ML20043D533

Regulatory and Technical Reports (Abstract Index Journal). Compilation for First Quarter 1990,January-March

ML20043D533
Person / Time
Issue date:	05/31/1990
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V15-N01, NUREG-304, NUREG-304-V15-N1, NUDOCS 9006080185
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4 Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.

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Vol.15, No.1 Regulatory and Technical Reports I

(Abstract Index Journal)

Compilation for First Quarter 1990 January - March Date Published: May 1990 Regulatory l'ublications liranch Division of Freedom ofInformation and l'ublications Services Omce of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 p

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CONTENTS Preface...................,............................................................v Index

.Tob Main Citations and Abstracts............................................................ 1

• Staff Reports

• Conference Proceedings

• Contractor Reports

• Intemational Agreement Reports Secondary Report N umber inde x......................................................... 2 Personal Aut hor in de x.................................................................. 3 S u bject i nde x......................................................................... 4 N RC Originating Organization index (Staff Reports)......................................... 6 NRC Originating Organization Index (International Agreements)............................... 6 NRC Contract Sponsor index (Contractor Reports).......................................... 7 Contra ctor inde x....................................................................... 8 International Organization Inde x..........................................................

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Licensed Fa cilit y inde x.................................................................. 10 I

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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) Statf and its contractors it in NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to:

Division of Publications Services Policy and Publicetions Management Branch Publishing and Translations Section Woodmont 537 U.S. Nuclear Regulatory Commission Washington, D.C. 20b55 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP XXXX, NUREG/CR XXXX, and NUREG/lA XXXX. These precede the following indbxes:

Secondary Report Number Index Personal Author index 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 l

NUREG@08. MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

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

l 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 internal 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. Argonno National Laboratory. May 1981.141 pp. 8105280299. ANL 813. 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 Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC intemal use).

l Contractor Report NUREG/CR 1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER l

REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.: BENNETT, P.R.

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

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

l authors or pub'isher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use).

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Intemational Agreement Report NUREG/lA-0031: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA. NEUMANN, U. Kraftwerk Union. AuDust 1996. 223 pp. 8608270424, 37659:138.

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

DRFT - draft ERR errata N - number R

revision S

supplement V

volume

_ Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the following address:

i 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)275-2000 or (202)275-2171. Non U.S. customers must make payment in advance either by International Postal Money Order, payable to the Superintendent of Documents, or by draf t on a United States or Canadian bank, payable to the Superintendent of Documents.

NRC Report Codes The NUREG designation, NUREG-XXXX, in^:ates that the document is a formal NRC staff generated report. Contractor prepared termal NRC reports carry the report code NUREG/CR XXXX. This type uf i

identification replaces contracor-established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported.

In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC sponsored conference proceedings and NUREG/lA is used for intemational agreement repcrts.

All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Publications Services, vi

^

3 Main Citations and Abstracts The report listin s in this compilation are arranged by report number, where NUREG XXXX is i

an NRC staff-inated report, NUREG/CP XXXX is an NRC sponsored conference report, NUREG/CR XX is an NRC contractor prepared report, and NUREG/lA XXXX is an inter-national agreement re sort. The bibliographic information (see Preface for details) is followed by a brief abstract of t11s report.

)

NUREG 0020 V13 N11: LICENSED OPERATING REACTORS ty. The other four abnormal occurrences took placs under other STATUS EUMMARY REPORT. Data As Of October NRC 6ssued beenses: the first involved a medcal diagnostic 30,1989.(Gray Book () LOVELACE.W.H. Division of Computer &

misadministration; the second involved a medcal therapy rns Telecommuncatons Services (Post 890205). January 1990.

administraton; the third 6nvolved a radiaton overuposure of a

$43pp. 9002120335. $2579130 radographer; and the fourth involved a significant breakdow*t THE OPERATING UNITS STATUS REPORT LICENSED and careless disregard of the radiation safety program at three OPERATING REACTORS provides data on the operaton of nu-of a heensee's manufactunng facihtees. The Agreement States clear units as timely and accurately as possible. This informa-reported no abnormal occurrenccs durinD the reporting period.

tion is collected by the Offee of Informaton Resources Man-The report also contains informat.on that updates some prev 6-agement from the Headquarters staff of NRC's Offee of En-ously reported abnormat occurrences.

forcement (OE), from NRC's Re9;onal Offees, and from utikties.

j The three sections of the report are: monthly highhghts and sta-NUREG 0304 V14 N04: REGULATORY AND TECHNICAL RE.

tistes for commere,al operating units, and errata from previously PORTS (ABSTRACT INDEX JOURNAL). Annual Compliaton reported data; a compliahon of detailed information on each For 1989.

• Dvision of Freedom of information & Pubhcations unit, provided by NRC's Re0ional Offees. OE Headquarters and Services (Post 890205). March 1990.153pp. 9004030145, the utthties; and an appendix for miscellaneous informaton such 53228.250.

as spent fuel storage capabihty, reactor years of exponence and This joumal includes all formal reports in the NUREG senes non power reactors in the U.S. It is hoped the report is helpfut prepared by the NRC staff and contractors; proceedings of con-to all agencies and 6ndiv! duals interested in maintaining an forences and workshops; as well as intomational agreement re-awareness of the U.S. energy situaton as a whole.

ports. The entnes in this compilation are indexed for access by title and abstract, secondary report number, personal author, NUMEG 0020 V13 N12: LICENSED OPERATING REACTORS subject, NHC organtration for statt and international agree-STATUS

SUMMARY

REPORT. Data As Of November monts, cowactor, intematonal organnaton, and heensed facib 30,1989.(Gray Dook 1) LOVELACE W.H. Division of Computer &

Y Telecommunicatons Services (Post 8D0205). January 1990.

558pp. 9003070140. 52809.125.

NUREG 0386 DOS R05: UNITED STATES NUCLEAR REGULA-See NUREG 0020,V13,N11 abstract.

TORY COMMISSION STAFF PRACTICE AND PROCEDURE NUREG 0020 V14 Not: LICENSED OPERATING REACTORS DIGEST. Commission, Appeal Board And Licensing Board STATUS

SUMMARY

REPORT. Data As Of December Decisions. July 1972. September 1989.

• Offee of the General 31,1989.(Gray Dook !) LOVELACE,W.H. Dvision of Computer &

Counsel (Post 860701). March 1990. 600pp. 9004110232.

Telecommuncations Services (Post 890205). February 1990.

53351:319.

550pp. 9003130117, 52885:090.

This Revision number 5 of the fifth edition of the NRC Prac-See NUREG 0020,V13,N11 abstract.

tee and Procedure Dgest contains a digesi of a number of Commission, Atomic Safety and ucensing Appeal Board, and NUREG 0040 V13 N04: LICENSEE CONTRACTOR AND Atomic Safety and Licensing Doard decisions issued during the VENDOR INSPECTION STATUS REPORT. Quarterly penod, July 1,1972 to September 30,1989, 6nterpreting the Report,0ctober December 1989.(White Book) ' Dvision of Re-NRC s Rules of Practee in 10 CFR Part 2.

actor inspecton & Safeguards (Post 870411). January 1990.

91pp. 9003070134. 52808.208.

NUREG 0540 V11 N10: TITLE LIST OF DOCUMENTS MADE This periodcal covers the results of Inspections performed by PUBLICLY AVAILABLE. October 1 31,1989.

• Division of Free-the NRC a Vendor Inspecton Branch that have been distnbuted dom of informaton & Pubhcations Services (Post 890205). Jan-uary M 326pp. W2120142. 52580251 198 thr h c This document is a monthly pubhcation containing descrip.

NUREG 0090 V12 NO3: REPORT TO CONGRESS ON ABNOR-tions of 6nformation received and Generated by the U.S. Nuclear MAL OCCURRENCES. July September 1989.

• Offee for Analy-Regulatory Commission (NRC). This information includes (1) sis & Evaluaton of Operatonal Data, Director. January 1990 doelieted material associated with civihan nuclear power plants 34pp. 9003070121. 52814:186.

and other uses of radioactive materiais, and (2) nondocketed Section 208 of the Energy Reorganization Act of 1974 loenti-matenal received and generated by NRC pertinent to its role as fies an abnormat occurrence as an unscheduled incident or a regulatory agency. The following indexes are included. Per-event whch the Nuclear Regulatory Commission determines to sonal Author, Corporate Source, Report Number, and Cross I

be signifcant from the standpoint of pubhc health and safety Reference to Pnnespal Documents.

and requires a Quarterly report of such events to be made to Congress. This report covers the penod July 1 through Septem-NUREG 0540 V11 N11: TITLE LIST OF DOCUMENTS MADE ber 30,1989. For this reporting penod, there were five abnormal PUBLICLY AVAILABLE. NOVEMBER 130, f 989.

• Division of occurrences. One obnormal occurrence took place at a heensed Freedom of Information & Pubhcations Servees (Post 890205).

f

!W nuclear power plant and involved signifcant defciencies associ-March 1990. 336pp. 9004090177. 53309.236 ated with the containment recirculation sump at the Trojan facili-See NUREG 0540,V11,N10 abstract.

l 1

2 Main Citations and Abstracts NUREG 0760 C102: INDEXES TO NUCLEAR REGULATORY 11,13,14, and 15 through 20 are not included in this supple-COMMISSION ISSUANCES. January 1,1980 Through December rnent, except to the extent that they affect the appleant's Final J1,1985.

• Drvision of Freedom of informaton & Pubicatens Safety Anatysis Report.

Servees (Post 890205). November 1989.1095pp. 9003160045.

52958.212.

NUREG-0797 823: SAFETY EVALUATION REPORT RELATED Ogests and indexes for issuances of the Commission, the TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-Atome Safety and Licensing Appeal Panel, ttw Atome Safety fRIC STATION.UNJTS 1 AND 2,Docaet Nos. 50445 And 50-and Licensing Board Panel, the Admirustrettve Law Judges, the 446.(Texas Utshtees Electric Company,et at)

• Comanche Peak Drectors' Decisions, and the Denials of Petitsons for Rulemak-Project Dvision. February 1990. 102pp. 9003070170.

InD are presented.

52808:299.

NUREG 0760 V30101: INDEXES TO NUCLEAR REGULATORY Supplement 23 to the Safety Evaluation Report related to the COMMISSION ISSUANCES. July September 1989.

• Dvision of operaton of the Comanche Peak Steam Electre Staton, Units Freedom of Information & Pubhcations Servces (Post 89020S.

1 and 2 (NUREG 0797), has been prepared by the Offee of Nu-February 1990. 50pp. 9004100439. 53350.238.

clear Reactor Regulation of the U S. Nuclear Re(,ulatory Com-See NUREG 0750,Cl02 abstract.

mission. The facihty is located in Somervell County, Texas, ap-proximate!y 40 trilles southwest of Fort Worth, Texas. This sup-NUREG 0760 V30 N01: NUCLEAR REGULATORY COMMISSION piement reports the status of certain issues that had not been ISSUANCES FOR JULY 1989 Pages 184.

• Dvision of Free-dom of Informaton & Pubicatons Services (Post 890205). Jarb resotved at the time of pubhcation of the Safety Evaluaton Report and Supplements 1, 2, 3, 4, 6,12, 21, and 22 to that uary 1990. 89pp. 9004030109. 53229:208.

report. This supplement also includes the evaluatons for heens.

Legalissuances of the Commission, the Atomic Safety and Li-censing Appeal Pane 4, the Atomic Safety and Licensing Board ing items resolved since Supplement 22 was issued. Supple-ment 5 has been cancelled. Supplements 7 through it were Panel, the Administrattve Law Judges, and NRC Program Of-hmited to the stati evaluaton of allegatons investigated by the fees are presented.

NRC Technical Review Team. Supplement 13 presented the NUREG-0750 V30 N02: NUCLEAR REGULATORY COMMISSION staff's evaluaton of the Comanche Peak Response Team ISSUANCES FOR AUGUST 1989. Pages 85165.

• Dvision of (CPRT) Program Plan, which was fomulated by the appleant to Freedom of informaton & Pubhcations Servces (Post 890205).

resolve vanous constructen and design issues raised by January 1990. 85pp. 9004030047, 53230:007, sources external to the applicant. Supplements 14 through 20 See NUREG 0750,V30,N01 abstract.

presented the staff's evaluaton of the appicant's Corrective

^

NUREG 0760 V30 NO3: NUCLEAR REGULATORY COMMISSION nn 7 0, 0, 1 13 o t

h re luded ISSUANCES FOR SEPTEMBER 1989. Pages 167 229.

• Dvi-l son of Freedom of information & Pubicatons Servces (Post this supplement, except to the extent that they affect the apph-890205). January 1990. 94pp. 9004030101. 53229.299.

cant's Final Safety Analysis Report.

See NUREG 0750,V' 0,N01 abstract.

J NUREG-0837 VJ9 N04: NRC TLD DIRECT RADIATION MONI-NUREG-0760 V30 N04: NUCLEAR REGULATORY COMMISSION TORING NETWORK. Progress Report. October December 1989.

ISSUANCES FOR OCTOBER 1989. Pages 231323.

• Dvision STRUCKMEYER R.; MCNAMARA,N. Region 1. Ofc of the Drec-of Freedom of information & Pubhcations Services (Post tor. March 1990. 300pp. 9004100322. 53343.078.

890205). January 1990. 97pp. 9004030044. 53230.096.

This report provides the status and results of the NRC Ther.

See NUREG 0750,V30,N01 abstract.

moluminescent Dosimeter (TLD) Drect Radiaton Monttonng "E***""

NUREG-0750 V30 N06: NUCLEAR REGULATORY COMMISSION of nsed facW ses WW N CW W N ISSUANCES FOR NOVEMBER 1989. Pages 32S 708. ' Dvisson of Freedom of information & Pubhcations Services (Post

@aw of M l

890205). March 1990. 389pp. 9004100444. 53350.286.

j See NUREG-0750,V30,N01 abstract.

NUREG-0896 S09: SAFETY EVALUATION REPCRT RELATED TO THE OPERATION OF SEABROOK STATION, UNITS 1 AND NUMEG 0797 822: SAFETY EVALUATION REPORT RELATED

2. Docket Nos. STN 54443 And STN 60 444.(Pubhc Service TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-Company Of New Hampshire)

• Dvision of Reactor Protects. l/

TRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50 ll (Post 870411). March 1990.182pp. 9004030159. 53229.042.

446.(Texas Utahties Electne Company,et al.)

• Comanche Peak This report is Supplement No. 9 to the Safety Evaluaton Project Dvision. January 1990. 369pp. 9003070163. 52814:220.

Report (SER) (NUREG-0896, March 1983) for the apphcaton l

Supplement 22 to the Safety Evaluation Report related to the filed by the Pubhc Service Company of New Hampshire, et al, l

operation of the Comanche Peak Steam Electre Station, Units for licenses to operate Seabrook Staten, Units 1 and 2 (Docket l

1 and 2 (NUREG 0797), has been prepared by the Offce of Nu-Nos. STN $0-443 and STN 50-444). It has been prepared by clear Reactor Regulation of the U.S. Nuclear Regulatory Com-the Offee of Nuclear Reactor Regulation of the U.S. Nuclear mission. The facihty is located in Somervell County, Texas, ap.

Regulatory Commission and provides recent information on proximately 40 miles southwest of Fort Worth, Texas. This sup-open items identified in the SER. The facihty is located in Sea-plement reports the status of certain issues that had not been brook, New Hampshire. Subject to favorable resoluton of the resolved at the time of pubhcaton of the Safety Evaluaton items discussed in this report, the staff concludes that the facili-Report and Supplements 1,2,3,4,6,12, and 21 to that report.

ty can be operated Dy the apphCant without endangering the This supplement also includes the evaluatons for heensing health and safety of the pubhc.

l Items resolved since Supplement 21 was issued. Supplement 5 has been cancelled. Supp;ements 7 through 11 were hmited to NUREG-0936 V08 N04: NRC REGULATORY AGENDA Ouarterly the staff evaluation of allegatons investigated by the NRC Report, October December 1989

• Dvision of Freedom of Infor.

Techncal Review Team. Supplement 13 presented the staffs mation & Pubhcations Services (Post 890205). January 1990.

evaluation of the Comanche Peak Response Team (CPRT) Pro-134pp. 9003070131. 52808:074.

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gram Plan, which was formulated by the applicant to resolve The NRC Regulatog Agenda is a compilation of all rules on l

various construction and design issues raisod by sources extor-whch the NRC has proposed or is considering acton and all nal to the applicant. Supplements 14 through 20 p'esented the petitons for rulemalung which have been received by the Com-staff's evaluation of the appheant's Corrective Action Program mission and are pending disposition by the Commission. The and CPRT activities. Items rientified in Supplements 7,8,9,10, Regulatory Agenda is updated and issued each rtuarter.

J

l Main Citations and Abstracts 3

NUMEO 1100 V08: BUDGET ESTIMATES. Fiscal Year 1991.

• D-censed by the Commation. This is an annual pubicaton for the veeon of Budget & Analyss (Post 890205). January 1990.

general use of the NRC Staff and a available to the pubic. The 208pp. 9002120328. 52610.231, digest 6s omded 6nto two perts: the first presents an overview of The report contains the f scal year budget justifcahons to the U.S. Nuclear Regulatory Commission and the second pro-Congress. The budget provkies estimates for salares and on-vedes data on NRC commeret ' nuclear reactor hconse9s and penses and for the Offee of the inspector General for focal Commercial nuclear power reactors worldwde.

NUREG 1881: TECHNICAL SPECIFICATIONS, COMANCHE NUREG 1214 R06: HISTORICAL DATA

SUMMARY

OF THE SYS-PEAK STEAM ELECTRIC STATION, UNIT 1. Docket No. 50-TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE-445, Appendix *A" To License No. NPF 28.

• Comanche Peak ALLENSPACH,F.; NEASE.R. Dvision of Licensee Performanes Project Dev6sion. February 1990. 350pp. 9003120783.

& Ovahty Evaluaton (Post 870411). February 1990. 114pp.

52883:015.

9003070137. 52807:320.

The Techn6 cal Specifcatons for Comanche Peak Steam The Historcal Data Summary of the Systemate Assessment Electric Staton, Unit 1 were prepared by the U.S. Nuclear Reg-of Licensee Performance (SALP)is produced porodcally by the yi.ory Commiss6on. They set forth the hmits, operating Conde-U S. Nuclear Regislatory Commesbn. This summary provides tons, and other requ6rernents apphcable to a nuclear reactor fa.

the results of the assessment fot each facility by NRC reg 60n c6hty, as set forth 6n Secton 50.36 of Title 10 of the Code of and is further dmded into the following sectons: $0ction 1 pro.

Federal Regulatons Part 50, for the protecton of the health and sents the most recent SALP report ratings for facihtes in oper-safety of the pubic, ston and under constructon. These rabngs are grouped by Regon showing rating based on the revised Manual Chapter NOREG 1586: TECHNICAL SPECIFICATIONS FOR SEABROOK 0516 functonal areas, then the pre revised Manual Chapter STATION, UNIT 1. Appendix "A" To Ucense No. NPF 86.

• D-0516 functonal areas and then for reactors under constructon.

veion of Reactor Propets.1/ll (Post 970411). March 1990.

Secton 2 presents a chronological heting of all SALP report rat-400pp. 9004090200. 53307:001.

Ings for each operating feelhty. These reungs are also grouped The Seabrook Staton, Unit i Technical Specifcatone were by Regon showing ratings based on the revised Manual Chap prepared by the U.S. Nuclear Regulatory Commeston to set 1

ter 0516 functonal areas and then the pre-revised Manual forth the hrrdts, operating condibons, and other r0Querements ap-Chapter funchonal areas. Sechon 3 presents a chrono;vyc-picable to a nuclear reactor fac6hty as set forth in Secton 50.36 hohng of all SALP report ratings for each f acihty under construc-of 10 CFR Part 50 for the protecton of the health and safety of tion. For historical purposes, past constructon ratings for facih-the pubic.

les that recently have been hcensed also are hated in Secton JUNE 1988 GEORGIA RSI INCIDENT.

• Offee of Govemmental NUftEG 1232 V04: SAFETY EVALUATION REPORT ON TEN-

& Pubhc Affairs (Post 870413). SETSER.J.L Georgia, State of.

NESSEE VALLEY AUTHORITY: WATTS BAR NUCLEAR PER*

February 1990. 45pp. 9003190259. 53005:245.

FORMANCE PLAN. AULUCK.R. TVA Projects DWinion. January On June 6,1988, operators of a pool irradiator in Decatur, 1990. 84pp. 9003070159. 52809.041 Georgia were prevented by a safety system from raising That safety evaluaton report on the informat:on submitted by sources from ths pool. Radiabon levels of 60 milhrem per hour the Tannessee Valley Authonty in its Nuclear Performance Plan at the surf ace of the pool water were found, indcatrve of a lonk for the Watts Bar Nuclear Plant and in supporting documents of one or more of the 252 Cs 137 source capsules used at the has been prepared by the U.S. Nuclear Regulatory Commisson irradetor. Because of the concems whch arose out of this bci-staff. The plan addresses the plant specife corrective actons dont, the State of Georgia and the Conforerce of Radiation as part of the recovery program for iconsing of Unit 1. The staff Control Program Directors, Inc. decided it should be reviewed in wil: be monitoring and inspecting the 6mplementation of the pro-depth. Georgia Governor, The Honorable Joe Frank Hams, cre-grams. The plan does not address all heensing matters that will ated an incident Evaluaton Task Force and charged it with col-be required for fuel load and operabon of Unit 1. Those remain

• lecting information on the incident, maintaining commun6 cations ing hcensing matters have been addressed in previous safety with the DOE Investigative Board and preparing a wntten report evoluations or will be addressed in accordance with routine of lessons learned. Since the incident and responses to it are NRC heensing prachces-still ongoing a hnal report of the task force Is expected at a NUREG 1316: TECHNICAL FINDINGS AND REGULATORY later date. A summary of the Task Force's First Intenm Report ANALYSIS RELATED TO GENERIC ISSUE 70 Evaluaton Of has been prepared for persons needing an overview of the inci-Power Operated Rehef Valve And Block Valve Reliability in dont and lessons leamed to date. The Conference estabhshed PWR Nuclear Power Plants. KIRKWOOD.R. Division of Safety an incident Review Team whch agreed to assume the responsb Issue Resoluton (Post 880717). December 1989. 26pp.

bihty from the Georgia task force to discuss the role of the States in regulating irradiators. Its Intenm Report provides a 90021P0f 48. 52611:263.

This report summanzes work periormed by the Nuclear Regu.

summary of Agreement States' views and recommendations an latory Commission staff to resolve Genenc issuc 70, "Techncal some of the issues raised by the incident.

Findings and Regulatory Analysis Related to Genere issue 70

• NUREG-1396 DRFT: INDUSTRY PERCEPTIONS OF THE Power Operated Rehet Valve and Block Valve Reliabihty." The IMPACT OF THE U.S. NUCLEAR REGULATORY COMMISSION report evaluates the reliability of PORVs and block valves and ON NUCLEAR POWER PLANT ACTIVITIES. Draft Report.

their safety signifcance in PWR nuclear power plants. The DAVIS,A.B.; PEDERSON.C.D. Region 3. Ofc of the Director.

report identifies those safety related functions that may be per.

March 1990.186pp. 9004090218. 53313:190.

formed by PORVs and describes ways in whch PORVs and Teams of senor managers from the NRC surveyed Icensee block valves may be improved. This report also presents the staff members representing 13 nuclear power utihties from regulatory analysis for Genere issue 70.

across the country to obtain their candad views of the effective-NUREO 1360 V02: NUCLEAR REGULATORY COMMISSION IN-ness and impact of NRC regulatory activities. Licensee com-FORMATION DIGEST.1990 Editon. OLIVE KL Division of monts addressed the full scope of NRC activtties and the Budget & Analysis (Post 890205). March 1990. 110pp.

impact of agency actions on heensee resources, staff perform.

l 9004090296. 53315:278 ance, planning and scheduhng, and organizational effecttve-l The Nuclear Regulatory Commission information Digest pro-ness. The pnncipal themes of the survey respondents' com-l vides summary information regarding the U.S. Nuclear Regula-ments are that (1) hcensees acqu:esce to NRC requests to tory Commission, its regulatory responsibihties, and areas h-avoid poc? ratings on NRC Systemate Assessment of Licensee r

l

4 Main Citations and Abstracta Performance (SALP) reports and the consequent financial and NUREG/CP.0110: PROCEEDINGS OF THE INTERNATIONAL pubhc percepbon problems that result, even if the requests re-WORKSHOP ON NEW DEVELOPMENTS IN OCCUPATIONAL Quire the expendsture of signifcant resources on matters of mar-DOSE CONTROL AND ALARA IMPLEMENTATION AT NUCLE-96nal safety sigruicance, and (2) HRC so dominates licensee re.

AR POWER PLANTS AND SIMILAR FACILITIES. BAUM.J.W.;

sources through its existing and changing formal and informal DIONNE.B.J.; KHAN.T.A. Brookhaven Natonal Laboratory. Feb-requirements that beensees belove that their plants, though not rua'y 1990. 600pp. 9004090237. BNL NUREG 52226.

unsafe, would have better reliability, and may even acheve a 53312:016.

higher degree of safety, if beensees were freer to manage their TNs report contains summares of papers and discussions own resources. This draft report does not attempt to defend any presented at the Intematonal Workshop on New Developments NRC positon; endorse or refute beensee perceptions; or explain in Occupatonal Dose Control and ALARA implementaten at any actiun taken by NRC in tuffilhng its responsibihtes to protect Nuclear Power Plants and Similar Facilites bek! at Brookhaven the health and safety of the public. Senior NRC managers have National Laboratory, Upton, New York September 16 21.1969.

made a prehminary evaluation of the information in tnis report and have made recommendatons to address bcontee concems The purpose of tNs workshop was to bring together scientists, engineers, regulators, and administrators who ars involved with 6n some areas. The final evaluaton and recommendatons will occupatonal dose control at nuclear facilites to exchange infor-be pubhshed at a later date as the finst NUREG.

mation on recent developments from their countnes. The eleven NUREQ/CP 0106 V01: PROCEEDINGS OF THE SEVENTEENTH countnes represented included: Canada, Finland, France, Ger-many, italy, Japan, Luxembourg Sweden, Switzerland, the WATER REACTOR SAFETY INFORMATION MEETING.

United Kingdom, and the United States of Amer 6ca. TNs work-WEISS.A.J. Brookhaven National Laboratory. March 1990.

606pp. 9004030082. 53228:322.

shop was sponsored jointly by the U.S. Nuclear Regulatory This three volume report Contains 64 papers out of the 111 Commission and the U.S. Department of Energy,in cooperaton with the Organization for Economic Cooperation and Develop-that were presented at the Seventeenth Water Reactor Safety ment, Nuclear Energy Agency.

Inforfatton Meeting held at the Hohday Inn Crowne Plaza, Rockville, Maryland, during the week of October 23 25, 1989.

NUREQ/CR 2000 V00N12: LICENSEE EVENT REPORT (LER)

The papers are pnnted in the order of their presentaton in each COMPILATION.For Month Of December 1989.

• Oak Ridge Na-sosslon and describe progress and results of programs in nucle.

tional Laboratory. January 1990. 70pp. 9002120164. ORNL/

j ar safety research conducted in this country and abroad. For.

NSIC.200. $2611:131.

j eign partcipaton in the meeting included ten different papers This monthly report contains Licensee Event Report (LER) presented by researchers from France, Germany, Japan, operatonal information that was processed into the LER data Norway and the United Kingdom. The titles of the papers and file of the Nuclear Safety informaton Center (NSIC) dunng the the names of the authors have been updated and may differ one month perod identified on the cover of the document. The trom those that appeared in the final program of the meeting.

LERs, from whch this information is derived, are submitted to the Nuclear Regulatory Commission (NRC) by nuclear power NUREQ/CP 0106 V02: PROCEEDINGS OF THE SEVENTEENTH plant heensees in accordance with federal regulations. Proce-WATER REAC10R SAFETY INFORMATION MEETING.

dures for LER reporting for revisions to those events occurring WEISS,A.J. Brookhaven Natonal Laboratory. March 1990.

prior to 1984 are described in NRC Regulatory Guide 1.16 and

$25pp. 9004090184. 53310.211, NUREG 0161, "Instructons for Preparaton of Data Entry See NUREG/CP 0105,V01 abstract.

Sheets for Licensee Event Reports." For those events occurring on and after January 1,1964 LERs are being submitted in ac-NUREQ/CP-0106 V03: PROCEEDINGS OF THE SEVENTEENTH cordance with the revised rule contained in Title 10 Part 50.73 WATER REACTOR SAFETY INFORMATION MEETING.

of the Code of Federal Regulatons (10 CFR 50.73 Licensee WEISS,A.J, Brookhaven Natonal Laboratory. March 1990.

Event Report System) which was published in the Federal Reg-550pp. 9004090193. 53308:060.

See NUREG/CP 0105,V01 abstract.

inter (Vol. 48, No.144) on Juty 26,1983. NUREG 1022, U-consee Event Report System. Descripton of Systems and NUREG/CP-0109: PROCEEDINGS OC THE SEMINAR ON LEAK-Guidelines for Reporting," provides supporting guidance and in-BEFORE. BREAK.Further Developments in Regulato*y Pohcies formation on the revised LER rule. The LER surnmaries in this And Supporting Research. WILKOWSKl,G.M. Battelle Memonal report are arranged alphabetcally by facihty name and then institute, Columbus Laboratories. CHAO K. S. Taiwan Power Co.

chronologically by event date for each facihty. Component, February 1990. G54pp. 9003190303. 53043:231.

system, keyword, and component vendor indexes follow the The fourth in a senes of International Leak Before-Break summanes. Vendors are those Identif ed by the utlhty when the (LBB) Seminars supported in part by the U.S. Nuclear Regula-LER form is initiated, the keywords for the component, system, tory Commission was held at the Natonal Central Library in and general keyword indexes are assigned by the computer Taipei, Taiwan on May 11 and 12,1989. The seminar updated using correlaton tables from the Sequence Coding and Search System.

the Intemational polcies and supporting research on LDB. At-tendees included representatsves from regulatory agencies, NUREG/CR 2000 V09 N1: LICENSEE EVENT REPORT (LER) electric utihty representatries, nuclear power plant fabreators, COMPILATION.For Month Of January 1990.

• Oak Ridge Na-research organizations, and academic institutions. Regulatory tonal Laboratory. February 1990.109pp. 9004090292. ORNL/

policy was the subject of presentatons by Mr. G. Arlotto (U.S.

NSIC 200. 53315:166.

NRC, U.S.A.), Dr. B. Jarman (AECB, Canada), Dr. P. Milella See NUREG/CR 2000,V08,N12 abstract.

(ENEA DISP, itaty), Dr. C. Faidy (EDF/Septen, France), and Dr-K Takumi (NUPEC, Japan). A paper by Mr. K. Wichman and Mr.

NUREG/CR 2000 V09 N2: LICENSEE EVENT REPORT (LER)

S. Lee of the U.S. NRC Offee of Nuclear Reactor Regulation is COMPILATION.For Month Of February 1990.

• Oak Ridge Na-ancluded as background material to these proceedings; it dis-tional Laboratory. March 1990.120pp. 9004090225. ORNL/

cusses the history and status of LBB apphcatons in U.S. nucle-NSIC 200. 53314:122.

I ar power plants. In addition, several papers on the supporting See NUREG/CR.2000,V08,N12 abstract research programa desenbod regulatory policy or industry stand-NUREG/CR 2331 V09 N3: SAFETY RESEARCH PROGRAMS j

ards for flaw evaluatons, e g., the ASME Secton XI code pro.

SPONSORED BY OFFICE OF NUCLEAR REGULATORY cedures. Supporting research programs were reviewed on the RESEARCH. Progress Report. July September 1989. WEISS A.J.

first and second day by several perteipsnts from Taiwan, U.S.,

Brookhaven Natonal Laboratory. February 1990. 104pp.

l Japan, Canada, Italy, and France.

9003070194. BNL NUREG.51454. 52812.282.

Main Chations and Abstracts 6

This progress report describes current actwtties and techn6 cal nsk are dertved largely from informaton summanzed in BEIR progress in the progrems at Brookhaven Natonal Labork;ory ill. with some adlustment to reflect more recent studies. The aponsored by the Dwison of Regulatory Appleatons, Dwison effect of the revmed dosimetry in Hiroshima and Nagasaki has of Engineenng, Dwison of Safety issue Resoluton, and Dwision not been conadored. Linear snd knear-Quadrate models are of Systems Research of the U.S. Nuclear Regulatory Commis-also recommended for assessin.g genetic nsks. Five classes of son, Offee of Nuclear Regulatory Research following the roof-genetic disease-dominant, x-hrked, ancuploidy, unbalanced genizaton in July 1988. The previous reports have covered the trar.slocahons and muttifactonal diseases-are consdered. In pered October 1,1976 through June 30,1989.

additon, the 6mpact of rachaton induced genetc damage on the 6nedence of porkimplantaton embryo losses is discussed. The NUREQ/CR 3145 V08: GEOPHYSICAL INVESTIGATIONS OF uncertainty n modehng radiologcal nealth noks is addressed by i

THE WESTERN OHIO.lNDIANA REGf0N. Annual Providing central, upper, and tower ostimmtes of all model pa-Report. October 1988. September 1989. YOUNG.C.J.; LAY,T,;

remeters. Data are provided which thould enable analysts to JACOBSON.J. Mchigan, Unty, of, Ann Arbor, Mt. February consider the timing and seventy of each type of health nsk.

1990. 62pp. 900307P54. 52802.257, Earthquake actMty in the Western Oho Indiana regon has NUREG/CR-4469 V08: NONDESTRUCTIVE EXAMINATION been monitored with a precision seismograph network consist-(NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS. Semiannual Report,0ctober 1987 March ons loca ed in ana al eart a a e been 1988. DOCTOR,S.R. DEFFENBAUGH,J.; GOOD.M.S.; et al.

corded during this report penod, with magrutudes ranging from Battelle Memorial institute, Pacife Northwest Laboratory. Octo-0.7 to 2.2. This may represent a return to a more normal local bor 19% 67pp. 9002t20167, PNL-5711,52tiO3.122.

level of seismcity after an anoma%usty low level in the 2 years The Evaluaton and improvement of NDE Rehabikty for in-following the occurrence of the m(b) = 4.6 earthquatte, in St.

servce inspecton of Light Water Reactors (NDE Rehability)

Marys, Ohio on July 12,1986. If the quiescent rerod reflected Program at the Pacific Northwest Laboratory wcs estabhshod by a release of most of the accumulated crustat strain by the St.

the Nuclear Regulatory Commissbn to determine the reliability Marys event, then the increased level of seismcity mky mark of current nservce inspecton (ISI) techniques and to develop the beginning of a new phase of strain accumulaton. Four re.

recommendatons that will ensure a suitably high inspection reh-gonal events were also well recorried by the array statens abihty. The objectrves of this program 6nclude determining the during this year. All of these events occurred in regens with rehabihty of ISI performed on the pnmary systems of commer-well-estabhshed histones of seismcity. Their magnitudes range cial hght water reactors (LWRs), using probabiliste t acture me-from 3.1 to 4,4.

chancs analysis to determine the irepact of NDE unrohability on NUMEG/CR 3668: MINET CODE DOCUMENTATION. VAN eystem safety; and evaluating rehability improvements that can BNL r,Nat onal be acNeved with improved and advanced technology. A finaf TUYLE,0.J.; NEPSEE T.C.: GUPPY,J G. Drookhaven wnEG' obschve is to formulate recommended revisions to MME Code Laboratory. December 19t\\9. 321pp. 9002120129.

61742, 52582.219.

and Regulatory requirements, based on matenal properties.

The MINET computer code, developed for ;ne transient anaty' servce conditions, and NDE capabilibes end uncertainties. The ses of fluid flow and heat transfer, is documented in tNs four-program scope is hmated to ISI of the pnmary systems including part reference. In Part 1, the MINET models, wNeh are based the piping, vessel, and other insoected components. This is a on a momentum integrat network method, are described. The progress report covenng the programmate work from Octobc-various aspects of utilizing the MINET code are discussed in 1987 through March 1988~

Part 2, The User's Manual. The tNrd part is a code desenption, detalhng the basic code structure and the various subrouhnes NUREG/CR 4469 VD9: NONDESTRUCTIVE EXAMINAllON and functions that make up MINET. In Part 4, example input (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT decks, as well as recent vahdation studies and apphcations of WATER REACTORS. Semiannual Report. April-September 1988.

MINET are summartred.

DOCTOR,S.R.; DEFFENBAUGH.J.; GOOD,M.S.: et al. Dattelle NUREQ/CR-4214 R01 Pl: HEALTH EFFECTS MODELS FOR NU.

Memonal Institute, Pacife Northwest Laboratory. November CLEAR POWER PLANT ACCIDENT CONSEOVENCE 1989.115pp. 9002120206. PNL 5711. 52597:100.

ANALYSIS. Low LET RadiatonPart I:Introducton, integraton Evaluation and improvement of NDE Rehabihty for inservice inspechon of Light Water Reactors (NDE Rehability) Program at And Summary. EVANS,J.S. Harvard School of Pubhc Health, the Pacife Northwest Laboratory was estabhshed by the Nucle-Boston, MA

• Sandia National Laboratones. January 1990.

133pp. 9002120313. SAND 85 7185. 52597:294.

ar Regulatory Commiss6on to determine the rehabihty of current TNs report provides dose-response models intended to be inservce inspechon (ISI) techniques and to develop recommen-used in estimating the radiological health effects of nuclear dabons that will ensure a suitably high inspection reliabihty. The power plant accidents. Models of earty and continuing effects, objectives of tNs program include determining the rehabikty of cancers and thyroid nodules, and genetic effects are provided.

ISI performed on the pnmary systems of commercial hght water Two parameter Weibull harard functons are recommended for reactors (LWRs); using probabiliste fracture mechanes analysis estimahng the nsks of earty and continuing health offects. Three to determine the impact of NDE unrohabihty on system safety; potentially lethal early effects-the hematopoiebc, pulmonary and evaluating rehabihty improvements tha; can be achieved end pastrointestinal syndromes-are considered. In addition, witn improved and advanced technology. A final objective is to models are provided for assessing the risks of several non-formulate recommended revisions to ASME Code and Regula-lethal early and continuing effects-including prodromal vomiting tory requirements, based on material properties, soryce condi-and diarrhea, hypothyroidism and radiaton ti,yroiditis, skin tions, and NDE uncertainties. The program scope is hmite:I to burns, reproductive effects, and spontaneous abortons. Linear ISI of the pnmary systems including the piping, vessel, and and knear quadratic models are recommended for estimahng other inspected components. TNs is a progress report covenng cancer risks. Parameters are given for analyzing the risks of the programmatic work from April 1988 through September seven types of cancer in adults-leukemia, bone, lung breast, 1988.

gastrointestinal, thyroid and "other", The category "other" can-NURECOR 4550 V01 R1: ANALYSIS OF CORE DAMAGE FRE.

cers, is intended to reflect the combined risks of muftiple mye-loma. fymphoma, and cancers of the bladder, kidney, brain.

QUENCY:

INTERNAL EVENTS METHODOLOGY.

ovary, uterus and cervix. Models of childhood cancers due to ER!CSON.D M. ERC Environmental & Energy bervces, Inc.

"in utero" exposure are also provided. For most cancers, both WHEELER,T.A.; SYPE.T.T.; et al. Sandia Nabonal Laboratones.

incidence and mortahty are addressed. The models of cancer January 1990. 482pp 9002120309. SAND 86 2084. 52576:007, u

1 6

Main Citations and Abstracts N')t4EG 1150 examines the risk to the pubhc from a selected local vibraton is excited dunng the puncture process. Further group of nuclear power plants. This report describes the meth.

Study of this local vibraton is needed.

odology that evolved as the intomal event core damage tre-Oversies for four plants were generated in support of NUREG.

NUMEG/CR 4666 V03: SEISMIC FRAGILITY OF NUCLEAR 1150. The objectrve is to perform an analysis that closely ap.

POWER PLANT COMPONENTS (PHASE II).Switchgear,14C prox 6 mates a etsteof the-art Level i Probabihstc Rtsk Assess.

Panels (NSSS)

And Reisys.

BANDYOPADHYAY; ment (PRA). Therefore, in pnncole, it is similar to those used in HOFMAYER.C.H.; KASSIR.M.K.: et al. Brookhaven Natonal previous PRAs. However, this methodology, based upon prev 6 Laboratory. February 1990. 73pp. 9004030128. BNL.NUREG-ous stuees and using analysts experienced in these techniques, 52007. 53228:101, allows the analysis to be focused upon selected areas. With this As part of the Component Frag 6hty Program which was initiet.

approach only the most important systems and failure modes ed in FY 1986, three adetonal equipment classes have been are emphasized and modeled in detall, and the data and human evaluated. This report contains the fragility results and 6scus.

reliability anstyses are simphfed. An analysis e'nploying this sions on these equipment classes which are awltchgear,180 methodology (exclusive of external revews) can be completed panels and reisys. Both iow and me$um voltage switchgear es.

In nine to twelve months uhing two or three full time experi.

sembhes have been Considered and a separate fragility esti-enced systems ana'ysts and part time personnel in other areas, mate for each type is provided. Test data on cabinets from the such as data analyhis and human rehabill'y analysis. This is sig, nuclear instrumentation / neutron monitoring system, plant / proc.

nifcantly faster and less expenstve than provous analyses, but ess protecten system, sohd state protecttve system and engl.

even so, most of the insights that are obtained by the more ex.

neered safeguards test system Compnoe the BNL data base for pensive studies are still provided.

l&C panels (NSSS). Fragihty levels have been determined for various failure modes of switchgear and l&C panels, and the de-NUMEG/CR-4664 V06: SCANS (SHIPPING CASK ANALYSIS terministe results are presented in terms of test response spec.

SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM tra. In additon, the test data have been evaluated for estimatmg FOR SHIPPING CASK DESIGN REVIEW. Volume 6 Theory the respectrve probabikste fragihty levels which are expressed Manual Buckhng Of Circular Cylindrcal Shells. LO.T.; MOK,0.C.;

in terms of a median value, an uncertainty coeffcient, a ran-CHINN.D.J. Lawrence LNormore Natonal Laboratory. February domness coefficient and an HCLPF value. Due to a wide varia.

1990. 76pp. 9004030139. UCID-20674. 53228:173.

ton of relay design and the fragility level, a genene fragibty level A c0mputer system called SCANS (Shipping Cask ANatysis cannot be estabhshed for relays, System) is being developed for the statt of the U.S. Nuclear Regulatory Commission to perform confirmatory hcensing review NUREG/CR 4661: CLOSEOUT OF IE BULLETIN 85-03: MOTOR.

analy^.es. SCANS can handle problems associated with impact, OPERATED VALVE COMMON MODE FAILURES DURING heat transfer, thermal stress, internal or external pressure loads, PLANT TRANSIENTS DUE TO IMPROPER SWITCH SET.

and lead slump. A new capabihty implemented in SCANS is TINGS. FOLEY,W.J.; DEAN.R.S.; STEINBRECHER H.; et al. PA.

buckhng analysis of the steel shells of a Spent fuel shipping RAMETER, Inc. February 1990.150pp. 9003120769. PARAME.

cask during a postulated impact with an unyieldireg suriace, TER IE158. 52882:267.

Three sets of buckhng analysis formulas are included: (1) Code Documentaton is provided in this report for the cloteout of IE Case N 284 of the ASME Boiler and Pressare Vessel Code, (2)

Bulletin 85 03. The purpose of the bulletin was to request h-American Petroleum inst!tute Bullehn 2U an upgrade of N 284 censees to develop and 6mplement programs to ensure that the that includes test results avallatie after N 284 was written in switch settings on certain safety related, motor-operated valves 1979, (3) formulas frequently used by the piping and pressure are selected, set, and maintained so as to ensure reliab'e valve vessel industry and formulas recommended by the Structural operaton when valves are subjected to maximum differential Stabikty Researcti Council. To be compatibie with the ASME pressures expected during normal operation and the design Code, the first set is focommended for use in shipping cask basis events. The report includes documentation and status of evaluation and this set is implemented in SCANS. The second the review of Action item e of the bulletin completed up to the and third sets are recommended references for SCANS users.

issuance of Generic Letter 8910 on June 28, 1989. Ucensee actons for Action item o were completed satisfactorily for 102 NUREG/CR 4654 V07: SCANS (SHIPPING CASK ANALYSIS (86%) of the 119 facihtees for which actons were required. Sat.

SYSTEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM isfactory completion of the remaining 17 facihties is enwred by FOR SHIPPING CASK DESIGN REVIEW. Volume 7: Theory the genede letter. In the report conclusions, the inter relation-Manual Puncture Of Shipping Casks. LO,T. Lawrence Livermore ship of the generic letter and the bulletin is presented. Back-Natonal Laboratory. February 1990. 65pp. 9004030060. UCID-ground information is supp' led in the introducton and Appendix 20674. 53226:100.

A.

Under current regulatory requirements, a shipping cask should be designed for a senos of hypothetscal eccident condi-NUREG/CR 4668: DAMAGED FUEL EXPERIMENT OF 1.Results tons. These test conditions include a 40-inch free drop of the And Analyses. GASSER,R.D.; FRYER,C.P.; GAUNTT R.O.; et al.

cask onto a 6-inch diameter puncture pin. In this Study, existing Sandia National Laboratories. January 1990. 269pp.

punctre test data was examined. Simple formulas based on 9003120772. SAND 861030. 52884:009.

test data were proposed for puncture evaluaton of shipping A senes of in pile expenments addressing LWR severe fuel casks. Dynamic and static nonhnear finite elemer<t analyses damage phenomena has been conducted in the Annular Core were performed to correlate anaiyses with the existing test data.

Research Reactor (ACRR) at Sandia Natonal Latoratones. The in this analytcal approach, three puncture failure predicton ACRR Debris Formaton and Relocation (DF) experiments are methods were proposed and their apphCability was evaluated.

Quasi. separate effects tests that provide a data base for the de-The analytical approach provides an alternatrve to teshng Doth velopment and venficaton of models for LWR severe core laminated and sohd wall shipping casks were analysed. In the damage accidents. The first expenment in this series DF 1, was study of laminated casks, the effects of the inner shell on the performed on March 15,1984, and the results are presented in puncture of the outer shell were examined, as were the effects this report. The DF 1 experiment examined the effects of low of matenal strength of the puncture pin. The study of geometric initial clad oxidation conditions on fuel damage and relocahon scahng of casks indicated that the normah2ed incipient puncture processes. The DF 1 test assembly consisted of a nine rod energy is insensitive to vanatons in the scale factors. This cord square-matrix bundle that employed PWR type fuel rods with a closion indicates that the proposed artatytical approach of Com-0.5 m fissile length. The fuel rods were composed of 10% en-bin,ng finite element analysis and failure prediction methods is nched UO(2) pellets within a rircaloy 4 cladding. Steam flowed Consistent with the similanty principles in tests, except when through the test bundle at flow rates varying between 0.5 and 3 l

1 I

I Main Citations and Abstracts 7

g/s, and the ACAR maintained a peak power level of 1.5 MW NUREG/CR 4691 V02: MELCOR ACCIDENT CONSEQUENCE dunng the high temperature oxdaten phase of the test inducing CODE SYSTEM (MACCStvolume 2: Model Desenptiort about N8.5 kW fission power and about ~20 kW peak oxidation JOW,H.; SPRUNG,J.L.; ROLLSTIN,J.A.; et al. Sandia Natonal power in the assembly. Visual observaton showed ear'y clad re-Laboratones. February 1990.183pp. 9003193289. SAND 86-locaton and partial blockage formaton at the gnd spacer loca-1562. 53005:059 ton accompanied by producton of a dense serosof. Posttest This report desenbos the MACCS computer code. The pur-cross sectons show 160uefaction losses of fuel in excess of 10 pose of this code is to simulate the impact of severe accidents volume percent, as well as large fractional losses of cladding at nuclear powet plants on the surrounding environment.

matenal trom the upper two-thirds of the bundle. The quantity of MACCS has been developed for the U.S. Nuclear Regulatory hydrogen measured dunng the test was consistent with the ob-Commission to replace the provous CRAC2 code and it incor-served magnitude of ciaoding oxidaten. Oxidation dnven heat-porates many improvements in modeling flexibility in companson ing rates of 25 K/s and peak temperatures in excess of 2525 K to CRAC2. The pnnespal phenomena considered in MACCS are were observed. The analyses, interpretateon, and appication of atmospheric transport, mi$ fee actions based on dose projec-these results to severe fuel damage accidents are discussed-ton, dose accumulaton by a number of pathways including food NUREG/CR 4471: THE DF 4 FUEL DAMAGE EXPERIMENT IN and water ingeston, earty and latent health effects, and eco.

ACRR WITH A BWR CONTROL BLADE AND CHANNEL BOX.

nomic costs. The MACCS code can be used for a variety of ap-GAUNTT,R.O.; GASSER R.D. Sandia Natonal Laboratones.

pications. These include (1) probabikste risk assessment (PRA)

OTT,L.J. Oak R dge National Laboratory. November 1989.

of nuclear power plants and other nucleai facilities, (2) sensitivi-j 39?pp. 9003070275. SAND 861443. 52806.283.

ty studies to gain a better understanding of the pararneters im.

The DF 4 test was an expenmental investigation into the melt portant to PRA, and (3) cost benefit analysis. This report is progression behavior of boiling water reactor (BWR) core com-composed of tnree volumes. Volume I, the User's Guide, de-ponents unoer high temperature severe core damage condi-scribes the input data requirements of the MACCS code and tons. In this study 14 rircaloy clad UD(2) fuel rods, and repre-provides directions for its use as illustrated by three sample sentatons of the rircaloy fuel canister and stainless steel /B(4)C problems Volume II, the Model Desenption, describes the un-control blade were assembled into a 0.5 m long test t:andle, dertying models that are implemented in the code, and Volume The test bunale was fission heated in a flowing steam environ-Ill, the Programmer's Reference Manuat, describes the code's ment, using the Annotar Core Research Reactor at Sandia Lab-structure and database management.

oratones, simulating the environmental conditions of an uncov-NUREG/CR 4691 V03: MELCOR ACCIDENT CONSEQUENCE ered BWR core exponencing high temperature damage as a CODE SYSTEM (MACCS). Volume 3: Programmer's Reference result of residual fisson product decay heating. The experimen.

Manual. ROLLSTIN,J.A.; CHANIN D.l.; JOW,H. Sandra Natonal tal results provide information on the thermal response of the test bundle components, the rapid exothermc oxdaten of the Laboratones. February 1990. 338pp. 9003190293 SAND 86-1562. 53042:250.

gircaloy fuel cladding and canister, the production of hydrogen from metal steam oxidaton, and the failure behaver of the pro-See NUREG/CR 4691,V02 abstract.

grossively melting bundle components. This information is pro-vided in the form of thermocouple data, steam and hydrogen NUREG/CR-4704 V03: RELATIVE DIOLOGICAL EFFECTIVE-flow rate data, test bundle fiss4on power data and visual obser-NESS (RBE) OF FISSION NEUTRONS AND GAMMA RAYS AT vaton of the damage progression. In addition to BWR back-OCCUPATIONAL EXPOSURE LEVELS. Studies On The Gross ground informaton, this document contains a desenpton of the And Microscopic Pathology Observed Al Death Of Mice Ex.

expenmental hardware with details on how the experiment was posed To 60 Equal Once Weekty Doses Of Fission....

Instrumented and diagnosed, a desenption of the test progres-GRAHN.D.; THOMSON,Jf.; CARNES B.A. Argonne Natonal son, and a presentaton of the on-line measurements. Also in Laboratory. January 1990, 111pp. 9004090170. ANL 86 33, this report are the results of a thermal anatysis of the fueled

$3314:010.

test section of the expenment demonstrating an overall consist-Dose-response analyses of the pathologc consequences of ency in the measurable quantities from the test. A discussion of exposure to 60 equal once weekly doses of fission neutrons or the results is provided.

60C0 gamma rays were used to generate relative bological ef-fect'veness (RBE) values for the major causes of death or for NUREG/CR 4674 V09: PRECURSORS TO POTENTIAL SEVERE tumor occurrences. Cumuistrve probabilities of death or occur.

CORE DAMAGE ACCIDENTS:f 988 A STATUS REPORT. Main rence were generated for 15 categories of neoplastic disease in Report And Appendix A. MINARICK,J.W.; CLETCHER.J.W.;

the interva! 800 999 days since first exposure for each dose, DLAKE A.A. Oak Ridge National Laboratory. February 1990.

sex, and radiation quality. Depending upon the pathologic end-169pp.9003070234. ORNL/NOAC 232. 52810.323 point, RBE values vaned from about 10 to 50. Vanation in RBE Thirty two operatonal events with conditonal probabillbes of values was not random or nonsystematic. RBE values were core damage of 1.0 x 10( 6) of higher occumng at commerciat lower for tumors involving the connective tissues (about 10 to light water reactors dunng 1988 are considered to be precut-

15) than for the epithelial t:ssue tumors (about 20 to 50). Fe-sors to potent:al severe core damage. These are desenbed males did have significantly higher nsk coeffeients for many along with associated signifcance estimates, categonzaton, endpoints. The results suggest that the RBE value of 20 for hte and subsequent analyses. This study is a cordnuaton of earher Stortening from all causes of death is a weighted average of work, whch evaluated the 1969-1981 and 19841987 events.

the RBE values representing the individual neoplaste diseases The report discusses (1) the general rationale for this study, (2)

(approximately 90% of the excess nsk of mortality induced by the selecton and documentaton of events as precursors. (3) rad ation at low doses). These results in conjunction with the re.

the estimation and use of conditional probabilit:es of subso-suits of large-scale studies funded by the Department of Energy quent severe core damage to rank precursor events, and (4) the and using the same facilities, support the concluson that the plant models used in the evaluaton process.

quality factor (O) for fission neutrons should be raised from the NUREG/CR-4674 V10: PRECURSORS TO POTENTIAL SEVERE present value of 10 to a value of 20.

COhd DAMAGE ACCIDENTS:1968 A

STATUS REPORT. Appendixes B

And C.

MINARICK,J W.;

NUREG/CR-4731 V02: RESIDUAL LIFE ASSESSMENT OF CLETCHER.J.W,; BLAKE.A.A. Oak Rdge Natonal Laboratory.

MAJOR LIGHT WATER REACTOR COMPONENTS OVER.

February 1990. 473pp. 90030701B4 ORNL/NOAC 232.

VIEW. SHAH,V.N.; MACDONALD.P.E.; AMAR.A.S.; et al. EG&G i

52813 073.

Idaho, Inc. (subs of EG&G, Inc.). November 1989. 467pp.

See NUREG/CR 4674,V09 abstract 9002120329 EGG 2469. 52578.023.

1

\\

n 1

8 Main Citations and Abstracts This report presents an assessment of the aging (time-d&

ing the qualifcation processes on a national (rather than local pendent degradation) of selected major light water reactor com-employer) basis ponents and structures The stressors, pc 'able degradation sites and mechanisms, potential failure rhodes and currer.t in-NUREG/C45169: MO'3ILIZATION AND TRANSPORT OF URA-service inspection requirements are discussed for eleven major NIUM AT URANIUM MILL TAILINGS DISPOSAL hght water reactor components reactor coolant pumps, pressur-SITES Apphcaton Of A Chemcal Transport Model ized water reactor (PWR) pressunzers, PWR pressurtzer surge ERGN R L ; HOSTETLER.C J ; KEMNER.M L Battelle Me-and spray lines, PWR reactor coolant sys*em charging and monet institute, Pacsfec Northwest Laboratory January 1990.

safety infecten nozzles, PWR feedwater lines, PWR control rod 116pp 9002120263 PNL-7154 52596.210.

dnve mechanisms and reactor internals, besling water reactor (BWR) containments, DWR teedwater anc main steam lines' The geochemca! processes of aqueous specifcation, precipe.

BWR control rod dnve mechanisms and reactor internals, elec-tation, dissolution, and adsorpton influence the transport of ura-incal cables and connections, and emergency diesel genera-neum at uranium mill tailings disposal sites Traditional transport tors Unresolved technical issues related to understanding and models involve the use

• a single parameter, the retardation managing the aging of these maior components are identified Single parameter models are most appicable to field situations NUREG/CR-4744 V03 N1: LONG-TERM EMBRITTLEMENT OF exhibiting no changes in major element chemistry along the flow CAST DUPLEX ST AINLESS STEELS IN LWR path. Because of the changes in major element chemistry that SYSTEMS Semiannual Report, October 1987 March 1988 CHOPRA,0X; CHUNG H M Argonne National Laboratory Feb-occur when acidic leachate contacts a neutralizing soil, a single ruary 1990 58pp. 90030702** AHL 89/22 52802198 parameter transport model cannot accurately capture the details Tt. progress repon

-n20s work performed by Argonne of uranium migration td a number of disposal sites We have g

g ggg National Laboratory t wm embattlement of cast duplex stainless steels in LW,, oms dunng the six mo..

, Oc-g m ithe result is a generalized conceptual model that can reproduce the tober 1987 to March 1986 A mechanistic unde,

activation energy of aging is descobed on the ; &st no re.

features observed at a number of uranium mill tailings disposal sults of microstructural charactenzation of vai v

, oats of sites Grades CF 3, CF-8. and CF 8M stainless steel thai.:ere used in aging 6tudies at different laboratones The kinetics of the NUREG/CR 5229 V02: TMI 2 EPICOR-il RESIN / LINER INVESTI-spinoda decomposition of femte (i e, the primary mechanism of GATION LOW LEVEL WASTE DATA DASE DEVELOPMENT agine, embnttlement) appear to be strongly safhsenced by a syn.

PROGRAM FOR FISCAL YEAR 1989 Annual Report.

ergoc effect of G-phase nuc6eation and grown Wnen the acts MCCONNELL,J W., ROGERS R D EG&G Idaho, Inc. (subs of vation energies (ranging from 18 to 50 kcal/ mole) were plotted EGAG, 'nc ) DAVIS.E.C., et al Oak Ridge National Laboratoiy as a funcLon of the volume traction of G phase produced dunng February 1990 52pp 9003120764. EGG-2577 52882124 accelerated aging. a good conelation was obtained regardless The EPICOR Il Resin / Liner investigation Low-Level Waste c' vanations in grade, bulk chemical composition, and fabnca-Data Base Development Program, funded by the U.S Nuclear tion process Spinodal-like decomposition of austenite in heats Regulatory Commission (NRC), is studying tne degradation ef-containing a relatively high level of Ni has also been investigat-fects in EPICOR il organic son exchange resins caused by radi-od. Charpy impact data for thermally aged cast sta:niess steel ation; examining the adecuacy of tost procedures recommend-were analyzed to determine the kinetics and extent of embnttle-ed in the Branch Technical Position on Waste Forms to meet m?nt The fernte morphology had a strong effect on the extent of embnttlement, whereas the matenal composition influenced the requirement 6 of 10 CFR 61 using solidified EPICOR-il resin, the kinetics of embntt;ement Results obtained from the present obtaining performance information on sohdified EPICOR ll son study of mechanical properties, and data of other investigators exchange resins in a disposal environment, and determining the were analyzed to develop the procedu e and correlations for condition of EPICOR-Il liners This report summanzes accom-predicting tne kinetics and extent of embnttlement, under reac-plishments of Fiscal Year 1989.

tor operating conditions, from the matenal parameters NUREG/C45256: COMPONENTS OF AN OVERALL PERFORM.

NUREG/CR 4882: OUALIFICATION PROCESS FOR ULTR ASON-ANCE ASSESSMENT METHODOLOGY DAVIS.P. A,

IC TESTING IN NUCLEAR INSERVICE INSPECTION APPLICA-PRICE,L.L., WAHl,K.K., et al Sandia National Laboratones Feb-TIONS SPANNER,J.C : DOCTOR,S R., TAYLOR T T, et al Bat-ruary 1990.101pp 9003190287. SAND 88-3020 53012:285 telle Memonal institute, Pacific Northwest Laboratory March 1990 150pp 9004100344 PNL-6179 53350 090 Both the U S Environmental Protection Agency (EPA) and The report documents one of the tasks conducted under a the U.S Nuclear Regulatory Commission (NRC) have promul-Pacific Northwest Laboratory (PNL) progrcm entitled Evaluation gated regulations rega' ding the periormance of geologic reposs-and improvement of NDE Rehability for Inservice inspection of tones for the disposal of high-level nuclear waste Specifically, Light Water Reactors (NDE Ru sality Program The objective the EPA has prontuigated three quantitative, postclosure re-of this task wts to develop recommended requirements and eMW Wm M N processes for qualifying the UT/ISI systems (personnet equip-NRC's three quantitative, postclosure requirements apply only ment, and procedures) for inservice inspection of nuclear power to particular subsystems of the repository To assess comph-plant components This report desenbes an overall qualification ance with all six of these quantitative requirements, the phe-process intended to achieve statistically designed performance nomena that can affect the performance of the repository, the validations, as well as include the prerequisite training and other processes by which these phenomena are produced, and the qualification recommendations The document also contains parameters associated with these processes will have to be recommendatior's for the test specimens, environment, and identihed and quantified in addition, the analyses performed to other conditions under which the quakfication processes should assess compliance will have to be conducted in accordance be conducted in general, the recommendations described in with a performance assessment methodology to ensure that all this document are more stnngent than the current industry re.

regulatory cntena are addressed A performance assessment querements however, there are exceptions The major areas methodology proposed by Sandia National Laboratones is com-whnre specific enhancements are recommenced include more posed of scenano development and screening. consequence stnngent cntena for Level lli qualifications explicit recommenda.

analysis, uncertanty analysis and sensitivity analysis This tions for r9 qualification, greater emphasis on penodic (annuai) methodology can be used to assess compliance with the EPA's training. and recommendations for coordinating and administer-and NRC s requirements b

___1__..__...

k

Main Citations and Abstracts 9

NUREO/CR4273 V04:

SCDAP/RELAPS/ MOD 2 CODE subject of this bulletin Background information is supplied in MANUAL, VOLUME 4 MATPRO A LIBRARY OF MATERIALS the introduction and Appendix A PROPERTIES FOR LIGHT WATER REACTOR ACCIDENT NUREQ/CRD02: CLOSEOUT OF IE BULLETIN 80-10 CON-Nr ANALYSIS BUCCAFURNI.A.. CARLSON E R, CHAMBERS R.

TAMINATION OF NONRADIOACTIVE SYSTEM AND RESULT-

.NG POTENTIAL FOR UNMONITORED, UNCONTROLLED RE-lJ 902'6 E G 255 53 LEASE OF RADIOACTlvlTY TO ENVIRONMENT FOLEY,W.J,

This report describes the materials properties conelatons and DE AN.R.S, HENNICK,A PARAMETER, Inc February 1990 E

computer subcodes (MATPRO) developed for use with vanous 30pp 9003120758 PARAMETER IE193 52882.237.

kght water reactor (LWR) accident analysis computer programs Documentation is provided in this report for the closeout of IE Formulation of the matenats properties are generalty semiempir, Bulle9n 80-10 regarding contamination of nonradioactive sys-Icai in nature The matenals properties subcodes contained in tems renulting in the potential for unmonitored, uncontrolled re-this document are for uranium, uranium diox!de, mixed uranium-lease of radioactivity to the environment Closeout is based on plutonium dioxide fuel, zircaloy cladding, zirconium dioxide.

the oocumentation and ve'ificaton of four actions required by stainless steel, stainless steel oxide, silver-endium cadmiurn the bul!etin for holders of an operating license for a nuclear alloy, boron carbide, Inconel 718, zirconium-uranlurt orygen power tacility at the time the bulletin was issued (05-06 80) The melts, and fitt gas mixtures bulletin was issued for informaton to holders of a constructon NUREO/CR 6286: CLOSEOUT OF IE BULLETIN 79-17. PIPE permit for a nuclear power facility Evaluation of utility re-CRACKS IN STAGNANT BORATED WATER SYS1 EMS AT sponses and NRC/ Region inspection reports in accordance PWR PLANTS FOLEY,W J, DEAN,R Sa HENNICK,A PARAM-with tho close out entenon indicates that the bulletin is closed ETtiR, Inc February 1990 35pp 9004090172 PAR AMETE R for 66 (98%) of the 66 nuclear power fadlities to which it was IE 177. 53314.237 issued for action A follow up item is proposed for the facility Documentation is provided in this report for the closcout of IE with open status, for the use of NRC regional inspectors in en-Bulletin 79-17 and its revision on the safety related subject of sunng successful completion of required actions When the bul-pipe cracks sn stagnant borated water systems at operating letin is closed as anticipated for the facility which requires plants with pressunzed water reactors (PWRs) Closecut is follow-up (see page C-1) the concerns of the bulletin will have based on the implementation and verification of actions required been resolved completely Background informaton is supplied by the bullotm Evaluation of utmty responses and NRC/ Region in the introduction and Appendix A inspect.on reports indicates that the bulletin is closed for all of NUREG/CR-5307: CLOSEOUT OF IE BULLETIN 80-02 INAD-the 41 operating PWRs to which it was issued for action it is EQUATE QUALITY ASSURANCE FOR NUCLEAR SUPDLIED concluded that the concerns of the bulletin have been resolved EOUIPMENT FOLEY,W.J. DEAN,R S, HENNICK,A PARAME.

Background informaton is supplied in the introduction and Ap.

TER, Inc December 1989 15pp 9002120237 PARAMETER pendex A IE198 52611204 NUHEGiCR 5209: CLOSE"lT OF IE BULLETIN 79-23 POTEN-Documentation is provided in this report for the closcout of IE TIAL FAILURE OF EME60ENCY DIESEL GENERATOR FIELD Bulletin 80-02 The subject is inadequate quality assurance for EXCITER TRANSFORMER FOLEY,W J DEAN,RS-nuclear equipment supplied by the Marvin Engineenng Compa.

HENNICK,A PARAMETER, Inc March 1990 27pp ny The equipmant or concern is supplied either directly or 9004090279 PARAMETER IE180 53330.298 through other suppliers for use in General Electnc boiling water Documentation is provided in this report for the closeout of IE reactors (BWRs) Closecut is based on the implementation and Bulletin 79 23 regarding the potential failu'e of emergency venfication of three (3) required actions Evaluation of utility re-diesel generator ficki exciter transformers Closeout is based sponses and NRC/ Region inspection reports indicates that the upon the implementaten and venficaton of three actons re-bulletin is closed for all of the 38 BWR facilities to which it was quired by the bulletin Evaluation of utility responses and NRC/

issued for action it is concluded that the safety concerns re-Region inspection reports indicates that the bulletin is closed to' flected in the !E Bulletin 80-02 were adequately resolved by the all of the 119 nuclear power facihties with an operating license actions taken by licensees and ventied by NRC inspectors or a construction permit at the time the bulletin was issued.

Background information is presented in the introduction and Ap-September 12, 1979 Deviai.ons from bulletin testing require-pendix A ments, along with licensee justifications, are listed it is conclud-ed that the problem of concem was not genenc, since only two NUREG/CR 5316: V tT PROGRESSION.OxtDATION, AND NATURAL CONVECTION IN A SEVERELY DAMAGED REAC-plants, Nine Mile Point 1 and Turkey Point, required modifica-TOR CORE DOSANJH,S S Sandia National Laboratones Feb-tons to correct conrections which could cause high circulating currents. The problem at Turkey Point is desenbed in the bulle ruary'1990 93pp 9003070229 SAND 88 3476 53361156 In the revised Severe Accident Research Program plars, the tin (see page A 1) Background information is supplied in the in.

troduction and Appendix A U S Nuclear Regulatory Commission places a great emphasis NUREG/CR 5298: CLOSEOUT OF IE BULLETIN 85 01 STEAM BINDING OF AUXILIARY FEEDWATER PUMPS FOLEY W J DE AN.R.S. HENNICK,A PARAMETER, Inc January 1990 g al O eM M is N W m d m M Q 37pp 9003070241 PARAMETER IE189 52803 268 grossion A model that treats some of the important physica, Documentation is provided in this report for the closeout of IE processes that can occur dunng this phase of the accident is Bulletin 85 01 regarding steam binding of auxiliant feedwater descnbed herein A number of straightforward examples are pumps for certain pressun2ed water reactors (PWRs) in nuclear given to illustrate the utility of the model and to identity the power facilities Individual facility closecuts are based on the im-dominant physical processes piementaton and venfication of the three actions required by the bulletin Evaluaton of utility responses and NRC/Regon in NUREG/CR 5368: RE ACTIVITY ACCIDENTS A Reassessment Of spection reports indicates that the bulletin is closed for 44 The Design Basis Events DIAMOND.D J.

HSU.C J,

(92%) of the 48 facilities to which it was issued for action Foi-FIT ZP A T RICK.R Brookhaven National Laboratory January lowup :tems are proposed for the use of NRC regional inspec-1990 135pp 9002120269 BNL-NUREG 52198 52596 325 tors in ensunng satisfactory completion of required actions for This report documents a study of light water reactor event se the four f acilities with open bulletin status Conciusions are sum-Quences which have been investigated for their poteritial to man 2ed in accordance with Genonc Letter 88-03 which an result in reactivity accidents with severe consequences The nounced the NRC Staff's resoluton of Genenc issue 93 on the study is an outgrowth of the concern which arose after the acci-A A_......___.....

4 l

10 Main Citations and Abstracts dont ai Chernobyl and was recommended by the report of the the Nuclear Plant Aging Research Program of the U.S. Nuclear U.S. Nuclear Regulatory Comission (NRC) on the implcations of Rego! story Commission with the Pacifc Northwest Laboratory that accident (NUREG 1251). The work was done for the NRC (PNL) as the prime contractor, Research conducted by PNL to reconfirm or bring into questson previous judgements on reac-under Phase i provided an initial assessment of snubber operat-l trwty events which must be analyzed for hcensing. Event se-ing exponence and was primarily based on a review of Icensee quences were defined and then a probabiliste assessment was event reports. The work proposed is an extension of Phase I completed to estimate the frequency of the reactivity event and includes research at nuclear power plants and in test lab-and/or a deterministc calculaton was completed to estsmate oratories. Included is technical background on the design and the Consequences to the fuel. Using the results of this analysis use of snubbers in commweial nuclear pows apphcahons; the done by others, and a set of screening criteria developed within primary faiture modes of both hydraulic and mechanical snub-l 1his study, judgements were made for each sequence as to its bers are discussed. The anticipated safety, technical, and regu-importance, and recommendations were made as to whether latory benefits of the work, along with concerns of the NRC and the NRC ought to be considering the importar.t sequences as

' ** NO i

part of the design basis or for further, more detailed, investiga.

tion.

NUREQ/CR 6395 V02: MULTILOOP INTEGRAL SYSTEM TEST (MIST): FINAL REPORT. Test Group 30, Mapping Tests.

GEISSLER.G.O. Babcock & Wilcox Co. December 1989.

NUREG/CR 6376: OUALITY ASSURANCE AND VERIFICATION 1,058pp. 9003070265. EPRl/NP-6480. 52803:305.

OF THE MACCS CODE, Version 1.5.

DOBBE.CA.

MARWIL,E.S.; CARLSON,EA; et al. EG&G Idaho, Inc. (subs. of The Muttiloop Integral System Test (MIST) is part of a multi-phase program started in 1983 to address small-break loss-of-EG&G, Inc.). February 1990. 61pp. 9003120744. EGG-2566.

$2882:176.

coolant accidents (SBLOCAs) specife to Babcock and Wilcox designed plants. MIST is sponsored by the U.S. Nuclear Regu-An independent quality assurance (QA) and venfcaton of latory Commission, ine Babcock & Wilcox Owners Group, the Version 1.5 of the MELCOR Accident Consequence Code Electric Power Research Institute, and Babcock and Wilcox.

System (MACCS) was performed. The OA and verification in.

The unique featv es of the Babcock and Wilcox design, specif6-volved examinaton of the code and associated documentation cally the hot leg U-hends and steam generators, prevented the for consistent and correct implementation of the models in an use of er%in0 integral system data or existing integral facihties i

error. free FORTRAN computer code. The OA and verification to ad6ess the thermal-hydraulic SBLOCA questions. MIST and l

was not intended to determine either the adequacy or appropri.

two other supporting facilities were specifically designed and l

ateness of the models that are used in MACCS 1.5. The re.

Constructed for this program, and an existing fac!!ity the Once views uncovered errors which were fixed by the SNL MACCS Through Integral System (OTIS)-was also used.' Data from code development staff pnor to the release of MACCS 1.5.

MIST and the other facilities will be used to benchmark the ade-Some difficutties related to documentation improvement and quacy of system codes, such as RELAPS and TRAC, for pre-code restructunng are also presented. The OA and venfication dichng abnormal plant transients. The MIST program is reported process concluded that Version 1.5 of the MACCS code, within in 11 volumes. The program is summartred in Volume 1; Vo!-

the scope arvi limitations of the models implcmented in the umes 2 through C describes groups of tests by test type; code, is essentially error free and ready for widespread use-Volume 9 presents inter group comparisons; Volume 10 pro.

vides comparisons between the calculations of RELAPS/ MOD 2 NUREG/CR 5381: ECONOMIC RISK OF CONTAMINATION and MIST observations, and Volume 11 presents the later l

CLEANUP COSTS RESULTING FROM LARGE NONREACTOR Phase 4 tests. TNs Volume 2 pertains to MIST mapping tests NUCLEAR MATERIAL LICENSEE OPERATIONS. PHILBIN.J.S' trasse m ea@ posNW mnts s%

Sandia National Laboratories. SALOIOAH. ERC Environmental sts inegaW N N d Msmst vanns in

& Energy Services, Inc, ROLLSTINA Gram, Inc, March 1990.

boundary system controls, and only the primary fluid mass i

177pp. 9004100430. SAND 89-1302. 53344:114.

vaned dunng a specific test in this test group.

Several potential inclient scenanos involving the accidental NUREG/CR 5395 V10: MULTILOOP INTEGRAL SYSTEM TEST release of radioactive material at five reference, nonreactor nu.

(MIST): FINAL REPORT. RELAPS/ MOD 2 MIST Analysis Com-clear material licensees are analyzed in this report. The eco, parisons. KLINGENFUS,J.A.; PARECE,M.V. Babcock & Wilcox nomic risk ($/ licensee /yr) of decontamination is evaluated for Co. December 1989. 300pp. 9002120143. EPRl/NP 6480.

each reference hcensee. Although most releases and cleanup 52 costs are minor, some less frequent incidents may result in very oftiloop Integrol System Test (MIST) is part of a multi-l high cleanup costs that dominate the economic nsk of decon.

phase program started in 1983 to address smalbbreak loss of.

l tamination of a particular licensee. The economic risk for the 5 coolant accidents (SBLOCAs) specific to Dabcock and Wilcox i

l plants ranged from a low of $14,000 per heensee per year to a designed plants. MIST is sponsored by tne U.S. Nuclear Regu-high of $104.000 per heensee per year. This report is the l

second of two reports by Sandia National Laboratories on the economic nsk of nonreactor nuclear material licensee oper-nq e tu of e a ka I o desi cf l

ations. This report provides technical basis for a proposed fi-cally the hot leg U-bends and steam generators, prevented the l

nancial responsibility rulemaking for nonreactor nuclear material use of existing integral system data or existing integral facilities hcensees.

to address the thermal-hydraulic SBLOCA questions. MIST and two other supporting facilities were specifically designed and NUREG/CR 5386: BASIS FOR SNUBBER AGING RESEARCH:

constructed for this program, and an existing facihty-the Once Through Integral System (OTIS) was also used. Data from NUCLEAR PLANT AGING RESEARCH PROGRAM.

BROWN,D.P. Lake Engineering, Inc. PALMER.G.R. Wyie Lab-MIST and the other facilities will be used to benchmark the ade-oratories. WERRY,E.V.; et al. Battelle Mernorial Institute, Pacific quacy of system codes, such as RELAPS and TRAC, for pre-dicting abnormal plant transients. The MIST program is reported Northwest Laborato'y. January 1990.119pp. 9002120341. PNL-in 11 volumes. The program is summarized in Volume 1; Vol-6911. 52601:006. This report describes a research plan to ad-umes 2 through 8 desenbes groups of tests by test type; aress the safety concerns of aging in snubbers used on piping Volume 9 presents inter group comparisons; Volume to pro-gnd equipment in commercial nuclear power plants. The work is vides comparisons between the calculations of RELAPS/ MOD 2 to be performed under Phase 11 of the Snubber Aging Study of and MIST ooservations, and Volume 11 presents the later Phase 4 tests. The comparisons of RELAPS/ MOD 2 against the i:

1 h

^

i Main Citations and Abstracts 11 MIST data and conclusions reacned are the subject of this property data. Separate finito element models were used to evaluate the overall free-field behavior of the structure and the volume.

localized behavior at a specific penetration location. Three sce-NUREQ/CR 5394: TECHNICAL BASIS FOR REVIEW OF HIGH-narios of static intemal pressurizaton, based on gases building LEVEL WASTE REPOSITORY MODELING. PR!CE LL.;

up slowly within the containment shell donng a severe accident, WAHl,K.K.; GALLEGOS,0.P.; et al. Sandia National Laborato-were evaluated. The scenarios included pressure loading with ries. March 1990. 43pp. 9004100413 SAND 89-1557.

temperatures uniformly increasing in the finite element model in 53344!042 conespondence to the properties of pressurized saturated Both the U.S. Environmental Protection Agency (EPA) and steam, pressure loading with non-uniformly increasing tempera-the U.S Nuclear Regulatory Commission (NRC) have promul-s, a Wm baM @d a mpW hp gated regulatons rogarding the performance of geologic reposi-ture increase, which was done for comparison with earher pub-tories for the disposal of high level nuclear waste. One of the lished analyses. It is concluded that thermal effects do not responsibihties of the U.S. Department of Energy (DOE) is to change the overall response of the structure or the general demor,3frate comphance with the appropriate regulatons. The sheit failure mode, compared to the response due to pressunza-DOE will most likely use extensive numencal r'iodeling to show ton at ambient temperature. The reduction in the predicted irk compliance with the various quantitabve requirements. Theses tornal. pressure capacity of the Containment building at tempora-analyses will then be evaluated by the NRC. There are different ture corresponds to the reduction in the ultimate strength of the levels of evaluation: peer review, conservative estimates, use of A516 Grade 60 steel due to tne temperature increase.

exishng modets/ codes, and development of modelskodes by the NRC. The intensity of the review wilt vary from analysis to NUREG/CR 5419: AGING ASSESSMENT OF INSTRUMENT AIR anafysis, depending on the importance of the analysis, the ac-SYSTEMS IN NUCLEAR POWER PLANTS. VILLARAN,M.;

ceptability of the conceptual model beNnd the analysis and the FULLWOOD,R.; SUDUDHi,M. Drookhaven National Lateratory.

solution technique used, and the potenhal for increasing confi-January 1990. 139pp. 9003070250. BNL NUREG 52212.

dence in the system desenption, should the NRC decide to de-52003:129.

velop its own models/ codes. An appropnate level of review can NRC Generic issue 43, "Contaminahon of instrument Air be determined by applying these four criteria in a specific Lines", has been unresolved since 1980. The potential sonous-

manner, ness of this issue was reinforced in a 1987 study by the Office NUREG/CR-5404 V01: AUXILLARY FEEDWATER SYSTEM for Analysis and Evaluation of Operational Data. Aging of com.

AGING STUDY. CASADA.D.A. Oak Ridge National Laboratory, ponents within compressed air systems, leading to degraded March 1990.183pp. 9004090260. ORNL-6566. 53314:311.

function of the system, is the sublect of this study. This work This report documents the results of a study of the Auxihary was performed under the auspices of the NRC's Office of Hu-Feedwater (AFW) System that has been conducted for the U.S.

clear Regulatory Research as part of the Nuclear Plant Aging Nuclear Regulatory Commission's Nuclear Plant Aging Re-Research (NPAR) Program. The objective of this study was to search Progam. The study reviews histoncal failurs data avail-identify all the aging modes and their causes, which should be able from the Nuclear Plant Reliabihty Data System, Licensee mitigated to achieve a reliable operation of all safety related o' Event Report Sequence Coding and Search System, and Nucle-equipment. Also included is an interim review of typical mainte er Power Exponence data bases. The failure histories of AFW nance activities for air systems in the nuclear power industry.

System components are considered from the perspectives of The Phase 2 effort of this study will make recommendations for how the failures were detected and the significance of the fail.

developing an effective maintenance program industry wide to ure. Results of a detailed review of operating and monitonng counter the effects of aging. The analysis of operating experi-practices at a plant owned by a cooperating utility are present.

ence data revealed that aging degradation occurs in the com-ed. General system configurabons and pert:nont data are pro-pressed air system, and becomes a factor as the system ages.

Normal wear of the system and contamination of the air domi-vided for Westinghouse and Dabcock and Wilcox units.

nate the probiems of system failure. Existing maintenance pro-NUREO/CR 5405: ANALYSIS OF SHELL RUPTURE FAILURE grams within the industry lack uniformity, and quality assurance DUE TO HYPOTHETICAL ELEVATED TEMPERATURE PRES-G""**

"Y" SUR12ATION OF THE SEQUOYAH UNIT 1 STEEL CONTAIN.

MENT BUILDING. MILLER.J,D. Sandia National Laboratones.

NUREG/CR 5421: LAPUR USER'S GUIDE. OTADUY,P.J.

February 1990.100pp. 9003070222. SAND 891650. 52811:132.

MARCH LEUBA.J. Oak Ridge National Laboratory. January Sardia National Laboratones, as part of the Containment in.

1990.77pp.9003070237. ORNL TM/11285. 52801:001, tegrity Programs under the sponsorship of the Nuclear Regula-LAPUR, a computer program in FORTRAN IV,is a mathemat-tory Commission (NRC), has developed analytical techniques ical desenption of the core of a boiling water reactor. Its two for predicting the performance of hght water reactor steel con-hnked modules, LAPURX and LAPURW, respectively solve the tainment buildings subject to loads beyond the design basis.

peah state governing equabons for the coolant and fuel and The analytical techniques are based on experience with large-C e dynamic eQuebons for the coolant, fuel, and neutron field in scale steel conta!nment model tests that provided imporMnt in-

'.no frequency domain. General implementabon desenptions are sights and expenmental vahdation of the analytical methods. As followed by a detailed desenption of input and output param-a means of demonstrating these analytical techniques, the NRC eters of LAPURX and LAPURW. Sample inputs are included asked Sandia to conduct a structural evaluation of an actual and stability benchmarks are noted.

steel con?inment building 1he object!ve of the analysis was to determine the actual pressure capacity and the mode, location, NUREO/CR-5424: ELICITING AND ANALYZING EXPERT and size of failure, where a functional definition of failure is JUDGEMENT.A Practical Guide. MEYER.M.A.; DOOKER.J.M.

used. The purpose of this report is to document the calculat:ons Los Alamos National Laboratory. January 1990. 424pp.

performed to determine the pressure hmits for the shell rupture 9002120292. LA-11667 MS. 52574.303.

mode of failure. General failure of the containment shell is pre-In this book we desenbe how to elicit and anaiyze expert dicted by apphcation of a failure entenon to the results from judgment. Expert judgment is defined here to include both the finite element situctural analyses. The failure entenon relates experts' answers to technical questions and their mental proc-the calculated values of strain in the containment plates, due to esses in reaching an answer. It refers specifically to data that Internal-pressurization loading. to the ultimate strain hmit of the are obtained in a deliberate, structured manner that makes use steel. Included in the failure cntenon are adjustments for factors of the body of research on human cognition and communica-inherent in finito element analysis, such cs level of detail and tion. Our aim is to provide a guide for lay persons in expert element size of finite element model and variations in matenal judgment. These persons may be from physical and engineenng n.

____-A--_____-

12.

Main Citations and Abstracts sciences, mathematics and statistics, business, of the military.

ity calculations were performed when possible. Generically, IA

- We provide background on the uses of expert judgment and on was found to Contribute less to total risk than many safety sys-the processes by which humans solve problems, including those tems; however, specific design weaknesses in safety systems, that lead to bias. Detailed guidance is offered on how to elicit non-safety systems, and the tA system were found to be signifi-expert judgment ranging from selecting the questions to be cant in risk.

posed of the experts to selecting and motivating the experts t setting up for and conducting the ehcitation. Analysis proce.

NUREG/CR 5474: ASSESSMENT OF CANDIDATE ACCIDENT dures are introduced and guidance is given on how to under.

MANAGEMENT STRATEGIES. LUCKAS W.J.; VANDENKIE-BOOM; LEHNER,J.R. Brookhaven National Laboratory. March I a aggregate h udg ents 1990.46pp.9004100337. BNL NUREG 52221,53343:029.

A set of selected candidate accKient management strategies, NUREG/CR 5436: ENVIRONMENTAL EFFECTS ON CORRO-whose purpose is to prevent or mitigate in-vessel core damage, l

SiON IN THE TUFF REPOSITORY, BEAVERS,J.A.;

were developed from various NRC and industry rep'cas. These THOMPSON,N.G. Cortest Columbus, Inc. February 1990.

strategies have been grouped in this report by the challenges 149pp. 9003190298. 53012:133.

they are intended to meet, and assessed to provide information Cortest Columbus is investigating the long term performance which may be useful to individual hcensees for consideration j

of container matenals used for highlevel waste packages as Uhen they perform their Individual Plant Examinations. Each as-part of the information needed by the Nuclear Regulatory Com-

% sament focused on describing and explaining the strategy.

mission to assess the Department of Energy's application to

. ensidering its relationship to existing requirements and prac-construct a geologic repository for high-level radioactive waste.

tices as well as identifying possible associated adverse effects.

The scope of work consists of employing short term techniques, such as electrochemical and slow strain rate mechanical test NUREG/CR 5476: POSTTEST ANALYSIS OF A 1:6-SCALE RE-techniques, to examine a wide range of oossible failure modes.

INFORCED CONCRETE REACTOR CONTAINMENT BUILD-Long-term tests are being used to verify and further examine ING. WEATHERBY,J.R. Sandia National Laboratories. February specific failure modes identified as important by the short term 1990. 96pp. 9004030052. SAND 89-2603. 53226:001.

studies. This report summarizes the results of a literature survey in an experiment conducted at Sandia National Laboratones, performed under Task 1 of the program. The survey focuses on a 1:6 scale model of a reinforced concrete light water reactor the influence of environmental variables on the corrosion be.

containment build:ng was pressurtzed with nitrogen gas to rr. ore havior of candidate container materials for the Tuff repository.

than three times its design prossure. The pressurtzation pro.

Environmental variables considered included: radiation, thermal duced one large tear and several smaller tears in the steel hner and microbial effects.

plate that functioned as the pnmary pneumatic seal for the structure. This report describes posttest finite element analyses NUMG/CR-5460:

HIGH TEMPERATURE CRACK ARREST TESTS USING 152 MM THICK SEN WIDE PLATES OF LOW.

of the 1:6 scale model test and compares pretest predictions of UPPER SHELF BASE MATERIAL: TESTS WP-2.2 AND WP 2.6.

the structural response to the experimental results. Strains and NAUS,D.J.; KEENEY WALKER; BASS,0.R.; et at Oak Ridge displacements calculated in axisyme.tric finite element analy-ses of the 1:6-scale model are compared to strains and dis-National Laboratory. February 1990. 143pp. 9003070238.

ORNL/TM 11352, 52802:055.

placements measured in the expenment. Detailed analyses of Two 152 mm-thick wide plate crack arrest tests (WP 2 senes) the liner plate are also descnbed in the report. The results from are discussed in this report. Each test used a 1 x 1 x 0.15 m those analyses indicate that the primary mechanisms that initiat-thick single edge-notch spectmen (a/w = 0.2), fabricated from ed the tear can be captured in a two-dimensional finite element a low upper-shelf base material, that was subjected to a linear model. Furthermore, the analyses show that studs which were thermal gradient along the plane of crack propagation. The used to anchor the liner to the concrete wall, played an impor-tests were conducted at the National Institute of Standards and tant role in initiating the liner tear.

Technology and wero designed to provide fracture toughness NUREQ/CR 5477: AN EVALUATION OF THE RELIABILITY AND measurements at temperatures approaching or above the onr.st USEFULNESS OF EXTERNAL. INITIATOR PRA METHODOLO-of the Charpy upper shelf regime in a rising toughness region GIES. BUDNITZ,R.Ja LAMBERT,H.E. Future Resources Associ-and with an increasing driving force. Results obtained from ates, Inc. January 1990.111pp. 9002120348. 52577.129.

these tests have produced crack arrest toughness values well This report, prepared to assist pokcy-icvel decision-makers, above the kmit recognized by the current ASME guidehnes (220 evaluates the extent to which each cakgory of extemal-initia.

MPa

• VTn) with arrests occuring at 44 to 102 degrees tors PRA methodology produces reliable and useful results and C above the matenal DW(NDT) (60 degrees C). The fracture insights, at its current state of the-art level This report address-data support (1) the use of fracture mechanics concepts to ana-es this need in the following five categoriec of extema! initiators:

lyze cleavage run arrest events, (2) the treatment of cleavage (1) earthquakes; (2) intemal Ores; (3) extemal floods; (4) ex-run-arrest and ductile fracture modes as separate events, and treme winds; and (5) transportation accidents. Each initiator is (3) the fact that cleavage arrest occurs above the ASME hmit.

examined separately. The thrust is to identify and describe the

- NUREG/CR 5472: A RISK BASED REVIEW OF INSTRUMENT pnncipal aspects of the current state-of the-art PRA methodolo-P NR SYSTEMS AT NUCLEAR POWER PLANTS. DEMOSS G.;

h,*in gh s* a id less robust and therefore provide less reh-LOFGREN,E.: ROTHLEDER.B.; et al. Science Applications Intemational Corp. (formerly Science Apphcations. Inc.). January NUREG/CR 5478: IMPROVED EDDY CURRENT INSPECTION 1990.165pp.9003070261. BNL NUREG 52220. 52802:324.

FOR STEAM GENERATOR TUBING. Progress Report For Jan.

The broad objective of this analysis was to provide risk based uary 1985 To December 1987. DODD.C.V.; DEEDS,W.E.;

information to help focus regulatory actions related to Instru-MCCLUNG,R.W. Oak Ridge National Laboratory. January 1990.

ment Air (IA) systems at operating nuclear power plants. We 36pp.9002120284. ORNL/TM-11389. 52597:255.

first created an extensive data base of summarized and charac.

A major limitation of eddy current inspection of steam genera-tenzed IA related events that gave a qualitative indication of the tor tubing is that small flaw signals can be masked by the of.

nature and severity of these events. Additionally, this data base fects of benign vanables, such as tube supports. To identify the was used to calculate the frequencies of certain events, which cntical flaw properties accurately and reliably in the presence of were used in the risk analysis. The nsk analysis consisted of re-signals caused by these other property vanations, we must have viewing published PRAs and NRC Accident Sequence Precursor enough information to distinguish the flaw signals from the ex-reports for IA-initiated accident sequences. IA interactions with traneous ones. Therefore, we developed instrumentation to tronthne systems, and lA-related risk significant events. Sensitiv-measure both the amplitude and the phase of the eddy-current o

n A

Main Citations and Abstracts 13 signal at several different frequencies, as well as computer hibit a nearly Nernshan response to pH, no hysteresis effects, equipment to process the data quickly and rehably. This need to and minimal response to lonic interferences. Sensitwity to cer-detect small flaws in the presence perturbing property variatons tain redox speces is observed, however. In addition, methods has also required the development of mo'e sensitue and more are discussed for preparing model indium oxide sensor surfaces complicated probes, such as pancake and reflection probes.

for ultraNgh vacuum surface analytical studies. Stoichometric i

These smaller coils can detect much smaller flaws and are less IrO(2)-like surfaces are shown to be relatively inert to gas phase sensttive to artifacts outside the tube, such as tube supports, water. However, hydroxylation of IrO(2)-hke surfaces can be ird magnetite, or cooper, By being pressed against the tube wall, duced by rf water plasma treatment.

they also avoid httoff effects. To increase the inspecton speed NUREG/CR 5491; SHIPPINGPORT STATION AGING EVALUA-an array of these small coils has been constructed and tested.

Finally, new and more comphcated tube standards were con-TION. ALLEN,R.P.: JOHNSON,A.B. Battelle Memorial inststute, structed to include the range of property variations.

Pacific Northwest Laboratory. January 1990. 143pp.

9002120344. PNL 7191. 52577:240.

NUREG/CR 5479: CURRENT APPLtCATIONS OF VIBRATION The Shippingport Atomic Power Station, the first U.S. large-MONITORING AND NEUTRON NOISE ANALYSIS.Detecton scale, central-station nuclear plant, now in the hnal stages of i

And Analysis Of Structural Degradaten Of Reactor Vessel inn-decommissioning, has been a major source of naturally aged nals From Operational Aging. DAMIANO.B.; KRYTER.R.C. Oak equipment for the Nuclear Plant Aging Research (NPAR) and j

. Ridge Nahonal Laboratory. February 1990. 43pp. 9004090190.

other U.S. Nuclear Regulatory Commission (NRC) programs.

ORNL/TM 11398. 53314:270.

The evaluaten of naturalty aged components is an integral part This report, which was prepared under the Nuclear Plant of the NPAR program strategy. Because naturalty aged compo-l Aging Research Program sponsored by the Unitod States Nu-nents and matenals expenence the actual service-related exter-j clear Regulatory Commission, discusses the apphcahon of vp nel stressors, corrosion and wear, testing procedures, and main-

]

bration monitonng and neutron noise analysis for monitonr:g tenance practices, their evaluaton is valuable in verifying degra-tight water reactor (LWR) vessel tr'temais. The report begins by dation models, vahdeting aging projectons based on the ex-descnbing the effects of loss of structural integrity on intemals trapolaton of accelerated test data, and detecting unexpected vibration and how sensible parameters can be used to detect aging mechanisms (surpnses) ihat could significantly impact and track the progress of degradation. This is followed by a de.

component or system safety performance. As part of the Ship-scripton and companson of vibraton monitonng and neutron pingport Stahon aging evaluaton work, more than 200 items, j

noise analysis, two methods for monitoring the mechanical in.

rangeng in size from small instruments and materials samples to tegnty of reactor vessel internal components. The major section one of the main coolant pumps, have been removed and J

of the report describes the status of reactor vessel intemals shipped to designated NRC contractors. Although detailed eval-condihon monitoring programs in the United States, Federal Re.

uahons of the components and material from the Shippingport I

pubhc of Germany, and France, three countries having substan.

Stahon are just beginning, the preliminary results from the stud-hal commitments to nuclear power. The last section presents los conducted to date are indicatwe of the value of the aging guidelines for U.S. Utihties wishing to estabhsh reactor intemals infomiation that ultimatefy may be obtained.

condihon monitoring programs, NUREG/CR-5482; LABORATORY ANALYSIS OF FLUID FLOW NUREG/CR 5492:

INVESTIGATIONS OF IRRADIATION-AND SOLUTE TRANSPORT THROUGH A VARIABLY SATU-ANNEAL REIRRADIATION (IAR) PROPERTIES TRENDS OF RATED FRACTURE EMBEDDED IN POROUS TUFF.

RPV WELDS. Phase 2 Final Report. HAWTHORNE J.R.;

CHUANG,Y,; HALDEMAN,W.R.; RASMUSSEN.T.C.; et al. Artzo.

HISER,A.L. Matenals Engineering Associates, Inc. January na, Univ. of, Tucson, AZ. February 1990. 332pp. 9003070258.

1990. 302pp. 9002120277. MEA-2088. 52574:001.

52 Notch ductility, fracture toughness and tensile propwty trends atory techniques are developed that allow concurrent U'

1. PE" 8"

measurement of unsaturated matrix hydrauhc conductivity end ps C inaMon % M Mas C posWaQ unneabg fracture transmissivity of fractured rock blocks. Two Apache W aM 288 %e C wadah M m m@aM Leap tuff blocks with natural fractures were removed from near Pmnary @ctNes mcW h when of wM Mal Superior, Artzona, shaped into rectangular prisms, and instru-reembnttlement rate with fluence following an anneal and the mented in the laboratory. Porous ceramic plates provided solu-ence d kat @ Hece M @ w W m mM&

tion to block tops at regulated pressures. Infiltration tests were ment susceptibikty. The welds were commercially made using a performed on both test blocks. Steady flow testing of the satu-single lot of filler wire and two welding fluxes (Linde 80 and rated first block provided estimates of matrix hydraulic conduc-Linde 0091). A relatively rapid reembnttlement of both weld tivity and fracture transmissivity. Fifteen centimeters of suction WS was esM M Mal remadam hw, N wn-apphed to the second block top showed that fracture flow was bnttlement rate decreased markedly after a reirradiahon fluence minimal and matrix hydraulic conductivity was an order of mag-of aw W x M n/c4 E gmater man 1 h m un, nitude less than the first block saturated matrix conductivity.

the benefit of IA procedures toward reduct ig total properties Coated wire ion-selective electrodes monitored aqueous chlorid-change with fluence was retained. The reembnttlement trend ed breakthrough concentrations. Minute samples of tracer solu-was independent of first cycle fluence level for the range inves-tion were collected with filter paper. The techniques worked well tigated. Residual embrittlement after 399 degrees C.168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> for studying transport behavior at near saturated flow conditions anneahng, indexed by 41 J temperature, appears independent and also appear to be promising for unsaturated conditions.

of the pre-anneal condition (I or IAR); however, wide vanability Breakthrough curves in the fracture and matrix, and a concen-among welds in percent recovery is indicated. Trends in IAR be-tration map of chloride concentrations within the fracture, sug-havior from Charpy-V specimen tosts were found to be rein-gest preferenhal flow paths in the fracture and substantial diffu-forced by tensile specimen and fracture toughness (0.5T CT s6on into the matnx. Average travel velocity, dispersion coeffi-sWmen) kst hndings.

cient and longitudinal dispersivity in the fracture are obtained.

NUREG/CR 5484: PH SENSORS BASED ON IRIDIUM OXIDE.

NUREG/CR 5493: INFLUENCE OF FLUENCE RATE ON RADI-ATION INDUCED MECHANICAL PROPERTY CHANGES IN RE-TARLOV,M.J.; KREIDER,K.G.; SEMANClK S.; et al. National in-ACTOR PRESSURE VESSEL STEELS. Final Report On Explora-stitute of Standards & Technology (formerly National Bureau of Standa. March 1990. 2tpp. 9004100360. 53344:092.

tory Expenments. HAWTHORNE,J R.; HISER,A.L. Materials En-Results are presented on the pH potenhal response of d c.

gineering Associates, Inc. March 1990. 300pp. 9004100351.

MEA-2376. 53344:299.

magnetron reactively sputtered indium oxide films. The films ex-a n

g

14 Main Citations and Abstracts This report describes a not of experiments undertaken using a mm-thick compact specimens (0.5TCS), to a fluence of 2A1 x 2 MW test reactor, the UBR, to quahty the significance of 10(19) neutrons /ctn(2) (greater man i MeV), resulted in de-fluence rate to the extent of embnttlement produced in reactor creases in the initiation fracture toughness, J(Ici, and the tear-pressure vessel steels at their service temperature. The test ing modulua, materials Included two reference plates (A 302 B, A 533 B steel) and two submerged are weld deposits (unde 80, Linde NUREG/CR 5512 DRF FC: RESIDUAL RADIOACTIVE CONTAMI-i 0091 welding fluxes). Charpy V (C(v)), tension and 0.5T CT NATION FROM DECOMMISSIONING. Technical Basis For l

compact specimens were employed for notch ductility, strength Translating Contamination Levels To Annual Dose. Draft Report and tracture toughness (J R curve) determinations, respectively.

For Comment. KENNEDY,W.E.; PELOQUIN,R.A. Battelle Memo-i Target fluence rates were B x 10(10),6 x 10(11) and 9 x 10(12) rial Institute, Pacific Northwest Laboratory. January 1990.

n/cm(2)-s( 1). Specimen fluences ranged from 0.5 to 3.8 x 317pp. 9003070227, PNL 721? 02811:232.

j 10(19) n/cm(2), E Greater than 1 Mev. W %s describe a This document describes the genefic modehng of the total ef.

fluence rate effect which may extend to powet reactor surveil-fective dose equivalent (TEDE) to an individual in a population lance as well as test reactor facilities now in use The depend-from a unit concentration of residual radioactive contamination.

ence of embrittlement sensitivity on fluence rate appears to Radioactive contamination inside buildings and soil contamina-difier for plate and weld deposit matenals. Relatively good tion are considered. Unit concentration TEDE factors by radio-l agreement in fluence-rate effects definition was observed nuclide, exposure pathway, and exposure scenario are calculat.

among the three test methods.

ed. Reference radiaton exposure scenarios are used to derive i

NUREG/CR-5W6: PRELIMINARY STRUCTURAL EVALUATION W concenkaton TEDE factors for about 200 inmal radom OF TROJAN RCL SUBJECT TO POSTULATED RPV SUPPORT uchdes and parent-daughter mixtures. For buildings, these unit FAILURE. LU,S.C. Lawrence Livermore Nabonal Laboratory.

concentration factors hat the annual TEDE for volume and sur-January 1990. 35pp. 9002120234. UCID-21831. 52611:224.

face contamination situatons. For soil, annual TEDE factors are This report desenbes a preliminary structeral evaluation made presented for unit concentrations of radionuchdes in soil during to determine whether the reactor coolant loop (RCL) piping of residential use of contaminated land and the TEDE per unit the Trojan nuclear power plant is capable of transferring the totalinventory for potential use of drinking water from a ground-loads normally carned by the reactor pressure vessel (RPV) water source. Because of the genenc treatment of potentially supports to other component supports in the RCL system if the complex ground-water systems, the annual TEDE factors for RPV supports should fait, say from radiation damage. For the drinking water for a Otven inventory may only indicate when ad-evaluahon, we use the computer model of the RCL system of diticnal site data or modeling sophistication are warranted. De-l Unit 1 of the Zion nuclear power plant because it is readily senptions are provided of the models, exposure pathways, ex-available; the RCL systems of these two plants closely resem, posure scenanos, parameter values, and assumptons used. An ble each other. As a bounding case in the evaluation we postu, analysis of the potential annual TEDE resulting from reference late that all four RPV supports have failed. Two load combina.

mixtures of residual radionuchdes is provided to demonstrate tions are evaluated: (1) the combinahon of dead weight, operat.

application of the TEDE factors.

Ing pressure, and the safe-6hutdown earthquake, and (2) the combinaton of dead weight, operating pressure, and a loss of.

NUREG/CR 5516: CAUSES OF FAILING THE DRAFT ANSI coolant accident. Both load combinations are classified as Level STANDARD N13.30 RADIOBIOASSAY PERFORMANCE CRI.

l 0 Service Limits in accordance with the ASME Boiler and Pres-TERION FOR MINIMUM DETECTABLE AMOUNT I

sure Vessel Code. Static and dynamic hnear elastic analyses MACLELLAN.J.A. Battelle Memoria' Institute, Pacific Northwest are conducted to compty with rules specified by Subsection NB Laboratory February 1990. 43pp. 9003070190. PNL 7217.

In conjunction with Appendix F Division 1, Secten lit of the 52813:026.

ASME Code. Results of this preliminary evaluation indicate that The test methods used for PNL bioassay performance tests ASME Code Appendix F requirements are satisfied by each of were evaluated by comparing the MDA based on performance the load combinations considered in the analysis, leading to the tests results with the MDA calculated by PNL using the bioas-conclusion that the Trojan RCL piping is capable of transfemng say laboratory's own quality control (OC) data. Two in vitro lab-l I

the RPV support loads to the steam generator and reactor cool-oratories and two in vivo laboratories were studied and a corre.

ant pump supports.

laton between the perfomance test MDA estimates and OC NUREG/CR 5511: IRRADIATION EFFECTS ON STRENGTH AND data was demonstrated. However, it was often necessary to ex.

amine the OC data to identify important characteristics of the TOUGHNESS OF THREE WIRE SERIES-ARC STAINLESS blank distribution that affect the MDA calculation. Since the STEEL WELD OVERLAY CLADDING.

HAGGAG,F.M.;

MDA equahon must be based on the specific analysis and cal-CORWIN,W.R.; NANSTAD R.K. Oak Ridge Natonal Laboratory.

culational methods of the procedure evaluated. Even when the February 1990. 155pp. 9003190328. ORNL/TM 11439.

53044:223.

correct MDA equaton is applied, the MDA calculated will have The potential for stainless steel cladding to improve the frac-a retabvely large confidence interval when only a few replicates ture behavior of operahng nuctear reactor pressure vessel.

are used to estimate the standard deviation. For this reason, a particularly dunng certain overcooling transients, may depend relabvely precise estimate of the MDA is generally only avail-greatly on the properties of the irradiated cladding. Therefore.

able when Poisson statistics may be applied. It was concluded three-wire stainless steel cladding irradiated at temperatures that performance testing alone cannot provide all the informa-tion necessary to make an accurate estimate of the measure-and to fluences relevant to power reactor operaton 5 e exam-ined. Posbrradiation tensile testing resutts show that, trot,125 mer,t nrocess MDA. Review of the laboratory's OC data and the to 288 degrees C, the yield strength increased by 8 to 30%,

entre measurement procedure wi!! be necessary. Specific rec-ommendations for changes to draft ANSI N13.30 " Performance ductility increased insignificantly, with alrnost no change in ulti-Cnteria for Radiobcassay" are given.

mate tensile strength. All cladding exhibited ductile-to-bnttle transition behavior dunng Charpy impact testing, because of the NUREG/CR 5527: RISK SENSITIVITY TO HUMAN ERROR IN dominance of detta fernte failures at low temperatures. On the THE LASALLE PRA. WONG,S.; HIGGINS.J.; O'HARA,J.; et al.

upper shelf, energy was reduced 15 and 20%, and lateral ex-Brookhaven National Laboratory. March 1990. 152pp.

pansion 43 and 41%, owing to irradiation exposure of 2 and 5 x 9004030071. BNL NUREG-52228. 53226:169.

10(19) neutrons /CM(2) (greater than l MeV), respectively. In A sensitivity evaluation was conducted to assess the impact addition, radialon damage resulted in 13 and 28 degrees C of human errors on the intemal event risk parameters in the La-Shifts of the Charpy impact transition temperature fcr the low Salle plant. The results provide the variahon in the risk param-and high fluences, respectively. Irradiation exposure of 12.5 eters nameiy, core melt frequency and accident sequence fre-l G

n n

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

Main Citations and Abstracts 16 quencies, due to hypothetical changes in human errer probabil-NUHEQ/lA 0013: RELAPS/ MOD 2 CALCULATIONS OF OECD-itses. Also provided are insights derrved from the results, which LOFT TEST LP SB 03. HARWOOD,C.; BROWN.G. Central Elec-highlight important areas for concentration of risk limitation ef-tricity Generating Board. January 1990. 48pp. 9002120316. GD/

I forts associated with human performance.

PE N/535. 52611:083.

This report compares the results of the RELAPS/ MOD 2 anal-NUREQ/lA 0012: RELAP5/ MOD 2 CALCULATIONS OF OECD-ysis with expenmental measurements. A simuistion of test LP-LOFT TEST LP SB 01. HALL,P.C.: BROWN,G. Central Electnci.

SB 03 was previously carried out at GDCD using the RELAPS/

ty Generating Board. January 1990. 40pp. 90021202B3. GD/PE.

MODI code. RELAPS/ MOD 2 was developed from RELAPS/

N/544. 5259n215.

MODI and contains more sophisticated hydraulic models and To assist CEGB in assessing the capabilities and status of RELAPS/ MOD 2, the code has tnen used to simulate SBLOCA const:tutive relationships. Comparison of the RELAP5/ MOD 2 test LP SB-01 carried out in the LOFT experimental reactor and MOD 1 calculations show that RELAP5/ MOD 2 perioms under the OECD LOFT programe. This test simulated a 1.0%

better than RELAPS/ MODI in a number of key areas; notably hot tog break ip a PWR, with early tnpping of the primary cool-mass errors are much reduced, there is improved numerical sta-ant circulating pumps. This report compares the results of the bility, and improved separator modelling and modelling of accu-RELAP5/ MOD 2 ana ysis with expenmental measurements.

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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 compRation. Each code is cross-referenced to the NUREG number for the report and to the 10 digit NRC Document Control System accession number.

J9 h}

SECONDARY REPORT NUMSER REPORT NUMSER SECONDARY REPORT NUMBER REPORT NUMBER ANL 86-33

• NUREG/CR4704 V03 ORNL/NSIC-200 NUREG/CR-2000 V08N12 gi ANL 89/22 NUREG/GR-4744 V03 N1 OANL/kiSC 200 NUREGICR-20no V09 N2 ORNL/N510-200 NUREG/CR 2000 V09 N1 BAW-2078 NUREG/CR-5395 V10 OANLMM-11352 NUREGICR-5450 BAW 2080 NUREG/CR 5395 V02 BNL NUREG 51454 NUREG/CR 2331 V00 N3 M'M$

BNL NUREG-51742 NUREG/CR 3668 OANL/TM-11439 NUREG/CR 5511 BNL NUREG 52007 NUREG/CR46.49 V03 PARAMETER IE158 NUREGICR4661 DNL NUREG 52198 NUREG/CR4308 r ARAMETER IE177 NUREG/CR 5286 DNL NUREG-52212 NUREG/CR-5419 PARAMETE R IE180 NUREG/CR 5289 DNL NUREG-52220 NUREG/CR-5472 PARAMETER IE189 NUREG/CR-5298 BNL NUREG-52221 NUREG/CR 5474 PARAMETER IE 193 NUREG/CR-5302 BNL NUREG 52226 NUREG/CP-0110 PARAMETER IE198 NUREG/CR-5307 BNL NUREG 52228 NUREG/CR 5527 PNL-5711 NUREG/CR4469 V08 I;s 6711 NUREGICR-4469 V09 EGG-2469 NUREG/CR 4731 V02 PNL 6179 NUREG/CR 4882 EGG 2555 NUREG/CR-5273 V04 EGG 2566 NUREG/CR 5376 k

h Ug EGG 2577 NUREG/CR 5229 V02 PNL 7 tot NUREG/CR-5491 EPRl/NP 6480 NUREG/CR-5395 V10 PNL 7212 NUREG/CR-5512 DAF FC EPRl/NP-6480 NUREG/CR 5395 V02 PNL 721.'

NUREG/CR-5516 GD/PE-N/535 NUREG/iA 0013 SAND 85-7185 NUREG/CR4214 R0f Pt GD/PE N/544 NUREG/lA-0012 SAND 86-1030 NUREG/CR-4668 IEB 70-17 NUREG/CR-5286 SAND 86-1443 NUREG/CR4671 IEB-79-23 NUF/G/CR 5289 SAND 661562 NUREG/CR 4691 V02 IEB-80 02 NUaEG/CR-5307 SAND 661562 NUREG/CR-4691 V03 SAND 86-20S4 NUREGICR4550 V01 R1 IEB45 001 NUREG/CR-5298 SAND 88-3020 NUREG/CR-5256 i

IEB-85-0G3 NUREG/CR-4661 08 LA 11667 MS NUREG/CR-5424 08 -

I MEA-2088 NUREG/CR-5492 SAND 89-1557 NUREG/CR 5398 MEA 2376 NUREG/CR4493 SAND 89-1650 NUREG/CR-5405 ORNL 6566 NUREG/CP 5404 VOI SAND 89-2603 NUREG /CR-5476 ORNL TM/11285 NUREG/CR-5421 0C10-20674 NUREG/CR4554 V07 ORNL/NOAC 232 NUREG/CR4674 V10 UCID-20674 NUREG/CR-4554 V06 ORNL/NOAC 232 NUREG/CR 4674 V09 UCID 21831 NUREG/CR-5506 l

i 17

i I

i I.

i l

i I

=

M m

Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of

.the report (s) prepared by the author, if further information is needed, refer to the main cita-tion by the NUREG number, ALLEN,R.P.

NUREG/lA-0013 RELAP5/ MOD 2 CALCULATIONS OF OECD LOFT NUREG/CR 5491: SHIPPINGPORT STATION AGING EVALUATION-TEST LP.SB-03.

ALLENSPACH,F.

DUCCAFURNI,A.

NUREG 1214 RO5: HISTORICAL DAT A

SUMMARY

OF THE SYSTEMAT*

NUREG/CR-5273 V04.

SCDAP/RELAP5/ MOD 2 CODE IC ASSESSMENT OF LICENSEE PERFORMANCE.

MANUALVOLUME 4 MATPRO A LIBRARY OF MATERIALS PHOP.

ERTIES FOR LIGHT. WATER REACTOR ACCIDENT ANALYSIS.

NUREG/CR-4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR BUDNITZ,R.J.

LIGHT WATER REACTOR COMPONENTS OVERVIEW.

NUREG/CR 5477: AN EVALUATION OF THE RELIABILITY AND USE.

FULNESS OF EXTERNAL INITIATOR PRA METHODOLOGIES.

AULUCK,H.

NUREG 1232 V04: SAFETY EVALUATION REPORT ON TENNESSEE J

VALLEY AUTHORITY: WATTS BAR NUCLEAR PERFORMANCE BUESCHER B.d731 NUHEG/CR V02: RESIDUAL LIFE ASSESSMENT OF MAJOR PLAN LIGHT WATER REACTOR COMPONENTS OVERVIEW.

BAKR,M.H.

CA NUREG/CR-4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR NUR CR 4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-LIGHT WATER PEACTOR COMPONENTS OVERVIEW.

CY: INTERNAL EVENTS METHODOLOGY.

BANDYOPADHVAY CARLSON E.P.

NUREG/CR 4659 Vn3-SF.;SMIC FRAGIUTY C* NUCLEAR POWER NUREG/Ch-o273 V04:

SCDAP/RELAPS/ MOD 2 CODE PLANT COMPONENTS (PHASE II)Setchgear, l&C Panels (NSSS)

MANUAL, VOLUME 4 MATPRO, A LIBRARY OF MATERIALS PROP-And Relays ~

ERTIES FOR LIGHT. WATER. REACTOR ACCIDENT ANALYSIS.

NUREG/CR 5376: OVALITY ASSURANCE AND VERIFLCATION OF THE BASS,B.R.

NUREG/CR-5450: HIGH-TEMPERATURE CRACK. ARREST TESTS MACCS CODE, Version 1.5 USING 152 MM-THICK SEN WlDE PLATES OF LOW UPPER-SHELF CARhE BASE MATERIAL: TESTS WD 2.2 AND WP 2.6.

ped 4704 V03: RELATIVE BIOLOGICAL EFFECTIVENESS (RBE)

OF FISSION NEUTRONS AND GAMMA RAYS AT OCCUPATIONAL BAUM J.W.

NUREG/CP 0110: PROCEEDtNGS OF THE INTERNATIONAL WORK.

EXPOSURE LEVELS Studies On The Gross And Microscopic Patholo-SHOP ON NEW DEVFLOPMENTS IN OCCUPATIONAL DOSE CON-gy Observed At Death Of Mice Exposed To 60 Equal Once Weekly TROL AND ALARA IMPLEMENTATION AT NUCLEAR POWER Doses Of Fossion....

PLANTS AND SIMILAR FACILITIES.

CASADA.D.A.

SEAUDOlN.B.F.

N1. REG /CR 5404 V01: AUXIUARY FEEDWATER SYSTEM AGING NUREG/CR-4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR STUDY.

LIGHT WATER REACTOR COMPONENTS OVERVIEW.

CHAMBERS,R.

BEAVERS,J.A-NUREG/CR-5273 V04:

SCDAP/RELAPS/ MOD 2 CODE NUREG/CR-5435. ENVIRONMENTAL EFFECTS ON CORROSION IN MANUAL, VOLUME 4 MATPRO A LIBRARY OF MATERIALS PROP-THE TUFF REPOSITORY.

ERTIES FOR LIGHT WATER REACTOR ACCIDENT ANALYSIS.

BLAKE,A.A.

CHANIN,0.L NUREG/CR-4674 V09: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR 4691 V02: MELCOR ACCIDENT CONSEOVENCE CODE DAMAGE ACCIDENTS:1988 A STATUS REPORT. Main Report And AP SYSTEM (MACCS) Volume 2: Model Desenption.

pendix A NUREG/CR 4691 V03: MELCOR ACCIDENT CONSEOUENCE CODE NUREG/CR-4674 V10: PRECURSORb TO POTENTIAL SEVERE CORE SYSTEM (MACCS).Volutte 3: Programmer's Reference Manual.

DAMAGE ACCIDENTS:1968 A STATUS REPORT.Appen:lixes B And C.

CHAO,K. S-NUREG/CP-0109 PROCEEDtNGS OF THE SEM!NAR ON LEAK-UREG CH-5256: COMPONENTS OF AN OVERALL PERFORMANCE gyp p

e ASSESSMENT METHODOLOGY.

CHINN,0 J WUREd/CR 4554 V06. SCANS (SHlPPING CASK ANALYSIS SYSTEM). A UAE 5424-EUCITING AND ANALYZING EXPERT MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING JUDGEMENTI Practical Guide' CASK DESIGN RF. VIEW. Volume 6 Theory Manual Buckhng Of Circu-BROSSEAU,0.A.

lar Cyhndncal Shells NUREG/CR-5398 TECHNICAL BASIS FOR REVIEW OF HIGH-LEVEL CH WASTi. REPOSITORY MODELING.

NUR G CR 73 V04-SCDAP/RELAPG/ MOO 2 CODE MANUAL, VOLUME 4 MATPRO A LIBRARY OF MATERIALS PROD-BROWN,D.P.

NUREG/CR-5386: BASIS FOR SNUBBER AGING RESEARCH NUCLE.

ERTIES FOR LIGHT WATER-REACTOR ACCIDENT ANALYSIS AR PLANT AGING RESEARCH PROGRAM.

NUREG/CR-4744 V03 N1. LONG TERM EMBRITTLEMENT OF CAST BROWN,G.

NUREG/lA 0012. RELAP5/ MOD 2 CALCULATIONS OF OECD LOFT DUPLEX STAINLESS STEELS IN LWR SYSTEMS Sermannual TEST LP-SB 01.

Report. October 1987. March 1988.

L 19 l

20 Personal Author Index

. CHUANG,Y.

DEMOSS.G.

NUREG/CR 5482: LABORATORY ANALYSIS OF FLUID FLOW AND NUREG/CR-5472: A RISK BASED REVIEW OF INSTRUMENT AIR SYS-SOLUTE TRANSPORT THROUGH A VARIARLY SATURATED FRAC-TEMS AT NUCLEAR POWER PLANTS.

'TURE EMBEDOED IN POROUS TUFF.

DEWIT.R.

CHUNG.H.M.

NUREG/CR-5450: HIGH-TEMPERATURE CRACK. ARREST TESTS NUREG/CR 4744 V03 N1: LONG-TERM EMBRITTLEMENT OF CAST USING 152 MM-THICK SEN WIDE PLATES Or LOW. UPPER-SHELF DUPLEX STAINLESS STEELS - IN LWR SYSTEMS Sorniennual Report October 1987. March 1988.

BASE MATERIAL: TESTS WP 2 2 AND WP 2.6.

CLETCHER.J.W.

DIAMOND.D.J.

NUREG/CR-5368: REACTIVITY ACCIDENTS.A Reassessment Of The NUREG/CR4674 V09, PRECURSORS TO POTENTIAL SEVERE CORE Desgr> Basis Events.

DAMAGE ACCIDENTS:1968 A STATUS REPORT. Main Report And Ap-perds A DIONNE.8.J.

NUREG/CR4674 V10: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCOENTS:1968 A STATUS REPORT. Appendixes B And NUREG/CP 0110: PROCEEDINGS OF THE INTERNATONAL WORK.

C.

SHOP ON NEW DEVELOPMENTS IN OV,UPATIONAL DOSE CON-TROL AND ALARA iMPLEMENTATON AT NUCLEAR POWER ggyg PLANTS AND SIMILAR FACILITIES.

NUREG/CR4731 V02. RESIDUAL LIFE ASSESSMENT OF MAJOR LIGHT WATER REACTOR COMPONENTS. OVERVIEW-DOOSE,C.A.

NUREG/CR 5376: OUALITY ASSURANCE AND VERIFICATION OF THE CORWIN.W.R.

MACCS CODE, Version 1.5.

NUREG/CR-5511: IRRADIATION EFFECTS ON STRENGTH ANO TOUGHNESS OF THREE WIRE SERIES ARC STAINLESS STEEL DOCTOR.S.R WELD OVERLAY CLADDING.

NUREG/CN 4469 V08. NONDESTRUCTIVE EXAMINATON (NDE ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER CRAMOND,W'4550 V01 RI: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR4469 Vos. NONDESTRUCTIVE EXAMINATON (NDE) RELL R.

REACTORS.Sermannual Report. October 1987. March 1988.

NUREG/CR CY: lNTERNAL EVENTS METHODOLOGY.

A9!LITY FOR INSERVICE INSPECTON OF LIGHT WATER REACTORS.Samannual Report.AprlLSeptember 1988.

CROUCH D.

NUREG!CR 4882: QUALIFICATION PROCESS FOR ULTRASONIC NUREG/CR 5527: RISK SENSITIVITY TO HUMAN ERROR IN THE LA.

TESTING IN NUCLEAR INSEiWICE INSPECTON APPLICATONS.

I SALLE PRA.

DOOD C.V.

DAMIANO,n.

NUPEG/CR 5 678. IMPROVED EDDYCURRENT INSFECTION FOR NUREG/C45479: CURRENT APPUCATIONS OF VIBRATION MONI.

STEAM GENERATOR TUBING. Progress Report For January 1985 To TORING AND NEUTRON NOISE ANALYSIS. Detection And Analysis Of December 1987, I

Structural Degradation Of Reactor Vessel internals Frorn Operational DOSANJH.S.S.

NUREG/C45318. MELT PROGRESSION. OXIDATION, AND NATURAL DAVIS,A.8.

CONVECTION IN A SEVERELY DAMAGED REACTOR CORE.

NUREG 1395 DR T: INDUSTRY PERCEPTIONS OF THE IMPACT OF THE U.S. NUCLEAR REGULATORY COMMISSION ON NUCLEAR DRAHOS,F.R.

POWER PLANT ACTIVITIES. Draft Report NUREG/CR4731 V02: RESIDUAL UFE ASSESSMENT OF MAJOR DAVIS,E.C.

LIGHT WATER REACTOR COMPONENTS. OVERVIEW.

NUREG/CR-5229 V02: TMb2 EPICOR-il RESIN /UNER INVtiSTIGATiON:

DROUIN.M T.

LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR4550 V01 Ri: ANALYSIS OF CORE DAMAGE FREOUEN-FISCAL YEAR 1989. Annual Report.

CY: INTERNAL EVENTS METFOOOLOGY.

DAVIS.P.A.

ERICSON,0.M.

NUREG/CR-5256: COMPONENTS OF AN OVERALL PERFORMANCE NUREG/CR4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-ASSESSMENT METHODOLOGY.

CY: INTERNAL EVENTS METHOOOLOGY, DEAN.R.S.

ERIKSON R.L.

NUREG/C4466t: CLOSEOUT OF IE BULLETIN 85-03: MOTOR OPER.

NUREG/CR-5169: MOBILf2ATON AND TRANSPORT OF URANIUM AT ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-URANIUM MILL TAluNGS DISPOSAL SITES. Application Of A ChemL SIENTS DUE TO IMPROPER SWITCH SETTLNGS.

cal Transport Model.

NUREG/C45286: CLOSEOUT OF IE BULLETIN 7917: PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS EVANS,0D.

NUREG/CR-5289 CLOSEOUT OF IE BULLETIN 79 23: POTENTIAL NUREG/CR 5482: LABORATORY ANALYSIS OF FLUlO FLOW AND FAILURE OF EMERGENCY DIESEL GENERATOR FIELD EXCITER SOLUTE TRANSPORT THROUGH A VARIABLY SATURATED FRAC.

TRANSFORMER.

TURE EMBEDOED IN POROUS TUFF.

NUREG/CR-5298. CLOSEOUT OF lE BULLETIN 85-01; STEAM BIND-ING OF AUXIUARY FEEDWATER PUMPS.

EVANS.J.S.

NUREG/CR-5302 CLOSEOUT OF IE BULLETIN 8010: CC7TAMINA.

NUREG/CR-4214 R01 Pl. HEALTH EFFECTS MODELS FOR NUCLEAR TION OF NONRADIOACTIVE SYSTEM AND RESULTING POTENTIAL POWER PLANT ACCOENT CONSEOUENCE ANALYSIS Low LET FOR UNMONITOREO, UNCONTROLLED RELEASE OF RAOlOACTiv.

ITY TO ENVIRONMENT.

Radiation.Part I: Introduction, integration And Summary.

NUREG/C45307: CLOSEOUT OF IE BULLETIN 8002: INADEOUATF FIELDS.R.J.

QUAUTY ASSURANCE FOR NUCLEAR SUPPUED EQUIPMENT, NUREG/CR5450: HIGH-TEMPERATURE CRACK. ARREST TESTS DEEDS,W.E.

USING 152.MM THICK SEN WlOE PLATES OF LOW UPPE4 SHELF BASE MATERIAL: TESTS WP 2.2 AND WP.2.6.

NUREG/C45478. IMPROVED E00Y. CURRENT INSPECTION FOR STEAM GENERATOR TUBING. Progress Report For January 1985 To FITZPATRICK.R.

December 1987.

NUREG/C45368: REACTIVITY ACC OENTS A Reassessment Of The Desagr> Basis Events.

NUREG/CR4469 V08: NONDESTRUCTIVE EXAMINATON (NDE) RELi-FOLE Y,W.J.

ABlUTY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR-4661: CLOSEOUT OF IE BULLETIN 85 07 MOTO40PER-REACTORS Semaannual Report. October 1987 March 1988.

NUREG/CR-4469 V00 NONDESTRUCTIVE EXAMINATION (NDE) REll-ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-SIENM DUE TO IMPROPER SWITCH SETTINGS.

ABILITY FOR INSERVICE INSPECTON OF UGHT WATER NUREG/CR-5286: CLOSEOUT OF IE BULLETIN 7917. PIPE CRACKS IN REACTORS Sernannual Report.Apni-September 1988.

STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS.

Personal Author index 21 NUREG/CR-5269: CLOSEOUT OF IE BULLETIN 79 23 POTENTIAL '

HAGGAG,F.M.

FAILURE OF EMERGENCY DIESEL GENERATOR FIELD EXCITER NUREG/CR-5511; IRRADIATON EFFECTS ON STRENGTH AND TOUGHNESS OF THREE WIRE SERIES ARC ST AINLESS STEEL TRANSFORMER.

1 NUREG/CR-5298 CLOSEOUT OF IE BULLETIN 85 01: STEAM BIND-WELD OVERLAY CLADDING.

ING OF AUXILIARY FEEDWATER PUMPS NUREG1CR-5302: CLOSEOUT OF IE BULLETIN 8010: CONTAMINA.

HAGRMAN,D.L.

TON OF NONRADIOACTIVE SYSTEM AND RESULTING POTENTIAL NUREG/CR-5273 V04:

SCDAP/RELAP5/ MOD 2 CODE FOR UNMONITORED, UNCONTROLLED RELEASE OF RADIOACTIV' MANUALVOLUME 4.MATPRO - A UBRARY OF MATERIALS PROP-NURE CR 5 7 CL E'OUT OF IE BULLETIN 80 02: INADEQUATE QUALITY ASSURANCE FOR NUCLEAR SUPPLIED EQUIPMENT.

HALDEMAN,WA NUREG/CR-5482 LABORATORY ANALYSIS OF FLUID FWN AND SOLUTE TRANSPORT THROUGH A VARIABLY SATURAT& FRAC.

NUREG CR-4668: DAMAGED FUEL EXPERIMENT OF 1.Results And TURE EMBEDDED IN POROUS TUFF.

Anahses.

HALL,P.C.

FULLWOOD R.

NUREG/lA 0012: RELAPS/ MOD 2 CALCULATIONS OF OECD LOFT NUREG/CR-5419-AGING ASSESSMENT OF INSTRUMENT AIR SYS, TEST LP-SB-01 TEMS IN NUCLEAR POWER PLANTS.

HAMPTON,N.L.

GALLEGOS.D.P.

NUREG/CR-5273 V04:

SCDAP/REL AP5/ MOD 2 CODE HUREG/CR-5256: COMPONENTS OF AN OVERALL PERFORMANCE MANUALVOLUME 4.MATPRO A U6RARY *)F MATERIALS PROP.

ASSESSMENT METHODOLOGY.

ERTIES FOR LIGHT WATER-REACTOR ACCIL ENT ANALYSIS.

NUREG/CR-5398: TECHNICAL BASIS FOR REVIEW OF HIGH-LEVEL WASTE REPOSITORY MODEUNO.

HARPER.F.7, NUREG/CR-4550 V01 R1: ANALYSIS OF COR'i DAMAGE FREQUEN-GARDNER,J.B.

NUREG/CR-4731 V02. RESIDUAL LIFE ASSESSMENT OF MAJOR CY: INT ERNAL EVENTS METHODOLOGY.

LIGHT WATER REACTOR COMPONENTS OVERVIEW.

HARWOOD,C, GARNER,R.W.

NUREG/lA 0013. RELAPS/ MOD 2 CALCULN IONS OF OECD-LOFT NUREG/CR 4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR TEST LP SB-03.

UGHT WATER REACTOR COMPONENTS OVERVIEW.

MAWTHORNE,J.R.

GA M M NUREG/CR-5492: INVESTIGATIONS OF MRADIATON-ANNEAL REIR-NUREG/CR 4668. DAMAGED FUEL EXPERIMENT DF 1.Results And RADIATION (IAR) PROPERTIES TRENDS OF RPV WELDS. Phase 2 NU E / R 4671: THE OF 4 FUEL DAMAGE EXPERIMENT IN ACAR NUR C 5493: INFLUENCE OF FLUENCE RATE ON RADIATION-IN.

WITH A BWR CONTROL BLADE AND CHANNEL BOX.

DUCED MECHANICAL PROPERTY CHANGES IN REACTOR PRES-SURE VESSEL STEELS Final Report On Exploratory Expenrnents.

GAUNTT,R O.

NU G/CR 4668. DAMAGED FUEL EXPERIMENT OF.1.Results And HEASLE R,P.G.

NUREG/CR-4469 V08: NONDESTRUCTl/E EXAMINATION (NDE) RELi-NURE R 4671: THE DF 4 FUEL DAMAGE EXPER: MENT IN ACAR ABidTY FOR INSERVICE INSPECTION OF UGHT WATER WITH A BWR CONTROL BLADE AND CHANNEL, BOX.

REACTORS.Sermannual Report, October 1987 March 1988 NUREG/CR-4469 V09: NONDESTRUCTIVE EXAMINATION (NDE) REU NUREG/CR 5395 V02: MULTILOOP INTEGRAL SYSTEM TEST (MIST):

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER GEISSLER.G.O.

REACTORS.Semannual Report, April-September 1988.

FINAL REPORT. Test Group 30. Mapping Tests.

HENNICK.A.

GOOD.M.S.

NUREG/CR 4661: CLOSEOUT OF IE BULLETIN 85 03: MOTOR OPER-NUREG/CR 4469 V08: NONDESTRUCTIVE EXAMINATION (NDE) REll-ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS.Semannual Report, October 1987 March 1988 SIENTS DUE TO IMPROPER SWITCH SETTINGS.

NUREG/CR-4469 VO9: NONDESTRUCTIVE EXAMINATION (NDE) REU-NUREG/CR-5286: CLOSEOUT OF lE BULLETIN 79-17. PIPE CRACKS IN ABILITY FOR INSERVICE INSPECTION OF UGHT WATER STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS.

l REACTORS. Semiannual Report,Apnl Septernber 1988-NUREG/CR 5289-CLOSEOUT OF IE BULLETIN 79-23: POTENTIAL l

FAILURE OF EMERGENCY DIESEL GENERATOR FIELD EXCITER GOODmCH,M.T.

TRANSFORMER.

NUREG/CR-5256: COMPONENTS OF AN OVERALL PERFORMANCE NUREG/CR-5298. CLOSEOUT OF IE BULLETIN 8501: STEAM BIND-ASSESSMENT METHODOLOGY.

ING QF AUXILIARY FEEDWATER PUMPS.

NUREG/CR-5398: TECHNICAL BASIS FOR REVIEW OF HIGH-LEVEL NUREG/CR-5302: CLOSEOUT OF IE BULLETIN 80-10' CONTAMINA-WASTE REPOSITORY MODEUNG.

TION OF NONRADIOACTIVE SYSTEM AND RESULTING POTENTIAL FOR UNMONITORED, UNCONTROLLED RELEASE OF RADIOACTIV-GRAHN,D.

NUREG/CR-4704 V03: RELATIVE BIOLOGICAL EFFECTIVENESS (RBE)

NU E CR 07 CLOSE'OUT OF IE BULLETIN 8002: INADEOUATE OF FISS!ON NEUTRONS AND GAMMA RAYS AT OCCUPATIONAL QUAUTY ASSURANCE FOR NUCLEAR SUPPUED EQUIPMENT.

EXPOSURE LEVELS. Studies On The Gross And Microscopic Patholo-gy Observed At Death Of Mice Exposed To 60 Equal Once Weekly HIGGINS J.

Doses Of Fismon...

NUREG/CR-5527: RISK SENSITIVITY TO HUMAN ERROR IN THE LA-SALLE PRA.

GREEN E.R.

NUREG/CR.4469 V08. NONDESTRU':TIVE EXAMINATION (NDE) REU-HISER,A.L NUREG/CR 5492: INVESTIGATIONS OF IRRADIATION-ANNEAL REIR-ABluTY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS Semiannuai neport.0ctober 1987 March 1988 RADIATION (IAR) PROPERTIES TRENDS OF RPV WEl,DS. Phase 2 NUREG/CR-4409 V09 NONDESTRUCTIVE EXAMINATION (NDE) RELI.

Final Report ABluTY FOR INSERVICE INSPECTION OF UGHT WATER NUREG/CR 5493: INFLUENCE OF FLUENCE RATE ON RADIATION-lN-REACTORS.Semannual Report.Apnt-September 1988 DUCED MECHANICAL PROPERTY CHANGES IN REACTOR PRES-SURE VESSEL STEELS Final Report On Exploratory Experiments.

GUPPY,J.G.

NUREG/CR-3668. MINET CODE DOCUMENTATION.

NUREG/CR-4659 V03 SEISMIC FRAGILITY OF NUCLEAR POWER GU20WSKl,R.V.

PLANT COMPONENTS (PHASE II)Setchgear. l&C Panels (NSSS)

NUREG/CR-5256; COMPONENTS OF AN OVERALL PERFORMANCE ASSESSMENT METHODOLOGY.

And Relays

22 Personal Author index HOHORST.J.K.

Structural Degradaten Of Reactor vessel Internais From Operatonal NUREG/CR-5273 V04:

SCDAP/RELAP5/ MOD 2 CODE Aging MANUAL, VOLUME 4 MATPRO. A LIBRARY OF MATER!ALS PROP.

ERTIES FOR LIGHT WATER-REACTOR ACCIDENT ANALYSIS.

LAATS.E.T.

NUREG/CR-5273 V04:

SCDAP/RELAP5/ MOD 2 CODE HOSTETLER.CJ.

NUREG/CR-5169-MOBILIZATON AND TRANSPORT CF URANIUM AT MANUAL VOLUME 4.MATPRO. A LIBRARY OF WATERIALS PROP.

ERTIES FOR LIGHT WATER REACTOR ACCIDENT ANALYSIS.

URANIUM MILL TAIUNGS DISPOSAL SITES Applicaten Of A Chemi-cal Transport Model.

LAM 8ERT.H.E.

HSU.C.J NVREG/CR 5477: AN EVALUATION OF THE REllABILITY AND USE.

HUREG/CR-5368. REACTIVITY ACCOENTS A Reasoessment Of The FULNESS OF EXTERNAL INITIATOR PRA METHODOLOGIES.

DeseBasis Events.

gay,y, HUANO.P.

NUREG/CR-3145 V08: GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-5484 PH SENSORS BASED ON IROlUM OXIDE.

WESTERN OHIOINDIANA REGION. Annual Report. October 1968 September 1989.

ISKANDER,S.K.

NUREG/CR-5450- HIGH TEMPERATURE CRACK ARREST TESTS LEHNFR J.R.

USING 152-MM-THICK SEN WOE PLATES OF LOW. UPPER-SHELF NUREG/CR-5474: ASSESSMENT OF CANDIDATF ACCIDENT MAN-BASE MATERIAL: TESTS WP 2.2 AND WP-2 6.

AGEMENT STRATEGIES.

JACO8 SON.J.

LO,T.

NUREG/CR-3145 V08-GEOPHYSICAL INVESTIGATONS OF THE NUREG/CR-4554 V06: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A WESTERN OHIOINDIANA REGON Annual Reoort,0ctober 1988 MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING September 1989.

CASK DESIGN REVIEW. Volume 6. Theory Manual Buckling Of Circu-JASTROW,J.D.

lar Cylmdncal Shells NUREG/CR 4554 V07: SCANS (SHIPPtNG CASK ANALYSIS SYSTEM)

NUREG/CR-5229 V02: TMI-2 EPICOR-il RESIN /UNER INVESTIGATON:

LOW LEVEL WASTE DATA _.. DEVELOPMENT PROGRAM FOR A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING FISCAL YEAR 1989. Annual Report.

CASK DE3iGN REVIEW. Volume 7: Theory Manual Puncture Of Shq>.

ping Casks.

JOHNSON,A.S.

LOFOREN E.

NURM/C45491: SHIPPINGoORT ST ATON AGING FVALUATON.

NUREG/CR 5472. A RISWBASED REVIEW OF INSTRUMENT AIR SYS-JOW.H.

TEMS AT NUCLEAR POWER PLANTS.

NUREG/CR-4691 V02: MELCOR ACCIDENT CONSEQUENCE CODE C DNT CONSEQUENCE CODE NUR G O 20 V13 N11: UCENSED OPERATING REACTORS STATUS NR R 60 L

SYSTEM (MACCS) Volume 3: Programmer's Reference Manual.

SUMMARY

REPORT. Data As Of October 30,1989.(Gray Book 1)

NUREG-0020 V13 N12. UCENSED OPERATING P2 ACTORS STATUS KASSIR.M.K.

SUMMARY

REPORT. Data An Of November 30, tree (Gray Book I)

NUREG/9R-4659 V03. SEISMIC FRAGluTY OF NUCLEAR POWER NUGEG 0020 V14 NO1: UCENSED OPERATING REACTORS STATUS PLANT COMPONENTS (PHASE il).Switchgear. I&C Panels (NSSS)

SUMMARY

REPORT. Data As Of December 311989.(Gray Book I)

LOW.S.R-KEENEY. WALKER NUREG/CR-5450: HIGH TEMPERATURE GRACK ARREST TESTS NUREG/CR-5450- HIGH TEMPERATURE CRACK ARREST TESTS USING 152 MM-THICK SEN WIDE PLATE! OF LOW UPPER SHELF USING 152.MM TH!CK SEN WIDE PLATES OF LOW UPPER SHELF BASE MATERIAL: TESTS WP.2.2 AND WP 2 6.

BASE MATIRIAL: TESTS WP 2.2 AND WP 2.6.

LU.S.C.

KEMNER M.L.

NUREG/CR-5506: PRELIMINARY STRUCTUTAL EVALUATION OF NUREG/CR 5169. MOBlU2ATON AND TRANSPORT OF URANIUM AT TROJAN RCL SUBJECT TO POSTULATED R'V SUPPORT FAILURE.

URANIUM MILL TAluNGS DISPOSAL SITES Applicaten Of A Chemi-cal Transport Model LUCKAS,W.

KENNEDY,W.E.

NUREG/CR-5527: RISK SENSITIVITY TO HUMAN ERROR IN THE LA.

SALLE PRA^

NUREG/CR-5512 DRF FC-RESOUAL RADIOACT.A CONTAMINATION FROM DECOMMISSIONING. Technical Basis For Translatin0 Contami-LUCKAS,W.J.

nat on Levels To Annual Dose. Draft Report For Comment.

NUREG/CR-5474: ASSESSMENT OF CANDIDATE ACCOENT MAN-KHAN T.A.

AGEMENT STRATEGIES.

NUREG/CP-0110: PROCEEDINGS OF THE INTERNATIONAL WORK-MACDONALD,P.E.

SHOP ON NEW DEVELOPMENTS IN OCCUPATIONAL DOSE CON

• NUREGICR-4731 V02: RESOUAL UFE ASSESSMENT OF MMOR TROL AND ALARA iMPLEMENTATON AT NUCLEAR POWER PLANTS AND SIMILAR FACluTIES.

UGHT WATER REACTOR COMPONENTS. OVERVIEW.

KIRKWOOD,8.J.

MACLELLAN J.A.

NUREG/C44731 V02: RESIDUAL UFE ASSESSMENT OF MAJOR NUREG/C45516. CAUSES OF FAluNG THE DRAFT ANSI STANDARD UGHT WATER REACTOR COMPONENTS OVERVIEW.

Nt3 30 RADIOBIOASSAY PERFORMANCE CRITERION FOR MINI-MUM DETECTABLE AMOUNT.

KIRKWOOD R.

NUREG 1316: TECHNICAL FINDINGS AND REGULATORY ANALYSIS MA RELATED TO GENERIC ISSUE 70 Evaluation Of Power Operated UREG/CH-4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-Rehef Valve And Block Valve Rehabihty in PWR Nuclear Power Plants.

CY: INTERNAL EVENTS METHODOLOGY.

KUNGENFUS,J.A.

MARCH-LEUBA.J.

NUREG/CR-5395 V10. MULTILOOP INTEGRAL SYSTEM TEST (MIST).

NUREG/C45421: LAPUR USER'S GUIDE.

FINAL REPORT. RELAPS/ MOD 2 MIST Analysis Compansons.

MARSHALL,A.C.

KREIDER,K.G.

NUREG/CR4668. DAMAGED FUEL EXPERIMENT OF 1.Results And NUREG/CR-5484. PH SENSORS BASED ON IROlUM OxlDE.

Analyses KRYTER,R.C.

MARSHALL,N.H.

NUREG/CR-5479. CURRENT APPLICATIONS OF VlBRATON MONI-NUREG/CR 5376: OVALITY ASSURANCE AND VERIFICATION OF THE TORING AND NEUTRON NOISE ANALYSIS Detection And Analysis Of MACCS CODE. Verson 1.5.

I l

p.

Personal Author Index 23 NEPSEE,T.C.

MARWIL,E.8.

NUREG/CR-5376 OVALITY ASSURANCE AND VERIFICATION OF THE NUREG/CR-3668. MINET CODE DOCUMENT ATION MACCS CODE, Version 1.5.

O'NARA,J.

N G CF15273 V04.

SCDAP/RELAPS/ MOO 2 CODE S E PR MANUALVOLUME 4 MATPRO. A LIBRARY OF MATERIALS PROP.

ERTIES FOR LIGHT. WATER-REACTOR ACCIDENT ANALYSIS.

OLAOUE.N.E.

NUREG/CR-5398 TECHNICAL DASIS FOR REVIEW OF HIGH. LEVEL MCCLUNG,R.W.

WASTE REPOSITORY MODELING.

NUREG/CF1-5478. IMPROVED EDDY CURRENT INSPECTION FOR STEAM GENERATOR TUBING Progress Report For January 1985 To OLIVE,K L NUREG 1350 V02-NUCLEAR REGULATORY COMMISSION INFORMA.

TION DIGEST.1990 Edihort MCCOMAS M.L NUREG/CF1-5273 V04.

SCDAP/RELAP5/ MOD 2 CODE OL8EN C.S.

MANUALVOLUME 4 MATPRO. A LIB 81ARY OF MATERIALS PROP.

NOREG/CR-5273 V04.

SCDAP/RELAP5/ MOD 2 CODE ERTIES FOR LIGHT. WATER. REACTOR ACCIDENT ANALYS!S.

MANUALVOLUME 4.MATPRO. A LIBRAFlY OF MATERIALS PROP.

ERTIES FOR LIGHT WATER. REACTOR ACCIDENT ANALYSIS.

MCCONNELL.J.W, NUREG/CR-5229 V02: TMI.2 EPICOR ll RESIN / LINER INVESTIGATION.

LOW. LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR OT ADUY,P.J.

NUREG/CR 6421: LAPUR USER'S GUIDE.

FISCAL YEAR 191; Annual Report BCNAMARA.N.

OTT,LJ.

.I NUREG-0837 V09 N04: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4671: THE OF 4 FUEL DAMAGE EXPERIMENT IN ACAR NETWORK. Progress Flepod October. December 1989.

WITH A DWR CONTROL DLADE AND CHANNEL DOX.

MCNEIL,K.A*

PALMER G.R NUfiEG/CR-5273 V04-SCDAP/FIELAPS/ MOD 2 CODE NUREd/CR 5386: DASIS FOR SNUDBER AGING RESEARCH: NUCLE.

MANUALVOLUME 4 MATPRO A LIBRARY OF MATERIALS PROP.

AR PLANT AGING RESEARCH PROGRAM' ERTIES FOR LIGHT. WATER.FIEACTOR ACCIDENT ANALYSIS.

PARECE.M.V.

MEYER.LC NUREG/CR-5395 V10: MULTILOOP INTEGRAL 3YSTEM TEST (MIST):

NUREG/CR4731 V02: RESIDUAL LIFE ASSESSMENT OF MMOR FINAL REPORT, RELAP5/ MOD 2 MIST Analysis Compensons.

LIGHT WATER REACTOR COMPONENTS OVERVIEW.

PEDEFtSON,C.D.

MEYER,M.A.

NUREG-1395 DAFT: INDUSTRY PERCEPTIONS OF THE IMPACT OF NUREG/CR-5424.

EllCITING AND ANALYZING EXPERT THE U S NUCLEAR REGULATORY COMMISSION ON NUCLEAR JUDGEMENT.A Pracacal Gurde POWER PLANT ACTIVITIES. Draft Report MILLER.J.D.

NUREG/CR 5405: ANALYSIS OF SHELL. RUPTURE FAILURE DUE TO PELOQUIN,R.A.

HYPOTHETICAL ELEVATED. TEMPERATURE PF1ESSUR12ATION Of NUREG/CA.5512 DAF FC RESIDUAL RADIOACTIVE CONTAMINATION THE SEQUOYAH UNIT 1 STEEL CONTAINMENT DUILDING.

FROM DECOMMISSIONING. Techrucal Basis For Translahng Contami-nabon Levels To Annual Dose. Draft Repori For Ccmment MILLER,R.L NUREG/CF1-5273 V04.

SCDAP/RELAP5/ MOD 2 CCDE PEPPER.S.E.

MANUALVOLUME 4 MATPF10 A LIBRARY OF MATERIALS PPOP' NUREG/CR 4659 V03. SEISMIC FRAGlLITY OF NUCLEAR POWER EFITIES FOFI LIGHT. WATER. REACTOR ACCIDENT ANALYSIS.

PLANT COMPONENTS (PHASE il)Switchgoar, l&C Panels (NSSS)

And Relays.

MINARICK,J.W.

NUREG/CR4674 V09' PRECURSORS TO POTENTIAL SEVERE CORE PHILSIN

.S.

DA CE ACCIDENTS:1988 A STATUS REPORT. Main Fleport And Ap-COSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE-NU EG/CR-4674 VIO: PRECURSORS TO POTENTIAL SEVERE CORE RJAL LICENSEE OPERATIONS.

DAMAGE ACCIDENTS:1968 A STATUS FIEPORT.Appendixen D And PRICE,LL NUREG/CR-5256. COMPONENTS OF AN OVERALL PERFORMANCE MOK,0.C.

NUREG/CF14554 V06: SCANS (SHIPPING CASK ANALYSIS SYSTEM). A ASSESSMENT METHODOLOGY.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING NUREGICR-5398: TECHNICAL BASIS FOR REVIEW OF HIGH-LEVEL CASK DESIGN FIEVIEW. Volume 6 Theory Manual Buckhng Of Circu.

WASTE REPOSITORY MODELING.

RASMUSSEN,T.C.

NUREG/CR.5482. LABORATORY ANALYSIS OF FLUID FLOW AND MORGAN.W.A.

NUREG/CR 5273 V04:

SCDAP/RELAP5/ MOD 2 CODE SOLUTE TRANSPORT THROUGH A VARIABLY SATURATED Ff1AC.

MANUALVOLUME 4 MATPRO A LIBRARY OF MATERIALS PROP-TURE EMBEDDED IN POROUS TUFF.

ERTIES FOR LIGHT WATER REACTOR ACCIDENT ANALYSIS.

RElLK.O.'

UREG C 1-4882. QUALIFICATION PROCESS FOR ULTRASONIC n ses TESTING IN NUCLEAR INSERVICE INSPECTION APPLICATIONS.

REYMANN,0.A.

NUFIEG/CR-5273 V04 SCDAP'/RELAPS/ MOD 2 CODE NUREG/

511: IRRADIATION EFFECTS ON STRENGTH AND MANUALVOLUME 4 MATPRO A L10RARY OF MATERIALS PROP-TOUGHNESS OF THREE. WIRE SERIES ARC STAINLESS STEEL ERTIES FOR LIGHT. WATER REACTOR ACCIDENT ANALYSIS.

WELD OVERLAY CLADDING.

RITCHIE.LT, NAUS,0.J-NUREGICR-4691 V02: MELCOR ACCIDENT CONSEOUENCE CODE NUREG/CR-5450- HIGH-TEMPERATURE CRACK. ARREST TESTS USING 152 MM-THICK SEN WIDE PLATES OF LOW-UPPER SHELF SYSTEM (MACCS) Volume 2. Model Desenpbon.

BASE MATERIAL: TESTS WP 2.2 AND WP 2.6 ROGERS R.D.

NUREG/CR 5229 V02. TMi-2 EPICOR.li RESIN!LINEH INVESTIGATION.

NEASE.R.

LOW LEVEL WASTE DATA DASE DEVELOPMENT PROGRAM FOR NUREG 1214 ROS HISTOf'ICAL DATA

SUMMARY

OF THE SYSTEMAT.

IC ASSESSMENT OF LICENSEE PERFORMANCE FISCAL YEAR 1989 AMuaf Report

24 Personal Author index ROLLSTINJ.

SVPE.T.T.

NUREG/CR 5381: ECONOMIC RISK OF CONTAMINATION CLEANUP NUREG/CR-4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-UOSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE-CY: INTERNAL EVENTS METHODOLOGY.

3 RIAL LICENSEE OPERATONS.

TARLOY.M.J.

ROLLSTINJ.A.

NUREG/CR-5484: PH SENSORS BASED ON IRIDIUM OXIDE.

NUREG/CR-4691 V02: MELCOR ACCIDENT CONSEQUENCE CODE SYSTEM (MACCS) Volume 2: Model Desenption.

TAYLOR T.T.

NUREG/CR-4691 V03: MELCOR ACCIDENT CONSEQUENCE CODE SYSTEM (MACCS) Volume 3. T,vy e -.'s Reference Manual NUREG/CR-4469 V08. NONDESTRUCTIVE EXAMINATION (NDE) RELL ABILITY FOR INSERVICE INSPECTION OF UGHT WATER ROTHLEDER,8.

REACTORS.Sermannual Report. October 1987 March 1988.

NUREG/CR-5472. A RISK BASTD REVIEW OF INSTRUMENT AIR SYS-NUREG/CH-4469 VDO. NONDESTRUCTIVE EXAMINATON (NDE) RELL TEMS AT NUCLEAR POWER PLANTS.

ABILITY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS.Sermannual Report, April Boptember 1988.

RUGER C.

NUREG/CR4882: OVAUFICATON PROCESS FOR ULTRASONIC NUREG/CR-5472: A RISK BASED REVIEW OF INSTRUMENT AIR SYS-TESTING IN NUCLEAR INSERVICE INSPECTION APPUCATONS.

TEMS AT NUCLEAR POWER PLANTS.

THOMPSON M gALogo,J,H.

NUREG/CR-5435: ENVIRONMENTAL EFFECTS ON CORROSION IN NUREG/CR 5381: ECONOMIC RISK OF CONTAMINATION CLEANUP THE TUFF REPOSITORY.

COSTS RESULilNG FROM IAPGE NONRE. ACTOR NUCLEAR MATE.

RIAL UCENSEE OPERATONS.

THOMSONJ.F.

NUREG/CR-4704 V03: RELATIVE BIOLOGICAL EFFECTIVENESS (RBE)

SEMANCin,5.

OF FISSION NEUTRONS AND GAMMA RAYS AT OCCUPATIONAL NUREG/CR-5484. PH SENSORS BASED ON IRIDIUM OXOE.

EXPOSURE LEVELS Studes On The Gross And Microscopec Pathoto-SERVER.W.L Oy Otmarved At Death Of Mice Erposed To 60 Equal Once Weckly Dosos Of Fission.

NUREG/CM 731 V02: RESIDUAL UFE ASSESSMENT OF MAJOR UGHT WATER REACTOR COMPONENTS. OVERVIEW.

TOLLIJ.E.

NUREG/CR 5376: OVAUTY AS$URANCE ANO VERIFICATON OF THE NU E I1392: LEAKAGE OF AN IRRADIA10R SOURCE THE JUNE 1988 GEORGIA RSIINCIDENT.

VAN TUYLE,GJ.

SHAH.V.N.

NUREG/CR-3668: MINET CODE DOCUMENTATION.

NUREG/CR 4731 V02 RESIDUAL LIFE ASSESSMENT OF MAJOR UGHT WATER REACTOR COMPONENTS. OVERVIEW.

VANDENKlEBOOM NUREG/CR-5474. ASSESSMENT OF CANDIDATE ACCOENT MAN.

84EGEL.E.A. -

AGEMENT STRATEG!ES.

NUREG/CM731 V02: RESIDUAL UFE ASSECSMENT OF MAJOR UGHT WATER REACTOR COMPONENTS. OVERVIEW.

Y

'MCbtMWNNWMWNMM SIMONENJ.A.

TEMS IN NUCLEAR POWER PLANTS.

HUREG/CR-4469 V08. NONDESTRUCTIVE EXAMINATON (NDE) F.EU.

NUREC/CR 5472: A RISK. BASED REVIEW OF INSTRUMENT AIR SYS-ABluTY FOR INSERVICE INSPECTON OF UGHT WATER TEMS AT NUCLEAR POWER PLANTS.

REACTORS.Sernannual Report. October 1987. March 1988.

NUREG/CR 4469 V09. NONDESTRUCTIVE EXAMINATON (NDE) REU.

WAHLK.K.

ABluTY FOR INSERVICE INSPECTON OF UGHT WATER NUREG/CR 5256: COMPONENTS OF AN OVERALL PERFORMANCE REACTORS.Sermannuai neport.Apnl-September 1988 ASSESSMENT METHODOLOGY.

NUREG/CH 5398: TECHNICAL BASIS FOR REVIEW OF HIGH4EVEL SINHA.U.P.

WASTE REPOSITORY MODEUNG.

NUREG/CM731 V02: RESIDUAL UFE ASSESSMENT OF MAJOR LIGHT WATER REACTOR COMPONENTS OVERVIEW.

WARE.A.G.

SFANNERJ.C.

NUREG/CR 4731 V02: RESIDUAL UFE ASSESSMENT OF MAJOR NUREG/CR 4469 V08: NONDESTRUCTIVE EXAMINATION (NDE) RELl-UGHT WATER REACTOR COMPONENTS. OVERVIEW.

ABluTY FOR INSERVICE INSPECTON OF UGHT WATER WEATHERBY,J.R.

RE ACTORS Sermannual Report October 1987. March 1988.

NUREG/CR-5476: POSTTEST ANALYSIS OF A 1.6-SCALE REIN-NUREG/CR-4469 V09: NONDESTRUCTIVE EXAMINATON (NDE) RELl*

FORCED CONCRETE REACTOR CONTAINMENT BUILDING.

ABILITY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS.Sermannual Report.Apnt September 1988.

WEISS,A.J.

NUREG/CR-4882: OVAUFICATION PROCESS FOR ULTRASONIC NUREG/CP4105 V01: PROCEEDINGS OF THE SEVENTEENTH TESTING IN NUCLEAR INSERVICE INSPECTION APPUCATONS.

WATER REACTOR SAFETY INFORMATION MEETING.

SPRUNG.L NUREG/CP-0105 V02; PROCEEDINGS OF THE SEVENTEENTH NUREG/CR-4691 V02: MELCOR ACCOENT CONSEQUENCE CODE WATER REACTOR SAFETY INFORMATION MEETING SYSTEM (MACCS). Volume 2: Model Descr:pton.

NUREG/CP 0105 V03: PROCEEDtNGS OF THE SEVENTEENTH WATER REACTOR SAFETY INFORMATION MEETING.

STALKER,K.T.

NUREG/CR 2331 V09 N3: SAFETY RESEARCH PROGRAMS SPON-NUREG/CM668: DAMAGED FUEL EXPERIMENT DF.1.Results /.nd SORED BY OFFICE OF NUCLEAR REGULATORY Analyses.

RESEARCH. Progress Report. July September 1989.

STEINBRECHER H.

WERRY,E.V.

NUREG/CM661: CLOSEOUT OF IE BULLETIN 85-03: MOTOR OPER.

NUREG/CR 5386: BASIS FOR SNURBER AGING RESEARCH: NUCLE.

ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN.

AR PLANT AGING RESEARCH PROGRAM.

SIENTS DUE TO IMPROPER SWliCH SETTINGS.

WHEELE R,T.A.

STRUCKMEYER,R.

NUREG/CR-4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG4837 V09 N04. NRC TLD DIRECT RADIATON MONITORING CY:lNTERNAL EVENTS METHODOLOGY.

NETWORK. Progress Report. October December 1989.

SU8UDHl.M.

NUREG/CP-0IO9 PROCEEDtNGS OF THE SEMINAR ON LEAK.

NUREGrCR 54i3. AGING ASSESSMENT OF INSTRUMENT AIR SYS-BEFORE BREAK.Furthc Developments in Regulatory Polectos And TEMS IN NUCLEAR POWER PLANTS-Supporting Research.

l

f Personal Author inden.

25 WOIISA VOUno,CJ.

NUMEG/CR 6627; RISK SENSITIVITY TO HUMAN ERROR IN THE LA. -

NUREG/CR4145 V06. GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIOINDIANA REGION. Annual Report,0ctober 1986 SALLE PRA,

September 1960.

I.

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

ALARA NUREG/CR-5421: LAPUR USER'S GUIDE.

NUREG/CP 0110: PROCEEDINGS OF THE INTERNATIONAL WORK.

SHOP ON NEW DEVELOPMENTS IN OCCUPATIONAL DOSE CON-B6oseeay TROL AND ALARA IMPLEMENTATION AT NUCLEAR POWER NUREG/CR-5516: CAUSES OF IAILING THE DRAFT ANSI STANDARD PLANTS AND SIMILAR FACILITIES.

N13.30 RADIOBIOASSAY PERFORMANCE CRITERION FOR MINI-MUM DETECTABLE AMOUNT, ANSI Standerd N13 NUREG/CR-5516. CAUSES OF FAILING THE DRAFT ANSI STANDARD D6olog6 cal Redestion Effect N1330 RADDDIOASSAY PERFORMANCE CRITERION FOR MINI-NUREG/CR-4704 V03: RELATIVE BIOLOGICAL EFFECTIVENESS (ABE)

MUM DETECTABLE AMOUNT.

OF FISSION NEUTRONS AND GAMMA RAYS AT OCCUPATIONAL EXPOSURE LEVELS Studies On The Gross And Mgroscopic Patholo-w A oa wed to 60 Ewal OeWW V1 NO3: REPORT TO CONGRESS ON ABNORMAL U""

OCCURRENCES July September 1989 S W Wa W Rea N Abstract NUREGICR4671: THE DF-4 FUEL DAMAGE EXPERIMENT IN ACRR NUREG 0304 Vid Nod REGULATORY AND TECHNICAL REPORTS

^

HANNEL BOX.

(ABSTRACT INDEX JOURNAL). Annual Compilabon For 1989.

NU

/C US Acc6 dent NUREG/CR-4214 Rol PL HEALTH EFFECTS MODELS FOR NUCLEAR Budget POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS. Low LET NUREG 1100 V06: BUDGET ESTIMATES Fiscal Year 1991.

NAP / ELANOD2 CODE Cleevege Fracture NU L C 527 V

MANUAL, VOLUME 4 MATPRO A LIBRARY OF MATERIALS PROP.

NUREG/CR-5450- HIGH-TEMPERATURE CRACK ARREST TESTS ERilES FOR LIGHT WATER-REACTOR ACCIDENT ANALYSIS.

USING 152 MM-THICK SEN WIDE PLATES OF LOW UPPER-SHELF BASE MATERIAL: TESTS WP-2.2 AND WP 2.6.

Accident Consequence Code System NUREG/CR 4691 V02: MELUOR ACCIDENT CONSEQUENCE CODE Clooeout SYSTEM (MACCS) Volume 2 Model Desenption NUREG/CR4661: CLOSEOUT OF IE BULLETIN 85 03. MOTOR-OPER.

NUREG/CR-4691 V03: MELCOR ACCIDENT CONSEQUENCE CODE ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-SYSTEM (MACCS) Volume 3. Programmer's Reference Manuel-SIENTS DUE TO IMPROPER SWITCH SETTINGS.

NUREG/CR 5286. CLOSEOUT OF IE BULLETIN 7917. PIPE CRACKS IN NU E C 7 ASSESSMENT OF CANDIDATE ACCIDENT MAN

• NURE /CR 2. CL E T OF IE BULLE 79 2 TENTIAL AGEMENT STRATEGIES.

FAILURE OF EMERGENCY DIESEL GENERATOR FIELD EXCITER TRANSFORMER.

Accident Sequence NUREG/CR-5298. CLOSEOUT OF IE BULLETIN 8501: STEAM BIND.

l NUREG/CR-4674 V09: PRECURSORS TO POTENTIAL SEVERE CORE GE ACCIDENTS:1988 A STATUS REPORT. Main Report And Ap-N R G/C 5 CL UT F1 ETIN 8010 CONTAMINA.

NUREG/CR 4674 V10 PRECURSORS TO POTENTIAL SEVERE CORE TION OF NONRADIOACTIVE SYSTEM ANO RESULTING POTENTIAL DAMAGE ACCIDENTS:1968 A STATUS REPORT. Appendixes B And FOR UNMONITORED, UNCONTROLLED RELEASE OF RADIOACTIV.

C.

ITY TO ENVIRONMENT.

NUREG/CR-5307: CLOSEOUT OF IE BULLETIN 80-02: LNADEQUATE Ag6ng OUALITY ASSURANCE FOR NUCLEAR SUPPLIED EQUIPMENT.

NUREG/CR4731 V02. RESIDUAL LIFE ASSESSMENT OF MAJOR LIGHT WATER REACTOR COMPONENTS. OVERVIEW Computer Code NUREG/CR-5419 AGING ASSESSMENT OF INSTRUMENT AIR SYS-NUREG/CR4554 V06: SCANS (SHIPPING CASK ANALYSIS SYSTEMEA TEMS IN NUCLEAR POWER PLANTS.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN REVIEW. Volume 6 Theory Manual Bucilling Of Circu.

Aging Study lar Cyhndncal Shells NUREG/CA-5404 VOI: AUXILIARY FEEDWATER SYSTEM AGING NUREG/CR4554 V07: SCANS (SHIPPING CASK ANALYSIS SYSTEM)

STUDY, A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN REVIEW. Volume 7: Theory Manual Puncture Of Ship-Air Flow NUREGICR 5419. AGING ASSESSMENT OF INSTRUMENT AIR SYS-ping Cash TEMS IN NUCLEAR POWER PLANTS.

Concrete Containment NUREG/CR 5476: POSTTEST ANALYSIS U A 1.6-SCALE REIN-Ausl66ery Feedwater NUREG/CR-5404 VOI: AUXILIARY FEEDWATER SYSTEM AGING FORCED CONCRETE REACTOR COMAINMENT BUILDING.

Containment Auxiliary Feedwater Pump NUREG/CR-5405. ANALYSIS OF SHELL RUPTURE FAILURE DUE TO NUREG/CR 5298. CLOSEOUT OF lE BULLETIN 85-01: STEAM BIND-HYPOTHETICAL ELEVATED TEMPERATURE PRESSURIZATION OF ING OF AUXILIARY FEEDWATER P'VPS.

THE SEQUOYAH UNIT 1 STEEL CONTAINMENT BUILDING.

BWR Contamination NUREG/CR4671. THE DF 4 FUEL DAMAGE EXPERIMENT IN ACAR NUREG 1302. LEAKAGE OF AN IRRADIATOR SOURCE - THE JUNE WITH A BWR CONTROL BLADE AND CHANNEL BOX.

1988 GEORGIA RSI INCIDENT.

27

i 28 Subject index NUREGICR-5512 DRF FC: RESIDUAL MDIOACTIVE CONTAMINATON Fracture FROM DECOMMISSIONING. Technical 9ees For Translating Contami-NUREG/CP 0109-PROCEEDINGS OF THE SEMINAR ON LEAK.

nation Levels To Annual Dose Draft Repet For Comment BEFORE-BREAK.Further Developments in Regulatory Policies And Supporting Researctt

~ Contemenet6on Cloenup NUREG/CR 5482: LABORATORY ANALYSIS OF FLUID FLOW AND NUREG/CR 5381: ECONOMIC RISK OF CSNTAMINATON CLEANUP SOLUTE TRANSPORT THROUGH A VARIABLY SATURATED FRAC-COSTS RESULTING FROM LARGE NONR.MCTOR NUCLEAR MATE-TURE EMBEDDED IN POROUS TUFF.

RIAL LICENSEE OPERATONS.

Fracture Mechanece Core Damage NUREG/CR-5450- HIGH-TEMPERATURE CRACK ARREST TESTS NUREG/CR-4550 VO) Rt: ANALYSIS OF CC 9E DAMAGE FREQUEN-USING 152 MM THICK SEN WIDE PLATES OF LOW UPPER SHELF i

CY: INTERNAL EVENTS METHOOOLOGY.

BASE MATERIAL: TESTS WP-2.2 AND WP 2.6.

NUREG/CR 4674 V09: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1986 A STATUS REPORT. Man Report And AP-Fracture Toughness i

pendix A.

NUREG/CR 4744 V03 N1: LONG TERM EMBRITTLEMENT OF CAST NUREG/CR4674 V10 PRECURSORS TO POTENTIAL SEVERE CORE DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual DAMAGE ACCIDENTS:1988 A STATUS REPORT.Appendines B And Report. October 1987 March 1988.

C.

NUREG/CR-5492: INVESTIGATIONS OF IRRADIATION-ANNEAL REIR-Conce60n RADIATON (IAR) PROPERTIES TRENDS OF RPV WELDS. Phase 2 Final Report.

NUREG/CR-5435: ENVIRONMENTAL EFFECTS ON CORROSOL IN NUREG/CR 5493: INFLUENCE OF FLUENCE RATE ON RADIATION IN.

THE TUFF REPOSITORY.

DUCED MECHANICAL PHOPERTY CHANGES IN REACTOR PRES-Crock Arrest Toughnese SURE VESSEL STEELS.FnsJ Report On Exploratory Expenments.

NUREG/CR-5511: IRRADtATON EFFECTS ON STRENGTH AND NUREG/CR-5450: HIGH-TEMPERATURE CRACK ARREST 1ESTS USING 152-MM4 HICK SEN WIDE PLATES OF LOW UPPE 4 SHELF TOUGHNESS OF THREE WIRE SERIES-ARC STAINLESS STEEL BASE MATERIAL TESTS WP 2.2 AND WP 2.6, WELD OVERLAY CLADOING.

Decommlee4on6ng NUREG/CH-6491: SHIPPINOPORT STATON AGING EVALUA TON.

NUREG/CR4668. DAMAGED FUEL EXPERIMENT OF 1 Results And A"'I '"-

NUREG/CR 65'2 DRF FC RESIDUAL RADIOACTIVE CONT / MINATION Y

FROM DECOMMISSIONING. Technical Bases For Translatu g Contami-g

,g,,,,,

nation Levels To Annual Dose Draft Report For Comment NUREG/CR4704 V03. RELATIVE BIOLOGICAL EFFECTIVENESS (RBE)

Design See6e Accident OF FISSON NEUTRONS AND GAMMA RAYS AT OCCUPATONAL NUREG/CR-5368. REACTIVITY ACCIDENTS A Reassessment Of The EXPOSURE LEVELS. Studies On The Gross And Microscopic Patholo-Desegn-Basis Events.

Oy Observed At Death Of Mice Exposed To 60 Equal Once Weekly Doses Of F:ssion..

D60eet NUREG 0386 005 R05: UNITED STATES NUCLEAR REGULATORY Generic leave 70 COMMISSION STAFF PRACTICE AND PROCEDURE NUREG 1316: TECHNICAL FINDINGS AND REGULATORY ANALYSIS DIGEST. Commission. Appeal Board And Licensing Board RELATED TO GENERIC ISSUE 70. Evaluation Of Power Operated Decisions. July 1972 September 1989 Relief Valve And Block Valve Reliability in PWR Nuclear Power Plants.

Earthquake Geologic Repository NUREG/CR-5477: AN EVALUATION OF THE RELIABILITY AND USE.

NUREG/CR 525P COMPONENTS OF AN OVERALL PERFORMANCE FULNESS OF EXTERNAL INITIATOR PAA METHODOLOGIES.

ASSESSMENT METHODOLOGY.

NUREG/CR-5398: TECHNICAL BASIS FOR REVIEW OF HIGH LEVEL Economic M6ek WASTE REPOSITORY MODELING.

NUREG/CR 5381: ECONOMIC RISK OF CONTAMINATON CLEANUP COSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE-Poettt' Effect RIAL LICENSEE OPERATIONS.

NUr.EG/CR4214 R01 Pl; HEALTH EFFECTS MODELS FOR NUCLEAR POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS. Low LET Eddy Current Radiation.Part I. introduction, integration And Summary.

NUREG/CR-5478. IMPROVED EDOY CURRENT INSPECTION FOR STEAM GENERATOR TUBING. Progress Report For January 1985 To Heat Transfer December 1987.

NUREG/CR-3668: MINET CODE DOCUMENTATION.

Embrittlement H6gh-Level Weste NUREG/CR 4744 V03 NI: LONG TERM EMBRITTLEMENT OF CAST NUREG/CR-5398: TECHNICAL BASIS FOR REVIEW OF HIGH-LEVEL DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual WASTE REPOSITORY MODELING.

Report. October 1987 March 1988.

Human Error Environmental Effect NUREG/CR-5527; RISK SENSITIVITY TO HUMAN ERROR IN THE LA-NUREG/CR 5435: ENVIRONMENTAL EFFECTS ON CORROSION IN sal LE PRA, THE TUFF REPOSITORY.

Esport Judgement NUREG/CR 5288: CLOSEOUT OF IE BULLETIN 79-17. PIPE CRACKS IN NUREG/CR-5424:

EllCITING AND ANALY2ING EXPERT STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS.

JUDGEMENT.A Practical Guide.

IE Bulletin 79-23 Flecal Year NUREG/CR 5289: CLOSEOUT OF IE BULLETIN 79-23 POTENTIAL NUREG-1100 V06: SUDGET ESTIMATES Fiscal Year 1991.

FAILURE OF EMERGENCY DIESEL GENERATOR FIELD EXCITER TRANSFORMER.

NUREG/CR-5493: INFLUENCE OF FLUENCE RATE ON RADIATION IN.

IE Bulletin 80 02 DUCED MECHANICAL PROPERTY CHANGES IN REACTOR PRES-NUREG/CR 5307; CLOSEOUT OF IE BULLETIN 8002: INADEOUATE SURE VESSEL STEELS Final Report On Exploratory Expenments.

OVAUTY ASSURANCE FOR NUCLEAR SUPPLIED EQUIPMENT.

Fluid Flow IE Bulletin 8010 NUREG/C9-3668. MINET CODE DOCUMENTATION NUREG/CR-5302: CLOSEOUT OF IE BULLETIN 80-10: CONTAMINA.

NUREG/CR-5482. LABOHATORY ANALYSIS OF FLUID FLOW AND TION OF NONRADIOACTIVE SYSTEM AND RESULTING POTENTIAL SOLUTE TRANSPORT THROUGH A VARIABLY SATURATED FRAC-FOR UNMONITORED. UNCONTROLLED RELEASE OF RADIOACTIV.

TURE EMBEDDED IN POROUS TUFF.

ITY TO ENVIRONMENT.

Subject index 29 IE Bulletin 86 01 NUREG/CR-5476: POSTTEST ANALYSIS OF A 16-SCALE REIN-NUREGICR 5298 CLOSEOUT OF IE BULLETIN 8501: STEAM BIND-FORCED CONCRETE REACTOA CONTAINMENT BUILDING.

!NG OF AUXILIARY FEEDWATER dlMPS s

i Leek hate IE Bulletin 86 03 NUREG/CP 0109. PROC *EDINGS OF THE SEMINAR ON LEAK-NURt'G/CR4661: CLOSEOUT OF IE BULLETIN 85-03: MOTOR OPER-BEFORE BREAK.Furth' Developments in Regulatory Policies And ATED VALVE COMMON TiODE FAILURES DURING PLANT TRAN-Supporting Research.

SIENTS DUE TO IMPROPE i SWITCH SETTINGS.

Leek-tetore4 reek infonnehon Digeet NUREG/CP 0109: PROCEEDINGS OF THE SEMNAR ON LEAK.

NURfG 1350 V02: NUCLEAR REGUl ATORY COMMISSION INFORMA-BEFORE BREAK.Further Developments In Regulatory Pohcies And TON DOEST.1990 Edition.

Supoortng Research inservice inspect 6on NUREG/CR4469 V08. HONDESTRUCTIVE EXAMINATON (NDE) REll-L"M"8 8'"'"

NUREG-1392. LEAKAGE OF AN IRRADIATOR SOURCE. THE JUNE ABILITY FOR INSERVICF INSPECTION OF LIGHT WATER 1988 GEORGIA RSI INCIDENT, RF ACTORS Sermannual Report October 1987 - March 1988.

NUREG/CR4469 V09 NONDESIRUCTIVE EXAMINATION (NDE) REll-A V30101: INDEXES TO NUCLEAR REGULATORY COM-AC OR mi t mber 988 NUREG/CR 4731 V02 RE AL U ASSESSMENT OF MAJOR MISSION ISSUANCES. July-September 1989.

NUREG 0750 V30 NOI: NUCLEAR REGULATORY COMMISSON IS-LIGHT WATER REACTOR COMPONENTS - OVERVIEW.

NUREG/CR4882: OUAL IFICATION PROCESS FOR ULTRASONIC SUANCES FOR JULY 1989 Pages 184.

TESTING IN NUCLEAR INSERVICE INSPECTION APPLICATIONS.

NUREG 0750 V30 NO2: NUCLEAR REGULATORY COMMISSON IS-SUANCES FOR AUGUST 1969. Pages 85165.

. Inspection NUREG 0750 V30 NO3: NUCLEAR REGULATORY COMM!SSON IS-NUREG 0040 V13 N04: LICENSEE CONTRACTOR AND VENDOR IN-SUANCES FOR SEPTEMBER 1989. Pages 167 229.

SPECT;ON ST ATUS REPORT. Quartarly Report. October December NUREG-0750 V30 N04: NUCLEAR REGULATORY COMMISSION IS-1989.(White Book)

SUANCES FOR OCTOBER 1989 Pagos 201323.

NUREG-0750 V30 N05: NUCLEAR REGULATORY COMMISSION IS-Instrument Air System SUANCES FOR NOVEMBER 1989. Pages 325-708.

NUREG/CR-5419. AGING ASSESSMENT OF INSTRUMENT AIR SYS-TEMS IN NUCLEAR POWER Pl. ANTS.

Licensed Operating Reactors NUREG/CR-5472: A RISK BASED REVIEW OF INSTRUMENT AIR SYS-NUREG-0020 V13 N11: UOENSED OPERATING REACTORS STATUS TEMS AT NUCLEAR POWER PLANTS.

SUMMARY

REPORT. Data As Of October 30,1989-(Gray Book I)

NUREG-0020 V13 N12; LICENSED OPERATING REACTORS STATUS Mum M

SUMMARY

REPORT. Data As Of November 30.1989 (Gray Book I)

NUREG/CR-5484; PH SENSORS BASED ON IRIDIUM OXIDE.

NUREG-0020 Vid N01: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of December 31,1989.(Gray Book 1)

Irrediation NUREG/CR-5492: INVESTIGATONS OF IRRADIATION-ANNEAL REIR-ATON (IAR) PROPERTIES TRENDS OF RPV WELDS. Phase 2 N

CR 08N12; LICENSEE EVENT REPORT (LER)

NUREG/C 5511: IRRADIATON EFFECTS ON STRENGTH AND COMPILATON.For Month Of December 1989.

NUREG/CR 2000 V09 N1: LICENSEE EVENT REPORT (LER)

TOUGHNESS OF THREE WIRE SERIES-ARC STAINLESS STEEL COMPILATION For Month Of January 1990.

WELD OVERLAY CLADDING' NUREG/C42000 V09 N2: UCENSEE EVENT REPORT (LER)

Irred6ator COMPILATION For Month Of February 1990.

NUREG 1392: LEAKAGE OF AN IRRADIATOR SOURCE - THE JUNE 1988 GEORGIA RSI INCIDENT.

Light Water Reactor NUREG/CR4469 V08: NONDESTRUCTIVE EXAMINATION (NDE) REll-LAPUR User's Guide ABILITY FOR INSERVICE INSPECTION OF UGHT WATER NUREG/CR 5421: LAPUR USER'S GUIDE.

REACTORS Semiannual Report,0ctober 1987 March 1988.

NUREG/CR4469 V09: NONDESTRUCTIVE EXAMINATON (NDE) REll-LER ABluTY FOR INSERVICE INSPECTION OF UGHT WATER NUREG/CR 2000 VOBN12: UCENSEE EVENT REPORT (LER)

REACTORS. Semiannual Report,Apni-September 1988.

COMPILATON For Month Of December 1989.

NUREG/CR-4674 V09: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-2000 V09 N1: UCENSEE EVENT REPORT (LER)

DAMAGE ACCIDENTS:1968 A STATUS REPORT. Main Report And Ap-COMPILATION For Month Of January 1990.

rex A NUREG/CR 2000 V09 N2; UCENSEE EVENT REPORT (LER)

NU EG/CR4674 VIO: PRECURSORS TO POTENTIAL SEVERE CORE COMPILATION For Month 0f February 1990.

DAMAGE ACCIDENTS:1988 A STATUS REPORT.Appardxes B And C.

LWR NUREG/CR-4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR NUREG/CR-4469 V08 NONDESTRUCTIVE EXAMINATION (NDE) RELl, LIGHT WATER REACTOR COMPONENTS OVERVIEW.

ABlWTY FOR INSERVICE INSPECTION OF UGHT WATER NUREG/CR-4744 V03 N1: LONG-TERM EMBRITTLEMENT OF CAST REACTORS Semiannual Report. October 1987 March 1988 NUREG/CR-4469 V09' NONDESTRUCTIVE EXAMINATION (NDE) REU.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual ABluTY FOR INSERVICE INSPECTION OF UGHT WATEH Report, October 1987 March 1988.

NUREG/CR-5273 V04:

SCDAP/RELAPS/ MOD 2 CODE RE ACTORS Semiannual Repod Apnf September 1988.

NUREG/CR 4674 V09 PRECURSOHS TO POTENTIAL SEVERE CORE MANUAL. VOLUME 4 MATPRO - A UBRARY OF MATERIALS PROP-DAMAGE ACCIDENTS 1968 A STATUS REPORT.Mam Heport And Ap.

ERTIES FOR UGHT WATER-REACTOR ACCIDENT ANALYSIS.

NUREG/CR-5405: ANALYSIS OF SHELL RUPTUPE FAILURE DUE TO perdx A.

NUREG/CR 4674 VIO: PRECURSORS TO POTENTIAL SEVERE CORE HYPOTHETICAL ELEVATED-TEMPERATURE PRESSUR12ATON OF DAMAGE ACCIDENTS.1988 A STATUS REPORT.Appendinei B And THE SEQUOYAH UNIT I STEEL CONTAINMENT BUILDING.

C.

NUREG/CR-5478: POSTTEST ANALYSIS OF A 16 SCALE REIN-NUREG/CR 4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR FORCED CONCRETE REACTOR CONTAINMENT BUILDING.

LIGHT WATER REACTOR COMPONENTS OVERVIEW NUREG/CR-4744 V03 N): LONG TERM EMBRITTLEMENT OF CAST Loes Of Coolant DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUREG/CR-5474. ASSESSMENT OF CANDtDATE ACCIDENT MAN-Report. October 1987 March 1988.

AGEMENT STRATEGIES NUREG/CR-5273 V04:

SCDAP/RELAP5/ MOD 2 CODE MANUAL, VOLUME 4 MATPRO A LIBRARY OF MATERIALS PROP.

Lose Of Coolant Accident ERTIES FOR UGHT WATER-REACTOR ACCIDENT ANALYSIS NUREG/CR-5395 V02: MULTILOOP INTEGRAL SYSTEM TEST (MIST)

NUREG/CR-5405: ANALYSIS OF SHELL RUPTURE FAILURE DUE TO FINAL REPORT. Test Group 30.Mappmg Tests.

HYPOTHETICAL ELEVATED TEMPERATURE PRESSURIZATION OF NUREG/CR 5395 V10: MULTILOOP INTEGRAL SYSTEM TEST (MIST)-

THE SEQUOYAH UNIT 1 STEEL CONTAINMENT BUILDING FINAL REPORT. RELAPS/ MOD 2 MIST Analysis Compensons

30 Subject index Low-Level Weste Data Base Operational Event NUREG/CR 5229 V02: TM1-2 EPICOR-ll RESIN / LINER INVESTIGATION.

NUREG/CR-4674 V09 PRECURSORS TO POTENTIAL SEVERE CORE LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR DAMAGE ACCIDENTS 1988 A STATUS REPORT. Main Report And Ap-FISCAL YEAR 1989. Annual Report ponds A.

NUREG/CR 4674 V10- PRECURSORS TO POTENTW L SEVERE CORE MACC8 DAMAGE ACCIDENTS:1968 A STATUS REPORT. Appendixes B And NUREG/CR-4691 V02: MELCOR ACCIDENT CONSEQUENCE CODE C,

SYSTEM (MACCS) Volume 2. Model Desenption.

NUREG/CR4691 V03. MELCOR ACCIDENT CONSEQUENCE CODE PRA SYSTEM (MACCS). Volume 3: Programmer's Reference Manual NUREG/CR-4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-CY: INTERNAL EVENTS METHODOLOGY, MACC8 Code NUREG/CR 5477: AN EVALUATON OF THE RELIABILITY AND USE.

NUREG/CR-5376: OUALITY ASSURANCE AND VERifvCATION OF THE FULNESS OF EXTERNAL INITIATOR PRA METHODOLOGIES.

MACCS CODE, Version 1.5-NUREG/CR-5527: RISK SENSITIVITY TO HUMAN ERROR IN THE LA-SALLE PRA.

NUREG/CR-4601 V02: MELCOR ACCOENT CONSEQUENCE CODE PWR SYSTEM (MACCS). Volume 2: Model Desenpton-NUREG 1316 TECHNICAL FINDINGS AND REGULATORY ANALYSIS NUREG/CR 4691 V03: MELCOR ACCIDENT CONSEQUENCE CODE RELATED TO GENERIC ISSUE 70.Evaluston Of Power Operated SYSTEM (MACCS). Volume 3. Programmer's Reference Manual-Rehef Valve And Block Vatve Reliability in PWR Nuclear Power Plants.

NUREG/CR-5298: CLOSEOUT OF IE BULLET N 8501: STEAM BIND-MELPROG Code ING OF AUXILIARY FEEDWATER PUMPS.

NUREG/CR-5316. MELT PROGRESSION. OXIDATION. AND NATURAL NUREG/lA 0012: RELAPS/ MOD 2 CALCULATIONS OF OECD. LOFT CONVECTION IN A SEVERELY DAMAGED R%CTOR CORE.

TEST LP-SB 01.

MINET Code Performance Assosoment NUREG/m3668 MINET CODE DOCUMETATON.

NUREGICR-5256: COMPONENTS OF AN OVERALL PERFORMANCE MIST ASSESSMENT METHODOLOGY.

NUREG/CR-5305 V02: MULTILOOP INTEGRAL SYSTEM TEST (MIST):

Performance HIMory FINAL REPORT. Tost Group 30, Mapping Tests.

NUREG 1214 ROS: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

Mett Progression IC ASSESSMENT OF LICENSEE PERFORMANCE.

NUR CR 4668; DAMAGED FUEL EXPERIMENT DF 1.Results And Pene For RuMng l

NUREG/CR4671: THE DF-4 FUEL DAMAGE EXPERIMENT IN ACAR NUREG-0936 V04 N04 NRC REGULATORY AGENDA.Ouarterty l

WITH A BWR CONTROL BLADE AND CHANNEL BOX.

ReportOcW Oecember 1981 l

NUREG/CR-5316: MELT PROGRESSION.OXIDATON, AND NATURAL CONVECTION IN A SEVERELY DAMAGED REACTOR CORE' N A G/CR 66t: CLOSEOUT OF IE BULLETIN 85-03: MOTOR OPER.

Motor Operated Valve ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-l NUREG/CR 4661: CLOSEOUT OF IE BULLETIN 85 03: MOTOR OPER.

SIENTS DUE TO IMPROPER SWITCH SETTINGS.

l ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-SIENTS DUE TO IMPROPER SWITCH FETTINGS.

Porous Tuf Multiloop integral System Test SOLUTE TRANSPORT THROUGH A VARIAB1,Y SATURATED FRAC.

NUREG/CR-5395 V02: MULTILOOP INTEGRAL SYSTEM TEST (MIST).

TURE EMBEDDED IN POROUS TUFF.

FINAL REPORT. Test Group 30, Mapping Tests.

i I

NUREG/CR-5395 V10 MULTILOOP INTEGRAL SYSTEM TEST (MIST):

Practice And Procedure D6 gest FINAL REPORT. RELAPS/ MOD 2 MIST Analysis compansons-NUREG-0386 005 R05: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE NSSS OlGEST.Commisson, Appeal Board And Licensing Board NUREG/CR-4659 V03: SEISMIC FRAGILITY OF NUCLEAR POWER Decisons. July 1972 September 1989.

PLANT COMPONENTS (PHASE II).Switchgear, IAC Panels (NSSS)

And Relays Pressure Vessel NUREG/CR-5450: HIGH-TEMPERATURE CRACK ARREST TESTS Neutron Source USING 152 MM-THICK SEN WIDE PLATES OF LOW UPPER SHELF NUREG/CR-4704 V03: RELATIVE BIOLOGICAL EFFECTIVENESS (RBE)

BASE MATERIAL: TESTS WP-2.2 AND WP-2.6.

OF FISSON NEUTRONS AND GAMMA RAYS AT OCCUPATIONAL EXPOSURE LEVELS Studies On The Gross And Microscopic Pathoto.

Pressurtzed Water Reactor gy Observed At Death Of Mice Exposed To 60 Equal Once Weekly NUREG 1316: TECHNICAL FINDINGS AND REGULATORY ANALYSIS Ooses Of Fission..,,

RELATED TO GENERIC ISSUE 70.Evaluaton Of Power Operated Rehef Valve And Block Valve Reliability in PWR Nuclear Power Plants.

Nondestructive Examination NUREG/CR.4668: DAMAGED FUEL EXPERIMENT OF 1.Results And NUREG/CR 4469 VOS: NONDESTRUCTIVE EXAMINATON (NDE) RELI-Anatyses.

ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER NUREG/CR-5298: CLOSEOUT OF IE BULLETIN 85 01: STEAM BIND-REACTORS. Semiannual Report. October 1987 March 1988.

ING OF AUXILIARY FEEDWATER PUMPS.

NUREG/CR 4469 V09. NONDESTRUCTIVE EXAMINATION (NDE) RELI-NUREG/lA 0012: RELAP5/ MOO 2 CALCULATIONS OF OECDLOFT ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER TEST LP.SB-01.

REACTOPS. Semiannual Report. April-Septeniber 1988.

Probabilistic R6ek Assessment Nuclear Plant Aging NUREG/CR-4550 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR 5491: SHIPPINGPORT STATION AGING EVALUATION.

CY: INTERNAL EVENTS METHODOLOGY.

Nuclear Plant Aging Research Program Ouality Assurance NUREG/CR-5386: BASIS FOR SNUBBER AGING RESEARCH: NUCLE-NUREG/CR-5376 QUALITY ASSURANCE AND VERIFICATION OF THE AA PLANT AGING RESEARCH PROGRAM.

MACCS CODE, Version 1.5.

Occupational Dose Control RELAP5/ MOD 2 NUREG/CP-0110 PROCEEDINGS OF THE INTERNATIONAL WORK-NUREG/lA 0012: RELAP5/ MOD 2 CALCULATIONS OF OECD LOFT SHOP ON NEW DEVELOPMENTS IN OCCUPATIONAL DOSE CON-TEST LP SB-01.

TROL AND ALARA lMPLEMENTATION AT NUCLEAR POWER NUREG/lA 0013. RELAP5/ MOD 2 CALCULATONS OF OECD-LOG PLANTS AND SIMILAR FACILITIES TEST LP-SB-03

Subject index 31 Radiation =

Regulatory And Techn6 cal Report

- NUREG/CR 4214 R01 PL HEALTH EFFECTS MODELS FOR NUCLEAR NUREG-0304 V14 N04. REGULATORY AND TECHNICAL REeDHTS POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS. Low LET (ABSTRACT INDEX JOURNAL). Annual Compilation For 1989.

Ra$ahortPart 1:Introdction, integrabon And Summary.

Regulatory impact Nedention Protect 6on NUREG 1395 DRFT: INDUSTRY PERCEPTONS OF THE IMPACT OF NUREG'CP 0110 PROCEEDINGS OF THE INTERNATONAL WORK-THE U.S. NUCLEAR REGULATORY COMMISSION ON NUCLEAR SHOP ON NEW DEVELOPMENTS IN OCCUPATONAL DOSE CON' POWER PLANT ACTIVITIES. Draft Report TROL AND ALARA IMPLEMENTATION AT NUCLEAR POWER PLANTS AND E!MILAR FACILITIES.

Relay NUREG/CR-4659 V03: SEISMIC FRAGILITY OF NUCLEAR POWER Rad 6oactive Weste.

PLANT COMPONENTS (PHASE tl).Switchgear, l&C Penets (NSSS)

NUREGICR-5381: ECONOMIC RISK OF CONTAMINATION CLEANUP COSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE, And Relays RIAL LICENSEE OPERATIONS.

pg ygy, Radioactivity NUREG 1316. TECHNICAL FINDINGS AND REGULATORY ANALYSIS NUREG/CR5302: CLOSEOUT OF IE BUL.LETIN 8010: CONTAMINA.

RELATED TO GENERIC ISSUE 70 Evaluation Of Power Operated TION OF NONRADIOACTIVE SYSTEM AND RESULTING POTENTIAL Rehef Vatve And Block Valve Reliatxhty in PWR Nuclear Power Plants.

FOR UNMONITORED. UNCONTROLLED RELEASE OF RADIOACTIV.

(TY TO ENVIRONMENT.

Report To Congress NUREG 0090 V12 NO3 REPORT TO CONGRE iS ON ABNORMAL RadioD4olo0Y OCCURRENCES July September 1989 NUREG/CR-4704 V03: RELATIVE BIOLOGICAL EFFECTIVENESS (RBE) j

. OF FIS$10N NEUTRONS AND GAMEA RAYS AT OCCUPATIONAL Resideel Red 6oactive EXPOSURE LEVELS. Studies On 7ne Gevas And Microscopic Pathoto-NUREG/CR5512 DAF FC RESIDUAL RADCACTIVE CONTAMINATON gy Observed Al Deatt. Of Mce Exposed To 60 Equal Once Weekly FROM DECOMMISSIGNING Techncal Basis For Translating Contami-Dose Of Fission.. -

nation Lesets To Annual Dose. Draft Report For Comment f

Reectivity Accident Risk Assessment NUREG/CR-5368 REACTIVITY ACCIDENTS.A Resssessment Of The NUREG/CP 0105 V01: PROCEEDINGS OF THE SEVENTEENTH Design Basis Events WATER REACTOR SAFETY INFORMATION MEETING.

NUREG/CP 0105 V02: PROCEEDINGS OF THE SEVENTEENTH Reactor Accident NUREG/CP 0105 V01: PROCEED:NGS OF THE SEVENTEENTH WATER REACTOR SAFETY INFORMATON MEETING.

WATER REACTOR SAFETY INFORMATON MEETING NUREG/CP 0105 V03: PROCEEDINGS OF THE SEVENTEENTH NUREG/CP 0105 V02: PROCEEDINGS OF THE SEVENTEENTH WATER REACTOR SAFETY INFORMATION MEETING WATER REACTOR SAFETY INFORMATON MEETING NUREG/CP 0105 V03: PROCEEDINGS OF THE SEVENTEENTH Rules

)

WATER REACTOR SAFETY INFORMATON MEETING NUREG-0936 V08 N04: NRC REGULATORY AGENDA.Ouarterty NUREG/CR-4691 V02: MELCOR ACCIDENT CONSEQUENCE CODE Report. October December 1989.

SYSTEM (MACCS)03. MELCOR ACCIDENT CONSEQUENCE CODE Volume 2 Model Descnotion.

q NUREG/CR4601 V Ru6es Of Practice i

SYSTEM (MACCS) Volume 3 Programmer's Reference Manual.

NUREG-0386 DOS ROS: UNITED STATES NUCLEAR REGULATORY j

NUREG/CR-5474. ASSESSMENT OF CANDIDATE ACCIDENT MAN-COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST.Commiss.on, Appeal Board And Licensing Board NUREG/ 5 S SENSITIVITY TO Hl' MAN ERROR IN THE LA.

4 Decisions. July 1972. September 1989.

SALLE PRA.

SCANS

]

Reactor Componente NUREG/CR-5419: AGING ASSESSMENT OF INSTRUMENT AIR SYS, NUREG/CR 4554 V06: SCANS (SHIPPING CASK ANA'.YSIS SYSTEM):A TEMS IN NUCLEAR POWER PLANTS.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN REVIEW. Volume 6. Theory Manual Buckhng Of Circu.

Reactor Containment lar Cytindncal Shells.

NUREG/CRS476: POSTTEST ANALYSIS OF A 1:6 SCALE REIN-NUREG/CR 4554 V07: SCANS (SHIPPING CASK ANALYSIS SYSTEM)

FORCED CONCRETE REACTOR CONTAINMENT BUILDING.

A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN FiEVIEW. Volume 7: Theory Manual Puncture Of Ship-R e rcom ping Cads.

NUREG/CR5316: MELT PROGRESSION. OXIDATION. AND NATURAL CONVECTION IN A SEVERELY DAMAGED REACTOR CORE.

SCDAP/RELAP5/MGD2 NUREG/CR-5273 V04-SCDAP/RELAPS/ MOD 2 CODE Reactor Pressure Vessel MANUAL. VOLUME 4.MATPRO. A LIBRARY OF MATERIALS PROP.

NUREG/CR5493; INFLUENCE OF FLUENCE RATE ON RADIATION-IN, ERTIES FOR LIGHT WATER-REACTOR ACCIDENT ANALYSIS.

DUCEO MECHANICAL PROPERTY CHANGES IN REACTOR PRES.

E b ATI OF Safety EvaJustion Report NU E / 55 L1 I i U TROJAN ACL SUBJECT TO POSTULATED RPV SUPPORT FAILURE.

NUREG-0797 S22. SAFETY EVALUATION REPORT RELATED TO THE CPERATION OF COMANCHE PEAK STEAM ELECTRIC Reactor Protection System STATION. UNITS 1 AND 2. Docket Nos. 50-445 And 50-446.(Texas Utili-A NUREG/CR5474. ASSESSMENT OF CANDIDATE ACCIDENT MAN-ties Electnc Company.et all

)

AGEMENT STRATEGIES.

NUREG 0797 S23: SAFETY EVALUATION REPORT RELATED TO THE l

OPERATION OF COMANCHE PEAK STEAM ELECTRIC ties Electnc Coms.'any.et al.)

l REG R 5368. REACTIVITY ACCIDENTS.A Reassessment Of The NUREG 0896 SO9-SAFETY EVALUATION REPORT RELATED TO THE l

NU E C 72 A ISK BASED REVIEW OF INSTRUMENT AIR SYS-OPERATION OF SEABROOK STATION. UNITS 1 AND 2. Docket Nos.

TEMS AT NUCLEAR POWER PLANTS' STN 54443 And STN-50-444 (Public Service Company Of New Hamp-Reactor Vessel shirel NUREGIC45479 CURRENT APPLICATIONS OF VIBRATION MONI.

NUREG 1232 V04. SAFETY EVALUATION REPORT ON TENNESSEE TORING AND NEUTRON NOISE ANALYSIS. Detection And Analysis Of VALLEY AUTHORITY: WATTS BAR NUCLEAR PERFORMANCE Structural Degradation Of Reactor Vessel Internals From Operational PLAN.

ng Safety Research Program Regulatory Agenda NUREG/CR2331 V09 N3. SAFETY RESEARCH PROGRAMS SPON.

NUREG-0936 V08 N04-NRC REGULAT ORY AGENDA.Ouarterty SORED BY OFFICE OF NUCLEAR REGULATORY

]

Report. October-December 1989 RESEARCH Progress Report. July September 1989.

j I

n n-

32 Subject index Setemic Fre0616ty Systemet6c Assessment Of Licensee Performance NUREG/CR 4659 V03: SEISMIC FRAGILITY OF NUCLEAR POWER NUREG-1214 RU5: HISTORICAL DAT A

SUMMARY

OF THE SYSTEMAT-PLANT COMPONENTS (PHASE II)Switchgear, l&C Panels (NSSS)

IC ASSESSMENT OF LICENSEE PERFORMANCE.

And Relays-TLD Setem6cny NUREG 0837 V09 N04: NRC TLD DIRECT RADIATON MONITORING NUREG/CR-3145 V08. GEOPHYSICAL INVESTIGATIONS OF THE NETWORK. Progress Report October December 1989.

WESTERN OHIOINDIANA REGON Annual Report,0ctober 1988 -

September 1989-TMb2 EPICOR Il Senor NUREG/CR-5229 V02: TMI-2 EPICOR-il RESIN / LINER INVESTIGATION.

NUREG/CR-5484. PH SENSORS BASED ON IRIDIUM OXIDE.

LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FISCAL YEAR 1989 Annual Report Severe Fuel Damage NUREGtCR-4671: THE OF 4 FUEL DAMAGE EXPERIMENT IN ACRR Toerine Resistence WITH A BWR CONTROL BLADE AND CHANNEL BOX-NUREG/CR 5511: IRRADIATON EFFECTS ON STRENGTH AND TOUGHNESS OF THREE WIRE SERIES-ARC STAINLESS STEEL Shell-Rupture Failure WELD OVERLAY CLADDING.

NUREG/CR-5405: ANALYSIS OF SHELL RUPTURE FAILURE DUE TO HYPOTHETICAL ELEVATEDTEMPERATURE PRESSURi2ATON OF Technical Specificetton THE SEOUOYAH UNIT 1 STEEL CONT AINMENT BUILDING.

NUREG-1381: TECHNICAL $PECIFICATIONS, COMANCHE FEAK STEAM ELECTRIC STATON UNIT 1. Docket No. 50-445,Appendu "A" C**"

To License No. NPF 28 NUREG/CR-4554 V06: SCANS (SHIPPING CASK ANALYSIS SYSTEM) A NUREG-1386: TECHNICAL SPECIFICATONS FOR SEABROOK STA-MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING TION, UNIT 1. Aopend x "A" To L6 cense No. NPF 86.

CASK DESIGN REVIEW. Volume 6. Theory Manual Buckhng Of Circu.

lar Cylindncal Sheits.

Tectonico NUREGICR-4554 VJ7: SCANS (SHIPPING CASK ANALYSIS SYSTEM)

NUREG/CR 3145 V06. GEOPHYSICAL INVESTIGATIONS OF THE A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIFPING CASK DESIGN REVIEW. Volume 7: Theory Manual Puncture Of Ship-WESTERN OHIO INDIANA REGON. Annual Report, October 1988 Septh 1989.

ping Casks.

g g,,g Thermoluminacent Doe 4 meter NUREG/lA 0012 RELAP5/ MOD 2 CALCULATONS OF OECD LOFT NUREG-0837 V09 N04: NRC TLD DIRECT RADIATION MONITORING TEST LP-SB-01.

NETWORK. Progress Report October December 1989.

NUREG/lA 0013: RELAP5/ MOD 2 CALCULATONS OF OECD LOFT TEST LP-SB43.

Title List NUREG 0540 V11 N10 TITLE LIST OF DOCUMENTS MADE PUDLICLY Snubner AVAll ABLE. October 1 31,1989.

NUhEG/CR-5386: BASIS FOR SNUDOER AGING RESEARCH; NUCLE-NUREGC40 V11 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY AR PLANT AGING RESEARCH PROGRAM.

AVAILADLE. NOVEMBER 1 30,1989.

Solute Transport Transport NUREG/CR-5482: LADORATORY ANALYSIS OF FLUID FLOW AND NUREG/CR-5169: MOBILIZATION AND TRANSPORT OF URANtUM AT SOLUTE TRANSPORT THROUGH A VARIADLY SATURATED FRAC-URANIUM MILL TAILINGS DISPOSAL SITES Apphcation Of A ChemL TURE EMBEDDED IN POROUS TUFF.

cal Transport Model Statniese Steel Transportation Accident NUREG/CR-4469 V09: NONDESTRUCTIVE EXAMINATON (NDE) RELl-NUREG/CR-5477: AN EVALUATION OF THE REllADILITY AND USE.

ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER FULNESS OF EXTERNAL INUIATOR PRA METHODOLOGIES.

REACTORS Semiannual ReportApnbSeptember 1988.

NUREG/CR 4744 V03 N1: LONG TERM EMBRITTLEMENT OF CAST D

X S INLE STE IN LWR SYSTEMS. Semiannual N E 8: IMPROVED EDDY CURRENT INSPECTION FOR STEAM GENERATOR TUDING Progress Report For January 1985 To Sta6neese Steel Wold December 1987.

NUREG/CR-5511: IRRADIATION EFFECTS ON STRENGTH AND TOUGHNESS OF THREE WtRE SERIES ARC STAINLESS STEEL Tuff Repository WELD OVERLAY CLADDING.

NUREG/CR-5435: ENVIRONMENTAL EFFECTS ON CORROSION IN THE TUFF REPOSITORY.

Station Stack Out NUREG/CR-5395 V02: MULTILOOP INTEGRAL SYSTEM TEST (MIST):

Ultrasonic Testing FINAL REPORT. Test Group 30. Mapping Tests.

NUREG/CR-4882: QUALIFICATION PROCESS FOR ULTRASONIC NUREG/CR-5395 V10 MULTILOOP INTEGRAL SYSTEM TEST (MIST).

TESTING IN NUCLEAR INSERVICE INSPECTON APPLICATIONS.

FINAL REPORT. RELAP5/ MOD 2 MIST Analysis Compansons.

Uren8um Steam Generator NUREG/CR-5169. MOBILIZATICN AND TRANSPORT OF URANtUM AT NUREG/CR-5478-IMPROVED EDDY-CURRENT INSPECTON FOR URANIUM MILL TAILINGS DISPOSAL SITES Apphcation Of A Chemi-STEAM GENERATOR TUBING. Progress Report For January 1985 To cal Transport Modet.

December 1987.

Structural Degradation Uranium M611 Tailing

- NUREG/CR 5479. CURRENT APPLICATONS OF VIBRATON MONI-NUREG/CR-5169: MOBill2ATION AND TRANSPORT OF URANIUM AT TORING AND NEUTRON NOISE ANALYSIS. Detection And A ialyms Of URANIUM MILL TAILINGS DISPOSAL SITES.Apphcatx>n Of A Choms-Structural Degradation Of Reactor Vessel internals From 06 erational cal Transpod mom

1. '"9 9

Ventilation System Support Feiture NUREG/CR 5472: A RISK-BASED REVIEW OF INSTRUMENT AIR SYS-NUREG/CR-5508. PRELIMINARY GTRUCTURAL EVALUATION OF TEMS AT NUCLEAR POWER PLANTS TROJAN RCL SUBJECT TO POSTULATED RPV SUPPORT F ALLURE.

Switchgear NUREG/CR-5470. CURRENT APPLICATIONS OF VIBRATION MONI-NUREG/CR-4659 V03: SEISMIC FRAGILITY OF NUCLEAP POWER TORING AND NEUTRON NOISE ANALYSIS. Detection And Andysis Of PLANT COMPONENTS (PHASE II)S*:tchgear. I&C Penna (NSSS)

Structural Degradation Ot Reactor Vessel internals From Operational And Relays.

/Lging

__o s.

A

+.

-i Subject inden 33 Weste Form Testing NUREG/CP 0106 V02: PROCEEDINGS OF THE SEVENTEENTH-NUREG/CR-5229 V02: TML 2 EPICOR il RESIN / LINER INVESTIGATION.

WATER REACTOR SAFETY INFORMATION MEETING NUREG/CP-0105 V03: PROCEEDINGS - OF THE SEVENTEENTH LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR WATER REACTOR SAFETY INFORMATION MEETING.

FISCAL YEAR 1989. Annual bport Wetal

' Water Reector teWy -

NUREG/CR 5492: WVESTIGATIONS OF IRRADIATION ANNEAL REIR.

NUREG/CP.0106 V01: PROCEEDINGS OF THE SEVENTEENTH RADIATION (IAR) PROPERTIES TRENDS OF RPV WELDS Phase 2 WATER REACTOR SAFETY INFORMATION MEETING.

Final Repart.

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NRC Originating Organization Index (Staff Reports)

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

OFF6CE OF EXECUTIVE D4 RECTOR FOR OPERATIONS (EDO)

NUREG 0020 V13 N12: LICENSED OPERATING REACTORS STATUS REGtON 1, OFC OF THE DIRECTOR

SUMMARY

RF. PORT. Data As Of November 30,1989.(Gray Book l)

NUREG-0837 V09 N04: NRC TLD DIRECT RADIATON MONITORING NUREG-0020 V14 N01: LICENSED OPERATING REACTORS STATUS Pr e

c ober December 1989-

SUMMARY

REPORT. Data As Of December 31,1989.(Grey Book 1) gg%]O 0

E Di NUREG 1395 DRFT: INDUSTRY PERCEPTONS OF THE IMPACT OF U.S. NUCLEAR REOULATORY COMMISSION THE U.S. NUCLEAR REGULATORY COMMISSION ON NUCLEAR OFFICE OF THE GENERAL COUNSEL (POST B00701)

POWER PLANT ACTIVITIES. Draft Report NUREG-0388 005 ROS: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE EDO. OFFICE OF ADMONISTRATION (PRE 870413 & POST 880205)

DIVISION OF FREEDOM OF INFORMATON & PUBUCATIONS SERV.

OlGEST.Commassen, Appeal Board And Licensing Bord ICES (POST 890205 Decisions. July 1972 September 1989.

NUREG-0304 V14 N04: REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL). Annual Compilation For 1989.

EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

NUREG-0540 V11 NIO: TITLE LIST OF DOCUMENTS MADE PUBLIC.

DIVISON OF ENGINEERWG (POST 870413)

LY AVAILABLE. October 1 31,1989.

NUREG/CR-4882: QUALIFICATON PROCESS FOR ULTRASONIU

~

NUREG 0540 V11 N11: TITLE LIST OF DOCUMENTS MADE PUBLIC-TESTING IN NUCLEAR INSERVICE INSPECTION APPLICATIONS.

LY AVAILABLE. NOVEMBER 130,1989.

DIVISION OF SAFETY ISSUE RESOLUTION (POST 880717)

NUREG-0750 C102: INDEXES TO NUCLEAR REGULATORY COMMIS.

NUREG-1318: TECHNICAL FINDINGS AND REGULATORY ANALYSIS SiON ISSUANCES. January 1.1980 Through December 31,1985.

RELATED TO GENERIC ISSUE 70.Evaluahon Of Power Operated NUREG-0750 V30101: INDEXES TO NUCLEAR REGULATORY COM.

Rehef Valve And Block Valve Rellatxhty in PWR Nuclear Power MISSION ISSUANCES. July September 1989.

Plants.

NUREG-0750 V30 N01: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR JULY 1989 Pages 184.

EDO. OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

NUREG 0750 V30 NO2: NUCLEAR REGULATORY COMMISSION IS' DIVISION OF REACTOR PROJECTS. l/Il (POST 870411)

SUANCES FOR AUGUST 1989. Pages85-165.

NUREG 0898 SO9: SAFETY EVALUATON REPORT RELATED TO NUREG 0750 V39 NO3: NUCLEAR NEGULATORY COMMISSON IS-THE OPERATON OF SEABROOK STATON, UNITS 1 AND

2. Docket Nos. STN-50 443 And STN-50 444.(Pubhc Sennce Compa-N RE 30 N N LEA RE UL R COMMISSION IS-SUANCES FOR OCTOBER 1989 Pa es 231323

y Of New u apshire)

NUREG 0750 V30 N05: NUCLEAR RfGULATORY COMMISSION IS-NUREG-138n TECHNICAL SPECIFICATIONS FOR SEABROOK STA-SUANCES FCR NOVEMBEH 1989 Pages 325 708.

TON, UNIT 1. Appendix "A" To License No. NPF 88.

NUREG-0938 V08 N04. NRC REGULATORY AGENDAOuarterly COMANCHE PEAK PROJECT DIVISION Report, October December 1989.

NUREG 0797 S22: SAFETY EVALUATION REPORT RELATED TO 1r4E OPERATION OF COMANCHE PEAK STEAM ELECTRIC EDO. OFFICE OF THE CONTROLLER (PRE 820418 & POST 880205)

STATON. UNITS 1 AND 2. Docket Nos. 50 445 And 50 446(Texas OlVISION OF BUDGET & ANALYSIS (POST 890205) ut,hties Electric Company.et al.)

m 21 SAW MMm MM MWG M l

R 2: NUCLEA EGU T ON INFOR, THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC

(

MATON DIGEST.1990 Editen.

STATON, UNITS 1 AND 2. Docket Nos. 50-445 And 50 448.(Texas EDO. OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL Utihties Electnc Company.et al.)

DATA NUREG-1381: TECHNICAL SPECIFICATIONS, COMANCHE PEAK OFFICE FOR ANALYSIS & EVALUATON OF OPERATIONAL DATA, D4-STEAM ELECTRIC STATION, UNIT 1. Docket No. 50-445, Appendix RECTOR "A" To Ucense No. NPF 28.

NUREG-0090 V12 NO3: REPORT TO CONGRESS ON ABNORMAL TVA PROJECTS DIVISION OCCURRENCES. July September 1989.

NUREG-1232 V04. SAFETY EVALUATON REPORT ON TENNESSEE OFFICE OF GOVERNMENTAL & PUBLIC AFFAIRS POST 870413)

N DIVISION ' OF REACTOR INSPECTON & SAFEGUARDS (POST N E 3 : LEA G N

D AT SOU T

UNE 1988 GEORGIA RSI INCIDENT, NU G 40 V13 N04: LICENSEE CONTRACTOR AND VENDOR IN-EDO. OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM SPECTION STATUS REPORT. Quarterty Report,0ctober December (870413-890204) 1989.(White Book)

DIVISION OF COMPUTER & TELECOMMUNICATIONS SERVICES DIVISION OF LICENSEE PERFORMANCE & QUALITY EVALUATION (POST 890205)

(POST 870411)

NUREG-0020 V13 N11: LICENSED OPERATING REACTORS STATUS NUREG-1214 ROS. HISTORICAL DATA

SUMMARY

OF THE SYSTEM-

SUMMARY

REPORT. Data As Of October 30.1989(Gray Book f)

ATIC ASSESSMENT OF LICENSEE PERFORMANCE.

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I NRC Originating Organization index (Intemational Agreements)

.This index lists those NRC organizations that have published international agreement re-

!) orts, The index is arranged alphabetically by major NRC organizations (e.g., program of-Wes) and then by subsections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

T h Tccw"ucYt 8AEo"va*Y"on'OEsI'EcEn?8o'nW

" $ 'si'$ $ $i"' " "' " ' ** " ' " * " " *

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A s

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 organ!zation (e.g., program office) and 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) areaared by that organi-zation. If further information is needed, refer to the main citation by tie NUREG/CR number.

EDO. OFFICE FOR ANALY818 & EC

' 4N OF OPEhATIONAL NUREG/CR 6229 V02. TML2 IPICOR il RESIN / LINER INVESilGA-DATA TION LOW 4EVEL WASTE DATA BASE DEVELOPMENT PF10-OFFICE FOR ANALYSIS & (W ATONCe IRATONAL DATA. Dl-GRAM FOR FISCAL YE AR 1989 Annual Report RECTOR NURE G/CR 6386 DASIS FOR SNUDBEFI AGING FIESE ARCH, NU-NUREG/CR 2000 V06N12, UCENMT afVENT Fi[ PORT (LE R)

CLEAR PLANT AGING RESEAF4CH PROGRAM COMPILATION F or Month t.

w 19tt9 NUREG/CR 6404 V01: AUXILLARY FEEDWATER SYSTEM AGING NUREG/CR-P000 V09 N 1.

. p

/ EVENT FtEPORT (LER)

STUDY.

COMPILATON For Month ol A 1993 NUREG/CR 6406. ANALYSIS OF BHELL RUPTURE FAILURE DUE NUHEG/CRM00 V09 N2-Lit. 6EE EVE NT REPORT (LER)

TO HYPOTHETICAL ELEVATE D TEMPERATUFIE PRESSURIZA-COMPIL ATION F or Month Of F ettuary 1990.

flON OF THE SECVOYAH UNIT 1 STEEL CONTAINMENT DUILD-DIVISION OF SAFETY PROGRAMS (POST 870413)

ING NURE G/CF14674 V09 PRE CURSOAS TO POTENTIAL SEVERE NUREQ/CR 6419 AGING ASSESSMENT OF INSTRUMENT AIR SYS-CORE DAMAGE ACCIDENT S 1986 A STATUS REPORT Main TE MS IN NUCLE AFI POWE R PL ANTS Report Aruf APPerds A NUREG/CR 6436 ENVIRONMENTAL EFFECTS ON CORROSION IN NURE G/CR4674 V10 Ptt 1Ul4 SONS TO Poif NTIAL SEVERE THE TURF REPOSITORY.

CORE DAMAGE ACCIDEnf 6198B A STATUS ALPORT.Appermes NUREG/CFI 6460 HOH TEMPERATURE CRACK-ARREST TESTS B And C.

USING 162-MM THICK SEN WIDE PLATES OF LOW UPPER SHELF BASE MATERIAL TESTS WP 2 2 AND WP 2 6.

EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CR-6476 POSTTEST ANALYSIS OF A 16 SCALE FIEIN.

DIVISION OF HOH LEVEL WAbfl MANA3EME NT (POST 670413)

FORCED CONCFIETE REACTOR CONTAINMENT BUILDING NUREGICR 62f4 COMPONENTS Or AN OVERALL PERFORMANCE NUREG/CR 6478 IMPROVED EDDY CUFIRENT INSPFCTON FOR NU E R 338 TE HN SIS FOR FlEVIEW OF HIGH. LEVEL m

97 WASTE FIEPOSITORY MODELING NUREG/CR-6482. LADORATORY ANALYSi$ OF fluid FLOW AND' DIVISON OF i.OW4EVEL WASTE MANAGEMENT & DECOMMIS$10N.

SOLUTE TRANSPORT lHROUGH A VARIABLY SATURATED ING (POST 870413)

FRACTURE EMBEDDED IN POROUS TUF F.

NUREG/CRA169 MODtL12ATION AND TRANSPORT OF VRANIUM NUREG/CR b484 PH SENSORS DASED ON 1RIDIUM OXIDE AT URANIUM MILL T AILINGS DISPOSAL SITES.Applicahon Of A NUREG/CR 6491: SHIPPINGPORT ST ATION AGING EVALUATION Chemical Transport Madet.

NUFtEG/CR-64?2 INVE STIGATIONS OF IRFIADIATION ANNEAL-NUREG/CR 63*1. ECONOMIC FilSN OF CONTAMINATON CLEANUP REIRRADIATION (IAR) PROPERTIES TRENDS OF FtPV COSTS RESULTING FROM LAF1GE NONREACTOR NUCLEAR MA.

WELDS Phase 2 F6nal Report TERIAL LICENSEE OPERATIONS-NUREGICR-b493. INFLUENCE OF FLUENCE RATE ON RADIATON-INDUCED MECHANICAL PROPEFITY CHANGES IN REACTOR EDO. OFFICE Or NUCLE AR REGULATORY RESE AHCH (POST 820406 PRESSURE VESSEL STEELSFinal Report On EuPloratory Expen-OFFICE OF NUCLEAR REGULATORY RESEARCH. DIRECTOR (POS monts 860720)

NUREG/CR-6506 PRELIMINARY STF4UCTURAL FVALUATON OF NUREG/CM331 V09 N3 SAFETY Fif SEA 1CH PROGRAMS SPON-TROJAN FICL SUBJECT TO POSTULATED RPY SUPPORT FAIL-SORED BY OFFICE OF NUCLEAR HEGULATORY URE RE SE ARCH Pr ress Report. July-September 1989 NUREG/CR-6511: IRRADIATON EFFECTS ON STRENGTH AND To ESS OF NU lE CR 3 V

E HYS L VESTIGATONS OF THE w

gp Ay C WESTERN OHIOlNDiANA FIEGION Annual Report, October 1988 '

DIV! SON OF REGULATORY APPLICATIONS (POST 870413)

Septemter 1989 NUREG/CR-3668 MINET CODE DOCUMENTATION NUREG/CR 4469 V0B NONDESTRUCTIVE E AAMINATON (NDE) RE' NUREG/CR4214 R01 PL HEALTH [FFECTS MODELS FOR NUCLE.

LIABILITY FOR INSERVICE INSPECTION OF LIGHT WATER AR POWER PLANT ACCIDENT CONSEQUENCE ANALYSISlow RE ACTORS Senuannual Rerert.Octoter 19R7 March 1988 LET Radiation Pert 1introduchon. Integration And Summary.

NURES/CR 4469 V09 NONDESTRUCTIVE E AAMINATION (NDE) RE-NUREG/CR 4704 V03 A'!LATIVE BIOLOGICAL EFFECTIVENESS LIABILITY FOR INSEF1VICE INSPECTION OF LIGHT WATER (RBE) OF FISSION NEUTRONS AND GAMMA RAYS AT OCCUPA-RE ACTORS Semannual Report, Apnl Septemter 1988 TONAL EXPOSURE LEVELS Studes On The Gross And Microscop.

NUREG/CR 4554 V06 SCANS (SHIPPING CASK ANALYSIS 4c Pathology Observed At Death Of Mice Exposed To t>0 Equal SYSTEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR Once Weekly Dosos Of Fission...

SHIPPING CASK DESIGN REVIEW Volume 6 Thenrv Manuel Buck-NUREG/CR-6516 CAUSES OF FAILING THE DRAFT ANSI STAND-knq Of Cecular Cylindncal Shells ARD N13 30 RADOBIOASSAY PERFORMANCE CRITERON FOR NUALGICR 4554 V07. SCANS (SHIPPING CASK ANALYGtS MINIMUM DETECT ABLE AMOUNT.

SYSTEMI A MICROCOMP!)TER BASED ANALYSIS SYSTEM FOR DIVISION OF SYSTEMS RESE ARCH (POST 880717)

SHIPPING CASK DESIGN REVIEW Volume 7. Theory Manual Punc-NUREG/CR-4550 V01 RI: ANALYSIS OF CORE DAMAGE FREQUEN.

tura Of SNpping Casks CY:INTEhNAL EVENTS METHOOOLOGY.

NUREG/CR-4659 V03. SEISMIC FRAGILITY OF NUCLEAR POWER NUREG/CR4668 DAMAGED FUEL EXPERIMENT DF 1.Results And PLANT COMPONENTS (PitASE 11)Selchgear, l&C Panels (NSSS)

Ana%es And Relays NUREG/CR4671 THE OF4 FUEL DAMAGE EXPERIMENT IN ACRR NUREG/CR-4731 V02 RESIDUAL LIFE ASSESSMENT OF MAJOR WITH A BWR CONTROL BLADE AND CHANNEL BOX.

LIGHT WATE R REACTOR COMPONENTS - OVE RVIEW NUREG/CR 4691 V02. MELCOR ACCIDENT CONSEQUENCE CODE NUFIEG/CR 4744 V03 N1: LONG TERM EMBRITTLEMENT OF CAST SYSTEM (MACCS) Voiume 2 Model DescriPhon.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual NUREG/CR4691 V03 MELCOR ACCIDENT CONSEQUENCE CODE Report.0ctoter 1987 March 1988 SYSTE M (MACCS) Volume 3 Programmer's Reference Manual NUREG/CR4882 OUALIFICATION PROCESS FOR UL1RASONIC NURE G/CR-6273 V04 SCDAP/RELAP5/ MOD 2 CODE TESTING IN NUCLE AR INSERVICE INSPECTON APPLICATIONS MANUALVOLUME A MATPRO A LIBRARY OF MATERIALS 39

40 NRC Contract Sponsor index PROPERTIES FOR LIGHi-WATER-RE ACTOP. ACCIDENT ANALY.

EDO OFFICE OF MUCLEAR REACTOR R$00LAtt0N T 4/st/te) bl$

OlVISION OF OPERATIONAL EVENTE AbSESSMENT POST 970411)

NURE G/CR-6316 WELT PRO 3RES$lON.OKIDATION, AND NATu-NUREG/CR-4661 CLOSEOUT OF BluLLETIN $6-03 IAOTOR OP.

RAL CONVECTION IN A SEVERELf DAMAGED REACTOR CORE ERATED VALVE COMMON 4AOL ALLURES DURING PLANT NUREG/CR 6368 REACTIVITY ACT,lDENTS.A Rossnesortent Of The R^ $'

8 DU H

N AE /r 296 EOUT LL Tl 7 IPE CRACKS Den 4 Bases Gents IN ST AGNANT BORATED WATE5, JYSTEMS AT PWR PLANTS NURE /CR 6376 OVALITY ASSURANCE AND VERIFICATION OF NUREG/CR-6289 CLOSEOUT OF IE BULLETIN 7923 POTENTIAL THE MACCS VODE verson 16 F AILURE OF EMERGENCY DIESEL GENERATOR FIELD EXCITER NURE GICR-6396 V02 MULTILOOP INTEGRAL SYSTEM TEST TRANSFORMER (MIST) FINAL REPORT Test Group 30.Mapprig Tests NUREG/CR 6?pe CLOSEOUT OF IE BULLETIN 86-01: STEAM BINDw NUREG/CR 6396 V10 MULTILOOP INTEGRAL SYSTEM TEST ING OF AUXIUARY FEEDWATER PVWPS (MIST) FINAL REPORT RELAP6/MODP MIST Analyms Compen.

NUREG/CR 6302 CLOSEOUT OF IE BULLETIN 9040 CONTAMINA.

sons TlON OF NONRADIOACTIVE SYSTEM AND RESULTING POTEN-NUREG/CR 6424 ELICITING AND ANALY7ING EXPERT TLAL FOR UNMONITORED. UNCONTROLLED RELEASE OF RA-JUDGE MENT. A Practcal Gm DIOACTIVITY TO ENVIRONMENT NUREG/CR 6472 A RISK-SASED FtEVIEW OF INSTRUMENT AIR NUREG/CR-6307 CLOSEOUT OF IE BULLETIN 9002 INADEQUATE SYSTEMS AT NUCLE AR POWER PLANTS OVALITY ASSURANCE FOR NUCLEAR SUPPUED EQUIPMENT E

N EM g27f0N MONI-NUREG/CR 6474 ASSESSMENT OF CANDIDATE ACCIDENT MAN.

R /R pp A

NUREGICR 7 A EV LUATION OF THE RELIABILITY AND USE-og S:

o, tor V niernais FULNESS OF EXTE81NAL INITIATOR PRA ME'THODOLOGIES ew A DivlSION OF REGULATORY APPLICATIONS (960720870413)

DIViS60N OFNSTEMS TECHNOLOGY (POST 890827)

NUREG/CR-6612 DAF FC RESIDUAL RADIOACTIVE CONT AMINA-NUREG/CR 6421 LAPUR USER'S GUIDE TION FROM DECOMMISSIONING Techncal Bases For Translatsng DIVISION OF RADLATION PROTECTION & EME.RGENCY PREPARED $

Conterrwneten Levels To Annual Dose Draft Repor1 For Comment N&

7 11 q

LASALLE PRA 5

um-mi-n muss

i 1

l 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.

amam8 NATIONAL LADORATORY NUREG/CR4308: REACTIVITY ACCIDENTS A Resteosoment Of The NURES/CR 4704 V03 RELATIVE BOLOGICAL EFFECTIVENESS (RBE)

Design-Been Events OF FISSON NEUTRONS AND GAMMA RAYS AT OCCUPATONAL NUREGICR4419. AGING ASSESSMENT OF INSTRUMENT AIR SYS-EXPOSURE LEVELS Studies On the Grote And Microscopic Pethoto.

TEMS IN NUCLEAR POWER PLANTS.

Oy Oteerved At Doom Of Mice Erposed To 60 Equal Once-weekty NUREG/CR4472. A RISK BASED REVIEW OF INSTRUMENT AIR SYS-Doese Feeiort...

TEMS AT NUCLEAR POWER PLANTS NUREG/pR 4744 V03 N1: LONG TERM [WBRITTLEMENT OF CAST NUREG/CR4474: ASSESSMENT OF CANDIDATE ACCIDENT MAN.

w DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Sentennual AGEMENT STRATEGIES.

Report. October 1987. Merch 1998 NUREG/CR4527; RISK SENSITIVITY TO HUMAN ERROR IN THE LA-NUREG/CR 5229 V02: TMi-2 EPICOR Il RESIN / LINER INVESTIGATON SALLE PRA.

LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FISCAL YEAR 1989. Annual Report CORTEST COLUMBUS, INC.

NUhEG/CR4435: ENVIRONMENTAL EFFECTS ON CORROSION IN AMIONA, M. 06 TUCSON, AZ THE TUFF REPOSITORY

• NUREG/CR4482; LABORATORY ANALYSIS OF FLUID FLOW AND SOLUTE TRANSPORT THROUGH A VARIABLY SATURATED FRAC.

8060 IDAHO,1NC (SUS $. 0F E060,1NC.)

TURE EMBEDDED IN POROUS TUFF

• NUREGICR-4731 V02: RESIDUAL LIFE ASSESSMENT OF MAJOR LIGHT WATER REACTOR COMPONENTS OVERVIEW.

DADCOCK & WILCOM CO NUREG/CR4229 V02 TWi-2 EPICOR-Il RESIN / LINER INVESTIGATION.

NUREG/CR4395 V02. IluLTILOOP INTEGRAL SYSTEM TEST (MIST).

LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FINAL REPORT. Test Gro@ 30 Mappi Tooto NUREG/CR4395 VIO: MULTILOOP IN ~ ORAL SYSTEM TEST (M6ST):

FISCAL YEAR 1989. Annual Report FINAL REPORT. HELAPS/ MOD 2 MIST Analysse Cornpensons NUREG/CR4273 V04-SCDAP/RELAP5/ MOD 2 CODE M ANUAL. VOLUME 4 MATPRO. A LIBRARY OF MATERIALS PROP-BATTELLE MEMORIAL INSTITUTE, COLUMSUS LADORATORIES ERTIES FOH t.lGHT WATER. REACTOR ACCIDENT ANALYSIS.

NUREG/CP4109 PROCEEDINGS OF THE $[MINAR ON LEAK.

NUREG/CR4376: QUALITY ASSURANCE AND VERIFICATION OF THE BEFORE BREAK.Further Developmente in Regulatory Pohcies And MACCS CODE, Version 1.6 SuppDrtmg ReseerCPL ERC ENVIRONMENTAL & ENERGY SERVICES,INC.

BATTELLE MEMORIAL INSTITUTE. PACIFIC le0RTHWEST NUREG/CR-4550 V01 RI: ANALYSIS OF CORE DAMAGE FREQUEN-LADORATORY CY: INTERNAL EVENTS METHODOLOGY.

NUREGICR4469 V00 NONDESTRUCTIVE EXAMINATON (NDE) RELL NUREG/CR4381: ECONOMIC RISK OF CONTAMINATION CLEANUP ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER COSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE-NUEGC 4 09 N T UC V M A NDE)RELL ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER FUTURE RESOURCES ASSOCIATES,1NC.

NUREG/CR4477: AN EVALUATION OF THE RELIABILITY AND USE.

lKW S

R ULTRASONIC FULNESS OF EXTERNAL INITIATOR PRA METHODOLOGIES.

N PE /CR 882 OU TESilNG IN NUCLE AR INSERVICE INSPECTION APPLICATIONS NUREG/CR4169 MOBill2ATION AND TRANSPORT OF URANIUM AT 0

8T URANIUM MILL l LINGS DISPOSAL SITES. Application Of A Chemh G 139 E KAGE OF AN IRRADIATOR SOURCE. THE JUNE NUREQ/

86. BASIS FOR SNUBBER AGING RESEARCH. NUCLE.

1988 GEORGIA RSI INCIDENT.

AR PLANT AGING RESEARCH PROGRAM.

NUREG/CR4491: SHIPPINOPORT STATON AGING EVALUATON ORAM, INC.

NUREG/CR 5512 DRF FC RESIDUAL RADIOACTIVE CONTAMINATON NUREG/CR4381: ECONOMIC RISK OF CONTAMINATION CLEANUF FROM DECOMMISSIONING. Techn6 cal Basie For Trenatating Contann-COSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE-nation Levelo To Annual Dose Draft Rept /t For Comment RIAL LICENSEE OPERATIONS.

NUREG/CFv5516-CAUSES OF F AILING THE DRAFT ANSI STANDARD N13 30 RADIODOASSAY PERFORMANCE CRITERON FOR MINI-HARVARD SCHOOL OF PUBLIC HEALTH. DOSTON, MA MUM DETECTABLE AMOUNT.

NUREQ/CR-4214 ROI Pl. HEALTH EFFECTS MODELS FOR NUCLEAR POWER PLANT ACCOENT CONSEOVENCE ANALYSIS Low LET E

1 A

GS OF THE SEVENTEENTH WATER RE ACTOR SAFETY INFORMATION MEETING LAKE ENQlNEERING,INC.

SEVENTEENTH NUREG/CR4386: BASIS FOR $NUBBER AGING RESEARCH: NUCLE-WATER R OR TETY N O MA ON MEE NO AR PLANT AGING RESEARCH PROGRAM.

NUREG/CP-0105 V03. PROCEEDINGS OF THE SEVENTEENTH LAWRENCE WmmE NANAL LAMAN HU G/CP 1 R EE S

TH I A ONAL WORK.

NUREG/CR 4554 V06 SCANS (SHIPPING CASK ANALYSIS SYSTEM) A SHOP ON NEW DEVELOPMENTS IN OCCUPATONAL DOSE CON.

TROL AND ALARA IMPLEMENTATION AT NUCLEAR POWER MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN REVIEW Volume 6 Theory Manual Buckkng Of Cwcu-PLANTS AND SIMILAR FACILITIES NUREG/CR 2331 V09 N3. SAFETY RESEARCH PROGRAMS SPON.

lar Cylindrtal Shells SORED BY OFFICE OF NUCLEAR REGULATORY NUREGICR-4554 V07: SCANS (SHIPPING CASK ANALYSIS SYSTEM)

A MtCROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING RESE ARCH Progroot R

.J Septernbar 1969 NUREG/CR 3868. MINET UWENTATION CASK DESIGN REVIEW Volume 7: Theory Manual Pureture Of Ship-NUREG/CR 4659 V03 S ISMIC FRAGILITY OF NUCLEAR POWER p6ng Casks.

PLANT COMPONENTS (PHASE II)Switchgear LAC Penels (NS$$)

NUREGICR4506: PRELIMINARY STRUCTURAL EVALUATON OF And Reiere.

TROJAN RCL SUBJECT TO POSTULATED RPV SUPPORT FAILURE.

41

42 Contractor index LOS ALAlfo$ NATIONAL LADORATORY PARAtatTER, INC.

NUREG/CR-6424 ELICITING AND ANAf.YZING EXPERT NUREG/CR 4661: CLOSEOUT OF IE BULLETIN 86 03 MOTOR OPER-JUDGEMENT.A Practical Gale.

ATED VALVE COMMON MODE FAILURES DURING PLANT TRAN-SIENTS DUE TO IMPROPER SWITCH SETTINGS, 98478 RIALS 8000lNEERIIe8 At90CIAftl,INC.

NUREG/CR 6286. CLOSEOUT OF IE BULLETIN 79-17. PIPE CRACKS IN NUREG/C46492. INVESteGATIONS OF IRRADtATION,ANNEALJIEIR-STAGNANT BORATED WAT{R SYSTEMS AT PWR PLANTS NUR G CR-89 OSEOU IE B LLET N 79 23 POT A

ATION BAR) PROPERTtES TRENDS OF RPV WELDS. Phase 2 NUREG/CR 6493. INFLUENCE OF FLUENCE RATE ON RADIATION-IN-NhE

/CR 2 CLOSEOUT OF IE BULLETIN 86-01: STEAM BIND-DUCED MECHAN CAL PROPERTY CHANGES IN REACTOR PRES" ING OF AUXILIARY FEEDWATER PUMPS.

SURE VESSEL STEELS.Finni Report On Exploretory Expenments.

NUREGiCR 6302: CLOSEOUT OF lt BULLETIN 6010. CONTAMiNA.

TION OF NONRADIOACTIVE SYSTEM AND RESULTING POTENTIAL 801CWitAN. UNIV. OF ANN AR80R, tel FOR UNMONITORED. UNCONTROLLED RELEASE OF RADIOACTly.

NNREG/CR 3146 V08. GEOPHYSICAL INVESTIGATIONS OF THE ITY TO ENVIRONMENT.

WESTERN OHIOIND6ANA REGION. Annual ReportOctober 1968 NUREQ/CR-6307. CLOSEOUT OF IE BULLETIN 8002: INADEQUATE September 1989.

QUALITY ASSURANCE FOR NUCLEAR SUPPLIED EQUIPMENT.

NAfl0NAL IIISTffVit OF ST ANDARDS & TECNIt0 LOGY (POR00ERLY SANDIA NATIONAL LADORATORIES NAfl004AL DUREAU Op NUREG/CR-4214 R01 PI: HEALTH EFFECTS MODELS FOR NUCLEAR NUREG/CR 6450. HIGH-TEMPERATURE CRACK ARREST TESTS POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS. Low LET N

bbm S AGE N M N E /

6 E MATE L TESTS P2 N P26 NUREG/CR-6404. PH SENSORS BASED ON 1RIDIUM OXIDE.

whnNefn7,NT4(4o[%EiETP[RIMENT DF 1.Results And 8

OAK R100E NAYlostAL LAtoRATORY NURYNR4671: THE DF 4 FUEL DAMAGE EXPERIMENT IN ACRR NUREG/CR-2000 V00N12: LICENSEE EVENT REPORT (LER)

WITH A BWR CONTROL BLADE AND CHANNE BOX.

COMPILATION For Month 0f December 1989.

NUREG/CR4691 V02. MELCOR ACCIDENT C5' NSEOuENCE CODE NUREG/CR-2000 V09 N1: LICENSEE EVENT REPORT (LER)

NUREG/CR 4691 SYSTEM (MACCS)03Vo6ume 2 Model Deserptson COMPILATION.For Month Of January 1990.

V MELCOR ACCIDENT CONSEQUENCE CODE NUREG/CR 2000 V00 N2: LICENSEE EVENT REPORT (LER)

SYSTEM MACCS) Volume 3 Programmer's Reference Manuel NUREG/CR(6266. COMPONENTS OF AN OVERALL PERFORMANCE COMPILATION.For Month Of February 1990 NUPEG/CR4671: THE DF 4 FUEL DAMAGE EXPERIMENT IN ACRR ASSESSMENT METHQDOLOGY.

WITH A BWR CONTROL BLADE AND CHANNEL BOX NUREG/CR 6316: MELT PROGRESSION. OXIDATION, AND NATURAL NUREG/CR 4674 V09 PRECURSORS TO POTEmlAL ' SEVERE CORE NUNG/C INC lbi AbN T DAMAGE ACCIDENTS.1988 A STATUS REPORT Main Report And Ap' CLEANUP COSTS RESULTING FROM LARGE NONREACTOR NUCLEAR MATE-NU CR 4674 V10. PRECURSORS TO POTENTIAL SEVERE CORE HvYoN3,E 8

o C C BASIS FOR REVIEW OF HIGH LEVEL DAMAGE ACCIDENTS 1968 A STATUS REPORT.Appendimos B And WASTE REPOSITORY MODELING C-NUREG/CR-6406: ANALYSIS OF SHELL RUPTURE FAILURE DUE TO NUREQ/CR 6220 V02: TM6-2 EPICOR li RESIN / LINER INVESTIGATION.

HYPOTHETICAL ELEVATED. TEMPERATURE PRESSURIZATION OF LOW LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR THE SEOOOYAH UNIT 1 STEEL CONTAINMENT BUILDING.

FISCAL YEAR 1989. Annual Report NUREG/CR-6476-POSTTEST ANAL \\$18 OF A 1:6-SCALE REIN-6"JREG/CR 6404 V01: AUXIUARY FEEDWATER SYSTEM AGING FORCED CONCRETE REACTOR CONTAINMENT BUILDING.

STUDY.

NUREG/CR 6421:LAPUR USER'S GUIDE.

SCIEleCE APPLICAtl0048 lefTERNATioNAL CORP. (FonteERLY NUREG/CR-6460. HIGH TEMPERATURE CRACK-ARREST TESTS SCitteCE APPLICATIONS, USING 162 MM THICK SEN WIDE PLATES OF LOW UPPER-SHELF NU9EG/CR 4660 V01 R1: ANALYSIS OF CORE DAMAGE FREQUEN-BASE MATERIAL TESTS WD 2.2 AND WP 2 8 CY: INTERNAL EVENTS METHODOLOGY.

NUREG/CR 6478. IMPROVED EDDYCURRENT INSPECTION FOR NUREG/CR-6472: A RISK BASED REVIEW OF INSTRUMENT AIR SYS-STEAM GENERATOR TUDING. Progress Report For January 1986 To TEMS AT NUCLEAR POWER PLANTS.

December 1987.

NUREG/CR.6479 CURRENT APPLICATIONS OF VIBRATION MONI.

TAlWAN POWER C0' PROCEEDINGS OF NUREG/CP 0109.

THE SEMINAR ON LEAK.

TORING AND NEUTRON NOISE ANALYSIS. Detection And Analys6s Of DEFORE BREAK.Further Jovelopmenta in Regulatory Policies And Structural Degradation Of Reactor vessel Intemals From Operational Supporting Research.

NU E /C46611: IRRADIATION EFFECTS ON STRENGTH AND WYLE LADORATORIES TOUGHNESS OF THREE-WIRE SERIES ARC STAINLESS STEEL NUREG/CR 6386: BASIS FOR SNUBBER AGING RESEARCH NUCLE-WELD OVERLAY CLADD4NG.

AR PLANT AGING RESE ARCH PROGRAM.

E a

. = :

r international Organization index This index lists, in alphabetical order, the countries and performing organizations that pre-1:ered the NUREG/lA reports listed in this compilation. Listed below each country and per.

lorming organization are the NUREG/lA numberc and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number.

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Licensed Facility inder, 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 and followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

mas C-. Poe $isei t== mm = i. mum ser av.

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Vol.15, No.1 a.ioa anosuamti Regulatory and Technical Repot ts (Abstract Index Journal) 3 o*YE REPOR1 PUBLISH { o l

39h0 Compilation for g,y First Quarter 1990

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January - March 6,1vPL o, REPOR1

6. AU1Horils4 Reference F. FL Riou Lov L h L o us.ww v..eis January-March 1990 e Pi n e.on.uiwo.onc.awizai ion.==vi *~o avoa i >6 in a.c,,-. o o"...- v2 =- -...-

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10. svPPLEMLh1 ANY NOTE %

t1. ABsinACI tiuo

,m er mu, This journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceedings of conferences and workshops; as well as 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, u....... n u u -

sa a t v wunos u scs + i ca s.....-. -,,.,~,..u m.

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Main Citations and Abstracts

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LICATIONS SVCS I Secondary Report P-223 WASHINGTON OC 20555 Number index Personal Author index Subject Index NRC Originating Organization h.

Index (Staff Reports)

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NRC Originating Organization Q

index (International Agreements)

NRC Contractor SponsorIndex Contractor index international Organization 3

Index l

Licensed Facility index.

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