimens were also tested, and a 41-NUREG/CR4133: FRAGMENTATION AND OUENCH BEHAVIOR J transiton temperature stift was determined from these speci-OF CORIUM MELT STREAMS IN WATER. SPENCER.B.W.;

mens. "Mean" curves of the same form as the ASME K(la)

WANG.K.; BLOMOUIST,C.A.; et al Argonne Natonal Laborato-curve were fit to all the data with only the reference tempera-ry. February 1994. 407pp. 9404010160. ANL-93/32. 78717:066.

ture as a parameter. The sNft between the mean curves agrees The interaction of molten core materials with water has been well with the 41 J transition temperature sNft obtained from the investigated for the pour stream mixing rnode. This interaction Charpy V-notch specimen tests.

plays a crucial role dunng the later stages of n-vessel core melt progresson inside a light water reactor such e dunng the TMI.

NUREG/CR4142: TENSILE-PROPERTY CHARACTERIZATION 2 accident. The key issues wNch anse dunng t.% molten core OF THERMALLY AGED CAST STAINLESS STEELS.

relocaton include: i) the thermal attack and possible damage to MICHAUD,W.F.; TOBEN,P.T.; SOPPET,W.K.; et al. Argonne Na-the RPV lower head from the impingsng molton fuei stream and/

tonal Laboratory. February 1994. 257pp. 9402250117. ANL-93/

or the debns bed, ii) the molten fuel relocation pathways includ-

35. 78278:001.

ing the effects of redistnbuton due to core support structure The effect of thermal aging on tensile properties of cast stain-Cnd the reactor lower internals, iii) the quench rate of the less steels during service in light water reactors has been evalu-molten fuel through the water in the lower plenum, iv) the steam ated. Tensile data for several experimental and commercial generation and hydrogen generaton during the interacton, v) heats of cast stainless steels are presented. Thermal aging in-the transient pressurization of the pnmary system, and vi) the creases the tensile strength of these steels. The high-C Mo-possibihty of a steam explosion. In order to understand these bearing CF-8M steels are more susceptible to thermal aging issues, a senes of six experiments (designated CCM-1 through -

than the Mo-free CF-3 or CF-8 steels. A procedure and correla-

6) was performed in which molten conum paswd through a deep pool of water in a long, slender pour stream mode. Re-tions are presented for predecting the change in tensile flow and suits discussed include the transient temperatures and pres-yield stresses and engineenng stress-vs.-strain curve of csst gg g,,,

g gg g

The knsde properties of aged cast stainless steel are estimrated dt pos t ch r s

from known material information, i.e., chemical corr. position and NUREG/CR4158: USER'S GUIDE FOR SIMPLIFIED COMPUTER the initial tensile strength of the steel. The cotelations de-MODELS FOR THE ESTIMATION OF LONG-TERM PERFORM-scribed in tfws report may be used for assessing foermal embrit-ANCE OF CEMENT-PASED MATEHIALS. PLANSKY,LE.;

tiement of cast stainless steel components.

SEITZ,R.R. EG&G Elaho, Inc. February 1994. 94pp.

9403140184. EGG-27i9. 78456:092.

i

Main Citations and Abstracts 33 NUREG/CR-6143 V02 PIA: EVALUATION OF POTENTIAL NUREG/CR-6143 V02PT4: EVALUATION OF POTENTIAL SEVERE ACCIDENTS DURING LOW POWER AND SHUT-SEVERE ACCIDENTS DURING LOW POWER AND SHUT-DOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of DOWN OPERATIONS AT GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From Internal Events For Plant Oper-Core Damage Frequency From Internal Events For Plant Oper-ational State 5 Dunng A Refuehng Outage.Sectons 1-9.

ational State 5 Dunng A Refueling Outage.intemal...

WHITEHEAD.D.W. Sandia National Laboratones. DARBY,J. Sci-FORESTER.J. Science Applications intemational Corp. (formerly ence & Engineenng Associates, Inc. YAKLE,J.; et al. Science Science Applications, Inc.). WHITEHEAD D.W. Sandia National Applications International Corp. (formerly Science Applications, Laboratories. DARBY,J.; et al. Science & Engineenng Associ-inc.).

June 1994. 284pp. 9408030161. SAND 93 2440.

ates, Inc. June 1994. 889pp. 9408030243. SAND 93-2440.

80442:001.

80439:001.

This document contains the accident sequence analysis of in-See NUREG/CR-6143,V02,PT3 abstract.

ternally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 dunng NUREG/CR-6143 V03: EVALUATION OF POTENTIAL SEVERE a refueling outage. The report documents the methodology ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-used dunng the analysis, describes the results from the applica.

ATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Damage tion of the methodology, and compares the results with the re.

Frequency From Internal Events For Plant C perational State 5 suits from two full power analyses performed on Grand Gulf.

During A Refueling Outage. LAMBRIGHT.J. Sandia National Laboratories. ROSS,S.; LYNCH,J.; et al. Science & Engineenng NUREG/CR-6143 V02PIB: EVALUATION OF POTENTIAL Associates, Inc. July 1994.112pp. 9408180206. SAND 93-2440.

SEVERE ACCIDENTS DURING LOW POWER AND SHUT-80613:244.

DOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Thrs report presents the details of the analysis of core Core Damage Frequency From internal Events For Plant Oper-damage frequency due to fire during shutdown Plant Operation-ational State 5 Dunng A Refueling Outage.Section 10.

at State 5 at the Grand Gulf Nuclear Staten. Insights from pre.

WHITEHEAD,0.W. Sandia National Laboratones. DARBY,J. Sci.

vious fire analyses (Peach Bottom Surry, LaSalle) were used to ence & Engineenng Associates, Inc. YAKLE,J.; et al. Science the greatest extent possble in this analysis. The fire analysis Applications International Corp. (formerly Science Applications, was fully integrated utihzing the same event trees and fault Inc.).

June 1994. 997pp. 9408030170. SAND 93-2440.

trees that were used in the internal events analysis. In assess-80426.001, ing shutdown nsk due to fire at Grand Gulf, a detailed screening See NUREG/CR-6143,V02 PI A abstract, was performed which included the following elements: (a) Com-puter-aided vital area analysis, (b) Plant inspections, (c) Credit NUREG/CR-6143 V02PIC: EVALUAT!ON OF POTENTIAL for automatic fire protection systems, (d) Recovery of random SEVERE ACCIDENTS DURING LOW POWER AND SHUT-failures, and (e) Detailed fire propagation modehng. This screen-DOWN OPERATIONS AT GRAND GULF UNIT 1. Analysis Of ing process revealed that all plant areas had a neglgible Core Damage Frequency From Intemal Events For Plant Oper-(<l.0E4 per year) contnbuten to fire-induced core damage ational State 5 Dunng A Refueling Outage Main Report-frequency.

WHITEHEAD D.W. Sandia National Laboratones. DARBY,J. Sci-ence & Engineering Associates, Inc. YAKLE J.; et al. Science NUREG/CR-6143 V04: EVALUATION OF POTENTIAL SEVERE Applications International Corp. (formerly Science Apphcatons, ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-inc.).

June 1994. 373pp. 9408030176. SAND 93 2440.

ATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Damage 80443:001.

Frequency From Internally Induced Flooding Events For Plant See NUREG/CR4143,V02,P1 A abstract.

Operatonal State 5 During a Refueling...

DANDINI,V.;

STAPLE,B.D.;

et al.

Sandia National Laboratones.

NUREG/CR-6143 V02PT2: EVALUATION OF POTENTIAL FORESTER.J. Science Apphcatons Intemational Corp. (formerfy SEVERE ACCIDENTS DURING LOW POWER AND SHtJT-Science Applications, Inc.). Jufy 1994. 472pp. 9409070061.

DOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of SAND 93-2440. 80787:001.

Core Damage Frequency From intemal Events For Plant Oper-An estimate of the contributen of internal flooding to the ational State 5 During Refueling Outage. internal.. DARBY,J.

mean core damage frequency at the Grand Gulf Nuclear Station Science & Engineering Associates, Inc. WHITEHEAD,D.W.;

was calculated for Plant Operatonal State 5 during a refuehng STAPLE,8.D.; et al. Sandia Natonal Laboratories. June 1994.

outage. Pursuant to this objective, flood zones and sources G09pp. 9408030187. SAND 93-2440. 80432:001.

were identified and flood volumes were calculated. Equipment This document contains the accident sequence analysis of in.

necessary for the maintenance of plant safety was identified ternally initiated events for Grand Gulf, Unit 1 as it operates in and its vulnerabihty to flooding was determined. Event trees and the Low Powar and Shutdown Plant Operational State 5 dunng fault trees were modified or developed as required, and PRA a refueling outage. The report documents the methodology quantification was performed using the IRRAS code. The mean used dunng the analysis, describes the results from the apphca-core damage frequency estimate for GGNS dunng POS 5 was tion of the methodology, and compares the results with the re-found to be 2.3 E-8 per year.

suits from two full power analyses performed on Grand Gulf.

NUREG/CR 6143 V05: EVALUATION OF POTENTIAL SEVERE

'AUREG/CR-6143 V02PT3: EVALUATION OF POTENTIAL ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER.

SEVERE ACCIDENTS DURING LOW POWER AND SHUT-ATIONS AT GRAND GULF, UNIT 1, Analysis Of Core Damage DOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Frequency From Seismic Events Dunng Mid-Loop Core Damage Frequency From Internal Events For Plant Oper.

Operations. Main Report. BUDNITZ,R.J. Future Resources Asso-ational State 5 During A Refuehng Outage.IntemaL.. YAKLE J.

ciates, Inc. DAVIS.P.R. PRD Consultsng RAVINDRA,M.K.; et al.

Science Apphcations intemational Corp. (formerly Science Ap-EOE, Inc. August 1994.110pp. 940923C278. 81006:083.

phcations, Inc.). DARBY,J. Science & Engineering Associates.

During 1989, the Nuclear Regulator / Commission (NRC) initi-Inc. WHITEHEAD,0.W.; et al. Sandia National Laboratories.

ated an extensive program to carefully examine the potential June 1994. 834pp. 9408030235. SAND 93-2440. 80429:001.

risks dunng low power and shutdown operations. The program This report provides supporting documentation for various includes two parallel projects ber.g performed by Brookhaven tasks associated with the performance of the probabihstic nsk Natonal Laboratory (Mi.) and Sandia Natonal Laboratories assessment for Plant Operational State 5 during a refuehng (SNL). Two plants, Surry (pressunzed water reactor) and Ge-~t outage at Grand Gulf, Unit I as documented in Volume 2, Part 1 Gulf (boiling water reactor), were selected as the plants to be of NUREG/CR-6143.

studed. The objectives of the program are to assess the risks

34 Main Citations and Abstracts of severe accidents irutiated dunng plant operational states SEVERE ACCIDENTS DURING LOW POWER AND SHUT-other than full power operation and to compare the estimated DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core core damage frequencies, important accident sequences and Damage Frequency From Intemal Events During Mid-Loop other qualitative and quanbtatue results with those accidents Operations. Appendices E (Sectons E.9-E.16).

CHU,T.L.;

irutiated dunng full power operabon as assessed in NUREG-MUSICKl.Z.; KOHUT,P.; et al. Brookhaven National Laboratory.

1150. The objectNe of this report is to document the approach June 1994.

502pp. 9408150234. DNL-NUREG-52399.

i utilized in the Grand Gulf plant and discuss the results obtained.

80 2

)

A parallel report for the Surry plant is prepared by SNL NUREG/CR4144 V02P4: EVALUATION OF POTENTIAL NUREG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE ACCIDENTS DURING LOW POWER AND SHUT.

SEVERE ACCIDENTS DURING LOW POWER AND SHUT-DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal Events Dunng M4 Loop Damage Frequency From Intemal Events During Mid-Loop Operatons. Appendices F-H. CHU,T.L; MUSICKI,Z.; KOHUT,P.;

Operatons. Main Report (Chapters 1-6). CHU,T.L; MUSICKl.Z.;

et al. Brookhaven National Laboratory. June 1994. 612pp.

KOHUT.P.; et al. Brookhaven Natonal Laboratory. June 1994.

9408150246. BNL-NUREG-52399. 80559:001.

494pp.9408120219. BNL-NUREG-52399. 80515:001.

See NUREG/CR-6144,V02,P1 A abstract.

Dunng 1989, the Nuclear Regulatory Commission (NRC) iruti-NUREG/CR4144 V02P5: EVALUATION OF POTENTIAL ated an extensue program to carefully examine the potentsal SEVERE ACCIDENTS DURING LOW POWER AND SHUT-nsks dunng low power and shutdown operatons. The program DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core includes two parallel protects being performed by Brookhaven Damage Frequency From internal Events Dunng Mid-Loop Natonal Laboratory (BNL) and Sandia National Laboratones Operatons. Appendices 1. CHU,T.L.; MUSICKl,Z.; KOHUT,P.; et (SNL). Two plants, Surry (pressunzed water reactor) and Grand al. Brookhaven National Laboratory. June 1994. 313pp.

Gulf (boshng water reactor), were selected as the plants to be 9408150284. BNL-NUREG-52399. 80558:001.

studied. The ob ectives of the program are to assess the risks See NUREG/CR-6144,V02 PI A abstract.

l of severe accidents initiated dunng plant operational states other than full power operation and to compare the estimated NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE ACCIDENTS DURING LOW POWER AND SHUT.

core damage frequencies, important accident sequences and DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core other qualitabve and quantitative results with those accidents initiated dunng full power operaton as assessed in NUREG-Damage Frequency From Internal Fires Dunng M 4 Loop Operatons. Main Report. MUSICKl.Z.; CHU,T.L Brookhaven Na-1150. The obtectue of this report is to document the approach tonal LabwaW HOW et at M, m Um MaM.

utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL This N REG 5 8 8 study shows that the core. damage frequency dunng mid-looo See NUREG/CR-6144,V02,P1 A abstract.

operabon at the Surry plant is comparable to that of power op-eration. We recogns2e that there is very large uncertainty in the NUREG/CR4144 V03 P2: EVALUATION OF POTENTIAL human error probabilibes in this study. This study identified that SEVERE ACCIDENTS DURING LOW POWER AND SHUT.

only a few procedures are available for mitigating accidents that DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core

(

may occur dunng shutdown. Procedures wntten specifically for Damage Frequency From Internal Fires Dunng Mid Loop shutdown accidents would be useful.

Operabons. Appendices. buSICKI,Z.; CHU,T.L Brookhaven Na-tonal Laboratory. HO,V.; et al. PLG, Inc. (formerly Pickard, NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL Lowe & Gamck, Inc.). July 1994. 400pp. 9408180149. BNL-SEVERE ACCIDENTS DURING LOW POWER AND SHUT-NUREG-52399. 80616:001.

DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core See NUREG/CR-6144,V02,P1 A abstract, I>amage Frequency From internal Events Dunng M4Lo p NUREG/CR4144 V04: EV/,LUATION OF POTENTIAL SEVERE Operatons. Main Report (Chapters 712). CHU,T.L; MUSICKI,2.;

ACCIDENTS DURING LCW POWER AND SHUTDOWN OPER-KOHUT,P.; et al. Brookhaven Natonal Laboratory. June 1994.

ATIONS AT SURRY,UNIf 1. Analysis Of Core Damage Frequen-630pp.9408150181. BNL-NUREG-52399. 80538.001.

cy From Internal Floods Ouring M4 Loop Operabons. KOHUT,P.

l See NUREG/CR-6144,V02,P1 A abstract.

Brookhaven National Laboratory. July 1994.

200pp.

9408180160. BNL-NUREG-52399. 80614.060.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL See NUREG/CR-6144,V02,P1 A abstract.

SEVERE ACCIDENTS DURING LOW POWER AND SHUT-NUREG/CR4144 V05: EVALUATION OF POTENTIAL SEVERE DOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-Damage Frequency From intemal Ev6a's Dunng M4 Loop ATIONS AT SURRY, UNIT 1. Analysis Of Core Damage Fre-Operations. Appendices A-D. CHU,T.L; MUSICKI,Z.; KOHUT,P.;

qmcy % %smic Ms Nnng M@p @ahsMaen et al. Brookhaven Nabonal Laboratory. Une 1994. 468pp.

i Report. BUDNITZ,R.J. Future Resources Associates, Inc.

9408150192. BNL-NUREG-52399. 80540:091.

i DAVIS,P.R. PRD Consulting. RAVINDRA M.K.; et al. EOE, Inc.

See NUREG/CR-6144,V02.P1 A' abstract.

August 1994.114pp. 9409230293. 81010:116.

Dunng M89, N Nuclear RegulaWy Canmisson (NRQ M l

NUREG/CR4144 V02P3A: EVALUA00N OF POTENTIAL E* *

• SEVERE ACCIDENTS DURING LOW POWER AND SHUT-risks dunng low power and shutdown opera'ons. The program DOWN OPERATIONS AT SURP,f, UNIT 1. Analysis Of Core includes two parallel protects being performed by Brookhaven j

DamaQe Frequency From Intunal Events Dunng M4Lwp Nabonal Laboratory (BNL) and Sandia Fatonal Laboratories i

Operatons. Appendices E (Soctions E.1 -E.8).

CHU T.L; (SNL). Two plants, Surry (pressurized water roactor) and Grand MUSICKI,Z.; KOHUT,P.; et al. Brookhaven Natonal Laborato y.

Gulf (boiling water reactor), were selected as the plants to be June 1994.

560pp.

9408150228.

BNL-NUREG-52399.

studied. The ob}ectives of the program are to assess the nsks 80555 001.

of severe accidents initiated dunng plant operational states See NUREG/CR-6144,V02,P1 A abstract.

other than full power operaton and to compare the estimated core damage frequencies, important accident sequences and NUREG/CR4144 V02P38: EVALUATION OF POTENTIAL other quahtative and quantitative results with those accidents l

Main Citations and Abstracts 35 initiated dunng full power operation as assessed in NUREG.

gions. The report also provides information characterizing the 1150. The oblective of this report es to document the approach methods and facilities used to treat and dispose of non-radioac-utihzed in the Surry plant and discuss the results obtained. A tive waste, including industrial, municipal, and hazardous waste parallel report for the Grand Gulf plant is prepared by SNL regulated under Subparts C and D of RCRA. The information in-NUREG/CR4145: VERIFICATION AND VALIDATION OF THE cludes a hst of disposal optons, the geograpNcal locatons of SAPHIRE VERSION 4.0 PRA SOFTWARE PACKAGE, such facilities, and a desenption of such processing and dispos-BOLANDER T.W.; CALLEY,M.B.; CAPPS,E.L; et al. Idaho Na-al fams.

tional Engineenng Laboratory. February 1994. 331pp.

NUREG/CR4147 V02: CHARACTERIZATION OF CUSS A 9404010225. EGG-2713. 78722:289.

LOW-LEVEL RADIOACTIVE WASTE 1986-1990. Main Report-A venfcation and vahdaten (V&V) process has been per-Part A. DEHMEL,J-C.; LOOMIS,D.; MAURO.J.; et al. S. Cohen &

formed for the System Analysis Programs for Hands-on Integrat-Associates, Inc. January 1994. 300pp. 9403140349. 78455:001.

ed Reliabihty Evaluation (SAPHIRE). SAPHIRE is a set of four See NUREG/CR-6147,V01 abstract.

computer programs that the Nuclear Regulatory Commission l

(NRC) developed to perform probabehste nsk assessments NUREG/CR4147 V03: CHARACTERIZATION OF CLASS A 1

(PRAs). These programs allow an analyst to create, quantify LOW-LEVEL RADIOACTIVE WASTE 1986-1990 Main Report.

and evaluate the risk associated with a facihty or process being Part B. DEHMEL,J-C.; LOOMIS.D.; MAURO J.; et al. S. Cohen &

analyzed. The programe included in this set are Integrated Reh-Associates, Inc. January 1994. 300pp. 9403140351. 78452:001.

abikty and Risk Anr:ysis System (IRRAS), System Analysis and See NUREG/CR-6147,V01 abstract.

Risk Assessmerd gSARA), Models and Results Database (MAR-NUREG/CR4147 V04: CHARACTERIZATION OF CLASS A D), and Faaft Tree / Event Tree / Piping and instrumentation Dia-LOW-LEVEL RADIOACTIVE WASTE 1986-1990. Appendices A-gram (FEP) graphcal editor. The V&V steps included a V&V E. DEHMEL,J-C.; LOOMIS.D.; MAURO,J.; et al. S. Cohen & As-plan to desenbe the process and entena by which the V&V sociates, Inc. January 1994, 300pp. 9403140353. 78475:001.

would be performed; a software requirements documentabon See NUREG/CR-6147,V01 abstract.

review to determine the correctness, completeness, and tracea-bihty of the requirements; a user survey to determine the useful.

NUREG/CR4147 V05: CHARACTERIZATION OF CLASS A ness of the user documentation, identificaton and testing of LOW-LEVEL RADIOACTIVE WASTE 1986-1990. Appendix F.

vital and non-vrtal features, and documentat>on of the test re.

DEHMEL,J-C.; LOOMIS.D.; MAURO,J.; et al. S. Cohen & Asso-suits.

ciates, Inc. January 1994. 600pp. 9403140355. 78473:012.

NUREG/CR4146: LOCAL CONTROL STATIONS: HUMAN ENGI-NEERING ISSUES AND INSIGHTS.

BROWN,W.S.;

NUREG/CR4147 V06: CHARACTERIZATION OF CMSS A HIGGINS.J.C.; O'HARA.J.M. Brookhaven National Laboratory.

LOW-LEVEL RADIOACTIVE WASTE 1986-1190. Appendices G-September 1994. 74pp. 9411080056. BNL-NUREG-52400.

J. DEHMEL.J-C.; LOOMIS,0.; MAURO.J.; et al S, Cohen & As.

81648:146.

sociates, Inc. January 1994. 500pp. 9403140360. 78471:114.

The ooiective of this research project was to evaluate current See NUREG/CR-6147,V01 abstract.

human engineenng at local control statens (LCSs) in nuclear NUREG/CR4147 V07: CHARACTERIZATION OF CLASS A power plants, and to identify good human engineenng practices relevant to the design of these operator interfaces. General ht.

LOW-LEVEL RADIOACTIVE WASTE 1986-1990. Appendices K-erature and reports of operatsng expenence were reviewed to P. DEHMEL,J-C.; LOOMIS,0.; MAURO.J.; et al. S. Cohen & As-determine the extent and type of human engineenng deficien-sociates, Inc. January 1994. 400pp. 9403140364. 78470:005.

caes at LCSs in nuclear power plants. In-plant assessments See NUREG/CR-6147,V01 abstract.

i were made of human engineenng at single-function as well as NUREG/CR4149: APPLICATIONS OF FIBER OPTICS IN PHYSI-multifuncton LCSs. Besides confirming the existence of human CAL PROTECTION. BUCKLE,T.H. Sandia National Laboratones.

engineenng deficiencies at LCSS, the in-plant assessments pro-March 1994. 51pp. 9404080104. SAND 93-2478. 78794:261, j

vided informaton about the human engineering upgrades that The purpose of this report is to provide techrncal informaton i

have been made at nuclear power plants. Upgrades were typi-useful for the development of fiber-optc communcatons and cally the result of any of three influences - regulatory activity, intrusion detecton subsystems relevant to physcal protection.

broad industry irutatives such as INPO, and specific in-plant There are major sectons on fiber-opte technology and applica-programs (e g., activities related to training). It is concluded that tions. Other topics include fiber-optc system components and l

the quality of LCSs is quite variable and might be improved if systems engineenng. This document also contains a glossary, a l

there were greater awareness of good practices and existing list of standards and specifcations, and a hst of fiber-opte ven-human engineenng guidance relevant to these operator inter-dors.

faces, which is available from a variety of sources. To make I

such human engineenng guidance more readdy accessible, NUREG/CR-6151: FEASIBILITY OF DEVELOPING RISK-BASED guideh,nes were compeled from such sources and included in the RANKINGS OF PRESSURE BOUNDARY SYSTEMS FOR IN-report as an appe@.

SERVICE INSPECTION. VO.T.V.; SMITH.B.W.; SIMONEN,F.A.;

et al. Batteile Memonal Institute, Pacific Northwest Laboratory.

NUREG/CR4147 V01: CHARACTERIZATION OF CLASS A August 1994. 86pp. 9410130086. PNL-8912. 81266:091.

LOW-LEVEL RADIOACTIVE WASTE 1986-1990. Executive Sum-Goals of the Nondestructive Evaluaton Reliability Program mary. DEHMEL.J-C.; LOOMIS.D.; MAURO J.; et at S. Cohen &

sponsored by %e NRC at PNL are to 1) assess current inspec-Associates, Inc. January 1994. 92pp. 9403140342. 78456.00).

ton technittees and requirements of all pressure boundary sys-This report describes the physical, chemical, and radological tems and swp06 ants, 2) deterrrune if improvements are proporties of Class A low-level radcactive waste using data ned, aria 't) if necessary, develop recommendations for re-contained in the Manifest Information Management System vising the apphc able ASME Codes and regulatory requirements.

(MIMS). Other sources of informaton include reports prepared in evaluating approaches that could be used to provide a tech-by the NRC, DOE, low-level waste Compacts and States, and nical basis for improved inservce inspecton (IS!) plans, PNL trade industnes. The database characterizes low-level waste has developed and apphed a method that uses results of prob-shipped for disposal from 1986 to 1990. A computer program abeliste risk assessment (PRA) to rank pressure boundary sys-was developed to analyze the data, with the results summanzed tems for ISI. In the PNL program, the feasibihty of developing in tables, histograms, and cumulative distributen curves pre-nsk-based genenc ISI requirements is being conducted in two senting radionuclide concentratson distributons in Class A waste phases. Phase i involves identifying and pnontmng the systems as a function of waste streams, waste generators, and by re-most relevant to plant safety. The results of these evaluatons I

1

36 Main Citations and Abstracts will be consolidated into requirements for comprehensive piping NUREG/CR-6156:

SUMMARY

OF COMMENTS RECEIVED system inspections that will be developed in Phase 11. This FROM WORKSHOPS ON RADIOLOGICAL CRITERIA FOR DE-report presents the Phase I results of evaluations for eight se-COMMISS!ON!NG. CAPLIN.J; PAGE.G.; SM!TH.D.; et al.' Ad-lected plant PRAs. Based on the results of this study,it appears vanced Systems Technology, Inc. January 1994. 200pp.

that there are generic insights that can be extrapolated from the 9402150300. 78122:101, selected plants to specific classes of light water reactors.

The Nuclear Regulatory Commission (NRC) is conducting an enhanced participatory rulemaking to establish radaologscal crite-NUREG/CR4152: EXPERIMENTS TO tiNESTIGATE DIRECT ria for site cleanup and decommissioning of NRC-licensed facili.

CONTAINMENT HEATING PHENOMENA WITH SCALED ties, Open public meetings were held dunng 1993 in Chicago, MODELS OF THE SURRY NUCLEAR POWER PLANT, IL, San Francisco, CA. Boston, MA, Dallas, TX, Philadelphia, BLANCHAT,T.K.; ALLEN,M.D.; PILCH,M.M.; et at Sandia Na-PA, Atlanta, GA, and Washington, DC. Interested parties were tional Laboratones. June 1994. 206pp. 9407260019. SAND 93-invited to provide input on the ru!emaking issues before the 2519. 80344:001.

NRC staff develops a draft proposed rule. This report summa-The Contairmt Technology Test Facility (CTTF) and the rtzes 3,635 comments categorized from transcripts of the seven Surtsey Test Facility at Sandia National Laboratories are used workshops and 1,677 commerits from 100 NRC docketed let.

~ to perform scaled experiments that simulate Hsgh Pressure Melt ters from individuals and organizations. No analysis or response Ejection accidents in a nuclear power plant (NPP). These ex-to the comments is included. The comments reflect a broad periments are designed to investigate the effects of direct cord spectrum of view-points on the issues related to radiological cri-tainment heating (DCH) phenomena on the containment load.

teria for site cleanup and decommissioning. The NRC also held High-temperature, chem #cally reactive melt (thermite) is ejected public meetings on the scope of the Generic Environmental by high-pressure steam into a scale rnodel of a reactor cavity.

Impact Statement (GEIS) during July 1993. The GEIS meetings Debris is entrained by the steam blowdown into a containment were held in Washington, DC., San Francisco, CA, Oklahoma model where specific phenomena, such as the effect of sub.

City, OK, and Cleveland, OH. Related comments from these compartment structures, prototypic air / steam / hydrogen atmos.

meetings were reviewed and comments which differed substan-pheres, and hydrogen generation and combustion, can be stud-tially from those from the workshops are also summarized in the ied. Four integral Effects Tests (iETs) have been performed body of the report. A summary of the comments from the GEIS with scale models of the Surry NPP to investigate DCH phe-scoping meetings is included as an Appendix.

nomena. The 1/6 scale integral Effects Tests (IET-9. IET 10, NUREG/CR4157: SURVEY AND EVALUATION OF AGING RISK and IET-11) were conducted in CTTF, which is a 1/6(th) scale ASSESSMENT METHODS AND APPLICATIONS, SANZO D.;

model of the Surry reactor containment building (RCB). The 1/

KVAM,P Los Alamos National Laboratory. APOSTOLAKIS.G.;

10(th) scale IET test (lET-12) was performed in the Surtsey et at Advanced Systems Concepts Associates. November vesset, which had been configured as a 1/10 scale Suny RCB.

1994.180pp. 9412130107. LA-12715-MS. 82007:186.

Scale rnodels were constructed in each of the facilities of the The US Nuclear Regulatory Commission irwtiated the nuclear Surry structures, including the reactor pressure vessel, reactor suppoit skirt, control rod dnve missde shield, biological shidd power plant aging research program about 6 years ago to wall, dtv, instrument tunnel, residual heat removal platform gather information about nuclear power plant aging. Since then, M OTgu has collected a sign ficant amount of information, and heat exchangers, seal table room and seal table, operating lav., qualitative, on plant aging and its potential effects on deck, and crane wall This report describes these experiments plant safety. However, this body of knowledge has not yet been and gives the results.

integrated into formalisms that can be used effectively and sys-NUREGICR4154 V01: EXPERIMENTAL RESULTS FROM CON-tematically to assess plant nsk resulting from aging, although TAINMENT PIP!NG BELLOWS SUBJECTED TO SEVERE AC-models for assessing the effect of increasing fadure rates on CIDENT CONDITIONS.Results From Bellows Tested in " Uke-core damage frequency have been proposed. This report sur-New" Conditions. LAMBERT,LD.; PARKS M.B. Sandia National veys the work on the aging of systems, structures, and compo-nents (SSCs) of nuclear power plants, as well as associated Laboratories. September 1994. 81pp. 9411000061. SAND 94-data bases. We take a critical look at the need to revise prob-1711. 81648:221.

abilistic risk assessments (PRAs) so that they will include the Bellows are an integral part of the containment pressure contribution to risk from plant aging, the adequacy of existing boundary in nuclear power plants. They are used at piping pen-me s W evabatsg Ws contrW and me adequacy of etrations to allow relative movement between piping and the e data mat have been used in mese, evaluation methods. We containment wall, wfule minimizing the load imposed on the W a Wnmay kamewM W integraMg h aW of piping and wall Piping bellows are prirnarily used in steel con-tainments; however, they have received limited use in some d

g concrete (reinforced and prestressed) containments in a severe accident they may be subjected to pressure and temperature NUREG/CR-6158: IMPLICATIONS FOR ACCIDENT MANAGE-conditions that exceed the design values, along with a combina-MENT OF ADDtNG WATER TO A DEGRADING REACTOR tion of axial and lateral deflections. A test program to determine CORE, KUAN P.; HANSON,0.J.; PAFFORD.D.J.; et at EG&G the leak tight capacrty of containment penetration bellows is Idaho, Inc. February 1994. 220pp. 9404010230. EGG-2644.

being conducted under the sponsorship of the U.S. Nuclear 78723.256.

Regulatory Cv,....G, at Sandia National Laboratories. Sever

• This report evaluates both the positive and negative cone al different bellows geometries, representative of actual contain-quences of adding water to a degraded reactor core during a ment bellows, have been subjected to extreme deflections severe accident. The evaluation discusses the earliest possible along with pressure and temperature loads. The bellows ge-stage at which an accident can be terminated and how plant ometries and loading conditions are desenbed along with the personnel can best respond to undesired results. Specifically testing apparatus and procedures. A total of thirteen bellows discussed are (a) the potential for plant personnel to add water have been tested, all in the 'like-new' condition. (Additional for a range of severe accidents, (b) the time available for plant tests are planned of bellows that have been subjected to corro-personnel to act (c) possible plant responses to water added sion.) The tests showed that bellows are capable of withstand-during the various stages of core degradation, (d) plant instru-ing relatively large deformations, up to, or near, the point of full mentation available to understand the core condition and (e) the compression or elongation, before developing leakage. The test expected response of the instrumentation during the various data is presented and discussed stages of severe accidents.

Main Citations and Abstracts 37 NUREG/CR-6160:

SUMMARY

OF IMPORTANT RESULTS AND crack tip opening displacement at initiation are considered in SCDAP/RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT plane strain, finite element computatens. The finite element re-l LP-FP-2. CORYELL.E.W.; AKERS,D.W.; ALLISON.C.M.; et al.

suits demonstrate a sgnifcant elevation in crack-tip constraint EG&G Idaho, Inc. April 1994.165pp. 9405310168. NEA-CSN1-due to macroscopc " sharpening" of the extending tip relative to R(94)3. 79591:001.

the blunt tip at the irutiaton of growth. However this effect is This report summarizes sigrufcant techncal findings from the offset partially by the additional plastic deformation associated LP FP-2 Expenment sponsored by the Organization of Econom-with the increased applied J required to grow the crack. The ini-ic Cooperaton and Development (OECD). It was the second, tial a/W ratio, tearing modulus, strain hardening exponent and and final, fisson product expenment conducted in the Loss of-specimen size interact in a complex manner to define the evolv-Fluid Test (LOFT) facility at the Idaho National Engineenng Lab-ing near-tip conditions for cleavage fracture. The paper explores oratory. The overall technical objective of the test was to con-development of the new model, provides necessary graphs and f

tribute to the understanding of fuel rod behavior, hydrogen gen-procedures for its applicaton and demonstrates the effects of I

eraton, and fisson product release, transport, and deposition the model on fracture data sets for two pressure vessel steels ounng a V sequence accident scenano that resulted in severe (A533B and A515).

core damage. An 11 by 11 test bundle, compnsed of 100 prepressunzed fuel rods,11 control rods, and 10 instrumented NUREG/CR-6164: RELEASE OF RADIONUCLIDES AND CHE-guide tubes, was surrounded by an insulating shroud and con-LATING AGENTS FROM CEMENT SOLIDIFIED DECONTAMI-tained in a specially designed central fuel module, that was in-NATION LOW-LEVEL RADIOACTIVE WASTE COLLECTED sorted into the LOFT reactor. The simulated transient was a V-FROM THE PEACH BOTTOM ATOMIC POWER STATION sequence loss-of-coolant accident scenario featunng a pipo UNIT 3. AKERS.D.W.; KRAFT,N.C.; MANDLER.J W. EG&G break in the low pressure injection system line attached to the Idaho, Inc. March 1994. 71pp. 9405050409. EGG-2722.

hot leg of the LOFT broken loop piping. The transient was ter-79165:130.

minated by reflood of the reactor vessel when the outer wall As part of a study being performed for the Nuclear Regulatory shroud temperature reached 1517 K. With sustained fisson Commission (NRC), small-scale waste-form specimens were j

power and heat from oxidaten and metal-water reactons, ele-collected dunng a low oxidation-state transition-metal ion l

vated temperatures resulted in zircaloy melting, fuel liquefacton, (LOMI)-nitnc permanganate (NP)-LOMI solidification performed material relocaton, and the release of hydrogen, aerosols, and in October 1989 at the Peach Bottom Atome Power Station fission products. A desenption and evaluation of the mator phe-Unit 3. The purpose of this program was to evaluate the per-l nomena, based upon the response of on-line instrumentation, formance of cement solidified decontamination waste to meet

)

analysis of fission product data, postarradiaton examinaton of the low-level waste stability requirements defined in the NRC's the fuel bundle, and calculations using the SCDAP/RELAPS

" Technical Position on Waste Forrn," Revision 1. The samples computer code, are presented.

were acquired and tested because little data have been ob-NUREG/CR-6161: BUCKLING EVALUATION OF SYSTEM tained on the physcal stability of actual cement solidified de-80 + (TM) CONTAINMENT. GREIMANN,L; FANOUS,F.;

contamination ion-exchange resin waste forms and on the lea-SAFAR.S.; et al. Iowa State Unw., Ames, IA. August 1994.

chability of radionuchdes and chelating agents from those waste 145pp. 9411080051. IS-5103. 81648 001.

forms. The Peach Bottom waste-form specimens were subject-The System 80+ (TM) containment may be subjected to com.

ed to compressive strength, immersion, and leach testing in ac-pressive forces whch could cause it to become unstable. The cordance with the NRC's "Techncal Postten on Waste Form,"

stability of the containment shell under presenbed loading com.

Revision 1. Results of this study indcete that the specimens binations was investigated with two analysis levels: axisymme.

withstood the compresson tests (>500 psi) before and after inc and three dimensional. An axisymmetnc shell model, includ-immerson testing and leaching, and that the leachability index-ing additional mass to account for penetratons and the spray es for all radonuchdes, including (14)C, (99)Tc, and (129)l, are heador system, was analyzed using BOSOR4 and BOSOR5 well above the leechebety index requirement of 6.0, required by l

finite difference codes. Loading combinatens with pressure, the NRC's "Techrwcal Positen on Waste Form," Revision 1.

temperature, self weght, and seisme satisfied the Amercan So-ciety of Mechancal Engineers (ASME) stress allowables. The NUREG/CR-6166: RISK IMPACT OF TECHNICAL SPECIFICA-buckling assessment was performed using the worst mendian TIONS REQUIREMENTS DURING SHUTDOWN FOR BWRS.

I assumpton, including matenal nonlineanties and a sinusoidal STAPLE,B.D.;

KIRK,H.K.

Sandia National Laboratories.

axisymmetnc imperfecton. The minimum factor of safety for YAKLE,J. Science Appleations intematonal Corp. (formerly Sci-l Servce Level C was 2.35. A SSE seismic margin of 2.91 was ence Apphcatons, inc1 October 1994. 160pp. 9411040066.

calculated. The ABAOUS finite element code was selected for 81646:030.

the three dimensional analysis and tested with classical and This report presents an apphcahon of probabikstic models BOSOR solutons. The three dimensenal model included the and nsk based cntena for determining the risk impact of the equipment hatch, two personnel airlocks, and additional mass Limiting Conditons of Operations (LCOs) in the Technical Spec-l for the spray system and small penetrations. Matenal nonhnear.

'f'catons (TSs) of a boiling water reactor dunng shutdown. This ity and an axisymmetnc sinusoidal imperfecten were incorporat.

analysis studied the risk impact of the current requirements of ed. A minimum factor of safety of 1.91 was predicted, which Anowed Outage Times (AOTs) and Sminance Test intowals does not satisfy ASME Secton NE3222.1 or Regulatory Guide (STis) in eight Plant Operational States (POSS) which encom.

1.57. Code Case N-284 is satisfied. The analysis is conservative pass power operatons, shutdown, and refueling. This report pnmanly because the SRSS 10% method provides a conserva, also discusses insights concerning TS acton statements.

tue estimate of modal coupling.

NUREG/CR-6168: DIRECT CONTAINMENT HEATING INTEGRAL NUREG/CR-6162: EFFECTS OF PRIOR DUCTILE TEARING ON EFFECTS TESTS AT 1/40 SCALE IN ZION NUCLEAR POWER CLEAVAGE FRACTURE TOUGHNESS IN THE TRANSITION PLANT GEOMETR'.

BINDER,J.L; MCUMBER,L.M.;

l REGION. DODDS.R.H.; TANG,M. Illinois, Unw. of, Urbana, IL.

SPENCER,B.W. Argonie Natonal Laboratory. September 1994.

ANDERSON,T.L. Texas A&M Unw., College Staten, TX. June 112pp. 9410120216. Al!L 94/18. 81254:150.

1994.45pp.9407250325. UILU-ENG93-2014. 80329-232.

The results of Direct Containment Heating (DCH) integral ex-Prevous work by the authors desenbed a micromechants penments are presented. The expenments simulated a high fracture model to correct measured J(c)-values for the mecha.

pressure melt ejecten in the Zion Nuclear Power Plant. Expen-l neste effects of large-scale yielding. This new work extends the ments were conducted in a 1/40 scale model of the Zon con-model to also include the influence of ductile crack extenson tainment. The model included the vessel lower head, cavity and pror to Cleavage. Ductile crack extensons of 10-15 X the initial instrument tunnel, and the lower containment structures. The 1

l l

1 l

38 Main Citations and Abstracts melt ejectons were driven by steam. There were two main ob-See NUREG/CR.6174,V01,DFC abstract.

jectwes of these experiments. The first was to invesbgate the effect of scale on DCH phenomena. The IET test senes ad.

NUREG/CR4176: REVIEW OF ENVIRONMENTAL EFFECTS ON dressed this by conducting counterpart integral tests in a 1/40 FATIGUE CRACK GROWTH OF AUSTENITIC STAINLESS scale facihty at Argonne Natx:inal Laboratory and in a 1/10 STEELS. SHACK,W.J.; KASSNER,T.F. Argonne National Labo-scale facihty at Sandia National Laboratories. Iron / alumina ther-ratory, May 1994. 38pp. 9406200315. ANL-94/1, 79830:287, m te with chromium was used as a core melt simulant in the IET Fabgue and environmentally assisted cracking of piping, pres-test series. The second objective was to address potential ex-sure vessel cladding, and core components in light water reac-penment distortions introduced by the use of non-prototypic tors are potential concerns to the nuclear industry and regula-iron / alumna thermete. The second objective was met in the U tory agencies. The degradation processes include intergranular series of tests which utilized a prototypic core melt. Corium ex*

stress corrosion cracking of austenstic stainless steel (SS) periments were conducted that were counterpart to the IET-piping in boiling water reactors (BWRs), and propagaton of fa-1RR and IET-6 iron / alumina tests-tigue or stress corrosion cracks (which inibate in sensitized SS NUREG/CR4169: RELAY TEST PROGRAM.Senes 11 cladding) into low-alloy femtic steels in BWR pressure vessels.

Tests. Integral Testing Of Relays And Circuit Breakers.

Crack growth data for wrought and cast austenitic SSs in samu-BANDYOPADHYAY,K; KUNKEL,C.; SHTEYNGART,S. Brookha.

lated BWR water, developed at Argonne National Laboratory ven National Laboratory. February 1994.176pp. 9403140191.

under U.S. Nuclear Regulatory Commission sponsorship over BNL-NUREG 52406. 78476:001.

the past 10 years, have been compiled into a data base along

(

TNs report presents the results of a relay test program con-with similar data obtained from the open literature. The data I

ducted by Brookhaven Nabonal Laboratory (BNL) under the were analyzed to develop corrosion-fatigue curves for austenluc sponsorship of the U.S. Nuclear Regulatory Commission (NRC).

SSs in aqueous environments corresponeng to normal BWR I

The program is a conhnuaton of an earher test program the re-water chemistries, for BWRs that add hydrogen to the feed-I suits of which were published in NUREG/CR-4867. The current water, and for pressurized water reactor (PWR) primary-system-l program was camed out in two phases: electrical testing and vi-coolant chemistry. The corrosion-fatigue data and curves in bruton teshng. The objective was pnmarily to focus on the elec-water were compared with the air hne in Secton XI of the tncal discontinuity or condnuity of relays sublect to electncal ASME Code, pulses and vibration loads. The electncal testing was conducted by KEMA-Powertest Company and the vibraton testing was per-NUREG/CR-6177: ASSESSMENT OF THERMAL EMBRITTLE-formed at Wyle Laboratones, Huntsville, Alabama. This report MENT OF CAST STAINLESS STEELS. CHOPRA,0.K.;

discusses the test procedures, presents the test data, includes SHACK,W.J. Argonne Natonal Laboratory. May 1994. 35pp.

an analysis of the data and provides recommendations regard-9406200312. ANL 94/2. 79830:326.

ing reliable relay teshng.

A procedure and correlatons are presented for assessing NUREG/CR4174 V1 DFC: REVISED ANALYSES OF DECOM.

thermal embrittlement and predicting Charpy-impact energy and MISSIONING FOR THE REFERENCE BOILING WATER REAC.

fracture toughness J-R curve of cast stainless steel compo-TOR POWER STATION. Effects Of Current Regulatory And nents under light water reactor operahng conditons from known Other Consderabons On The Financial Assurance Require.

material informaton. The "saturaton" impact strength and frac-ments Of The Decommissorung Rule And....

SMITH,R1; ture toughness of a specife cast stainless steel, i.e., the mini-BIERSCHBACH.M.; KONZEK,G.J.; et al. Battelle Memonal insti.

mum value that would be achieved for the material after long-tute, Paofic Northwest Laboratory. September 1994. 150pp.

term service, is estimated from the chemical compositen of the 9411080088. 81664:108.

steel. Fracture properties as a functon of time and temperature With the issuance of the Decommissioning Rule (June 27, of reactor service are eshmated from the kinetics of embrittle-i 1988), nuclear power plant hcensees are required to submit to ment which are also determined from chemical composition. A the U.S. Nuclear Regulatory Commsson (NRC) for review, de-common " predicted lower bound" J-R curve for cast stainless commissoning plans and cost estimates. This reevaluation steels of unknown chemical composition is also defined for a study provides some of the needed bases documentaten to the given grade of steel, femte content, and temperature. Examples NRC staff that will assist them in assessing the adequacy of the of estimahng fracture toughness of cast stainless steel compo-hcensee subattals. This report presents the results of a review nents dunng reactor service are presented.

and reevaluation of the PNL 1980 decommissioning study of the WNP-2 nuclear power plant for the DECON, SAFSTOR, and NUREG/CR 6178: LABORATORY CHARACTERIZATION OF ENTOMB decommissioning alternat ves. These attematives now ROCK JOINTS. HSIUNG S.M.; KANA,D.D.; AHOLA,M.P.; et al.

include an arvtial S-7 year period during which the spent fuel is Center for Nuclear Waste Regulatory Analyses. May 1994.

stored in the spent fuel pool, pnor to beginning major disassem-240pp. 9407260012. CNWRA 93-013. 80342:001, bly or extended safe storage of the plant. This report also ire A laboratory characterizaton of the Apache Leap tuff joints ciudes conssderaton of the NRC requirement that decomms-under cyche pseudostatic and dynamic loads has been under-sioning activrbes leadng to terminaton of the nuclear hcense be taken to obtain a better understanding of dynamic joint shear completed within 60 years of final reactor shutdown, consider-behavior and to generate a complete data set that can be used aton of packaging and disposal requirements for Greater-Thark for validation of existing rock-joint models. Study has indicated Class C low-level waste, and reflects all costs in 1993 dollars.

that available methods for determining joint roughness coeffi-Sensstivity of the total hcense terminaten cost to the disposal cient (JRC) significantly underestimate the roughness coefficient costs at different low-level radioactrve waste disposal sites, and of the Apache Leap tuff joints, that will lead to an undereshma-to different depths of contaminated concrete surface removal ton of the joint shear strength. The results of the direct shear with the facdsties are also examned-tests have indicated that both under cyclic pseudostatic and dy-NUREG/CR4174 V2 DFC: REVISED ANALYSES OF DECOM.

namic loadings the joint resistance upon reverse shearMg is MISSIONING FOR THE REFERENCE BOILING WATER REAC-smaller than that of forward sheanng and the joint ddation re.

TOR POWER STATION Effects Of Current Regulatory And sutting from forward shearing recovers during reverse sheanng.

Other Considerabons On The Financial Assurance Require-Within the range of variaton of shearing velocity used in these ments Of The Decommasoning Rule And....

SMITH,R1; tests, the shearing velocity effect on rock-jotnt behavior seems BIERSCHBACH.M.; KONZEK,G.J.; et al. Battelle Memorial Inste to be minor, and no nobceable effect on the peak joint shear tute, Pacific Northwest Laboratory. September 1994. 300pp.

strength and the joint shear strength for the reverse sheanng is 9411080106. 81664:258.

observed.

Main Citations and Abstracts 39 NUREG/CR4180: HYDROGEN MIXING STUDIES (HMF): USER'S input file and perform data checking. This capability increases MANUAL LAM.K.L; WILSON,TL Los Alamos Natr.,nal Labora-productivity and decreases the chance of user error, tory. TRAVIS.J R. Science Apphcations internatioral Corp. (for-merly Science Applications, Inc.). DecrEbe 1994. 130pp.

NUREG/CR 8182 V02: OFFSCALE: A PC INPUT PROCESSOR 9412300171. LA-12741 M. 82159:052.

FOR THE SCALE CODE SYSTEM. Volume 2: The ORIGNATE Hydrogen Mixing Studies (HMS) is a best-estimate analysis Processor for ORIGEN-S. BOWMAN,S.M. Oak Ridge National tool for predicting the transport, mixing, and combustion of hy-Laboratory. November 1994. 27pp. 9412070073. ORNL/TM.

drogen and other gases in nuclear reactor containments and 12663. 81956:171, other facahbes. It can model geometrically complex facilities OFFSCALE is a suite of personal computer input processor having multiple compartments and internal structures. The code can simulate the effects of steam condensation, heat transfer to programs developed at Oak Ridge Nabonal Laboratory to pro-walls and internal structures, chemcal kinetics, and fluid turb vide an easy-to-use interface for modules in the SCALE-4 code lence. The gas mixture may consist of components included in system. ORIGNATE is a program in the OFFSCALE suite that a buelt-in hbrary of 20 species. TNs manual describes how to serves as a user friendly interface for the ORIGEN-S isotopic use the code. It explains how to set up the model geometry.

generation and depletion code. It is designed to assist an define walls and obstacles, and specify gas species and maten-ORIGEN-S user in prepanng an input file for execution of light-al properties. Definitions of initial and boundary conditions are water reactor (LWR) fuel depletion and decay cases. ORIG-NATE generates an input file that may be used to execute also described. The manual also describes vanous physical ORIGEN-S in SCALE-4. ORIGNATE features a pulldown menu model and numercal procedure options, as well as how to turn them on. The reader also learns how to specify different out-system that accesses sophistcated data entry screens. The puts, especially graphical display of solution variables. Finally program allows the user to quickly set up an ORIGEN-S input sample problems are included to illustrate some applications of file and perform error checking. This capabihty increases pro.

the code. An input deck that illustrates the minimum required ductivity and decreases the chance of user error, data to run HMS is given at the end of this manual.

NUREG/CR4183: PEER REVIEW OF THE TMI-2 VESSEL IN-NUREG/CR4181: A PILOT APPLICATION OF RISK-BASED VESTIGATION PROJECT METALLURGICAL EXAMINATIONS.

METHODS TO ESTABLISH INSERVICE INSPECTION PRIOA*

BOHL.R.W.; GAYDOS.R.G.; VANDER VOORT,G.; et al. Ar-ITIES FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NU-gonne National Laboratory. July 1994. 45pp. 9408220048. ANL-CLEAR POWER STATION.

VO,T.V.;

GORE B.F.;

94/3. 80637:206 SIMONEN F.A.; et al. Battelle Memorial Institute, Pacife North.

Fifwn @ h hom h lower Med W h Three s

atory. August 1994. 72pp. 9409090115 PNL-9020.

Mde letand (TMI) Unis 2 nucesor reactor pressure vessel were 2 M aded W m by h idaho As part of the NDE Reliabihty Program sponsored by the Notonal Engmeonng Laboratory. Argonne National Laboratory NRC, PNL is developing a method that uses risk-based ap-(ANL) and several of the European particspents. These exami-proaches to estabhsh insennce inspecten plans for nuclear nations delennmed that a p of tie lower head reached power plant components. This method uses probabiliste risk as.

temperatures as high as 1100 degrees C Mng the accident sessnnnt (PRA) results and fadure Modes and Effects Analysis and cooled from these largeratures at =10-100 degrees C/

techr'iques to identify and pnontize the most risk-important sys.

min. The rememder of the lower head was W to have re-tems and components for inspecton. The Surry Nuciear Power momed below 727 degrees C. A panel of three outesde peer re-Station Unit 1 was selected for pdot apphcations of this method.

Viewers was tormed to conduct an E A review of this The specific systems addressed in this report are the reactor metallurgeal analyses. Thee review determmed that the conclu-pressure vessel, reactor coolant, low-pressure injection, and eions resultmg from the INEL study were fundamentaNy correct.

auxiliary feedwater. The results provide a risk based ranking of in particular, the panel recffirmed that four lower heed semples components within these systems and relate the target nsk to attamed temperatures as high as 1100 degrees C, and perhaps target failure probabihty values for individual components. These as high as 1150-1200 degrees C in one case, during me acci-results will be used to guide the development of improved in-specton plans for nuclear power plants. To develop mspection dent. They concluded that these samples subsequere/ cooled plans, the acceptable level of risk from structural failure for im-at a rate of =50125 degrees C/ min in the temperate range of 600-400 degrees C. The reviewere also agreed that the re-portant systems and w,@,+its wdl be apportened as a small fraction of the total PRA. estimated nsk for core damage. This maander of the lower heed samples had not exceeded 727 de-process wdl determine acceptable target risk and target failure grees C during the acoderd and suggested several refinemere probabdity values for indrvidual components. Inspection require-and alternative procedures that could have been employed in be se t le s to assure that acceptable failur N Winal analysis.

NUREG/CR4185: TMI-2 INSTRUMENT NOZZLE EXAMINA.

TlONS AT ARGONNE NATIONAL LABORATORY. February NUREG/CR4102 V01: OFFSCALE: A PC INPUT PROCESSOR 1991 - June 1993. NEIMARK,LA.; SHEARER.T.L; PUROHIT,A.;

FOR THE SCALE CODE SYSTEM. Volume 1: The CSASIN et al. Argonne National Laboratory. March 1994. 158pp.

Processor For The Cntcahty Sequences. BOWMAN.S.M. Oak 9405050325. ANL-94/5. 79162:155.

Rdge National Laboratory. November 1994.21pp.9412070069.

ORNL/TM 12663. 81956:148.

Six of the 14 instrument-penetraton-tube nozzles removed OFFSCALE is a suite of personal computer input processor from the lower head of TMI-2 were examined to identify damage programs developed at Oak Ridge National Laboratory to pro-mechanisms, provide insight to the fuel relocation scenano, and vide an easy-to-use interface for rnodules in the SCALE-4 code provide input data to the margin-to-failure anatysis. Visual in-spection, gamma scanning, metallography, merohardness system. CSASIN (formerly known as OFFSCALE)is a program measurements, and scanning electron microscopy were used to m the OFFSCALE suite that serves as a user fnendly interface obtain the desired information. The results showed varying de-for the Cnticahty Safety Analysis Sequences (CSAS) available in grees of demage to the lower head nozzles, from *SO% molt-SCALE-4. It is designed to assist a SCALE-4 user in preparing off to no damage at all to near-neighbor nozzles. The elevations an input fde for execution of cnticahty safety problems. Output of nozzle damage suggested that the lower elevatons (near the from CSASW vrates an input fue that may be used to exe-cute the CSAL wouol module in SCALE-4. CSASIN features a lower head) were protected from molten fuel, apparently by an pulldown rnenu system that accesses Sophisticated data entry insulating layer of fuel debris. The pattern of nozzle damage was consistent with fuel movement toward the hot-spot locaton screens. The program allows the user to quickly set up a CSAS identified in the vessel wall. Evidence was found for the exist.

40 Malf) Citations and Abstracts ence of a significant quanbty of control assembly debns on the addressed are 12,18, 24, and 30 inch-thick reinforced con-lower head before the massive relocabon of fuel occurred.

crete slabs with reinforcing ratios of 0.2, 0.4, 0.6, 0.8.1.0 per.

cmL NUREG/CR4187: RESULTS OF MECHANICAL TESTS AND SUPPLEMENTARY MICROSTRUCTURAL EXAMINATIONS OF NUREG/CR4190 V01 R1: PROTECTION AGAINST MALEVO-THE TMI.2 LOWER HEAD SAMPLES. DIERCKS,D.R.;

LENT USE OF VEHICLES AT NUCLEAR POWER f 4EIMARK L.A. Argonne Natonal Laboratory. March 1994.

PLANTS.Vehiclo Barrier System Selection Guidance For Blast 143pp. 9405040214. ANL 94/8. 79104:205.

Protection. NEBUDA,D.T. Army, Dept. of, Corps of Engineers.

Metallographic examinations of 15 samples from the lower December 1994. 20pp. 9501180290. 82341:256.

head of me TMI-2 pressure vessel confirmed that four semples This manual provides for detemuning the minimum safe attained tempermanes as high as 1100 degrees C during the ac-standoff distance between vital safety related equipment and i

cident and cooled at =10-100 degrees C/ min. Portions of two the design basis vehicle bomb threat adopted by the U.S. Nu-adeoent samp6es, and poseably a tNrd sample away from the clear Regulatory Commission. Vital safety related equipment l

l I

hot spot, also exceeded 727 degrees C. Results from tonesle should survive the design basis veNele bomb attack when the tests conducted on tNs material at 600-1200 degrees C gener-minimum safe standoff distance is provided. Guidance is provid-j ally agreed well with literature data on A533, Grade B steel. The ed for exposed vital safety related equipment and for equipment l

material from the hot spot exfubited higher strengths then the housed witNn vital area bamers.

remaining material, reflecting the heat treatment received dunng l

the accident. Charpy V-notch impact tests similerty found signef6-NUREG/CR4190 V02: PROTECTION AGAINST MALEVOLENT cantly lower upper sheet energies and Ngher transition tempera-USE OF VEHICLES AT NUCLEAR POWER PLANTS. Vehicle tures for the matcrial from the hot spot. However, creep tests Barrier System Siting Guidance For Blast Protection.

conducted at

• 600-1200 degrees C revealed little difference NEBUDA.D.T. Army, Dept. of, Corps of Engineers. August 1994.

between metonal at and away from the hot spot. Cracks were 41pp. 9409200303. 80954:236.

found in the stainless steel cladding of boet samples from the See NUREG/CR-6190,V01 abstract hot spot. The cracks appeared to be the result of hot-teanng, NUREG/CR4190 V02 R1: PROTECTION AGAINST MALEVO-probably seeisted by intergranular penetration of liquid Ag-Cd.

LENT USE OF VEHICLES AT NUCLEAR POWER Crack propagation into the A533 vessel steel was a maximurn PLANTS. Vehicle Barrier System Selection Guidance.

of a 6 mm. Metenate in the cracks suggest the presence of NEBUDA.D.T, Army, Dept. of, Corps of Engineers. December control-assembly debne on the lower head before the massive 1994. 41pp. 9501180287, 82341:215.

fuse flow amved.

This manual provides a simplified procedure for selecting land l

NUREG/CR4148 V01: MICROBIAL DEGRADATION OF LOW-vehicle barriers that will stop the design basis vehicle threat LEVEL RADIOACTIVE WASTE. Annual Report For FY 1993-adopted by the U. S. Nuclear Regulatory Commission. Proper l

ROGERS,R.D.: HAMILTON,M.A.; VEEH,R.H.; et al. EG&G selection and construction of vehicle bamers should prevent in-l Idaho, Inc. Apnl 1994. 63pp. 9405040133. EGG 2730.

trusion of the design basis veNele. In addition, vital safety relat-l 79103:259 ed equipment should survive a design basis vehicle bomb l

The Nuclear Regulatory Commission stipulates that disposed attack when vehicle barriers are property selected, sited, and low-level radioactive waste (LLW) be stabilized. Because of aP-constructed. This manual addresses passive vehicle barriers, parent ease of use and normal structural integnty, cement has actrve vehicle bamers, and site design features that can be been widely used as a binder to solidify LLW. However, the re-used to reduce vehicle impact velocity, suiting waste forms are somebmes suscepbble to failure due to the actions of waste conshtuents, stress, and environment. This NUREG/CR4193: PRIMARY SYSTEM FISSION PRODUCT RE-report rewews laboratory efforts that are being developed to ad.

LEASE AND TRANSPORT.A State-Of-The-Art Report To The dress the effects of HN.wbvv0gically influenced chemical attack Committee On The Safety Of Nuclear installations.

on cement-solidified LLW. Groups of microorganisms are being WRIGHT.A L Oak Ridge National Laboratory. June 1994.

employed that are capable of metabolically converting organic 242pp. 9407010317. NEA/CSNt/R(94)2. 80031:001.

and inorganic substrates into organic and mineral acids. Such This report presents a summary of the status of research ac-acids aggressively react with cement and can ultimately lead to tivities associated with fission product behavior (release and structural failure. Results on the application of mechanisms irp transport) under severe accident conditions within the pnmary herent in macrobially influenced degradaten of cement-based systems of water-moderated and water cookd nuclear reactors.

l matenal are the focus of tNs report. Sufficient data-validated For each of the areas of fission product release and fission l

evidence of the potential for macrobially influenced deteriorabon product transport, the report summarizes relevant informat on of cement-solidified LLW has been developed during the course on important phenomena, major expenments performed, rele-of tNs study. These data support the continued development of vant computer models and codes, comparisons of computer appropnate tests necessary to determine the resistance of code calculabons with expenmental results, and general conclu-cement solidified LLW to microbially induced degradation that sions on the overall state of the art. Finally, the report provides could impact the stability of the waste form. They also justify an assessment of the overall importance and knowledge of pri-the conbnued effort of enumeration of the conditions necessary mary system release and transport phenomena and presents to support the nm.wvvvgical growth and population expansion.

major conclusions on the state of the art.

NUREG/CR4190 V01: PROTECTION AGAINST MALEVOLENT NUREG/CR4194: METALLOGRAPHIC AND HARDNESS EXAMI.

USE OF VEHICLES AT NUCLEAR POWER PLANTS.VeNele NATIONS OF TMI-2 LOWER PRESSURE VESSEL HEAD SAM-Bamer System Siting Guidance For Blast Protecton.

PLES. KORTH.G.E. EG&G Idaho, Inc. March 1994. 124pp.

NEBUDA,0.T. Army, Dept. of Corps of Engineers. August 1994.

9404110355. TMI V(92)EG01. 78817:224.

18pp. 9409200299. 80954:218.

Fifteen steel samples were removed from the lower pressure TNs manual provides for determining the minimum safe vessel head of the damaged TMI.2 nuclear reactor to assess standoff distance between vital safety related equipment and the thermal threat to the head posed by 15 to 20 metnc tons of the design basis vehicle bomb threat adopted by the U.S. Nu-molten core debns relocabng there during the accident. Full clear Regulatory Commission. Vital safety related e;/.prrent sections of thirteen of the samples and partial sections of the should survive the design basis veNele txeb attack when tre other two samples underwent hardness and metallographic ex-minimum safe standoff distance is provided. Gddance is provb-aminat ons at the Idaho National Engineenng Laboratory. These ed for exposed vital safety related equipment ano ior equip 9ent exarranations have shown that eleven of the fifteen samples did housed within vital area bamers. The types of vital area camers not exceed the femte.austenste transformation temperature of

i I

l Main Citations and Abstracts 41 l

727 degrees C dunng the accident. The remaining four samples suits as they relate to the damage to the reactor vessel and to did show evidence of having a much more severe thermal histo-the development of a core relocation scenario. Not all examina-ry. The samples from core grid positions F 10 and G-8 are be-tions originally proposed as part of this program were complet-heved to have experienced temperatures of 1,040 to 1,060 de-ed due to facihty problems at the INEL Consequently, only the grees C for about 30 minutes. Samples from positions E-8 and results of completed aspects of the examination program are E 6 appear to have been subjected to 1,075 to 1,100 degrees C presented.

for approximately 30 minutes.

NUREG/CR4195: EXAMINATION OF RELOCATED FUEL NUREG/CR4200: UNCERTAINTY ANALYSIS OF SUPPRESSION DEBRIS ADJACENT TO THE LOWER HEAD OF THE TMI-2 POOL HEATING DURING AN ATWS IN A BWR 5 PLANT.An REACTOR VESSEL.

AKERS.D.W.;

JENSEN.S.M.;

Application Of The CSAU Methodology Using The BNL Engi-SCHUETZ B.K.

EG&G Idaho, Inc. March 1994. 143pp.

neering Plant Analyzer. WULFF,W.; CHENG,H.S.; MALLEN A.N.;

9405050363. TMI V(92)EG10. 79193:304, et al. Brookhaven National Laboratory. March 1994. 62pp As part of the Three Mila Island Unit 2 (TMI-2) Vessel Investi-9405040189. BNL NUREG 52412. 79104:144.

gabon Project, funded by the Organizabon for Economic Coop-The uncertainty has been estirnated of predicting the peak eration and Development, physical, metallurgical, and radioche-temperature in the suppression pool of a BWR power plant, mical examinations were performed on samples of previously which undergoes an NRC postulated Anticipated Transient monen matenal that had relocated to the lower plenum of the Without Scram (ATWS). The ATWS is initiated by recirculabon TMI-2 reactor dunng the accident of 28 March 1979. This report pump trips and then leads to power and flow oscillations as presents the results of those examinabons and some limited they had occurred at the LaSalle-2 Power Stabon in March of analysis of these results as required for the interpretaten of the data. Pnncipal conclusions of the examinations are that the bulk 1988. After limit cycle oscillations have been established, the W

Mm MSIV l ste lower head debns is homogeneous and composed pnmanly of (U,Zr)O(2). This molten material reached temperatures greater discharge through the turbine bypass into the condenser Postu-than 2,600 degrees C and probably reached the lower head as lated operator actions, namely to lower the reactor vessel pres-a liquid or slurry at temperatures below the peak temperature. A sure and the level elevabon in the downcomer, are simulated by debns bed was formed, which was composed of particulate a robot model, which accounts for operator uncertainty. All bal-debris above a monohthic melt that sohdified on the lower head, ance of plant and control systems modeling uncertainbes were NUREG/CR4196: CALCULATIONS TO ESTIMATE THE MARGIN part of the statistical uncertainty analysis, that was patterned TO FAILURE IN THE TMI-2 VESSEL. STICKLER L.A -

ah h Code Mng, Ap@W aM Uncnnh pm REMPE.J.L; CHAVEZ,S.A.; et al. EG&G Idaho, Inc. Marc {

eduaton rneMogy N anaWs he M N WW 1994. 300pp. 9404060281. TMI V(93)EG01, 78759:001, suppression pool peak temperature of 329.3K (133 degrees F)

As part of the OECD-sponsored Three Mile Island Unit 2 has a 95-percenble uncertainty of 14.4K (26 degrees F), and (TMI-2) Vessel Investigation Project (VIP), margin-to fa.: 're cal.

that the size of this uncertainty bracket is dominated by the ex-culatons for machanisms having the potential to tt" eaten the perimental uncertainty of measunng Safety and Rehef Valve integnty of the vessel were performed to improve understanding mass flow rates under enbcal flow conditions. The analysis of events that occurred dunng the TMi-2 accident. Analyses showed also, that the probability of exceeding the suppression considered four failure mechanisms: tube rupture, tube ejection, pool temperature timet of 352.6K (175 degrees F) is most likely global vessel fadure, and localized vessel failure. Calculational zero (it is estimated as < 510(-4)). The equere fora of the i

input was based on data from the TMI 2 VIP examinabons of sum of the squares of stil the computed peak pool temperatures the vessel steel samples, the instrument tube nozzles, and sam-is 350.7K (171.6 degrees F).

)

pies of the hard layer of debns found on the TMI-2 vessel lower head. Sensitnnty studies were performed to invesbgate the un-NUREG/CR4201: COMPRESSION AND IMMERSION TESTS certainties in key parameters for these analyses.

AND LEACHING OF RADIONUCLIDES STABLE METALS, AND NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT IN-CHELATING AGENTS FROM CEMENT-SOLIDIFIED DECON-TEGRATION REPORT.

WOLF,J R.;

REMPE,J.L.

TAMINATION WASTE COLLECTED FROM NUCLEAR POWER STICKLER L.A.; et al. EG&G Idaho, Inc. March 1994. 168pp STATIONS. AKERS,D.W.; KRAFT,N C.; MANDLER.J.W. EG&G 9405050346. TMI V(93)EG10. 79164:142.

Idaho, Inc. June 1994. 250pp. 9407070300. E3G 2736.

The Three Mile Island Unit 2 (TMI-2) Vessel Investigetion 80113:304.

Project (VIP) was an international effort that was sponsored by A study was performed for the Nuclear Regulatory Commis-the Nuclear Energy Agency of the Organizabon for Economic sion (NRC) to evaluate structural stabihty and teachabihty of ra-Cooperation and Development The pnmary objectives of the donuchdes, stable metals, and chelating agents frorn cement-VIP were to extract and examine samples from the lower head sohdified decontamenation ion-exchange resin wasten collected and to evaluate the potential modes of failure and the margin of from seven commercial bothng water reactors and orie pressur.

structural integnty that remained in the TMI 2 reactor vessel ized water reactor. The decontamination methods tsed at the during the accident. This report presents a summary of the reactors were the CarFDecon, AP/Citrox, Dow NS-1 and LOMI maior findings and conclusions tt'at were developed from re-processes. Samples of untreated resin waste and solidified search dunng the VIP. Results from the various elements of the waste forms were subjected to immersion and compressive project are integrated to form a cohesive understanding of the strength testing. Some waste-form samples were leach-tested vessel's condition after the accident.

using simulated groundwaters and simulated seawater for com-i l

NUREG/CR4198: TMi-2 INSTRUMENT NOZZLE EXAMINA.

panson with the deionized water tests that are normally per.

TIONS PERFORMED AT THE INEL.

AKERS.D.W.;

formed to assess waste-form leachability. This regort presents SCHUETZ.B K.

EG&G Idaho, Inc. March 1994.. 130pp.

the results of these tests and assesses the effectu of the vari-9404010249. EGG-2735. 78728.034.

ous decontaminaton methods, waste form formulabons, lea-As part of the Three Mde Island Unit 2 (TMI-2) Vessel investe chant chemical compositons, and pH of the lea:hant on the gaton Prrvet, unocr the auspeces of the Orgaruzaton for Eco.

structural stabihty and leachability of the waste forms. Resufts nomic Cooperation end Development, examinatons were per.

indicate that releases from intact and degraded wtste forms are formed at the Idaho Natonal Engineering Laboratory on eight similar and that the behavior of some radionucides such as nozzles and one guide tube from the TMI-2 reactor vessel. This (55)Fe, (60)Co, and (99)Tc were similar. In addition, the teacha-docurnent desenbes the examinaton methodology, summanzes bility indexes are greater than 6.0, which meets the requirement the examinaten results, and presents interpretatens of the re-in the NRC's " Technical Positon on Waste Forn " Revison 1.

(

l l

42 Main Citations and Abstracts NUREG/CR4203: VALIDATION STUDIES FOR ASSESSING UN-measured resutts for high-energy threshold reactions in the SATURATED FLOW AND TRANSPORT THROUGH FRAC-cavity by up to 41% and thermal reactons by up to a factor of TURED ROCK.

BASSETT,R.L; NEUMAN,S P.;

30. The transport calculations performed with the ongenal RASMUSSEN,T.C.; et al. Anzor'a, Univ of, Tucson, AZ. August SAILOR cross-section library (based on ENDF/B IV data) over-1994. 219pp. 9409090105. 80816.111.

estimate measured threshold reactions by onfy 15% and the The obectrves of this contract are to examine hypotheses thermal reactions by about a factor of 2.50. These results are t

End conceptual models concerning unsaturated flow and trans-inconsistent with those obtained in earteer studies that com-port through heterogeneous fractured rock and to design and pared transport calculatons done with SAILOR vs ENDF/8 VI, execute confirmatory field and laboratory expenments to test whch indicate that SAILOR tends to underestimate cavity actwi-these hypotheses and conceptual models. Important new infor-ties for threshold reactions, while the ENDF/B V1 values usually mation is presented such as the appicaten and evaluaton of agree better with expenmental results. One factor that probably procedures for estimating hywaulic, pneumate, and solute contnbutes to the rather large discrepancy between the comput-transport coeffcients for a range of thermal regimes. A field ed and measured activities is the core power distnbution used heater expenment was designed that focused on identifying the in the calculations. Because of unavailabelity of plant specife suitability of existing monitonng equipment to obtain required data, a generic power distnbution provided by Westinghouse data. A reliable method was developed for conducting and inter-was used. Since the calculated cavity flux levels appear to be preting tests for air permeability using a straddle-packer ar-over-estimated, the results estimated for the exposure to the l

rangement. Detailed studies of fracture flow from Queen Creek support structure should be conservatwo.

I into the Magma Copper Company ore haulage tunnel have been irwtiated. These studies will provide data on travel time for NUREG/CR4208: AN EMPIRICAL INVESTIGATION OF OPERA.

transport of water and solute en unsaturated tuff. The collection TOR PERFORMANCE IN COGNITIVELY DEMANDING blMU-of rainfall, runoff, and infiltration data at two small watersheds at LATED EMERGENCIES. ROTH,E.M.; MUMAW,R.J. Westing-the Apache Leap Tuff Site enabled us to evaluate the quantity house Electnc Corp. LEWIS P.M.

Human Factors Branch s,nd rate of water infiltrating into the subsurface via oither frac-(880717-941217). July 1994.143pp. 9407250208. 80328:005.

tures or matnx. Charactenzaten methods for hydraulc param.

This report documents the results of an empercal study of nu-eters relevant to high level waste transport, including fracture clear power plant operator performance in cognitwely demand-apertures, transmisswity, matnx porosity, and fracture wetting ing simulated emergencies. Dunng emergencies operators front propagation velocities, were developed follow highly presenptwe wntten procedures. The ob ectwes of t

NUREG/CR4204: QUESTIONS AND ANSWERS BASED ON RE.

the study were to understand and document what role higher.

VISED 10 CFR PART 20. BORGES T.; STAFFORD R.S_;

level cognitwo actwitses such as diagnosis, or more generally LU,P.Y.; et al. Oak Ridge Natonal Laboratory. May 1994.

' situation assessment', play in guiding operator periormance, 111pp.9405310172. ORNL/TM 12690. 79590.226.

gwen that operators utilire procedures in responding to the NUREG/CR-6204 is a collecton of questons and answers events. The study examined crew performance in variants of that were onginally issued in seven sets and which pertain to two emergencies: (1) an interfacing System Loss of Coolant Ac-revised 10 CFR Part 20. The questions came from both outside cadent and (2) a Loss of Heat Sink scenano. Data on operator and within the NRC. The answers were compiled and provided performance were collected using training simulators at two by NRC staff within the offces of Nuclear Reactor ReguWion, plant sites Up to 11 crews from each plant partecipated in each Nuclear Matenal Safety and Safeguards, Nuclear Regulatory of two simulated emergencies for a total of 38 cases. Crew per.

Research, the Office of State Programs, and the fue regional formance was videotaped and partial transcnpts were produced offces. Although all of the questions and answers have been and analyzed. The results revealed a number of instances reviewed by attorneys in the NRC Office of the General Coun-where hegher-level cognitwe actuities such as situaton assess-sel, they do not constitute official legal interpretations relevant ment and response planning enabled crews to handle aspects to revised 10 CFR Part 20. The questens and answers do, of the situaton that were not fully addressed by the procedures.

however, reflect NRC staff decisions and techncal options on This report documents these cases and discusses their implica-aspects of the revised 10 CFR Part 20 regulatory requirements.

tions for the development and evaluaton of training and control This NUREG is being made available to encourage communca.

room aids, as well as for human reliability analyses.

ton among the public, industry, and NRC staff concerning the malor revisions of the NRC's standards for protecton against NUREG/CR-6209: MEMPHIS AREA REGIONAL SEISMIC radiation.

NETWORK. Final Report, October 1986 September 1992.

CHlU,J M : JOHNSTON,A.C. Memphis State Unw., Memphis, NUREG/CR-6205: VALVE ACTUATOR MOTOR DEGRADATION.

TN March 1994. 69pp. 9404110348. 78831:143.

KUECK.J D. Oak Rdge National Laboratory. December 1994.

The Memphis Area Regional Setsmic Network (MARSN) has 41pp. 9501180329. OHNL-6796. 82347:191.

provided important southem coverage of the New Madnd seis-Valve actuator motor degradation and failure has been a sig-mic zee (NMSn The mtwork hew to WnW h Men nifcant, but little studied, problem in the nuclear industry. This County fault zone located east of and parallel to the SW seg-study provides a discusson of the pnrnary failure mode tuer.

ment of the NMSZ. MARSN data has also been added into a mal degradation - and reviews the basis for the soluton of ther-database obtained by the PANDA expenment for a comprehen-mal degradation - thermal protection. The study also provides sue seismological study of the NMSZ. Results demonstrate that reviews of vanous industry data bases, discusses effects of earthquakes h the MZ am mai@ cmW @ @s other failure modes such as corroson, and provides a review of other consideratons the user should entertain when assessing ang g o o

m NW n N ments of the ed central NMSZ, however, is very compicated. The northcentral NUREG/CR-6206: TRANSPORT CALCULATIONS OF RADI-section is characten2ed by a well-defined planar feature dipping ATION EXPOSURE TO VESSEL SUPPORT STRUCTURES IN

~31 degrees SW which shows dominantly normal faulting. The THE TROJAN REACTOR. ASGARI.M ; WILLIAMS.M.L. Louisi-southcentral secten shows a ~48 degrees SW dipping fault ana State Unw., Baton Rouge, LA. KAM F.B K.; et al. Oak Ridge with dominantly reverse faulting Although the E-W regional National Laboratory. Jufy 1994. 74pp. 9412020031. ORNL/TM-stress may play an important role in fault movements in the 12693. 81941:001.

NMSZ, the complication in focal mechanisms suggests that Cornpansons of transport calculations of the dosimeter actwo-other factors including postseismic relaxation by the 1811 1812 ties with the expenmental measurements shows that the values earthquakes, interactions between ad acent fault segments, or i

obtained with ENDF/8-VI cross-section data overestimate the features such as the nght-lateral stnke-slip Crittenden County

4 Main Citations and Abstracts 43 fault cannot be overlooked in future tectonic studies of the NUREG/CR4216: EVALUATION OF ROCK JOINT MODELS P!D

NMSZ, COMPUTER CODE UDEC AGAINST EXPERIMENTAL RE-NUREG/CR4211: INTEGRATED FUEL-COOLANT INTERAC-SULTS. HSIUNG,S.M.; GHOSH.A.; CHOWDHURY,A.H.; et al.

TION (IFCI 6.0) CODE. User's Manual DAVIS,F.J.; YOUNG,M.F.

Center for Nuclear Waste Regulatory Analyses. November Sandia Nabonal Laboratones. Apnl 1994. 68pp. 9405310193.

1994.116pp. 9411280288. CNWRA 93-024. 81851:276.

SAND 94-0406. 79590:044.

The Mohr-Coulomb, Barton-Bandis, and Continuously-Yielding The Integrated Fuel-Coolant interacton (IFCl) computer code rock joint models and their numencal implementaten in the is being developed at Sandia National Laboratories to investi.

UDEC code were evaluated for their ability to simulate joint be-gate the fuel-coolant interacton (FCI) problem at large scale havior under cyclic pseudostatic and dynamic loading condi-using a two-dimensonal, four-field hydrodynamic framework and tions. Some deficiencies of these joint models and their imple-physically based models. IFCI will be capable of treating all mentaten in UDEC were identified. These deficiencies include major FCI processes in an integrated manner. This document is that the rock joint models under evaluaton may not be able to a product of the effort to generate a stand-alone version of sufficiently predict the joint shear and dilation behavior during IFCl, IFCI 6.0. The User's Manual describes in detail the hydro-werse joint shearing. Both joent forward and reverse shearing dynamic method and physical rnodels used in IFCI 6.0. Appen-

,.3 important phenomena of a rock joint behavior. Reverse dix A is an input manual, provided for the creaton of working shearing can result from earthquake, thermal load, or both - all decks.

of which are expected to be experienced during the hfe of a NUREG/CR4212: VALUE OF PUBLIC HEALTH AND oAFETY high-level waste reposstory. These deficiencies could result in ACTIONS AND RADIATION DOSE AVOIDED. BAUM J.W.

an overestimation of the stability of emplacement drifts and ern-

)

Brookhaven National Laboratory. May 1994. 54pp. 9407260025.

placement boreholes and predicton of incorrect near-field flow i

BNL-NUREG-52413. 80342:241, pattern (including preferential pathways for water and gas).

The values judged best to reflect the wilkngnens of society to pay for the avoidance or reducton of risk were deduced from NUREG/CR4218: A REVIEW OF THE TECHNICAL ISSUES OF studies of costs of health care, transportaten safety, consumer AIR INGRESSION DURING SEVERE REACTOR ACCIDENTS.

product safety, govemment agency actons, wage-risk compen-POWERS,D.A.; KMETYK,LN.; SCHMIDT R.C. Sandia National sation, consumer behavior (market) studies, and wilkngness-to.

Laboratones. September 1994. 84pp. 9409270204. SAND 94-pay surveys. The results ranged from $1,400,000 to $2,700.000 0731, 81046:023.

per hfe saved. Applying the mean of these values ($2.100.000)

Severe reactor accident scenarios involving air ingression into and the latest risk per unit dose coefficents used by the ICRP the reactor coolant system are described. Evidence from (1991), which take into account risks to the generrl public, in-modem reactor accident analyses and from the accident at cluding genetic effects and non-fatal cancers, yields a value of Three Mile Island shows residual fuel will be present in the core j

dose avoided of $750 to $1,500 per person-cSv for public expo-region when air ingresson is possible. This residual fuel can sures. The lower value applies if adjustrnents are made for interact with the air. Exploratory calculatons with the MELCOR years of hfe lost per fatahty. A nominal value of $1,000 per person-cSv seems appropnate in light Jf the many uncertainties code of station blackout accidents dunng shutdown conditons involved in deducing these values. These values are consistent and during operatons are used to examine clad oxidation by air and ruthonium release from fuel in air. Extensive ruthenium re-with values recommended by several European countnes for in-devidual doses in the region of 1 mSv/y (100 mrem /y). Below ase is Meted when ak @sson raks exceed about 10 this dose rate, most countnes have values a factor of 7 to 10 moles /s. Past studies of air interactons with irradiated reactor i

lower, based on the assumpton that society is less concemed fuel are reviewed. Effects air ingression may have on fisson l

with fatahty nsks below about 10(-4)/y.

product release, transport, 46,i, and revaponzation are i

discussed. Perhaps the most important effects of air ingression j

NUREG/CR4213: HIGH-TEMPERATURE HYDROGEN-AIR-are expected to be enhanced release of ruthenium from the fuel STEAM DETONATION EXPERIMENTS IN THE BNL SMALL" and the formation of copious amounts of aerosol from uranium SCALE DEVELOPMENT APPARATUS. CICCARELLI.G.;

oxide vapors. Revaporization of iodine and tellurium retained in 1,

GINSBURG,T.; BOCCIO,J.; et al. Brookhaven Natonal Labora-the reactor coolant system might be expected.

tory. August 1994. 99pp. 9409230300. BNL-NUREG-52414.

81011 M NUREG/CR-6221: THE VALLES NATURAL ANALOGUE j

The Small-Scale Development Apparatus tSSDA) was con-PROJECT, STOCKMAN.H.W.; KRUMHANSL,J.L; HO.C.K.; et al.

structed to provide a prehminary set of expenmental data to Sandia Natonal Laboratories. December 1994. 126pp.

charactenze the effect of temperature on the abihty of hydro-9412300204. SAND 94 0650. 82162:197.

gert-at-stoam mixtures to undergo detonatens and, equally im-portant, to support design of the larger scale High-Temperature The contact between an obsidian flow and a steep-walled tuff Combuston Facihty (HTCF) by providing a test bed for soluton canyon was examned as an analogue for a Wel waste re-of a number of high-temperature design and operatonal prob-pository. The analogue site is located in the Valles Caldera in lems. The SSDA, the central element of which is a 10<m inside New Mexico, where a massive obsidian flow filled a paleocan-diameter,6.1-m long tubular test vessel desegned to permit det.

yon in the Battleship Rock tuff. The obsidian flow provided a onaten experrnents at temperatures up to 700K, was employed heat source, analogous to waste panels or an igneous intrusion j

to study self sustained detonations in gaseous mixtures of hy.

in a repository, and caused evaporation and migration of water.

I drogen, ar, and steam at temperatures between 300K and The tuff and obsidian samples were analyzed for major and 650K at a fixed initial pressure of 0.1 MPa. Detonation cell size trace elements and mineralogy by INAA, XRF, X-ray diffracton, i

measurements provide clear evidence that the effect of hydro-and scanning electron microscopy and electron microprobe.

gen-air gas mixture temperature, in the range 300K 650K, is to Samples were also analyzed for D/H and (39)Ar/(40)Ar isotopic decrease cell size and, hence to increase the sensitivity of the compositon. Overall, the effects of the heating event seem to mixture to undergo detonatons. The effect of steam content, at have been shght and hmited to the tuff nearest the contact.

any given temperature, is to increase the cell size and, thereby, There is some evidence of devitnficaton and migration of vola-to decrease the sensitmty of stoichiometnc hydrogen-air mix-tiles in the tuff within 10 meters of the contact, but vanatons in tures. The hydrogen-ar detonabihty hmits for the 10-cm inside mtior and trace element chemistry are small and difficult to dis-diameter SSDA test vessel, based upon the onset of single-tinguish from the natural (pre-heating) variability of the rocks.

I head spin, decreased from 15 percent hydrogen at 300K down i

to between 9 and 10 percent hydrogen at 650K.

I 1

44 Main Citations and Abstracts NUREG/CR-6223: REVIEW OF THE PROPOSED MATERIALS dertaken. Virtualty every U.S. nuclear pipe fernte steel grade OF CONSTRUCTION FOR THE SBWR AND AP600 AD-tested had pipe lengths that were susceptible to DSA and un-VANCED REACTORS.

DIERCKS.D.R.;

SHACK,W.J.;

stable crack Jumps at 288 degrees C (550 degrees F). Data CHUNG H.M.; et at Argonne National Laboratory May 1994.

from specimens and pipe tests showed that the ins l abilities are 113pp. 9406210280. ANL-94/13. 79836:215.

random and the number of or precise locatsons of crack jumps The General Electre Simphfied Dailing Water Reactor cannot be predicted accurately, however the susceptible of a (SBWR) and the Westinghouse Advanced Passive 600 MWe steel to crack jumps occurring can be predicted. A screening Reactor (AP600) have been reviewed in detail by Argonne Na-cntena based on the rato of the hardness at operating versus tional Laboratory. The objectives of these reviews were to (a) room temperature was developed to make these predictons.

evaluate proposed advanced-reactor designs and the matenals Since expenments at LWR temperatures on U.S. femtic nuclear of construction for the safety systems, (b) identify all aging and pipe steel have been shown that large crack jumps occurred in environmentally related degradaton mechanisms for the materi-pipe and laboratory expenments, femte steels that are highly als of construction, and (c) evaluate from the safety viewpoint sensitive to crack jumps, should be used with cauten. The DSA the suitabihty of the proposed matenals for the design apphca-hardness screening entenon is relatively simple to implement ton. The matenals selected for both reactors were generally and could be used in situ on existing plant piping systems, or as sound, and no major selection errors were found. It was appar-a mill quality control test when purchaserg new matenals. The l

ent that considerable thought had been given to the matenals data suggest that it may be possible to manufacture femte selection process, making use of lessons learned from previous steel pipe that is less susceptible to DSA, whch would be bene-LWR expenence. The review resulted in the suggeston of alter-ficia! for future advanced light water reactor plants or replace-l nate and possibly better matenals choces in a number of ment piping in existing plants.

I cases, and several potential problem areas have been cited.

The review of the AP600 matenals of constructen was impaired NUREG/CR-6228: PRELIMINARY ASSESSMENT OF THE FRAC-by the fact that the matenals designatens given in the Standard TURE BEHAVIOR OF WELD MATERIAL IN FULL THICKNESS Safety Analysis Report (SSAR) were often too vague to identify CLAD BEAMS. KEENEY,J.A.; BASS.B.R.; MCAFEE W.J.; et al.

the specific alloy to be used. The SSAR for the SBWR generally Oak Ridge Natonal Laboratory. October 1994. 64pp.

l gave more detailed matenals mformaton.

9411040076. ORNL/TM 12735. 81646:193.

A testing program is desenbed that utshzes %ill-thickness clad NUREG/CR-6224 DFC: PARAMETRIC STUDY OF THE POTEN-beam speime to quantify fracture toughness for shallow TIAL FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA cracks in matenal for which metallurgical co1ditions are proto-GENERATED DEBRIS. Draft For Comment. ZlGLER.G.;

typic of those found in reactor pressure vensels (RPVs). The BEATY,R.; BRIDEAU,Ja et al Science & Engineenng Associ-beam specimens are fabreated from a section of an RPV wall I

ates, Inc. August 1994. 238pp. 9408310226. SEA 93-554-06-A1.

80719:001.

(removed from a canceled nuclear plant) that includes weld, This report documents a plant-specife study for a BWR 4 plate and clad material. Metallurgcal factors potentially influenc-with Mark l Containment, that evaluated LOCA-generated debns ing fracture toughness for shallow cracks in the beam speci-phenomena and the potential for losing recirculation coohng ca, mens include matenal gradients due to welding and cladding pability due to ECCS pump sucten strainer clogging and loss of appleatens, as well as matenal inhomogenetties in welded re-net positive suction head (NPSH) margin. The major elements gens due to reheating in multiple weld passes. Summanes of of the study were: (1) acquisition of data from a reference BWR; the testing program and analytcal studies are provided. Frac-(2) analysis of weld failure frequencies to estimate the LOCA ture toughness data from the three clad beam specimens are frequency; (3) development of BWR debns generaton and compared with other shallow-and deet > crack uniaxial beam debns transport models; (4) modehng debns transport en the and cruciform data generated previously from A 533 Grade B suppresson pool (5) estimation of strainer blockage frequency plate matenal. Diffculties with interpreting lower-bound fracture and loss of NPSH margin; and (6) estimaton of core damage toughness curves constructed from the shallow-crack data are frequency attnbutable to a loss of ECCS following a large essentially resolved by adopting a sing le normalizing tempera-l LOCA. A point estimate overall DEGB pipe break frequency (per ture parameter, namely, the nil ductihtr temperature (NDT).

Rx year) of 1.6E-04 was calculated with a corresponding overall When normalized to NDT, the combermd shallow-crack data i

DCCS loss of NPSH frequency (per Rx year) of 1.5E-04. The base from the plate and weld matenal exhibst an elevated mean l

point estimate of core damage frequency (por Rx-year) due to fracture toughness in the transrtion temperature region, accom-l blockage-reiated accident sequences for the reference plant Panied by increased data scatter that tends toward the same ranged from 4.2E-06 to 2 SE 05. The results of this study show lower hound associated with highly-constrained deep-crack that severe strainer blockage and loss of NPSH can occur data. Additional full-thickness c'ad beam testing is proposed to within the first 30 minutes of the LOCA if partculates are complete the investigation of shallow-crack fracture toughness present en additon to the LOCA-generated debns. This study 61 behaver in prototypic weld and plate matenal. This report is lustrates the competing effects of debns settling versus turbu.

designated as HSST Report No.123.

lence effects, whch can take place in the suppresson pool.

NUREG/CR-6231: A COMPARISON OF THE RELATIVE IMPOR.

NUREG/CR-6226: EFFECT OF DYNAMIC STRAIN AGING ON TANCE OF COPPER PRECIPITATES AND POINT DEFECT THE STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC CLUSTERS IN REACTOR PRESSURE VESSEL EMBRITTLE-PIPING AT LWR TEMPERATURES. MARSCHALLC.W.;

MENT. STOLLER,R.E. Oak Ridge National Laboratory. Decem-MOHAN.R.; KRISHNASWAMY,P.; et al. Battelle Memonal Inste-ber 1994. 52pp. 9501180295. ORNL-6811. 82341.276.

tute.

Columbus Laboratones. October 1994.

102pp.

Embnttlement of reactor pressure vessel (RPV) steels is be-9411160005. BMI-2176. 81756 201.

heved to anse pnmanly from matnx hardening due to the forma-This topecal report is on the phenomenon of dynamic strain ton of radiation-induced point defect clusters (PDC) and radi-aging (DSA) in femte nucleat piping steels and its effect on aton-enhanced, copper-rch precipitates ICRP). A model has fracture at LWR temperatures. The report was a dehverable been developed to investigate the relative importance of PDC from the U S. NRC's program entitled "Short Cracks in Piping and CRP in RPV embnttlement. The model has been used to and Piping Welds". The oblectrve of th,s work was to predet the determine the influence of a range of irradiaton and matenal occurrence of and evaluate the effects of ductile crack instabil-parameters on the predicted change in yield strength. The re-itses, which occur frequently in femtic steel pipe fracture tests at cults indcate that there are temperature and displacement rate 288 degrees C (550 degrees F), and are beheved to be due to regimes wherein either CRP or PDC can dominate the maten-dyname strain aging Numerous laboratory tests and one nu-al's response to arradiaton. Both interstitial land vacancy type mercal simulation of a C(T) test with crack instabihties were un-defects contribute to the PDC componert. The different de-

6 Main Citations and Abstracts 45 l

pendencies of the CRP and PDC on temperature and displace-fracture mechanics analysss may be more appropriate than a ment rate indcate that simple data extrapolations could lead to displacement-controlled analysis for these tests.

poor predictions of RPV embrittlement. Fortunately, the yield l

strength changes predcted by the composite PDC /CRP model NUREG/CR4234: VALIDATION OF ANALYSIS METHODS FOR 1

exhibit very httle dependence on displacement rate below about ASSESSING FLAWED PIPING SUBJECTED TO DYNAMIC l

5 x 10(-9) dpa/s " this result is confirmed, concems about ac.

LOADING. OLSON,R.J.; WOLTERMAN,R.L; WILKOWSKI,G.M.;

l celerated Gplacement rates in power reactor surveillance pro-et al. Battelle Memorial Institute, Columbus Laboratories. August I

greas should be minimized. The sensitivity of the model to me-1994. 68pp. 9409060227. ANL-94/22. 80785:028.

crostructural parameters highlights the need for more detailed Argonne Natonal Laboratory and Battelle have jointly con-charactenzation of RPV steels.

ducted a research program for the USNRC to evaluate the atxli-l

" *"O "0*

NUREG/CR-6232: ASSESSING THE ENVIRONMENTAL AVAIL-art arWysis method to predict the behavior of circumferentially ABILITY OF URANIUM IN SOILS AND SEDIMENTS.

surface-ciacked pipe system water-hammer expenments. The AMONETTE.J E.; HOLDREN G R.; KRUPA,K.M.; et al. Battelle exper nental data used in the evaluabon were from the HOR e

Memorial Institute, Pacife Northwest Laboratory. June 1994.

Tee; Group E31 senes conducted by the Kemforschungszen-l l

116pp. 9406290309. PNL-9750. 80014:097.

,/um Karlsruhe (KfK) in Germany. The incentive for this evalua-This report reviews existing approaches to determine their po-ton was that simplified engineering methods, as well as newer tential for assessing the environmental availability (i.e., aqueous

.. state-of-the-art" fracture analysis methods, have been typically l

solubility) of uranium in bulk soils or sediments. Direct empirical validated only with state expenmental data. Hence these dy-approaches that involve extractons by aqueous solutens pro-namic expenments were of high interest. High-rate dyname vide estimates of the solubility of operatonalty defined compo-loading can be classified as either repeating, e.g., seismic, or nents. Altemately, indirect approaches based on advanced in-nonrepeating, e g., water hammer. Development of experimental j

strumental techniques can be used to identify specific forms of data and validation of cracked pipe analyses under seisme uranium; the contnbuten of each form to the solubility is then loading (repeating dynamic loads) are being pursued separately inferred from geochemcal equilibnum and kinetc modera. At within the NRC's International Piping integnty Research Group present, the direct emparcal approaches are more likely to pro-(IPIRG) program. This report described developmental and vali-vide useful estimates of environmentally available uranium at dation efforts to predct crack statulity under water-hammer reasonable costs. For the long term, we recommend develop-loading, as well as compansons using currently used analysis ment of a flow-cell procedure that incorporates dissolution-rate rocedures. Current fracture analysis methods use the elastc information into the assessment of environmental availability, in the intenm, we recommend development of a batch wet-cheme-Y g

cal procedure based on a combinaten of standard and non-analysis, while stateof-the-art methods employ nonhnear standard methods. Both procedures require laboratory testing cracked-pipe time-history finite element analyses. The results showed that the current decoupled methods were conservative and correlaton with field data before being used for regulatory purposes. Lastly, we have tabulated literature oata on the aque-in their predictons, whereas the cracked pipe finite element ous complexes of uranium and major uranium minerals, shown analyses were more accurate, yet shghtly conservative. The that the environmental availatxlity of uranium cannot be predict, nonhnear time-history cracked-pipe finite element analyses con-ed from thermodyname solubility data stone, and compiled a last ducted in this program were also attractive in that they were of laboratones capable of performing envenmental availabihty done on a small Apollo DN5500 workstation, whereas other determinatens.

cracked-pipe dynamic analyses conducted in Europa on the same experiments required the use of a CRAY2 supercocouter, NUREG/CR4233 V01: STABILITY OF CRACKE0 PIPE UNDER and were less accurate.

INERTIAL STRESSES. Subtask 1.1 Final Repot SCOTT,P.M.;

WILSON.M.; OLSON.R.J.; et al. Battelle Memonal Institute, Co-NUREG/CR4236: SEISMIC INVESTIGATIONS OF THE HDR l

lumbus Laboratones. August 1994.195pp. 9409260057. BMI-SAFETY PROGRAM Summary Report.

MALCHER,L; l

2177. 81043.001.

SCHRAMMEL.D.; STEINHILDER.R; et al. Germany, Federal l

This report presents the results of the pipe fracture experi-Republic of. August 1994. 95pp. 9409060242. ANL 94/20.

i ments, analyses, and matenal charactenzation efforts performed 80785:249.

I within Subtask 1.1 of the IPIRG Program. The objective of Sub-The primary objective of the seismic investigations, performed task 1.1 was to expenmentally venfy the analysis methodologies at the HDR facility in Kahl/ Main, FRG was to validate calcula-for circumferentially cracked pipe subjected primanly to inertial tional methods for the seismic evaluation of nuclear-reactor sys-stresses. Eight cracked-pipe expenments were conducted on 6-tems, using expenmental data from an actual nuclear plant.

inch nominal diameter TP304 and A106B pipe. The expenmen-Using eccentnc mass shaker excitation the HDR soil / structure tal procedure was developed using nonhnear time-history finite system was tested to incipient failure, exhibiting highly nonhnear element analyses whch included the nonhnear behavior due to response and demonstrating that structures not seismcally de-the crack. The model did an excellent job of predcting the dis-signed can sustain loads equivalent to a design basis earth-placements, forces, and times to maximum moment. The com-quake (DBE). Load transmission from the structure to piping /

panson of the expenmental loads to the predicted loads by the equipment indcated signifcant response amphfcations and Net-Section-Collapse (NSC), Dimensionless Plastc Zone Pa-shifts to higher frequencies, while the response of tanks /ves-rameter, J-estimaton schemes, R6, and ASME Section XI in-seis depended mainly on their support conditons. The evalua-service flaw assessment cntena tended to underpredict the ton of vanous piping support configurations demonstrated that measured bending moments except for the NSC analysis of the proper system design (for a given spectrum) rather than number A106B pipe. The effects of flaw geometry and loading history of supports or system stiffness is important to limiting pipe on toughness were evaluated by calculating the toughness from stresses. Piping at loads exceeding the DBE eightfold still had the pipe tests and companng these results to C(T) values.

significant margins and failure is improbable in spite of multiple These effects were found to be variable. The surface-crack ge-support failures. The mean value for pipe damping, even under ometry tended to increase the toughness (relative to C(T) re-extreme loads, was found to be about 4%. Companson of linear suits), whereas a negative load-rato significantly decreased the and nonhnear computatonal results with piping response meas-TP304 stainless steel surface-cracked pipe apparent toughness.

urements showed that predictions have a wide scatter and do The inertial expenments tended to achieve complete failure not necessarily yield conservatrve responses underpredicting, in within a few cycles after reaching maximum load in these rela-particular, peak support forces. For the soil / structure system tively small diameter pipe expenments. Hence, a load-controlled the quakty of the predctions did not depend so much on the

46 Main Citations and Abstracts complexity of the modeling, but rather on whether the model NUREG/CR-6249: UNIRRADIATED MATERIAL PROPERTIES OF captured the sahent features and nonhnearities of the system.

MIDLAND WELD WF-70. MCCADE.D.E.; NANSTAD,R.K.;

ISKANDER,S.K.; et al. Oak Ridge National Laboratory. October NUREG/CR-6237: STATISTICAL ANALYSIS OF FATIGUE 1994. 93pp. 9411160151. ORNL/TM-12777. 81756:303, STRAIN-LIFE DATA FOR CARBON AND LOW-ALLOY STEELS.

Wold metal, designated WF-70, taken from the nozzle course KEISLER J ; CHOpRA,0.K.; SHACK W.J. Argonne National Lab-and beltline welds of the Midland Reactor, Unit 1, has been oratory. August 1994. 52pp. 9409260059. ANL-94/21.

grven a preliminary evaluation using the conventional Charpy V-81043:196.

notch (CVN), drop-weight (DWT), and chemical analyses. There The existing fatigue strain vs.hfe (S-N) data, foreign and do-was a significant difference in copper content, nominally 0.25%

mestic, for carbon and low-alloy steels used in the construction versus 0.40% Because the objective of this study was to evalu-of nuclear power plant c,0,npcwnts have been compiled and ate the before-and-after irradiation properties, these are regard-categorized according to material, loading, and environmental ed as different materials. This rerort summarizes material char-conditions. A statistcal model nas been developed for estimat-acterization results and present". the reiutts of fracture mechan-ing the effects of the various test conditions on fatigue Efe. The ics tests on the urwradiated material to istablish baseline tran-results of a ngorous statisteal analysis have been used to estL sit >on temperature and J-R curves. Turade properties were also mate the probability of initiating a fatigue crack. Data in the liter.

determined. Reference nil-ductility temperatures, RD(NDT), de-ature were reviewed to evaluate the e'fects of size, geometry, termined from CVN transition curves (RT(NOT)) method specific and surface finish of a wingv e6t on its fatigue hfe. The fatigue to low upper-shelf energy materials) varied from -20 to + 37 de-S N curves for cea,pceents have been determined by applying grees C (-4 to 99 degrees F) at various locations in the belthne design margins for size, geometry, and surface finish to crack weld Reference temperatures using a fracture mechanics.

initiation curves estimated from the model The significance of based transition temperature rnodel were -60 degrees C ( 76 the effect of environment on the current Code design curve and degrees F) for the beltline weld and -33 degrees C (.27 degrees on the proposed interim design curves for carbon and low-alloy F) for the nozzle weld. Tensile tests indicated the nonle weld j

steels presented in NUREG/CR-5999 is discussed.

had higher strength than the beltline weld. J-R curves were de-veloped at 21, 50, and 288 degrees C (70,32, and 550 degrees NUREG/CR-6241: TECHNICAL GUIDELINES FOR ASEISMIC F). The predeted J-R curves matched the expenmental J-R DESIGN OF NUCLEAR POWER PLANTS. Translation Of JEAG curves reasonably well. Crack-arrest tests were conducted, but 4601 1087. PARK,Y.J.; HOFMAYER C.H. Brookhaven Nabonal the specimens failed to develop Amencan Society for Testing Laboratory. June 1994. 962pp. 9407080256. BNL-NUREG-and Matenals vahd data in all but one test. More expenmenta-

$2422. 80129:001.

tion and test rnethod development are recommended. Five ex-This document is a translation, in its entirety, of the Japan penmental objectives to be accomplished from the teshng of ir-Electric Association (JEA) pubhcation enhtled " Technical Guide.

radiated matertal were identified.

knes for Aseisme Design of Nuciear Power Plants JEAG NUREG/CR-6250:

SUMMARY

OF COMMENTS RECEIVED ON 4601-1987. This guidehne describes in detail the asessme STAFF DRAFT PROPOSED RULE ON RADIOLOGICAL CRITE-desegn techniques used in Japan for nuclear power plants. It RIA FOR DECOMMISSIONING. CAPLIN J.; PAGE,G.; SMITH.D.;

contains chapters deahng with: (a) the selection of earthquake ground motions for a este, (b) the investigation of foundation and et al. Advanced Systems Technology, Inc. August 1994.123pp.

9409060247,80786:001'ry Cun,e.,06 (NRC) is conducting a bedrock conditions, (c) the evaluation of ground stabihty and the The Nuclear Regulato effects of ground movement on buried peping and structures, (d) the analysis and desgn of structures, and (e) the analysis and enhanced partcapatory rulemaking to establish radeological crite-ria for the decommissioning of NRC-licensed facihties. The NRC design of equipment and distnbution systems (piping, electrical obtained comments on the scope, issues, and approaches raceways, instrumentation, tubing and HVAC duct). The guide-through a senes of workshops (57 FR 58727), Generic Environ-kne also includes appendees which summarize data, informa~

rnental impact Statement (GEIS) scoping meetings (58 FR tson and references related to aseismic design technology.

33570), a dedicated electronic bulletin board system (58 FR

,a wn n ases. A swnmary of Wahop and EW/CR-6245: ASSESSMENT OF PRESSURIZED WATER RLTCTOR CONTROL ROD DRIVE MECHANISM NOZZLE scenMng caneh was pub shed as WREER-6M n e

2. W, the hse @M h h %al CRACKING. SHAH,V.N.; WARE.A.G.; PORTER,A.M. EG&G

?

E"E**

Idalrc Inc. October 1994. 67pp. 9412300191, EGG-2715-821'19:255*

a staH & aft' proposed mie on radologeal cntena fw h missioning Copees of the staff draft were distributed to the Th s report surveys the field experience related to cracking of Agreement States, participants in the earher meetings, and pressurized water reactor (PWR) control rod drive rnechanism other interested parties for comment. This report summartzes nozzles (Alloy 600 rnatenal); evaluates design, fabrication, and the comments identified from the 96 docketed letters received operating conditons for the nozzles in U.S. PWRs; and evalu-on the staff draft. No analysts or response is included in this ates the safety ssgnificance of nozzle cracking. Inspection at 78 report. The comments reflect a broad spectrum of viewpoints.

overseas and one U.S. PWR has revealed mainly axial cracks in Two subjects on which the commenters were in general agree.

101 nozzles. The cracking is caused by pnmary water stress ment were (1) that the enhanced partcipatory rulemaking corrosion cracking, which requires the samuttaneous presence of should proceed, and (2) that the forthcoming GEIS and guid-high tensale stresses, high operating temperatures, and suscep-tible rnicrostructure. CRDM nozzle craciung is not a short-term ance documents are needed for better understanding of the draft rule' safety issue. An axial crack is not hkely to grow above the vessel head to a critical length because the stresses are not NUREG/CR-6252: LESSONS LEARNED FROM THE THREE high enough to support the growth away from the attachment MILE ISLAND-UNIT 2 ADVISORY PANEL LACH D.;

weld. Primary coolant leaking through an axial crack could DOLTON,P.; DURBIN.N.; et at Battelle Seattle Research cause a short circumferential crack on the outside surface.

Center. August 1994. 53pp. 9409060253. PNL-9871. 80786:124.

However, this crack is not likely to propagate through the nozzle in response to public concern about the cleanup of the Three wall to cause rupture. Leakage of the primary coolant from a Mile Island, Unit 2 (TMI.2) facihty after an accident on March through-wall crack could cause boric acid corrosion of the 28, 1979 involving a loss of reactor coolant and subsequent vesset head and challenge the structural integnty of the head, damage to the reactor fuel, twelve cittzens were asked to serve but it is very unlikety that the accumulated deposits of bonc acid on an independent Advisory Panel to consult with the Nuclear crystals resulting from such leakage could remain undetected.

Regulatory Commission (NRC) on the decontamination and

Main Citations and Abstracts 47 cleanup of the facihty. The panel met 78 times over a period of Amencan, where the recurrence interval of large earthquakes is thirteen years, holding public meetings in the vicinity of TMI-2 longer than the histore record of earthquakes. Because surface and meeting regularly with NRC Commissioners in Washington, traces of seismogenic faults have been diffeult to identify in D.C. This report desenbes the results of a project designed to eastern North Arnenca, most paleoseismetty sbdies have em-identfy and desenbe the lessons learned from the Advisory ployed features resulting from hquefaction. The goals of paleoli-Panel and place those lessons in the context of what we gener.

quefaction studies are to determine the recurrence intervals, ally know about citizen advisory groups. A summary of the em-magnitudes, and source areas of prehistoric earthquakes. To pirical hterature on cat zen advisory panels is followed by a bnef accomplish these goals, one must be able to idenbfy earth-history of the TMI-2 Advisory Panel. The body of the report con-quake-induced hquefaction features, determine their ages, and tains the analysis of the lessons teamed, preliminary conclu-map their distribution. This report reviews (1) sediment deforma-sions about the effectiveness of the Panel, and impicabons for tion structures, (2) methods for dating hquefaction features, and the NRC in the use of advisory panels. Data for the report in-(3) relationships between liquefaction and the magnitude and clude meeting transcnpts and interviews with past and present distance of causative earthquakes. Recent studies by the author Panel partespants.

in Quebec Province, Canada and in the New Madrid seismic NUREG/CR4254: SOUTHERN APPALACHIAN REGIONAL SEIS-r ne of the central US provide the basis for this report.

MIC NETWORK. CHlU.S-C.C.; JOHNSTON A.C.; CHlU,J-M.

NUREG/CR4262: CLEAVAGE BEHAVIORS IN NUCLEAR Memphes State Univ., Memphis, TN. August 1994. 81pp.

VESSEL STEELS. IRWIN,G.R.; ZHANG,X.J. Maryland, Univ. of, 9409230288.81010:035.

College Park, MD

• Oak Ridge National Laboratory. November Memphis State University has monitored the seismicity of the 1994. 27pp. 9412300223. ORNLSUB79777811. 82163:001, southern Appalachian area since late 1979 by means of the Cleavage behaviors of nuclear vessel steels in the transition Southem Appalachian Regional Seismic Network (SARSN),

temperature range are reviewed. Viewpoints are presented to which has provided good spatial coverage for earthquake loca-assist understanding of cleavage crack speed, cleavage initi-tion. Activity is more heavily concentrated in the Valley and ation, cleavage arrest, and the sensstrvity of fracture toughness Ridge province (VR) of eastem Tennessee than in the Blue to constraint and temperature. The importance of high local Ridge (BR) or Piedmont (P). The maionty of these events he be-stress elevabons by high strain rate is emphasized. Thes report tween the New York-Alabama and the Chngman/Ocoee linea-is designated as HSST Report No.149.

ments, magnetc anomalies of deep seated basement struc-tures. Thus, SARSN has been able to define the first order NUREG/CR4267: AIR-WATER SIMULATION OF PHENOMENA charactenstes of the Southem Appalachian seismic zone. The OF CORIUM IASPERSION IN DIRECT CONTAINMENT HEAT.

local depths of the earthquakes are concentrated between 8 ING. ISHil M.; REVANKAR,S.T.: ZHANG,G.; et al. Purdue Univ.,

and 16 km, pnncipally beneath the Appalachian overthrust. In West Lafayette, IN. October 1994.126pp. 9411160157. PU NE-cross-secten the average seismicity is shallower beneath the 93/1, 81757:031.

BR and P provinces than the VR and North Amencan craton.

The degree of conum dispersion has not only the strongest Results of focal mechanism studies for events that occurred be-parametnc eftsct on the containment pressunzation but also I

tween October 1986 and December 1991, indcate that the has the highest uncertainty in predteting it. In view of this, sepa-basement of the VR province is under honzontal, NE-SW com-rate effect tests on the corium dispersion mechanisms in the re-pressive stress. Right-lateral stnke-slip faulting on nearty N-S actor cavity and the subcompartment trapping mechanisms faults es preferred as it agrees with the trend of the regional were carried out. Four major objectives of this conum dispersion magnetic anomaly pattem-study are: (1) to perform a detailed scaling study using the NUREG/CR425$: DESIGN OF AN OPEN ARCHITECTURE SEIS.

newly proposed step-by-step integral scahng method, then to MIC MONITORING SYSTEM. GHAllB.H.A.; LEONARD.S K.;

evaluate existing models for entrainment, parbcle stze and trap-KRAFT,G.D.

ENSCO, Inc.

September 1994.

125pp.

ping. (2) to perform carefully desegned simulaton expenments 9410060315. DCS-94-024. 81223 235.

using water-ak and woods metal-air in a 1/10 knear scale Thcs document presents a top level design of an Open Archi.

model, (3) to develop rehable mechaniste models and correla-tecture Seismic Monitonng (OASM) system intended to auto-tions for corium dispersions, which can be used to predict matically monitor local and regional seisme activibes. The corium jet disintegraton, entrainment, drop size, hquid film carry system is designed to process single and three component data over, and subcompartment trapp6ng, and (4) to use the models serrular to those recorded by the U.S. Natonal Seismc Network to perform stand alone calculatons for typical prototypic condi-and local stabons operated by academic and research institu_

tons. The results of the expenments that were conducted using tions. The main components of the system are time senes proc-as-water are preseded.

essing, event formabon, event classifcaton, and hazard as.

NUREG/CR4269: A PLAN FOR THE MODIFICATION AND AS-sessment. These fully independent modules are complemented SESSMENT OF TRAC-PF1/ MOD 2 FOR USE IN ANALYZ1NG by operatonal and research databases. Signal detection, onset CANDU 3 TRANSIENT THERMAL-HYDRAULIC PHENOMENA.

time estimation and signal charactenzaten are functons of the time senes module. Signal associaton and event location are SIEBE,D.A.; BOYACK,B.E.; GIGUERE,P.T. Los Alamos National performed in tha event formation module. The seismic source is Laboratory. November 1994. 73pp. 9412130154. LA-12853-MS.

82008:048 charactenzed in the event classifcaton module. The seismic hazard assessment module estimates basc parameters for use This report presents the results of the review and planning in seisme nsk analysis. The operatonal database acts as the done for the United States Nuclear Regulatory Commission to central facahty through whch access to the seisme data and identrty the thermal-hydraulc phenomena that could occur in communicaton between the system's modules are accom-the CANDU 3 reactor design dunng transient conditions, plan modif^ ations to the TRAC-PF1/ MOD 2 (TRAC) computer code c

plished. The functon of the research database is to provide a seisme bulletin and segmented waveforms for events.

needed to adequately predict CANDU 3 transient thermal-hy-draulic phenomena, and identify an assessment program to NUREG/CR4254: THE LIQUEFACTION METHOD FOR ASSESS-venty the abinty of TRAC, when modified, to predet these phe-ING PALEOSEISMICITY. TUTTLE,M.P. Lamont Doherty Earth nomena. This work builds on analyses and recommendatons Observatory. December 1994. 47pp. 9501180164. 82348.262.

produced by the Idaho Natonal Engineering Laboratory (INEL).

Paleoseismesty studies expand our knowledge of setsme ac-To identify the thermal-hydraulc phenomena, a large-break trvity into the prehtstonc penod and thereby can improve our un-loss-of-coolant accident simulation, performed as part of earter derstanding of the earthquake potential of various regens. Pa-work by INEL with an Atomc Energy of Canada, Lirruted (AECL) looseistnology is proving especially useful in eastern North thermal-hydrauhc computer code (CATHENA), was anatyred in i

48 Main Citations and Abstracts detail Other accident scenarios were examined for additional NUREG/CR4281: A SIMPLIFIED LEAK-BEFORE-BREAK EVAL-phenomena. A group of Los Alamos National Laboratory reactor UATION PROCEDURES FOR AUSTENITIC AND FERRITIC thermal-hydrauhes experts ranked the phenomena to produce a STEEL PIPING. GAMBLE.R.M.; ZAHOOR A.: GHASSEMI.B. No-preliminary phenomena identification and ranking table (PIRT).

vetech Corp. October 1994. 48pp. 9411280283. 81851:226.

This PIRT provided an adequate foundation for planning a pro-A simphfied procedure has been defined for computing the al-gram of code modifications. A plan for code modificatons was towable circumferential throughwall crack length as a function of developed based on this PIRT and on information about the applied loads in piping. This procedure has been defined to modehng methodologies for CANDU-specific phenomena used enable leak-before-break (LBB) evaluatons to be AHW in AECL codes. AECL thermal-hydraulic test facihties and pro-without complex and time consumeng analyses. The develop-grams were reviewed and the informaton used in developing an ment of the LBB evaluation procedure is similar to that now (ssessment plan to ensure that TRAC-PF1/ MOD 2. when modi-used in Section XI of the ASME Code for evaluaten of part-fied, will adequately predet CANDU 3 phenomena.

throughwall flaws found in piping. The LBB evaluaton proce-dure was bench marked using experimental data obtained from NUREG/CR4270 DRF FC: ESTIMATING BOILING WATER RE-pipes having circumferential throughwall flaws. Comparisons of ACTOR DECOMMISSIONING COSTS.A User's Manual For The the expenmental and prodcted load carrying capacities indcate BWR Cost Estimating Computer Program (CECP) Software Draft that the method has a conservative bias, such that for at least Report For Comment BIERSCHBACH,M. Battelle Memonal in-97% of the experiments the experimental load is equal to or stitute, Pacsfc Northwest Laboratory. November 1994.188pp.

greater than 90% of the predicted load.

9412300195. PNL-10086. 82160:278.

With the issuance of the Decommissioning Rule (July 27 NUREG/CR4288: GEOCHEMICAL INVESTIGATIONS RELATED 1988), nuclear power plant hcensees are required to submit to TO THE YUCCA MOUNTAIN ENVIRONMENT AND POTEN-the U.S. Regulatory Commesson (NRC) for review, decommis-TIAL NUCLEAR WASTE REPOSITORY, MURPHY,W.M.;

sioning plans and cost estimates. This user's manual and the PABALAN,R.T. Center for Nuclear Waste Regulatory Analyses.

accompanying Cost Estimating Computer Program (CECP) soft-November 1994. 200pp. 9412300218. 82161:098.

ware provide a cost-calculating methodology to the NRC staff Research is reported which focused on expenmental determi-that will assist them in assessing the adequacy of the hcensee nations and theoreticalinterpretatons of fundamental thermody-submittals. The CECP designed to be used on a personal com-namic and kinetc properties of minerals and reactons that puter, provides estimates for the cost of decommissioning BWR characterize geochamcal processes at the proposed nuclear power statons to the point of license terminaton. Such cost es-waste repository site at Yucca Mountain and that could affect timates include component, piping. and eqaipment removal the capacity of the site to isolate nuclear waste. Cation ex-costs; packaging costs; decontaminaten cests; transoortation change equikbna were determined for clinoptilohte and binary costs; bunal costs; and manpower costs. In additon to costs.

solutions of NA(+) with K(+), Ca(2+), and Sr(2+). Results the CECP also calculates burial volames, person-hours, crew-were interpreted using a Margules solid solution model Dissolu-hours, and exposure person-hours associated,J..h de,omrnes' tion rate data for analcime can be rationalized by mechanisms soning-whch invoke either ultrareactive material or rate dependence NUREG/CR4276: QUALITY MANAGEMENT IN REMOTE AF.

on aqueous aluminum. Solubihty determinatens were interpret.

TERLOADING BRACHYTHERAPY.

TORTORELLI,J.P.;

ed to obtain equihbnum constants for dissoluton reactions of SIMION,G.P.; KOZLOWSKl S.D. EG&G Idaho. Inc. October analcime and of Na-chnoptilohte. Chemical analytcal data indi-1994. 93pp. 9412130077. EGG-2746. 82006:211.

cate that natural groundwaters at Yucca Mountain appear to be Over a dozen govemment and professional organizations in at equilibnum with analcime. Employing pnnciples of thermody-the United States and Europe have issued regulations and guid.

names, reaction kinetes, rnass transfer, an mass transport, ance concerning quahty management in the practice of remote computatonal models were developed using EO3/6 to explore afterloading brachytherapy. Informator, from the pubications of the evoluton of the natural geochemical system, and the geo-these organizations was collected vid collated for this report.

chemical processes that are hkely to occur under repository This report provides the brachytheiapy licensee access to a conditons.

broad field or quality management information in a single, tops-NUREG/CR4289: RECONCENTRATION OF RADIOACTIVE MA-cany organzed document.

TERIAL RELEASED TO SANITARY SEWERS IN ACCORD-NUREG/CR4278: SURVEY OF INDUSTRY METHODS FOR ANCE WITH 10 CFR PART 20. AINSWORTH,C.C.; HILL,RL; PRODUCING HIGHLY RELIABLE SOFTWARE.

CANTRELL,K.J.; et al. Battelle Memonal Institute, Pacific North-LAWRENCE,J.D.; PERSONS WL Lawrence Livermore National west Laboratory. December 1994. 200pp. 9412300213. PNL-Laboratory. November 1994. 47pp. 9412130117. UCRL-ID.

10193. 82161:294.

117524. 82008:001.

The U.S. Nuclear Regulatory Commission (NRC), in accord-The Nuclear Reactor Regulaton Offee of the U.S. Nuclear ance with 10 CFR 20, and agreement states, in accordance Regulatory Comrnisson is charged with assessing the safety of with state regulations, regulates the discharge of radcactive new instrument and control designs for nuclear power plants materials into sarvtary sewer systems. A one-year study was whch may use computer-based reactor protecten systems.

conducted by Pacife Northwest Laboratory (PNL) for the NRC i

Lawrence Livermore National Laboratory has evaluated the to assess whether radcactive materials that are descharged to l

latest techruques in software rehability for measurement, estuna-sanitary sewer systems undergo signifcant reconcentraton l

ton, error detecton, and predction that can be used dunng the within the wastewater treatment plants (WWTP) and to deter.

software hfe cycle as a means of nsk assessment for reactor mine the physcal and/or chemical process that may result in protection systems. One aspect of this task has been a survey radionuchde reconcentration within the WWTPs. The study ob-of the software industry to collect information to help identify jectives were addressed by collect #ng informaton and data on the design factors used to improve the rehabikty and safety of wastewater treatment, relevant geochemical processes, and in-software. The intent was to discover what practces really work dividual radionuchde behavior in WWTPs from the open htera-in industry and what design factors are used by industry to ture, NRC reports, EPA surveys, and interviews with NRC h-achieve highly rehable software. The results of the survey are censees and staff of WWTPs that may be impacted by these documented in this report. Three companies partcipated an the discharges. Redonuclide mass balance and removal effcien-survey-Computer Sciences Corporaton, Internatonal Business coes were calculated for WWTPs at Oak Ridge, TN; and Erwin, Machines (Federal Systems Company), and TRW. Discussions TN, but were not shown to be rehable since the licensee re-were also held with NASA Software Engineenng Lab, University lease data generally underestimated the mass of redonuchde of Maryland, CSC, and the AIAA Software Rehability Propect.

that was ultimately found in the sludge. This drapanty may be i

1

Main Citations and Abstracts 49 4

due, in part, to the fact that data available for use in this study NUREG/GR-0008: VALIDATION OF SEISMIC PROBABILISTIC were collected to address regulatory concems and not to per-RISK ASSESSMENTS OF NUCLEAR POWER PLANTS.

form mass balance calculatons. A hmited modeling study ELLINGWOOD.B. Johns Hopkins Univ., Baltimore, MD. January showed some promise for predcting radionuchde behavior in 1994.134pp. 9403150302. 78501:001.

1 WWTPs, however, the general applicability of using these em-A seismic probabiliste nsk assessment (PRA) of a r xlear j

pincal models remains uncertain. With the data and models cur-plant requires identification and information regarding the seis-rently available, it is not possible to quantitatively determine the mic hazard at the plant site, dominant accident sequences lead-physical and chemical processes that cause reconcentraton or ing to core damage, and structure and equipment fragilities. Un-to calculate, a pnon, reconcentration factors for specific WWTP unit processes or WWTPs in general.

certainties are associated with each of these ingredients of a PRA. The sensitivity of accident sequence probabilities and NUREG/CR-6290:

KEY ANALYS!S SYSTEM USER'S high-confidence, low probability of failure (HCLPF) plant fragih-GUIDE. Version 2.0. MASSE,R.P. Intenor, Dept. of, Geologcal ties to seismic hazard and fragibty modehng assumptions was Survey. November 1994.146pp. 9412070007. 81956:001.

examined for three nuclear power plants. Mean accident se-The KEY analysis system is a software program designed to quence probabihties were found to be relatively insensitive (by a process digital waveform data from the United States Natonal factor of two or less) to: uncertainty in the coefficient of vana-Seismograph Network. The KEY system performs many data tion (loganthmic standard deviation) desenbing inherent random-processing and scientifc analysis functons. Detailed operating ness in component fragility; truncation of lower tail of fragibty; procedures for the KEY analysis system are provided in this uncertainty in random (non-seismic) equipment failures (e.g.,

User's guide.

diesel generators); correlation between component capacities; and functonal form of fragility family. Accident sequence prob-NUREG/CR-6294: DESIGN FACTORS FOR SAFETY-CRITICAL abilities, expressed in the form of a frequency distribution, are SOFTWARE. LAWRENCE.J D.; PRECKSHCT G G. Lawrence affected signifcantly by the seismic hazard modehng. When the Lsvermore Natonal Laboratory. December 1994. 22pp.

9412300220. 82162:323.

fragibty modeling and plant logic are effectively uncoupled from This report is the fourth of a senes of reports prepared for the the seismc hazard analysis in a seismic margin study, the influ-Nuclear Regulatory Commission Offee of Nuclear Reactor Reg-ence of the large uncertainty in the seismic hazard is eliminal-ulation, and provides the summary and concluson for this task.

ed. In seismic margin studies, uncertainties in fragibty modehng It is widely believed in the software engineenng community that assume a different si9nifcance than they do in seismic PRA almost anything can affect the abihty of software to tehably per.

Studies Gross design and construction errors and their impact form its tasks, partcularly when safety is at issue. While this on estimates of seismic nsk are dealt with in approximation by statement is true, both in the abstract and in specific instances, postulating vanous error scenanos and their effect on compo-it is not particulady helpful It remains neceur, for auditors nent fragihties and recalculating the core damage probabikties.

and other reviewers to assure themselves a.nd the pubhc that Plausible design /constructon errors appear to have httle effect safety-cntcal software has sufficiently low probability of taihng in on mean core damage probabiaties.

such a way as to cause death or injury to perrrt it to be used in safety-cntcal apphcations. Achieving this ar.surance is best NUREG/GR-0011: INFORMATION BLAS AND UFETIME MOR-done by using a well-planned, methodcal apitoach. A possible TALITY RISKS OF RAD!ATION-INDUCED CANCER. Low LET approach is to concentrate on those attnbutis of the software Radiation. PETERSON LE.; SCHULL,WJ.; DAVIS,B.R.; et al.

and the development process (desgn factors) that are most n-Texas, Univ. of. Houston, TX. April 1994.182pp. 9405310162.

fluential in achieving dependable software. Seventy-four design 7959t163.

factors are identified in this report, divided into nine categones, Additive and mutJ;166catrve models of relative risk were used Seven categones relate to the development process, and one to measure the effect of cancer miscfassification and DS86 category relates to the products of that process. The remaining random errors on Irfetime nsk protections in the Ufe Span Study category contains negative factors whose presence should be (LSS) of Hiroshima and Nagasaki atome bomb survivors. The regarded as cause for intense scrutiny of the development proc-true number of cancer deaths in each stratum of the cancer ess. Seven of the design factors should be considered manda-tory for any organization responsible for developing safety-cnti-mortality cross-classifcaten was eshmated us#ng sufficient sta-tistcs from the EM algonthm. Average survivor doses in the cal software. An additional nine factors are considered essential strata were corrected for DS86 random error (u=0.45) by use to safety, but not as important as the first seven. The remaining of reduchon factors. Poisson regreeson was used to model the desgn factors can provide additional important indications of corrected and uncorTected mortality rates with covenates for the quality of the development effort and the software resulting age at-time-oftombing. age at-twne-of-death and gender from that eHort.

Excess naks were in good agreement with noks in RERF Report NUREG/CR-6303: METHOD FOR PERFORMING DIVERSITY 11 (Part 2) and the BEIR-V Report. Bias due to DS86 random AND DEFENSE-IN-DEPTH ANALYSES OF REACTOR PRO.

error typcally ranged from -15% to -30% for both sexes, and all TECTION SYSTEMS. PRECKSHOT,G.G. Lawrence Uvermore S8tes and models. The total bias, including diagnoste misclassifi-National Laboratory. December 1994. 45pp. 9501180332.

cation, of excess nsk of nonieukemia for exposure to 1 Sv from UCRL lD-119239. 82347:226.

age 18 to 65 under the non-constant relative protection model The purpose of this NUREG is to desenbe a method for ana-was 37.1% for males and -23.3% for females. Total excess lyzing computer-based nuclear reactor protection systems that risks of leukemia under the relative projection rnodel were discovers design vulnerabihties to common-mode failure. The biased -27.1% for males and -43 4% for females. Thus, nonieu-potential for commorkmode failure has become an important kemia nsks for 1 Sv from ages 18 to 65 (DRREF=2) increased issue as the software content of protection systems has erk from 1.91%/Sv to 2.68%/Sv among males and from 3.23%/Sv creased. This potential was not present in earlier analog protec-to 4.02%/Sv among females. Leukemia excess nsks increased tion systems because it could usuany be assumed that from 0.87%/Sv to 1.10%/Sv among males and trom 0.73%/Sv j

common-mode failure, if it did occur, was due to slow process-to 1.04%/Sv among females. Bias was dependent on the i

es such as corrosion or premature wear-out. This assumption is gender, site, correction method, exposure profile and projection I

no longer true for systems containing software. It is the purpose model considered Future studies that use LSS data for U.S. nu-l of the analysis method desenbed here to determine points of a clear workers may be downwardly biased if lifetime nsk protec-l design for whch credible common-mode failures are uncompen-tions are not adjusted for random and systemate errors.

sated either by diversity or defense-in-depth.

l

50 Main Citations and Abstracts NUREG/GR 0013: APPLICATIONS OF A NEW MAGNETIC MON-seals. The code results are compared to experimental load ITORING TECHNIOUE TO IN SITU EVALUATION OF FATiOUE measurements performed at the Combustion Engineering Labo-DAMAGE IN FERROUS COMPONENTS.

JILES,0.C.;

rutory in Windsor (U.S.A.). Those measurements were part of BINER.S B.; GOVINDARAJU M.; et al. Iowa State Univ., Ames, the PWR Valve Test Program undertaken by EPRI after the IA. June 1994. 41pp. 9407250286. 80328:195.

TMI 2 accident. This particular kind of transients challenges the The work undertaken in this project consisted of research into apphcabihty of the following code models: two-phase choked the use of magnetic inspection methods for the estimation of fa-discharge; interphase drag in conditions with large density gradi-t9ue hfe of nuclear pressure vessel steel The rationale for this ents; heat transfer to metallic structures in fast changing condi-E Magnete materials are closely mterrelated, and therefore the work was that the mecharucal and magnetic properties of ferro-tions; and two-phase flow at abrupt expansions.

NUREG/lA-0114: ASSESSMENT OF RELAP5/ MOD 3 WITH THE measurements of the magnete properties could be used t LOFT L9-1/L3-3 EXPERIMENT SIMULATING AN ANTICIPAT-morntor the evolubon of fabgue damage in these specimens as ED TRANSIENT WITH MULTIPLE FAILURES. BANG,Y.S.;

they were subjected to cycle loading. The results of the work SEUL,K.W.; KIM,H.J. Korea Institute of Nuclear Safety. February have shown that it ts possible to monitor the fabgue damage 1994.112pp. 9404010254. ICAP00196. 78728:155.

nondestructrvely by magnetc techruques. For example, m load-The RELAPS/ MOD 3 Sm5 code was assessed using the L9-1/

controlled high-cycle fabgue tests, it has been found that the L3-3 test carried out in the LOFT facility, a 1/60-scaled experi-plashc stram and coercmty accumulate loganthmically during mental reactor, simulating a loss of feodwater accident with the fabgue process. Thus a quantitabve relatonship between multiple failures and the sequentially-induced small break loss-coercivity and the number of fabgue cycles could be estabhshed of coolant accident. The code predictability was evaluated for based on two empercal coeffcients, which can be determined the four separated sub-penods with respect to the system re-from the test condebons and matenal properties Also it was s onse; irutial heatup phase, spray and PORV cycling phase, found that prediction of the onset of fatigue failure in steels was blowdown phase and recovery phase. Based on the cocnpari-possible under certain conditions. In strain-controlled low cycle sons of the results from the calculation with the experiment fabgue, entical changes in Barkhausen emissions, coercevity and data, it is shown that the overall thermal-hydraulic behavior irn-eres loss occurred in the last ten to twtaty percent of fa.

ortant to the scenario such as a heat removal between the pr>

mary side and the secondary side and a system depressunza-NUREG/lA 0093: RELAPS/ MOD 3 ASSESSMENT FOR CALCU-tion was well-predcted and that the code could be apphed to LATION OF SAFETY AND RELIEF VALVE DISCHARGE the full-scale nuclear power plant for an anticipated transient PIPING HYDRODYNAMIC LOADS.

ETUBBE.E.J.;

with multiple failures within a reasonable accuracy. The minor VANHOENACKER,L; OTERO,R. TRACTEBEL Febrerv 1994.

discrepancies between the predetion and the expenment were 300pp. 9405040180. ICAP00 t S9. 79105:001.

identified in reactor scram time, post-scram behavior in the ini-This report presents an assessment study for the use of the bal heatup phase, excessive heatup rate in the cychng phase, code RELAP5/ MOD 3/$M5 in the calculation of transient hydro-insuffeient energy convected out the PORV under the hot leg dynarmc loads on safety and rehef discharDe pipes. Its prede-stratified condition in the saturated blowdown phase and void cessor, RELAPS/ MODI, was found adequate for this lund of distribution in secondary side in the recovery phase. This may calculabon by EPRI. The hydrodynamic loads are very important come from the code uncertainbes in predcting the spray mass for the discharge piping desegn because of the fast opening of flow rate, the associated condensation in pressurt2er and junc-the valves and the presence of hquid in the upstream loop tion fluid density under stratified condsbon.

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 repo-t and to the 10-digit NRC Document Control System accession number, t

i SECONDARY REPORT NUMBER REPORT NUMBER SECONDA3Y REPORT NUMSER REPORT NUMSER r

04-4448 Y1T1 NUREG/CR4074 V01 EGG-2715 NUREG/CR4245 ANL 93/22 NUREG/CR-4513 R01 EGG-2716 NUREG/CR4116 V01 ANL 93/32 NUREG/CR4133 EGG-2716 NUREG/CR4118 V02 ANL-931'15 16,MG/CR4142 EGG-2716 NUREG/CR4116 V03 ANL-94/1 NUREG/CR4176 EGG 2716 NUREG/CR4116 V04 ANL-94/13 NUREG/CR4223 EGG-2716 NUREG/CR4116 VOS l

ANL-94/16 NUREG/CR-4667 V17 EGG 2716 NUREG/CR4116 V07 ANL-94/18 NUREG/CR4168 EGG 2716 NUREG/CR4116 VOS ANL-94/2 NUREG/CR4177 EGG-2717 NUREG/CR4121 ANL-94/20 NUREG/CR4236 EGG-2719 NUREG/CR4138 ANL-94/21

4UREG/CR4237 EGG-2721 NUREG/CR4160 ANL 94/22 NUREG/CR4234 EGG-2722 NUREG/CR4164 ANL 94/3 NUREG/CR4183 EGG-2730 NUREG/CR4188 V01 ANL-94/5 NUREG/CR4185 EGG-2731 NUREG/CR4194 1

ANL-94/8 NUREG/CR4187 EGG-2732 NUREG/CR4195 ANL/EES-T4364 NUREG/CR-5344 RO1 EGG-2734 NUREG/CR4197 BML2164 NUREG/CR 5128 R01 EGG-2735 NUREG/CR4198

)

BML2173 NUREG/CR-4599 V03 N2 EGG-2736 NUREG/CR4201 BML2176 NUREG/CR4226 EGG-2742 NUREG/CP 0137 V02 BML2177 NUREG/CR4233 V01 EGG-2742 NUREG/CP-0137 V01 s

BML2178 NUREG/CR-6234 EGG-2746 NUREG/CR4276 BNL-NUREG-51581 NUREG/CR-2907 V12 ICAP00159 NUREG/lA 0093 1

BNL NUREG-51934 NUREG/CR4409 VOS ICAP00196 NUREG/tA4114 BNL-NUREG-52288 NUREG/CR-5726 IS-5103 NUREG/CR4161 BNL NUREG-52309 NUREG/CR-5812 KFK/PHDR 99E-91 NUREG/CR4236 l

BNL-NUREG-52319 NUREG/CR4850 LA 12649-MS NUREG/CR4104 BNL-NUREG-52333 NUREG/CR-5908 V02 LA 12715-MS NUREG/CR4157 BNL NUREG-52333 HUMG/CR 5908 V01 LA 12741-M NUREG/CR4180 BNL NUREG-52345 NUREG/CR-5939 LA-12853-MS NUREG/CR4269 BNL NUREG-52353 NUREG/CR 5967 NEA-CSNLR(94)3 NUREG/CR4160 BNL NUREG-52359 NUREG/CR4990 NEA/CSNt/R(93)8 NUREG/CP-0127 BNL NUREG-52363 NUREG/CR-5994 NEA/CSNI/R(94)2 NUREG/CR4193 BNL-NUREG-52380 NUREG/CR 6053 NIST SP 500-216 NUREG/CP-0136 BNL NUREG-52385 NUREG/CR4086 ORNL-6698 NUREG/CR-5625 BNL-NUREG-52387 NUREG/CR 6094 ORNL4796 NUREG/CR4205 BNL NUREG-52388 NUREG/CR 6093 ORNL4811 NUREG/CR4231 BNL-NUREG 52390 NUREG/CR-6087 ORNL4820 NUREG/CR-5247 V01 R2 BNL-NUREG-52391 NUREG/CR4105 OnNUNOAC 232 NUREG/CR4674 V17 BNL NUREG-52394 NUREG/CR4112 DRF FC ORNL/NOAC-232 NUREG/CR-4674 V18 BNL NUREG 52396 NUREG/CR4128 ORNUNOAC.232 NUREG/CR 4674 V19 BNL NUREG-52399 NUREG/CR4144 V02P1 A ORNUNOAC-232 NUREG/CR-4674 V20 BNL-NUREG-52399 NUREG/CR4144 V02PIB ORNUTE10128 NUREG/CR4816 R02 BNL-NUREG-52399 NUREG/CR4144 V02P2 ORNL/TM-11)$2 NUREG/CR-5359

~

BNL NUREG-52399 NUREG/CR4144 V02P3A ORNL/TE11568 NUREG/CR-5591 V02 N1 BNL-NUREG-52399 NUREG/CR4144 V02P3B ORNUTM11568 NUREG/CR-5591 V02 N2 BNL-NUREG42399 NUREG/CR4144 V02P4 ORNL/TE12067 NUREG/CR-5569 ROI BNL-NUREG-52399 NUREG/CR4144 V02P5 ORNL/TE12164 NUREG/CR-5904 BNL NUREG-52399 NUREG/CR4144 V03 P1 ORNL/TE12221 NUREG/CR-5941 BNL-NUREG-52399 NUREG/CR4144 V03 P2 ORNL/TM-12415 NUREG/CR4076 BNL-NUREG-52399 NUREG/CR4144 V04 ORNUTM-12416 NUREG/CR4077 BNL-NUREG-52400 NUREG/CR4146 ORNUTE12460 NUREG/CR4102 BNL NUREG-524G8 NUREG/CR4169 ORNL/TE12498 NUREG/CR4132 BNL-NUREG-52409 NUREG/CP-0135 ORNL/TW12513 NUREG/CR4139 BNL-NUREG-52412 NUREG/CR4200 ORNL/TE12663 NUREG/CR4182 V01 BNL-NUREG42413 NUREG/CR4212 ORNUTE12663 NUREG/CR4182 V02 BNL-NUREG-52414 NUREG/CR4213 ORNL/TE12681 NUREG/CR4193 BNL-NUREG42422 NUREG/CR4241 ORNUTM12690 NUREG/CR4204 BSRC-700/94/016 NUREG/CR-5758 V04 ORNUTM12693 NUREG/CR4206 BSRC400/94/014 NUREG/CR-6252 ORNUTE12735 NUREG/CR4228 CNWRA 92 007 NUREG/CR-5919 ORNUTE12777 NUREG/CR4249 CNWRA 93 013 NUMEG/CR4178 OANL/TE9593 NUREG/CR4219 V*a N1 CNWRA 93 024 NUREG/CR4216 ORNLSUB79-77789 NUREG/CR 5861 DCS-94-024 NUREG/CR4255 ORNLSUB79777811 NUREG/CR4262 EGG.2458 NUREG/CR-4439 V5R4P3 PNL 10086 NUREG/CR4270 DRF FC EGG-2458 NUREG/CR4639 V5R4P2 PNL 10193 NUREG/CR4289 EGG-2562 NUREG/CR 5314 V05 PNL-4221 NUREG/CR-2850 V12 EGG 2577 NUREG/CR4229 V06 PNL-5210 NUREG/CR-3950 V09 EGG-2596 NUREG/CR-5535 V07 PNL4462 NUREG/CR-5161 V02 EGG-2644 NUREG/CR4158 PNL 7784 NUREGICR 5830 EGG-2686 NUREG/CR-5935 PNL4428 NUREG/CR4122 EGG-2705 NUREG/CR-5535 V06 PNL4462 NUREG/CR-5973 RO1 EGG-2707 NUREG/CR4088 PNL4564 NUREG/CR4123 EGG-2713 NUREG/CR4145 PNL-8844 NUREG/CR-5963 3

51 4

52 Secondary Report Number index SECONDARY REPORT NUMSER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUM8ER PNL4912 NUREG/CR4151 SAND 93-2440 NOREG/CR4143 V02PT2 PNL4919 NUREG/CR-5985 SAND 93-2440 NUREG/CR 6143 V02PT3 PNL-9020 NUREG/CR4181 SAND 93-2440 NUREG/CR4143 V02PT4 PNL-9750 NUREG/CR4232 SAND 93 2440 NUREG/CR4143 V03 SAND 93 2440 NUREG/n4143 V04 PNL 9671 NUREG/CR4252 SAND 93 2478 NUREGX 44149 PNL-9985 NUREG/CR-5758 VD4 SAND 93-2519 NUREG/CR4152 PU NE-93/1 NUREG/CR4267 SAND 93-7107 NUREG/CR4103 SAND 66-1309 NUREG/CR-4551 V01 Rg SAND 94-0406 NUREG/CR4211 SAND 68-1887 NUREG/CR-4838 SAND 93-0234 NUREG/CR-5407 3

SAND 93-0971 NUREG/CR4042 SAND 941711 NUR6G/CR4154 V01 SAND 911049 NUREG/CR4044 SEA 93 55446-A1 NUREG/CR4224 DFC SAND 93-1535 NUREG,R4075 S01 TMI V 92)EG01 NUREG/CR4194 SAND 931535 NUREG/CR4075 TMI V 92)EG10 NUREG/CR4195 SAND 93-1535 NUREG/CR4075 S01 TMl V 93)EG01 NUREG/CR4196 SAND 911535 NUREG/CR4075 TMI V(93)EG10 NUREG/CR4197 SAND 93-1737 -

NUREG/CR 4092 TMI V(94)EG01 NUREG/CR4198 NUREG/CR4095 TMIV(93)ALO1 NUREG/CR 4185 E

SAND 93-1803 SAND 931804 NUREG/CR4093 TMIVt93)ALO2 NUREG/CR4187 SAND 93-2042 NUREG/CR4107 UCRL-ID-117524 NUREG/CR4278 SAND 93 2440 NUREG/CR4143 V02P1 A UCAL-ID-118245 NUREG/CR-5403 SAND 93-2440 NUREG/CR4143 V02PIB UCRL-ID-119239 NUREG/CR4303 SAND 93-2440 NUREG/CR4143 V02P1C UILU-ENG93-2014 NUREG/CR4162 l

I 4

4 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-tion by the NUREG number.

1 i

AABERG,RL AMARASOORlVA,W.

NUREG/CR4289: RECONCENTRATON OF RADIOACTIVE MATERIAL NUREG/CR-5960: STEAM EXPLOSIONS: FUNDAMENTALS AND EN-RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR GERGETIC BEHAVIOR.

PART 20.

AMONETTE,J.E.

ABU-ELD,R.

NUREG/CR4232: ASSESSING THE ENVIRONMENTAL AVAllABILITY NUREG 1486: FINAL SAFETY EVALUATON REPORT TO UCENSE OF URANIUM IN SOILS AND SEDIMENTS.

THE CONSTRUCTON AND OPERATON OF A FACILITY TO RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL ANDERSON.BL NEAR CLIVE UTAH. Docket No. 404989 (Envrocare of Utah,inc.)

NUREG/CR-5403: PREDICTING THE PRESSURE DRIVEN FLOW OF GASES THROUGH MICRO-CAPlLLARIES AND MICRO-ORIFICES.

AHMAD,J.

RAT E ATO L

N RE CR4162: EFFECTS OF PRIOR DUCTILE TEARING ON CLEAV-AGE FRACTURE TOUGHNESS IN THE TRANSITION REGION.

AHOLA,M.P.

NUR G/CR4178. LABORATORY CHARACTERIZATION OF ROCK A

UR R-5985: EVALUATON OF COMPUTER-BASED ULTRASONIC NUREG/CR4216. EVALUATON OF ROCK JOINT MODELS AND COM.

INSERVICE INSPECTION SYSTEMS.

PUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS ANGELINI,5.

AINSWORTH.C.C.

NUREG/CR 5960: STEAM EXPLOSONS: FUNDAMENTALS AND EN-NUREG/CR4289 RECONCENTRATION OF RADIOACTIVE MATERIAL GERGETIC BEHAVOR.

RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR ART M ANKRUM A.R.

NUREG/CR-5973 R01: CODES AND STANDARDS AND OTHER GUID-AKERS,0.W.

ANCE CITED IN REGULATORY DOCUMENTS.

NUREG/CR4160:

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP 2-APOSTOLAKIS.G.

NUREG/CR4164 RELEASE OF RADIONUCUDES AND CHELATING NUREG/CR4157: SURVEY AND EVALUATION OF AGING RISK AS.

AGENTS FROM CEMENT-SOUDIFIED DECONTAMINATION LOW-S S M Mmpi LEVEL RADOACTIVE WASTE COLLECTED FROM THE PEACH BOTTOM ATOMC POWER STATON UNIT 3.

ASFURA.A.P.

NUREG/CR4195: EXAMINATON OF RELOCATED FUEL DEBRIS AD.

NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DEGRADED JACENT TO THE LOWER HEAD OF THE TMt 2 REACTOR VESSEL SHEAR WALL STIFFNESSES ON SEISMIC PLANT RISK AND SEIS-NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE.

MIC DESIGN LOADS ~

GRATON REPORT.

ASGAR1,M.

NUREG/CR4198 TML2 INSTRUMENT NOZZLE EXAMINATIONS PER*

NUREG/CR4206: TRANSPORT CALCULATIONS OF RADIATION EX-NUR G/

1:

MPRESSION ANO IMMERSON TESTS AND LEACHING OF RADIONUCUDES. STABLE METALS AND CHELATING CT '

AGENTS FROM CEMENT-SOUDIFIED DECONTAMINATION WASTE ATHEY,G.F.

COLLECTED FROM NUCLEAR POWER STATIONS.

NUREG/CR-5247 V01 R2: RASCAL VERSION 2.1 USER'S GUIDE.

NUREG/CR-5247 V02 R2: RASCAL VERSION 2.1 WORKBOOK.

NUREG/CR-6044; EXF".RIMENTS TO INVESTIGATE DIRECT CON-ATWOOD,CL TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NUREG/CR4116 VO1: SYSTEMS ANALYSIS PROGRAMS FOR ZON NUCLEAR POWER PLANT IN THE SURTSEY TEST FACluTY.

HANDS-ON INTEGRATED REUABlWTY EVALUATONS (SAPHIRE)

NUREG/CR-6075 S01: THE PROBABILITY OF CONTAINMENT FAIL-VERSION 5 0. Technical Fleference Manual.

UAE BY DIRECT CONTAINMENT HEATING IN ZION.

NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DIRECT CON.

AZARM,A.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NUREG/CR-5939 THE EFFECTS OF AGE ON NUCLEAR POWER SURRY NUCLEAR POWER PLANT.

PLANT CONTAINMENT COOUNG SYSTEMS.

ALLENSPACH,F.

BAIRD,0.B.

NUREG 1214 R13: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

NUREG/CR-5161 V02: EVALUATION OF SAMPUNG PLANS FOR IN-IC ASSESSMENT OF UCENSEE PERFORMANCE.

SERVICE INSPECTON OF STEAM GENERATOR ALLISON,C.M.

TUBES Comprehensive Analytical And Monte Carlo Simulation Results For Several Samphng Plans.

NUREG/CR4160

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP 2.

BAKER C.C.

NUREG/CR-5908 V02: ADVANCED HUMAN-SYSTEM INTERFACE CLLISON.D.P.

DESIGN REVIEW GUIDELINE. Evaluation Procedures And Guidehnes NUREG 1022 RO1 DR FC: EVENT REPORTING GUIDELINES For Human Factors Engineenng Reviews.

10CFR50.72 AND 50.73.Second Draft For ComrnenL B A K E R.D.A.

ALVIS J.M.

NUREG/CR-2850 V12. DOSE COMMITMENTS DUE TO RADIOACTIVE NUREG/CR-3950 V09 FUEL PERFORMANCE REPORT FOR 1991.

RELEASES FROM NUCLEAR POWER PLANT SITES IN 1990.

53

54 Personal Author index BANER K.

BEYER,C.E.

NUREG/CR4122: STAFFING DECISION PROCESSES AND NUREG/CR-3950 V09: FUEL PERFORMANCE REPORT FOR 1991.

ISSUES. Case Studies Of Seven U.S. Nuclear Power Plants.

DEZLER,P.

SANDYOPADHYAY,K NUREG/CR4128: PIPING BENCHMARK PROBLEMS FOR THE ABB/

NUREG/CH4169: RELAY TEST PROGRAM. Series il Tests. Integral Test-CE SYSTEM 80+ STANDARDtIED PLANT, ing Of Re6eys And Crcuit Breakers.

BIER $CHSACH,M.

gy,g, NUREG/CR4174 V1 DFC REVISED ANALYSES OF DECOMMISSION-NUREG/lA-0114: ASSESSMENT OF RELAP5/ MOD 3 WITH THE LOFT ING FOR THE REFERENCE BOfLING WATER REACTOR POWER L9-1/L3-3 EXPERIMENT SIMULATING AN ANTICIPATED TRANSIENT STATON. Effects Of Current Regulatory And Other Consideratione On WITH MULTIPLE FAILURES

• The Fnarnal Assurance Requirernents Of The Deconmssiorung Rule SARCHI,T.

And....

NUREG-1415 V06 NO2: OFFICE OF THE INSPECTOR NUREG/CR6174 V2 DFC REVISED ANALYSES OF DECOMMISSION-GENERALSermannual Report,0ctober 1.1993. March 31,1994.

ING FOR THE REFERENCE BOILING WATER REACTOR POWER STATION Effects Of Current Regulatory And Other Consideratons On BARNES,V.

The Fnarcel Assurance Requrements Of The Decomrmessorung Rule NUREG/CR-5600 V01: THE IMPACT OF EWIRONMENTAL CONDI-And.

TONS ON HUMAN PERFORMANCE. A Handbook Of Er,a...

.tal NUREG/CR4270 DAF FC: ESTIMATING BOluNG WATER REACTOR l

Exposures.

DECOMMISSIONING COSTS.A User's Manuel For The BWR Cost Estl-NUREG/CR 5600 V02: THE IMPACT OF ENVIRONMENTAL CONDI-rnateng Computer Program (CECP) SoftwareDraft Report For Com-TlONS ON HUMAN PERFORMANCE. A Cntical Rewsw Of The utera-rnent.

i i

ture.

BINDER,JL BARMERE,E NUREG/CR4168: DIRECT CONTAINMENT HEATING INTEGRAL EF.

NUREG/CR-6093: AN ANALYSIS OF OPERATIONAL EXPERIENCE FECTS TESTS AT 1/40 SCALE IN ZION NUCLEAR POWER PLANT DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-GEOMETRY.

DRESSING HUMAN RELIABILITY ASSESSMENT ISSUES.

S4NER,S.B.

BASS,B.R.

NUFtEG/CR4132: BIAXIAL LOADING AND SHALLOW. FLAW EFFECTS NUREG/GR 0013: APPUCATIONS OF A NEW MAGNETIC MONITOR-ING TECHNIQUE TO IN SITU EVALUATION OF FATIOUE DAMAGE ON CRACK-TIP CONSTRAINT AND FRACTURE TOUGHNESS.

NOREG/CR4226: PREUMINARY ASSESSMENT OF THE FRACTURE IN FERROUS COMPONENTS.

BEHAYlOR OF WELD MATERLAL IN FULL THICKNESS CLAD

BEAMS, SITTNER,A.

NUREG/CH 5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI-

+

BASSETT,RL TIONS ON HUMAN PERFORMANCE. A Handbook Of Envronmental NUREG/CR4203. VALIDATION STUDIES FOR ASSESSING UNSATU-Exposures.

RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

NUREG/CR-5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI-TIONS ON HUMAN PERFORMANCE. A Cntical Recew Of The uters-BAUM M I"

NUREG/CR4212: VALUE OF PUBUC HEALTH AND SAFETY ACTONS AND RADIATION DOSE AVOIDED-BLANCHAT.T.K.

NUREG/CR4044: EXPERIMENTS TO INVESTIGATE DIRECT CON-DEATY,R.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL ZON NUCLEAR POWER PLANT IN THE SURTSEY TEST FACluTY.

FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT-NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DIRECT CON-ED DEBRIS Draft For Comment.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE DEHLING.K.

SURRY NUCLEAR POWER PLANT.

NUREG-1492 DFC: REGULATORY ANALYSIS ON CRITERIA FOR THE RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE SLEY,0.

MATERIALDraft Report For Comment, NUREG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DEHLING,U.H.

SURRY, UNIT 1.Analy1ue Of Core Damage Frequency From Intemal s

NUREG 1492 DFC: REGULATORY ANALYSIS ON CRITERIA FOR THE Events Dunno M4 Loop Operatons Main Report (Chapters 14).

RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE NUREG/CR 6144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-MATERLALDran Report For Comrnent.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY,UNfT 1. Analyses Of Core Damage Frequency From Intemal DELTRACC E NUREG/CP 0136' PROCEEDINGS OF THE DIGITAL SYSTEMS REll-NUR CR 44 Vb E LU OF PO R ACCb ABluTY AND NUCLEAR SAFETY WORKSHOP. September 13-14 DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT 1993,Rockville Crowne Plaza Hotel,Rockvillo. Maryland.

SURRY, UNIT 1. Analyses Of Core Damage Frequency From Intemal DENAVIDES,0.

Events During Mid-Loop Operatione. Appendices A D.

NUREG/CR4143 V02 PIA EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4144 V02P3A: EVALUATON OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GRAND GULF, UNIT 1.Anatyens Of Core Damage Frequency From In-SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal temal Events For Plant Operatonal State 5 Dunng A Refuehng Events During M4 Loop Operat ons. Appendices E (Sectons E.1 E.8L NUREG/CR4144 V02P3B: EVALUATION OF POTENTIAL SEVERE AC-Outaos Sectons 10 NUREG/CR4143 V02P18: EVALUATON OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AV CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From intemat GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From in.

Events Dunng Mid-Loop Operatons. Appendices E (Sectone E.9-E.16).

NUREG/CR4144 V02P4: EVALUATION OF POTENTIAL SEVERE ACCL temal Events For Plant Operational State 5 Dunng A Refuehng DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Outage.Secten 10.

NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC.

SURRY, UNIT 1. Analyses Of Core Damage Frequency From intomal COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Events Dunng Meloop Operations. Appendices F-H.

GRAND GULF,0 NIT 1. Analysis Of Cnre Damage Frequency From in.

NUREG/CR4144 V02P5: EVALUATION OF POTENTIAL SEVERE ACCl-temal Events For Plant Operational State 5 Dunng A Refuehng DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Outage. Main Report.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From intemal Events During M4 Loop Operations. Appendices L SERK,5.

NUREG/CR4123: AN INTERNATIONAL COMPARISON OF COMMER-BLOMQUIST.C A.

CIAL NUCLEAR POWER PLANT STAFFING REGULATIONS AND NUREG/CR4133. FRAGMENTATION AND OUENCH BEHAVIOR OF PRACTICE.19801990.

CORIUM MELT STREAMS IN WATER.

Personal Author index 56 l

90ARDMAN,J.R.

NUREG/CR4144 V02P5: EVALUATION OF POTENTIAL SEVERE ACCI-NUREG-1275 V10- OPERATING EXPERIENCE FEEDBACK REPORT -

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT l

RELIABILITY OF SAFETY RELATED STEAM TURBINE-DRIVEN SURRY, UNIT 1. Analyse Of Core Damage Frequency Frorn Intemal STANDBY PUMPS. Commencal Power Reactors.

Events Dunng Mid-Loop Opershons.Appendees I.

(

90CCIOJ.

BRAMWELL,0.L.

NUREG/CR4213: HtGH-TEMPERATURE HYDROGEN-AIR-STEAM NUREG/CR-5935:

SUMMARY

OF WORK COMPLETED UNDER THE DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-i l

MENT APPARATUS.

ENVIRONMENTAL AND DYNAMC EQUIPMENT QUALIFICATION RE-l SEARCH PROGRAM (EDOP).

BODENSTEINER,J.

BRAVERMAN.J.

NUREG 1415 V07 N01: OFFICE OF THE INSPECTOR NUREG/CR4128: PIPING BENCHMARK PROBLEMS FOR THE A8B/

GENERAL $ermannual Report, April 1 -September 30,1994.

CE SYSTEM 80+ STANDARDIZED PLANT.

kMW-hUREG/CR4183. PEER REVIEW OF THE TMi-2 VESSEL INVESTIGA-BREEDING.RJ.

TION PROJECT METALLURGICAL EXAMINATIONS.

NUREG/CR-4551 V01 R1: EVALUATION OF SEVERE ACCIDENT RISKS: METHODOLOGY FOR THE CONTA!NMENT. SOURCE mg TERM. CONSEQUENCE, AND RISK INTEGRATION ANALYSES.

NUREG/CP-5407: ASSESSMENT OF THE IMPACT OF DEGRADED l

SHEAR WALL STIFFNESSES ON SEISMIC PLANT RISK AND SEIS-SRIDEAU,J NUREG/bR4224 DFC: PARAMETRC STUDY OF THE POTENTIAL NUR G 5

E lEW OF THE DIABLO CANYON PROBABILISTIC FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

RISK ASSESSMENT.

ED DEBRIS. Draft For Comment.

DOLANDER,T.W.

BROWN.C.

NUREG/CR 6145: VERIFICATION AND VALIDATION OF THE SAPHIRE NUREG-1504: REVIEW CRITERIA FOR THE PHYSICAL FITNESS VERSION 4.0 PRA SOFTWARE PACKAGE.

TRAINING REQUIREMENTS IN 10 CFR PART 73.

BOLTON,P.

BROWN,T.D.

NUREG/CR4252: LESSONS LEARNED FROM THE THREE MILE NUREG/CR-4551 V01 R1: EVALUATION OF SEVERE ACODENT ISLAND UNIT 2 ADVISORY PANEL RISKS: METHODOLOGY FOR THE CONTAINMENT. SOURCE 80NGARRA J.P.

TERM. CONSEQUENCE, AND RISK INTEGRATION ANALYSES.

NUREG/CR4143 V02P1A EVALUATION OF POTENTIAL SEVERE AC-NUREG-0711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT MODEL-GRAND GULF,0 NIT 1. Analyse Of Core Damage Frequency From In-DOROES.T.

temal Events For Plant Operatonal State 5 Dunno A Refuehng NUREG/CR-5569 R01: HEALTH PHYSCS POSITIONS DATA BASE.

Outage.Sectons 19 NURE /CR OUESTIONS AND ANSWERS BASED ON REVISED NUREG/CR4143 V02918: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-BOWMAN,S.M.

temal Events For Plant Operatonal State 5 Dunng A Refueling NUREG/CR4102 VALfDATION OF THE SCALE BROAD STRUCTURE Outage.Secten 10.

44 GROUP ENDF/B-Y CROSS-SECTION LIBRARY FOR USE IN NUREG/CR4143 V02P1C: EVALUATION OF POTENTIAL SEVERE AC-CRITICALITY SAFETY ANALYSES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4182 V01: OFFSCALE: A PC INPUT PROCESSOR FOR GRAND GULF. UNIT 1. Analyses Of Core Damage Frequency From In-THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For temal Events For Plant Operatonal State 5 Dunng A Refueling The Crmcality Sequences.

Outage. Main Report NUREG/CR4f82 V02: OFFSCALE: A PC INPUT PROCESSOR FOR NUREG/CR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE AC.

THE SCALE CODE SYSTEM. Volume 2: The ORIGNATE Processor for CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ORIGEN-S.

GRAND GULF, UNIT 1. Analyse Of Core Damage Frequency From in.

sOYACK,5.E.

temal Events For Plant Operatonal State 5 Dunng A Refuehng Outage.intemet..-

NUREG/CR-6269: A PLAN FOR THE MOOlFICATION AND ASSESS-MENT OF TRAC-PF1/ MOO 2 FOR USE IN ANALYZING CANDU 3 BROWN.W.S.

i TRANSIENT THERMAL-HYDRAULIC PHENOMENA.

NUREG/CR-5908 V02-ADVANCED HUMAN-SYSTEM INTERFACE DOZOKI,0.E.

DESIGN REVIEW GUIDELINE. Evaluaton Procedures And Guidelines For Human Factors Engineenng Reviews.

NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABILISTIC NUREG/CR4105: HUMAN FACTORS ENGINEERING GUIDANCE FOR RISK ASSESSMENT.

THE REVIEW OF ADVANCED ALARM SYSTEMS.

NUREG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4148: LOCAL CONTROL STATIONS: HUMAN ENGINEER-QDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ING ISSUES AND INSIGHTS.

SURRY,0 NIT 1. Analyse Of Core Damage Frequercy From Intamal Events Dunng Mid-Loop Operatons Marn Report (Chapters 14)

BRUMMETT,E.

NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-NUREG-1486: FINAL SAFETY EVALUATION REPORT TO LICENSE

)

j l

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT THE CONSTRUCTION AND OPERATION OF A FAOLITY TO SURRY, UNIT 1. Analyse Of Core Damage Frequency From internal RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL j

Events Dunng MLoop Operatons Main Report (Chapters 712.

NEAR CLIVE, UTAH. Docket No. 404989.(Envrocare of Utah,incJ NUREG/CR4144 V02P2 EVALUATION OF POTENTIAL SEVER ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT BRUST,F.

SURRY. UNIT 1.Analysn Of Core Damage Frequency From Intemal NUREG/CR-4599 V03 N2; SHORT CRACKS IN PIPING AND PIPING Events Dunng Mid-Loop Operations A-D.

WELDS. Semiannual Report. October 1992 - March 1993.

NUREG/CR4144 V02P3A: EVALUAT OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT BRYSON,J.W.

SURRY UNIT 1. Analyse Of Core Damage Frequency From Intemal NUREG/CR4132: BIAX1AL LOADING AND SHALLOW-FLAW EFFECTS Events Dunn0 Med-Loop Operatons.

' E (Sectons E.1-E.8L ON CRACK-TIP CONSTRAINT AND FRACTURE TOUGHNESS.

NUREG/CR4144 V02P38: EVALUAT OF POTENTIAL SEVERE AC-QDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DUCKLE.T.H.

SURRY. UNIT 1. Analyse Of Core Damage Frequency From Intamal NUREG/CR4149: APPLICATIONS OF FIBER OPTICS IN PHYSICAL Events Dunng Mid-Loop Operations. Appendices E (Sectons E.9-E.16)

PROTECTION.

NUREG/CR4144 V02P4: EVALUATION OF POTENTIAL SEVERE ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT BUDNITZ,R.J.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal NUREG/CR4143 VOS: EVALUATION OF POTENTIAL SEVERE ACCI-Events During Mid-Loop Operatsons Appendees F.H.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT l

l

56 Personal Author index GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From CHEN,Z.J.

Seesmic Events Dunng M4 Loop Operabons Main Report.

NUREG/GR4013. APPLICATONS OF A NEW MAGNETIC MONITOR-NUREG/CR4144 V05: EVALUATION OF POTENTIAL SEVERE ACC4-ING TECHNIQUE TO IN SITU EVALUATION OF FATOUE DAMAGE DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT IN FERROUS COMPONENTS.

SURRY, UNIT 1. Analysis Of Core Darnage Frequency Frorn Seesmic Events Dunng M4 Loop Operations.Mam Report CHENG,H.S.

NUREG/CR4200: UNCERTAINTY ANALYSIS OF SUPPRESSION POOL BUFFLER,P.A.

HEATING DURING AN A1WS IN A BWR-5 PLANT.An Application Of NUREG/GR-0011: INFORMATON BIAS AND LIFETIME MORTALITY The CSAU Methodology Usmo The BNL Engmeenng Plant Analyrer.

RISKS OF RADIATON-INDUCED CANCER Low LET Radiation.

CH8U,J-M.

SUMGARDNER,J.D.

NUREG/CR4209:

MEMPHIS AREA REGIONAL SEISMIC NUREG/CR-5830. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NETWORK Final R October 1986 September 1992.

i SPECTION GUIDE FOR THE MCGUtRE NUCLEAR POWER PLANT, NUREG/CR4254; THERN APPALACHIAN REGIONAL SEtSMIC 4

NETWORK.

f.YKOSKI,L NUREG-1486: FINAL SAFETY EVALUATON REPORT TO LICENSE CHtU.S.C.C.

THE CONSTRUCTION AND OPERATON OF A FACILITY TO NUREG/CR4254: SOUTHERN APPALACHIAN REGIONAL SEISMIC RECEIVE STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NETWORK.

NEAR CUVE, UTAH. Docket No. 40-8989 (Envirocare of Utah,inc.)

CALLEY,M.8.

NUREG/CR-4513 R01: ESTIMATON OF FRACTURE TOUGHNESS OF j

NUREG/CR4145: VERIFICATION AND VAllDATON OF THE SAPHIRE CAST STAINLESS STEELS DURING THERMAL AGING IN LWR SYS-j VERSION 4.0 PRA SOFTWARE PACKAGE.

NU EG CR 4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN CAMP A.L LIGHT WATER REACTORS. Semiannual Report, April 1993 Septem-NUREG/CR4042 PERSPECTIVES ON REACTOR SAFETY, ber1993.

NUREG/CR4142: TENSILE. PROPERTY CHARACTERIZATION OF CANTRELL,K.J.

THERMALLY AGED CAST STAINLESS STEELS.

NUREG/CR4289: RECONCENTRATION OF RADIOACTIVE MATERIAL NUREG/CR4177: ASSESSMENT OF THERMAL EMBRITTLEMENT OF RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR CAST STAINLESS STEELS.

PART 20.

NUREG/CR4237: STATISTICAL ANALYSIS OF FATIGUE STRAIN-LIFE DATA FOR CARBON AND LOW ALLOY STEELS.

CAPLIN,J.

NUREG/CR4156:

SUMMARY

OF COMMENTS RECEIVED FROM CHOWDHURY,A.H.

WORKSHOPS ON RADIOLOGICAL CRITERIA FOR DECOMMISSION-NUREG/CR.5919. REPOSITORY OPERATIONAL CRITERIA COMPARA.

ING.

TIVE ANALYSIS.

NUREG/CR4250

SUMMARY

OF COMMENTS RECEIVED ON STAFF NUREG/CR-6178 LABORATORY CHARACTERIZATON OF ROCK DRAFT PROPOSED RULE ON RADIOLOGICAL CRITERIA FOR DE.

JOINTS.

NUREG/CR4216: EVALUATON OF ROCK JOINT MODELS AND COM-COMMISS ONING.

PUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS.

CAPPS,E.L NUREG/CR4145: VERIFICATON AND VALIDATON OF THE SAPHIRE CHU,T.L VERSION 4.0 PRA SOFTWARE PACKAGE.

NUREG/CR4144 V02P1 A: EVALUATON OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT CAReOetEAU,M.L SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intemal NUREG/CR4160.

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

Events Dunno M4 Loop Operations Man Report (Chapters 94).

RELAP5 ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP 2.

NUREG/CR4144 V02P18: LVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT CARDtLE.F.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intemal NUREG-1500: WORKING DRAFT REGULATORY GUIDE ON RELEASE Events Durmg M4 Loop Operatons Man Report (Chapters 7-12).

CRITERIA FOR DECOMMISSONING. NRC STAFF'S DRAFT FOR NUREG/CR4144 V02P2; EVALUATION OF POTENTIAL SEVERE ACCI-COMMENT.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY UNIT 1. Analyses Of Core Damage Frequency From Intemal NU 1

ASSESSMENT OF DATABASES AND MODELING CAPA-NU CR4 4 V02 A.

T POT NTIAL SEVERE AC-BILITIES FOR THE CANDU 3 DESIGN.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT O***9*

'4"*"CY CARLSON,R.W NUREG/CR$403: PREDICTING THE IPESSURE DRIVEN FLOW OF NUR C 4744 V02P3B N

POTE A E C-GASES THROUGH MICRO-CAPILLARIES AND MICRO ORIFICES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY.UNlT 1. Analysis Of Core Damage Frequency From Intemal l

CARROLL.D.P NUREG/CR 5990 THE EFFECTS OF SOLAR-GEOMAGNETICALLY IN-NUF C

4 P E ION E lAL ERE DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT POWER STATIONS.

l SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intamal Events Dunng M4 Loop Operations. Appendices F-H.

C1RTER.D NUREG/CR-6144 V02P5: EVALUATON OF POTENTIAL SEVERE ACCl-i l

NUREG/CR-5569 RO1: HEALTH PHYSICS POSITONS DATA BASE.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4204. QUESTIONS AND ANSWERS BASED ON REVISED SURRY UNIT 1 Analyses Of Core Damage Frequency From intomal 10 CFR PART 20.

Events Dunng M4 Loop Operations. Append 6ces I.

CELLA.M.A.

NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR 5965 MODELING FIELD SCALE UNSATURATED FLOW CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT AND TRANSPORT PROCESSES.

SURRY, UNIT 1 Analysis Of Core Damage Frequency From intemal Fores Durmg M4 Loop Operationa Man Report.

CHAVE2.S.A.

NUREG/CR-6144 VO3 P2: EVALUATION OF POTENTIAL SCVERE AC.

NUREG/CR4196. CALCULATONS TO ESTIMATE THE MARGIN TO CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT FAILURE IN THE TMI-2 VESSEL SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/CR 6197: TML-2 VESSEL INVESTICATON PROJECT INTE-Fires Dunng M4 Loop Operations. Appendices.

GRATION REPORT.

CHUNG,H.M.

CHEN.X.

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5960. STEAM EXPLOSIONS FUNDAMENTALS AND EM-LIGHT WATER REACTORS. Semiennual Report.Aprd 1993. Septem.

GERGETIC BEHAVOR.

bor1993.

i

Personal Author Index 57 NUREG/CR4223: REVIEW OF THE PROPOSED MATERIALS OF CON-CROSS OtAL,A.E.

STRUCTION FOR THE SBWR AND AP600 ADVANCED REACTORS NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Man Report And CICCARELLI.G-Appendix A.

NUREG/CR4213: HIGH-TEMPERATURE HYDROGEN-AIR-STEAM NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-DAMAGE ACCIDENTS: 1992 A STATUS REPORT.Apperdcas B. C, D.

MENT APPARATUS.

E, F, And G.

CLARK,R.L CUTHILL.B.B.

NUREG/CR-5904: FUNCTIONAL ISSUES AND ENVIRONMENTAL NUREG/CP-0136: PROCEEDINGS OF THE DIGITAL SYSTEMS RELI-OUALIFICATION OF DIGITAL PROTECTION SYSTEMS OF AD-VANCED LIGHT WATER NUCLEAR REACTORS.

A81UTY AND NUCLEAR SAFETY WORKSHOP. September 1314 1993,Rockville Crowne Plaza Hotel,Rockville, Maryland.

CLETCHER,J.W.

C NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE G CR 74 V01: SEALED SOURCE AND DEVICE DESIGN DAMAG ACCIDENTS: 1992 A STATUS REPORT. Man Report And SAFETY TESTING.Techrucal Report On The Fmdmgs Of Task 1.Octo-NU CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE ber 1991 - Apnl 1993.

DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Appendices B. C, D, DAILY,M.C.

N ik CR 674 V19: PRECURSORS TO POTENMAL SEVERE CORE NUREG-1500 WORKING DRAFT REGULATORY GutDE ON RELEASE CRITERIA FOR DECOMMISSIONING: NRC STAFF'S DRAFT FOR DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Mam Report And NU 67 V20: PRECURSORS TO POTENTIAL SEVERE CORE DA GE ACCIDENTS: 1993 A STATUS REPORT.Appendences E NU CR4143 V02 PIA EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT COLLINS.J.L.

GRAND GULF, UNIT 1.Analyes Of Core Damage Frequency Frorn in-NUREG/CR4077: DATA

SUMMARY

REPORT FOR FISSION PRODUCT ternal Events For Plant Operational State 5 Dunng A Refueling RELEASE TEST Vl4.

Outage. Sections 1-9.

NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC-COLLINS,M.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG-1460 RO1: GUIDE TO NRC REPORTING AND RECORDKEEP-GRAND GULF. UNIT 1.Analyss Of Core Damage Frequency From In-ING REQUIREMENTS. Compded From Requirements in Title 10 Of temal Events For Plant Operational State 5 Dunng A Refuelmg The U.S. Code Of Federal Regulations As Codifed On December 31 Outage Section 10.

1e93.

NUREG/CR4143 V02PtC: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT COMES,L GRAND GULF UNIT 1.Analyss Of Core Damage Frequency From In-NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL temal Events For Plant Operational State 5 Dunno A Refuelmg FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

Outage.Mam RW ED DEBRIS. Draft For Comment.

NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVERE AC-CONGEMW CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-2907 V12 RADIOACTIVE MATERIALS RELEASED FROM GRAND GULF, UNIT 1.Anatysis Of Core Damage Frequency From In-NUCLEAR POWER P. ANTS. Annual Report 1991.

temal Events For Plant Operational State 5 Dunng Refuelmg Outage.intemal COPINGER,0.A.

NUREG/CR4143 V04: EVALUATION OF POTENTIAL SEVERE ACCI-NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Man Report And GRAND GULF. UNIT 1.Analyss Of Core Damage Frequency From In-Apperdx A.

temally induced Floodmg Events For Plant Operational State 5 Durno NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE a Refuelmg~..

DAMAGE ACCIDENTS: 1992 A STATUS REPORT.Apperdces B, C, D, E F, And G DANG,V.

NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-6144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Mam Report And CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Appareces A-D.

SURRY. UNIT 1.Analyes Of Core Damage Frequency From Intamal NUAEG/CR 4874 V20 PRECURSORS TO POTENTIAL SEVERE CORE Events Dunng M4 Loop Operations. Man Report (Chapters 14).

DAMAGE ACCtDENTS: 1993 A STATUS REPORT,Appendences E NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-A id F.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1.Analyes Of Core Damage Frequency From Intemal CORRADINI,M.L Events Dunng M4 Loop Operations Man Report (Chapters 712).

NUREG/CR4196: CALCULATIONS TO ESTIMATE THE MARGIN TO NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCI-FAILURE IN THE TMI-2 VESSEL DENTS DURING LOW POWER AND SHUTDCWN OPERATIONS AT NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PRCUECT INTE-SURRY, UNIT 1. Analysis Of Core Damage frequency From Intemal GRATION REPORT.

Events Dunng M4 Loop Operataons.Apperdeen A-D.

NUREG/CR4144 V02P3A EVALUATION OF COTENTIAL SEVERE AC-S N LOW POWER AND SHmW OPERAWS AT UREG 5:.41 V02 N1: HEAVY SECTION STEEL IRRADIATION SURRY, UNIT 1.Analyms Of Core Damage Frauency From intemal PRO, GRAM.Sermannual Progress Report For October 1990. March Events Dunng M4 Loop Operations.Appareces t (Sections E.1.E 8).

,gg NUREG/CR 5591 V02 N2: HEAVY SECTION STEEL IRRADtATION NUREG/CR4144 V02P38: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PROGRAM.Swmiannual Pr oss R 1 For it-September 1991.

NUREG/CR4139. CRACK RESTYESTS TWO IRRADIATED SURRY UNIT 1.Analyms Of Core Damage Frequoney From intomal HIGH. COPPER WELDS. Phase 11: Results Of Duplex Type Specimens NI R CR 4 V02P EV ION ENTIAL VE E CORYEL1,E.W.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4160

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP.FP-2.

Events Dunng M4 Loop Operations.Apperdces F-H.

NUREG/CR4144 V02P5: EVALUATION OF POTENTIAL SEVERE ACCI-COX,D.F.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE SURRY, UNIT 1. Ann'ysis Of Core Demage Fromency From Intemal DAMAGE ACCIDENTS: 1992 A STMUS REPORT.Mam Report And Events Dunng M4 Loop Operations.Apperdces t

'R 4674 V18: PRECURSC RS TO POTENTIAL SEVERE CORE DANIEL,S.

DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Appendices B. C, D.

NUREG/CR4143 V02PtA: EVALUATION OF POTENTIAL SEVERE AC-E. F, And G.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT

58 Personal Author Index GRAND GULF, UNIT 1. Analysis Of Core Darnage Frequency Frorn in-DEHART,M.D.

ternal Events For Plant Operational State 5 Durmg A Refuehng NUREG/CH-6102. VALIDATION OF THE SCALE BROAD STRUCTURE Outage Sectons 19.

44. GROUP ENOF/B.Y CROSS-SECTION LIBRARY FOR USE IN NUREG/CR4143 V02P18 EVALUATION OF POTENTIAL SEVERE AC CRITICALITY SAFETY ANALYSES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GRAND GULF,0 NIT 1.Analyes Of Core Darnage Frequency Frorn in-DEHMEL,J-C.

l temal Events For Plant Opershonal State 5 Dunng A Refuehng NUREG/CR4147 V01: CHARACTER 12ATION OF CLASS A LOW-LEVEL Outage Section 10.

RADIOACTIVE WASTE 1986-1990 Esecutive Sumrnary.

l NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4147 V02: CHARACTERl2ATON OF CLASS A LOW 4EVEL i

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RADIOACTIVE WASTE 1986-1990.Mam Report.Part A.

l GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency Frorn in.

NUREG/CR-6147 V03: CHARACTERl2ATION OF CLASS A LOW. LEVEL teenal Events For Plant Operatonal State 5 Dunng A Refuelmg RADOACTIVE WASTE 19861990 Mam Report-Par 1 B Outage Man Report NUREG/CR4147 V04. CHARACTERl2ATION OF CLASS A LOW-LEVEL RADOACTIVE WASTE 19861990.Appeneces A.E.

DAR8Y,J.

NUREG/CR4147 V05: CHARACTERl2ATON OF CLASS A LOW LEVEL NUREG/CH4143 V02PI A: EVALUATION OF POTENTIAL SEVERE AC-RADIOACTIVE WASTE 19861990.Appendu F.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4147 V06. CHARACTER 12ATION OF CLASS A LOW LEVEL GRAND GULF UNIT 1. Analysis Of Core Damage Frequency From in.

RADIOACTIVE WASTE 1986-1190.Appeneces G-J temal Events For Plant Operatonal State 5 Dunng A Refuelmg NUREG/CR4147 V07; CHARACTERl2ATION OF CLASS A LOW-LEVEL NU CR VO P10 EVALUATON OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DEWALL,K.Q.

GRAND GULF. UNIT 1 Analyses Of Core Damage Frequency From in.

NUREG/CH-5935:

SUMMARY

OF WORK COMPLETED UNDER THE temal Events For Plant Operatonal State 5 Dunng A Flefuehng ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUALIFICATON RE-b NU E CR 3 2PIC EVALUATON OF POTENTIAL SEVERE AC-l CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DIAMOND D.J.

GRAND GULF. UNIT 1.Anahms Of Core Damage Frequency From In-NUREG/CR 6144 V029tA: EVALUATION OF POTENTIAL SEVERE AC-temal Events For Ptant Operahonal State 5 Dunng A Refuelmg CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY, UNIT 1.Ana'yms Of Core Damage Freq;ency From Intemal NU E 143 2PT2. EVALUATION OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NU C

44 2 B AL T TE S

RE AC-GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From in.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT temal E For Plant Operational State 5 Dunng Refuehng SURRY, UNIT 1.Analyms Of Core Damage Fregoency From Internal NUREG CR 6143302PT3 EVALUATON OF POTENTIAL SEVERE AC-Events Durmg MdLoop Operatons Mam Report (Chapters 712).

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCl-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GRAND GULF. UNIT 1.Analyms Of Core Damage Frequency From in.

SURRY, UNIT 1.Analyes Of Core Damage Frequency From Internal temal Events For Plant Operatonal State 5 Dunng A Refue4mg Events Durmg Meloop Operatons Appeneces A-D.

Oum Intanal NURE CR4143302PT4. EVALUATION OF POTENTIAL SEVERE AC, NUREG/CH-6144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GRAND GULF, UNIT 1.Analyss Of Core Damage Frequency From in.

SURRY,0 NIT 1.Analyms Of Core Damage Frequency From internal temal Events For Plant Operahonal State 5 Durmg A Refuehng Events Durmg M4 Loop Operatons.Appendees E (Sectone E.1-E.8).

NUREG/CR4144 V02P38. EVALUATON OF POTENTIAL SEVERE AC.

Outage intenal _

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATlONS AT DAVIDSON,0.R.

SURRY. UNIT 1. Analyse Of Core Damage Frequency From Intamal NUREG/CR-6203. VALIDATON STUDIES FOR ASSESSING UNSATU-Events Dunng M4 Loop Opershons.Appendces E (Ser; tons E.9 E.16)

RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

NUREG/CR.6144 V02P4 EVALUATON OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT D#.T.R.R.

SURRY, UNIT 1 Analysis Of Core Damage Frequency Fram Irnemal NUREt. 7 R M011: INFORMATION C!AS AND LIFETIME MORTALITY Events Dunng MdLoop Operations Appendcas F-H RISKS & UrMT;SINDUCED CAUCER Low LET Radafon.

NUREG/CR 6144 V02P5 EVALUATION OF POTENTIAL SEVERE ACCI.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DAVIS).J.

SURRYMNIT 1. Analysis Of Core Damage Frequency From internal NUREG/CR4211: INTEGATED FUEL-C(XnLANT INTERACTION (IFCI Events Dunng M4 Loop Operatonn.Apperxkes t 6.0) CODE User's Manual.

DIBIASIO A.

DAYlS,J.

NUREG/CR-5939 THE EFFECTS OF AGE ON NUCLEAR POWER NUREG/CR 5939-THE EFFECTS OF AGE ON NUCLEAR POWER PLANT CONTAINMENT COOLING SYSTEMS PLANT CONTAINMENT COOLING SYSTEMS DIEFlCKS,0.R.

N R'E [CR4143 V05. EVALUATION OF POTENTIAL SEVERE ACCL G/CR4181 N WW & M W WSm MSM DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUR /CR 6 87 ESULTS MEC Ni L TS AND SUPPLE-GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From MENTARY MICROSTRUCTURAL EXAMINATIONS OF THE TMI2 Seesme Events Durmg M4 Loop Operstons Mam Repod LOWER HEAD SAMPLES-NUREG/CR 6144 V05: EVALUATION OF POTENTIAL SEVERE ACCI-NUREG/CR-6197. TMI-2 VESSEL INVESTIGATION PROJECT INTE-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY UNIT 1 Analyss Of Core Damage Frequency From Seesmic NUREG 422 R YlEW OF THE PROPOSED MATERIALS OF CON-Events Durmg M4 Loop Operatons Man Repod STRUCTON FOR THE SBWR AND AP600 ADVANCED REACTORS.

DAVIS,R.E.

DINOMAN,S.E.

NUREG/CR4004 CALCULATONS IN SUPPORT OF A POTENTIAL NUREG/CH-6092: RISK ASSESSMENT FOR THE INTENTONAL DE.

DEFINITION OF LARGE RELEASE.

PRESSURIZATION STRATEGY IN PWRS.

DAWSON,J.F.

NUREG/CR 5963 CONTINUOUS AE CRACK MONITORING OF A DIS-DOCTOR,S.R.

SIMILAR METAL WELDMENT AT LIMERICK UNIT 1.

NUREG/CR 5985. EVALUATON OF COMPUTER. BASED ULTRASONIC INSERVICE INSPECTION SYSTEMS.

DEGRASSI.D.

NUREG/CR 6181: A PILOT APPLICATON OF RISK-BASED METHODS NUREG/CR4128 P1 PING BENCHMARK PROBLEMS FOR THE ABB/

TO ESTABLISH INSERVICE INSPECTON PRIORITIES FOR NUCLE-CE SYSTEM 80 + STANDARDIZED PLANT.

AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATON

Personal Author Index 59 DoDDs,R.H.

ELUNGWOOD,5.

NUREG/CR4162: EFFECTS OF PRIOR DUCTILE TEARING ON CLEAV-NUREG/GR4008: VALIDATION OF SEISMIC PROBA8ILISTIC RISK AS.

AGE FRACTURE TOUGHNESS IN THE TRANSITION REGION.

SESSMENTS OF NUCLEAR POWER PLANTS.

DOLAN,8.W.

ELUOT,5.

NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE NUREG 1511: REACTOR PRESSURE VESSEL STATUS REPORT.

DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Main Report And Appendur A.

EMRIT.R.

NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE NUREG-0933 S17: A PRORIT12ATION OF GENERIC SAFETY ISSUES.

DAMAGE ACCIDENTS 1992 A STATUS REPORT.Appendees B. C, D, E, F, And G.

EWING,P.D.

NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-5941: TECHNICAL BASIS FOR EVALUATING ELECTRO-DAMAGE ACCIDENTS.1993 A STATUS REPORT. Man Report And MAGNETIC AND RADIO FREQUENCY INTERFERENCE IN SAFETY.

Appendees A-D.

RELATED IAC SYSTEMS.

NUREG/CR-4674 V20: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Appendences E FADDEN,M.

And F.

NUREG-0525 V02 802: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL). January 1,1990 Through December 31,1993.

NUREG 1368: PREAPPLICATION SAFETY EVALUATON REPORT FOR FAIRBANKS,C.

THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

NUREG 1511: REACTOR PRESSURE VESSEL STATUS REPORT.

LOUtD-METAL REACTOR Final Report.

DONOHEW,J.M.

FANOUS,F.

NUREG/CR4161: BUCKLING EVALUATION OF SYSTEM 80+(TM)

NUREG-1368 PREAPPLICATION SAFETY EVALUATION REPORT FOR CONTAINMENT.

THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

LIQUID METAL REACTOR. Final Report.

FARRAR.C.R.

NUREG/CR 6104: SHEAR WALL ULTIMATE DRIFT LIMITS.

NUREG 1415 V07 N01: OFFICE OF THE INSPECTOR FAWCETT-LONG.J.

GENERALSemiannual Report.Apnl 1 September 30,1994-NUREG/CR-5680 VOI: THE IMPACT OF ENVIRONMENTA '. CONDI-DOR W.

TONS ON HUMAN PERFORMANCE. A Handbook Of Env onmental NUREG/CR4074 V01: SEALED SOURCE AND DEVICE DESIGN NU C 15680 V02: THE IMPACT OF ENVIRONMENTN. CONDI-SAFETY TESTING Techncal Report On The FindinDs Of Task 1.Octo*

ber 1991 Apnl 1993.

TONS ON HUMAN PERFORMANCE. A Critcal Review Of Re Litera-g DOTY,K.

^

N RhG/CR-5758 V04: FITNESS FOR DUTY IN T-IE NUCLEAR POWER N LE R E I N A

R 1'

INDUSTRY. Annual Summary Of Program P.ervormance Reports CY DOYLE D.J.

1993.

NUREG/CR-5407. ASSESSMENT OF THE IMPACT OF DEGRADED EA WALL ST NESSES ON SEISMIC PLANT RISK AND SEIS-FINFR DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP.

DRAPER.D.

MENT APPARATUS.

NUREG/CR 5680 VO1: THE IMPACT OF ENVIRONMENTAL CONDI-l TIONS ON HUMAN PERFORMANCE. A Handbook Of Ermronmental FISCHER,L.E.

Exposures NUREG/CR-5403' PREDICTING THE PRESSURE DRIVEN FLOW OF NUREG/CR-5580 V02: THE IMPACT OF ENVIRONMENTAL CONDi-GASES THROUGH MICRO-CAPILLARIES AND MICRO-ORIFICES.

S ON HUMAN PERFORMANCE. A Cntcal Revew Of The Litera-NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABILISTIC DUFFEY,T.A.

RISK ASSESSMENT.

j NUREG/CR-6104 SHEAR WALL ULTIMATE DRIFT LIMITS FLANIGAN,L.F.

DUR81N,N.

NUREG/CR-5128 R01: EVALUATON AND REFINEMENT OF LEAK.

NUREG/CR 5758 V04. FITNESS FOR DUTY IN THE NUCLEAR POWER RATE ESTIMATON MODELS.

INDUSTRY. Annual Summary Of Program Performance Reports CY 1993.

FLIEGEL,M.

NUREG/CR4122: STAFFING DECISION PROCESSES AND NUREG-1486 FINAL SAFETY EVALUATION REPORT TO LICENSE ISSUES Case Studes Of Seven U S. Nuclear Power Plants.

THE CONSTRUCTION AND OPERATON OF A FACILITY TO NUREG/CR4252-LESSONS LEARNED FROM THE THREE MILE ISLAND-UNIT 2 ADVISORY PANEL RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NEAR CLiVE. UTAH Docket No. 40-8989 (Ermrocare of Utah.Inc )

DWYER,P.A.

NUREG-1497. INTERIM LICENSING CRITERIA FOR PHYSICAL PRO-FORESTER,J.

TECTON OF CERTAIN STORAGE OF SPENT FUEL-NUREG/CR4143 V02PI A: EVALUATON OF POTENTIAL SEVERE AC.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT I

ECHEVERRIA.D GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-I NUREG/CR-5680 V01. THE IMPACT OF ENVIRONMENTAL CONDl_

ternal Events For Plant Operational State 5 Dunng A Refuehng 3

TONS ON HUMAN PERFORMANCE. A Handbook Of Envronmental NU E CR Vb2P18. EVALUATION OF POTENTIAL SEVERE AC-NI

-5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI, CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT TONS ON HUMAN PERFORMANCE. A Cntcal Revew Of The Litera.

GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-ture.

ternal Events For Plant Operational State 5 Dunng A Refuehng Section 10.

ECKENRODE,R.J.

NURE CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC-NUREG-0711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT MODEL.

GRAND GULF, UNIT 1. Analysis Of Core Dama9e Frequency From In-temal Events For Plant Operational State 5 Dunng A Refuehng ECONOMOS.C.

Outage Ma n Report.

NUREG/CR4213' HIGH-TEMPERATURE HYDROGEN AIR-STEAM NUREG/CR4143 V02PT4. EVALUATON OF POTENTIAL SEVERE AC.

DETONATION EXPERIMENTS IN THE BNL SMALL SCALE DEVELOP-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT MENT APPARATUS.

GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From In-

l i

L 60 Personal Author Index ternal Events For Plant Operational State 5 Dunng A Refueling NUREG/CR4216: EVALUATION OF ROCK JOINT MODELS AND COM-l Outage. Internal PUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS.

NUREG/CR4143 V04: EVALUATON OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT OiOUERE.P.T.

GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From tr>

NUREG/CR-6269: A PLAA FOR THE MODIFICATION AND ASSESS-tornelly induced Flooding Events Far Plant Operational State 5 Dunng MENT OF TRAC-PF1/ MOD 2 FOR USE IN ANALYZING CANDU 3 a Refueling.

TRANSIENT THERMAL HORAUUC PHENOMENA.

FORSLUNDA GILSERT,5.G.

NUREG/CR 5758 V04: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/CR4639 V5R4P2-NUCLEAR COMPUTERl2ED UBRARY FOR INDUSTRY. Annual Summary Of Program Performance Reports CY ASSESSING REACTOR RELIABILITY (NUCLARR) Volume 5: Data NURE /CR4122-STAFFING DECISION PROCESSES AND NURE 9V RIZED UBRARY FOR ISSUES. Case Studies Of Seven U S. Nuclear Power Plants.

ASSESSING REACTOR RELIABluTY (NUCLARR) Volume 5: Data Manual Part 3. Hardware Component Feelure Data.

FRANCINI,R. ~

NUREG/CR-4599 V03 N2 SHORT CRACKS IN PIPING AND PIPING O

HIGH-TEMPERATURE HYDROGEN-AIR-STEAM WELDS. Semiannual Report, October 1992 March 1993.

4213:

FREDERICK,L, DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-MmEG-1415 V07 N01: OFFCE OF THE INSPECTOR MENT APPARATUS.

GF.NERALSemiannual Report.Agni 1. September 30,1994.

OOLDIN,0.

FR M A NUREG-1492 OFC-MGULATORY ANALYSIS ON CRITERIA FOR THE NUREG/CR4812: MAP' AGING AGING IN NUCLEAR POWER RELEASE N PATIENTS ADMINISTERED RADIOACTIVE PtNN. insights From NRC Maintenance Team inspecton Reports-MATERiALDraft Reprt For Comment.

FRIESEL,M.A.

GOLDMANA NUREG/CR 5963: CONTINUOUS AE CRACK MONITORING OF A DIS-NUREG/CR4104: SHEAR WALL ULTIMATE DRIFT UMITS.

SIMILAR METAL WELDMENT AT UMERCK UNIT 1.

GOLUS,0.R.

GALLETTI.G.S.

NUREG-1368: PREAPPUCATON SAFETY EVALUATION REPORT FOR NUREG4711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

MODEL UQUID-METAL REACTOR. Final Report.

GALYEAN,W.L NUREG/CR-6116 V01: SYSTEMS ANALYSIS PROGRAMS FOR OONZALEZ,M.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR4074 V01: SEALED SOURCE AND DEVICE DESIGN VERSION 5 0.Tectwucal Reference Manual.

SAFETY TESTING. Technical Report On The Findings Of Task 1.0cto-NUREG/CR4145: VERIFICATION AND VAUDATION OF THE SAPHIRE ber 1991. April 1993.

VERSION 4.0 PRA SOFTWARE PACKAGE.

GOODMAN,C.

NUREG-0711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW

-6281: A SIMPUFIED LEAK-BEFORE-BREAK EVALUATON MODEL l

PROCEDURES FOR AUSTENITC AND FERRITC STEEL PIPING.

GORCAF.

l GARCIA.P.

NUREG-1486: FINAL SAFETY EVALUATION REPORT TO UCENSE NUREG/CR-5830: AUXILIARY FEEDWATER SYSTEM RISK BASED IN-SPECTON GUOE FOR THE MCGUIRE NUCLEAR POWER PLANT.

THE CONSTRUCTON AND OPERATION OF A FACILITY TO NUREG/CR4122.

STAFFING DECISION PROCESSES AND RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL ISSUES. Case Studies Of Seven U S. Nuclear Power Plants.

NEAR CUVE. UTAH. Docket No. 40-8989.(Envirocare of Utah,Inc.)

NUREG/CR4123: AN INTERNATIONAL COMPARISON OF COMMER.

GAYDOS.R.G.

CIAL NUCLEAR POWER PLANT STAFFING REGULATIONS AND NUREG/CR 6183: PEER REVIEW OF THE TMi-2 VESSEL INVESTIGA-PRACTICE.19801990.

TION PROJECT METALLURGICAL EXAMINATONS.

NUREG/CR4151: FEASIBluTY OF DEVELOPING RISK BASED RANK.

INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVICE IN-GELHAR.LW.

SPECTION.

NUREG/CR-5965: MODELING FIELD SCALE UNSATURATED FLOW NUREG/CR4181: A PILOT APPLCATION OF RISK-BASED METHODS BS EM MCm mmM M NN NU EG CR FOR ANCE ASSESSMENT OF A HYPO-AR COWONWS AT MRY W 1 NUCLEAR POWER STATON THETICAL LOW LEVEL WASTE FACIUTY. Groundwater Flow And l

Transport Sanulaton-GORHAM E.D.

NUREG/CR-4,51 V01 R1: EVALUATION OF SEVERE ACCIDENT l

OERLACH,L.

NUREG/CR4213: HIGH TEMPERATURE HYDROGEN-AIR-STEAM RISKS: MITHODOLOGY FOR THE CONTAINMENT, SOURCE TERM.CONUEQUENCE, AND RISK INTEGRATION ANALYSES.

DETONATON EXPERIMENTS IN THE BNL SMALL SCALE DEVELOP.

MENT APPARATUS.

OOUDAEN,F, OHADIALIA NUREG/CR-3145 V10- GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPING AND PIPING WESTERN OHIO. INDIANA REGION. Final Report October 1986-Sep-l WELDS. Semiannual Report. October 1992 March 1993.

tomber 1992.

OHALIS,H.A.

GOVINDARAJU,M.

l NUREG/CR4255: DESIGN OF AN OPEN ARCHITECTURE SEISMC NUREG/GR 0013: APPLICATIONS OF A NEW MAGNETIC MONITOR-MONr ORING SYSTEM-ING TECHNIQUE TO IN SITU EVALUATION OF FATIQUE DAMAGE IN FERROUS COMPONENTS.

l OMM W A PrAEG/CR4281: A SIMPUFIED LEAK.BEFORE. BREAK EVALUATON ORANDA.T.M.

PROCEC'mES FOR AUSTENITC AND FERRITIC STEEL PIPING.

NUREG/CR-5908 V02: ADVANCED HUMAN-SYSTEM INTERFACE DESIGN REVIEW GUOEUNE. Evaluation Procedures Aad Gur'44nea ggoggggg NUREG/CR4157: SUN /EY ANO EVALUATION OF a,GING RISK AS.

For Human Factors Engmoenng Reviews.

SESSMENT METHODS A.3 af9LC.* TONS.

OHOSH,A.

NUREG/CR-5758 V04: FITNESS FOR DUTY IN THE saUCLEAR POWER HUREG/CR4178; LABORATORY CHARACTERIZATION OF ROCK INDUSTRY. Annual Summary Of Program Performam Repests CY JOINTS _

1993.

Personal Author index 61 GRAVES,N.L HARTY,R.

NOVG/CR-4838: ItCROCOMPUTER APPLICATONS OF, AND MODI-NUREG/CR4252 LESSONS LEARMED FROM THE THREE MILE FIQwG TO, T.4E MODULAR FAULT TREES.

ISLAND-UNIT 2 ADVISORY PANEL GREIMANKL.

HASKINf.E.

NUREG/CR4161: BUCKLING EVALUATON OF SYSTEM 80+(TM)

NUREG/CR4042 PERSPECTIVES JN REACTOR SAFETY.

CONTAINMENT.

HAUTHJ.

1 GPMITH,R.O.

NUREG/CR4122: STAFFING DECISION PROCESSES AND NUREG/CR4044: EXPERIMENTS TO INVESTIGATE DIRECT CON-ISSUES. Case Studies Of Seven U.S. Nuclear Power Plants.

i TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NUREG/CR4123: AN INTERNATIONAL COMPARISON OF COMMER.

ZION NUCLEAR POWER PLANT IN THE St TY TEST FACILITY.

CIAL NUCLEAR POWER PLANT STAFFING REGULATIONS AND PRACTICE.19801990.

NUREG/CR-5812-MANAGING AGING IN NUCLEAR POWER HEASLER,P.G.

l PLANTS Insights From NRC Masntenance Team inspecta Reports.

NUREG/CR-5161 V02: EVALUATION OF SAMPLING PLANS FOR IN-SERVICE INSPECTION OF STEAM GENERATOR N

CR4157: SURVEY AND EVALUATION OF AGING RISK AS.

TUBES.Comprehenarve Anahtical And Monte Carlo Simulation Results I

SESSMENT METHODS AND APPLICATIONS.

GUNTHER,W HELT ONJ.C.

NUREG/CdO135: WORKSHOP ON ENVIRONMENTAL QUALIFICATION NUREG/CR-4M1 V01 R1: EVALUATON OF SEVERE ACCIDENT OF ELECTRC EQUIPMENT.

RISKS: Mf'THODOLOGY FOR THE CONTAINMENT SOURCE NUREG/CR-5812-MANAGING AGINf, IN NUCLEAR POWER TERM,CONSEG;CN. AND RG INTEGRATION ANALYSES.

PLANTS.Insaghts From NRC Maintenance, team inspection Reports' HERMANN,0.W.

GUTIERRE%L NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSi3 DECWf NUREG/CR e?56 V04: FITNESS FOR DUTY IN TME NUCLEAR POWER HEAT GUIDE USING SAS2H/OPIGEN-S DATA.

INDUSTRY. Annual Sumrqary Of Program Forformance Reports CY tirs3.

HIGGINS).C.

NUREG 0711: HUMAN FACTORS ENGINEERING 'ROGRAM REVIEW Guma a.

NUREG/CR4105: HUMAN FACTOFtS ENGINEERIN3 GUIDANCE FOR MODEL NUREG/CR4203: VALIDATION STUDIES FOR ASSESSING UNSATU-RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

THE REVIEW OF ADVANCED ALARM SYSTEMS.

NUREG/CR-6146: LOCAL CONTROL STATIONS: HtlMAN ENGINEER-CWALTNEY,R.C.

ING ISSUES AND INSIGHTS.

NUREG/CR-5359 REVIEW OF ELASTC STRESS AND FATGUE TO.

FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA.

HILL,R.L.

TON TO ASME DESGN CRITERIA FOR NUCLEAR POWER PIPING NUREG/CR4289: RECONCENTRATON OF RADIOACTIVE MATERIAL SYSTEMS.

RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR HACKETT,E.

NUMEG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

HILLMAN,K.

NUREG/CR-6086: SELECTED FAULT TESTING OF ELECTRONIC ISO.

NR CR 5919-REPOSITORY OPERATONAL CRITERIA COMPARA.

TIVE ANALYSIS.

HILLS,R.G.

NUREG/CR-6063. INTRAVAL PHASE 11 MODEL TESTING AT THE LAS UR G 486: FINAL SAFETY EVALUATION REPORT TO LICENSE NU E /

120 C OLLED FIELD W TUDY FOR VALIDATION OF THE CONSTRUCTON AND OPERATON OF A FACIUTY TO VADOSE ZONE TRANSPORT MODELS.

RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NEAR CLIVE. UTAH. Docket No. 404969.(Envirocare of Utah Inc.)

HIMESJ.

NUREG-1471: CONCEPT OF OPERATIONS WITH ORGANIZATON

• • E "**

N G

188 V01: MICROBIAL DEGRADATON OF LOW-LEVEL RADCACTIVE WASTE. Annual Report Fo-FY 1993.

HINS,A.G.

NUREG/CR4185: TMI-2 INSTRUMENT NOZZLE EXAMINATONS AT l

I A GmNE NATN WMAME% M -June m N

C 5973 R01: CODES AND STANDARDS AND OTHER GUID-ANCE CITED IN REGULATORY DOCbl.*ENTS.

HISER A.L.

j NUREG-1426 V02: COMPILATION OF REPORTS FROM RESEARCH

]

HANSON,A.L

^

RANCH ISim ENGINEERING 1

F NITION OF LAR E SF HANSOIW.

HO,C1 NUREG/CR4158: IMPLICATONS FOR ACCIDENT MANAGEMENT OF NUREG/CR4221. T>dE VALLES NATURAL ANALOGUE PROJECT.

ADDING WATER TO A DEGRADING REACTOR CORE.

go,y, HARPERI.T.

NUREG/CR-6144 V00 P1: EVALUATON OF PC TENTIAL SEVERE AC.

NUREG/CR-4551 V01 Rt: EVALUATION OF SEVERE ACCIDENT CIDENTS DURING L')W POWER AND SHUTEOWN OPERATIONS AT i

RISKS: METHODOLOGY FOR THE CONTAINMENT SOURCE SURRY, UNIT 1.Analyas Of Core Damage Firauency From Internal TERM. CONSEQUENCE. AND RISK INTEGRATON ANALYSES.

NU bG 1 V P VA A ON POTENTIAL SEVERE AC.

HARPER,M.R.

CIDENTS DURING LOW POWER AND SI UTDOWN OPERATIONS AT i

NUREG-1022 R01 DR FC: EVENT REPORTING GUIDELINES SURRY, UNIT 1. Analysis Of Core Darny Frequency From internal 10CFR50.72 AND 50.73.Second Draft For Comment.

Fires Dunng Mid-Loop Operatons.Appendl m s.

HARRIS.R.V.

HOSSINS,R.R.

NURiG/CR-5965: EVALUATION OF COMPUTER-BASED ULTRASONIC NUREG/CR4160

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

INSERVICE INSPECTON SYSTEMS.

RELAP5 ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP-2.

1 HARTFIELD.R.A.

HOFMAYER.C.H.

1 NUREG-0020 V18: LICENSED OPERATING REACTORS STATUS SUM-NUREG/CR4241: TECNNCAL GUIDELINES FOR ASEISMC DESIGN MARY REPORT.Deta As Of December 31,1993.(Gray Book 1)

OF NUCLEAR POWER PLANTS. Translation Of JEAG 46011987.

62 Personal Author Index HOHORST,J K.

NUREG/CR4144 V02P5: EVALUATON OF POTENTIAL SEVERE ACO-NUREG/CR4180

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP-2 SURRY, UNIT 1.Analysse Of Core Damage Frequency From internal Events Dunng M4 Loop Operatons.Apperwhees 1.

EN NUREG/CR4232: ASSESSING THE ENVIRONMENTAL AVAILABillTY HUBER D.

OF URANIUM IN SOILS AND SEDIMENTS NUREG-1415 V06 NO2: OFFICE OF THE INSPECTOR HOWESA GENERALSemiannual Report,0ctober 1,1993 March 31,1994.

NUREG-1415 V07 Not: OFFICE OF THE INSPECTOR NUREG/CR4144 V02 PIA EVALUATON OF POTENTIAL SFV'aE AC-CIDENTS DURING LOW POWER AND SHUTDOWN 0%ATIONS AT GENERALSemannual Report, April 1 -September 30,1994.

SURRY, UNIT 1. Analysis Of Core Damage Free,ency From intemal HUFFERT,A.M.

Events Dunng Med-Loop Operabons Maan Report (Chapters 14)-

NUREG-1500: WORKING DRAFT REGULATORY GUIDE ON RELEASE NUREri/CR4144 V02P18: EVALUATION OF POTLNTIAL SEVERE AC-CRITERIA FOR DECOMMISSIONING: NRC STAFF'S DRAFT FOR QDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT COMMENT.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal NUREG-1501 DRFT: BACKGROUND AS A RESIDUAL RADCACTIVITY Events Dunng M4 Loop Operations Main Report (Chapters 712)-

NUREG/CR4144 V02P2 EVALUATION OF POTENTIAL SEVERE ACO-CRITERION FOR DECOMMISSONING Appendix A To The Draft Go-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT nenc Envronmental impact Statement in Support Of RutemeMng On SURRY, UNIT 1. Analyses Of Core Darnage Frequency Frorn Internal Radiological Critere For Deconwnissaaning Of NRC.

Events Dunng M4 Loop Operatons es A-D HUNT,P.

NUREG/CR4144 V02P3A EVALUAT OF POTENTIAL SEVERE AC-NUREG/CR-6122:

STAFFING DECISION PROCESSES AND QDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT ISSUES Case Studees Of Seven U.S. Nuclear Power Plants.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From intomal Events Dunng M4 Loop Operations. Appendices E (Sectons E.1 E.8).

HUTTON P.H.

NUREG/CR-6144 V02P38: EVALUATlun OF POTENTIAL SEVERE AC' NUREG/CR-5963: CONTINUOUS AE CRACK MONITORING OF A DIS-ODENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SIMILAR METAL WELDMENT AT LIMERICK UNIT 1.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal Events Dunng M4 Loop Operations. Appendices E (Sectons E.9-E,16).

ILBERO.D.

NUREG/CR4144 V02P4: EVALUATON OF POTENTIAL SEVERE ACO-NUREG/CR4144 V02PI A EVALUATON OF POTENTIAL SEVERE AC.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY,0 NIT 1. Analysis Of Core Damage Frequency From Internal SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal Events Dunng M4Loor Operatons Appendices F-H-NUREG/CR4144 V02P5: EVALUATON OF POTENTIAL SEVERE ACCl-Events Dunng M4 Loop Operations Main Report (Chapters 14)

NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVEP.E AC-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIOGo AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal Events Dunng M4 Loop Operations. Appendices I.

SURRY, UNIT 1. Analysis Of Core Damage Fremency From intemal Events Dunng M4 Loop Operations. Main Report (Chsspters 712).

HORAAC.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACO-NUREG/CR-4551 V01 R1: EVALUATON OF SEVERE ACCOENT DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RISKS: METHODOLOGY FOR THE CONTAINMENT. SOURCE SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal TERM CONSEQUENCE, AND RISK INTEGRATION ANALYSES.

NUR CR 44 V02 A AL T POTE TIAL SEVtX AP HOU,Y,-M.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Events Dunng Mid-Loop Operations, Appendices E (Sections E.1.E.8).

SURRY, UNIT 1.Analyms Of Core Damage Frequency From Intemal NUREG/CR4144 V02P38: LVALUATION OF POTENTIAL SEVERE AC-Fres Dun ~na M4 Loop Operations Main Report.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4144 V03 P2: EVALUATION OF POTENTIAL SEVERE AC.

SURRY, UNIT 1. Analyses Of Cue Damage Frequency From Internal CIDENTS DURING LOW POWER AND SHUTDOWN OPEMTONS AT Events Dunng M4 Loop Operates. Appendices E (Sectons E.9-E.16).

SURRY, UNIT 1. Analysis Of Core Damage Frequent / From internal NUREG/CR4144 V02P4: EVALUAMON OF POTENTIAL SEVERE ACO-Fres Dunng M4 Loop Operatons. Appendices.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal HSIUNG,5.M.

Events During M4 Loop Operations. Appendices F H.

NUREG/CR4178: LABORATORY CHARACTER 12ATION OF ROCK NUREG/CR4144 V02P5. EVALUATON OF POTENTIAL SEVERE ACO.

K4NTS DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUF EG/CR4216: EVALUATON OF ROCK JOINT MODELS AND COM-SURRY, UNIT 1. Analyses Of Core Damage Frequency From internal PL TER CODE UDEC AGAINST EXPERIMENTAL RESULTS.

Events Dunng M4 Loop Operatons. Appendices L HSU.C J.

IPPOLITO L.M-NU8",dG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CP 0136: PROCEEDINGS OF THE DIGITAL SYSTEMS REU-CDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT ABILITY AND NUCLEAR SAFETY WORKSHOP. September 13-14, SURRY, UNIT 1. Analyses Of Core Damage Frequency Frorn intemal 1993,Rockville Crowne Plaza Hotel,Rockville. Maryland Events Dunng M4 Loop Operations Main Report (Chapters 14)

NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-IRWIN,G.R.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4262: CLEAVAGE BEHAVORS IN NUCLEAR VESSEL SURRY. UNI,r 1. Analysis Of Core Damage Frequency From Intemal STEELS.

Events Dunno M4 Loop Operations.Mann Report (Chapters 7-12)

?.'UREG/CR4144 V02P2. EVALUAlON OF POTENTIAL SEVERE ACCl-ISHit M.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4267. AIR-WATER SIMULATION OF PHENOMENA OF SURRY, UNIT 1 Analysis Of Core Damage Frequency From intemal CORIUM DISPERSON IN DIRECT CONTAINMENT HEATING.

Events Dunng M4 Loop Operatons. Appendices A.D.

NUREG/CR4144 V02P3A. EVALUATION OF POTENTIAL SEVERE AC-ISKANDERAK.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4139: CRACK. ARREST TESTS ON TWO IRRADIATED SURRY, UNIT 1.Analps Of Core Damage Frewency From Intemal HIGH COPPER WELDS. Phase 11 Results Of Duplex Type Specimens Events Dunng M4 Loop Operations.

es E (Sectons E.1-E.8).

NUREG/CR4228: PRELIMINARY ASSESSMENT OF THE FRACTURE NUREG/CR-6144 V02P38. EVALUAT OF POTENTIAL SEVERE AC-BEHAVIOR OF WELD MATERIAL IN FULL. THICKNESS CLAD CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT BEAMS.

SURRY UNIT 1. Analyses Of Core Damage Frequency Frorn internal NUREG/CR.6249: UNIRRADIATED MATERIAL PROPERTIES OF MID-Events Dunng M4 Loop Operations Ap2endices E (Sections E.9-E.16)

LAND WELD WF 70.

NUREG/CR4144 V02P4. EVALUATION UF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ISRAEt,S.

SURRY. UNIT 1. Analyses Of Core Damage Frequency From Intemal NUREG-1022 ROI DR FC: EVENT REPORTING GUIDELINES Events Dunng M4 Loop Operatsons. Appendices F.H.

10CFR50 72 AND 50 73.Second Draft For Comment

Personal Author index 83 l

- JACKSON,K.

NUREG/CR4254: SOUTHERN APPALACHIAN REGIONAL SEISMIC NUREG 1471: CONCEPT OF OPERATIONS WITH ORGANIZATION NETWORK.

CHARTS.NRC inca $ ant Response JOHNSTON,J.J.

JACOSUS,MA NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DEGRADED NUREG/CR4095: AGING. LOSSOF400LANT ACODENT (LOCA).

SHEAR WALL STIFFNESSES ON SEISMIC PLANT RISK AND SEIS-1 AND HIGH POTENTIAL TESTING OF DAMAGED CABLES-MIC DESIGN LOADS.

JANSEN,J.M' 4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-JONESM.

NUREG/CR4145: VERIFICATION AND VALIDATION OF THE SAPHIRE DAMAGE ACQDENTS: 1992 A STATUS REPORT.Maen Report And VERSION 4.0 PRA SOFTWARE PACKAGE.

74 V18: PRECURSORS TO POTENTIAL SEVERE CORE D

CCOENTS: 1992 A STATUS REPORT.Appendces B, C, D.

RE 022 R01 DR FC: EVENT REPORTING GUIDELINES 10CFR50.72 AND 50.73.Second Daft For Cornrnent.

JASTROW,J.D.

NUREG/CR-5229 V06: FIELD LYSIMETER INVESTIGATONS: LOW.

JOY,D.R.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG-0430 Vt3: LICENSED FUEL FAOLITY STATUS FISCAL YEAR 1993. Annual Report REPORT. inventory Dfference Data. July 1,1992. June 30,1993.(Gray g

JENSEN.S.M.

NUREG/CR4160:

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

JOYCE,J.A.

I RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP 2.

NUREG/CR4051: EFFECTS OF TENSILE LOADNG ON UPPER SHELF NUREG/CR4195: EXAMINATION OF RELOCATED FUEL DEBRIS AD.

FRACTURE TOUGHNESS.

JACENT TO THE LOWER HEAD OF THE TMI 2 REACTOR VESSEL KAM,F.s.K.

UR GR-0013. ;JT1.lCATIONS OF A NEW MAGNETIC MONITOR-0 AB E R a

ING TECHNIQUE TD IN SITU EVALUATION OF FATIOUE DAMAGE IN FEMIOUS COMPONENTS.

NUREG/CR4076: TR-EDB: TEST REACTOR EMBRITTLEMENT DATA BASE.VERSON 1.

JOHNSEN G.W.

NUREG/CR4206 TRANSPORT CALCULATIONS OF RADIATON EX-rdHEG/CR4l00 UNCERTAINTY ANALYSIS OF SUPPRESSION POOL POSURE TO VESSEL SUPPORT STRUCTURES IN THE TROJAN RE.

NF4 TING CJRING AN ATWS IN A BWR-5 PLANT.An Applicaton Of ACTOR.

Tl i SAU ilethodology UmnD he BNL Engmoenng Plard Analyrer.

T KANA D.D.

JOHNSON.D.

NUREG/CR4178: LABORATORY CHARACTERtZATION OF ROM NURiG/CR4141 V02P1A EVALUATON OF POTENTIAL SEVERE AC.

JOINTS.

CIDENTS DUA NG LOW POWER AND SHUTDOWN OPERATONS AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal KAPLAN D.L Events Dunng k4 Loop Operations.Mam Report s14).

NUREG/CR4289: RECONCENTRATION OF RADIOACTIVE MATERIAL NUREG/CR4144 '/02P18: EVALUATION OF POT NT L SEVERE AC.

RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PART 20.

SURRY UNIT 1.Analyms Of Core Damage Frequency From intemal Events During M4 Loop Operations Man Report (Chapters 712).

KAPLAN,M.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACO.

NUREG/CR4147 V01:CHARACTERIZATON OF CLASS A LOW-LEVEL i

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RADIOACTIVE WASTE 19861990 Executive Sumrnary.

SURRY, UNIT 1. Analyses Of Core Damage Frequency From intemal NUREG/CR4147 V02: CHARACTERIZATION OF CLASS A LOW LEVEL Events Deng Mid-Loop atons se A-0 RADIOACTIVE WASTE 1986-1990. Man Report-Part A.

NUREG/CR4144 V02P3A: VALUAT OF POTENTIAL SEVERE AC-NUREG/CR4147 V03: CHARACTERIZATION OF CLASS A LOW-LEVEL ODENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT RAD 60 ACTIVE WASTE 1986-1990 Man Report-Part B.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intamal NUREG/CR4147 V04:CHARACTERIZATON OF CLASS A LOW-LEVEL Events Dunng M4 Loop Operatons.

s E LSectons E 1-E 8).

RADIOACTIVE WASTE 198&1990.Appendcas A-E.

NUREG/CR4144 V02P38: EVALUAT OF POTENTIAL SEVERE AC-NUREG/CR4147 V05: CHARACTERIZATION OF CLASS A LOW-LEVEL ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RADOACTIVE WASTE 19661990 Appendix F.

SURRY, UNIT 1.Analyms Of Core Damage Frequency Frorn intomal NUREG/CR4147 V06: CHARACTERIZATION OF CLASS A LOW-LEVEL Events Dunng M4 Loop Operations Appendcas E (Sectons E 9-E.16).

RADIOACTIVE WASTE 1986-1190.Appendees G4 NUREG/CR4144 V02P4: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR4147 V07: CHARACTERIZATION OF CLASS A LOW-LEVEL DENTS DUR!NG LOW POWER AND SHUTDOWN OPERATIONS AT RADCACTIVE WASTE 198&1990. Appendices K.P.

SURRY, UNIT 1.Analyms Of Core Damage Frequency From Intemal Events Dunng MdLoop Operations Appendees F.H.

KARLSEN,T.

NUREG/CR4144 V02PS: EVALUATON OF POTENTIAL SEVERE ACO-NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LIGHT WATER REACTORS. Semiannual Repor1.Apnl 1993. Septem.

SURRY, UNIT 1.Analyms Of Core Damage Frequency Frorn Intemal ber 1993.

=

Events Dunng M4 Loop Operatons.Appendees 1.

KASSNER,T.F.

JOHNSONAJ.

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5726. REVIEW OF THE DIABLO CANYON PROBABILISTIC RISK ASSESSMENT.

LIGHT WATER REACTORS. Semiannual Report,Apnl 1993. Septem-ber 1993.

JOHNSO%R L NUREG/CR4176: REVIEW OF ENVRONMENTAL EFFECTS ON FA-NUREG 1E95: OVERALL REVIEW STRATEGY FOR THE NUCLEAR TIGUE CRACK GROWTH OF AUSTINITIC STAINLESS STEELS.

REGULATORY COMMISSON'S HIGH-LEVEL WASTE REPOSITORY NUREG/CR4223: REVIEW OF THE FHOPOSED MATERIALS OF CON-PROGRAM.

STRUCTION FOR THE SBWR ANil AP600 ADV/NCED REACTORS.

JOHNSON,T.L.

KASTUR1,S.

NUREG-1486: FINAL SAFETY EVALUATON REPORT TO LICENSE NUREG/CR.5990 THE EFFECTS OF MLAR-GEOMAGNETICALLY IN-THE CONSTRUCTON AND OPERATION OF A FACILITY TO DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL POWER STATIONS NEAR CLlVE. UTAH Docket No. 404989 (Ermrocare of Utah,Inc.)

JOHNSTON.A.C.

NUREG/CR4228: PRELIMINARY ASSESSMENT OF THE FRACTURE NUREG/CR4209-MEMPHIS AREA REGIONAL SEISMIC BEHAVIOR OF WELD MATERIAL IN FULL THICKNESS CLAD i

NETWORK Fanal Report. October 1986 - September 1992.

BEAMS.

4

64 Personal Author index l

KEISLER,J.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCl-NUREG/CR4237: STATISTOAL ANALYSIS OF FATIGUE STRA!N-LIFE DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT l

DATA FOR CARBON AND LOW-ALLOY STEELS.

SURRY, UNIT 1. Analysis Of Core Damage Frequency Frorn internal i

Events During M4 Loco Operations.Appareces A-D.

KENDRICK,E.D.

NUREG/CR4144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-l NUREG/CR4950 V09: FUEL PERFORMANCE REPORT FOR 1991-CIDENTS DURING LOW POWER AND SHtrTDOWN OPERATONS AT l

SURRY, UNIT 1. Analysis Of Core Dama9e Frequency From internal KENNEALLY R.M.

NUREG-1368: PREAPPUCATON SAFETY EVALUATON REPORT FOR NUR CR 8

AU TENT SE E

C-THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LIOutD-METAL REACTOR. Final Report SURRY, UNIT 1. Analyses Of Core Damage Frequency From Intemal KERR.G.D.

Events Dunng M4 Loop Operations. Appendices E (Sectons E.9-E.16).

NUREG/CR4144 V02P4: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR-5569 Rot: HEALTH PHYSICS POSITONS DATA BASE.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT I

KHAM.T.A.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal NUREG/CR-4409 V05: DATA BASE ON DOSE REDUCTON RE.

Events Dunn0 M4 Loop Operations.Appereces F-H.

NUREG/CR4144 V02PS: EVALUATON OF POTENTIAL SEVERE ACC8-SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT KlWNSKIT.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From intamal NUREG/CR-4599 V03 N2-SHORT CRACKS IN PIPING AND PlPING Events Dunng M4 Loop Operatens. Appendices L WELDS.Sermantmal Report. October 1992 March 1993.

NUREG/CR4144 V04: EVALUATION OF POTENTIAL SEVERE ACC4-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT KIM H.A SURRY UNIT 1. Analysis Of Core Damage Frequency From Intemai j

NUREG/lA 0114 ASSESSMENT OF RELAP5/ MOD 3 WITH THE LOFT Floods During M4 Loop Operations.

L9-1/LS-3 EXPERIMENT SIMULATING AN ANTOIPATED TRANSIENT l

WITH MULTIPLE FAILURES.

KONZEK,0.J.

I NUREG/CR4174 V1 DFC: REVISED ANALYSES OF DECOMMISSION-KIMA ING FOR THE REFERENCE BOILING WATER REACTOR POWER NUREG/CR-5994: EMERGENCY OtESEL GENERATOR: MAINTENANCE STATION, Effects Of Current Regulatory And Other Consideratons On AND F.ULURE UNAVAILABluTY, AND THEIR RISK IMPACTS.

The Financial Assurance Requrements Of The Decommessonmg Rule ONOSHITA M.

NU 5'/CR4174 V2 DFC: REVISED ANALYSES OF DECOMMISSON-NUREG/CRJ213: HIGH-TEMPERATURE HYDROGEN-AIR STEAM ING FOR THE REFERENCE BOILING WATER REACTOR POWER DETONATION EXPERIMENTS IN THE DNL SMALL-SCALE DEVELOP.

STATON Effects Of Current Regulatory And Other Considerstone On MENT APPARATUS.

The Financial Assurance Requirements Of The Decommenssonog Rule KIRK.H.K.

And...

NUREG/CR4143 V02P1A EVALUATION OF POTENTIAL SEVERE AC.

CIDENTS DURING LOW POWER AND SHl/TDOWN OPERATONS AT E CR-5904. FUNCTIONAL ISSUES AND ENVIRONMENTAL GRAND GULF UNIT 1. Analysis Of Core Damage Frequency From in.

QUALIFICATON OF DIGITAL PROTECTON SYSTEMS OF AD-temal Events For Plant Operatonal State 5 Dunng A Refuelmg VANCED LIGHT. WATER NUCLEAR REACTORS.

NUREgsy/,e.Sectons 19'P18: EVALUATION OF POTENTIAL SEVERE AC-Outa NUREG/CR 5941: TECHNICAL BASIS FOR EVALUATING ELECTRO-

- CR4143 V02 MAGNETIC AND RADIO-FREQU5NCY INTERFERENCE IN SAFETY-i CIDENTS DURiNG LOW POWER AND SHUTDOWN OPERATONS AT RELATED l&C SYSTEMS.

j GRAND GULF, UNIT 1.Anatysis Of Core Damage Frequency From in.

j ternal Events For Plant Operational State 5 Dunng A Refueling KORTH,0.E.

i N-W 10.

NUREG/CR4194: METALLOGRAPHC AND HARDNESS EXAMINA.

NUREd/CR4143 V02P1C: EVALUATON OF POTENTIAL SEVERE AC-TONS OF TMI-2 LOWER PRESSURE VESSEL HEAD SAMPLES, CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4197: TMI2 VESSEL INVESTIGATION PROJECT INTE.

GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From in.

GRATION REPORT.

temal Events For Plant Operational State 5 Dunng A Refueling Outage Mam Report.

KOS,J.A.

NUREu/CR4143 V04: EVALUATON OF POTENTIAL SEVERE ACCl*

NUREG/CR 6196: CALCULATIONS TO ESTIMATE THE MARGIN TO DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT FAILURE IN THE TMI 2 VESSEL GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-NUREG/CR4197: TMI-2 VESSEL INVESTIGATON PROJECT INTE-temally Induced Flooding Events For Plant Operatonal State 5 Dunng GRATION REPORT.

NL RE /

1"66: RtSK IMPACT OF TECHNICAL SPECIFICATONS RE-KOT,C.A.

QUIREMENTS DURING SHUTDOWN FOR BWRS.

NUREG/CR4234: VALIDATION OF ANALYSIS METHODS FOR AS.

l SES$1NG FLAWED PlPING SUBJECTED TO DYNAMIC LOADING.

KLAMERUS.E.W.

NUREG/CR4236: SEISMIC INVESTIGATONS OF THE HDR SAFETY NUREG/CR-5407: ASSESSMENT OF THE IMPACT O' DEGRADED PROGRAM. Summary Report SHEAR WALL STlFFNESSES ON SEISMO PLANT Rf.,K AND SEIS-MO DESIGN LOADS.

KOZLOWSKI.S.D.

NUREG/CR4276: OUALITY MANAGEMENT IN REMOTE AFTERLOAD-KMETYK.L.N.

ING BRACHYTHERAPY.

NUREG/CR4107:

SUMMARY

OF MELCOR 1.8.2 CALEULATIONS FOR THREE LOCA SEQUENCES (AG.S2D & S30) AT THE SURRY PLANT.

KRAFT,G.D.

i NUREG/CR4218: A REVIEW OF THE TECHNICAL ISSlJES OF AIR IN-NUREG/CR4255: DESIGN OF AN OPEN ARCHITECTURE SEISMO l

GRESSON DURING SEVERE REACTOR ACCIDENTS.

MONITORING SYSTEM.

f KNUD8ON.D.L.

KRAFT,N.C.

l NUREG/CR4075 S01: THE PROBABILITY OF CONTAINMENT FAIL.

NUREG/CR4164: RELEASE OF RADONUCLIDES AND CHELATING UAE BY DIRECT CONTAINMENT HEATING IN ZON.

AGENTS FROM CEMENT.SOUDlFIED DECONTAMINATON LOW.

l LEVEL RADIOACTIVE WASTE COLLECTED FROM THE PEACH KOHUT,P.

BOTTOM ATOMO POWER STATON UNIT 3.

I NUREG/CR4144 V02PtA-EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4201: COMPRESSON AND IMMERSION TESTS AND l

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT LEACHING OF RADIONUCLOES. STABLE METALS, AND CHELATING SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intamal AGENTS FROM CEMENT. SOLIDIFIED DECONTAMINATION WASTE Events Dunng M4 Loop Operations.Mam Report (Chapters 14)

COLLECTED FROM NUCLEAR POWER STATONS.

NUREG/CR4144 V02P18. EVALUATON OF POTENTIAL SEVERE AC-l CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT KRISHNAMURTMf i

SURRY UNIT 1.Analysee Of Core Damage Frequency From Intemal NUREG/CR-5535 V06: RELAP5/ MOD 3 CODE MANUALValidaten Of

(

Events Dunng M4 Loop Operstone Man Report (Chapters 712).

Numencal Techruques in RELAPS/ MOD 3.

l l

Personal Author index 65 KRISHNASWAMY,P.

LEAHY,TJ.

NUREG/CR-4599 V03 N2-SHORT CRACKS IN PIPING AND PIPING NUREG/CR4088.

SUMMARY

OF 1991 1992 MISADMINISTRATON WELDS.Sermannual Report, October 1992 - March 1993.

EVENT INVESTOATONS.

NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPING AT LEE,8.8.

LWR TEMPERATURES.

NUREG/CP.0135: WORKSHOP ON ENVIRONMENTAL QUALFICATION OF ELECTRIC EQUIPMENT.

KRUMHANSLJ.L.

NUREG/CR4087: THE EFFECTS OF AGING ON BOtuNG WATER RE-NUREG/CR4221: THE VALLES NATURAL ANALOGUE PROJECT.

ACTOR CORE ISOLATION COOUNG SYSTEMS.

KRUPA,K.M.

LEE,M.P.

NUREG/CR4232: ASSESSING THE ENVIRONMENT AL AVAll ABILITY NUREG-1494: STAFF TECHNOAL POSITON ON CONSIDERATION OF OF URAN 1UM IN SOILS AND SEDIMENTS.

FAULT OtSPLACEMENT HAZARDS IN GEOLOGIC REPOSITORY KUAN.P.

DESO N.

NUREG/CR4158: IMPUCATONS FOR ACCOENT MANAGEMENT OF LELLOUCHE G.S.

ADDING WATER TO A DEGRADtNG REACTOR CORE.

NUREG/Clk4200- UNCERTAINTY ANALYSIS OF SUPPRESSION POOL KUECK J.D.

HEATING DURING AN ATWS IN A BWR-5 PLANT.An Application Of NUREG/CR4205: VALVE ACTUATOR MOTOR DEGRADATION.

The CSAU Methodology Using The BNL Engmeenng Plant Analyzer.

KUNKEL,C.

LEONARD,S.K.

NUREG/CR4169: RELAY TEST PROGRAM. Sanes il Tests integral Test.

NUREG/CR4255: DESIGN OF AN OPEN ARCHITECTURE SEISMIC ing Of Relays And Circuit Breakers.

MONITORING SYSTEM.

KURT2,RJ.

LEVERT,F.E.

NUREG/CR-5161 V02: EVALUATION OF SAMPUNG PLANS FOR IN.

NUREG/CR-4833: LARGE AREA SELF-POWERED GAMMA RAY SERVICE INSPECTION OF STEAM GENERATOR DETECTORPhase il Development Of A Source Position Monitor For l

TUBES.Cornprehensrve Analytical And Monte Carlo Sunulation Residts Use On industrial Radiographic Urvts.

For Several Sarnpling Plans.

LEWIS,P.M.

KVAM,P.

NUREG/CR-6208: AN EMPIRICAL INVESTIGATON OF OPERATOR NUREG/CR4157: SURVEY AND EVALUATION OF AGING RISK AS-PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER-SESSMENT METHODS AND APPUCATONS.

GENCIES.

KVARFORDT.KJ.

LIN,J.

NUREG/CR4118 V02: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4144 V02P1A: EVALUATON OF POTENTIAL SEVERE AC-HANDS-ON INTEGRATED RELIABluTY EVALUATONS (SAPHIRE)

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT VERSION 5.0. integrated Rehability And Risk Analysis Systern (IRRAS)

SURRY. UNIT 1Analysm Of Core Damage Frequency From internal Reference Manual Events Dunng MdLoop Operations Main Report (Chapters 14).

NUREG/CR4116 V04: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4144 V02P18: EVALUATON OF POTENTIAL SEVERE AC.

HANDSON INTEGRATED RELIABILITY EVALUA'ilONS (SAPHIRE)

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT VERSION 5.0. Systems Anatysis And Risk Assessment (SARA) Refer-SURRY, UNIT 1. Analysis Of Core Damage Frequency Frorn internal ence Manual.

Events During MdLoop Operations. Man Report (Chapters 712).

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCI-LACH.D-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4252-LESSONS LEARNED FROM THE THREE MILE SURRY, UNIT 1 Analyse Of Core Damage Frequency From Intemal ISLANDUNIT 2 ADVtSORY PANEL Events During MdLoop stens.

es A-D.

NUREG/CR4144 V02P3A-VALUAT OF POTENTIAL SEVERE AC.

WOME,R CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-3145 V10- GEOPHYSICAL INVESTIGATONS OF THE SURRY, UNIT 1Analysm Of Core Damage Frequency Frorn Internal WESTERN OHIO-INDIANA REGON.Fmal Repor1, October 1988-Sep-Events During MdLoop Operatons.

s E (Sections E.1.E.8).

tember 1992-NUREG/CR4144 V02P38: EVALUAT OF POTENTIAL SEVERE AC.-

LAM A CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4180: HYDROGEN MIXING STUDIES (HMS): USER'S AnaWe O Com Damage Requency Rorn intamal j

MANUAL Ennts Dunng Noop Operatons.

E (Sections E.9-E.16).

NUREG/CR4144 V02P4: EVALUATON POTENTIAL SEVERE ACCl-LAMBERT,LD.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4154 VOI: EXPERIMENTAL RESULTS FROM CONTAIN-SURRY, UNIT 1Anatysis Of Core Damage Frequency From internal MENT PIPING BELLOWS SUBJECTED TO SEVERE CONDITONS.Results From Bellows Tested in,' Uke-New', ACCOENTNU CR 44 E LU 1 EN l'AL SEVERE ACCl-Conditons.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT LAMBRIGHT.J.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/CR4143 V03: EVALUATION OF POTENTIAL SEVERE ACC3 Events Dunng MdLoop Operations Appendices i DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4144 V03 PI: EVALUATION OF POTENTIAL SEVERE AC-GRAND GULF. UNIT 1.Analysrs Of Core Damage Frequency From in.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT temal Events For Plant Operational State 5 Dunng A Refuelmo Outage.

SURRY, UNIT 1 Analyses Of Core Damage Frequency From Intemal Fres MdLoop Operations Mam R LARA,J.

NUREG/CR 144 V03 P2: EVALUATION POTENTIAL SEVERE AC-NUREG/CR4086: SELECTED FAULT TESTING OF ELECTRONIC ISO.

CIDENTS DURihG LOW POWER AND SHUTDOWN OPERATONS AT LATION DEVICES USED IN NUCLEAR DOWER PLANT OPERATION, SURRY, UNIT 1.A talysis Of Core Damage Frequency From Intemal Free Dunng MdLcop Operatons. Appendices NUREG/CR-4874 V17: PRECURSORS TO POTENTIAL SEVERE CORE LINDENMEIER,C.

DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Man Report And NUREG/CR4232: ASSESSING THE ENVIRONMENTAL AVAILABILITY Appendix A.

OF URANIUM IN SOILS AND SEDIMENTS.

NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1992 A STATUS REPORT.Appereces 8. C, D.

LINK,R.E.

E F, And G.

NUREG/CR-6051: EFFECT3 OF TENSILE LOADING ON UPPER SHELF FRACTURE TOUGHNESS.

LAWRENCE.J.D.

NUREG/CR4278. SURVEY OF INDUSTRY METHODS FOR PRODUC.

LIU,C.

ING HIGHLY RELIABLE SOFTWARE.

NUREG/CP 0138: PROCEEDINGS OF WORKSHOP 1 IN ADVANCED NUREG/CR4294 DESIGN FACTORS FOR SAFETY-CRITICAL SOFT.

TOPICS IN RISK AND RELIAMLITY ANALYSIS Model Uncertanty: Its WARE.

Charactenzaten Ard Quantrhcation.

66 Personal Author index LLOYD,R.C.

MALCHER,L.

NUREG/CR-5830: AUXtLIARY FEEDWATER SYSTEM RISK. BASED IN.

NUREG/CR4236: SEISMIC INVESTIGATONS OF THE HDR SAFETY j

SPECTION GUIDE FOR THE MCGUIRE NUCLEAR POWER PLANT.

PROGRAM. Summary Report.

LOFARO,R.

MALLEN,A.N.

NUREG/CP-0135: WORKSHOP ON ENVIRONMENTAL QUAUFICATION NUREG/CR4200- UNCERTAINTY ANALYSIS OF SUPPRESSION POOL OF ELECTRIC EQUIPMENT.

HEATING DURING AN ATWS IN A BWR-5 PLANT.An Applicaton Of NUREG/CR-5939: THE EFFECTS OF AGE ON NUCLEAR POWER The CSAU Methodology Usmg The BNL Ergpneenng Plant Analyzer.

PLANT CONTAINMENT COOUNG SYSTEMS.

MANDLER.J.W.

LOHRSTORFER,C.

NUREG/CR4164: RELEASE OF RADIONUCLIDES AND CHELATING NUREG/CR4203: VALIDATION STUDIES FOR ASSESSING UNSATU-AGENTS FROM CEMENT-SOLIDIFIED DECONTAMINATION LOW-RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

LEVEL RADIOACTIVE WASTE COLLECTED FROM THE PEACH BOTTOM ATOMIC POWER STATION UNIT 3.

@8' REG 1511:

I-NUREG/CR4201: COMPRESSON AND IMMERSON TESTS AND NU REACTOR PRESSURE VESSEL STATUS REPORT.

LEACHING OF RADIONUCLOES. STABLE METALS, AND CHELATING AGENTS FROM CEMENT-SOUDiFIED DECONTAMINATION WASTE LOOMIS,0.

NUREG/CR4147 V01: CHARACTERl2ATON OF CLASS A LOW-LEVEL COLLECTED FROM NUCLEAR POWER STATIONS.

RADOACTIVE WASTE 1986-1990.Executwo esam a i

^

NUREG/CR 3950 V09: FUEL PERFORMANCE REPORT FOR 1991.

R WASTE 986

-Part A.

NUREG/CR4147 V03: CHARACTER 12AT OF Cl. ASS A LOW LEVEL C.W NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPING AND PIPING NU CR4 7V OF A LOW-LEVEL WELDS.Serniennual Report. October 1992. March M RADIOACTIVE WASTE 1

A-E.

NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE NUREG/CR4147 V05:

ACT R OF CLASS A LOW-LEVEL STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPING AT RADIOACTIVE WASTE 1986-1990.

F.

NUREG/CR4147 V06: CHARACTER TlON OF CLASS A LOW-LEVEL LWR TEMPERATURES.

NUREG/CR4233 V01: STABIUTY OF CRACKED PIPE UNDER INER.

RADCACTIVE WASTE 1986-1190. Appendices G-J.

NUREG/CR4147 V07: CHARACTERl2ATION OF CLASS A LOW-LEVEL TIAL STRESSES. Subtask 1.1 Fmal Report.

RADCACTIVE WASTE 19861990. Appendices K-P' MWA LOPRESTI,F.

NUREG/CR4290 KEY ANALYSl? oYSTEM USER'S GUIDE. Version 2.0.

NUREG-1471: CONCEPT OF OPERATIONS WITH ORGANIZATON

• • P ""*

' CR4147 V01: CHARACTER 12ATON OF CLASS A LOW-LEVEL N

/

LORENZ,R.A.

RADIOACTIVE WASTE 19861990 Esecutive Summary.

NUREG/CR4077; DATA

SUMMARY

REPORT FOR FISSON PRODUCT NUREG/CR4147 V02 CHARACTER 12ATION OF CLASS A LOW-LEVEL RELEASE TEST Via' RADIOACTIVE WASTE 19861990.Mac Repor.t.Part A.

NUREG/CR4147 V03: CHARACTERt2ATION OF CLASS A LOW-LEVEL LOWRY,W.

RADOACTIVE WASTE 1986-1990.Mam Report-Part B.

NUREG/CR-6103: PRORIT12ATON OF REACTOR CONTROL COMPO.,

NUREG/CR4147 V04: CHARACTER 12ATION OF CLASS A LOW-LEVEL N

S SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEQUENCE OF R

CT W T 1 A

RADOACTIVE WASTE 1986 1990.

x F.

LU,P.Y.

NUREG/CR4147 V06: CHARACTER TION OF CLASS A LOW LEVEL NUREG/CR-5569 RO1: HEALTH PHYSICS POSITIONS DATA BASE.

RAD OACTIVE WASTE 1986-1190 Appendices G.J.

NUREG/CR-6204: OUESTIONS AND ANSWERS BASED ON REVISED NUREG/CR4147 V07: CHARACTERMATION OF CLASS A LOW-LEVEL 10 CFR PART 20.

RADOACTIVE WASTE 19861990. Appendices K-P.

LUCKAS W.

MAYFIELD M.

NUREG/CR4093: AN ANALYSIS OF OPERATONAL EXPERIENCE NUREG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD.

DRESSING HUMAN REUABluTY ASSESSMENT ISSUES.

N RE 4 228: PREUMINARY ASSESSMENT OF THE FRACTURE LUIS,S.

BEHAVOR OF WELD MATERIAL IN FULL THICKNESS CLAD NUREG/CR4063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS BEAMS.

CRUCES TRENCH SITE.

LURIE D.

NUREG/CR4249 UNIRRADIATED MATERIAL PROPERTIES OF MID-NUREG-1475: APPLYING STATISTICS.

LAND WELD WF 70.

LYNCH,J.

MCCONNELL,J.W.

NUREG/CR4143 V03: EVALUATION OF POTENTIAL SEVERE ACCl.

NUREG/CR 5229 V06: FIELD LYSIMETER INVESTIGATIONS: LOW-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From in.

FISCAL YEAR 1993 Annual Report.

ternal Events For Plant Operational State 5 Dunng A Refuehng Outage.

NUREG/CR4188 VO1: MICROBIAL DEGRADATION OF LOW-LEVEL RADIOACTIVE WASTE. Annual Report For FY 1993.

MACKINNON JA.

NUREG-1022 RO1 DR FC: EVENT REPORTING GUIDEUNES MCCONNELL,KJ.

10CFR50.72 AND 50.73.Second Draft For Comment.

NUREG-1494: STAFF TECHNICAL POSITION ON CONSIDERATON OF FAULT DISPLACEMENT HAZARDS IN GEOLOGIC REPOSITORY MADNt,LK.

DESIGN.

NUREG/CR-5850: ANALYSIS OF LONG-TERM STATION BLACKOUT WITHOUT AUTOMATIC DEPRESSUR12ATION AT PEACH BOTTOM MCCONNELL,V.S.

USING MELCOR (VERSION 1.8).

NUREG/CR4221: THE VALLES NATURAL ANALOGUE PROJECT.

MAOUIRE-MOFFITT MCDUFFIE,P.N.

NUREG/CR-5973 R01: CODES AND STANDARDS AND OTHER 0U0 NUREG/CR4174 V1 DFC: REVISED ANALYSES OF DECOMMISSION-ANCE CITED IN REGULATORY DOCUMENTS.

ING FOR THE REFERENCE BOluNG WATER REACTOR POWER STATION Effects Of Current Regulatory And Other Consadorations On WALARO,J.C.

The Fmancial Assurance Requirements Of The Decommisssonmg Rule NUREG-1500- WORKING DRAFT REGULATORY GUIDE ON RELEASE And....

CRITERIA FOR DECOMMISSONING: NRC STAFF'S DRAFT FOR NUREG/CR4174 V2 DFC: REVISED ANALYSES OF DECOMMISSION.

COMMENT.

ING FOR THE REFERENCE BOluNG WATER REACTOR POWER

I Personal Author index 67 STATION. Effects Of Current Regulatory And Other Consdershons On NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC.

The Fmancial Aneurance Requrements Of The Decommissenmg Rule CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT And.

GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From Ir>

MCGARRY,E.D.

tornal Events For Plant Operational State 5 Dunng A Refueleg NUREG/CR4206 TRANSPORT CALCULATONS OF RADIATION EX-Outage.Secton 10.

POSURE TO VESSEL SUPPORT STRUCTURES IN THE TRO,lAN & E-NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC.

A OR CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Dema0e Frequency From in.

MCOUens a a tornal Events For Plant Opershonal State 5 Dunng A Refuelmg NUREG-1492 DFC: REGULATORY ANALYSIS ON CRITERIA FOR THE OutageMain Report RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE MATERIALDraft Report For Comment.

MINANCW MCKAY,M.K.

NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Man Report And NUREG/CR-6116 V07: SYSTEMS ANALYSIS PROGRAMS FOR Appendix A.

HANOS-ON If(TEGRATED RELIABluTY EVALUATIONS (SAPHIRE)

VERSION 5.0. Fault Tree, Event Tree, And Pipeg & Instrumentaten NUREG/CR 4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Appendices B, C, D, Degram (FEP) Editors Reference Manual.

E, F. And G.

MCKENNA,TJ.

NUREG/CR-4674 Vf 9: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-5247 V01 R2 RASCAL VERSION 2.1 USER'S GUIDE.

DAMAGE ACCIDENTS: 1993 A STATUS REPORT. Main Report And NUREG/CR-5247 V02 R2 RASCAL VERSION 2.1 WORKBOOK.

Appendices A-0.

MCLAUGHLIN.D-NUREG/CR-4674 V20 PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Appendences E NUREG/CR-5965: MODEUNG FIELD SCALE UNSATURATED FLOW And F.

AND TRANSPORT PROCESSES.

NUREG/CR-8063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS CRUCES TRENCH SITE.

MffCHELL.D.

NUREG/CR 6143 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-MCUMBER L.M.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4133: FRAGMENTATON AND OUENCH BEHAVOR OF GRAND GULF, UNIT 1 Analyses Of Core Damage Frequency From In-CORIUM MELT STREAMS IN WATER.

temal Events For Plant Operatonal State 5 Dunno A Refuelmg NUREG/CR4168: DIRECT CONTAINMENT HEATING INTEGRA! EF-Outage. Sections 10.

FECTS TESTS AT 1/40 SCALE IN ZION NUCLEAR POWER FLANT NUREG/CR4143 V02PIB: EVALUATION OF POTENTIAL SEVERE AC.

GEOMETRY.

COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-1501 DRFT: BACKGROUND AS A RESI UAL RADOKTIVITY

. 'n D

r CRITERION FOR DECOMMISSONING. Appendix A To The l'iraft Go-nonc Envronmental trnpact Statement in Support Of Rulemakir.g On NUREd/CR-6143 V02PIC: EVALUATON OF POTENTIAL SEVERE AC-Radological Cntene For Decommessonog Of NRC.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF UNIT 1. Analysis Of Core Damage Frequency From In-MEDHEKAR.S.

ternal Events For Plant Operational State 5 Dunng A Refueling NUREG/CR-5980: STEAM EXPLOSIONS: FUNDAMENTALS AND EN.

Outage.Mam Report GERGETIC BE 4AVIOR.

MITCHELL.M.

MEINHOLD.C.S.

NUREG/CR4.12 DRF FC: IMPACT OF REDUCED DOSE LIMITS ON NUREG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

NRC UCFASED ACTIVITIES. Mabr issues in The implementaten Of MODRO.S.M.

ICRP/NFRP Dose limit Recommendations. Draft Report For Comment.

NUREG/CR4160-

SUMMARY

OF lMPORTANT RESULTS AND SCDAP/

MELSER,3.

FIELAP5 ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP.2, NUREG/(,R4122-STAFFING DECISON PROCESSES AND ISSUE A Case Studies of Seven U S. Nuclear Power Plants.

goppgyy'y'g, NUREO,'CR4123: AN INTERNATIONAL COMPARISON OF COMMER.

NUREG/CR 5830: AUXIUARY FEEDWATER SYSTEM RISK-BASED IN.

Of" NUCLEAR POWER PLANT STAFFING REGULATIONS AND SPECTION GUIDE FOR THE MCGUlRE NUCLEAR POWER PLANT.

PRACTICE.1980-1990.

MOFFITT,R.

MEYER,R.0.

NUREG/CR-5758 V04: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG 1502 ASSESSMENT OF DATABASES AND MODEUNG CAPA-BluTIES FOR THE CANDU 3 DESIGN.

INDUSTRY. Annual Summary Of Program Performance Reports CY 1993.

MICHAUD,W.F.

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN MOHA"'R*

LIGH1 WATER REACTORS. Senuannual Report, April 1993 - Septem.

NUREG/CR-4599 V03 N1 SHORT CRACKS IN PtPING AND PLPING ber 1993.

WELDS.Somiannual Reoort. October 1992 March 1993.

NUREG/CR4142-TENSILE-PROPERTY CHARACTERIZATON OF NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE THERMALLY AGED CAST STAINLESS STEELS.

STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PtPING AT MIUCl,T.

LWR TEMPERATURES.

NUREG/CR4157: SURVEY AND EVALUATON OF AGING RISK AS-MONTELEONE,S.

SESSMENT METHOOS AND APPUCATIONS.

NUREG/CP.0133 V01: PROCEEDINGS OF THE TWENTY-FIRST MIRER.K.M.

WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-NUREG 1501 DRFT: BACKGROUND AS A RESIDUAL RADCACTIVITY sion; Advanced Reactor Research; Advanced Control System Technol-CRITERlON FOR DECOMMISSONING Appendix A To The Draft r'e.

ogy; Advanced instrumentaten & Control Hardware; Human Factors..

neric Envronmental impact Statement in Support Of Rulemakog On NUREG/CP-0133 V02-PROCEEDINGS OF THE TWENTY FIRST Radological Cntena For Decommessenmg Of NRC...

WATER REACTOR SAFETY INFORMATION MEETING. Severe Acci-dont Research.

MILLER,S.

NUREG/CP-0133 V03: PROCEEDINGS OF THE TWENTY FIRST NUREG/CR-6143 V02 PIA: EVALUATION OF POTENTIAL SEVERE AC.

WATER REACTOR SAFETY INFORMATON MEETING. Primary CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT System Integrity: Agm0Research, Products & Applicetone: Structural &

GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From in.

Seisnuc Engmeenng; seismology & Geology ternal Events For Plant Operatonal State 5 Dunng A Refuelang NUREG/CP-0139 TRANSACTIONS OF THE TWENTY.SECOND WATER Outage.Sectons 14 REACTOR SAFETY INFORMATON MEETING.

68 Personal Author index MOORE,C.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/C45680 VOI: THE IMPACT OF ENVIRONMENTAL CONDI-Events Dunng M4 Loop Operations.Appereces E (Sectons E.1 E.8).

TONS ON HUMAN PERFORMANCE. A Handbooit Of Environmental NUREG/CR4144 V02P30: EVALUATION OF POTENTIAL SEVERE AC-Exposures.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-5680 V02: THE IMPACT OF ENVIRONMENTAL CONDl-SURRY. UNIT 1. Analysis Of Core Damage Frequency From Internal i

TIONS ON HUMAN PERFORMANCE. A Cntcal Review Of The Litera-Events Dunng M4 Loop Operations Appereces E (Sections E.9-E.16).

ture.

NUREG/CR-6144 V02P4: EVALUATION OF POTENTIAL SEVERE ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1.Anafysts Of Core Damage Frequency From internal G-368: PREAPPLICATION SAFETY EVALUATION REPORT FOR THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

NU G CR 44 EV LU ION ENTIAL SEVERE ACCl-LIQUID-METAL REACTOR. Final RW DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intemal MOORE,R.H.

NUREG-1475: APPLYING STATISTICS.

Events During Mid-Loop Operations.Apperdcas t NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-MOORE S.E.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-5359 FIEVIEW OF ELASTIC STRESS AND FATIGUE TO-SURRY, UNIT 1 Analysis Of Core Damage Frequency From internal FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA-Fires Dunno M4 Loop Operatons. Main Report.

TION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER P1PtNG NUREG/CR4144 V03 P2 EVALUATION OF POTENTIAL SEVERE AC-SYSTEMS.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SWRY,W LAnaW G Cm Damage Ngueng M inW MORISSEAU.D.

Fires During M4 Loop OperatonsApperdcas.

NUREG/CH-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDl-TIONS ON HUMAN PERFORMANCE. A HandbotA Of Environmental NANSTAD,R.K.

C$5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI-NUREG/CR4139: CRACK. ARREST TESTS ON TWO IRRADtATED HIGH-COPPER WELDS. Phase 11: Results Of Duplex-Type Specimens.

TIONS ON HUMAN PERFORMANCE. A Cntical Review Of The Litera.

NUREG/CR.6249. UNIRRADIATED MATERIAL PROPERTIES OF MO-ture.

LAND WELD WF-70.

MORRIS,R.H.

NUREG/CR-4674 VIT: PRECURSORS TO POTENTIAL SEVERE CORE NEBUDA.D.T.

DAMAGE ACCOENTS: 1992 A STATUS REPORT. Main Report And NUREG/C46190 V01: PROTECTON AGAINST MALEVOLENT USE OF 1

Apperdx A.

VEHICLES AT NUCLEAR POWER PLANTS. Vehicle Bamer System NUREG/CR-4674 V18 PRECURSORS TO POTENTIAL SEVERE CORE Sitino Guidance For Blast Protection.

j DAMAGE ACCIDENTS: 1992 A STATUS REPORTAppereces B, C D, NUREG/CR4190 V01 R1: PROTECTION AGAINST MALEVOLENT USE i

E. F, And G.

OF VEHICLES AT NUCLEAR POWER PLANTS.Vehole Bamer System Selection Guidance For Blast Protection.

N EM6m M NECW AWM MNM RM &

EG CP-0138: PROCEEDINGS OF WORKSHOP 1 IN ADVANCED C

E O

PWSNM h System TOPICS IN RISK AND RELIABILITY ANALYSIS.Model Uncertainty Its p g harmten And hton.

NUREG/CR6190 V02 R1: PROTECTION AGAINST MALEVOLENT USE MUBAYI,V.

OF VEHICLES AT NUCLEAR POWER PLANTS.Vehele Bamer System NUREG/CR-6094. CALCULATIONS IN SUPPORT OF A POTENTIAL Selection Guidance DEFINITION OF LARGE RELEASE.

MULLEY,G.

NUREG/CR4185: TMI-2 INSTRUMENT NOZZLE EXAMINATONS AT NUREG-1416 V07 N01: OFFICE OF THE INSPECTOR ARGONNE NATIONAL LABORATORY. February 1991 - June 1993.

GENERALSemiannual Report, April 1. September 30,1994.

NUREG/CR-6187: RESULTS OF MECHANICAL TESTS AND SUPPLE-MENTARY MICROSTRUCTURAL EXAMINATIONS OF THE TMI-2 M AW,M LOWER HEAD SAMPLES NUREG/CR6126: COGNITIVE SKILL TRAINING FOR NUCLEAR NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE.

POWER PLANT OPERATONAL DECISION-MAKING.

GRATION REPORT.

NUREG/CR4127: THE EFFECTS OF STRESS ON NUCLEAR POWER PLANT OPERATONAL DECISION MAKING AND TRAINING AP.

NEUMAN,5 P NUREG/CN4203: VALIDATON STUDIES FOR ASSESSING UNSATU.

NUR G CR ICA f S ATON OF OPERATOR PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER.

RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

GENCIES.

NEYMO m MURFIN.W.B.

NUREG/CR6053: COMPARISON OF MACCS USERS CALCULATONS NUREG/CR-4551 V01 R1: EVALUATION OF FEVERE ACCIDENT FOR THE INTERNATIONAL COMPARISON EXERCISE ON PROBABI-RISKS. METHODOLOGY FOR THE CONTAINMENT. SOURCE LISTIC ACCIDENT CONSEQUENCE ASSESSMENT CODES.

TERM.CONSEOUENCE, AND RISK INTEGRATION ANALYSES.

NICHOLS,R.T.

MURPHY,W.M.

NUREG/CR-6044: EXPERIMENTS TO INVESTIGATE DIRECT CON.

NUREG/CR4288: GEOCHEMICAL INVESTIGATIONS RELATED TO TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NUCLE

• ZON NUCLEAR POWER PLANT IN THE SURTSEY TEST FACILITY.

AR WASTE REPOSITORY.

NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DIRECT CON-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE MUSICKI,Z.

SURRY NUCLEAR POWER PLANT.

NUREG/CR4144 V02P1 A. EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT REG CR-5973 Rot: CODES AND STANDARDS AND OTHER GUO.

s oop aio s aan R ESE RE AC.

ANCE CITED IN REGULATORY DOCUMENTS.

NUREG/CR4144 V02PIB ALUATON OF T COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NORTON,L.

SURRY, UNIT 1 Analysis Of Core Damage Frequency From Intemal NUREG-1415 V06 NO2 OFFICE OF THE INSPECTOR Events Dunng M6 Loop Operations Main Report (Chapters 7-12).

NUREG/CR4144 V02P2. EVALUATON OF POTENTIAL SEVERE ACCl-GENERALSemiannual Report October 1,1993. March 31,1994.

DENTS DURING LOW POWER AND SHUTDOWN OPEFiATIONS AT SURRY. UNIT 1 Analyses Of Core Damage Frequency From Intemal NORTON,M.V.

Events During Mid-Loop Operatons Apperd'cet A-D_

NUREG/CR4289: RECONCENTRATION OF RADIOACTIVE MATERIAL NUREG/CR6144 V02P3A EVALUATION OF POTENTIAL SEVERE AC.

RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR CIDENTS DURING LC# POWER AND SHUTDOWN OPERATONS AT PART 20.

1 Personal Author index 69 j.

NOvaK.s.D.

PARKS,M.s.

j NUREG/CR4088:

SUMMARY

OF 1991 1992 MISADMINISTRATION NUREG/CR4154 V01: EXPERIMENTAL HESULTS FROM 'CONTAIN-

- EVENT INVESTIGATIONS.

MENT PIPING BELLOWS SUBJECTB) TO SEVERE ACCIDENT NUREG/CR4145; VER6FICATION AND VALCATION OF THE SAPHIRE CONDITIONS.Resulte From Bellows Tee.ed in "Like-New" Condelsons.

VERSION 4.0 PRA SOFTWARE PACKAGE.

4 PAUL.D.D.

j-NOWWN,8.

NUREG/CR4103: PRORITIZATON OF REACTOR CONTROL COMPO- '

NUREG/CR-5128 R01: EVALUATION AND REFINEMENT OF LEAK.

RATE ESTIMATION MODELS.

NENTS SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEQUENCE OF AGING-PAYNE.A.C.

NUREG/CR-4838: MICROCOMPUTER APPLICATIONS OF, AND MODI-CR4287: AIR. WATER SIMULATION OF PHENOMENA OF J

CORIUM DISPERSION IN DIRECT CONTAINMENT HEATING.

PAYNE.G.A.

NOREG/CR-3950 V00: FUEL PERFOROANCE REPORT FOR 1991.

O'Num I,

NUREG/CR4918 V0h CONTROL OF WATER INFILTRATION INTO PENNELL,W.E. -

NEAR SURFACE LLW DISPOSAL UNITS.Progrees Report On Field Ex-4 NUREG/CR4219 V10 N1: HEAVY f.ECTION STEEL TECHNOLOGY penments At A Humid Region $de.BeltsWie.Marpend.

PROGRAM.Sennennual Progrees Roport For October 1992

• March O'HARA,J.M.

1993.

NUREG0711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW PENOYARJ.

j NUR G/CR-5908 V01: ADVANCED HUMAN-SYSTEM INTERFACE NUREG/CR-5994: EMERGENCY DEf.EL GENERATOR: MAINTENANCE DESIGN REVIEW GUOEUNE. General Evaluation Model, Technical AND FAILURE UNAVAILA81UTY, A ND THEIR RISK IMPACTS.

P E FCR5 VO V

H'UMAN-SYSTEM INTERFACE DESIGN REVIEW GUIDEUNE. Evaluation Procedures And Guidelines NUREG/CR-5800 V01: THE IMPAm OF ENVIRONMENTAL CONDI-For Human Factors E Review 1L-TIONS ON HUMAN PERFORMANCE. A Handbook of Envuonmental NUREG/CR4105:

F ORS ENGINEERING GUIDANCE FOR Exposures.

THE REVIEW OF ADVANCED ALARM SYSTEMS' NUREG/CR-5880 V02: THE IMPACT OF ENVIRONMENTAL CONDI-NUREG/CR4148: LOCAL CONTROL STATICNS: HUMAN ENGINEER-TiONS ON HUMAN PERFORMAiCE. A Cnlical Review Of The Uters-d ING ISSUES AND INSIGHTS.

ture.

OLSON,R.J.

PERSONS,W.L NUREG/CR4233 V01: STA8fuTY OF CRACKED PlPE UNDER INER.

NUREG/CR4278: SURVEY OF IfiDUSTRY METHODS FOR PRODUC-TIAL STRESSES. Subtask 1.1 Final Report.

ING HIGHLY REUABLE SOFTW ARE.

J NUREG/CR4234: VAUDATION OF ANALYSIS METHODS FOR AS.

SESSING FLAWED PIPING SUBJECTED TO DYNAMC LOADING.

PETERSON.LE.

NUREG/GR4011: INFORMATCN BIAS ANO LIFETIME MORTAUTY I

s j

00SDRNE,M.F.

RISKS OF RADIATION-INDUCID CANCER. Low LET Radiatiort J

NUREG/CR4077: DATA

SUMMARY

REPORT FOR FISSION PRODUCT j

RELEASE TEST Vl4.

PILCH,M.M.

NUREG/CR4044: EXPERIMENTS TO INVESTIGATE DRECT CON.

OSTROM L.T.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE

]

NUREG/CR4088:

SUMMARY

OF 1991 1992 MISADMINISTRATION ZION NUCLEAR POWER PLANT IN THE SURTSEY TEST FACILITY.

. EVENT INVESTIGATIONS.

NUREG/CR4075: THE PROSABluTY OF CONTAtNMENT FAILURE BY.

2 4

DIRECT CONTAINMENT HEATING IN ZION.

i OTEROA NUREG/CR4075 S01: THE PROBABluTY OF CONTAINMENT FAIL.

NUREG/lA-0093: RELAPS/ MOD 3 ASSESSMENT FOR CALCULATION URE BY DIRECT CONTAINMENT HEATING IN ZON.

OF SAFETY AND REUEF VALVE DISCHARGE PlPING HYDRODY-NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DIRECT CON-l NAMC LOADS.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE SURRY NUCLEAR POWER PLANT.

7 l

NUREG/CR4288, GEOCHEMICAL INVESTIGATONS RELATED TO PITTIGUO.L '

j THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NUCLE-NUREG-1488: FINAL SAFETY EVALUATION REPORT TO UCENSE AR WASTE REPOSITORY.

THE CONSTRUCTION AND OPERATION OF A FACluTY TO i

PAPFORDAA RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL

)

I A C dom M 404WEnecare W Mn4 NUREG/CR4158: IMPUCATIONS FOR ACCOENT MANAGEMENT OF ADDING WATER TO A DEGRADtNG REACTOR CORE.

PLANSKY,L.E.

?

NUREG/CR4138: USER'S GUIDE FOR SIMPLIFIED COMPUTER PAGE,G.

NUREG/CR4158:

SUMMARY

OF COMMENTS RECElVED FROM MODELS FOR THE ESTIMATION OF LONG TERM PERFORMANCE WORKSHOPS ON RADIOLOGCAL CRITERIA FOR DECOMMISSION-OF CEMENT-BASED MATERIALS.

N G/CR4250

SUMMARY

OF COMMENTS RECEIVED ON STAFF 1

DRAFT PROPOSED RULE ON RADIOLOGCAL CRITERIA FOR DE-NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC-j COMMISSONING.

TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING.

I PAINTER.C.L POWERSAA.

1 NUREG/CR-3950 V00 FUEL PERFORMANCE REPORT FOR 1991.

NUREG/CR4218: A REVIEW OF THE TECHNICAL ISSUES OF AIR IN-GRESSION DURtNG SEVERE REACTOR ACCIDENTS.

t PARK,W.R.

5 NUREG/CR-5985: EVALUATION OF COMPUTER-BASED ULTRASONC PRECKSHOT,G.G.

INSERVICE INSPECTION SYSTEMS.

NUREG/CR4294: DESIGN FACTORS FOR SAFETY 4RITICAL SOFT.

WARE.

PARK,YJ.

NUREG/CR4303: METHOD FOR PERFORMING DIVERSITY AND DE.

NUREG/CR4241: TECHNICAL GUIDEUNES FOR ASEISMC DESIGN FENSE-IN-DEPTH ANALYSES OF REACTOR PROTECTION SYS-OF NUCLEAR POWER PLANTS.Trenelation Of JEAG 4801 1987.

TEMS.

PARKS.C.V.

PUROMIT,A.

NUREG/CR-5825: TECHNCAL SUPPORT FOR A PROPOSED DECAY NUREG/CR4185 TMI-2 INSTRUMENT NOZZLE EXAMINATIONS AT HEAT GUIDE USING SAS2H/ORIGEN-S DATA.

ARGONNE NATIONAL LABORATORY. February 1991. June 1993.

J l

4

70 : Personal Author index mas er a NUREG/CR-4639 V5R4P3: NUCLEAR COMPUTERIZED UBRARY FOR NUREG/CR4158: IMPUCATIONS FOR ACCIDENT MANAGEMENT OF ASSESSING REACTOR REUABluTY (NUCLARR). Volume 5: Data ADDING WATER TO A DEGRADING REACTOR CORE.

Manual.Part 3: Herdeere Component Failure Deta.

RAlet4N,S.

RtDEVAW.

NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPtNG AND PIPING NUREG/CR-4918 V0h CONTROL OF WATER INFILTRATION INTO WELDS. Semiannual Report, October 1992. March 1993.

NEAR SURFACE LLW DISPOSAL UNITS.Progrees Report On Field Ex.

penments At A Humid Region Ser.Sellevise Maryland.

NUREG 1416-OPERATIONAL EXPERIENCE AND MAINTENANCE PRO-ROCKHOLD,tL GRAMS OF TRANSAMERICA DELAVAL, INC., DIESEL GENERA.

NUREG/CR-8063: INTRAVAL PHASE il MODEL TESTING AT THE LAS TORS.

CRUCES TRENCH SITE.

RAIM-8EN A-R00ASAUGH,E.C.

NUREG/CR4093: AN ANALYSIS OF OPERATIONAL EXPERIENCE NUREG/CR-5359: REVIEW OF ELASTIC STRESS AND FATIGUE-TO-DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD.

FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA-DRESSING HUMAN REUABlWTY ASSESSMENT ISSUES.

. TON TO ASME DESIGN CRITERIA FOR NUCLEAR POWER PlPING SYSTEMS.

pamaanse n.i y, NUREG/CR-5247 V01 R2 RASCAL VERSON 2.1 USER'S GUIDE.

ROGERSAD.

RA0000M,V.M.

NUREG/CR-5229 V06: FIELD LYSIMETER INVESTIGATIONS: LOW-NUREG/CR4535 V06: RELAPS/ MOD 3 CODE MANUALValidaeon Of LEVEL WASTE DATA BASE DEVELOPMENT. PROGRAM FOR TN in RELAP5A4003.

. NU G C 4 86 DEGRADATION OF LOW-LEVEL RAO,D.V.

RADIOACTIVE WASTE. Annual Report For FY 1993.

NUREG/CR4224 DFC: PARAMETRO STUDY OF THE POTENTIAL DE S Dr p EG-1486: FINAL SAFETY EVALUATION REPORT TO UCENSE THE CONSTRUCTON AND OPERATION OF A FACluTY TO RAO.M.C.

RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NUREG/CR4132-BIAXIAL LOADING AND SHALLOW. FLAW EFFECTS NEAR CUVE. UTAH. Docket No. 40-8969.(Erwirocere of Utah,Inc.)

ON CRACK-TIP CONSTRAINT AND FRACTURE TOUGHNESS.

R00ENFIELD,A.

mansanam n as NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPtNG AND PIPING NUREG/CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR

. WELDS. Semiannual Report, October 1992. March 1993.

HANDS-ON INTEGRATED REUABluTY EVALUATONS (SAPHIRE)

VERSION 5.0. Technical Reference Manual.

ROSS.S.

i NUREG/CR4116 V02-SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4143 V03: EVALUATION OF POTENTIAL SEVERE ACCl-HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT VERS 8ON 5.0. Integrated Reliebety And Rask Analyse System (IRRAS)

GRAND GULF UNIT 1.Analysie Of Core Damage Frequency From In-Roterance Marmal.

- tornal Events For Plant Operational State 5 Dunng A Refuehng Outage.

pamasinessel T.C.

ROTH.E.al.

NUREG/CR4203: VAUDATION STUDIES FOR ASSESSING UNSATU-NOREG/CR4126: COGNITIVE SKILL TRAINING FOR NUCLEAR RATED FLOW ANO TRANSPORT THROUGH FRACTURED ROCK.

POWER PLANT OPERATIONAL DECISION-MAKING.

NUREG/CR4206: AN EMPIRICAL INVESTIGATION OF OPERATOR RAVINDRA M.0L PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER-NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABluSTO GENCIES..

RISK ASSESSMENT.

NUREG/CR4143 V05: EVALUATION OF POTENTIAL SEVERE ACCl-ROUSSEL.A.

i DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4122- ' STAFFING DECISON PROCESSES AND

)

GRAND GULF, UNIT 1.Analyes Of Core Dama9e Frequency From ISSUES. Case Studies Of Seven U.S. Nuclear Power Plants.

Saamic Events Dunng Mid-Loop Operations. Main Report.

NUREG/CR4144 V05: EVALUATION OF POTENTIAL SEVERE ACCl-RUFF,L, -

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-3145 V10- GEOPHYS4 CAL'. INVESTIGATIONS OF THE SURRY UNIT 1.Analyes Of Core Dama9e Frequency From Seismic WESTERN OHIOINDIANA REGION.Faial Report, October 1986-Sep-Events Dunng Mid-Loop Opershons. Main Report.

tomber 1992.

REECE.WJ.

RUIZ,80.

NUREG/CR-4639 V5R4P2 NUCLEAR COMPUTERIZED UBRARY FOR NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL ASSESSING REACTOR REUABluTY (NUCLARR). Volume 5: Data FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

NUR P3 NUCL ZED UBRARY FOR ASSESSING REACTOR REUABluTY (NUCLARR). Volume 5: Data RUSSELL,ILO.

Manual.Part 3: Hardware Component Failure Data.

NUREG/CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHtRE) i REMPE A NUREG CR 1 02 S TNS ANALYSIS PROGRAMS FOR NUREG/CR4196: CALCULATIONS TO ESTIMATE THE MARGIN TO N

G 1

M VES l' INVESTIGATION PROJECT INTE-HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHtRE)

GRATION REPORT,.

. VERSON 5.0. Integrated Rehability And Riek Analyes System (IRRAS)

Reference Manual.

REINER AP.

NUREG/CR4116 V03: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR 5625: TECHNICAL SUPPORT FOR A PROPOSED DECAY HANDS-ON INTEGRATED REUABlWTY EVALUATIONS (SAPHIRE)

HEAT GUIDE USING SAS2H/ORtGEN-S DATA.

VERSON 5.0. Integrated Rehabihty And Reak Analyse System (IRRAS)

Tutonal Manual.

REvaamr ast,S.y,

. NUREG/CR4116 V04: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4267: AIR-WATER SIMULATON OF PHENOMENA OF HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHlRE)

CORIUM OtSPERSION IN DIRECT CONTAINMENT HEATING.

VERSION 5.0. Systems Analysis And Riek Aaesamment (SARA) Refer-ence Manual RICHARDS,R.E.

NUREG/CR4116 V05: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR-4639 V5R4P2; NUCLEAR COMPUTERIZED UBRARY FOR HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHsRE)

ASSESSING REACTOR REUABluTY (NUCLARR). Volume 5: Date VERSION 5.0. Systems Analyse And Reek Aseseement (SARA) Tutonal Manuel.Part 2: Human Error Probabsty (HEP) Dete.

Manual.

f I

Personal Author index 71 j

NUREG/CR4116 V08: SYSTEMS ANALYSIS PROGRAMS FOR SCHNEIDER,8.

1 HANDS ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

NUREG-1492 DFC: REGULATORY ANALYSIS ON CRITERIA FOR THE VERSON 5.0 Models And Results Database (MAR-D) Reference RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE Manual MATERIALDraft Report For Comment d

RUTHER,W.E.

SCHRAMMEL D.

NUREG/CR-4667 V17 ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4236: SEISMIC INVESTIGATIONS OF THE HDR SAFETY LIGHT WATER REACTORS. Semiannual Report. April 1993. Septem-PROGRAM. Summary Report.

4 bei1993.

(

SABEK,M.G.

SCHUETZ,8.K.

NUREG/CR4195: EXAMINATION OF RELOCATED FUEL DEBRIS AD.

NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABluSTIC JACENT TO THE LOWER HEAD OF THE TMi-2 REACTOR VESSEL RISK ASSESSMENT, NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE-SAFAR.S.

GAATION REPORT.

NUREG/CR4198. TMI 2 INSTRUMENT NOZZLE EXAMINATIONS PER-NUREG/CR4161: BUCKLING EVALUATION OF SYSTEM 80+(TM)

FORMED AT THE INEL CONTAINMENT.

SCHULL,W.J.

J SALYER,W.D.

NUREG/GR-0011: INFORMATION BIAS AND UFETIME MORTALITY NUREG/CR-4674 V17; PRECURSORS TO POTENTIAL SEVERE CORE RISKS OF RADIATON-INDUCED CANCER. Low LET Radiation.

DAMAGE ACCOENTS: 1992 A STATUS REPORT. Main Report And Apperdx A.

SCHULTZ,R.R.

)

?,"JRt!G/CR-4874 V18: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-5535 V07: 4 LAP 5, MOD 3 CODE MANUALSummanes And DAMAGE ACCOENTS: 1992 A STATUS REPORT.Appendces B C, D.

Reviews Of Independent Code Assessment Reports.

E, F. And G.

SCHULZ,R.K.

2 SAMANTA.P.

NUREG/CR-4918 V07: CONTROL OF WATER INFILTRATION INTO NUREG/CR-5967: DEVELOPMENT AND APPUCATION OF DEGRADA.

NEAR SURFACE LLW DISPOSAL UNITS. Progress Report On Field Ex-TION MODELNG TO DEFINE MAINTENANCE PRACTICES.

penments At A Humid Region Sato.Bettsville. Maryland.

j NUREG/CR-5994: EMERGENCY DIESEL GENERATOR: MAINTENANCE AND FAILURE UNAVAILABILITY, AND THElR RISK IMPACTS.

SCHUSTER,G J.

i' SANDIN,S.

NUREG/CH-5985. EVALUATION OF COMPUTER-BASED ULTRASONIC INSERVICE INSPECTION SYSTEMS.

NUREG 102E A01 DR FC: EVENT REPORTING GUIDEUNES

%%U.72 AND 50.73.Second Draft For Comment.

SCHWARTZ,C.W.

i NUREG/CR-5861: CRACK. SPEED RELATIONS INFERRED FROM SANDS.S.P, LARGE SINGLE-EDGE-NOTCHED SPECIMENS OF A 533 B STEEL.

NUREG-1368. PREAPPLICATON SAFETY EVALUATON REPORT FOR THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

SCIACCA,F.

LIQUO-METAL REACTOR Final Report NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL

]

SANECKl J.E.

FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT-ED DEBRIS.Ctatt For Comment.

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS. Semannual Report,Aprd 1993. Septem.

SCOTT,P.M.

ber 1993.

NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPING AND PIPING SANFORD,W.E.

WELDS.Sern annual Report. October 1992 March 1993.

NUREG/CR-f s28 ROI: EVALUATION AND REFINEMENT OF LEAK-NUREG/CR-5229 V06. FIELD LYSIMETER INVESTIGATIONS: LOW.

RATE ESTMATION MODELS.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR 6233 V01: STABiUTY OF CRACKED PIPE UNDER INER-FISCAL YEAR 1993. Annual Report.

TIAL STF ESSES. Subtask 1.1 Final Report.

SANZO D.

StlTZ R.R.

4 l

NUREG/CR4157: SURVEY AND EVALUATON OF AGING R!SK AS.

M.E/CR4138: USER'S GUIDE FOR SIMPUFIED COMPUTER SESSMENT METHODS AND APPUCATONS.

PQELS FOR THE ESTIMATON OF LONG. TERM PERFORMANCE j

SATO.K.

.lF CEMENT-BASED MATERIALS.

i NUREG/CR4213: HIGH TEMPERATURE HYDROGEN-AIR-STEAM SEUL K.W.

DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP.

NUREG/lA-0114: ASSESSMENT OF RELAP5/ MOD 3 WITH THE LOFT MENT APPAAATUS.

L9-1/L3-3 EXPERIMENT SIMULATING AN ANTICIPATED TRANSIENT SATTISON M.B.

WITH MULTIPLE FAILURES.

NUREG/CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR SHACK W.J.

HANDSON INTEGRATED RELIABILITY EVALUATONS (SAPHIRE)

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN VERSION 5 0.Techncal Reference Manual UGHT Y ATER REACTORS. Semiannual Report.Aprd 1993. Septem.

NUREG/CR4116 V05: SYSTEMS ANALYSIS PROGRAMS FOR ber 1993 i

HANDSON INTEGRATED RELIABluTY EVALUATONS (SAPHIRE)

NUREG/CF 4176: REVIEW OF ENVIRONMENTAL EFFECTS ON FA.

VERSION 5.0 Systems Analyses And Risk Assessment (SARA) Tutonal TiGuti CAACK GR7WTH OF AUSTENITIC STAINLESS STEELS.

i Manual.

NUREG/CR4177: ASSESSMENT OF THERMAL EMBRITTLEMENT OF CAST STAINLESS STEELS.

SCANLON,8.

NUREG/CR4223: REVtEW OF THE PROPOSED MATERIALS OF CON-NUREG/CR4063: INTRAVAL PHASE If MODEL TESTING AT THE LAS STRUCTON FOR THE SBWR AND AP600 ADVANCED REAR TORS.

CRUCES TRENCH SITE.

NUREG/CR4237: STATISTICAL ANALYSIS OF FATlGUE STP.# LIFE DATA FOR CARBON AND LOW-ALLOY STEELS.

SCHMfDT,R.

NUREG/CR4233 V01: STABluTY OF CRACKED PIPE UNDER INER.

SHAFFER.C.

TIAL STRESSES Subtask 1.1 Final Report.

NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL SCHMIDT,R.C.

FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

ED DEBRtS Draft For Comment.

NUREG/CR4218 A REVIEW OF THE TECHNICAL ISSUES OF AIR IN-j GRESSION DURING SEVERE REACTOR ACCIDENTS.

SHAH,V.N.

NUREG/CR-5314 VOS INSIGHTS FOR AGING MANAGEMENT OF SCHNEtDER,J.P.

MAJOR LWR COMPONENTS METAL CONTAINMENTS.

NUREG/CR4133 FRAGMENTATON AND QUENCH BEHAVOR OF NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC.

CORIUM MELT STREAMS IN WATER.

TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING.

1 i

72 Personal Author index SHEARER,TL SURRY, UNIT 1. Analysis Of Core Damage Frequercy From internal NUREG/CR4185: TMl-2 INSTRUMENT NOZZLE EXAMINATIONS AT Fires Dunng Mid-Loop Operations. Appendices.

ARGONNE NATIONAL LABORATORY. February 1991 - June 1993.

NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE-SJOREEN,AL GRATION REPORT.

NUREG/CR-5247 V01 R2 RASCAL VERSION 2.1 USERS GUIDE.

NUREG/CR 5247 V02 R2: RASCAL VERSION 2.1 WORK 9OOK.

SHELT M NUREG-1460 R01: GUIDE TO NRC REPORTING AND RECORDKEEP*

SKINNER,NL ING REQUIREMENTS. Compiled From Requrements in T/le 10 Of NUREG/CR4116 V02: SYSTEMS ANALYSIS PROGRAMS FOR The U.S. Code Of Federal Regulations As Codifed On December 31' HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

W3.

VERSION 5.0. Integrated Rehabdity And Risk Analysis System (IRRAS)

Reference Manual SHENG,8.

NUREG/CR4116 V03: SYSTEMS ANALYSIS PROGRAMS FOR NUREG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

HANDSON INTEGRATED RELIABluTY EVALUATIONS (SAPHIRE)

SHERFEY,LL VERSION 5.0. Integrated Rehabiltly And Risk Analyses System (IRRAS)

NUREG/CR 5973 R01: CODES AND STANDARDS AND OTHER GUID-Tutonal Manual.

ANCE OTED IN REGULATORY DOCUMENTS.

NUREG/CR4116 V04: SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

UR CR-5535 V06: RELAP5/ MOD 3 CODE MANUALVakdation Of y

Numencal Techniques m RELAPS/ MOD 3-NUREG/CR41'16 V05: SYSTEMS ANALYS!S PROGRAMS FOR HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

SHTEYNGART,S.

VERSION 5.0. Systems Analysis And Renk Assessment (SARA) Tutonal NOREG/CR4169 RELAY TEST PROGRAM. Seres il Tests. integral Test.

NURF 46116 V07: SYSTEMS ANALYSIS PROGRAMS FOR HAS4C4 UN INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

SIESE,D.A.

NUREG/CR4269-A PLAN FOR THE MODIFICATION AND ASSESS-VEnv.* 5.0. Fault Tree. Event Tree, And Piping & Instrumentat.on MENT OF TRAC.PF1/ MOD 2 FOR USE IN ANALYZING CANDU 3 Diagram (FEP) Editors Refererw:e Manual.

TRANSIENT THERMAL-HYDRAUUC PHENOMENA.

NUREG/CR4116 V08: SYSTEMS ANALYSIS PROGRAMS FOR HANDSON INTEGRATED REUABILITY EVALUATIONS (SAPH;RE)

CR4276: OVAUTY MANAGEMENT IN REMOTE AFTERLOAD-ING BRACHYTHERAPY.

SLAYlCH,A.

Sit 0DNEN,F.A.

NUREG/C45680 VOI: THE IMPACT OF ENVIRONMENTAL CONDI-NUREG/CR4151: FEASlBluTY OF DE4LCWs RISK-BASED RANK-TIONS ON HUMAN PEPfW4ANCE. A Handbook Of Envronmental INGS OF PRESSURE BOUNDARY F ' W FOR INSERVICE IN-Exposures SPECTION NUREG/C45680 V02: THIE f/?ACT OF ENVIRONMENTAL CONDI-NUREG/CR4181: A PILOT APPUCATION OF RISK-BASED METHODS TIONS ON HUMAN PERFORMANCE. A Cntecal Review Of The Litera-TO ESTABUSH INSERVICE INSPECTION PRIORITIES FOR NUCLE-AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWE9 STATION.

ture.

SINHA.U.P.

SLOAN,S.M.

NUREG/CR-5314 V05: INSIGHTS FOR AGING MANAGEMENT OF NUREG/CR-5535 V07: RELAPS/ MOD 3 CODE MANUALSummanes And MAJOR LWR COMPONENTS METAL CONTAINMENTS.

Reviews Of Independent Code Assessment Reports.

SIU,N.

SMcDTS C.

NUREG/CP-0138: PROCEEDINGS OF WORKSHOP l IN ADVANCED NUREG/CP 0138: PROCEEDINGS OF WORKSHOP 1 IN ADVANCED TOPICS IN RISK AND REUABluTY ANALYSIS.Model Uncertainty its TOPICS IN RISK AND RELIABluTY ANALYSIS.Model Uncertainty Its NU G CR41 V02P L TION OF POTENTIAL SEVERE AC-ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SMITH,S.W.

SURRY, UNIT 1. Analysis 01 Core Damage resquency From Intamal NUREG/CR4151: FEANITY OF DEVELOPING RISK-BASED RANK-NR C

44 2 B VL T OF T S

RE AC-CT '

ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1. Analysis Of Core Damage Frequency From intomal StelThC.L.

L$)YlON OF PO Eb SEVdR ACCl-

^

NR C

44 E

VERS 0

PC DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1. Analysis Of Core Damage Frequency Frorn intamal SMITH,D.

NUREG/CR4156:

SUMMARY

OF COMMENTS RECEIVED FROM NR CR 44 A

L POTENTIAL SEVERE AC-WORKSHOPS ON RADIOLOGICAL CRITERIA FOR DECOMMISSION-ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1. Analyses Of Core Demoge Frequency From IntW N

G/CR4250:

SUMMARY

OF COMMENTS RECEIVED ON STAFF LAAFT PROPOSED RULE ON RADIOLOGICAL CRITERIA FOR DE-NR CR 44 02 B L T PO E SE RE C-COMMISSIONM ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1. Analyse Of Core Damage Frequency From Intemal SMITH,L NR CR 4

E lON E

E NUREG/CR-6107:

SUMMARY

OF MELCOR 1.8.2 CALCULATIONS FOR DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT THREE LOCA SEQUENCES (AG.S2D & S3D) AT THE SURRY PLANT.

SURRY, UNIT 1 Analyses Of Core Drenage Frequency From inlemal SMffW NP CR 44 E ALU EN lAL SEVERE ACCg.

NUREGICR4174 V1 DFC: REVISED ANALYSES OF DECOMMISSION-ING FOR THE REFERENCE BOILING WATER REACTOR POWER uEMS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1 Analyes Of Core Demage Frequency From Intemal STATION Effects Of Current Regulatory And Other Conesderations On The inancial Assurance Reparements Of The Decomtrwsesoning Rule c

Events Dunng bad-Loop Operations Appendices 1.

NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC.

And-..

ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4:74 V2 DFC: REVISED ANALYSES OF DECOMMISSION.

SURRY, UNIT 1.Analyss Of Core Damage Frequency From internal ING FOR TF.E FIECQENCE BoluNG WATER REACTOR POWER Free Dunno Mid-Loop Operatons.Masn Report.

STATION. Effects Of Current Regulatory And Other Coneederations On NURFG/CR4144 v03 P2: EVALUATION OF POTENTIAL SEVERE AC-The Financel Assurance Requrements Of The Decommessoning Rule CILuNTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT And...

r

Personal Author index 73 SMITH,SX STEELE,R.

NUREG/CR-5314 V05: INSGHTS FOR AGING MANAGEMENT OF NUREG/CR-5935

SUMMARY

OF WORK COMPLETED UNDER THE MMOR LWR COMPONENTS METAL CONTAINMENTS.

ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUALIFICATION RE-SEARCH PROGRAM (EDOP).

NUREG/CR4196: CALCULATONS TO ESTIMATE THE MARGIN TO STEINHILBER,H.

FAILURE IN THE TMI-2 VESSEL NUREG/CR4236: SEISMC INVESTIGATONS OF THE HDR SAFETY SOS M PROGRAM Summary Report.

NUREG 1488: REVISED LIVERMORE SEISMIC HAZARD ESTIMATES STETAR,E.A.

FOR SIXTY-NINE NUCLEAR POWER PLANT SITES EAST OF THE NUREG/CR4289: FIECONCENTRATION OF RADIOACTIVE MATERIAL ROCKY MOUNTAINS Foal Report RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR SOPPET,W X PAM M NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN STICXLER.LA.

LIGHT WATER REACTORS. Semiannual ReporLApnl 1993 Septem-ber 1993.

NUREG/CR4196: CALCULATONS TO ESTIMATE THE MARGIN TO FAlLURE IN THE TML2 VESSEL NUREG/CR.6142-TENSILE-PROPERTY CHARACTERIZATON OF NUREG/CR 6197: TMI-2 VESSEL INVESTIGATON PROJECT INTE-THERMALLY AGED CAST STAINLESS STEELS.

GRATON REPORT.

SPENCER.B.Vr.

NUREG/CR4133: FRAGMENTATION AND OVENCH BEHAVOR OF STOCK,D NUREG/CR-5967: DEVELOPMENT AND APPLICATON OF DEGRADA-NU 61 RE A MENT HEATING INTEGRAL EF-TION MODELING TO DEFINE MAINTENANCE PRACTICES.

FECTS TESTS AT 1/40 SCALE IN ZION NUCLEAR POWER PLANT GEOMETRY.

STOCKMAN,H.W.

NUREG/CR4221: THE VALLES NATURAL ANALOGUE PROJECT.

STADLER,L STO R,R.E.

NUR 3

L REGULATORY COMMISSION INFORMA-N OF COPPER PRECIPITATES AND POINT DEFECT CLUSTERS IN RE-STAFFORD,R.S.

ACTOH PRESSURE VESSEL EMBRITTLEMENT.

NUREG/CR 5569 RO1: HEALTH PHYSICS POSITONS DATA BASE.

NUREG/CR4204: QUESTIONS AND ANSWERS BASED ON REVISED STROSNIDER.J.

10 CFR PART 20.

NUREG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

STALLMANN,F.W.

STRUCKMEYER,R.

NUREG/CR-4816 R02: PR-EDB: POWER REACTOR EMBRITTLEMENT NUREG 0837 V13 N04: NRC TLD D! RECT RADIATON MONITORING DATA BASE, VERSION 2. Program Descnptort NETWORKProgress Report October December 1993.

NUREG/CR4076: TR-EDB: TEST REACTOR EMBRITTLEMEW DATA NUREG-0837 V14 N01: NRC TLD DIRECT RADIATON MONITORING BASE,VERSON 1.

NETWORKProgress Report January-March 1994.

NUREG-0837 V14 NO2 NRC TLD DIRECT RADIATION MONITORING STAMPS,D.W.

NETWORK. Progress Report April-June 1994.

NUREG/CR4075 S01: THE PROBABILITY OF CONTAINMENT FAIL.

NUREG-0837 V14 NO3. NRC TLD DIRECT RADIATON MONITORING URE BY DIRECT CONTAINMENT HEATING IN ZON.

NETWORKProgress Report July-September 1994.

STAPLE.8 A STUB 8E EJ.

NUREG/CR4143 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/lA-0093: RELAP5/ MOD 3 ASSESSMENT FOR CALCULATON ODENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT OF SAFETY AND RELIEF VALVE DISCHARGE PIPING HYDRODY-Gr AND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-NAMIC LOADS.

• ernal Events For Plant Operational State 5 Dunng A Refuskng Outage Sections 19.

STUSLER,W.8 AUREG/CR4143 V02P18: EVALUATON OF POTENTIAL SEVERE AC-NUREG-0711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT MODEL GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency Frorn in.

temal Events For Plant Operational State 5 Dunng A Refuehng SU.R.F.

Outage.Section 10.

NUREG/CR4144 V02P1A: EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR4143 V02P1C: EVALUATON OF POTENTIAL SEVERE AC.

ODENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From ytemal GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From in.

Events Dunng M4 Loop Operatons. Main Report (Chapters temal Events For Plant Operational State 5 Dunng A Refuehng NUREG/CR4144 V02PIB. EVALUATON OF POTENTIAL N 4 i*

AC-Outage.Mam Report.

CIDENTS DURING LOW POWER AND SHUTDOWN OP6cA 'ivNS AT NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVERE AC.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Events Dunng M4 Loop Operations. Main Report (Chapters 712L GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From in.

NUREG/CR4144 V02P2: EVALUATON OF POTENTIAL SEVERE ACO-temal Events For Plant Operational State 5 Dunng Refuehng DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Outage.IntemaL.

SURRY, UNIT 1. Analysts Of Core Damage Frequency From intamal NUREG/CR4143 V02PT3; EVALUATION OF POTENTIAL SEVERE AC-Events Dunng MdLoop Operations. Appendices A-D.

CtDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4144 V02P3A: EVALUATON OF POTENTIAL SEVEPE AC.

GRAND GULF, UNIT 1.Anatysis Of Core Damage Frequency From in.

CIDENTS DURING LOW POWER AND SHilTDOWN OPERAlCNS AT temal Events For Plant Operational State 5 Dmng A Refuehng SURRY, UNIT 1. Analyses Of Core Dwage Frequency from Internal Outage Intemal..

Events Dunng M4 Loop OperatonsAapendices E (SecWe E.1 E.8L NURFG/CR4143 V02PT4: EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR4144 V02P3B: EVALUATOW 7 PCTEMilAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GRAND GULF,0 NIT 1. Analysis Of Core Damage Frequency From In-SURRY. UNIT 1. Analysis Of Cure Damage Frequency Frorn infomal temal Events For Plant Operatonal State 5 Dunng A Refuehng Events Dunng M4 Loop Operations. Appendices E (Sections E.9-E.16L Outage.IntemaL.

NUREG/CR4144 V0W4: EVALUATON OF POTENTIAL SEVERE ACCI-NUREG/CR-6143 V04: EVALUATION OF POTENTIAL SEVERE ACCl-DENTS DURtNG LOW POWER AND SHUTDOWN OPERATONS AT DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY,lklT 1. Analyses Of Core Damage Frequency From intemal GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-Events Dunng M4 Loop Operatons.Apperdcas F.H.

tema!Iy indM Fboding Events For Plant Operational State 5 Dunng NUREG/CR4144 V02PS EVALUATON OF POTENTIAL SEVERE ACCl-a Retueleng...

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4166 RISK IMPACT OF TECHNICAL SPECIFICATIONS RE-SURRY, UNIT 1. Analysis Of Cose Damage Frequency From Intemal OutREMENTS DURING SHUTDOWN FOR BWRS.

Events Dunng Mid-Loop Operatons. Appendices t I

i

~ -

74 Personal Author index SUSUDHI,M.

NUREG/CR 6197: TMI 2 VESSEL INVESTIGATION PROJECT INTE-NUREG/CR-5812: MANAGING AGING IN NUCLEAR POWER GRATION REPORT.

PLANTS Insaghts From NRC Mantenance Team inspection Reports.

NUREG/CR5939: THE EFFECTS OF AGE ON NUCLEAR POWER THOMAS,W.

Pt ANT CONTAINMENT COOLING SYSTEMS.

NUREG/CR4224 DFC-PARAMETRIC STUDY OF THE POTENTIAL NUREG/CR5990: THE EFFECTS OF SOLAR-GEOMAGNETICALLY IN.

FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR ED DEBRIS Draft For Comment.

POWER STATIONS.

THOMAS,W.A.

SULUVAN.T.M.

NUREG/CR4126. COGNITIVE SKILL TRAINING FOR NUCLEAR NUREG/C45229 V06: FIELD LYSIMETER INVESTIGATIONS: LOW-POWER PLANT OPERATIONAL DECISIORMAKING.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FISCAL YEAR 1993. Annual Report THORSON,R.

NUREG/CR3145 VIO: GEOPHYSICAL INVESTIGATONS OF THE SWAIN.R1.

WESTERN OHIOINDIANA REGON.Fenal ReFort. October 1986-Sep.

NUREG/CR4249 UNIRRADIATED MATERIAL PROPERTIES OF MID.

tomber 1992.

LAND WELD WF.70, THROMED.

SWATZLER.D.

NUREG.1368: PREAPPLICATION SAFETY EVALUATON REPORT FOR NUREG/CR4126: COGNITIVE SKILL T RAINING FOR NUCLEAR THE POWER REACTOR INNOVATIVE SMALL MODULE Fe3;a.'

POWER PLANT OPERATIONAL DECISON-MAKING.

LIQUID-METAL REACTOR Fool Report I

TADtOSEL NUREG/CR407!, S01: THE PROBABILITY OF CONTAINMENT FAIL-TICHLERA NUREG/CR-2907 V12: RADOACTIVE MATERIALS RELEASED FROM URE BY DIRECT CONTAINMENT HEATING IN ZON.

NUCLEAR POWER PLANTS. Annual Report 1991.

TALSOTT,M.E.

NUREG/CR4114 V03. PERFORMANCE ASSESSMENT OF A HYPO.

TOSEN,P.T.

THETICAL LOW-LEVEL WASTE FACILITY. Groundwater Flow And NUREG/CR4142: TENSILE PROPERTY CHARACTERIZATION OF Transport SenAstort THERMALLY AGED CAST STAINLESS STEELS.

TAM.P.S.

TONG,W.H.

NUREG-0647 S13. SAFETY EVALUATON REPORT RELATED TO THE NUREG/CR4143 V05 EVALUATION OF POTENTIAL SEVERE ACCI.

OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG.0647 S14 SAFETY EVALUATION REPORT RELATEu) TO THE GRAND GULF, UNIT 1 Analysas Of Core Darnage Frequency From 2 Docket Nos 50 390 And 50-391 (Tennssee Valley Authonty Seismc Events Dunng M4 Loop Operatsons Mam Report.

OPERATION OF WATTS BAR NUCLEAR PLANT. UNITS 1 AND NUREG/CR4144 V05. EVALUATION OF POTENTIAL SEVERE ACCI.

I

2. Docket Nos. 50-390 And 50 391 (Tennessee Valley Authonty)

DENTS DURiNG LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, M LAnaW O Core Wnage Rm W bame TANG.M vents Dunng o p Operatons hn RW NUREG/CR4162 EFFECTS OF PROR DUCTILE TEARING ON CLEAV.

AGE FRACTURE TOUGHNESS IN THE TRANSITION REGON TORTORELU 4P.

NUREG/CR4276. OUALITY MANAGEMENT IN REMOTE AFTERLOAD.

TAYLOR.BJ.

i ING BRACHYTHERAPY.

NUREG/CR4816 R02: PR-EDB. POWER REACTOR EMBRITTLEMENT j

DATA BASE.VERSON 2. Program W TRAVISAR.

NUREG/CR4077: DAT A

SUMMARY

REPORT FOR FISSON PRODUCT TAYLOR,J.

NUREG/CP4135. WORKSHOP ON ENVIRONMENTAL QUALIFICATON RELEASE TEST Vl4 OF ELECTRIC EQUIPMENT.

NUREGICR4180: HYDROGEN MIXING STUDIES (HMSLUSER'S NUREG/CR-5812 MANAGING AGING IN NUCLEAR POWER MANUAL PLANTS s From NRC Mamtenance Team inspection Reports NUREG/CR SELECTED F AULT TESTING OF ELECTRONIC ISO.

TRAVIS.R.

LATION DEVICC 88 SED IN NUCLEAR POWER PLANT OPERATON.

NUREG/CR-5939. THE EFFECTS OF AGE ON NUCLEAR POWER PLANT CONTAINMENT COOLING SYSTEMS TAYLOR,T.T.

NUREG/C45985' EVALUATION OF COMPUTER BASED ULTRASONIC TRUSTY,A.D.

INSERVICE INSPECTON SYSTEMS NUREG/CR4145 VERIFICATON AND VAllDATON OF THE SAPHiRE VERSION 4 0 PRA SOFTWARE PACKAGE.

NUREG/CR5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI-TSAO,J.

TIONS ON HUMAN PERFORMANCE. A Handbook Of Erwonmental NUREG 1511: REACTOR PRESSURE VESSEL ST ATUS REPOHT.

NL WCR 5680 V02 THE IMPACT OF LNVIRONMENTAL CONDl-TUYTLE.M.P.

l TIONS ON HUMAN PERFORMANCE A Crvtcal Review Of The Latera-NUREG/CR4258: THE LIOUEFACTON METHOD FOR ASSESSING PA.

ture LEOSEISMICITY.

TERRILL.E.

URYASEV,S.

NUREG/CR4122 STAFFING DECISION PROCESSES AND NUREG/C45994: EMERGENCY DIESEL GENERATOR. MAINTENANCE

/CR472 i TE T

LC A SON OF COMMER NL) l CIAL NUCLEAR POWER PLANT STAFFING REGULATONS AND VANDEN HEUVEL.L l

PRACTICE.1980 *990 NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE E CR4132: BIAXlAL LOADING AND SHALLOW. FLAW EFFECTS A

I ON CHACK TIP CONSTRAINT AND FRACTURE TOUGHNESS.

NU 44674 V18. PHECURSORS TO POTENTIAL SEVERE CORE THEOFANOUS,T.G.

• F* A G'

NUREG/CR5900: STEAM EXPLOSIONS FUNDAMENTALS AND EN-l GERGETIC BEHAVIOR NUREG/CR4075 THE PROBABillTY OF CONTAINMENT FAILURE BY NUREG/CR-4674 V19 PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS 1993 A STATUS REPORT. Man Report And DIRECT CONTAINMENT HEATING IN ZION.

Appendices A-D.

THINNES,OL NUREG/CR-4674 V20 PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4196 CALCULATONS TO ESTIMATE THE MARGIN TO DAMAGE ACCIDENTS.1993 A STATUS REPORT.Appendences E FAILURE IN THE TMI 2 VESSEL And F.

)

i i

Personal Author index 75 VANDER VOORT G.

NUREG/CR4143 V02P1C: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4183: PEER REVIEW OF THE TMI-2 VESSEL INVESTIGA-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TION PROJECT METALLURGICAL EXAMINATIONS.

GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-VANHOENACKER,L.

ternal Events For Plant Operational State 5 Dunng A Refuehng NUREG/lA-0093. RELAPS/ MOD 3 ASSESSMENT FOR CALCULATION Outage Main RW t

NUREG/CR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE AC-OF SAFETY AND REUEF VALVE DISCHARGE PIPtNG HYDRODY*

NAMIC LO/,DS.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1.Analysse Of Core Damage Frequency From In-VANHORN,R.L.

ternal Events For Plant Operational State 5 Dunng A Refuehng NUREG/CR-6116 V03: SYSTEMS ANALYSIS PROGRAMS FOR Outage.intemaL HANDS-ON INTEGRATED RELIABluTY EVALUATIONS (SAPHIRE)

1. "'N' #"

^I N

CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL NUREG/CR4145: VERIFICATION AND VAUDATION OF THE SAPHIRE FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

VERSION 4.0 PRA SOFTWARE PACKAGE.

ED DEBRIS. Draft For Comment.

VANKUIKEN.J.C.

WANG,J.A.

NUREG/CR-5344 RO1: REPLACEMENT ENERGY COST ANALYSIS NUREG/CR 4816 R02: PR EDB: POWER REACTOR EMBRITTLEMENT PACKAGE (RECAP): USER'S GUIDE.

DATA BASE, VERSION 2. Program Description.

NUREG/CR4076: TR-EDB: TEST REACTOR EMBRITTLEMENT DATA NU E'G/CR4188 V01: MICROBIAL DEGRADATION OF LOW-LEVEL

^

RADIOACTIVE WASTE. Annual Report For FY 1993.

w Ang,g, VE9ELY,W.

NUREG/CR4133: FRAGMENTATION AND OUENCH BEHAVIOR OF NUREG/CR-5967: DEVELOPMENT AND APPUCATION OF DEGRADA-CORlUM MELT STREAMS IN WATER.

TION MODELING TO DEFINE MAINTENANCE PRACTICES.

NUREG/CR 5994: EMERGENCY DIESEL GENERATOR: MAINTENANCE WANG,Y.K.

AND FAILURE UNAVAILABluTY, AND THEIR RISK IMPACTS.

NUREG/CR4128: PIPING BENCHMARK PROBLEMS FOR THE ABB/

CE SYSTEM fl0 + STANDARDt2ED PLANT.

VIGIL,R.A.

NUREG/CR-6095. AGING. LOSS-OF COOLANT ACCIDENT (LOCA),

WARE,A.G.

AND HIGH POTENTIAL TESTING OF DAMAGED CABLES.

NUREG/CR4121: COMPONENT EVALUATION FOR INTERSYSTEM y,

LOSS OF-COOLANT ACCIDENTS IN ADVANCED UGHT WATER RE-NU EG/CP 015 WOR OP ON ENVIRONMENTAL OUALIFICATION NUR G CN4245: ASSESSMENT OF PRESSURIZED WATER REAC.

NUREG/CR4086: SELECTED FAULT TESTING OF ELECTRONIC ISO-TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING.

LATION DEVICES USED IN NUCLEAR POWER PLANT OPERATION.

WARNER,R.D.

VINGELIS.P.J.

NUREG/CR-5973 Rot: CODES AND STANDARDS AND OTHER GUID-NUREG/CR4908 V02: ADVANCED HUMAN-SYSTEM INTERFACE ANCE CITED IN REGULATORY DOCUMENTS.

DESIGN REVIEW GUIDELINE. Evaluation Procedures And Guidehnes q

i For Human Factors Engineenng Reviews.

WARRICK,A.W.

(

NUREG/CR4120: CONTROLLED FIELD STUDY FOR VAUDATION OF VINTHER,R.W.

VADOSE ZONE TRANSPORT MODELS.

NUREG/CR4973 RO1: CODES AND STANDARDS AND OTHER GUID.

ANCE CITED IN REGULATORY DOCUMENTS.

WASTLER,S.

y,ggg NUREG-1486: FINAL SAFETY EVALUATION REPORT TO UCENSE NUREG/CR-6103: PRIORITlZATION OF REACTOR CONTROL COMPO-THE CONSTRUCTION AND OPERATION OF A FACIUTY TO NENTS SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEOVENCE OF RME.SME AND DISPOSE OF HE(2) BNM MATERE AGING.

NEAR N,WAHMet M 6894Enwrocare of WaMnc)

YO.T.V.

WATKINS J.C.

NUREG/CR-5830: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

NUREG/CR4935:

SUMMARY

OF WORK COMPLETED UNDER THE SPECTION GUtDE FOR THE MCGUIRE NUCLEAR POWER PLANT.

ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUAUFICATION RE-NUREG/CR4151: FEASIBluTY OF DEVELOPING RISK BASED RANK.

SEARCH PROGRAM (EDOP).

INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVICE IN-SPECTION WElidR,M.

NUREG/CR4181: A PILOT APPUCATION OF RISK-BASED METHODS NUREG-1486: FINAL SAFETY EVALUATION REPORT TO UCENSE TO ESTABUSH INSERVICE INSPECTION PRIORITIES FOR NUCLE-THE CONSTRUCTION AND OPERATION OF A FACluTY TO AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATIOM RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NEAR CUVE, UTAH. Docket No. 40 8989 (Envirocare of Utah,inc.)

NUREG/CR-3145 V10 GEOPHYSICAL INVESTIGATIONS OF THE WE85TER,C.S.

WESTERN OHIO-INDIANA REGION Final Report, October 1986.Sep-NUREG/CR4077: DATA

SUMMARY

REPORT FOR FtSSION PRODUCT tomber 1992.

RELEASE TEST VI-6.

WALLACE.D.R.

NUREG/CP4136: PROCEEDINGS OF THE DIGITAL SYSTEMS REU-WELCHR NUREG/CR 5908 V02: ADVANCED C.iAN-SYSTEM INTERFACE ABluTY AND NUCLEAR SAFETY WORKSHOP. September 13-14, 1993 Rockvitae Crowne Plaza Hotel,Rockwelle. Maryland.

DESIGN REVIEW GUIDELINE. Ev':duation Prowdures And Guidehnes For Human Factors Engmeenng 6eview!L CALSH.B.

NUREG/CR4143 V02P1A' EVALUATION OF POTENTIAL SEVERE AC.

WESTRA.C.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-5680 Voi: THE IMr ACT OF ENVIRONMENTAL CONDl-GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From in.

TiONS ON HUMAN PERFORwANCE. A Handbook Of Environmental temal Events For Plant Operational State 5 Dunn0 A Refuehng Exposures.

Outage Secisons 19.

NUREG/CR-5680 V02: THE IMF ACT OF ENVIRONMENTAL CONDI-NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC-TIONS ON HUMAN PERFORMANCE. A Crtlical Review Of The Uters.

CfDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ture GRAND GULF UNIT 1. Analysis Of Core Damage Frequency From In-NUREG/CR 5758 V04: FITNESS FOR DUTY IN THE NUCLEAR POWER temal Events For Plant Opershonal State 5 Dunng A Refuehng INDUSTRY. Annual Summary Of Program Pertonnarco Reports CY Outage.Section 10.

1993.

76 Personal Author index l

WETZEL,8.A.

NUREG/CR-6233 Vot: STABILITY OF CRACKED PIPE UNDER INER-l NUREG 1368: PREAPPLICATON SAFETY EVALUATION REPORT FOR TIAL STRESSES Subtask 1.1 Final Report.

THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

NUREG/CR-6234: VALOATON OF ANALYSIS METHODS FOR AS-LIQUID METAL REACTOR. Final Report.

SESSING FLAWED PIPING SUBJECTED TO DYNAMIC LCADING.

l WHITEHEAD,0.W.

WILLIAMS,M.L NUREG/CR-4838. MICROCOMPUTER APPLICATIONS OF, AND MODI-NUREG/CR4206: TRANSPORT CALCULATIONS OF RADIATION EX-FICATIONS TO. THE MODULAR FAULT TREES.

POSURE TO VESSEL SUPPORT STRUCTURES IN THE TROJAN RE-NUREG/CR-6093: AN ANALYSIS OF OPERATIONAL EXPERIENCE ACTOR.

DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-i DRESSING HUMAN RELIABILITY ASSESSMENT ISSUES WILLING,0.L l

NUREG/CR4143 V02 PIA: EVALUATON OF POTENTIAL SEVERE AC-NUREG/CR.5344 RO1: REPLACEMENT ENERGY COST ANALYSIS l

COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT PACKAGE (RECAP): USER'S GUIDE.

GRAND GULF, UNIT 1. Analyses Of Core Darnage Frequency Frorn in-l ternal Events For Ptant Operatonal State 5 Dunng A Refuehng WILSON,0.E.

Outage Sectons 19.

NUREG/CR-5535 V07: RELAP5/ MOD 3 CODE MANUALSurnrnanes And NUREG/CR4143 V02P18 EVALUATON OF POTENTIAL SEVERE AC-Reviews Of Independent Code Assesernent Reports.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT l

GRAND GULF, UNIT 1. Analyses Of Core Darnage Frequency Frorn in.

WILSON,M.

l ternal Events For Plant Operatonal State 5 Dunng A Refueling NUREG/CR-6233 V01: STABILITY OF CRACKED PIPE UNDER INER-l Outage.Section 10.

TIAL STRESSES. Subtask 1.1 Final Report NUREG/CR-6143 V02PIC: EVALUATON OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT WILSON,R.

GRAND GULF, UNIT 1.Analyes Of Core Darnage Frequency Frorn in-NUREG/CR-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI-ternal Events For Plant Operatonal State 5 Dunng A Refuehng TONS ON HUMAN PERFORMANCE. A Handbook Of Envronrnental Outage.Mam Report.

Exposures.

NUREd/CR4143 V02PT2-EVALUATON OF POTENTIAL SEVERE AC-NUREG/CR-5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI.

COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT TONS ON HUMAN PERFORMANCE. A Cntscal Review Of The Liters-GRAND GULF, UNIT 1.Analyss Of Core Damage Frequency From In-ture.

ternal Events For Plant Operational State 5 Dunng Refuehng Outage. internal.

WILSON,T.L.

NUREG/CR4143 V02PT3: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-6180. HYDROGEN MIXING STUDIES (HMS) USER'S CtDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT MANUAL l

GRAND GULF, UNIT 1.Analyss Of Core Darnage Frequency From in-l temal Events For Plant Operatonal State 5 Dunng A Refuehng WITT,R.J.

Outage lntemal..

NUREG/CR4196: CALCULATONS TO ESTIMATE THE MARGIN TO I

l NUREG/CR4143 V02PT4: EVALUATON OF POTENTIAL SEVERE AC-FAILURE IN THE TMI-2 VESSEL COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE-GRAND GULF,0 NIT 1.Analvss Of Core Damage Frequency From In-GRATON REPORT.

temel Events For Plant Operatonal State 5 Dunng A Refuehng Outage.intamaL..

WITTMEYER,G.

NUREG/CR4143 VO4 EVALUATON OF POTENTLAL EEVERE ACCl-NUREG/CR-6063 INTRAVAL PHASE 11 MODEL TESTING AT THE LAS l

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT CRUCES TRENCH SITE.

GRAND GULF UNIT 1 Analysis Of Core Damage Frequency From Irv temally induced Flooding Events For Plant Operatonal State 5 Dunng WOLF,J.R.

a Refuehng_

NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE-GRATON REPORT.

WIBLIN,C.

NUREG/CR4156

SUMMARY

OF COMMENTS VALVED FROM WOLFRAM.L.M.

WORKSHOPS ON RADOLOGICAL CRITERIA FO 4 DECOMMISSON-NUREG/CR4145: VERIFICATION AND VALIDATON OF THE SAPHIRE ING.

VERSON 4 0 PRA SOFTWARE PACKAGE.

i NUREG/CR4250:

SUMMARY

OF COMMENTS 1ECEIVED ON STAFF i

DRAFT PROPOSED RULE ON RADOLOGIC% CRITERIA FOR DE-WOLTERMAN,R.L.

COMMISSIONING NUREG/CR4234: VALIDATON OF ANALYSIS METHODS FOR AS-SESSING FLAWED PtPING SUBJECTED TO DYNAMIC LOADING.

WICHMAN,K.

NUREG-1511: REACTOR PRESSURE VESSE6 STATUS REPORT.

WONG,S.M.

NUREG/CR4144 V02PI A: EVALUATON OF POTENTIAL SEVERE AC-WlERENGA.P.J.

COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal CRUCES TRENCH SITE-Events Dunng MdLoop Operatens Mam Report (Chapters 14)

NUREG/CR4120 CONTROLLED FIELD STUDY FOR VAllDATON Or NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-VADOSE ZONE TRANSPORT MODELS COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1.Analyes Of Core Damage Frequency From Internal WlERINGA.D.

NUREG/CR-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDL-CR4744 P2 E L OF PO E L

ACCI-NUI TIONS ON HUMAN PERFORMANCE. A Handbook Of Envronmental DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NU CR 5680 V02: THE IMPACT OF ENVIRONMENTAL CONDl-SURRY, UNIT 1.Anatysis Of Core Damage Frequency From intomal TONS ON HUMAN PERFORMANCE. A Cntical Revew Of The Litera-NU C

44 0 A.

T POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT WILHELM,W.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/CR-6086: SELECTED FAULT TESTING OF ELECTRONIC 150 Events During M4 Loop Operatons.Appeneces E (Sectons E.1-E.8).

NUREG/CR4144 V02P38: EVALUATION OF POTENTIAL SEVERE AC.

LATON DEVICES USED IN NUCLEAR POWER PLANT OPERATION.

COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT wtLKOWSKI.G.M.

SURRY, UNIT 1.Analyes Of Core Damege Frequency From intemal NUREG/CR-4599 V03 N2. SHORT CRACKS IN PIPING AND PIPING Events Dunng MdLoop Operatons Appendices E (Sections E.9-E.16).

WELDS Semsannual Report. October 1992. March 1993 NUREG/CR4144 V02P4: EVALUATION OF POTENTIAL SEVERE ACCI-NUREG/CR-5129 R01: EVALUATON AND REFINEMENT OF LEAK-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RATE ESTIMATON MODELS SURRY. UNIT 1.Analyes Of Core Demage Frequency Frorn intemal NUREG/CR4226' EFFECT OF DYNAMIC STRAIN AGING ON THE Events Dunng MdLoop Operatons.Appendees F-H.

STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PlPING AT NUREG/CR-6144 V02PS: EVALUATON OF POTENTIAL SEVERE ACCl-LWR TEMPERATURES.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT I

1 l

Personal Author index 77 SURRY UNIT 1.Analyss Of Core Damage Frequency From internal SURRY. UNIT 1.Analyms Of Core Damage Frequency From intemal Events Dunng M4 Loop Operations.Appeneces I.

Events Dunng M4 Loop Operataons. Main Report (Chapters 14).

NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-WOOD,RT.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-5904: FUNCTIONAL ISSUES AND ENVIRONMENTAL SURRY. UNIT 1 Analysis Of Core Damage Frequency From intemal OUALiFICATION OF DIGITAL PROTECTION SYSTEMS OF AD.

Events Dunng M4 Loop Operations.Mari Report (Chapters 712).

VANCED LIGHT-WATER NUCLEAR REACTORS.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT woog,g,y, SURRY. UNIT 1.Analyms Of Core Damage Frequency From intomal NUREG/CR4116 V02-SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHtRE)

Events Dunng M4 Loop Operations.Appendces A D.

I NUREG/CR4144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-VERSION 5.0. Integrated Rekatzhty And Rask Analysis System (IRRAS)

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Reference Manual-NUREG/CR4116 V04: SYSTEMS ANALYSIS PROGRAMS FOR SURRY, UNIT 1.Analyes Of Core Damage Frequency From Intamal l

HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPH!RE)

Events Dunng M4 Loop Operations.Appendees E (Sections E.1 E.8).

i VERSION 5.0. Systems Analysis And Fosk Assessment (SARA) Refer.

NUREG/CR4144 V02P38: EVALUATION OF POTENTIAL SEVERE AC-ence Manual CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR SURRY, UNIT 1. Analysis Of Core Damage Frequency From intemal I

HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

Events Dunng M4 Loop Operations.Appendcas E (Sections E.9-E.16).

VERSION 5.0. Fault Tree, Event Tree And Piping & instrumentation NUREG/CR4144 V02P4: EVALUATION OF POTENTLAL SEVERE ACCI-Diagram (FEP) Editors Reference Manual DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1.Analyms Of Core Damage Frequency From intemal WRIGHTAL Events Dunng Mid-Loop Operations.Appendces F-H.

NUREG/CR4193: PRIMARY SYSTEM FISSION PRODUCT RELEASE NUREG/CR4144 V02P5: EVALUATION OF POTENTIAL SEVERE ACCI-AND TRANSPORT.A State-Of-The-Art Report To The Conmttee On DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT The Safety Of Nuclear Installations.

SURRY. UNIT 1.Analyws Of Core Damage Fren.sency From intemal Events During M4 Loop Operations.Appendces I.

994 NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-SU AN V U ION OF AGING RISs( AS.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ES SURRY. UNIT 1.Analyss Of Core Damage Frequency From Intemal I

WU,Q.

Fires Dunng M4 Loop Operations. Main Report.

HUREG/CR4267: AIR-WAT ER SIMULATION OF PHENOMENA OF NUREG/CR4144 V03 P2: EVALUATION OF POTENTIAL SEVERE AC-CORIUM DISPERSG IN DIRECT CONTAINMENT HEATING.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1.Analyms Of Core Damage Frequency From intemal WULFF,W.

Fres Dunng M4 Loop Operations.Appeneces.

NUREG/CR4200: UNCERTAINTY ANALYSIS OF SUPPRESSION POOL HEATING DURING AN ATWS IN A BWR-5 PLANT.An App 6 cation Of YARDUMtAN J.

The CSAU Methodology Usng The BNL Engineenng Plant Analyzer.

NUREG-0525 V02 R02: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL) January 1,1990 Through December 31,1993.

NUREG/CR4063: INTRAVAL PHASE 11 JODEL TESTING AT THE LAS YEH.T.C.

CRUCES TRENCH SITE.

NUREG/CR4120: CONTROLLED FIELD STUDY FOR VALIDATION OF j

VADOSE ZONE TRANSPORT MODELS.

j i

VAKLE,J.

NUREG/CR4143 V02PI A. EVALUATION OF POTENTIAL SEVERE AC-YOUNG,M.F*

CfDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4211: INTEGRATED FUEL-COOLANT INTERACTION (IFCI GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-6.0) CODE. User's Manual temal Events For Plant Operational State 5 Dunng A Refueling j

Outage Sections 1-9.

i NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC-YU,C.K.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-4409 V05: DATA BASE ON DOSE REDUCTION RE-SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

GRAND GULF. UNIT 1. Analyses Of Core Damage Frequency From in.

l temal Events For Plant Operational State 5 Dunng A Refueling YUEN W.W.

Outage.Section 10.

NUREG/CR4143 V02PIC-EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-5960: STEAM EXPLOSIONS: FUNDAMENTALS AND EN-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GERGETIC BEHAVIOR.

GRAND GULF. UNIT 1.Analysas Of Core Damage Frequency From in.

temal Events For Plant Operational State 5 Dunng A Refuehng ZAHOORA Outage. Main Report.

NUREG/CR4281: A SIMPLIFIED LEAK-BEFORE-BREAK EVALUATION NUREG/CR4143 V02PT3: EVALUATION OF POTENTIAL SEVERE AC-PROCEDURES FOR AUSTEN1 TIC AND FERRITIC STEEL PIPLNG.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF. UNIT 1. Analyses Of Core Damage Frequency From in.

ZEIGLER,S L.

ternal Events For Plant Operational State 5 Dunng A Refuehng NUREG/CR4145: VERIFICATION AND VALIDATION OF THE SAPHIRE i

Outage.IntemaL.

VERSION 4.0 PRA SOFTWARE PACKAGE.

I NUREG/CR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ZEITOUN.A.

GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency Frorn in-NUREG-1484 V01: FINAL ENVIRONMENTAL IMPACT STATEMENT temal Events For Plant Operatsonal State 5 Dunng A Rehsehng FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-Outage Intemal--

RICHMENT CENTER. HOMER, LOUISlANA. Docket No. 70-3070 Louise.

NUREG/CR4143 V03: EVALUATION OF POTENTIAL SEVERE ACCl-ana Energy Servicet LP. Ermronmental impact Statement DENTS DURING LOW POWER AND SHt/TDOWN OPERATIONS AT NUREG-1484 V02. IINAL ENVIRONMENTAL IMPACT STATEMENT l

GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-FOR THE CPNS1RUCTION AND OPERATION OF CLAIBORNE EN-N

/CR 66 R SK IM TE fICAL IICAlON Y **'**'

' " " ' * " ' " ^ "

"E"'**'

OUIREMENTS DURING SHUTDOWN FOR BWFIS.

OAND YAM.H.

NUREG/CR4075: THE PROBABILITY OF CONTAINMENT FAILURE BY NUREG/CR4267. AIR-WATER SIMULATION OF PHENOMENA OF DIRECT CONTAINMENT HEATING IN ZION CORIUM DISPERSION IN DIRECT CONTAINMENT HEATING.

l l

YANG.J.

ZHANG,X.J.

I NUREG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4262. CLEAVAGE BEHAVIORS IN NUCLEAR VESSEL CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT STEELS i

l l

i i

78 Personal Author index Z)GLER,0.

ZIMMERMAN,T.L NUREG/CR4224 DFC: PARAMETRIC STUOY OF THE POTENTIAL NUREG/CR4838. MICROCOMPUTER APPLICATIONS OF, AND MODI-FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

FICATIONS TO, THE MODULAR FAULT TREES.

ED DEBRIS Draft For Comment.

t I

4 l

l t

i i

l l

Subject Index This index was developed frorn 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-sonable thesaurus has been developed through experience. Suggestions for improvements cre welcome.

10CFR20 NUREG/CR-6092: RISK ASSESSMENT FOR THE INTENTONAL DE-NUREG/CR4204: QUESTONS AND ANSWERS BASED ON REVISED PRESSURIZATION STRATEGY IN PWRS.

10 CFR PART 20 NUREG/CR4158: IMPUCATONS FOR ACCIDENT MANAGEMENT OF ADDING WATER TO A DEGRADING REACTOR CORE.

NUREG/CR4102: VAllOATION OF THE SCALE BROAD STRUCTURE Accident Sequence 44 GROUP ENDF/B-Y CROSSSECTON UBRARY FOR USE IN NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE CRITICALITY SAFETY ANALYSES.

DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Man Report And Apperdu A.

A 633 8 Steel NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-5861: CRACK-SPEED RELATIONS INFERRED FROM DAMAGE ACCIDENTS: 1992 A STATUS REPORT.Appereces B, C, D, LARGE SINGLE EDGE-NOTCHED SPECIMENS OF A 533 8 STEEL.

E. F. And G.

A533 Grade 5 Claes 1 Steet NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4132: BIAxlAL LOADING AND SHALLOW FLAW EFFECTS DAMAGE ACCIDENTS: 1993 A STATUS REPORT. Man Report And ON CRACK-TIP CONSTRAINT AND FRACTURE TOUGHNESS.

NUR 674 V20 PRECURSORS TO POTENTIAL SEVERE CORE ASWR DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Appendences E NUREG-1503 V01: FINAL SAFETY EVALUATON REPORT RELATED And F.

TO THE CERTIFICATION OF THE ADVANCED BC4UNG WATER RE-ACTOR DESIGN Docket No. 524011 General Electnc Nuclear Energ)

Acoustic Emission NUREG-1503 V02: FINAL SAFETY EVALUATON REPORT RELATw NUREG/CR-5963: CONTINUOUS AE CRACK MONITORING OF A DIS-TO THE CERTIFICATION OF THE ADVANCED BOluNG WATER RE-SIMILAR METAL WELDMENT AT UMERICK UNIT 1.

ACTOR DESIGN. Appendices. Docket No.52-001.4 General Electnc Nu-clear Energy)

Advanced Dolling Water Reactor NUREG-1503 V01: FINAL SAFETY EVALUATON REPORT RELATED ACRS Report TO THE CERTIFICATION OF THE ADVANCED BOluNG WATER RE-NUREG-1125 V15. A CCEP1LATION OF REPORTS OF THE ADVISORY ACTOR DESIGN Docket No. 524014 General Electnc Nuclear Energy)

COMMITTEE ON REACTOR SAFEGUARDS.1993 Annual.

NUREG-1503 V02 FINAL SAFETY EVALUATON REPORT RELATLD TO THE CERTIFICATION OF THE ADVANCED BOluNG WATER RE-

^

N REG-1272 V08 N01: OFFICE FOR ANALYSIS AND EVALUATION OF OPERATONAL DATA.1993 Annual Report Power Reactors.

c h h gd ALWR Advanced Light Water Reactor NUREG/CR-6121: COMPONENT EVALUATON FOR INTERSYSTEM NUREG-1242 V03 FT01: NRC REVIEW OF ELECTRIC POWER RE-LOSSOF-COOLANT AC4fDENTS IN ADVANCED UGHT WATER RE-SEARCH INSTITUTE'S ADVANCED UGHT WATER REACTOR UTiu-ACTORS.

TY REQUIREMENTS DOCUMENTS.Pasarve Plant Desegns Chapter 1.Prolect Number 669.

i APG00 Advanced Reactor NUREG-1242 V03 PT02 NRC REVIEW OF ELECTRIC POWER RE-NUREG/CR4223 REVIEW OF THE PROPOSED MATERIALS OF CON-SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILl-STRUCTON FOR THE SBWR AND AP600 ADVANCED REACTORS.

TY REQUIREMENTS DOCUMENT.Passrve Plant DessgnsChapters 2 13 Protect Number 669 ASME Criteria NUREG/CR4121: COMPONENT EVALUATON FOR INTERSYSTEM NUREG/CR-5359: REVIEW OF ELASTIC STRESS AND FATIGUE TO-LOSSOF-COOLANT ACCOENTS IN ADVANCED UGHT WATER RE-FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA-ACTORS.

TION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER PIPING SYSTEMS.

Advanced Reactor Research NUREG/CP4133 V01: PROCEEDINGS OF THE TWENTY FIRST ATWS WATER REACTOR SAFETY INFOFIMATON MEETING Plenary Ses-NUREG/CR4200: UNCERTAINTY ANALYSIS OF SUPPRESSON POOL soon; Advanced Reactor Research; Advanced Control System Technol.

HEATING DURING AN ATWS IN A BWR-5 PLANT.An Appleahon Of ogy: Advanced instrumentaban & Control Hardware; Human Factors..

The CSAU Methodology Usmg The BNL Engneermg Plant Anayrer.

Advanced Reactors System 40 +

V NO3: REPORT TO CONGRESS ON ABNORMAL C SYS M S

OR ED OCCURRENCES ember 1993.

NUREG 0090 V16 N. EPORT TO CONGRESS ON ABNORMAL Advloory Panel NUREG/CR-6252 LESSONS LEARNED FROM THE THREE MILE NU V

F POR TO GRESS ON ABNORMAL ISLAND UNIT 2 ADVISORY PANEL OCCURRENCES. January-March 1994 NUREG 0090 V17 NO2: REPORT TO CONGRESS ON ABNORMAL A

OCCURRENCES Aprd-June 1994 UAEG/CR-5939 THE EFFECTS OF AGE ON NUCLEAR POWER Accident identification PLANT CONTAINMENT COOUNG SYSTEMS.

NUREG/CR4157: SURVEY AND EVALUATON OF AGING RISK AS-A SESSMENT METHOOS AND APPUCATIONS.

Accident Management WATER REACTOR SAFETY INFORMATION MEETING Plenary Ses-NUREG/CP-0127: PROCEEDINGS OF THE CSNI SPECIALISTS MEET.

saon; Advanced Reactor Research; Advanced Control System Technol-j ING ON FUEL COOLANT INTERACTONS.

ogy; Advanced instrumentabon & Control Hardware; Human Factors..

j 79

a s

i 80 Subject Index j

J NUREG/CP 0133 V02: PROCEEDINGS OF THE TWENTY-FIRST NUREG/CR4143 V02PI A: EVALUATION OF POTENTIAL SEVERE AC.

j WATER REACTOR SAFETY INFORMATION MEETING Severe Acce CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT dont Research.

GRAND GULF UNIT 1.Analysse Of Core Damage Frequency From In-NUREG/CP 0133 V03: PROCEEDINGS OF THE TWENTY-FIRST temal Events For Plant Operational State 5 Dunng A Refueling WATER REACTOR SAFETY INFORMATION MEETING.Pnmary Outage Sections 19.

System Integnty; Aging Research Products & Apohcations; Structural &

NUREG/CR4t43 V02P18: EVALUATION OF POTENTIAL SEVERE AC-1 Sersmic E

& Geology CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT 2

NUREG/CP-Ot35:

KSHOP ENVIRONMENTAL QUALIFICATION GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From irk NU E CR 3 4 V N GHTS FOR AGING MANAGEMENT OF tornal Events For Plant Operational State 5 Dunng A Refueling MAJOR LWR COMPONENTS METAL CONTAINMENTS Outage.Secton ta i

NUREG/CR4143 W2PIC: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR-5812: MANAGING AGING IN NUCLEAR POWER CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT j

PLANTS From NRC Maintenance Team inspeuon R s

NUREG/CR-5. FUNCTIONAL ISSUES AND ENV1

.FNTAL GRAND GULF. UNIT 1.Analysse Of Core Damage Frequency From In-OUALIFICATION OF DIGITAL PROTECTION SYSTEMS Of AD-tornal Events For Plant Opershonal State 5 Dunng A Refuehng VANCED LIGHT-WATER NUCLEAR REACTORS.

Outage.Maan Report j

NUREG/CR-5967. DEVELOPMENT AND APPLICATION OF DEG.tOA-NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVERE AC.

j TION MODELING TO DEFINE MAINTENANCE PRACTICES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6087: THE EFFECTS OF AGING ON BOtLING WATER RE-GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-tenial vents nt Operatonal Staw 5 Dunng Wing NUR G CR-60 5:

ING LOS T ' ACCOENT (LOCA),

NU 4 3/02PT3. EVALUATION OF POTENTIAL SEVERE AC-j NUREG 4 03:

1I TION OF REA OL COMPO-NENTS SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEQUENCE OF CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT i

AGING GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From frF 4

NUREG/CR 6157: SURVEY AND EVALUATION OF AGING RISK AS.

temal Events For Plant Operational State 5 Dunng A Refueling 2

SESSMEldT METHODS AND APPLICATIONS Outage. internal-NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE NUREG/CR4143 V02PT4. EVALUATION OF POTENTIAL SEVERE AC-STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPING AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERADONS AT l

LWR TEMPERATURES-GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From fru 4

temal Events For Plant Operational State 5 Dunno A Refuehng Air ingression Outage Intemal j

NUREG/CR4218 A REVIEW OF THE TECHNICAL ISSUES OF AIR IN-NUREG/CR4166IRISK IMPACT OF TECHNICAL SPECIFICATIONS RE.

GRESSION DURING SEVERE REACTOR ACCIDENTS.

QUIREMENTS DURING SHUTDOWN FOR BWRS.

I NUREG/CR4174 V1 DFC: REVISED ANALYSES OF DECOMMISSION-gg, NUREG/CR4203: VALIDATION STUDIES FOR ASSESS!NG UNSATU-ING FOR THE REFERENCE BOILING WATER REACTOR POWER i

RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

STATION Effects Of Current Regulatory And Other Conssderations On i

The Financial Assurance Requrements Of The Decommissioning Rule l

Air-Misture And..

j NUREG/CR4213-HIGH-TEMPERATURE HYDROGEN-AIR-STEAM NUREG/CR4174 V2 DFC REVISED ANALYSES OF DECOMMISSION-4 DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-ING FOR THE REFERENCE BOILING WATER REACTOR POWER MENT APPARATUS.

STATION Effects Of Current Regulatory And Other Consrierahons On The Financial Assurance Requrements Of The Decommisseorung Rule i

NUREG/CR-2907 V12: RADIOACTIVE MATERIALS RELEASED FROM NUREd/CR4200: UNCERTAINTY ANALYSIS OF SUPPRESSION POOL NUCLEAR POWER PLANTS. Annual Report 1991.

HEATING DURING AN ATWS IN A BWR-5 PLANT.An Apphcation Of The CSAU Methodology Using The BNL Enginoenny Plant Analyzer.

gg,,, gy,,,,

NUREG/CR4105: HUMAN FACTORS ENGINEERING GUIDANCE FOR NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT-f THE REVIEW OF ADVANCED ALARM SYSTEMS.

ED DEBRIS. Draft For Comment.

Aneger NUREG/CR4270 DAF FC: ESTIMATING BOILING WATER REACTOR NUREG-1499-REASSESSMENT OF THE NRC'S PROGRAM FOR PRO.

DECOMMISSIONING COSTS A User's Manual For The BWR Cost Est-j TECTING ALLEGERS AGAINST RETAUATION.

mating Computer Program (CECP) Software Draft Report For Com-j Annual Report NUREG-1145 V10- U S NUCLEAR REGULATORY COMMISSION 1993 Battleship Roce Tuff 4

}

ANNUAL REPORT.

NUREG/CR4221: THE VALLES NATURAL ANALOGUE PROJECT.

j Antecipated Trenaient payeeten Stettetic a

NUREG/lA-0114. ASSESSMENT OF RELAPS/ MOD 3 WITH THE LOFT NUREG/CP 0138. PROCEEDINGS OF WORKSHOP 1 IN ADVANCED

]

L9-1/L3-3 EXPERIMENT SIMULATING AN ANTICIPATED TRANSIENT TOPICS IN RISK AND RELIABILITY ANALYSIS Model Uncertainly: its WITH MULTIPLE FAILURES.

Charactenzaton And Quantification.

4 Aeoiemic Design g

NUREG/CR4241: TECHNICAL GUIDELINES FOR ASEISMIC DESIGN NUREG/CR4154 vo1: EXPERIMENTAL RESULTS FROM CONTAIN-OF NUCLEAR POWER PLANTS Translation Of JEAG 46011987.

MENT PtPING BELLOWS SUBJECTED TO SEVERE ACCIDENT 4

Atomopher6c Transport CONDITIONS Results From Bellows Tested in "LAe-New" Conditons NUREG/CR-5247 V01 R2: RASCAL VERSION 2.1 USER'S GUIDE.

a NUREG/CR-5247 V02 R2: RASCAL VERSION 2.1 WORKBOOK.

84 ental I amertg 4

1 NUREG/CR4132. BIAXIAL LOADING AND SHALLOW-FLAW EFFECTS 1

Automouc Detechon ON CRACK TIP CONSTRAINT AND FRACTURE TOUGHNESS.

l NUREG/CR4255-DESIGN OF AN OPEN ARCHITECTURE SEISMIC MONITORING SYSTEM.

Biogeochemical Process NUREG/CR4289. RECONCENTRATION OF RADIOACTIVE MATERIAL AS S

SEW 8S W AMDM M M W NU G CR A

ARY FEEDWATER SYSTEM RISK-BASED IN-PART 20.

SPECTION GUIDE FOR THE MCGUIRE NUCLEAR POWER PLANT.

Blast Protechen Auuntary System NUREG/CR-5994. EMERGENCY DIESEL GENERATOR. MAINTENANCE NUREG/CR4190 V01: PROTECTION AGAINST MALEVOLENT USE OF AND FAILURE UNAVAJLABluTY, AND THEIR RISK IMPACTS VEHICLES AT NUCLEAR POWER PLANTS Vehicle Bamer System Sating Guidance For Blast Protectson.

SWR NUREG/CR4190 V01 Rt: PROTECTION AGAINST MALEVOLENT USE NUREG/CR4087: THE EFFECTS OF AGING ON BOlUNG WATER RE-OF VEHICLES AT NUCLEAR POWER PLANTS Vehicle Bamer System ACTOR CORE ISOLATION COOLING SYSTEMS.

Selecten Guidance For Blast Protection.

l 1

Subject index 81 NUREG/CR4190 V02: PROTECTION AGAINST MALEVOLENT USE OF CaplNary SerHer VEHICLES AT NUCLEAR POWER PLANTS.VeNele Batner System NUREG/CR-4918 V07: CONTROL OF WATER INFILTRATION INTO Seh'ng Gnadence For Blast Protecton.

NEAR SURFACE LLW DISPOSAL UNITS.Progrees Report On Field Ex-NUREGICR4190 V02 R1: PROTECTION AGAINST MALEVOLENT USE portmente At A Humid Regen S4e,Besteville Marytend.

I OF VEHICLES AT NUCLEAR POWER PLANTS.VeNcio Bemer System l

Seiecton Giedence Carmen steet NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE I

NURE CR408 HE EFFECTS OF AGING ON BOILING WATER RE-ACTOR OORE 880LATION COOLING SYSTEMS.

NUREG/CR4143 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-Cast Stehdene Steel CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4142: TENSILE-PROPERTY CHARACTERIZATION OF

{

GRAND GULF, UNIT 1.Analves Of Core Damage Frequency From In-THERMALLY AGED CAST STAINLESS STEELS.

tornal Events For Plant Operemonal State 5 Dunng A Refueling Outage.Sectons 19.

Coment NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4188 V0t MICROBIAL DEGRADATION OF LOW-LEVEL CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RADIOACTIVE WASTE. Annual Report For FY 1993.

GRAND GULF, UNIT 1. Analyse Of Core DemeGe Frequency From In-temal Evente For Plant Operatonal Stele 5 Dunng A Refuehn0 Cement-SetdNeed Outogo.Secton 10.

NUREG/CR4164: RELEASE OF RADIONUCLIDES AND CHELATING -

NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC.

AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATION LOW-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LEVEL RADIOACTIVE WASTE COLLECTED FROM THE PEACH i

GRAND GULF, UNIT 1.Analyes Of Core Damage Frequency From in.

BOTTOM ATOMIC POWER STATION UNIT 3.

f femal Evente For Plant Operatenel State 5 Dunng A Refuehn0 Outage.uem Report.

Cerenselse of e-r"---*

NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVERE AC-NUREG4383 V01 R17: DIRECTORY OF CERTIFICATES OF COMPLi-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC GRAND GULF. UNIT 1.Analysm Of Core Damage Frequency From In-Approved %get tornal Events For Plant Operatonal - Stele 5 Dunng Refuehng NUREGM V02 R17: DIRECTORY OF CERTIFICATES OF COMPLi-

~

NURE CR 3 V02PT3: EVALUATION OF POTENTIAL SEVERE AC.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG4383'V03 R14: DIRECTORY OF CERTIFICATES OF COMPL1-GRAND GULF, UNIT 1.Analyes Of Core Damage Frequency From In-ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC Events For Plant Operatonal State 5 Dunn0 A Refuehng Approved Queri ssurance Programs For R% Matensis Pack.

A NUREG/CR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE AC-

• P"'

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT

% Agent GRAND GULF, UNIT 1 Analyse Of Core Damage Frequency From In-NUREG/CR4164: RELEASE OF RADIONUCLIDES AND CHELATING torned Events For Plant Operational State 5 During A Refuehng AGENTS FROM CEMENT-SOLIDIFIED DECONTAMINATION LOW-NURE CR 66': RISK IMPACT OF TECHNICAL SPECIFICATIONS RE'

~

BO OM R TAT 3~

l OUtREMENTS DURING SHUTDOWN FOR BWRS.

NUREG/CR4174 V1 DFC: REVISED ANALYSES OF DECOMMISSION-Groult Bresher ING FOR THE REFERENCE BOluNG WATER REACTOR POWER STATION Effects Of Current Regulatory And Other Conadorations On NUREG/CR4169: RELAY TEST PROGRAM.Senes 11 Tests. Integral Teet-4 Of W And M hm' The Financial Aneurance Requrements Of The CAu

_ i Rule

?

NUR E/CR4174 V2 DFC* REVISED ANALYSES OF DECOMMISSION.

f484 ENVIRONMENTAL IMPACT STATEMENT ING FOR THE REFERENCE BOILING WATER REACTOR POWER FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-I STATION.Efeects Of Current Regulatory And Other Canaderations On R:CHMENT CENTER, HOMER. LOUISIANA. Docket No. 70 3070. Lou:si-

)

The Financial Aneurance Requrements Of The Decommmesonmg Rule ans Energy Senaces. LP. Environmental impact Statement.

[

NUREd/CR4200' UNCERTAINTY ANALYSIS OF SUPPRESSION POOL MEM& W2: N EWRONMEM WM SWEM l

FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-HEATING DURING AN ATWS IN A BWR-5 PLANT.An Apptceton Of R

M ER IStA Docket Nc. M70M 6

The CSAU Methodology Using The BNL Engmeenng Plant Analyzer.

NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL ane nergy es,\\..

Responses FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT-ED DEBRIS. Draft For Comment.

NUREG/CR4270 DAF FC: ESTIMATING BoluNG WATER REACTOR NUREG/CR4262: CLEAVAGE BEHAVIORS IN NUCLEAR VESM.L DECOMMISSIONING COSTS.A User's Manuel For The BWR Cost Eati-STEELS.

metmg Computer Program (CECP) Software. Draft Report For Com" Code NUREG/CR-5973 R01: CODES AND STANDARDS AND OTHER GUID-Breelytherapy ANCE CITED IN REGULATORY DOCUMENTS.

]

NUREG/CR4276: OUAUTY MANAGEMENT IN REMOTE AFTERLOAD-

/ 4127: THE EFFECTS OF STRESS ON NUCLEAR POWER Suck 8ng Evoluellen PLANT OPERATIONAL DECISION MAKING AND TRAINING AP-NUREG/CR4161: BUCKUNG EVALUATION OF SYSTEM 80+(TM)

PROACHES TO REDUCE STRESS EFFECTS.

CONTAINMENT.

gugget NUREG/CR4126: COGNITIVE SKILL TRAINING FOR NUCLEAR NUREG 1100 V10 BUDGET ESilMATES. Fiscal Year 1995.

POWER PLANT OPERATIONAL DECISeON-MAKING.

CA8WU 3 Cesc.ter Coste NUREG/CR4269-A PLAN FOR THE MODIFICATION AND ASSESS-NUMG/CR4128: PIPING BENCHMARK PRO 6LEMS FOR THE ABB/

MENT OF TRAC-PF1/ MOD 2 FOR USE IN ANALYZING CANDU 3 CE SYSTEM 80+ STANDARDIZED PLANT.

TRANSIENT THERMAL-HYDRAUUC PHENOMENA.

Cenerete Berrter CANDU 3 Design NUREG/CR4138. USER'S GUIDE FOR SIMPLIFIED COMPUTER l

NUREG-1502: ASSESSMENT OF DATABASES AND MODEUNG CAPA-MODELS FOR THE ESTIMATION OF LONG. TERM PERFORMANCE BluTIES FOR THE CANDU 3 DESIGN.

OF CEMENT. BASED MATERIALS.

n

._ _.~

82 Subject Index Concrete Cracking NUREG/CR4144 V02PIB EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR-6236. SEISMIC INVESTIGATONS OF THE HDR SAFETY CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PROGRAM.Sumrnary Report SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal Events Dunng Mel oop Operabons. Main Report (Chapters 712).

ConWauon Management NUREG/CR 6144 V02P2: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CP-0136: PROCEEDINGS OF THE DIGITAL SYSTEMS REU*

DENTS DURING LOW FOWER AND SHUTDOWN OPERATONS AT ABluTY AND NUCLEAR SAFETY WORKSHOP. September 13-14' SURRY, UNIT 1.Analyss Of Core Damage Frequency From internal 1993,Rockville Crowne Plaza Hotel,Rockville, Maryland.

Events Dunng M4 Loop Operabons.Appereces A-D.

NUREG/CR4144 V02P3A EVALUATON OF POTENTIAL SEVERE AC-ConstrucWon CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-6223: REVIEW OF THE PROPOSED MATERIALS OF CON.

STRUCTON FOR THE SBWR AND AP600 ADVANCED REACTORS.

SURRY, UNIT 1 Analysis Of Core Damage Frequency From internal Events Durmg MdLoop Opershons.Apperdces E (Secnons E.1.E.8).

Containment NUREG/CR4144 V02P38: EVALUATON OF POTENTIAL SEVERE AC-NUREG/CR-5939: THE EFFECTS OF AGE ON NUCLEAR POWER CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT PLANT CONTAINMENT COOLING SYSTEMS.

SURRY, UNIT 1.Analyss Of Core Damage Frequency From intemal NUREG/CR4075; THE PROdABILITY OF CONTAINMENT FAILURE BY Events During M4 Loop Opershons.Apperdcas E (Sections E9-E.16).

DIRECT CONTAINMENT HEATING IN ZION.

NUREG/CR4144 V02P4:EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR4075 S01: THE PROBABILITY OF CONTAINMENT FAIL-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT URE BY DIRECT CONTAINMENT HEATING IN ZION.

SURRY, UNIT 1.Analyss Of Core Damage Frequency From internal NUREG/CR4161: BUCKUNG EVALUATON OF SYSTEM 80+(TM)

Events During M4 Loop OperabonsApperdces F-H.

CONTAINMENT

• NUREG/CR4144 V02P5: EVALUATION OF POTENTIAL SEVERE ACCl-l l

NUREG/CR4213: HIGH-TEMPERATURE HYDROGEN-AIR-STEAM DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DET EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-SURRY, UNIT 1.Analyss Of Core Damage Frequency Frorn Intemal Events Durmg M4 Loop OpersbonsApperdces L NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-Containment Heating NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DtRECT CON.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE SURRY, UNIT 1 Analysis Of Core Darnage Frequency From internal SURRY NUCLEAR POWER PLANT.

Fires Dunng M4 Loop Operanons. Main Report NUREG/CR4144 V03 P2 EVALUATON OF POTENTIAL SEVERE AC-Containment Mp6ng CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4154 V01: EXPERIMENTAL RESULTS FROM CONTAIN-SURRY, UNIT 1 Analyss Of Core Damage Frequency From internal MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT Fires Dunng M4 Loop Operabonshpereces.

CONDITIONS.Results From Bellows Tested in " Uke-New" Cordtsons.

NUREG/CR4144 V04: EVALUATION OF POTENTIAL SEVERE ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Containment System SURRY. UNIT 1Analyss Of Core Damage Frequency From internal NUREG/CR-6180- HYDROGEN MIXING STUDIES (HMS): USER'S Floods Dunng M4 Loop Operations.

MANUAL NUREG/CR4144 V05: EVALUATON OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Control Equipment SURRY. UNIT 1 Analysis Of Core Damage Frequency From Seestruc NUREG/CR4146. LOCAL CONTROL STATONS HUMAN ENGINEER.

ING ISSUES AND INSIGHTS.

Events Dunng M4 Loop Operabons. Main Report Control Rod Drtve Core Degradation NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC.

NUREG/CR4218: A REVIEW OF THE TECHNICAL ISSUES OF AIR IN-TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING.

GRESSION DURING SEVERE REACTOR ACCIDENTS.

Control Room Corium NUREG/CR-5908 V01: ADVANCED HUMAN-SYSTEM INTERFACE NUREG/CR4133: FRAGMENTATON AND OUENCH BEHAVOR OF DESIGN REVIEW GUIDEUNE. General Evatusbon Model, Techrweal CORIUM MELT STREAMS IN WATER.

Development. And Guideline Desenption.

NUREG/GR-5908 V02-ADVANCED HUMAN-SYSTEM INTERFACE Corium Diepersion DESIGN REV!EW GUIDEUNE. Evaluation Procedures And Guidelines NUREG/CR4267: AIR-WATER SIMULATON OF PHENOMENA OF For Human Factors Engmeenng Reviews.

CORIUM DISPERSION IN DIRECT CONTAINMENT HEATING.

Control Stat 6on Corrosion Fatique NUREG/CR4146: LOCAL CONTROL STATONS: HUMAN ENGINEER.

NUREG/CR-4667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN ING ISSUES AND INSIGHTS-LIGHT WATER REACTORS. Senwannual Report.Apnl 1993 Septem.

• N Control System NUREG/CR4105: HUMAN FACTORS ENGINEERING GUOANCE FOR Crack THE REVIEW OF ADVANCED ALARM SYSTEMS.

NUREG/CR-4599 V03 N2: SHORT CRACKS IN PIPING AND PIP 1NG Cooling System WELDS. Semiannual Report. October 1992 - March 1993.

NUREG/CR5939: THE EFFECTS OF AGE ON NUCLEAR POWER NUREG/CR-5128 ROI: EVALUATON AND REFINEMENT OF LEAK-PLANT CONTAINMENT COOLING SYSTEMS.

RATE ESTIMATON MODELS.

l Core Damage Crack Arrest NUREG/CR4143 V03: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR.5591 V02 N1: HEAVY-SECTON STEEL 1RRADIATON DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT PROGRAM Senuannual Progress Report For October 1990 - March GRAND GULF. UNIT 1 Analyses Of Core Damage Frequency From in.

1991.

tornal Events For Plant Operahonal State 5 Dunng A Refuehng Outaos.

NUREG/CR-6139: CRACK-ARREST TESTS ON TWO IRRADIATED NUREG/CR4143 V04: EVALUATON OF POTEnilAL SEVERE ACCl-HiGH COPPER WELDS Phase ll: Results Of Duplex-Type Specimens.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF. UNIT 1 Analysis Of Core Damage Frequency From In-Crack Growth temally induced Floodmg Events For Plant Operational State 5 Dunng NUREG/CR4182: EFFECTS OF PROR DUCTILE TEARING ON CLEAV-a Refuehng.....

AGE FRACTURE TOUGHNESS IN THE TRANSITON REGION.

NUREG/CR4143 V05: EVALUATON OF POTENTIAL SEVERE ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Crack StatWitty l

GRAND GULF. UNIT 1 Analysis Of Core Damage Frequency From NUREG/CR4234: VALOATION OF ANALYSIS METHODS FOR AS-Seistrue Events Dunno M4 Loop Opershons Mam Report SESSING FLAWED PIPING SUBJECTED TO DYNAMIC LOADING.

NUREG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Cracked Pipe SURRY. UNIT 1.Analyss Of Core Damage Frequency From Internal NUREG/CR4233 VOI: STABlUTY OF CRACKED PIPE UNDER INER.

Events Dunng Mid Loop Operations Main Report (Chapters 14).

TIAL STRESSES. Subtask 1.1 Final Report I

i Subject index 83 Crew Perfremance NUREG/CR-6250'

SUMMARY

OF COMMENTS RECEIVED ON STAFF NURE&CR-6126: COGNITIVE SKILL TRAINING FOR NUCLEAR DRAFT PROPOSED RULE OF RADIOLOGICAL CRITERIA FOR DE-POVIER PLANT OPERATONAL DECISION-MAKING.

COMMISSIONING.

Criticality Decommiseloning Cost NUREG/CR4102: VALOATION OF THE SCALE BROAD STRUCTURE NUREG/CR4270 DRF FC: ESTIMATING BOfLING WATER REACTOR 44-GROUP ENDF/B-Y CROSS-SECTION LIBRARY FOR USE IN CRITICALITY SAFETY ANALYSES.

DECOMMISSONING COSTS.A user's Manual For The BWR Cost Esti-mating Computer Program (CECP) Software. Draft Report For Com-Criticality Sequence NUREG/CR-6182 V01: OFFSCALE: A PC INPUT PROCESSOR FOR Decontamination THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For The CnhcaW Sequences.

NUREG/CR4164: RELEASE OF RADIONUCUDES AND CHELATING AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATON LOW.

Crustal Structure LEVEL RADIOACTIVE WASTE COLLECTED FROM THE PEACH NUREG/CR-3145 V10- GEOPHYSICAL INVESTIGATIONS OF THE NjE / R WESTERN Ctf;O4NDtANA REGON. Final Report. October 1986 Sep-CO PR SS AP M ERSION TESTS AND tember m2.

LEACHING OF RADONUCLCES STABLE METALS, AND CHELATING AGENTS FROM CEMENT-SOLIDIFIED DECONTAMINATON WASTE Damaged Cable COLLECTED FROM NUCLEAR POWER STATIONS.

NUREG/CR4095: AGING, LOSS OF COOLANT ACCIDENT (LOCA),

AND HIGH POTENTIAL TESTING OF DAMAGED CABLES.

Defect Clueter NUREG/CR-6231: A COMPARISON OF THE RELATIVE IMPORTANCE Data Base OF COPPER PRECIPITATES AND POINT DEFECT CLUSTERS IN RE-NUREG/CR-4409 V05: DATA BASE ON DOSE REDUCTON RE-ACTOR PRESSURE VESSEL EMBRITTLEMENT.

SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

NUREG/CR-4816 R02: PR-EDB: POWER REACTOR EMBRITTLEMENT Defonoo-In-Depth DATA BASE,VERSON 2. Program Desenphon.

NUREG/CR4303: METHOD FOR PERFORM;NG OfVERSITY AND DE-Data Transmisanon System

~

y3~

NUREG/CR4149: APPLICATONS OF FIBER OPTICS IN PHYSICAL PROTECTON.

Degradation l

De h NUREG/CR-6205: VALVE ACTUATOR MOTOR DEGRADATON, l

NUREG-1502 ASSESSMENT OF DATABASES AND MODELING CAPA-l BILITIES FOR THE CANDU 3 DESIGN.

Degradation Modeling I

NUREG/CR-5967: DEVELOPMENT AND APPLICATION OF DEGRADA-Debris TON MODELING TO DEFINE MAINTENANCE PRACTICES.

NUREG/CR4224 DFC: PARAMETRO STUDY OF THE POTENTIAL F

B STR LOCKAGE DUE TO LOCA GENERAT-U

/CR4158. IMPLICATIONS FOR ACCIDENT MANAGEMENT OF ADDING WATER TO A DEGRADING REACTOR CORE.

Debrie Cooling N

G 42 INT ATED FUEL COOLANT INTERACTION (IFCI De DE-PRESSURIZATON STRATEGY IN PWRS.

Decay Heat NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSED DECAY Depressurtration System HEAT GUIDE USING SAS2H/ORIGEN.S DATA.

NUREG/CR4850: ANALYSIS OF LONG-TERM STATON BLACKOUT DecimMaMn9 WITHOUT AUTOMATIC DEPRESSURl2ATON AT PEACH BOTTOM USING MELCOR (VERSION 1.8),

NUREG/CR4126: COGNITIVE SKILL TRAINING FOR NUCLEAR POWER PLANT OPERATONAL DECISON. MAKING.

Design Factor NUREG/CR4294: DESIGN FACTORS FOR SAFETY-CRITICAL SOFT.

Decommissionin9 WARE.

NUREG-1307 R04: REPORT ON WASTE BURIAL CHARGES. Escalation Of Dw-.m Waste Disposai Costs At Low-Level Waste Bunal Detonations-Experimental Data Facilities.

NUREG/CR4213. HIGH-TEMPERATURE HYDROGEN-AIR-STEAM NUREG-1496 V1 DFC: GENERIC ENVIRONMENTAL IMPACT STATE-MENT IN SUPPORT OF RULEMAKING ON RADOLOGICAL CRITE-DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-MENT APPARATUS RtA FOR DECOMMISSONING OF NRC-LICENSED NUCLEAR FACILITIES Main Report Draft Report For Comment.

Devitrification NUREG-1496 V2 DFU: GENERIC ENVIRONMENTAL IMPACT STATE-NUREG/CR4221: THE VALLES NATURAL ANALOGUE PROJECT.

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-RIA FOR DECOMMISSIONING OF NRC-LICENSED NUCLEAR D6eset Generator FACILITIES.Apperxh'ees. Draft Report For Comment NUREG-1416: OPERATONAL EXPERIENCE AND MAINTENANCE PRO-NUREG 1500 WORKING DRAFT REGULATORY GUOE ON RELEASE GRAMS OF TRANSAMERICA DELAVAL, INC DIESEL GENERA-CRITERIA FOR DECOMMISSONING: NRC STAFF'S DRAFT FOR TORS.

COMMENT.

NUREG-1501 DRFT: BACKGROUND AS A RESOUAL RADOACTIVITY D6gital Protection System CRITERON FOR DECOMMISSIONING Appendix A To The Draft Go-NUREG/CR 5904: FUNCTONAL ISSUES AND ENVIRONMENTAL neric Ermronmental impact Statement in Support Of Rulemalung On OVAUFICATION OF DIGITAL PROTECTON SYSTEMS OF AD-R Cntena For Cmu,mm w NMENTS

- Of NRC....

VANCED LIGHT WATER NUCLEAR REACTORS.

NUR 156:

SUMMARY

OF CO RECEIVED FROM WORKSHOPS ON RADOLOGICAL CRITERIA FOR DECOMMISSON-Digital System ING NUREG/CR4174 V1 OFC: REVISED ANALYSES OF DECOMMISSON-NUREG/CP 0136: PROCEEDINGS OF THE DIGITAL SYSTEMS REll.

ING FOR THE REFERENCE BOILING WATER REACTOR POWER ABILITY AND NUCLEAR SAFETY WORKSHOP. September 13-14, STATON.Ef'ects Of Current Regulatory And Other Considerabons On 1993,Rockville Crowne Plaza Hotel,Rockville. Maryland.

The Financial Assurance Requirements Of The Decommissiorung Rule Direct Containment Heating And....

NUREG/CR4044: EXPERIMENTS TO INVESTIGATE DIRECT CON-NUREG/CR-6174 V2 DFC: REVISED ANALYSES OF DECOMMISSION-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE ING FOR THE REFERENCE BOILING WATER REACTOR POWER ZION NUCLEAR POWER PLANT IN THE SURTSEY TEST FACILITY.

STATON Effects Of Current Regulatory And Other Considerations On NUREG/CR4168: DIRECT CONTAINMENT HEATING INTEGRAL EF.

The Financial Assurance Requirements Of The Decomrnissionsng Rule FECTS TESTS AT 1/40 SCALE IN ZON NUCLEAR POWER PLANT And....

GEOMETRY.

~

M

$8lSC% lAdSM NUREG/CR4267: AIR-WATER SIMULATION OF PHENOMENA OF Elmete Strees CORIUM DISPERSION IN DIRECT CONTAWMENT HEATING.

NUREG/CR-5350: REVIEW OF ELASTC STRESS AND FATIGUE-TO-FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA-

/lA4093: RELAP5/ MOD 3 ASSESSMENT FOR CALCULATION TE '

OF SAFETY AND REUEF VALVE DISCHARGE PIPING HYDRODY.

NAMIC LOADS.

an g,sede NUFiEGJ:R-6881: CRACK 8 PEED RELATIONS INFERRED FROM mg LARGE WWENMCHED SMMENS & A 533 8 NER NUREG/CR-5083: CONTINUOUS AE CRACK MONITORNG OF A DIS-SIMILAR METAL WELDMENT AT LIMERCK UNIT 1.

gg,,g,g, gw NUREG/CP 013: WORKSHOP ON ENVIRONMENTAL QUALIFICATION Olaerelly NUREGICR4303: METHOD FOR PERFORMMG DIVERSITY AND DE.

OF ELECTRIC (OUIPMENT, F

-M-DEPTH ANALYSES OF REACTOR PROTECTION SYS-NUREG/CR4103: PRIORITIZATION OF REACTOR CONTROL COMPO-NENTS SUSCEPTIOLE TO FIRE DAMAGE AS A CONSEQUENCE OF DoesP -

NUREG/CR-2950 V12-DOSE COMMITMENTS DUE TO RADIOACTIVE AGING.

RELEASES FROM NUCLEAR POWER PLANT SITES M 1900.

Elecerteel spetem Does Reduellen NUREG/CR-5000: THE EFFECTS OF SOLAR-GEOMAGNETICALLY IN-NUREG/CR-4409 VOS: DATA BASE ON DOSE REDUCTION RE-DUCED CURRENTS ON ELECTRICAL SYSTEMS W NUCLEAR SEARCH PROJECTS FOR NUCLEAR POWER PLANTS-POWER STATIONS.

Oran Eastruenental impoet Statement Eineerenagneen laterterenes NUREG-150s DRAFT ENVIRONMENTAL IMPACT STATEMENT TO NUREGICR-Seet: TECHNCAL BASIS FOR EVALUATING ELECTRO.

CONSTRUCT AND OPERATE THE CROWNPOINT SOLUTION MAGNETIC AND RADIOFREQUENCY INTERFERENCE IN SAFETY.

MINMG PROJECTCROWNPOINT, NEW MEXICO. Docket No. 4 RELATED i&C SYSTEMS.

8908.(Hydro Resources.Inc.)

E"*"'E"***"-

DreR proposed Rule NWEGM500k M NCTS W M&MNY E NUREG/CR4250'

SUMMARY

OF COMMENTS RECEIVED ON STAFF M C M M ON N WS N N E N DRAFT PROPOSED RULE ON RADIOLOGICAL CRITERIA FOR DE-POWER STATIONS.

COMMISSIONWG J

Electrente lashWen Destes i

gyeit Regulatory Gedde NUREG/CR4006: SELECTED FAULT TESTING OF ELECTRONIC ISO-NUREG-1500- WORKING DRAFT REGULATORY GUIDE ON RELEASE LATION DEVICES USED M NUCLEAR POWER PLANT OPERATION.

CRITERIA FOR DECOMMISSIONING NRC STAFF *S DRAFT FOR COMMENT.

NUREG/CR-4016 R02: PREDB: POWER REACTOR EMBRITTLEMENT a

Orlet Lhelt i

NUREG/CR4104: SHEAR WALL ULTIMATE DRIFT UMITS.

DATA 8ASE VERSION 2. Proipam NUREG/CR-5601 V02 N1: IEAVY ION STEEL IRRADIATION Ductile Toerine PROGRAM Seneannual Prolpees Report For Octatrer 19eo - March i

NUREG/CR4162 EFFECTS OF PROR DUCTILE TEARNG ON CLEAV-1301.

AGE FRACTURE TOUGHNESS IN THE TRANSITION REGION NUREG/CR-5601 V02 N2: HEAVY SECTION STEEL IRRADIATION PROGRAM.Seneannual Report For W ^_

1991.

G 4128: PtPtNG BENCHMARK PRO 6LEMS FOR THE AB8/

B VE CE SYSTEM 80 + STANDARDIZED PLANT.

NUREG/CR4231: A COMPAR; SON OF THE RELATIVE IMPORTANCE NUREG/CR-6216: EVALUATION OF ROCK JO6NT MODELS AND COM-OF COPPER PRECIPITATES AIO POINT DEFECT CLUSTERS W RE-a ACTOR PRESSURE VESSEL EMBRITTLEMENT.

NUR CR 41 E GUI UNES F DESIGN OF NUCLEAR POWER PLANTS.Transiston Of JEAG 4801-1987.

g,,,,,,,, gg,,,g m NUREG/CR-6004: EMERGENCY DIESEL GENERATOR: MAINTENANCE Dynamic Leading AND FAILURE UNAVAILASIUTY, AND THEIR RISK RIPACTS.

[

NUREG/CR4178; LABORATORY CHARACTER 12ATION OF ROCK JOINTS Enfereement Aouen NUREG/CR4234, VAUDATION OF ANALYSIS METHODS FOR AS.

NUREG4940 V12 NO3: ENFORCEMENT ACllONS: SIGNIFICANT AC.

SESSNG FLAWED PtPING SUEUECTED TO DYNAMIC LOADMG.

TIONS RESOLVED.

Report. July-Seplomber 1993.

Dynamic Quaellesmen NUREG-0040 V12 ND4: E MENT ACTIONEk SIGNIFICANT AC-NUREG/CR-5035:

SUMMARY

OF WORK COMPLETED UNDER THE TIONS RESOLVED.Ouarterly Progrees Report,0ctober-December i

+

ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUAUFICATION RE.

1993.

SEARCH PROGRAM (EDOP).

NUREG4940 V13N01P01: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED REACTOR UCENSEES. Quarterly Progrees Dynounc-Arrest Toughnese

-March 1994.

NUREG/CR-5061: CRACK-SPEED RELATIONS NFERRED FROM G 4e40 13N01P02: ENFORCEMENT ACTIONS-SIGNIFICANT L

LARGE SINGLE-EDGE-NOTCHED SPECIMENS OF A 533 S STEEL ACTIONS RESOLVED ~ MEDCAL UCENSEES.Ouarterly Progrees i

ECCS 1

ENFORCEMENT ACTIONS: SIGNIFICANT NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL ACTIONS RESOLVED INDUSTRIAL UCENSEES.Ouarterly Progrees FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT'

-March 1994.

ED DEBRIS. Draft For Comment G-Oe40 13N02P01: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED-REACTOR UCENSEES.Ouarterty Progrees EPIDDR N NUREG/CR 5229 V06: FIELD LYSIMETER INVESTIGATIONS: LOW-

. ENFORCEMENT ACTIONS SIGNIFICANT LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR ACTIONS RESOLVED MEDICAL UCENSEES.Ouarterly Proyees FISCAL YEAR 1993. Annual Report.

MJune 1994.

G4940 V13N02P03. ENFORCEMENT ACTIONS: SIGNIFICANT Earthquake ACTIONS RESOLVED INDUSTRIAL LICENSEES. Quarterly Proyees NUREG/CR4241: TECHNICAL GUIDELNES FOR ASEISMC DESIGN i

CR4290 ANA E USE GU DI 2.0.

NUR 1: ENFORCEMENT ACTIONS: SIGNIFICANT NUREG/GR-0008: VALIDATION OF SEISMC PROBASluSTIC RISK AS-ACTIONS RESOLVED. REACTOR UCENSEES. Quarterly Progrees SESSMENTS OF NUCLEAR POWER PLANTS.

Report, July September 1994.

r

Subject Index 85 NUREG-0940 V13NO3P02: ENFORCEMENT ACTIONS: SIGNIFICANT Fault Testing ACTIONS RESOLVED. MEDICAL LICENSEES Ouarterty Progress NUREG/CR-6086: SELECTED FAULT TESTING OF ELECTRONIC ISO.

Report. July-Septernber 1994.

LATION DEVICES USED IN NUCLEAR POWER PLANT OPERATION.

NUREG4940 V13NO3P03: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED MATERIAL LICENSEES (NON-Fault Tree MEDICAL) Quarterty Progress Report. July-September 1994.

NUREG/CR 4838. MICROCOMPUTER APPLICATIONS OF AND MOU FICATIONS TO, THE MODULAR FAULT TREES.

Engineered Safety System NUREG/CR 6116 V07: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR-6087: THE EFFECTS OF AGING ON BOILING WATER RE-HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

ACTOR CCRE ISOLATION COOLING SYSTEMS-VERSION 5.0. Fault Tree. Event Tree, And Ppng & Instrumentation Environmental 9'*'"

NUREG/CR4232: ASSESSING THE ENVIRONMENTAL AVAILABILITY Federal Regulation OF URANIUM IN SOILS AND SEDIMENTS.

NUREG 1460 R01: GUIDE TO NRC REPORTING AND RECORDKEEP.

ING REQUIREMENTS. Compiled Frorn Requirements in Title 10 Of 1 THE 'MPACT OF ENVIRONMENTAL CONDI-The U S. Code Of Federal Regulations As Codified On December 31 TIONS ON HUMAN PERFOI:IMANCE. A Handbook Of Envvonmental 1891 Exposures-NUREG/CR 5680 V02-THE IMPACT OF ENVtRONMENTAL CONOt-Ferrous Component TIONS ON HUMAN PERFORMANCE. A Cntcal Review Of The Litera^

NUREG/GR 0013: APPLICATONS OF A NEW MAGNETIC MONITOR-ING TECHNIQUE TO IN SITU EVALUATION OF FATIQUE DAMAGE IN FERROUS COMPONENTS.

Environmental Qualification NUREG/CP 0135. WORKSHOP ON ENVIRONMENTAL QUALIFICATION I

OF ELECTRC EOUlPMENT.

pyg y g gg g gg NUREG/CR-5904: FUNCTONAL ISSUES AND ENVIRONMENTAL PROTECTION.

QUALIFICATION OF DIGITAL PROTECTION SYSTEMS OF AD-NUR 593 MMA W

PLETED UNDER THE UR CR4120- CONTROLLED FIELD STUDY FOR VAllDATION OF ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUALIFICATION RE-VADOSE ZONE TRANSPORT MODELS.

SEARCH PfsOGRAM (EDOP)-

FinW DW Approvd Event Reporting Guidehne NUREG-1242 V03 PT01: NRC REVIEW OF ELECTRIC POWER RE-NUREG-1022 R01 DR FC: EVENT REPORTING GUIDELINES SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILI-10CFR50.72 AND 50.73.Second Draft For Comment.

TY REQUIREMENTS DOCUMENTS. Passive Plant Designs. Chapter 1.Propect Number 669.

Event Tree NUREG 1242 V03 PT02-NRC REVIEW OF ELECTRIC POWER RE.

NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILl-HANDS-OH INTEGRATED REllABILITY EVALUATIONS (SAPHIRE)

TY REQUIREMENTS DOCUMENT. Passive Plant Dessgns. Chapters 2-VERSION 5.0. Fault Tree. Event Tree, And Ppng & Instrumentation 13 Protect Number 669 Diagram (FEP) Editors Reference Menual NUREG-1462 Vot: FINAL SAFETY EVALUATION REPORT RELATED TO THE CERTIFICATION OF THE SYSTEM 80+ DESIGN Chapters 1 Examiner Standard

14. Docket No. 52402. (Asea Brown Boven Comtzstion Engineenng)

J NUREG-1G21 R07 S01: OPERATOR LICENSING EXAMINER STAND-NUREG-1462 V02: FINAL SAFETY EVALUATION REPORT RELATED ARDS.

TO THE CERTIFICATION OF THE SYSTEM 80 + DESGN Chapters NUREG-1478: NON POWER REACTOR OPERATOR LICENSING EXAM-15-22 And Appendees. Docket No. 52402.(Asea Brown Bover6Com-INER STANDARDS.

bustion Engineenng)

Emportment LP-FP-2 Final Environmentalimpact Statement NUREG/CR4160

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

NUREG-1484 V01: FINAL ENVIRONMENTAL IMPACT STATEMENT RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP-2.

FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-FEP RICHMENT CENTER, HOMER, LOUIStANA. Docket No. 70-3070.Louisi-ana Energy Servces. L.P. Envronmental impact Statement.

I NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR NUREG-1484 V02: FINAL ENVIRONMENTAL IMPACT STATEMENT HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-VERSION 5.0 Fautt Tree, Event Tree. And Ppng & Instrumentaten Diagram (FEP) Editors Reference Manual RICHMENT CENTER. HOMER, LOUISIANA Docket No 70-3070 Louisi-ana Energy Servces. L P. Comments And Responses.

EG-1426 V02 COMPILATON OF REPORTS FROM RESEARCH RE S0 C F NAL ENVIRONMENTAL STATEMENT RELAT-SUPPORTED BY THE MATERIALS ENGINEERING ED TO THE OPERATION OF WATTS BAR NUCLEAR PLANT UNITS 1 NUR C 535 E

E S TE AND FATIGUE TO.

AND 2 Draft Report For Comment. Docket Nos. 50-390 And 50-FAILURE DATA FOR BRANCH CONNECTONS AND TEES IN RELA.

391.(Tennessee Valley Authonty)

TION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER PIPING SYSTEMS-Final Safety Evaluation Report NUREG-1462 V01. FINAL SAFETY EVALUATON REPORT RELATED Fatigue Crack TO THE CERTIFCATON OF THE SYSTEM 80+ DESIGN. Chapters 1-14 Docket No. 52402. (Asea Brown Boven Combustion Engineenng)ED NUREG/CR4176: REVIEW OF ENVIRONMENTAL EFFECTS ON FA, NUREG 1462 V02: FINAL SAFETY EVALUATION REPORT RELAT TGUE CRACK GROW 1H OF AUSTENITC STAINLESS STEELS.

TO THE CERTIFICATON OF THE SYSTEM 80+ DESIGN Chapters Fatigue Crack initiation 15-22 And Appendees. Docket No. 52 002.(Asea Brown Boven-Com-NUREG/CR4237: STATISTICAL ANALYSIS OF FATIGUE STRAIN-LIFE bustion Engineenng)

DATA FOR CARBON AND LOW-ALLOY STEELS NUREG 1486. FINAL SAFETY EVALUATION REPORT TO LICENSE THE CONSTRUCTION AND OPERATON OF A FACILITY TO Fettgue Damage RECEIVE STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NOREG/GR-00 t3: APPLEATIONS OF A NEW MAGNETIC MONITOR.

NEAR CLIVE.UT AH Docket No. 40-8989 (Envrocare of Utah.inc.)

ING TECHNOUE TO IN SITU EVALUATON OF FATIQUE CAMAGE NUREG-1503 V01: FINAL SAFETY EVALUATION REPORT RELATED IN FERROUS COMPONENTS.

TO THE CERTIFICATION OF THE ADVANCED BOILING WATER RE.

ACTOR DESIGN Docket No. 52 001 (General Electre Nuclear Energy)

Faun thrheent Hazard NUREG-1503 V02: FINAL SAFETY EVALUATON REPORT RELATED NUREG-1494: STAFF TECHNICAL POSITION ON CONSIDERATION OF TO THE CERTIFICATION OF THE ADVANCED BOILING WATER RE-FAULT DISPLACEMENT HAZARDS IN GEOLOGIC REPOSITORY ACTOR DESIGN Appendcas. Docket No.52-001.(General Electrc Nu-DESIGN.

clear Energy)

86 Subject index Financial Statement Fuel Damage NUREG-1470 V03: FINANCIAL STATEMENT FOR FISCAL YEAR 1993-NUREG/CR-6077: DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST Vl4.

Fire Damage NUREG/CR4103: PRIORITIZATION OF REACTOR CONTROL COMPO-Fud Debrie N TS SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEQUENCE OF NUREG/CR4185: TMI-2 INSTRUMENT NO22LE EXAMINATIONS AT ARGONNE NATIONAL LABORATORY. February 1991 - June 1993.

NUREG/CR-6105: EXAMINATION OF RELOCATED FUEL DEBRIS AD-Fiscal Year NUREG 1100 V10 BUDGET ESTIMATES. Fiscal Year 1995.

JACENT TO THE LOWER HEAD OF THE TMI-2 REACTOR VESSEL Flee 6on Product Fuel Performance Report NUREG/CR4077: DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR-3950 V09: FUEL PERFORMANCE REPORT FOR 1991.

RELEASE TEST VI4.

Fuel-Coolant interaction NRG P 27 EE S OF CSNI SPECIAUSTS MEET.

E CR 1 P MARY SYSTEM FISSION PRODUCT RELEASE AND TRANSPORT.A State Ol The-Art Report To The Comrmttee On NUREG/CR4211: INTEGRATED FUEL-COOLANT INTERACTION (IFCI The Safety Of Nuclear installabons.

6.0) CODE. User's Manual.

Floelon Product Transport NUREG/CR4193: PRIMARY SYSTEM FISSION PRODUCT RELEASE Full-Th6cknese Clad Bedm AND TRANSPORT.A State-Ol-The-Art Report To The Comrmttee On NUREG/CR4228: PREUMINARY ASSESSMENT OF THE FRACTURE, The Safety OI Nuclear installations.

BEHAVIOR OF WELD MATERIAL IN FULL THICKNESS CLAD BEAMS.

Fitnese For Duty NUREG/CR-5758 V04: FITNESS FOR DUTY IN THE NUCLEAR POWER Ganna Ray Detector INDUSTRY. Annual Summary Of Program Performance Reports CY NOREG/CR-4833: LARGE AREA SELF-POWERED GAMMA RAY 1993-DETECTOR. Phase 11 Development Of A Source Position Morwtor For Use On industnal Radographe Uruts.

Fhed %

NUREG/CR4234. VALIDATION OF ANALYSIS METHODS FOR AS-SESSING FLAWED PIPtNG SUOJECTED TO DYNAMIC LOADING.

UREG/CR-5403: PREDICTING THE PRESSURE DRIVEN FLOW OF Fracture Mechanice GASES THROUGH MICRO-CAPILLARIES AND MICRO-ORIFICES.

NUREG-1426 V02: COMPtLATION OF REPORTS FROM RESEARCH SUPPORTED BY THE MATERIALS ENGINEERING General Deelgrt Criterla d BRANCH,DfVISION OF FNGINEERING 1991-1993.

NUREG/CR4281: A SIMPUFIED LEAK.BEFORE-OREAK EVALUATION NUREG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

PROCEDURES FOR AUSTENITIC AND FERRITIC STEEL PIPING.

NUREG/CR-4219 V10 N1: HEAVY SECTICN STEEL TECHNOLOGY PROGRAM Sermannual Pregress Report For October 1992 - March Generalized Seamforming 2E MM & M OM MNMM WM NL G/CR-4599 V03 N2: SHOrtr CRACKS IN PIPING AND PtPING MONITORING SYSTEM.

WELDS Sermannual R October 1992. March 1993.

NUREG/

H01 LUATION AND REFINEMENT OF LEAK.

Generic Environmentallmpact Statement NUREG/CR-5591 V02 N1: HEAVY-SECTION STEEL IRRADIATION NUREG 1496 V1 DFC: GENERIC ENVIRONMENTAL IMPACT STATE.

PROGRAM Sermannual Progress Report For October 1990 - March MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE.

1991.

RIA FOR DECOMM!SSIONING OF NRC-UCENSED NUCLEAR NUREG/CR-5591 V02 N2: HEAVY SECTION STEEL IRRADIATION FACILITIES Main Report. Draft Report For Comment.

PROGRAM Sarmannual Progress Report For Apn'i-September 1991.

NUREG 1496 V2 DFC: GENERIC ENVIRONMENTAL IMPACT STATE.

NUREG/CR4228. PREUMINARY ASSESSMENT OF THE FRACTURE MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-BEHAVIOR OF WELD MATERIAL IN FULL THICKNESS CLAD RIA FOR DECOMMISSIONING OF NRC-UCENSED NUCLEAR N

/CR4233 V01: STABlUTY OF CRACKED PIPE UNDER INER' NUR 1

KGR AS A AL RADIOACTIVITY CRITERION FOR DECOMMISSIONING. Appendix A To The Draft Go-NU EG 4

L D Tl OF YSIS METHODS FOR AS.

SESSING FLAWED PIPING SUBJECTED TO DYNAMIC LOADING nonc Enwonmental impact Statement in Support Of Rulemalun9 On Radiological Cntena For Decommissiorun9 Of NRC..

Fracture Toughnese NUREG/CR-4513 R01: ESTIMATION OF FRACTURE TOUGHNESS OF Generic Safety lseuse CAST STAINLESS STEELS DURING THERMAL AGING IN LWR SYS-NUREG 0933 S17: A PRIORITI2ATION OF GENERIC SAFETY ISSUES.

TEMS.

NUREG/CR4051: EFFECTS OF TENSILE LOADING ON UPPER SHELF Geochemistry FRACTURE TOUGHNESS.

NUREG/CR4288 GEOCHEMICAL INVESTIGATIONS RELATED TO NUREG/CR4132; BIAXIAL LOADING AND SHALLOW FLAW EFFECTS THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NUCLE-ON CRACK TIP CONSTRAINT AND FRACTURE TOUGHNESS-AR WASTE REPOSITORY' NUREG/CR4139-CHACK-ARREST TESTS ON TWO IRRADIATED O'0 D 8 N'PO' D NU 162 OF IL A

NUREG.1323 R00: LICENSEE APPUCATION REVIEW PLAN FOR A AGE FRACTURE TOUGHNESS IN THE TRANSITION REGION.

NUREG/CR4177: ASSESSMENT OF THERMAL EMBRITTLEMENT OF GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-CAST STAINLESS STEELS.

LEVEL RADIOACTIVE WASTE.

NUREG/CR4228: PREUMINARY ASSESSMENT OF THE FRACTURE NUREG-1494: STAFF TECHNICAL POSITION ON CONSIDERATION OF BEHAVIOR OF WELD MATERIAL IN FULL. THICKNESS CLAD FAULT DtSPLACEMENT HAZARDS IN GEOLOGIC REPOSITORY BEAMS DESIGN.

NUREG/CR4249: UNIRRADIATED MATERIAL PROPERTIES OF MID-NUREG 1495: OVERALL REVIEW STRATEGY FOR THE NUCLEAR NU

/CR C EAVAGE BEHAVIORS IN NUCLEAR VESSEL p

gy STEELS.

NUREG/CR-5919 REPOSITORY OPERATIONAL CRITERIA ODMPARA.

TIVE ANALYSIS.

Fractured Rock NUREG/CR.6203. VAUDAflON STUDIES FOR ASSESSING UNSATU-RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

Fragmentation WATER REACTOR SAFETY INFORMATION MEETING.Pnmary NUREG/CR4133: FRAGMENTATION AND OUENCH BEHAVIOR OF System lntegnty; Aging Research, Products & Appicat ons: Structural &

CORIUM MELT STREAMS IN WATER Seestnic Engineenng; Seismology & Geology.

Subject index 87 M. _ _ _ Redlet6on Human Factore Engineering Program NUREGICR-5990. THE EFFECTS OF SOLAR-GEOMAGNETICALLY IN-NUREG-0711: HUMAN FACTORS ENGINEERING PROGRAM REVIEW DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR MODEL POWER STATIONS.

Human Performance k oetigonen NUREG/CR-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI.

. NUREG/CR-3145 V10- GEOPHYSICAL INVESTIGATIONS OF TH'i WESTERN OHIO-INDIANA REGION. Final Report, October 1986 Sep.

TlONS ON HUMAN PERFORMANCE. A Handbook Of Environmental tanbw m2.

Nu CR 5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI.

Ground MoWen TIONS ON HUMAN PERFORMANCE. A Cnecal Review Of The Usera-NUREG-1488: REVISED UVERMORE SEISMIC HAZARD ESTIMATES ture.

FOR SIXTY-NINE NUCLEAR POWER PLANT SITES EAST OF THE ROCKY MOUNTAINS. Final Report.

Human m NUREG/CR4093: AN ANALYSIS OF OPERATIONAL EXPERIENCE Groundweter DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-NUREG/CR4114 V03: PERFORMANCE ASSESSMENT OF A HYPO-DRESSING HUMAN REUABluTY ASSESSMENT ISSUES.

THETICAL LOW-LEVEL WASTE FACIUTY. Groundwater Flow And NUREG/CR4208: AN EMPIRICAL INVESTIGATION OF OPERATOR Transport Semulebon, PERFORMANCE IN COGNITIVELY DEMANDtNG SIMULATED EMER.

g GENCIES.

NUREG/CR4236: SEISMIC INVESTIGATIONS OF THE HDR SAFETY Human-System lnterfeos l

PROGRAM.Sunnary Report NUREG/CR-5908 V01: ADVANCED HUMAN-SYSTEM INTERFACE HPCIPump DESIGN REVIEW GUIDELINE. General Evaluation Model, Tectwcal NUREG-1275 V10: OPERATING EXPERIENCE FEEDBACK REPORT -

Developrnent And NW W RELIA 81UTY OF SAFETY-RELATED STEAM TURBINE-DRIVEN NUREG/CR-5906 V02: ADVANCED HUMAN-SYSTEM INTERFACE STANDBY PUMPS. Commencal Poww Ream DESIGN REVIEW GUIDELINE. Evaluation Procedures And Gudennes For Human Factors Ergneenne Reviews.

Health Phyence NUREG/CR-5569 RO1: HEALTH PHYSICS POSITONS DATA BASE.

,Mr----

Load NUREG/CR4204: QUESTIONS AND ANSWERS BASED ON REVISED ' NUREG/LA 0093: RELAPS/ MOD 3 ASSESSMENT FOR CALCULATION 10 CFR PART 20.

OF SAFETY AND RELIEF VALVE DISCHARGE PIPING HYDRODY-NAMIC LOADS.

T% W NUREG/CR-4219 V10 N1: HEAVY SECTION STEEL TECHNOLOGY Hydrogen Mining Study PROGRAM.Sermannual Progress Report For October 1992 March M3.

NUREG/CR4180: HYDROGEN MIXING STUDIES (HMS): USER'S MANUAL

"__., t _ _ _i Steelirredletion Program NUREG/CR-5591 V02 N1: HEAVY SECTON STEEL IRRADIATION "N

'I PROGRAM.Semaannual Progress Report For October 1990. March NUREG/CR-4918 VOT: CONTROL OF WATER INFILTR/u.ON INTO iggj.

NEAR SURFACE LLW DISPOSAL UNITS. Progress Report On Field Ex.

NUREG/CR-5591 V02 N2. HEAVY SECTON STEEL IRRADIATON penments At A Hurmd Region Sste,Beltsville Maryland.

PROGRAM Serrmannual Progress Report For Apni-September 1991, p

High Pressure Meet Election NUREG/lA 0093: RELAP5/ MOD 3 ASSESSMENT FOR CALCULATION NUREG/CR-6044. EXPERIMENTS TO INVESTIGATE DIRECT CON-OF SAFETY AND REUEF VALVE DISCHARGE PIPING HYDRODY-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NAMIC LOADS.

IlON NUCLEAR POWER PLANT IN THE SURTSEY TEST FACluTY.

NUREG/lA 0114: ASSESSMENT OF RELAP5/ MOD 3 WITH THE LOFT NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DIRECT CON-(g.1/(3 3 EXPERIMENT SIMULATING AN ANTICIPATED TRANSIENT TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE WITH MULTIPLE FALLURES.

SURRY NUCLEAR POWER PLANT.

High Level Redloactive Weste NUREG-1323 ROO LICENSEE APPUCATON REVIEW PLAN FOR A NUREG/CR4063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-CRUCES TRENCH SITE.

LEVEL RADIOACTIVE WASTE.

NUREG-1495: OVERALL REVIEW STRATEGY FOR THE NUCLEAR ggggg REGULATORY COMMISSON'S HIGH-LEVEL WASTE REPOSITORY NUREG/CR.6116 V01: SYSTEMS ANALYSIS PROGRAMS FOR PROGRAM' HANDS-ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

VERSION 5.0.Techrucal Reference Manual.

i Human Error NUREG/CR4116 V02: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR-4639 V5R4P2 NUCLEAR COMPUTERIZED UBRARY FOR HANDS-ON INTEGRATED REUABluTY EVALUATONS (SAPHIRE)

ASSESS!NG REACTOR REUABluTY (NUCLARR). Volume 5: Data VERSION 5 0. Integrated Rehability And Resk Analysis System (tARAS) 1 Manual.Part 2 Human Error Probatzlety (HEP) Data.

Reference Manual.

j NUREG/CR-4639 V5R4P3: NUCLEAR L'OMPUTER12ED UBRARY FOR NUREG/CR4116 V03: SYSTEMS ANALYSIS PROGRAMS FOR ASSESSING REACTOR REUA8luTY (NUCLARR). Volume 5: Data HANDS-ON INTEGRATED REUA81UTY EVALUATIONS (SAPHIRE) 1 Manual.Part 3: Hardware Component Fadure Data.

VERSION 5 0. integrated Rehabehty And Rash Analysis System (IRRAS)

NUREG/CR4208. AN EMPIRICAL INVESTIGATION OF OPERATOR Tutonal Manual.

PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER-NUREG/CR4145: VERIFICATION AND VAUDATON OF THE SAPHIRE GENCIES.

VERSION 4 0 PRA SOFTWARE PACKAGE.

Human Factor ISLOCA NUREG/CR4105: HUMAN FACTORS ENGINEERING GUIDANCE FOR THE REVIEW OF ADVANCED ALARM SYSTEMS.

WMRWt WOW WE W M WNWW LOSS OF COOLANT ACCIDENTS IN AD ' JCED UGHT WATER RE-Human Factore Engineering ACTORS.

NUREG/CR-5908 V01: ADVANCED HUMAN-SYSTEM INTERFACE DESIGN REVIEW GUIDEUNE. General Evaluabon Model. Tectncal Inc6 dent Roeponse Development. And Guidehne Descnotion.

NUREG-1471: CONCEPT OF OPERATONS WITH ORGAN 12ATON NUREG/GH-5908 V02. ADVANCED HUMAN-SYSTEM INTERFACE CHARTS.NRC incadent Response.

DESIGN REVIEW GUIDELINE. Evaluahon Procedures And Gudelines For Human Factors E Revews.

Independent Code Asaesoment NUREG/CR4146: LOCA CON ROL STATONS. HUMAN ENGINEER-NUREG/CA 5535 V07: RELAP5/ MOD 3 CODE MANUAL $ummenes And

)

ING ISSUES AND INSIGHTS.

Reviews Of independent Code Assessment Reports.

)

4 I

1

88 Subject index Industry Survey NUREG/CR-6144 V02PIB: EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR4278: SURVEY OF INDUSTRY METHODS FOR PRODUC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ING HIGHLY REUABLE SOFTWARE.

SURRY, UNIT 1. Analysis Of Core Damage Frequency Frorn Intemal Events Dunng M4 Loop Operations Main Report (Chapters 712).

Strese NUREG/C46144 V02P2: EVALUATON OF POTENTIAL SEVERE ACCI-NUREG/CR4233 Vot: STABluTY OF CRACKED PIPE UNDER INER.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT TIAL STRESSES. Subtask 1.1 Final Report SURRY, UNIT 1.Analyss Of Core Damage Frequency From intemal Events Dunng M4 Loop Operations.Appendeces A-D.

Information Bias NUREG/CR4144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/G40011: INFORMATION BIAS AND UFETIME MORTAUTY CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT RISKS OF RADtATON-INDUCED CANCER. Low LET Radatiott SURRY, UNIT 1.Analyss Of Core Damage Frequency From Internal Events Dunng M4 Loop Operatons Appendices E (Sections E.1-E.8) information D60est NUREG/CR4144 V02P38: EVALUATION OF POTENTIAL SEVERE AC-NUREG-1350 V06: NUCLEAR REGULATORY COMMISSION INFORMA, CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT TON DIGEST.1994 Edition.

SURRY, UNIT 1.Analyss Of Core Damage Frequency From intamal input Processor Events Dunng M4 Loop Operations Appendices E (Sections E 9-E.16).

NUREG/CR4182 V01: OFFSCALE: A PC INPUT PROCESSOR FOR NUREG/CR4144 V02P4: EVALUATION OF POTENTIAL SEVERE ACCl-THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT The Cnticality Sequences.

SURRY, UNIT 1.Analyss Of Core Damage Frequency Frorn :ntemal NUREG/CR4182 V02: OFFSCALE: A PC INPUT PROCESSOR FOR Events Dunng M4 Loop Operations. Appendices F-H.

THE SCALE CODE SYSTEM Volume 2: The ORIGNATE Processor for NUREG/CR4144 V02PS: EVALUATION OF POTENTIAL SEVERE ACCl-ORIGEN-S.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY, UNIT 1 Analyss Of Core Damage Frequency Frorn internal G/

6 V02-EVALUATON OF SAMPUNG PLANS FOR IN-SERVICE INSPECTION OF STEAM GENERATOR Internal Fire TUBES. Comprehensive Analytical And Monte Carlo Simulation Results NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-For Several Sampling Plans.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/C45985: EVALUATON OF COMPUTER-BASED ULTRASONIC SURRY, UNIT 1. Analyses Of Core Damage Frequency Frorn Intemal INSERVICE INSPECTON SYSTEMS.

Fires Dunno M4 Loop Operations Main Report NUREG/CR4151: FEASI8luTY OF DEVELOPING RISK-BASED RANK

• NUREG/CR-6144 V03 P2: EVALUATION OF POTENTIAL SEVERE AC-INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVICE IN.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY, UNIT 1.Analyss Of Core Damage Frequency Frorn Intemal NU G CR 6181: A PILOT APPUCATION OF RISK BASED METHODS Fires Dunng M4 Loop Opsatons. Appendices.

TO ESTABUSH INSERVICE INSPECTON PRIORITIES FOR NUCLE-AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION.

Internal Flood NUREG/CR4144 V04: EVALUATION OF POTENTIAL SEVERE ACCl-inserv6ce Testing DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CP-Ol37 V01: PROCEEDINGS OF THE THIRD NRC/ASME SURRY, UNIT 1. Analysts Of Core Damage Frequency From Intemal SYMPOSIUM ON VALVE AND PUMP TESTING. Held At The Hyatt Re_

Floods Dunng M4 Loop Operations.

gency Hotel, Washmgton,DC, July 18-21,1994.Sesson 1 A Sessson 2C I"I*'Ph***

'O NUREG/CP-0137 V02: PROCEEDINGS OF THE THIRD NRC/ASME NUREG/CR-5535 V07: RELAP5/MODJ CC')E MANUALSummenes And SYMPOslUM ON VALVE AND PUMP TESTING.Hetd At The Hyatt Re-Rm Q Independent Code Anesswt Reports.

gency Hotel, Washington.DC. July 18-21,1994. Session 3A -Session

1. 0-Investigation NUREG/CR4088:

SUMMARY

OF 1991 1992 MISADMINISTRATION inspect 6on Gu6de EVENT INVESTIGATONS.

NUREG/CR-5830: AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-SPECTON GUCE FOR THE MCGUIRE NUCLEAR POWER PLANT.

lodine 131 NUREG-1492 DFC. REGULATORY ANALYSIS ON CR.TERIA FCR THE inspection Reporg RELEASE OF PATIENTS ADMINISTERED RADOACTIVE NUREG/C45812-MANAGING AGING IN NUCLEAR POWER MATERIALDraft Report For Comment.

PLANTS. Insights From NRC Maintenance Team inspection Reports Irrad6ated High Copper Wold Instrument Nozzle NUREG/CR4185: TMI-2 INSTRUMENT NOZZLE EXAMINATONS AT NUREG/CR-6139: CRACK ARREST TESTS ON TWO IRRADIATED HIGH-COPPER WELDS. Phase 11: Results Of Duplex Type Specamens.

ARGONNE NATIONAL LABORATORY February 1991. June 1993 NUREG/CR 6198: TMI-2 INSTRUMENT NOZZLE EXAMINATONS PER-KEY Analysie System FORMED AT THE INEL NUREG/CR4290 KEY ANALYSIS SYSTEM USER'S GUIDE. Version 2.0.

Instrumentation And Control System NUREG/CR-5941: TECHNICAL BASIS FOR EVALUATING ELECTRO, LOCA "4UREtbCR4107:

SUMMARY

OF MELCOR 1.8.2 CALCULATONS FOR MAGNETIC AND RADIO-FREQUENCY INTERFERENCE IN SAFETY-

! sREE LOCA SEQUENCES (AG S2D & S30) AT THE SURRY PLANT.

RELATED I&C SYSTEMS' NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL Intergranular Strees Corroeion Cracking FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT-NUREG/C45314 V05. INSIGHTS FOR AGING MANAGEMENT OF ED DEBRIS.Dra't For Comment MAJOR LWR COMPONENTS METAL CONTAINMENTS.

Interim Ucensing Criteria NUREG/C44513 RO1: ESTIMATION OF FRACTURE TOUGHNESS OF NUREG-1497: INTERIM UCENSING CRITERIA FOR PHYSICAL PRO.

CAST STAINLESS STEELS DURING THERMAL AGING IN LWR SYS-TECTON OF CERTAIN STORAGE OF SPENT FUEL TEMS.

NUREG/C44667 V17: ENVIRONMENTALLY ASSISTED CRACKING IN Internal Event UGHT WATER REACTORS. Semiannual Report Apr91993 Septem-NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABILISTIC ber 1993.

RISK ASSESSMENT.

NUREG/CR-5314 V05: INSIGHTS FOR AGING MANAGEMENT OF NUREG/CR6143 V03: EVALUATION OF POTENTIAL SEVERE ACCl-MAJOR LWR COMPONENTS METAL CONTAINMENTS.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From In-STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPtNG AT temal Events For Plant Operationsa State 5 Durmo A Refuehno Outage.

LWR TEMPERATURES.

NUREG/CR4144 V02PI A EVALUATON OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Large Rolesee SURRY, UNIT 1. Analysis Of Core Damage Frequency From intemal NUREG/C46094: CALCULATONS IN SUPPORT OF A POTENTIAL Events Dunn0Mid-Loop Operations. Main Repwt (Chapters 14)

DEFINITION OF LARGE RELEASE.

Subject index 89 Las Cruces Trench Uconsed Fuel NJitty Status Report NUREG/CR-6063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS NUREG-0430 v13:

LICENSED FUEL FACILITY STATUS CRUCES TRENCH SITE.

REPORT.Inventos/ Difference Data. July 1,1992 June 30,19931 Gray Book it)

NUREG/CR-6164; RELEASE OF RADIONUCUDES AND CHELATING Ucensed Operating Reactom AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATION LOW-NUREG-0020 V18: UCENSED OPERATING REACTORS STATUS SUM-LEVEL hADIOACTIVE WASTE COLLECTED FROM THE PEACH MARY MDORT. Data As Cf December 31,1993 (Gray Book l)

BOTT JM AlOMIC POWER STATON UNIT 3.

NUREG 'CR4201. COMPRESSION AND IMMERSION TESTS AND Ucensee Ennt Report LEAC,11NG OF RADIONUCLOES. STABLE METALS, AND CHELATING NUREG/CR-4674 V17: PRECURSOPS TO POTENTML SEVERE CORE AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATION WASTE DAMA05 ACCIDENTS: 1992 A S,ATUS REPORT. Vain Report And COLLECTED FROM NUCLEAR POWER STATONS.

Appenda A.

NUHEG/OR 4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE Leak Rate DAMAGE ACCOENTS: 1992 A STATUS REPORT.Apisereces B, C, D, NUREG/CR-5128 ROI: EVALUATION AND REFINEMENT OF LEAK-E. F. Ar.1 G.

RATE EST!MATION MODELS.

Ucensee Performance Leek-Before-8 reek NUREG/CR-5680 VOI: THE IMPACT Ol' ENVIF,0NMENTAL &lNDl-NUREG/CR4281: A SIMPLIFIED LEAK-BEFORE-BREAK EVALUATION TIONS ON HUMAN PERFORMANCE. A Handtook Of Environmental PROCEDURES FOR AUSTENITIC AND FERRITIC STEEL PIPtNG.

Exposures.

NUREG/CR-5680 V02: THE IMPACT OF ENVIRONMENTAL CONDl-Legalleeuences TIONS ON HUMAN PERFORMANCE. A Cntical Revww Of The Uters-NUREG4750 V37: NUCLEAR REGULATORY COMMISSION ture.

ISSUANCES.Opnons And Decmons Of The Nuclear Regulatory Corn-russion With Selected Orders. January June 1993.

Ught Water Reactor NUREG-0750 V38-NUCLEAR REGULATORY COMMISSON NUREG/CR 4513 R01: ESTIMATON OF FRACTURE TOUGHNESS OF ISSUANCES.Opnois And Decrasons Of The Nuclear Regulatory Com.

CAST STAINLESS STEELS DURING THERMAL AGING IN LWR SYS-rmssson With SeleM rirders. M*-Oecember 1993.

TEMS.

NUREG-0750 V38101: INutAES 'TO NUCLEAR REGULATORY COM.

NUREG/CR-4667 V17. ENVIRONMENTALLY ASSISTED CRACKING IN MISSON ISSUANCES. July-September 1993.

UGHT WATER REACTORS. Sermannual Report.Apnl 1993 - Septem-NUREG4750 V38102-INDEXES TO NUCLEAR REGULATORY COM.

ber 1993.

MISSON ISSUANCES. July Oscomber 1993.

NUREG/CR-5314 VOS: INSIGHTS FOR AGING MANAGEMENT OF NUREG4750 V38 N05: NUCLEAR REGULATORY COMMISSON IS.

MAJOR LWR COMPONENTS METAL CONTAINMENTS.

SUANCES FOR NOVEMBER 1993. Pages 187 268.

NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE NUREG-0750 V38 N06: NUCLEAR REGULATORY COMMISSON IS-STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPING AT SUANCES FOR DECEMBER 1993. Pages 289-391.

LWR TEMPERATURES.

NUREG-0750 V39101: INDEXES TO NUCLEAR REGULATORY COM-MISSION ISSUANCES January -March 1994.

Uquefaction Method NUREG-0750 V39102: INDEXES TO NUCLEAR REGULATORY COM_

NUREG/CR4258: THE UQUEFACTION METHOO FOR ASSESSING PA-MISSON ISSUANCES. Ja June 1994.

LEOSEISMICITY.

NUREG 0750 V39 N01:

AR FIEGULATORY COMMISSION IS-SUANCES FOR JANUARY 1994. Pages 145.

Uguld Effluent NUREG 0750 V39 NO2 NUCLEAR HEGULATORY COMMISSON IS-NUREGK12907 V12: RADIOACTIVE M*4ERIALS RELEASED FROM SUANCES FOR FEBRUARY 1994. Pages 47 90.

NUCLEAR POWER PLANTS. Annual 6eport 1991.

NUREG-0750 V39 NO3: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR MARCH 1994.P 91 186.

Uquideetal Reactor NUREG-0750 V39 N04: N REGULATORY COMMISSON IS-NUREG-1368: PREAPPUCATION SAFE'Y EVALUATON REPORT FOR SUANCES FOR APRIL 1994. Pages 187 247.

THE POWER REACTOR INNOVATi\\E SMALL MODULE (PRISM)

NUREG 0760 V39 N05: NUCLEAR REGULATORY COMMISSON IS.

UQUID' METAL REACTORFinal V'.

SUANCES FOR MAY 1994.P 249-284.

NUREG-0750 V39 N06: N AR REGULATORY COMMISSION IS-Long-Term Performance SUANCES FOR JUNE 1994 Pages 285 390.

NUREG/CH-6138: USEF. S GUIDE FOR SIMPUFIED COMPUTER NUREG-0750 V40101: INDEXES TO NUCLEAR REGULATORY COM-MODELS FOR THE ES11MATION OF LONG-TERM PERFORMANCE MISSON ISSUANCES. July-September 1994.

OF CEMENT BASED MATERIALS.

NUREG-0750 V40 NOI: NUCLEAR REGULATORY COMMISSON IS-SUANCES FOR JULY 1994.P 1-41.

LoseOf4oolant Accident NUREG-0750 V40 N02 REGULATORY COMMISSION IS-NUREG/CR4095: AGING. LOSS-OF-COOLANT ACCIDENT (LO%),

SUANCES FOR AUGUST 1994 Pages43-132.

AND HIGH POTENTIAL TESTING OF DAMAGED CABLES.

NUREG4750 V40 NO3: NUCLEAR REGULATORY COMMISSON iS-NUREG/CR4121: COMPONENT EVALUATION FOR INTs:RSYSTEM SUANCES FOR SEPTEMBER 1994. Pages 133-145.

LOSS-OF COOLANT ACCOENTS IN ADVANCED UGHT WATER RE-NUREG-0750 V40 N04. NUCLEAR REGULATORY COMMISSION IS.

ACTORS.

SUANCES FOR OCTOBER 1994. Pages 147-167.

gp Leesons Learned NUREG/CR4093: AN ANALYSIS OF OPERATIONAL EXPERIENCE NUREG/CR4252-LESSONS LEARNED FROM THE THREE MILE DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-ISLAND 4) NIT 2 ADVISORY PANEL DRES$4NG HUMAN FIELIABILITY ASSESSMENT ISSUES.

NUREG/CR4143 V02P1A: EVALUATON OF POTENTIAL SEVERE AC-Lethel Radiation Does CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4094: CALCULATONS IN SUPPORT OF A POTENTIAL GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-DEFINITON OF LARGE RELEASE.

temal Events For Plant Operational State 5 Dunng A Refuehng Outape Sections 19.

Ucense ?_;, _ _. _.

NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC.

NUREG-1200 R03: STANDARD REVIEW PLAN FOR THE REVIEW OF A CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LICENSE APPLICATON FOR A LOW-LEVEL RADIOACTIVE WASTE GRAND GULF,UNfT 1.Analysas Of Core Damage Frequency Frorn in.

DISPOSAL FACIUTY.

temal Events For Plant Operational State 5 Dunng A Refuehng th= hm 10.

Ucense Application Review Plan NUREd/CR4143 V02P1C EVALUATION OF POTENTIAL SEVERE AC-NUREG-1323 F100 UCENSEE APPUCATON REVIEW PLAN FOR A CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HtGH-GRAND GULF, UNIT 1.Analyssa Of Core Damage Frequency From in.

LEVEL RADCACTfVE WASTE.

temal Events For Plant Operational State 5 Dunng A Refuehng NUREG-1495: OVERALL REVIEW STRATEGY FOR THE NUCLEAR Outage Mam Report REGULATORY COMM!SSION'S HIGH-LEVEL WASTE REPOSITORY NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVERE AC.

PROGRAM COENTS DUR!NG LOW POWER AND SHUTDOWN OPERATONS AT s

1

,i i,

90 Subject index GRAND GULF. UNIT 1.Analyws Of Core Darnage Frequency Frorn In-NUREG/CR4147 V03 CHARACTER 12AT!ON OF CLASS A LOW-LEVEL 7

tornal Events For Plant Operational State 5 Dunng Refuehng RADIOACTIVE WASTE 1986-1990 Main Report Part B i

Outage inteenal...

NUREG/CR4147 V04. CHARACTERIZATION OF CLASS A LOW LEVEL NUREG/CR4143 V02PT3. EVALUATION OF POTENTIAL SEVERE AC+

RADIOACTIVE WASTE 1986-1990 Appendices A-E.

1 CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6147 V05: CHARACTERIZATION OF CLASS A LOW. LEVEL GRAND GULF, UNIT 1.Analyws Of Core Damage Frequency Frorn In.

RADIOACTIVE WASTE 1986-1990 Apperds F.

tornal Events For Plant Operational State 5 Dunng A Refuehng NUREG/CR 6147 V06 CHARACTERIZATION OF CLASS A LOW. LEVEL Outage intemel....

RADOACT!YE WASTE 19861190 Apperdces G-J i

NUREG/CR4143 V02PT4 EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR-6147 V07: CHARACTERl2ATION OF CLASS A LOW LEVEL COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT RADOACTIVE WASTE 1986 1990.

es K-P.

3 GRAND GULF. UNIT 1. Analyses Of Core Darnage Frequency Frorn in-NUREG/CR-6164. RELEASE OF R UCLIDES AND CHELATING l

tornal Events For Plant Opershonal State 5 Dunng A Refueling AGENTS FROM CEMENT-SOLIDIFIED DECONTAMINATION LOW-Outage InfomaL.

LEVEL RADOACTIVE WASTE COLLECTED FROM THE PEACH 2

NUREG/CR41G VJ3: EVALUATION OF POTENTIAL SEVERE ACCI-BOTTOM ATOMIC POWER STATION UNIT 3 i

DENTP DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6188 VOI: MICROBIAL DEGRADATION OF LOW LEVEL j

GRANO GULF. UNIT 1. Analyses Of Core Damage Frequency From in-RADCACT!VE WASTE Annual Fmport for FY 1993 Iernal Events For Plant Opershonal State 5 Dunng A Refuelmo Outge.

NUREG/CR 6143 V04 EVALUATON OF POTENTIAL SEVERE SCCl-Low-Level Vseste Data Bees 4

DENTS DURING LOW POWER AND SHUTDOWN OPERATIO4S AT NUREG/CF05229 V06: FIELD LYSIMETER INVESTIGATONS. LOW.

j GRAND GULF. UNIT 1. Analysis Of Core Damage Frequency From in-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR ternally Induced Fluoding Events For Plant Operational State 5 Dunne FISCA L YEAR 1993. Annual Floport.

a a Re'uoling...

4 NUREGICR-6143 VOS: EVALUATION OF POTENTIAL SEVERE ACCl-Low-Levei Weste thraad

{

DENTS D'JRING LOW POWER AND SHUTDOWN OPERATONS AT NUREGi";R 4918 V07: CONTROL OF WATER INFILT^lATON INTO r

GRAND GJLF, UNIT 1.Analysee Of Coes Damage Frequency From NEAR GURFACE LLW DISPOSAL UN'f1 Progress 3eport On Field Em 2

j Seismic Ev mts Dunna M4 Leap Operatior s Maen Report.

perrnents At A Humed Region Sste. Belt. ' "dary'ard NUREG/CR4144 V02FI A: F'. ALUATION OF POTENTIAL SEVERE AC-a 3

CIDENTS DURING LOW ROWER AND SHUT 00WN OPERATONS AT Low-Level Weste Desposal FacNNy j

SURRulNIT 1 Analysw Of Core Damage Fisquency From internal NUREG 1200 R03: STANDARD REVIEW PLAN FOf 4 THE REVIEW OF A l

Events Dunng M4 Loop Operations Main Report ' Chapters t 4)

LICENSE APPLICATION FOR A LOW. LEVEL FLW40 ACTIVE WASTE j

NUREG/CR-6144 VO2PIB EVALUATION OF POTENTIAL SEVERE AC-OtSPOSAL FACILITY.

4 CIDENTS DURING LOW POWER AND SHUTDOWts OPERATIONS AT SURRY. UNIT 1. Analyses Of Core Damage Frequency Crom Intema' Lower Head Events Dunng M4 Loop Operations Man Report (Chapters 7.t?)

N9 REG /CR-6181: RESULTS OF MECHANICAL TESTS AND SUPPLE-NUREG/CR 6144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCI MENTARY MICROSTRUCTURAL EXAMINATIONS OF THE TMI-2 DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS A7 LOWER HEAD SAMPLES.

i SURRY. UNIT t. Analyses Of Core Damage Frequency From Interral i

Events Dunng Mid Loop Operations Appendices A-D L w,er Pressure Vessel Head i

NUREG/CR 6144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-6194 METALLOGRAPHIC AND HARDNESS EXAMINA.

CIDENTS DURN 3 LOW POWER AND SHUTDOWN OPERATONS AT TIONS OF TMI-2 LOWER PRESSURE VESSEL HEAD SAMPLES.

SURRY. UNIT t. Analyses Of Core Damage Frequency From 6ntemai Events Dunng i A6 Loop Operations es E (Sectior s li1-E 8)

Lysimeter NUP dG/CR-6144 V02P30. EVALUAT OF POTENTIAL SEVERE AC-NUREG/CF05229 V06. FIELD LYSIMETER INVESTOATIONS LOW-CIDENTS DURidG LOW POWER AND SHUTDOWN OPERATIONS AT LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR 3

SURRY. UNIT 1 Analyses Of Core Damage Frequency From intemal FISCAL YEAR 1993 Annual Report NR CR 4VOP E ON POYENT E

gagg

{

DENTS DURING LOW POWER ANO SHUTDOWN OPERATIONS AT NUROG/CF06053: COMPARISON OF MACCS USERS CALCULATONS i

SURRY UNIT 1 Analyses Of Core Damage Frequency From internal FOR THE INTERNATIONAL COMPARISON EXERCISE ON PROBABI-I LISTIC ACCIDENT CONSEQUENCE ASSESSMENT CODES.

NR CR 44 VO E L PO E lAL SEVERE ACCI-j DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT MELCOR l

SURRY. UNIT 1 Analyses Of Core Damage Frequency From intemal NUREG/CR 5850- ANALYSIS OF LONG-TERM STATION BLACKOUT j

WITHOUT AUTOMATO DEPRESSURl2ATON AT PEACH BOTTOM NR CR 44 V ALU OF ENTIAL SEVERE AC.

j CIDENTS DURING LOW POWER AND SHUTOOWN OPERATIONS AT NUR C

AJ MA MELCOR 1.8 2 CALCULATIONS FOR SURRY. UNIT 1 Analyses Of Core Damage Frequency From internal a

THREE LOI:A CEOUENCES (AG.S2D & S3D) AT THE SURRY PLANT.

Fres Dunng M4 Loop Operations Main Report.

1 i

NUREG/CR 6144 V03 P2: EVALUATION OF POTENTIAL SEVERE AC.

Me asurement j

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT

~

NUREG/GM0tE APPLICATONS OF A NEW MAGNFTIC %40NITOR.

SURRY. UNIT I Ana6ysis Of Core Damage Frequency From Intemal ING TECHNIGUE TO IN SITU EVALUATION OF FATOUE GAMAGE Fres Our Md Loop Opershone Appereces i

NUREG/CR1144 V04 EVALUATION OF POTENTIAL SEVERE ACCI-IN FERROUS COMPONENTS.

j DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Meinknance Practh SURRY,0 NIT 1 Analyses Of Core Damage Frequency From Intemal NUREGICR 5967: DEVELOPMENT AND APPLCATON OF DEGRADE-TON MODELING TO DEFINE MAINTENANCE PRACTICES.

NUR C

44 05 V A ON OF POTENTIAL SEVERE AC.;l-DENTS DURING LOW POWER AND SHUTDOWN OPERATO% AT 4

Mainknence Program SURRY, UNIT 1 Analyses Of Core Damage Frequency From ',eismic NUREG 1416. OPERATONAL EXPERIENCE AND MAINTENANCE PRO.

Events Dunng M4 Loop Operations Main Report-GRAMS OF TRANSAMERCA DELAVAL, INC., DIESEL GENERA-TORS Low-Alloy Heeg NUREG/CR-6237: STATISTICAL ANALYSIS OF FATIGUE SVT!ARLirE NUREG/CR-5812-MANAGING AGING IN NUCLEAR POWER DATA FOR CARBON AND LOW ALLOY STEELS PLA14TS Insights From NRC Mamtenance Team inspection Reports tow-Level Medloactive Weste Makveient NUREG/CR 5965-MODEUNG FIELD SCALE UNSATURATED FLOW NUREG/CR-6190 V01: PROTECTON AGAINST MALEVOLENT USE OF l

AND TRANSPORT PROCESSES.

VEHCh ES AT NUCLEAR POWER PLANTS. Vehicle Bamer System j

NUREG/CR4114 V03. PERFORMANCE ASSESSMENT OF A HYPO.

Seting G<mdance For Blast Protectort 1

THETCAL LOW LEVEL WASTE FACILITY. Groundwater Fkiw And NUREG/CW6190 V01 RI: PROTECTON AGAINST MALEVOLENT USE J

Transport Smalation OF VEHIC LES AT NUCLEAR POWER PLANTS Vehicle Barner System NUREG/CR-6147 V01 CHARACTERl2ATION OF CLASS A LOWLEVEL Selection bs1ance For Blast Protecton.

RADIOACTIVE WASTE 19661990 Executive Summary NUREG/CR-6190 V02: PROTECTION AGAINST MALEVOLENT USE OF J

NUREG/Ch6147 V02-CHARACTERIZATON OF CLASS A LOW LEVEL VEHICLES AT NUCLEAR POWER PLANTS Vehicle Bamer System RADOACTIVE W AS TE 19861990 Main Report-Part A S4hng GuKlance For Blast Protection.

1 1

1 l----

Subject index 91

• " JAM /CR4190 V02 R1: PROTECTION AGAINST MALEVCAENT USE Neutron Fluerce OF VErNCLES AT NUCLEAR POWER PLANTS VeNele Barnw System NUREG/CR 6139: CRACK-ARREST TESTS ON TWO IRRADIATED W w % =$ence.

HIGH-COPWI WELDS. Phase it; Results Of Duplex Type Specunens.

MostwM w eties Nondestructive Evolustion ND E M HESULTS OT MECHANICAL TESTS AND SUPPLE-NUREG/CR 5965: EVALUATION OF COMPUTER-BASED ULTRASONIC MEnfARY k:"lROSTRUCTURAL EXAMINATIONS OF THE TMl-2 INSERVICE INSPECTION SYSTEMS.

LOWER HEAD NLES.

NUREG/CR4151: FEASIBILITY OF DEVELOPING RISK 8ASED RANK-m INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVICE IN-NURE'GM4187: RESV FS OF MECHANICAL TESTS AND SUPPLE

• SPECTION NUREG/CR-6181: A PILOT APPLCATION OF R!SK-BASED METHODS MENTARY MICROSTRl/TURAL EXAMINATIONS OF THE TMI-2 TO ESTABLISH INSERVICE INSPECTION PRORITIES FOR NUCLE-LOWER HEAD SAMPLES-AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION.

NUREG/CR4194: METALLOGRAPHC AND HARDNESS EXAMINA*

NUREG/GR-0013: APPLICATIONS OF A NEW MAGNETC MONITOR.

TONS OF TMI 2 LOWER PRESSURE VESSEL HEAD SAMPLES-NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PRCLIECT INTE-ING TECHNIQUE TO IN SITU EVALUATION OF FATIQUE DAMAGE GRATON REPORT.

IN FERROUS COMPONENTS.

Nondestructive Testing

(

NUREG/CR4183: PEER REVIEW OF THE TML-2 VESSEL INVESTtGA-N TON PROJECT METALLURGCAL EXAMINATONS.

L L

AT LIME K 1'

Nosaie Crooking U' REG /CR 5403: PREDCTING THE PRESSURE DRIVEN FLOW OF NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC.

GASES THROUGH MCRO CAPILLARIES AND MORO ORIFICES.

TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING.

adscreestal Degradesion Nuclear Computertred Lahrery i

NUREG/CR4188 V01: MOROBIAL DEGRADATION OF LOW-LEVEL NUREG/CR-4639 V5R4P2: NUCLEAR COMPUTERIZED LIBRARY FOR RADIOACTIVE WASTE Annuel Report For FY 1993.

ASSESSING REACTOR RELIABILITY (NUCLARR). Volume 5: Date i

Manual.Part t Human Error Probetality (HEP) Deta.

I astrco Ortnce NUREG/CR 4639 V5R4P3: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR 5403: PREDICTING THE PRESSURE DRIVEN FLOW OF ASSESSING REACTOR RELIABILITY (NUCLARR). Volume 5: Data i

(

GASES THROUGH MICRO CAPtLLARIES AND MtCRO ORIFICES.

Manual. Pert 3: Hardware Component Festure Data.

teleedministration Event Nuoleer Forrttic Piping NUREG/CR-8088:

SUMMARY

OF 1991 1992 MISADMINISTRATION NUREG/CR4226: EFFECT OF DYNAMC STRAIN AGING ON THE EVENT INVESTnGATIONS.

STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PtPING AT LWR TEMPERATURES.

i teodel Uncertainty NUREG/CP-0138. PROCEEDINGS OF WORKSHOP 1 IN ADVANCED Nucteer RAsterial l

TOPICS IN RISK AND RELIABILITY ANALYSIS Model Uncertainty its NUREG/CR4195: EXAMINATON OF RELOCATED FUEL DEBRIS AD.

Charactenzation And Quantificatort JACENT TO THE LOWER HEAD OF THE TMI-2 REACTOR VESSEL Modele And Resulte Detehene Nuclear Power Plant NUREG/CR 4116 V08: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/GR-0008' VALIDATION OF SEISMIC PROBABILISTIC RISK AS-HANDSON INTEGRATED RELIABILITY EVALUATONS (SAPHIRE)

SESSMENTS OF NUCLEAR POWER PLANTS.

VERSON 5.0 Mudele And Results Database (MAR 0) Reference Manual.

Nuclear Reguietory Reeeerch Isodular Logic NUREG-1266 V08: NRC SAFETY RESEARCH IN SUPPORT OF REGU-LATON. FY 1993.

NUREG/CR-4838: MICROCOMPUTER APPLCATIONS OF. AND MODI-FICATIONS TO, THE MODULAR FAULT TREES.

Nuclear Safety MoNeo Core NUREG/CP-0136: PROCEEDINGS OF THE DIGITAL SYSTEMS RELI-NUREG/CR-6133: FRAGMENTATON AND QUENCH BEHAVOR OF ABluTY AND NUCLEAR SAFETY WORKSHOP. September 13-14 CORIUM MELT STREAMS IN WATER.

1993.Rockville Crowne Plaza Hotel,Rockville. Maryland.

NUREG/CP-0139: TRANSACTIONS OF THE TWENTY-SECOND WATER Hortemy Riek REACTOR SAFETY INFORMATION MEETING.

NUREG/GR-0011: INFORMATON BIAS AND LIFETIME MORTALITY RISKS OF RADIATION-INDUCED CANCER. Low LET Radiaton.

Nuclear Weste Repomory NUREG/CR4288: GEOCHEMICAL INVESTIGATIONS RELATED TO Motor Operated Velve THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NUCLE.

NUREG/CR4205 VALVE ACTUATOR MOTOR DEGRADATION.

AR WASTE REPOSITORY.

esonor-Operated Vesve Nunwrtcal Technique NUREG/CP 0137 V01: PROCEEDINGS OF THE TH:RD NRC/ASME NUREG/CR-5535 V06: RELAPS/ MOD 3 CODE MANUALValidaten Of SYMPOSIUM ON VALVE AND PUMP TESTING. Held At The Hyatt Re.

Numencal Techniques in RELAP5/ MOO 3.

Hotel. Washington.DC, July 18-21,1994. Session 1 A - Sesson OFF9W NUREG/CP4137 V07 PROCEEDINGS OF THE THIRD NRC/ASME NUREG/CR4182 V01: OFFSCALE: A PC INPUT PROCESSOR FOR SYMPOSIUM ON VALVE AND PUMP TESTING. Held At The Hyatt Re-THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For Hotel, Washington.DC. July 18-21,1994. Session 3A -Sesson NUREG 68 2

FSCALE: A PC INPUT PROCESSOR FOR THE SCALE CODE SYSTEM. Volume ? The ORIGNATE Processor for heultipient Action leeue ORIGEN-S.

NUREG-1435 S03: STATUS OF SAFETY ISSUES AT LICENSED POWER PLANTS.TMI Acton Plan Requirements Unresolved Safety ORIGEN-S Issues.Genenc Safety issues.Other Multsplant Acton issues.

NUREG/CR4182 V02: OFFSCALE: A PC INPUT PROCESSOR FOR THE SCALE CODE SYSTEM. Volume 7. The ORIGNATE Processos for NRC's Program ORIGEN-S.

NUREG 1499-REASSESSMENT OF THE NRC'S PROGRAM FOR PRO-TECTING ALLEGERS AGAINST RETALIATION.

Occupational Esposure NUREG/CR-5600 V01: THE IMPACT OF ENVIRONMENTAL CONDI-Natural Analogue TIONS ON HUMAN PERFORMANCE. A Handbook Of Enwonmental NUREG/CR-6221: THE VALLES NATURAL ANALOGUE PRCklECT.

Exposures.

l i

l

92 Subject index NUREG/CR-5680 V02 THE IMPACT OF ENVIRONMENTAL CONDl-Pesolve Safety System TIONS ON HUMAN PERFORMANCE. A Cntical Review Of The Liters-NUREG/CR-6223: REVIEW OF THE PROPOSED MATERIALS OF CON.

ture.

STRUCTON FOR THE SBWR AND AP600 ADVANCED REACTORS.

NUREG/CR-6112 DAF FC IMPACT OF REDUCED DOSE LIMITS ON NRC LCENSED ACTIVITIES. Major issues in The implementation Of Poor Review ICRP/NCRP Dose Lmt Recommendations. Draft Report For Comment.

NUREG/CR4183: PEER REVIEW OF THE TMi-2 VESSEL INVESTIGA-TlON PROJECT METALLURGICAL EXAMINATIONS.

NUREG/CR4212: VALUE OF PUBLIC HEALTH AND SAFETY ACTONS Performance Aseeemment AND RADIATION DOSE AVOlDED.

NUREG/CR 5965: MODELING FIELD SCALE UNSATURATED FLOW AND TRANSPORT PROCESSES.

W Of N Ww hW NUREG-1415 V06 NO2-OFFICE OF THE INSPECTOR Performance History GENERALSermannual Recort. October 1,1993 March 31.1994 NUREG-1214 R13: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

NUREG 1415 VO7 N01: OFFCE OF THE INSPECTOR IC ASSESSMENT OF LCENSEE PERFORMANCE.

GENERALSermannual Report.Apnl 1 -September 30,1994.

Performance Testing Operating Empertence NUREG/CR-5994: EMERGENCY DIESEL GENERATOR: MAINTENANCE NUREG 1272 V08 NOI: OFFICE FOR ANALYSIS AND EVALUATION OF

^" F U

TY TE OPERATONAL DATA.1993 Annual Report - Power Reactors.

NUR G,A 6

9-Operating Empertence Feedback Report in9 Of Relays And Circuit Breakers.

NUREG-1275 V10: OPERATING EXPERIENCE FEEDBACK REPORT.

Petit 6one For Rulerneking RELIABILITY OF SAFETY RELATED STEAM TURDINE DRIVEN NUREG-0936 V12 N04: NRC REGULATORY AGENDA.Ouarterly STANDBY PUMPS. Commencal Power Reactors.

Report October-December 1993.

NUREG-0936 V13 Not: NRC REGULATORY AGENDA.Ouarterty Operettonal Event NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE Report. January, March 1994.

NUREG-0936 V13 NO2-NRC REGULATORY AGENDA.Ouarterly DAMAGE ACCOENTS: 1992 A STATUS REPORT. Main Report And Appener A.

Report,Apnl-June 1994.

NUREG/CR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCOENTS: 1992 A STATUS REPORT.Apperdces B. C, D, Phyelcol FMnees training NUREG 1504: REVIEW CRITERIA FOR THE PHYSCAL FITNESS E, F. And G.

NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE CORE TRAINING REQUIREMENTS IN 10 CFR PART 73.

DAMAGE ACCIDENTS: 1993 A STATUS REPORT. Main Report And Append'cas A-D.

Physical Protection NUREG/CR-4674 V20: PRECURSORS TO POTENTIAL SEVERE CORE NUREG-1497: INTERIM LICENSING CRITERIA FOR PHYSICAL PRO-DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Appendences E TECTION OF CERTAIN STORAGE OF SPENT FUEL NUREG/CR4149: APPLCATIONS OF FIBER OPTICS IN PHYSICAL And F.

PROTECTON.

Operat6onal Expertence NUREG/CR4093. AN ANALYSIS OF OPERATONAL EXPERIENCE Pipe DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD.

NUREG/CR-4599 V03 N2 SHORT CRACKS IN PIPtNG AND PIPING DRESSING HUMAN RELIABluTY ASSESSMENT ISSUES.

WELDS.Sermannual Report. October 1992. March 1993.

NUREG/CR-5128 Rot: EVALUATION AND REFINEMENT OF LEAK-Operator Licensing RATE ESTIMATION MODELS.

NUREG-1021 R07 S01: OPERATOR LCENSING EXAMlNER STAND-NUREG/CR4128: P1 PING BENCHMARK PROBLEMS FOR THE ABB/

ARDS.

CE SYSTEM 80+ STANDARDl2ED PLANT.

Operator Performance Pipe Strees NUREG/CR4127: THE EFFECTS OF STRESS ON NUCLEAR POWER NUREG/CR4236: SEISMIC INVESTIGATONS OF THE HDR SAFETY PLANT OPERATIONAL DECISION MAKING AND TRAINING AP-PROGRAM Summary Report. '

PROACHES TO REDUCE STRESS EFFECTS.

NUREG/CR4208: AN EMPIRCAL INVESTIGATION OF OPERATOR Piping PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER.

NUREG 1426 V02: COMPILATION OF REPORTS FROM RESEARCH GENCIES.

SUPPORTED BY THE MATERIALS E NGINEERING DRANCH.DIVISON OF ENGINEERING.1991-1993.

Orgentastion Chart NUREG/CR-5935:

SUMMARY

OF WORK COMPLE'ED UNDER THE NUREG0325 R17: U.S. NUCLEAR REGULATORY COMMISSION OR*

ENVIRONMENTAL AND DYNAMC EQUIPMENT OlALIFICATION Nt-GANIZATION CHARTS AND FUNCTIONAL STATEMENTS. October SEARCH PROGRAM (EDOP).

3.1994.

NUREG/CR4121: COMPONENT EVALUATON FOR INTERSYSTEM NUREG 1471: CONCEPT OF OPERATIONS WITH ORGANIZATON Loss.op. COOLANT ACCOENTS IN ADVANCED LIGHT WATER RE-CHARTS.NRC incident Response-ACTORS.

NUREG/CR4181: A PILOT APPLCATON OF RISK. BASED METHOOS PRA TO ESTABLISH INSERVCE INSPECTION PRORITIES FOR NUCLE.

NUREG/CR-4838: MCROCOMPUTER APPLICATONS OF. AND MODl' AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION, FICATIONS TO. THE MODULAR FAULT TREES.

NUREG/CR4281: A SIMPLlFIED LEAK BEFORE-BREAK EVALUATON PROCEDURES FOR AUSTENITIC AND FERRITIC STEEL PtPING.

PRISM NUREG-1368: PREAPPLICATION SAFETY EVALUATON REPORT FOR T

POWER RE OR I NOVATIVE SMALL MODULE (PRISM)

URE 5359: REVIEW OF ELASTIC STRESS AND FATIGUE 40 RW FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA TION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER PlPING Pwn NUREG/CR4075: THE PROSANTY OF CONTAINMENT FAILURE BY SYSTEMS.

NUREG/CR4151: FEASIBILITY OF DEVELOPlNG RISK BASED RANK.

DIRECT CONTAINMENT HEATING IN ZION NUREG/CR4075 S01: THE PROBABluTY OF CONTAINMENT FAIL.

INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVICE IN-UAE BY DIRECT CONTAWMENT HEAMiG IN ZION.

SPECTON.

NUREG/CR-8092: R;SK ASSESSMthi FOR THE INTENTIONAL DE-PRESSURIZATON STRATEGY IN PWAS_

Plant Organisation NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC.

NUREG/CR4122 STAFFING DECISON PROCESSES AND TOR CONTROL ROO DRIVE MECHANISM NOZZLE CRACKING.

ISSUES. Case Studies Of Seven U.S. Nuclear Power Plants.

Peleceolemiemy Power Reactor NUREG/CR-6258: THE LIOUEFACTION METHOD FOR ASSESSING PA-NUREG/CR 4816 R02: PR-EDB: POWER REACTOR EMBRtTTLE* LENT LEOSEISMCITY.

DATA BASE.VERSON 2. Program Desenphon

1 l

Subject index 93 Power Surges VERSION 5.0. Systems Analyses And Risk Assessment (SARA) Refer.

NUREG/CR-5941: TECHNICAL BASIS FOR EVALUATING ELECTRO-ence Manual.

MAGNETC AND RADIO-FREQUENCY INTERFERENCE IN SAFETY-NUREG/CR4116 V05: SYSTEMS ANALYSIS PROGRAMS FOR RELATED i&C SYSTEMS.

HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

Preapplication Safety Evaluation Report VERSION 5 0. Systems Analysis And Risk Assessment (SARA) Tutonal Manual.

NUREG-1368: PREAPPUCATION SAFETY EVALUATON REPORT FOR NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR THE POWER REACTOR INNOVATIVE SMALL MODULE (PRISM)

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

UQUID-METAL REACTOR. Final Report VERSION 5.0. Fault Tree, Event Tree, And Piping & Instrumentation Pressure Boundry System NUREG/CR4151: FEASIBILITY OF DEVELOPING RISK-BASED RANK.

NU

/C 4116 V08 S M YSIS PROGRAMS FOR INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVCE IN-HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

SPECTION.

VERSION 5.0.Models And Results Delabase (MAR.0) Reference NUREG/CR4181: A PILOT APPUCATION OF RISK-BASED METHODS TO ESTABUSH INSERVICE INSPECTION PRIORITIES FOR NUCLE-NURE CR4145: VERIFCATION AND VAUDATION OF THE SAPHIRE VERSON 4.0 PHA SOFTWARE PACKAGE AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION.

NUREG/CR4166: RISK IMPACT OF TECHN' CAL SPECIFCATIONS RE.

Presoure Veeeel OUIREMENTS DURING SHUTDOWN FOR BWRS.

NUREG/CR-5591 V02 N1: HEAVY-SECTON STEEL IRRADIATION NUREG/CR4181: A PILOT APPUCATON OF RISK-BASED METHODS TO ESTABUSH INSERVICE INSPECTION PRIORITIES FOR NUCLE-PRO, GRAM.Somannual Progress Report For October 1990 - March 999 AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION.

NUREG/CR-5591 V02 N2: HEAVY-SECTION STEEL IRRADIATION NUREG/GR-0008: VAUDATON OF SEISMC PROBABluSTIC RISK AS-PROGRAM.Semannual Progress Report For Apni-September 1991.

SESSMENTS OF NUCLEAR POWER PLANTS.

Presourtrod Thermal Shock Program Performance NUREG-1511: REACTOR PRESSURE VESSEL STATUS REPORT.

NUREG/CR-5758 V04: FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTRY. Annual Summary Of Program Performance Reports CY Pressurtaed Water Reector 1993.

NUREG/CR4075: THE PROBABluTY OF CONTAINMENT FAILURE BY DIRECT CONTAINMENT HEATING IN ZION.

Public Health NUREG/CR4075 S01: THE PROBABluTY OF CONTAINMENT FAIL-NUREG/CR4212: VALUE OF PUBUC HEALTH AND SAFETY ACTIONS URE BY DIRECT CONTAINMENT HEATING IN ZION.

AND RADIATION DOSE AVOIDED.

NUREG/CR4092 RISK ASSESSMENT FOR THE INTENTIONAL DE-PRESSURIZATON STRATEGY IN PWRS.

Pump Testing NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC-NUREG/CP-0137 V02 PROCEEONGS OF THE THIRD NRC/ASME TOR CONTROL ROD DRIVE MECHANISM NCZZLE CRACKING.

SYMPOSlUM ON VALVE AND PUMP TESTING. Heed At The Hyatt Re.

4 Thal SMck gency Hotel, Wasivngton,0C. July 18-21,1994 Sessen 3A -Session NUREG/CR-4219 V10 N1: HEAVY SECTON STEEL TECHNOLOGY d O' PROGRAM.Sermannual Progress Report For October 1992 March N

Quahty Assurance NUREG 1475: APPLYING STATISTCS.

ProbabWetic Fw NUREG-1495: OVERALL REVIEW STRATEGY FOR THE NUCLEAR NUREG/CR4075: THE PROBABluTY OF CONTA!NMENT FA! LURE BY REGULATORY COMMISSON'S HIGH-LEVEL WASTE REPOSITORY NUR CR 5

1 H A UTY OF CONTAINMENT FAIL.

NU E / 0'135: WORKSHOP ON ENVIRONMENTAL QUAUFICATION URE BY DIRECT CONTAINMENT HEATING IN ZION.

OF ELECTRIC EQUIPMENT.

Probabsetic Riek Aseeeement QueNty "./ W T.t NUREG-1489 A REVIEW OF NRC STAFF USES OF PROBABILISTC NUREG/CR4276: OUAUTY MANAGEMENT IN REMOTE AFTERLOAD-RISK ASSESSMENT.

ING BRACHYTHERAPY, NUREG/CP4133 V01: PROCEEDNGS OF THE TWENTY FIRST WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-RASCAL Version 2 sion; Advanced Reactor Research; Advanced Control System Technoi NUREG/CR-5247 V01 R2: RASCAL VERSION 2.1 USER'S GUIDE.

Advanced instrumentaten & Control Hardware; Human Factors....

NUREG/CR-5247 V02 R2: RASCAL VERSON 2.1 WORKBOOK.

NU G/CR-4551 V01 R1: EVALUATION OF SEVERE ACCIDENT RISKS: METHODOLOGY FOR THE COffTAINMENT, SOURCE RCIC Pump TERM, CONSEQUENCE, AND RISK INTEGRATION ANALYSES.

NUREG-1275 V10: OPERATING EXPERIENCE FEEDBACK REPORT -

NUREG/CR-4639 V5R4P2: NUCLEAR COMPUTER! ZED UBRARY FOR REUABluTY OF SAFETY-RELATED STEAM TURBINE DRIVEN ASSESSING REACTOR RELIABIUTY (NUCLARR). Volume 5 Data STANDBY PUMPS. Commencal Power Reactors.

Manual.Part 2 Human Error Pr Deta.

NUREG/CR-4639 V5R4P3: NUCLEAR ERIZED LIBRARY FOR RCIC System ASSESSING REACTOR REUABlWTY (NUCLARR). Volume 5. Data NUREG/CR4087: THE EFFECTS OF AGING ON BOIUNG WATER RE-Manual. Pert 3. Hardware Failure Data.

ACTOR CORE ISOLATION COOLING SYSTEMS.

NUREG/CR-5407: ASSESSME T OF THE IMPACT OF DEGRADED EA ST NESSES ON SEISMC PLANT RISK AND SEIS-RE 3

NUREG/CR-5726: REYlEW OF THE DABLO CANYON PROBABILISTC Numencal Techniques in RELAP5/ MOD 3.

RISK ASSESSMENT.

NUREG/CR-5535 V07: RELAPS/ MOD 3 CODE MANUALSummanes And NUREG/CR4092: RISK ASSESSMENT FOR THE INTENTIONAL DE.

Reviews Of independent Code Assessment Reports.

PRESSURIZATION STRATEGY IN PWRS NUREG/lA 0093: RELAP5/ MOD 3 ASSESSMENT FOR CALCULATION NUREG/CR4104. SHEAR WALL ULTIMATE DRIFT LIMITS.

OF SAFETY AND REUEF VALVE DISCHARGE PIPING HYDRODY-NUFIEG/CR4116 vot: SYSTEMS ANALYSIS PROGRAMS FOR NAMIC LOADS.

HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

NUREG/1/ ?114: ASSESSMENT OF RELAP5/ MOD 3 WITH THE LOFT VERSION 5 0. Technical Reference Manual.

L9-1/ Liv JPERIMENT SIMULATING AN ANTICIPATED TRANSIENT NUREG/CR4116 V02: SYSTEMS ANALYSIS PROGRAMS FOR WITH MULTIPLE FAILURES.

HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

VERSON 5 0. Integrated Reliabrhty And Risk Analysie System (IRRAS)

Radiation Does Reference Manual.

NUREG/CR4212-VALUE OF PUBUC HEALTH AND SAFETY ACTIONS NUREG/CR4116 V03: SYSTEMS ANALYSIS PROGRAMS FOR AND RADIATION DOSE AVOCED.

HANDSON INTEGRATED RELIABluTY EVALUATONS (SAPOIRE)

VERSION 5 0. Integrated Reliat>hty And Risk Analysis System (IRTAS)

Radletion Effect Tutonal Manual NUREG/CR4053: COMPARISON OF MACCS USERS CALCULATIONS

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NUREG/CR4116 V04: SYSTEMS ANALYSIS PROGRAMS FOR FOR THE INTERNATONAL COMPARISON EXERCISE ON PROBABS-HANDS-ON INTEGRATED RELIABlWTY EVALUATONS (SAPHIRE)

USTIC ACCOENT CONSEQUENCE ASSESSMENT CODES.

m

94 Subject index Radiation Embrtttlement RIA FOR DECOMMISSONING OF NRC-LOENSED NUCLEAR NUREG 1426 V02: COMPILATION OF REPORTS FROM RESEARCH FACILITIES Appendices Drac Report For Comment.

SUPPORTED BY THE MATERIALS ENGINEERING NUREG/CR4156:

SUMMARY

OF COMMENTS RECElVED FROM BRANCH, DIVISION OF ENGINEERING 1991 1993.

WORKSHOPS ON RADIOLOGIGAL CRITERIA FOR DECOMMISSON-NUREG IS11: REACTOR PRESSURE VESSEL STATUS REPORT.

ING.

NUREG/CR 6250:

SUMMARY

OF COMMENTS RECEfVED ON STAFF

/

V1 DFC REVISED ANALYSES OF DECOMMISSION-g' ING FOR THE REFERENCE BOLLING WATER REACTOR POWER STATION Effects Of Current Regulatory And Other Considerations On Redonuclide The Financial Assurance Requirements Of The Decommissioning Rule NUREG/CR4063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS NUR /CR4174 V2 DFC: REVISED ANALYSES OF DECOMMISSON-ING FOR THE REFERENCE BOILING WATER REACTOR POWER Redonuclide Behavtor STATION Effects Of Current Regulatory And Other Considerations On NUREG/CR4201: COMPRESSION AND IMMERSON TESTS AND The Fmancial Assurance Requwements Of The Cm... m Rule LEACHING OF RADIONUCLlDES. STABLE METALS. AND CHELATING And-AGENTS FROM CEMENT-SOLOirlED DECONTAMINATION WASTE Radiation Hazard COLLECTED FROM NUCLEAR POWER STATONS.

NUREG/CR4053: COMPARISON OF MACCS USERS CALCULATIONS FOR THE INTERNATONAL COMPARISON EXERCISE ON PROBABl.

NUREG/CR4187: RESULTS OF MECHANICAL TESTS AND SUPPLE-LISTC ACCIDENT CONSEQUENCE ASSESSMENT CODES-MENTARY MICROSTRUCTURAL EXAMINATIONS OF THE TMI 2 Radiation Monitortng LOWER HEAD SAMPLES.

NUREG/CR4112 DRF FC-IMPACT OF REDUCED DOSE LIMITS ON NRC LICENSED ACTIVITIES. Ma}or issues in The implementation Of Reactor Accident ICRP/NCRP Dose Last Recommendations. Draft Report For Comment.

NUREG/CR.5247 V01 R2: RASCAL VERSION 2.1 USER'S GUIDE.

NUREG/CR-5247 V02 R2: RASCAL VERSON 2.1 WORKBOOK.

Radiation Protection NUREG/CR4044: EXPERIMENTS TO INVESTIGATE DRECT CON-NUREG/CR-4409 V05: DATA BASE ON DOSE REDUCTON RE-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

ZON NUCLEAR POWER PLANT IN THE SURTSEY TEST FACILITY.

NUREG/CR-5569 R01: HEALTH PHYSICS POSITIONS DATA BASE.

NUREG/CR4094: CALCULATONS IN SUPPORT OF A POTENTIAL NUREG/CR4204. QUESTIONS AND ANSWERS BASED ON REVISED DEFINITION OF LARGE RELEASE.

10 CFR PART 20-NUREG/CR4144 V02 PIA: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4212 VALUE OF PUBLC HEALTH AND SAFETY ACTONS CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT AND RADATON DOSE Av00ED.

SURRY. UNIT 1 Analysis Of Core Damage Frequency From internal Radiation Therapy Events Dunng Mid-loop Operations. Main Report (Chapters 14).

NUREG/CR4088:

SUMMARY

OF 1991 1992 MISADMINISTRATON EVENT INVESTIGATIONS' RmW Component NUREG/CR-5812-MANAGING AGING IN NUCLEAR POWER Radiattorwanduced Cancer PLANTS. Insights From NRC Mantenance Team inspection Reports.

NUREG/GR4011: INFORMATON BIAS AND LIFETIME MORTALITY NUREG/CR.5967: DEVELOPMENT AND APPLICATION OF DEGRADA-RISKS OF RADIATION-INDUCED CANCER Low LET Radiatiort ilON MODELING TO DEFINE MAINTENANCE PRACTICES.

Radso Frequency Interference Reactor Control NUREG/CR4941: TECHNOAL BASIS FOR EVALUATING ELECTRO-NUREG/CR.6103. PRIORITIZATON OF REACTOR CONTROL COMPO-MAGNETIC AND RADIO. FREQUENCY INTERFERENCE IN SAFETY.

NENTS SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEQUENCE OF RELATED IAC SYSTEMS.

AGING.

Rad 6oactive Material Reactor Control System NUREG.0383 V01 R17: DRECTORY OF CERTIFICATES OF COMPLi-NUREG/CR4146: LOCAL CONTROL STATIONS: HUMAN ENGINEER-ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC ING ISSUES AND INSIGHTS.

I i

NU G 038 17: DIRECTORY OF CERTIFICATES OF COMPLi-Reactor Coolant Systern ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certrhcates Of NUREG/CR-5535 V06. RELAP5/ MOO 3 CODE MANUALVahdation Of e-Numencal Techniques in RELAP5/ MOD 3 NURE V03 R14: DIRECTORY OF CERTIFICATES OF COMPL1-NUREG/CR4193 PRIMARY SYSTEM FISSION PRODUCT RELEASE ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC AND TRANSPORT.A State 4f The-Art Report To The Comrmttee On Approved Oumhty Assurance Programs For Radioactsve Matenals Pack-The Safety Of Nuclear installabons.

NkEG.1492 DFC REGULATORY ANALYSIS ON CRITERIA FOR THE Reactor instrumentation RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE NUREG/CR4105: HUMAN FACTORS ENGINEERING GUIDANCE FOR MATERIALDraft Report For Comment.

THE REVIEW OF ADVANCED ALARM SYSTEMS.

NUREG/CR-2907 V12: RAPOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Anrwal Report 1991.

Reactor Material interaction NUREG/CR4289. RECONCENTRATION OF RADOACTIVE MATERIAL NUREG/CR4198: TMI-2 INSTRUMENT NOZZLE EXAMINATONS PER-RELEASED TO SANITARY SEWERS IN ACCORDANCE WTTH 10 CFR FORMED AT THE INEL PART 20.

RadioactM Releau Reactor Operator NUREG/CR.2850 V12: DOSE COMMITMENTS DUE TO RADOACTIVE NUREG/CR.5908 V01: ADVANCED HUMAN-SYSTEM INTERFACE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1990.

DESIGN REVIEW GUIDEUNE. General Evaluation Model. TecMcal Ca.e.nl And Gu6deline Descnption.

Radlographic NUREG/CR 5908 V02: ADVANCED HUMAN-SYSTEM INTERFACE NUREG/CR-4833: LARGE AREA SELF-POWERED GAMMA RAY DESIGN REVIEW GUIDEUNE. Evaluation Proce&res And Guidelines DETECTOR. Phase 11 Development Of A Source Posation Monitor For For Human Factors Engmeermg Reviews.

Use On Irw.justnal Radiographic Urvts.

NUREG/CR4122: STAFFING DECISION PROCESSES AND ISSUES Case Studies Of Seven U S. Nuclear Preer Plants.

Radlolog6 cal Criterta NUREG/CR4123. AN INTERNATIONAL COMPARISON OF COMMER-(

NUREG 1496 V1 DFC: GENERC ENVIRONMENTAL IMPACT STATE-CIAL NUCLEAR POWER PLANT STAFFING REGULATIONS AND i

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE.

PRACTICE.19801990.

l RIA FOR DECOMMISSONING OF NRC LICENSED NUCLEAR F ACILITIES Mac Report Draft Report For Comment.

Reactor Operator Licensing NUREG-1496 V2 DFC: GENERC ENVIRONMENTAL IMPACT STATE-NUREG 1478: NON POWER REACTOR OPERATOR LICENSING EXAM.

MENT IN SUPPORT OF RULEMAKING ON RADOLOGICAL CRITE-INER STANDARDS.

l l

l

m i

j Subject Index 95 4

Reactor Pressure Vessel NUREG-0936 V13 NO2: NRC REGULATORY AGENDA.Ouarterly 1

NUREG-1426 V02-COMPILATION OF REPORTS FROM RESEARCH Report, April-June 1994.

SUPPORTED BY THE MATERIALS ENGINEERING 4

BRANCH. DIVISION OF ENGINEERING 1991 1903.

Regulatory And Technical Report NUREG 1511: REACTOR PRESSURE VESSEL STATUS REPORT.

NUREG-0304 V18 N04: REGULATORY AND TECHNICAL REPORTS NUREG/CR4231: A COMPARISON OF THE RELATIVE IMPORTANCE OF COPPER PRECIPITATES AND POINT DEFECT CLUSTERS IN RE-(ABSTRACT INDEX JOURNAL). Annual Compdaten For 1993.

NUREG-0304 Vig N01: REGULATORY AND TECHNICAL REPORTS j

NUR G CR I RA TE A ROPERTIES OF MID-I^

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• E * "

g l

NU

/

21 EAVAGE BEHAVIORS IN NUCLEAR VESSEL NUREG-0304 V19 NO2: REGULATORY AND TECHNICAL REPORTS STEELS.

(ABSTRACT INDEX JOURNAL). Compdaten For Second Ouarter 1994,Aprildune.

Reactor Protection system NUREG/CR4303: METHOD FOR PERFORMING DIVERSITY AND DE.

Regulatory Prostucte l;

FENSE-IN-DEPTH ANALYSES OF REACTOR PROTECTION SYS-NUREG-1266 V08: NRC SAFETY RESEARCH IN SUPPORT OF REGU-TEMS.

LATION - FY 1993.

Reactor ReilatWitty Regulatory statWilsation NUREG/CR-4639 V5R4P2: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG-1242 V03 PT01: NRC REVIEW OF ELECTRIC POWER RE-l ASSESSING REACTOR RELIABILITY (NUCLARR) Volume 5: Data SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILl-Manual Part 2: Human Error HEFLDERIZED LIBRARY FORTY REOUIREMENTS DOCUMENTS.Passrve Plant DesegnsChapter eta.

1 NUREG/CR-4639 V5R4P3: NUCLEAR Pui 1.Pr@t Numhr 669.

~,

ASSESSING REACTOR FIELIABILITY (NUCLARR). Volume 5: Data NUREG-1242 V03 PT02: NRC REVIEW OF ELECTRIC POWER RE-l Manual.Part 3: Hardware Component Fadure DataL SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILI-g

,g, TY REQUIREMENTS DOCUMENT.Passeve Plant Desagns. Chapters 2-j NUREG/CP-0133 V01: PROCEEDINGS OF THE TWENTY-FIRST 135t Nurnber 661 WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-Relay son; Advanced Reactor Research; Advanced Control System Technol-4 Advanced Instrumentation & Control Hardware: Human Factors....

NU. REG /CH4169: RELAY TEST PROGRAM Senes il Tests integral Test-NU G/CP-0133 V02-PROCEEDINGS OF THE TWENTY-FIRST ing Of Relays And Cucult Breakers.

WATER REACTOR SAFETY INFORMATION MEETING. Severe Acce dont Research.

Crkr6e NUREG/CP 0133 V03: PROCEEDINGS OF THE TWENTY-F!RST NUREG-1500: WORKING DRAFT REGULATORY GUIDE'ON RELEASE WATER REACTOR SAFETY INFORMATION MEETING Pnmary CRITERIA FOR DECOMMISSIONING: NRC STAFF'S DRAFT FOR System integnty; Agmg Research, Products & Apphcations; Structural &

COMMENT.

NO C

NN S OF TWENTY SECOND WATER RollatWiny REACTOR SAFETY INFORMATION MEETING.

NUREG/CR4278: SURVEY OF INDUSTRY METHODS FOR PRODUC-NUREG/CR-4551 V01 R1: EVALUATION OF SEVERE ACCIDENT ING HIGHLY RELIABLE SOFTWARE.

RISKS: METHODOLOGY FOR THE CONTAINMENT. SOURCE TERM. CONSEQUENCE AND RISK INTEGRATION ANALYSES.

Rei6alWilty Analyste i

NUREG/CR-5994. EMERdENCY DIESEL GENERATOR. MAINTENANCE NUREG/CP 0138: PROCEEDINGS OF WORKSHOP 1 IN ADVANCED j

AND FAILURE UNAVAILABILITY, AND THElR RISK IMPACTS.

TOPICS IN RISK AND RELIABILITY ANALYSIS Model Uncertamty: Its j

NUREG/CR4042: PERSPECTIVES ON REACTOR SAFETY.

Characternaton And Quantification.

NUREG/CR4213: HIGH-TEMPERATURE HYDROGEN. AIR-STEAM DETONATION EXPERIMENTS IN THE BNL SMALL SCALE DEVELOP-Revnote AM. ".,

MENT APPARATUS.

NUREG/CR4276: OUALITY MANAGEMENT IN REMOTE AFTERLOAD-Reeckr Shutdown ING BRACHYTHE.

'Y.

NUREG/CR-5344 RO1: REPLACEMENT ENERGY COST ANALYSIS m

PACKAGE (RECAP): USER'S GUlOE.

NUREG/CR 5344 R0 REPLACEMENT ENERGY COST ANALYSIS Reconcentration PACKAGE (RECAP): USER'S GUIDE.

1 NUREG/CR4289 RECONCENTRATION OF RADIOACTIVE MATERIAL RELEASED TO SANITARY SEWFRS IN ACCORDANCE WITH 10 CFR Report To Covees PART 20.

NUREG-0090 V16 NO3: REPORT TO CONGRESS ON ABNORMAL OCCURRENCESJuly-September 1993.

j Recordheeping M_' _....

NUREG 0090 V16 N04: REPORT TO CONGRESS ON ABNORMAL NUREG 1460 R01: GUIDE TO NRC REPORTING AND RECOROKEEP-OCCURRENCES October-December 1993.

ING REQUIREMENTS. Compded From Requrements in Title 10 Of NUREG4090 V17 N01: REPORT TO CONGRESS ON ABNORMAL j

The U S. Code Of Federal Regulatons As Codified On December 31, OCCURRENCES. January March 1994.

i 1993.

NUREG-0090 V17 NO2: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. April-June 1994.

NUREG/CR4112 DAF FC IMPACT OF REDUCED DOSE UMITS ON Reporting i m_

NRC UCENSED ACTIVITIES. Ma#ar issues In The implementaten Of NUREG 1460 R01: GUIDE TO NRC REPORTING AND RECOROKEEP.

j ICRP/NCAP Dose Lunet Recommendations Draft Report For Comment-ING REQUIREMENTS. Compeled From Requrements in Title 10 Of Refueling Outage The U.S. Code Of Federal Regulatons As Codified On December 31, 3993' NUREG/CR-6143 V04: EVALUATION OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Repoeltory Operet6ona Cetteria j

GRAND GULF.UNtT 1.Anatysis Of Core Damage Frequency From in.

NUREG/CR 5919-REPOSITORY OPERATIONAL CRITERIA COMPARA-ternally Induced Floodmg Events For Plant Operational State 5 Dunng TIVE ANALYSIS.

1 a Refuelmg -

Residuel Radioactivity 4

G 4209:

MEMPHIS AREA REGIONAL SEISMIC NUREG-1501 DRFT: BACKGROUND AS A RESIDUAL RADIOACTIVITY q

NETWORK Final Report, October 1966 - September 1992.

CRITERION FOR DECOMMISSIONING.Appenen A To The Draft Go-neric Envronmental impact Statement In Support Of Rulemaking On Regulatory Agenda Radiological Cntena For Deconmssonino Of NRC-i 1

NOREG-0936 V12 N04. NRC REGULATORY AGENDA.Ouarteriy Re. port. October-December 1993 Retallation NUREG-0936 V13 NOI: NRC REGULATORY AGENDA.Ouarterly NUREG-1499. REASSESSMENT OF THE NRC'S PROGRAM FOR PRO-Report. January-March 1994.

TECTING ALLEGERS AGAINST RETAl.lATION I

i i

i

96 Subject index Review Criteria SARA NUREG 1504: REVIEW CRITERIA FOR THE PHYSICAL FITNESS NUREG/CR-6116 V04. SYSTEMS ANALYSIS PROGRAMS FOR TRAINING REQUIREMENTS IN 10 CFR PART 73.

HANDS-ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

VERSION 5.0. Systems Analysis And Risk Assessment (SARA) Refer.

Rtsk Analyste once Manual.

NUREGICR-4874 V19-PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-6116 V05: SYSTEMS ANALYSIS PROGRAMS FOR DAMAGE ACCIDENTS: 1993 A STATUS REPORT. Main Report And HANDS-ON INTEGRATED REUABILITY EVALUATONS (SAPHIRE)

Appendices A-D.

VERSION 5.0. Systems Analysis And Risk Assessment (SARA) Tutor 6al NUREG/CR-4674 V20 PRECURSORS TO POTENTIAL SEVERE CORE Manual DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Appendences E And F.

SAS2H/ORIGEN-S Data NUREG/CR 5625. TECHNICAL SUPPORT FOR A PROPOSED DECAY Risk Aseeeement HEAT GUOE USING SAS2H/ORIGEN-S DATA.

NUREG/CR 6094: CALCULATIONS IN SUPPORT OF A POTENTIAL DEFINITON OF LARGE RELEASE.

SCALE NUREG/CR4112 DRF FC: IMPACT OF REDUCED DOSE UMITS ON NUREG/CR-6102: VALOATION OF THE SCALE BROAD STRUCTURE NRC UCENSED ACTMTIES. Maior lasues in The implementahon Of 44-GROUP ENDF/B Y CROSS SECTION LIBRARY FOR USE IN ICRP/NCRP Dose Umst Recommendahons. Draft Report For Comment CRITICALITY SAFETY ANALYSES.

Riek Aseeeement Method SCALE Code System NUREG/CR4157: SURVEY AND EVALUATION OF AGING RISK AS-NUREG/CR4162 V01: OFFSCALE. A PC INPUT PROCESSCR FOR SESSMENT METHODS AND APPUCATONS.

THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For The Cnticality Sequences.

Rock Joint NUREG/CR4182 V02: OFFSCALE: A PC INPUT PROCESSOR FOR NUREG/CR4176. LABORATORY CHARACTERl2ATON OF ROCK THE SCALE CODE SYSTEM. Volume 2 The ORIGNATE Processor for JOlNTS-ORIGEN-S.

NUREG/CR4216: EVALUATION OF ROCK JOINT MODELS AND COM-PUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS.

SCOAP/RELAPS NUREG/CR4160

SUMMARY

OF IMPORTANT RESULTS AND SCDAP/

Rhing RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-FP 2.

NUREG-1490 V1 DFC: GENERIC ENVIRONMENTAL IMPACT STATE.

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE.

Safeguards Summary Event Uet RlA FOR DECOMMISSIONING OF NRC-UCENSED NUCLEAR NUREG 0525 V02 R02-SAF EGUARDS

SUMMARY

EVENT US1 FACluTIES Mam Report. Draft Report For Comment (SSEL). January 1,1990 Tivough December 31,1993.

NUREG-1496 V2 DFC: GENERIC ENVIRONMENTAL IMPACT STATE.

MENT IN SUPPORT OF RULEMAKING ON RADOLOGICAL CRITE-Safety Evaluetlon Report RlA FOR "'.AJMMISSIURING OF NRC.UCENSED NUCLEAR NUREG-0647 S13: SAFETY EVALUATION REPORT RELATED TO THE FACluTIES Appendcas. Draft Repurt For Comment.

OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND NUREG 1501 DRFT: BACKGROUND AS A RESIDUAL RADIOACTIVITY

2. Docket Nos. 50-390 And 50 391.(Tennesee Valley Aumonty)

CRITERON FOR DECOMMISSON!NGAppendix A To The Draft Go-NUREG-0847 S14: SAFETY EVALUATON REPORT RELATED TO THE nonc Envvenmental impact Statement in Support Of Rulemaking On OPERATON OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND RaGological Cnlena For Decomrrusserwng Of NRC...

2 Docket Nos. 50-390 And 50-391.(Tennessee Valley Auttenty)

NUREG-1491: SAFETY EVALUAllON REPORT FOR THE CLAIBORNE Rules CNRICHMENT CENTER. HOMER. LOUISIANA. Docket No. 70 3070, Lou-NUPEG4936 V12 N04. NRC REGULATORY AGENDA.Ouarter4 isiana Energy SerMces,LP.

Report,0ctober-December 1993.

NUREG4936 V13 N01: NRC REGULATORY AGENDA Ouarterly Safety laeue Report, January. March 1994.

NUREG 1435 S03. STATUS OF SAFETY ISSUES AT UCENSED NUREG4936 V13 NO2: NRC REGULATORY AGENDA.Ouarterty POWER PLANTS.TM) Acton Plan Requrements. Unresolved Safety Report,Apni-June 1994.

Issues.Generte Safety issues.Other Multiplant Action lasues.

SAPHIRE Safety Resoerch NUREG/CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR NUREG 1266 V06: NRC SAFETY RESEARCH IN SUPPORT OF REGU-HANDSON INTEGRATED REUA81UTY EVALUATONS (SAPHiRE)

LATION - FY 1993.

VERSON 5 0.Techrucal Reference Manual.

NUREG/CR4116 V02: SYSTEMS ANALYS!S PROGRAMS FOR Safety Testing HANDS-ON INTEGRATED RELIABluTY EVALUATONS (SAPHIRE)

NUREG/CR4074 V01: SEALED SOURCE AND DEVICE DESIGN VERSION 5.0. Integrated Reliability And Risk Analysis System (IRRAS)

SAFETY TESTING.Techrucal Report On The Findinos Of Task 1.Octo-Reference Mammal.

bet 1991. Apnl 1993.

NUREG/CR4116 V03. SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHiRC)

Sampling Plan VERSON 5.0 Integrated Achatxiety And Resk Analysss System (IRRAS)

NUREG/CR-5161 V02. EVALUATON OF SAMPUNG PLANS FOR IN-Tutonal Manual.

SERVICE INSPECTON OF STEAM GENERATOR NUREG/CR4116 V04: SYS1 EMS ANALYSIS PROGRAMS FOR TUDES. Comprehensive Anatytical And Monte Carlo Swnulation Results HANDSOH INTEGRATED RELIABluTY EVALUATONS (SAPHlRE)

For Several Samphng Plans.

VERSON 5.0. Systems Analyses And Risk Assessment (SARA) Refer-ence Manual.

Sequeney Earthquake NUREG/CR4116 V05: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4258. THE UQUEFACTION METHOD FOR ASSESSING PA-HANDS.ON INTEGRATED RELIA 81UTY EVALUATONS (SAPHIRE)

LEOSEISMICITY.

VERSION 5 0 Systems Anaysis And Risk Assessment (SARA) Tutonal Manual.

Saturated Flow NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4114 V03: PERFORMANCE ASSESSMENT OF A HYPO-HANDS-ON INTEGRATED RELIADIUTY EVALUATIONS (SAPHIRE)

THETICAL LOW. LEVEL WASTE FACILITY. Groundwater Flow And VERSION 5.0. Fault Tree, Event Tree, And Pipm0 & Instn,mentaton Transport Swnulaten.

Diagram (FEP) Edtors Refesence Manual.

NUREG/CR4116 V06. SYSTEMS ANALYSIS PROGRAMS FOR Scaling HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

NUREG/CR4267: AIR-WATER SlMULATION OF PHENOMENA OF VERSION 5.0.Models And Results Database (MAR.0) Reference CORIUM DISPERSION IN DIRECT CONTAINMENT HEATING.

Manual.

Sealed Source SAPHIRE Version 4 NUREG/CR4074 V01: SEALED SOURCE AND DEVICE DESIGN NUREG/CR4145. VERIFICATION AND VALOATON OF THE SAPHIRE SAFETY TESTING Techrucal Report On The Findings Of Task 1.Octo-VERSON 4 0 PRA SOFTWARE PACKAGE.

ber 1991. April 1993.

1 Subject index 97 Sediment NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4232-ASSESSING THE ENVIRONMENTAL AVAILABluTY CIDENTS DUR!NG LOW POWER AND SHUTDOWN OPERATIONS AT OF URANIUM IN SOfLS AND SEDIMENTS.

GRAND GULF, UNIT 1. Analysis Of Core Darnage Frequency Frorn in-temal Events For Plant Operational State 5 During A Refuehng Outage.Section 10.

NUREG/CR4104: SHEAR WALL ULTIMATE DRIFT UMITS.

NUREG/CR4143 V02P1C: EVALUATION OF POTENTIAL SEVERE AC-Seiende Data CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4290 KEY ANALYSIS SYSTEM USER'S GUIDE. Version 2.0.

GRAND GULF, UNIT 1. Analysis Of Core Darnage Frequency Frorn in-temal Events For Plant Operational State 5 During A Refueling Setemic Design Load Outage. Main Report.

)

NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DEGRADED NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVERE AC-1 SHEAR WALL STIFFNESSES ON SEISMIC PLANT RISK AND SEiS-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT MIC DESIGN LOADS.

GRAND GULF. UNIT 1. Analysis Of Core Darnage Frequency From In-ternal Events For Plant Operational State 5 During Refueling Solemic Effect Outa0e. internal.

I NUREG/CR4169 RELAY TEST PROGRAM. Series il Tests. Integral Test-NUREG/CR4143 V02PT3: EVALUATION OF POTENTIAL SEVERE AC-Ing Q Relays And Orcuit Dreakers.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Seismic Event GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABILISTC temal Events For Plant Operational State 5 Dunng A Refueling RISK ASSESSMENT Outage.intemst-i NUREG/CR4143 VOSI EVALUATION OF POTENTIAL SEVERE ACUI-NUREG/CR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE AC.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency Frorn GRAND GULF,0 NIT 1. Analyses Of Core Damage Frequency From In-Seestnic Events Dunng M4 Loop Operations. Main Report temal Events For Plant Operational State 5 Dunng A Refueling NUREG/CR4144 V05: EVALUATION OF POTENTIAL SEVERE ACCI-Outage.intemal.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4143 V03: EVALUATON OF POTENTIAL SEVERE ACCl-SURRY, UNIT 1. Analysis Of Core Damage Frequency From Seismic DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Events Dunng M4 Loop Operations.Maan Report GRAND GULF. UNIT 1.Analysta Of Core Damage Frequency From In-i temal Events For Plant Operational State 5 Dunng A Refueling Outage.

)

Setemic Hazard NUREG/CR4143 V04: EVALUATION OF POTENTIAL SEVERE ACCl-NUREG 1488: REVISED UVERMORE SEISMIC HAZARD ESTIMATES DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT FOR SIXTY-NINE NUCLEAR POWER PLANT SITES EAST OF THE GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-ROCKY MOUNTAINS. Final Report.

temally induced Flooding Events For Plant Operational State 5 Dunng 8*'**"

NUREG 553 V05: EVALUATION OF POTENTIAL SEVERE ACCl-N E CR4254: SOUTHERN APPALACHIAN REGONAL SEISMIC DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NOREG/CR4255: DESIGN OF AN OPEN ARCHITECTURE SEISMIC GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From MONITORING SYSTEM' S*'**'c Events Dunno M4 Loop Operations. Main Report.

NUREG/CR-6144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-Setemic Network CIDENTS DURiNG LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4209:

MEMPHIS AREA REGIONAL SEISMIC SURRY, UNIT 1 Analysis Of Core Damage Frequency From intemal NETWORKJinal R

, October 1986 - Septembut t 992.

Events Dunno M4 Loop Operations Main Report (Chapters 712).

NUREG/CR4254:

HERN APPALACHIAN REGIONAL SEISMIC NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCl-NETWORK-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1.Analysas Of Core Damage Frequency From Intemal Seiernic Tunng Events Dunng M4 Loop Operations. Appendices A-0.

NUREG/CR 5935:

SUMMARY

OF WORK COMPLETED UNDER THE NUREG/CR4144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-ENVIRONMENTAL AND DYNAMIC EQUIPMENT OUAUFICATION RE-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT

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'9"*"CY N1

/CR4r E

NVESTIGATIONS OF THE HDR SAFETY PROGRAM'Summe Rm Events Dunng M4 Loop Operations. Appendices E (Sectione E.1.E.8).

NUREG/CR4144 V02P3B: EVALUATON OF POTENTIAL SEVERE AC.

Seismicity CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-3145 V10- GEOPHYSICAL INVESTIGATIONS OF THE SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal WESTERN OHIO INDIANA REGONFinal Roport October 1986-Sep-Events Dunng M4 Loop OperationsAppendices E (Sectione E 9-E.16).

tomber 1992.

NUREG/CR4144 V02P4: EVALUATON OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT I- -2 57 SURRY, UNIT 1 Analysis Of Core Darnage Fromency From internal NUREG/CP4133 V03: PROCEEDINGS OF THE TWENTY-FIRST Events Dunng Mid Loop OperationsAppendices F-H.

WATER REACTOR SAFETY INFORMATION MEETING.Pnmary NUREG/CR4144 V02P5: EVALUATON OF POTENTIAL SEVERE ACCI-System Integnty; Aging Research, Products & Applications; Structural &

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Seismic Engineenng-Seismology & Geology SURRY, UNIT 1 Analysis Of Core Damage Frequency From Intemal Events Dunno M4 Loop OperationsAppendices L Some Accident NUREG/CR4144 V03 P1: EVALUATON OF POTENTML SEVERE AC-NUREG-1502 ASSESSMENT OF DATABASES AND MODEUNG CAPA*

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT f,

7 ***""

N E/

33 S OF THE TWENTY FIRST op,,

WATER REACTOR SAFETY INFORMATON MEETING. Severe AccL NUREG/CR4144 V03 P2 EVALUATION OF POTENTIAL SEVERE AC.

NU G 51 V01 R1: EVALUATION OF SEVERE ACCOENT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RISKS: METHODOLOGY FOR THE CONTAINMENT SOURCE SURM,UNU LAnahses Of Com Damage Fmquency Nrn Intemal s % MW Operatens.W.

TERM. CONSEQUENCE AND RISK INTEGRATION ANALYSES NUREG/CR4075: THE PROBABlWTY OF CONTAINMENT FAILURE BY NUREG/CR4144 V04: EVALUATION OF POTENTIAL SEVERE ACCl-DIRECT CONTAINMENT HEATING IN ZION.

DENTS DURING LOW POWER AND SHUTDONN OPERATIONS AT NUREG/CR4075 S01: THE PROBABluTY OF CONTAINMENT FAIL.

SURRY, UNIT 1 Analysis Of Core Damage Frequency from Intemal URE BY DIRECT CONTAINMENT HEATING IN ZION.

Floods Dunng Mid-Loop Operations.

NUREG/CR4107:

SUMMARY

OF MELCOR 1.8 2 CALCULATONS FOR NUREG/CR4144 V05: EVALUATON OF POTENTIAL SEVERE ACCl-THREE LOCA SEQUENCES (AG.S2D & S3D) AT THE SURRY PLANT, DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4143 V02PI A: EVALUATION OF POTENTIAL SEVERE AC-SURRY, UNIT 1 Analysts Of Core Damage Frequency From Seismic CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Events Dunno M4 Loop Operatons Main Report GRAND GULF, UNIT 1 Analyses Of Core Damage Frerp.sency Fro n irk NUREG/CR4154 V01: EXPERIMENTAL RESULTS FROM CONTAIN-temal Events For Plant Operational State 5 Dunng A Refueling MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT Outage. Sections 19.

CONDITONS.Results From Bellows Tested in " Uke-New" Conditions.

98 Sul:Joct Index NUREG/CR-6158: IMPLICATIONS FOR ACCIDENT MANAGEMENT OF tornal Events For Plant Operatonal State 5 Dunng A Refuelmg ADONG WATER TO A DEGRADNG REACTOR CORE.

Outage Man Report NUREG/CR4168: DIRECT CONTAINMENT HEATING INTEGRAL EF-NUREG/CR-6143 V02PT2-EVALUATION OF POTENTIAL SEVERE AC-FECTS TESTS AT 1/40 SCALE IN ZION NUCLEAR POWER PLANT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GEOMETRY.

GRAND GULF, UNIT 1 Analyes Of Core Damage Frequency From In-NUREG/CR.6.33: PRIMARY SYSTEM FISSION PRODUCT RELEASE ternal Events For Plant Operatonal State 5 Durmg Refuelmg AND TRANSPORT.A State-Of The Art Report To The Comrnttee On Outage internsL..

The Safety Of Nuclear Installatons.

NUREu/CR-6143 V02PT3: EVALUATON OF POTENTIAL SEVERE AC-NUREG/CR4267: AIR-WATER SIMULATION OF PHENOMENA OF CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT CORIUM DISPERSON IN DIRECT CONTAINMENT HEATING.

GRAND GULF,tlNIT 1 Analyta Of Core Damage Frequency From In-ternal vents nt Opwam Sale 5 Dung A Meng Severe Accident h'OGNITIVE a.it NUREG/CR4126: O SKILL TRAINING FOR NUCLEAR NU E 4

"V02PT4 EVALUATION OF POTENTIAL SEVERE AC-POWER PLANT OPERATONAL DECISON MAKING.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Severe Core Damage GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency From In-NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE ternal Events For Plant OperatKmal State 5 Durmg A Refuelmg DAMAGE ACODENTS: 1992 A STA10S REPORT. Main Report Ard NU

/03: EVALUATON OF POTENTIAL SEVERE ACO-NU CR 4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DAMAGE ACCIDENTS' 1992 A STATUS REPORT. Appendices B, C, D.

GRAND GULF UNIT 1.Analyes Of Core Damage Frequency From In-E F And G tornal Events For Plant Operational State 5 Dunng A Refueleg Outage.

NUREG/ chi 674 V19 PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4143 V04: EVALUAllON OF POTENilAL SEVERE AGGi-DAMAGE ACCIDENTS 1993 A STATUS REPORT. Main Report And DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT

%-A.o GRAND GULF, UNIT 1 Analyss Of Core Damage Frequency From In-NUREG/CR-4874 V20: PRECURSORS TO POTENTIAL SEVERE CORE ternally Induced Floodirq Events For Plant Operational State 5 Durmg MAGE ACODENTS: 1993 A STATUS REPORTAppendences E fRe og 61 3 V05: EVALUATION OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Severe Reactor Accident GRAND GULF, UNIT 1 Anatyss Of Core Damage Frequency From NUREG/CR4152-EXPERIMENTS TO INVESTIGATE DIRECT CON.

Seistmc Events Dunng M4 Loop Operatons Mam Report.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NUREG/CR4144 V02 PIA: EVALUATION OF POTENTIAL SEVERE AC-SURRY NUCLEAR POWER PLANT.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4211: INTEGRATED FUEL COOLANT INTERACTION (IFOI SURRY, UNIT 1 Analyses Of Core Damage Frequency From Internal 6 0) CODE. User's Manual Events Dunng M4 Loop Operatens Man Report (Chapters 16)

NUREG/CR-6218 A REVIEW OF THE TECHNICAL ISSUES OF AIR Id-NUREG/CR4144 V02P18 E.VALUATON OF POTENTIAL SEVERE AC-GRESSON DURING SEVERE REACTOR ACCIDENTS.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY. UNIT 1 Analyss Of Core Damage Frequency From Internal Seww D6sposal Events Dunng M4 Loop Operations Man Report (Chapters 712)

NUREG/CR4289-RECONCENTRATION OF RADIOACTIVE MATERIAL NUREG/CR4144 V02P2 EVALUATION OF POTENTIAL SEVERE ACCl-RELEASED TO SANITARY SEWERS IN ACCORDANCE W1TH 10 CFR DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT PART 20-SURRY, UNIT 1 Analyss Of Core Damage Frequency From internal Events Dunng M4 Loop Operations A es A-D Shear Stresa NUREG/CR4144 V02P3A: CVALUAT OF POTENTIAL SEVERE AC.

NUREG/CR-6178, LADORATORY CHARACTER 12ATON OF ROCK CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT 0

^

• 1. "*8*

'9"*"#Y NU EG/CR4216. EVALUATON OF ROCK JOINT MODELS AND COM-PUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS.

N CR 4 VO 8 A

T TEN SE Shear Wall CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DEGRADED SURRY, UNIT 1 Analyses Of Core Damage Frequency From internal SHEAR WALL STlFFNESSES ON SEISMIC PLANT HISK AND SEIS-NR CR4 4 P E lON PO ENTI VE E NL GE 4 SH AR WALL ULTIMATE DRIFT LIMITS.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1. Analyses Of Core Damage Frequency From Internal Shift Composition Events Dunng M4 Loop Operations Appereces F H.

NUREG/CR4122-STAFFING DEOSiON PROCESSE, 9

NUREG/CR4144 V02P5. EVALUATON OF POTENTIAL SEVERE ACCl-ISSUES Case Studies Of Seven U S Nuclear Power Plants.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR 6123 AN INTERNATIONAL COMPARISON OF COMMEH-SURRY, UNIT 1 Analyses Of Core Damage Frequency From internal CAL NUCLEAR POWER PLANT STAFFING REGULATIONS AND Events Durmg M4 Loop Operations Appereces I PRACTICE 1980-1990.

NUREGICR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-ODENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Shock Wave SURRY, UNIT 1 Analyse Of Core Damage Frequency From internal NUREG/CR 5960. STEAM EXPLOSIONS. FUNDAMENTALS AND EN-Fires Dunna M4 Loop Operations Man Report.

GERGETIC BEHAVIOR.

NUREG/CM144 V03 P2: EVALUAllON OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Shutdown SURRY, UNIT 1 Analyse Of Core Damage Frequency From internet NUREG/CR4093. AN ANALYSIS OF OPERATIONAL EXPERIENCE Fires Durno M4 Loop Operations AppenGees.

DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD.

NUREG/CH 6144 V04 EVALUATION OF POTENTIAL SEVERE ACCI-DRESSING HUMAN REllABILITY ASSESSMENT ISSUES.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1 Analyws Of Core Damage Frequency From internal Shutdown Opwation Floods Dunno M4 Loop Operatiora NUFIEG/CR4143 V0291A EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR 6144 Voi EVALUATON OF POTENTIAL SEVERE ACCl-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF. UNIT 1Analywa Of Core Damage Frequency From In-SURRY, UNIT 1 Ana'yms Of Core Damage Frequency From Seismic ternal Events For Plant Operational State 5 Dunng A Refueling Events Dunng M4 Loop Operations. Man Report Outage Sections 1-9 NUREGICR4143 V02P18 EVALUATON OF POTENTIAL SEVERE AC-86mpimed Bo416ng Water Reactor CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4223 REVtEW OF THE PROPOSED MATERIALS OF CON-GRAND GULF UNIT 1 Analysis Of Core Damage Frequency From In-STRUC180N FOR THE SBWR AND AP600 ADVANCED REACTORS, ternal Events For Plant Operational State 5 Dunng A Refusing Outage Section 10 Simulated Emergencies NUREG/CR4143 V02P10: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4208: AN EMPIRICAL INVESTIGATON OF OPERATOR COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER.

GRAND GULF UNIT 1 Analysts Of Core Damage Frequency From In-GENCIES.

Subject index 99 Software Stress NUREG/CR4278: SURVEY OF INDUSTRY METHODS FOR PRODUC-NUREG/CR-6127: THE EFFECTS OF STRESS ON NUCLEAR POWER ING HIGHLY RELIABLE SOFTWARE.

PLANT OPERATIONAL DECISION MAKING AND TRAINING AP-NUREG/CR4294: DESIGN FACTORS FOR SAFETY-CRITICAL SOFT.

PROACHES TO REDUCE STRESS EFFECTS.

WARE.

S0il Stress Corrosion CracWng NUREG/CR4067 V17; ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6232: ASSESSING THE ENVIRONMENTAL AVAILABILITY OF URANIUM IN SOILS AND SEDIMENTS.

LIGHT WATER REACTORS. Serruannual Report, April 1993 - Septem-ber 1993.

M Waste Disposal NUREG/CR4176: REVIEW OF ENVIRONMENTAL EFFECTS ON FA.

NUREG/CR-2907 V12: RADIOACTIVE MATERIALS RELEASED FROM TIGUE CRACK GROWTH OF AUSTENITC STAINLESS STEELS.

NUCLEAR POWER PLANTS. Annual Report 1991.

Suppression Pool Heating Source Poettion Monitor NUREG/CR-6200: UNCERTAINTY ANALYSIS OF SUPPRESSION POOL HEATING DURING AN ATWS IN A BWR-5 PLANT.An Apphcataan Of NUREG/CR-4833: LARGE AREA SELF-POWERED GAMMA RAY DETECTOR. Phase il CE:,.-a Of A Source Posthon Morntor For The CSAU Methodology Using The BNL Engineenng Plant Analyzer.

Use On Industnal Radiographic Uruts.

Surtsey Test Faclitty NUREG/CR4152; EXPERIMENTS TO INVESTIGATE DIRECT CON.

SOUT R APPALACHIAN REGONAL SEISMC TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NETWORK.

SURRY NUCLEAR POWER PLANT.

Spent Fuel System 80+

NUREG 1323 ROO: LCENSEE APPLICATION REVIEW PLAN FOR A NUREG/CR 6161: BUCKLING EVALUATION OF SYSTEM 80+(TM)

GEOLOGC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-CONTAINMENT.

LEVEL RADIOACTIVE WASTE-NUREG 1497: INTERIM LCENSING CRITERLA FOR PHYSCAL PRO-System 80+ Design TECTION OF CERTAIN STORAGE OF SPENT FUEL NUREG-1462 V01: FINAL SAFETY EVALUATION REPORT RELATED TO THE CERTIFICATION OF THE SYSTEM 80 + DESIGN. Chapters 1 Spent Fuel Storage 14 Docket No. 52 002. (Asea Brown Boven-Combustion Engineenng)

NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSED DECAY NUREG-1462 V02 FINAL SAFETY EVALUATION REPORT RELATED HEAT GUIDE USING SAS2H/ORIGEN-S DATA TO THE CERTIFCATION OF THE SYSTEM 80+ DESIGN. Chapters 15-22 And Appendices. Docket No. 52402.(Asea Brown BoveriCom-Staffing busbon Engineenng) i NUREG/CR4122: STAFFING DECISION PROCESSES AND ISSUES. Case Studies Of Seven U S. Nuclear Power Plants.

System Reliablitty NUREG/CP4136: PROCEEDINGS OF THE DIGITAL SYSTEMS REll-Starting Regulation ABILilY AND NUCLEAR SAFETY WORKSHOP. September 13-14, NUREG/CR4123: AN INTFRNATIONAL COMPARISON OF COMMER-1993,Rockvihe Crowne Plaza liotel.Rockville, Maryland.

CIAL NUCLEAR POWER PLANT STAFFING REGULATIONS AND I

PRACTICE.1980-1990 System Safety I

NUREG/CR-6294: DESIGN FACTORS FOR SAFETY CRITCAL SOFT.

Stain-Life Curve WARE.

NUREG/CR4237: STATISTICAL ANAL) 6 GF FATIGUE STRAIN-LIFE DATA FOR CARBON AND LOW-ALLOY STEELS.

Systematic Assessment Of Ucensee Performance Stainless $6 eel NUREG 1214 R13: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

IC ASSESSMENT OF LICENSEE PERFORMANCE.

NUREG/CR-4513 R01: ESTIMATION OF FRACTURE TOUGHNESS OF CAST STAINLESS STEELS DURING THERMAL AGING IN LWR SYS-TLD TEMS-NUREG-0837 VIJ N04: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-6176: REVIEW OF FWT MENTAL EFFECTS ON FA-NETWORK. Progress Report. October December 1993.

[

TIGUE CRACK GROWTH OF A (?Wi STAINLESS STEELS.

I NUREG/CR4177. ASSESSMENT P *WMAL EMBRITTLEMENT OF NUREG4837 V14 N01: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report. JanuaryMarch 1994.

NU

/

VO S ILITY OF CRACKED PIPE UNDER INER-NE ss e TIAL STRESSES. Subtask 1.1 Fanal Report.

NUREG-0837 V14 NO3. NRC TLD DIRECT RADIATION MONITORING gw p,g pg,,

NETWORK. Progress Report. July September 1994, NUREG-1200 R03: STANDARD REVIEW PLAN FOR THE REVIEW OF A TMt Action Plan LICENSE APPLCATION FOR A LOW-LEVEL RADIOACTIVE WASTE NUREG-1435 S03: STATUS OF SAFETY ISSUES AT LICENSED NUR CR 59 CODES AND STANDARDS AND OTHER GUID-POWER PLANTS.TMI Actson Plan Requirements. Unresolved Safety ANCE CITED IN REGULATORY DOCUMENTS.

Issues Genenc SaMy homer Want Achon issues.

Stat 6on Blackout 1MF2 NUREG/CR 5850 ANALYSIS OF LONG-TERM STATION BLACKOUT NUREG/CR4183: PEER REVIEW OF THE TMI.2 VESSEL INVESTIGA-WITHOUT AUTOMATIC DEPRESSURIZATION AT PEACH BOTTOM TION PROJECT METALLURGICAL EXAMINATIONS.

USING MELCOR (VERSION 1.8).

NUREG/CR-6185: TMI-2 INSTRUMENT NOZZLE EXAMINATIONS AT ARGONNE NATIONAL LABORATORY. February 1991 - June 1993.

Statistics NUREG/CR6187: RESULTS OF MECHANICAL TESTS AND SUPPLE-NUREG M, ePLYING STATISTICS.

MENTARY MICROSTRUCTURAL EXAMINATIONS OF THE TMI-2 LOWER HEAD SAMPLES.

Steam L%m NUREG/CR4195. EXAMINATION OF RELOCATED FUEL DEBRIS AD-NUREGWP 0127: PROCEEDINGS OF THE CSNI SPECIALISTS MEET.

JACENT TO THE LOWER HEAD OF THE TMI 2 REACTOR VESSEL ING ON Net-COOLANT INTERACTIONS.

NUREG/CR4197: TMI 2 VESSEL INVESTIGATION PROJECT INTE.

NUREG/CR406& STFAM EXPl.OfWWS: FUNDAMENTALS AND EN.

GRATION REPORT.

GERGETIC BEHAVIOR.

NUREG/CR-6198 Thl-2 INSTRUMENT NOZZLE EXAMINATIONS PER-FORMED AT THE INEL.

Steam Generator Tut >e NUREG/CR.51$1 V02: EVALUATION OF SAMPLING PLANS FOR IN-TRAC-PF1/ MOD 2 SERVICE INSPECTION OF STEAM GENERATOR NUREG/CR.6269: A PLAN FOR THE MODIFICATION AND ASSESS.

TUBES Comprehenssve AnaYbcal And Monte Carto Simulation Results MENT OF TRAC-PF1/ MOD 2 FOR USE IN ANALYZING CANDU 3 For Several Samping Plans.

TRANSIENT THERMAL-HYDRAULIC PHENOMENA.

1 I

1 M M lOdSE

- Technical tuWenne NUREG-0540 V15 N12: TITLE UST OF DOCUMENTS MADE PUBUCLY NUREG/CR4241: TECHNICAL GUIDEUNES FOR ASEISMIC DESIGN AVMLABLE. December 1 31,1993.

' ]

OF NUCLEAR POWER PLANTS Treneleton Of JEAG 4001 1987.

NUREG-0540 Vie N01: TITLE UST OF DOCUMENTS MADE PUBUCLY AVAILA8LE.Jonuary 1 31,1994.-

Technical SpeeWisemen NUREG 0540 V16 NO2: TITLE UST OF DOCUMENTS MADE PUBUCLY j

NUREG/CR4198: RISK IMPACT OF TECHNICAL SPECIFICATIONS RE-AVMLABLE.Felwuery 148,1994.

OUIREMENTS DURNG SHUTDOWN FOR SWRS.

NUREG 0540 V16 NO3: TITLE UST OF DOCUMENTS MADE PUBLICLY 5

I AVAILABLE. March 1 31,1994.

Technical Training Center

' NUREG4640 V16 N04: TITLE UST OF DOCUMENTS MADE PUBLICLY j

NUREG/CR4042: PERSPECTIVES ON REACTOR SAFETY, AVAILABLE. April 1 30,1994.

NUREG-0640 V16 N05: TITLE UST OF DOCUMENTS MADE PUSUCLY 1

Tenelle AVMLABLE.Mey 1 31,1994.

1 NUREG/CR4249; UNIRRADIATED MATERIAL PROPERTIES OF MID-NUREG4640 V16 N06: TITLE UST OF DOCUMENTS MADE PUSUCLY

.j LAND WELD WF-70.

AVAILABLE. June 1 30.1994.

I NUREG4640 V16 N07: TITLE UST OF DOCUMENTS MADE PUBLICLY Tennes Leseng AVMLABLE. July 1 31,1994.

NUREG/CR4051: EFFECTS OF TENSILE LOADMG ON UPPER SHELF NUREG4640 V16 NOS: TITLE UST OF DOCUMENTS MADE PUSUCLY -

FRACTURE TOUGHNESS.

AVMLASLE. August 1 31,1994.

NUREG4640 V16 N00: TITLE UST OF DOCUMENTS MADE PUBUCLY i

Teneto Propony

. AVMLABLE. September 1 30, 1994.

f NUREG/CR4142: TENSILE PROPERTY CHARACTERIZATION OF NUREG4540 V16 N10 TITLE UST OF DOCUMENTS MADE PUSUCLY THERMALLY AGED CAST STANLESS STEELS.

AVAILABLE. October 131,1994.

Test Reac W yeggens gege,g NUREG/CR4076: TR EDB: TEST REACTOR EMBRITTLEMENT DATA NUREG-0300 Vos N01: TOPICAL REPORT REVIEW' STATUS.(Blue

.r BASE, VERSION 1.

goag

[

Therapeuse m Training NUREG 1492 DFC: REGULATORY ANALYSIS ON CRITERIA FOR THE NUREG/CR4127: THE EFFECTS OF STRESS ON NUCLEAR POWER RELEASE-OF PATIENTS ADMINISTERED RADIOACTIVE PLANT OPERATIONAL DECISION MAKING AND TRAINING AP-MATERIALDraft Report For Comment PROACHES TO REDUCE STRESS EFFECTS. -

l Thennel Aging Treneamertes Delevel,Inc NUMEG/CR4513 R01: ESTIMATION OF FRACTURE TOUGHNESS OF NUREG-1416: OPERATIC 04AL EXPERIENCE AND MAINTENANCE PRO-

)

CAST STAINLESS STEELS DURNG THERMAL AGING IN LWR SYS -

GRAMS OF TRANSAMERICA DELAVAL, NC., DOESEL GENERA.

TORS.

/CR4142: TENSILE-PROPERTY CHARACTERIZATION OF THERMALLY AGED CAST STAINLESS STEELS.

Trenelent Anaspels l

NUREG/CR4177: ASSEnnMFNT OF THERMAL EMBRITTLEMENT OF NUREG 1902: ASSESSMENT OF DATABASES AND MODELNG CAPA.

CAST STAINLESS STEELS.

. SILITIES FOR THE CANOU 3 DESIGN.

Damage y,,,,,,,, gg,,g,gg,,

3 NUREG/CR4194: METALLOGRAPHIC AND HARDNESS EXAMINA NUREG/CR-5636 V06: RELAP5/ MOO 3 CODE MANUALVeluiston Of

l TIONS OF TML-2 LOWER PRESSURE VESSEL HEAD SAMPLES.

Numencel Techruques in RELAP5/ MOO 3' NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE.

GRATION REPORT.

Treneluen Temperature.

NUREG/CR4262: CLEAVAGE BEHAVIORS IN NUCLEAR VESSEL Thermal Delosamen STEELS.

NUREG/CR-See0 STEAM EXPLOSIONS: FUNDAMENTALS AND EN-GERGETIC BEHAVIOR.

TW NUREG/CR4114 V03: PERFORMANCE ASSESSMENT OF A HYPO-1 Thenned Embransment THETICAL LOW LEVEL WASTE FACluTY. Groundwater Flow And NUREG/CR4142: TENSILE PROPERTY CHARACTERIZATION OF THERMALLY AGED CAST STANLESS STEELS Tran p rt h NUREG/CR4177: ASSESSMENT OF THERMAL ' EMBRITTLEMENT OF NUREG/CR4203: VAUCATION STUDIES FOR ASSESSNG UNSATU-RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK -

CAST STAINLESS STEELS.

Thennel Overtseg Transperheen NUREG4383 V01 R17: DIRECTORY OF CERTIFICATES OF COMPU.

NUREG/CR4205: VALVE /;;TUAMH enOTOR DEGRADATION.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC Thermes Response Apoved Pachegos NUREG4383 V02 R1'7: DIRECTORY OF CERTIFICATES OF COMPU-NUREGICR4198: CALCULATIONS TO ESTIMATE THE MARGW TO FAILURE IN THE TMI-2 VESSEL ANCE FOR RADIOACTIVE MATERIALS PACKAGESCerencesse Of Thenno64tystrends Phenomens N

43liEV03 R14; DIRECTORY'OF CERTIFICATES OF COMPU.

NUREG/CR42e9 A PLAN FOR l'HE MODIFICATION AND ASSESS.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC MENT OF TRAC.PF'%O02 FOR USE IN ANALYZING CANDU 3 Approved QueNy Assurance Programs For R% Malenels Peck.

TRANSIENT THERMAL +fYDRAUUC PHENOMENA.

eens.

Thennelumineesent Destmeter Tube Feture NUREG4837 V13 N04: NRC TLD DIRECT RADIATION MONITORING NUREG/CR4196: CALCULATIONS TO ESTIMATE THE MARGIN TO NETWORK.Progrees Report the=@ecember 1993.

FAILURE IN THE TML-2 VESSEL NUREG4837 V14 Not: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report.

44erch 1994.

UDEC NUMEG-OS37 V14 N02: NRC TLD CT RADIATION MONITORING NUREG/CR4216: EVALUATION OF ROCK JOINT WODELS AND COM-NETWORK.Proyees Report AprB. June 1994.

PUTER CODE UDEC AGAINST EXPERRIENTAL RESULTSc NUREG0037 V14 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK.Proyese Report. July "

_ 1994.

Untrredleted Meterial NUREG/CR4249: UNIRRADIATED MATERIAL PROPERTIES OF MID-Three Ime latend LAND WELD WF-70.

NUREG/CR4252: LESSONS LEARNED FROM THE THREE MILE ISLAND. UNIT 2 ADVISORY PANEL Unastureted Floor NUREG/CR-6086: MODELING FIELD SCALE UNSATURATED FLOW Title List AND TRANSPORT PROCESSES.

NUREG4540 V15 N11: TITLE UST OF DOCUMENTS MADE PUBLICLY NUREG/CR4203: VALIDATION STUDIES FOR ASSESSING UNSATU-AVAILABLE November 1 30,1993.

RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCet

Subject Index 101 Upper Shelf Vibration Experiment NUREG/CR4051: EFFECTS OF TENSILE LOADING ON UPF ER SHELF NUREG/CR4236: SEISMIC INVESTIGATIONS OF THE HOR SAFETY FRACTURE TOUGHNESS.

PROGRAM. Summary Report.

Urarnum Vital Barrier NUREG/CR4232: ASSESSING THE ENVIRONMENTAL AVAILABluTY NUREG/CR4190 V01 R1: PROTECTON AGAINST MALEVOLENT USt:

OF URANIUM IN SOILS AND SEDIMENTS.

OF VEHICLES AT NUCLEAR POWER PLANTS.Vetuele Bamer System Selecton Cudance For Blast Protecton.

UtNety Requirements Document NUREG/CR4190 V02 RI: PROTECTON AGAINST MALEVOLENT USE NUREG-1242 V03 PT01: NRC REVIEW OF ELECTRIC POWER RE.

OF VEHICLES AT NUCLEAR POWER PLANTS.Vetuele Bamer System SEARCH INSTITUTE'S ADVANCED UGHT WATER RFACTOR UTILI.

Selecton Gudance.

TY REQUIREMENTS DOCUMENTSPasswo Plant Designs. Chapter Watte Surial EM307 m REN ON WASTE BM CHAEMacalam NUR 42 PT02: NRC REVIEW OF ELECTRIC POWER RE-SEARCH INSTITUTE'S ADVANCED UGHT WATER REACTOR UTill-Decomnussoning Waste Disposal Costs At Movel Waste Bunal FacM es.

TY REQUIREMENTS DOCUMENT. Passive Plant Des #gns. Chapters 2-

13. Protect Number 669.

Weste Compact NUREG/CR4147 V01: CHARACTER 12ATION OF CLASS A LOW LEVEL Vadose hne Transpwt RADIOACTIVE WASTE 19861990 Executwo Summary NUREG/CR 6120: CONTROLLED FIELD STUDY FOR VALIDATION OF NUREG/CR4147 V02: CHARACTERIZATION OF CLASS A LOW-LEVEL VADOSE ZONE TRANSPORT MODELS-RADIOACTIVE WASTE 1986-1990 Main Report-Part A.

NUREG/CR4147 V03: CHARACTERtZATION OF CLASS A LOW-LEVEL Valve RADOACTIVE WASTE 1986-1990. Main Report-Part B NUREG/CR-5935:

SUMMARY

OF WORK COMPLETED UNDER THE NUREG/CR4147 V04: CHARACTERIZATION OF CLASS A LOW LEVEL ENVIRONMENTAL AND DYNAMIC EOUIPMENT OUALIFICATION RE-RADIOACTIVE WA3TE 1986-1990.

dces A-E.

SEARCH FROGRAM (EDOP).

NUREG/CR4147 V05: CHARACTERI TlON OF CLASS A LOW LEVEL RADIOACTIVE WASTE 19861990 Apperdx F.

Vahre And Pump Testing NUREG/CR4147 V03: CHARACTERizATON OF CLASS A LON LEVEL NUREG/CP0137 VOI: PROCEEDINGS OF THE THIRD NRC/ASME RADIOACTIVE WASTE 19861190. App Mices G4 SYMPOSIUM ON VALVE AND PUMP TESTING. Held At The Hyatt Re-NUREG/CH4147 V07: CHARACTERIZATION C O.?O a LOW-LEVEL gency Hotel, Washington.DC, July 18-21,1994.Sesson 1 A. Sesson RADIOACTIVE WASTE 19861990.Appereces K P.

Waste Generator Vahre Testing NUREG/CR 6147 VJ1: CHARAvTERIZATION OF CLASS A LOW-LEVEL NUREG/CP-0137 V02-PROCEEDINGS OF THE THIRD NRC/ASME NURE 4 47V :

A ION OF C A LOW-LEVEL SYMPOSIUM ON VALVE AND PUMP TESTING.Hald At The Hyatt Re-RADIOACTIVE WASTE '9861990 Main Report-Part A.

gency Hotel, Washington,DC. July 18-21,1994.Sesson 3A -Sesson NUREG/CR4147 V03: CH ARACTERIZATON OF CLASS A LOW-LEVEL 40-RADIOACTIVE WASTE ',986-1990 Mesn Aeport-Part B.

NUREG/CR4147 V04: Cf(ARACTERIZATION OF CLASS A LOW-LEVEL Vapor Transport RADIOACTIVE WASTr. 1986-1990.Apperdces A-E-NUREG/CR 5965: MODEUNG FIELD SCALE UNSATURATED FLOW NUREG/CR4147 V05 CHARACTER 12ATION OF CLASS A LOW-LEVEL AND TRANSPORT PROCESSES-RADIOACTIVE WNTE 1986-1990.Apperex F.

NUREG/CR4147 VJ6. CHARACTERl2ATION OF CLASS A LOW LEVEL Vehicle Barrier System RADIOACTIVF WASTE 19861190. Appendices G4 NUREG/CR4190 V01: PROTECTON AGAINST MALEVOLLMT USE OF NUREG/CR4?47 V07: CHARACTER 12ATON OF CLASS A LOW LEVEL VEHICLES AT NUCLEAR POWER PLANTSVehicle Bamer System RADIOAC'etVE WASTE 1986-1990 Appereces K-P.

SetinDGuidance For Blast Protection.

NUREG/CR4190 V01 Rt: PROTECTION AGAINST MALEVOLENT USE Water Flow OF VEHICLES AT NUCLEAR POWER PLANTS Vehicle Samer System NUREG/CR4063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS Selecton Guidance For Blast Protecton.

CRUCES TRENCH SITE.

NUREG/CR4190 V02-PROTECTON AGAINST MALEVOLENT USE OF W "

VEHICL TN POW PLANTSVetucle Bamer System R

C -4 18 V07: CONTROL OF WATER INFILTRATION INTO NU

/CR4190 V02 R1: PROTECTON AGAINST MALEVOLENT USE NEAR SURFACE LLW DISPOSAL UNITS. Progress Fieport On Field Ex-OF VEHICLES AT NUCLEAR POWER PL ANTE.Velucle Bamer System penments At A Hurrud Regon Site.Beltsvale. Maryland.

Selecten Guidance.

Weld Vendw inspection NUREG/CR4228. PREUMINARY ASSESSMENT OF THE FRACTURE NUREG4040 V17 N04: UCENSEE CONTRACTOR AND VENDOR IN_

BEHAVIOR OF WELD MATERIAL IN FULL THICKNESS CLAD DEAMS.

SPECTION STATUS REPORT. Quarterly Report,0ctober December 1D93.(White Book)

Weld WF-70 NUREG4040 V18 N01: LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR4249. UNIRRADIATED MATERIAL PROPERTIES OF MID-SPECTION STATUS REPORT. Quarterly Report. January-March LAND WELD WF 70.

1994.(White Book)

NUREG-0040 V18 NO2-LICENSEE CONTRACTOR AND VENDOR IN-Whletleblower SPECTION STATUS REPORT. Quarterly Report.AprilJune 1994.(Wtuto NUREG-1499: REASSESSMENT OF THE NRC'S PROGRAM FOR PRO-Book)

TECTING ALLEGERS AGAINST RETALIATION.

NUREG4040 V18 NO3. LICENSEE CONTRACTOR AND VEICOR IN-SPECTION STATUS REPORT. Quarterty Report. July September Workshop 1994.(Wtute Book)

NUREG/CH4156:

SUMMARY

OF COMMENTS RECE!VED FROM WORKSHOPS ON RADIOLOGICAL CRITERIA FOR DECOMMISSION-Vessel ING.

NUREG/CR4195: EXAMINATON OF RELOCATED FUEL DEBRIS AD.

JACENT TO THE LOWER HEAD OF THE TMI-2 REACTOR VESSEL.

Yucca Mountain NUREG/CR4196: CALCULATIONS TO ESTIMATE THE MARGIN TO NUREG-1494. STAFF TECHNICAL POSITION ON CONSOERATON OF FAILURE IN THE TMI-2 VESSEL FAULT DISPLACEMENT HAZARDS IN GEOLOGIC REPOSITORY DESIGN.

Vessel investigation Project NUREG/CR4288. GEOCHEMICAL INVESTIGATIONS RELATED TO NUREG/CR-8197: TMI-2 VESSEL INVESTIGATION PROJECT INTE.

THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NUCLE-GRATON REPORT.

AR WASTE REPOSITORY.

s 4

1 I

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4 4.

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E I

NRC Originating Organization Index (Staff Reports)

This index lists those NRC organizations that have published staff reports. The index is ar-ranged alphabetically by major NRC organizations (e.g., program offices) and then by sub-sections 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.

ADVISORY COMMITTEE (S)

NUREG-0540 V15 N12: TITLE LIST OF DOCUMENTS MADE PUBLIC-ACRS ADVISORY COMMITTEE ON REACTOR SAFEGUARDS LY AVAILABLE. December 1-31,1993.

i NUREG-1125 V15: A COMPILATON OF REPORTS OF THE ADVISO-NUREG-0540 V16 N01: TITLE LIST OF DOCUMENTS MADE PUBLC RY COMMITTEE ON REACTOR SAFEGUARDS.1993 Annual.

LY AVAILABLEJanuary 1 31, 1994 j

NUREG4540 V16 NO2: TfTLE UST OF DOCUMENTS MADE PUBLIC-l OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

OFC OF THE EXECUTIVE DIRECTOR FOR OPERATIONS LY AVAILABLE February 1 28.1994.

[

NUREG-1489: A REVIEW OF NRC STAFF USES OF PROBABIUSTIC NUREG-0540 V16 NO3: TITLE UST OF DOCUMENTS MADE PUBLC RISK ASSESSMENT.

LY AVAILABLE March 1 31,1994.

NUREG 0540 V16 N04: TITLE UST OF DOCUMENTS MADE PUBLC NUREG 1499: REASSESSMENT OF THE NRC'S PROGRAM FOR PROTECTING ALLEGERS AGAINST RETAUATION.

LY AVAILABLE.Apnt i 30.1994.

NUREG 0750 V37; NUCLEAR REGULATORY COMMISSION NU G 7 13 04 NRC TLD DIRECT RADIATION MONITORING NETWORK Progress Reoort October-December 1993-s W 6ected Or ry-J 1993 NUREG-0837 V14 N01: NRC TLD DIRECT RADIATION MONITORING NUREG4750 V38: NUCLEAR REGULATORY COMMISSON l

NETWORK. Progress Report January-March 1994 ISSUANCES.Operwons And Decssions Of The Nuclear Regulatory l

NUREG4837 V14 NO2: NRC TLD DIRECT RADIAT' ION MONITORING Commission With Selected OrdersJulL-December 1993 NUREG-0750 V38101: INDEXES TO NUULEAR REGULATORY COM-NURE 083 1 3N T RC DIATION MONITORING NU 8 102 ND X T

U AR REGULATORY COM-NETWORK Pr ess Report July-September 1994.

M SSION ISSUANCESJuly December 1993.

NU G V12 3i R

E T ACTIONS. SIGNIFICANT AC-NUREG-0750 V38 N05: NUCLEAR FIEGULATORY COMMISSION IS.

{

N E A

E R

WSSG G NUR 0 12 FORCE E A ION S I

N TIONS RESOLVED.Ouarterty Progress Report,0ctober-Decembe' NUREG V9 01 LEAR RE L COMMISSION IS-NUREG-0940 V13N01P01: ENFORCEMENT ACTIONS. SIGNIFICANT NU EG V39 2N R

UL ORY COMMISSION IS-ACTONS RESOLVED REACTOR UCENSEESQuarterly Progress SUANCES FOR FEBRUARY 1994. Pa a 47 90.

NUR 3 IP02 NFORCEMENT ACTIONS SIGNIFICANT SU N F

P 91 6 ACTONS RESOLVED MEDICAL UCENSEEJOuerterty Progress Report. January-March 1994.

NUREG-0750 V39 N04: NUCLEAR EGULAT'ORY COMMISSION IS.

NUREG 0940 V13N01P03: ENFORCEMENT ACTIONS SIGNtFCANT SUANCES FOR APRIL 1994 Pages 187 247.

NUREG-0936 V12 N04: NRC REGULATORY AGENDA.Ouarterty ACTONS RESOLVED INDUSTRIAL UCENSEES Quarterly Progress Report.Octobw-Deewnbw 1993.

NUR 37 02P01 NFORCEMENT ACTONS: SIGNIFICANT

[gg p

]

l ACTIONS RESOLVED REACTOR UCENSEES.Ouartetty Progress OFFCE OF ADM ISTRATION DIRECTOR (POST 940714) i Report. April-June 1994.

NUREG 1145 V10: U S NUCLEAR REGULATORY COMMISSION i

NUREG-0940 V13N02P02. ENFORCEMENT ACTIONS' SIGNIFICANT 1993 ANNUAL REPORT.

ACTONS RESOLVED MEDICAL LICENSEESouarterty Progress OlVISION OF FREEDOM OF INFORMATION & PUBLICATIONS SERV-Report.Apnt-June 1994 CES (POST 940714 NUREG-0940 Vt3N02P03: ENFORCEMENT ACTIONS: SIGNIFICANT NUREG-0304 V19 NO2: REGULATORY AND TECHNICAL REPORTS ACTONS RESOLVED INDUSTRIAL UCENSEES Quarterty Progress (ABSTRACT INDEX JOURNAL). Compitabon For Second Quarter Report.Aoni-June 1994 1994, April. June.

NUREG-0940 V13NO3P01: ENFORCEMENT ACTIONS SIGNIFICANT NUREG 0540 V16 N05: TITLE UST OF DOCUMENTS MADE PUBUC-ACTIONS RESOLVED. REACTOR UCENSEES Quarterly Progress LY AVAILABLE May 1 31,1994 Report. July-September 1994.

NUREG-0540 V16 N06: TITLE UST OF DOCUMENTS MADE PUBLC NUREG-0940 V13NO3P02: ENFORCEMENT ACTIONS' SIGNIFICANT LY AVAILABLEJune 1 30,1994.

ACTIONS RESOLVED. MEDICAL UCENSEES Ouarterty Progress NUREG4540 V16 N07: TITLE LIST OF DOCUMENTS MADE PUBLC Report. July-September 1994.

LY AVAILABLEJuly 1 31,1994 NUREG0940 V13NO3P03 ENFORCEMENT ACTIONS-SiGNIFICANT NUREG-0540 V16 N08: TITLE UST OF DOCUMENTS MADE PUBLIC-ACTIONS RESOLVED. MATERIAL UCENSEES (NON-LY AVAILABLE. August 1 31,1994.

MEDICAL)Quarterfy Progress Report. July-September 1994 NUREG 0540 V16 N09: TITLE UST OF DOCUMENTS MADE PUBLC OFC OF PERSONNEL (POST B70413)

LY AVAILABLE. September 1 30.1994 NUREG-0325 R17: U.S. NUCLEAR REGULATORY COMMISSON OR-NUREG 0540 V16 N10: TITLE UST OF DOCUMENTS MADE PUBLC GANIZATION CHARTS AND FUNCTONAL STATEMENTS October LY AVAILABLE. October 1 31,1994.

3.1994.

NUREG 0750 V39101: IPOEXES TO NUCLEAR REGULATORY COM-MISSION ISSUANCESJanuary -March 1994.

EDO - OFFICE OF ADMINISTRATION (PRE 870413 & POST 890205)

NUREG 0750 V39102. INDEXES TO NUCLEAR REGULATORY COM-DIVISION OF FREEDOM OF INFORMATION & PUBUCATIONS SERV-MISSION ISSUANCES. January June 1994 1CES (890206 9407 NUREG 0750 V39 N05: NUCLEAR REGULATORY COMMISSION IS-NUREG 0304 V18 N04. REGULATORY AND TECHNICAL REPORTS SUANCES FOR MAY 1994 Pages 249-284.

(ABSTRACT INDEX JOURNAL) Annual Compilabon For 1993.

NUREG-0750 V39 N06: NUCLEAR REGULATORY COMMISSION IS-NUREG 0304 V19 N01: REGULATORY AND TECHNICAL REPORTS SUANCES FOR JUNE 1994 Pages 285-390.

(ABSTRACT INDEX JOURNAL). Compdahon For Frst Quarter NUREG 0750 V40101: INDEXES TO NUCLEAR REGULATORY COM-1994. January-March.

MISSION ISSUANCESJuty-September 1994.

NUREG 0540 V15 NII: TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG-0750 V40 N01: NUULEAR REGULATORY COMMISSION IS-LY AVAILABLE November 1 30,1993 SUANCES FOR JULY 1994 Peges 1-41.

103

104 NRC Originating Organization Index (Staff Reports)

NUREG-0750 V40 NO2: NUCLEAR REGULATORY COMMISSION IS-NUREG-1486: FINAL SAFETY EVALUATION REPORT TO LICENSE SUANCES FOR AUGUST 1994. Pages43-132 THE CONSTRUCTION AND OPERATION OF A FACILITY TO NUREG4750 V40 NO3. NUCLEAR REGULATORY COMMISSION IS RECEIVE. STORE AND DISPOSE OF 11E(2) BYPRODUCT MATERI-SUANCES FOR SEPTEMBER 1994 Pages 133145.

AL NEAR CLlVE. UTAH Docket No. 40 8989(Envrocare of Utah.Inc)

NUREG 0750 V40 N04. NUCLEAR REGULATORY COMMISSION IS.

OlVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207)

SUANCES FOR OCTOBER 1994. Pages 147167.

NUREG-1491: SAFETY EVALUATION REPORT FOR THE CLAl-NUREG 0936 V13 NO2: NRC REGULATORY AGENDA Ouartetty BORNE ENRICHMENT CENTER, HOMER.LOUlSiANADocket No.

Report.Apni June 1994.

70-3070, Louisiana Energy Sennces.L.P.

NUREG-1497: INTERIM LICENSING CRITERIA FOR PHYSICAL PRO-EDO - OFFICE OF THE CONTROLLER (PRE 820410 & POST 890205)

TECTION OF CERTAIN STORAGE OF SPENT FUEL G7FICE OF THE CONTROLLER (POST 890205)

NUREG 1504. REVIEW CRITERIA FOR THE PHYSICAL FITNESS NUREG 1470 V03. FINANCIAL STATEMENT FOR FISCAL YEAR TRAINING REQUIREMENTS IN 10 CFR PART 73.

1993.

NUREG 1475: APPLYING STATISTICS.

OPERATIONS BRANCH DIVISION OF BUDGET & ANALYSIS (POST 890205)

NUREG 0525 V02 R02: SAFEGUARDS

SUMMARY

EVENT LIST NUREG-1100 V10: BUDGET ESTIMATES Ftscal Year 1995.

(SSEL) January 1.1990 Through December 31,1993.

NUREG-1350 V06: NUCLEAR REGULATORY COMMISSION INFOR-DIVISION OF WASTE MANAGEMENT (NMSS 940403)

MATION DIGEST.1994 Edition.

NUREG 1323 R00: LICENSEE APPUCAllON REVIEW PLAN FOR A GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL HIGH-LEVEL RADIOACTIVE WASTE.

DATA NUREG 1494: STAFF TECHNICAL POSITION ON CONSIDERATION OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA. DI-OF FAULT DISPLACEMENT HAZARDS IN GEOLOGIC REPOSI-NU G 6

RE TO CONGRESS ON ABNORMAL NUR G 1 95.

ERALL REVIEW STRATEGY FOR THE NUCLEAR NUREG-0090 V16 N04 FhEPORT TO CONGRESS ON ABNORMAL REGULATORY COMMISSION'S HIGH-LEVEL WASTE REPOSI-TORY PROGRAM.

OCCURRENCES October-Decernber 1993 NUREG 1508. DRAFT ENVIRONMENTAL IMPACT STATEMENT TO NUREG-0090 V17 NOI: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. January-March 1994 CONSTRUCT AND OPERATE THE CROWNPOINT SOLUTION NUREG-0090 V17 NO2; REPORT TO CONGRESS ON ABNORMAL MINING PROJECT.CROWNPOINT, NEW MEXICO. Docket No. 40-OCCURRENCES Apr4-June 1994.

8968 (Hydro Resources.Inc.)

NUREG 1022 RO1 DR FC: EVENT REPORTING GUIDELINES 10CFR50 72 AND 50 73 Second Draft For Comment.

U.S. NUCLEAR REGULATORY COMMfSSION NOREG-1272 V08 N01: OFFICE FOR ANALYSIS AND EVALUATION OFFICE OF THE INSPECTOR GENERAL (POST 890417)

OF OPERATIONAL DATA.1993 Annual R

- Power Reactors.

NUREG-1415 V06 NO2: OFFICE OF THE INSPECTOR DIVISION OF OPERATIONAL ASSESSMENT (

T 870413)

GENERAL. Semiannual Report. October 1.1993 - March 31,1994.

NUREG-1471: CONCEPT OF OPERATIONS WITH ORGANIZATION NUREG-1415 V07 No t.

OFFICE OF THE INSPECTOR NE ANCH INCL RE NUREG/CR-5247 VD1 R2; RASCAL VERSION 2.1 USER'S GUIDE.

EDO - OFFICE OF NUCLEAP. FEGULATORY RESEARCH (POST 820405)

K OFFICE OF NUCLEAR REGULATORY RESEARCH (860720 941217)

DI i OF S GRA S ST 04 3 NUREG-1266 V08 NRC bAFETY RESEARCH IN SUPPORT OF REG-NUREG 1275 V10' OPERATING EXPERIENCE F EDBACK REPORT -

^

NUR /CR 5569 0 HEALTH PHYSICS POSITIONS DATA BASE.

TN MPS.

Rea DIVISION OF ENGINEEHING (870413-941217)

TRENDS & PATTERNS ANALYSIS BRANCH NUREG-1426 V02: CCMPILATION OF REPORTS FROM RESEARCH NUREG/CR-6116 V01: SYSTEMS ANALYSIS PROGRAMS FOR HANDSON INTEGRATED RELIABluTY EVALUATIONS (SAPHIRE)

SUPPORTED BY THE MATERIALS ENGINEERING BRANCH, DIVISION OF ENGINEERING 1991 1993.

VERSION 5 0. Technical Reference Manual.

NUREG/CR-6116 V02: SYSTEMS ANALYSIS PROGRAMS FOR ELECTRICAL & MECHANICAL ENGINEERING BRANCH (870717 HANDS-ON INTEGRATED REUABILITY EVALUATONS (SAPHIRE) 941217)

NUREG/CP4138: PROCEEDINGS OF WORKSHOP 1 IN ADVANCED VERSION 5.0 integrated Rehability And Risk Analysm System TOPICS IN RISK AND RELIABILITY ANALYSIS Model Uncertainty-(IRRAS) Reference Manual Its Characteruation And Quantification.

EDO OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM DIVISION OF REGULATORY APPUCATIONS (870413-941217)

(POST e61100)

NUREG 1307 R04:

REPORT ON WASTE BURIAL OFFICE OF INFORMATON RESOURCES MANAGEMENT (POST CHARGES. Escalation Of Decommisseorung Waste Disposal Costs At i

890205)

Low-Level Waste Bural Facilities.

(

NUREG4020 V18. LICENSED OPERATING REACTORS STATUS NUREG-1492 DFC: REGULATORY ANALYSIS ON CRITERIA FOR

SUMMARY

REPORT. Data As Of December 31,1993(Gray Book 1)

THE RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE NUREG-1460 R01: GUIDE TO NRC REPORTING AND RECORD-MATERIAL. Draft Report For Comment.

l KEEPING REQUIREMENTS. Compiled From Requrements in Title NUREG-1496 V1 DFC: GENERIC ENVIRONMENTAL lMPACT STATE-10 Of The U S. Code Of Federal Regulations As Codifed On Decem-MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-bor 31,1993.

RIA FOR DECOMMISS40NING OF NRC-UCENSED NUCLEAR NURE 1 96 2 C NER I I

' IMPACT STATE-8 OF NUCL RMT IAL EY AEU MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-NUREG4430 V13.

LICENSED FUEL FACluTY STATUS RIA FOR DECOMMISSIONING OF NRC-LICENSED NUCLEAR REPORT. inventory Difference Data. July 1,

1992 June 30, FACILITIES. Appendices. Draft Report For Comment.

1993 (Gray Book fl)

DfVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST NUREG 1500: WORKING DRAFT REGULATORY GUIDE ON RE-LEASE CRITERIA FOR DECOMMISSIONING: NRC STAFF'S DRAFT 870729)

NUREG4383 V01 R17: DIRECTORY OF CERTIFICATES OF COMPU-FOR COMMENT.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC NUFiEG 1501 DRFT: BACKGROUND AS A RESIDUAL RADIOACTIV.

ITY CRITERION FOR DECOMMISSIONING. Appendix A To The Draft Approved Packages NUREG-0383 V02 R17: DIRECTORY OF CERTIFICATES OF COMPU-Genenc Envronmental Impact Statement in Support Of Rulemaking ANCE FOR RADIOACTIVE MATERIALS PACKAGESCer1Acates Of On Radiological Cnterm For Decommesseorung Of NRC..

Comphance.

DIVISON OF SAFETY ISSUE RESOLUTION (880717 941217)

NUREG4383 V03 R14. DIRECTORY OF CERTIF6 CATES OF COMPU-NUREG4933 S17: A PRIORITIZATION OF GENERIC SAFETY ANCE FOR RADCACTIVE MATERIALS PACKAGES Report Of NRC ISSUES Approved Quakty Assurance Programs For Radioactrve Matenals DIVISION OF SYSTEMS RESEARCH (880717 941217)

NUREG 1502: ASSESSMENT OF DATABASES AND MODEUNG CA.

(

Packages.

DIVISON OF LOW-LEVEL WASTE MANAGEMENT & DECOMMISSION-PABILITIES FOR THE CANDU 3 DESIGN.

ING (870413-940402 HUMAN FACTORS BRANCH (880717-941217)

NUREG-1200 R03: STANDARD REVIEW PLAN FOR THE REVIEW OF NUREG/CP 0136. PROCEEDINGS OF THE DIGITAL SYSTEMS REll-A UCENSE APPUCATION FOR A LOW-LEVEL RADIOACTIVE ABILITY AND NUCLEAR SAFETY WORKSHOP. September 13-14 WASTE DISPOSAL FACILITY.

1993.Rockville Crowne Plaza Hotel Rockville. Maryland.

l l

l NRC Originating Organization index (Staff Reports) 105 NUREG/CR-5680 VO1: THE IMPACT OF ENVIRONMENTAL CONDI.

NUREG-1462 V02: FINAL SAFETY EVALUATION REPORT RELATED TIONS ON HUMAN PERFORMANCE. A Handbook Of Envronmental TO THE CERTIFICATION OF THE SYSTEM 80+ DESIGN. Chapters Exposures.

15-22 And Appendices. Docket No.52-002.(Asea Brown Boveri.

NUREG/CR-5680 V02: THE IMPACT OF ENVIRONMENTAL COND4-Cornbustion Engineenng)

TONS ON HUMAN PERFORMANCE. A Cntical Review Of The Liter.

NUREG-1503 V01: FINAL SAFETY EVALUATION REPORT RELATED sture.

TO THE CERTIFICATION OF THE ADVANCED BOILING WATER NUREG/CR-6208 AN EMPIRICAL INVESTIGATION OF OPERATOR REACTOR DESIGN. Docket No. 52 001.(General Electnc Nuclear PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED Energy)

EMERGENCIES.

NUREG 1503 V02: FINAL SAFETY EVALUATION REPORT RELATED WASTE MANAGEMENT BRANCH (POST 941217)

TO THE CERTIFICATION OF THE ADVANCED BOfLING WATER NUREG/CR-4918 V07: CONTROL OF WATER INFILTRATON INTO REACTOR DESIGN. Appendices. Docket No. 52 001.(General Elec-NEAR SURFACE LLW DISPOSAL UNITS. Progress Report On Field tnc Nuclear Energy)

Expenments At A Hurmd Region Sete.Beltsville, Maryland.

DIVISON OF REACTOR PROJECTS.1/Il (POST 870411)

NUREG 0847 S13: SAFETY EVALUATION REPORT RELATED TO EDO - OFFICE OF NUCLEAR REACTOR REGULATION POST 800428)

THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 UR 4 SA L

EPORT EL TED TO DIVI OF REA OR ROL HU ACTORS T

THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND T D ECT FOR EbEAC N

1 HUMAN FACTORS ENGINEERING PROGRAM SE RE-NURE 98 S DF FINAL ENVIRONMENTAL STATEMENT RE-NUREG 1021 A07 S01: OPERATOR LICENSING EXAMINER STAND-LATED TO THE OPERATION OF WATTS BAR NUCLEAR PLANT NURE 21478: NON-POWER REACTOR OPERATOR LICENSING EX-UNITS 1 AND 2 Draft Report For Comment. Docket Nos. 50 390 And AMINER STANDARDS.

S39$ennessee Valley Authonty)

PROBABILISTIC SAFETY ASSESSMENT BRANCH OFFICE OF NUCLEAR REACTOR REGULATION, DtRECTOR (POST NUREG/CR-6093: AN ANALYSIS OF OPERATONAL EXPERIENCE 870411)

NUREG4040 V18 NO3: LICENSEE CONTRACTOR AND VENDOR IN-DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-DRESSING HUMAN RELIABILITY ASSESSMENT ISSUES.

SPECTON STATUS REPORT. Quarterty Report, July-September REACTOR SYSTEMS BRANCH 1994 (White Book)

NUREG/CR-3950 V09: FUEL PERFORMANCE REPORT FOR 1991.

PROGRAM MANAGEMENT. POLICY DEVELOPMENT & ANALYSIS DIVISION OF REACTOR INSPECTON & LICENSEE PERFORMANCE STAFF (POST 870411)

(POST 921004)

NUREG-0390 V08 N01: TOPICAL REPORT REVIEW STATUS (Blue NUREG4040 V17 N04: LICENSEE CONTRACTOR AND VENDOR IN-NU 1435 S03: STATUS OF SAFETY ISSUES AT LICENSED 3

e POWER PLANTS.TMt Action Plan Requirements. Unresolved Safety NUREG4)040 VtB N01: LICENSEE CONTRACTOR AND VENDOR IN-AS E

CT FOR D NC D C OR CENSE RE-NEWAL (POST 910918 NUREG 1242 V03 PT01: NRC REVIEW OF ELECTRIC POWER RE-NUREG 0040 V18 NO2: LICENSEE CONTRACTOR AND VENDOR IN-SPECTION STATUS REPORT. Quarterly Report.Apni June SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILITY REQUIREMENTS DOCUMENTS. Passive Plant j

NUREG 4R ISTORICAL DATA

SUMMARY

OF THE SYSTEM-NU 2

0 RC EV1 OF ELECTRIC POWER RE-RAD A O

ION B SEARCH INSTITUTE *S ADVANCED LIGHT WATER REACTOR NUREG/CR-6204: QUESTIONS AND ANSWERS BASED ON RE.

UTILITY REQUIREMENTS DOCUMENT. Passive Plant VISED 10 CFR PART 20.

Dessgns Chapters 213. Project Number 669.

DIVISION OF ENGINEERING (POST 921004) l NUREG-1368: PREAPPLICATON SAFETY EVALUATON REPORT NUREG-1416: OPERATIONAL EXPERIENCE AND MAINTENANCE FOR THE POWER REACTOR INNOVATIVE SMALL MODULE PROGRAMS OF TRANSAMERICA DELAVAL. INC. DIESEL GEN-(PRISM) LOutD-METAL REACTOR. Final Report.

ERATORS.

NUREG 1462 V01: FINAL SAFETY EVALUATON REPORT RELATED NUREG-1488: REVISED LIVERMORE SEISMIC HAZARD ESTIMATES l

TO THE CERTIFICATION OF THE SYSTEM 80+ OESIGN Chapters FOR SIXTY-NINE NUCLEAR POWER PLANT SITES EAST OF THE i

114. Docket No.52-002. (Asea Brown Boven-Combustion Ergneer-ROCKY MOUNTAINS.Fmat Report.

ing)

NUREG-1511: REACTOR PRES $URE VESSEL STATUS REPORT.

l l

l

~e A.

4m 5

NRC Originating Organization index (international Agreements)

This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically b fices) and then by subsections of these (e.g.,y major NRC organizations (e.g., pr,ogram divisions, branches) where appropnate. Each entry is followed by a NUREG number and title of the report (s). If further information is j

needed, refer to the main citation by NUREG number.

J 4

J l

EDO OFFICE OF NUCLEAR REGULATORY RfSEARCH 820405)

NUREG/lA 0114: ASSESSMENT OF RELAPS/ MOD 3 WITH THE LOFT NU EG IA-009 M

AS CAL L ON SE WT M L PL FA UR S OF SAFETY AND REUEF VALVE DISCHARGE PIPtNG HYDRODY.

NAMIC LOADS.

p 107

j

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program office) cnd then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) presared by that organi-zation. If further information is needed, refer to the main citation by the 9UREG/CR number.

1 EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR-4816 R02: PR-EDB: POWER REACTOR EMBRITTLE-DATA MENT DATA BASE. VERSION 2. Program Desenption.

INCOENT RESPONSE BRANCH NUREG/CR-5128 R01: EVALUATION AND REFINEMENT OF LEAK-NUREG/CR 5247 V01 R2: RASCAL VERSION 2.1 USER'S GUIDE-RATE ESTIMATON MODELS WORKBOOK.

NUREG/CR 5161 V02: EVALUATION OF SAMPUNG PLANS FOR IN-DI IS OF F

GRA P ST 04 3)

NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE SERVICE INSPECTION OF STEAM GENERATOR CORE DAMAGE ACCIDENTS: 1992 A STATUS REPORT. Main TUBES Comprehensive Analytscal And Monte Carlo Sanulation Re-suits For Several Sarnphng Plans.

Report And Appendix A' PRECURSORS l

NUREG/CR-4674 V18:

TO POTENTIAL SEVERE NUREG/CR-5314 VOS: INSIGHTS FOR AGING MANAGEMENT OF l

CORE DAMAGE ACCIDENTS:

1992 A

STATUS MAJOR LWR COMPONENTS METAL CONTAINMENTS.

I REPORT.Apperdces B, C. D. E. F, And G NUREG/CR-5359: REVIEW OF ELASTIC STRESS AND FATIGUE-TO-l NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RE-CORE DAMAGE ACCOENTS: 1993 A STATUS REPORT Main LATION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER Report And Appereces A D.

PIPING SYSTEMS.

NUREG/CR-4674 V20- PRECURSORS TO POTENTIAL SEVERE NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DEGRADED CORE DAMAGE ACCIDENTS 1993 A

STATUS SHEAR WALL STlFFNESSES ON SEISMIC PLANT RISK AND SEIS-REPORT.Appendences E And F.

MIC DESIGN LOADS.

NUREG/CR4245: ASSESSMENT OF PRESSURIZED WATER REAC-NUREG/CH-5591 V02 N1; HEAVY-SECTION STEEL IRRADIATON TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING.

PROGRAM Serniannual Progress Report For October 1990 - March 1991 EDO - OFFICE OF INFORMATON RESOURCES MANAGEMENT & ARM NUREG/CR-5591 V02 N2: HEAVY SECTION STEEL IRRADIATION O ICE O FORMATON RESOURCES MANAGEMENT (POST NURE / 58 2 AG G AG I

A PO Fi N RE / R-2907 V12: RADtOACTIVE MATERIALS RELEASED F90M PLANTS.Insaghts From NRC Maintenance Team inspecten Reports.

NUCLEAR POWER PLANTS Annual Repor11991.

NUREG/CR-5861: CRACK-SPEED RELATONS INFERRED FROM LARGE SINGLE EDGE-NOTCHED SPECIMENS OF A 533 B STEEL.

EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CR-5904: FUNCTIONAL ISSUES AND ENVIRONMENTAL DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST QUALIFICATION OF DIGITAL PROTECTON SYSTEMS OF AD-870729)

VANCED LIGHT WATER NUCLEAR REACTORS.

NUREG/CR 5403. PREDICTING THE PRESSURE DRIVEN FLOW OF NUREG/CR-5935:

SUMMARY

OF WORK COMPLETED UNDER THE GASES THROUGH MICRO-CAPI' ARIES AND MICRO-ORIFICES.

ENVIRONMENTAL AND DYNAMPC EQUIPMENT OU/.LIFICATION l

NUREG/CR-6074 V01: SEALED.nOURCE AND DEVICE DESIGN RESEARCH PROGRAM (EDOP).

SAFETY TESTING.Techrucal Report On The Fudngs Of Task 1.0c-NUREG/CR-5939: THE EFFECTS OF AGE ON NUCLEAR POWER tober 1991 - April 1993-PLANT CONTAINMENT COOLING SYSTEMS NUREG/CR-6088.

SUMMARY

OF 1991 1992 MISADMINISTRATION NUREG/CR 5941: TECHNICAL BASIS FOR EVA!UATING ELECTRO-NU 01 OF SCALE: A PC INPUT PROCESSOR FOR AFETY RELAT ISC SY EMS THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For NUREG/CR-5963: CONTINUOUS AE CRACK MONITORING OF A NUR G CR 82 FSCALE: A PC INPUT PROCESSOR FOR DISSIMILAR METAL WELDMENT AT LIMERICK UNIT 1.

THE SCALE CODE SYSTEM. Volurne 2: The ORIGNATE Processor NUREG/CRM DEVEMENT AND ANATON OF DEGRA-for ORIGEN-S DATON MODEUNG TO DEFINE MAINTENANCE PRACTICES.

DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207)

NUREG/CR-5985. EVALUATION OF COMPUTER-BASED ULTRA.

NUREG/CR4149: APPLICATONS OF FIBER OPTICS IN PHYSICAL SONIC INSERVICE INSPECTION SYSTEMS.

PROTECTION NUREG/CR-5990: THE EFFECTS OF SOLAR GEOMAGNETICALLY I

DIVISION OF WASTE MANAGEMENT (NMSS 940403)

INDUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR NUREG/CR-5919: REPOSITORY OPERATONAL CRITERlA COM-POWER STATONS.

PARATIVE ANALYSIS NUREG/CR4051: EFFECTS OF TENSILE LOADING ON UPPER NUREG/CR4232: ASSESSING THE ENVIPONMENTAL AVAILABIL-SHELF FRACTURE TOUGHNESS.

ITY OF URANIUM IN SOILS AND SEDIMENTS NUREG/CR4076. TR EDB. TEST REACTOR EMBRITTLEMENT DATA BASE. VERSION 1.

EDO - OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

NUREG/CR4066 SELECTED FAULT TESTING OF ELECTRONIC NUR /CR 3 0 E YIA VESTIGATONS OF THE ISOLATION DEVICES USED IN NUCLEAR POWER PLAN 1 OPER.

W 0-INDIANA REGION Final Report. October 1986-NUR CR4087. THE EFFECTS OF AGING ON BOluNG WATER NUREG/CR-4219 'V10 N1: HEAVY SECTION STEEL TECHNOLOGY REACTOR CORE ISOLATION COOUNG SYSTEMS.

PROGRAM.Sermannual Progress Report For October 1992 March NUREG/CR4103. PRIORITIZATON OF REACTOR CONTROL COM-1993 PONENTS SUSCEP11BLE TO FIRE DAMAGE AS A CONSE-NUREG/CR-4513 ROI: ESTIMATION OF FRACTURE TOUGHNESS QUENCE OF AGING.

OF CAST STAINLESS STEELS DURING THERMAL AGING IN LWR NUREG/CR4104. SHEAR WALL ULTIMATE DRIFT UMITS.

l SYSTEMS.

NUREG/CR4121: COMPONENT EVALUATON FOR INTERSYSTEM i

NUREG/CR 4599 V03 N2: SHORT CRACKS IN PIPING AND PfPING LOSS-OF-COOLANT ACCOENTS IN ADVANCED UGHT WATER l

WELDS Sermannual Report. October 1992 - March 1993.

REACTORS.

NUREG/CR4667 V17; ENVIRONMENTALLY ASSISTED CRACKING NUREG/CR4132 BIAXIAL LOADING AND SHALLOW-FLAW EF-IN UGHT WATER REACTORS. Sermannual Report.Apnl 1993 Sep.

FECTS ON CRACK TIP CONSTRAINT AND FRACTURE TOUGH-ternber 1993 NESS.

109 l

110 NRC Contract Sponsor Index l

NUREG/CR4139. CRACK ARREST TESTS ON TWO IRRADIATED NUREG/CR 6102: VALIDATON OF THE SCALE BROAD STRUC-l l

HIGH-COPPER WELDS. Phase IL Results Of Duplex Type Spece TURE 44 GROUP ENDF/B Y CROSS-SECTION LIBRARY FOR USE i

rnent.

IN CRITICAUTY SAFETY ANALYSES.

l NUREG/CR4142: TENSILE PROPERTY CHARACTER 12ATON OF NUREG/CR4112 DRF FC: IMPACT OF REDUCED DOSE LIMITS ON THERMALLY AGED CAM STAINLESS STEELS.

NRC LICENSED ACTIVITIES. Mapr issues in The implementaten Of NUREG/CR4151: FdASIBluTY OF DEVELOPtNG RISK-BASED ICRP/NCRP Dose Lmt Recornrnendatons Draft Report For Corn-RANKINGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERV.

rnent.

4 ICE INSPECTON.

NUREG/CR-6114 V03 PERFORMANCE ASSESSMENT OF A HYPO.

l NUREG/CR4154 V01: EXPERIMENTAL RESULTS FROM CONTAIN.

THETICAL LOW-LEVEL WASTE FACluTY. Groundwater Flow And i

MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT Transport Sanulation CONDITIONS.Results From Bellows Tested in " Uke-New" Condi-NUREG/CR4120- CONTROLLED FIELD STUDY FOR VALIDATON tions OF VADOSE ZONE TRANSPORT MODELS NUREG/CR4162: EFFECTS OF PRIOR DUCTILE TEARING ON NUREG/CR4138: USER'S GUIDE FOR SIMPLIFIED COMPUTER CLEAVAGE FRACTURE TOl>GHNESS IN THE TRANSITION MODELS FOR THE ESTIMATON OF LONG TERM PERFORM-REGION.

ANCE OF CEMENT BASED MATERIALS.

NUREG/CR4169. RELAY TEST PROGRAM.Senes 11 Testsintegral NUREG/CR4147 V01: CHARACTER 12ATION OF CLASS A LOW-Testing Of Relays And Creust Breakers.

LEVEL RADIOACTIVE WASTE 1986-1990 Executrve Summary.

NUREG/CR4181: A PILOT APPLICATON OF RISK-BASED METH-NUREG/CR4147 V02: CHARACTERl2ATION OF CLASS A LOW-ODS TO ESTABLISH INSERVICE INSPECTON PRIORITIES FOR LEVEL RADIOACTIVE WASTE 1986-1990.Masn Report Part A.

NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER NUREG/CR4147 V03: CHARACTERIZATION OF CLASS A LOW-STATION LEVEL RADIOACTIVE WASTE 19861990. Main Report-Part B.

NUREG/CR4183: PEER REVIEW OF THE TMi-2 VESSEL INVESTl-NUREG/CR-6147 V04: CHARACTERIZATON OF CLASS A LOW-GATION PROJECT METALLURGICAL EXAMINATONS.

LEVEL RADIOACTIVE WASTE 19861990. Appendices A-E.

NUREG/CR4196: TMi-2 INSTRUMENT NOZZLE EXAMINATIONS NUREG/CR4147 V05: CHARACTER 12ATON OF CLASS A LOW-PERFORMED AT THE INEL.

LEVEL RADIOACTIVE WASTE 19661990. Appendix F.

NUREG/CR4205. VALVE ACTUATOR MOTOR DEGRADATION NUREG/CR4147 V06: CHARACTERIZATION OF CLASS A LOW-NUREG/CR4200: TRANSPORT CALCULATONS OF RADIATON EX-LEVEL RADIOACTIVE WASTE 1986-1190. Appendices G-J.

POSURE TO VESSEL SUPPORT STRUCTURES IN THE TROJAN NUREG/CR4147 V07. CHARACTERIZATION OF CLASS A LOW-REACTOR-LEVEL RADIOACTIVE WASTE 19861990 Appendices K P.

NUREG/CR4209:

MEMPHIS AREA REGIONAL SEISMIC NUREG/CR4156;

SUMMARY

OF COMMENTS RECEIVED FROM NETWORK Final Report. October 1986 - Septernber 1992.

WORKSHOPS ON RADIOLOGICAL CRITERIA FOR DECOMMIS-NUREG/CR4221:THE VALLES NATURAL ANALOGUE PROJECT.

SiONING.

NUREG/CR4223 REVIEW OF THE PROPOSED MATERIALS OF NUPEG/CR4164. RELEASE OF RADIONUCLIDES AND CHELATING CONSTRUCTON FOR THE SBWR AND AP600 ADVANCED REAC-AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATION LOW-TORS LEVEL RADIOACTIVE WASTE COLLECTED FROM THE PEACH NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE BOTTOM ATOMIC POWER STATON UNIT 3.

STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPING NUREG/CR 6174 V1 DFC: REVISED ANALYSES OF DECOMMIS-AT LWR TEMPERATURES.

SIONING FOR THE REFERENCE BOILING WATER REACTOR NUREG/CR4228. PRELIMINARY ASSESSMENT OF THE FRACTURE POWER STATION Effects Of Current Regulatory And Other Consad-BEHAVOR OF WELD MATERIAL IN FULL THICKNESS CLAD erations On The Financial Assurance Requnements Of The Decom-NU 41 5FC REVISED ANALYSES OF DECOMMIS-NU E /CR4231: A COMPARISON OF THE RELATIVE IMPOR-TANCE OF COPPER PRECIPITATES AND POINT DEFECT CLUS-SiONING FOR THE REFERENCE BOluNG WATER REACTOR i

POWER STATON Effects Of Current Regulatory And Other Consed-ll INER.

N E /CR V : TA I KED UN orations On The Financial Assurance Requirements Of The Decom-TIAL STRESSES Subtask 1.1 Final RW NUREG/CR4234: VALIDATION OF ANALYSIS METHODS FOR AS-NL E 41 50RATORY CHARACTERIZATON OF ROCK SESSING FLAWED PlPING SUBJECTED TO DYNAMIC LOADING.

NURE / 4 236: SEISMIC INVESTIGATONS OF THE HDR SAFETY NU EG/C!R4188 V01: MICROBIAL DEGRADATION OF LOW-LEVEL RADIOACTIVE WASTE. Annual Report For FY 1991 NUREG/CR4 N L GUOELINES FOR ASEISMIC DESIGN NUREG/CR4201: COMPRESSION AND IMMERSION TESTS AND i

OF NUCLEAR POWER PLANTS Translation Of JEAG 4601-1987 LEACHING OF RADONUCLIDES. STABLE METALS, AND CHELAT-l NUREG/CR4249: UNIRRADIATED MATERIAL PROPERTIES OF MID.

ING AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATION LAND WELD WF 70.

NUREG/CR4254. SOUTHERN APPALACHIAN REGIONAL SEIEMIC WASTE COLLECTED FROM NUCLEAR POWER STATONG.

NUREG/CR-6203: VALIDATION STUDIES FOR ASSESSING UN-i NETWORK.

NUREG/CR4255. DESIGN OF AN OPEN ARCHITECTURE SEISMIC SATURATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

MONITORING SYSTEM l

NUREG/CR4258: THE LIOUEFACTION METHOD FOR ASSESSING NUREG/CR4204 OUESTIONS AND ANSWERS BASED ON RE-VISED 10 CFR PART 20.

PALEOSEISMICITY.

NUREG/CR4262: CLEAVAGE BEHAVIORS IN NUCLEAR VESSEL NUREG/CR4212: VALUE OF PUBLIC HEALTH AND SAFETY AC.

TONS AND RADIATON DOSE AVOIDED.

STEELS.

NUREG/CR4290 KEY ANALYSIS SYSTEM USER'S GUlOE. Version NUREG/CR4216: EVALUATON OF ROCK JOINT MODELS AND COMPUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS.

I 2.0.

NUREG/CR4250:

SUMMARY

OF COMMENTS RECEIVED ON ST AFF

(

DtVISON OF REGULATORY APPLICATIONS (870413-941217) ON RE-DRAFT PROPOSED RULE ON RADIOLOGICAL CRITERIA FOR DE-NUREG/CR4409 V05: DATA BASE ON DOSE REDUCTI SEARCH PRCMECTS FOR NUCLEAR POWER PLANTS COMMISSIONING.

NUREG/CR4833: LARGE AREA SELF-POWERED GAMMA RAY NUREG/CR4270 DRF FC: ESilMATING BOluNG WATER REACTOR DETECTOR. Phase 11 Development Of A Source Poset on Monitor For DECOMMISSONING COSTS.A User's Manual For The BWR Cost Use On industnal R aptec Uruts Estimating Computer Program (CECP) Software Draft Report For NUREG/CR-4918 V07:

ROL OF WATER INFILTRATON INTO Comrnent.

NEAR SURFACE LLW DISPOSAL UNITS Progress Report On Field NUREG/CR4288. GEOCHEMICAL INVESTIGATIONS RELATED TO Expenments At A Humid Region Site.Bettswine. Maryland.

THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NU.

NUREG/CR-5229 V06: FIELD LYSIMETER INVESTIGATIONS: LOW-CLEAR WASTE REPOSITORY.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR4289. RECONCENTRATION OF RADIOACTIVE MATERI-FISCAL YEAR 1993. Annual R AL RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH NUREG/CR 5344 R01: REP EMENT ENERGY COST ANALYSIS 10 CFR PART 20 l

PACKAGE (RECAP): USER'S GUIDE.

DIVISION OF SAFETY ISSUE RESOLUT60N (880717 941217)

NUREG/CR4569 RO1: HEALTH PHYSICS POSITIONS DATA BASE.

NUREG/CR4551 V01 R1: EVALUATON OF SEVERE ACCIDENT NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSED DECAY RISKS: METHODOLOGY FOR THE CONTAINMENT. SOURCE l

HEAT GUIDE USING SAS2H/ORIGEN-S DATA.

TERM. CONSEQUENCE, AND RISK INTEGRATION ANALYSES.

I NUREG/CR4965: MODELING FIELD SCALE UNSATURATED FLOW NUREG/CR4836. MICROCOMPUTER APPLICATIONS OF, AND AND TRANSPORT PROCESSES.

MODIFICATIONS TO. THE MODULAR FAULT TREES.

NUREG/CR4063. INTRAVAL PHASE il MODEL TESTING AT THE NUREG/CR-5726, REVIEW OF THE DIABLO CANYON PROBABILIS-LAS CRUCES TRENCH SITE.

TIC RISK ASSESSMENT.

l l

l

NRC Contract Sponsor index 111 NUREG/CR4042: PERSPECTIVES ON REACTOR SAFETY.

Frorn internal Events Dunng M4 Loop Operations.Mann Report NUREG/CR4093: AN ANALYSIS OF OPERATONAL EXPERIENCE (Chapters 14).

DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-NUREG/CR4144 V02PIB: EVALUATON OF POTENTIAL SEVERE DRESSING HUMAN REUABIUTY ASSESSMENT ISSUES ACCOENTS DURING LOW POWER AND SHUTDOWN OPER-NUREG/CR4094: CALCULATONS IN SUPPORT OF A POTENTIAL ATONS AT SURRY, UNIT 1. Analysis Of Core Darnage Frequency I

NU E /

1 SU ARY AELCOR 1.8.2 CALCULATIONS pt 2)

FOR THREE LOCA SEQUENCES (AG,S2D & S30) AT THE SURRY NUREG/CR4144'V02P2: EVALUATION OF POTENTIAL SEVERE AC-NU CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

AT SURRY, UNIT 1. Analysis Of Core Damage Frequency From inter.

nal Events Dunng M4 Loop Operations Appendices A-D.

NU E /CR 1

SYS AN SIS PROGRAMS FOR EWM44 W2m ENAW & NN ME A O S

RING LOW POWER AND SHWN N HANDS-ON INTEGRATED RELIABluTY EVALUATONS (SAPHIRE) j A N A S RY, UNIT LAnahms O Core Dang Requency VERSION 5.0. Integrated Rehablity And Rd Analysis System (IRRAS) Reference Manual.

Frorn Internal Events Dunng M4 Loop Operations. Appendices E NUREG/CR4116 V03: SYSTEMS ANALYSIS PROGRAMS FOR HANDSON INTEGRATED REUABluTY EVALUATONS (SAPHIRE)

N CR 14 2P38: EVALUATION OF POTENTIAL SEVERE l

SON 5 In egr ed Rehatnirty And Risk Anaysis System ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-ATIONS AT SURRY, UNIT 1.Analyss Of Core Damage Frequency NUREG R4116 V04: ' SYSTEMS ANALYSIS PROGRAMS FOR From intemal Events Dunng M4 Loop OperahornAppendices E HANDSON INTEGRATED REUABluTY EVALUATONS (SAPHIRE)

(Sectons E.9-E.16).

)

VERSION 5.0. Systems Analysis And Risk Assessment (SARA) Refer-NUREG/CR4144 V02P4: EVALUATON OF POTENTIAL SEVERE AC-ence Manual CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS NUREG/CR4116 VOS: SYSTEMS ANALYSIS PROGRAMS FOR AT SURRY. UNIT 1.Analyms Of Core Damage Frequency From inter.

HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE) nel Events Dunng M4 Loop Operations. Appendices F-H.

VERSION 5.0. Systems Analyses And Risk Assessment (SARA) Tuto.

NUREG/CR4144 V02P5: EVALUATON OF POTENTIAL SEVERE AC nel Manual.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS i

NUREG/CR4116 V07: SYSTEMS ANALYSIS PROGRAMS FOR AT SURRY, UNIT 1.Analyms Of Core Damage Frequency From inter.

HANDS.ON INTEGRATED REUABILITY EVALUATONS (SAPHIRE) nel Events Durin0 M4 Loop Operations Appendices L VERSON 5.0. Fault Tree Event Tree, And Piping & instrumentation NUREG/CR4144 V03 Pt: EVALUATION OF POTENTIAL SEVERE Diagram (FEP) Editors Reference Manual ACCOENTS DURING LOW POWER AND SHUTDOWN OPER-NURLG/CR4116 V08: SYSTEMS ANALYSIS PROGRAMS FOR ATIONS AT SURRY, UNIT 1.Anatyms Of Core Damage Frequency HANDSON INTEGRATED REUABluTY EVALUATONS (SAPHIRE)

From intemal Fres Dunng M4 Loop Operations.Ma n Report VERSION 5.0.Modols And Results Database (MAR.D) Reference NUREG/CR4144 V03 P2: EVALUATION OF POTENTIAL SEVERE Manual ACCOENTS DURING LOW POWER AND SHUTDOWN OPER-NUREG/CR4143 V02P1A: EVALUATON OF POTENTIAL SEVERE ATONS AT SURRY. UNIT 1. Analysis Of Core Dama0e Frequency ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-From Intamal Fires Dunng Mid-Loop Operations. Appendices.

ATONS AT GRAND GULF, UNIT 1.Analyms Of Core Damage Fre.

NUREG/CR4144 V04: EVALUATON OF POTENTIAL SEVERE ACCI-quency From Internal Events For Plant Operational State 5 Dunng A DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT Refuehng Outage. Sections 19.

SURRY, UNIT 1. Analysts Of Core Damage Frequency From Internal NUREG/GR4143 V02PIB. EVALUATON OF POTENTIAL SEVERE Floods Dunng M4 Loop Operations.

ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-NUREG/CR4144 V05: EVALUATION OF POTENTIAL SEVERE ACCI-ATIONS AT GRAND GULF. UNIT 1.Analyms Of Core Damage Fre.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT quency From Intemal Events For Plant Operational State 5 Dunng A SURRY, UNIT 1.Analyms Of Core Damage Frequency From Seismic Refuehng Outage Section 10.

Events Dunng M4 Loop Operations Main Report.

NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE NUREG/CR4145: VERIFICATION AND VALIDATION OF THE SA-ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-PHIRE VERSION 4.0 PRA SOFTWARE PACKAGE.

ATONS AT GRAND GULF, UNIT 1.Analyms Of Core Damage Fre-NUREG/CR4157: SURVEY AND EVALUATION OF AGING RISK AS.

quency Frorn intemal Events For Plant Operational State 5 Dunng A SESSMENT METHODS AND APPUCATONS.

Refuehng Outage Masn Report.

NUREG/CR4224 DFC: PARAMETRIC STUDY OF THE POTENTIAL NUREG/GR4143 V02PT2: EVALUATON OF POTENTIAL SEVERE FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENER-ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-ATED DEBRIS. Draft For Comment.

ATIONS AT GRAND GULF. UNIT 1.Analyms Of Core Damage Fre-NUREG/CR4237: STATISTICAL ANALYSIS OF FATIGUE STRAIN-quency From intemal Events For Plant Operational State 5 Dunng LIFE DATA FOR CARBON AND LOW-ALLOY STEELS.

Refuehng Outage hitemaL.

PROBABILISTIC RISK ANALYSIS BRANCH (9108%941217)

NUREG/CH4143 V02PT3: EVALUATON OF POTENTIAL SEVERE NUREG/CR4053. COMPARISON OF MACCS USERS CALCULA.

ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-TONS FOR THE INTERNATIONAL COMPARISON EXERCISE ON ATIONS AT GRAND GULF. UNIT 1.Anatyms Of Core Damage Fre-PROBABluSTIC ACCIDENT CONSEQUENCE ASSESSMENT quency From Intemal Events For Plant Operahonal State 5 Dunng A CODES.

Refueling Outage intemaL.

DIVISON OF SYSTEMS RESEARCH (880717 941217)

NUREG/GR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE NUREG/CR-4639 V5R4P2: NUCLEAR COMPUTERf2ED UBRARY ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-FOR ASSESSING REACTOR REUABluTY (NUCLARR) Volume 5:

ATONS AT GRAND GULF, UNIT 1.Analyms Of Core Damage Fre-Data Manual Part 2: Human Error Probabehty (HEP) Data.

quency From internal Events For Plant Operational State 5 Dunng A NUREG/CR-4839 V5R4P3: NUCLEAR COMPUTERIZED UBRARY Refuehng Outage IntemaL.

FOR ASSESSING REACTOR REUABluTY (NUCLARR). Volume 5.

NUREG/CH4143 V03: EVALUATION OF POTENTIAL SEVERE ACCI-Data ManuaLPart 3. Hardware Component Failure Data.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-5535 V06: RELAPS/ MOD 3 CODE MANUAL.Vahdation Of GRAND GULF. UNIT 1.Analyms Of Core Damage Frequency From in.

Numencal Tectwupes in RELAPS/ MOD 3 temal Events For Plant Operational State 5 Dunng A Refuehng NUREG/CR-5535 V07; RELAP5/ MOD 3 CODE MANUALSummanes Outage.

And Reviews Of Independent Code Assessment Rep;rts.

NUREG/CR4143 V04: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TIONS ON HUMAN PERFORMANCE. A Handbook Of Envronmental GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From In-Exposures.

temally induced Floodmg Events For Ptar:t Operational State 5 NUREG/CR-5600 V02: THE IMPACT OF ENVIRONMENTAL CONDI-Dunng a Refuehng-...

TONS ON HUMAN PERFORMANCE. A Cntical Rewow Of The bier-NUREG/CR4143 VOS: EVALUATION OF POTENTIAL SEVERE ACCl-ature.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-5850- ANALYSIS OF LONG TERM STATON BLACKOUT GRAND GULF. UNIT 1.Analyes Of Core Damage Frequency From WITHOUT AUTOMATIC DEPRESSUR12ATON AT PEACH BOTTOM Seismic Events Dunng Mid-Loop Opershons Main Report.

USING MELCOR (VERSION 1.8).

NUREG/CR4144 V02PI A: EVALUATION OF POTENTIAL SEVERE NUREG/CR4908 V01: ADVANCED HUMAN-SYSTEM INTERFACE ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-DESGN REVIEW GUIDELINE. General Evaluation Model, Technical ATIONS AT SURRY. UNIT 1. Analyses Of Core Damage Frequency Development. And Guidehne Descretion.

112 NRC Contract Sponsor index NUREG/CR-5908 V02: ADVANCED HUMAN-SYSTEM INTERFACE NUREG/CR 6208 AN EMPIRICAL INVESTIGATION OF OPERATOR DESGN REVIEW GUIDELINE. Evaluation Procedures And Gude.

PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED hnes For Human F actors Engneenng Revews.

EMERGENCIES NUREG/CR 5960- STEAM EXPLOSIONS: FUNDAMENTALS AND EN.

NUREG/CR4211: INTEGRATED FUEL-COOLANT INTERACTON GERGETIC BEHAVOR (IFCI 6 0) CODE User's Manual NUREG/CR-5994: EMERGENCY DIESEL GENERATOR: MAINTE.

NUREG/CR4213. HIGH-TEMPERATURE HYDROGEN-AIR-STEAM DETONATION EXPERIMENTS IN THE BNL SMALL SCALE DEVEL-NANCE AND FAILURE UNAVAILABILITY, AND THEIR RISK IM-OPMENT APPARATUS.

PACTS NUREG/CR4218: A REVIEW OF THE TECHNICAL ISSUES OF AIR NUREG/CR4044. EXPERIMENTS TO INVESTIGATE DIRECT CON.

INGRESSION DURING SEVERE REACTOR ACCOENTS.

TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF NUREG/CR4267: AIR-WATER SIMULATON OF PHENOMENA OF THE ZON NUCLEAR POWER PLANT IN THE SURTSEY TEST FA-CORIUM DISPERSION IN DIRECT CONTAINMENT HEATING.

26 NU E CR-6075: THE PROBABILITY OF CONTAINMENT FAILURE R PF F R S YIGC BY DIRECT CONTAINMENT HEATING IN ZON TRANSIENT THERMAL HYDRAUUC PHENOMENA NUREG/CR4075 S01: THE PROBABILITY OF CONTAINMENT FAIL.

NUREG/CR 6276. QUALITY MANAGEMENT IN REMOTE AFTER-URE BY DIRECT CONTAINMENT HEATING IN ZON.

LOADING BRACHYTHERAPY' NUREG/CR4077: DATA

SUMMARY

REPORT FOR FISSION PROD-UCT RELEASE TEST VI-6 gDO. OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428)

NUREG/CR-6092: RISK ASSESSMENT FOR THE INTENTONAL DE-OFFICE OF NUCLEAR REACTOR REGULATON (POST 941001)

PRESSUR12ATON STRATEGY IN PWRS.

NUREG/CR4190 V01 R1: PROTECTION AGAINST MALEVOLENT NUREG/CR 6105: HUMAN FACTORS ENGINEERING GUIDANCE USE OF VEHICLES AT NUCLEAR POWER PLANTS.Vetucle Bamer FOR THE REVIEW OF ADVANCED ALARM SYSTEMS.

System Selection Gudance For Blast Protection.

NUREG/CR4122 STAFFING DEC1 DON PROCESSES AND NUREG/CR4190 V02 R1: PROTECTION AGAINST MALEVOLENT ISSUES Case Studes Of Seven U S. Nuclear Power Plants USE OF VEHICLES AT NUCLEAR POWER PLANTS Vetucle Barrer NUREG/CR-6123 AN INTERNATONAL COMPARISON OF COM-System Selection Guidance MERCIAL NUCLEAR POWER PLANT STAFFING REGULATIONS OFFICE OF NUCLEAR REACTOR REGULATON, DIRECTOR (POST AND PRACTICE 1980-1990.

870411)

NUREG/CR4126. COGNITIVE SKILL TRAINING FOR NUCLEAR NUREG/CR-2850 V12: DOSE COMMITMENTS DUE TO RADIOAC-TlVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1990.

POWER PLANT OPERATONAL DECISION. MAKING.

NUREG/CR4176. REVIEW OF ENVIRONMENTAL EFFECTS ON FA-NUREG/CR4127; THE EFFECTS OF STRESS ON NUCLEAR TIGUE CRACK GROWTH OF AUSTENITIC STAINLESS STEELS POWER PLANT OPERATONAL DECISION MAKING AND TRAIN.

NUREG/CR 4177: ASSESSMENT OF THERMAL EMBRITTLEMENT ING APPROACHES TO REDUCE STRESS EFFECTS.

OF T TA E E

NUREG/CR4133: FRAGMENTATION AND OUENCH BEHAVIOR OF CORIUM MELT STREAMS IN WATER.

NUREG/CR4146 LOCAL CONTROL STATIONS HUMAN ENGi-NU E CR 5 3 R0 CODES AND STANDARDS AND OTHER XPER M NTS TO INVESTIGATE DIRECT CON-DIVI I OF OP A ING AC SU T P ST 921004 NUREG CR TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF NUREG/CR4252: LESSONS LEARNED FROM THE THRE MILE THE SURRY NUCLEAR POWER PLANT.

ISLAND UNIT 2 ADVISORY PANEL NUREG/CR4158 IMPUCATIONS FOR ACCIDENT MANAGEMENT DIVISION OF REACTOR CONTROLS & HUMAN FACTORS (POST OF ADDING WATER TO A DEGRADING REACTOR CORE-921004)

NUREG/CR-616&

SUMMARY

OF IMPORTANT RESULTS AND NUREG/CR4095: AGING, LOSS-OF COOLANT ACCIDENT (LOCA),

SCDAP/RELAPS ANALYSIS FOR OECD LOFT EXPERIMENT LP-AND HIGH POTENTIAL TESTING OF DAMAGED CABLES F P-2.

NUREG/CR4278 SURVEY OF INDUSTRY METHODS FOR PRO-l NUREG/CR4166 RISK IMPACT OF TECHNICAL SPECIFICATIONS DUCING HIGHLY REllABLE SOFTWARE.

REQUIREMENTS DURING SHUTDOWN FOR BWRS.

NUREG/CR4294. DESIGN F ACTORS FOR SAFETY CRITICAL SOFT-NUREG/CR4168. DIRECT CONTAINMENT HEATING INTEGRAL EF-WARE.

FECTS TESTS AT 1/40 SCALE IN ZON NUCLEAR POWER PLANT NUREG/CR4303 METHOD FOR PERFORMING DIVERSITY AND DEFENSE lN DEPTH ANALYSES OF REACTOR PROTECTION GEOMETRY.

NUREG/CR4180 HYDROGEN MIXING STUDIES (HMS)-USER'S SYSTEMS.

DIVISION OF SYSTEMS SAFETY & ANALYSIS (POST 921004)

MANUAL.

NUREG/CR-3950 V09 FUEL PERFORMANCE REPORT FOR 1991 NUREG/CR4185' TMl-2 INSTRUMENT NOZZLE EXAMINATIONS AT NUREG/CR-5830: AUXILIARY FEEDWATER SYSTEM RISK-BASED ARGONNE NATIONAL LABORATORY Februa 1991 June 1993 NUREG/CR4187: RESULTS OF MECHANICALTESTS AND SUPPLE-INSPECTION GUIDE FOR THE MCGUIRE NUCLEAR POWER ARY M CROSTR URAL EXAMINATIONS OF THE TMl-2 DIVI N OF RADIATION SAFETY & SAFEGUARDS (POST 92 b

SS W M W THE NUREG/CR-6193 PRIMARY SYSTEM FISSION PRODUCT RELEASE Annual Summan am ed rmance Re-AND TRANSPORT.A State-Of-The-Art Report To The Committee On y 9 NUREG/CR4190 V01: PROTECTON AGAINST MALEVOLENT USE NUR G 6 94 MET L IC AND HARDNESS EXAMINA.

TONS OF TMI 2 LOWER PRESSURE VESSEL HEAD SAMPLES.

NUREG/CR4195. EXAMINATON OF RELOCATED FUEL DEBRIS NU GC 02 R T A A NST MALEVOLENT USE j.g ADJACENT TO THE LOWER HEAD OF THE TMi-2 REACTOR OF VEHICLES AT NUCLEAR POWER PLANTS verucle Barner NU E N6196 CALCULATONS TO ESTIMATE THE MARGIN TO Divifi$[ofhGfN l

Rf G P T91

)

FAILURE IN THE TMI-2 VESSEL NUREG/CR4128 PIPING BENCHMARK PROBLEMS FOR THE ADB/

NUREG/CR4197: TMI-2 VESSEL INVESTIGATION PROJECT INTE-CE SYSTEM 80 + STANDARDIZED PLANT.

GRATION REPORT.

NUREG/CR4161 BUCKLING EVALUATION OF SYSTEM 80+(TM1 NUREG/CR-6200- UNCERTAINTY ANALYSIS OF SUPPRESSION CONTAINMENT.

POOL HEATING DURING AN ATWS IN A BWR-5 PLANT.An Apple-NUREG/CR4281. A SIMPLIFIED LEAK-BEFORE. BREAK EVALUA-caton Of The CSAU Methodology Usmg The BNL Engmeenng Plant TION PROCEDURES FOR AUSTENITIC AND FERRITIC STEEL Analyzer.

PlPING

Contractor index This index lists, in alphabetical order, 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 needed, refer to the main citation by the NUREG/CR number.

ADVANCED SYSTEMS CONCEPTS ASSOCIATES NUREG/CR4187; RESULTS OF MECHANICAL TESTS AND SUPPLE-NUREG/CR4157: SURVEY AND EVALUATON OF AGING RISK AS-MENTARY MICROSTRUCTURAL EXAMINATIONS OF THE TMI2 SESSMENT METHODS AND APPLICATIONS.

LOWER HEAD SAMPLES.

NUREG/CR4223: REVIEW OF THE PROPOSED MATERIALS OF CON-ADVANCED SYSTEMS TECHNOLOGY,INC.

STRUCTION FOR THE SBWR AND AP600 ADVANCED REACTORS.

i NUREG/CR4156:

SUMMARY

OF COMMENTS RECElVED FROM NUREG/CR-6234. VALIDATION OF ANALYSIS METHODS FOR AS-WORKSHOPS ON RADIOLOGICAL CRITERIA FOR DECOMMISSION-SESSING FLAWED PIPING SUBJECTED TO DYNAMIC LOADING.

ING.

NUREG/CR-6236. SEISMIC INVESTIGATIONS OF THE HDR SAFETY NUREG/CR4250

SUMMARY

OF COMMENTS RECEIVED ON STAFF PROGRAM. Summary Report.

DAAFT PROPOSED RULE ON RADIOLOGICAL CRITERIA FOR DE-NUREG/CR4237: STATISTICAL ANALYSIS OF FATIGUE STRAIN-LIFE COMMISSIONING.

DATA FOR CARBON AND LOW-ALLOY STEELS.

AEA TECHNOLOGY ARIZONA STATE UNIV., TEMPE, A2 NUREG/CR4144 V02PI A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-4551 V01 R1: EVALUATON OF SEVERE ACCIDENT CIDENTS DURiNG LOW POWER AND SHUTDOWN OPERATIONS AT RISKS-METHODOLOGY FOR THE CONTAINMENT. SOURCE SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal TERM. CONSEQUENCE. AND RISK INTEGRATION ANALYSES.

Events Dunng M4 Loop Operations Mam Report (Chapters 14).

NUREG/CR4144 V02PID EVALUATION OF POTENTIAL SEVERE AC-ARIZONA, UNIV. OF, TUCSON, AZ CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6063: INTRAVAL PHASE il MODEL TESTING AT THE LAS SURRY UNIT 1 Analysis Of Core Damage Frequency From Internal CRUCES TRENCH SLTE.

Events Dunng Md Loop Operations Man Report (Chapters 712)

NUREG/CR-6120: CONTROLLED FIELD STUDY FOR VALIDATION OF NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCl-VADOSE ZONE TRANSPORT MODELS.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4203: VALIDATON STUDIES FOR ASSESSING UNSATU-SURRY. UNIT 1. Analysis Of Core Damage Frequency From Internal RATED FLOW AND TRANSPORT THROUGH FRACTURED ROCK.

Events Dunng M4 Loop Operations. Appendices A-D NUREG/CR 6144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-ARMY, DEPT. OF, CORPS OF ENGINEERS CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4190 V01: PROTECTION AGAINST MALEVOLENT USE OF SURRY. UNIT 1. Analysis Of Core Damage Frequency From intomal VEHICLES AT NUCLEAR POWER PLANTS. Vehicle Barner System Events Dunng M4 Loop Operations Appendices E (Sections E 1-E 8)

NUREG/CR 6144 V02P38: EVALUATION OF PO1ENTIAL SEVERE AC-S tmg Guidance For Blast Protection.

NUREG/CR-6100 V01 R1: PROTECTION AGAINST MALEVOLENT USE CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT OF VEHICLES AT NUCLEAR POWER PLANTS Vehicle Bamer System SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal Selection Guidance For Blast Protectiort Events Dunng M4 Loop Operations Appendices E (Sections E 9-E.16)-

NUREG/CR4190 V02: PROTECTION AGAINST MALEVOLENT USE OF NUREG/CR4144 V02P4 EVALUATION OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT VEHICLES AT NUCLEAR POWER PLANTS Vetucle Barner System Siting Guidance For Blast Protection.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From intemal NUREG/CR4190 V02 RI: PROTECTON AGAINST MALEVOLENT USE O

NR C

44 VO E L ION PO E lAL SEVERE ACCI-e DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT SURRY. UNIT 1. Analysis Of Core Damage Frequency From internal ATHEY CONSULTING Events Dwmg M4 Loop Operations. Appendices l-NUREG/CR-5247 V01 R2. RASCAL VERSION 51 USER'S GUIDE.

NUREG/CR 5247 V02 R2. RASCAL VERSION 2.1 WORKDOOK.

ALASKA, UNIV. OF, FAIRBANKS, AK NUREG/CR-6221: THE VALLES NATURAL ANALOGUE PROJECT.

BATTELLE HUMAN AFFAIRS RESEARCH CENTERS ARGONNE NATIONAL LABORATORY NUREG/CR-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI-NUREG/CR-4513 ROI: ESTIMATON OF FRACTURE TOUGHNESS OF TIONS ON HUMAN PERFORMANCE. A Handbook Of Environmental T STAINLESS STEELS DURING THERMAL AGING IN LWR SYS-s[5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI-NU NUREG/CR-4667 V17. ENVIRONMENTALLY ASSISTED CRACKING IN TlONS ON HUMAN PERFORMANCE. A Cntical Revww Of The Liters.

T ATER REACTORS Semiannual Report.Apnl 1993 Septem-NL G/CR-6122-STAFFING DECISION PROCESSES AND NUREG/CR-5344 RO1: REPLACEMENT ENERGY COST ANALYSIS ISSUES Case Studies Of Seven U S. Nuclear Power Plants.

PACKAGE (RECAP): USER'S GUIDE.

NUREG/CR-6123: AN INTERNATIONAL COMPARISON OF COMMER-NUREG/CR4133. FRAGMENTATION AND QUENCH BEHAVOR OF CIAL NUCLEAR PCWER PLANT STAFFING REGULATONS AND CORIUM MELT STREAMS IN WATER.

PRACTICE.19801990.

NUREG/CR4142: TENSILE. PROPERTY CHARACTERl2ATION OF THERMALLY AGED CAST STAINLESS STEELS.

BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES NUREG/CR4168: DIRECT CONTAINMENT HEATING INTEGRAL EF-NUREG/CR 4599 V03 N2: SHORT CRACKS IN PIPING AND PIPING FECTS TESTS AT 1/40 SCALE IN ZON NUCLEAR POWER PLANT WELDS. Semiannual Report. October 1992 March 1993.

GEOMETRY.

NUREG/CR-5128 RO1 EVALUATION AND REFINEMENT OF LEAK-NUREG/CR4176: REVIEW OF ENVIRONMENTAL EFFECTS ON FA-RATE ESTIMATON MODELS TIGUE CRACK GROWTH OF AUSTENITIC STAINLESS STEELS NUREG/CR4226: EFFECT OF DYNAMIC STRAIN AGING ON THE NUREG/CR-6177: ASSESSMENT OF THERMAL EMBRITTLEMENT OF STRENGTH AND TOUGHNESS OF NUCLEAR FERRITIC PIPING AT CAST STAINLESS STEELS.

LWR TEMPERATURES.

NUREG/CR4183 PEER REVIEW OF THE TMI-2 VESSEL INVESTIGA-NUREG/CR-6233 V01: STABILITY OF CRACKED PIPE UNDER INER-TON PROJECT METALLURGICAL EXAMINATIONS TIAL STRESSES Subtask 1.1 Foal Report.

NUREG/CR4185: TMb2 INSTRUMENT NOZZLE EXAMINATIONS AT NUREG/CR4234: VALIDATON OF ANALYSIS METHODS FOR AS-ARGGi4NE NATIONAL LABORATORY. February 1991 June 1993.

SESSING FLAWED PIPING SU6JECTED TO DYNAMIC LOADING.

113

i 114 Contractor Index l

BATTELLE MEMORIAL INSTITUTE. PACIFIC NORTHWEST NUREG/CP 0139: TRANSACTIONS OF THE TWENTY SECOND WATER LABORATORY REACTOR SAFETY INFORMATION MEETING.

NUREG/CR-2850 V12: DOSE COMMITMENTS DUE TO RADIOACTIVE NUREG/CR-2907 V12: RADIOACTIVE MATERIALS RELEASED FROM RELEASES FROM NUCLEAR POWER PLANT SITES IN 1990.

NUCLEAR POWER PLANTS. Annual Report 1991.

NUREG/CR-3950 V09: FUEL PERFORMANCE REPORT FOR 1991.

NUREG/CR-4409 V05: DATA BASE ON DOSE REDUCTION RE-i NUREG/CR 5161 V02: EVALUATON OF SAMPLING PLANS FOR IN-SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

i SERVICE INSPECTON OF STEAM GENERATOR NUREG/CR-5726: REVIEW OF THE DIABLO CANYON PROBABILISTIC i

TUBES. Comprehensive Anatytical And Monte Carlo Sunulaton Results RISK ASSESSMENT.

For Several Sarnphng Plans.

NUREG/CR 5812: MANAGING AGING IN NUCLEAR POWER NUREG/CR 5247 V01 R2: RASCAL VERSION 2.1 USER'S GUIDE-PLANTS Insights From NRC Mamtenance Tearn inspection Reports.

NUREG/CR-5680 V01: THE IMPACT OF ENVIRONMENTAL CONDI-NUREG/CR-5850: ANALYSIS OF LONG-TERM STATION BLACKOUT TONS ON HUMAN PERFORMANCE. A Handbook Of Environmental WITHOUT AUTOMATIC DEPRESSURIZATON AT PEACH BOTTOM Exposures.

USING MELCOR (VERSION 1.8).

NUREG/CR-5680 V02: THE IMPACT OF ENVIRONMENTAL CONDI-NUREG/CR-5908 V01: ADVANCED HUMAN-SYSTEM INTERFACE TONS ON HUMAN PERFORMANCE. A Cntical Revww Of The Uters-DESIGN REVIEW GUIDELINE. General Evaluation Model. Technscal tura Development. And Guidelme Desenption.

NUREG/CR-5758 V04: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/CR 5908 V02: ADVANCED HUMAN-SYSTEM INTERFACE INDUSTRY. Annual Summary Of Program Performance Reports CY DESIGN REVIEW GUlOELINE. Evaluahon Procedures And Guedeines NUREG/CR-5830- AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-NURE CR 5 THE OF AGE ON NUCLEAR POWER SPECTION GUIDE FOR THE MCGUIRE NUCLEAR POWER PLANT.

PLANT CONTAINMENT COOLING SYSTEMS NUREG/CR-5963. CONTINUOUS AE CRACK MONITORING OF A DIS.

NUREG/CR-5967: DEVELOPMENT AND APPLICATION OF DEGRADA-TION MODELING TO DEFINE MAINTENANCE PRACTICES.

Y G R 597 R :CO NA NO OTHER GUID.

EGGN WE WGS & W%NWWW h ANCE CITED IN REGULATORY DOCUMENTS DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR j

NUREG/CR-5985: EVALUATION OF COMPUTER-BASED ULTRASONIC INSERVICE INSPECTION SYSTEMS POWER STATIONS.

NUREG/CR4063: INTRAVAL PHASE 11 MODEL TESTING AT THE LAS NUREG/CR-5994: EMERGENCY DIESEL GENERATOR: MAINTENANCE CRUCES TRENCH SITE.

AND FAILURE UNAVAILABILITY, AND THEIR RISK IMPACTS.

NUREG/CR4122: STAFFING DECISION PROCESSES AND NUREG/CR4053 COMPARISON OF MACCS USERS CALCULATIONS ISSUES Case Studes Or Savon U S. Nuclear Power Plants.

FOR THE INTERNATONAL COMPARISON EXERCISE ON PROBABI-NUREG/CR 4123: AN INTERNATIONAL COMPARISON OF COMMER-LISTIC ACCIDENT CONSEQUENCE ASSESSMENT CODES.

CIAL NUCLEAR POWER PLANT STAFFING REGULATONS AND NUREG/CR4086: SELECTED FAULT TESTING OF ELECTRONIC ISO.

PRACTICE 1980-1990.

LATION DEVICES USED IN NUCLEAR POWER PLANT OPERATION.

NUREG/CR4151: FEASIBILITY OF DEVELOPING RISK-BASED RANK.

NUREG/CR4087: THE EFFECTS OF AGING ON BOILING WATER RE-INGS OF PRESSURE BOUNDARY SYSTEMS FOR INSERVICE IN-ACTOR CORE ISOLATION COOLING SYSTEMS.

SPECTON.

NUREG/CR4093: AN ANALYSIS OF OPERATIONAL EXPERIENCE NUREG/CR4174 V1 DFC: REVISED ANALYSES OF DECOMMISSON-DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-ING FOR THE REFERENCE BOILING WATER REACTOR POWER ORESSING HUMAN RELIABILITY ASSESSMENT ISSUES.

STATON Effects Of Current Regulatory And Other Cons dershons On NUREG/CR-6094 CALCULATONS IN SUPPORT OF A POTENTIAL The Fmancial Assurance Requirements Of The Decv..... w g Rule DEFINITION OF LARGE RELEASE.

And-..

NUREG/CR4105: HUMAN FACTORS ENGINEERING GUIDANCE FOR NUREG/CR-6174 V2 DFC: REVISED ANALYSES OF DECOMMISSION-THE REVIEW OF ADVANCED ALARM SYSTEMS.

ING FOR THE REFERENCE BOILING WATER REACTOR POWER NUREG/CR-6112 DAF FC: IMPACT OF REDUCED DOSE LIMITS ON STATION Effects Of Current Regulatory And Other Conssderabons On NRC LICENSED ACTMTIES. Major lasues in The implementahon Of The Fmancial Assurance Requwements Of The Dm.v..-.

v.

g Rule ICRP/NCRP Dose Lmt Recommendahans. Draft Report For Comment.

And...

NUREG/CR4128: PIPING BENCHMARK PROBLEMS FOR THE ABB/

NUREG/CR 6181: A PILOT APPLICATION OF RISK-BASED METHODS CE SYSTEM 80 + STANDARDIZED PLANT.

TO ESTABLISH INSERVICE INSPECTION PRIORITIES FOR NUCLE.

NUREG/CR4144 V02P1A EVALUATON OF POTENTIAL SEVERE AC-AR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATON.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4232: ASSESSING THE ENVIRONMENTAL AVAILABILITY SURRY UNIT 1. Analysis Of Core Damage Frequency From Internal OF URANIUM IN SOILS AND SEDIMENTS.

Events Dunno M4 Loop Operations Main Report (Chapters 14).

NUREG/CR4252: LESSONS LEARNED FROM THE THREE MILE NUREG/CR4144 V02P18. EVALUATON OF POTENTLAL SEVERE AC-ISLAND UNIT 2 ADVISORY PANEL CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4270 DRF FC: ESTIMATING BOILING WATER REACTOR SURRY, UNIT 1.Analyes Of Core Damage Frequency From internal DECOMMISSIONING COSTS.A user's Manual For The BWR Cost Este Events Dunno M4 Loop Opershons Main Report (Chapters 712).

mahng Computer Program (CECP) Software Draft Report For Com-NUREG/CR4144 V02P2. EVALUATON OF POTENTIAL SEVERE ACCI-ment DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4289: RECONCENTRATON OF RADIOACTIVE MATERIAL SURRY UNIT 1. Analysis Of Core Damage Frequency From intamal RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR Events Dunng M4 Loop Operahons Appereces A-D.

PART 20.

NUREG/CR4144 V02P3A: EVALUATON OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT BATTELLE SEATTLE RESEARCH CENTER SURRY, UNIT 1.Arialysis Of Core Damage Frequency From internal NUREG/CR4252: LESSONS LEARNED FROM THE THREE MILE Events Dunng M4 Loop Opershone.Appereces E (Sections E.1-E.8).

ISLAND-UNIT 2 ADVISORY PANEL-P.VREG/CR.6144 V02P30' EVALUATION OF POTENTIAL SEVERE AC-C'OENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT BONNEVILLE POWER ADMINISTRATION NUREG-1475. APPLYING STATISTICS.

SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal Events Dunng M4 Loop Operations.Appereces E (Sections E 9-E.16).

BROOKHAVEN NATIJNAL LABORATORY NUREG/CR4144 V02P4. EVALUATION OF POTENTIAL SEVERE ACCl-NUREG-0711: Hl., MAN FACTORS ENGINEERING PROGRAM REVIEW DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT MODEL SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intamal NUREG/CP 0133 V01: PROCEEDtNGS OF THE TWENTY FIRST Events Dunng M4 Loop Opershons Appereces F H.

WATER REACTOR SAFETY INFORMATION MEETINGiMenary Ses.

NUREG/CR-6144 V02P5. EVALUATION OF POTENTIAL SEVERE ACCI-soon; Advanced Reactor Research; Advanced Control System Technoi.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT ogy: Advanced instrumentation & Contro' Hardware; Human Factors...

SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/CP 0133 V02: PROCEEDINGS OF THE TWENTY-FIRST Events Dunno Mid-Loop Opershons Appendices L WATER REACTOR SAFETY INFORMATION MEETING Severe Acep NUREG/CR-6144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC-dent Research.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CP-0133 V03. PROCEEDINGS OF THE TWENTY-FIRST SURRV, UNIT 1. Analysis Of Core Damage Frequency Frorn internal WATER REACTOR SAFETY INFORMATON MEETING.Pnmary Fres Dung M4 Loop Opershons Man Report.

System Integnty, Agmg Research Products & Applicehons; Structural &

NUREG/CR-6144 V03 P2: EVALUATON OF POTENTLAL SEVERE AC-Seismic Engmeenng. Seismology & Geology CIDENTS DURING LOW POWE':I AND SHUTDOWN OPERATONS AT NUREG/CP-0135: WORKSHOP ON ENVIRONMENTAL QUALIFICATON SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal OF ELECTRIC EQUIPMENT.

Fres Durmg Mid-Loop Operahone Appendices

l Contractor index 115 NUREG/CR41A4 V04: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR-4639 V5R4P3: NUCLEAR COMPUTERIZED UBRARY FOR DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ASSESSING REACTOR RELIABILITY (NUCLARR).Volurne 5: Data SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal Manual.Part 3 Hardware Component Failure Data Floods Dunng MdLoop Operations.

NUREG/CR-5229 V06. FIELD LYSIMETER INVESTIGATIONS: LOW-NUREG/CR4146: LOCAL CONTROL STATONS: HUMAN ENGINEER-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR ING ISSUES AND INSIGHTS FISCAL YEAR 1993. Annual Report NUREG/CR4169: RELAY TEST P60 GRAM.Senes il Tests integral Test-NUREG/CR4535 VO6: RELAP5/ MOD 3 CODE MANUALValidation Of ang Of Relays And Circuit Breakers.

Numencal Techniques in RELAPS/ MOD 3.

NUREG/CR4200- UNCERTAINTY ANALYSIS OF SUPPRESSION POOL NUREG/CR-5535 V07: RELAP5/ MOD 3 CODE MANUALSummanes And HEATING DURING AN ATWS IN A SWR-5 PLANT.An Application Of Reviews Of Independent Code Assessment Reports The CSAU Methodology Using The BNL Ergneenng Plant Analyrer.

NUREG/CR-5935.

SUMMARY

OF WORK COMPLETED UNDER THE NUREG/CR4212 VALUE OF PUBUC HEALTH AND SAFETY ACTIONS ENVIRONMENTAL AND DYNAMIC EQUIPMENT QUAUFICATION RE.

AND RADIATON DOSE AVOIDED SEARCH PROGRAM (EDOP).

NUREG/CR 6213: HIGH-TEMPERATURE HYDROGEN-AIR-STEAM NUREG/CR4088:

SUMMARY

OF 1991 1992 MISADMINISTRATION DETONATION EXPERIMENTS IN THE BNL SMALL SCALE DEVELOP-EVENT INVESTIGATIONS.

MENT APPARATUS NUREG/CR4116 V01: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4241: TECHNICAL GUIDELINES FOR ASEISMIC DESIGN OF NUCLEAR POWER PLANTS. Translation Of JEAG 46011987.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

VERSION 5 0 Tece zal Reference Manual.

NUREG/CR4116 v02: SYSTEMS ANALYSIS PROGRAMS FOR CA E E NR R

AS AND UFETIME MORTAUTY HANDS-ON INfEGRATED RELIABILITY EVALUATONS (SAPHIRE)

RISKS OF RADIATION-INDUCED CANCER Low LET Radiation.

Reference Manual.

CALIFORNIA, UNIV. OF, LOS ANGELES, CA NUREG/CR4116 V03: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR-4918 V07: CONTROL OF WATER INFILTRATION INTO HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NEAR SURFACE LLW DISPOSAL UNITS. Progress Report On Field Ex.

VERSION 5 0. Integrated Reliability And Reak Analysis System (IRRAS) penments At A Humid Region Site.Beltsville. Maryland NURE 4

V04: SYSTEMS ANALYSIS PROGRAMS FOR CALIFORNIA, UNIV. OF, SANTA BARBARA, CA HANDS-ON INTEGRATED RELIABILITY EVALUATONS (SAPHIRE)

NUREG/CR-5960: STEAM EXPLOSIONS: FUNDAMENTALS AND EN-VERSION 5.0. Systems Analysis And Risk Assessment (SARA) Refer.

GERGETIC BEHAVIOR.

ence Manual.

NUREG/CR4075. THE PROBABILITY OF CONTAINMENT FAILURE BY NUREG/CR4116 V05: SYSTEMS ANALYSIS PROGRAMS FOR DIRECT CONTAINMENT HEATING IN ZION HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

CARLOW INTERNATIONAL, INC.

VERSION 5.0. Systems Analysis And Risk Assessment (SARA) Tutonal Manual.

NUREG/CR-5908 V02: ADVANCED HUMAN-SYSTEM INTERFACE NUREG/CR-6116 V07; SYSTEMS ANALYSIS PROGRAMS FOR DESIGN REVIEW GUOELINE. Evaluation Procedures And Guidelines HANDS-ON INTEGRATED REUABlUTY EVALUATONS (SAPHIRE)

For Human Factors Engineenng Heviews VERSION 5.0. Fault Tree. Event Tree, And Piping & Instrumentation CENTER FOR NUCLEAR WASTE REGULATORY ANALYSES G'/

NT M N

6116 08 ALYSIS PROGRAMS FOR NUREG/CR-5919 REPOSITORY OPERATIONAL CRITERIA COMPARA.

HANDS-ON INTEGRATED RELIABluTY EVALUATIONS (SAPHIRE)

Nt EG INTRAVAL PHA3E 11 MODEL TESTING AT THE LAS VERSON 5 0Mo@s And ResuRs DeWse pq Rehence E @CR WOW MMM M WMWW NURE CR 617 L

ATORY CHARACTERl2ATON OF ROCK JOINTS LOSS OF. COOLANT ACCIDENTS IN ADVANCED UGHT WATER RE-NUREG/CR4216 EVALUATON OF ROCK JOINT MODELS AND COM, ACTORS PUTER CODE UDEC AGAINST EXPERIMENTAL RESULTS NUREG/CR-6138 USER'S GUIDE FOR SIMPUFIED COMPUTER NUREG/CR-6288. GEOCHEMICAL INVESTIGATIONS RELATED TO MODELS FOR THE ESTIMATION OF LONG-TERM PERFORMANCE THE YUCCA MOUNTAIN ENVIRONMENT AND POTENTIAL NUCLE.

OF CEMENT BASED MATERIALS.

AR WASTE REPOSITORY.

NUREG/CR4158. IMPLICATIONS FOR ACCIDENT MANAGEMENT OF ADDING WATER TO A DEGRADING REACTOR CORE.

CONSULTING ENGINEER NUREG/CR4160:

SUMMARY

OF iMPORTANT RESULTS AND SCDAP/

NUREG/CR4104 SHEAR WALL U.TIMATE DRIFT UMITS RELAP5 ANALYSIS FOR OECD LOFT EXPERIMENT LP FP-2.

NUREG/CR4164: RELEASE OF RADIONUCUDES AND CHELATING EASTERN RESEARCH GROUP,1NC.

AGENTS FROM CEMENT SOUDIFIED DECONTAMINATION LOW-NUREG/CR4147 V01:CHARACTERIZATON OF CLASS A LOW LEVEL LEVEL RADIOACTIVE WASTE COLLECTED FROM THE PEACH RADIOACTIVE WASTE 1986-1990 Executive Summary BOTTOM ATOMIC POWER STATION UNIT 3 NUREG/CR4147 V02: CHARACTERIZATON OF CLASS A LOW-LEVEL NUREG/CR-6188 V01: MICROBIAL DEGRADATION OF LOW LEVEL RADIOACTIVE WASTE 1986-1990 Main Report-Part A RADIOACTIVE WASTE. Annual Report For FY 1993 NUREG/CR-6147 V03: CHARACTERIZATION OF CLASS A LOW-LEVEL NUREG/CR 6194. METALLOGRAPHIC AND HARDNESS EXAMINA-RADIOACTIVE WASTE 1986-1990 Main Report-Part B TIONS OF TMI 2 LOWER PRESSURE VESSEL HEAD SAMPLES NUREG/CR4147 V04 CHARACTERIZATION OF CLASS A LOW-LEVEL NUREG/CR4195: EXAMINATION OF RELOCATED FUEL DEBRIS AD-RADIOACTIVE WASTE 1986-1990. Appendices A-E.

JACENT TO THE LOWER HEAD OF THE TMI-2 REACTOR VESSEL NUREG/CR4147 V05. CHARACTERIZAllON OF CLASS A LOW-LEVEL NUREG/CR4196. CALCULATIONS TO ESTIMATE THE MARGIN TO RADtOACTIVE WASTE 1986-1990.Appenen F.

FAILURE IN THE TML2 VESSEL NUREG/CR4147 V06: CHARACTERIZATION OF CLASS A LOW LEVEL NUREG/CR 6197: TMI-2 VESSEL INVESTIGATION PROJECT INTE-RADIOACTIVE WASTE 19861190 Appeneces G-J.

GRATON REPORT.

NUREG/CR4147 V07: CHARACTERIZATION OF CLASS A LOW-LEVEL NUREG/CR4198 TMI-2 INSTRUMENT NOZZLE EXAMINATONS PER.

RADIOACTIVE WASTE 1986-1990. Appendices K-P.

FORMED AT THE INEL.

NUREG/CR 6201: COMPRES$10N AND IMMERSON TESTS AND EG4G IDAHO, INC-LEACHING OF RADIONUCUDES. STABLE METALS, AND CHELATING NUREG/CP-0137 V01; PROCEEDINGS OF THE THIRD NRC/ASME AGENTS FROM CEMENT SOLIDIFIED DECONTAMINATION WASTE SYMPOSIUM ON VALVE AND PUMP TESTING Held At The Hyatt Re.

COLLECTED FROM NUCLEAR POWER STATIONS gency Hotel, Washington,DC. July 18-21. 1994 Session 1 A. Session NUREG/CR4245. ASSESSMENT OF PRESSURIZED WATER REAC-2C.

TOR CONTROL ROD DRIVE MECHANISM NOZZLE CRACKING NUREG/CP 0137 V02: PROCEEDINGS OF THE THIRD NRC/ASME NUREG/CR 6276 OUALITY MANAGEMENT IN REMOTE AFTERLOAD.

SYMPOSIUM ON VALVE AND PUMP TESTING Held At The Hyatt Re-ING BRACHYTHERAPY, gency Hotel, Washington.DC. July 18-21,1994 Session 3A -Session 4B EGYPT, GOVT.OF NUREG/CP-0138 PROCEEDINGS OF WORKSHOP l IN ADVANCED NUREG/CR 5726 REVIEW OF THE DIABLO CANYON PROBABILISTIC TOPICS IN RISK AND REUABILITY ANALYSIS Model Uncertainty: Its RISK ASSESSMENT.

Charactenzation And Quantsfication NUREG/CR 4639 V5R4P2; NUCLEAR COMPUTERIZED UBRARY FOR ENERGY, INC.

ASSESSING REACTOR REUABILITY (NUCLARR) Volume 5 Data NUREG/CR-4838 MICFtOCOMPUTER APPUCATIONS OF, AND MODI-Manual Part 2. Human Error Probatnhty (HEP) Data FICATIONS TO. THE MODULAR FAULT TREES.

116 Contractor index ENERGY, DEPT.0F, ENVIRONMENTAL MEASURE.11ENTS LABORATORY K.E.M.P. CORP.

NUREG-1501 DRFT: BACKGROUND AS A RESIDUAL RADIOACTIVITY NUREG/CR-4833. LARCE AREA SELF-POWERED GAMMA RAY CRITERlON FOR DECOMMISSIONING. Appendix A To The Draft Go-DETECTOR. Phase il Deolopment Of A Source Posita Monitor For J

nerc Envronmental impact Statement in Support Of Rulemalmg On Use On industnal Radiogn phe Uruts.

Radiologmal Cntena For Decomrmassorung Of NRC.

KTECH CORP, ENSCO, INC.

NUREG/CR.6044: EGtiRIMENTS TO INVESTIGATE DIRECT CON-NUREG/CR4255: DESIGN OF AN OPEN ARCHITECTURE SEISMIC TAINMENT HEATi 43 PHENOMENA WITH SCALED MODELS OF THE MONITORING SYSTEM.

ZION NUCLEAR POWER PLANT IN THE SURTSEY TEST FACluTY.

NUREG/CR4152: EXPERIMENTS TO INVESTIGATE DIRECT CON-EQE ENGINEERING CONSULTANTS (FORMERLY EOE ENGINEERING, TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THF i

SURRY NUCLEAR POWER PLANT.

NU

/CR-5407: ASSESSMENT OF THE IMPACT OF MRADED SHEAR WALL STIFFNESSES ON SEISMIC PLANT RIS) 1 SE!S.

LAMONT-DOHERTY EARTH OSSERVATORY NUREG/CR4258: THE LOUEFACTION METHOD FOR ASSESSING PA-NUR G 5

E lEW OF THE DIABLO CANYON PRC 3.lSTIC LEOEEMM RISK ASSESSMENT.

LAWREICE LIVERMORE NATIONAL LADORATORY EQE, NdC.

NUREG/CR4143 V05: EVALUATION OF POTENTIAL SEVERE ACCb NUREG/CR-5403: PREDICTING THE PRESSURE DRIVEN FLOW OF G ASES THROUGH MICRO CAPlLLARIES AND MICROGRIFICES.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUEG/CR4278: SURVEY OF INDUSTRY METHODS FOR PRODUC-GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency Frorn N

GC 4

EV LU

)F PO EVERE ACCL N

/R S N FOR SAFETY CRITICAL SOFT.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT WA E REG /CR4303: METHOD FOR PERFORMING DIVERSITY AND DE-SURRY. UNIT 1. Analyses Of Core Damage Frequency From Seisme FENSE-IN-DEPTH ANALYSES OF REACTOR PROTECTON SYS-Events Omng M4 Loop Operations.Mam RW TEMS.

l FLORIDA, UNIV. OF, GAINESVlLLE, FL i

NUREG/CR-5990: THE EFFECTS OF SOLAR-GEOMAGNETICALLY IN-LOS ALAMOS NATIONAL LASORATORY I

DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR NUREG/CR-6104: SHEAR WALL ULTIMATE DRIFT UMITS.

j NUREG/CR4157: SURVEY AND EVALUATION OF AGING RISK AS-POWER STATIONS.

SESSMENT METHODS AND APPUCATIONS.

j FUTURE RESOURCES ASSOCIATES. INC.

NUREG/CR4180- HYDROGEN MIXING STUDIES (HMS): USER'S i

NUREG/CR4143 V05: EVALUATION OF POTENTIAL SEVERE ACCI-MANUAL I

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4269: A PLAN FOR THE MODIFICATION AND ASSESS-1 GRAND GULF, UNIT 1 Analysis Of Core Damage Frequency From MENT OF TRAC-PF1/ MOD 2 FOR USE IN ANALYZING CANDU 3 Seismc Events Dunng MdLoop Operations Mam Report TRANSIENT THERMAL-HYDRAUUC PHENOMENA.

NUREG/CR4144 V05: EVALUATION OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LOUISIANA STATE UNIV., BATON ROUGE, LA SURRY, UNIT 1. Analysis Of Core Damage Frequency Frorn Seisme NUREG/CR4206: TRANSPORT CALCULATIONS OF RADIATION EX-l Events Dunng Mid-Loop Operations.Mam Report.

POSURE TO VESSEL SUPPORT STRUCTURES IN THE TROJAN RE-ACTOR.

GEO CENTERS. INC.

NUREG/CR4107:

SUMMARY

OF MELCOR 1.8.2 CALCULATIONS FOR MARYLAND, UNIV. OF, COLLEGE PARK, MD l

THREE LOCA SEQUENCES (AG.S2D & S3D) AT THE SURRY PLANT.

NUREG/CP-0138: PROCEEDINGS OF WORKSHOP 1 IN ADVANCED TOPICS IN RISK AND REUABluTY ANALYSIS.Model Uncertainty-Its GERMANY, FEDERAL REPUBUC OF Characternation And Quantifcation.

NUREG/CR4236: SEISMIC INVESTIGATIONS OF THE HDR SAFETY NUREG/CR-4918 V07: CONTROL OF WATER INFILTRATON INTO PROGRAM. Summary Report NEAR SURFACE LLW DISPOSAL UNITS. Progress Report On Field Ex-NUR G CR R1: EVALUATION OF SEVERE ACCIDENT RISKS: METHODOLOGY FOR THE CONTAINMENT. SOURCE TERM. CONSEQUENCE, AND RISK INTEGRATON ANALYSES.

MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAM 8 RIDGE, MA NUREG/CR-5965: MODEUNG FIELD SCALE UNSATURATED FLOW IDAHO NATIONAL ENGINEERING LA80RATORY NUREG/CR 5314 V05: INSIGHTS FOR AGING MANAGEMENT OF NUREG/CR4063 INTRA L H SE 11 MODEL TESTING AT THE LAS MAJOR LWR COMPONENTS METAL CONTAINMENTS.

NUREG/CR4075 S01: THE PROBABluTY OF CONTAINMENT FAIL.

CRUCES TRENCH SITE.

NUREG/CR4114 V03: PERFORMANCE ASSESSMENT OF A HYPO-URE BY DIRECT CONTAINMENT HEATING IN ZION.

NUREG/CR4145: VERIFICATION AND VAUDATION OF THE SAPHIRE THETICAL LOW-LEVEL WASTE FACluTY. Groundwater Flow And VERSION 4 0 PRA SOFTWARE PACKAGE.

Transport Simulation.

NUREG/CR4200 UNCERTAINTY ANALYSIS OF SUPPRESSION POOL NUREG/CR4144 V02P1A: EVALUATION OF POTENTIAL SEVERE AC.

HEATING DURING AN ATWS IN A BWR-5 PLANT An Appication Of CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT The CSAU Methodology Usmg The BNL Engmeermg Plant Analyrer.

SURRY, UNIT 1. Analyses Of Core Damage Frequency From Intemal Events Dunng M4 Loop Operations.Mem Report (Chapters 14).

ILLINOIS, UNIV. OF, URSANA, IL NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4133: FRAGMENTATION AND OUENCH BEHAVIOR OF CtDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT CORIUM MELT STREAMS IN WATER.

SURRY, UNIT 1 Analysis Of Core Damage Frequency From intemal NUREG/CR-6162: EFFECTS OF PRIOR DUCTILE TEARING ON CLEAV-Events During M4 Loop Operatiorts Mem Report (Chapters 712)

AGL r AACTURE TOUGHNESS IN THE TRANSITION REGON.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1. Analyses Of Core Damage Frequency From internal U

CR ikE AN S S ST USER'S GUIDE. Version 2.0.

Events Dunng M4 Loop Operations. Appendices A-D.

NUREG/CR4144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-IOWA STATE UNIV AMES,lA CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4161: BUCKUNG EVALUATON OF SYSTEM 80+(TM)

CDNTAINMENT.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From Intemal NUREG/GR4013: APPLICATIONS OF A NEW MAGNETIC MONITOR.

Events Dunng M4 Loop Operations.Appendees E (Sections E.1-E.8)

ING TECHNIQUE TO IN SITU EVALUATION OF FATOUE DAMAGE NUREG/CR4144 V02P38: EVALUATION OF POTENTIAL SEVERE AC.

IN FERROUS COMPONENTS.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY. UNIT 1.Anatysea Of Core Damage Frequency From Intemal JONNS NOPKINS UNIV., BALTIMORE MD Events Dunng M4 Loop Operations Appendees E (Sections E 9.E.16).

NUREG/GR4008: VAUDATION OF SEISMIC PROBABluSTIC RISK AS-NUREG/CR4144 V02P4; EVALUATION OF POTENTIAL SEVERE ACCl-SESSMENTS OF NUCLEAR POWER PLANTS.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT I

l l

\\

\\

Contractor index 117 SURRY, UNIT 1. Analyses Of Core Damage Frequency From internal NUREG/CR-5591 V02 N2: HEAVY-SECTION STEEL IRRADIATON

[

Events Dunng M4 Loop Operatons Appendees F-H PROGRAM Semsannual Progress Report For Aprd-September 1991.

NUREG/CR4144 YO2P5: EVALUATION OF POTENTIAL SEVERE ACCI-NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSED DECAY DENTS DUR:?.J LOW POWER AND SHUTDOWN OPERATIONS AT HEAT GUIDE USING SAS2H/ORIGEN-S DATA.

SURRY. UNIT 1. Analyses Of Core Damage Frequency From intemal NUREG/CP-5861: CRACK SPEED RELATIONS INFERRED FROM Events 9unng M4 Loop Operations Appendees 1.

LARGE SINGLE-EDGE-NOTCHED SPECIMENS OF A 533 D STEEL NUREGiCR4144 V03 Pt: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR 5904: FUNCTONAL ISSUES AND ENVIRONMENTAL CIDiNTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT QUALIFCATION OF DIGITAL PROTECTION SYSTEMS OF AD-SU1RY, UNIT 1. Analysis Of Core Damage Frequency From Intemal VANCED LIGHT WATER NUCLEAR REACTORS.

Faes Dunng M4 Loop Operations.Mam Report.

NUREG/CR-5941: TECHNICAL BASIS FOR EVALUATING ELECTRO-NUREG/CR4144 V03 P2: EVALUATION OF POTENTIAL SEVERE AC-MAGNETIC AND RADIO-FREQUENCY INTERFERENCE IN SAFETY-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT RELATED IAC SYSTEMS.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/CR4076 TR-EDB: TEST REACTOR EMBRITTLEMENT DATA Fires Dunng MeLoop Operations.Appendees.

BASE, VERSION 1.

NUREG/CR4077: DATA

SUMMARY

REPORT FOR FISSON PRODUCT i #GPHIS STATE UNIV, MEMPHtS, TN RELEASE TEST VI-6.

P JREG/CR4209.

MEMPHIS AREA REGIONAL SEISMIC NUREG/CR4102: VALIDATION OF THE SCALE BROAD STRUCTURE NETWORK Final Report, October 1986 - September 1992.

44 GROUP ENDF/B-Y CROSS-SECTON LIBRARY FOR USE IN NUl 'EG/CR4254: C HERN APPALACHIAN REGIONAL SEtSMIC CRITICALITY SAFETY ANALYSES.

NETWORK-NUREG/CR4132-BIAXIAL LOADING AND SHALLOW-FLAW EFFtCTS ON CRACK TIP CONSTRAINT AND FRACTURE TOUGHNESS.

MICHIGAN, UNIT. OF, ANN ARSOR, MI NUREG/CR4139: CRACK-ARREST TESTS ON TWO IRRADIATED NUREG/CR-3 45 VIO: GEOPHYSICAL INVESTIGATIONS OF THE HIGH4OPPER WELDS Phase 11: Results Of Duplex Type Specimens.

WESTERN OFIO-INDIANA REGON Final Report October 19864 NUREG/CR4182 V01: OFFSCALE: A PC INPUT PROCESSOR FOR termer 1992.

THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For MOS,WC.

The Cntcahty Sequences.

NUREG/CR4182 V02: OFFSCALE: A PC INPUT PROCESSOR FOR NUREG/CR-5990: THE EFFECTS OF SOLAR GEOMAGNETICALLY IN-THE SCALE CODE SYSTEM Volume 2: The ORIGNATE Processor for DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR POWER STATONS.

ORIGEN-S NUREG/CR 6193: PRIMARY SYSTEM FISSION PRODUCT RELEASE NATONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERLY AND TRANSPORT.A State Of-The-Art Report To The Commettee On NATIONAL SUREAU OF The Safety Of Nuclear instanations.

NUREG/CP4136: PROCEEDINGS OF THE OtGITAL SYSTEMS REll-NUREG/CR4204: OUESTIONS AND ANSWERS BASED ON REVISED ABILITY AND NUCLEAR SAFETY WORKSHOP. September 13-14 N E/

5 VALVE ACTUATOR MOTOR DEGRADATON.

REG /GM MANMT NMN & MM M N

G CR NS T

'l LT S OF RADIATION EX-POSURE TO VESSEL SUPPORT STRUCTURES IN THE TROJAN RE-U AN ACTOR.

NUREG/CR4228: PRELIMINARY ASSESSMENT OF THE FRACTURE NEW MEXICO STATE UNIV, LAS CRUCES, NM BEHAVIOR OF WELD MATERIAL IN FULL THICKMS CLAD NURE TRAVAL PHASE il MODEL TESTING AT THE LAS N

/CR4231: A COMPARtSON OF THE RELATIVE IMPORTANCE NUREG/CR4120- CONTROLLED FIELD STUDY FOR VALIDATION OF OF COPPER PRECIPITATES AND POINT DEFECT CLUSTERS IN RE.

VADOSE ZONE TRANSPORT MODELS.

ACTOR PRESSURE VESSEL EMBRITTLEMENT.

NUREG/CR4249. UN1RRADIATED MATERIAL PROPERTIES OF MID.

NEW MEXICO, UNIV. OF, ALBUQUEROUE, NM LAND WELD WF 70.

NUREG/CR4042: PERSPECTIVES ON REACTOR SAFETY.

NUREG/CR4262: CLEAVAGE BEHAVORS IN NUCLEAR VESSEL STEELS NOVETECH CORP, NUREG/CR4281: A SIMPUFIED LEAK-BEFORE BREAK EVALUATION OODFN ENVIRONIAENTAL & ENERGY SERVICES (FORMERLY PROCEDURES FOR AUSTENITIC AND FERRITIC STEEL PIPING.

tWLTIPLE DYNA 800CS CORP)

NL REG /CR-5314 V05. INSIGHTS FOR AGING MANAGEMENT OF NUCLEAR POWER ENGINEERING CORP.

16 AJOR LWR COMPONENTS METAL CONTAINMENTS.

NUREG/CR4213-HIGH-TEMPERATURE HYDROGEN-AIR-STEAM DETONATION EXPERIMENTS IN THE BNL SMALL-SCALE DEVELOP-ORGAW7.ATION FOR ECONOtsic COOPERATION & DEVELOPteENT MENT APPARATUS.

NUREGMP.0127: PROCEEDINGS OF THE CSNI SPECIALISTS MEET.

"^

OAK RIDGE NATIONAL LABORATORY NUREG/CR-4219 V10 N1: HEAVY SECTON STEEL TECHNOLOGY PACIFIC-NUCLEAR CO.

PROGRAM Semiannual Progress Report For October 1992 - March NUREG/CR-5535 V06: *tELAP5/ MOD 3 CODE MANUAL.Vahdatson Of NUREG/CR-4674 V17. PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS.1992 A STATUS REPORT. Main Report And PERFORMANCE TECHNOLOGY GROUP,INC.,

Appendix A-NUREu CR4282 RECONCENTRATION OF RADOACTIVE MATERIAL NUREG/CR-4674 V18 PRECURSORS TO POTENTIAL SEVERE CORE RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR DAMAGE ACCIDENTS.1992 A STATUS REPORT. Appendices B, C, D.

PART 20 E. F. And G.

NUREG/CR-4674 V19 PRECURSORS TO POTENTIAL SEVERE CORE ptG, INC. (FOResERLY PICKARD, LOWE A GARRICK,INC.)

DAMAGE ACCIDENTS: 1993 A STATUS REPORT. Man Report And NUREG/CR4144 V02P1A EVALUATON OF POTENTIAL SEVERE AC-NU 6

V20- PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS.1993 A STATUS REPORT.Appendences E SURRY, UNIT 1. Analyses Of Core Damage Frequency From Internal And F.

Events Dunng M4 Loop Operatsons Main Report (Chapters t4).

NUREG/CR 4816 R02: PR-EDB POWER REACTOR EMBRITTLEMENT NUREG/CR4144 V02P18. EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SH'JTDOWN OPERATIONS AT NURE /CR 52 VO 2R V

10SE R'S GUOE.

SURRY. UNIT 1. Analyses Of Core Damage Frequency From Intemal NUREG/CR-5247 V02 R2: RASCAL VERSION 21 WORKBOOK Events Dunng M4 Loop Operations Main Report (Chapters 712).

NUREG/CR-5359 REVIEW OF ELASTIC STRESS AND FATIGUE.TO.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCl-FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER PIPING SURRY,UNtT 1. Analyses Of Core Damage Frequency From Intemal SYSTEMS Events Dunng M4 Loop Operations es A.D.

NUREG/CR-5569 RO1: HEALTH PHYS 8CS POSITIONS GATA S*SE NUREG/CR4144 V02P3A-EVALUAT OF POTENTIAL SEVERE AC.

NUREG/CR-5591 V02 N1: HEAVY SECTON STEEL IRRADIATON CfDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PROGRAM.Senuannual Progresa Report For October 1990 - March SURRY. UNIT 1. Analysis Of Core Damage Frequency From internal 1991.

Events Dunng M4 Loop Operataons.Appendw es E (Sections E.1-E 8)

Contractor index 117 SURRY, UNIT 1. Analysis Of Core Damage Frecpency Frorn internal NUREG/CR-5591 V02 N2. HEAVY SECTION STEEL IRRADIATION Events Dunng M4 Loop Operations Appendices F H PROGRAM Sermannual Progress Report For Apr$ September 1991.

NUREG/CR4144 V02P5. EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR-5625: TECHNICAL SUPPORT FOR A PROPOSED DECAY DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT HEAT GUIDE USING SAS2H/ORIGEN-S DATA.

SURRY. UNIT 1. Analysis Of Core Damage Frequency From internal NUREG/N61: CRACK SPEED RELATONS INFERRED FROM Events Dunng M4 Loop Operations.Apper6ces 1.

LARiE SINGLE EDGE NOTCHED SPECIMENS OF A 533 8 STEEL NUREG/CR4144 V03 P1: EVALUATION OF POTENTIAL SEVERE AC.

NURE01/CR 5904: FUNCTIONAL ISSUES AND ENVIRONMENTAL CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT OU, LIFICATION OF DIGITAL PROTECTON SYSTEMS OF AD-SURRY, UNIT 1. Analysis Of Core Da nege Frequency From inteTial VANCED UGHT WATER NUCLEAR REACTORS.

i Fres Dunno M4 Loop Operations Main Report.

NUREGnR-5941: TECHNICAL BASIS FOR EVALUATING ELECTRO-NUREG/CR4144 V03 P2-EVALUATON OF POTENTIAL SEVERE AC.

MAGNE'IC AND RADIO. FREQUENCY INTERFERENCE IN SAFETY-l CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RELATEt; l&C SYSTEMS.

SURRY, UNIT 1. Analyses Of Core Damage Frequancy From Internal NUREG/CH-6076: TR-EDB: TEST REACTOR EMBRITTLEMENT DATA Fres Dunng M4 Loop Operations. Appendices.

BASE, VERSION 1.

NUREG/CR4077: DATA

SUMMARY

REPORT FOR FISSION PRODUCT MEMPHIS STATE UNIV., MEMPHIS, TN RELEASE TEST Vl4.

NUREG/CR4209:

MEMPHIS AREA REGIONAL SEISMC NUREG/CR4102: VALIDATION OF THE SCALE BROAD STRUCTURE NETWORK Final A

, October 1906 Ceptomber 1992.

44-GROUP ENDF/B-Y CROSS-SECTION LIBRARY FOR USE IN

,i NUREG/CR4254:

HERN APPALACHIAN REGIONAL SEISMIC CRITICALITY SAFETY ANALYSES.

NETWORK.

NUREG/CR4132: BIAXIAL LOADING AND SHALLOW.FuW EFFECTS ON CRACK TIP CONSTRAINT AND FRACTURE TOUGHNESS MICHIGAN, UNIV. OF, ANN AR80R, MI NUREG/CR4139. CRACK-ARREST TESTS ON TWO IRRADIATED NUREG/CR-3145 VIO: GEOPHYSCAL INVESTIGATONS OF THE HIGH-COPPER WELDS. Phase 11: Results Of Duplex-Type Specimens.

l WESTERN OHIO-INDIANA REGION Final Report. October 1986-Sep.

NUREG/CR4182 V01: OFFSCALE: A PC INPUT PROCESSOR FOR tembw 1992.

THE SCALE CODE SYSTEM. Volume 1: The CSASIN Processor For The Cnticality Sequences.

MOS, NiC'CR-5990 THE EFFECTS OF SOLAR-GEOMAGNETCALLY IN-82 M NW1 A N N MNW M NUREG/

DUCED CURRENTS ON ELECTRICAL SYSTEMS IN NUCLEAR POWER STATONS.

NU 6193: PRIMARY SYSTEM FISSION PRODUCT RELEASE AND TRANSPORT.A State Of-The-Art Report To The Committee On NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERLY NATIONAL SUREAU OF The Safety O Nuclear instanations.

NUREG/CR4204: OVESTIONS AND ANSWERS BASED ON REVISED NUREG/CP-0136. PROCEEDtNGS OF THE DIGITAL SYSTEMS RELi-ABILITY AND NUCLEAR SAFETY WORKSHOP. September 13 14, NU E /

VALVE ACTUATOR MOTOR DEGRADATION.

NURE N PCfR NWEG/W206. MANSNRT CAmAWS & RANW EX-L LT RADIATON EX-W TO WSSR SUNT SmmMS W M MAN E POSURE TO VESSEL SUPPORT STRUCTURES IN THE TROJAN RE.

ACTOR.

ACTOR-NUREG/CR4228: PRELIMINARY ASSESSMENT OF THE FAACTURE NEW MEXCO STATE UNIV., LAS CRUCES, NM BEHAVIOR OF WELD MATERIAL IN FULL-THICKNESS CLAD NURE TRAVAL PHASE 11 MODEL TESTING AT THE W NUREG/CR4231: A COMF ARISON OF THE RELATIVE IMPORTANCE NUREG/CR4120 CONTROLLED FIELD STUDY FOR VALIDATON OF OF COPPER PRECIPITA1ES AND POINT DEFECT CLUSTERS IN RE-VADOSE ZONE TRANSPORT MODELS.

ACTOR PRESSURE VES!'EL EMBRITTLEMENT.

NUREG/CR4249: UNIRRAL,'ATED MATERIAL PROPERTIES OF MID-NEW MEXICO, UNIV. OF, ALSUQUEROUE, NM LAND WELD WF-70.

NUREG/CF16042 PERSPECTIVES ON REACTOR SAFETY.

NUREG/CR4262: CLEAVAGE BEHAVIORS IN NUCLEAR VESSEL STEELS.

NOVETECH CORP.

I NUREG/CR4281: A SIMPLIFIED LEAUr ; FORE-BREAK rVALUATION OGDEN ENVIRONMENTAL & ENERGY SERVICES (FORMERLY PROCEDURES FOR AUSTENITIC Af0 FERRITIC S' EEL PIPING.

MULTIPLE DYNAMICS CORP) l NUREG/CR-5314 V05: INSIGHTS FOR AGING MANAGEMENT OF l

NUCLEAR POWER ENWNEERING CORP.

MAJOR LWR COMPONENTS METAL CONTAINMENTS.

NUREG/CR4213 HIGH-TEMPERATURE HYDROGENdlR-STEAM DETONATON EXPERIMENTS IN THE BNL SMALL-SCALF DEVELOP-ORGANIZATION FOR ECONOMIC COOPERATION & DEVELOPMENT MENT APPARATUS NUREG/CP 0127: PROCEEDINGS OF THE CSNI SPECIALISTS MEET-OAK RIDGE NATIONAL LA80RATORY NUREG/CR-4219 V10 N1: HEAVY SECTION STEEL TECHNOLOGY PACIFIC-NUCLEAR CO.

PROGRAM Semiannual Progress Report For October 1992 March NUREG/CR-555 W8L RELAPS/ MOD 3 CODE MANUAL. Validation Of NUR G/CR-4674 V17. PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS 1992 A STATUS REPORT Main Report And PERFORMANCE TECHNOLOGY GROUP, INC.,

8 A.

NUREG/CR4289 RECONCENTRATION OF RADOACTIVE MATERIAL N

G/CR4674 V18. PRECURSORS TO POTENTIAL SEVERE CORE RELEASED TO SANITARY SEWERS IN ACCORDANCE WITH 10 CFR DAMAGE ACCIDENTS: 1992 A STATUS REPORT.Appyrxhces B, C, D.

PART 20.

E, F, And G NUREG/CR-4674 V19: PRECURSORS TO POTENTIAL SEVERE CORE PLO, INC. (FORMERLY PICKARD, LOWE 4 GARRICK, INC.)

DAMAGE ACCIDENTS: 1993 A STATUS REPORT Main Repat And NUREG/CR4144 V02P1A-EVALUATON OF POTENTIAL SEVERE AC-

^" "

^

NU 67 V20- PRECURSORS TO POTENTIAL SEVERE CORE

"*"'Y DAMAGE ACCIDENTS: 1993 A STATUS REPORT.Appendences E Events Dunng M4 Loop ations Main Report (Chapters (4).

44 2% NATON & MENM MRE &

NUR /CR-4816 R02: PR-EDB POWER REACTOR EMBRITTLEMENT CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NURE /CRSi V R2 R VE 1 USER'S GUIDE.

SURRY. UNIT 1. Analyses Of Core Damage Frequency From Internal NUREG/CR-5247 V02 R2: RASCAL VFRSON 21 WORKBOOK _

Events Dunng Mid-Loop Operations Main Report (Chapters 712)

NUREG/CR-5359-REVtEW OF ELASTO STRESS AND FATIGE TO.

NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVEDE ACCl-FAILURE DATA FOR BRANCH CONNECTIONS AND TEES IN RELA.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TION TO ASME DESIGN CRITERIA FOR NUCLEAR POWER PIPING SURRY, UNIT 1.AnaYsis Of Core Damage Frequency From Internal SYSTEMS Events Dunng Mid-Loop Operations. Appendices A-D NUREG/CR-5569 ROI: HEALTH PHYSICS POSITIONS DATA BASE.

NUREG/CR4144 V02P3A: EVALUATON OF POTENTIAL SEVERE AC-NUREG/CR-5591 VG2 N1: HEAVY-SECTON STEEL tRRADIATON CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PROGRAM.Semiennual Progress Report For October 1990 March SURRY, UNIT 1 Analysrs Of Core Demage Frequency From internal 1991.

Events Dunng M4 Loop Operations Appendices E (Sections E.1 E.8)

118 Contractor index NUREG/CR 6144 V02P30: EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR-6075 S01: THE PROBABILITY OF CONTAINMENT FAIL-COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT UAE BY DIRECT CONTAINMENT HEATING IN ZION.

SURRY, UNIT 1. Analysis Of Core Damage Frequency Frorn Internal NUREG/CR.6092: RISK ASSESSMENT FOR THE INTENTIONAL DE-Events Dunng M4 Loop Operations Appendices E (Sections E 9-E.16).

PRESSURIZATION STRATEGY IN PWRS.

NUREG/CR4144 V02P4: EVALUATION OF POTENTIAL SEVERE ACCl-NUREG/CR4093: AN ANALYSIS OF OPERATIONAL EXPERIENCE DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DURING LOW POWER AND SHUTDOWN AND A PLAN FOR AD-SURRY. UNIT 1 Analyss Of Core Damage Frequency Frorn internal DRESSING HUMAN RELIABILITY ASSESSMENT ISSUES.

F*ents Dunng M4 Loop Operations Appendices F-H_

NUREG/CR4095; AGING. LOSS OF. COOLANT ACCIDENT (LOCA),

NU9FG/CP4144 V02P5; EVALUATON OF POTENTIAL SEVERE ACCl.

AND HtGH POTENTIAL TESTING OF DAMAGED CABLES.

DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4103: PRIORITI2ATION OF REACTOR CONTROL COMPO-SURRY, UNIT t Analyss of Core Darnage Frequency From intemal NENTS SUSCEPTIBLE TO FIRE DAMAGE AS A CONSEQUENCE OF Events Dunng Mea-Loop Operations Appendices 1.

AGING.

NUREG/CR4144 V03 PI: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR-6107:

SUMMARY

OF MELCOR 1.8.2 CALCULATIONS FOR CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT THREE LOCA SEQUENCES (AG.S2D & S3D) AT THE SURRY PLANT.

SUaRY, UNIT 1Analyss Of Core Damage Frequency From intemal NUREG/CR-6143 V02P1A: EVALUATION OF POTENTIAL SEVERE AC.

Fires Dona M4 Loop Operations Main Report.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/Ch4144 V03 P2: EVALUATON OF POTENTIAL SEVERE AC-GRAND GULF. UNIT 1 Anahsis Of Core Damage Frequency Frorn in.

CIDENTS JURING LOW POWER AND SHUTDOWN OPERATONS AT ternal Events For Plant Operational State 5 Dunng A Refuehng SURRY,DNIT 1.Analyss Of Core Damage Frequency From intemal Outage. Sections 19.

Fires Dunng M4 Loop Operations Appendices.

NUREG/CR-6143 V02PtB: EVALUATON OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT PRO CONSULTING GRAND GULF, UNIT 1. Analysis of Core Damage Frequency Frorn in.

NUREG/CR4143 V05: EVALUATION OF POTENTIAL SEVERE ACCI-ternal Events For Plant Operational State 5 Dunng A Refuehng DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Outage Section 10 GRAND GULF, UNIT 1.Analyss Of Core Damage Frequency From NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC.

Seismic Events Dunng M4 Loop Operations.Mann Report CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-6144 V05: EVALUATON OF POTENTIAL SEVERE ACCl-GRAND GULF, UNIT 1 Analysis Of Core Damage Frequency From In-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ternal Events For Plant Operahonal State 5 Dunn0 A Refuehng SURRY. UNIT 1 Analyss Of Core Darnage Frequency From Seismic Outape. Main Report Events Dunng M4 Loop Operahons. Main Report NUREG/CR4143 V02PT2-EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT

• • • 9*

NU EG 553 V06 RE CODE MANUALVahdation Of temal ena ant OperaW he 5 Dunng NW Numencal Techniques in REl.AP5/ MOD 3 NUREG/CR-6267. AIR-WATER SIMULATON OF PHENOMENA OF OuagelntemaL CORIUM DISPERSON IN DIRECT CONTAINMENT HEATING' REG /GM W2m ENATON & UEm SEM C CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT OUANTUM TECHNOLOGIES,INC.

GRAND GULF, UNIT 1 Analysis Of Core Damage Frequency From In-NUREG/CR4126. COGNITIVE SKILL TRAINING FOR NUCLEAR ternal Events For Plant Operational State 5 Dunng A Refuehng POWER PLANT OPERATONAL DECISION-MAKING.

Outage.intemal....

NUREG/CR4143 V02PT4: EVALUATION OF POTENTIAL SEVERE AC.

RELIABILITY AND PERFORMANCE AFSOCIATES COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-4674 V17: PRECURSOR;d TO POTENTIAL SEVERE CORE GRAND GULF. UNIT 1.Analyss Of Core Damage Frequency From Irt.

DAMAGE ACCOENTS: 1992 A STATUS REPORT. Main Report And temal Events For Plant Operataonal State 5 Dunng A Refuehng Apperdu A.

Oui 8 ntemal _.

I NUREG/GR-4674 V18: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/ CRC 43 V03 EVALUATON OF POTENTIAL SEVERE ACCl-DAMAGE ACCIDENTS 1992 A STATUS REPORTAppendices B. C, D.

DENTS DURh4G LOW POWER AND SHUTDOWN OPERATIONS AT E, F, And G.

GRAND GULF,VNIT 1 Analysis Of Core Damage Frequency From in-S. COMEN & ASSOCIATES, INC.

ternal Events For Plant Operational State 5 Dunng A Refueling Outa_ge.

NUREG/CR4143 V04 EVALUATION OF POTENTIAL SEVERE ACCI-NUREG 1492 OFC: REGULATORY ANALYSIS ON CRITERIA FOR THE DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE GRAND GULF, UNIT 1 Analysis Of Core Damage Frequency From In-CHAkA NURE /

47 VO ZATION OF CLASS A LOW-LEVEL aR t

Lb A LOW-LEVEL

^

^

NURE 7 0:

ZA ION O PROTEC R /

NURE 7

A T RZ 10 OF C S A LOW LEVEL S TO MESMATE DIRECT N RADIOACTIVE WASTE 19861990 Main Report-Part B.

TAINMENT HiATING PHENOMENA WITH SCALED MODELS OF THE

?

^

l NUREG/C VO R

N'TAL RESULTS FROM CONTAIN-RAD CT WAST 1 9

es A E NUREG/CR4147 V05: CHARACTER A70 OF CLASS A LOW-LEVEL MENT PIPING *LLOWS SUBJECTED TO SEVERE ACCOENT RADCACT!VE WASTE 19861990 Apperdx F CONDITIONS Results From Benows Tested in "Like-New" Cordhons.

NUREG/CR-6147 V06: CHARACTERIZATION O'F CLASS A LOW LEVEL NUREG/CR-6166. RISK IMPACT OF TECHNICAL SPECIFICATIONS RE-RADIOACTIVE WASTE 19861190 Appereces G-J OUIREMENTS DURING SHUTDOWN FOR BWRS.

NUREG/CR4147 V07: CHARACTER!2ATON OF CLASS A LOW-LEVEL NUREG/CR-6211. INTEGRATED FUEL COOLANT INTERACTON (IFCI RADOACTIVE WASTE 19861990Appendeces K-P.

6 0) CODE. User's Manual NUREG/CR-6218 A REVIEW OF THE TECHNICAL ISSUES OF AIR IN.

SANDIA NATIONAL LABORATORIES GRESSION DURING SEVERE REACTOR ACCIDENTS.

NUREG/CR 4551 V01 RI: EVALUATON OF SEVERE ACCIDENT NUREG/CR-6221 THE VALLES NATURAL ANALOGUE PROJECT.

RISKS: METHODOLOGY FOR THE CONTAINMENT. SOURCE TERM. CONSEQUENCE, AND RISK INTEGRATION ANALYSES SCIENCE & NNEEM ANATES, E NUREG/CR-4838 MICROCOMPUTER APPLICATIONS OF, AND' MODI-NUREG/CR4095. AGING, LOSS-OF-COOLANT ACCIDENT (LOCA).

FICATICNS TO. THE MODULAR FAULT TREES AND HIGH POTENTIAL TFSTING OF DAMAGED CABLES NUREG/CR-5407: ASSESSMENT OF THE IMPACT OF DEGRADED NUREG/CR4103. PRIOR:TIZATION OF REACTOR CONTROL COMPO-SHEAR WALL STIFFNESSES ON SEISMIC PLANT RISK AND SEIS-NENTS SUSCEPTIEf E TO FIRE DAMAGE AS A CONSEQUENCE OF MIC DESIGN LOADS AGING NUREG/CR-5726 REVIEW OF THE DIABLO CANYON PROBABILISTIC NUREG/CR-6143 V02P1C EVALUATION OF POTENTIAL SEVERE AC-RISK ASSESSMENT, COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4042. PERSPECTIVES ON REACTOR SAFETY.

GRAND GULF, UNIT 1 Analysis Of Core Damage Frequency From In-NUREG/CR-6044 EXPERIMENTS TO INVESTIGATE DIRECT CON-temal Events For Plant Operational State 5 Dunng A Refuehng TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE Outage Sections 19 ZION NUCLEAR POWER PLANT IN THE SURTSEY TEST FACILITY.

NUREG/CR4143 V02P18: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4075: THE PROBABILITY OF CONTAINMENT FAILURE BY COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DIRECT CONTAINMENT HEATING IN ZION.

GRAND GULF. UNIT 1 Analysis Of Core Damage Frequency From In-

Contractor index 119 temal Events For Plant Operatonal State 5 Dunng A Refuehng GRAND GULF UNIT 1 Analysis Of Core Damage Frequency From In-Outage Secton 10.

NUREG/CR4143 V02PIC: EVALUATION OF POTENTIAL SEVERE AC-tema! Events For Plant Operational State 5 Dunng A Refuehng Outage.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4143 V04 EVALUATON OF POTENTIAL SEVERE ACCl-GRAND GULF, UNIT 1. Analysts Of Core Damage Frequency From In-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Analysis Of Core Damage Frequency From in-ternal Events For Plant Operational State 5 Dunng A RefurJng ternally induced Flooding Events For Plant Operational State 5 Durmg Outage Man Report NUREG/CR4143 V02PT2: EVALUATION OF POTENTIAL SEVEME #C-a Retuelsng.....

COENTS DURING LOW POWER AND SHUTDOWN OPERMONS A' NUREG/CR4166: RISK IMPACT OF TECHNICAL SPECIFICATIONS RE-GRAND GULF UNIT 1 Analysis Of Core Damage Frequerc) From In-OUtREMENTS DURING SHUTDOWN FOR BWRS.

NUREG/CR 4180: HYDROGEN MIXING STUDIES (HMS). USER'S ternal Events For Plant Operatonal State t; Dunnp Refuelmg MANUAL.

Outage.IntemaL..

NUREG/CR4143 V02PT3: EVALUATION OF POTENTIAL SE)EFE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERAROM AT SOREQ NUCLEAR RESEARCH CENTER, YAVNE, ISRAEL GRAND GULF, UNIT 1. Analyses Of Core Damage Frequency ??cm In-NUREG/CR4144 V02P1 A: EVALUATION OF POTENTIAL SEVERE AC-temal Events For Plant Operatonal State 5 Dunng A Refuehng CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Outace intemaL SURRY, UNIT 1 Analyses Of Core Damage Frequency From intemal NUREG/CR4143 V02PT4. EVALUATION OF POTENTIAL SEVERE AC-Events Dunng M4 Loop Operations Mam Report (Chapters 14L COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR4144 V02P18: EVALUATION OF POTENTIAL SEVERE AC.

GRAND GULF, UNIT 1 Analysis Of Core Damage Frequency From In-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT temal Events For Plant Operational State 5 Dunng A Refuelmg SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal Events Dunng M4 Loop Operatons. Main Report (Chapters 712).

NU 303. EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR4144 V02P2: EVALUATION OF POTENTIAL SEVERE ACCI-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT GRAND GULF. UNIT 1. Analysis Of Core Dama9e Frequency From In-SURRY, UNIT 1. Analysis Of Core Demage Frequency From Intemal temal Events For Plant atonal State 5 Events Dunng M4 Loop Operations. Appendices A 0.

A RefueshOutaENTI t'NUREG/CR4144 V02P3A: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4224 DFC:

RAMETAIC STU THE P FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERAT.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT ED DEDRIS. Draft For Comment.

SURRY, UNIT 1. Analyses Of Core Damage Frequency Frorn intemal Events Durmg M4 Loop Operatons.Appereces E (Sections E.1 E.8).

SCIENCE APPUCATIONS INTERNATIONAL CORP,(FORMERLY NUREG/CR4144 V02P38. EVALUATION OF POTENTIAL SEVERE AC-SCIENCE APPLICATIONS, CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG-1484 V01: FINAL ENVIRONMENTAL IMPACT STATEMENT SURRY, UNIT 1. Analysis Of Core Damage Frequency From internal FOR THE CONSTRUCTION AND OPERATON OF CLAIBORNE EN-Events Dunng M4 Loop Operations.Appereces E (Sections E.9-E.16).

RICHMENT CENTER. HOMER, LOUISIANA. Docket No 70-3070 Louise NUREG/CR-6144 V02P4. EVALUATION OF POTENTIAL SEVERE ACCI-ana Enerov Services, LP. Environmental lmpact Statement.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG-14M V02: FINAL ENVIRONMENTAL IMPACT STATEMENT FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-SURRY, UNIT 1. Analysis Of Core Damage Frequency From Intemal Events Dunng M4 Loop Operations Appereces F-H.

RICHMENT CENTER, HOMER. LOUISIANA. Docket No. 70'3070.Louish ana Energy Sennces, L P. Comments And Responses.

NUREG/CR4144 V02P5. EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR-4674 V17: PRECURSORS TO POTENTIAL SEVERE CORE DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DAMAGE ACCIDENTS: 1992 A STATUS REPORT Main Report And SURRY, UNIT 1. Analysis Of Core Damage Frequency From Internal Events Dunng M4 Loop Operations Appereces 1.

NU C -4674 V18. PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS.1992 A STATUS REPORT.Apperdces B, C, D, SOUTHWEST RESEARCH INSTITUTE NUREG/CR-6074 V01: SEALED SOURCE AND DEVICE DESIGN NL G R-4674 V19: PRECURSORS TO POTENTIAL SEVERE CORE SAFETY TESTING Technical Report On The Findings Of Task 1.Octo.

DAMAGE ACCIDENTS 1993 A STATUS REPORT Masn Report And 1991 Apnl 1993.

NU 4674 V20: PRECURSORS TO POTENTIAL SEVERE CORE TECHNADYNE ENGINEERING CONSULTANTS,INC.

DAMAGE ACCIDENTS.1993 A STATUS REPORT.Appendences E NUREG/CR-4551 VOI Rf: EVALUATION OF SEVERE ACCIDENT And F.

RISKS. METHODOLOGY FOR THE CONTAINMENT, SOURCE NUREG/CR-5939 THE EFFECTS OF AGE ON NUCLEAR POWER TERM. CONSEQUENCE, AND RISK INTEGRATION ANALYSES.

PLANT CONTAINMENT COOllNG SYSTEMS.

NUREG/CR-5967. DEVELOPMENT AND APPLICATON OF DEGRADA.

TEXAS A&M UNIV, COLLEGE STATION, TX TION MODELING TO DEFINE MAINTENANCE PRACTICES NUREG/CR-6143 V02P1A EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR 6162: EFFECTS OF PRIOR DUCTILE TEARING ON CLEAV.

ClDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT AGE FRACTURE TOUGHNESS IN THE TRANSITON REGION GRAND GULF. UNIT 1. Analyses Of Core Damage Frequency From in.

TEXAS, UNIV. OF, AUSTIN, TX ternal Events For Plant Operational State 5 Dunng A Refuehng NUREG/CR4063 INTRAVAL PHASE ll MODEL TESTING AT THE LAS Outage Sections 19 CRUCES TRENCH SITE.

NUREG/CR-6143 V02PIB. EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TEXAS, UNIV. OF, HOUSTON, TX GRAND GULF. UNIT 1 Ana!ysis Of Core Damage Frequency From In-NUREG/GR.0011: INFORMATON BIAS AND LIFETIME MORTALITY temal Events For Plant Operational State 5 Dunng A Refuehng RISKS OF RADIATION-lNDUCED CANCER Low LET Radiation.

Outage Secton 10.

NUREG/CR4143 V0291C: EVALUATON OF POTENTIAL SEVERE AC.

U.S. NAVAL ACADEMY, ANNAPOUS, MD CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4051: EFFECTS OF TENSILE LOADING ON UPPER SHELF GRAND GULF. UNIT 1 Analysis Of Core Damage Frequency From In-FRACTURE TOUGHNESS.

temal Events For Plant Operational State 5 Dunng A Refuehng Outage Main Report.

WESTINGHOUSE ELECTRIC CORP.

NUREG/CR4143 V02PT3. EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4126-COGNITIVE SKILL TRAINING FOR NUCLEAR CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT POWER PLANT OPERATIONAL DECISION. MAKING GRAND GULF, UNIT 1 Analyses Of Core Damage Frequency From in.

NUREG/CR-6127. THE EFFECTS OF STRESS ON NUCLEAR POWER ternal Events For Piant Operatonal State 5 Dunng A Refueling PLANT OPERATONAL DECISION MAKING AND TRAWING AP-Outage inteman...

PROACHES TO REDUCE STRESS EFFECTS.

NUREG/CR4143 V02PT4 EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-6208. AN EMPIRlCAL INVESTIGATON OF OPERATOR CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PERFORMANCE IN COGNITIVELY DEMANDING SIMULATED EMER-GRAND GULF. UNIT 1. Analyses Of Core Damage Frequency From In-GENCIES.

temal Events For Plant Operatonal State 5 Durmg A Refuehng Outage intemal....

WISCONSIN. UNIV. OF. MADISON, WI NUREu/CR 6143 V03 EVALUATION OF POTENTIAL SEVERE ACCl-NUREG/CR-6196. CALCULATIONS TO ESTIMATE THE MARGIN TO DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT FAILURE IN THE TMI-2 VESSEL l

.e.m'--ao u

W 1

r B

I l

l i

f i

i

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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.

SELGIUM REPUBLIC OF MOMA TRACTEBEL KOREA INSTITUTE OF NUCLEAR SAFETY NUREG/lA4093: RELAP5/ MOD 3 ASSESSMENT FOR CALCULATION NUREG/lA4114: ASSESSMENT OF RELAPS/ MOD 3 WITH THE LOFT OF SAFETY AND RELIEF VALVE DISCHARGE PIPING HYDRODY.

L9-1/L3-3 EXPERMENT SIMULATING AN ANTICIPATED TRAN.

NAMIC LOADS.

SIENT WITH MULTIPLE FAILURES.

m 121

z--

a a

s

_,,, ~w-s, 4

.a am

-,s1,su f,

h 4

6 e

l 4

4

)

i P

h.

i 4

k I

i 4

4 i

B

l Licensed Facility Index l

This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number 1

and followed by the report nurnber, if further information is needed, refer to the main citation by the NUREG number.

l 52403 AP600 Standard Plant Desgn, Weshn0 house NUREG/CR4223 50 280 Suny Power Stason, Umt 1, Vrgne Electnc & NUREG/CR4152 Electnt Corp, Power Ca 50 275 DetNo Canyon Nuclear Pces Plant, Umt 1, NUREG/CR 5726 54280 Suny Power Staban, Umt 1 Vrgna Electnc 4 NUREG/CR-6181 Pacmc Gas & Eiscine Co Poww Co.

% 323 Diablo Canyon Nuclear Power Plant, Umt 2, NUREG/CR 5726 50-281 Suny Power Staton, Umt 2. Vrgna Electnc & NUREG/CR 6107

• acdc Gas & Electc Co Power Ca Protect 469 EPRI. ALWR NUREG-1242 V03 PTot 50-261 Suny Power Staton, Umt 2. Vrgna Electnc & NUREG/CR 6152 Protect 469 EPRf. ALWR NUREG1242 W3 PT02 Poww Ca ak WR Desgn, 52 002 System 80+ Standerked Nuclear Power Plant NUREG1462 V01 al Ca, N YO1 Senphhed (SB ) Desgn, al

/CR-6223 52 002 Systen 80 Standerk Nuclear Power Plant NUREG1462 V02 Electnc Ca, Des., Combusbon Engmee

% 416 Gand Gull Nuclear Staton, Und 1, usssspp NUREG/CR4143 V02PtA 52402 Syskm 80+ Standarted Nuclear Pcwor Plant NUREG/CR-6128 Power & Lght Co Des., Combuston Enginee 50 416 Grand Gu# Nuclear Stanon, Urd 1. Essrmp NUREG/CR4143 V02P1g 52 002 System 80+ Slander &ed Nuclear Power Plant NUREG/CR4161 Power & Lght Co Des., Comnuston Engree 54416 Grand Guf Nucles Staban, Urut 1, Hsssspp NUREG/CR4143 V02PtC

% 320 Three Mie laiand Nuciss Stanon, Umt 2.

NUREG/CR4183 Power & LgM Co General PutAc Untes 50 416 Grand gun Nuclear Stabon, Umt 1, Esssspp NUREG/CR4143 V02PT2 50 320 Three Mie bland Nuclear Stahon, Umt 2,.

NUREG/CR4185 Power & Lght Co General PutAc Uthes

% 416 Grand Gup Nuclear Stahon, Und 1, Massspp NUREG/CR4143 V02PT3 W320 Three hie Island Nuclear Stabon, Umi 2, NUREG/CR4187 Power & L;ght Co General Pubbt Untes 50416 Grand Gun Nuclear Staton, Und 1. Essaspp NUREG/CR 6143 V02PT4 50-320 Three ule island Nucles Staban, Umt 2, NOREG/CR4194 Power & (ght Co General Pub 4c UHtes 50 416 Grand Guit Nuclear Stabon, Umt 1, Msssspp NUREG/CR4143 V03 50 320 Three his island Nuclear Staton, Umt 2, NUREG/CR4195 Power & Lght Co General Ph Unbes 54416 Grand gun Nuclear Staton, Umt 1, Esssspp NUREG/CR-6143 V04 54320 Three hie Island Nuclear Station Urvt 2, NUREG/CR-6196 1

Power & (ght Co-General Pubhc Uthes

$4416 Grand Guti Nucles Stahon, Und 1. Essssipp NUREG/CR4143 V05 50 320 Three Wie Island Nuclear Staton, Urut 2, NUREG/CR4197 Poww & Lght Ca General Pubhc Utikbes 44a968 Hydro Resources. Inc., Dallas, TE NUREG1508 50-320 Three use Island Nuclear Staton, und 2.

NUREG/CR4198 j

743070 Louisana Energy Sernces, WasNngton, DC, NUREG1484 V01 ggy WaWangt n, DC' Genwal Pubhc Umbes DC W320 Three his Island Nuclew Staten, Umt 2, NUREG/CR4252

% 329 Edland Plant, rut 1, Consumer Co NUREG/CR4249 General Pub 4c Umkes

% 330 udland Plant, Und 2. Consumers Power Co NUREG/CR 6249 5434 Troian Nuclear Plant Portland General Elecinc NUREG/CR4206 54171 Peach Bottom Alome Power Staton Umt 1,'

NUREG/CR 5850 50 390 Watts Bar Nuctsar Plant, Umt 1 Tennessee NUREG4496 S01 DFC Phiadelpma Electnc Co Valley Aumonty r

l 50 277 Peach Bottom Alomic Power Staton, Umt 2, NUREG/CR-5850 50 390 Watts tw Nuclear Plant, Urd 1. Tennessee NUREG0847 S13 Phfadeghe Electnc Ca V

WanQsAumonty ES278 Peach Bottom Atome Power Stahon, Und 3.

NUREG/CR 5850 2 390 Nuclear Plant Und 1. Tennessee NUREG4847S14 Phtadegha Eoc!nc Ca valley Authordy 54278 Peach Bottom Atome Power Staton, Umt 3, NUREG/CR4164 50 391 Watts Bar Nuclew Plant. Urut 2, Tennessee NUREG0498 S01 DFC Phiadelpha Electnc Co.

Valley Authonty (S 280 Suny Poww Slahon, Und 1. Vrgna Electnc & NUREG/CR4107 54391 Watts Bar Nuclear Plant, Umt 2 Tennessee NUREG4647 St3 PowerCo Valley Authonty b280 Suny Power Stanon. Und 1 Vrpms Elecinc & NUREG/CR4144 V02 PIA 54391 Watts Bar Nuclear Plant, Und 2. Tennessee NUREG 0647 S14 Power Co.

Valley Aumarty 5S 280 Surry Power Staton, Umt 1. Vrgna Elecinc & NUREG/CR4144 V02PIB 50 369 Wdham B. McGure Nuclear Staton, Und 1 Duke NUREG/CR-5830 Power Co Power Co (S 280 Suny Power Stanon Umi 1. Vagna Eiectnc & NUREGICR4.44 V02P2 50 370 W4 ham 8 McGure Nuclear Stahon, Urut 2, Duke NUREG/CR-5830 Power Co Power Co

% 200 Suny Power Staton, Und 1, Vrgna Electre & NUREG/CR4144 V02P3A 2 295 Zon Nuclear Power Staten, Und 1 NUREG/CR4044 i

Power Ca

? - - --- alth Eeson Co 54280 Sry Power Staton, Und 1, Vrgna Elo inc 4 NUREG/CR4144 V02P3B E295 Zon Nuclear Power Slabon, Umt 1 NUREG/CR4075 Power to G-

.2 Eeson Co 54200 Suny Power Staton, Und 1. Vrgne Electnc & NUREG/CR4144 V02P4 54295 Zon Nutteer Power Staton, Und 1 NUREG/CR4075 S01 Power Ca Commonweanh Eeson Co.

ES280 Suny Power Stabon, Umt 1 Vrgna Electnc & NUREG/CR6144 V02P5

% 295 Zon Nuclear Power Staton, Und 1 NUREG/CR4168 PowerCo Commonweanh Edson Ca 50 280 Surry Power Staton, Und 1, Vegna Electnc & NUREG/CR4144 V03 P1 54304 Zon Nuclear Power Stanon Und 2.

NUREG/CR4044 Power Ca

?_

._1. Eeson Co

% 280 Suny Power Staten, Umt 1, Vrgma Elecine & NUREG/CR4144 V03 P2 50 304 Zon Nuclear Power Staton, Umt 2, NUPEG/CR-6075 Power Co i

Commonwealm Eeson Co fS280 Suny Power Stabon, Umt 1, Vrgna Elecinc & NUREGICR4144 V04 54304 Zon Nuclear Power Stanon, Und 2.

NUREG/CR4075 S01 Power Co Commonweann Edson Ca SS280 Suny Pcuer Stahon, Und 1 Vrgna Electnc & NUREG/CR4144 V05 54304 Zon Nuclear Power Stahon, Und 2.

NUREG/CR 6168 Power Ca Commonweanh Eeson Co 123 l

l l

I l

NRC FOC 336 U.S. NUCLEAR REGULATOQY COMMISSION

1. WEPORT NUMBER kR 1102 any noi. no2 BIBLIOGRAPHIC DATA-SHEET NUfG 04 tsee inuunctions on the mwse,

2. TITLE AND SUBTITLE Regulatory and Technical Reports (Abstract Index Journal) 3.

DATE REPORT PUBLISHED Annual Compilation for 1994 l

"o"

• vtaa March 1995

4. FIN OR GR ANT NUMBER YAUTHOR(S)

6. TYPE OF REPORT

7. PE RIOD COV E R E D (sacAusne ostess 8.

F RMING NIZATION - NAME AND ADDRESS (ss Nnc. orovie oivaion. oHue er neeron U.5 Nuckar nosuintory Commusuon. emt matern, anheav st eontractor. orovkse Division of Freedom of Information and Publications Services Office.of Administration U.S. Nuclear Regulatory Commission Bashington, DC 20555-0001 9.

R GANIZATION - NAME AND ADDR ESS tor Nac. type *kme as sboue";itcontrernor. orovirk Nnc Divuton. orike er ne,kn. v.1 Nacnone nenutatory c_-

Same as 8, above.

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 more or hast This journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors, proceedings of conferences and workshops, grants, and international agreement reports. The entries in this compilation are indexed for access by title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agreements, contractor, international organization, and licensed facility.

12. KE Y WORDS/DESCR:PTORS (Usr were orperom sner evim asser resenners m sacer/np rae roporr.)

13. AVAILAe:UT y 4T AIEMENT compilation, abstract index unlimited

14. SECURaiV CLA$$1FICAT604 f rhk Pneel unclassified erna nni unclassified

16. NUMBER OF PAGE S

16. PRICE NRC FORM 335 (249)

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Printed

• on recycled paper l

Federal Recycling Program

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8 gg Main Citations E

za and Abstracts

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$"e k

15 855 Secondary Report El ;

i!

Number index gE I

6!

"E Personal Author index 0

Subject index NRC Originating Organization Index (Staff Reports)

NRC Originating Organization

/

Index (International Agreements)

NRC Contractor Sponsor index Contractor index International Organization index g

A W3 aHne$

i Pn

,o.

Licensed Facility 6g_?g y f,tj index j

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