ML20245F773

Regulatory and Technical Reports (Abstract Index Journal). Compilation for First Quarter 1989,January-March

ML20245F773
Person / Time
Issue date:	07/31/1989
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V14-N01, NUREG-304, NUREG-304-V14-N1, NUDOCS 8908150066
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NUREG-0304 Vol.14, No.1 Regulatory and Technical Reports (Abstract Index Journal) I

' Compilation for First Quarter 1989 January - March U.S. Nuclear Regulatory Commission Office of Administration se arco j

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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 I this publication. I Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 l

e - NUREG-0304

. Vol.14, No.1 .

Regulatory and Technica Reports (Abstract Index Journal)

Compilation for First-Quarter 1989 January- March ,

Date Published: July 1989

.. Regulatory Publications Branch I Division of Freedom ofInformation and Publications Services Office of Administration

' U.S. Nuclear Regulatory Commission ,

.. Washington, DC 20555

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3 CONTENTS J Pref ace . . . . . . . . . . . . . . . . . . . . . .

Index Tab

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. .. . . I Main Citations and Abstracts . . . . . . . . . . . . . . .

• Staff Reports

• Conference Proceedings

• Contractor Reports 2

• International Agreement Reports ... ... . ... . .

.. 3 Secondary Report Number index . . . ... .. ........

Personal Author index . . . . . . . . . . . . . . . . . . . ....... . ... . ... .

. .. ........... .. ... .. 45 Subject inde x . . . . . . . . . . . . . . . . . . . . . . . . . . .. ......... ... .... 6 NRC 0riginating Organization index (Staff .Reports) . .... .. ..

. .. .. .NRC Originating Orga 7

8 NRC Contract Sponsor index (Contractor Reports) .

9 Contractor index . . . . . . . . . . . . . . . . . . .... ..... . . .. .

...... 10 International Organization index . . . .. .

Licensed Facility index . . . . . ..... .

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PREFACE .l 1

i This compilation consists of bibliographic data and abstreets for the formal regulatory and technical  !

reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors, it is NRC's i 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 Publications Management Branch  !

Publishing and Translations Section l' Woodmont 537

'U.S. Nuclear Regulatory Commission Washington, D.C. 20555 j i

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

I Secondary Report Number Index i 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 i Licensed Facility index A detailed explanation of the entries precedes each index.

-- The bibliographic elements of the main citations are the following:

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Staff Report NUREG-0808: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. 2 ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048. 09570:200. ,

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

Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National ,

Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070. l l

Where the entries are (1) regort 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 Cantrol System accession number, (8) the repott number of the originating organization, (9) the ,

microfiche address (for NRC internal use), j j

Contractor Report {

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

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

Where the entries are (1) report nu'mber, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) datc icport 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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1 International Agreement Report NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA. NEUMANN, U. Kraftwerk Union. August 1986. 223 pp. 8608270424, 37659:138.

Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use).

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

ADD - addendum APP - appendix DRFT - draft ERR - errata N - ramber 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 foliowing address:

Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275-2171. Non-U.S. customers must make payment in advance either by Intemational Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents.

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

in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/lA is used for international agreement reports.

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

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Main Citations and Abstracts

. The report listings in this compilation are arranged by report number, where NUREG-XXXX is en NRC staff-onginated report, NUREG/CP-XXXX is an NRC-sponsored conference report, NUREG/CR-XXXX is an NRC contractor-prepared report, and NUREG/;A XXXX is an inter-n:tional agreement report. The bibliographic information (see Preface for details) is followed by a brief abstract of this report.

NUREG-0020 V12 N12: LICENSED OPERATING REACTORS NUREG-0383 V01 R11: DIRECTORY OF CERTIFICATES OF STATUS

SUMMARY

REPORT. Data As Of November COMPLIANCE FOR RADIOACTIVE MATERIALS 30,1988.(Gray Book 1) SCHWARTZ,1. Policy & Program Manage- PACKAGES. Report Of NRC Approved Packages.

• Division of ment Staff (870413-890204). January 1989 539pp. Gafeguards & Transportation (Post 870413). December 1988.

8902080482.48399:126. 505pp. 8901230159. 40180:234.

The OPERATING UNITS STATUS REPORT - LICENSED OP- This directory contains a Report of NRC Approved Packages ERATING REACTORS provides data on the operation of nucle. (Volume 1), Certificates of Compliance (Volume 2), and a Er units as timely and accurately as possible. This information is Reoort of NRC Approved Quality Assurance Programs for Ra-collected by the Office of Administration and Resources Mars- dioactive Matenal Packages (Volume 3). The purpose of this di-tgement from the Headquarters staff of NRC's Office of En- rectory is to make available a convenient source of information forcement (OE), from NRC's ReDional Offices, and from utilities. on Quality Assurance Programs and Packagings which have The three sections of the report are: monthly highlights and sta. been approved by the U.S. Nuciear Regulatory Commission.

tistics for cominercial operating units, and errata from previously Shipments of radioactive material utilizing these packagings reported data; a compilation of detailed information on each must be in accordance with the provisions of 49 CFR 173.471 unit, provided by NRC's Regional Offices, OE Headquarters and and 10 CFR Part 71, as applicable. In satisfying the require-the utilities; and an appendix for miscellaneous information such ments of Section 71.12. It is the responsibility of the licensees Es spent fuel storage capability, reactor years of experience and to insure themselves that they have a copy of the current ap-non power reactors in the U.S. It is hoped the report is helpful preval and conduct their transportation activities in accordance to all agencies and individuals interested in maintaining an with an NRC apprued quality assurance program. Copies of awareness of the U.S. energy si'uution as a whole. the current approval may be obtained from the U.S. Nuclear Regu(atory Commission's Public Docket Room files (see Docket NUREG-0020 V13 N01: LICENSED OPERATlHG REACTORS No. listed on each certificate) at 2120 L Street N.W., Washing-STATUS

SUMMARY

REPORT. Data As Of December ton, DC 20555. Note that the general license of 10 CFR 71.12 31,1988.(Gray Book I) SCHWARTZ,t. Policy & Program Manage. does not authorize the receipt, possession, use or transfer of ment Staff (870413-890204). February 1989. 600pp. byproduct source, or special nuclear material; such authonzation 8904030094.49170:043. must be obtained pursuant to 10 CFR Parts 30 to 36,40,50 or See NUREG 0020,V12,N12 abstract. 70.

NUREG-0020 V13 N02: LICENSED OPERATING REACTORS NUREG-0383 V02 R11: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADICACT!VE MATERIALS STATUS

SUMMARY

REPORT. Data As Of January 31,1989.(Gray Book l) SCHWARTZ,t. Division of Computer &

PACKAGES. Certificates Of Compliance. Division of Safe-a ds napo tion (Post 870413). December 1988. 660pp.

Telecommunications Services (Post 890205). March 1989. [g01p3 68 500pp. 8903210207. 48938.053.

See NUREG-0383,V01,R11 abstract.

See NUREG-0020,V12,N12 abstract.

NUREG 0090 V11 NO3: REPORT TO CONGRESS ON ABNOR-MAL OCCURRENCES. July-September 1988.

• Office for Analy.

MPL A CE FR ADI ACTIVE MATERIA PACKAGES. Report Of NRC Approved Quality Assurance Pro-sis & Evaluation of Operational Data, Director. January 1989. grams For Radioactive Matonal Packages.

• Division of Safe-4Bpp. 8903140231. 48B49.289 guards & Transportation (Post 870413). December 1988. 203pp.

l Section 208 of the Energy Reorganization Act of 1974 edenti- 8901230181. 48174:300.

fies an abnormal occurrence as an unscheduled incident or See NUREG-0383,V01,R11 abstract.

event which the Nuclear Regulatory Commission determines to i

be significant from the standpoint of public health and safety NUREG-0386 D04 R11: UNITED STATES NUCLEAR REGULA-tnd requires a Quarterly report -of such events to be made to TORY COMMISSION STAFF PRACTICE AND PROCEDURE Congess. This report covers the penod July 1 to September DIGEST. July 1972 - Merch 1988.

• Office of the General Coun-30,1983. For this reporting period, there were no abnormal oc- sel. December 1968. 460pp. 8901230136. 48182:019.

currences at nuclear power plants licensed to operate. There This Revision 11 of the fourth edition of the NRC Staff Prac-were two abnormal occurrences under other NRC-issued li- tice and Procedure Digest contains a digest of a number of conses: multiple medical therapy misadministration at a single Commission, Atomic Safety and Licensing Appeal Board, and hospital and a mede.al diagnostic misadministration. Therm was Atomic Safety and Licensing Board decisions issued during the one abnormal occurrence reported by an Agreement State period from July 1,1972, to March 31, 1988, interpreting the (Texas) involving a medical diagnostic misadministration. The NRC's Rules of Practice in 10 CFR Part 2. This Revision 11 re-report also contains information updating some previously re. places, in part, earlier editions and supplements and includes ported abnormal occurrences. appropriate changes reflecting the amendments to the Rules o' Practice effective through March 31,1988.

1

2 M:In Cit:tirns cnd Abstrscts "

NUREG 0430 V09 N01: LICENSED FUEL FACILITY STATUS in April 1986 the staff of the U.S. Nuclear Regulatory Com-REPORT. inventory Difference Data. January-June 1988.(Gray mission issued its Safety Evaluation Report (NUREG-0781) re-Book it) Office of Nuclear Material Safety & garding the application of Houston Lighting and Power Compa-Safeguards. Director. March 1989. 14pp. 8904110417. ny (applicant and agent for the owners) for a license to operate 49249:115. South Texas Project, Units 1 and 2 (Dockets Nos. 50 498 and NRC is committed to the periodic pubhcation of licersed fuel 50-499). The facility is located in Matagorda County, Texas, facilities inventory difference data, following agency review of west of the Colorado River,8 miles north northwest of the town the information and completion of any related NRC investiga- of Matagorda and about 80 miles southwest of Houston. The tions. Information in this report includes inventory difference first supplement to NUREG.0781 was issued in September data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched 1986, the second supplement in January 1987, the third supple-uranium, plutonium, or uranium-233. ment in May 1987, the fourth supplement in July 1987 and the fifth supplement in March 1988. This sixth supplement provides NUREG-0540 V10 N10: TITLE LIST OF DOCUMENTS MADE updated information on the issues that had been considered PUBLICLY AVAILABLE. October 1 31,1988.

• Division of Free- previously as well as the evaluation of issues that have arisen dom of Information & Publications Services (880515-890204). since the fifth supplement was issued. The evaluation resolves January 1989. 374pp. 8903030364. 48704:001. all the issues necessary to support the issuance of a low-power This document is a monthly publication containing descrip. license for Unit 2.

tions of information received and generated by the U.S. Nuclear Regulatory Commission (NRC). This information includes (1) NUREG-0781 807: SAFETY EVALUATION REPORT RELATED docketed material associated with civilian nuclear power plants TO THE OPERATION OF SOUTH TEXAS PROJECT, UNIT and other uses of radioactive matenals, and (2) nondocketed 2. Docket No. 50-499.(Houston Lighting And Power Company)

• material received and generated by NRC pertinent to its role as Division of Reactor Projects - lif fV,V & Special Projects (Post a regulatory agency. The following indexes are included: Per- 870411). March 1989. 58pp. 8904110402. 49272:158.

sona! Author, Corporate Source, Report Number, and Cross Reference to Pnncipal Documents, in April 1986 the staff of the U.S. Nuclear Regulatory Com-mission issued its Ssfety Evaluation Report (NUREG-0781) re-NUREG-0540 V10 N11: TITLE LIST OF DOCUMENTS MADE Oarding the application of Houston Lighting and Power Compa-PUBLICLY AVAILABLE. November 1 30,1988.* Division of Free, ny (applicant and agent for the owners) for a license to operate dom of information & Publications Services (880515-890204). South Texas Project, Units 1 and 2 (Docket Nos. 50-498 and January 1989. 423pp. 8903080549. 48747:021. 50-499). The facility is located in Matagorda County, Texas.

See NUREG-0540,V10,N10 abstract. west of the Colorado River,8 miles north-aorthwest of the town NUREG 0540 V10 N12: TITLE LIST OF DOCUMENTS MADE of Matagorda and about 89 miles southwest of Houston. The PUBLICLY AVAILABLE. December 1 31,1988.

• Division of first supplement to NUREG-0781 was issued in September Freedom of information & Publications Services (880515- 1986, the second supplement in January 1987, the third supple-890204). March 1989 325pp. 8903210193. 48937:094. ment in May 1987, the fourth supplement in July 1987, the fifth See NUREG-0540,V10,N10 abstract. supplement in March 1988, and the sixth supplement in January 1989. This seventh supplement provides updated information on NUREG-0750 V28101: INDEXES TO NUCLEAR REGULATORY the issues that had been considered previously as well as the COMMISSION 1SSUANCES. July-September 1988.

• Division of evaluation of issues that have arisen since the sixth supplement Freedom of information & Publications Services (880515- was issued. The evaluation resolves all the issues necessary to 890204). January 1989. 54pp. 8902080321. 48405:295.

Digests and indexes for issuances of the Commission, the support the issuance of a full-power license for Unit 2.

Atomic Safety and Licensing Appeal Panel, the Atomic Safety NUREG-0800 05.2.2 R2: STANDARD REVIEW PLAN FOR THE and Licensing Board Panel, the Administrative Law Judge, the Directors' Decisions, and the Denials of Petitions for Rulemak- REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR ing are presented. POWER PLANTS. LWR Edition. Revision 2 To SRP Section 5.2.2, " Overpressure Protection," And Revision 1 To Branch NUREG 0750 V28 N05: NUCLEAR REGULATORY COMMISSION Technical Position RSB 5-2.

• Office of Nuclear Reactor Regu-ISSUANCES FOR NOVEMBER 1988. Pages 499-566.

• Division lation, Director (Post 870411). January 1989. 11pp.

of Freedom of Information & Publications Services (860515- 8902140010. 48457:123, 890204). January 1989. 76pp. 8902080470. 48402:098. Revision 2 to SRP Section 5.2.2 and Revision i to l's associ-Legal issuances of the Commission, the Atomic Safety and Li- ated Branch Technical Position RSB 5-2 incorporate a definition censing Appeal Panel, the Atomic Safety and Licensing Board of the phrase "at low temperatures" that is used in the require-Panel, the Adm6nistrative Law Judge, and NRC program offices ment for a low temperature overpressure protection (LTOP) are presented.

system. The phrase is defined by adding a new Branch Posrtion NUREG-0750 V28 N06: NUCLEAR REGULATORY COMM?SSION B 2, which specifies the enable temperature below which the ISSUANCES FOR DECEMBER 1988. Pages 567-833.

• Division LTOP system must be operable. Fracture considerations were of Freedom of Information & Publications Services (Post the basis for establishing the enable temperature. This revision 890205). March 1989. 225pp. 8904110445. 49248:089. provides relief for those plants whose licensees were interpret.

See NUREG 0750,V28,N05 abstract. ing the phrase "at low temperatures" in an overconservative NUREG-0750 V29 N01: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JANUARY 1989. Pages 1-87.

• Division of NUREG-0800 06.5.2 R2: STANDARD REVIEW PLAN FOR THE Freedom of information & Publications Services (Post 890205)- REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR March 1989. 94pp. 8904110415. 49250:239.

POWER PLANTS LWR Edition. Revision 2 to SRP Section 6.5.2, See NUREG 0750,V28,N05 abstract.

" Containment Spray As A Fission Product Cleanup System."

• NUREG-0781 S06: SAFETY EVALUATION REPORT RELATED Offee of Nuclear Reactor Regulation, Director (Post 870411).

TO THE OPERATION OF SOUTH TEXAS PROJECT, UNIT January 1989,16pp. 8902080267. 48405:279.

2. Docket No. 50-499 (Houston Lighting And Power Company)

• This revision eliminated overly conservative assumptions in Division of Reactor Projects liljV,V & Special Projects (Post estimating the effectiveness of post-accident fission product re-870411). Jariuary 1989.161pp. 8902080310. 48402:274. moval by containment spray systems.

Main Citations and Abstracts 3 NUREG-0800 06.5.4 R3: STANDARD REVIEW PLAN FOR THE NUREC-1100 V05: BUDGET ESTIMATES Fiscal Years 1990-REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR 1991.

• Division of Budget & Anhlysis (870413 890204). January POWER PLANTS. LWR Edition. Revision 3 To SRP Section 1989.181pp. 8901230046. 48178:009.

6.5,4, " Ice Condenser As A Fission Product Cleanup System."

• This report contains the f: scal year budget justifications to Office of Nuclear Reactor Regulation, Director (Post 870411). Congress. The budget provides estimates for salaries and ex-January 1989. 8pp. 8902130155. 48457:107, penses and for the Office of the Inspactor General for fiscal This revision incorporates the pH range specified in SRP Soc- years 1990-1991, tion 6.5.2, Revision 2, for recirculating solutions.

NUREG-1137 S08: SAFETY EVALUATION REPORT RELATED NUREG-0800 06.5.5 R0: STANDARD REVIEW PLAN FOR THE TO THE OPERATION OF VOGTLE ELECTRIC GENERATING REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Geor-POWER PLANTS. LWR Edition. Revision 0 To SRP Section gia Power Company et al)

• Division of Reactor Projects - 1/11 6.5.5, " Pressure Suppression As A Fission Product Cleanup (Post 870411). February 1989.109pp. 8903140114. 48840:233.

System."

• Office of Nuclear Reactor Regulation, Director (Post in June 1985, the staff of the Nuclear Regulatory Commission 870411). January 1989. 7pp. 8902130151. 48457:116. issued its Safety Evaluation Report (NUREG-1137) regarding This section resolved an inconsistency in NRC regulatory the application of Georgia Power Company, Municipal Electnc guidance by giving credit to pressure suppression pools as post- Authority of Georgia, Oglethorpe Power Corporation, and the accident fission product cleanup systems in boiling water reac- City of Dalton, Georgia, for licenses to operate the Vogtle Elec-tors. tnc Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 50- 425). Supplement 1 to NUREG-1137 was issued by the staff NUREG-0837 V08 NO3: NRC TLD DIRECT RADIATION MONI- in October 1985, Supplement 2 was issued in May 1986, Sup-TORING NETWOR A. Progress Report. July-September 1988. plement 3 was issued in August 1986, Supplement 4 was STRUCKMEYER RJ MCNAMARA,N. Region 1, Ofc of the Direc- issued in December 1986, Supplement 5 was issued in January tor. January 1989. 232pp. 8902080504. 48406:267. 1987, Supplement 6 was issued in March 1987, and Supple-This report orovides the status and results of the NRC Ther- ment 7 was issued in January 1988. The facility is located in moluminescent Dosimeter (TLD) Dir9ct Radiation Monitonng Burke County, Georgia, approximately 26 miles south-southeast Network, it presents the radiation levels measured in the vicinity of Augusta, Georgia. and on the Savannah River. This eighth of NRC licensed facility sites tuoughout the country for the suoplement to NUREG-1137 provides recent information regard-quarter of 1988. ing resolution of some of the open and confirmatory items that remained unresolved following issuance of Supplement 7. SSER NUREG-0936 V07 N04: NRC REGULATORY AGENDA Quarterly 8 is the final supplement to the SER in support of the issuance Report,0ctober December 1988.

• Division of Freedom of Infor- of a low-power license for Vogtle Unit 2 and provides the con-mation & Publications Services (880515-89C204). January 1989. clusions, on the basis of the staff's review of the Final Safety 121pp. 8902100122. 48442:151. Analysis Report (FSAR) through Amendment 39, that Unit 2 The NRC Regulatory Agenda is a compilation of all rules en may be issued a license authonang power up to 5 percent.

which the NRC has proposed or is considering action and all petitions for rulemaking which have been received by the Com. NUREG 1137 SO9: SAFETY EVALUATION REPORT RELATED mission and are pending disposition by the Commission. The TO THE OPERATION OF VOGTLE ELECTRIC GENERATING Regulatory Agenda is updated and issued each quarter. PLANT, UNITS 1 AND 2 Docket Nos. 50-424 And 50 425(Geor-gia Power Company et al)

• Division of Reactor Projects - 1/11 NUREG-0940 V07 N04: ENFORCEMENT ACTIONS:SIGNIFICANT (Post 870411). March 1989 22pp. 8904110499. 49224.338.

ACTIONS RESOLVED.Ouarterly Progress Report, October-De- In June 1985, the staff of the Nuclear Regulatory Commission cember 1988.

• Ofc of Enforcement (Pest 870413). February issued its Safety Evaluation Report (NUREG 1137) regarding 1989. 400pp. 8904030097. 49169.001- the application of Georgia Power Company, Municipal Electnc This compilation summarizes significant enforcement actions Authonty of Georgia. Oglethorpe Power Corporation, and the that have been resolved dunng one quarterly period (October - City of Dafton, Georya, for licenses to operate the Vogtle Elec-December 1988) and includes copies of letters, Notices, and tric Generating Plant, Units 1 and 2 (Docket Nos. 50 424 and Orders sent by the Nuclear Regulatory Commission to licensees 50- 425). Supplement 1 to NUREG 1137 was issued by the staff with respect to these enforcement actions. It is anticipated that in October 1985, Supplement 2 was issued in May 1986 Sup-the information in th s publication will be widely disseminated to piement 3 was issued in August 1986, Supplement 4 was mariagers and employees engaged in activities licensed by the issued in December 1986, Supplement 5 was issued in January NRC, so that actiuns can be taken to improve salety by avoid- 1987, Supplement 6 was issued in March 1987, Supplement 7 ing future violations similar to those described n this publica- was issued in January 1988, and Supplement B vcas issued in tion. February 1989. The facility is located in Burke County, Georgia, approximately 26 miles south southeast of Augusta, Georgia, NUREG-1021 R05: OPERATOR LICENSING EXAMINER STAND- and on the Savannah River. This ninth supplement to NUREG-ARDS. SHIRAKl.C. Division of Licensee Performance & Quality 1137 provides recent information regarding resolution of condi-Evaluation (Post 870411). January 1989. 257pp. 8903030290. tional items following issuance of Supplement 8.

48721:245.

The Operator Licensing Examiner Standards provide policy NUREG-1230: COMPENDIUM OF ECCS RESEARCH FOR REAL-and guidance to NRC examiners and establish the procedures ISTIC LOCA ANALYSIS. Final Report.

• Division of Systems Re-and practices for uamining and licensing of applicants for NRC se6fch (Post 880717) December 1988 1,354pp. 8903030340.

operator licenses pursuant to Part 55 of Title 10 of the Code of 48710 041.

Federal Regulations (10CFR 55). They are intended to assist Emergency Core Cooling Systems (ECCS) are required on all NRC examiners and facilrty licensees to understand the exame light wator teactors {LWRs) in the United States to provide cool-nation process better and to provide for equitable and consist. ing of the reactor core in the event of a break in the reactor ent administration of examinations to all applicants by NRC ex- piping These accidents are called loss- of coolant accidents aminers These standards are not a substitute for the operator (LOCA), and they range from small leaks to a postulated full licensing regulations and are subject to revision or other internal break of the largest pipe in the reactor cooling system Federal operator examination licensing policy changes. As appropnate, government regulations require that calculations of the LOCA these standards will be revised penodically to accommodate be performed to show that the ECCS will maintain fuel rod clad-comments and reflect new information and exponence ding temperatures. cladding omdation, and hydrogen production

4 Main Citations and Abstracts Wthin certain hmits. The Nuclear Regulatory Comrnission (NRC) 50 is not a regulatory requirement for an LLW disposal facility, and others have cortgleted an extensive ir.vestigation of futil the criteria that were developed for 10 CFR Part 50 are basic to rod behavior and ECCS performance. The technology has been any OA program. The document specifically establishes OA

}

advanced to the point that it is now possible to make a realistic testirr. ate of ECCS performance dunng a LOCA and to quantdY guidance for the design, construction, and operation of those structures, systems, components, as well as, for site character-the uncertainty of this calculation. This report serves as a gen- Ization activities necessary to meet the performance objectives eral reference for ECCS research. The report (1) summarizes of 10 CFR Part 61 and to limit exposure to or release of radio-the understarWng of LOCA phenomena in 1974, (2) reviews ex- activity.

perimental and anah. cal programs developed to address the phenomena, (3) desc tbes best. estimate computer codes devel-NUREG-1319: A PRIORITIZATION OF RESEARCH ACTIVITIES.

Oped 4 l the NRC, (Q iiscusses the salient technical aspects of Plihn,J.vt Division of Regulatory Applications (Post LOCA pher~nena an0 our current understanding of them, (5) 870413) December 1988.106pp. 8902090061. 48440:045.

discusses probabilistic rNk studies and (6) examines the impact of research on the ECCS regulations. This report presents the results of the efforts in priorttizing 'he activities in the Office of Research. The purpose of this pnoriti-NUREG-1232 V02 S01: SAFETY EVALUATION REPORT ON Zation is to provide a basis on which to make management de-TENNESSEE VALLEY AUTHORITY: SEQUOYAH NUCLEAR cisions. The report describes also the methodology and criteria PERFORMANCE PLAN.Sequoyah Unit 1 Restart.

• Associate upon which the prionty rankings are based. Each activity was Nrector for Special Projects (Post 890101). January 1989. evaluated against four attributes. These attributes ars: safety 84pp. P'03080538. 48748:084. resurance, usefulness, appropriateness, and resources. This Th< 3ae Evaluation Report (SER) on the Sequoyah Nucle- report will be penodically upda'ed to include the prioritization of ar Performance Wn, NUREG-1232, Volume ?, was based on new activities, deletion of completed activities and to reflect the information sut mitted by the Tennessee Valley Authonty changes in budget a; locations and projections.

(TVA) it, its Segoyah Nuclear Performance Plan (SNPP),

through Revision 2, and on copporting documents. It was issued NUREG 1334: TECHNICAL SPECIFICATIONS FOR SOUTH on May 18,1988 by the U.S. Nuclear Regulatory Commission TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-staff for the restart of Sequoyah Unit 2. The SNPP addresses 4F.(Houston Ughting And Power Company)

• Pvision of Reac-the plant-specific concems requiring resolution beiore startup of either of the fequoyah unds. In particular, the SER addressed tor Projects - lit,lV,V & Special Projects (Post 870411). January 1989. 452pp. 8902080323. 48409:098.

required acticos for Unit 2 restart. In most cases, the program-matic aspocts for Unit 1 are identical to those for Unit 2. TVA The South Texas Project, Unit Nos.1 and 2, Technical Speci-described the differences in programs between Unit 1 and Unit fications were prepared by the U.S. Nuclear Regulatory Com-2 in Revision 3 of the SNPP. This was submitted by TVA in its mission to set forth the limits, operating conditions, and other letter dated May 9,1966. Where the Unit 1 program is different, requirements applicable to a nuclear reactor facility as set forth the staff % evaluation is provided in this supplement to the in Section 50.36 of 10 CFR Part 50 for the protection of the staff's SER in NUREG-1232 Volurr.e 2. On the basis of its health and sefety of the public.

review. % staff cont'ludes the Sequoyah-specific issues have r to the extent that would support the restart of NUREG-1335 DRFT FC- INDIVIDUAL PLANT EXAMINATION. SUBMITTAL GUIDANCE. Draft Report For Com-ment.

• Office of Nuclear Regulatory Research, Director (Post NURLG 975 V00 OPERATING EXPERIENCE FEEDBACK 860720).

• Office of Nuclear Reactor Regulation, Director (Post RE W TECHNICAL SPECIFICATIONS. Commercial Pown 870411). January 1989. 27pp. 8903060176. 48762:137.

Reactors. O'RE!LLY.P.D.; PLUMLEE.RL Division of Safetv Pro-Based on the Policy Staternent on Severe Reactor Accidents gra Post 870413). March 1989. 108pp. 890C 10441.

Regarding Future Designs and Existing Plants, the performrice This report documents the results of a trends and pattems of a plant examinahon is mqM h N hnse of end nuh analysis of operational exponence weh Technical Specifications ar power plant. The plant examination looks for severe accident at commercial nuclear power reactors in the U.S., pnmanly vulnerabilities and cost-effective safety improvemerrts that danng the period 1984-86. Limited use was made of preliminary would reduce or eliminate any discovered vulnerability. This results from 1987 data to provide additional insights regarding document delineates the guidance for reporting the results of a trends in Technical Specificatun violation rates. Major objec. plant examination.

tives of this report prepared by the NRC's Office for Analysis and Evaluation of Opera'ional Data are to: (1) identify and cata- NUREG 1336: INTERIM GUIDANCE ON THE STANDARD log technical specification- related licensee event reports FORMAT AND CONTENT OF FINANCIAL ASSUAANCE (LERs), (2) categonze and evatuste the events reported in MECHANISMS REOVRED FOR DECOMMISSIONING UNDER these LERs, (3) identify any issmis arising from the evaluation 10 CFR PARTS 30.40,AND 70.

• Division of Low Level Waste which appear to have generic safety significance, or wt ;h Management & Decommissioning (Post 870413). December relate to the on going Technical Specification improve n nt 1988,126pp. 8902080337. 48430:272.

Program, and (4) trerst thc results of the arialysis of th( Scta Interim Guidance on the Standard Format and Content of Fi-obtained in (1) thro >agh (3). nancial Assurance Mechanisms Required for Decommissioning under 10 CFR Parts 30, 40, and 70, NUREG-1336, discusses NUREG-1293: OUALITY ASSURANCE GUIDANCE FOP LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FAC11 W. Final the information to be provided in a license application and es-Aepot. PITTIGLIO,C.L. Division of Low Level Waste , mage- tablishes a uniform format for presenting the information re-ment & Decommissioning (Post 870413). January 1989. 28pp. quired to .neet the decommissioning licensing requirements.

8901230043.48177:340. The use of the Standard Format will (1) help ensure that the h.

This document provides guidance to an applicant on meeting cense application contains the information required by the regu-the Quality control (OC) requirements for a low-level waste lations, (2) aid the applicant in ensunng that the information is (LLW) disposal facility of 10 CFR 61.12. The OC requirements complete. (3) help persons reading the Standard Format to are the basis for developing of a quality assurance (OA) pro- locate information, and (4) contribute to shortening the time re-gram and the guidance provided horein. The entena developed quired for the review process. The Standard Format and Con-for this document are similar to the entena developed for 10 tent (NUREG-1336) ensures that the information required to CFR Part 50 Appendix B. Although Appendix B of 10 CFR Part perform the review is provided, and in a useable format.

MIin Citati:ns rmd AbstrCcts 5 NUREG-1337: INTERIM GUIDANCE ON THE STANDARD ciated with several alternatives considered for resolution of the REVIEW PLAN FOR THE REVIEW OF FINANCIAL ASSUH- Generic issue 99.

ANCE MECHANISMS FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,AND 70.

• Division of Low Level Waste NUREG-1343: TECHNICAL SPECIFICATIONS FOR VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 And 2. Docket Nos.

Management & Decommissioning (Post .870413). December 1988. 36pp. 8902080330. 48431:038. 50-424 And 50-425.(Georgia Power Company)

• Division of Re-actor Projects - 1/11 (Post 870411.). February 1989. 472pp.

interim Guidance on the Standard Review Plun (SRP) for the Review of Financial Assurance uechanisms for Decommission, 8903140124. 48852.t02.

Ing under 10 CFR Parts 30,40, t i 70 is prepared for the guid- The Vogtle Electric Generating Plant, Unit Nos.1 and 2, ance of Nuclear Regulatory Commission staff reviewers in per. Technical Specifications were prepared by the U.S. Nuclear forming rev4ws of apphcations from material licensees affected Regulatory Commission to set forth the hmits, operating condi-by the June 27,1980 (53FR24018). The pnncipal purpose of tions, and other requirements applicable to a nuclear facility as the SRP is to assure the quakty and uniformity of staff reviews set forth in Section 50.36 of 10 CFR 50 for the protection of the _

and to present a base from wtJch to evaluate the financial as- health and safety of the public.

surance aspects of the applications. NUREG-1337 identifies NUREG-1345: REVIEW OF EVENTS AT LARGE POOL-TYPE IR-who performs the review, the matters that are reviewed, the RADIATORS. TRAGEH,E.A. Division of Safety Programs (Post basis of the review, how the review is performed, and the con- 870413). March 1989. 64pp. 8904110450. 49272:097.

clusions that are sought. .Large pool-type gamma irradiators are 'used in applications such as the " cold" sterihzation of medical and pharmaceutical NUREG-1338: DRAFT PREAPPLICATION SAFETY EVALUATION REPORT FOR THE MODULAR HIGH-TEMPERATURE GAS- supplies, and recent changes in federri regulations make it pos-COOLED REACTOR. WILLIAMS,P.M.; KING,T.L.; WILSON J..'. sible they will be used extensively in the pmservation of food-Division of Regulatory Applications (Post 870413). March 1989. stuffs Becausu of this possible large increase in the use of irra-300pp. 8903280333. 49070:011. distors, the Office of Nuclear Materials Safsty and Safeguards This draft safety evaluation report (SER) presents the prelimi- was interested in knowing what events had occurred at irradia-nary results of a preapplication design review for the standard tors. The event data would be used as background in develop-modular high- temperature gas-cooled reactor (MHTGR) ing new regulations on irradiators. Therefore, AEOD began a (Project 672) The MHTGR conceptual design was submrtted by study of the opers'ing experience at large, wet-source storage the U.S. Department of Energv (DOE) in accordance w!!h the gamma irradiators. The scope of the study was to assess all U.S. Nuclear Regulatory Commission (NRC) " Statement of available operating information on targe (more than 250,000 Policy for the Regulation of Advanced Nuclear Power Plants" curie), pooleype trradiatois bcensed by both the NRC and the (51 FR 24G43), which provides for early Commissian teview and Agreement States, and events at foreign facilities. The study found thet abcut 0.12 events heve been reported per irradiator-interaction. The standard MHTGR consists of four identical re-actor modules, each with a thermal output of 350 MWt, coupled year. Most of these events were precursor events, in that there with two steam turbine-generator sets to produce a total plant was no evidence of damage to the radioactive sources or oeg-electrk:al output of 540 MWe. The reactors are helium cooled radation h the level of safety of the facility. Events with more and graphite moderated and utilize ceramically coated particle- significant impacts had a reported frequency of about 0.01 type nuclear fuel. The design includes passiv3 reactor-shutdown event per irradiator year. However, the actual rate of occurrence and decay. heat-removal features The staff and its contractors cf events of concern to the staff may be higher because there at the Oak Ridge National Laboretory and the Brookhaven Na. are few specific reporting requirements for events at irradiators.

tional Laboratory have reviews.d this design with emphasis on NUREG-1346: TECHNICAL SPECIFICATIONS FOR SOUTH those unique provisions in the design that accomplish the key TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-Safety functions of reactor shutdown, decay-hes' mmoval, and 499.(Houston Lighting And Power Companyt

• Division of Reac-containment of radio %ct,ve material. Final gu, dan e sn the ac' tor Projects - lil,lV,V & Special Projects (Post 870411). March ceptabihty of the MHTGR standard design is contnyent on re- 1989. 400pp. 8904110419. 49247:001.

ceipt and evaluation of additional information requested from The South Texar. Project. Unit Nos.1 and 2, Technical Speci-DOE pertaining to the adequacy of the containment desyn. fications were prepared by the U.S. Nuclear Regulatory Com-mission to set forth the hmits, operating conditions, and other WREG-1340: REGULATORY ANALYSIS FOR THE RESOLU-TON OF GENERIC ISSUE 99, " LOSS OF RHR CAPABillTY IN requirements applicable to a nuclear reactor faciley as set forth l- PWRS." SPANO,A.fl. Division of Safety issue Reso'utior! (Post in Section 50.36 of 10 CFR Part 50 for the protection of the 880717). February 1989. 785p. 8903080541. 48748.2m health and safety of the public.

Genenc issue 99 is concemed with the loss of Osidual heat NUREG-1350 VD?: NUCLEAR REGULATORY COMMISSION removal (RHR) capability in pressunzt'd water re rctors dunng 1939 5 ry , 7,0N DIGEST. MCGRATH,K.A. Division of cold-plant outage operations. The issue foc':ses on two nsk-sig. Bur $ 5 A .c (Post 890205). March 1989. 9Bpp.

nificant common cause failure modes Of the RHR system: (1) air 89tM210223. 48936:354.

binding of the RHR pumps during rc.bceo- enventoy op3ratior:3 The Nuclear Regulatory Commission 1989 Information Digest and (2) spunous closunr of 8he Rid suction valves due to mis

• provides summary information regarding the US NRC, its regula-apphcation of the autoclosure interlocks. Resolution of this tory responsibikties, and the areas keensed by the Commission.

issue involves consideration of the adequacy of plant capabili- This is the first of an annual puhhcation for the general use of ties for (1) preventing losses of EHR, (2) responding promptly the NRC staff and is aval:able to the public. The Digest is divid-and effectively to such challenges in order to prevent core ed into two parts: the first presents an overview of the Nuclear damage, and (3) ensu-ing timely containment protect.on against Regulatory Comm rion and the second provides data on com.

the release of radioactivity to the environment in the unkkely mercial nuclear reactor hcensees, as well as on worldwide com-evont of core damage due to loss of shutdown cooling. This en- mercial nuclear reactors.

tails examination of (1) relevant operational and accident re.

sponse procedures, (2) the instrurnentation available to the op- NUREG-1355: THE STATUS OF RECOMMENDATIONS OF THE erator for accident diagnosis and mitigation, and (3) administra- PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE ttve controls availatne for ensunng control room cognizance of MILE ISLAND.A Ten-Year Review,

• Ofc of the Executive Direc-ongoing maintenance activities that could potentially affect sta- tor for Operations. March 1989. 94pp. 8903280349. 49119:160.

bihty of the reactor coolant system. This regulatory analysis pro- This report summanzes the status of actions taken by the vides quantitative assessments of the costs and benefits asso- U.S. Nuclear Regulatory Commission ;NRC) in response to rec.

, 6 Miln Citatirns end Abstr: cts l

commendations made by the Presidential Commission on the Ac- The Advanced and Water Reactor Safety Research Programs cl dent at Three Mile Island in the 10 years since the accident Quarterly Progress Reports have been combined and are in-occurred in March 1979. It also updates NRC's initial response cluded in this report entitled, " Safety Research Programs Spon-to tra Presidential Commission's recommendations contained in sored by the Office of Nuclear Regulatory Research - Progress "NRC Views and Analysis of the Recommendations of the Report." This progress report wit! describe current activities and President's Commission of the Accident at Three Mile Island" technical progress in the programs at Brookhaven National Lab-(NUREG-0632), issued in 14ovember 1979. The status of ongo- oratory sponsored by the Division of Regulatory Applications, ing initiatives for actions not yet complete are also reported for Division of Engineenng. Division of Reactor Accident Analysis, reference purposes. On the basis of its analysis of NRC and the and Division of Reacar and Plant Systems of the U.S. Nuclear industry, the Pr%idential Commission found meny then-current Regulatory Commission, Office of Nuctsar Regulatory Research practices inadeq.9te and in need of improvement. As a result of following the reorganization in February 1987, its recommendat.uns and of guidance from other studies, sub-stant.al changes have been made in the 10 years since the ac- NUREG/CRQ331 V08 N3: SAFETY RESEARCH PROGRAMS cident. This report reflects how, based on Presidential Commis- SPONSORED BY OFFICE OF NUCLEAR REGULATORY Bion recommendations and continued work, revised practices RESEARCH.Pr09ress Report July September 1988. WEISS.A.J.

and standards are now being implemented by NRC and Brookhaven National Laboratory. February 1989. 200pp.

throughout the industry. 8903090070. BNL-NUREG 51454. 48757:234.

This progress report describes,cu' rent activities and technical NUREG-1359: TECHNICAL SPECIFICATIONS FOR VOGTLE progress in the programs at Brookhaven National Laboratory ELECTRIC GENERATING PLANT, UNITS 1 AND 2. Docket Nos. cponsored by the Dwision of Regulatory Applications, Division 50-424 And 50-425 f3eorgia Power Company,et al)

• Division of of Engineenng, Division of Safety issue Resolution, and Division Reactor Projects - 1/Il (Post 870411). March 1989. 400pp. of Systems Research of the U S. Nucleat Regulatory Commis-6904110511,49249.130. sion, Office of Nuclear Regulatory Research followmg the reor-The Vogtle Electric Generating Pinnt, Unit Nos.1 and 2, ganization in July 1988. The previous reports have covered the Technical Specifications were prepared by the U.S. Nuclear penod October 1,1976 through June 30,1988.

Regulatory Commission to set forth the limits, operating condi-tinns, and other requirements applicable to a nuclear facility as NUREG/CR-3006: DAMPING IN BUILDING STRUCTURES set forth in Section 50.36 of 10 CFR 50 for the protection of the DURING EARTHQUAKES. Test Data And Modelin0 health and safety of the public. COATS,0.W. Lawrence Livermore National Laboratory. January 1989. 262pp. 8902090079. UCRL 53043. 48440:151.

NUREG/CR-2000 V07N12: LICENSEE EVENT REPORT (LER) A review and evaluation of the stete-of-the-art of damping in COMPILATION.For Month Of December 1988.

• Oak Ridge Na- building structures during earthquakes is presented. The primary tional Laboratory. January 1989. 92pp. 8902100080. ORNL/ emphasis is in the following areas: (1) the evaluation of com-NSIC-200. 48445:006, monly used mathematical techniques for incorporating damping This monthly report contains Licensee Event Report (LER) effects in both simple and complex systems, (2) a compilation operational information that was processed into the LER data and interpretation of damping test data, (3) an evaluation of file of the Nuclear Safety information Center (NSIC) dunng the structure testirg methods and building instrumentation prac-ona month penod identifeed on the cover of the document. The tices, and (4) an investigation of rigid- body rotation effects on LERs. from which this information is denved, are submitted to damping values from fsst data. A literature review provided the the Nuclear Regulato'y r Commission (NRC) by nuclear power buas for evaluating mathematical techniques used to incorpo-plant licensees in accordance with federal regulations. Proce- rate earthquake irtuced damping effects in simple and cornplex dures for LER reponng for revisions to those events occurnng systems SirrNe mathematical models for viscous, hysteretic, pnot to 1964 are desenbod in NRC Regulatory Guide 1.16 and and Coulomb damping are presented. Test data from a wide NUREG 1061, "Instractnns for Preparation of Data Erdry range of sources, has been compiled ano interpreted for build-Sheets for Licensee Event Reports? For those events occurring ings, nuclear power plant htructures, piping, equipm69t, and iso-on and after January 4,1984, LERs are being submitted in ac- lated structurat elements. Test methods used to determme cordance with the revised rde contained in Title 10 Part 50.73 damping and frequency parameters are discussed. A discussion of the Codn of Feveral Regulations (10 CFR 50.73 - Licensne of identification techniques typically used to determine building Event Report System) which was publisi.ed in the Federal Reg- parameters (frequency and damping) from strong motion ister (Vol. 48, No.14 j on July 26,196% NUREG 1022, "Li- records is included.

consee Event Report System Desenp son of Systems and Guidelines for Reporting, provides suppc,hng guidance and in- NUREG/CR-3037: USER'S MANUAL FOR FIRIN.A Computer formation on the revised LER rule The ! ER summanes in this Code To Estimate Accidental Fire And 9adioactive Airborne Re-report are arranged alphabetically by is tility name and then lease in Nuclear Fuel Cycle Facitaties. CHAN.M.K.;

chronologically by event date for eact facility. Component, BALLINGER,M.Y.: OWCZARSKI,P.C. Battelle Momonal institute, system, keyword, and component vende r indexes follow the Pacific Northwest Laboratory. February 1989. 212pp.

summanes. Vendors are those identstad iy the utility when the 89 2%201 %M EMM r LER for - is initiated, the keywords for tt e component, system The U.S. Nuclear Regulatory Commission has sponsored a l

and g% oral keyword indexes are assigned by the computer msearch program to bettcr estimate the source term, or amount using correlation tables from the Sequence Coding and Search and charactenstics cf radioactive matenal made airborne, from System. several types of accidents. The most concentrated work has been done on fires, reeulting in the development of FIRIN, a NUREG/CR4000 V08 N1: LICENSEE EVENT REPORT (LER) compartm.11 fire code that estimates the release and distribu.

COMPILA1 ION For Month Of January 1989.

• Oak Ridge Na- tion of radioectwo matenals within a room from fires involving [

tional Laboratory. February 198L 130pp. 8903140098. ORNL/ radioactive matenals. The User's Manual describes the techni-NSIC-200. 48851'332. cat basis of the code and provides input and output inforrr.ation See NUREG/CR-2000307,N12 abstract for code users.

NUREG/CR-2331 V07 N4: SAFETY RESEARCH PROGRAMS NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF SPONSORED BY OFFICE OF NUCLEAR REGULATORY THE WESTERN OHIO-INDIANA REGION. Annual RESEARCH. Progress Report. October-December 1987. Report. October 1987 - September 1988 SCHWARTZ,S.Y.;

WEISS A.J. Brookhaven National Laboratory. September 1988 LOY,T.; YOUNG.C.J. Michigan. Univ. of, Ann Arbor, Mt. Decem-135pp.8903090103. BNL-NUREG-51454. 48769:001. bor 1988. 68pp. 8904110431. 49221:147.

Miln Cit:tions and Abstricts 7 Earthquake activity in the Western Ohio-Indiana region was Knowledge of the return perice of large floods is required to

' monitored with a precision seismograph network consisting of make risk analyses fr nuclear rower plants subject to flooding nine statiors located in west- central Ohio and four stations lo- trom rivers. The system rsported here combined the stochastic cated in Indiana. No local earthquakes were recorded during simulation of hourly rainfall data and daily pan evaporation data this report period. The low level of local seismicity in the last 2 with the deterministic simulation of streamflow by using the syn-years suggests that the occurrence of the m(b) = 4.5 earth- thetic rcinfall and evaporation data as Mput to a calibrated rain-quake, in St. Mary's, Ohio on June 12,1986, releastd most of fan-runoff model. The sequence of annual maximum flood the crustal strain accumulated. Four regionat events were well peaks from a synthetic record of 10,000 years or more was recorded by the array stations during this year. Their magni- then analyzed to obtain estimates of flood frequency, The rea-iudes range from m(b!g) - 3.2-4.5. The largest of these events sonableness of the flood frequency results must be evaluated (September 7,1988 in northeastern Kentucky, m(b) - 4.5) had niinor damege toported. Travel time data from regional earth- on the degree of mimicry of the key characteristics of the ob-quakes were analyted to gas insight into the local velocity served rainfall data and the abihty of the rainfall-runoff model to structure. P(n) velocities computed using travel times from three mimic the observed flood frequency dunng the calibration events in northeastern Ohio and four events from southern lik- eriod. On this basis, the flood frequency results appeared to nois and southeastern Missouri reveal comparable values of PO 9 8.41 km/s. This value is consistent regardless of the direction model parameters. There is a need to develop regional param-of the arrival, requring httle or no nortfrasterly or southwesterly eters for the stochastic models and to conduct research on the dip to the Moho beneath t'ie stations. The P(n) velocity ob. relationship between the stochastic structure of ra% fall and sto-tained is comparable to values computed in previous studies chastic structure of flood frequoney. The methodology is appli-from earthquakes in those regions. cable, assuming a highly skilled analyst, to watersheds s milar to tnose already tested.

NUREG/CR-4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT NUREG/CR 4554 V01: SCANS (SHIPPING CASK ANALYSIS WATER REACTORS. Semiannual Report. April-September 1687. SYSTEM):A MICROCOMPUTER BASED ANALYSIS SYSTEM DOCTOR,S.R.; DEFFENBAUGH,J.; GOOD.M.S.; et al. Battelle FOR SHIPPING CASK DESIGN REVIEW. Volume 1: User's Memorial Institute, Pacific Northwest Laboretory. November Manual To Version la (including Program Reference) 1988.104pp. 6901230058. PNL-5711. 48180:130. GERHARD,M.A.; TRUMMER,D.J.; JOHNSON,G.L.; et al. Law-Evaluation and improvement of NDE Reliabihty for Inservice rence Livermore National Laboratory. February 1989. 168pp.

Inspection of Light Water Reactors (NDE Reliability) Program at 8904110484. UCID-20674. 49271:265.

the Pacific Northwest Laboratory was estabhshed by the Nucle- SCANS (Shipping Cask Analysis System)is a microcomputer-ar Regulatory Commissbn to determine the rehability of current based system of computer programs and databases developed inservice inspection OSI) techniques and to develop recommen-at the Lawrence Livermore National Laboratory (LLNL) for eval-dations that will ensure a suitably high inspect On reliability. The objectives of this program include determining the reliability of uating safety analysis reports on spent fuel shipping casks.

ISI Performed on the prir iary systems of commercial light-water SCANS is an easy-to-use system that calculates the global re-reactors (LWRs); using probabihstic fracture mechanics analysis s onse to impact loads, pressure loads and tnermal conditions, to determine the impact of NDE unreliability on system safety; providing reviewers with an independent check on anatyses and evaluating reliabikty improvements that can be achieved submitted by licensees. SCANS is based on microcomputers wdh improved and advanced technology. A final objective is to compatible with the IBM-PC family of computers. The system is formu' ate recommended revisions to ASME Code and Regula. composed of a series of menus, input programs, cask analysis tory requirements, based on material proporties, service condi. programs, and output display programs. All data is entered tions, and NDE uncertainties. The program scope is limited to through fill-in- the-blank input screens that contain descriptive ISI of the primary systems inch. ding the piping, vessel, and data requests. Analysis options are based on regulatory cases other inspected components. This is a progress report covering described in the Code of Federal Regulations (1983) and Regu-the programmatic work from April 1987 through September latory Guides pubhshed by the U.S. Nuclear Regulatory Com-1987. mission in 1977 and 1978.

NUREG/CR-4478: UPDATE - A FORTRAN 77 SOURCE FILE W97 4/CR-4554 V02: SCANS (SHIPPING CASK ANALYSIS MANIPULATOR. Adapted For The Data General MV Senes i v.EM):A MICROCOMPUTER BASED ANALYSIS SYSTEM Eclipse Computers Under AOS/VS. KIRK,B.L. Oak Ridge Na- FOR SHIPPING CASK DESIGN REVIEW. Volume 2: Theory tional Laboratory. January 1989. 88pp. 8903080544. ORNL/ Manual - Impact Analysis. CHUN.R.C.; TRUfL ER,D.J.;

TDMC-4. 48748:172- NELSON,T.A. Lawrence Livermore National Laboratory. Febru-The primary purpose of UPDATE, which onginated at Law' ary 1989. 52pp. 8904110406. UCID-20674. 49223:176.

rence Derketoy Laboratory, is to provide a capabihty for main. The Lawrence 'Livermore National Laboratory, under contract taining source files. This portable version, which has the same to the U.S. Nuclear Regulatory Commission's Office of Nuclear name and function as the original CDC utihty program, serves as a file manager and can be a valuable tool in computer code Material Safety and Safeguards, has developed a system of development work. Wntion in FORTRAN 77, its general function computer codes called the Shipping Cask Analysis System In to manipulate source files by allowing the user to modify the (SCANS) to perform keensing analyses for shipping containers.

source code through a series of commands called directives' g

The source file with UPDATE directives is initial'y copied to an- Pcirts: IMPASC (Impact AnalY sis of Shippin9 Containers), and other file called a program library. Changes are then maoe to OUASC (OUasi-static Analysis of Shipping Containers). It com-the program hbrary. From this program hbrary is extracted the bines both cynamic and quass-static computer analyses and is

' compile ready

• file, that is, the final input to a FORTRAN 77 operational on the IBM PC and compatible computars. This compiler. report presents a discussion of the theoretical background behind IMPASC and desenbes its approach to handkng large NUREG/CR 4496 VD2: A SYSTEM FOR GENERATING LONG ngid body motions that are common rn oblique angie drops.

STREAMFLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN PERIOD. Phase 11. FRANZ,D.D.; NUREG/CR-4554 V03: SCANS (SHIPPING CASK ANALYSIS KRAEGER,B.A.; LINSLEY,R.K. Linsley, Kraeger & Associates, SYSTEM):A MICROCOMPUTER BASED ANALYSIS SYSTEM Ltd. February 1989. 222pp. 8903280351. 49069:099 I

i

8 Miln Citttirns cnd AbstrtCts  !

FOR SHIPPING CASK DESIGN REVIEW. Volume 3: Tt eory standing and guidance to help improve the application of statis-Manua! - Lead Slump in impact Analysis And Verification Of tical methods to nuclear material management problems.

Impact Analysis. CHUN.R.C. LO,T.; MOK,G.C.; et al. Lawrence Livermore National Laboratory. February 1989. 102pp. NUREG/CR-4618: EVALUATION OF RELIABILITY TECHNOLO-8904110395. UCID 20674. 49223:077. GY APPLICABLE TO LWR OPERATIONAL SAFETY.

A computer system called SCANS (Shipping Cask Analysis AZARM,M.A. Brookhaven National Laboratory. LOFGREN,E.V.

System) has been developed for the staff of the U.S. Nuclear Science Applications International Corp. (formerly Science Ap-Regulatory Commission to perform confirmatory licensing review plications, Inc.). August 1988. 200pp. 8904030084. BNL-arialyses. The SCANS can handle problems associated with NUREG-51995. 49171:216.  ;

impact, heat transfer, thermal stress, ard pressure. A new This report describes the technical work performed at Brook.

methodology is developed for SCANS to analy70 the lead slump haven National Laboratory under the auspices of the Nuclear behavior of lead shielded casks during a postulated impact with Regulatory Commission to identify the task and to evaluate the an unyielding surface. The methodology is an expansion of the technology for an operational reliabihty program applicable to existing lumped-parameter impact analysis method. It is as- the safety of operating nuclear power plants. This technical sumed that the leao and the steel cylinders are not bonded as work was performed through a many fauted approach to ac-opposed to the existing bonded-lead assumption. The lead and complish the following: (1) define a reliability structure and tasks the steel cylinders are treated as an elastic medium and thin to help maintain a low core-melt frequency and to facilitate im-elastic shells, respectively. The shielding reduction, or the per. plementation of performed-based regulation. (2) charactuize in-manent lead slump at the opposite end of impact, is obtained dustry application of reliability technology for improving oper-by converting the elastic strain energy of the lead into plastic ations and safety, (3) determine the impact that reliability pro-deformation The interface pressure between the lead and the grams could potentially have on resolving safety issues, (4) steel, the hoop stress in the steel shells, and the reduction in evaluate the use of reliability technology of the kind that would shielding are among items that can be calculated. The adequa- be applicable for a reliability program on a nuclear power plant cy of this lead stump methodology is established by companng system, and (5) explore the feasibility of techniques for monitor-results with those obtained from rigorous finite element analy. ing plant safety performance as a potential aspect of perform-ses and from cask impact tests. ance-based regulation. This technical evaluation of a reliability technology was prepared to support resolution of Generic Issue NUREG/CR-4554 V04: SCANS (SHIPPING CASK ANALYSIS II.C.4, " Reliability Engineering."

SYSTEM);A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN REVIEW. Volume 4: Theory NUREG/CR-4627 R01: GENERIC COST ESTIMATES. Abstracts Manual - Thermal Analysis. JOHNSON G.L.; MAPIRO,A.B. From Generic Studies For Use in Preparing Regulatory impact Lawrence Livermore National Laboratory. February 1989.67pp. Analyses. CLAIBORNE,E.; SCIACCA,F.; SIMION.G.; et at Sci.

8904110465. UCID 20674. 49238:330. ence & Engineenng Associates, Inc. February 1989. 194pp.

TOPAZ is the two dimensional, implicit, finite-element comput. 8902100184. 48442:272.

er code included in the SCANS cask analysis system for heat The Nuclear Regulatory Commission has sponsored a conduction calculations. TOPAZ, a code developed on LLNL number of generic cost estimating studies. These studies were mainframes, has been implemented on IBM PC computers. This prepared to aid NRC analysts in propanng Regulatory impact report provides documentat;on of TOPAZ controls and variables Analyses (NA's). These generic studies provide cost estimates and a desenption of the numerical algorithms used. Sample that would have wide application to a large number of Regula-problems with anaiytical solutions are presented. tory Analyses being performed throughout the NRC and deal pnmarily with repair and modification activities that may be im.

NUREG/CR 4554 V05: SCANS (SHIPPING CASK ANALYSIS posed on nuclear plants as a result of regulatory actions. Ab-SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM stracts of each of the generic cost estimating studies have been FOR SHIPPING CASK DESIGN REVIEW. Volume 5: Theory prepared and assembled in this catalog. These abstracts Manual - Thermal / Pressure Stress Analysis. TRUMMER,D.J.; present t*) results of the more detailed studies in a compact, GERHARD,M.A. Lawrence Livetmore National Laboratory. Feb- easily Werstood and readily useable format. Individual ab-ruary 1989. 34pp. 8904110456. UCID 20674. 49238.292. stracts have been developed to treat the main-line topics of the This report describes the SCANS modules for thermal and generic studies. In addition, abstracts have been prepared cov- 1 pressure stress analyses and includes validation and bench- ering important sub-topics or " stand-alones" which are of broad mark examples. SCANS (Shipping Cask Analysis System) is a microcomputer-based system of computer programs and data-interest in RIA preparation. This abstract catalog updates and revises information presented in NUREG/CR 4627. The catalog

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bases developed at the Lawrence Livermore National Laborato- will be expanded and modified as additional genenc cost stad-ry (LLNL) for evaluating safety analysis reports on spent fuel ies are completed and as abstracts are modified to reflect up-shipping casks. Thermal and pressure stress cases are based dated conditions. l on regulatory cases desenbod in the Code of Federal Regula-tions (1983) and Regulator / Guides published by the U.S. Nu- NUREG/CR-4639 V5P2R1: NUCLEAR COMPUTERIZED Ll-clear Regulatory Commission (1977,1978). BRARY FOR AS3ESSING REACTOR RELIABILITY i (NUCLARR). Data Manual.Part 2: Human Error Probability (HEP)

NUREG/CR 4604: STATISTICAL METHODS FOR NUCLEAR MA- Estimates. GERTMAN D.I.; GILBERT,B.G.; GILMORE,W.E.; et TERIAL MANAGEMENT. BOWEN.W M.; BENNETT,C.A. Battelle al. EG&G Idaho, Inc. (subs. of EG&G, Inc.). January 1989.

Memorial instiNte, Pacific Northwest Laboratory. December 482pp. 8903090033. EGG-2458. 48772:238.

1988.1,020pp. 8903210228. PNL 5849. 48934:062. This volume of a five-volume series summanzes those data i This book is intended as a reference manual of statistical currently resident in the first release of the Nuclear Computer- l l methodology for nuclear matenal management practitioners. It ized Library for Assessing Reactor Reliability (NUCLARR) data l desenbes statistical methods currently or potentially important in base. The raw human error probability (HEP) and hardware ,

nuclear matenal management, explains the choice of methods component failure data (HCFD) contained herein are accompa- )

for specific applications, and provides examples of practical ap- nied by a glossary of terms and the HEP and hardware taxono- >

plications to nuclear material management problems. Together mies used to structure the data. Instructions are presented on with the accompanying training manual, which contains fully how the user may navigate through the NUCLARR data man-worked out problems keyed to each chapter, this book can 3?so agement system to find anchor values to assist in solving nsk-be used as a textbook for courses in statistical methods for nu- related problems. Volume V: Data Manual will be updated on a j c ear matenal management, it should provide increased under- penode basis so that nsk analysts without access to a comput- l l

l

Miln CitIti:ms cnd Ab;tr2 Cts 9 l

or may have access to the latest NUCLARR data. Those users This report summarizes a probate,isac reliability evaluation of '

wishing to learn more regarding the computer-based interactive BWR reactor coolant piping performed for the U.S. Nuclear search and report-generation capabilities of the NUCLARR Regulatory Cornmission (NRC) by the Lawrence Livermore Na-system are referred to the other volumes in the NUREG/CR- tsonal Laboratory (LLNL). In this evaluation, LLNL estimated the 4639 series, e.g., Volume 1: Summary Description or Volume IV: probability of a double-ended guillotine break (DEGB) in the User's Guide.

main steam, feedwater, and recirculation loop piping of a repre-NUREG/CR-4639 V5P3R1: NUCLEAR COMPUTERIZED Ll. sentative Mark I BWR plant. Two causes of pipe break were BRARY FOR ASSESSING REACTOR RELIABILITY considered: crack growth at welded joints, and the earthquake-(NUCLARR). Data Manual.Part 3: Hardware Component Failure induced failure of supports for piping and components. A prob-Data (HCFD). GERTMAN,D.I.; GILBERT B.G.; GILMORE,W.E.; abilistic fracture mechanics model, including intergranular stress i et al. EG&G Idaho, Inc. (subs. of EG&G, Inc.). January 1989. corrosion cracking (IGSCC) in Types 304 and 316NG stainless 635pp. 8903090022. EGG-2458. 48769:136. steels, was used to estimate the probability of crack-induced See NUREG/CR 4639,V05,P02,R01 abstract. pipe break. The probability of pipe break indirectly caused by NUREG/CR-4639 V5P4Rt: NUCLEAR COMPUTERIZED Ll- suppott failure was estimated by applying reliability techniques BRARY FOR ASSESSING REACTOR RELIABILITY to supports for " heavy components," such as the reactor pres-(NUCLARR). Data Manual.Part 4: Summary Aggregations. sure vessel, as well as to conventional pipe supports such as GERTMAN.D.L; GILBERT,B.G.; GILMORE,W.E.; et al. EG&C spring hangers and snubbers. Our probabilistic fracture rnechan-Idaho, Inc. (subs, of EG&G, Inc.). January 1989. 340pp. ics evaluation found that the probability of crack-induced DEGB 8P03090032. EGG-2458. 48771:257. in main steam, feedwater and, H IGSCC is nc* a tactor, recircu-See NUREG/CR 4639,V05,P02,R01 abstret. lation piping is very low, in IGSCC susceptible Type 304SS piping. stress corrosion dominates the probability of DEGB due NUREG/CR-4667 VDS: ENVIRONMENT ALLY ASSISTED CRACK.

ING IN LIGH1 WATER REACTORS. Semiannual Rep;, April- mainly to cracks that initiate during the first few years of plant September 1987. SHACK,W.J.; KASSNER,*f.F.; MAlYA.P.S.; et life; replacing Type 304 piping with IGSCC-resistant Type al. Argonne Natonal Lab ^ratory. February 1989. 72pp. 316NG lowers DEGB probabilities by several orders of magni-8903280340. ANL-88-32. 49070:331. tude. We also found that the probability of pipe break caused by This report sumrnanzes work performed by Argonne National seismical!y-induced support faifure is low regardless of whether Laboratory on environmentally assisted cracking in light water " heavy component" supports or conventional pipe supports are reactors dunng the six months from April to September 1987. being considered.

The stress . corrosion cracking (SCC) behavior of Types 304, 316NG, and 347 stainless steel (SS) was investigated by means NUREG/CR-4792 V02: PROBABILITY OF FAILURE IN BWR RE-of slow-strain-rate and fracture-mechanics crack-growth rate ACTOR COOLANT PIPING. Pipe Failure Induced By Crack tests in high-temperature water. The relative susceptibility of the Growth And Failure Of Intermediate Supports. LO,T.;

solution-annealad Type 304 SS to crack initiation was deter- BUMPUS S.E.; CHINN,D.J.; et al. Lawrence Livermore National mined in water with two dissolved- oxygen concentrations. The Laboratory. March 1989. 200pp. 8903280308. UCID-20Si4.

effect of dissolved copper and urganic impurities on the SCC of 49072:187.

sensitized Type 304 SS was also investigated. Fatigue tests are The U.S. Nuclear Regulatory Commission (NRC) contracted being conducted on Type 316NG SS in air at room temperature with the Lawrence Livermore National Laboratory (LLNL) to to provide baseline data for comparison with results tha' will be conduct a study to determine if the probability of occurrence of obtained in high temperature water. The susceptibility of sever- a double +,tded guillotine break (DEGB) in tne major coolant al heats of different grades of low-alloy steel to transgranular piping systems of nuclear power plants is large enough to war-SCC was explored in slow-strain rate tests at 289 degrees C, rant the current stringent (ssign requirements of designing and the vanation in the crack growth rates was attributed to dif- against the postulated e>fects of a DEGB. The study includes ferences in the sulfur content and sulfide includion distr'butions both the PWR (Pressurized Water Reactor) and the BWR (Boil-in the materials.

ing Water Reactor) plants in the United States. Earlier efforts NUREG/CR-4780 V02: PROCEDUHES FOR TREATING concentrated on the reactor coolant loops of PWR (nnts, and COMMON CAUSE fan.URES IN SAFETY AND RELIABILITY the results indicated that the DEGB probability in thekiilactor STUDIES.Anatytical Background And Techniques. MOSLEH,A.; coolant loops (RCLs) was very small. Following the study of the FLEMING.K.; PARRY,Ga et al. Packard, Lowo & Garrick, Inc. PWR plants, a study of BWR reactor coolant piping was per.

January 1989.132pp. 8902090032. EPRI NP-5613. 48441:081. formed. The Brunswick Steam Electric Plant at Southport, North This report presents a framework for the inclusion of the Carolina wat selected as the pilot plant for the BWR evaluation.

impact of common cause failures in risk and reliability evalua- The probability of pipe failure in three major coolant pipings was tions. Common cause failures are defined as that subset of de- assessed: the recirculation loops, the pnmary steam lines, and pendent failures for which causes are not explicitly included in the main feedwater lines, in the case of recirculation loops, both the logic model as basic events. The emphasis hera is on pro- the existing and a proposed replacement system were studied.

viding procedures for a practical, systematic approach that can A probabilistic fracture mechanics approach was used in this be used to perform and clearly document the ana'ysis. The t

study to estimate the crack growth and to assess the crack sta-l framework compnses four major stages: (1) system logic modei

' bility in the piping systems throughout the lifetime of the plant.

development, (2) identification of common cause component groups, (3) common cause modeling and data analysis, and (4)

The effects of the failure of intermediate pipe supports were system quantification and interpretation of results. The frame- also examined. The results of the assessment indicated that the work and the methods discussed for performing the clfferent i robability of occurrence of DEGB due to crack growth and in-stages of the analysis integrate insights obtained from engineer- stabihty is small if the problem of intergranular St.ess corrosion ing assessments of the system and the historical evidence from crackin9 (lGSCC) is resolved by the use of the replacement multiple failure events into a systematic, reproducible, and de. system. The study of inte meFate support failure yielded some fensible analysis. gu:delines for significant reduction of effort in assessing the ef-fects of seismically induced failure of intermediate supports.

NUREG/CR-4792 VD1: PROBABILITY OF FAILUh6 IN BWR RE-ACTOR COOLANT PIPING. Summary Report. HOLM AN,G.S.; NUREG/CR-4835: COMPARISON AND APPLICATION OF QUAN-CHOU.C.K. 8.awrence Livermore National Laboratory. March 1989.134pp. 8903280347. UCID-20914. 49068:327.

10 MIin Cit:tisns End Ab:tr: cts TITATIVE HUMAN RELIABILITY ANALYSIS METHODS FOR Before a license can be obtained to construct a facihty for the THE RISK METHODS INTEGRATION AND EVA1 UATION PRO- shallow- land burial of low-level v.aste, the U.S. Nuclear Regula-GRAM (RMIEP).Fina: Report. HANEY,L.N.; BLACKMAN,H.S. tory Commission must be assured that the facilities will meet EG&G Idaho, Inc. (suos. of EG&G, Inc.). BELL,B.J.; et al. Bat- both performance objectives and prescriptive requirements set telle Memonal Institute, Columbus Laboratones. January 1989. forth in 10CFR61, " Licensing Requirements for Land Disposal 197pp. 8902080316. EGG-P485. 48408:261. of Radioactive Waste." Subpart D. Section 61.50(a)(2) of This report documents the human reliabihty analysis (HRA) $*EFR61 states that a "oisposal site shall be capable of being portion of the Risk Methods Integration and Evaluation Program r%racteind, modeled analyzed and monitored." In order to (RMIEP) sponsored by the U.S Nuclear Regulatory Commis- test the concopt o' " site modelability," a 30-year old low-level sion. A literature search identified 20 HRA methods for consid- radicoctive waste disposal site at Chalk Rrn.r Nuclear Laoorato-eration. Twelve methods were evaluated as appropriate for use ries (CRNL). Canada, was used for evaluating the process of in probabilistic risk assessment (PRA) for nuclear power plants site characterization and the subsequent modeling predictions (NPPs) by using a cnteria set develcped for that purpose. Data of radionuclides transport from the site by groundwater. This were collected at a commercial NPP witn operators responding evaluation was performed by comparing the actual measured to walkthroughs of four severe accident scenarios. An HRA sys- radionuclides migratiori with predicted migration estbated from tems analysis was also performed for the plant. The data were hydrologic /radionaclide transport models. This comparison has used in the application and quantitative comparison of selected provided valuable insights into the soplicability of transport mod-HIIA methods. Possible reasorts for observed quantitative differ- eling, and in determining what level of effort is needed in site ences are discussed. Qualitative compansons of the 12 meth- characterization at locations similar to the "A" Disposal Area to ods are provided via attribute descriptions and utility ratings in provide the ossired degree of predictive capabihties. The trans-an HRA selection tool designed to help analysis select appropri- port modeling using the limited site characterization data set sie HRA methods based on their goals and available resources. was in reasonable agreement with the actual radionuclides mi-An assessment of ithe current state of the art of HRA for PRA is gration, thus giving confidence ir the use of site characterize-piesented as well as recommendations for the future use of tion data for predicting radionuclides migration.

HRA in PRA.

NUREG/CR-4940: TECHNICAL FINDINGS RELATED TO GE-NUREG/CR-4849 V0t STEAM GENERM OR GROUP NERIC ISSUE 23. REACTOR COOLANT PUMP SEAL FAILURE.

PROJECT. Task 9 Final Report: Nondestructive Evaluation RUGER,C.J.; LUCKAS,W.J. Brookhaven National Laboratory.

Round Robin. Volume 1: Desenption And Summary Data- March 1989. 122pp. 8904110522. BNL-NUREG-52144.

DOCTOR,P.G.; HARTY,H.; FERRIS R.H.; et al. Battelle Memori-49239.235.

al Institute, Pacific Northwest Laboratory. January 1989. 350pp- Reactor coolan', pumps contain mechanical seals to limit the 8903280277. PNL 5868. 49071:110. leakage of pressurized coolant from the reactor coolant system The Steam Generator Group Project (SGGP) is using the re- to the containment. These seals have the potential to leak, and tired-from- service Surry 2A pressurized water reactor steam a few .have degraded and even failed resulting in a small break generator as a test bed to investigate the reliability and effec- loss of coo l ant accident (LOCA). As a result, " Reactor Coolant tiveness of inservice nondestructive eddy current inspection Pump Seal Failure," Generic issue 23, was established. This equipment and procedures. The information developed will pro- report summanzes the findings of a technical investigation gen-vide tl.e technical basis for updating the Regulatory Guides gov- ersted as part of the program to resolve this issue. These tech-erning inservice inspection and tube plugging critona of steam nical findings address the vanous fact-finding issue tasks devel-generators. This report describes Task 9 of the multi-task oped for the action plant associated with the genenc issue, project. The objective of Task e was to plan, perform and ana- namely background information on seal failure, evaluation of lyze results of four round robin nondestructive examinations on seal cooling, and mechanical- and maintenance-induced failure a subset of tubes from the Surry generator. A descnption of the mechanisms.

obNtives, conduct and analysis of results for each round robin is presented. Validation of inspection results will be provided by NUREG/CR-4b0 V03: THE SHORF.LINE ENVIRONMENT AT-removal of specimens from the generator for destructive and MOSPHERIC DISPERSION EXPEnlMENT (SEADEX). Airborne out of. generator charactenzation (in progress). Preliminary com- LIDAR Data. JOHNSON,W.B.; CANTRELL,B.K.; MORLEY,B.M.;

pansons of the inspection results in terms of agreement among et al. SRI International. December 1988. 223pp. 8902080443.

teams of defect detection and sizing are given for each round 48406:044.

robin. The majority of indications were reported at the hot leg The SEADEX atmospheric dispersion field study was conduct-top of the tube sheet. The best results on detection agreement ed during the penod May 29 June 8,1982, in northeastern ranged between 70 and 90 percent, bu' reported defect sizes Wisconsin, in the vicinity of the Kewaunee Powei Plant on the vaned significantly between teams for all round robins. The vari- western shore of Lake Michigan. The specific objectives of ability in detection and sizing appears to be due to analyst inter- SEADEX were to charactenze (1) the atmospheric dispersion pretation of the complex eddy current signals rather than differ- and (2) the meteorological conditions influerscing this dispersion ences in inspection equipment. as completely as possible during the test penod. This field study NUREG/CR-4849 V02: STEAM GENERATOR GROUP included a series of controlled tracer tests utilizing state of-the.

PROJECT. Task 9 Final Report: Nondestructive Evaluation art tracer measurement technology to determine honzontal and Round Robin. Volume 2: Raw Inspection Data. DOCTOR,P.G.; vertical dispersion over both land and water. Extensive meteor-HARTY,H ; FERRIS.R H.; et al. Battelle Memonal Institute, Pa- ological measurements were obtained to thoroughly character-cific Northwest Laboratory. March 1989. 710pp. 8903280281. ize the three dimensional structure of the atmosphenc boundary PNL 5868. 49067:001. layer controlling the dispersion process. This volume presents See NUREG/CR 4849,V01 abstract. the airborne LIDAR data collected dunng the field study.

NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE NUREG/CR-5030: AN ASSESSMENT OF STEAM-EXPLOSION.

MODELING OF A LOW LEVEL WASTE SHALLOW-LAND INDUCf.D CONTAINMENT FAILURE. THEOFANOUS,T.G.;

BURIAL SITE.A Companson Of Predictive Radionuclides Trans- AMARASOORlYA.W.; NAJAFl.B.; et al. California, Univ. of, l port Modeling Versus Field Ot'servations At The "A" Disposal Santa Baroara, CA. February 1989. 206pp. 8903090145.

l Area.Cnalk River Nuclear Laboratories. ROBERTSON.D E.; 48771:051.

BERGERON.M.P.; HOLFORD.D.; et al. Battelle Memorial Insti- A vanety of probabihstic models to quantify the likelihood of tute, Pacific Northwest Laboratory. March 1989. 157pp. steam explosion induced ( mode) containment failure from core 8904110443. PNL-6175. 49268.311. melt accidents in commercial light water reactors have been

Miln Cit tirns cnd AbstrcCts 11 proposed in the past. In many respects, these models and as- NUREG/CR-5069: TRAC-PF1/ MOD 1. Correlations And Models.

sociated mechanistic considerations w are complimentary. LfLES.D.R.; SPORE,J.W.; KNIGHT,T.D.; et al. Los Alamos Na-Based on this and taking into account recent research efforts in tional Laboratory. December 198(f. 558pp. 8903030322. LA- l this area, we perceived a need to consolidate a common co- 11208-MS. 48720:045.

proach. A new probabilistic framework is pcoposed for this pur. The Los Alamos National Laboratory is developing the Tran-  ;

pose. Quantification of all inputs required by this framework is sient Reactor Analysis Code (TRAC) to provide advanced best- l carried out for the case of a low pressure core melt scenario in estimate simulations of real and postulated transients in pres-a pressurized water reactor. in particular, this quantification is surized light-water reactors and for many related thermal-hy-based on detailed mechaniste considerations of premixing and draulic facilities. The TRAC-PF1/ MOD 1 program is the latest re-energy partitions during the early expansion and later impact leased version. The code features a one-and/or three-dimen-phases of the explosion. Key findings of these studies are: 1. sional, two-fluid treatment for the thermal hydraulics, together Premixing of large corium pours is limited by strong coolant de. with other necessary modeling capabilities to describe a reactor pletion phenomena, 2. Explosions which may be energetic system.

enough to be considered challenging to the upper head vessel NUREG/CR-5074: DEVELOPMENT OF A PHENOMENA IDENTI-integrity would also fall the lower head, thus substantially reduc-FICATION AND RANKING TABLE (PIRT) FOR THERMAL-HY-ing the upward-risrected mechanical energy release, and 3.

, DRAULIC PHENOMENA DURING A PWR LARGE-BREAK Up ser vessel head loading is dominated by a solid-solid impact LOCA. SHAW R.A.; ROUHANI,S.Z.; LARSON,T.K.; et al. EG&G (i.e., core support plate and associated structures impacting the Idaho, Inc. (subs. of EG&G, Inc.). November 1988. 'iOOpp.

upper head), which produces a highly concentrated, in time, im- 8002080515. EGG-2527. 48402:174.

pulse. The results of the probabilistic combination of these The U.S. Nuclear Regulatory Commission (blRC)is proposing inputs provide additional support to the generally prevalent ex- to amend its regu!stions to permit the use of best-estimate pectation that steam explosions do not pose a significant threat safety analysis codes to demonstrate that the emergency core to the containment. To fully appreciate the qualitative limitations cooling system would protect the reactor core during a postulat-on this result the compiete study must be carefully considered. ed loss-of coolant accident (LOCA). A l'ey feature v this pro-posed rule in that the licensee will be required to quantify the NUREG/CR 5042 S02: EVALUATION OF EXTERNAL HAZARDS uncertainty of the best-estimate calculations and include that TO NUCLEAR POWER PLANTS IN THE UNITED uncertainty when comparing the calculated resuits with the Ap-STATES.Other External Events. KIMURA,C.Y.; pendix K hmits. The NRC has further proposed a Code Scaling, PRASSINOS P G. Lawrence Livermore National Laboratory. Applicability, and Uncertainty (CSAU) evaluation methodology.

February 1989. 54pp. 6903060025. UCID-21223. 48762:080. One of the cornerstones of that methodology is the identifica-in suppo"t of implementation (,f the Nuclear Regulatory Com- tion and ranking of all the processes that occur during a specific mission's Severe Accident Policy, the Lawrence Livermore Na- scenano. The ranking is done according to importance during tional Laboratory (LLNL) has performed a study of the risk of the .cenario and is used to limit the uncertainty analysis to a core damage to nuclear power plants in the United States due sufficient, but cost effective scopa. The work reported in this to "otner external events." The broad objective has teen to document identif.ea the thermal-hydraulic phenomena that occur gain an understanding 3 of whether "other external events" (the dunng a large-break loss-of-coolant accident (LBIOCA) in a hazards not covered by previous reports) are among the major Westa.ghouse four loop pressurized water reactor and ranks tne potential accident initiators that may pose a threat of severe re- relative importance of ecch with respect to peak cladding tem-actor core damage or of a large radioactive release to the envi- perature.

ronment from the reactor. The "other extemal events" covered in this report are nearby industrial / military facility accit.ents, on NUREG/CR 5088: FIRE RISK SCOPING STUDY: INVESTIGATION site hazardous material storage accidents, severe temperature OF NUCLEAR POWER PLANT FIRE RISK,1NCLUDING PREVI-OUSLY UNADDRESSED ISSUES. LAMBRIGHT,J.A.;

transients, severe weather storms, lightning strikes, external fires, extraterrestrial activity, volcanic activity, earth movement, NOWLEN,S.P.: NICOLETTE,V.F.; et al. Sandia National Labora-and abrasive windstorms. The analysis was based on two fig. tories. , January 1989. 427pp. 8902100146. SAND 88-0177.

ures of-merit, one based on core damage frequency and the 48443:299.

Other based on the frequency of large radioactive releases. An investigation of nuclear power plant fire nsk issues raised as a result of the USNRC sponsored Fire Protection Research NUREG/CR 5046: A "INITE ELEMENT ANALYSTS OF A REAC. Program at Sandia National Laboratories has been performed.

TOR PRESSURE , SEL DURING A SEVERE ACCIDENT. The spot.ific objectives of this study were (1) to review and re-CHAMBERS.R.S. Sandia National Laboratories. February 1989. quantity fire risk scenarios from four fire probabilistic risk as-41pp. 8903030367. SAND 87-2183. 48713:315. assments (PRAs) in light of updated data bases made avail-

.ote as a result of USNRC sponsored Fire Protection Research Finite element analyses have been conducted to investigate Program and updated computer fire modeling capabilities, (2) to some of the salient features associated with the failure of a identify potentially significant fire risk issues that have not been pressunzed nuclear reactor vessel. In this limited study, the fire risk context and to quantify the temperature dependent elastic plastic response of the steel has previously addressed potential impact of those i in a,dentified fire risk issues where possi-been considered without regard to any embrittlement or coupled ble, and (3) to review current fire regulations and plant imple-creep phenomena demonstrated by the material. The primary mentation practices for relevance to the identified ut. addressed purpose of these calculations is to determine the relative impor- ssues. h nm W aswes wme @,nM wM had nd e

tance of thermal strains in establishing the net time to failure. previously been addressed in a fire r sk context. These six The presence of molten material in the bottom of the pressure issm a c n rol eswns intmactions, (2) sssrnW inW-vessel is crudely simulated by a teruperature boundary condi- a s,@ m W W QWg & & ms W @ s e tion. Thermal properties are assumed to be temperature inde- otroll (4) total environment equipment survival pncluding the pendent and the resulting linen Seat conduction Problem is spurious operation of fire suppression systems), (5) the adequa-solved independently (uncouples trorr the mechanical problem..

cy of fire barrier qualification methods, and (6) the adequacy of Although the pretence of thermal strahs dramatically increases analytical tools for fire.

the magnitude of the sag in the bottom of the pressure vessel and changes the transient stress distribution occurring through WUREG/CR-5100: INTEGRATING FIBER OPTIC RADIATION DO-the thickness of the shell, it does not significant!y alter the time SIMETER. SOLTANI.P.K.; WRIGLEY,C.Y.; STORTI,G.M.; et al.

to fa:Iure. Quai,.ex Corp. March 1989. 53pp. 8903260361. 49069:320.

l

'12 Rin Cit:ti:ns end Ab;tr: cts The purpose of this research effort was to determine the fea- fuel performance, and radiation buildup. A comparison of Nucle-sibility of forming a radiation sensor coupled to an optical fiber er Regulatory Commission (NRC) Regulatory Guide 1.56, the l capable of measuring gamma photon, X-ray, and beta particle Boiling Water Reactor Owners Group (BWROG) Water Chemis- j l dose rates and integrated dose, and to construct a prototype try Guidelines, and Plant Technica! Specifications showed that

{

dosimeter read-out system utilizing the fiber optic sensor. The the BWROG Guidelines are more stringent than the NRC Regu- j key component of the prototype dosimeter system is a newly- lutory Guide, which is almost identical to Plant Technical Speci- 1 developed radiation sensitive storage phosphor. When this fications. Plant performance with respect to BWR water chemis-phosphor is excited by energetic radiation, a proportionate pop- try has shown dramatic improvements in recent years. Up until ulation of electron-hole pairs are created which become trapped 1979 BWHs experienced an average of 3.0 water chemistry in-at specific impurities within the phosphor. These can subse- Cidents per reactor-year. Since 1979 the water chemistry techni-quently be stimulated optically with near infrared at abcut 1 mi- ca! specifications have been violated an average of on!y p.2 crometer wavelength producing a characteristic luminescence times per reactor-year, with the most recent data from 1986-which is directly proportional to the energetic radiation input. By 1987 showing only 0.05 violations per reactor-year. The data attaching the phosphor to the end of an optic # fiber, it is possi- clearly demonstrate the industry-wide commitment to improving ble to transmit both the IR optical stimulation and the character, water quality in BWRs. In addition to improving water quality, istic phosphor luminescence through the fiber to and from the domestic BWRs are beginning to switch to hydrogen water read-out instrument, which can be located far (e.g, kilometers) chemistry (HWC), a remedy for intergranular stress corrosion from the radiation field. This document reports on the specific cracking. Three domestic BWRs are presently operating on design of the prototype system and its operating characteristics, HWC, and fourteen more have either performed HWC mini tests including its sensitivity to various radiation dose rates and ener- or are in various stages of HWC implementation. This report in-gies, its dynamic range, signal-to-noite ratio at various radiation cludes a detailed review of HWC science and technology as intent.4 ties, and other system charactelstics. Additionally, the re- well as areas in which further research on BWR water chemistry dietion hardness of the phosphor and fiber are evaluated. may be needed.

NUREG/CR-5102: INTERFACING SYSTEMS NUREG/CR-5116: SURVEY OF PWR WATER CHEMISTRY.

LOCA: PRESSURIZED WATER REACTORS. BOZOLI.G.; GORMAN,J. Dominion Engineering, Inc.

• Argonne National KOHUT.P.; FITZPATRICK R. Brookhaven National Laboratory- Laboratory. February 1989.138pp. 8903140138. ANL 88-43.

February 1989. 350pp. 8904110373. BNL-NUREG-52135. 48838:281.

49223.221.

This report surveys available information regarding primary This report summarizas a study performed by Brookhaven and secondary water chernistnes of pressun2ed water reactors National Laboratory for t.1e Reactor and Plant Safety issues (PWRs) and the impact of these water chemistries on reactor Branch. RES, U.S. Nuclear Regulatory Corr?ntssion. This study operation. The emphasis of the document is on aspects of was requested by the NRC in order to provide a technical basis water chemistry that , ' ct the integrity of the primary pressure for the resolution of Genenc Issue 105, " Interfacing LOCA at boundary and the radiation dose associated with maintenance LWRs." This report deals with pressurized water reactors and operation. The report provides an historical overview of the I (PWRs). A parallel report was also accomplished for boiling 6evelopment of pnmary and secondary water chemistries, and water reactors. This study focuses on three representative describes practices currently being followed. Current problems PWRs and extrapolates the plant specific findings for their ge- and areas of research associated with water chemistry are de-neric applicability. In addition, a generic analysis was performed scribed. Recommendations for further research are included.

to investigate the cost-benefit aspects of imposing a tesdng pro-gram that would require some minimum level of leak testing of NUREG/CR-5124: INTERFACING SYSTEMS LOCA. BOILING the pressure isolation valves on plants that presently have no WATER REACTORS. CHU,T-L.; STOYANOV,S.;  !

such requirements. FITZPATRICK.R. Brookhaven National Laboratory. February l NUREG/CR-5112: EVALUATION OF BOILING WATER REAC- 1989.288pp.8903140208. BNL NUREG-52141. 48849:001.

TOR WATER. LEVEL SENSING LINE BREAK AND SINGLE TNs report summarizes a study performed by Brookhaven FAILURE.Genene issue 101 Boiling Water Reactor Level Re- National Laboratory for the Office of Regulatory Research, Re- l dundancy-Technical Findings. BRUSKE.S.J.; COLLINS,B.L; actor and Rant Safety issues Branch, Division of Reactor and j MONNIE.D.L; et al. EG&G Idaho, Inc. (subs. of EG&G, Inc.). Plant Systems, U.S. Nuclear Regulatory Commission. This study i March 1989.100pp. 8903280339. EGG 2536. 49072:096. was requested by the NRC in order to provide a technical basis This report presents an evaluation of the potential safety con _ t sr the resolution of Generic issue 105, " Interfacing LOCA at corns identified in Genenc issue 101, related to Boiling Water L VRs." This report deals with boiling water reactors (BWRs). A Reactor (BWR) level sensing line breaks. For this review, failure PUallel report was also accomplished for pressurized water re-combinations and transients were evaluated to assure that ex. ac ors. This study focusses on three representative BWRs and isting BWR plants could be safely shut down under postulated ewapolates the plant specific findings for their genenc apphca-conditions of a break or leak in the instrument line of the reac- bikty. In addition, a generic analysis was performed to investi-tor vessel level instrumentation plus an independent single fail, gate the cost-benefit aspects of imposing a testing program that ure in any protection system. The review evaluated all the de- would require some minimum level of leak testing of the pres-signs currently employed in bothng water reactor plants. Part I sure isolation valves on plants that presently have no such re-of this repo"t describes the methodology used to evaluate the quirements.

vanous designs and provides technical findings. Part 11 presents the value/ impact analysis performed to evaluate the vanous al- NUREG/CR-5143: APPLICATION OF THE J-INTEGRAL AND tornatives that we's considered to improve plant response to a THE MODIFIED J-INTEGRAL TO CASES OF LARGE CRACK EXTENSION. JOYCE,J.A.; DAVIS.D.A.; HACKETT,E.M.; et al.

postulated water-level sensing hne break and single failure' U.S. Naval Academy, Annapolis, MD. February 1989. 45pp.

NUREG/CR 5115: A REVIEW OF BOILING WATER REACTOR 8903280050,49080:087.

WATER CHEMISTRY. Science, Technology,And Performance. The J. integral is widely accepted as a measure of elastic-FOX,M.J. Argonne National Laboratory.

• Aptech Engineering plastic fracture toughness of engineering alloys. Specimen size Services. February 1989 76pp 8903280321. ANL-88-42. and geometry dependence were first noted in fracture tough-49071:039. ness measurements using the ASTM. Est3 standard calculation Bosling water reactor (BWR) water chemistry (science, tech- of 'he deformation J-integral (J(d)) by McCabe, et al. in 1983.

nology, and performance) has been reviewed with an emphasis The modified J integra! (J(m)) was introduced by Ernst to at- i on the relationships between BWR water quality and corrosion, tempt to minimize or ehminate size and geometry dependence i l

M:In Citations and Abstrccis 13 when crack growth was appreciable. Since J(m) was introduced, were loss of pool cooling and make-up capability, fuel rack questions have ansen regarding the proper parameter to de- damage and loss of liner integrity due to a cask drop accident.

acribe the response of a flawed body to loading. The objective of this research task is to investigate the crack growth and NUREG/CR-5186: VALUE/ IMPACT ANALYSIS OF GENERIC specimen J capacity limitations of J(d) and to verify the accura- ISSUE 94. "/,DDITIONAL LOW TEMPERATURE OVERPRES-cy and specimen independence of the current J(m) formulation. SURE PROTECTION FOR LIGHT WATER REACTORS."

J-R curve tests have been conducted on 1/2T,1T, and 2T GORE.B.F.; VO,T V.; COLBURN A.J.; et al Battelle Memorial in-compact tension specimens of materials having entical fracture stitute, Pacific Northwest Laboratory. November 1988. 173pp.

toughness values ranging from J(Ic) = 800 in-lb/in(2) to J(Ic) = 8902080406. PNL-6589, 48410:100.

2600 in-lb/in(2). These tests were conducted according to This document prcesnts a value/ impact analysis of the ex-ASTM E1152-87 with the exception that the specirnen loading pected risk reduction and associated costs for seven regulatory was continued until large crack extensions were present in alternatives which have been proposed to resolve the generic many cases exceeding 50% of the initial uncracked ligament. issue of LTOP transients (Gl 94). It also presents an analysis of Additional analysis was conducted as wellin terms of the modi- the reduction of public risk which was accomplished by the fied J integral (J(m)) using equations proposed by Ernst and 1979 imposition of requirements for LTOP transient protection, Landes. The test procedure was directes toward finding the when LTOP had previously been designated as a generic issue useful limits of the two J quantities for the /naterials listed (GI A-26). The alternatives evaluated were: Alternative 1, no above, or for materials of equivalent strength and toughness. action (base caso evaluation of risks and consequences for NUREG/CR-5173: REVIEW AND ASSESSMENT OF THERMO- bTOP events); Alternative 2, prohibit operations with the RCS DYNAMIC AND TRANSPORT PROPERTIES FOR THE CON-water solid" except when it is depressurized and vented, and TAIN CODE. VALDEZ,G.D. Sandia National Laboratories. De- require all operating reactors to maintain a bubble of steam or cember 1988.171pp. 8901230131. SAND 88-7101. 48184:238. noncondensible gas (N(2)) in the pressunzer when the RCS is A study was carried out to review available data and correla- not vented; A!temative 3, prohibit operation of the RCS in a tions on the thermodynamic and transport properties of materi. water-solid condition when either train of the OMS is out of als applicable to the CONTAIN computer code. CONTAIN is the service; Alts, native 4, proliibit operation with the RCS in a NRC's best-estimata, mechanistic computer code for modehng wate]olid condition when a high pressure safety injection containment response to a severe accident. Where appropnate, pump is in service; Alternative 5, prohibit restart of a reactor recommendations have been made for sittable approximations coolant pump when the RCS is in a water solid condition, After-for material properties of interests. Based on a modif6d Bene. native 6, require that the pressure setpoint for automatic isola-dict-Webb-Rubin (BWR) equation of state, a procedure is intro. tion of the residual heat removal system be raised at ove the duced for calculating thermodynamic properties for common setpoint for residual heat removal safety relief valve opening to gases in the CONTAIN code. These gases are nitrogen, maintain this path as backup to the OMS; and Altemative 7, re-oxygen, hydrogen, carbon dioxide, carbon monoxide, steam, quire the OMS to be safety grade. The risk analysis of each al-helium, and argon. The thermodynamic equations foi density, ternative estimated the public nsk from the operation of the 63 currently represented in CONTAIN by relatively simple fits, were presently operating PWRs summed over the period from the independently checked and are recommended to bo replaced present to EOL.

by the Lee-Kesier equation of state which substantially im-proves accuracy without too much secnfice in computational ef- NUREG/CR-5195: FATIGUE STRENGTH OF ASME SA 106-B ficiency. The accuracy of the calculated values have been found WELDED STEEL PIPES IN 288 DEGREES C AIR ENVIRON-to be generally acceptable. Various correlations and models for MENTS. TERRELL,J.B. Materials Engineenng Associates, Inc.

single component gas transport properties, viscosity and ther. December 1988. 50pp. 8901230111 MEA-2307. 48177:289.

mal conductivity, were also assessed with available expenmen. Fatigue hfe tests were performed on 102-mm (4-in.) girth butt tal data. When a suitable correlation or model was not available, welded SA 106-B steel pipes in 288 degrees C air under axial transport properties were obtained by performing least-sqw. ires loading with a load ratio (R) of -1.00. The fatigue strength at fit on experimental data. Twenty-two additional matenals con- 10(7) cycles was determined to have an extrapolated value of sisting of condensed phases and miscellaneous vapors were 66.9 MPzi (9.7 ksi), and the fatigue strength reduction factor also evaluated. The approach ranged from (1) a least square fit (K(f)) ws.s determined to be 3.47. Crack initiation usually oc-to available data using a polynominal or other function having curred at the notch located at the internal weld fusion hne on little or no theoretical basis, to (2) a semiempiricLI correlation the inside diameter of each pipe, at the point of weld bead emploving an analytical expression suggested by theory with ov1!rlay. In all cases, the fatigue crack progressed on a plane constants determined by comparison with data. normal to the pipe axis. The presence and relatwo number of secondary initiation sites, observed at internal weld fusion hne NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALY. locations around the pipe circumference, appeared to be a SES OF THE SPENT FUEL POOLS AT TWO REPRESENTA- function of applied stress amplitude. ASME Section lit analysis TIVE NUCLEAR POWER PLANTS. PRASSINOS.P.G.; of the data, usirg the procedures given in NB-3228.5 and NB-KIMURA,C.Y,; MCCALLEN,D.B.: et al. Lawrence Livermore Na- 3653.5, resulted in alternating stress (S(alt)) values which were tional Laboratory. January 1989.167pp. 8902090016. UCfD- conservatwe when compared to the ASME Section til mean 21425.48441:213. data line for caton steels. The results of this study successfully This report discusses work done in support of the resolution provide baseline data for comparison with proposed fatigue life of Generic issue-82, "Beyond Design Basis Accidents in Spent tests of girth butt welded pipes containing 288 demees C PWR Fuel Pools." Specifically the probability of spent fuel pool failure environments at 13.8 MPa, under axial loading r; a cyche fre-due to earthquakes was determined for the pools at the Ver- quency of 0.017 Hz.

mont Yankee Nuclear Power Station (BWR) and the H. B. Rob-inson S. E. Plant. Unit 2 (PWR). The dom:nant failure mode for NUREG/CR-5197: EVALUATION OF GFNERIC ISSUE 115. "EN-each pool was gross structural failure caused by esismic motion HAf4 CEMENT OF THE RELIAB18 'iY OF WESTINGHOUSE resulting in the loss of pool hner integnty. The resulting sudden SOLID STATE PROTECTION SYSTEM / RENY,D. A.;

loss of water was then assumed to lead to a self-propagating BRUSKE S.J.; VALENTI,L.N.; et al EG8G Idaho, Inc. (subs. of cladding failure and fission product inventory release from 'the EG8G, lac ). January 1989.134pp. 8902080304. EGG-2546.

spent fuel elements in the pool. The mean annual frequency of 48408.127.

failure due to this failure mode was found to be 6.7E-06 at Ver- This report presents an evaluation of the potential safety con-mont Yankee and 1.8E-06 at H. B. Robinson. Other earthquake aorns identified in Genenc issue 115, enhancement of the reli-induced failure modes studied but found to be less important ability of the Westinghouse solid state protection system. Sever-b

14 M In Cittti ns Cnd Abstrzets al options for improving the design of the reactor trip system oi 10 CFL Part 61 radionuclides, selected transition metals, and were identified and evaluated for their effect on core damage organic cheisting and complexing agents. The organic acids in-frequency. A consequence snalysis, cost analysis, and cost / cluded oxalic acid, citric acid, formic acid, picolinic acid, ethylen-benefits analysis were performed. ediaminetetrascetic acid (EDTA), and diethylenetnaminepentaa-NUREG/CR-5206: A PROBABILIS cdc acKi (DPTA). Seven solidified resin waste samples were EVALUATION OF THE SAFETY OF BABCOCK & Wil NUCLEAR REACTOR leached in deionized water following the ANS 16.1 leach-test procedure. Release rates and teachability Indexes of radionu-POWER PLANTS WITH EMPHAS. ON HISTORICALLY OB-SERVED OPERATIONAL chdet vansition metals, and organic acids were determined.

EVENTS. HSU.C.J.:

YOUNGBLOOD,R.W.; FITZPATRICK R.; et al. Brookhaven Na- Analytical methods are described which were used to rruasure tional Laboratory. March 1989. 170pp. 89041104FB. BNL- organic acids in resin wastes ard ir' leachate solutions, generat-NUREG-52161. 49224:176. ed during leach-testing of waste form samples. Gas-hquid and This report %mmarizes a study performed by Brookhaven ion chromatography methods were developed which allowed National Labc. :ory for the Office of Nuclear Reactor Regula- detection of organic acids in the tenths- of-a ppm range.

tion Division of Engineering & System Technology (A/D for NUREG/CR-5234: VALUE/ IMPACT ANALYSIS FOR GENERIC Systems), U.S. Nuclear Regulatory Commission. This study was ISSUE 51: IMPROVING THE RELIABILITY OF OPEN-CYCLE requested by the NRC to assist their staff in assessing the nsk SERVICE-WATER SYSTEMS. DALING,P.M.; STILES,0.L.;

significance of features of the Babcock & Wilcox (B&W) reactor WEAKLEY,S.A.; et al. Battelle Memonal institute, Pacific North-plant design in the light of recent operational events. This study west Laboratory. February 1989.127pp. 8903060040. PNL-focuses on a critical review of submissions from the B&W 6668.48777:253.

Owners Group (BWOG) and as an independent assessment of The value/ impact analysis undertaken in this study indicates the risk significance of " Category C" events et each oporating that implemen;ing a fouing surveillance.e and control program at B&W reactor, Category C events are those in which system nuclear power plants would provide a reduction in public health conditions teach limrts which require significant safety system risk for acceptable costs. A basic program would include contin.

and timely operator response to mitigate. A precursor study for uous chlorination of open-cycle service- water systems, periodic l each of the mejor P&W historical Category C events als.o was flushing and flow-testing of the safaty related heat exchangers carried out. In adoition, selected PRAs for B&W reactor plants cooled by the service-water system, inspection of the intake and plants with other pressurized water reactor (PWR) designs structure, and chlorination of the fire protection system. Imple-were revtewed to appraise their handling of Category C events, mentation of more comprehensive fouling programs was shown thereby establishing a comparison between the risk profiles of to reduce public health nsks only slightly more than the basic B&W reactor plants and those of other PWR designs. The ef- program but was shown to increase costs dramatically.

festiveness of BWOG recommendations set forth in Appendix J of the BWOG SPIP (Safety and Performance improvement Pro- NUREG/CR-5239: FLUID FLOW AND SOLUTE TR ANSPORT gram) report (BAW-19t 9) also was evaluaMd. MODELING THROUGH THREE-DIMENSIONAL NETWORKS NUREG/CR-5221: BIODEGRADATION OF ION EXCHANGE OF VARIABLY SATURATED DISCRETE FRACTURES.

MEDIA. BOWERMAN,B.S.; CLINTON.J.H.; COWDERY,S.R. RASMUSSEN,T.C.; EVANS D.D. Arizona, Univ. of, Tucson, AZ.

Brookhaven National Laboratory. August 1988. 54pp. Janitary 1989. 212pp. 8902080497. 48401:246, 8902080298. BNL NURTG-52tB3. 48405:349. The boundary integrr* othod is paed to estimate hydraulic The purpose of this Etudy was to investigate some of the and solute transport prr w , s of unsaturated, fractured rock by more basic aspects of biodegradation of ion-exchange media, solving the boundary vm , '.blem within intersecting fracture ,

specifically to evaluate the ability of micro. organisms to utilize planes. Flow through botG imf>9rmeable and permeable rock is the ion-exchange media or materials sorbed on them as a food determined using two and three dimensional formulations, +

source. The ASTM-G22 teist, alone and combined with the spectively. Synthetic fracture networks are created to perform Bartha Pramer respirometric method, failed to indicate the bio, sensitivity F.tudies, results of which show that: (1) The global hy-degradability of the ion-exchange media. Subsequently, a mixed drolic conductivity is linearty dependent on the product of frac-microbial culture, grown from resin waste samples obtained ture transmissivity and density fnr fractures of infinite length; (2) from the BNL High Flux Beam Reactor, was used to evaluate The effect of correlation between fracture length and transmissi-the susceptibility of different types of ion-exchange media to bi. vity is to increase the global hydraulic conductivity; and (3) Sim.

ological attack. Mixed cation and anion (IRN 77 and IRN 78) ulated ?ow through a fractured permeWe matrix compare fa-resin beads proved to be more susceptible to biological attack vorably with analytic results. Flow through variably saturated after being subjected to gamma irradiation and/or loaded with i4ctures is modeled using a constant capillary head within ind%

organic acid anions such as EDTA, citrate, and oxalate. Mixed vu ual fractures. A simulated free surface compares favorably tRN 77 and IONAC A 365 resin beads supported less abundant u an approximate analytic solution and with laboratory re-growths than the IRN-77 and IRN-78. Irradiation and loading suits. Simulations indicate zones of water under both positive with alcolinate and formate resulted in growths that were ap- and negative pressure, as well as regions of air-filled voids.

proximately equal or less abundant than found in the untreated Travel times ard breakthrough curves are determined by inte-IRN-77 and lONAC A-365 mixture. grating the inverse velocity over a streamline, and then sum-ming over all streamlines. Faster travel times are noted as frac-NUREG/CR-5224: THE TEACHABILITY OF DECONTAMINATION ture saturation decreases for the fracture network examined.

ION-EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPER-ATING NUCLEAR POWER PLANTS. MCISAAC,0.V.; NUREG/CR 5243: BOREHOLE CLOSURE IN SALT.

MANDLER,J.W. EG&G Idaho, Inc. (subs. of EG&G, Inc.). March FUENKAJORN K.; DAEMEN J.J.K. Arizona, Univ. of, Tucson, 1989. 208pp. 89041104?9. EGG 2549. 49239:041. AZ. December 1988. 479pp. 8101230119, 48133:119.

Data are presented on the release of radionuclides, stable Constitutive parameters are jeteimined from characterization metals, and organic reagents froni decontamination ,on-ex- expenments to predict borehot closure in salt subjected to vari-change resin wastes solidified in Por 'arrd cement. Both solidi- ous loading configurations. Paeological, empirical, and physical fied and unsolidified decontammation resin waste samples were theory models are based on uniarial creep tests, strain and collected from five coinmercial light water reactors following stress rate controlled uniaxial tests, constant strain rate triaxial chemical decontamination of primary coolant systems. The de- tests, cyclic loading tests, and seismic velocity measurements.

contaminations were performed using the Can-Decon, AP/ Solutions for a thick-walled cylinder and for a circular hole in an Citrox, Dow NS-1, and LOMI processes. Samples of unsolidified infinite plate are derived from linear viscoelastic models and decontamination resin waste were analyzed f7 concentrations empincal laws, Salt behaves as an elastic-viscoelastic matenal. ,

Miln Citati:ns and Abstr: cts 15 The elastic behavior tends to be linear and tirr,e-independent. NUREG/CR-525C V02: SEISMIC HAZARD CHARACTERIZATION the plastic deformation is time-dependent. The stress in re. OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY ment to strain rate increment ratio gradually decreases as tho MOUNTAINS.Results And Discussion For lhe Batch 1 Stes, stress level increases. The transient potential creep law BERNREUTER,D.L.; SAVY,J.B.; MENSING R.W.; et al. Law-gives the simplest satisfactory description of the transient visc9 rence Livermore National Laboratory. January 1989. 323pp.

plastic behavior. Variation of indinsic properties, attributed to 8903030394 UCID-21517. 48719:082, nonuniform distribution of intercrystalline gaps and air voids, See NUF<EG/CR-5250,V01 abstract.

plays a more significant role in instantaneous than in transient deformation. Mechanisms governing time-dependent deforma- NUREG/CR-5250 V03: SEISMIC HAZARD CHARACTERIZATION tion are fracture propagation, plastic flow and dislocation of satt OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results And Dircussion For The Batch 2 Sites.

crystals, and healing of the intercrystalline gaps and induced BERNREUTER.D.L.; SAVY,J.B.; MENSING R.W.; et al. Law-fractures. Different test parameters induce different combina- rence Livermore National Laboratory. January 1989. 293pp.

tions of reformational mechanisms. which lead to different po- 8903030348. UCID-21517. 48714:118.

tential law parameters. The transient potential creep Ndel See NUREG/CR 5250,V01 abstract.

does not accurately predict borehole closure, probably due t snadequate or incomplete predictive capability of the model for NUREG/CR-5250 V04: SEISMIC HAZARD CHARACTERIZATION these loading and deformation configurations or due to the OF 69 NUCLEAR PLANT SITES EAST OF THE ROCK' MOUNTAINS.Results And Discussion For The Batch 3 Siter methods used in multiaxial formulation, or both. Since model pa-rameters depend upon the mechanisms goveming creep, pro- BERNREUTER,D.L.; SAVY,J.B.; MENSING,R.W.; et al. Lav diction of borehole deformation by using only one s(* of param- rence Livermore National Laboratory. January 1989, 284p eters may be inadequate. 8903030308. UCID 21517. 48715:051.

See NUREG/CR-5250,V01 abstract.

NUREG/CR-5245: A REVIEW OF THE CRYSTAL RIVER UNIT 3 NURES/CR 5250 V05: SEISMIC HAZARD CHARACTERIZATK PROBABILISTIC RISK ASSESSMENT. internal Events. Core OF 69 NUCLEAR PLANT SITES EAST OF THE ROC' Damage Frequency. HANAN.N.A.; HENLEY,D.R. Argonne Na- MOUNTAINS.Results And Discussion For The Batch 4 Sit tional Laboratory. January 1989.193pp. 8902100096. ANL BERPREUTER,D.L.; SAVY,J.B.; MENSING,R.W.; et al. Law-

40. 48443:106. rence Livermore National Laboratory. January 1989. 288pp.

A review of the Crystal River Unit 3 Probabilistic Risk Assess. 8903140041. UCID-21517. 48839:059.

ment (CR 3 PRA) was performed with the objective of evaluat- See NUREG/CR-5250,V01 abstract.

ing the dominant accident sequences ar d major contributions to the core damage frequency from internally generated initiators. NUREG/CR-5250 V06: SEISMIC HAZARD CHARACTER!ZATION This review included not only an assessment of the assumption OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS. Regional Comparison Between Sites. Site and methods used in the CR-3 PRA, but also included a quanti- Effects. General Discussion And Conclusions.

tative analysis of the accident initiators. and accident so- BERNREUTER,D.L; SAVY,J.B.; MENS!NG,R.W.; et al. Law-quences resulting in core damago. The effects of data uncer- rence Livermore National Laboratory. January 1989. 13Spp.

tainties on the core damage frequency were quantified and sen- 8903140099. UCID-21517, 48838:145.

sitivity analysis wat, also performed. See NUREG/CR-5250,V01 abstract.

NUREG/CR-5250 V01: SEISMIC HAZARD CHARACTERIZATION HUREG/CR-5250 V07: SEISMIC HAZARD CHARACTER 1ZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS. Methodology,lnput Data And Comparisons To Pro. MOUNTAINS. Questionnaires. BERNREUTER.D.L.; SAVY,J.B.;

vious Results For Ten Test Sites. BERNREUTER,0.La MENSING,R.W4 ot al. Lawrence Livermore National Laboratory.

SAVY,J B.; MENSING,R.W.; et al. Lawrer.ce Livermore National January 1989. 605pp. 8903140119. UCID-21517. 48849.332.

Laboratory. January 1989. 350pp. 8903030423. UCID-21517. See NUREG/CR-5250,V01 abstract.

48718:092.

NUREG/CR-5250 V08: SEISMIC HAZARD CHARACTERIZATION The EUS Seismic Hazard Characterization Project (SHC) is OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY the outgrowth of an earlier study 9eriormed as part of the U.S. MOUNTAINS. Supplementary Seismic Hazard Results For Sites Nuclear Regulatory Commission's (NRC) Systema %ic Evaluation With Multiple Soil Conditions. BERNREUTER,D.L.; SAVY,J.B.;

Program (SEP). The objectives of the SHC were: (1) to develop MENSING,R.W.; et al. Lawrence Livermore National Laboratory.

a seismic hazard characterization methodology for the region January 1989. 211pp. 8903140068. UCID 21517. 48839:347.

east of the Rocky Mountains (EUS), and (2) the M,) plication of See NUREG/CR-5250,V01 abstract.

the methodology to 69 sate locations, some of them with several local soil conditions. The method developed uses expert opin- NUREG/CR-5251: WELLFIELD INSTALLATION AND ions to obtain the input to the analyses. An important aspect of INVESTIGATIONS.CRESTON STUDY AREA, EASTERN WASH-the elicitation of the expert opinion process was the holding of INGTON. PASCHIS,J.A.; KUNKEL,J.R.; KOENIG,R.A. In-Situ, two feedback meetings with all the experts in order to finalize inc. November 1988. 468pp. 8902080373. 48438 210.

the methodo80gy and the input data bases. The hazard esti- This report summanzes the design and implementation of field testing used to charactenze the Roza Member basalt aqui-mates are reported in terms of peak ground acceleration (PGA) fer of the Wanapum Formation of the Columbia River Basalt and 5% damping velocity response spectra (PSV). A total of Group. A research wellfield was implemented and tests con-eight volumes make up this report which contains a thorough description of the methodology, the expert opinion's solicitation ducted to lithologically classify subsurface co idations underlying an 800-by 1400-foot area located in the northwest quarter of process, the input data base as well as a discussion, compari- Section 16. T.25N R.34E, six miles south of Creston in Lincoln son and summary volume (Volume VI). Consistent with previous County, Washington. The relatively high-yielding aquifers of the analyses, this study finds that there are large uncertainties as- Roza Member basalt are separated by claystone aquitards and sociated with the estimates of seismic hazard in the EUS. and it are hydrologically interrupted by at least two different subsur-identifies the ground motion modeling as the prime contnbutor face hydrologic structures. Water-level data were used to deter-i to those uncertainties. The data bases and software are made mine the regional and local hydraulic grauient. Wellbore slug l available to the NRC and to the public uses through the Nation- tests and passive (nonpumping) tracer tests permitted calcula-a1 Energy Software Center (Argonne, lilinois). tion of equivalent hydraulic apertures and effective porosities for h _ _ _ _ _ _ _ _ _ _ -

16 Main Cit:ti:ns Cnd Abstructs the Roza basalt flow top and flow interiors Anafysis of data data pertaininb J the participating minerals and aqueous spe-from pumping testa permitted calculation of transmissivities and cies. In partbular, it is important that the thermodynamic proper.

storage coefficients for the Roze Msalt. Applicable wellfield aq- ties of the aluminate ion be accurately determined, because uifer analyses along with model a adies will provide guidelines most rock forming minerals in the earth's crust are aluminosili-for use in evaluating proposed sitos for disposing of high-level cates, and most groundweters are neutral to slightly alkaline, nuclear waste in saturated fractured geologic media- where the aluminate ion is tr.e predominant aluminum species in NUREG/CR-5265: SIZE EFFECTS ON J-R CURVES FOR A 302 solution. Without a precist knowledge of the thermodynamic B PLATE. HISER,A.L.; TERRELL,J.B. Materials Engineering As- properties of the aluminate ion, aluminosilicate mineral solubili-sociates, Inc. January 1989. 301pp. 8902030475. MEA-2320. ties cannot be determined.

48400 305. .

This study was conceived to determine J-R curves from vart- HUREG/CR-5276: SADDE (SCALED ABSORBED DOSE DISTRI-ous sizes of specimens to investigate data extrapolation from BUTION EVALUATOR).A Code lo Generate input For VARS-small size specimens. This study resulted in the finding of a sig- KIN. REECE,W.D.; MILLER,S.D.; DURHAM.J.S. Battelle Memori-nificant size effect or dependence for the low toughness A 302- al Institute, Pa:ific Northwest Laboratory. January 1989. 6 Sop. ~

B plate inustigated. The magnitude of this size dependence is 8902100120. PNL-6761,4B442:086.

unprecedented for reactor pressure vessel steels. The observed The VARSKIN computer code has been limited to the iso-size dependence results in vastly reduced J R curve toughness topes for which the scaled absorbed dose distributions were levels with increased specimen size, for compact specimens previded by the Medical internal Radiation Dose (MIRD) Com-ranging in thickness from 12.7 to 152.4 mm, with all specimens mittee or - data that could be interpolated from isotopes that proportional in terms of dimensions. The plate used in this study had similar spectra. This document describes the methodology was specially made to duplicate early production A 302-B to calculate the scaled absorbed dose distribution data for any plates, which typically exhibit low Charp-V upper shelf ene gy levels. The minimal ct:ss-rolling applied to the plate and b.e isciope (ineuding emissions by the daughter isotopes) and its high sulfur content result in a high proportion of manganese-sul- implementacon by a computer code called SADDE (Scaled Ab-fide inclusions. The resultant microstructure is one possible ex. sorbed Dose Distribution Evaluator). The SADDE source code is p!anation for the unexpected results for this plate. Other causes providsd along with input examples and verification calculations.

and ideas for future work are also described. An additional ob-servation is that initial crark length-to-width ratio (a/W) did hsve NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR j an int;uence on tr e J.R curves for this plate. Specifically, a long MARK I CONTAINMENT IMPROVEMENTS. CLAIBORNE E.; i crack iength (a/W-0.6) can give somewhat higher J-R curve SIMlON,G.; KNUDSON,R.; et al. Science & Engir.eering Associ-levels than a short crack length (a/W-0.5), for large a levels, atos, Inc. January 1989. 70pp. 8902090045 SEA 87 253-07A1.

This finding has direct impact on the app;icability of J(D) and 48442:020.

J(M), and data extrapolation. The Nuclear Regulatory Commission (NRC) has edentified several potential improvements to BWRs with Mark I contain-NUREG/CR-5266: EXAMINATION OF TWO 3M TYPE 992F STATIC ELIMINATORS. CZAJKOWSKI.C. Brookhaven National ments. The Mark I containments have been chosen for this an!-

Laboratory. January 1969. 27pp. 8902090056. BNL-NUREG. tial effort because they appear to have a relatively high condi-52175.48457:001. tional probabi!ity of failure for severe accidents. This study is a Various incidents of facility contamination by polonium-210 cost analysis of three potential art,as of Mark I containment im-microspheres have been previously reported A previously per. provement. Considered were: (1) Automatic Depressurization formed fa0ure investigation by Brookhaven National Laboratory System (ADS) enhancements, (2) Back-up water supply en-on six state eliminators concluded that non-uniform and imper- hancements, sod (3) Wetwell venting enhancements. These fact microspheres have been produced and installed in static cost analyses are comprehensive, in addition to the ces's of shminatora manufactured by the Minnesota Mining and Manu- physical mtKlifications, they also inck'de aspects such as 6ngi-facturing Company (3M). This report details the investigation of necring and quality assurance, radiation exposure, health phys-two type 902F static eliminators. The report essentially confirrns ics support, and anti-contami'stion clothing. Also considered the prior report's conclusions that imperfect microspheres are are licensee costs associated with technical specification being produced and that the epory binder matenal is subject to Shanges, writing or rewriting operating procedures, and staff cracking even under ambient condihons. tairnng. The analyses show that the industry costs per reactor NUREG/CR-5270: ASSESSMENT OF SEISMIC MARGIN CALCU. for installation of all improvements range from about $1.6 mil-LATION METHODS. KENNEDY.RP.; MURRAY,R.C.; lion for the low cost options to $3.15 million for the high cost l RAVINDRA,M.K.; et al. Lawrence Liveimore National Laborato- options. In addition, these physical modifications are estimated ry. March 1989. 600pp. 8904110383. UCID-21572. 49221:217. to resuR in occupational radiation exposures ranging from about Seismic margin review of nuclear power plants ieQuires that 20 person. rem to 150 person-rem per reactor, the Hsgh Confidence of Low Probability of Failure (HCLPF) ca-pacity be calculated for certain components. The candidate NUREG/CR-5292: CLOSEOUT OF IE BULLETIN 80-23: FAIL-methods for calculating the HCLPF capacity as recommended URES OF SOLENOID VALVES MANUFACTURED BY VALCOR by the Expert Panel on Quantification of Seismic Margins are ENGINEERING CORPORATION. FOLEY,W.J.; DEAN R.S.;

the Conservative Determiriistic Failure Margin (CDFM) method HENNICK A. Parameter, Inc. February 1989.31pp.8903140089.

and the Fragildy Analysis (FA) method. The preseni study evalu- PARAMETER IE183. 48840:198.

tited these too methods using some representative components Documentation is provided in this report for the closecut of IE in order to provide further guidance in conductira seismic Bulletin 80-23 regarding failures in safety-related systems of nu-margin reviews. It is concluded that either of the two methods clear power plants due to solenoid valves manufactured by could be used for calculating HCLPF capacities. Valcor Engineenng Corporation. Closeout is based on the imple-NUREG/CR-5271: THERMOCHEMICAL PROPERTIES OF mentation and venfication of three actions by licensees of oper.

l GIBBSITE,BAYERUE,80EHMITE, DIASPORE AND THE ALU- ating plants, or one action by holders of construction permits.

MINATE lON BETWEEN O AND 350 DEGREES C. APPS.J.A.; Evaluation of utility responses and NRC/Regon inspection re-MCNEIL.J.M.; JUN,C.-H. Lawrence Berkeley Laboratory. Janu. ports indicates that the bulletin is clobed out for all of the 120 ary 1989. 348pp. 8902060462. LBL 21482. 48407:139. facilites to which it was issued for action. Background informa-A requirement for modelling the chemical behavior of ground- tion is supplied in the introduction and Appendix A of this water in a nuclear waste repository rs accurate thermodynamic report.

1 i

MCin Citations and Abstracts 17 NUREG/CR 5311: AN AS$ESSMENT OF RELAP5/ MOD 2 APPLi- cation of the HFIR data, a survey of all LWR vessel support da-CABILITY TO LOSS-OF-FEEDWATER TRANSIENT ANALYSIS signs, and a structural and fracture-mechanics analysis of the' IN A BxdCOCK AND WILCOX REACTOR PLANT. supports for two specific PWR plarits of particular interest with DIMENNA.R.A.; HALL.D.G.; KULLBERG CX; et al. EG&G regard to a potential for support failure as a result of propaga-

. Idaho, Inc. (subs. of EG&G, Inc.). February 1989. 115pp. tion of flaws. Calculations performed thus far indicate best-esti-8903140164. EGG-2560. 48851:217. mate critical f!aw sizes, corresponding to 32 EFPY,of ~0.2 in for The applicability and scaling capability of RELAPS/ MOD 2 one plant and-0.4 in for the other. These flaw sizes are small when applied to a Babcock and Wilcox (B&W) loss of feedwater enough to be of concern. However, it appears that low-cycle fa-

transient is essessed using a code applicability methodology. A tigue is not a viable mechanism for creation of such flaws, and l

~

loss-of-feedwater test with a feed-and-bleed recovery wes se- thus, presumably the flaws would have to exist at the time of lected from the once-through integral system (OTIS) test data fabrication.

as a reference transient. Nondimensional comparisons are made between code assessment ca!culations and cods applica. NUREG/CR-5324: DETECTING COMPONENT FAILURE POTEN-tions calculations using computer code models scaled accord- TIAL USING PROPORTIONAL HAZARD MODEL. CAVER,D.P.;

ing to scahng criteria derived from the work of Ishd and others. SAMANTA,P.K. Brookhaven National Laboratory. February The results , indicate that RELAPS/ MOD 2 can scale the phe- 1989.37pp.8903030377. BNL-NUREG 52188. 48710:001.

nomena observed in the experiment and that the code is appli- The proportional hazard model (Cox,1972) provides a frame-cable for transients for whicn phenomena are within this enve- work for incorporating measurable parameters (covariates) into lope. The results also demonstrate the usefulness of the code the description of failure rates. In this report a methodology is applicability methodology for interpreting and venfying code cal-' pre.sented for calculating the failure probability based on param-culations. eters measured at a routine surveillance test of the component.

NUREG/CR 5313: EOUIPMENT QUALIFICATION (EO)-RISK These parameters are expected to vary in an identifiable way SCOPING STUDY. BUSTARD,LD. Sandia National Laborato- before a failure indicating an impending failure, and are termed nes. CLARK,J.; MEDFORD,G.T,; et el. Science Applications symptomatic emitters. A decision framework for a repair /re-Internatiorial Corp. (formerly Science Applications, Inc.). January placement policy based on such information is presented. The 1989. 280pp. 8903080553. SAND 88-3330. 48746.101. use of symptomatic emitter information in determining a compo-The objective of the EO-Risk Scoping Study was to use prob _ nent failure potential has, great utility, particularly in highly reli-abilistic risk assessment (PRA) techniques (1) to assess the able systems like those in nuclear power plants. Further work impact of the electrical equipment environmental qualification or on application of this methodology is recommended. If actual lack thereof on reactor fisk and its uncertainties, and (2) to data arp not readily available, simulated data can be developed identify any analyses or testing that may be necessary to using the simulation e,pproach presented in this report.

reduce the risk or its uncertainties stemming from lack of quahfi.

cation of equipment important to safety. To achieve these ob- NUREG/CR 5325: A MULTIVARIATE STATISTICAL STUDY ON A Joctaves, PRA techniques and , insights were employed to identify DIVERSTIED DATA GATHLRING SYSTEM FOR NUCLEAR equipmont that must function in accident-induced harst environ- POWER PLANTS SAMANTA,PX- ' TEICHMANN,T.;

ments and whose failure would be risk significant. Sevval com- LEVINE,fLM.; et al. Brookhaven Nationai Laboratory. February ponents from the resultant list were then st-locted for more de- 1989.130pp.8903030352. BNL-NUREG 52169.

tailed analyses Accdont scenarios and environments, which in this report, multivariate statistical motnods are presented PRAS suggest are nsk significant, were determined for each se- and applied to demonstrate their use in analyzing nuclear power le'M eqtnpment operation. For these accident lant operational data. For analyses of nuclear power plant eqaipment qualification research and testences expen,wereconditions, ex. both approaches are Presented for detecting malfunctions events, . .

emined to determine whether equipment accident reliabil.ty and degradations within the course of the event At the system might dt'for substantially from the reliability values based on Iml, approaches are investigated as a means of d:egnosis of normal operation conditions employed in past PRA analyses. system level performance. Ths involves the detection of devF Note, accident reliability information is generally unaveilc',le. ations from normal performance of the system. The input data Where significant differences were considered probable, para- analyzed are the measu'able physical parameters, such as metric risk achievement analyses were used to assess the po- steam generator level, pressurizer water level, auxiliary feed-tential risk impact of the equipment felures in addition, those water flow, etc. The stady provides the methodology and illus-equipment qualification practices and outstanding research trative examples based on data gathered from simulation of nu-issues that potentially could impact the accident equipment reli- Clear power plant transients and computer simalation of a plant ability were noted. This information, when combined wth per- system performance (due to lack of easily accessible operation-spectives regarding potential equipment risk impact, provided a al data). Such an approach, once fully developed, can be used basis for assessing the potential nsk importance of various EO t sxpl re staPshcally the detection of failure trends and pat-pwctices and issues. Additional discussion regarding the study's terns and prevention of conditions with serious safety imphca-approach, conclusions, and recommendations is provided. tions.

IMREG/CR-5320: IMPACT OF RADIATION EMBRITTLEMErlT NUREG/CR-5326: FRACTURE EVALUATION OF SURFACE ON INTEGRITY OF PRESSURE VESSEL SUPPORTS FOR CRAOKS EMBEODED IN REACTOR VESSEL CLADDING TWO PWR PLANTS. CHEVERTON,R D ; PENNELL,W.E.; MCCABE,D E. Materials Engineering Associates, Inc. March ROBINSON G.C.; et af. Oak Ridge National Laboratory. January 489. 33pp. 8904110388. MEA-2029. 49271:121.

1969.276pp.8903090048. GRNUTM-10966. 48758.074. The surface crack embedded in the clad layer of a teactor Recent data irom the HFIR vessel surveillance program indi- pressure vessel (RPV) has been identified as a critical safety cate a substantial radiation emonttlement rate eFect at low erra- assessment condition relative to the pressurized thermal shock diation temperatures b120 degrees F) fy A212-B, A350-LF3 accident scenano. This project was initiated to determine the A105 Il and corresponding welds. PWR vessel supports are fab- severity of such cracks experimentally. It is the first study to ;n-ricated of similar matenals and are subjected to the same low vestigate irradiated, clad vecal steel, and to identify the materi-temperatures and fast neutron fluxes (10(8) 10(9) n/cm(2) s, E al property and stress conditions in the local region of the crack 1.0 MeV) as those in the HFIR vessel. Thus, the embnttlement that are significant to the analysis. Bend bar tests provided the rate of these stMtures may be greater than previously antici- expenmental simulatson of the subject RPV surface crack This pated. A study sponsored by the NRC is under way at ORNL to report covers analpis techniques used and presents the find-determine the impact of the rate effect on PWR vessel support ings indicated by the experimental resuits for irradiated and Jnir-life expectancy, The scope includes the interpretation and appli- radiated materials.

18 M:In Cititi:ns and Abstracts NUREG/CR-5328: CORRELATIONS BETWEEN POWER AND This report presents the results of a numerical analysis to de-TEST REACTOR DATA BASES GUTHRIE,G.L; SIMONEN,E.P. termine the stabihty of waste disposal rooms for vertical and Battelle Memonal Institute, Pacific Northwest Laboratory. Febru- horizontal emplacement during the period of waste retrieval. It is ary 1989. 41pp. 5!903290011. PNL-6793. 49128:001, assumed that waste retrieval starts 50 years after the initial em-Ddferences between power reactor and test reactor data placement of the waste, and that access to and retrieval of the bases have been evaluated. Charpy shift data has been assem- waste containers take place through the disposal rooms. It is bled from specimens irradiated in both high-flux test reactors her assumed that the disposal rooms are not backfilled.

and low-flux power reactors. Preliminary tests for the existence Wective cooling of the disposal rooms in preparation for of a bias between test and power reactor data bases indicate a ,. 'ite retrieval is included in the possible bias between the weld data bases. The bias is noncon- meters used were taken from the analysis.

Nevada Nuclear Conditons Wasteand pa- !

i servative for power reactor predictive purposes, using test reac. Storage investigation (NNWSI) Project Site Characterization tor data. The lesser shift for test reactor data compared to Plan Conceptual Design Report (MacDougall et al.,1987). Ther-power reactor data is interpreted primarily in terms of greater mal results are presented which illustrate the heat transfer re-poh-t defect recombination for test reactor fluxes compared to sponse of the rock adjacent to the disposal rooms. Mechanical 2' power reactor fluxes. The possibility of greater thermal aging ef- results are presented which illustrate the predicted distribution fects dunng lower damage rates is also discussed. of stress, joint slip and room deformation s for period of time in-vestigated. Under the assumption that the host rock can be NUREG/CR-5335: STABILITY OF DISPOSAL ROOMS DURING classified as " fair to good" using the Geomechanics Classifica-WASTE RETRIEVAL. BRANDSHAUG.T. ITASCA Consulting tion system (Bieniawski,1974), only light ground support would Group, loc. March 1989.101pp. 8904110438. 49271:160. appear to be necessary for the disposal rooms to remain stable.

k l

l

7 Secondary Report Number Index

' This index lists, in alphabetical order, the performing organization-issued report codes for the L -NRC contractor and international agreement reports in this compilation. Each code is cross-L referenced to the NUREG number for the report and to the 10-digit NRC Document C"' i t  !

System accession number.

SECONDARY REPORT NUMBER Ms! PORT NUMBER - SECONDARY REPORT NUMBER REPORT NUMBER ANL 86 32 NUREG/CR-4667 V05 ORNL/TM 10966 NUREG/CR-5320 ANL-88-40 NUREG/CR 5245 PARAMETER IE183 NUREG/CR-5292 ANL 88-42 - NUREG/CR-5115 PLG 0547 NUREG/CR-4780 V02 C.NL-88-43 NUREG/CR-5116 PNL-4532 NUREG/CR-3037 NUREG/CR-4835 PNL-5711 NUREG/CR-4469 V07 BMI-2159 NUi4EG/CR 4604 BNL-NUREG-51454 NUREG/CR-2331 V08 N3 PNL 5849 PNL-5868 NUREG/CR 4849 V01 BNL NUREG-51454 NUREG/CR-2331 V07 N4 P BNL-NUREG 51995 NUREG/CR-4618 P 6 j BNL-NUREG-52135 NUREG/CR 5102 PNL 6589 NUREG/CR-5186 BNL-NUREG 52141 NUREG/CR-5124 PNL 6668 NUREG/CR-5234 BNL-NUhEG-52144 NUREG/CR-4948 PNL 6761 NUREG/CR-52,6 BNL NUREG-52181 NUREG/CR-5206 PNL4793 NUREG/CR-5328 -

BNL-NUREG-52163 NUREG/CR-5221 SAND 87 2183 NUREG/CR 5046 BNL-NUREG 52175 NUREG/CR 5266 SAND 88 0177 NUREG/CR-5088

, BNL-NUREG 52188 NUREG/CR-5324 SAND 88 3330 NUREG/CR 5313 l BNL-NUREG 52189 NUREG/CR-5325 SAND 88-7101 NUREG/CR-5173 i EGG-2458 NUREG/CR-4639 V5P3R1 SEA 87 253-07A1 NUREG/CR-5278 NUREG/CR-4554 V03 EGO-2458 NUREG/CR 4639 V5P4R1 UCID 20074 EGG 2458 NUREG/CR-4639 V5P2R1 UCID 20674 NUREG/CR-4554 V02 EGG-2485 . TIREG/CR 4835 UCID-20674 NUREG/CR-4554 VOS EGG-2527 ,1LREG/CR-5074 UCID-20674 NUREG/CR-4554 V04 EGG 2536 NUREG/CR-5112 UCID 20674 NUREG/CR-4554 V01 EGG-2,546 NUREG/CR-5197 UCID-20914 NUREG/CR 4792 V02 UCID-20914 NUREG/CR-4792 V01 EGG-2549 NUREG/CR-5224 NUREG/CR-5042 S02 0C10-21223 EGG NUREG/CH-5311

' EPR[2560 NP 5613 NUREG/CR-4780 V02 $$ E h ,^ $ yo4

- IEB 80 23 NUREG/CR 6292 UCID 21517 NUREG/CR-5250 V03 LA 11208-MS . NUREG/CR-5069 UCID-21517 NUREG/CR 5250 V02 LBL-21482 NUREG/CR-5271 UCID 21517 NUREG/CR-5250 V01 MEA 2307 NUREG/CR-5195 UCID-215t7 NUREG/CR 5250 VOS MEA 2320 NUREG/CR-5265 UCID 21517 NUREG/CR-5253 V08 MFA-2329 -

NUREG/GR-5326 UCID-21517 NUREG/CR-5250 V06 ORNL/NSIC 200 NUREG/CR-2000 V07N12 UCID-21517 NUREG/CR 5250 V07 l ORNL/NSIC-200 NUREG/CR 2000 V08 N1 UCID-21572 NUREG/CR 5270 1

- ORNL/TDMC-4 NUREG/CR 4478 UCRL 53043 NUREG/CR 3006 l

l I

l 19

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

<l

'n'

.S

' ,J . 5 i 5

'I i

l i .- ,

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.

ABEL,K.H. NUREG/CR-5250 VOS MISMIC HAZARD CHARACTERIZATION OF 69 NUREG/CR4879 V02: DEMONSTRATION OF PERFORMANCE MOD- NUCLEAR PLANT SITL 5 f AST OF THE ROCKY MOUNTAINS.Results ELING OF A LOW LEVEL WASTE SHALLOW 4AND BURIAL SITE.A And Dscussion For The Batch 4 Sites.

Companson Of Predictive Radeonuclide Transport Modeling Versus NUREG/CR-5250 V06: SEISMIC HAZARD CHARACTER 12ATION OF 69 Field Observations At The "A" Dsposal Area. Chalk River Nuclear Lab- NUCLEAR PLANT SITES EAST OF THE ROCKY l oratories. MOUNTAINS.4egional Comparison Between Sites. Site Effects. General )

Dscusson And Conclusions.

ABOLFADL,M.A. NUREG/CR-5250 V07: SEISMIC HAZARD CHARACTERIZATION OF 69 NUREG/CR-5030; AN ASSESSMENT OF STEAM-EXPLOSION-IN' PLANT SITES OF THE ROCKY NUCLEAR EAST DUCED CONT AINMENT FAILURE. MOUNTAINS.Ouestonnaires.

AMARASOORlYA,W. NUREG/CR-5250 V08: SEISMIC HAZARD CHARACTERIZATION OF 69 NUREG/CR-5030: AN ASSESSMENT OF STEAM-EXPLOSION-IN. NUCLEAR PLANT SITES EAST OF THE ROCKY DUCED CONT AINMENT FAILURE. MOUNTAINS. Supplementary Seismic Hazard Results For Sites With Multiple Soil Conditions.

AMICO,P.J.

NUREG/CR 5206: A PROBABILISTIC EVALUATION OF THE SAFETY BIRKAA.S.

OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS NUREG/CR 4849 V01: STEAM GENERATOR GROUP PROJECT. Task 9 WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL Final Report Nondestructive Evaluation Round Robin. Volume 1: De-EVENTS. senpton And Summary Data.

NUREG/CR4849 V02: STEAM GENERATOR GROUP PROJECT. Task 9 URE /CR 5271: T riERMOCHEMICAL PROPERTIES OF t GlBBSITE.BAYERITE.BOEHMITE. DIASPORE AND THE ALUMINATE lON BETWEEN O AND 350 DEGREES C. BLACKMAN,H.S.

AZARM,M.A. NUREG/CR dD35: COMPAR. SON AND APPLICATION OF QUANTITA-NUREG/CR-4618: EVALUATION OF RELIABILITY TECHNOLOGY Ap. TlVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK PLICABLE TO LWR OPERATIONAL SAFETY. METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP). Final Report.

BACA.G.

NUREG/CR4627 R01: GENERIC COST ESTIMATES. Abstracts From BOHN.M.P.

Generic Studies For Use in Prepanng Regulatory impact Analyses. NUREG/CR-5008: FIRE RISK SCOPING STUDY: INVESTIGATION OF NUCL EAR POWER PLANT FIRE i<lSK,lNCLUDING PREVIOUSLY UN-BALLINGER.M Y. ADDAESSEDISSUES.

NUREG/CR-3037: USER'S MANUAL FOR FIRIN.A Computer Code To Estimate Accidentat Fue And Radoactwe Airborne Release in Nuclear BOWEN,W.M.

Fuel Cycle Facilities. NUREG/CR4604: STATISTICAL METHODS FOR NUCLEAR MATERIAL BELL,B.J. MANAGEMENT.

NUREG/CR4835: COMPARISON AND APPLICATION OF OUANTITA. BOWEM,BS TlVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE REK METHODS INTEGRATION AND EVALUATION PROGRAM NUREG/CR-5221. BIODEGRADATION OF ION-EXCHANGE MEDIA.

(RMIEP) Fir 4 t Report BENNETT,C.A. NUREG/CR-5102: INTERFACING SYSTEMS LOCA: PRESSURIZED NUREG/CR4604: STAllSTICAL METHODS FOR NUCLEAR MATERIAL WATER REACTORS.

MANAGEMENT.

BRANDSHAUG,T.

BERGERON.M.P. N NUREG/CR 5335: STABILITY OF DISPOSAL ROOMS DURING WASTE NUREG/CH4879 V02: DEMONSTRATION OF PERFORMANCE MOO- RETRIEVAL.

ELING OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A Comparison Of Predictive Ratteonuclide Transport Modeling Versus BRUSKE,S.J.

Field Observations At The "A" Dsposal Area, Chalk Rwer Nuctear Lab- NUREG/CR 5112; EVALUATION OF BOILING M TER REACTOR oratones. WATER-LEVEL SENSING LINE BREAK AND SINGLE BERNREUTER,0.L. FAILURE.Genenc issue 101 Boiling Water Reactor Level Redundancy.

^ "^"^ ^

NUR G C -5 E /ALUATION OF GENERIC ISSUE 115. " ENHANCE-N LEAR P SE E ST OF HE CKY MENT OF THE RELIABILITY OF WESTINGHOUSE SOLID STATE MOUNTAINS Methodology.lnput Data And Compensons To Prevtous Results For Ten Test Sites. PROTECTION SYSTEM NUREG/CR-5250 V02: SEISMIC HAZARD CHARACTERIZATION OF 69 se a n For B tch 1 tes NUREG R 4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR NUREG/CR-5250 V03: SE!SMIC HAZARD CHARACTERIZATION OF 69 COOLANT PIPING. Pipe Failure Induced By Crack Growth And Failure NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS Results Of intermediate Supports.

And Dscussion For The Batch 2 Sites.

NUREG/CH4250 V04: SEISMIC HAZARD CHARACTERIZATION OF 69 BUSTARD,LD.

NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Rosults NUREG/CR 5313: EQUIPMENT OVALIFICATION (EO). RISK SCOPING And Dscussion For The Batch 3 Sites. STUDY.

21

1 22; Personal Author index -

l' CAMPSELL,R.D. -

NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK f NUREG/CR-5178. SEISMIO FAILORE AND CASK DROP ANALYSES OF CONTAINMENT IMPROVEMENTS.

= THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR POWER PLANTS." CLARK,J.

CANTRELL,s.K- NUREG/CR-5313: EOUIPMEN'T OUAllFICATION (EO).RtSK SCOPING STUDY' NUREG/CR.4950 V03: THE SHORELINE ENVIRONMENT ATMOS.

PHERIC DISPERSION EXPERIMENT (SEADEX).Asrborne LIDAR Data- CLINTON,J.H.

CAPPIELLO,M.W. NUREG/CR 5221: BIODEGRADATION OF lON EXCHANGE MEDIA.

NUREG/CR-5069: TRAC.PF1/ MOD 1. Correlations And Models.

COATS.D.W.

CHAMBERS,R.S. NUREG/CR-3006: DAMPING IN BUILDING STRUCTURES DURING NUREG/CR-5046: A FINITE ELEMENT ANALYSIS OF A REACTOR EARTHQUAKES. Test Data And Modeling.

. PRESSURE VESSEL DURING A SEVERE ACC DENT.

COHEN,S.

CHAMP.D.R.

NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK 1 NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD- CONTAINMENT IMPROVEMENTS.

ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL SITE.A Comparison Of Predictive Radonuchde Transport Modehng Versus COLBURN,A.J.

Field Observations Al The "A" Dsposal Area. Chalk River Nuclear Lab-NUREG/CR-5186: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE 94, oratones

" ADDITIONAL LOW TEMPERATURE OVERPRESSURE PROTECTION CHAN,M.K. FOR LIGHT WATER REACTORS."

NUREG/CR.3037: USER'S MANUAL FOR FIM A Computer Code To e Ac I Fire And Radioactive Airbome Release in Nuclear COLLINS 8.L

' NUREG/CR-5112: EVALUATION OF BOILING WATER REACTOR WATER-LEVEL SENSING LINE BREAK AND SINGLE CHEN.J.C. - FAILURE. Generic lasue 101 Boiling Water Reactor Level Redundancy.

NUREG/CR-5250 V01: SEISMIC HAZARD CHARACTERIZAT.'ON OF 69 Technical Findings.

NUCLEAR PLANT SITES EAST OF THE ROCKY

' MOUNTAINS. Methodology. Input Data And Comparisons To Previous COWDERY,S.R.

Results For Ten Test Snes. NUREG/CR-5221: BIODEGRADATION OF ION EXCHANGE MEDIA.

NUREG/CR-5250 V02: SEISMIC HAZARD CHARACTERIZATION OF 69 i NUCLEAR PLANT SITES E AST OF THE ROCKY MOUNTAINS.Results CREAGER,R.E.

And Dscussion For The Batch 1 Sites. NUREG/CR-5100: INTEGRATING FIBER OPTIC RADIATION DOSIME-NUREG/CR-5250 V03: SEISMIC HAZARD CHARACTERIZATION OF 69 TER NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results And Dscussion For The Batch 2 Sites- CZAJKOWSKl,C.

NUREG/CR-5250 V04: SEISMIC HAZARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EASY OF THE ROOKY MOUNTAINS.Results NUREG/CH 5266. EXAMINATION OF TWO 3M TYPE 902F STATIC And Dscusson For The Batch 3 Sites. ELIMINATORS' NUREG/CR-5250 V05: SEISMIC HAZARD CHARACTERIZATION OF 69

^ " ES AST THE ROCKY MOUNTAINS.Results DAEMEN.J.J K c F NUREC/CRd243; BOREHOLE CLOSURE IN SALT.

NUREG/CR 5250 V06: SEISMIC HAZARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY DANP.E MOUNTAINS.Re9enal Compenson Dehveen Sites, Site Effects.Generel NUREG/CR-5234: VALUE/ IMPACT ANALYSIS FOR GENERIC ISSUE (Acusuon And Conclusions. 51: IMPROVING THE RELIABILITY Or OPEN-CYCLE SERVICE-NUREG/CR-5250 V07: SEISt4tC HAZARD CHARACTERIZATION OF 69 WATER SYSTEMS.

NUCLEAR PLANT SITES EAST OF THE ROCKY '

MOUNTAINS.Ouestonnaires. DAVIS,D.A. '

NUREG/CR-5250 V08: SEISMIC HAZARD CHA14ACTERt2ATION OF 6d NUREG/CR-5143: APPLICATION OF THE J-INTEGRAL AND THE MODl-NUCLEAR PLANT SITES EAST OF THE ROCKY FIED J-INTEGRAL TO CASES OF LARGE CPACK EXTENSION.

MOUNTAINS. Supplementary Seismic Hazard Results For Sites With Multiple Soil Conditions. DAYlS.M.L CHEVERTON,R.D. NUREG/CR-5197: EVALUATION OF GENERIC ISSUE 115. " ENHANCE-l MENT OF THE RELIABILITY OF WESTINGHOUSE SOLID STATE NUREG/CR 5320: IMPACT OF RADIATION EMBRITTLEMENT ON IN*

PROTECTION SYSTEM."

TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR PLANTS-DEAN,R.S.

CHlNN,0.J. NUREG/CR 5292: CLOSEOUT OF IE BULLETIN 8023: FAILURES OF I NUREG/CR-4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEERING COOLANT PIPING. Pipe Failure Induced By Crack Growth And Failure CORPORATION.

Of Intermediate Supports.

DEFFENSAUGH,J.

I CHOU.C.K. .

NUREG/CR-4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) RELI- i NUREG/CR-4792 Vot: PROBABILITY OF FAILURE IN BWR REACTOR ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER I COOLANT PIPING. Summary Report. REACTORS. Semiannual Report. April September 1987.  !

CHU,T-L. DIMENNA,R.A.

NUREGICR.5124: INTERFACING SYSTEMS LOCA. BOILING WATER NUREG/CR-5074: DEVELOPMENT OF A PH NOMENA IDENTIFICA-REACTORS.

TION AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAULIC CHUN.R.C. PHENOMENA DURING A PWR LARGE.BRE/ K LOCA.

NUREG/CR 4554 V02: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A NUREG/CR-5311: AN ASSESSMENT OF REl APS/ MOD 2 APPLICABIL.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING ITY TO LOSS.0F FEEDWATER TRANSIEP T ANALYSIS IN A BAB.

CASK DESIGN REVIEW. Volume ?: Theory Manual - Impact Analysis. COCK AND WILCOX REACTOR PLANT.

NUREG/CR 4554 V03: SCANS (SHIPPING CA5K ANALYSIS SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSliM FOR SHIPPING DOCTOR M CASK DESIGN REVIEW. Volume 3: Theory Manuat - Lead Slump in NUREG/CR4849 V01: STEAM GENERP.(OR GROUP PROJECT. Task 9 Impact Analysis And Venfication Of impact Analysis. Final Report: Nondestructive Ev6,siten Round Robin. Volume 1: De-senpton And Summary D*

CI Al80RNE.E. NUREG/CR4849 V02 W .. Gr.NERATOR GROUP PROJECT. Task 9 NUREG/CH-4627 R01: GENERIC COST ESTIMATES. Abstracts Frorr, Final Report: Nonda rJClive Evaluaton Round Robin. Volume 2: Raw Genenc Studies For Use In Prepanng Regulatory impact Anatyses. Inspecton Data.

PIrsrn11 Author Index 23 DOCTOR,5JL - GERTMAN,D.L NUREG/CR4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) FtEll- NUREG/CR-4639 V5P2A1: NUCLEAR COMPUTERIZED LIBRARY FOR ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 2:

REACTORS. Semiannual Report. April-September 1987. Human Error Probability (HEP) Eshmates.

NUREG/CR-4639 V5P3R1: NUCLEAR COMPUTERIZED LIBRARY FOR dC 5069: TRAC PF1/ MOD 1. Correlations And Models. #[dw H aueO CFD PURHAM J.S NUREG/CR 4639 V5P4R1- NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CFi 5276: SADDE (SCALED ABSORBED DOSE DISTRIBUTION ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part EVALUA(OR).A Code To Generate input For VARSKIN. 4: Summary Aggregations.

ESCHGACH,E.J. GILBERT,8.G.

NUREG/CR-5186: VALUE/ IMPACT ANALYSIS Or GENERIC ISSUE 94, NUREG/CR 4639 V5P2R1: NUCLEAR COMPUTERIZED LIBRARY FOR

" ADDITIONAL LOW TEMPERATURE OVERPRESSURE FOOTECTION ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 2:

FOR LIGHT WATER REACTORS

• Human Error Probability (HEP) Estimates.

N'JREG/CR4639 V5P3R1: NUCLEAR COMPUTERIZED LIBRARY FOR EVANS,0.D.

ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 3:

NUREG/CR-5239: FLUID FLOW AND SOLUTE TRANSPORT MODEL- Hardware Component Failure Data (HCFD).

ING THROUGH THREE-DIMENSIONAL NETWORKS OF VARIABLY NUREG/CR4639 V5 PAR 1: NUCLEAR COMPUTERIZED LIBRARY FOR SATURATED DISCRETE FRACTURES. ASSESSING REACTOR RELIABILITY (NUCLAR.R). Data Manel.Part 4: Summary A9gregahons.

FERRIS R.H.

NUREG/CR4849 V01: STEAM GENERATOR GROUP PROJECT. Task 9 GILMORE,W.E.

Final Report: Nondestructive Evaluaton Round Robin. Volume 1: De' NUREG/CR4639 V5P2RI: NUCLEAR COMPUTERIZED LIBRARY FOR NU R 84 0 TEAM GENERATOR GROUP PROJECT. Task 9 ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual.Part 2:

Final Report: Nondestructive Evaluation Round Robin. Volume 2: Raw H "

NURE C 63 UC AR MPUTERIZED LIBRARY FOR

'""P' " "#

ASSESSING hEACTOR RELIABILITY (NUCLARR). Data Manual. Fart ":

FITZPATRICK,R. Hardware Component Failure Data (HCFD).

NUREG/CR-5102: INTERFACING SYSTEMS LOCA. PRESSURIZED '

NUREG/CR4639 VCP4R1: NUCLEAR COMPUTERIZED LIBRARY FOR WATER REACTORS. ASSESSING REACTOR RELIABILITY (NUCt.ARR). Data Manual,Part NUREG/CR-5124: INTERFACING SYSTEMS LOCA: BOILING WATER 4: Summary A. aggregations.

REACTORS.

NUREG/CR 5206: A PROBABit'STIC EVALUATION OF THE SAFETY GODFREY,P.

OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK 1 WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL CONTAINMENT IMPROVEMENTS.

EVENTS.

GOOD,M.S.

FLEMING.K. NUREG/CR4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) RELi-NUREG/CR 4780 V02: PROCEDURES FOR TREATING COMMON ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER CAUSE FAILURES IN SAFETY AND RELIABILITY STUDIES.Analyhcal REACTORS. Semiannual Report, April September 1987.

Background And Techniques.

GORE.8.F.

NUR UR 5292: CLOSEOUT OF IE BULLETIN 8023: FAILURES OF G 5186: MENW WMS & WN ESW %

SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEERING FOR LIGHT WATER REACTORS.

CORPORATION.

FOX,M.J. GORMAN,J. .I NUREG/CR-5115: A F.EVIEW OF BOILING WATER REACTOR WATER NUREG/CR 5116: SURVEY OF PWR WATER CHEMISTRY.

CHEMISTRY. Science, Technology,And Performance.

GREEN,E.R.

FRANZ,0.D. NUREG/CR4469 V07. NONDESTRUCTIVE EXAMINATION (NDE) RELi- i NUREG/CR4496 V02: A SYSTEM FOR GENERATING LONG STREAM- ABILITY FOR IN3ERVICE INSPECTION OF LIGHT WATER i FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN REACTORS.Semiar.nual FleporLApril-September 1987, PERIOD. Phase IL GREGG,R.E.

FUENKAJORN,K. elVREG/CR S112: EVALUATION OF BOILING WATER REACTOR NUREG/CR-5243: BOREHOLE CLOSURE IN SALT. WATER LEV' d L SENSING LINE BREAK AND SINGLE l FAILURE Genenc lasue 101 Boiling Water Reactor Level Redundancy-  ;

GALYEAN,W.J.

NUREG/CR-4639 V5P2RI: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/C -51 E /ALUATION OF GENERIC ISSUE 115. " ENHANCE.

ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual Part 2:

Human Error Probability (HEP) Eshmates. MENT OF THE RELIAB,,ILITY OF WESTINGHOUSE SOLID STATE NUREG/CR4639 V5P3R1: NUCLEAR COMPUTERIZED LIBRARY FOR PROTECTION SYSTEM.

ASSESSING REACTOR RELIAalLITY (NUCLARR). Data ManualPart 3: GREINER,S.E NUF /C ' R L PUTERIZED Ll83RARY FOR NUREG/Cfh5069: TRAC.PF1/ MOD 1.Correlabons And Models.

ASSESSING REACTOR RELIABILITY (NUCLARR)Da0 Manual? art GUFFEE,LA.

4-Summary Aggregacns.

NUREC/CR-5%~t TRAC @Ft/ MOD 1.Correlabons And Models.

GAVER,D.P.

NUREG/CR 5324: DETECTING COMPONENT FAILURE POTENTIAL GUTHRIE,0.L USING PROPORTIONAL HAZARD MODEL NUREG/CR-5328. CORRELATIONS BETWEEN POWER AND TEST RE.

ACTOR (%Ia BASES.

GE6HARD,M.A.

i NUREG/CR4554 VC1: SCANS (SHIPPING CASK ANALYSIS SYSTEM) A HACKETT E.M.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING NUREG/CR-5143: APPLICATION OF THE J-INTEGR4. AND THE MODI-CASK DES:3N REVIEW. Volume 1: User's Manual To Versson la (In- FIED J-lNTEGRAL TO CASES OF LARGE CRACK EXTENSION. l NI hC kV05 C S (SWIPPING CASK ANALYSIS SYSTEM) A HALL.D.G.

MICROCOMF'JTER SASED ANALYSIS JYSTFM FOH SHIPPING NUREG/CR-5311: AN ASSESSMENT OF RELAP5/ MOD 2 APPLICABIL.

CASK C.ESIGN REVIEW. Volume 5: Theory Manual - Thermal / Pressure ITY TO LOSS OF FLEDWATER 1RANSIENT ANALYSIS IN A BAB-Stress Analysis. COCK AND WILCOX REACTOR PLANT.

--__ ._-__________ _ _ . __ ._. _ __ __ _0

24 Person:1 Authrr ind3x HANAN N A. JOHNSON,0.L NUREG/CR 524i A REVIEW OF THE CRYSTAL RIVER UN'T 3 PROB- NUREG/CR-4554 V01: SCANS (SHIPPING CASK ANALYSIS SYSTEM)A l ABillSTIC RISK ASSESSMENT. Internal Events, Core Damage Frequen- MICROCOMPUTER BASED ANALYLIS SYSTEM FOR SHIPPING 1 cy. CASK DESIGN REVIEW. Volume 1: User's Manual To Version 1a (In- j ciuding Program Reference) <

HANEY,LN.

NUREG/CR-4554 V04: SCANS (SHIPP.NG CASK ANALYSIS GYSTEM)A NUREG/CR-4835: COMPARISON AND APPLICATION OF OUANTITA- MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING TIVE HUMAN REUABl1JTY ANALYSIS METHODS FOR THE RISK CASK DESIGN REVIEW. Volume 4: Theory Manual . Thermal Analysis.

METHODS. ' INTEGRATION AND EVALUATION PROGRAM JRMIEP) Final Report. JOHNSON,K.L NUREG/CR 5234: VALUE/ IMPACT ANALYSIS FOR GENERIC ISSUE HARRIS,M.S. 51: IMPROVING THE REUABILITY OF OPEN-CYCLE SERVICE.

NUREG/CR-5186: VALUE/lMPACT ANALYSIS OF GENERIC ISSUE 94, WATER SYSTEl IS.

" ADDITIONAL LCr# TEMPERATURE OVERPRESSURE PROTECTION FOR UGHT WATER REACTORS." . JOHNSON,W.B.

NUREG/CR-4950 V03: THE SHORELINE ENVIRONMENT ATMOS.

H?,RTY,H.

PHERIC DISPERSION EXPERIMENT (SEADEX).Asrborne UDAR Data.

NUREG/CR-4849 V01: STEAM GENERATOR GROUP PROJECT. Task 9 Final Report: Nondestructive Evaluation Round Robin. Volume 1: De. JOYCE.J.A.

scription And Summary Data. NUREG/CR-5143: APPLICATION OF THE J INTEGRAL AND THE MODI-NUREG/CR-4849 V02: STEAM GENERATOR GROUP PROJECT. Task 9 FIED J INTEGRAL TO CASES OF LARGE CRACK EXTENSION.

Firial Report: Nondestructive Evaluation Round Robin. Volume 2: Raw inspection Data, JUN,C.-H.

NUREG/CR-5271: THERMOCHEMICAL . PROPERTIES OF HASHIMOTO.P.S. GIBBSirE BAYERITE,00EHMITE, DIASPORF AND THE ALUMINATE NUREG/CR-5176: SEISMIC FAILURE AND CM DROP ANALYSES OF IOff BETWEEN O AND 350 DEGREES C.

THE SPENT FUEL Pont 5 AT TWO REPRESENTATIVE NUCLEAR POWER PLANTS. KASSNER,T.F.

NUREG/CR 4667 V05: ENVIRONMENTALLY ASSISTED CRACKING IN HAY S,R.A. UGHT WATEP REACTORS. Semiannual Rept, April-September 1987.

NUREG/CR-5143: APPLICATION OF THE J-INTEGRAL AND THE MODI-KATO,W.Y.

FIED J-INTEGRAL TO CASES OF LARGE CRACK EXTENSION.

NUREG/CP 5325: A MULTIVARIATE STATISTICAL STUDY ON A Dl-HEASLER.P.G. . VERSIFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER NUREG/CR-4460 V07: NONDESTRUCTIVE EXAMINATION (NDE) REU. PLANIS.

ADluTY FOR INSERVICE INSPECTION OF UGM WATER I REACTORS.Seminnnual Report. April-September 1987. KENNEDY,R P. '

HENLEY,D R. METHODS.

NUREG/CR-5245: A REVIEW OF THE CRYSTAL RIVER UNIT 3 PROB-KI Y R.W.

ABILISTIC RISK ASSESSMENT. internal Events. Core Damage Frequen. g EUNG OF A LOW LEVEL WASTE SHALLOWsLAND BURIAL SITE.A HENNICK,A. Comparison Of Predictwe Radeonuclih Transport Modeling Versus NUREG/CR-5292: CLOSEOUT OF IE 0ULLETIN 80-23: FAILURES OF Field Otmervahons At The "A" D%oosal Area. Chalk Rwer Nuclear Lab.

SOLENOID VALVES MANUFACTURED BY VALCOFi ENGINEEFUNC oroturies.

KIMURA,C.Y.

HE'8SF D.J. NUREG/CRP42 S02: EVALUATION OF EXTERNAL HAZARDS TO NU-NUREG/CR-4635: COMPARISON AND APPUCATION OF QUANTITA. CLEAR POWER PLANTS IN THE UNITED STATES Other External TiVE HUtdAN HEUABILITY ANALYSIS METHODS FOR THE RISK Everits l METrtOOS INTEGRATION AND EVALUATION PROGRAM NUREG/CR-5176: SEISMIC F ALLURE AND CASK DROP ANALYSES OF ]

(RMIEP). Final Report. THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR l POWER PLANTS.

HtSER.A.L NUR "'N T CR-5265: SIZE EFFECTS ON J-R CURVES FOR A 302-D N

REPORT FOR THE MOOULAR HIGH-TEMPERATURE GAS COOLED HOLFORD.D. REACTOR.

NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD-EUNG OF A LOW LEVEL WASTE SHALLOW-LAND DURIAL SITE.A Companson Of Predictwe Radionuchde Transport Modehng Versus REG /CR-4478: UPDATE A FORTRAN 77 SOURCE FILE Field Observehnns At The "A" Disposal Area. Chalk Piver Nuclear Lal MANIPULATOR Adap*ed For The Data General MV Senes Eclipse oratones. Computers Under AOS/VS.

KNIGHT T.D.

NUREG b 4792 V01: PROBABluTY OF FAILURE IN BWR REACTOR * "" "'

COOLANT PIPING Summary Report.

KNUDSON R.

NUREG/CR4702 V02: PRO 8ADIUTY OF FAILURE IN BWR REACTOR NUREG/CR-5278 COST ANALYSIS FOR POTENTIAL BWR WL M i '

COOLANT PIPING Pipe Fadare indLced By Crack Grow'h And Sailure Of intermediate Supports. CONTAINMENT IMPROVEMENTS.

H8U,C.J. KOENIG R.A.

NUREG/CR-5251: WELLFIELD INSTALLATION AND iWEEG/CR-5206: A PROBABILISTIC EVALUATION OF THE SAFCTY INVESTIGATIONS CRESTON STUDY AREAE6 STERN WASHING-OF OAEOCK & WILCOX NUCLEAR REACTOR POWER PLANTS TON.

WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL EVENTS, KOHUT.P. l JENKiNS.J.P. NUREG/CR.5102: INTERFACING SYSTEMS LOCA.PRESSURt2ED i WATER REACTORS.

NUREG/CR4f05: COMPARISON AND APPUCATION OF QUANTITA-TIVE HULJN REUADIUTY ANALYSIS METHODS FOR THE RISK KOLACZKOWSKI.A.

METHOOS INTEGRATION AND EVALUATION PROGRAM NUREG/CH4313. EQUIPMENT QUAUFICATION (EO)-RISK SCOPING (RMIEP) Final Report. STUDY. j

PIrsonIl Author index 25 t i

KRAEGER.B.A. MCCALLEN D.B.

. NUREG/CR-4496 V02: A SYSTEM FOR GENERATING LONG STREAM, NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR PERIOD. Phase 11. POWER PLANTS.

KULLBERG.C.M. MCGRATH,K.A.

NUREG/CR-5311: AN ASSESSMENT OF RELAP5/ MOD 2 APPLICABIL- NUREG-1350 V01: IMICLEAR REGULATORY COMMISSION 1989 IN-(TY TO LOSSOF-FEEDWATER TRANSIENT ANALYSIT IN A BAB- FORMATION DIGEST' COCK AND WILCOX REACTOR PLANT.

MCISAAC,0.V.

~

URE / -5251- WELLFIELD INSTALLATION AND EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPERATING NU-INVESTIGATIONS,CRESTON STUDY AREA, EASTERN WASHING- CLEAR pop /ER PLANTS.

TON.

MCNAMARA,N.

LAMBRIGHTJ.A.

NUREG/CR-5088. FIRE RISK SCOPING STUDY: INVESTIGATION OF NUREG 0837 V08 NO3: NRC TLD DIRECT RADIATICN MONITORING h' NUCLEAR POWER PLANT FIRE RISK.lNCLUDING PREVIOUSLY UN. NETWORK. Progress Report. July-Septereber 1988.

f. ADDRESSEOISSUES.

MCNEILJ.M.

l LAR80NJ.R. NUREG/CR-5271: THERMOCHEMICAL PROPERTIES OF l NUREG/CR 53tt: AN ASSESSMEN7 OF RELAPS/ MOD 2 APPLICABIL. GIBBSITE.BAYERITE.BOEHMITE. DIASPORE AND THE ALUMINATE ITY TO LOSS.OF-FEEDWATER TRANSIENT ANALYSIS IN A BAB- ION BETWEEN D AND 350 DEGREES C.

COCK AND WILCOX REACTOR PLANT, LARSON.T.K. NUREG/CR 5313: EQUIPMENT QUAUFICATION (EO)-RISK SCOPING NUREG/CR-5074: DEVELOPMENT OF A PHENOMENA IDENTIFICA- STUDY.

TION AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAUUC PHENOMENA DUR$3 A PWR LARGE-BREAK LOCA. MENSING.R.W.

NUREG/CF4311: AN ASSESSMENT OF RELAP5/ MOD 2 APPLICABIL- NUREG/CR4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR ITY TO LOSSOF-FEEDWATER TRANSIENT ANALYSIS IN A BAB- COOLANT PlPING Pipe Failure Induced By Crack Growth And Failure COCK AND WILCOX REACTOR PLANT. Of intumediate Supports.

NUREG/CR-5250 V01: SEISMIC HAZARD CHARACTER 12ATION OF 69 N E'G R 5325: A MULTIVARIATE STATISTICAL STUDY ON A DI. M UNTAINS. Methodology, input Data And Compensons To Previous I

VERSlFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER Results For Ten Test Sites.

PLANTS. NUREG/CR-5250 V02: SEISMIC HAZARD CHARACTER 12ATION OF 69 ULES,D.R. NUCLEAR PLANT SITES EAS* OF THE ROCKY MOUNTAINS.Results NUREG/CR-5069: TRACsPF1/ MODI Cnr clations And Modele.

NUR G 5 50 V$3 ESM HA A D CHARACTERl2ATION OF 69 l LINSLEY,R.K. NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results NUREG/CR 4496 V02: A SYSTEM FOR GENERATING LONG STREAM. And Discussion For The Batch 2 Sites.

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETilRN NUREG/CR-5250 V04: SEISMIC HAZARD CHARACTER 12ATION OF 69 PERIOD. Phase 11. NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results  ;

Arvi Discussion For Tha Batch 3 Sites. j LO.T. NUREG/CR 5250 V05: SEISMIC HAZARD CHARACTERIZATION OF 69 l NUREG/CR-4554 V03: SCANS (SHIPPING CASK ANALYSIS SYSTEMLA NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results i MICROCOMPUTER DASED ANALYSIS SYSTEM FOR SHIPPING And Oscussio1 For The Batch 4 Sites.

CASK DESIGN REVIEW. Volume 3: Theory Manual - Lead Slump in NUREG/CR-5250 V06: SEISMIC HAZAAD CHARACTER 12ATION OF 69 Impact Analysis And Verification Of impact Analysis. NUCLEAR PLANT CITES EAST OF THE ROCKY NUREG/CR 4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR MOUNTANS. Regional Comparison Between Sites. Site Effects, General COOLANT PIPING. Pipe Failure Induced By Crack Growth And Failure Dscussion And Conclusions.

Of Intermediate Supports. NUREG/CR-5250 V07: SEISMIC HAZARD CHARACTERIZATION OF 69 )

NUCLEAR PLANT SITES EAST OF THE ROCKY j LOFGREN.E.V NUREG/CRI4618: EVALUATION OF RELIABILITY TECHNOLOGY AP. NUREG SEI iC HAZARD CHARACTERIZATION OF 69 PLICABLE TO LWR OPERATIONAL SAFETY. NUCLEAR PLANT SITES EAST OF THE ROCKY LOY.T.

MOUNTAINS. Supplementary Sessmic Hazard Results For Sites With NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF THE Multiple Soil Conditions.

WESTERN OHIO-INDIANA REGION. Annual Report. October 1987 September 1988' MILLER.B.D.

NUREG/CR-5276: SADDE (SCALED ABSORBED DOSE DISTRIBUTION LUCAS.G.E. EVALUATORtA Code To Generate input For VARSKIN.

NUREG/CR-5030: AN ASSESSMENT OF eTEAM-EXPLOSION IN-DUCED CONTAINMENT FAlLURE. MWTON.L.A.

NUREG/CR-4835: COMPARISON AND APPLICATION OF QUANTITA-LUCKAS,W.J. TIVE HUMAN RELIABILITY ANALYSIS METHODS FOP THE RISK NUREG/CR-4946: TECHNICAL FINDINGS RELATED TO GENERIC METHODS INTEGRATION AND EVALUATION PROGRAM ISSUE 23: REACTOR COOLANT PUMP SEAL FAILURE. (RMIEP) Final Report.

MAHAFFYJ.H. MOK,0.C.

NUREG/CR 5069: TRAC.PF1/ MOD 1. Correlations And Models. NUREG/CR4554 V01: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A MICROCOMPUTER BASED A11ALYSIS SYSTEM FOR SHIPPING CASK CESIGN REVIEW Vetume C Users Manyal To Version 1a (In-MAlYA.P.S' NUREG/C R 4667 V05: ENVIRONMENTALLY ASSISTED CRACKING P R INe LIGHT WATER REACTORS. Sermannual Ropt. April-September 1987.

NU C y e((GHIPPING N CASK ANALYSIS )

SYSTEM MANDLET.J.W. MICROCOMPUTER BASED ANMYSIS SYSTEM FOR SHIPPING NUIREG/CR-5224: THE TEACHABILITY OF DECONTAMINATION ION. CASK DESIGN REVIEWVolume 3: Theory Manua! - Lead Stump in EXCHANGE RESINS SOLIDIFIED 'N CEMENT AT OPERATING NU- Impact Analysis And Venfication Of impact Analysis.

CLEAR POWER PLANTS.

MOLTYANER.G.L MCCABE.D.E. NURfG/CR 46?9 V02: DEMONSTRATION OF PERFORiJIAf4CE MOD-NUREG/CR-5326: FRACTURE EVALUATION OF SURFACE CRACKS EUNG OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A EMBEDDED IN REACTOR VESSEL CLADDING. Companson Of Predictive Radionuclides Transport Modeling Versus i

l 1

J

26 . Persord Auth:r index z Field Observations At The *A" Disposal Area, Chalk River Nuclear Lab- PARRY,0.

oratones.

NUREG/CR-4700 V02: PROCEDURES FOR TREATING COMMON CAUSE FAILURES IN SAFETY AND REUABluTY STUDIES. Analytical a ground And Techniques.

G/ R-5112: EVALUATION OF BOILING WATER REACTOR WATER-LEVEL SENSING . LINE BREAK AND SINGLE PASAMEHMETOOLU F ALLURE. Generic lasue 101 Boiling Water Reactor Level Redundancy.

Technical Firdngs. NUREG/CR 5069: TRAC-PF1/ MOD 14.,&lr' ions And Models.

MORLEY,8.M. PASCHIS,J.A.

i^

NUREG/CR4950 V03: THE SHOREUNE ENVIRONMENT ATMOS- NUREG/CR-5251: WELLFIELD INSTALLATION AND PHERIC DISPERSION EXPERIMENT (SEADEX).Arborne UDAR Data. INVESTIGATtGNS,CRESTON STUDY AREA, EASTERN WASHING.

TON- i MOSLEH.A. "

t NUREG/CR4780 V02: PROCEDURES FOR TREATING COMMON PAULA.H.

CAUSE FAILURES IN SAFETY AND REUABILITY STUDIES. Analytical Background And Technklues.

NUREG/CR-4780 V02: PROCEDURES FOH TREATING COMMON CAUSE FAILURES IN SAFETY AND REUABluTY STUDIES. Analytical j L'URRAY,R.C. Background And Techniques. -

NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF PENNELL,W.E T E SPE I UEL POOLS AT TWO REPRESENTATIVE NUCLEAR NUREG/CR 5320; IMPACT OF RADIATION EMBRITTLEMENT ON IN- 0 NUREG/CR-5270. SSESSMENT OF SEtSM'C MARGIN CALCULATION TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO P5NR ' )

METHODS. PLANTS.

t,5YERS,0.A. PITTIGUO,C.L.  ;

NUREG/CR4879 V02: DEMONSTRATION OF PERFORMANCE MOD- NUREG 1293: OUAUTY ASSURANCE GUIDANCE FOR LOW-LEVEL i ELING OF A LOW LEVEL WASTE SHALLOW. LAND BURIAL SITE.A RADIOACTIVE WASTE DISPOSAL FAplLITY. Final Report.

Compenson Of Predictrve Radionuclides Transport Modeling Versus s Field Observations At The "A" Disposal Area. Chalk River Nm: lear Lats PITTMAN.J.W.  !

oratoneS. NUREG 1319: A PRIORITIZATION OF RESEARCH ACTIVITIES.

NAFDAY,A.M. PLUMLEE,G.L NUREG/CR 5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF NUREG-1275 V04: OPERATING EXPERIENCE FEEDBACK REPORT -

THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR POWER PLANTS. TECHNICAL SPECIFICATIONS. Commercial Power Reactors.

NAJAFI,B. N8KM NUREG/CR-5030: AN ASSESSMENT OF STEAM-EXPLOSION-IN- NUREG/CR 5112: EVALUATION OF BOILING WATER REACTOH DUCED CONTAINMENT FAILURE. WATER LEVEL SENSING LINE B9EAK AND SINGLE FAiuME. Generic tasue 101 Boiling Water Recctor Level Redundancy-NANSTADAK. Technical Findings. 1 NUREG/CR.5320: IMPACT OF RADIATION EMBRITTLEMENT ON IN-TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR PRASSINOS,P.G. .

PLANTS. NUREG/CA 5042 S02: EVALUATION OF EXTERNAL HAZARDS TO NU- i NELSONAA. vent

'NUREG/CR-5069: TRAC-PF1/MODt.Conciations AnJ Mode;s.

NUREG/CR 5176: SEISMIC FAILURE AND CASK OROP ANALYSES OF NELSON,T.A. THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR NUREG/CR-4554 V02: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A POWER PLANTS.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING CASK DESIGN REVIEWVolume 2: Theory Manual impact Analysis. M N RE C'R 780 ' V02: PROCEDURES FOR TREATING COMMON NICOLETTE V.F. CAUSE FAILURES IN SAFETY AND REUABtLITY STUDIES Analytical NUREG/CR 5088: FIRE RISK SCOPING STUDY: INVESTIGATION OF Background And Techniques.

NUCLEAR POWER PLANT FIRE RISK,1NCLUDING PREVIOUSLY UN-ADDRESSEDISSUES. RASMUSSEN,T.C.

NUREG/CR-5239: FLUID FLOW AND SOLUTE TRANSPORT MODEL.

N E/CR4950 V03: THE SHOREUNE ENVIRONMENT ATMOS- S TURAT SR F URE PHERIC DISPER$10N EXPERIMENT (SEADEX).Airborno LIDAR Data.

NOWLEN.S.P. RAVINDRA.M.K.

NUREG/CR 5088: FIRE RISK SCOPING STUDY: INVESTIGATION OF NUREG/CR-5176: SEtSMIC FAILURE AND CASK DROP ANALYSES.0F NUCLEAR POWER PLANT FIRE RISK,1NOLUDING PREVIOUSLY UN- THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUOLEAR ADDRESSEDISSUES. POWER PLANTS.

NUREG/CR-5270: ASSESSMENT OF SEISMIC MARGIN ( ALCULATION O'RElLLY,P.D. METHODS.

NUREG-1275 V04: OPERATING EXPERIENCE FEEDBACK REPORT .

TECHNICAL SPECIFICATIONS Comme cial Power Reactors. REECE,W.D.

NUREG/CR-5276: SADDE (SCALED AB$ ORBED DOSE DISTRIBUTION URE CR4879 V02: DEMONSTRATION OF PERFORMANCE MOD-

^ ' " " * '"

EUNG OF A LOW LEVEL WASTE SHALLOW. LAND BUill&L SITE.A REED.J.W.

Compenson Of Predictive Radiosiuclide Transport Modeling Versus Aeid Observahons At The "A" Disposal Area, Chalk River Nuc6 .A* NUFIEG/CR-5270: ASSESSMENT OF SEISMIC MARGIN CALCULATION WETHOM oratoriett OWC 1.AR$1u,P.U REN S A.

NUREG/CR 3037; USER'S MANUAL FOR FIRIN.A Computer Code To NUfiEG/CR.5197: EVALUATION OF GENERIC ISSUE 115. " ENHANCE-Estimate Accidental Fire And Radoective Airbome Release in Nuclear MENT OF THE REUABILITY OF WESTINGHOUSE SOUD STATE Fuel Cycle Facilibes. PROTECTION SYSTEM?

PARK,J.Y. RIORDAN.B.

NUREG/CR4667 V05: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK l UGHT WATFR REACTORS. Semaannual Rept.Apnt-September 1987. CONTAINMENT IMPROVEMENTS.

P;rsonal Author inder 27 ROBERTSON,D.E. NUREG/CR-5278: COST ANALYSIS FOR SOTENTIAL BWR MARK I NUREG/CR 4879 V02: DEMONSTRATION OF PERFORMANCE MOD- CONTAINMENT IMPROVEMENTS.

ELING OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A {-

Companson Of Predictrve Radionuclides Transport Modehng Versus SHACK.W.J.

Field Observations At The "A" Disposal Area. Chalk River Nuclear Lab- NUREG/CR-4687 V05: ENVIRONMENTALLY ASSISTED CRACKING IN oretones. LIGHT WATER REACTORS. Semiannual Rept April September 1987.

ROSINSON,G.C. SHAPIRO,A.P.

NUREG/CR-5320: IMPACT OF RADIATION EMBRITTLEMENT ON IN. NUREG/CR-4554 V04: SCANS (SHIPPING CASK ANALYSIS SYSTEM):A TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING PLANTS. CASK DESIGN REVIEW. Volume 4: Theory Manual Thermal Analysis.

ROSE,8.E SHAW,R.A.

NUREG/CR-4835: COMPARISON AND APPLICATION OF QUANTITA.

NUREG/CR-5074. DEVELOPMENT OF A PHENOMENA IDENTIFICA-TlVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK METHODS INTEGRATION AND EVALUATION PROGRAM TION AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAULIC (RMIEP) final Report. PHENOMENA DURING A PWR LARGE-BREAK LOCA.

ROUHANI,8.Z. SHERWOOD,K.B.

NUREG/CR 5074: DEVELOPMENT OF A PUENOMENA IDENTIFICA. NUREG/CR-5069: TRAC-PFl/ MOD 1. Correlations And Models.

TION AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAULIC PHENOMENA DURING A PWR LARGE-BREAK LOCA. SHINA -

l RUGERAJ.

NUREG/CR-4948: TECHNICAL FINDINGS RELATED TO GENERIC SHIRE,P.R.

ISSUE 23 REACTOR COOLANT PUMP SEAL FAILURE. NUREG/C't 069: TRAC-PF1/ MOD 1. Correlations And Models.

RUMBLE,E. SIMlON,G.

NUREG/CR 5030: AN ASSESSMENT OF ETEAM-EXPLOGION-IN- NUREG/CR 4627 R01: GENERIC COST ESTIMATES. Abstracts From DUCED CONTAINMENT FAILURE- Generic Studies For Uss in Prepring Regulatory impact Analyses.

NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK l CONTAINMENT IMPROVEMENTS.

red CR'-4667 VOS: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual Rept April-September 1987. SIMONEN,E.P.

SAWANTA.P.K. NUREG/CR-5328 CORRELATIONS BETWEEN POWER AND TEST RE-NUREG/Cli-5324: DETECTING COMPONENT FAILURE POTENT;AL ACTOR DATA BASES.

USING PROPORTIONAL HAZARD MODEL.

NUREG/CR-5325: A MULTIVARIATE STATISTICAL STUDY ON A DI. Sl'AONEN,F.A.

VERSIFIED DATA GATHERING SYSTE'M FOR NUCL EAR POWER NUREG/CR-4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) TtEn PLANTS. ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report, April-September 1987.

NUREG/CR-5250 V01: SEISMIC HAZARD C4ARACTERIZATION OF 69 SOLTANI.P.K.

NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR-5100: INTEGRATING FIBER OPTIC RADIATION DOSIME.

MOUNTAINS. Methodology. input Data And Cvnpansons To Previous TER.

Resutta For Ten Test Sites.

NUREG/CR-5250 V02: SEISMIC HAZARD CHARAT.TERIZATION OF 69 SPANNER.J.C.

NUCLEAR PLANT SITES EAST OF THE ROCKY IJOUNTAINS.Resuta NUREG/CR4d69 V07; NONDESTRUCTIVE EXAMINATION (NDE) REll-NF / 5 V3 ES HA A ID CHARACTERIZATION OF 60

^

REACTORS. Semiannual Report,Apni September 1987.

NUCLEAR PLANT SITES EAST OF THE ROCKY MOJNTAINS Results And Dmcursion For The Batch 2 Sites SPANO,A H UC E LANT T AT K UNTAI S Resu t NUREG 4 REGULATORY ,

ALYSt FR HE RE L ION OF NUREG/CR 5250 V05: SEISM!C HAZARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS Results SPORE J.W.

And Discussion for The Batch 4 Sites. NUREG/CR-5069: TRAC-PF1/ MOD 1. Correlations And Models.

NUREC/CR-5250 V06: SEISMIC HAZARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY STEtNKE R.G.

MOUNTAINS Flegional Compenson Between Sites. Site Effects, General NUREG/CR-5069: TRAC-PF1/ MOD 1. Correlations And Models.

Discussion And Conclusions.

NUREG/C15250 V07: SEISMIC HAZARD CHARACTERIZATION OF 69 STEVENSON,J.D.

NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR-5270: ASSESSMENT OF SEISMIC MARGIN CALCULATION MOUNT AINS Questionnaires. METHODS.

NUREG/CR-5250 V08. SEISMIC HAZARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY STILE S,0.L MOUNTAINS Supplementary Sesmic Hazard Results For Sites With NUREG/CR-5234. VALUE/ IMPACT ANALYSIS FOR GENERIC ISSUE Multiple Soll Conditions $1:IMPROVtNG THE RELIABILITY OF OPEN. CYCLE SERVICE.

SCHWARTZ,1.

NUREG-0020 V12 N12: LICI'NSED OPERATING REACTOR! STATUS STORTI,G.M.

NL EG 20 V 1 fl S 0 Oki T

SUMMARY

REPORT. Data As Of December 309t)8.

b ot !

TAT.US TER'

^

NUREG.0020 V13 NO2: LICENSED OPERATING RE S T TUS STOYANOVS SUMnAARY BEPORT. Data As Of January 31,1989 (Gray Book I)

NUREG/dA 5124. INTERFACING SYSTEMS LOCABOILING WATER SCHWART1 S.y. REACTORS.

NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS C'F THE WESTERN OHIOINDIANA REClON Annual Report, October 1987 STRUCKMEYER,R.

September 1988. NUREG 0837 V06 NO3. NRC TLD DlHECT RADIATION (AONITORING NETWORK Progress Report. My. September 1988.

SCIACCA.F. l NUREG/CR-4627 R01: GENERIC COST ESTIMATES. Abstracts From STUMPF,H.J. 1 Genenc Studies For Use in Prepanng Regulatory impact Analyses. NUREG/CR.5069 TRAC-PF1/MODifstrela'tsons And Models. ]

1 l

. _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. J

28 P rson:1 Auth;r Ind;x .

l TAYLOR,T.T. W A TKINS,J.C.

NUREG/CR 4409 V07. NONDESTRUCTIVE EXAMINATION (NDE) RELl- NUREG/CR-5311: AN ASSESSMENT OF RELAPL/ MOD 2 APPLICABIL-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER ITY TO LOSSOF FEEDWATER TRANSIENT ANALYSIS IN A BAB-REACTORS.Semiannua! Report.Apra-September 1987.

COCK AND WILCOX REACTOR PLANT.

TEICHMANN,T.

WEAKLEY,S.A.

NUREG/CR 5325 A MULTIVAA1 ATE STATISTICAL STUDY ON A Di- ,

VERSIFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER NUREG/CR 5234: VALUE/ IMPACT ANALYSl$ FOR GENERIC ISSUE PLANTS 61:IMPDOV'NG THE RELIABILITY OF OPEN-CYCLE SERVICE-WATER ~ 11 EMS.

TERRELL.J.B.

NUREG/CR-5195: FATIGUL' STRENGTH OF ASME SA 106-8 WELDED WEISS,A.J.

MFETY RESEARCH PROGRAMS SPON-NU C $20 Ef hS bt JR OR A 302 B SORED BY OFFICE OF NUCLEAR REGULATORY PLATE' RESEARCH. Progress Report October December 1987.

THEOFANOUS,T.G. NUREG/CR 2331 V08 Nr. SAFETY RESEARCH PROGRAMS SPON.

NUREG/CR-5030: tsN ASSESSMENT OF STEAM-EXPLOSION-IN. SORED BY OFFICE OF NUCLEAR REGULATORY j DUCED CONTAINMENT F A! LURE. RESEARCH. Progress Report. July-September 1988.

THOMAS.C.W. WILLIAMS,P.M.

NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD- NUREG-1338. DRAFT PREAPPLICATION SAFETY EVALUATION ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL $1TE.A REPORT FOR THE MODULAR HIGH-TEMPERATURE GASCOOLED Compenson Of Predictive Radionuclido Transport Modeling Versus REACTOR.

f~ield Observations At The "A" Disposal Area. Chalk River Nuclear Lab-oratones. WILSON,J.N.

TONG,W.H. f4UREG-1333: DRAFT PREAPPLICATION SAFETY EVALUATION NUREG/CR 5170: SEISMIC FAILURE AND CASK DROP ANALYSES OF REPORT FOR THE MODULAR HIGH-TEMPERATURE GASS JOLED THE SPENT FUEL POOL $ AT TWO REPRESENTATIVE NUCLEAR REACTOR.

POWE9 PLANTS.

WITTE,M.C.

TRAGER.E.A.

NUREG/CR 4554 V03. SCANS (SHIPPING CASK ANALYSIS SYSTEM).A NURFG 1345: REVIEW OF EVENTS AT LARGE POOL TYPE IRRADIA- MICROCOMPUTER BASED ANALYSIS SYSTEM FCR SHIPPING TORS. CASK DESIGN REVIEW Volume 3. Theory Manual . Lead Slump Ir' TRUMMER.D.J. Impact Ahab/ nis And Ventilation Of Impact Analysis.

NUREG;0R-4554 VOL SCANS (SHIPPING CASK ANALYSIS SYSTEM).A MICROODMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING WORLEDGG CASK DGIGN REVIEW.V0Wme 1: User's Manual To Vessen 10 (In. NUREG/CR-4780 V02: PROCEDURES FOR TREATING COMMON cioding Progrkm Reference) CAUSE FAILURES IN SAFETY AND RELIABILITY STUDIES. Analytical NUREC/CR.4554 V02: SCAN. S (SHIP' PING CASK ANALYSIS SYSTEM).A Background And Techniques.

MICROCOMPUTER BASED ANALY3IS SYSEM 00R SHIPPING CASK DESIGN REVIEW Volume 2 Theory Man sal - Impact Analysis. WRIGLEY,C.Y.

NUPLG/CH 4554 V05. SCANS (SHIPPING CASK ANALYSIS SYSTEM).A NUREG/CR 5100: INTEGRATING FIBER OPTIC RADIATION DOSIME-MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING TER CASK DESIGN REVIEW. Volume $: Theory Manual . Thermal / Pressure Stress Analysis. YOUNG,C.J.

UTHE,E.E, NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF THE NUREG/GR-4050 V03. THE SHORELINE ENVIRONMENT ATMOS. WESTERN OHlO-INDIANA HEGION Annual Report, October 1987 PHERiC DISPER$10N EXPERIMENT (SEADEX) Airborne LIDAR Dufa. September 1988.

VALDEZ,G.D. YOUNG,J.L.

NUREG/CR-5173 REVIEW AND ASSESSMENT OF THERMODYNAMIC NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD-AND TRANSPORT PROPERTIES FOR THE CONTAIN CODE. ELING OF A LOW L5 VEL WASTE SHALLOW L/ND BURIAL SITE.A VALENTI,L N Companson Of Predictive Radionuclida Transport Modeling Versus Fdd Observations At The "A" Disposal Area. Chalk River Nuclose Lab.

NUREG/CR 5197: EVALUATION OF GENERIC ISSUE 115. " ENHANCE-oratonn MENT OF THE RELIABILITY OF WESTINGHOUSE SOLID STATE  ;

PROTECTION SYSTEM "

YOUNGBLOOD,R.W.

VO,T.V. NUREG/CR 5200, A PROBABILISTIC EVALUATION OF THE SAFETY I NUREG/CR-5166: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE 94 OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS l

" ADDITIONAL LOW TEMPERATURE OVERPRESSURE PROTECTION WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL FOR LIGHT WATER REACTORS." EVENTS l

i m__._._

j

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

! are welcome, i.

A 302-8 Plate Basalt NUREG/CR-5265: SIZE EFFECTS ON J-R CURVLS FOP A 302-B NUREG/CR-5251: WELLFIELD INSTALLATION AND PLATE. INVESTIGATIONS.CRESTON STUDY AREA. EASTERN WASHING-TON.

AOS/VS NUREG/CR4478: UPDATE - A FORTRAN 77 SOURCE FILE Bayertte MANIPULATOR. Adapted For Tre Data General MV. Series Echpse NUREG/CR 5271: THERMOCHEMICAL PROPERTIES OF Computers Under AOS/VS. GIBBSITE,BAYERITE,BOEHMITE, DIASPOPE AND THE ALUMINATE ION BETWEEN O AND 350 DEGREES C.

ASME Code NUREG/CR 5195: FATIGUE STRENGTH OF ASME SA 106-0 WELDED Biodegradation STEEL PIPES IN 288 DEGREES C AIR ENVIRONMENTS.

NUREG/CR 5221: BIODEGRADATION OF ION-EXCHANGE MEDIA.

Abnormal Occurrence l

^ N E R 47 R NC .J ember 19 . 1: PROBABluTY OF FAILURE IN BWR REACTOR COOLANT PIPING. Summary Report Accident NUREG/CR4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR COOLANT PIPING. Pipe Failure induced By Crack Growth And Failure NUREG 1355: THE STATUS OF RECOMMENDATIONS OF THE PRESI-Of intermediate Supports.

DENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND.A Ten Year Review. NUREG/CR-5112: EVALUATION OF BOluNG WATER REACTOR WATER-LEVEL SENSING LINE BREAK AND SINGLE ,

Accidental Fire FAILURE.Genene issue 101 Bolling Water Reactor :.evel Redundancy-  !

NUREG/CR 3037: USER'S MANUAL FOR FIRIN.A Computer Code To Technical Findings.

Estimate Accidentaf Fire And Radioactive Airborne Release in Nuclear NUREG/CR-5115: A REVIEW OF BOILING WATER REACTOR WATER Fuel Cycle Facilities. CHEMISTRY. Science, Technology,And Performance.

NUREG/CR-5124: INTERFACING SYSTEMS LOCA:HOlUNG WATER Al borne Sampling REACTORS.

NUREG/CR-4950 V03: THE SHORELINE ENVIRONMENT ATMOS. NUREG/CR 5278: COST ANALYSIS FOR POTENTIAL BWR MAFX 1 PHERIC DISPERSION EXPERIMENT (SEADEX). Airborne UDAR Data. CONTAINMENT IMPROVEMENTS.

Aquifer Test Borehole NUREG/CR-5251: WELLFIELD WSTALLATION AND NUREG/CR-5243; BOREHOLE CLOSURE IN SALT, INVESTIGATIONS,CRESTON STUDY AREA EASTERN WASHING-TON. Budget NUREG-1100 V05: BUDGET ESTIMATES. Fiscal Years 1990-1991.

NUREG/CR4950 V03: THE SHORELINE ENVIRONMENT ATMOS- Building Structure l l

PHERIC DISPERSION EXPERIMENT (SEADEX).Asrborne UDAR Data. NUREG/CR-3006: DAMPING IN BUILDING STRUCTURES DURING EARTHQUAKES. Test Data And Modeling.

l SWR ]

NUREG/CR-4792 V01: PROBABluTY OF FAILURE IN BWR REACTOR CONTAIN Code COOLANT PIPING. Summary Report. NUREG/CR-5173: REVIEW AND ASSESSMENT OF THERMODYNAMIC l NUREG/CR4792 V02: PROBABILITY OF FAILURE IN BWR PEACTOR AND TRANSPORT PROPERTIES FOR THE CONTAIN CODE.

i COOLANT PIPING. Pipe Failure induced By Crack Crowth And Failure l Of Intermediate Suppwis. Caek Drop Accident NUREG/CR-5112: EVALUATION - TOluNG WATER REACTOR NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF WATER-LEVEL SENSINO 2NE BREAK AND SINGLE THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR FAILURE.Genene lasue 101 Boili. , Watu 1eactor Level Redundancy.

POWER PLANTS-Technical Findings.

NUREG/CR-5115: A REVIEW OF BOluNG WATER REACTOR WATER SY TEM LO :BOluNG WATER NUR G/CR-5224: THE LEACHABlWTY OF DECONTAMINATION TON-NUREG/ 5 24 f ERFA REACTORS. EXCHANGE RES NS SOUDIFIED IN CEld NT AT OPERATING NU.

NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK I CLEAR POWER PLANTS.

CONTAINMENT IMPROVEMENTS. WCg h Babcock And Wilcox Power Plant NUREG 0381 V01 R11: DIRECTORY OF CERTIFICATES OF COMPLl-i NUREG/CR-5206: A PROBABILISTIC EVALUATION OF THE SAFETY ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC l OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS Approved Packages.

WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL NUREG-0383 VO? R11: DIRECTORY OF CERTIFICATES OF COMPLl-EVENTS. ANCE FOR RADIOAC:.VE MATERIALS PACKAGESCertificates Of Compliance.

jL Babcock And WIlcox Reactor Plant NUREG-0303 V03 ROB: DIRECTOFd OF CERTIFICATES OF COMPU-NUREG/CR-5311: AN ASSESSMENT OF RELAPS/ MOD 2 APPLICABIL- ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC l

. ITY TO l.OSS OF-FEEDWATER TRANSIENT ANALYSIS IN A BAB- Approved Quahty Assurance Programs For Radioactive Matenal Pack-

! COCK AND WILCOX REACTOR PLANT. ages.

l 29 o _.o j

30' Subject trulex case out t: _  : ,:,,

NUREG/CR 5292 CLOSEOUT OF IE DULLETIN 80 23. FAILURES OF NUREG 1336: INTERIM GUID/qCE ON THE FTANDARD FORMAT AND SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEERING CONTENT OF FINANCIAL ASSURANCE MlCHANISMS REQUIRED CORPORATION- FOR DECOMMISSIONING UNDER 10 CFR PARTS 30.40,AND 70.

NUREG-1337: INTERIM GUIDANCE ON THE STANDARD REVIEW Component Fanure PLAN FOR THE REVIEW OF FINANCIAL ASSURANCE MECHANISMS NUREG/CR-5324: DETECTING COMPONENT FAILURE POTENTIAL FOR DECOMMIS$10NING UNDER 10 CFR PARTS 30,40,AND 70.

USING PROPORTIONAL HAZARD MODEL Decontamination 0888 NUREG/CR-5224: THE TEACHABILITY OF DECONTAMINATION ION-NUREG/CR4639 V5P2R1: NUCLEAR COMPUTERIZED LIBRARY FOR EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPERAllNG NU-ASSESS:NG REACTOR RELIABILITY (NUCLARR).Deta Manual.Part 2: CLEAR POWER PLANTS.

Human Error Probability (HEP) Eatimates.

NUREG/CR-4639 V5P3R1: NUCLEAR COMPUTERIZED LIBRARY FOR D6 gest ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 3: NUREG-0386 004 Rit: UNITED STATES NUCLEAR REGULATORY NU E /CR-463 P R1 U LE PUTERIZED LIBRARY FOR ch 1 8' t ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part 4: Summary Aggregations. Disposal Roorr-Containment Fabure elUREG/CR-5335: STABILITY OF DISPOSAL ROOMS DU91NG WASTE NUREG/CR-5030: AN ASSESSMENT OF STEAM-EXPLOSION-IN- RETRIEVAL DUCED CONTAINMENT FAILURE. Dose Calculation.

Containment Spray NUREG/CR-5M6: SADDE (SCALED ABSORBED DOSE DISTRIBUTION EVALUATOFQ.A Code To Generate input For VARSKIN.

NUREG-0800 06.52 R2: STANDARD REVIEW PLAN FOR THE PEVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER

' Spray AFs nup en EG/CROOO: INTEGRATING FIBER OPTIC RADIATION DOSIME-TER.

Contaminetton NUR G-1345: REVIEW OF EVENTS AT LARGE POOL TYPE IRRADIA- E S NUREG/CR-5266: EXAMINATION OF TWO 3M TYPE 902F STATIC LOCA ANALYSIS Final Report. I ELIMINATORS-Eannou .

),

Core Damage NUREG/CR-3006: DAMPlNG IN BUILDING STRUCTURES DURING 1 NUREG/CR-5042 SO2: EVALUATION OF EXTERNAL HAZARDS TO NU. EARTHQUAKES. Test Data And Modeling.

CLEAR POWER PLANTS IN THE UNITED STATES.Other Extemal NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF Events. .

THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR NUREG/CR-$197: EVALUATION OF GENERIC ISSUE 115. " ENHANCE. POWER PLANTS.

MENT OF THE REUABILITY OF WESTINGHOUSE SOLID STATE PROTECTION SYSTEM? Eddy Current NUREG/CR-5245: A REVIEW OF THE CRYSTAL RIVER UNIT 3 PROB- NUREG/CR-484R V01: STEAM GENERATOR GROUP PROJECT Task 9 ABIL10 TIC RISK ASSESSMENT.intemal Events. Core Damage Frequen. Final Report: Nondestructive Evaluation Round Robin. Volume 1: De.

cy. senption And Summary Data.

NUREG/CR-4849 V02: STEAM GENERATOR GROUP PROJECT. Task 9 Core Meltdown Fenal Report: Nondest uctwe Evaluation Round Robin. Volume 2: Raw

- NUREG/CR-5046: A FINITE ELEMENT ANALYSIS OF A REACTOR inspection Data.

PRESSURE VESSEL DURING A SEVERE ACCIDENT.

Emergency Core Cooling System Core-Melt Accident NUREG/CR-5074: DEVELOPMENT OF A PHENOMENA IDENTIFICA-NUREG/CR-5030- AN ASSESSMENT OF STEAM-EXPLOSION-IN- TION AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAULIC DUCED CONTAlWMENT FAILURE. PHENOMENA DURING A PWR LARGE-BREAK LOCA.

Corrosion Enforcement Act6on NUREG/CR-4667 V05: ENVIRONMENTALLY ASSISTED . CRACKING IN NUREG 0940 V07 N04: ENFORCE. MENT ACTIONS:SIGNIFICANT AC-LIGHT WATER REACTORS Semiannual Rept, April-September 1987. TIONS RESOLVED'Ouarterly Progress Report,0ctober December NVREG/CR-6115: A REVIEW OF BOILING WATER REACTOR WATER 1988' CHEMISTRV.Scence, Technology And Performance.

NUREG/CR-5234- VALUE/ IMPACT ANALYSIS FOR GENERIC ISSVE Equipment Guanficat6on 51: IMPROVING THE RELIABILITY OF OPEN-CvCLE aERVICE- NUREG/CR 5313: EQUIPMENT QUALIFICATION (EOFRISK SCOPING WATER SYSTEMS. STUDY.

Crack "

NUREG/CR-5143: APPLICATION OF THE J-INTEGRAL AND THE MODI- N REG.134*,: REVIEW OF EVENTS AT LARGE POOL TYPE IRRADIA-FIED J INTEGRAL TO CA9ES OF LARGE CRACK EFENSION TORS.

Creston Study Area AND External Hazard NUREG/CR 5251: WELLFIELD INSTALLMION INVESTIGATIONS CRESTON STUDY AREA. EASTERN WASHING- NUREG/CR-5042 S02: EVALUATION OF EXTERNAL HAZARDS TO NO-TON. CLEAR POWER PLANTS IN THE ONITED STATES.Other Extemat Events Demping NUREG/CR-3006: DAMPING IN BUILDING STRUCTURES DURING FORTRAN 77 EARTHQUAKES. Test Data And Modehng. NUREG/CR4478: UPDATE A FORTRAN 77 SOURCE FILE MANIPULATOR. Adapted For The Data General MV Senes Eclipse Decay Heat Removal Computert Under AOS/VS. _

NUREG 1338: DRAFT PREAPPLICATION SAFETY EVALUATION REPORT FOR THE MODULAR HIGH-TEMPERATURE GAS-COOLED Failure REACTOR. NUREG/CR-4780 V02; PROCEDURES FOR TREATING COMMON NUREG 1340: REGULATORY ANALYSIS FOR THE RESOLUTION OF CAUSE FAILURES IN SAFETY AND RELIABILITY STUD!ES. Analytical GENERIC ISSUE 99. " LOSS OF RHR CAPABILITY IN PWRS " Background And Techniques.

I L _ _ _ _ _ _ _ _ _ _ _ _ _ . _

Subject ind2x 31 Fellgue Fracture Toughnese NUREG/CR-5143: APPLICATION OF THE J lNTEGF,AL AND THE MODI-NUREG/CR-5195: FATIGUE STRENGTH OF ASME SA 106-B WELDED STEEL PIPES IN 288 DEGREES C AIR ENVIRONMENTS. FIED J-INTEGRAL TO CASES OF LARGE CRACK EXTENSION.

Falser Optic Fractured Rock NUREG/CR-5100: INTEGRATING FIBER OPTIC RADIATION DOSIME- NUREG/CR-5239: FLUID FLOW AND SOLUTE TRANSPORT MODEL-TER. ING THROUGH ThREE-DIMENSIONAL NETWORKS OF VARIABLY SATURATED DISCRETE FRACTURES.

Financial Aneurence NUREG/CR-5251' WELLFIELD INSTALLATION AND INVESTIGATIONS,CRESTON S'UDY AREA ~ EASTERN WASHING- 1 NUREG 1336: INTERIM GUIDANCE ON THE STANDARD FORMAT AND CONTENT OF FINANCIAL ASSURANCE MECHANISMS REQUIRED TON' FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,AND 70.

NUREG-1337: INTERIM GUIDANCE ON THE STANDARD REVIEW Mooled Rector PLAN FOR THE REMEW OF FINANCIAL ASSURANCE MECHANISMS NUREG-1338: DRAFT PREAPPLICATION SAFETY EVALUATION FOR DECOMMISSIONING UNDER 10 CFR PARTS 30.40,AND 70. REPORT FOR THE MODULAR HIGH TEMPERATURE GAS COOLED REACTOR.

par, -

A NUREG/C43037: USER'S MANUAL FOR FIRIN.A Computer Ccde To k eA Fire And Rt.$oactive Airbome Release in Nuclear N EGI 6 Rot: GENERIC COST ESTIMATES.Abstreets From Genenc Studies For Use in Prepanng Regulatory impact Analyses.

Fire Barrier Generic leeue 023 NUREG/CR-5088: FIRE RISK SCOPING STUDY:! INVESTIGATION OF NUREG/CR-4948 TECHNICAL FINDINGS RELATED TO GENERIC NUCLEAft POWER PLANT FIRE RISK,1NCLUDING PREVIOUSLY UN.

ISSUE 23 REACTOR COOLANT PUMP SEAL FAILURE.

ADDRESSEDiSSUES.

Fir, Risk Generic leeue 061 NUREG/CR-5234: VALUE,lMPACT ANALYSIS FOR GENc.RIC ISSUE NUREG/C45088: FIRE RISK SCOPING STUDY. INVESTIGATION OF 51; IMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE.

NUCLEAR POWER PLANT FIRE RISK. INCLUDING PREVIOUSLY UN-WATER SYSTEMS.

ADDRESSED ISSUES.

Fire Betety Generic leeue 094 NUREG/CR-5186: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE 94, NUllEG/CR-5088: FIRE RISK SCOPING STUDY: INVESTIGATION OF N'JCLEAR POWER PLANT FIRE RISK,1NCLUDING PREVIOUSLY UN. "ALOITIONAL LOW TEMPERATURE OVERPRESSURE PROTECTION ADDRESSEDISSUES. FOR LfGHT WATER REACTORS."

Fire Suppression System Generic loeue 099 NUREG/C45088: FIRE RISK SCOPING STUDY; INVESTIGATION OF NUREG-1340: REGULATORY ANALYSIS FOR THE RESOLUTION OF NUCLEAR POWER PLANT FIRE RISK.lNCLUDING PREVIOUSLY UN- GENERIC ISSUE 99. " LOSS OF RHR CAPABILITY IN PWRS."

ADDRESSED ISSUES.

Generic leeue 101 Flocal Year NUREGICR 5112: EVALUATION OF BOILING WATER REACTOR NUREG-1100 V05: BUDGET ESTIMATES. Fiscal Years 1990-1991- WATER LEVEL SENSING LIN2 BREAK AND SINGLE Flee 4on Product Cleanup hn NUREG-0800 06.5.2 R2: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Generic leeue 105 PLANTS.iWR Editon.Rewsion 2 to SRP Secton 6.5.2. " Containment NUREG/C45102: INTERFACING SYSTEMS LOCA; PRESSURIZED NU E$.0 0 R S kR I PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Generic laeus 115 PLANTS. LWR Edilon.Revson 3 To SRP Secten 6.5 4," Ice Condens- NUREG/CR-5197: EVALUATION OF GENERIC ISSUE 115. " ENHANCE.

er As A Fisson Product Cleanup Svstem." MENT OF THE RELIABILITY OF WESTINGHOUSE SOLID STATE NUREG-0000 06.5.5 RO: STANDARD FIEVIEW PLAN FOR THE REVIEW PROTECTION SYSTEM."

OF SAFETY ANALYSIS REFORTS FOR NUCLEAR POWER PLANTS. LWR Editen.Rewsion 0 To SRP Secton 6.5 5, " Pressure Sup- Generic Safety leaue presson As A Fission Product Cleanup System." NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR REG /CR-4496 V02: A SYSTEM FOR GENERATING LONG STREAM-FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN Gibbelte PERIOD. Phase ll. NUREG/CR-5271: THERMOCHEMICAL PROPERT!ES OF GIBBSITE.BAYERITE,BOEHMITE. OlASPORE AND THE ALUMINATE BETW 0 W 350 DENS C U /CR 5239: FLUID FLOW AND SOLUTE TRANSPORT MonEL-ING THROUGH THREE-DIMENSIONAL NETWORKS OF VAP8t Ly Ground Motion SATURATED DISCRETE FRACTURES. NUREG/CR4250 V01: SEISMIC HAZARD CHARACTER 12ATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY Fracture MOUNTAINSMethodology. input Data And Comparisons To Prevous NUREG/CR-5326: FRACTURE EVALUATION OF SURFACE CRACKS EMBEDDED IN REACTOR VESSEL CLADDING ~ NUR C 20 MIC HAZARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS Results Fracture Mechenica And Discussion For The Batch 1 Sites NUREG/CR 4792 V01: PROBABILITY OF FAILURE IN BWR REACTOR NUREG/CR-5250 V03: SEISMIC HAZARD CHARACTERl2ATION OF 69 COOLANT PIPING Summary Report.

NUCIEAR PLANT SITES EAST OF THE ROCKY MOUNT AINS Results NUREG/C44792 V02. PROBABILITY OF FAILURE IN BWR REACTOR And Discussion For The Batch 2 Sites.

COOLANT PIPING. Pipe Failure induced By Crack Growth And Fellure NUREG/CR-5250 V04: SEISMIC HAZARD CHARACTER 12ATION OF 69 Of intenmdiate Supports NUCLEAR PLANT SITES EAST OF THE ROCKY MOUN1 AINS Results NUREG/CR-5143: APPLIC' TION OF THE J-lNTEGRAL AND THE MODI.

And Discussion For The Batch 3 Sites.

FIED J INTEGRAL TO CASES OF LARGE CRACK EXTENSION NUREG/C45265: SIZE EFFECTS ON J.R CURVES FOR A 302-B NUREG/CR-5250 V05: SEISMIC HAZARD CHARACTERIZATION OF 69 PLATE.

NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNT AINS Results NUREG/CR 5320: IMPACT OF RADIATION EMBRITTLEMENT ON IN- And Discussion For The Batch 4 Sites TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR NUREG/CR4250 V06. SElSMIC HAZARD CHARACTERIZATION OF 69 NUCL EAR PLANT SITES EAST OF THE ROCMY PLANTS.

32 Subject ledex

' MOUNTAINS Regional Comparison Between Sites. Site Effects, General LER Discussion And Conclusions. NUREG/CR-2000 V07N12: LICENSEE EVELT REPORT (LER)

NUREG/CR-5250 408: SEISMIC HAZARD CHARACTER 12ATION OF 69 COMPILATION For Month Of December 1988.

NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINSSmplementary Seismic Hazard Results For Sites With LIDAR Mattiple Soil Condeons. NUREG/CR 4950 V03: THE SHORELINE ENVIRONMENT ATMOS.

PHERIC DISPERSION EXPERIMENT (SEADEX).Asrborne LIDAR Data.

NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD- LOCA ELING OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL SITE.A NUREG-1230: COMPENDIUM OF ECCS RESEARCH FOR REALISTIC Companson Of Predictive Radionuclides Transport Modeling Versus LOCA ANALYSIS. Final Report.

Field Observations At The "A" Disposal Area, Chalk River Nuclear Lab-NUREG/CR-5074 DEVELOPMENT OF A PHENOMENA (DENTIFICA-oratones.

TlON AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAULIC High-Level Waste PHENOMENA DURING A PWR LARGE BREAK LOCA.

NUREG/CR-5271: THERMOCHEMICAL NUREG/CR 5102: INTERFACING SYSTEMS LOCA. PRESSURIZED PROPERTIES OF WATER REACTORS.

GIBBSITE,BAYERITE.BOEHMITE, DIASPORE AND THE ALUMINATE NUREG/CR-5124: INTERFACING SYSTEMS LOCA BOILING WATER (

ION BETWEEN 0 AND 350 DEGPEES C- REACTORS.

Human Error Probability LWR NUREG/CR-4639 V5P4R1: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual,Part NUREG-0800 06.5.5 RO. STANDARD REVIEW PLAN FOR THE REVIEW

4. Summary Aggregations. OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition.Revison 0 To SRP Section 6.5 5 " Pressure Sup-pression As A Fesuun Product Cleanup System."

N REG / R 35 AR! SON AND APPL! CATION OF OUANTITA. NUREG-1230: COMPENDIUM OF ECCS RESEARCP FOR REALISTIC TIVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK ' F "P METHODS INTEGRATION AND EVALUATION PROGRAM R /R 69 V07 0NDE TRUCTIVE EXAMINATION (NDE) REll-(RMIEP). Final Report' ABILITY FOR INSERVICE INSPECTION Or LIGHT WATER REACTORS Semiannual Report April-September 1987.

IE Bulletin 80-23 NUREG/CR-46*8. EVALUATION OF RELIABILITY TECHNOLOGY AP-NUREG/CR-5292: CLOSEOUT OF IE BULLETIN 80-23. FAILURES OF PLtCABLE TO LWR OPERATIONAL SAFETY. I SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEER!NG NUREG/CR-5186. VALUE/ IMPACT ANALYSIS CV OENERIC ISSUE 94 CORPORATION' " ADDITIONAL LOW TEMPERATURE OVERPRESSURE PslOTECTION FOR LIGHT WATER REACTORS."

Ice Condenser NUREG-0800 06.5 4 R3: STANDARD REVIEW PLAN FOR THE REVIEW Leak Test OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR-5124: INTERFACING SYSTEMS LOCA BOILING WATER PLANTS LWR Edition Revision 3 To SRP Section 6.5 4 " Ice Condens- REACTORS.

er As A Fission Product Cleanup Syttem."

Legal issuances Individual Plant Examination NUREG 0750 V2B 101: INDEXES TO NUCLEAR REGULATORY COM-NUREG-1335 DRFT FC- INDIV! DUAL PLANT MISSION ISSUANCES July-September 1988.

EXAMINATION SUBMITTAL GUIDANCE. Draft Report For Comment NUREG-0750 V2B N05: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR NOVEMBER 1988. Pages 499-566.

InformatM Digest NUREG-0750 V2B N05 NUCLEAR REGULATORY COMMISSION IS-NUREG-1350 VD1: NUCLEAR REGULATORY COMMISSION 1989 IN- SUANCES FOA NOVEMBER 1988 Pages 499-566.

FORMATION DIGEST. NUREG 0750 V2B N06 NUCLEAR REGULATORY COMMISSION IS-NUREG 1350 V01: NUCLEAR REGULATORY COMMISSION 1989 IN- SUANCES FOR DECEMBER 1988. Pages 567-833.

FORMATION DIGEST.

Licensed Fuel Facility inservice inspection t4UREG 0430 V09 N01: LICENSED FUEL F ACILITY STATUS NUREG'CR-4469 V07. NONDESTRUCTIVE EXAMINATION (NDE) REll- REPORT. inventory Ditterence Data. January June 1988.(Gray Book H) nBILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Cam 4 annual Report.Apol September 1987. Licensed Operating Reactors instrument Line Break NUREG-0020 V12 N12: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT Data As Of November 30,1988 (Gray Boph f)

NUREGICR-5112: EVALUATION OF BOILING WATER REACTOR NUREG-0020 V13 N01: LICENSED OPERATING REACTORS STATUS WATER-LEVEL SENSING LINE BREAK AND SINGLE

SUMMARY

REPORT Data As Of December 31,1988.(Gray Book t)

FAILURE Genene issue 101 Boiling Water Reactor Level Redundancy-Technical Findmgs NUREG 0020 V13 NO2: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of January 31.1989.(Gray Book t)

Inventory Difference Data Ucensee Event Report NUREG-0430 V09 N01: LICENSED FUEL FACILITY STATUS NUREG-1275 V04: OPERATING EXPERIENCE FEEDDACK REPORT -

REPORT. Inventory Difference Data January June 19BB.(Gray Book ll)

TECHNICAL SPECIFICATIONS Commercial Power Reactors.

NUREC/CR 2000 V07N12. LICENSEE EVENT REPORT (LER)

NU EG/CR 5224. THE TEACHABILITY OF DECONTAMINATION ION- " "

EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPERATING NU' NURE 0O 8 bkN E EVENT REPORT (LER)

CLEAR POWER PLANTS. COMPILATION for Month Of January 1989 Ught Water Reactor NU EG/CR 522 BIODEGRADATION OF ION EXCHANGE MEDIA. NUREG.0800 06 5 5 R0 STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER trradiator PLANTS. LWR Edition. Revision 0 To SRP Section 6 5 5, " Pressure Sup-NUREG-1345 REVIEW OF EVENTS AT LARGE POOvTYPE IRRADIA. pression As A Fission Product C!oanup System "

TORS NUREG-1230 COMPENDIUM OF FCCS RESEARCH FOR REALISTIC LOCA ANALYSIS Final Report.

J-integral NUREG/CR-4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) RELI.

NUREG/CR 5143 APPLICATION OF THE J-INTEGRAL AND THE MODI. ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER FIED J-INTEGRAL TO CASES OF LARGE CRACK EXTENSION RE ACTORS Semiannual Report Apol-September 1987.

NUREG/CR 4618. EVALUATION OF RELIABILrry TECHNOLOGY AP.

J-R Curve PLICABLE TO LWR OPERATIONAL SAFETY.

NUREG/CR-5265 SIZE EFFECTS ON J.R CURVES FOR A 302-8 NUREG/CR-4667 V05. ENVIRONMENTALLY ASSISTED CRACKING IN PLATE LIGHT WATER REACTORS Sermannual Rept.Apnl September 1987.

l Subject Ind;x 33 NUREG/CR 5186: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE C4, Operating Expertence

" ADDIT!ONAL LOW TEMPERATURE OVERPRESSURE PROTECTION NUREG-1275 V04: OPERATING EXPERIENCE FEEDBACK REPORT -

FOR LIGHT WATER REACTORS." TECHNICAL SPECIFICATIONS Conimercial Power Reactors.

Loeo Of-Feedwater Operational Event NUREG/CR 5311: AN ASSESSMENT OF RELAPS/ MOD 2 APPLICABIL- NUREG/CR-5200: A PROBABILISTIC EVALUATION OF THE SAFETY ITY TO LOSSOF FEEDWATER TRANSIENT ANALYSIS IN A BAB- OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS COCK AND WILCOX REACTOR PLANT- WITH EMPHASIS ON HISTORICALLY OBSU4VED OPERATIONAL EVENTS.

Love Level Weste NUREG/CR-5325: A MULTIVARIATE STATISTICAL STUDY ON A DI.

NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD. VERSIFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER i ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL SITE.A PLANTS.'

I Compartson Of Predictive Radionuclides Transport Modehng Versus l Field Observations At The "A" Disposa! Area. Chalk River Nuclear Lab- Operational Safety f oratories. NUREG/CR-4618: EVALUATION OF RELIABILITY TECHNOLOGY AP- l

1. b^

i Low Temperature Overpressure i NUREG/CR-5186: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE 94 Operator Ucenalng Examiner i

" ADDITIONAL LOW TEMPERATURE OVERPRESSURE PROTECTION NUREG 1021 R05. OPERATOR LICENSING EXAMINER STANDARDS.

FOR LIGHT WATER REACTORS."

Overpressure Protection  !

Low-Level Waste Disposal Facility NUREG-0800 05.2.2 R2: STANDARD REVIEW PLAN FOR THE REVIEW NUREG 1293: OUALITY ASSURANCE GUIDANCE FOR LOW LEVEL OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER  !

RADIOACTIVE WASTE DISPOSAL FACILITY. Final Repoa PLANTS. LWR Edition.Revitiion 2 To SRP Section 5.2.2," Overpressure l Protection," And Revision 1 To Branch Technical Position RSB 5-2.

Mark i Containment NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK l P-Wave Res ual CONTAINMENT IMPROVEMENTS.

WESTERN OHIO-INDIANA REGION. Annual Report, October 1987 I Microsphere September 1988.

NUREG/CR-5266: EXAMINATION OF TWO 3M TYPE 902F STATIC ELIMINATORS. p Mitigation System NUREG/CR-4835: COMPARISON AND APPLICATION OF OUANTITA-NUREG/CR-5186: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE 94, TIVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK

" ADDITIONAL LOW TEMPERATURE OVERPRESSURE PROTECTION METHODS INTEGRATION AND EVALVATION PROGRAM FOR LIGHT WATER REACTORS." (RMIEP) Final Report.

NUREG/CR-5245: A REVIEW OF THE CRYSTAL RIVER UNIT 3 PROB-Multivariate Statistical Analysia ABILISTIC RISK ASSESSMENT. internal Events. Core Danage Frequer NUREG/CR 5325: A MULTIVARIATE STATISTICAL STUDY ON A Dl- cy VERSIFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER NUREG/CR 5313. EQUIPMENT QUALIFICATION (EO). RISK SCOPING PLANTS. STUDY.

NUCLARR PWR NUREG/CR 4639 V5P2R1: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5074. DEVELOPMENT OF A PHENOMENA IDENTIFICA-ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 2: TION AND RANKING TABLE (PIRT) FOR THERMAL HYDRAULIC Human Error Probability (HEP) Estimates- PHENOMENA DURING A PWR LARGE BREAK LOCA.

NUREGICR 4639 V5P3R1. NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5102: INTERFACING SYSTEMS LOCA. PRESSURIZED ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 3: WATER REACTORS.

Hardware Component Failure Data (HCFD). NUREG/CR-5116: SURVEY OF PWR WATER CHEMISTRY.

NUREG/CR 4639 V5P4R1: NUCLEAR COMPUTER! ZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part Performance Monitoring

4. Summary Aggregations NUREG/CR 4618 EVALUATION OF RELIABILITY TECHNOLOGY AP.

Nondestructive Evaluation NUREG/CR-4849 V01: STEAM GENERATOR GROUP PROJECT. Task 9 Petitions For Rulemaking Final Report: Nondestructive Evaluation Round Robin. Volume 1: De- NUREG-0936 V07 N04. NRC REGULATORY AGENDA Ouarterly NU E kR 849 02 TEA GENERATOR GROUP PROJECT. Task 9 Final Report Nondestructive Evaluation Round RotanVolume 2: Raw Piping Failure inspection Data- NUREG/CR-4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR Nondestructive Examination CWM NM % Fake inW W Qad GM W Faw O inwrmediate Supports.

NUREGIC44469 V07: NONDESTRUCTIVE EXAMINATION (NDE) REll-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Practice And Procedure Digest REAR' TORS. Semiannual Report April-September 1987. NUREG 0386 D04 R11: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July Nuclear Fuel Cycle 1972 March 1988, NUREG/CR-3037: USER'S MANUAL FOR FIRIN A Computer Code To Estimate Accidental Fire And Radioactive Airborne Release In Nuclear Pressure Suppression Pool Fuel Cycle Facilities NUREG-0800 06.5.5 RO. STANDARD REVIEW PL AN FOR THE REVIEW Nuclear Material Management OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS LWR Edition Revision 0 To SRP Section 6 5.5. " Pressure Sup-NUREG/CR 4004 STATISTICAL METHODS FOR NUCLEAR MATERIAL pression As A Fission Product Cleanup System."

MANAGEMENT.

Nuclear Safety PressJre Vessel NUREG-1355 THE STATUS OF RECOMMENDATIONS OF THE PRESI. NUREG/CR.5320: IMPACT OF RADIATION EMBRITTLEMENT ON IN-DENT'S COMMISSION ON THE ACCIDENT AT THREE MILE TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR ISLAND.A Ten Year Review PLANTS Nuclear Waste Pressure Vessel Embrittlement NUREG/C,45328 CORRELATIONS DETWEEN POWER AND TEST RE-NUREG/CR-5335. STABILITY OF DISPOSAL ROOMS DURING WASTE RETRIEVAL. ACTOR DATA BASES.

1

34 Subj:Ct Ind2x Pressurized Thermal Shock NUREG/CR-5326: FRACTLTIE EVALUATION OF SURFACE CRACKSNUREG-0383 V03 R08: DIRECTORY OF CERTIFICATES OF COMPLI.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC EMBEDDED IN REACTOR VESSEL CLADDING.

Approved Quality Assurance Programs For Radioactive Matenaf Pack-Pressurtred Water Reactor *#*'

NUREG 1340: REGULATORY ANALYSIS FOR THE RESOLUTION OF Radioactive Particle GENERIC ISSUE 99. " LOSS OF RHR CAPABILITY IN PWRS "

i NUREG/C45074: DEVELOPMENT OF A PHENOMENA IDENTIFICA. NUREG/CR 5276: SADDE (SCALED ABSORBED DOSE DISTRIBUTION TION AND RANKING TABLE (PIRT) FOR THERMAL HYDRAULIC EVALUATOR).A Code To Generate input For VARSKIN.

PHENOMENA DURING A PWR LARGE-BREAK LOCA.

Radionuclides Transport NUREG/CR 5102: INTERFACING SYSTEMS LOCA. PRESSURIZED NUREG/C NUREG/CR-4879 V02: DEMONSTRATION OF PERFORMANCE MOD-6 SU VEY OF PWR WATER CHEMISTRY. ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL $1TE,A Probabilistic Risk Assessment Companson Of Predictive Radionuclides Transport ModehnD Versus NUREG 1335 DRFT FC:

Field Observations At The "A" Disposal Area. Chalk River Nuclear Lab-INDIVIDUAL PLANT oratones.

EXAMINATION: SUBMITTAL GUIDANCE Draft Report For Comment NUREG/CR 4835: COMPARISON AND APPLICATION OF QUANTITA.

(

Reactor Accident TlVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK METHODS INTEGRATION AND EVALUATION PROGRAM NUREG/CR-5046. A FINITE ELEMENT .4NALYSIS OF A REACTOR (RMIEP) Final Report. PRESSURE VESSEL DURING A SEVERE ACCIDENT.

NUREG/CR-51/6: SEISMIC FAILURE AND CASK DROPANALYSES OF Reactor Coolant Piping THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLEAR POWER PLANTS. NUREG/C44702 V01: PROBABILITY OF FAILURE IN BWR REACTOR COOLANT PIPING Summary Report NUREG/CR-5245: A REVIEW OF THE CRYSTAL RIVE 1 UNIT 3 PROB-ABILISTIC RISK ASSESSMENT.internat Events, Core vamage Frequen- NUREG/CR 4792 V02: PROBABILITY OF FAILURE IN BWR REACTOR COOLANT PlPING. Pipe Failure Induced By Crack Growth And Failure NU$EG/CR-5313: EQUIPMENT QUALIFICATION (EO)-RISK SCOPING STUDY.

Reactor Coolant Pump A NUREG/CR-4948. TECHNICAL FINDINGS RELATED TO GENERIC R 5324 D CTING COMPONENT FAILURE POTENTIAL USING PROPORTIONAL HAZARD MODEL, Reactor Data Base NUREG/CR-5328: CORRELATIONS t3ETWEEN POWEH AND TEST RE-NUR G/CR-4948: TECHNICAL FINDINGS RELATED TO GENERIC ACTOR DATA BASES.

ISSUE 23 REACTOR COOLANT PUMP SEAL FAILURE.

Reactor Plant Design Quality Assurance NUREG 1293; QUALITY ASSURANCE GUIDANCE FOR LOW LEVEL NUREG/CR-5206: A PROBABILISTIC EVALUATION 01 THE SAFETY RADIOACTIVE WASTE DISPOSAL FACILITY.F:nal Report. OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS WITH EMPHAS'S ON HISTORICALLY OBSERVED OPERATIONAL Quality Assurance Program EVENTS.

NUREG-0383 W1 R11: DIRECTORY OF CERTIFICATES OF COMPU.Reactor Pressure Vessel ANCE FOR F ADIOACTIVE MATERIALS PACKAGES Report Of NRc Approved Pacugei NUREG/CR 5046 A FINITE ELEMENT ANALYSIS OF A REACTOR NUREG 03B3 W R11: DIRECTORY OF CERTIFICATES OF COMPU- PRESSURE VESSEL DURING A SEVERE ACCIDENT.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certshcates Of Reactor Safety Comphance.

NUREG/CR 5325; A MULTIVARIATE STATISTICAL STUDY ON A DI-NUREG 0383 V03 ROB: DIRECTORY OF CERTIFICATES OF COMPLl-ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC VERSIFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER PLANTS.

Approved Quality Assurance Programs For Radioactive Matenal Pack-ages.

Reactor System NUREG/CR-5069. TRAC-PF1/ MOD 1.Corre.atsons And Models.

NUREG/CR 5311: AN ASSESSMENT OF RELAPS/ MOD 2 APPLICABiL- Reector Trip System ITY TO LOSS OF-FEEDWATER TRANSIENT ANALYSIS IN A BAB-COCK AND WILCOX REACTOR PLANT. NUREG/CR-5197: EVAU TlON OF GEt RIC ISSUE 115. " ENHANCE-MENT OF THE RELIABlUTY OF WESTINGHOUSE SOLID STATE Ridlation Damage NUREG/CR 5221: BIODEGRADATION OF ION-EXCHANGE MEDIA. Reactor Vesset

• k$/CR 5 NUREG 0800 05.22 R2: STANDARD REVIEW PLAN FOR THE REVIEW INTEGRATING FIBER OPTIC RADIATION DOSIME- OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER TER' PLANTS LWR Edition. Revision 2 To SRP Section 5 2.2, " Overpressure R:dlation Embrittlement Protection." And Revision 1 To Branch Technical Position RSB 5 2 NUREG/CR-5320: IMPACT OF RADIATION EMBRITTLEMENT ON IN. Reactor Vessel Cladding TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR NUREG/CR-5326. FRACTURE EVALUATION OF SURFACE CRACKS PLANTS EMBEDDED IN REACTOR VESSEL CLADDING R dlation Monitoring Network Regulatory Agenda NUREG-0837 V0B NO3- NRC TLD DIRECT RADIATION MONITORING NUREG 0036 V07 N04. NRC REGULATORY AGENDA.Ouarterly NETWORK. Progress Report July-September 1968 Report. October December 1988 Ridloactive Airborne Release ReDulatory Impact Analyses NUREG/CR-3037: USER'S MANUAL FOR FIRIN A Computer Code To NUREG/CR 4627 RO1: GENERIC COST ESTIMATES Abstracts From Estimate Accidental Fire And Radioactive Airbome Release in Nuclear Fuel Cycle Facihties. Genent Studies For Use in Prepanng Regulatory impact Analyses Reliability Ridloactive Materiale Package NUREG-0383 VD1 R11. DIRECTORY OF CERTIFICATES OF COMPU. NUREG/CA-4780 V02: PROCEDURES FOR TREATING COMMON ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC CAUSE FAILURES IN SAFETY AND RELIABILITY STUDIES Analyt< cal A Background And Techneens NUkroved Packag$1:

G 0383 V02 DIRECTORY OF CERTIFICATES OF COMPU- Report To Congress ANCE FOR RADIOACTIVE MATERIALS PACKAGES Cemhcates Of Comphance NUREG 0090 V11 NO3' REPORT TO CONGRESS ON ABNORMAL OCCURRENCESJuly-September 1988

I SubDct ind:;x 35 i

Research Acthrity . Scaled Absorbed Doec Distribution NUREG/CR-5276: SADDE (SCALED ABSORBED DOSE DISTRIBUTION I NUREG-1319: A PRIORITIZATION OF RESEARCH ACTIVITIES.

EVALUATOR).A Code To Generate input For VARSKIN.

f NUREG-1340: REGULATORY ANALYSIS FOR THE RESOLUTION OF Sed 6 ment  !

GENERIC ISSUE 99," LOSS OF RHR CAPABILITY IN PWRS." NUREG/C45234: VALUE/ IMPACT ANALYSIS FOR GENERIC ISSUE  :

51: IMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE-Resin Weste WATER SYSTEMS-NUREG/CR-5221: BIODEGRADATION OF ION-EXCHANGE MEDIA.

NUREG/CR-5224: THE LEACHABluTY OF DECONTAMINATION ION- Solemic Capacity '

EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPERATING NU. NUREG/CR-5270: ASSESSMENT OF SEISMIC MARGIN CALCULATION CLEAR POWER PLANTS- METHODS.. .

Rules Setemic Hazard NUREG 0936 V07 N04: NRC - REGULATORY AGENDA.OuarteriY NUREG/CR-5250 V01: SEISMIC HAZARD CHARACTER 12ATION OF 69

- Report, October December 1988- NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS. Methodology, Input Data And Compansons To Previous Rube Of PMW ~

NUREG 0386 D04 R11: UNITED STATES NUCLEAR REGULATORY NUR CR S250 V MIC HAZARD CHARACTER 12ATION OF 69 '

COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST July NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results 1972. March 1988. And Discussion For The Batch 1 Sites.

NUREG/CR-5250 V03: SEISMIC HAZARD CHARACTERIZATION OF 69

. SA 1064 SW NUCLEAR PLAN 1 SITES EAST OF THE ROCKY MOUNTAINS.Results NUREG/CR-5195: FATIGUE STRENGTH OF ASME SA 106-B WELDED ^" "* " Fcr STEEL PIPLS IN 288 DEGREES C AIR ENVIRONMENTS. .NUR GkR 5 5O vo4ES HA ARD CHARACTERIZATION OF 69 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results SCANS And Discussion For The Batch 3 Sites.

NUREG/CR 4554 V01: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A NUREG/CR 5250 V05: SEISMIC HAZARD CHARACTERIZATION OF 69 MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results CASK DESIGN REVIEW. Volume 1: User's Manual To Version la (In- And Discussion For The Batch 4 SLs.

Pr Ref NUREG/CR-5250 V06: SEISMIC HAZARD CHARACTERIZATION OF 69 NUR CR V02 C S (SHIPPING CASK ANALYSIS SYSTEM).A NUCLEAR PLANT SITES EAST OF THE ROCKY MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING MOUNTAINS. Regional Companun Between Sites. Site Effects, General CASK DESIGN REVIEW. Volume 2: Theory Manual - Impact Analysis.

Discussson And Conclusions.  ;

NUREG/CR 4554 V03: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A NUREG/CR-5250 V07: SEISMIC HAZARD CHARACTERIZATION OF 69 1 MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING NUCLEAR PLANT SITES EAST OF THE ROCKY CASK DESIGN REVIEW. Volume 3: Theory blanual . Leari Slump In NUR G - SElb C HAZARD CHARACTERIZAllON OF 69 N G S S ShH A ALYF4S %1EM).A EAST OF THE ROCKY NUCLEAR PLANT SITES MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING MOUNTAINSSupplementary Seismic Ha:ard Results For Sites With CASK DESIGN REVIEW. Volume 4: Theory Manual . Thermal Analysis.

Multiple So# Condittons.

NUREG/CR 4554 V05: SCANS (SHIPPING CASK ANALYSIS SYSTEM);A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING SK DE IGN REVIEW. Volume 5: Theory Manual Thermal / Pressure UREG/C 270: ASSESSMENT OF SEISMIC MARGIN CALCULATION METHODS.

j Safety SeismicNy NUREG/CR-4780 V02: PROCEDURES FOR TREATING COMMON NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF THE <

CAUSE FAILURES IN SAFETY AND RELIABILITY STUDIES. Analytical WESTERN OHlOINDIANA REGION. Annual Report, October 1987.-

Background And Techruques. September 1968.

Safety Assurance Service-Water System ,

NUREG 1319: A PRIORITIZATION OF RESEARCH ACTIVITIES. 1 NUREG/CR-5234: VALUE/lMPACT ANALYSIS FOR GENERIC ISSUE

- Safety Evaluation Report 51.lMPROVING THE RELIABILITY OF OPEN CYCLE SERVICE-NUREG 0781 S06: SAFETY EVALUATION REPORT RELATED TO THE WATER SYSTEMS.

OPERATION OF SOUTH TEXAS PROJECT, UNIT 2. Docket No. 50-Severe Accident 4991 Houston Ughting And Power Company) DRFT FC: INDIVIDUAL PLANT NUREG 0781 S07: SAFETY EVALUATION REPORT RELATED TO THE NUREG-1335 OPERATION OF SOUTH TEXAS PROJECT, UNIT 2. Docket No. 50- EXAMINATION. SUBMITTAL GUIDANCE. Draft Report For Comment.

499 (Houston LighUng And Power Company) NUREG/CR 4835. COMPARISON AND APPLICATION OF OUANTITA-NUREG-1137 SOB. SAFETY EVALUATION REPORT RELATED TO THE TIVE HUMAN FIEllABILITY ANALYSIS METHODS FOR THE RISK OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 METHODS INTEGRATION AND EVALUATION PROGRAM AND 2. Docket Nos. 50-424 And 50-425.(Georgia Power Company.et af) (RMIEP) Final Report.

NUREG 1137 SO9: SAFETY EVALUATION REPORT RELATED TO THE NUREG/C45030: AN ASSESSMENT OF STEAM-EXPLOSION-IN-OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 DUCED CONTAINMENT FAILURE.

AND 2. Docket Nos. 50 424 And 50-425 (Georgia Power Company.et al) NUREG/CR 5042 S02: EVALUATION OF EXTERNAL HAZARDS TO NU-NUHEG 1232 V02 S01: SAFETY EVALUATION REPORT ON TENNES. CLEAR POWEFI PLANTS IN THE UNITED STATESOther External SEE VALLEY AUTHORITY: SEQUOYAH NUCLEAR PERFORMANCE Events.

PLAN Sequoyah Unit 1 Restart. NUREG/CR5046. A FINITE ELEMENT ANALYSIS OF A REACTOR NUREG 1338: DRAFT PREAPPLICATION SAFETY EVALUATION PRESSURE VESSEL DURING A SEVERE ACCIDENT.

REPORT FOR THE MODULAR HIGH-TEMPERATURE GAS-COOLED NUREG/CR 5173 REVIEW AND ASSESSMENT OF THERMODYNAMIC '

AND TRANSPORT PROPERTIES FOR THE CONTAIN CODE.

FIEACTOR.

Safety Research Program Shipping Cask ,

i NUREG/CR-2331 V07 N4: SAFETY RESEARCH PROGRAMS SPON. NUREG/CR 4554 V01 SCANS (SHIPPING CASK ANALYSIS SYSTEM).A J SORED BY OFFICE OF NUCLEAR REGULATORY MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING RESEARCH. Progress Report. October December 1987. CASK DESIGN REVIEW Volume 1: User's Manual To Version la (In.

NUREG/CR 2331 V08 N3. SAFETY RESEARCH PROGRAMS SPON- ciuding Program Reference)

SORED - BY OFFICE OF NUCLEAR REGULATOqy NUREG/CR4554 V02. SCANS (SHIPPING CASK ANALYSIS SYSTEM).A RESEARCH Progress Report July. September 1988. MICROCOMPUTER BASED ANALYSIS SYSTEM FOH SHIPPING CASK DESIGN REVIEW Volume 2 Theory Manual - Impact Analysis.

Salt NUREG/CR4554 V03: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A NUREG/C45243: DOREHOLE CLOSURE IN SALT. MICROCOMPUTER BMED ANALYSIS SYSTEM FOR SHIPPING

o

' 36 ' Subject index CASK DESIGN REVIEW. Volume 3: Theory Manual w Lead Slump in Steam Generator impact Analyssa And Verification Of impact Analysis. NUREG/CR-48A9 V0t STEAM GENERATOR GROUP PROJECT, Task 9 NUREG/CR4554 V04: SCANS (SHIPPING CASK ANALYSIS SYSTEMLA Final Report: Nondestructive Ev61uation Round Robin. Volume 1; De-l MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING scription And Summary Data.

CASK DESIGN REVIEW. Volume 4: Theory Manual Thermal Analysis. NUREG/CR-4849 V02: STEAM GENERATOR GROUP PROJECT. Task 9 l NUREG/CR-4554 V05: SCANS (SHIPPING CASK ANALYSIS SYSTEM).A Final Report Nondestructive Evaluation Round Robin. Volume 2: Raw

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING Inspection Data.

. CASK DESIGN REVIEW. Volume 5: Theory Manual . Thermal / Pressure Stress Analysis. . Steel Pipe '

NUREG/CR.5195: FATIGUE STRENGTH OF ASME SA 106 B WELDED Shore 46ne Environment STEEL PIPES IN 268 DEGREES C AIR ENVIRONMENTS.

NUREG/CR4950 V03: THE SHORELINE ENVIRONMENT ATMOS-PHERIC DISPERSION EXPERIMENT (SEADEX).Airbome LIDAR Data. Stochastic Hydrology NUREG/CR-4496 V02: A SYSTEM FOR GENERATING LONG STREAM.

Solenoid Valve -

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN NUREG/CR-5292: CLOSEOUT OF IE BULLETIN 80-23: FAILURES OF PERIOD. Phase 11, 4

SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEERING CORPORATION Stochastic Rainfall NUREG/CR-4496 V02: A SYSTEM FOR GENERATING LONG STREAM-Solute Transport . .

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN NUREG/CR-5239: FLUlO FLOW AND SOLUTE TRAfMORT MODEL- PERIOD. Phase il ING THROUGH THREE-DIMENSIONAL NETWORK 6 OF VARIABLY SATURATED DISCRETE FRACTURES. Stress Corrosion Cracking NUREG/CR.4667 V05: ENVIRONMENTALLY ASSISTED CRACKING IN Source File LIGHT WATER REACTORS. Semiannual Rept, April-September 1987.

NUREG/CR-4478: UPDATE - A FORTRAN 77 SOURCE FILE NUREG/CR.4702 V01: PROBABILITY OF FAILURE IN BWR REACTOR MANIPULATOR. Adapted For The Data General MV Sones Eclipse COOLANT PIPING. Summary Report.

Computers Under AOS/VS. NUREG/CR-5115: A REVIEW OF BOILING WATER REACTOR WATER CHEMISTRY. Science, Technology And Performance.

Source Leakage NUREG/CR-5116: SURVEY OF PWR WATER CHEMISTRY, NUREG 1345: REVIEW OF EVENTS AT LARGE POOL-TYPE IRRADIA-NUREG/CR-5326. FRACTURE EVALUATION OF SURFACE CRACKS >

Spent Fuel Poof EMBEDDED IN REACTOR VESSEL CLADDING.  ;

NUREG/CR-5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF T E SPE UEL POOLS AT TWO REPRESENTATIVE NUCLEAR E  !

l

' ' NUREG 0837 V08 NO3: NRC TLD DIRECT RADIATION MONITORING .!

NETWORK. Progress Report. July-September 1988.

Sta6niego Steet NUREG/CR-4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) REll.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NU G/CR 5069: TRAC-PF1/ MOD 1.Cortys And Models.  !

l REACTORS Semiannual Report, April. September 1987, Technical Specification NUREG 1275 V04: OPERATING EXPERIENCE FEEDBACK REPORT .

N G 13 NTERIM GUIDANCE ON THE STANDARD FORMAT AND

^ P ^

CONTENT OF FINANC!AL ASSURANCE MECHANISMS REOUIRED N E 33}TE CA IC 1 NS F U H TEX /.S FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,AND 70- PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-499.(Houston Lightino And Power Company)

NUREGd343: TECHNICAL SPECIFICATIONS FOR VOGTLE ELECTRIC N G GENERATING PLANT, UNITS 1 And 2. Docket Nos. 50424 And 50-5.2 2 R2: STANDARD REVIEW PLAN FOR THE REVIEW ,

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER g4M [33"4f YC N ECIF1 CATIONS FOR SOUTH TEXAS PLANTS LWR Edition. Revision 2 To SRP Section 5.2.2, " Overpressure Protection," And Revision 1 To Branch Technicat Position RSB 5 2. PROJECT, UNITS 1 AND 2. Docket Nos. 50 498 And 50499 (Houston Ligh NUREG-0800 06.5.2 Fl2: STANDARD REVIEW PLAN FOR THE REVIEW NUHEh359: TECHNICAL Sand Power Compan CIFICATIONS FOR VOGTLE ELECTRIC

- OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER GENERATING PLANT, UNITS 1 AND 2. Docket Nos. 50424 And 50 PLANTS. LWR Edition. Revision 2 to SRP Section 6.5.2, " Containment Sprav As A Fassion Product Cleanur System." 425.(Georgia Power Company,et al)

NUREG 0800 06.5.4 R3: ETANDARD REVIEW PLAN FOR THE REVIEW Tectonic OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF THE PLANTS 1WR Edition. Revision 3 To SRP Section 63 4, " Ice Condens-WESTERN OHIO-INDIANA REGION. Annual Report. October 1967 +

er As A Fission Product Cleanup System."

Septemtaer 1988.

NUREG-0800 06.5.5 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Thermal Aning PLANTS LWR Edition. Revision O To SRP Section 6 5.5, " Pressure Sup- NUREGiCR-5328. CORRELATIONS BETWEEN POWER AND TEST RE-pression As A Fission Product Cleanup System." ACTOR DATA BASES.

NUREG-1337: INTERIM GUIDANCE ON THE STANDARD REVEN PLAN FOR THE RE' VIEW OF FINANCIAL ASSURANCE MECHANISMS Thermal Hydraulle FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,AND 70. NUREG/CR 5069: TRAC PF1/ MOD 1. Correlations And Models Static Eliminator Thermal-Hydraulic Phenomena NtlREG/CR 5266. EXAMINATION OF TWO 3M TYPE 90PF STATIC NUREG/CR-5074. DEVELOPMENT OF A PHENOMENA IDENTIFICA-ELIMINATORS. TION AND RANKING TABLE (PIRT) FOR THERMAL HYDRAULIC PHENOMENA DURING A PWR LARGE-BREAK LOCA.

Station Blackout NUREG/CR-4948: TECHNICAL FINDINGS RELATED TO GENERIC Thermodynamic l ISSUE 23: REACTOR COOLANT PUMP SEAL FAILURE. NUR2G/CR5173 REVIEW AND ASSESSMENT OF THERMODYNAMIC  !

AND TRANSPORT PROPERTIES FOR THE CONT AtN CODE Statistical Method NUREGICR-5271: THERMOCHEMICAL PROPERTIES OF 14UREG/CR4604: STATISTICAL METHODS FOR NUCLEAF, MATERIAL GIBBSITE.BAYERITE.BOEHMITE, DIASPORE AND THE ALUMINATE MANAGEMENT. ION BETWEEN O AND 350 DEGREES C.

i Steam Explosion Thermoluminescent NUREG/CR-5030: AN ASSESSMENT OF STEAM EXPLOSION-IN. NUREG-0837 V00 NO3 NRC TLD DIRECT RADIATION MONITORING DUCED CONTAINMENT FAILURE _ NETWORK Progress Report July Septernber 1988

Subjtet Ind:;x 37 Three Mlle Island unsaturated Zone NUREG-1355: THE STATUS OF RECOMMENDATIONS OF THE PRESI- NUREG/CR-5239: FLUID FLOW AND SOLUTE TRANSPORT MODEL-DENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ING THROUGH THREE-DIMENSIONAL NETWORKS OF VARIABLY ISLAND.A Ten Year Rewew. SATURATED DISCRETE FRACTURES.

Title List VARSKIN NUREG-0540 V10 NIO: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILADLE. October 1-31.1988 NUREG/CR-5276. SADDE (SCALED ABSORBED DOSE DISTRIBUTION NUREG 0540 V10 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY EVALUATOR).A Code To Generate input For VARSKIN.

AVAILABLE. November 1 30 1988.

NUREG 0540 V10 N12: TITLd LIST OF DOCUMENTS MADE PUBUCLY Waste Retrieval AVAILABLE. December 1-31,1988. NUREG/CR-5335: STABILITY OF DISPOSAL ROOMS DURING WASTE RETRIEVAL Transient Reactor Analysis Code NUREG/CR-5069: TRAC-PF1/ MOD 1.Correlahons And Models.

Transportation NUREG/CR-4667 V05: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG 0383 V01 R11: DIRECTORY OF CERTIFICATES OF COMPLI- UGHT WATER REACTORS. Semiannual Rept, April September 1987.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC NUREG/CR-5115: A REVIEW OF BOILING WATER REACTOR WATER CHEMIST RY.Scence, Technology.And Performance.

NU G 8 Yt1: DIRECTORY OF CERTIFICATES OF COMPU- NUREG/CH-5116: SURVEY OF PWR WATER CHEMISTRY.

ANCE FOR RADIOACTIVE MATERIALS FACKAGES. Certificates Of Westinghouse Solid State Protection System NU $ 3NV03 R08: DIRECTORY OF CERTIFICATES OF COMPLl-ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC NUREG/CR-5197: EVALUATION OF GENERIC ISSUE 115. " ENHANCE-Approved Quality Assurance Programs For Radioactive Material Pack- MENT OF THE RELfABluTV OF WESTINGHOUSE SOUD STATE ages. PROTECTION SYSTEM "

r,-,,,-- - - -- - - - - , ,

3 ,

I I

i l

. NRC Originating Organization Index (Staff Reports)  ;

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

I OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) EDO - OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS OFC OF THE EXECUTIVE DIRECTOR FOR OPERATIONS OFFICE OF NUCLEAR MATERIAL SAFETY &  :

l NUREG 1355: THE STATUS OF RECOMMENDATIONS OF THE SAFEGUARDS. DIRECTOR PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE NUREG 0430 V09 N01: LICENSED FUEL FACILITY STATUS ISLAND.A Ten-Year Review. REPORT. inventory Difference Data. January-June 1988.(Gray Book II)

REGION 1. OFC OF THE DIRECTOR DIVISION OF SAFEGUARDS & TRANSPORTATION (POST 870413)

NUREG 0837 V08 NO3: NRC TLD DIRECT RADIATION MONITORING NUREG 0383 V01 R11: DIRECTORY OF CERTIFICATES OF COMPL1- i NETWORK. Progress Report. July-September 1988. ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC 1 OFC OF ENFORCEMENT (POST 870413) Approved Packages. ,

NUREG-0940 V07 N04: ENFORCEMENT ACTIONS:SIGNIFICANT AC- NUREG 0383 V02 R11: DIRECTORY OF CERTIFICATES OF COMPLl- 1 TIONS RESOLVED.Ouarterly Progress Report. October-December ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of Compliance.

1988' l' NUREG 0303 V03 ROB: D: RECTORY OF CERTIFICATES OF COMPLI.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC f EDO- OFFICE OF ADMINISTRADON (PRE 870413 & POST 890205)

Approved Quality Assurance Programs For Radioactive Matenal l DIVISION OF FREEDOM OF INFORMATION & PUBLICATIONS SERV.

Packages.

ICES (POST 890205 NUREG-0750 V28 N06: NUCLEAR REGULATORY COMMISSION IS. DIVISION OF LOW LEVEL WASTE MANAGEMENT & DECOMMISSION.

N A YAS NCE GUID CF FOR OW LEVEL NUREG 50 V29 NO NUCLE RE ULA COMMISSION IS-SUANCES FOR JANUARY 1989. Pages 187.

NUREG 1336: INTERIM GUIDANCE ON THE STANDARD FORMAT AND CONTENT OF FINANCIAL ASSURANCE MECHANISMS RE-EDO - OFFICE OF THE CONTROLLER (PRE 820418 & POST 890205)

QUIRED FOR DECOMMISSIONING UNDER 10 CFR PARTS DIVISION OF BUDGET & ANALYSIS (POST 890205)

NUREG 13 V01 CLEAR REGULATORY COMMISSION 1989 IN-N E 37. NTERIM GUIDANCE ON THE STANDARD REVIEW PLAN FOR THE REVIEW OF FINANCIAL ASSURANCE MECHA.

EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NISMS FOR DECOMMISSIONING UPDER 10 CFR PARTS DATA 30,40.AND 70.

OFFICE FOH ANALYSIS & FVALUATION OF OPERATIONAL DATA, DI-U.S. NUCLEAR REGULATORY COMMISSION NUREG-0090 V11 NO3: REPORT TO CONGRESS ON ABNORMAL N RE 386 R11 U TED ATES NUCLEAR REGULATORY DIVIS OF SA R AM ( OST 870413) fg7 h, ggg' NUREG-1275 V04: OPERATING EXPERIENCE FEEDBACK REPORT -

TECHNICAL SPECIFICATIONS Commercial Power Reactors. EDO - OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

NUREG 1345; REVIEW OF EVENTS AT LARGE POOL-TYPE IRRA- OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR (POST DIATORS. 860720)

NUREG 1335 DRFT FC: INDIVIDUAL PLANT EDO. OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM EXAMINATION.SUDMITTAL GUIDANCE. Draft Report For Comment.

(870413-890204) DIVISION OF REACTOR & PLANT SYSTEMS (870413 880716)

DIVISION OF COMPUTER & TELECOMMUNICATIONS SERVICES NUREG/CR-4835: COMPARISON AND APPLICATION OF OUANTITA-(POST 890205) TIVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK NUREG 0020 V13 NO2. LICENSED OPERATING REACTORS STATUS METHODS INTEGRATION AND EVALUATION PROGRAM

SUMMARY

REPORT. Data As Of January 31.1989.(Gray Book i) (RMIEP). Final Report.

POLICY & PROGRAM MANAGEMENT STAFF (870413 890204) DIVISION OF REGULATORY APPLICATIONS (POST 870413)

NUREG-0020 V12 N12: LICENSED OPERATING REACTORS STATUS NUREG-1319: A PRIORITIZATION OF RESEARCH ACTIVITIES.

SUMMARY

REPORT. Data As Of November 30.1988.(Gray Book 1) NUREG-1338. DRAFT PREAPPLICATION SAFETY EVALUATION NUREG 0020 V13 N01: LICENSED OPERATING REACTORS STATUS REPORT FOR THE MODULAR HIGH-TEMPERATURE GAE-

SUMMARY

REPORT. Data As Of December 31.1988.(Gray Book 1) COOLED REACTOR.

OlVISION OF FREEDOM OF INFORMATION & PUBLICATIONS SERV. OlvlSION OF SAFETY ISSUE RESOLUTION (POST 880717)

NUREG-1340: REGULATORY ANALYSIS FOR THE RESOLUTION OF ICES (880515 8902 NUREG.0540 V10 N10: TITLE LIST OF DOCUMENTS MADE PUBLIO. GENERIC ISSUE 99. " LOSS OF RHR CAPABILITY IN PWRS "

D DIVISION OF SYSTEMS RESEARCH (POST 880717)

NUREG0 V 0 N11 ITLE ST OF DOCUMENTS MADE PUBLIC-OCA NA YS F e LY AVAILABLE. November 1 30.1988.

NUREG.0540 V10 N12: TITLE LIST OF DOCUMENTS MADE PUBLIC- gog , OFFICE OF NUCLEAR REAC10R REGULATION (POST 4/28/80)

LY AVAILABLE. December 1-31,1988 OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST j NUREG-0750 V28101: INDEXES TO NUCLEAR RE7UL ATORY COM- 870411) l MISSION ISSUANCES July. September 1988. NUREG.0000 05.2.2 R2: STANDARD REVIEW PLAN FOR THE NUREG 0750 V2B N05: NUCLEAR REGULATORY COMMtSSION IS- REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR SUANCES FOR flOVEMBER 1988 Pages 499-566. POWER PLANTS LWR Edition Revision 2 To SRP Section 5.2.2, NUREG 0936 V07 N04. NRC REGULATORY AGENDA.Ouarterly " Overpressure Protection," And Revision 1 To Branch Technical Po-Report Octotver December 1988 sition RSB 42. I DIVISION OF BUDGET & ANALYSIS (870413 890204) NUREG 0800 06.5.2 R2. STANDARD REVIEW PLAN FOR THE I NUREG.1100 V05. BUDGEl ESTIMATES Fiscal Years 1990-1991. REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR 39 I

1

40 NRC Origin; ting Org niz:ti:n Ind:;x (St;ff R;ptrts)

POWER PLANTS. LWR Edition. Revision 2 to SRP Sectron 6.5.2, NUREG 1359: TECHNICAL SPECIFICATIONS FOR VOGTLE ELEC-

" Containment Spray As A Fission Product Cleanup System." TRIC GENERATING PLANT, UNITS 1 AND 2. Docket Nos. 50-424 NUREG 0800 06.5 4 R3: STANDARD REVIEW PLAN FOR THE And 50-425 (Guorgia Power Company,et al)

REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR DIVISION OF REACTUR PROJECTS - Ill.IV,V & SPECIAL PROJECTS POWER PLANTS. LWR Edition.Rowsion 3 To SRP Secton 6.5.4. " Ice (POST 870411 Oc.idenser As A Fisson Product Cleanup System." NUREG 0781 S06: SAFETY EVALUATION REPORT RELATED TO NUREG 0800 06 5.5 RO: STANDARD REVIEW PLAN FOR THE THE OPERATION OF SOUTH TEXAS PROJECT, UNIT 2. Docket No.

REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR 50-499.(Houston Lighting And Power Company)

POWER PLANTS. LWR Editon.Rewsion 0 To SRP Section 6.5.5 NUREG-0781 S07: SAFETY EVALUATION REPORT RELATED TO

" Pressure Suppresson As A Fesson Product Cleanup Systern." THE OPERATION OF SOUTH TEXAS PROJECT, UNIT 2. Docket No.

NUREG-1335 DRFT FC: INDIVIDUAL PLANT EXAMINATION. SUBMITTAL GUIDANCE. Draft Report For Comment. NU E 1 'l TEC lhL eP C F C 10 S FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 Ar d 50 499.(Houston DIVISION OF REACTOR PROJECTS - t/ll (POST 870411)

NUREG 1137 S08. SAFETY EVALUATION REPORT RELATED TO Lgl hti And Power Com THE OPERATION OF VOGTLE ELECTRIC GENERATING A any)

S ECEAMS M SM EW PLANT. UNITS 1 AND 2. Docket Nos. 50 424 And 50-425.(Georgia . S1A 2ht M 54498 AM 5449&(Houston Power Company.et al)

ASS A D E R S CIAL PROJECTS (POST 890101)

NUREG-1137 SO9. SAFETY EVALUATION REPORT RELATED TO NUREG-1232 V02 S01: SAFETY EVALUATION REPORT ON TEN-THE OPERATION OF VOGTLE ELECTRIC GENERATING NESSEE VALLEY AUTHORITY: SEQUOYAH NUCLEAR PERFORM-PLANT. UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Georgia ANCE PLAN.Sequoyah Unit 1 Restart Power Company.et al) DIVISION OF LICENSEE PERFORMANCE & OUALITY EVALUATION NUREG-1343: TECHNICAL SPECIFICATIONS FOR VOGTLE ELEC- (PDST B70411)

TRIC GENERATING PLANT, UNITS 1 And 2. Docket Nos. 50 424 And NUREG-1021 ROS: OPERATOR LICENSING EXAMINER STAND.

50 425.(Georgia Power Company) ARDS.

,6 l

l l

l

m. _ _ - _

NRC Originating Organization index-(International Agreements)

This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each Gntry 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.

k Yhere were no NUREG/lA reports for this quarter.

l 1

l l

1 41

l 1

I Y

1

...m.__,_mm. _ _ _ _______ .-- --

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organization'ns that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (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) arepared by that organi-

.zation. If further information is needed, refer to the main citation by tie NUREG/CR number.

L .

EDO . OFFICE FOR ANALYSIS & EVALUATION OF OPEf1ATIOML NUREG/CR-4639 V5P3R1: NUCLEAR COMPUTERIZED LIBRARY DATA FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data OFFICE FOR ANALYSTS & EVALUATION OF OPERATIONAL DATA DI- Manual.Part 3: Hardware Component FM Data {HCFD).

RECTOR NUREG/CR-4839 V5P4R1: NUCLEAR u;.JPUTERIZED LIBRARY NUREG/CR-2000 V07N12- LICENSEE EVENT REPORT (LER) FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data COMPILATION For Month Of December 1988. Manual. Pan 4 Summary Aggregations.

NUREG/CR-2000 V08 N1: LICENSEE EVENT ' REPORT (LFR) NUREG/CR-5116: SURVEY OF PWR WATER CHEMISTRY.

COMPILATION:For Month Of January 1989 NUREG/CR4278: SADDE (SCALED ABSORBED DOSE DISTRIBU-DIVISION OF SAFETY PROGRAMS (POST 870413) TION EVALUATOR) A Code Tn Generate input For VARSKIN.

NUREG/CR 5324: DETECTING wMPONENT FAILURE POTENTIAL DIVISION OF ENGINEERING (POST 870413)

USING PROPORTIONAL HAZARD MODEL NUREG/CR-3008; DAMPING IN BUILDING STRUCTURES DURING NUREG/CR-5325: A MULTIVARIATE STATISTICAL STUOY ON A Dl- EARTHQUAKES. Test Data And Mode:ing.

VERSIFIED DATA GATHERING SYSTEM FOR NUCLEAR POWER NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF THE PLANTS- WESTERN OHlOINDIANA REGION. Annual Report, October 1987 -

EDO OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM NU 9 5/07: NONDESTRUCTIVE EXAMINATION (NDE) RE.

LIABILITY FOR INSERVICE INSPECTION OF LIGHT WATE9 OlV Fl F RMATION SUPPORT SERVICES (870413 890204) REACTORS. Semiannual Report. April-September 1987, NUREG/CR-4478: UPDATE .A FORTRAN 77 SOURCE FILE MANIPULATOR. Adapted For The Data General MV Senes Eclipse NUREG/CR4496 V02: A SYSTEM FOR GENERATING LONG Computers Under AOS/VS. STREAMFLOW RECORDS FOR STUC'Y OF FLOODS OF LONG RETURN PERIOD. Phase IL EDO-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CR 4554 V04: SCANS (SHIPPING CASK ANALYSIS OFFICE OF NUCLEAR MATERIAL SAFETY & SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR t SAFEGUARDS. DIRECTOR SHIPPING CASK DESIGN REVIEW. Volume 4: Theory Manual Ther-NUREG/CR-5271: THERMOCHEMICAL PROPERTIES OF mal Anafysia j GIBBSITE.BAYERITE, BOEHMITE, DIASPORE AND THE ALUMI- NUREG/CR 4554 V05: SCANS (SHIPPING CASK ANALYSIS NATE lON BETWEEN O AND 350 DEGREES C. SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR DIVISION OF SAFEGUARDS & TRANSPORT ATION (POST 870413) SHIPPING CASK DESIGN REVIEW. Volume 5. Theory Manual Ther-NUREG/CR-4554 VOI: SCANS (SHIPPING CASK ANALYSIS mal / Pressure Stress Analysis.  :

SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR NUREG/CR 4067 V05: ENVIRONMENTALLY ASSISTED CRACKING SHIPPING CASK DESIGN REVIEW. Volume 1: User's Manual To Ver- IN LIGHT WATER REACTORS Semiannual Rept,Apni-September soon 1a (including Program Reference) 1987. o NUREG/CR4554 V02: SCANS (SHIPPING CASK ANALYSIS NUREG/CR-4792 V01: PROBABILITY OF FAILURi IN BWR REAC-  !

SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM FOd TOR COOLANT PAPING. Summary Report SHIPPING CASK DESIGN REVIEW. Volume R Theory Manual - NUREG/CR 4702 V02: PRODAB;LITY OF FAILURE IN BWR REAC-Impact Antlysis. TOR COOLANT PIPING Pipe Failure induced By Crack Growth And i NUREG/CR4554 V03: SCANS (SHIPPING CASK ANALYSIS Failure Of intermediate Supports. l SYSTEM).A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR NUREG/CR 4849 V01: STEAM GENERATOR GROUP PROJECT. Task SHIPPING CASK DESIGN REVIEW Volume 3. Theory Manual Lead 9 Final Report Nondestructive Evaluation Round RobinVolume 1:

Slump in impact Analysis And Venfication Of impact Analysis. Desenption And Summary Data.

DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST NUREG/CR 4849 V02: STEAM GENERATOR GROUP PROJECT. Task NU G/ R 3037: USER'S MANUAL FOR FIRIN.A Computer Code To Ra i sp Da a Estimate Accidental Fire And Radioactive Airborne Release in Nucle- NUREG/CR4879 V01 DEMONSTRATION OF PERFORMANCE MOD-er Fuel Cycle Facihties- EL NG OF A LOW LEVEL WASTE SHALLOW-LAND BURIAL $1TE.A ,

I DIVtSION OF FUEL CYCLE, MEEtCAL, ACADEMIC & COMMERCIAL Companson Of Predictve Radionuclides Transport Modeling Versus NUREG CR 5266 XAMINATION OF TWO 3M TYPE 902F STATIC DIVI I N LEVEL WASTE MANAGEMENT (POST 670413) PHERIC DISPERSION EXPERIMENT (SEADEX) Airborne LIDAR NUREG/CR-5335: STABILITY OF DISPOSAL ROOMS DURING WASTE RETRIEVAL. NUF EG/CR-5115: A REVIEW OF BOILING WATER REACTOR WATER CHEMISTRY. Science. Technology.And Performance. j EDO. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

OFFICE OF NUCLEAR REGULATORY RESEARCH. DIRECTOR (POST NUREG/CR-5143. APPLICATION OF THE J lNTEGRAL AND THE 'l 860720) MODIFIED J-INTEGRAL TO CASES OF LARGE CRACK EXTEN- .

1 NUREG/CR4331 V07 N4: SAFETY RESEARCH PROGRAMS SPON. SION.

SORED BY OFFICE OF NUCLEAR REGULATORY NURr.G/CR-5105: FATIGUE STAFNGTH OF ASME SA 106 B RESEARCH. Progress Report October December 1987. WELDED STEEL PIPES IN 288 DEGREES C AIR ENVIRONMENTS.

NUREG/CR 2331 V06 N3. SAFETY RESEARCH PROGRAMS SPON. NUREG/CR-5221: BIODEGRADATION OF ION-EXCHANGE MEDIA. 4 SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-5224 THE TEACHABILITY OF DECONTAMINATION lON-RESE ARCH Progress Report JulvAeptember 1988 EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPERATING NU.

NUREG/CR-4604- STATISTICAL METHODS FOR NUCLEAR MATERI- CLEAR POWER PLANTS AL MANAGEMENT. NUREG/CR-5239 FLUID FLOW AND SOLUTE TRANSPORT MODEL-NUREG/CR4639 V5P2R1. NUCLEAR COMPUTERIZED LIBRARY ING THROUGH THREE-DIMEGIONAL NETWORKS OF VARIADLY FOR ASSESSING REACTOR RELIABILITY (NUCLARR) Deta SATURATED DISCRETE FRACTURES.

Manual Part 2: Human Error Probability (HEP) Estimates NUREG/CR-5243 BOREHOLE CLOSURE IN SALT.

43

44 NRC Contratet Sponsor ind2x NUREG/CR-5251: WELLFIELD INSTALLATION AND NUREG/CR-4780 V02: PROCEDURES FOR TREATING COMMON INVESTIGATIONS,CRESTON STUDY AREA, EASTERN WASHING- CAUSE FAILURES IN SAFETY AND RELIABILITY TON. STUDIES. Analytical Background And Technsques.

NUREG/CR-5265: SIZE EFFECTS ON J-R CURVES FOR A 302-8 NUREG/CR-5030: AN ASSESSMENT OF STEAM EXPLO9 TON-IN-PLATE. DUCED CONTAINMENT FAILURE NUREG/CR 5270: ASSESSMENT OF SEISMIC MAFsGIN CALCULA. NUREG/CR-5046: A FINITE ELEMENT ANALYSIS OF A REACTOR TION METHODS. PRESSURE VESSEL DURING A SEVERE ACCIDENT.

NUREG/CR-5320: IMP / CT OF RADIAT.ON EMBRITTLEMENT ON IN- NUREG/CR 5069: TRAC PF1/ MOD 1.Conelations And Models.

TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR NUREG/CR-5074: DEVELOPMENT OF A PHENOMENA (DENTIFICA-PLANTS TION AND RANKING TABLE (PIRT) FCA THERMAL HYDRAULIC NUREG/CR 5326: FRACTURE EVALUATION OF SUF. FACE CRACKS PHENOMENA DURING A PWR LARGE-BREAK LOCA.

EMBEDDED IN REACTOR VESSEL CLADDING NUREG/CR-5173: REVIEW AND ASSESSMENT OF THERMODY.

NUREG/CR-5328: CORRELATIONS BETWEEN POWER AND TEST NAMIC AND TRANSPORT PROPERTIES FOR THE CONTAIN DIVI ON O R OR & LANT SYSTEMS (870413-880716) N REG CA-E245: A REVIEW OF THE CRYSTAL RIVER UNIT 3 NUREG/CR4835: COMPARISON AND APPLICATION OF QUANTITA, PROBABILISTIC RISK ASSESSMENT. internal Events, Core Damage TIVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE RISK METHODS INTEGRATION AND EVALUATION PROGRAM NU G CYS311: AN ASSESSMENT OF RELAP5/ MOD 2 APPUCA-BILITY TO LOSS-OF FEEDWATER TRANSIENT ANALYSIS IN A DIV E ULA ORY APPLICATIONS (POST 870413) BABCOCK AND WILCOX REACTOR PLANT.

NUREG/CR 4627 R01: GENERIC COST ESTIMATES. Abstracts From Genenc Studies For Use in P epanng Regulatory impact Analyses- EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

PROGRAM MANAGEMENT, POLICY DEVELOPMENT & ANALYSIS NUREG/CR 5088: FIRE RISK SCOPING STUDY: INVESTIGATION OF NUCLEAR POWER PLANT FIRE RISK,1NCLUDING PREVIOUSLY STAFF (POST B70411)

NUREG/CR-5206: A PROBABlUSTIC EVALUATION OF THE SAFETY UNADDRESSEDISSUES. OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS NUREG/CR-5100: INTEGRATING FIBER OPTIC RADIATION DOSIME-TER WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL EVENTS.

NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK i DIVISION OF OPERATIONAL EVENTS ASSESSMENT (POST 870411)

CONTAINMENT IMPROVEMENTS. NUREG/CR-5292: CLOSEOUT OF IE BULLETIN 80-23. FAILURES OF NUREG/CR-5313: EQUIPMENT QUALIFICATION (EO)-RISK SCOPING SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEERING STUDY- CORPORATION.

DIVISION OF SAFETY ISSUE RESOLUTION (POST 880717) DIVISION OF ENGINEERING & SYSTEMS TECHNOLOGY (POGT NUREG/CR-4948. TECHNICAL FINDINGS RELATED TO GENERIC 870411)

ISSUE 23: REACTOR COOLANT PUMP SEAL FAILURE. NUREG/CR-5250 V01: SEISMIC HAZARD CHARACTERl2ATION OF NUREG/CR-5042 S02: EVALUATION OF EXTERNAL HAZARDS TO 69 NUCLEAR PLANT SITES EAST OF THE ROCKY NUCLEAR POWER PLANTS IN THE UNITED STATES.Other Exter- MOUNTAINS. Methodology, input Data And Compansons To Previous nal Events. Results For Ten Test Sites.

NUREG/CR 5102: INTERFACING SYSTEMS LOCA. PRESSURIZED NUREG/CR-5250 V02: SEISMIC HAZARD CHARACTERIZATION OF WATER REACTORS. 69 NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR-5112: EVALUATION OF BOILING WATER REACTOR MOUNTAINS.Results And Discussson For The Batch 1 bites WATER-LEVEL SENSING LINE BREAK AND SINGLE NUREG/CR-5250 V03: SEISMIC HAZARD CHARACTERIZATION OF FAILURE. Generic lasue 101 Boiling Water Reactor Level Redunden- 69 NUCLEAR PLANT SITES EAST OF THE ROCKY cy-Technical Findings. MOUNTAfNS.Results And Discussion For The Batch 2 Sitet.

NUFIEG/CR-5124: INTERFACING SYSTEMS LOCA. BOILING WATER NUREG/CR-5250 V04: SEISMIC HAZARD CHARACTENZATION OF REACTORS. 69 NUCLEAR PL/.NT SITES EAST OF THE ROCKY NUREG/CR 5176: SEISMIC FAILURE AND CASK DROP ANALYSES MOUNTAINS Results And Discussion For The Datch 3 Sites OF THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLE- NUREG'CR 5250 V05: SEISMIC HAZARD CHARACTERIZATION OF AR POWER PLANTS. 69 NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR-5186: VALUE/ IMPACT ANALYSIS OF GENERIC ISSUE MOUNTAIN 1Results And Discussion For The Batch 4 Sites.

94, " ADDITIONAL LOW TEMPERATURE OVERPRESSURE PRO. tvUREG/CR 5250 V06: SEISMIC HAZARD CHARACTERIZATION OF TECTION FOR LIGHT WATER REACTORS." 69 NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR-5197: EV%LUATION OF GENERIC (SSUE 115. "EN- MOUNTAINS. Regional Companson Between Sites, Site HANCEMENT OF THE RELIABILITY OF WESTINGHOUSE SOLIO Effects. General Discussion And Conclusions.

STATE PROTEC~lON SYSTEM." NUREG/CR 5250 V07: SE!SMIC HAZARD CHARACTERi2ATION OF NUREG/CR-5234: VALUE/lMPACT ANALYSIS FOR GENERIC ISSUE 69 NUCLEAR PLANT SITES EAST OF THE ROCKY

$1.lMPROVING THE REUABILITY OF OPEN CYCLE SERVICE- MOUNT AINS Ouestionnaves.

WATER SYSTEMS NUREG/CR-5250 V08. SEISMIC HAZARD CHARACTERIZATION OF DIVISION OF SYSTEMS RESEARCH (POST 880717) 69 NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR-4618: EVALUATION OF RELIABILITY TECHNOLOGY AP. MOUNTAINS Supplementary Seismic Hazard Results For Sites With PLICABLE TO LWR OPERATIONAL SAFETY. Multiple Soll Conditions.

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 i titles of their. reports. If further information is needed, refer to the main citation by the 1 1

' NUREG/CR number.

iL APPLIED RISK TECHNOLOGY CORP. NUREG/CR-4618: EVALUATION OF RELIABILITY TECHNOLOGY AP-PLICA 9tE TO LWR OPERATIONAL SAFETY, NUREGICR-5206: A PROBABILISTIC EVALUATION OF THE SAFETY OF BABCOCK & WILCOX NUCLEAR REACTOR POWER PLANTS NURc!G/CR-4948: TECHNICAL FINDINGS RELATED TO GENERIC WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL ISSUE 23: REACTOR COOLANT PUMP SEAL FAILURE.

EVENTS. NUREG/CR-5102: INTERF ACING SYSTEMS LOCA: PRESSURIZED WATER REACTORS.

APTECH ENGINEERING SERVICES NUREG/CR 5124: INTERFACING SYSTEMS LOCA: BOILING WATER NUREG/CR 5115: A REVIEW OF BOtLING WATER REACTOR WATER REACTORS.

CHEMISTRY. Science, Schnology,And Performance, NUREG/CR-5206: A PROBAB LtSTIC EVALUATION OF THE SAFETY i

OF BABCOCX & WILCOX NUCLEAR REACTOR POWER PLANTS l ARCONNE NATIONAL LAi> ORATORY WITH EMPHASIS ON HISTORICALLY OBSERVED OPERATIONAL MUREG/CR4667 V05: ENVIRONMENTALLY ASSISTED CRACKING IN O Of G Ti hANOR NER NUR G/C 5221: BIODEGRADAT!ON OF ION-EXCHANGE MEDIA.

N E/R 15: RV NUREG/CR-5266: EXAMINATION OF TWO 3M WPE 902F STATIC CHEMISTRY. Science, Technology.And Performance NUREG/CR-5116: SURVEY OF PWR WATER CHEMISTRY. ELIMINATORS.

NUREG/CR-5245: A REVIEW OF THE CRYSTAL RIVER UNIT 3 PROB- NUREG/CR-5324: DETECTING COMPONENT FAILURE POTENTIAL ABILISTIC RISK ASSESS 84ENT. internal Events. Core Damage Frequen- USING PROPORTIONAL HAZARD MODEL.

l cy. NUREG/CR 5325: A MULTIVARIATE STATISTICAL STUDY ON A Dl-ARIZONA, UNIV. OF, TUCSON, A2 NUREG/CR-5239: FLUID FLOW AND SOLUTE TRANSPORT MODEL-f' ING THROUGH THREF-DIMENSIONAL NETWOciKS OF VARIABLY C ALIFORNIA, UNIV. OF. SANTA BARBARA, CA NUREG/CR-5030: AN ASSESSMENT OF STEAM-EXPLOSION-lN-iN C S4 BO OLE C RE IN SALT. DUCED CONTA'NMENT FAILURE.

RATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES DAVID W. TAYLOR NAVAL RESEARCH & DEVELOPMENT CEhTER NUREG/CR4835: COMPARISON AND APPLICATION OF OUANTITA.

NUREG'CR 5143: APPLICATION OF THE JINTEGRAL AND THE MODI-TiVE HUMAN RELIABMJTY ANALYSIS METHODS FOR THE RISK FIED J INTEGRAL TO CASES OF LARGE CRACK EXTENSION.

METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Final Report. DOMINION ENGINEERING,INC.

CATTELLE MEMORIAL INSTITUTE, PACtFIC NORTHWEST NUREG/CR-5116: SURVEY OF PWR WATER CHEMISTRY.

LABORATORY NUREG/CR-3037: USER'S MANUAL FOR FIRIN.A Computer Code To EG4G IDAHO,1NC. (SUBS. 0F EG&G,1NC.)

Estimate Accidental Fire And Radioactive Airborne Release in Nuclear NUREG/CR 4639 V5P2R1: NUCLEAR COMPUTERIZED LIBRAl.Y FOR Fuel Cycle Facilities. ASSESSING REACTOR RELIABillYY (NUCLARR). Data Manual.Part 2:

- NUREG/CR4469 V07: NONDESTRUCTIVE EXAMINATION (NDE) RELi- Hume Error Probabihty (HEP) Estimates.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR4639 V5P3R1: NUCLEAR COMPUTERIZED LIBRARY FOR AS SS A ABW WM%ata ManuaN 1 NU EG R6 TYT ISL OYS$O V EAR MATERIAL Hardware Component Failure Data NCFD).

MANAGEMENT NUREG/CR4639 V5P4R1: NUCLEAR COMPUTERIZED LfDRARY FOR

. NUREG/CR4649'V01: STEAM GENERATOR GROLM PROJECT. Task 9 ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual,Part Final Report: Nondestructive Evaluation Round R(bin'v'olume 1: De. 4. Summary Aggregations.

senption And Suramary Data NUREG/CR4835: COMPARISON AND APPLICATION OF OUANTITA-NUREG/CR4849 V02: STEAM GENERATOR GROUP PROJECT. Task g Fha! Report: Nondestructive Evaluation Ruund Robin. Volume E Raw TlVE HUMAN RELIABILITY ANALYSIS METHODS FOR THE FISK Inspection Data.

METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Final Report.

NUREG/Er4879 V02: DEMONSTRATION OF PERFORMANCE MOD.

ELING OF A LOW LEVEL WASTE SHALLOW LAND BURIAL SITE.A NUREG/CR-5074 DEVELOPMENT OF A PHENOMENA IDENTIFICA-Compenson Of Predictive Radionuchde Transport Modeline Versus TION AND RANKING TABLE (PIRT) FOR THERMAL-HYDRAULIC Field Observations At The "A" Disposal Area. Chalk River Nuclea, ab- PHENOMENA DURING A PWR LARGE BREAK LOCA.

oratones NUREG/CR-5112: EVALUATION 0F BOILING WATER REACTOR NUREG/CR 5186: VALUE/lMPACT ANALYSIS OF GENERIC ISSU2 94 WATER-LEVEL SENSING LINE BREAK AND SINGLE

1. ADDITIONAL LOW TEMPERATURE OVERPRESSURE PROTE0 TION FAILURE. Generic issue 101 Boiling Water Reactor Level Redundancy- ,

FOR LIGHT WATER REACTORS " Technical Findings '

NUREG/CR 5234: VALUE/ IMPACT ANALYSIS FOR GENERIC ISSVE NUREG/CR-5197: E5/ALUATION OF GENERIC ISSUE 115. " ENHANCE. i 51: IMPROVING THE RELIABILITY OF OPEN-CYCLE SERVICE

• MENT OF THE RELIABILITY OF WESTINGHOUSE SOLt0 STATE NU G/ S4DDE (SCALED ADSORBED DOSE DISTRIBUTION NUREG/CR 5 4 THE ACHABILITY OF DECONTAMINATION ION.

EXCHANGE RESINS SOLIDIFIED IN CEMENT AT OPERATING NU-N 5 8 IkREL i Bb NP R AND TEST RE-CLEAR POWER PLANTS ACTOR DATA BASES ^ NUREG/CR-5311: AN ASSESSMENT OF RELAPS/ MOD 2 APPLICABIL, iTY TO LOSS-OF FEEDWATER TRANSIENT ANALYSIS IN A BAB-EROOKHAVEN NATIONAL LABORATORY NUREG/CR 2331 V07 N4: SAFETY RESEARCH PROGRAMS SPON- COCK AND WILCOX REACTOR PLANT.

SORED BY OFFICE OF NUCLEAR REGULATORY E OE, INC.

RESEARCH Progress Report,0ctober Lecember 1987, NUREG/C42331 V08 N3. SAFETY RESEARCH PROGRAMS SPON- NUREG/CR-St76 SEISMIC FAILURE AND CASK DROP ANALYSES OF SORED BY OFFICE OF NUCLEAR REGULATORY THE SPENT FUEL POOLS AT TWO REPRESENTATIVE NUCLFAR RESEARCH Progress Report, July-September 1988. POWER PLANTS.

45

=_ _ _ - - - _ - - _ _ _ - _ _ _ - - _ _ _ _ - _ _ _ - _ - - _ _ - _ _ _ _ _ _ _ _ - _ _____- _ - _______ _ - - _ - _ _ _ _ -.

i l

L 46 C ntr:ct:r Ind;x IN-SITU, INC. MATERIALS ENGINEERING ASSOCIATES, INC.

NUREG/CR 5251: WELLFIELD INSTALLATION AND NUREG/CR-5195: FATIGUE STRENGTH OF ASME SA 106-B WELDED INVESTIGATIONS CRESTON STUDY AREA. EASTERN WASHING- STEEL PIPES IN 288 DEGREES C AIR ENVIRONMENTS.

TON. NUREG/CR-5265: SIZE EFFECTS ON J R CURVES FOR A 302-8 PLATE.

ITASCA CONSULTING GROUP, INC. NUREG/CR-5326: FRACTURE EVALUATION OF SURFACE CRACKS NUREG/CR-5335: STABILITY OF DISPOSAL ROOMS DURING WASTE EMBEDDED IN REACTOR VESSEL CLADDING.

RETRIEVAL MATHTECH, INC.

LAWRENCE BERKELEY LABORATORY NUREG/CR-4627 RO1: GENERIC COST ESTIMATES. Abstracts From NUREG/CR4271: THERMOCHEMICAL PROPERTIES OF Generic Studies For Use in Prepanng Regulatory impact Analyses.

GIBBSITE.EiAYERITE,BOEHMITE, DIASPORE AND THE ALUMINATE NUREG/CR 5278: COST ANALYSIS FOR POTENTIAL BWR MARK I ION DETWEEN 0 AND 350 DEGREES C. CONTAINMENT IMPROVEMENTS.

LAWRENCE LIVERMORE NATIONAL LABORATORY MICHIGAN, UNIV. OF ANN AFIBOR, MI NUREG/CR-3006: DAMPING IN BUILDING STRUCTURE 0 DURING NUREG/CR-3145 V07: GEOPHYSICAL INVESTIGATIONS OF THE ,

WESTERN OHIO-INDIANA REGION. Annual Report. October 1987 4 NURE /CR-45 01 SC P CASK ANALYSIS SYSTEM).A September 1988.

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING j CASK DESIGN REVIEW. Volume 1: User's Manual To Version la (In- OAK RIDGE NATIONAL LABORATORY N C 55 V02 SC S (SHIPPING CASK ANALYSIS SYSTEM).A NUREG/CR-2000 V07N12: LICENSEE EVENT REPORT (LER)

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING

's NU EG/C V08 L E EE E /ENT REPORT (LER)

COMPILAllON:For Month Of January 1989.

NURE / 45 3 4N ( lPN ASK Y lh T I ). ^

MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING MANIPULATOR. Adapted For The Data General MV Sanes Eclipse CASK DESIGN REVIEW. Volume 3: Theory Manual Lead Slump in N G CR 4 S S H N NU E 5 I OF RADIATION EMBRITTLEMENT ON IN-ALYSIS SYSTEM):A MICF.OCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING TEGRITY OF PRESSURE VESSEL SUPPORTS FOR TWO PWR CASK DESIGN REVIEW. Volume 4: Theoiy Manual - Thermal Analysis. PLANTS.

NUREG/CR44554 V05: SCANS (SHIPPING CASK ANALYSIS SYSTFM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIPPING SK D SIGN REVIEW. Volume 5: Theory Manus: Thermal / Pressure NUREG/ R 2 b CLOSEOUT OF IE BULLETIN 80-23. FAILURES OF SOLENOID VALVES MANUFACTURED BY VALCOR ENGINEERING NUREG/CR-4792 V01: PROBABILITY OF FAILURE IN BWR REACTOR CORPORATION.

COOLANT PIPING. Summary Report PICKARD, LOWE & GARRICK, INC.

NUREG/CR-4702 V02. PROBABILITY' OF FAILURE IN BWR RF NCTOR COOLANT PIPING Pipe Failure induced By Crack Growth A..a Failure NUREG/CR-4780 V02: PROCEDURES FOR TREATING COMMON Of intermediate Supports. CAUSE F AILURES IN SAFETY AND RELIABILITY STUDIES. Analytical NUREG/CR-5042 S02. EVALUATION OF EXTERNAL HAZARDS TO NU. Background And Techniques.

CLE POWER PLANTS IN THE UNITED STATESOther External OUANTEX Cr,elp. j NUREG/CR 5176: SEISMIC FAILURE AND CASK DROP ANALYSES OF NUREG/CF 5101 INTEGRATING FIBER OPTIC RADIATION DOSIME-THE SPENT FUEL. POOLS AT TWO REPRESENTATIVE NUCLEAR TER.

POWER PLANTS NUREG/CP-5250 V01: SEISMIC HAZARD CHARACTERIZATION OF 69 SANDIA NATIO (AL LABORATORIES NUCLEAR PLANT SITES EAST OF THE ROCKY NUREG/CR4046: A FINITE ELEMENT ANALYSIS OF A REACTOR MOUNTAINS. Methodology, Input Data And Compansons To Previous PRESSURE VESSEL DURING A SEVERE ACCIDEN L NUREG/CR-5088: FIRE RISK SCOPING STUDY, INVESTIGATION OI Results For Ten Test Sites.

NUREG/CR-5250 V02: SEISMIC HAZARD CHARACTERIZATION OF 69 NUCLEAR POWER PLANT FIRE RISK,1NCLUDING PREVIOUSLY UN-NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results ADDRESSEDISSUES NUREG/CR-5173; REVIEW AND ASSESSMENT OF THERMODYNAMIC And Discussion For The Ba*ch 1 Sites NUREG/CR-5250 V03: SEISMIC HAZARD CHARACTERIZATION OF cg AND TRANSPORT PROPERTIES FOR THE CONTAIN CODE.

NUCLEAR PLANT RITES EAST OF THE ROCKY MOUNTAINS ReLufts NUREG/CR-5313. EOUIPMENT QUALIFICATION (EO)-RISK SCOPlNG j And Discuse:on For The Batch 2 Sites. STUDY.

I NUREG/CR-5250 V04: SEISMIC HAZARD CHARACTER!ZATION OF 69 CC& A, INC.

NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS.Results And Discussion For The Batch 3 Sites. NUREG/CR 4627 RO1: GENERIC COST (STIMATES. Abstracts From NUREG/CR-5250 V05: SEISMIC HAZARD CHARACTERIZATION OF 69 Genenc Studies For Use in Prepanng Regulatory impact Analyses 4 NUCLEAR PLANT SITES EAST OF THE ROCKY MOUNTAINS Results NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK i l And Discussion For The Batch 4 Sites CONT AINMENT IMPROVEMENTS ]

NUREG/CR 5250 V06: SEISMIC HAZARD CHARACTERIZATION OF 69 d NUCLEAR PLANT SITES EAST OF THE ROCKY SCIENCE & ENGINEERING ASSOCIATES,INC.

l MOUNTAINS. Regional Companson Between Sites Site Effects, General NUREG/CR 4627 RO1: GENERIC COST ESTIMATES.Abstra;ts From Discussion Ar j Conclusions. Genenc Studies For Use in Prepanng Regulatory impact Analyses.

NUREG/CR-5250 V07: SEISMIC HAZARD CHARACTERl2ATION OF 69 NUREG/CR-5278: COST ANALYSIS FOR POTENTIAL BWR MARK l NUCLEAR PLANT SITES EAST OF JHE ROCKY CONTAINMENT IMPROVEMENTS MOUNTAINS Questionnaires.

NUREG/CR-5250 V08: SEISMIC HAZARD CHARACTERIZATION OF 60 SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY NUCLEAR PLANT SITES EAST OF THE ROCKY SCIENCE APPLICATIONS, MOUNTAINS Supplementary Seismic Hazard Results For Sites With NUREG/CR 4618. EVALUATION OF RELIABILITY TECHNOLOGY AP.

Muftiple Soil Conditions PLICABLE TO LWR OPERATIONAL SAFETY.  !

NUREG/CR-5270: ASSESSMENT OF SEISMIC MARGIN CALCULATION NUREG/CR-5313 EOUlPMENT QUALIFICATION (EO) RISK SCOPING l METHODS. STUDY.

LINSLEY, KRALGER & ASSOCIATES, LTD. SRI INTERN ATION AL NUREG/CR-4496 V02: A SYSTEM FOR GENERATING LONG STREAM- NUREG/CR-4950 V03. THE SHORELINE ENVIRONMENT ATMOS- )

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN PHERIC DISPERSION EXPERIMENT (SEADEX) Airborne LIDAR Data j~

PERIOD Phase 11.

U.S. NAVAL ACADEMY, ANN APOLIS, MD LOS ALAMOS NATIONAL LABORATORY NUREG/CR 5143 APPLICATION OF THE J-INTEGRAL AND THE MODI-NUREG/CR 5069 TRAC-PF1/ MOD 1. Correlations And Models. FiED J-INTEGRAL TO CASES OF LARGE CRACK EXTENSION 9

l International Organization index This index lists, in alphabetical order, the cour# ries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed below each country and per-

"orming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NLIREG/lA number.

There were no NUREG/lA reports for this quarter.

47

I I

-1 1

}

I 1

41 1 4

E L_.. -- _ __ __ __ _ _ _ - _ _ - - - _ -

Licensed Facility index i

This index lists the facilities that were the subject of NRC staff or contractor reports. The '

f cility names are arranged in alphabetical order. They are preceded by their Docket number ]

i cnd followed by the report number. If further information is needed, refer to the main citation J by the NUREG number.

l 1

50 424 AMn W. Vogtle Nuclear Plant, Unt 1, George fiURE41137 S08 50 327 Sequoyah Nudeer Plant, Unit 1, Tennesene NURE41232 V02 801 S 424 AMn Voglie Nuclear Plant, Unit 1. Geoqps NURE41137 S00 S 326 Plant Unit 2, Tennseems id)RE41232 V02 801 ma av% e unit.% . E4,us $mm9. 43^.".% = 1. Nouston u,,.ng & .E4 4 1 54414 AMn Voglio Nucteer Plant, Ura 1, Geoqpa NURE41359 STN.60498 South e Pngsct, Unit 1. Nouston Ughtng & NURE41M5 Power Co. Power Co.

S425. AMn W. Voglie Nuclear Plant, Unit 2, Geoqpa NURE41137 SOB ST454499 South Tauss Proinct, Unit 2, Houston Ughtng & NURE4041308 50 425 AMn Vogtle Nuclear Plant, Unn 2, Geoqpa NURE41137 SO9 STN 50499 South Pngsct, Unit 2, Houston Ughtmg & NUREGC781807 Power Co. Power Co.

50425 AMn W. Voglio Nuclear Plant, Unit 2, Geoqpa NURE41343 STN 50499 South Texas Pngsct, Unit 2, Houston Ughtng & NURE41334 Power Co. Power Co.

54425 Ahn W. Vogtle Nuclear Plant, Und 2, Gentgla NUREG1359 STF54'89 South Texas Project. Unn 2, Houston UghtnD& NURE41346 503c2 Crys YR Nuclear Plant, Unit 3, Flonda Power NUREG/CR 524! 54320 Thr Wile telend N NURE41355 Coq,.

- u.udear Seton .e. Unit 2, I

49

_._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - . - - - _ . - _ . . _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ a

- - ~ -.

NRC $0RM 335 - U.S. NUCLE AR REGUL AT ORY COMMIS510N 1 REPORT NUMBEH L*!a iio2.- o'um N%m [esNfy7 mum BIBLIOGRAPHIC DATA SHEET (See instructions on the reverse) N0R(0-0304 .

A

2. m LE AND $USMLE Vol. 14, No. 1 Regulatory and Technical Reports (Abstract Index Journal) y g,.rE ReeORT eustissED uw n Au Compilation.for First Quarter 1989 July l 1989~

January - March - UtN OR MANT NUYSE R S. AUTHOR (S) 6 T YPE OF REPORT Reference l

7. VE R IOD COV E H E D nnciossro. D.orw January - March 1989

8. PE R F OR MING ORG ANIZ AT lON - N AME AND ADDREYS tar NRc. provoo+ 0ovoston. orroce ur Regron. U.S Nuctw negulatory coormossoon. and mornny eJdreu.'s! contr<n too. ProvWr.

name and marinns ocorem t Division of Freedom of Information and Publications Services Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555

9. SPONSORING ORG ANtZATION - NAME AND AODRESS ur NRC. type %me ns atoove" dicontractor, provnt* NRC Dtvisron Office er negion. U S Nucivor negulatorv Commwon end mentmg audron t I

I Same as e, above.

10. SUePLEME NT ARY NOTES

11. ABST R ACT (200 words or erw/

This journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceedings of conferences and workshops; a' 11 as international agreement reports. The entries in this compilation are indeLJ for access by title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agreements, contractor, international organization, and licensed facility.

12, KE Y WORDS,'UESCR P T ORS {L uo words or onesses anot wesi assist researchers m iocorme tne reporn 1 13 AV A'Lkts' u l ' b i n i L Mk k l~

Unlimited compi1ation a ~ ~ u*ssa-*~

abstract inde:. " ~, u,~

Unclassified tin., n,mru Unc1assified Ib NUMBEH Of FAGEb 16 PHILL NRC FORM 335 Q.891

wucuAn nUUIEr'oYv*c!>uusssion '"W^EU$YIE'"I sN $

WASHINGTON, D.C. 20566 w.m:==. - g

.E Main Citatione y and Abstracts [

w PIV F y' IANIAC29ty99 r ;509 PUR~tuRLfICATiang gyr Secondary Report WAsHINGrote Number index DC 20599 Personal Author index E

o C

Subject Index g Cl E

>l 5$

NRC Originating Organization Zq Index (Staff Reports) $R_-

se .

<E I -

>Q:.

5r E:e.

NRC Originating Organization  :: Q index (International Agreements) g w

n-O 2:

c NRC Contractor g SponsorIndex g C

z Contractor index International Organization Index ,

E G

5 Licensed Facility Index