ML20137A618

Regulatory and Technical Reports.Compilation for Third Quarter 1985,July - September

ML20137A618
Person / Time
Issue date:	10/31/1985
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V10-N03, NUREG-304, NUREG-304-V10-N3, NUDOCS 8511260053
Download: ML20137A618 (72)

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NUREG-0304 Vol.10, No. 3 Regulatory and Technical Reports (Abstract Index Journal;'

Compilation for Third Quarter 1985 July - September U.S. Nuclear Regulatory Commission Offico of Administration

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Available from Superintendent of Documents U.S. Government Printing Office p' -

Post Office Box 37082 Washington, D.C. 20013 7082 A year's subscription consists of 4 issues for this publication.

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

NUREG-0304 Vol.10, No. 3 Regulatory and Technical Reports (Abstract Index Journal)

Com ailation for I Thirc Quarter 1985 July - September Date Published: October 1985 Policy and Publications Management Branch Division of TechnicalInformation and Document Control Offico of Administration U.S. Nuclear Regulatory Commission Washington, D.C. 20666 f

1 CONTENTS Preface v Index Tab Main Citation and Abstracts 1 Staff Reports. ... . . .

Conference Proceedings . . .

Contractor Reports . . . . . . . .

Contractor Report Number Index . . 2 Personal Author Index , . . 3 Subject index ... .............. .. 4 NRC Originating Organization index (Staff Reports) 5 NRC Contract Sponsor Index (Contractor Reports) . . . 6 Contractor index . . . . 7 Licensed Facility Index . . . 8 iii i

1

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

i Division of Technical Information and Document Control Policy and Publications Management Branch Publishing and Translations Section Woodmont 501 l U.S. Nuclear Regulatory Commission i Washington, D.C. 20555 l l

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

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

NRC Contract Sponsor Index (Contractor Reports)

Contr9ctor Index Licensed Facility Index A detailed explanation of the entries precedes each index.

1 The bibliographic elements of the main citations are the following.

Staff Report NUREG-0508: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

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

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

Conference Report

, NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND l l RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070.

Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC ir:ternal use).

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

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

Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if

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

i l

l l v

1 i

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

ADD - addendum APP - appendix DRFT - draft l ERR - errata N - number l R -

revision i

S - supplement V - voiume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the NRC-GPO Sales Office or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the NRC-GPO Sales Office send a check or money order, payab!e to the Superintendent of Documents, to the following address:

U.S. Nuclear Regulatory Commission ATTN: Sales Manager Washington, D.C. 20555 You may charge any purchase to your GPO Deposit Account, Master Charge card, or VISA charge card by calling the NRC-GPO Sales Office on (301) 492-9530. Non-U.S. customers must make payment in advance either by International 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.

All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Technical Information and Document Control.

vi

Main Citations and Abstracts The report listings in this com ailation are ar- is an NRC contractor-prepared report. The ranged by report number, where NUREG- bibliographic information (see Prefaco for XXXX is an NRC staff-originated report, details) is followed by a brief abstract of this NUREG/CP-XXXX is an NRC-sponsored report.

conference report, and NUREG/CR-XXXX NUREG-0020 V09 N06: UCENSED OPERATING REACTORS Agreement State (Texas). Three events involved radiation over.

STATUS

SUMMARY

REPORT. Data As Of May 31,1985.(Gray exposures; the other event involved a well logging source wnsen Book I) ROSS,P.A.; BEEBE M.R. Division of Budget & Analysis. was apparently stolen, but later was recovered. The report also Jufy 1985. 437pp. 8508190629. 32261:001. contains information updating some previously reported abnor.

The OPERATING UNITS STATUS REPORT LICENSED OP- mal occurrences.

ERATING REACTORS provides data on the operation of nucle-ar units as timely and accurately as possible. This information is NUREG-0304 V10 N02: REGULATORY AND TECHNICAL collected by the Office of Resource Management from the REPORTS. Compilation For Second Quader 1985.

• Division of Headquarters staff of NRC's Office of Inspection and Enforce- Technical Information & Document Control. July 1985. 88pp.

ment, from NRC's Regional OHices, and from utilities. The three 8508150008. 32198:129.

sections of the report are: monthly highlights and statistics for This joumal lists all formal reports in the NUREG series pre-commercial operating units, and errata from previously reported pared by the NRC staff and contractors, as well as proceedings data; a compilation of detailed information on each unit, provid- of conferences and workshops. The entries in the compilation ed by NRC's Regional Offices, IE Headquarters and the utilities; are indexed for access by title and abstract, contractor report and an appendix for miscellaneous mformation such as spent number, personal author, subject NRC organization, contractor, fuel storage capability, reactor-years of experience and non-power reactors in the U.S. It is hoped the report is helpful to all and licensed facility' agencies and indmduals interested in maintaining an awareness of the U.S. energy situation as a whole.

NUREG-0386 D03: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE NUREG-0020 V09 N07: UCENSED OPERAT!NG REACTORS DIGEST. JULY 1972 - SEPTEMBER 1983.

• Office of the Execu-STATUS

SUMMARY

REPORT. Data As Of June 30.1985.(Gray tive Legal Director.

• Aspen Systems, Inc. Jufy 1985. 800pp.

Book 1) ROSS P.A.; BEEBE M.R. Division of Budget & Analysis. 8508210006. 32303:334 August 1985. 426pp. 8509130022. 32620:001. This edition of the NRC Staff Practice and Procedure Digest See NUREG-0020,V09,N06 abstract. contains a digest of a number of Commission, Atomic Safety NUREG-0020 V09 N08: LICENSED OPERATING REACTORg and Licensing Appeal Board, and Atomic Safety and Ucensing STATUS

SUMMARY

REPORT. Data As Of July 31,1985.(Gray Board decisions issued during the period from July 1,1972 to Book I)

• Division of Budget & Analysis. September 1985. September 30,1983 nterpreting the NRC's Rules of Practice in 405pp. 8510070175. 32902:037. 10 CFR Part 2. This edition replaces earlier editions and supple-See NUREG-0020,V09,N06 abstract. ments and includes appropriate changes reflecting the amend.

NUREG-0040 V09 N02: UCENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Quarterty NUREG-0540 V07 N05: TITLE UST OF DOCUMENTS MADE Report. April-June 1985. (White Book)

• Division of OA, Vendor PUBLICLY AVAILABLE.May 1-31, 1985.

• Division of Technical

& Technical Training Center Programs (Post 850212). August Information & Document Control. July 1985. 489pp.

1985. 245pp. 8509060260. 32503:078.

8507250203. 31790:011 This pe'iodical covers the results of inspections performed by the NRC s Vendor Program Branch that have been distnbuted This document is a monthly publication containing de^icrip-to the inspected orr,anizations during the period from April 1985 tions of information received and generated by the U.S. NRC.

through June 1985. Also included in this issue are the results of This information includes (1) docketed material associated with certain inspections performed prior to April 1985 that were not civilian nuclear power plants and other uses of radioactive ma-included in previous issues of NUREG-0040. terials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following NUREG-0000 VOS Not: REPORT TO CONGRES3 ON ABNOR- indexes are included: Personal Author IndeF, Curporate Source MAL OCCURRENCES.Janurary-March 1985.

• AEOD, Director's Index, Report Number index, and Cross Reference to Principal Office. August 1985. 46pp. 8509060189. 32505:273.

Documents Index.

Section 208 of the Energy Reorganization Act of 1974 identi-fies an abnormal occurrence as an unscheduled incident or NURrg.0540 V07 N16: TITLE UST OF DOCUMENTS MADE event which the Nuclear Regulatory Commission determines to PUBilCLY AVAILABLE. June 1-30, 1985.

• Division of Technical be significant from the standpoint of public health and safety Information & Document Control. July 1985. 577pp.

and requires a quartwly report of such events to be made to 8508150439.32218:238.

Congress. This report covers the period January 1 to March 31, See NUREG-0540,V07,N05 abstract.

1985. During the report period, there was one abnormal occur-rencs at the nuclear power plants licensed to operate; the event NUREG-0540 V07 N07: TITLE LIST OF DOCUMENTS MADE involved a premature criticality during teactor startup. There PUBLICLY AVA! LADLE. July 1-31, 1985.

• Division of Technical were three abnormal occurrences at the other NRC licensees.

Two events involved diagnostic medical misadministrations and Information & Document Control. August 1985. 638pp.

the other event involved unlawful possession of radioactive ma- 8509190103.32670:122' terial There were four abnormal occunences reported by an See NUREG-0540,V07.N05 abstract.

1

2 Main Citations and Abstracts NUREG-0540 V07 N08: TITLE LIST OF DOCUMENTS MADE NUREG-0750 V22 N01: NUCLEAR REGULATORY COMMISSION PUBLICLY AVAILABLE. August 1-31, 1985.

• Division of Techni- ISSUANCES FOR JULY 1985. Pages 1176.

• Division of Tech-cal Information & Document Control. September 1985. 610pp. nical information & Document Control. August 1985. 184pp.

8510070182. 32903:117, 8509300527. 32787:308.

See NUREG 0540,V07,N05 abstract. See NUREG-0750,V21,N05.

NUREG-0606 V07 NO3: UNRESOLVED SAFETY ISSUES NUREG-0798 S06: SAFETY EVALUATION REPORT RELATED

SUMMARY

. Data As Of August 16,1985. (Aqua Book)

• Drvision TO THE OPERATION OF FERMI-2. Docket No. 50-341.(Detroit of Safety Technology. August 1985. 55pp. 8509260507. Edison Company)

• Division of Licensing July 1985. 56pp.

11631:363. 8508210034. 32302:259.

This report provides an overview of the status of the progress Supplement No. 6 to the Safety Evaluation Report (SER) re-and plans for resolution of the genenc tasks addressing "Unre. lated to operation of the Fermi-2 facility addresses items perti-solved Safety issues" as reported to Congress. nent to the issuance of the full power license for Fermi-2. The Fermi-2 facility is located on Lake Ene in Monroe County, NUREG-0675 S32: SAFETY EVALUATION REPORT RELATED almost 8 miles east-northeast of Monroe, Michigan.

TO THE OPERATION OF DIABLO CANYON NUCLEAR NUREG-080013.5.2 R1: STANDARD REVIEW PLAN FOR THE POWER PLANT, UNITS 1 AND 2. Docket Nos.50 275 And 50-323.(Pacific Gas and Electric Company)

• Division of Licensing. REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 1 To Section 13.5.2, August 1985. 50pp. 8508210398. 32337:324.

.. Operating And Maintenance Procedurea," and Revision 0 of Supplement 32 to the Safety Evaluation Report for the apple.-

Appendix A to Section 13.5.2, " Review... "

• Office of Nuclear cation by Pacific Gas and Electric Company for licenses to op erate Diablo Canyon Nuclear Power Plant, Units 1 and 2 Reactor Regulation, Director. July 1985. 29pp. 8508150055.

(Docket Nos. 50-275/323) has been prepared by the Office of 32195:180' No.1 to Section 13.5.2 and Revision 0 to Append Revision Nuclear Reactor Regulation of the U.S. Nuclear Regulatory A of Section 13.5.2 of the Standard Review Plan incorporates Commission. This supplement provides the staff evaluation of those matters that require an appropriate re changes that have been developed since the original issuance power operation of Un t 2 and updates prev, solution ious supplements t pnor in July to full- 1981. This revision incorporates guidelines of Task the Safety Evaluation Report. Action Plan items f.C.1 and I.C.9 of NUREG-0660 as clanfied in Supplement 1 of NUREG-0737. Appendix A to SRP Section NUREG-0748 V05 N05: OPERATING REACTORS LICENSING 13.5.2 was formerfy NUREG-0899.

ACTIONS

SUMMARY

. Data As Of May 31,1985. (Orange Book) NUREG-0837 V04 N04: NRC TLD DIRECT RADIATION MONI.

• Management Support Branch. July 1985.357pp.8507250167. TORING REPORT. Progress Report, October December 1984.

31792:028. JANG J.; KRAMARIC.M.; COHEN L Region 1. Office of Direc-The Operating Reactors Licensing Actions Summary is de- tor. Juty 1985. 316pp. 8508010750. 31924:148.

signed to provide the Management of the Nuclear Regulatory This report provides the stetus and results of the NRC Ther-Commission (NRC) with an overview of licensing actions dealing moluminescent Dosimeter (TLD) Direct Radiation Monitoring with the operating power and nonpower reactors. Network. It presents the radiation levels rneasured in the vicinity of NRC licensed facility sites throughout the country for the NUREG-0748 V05 N06: OPERATING REACTORS LICENSING fourth quarter of 1984.

ACTrONS

SUMMARY

. Data As Of June 30,1985. (Orange Book)

Management Support Branch. August 1985. 358pp. NUREG-0837 V05 N01: NRC TLD DIRECT RADIATION MONI-8508210039.32302:336. TORING NETWORK. Progress Report, January-March 1985.

See NUREG-0748,V05,N05 abstract. JANG,J.; KRAMAR;C,M.; COHEN,L Region 1, Office of Direc-tor. July 1985.152pp. 8508020373. 31961:073.

NUREG-0748 V05 N07: OPERATING REACTORS LICENSING This report provides the status and results of the NRC Ther.

ACTIONS

SUMMARY

. Data As Of July 31,1985.(Orange Book)

• molumincscent Dosimeter (TLD) Direct Radiation Monitoring Management Support Branch. September 1985. 475pp. Network. It presents the radiation levels measured in the vicinity 8510030438. 32846:287. of NRC licensed facility sites throughout the country for the first See NUPEG4748.V05.N05 abstract. quarter of 1985.

NUREG-0750 V21102: INDEXES TO NUCLEAR REGULATORY NUREG-0837 V05 NO2: NRC TLD DIRECT RADIATION MONI-COMMISSION ISSUANCES. January-June 1985.

• Division of TORING NETWORK. Progress Report, Apni-June 1985.

Technical information & Document Control September 1985. JANG,J.; KRAMGIC,M. COHEN.L Region 1. Office of Direc-108pp. 8510030126. 32855:009. tor. September 1985. 225pp. 8510020232. 32828:336.

Digests and indexes for issuances of the Commission, the This report provides the status and results of the NRC Ther-Atomic Safety and Licensing Appeal Pan 31, the Atomic Safety mofuminescent Dosimeter (TLD) Direct Radiation Monitonng and Licensing Board Panel, the Administrative Law Judge, the Network. It piesents the radiation levels measured in the vicinity Directior's Decisions, and the Denials of Petitions for Rulemak- of NRC licensert facility sites throughout the country for the ing are presented. second quartor of 1985.

NUREG-0750 V21 N05: NUCLEAR REGULATORY COMMISSION NUREG-0856 DRFT FC: REASSESSMENT OF THE TECHNICAL ISSUANCES FOR MAY 1985. Pages 1,043-1,567.

• Division of BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For Technical Information & Document Control. July 1985. 524pp. Comment) SILBERBERG,M.; PASEDAG W.F.; RYDER,C.P.; et 8508260300. 32370:304. al. Accident Source Term Program Office. Jufy 1985. 265pp.

Legalissuances of the Commission, the Atomic Safety and Li. 8508190634. 32262:085.

censing Appeal Panel, the Atomic Safety and Licensing Board NUREG-0956 describes the NRC staff and contractor efforts Panel, the Administrative Law Judge, and NRC Program Offices. to reassess and update the agency's analytical procedures for estima'.ing accident source terms for nuclear power plants. Tht, NUREG-0750 V21 N06: NUCLEAR REGULATORY COMMISSION effort included development of a new source term analytical ISSUANCES FOR JUNE 1985. Pages 1,569-1,786.

• Division of procedure - a set of computer codes - that is intended to re-Technical information & Document Control. August 1985. place the methodology of the Reactor Safety Study (WASH-215pp. 8509230739. 32702:248. 1400) and to be used in reassessing the use of TID-14844 as-See NUREG-0750,V21,N05 abstract. sumptions (10 CFR 100). NUREG-0956 describes the develop-

MaM Citations and Abstracts 3 ment of these codes, the demonstration of the codes to calcu- sumptions (10 CFR 100). NUREG 0956 desenbes the develop-late source terms for specific cases. the peer review of this ment of these codes, the demonstration of tha codes to caku-work, some perspectNes on the overall impact of new source late source terms for specific cases, the peer review of this terms on plant risks, the plans for related research projects, and work, some perspectives on the overall impact of new source the conclusions and recommendations resulting from the effort. terms on plant risks, the plans for related research projects, and NUREG-0896 S03: SAFETY EVALUATION REPORT RELATED the conclusions and recommendations resulting from the effort.

TO THE OPERATION OF SEABROOK STATION, UNITS 1 AND

2. Docket Nos. 50-443 And 50-444.(Public Service Company of NUREG 0979 SO4: SAFETY EVALUATION REPORT RELATED New Hampshire,et al)

• Division of Licensing. July 1985.94pp. TO FINAL DESIGN APPROVAL OF THE GESSAR ll.BWR/6 NUCLEAR ISLAND DESIGN. Docket No. 50-447.(General Elec-p eme t o 3 to the Safety Evaluation Report for the ap-plication filed by Public Service Company of New Hampshire, et 508 203 31961 al. for licerces to operate the Seabrook Station, Units 1 and 2, Supplement 4 to the Safety Evaluation Report (SER) for the located in Rockingham County, New Hampshire, has been pre- a ! cation filed by General Electnc Company for the final pared by the Office of Nuclear Reactor Regulation of the U.S. design approval for the GE BWR/6 nuclear island design Nuclear Regulation Commission. This supplement provides in- (GESSAR 11) has been prepared by the Office of Nuclear Reac-formation to update the status of the NRC review of the applica- tor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR 11 SER (NUREG-0979) issued in April 1983 summarizing the results of the staff's safety review NUREG-0933 S03: A PR:ORITIZATION OF GENERIC SAFETY of the GESSAR 11 BWR/6 nuclear island design; Supplement 1, ISSUES. EMRIT R.: MINNERS,W.: VANDERMOLEN,H.: et al. issued in July 1983; Supplement 2, issued in November 1984; Division of Safety Technology. July 1985. 269pp. 8508150024. and Suppiement 3, issued in January 1985. Subject to favorable 32194:252.

resolution of the items discussed in this supplement, the staff The report presents the prionty rankings for generic safety concludes that the GESSAR 11 design satisfactonly addresses issues related to nuclear power plants. The purpose of these the severe-accident concerns desenbed in the Commission's rankings is to assist in the timely and efficient allocation of NRC Policy Statement on Severs Reactor Accidents Regarding resources for the resolution of these safety issues that have a Future Designs and Existing Plants.

significant 5 otential for reducing risk. The safety prionty rankings are HIGH, MEDIUM LOW , and DROP and have been assigned NUREG-0989 S02: SAFETY EVALUATION REPORT RELATED on the basis of nsk significance estimates, safety issues were TO THE OPERATION OF RIVER BEND STATION. Docket No.

implemented, and the consideraton of uncertainties and other 50-458.(Gulf States Utilities Company, Cajun Electric Power Co-quantitative or qualitative factors. To the extent practical, esti- operative)

• Division of Licensing. August 1985. 247pp.

mates are quantitative.

8508210406. 32338:015.

NUREG-0936 V04 N02: NRC REGULATORY AGENDA.Ouarterty Supplement No. 2 to the Safety Evaluation Report for the ap-Report,ApniJune 1985.

• Division of Rules and Records. July plication filed by Gulf States Utilities Company as applicant and 1985. 216pp. 8508090725. 32101:305. for itself and Cajun Electric Power Cooperative, as owners, for a The NRC Regulatory Agenda is a compilation of all rules on license to operate River Bend Station has been prepared by the which the NRC has proposed or is considering action and all Office of Nuclear Reactor Regulation of the U.S. Nuclear Regu-petitions for rulemaking which have been received by the Com- latory Commission. The facility is located in West Feliciana mission and are pending disposition by the Commission. The Parish, near St. Francisville, Louisiana. This supplement reports Regulatory Agenda is updated and issued each quarter. The the status of certain items that had not been resolved at the Agendas for April and October are published in their entirety in time of publication of the Safety Evaluation Report and Supple-the Federal Register while a notice of availability is published in ment No.1.

the Federal Register for the January and July Agendas.

NR 89 SR SAM NMON REM REWED NUREG-0940 V04 NO2: ENFORCEMENT ACTIONS.SIGNIFICANT RESOLVED.Ouarterfy Progress Report,Apnt- TO THE OPERATION OF RIVER BEND STATION Docket No.

ACTIONS June,1985. , Director's Office Ofice of Inspection and Enforce-50-458.(Gulf States Utilities Company)

• Division o Licensing.

ment. July 1985. 341pp. 8508190616. 32260:001. August 1985. 305pp. 8509100337. 32529:148.

This compilation summanzes significant enforcement actions Supplement No. 3 to the Safety Evaluation Report fc' the ap-that have been resolved dunng one quarterfy period (Apnl - lication filed by Gulf States Utilities Company as appli: ant and June 1985) and includes copies of letters, Notices, and Orders for itself and Cajun Electric Power Cooperative, as owr ers, for a sent by the Nuclear Regulatory Commission to licensees with license to operate River Bend Station has been prepared by the respect to these enforcement actions and the licensees' re- Of$ce of Nuclear Reactor Regulation of the U.S. Nuclear Regu-sponses. It is anticipated that the information in this publication latory Commission. The facility is located in West Feliciana will be widely disseminated to managers and employees en- Parish, near St Francisville, Louisiana. The supplement reports gaged in actaiGes T.cerised by the NRC in the interest of pro- the status of Main items that had not been resolved at the moting public health and safety as well as common defense time of publication of the Safety Evaluation Report and Supple-and security. ment Nos. I and 2.

NUREG-0956 DRFT FC: REASSESSMENT OF THE TECHNICAL NUREG-0989 SO4: SAFETY EVAlUAT!ON REPORT RELATED BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For TO THE OPERATION OF RIVER BEND STATION. Docket No.

Comment). SILBERBERG.M.; MITCHELL,J.A.: MEYER,R.O.; et 50-458.(Gulf States Utilities Company, Cajun Electric Power Co-al. Accident Source Term Program Office. Jufy 1985. 265pp. operative)

• Division of Licensing. September 1985. 39pp.

8508190634. 32262:085. 8509260506. 32758:043.

NUREG 0956 describes the NRC staff and contractor efforts Supplement No. 4 to the Safety Evaluation Report for the ap-to reassess and update the agency's analytical procedures for plication fi,ad by Gulf States Utilities Company as applicant and estimating accident source terms for nuclear power plants. The for itself and Cajun Electric Power Cooperative, as owners, for a effort included development of a new source term analytical license to operate River Bend Station has been prepared by the procedure - a set of computer codes - that is intended to re- Office of the Nuclear Reactor Regulation of the U.S. Nuclear place the methodology of the Reactor Safety Study (WASH- Regulatory Coinmission. The facility is located in West Feliciana 1400) and to be used in reassessing the use of TID-14844 as- Parish, near St. Francisville, Louisiana.

4 Main Citations and Abstracts NUREG-0991 S05: SAFETY EVALUATION REPORT RELATED significant changes in design enteria and method for the seismic TO THE OPERATION OF LIMERICK GENERATING qualifcation of equipment over the years. Therefore, the seismic STATION, UNITS 1 AND 2. Docket Nos 50-352 And 50-353. qualification of equipment in operating plants should be reas-(Philadelphia Electric Company)

• Dmsion of Licensing. July sessed to deternine whether requalifcation is necessary. The 1985. 52pp. 8508020362. 31962:001. objective of technical studies performed under the Task Action in August 1983 the NRC issued its Safxy Evaluation Report Plan A-46 was to establish an explcit set of guidelines and ac-regarding the application for licenses to operate the Lamenck ceptance enteria to judge the adequacy of equipment under Generatmg Station, Units 1 and 2 located on a site in Montgom- seismic loading at all operating plants, in heu of requinng qualife-ery and Chester Counties, Pennsyfvania. Supplement 1 was cation to the current enteria that are applied to new plants.

issued in December 1983 and addressed several outstanding issues. SSER 1 also contains the comments made by the Advi. NUREG-1031 S02: SAFETY EVALUATION RELATED TO THE sory Committee on Reactor Safeguards in its interim report OPERATION Or MILLSTONE NUCLEAR POWER dated October 18,1983. Supplement 2 was issued in October STATION, UNIT 3. Docket No 50-423.(Northeast Nuclear Energy 1984. Supplement 3 was issued in October 1984 and addressed Company)

• Division of Licensing. September 1985. 100pp.

the remaining issues that required resolution before issuance of 8510010227. 32836:195.

the operabng license for Unit 1 On October 26,1984 a license The Safety Evaluation Repurt issued in August 1984 provided (NPF-27) for Unit I was restricted to a five percent power level the results of the NRC staff review of Northeast Nuclear Energy and contained conditions which required resolution prior to pro. Company's application for a imense to operate the Millstone Nu-ceeding beyond the five percent power level. Supplement 4 clear Power Station, Unit No. 3. Supplement No.1 to ,that issued in May 1985 addressed some of the technicalissues and report, issued in March 1985 updated the information contained their associated license conditions, which required resolution in the Safety Evaluation Report and addressed the ACRS prior to proceeding beyond the five percent power levet. SSER Report issued on September 10,1984. This Report, Supplement 4 also contained the comments made by the Advisory Commit. No. 2 update 2 the information contained in the Safety Evalua-tee on Reactor Safeguards in its report dated November 6, tion Report and Supplement No.1 and addresses prior unre-1984. This Supplement 5 to the SER addresses further issues solved items. The facility is located in Waterford Township, New that require resolution pnor to proceeding beyond the five per. London, Connecticut.

NUREG-1048 S02: SAFETY EVALUATION REPORT RELATED NUREG-0991 S06: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF HOPE CREEK GENERATING TO THE OPERATION OF LIMERICK GENERATING STATION. Docket No. 50-354.(Public Service Electric and Gas STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50- Company)

• Division of Licensing. August 1985. 92pp.

353.(Philadelphia Electric Company)

• Division of Licensing. 8508190624. 32302:143.

August 1985.19pp. 8508210037. 32302:317. Supplement No. 2 to the Safety Evaluation Report on the ap-in August 1983 the staff of the Nuclear Regulatory Commis- plication filed by Public Service Electric and Gas Company as sion issued its Safety Emfuation Report (NUREG-0991) regard- applicant for itself and Atlante City Electnc Company, as ing the application of the Philadelphia Electric Company (the li- owners, for a license to operate Hope Creek Generating Station censee) for heenses to operate the Limerick Generating Staticn, has been prepared by the Office of Nuclear Reactor Regulation Units 1 and 2 located on a site in Montgomery and Chester of the U.S. Nuclear Regulatory Commission. The facility is locat-Counties, Pennsytvania. Supplement 1 was issued in December ed in Lower Alloways Creek Township in Salem County, New 1983. Supplement 2 was issued in October 1984. Supplement 3 Jersey. This scpplement reports the status of certain items that was issued October 1984. Supp'ement 4 was issued in May had not been resolved at the time of publication of the Safety 1985. Suppiment 5 was issued in July 1985 and Supplement 6 Evalustion Report.

issued in August 1985. This supplement 6 addresses further issues, principally the status of offsite emergency planning, that NUREG-1070: NRC POLICY ON FUTURE REACTOR require resolution prior to proceeding beyond the five percent DESIGNS. Decisions On Severe Accident issues in Nuclear power level. Power Plant Regulation.

• Offee of Nuclear Reactor Regulation, NUREG-1022 S02: LICENSEE EVENT Director. July 1985.147pp. 8508150036. 32197:342.

REPORT SYSTEM. Evaluation Of First Year Results And Recommenda- On April 13,1983, the U.S. Nuclear Regulatory Commission tions For improvements. HEBDON,F.J. AEOD, Director's Office. issued for public comment a " Proposed Policy Statement on September 1985. 84pp. 8509230665. 32701:299. Severe Accidents and Related Views on Nuclear Reactor Regu-This report desenbes an evaluation of an industry. wide lation (48 FR 16014). This report presents and discusses the sample of Licensee Event Reports (LERs) that was conducted on SMS hnal vWon of mat poHey staenwnt now enWed, to determine whether or not these LERs were prepared in ac- Policy Statement on Severe Rea:: tor Accidents Regarding cordance with the requirements set forth in 10 CFR 50.73 Futum Mgns aM hsung RanC n pmwdes an onw of which became effective on January 1,1984. The study was per comments meeW kom me p@c and me Msg Cunh formed at the Idaho National Engineering Laboratory (INEL) by on Reactor Sateguards and the statt response to these. In adds-EGSG, Inc. The evaluation (NUREG/CR-4178) indicated that al- tion to tho Policf Statement, the report discusses how the poh-though the overall quality of tne LERs was good, many LERs s of hs statenwnt mfam to omer E p ograms, incWng the Severe Accident Research Program; tne { implementation of failed to meet all of the requirements. This supplementary report presents the methodology that was used to evaluate the LERs' safety measums msumng fmm kssons leamed in me accht the conclusions reathed conceming pfoblem areas in the re at Three Mile island; safety goal development; the resolution of ports, and suggestions as to how the overall quality and com- Unresolved Safety issues and other Generic Safety issues; and pleteness of reports can be improved. PossW msions of @s w mgulate mquimments msubg from the Severe Accident Source Term Program. Also dis-NUREG-1030 DRFT: SEISMIC QUALIFICATION OF EQUIPMENT cussed are the main features of a generic decision strategy for IN OPERATING NUCLEAR POWER PLANTS. Unresolved Safety resofving Regulatory Questions and Technics.1 issues relaung to issue A-46. Draft Report For Comment. CHANG T.Y. Division of severe accidents; the development and regulatory use of new Safety Technology. August 1985. 215pp. 8509180444. safety information; the treatment of uncertainty in severe acci-32665:158. dent decision making; and the development and implementation The margin of safety provided in existing nuclear power plant of a Systems Reliability Program for both existing and future equipment to resist seismically induced loads and perform their plants to ensure that the realized level of safety is commensu-intended safety functions may vary considerably, because of rate with the safety analyses used in regulatory decisions.

Main Citations and Abstracts 5 NUREG 1080 VJ2: LONG-RANGE RESEARCH PLAN FY 1986- The nuclear plant aging research descnbed in this plan is in-FY 1990.

• Office of Nuclear Regulatory Research, Drector. tended to resolve issues related to the aging and service wear August 1985.157pp. 8509130047, 32621:067. of equipment and systems at commercial reactor facilities and The Long-Range Research Plan (LRRP) was prepared by tho their possible impact on plant safety. Emphasis has been Offee of Nuclear Regulatory Research (RES) to assist the NRC placed on identification and charactenzation of the mechanisms in coordinating its long-range research planning with the short- of material and component degradation dunng sevice and eval-range budget cycles. The LRRP lays out programmatic ap- uation of methods of inspection, surveillance, conditico monitor-proaches for rese6;ch to help resolve regulatory issues. The ing and maintenance as means of mitigating such effects. Spe-plan wi!! be updated annually. cifically, the goals of the program are as follows: (1) To ident.fy and charactenze aging and service wear effects which, if un-NUREG-1094: FINAL ENVIRONMENTAL STATEMENT RELATED checked, could cause degradation of structures, components.

TO THE OPERATION OF BEAVER VALLEY POWER and systems and thereby impair plant safety, (2) To identify STATION. UNIT 2. Docket No. 50412. (Duquesne Light Compa- methods of inspection, surveillance and monstonng, or of evalu-ny)

• Division of Licensing September 1985. 300pp- ating residual life of structures, components and systems, 8509300559. 32792:017, which will assure timely detection of significant aging effects The Final Environmental Statement related to the operation prior to loss of safety function, and (3) To evaluate the effec-of Beaver Valley Power Station, Unit 2 by Duquesne Light Com- tiveness of storage, maintenance, repair and replacement prac-pany, et al (Docket No. 50-412), located in Beaver Cour ty- tices in mitigating the rate and extent of degradation caused by Pennsytvania, has been prepared by the Office of Nuclear Re- aging and service wear.

actor Regulaton of the U.S. Nuclear Regulatory Commission.

This statement reports on the staff's review of the impact of op. NUREG-1149: TECHNICAL SPECIFICATIONS FOR LIMERICK eration of the plant. Also included are comments of state and GENERATING STATION, UNIT 1. Docket No. 50-352. (Philadel-federal govemments, local agencies and members of the public phia Electric Company) MARTIN,R.E. Office of Nuclear Reactor on the Draft Environmental Statement for this project and staff Regulation, Drector. June 1985. 500pp. 8508270346.

responses to these comments. The NRC staff has concluded. 32380:272.

based on weighing of environmental, technical and other fac- The Limerick Generating Station, Unit No.1, Techncal Speci-tors, that an operating Icense could be granted. fications were prepared by the U.S. Nuclear Regulatory Com-mission to set forth the limits, operating conditions and other re-NUREG-1122: KNOWLEDGES AND ABILITIES CATALOG FOR quirements applicable to a nuclear reactor facility as set forth in NUCLEAR POWER PLANT OPERATORS. Pressurized Water Section 50.36 of 10 CFR Part 50 for the protection of the health Reactors.

• Dvision of Human Factors Safety. July 1985- and safety of the public.

400pp. 8508090488. 32120:352.

This document catalogs roughly 5300 knowledges and aom- NUREG-1151: TECHNICAL SPECIFICATIONS FOR DIABLO ties of reactor operators and senior reactor operators. It results CANYON NUCLEAR POWER PLANT UNlTS I AND 2. Docket from a reanytsis of a much larger job-task analysis data base Nos. 50-275 And 50-323.(Pacific Gas And Electnc Company)

• compiled by the Institute of Nuclear Power Operations (INPO). Division of Licensing. August 1985. 465pp. 8509100521.

Knowledges and abilities are cataloged for 45 major power 32535:001.

plant systems and 38 emergency evolutions, grouped according The Diablo Canyon 1 and 2 Technical Specif~ cations were to 11 fundamental safety functions (e.g., reactnnty control and prepared by the U.S. Nuc: car Regulatory Commission to set reactor coolant system inventory control). With appropriate sam- forth the limits, operating cond;tions, ar d other requinments ap-p!ing from this catalog, operator licensing examinaticns having plicable to a nuclear reactor facility as set forth in Section 50.36 content validity can be developed. A structural sampling proce- of 10 CFR Part 50 for the protection of the health and safety of dure for this catalog is under development by the Nuclear Regu- the public.

latory Commission (NRC) and will be published as a companion NUREG-1154: LOSS OF MAIN AND AUXILIARY FEEDWATER document, Examiners Handbook for Developing Operator Li- EVENT AT THE DAVIS-BESSE PLANT ON JUNE 9,1985.

• censing Examinations (NUREG-1121). The examinations devel-oped by using the catalog and handbook will cover those topics Office of the Executive Drector for Operations. July 1985.

listed under Titie 10, Code of Federal Regulations, Part 55.

une 9 85 To E son Company's Davis-Besse Nu-NUREG-1141: TECHNICAL SPECIFICATIONS FOR FERMI-2 clear Power Plant, located in Ottawa County, Ohio, experienced FACILITY. Docket No. 50-341. (Detroit Edison Company)

• Divi- a partial loss of feedwater while the plant was operating at 90%

sion of Licensing. July 1985. 430pp. 8508070371. 32058:001. power. Following a reactor trip, a loss of all feedwater occurred.

The Fermi-2 facility Technical Specifications were prepare 1 The event involved a number of equipment malfuntiws and ex-by the U.S. Nuclear Regulatory Commission to set forth the tensive operator actions, including operator etions outside the limits, operating conoitions, and other requirements applicable control room. Several operator errors also occurred during the to a nuclear reactor facility as set forth in Section 50.36 of to event. This report documents the fin 6ngs of at, NRC Team sent CFR Part 50 for the protection cf the health and safety of the to Davis-Besse by the NRC Executive Director for Operations in public. conformance with the staff-proposed incident investigation Pro-gram.

NUREG-1142: TECHNICAL SPECIFICATIONS FOR RIVER BEND STATION. Docket No. 50-458. (Gulf States Utilities Company) NUREG-1155 V01: RESEARCH PROGRAM PLAN. Reactor Ves.

BENEDICT,R Division of Licensing. August 30,1985. 541pp. sels. VAGINS.M. Division of Engineering Technology. July 1985.

8509180507. 32663:337. 41pp. 8508150435. 32220:110.

The River Bend Station Technical Specifications were pre- This document presents a plan for research in Reactor Ves-pared by the U.S. Nuclear Regulatory Commission to set forth seis to be performed by the Materials Engineering Branch, the limits, operating conditions and other requirements applica- MEBR, Divisioa of Engineering Technology, (DET), Office of Nu-ble to a nuclear reactor facilty i as set forth in Section 50.36 of clear Regulatory Research, it is one of four plans describing the 13 CFR Part 50 for the protection of the health and safety of ongoing research in the corresponding areas of MEBR actnnty, the public. which are being published simultaneously in four volumes as follows: Vol 1 Reactor Vessels Vol. 2 Steam Generators, Vol. 3 NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) Piping and Vol. 4 Non-Destructive Examination. These plans PROGRAM PLAN. MORRIS,B.M.; VORAJ.P. Division of Engi- have been updated and are more detailed expansions of those neering Technology. July 1985. 48pp. 8508210443. 32337:277. originally published as part of the Long Range Research Plan

6 Main Citations and Abstracts for the Off'ce of Nuclear Regulatory Research in NUREG-1080 Dose reduction data and experience from 28 foreign and 10 Vol.1. U.S. nuclear power plants was examined to determine causes NUREG-1155 V02: RESEARCH PROGRAM PLAN. Steam Genera-for the wide vanations in occupatonal dose from country to country. Major topics discussed were: steam generator and re-

. tors. MUSCARA.J.; SERPAN.C.Z. Division of Engineenng Tech-fuehng maintenance problems; utility ar'd suppher ALARA pro-report desh b t. R re ac p ram related to grams; effectiveness of dose 4educten moditcatens; attitudes steam generators Mainly it (liscusses the program for evalua- and training, current and future dose-reducten research. While tion of a removed-from-utvice degraded steam generator. Also many parametes conWe to mences of m@a.

discussed are projects to evaluate the vibration and wear that doses between plants from different riatons, st is clear that could result from chemical cleaning and NDE tasks for inservice most U.S. plants have higher collective dose equivalent per re-inspection of steam generators. actor per magawatt. year then most other countries, even for p8vts of sim*!a'sde and age. Worldwide Finnish and Swedish NUREG 1155 V03: RESEARCH PROGRAM Pl.AN. Piping. plants, both PWR and BWR, have achieved the lowest valuet VAGINS.M.; STROSNIDER,J. Division of Engineering Technolo- Major factors which cordnbute to low doses include: 1) minimi-gy. Juty 1985.16pp. 8508160080. 32230.056. raton of cobalt in >Amary system components exposed to This document presents a plan for research in Piping to be water,2) careful plant design layout and component segrega-performed by the Matenals Engineenng Branch. MEBR, Division tion and shielding,3) plant standardizabcn,4) selecton of com-of Engineering Technology, (DET), Offee of Nuclear Regulatory ponents and systems for incraased rehabdity, 5) management Research. It is one of 'our plans desenbing the ongoing re- interest and commitment, 6) minimum number of workers and search in the corresponding areas of MEBR actmty, which are indepth worker training, 7) careful control of pnmary system being published simultaneously in four volumes as follows: Vol. oxygen and pH,8) good primary system water purity to minimize 1 Reactor Vessels, Vol. 2 Steam Generators, Vol. 3 Piping, and corrosion product formaton,9) use of special tools and robot-Vol. 4 Non-Destructive Examination. These plans have been up- ics,10) decontamination and passivation of pnmary systems dated and are more detailed expansions of those originally pub- and components, and 11) extent of backfitting and mandated in-lished as part of the Long Research Plan for the Offee of Nu- spections.

clear Regulatory Research in NUREG 1080 Vol.1.

NUREG-1155 V04: RESEARCH PROGRAM PLAN.Non-Destruc- NUREG/CR-1677 V02: PIPING BENCHMARK tive Examination. MUSCARA,J. Division of Engineering Technol- PROBLEMS, VOLUME 11 DYNAMIC ANALYSIS INDEPENDENT ogy. July SUPPORT MOTION RESPONSE SPECTRUM METHOD.

I 1985.

This report 27pp.the desenbes 8508160074.

NRC research32230:073'ogram pr in non-de-BEZLER P.; SUBUDHl,M.; HARTZMAN.M. Brookhaven National structive evaluation. Projects are desenbed for the development Laboratory. August 1985. 401pp. 8509160031. BNL-NUREG-s,nd evaluation of techniques for periodic inservice inspection of 512 3 625 p Pf reactor components and for the continuous online rnonitonng of reactors. The areas of study described are ultrasonic, eddy cur- verifying the adequacy of computer programs used for the dy-rent testing and acoustic emission. namic analysis and design of elastic piping systems by the inde-pendent support motion, respor se spectrum method. The dy-NUREG-1157: ENVIRONMENTAL ASSESSMENT FOR RENEW- namic loading is represented by distinct sets of support excita-AL OF SOURCE MATERIAL LICENSE NO. SUB-1010. Docket tion spectra assumed to be induced by non-uniform excitation in No. 40-8027. (Sequoyah Fuels Corporation)

• Division of Fuel three spatial directions. Complete input descriptions for each Cycle & Matenal Safety. August 1985. 406pp. 8509060254 problem are provided and the solutions include predicted natu-32504:010. ral frequencies, participation factors, nodal displacements and in response to an applicaton for renewal of Source Material element forces for independent support excitation and also for License SUB-1010 for the Sequoyah Fuels Corporation facility. uniform envelope spectrum excitation. Solutions to the associat-the NRC staff prepared this Environmental Assessment. The ed anchor point pseudo-static displacements are not included.

Environmental Assessment includes discussions of the need for the proposed renewal action, alternatives to the action, and the NUREG/CR-2000 V04 N6: LICENSEE EVENT REPORT (LER) environmentalimpacts of the proposed action and alternatives. COMPILATION.For Month Of June 1985.

• Oak Ridge National NUREG/CP-0063: PROCEEDINGS OF THE 1984 STATISTICAL PP' SYMPOSIUM ON NATIONAL ENERGY ISSUES. KINNISON,R.; 3 DOCTOR P. Battelle Memorial Institute, Pacific Northwest Lab- This monthly report contains Licensee Event Report (LER) oratories. July operational information that was processed into the LER data 1985. 244pp.

The 1984 Statistical Symposium 8507230193.

on National Ene 31754 241' rgy fle Issaes of the Nuclear Safety Information Center (NSIC) during the was the tenth in a series of annual symposia bringing together one month period identified on the cover of the document. The statisticians and other interested parties who are actively en- LFRs, from which this information is derived, are submitted to gaged in the pursuit of solving the nation a energy problems. Ins- tne Nuclear Regulatory Commission (NRC) by nuclear power tially the symposium was sponsored by U.S. Department of plant licensees in accordance with federal regu'avans. Proce.

Energy (DOE) and named the DOE dures for LER reporting for revisions to those events occamng symposium is organized by a ng steen, Statistical committee rnade Symposium.

up of Theto 1984 are described in NRC Regulatory Guide 1.18 and prior representatives from the national laboratories. The 1984 sympo- NUREG-1061, instructions for Preparation of Data Entry Sheets sium was hosted by Pacific Northwest Laboratory, and it was or- for Ucensee Event Reports. For those events occurring on and ganized around four special topical sessions: (1) Assessing and after January 1,1984, LERs are being submitted in accordance Assuring High Reliability, (2) Spatial Statistical, (3) Quantification w th the revised rule contained in Title 10 Part 50.73 'of the of Informed Opinion, and (4) Health Effects of Energy Technol- Code of Federal Regulations (10 CFR 50.73 - Licensee Event ogies. These were chosen research and data analysis. Seve al Raport System) which was published in the Federal Regulations contributed papers were also presented. (Vol. 48 No.144) on July 26,1983. NUREG-1022, Licensee Event Report System - Description of Systems and Guidelines NUREG/CP-0066: PROCEEDINGS OF AN INTERNATIONAL for Reporting, provides supporting guidance and information on WORKSHOP ON HISTORIC DOSE EXPERIENCE AND DOSE the revised LER rule. The LER summaries in this report are ar-REDUCTION (ALARA) AT NUCLEAR POWER PLANTS.MAY ranged alphabetically by facility name and then chronologically 29-JUNE 1,1984. HORAN,J.R.; BAUM.J.W.; DIONNE,B.J. by event date for each facility. Component, system, keyword, Brookhaver National Laboratory. September 1985. 279pp. and component vendor indexes follow the summaries. Vendors 8510040389. BNL-NUREG-51901. 32856:242. are those identified by the utility when the LER form is initiated;

Main Citations and Abstracts 7 the keywords for the component, system, and general keyword NUREG/CR-2815 V02 R1: PROBABILISTIC SAFETY ANALYSIS indexes are assigned by the computer using correlation tables PROCEDURES GUIDE. Sections 8-12. MCCANN,M.; REED.J.W.;

from the Sequence Coding and Search System. RUGER C.; et al. Brookhaven National Laboratory. August 1985.

P NUREG/CR-2000 V04 N7: LICENSEE EVENT REPORT (LER) See NURE / 2815 0 ,R abstra COMPILATION:For Month Of Jufy 1985.

• Oak Ridge National Laboratory. August 1985.129pp. 8509060195. ORNL/NSIC- NUREG/CR-3091 V06: REVIEW OF WASTE PACKAGE VERIFI-200.32505:141. CATION TESTS. Semiannual Report Covenng The Penod Octo-See NUREG/CR-2000,V04,N6 abstract ber 1984 - March 1985. SOO.P. Brookhaven National Laborato-ry. July 1985. 171pp. 8508220318. BNL-NUREG-51630.

NUREG/CR-2000 V04 N8: LICENSEE EVENT REPORT (LER) 32 COMPILATION For Month Of August 1985.

• OavRidge Nation-( at S rnber 1985.147pp. 8510030309. ORNL/ he ential of WAPPA, a secondmeneration < taste package system code, to meet the needs of the regu;atory community See NUREG/CR-2000,V04,N6 ab2 tract. are analyzed. The analysis includes an indepth review of WAPPA's individual process models and a review of WAPPA's NUREG/CR-2800 S03: GUIDELINES FOR NUCLEAR POWER operation. It is concluded that the code is of limrted use to the PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DE- NRC in the present form. Recommendations for future improve-VELOPMENT. ANDREWS,W.D.; BICKFORD.W.E.; ment, usage, and implementation of the code are given. This COUNTS,C.A.; et al. Battelle Memorial Institute, Pacific North- report also desenbes the results of a testing program undertak-west Laboratories. September 1985. 143pp. 8510030435. en to determine the chemical environment that will be present 32847:289. near a high-level waste packege emplaced in a basa't reposi-This Supplemental report is the fourtn in a senes that docu- tory. For this purpose, low carbon 1020 steel (a current BWIP ment and use methods developed by the Pacific Northwest Lab- reference container material), synthetic basalte groundwater oratory to calculate, for prioritization purposes, the risk, dose and a mixture of bentonite and basalt were expnsed, in an auto-and cost impacts of implementing resolutions to reactor safety clave, to expected conditions some period after repository seal-iscues. The initial report in this series was published by An- ing (150 degrees centigrade, aporoximately 10 4 MPa). Param-drews et al in 1983 as NUREG/CR 2800. This supplement con- eters measured include changes in gas pressure with time and sists of two parts describing separate research efforts: (1) an al- gas composition, variation in dissolved oxygen (DO), pH and temative human factors methodology approach and (2) a priori- certain ionic concentrations of water in the packing material tization of the NRC's Human Factors Program Plan. The alter- across an imposed thermal gradient, mineralogic alteration of native human factors methodology approach may be used in the basalt / bentonite mixture, and carbon steel corrosion behay-specific future cases in which the methods identfied in the ini- ior. A second testing program was also initiated to check the tial report (NUREG/CR-2800) may not adequately assess tha likelihood of stress corrosion cracking of austenitic stainless proper impact for resolution of new safety issues. The attema. steels and incoloy 825 which are being considered for ese as tive methdology included in this supplement is entitled Method- waste container materials in the tuff repository program.

ology for Estmating the Public Risk Reduction Affected t'y Human Factors improvement. The prioritization section of this NUREG/CR-3145 V03: GEOPHYSICAL INVESTIGATIONS OF report is entided Priontzation of the U.S. Nuclear Regulatory THE WESTERN OHIO-INDIANA REGION - ANNUAL Commission Human Factors Program Plan. REPORT.(October 1982 - September 1983, Volume 3).

POLLACK,H.N.; CHRISTENSEN D.; WELC,J. Michigan, Univ. of, NUREG/CR-2815 V01 R1: PROBABILISTIC SAFETY ANALYSIS Ann Arbor, ML May 1985. 51pp. 8507230043. 31753:299.

PROCEDURES GUIDE. Sections 1-7 And Appendices- Earthquake activity in the Westem Ohio - Indiana region has BARI.R.A.; BUSLIK,A.J.; CHO.N.Z.; et at Brookhaven National been monitored with a precision seismograph network consist-Laboratory. August 1985. 203pp. 8509110037. BNL-NUREG. ing of nine stations lecated in west-central Ohio and four sta-51559. 32561:010.

tions sited in Indiana. Delve local and near-regional earth-A procedures guide for the performance of probabilistic safety quakes have been recorded and located during this report assessment has been prepared for interim use in the Nuclear period, ranging in magnitude frem 0.3 to 4.0 m(big). An event Regulatory Commission programs. It will be revised as com- which occurred on January 14,1bS4, in Toledo, Ohio, and two ments are received, and as experience is gained from its use. events on July 28 and August 29,1924, near Terre Haute, Indi-The probabilistic safety assessment studies performed are in- ana, were felt. Only minor damage was reported from these tended to produce probabilistic predictive models that can be events. Of the twelve events, four occurred in the center of the used and extended by the utilities and by NRC to sharpen the Ohic array, three occurred near the city of Toledo, Ohio, four focus of inquiries into a range of issues affecting reactor safety. occurred in Indiana (including one on the Indiana-Illinois border),

This first volume of the guide desenbes the determination of the and one was located near Chicago, Illinois. Teleseismic P-wave probability (per year) of core damage resulting from accident ini- residuals have been updated and evaluated by back projection tiators intemal to the plant (i.e., intrinsic to plant operation) and to various depthe in the lower crust. The residuals are found to from loss of off-site electric power. The scope includes human corresp0M audJy to magnet,c anomalies in the lower crust of reliability analysis, a determination of the importance of various mio. It is tho.aht that these magnetic anomalies may represent t core damage accident sequences, and an explicit treatment and the remains of an ancient rift zone or perhaps they are the sig- '

display of uncertainties for key accident sequences. The second nature of the Grenville Front complex which may cross through volume deals with the treatment of the so-called extemal events this area.

including seismic disturbances, fires, floods, etc. Ultimately, the i:;uide will be eugmented to include the plant-specific analysis of NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT in-piant processes (i.e., containment performance). This guide SEQUENCE INFORMATION. CATHEY,N.G.; KRANTZ,E.A.;

provides the structure of a probabilistic safety study to be per- POLOSKI,J.P.; et al EGaG Idaho, Inc. (subs. of EG&G, Inc.),.

formed, and indicates what products of the study are valuable August 1985. 329pp. 8509130115. EGG-2259. 32605:254.

for regulatory decision making. For intemal events, methodology Information conceming the dominant accident sequences b treated in the guide only to the extent necessary to indicate from twelve published probabilistic risk assessments (PRA) is the range of methods which is acceptable; ample reference is cataloged in this report, which is published as a part of the Ac-given to attemative methodologies which may be utilized in the cident Sequence Evaluation Program (ASEP). The purpose of performance of the study. For extemal events, more explicit this report is to provide users of PRA information a single refer-guidance is given. ence document. The cataloged results include plant operation l

8 Main Citations and Abstracts information, core-melt and sequenco frequencies, and a de- the coldleg downcomer and lower plenum of a pressurtzed senption of each dominant accident sequence. The report pro- water reactor under conditions of interest to the issues of pres-vides a consistant set of insigh;s on the factors that drive the surized thermal shock. Several cold-leg assembl*5 are modeled dominant accident sequences. ASEP has reconstructed the and the downcomer arrangement can be altered to match PRA fault tree models at the system or train level of detail and vendor-specific configuratons. The facility can be operated to requantified the sequence lilcelihoods to provide the consistent model flow rates based on Froude number of the injected flow insights. This work provides the information for the cther ASEP in the cold log and with steady or transient inlet boundary condi-actrvities on accident liliephood essenment for the cpectiac tions. Exter.sNe ins'rumentation is provided to measure flow and near. term cporating plants. rates, temperatures and pressure at the facility boundanes and for detailed measurements of temperatures, velocity and heat NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE transfer data in the cold-leg and downcomer models. The test DOSIMETRY IMPROVEMENT PROGRAM. LWR Power Reactor data are monitored and recorded by a computer data acquisi-Surveillance Physics-Dosimetry Data Base Compendium. ten sysum mat is also used W posttest reducten and plob MCELROY,W.N. Hanford Engineenng Development Laboratory, tirg. The planned test matrix includes 75 teets with variations in August 1985.533pp.8509110278. HEDL-TME-85-3. 32563.020.

This NRC physics-dosimetry compendium (Sections 1.0 cold leg and downcomer geometries, loop and H,Pt flow rates, cold-leg Froude number and loop to HPI density difference, through 4.0) is a colla

• ion of information and data developW Test msds d M mpamd b a senes of MM RWs.

from available research and commercial fight water reactor vessel surveillance program (RVSP) documents and related sur- NUREG/CR-3426 V02: THERMAL AND FLUID MIXING IN 1/2 veillance capsule reports The Section 4.0 data represents the SCALE TEST FACILITY. Data Report. VALENZUELA.J.A.;

results of the HEDL least-squares FERRET-SAND 11 Code re- DOLAN,F.X. Creare, Inc. September 1385. 208pp. 8510020217.

evaluation of exposure uruts and va'ues for 47 PWR and BWR EPRI NP-3802. 32838:069.

surveiliam.e v apsules. Using a consistent set of auxdiary data This report presents data from an expenmental study of fluid and dosimetry-adjusted reactor physics results, the revised mixing in a 1/2-scale model of the cold-leg, downcomer, lower fluence values for E > 1 MeV averaged 25% higher than the plenum, pump simulator, and loop seal typical of a Westing-originalty reported values. The range of fluence values (new/ house Pressunzed Water Reactor. The tests were transient old) was from a low of 0.80 to a high of 2.38. These HEDL-de- cooldown tests in that they simulated an extreme condition of rived FERRET-SAND 11 exposure parameter values are being Small-Break Loss-of-Coolant Accident (SOLOCA) during which used for NRC-supported HEDL and other PWR and BWR trend cold High Pressure injection (HPI) fluid is injected into stagnant, cut te data development and testing studies, which support Re- hot primary fluid with complete loss of natural circulation in the vision 2 of Regulatory Guide 1.99. These trend curves are used loop. Extensive temperature, velocity, and heat transfer coeffi-

, by the utsties and by the NRC tu account for neutron radiation cient data are presented at two cold-leg Froude numbers: 0.052 g

damage in settng pressure / temperature limits, in analysing frac

• and 0.076. The 1/2-scale data are compared with earlier data tures, and in predicting neutron-induced changes in reactor PV from a 1/5-scale.geometricatty similar facility to assess scaling steel fracture toughness and embnttlement during the vessel's principles

• service life. The status of the development and application of

new advancements in LWR reactor surveillance programs is dis- NUREG/CR-3442

RADTWO A COMPUTER CODE FOR SIMU-cussed, such as cavity physics-dosimetry for improving the reli- LATING FAST-TRANSIENT, TWO-DIMENSIONAL TWO-LAYER ability of current and end-of-life (EOL) predictions on the metal- RADIONUCLIDE CONCENTRATIOr4 CONDITIONS IN lurgical conditions of pressure vessels and their support struc- LAKES. RESERVOIRS. RIVERS. ESTUARIES AND COASTAL tures. REGIONS. ERASLAN,A.H.; DIAMENT,H. Oak Ridge National NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGl. Laboratory. July 1985. 444pp. 8509180502. ORNL/TM-8869.

CAL t.CCIDENTS. METHODS, COSTS AND SCHEDULES FOR 32666:013.

DECONTAMINATION. TAWIL,J.J.; BOLD,F.C.; HARRER.B.J.; et RADTWO is a computer code for predicting the transient, 1 al. Battelle Memorial Institute, Pacific Northwest Laboratones. two-dimensional transport of radionuchdes in receg water August 1985. 379pp. 8509110274. PNL-4790. 32562:001, bodies. The model formulation considers two coupled, depth-This report documents a data base and a computer program averaged transport equations for the water layer and the bottom for conducting a decontamination analysis of a large, radiologi. sediment layer. The coupling condiSons incorporate bottom cally contaminated area. The data base, which was compiled deposition and resuspension effects. The computer code uses a largely through interviews with knowledgeable persons both in discrete-element method which offers variable size grid cells, the public and prhate sectors, consists of the costs, physcal accurate shoreune representation, and numerical accuracy. A inputs, rates and contaminant removal efficiencies of a large sample applicatx:,n is pav&d for the problem of a hypothetical number of decontamination procedures. The computer program accidental release of radionuclides to the coastal environment.

utilizes this data base along with information specific to the con. Results are presented as contours of constant radionuclide con-taminated site to provide detailed information that includes the centration in the water layer and the bottom sediment layer at least costly method for effectively decontaminating each sur, various times during the model simulation period.

face at the site, various types of property losses associated with the contamination, the time at which each subarea within the NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINA-TIONS ON SOLIDIFICATION, WASTE DISPOSAL AND ASSOCl-site should be decontaminated to minimize these property EXPOSURE. DAVIS,M.S.;

ATED OCCUPATIONAL losses, the quality of various types of labor and equipment nec. FiClULO,P.L; BOWERMAN.B.S.; et al. Brookhaven National essary to complete the decontamination, dose to radiation work- Laboratory. July 1985. 102pp. 8507250157. BNL-NUREG-ers, the costs for surveying and monitoring activities, and the 51699. 31794:123 disposal costs associated with radiological waste generated This report describes work Conducted by BNL on the degra-dunng cleanup. The program and data base are demonstrated dation of simulator chemical decontamination wastes by com-with a decontamination analysis of a hypothetical site.

bustion and acid digestion. Both acid digestion and combustion NUREG/CR-3426 V01: THERMAL AND FLUlO MIXING IN 1/2- are capable of effecting 90% destruction of the materials stud-SCALE TEST FACILITY. Facility And Test Design Report. led, as measured by tne conversion of carbon compounds in DOLAN.F.X.; VALENZUELA,J. A. Creare, Inc. September 1985. the waste to carbon dioxide. Work on the direct solidtf' cation of 130pp.8510020227. EPRI NP-3802. 32839:028. simulated decortamination wastes in cement and vinyl ester.

This report desenbes the test facility and program designed to styrene is reported also. Laboratory scale waste forms were measure fluid mixing and heat transfer in a 1/2. scale model of prepared using these binders. However, process control pro-4 l

i

Main Citations and Abstracts 9 grams and full scale solidification studies are necessary to con- removed due to intwiia and gravitational settling. Depos: tion of firm the acceptability of the wastes. the sma!!est particles was favored ty the USe of low face velXi-ties. A fractional efficiency curve was determined for each ma-NUREG/CR-3481 V02: NUCLEAR POWER PLANT PERSONNEL OUALIFICATIONS AND TRAINING: TAPS - The Task Analysis al at eam WW W CWane Values d N @@

Profiling System. JORGENSEN,C.C. Oak Ridge National Labo- facW, Mn W@aMn@sm W, wem caWaM ritory. July 1435. 246pp. 8508090705. ORNL/TM-0308/V2- QuaW facWs wm kss W wM rnatenals man W % kss at high velocities rather than low; and best for the single-use respe-Sr dscuses en autometed tesk enalyms profihng ," 8fel ^

system (TAPS) designed to provde a linking tool between the behaviors of nuclear pcwer plant operators in performing their NUREG/CR 3609: EVALUATION 'OF NEUTRON DOSIMETRY tasks and the measurement tools necessary to evaluate their in- TECHNIQUES FOR plant performance. TAPS assists in the identification of the WELL LOGGING OPERATIONS.

CUMMINGS,F.M.; HAGGARD.D.L; ENDRES,G.W. Battelle Me-entry-level skill, knowledge, ability and attitude (SKAA) require- morial Institute, Pacirc Northwest Laboratones. July 1985.51pp.

ments for the various tasks and rapidly associates them with 8508010304. PNL-4942. 31928:170.

measurement tots and human factors pnnciples. This report Neutron dose and energy spectral measurements from (241) desenbes the development of TAPS and presents its first dem- AmBe and a 14 MeV neutron generator were performed at a onstration. It begins with characteristes of skilled human per- well-logging laboratory. The measurement technique included formance and proceeds to postulate a cognitive model to for- the tissue equivalent proportional counter, multisphere, two mally desenbe these charactenstics. A method is denved for types of remmeters and five types of personnel neutron dosi-linking SKAA characteristics to measurement tests. The entire meters. Several source configurations were used to attempt to process is then automated in the form of a task analysis com- relate data to field situations. The results of the measurements puter program. The development of the prngram is detailed and indcated that the thermoluminescent albedo dosimeter was the a user guide with annotated code listings and supporting test in- most appropriate personnel neutron dosimeter, and that the fonnaten is provided.

most appropriate calibration source would be the source nor.

NUREG/CR-3485: PRA REVIEW MANUAL EL-BASSIONI.A.; mally employed in the field with the calibration source being CHO.N.Z.; HANAN,N.; et al. Brookhaven National Laboratory. used in the unmoderated configuration.

ember 1985. 2pp. 8509230653. BNL-NUREG-51710.

NUREG/CR-3613 V03 N1: EVALUATION OF WELDED AND This PRA Review Manual desenbes the approach for r.eview. REPAIR WELDED STAINLESS STEEL FOR LWR ing a Level 1 PRA, i.e., one which carries the accident analysis SERVICE. Semiannual Report For October 1984 Through March to the point of determination of core damage frequency, but ex. 1985. ATTERlDGE D.G.; CHARLOT LA.; BRUEMMER,S.M.; et cludes questions of containment integrity (but does include con- al. Battelle Memorial Institute, Pacific Northwest Laboratones.

tdnment failure induced core damage) and of offsite conse. September 1985. 59pp. 8510040360. PNL-4941. 32856:184.

quences. The manual will be revised as comments are received' Pacife Nortliwest Laboratory, under the sponsorship of the and as expenence is gained from its use. The procedure in Division of Engineering Technology of the U.S. Nuclear Regula-volves three parts: The first (Phase 1) is concemed with the tory Conmission, is conducting a program to determine a formal aspects of the PRA. Phase 1 surveys its apparent com- method for eva!uating welded and repair-welded stainless steel pleteness, scrutabihty, and determines to what extent the PRA (SS) piping for light-water reactor service. Vahdated models, can usefully be further examined. It also identifies salient and based on experimental data, are being developed to predict mi-distinctive features, of the study, methods, and reported results. crostructural development (e.g., the degree of sensitization) and The second part (Phase 2) rewews the analyses in a compre- the stress-corrosion cracking (SCC) resistance in the heat-a,f-hensive and thorough but qualitative way, which is designed to fected zone of the SS weldments. Stress-corrosion cracking is focus on unusual or unsupported features, and to lay the caused by a combination of a susceptible microstructure, an ag-groundwork far further, more detailed studies. The final stage gmssive environment, and tensile stress. Control of any of (Phase 3) addresses details of issues and concems raised in these three factors can eliminate SCC in most practcal situa-the earher phases, and involves detailed quantitative examina- tions. This program will measure and model the development of tion of selected areas to ensure the overall validity. The first a susceptible microstructure as it pertains to welded and repair-part of this manual, dealing with " internal" event PRAs, handles welded SS pipe. Empirical correlations between material micros.

these phases sequentially as a whole. In the second part, which tructure and SCC will be determined using constant extension treats " external" events, the phases are identified within each rate tests. The successful completion of these tasks will result event section, while Chapter 9 gives a sequential summary of in a method for assessing the effects of welding / repairing pa-the end results for each event. rameters on the SCC resistance of component-specife nuclear reactor welds / repairs. The present report describes the NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY progress of these studies during the first half of the 1985 fiscal PROiECTION:lli. SUBMICRON DARTICLE TESTS AND SUM- year.

MARY OF QUAllTY FACTORS. PRICE,J.M.; COOPER.D.W.;

YEE.C.S.; et al. Sandia National Laboratones. September 1985. NUREG/CR-3633 V01 S1: TRAC-BD1/ MOD 1:AN ADVANCED 05pp. 8509260257. SAND 83-7450. 32757:118. BEST ESTIMATE COMPUTER PROGRAM FOR BOILING The efficacy of readily available materials, such as cotton fab- WATER REACTOR TRANSIENT ANALYSIS. TAYLOR,D.D.;

rics, toweling a surgical mask, and a single-use respirator, for SHUMWAY,R.W.; SINGER,G L; et al. EG&G Idaho, Inc. (subs.

providing emergency respiratory protection was evaluated by of EG&G, Inc.),. September 1985.122pp. 8510040411, EGG-determing the filtration efficiency as a function of aerosol parti- 2294.32855:116.

cle size over the size range of 0.001 to 5.0 mm and as a func- The TRAC-BD1/ MODI computer program provides a best-es-tion of filtration face velocity. Filtration face velocity was set at timate analysis capability for the anal/ sis of the full range of 1.5, 5.0, and 15.0 cm/s. This report describes the equipment postulated accidents in Boiling Water Reactor (BWR) systems and procedures used to obtain efficiency measurements for par- and related experimer;tal facilities. The program is described in teles 0.5 mm in diameter and smaller, and summarizes the re- four volumes: Volume 1, Code Description; Volume 2. User's suits of all three phases of this research. Particles with diame- Guide; Volume 3, Code Structure and Programming information; ters from 0.10 to 0.50 mm proved to be the most difficult sizes and Volume 4, Developmental Assessment. Volume 1 desenbes of particles to remove. Particles smaller than 0.10 mm were re- the thermal-hydraulic models, numerical mehtods, and compo-moved due to diffusion while particles larger than 0.50 mm were nent models available. Volume 2 desenbes the input and output

10 Main Citations and Abstracts of the TRAC-BD1/ MOD 1 code and provides guidelines for use NUREG/CR-3660 V04: PROBABILITY OF PIPE FAILURE IN RE-of the code modeling of BWR systems. Volume 3 is designed ACTOR COOLANT LOOPS OF WESTINGHOUSE PWH for the programmer or model developer who needs to imple- PLANTS. Volume 4. Pipe Failure Induced By Crack Growth in ment or modify the TRAC-BD1/ MOD 1 program. Volume 4 dis- West Coast Plants. CHINN.D.J.; HOLMAN.G.S.: LO.T.Y.: et al.

cusses the results of the development assessment calculations Lawrence Livermore National Laboratory. July 1985. 58pp.

performed with TRAC-BD1/ MOD 1. 8508020207. UCID-19988 V04. 31%2:143.

Tua U.S. Nuclear Regulatory Commission contracted w'th the NUREG/CR-3633 V04: TRAC-BD1/ MOD 1:AN ADVANCED BEST Lawrence Livermore Natonal Laboratory to conduct a study to ESTIMATE COMPUTER PROGRAM FOR BOLLING WATER determine if the probability of occurrence of a doub!e-ended REACTOR TRANSIENT ANALYSIS. Volume 4: Developmental guillotine break in primary coolant piping warrants the current Assessment. SHUMWAY,R.W. EG&G Idaho, Inc. (subs. of design requirements that safeguard against the effects of such EG&G,Inc.) September 1985.101pp. 8510040407. EGG-2294. a break. This report assesses the reactor-coolant-loop piping 32858:314. system of west coast Westinghouse plants. The results indicate This volume of the TRAC-BD1/ MOD 1 manual discusses the that directly induced DEGB is an unlikely event in the west results of developmental assessment calculations performed coast Westinghouse plants. For the Trojan plant, leak is far mainly with a preliminary version of TRAC-BD1/ MOD 1 (V21), more likely than a direct DEGB. Further, earthquakes have very and some selected cases performed with the official version of little effect on. the probabilities of leak and direct DEGB. At the TRAC-BD1/ MODI (V22), which differed from the preliminary Diablo Canyon plant, the increase in postulated seismic levels version due to few small corrections. Twenty one test cases due to reevaluation of the site to account for the Hosgri Fauft have been performed, ranging from simple single-effect flow has caused directly induced DEGB failure probability to be de-tests up to a full BWR/6 system calculation. The four groups of pendent on earthquake occurrences. The resulting direct DEGB tests are: separate effects hydraulic tests, steady state heat failure probability is still much lower than the indirect DEGB fail-transfer tests, transient heat transfer tests, and integral system ure probabdity for Diablo Canyon.

effects tests. The separate effects test cases were each NUREG/CR-3706: TRAC ANALYSES OF SEVERE OVERCOOL-chosen to exercise a specific hydraulic or heat transfer model in ING TRANSIENTS FOR THE OCONEE 1 PWR. IRELAND,J.R.

the code, while the integral system effects tests were chosen to Los Alamos Scientific Laboratory. August 1985. 261pp.

exercise the code as a whole. The TRAC code version initially 8509160026. LA-10055-MS. 32625:059.

used was TA021 for all cases. Code errors were evident in This report desenbes the results of several Transient Reactor some of the heat transfer runs. Analysis Code (TRAC)-PF1 calculations of overcooling tran-sients in a Babcock & Mcox bwersh,p, pmssM water I

NUREG/CR-3638: HYDROGEN-STEAM JET-FLAME FAOfLITY mactw (Omneed). De purpose of this study is to provide de-AND EXPERIMENTS. SHEPHERD.J.E. Sandia National Labora- tailed input on thermal-hydrauhc data to Oak Ridge National tories. July 1985. 138pp. 8508010764. SAND 84-0060. Laboratory for pressurized thermal-shock analyses. The tran-31925:104 sient calculations performed were plant specific in that details of As part of NRC-sponsored research on light-water reactor the primary system, the secondary system, and the plant.inte-safety, the high-temperature combustion of stJam-hydrogen jets graW control systwn of Oconeed wm Wed in N N in an air atmosphere is being investigated at Sandia. This re- m e s d N cakulahs Mcam mat N W search is onented at understanding the generic issues involved e as fa e ansd was N most swm in Ms in accident-generated jets and the specific probisms of usirm of ms ma% W W ppsatums in N h deliberate flaring from high-point vents to eliminate hydr 09'r] comer region of the vessel. The power-operated relief valve from the pnmary system. In this report we give some back. loss-of-coolant accident transient was the least severe in terms ground on diffusion-flame combustion, desenbe the expenmen- of downcomer liquid temperatuies because of vent-valve fluid tal facility constructed at Sandia to study high-temperature' mixing and near-saturated conditions in the pnmary system. It is steam-hydrogen jets and discuss our results. recommended that future calculations consider a wider range of pwatm achs to cww me spra of wmMg kanssnt NUREG/CR-3660 V01: PROBABILITY OF PIPE FAILURE IN THE sequences more completely.

REACTOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 1: Summary Report. HOLMAN,G.S.; CHOU.C.K. NUREG/CR-3710: LABORATORY STUDIES OF A BREACHED Lawrence Livermore National Laboratory, July 1985. 103pp. NUCLEAR WASTE REPOSITORY IN BASALT. SEITZ,M.G.;

8508010773. UCID-19988. 31923:247. BOWERS D.L; GERDING,T.J.: et al. Argonne National Labora-As part of its reevaluation of the double-ended guillotine tory. August 1985.150pp. 850E260304. ANL-84-16. 32368:071.

break (DEGB) of reactor coolant loop piping as a design basis Experiments are desenbed that combine backfill, radioactive event for nuclear power plants, the U.S. Nuclear Regulatory waste, and basalt rock in a single flowing groundwater stream in .

Commission (NRC) contracted with the Lawrence Livermore Na- a manner analogous to a hydraulic breach of a waste reposi-tional Laboratory (LLNL) to estimate the probability of occur- tory. The expenments were used to study chemical interactions rence of a DEGB, and to assess the effect that earthquakes that woul<f occur if repository components were breached by have on DEGB probability. This report describes a probabilistic fiowing water. The result of most significance to issues of re-evaluation of reactor coolant loop piping in PWR plants having pository performance was that uranium, neptunium, and plutoni-nuclear steam supply systems d6 signed by Westinghouse. Two um w(re found to move more rapidly through repository compo-causes of pipe break were considered: pipe fractJre due to the nents that were altered to represent aging then through fresh growth of cracks at welded joints (" direct" DEGB), and pipe materials. In contrast, cesium moved slower through altereo re-rupture indirectly caused by failure of component supports due pository raateriaIS, as had been deduced from previous work to an earthquake (" indirect" DEGB). The probabihty of direct using batch adsorption tests. Two other parameters studied ex-DEGB was estimated using a probabilistic fracture mechanics penmentally, the metal alloy used in the apparatus and an ioniz.

model. The probability of indirect DEGB was estimated by esti- ing radiation field imposed on the experimental apparatus, had mating support fragility and then convolving fraginty and seismic ittle or no measurable effect on radioactive elemerit transport hazard. The results of this study indicate that the probability of by flowing water. Inasmuch as the alteration of the repository a DEGB from either cause is very low for reactor coolant loop materials aging b an actual repository, we conclude that piping in these plants, and that NRC should therefore consider changes with age will detrimentally affect the ability of a reposi-eliminating DEGB as a design basis event in favor of more real- tory to isolate uranium, neptunium, and plutonium. Because istic cnteria. these elements have long-lived radioactive isotopes in nuclear

l l

Main Citations and Abstracts 11 l

l waste, the degradation with time is a major issue regarding the safety issues, (2) understand the progression of risk.ssgnificant performance of a nuclear waste repository in basalt, accident sequences and (3) conduct safety assessments. The collectue NRC-sponsored effort at Sandia National Laboratones NUREG/CR-3736: FIELD AND THEORETICAL INVEST!GA7lONS i OF FRACTURED CRYSTALLINE ROCK NEAR is directed at enha7Cing the technology base supporting licens-  ;

ORACLE ARIZONA. JONES,J.W.; SIMPSON,E.S.; ing decisions.

NEUMAN,S.P.; et al. Arizona, Univ. of, Tucson, AZ. August NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACILl-1985.115pp. 8508290526. 32410:254. TIES. RO",J.A.; DEWALL,K.G.; STEELE,R.; et al. EG&G A combination of geophysical and hydraulic testing has oeen Idaho. ir (sutes. of EGaG, Inc ),. Juty 1985. 51pp conducted in grarmte near Oracle Anzona. The purpose of the 85072501 -GG-2317. 31786:061.

work is to determine relationships, if any, ar1ong (1) fracture This reg ' resents the results of the survey of Aged Nuclear distribution, (2) geophysical propertes, and (3) hydraulic prcper- Power Pla . Facihties conducted for the USNRC Office of Nu-ties of fractured rock of low hydraulic conductivity. To date, clear Regulatory Pesearch. The results of this report recom-eight vertcal borings spaced 20 to 50 feet apart, ranging from mend methods to help formulate comprehensive research pro-250 to 300 feet in depth, have been drilled. The data obtained gram that will systematically identify aging and service wear ef-from neutron, gamma, acoustic-velocity, electncal-resistivity, and fects which are likely to affect plant safety. The survey centered acoustic. televiewer logs, with the results of over 100 single- on sa'ety related plant systems with regard to component fail-hole, straddle packer injection tests make possible a detailed ures from operating histories. The age related failure information description of the fracture system. Geophysical logs readily gathered from the plant histories was ana!yzed for reoccurnng detect fractures and are sensstwe to subtle lithologic variations failure patterns. Emphasis was on identfication of specific of the grarwte. Orientation and distnbution of individual fractures equipment with high failure rates and of failure mechanism rela-were determined from the interpretation of t7e acoustic-tele

• tionships. The data would not support spec:fic oquipment identi-viewer data, and from the analysis of core ottained from one fication. It did imply a direct relationship between failure and fail-borehole. Fracture densities over the 13-foct long straddle- ure mechanism. 70% of the failures reported were due to four packer test intervals did not correlate with measured hydraulic failure mechanisms. In addition there appeared to be a strong conductivity measurements. A strong correlaton between the correlation between cause of failure and the system in which neutron-log response and measured hydraulic conductivity does the component operates. This is venfied by detailed study of exist; it was used to supplant conductivity measurements. The several plant systems and corroborated by personnel inter-geostatistical technique of kriging provided a three-dimensional views. This survey indicates identfication and ehmination of map ci hydraulic conductivity that can be compared with sub- System level cause of component failure is a viable approach to surface interpretations of the geophysical logs. prevent and mitigate major reported aging effects.

NUREG/CR-3816 V03: REACTOR SAFETY RESEARCH.Ouarterty Report. July-September 19M

• Sandia National Laboratories. NUREG/CR-3851 V04: EVALUATION OF RADIONUCLIDE GEO-July 1985.190pp. 8507250114. SAND 84-1072. 31787.147. CHEMICAL INFORMATION DEVELOPED BY DOE HIGH-Sandia National Laboratones is conducting, under USNRC's LEVEL NUCLEAR WASTE REPOSITORY SITE sponsorship, phenomenological research related to the safety PROJECTS. Annual Progress Report For October 1983-Septem-of commercial nucicar power reactors. The overall objectives of ber 1984. KELMERS A.D.; KESSLER.J.H.; SEELEY,F.G.; et al.

this work is to provide NRC a comprehensive data base essen- Oak Ridge National Laboratory. September 1985. 66pp.

tial to (1) defining key safety issues, (2) understanding risk-sig- 8509260079. ORNL/TM-9191. 32760d 66.

nificant accident sequences, (3) developing and verifying Geochemical information relevant to the retention of radionu-models used in safety assessments, and (4) assunng the public clides by candidate high-level nuclear waste geologic repositor-that power reactor systems will not be licensed and placed in ies being characterized by Department of Energy (DOE) commercial service in the United States without appropriate rojects is being evaluated by Oak Ridge National Laboratory consideration being given to their effects on health and saf6ty. (ORNL) for the Nuclear Regulatory Commission (NRC). Empha-Together with other programs, the Sandia effcrt is directed at s s has been given to the experimental evaluation of key radion-assuring the soundness of the , technology base upon which h. uclides relevant to the Hanford Site being characterized by the censing decisions are tr.ade. This report describes progress in a Basalt Waste Isolation Project (BWIP). In work by the BWIP, hy-number of activities dealing with current safety issues relevant drazine was added to groundwater to simulate the reducing to both light water and breeder reactors. The work includes a redox condition expected in the repository. Such laboratory broad range of experiments to simulate accidental conditions to methodology may not adequately model in situ repository geo-provide the required data base to understand important acci- chemical conditions. We have beem employing anoxic redox dent sequences and to serve as a basis for development and conditions to a!!ow the basalt to establish the effective redox complex computer simuladon models and condition in batch contact sorption experiments. Sorption of venfication codes used in acci of the, dent analysis ad licensing ev,ews Such Np(V) aor Tc(Vil) by basalt from synthetic groundwater under program must include the development of aralytical modeis anoxic redox conditions may involve chemisorpticn reduction re-venfied by expenment, which can be used to predict reactor' actions on the basalt surface. Our sorption ratio for neptunium and sa ty tem performance under a broad variety of abnor- under oxic redox conditions does not compare favorably with the value published by the BWIP. The published solubility of technetium under the reducing redox conditions expected by NUREG/CR-3816 V04: REACTOR SAFETY RESEARCH.Ouarterfy 8WIP at the repository probably is based on calculations involv-Report, October-December 1984.

• Sandia NaSonal Laborato- ing inadequate thermodynamic data. Under oxic redox condi-ries. September 1985. 242pp. 8510040565. SAND 84-1072. tions, our uranium sorption ratio was much lower than values re-32855:238. ported by the BWIP. A mineralogical and chemical characteriza-Sandia National Laboratories is conducting, under the tion was completed for the three basalt samples used in our USNRC's sponsorship, phenomenological research related to work. Significant differences were seen in both the quantity and the safety of commercial nuclear power reactors. The research composition of the mesostasis. A potential deficiency in the in-includes expetiments to simulate the phenomenology of the ac- formation published by the BWIP is the absence of hthological cident conditions and the development of analytical models, information as well as mineralogical and chemical characteriza-verified by experiment, which can be used to predict reactor tion for the basalt samples. Our geochemical modeling work and safety systems performance and behavior under abnormal suggested that code-to-code evaluation for geochemical calcu-conditions. The objective of this work is to provide NRC requi- lations may be less important than a detailed evaluation of the site data bases and analytical methods to (1) identify and define data bases.

J l

12 Main Citations and Abstracts i

NUREG/CR-3876: PROBABILITY BASED LOAD COMBINATION NUREG/CR-3901: DOCUMENTATION AND USER'S GUIDE.GS2 CRITERIA FOR DESIGN OF CONCRETE CONTAINMENT & GS3 - VARIABLY SATURATED FLOW AND MASS TRANS-STRUCTURES. HWANG,H.; KAGAMI,S.; REICH,M.; et at PORT MODELS. DAVIS,L.A.; SEGOL,G. Water, Weste & Land, Brookhaven National Laboratory. August 1985. 99pp. Inc. June 1985. 305pp 8507250143. WWL/TM-17912.

8509110003. BNL-NUREG-51795. 32561:213, 31789:002.

This report describes a research effort for the development of This report presents documentation and user's manual for the probability-based load combination enteria for design of programs GS2 (two-dimensional version) and GS3 (three-dimen.

concrete containment structures. the proposed entena are en a sbnal wsim). Mathemata al equatbns and physka! principles lead and resistance factor design (LRFD) format. In order to utilized to develop the code are presented in Section 2. The nu-test the performance objectves of the proposed entena, four merica! approach used (Galerkin Finite Elements) is presented representative structurcs are selected using a Latin hypercube in Section 3. Secten 4 presets an overview of how problems sampling technique. Next, the reliability analysis method devel- should be set up to property use the code while detaied input ins *uctions are presented in Section 5. Output produend by the oped by Brookhaven National Laboratory is employed to assess code is discussed in Section 6. Three example problems, in-the reliability of these representative containments. Further-ciuding sample input data sets and output data, are presented more, an obisctive function is defined and a minimization tech-nique is developed to find the optimum load factors. The load in Secton I Program informabon is prom in W 8. A listing of important program variables along with complete pro-factors for accident pressure due to the design basis acMent gram listings are presented in the Appendices. This report was and safe shutdown earthquake are denved for three target limit prepared as part of a project in which NRC staff was presented state probabilities. Other load factors are also discussed on the a Waining use m hp to p% m m cower yogran basis of prior expenence with probability-based design cntena for ordinary building construction. The proposed load combina- Warns GS2,aN GS3 can M um to anaW h M mass transpM in msaturatM, pamah sauaM, or fuW sab tions are based on the best available data to date pertaining to ram Ha Wons. R h anMpaM mat N E d m h loads and resistances. If in the future the data base changes the developed methodol09Y can readily be utlized to update the c n am F W ha Msm Wwab uating attemative sites and designs for waste disposal, and for load factors resulting from these changes. companng their results with results from other methods of solu-NUREG/CR-3885 YO4: HIGH-TEMPERATURE GAS COOLED RE-ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED EVALUATION. Quarterly Progress Report October 1-December FROM TESTING THE ZB-1 INTERMEDIATE SCALE PRES-31,1984. BALLS.J.: CLEVELAND.J.C.; HARRINGTON,R.M.; et SURE VESSEL HUTTON P.H.; KURTZ,R.J.; PAPPAS,R.A.; et al. Oak Ridge National Laboratory. August 1985. 24pp. at Battelle Memonal Insttute, Pacific Northwest Laboratories.

j 8508210424. ORNL/TM-9267/V4. 32334:209. September 1985. 240pp. 8509300233. PNL-5184. 32811:317.

4 Mode 8ing and code development work on the modular High- Acoustic emission (AE) monitoring of flaw growth in an inter-Temperature Gas-Cooled Reactor (HTGR) continued with the mediate scale vessel during cyclic loading at 65 degrees centi-

,! grade and 288' degrees centigrade is described in this report.

development and testing of a thermal model of the upper reflec-l tor. The longer-term heatup accident scenario in which cavity The report deals with background, methodology, and results.

wall cooling is lost was also modeled. Sensitivity studies were The work discussed is of major significance in a program sup-run for vanations in soil conductivity and decay heat generation ported by NRC to develop and demonstrate application of AE rate. Fission-product (FP) release and transport experiments monitoring for continuous surveillance of reactor pressure

) boundaries to detect and evaluate growing flaws. Several areas were completed and initiated for several additional elements.

Progress was made in establishing an FP redistnbution capabil- of technical concern are addressed. Results support the feasi-j ity in the ORECA code. bility of effective continuous mon 1. onng.

l NUREG/CR-3900 V04: LONG-TERM PERFORMANCE OF MATE- NUREG/CR-3935: THERMAL-HYDRAULIC ANALYSES OF l RIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual OVERCOOLING SEQUENCES FOR THE H.B. RO ..NSON Report. April 1984 - Apri 1985. STAHL D.; MILLER,N.E. Battelle UN!T 2 PRESSURIZED THERMAL SHOCK STUDY.

FLETCHER,C.D.; DAVIS.C.B.; OGDEN.D.M. EG8G Idaho, Inc.

Memonal Institute, Columbus Laboratones. July 1985. 235pp.

(subs. of EG&G, Inc.). July 1985. E91pp. 8507250171. EGG-8508150047. BMI-2127. 32197:107.

2335.31785:130.

Waste form experimentation has focused on borosilicate Oak Ridge National Laboratory (ORNL), as a part of the Nu-glass, using the reference composition MCC 76-68. An experi- clear Regulatory Commission's (NRC's) pressurized thermal i ment investigated the influence of continuous contact between shock (PTS) integration study for the resolution of Unresolved the glass specimen and the leachate on the results of corrosion Safety issue A49, identified overcooling sequences of interest studies. It was found that precipitates formed during cooling can to the H.B. Robinson PTS study. For each sequence, reactor affect the results. Other experiments evaluated the influence of vesS# down-comer fiuid pressure and temperature histories crystallization on glass waste-form performance and the influ- were required for the two-hour period following the initiating ence of organic acid on the waste-form and radionuclide mobili- event. Analyses previously performed at the Idaho Natbn? En-ty in groundwater. Models were used to analyze glass dissolu- gineering Laboratory (INEL) fully investigated a lireited number tion, including the repreciptation of dissolved glass spedes. of the sequences using a detailed RELAPS model of the H.B.

< The effect of groundwater species on the electrochemistry of Robinson, Unit 2 (HBR-2) plant. However, a full investigation of steels is being analyzed to evaluate susceptability to pitting and all sequences using the detailed model was not economically stress-corrosion cracking. Species identified as potential crack- practical. New methods were required to generate results for ing agents are being investigated by slow strain rate experi- the remaining sequences. Pressure and temperature hstmes ments. Hydrogen embrittlemont studies of steel showed an- for these remaining sequences were generated at the INEL nealed cast steel to be more sensitive to embrittlement. Realis- through a process combining: (a) partial-length calculations tic general and pitting corrosion models are being developed, using the detailed RELAPS model, (b) full-length calculations based on known principles of rnass transport and radiolytic pro- using a simplified RELAPS model, and (c) hand calculations.

duction. Mechanical and water. chemistry-related stresses which This report documents both the methods used in this process influence mechanical degradation were evaluated. Groundwater- and the results. The sequences investigated contain significant radiofysis and water-chemistry studies are continuing as part of conservatisms conceming equipment failures, operator actions, the integrated system performance task. or both. Consequently, care should be taken in applying the re-

\

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

Main Citations and Abstracts 13 suits presented herein without an understanding of the conserv- Nuclear containments which will not leak significantly, that is, atisms and assumptions. The results of the thermal-hydraulic beyond technical specifications, dunng a design accident may analyses presented here, along with additional analyses of mul- leak severely during a severe accident when the pressure in-tidimensional and fracture mechanics effects, will be utilized by creases beyond the design level Small leaks which are visual-ORNL to assist the NRC in resolving the PTS unresolved safety ized as occumng at local details may occur before complete issue. vessel failure. Buckling of the hatch door, large deformations and ovaling of the hatch sleeve are potential causes of mis-NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPER- match at the sealing surface which can result in a leakage path.

ATIONAL TRANSIENT (OPTRAN) TESTS 1-1 AND 12 IN THE As a typical example of steel containments the Sequoyah ecuip-POWER BURST FACILITY. MCCARDELL,R.K.; PLOGER,S.A.; ment hatch was selected. If penetrations effects are neglected, l

MCCORMICK,R.D.; et at EGaG Idaho, Inc. (subs. of EG&G, gross yielding of the 1/2-inch shell plate near the springfine of Inc.) September 1985. 78pp. 8510030433. EGG-2297. the Sequoyah containment will occur at an internal pressuro of 32911:221. between 50 and 60 psi. The results of a finite element analys:s Operational transients occur occasionally in light water reac- showed that a maximum of 0.9 irch of 85 to 90 psi, far aoove tors when minor malfunctions of certain system components gross yielding of shell. Although buckling increased the relative affect the reactor core. This report presents the results of the seal motions, they remained sufficiently small to prevent leak-operational transient Test OPTRAN 1-1 and OPTRAN 1-2, in- age. The Sequoyah equipment hatch should not leak before cluding a comparison of the data with postlest calculat;ons and strains of several percent develop in the 1/2-inch contairement the postirradiation examination results. The OPTRAN11 tests shell plate near the springline, which occurs between 50 and 60 simulated operational transients with reactor scram. Four pro- psi. In the unlikely event of hatch buckling, postbuckling defor-gressrvely higher and broader power transients at a constant mations would not introduce leakage.

coolant flow rate were performed. The first transient simulated a BWR-5 turbine tnp without steam bypass, with fuel rods operat- NUREG/CR-3980 V04: LIGHT-WATER-REACTOR SAFETY FUEL ing near BWR.6 core average rod powers. The second transient SYSTEMS RESEARCH PROGRAMS. Quarterfy Progress simulated a generator load rejection w%out steam bypass, with Report, October-December 1984. CHUNG H.M.; REST,J. Ar-fuel rods operating near core average powers. The last two gonne National Laboratory. September 1985. 63pp.

transients were performed at higher core average peak rod 8510040302. ANL-84-61 V04. 32856.118.

powers than safety analyses predict to be possible in commor- This progress report summanzes the Argonne Natsonal Labo.

cial reactors to define failure threshold margins. Test OPTRAN ratory work performed during October, November, and Decem-1-2 was performed to evaluate the probability and extent of fuel ber 1984 on water reactor safety problems related to fuel and rod damage for the most severe BWR anticipated transient fuel cladding materials. The research and development areas without scram (ATWS) that results in boiling transition, a main covered are Transient Fuel Response and Fission Product Re-steam line isolation valve closure transient without scram. Two lease and Clad Properties for Code Venfication.

sets of two fuel rods were tested. In each set, an unerradiated fuel rod was used to heat the coolant to typical BWR conditions NUREG/CR-4006: CLOSEOUT OF IE BULLETIN 81-01. SURVEIL-for each prewously irradiated fue! rod. Fo!!cwing an extensive LANCE OF MECHANICAL SNUBBERS. FOLEY,W.J.;

fuel conditioning period of operation, a single power transient DEAN,R.S.; HENNICK.A. Parameter, Inc. August 1985. 82pp.

was performed that simulated the power history and coolant 8508260307, IE 145. 32368:219.

conditions calculated for a main steam line isolation valve clo. In the period from August 1974 to May 1980, failures of me-chanical snubbers were desenbed in event reports issued for sure ATWS.

nine facilities and in a NRC/IE study of the DOE Fast Flux Test NUREG/CR-3949 V02: EDDY-CURRENT INSPECTION FOR Facility. In most failures, the snubbers were frozen and would STEAM GENERATOR TUBING PROGRAM. Annual Progress not permit free piping motions during thermal transients. In Report For Period Ending December 31,1984. DODD.C.V.; some cases, the failed snubbers no longer provided seismic DEEDS,W.E.; SMITH J.H.; et al. Oak Ridge National Laboratory. shock restraint. Because of concern about the reported failures 4

August 1985. 16pp. 8509110016. ORNL/TM-9339/V2. of mechanical snubbers, standard techrucal specification revi-32564:288. sions for snubber surveillance were issued by NRC/DL on No.

Eddy-current inspection is the most suitable method for rapid vember 20, 1980. IE Bulletin 81-01 was issued January 27, boreside evaluation of steam generator tubing. However, small 1981 to require examination and testing of mechanical snubbers flaws can be masked by the effects of harmless variables, such in safety-related systems at licensed facilities and at selected as tube supports, To identify the critical properties accurately facilities under construction. Evaluation of utility responses and and reliably in the presence of extraneous signals caused by NRC/IE inspection reports indicates that the bulletin can be variations of unimportant properties, sufficient informatior is closed cut per specific criteria for 73 (95%) of the 77 fac!!ities needed to identify harmful variations and reject harmless or:es. to which it was issued for action. Followup items are proposed For this reason we have been developing instrumentation capa- for use by NRC/IE to ensure satisfactory completion of correc-ble of measuring both the amplitude and phase of the eddy-cur- tive actroa t! me remrining four (4) facilities.

rent signal at several different frequencies, as well as computer equipment capable of processing the data quickly anti reliably. EUREG/CR#37: DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST Hi-5. OSBORNE.M.F.;

Our probes and test conditions are also computer-optimized.

The most recent probe design embodies an array of arr:all flat COLUNS.J.L; LORENZ,R.A.; et al. Oak Ridge National Labora-

" pancake" coils and improves the detect.on of small flaws and tory. Juty 1985. 75pp P508090712. ORNL/TM 9437.

the rejection of tube support signals. We have also experimen. 32104:055.

tally verified the accuracy of our computer programs for calcu.

The fifth in a series of high-temperature fission product re-lease tests was conducted for 20 min at 1700 degrees centi-lating the signals produced by defects in tubing and are adapt.

ing our new IBM System 9000 computer to take and process grade in flowing steam. The test specimen, a 15.2-cm-long sec-tion of a fuel rod which had been irradiated to a bumup of 38.3 the larger amounts of data required by additional variables, su$

as copper coating and intergranular attack. mwd /kg, was heated in an induction 'urnace under simulated LWR accident conditions in a hot cell. Posttest inspection NUREG/CR-3952: SEQUOYAH EOUIPMENT HATCH _ SEAL showed severe oxidation and fragmentation of the fuel speci-LEAKAGE. GREIMANN,L; FANOUS,F.; BLUHM.D. Ames Labo- men, but no cladding melting was apparent. Analyses of test ratory, Energy & Mineral Resources Research Institute. July components showed total releases from the fuel of 19.9% for 1985. 49pp. 8508090579. IS-4862. 32105:296. (85)Kr. 22.4% for (129)l,18.0% for (110m)Ag, and 20.3% for l

14 Main Citations and Abstracts (137)Cs. A smaller fraction of the (125)Sb (0.326%) was re- signal validation techniques to core exit thermocouples and leased from the fuel, and 99% of the (110m)Ag and (125)Sb other rneasurement systems are made.

was retained in the fumace. Posttest analysis of the fuel speci-men indicated a (134)Cs relcase of 24.5%, which is reasonably NUREG/CR4081: ABSORPTION OF GASEOUS LODINE BY good agreement with the (137)cs data. These releases were WATER DROPLETS. ALBERT,M.F. Oak Ridge Na 9nal Labora-less than half those in test HI-2, where more oxidation and a tory. August 1985. 212pp. 8509110032. ORNL/TM-9488.

!argo axial crack probably we e significant factors in the release 32558:154.

of fission products. A new model has been developed for pred6cting the rate at which gaseous rnolecular iodine is aasorbed by water sprays.

NUREG/CR-4038: SENSITIVITY AND UNCERTAINTY STUDIES The model is a quasi-steady state mass transfer model that in-OF THE CRAC2 COMP'JTER CODE. KOCHER,D.C.;

cludes the iodine hydrolysis reactions. The part netors of the WARD.R.C.; KILLOUGH,G.G.; et al. Oak Ridge National Labora-tory. May 1985. 247pp. 8507250201. ORNL 6114. 31793.025. model are spray drop size, initial concentrabon of the gas and liquid phases, temperature, pressure, buffered or unbuffered This report presents a study of the sensitivity of reactor acci-spray solution, spray flow rate, containment diameter and drop dont conwquwws predic'ed by the CRAC2 computer code to uncertainties in selected trK>dels and parameters useu in the fall height. The results of the model were studied under many values of these parameters. Plots of concentration of iodine code. The sources of cucertainty that were investigated include (1) the model for plume rise, (2) the model for wet deposition, species in the drop versus time have been produced by varying the initial gas phase concentration of molecular iodine over the (3) the meteorological bin-sampling procedure for selecting range of 1 x 10 (-5) moles / liter to 1 x 10 (-10) moles / liter, a weather sequences involving rain, (4) the dose conversion fac-buffered pH of 7 or 9, and a drop size of 1000 microns. Results tors for inhalation as they are affected by uncertainties in the from the model are compared to results available from the Con-physical and chemical form of the released radionuclides, (5) ths weathering half-time for extemal ground-surface exposure- tainment Systems Expenments at Pacific Northwest Laboratory.

The difference between the model predictions and the experi-and (6) the transfer coefficients for terrestrial foodchain path-ways. The most irnportant sources of uncertainty in our analy- mental data ranges from 120% to 68% with the closest agree-ses were the choice of wet-deposition model, the dose conver- ment 7.7%. The new spray model is also compared to previous-ly existing spray models. At high concentrations of gaseous mo-sion factors for inhalation, and the weathering half-time for ground-surface exposure. The choice of pluma-rMa modal the lecular iodine, the new spray model is considered to be less ac-use of an attemative bin-sampling procedure, and uncertainties curate than the previous models. At low concentrations, the in terrestrial foodchain pathways usually had insignificant effects new model predicts results that are closer to the experimental on CRAC2 prediction. data. Inclusion of the iodine hydrolysis reactions is shown to be important for determining the removal of molecular iodine from NUREG/CR-4060: THE DC-1 AND DC-2 DEBRIS COOLABILITY gas phase by water sprays for most conditions.

AND MELT DYNAMICS EXPERIMENTS. HITCHCOCK.J.T.;

KELLY,J.E. Sandia National Laboratories. July 1985. 163pp. NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE 6508090648. SAND 84-1367. 32101:142. II. Semiannual Report, . October 1984 - March 1985.

The DC experiment series investigates the heatup and melt of WILKOWSKl,G.M.; AHMAD,J.; BARNES C.R.; et al. Battelle Me-dry reactor core detmis through nuciear heating of actual reactor morial Institute, Columbus Laboratories. July 1985. 367pp.

materials in order to obtain the thermal properties of dry debris, 8508090632. BMI-2120. 32100:136.

the nature of the transition from a debris bed to a molten pool, The efforts in this report are broken into six work packages and the thermal and kinetic behavior of molten pools. The pur, related to pipe-fracture research efforts and six work packages pose is to develop a data base in support of model develop- that are supporting research efforts. The pipe-fracture efforts in-ment. The work is jointly sponsored by the USNRC, the PNC volve only circumferential crack orientations. Twenty-six pipe ex-(Japan), and EURATOM. This report prcvides a description of periments have been conducted to date, with all but two at the two experiments in the DC series and documents the con- 500F (288C). Approximately 35 additional pipe experiments figuration and the data. These tests investigated dry debris beds from past programs were also analyzed. Analysis efforts include (2 kg) composed of pure UO(2) and mixed UO(2) and stainless limit load and elastic-plastic fracture mechanics analysis. Elas-steel. Heat transfer characteristics were studied at several tic-plastic fracture-mechanics analytical efforts concentrate on steady-state conditions below melt. The beds were then take, J-integral estimation schemes that can predict loads and dis-into melt to observe the growth of a molten pool in the UOr , placements (predictive J-estimation schemes), rather than those bed and the agglomeration and migration of steel in a compo.. that can only be used to calculate the toughness (rkfactor anal-ite bed. The peak measured temperaturo in the UO(2) bed was ysis). Finite-element analym a.r corducted in selected cases.

above 3000 degrees centigrade. Approximately 50% of the Supporting research efforts involve geomety effects on J-R urania formed a molten pool. In the mixed UO(2) and steel bed, curves, notch acuity effects, predicsing J-R curves with large the peak measured temperature was 2600 degrees cents:ade. amounts of crack growth from small specimens, development of With about 90% of the steel molten, material migration occurred a large compliant pepe test system, evaluation of cracks in resulting in a significant increase in the gross bed thermal con- welds, and procurernent of cracked (pe removed from service.

NUREG/CR-4085: USERS MANUAL FOR CONTA:N t.0.A Com.

NUREG/CR-4080: DETERMINATION OF THE AVAILABILIT/ OF puter Code for Severe Reactor Accident Containment Analysis.

CORE EXIT T11HMOCOUPLES DURING SEVERE ACCIDENT BERGERON.K.D.; CLAUSER,M.J.; HARRISON.B.O.; et al.

SITUATIONS. EDSON,J.L EG&G Idaho, Inc. (subs. of EG&G, Sandia N,%cnal Laboratories. July 1985. 354pp. 8508090656.

Inc.),. September 1985. 47pp. 8510030430. EGG-2366.

SAND 84-1204. 32099:142.

32848.082- The CONTAN 1.0 computer code is an integratid analysis This report presents the findings and recommer.uations of the tool for the physical, chemical, and radiological conditions inside Nuclear Power Plant Instrumentation Evaluation (NPPIE) pro- a containment building following the release of radioactive ma-gram conceming signal validation methods to determine the on- terial fror9 the pnmary system in a severe reactor accident. It line availability of core exit thermocouples during accident situa- can also predict the source term to the environment. The pur-tions. Methods of selecting appropriate signal validation tech- po3e of Ws User's Manualis to provide a basic understanding niques are discussed and sources of error identified. TI.is report of the features and models in CONTAIN 1.0 so that users can shows that through the use of these techniques the existence prepare reasonable input and understand the output and its sig-of high-temperature-caused errors may be detected as they nificance for particular applications. Besides input instructions, occur. Specific recommendations for application of seiected the User's M*nual also contains brief descriptions of the basic

Main Citations and Abstracts 15 features of the models. Both light. water reactors and liquid- analyses of complex computer models. In particular, this pro-metal reactors can be modeled with CONTAIN 1.0. The code gram is most useful in analyzing input-output relationships when includes atmospheric models for steam / air <hermodynamics, in- the input has been selected using the Latin hypercube sampling tercell flows, condensation /evaporabon on i turctures and aero- program developed at Sandia (Iman and Shortencaner,1984).

sols, aerosol behavior, hydrogen burning, Todium/ atmosphere The present computer program calculates the p,trtial correlation chemistry, sodium-spray fires, and sodium-pool fires. It also in* coefficients and/or the standardized regression coefficients cludes models for reactor cavity phenomena such as core / con- from the multivariate input to, and output from, a computer crete interactions, coolant-pool boiling, an1 sodium / concrete model. These coefficients can be calculated on either the origi-interactions. Heat conduction in structures, fission-product nal observations or on the ranks of the onginal observations.

decay and transport, radioactive heating, and the thermal-hy. The coefficients provide alternative measures of the relative Caulic and fission-product decontaminate aspects of engi- contribution (importance) of each of the vanous inputs to the neered saMy Wes are also MeM observed output variations. Relationships between the coeffi-NUREG/CR-4107: SEQUENTIAL TEST PROCEDURES FOR DE- cients and differences in their interpretates are identified. If TECTING PROTRACTED MATERi tLS LOSSES- the computer-model output has an associated time or spatial GOLDMAN,A.S. Los Alamos Scientific Laboratory. July 1985. history then the computer program will generate a graph of the 51pp. 8508150033. LA-10319-MS. 32196:285. coefficients over time or space for each input.vanable. output Sequential tests are required for detecting protracted (trickle) variable combination of interest, thus indicating the importance losses of strategic special nuclear matenals I om a single mate- of each input over time or space. The computer program is rials control unit (MCU). We compared applic,ible tosts including user-friendly and written in FORTRAN 77 to facditate portability.

modified versions of Page's test and power-o se procedures. We used simulated data from a MCU in a cor<ersion/ fabrication NUREG/CR-4125 Voi: GUIDELINES AND WORK 3OOK FOR AS-process that tool into account process variations, materials holdup, and measurement uncertainties. Compansons were SESSMENT OF ORGANIZATION AND ADMINISTRATION OF made over a 60-day accounting period under different loss sce- UTILITIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PLANT. Volume 1: Guidelines For Utihty Organization narios. Some important findings include. (1) N ) single procedure And Aoministration Plan. THURBER,J.A.; OLSON J.;

b best for all civersson scenanos. (2) Power-cie procedures are best for protracted losses that occur earty in the accounting OSBORN,R.N.; et al. Battelle Human Affairs Research Centers.

period and Page's test is best for late loss occurrence. (3) If August 1985. 41pp. 8508230324. PNL-5374. 32358:304.

holdup process variations are not included in'the Inventory Dif- This report is a partial response tc the requirements of item ference model but are present in the procets, then assuming i.B.1.1 of the "NRC Action Plan Developed as a Result of the steady-state conditions, flase-alarm probabilitie s can double. TMI-2 Accident," NUREG-0660, and is designed to serve as a NUREGICR-4116: NUFEGO-NP A DIGITAL COMPUTER CODE basis for replacing the earlier NUREG-0731, " Guidelines for FOR THE UNEAR STABIUTY ANALYSIS OF BOILING WATER Utility Management Structure and Technical Resources." The NUCLEAR REACTORS. PENG,S.J.; PODOWSKI,M.Z.; Guidelines are intended to provide guidance to the user in pre-LAHEY,R.T. Rensselaer Polytechnic Institute, Troy, NY. August paring a wntten plan for a proposed nuclear organization and 1985, 437pp. 8509060217. 32502:001. administration. The purpose of the Workbook is to guide the The phenomena of nuclear-coupled density wave oscillations NRC reviewer through a systematic review and assessment of a are of considerable importance in boiling water nuclear reactor proposed organization and administration. It is the NRC's inten-(BWR) stability analysis. A state-of-the-art :inear frequency tion to incorporate these Guidelines and Workbook into a future domain digMa! computer code, NUFREO-NP, has been devel- revision of the Standard Review Plan (SRP), NUREG-0800.

oped for either forced or natural circulation BWR stability analy- However, at this time the report is being published so that the sis. The NUFREO-NP code can be excited by many external material may be used on a voluntary basis by industry to sys-perturbations, including system pressure pedurbation. It is tematically prepare or evaluate their organization or administra-based on one dimensional drift-flux thermal hydraulics, and tion plans. Use of the report by the NRC would not occur until allows for subcooled boiling, arbitrary nonuniform axial and after it has been incorporated in the SRP.

radial power shapes, distnbuted local losses (e.g., spacers),

point or multi-dimensional neutron kinetics, and detailed fuel NUREG/CR-4125 V02: GUIDELINES AND WORKBOOK FOR AS-element dynamics. It has been compared with both out-of-core SESSMENT OF ORGANIZATION AND ADMINISTRATION OF and in-core data, and good agreement has been found. UTILITIES SEEKING OPERATING LICENSE FOR A NUCLEAR NUREG/CR-4119: INTEGRITY OF CONTAINMENT PENETRA. POWER PLANT. Volume 2: Workbook For Assessment Of Orga-TIONS UNDER SEVERE ACCIDENT CONDITIONS FY84 nization And Management. THURBER,J.A.; OLSON.J.;

ANNUAL REPORT. SUBRAMANIAN.C. Sandia National Labora- OSBORN,R.N.; et al. Battelle Human Affairs Research Centers.

tories. August 1985. 37pp. 8508210429. SAND 85-0016. August 1985. 92pp. 8508260008. PNL-5374. 32363:272.

32334:167. See NUREG/CR-4125,V01 abstract.

This document is an annual report for FY84 on the NRC-funded program titled " Integrity of Containment Penetrations NUREG/CR-4130: ICEDF:A CODE FOR AEROSOL PARTICLE Under Severe Accident Conditions." The purpose of this pro- CAPTURE IN ICE COMPARTMENTS. OWCZARSKI,P.C.;

gram is to evaluate the behavior of seals and ga-kets and SCHRECK,R.f.; WINEGARDNER,W. Batteile Memorial Institute, major fixed and operable penetrations. The scope and objec- Pacafic Nor%est Laboratoies. September 1585. 93pp.

tives of this program are discussed as well as the test matrix, SS10040377. PNL-5379. 32857:295.

test facilities, and test procedures. This report descrit,es the technical bases and use of comput-NUREG/CR-4122: A FORTRAN 77 PROGRAM AND USER'S er code ICEDF. ICDEF was developed to serve as a tool for GUIDE FOR THE CALCULATION OF PARTIAL CORRELATION calculating particle retention in pressunzed water reactor (PWR)

AND STANDARDIZED REGRESSION COEFFICIENTS. ice compartments during severe accidents. This report also IMAN.R.L; SHORTENCARIER; JOHNSON J.D. Sandia National serves as a complete user's guide for the most recent stand-Laboratories. August 1585. 54pp. 8506210437. SAND 85-0044. alone version of ICDEF. A complete code desenption, code op-32333:270. erating instructions, code listing, examples of the use of ICEDF, This document is for users of a computer program developed and a summary of a parameter sensitivity study support the use by the authors at Sandia National Laboratories. The computer of code ICEDF.

program is designed to be used in conjunction reth sensitivity 1

1 1

1 16 Main Citations and Abstracts NUREG/CR-4137; PRETEST PREDICTIONS FOR THE RE- the Auxitiety and Fuel Handling Buildings. The water was decon-SPONSE OF A 1:8-SCALE STEEL LWR CONTAINMENT taminated using a demineralization system called EPICOR-il de-BUILDING MODEL TO STATIC OVERPRESSURIZATION. veloped by Epicor,Inc. The Low-Level Waste Data Base Devel-CLAUSS.D.B. Sandia Nabonal Laboratories. Jufy 1985. 53pp. opment-EPICOR-il Resin / Liner invesbgation Project, funded by 8508120552. SA ND85-0175. 32147:107, the U.S. Nuclear Regulatory Commission, is studying the chemi-Tha analy*es usad to pra+ct tha babevice of a 1 B *cale ca: and phys; cal w,dibons of the synthetic ion cxchange rosins model of a steet LWR containment building to static overpres- found in several EPICOR-il prefilters. The work is being done by l surization are desenbed and results are presented. Finite strain, EG&G Idaho,Inc. at the Idaho National Engineering Laboratory.

targe displacement, and nonlinear material properties were ac- This report summarizes results and analyses of the first sam-counted for using finite element methods. Three-dimensional ping of ion exchange resins from EPICOR-il prefilters PF-8 and models were needed to analyze the penetrations, wtuch includ. 20. Results are compared with baseline data from tests per.

ed operable equipment hatches, personnel lock representabons, 9 formed on unirradiated Epicor, lie. resins to determine if degra-i and a constrained pipe. It was concluded that the scale model dation has occurred due to the high intemal radiation dose re-wwld fait due to leakage caused by large deformations of the equipment hatch sleever. ceived by the EPICOR-Il resins. Results also are compared with recent findings on resin degradation by Battelle Columbus Lab-NUREG/CR-4138: DATA ANALYSES FOR NEVADA TEST SITE oratones and Brookhaven National Laboratory.

1 (NTS) PREMIXED COMBUSTION TESTS. RATZEL,A.C. Sandia i National Laboratories. Jufy 1985.179pp. 8507250130. SAND 85- NUREG/CR-4151: INTEGRATION OF EMERGENCY ACTION 0135. 31794:341. LEVELS WITH COMBUSTION ENGINEERING EMERGENCY This report provides results from an indepth analysis of OPERATING PROCEDURES.By Use of Combustion Engineer-twenty-one premixed large-scate combustion experiments spon- ing Owners Group Emergency Operating Procedure Technical sored by the NRC and EPRI and conducted by EG8G at the Guidelines. FALETTI,D.W.; JAMISON.J.D. Battelle Memonal in-

Nevada Test S4te (NTS). These experiments were performed in s*,itute, Pacific Northwest Laboratories. September 1985.137pp.

a 2048 m(3) spherical vessel (hydrogen dewar) with mixtures of 8500250160. FNL-5392. 32753.295.

hydrogan, steam, and air Ignited by glow plugs or heated resist-Pacific Northwest Laboratory, under contract to the U.S. Nu.

, ance coils. Hydrogen concent stions ranged from 5 to 13% (by clear Regulatory Commission developed a method for integrat-volume) and steam concentrations from 4 to 40%. Several tests also incorporated spray systems and/or fans which enhanced ing Emergency Action Levels (EALs) with plant-specific Emer-the combustion rate and significantly altered the postcombus- gency Operating Procedures (EOPs) using the Combustion En-uon gas cooling. Data provided by EPRI from instrumentation gineenng Owner's Group Emergency Operating Procedure designed to charactenze the thermal environment in the dewar Technical Guidelines (CEOG EOPTGs). Using the Combustion during and following combustion have been evaluated. The data Engineering Owner's Group Technical Guidelines document, a i

reduction package SMOKE has been used to process data from set of emergency class definitions and enteria were developed

thin-film gauges, commercial heat flux gauges, capacitance ca- based on the status of the three main fission product bamers i lorimeters, gas and wall thermocouples, and pressure sensors. (fuel cladding, primary coolant system and containment). The i Local measurements of the heat transfer are provided from the EOPTGs were then annotated to point out where, in a symp-

. calonmetry, and global averages are inferred from the pressure. tom /funco ~ based EOP pattemed after the EOPTG, the in-Instrumentation " goodness for each test is assessed based on ferred p et condition is such that a specific EAL may have

the raw data and on comparisons of local and global resutts. been exceeded. After the EOPTGs have been annotated, the Graphical and tabular results are prooded for each test, and proposed method was tested by applying it to the reactor acci.

l trends observed from the results are reported. dent secuences that were shown in the reactnr safety study to

! NUREG/CR-4143: REVIEW AND EVALUATION OF THE MILL. dominate accident risk to determine if an EAL set linked to the STONE UNIT 3 PROBABILISTIC SAFETY STUDY.Contanment EOP annotations would produce timely and accurate classifica-Failure Modes. Radiological Source-Terms And Offsite Conse- tion of the risk dominant sequences. Additional annotations and quences. KHATIB-RAHBAR; PRATTW.; LUDEWIG,H.; et al. additions to the EOPTGs were developed and tN revised anno-I Brookhaven National Laboratory. September 1985- 75pp. tations were shown to produce timely and accurate event clas-8510020257. BNL-NUREG-51907. 32829:198. sifications for all the accident sequences.

l A technical review and evaluatic'1 of the Millstone Unit 3 Probabilistic Safety Study has been performed. It was deter. NUREG/CR-4182: VERIFICATION OF SOIL STRUCTURE INTER-mined that (1) long-term damage indices (latent fatalities, ACTION METHODS. MILLER,C.A.; COSTANTINO,C.J.; PHILIP.

persorwem, etc.) are dominated by late failure of the contain- PACOPOULO; et al. Brookhaven National Laboratory. July ment, (2) short-term damage indices (earty fatalities, etc.) are 1985.182pp.8508090676. BNL NUREG-51893. 32102:161.

dominated by bypass sequences for intemally initiated events. Soil-st ucture interaction (SSI) methods currently used by in-while severe seismic sequences can a'so contnbute significantly dustry to evalue:e the seismic response of nuclear power plant to early damage indices. These overall estimates of severe ac- facilities are reviewed veith thJ aim of evaluating those areas of cident risk are extremely low compared with other societal uncertainty which stil. exist in the analytic approaches. The pri-sources of risk. Furthermore, the risks for Mitistone-3 are com- mary methodologies used by vanous agencies generally can be parabie to nsks from other nuclear plants at high population grouped into three areas, namety, lumped parameter methods,

=itn Se4micci laduced acudents dominate the sevnr acci- finite element methods of combined soil / structure systems, aqd dent risks at Millstone-3. Potential rnitigative features were substructuring methods of analysis. Each of these are dis-shown not to be cost-effectue for internal events. Value-impact cussed in the report. In generai, it was found that lumped pa-analyses for seismic events showed that a manually actuated rameter approaches yield reasonable results provided that the containment spray system might be cost-effective.

site is relatively uniform and the seismic inputs are low enough NUREG/CR-4150: EPICOR-il RESIN DEGRADATION RESULTS such that nonlinear effects are unimportant. The flaite elsmont 1

FROM FIRST RESIN SAMPLES OF PF-8 AND PF-20. results are reasonable provided that extreme care is taken in MCCONNELL,J.W.; SANDERS,RA EG&G Idaho, Inc. (subs. of determining mesh geometry, boundary conditions, 3D effects.

EG&G, Inc.) July 1985. 52pp. 8508150017. EGG-2376. etc. Similar conclusions can be apphed to the structuring ap.

< 32190:267. proaches.

l The 28 Mrsch 1979 accident at Three Mile Island Unit 2 re-leased approximately $60,000 gallons of contaminated water to 1

i

Main Citations and Abstracts 17 NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA NUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR FOR ACCIDENTAL RADIONUCLlDE RELEASES FROM NU- POWER PLANT VALVE FAILURE-RATE VARIABILITY-SOME CLEAR REACTORS. RUNKLE,G E.; OSTMEYER,R.H. Sandia PRELIMINARY RESULTS. BECKMAN.R.J.; MARTZ,H.F. Los National Laboratones. September 1985. 61pp. 85100102:9. plamos Scientific Laboratory. July 1985.70pp.8508090719.LA-SAND 85-0283. 32833:329. 10396-MS. 32105:076.

This report reviews dosimetry models for estimahng the ab- Valve failure data from the In-Plant Reliability Data System soroed dose from ertemal and internal exposure to radionu- (IPRDS) are statstically analyzed using the Fadure Rate Aaaly-clides, important modeling parameters and assumptions are de- sis Code (FRAC). Data from the five fasiure modes, four of scribed. Recommendations for the dosimetry data to be used in which are time related and the other demand related, are ana-the MELCOR health and economic conseqt snce model are lyzed to determine which of the '.ectors--operating system, valve made. For estimating the dose from cloudshine and ground- size, vane type, operating type, and operabng mode--mnst shine, the models for extemal exposure developed by Kocher affect valve failure rates. A seurate analysis is given for each are recommended. The ICRP-30 models and metabohc param- of two plants, a pressunzed water reactor (PWR) and a boiling eters are recommended for estmating the dose from radionu- water reactor (BWR). For both plants and each failure mode, clades deposited intemally via inhalabon and ingestion. Dose mutteplicatrve adjustments for the mean are oblakiud fur catego-conversion factors calculated with these models for a variety of ries, such as nuclear or containment systems, of the vanous radionuclides, clearance classes, particle sizes and integrabon factors. These multipliers indicate whether a particular category periods were obtained from Oak Ridge National Laboratory for of a factor has a corresponding failure rate that is less than the use in the MELCOR health and economic consequence model. average failure rate (a multiplier less than one) or greater than Sources and magnitude of uncertainty in dose factors were average (a multiplier greater than one). Based on the multiplica-evaluated. Recommendations are made for assessing the un- tive adjustments, ball valves are shown to be the most reliable certainty in estmated consequences due to uncertainty in dose valves for the PWR plant. Globe and gate valves have the high-conversion factors, est failure rates for this plant. The average failure rate for the BWR plant is found to be hatf that of the PWR plant for three of NUREG/CR-4213. SET 3 REFERENCE MANUAL. MARELLR.B. the five fadure modes studied. In addibon to the multpliers.

Sandia National Laboratones. July 1985. 250pp. 8508090642. point eshmates and confidence intervals on the failure rates are SAND 83-2675. 32097:13 /. given for selected valve factor combinations. These estimates The Set Equation Transformation System (SETS) is used to and intervals are compared with several other estimates.

achieve the symbohc manipulation of Boolean equations. Sym-botic maniputabon involves changing equations from their ongi- NUREG/CR-4219 V01: HEAVY-SECTION STEEL TECHNOLOGY nal forms into more useful forms - partcularly by applying Bool. PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTO-ean identbes. The SETS program is an interpreter which reads, BER 1984 - MARCH 1985. PUGH.C.E. Oak Ridge Nabonal Lab-interprets, and executes SETS user prograw. The user wntes a oratory. July 1985. 216pp. 8507250168. ORNL/TM-9593/V1.

SETS user program specifying the processing to be achieved 31783:146.

and submits it, along with the required data, for execution by The Heavy-Secton Steel Technology (HSST) Program is an SETS. Because of the general nature of SETS, ie., the capabel- engineenng research actnnty conducted by the Oak Ridge Na-ity to manipulate Boolean equations regardless of their r f.% tional Laboratory for the Nuclear Regulatory Commission. The the program has been used for many different kinds of anarysis. program Comprises studies related to a!! areas of the technolo-gy of materials fabricated into thick-section primary-coolant con-NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR tainment systems of light-water-cooled nuclear power reactors.

POWER PLAtrT ACCIDENT CONSEQUENCE ANALYSIS.Part The investigation focuses on the behavior and structural integri-

1. Introduction, integration & Summary.Part II:Scientfic Basis For ty of steel pressure vessels containing cracklike flaws. Current Health Effects Models. EVANS.J.S.; MOELLER,D.W-; work is organized into ten tasks: (1) program management, (2)

COOPER.D.W.; et al. Harvard Univ., Cambndge, MA. August fracture-methodology and analysis, (3) material characterization 1985. 357pp. 8509190140. SAND 85-7185. 32681:109. and propt,rties, (4) environmentally assisted crack growth stud-Aristysis of the radiological health effects of nuclear power ies, (5) crack arrest technology, (6) irradiation effects studies, plant accidents requires models for predicting carly health ef- (7) cladding evaluations, (8) intermediate vessel tests and anal-fects, cancers and benign thyroid nodules, and genetic effects. ysis (9) thermal-shock technology, and (10) pressurized ther.

Since the publicabon of the Reactor Safety Study, addibonal in- mal-shock technology.

formation on radiological hesith effects has become available.

This report summarizes the effort of a program designed to pro- NUREG/CR-4227: HUMAN ENGINEERING GUIDELINES FOR vide revised health effects models for nuclear power plant acci- THE EVALUATION AND ASSESSMENT OF VISUAL DISPLAY dent consequence modelling. The new modeis for earh etWts UNITS. GILMORE,W.E. EG&G idaho, Inc. (subs. of EG&G, address four causes of mortality and nine categories of morbidi- Inc.).. July 1985. 535pp. 8508150085. EGG 2388. 32193:075.

ty. The models for early effects are based upon two pararreter This report provides the Nuclear Regulatory Commission with Weibell functions. They permit evaluation of the influence of a single source that documents known guidelines for conducting dose protraction and address the issues of variation in radiosen- formal Human Factors evaluabons of Visual Display Units sitivity among the population. The piecewise-linear dose-re- (VDUs). The handbook is a " cookbook" of acceptance guide-sponse models used in the Reactor Safety Study to predict can- lines for the reviewer faced with the task of evaluating VDUs al-cers and thyroid nodules have been replaced by finear and ready designed or planned for service in the control room. The linear-quadratic models. The new models reflect the most re- areas addressed are visual displays, controls, control / display in-cently reported results of the follow-up of the survrvors of the tegration, and workplace layout. Guidelines relevant to each of bombings at Hiroshima and Nagasaki and permit analysis of those areas are presented. The existence of supporting re-both morbidity and mcrtality. The new models for genetic ef- search is also indicated for each guideline. A comment section fects allow prediction of genet;c risks in each of the first five and Method for Assessment secbon are provided for each set generations a*ter an accident and include information on the rel- of guidelines.

ative severity of various classes of genetic effects. The uncer-tainty in modelling radiological health risks is addressed by pro- NUREG/CR-4232. THE RESPONSE OF VENTILATION viding central, upper and lower estimates of risks. An approach DAMPERS TO LARGE AIRFLOW PULSES. GREGORY,W.S.;

is outlined for summarizing the health consequences of nuclear SMITH.P.R. Los Alamos Scientific Laboratory. July 1985.71pp.

power plant accidents. 8507250121. LA-10413-MS. 31794.051.

18 Main Citations and Abstracts The results of an expenmental program to evaluate the re- tion conditions. An executive summary is provided including a sponse of ventilaton system dampers to simulated tomado statment of the findings and recommendations of the report.

transients are reported. Relevant data, such as damper re-sponse tme, flow rate and pressure drop, and flow / pressure vs NUREG/CR-4248: RECOMMENDATIONS FOR NRC POLICY ON blade angle, were obtained, and the response of one tomado SHIFT SCHrDULING AND OVERTIME AT NUCLFAR POWER protectrve damper to simulated tomado transients was evaluat. PLANTS. LEWIS.P.M. Battelle Memonal Institute, Pacific North-ed. Empincal relatonsNps that will allow the data to be integrat- west Laboratories. July 1985.150pp. 8508090710. PNL-5435.

ed into flow dynamics codes were developed. These flow dy- 32102:343.

namics codes can be used by safety analysts to p'redict the re- This report contains the Pacific Northwest Laboratory's sponse of nuclear facility ventilaten systems to tomado depres- (PNL's) recommendations to the U.S. Nuclear Regulatory Com-sunzatons. mission (NRC) for an NRC policy on shift scheduling and hours of work (including overtime) for control room operators and NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELEC. other safety-telated personnel in nuclear power plants. First, it

, TRIC MOTOR-OPERATED VALVES USED IN ENGINEERED is recommended that NRC make three additons to its present i SAFETY-FEATURE SYSTEMS OF NUCLEAR POWER pokey on overtime: 1) hmit personnel to 112 hours0.0013 days <br />0.0311 hours <br />1.851852e-4 weeks <br />4.2616e-5 months <br /> of work in a i PLANTS. GREENSTREET W.; MUR PHY,G.A.; 14-day period,192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br /> in 28 days, and 2.260 hours0.00301 days <br />0.0722 hours <br />4.298942e-4 weeks <br />9.893e-5 months <br /> in one EISSENBERG D.M. Oak Ridge National Laboratory. July 1985. year; exceeding these Imts would require plant manager ap-124pp. 8507250149. ORNL-6170/V1. 31808.069. proval,2) add a requirement that licensees obtain approval from 6 TNs is the first in a series of three reports on electnc motor. NRC if plant personnel are expected to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of operated valves (MOVs) to be produced under the U.S. Nuclear work in a 7-day period,132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br /> in 14 days,228 hours0.00264 days <br />0.0633 hours <br />3.769841e-4 weeks <br />8.6754e-5 months <br /> in 28 Regulatory C-,no6's Nuclear Plant Aging Research pro- days, and 2,300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> in one year, and 3) make the policy a gram. TNs progrem addresses the evaluaton and identiftation requirement, rather than a nonbinding recommendation.

of practical and cost-effective methods for detecting, monitor. Second, it is recommended that licensees be required to obtain ing, and assessing the seventy of time-dependent degradation NRC approval to adopt a routine 12-hour / day shift schedule.

(aging and service wear) of MOVs in nuclear plants. These Third, it is recommended that NRC add several nonbinding rec-methods are to provide capabihties for establishing degradation ommendations concerning routine 8-hour / day schedules. Final-

. trends pnor to failure and developing guidance for effective ly, because additional data can strengthen the basis for future maintenance. This report examines failure modes and causes NRC pokcy on overtime, five methods are suggested for collect-resulting from aging and service wear, manufacturer-recom- ing data on overtime and its effects, mended maintenance and surveillance practices, and measura-j ble parameters (including functional indicators) for use in as- NUREG/CR-4249: PRESSURE VESSEL FRACTURE STUDIES l sessing operational readiness, establishing degradation trends, PENETRATING TO THE PWR THERMAL SHOCK and detecting incipient failure. The results presented are based ISSUE: EXPERIMENTS TSE-5,TSE 5A AND TSE-6.

j on information derived from operating experience records, nu- CHEVERTON,R.D.; BALL.D.G.; BOLT.S.E.; et al. Oak Ridge Na-clear industry reports, manufacturer-supplied information, and tional Laboratory. July 1985. 255pp. 8507250152. ORNL-6163.

input from architect-engineer firms and plant operators. 31791:140.

Thermal-shock expenments TSE-5, TSE-6 were conducted for NUREG/CR-4239: ANALYSIS OF THE ABILITY OF CURRENT the purpose of investigating the behavior of surface flaws under HEALTH PHYSICS INSTRUMENTS TO PREDICT DOSE IN EX- precsurized-water-reactor (PWR) overcooling-accident condi-5 POSED INDIVIDUALS. ARMANTROUT,G.A. Lawrence Liver- tions. These experiments were the fifth, sixth, and seventh in a l more National Laboratory July 1985. 315pp. 8507250126. senes of thermal-nhock expenments conducted with large steel UCID-20398. 31788:001- cytinders (A508, class-2 chemistry; 001-mm OD x 70- and 152-l In this study, theoretical calculations of effective dose to body mm wall x 1.2-m length) as a part of the Heavy-Section Steel l tissue using Monte Carlo simulation techniques have been per- Technology (HSST) Program for this purpose. For each of these formed for both gamma ray and beta ray irradiation. Similar cal- experiments the initial flaw was on the inner surface and ex-culations for neutron irradiation by other workers have also tended the full length of the cyhnder. The thermal shock was been reviewed. Evaluations were made of the performance of a applied to the inner surface only, and this was accomplished by series of the more common health physics instruments. In this effectively dunking the test cylinder, initially at 93 degrees centi-

! evaluation, representative instruments for both gamma-ray and grade, into a large volume of liquid nitrogen. Results of the ex-l beta ray survey work were evaluated using a series of calibrated penments have confirmed that (1) linear elastic fracture me-

, radiation sources. These instrument evaluations were then com- chanics (LEFM) is vahd for thermal-shock loading, (2) crack pared against similar evaluations in the literature, and an eval- arrest will take place in accordance with recently developed uation of basic instrument response by type was performed. In crack-arrest concepts, (3) the crack-arrest loughness values for j addition, data on calculated health effects was used to evaluate rising and fahing KO tields are the same, (4) warm prestressing j the abihty of these instruments to predict healtt' effects. Key re. is effective in preventing crack initiation, (5) therrnal shock j sults for the gamma-ray, beta-ray, and neutrcn survey mcters alone cannot drive a flaw all the way through the wall (6) dy- '

are given. namic effects for PWR-vessel thermal-shock loading conditions are negligible (7) in the absence of cladding and under severe NUNG/CR-4240 V01: PHYSICS OF RJACTOR thermal-shock loading condi*, ions finite-length flaws will extend i SAFETY.Ocarterfy Report. January March 1985.

• Argonne Na- on the surface to becomo ver) long, and (8) there can be very tional Laboratory. July 1985. 27pp. 85080107)J. ANL-85-23 large scatter in small-specimen fracture-toughness data.

V01, 31923:219.

This quarterfy progress report summarizes work done during NUREG/CR-4250 VEHICLE BARRIERS. EMPHASIS ON NATU-the months of January-March 1985 in Argonne National Labora- RAL FEATURES. ADAMS,K.G.; ROSCOE 8.J. Sandia National l tory's Applied Physics and Components Technology Divisions Laboratories. September 1985.110pp. 8509300238. SAND 85-

! for the Divishn of Reactor Safety Research in the U.S. Nuclear 0935.32815:292.

i Regulatory Commission. The work in the Applied Physics Divi- The recent increase in the use of car and truck bombs by ter-

, sion includes reports on reactor safety modeling and assess- rorist organizations has led NRC to evaluate the adequacy of I

ment by members of the Reactor Safety Appraisals Section. licensee security against such threats. As part of this evaluation, Work on reactor core thermal-hydraulics is performed at ANL's one of the factors is the effectiveness of terrain and vegetation

Components Technology Divison, emphasizing 3-dimensional in providing barriers against the vehicle entry. The effectiveness code development for LMFBR accidents under natural convec- of natural features is presented in two cortexts. First, certain l

Main Citations and Abstracts 19 natural features are presented. In addition to the discusson of The purpose of this report is to provide the NRC and the nu-natural features, this report provides a discussion of methods to clear industry with information and data which will be useful for slow vehicles. Also included is an overview of man-made bamer occupatonal dose reduction at nuclear power plants. The objec-systems, with particular attention to ditches. tNes of this effort were to: 1. identify the repetitive high-dose jobs, related collective dose ranges and apphcable dose redoc.

NUREG/CR-4251 V01: MITIGATIVE TECHNIOUES FOR tion techniques,2. investigate and recommend improvements in GROUND WATER CONTAMINATION ASSOCIATED WITH the selection of high reliability and low maintenance equipment SEVERE NUCLEAR ACCIDENTS.Valume 1. Analysis Of Generic to assure that collective doses received dunng equipment repair Site Conditions. OBERLANDER.P.L; SKAGGS.R.L; is considerod, 3. recommend improved radioactive waste han-SHAFER,J.M. Battelle Memonal Institute, Pacific Northwest Lab- dling procedure and equipment which could reduce collective oratories. August 1985. 321pp. 8509110279. PNL-5461. dose equivalent, and 4. examine current ALARA incentives and 32559:004- recommend new positve steps whch will provide additional Pacific Northwest Laboratory evaluated the feasibility of using dose-reduction incentives. Ten nuclear sites were visited by two ground-water containment mitiga' ion techniques to control radt- Brookhaven health phys: cists to collect the needed dose-reduc-onuchde migration following a severe commercial nucitw power tion data and information. This report summanzes the findings reactor accident. The two types of severe commercial reactor and recommendatens on the above objectives.

accidents investigated are 1) containment basemat penetration of core melt debns, which slow!y cools and leaches radionu- NUREG/CR-4255 V21: AEROSAL RELEASE AND TRANSPORT clidas to the subsurface environment; and 2) containment base- PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTO-mat penetration of sump water without full penetration of the BER 1984 - MARCH 1985. ADAMS,R.E.; TOBIAS M.L Oak core mass. Six genenc hydrogeologic site classifications were Ridge National Laboratory. August 1985. 56PP. 8509110010.

developed from an evaluation of reported data pertaining to the ORNL/TM-9632/V1. 32565;009.

hydrogeologic properties of all epstng and proposed commer- This report summarizes progress for the Aerosol Release and cial reactor sites. One-dimensonal radonuclide transport analy- Transport Program sponsored by the Nuclear Regulatory Com-ses were conducted on each of the individual reactor sites to mission, Office of Nuclear Regulatory Research, Division of Ac-determine the generic characteristics of a radionuchde dis- cident Evaluation, for the period October 1984. March 1985.

charge to an accessiNe environment. Ground-water contain- Topecs discussed includa (1) steam-only experiments in the ment mitigation techniques that may be suitable for severe NSPP facility; (2) tests in small vessels to study thermal output, power plant accidents, depending on specife site and accident mass generation rates, and other operating features of plasma conditions, were identitied and evaluated. Feasible mitigative torch aerosol generators in support of the development of the techniques and associated constraints on feasibility were deter. Aerosol Moisture interaction Tests (AMIT) facihty and to support mined for each of the six hydrogeologic site classifcations. the LWR Aercsol Conta.nment Experiments (LACE) program at Three case studies were conducted at power p! ant sites located Hanford; (3) analysis of data from plasma torch aerosol genera-along the Texas Gulf Coast and the Ohio River. Mitigative strat- tor tests; (4) analysis of steam behavior in the NSPP vessel in egies were evaluated for their impact on containment transport. aerosol experiments and in steam only tests; and (5) a study of Results show that the techniques evaluated significantly in- the feasibility of expenments for shape factor measurements.

creased ground-water travel times and reduced contaminant mi-gration rates. NUREG/CR-4257: INSPECTION. SURVEILLANCE,AND MONI-TORING OF ELECTRICAL EQUIPMENT INSIDE CONTAIN-NUREG/CR-4251 V02: MITIGATIVE TECHNIQUES FOR MENT OF NUCLEAR POWER PLANTS--WITH APPLICATIONS GROUND-WATER CONTAMINATION ASSOCIATED WITH TO ELECTRICAL CABLES. AHMED.S.; CARFAGNO,S.P.

SEVERE NUCLEAR ACCIDENTS.Vcbme 2. Case Study Anaty- ARVIN/CALSPAN Advanced Technology Center. August 1985.

sis Of Hydrologic Characterization And Mitigative Schemes. 8509100530. 32536:234 OBERLANDER,P.L; SKAGGS,R.L; SHAFER.J.M. Battelle Me- 101pp' The general concepts of equipment condition monitoring as monal Inst:tute, Pacific Northwest Laboratones. August 1985 appliccble to the detection of age-related detenoration of 301pp. 8509100515. PNL-5461. 32538:094. safety-related equipment are desenbed. The goal is to detect See NUREG/CR-4215,V01 abstract. deterioration in the incipient stage, prior to in-service failure and NUf1EG/CR 4252: INDEPENDENT ASSESSMENT OF TRAC. prior to the point at which equipment can no longer be expect-PD2/ MOD 1 CODE WITH BCL ECC BYPASS TESTS. ed to perform its function when t.xposed to design basis acci-SLOVIK,G.C.; SAHA,P, Brookhaven National Laboratory. August dent conditions. The application of condition monitonng is dis-1985. 78pp. 8509180076. 32668:213. cussed specifically for electrical cables. The goal of cable con-This report presents the TRAC-PD2/ MOD 1 independent as. dition monit ri"1 h to d1termine the degree of cable degrada-sessment calculations performed at Brookhaven National Labo. tion and to predt the remairung useful life. In situ nondestruc-ratory (BNL) using the Emergency Cnre Cooling (ECC) bypass tive testing and destructive laboratory testing are discussed. In-experiments conducted in a 2/15 scale PWR vessel at Battelle terim ruommendations are given for the implementation of a Columbus Laborsteries (BCL). Both steady state experiments condition monitoring prograra.

with vanous ECC water subcoolings and transient tests with h ! NUREG/CR-4258: AN APPPOACH TO TEAM SKILLS TRAINING wall effects were simulated. Besides the base cases, severa sensitivity calculations were performed to study the effects of OF NUCLEAR DAVIS.LT.; PC TURNEY,J.R.

GADDY,C.D.; AER PLANT CONTROL General Physics Corp ROOM CRE nodalization, particularfy the relative locations of the hot leg July 1985. 83pp. 8508200093. GP-R 123022. 3227tCPO.

penetrations in the downcomer. In addition, calculations were An invesNation of cwent team skms tranng peactces and performed to determine the effect of slight increases in the re- research was conducted by General Physics Corporation for the verse core steam flow and the associated form losses due to Office of Nuclear Reactor Regulation. The methodology used in-the hot leg penetrations. Code corrections as received from the cluded a review of relevant team skills training literature and a code developers at Los Alamos National Laboratory (LANL) workshop to collect inputs from tearr training practitioners and were afso incorporated into this study, researchers from the public and priva,e sectors. The workshop NUREG/CR-4254: OCCUPATIONAL DOSE REDUCTION AND was attended by representatives from nuclear utility training or-ALARA AT NUCLEAR POWER PLANTS. Study On High-Dose ganizations, the commercial airline industry, federal agencies, Jobs,Radwaste Handling,And ALARA Incentives. DIONNE,B.J.; and slefense training and research commands. The literature re-BAUM.J.W. Brookhaven National Laboratory. July 1985.104pp. views and workshop results provided the input for a suggested 8508090697. BNL NUREG-51888. 32105:192. approach to team skills training that can be integrated into ex.

20 Main Citations and Abstracts isting training programs for control room operating crows. The tude above sea level, and an estimate of the overall uncertain-approach includes five phases: (1) team skills objectives devel- ties in dese-rate measurements; (2) beta-particle and nearty opment, (2) basic team skills training, (3) team task training, (4) monoonergetic electron spectra and their dependence on team skills evaluation, and (5) team training program evaluation. source configuration; and (3) degree of achievable uniformity of Supporting background information and a user-onented desenp- beam cross sections. Included also is a review of the results of tion of tne approach to team skills training are provided. a first attempt to predict instrument response to realistic beta-particle environments from their response to monoenergetic NUREG/CR 4259: TAILINGS NEUTRALIZATION AND OTHEr ALTERNATIVES FOR IMMOD!LfZING TOXIC MATERIALS 1, electrons and knowledge of the approximate beta-particle spec-TAILINGS. Final Report' OPITZ.B.E - SHERWOOD.D.R , tra. Attached to the repM am proposed gMm W estam DODSON;M E.; et al. Battelle Memorialinstitute, Pacific North: W secondary calibration laboratones for radiaten-protection in-west Laboratones. September 1985.131pp. 8510040347. PNL- skuments.

5467,32857 159 NUREG/CR 4268: RATIO METHODS FOR COST-EFFECTIVE This final document, in a senes of six, summarizes research FIELD SAMPLING OF COMMERCIAL RADIOACTIVE LOW-completed since the beginning of the project Three subtasks iFVEL WASTFS FRFRHARDT,L L : SIMMONS.M A ;

are included: Subtsk A Neutralization Metnods Selecten, THOMAS.J M Battelle Memonal Institute, Pacific Northwest Subtask B - Latwratey Analysis; and Subtask C Field Testing. Laboratones. July 1985. 81pp. 8508090570. PNL-5156.

Subtask A reviews treatment processes from other industnes to 32106:069.

evaluate if current waste technology from other fields is applica- An investigation of cost-effective methods for sampling at ble to the uranium industry. This task also identifies several rea- commercial radioactive low-level waste sites has been one goal gents that were tested for their effectiveness in treating acidic of this project. To that end, double sampling was investigated, tailings and tailings solution in order to immobilize the cuntami- and we found that the method appears useful when estimating nants associated with the acid waste. Subtask B desenbes the total radionuclide inventory in waste site environs. The methods laboratory batch and column treatment studies performed on are explained, decision enteria for cost effectiveness presented, sohd waste taihngs and tailings solution over the course of the and a worked example based on field data is provided. The sta-project. The evaluation of several ra gents identified in Subtask tistical basis for the conclusion that double sampling appears to A was based on three enteria: 1) treated effluent water quality, be robust and cost. effective is in separate sectons. Field tests

2) neutralized sludge handling and hydraulic properties, and 3) and addit onal estimates of " field instrument" errors are needed reagent costs and acid neutralizing efficiency. Subtask C pre- to substantiate the findings.

sents a field demonstration plan that will evaluate the effecti-venss, costs and benef ts of neutralizing acdc uranium mill tail- NUREG/CR-4272: RESPONSE TREE ings solution to reduce the potential leaching of toxic trace EVALUATION. EXPERIMENTAL ASSESSMENT OF AN EXPERT metals, radionuclides and macro ions from a tailings impound- SYSTEM FOR NUCLEAR REACTOR OPERATORS.

ment. NELSON,W.R.; BLACKMAN,H.S. EG&G ldaho, Inc. (subs. of NUREG/CR-4260: TORAC USER'S MANUALA Computer Code EG&G Inc),. September 1985. 68pp. 8510040397. EGG-2397.

For Analyzing Tornado-induced Flow And Material Transport in 32859.203.

Nuclear Facilities. ANDRAE,R.W.; TANG P.K.; MARTIN RA.; et The United States Nuclear Regulatory Commission (USNRC) al. Los Alamos Scientfic Laboratory. July 1985. 145pp. sponsored a project performed by EGSG ldaho, Inc., at the 8507250118. LA 10435-M. 31787.001. Idaho Natonal Engineering Laboratory (INEL) to evaluate differ.

This manual desenbes the TORAC computer code, which can ent display concepts for use in nuclear reactor control rooms.

model tomado-induced flows, pressures, and material transport included in this project was the evaluation of the response tree within structures. Future versons of this code will have im- computer-based decision aid and its associated displays. The proved analysis capabilities. In addition, it is part of a family of response tree evaluation task was designed to (a) assess the computer ccJes that is designed to provide improved methods ment of the response tree decision and and (b) develop a tech-of safety analysis for the nuclear industry. TORAC is directed nical basis for recommendaticns, guidelines, and criteria for the toward the analysis of facility ventilation systems, including design and evaluation of computertzed decision aids for use in interconnected rooms and corridors. TORAC is an improved reactor control rooms. Two major experiments have been con-version of the TVENT code. In TORAC, blowers can be turned ducted to evaluate the response tree system. This report em-on and off and dampers can be controlled with an arbitrary time phasizes the conduct and results of the second expenment. An function. The material transport capability is very basic and in- enhanced version of the response tree system, known as the cludes convection, depletion, entrainment, and filtration of mate, automated response tree system, was used in a controlled ex-rial. The input specifications for the code and vanety of sample periment using trained reactor operstos ans test sobjects. This problems are provided. report discusses the automated response 'ree system, the design of the evaluation expenment, and the quantitative results NUREG/CR-4266: STANDARD BETA. PARTICLE AND MONOEN- of the experiments. The results of the experiments are com-ERGETIC ELECTRON SOURCES FOR THE CALIBRATION OF pared to the results of the prwious expenments to provide an BETA-RADIATION PROTECTION INSTRUMENTATION ir,tegrated perspective of the response tree evaluation project.

EHRLICH,M.; PRUITT,J.S.; SOARES.C.G ; et al. Commerce, in additon, a subjective assessment of the rwts addresses Cept. of. National Bureau of Standards. Aursst 1985. 86pD- the impications for the use of advanced "intekigent" decision 8509060202. NBSIR 85-3169. 32505.057. aids in the reac'or control room.

In a project funded jointfy by the National Bureau of Stand-ards (NBS) and the Nuclear Regulatory Commission (NRC), NUREG/CR-4275: HEA'/Y-SECTION STEEL TECHNOLOGY NBS has developed a calibration facility for beta-particle instru- PROGRAM FIVE-YEAR PLAN FY 1984-1988.

• Oak Ridge ments and sources used in radiation-protection dosimetry. The National Laboratory. August 1985.160pp. 8509110024. ORNL/

facility consists of beta-particle and nearly monoenergetic elec- TM-9654. 32558.001.

tron beams characterized in terms of absorbed-dose rates to The second in en annual series of five-year program plan plastic and in terms of beta-particle spectra. A second phase of documents is presented for the Heavy-Section Steel Technolo-the project was concerned with establishing secondary calibra- gy program. The program is carried out by the Oak Ridge Na-tion laboratories for radiation-protection instruments. This final tional Laboratory for the Materials Engineering Branch, Division report includes a detailed discussion of (1) the determination of of Engineering Technology, Office of Nuclear Regulatory Re-absorbed-dose rates to plastic for each beta-particle and nearly search of the U.S. Nuclear Regulatory Commission. The pro-monoenergetic electron beam, dose-rate dependence on alti- gram is aimed at advancing the understanding and validation of

Main Citations and Abstracts 21 materials and structures behavior as they relate to light water energies greater than 1.0 MeV (F> 1.0 MeV) is the most widely reactor pressure vessel integrity. The program has nine techni- used parameter; however, current thinking favors displacements cat tasks and a management function. A background statement per atom (dpa) in iron as better related to the physical mecha-end a plan-of-action is given for each. The nine techrucal tasks nism of radiation damage. Fluences for energies greater than addrest fracture methodology and analysis, materials character- 0.1 MeV (F> 0.1 MeV) are also considered since neutrons in ization, crack growth, crack arrest, irradiation effects, cladding the 0.1 to 1.0 MeV range are likely to contnbute to the damage.

evaluations, intermediate-vessel testing, thermal-shock testing, in order not to prejudice future investigations, all three damage and pressurized thermal-shock expenments. parameters F> 1.0 MeV, F> 0.1 MeV, and dpa will be listed in I Ws rp Ms NUREG MX cone N won expose pa-NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERI- rameters W the 12 meaHurgical sme capsuM wM ENCE AND ASSISTANCE OF A SHIFT TECHNIOAL ADVISOR comprise the Fifth HSST Irradiation Series, in order to make l (STA) ON CREW PERFORMANCE IN CONTROL ROOM SIMU- ayadaMe tM dam M a % maner aM not to May N anah LATORS. BEARE,A.N.; DONOVAN,M.D.; LASSITER,D.L; et al.

ysis of capsules irradiated early he the series, it was decided to General Physacs Corp. Septr'mber 1985. 202pp. 8510040318. make this NUREG a loose-leaf document, The exposure values will be distnbuted as they become amiable.

.i de second expenment using a tranng simulator to evaluate effects of expenance level of Senior Re- NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING actor Operators (SRO) in the supennsor's role, and presence of IN. LIGHT WATER REACTORS. Annual Report, October 1983 -

0 Shrft Techrucal Advisor (STA) on performance of nuclear September 1984. SHACK,WJ.; KASSNER T.F.; MAlYA,P.S.; et power plant control room operators / crews. The experiment was al. Argonne National Laboratory. August 1985. 149pp.

conducted in a pressurized water reactor (PWR) plant-refer. 8508210430. ANL-85-33. 32338:265.

enced simulator. Data was collected on 20 three-man crews of This progress report summanzes work performed by the Ar-licensed operators, performing four sequences. Performance gonne National Laboratory and a subcontractor, E.F. Rybicki, measures were derived from task analyses of the sequences. Inc., on environmentally assisted cracking in light water reactors One set of measures focused on task performance; the second during the twelve months from October 1983 through Septem-set measured control of system parameters. Instructors' ratings ber 1984.

and performance scores of trainees were compared to scores of operators / crews, to validate the performance measures- NUREG/CR-4268: FOCAL MECHANISM ANALYSES FOR VIR-System parameters and control manipulations were recorded by GINIA AND EASTERN TENNESSEE EARTHOUAKES (1978-the simulator's enmputer. Communications and selected venfi- 1984). BILLINGER,G.A.; TEAGUE A.G.; MUNSEY,J.W.; et al.

cations were recorded on checklists and videotapes. Question- Virginia Polytechnic institute & State Univ., Blacksburg, VA.

naires recorded biographicalinformation and self reported work- June 1985. 91pp. 8508010243. 31928:001.

loads. No significant differences in overall performance were Focal mechanisms are presented for 11 earthquakes from the found attnbutable to expenence of supervisors, nor to presence Giles County, Virginia, seismic zone and its viciruty and for 12 of a STA. Results were sirrular to results of an earlier experi- eathquakes from the Central Virginial seismic zone. These ment paed with boiling water reactor (BWR) crews. These earthquakes (0<M<4) were monitored by local networks be-results are also reported. Reported workloads of supervisors as- tween January 1978 and October 1984. In Giles County, the sisted by STA/s were sigruficantly lower than workloads report- data base consists of 43 P-wave pocrities and 50 SB to P am-ed by those unassisted- piituds ratios (SV/P) that yielded six s ngle event focal mechan-NUREG/CR-4281: AN EMPIRICAL ANALYSIS OF SELECTED ims (SEFM's) and five composite erent focal mechanisms NUCLEAR POWER PLANT MAINTENANCE FACTORS AND (CFM's). In Central Virginia,79 P-wave polarities and 51 SV/P PLANT SAFETY, OLSON.J.; OSBORN R.N.; THURBER,J.A.; et ratios are used to determine 11 SEFM's and four CFM's. A al. Battelle Human Affairs Research Centers. July 1985.62pp. computer program FOCMEC was used to determine the focal 8508150065. PNL-5487,32196:224. mechanism solutions. The results for the Giles County seismic This report contains a statistical analysis of the relationship zone show mainly strike-slip mechanisms on steeply dipping (73 between selected aspects of nuclear power plant maintenance degrees plus minus 16 degrees) NNE (right lateral motion) and programs and safety related performance. The report identifies ESE (left lateral motion) trending nodal planes. However, some a large number of maintanance resources which can be expect. (4/11) counterclockwise. The P axes in Central Virginia are gen-ed to influence maintenance performance and subsequent plant erally northeast trending for shallow earthquakes (>8 km) and safety performance. The resources for which data were readily northwest trending for deeper ones (<8 km). In Giles County, available were related statisticalfy to two sets of performance in, where the seismic activity is occwring beneath me Appalachian dica

• ors: maintenance intermediate safety indicators, and final decollement, fautting tnd inferred stress orientations are more safety performance indicators. The results show that the admin. uniform than in Central Virginia, some 200 km away, where the istra"vo structure cf the plant maintenance program is a signif. seisnucity is occwnng near and above me decollement.

cant predictor of performance on both sets of indicators.

NUREG/CR-4290 V02: PROBABILITY OF PIPE FAILURE IN THE NUREG/CR-4284: NEUTRON EXPOSURE PARAMETERS FOR REACTOR COOLANT LOOPS OF BABCOCK AND WILCOX THE FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIA- PWR PLANTS. Volume 2: Guillotine Break Indirectly induced By TION SERIES. STALLMANN,F.W.; KAM.F.B.; BALDWIN,C.A. Earthquakes. RAVINDRA,M.K.; CAMPBELL R.D.; KIPP,T.R.; et Oak Ridge National Laboratory. August 30, 1985. 39pp. al. Law eve Livermore Na nonal Laboratory. July 1985.146pp.

8509110035. ORNL/TM-9664. 32564:337. 8508010757. UCRL-53644. 31924 001, The Nuclear Regulatory Commission's (NRC's) Hesvy Section The requirements to design nuclear power plants for the ef-Steel Teh;oy (HSST) Pretram is concemed with the inves- fects of an instantaneous double-ended guillotine break (DEGB) tigation of cracklike flaws in reactor pressure vessel steels. In of the reactor coolant loop (RCL) piping have led to excessive the fifth irradiation series, capsules containing a variety of met- design costs, interference with normal plant operation and main-alturgical test specimens were irrad;ated to fluences in the tenance, and unnecessary radiation exposure of plant mainte-range of 1.10(19) to 3.10(19) nuetw/cm(2) (E> 1.0 MeV). In nance personnel. This report desenbes an aspect of the NRC/

order to correlate radiation embrittlement to damage fluences, Lawrence Livermore National Laboratory sponsored research accurate determination of the neutron fluence spectra at the program aimed at demonstruing that the probability of DEGB in critical location of the test specimen is needed. The part of the RCL Piping of nuclear power plants is acceptably small and the neutron spectrum which is responsible for the radiation damage requirements to design for the DEGB effects (e.g., provision of b characterized as " damage exposure parameter." Fluences for pipe whip restraints) may be removed. This study estimates the

1 i l 22 Main Citations and Abstracts probability within the centaintnent of Babcock & Wilcox supplied flew detecton se discussed in this report. The penod covered is pressurizd water reactor nuclear power piants in the United October 1,1984, to Apnl 1,1985. Topes include final analysis States. The medium probability of indirect DEGB was estimated of ZB-1 vesset test data, prepantion for continuous AE monitor-to range between 6x10(-11) and 1x10(-7) per year. Using very ing of Watts Bar Unit 1 reactu dunng operabon, AE signal pat-conservative assumptions, the 90% subjective probability value tern recognition development, and development of an ASTM (confidence) of P(DEGB) was found to be less than 1x10(-5) per standard for applicabon of contnuous AE monitoring to pres-year.

sure boundaries.

NUREG/CR-4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSICAL BENCHMARKING OF P! PING SYSTEMS. NUREG/CR-4303: HIGH. LEVEL WASTE PRECLOSURE SYS-BEZLER.P.; SUBUDHI,M., SHTEYNGART,S.; et al. Brookhaven TEMS SAFETY ANALYSIS. Phase 1 Final Report. HARRIS P.A.;

National Laboratory. September 1985.105pp. 8509300512. LIGON,D.M.; STAMATELATOS.M.; et al. GA Technologies, incl BNL.NUREG-51897, 32792:322. General Atomic Co. September 1985. 330pp. 8509300521.

Physical benchmark evaluations were used to assess the ac. SAND 85-7192. 32810:001.

curacy and adequacy of the analysis methods and assumptions The major effort for this project has been on the gathering, used in typical piping quJification evaluations. Physical bench- organizing, and assembling of information pertnent to the safety mark evaluabons have been completed for six systems involving assessment of a nuclear waste repository curing preclosure op-both laboratory and in situ tested piping. In each evaluation erations. Specific issues addressed in this report are: 1. De-elastic finite element methods were used to predct the time his- tailed analysis of a conceptual basalt repository design in order tory response of a system for which physical test results were to identfy potential instaating event / accident scenarios capable available. In the analytcal simulations the measured support ex-of causing radiological and/or nonradiological consequences. 2.

citations and the measured damping propertes were used as Evaluation of radiological and nonradiological consequences rel-input and the acceleration and displacement response of piping interior points were predcted as output. The linear analysis evant to a nuclear repository and recommendation of an ap-methods were found to provide reasonable estimates of system roach for quanttative evaluabon of these consequences. 3.

response. For a near linear system and using conservative esti- Comparative evaluation of severalimportance ranking measures mates for system damping, a good correlation of response that had been used in the nuclear industry in order to select a traces and acceptable estimates of response peaks can be ex. measure to best meet the needs of the program. 4. Develop-

pected. Using realiste eshmates of ur9 form system damping, ment of event and fault tree models for those initiating events large underestimates of peak response components were ob- which have passed the preliminary screening process. 5. Com.

served and deviations of 100% or greater should be expected. pilation of specifs data such as initiating event frequencies, NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTING- component / system failure rates and repair times, personnel HOUSE REACTOR COOLANT PUMP. BOARDMAN T.; injury, and basic information necessary for more detailed radio-JEANMOUGIN,N.; LOFARO,R.; et al. Rockwell Intemational logical consequence evaluabons at a later bn* & Selechon of Corp. July 1985. 66pp. 8508020424. 85-ETEC-DRF-171. a set of accident scenarios to be quantified in the next Study 31961:006. phase to demonstrate the applicabilvy of the proposed method-An independent analysis was performed to determine seal ology that will identify and quantitatively prioritize structures, leakage rates for the Westinghouse Reactor Coolant Pump components, systems, and operations which are important to (RCP) during a postulated station blackout resulting from loss of safety dunng the preclosure phase of a HLW rep.

ac electric power. The analysis confirmed Wesbnghouse calcu-lations on RCP seals performance for the three conditions in. NUREG/CR-4304: PRESSURE VESSEL FRACTURE STUDIES vestigated: (1) all three seals function, (2) No.1 seal fails open PERTAINING TO THE PWR THERMAL-SHOCK while Nos. 2 and 3 seals function, and (3) all three seals fail ISSUE. Experiment TSE 7. CHEVERTON,R.D.; BALL.D.G.;

open. BOLT.S.E.; et al. Oak Ridge National Laboratory. September 1985.152pp. 8510040402. ORNL-6177. 32859.050.

NUREG/CR-4298: DESIGN AND INSTALLATION OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55. Thermal-shock experiment TSE 7 was conducted for the pur-LEWIS.J.R.; BYERS,K.R.; FLUCKlGER,J.D.; et al. Battelle Me. pose of investigating the behavior of surface flaws under pres-morial inshtute, Pacife Northwest Laboratories. Jufy 1985. surized-water-reactor (PWR) overcooling-accident condibons.

120pp. 8507250138. PNL 5490. 31794:221. This experiment was the eighth in a series of thermal-shock ex-The Pacific Northwest Laboratory has studied the design ard periments conducted for this purpose with large steel cylinders installation of computer managed systems that can help nuclear (A508, class-2 chemistry; 991-mm outside diameter x 152-mm power plant licensees to meet the physical security require- wall x 1.2-m length) as a part of the Heavy-Section Steel Tech-ments of 10 CFR 73 55 (for access control, alarm monitoring, nology (HSST) Program. The initial flaw for TSE-7 was a shal-and Warm recording.) Two objectives were to study the power low, semielliptical, inner surface, axially oviented, sharp crack lo-plaru secunty functions that could be aided by a computer-man- cated at midlength of the test cylinder. The thermal shock was aged physical secunty system and to eva;uato the safety and applied to the inner surface only, and this was accomplished by security considerations of such a system. A further objective effectively dunking the test cyli,1e'. initially at aquivalent 93 de-was to develop guidance on system design, selection, and in- grees centigrade, into a large volume of liquid nitrogen. The

stallation. The design guidance includes safety and security re- specific purpose of TSE-7 was to determine whether, in agree-quirements, design alternatives, computer secunty, workspace ment with analysis, a short and shallow surface flaw, in the ab-design and user interface design. Guidance is also provided on sence of cladding, would extend on the surface to effectively writing a system specification for procurement, bid review proce- become a very long flaw as a result of severe thermal-shock dures, and site preparation.

loading. During the experiment, there were three major initiation-NUREG/CR-4300 V01: ACOUSTIC EMISSION / FLAW RELATION- a. Test events. The first event consisted of some radial propaga.

SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRES- tion and very extensive surface extension, with many bifurca.

SURE VESSELS. Progress Report. October-March 1985. tions taking place. The second and third events consisted pri-HUTTON,P.H.; KURTZ,R.J. Battelle Memorial Institute, Pacific marily of radial propagation. A fourth initiation event was pre-Northwest Laboratories. August 1985. 27pp. 8508210022. PNL- vented by warm prestressing. These results were in good 5511, 32302:232.

l agreement with predictions.

Technical progress in developing continuous acoustic emis-sion (AE) monitoring of nuclear reactor pressure boundaries for 3

Main Citations and Abstracts 23 NUREG/CR-4305: COMMENTS ON THE LEAK-BEFORE-BREAK NUREG/CR-4325: A PARAMETRIC STUDY OF PWR PRESSURE CONCEPT FOR NUCLEAR POWER PLANT PlPING SYSTEMS. VESSEL INTEGRITY DURING OVERCOOLING RODABAUGH,E.C. E.C. Rodabaugh Associates, Inc.

• Oak ACCIDENTS.CONSIDERING BOTH 2-D AND 3 D FLAWS.

Ridge National Laboratory. September 1985. 58pp. CHEVERTON,R.D.; BALL.D.G. Oak Ridge National Labratory.

8510030312.32854:257. September 1985. 40pp. 8510040371. ORNL/TM-9682.

Leak-before-break entails the concept that, with a high 32860:335.

degree of probability, failure of the pressure boundary of piping A continuing analysis of the pressurized water reactor pres-systems will be signaled by a detectable leak which will provide surized thermal-shed prob'emi dest?S tht? th* previcWy ec-ample time to shut down and repair that leak. The status of the cepted degree of cr nservatism in the fracture-mechanics model leak-before-break concept is discussed in this report, including mds b be more Ocsey waluaW aM, d esh, WM c review of industrial are nuclear power plant expenence with Om feature mat was wewd to W consenaM ms me use respect to leak-before-break, fracture mecharucs and potential of w m nal as oppW m HMe-Q Haws. W degree of conservatism could not be adequately investigated elimination of postulated pipe breaks in ruclear power plant because of computational limitations and a lack of knowledge piping design. regarding !!aw behavior; however, that situation has changed to the extent that some cases involving finite-length flaws can be NUREG/CR-4314: BRIEF SURVEY AND COMPARISON OF studied. A flaw of particular interest is one that is located it' an COMMON CAUSE FAILURE ANALYSIS. WALLER,R.A. Los axial weld of a plate-type vessel. For those vessels that suffer Alamos Scientific Laboratory. August 1985. 28pp. 8509130416. relatively high radiation damage in the welds, the length of the LA-10474-MS. 32605:035. flaw will be no greater than the length of the weld, and recent This paper presents a bnet survey of methods and a list of calculations indicate that a deep flaw of that length (is equiva-references for analyzing common cause (mode) failure. Implicrt lent to 2m) is not effectively infinitely long, contrary to previous models, explicit modeling techniques, and computer aids are in- thinking. The benefit to be derived from consideration of the 2-cluded in the discussion. It is suggested that although current m flaw and also a semielliptical flaw with a length-to-depth ratio trends are emphasizing development of explicit models, a realis- of 6/1 was investigated by analyzing several postulated tran-tic assessment of data availability will force continued use of im- sients. In doing so the sensitivity of the benefit to a specified plicit or hybnd models in the immediate future. maximum of the analysis indicate that for some conditions the benefit in using the 2-m f aw is substantial, but it decreases with NUREG/CR-4317 V01: CANADIAN SEISMIC increasing pressure, and above a certain pressure there may be AGREEMENT.Techtwcal Report Covering 1979-1985. no benefit, depend;ag on the duration of the transient and the HAYMAN,R.B.; BASHAM.P.W.; WETTMILLER,R.J.; et al. limit on crack-arrest toughness.

Canada, Govt. of. July 1985. 71pp. 8508090583. 32106:001.

This Final Report provides a comprehensive summary of the NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAIL-act'vities of the Earth Physics Branch (EPB) in eastern Canada URES ON TRANSIENTS AND ACC! DENTS AT A 3-LOOP particularty with respect to the Eastern Canadian Telemetered WESTINGHOUSE PRESSURIZED WATER REACTOR. Main Network (ECTN). The report desenbes the seismographic Idaho, Inc. (sub's. of EG&G' , Inc.).. August 1985. 162pp.

system developed by EPB to morutor the regional seismicity. 8509100428. EGG-2405. 32530:093.

There are detailed descriptions of the logic used to select the This report documen's the evaluation of the effects of nonsa-various physical components of the system, the hardware and fety grade control system failures on a typical 3-loop Westing-software that constitute the recording system, and the data flow house pressurized water reactor plant. The methods utilized for from detection to archive in addition to the engineering details, this evaluation include a system level failure rnodes and effects there is a discussion of the scientific results from the analysis of analysis, deterministe computer analysis (utilizing a plant model the regional seismicity during the reporting period. Particular that includes the nuclear steam supply system, balance of plant emphasis has been placed on the Miramichi, New Brunswick systems and control systems), a review of plant occurrences, a earthquakes of January 1982. proteility analysis and a review of applicable Nuclear Regula-tory Ammission (NRC) cnteria pertaining to control systems.

NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH This study identified two gystem failures that could cause tran-PROGRAMS.Ouarterty Report, January-March 1985. EDLER,S.K- sients leading to a steam generator overfill and two system fail-Battelle Memorial Institute, Pacific Northwest Laboratories. ures that could lead to a reactor coolant cooldown of greater August 1985. 27pp. 8509100359. PNL-5516- 1. 32528:024 than 100 degrees fahrenheit per hour. It also identified two This document summarizes work performed by Pacific North- system failures that could lead to an overpressurization of low west Laboratory from January 1 through March 31,1985, for the temperatures and two steam generator tube rupture events that Division of Accident Evaluation and the Division of Engineering could be further aggravated by additional system failures. This Technology, U.S. Nuclear Regulatory Commission. PNL is oper. study concludod that the existing NRC criteria, conceming con-(ted for the U.S. Department of Energy by Battelle Memorial in, trol systems, adequa'ely addressed the potential problem areas stitute under Contract DE-AC06-76RLO 1830. Results from an that were identified during this evaluation. Based on the results instrumented fuel assembly irradiation program being performed of this study, it is recommended that the consequences and risk at Halden, Norway, are reported. Experimental data and analyti- associated with overfill and overcool transients be further inves-cal models are being provided to aid in decision making regard- tigated. It is also recommended that the piobabilities associated ing pipe-to-pipe impacts following postulated breaks in high- with the low temperature overpressurization and the steam gen-energy fluid system piping. Fuel assemblies and analytical sup- erator tube rupture sequences be evaluated by the NRC staff.

Port are being provided for experimental programs at the Power

• "*#8 0' '"Y "" W studies being pedormed on the effects of control system fail-Burst Facility, Idaho National Engineering Laboratory, Idaho ures to establish a position for resolution of Unresolved Safety Falls, Idaho. High. temperature materials property tests are issue A 47 (Safety implications of Control Systems).

being conducted to provide data on severe core damage fuel behavior. Thermal-hydraurc computer programs are providing NUREG/CR-4331: SIMPLIFIED SEISMIC PROBABILISTIC RISK best-estimate analyses for a variety of safety issues in light. ASSESSMENT. Procedures And Limitations. SHIEH LC.:

water reactors. Severe fuel damage tests are being conducted JOHNSON J.J.; WELLS,J E.; et al. Lawrence Uvermore National in the NRU Reactor, Chalk River, Canada. Laboratory. August 1985.198pp. 8508220290. UCID-20468.

32345:320.

24 Main C tations and Abstracts At the request of the U.S. Nuclear Regulatory Commission, categories of hypothesis: (A) P)pothces on the specific gaolog-the Lawrence Livermore National Laboratory has developed a ic structures that might cause large earthquakes in the Scuth-simphfied seismic probab;listic risk assessment (PRA) methodol- eastem Seaboard: (B) hypotheses on the seisrectectonic zones ogy. The purpose of tNs methodology is to reduce the costs in which large earthquakes might occur and (C) hypotheses on whde adequately performing seismic probabilistic risk assess- temporal vanations of seismicity in tt'e Southeastem Seaboard.

, ments of nuclear power plants. This report summanzes the de- Hypotheses that are representative of each category are sum-velopment of the samphfied methodology and explains guide- marized, and evidence for and against each hypothesis is given, unes for appiyirug the procedures. Tne development effort is part if such evidence is available. When data are interpreted in the of the scope of work of the Seismic Safety Margins Research ways that current'y seem to be the most straightforward, the hy.

Program (SSMRP). potheses that are supported by one kind of evidence are usual-ly opposed by another kind of evidence. ReacNng a consensu NUREG/CR-4333: SIE. GENEVIEVE FAULT ZONE. MISSOURI on the cause of the Charleston earthquake, and on thei-likel,s AND ILLINOIS. NELSON W.J.; LUMM D.K.1;linois, State of. July hood of such an eaequah anng at oh Waws of me Southea?'+m Seaboard, will therefore probably require the rec- l 1985.104pp.

The Ste. Genevieve8508090574.

Fault Zone 32103:255'a is major structural feature etion of what an% appear to M convary As of M whsch strikes NW-SE for about 190 km on the NE flank of the Ozark Dome. There is up to 900 m of vertical displacement on dence.

high angle normal and reverse faults in the fault zone. At both NUREG/CR-4350 V02: PROBABILISTIC RISK ASSESSMENT ends the Ste. Genevieve Fault Zone dies out into a monocline. COURSE DOCUMENTATION. Volume 2: Probability And Statis-Two periods of faulting occurred. The first was in late Middle tics For PRA Applications. IMAN,R.L; PRAIRIE,R.R. Sandia Na-Devonian time and the second from latest Mississippian through tional Laboratories. September 1985. 200pp. 8510040381.

j earty Pennsytvanian time, with possible minor post-Pennsylva. SAND 85-1495. 32858:029.

nian movement. No evidence was found to support the hypoth- This course is intended to provide the necessary probabilistic esis that the Ste. Genevieve Fault Zone is part of a north- aad statistical skills to perform a PRA. Fundamental background westward extension of the late Precambnan-early Cambrian information is reviewed, but the pnncipal purpose is to address Reelfoot Rift The magnetic and gravity anomalies cited in sup- specific techniques used in PRAs and to illustrate them with ap-port of 'he "St. Louis arm" of the Reelfoct Rift possibly reflect plications. Specific examples and problems are presented for deep crustal features undertying and older than the volcanic ter- most of the topics.

rain of the St. Francois Mountains (1.2 to 1.5 bilhon years old).

In regard to neotectonics no displacements of Ouaternary sedi- NUREG/CR-4352: SUGGESTED STATE REQUIREMENTS AND ments have been detected, but small earthquakes occur from CRITERIA FOR A LOW-LEVEL RADIOACTIVE WASTE DIS-time to time along the Ste. Genevieve Fault Zone. Many faults POSAL SITE REGULATORY PROGRAM. RATLIFF,R.A.;

in the zone appear capable of slipping under the current stress DORNSIFE,B.; AUTRY,V.; et al. Conference of Radiation Cort i regime of east-northeast to west-southwest horizontal compres- trol Program Directors, Inc. August 1983. 50pp. 8509060210.

sion. We conclude it'at the zone may continue to experience 32503:323.

small earth movements, but catastrophic quakes similar to Desenption of criteria and procedure for a state to follow in

! those at New Madrid in 1811-12 are unlikely. the development of a program to regulate a LLW disposal site.

NUREG/CR-4334: AN APPROACH TO THE OUANTIFICATION This would include identifying those portions of the NRC regula-OF SEISMIC MARGINS IN NUCLEAR POWER PLANTS. tions that should be matters of compatibility, identifying the vari-l BUDNITZ,R.J.; AMICO,P.J.; CORNELL.C.A.; et al. Lawrence ous expertise and disciplines that will be necessary to effective-Livermore National Laboratory. August 1985. 307pp. ly regulate a disposal site, identifying the resources necessary 8508260301, UCID-20444. 32369:001, for conducting a confirmatory monitonng program, and providing This report is the second report of the Expert Panel on Quan. suggestions in other areas which, based on experiences, would tification of Seismic Margins. The Panel's first report was enti. result in a more effective reg

• ary program.

tied, "NRC Seismic Design Margins Program Plan." The objec-tive of this report is to discuss progress to date in studying the NUREG/CR 4354: A STUDY OF SEISMICITY AND TECTONICS issue of quantification of seismic margins in nuclear power IN NEW ENGLANDFinal Report. EBEL.J.E. Boston College, plants. In particular, it deals with progress towards the establish- Chestnut Hill, MA. August 1985. 100pp. 8509100335.

ment of review guidelines that would be useful in studying how 32528:181.

much seismic margin exists. The guidelines themselves will be The operation by Weston Observatory of a seismic network in the subject of the next Panel report. The work presented in this New England frorn 1,974 to 1985 is desenbed, and the results of report is the result of a detailed study of seismic Probabilistic me seismic gui.w .g are summanzed. The netwo* cowage Risk Assesssments, historical earthquake performance of the of Weston Observatory increased from two operating stations in nuclear and non-nuclear facilities, and test data, augmented by 1974 to 36 stations in 1979 and was stabalized at 30 stations in the individual experience and expertise of the Panet members. me earh 1980's. The network was used to find the locations The major development discussed in this report is the HCLPF and magMudes of an eamquake activity detected during me concept, which demonstrates margin by showing that there is a study period. Most earmquakes from 1974 m 1985 were found High Confidence of a Low Probability of Failure for a given M occur in me same places as mose wNch have been h earthquake size. mented historically, although the activity appears to be random i both in space and time. Studies of aftershocks and detailed NUREG/CR-4339: A REVIEW OF RECENT RESEARCH ON THE monitoring in selected areas did not show any strong correta-SEISMOTECTONICS OF THE SOUTHEASTERN SEABOARD tions between the earthquake locations and mapped geologic AND AN EVALUATION OF HYPOTHESES ON THE SOURCE structures. It is concluded that the relationship among earth OF THE 1886 CHARLESTON. SOUTH CAROLINA EARTH- quakes, tetonic or structural zones and faults exposed on the

OUAKE. DEWEY,J.W. Interior Dept. of, Geological Survey. surface are not weit understood. The causes of the earthquake August 1985. 45pp. 8508290532. 32410

197, activity in the northeast are not clearfy established with the seis-In spite of extensive research on the source of the 1886 mic data which was gathered and analyzed.

, Charleston, S.C. Earthquake, there is not yet a consensus among earth scientists on the characteristics of the fault that NUREG/CR-4355 V01: 238 PU(IV) IN MONKEYS. Overview Of produced the earthquake or on the likelihood of future large Metabolism. DURBIN,P.W.; JEUNG N.; SCHMIDT.C.T. Lawrence earthquakes at other locatioms of the Southeastern Seaboard. Berkeley Laboratory. September 1985. 96pp. 8510020240. LBL-This report reviews the evidence from recent research on three 20022.32839:157.

i i

1

Main Citations and Abstracts 25 Complete balance studies were performed using 21 adult and equipment qualification procedures. This paper gives examples four adolescent Macaque monkeys (three species, both sexes) of the utility of density profiling for studying oxygen diffusion-lim-to define distnbution and retention of (238)Pu(IV) citrate from 2 ited degradation in both radiation and thermal aging environ-hr to 1100 d after parenteral injection. Experimental methods ments and in discovering / understanding chemical dose-rate ef-are desenbed in detail. Initial distnbution (6 adults, 7.5 plus fects in high energy radiation environments, minus 0.6 d) and retention (6 adults,711 plus minus 310 d) of Pu were, respectively, as follows: skeleton and teeth 28 and NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL 14% 10, liver 60 and 11% ID, other soft tissues 6 and 0.9% ID PURPOSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUA-and excrotion 5 and 74% ID. Initial Pu content of ovary and TION OF NUCLEAR REACTOR VESSELS AND PIPING COM-tests was 0.005 and 0.06% 10, respectively, and both declined PONENTS. GANAPATHY,S.; SCHMULT,B.; WU,W.S.; et al.

with T1/2 equivalent to 1 yr. Liver Pu was cleared mainly by ex- Michigan, Univ. of, Ann Artxir, MI. August 1985. 126pp.

cretion to feces, but also by recirculation, with an average T1/2 8509100506. 32536:106.

= 180 d. By 1100 d, most soft tissues lost 50 to 90% of their This report desenbes the design details of a special purpose initial Pu content. The irntial Pu concentration on trabecular system for real-time nondestructive evaluabon of reactor ves-bone surfaces in red marrow was calculated to be about 4.5 seis and piping components. The system consists of several times greater than on compact bone surfaces. About one-half of componen's and the report presents the results of the research initial skeletal Pu was eliminated in 1100 days, mainly from can-aicned a; the design of each component and recommendations cenous structures in red marrow. Implications for some changes in Internationi Commission on Radiological Protection metabohc based on the results. One major wnyg,6at of the NDE and dosametric models for Pu are noted. system, namely the real-time SAFT processor was designed with sufficient details to enable the fabricabon of a prototype by NUREG/CR-4357: THE FEASIBIUTY OF DETECTING THE GARD inc. under a subcontract from The Univers,ty of Michigan IMPORT OF UNAUTHORIZED RADIOACTIVE MATERIALS and the reptwt includes ttwr results and conclusions.

INTO THE UNITED STATES. BEE R.W.; GOFtDON,J.;

KWAN.O.; et at Aerospace Corp. September 1985. 235pp. NUREG/CR-4376 HEAT TRANSFER, CARRYOVER AND FALL l 8510040578. 32862:213. BACV IN PWR STEAM GENERATORS DURING TRANSIENTS.

l This report explores the feasibility of establishing and operat' LIAO, -H.; PArwOS,A.; GRIFFITH,P. Massachusetts Institute of ing a radiation monitoring system at U.S. borders to operate in Technology, Cambndge, MA. September 1985. 212pp.

! conjunction with normal U.S. Customs Service inspection prece- 8510020307. EPRI NP 4298. 32828:124.

dures to detect the presence of accidental radioactive contami- The cmcwn ovw Prehed Thwmal Shock (PTSL along

• **"Y 0 er concwns, indicates me need for accwate vertent contamination oa ctive ena ndus kW m. me steam generats beham. during me Now-al and commercial usage and explored how such materials might accidental?y enter new product manufacturing processes. down of the steam generator secondary side. To fulfill this need l Radiation inunnuung equipment necessary to detect such con- a computer program, SIT-SG (Simulator of Transient in Steam tamination was examined, and a technical system description Generator) is developed. This is a one-dimensional best.esti-and capital, operating, and maintenance costs were developed. mate code with the assumption that the vapor and liquid phases Hypothetical scenarios were developed to estimate a range of are in thermal e@ilibrium but not homogeneous. The drift flux costs for recall, cleanup, and disposal of contaminted products model is used to desenbe the relationship between the vapor wen after distnbution to consumers. Estimates were made of and the liquid phase velocity. No momemtum equation is re-added health risk to consumers associated with exposure to quired for SIT-SG because the detailed pressure distnbution in contaminant radiation if the contaminated products go undetect- the vessel is not important for the blowdown process. Based on ed. An economic analysis was performed to provide a common ao cefiv iis00s e between the code predictions and the data ob-i basis for their comparison. Because such aspects as frequency, tained from the expenments conducted in Battelle-Frankfurt and level of contamination, characteristics of poduct distribution, GE, the best flux model constants for various flow regimes are and health effects cannot be assessed using absolute bounding selected. SIT-SG has been used to predict the carryover, fall conditions, no direct wnverisons of courses of action can be back, and heat transfer for the MIT steam generator blowdown made. Rather, the study provides decision makers with various experiments. The results are encouraging. It is found that the dimensions of the problems, thus providing an improved basis measured dryout front is much higher than the calculated mix-for related decision making. ture level if the effective heat transfer are is dett Tnined from NUREG/CR-4358: APPUCATIONS OF DENSITY PROFluNG TO the mixture level, the primary-to-secondary heat teransfer will be EQUIPMENT OUALIFICATION ISSUES. GILLEN,K.T.; substantially underpredicted. From the result of the liquid hold CLOUGH.R.L; DHOOGE,N.J. Sandia National Laboratories. up study we would expect to find two mixture levels, one in the September 1985. 38pp. 8510040354. SAND 85-1557. bottom of the steam generator and one cbove 'he top tube sup-32861:012. port plate, provided that flooding occurs at all.

This paper reviews the density profiling technique, a new, in-expensive and versatile analytical method which can yield ex- NUREG/CR-4377: EVALUATIONS AND UTluZATIONS OF RISK tremely useful information on heterogenerbes in polymers. The IMPORTANCES. VESELY,W.E.; DAVIS.T.C. Battelle Memonal technique makes use of a dens:ty gradient column to measure Institute, Columbus Laboratories. August 1985. 131pp.

the density of a series of successivet/ cut slices across a 8509100355. BMI-2129. 32528:049.

sample. Since the density of very thin slices can easity be ob. This report presents approaches for utilizing Probabilistic Risk tained, density profiles across very small cross-sections Analysis (PRA) to determine risk importances. PRAs can be

(<1mm) are readily available. A major epplication of the tech- used to identify the importances of risk contributors or proposed nique involves oxidation studies of polymers, since oxidation re- changes to designs or operations. The objective of this report is actions usually lead to substantial increases in polymer density. to serve as a handbook and guide in evaluating and applying Diffusion-limited oxidation effects, which lead to heterogeneous- risk importances. The utilization of both qualitative risk impor-ly oxidized materials, are often present in polymer aging studies tances and quantitative risk importances is described in this in air. Since these effects are responsible for the commonly-ob. report. Qualitative risk importances are based on the logic served physical dose-rate effects in radiation aging environ- models in the PRA, while quantitative risk importances are ments and for non-Arrhenius behavior in thermal aging environ- based on the quantitative results of the PRA. Both types of im-ments, the availability of simple oxidation profiling techniques is portances are among the most robust and meaningfulinforma-a tremendous aid in validating the aging simulation aspects of tion a PRA can provide 1

26 Main Citations and Abstracts NUREG/CR-4379 V01: LONG TERM PERFORMANCE OF MATE- mission (NRC) with operational data that can be used in evalua-RIALS USED FOR HIGH-LEVEL WASTE PACKAGINGAst tion of plant designs for hquid and gaseous radwaste treatment Ouarterty Report, Year Four Apni-June 1985. STP _,0.; systems. Data presented were obtained at the Prairie Island N'i- .

MILLER,N.E. Battelle Memonal Instrute, Columbus 1 - rato- clear Generating Station, operated by Northern States Power, nes. September 1985.111pp. 8510020228. 32838.277. located near Red Wing Minnesota. In-plant measurements were High-leve! waste glass studies are being concluded and ef- conducted cunng the time penod from October 1980 to August forts are being directed towaro studying spent-fuel performance. 1981. This plant is the fifth in a senes of operating LWRs to be The effects of devitrification on glass teach rates are being in- studied.

vestgated, and salca dissolution was studied to provide data for the glass dissolution model. Prehminary data support this model. NUREG/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR A leach test using organic acids was conducted and leaching PART 50, APPENDIX J LEAX TESTS FOR PRIMARY AND I trends were observed. Real and simulated spent fuels are being SECONDARY CONTAINMENTS OF LIGHT WATER COOLED incorporated in integral tests using simulated groundwater in a NUCLEAR POWER PLANTS. SClACCA,F - NELSON W -

prototypic repository environment. The reactions of groundwater SIMPKINS,B.; et al. Science & Engineenng Associates, Inc$

species with steels are being analyzed to evaluate susceptabi. Se tember hty to petting and stress-corrosion cracking. Potental cracking 1985.

This report 92pp.the8510030131.

addresses 32849.332'en the existin differences betwe agents are being irivestigated by slow strain rata experiments. and proposed Appendix J and identfies eleven substantive eneral and @ng conosen models wwe furmer dweW areas where quantifiable impacts will likety result. The analysis based on known pnnciples of mass transport and radiolytic pro- indcated that there are four arcas of change which tend to duction. A simplified groundwater radiolysis model, developed iwe all others in farms of cost impacts The applicable for use with the corrosen models, was compared with other aragraph numbers from Draft E2 of the Appendix J revision mechanisms for species concentration predictions. and the nature of the change follows: Ill.A(4) & lit.A(6) - Test Pressure & Testing at Reduced Pressure No Lonaer Allowed, NUREG/CR-4337: IN-PLANT SOURCE TERM MEASUREMENTS til.A(7)(b)(;) Aaeptance Criteria 1.0 La Acceptable "As Found" AT PRAIRIE ISLAND NUCLEAR GENERATING STATION. Leakage; lil.A(8)(a) Retestng Following Failure of "As Found" MANDLER J.W.; STALKER,A.C.; CRONEY,S.T.; et al. EG&G Type A Test - Corrective Action Plan, and fl!.A(8Hb)(ii) Option to-Idaho, Inc. (subs. of EG&G, Inc.),. September 1985. 250pp. Do More Frequent Type B & C Testing Rather Than More Type 8510040592. EGG-2420. 32863:172. A Penalty Tests. The best estimate is that the proposed Appen-This report presents data obtained at Prairie Island as a part dix J would result in cost savings ranging from about $100 mil-of the in-Plant Source Term Measurement Program iri operating lion to $160 million, and increase routine occupational exposure light water reactors (LWRs). the work was conducted for the on the order of 10,000 person-rem. These estimates capture Office of Nuclear Regulatory Research (RES) en support of the the total impact to industry and tre NRC over the assumed op-Meteorology and Effluent Treatment Branch (METB) of the erating life of all existing and plar'ned future power reactors. All Office of Nuclear Reactor Regulation (NRR). The pnmary objec- dollar impacts projected to occur in future years have been tive of this program is to provide the Nuclear Regulatory Com- present worthed at discount rates ranging from 5% to 10%.

Contractor Report Number Index l This index lists, in alphabetical order, the NUREG/CR for the report and to the 10-contractor-issued report codes for the NRC digit NRC Document Control System acces-contractor reports in this compilation. Each sion number.

contractor code is cross-referenced to the SECONDARY REPORT NUMBER REPORT NUM8ER SECONDARY REPORT NUMSER REPORT NUMBER 85-ETEC-DRF-171 NUREG/CR-4294 ORNUNSC200 NUREG/CR-2000 V04 N6 i ANL-84-16 NUREG/CR-3710 ORNUNSC200 NUREG/CR-2000 V04 N7 l ANL-8441 V04 NUREG/CR-3980 V04 ORNUNSIC-200 NUREG/CR-2000 V04 N8 ANL45-23 V01 NUREG/CR-4240 VOI ORNUTM4869 NUREG/CR-3442 ANL45-33 NUREG/CR-4287 ORNUTM-9191 NUREG/CR 3851 V04 BMI-2120 NUREG/CR-4082 V02 ORNUTM-9267/V4 NUREG/CR-3885 V04 BMI-2127 NUREG/CR-3900 V04 BMI-2129 O CR NUREG/CR-4377 ORNL/TM-9437 BNL-NUREG-51267 NUREG/CH-1677 V02 NUREG/CR-4037 BNL-NUREG-51559 ORNUTM-9488 NUREG/CH-4061 NUREG/CR2815 V01 R1 ORNUTM-9593/V1 NUREG/CR-4219 V01 BNL-NUREG-51559 NUREG/CR-2815 V02 R1 ORNUTM-9632/V1 NUREG/CR-4255 V01 BNL-NUREG-51630 NUREG/CR 3091 V06 ORNUTM-9654 NUREG/CR-4275 BNL-NUREG-51699 NUREG/CR-3444 V02 ORNUTM-9660 NUREG/CR-4280 BNL-NUREG-51710 NUREG/CR-3485 ORNUTM-9664 NUREG/CR-4284 BNL-NUREG-51795 NUREG/CR-3876 ORNUTM-9682 NUREG/CR4325 BNL-NUREG-51888 NUREG/CR-4254 PNL-4790 NUREG/CR-3413 BNL-NUREG-51893 NUREG/CR4182 PNL-4941 NUREG/CR-3613 V03 N1 BNL-NOREG-51897 NUREG/CR-4291 PNL-4942 NUREG/CR-3609 BNL-NUREG-51901 NUREG/CP-0066 PNL-5156 NUREG/CR-4268 BNL-NUREG-51907 NUREG/CR-4143 PNL-5184 NUREG/CR-3915 CREARE TN-384 NUREG/CR-3426 V01 DNL-5374 NUREG/CR-4125 V01 CREAVE TN-384 NUREG/CR-3426 V02 PNL-5374 NUREG/CR-4125 V02 EGG-2259 NUREG/CR-3301 PNL-5379 NUREG/CR-4130 EGG-2294 NUREG/CR-3633 V04 PNL-5392 NUREG/CR-4151 EGG-2294 EGG-2297 NUREG/CR-3633 V01 S1 NUREG/CR-3948 k$

PNL-5461 hk/ $ V02 NUREG/CR-4251 V01 E 3G-2317 NUREG/CR-3819 EGG-2335 PNL-5467 NUR~G/CR-4259 NUREG/CR-3935 PNL-5487 NUREG/CR-4281 EGG-2366 NUREG/CR-4080 PNL 5490 NUREG/CR-4298 EGG-2376 NUREG/CR-4150 PNL-5511 NUREG/CR-4300 V01 EGG-2388 NUREG/CR4227 PNL-5516-1 NUREG/CR-4318 V01 EGG-2397 NUREG/CR-4272 SAND 83-2675 NUREG/CR-4213 EGG-2405 NUREG/CR-4326 V01 SAND 63-7450 NUREG/CR-3537 EGG-2420 NUREG/CR-4397 SAND 84 0060 NUREG/CR-3638 EPRI NP-3802 NUREG/CR-3426 V02 SAND 84-1072 NUREG/CR-3816 V03 EPRI NP-3802 NUREG/CR-3420 V01 SAND 84-1072 NUREG/CR-3816 V04 EPRI NP-4298 NUREG/CR-4376 SAND 84-1204 NUREG/CR-4065 GP-R-123022 NUREG/CR-4258 SAND 841367 NUREG/CR-4060 HEDL-TME45 3 NUREG/CR-3319 SAND 85-0016 NUREG/CR-4119 IE-145 NUREG/CR-4006 SAND 85 0044 NUREG/CR-4122 IEB-8141 NUREG/CR-4006 SAND 85 0135 NUREG/CR4138 IS4862 NUREG/CR-3952 SAND 854175 NUREG/CR-4137 LA-10055-MS NUREG/CR 3706

^

LA-10319-MS SA 09 4 NUREG/CR-4107 SAND 851495 LA-10396-MS NUREG/CR-4217 NUREG/CR-4350 V02 LA 10413-MS SAND 85-1557 NUREG/CR-4358 NUREG/CR-4232 SAND 85-7185 NUREG/CR-4214 LA 10435-M NUREG/CR-4260 SAND 65-7192 NUREG/CR-4303 LA-10474-MS NUREG/CR-4314 UCID 19988 NUREG/CR-3660 V01 LBL-20022 NUREG/CR4355 V01 UCID-19988 V04 NUREG/CR-3660 V04 NBSIR 85-3169 NUREG/CR-4266 UCID-20398 NUREG/CR-4239 ORNL4114 NUREG/CR-4038 UCID 20444 NUREG/CR-4334 ORNL4163 NUREG/CR-4249 UCID-20468 NUREG/CR-4331 ORNL4170/V1 NUREG/CR-4234 V01 UCRL-53644 NUREG/CR-4290 V02 ORNL4177 NUREG/CR-4304 WWUTM-1791-2 NUREG/CR-3901 27

l 4

Personal Author incia This index lists the personal authors of NRC report (s) prepared by the author. If informa-staff and contractor reports in abhabetical tion is needed, refer to the main citation by order. Each name is followec. by the the NUREG number.

NUREG number and the title of the ADAMS,K G. BALDWIN C.A.

NUREG/CR4250 VEHICLE BARRIERS. EMPHASIS ON NATURAL FEA- NUREC/CR-4284: NEUTRON EXPOSURE PARAMETERS FOR THE TURES. FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATION SERIES.

ADAMS R.E.

NUREG/CR-4255 V01: AEROSAL RELEASE AND TRANSPORT FRO. BALL.D.G.

GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 NUREG/CR-4249- PRESSURE VESSEL FRACTURE STUDIES PENE-MARCH 1985. TRATING TO THE PWR THERMAL-SHOCK ISSUE. EXPERIMENTS TSE-5.TSE 5A AND TSE-8.

AHMAD,J. NUREG/CR-4304: PRESSURE VESSEL FRACTURE STUDIES PER-TAINING TO THE PWR THERMAL-SHOCK ISSUE. Experiment TSE-7.

NUREG/CR4082 V02: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report. October 1984 March 1985. NUREG/CR-4325: A PARAMETRIC STUDY OF PWR PRESSURE VESSEL INTEGRITY DURING OVERCOOUNG ACCIDENTS,CONSIDERING BOTH 2-D AND 3-D FLAWS.

AHMED S.

NUREG/CR-4257: INSPECTION. SURVEILLANCE.AND MONITORING BALL.SJ-OF ELECTRICAL EQUIPMENT INSIDE CONTAINMENT OF NUCLEAR NUREG/CR-3885 V04: HIGH-TEMPERATURE GASCOOLED REACTOR POWER PLANTS-WITH APPUCATIONS TO ELECTRICAL CABLES. SAFETY STUDIES FOR THE DIVISION OF ACCIDENT AKERS,D.W.

EVALUATON.Ouarterty Progress Report, October 1-December 31,1984.

NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT PRAIRIE ISLAND NUCLEAR GENERATING STATON- SARI,R.A.

NUREG/CR-2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-REG -4081: ABSORPTON OF GASEOUS LODINE BY WATER DROPLETS. BARNES,C.R.

NUREG/CR-4082 V02: DEGRADED PlPING PROGRAM - PHASE ALPERT,DJ.

RSeW RW October 1984 - March 1985.

NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY PROTECTON.flL SUBMICRON PARTICLE TESTS AND

SUMMARY

BARRETT,R.

OF QUAUTY FACTORS. NUREG/CR4143: REVIEW AND EVALUATON OF THE MILLSTONE NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER UNIT 3 PROBABluSTIC SAFETY STUDY. Containment Failure PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part Modes. Radiological Source. Terms And Offsite Consequences.

1:Introdelntegration & Summary Part II. Scientific Basis For Health Effects Models. BASHAM,P.W.

NUREG/CR-4717 V01: CANADIAN SEISMIC AGREEMENT.Techrucal AMICO,PJ. Report Coviding 1979-1985.

NUREG/CR-4334:AN APPROACH TO THE QUANTIFICATION OF SEIS-MIC MARGINS IN NUCLEAR POWER PLANTS. BAUM .W.

ANDRAE,R.W. SHOP ON H:STORIC DOSE EXPERIENCE AND DOSE REDUCTON NUREG/CR-4260 TORAC USER'S MANUALA Computer Code For Ana- (ALARA) AT NUCLEAR POWER PLANTS,MAY 294UNE 1.1984.

tyzing Tornado-induced Flow And Matenal Transport in Nuclear Facili- NUREG/CR4254: OCCUPATONAL DOSE REDUCTON AND ALARA g,es. AT NUCLEAR POWER PLANTS. Study On High-Dose Jobs.Radweste Handling.And ALARA Incentives.

ANDREWS,W.B.

NUREG/CR-2800 S03. GUIDEUNES FOR NUCLEAR POWER PLANT NU G/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART SAFETY ISSUE PRIORITIZATON INFORMATION DEVELOPMENT.

50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY ARMANTROUT,G.A. CONTAINMENTS OF LIGHT WATER-COOLED. NUCLEAR POWER PLANTS.

NUREG/CR-4239- ANALYSIS OF THE ABluTY OF CURRENT HEALTH PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID- BEARE,A.N.

UALS. NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW ARNOLD W.D NUREd/CR53851 V04: EVALUATON OF RADIONUCLIDE GEOCHEMI-E IN MM M SW.AMS.

CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR BECKMAN,RJ. I WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For NUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER October 1983-September 1984- PLANT VALVE FAILURE-RATE VARIABluTY-SOME PRELIMINARY ATTERIDGE.D.G.

NUREG/CR-3813 V03 N1: EVALUATION OF WELDED AND REPAIR- BEE,R.W.

WEl.DED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report NUREG/CR-4357: THE FEASIBluTY OF DETECTING THE IMPORT OF For October 1984 Through March 1985. UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED STATES.

AUTRY,V.

NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE- BEEBE.M.R.

RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE NUREG-0020 V09 N08: UCENSED OPERATING REACTORS STATUS REGULATORY PROGRAM.

SUMMARY

REPORT. Data As Of May 31,1985.(Gray Book I) 29

30 Personal Author Index NUREG.0020 V09 N07: u FO OPERATING REACTORS STATUS SYERS,K.R.

SUMMARY

REPORT. Dab 4  % 30,1985.(Gray Book I) NUREG/CR-4298. DESIGN AND INSTALLATION OF COMPUTER SYS-TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

BE M M NUREG-1142- TECHNICA "

• CATONS FOR RIVER BEND CAMP 8 ELL,R.D.

STATION. Docket No. 504 a States Utilities Company) NUREG/CR-4290 V02: PROBABluTY OF PIPE FAILURE IN THE REAC-BERGERON,K.D TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR NUREG/CR-4585: USERS MANUAL FOR CONTAIN 1.0.A Computer PLANTS. Volume 2:Gudiotine Break Indirectly induced By Earthquakes.

Code for Severe Reactor Acodent Containment Analyms.

CARFAGNO.S.P.

SEZLER P. NUREG/CR-4257: INSPECTON. SURVEILLANCE AND MONITORING NUREG/CR-1677 V02: PIPING BENCHMARK PROBLEMS, VOLUME 11 OF ELECTRICAL EQUIPMENT INSIDE CONTAINMENT OF NUCLEAR DYNAMIC ANALYSIS INDEPENDENT SUPPOHT MOTION RE. POWER PLANTS-WITH APPUCATIONS TO ELECTRICAL CABLES.

SPONSE SPECTRUM METHOD.

NUREG/CR4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSt. CASHMAN,T.

CAL BENCHMARKING OF PIPING SYSTEMS. NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE-RlA FOR A LOW LEVEL RADCACTIVE WASTE DISPOSAL SITE SICKFORD,W.E. REGULATORY PROGRAM.

NUREG/CR-2800 S03: GUIDEUNES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATON INFORMATON DEVELOPMENT. CATHEY,N.G.

NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE-RE /CR4397: IN-PLANT SOURCE TERM MEASUREMENTS AT PRAIRIE ISLAND NUCLEAR GENERATING STATON. CHANG,T.Y.

NUREG-1030 DRFT: SEISMIC OUALIFICATON OF EQUIPMENT IN OP.

ERATING NUCLEAR POWER PLANTS. Unresolved Safety issue A-NR /CR 288. FOCAL MECHANISM ANALYSES FOR VIRGINIA 48. Draft Report For Comment.

AND EASTERN TENNESSEE EARTHOUAKES (1am.1984).

BLACKMAN H.S. CHARLOT,L.A.

NURFG/CR-4272- RESPONSE TREE EVALUATON EXPERIMENTAL NUREG/CR-3613 V03 N1: EVALUATION OF WELDED AND REPAIR.

ASSESSMENT GF AN EXPERT SYSTEM FOR NUCLEAR REACTOR WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report OPERATORS. For October 1984 Through March 1985.

BLUHM,D. CHEN,J.C.

NUREG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE. NUREG/CR-4331: SIMPLIFIED SEISMIC PROBABILISTIC RISK ASSESSMENT. Procedures And Limitations.

BOAADMAN.T.

NUREG/CR4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE CHEVERTON R.D.

REACTOR COOLANT PUMP. NUREG/CR-4249: PRESSURE VESSEL FRACTURE STUDIES PENE-TRATING TO THE PWR THERMAL SHOCK ISSUE. EXPERIMENTS NE /CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL NU G -4 E RE VESSEL FRACTURE STUDIES PER-ACCIDENTS. METHODS, COSTS AND SCHEDULES FOR DECON* TAINING TO THE PWR THERMAL. SHOCK ISSUE.Expenment TSE-7.

TAMINATION.

NUREG/CR4325: A PARAMETRIC STUDY OF PWR PRESSURE 80LT,S.E. VESSEL INTEGRITY DURING OVERCOOUNG NUREG/CR4249: PRESSURE VESSEL FRACTURE STUDIES PENE- ACCIDENTS.CONSIDERING BOTH 2-D AND 3-D FLAWS.

TRATING TO THE PWR THERMAL-SHOCK ISSUE: EXPERIMENTS NUR C -4304 RE VESSEL FRACTURE STUDIES PER. EG[CR-3660 V04: PROBABluTY OF PIPE FAILURE IN REACTOR TAINING TO THE PWR THERMAL-SHOCK ISSUE.Expenment TSE-7. COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4. Pipe Fadure Induced By Crack Growth in West Coast Ptants.

BOWERMAN.B.S.

NUREG/CR-0444 V02: THE IMPACT OF LWR DECONTAMINATIONS CHO N.Z.

ON SOUDIFICATION, WASTE DISPOSAL AND ASSOCIATED OCCU- NUREG/CR-2615 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-PATIONAL EXPOSURE. DURES GUOE. Sections 17 And Appendices.

NUREG/CR-3485: PRA REVIEW MANUAL 80WERS,D.I..

NUREG/CR-3710 LABORATORY STUDIES OF A BREACHED NUCLE. CHOU.C.K.

AR WASTE REPOSITORY IN BASALT. NUREG/CR 3660 V01: PROBABlUTY OF PIPE FAILURE IN THE REAC-TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume N RE /CR4082 V02: DEGRADED PIPING PROGRAM - PHASE ll. Semiannual Report, October 1984 - March 1985. CHRISTENSEN.D.

NUREG/CR-3145 V03: GEOPHYSICAL INVESTIGATIONS OF THE NURE 13 V03 N1: EVALUATION OF WELDED AND REPAIR-

" ^ *

• 82 Septeh M83, Wurne 3L WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report For October 1984 Through March 1985. CHUNG H.M.

BRUSKE.SJ NUREG/CR-3980 V04: UGHT WATER-REACTOR SAFETY FUEL SYS-NUREG/CN-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES TEMS RESEARCH PROGRAMS. Quarterty Progress Report,0ctober.

ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE December 1984.

PRESSURIZED WATER REACTORMain Report CLAUSERMJ.

BRUST.F. NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Computer NUREG/CR4082 V02- DEGRADED PIPING PROGRAM - PHASE Code for Severe Reacter Accident Contamment Analysis.

II. Semiannual Report October 1984 - March 1985.

CLAUS $,D.B.

BUDNITZ,R.J. NUREG/CR-4137: PRETEST PREDICTIONS FOR THE RESPOf4SE OF l NUREG/CR-4334: AN APPROACH TO THE QUANTIFICATION OF SEIS. A 1:8-SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO MIC MARGINS IN NUCLEAR POWER PLANTS. STATIC OVERPRESSURIZATION.

BUSLIK,A.J. CLEVELAND,J.C.

NUREG/CR-2815 V01 RI: PROBABILISTIC SAFETY ANALYSIS PROCE. NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR DURES GUIDE. Sections 17 And Appendices. SAFETY STUDIES FOR THE DIVISION OF ACCIDENT

Personal Author Index 31 EVALUATION.Quarterty Progress Report, October 1-December DAVIS.LT.

31,1984. NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU.

CLEAR POWER PLANT CONTROL ROOM CREWS.

NUREG/CR4358: APPUCATONS OF DENSITY PROFILING TO EQUIP- DAVIS,M.S.

MENT QUAUFICATION ISSUES. NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS ON SOUDIFICATION, WASTE DISPOSAL AND ASSOCIATED OCCU-COHEH,L PATIONAL EXPOSURE.

NUREG.0837 V04 N04: NRC TLD DIRECT RADIATON MONITORING REPORT ess Report. October-December 1984. DAVIS'T C.

NUR 4 83 5 N01 TLD DIRECT RADIATON MONITORING NUREG4837 Vk N02- NRCLD ECT RADIATION MONITORING PORTANCES.

NETWORK. Progress Report, Apr;l. June 1985. DAWSON,J.F.

COMEN,S. NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM NUREG/CR-4398: COST ANALYSIS OF REVISONS TO 10 CFR PART TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL SC. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY CONTAINMENTS OF UGHT-WATER-COOLED NUCLEAR POWER DEAN,R.S.

PLANTS. NUREG/CR-4006: CLOSEOUT OF IE BULLETIN 81-01. SURVEILLANCE OF MECHANICAL SNUBBERS.

COLLINS,J.L NUREG/CR4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT DEEDS,W.E.

RELEASE TEST HI-5. NUREG/CR-3949 V02: EDDY CURRENT INSPECTON FOR STEAM GENERATOR TUBING PROGRAM. Annual Progress Report For Period G-0933 S03: A PRIORITIZATON OF GENERIC SAFETY ISSUES.

DENNEHY,T.G.

NURFG/CR4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-E $3537: EXPEDIENT METHODS OF RESPIRATORY POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU-PROTECTIONII. SUBMICRON PARTICLE TESTS AND

SUMMARY

CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

OF QUALITY FACTORS.

NUREG/CR4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER Integr tion & Part I tific Ba s Ha NUR G -3819: SURVEY OF AGED NWER PLANT FACluTIES.

cts Ms. DEWEY,J.W.

CORNELL,C.A. NUREG/CR4339: A REVIEW OF RECENT RESEARCH ON THE SEIS.

NUREG/CR4334:AN APPROACH TO THE QUANTIFICATION OF SEIS- MOTECTONICS OF THE SOUTHEASTERN SEABOARD AND AN MIC MARGINS IN NUCLEAR POWER PLANTS. EVALUATION OF HYPOTHESES ON THE SOURCE OF THE 1886 CHARLESTON. SOUTH CAROUNA EARTHOUAKE.

CORNWELL.B.C.

NUREG/CR-3819- SURVEY OF AGED POWER PLANT FACluTIES. DHOOGE,N.J.

NUREG/CR-4358: APPUCATIONS OF DENSITY PROFluNG TO EQUIP-NUR CR 182: VERIFICATION OF SOIL STRUCTURE INTERACTION METHODS. DIAMENT,H.

NUREG/CR.3442: RADTWO.A COMPUTER CODE FOR SIMULATING COUNTS.C.A. FAST-TRANSIENT, TWO-DIMENS!ONAL,TWO-LAYER RADIONU-NUREG/CR-2800 S03: GUIDEUNES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATON INFORMATION DEVELOPMENT. RIVE S TUARIES,A D TAL REGIONS KES.RESERV CRONEY,S.T.

NUREG/CR 4397: IN-PLANT SOURCE TERM MEASUREMENTS AT DICK C.E.

PRAIR:E ISLAND NUCLEAR GENERATING STATION. NUREG/CR4266: STANDARD BETA-PARTICLE AND MONOENERGE-TIC ELECTRON SOURCES FOR THE CAUBRATION OF BETA-RADI-CUMMINGS,F.M. ATION PROTECTON INSTRUMENTATION.

NUREG/CR-3609: EVALUATON OF NEUTRON DOSIMETRY TECH-N10UES FOR WELL-LOGGING OPERATIONS. DIONNE,B.J.

NUREG/CP-0066: PROCEEDINGS OF AN INTERNATIONAL WORK.

CURRIE J.W. SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTON NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL (ALARA) AT NUCLEAR POWER PLANTS,MAY 29-JUNE 1,1984.

ACCIDENTS. METHODS, COSTS AND SCHEDULES FOR DECON- NUREG/CR-4254: OCCUPATIONAL DOSE REDUCTION AND ALARA TAMINATION. AT NUCLEAR POWER PLANTS Study On High-Dose Jobs,Radwaste Handling.And ALARA Incentrves.

CUTSHALL,N.H.

NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI- DOCTOR,P CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NUREG/CP4063. PROCEEDINGS OF THE 1984 STATISTICAL SYMPO-WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For SIUM ON NATIONAL ENERGY lSSUES.

October 1983-September 1984.

DODD C.V.

NU /CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM NUREGICR-3949 V02: EDDY. CURRENT INSPECTION FOR STEAM l GENERATOR TUBING PROGRAM. Annual Progress Report For Penod TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL Ending December 31,1984.

DAVIS,C.B.

DODSON,M.E.

NUREG/CR-393f: THERMAL-HYDRAULIC ANALYSES OF OVERCOOL.

ING SEQUENCES FOR THE H.B. ROBINSON UNIT 2 PRESSURIZED NUREG/CR-4259: TAluNGS NEUTRAUZATION AND OTHER ALTER-THERMAL SHOCK STUDY. NATIVES FOR IMMOBILIZING TOXIC MATERIALS IN TAlUNGS Final NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES Report.

ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE 4 PRESSURIZED WATER REACTOR. Main Report. DOLAN.F.X. I NUREG/CR-3426 V01: THERMAL AND FLUID MIXING IN 1/2 SCALE DAVIS.LA. TEST FACILITY. Facility And Test Desen Report.

NUREG/CR-3901: DOCUMENTATION AND USER'S GUIDE:GS2 & GS3 NUREG/CR-3426 V02: THERMAL AND FLUID MIXING IN 1/2 SCALE VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS. TEST FACluTY. Data ReporL

32 Personal Author index DONOVAN.M.D. FLETCHER,C.D.

NUREG/CR-4280- THE EFFECTS OF SUPERVISOR EXPERIENCE AND NUREG/CR-3935: THERMAL-HYDRAULIC ANALYSES OF OVERCOOL-ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW ING SEQUENCES FOR THE H B. ROBINSON UNIT 2 PRESSURIZED PERFORMANCE IN CONTROL ROOM SIMULATORS. THERMAL SHOCK STUDY. l DORNSIFE,8. FLUCKlGER,J.D.

NUREG/CR-4352: SUGGESTED STATE REQUIREMENTS AND CRITE- NUREG/CR-4298: DESIGN AND INSTALLATON OF COMPUTER SYS-RfA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

REGULATORY PROGRAM.

FOLEY,W.J.

N  : SENSITIVITY AND UNCERTAINTY STUDIES OF THE OF M CHAN SNUB S CRAC2 COMPUTER CODE. l DUROIN,P.W FRAGOLA.J-.

NUREG/CR 2815 V01 R1: PROBABILISTIC SAFETY ANALYSIS PROCE-NUREG/Clk-4355 VOI: 238 PU(IV) IN MONKEYS.Ovennew Of Metabo-DURES GUIDE. Sections 1-7 And Appendees.

EASLEY,P. GADDY,C.D.

NUREG/CR-4143: REVIEW AND EVALUATION OF THE MILLSTONE NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU-UNIT 3 PROBAtilllSTIC SAFETY STUDY.Contamment Fadure CLEAR POWER PLANT CONTROL ROOM CREWS.

Modes. Radiological Source Terms And Offsite Consequences.

GALLUCCl,R.H.

EBEL J.E. NUREG/CR-2800 S03: GUIDELINES FOR NUCLEAR POWER PLANT NUREG/CR4354: A STUDY OF SEISMICITY AND TECTONICS IN NEW SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.

ENGLAND. Final Report E8 ERHARDT,LL NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR.

NUREG/CR4268: RATO METHODS FOR COST-EFFECTIVE FIELD POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU.

SAMPUNG OF COMMERCIAL RADIOACTIVE LOW-LEVEL WASTES. CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

EDLER,S.K. GERDING,T.J.

NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH NUREG/CR-3710 LABORATORY STUDIES OF A BREACHED NUCLE-PROGRAMS.Quarterty ReportJanuary-March 1985. AR WASTE REPOSITORY IN BASALT.

EDSON,J.L GILLEN,K.T.

NUREG/CR 4080: DETERMINATON OF THE AVAILABluTY OF CORE NUREG/CR-4358: APPLICATONS OF DENSITY PROFILING TO EQUIP-EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATIONS. MENT QUALIFICATON ISSUES.

EHRLICH,M.

GILMORE,W.E NUREG/CR4266; STANDARD BETA-PARTICLE AND MONOENERGE*

NUREG/CR 4227: HUMAN ENGINEERING GUIDELINES FOR THE TIC ELECTRON SOURCES FOR THE CAllBRATION OF BETA-RADI-EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS.

ATION PROTECTON INSTRUMENTATON.

EISSENGERG,D.M. GODFREY,D.

NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC NUREG/CR4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART 50,APPENDfX J. LEAK TESTS FOR PRIMARY AND SECONDARY MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY FEA-TURE SYSTEMS OF NUCLEAR POWER PLANTS. CONTAINMENTS OF UGHT-WATER-COOLED NUCLEAR POWER PLANTS.

, EL-8ASSIONI,A.

NUREG/CR-2815 V01 R1: PROBABILISTIC SAFETY ANALYSIS PROCE- GOLDIN.D.

DURES GUIDE. Sections 17 And Appendices. NUREG/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART NUREG/CR-3485. PRA REVIEW MANUAL 50. APPENDIX J. LEAK TESTS FOR PR! MARY AND SECONDARY CONTAINMENTS OF LIGHT. WATER-COOLED NUCLEAR POWER ELLINGWOOD,8. ptANys, NUREG/CR-3876: PROBABluTY BASED LOAD COMBINATION CRITE-RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. GOLDMAN,A.S.

NUREG/CR-4107: SEQUENTIAL TEST PROCEDURES FOR DETECT.

U 5-0933 S03: A PRIORITIZATION OF GENERIC SAFETY ISSUES.

ENDRES,G.W N SO E L GIN OPERAT NS ^ ^ ^ ^ " "

STATES.

ERASLAN,A.H.

MAY,LH.

NUREG/CR-3442: RADTWO A COMPUTER CODE FOR SIMULATING FAST. TRANSIENT. TWO-DIMENSIONAL,TWO-LAYER RADIONU- NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERIENCE #AD CUDE CONCENTRATON CONDITIONS IN ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW LAKES RESERVOIRS. RIVERS. ESTUARIES,AND COASTAL REGIONS. PERFORMANCE IN CONTFIOL ROOM SIMULATORS.

EVANS,J.S. GREENSTREET,W.

NUREG/CR 4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER NUREG/CFI-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY-FEA-1:IntroductionIntegration & Summary.Part II:Soentific Basis For Health TURE SYSTEMS OF NUCLEAR POWER PLANTS.

GREGORY,W.S.

FALETTI,D.W. NUREG/CR4232. THE RESPONSE OF VENTILATION DAMPERS TO NUREG/CR-4151: INTEGRATON OF EMERGENCY ACTION LEVELS LARGE AIRFLOW PULSES.

WITH COMBUSTON ENGINEERING EMERGENCY OPERATING NUREG/CR-4260: TORAC USER'S MANUALA Computer Code For Ana-PROCEDURES By Use of Combustion Engineering Owners Group lyzing Tornado-induced Flow And Matenal Transport in Nuclear Facili-Emergency Operating Procedure Technical Guidelines. ties.

FANOUS,F. GREIMANN,L NUREG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE. NUREG/CR-3952 SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE.

l i

1

Personal Author index 33 GRIFFITU,P. HORAN,J.R.

NUREG/CR4376 HEAT TRANSFER. CARRYOVER AND FALL BACK IN )

NUREG/CP-0068: PROCEEDINGS OF AN INTERNATONAL WORK- i PWR STEAM GENERATORS DURING TRANSIENTS. SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTON 1 GRONEMYER,L (ALARA) AT NUCLEAR POWER PLANTS.MAY 294UNE 1,1984.

NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE- HOSKER,R.P. '

RlA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE NUREG/CR-4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE I REGULATORY PROGRAM.

CRAC2 COMPUTER CODE.

HAGGARD.D.L HUTTON,P.H.

NUREG/CR-3609- EVALUATON OF NEUTRON DOS 1 METRY TECH-NIQUES FOR WELL-LOGGING OPERATONS. NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL HALL,R.E. NUREG/CR4300 V01: ACOUSTIC EMISSON/ FLAW RELATONSHIP NUREG/CR-2815 V01 R1: PROBABluSTO SAFETY ANALYSIS PROCE- FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE DURES GUfDE.Sectons 17 And Appendices. VESSELS Progress Report October-March 1985.

HWANG,H.

N R G CR-4334:AN APPROACH TO THE QUANTIFICATION OF SEIS- NUREG/N3876: NBABW BASED M WBWM N-MIC MARGINS IN NUCLEAR POWER PLANTS. b ^

HANAN,N. ILDERG,D.

NUREG/CR-3485: PRA REVIEW MANUAL NUREG/CR-2815 V01 R1: PROBABiUSTIC SAFETY ANALYSIS PROCE-DURES GUIDE.Sectons 1-7 And Appendees.

HARRER,B.J.

NUREG/CR-3413 OFF-SITE CONSEQUENCES OF RADIOLOGICAL WANAL ACCIDENTSMETHODS, COSTS AND SCHEDULES FOR DECON- NUREG/CR4122 A FORTRAN 77 PROGRAM AND USER'S GUIDE TAMINATION' FOR THE CALCULATION OF PARTIAL CORRELATON AND STAND-ARDIZED REGRESSION COEFFICIENTS.

HARRINGTONAM.

NUREG/CR-4350 V02: PROBASIUSTIC RISK ASSESSMENT COURSE NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR DOCUMENTATION.Volurne 2: Probatxlity And Statistics For PRA Appli-SAFETY STUDIES FOR THE DIVISION OF ACCIDEFC catens.

EV TON.Quarterty Progress Report. October 1-December HARRIS,P.A. NUREG/CR-3706: TRAC ANALYSES OF SEVERE OVERCOOUNG TRANSIENTS FOR THE OCONEE 1 PWR.

NUREG/CR4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS SAFETY ANALYSIS.Ptass 1, Fr' .al Report. ISKANDER,5.K.

HARRISON.B.D- NUREG/CR-4249: PRESSURE VESSEL FRACTURE STUDIES PENE-TRATING TO THE PWR THERMAL-SHOCK ISSUE. EXPERIMENTS NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer TSE-5,TSE-5A AND TSE-8.

Code for Sev6te Reactor Accident Containment Analysis.

NUREG/CR-4304: PRESSURE VESSEL FRACTURE STUDIES PER-TAINING TO THE PWR THERMAL-SHOCK ISSUE.Expenment TSE-7.

NUREG/CR-1677 V02- PIPING BENCHhaARK PROBLEMS, VOLUME 11 JACKSON,D.H.

DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION RE- NUREG/CR-4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR SPONSE SPECTRUM METHOD.

POWSR PLANT MAINTENANCE FACTORS AND PLANT SAFETY.

HAYMAN,R.5-JACOBS,G.K.

NUREG/CR4317 V01: CANADIAN SEISMIC AGREEMENT.Techrucal Report Covenng 1979 1985. NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI-CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For NUREG S03: GUIDELINES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRORITIZATION INFORMATON DEVELOPMENT. JAMtSON J.D.

HEATON,H.T NUREG/CR-4151: INTEGRATION OF EMERGENCY ACTON LEVELS NUREG/Clk4266. STANDARD BETA-PARTICLE AND MONOENERGE- WRH CNBUSMN ENGWEERWG EMEMEC MMG TIC ELECTRON SOURCES FOR THE CAUBRATION OF BETA-RADI- PROCEDURES.By Use of Cornbustion Engineenne Owners Group ATON PROTECTON INSTRUMENTATION. gency Operaq Wocedwe TechNcal Wnes.

HEBOON,FJ. JANG,J.

NUREG 1022 S02 LICENSEE EVENT REPORT SYSTEM. Evaluation Of NUREG-0837 V04 N04: NRC TLD DIRECT RADIATON MONITORING First Year Results And Rm.v.o...s atons For improvements.

N REG NO C D AD TON MONITORING HENNICK.A. NETWORK. Progress Report, January-March 1985.

NUREG/CR4008: CLOSEOUT OF IE BULLETIN 8101:SURVEluANCE NUREG-0837 VOS N02: NRC TLD DIRECT RADIATON MONITORING OF MECHANICAL SNUBBERS. NETWORK. Progress Report, Apnl4une 1985.

HfCKS.8.5. JANKOWSKI,M.W.

NUREG/CR4038: SENSITMTY AND UNCERTAINTY STUDIES OF THE NUREG-0856 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES CRAC2 COMPUTER CODE. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

HITCHCOCK.J.T. JEANMOUGIN,N.

NUREG/CR4060: THE DC-1 AND DC 2 DEBRIS COOLABluTY AND NUREG/CR4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE i MELT DYNAMICS EXPERIMENTS. REACTOR COOLANT PUMP.

HOLMAN,G.S. )

JEUNG,M. I NUREG/CR-3660 V01: PROBABluTY OF PIPE FAILURE IN THE REAC-NUREG/CR4355 V01: 238 PU(IV) IN MONKEYS. Overview Of Metabo.

TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume fism.

1: Summary Report.

NUREG/CR-3660 V04: PROBABluTY OF PIPE FAILU9E IN REACTOR JOHNSON,A.C.

COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4. Pipe Failure induced By Crack Growth in West Coast Plants. NUREG/CR4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA AND EASTERN TENNESSEE EARTHOUAKES (1978-1984).

34 Personal Author index JOHNSON.J.D. KOLLAR,F, NUREG/CR4122- A FORTRAN 77 PROGRAM AND USER'S GUIDE NUREG/CR4317 V01: CANADIAN SEISMIC AGREEMENT.Technscal FOR THE CALCULATION OF PARTIAL CORRELATION AND STAND- Repor1 Covenng 1979-1985.

ARDIZED REGRESSON COEFFICIENTS.

KRAMARIC,M.

JOHNSON,J.J. NUREG-0837 VO4 N04: NRC TLD DIRECT RADIATON MONITORING NUREG/CR-4331: SIMPUFIED SEISMIC PROBA81USTIC RISK REPORT. Progress Report. October-December 1984.

ASSESSMENT. Procedures And Lirrutations. NUREG-0837 V05 N01: NRC TLD DIRECT RADIATON MONITORING NETWORK.Pr ess Report, January-March 1985.

JONES.J.W. NUREG-0837 NO2: NRC TLD DIRECT RADIATON MONITORING NUREG/CR-3736: FIELD AND THEORETICAL INVESTIGATONS OF NETWORK. Progress Report, Apni-June 1985.

FRACTURED CRYSTALUNE ROCK NEAR ORACLE, ARIZONA.

JORCENSEN C.C. NUREG/CR4082 V02: DEGRADED PIPING PROGRAM PHASE NUREG/CN-3481 V02: NUCLEAR POWER PLANT PERSONNEL QUAU- RSemannual Report, h M84 - M M81 FICATIONS AND TRAINING: TAPS - The Task Analysis Profilary System- KRANTZ,E.A.

KAGAMI,S. NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE-QUENCE INFORMATION.

NUREG/CR-3878: PROBABIUTY BASED LOAD COM81NATICN CRITE-RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. KU,J.Y.

KAM.F.S. NUREG/CR-4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE NUREG/CR-4284: NEUTRON EXPOSURE PARAMETERS FOR THE CRAC2 COMPUTER CODE.

FI HEAVY SECTON STEEL TECHNOLOGY IRRADIATON KURTZ,R.J.

NUREG/CR-3915: ACOUSTIC EMISSON RESULTS 08TAINED FROM

~

KAO,C. TESTING THE ZB 1 INTERMEDIATE SCALE PRESSURE VESSEL NUREG/CR-3876: PRO 8ABluTY BASED LOAD COMBINATION CRITE- NUREG/CR-4300 VOI: ACOUSTIC FMISSION/ FLAW RELATIONSHIP RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Progress Report, October March 1985.

KASSNER,T.F.

NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN KWAN,Q.

UGHT WATER REACTORS. Annual Report,0ctober 1983 - September NUREG/CR-4357: THE FEASIBluTY OF DETECTING THE IMPORT OF 1984. UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED STATES.

KELLY,J.E.

NUREG/CR-4060: THE DC 1 AND DC-2 DEBRIS COOLA81UTY AND LAHEY,R.T.

MELT DYNAMICS EXPERIMENTS- NUREG/CR-4116: NUFEGO-NP:A DIGITAL COMPUTER CODE FOR THE UNEAR STABluTY ANALYSIS OF 80 lung WATER NUCLEAR N red /CR4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU- LANDOW,M.

CLEAR REACTOR VESSELS AND PIPING COMPONENTS. NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE fl. Semiannual Report, October 1984 March 1985.

g NUREG/CR-3851 V04. EVALUATION OF RADIONUCUDE GEOCHEMi- LASSITER,D.L.

CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW October 1983-September 1984. PERFOF MANCE IN CONTROL ROOM SIMULATORS.

KENNEDY,R.P. LEE S.Y.

NUREG/CR4334:AN APPROACH TO THE QUANTIFICATON OF SEIS- NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI-MIC MARGINS IN NUCLEAR POWER PLANTS- CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR KESSLER,J.H.

WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For October 1983-September 1984.

NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI.

CAL INFORMATION DEVELOPED 8Y DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For LEWIS NUREG J.R[CR-4298: DESIGN AND INSTALLATON OF COMPUTER SYS-October M83-September 1984. TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

KEYS W.S. LEWIS,RM.

NUREG/CR-3738: FIELD AND THEORETICAL INVESTIGATONS OF FRACTURED CRYSTALUNE ROCK NEAR ORACLE, ARIZONA. NUREG/CR-4248: RECOMMENDATONS FOR NRC POLICY ON SHIFT SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS.

KHATIS-RAH 8AR NUREG/CR-4143: REVIEW AND EVALUATION OF THE MILLSTONE UAOL-H.

NUREG/CR4376; HEAT TRANSFER. CARRYOVER AND FALL 8ACK IN UNIT 3 PRO 8ABluSTIC SAFETY STUDY. Containment Failure Modes, Radiological Source-Terms And Offsite Consequences. PWR STEAM GENERATORS DURING TRANSIENTS.

KILLOUGH G.G. LIGON DM.

NUREG/CR-4038: SENTiTIVITY AND UNCERT UNTY STUDIES OF THE NUREG/CR-4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS CRAC2 COMPUTER CODE. SAFETY ANALYSIS. Phase 1, Final Report KINNISON,R. LO,T.Y.

NUREG/CP4063: PROCEEDINGS OF THE 1984 STATISTICAL SYMPO. NUREG/CR-3660 V04: PROBA81UTY OF PIPE FAILURE IN REACTOR SIUM ON NATIONAL ENERGY ISSUES. COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4. Pipe Failure Induced By Crack Growth in West Coast Plants.

NUREG/CR4290 V02: PROBABluTY OF PlPE FAILURE IN THE REAC- LOFARO,R.

l TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE

PLANTS. Volume 2. Guillotine Ore.k Indirectly Induced By Earthquakes. REACTOR COOLANT PUMP.

KOCHER.D.C. LOFGREN.E.

NUREG/CR4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE NUREG/CR 2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-CRAC2 COMPUTER CODE. DURES GUIDE. Sections 17 And Appendices.

Personal Author Index 35 LORENZ,R.A. MENSING,R.W.

NUREG/CR-4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR-3660 V04. PROBABluTY OF PIPE FAILURE IN REACTOR RELEASE TEST Hi-5. COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4. Pipe Fadure Induced By Crack Growth in West Coast Plants.

NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT MEYER,R.E.

PRAIRIE ISLAND NUCLEAR GENERATING STATON-NUREG/CR 3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI-LUDEWIG,H. CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NUREG/CR-4143: REVIEW AND EVALUATON OF THE' MILLSTONE WASTE REPOSITORY SITE PROJECTS Annual Progress Report For UNIT 3 PROBA81USTIC SAFETY STUDY. Containment FaAare October 1983-September 1984.

Modes, Radiological Source Terms And Offsste Consequences.

MEYER,R.O.

LUMM.D.K. NUREG 0956 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES NUREG/CR-4333: STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

UNOIS.

MILIAN.L LYONS,J.A. NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS NUREG/CR-4317 V01: CANADIAN SEISMIC AGREEMENT. Technical ON SOUDIFICATON. WASTE DISPOSAL AND ASSOOATED OCCU-Report Covenng 1979 1985. PATIONAL EXPOSURE.

MAlYA P S. MILLER,C.A.

NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-4182: VERIFICATON OF SOIL STRUCTURE INTERACTON UGHT WATER REACTORS. Annual Report. October 1983. September METHODS.

1984.

MANDLER,J.W

• NUREG/CRd397: IN-PU.NT SOURCE TERM MEAS'JREMENTS AT NURE'G/CR-3900 V04: LONG-TERM PERFORMANCE OF MATERIALS PRAIRIE ISLAND NUCLEAR GENERATING STATON. USED FOR HIGH-LEVEL WASTE PACKAGING Annual Report. April 1984. Apnl 1985.

MARSCHALL,C.W. NUREG/CR-4379 V01: LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR-4082 V02- DEGRADED PIPING PROGRAM - PHASE USED FOR HIGH-LEVEL WASTE PACKAGING First Quarterty ILSemiannual Report, October 1984. Maren 1985. Report, Year Four Apnt-June 1985.

MARTIN,R A. MILSTEAD,W.

NUREG/CR-4260: TORAC USER'S MANUALA Computer Code For Ana. NUREG-0933 S03: A PRIORITIZATION OF GENERIC SAFETY ISSUES.

Byzing Torrshinduced Flow And Material Transport in Nuclear Facdi-ties. MINNERS,W.

MARTIN,R.E. NUREG 0933 S03: A PRIORITI2ATION OF GENERIC SAFETY iSCUES.

NUREG-1149 TECHNICAL SPEOFICATONS FOR UMERICK GENER- MITCHELL,J.A.

ATING STATON, UNIT 1. Docket No. 50-352. (Philadelphia Electric NUREG-0956 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES Company) FOR ESilMATING SOURCE TERMS. (Draft Report For Comment).

MARTZ,H.F. MOAYERI,N.

NUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER NUREG/CR 4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-PLANT VALVE FAILURE-RATE VARIABluTY-SOME PREUMlNARY POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU-RESULTS-CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

NUREG CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE UREG/

ILSemiannual Report, October 1984 . March 1985. 214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER PLANT ACODENT CONSEQUENCE ANALYSIS.Part MC8 RIDE,K.C. I: Introduction, Integration & Summary.Part II.Soentific Basis For Health NUREG/CR-4298: DESIGN AND INSTALLATION OF COMPUTER SYS. Effects Models.

TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

MOHR.C.M.

MCCANN,M. NUREG/CR-3833 V01 SI: TRAC 801/ MOD 1.AN ADVANCED BEST ES-NUREG/CR-2815 V02 R1: PROBA81USTIC SAFETY ANALYSIS PROCE- TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR DURES GUOE. Sections 8-12. TRANSlENT ANALYSIS.

MCCANN M.W. MORRIS,8.M.

NUREG/CR-3485: PRA REVIEW MANUAL NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-MCCARDELL,R.K.

NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATIONAL MUNRO,P.S.

TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST FACluTY, NUREG/CR 4317 V01: CANADIAN SEISMIC AGREEMENT. Technical Report Covering 1979-1985.

MCCLUNG.R.W.

NUREG/CR-3949 V02: EDDY CURRENT INSPECTION FOR STEAM G R 8 OGRAM. Annual Progress Report For Penod NU EG 4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA AND EASTERN TENNESSEE EARTHOUAKES (1978-1984).

MCCONNELL J.W. MURATA K.K.

NUREG/CR-4150: EPICOR-il RESIN DEGRADATON RESULTS FROM NUREG/CR 4085: USERS MANUAL FOR CONTAIN 1.0.A Computer FIRST RESIN SAMPLES OF PF-8 AND PF-20. Code for Severo Reactor Accident Containment Analysis.

MCCORMICK,R.D. MURPHY,0.A.

NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATONAL NUREG/CR-4234 Vot: AGING AND SERVICE WEAR OF ELECTRIC TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY FEA.

FAQUTY. TURE SYSTEMS OF NUCLEAR POWER PLANTS.

MCELROY,W.N. MUSCARA,J.

l NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME- NUREG 1155 V02: RESEARCH PROGRAM PLAN. Steam Generators.

TRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveillanca NUREG-1155 V04: RESEARCH PROGRAM PLAN Non-Destructive Ex.

Pt ysics-Dossmetry Data Base Compendium. amination.

f 36 Personal Author index NAKAGAKI,M. OS80RN,R.N.

NUREG/CR-4082 V02- DEGRADED PIPING PROGRAM - PHASE NUREG/CR-4125 VOI: GUIDELINES AND WORKBOOK FOR ASSESS.

II.Sermannual Report. October 1984 . March 1985. MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER E /CR 249- PRESSURE VESSEL FRACTURE STUDIES PENE. p TRATING TO THE PWR THERMAL SHOCK ISSUE. EXPERIMENTS NUREG/CR-4125 V02 GUIDEUNES AND WORKBOOK FOR ASSESS-TSE-5,TSE-5A AND TSE4 MENT OF ORGANIZATON AND ADMINISTRATON OF UTluTIES NUREG/CR-4004: PRESSURE VESSEL FRACTURE STUDIES PER- SEEKING OPERATING UCENSE FOR A NUCLEAR POWER TAINING TO THE PWR THERMAL-SHOCK ISSUE.Expenment TSE-7. PLANT. Volume 2. Workbook For Assessrnent Of Organizaten And Man-agement NELSON W NUREG/CR-4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLCAR i NOREG/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. j

50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY CONTAINMENTS OF UGHT-WATER 400 LED NUCLEAR POWER OS80RNE,M.F.

E NUREG/CR-4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT NELSON,W.J. RELEASE TEST HI-5.

NUR /CR-4333: STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL. OSTMEYER,R.H.

NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC.

NELSON,W.R. CIDENTAL RADIONUCUDE RELEASES FROM NUCLEAR REAC-TORS.

NUREG/CR4272 RESPONSE TREE EVALUATON EXPERIMENTAL ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR OWCZARSKl.P.C.

OPERATORS NUREG/CR-4130: ICEDF:A CODE FOR AEROSOL PARTICLE CAP-NEUMAN,S.P. TURE IN ICE COMPARTMENTS.

NUREG/CR-3736: FIELD AND THEORETICAL INVESTIGATONS OF FRACTURED CRYSTALUNE ROCK NEAR ORACLE. ARIZONA. PAGE.R.E NUREG/CR-3613 V03 N1: EVALUATION OF WELDED AND REPAIR.

NICHOLS,F.A. WELDED STAINLESS STEEL FOR LWR SERVICE.Sermannual Report i NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN For October 1984 Through March 1985.

UGHT WATER REACTORS. Annual Report,0ctober 1983 September 1984. PAPASPYROPOULOS NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM . PHASE NICOLOSI,L. fl.Sermannual Report. October 1984 . March 1985.

NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATONS ON SOUDIFICATION. WASTE OfSPOSAL AND ASSOCIATED OCCU. PAPAZOOLOU,LA.

PATONAL EXPOSURE. NUREG/CR-2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE.

DURES GUIDE. Sections 1-7 And Appendices.

U dR4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST Hi 5. PAPPAS,R.A.

NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM REG [CR 2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE.

DURES GUIDE. Sections 17 And Appendices. PARK,J.Y.

NUREG/CR-3485: PRA REVIEW MANUAL NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Annual Report. October 1983 - September OSERLANDEM.P.L 1904-NUREG/CR-4251 VO*: MIT JATIVE TECHNIQUES FOR GROUND.

WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR CH 100 NUREG CR-4376: HEAT TRANSFER. CARRYOVER AND FALL BACK IN NUR CR 2 2: T AT E GROUND-PWR STEAM GENERATORS DURING TRANSIENTS.

WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR ACCIDENTS. Volume 2. Case Study Analysis Of Hydro 6ogic Character. PASEDAG W.F'

• 1. ^" * *' NUREGb856 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

OGDEN.D.M.

NUREG/CR-3935: THERMAL HYDRAULIC ANALYSES OF OVERCOOL- PASUPATH1,V.

ING SEQUENCES FOH THE H.B. ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY. NUREG/CR 4082 V02: DEGRADED PIPING PROGRAM - PHASE ll. Semiannual Report, October 1984 March 1985 NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE PEABODY,C.A.

PRESSURIZED WATER REACTOR. Main Report NUREG-0856 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES OLSON,J. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

NUREG/CR4125 V01: GUIDEUNES AND WORV800K FOR ASSESS-PE NG,S.J.

MENT OF ORGANIZATION AND ADMINISTRATON OF UTIUTIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR4116: NUFEGO-NP:A DIGITAL COMPUTER CODE FOR PLANT. Volume 1. Guidelines For Utility Orgarization And Administration THE UNEAR STABluTY ANALYSIS OF BOluNG WATER NUCLEAR Plan. REACTORS.

NUREG/CR4125 V02: GUIDEUNES AND WORKBOOK FOR ASSESS.

MENT OF ORGANIZATION AND ADMINISTRATION OF UTIUTIES PHILIPPACOPOULO SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR-4182: VERIFICATION OF SOIL STRUCTURE INTERACTION PLANT. Volume 2: Workbook For Assessment Of Organizaton And Man- METHODS.

NU -4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR PICIULO.P.L POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATONS ON SOUDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU.

OPITZ,5.E. PATONAL EXPOSURE.

NUREG/CR4259: TAluNGS NEUTRAUZATON AND OTHER ALTER-NATIVES FOR IMMOBluZING TOXIC MATERIALS IN TAluNGS. Final PITTMAN.J.

Report NUREG4933 S03: A PRIORITIZATION OF GENERIC SAFETY ISSUES.

l

+- _,

Personal Author index 37 PLOGER,SA RElCH,M. 4 NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATIONAL NUREG/CR-3876: PROBABlUTY BASED LOAD COMBINATON CRITE- l TRANSIENT (OPTRAN) TESTS 11 AND 1-2 IN THE POWER BURST RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

FACluTY. NUREG/CR-4182 VERIFICATION OF SOIL STRUCTURE INTERACTION PODOWSKi,M.Z.

NUREG/CR-4116: NUFEGO-NP:A DIGITAL COMPUTER CODE FOR REST,J.

THE LINEAR STABluTY ANALYSIS OF BOluNG WATER NUCLEAR NUREG/CR-3900 V04: LIGHT-WATER-REACTOR SAFETY FUEL SYS-l REACTORS. TEMS RESEARCH PROGRAMS Quarterly Progress Report,0ctober.

POLLACK,H.N. December 1984.

NUREG/CR-3145 V03: GEOPHYSICAL INVESTIGATONS OF THE REXROTH,P.E.

TE O-IN NA GON - ANNUAL REPORT 40ctober NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Computer Code for Severo Reactor Accident Contamment Analyss.

POLOSKI,J.P.

NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE. RIANI,L QUENCE INFORMATION. NUREG-0933 S03: A PRORITIZATON OF GENERIC SAFETY ISSUES.

POPELAR,C. RIGGS,R.

NUREG/CR-4062 V02 DEGRADED PlPING PROGRAM . PHASE NUREG4933 S03: A PRORITl2ATION OF GENERIC SAFETY ISSUES.

ILSerrmannual Report, October 1984 March 1985.

RIORDAN,5.

POWERS,TA NUREG/CR-4398: COST ANALYSIS OF REVISONS TO to CFR PART NUREG/CR-2800 S03: GUIDEUNES FOR NUCLEAR POWER PLANT 50 APPENDlX J. LEAK TESTS FOR PRIMARY AND SECONDAHY SAFETY ISSUE PRORITIZATON INFORMATION DEVELOPMENT. CONTAINMENTS OF UGHT WATER. COOLED NUCLEAR POWER PRAIRIE,R.R. PLANTS.

NUREG/CR4350 V02 PROBABILISTIC RISK ASSESSMENT COURSE RCOABAUGH,E.C.

DOCUMENTATON. Volume 2: ProbatMilly And Statistics For PRA @

NUREG/CR-4305: COMMENTS ON THE LEAK-BEFORE-BREAK CON-UEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS.

PRATT,W.

NUREG/CR-4143: REVIEW AND EVALUATON OF THE MILLSTONE ROSCOE,5.J.

UNIT 3 PROBABluSTIC SAFETY STUDY.Contamment Failure NUREG/CR-4250 VEHICLE BARRIERS. EMPHASIS ON NATURAL FEA-Modes. Radiological Source-Terms And Offsite Consequences. TURES.

PREVOSTJ. ROSE,JA NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACIUTIES.

REACTOR COOLANT PUMP.

ROSS P.A.

PRICE).M. NUREG4020 V09 N06: UCENSED OPERATING REACTORS STATUS NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY

SUMMARY

REPORT. Data As Of May 31,1965.(Grey Book l)

PROTECTON.llt. SUBMICRON PARTICLE TESTS AND

SUMMARY

NUREG-0020 V09 N07: UCENSED OPERATING REACTORS STATUS OF QUAUTY FACTORS-

SUMMARY

REPORT. Data As Of June 30,1965.(Gray Book 1)

PRUITTJ.S. RUGER,C.

NUREG/CR-4266: STANDARD BETA-PARTICLE AND MONOENERGE*

NUREG/CR-2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-TIC ELECTRON SOURCES FOR THE CAUBRATON OF BETA-RADS.

DURES GUIDE.Soctions 8-12.

ATON PROTECTION INSTRUMENTATION.

PUGH,C.E. RUNKLE,0.E.

NUREG/CR-4219 V01: HEAVY SECTON STEEL TECHNOLOGY PRO- NUREG/CR4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC-GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1964 - CIDENTAL RADIONUCUDE RELEASES FROM NUCLEAR REAC-MARCH 1985. TORS.

MANSOM C.S. RUTHER,W.E.

NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE UGHT WATER REACTORS. Annual Report,0ctober 1963. September PRESSURIZED WATER REACTOR Main Report. 1964.

RAO,K.S. RYDER,C.P.

NUREG/CR-4038. SENSITIVITY AND UNCERTAINTY STUDIES OF THE NUREG-0656 DAFT FC: REASSESSMENT OF THE TECHNCAL BASES CRAC2 COMPUTER CODE. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

RATLIFF,R.A. SAHA,P.

NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE- NUREG/CR-4252: INDEPENDENT ASSESSMENT OF TRAC.PD2/ MOD 1 RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE CODE WITH BCL ECC BYPASS TESTS.

REGULATORY PROGRAM.

SAMANTA.P.K.

N E '" ^

-4138: DATA ANALYSES FOR NEVADA TEST SITE (NTS) RES GUI 7 Ar d l' REMIXED COMBUSTION TESTS.

RAVINDRA,M.K. 8AEA8'AA NUREG/CR-4290 V02- PROBABluTY OF PIPE FAILURE IN THE REAC- NUREG/CR-4150 EPCOR-il RESIN DEGRADATON RESULTS FROM TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR FIRST RESIN SAMPLES OF PF 8 AND PF-20.

PLANTS. Volume 2.Guillotme Break Indirectty induced By Earthquakes.

SCHE N A l REED.J.W. NUREG/CR4065: USERS MANUAL FOR CONTAIN 1.0.A Computer NUREG/CR 2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE. Code for Severo Reactor Accident Contamment Analyse.

DURES GUIDE.Sectione 812.

NUREG/CR-3485: PRA REVIEW MANUAL SCHMIDT,C.T.

NUREG/CR-4334: AN APPROACH TO THE QUANTIFICATION OF SEIS. NUREG/QR4355 V01: 238 PU(IV) IN MONKEYS. Overview Of Metabo.

MIC MARGINS IN NUCLEAR POWER PLANTS. lism.

l l

38 Personal Author index SCHMULT,S. SHERWCOD,D.R. l NUREG/CR4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR- NUREG/CR-4259: TAIUNGS NEUTRAUZATON AND OTHER ALTER- 1 POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU- NATIVES FOR IMMOBluZING TOX:C MATERIALS IN TAluNGS Final CLEAR REACTOR VESSELS AND PIPING COMPONENTS. Report.

SCHRECK,R.L SHIEH,LC.

NUREG/CR-4130 ICEDF:A CODE FOR AEROSOL PARTICLE CAP- NUREG/CR-4331: SIMPLIFIED SEISMIC PROBABluSTIC RISK TURE IN ICE COMPARTMENTS. ASSESSMENT. Procedures And Urrutstens.

SCHWARTZ,R.S. SHtNOZUKA,M.

NUREG/CR-4268: STANDARD BETA-PARTICLE AND MONOENERGE. NUREG/CR 3876: PROBABILITY BASED LOAD COMBINATION CRITE.

TIC ELECTRON SOURCES FOR THE CALIBRATON OF BETA.RADI- RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

ATON PROTECTION INSTRUMENTATION. NUREG/CR4334: AN APPROACH TO THE QUANTIFICATION OF SEIS.

MIC MARGINS IN NUCLEAR POWER PLANTS.

SCIACCA,F.

SHIRE,P.R.

NUREG/CR-4398. COST ANALYSIS OF REVISIONS TO 10 CFR PART 50, APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer Code for Severe Reactor Acodent Containment Analyss.

CONTAINMENTS OF UGHT-WATER COOLED NUCLEAR POWER SH10,K.K.

SCIACCA,F.W. NUAEG/CR*2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer DURES GUIDE.Sectons 8-12.

Code for Severo Reactor Acodent Containment Analysis. NUREG/CR-3485: PRA REVIEW MANUAL SCOTT,P SHMTENCARIER NUREG/CR4122: A FORTRAN 77 <ROGRAM AND USER'S GUIDE NUREG/CR-4082 V02 DEGRADED PIPING PROGRAM - PHASE FOR THE CALCULATON OF PARTIAL CORRELATION AND STAND-II.Sermannual Report. October 1984 . March 1985.

ARDIZED REGRESSION COEFFICIENTS.

SEELEY,F.G. y NUREG/CR-3851 V04: EVALUATON OF RADONUCUDE GEOCHEMI- NUREG/CR4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSI-CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR CAL BENCHMARKING OF PIPING SYSTEMS.-

WASTE REPOSITORY SITE PROJECTS Annual Progress Report For October 1983-September 1984. SHUMWAY,R.W.

NUREG/CR-3633 V01 S1: TRAC-BD1/ MOD 1:AN ADVANCED BEST ES-SEGE,&

^

NUREG4933 S03: A PRORITIZATION OF GENERIC SAFETY ISSUES. fRANS E A ALYSIS SEGOL,4 NUREG/CR-3633 V04: TRAC 801/ MOD 1.AN ADVANCED BEST ESTI-NUREG/CR-3901: DOCUMENTATION AND USER'S GUIDE.GS2 & GS3 MATE COMMER NM M MM WATER MAM TRANSIENT ANALYSIS. Volume 4: Developmental Assessment.

VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS.

SILSERSERG,M.

SEITZ,M.4 N 6 N m WSSESSMENT & THE TECHN BASES NUREG/CR-3710: LABORATORY STUDISS OF A BREACHED NUCLE- FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

AR WASTE REPOSITORY IN BASALT.

I'" " " I'"'A" SENGLAUS,M.E.

NUREG/CR-4268: RATIO METHOOS FOR COST. EFFECTIVE FIELD NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer SAMPLING OF COMMERCIAL RADCACTIVE LOW. LEVEL WASTES.

Code for Severe Reactor Acodent Containment Analyss.

SIMPKINS,S.

"* NUREG/CR4398: COST ANALYSIS OF REVISONS TO 10 CFR PART NUREG/CR4259- TAILINGS NEUTRAUZATON AND OTHER ALTER. 50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY NATIVES FOR IMMOBluZING TOXIC MATERIALS IN TAluNGS Final CONTAINMENTS OF UGHT WATER. COOLED NUCLEAR POWER Report PLANTS.

SERPAN.C.Z. 84MPSON E.S.

NUREG-1155 V02 RESEARCH PROGRAM PLAN. Steam Generators. NUREG/CR 3738- FIELD AND THEORETICAL INVESTIGATONS OF SHACK,WJ. FRACTURED CRYSTALUNE ROCK NEAR ORACLE ARIZONA.

NUREG/CR4287; ENVIRONMENTALLY ASSISTED CRACKING IN SINGER,4L UGHT WATER REACTORS. Annual Report,0ctober 1983. September NUREG/CR-3633 V01 S1: TRAC 001/ MOD 1:AN ADVANCED BEST ES.

198L TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR TRANSIENT ANALYSIS.

SHAFER,J.M.

NUREG/CR4251 V01: MITIGAffVE TECHNIQUES FOR GROUND- SKAGGS.R.L.

WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR4251 V01: MITIGATIVE TECHNIQUES FOR GROUND-ACCIDENTS. Volume 1:Analyms Of Genenc Sde Conditons. WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR-4251 V02: MITIGATIVE TECHNIQUES FOR GROUND. ACCIDENTS. Volume 1 Analyse Of Generic Sale Conditions.

WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR4251 V02: MITIGATIVE TECHNIQUES FOR GROUND.

ACCIDENTS. Volume 2. Case Study Analysis Of Hydrologic Character. WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR traton And Mitigative Schemes ACCIDENTS. Volume 2 Case Study Analyse Of Hyctologic Character.

Ization And Mitigative Schemes.

NUREG/CR-3948. EXPERIMENTAL RESULTS OF THE OPERATONAL SKORPtK.J.R.

TRANSIENT (OPTRAN) TESTS 11 AND 1-2 IN THE POWER BURST NUREG/CR-3915: ACOUSTIC EMISSON RESULTS OBTAINED FROM FACIUTY, TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL SHAPIRO,S.J. SLOVIK,0.C.

NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCOENT SE. NUREG/CR4252: INDEPENDENT ASSESSMENT OF TRAC-PD2/ MOD 1 QUENCE INFORMATON. CODE WITH BCL ECC BYPASS TESTS.

SHEPHERD,J.E. SMITH,F.J.

NUREG/CR-3838: HYDROGEN-STEAM JET. FLAME FACluTY AND EX- NUREG/CR-3851 V04: EVALUATON OF RADIONUCUDE GEOCHEMI-PERIMENTS. CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR

Personal Author index 39 WASTE REPOSITORY SITE PROJECTSAnnual Progress Report For NUREG/CR 4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSI.

October 1983-September 1984. CAL BENCHMARKING OF PIPING SYSTEMS.

SMITH,J.H- SUES,R.H.

NUREG/CR-3949 V02 EDDY CURRENT INSPECTION FOR STEAM NUREG/CR4290 V02: PROBABluTY OF PIPE FAILURE IN THE REAC-GENERATOR TUBING PROGRAM. Annual Progress Report For Penod TOR COOLANT LOOPS OF DABCOCK AND WILCOX PWR Ending Decernber 31,1984. PLANTS. Volume 2 Guillotine Break Indirectty Induced By Earthquakes.

SMITH,P.D. SULUVAN,W.H.

NUREG/CR-4331: SIMPUFIED SEISMIC PROBABILfSTIC RISK NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE-

. ASSESSMENT. Procedures And Lmtations- QUENCE INFORMATION.

SMIM.P.R. TANG,P.K.

NUREG/CR4232- THE RESPONSE OF '.'"iTILATON DAMPERS TO NUREG/CR4260: TORAC USER'S MANUALA Computer Code For Ana.

LARGE AIRFLOW PULSES. lyzing Tomado-induced Flow And Matenal Transport in Nuclear Facdi-SOARES,C.G. ties.

NUREG/CR-4266; STANDARD BETA-PARTICLE AND MONOENERGE-TAWIL,J.J TIC ELECTRON SOURCES FOR THE CALIBRATON OF BETA-RADI.

ATON PROTECTION LMSTRUMENTATON. NUREG CR 3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL ACCIDENTS. METHODS, COSTS AND SCHEDULES FOR DECON.

SOMMERS.P.E. TAMINATION.

NUREG/CR-4125 V01: GUIDEUNES AND WORKBOOK FOR ASSESS-MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES TAMDA SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR-3633 V01 S1: TRAC-BD1/ MODI.AN ADVANCED BEST ES-PLANT. Volume 1: Guidelines For Utihty Organization And Adrmnistration TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR Plan. TRANSIENT ANALYSIS.

NUREG/CR4125 V02: GUIDEUNES AND WORKBOOK FOR ASSESS-TEAGUE,A.G.

MENT OF ORGANIZATION AND ADMINISTRATION OF UTIUTIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR-4288. FOCAL MECHANISM ANALYSES FOR VIRGINIA PLANT. Volume 2: Workbook For Assessment Of Organization And Man- AND EASTERN TENNESSEE EARTHOUAKES (1978-1984).

NU CR4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR EHMANN,T.

POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. NUREG/CR-2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-DURES GUIDE. Sections 17 And Appendices.

SOO,P. NUREG/CR 2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE.

NUREG/CR-3091 V06: REVIEW OF WASTE PACKAGE VERIFICATION DURES GUIDE. Sections 812.

TESTS.Sermannual Report Covering The Period October 1984 - March NUREG/CFi J485: PRA REVIEW MANUAL THOMAS J M.

STAHL.D. NUREG/CR-4268. RATIO METHODS FOR COST EFFECTIVE FIELD NUREG/CR-3900 V04: LONG-TERM PERFORMANCE OF MATERIALS SAMPUNG OF COMMERCIAL RADf0 ACTIVE LOW-LEVEL WASTES.

USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report.Apnl 1984 Apn'i 1985. THURSER,J.A.

NUREG/CR-4379 V01: LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR-4125 V01: GUIDELINES AND WORKBOOK FOR ASSESS-USED FOR HIGH-LEVEL WASTE PACKAGING.First Quarterty MENT OF ORGANIZATON AND ADMINISTRATON OF UTILITIES Report, Year Four ApntJune 1985. SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLANT. Volume 1: Guidelines For Utility Organization And Admirwstration STALKER A.C. Plan.

NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT NUREG/CR4125 V02: GUIDEUNES AND WORKBOOK FOR ASSESS-PRAIRIE ISLAND NUCLEAR GENERATING STATON. MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES l SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NU G/CFi 284: NEUTRON EXPOSURE PARAMETERS FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATON NU CR-4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR SERIES.

POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY.

STAMATELATOS,M.

TOGIAS,M.L NUREG/CR4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS NUREG/CR4255 V01: AEROSAL RELEASE AND TRANSPORT PRO-SAFETY ANALYSIS. Phase 1, Final Report GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

STEELE.R. MARCH 1985.

NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACluTIES. TRAVIS,J.R.

STITT,0.D. NUREG/CR4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES RELEASE TEST Hi-5.

ON TRANSIENTS AND ACCOENTS AT A 3-LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR Main Report " L N REG -4085: USERS MANUAL FOR CONTAIN 1.0.A Computer STROMBERG.H.M. Code for Severe Reactor Accident Containment Analysis.

NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE TU N ,J.R.

PRESSURIZED WATER REACTORMain Report NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU.

CLEAR POWER PLANT CONTROL ROOM CREWS.

STROSNIDER.J.

NUREG 1155 V03: RESEARCH PROGRAM PLAN. Piping. UNIONE,A.

NUREG/CR 2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE.

SUORAMANIAN,C. DURES GUIDE. Sections 17 And Appendices.

NUREG/CR-4119: INTEGRITY OF CONTAINMENT PENETRATIONS NUREG/CR 2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE.

UNDER SEVERE ACCIDENT CONDITIONS FY84 ANNUAL REPORT. DURES GUIDE. Sections 812.

SUGUOHI,M. VADEN,J.

NUREG/CR.1877 V02: PlPING BENCHMARX PROBLEMS. VOLUME 11 NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE.

DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION RE. RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE SPONSE SPECTRUM METHOD. REGULATORY PROGRAM.

40 Personal Author index VAGINS,M. WELC J.

NUREG 1155 VOI: RESEARCH PROGRAM PLAN. Reactor Vessels. NUREG/CR-3145 V03 GEOPHYSICAL INVESTIGATIONS OF THE NUREG 1155 V03: RESEARCH PROGRAM PLAN. Piping. WESTERN OHIO-INDIANA REGION ANNUAL REPORT.(October 1982 - September 1983. Volume 3).

NUREG/CR-3426 V01: THERMAL AND FLUID MixlNG IN 1/2-SCALE WELLS.J.E.

TEST FACluTY. Facity And Test Design Report. ,

NUREG/CR4331: SIMPUFIED SEISMIC PROBABlUSTIC RISK NUREG/CR-3428 V02: THERMAL AND FLUID MIXING IN 1/2-SCALE ASSESSMENT. Procedures And Limitations.

TEST FACluTY. Data Report.

WETTMILLER,RJ.

VANDEGRIFT,0.F. NUREG/CR-4317 V01: CANADIAN SEISMIC AGREEMENT. Technical NUREG/CR-3710' LABORATORY STUDIES OF A BREACHED NUCLE. Report Covenng 1979-1985.

AR WASTE REPOSITORY IN BASALT. WHATLEY,S.K.

VANDERMOLEN,H. NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI-NUREG4933 S03: A PRIORITIZATION OF GENERIC SAFETY ISSUES. CAL INFGAMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Annual Progress Report For VESELY,W. October 1983-September 1984.

NUREG/CR-2815 V01 R1: PROBABILISTIC SWETY AN ALYS!S PROCE-DURES GUtDE. Sections 17 And Appendices. WIDRIG,R D.

NUREG/CR4125 V01: GUIDELINES AND WORKBOOK FOR ASSESS-VESELY,W.E MENT OF ORGANIZATION AND ADMINISTRATION OF UTIUTIES NUREG/CR4377: EVALUATIONS AND UTlu2ATIONS OF RISK IM- SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PORTANCES. PLANT. Volume 1. Guidelines For Utility Orgaruzation And Administration Plan.

VICKROY,S.C. NUREG/CR-4125 V02: GUIDELINES AND WORKBOOK FOR ASSESS-NUREG/CR4298: DESIGN AND INSTALLATION OF COMPUTER SYS- MENT OF ORGANIZATION AND ADMINISTRATION OF UTluTIES TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55. SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLANT. Volume 2. Workbook For Assessment Of Orgaruzation And Man-VIERZ8A,E. agenient NUREG/CR4357: THE FEASIBluTY OF DETECTING THE IMFORT OF UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED WILKOWSKI,G.M.

STATES. NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM PHASE II. Semiannual Report October 1984. March 1985.

NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO- WILLIAMS,0.C.

GRAM PLAN.

NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Computer

' Code for Severs Reactor Accident Containment Analysis.

NUREG/CR-4314: BRIEF SURVEY AND COMPARISON OF COMMON WILSON.J.H.

CAUSE FAILURE ANALYSIS. NUREG/CR 3885 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT CALLO,A. EVALUATION Ouarterty Progress Report. October 1-December NUREG/CR4357: THE FEASIBluTY OF DETECTING THE IMPORT OF 31,1964.

UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED WINEGARDNER,W.

STATES.

NUREG/CR-413'): ICEDF:A CODE FOR AEROSOL PARTICLE CAP.

CANG,Y.K. TURE IN ICE COMPARTMENTS.

NUREG/CR4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSI-WO CAL BENCHMARKING OF PIPING SYSTEMS.

N G R 213: SETS REFERENCE MANUAL CARD.R.C. #

NU E CR4038 SEN ITY AND UNCERTAINTY STUDIES OF THE NUR G/CR4385: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU.

CATERMAN,M.E. CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE N G/CR-3537: EXPEDIENT METHODS OF RESPIRATORY PRESSURIZED WATER REACTORMain RW PROTECTION. lit. SUBMICRON PARTICLE TESTS AND

SUMMARY

WEAKLEY,S.A. OF OUAUTY FACTORS.

NUREG/CR-2800 S03. GUIDEUNES FOR NUCLEAR POWER PLANT YOUNG T.E.

SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT. NUREG/CR4397: IN-PLANT SOURCE TERM MEASUREMENTS AT PRAIRIE ISLAND NUCLEAR GENERATING STATION.

WEBER,C.F.

NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR YOUNG 8LOOD BJ.

SAFETY STUDIES FOR THE DivlSION OF ACCIDENT NUREG/CR-2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-EVALUATION.Ouarterty Progress Report, October 1 December DURES GUIDE. Sections 812.

31,1984.

YOUNGBLOOD R.W.

WE8 STER.C.S. NUREG/CR 2815 V01 RI: PROBABILISTIC SAFETY ANALYSIS PROCE-NUREG/CR4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT DURES GUIDE. Sections 17 And Appendices.

RELEASE TEST Hi 5. NUREG/CR 3485; PRA REVIEW MANUAL

Subject Index This index was developed from keywords moved later when a reasonable thesaurus end word strings in titles and abstracts. has been developed through experience.

During this development period, there will Suggestions for improvements are wel-be some redundancy, which will be re- come.

ALARA Acc6 dent Mitigation NUREG/CP-0066: PROCEEDINGS OF AN INTERNATIONAL WORK- NUREG/CR4143: REVIEW AND EVALUATION OF THE MILLSTONE SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTON UNIT 3 PROBABILISTIC SAFETY STUDY. Containment Fadure (ALARA) AT NUCLEAR POWER PLANTS,MAY 29,1UNE 1.1984. Modes,Radiologeal Source-Terms And Offsite Consequences.

i NUREG/CR-4254: OCCUPATIONAL DOSE REDUCTION AND ALARA

! AT NUCLEAR POWER PLANTS. Study On High-Dose Jobs Radwaste Acid Digestion Handkng.And ALARA incentives. NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS ON SOUD!FICATON. WASTE DISPOSAL AND ASSOCIATED OCCU-2 EG/Cn-3633 V01 SI: TRAC 801/ MOD 1:AN ADVANCED BEST ES-TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR Acoustic Emiselon TRANSIENT ANALYSIS- NUREG 1155 V04: RESEARCH PROGRAM PLAN.Nor> Destructive Ex-amination Abnormal Occurrence NUREG/CR'-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM NUREG-0090 V08 NOI: REPORT TO CONGRESS ON ABNORMAL TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL OCCURRENCES.Janurary-March 1985. NUREG/CR4300 Vot: ACOUSTIC EMISSION / FLAW RELATIONSHIP Abstract FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG-0304 V10 NO2: REGULATORY AND TECHNICAL VESSELS. Progress Report, October-March 1985.

REPORTS.Compdation For Second Quarter 1985.

Administration Accese Control NUREG/CR-4125 V01: GUIDELINES AND WORKBOOK FOR ASSESS-NUREG/CR-4298: DESIGN AND INSTALLATON OF COMPUTER SYS. MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.- SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PLANT. Volume'1: Guidelines For Utility Organization And Administiation Accident Plan.

NH"1EG4956 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES NUREG/CR-4125 V02- GUIDEUNES AND WORKBOOK FOR ASSESS-FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment). MENT OF ORGANIZATION AND ADMINISTRATON OF UTIUTIES NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE- SEEKING OPERATING LICENSE FOR A NUCLEAR POWER NU EG CR 1 OF S TE CONSEQUENCES OF RADIOLOGICAL ACCtDENTS. METHODS, COSTS AND SCHEDULES FOR DECON- 89'**

TAMINATION Aerosol EAC FETY RESEARCH Ouarterty NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY Report. July - ember 1984. PROTECTION.fil. SUBMICRON PARTICLE TESTS AND

SUMMARY

NUREG/CR . DETERMINATON OF THE AVAILABluTY OF CORE OF QUAUTY FACTORS.

EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATONS. NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer NUREG/CR 4085: USERS MANUAL FOR CONTAIN 1.0.A Computer Code for Severe Reactor Acodent Containment Analysis.

Code for Severe Reactor Acadent Containment Analyss. NUREG/CR-4130t ICEDFA CODE FOR AEROSOL PARTICLE CAP-NUREG/CR-4119: INTEGRITY OF CONTAINMENT PENETRATONS TURE IN ICE COMPARTMENTS.

UNDER SEVERE ACCIDENT CONDITONS FY84 ANNUAL REPORT. NUREG/CR-4255 V01: AEROSAL RELEASE AND TRANSPORT PRO-NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC- GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

CIDENTAL RADONUCLIOE RELEASES FROM NUCLEAR REAC- MARCH 1985.

TORS.

NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER Aging PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part NUREG 1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-1.IntroductionIntegration & Summary.Part ILScientific Bass For Health GRAM PLAN.

Effects Models. NUREG/CR-3710: LABORATORY STUDIES OF A BREACHED NUCLE-NUREG/CR4251 V01: MITIGATIVE TECHNIOUES FOR GROUND- AR WASTE REPOSITORY IN BASALT.

WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACluTIES.

ACCIDENTS. Volume 1 Ana_ lyss Of Genere Site Conditions. NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC NUREG/CR-4251 V02: MinGATIVE TECHNIQUES FOR GROUND- MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY-FEA-WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR TURE SYSTEMS OF NUCLEAR POWER PLANTS.

ACCIDENTS. Volume 2. Case Study Analysis Of Hydrologc Character-tration And Mitigative Schemes. Alterneting Current NUREG/CR-4294. LEAK RATE ANALYSIS OF THE WESTINGHOUSE NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE REACTOR COOLANT PUMP. REACTOR COOLANT PUMP.

NUREG/CR 4304: PRESSURE VESSEL FRACTURE STUDIES PER-TAINING TO THE PWR THERMAL-SHOCK ISSUE. Experiment TSE-7. Ambient Redletion Levele NUREG/CR4325: A PARAMETRIC STUDY OF PWR PRESSURE NUREG 0837 V04 N04: NRC TLD DIRECT RADIATON MONITOR;NG VESSEL INTEGRITY DUR'NG OVERCOOUNG REPORT. Progress Report. October December 1984.

ACCIDENTS,CONSIDERING BOTH 2-D A.40 3-0 FLAWS.

NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES Anticipated Tranelents Without Scram ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE NUREG/CR 3633 V01 SI: TRAC-BD1/ MOD 1:AN ADVANCED BEST ES-PRESSURIZED WATER REACTOR Main Report. TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR TRANSIENT ANALYSIS.

NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER Artificial Intelligence PLANT ACCIDENT CONSEQUENCE ANALYSIS Part NUREG/CR-3481 V02: NUCLEAR POWER PLANT PERSONNEL QUAU.

l. Introduction. Integration & Summary Part II: Scientific Bass For Health FICATONS AND TRAINING: TAPS - The Task Analysis Profiling Effects Models. System.

41

42 Subject index NUREG/CR-4272- PESPONSE TREE EVALUATION. EXPERIMENTAL CRT Display ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR NUREG/CR-4227: HUMAN ENGINEERING GUIDELINES FOR THE OPERATORS. EVALUATON AND ASSESSME*T OF VISUAL D: SPLAY UNITS.

Austenitic Steintese Steel Centration NUREG/CR-3613 V03 N1: EVALUATON OF WELDED AND REPAIR- NUREG/CR-4266: STANDARD BETA-PARTICLE AND MONOENERGE-WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report TIC ELECTRON SOURCES FOR THE CAUBRATON OF BETA-RAryg.

For October 1984 Through March 1985. ATON PROTECTION INSTRLMENTATON.

BWIP Canada NUREG/CR-3851 V04: EVALUATON OF RADIONUCUDE GEOCHEMI. NUREG/CR 4317 VL1: CANADIAN SEISMIC AGREEMENT.Techrmcal CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR Report Covenng 1979-1985.

WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For j October 1983-September 1984. Chemical Clooning i

=

NUREG-1155 V02: RESEARCH PROGRAM PLAN Steam Generators.

Beience Study NU G/CR-4355 V01: 238 PU(IV) IN MONKEYS. Overview Of Metabo. C GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

Saeeft MARCH 1985.

NUREG/CR-3710- LABORATORY STUDIES OF A BREACHED NUCLE. NUREG/CR-4275: HEAVY-SECTON STEEL TECHNOLOGY PROGRAM AR WASTE REPOSITORY IN BASALT. - FIVE YEAR PLAN FY 1984-1988.

NUREG/CR-3851 V04: EVALUATON OF RADIONUCUDE CEOCHEMI-CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR Clooecut WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For NUREG/CR-4006: CLOSEOUT OF IE BULLETIN 81-01. SURVEILLANCE October 1983-September 1984. OF MECHANICAL SNUBBERS.

NUREG/CR4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS SAFETY ANALYSIS. Phase 1. Foal Report. O REG-0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES

' genchmerti Probleme FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

NUREG/CR 1677 V02: PIPING BENCHMARK PROBLEMS, VOLUME 11 NUREG/CR-3091 V06: REVIEW OF WASTE PACKAGE VERIFIC% tot!

DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTON RE. TESTS. Semiannual Report Covenng The Penod October 1984 - Marct i SPONSE SPECTRUM METHOD.

NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME.

I Bentonite TRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveillance NUREG/CR-3710 LABORATORY STUDIES OF A BREACHED NUCLE. Physics-Dosametry Data Base Comperdum.

AR WASTE REPOSITORY IN BASALT. NUREG/CR 3901: DOCUMENTATION AND USER S GUIDE:GS2 & GS3

- VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS.

Beta Radiation NUREG/CR4130: ICEDF:A CODE FOR AEROSOL PARTICLE CAP-NUREG/CR-4266: STANDARD BETA-PARTICLE AND MONOENERGE. TURE IN ICE COMPARTMENTS.

TIC ELECTRON SOURCES FOR THE CALIBRATION OF BETA-RADI. NUREG/CR 4252: INDEPENDENT ASSESSMENT OF TRAC-PD2/ MODI ATON PROTECTION INSTRUMENTATON. CODE WITH BCL ECC BYPASS TESTS.

Body Waves Combustion NUREG/CR-4354: A STUDY OF SEISMICITY AND TECTONICS IN NEW NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS ENGLAND. Final Report. ON SOUDIFICATON. WASTE DISPOSAL AND ASSOCIATED OCCU-PATONAL EXPOSURE.

Booleen Algebra NUREG/CR-4213: SETS REFERENCE MANUAL Common Cause Fellure NUREG/CR-4314: BRIEF SURVEY AND COMPARISON OF COMMON Borosellcate Glees CAUSE FAILURE ANALYSIS.

i NUREG/CR 3900 V04: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report Aprd Compilation Of Rulee

'l 1984 + Apnl 1985. NUREG4936 V04 NO2: NRC REGULATORY AGENDA.Ouarterfy Report,Apnt-June 1985.

Buoyant Plume NUREG/CR-3426 V01: THERMAL AND FLUID MIXING IN 1/2 SCALE Component Degredation TEST FACluTY. Facdify And Test Desegn Report. NUREG 1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-NUREG/CH-3426 V02: THERMAL AND FLUID MIXING IN 1/2. SCALE GRAM PLAN.

TEST FACluTY. Data Report.

4 sypees Test NUREG/CR-3442- RADTWO:A COMPUTER CODE FOR SIMULATING NUREG/CR-4252: INDEPENDENT ASSESSMENT OF TRAC PD2/ MOD 1 FAST. TRANSIENT, TWO-DIMENSIONAL,TWO LAYER RADIONU-

, CODE WITH BCL ECC BYPASS TESTS. CUDE CONCENTRATON CONDITONS IN

LAKES. RESERVOIRS. RIVERS, ESTUARIES,AND COASTAL REGIONS.

NUREG/CR-3633 V01 St: TRAC-BD1/ MOD 1;AN ADVANCED BEST ES-COSRA NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH TIMATE COMPUTER PROGRAM FOR BOfuNG WATER REACTOR PROGRAMS.Quarterty Report, January-March 1985. TRANSIENT ANALYSIS.

, NUREG/CR-3633 V04: TRAC-BD1/ MODI AN ADVANCED BEST ESTI-CONTAIN MATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer TRANSIENT ANALYSIS Volume d: Dev tel Assessment.

Code for Severe Reactor Accident Conta nment Analysis. NUREG/CR4038: SENSITIVITY AND U RTAINTY STUDIES OF THE CRAC2 COMPUTER CODE.

CORCON NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0 A Computer f NUREG-0956 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES Code for Severe Reactor Accident Containment Analyses.

FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment). NUREG/CR-4116: NUFEGO NP:A DIGITAL COMPUTER CODE FOR 4 THE UNEAR STABILITY ANALYSIS OF BOluNG WATER NUCLEAR CORSOR REACTORS.

NUREG-0956 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC.

FOR ESTIMATING SOURCE TERMS. (Draft Repor1 For Comment). CIDENTAL RADIONUCUDE RELEASES FROM NUCLEAR REAC.

TORS.

CRAC2 NUREG/CR4260: TORAC USER S MANUAL.A Computer Code For Ane-NUREG/CR4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE lymng Tomado-induced Flow And Matenal Transport in Nuclear Facili-CRAC2 COMPUTER CODE. tes.

1 l

l

l Subject Index 43 NUREG/CR4318 V01: REACTOR SAFETY F1ESEARCH Control Room PROGRAMiOuarterly Report. January. March 1985. NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND Computer Modeling ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW PERFORMANCE IN CONTROL ROOM SlMULATORS.

NbREG/CR4251 V01: MITIGATIVE TECHN QUES FOR GROUND-WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR Control System ACCIDENTS.Vclume 1 Analyss Of Genenc Site Corxlitiorts.

NUREG/CR-4328 V01: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE

/ AN ENGINEERING GUIDEUNES FOR THE EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS. Coolant Solloway And Damage Ceph & NUREG/CR 4318 V01: REACTOR SAFETY RESEARCH PROGRAMS.Ouarterty Report. January March 1985.

NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADOLOGICAL ACCfDENTS. METHODS, COSTS AND SCHEDULES FOR DECON-Core NU E C 22: A FORTRAN 77 PROGRAM AND USER'S GUIDE

" ^

FOR THE CALCULATON OF PARTIAL CORRELATION AND STAND-N /C 94 EX ER E TAL E ULTS OF THE OPERATONAL ARDIZED REGRESSON COEFFICIENTS. TRANSIENT (OPTRAN) TESTS 1-1 AND 12 IN THE POWER BURST NUREG/CR-4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA FA NUpE 4080: DETERMINATION OF THE AVAILA81UTY OF CORE NUR G CR437 : TT SFER CAR Y EI L BACK IN EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATONS.

PWR ST *AM GENERATORS DURING TRANSIENTS.

Core Cooling I Computer Securtty NUREG/CR4060: THE DC-1 AND DC 2 DEBRIS COOLABIUTY AND

! MELT DYNAMICS EXPERIMENTS.

NUREG/CR-4298: DESIGN AND INSTALLATON 07 COMPUTER SYS.

TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

Core Deeruptive Accident Computer System NUREG/CR4240 V01: PHYSICS OF REACTOR SAFETY.Quarterty NUREG/CR4298: DESIGN AND INSTALLATON OF COMPUTER SYS. Report. January-March 1985.

TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

Concrete NUREG/CR4130 ICEDFA CODE FOR AEROSOL PARTICLE CAP-NUF.EG/CR-3878: PROBA81UTY BASED LOAD COMBINATION CRITE. TURE IN ICE COMPARTMENTS.

RfA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. NUREG/CR4251 V01: MITIGATIVE TECHNIQUES FOR GROUND-WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR Congrees ACCIDENTS. Volume 1 Anatyss Of Genenc Site Conditions.

NUREG4090 V08 N01: REPORT TO CONGRESS ON ABNORMAL NUREG/CR-4251 V02: MITIGATIVE TECHNIQUES FOR GROUND-OCCURRENCES.Janurary-March 1985. WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR y ___

ACCIDENTS. Volume 2. Case Study Analyss Of Hydrologic Character-i NUREG/CR-4 EVIEW AND EVALUATION OF THE MILLSTONE UNIT 3 PROBABluSTIC SAFETY STUDY.Containtnent Failure Corrective Maintenance Modes. Radiological Source Terms And Offsete Consequences.

NUREG/CR-4377: EVALUATIONS AND UTIU2ATONS OF RISK IM-Catelw PORTANCES.

NUREG/C43900 V04: LONG-TERM PERFORMANCE OF MATERIALS Corrooton

, USED FOR HIGH-LEVEL WASTE PACKAGING.Amual ReportApr#

1984. Apnl1985. NUREG 1155 V02: RESEARCH PROGRAM PLAN. Steam Generators.

NUREG/CR 4379 V01: LONG TERM PERFORMANCE OF MATER'ALS Corrosion Aeoleted Fetigue USED FOH HIGH-LEVEL WASTE PACKAGING.First Quarter y ReportYear Four Apnl4une 1985. NUREG 1155 V01: RESEARCH PROGRAM PLAN Reactor Vessele.

NUREG 1155 V03: RESEARCH PROGRAM PLAN Pipeg.

Containment NUREG/CR-3878: PROBA810TY BASED LOAD COM81NATON CRITE- RE 4398: COST ANALYSIS OF REVISIONS TO to CFR PART NU CR 9 2 SCOUO EO IP E 50, APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY ATCH SE L KA E NUREG/CR-4081: ABSORPTION OF GASEOUS IODINE BY WATER CONTAINMENTS OF UGHT WATER 400 LED NUCLEAR POWER DROPLETS. PLANTS.

NUREG/C44119- INTEGRITY OF CONTAINMENT PENETRATIONS UNDER SEVERE ACCIDENT CONDITONS FY84 ANNUAL REPORT Coote NUREG/CR-4137: " RETEST PREDICTONS FOR THE RESPONSE OF NUREG/CR-4268. RATIO METHOOS FOR COST EFFECTIVE FIELD A 1:8-SCALE f (EEL LWR CONTAINMENT BUILDING MODEL TO SAMPUNG OF COMMERCIAL RADCACTIVE LOW LEVEL WASTES.

STATIC OVER RESSURl2ATION. NUREG/CR4398. COST ANALYSIS OF REVISONS TO 10 CFR PART NUREG/CR-4143: REVIEW AND EVALUATON OF THE MILLSTONE 50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY UNIT 3 PROBA81USTIC SAFETY STUDY. Containment Failure CONTAINMENTS OF LIGHT WATER 400 LED NUCLEAR POWER Modes. Radiological Source-Terme And Offs to Consequences. PLANTS.

NUREG/C44151: INTEGRATION OF EMERGENCY ACTION LEVELS WITH COMBUSTION ENGINEERING EMERGENCY OPERATING Crock PROCEDURESBy Use of Combustion Engineenng Owners Group NUREG/C43880 V04: PROBABluTY OF PIPE FAILURE IN REACTOR Emergency Operating Procedure Technical Guidelir,es. COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4 Pipe NUREG/CR4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART Failure Induced By Crack Growth in West Coast Plants.

50. APPENDIX J, LEAK TESTS FOR PRIMARY AND SECONDARY NUREG/C44275: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM CONTAINMENTS OF UGHT WATERCOOLED NUCLEAR POWER . FIVE YEAR PLAN FY 1984 1988.

PLANTS. NUREG/C44305: COMMENTS ON THE LEAK-8EFORE. BREAK CON-CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS.

i Contaminetten NUREG/CR-4251 V01: MITIGATIVE TECHNIQUES FOR GROUND- Creek Growth WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR42tn V01: HEAVY SECTION STEEL TE;HNOLOGY PRO-ACCIDENTS. Volume 1-Analyse Of Genenc $<te Conditions.

GRAM eWANNUAL PROGRESS REPORT FOR OCTOOFR 1984 .

NUREG/CR-4251 V02. MITIGATIVE TECHNIOUES FOR GROUND- *

• We 1985.

WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAH NUREG/CR4287: ENVIRONMENTALLY ASSISTED CRACKING IN ACCIDENTS. Volume 2. Case Study Analysis Of Hydrologic Character. UGHT WATER REACTORS. Annual Report,0ctober 1983 September tretion And Metigettve Schemes. 1984.

I

4 1 44 Subject Index Crew Portormance esecrete nament NUREG/CR4280 THE EFFECTS OF SUPERVISOR EXPERIENCE AND NUREG/CR-3442: RADTWO A COMPLTER CODE FOR SIMULATING ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW FAST TRANSIENT, TWO-DIMENSICNAL.TWO LAYER RADIONU-PERFORMANCE IN CONTROL ROOM SIMULATORS. CUDE CONCENTRATON CONDITONS IN LAKES, RESERVOIRS, RIVERS. ESTUARIES.AND COASTAL REGIONS.

3 NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU- Disposed CLEAR POWER PLANT CONTROL ROOM CREWS. NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE-RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE DEGS

+ NUREG/CR-4290 V02: PROBASIUTY OF PIPE FAILURF IN THE REAC, REGULATOHY PROGRAM.

TOR COOLANT LOOPS OF BABCOCK AND 'WILCOX PWR g PLANTSvolume 2: Guillotine Break Indirectly induced By Earthquakes.

NUREG/CP 0066: PROCEEDINGS OF AN INTERNATONAL WORK-DemeGe Metigeson SHCP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION NUREG 1155 V03: RESEARCH PROGRAM PLAN. Piping. (ALARA) AT NUCLEAR POWER PLANTS.MAY 294UNE 1,1984.

NUREG/CR4239: ANALYSIS OF THE ABluTY OF CURRENT HEALTH Demper Response Time PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID-NUREGiCR-4232- THE RESPONSE OF VENTILATION DAMPERS TO UALS.

LARGE AIRFLOW PULSES. NUREG/CR4266: STANDARD BETA PARTICLE AND MONOENERGE-TIC ELECTRON SOURCES FOR THE CAL!BRATON OF BETA-RADI-REG /CR-4182 VERIFICATON OF SOIL STRUCTURE INTERACTON METHODS Does Reduction NUREG/CR4254: OCCUPATONAL DOSE REDUCTION AND ALARA Data W AT NUCLEAR POWER PLANTS. Study On High-Dose Jobs,Radwaste NUREG/CR4138. DATA ANALYSES FOR NEVACA TEST SITE (NTS) Handling,And ALARA incentives.

PREMIXED COMBUSTION TESTS.

Data Sese Doeimeter NUREG-1148: NUCLEAR POWER PLANT FIRE PROTECTON RE- NUREG/CR-4239: ANALYSIS OF THE ABluTY OF CURRENT HEALTH SEARCH PROGRAM, PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID-NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME- UALS.

TRY IMPAOVEMENT PROGRAM. LWR Power Reactor Survedlance Physcstesimetry Data Base Compendiurrt Doelmetry NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL NUREG/CR 3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME-ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON- TRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveillance TAMINATION. Physcs-Dossmetry Data Base Compendium.

NUREG/CR4185: AN ASSESSMENT OF DOslMETRY DATA FOR AC-E CR-4060- THE DC-1 AND DC-2 DEBRIS COOLABluTY AND RS' MELT DYNAMICS EXPERIMENTS.

Double-Ended Gulliotine greek Deciesen Aid NUREG/CR4200 V02: PROBABILITY OF PIPE FAILURE IN THE REAC-NUREG/CR-4272: RESPONSE TREE EVALUATION EXPERIMENTAL TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR PLANTS. Volume 2. Guillotine Break Indirectly induced By Earthquakes.

OPERATORS.

Drme Decontaminetton NUREG/CR-3413; OFF-SITE CONSEQUENCES OF RADIOLOGICAL NUREG/CR4258: AN APPROACH TO TEAM SKILLS TRAINING OF NO.

ACCIDENTS METilODS, COSTS AND SCHEDULES FOR DECON- CLEAR POWER PLANT CONTROL ROOM CREWS.

TAMINATON Dynamic Analysie NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATONS NUREG/CR 1677 V02: PIPING BENCHMARK PROBLEMS, VOLUME 11 ON SOUDIFICATONWASTE DISPOSAL AND ASSOCIATED OCCU, PATIONAL EXPOSURE. DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTON RE-SPONSE SPECTRUM METHOD.

Defect NUREG 1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO. EOCMEC GRAM PLAN. NUREG/CR4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA AND EASTERN TENNESSEE EARTHOUAKES (19781984).

Degradation NUREG 1155 V02: RESEARCH PROGRAM PLAN Steam Generators. EPICOR-il NUREG/CR4150 EPICOR-il RESIN DEGRADATON RESULTS FROM DominerseseMon System FIRST RESIN SAMPLES OF PF-8 AND PF-20.

NUREG/CR-4150 EPICOR-Il RESIN DEGRADATION RESULTS FROM FIRST RESIN SAMPLES OF PF 8 AND PF 20. ESF NUREG/CR-4130: ICEDF.A CODE FOR AEROSOL PARTICLE CAP-

1. " 8~

EG 358: APPLICATONS OF DENSITY PROFluNG TO EQUIP-MENT OVAUFICATON ISSUES. Earthqualie NUREG/CR-3876: PROBABluTY BASED LOAD COMBINATON CRITE-Donelty Wave RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

NUREG/CR-4118: NUFEGO-NP:A DOITAL COMPUTER CODE FOR NUREG/CR4290 V02: PROBABlWTY OF PIPE FAILURE IN THE REAC.

THE L.NEAR STABlWTY ANALYSIS OF BOluNG WATER NUCLEAR TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR REACTORS' PLANTS Volume 2 Gudlotine Break Indbrectly induced By Earthquakes.

j Deelen Accideng NUREG/CR-4334: AN APPROACH TO THE QUANTIFICATON OF SEIS-MIC MARGINS IN NUCLEAR POWER PLANTS.

NUREG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE.

l NUREG/CR-4339: A REVIEW OF RECENT RESEARCH ON THE SEIS-Deelen Geele Event MOTECTONICS OF THE SOUTHEASTERN SEABOARD AND AN NUREG/CR-3660 V01: PROBABluTY OF PIPE FAILURE IN THE REAC. EVALUATON OF HYPOTHESES ON THE SOURCE OF THE 1866 TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume CHARLESTON. SOUTH CAROLINA EARTHOUAKE.

1. Summary Report. NUREG/CR4354: A STUDY OF SEISMICITY AND TECTONICS IN NEW ENGLAND Final Report.

Deffusion Flame NUREG/CR 3638: HYDROGEN-STEAM JET FLAME FACluTY AND EX- Eddy Current PERIMENTS. NUREG-1155 V02: RESEARCH PROGRAM PLAN Steam Generators.

Subject Index 45 NUREG 1155 V04: RESEARCH PROGRAM PLAN.Non-Destructive Ex- FERRET SAND 11 ammabon. NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME-NUREG/CR-3949 V02 EDDY-CURRENT INSPECTION FOR STEAM TRY IMPROVEMENT PROGRAM LWR Power Reactor Surveillance GENERATOR TUBING PROGRAM. Annual Progress Report For Penod Physacs-Doswnetry Data Base Compendium.

Ending December 31.1984.

FORTRAN 77 Effluente NUREG/CR-4122: A FORTRAN 77 PROGRAM AND USER'S GUIDE NUREG/CR4397: IN-PLANT SOURCE TERM MEASUREMENTS AT FOR THE CALCULATON OF PARTIAL CORRELATON AND STAND-l PRAIRIE ISLAND NUCLEAR GENERATING STATION. ARDlZED REGRESSION COEFFICIENTS.

I Elastic-Pteetle Fracture Mechar4es FRAC

/ 2V2 E RADE P PIN P RAM - PHASE NUREG/CR4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER ll. Semiannual Report. October 1984 March 1985. PLANT VALVE FAILURE-RATE VARIA81UTY-SOME PREUMINARY RESULTS.

Electric Motor Operated Velves NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC FRAPCON MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY-FEA- NUREG/CR4318 V01: REACTOR SAFETY RESEARCH TURE SYSTEMS OF NUCLEAR POWER PLANTS. PROGRAMS.Quarterfy Report. January-March 1965-Emergency Core Cooung Failure NUREG/CR-4252 INDEPENDENT ASSESSMENT OF TRAC-PO2/ MODI NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-CODE WITH BCL ECC BYPASS TESTS. GRAM PLAN.

NUREG/CR-3660 V04: PROBABluTY OF PIPE FAILURE IN REACTOR Emergency Operating Procedure COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4 Pipe

, NUREG-080013.5.2 R1: STANDARD REVIEW PLAN FOR THE REVIEW Failure induced By Crack Growth in West Coast Plants.

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE.

PLANTS. LWR Edition.Rewsion 1 To Section 13.5.2, "Operabng And NUREG/CR-4217; A STATISTICAL ANALYSIS OF NUCLEAR POWER Mantenance Procedures," and Revision 0 of Appendix A to Section PLANT VALVE FAILURE RATE VARIA81UTY-SOME PREUMENARY 13 5.2,"RenewJ RESULTS.

NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC Emergency Preparednese MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY-FEA.

4 NUREG/CR4151: INTEGRATON OF EMERGENCY ACTON LEVELS TURE SYSTEMS OF NUCLEAR POWER Pt>NTS.

WITH COMBUSTION ENGINEERING EMERGENCY OPERATING NUREG/CR4305: COMMENTS ON THE LEAK-BEFORE-BREAK CON-PROCEDURESBy Use of Combustion Engineenng Owners Group CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS.

Emergency Operatmg Procedure Technical Guidelmes. NUREG/CR-4314: BRIEF SURVEY AND COMPARISON OF COMMON CAUSE FAILURE ANALYSIS.

Enforcement Acuon NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES

, NUREG-0940 V04 N02- ENFORCEMENT ACTONS SIGNIFICANT AC" ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE l

TONS RESOLVED,Ouarter?y Progress Report,Apni-June,1985. PRESSURIZED WATER REACTOR.Me n Report j Engineered Safety Feature pg,,, g,,%

1 NUREG/CR-4130: ICEDF:A CODE FOR AEROSOL PARTICLE CAP-

<' TURE IN ICE COMPARTMENTS. NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACluTIES.

NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC

^ " ^ ^

T RE Y MS N R R '

REG 4 7AS TISTICAL ANALYSIS OF NUCLEAR POWER 1 PLANT VALVE FAILURE-RATE VARIAB!UTY-SOME PREUMINARY Environment RESULTS.

NUREG/CR4138: DATA ANALYSES FOR NEVADA TEST SITE (NTS)

PREMIXED COMBUSTON TESTS. Fault i Environmental Asseeement MOTECTONICS OF THE SOUTHCASTERN SEABOARD AND AN NUREG 1157: ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF EVALUATON OF HYPOTHESES CN THE SOURCE OF THE 1886

_ SOURCE MATERIAL UCENSE NO. SUB-1010 Docket No. 40-8027. CHARLESTON. SOUTH CAROUNA EARTHOUAKE.

1 (Sequoyah Fuels Corporabon)

Fault Tree Equation Treneformation System NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE.

NUREG/CR4213. SETS REFERENCE MANUAL QUENCE INFORMATON.

Equipment Hatch NUREG/CR4137; PRETEST PREDICTIONS FOR THE RESPONSE OF Fault Zone A 1.8-SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO NUREG/CR-4333: STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL-STATIC OVERPRESSURIZATION. UNOIS.

Equipment Quanfication Ferrttic Steel NUREG/CR4358: APPLICATIONS OF DENSITY PROFIUNG TO EQUIP-NUREG/CR 4219 V01: HEAVY SECTON STEEL TECHNOLOGY PRO-MENT OUAUFICATION ISSUES.

GRAV SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984

• l Event Tree MARCH 1985 NUREG/CR-4377 EVALUATIONS AND UTluZATIOM OF RISK IM- NUREG/CR4275: HEAVY-SECTON STEEL TECHNOLOGY PROGRAM PORTANCES. FIVE YEAR PLAN FY 1984-1988.

Esemination Filtration NUREG 1122: KNOWLEDGES AND ABlWTIES CATALOG FOR NUCLE. NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY AA POWER PLANT OPERATORS Pressunred Water Reactors. PROTECTIONlit. SUBMICRON PARTICLE TESTS AND

SUMMARY

OF QUAUTY FACTORS.

Esport System NUHEG/CR4272: RESPONSE TREE EVALUATION EXPERIMENTAL Final Environmental Statement ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR NUREG 1094: FINAL ENVIRONMENTAL STATEMENT RELATED TO OPERATORS. THE OPERATION OF BEAVER VALLEY POWER STATON, UNIT j 2. Docket No. 50-412. (Duquesne UQht Company)

Esposure Finite Element Method NUREG/C43444 V02: THE IMPACT OF LWR DECONTAMINATIONS ON SOUDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU- NUREG/CR4182: VERIFICATON OF SOIL STRUCTURE INTERACTON l PAflONAL EXPOSURE- METHODS.

i, 4

_ . _ . _ _ _ _ _ _ _ , _ _ _._____m

t 46 Subject Index 1

Fire Protection Reeeerch Fuel Cycle NUREG-1144: NUCLEAR POWER PLANT FIRE PROTECTON RE- NUREG/CR-4232: THE RESPONSE OF VENTILATION DAMPERS TO SEARCH PROGRAM. LARGE AIRFLOW PULSES.

i Fiesion Product Fuel Damage NUREG/CR-3980 V04: LIGHT-WATER-REACTOR SAFETY FUEL SYS- NUREG/CR4037: DATA

SUMMARY

REPORT FOR FISSON PRODUCT TEMS RESEARCH PROGRAMS. Quarterty Progress Report,0ctober- RELEASE TEST Hi-5.

December 1984.

NUREG/CR4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT FuelRod RELEASE TEST HI S. NUREGICR4948- FDPERIMENTAL RESULTS OF THE OPERATIONAL NUREG/CR4081: ABSORPTION OF GASEOUS ODINE BY WATER TRANS!ENT (OPTRAN) TESTS 11-AND 12 IN THE POWER BURST 4 DROPLETS. FACluTY.

, NUREG/CR4085 USERS MANUAL FOR CONTAIN 1.0.A Computer NUREG/CR4037: DATA

SUMMARY

REPORT FOR FISSION PRODUCT Code for Severe Reactor Acodent Containment Analysis. RELEASE TEST Hi 5.

Flee 6on Product Retention s ukushima Data r NUREG/CR4130 ICEDF:A CODE FOR AEROSOL PARTICLE CAP- NUREG/CP.-4182: VERIFICATION OF SOIL STRUCTURE INTERACTON j TURE IN ICE COMPARTMENTS. METHODS.

Five-Year Reeeerch Plan 082 NUREG 1080 V02: LONG RANGE RESEARCH PLAN FY 1986 FY 1990. NUREG/CR-3901: DOCUMENTATON AND USER'S GUIDE.GS2 & GS3

. VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS.

NUREG/CR-3915: ACOUSTIC EMISSON RESULTS OBTAINED FROM ass TESTING THE 2B 1 INTERMEDIATE SCALE PRESSURE VESSEL NUREG/CR-3901: DOCUMENTATON AND USER'S GUIDE.GS2 & GS3 NUREG/CR4219 V01: HEAVY-SECTION STEEL TECHNOLOGY PRO- . VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS.

, GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 MARCH 1985. Generte Safety leaue NUREG/CR4275: HEAVY SECTION STEEL TECHNOLOGY PROGRAM NUREG-0933 S03. A PRIORITIZATON OF GENERIC SAFETY ISSUES.

. FIVE. YEAR PLAN FY 1984-1988.

NUREG/CR4284: NEUTRON EXPOSURE PARAMETERS FOR THE Genetic Effects j FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATION NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER j SERIES. PLANT ACCIDENT CONSEOUENCE ANALYSIS.Part NUREG/CR4304: PRESSURE VESSEL FRACTURE STUDIES PER- untroduction.Integrat on & Summary Part II. Scientific Basis For Health j TAINING TO THE PWR THERMAL SHOCK ISSUE Expenment TSE 7. Effects Models~

, NUREG/CR4325: A PARAMETRIC STUDY OF PWR PRESSURE VESSEL INTEGRITY DURING OVERCOOUNG Geochemical naa.eang j ACCIDENTS.CONSIDERING BOTH 2-D AND 3-D FLAWS- NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI-CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR

• • • E UR G CR4300 V01: ACOUSTIC EMISSION / FLAW RELATONSHIP

< FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE

] VESSELS. Progress Report, October. March 1985. GeophyelcalInterpretation

( NUREG/CR-3736. FIELD AND THEORETICAL INVESTIGATONS OF FRACTURED CRYSTALUNE ROCK NEAR ORACLE. ARIZONA.

A / 45 V03: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION . ANNUAL REPORT.(Octobe' Groundwater 1982. September 1983. Volume 3)- NUREQ/CR-3851 V04. EVALUATON OF RADIONUCUDE GEOCHEMI-NUREG/CR4288. FOCAL MECHANISM ANALYSES FOR VIRGINIA CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR AND EASTERN TENNESSEE EARTHOUAKES (19781984T WASTE REPOSITORY SITE PROJECTS Annual Progress Report For

' October 1983-September 1984 Fracture Mechanico NUREG/CR-4251 V01: MITIGATIVE TECHNIQUES FC3 GROUNO-

' NUREG 1155 V01:RESEARCH PROGRAM PLAN Reactor Vessels WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR4082 V02: DEGRADED PIPING PROGRAM . PHASE ACCIDENTS Volume 1: Analysis Of Genenc Site Conditions.

i 11 Sermannual Report October 1984 . Marcti 1985 h

NUREG/CR4219 V01: HEAVY SECTON STEEL TECHNOLOGY PRO, NUREG/CR4251 V02: MITIGATIVE TECHNIOUES FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR GRAM SEMtANNUAL PROGRES6 REPORT FOR OCTOBER 1984 ACCIDENTS. Volume 2 Case Study Analysis Of Hydrologic Character-MARCH 1985.

i NUREG/CR-4249 PRESSURE VESSEL FRACTURE STUDIES PENE. Ization And Mitgative Schemes.

TRATING TO THE PWR THERMAL SHOCK ISSUE EXPERIMENTS 0"'d'""

TSE-5,TSE 5A AND TSE 6 NUREG/CR-2800 $03: GUIDEUNES FOR NUCLEAR POWER PLANT NUREG/CR 4275: HEAVY SECTON STEEL TECHNOLOGY PROGRAM

' SAFETY ISSUE PRIORITIZATION INFORMATON DEVELOPMENT.

' . FIVE YEAR PLAN FY 1984 1988' NUREG/CR 4125 V01: GUIDEUNES AND WORKBOOK FOR ASSESS-NUREG/CR4304: PRESSURE VESSEL FRACTURE STUDIES PER.

1 TAINiNG TO THE PWR THERMAL SHOCK ISSUE Expenment TSE 7. MENT OF ORGANI2ATON AND ADMINISTRATION OF UTluTIES

/

NUREG/CR4305: COMMENTS ON THE LEAK BEFORE-BREAK CON. SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PLANT. Volume 1: Guidelines For Unlity Organization And Administration i CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS.

NUREG/CR4325: A PARAMETRIC STUDY OF PWR PRESSURE Plan.

4 VESSEL INTEGRITY DURING OVERCOOUNG NUREG/CR4125 V02: GUIDEUNES AND WORKBOOK FOR ASSESS.

ACCIDENTS.CONSIDERING BOTH 2-D ANO 3-D FLAWS. MENT OF ORGAN 12ATON AND ADMINISTRATON OF UTILITIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER 1

Froude Number PLANT. Volume 2:Wortibook For Assessment Of Organizahon And Man-i NUREG/CR-3426 Vot: THERMAL AND FLUID MIXING IN 1/2-SCALE agement.

TEST FACluTY. Fac4 sty And Test Design Report. NUREG/CR4151: INTEGRAflON OF EMERGENCY ACTON LEVELS NUREG/CR-3426 V02 THERMAL AND FLUlu MIXING IN 1/2-SCALE WITH COMBUSTON ENGINEERING EMERGENCY OPERATING TEST FACluTY.Deta Report. PROCEDURESBy use of Combustion Engineenng Owners Group i

Emergency Operahng Procedure Technical Guide 4ines .

Fuel Claddin9 NUREG/CR4227; HUMAN ENGINEERING GUIDEUNES FOR THE NUREG/CR-3900 V04: LONG TERM PERFORMANCE OF MATERIALS EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS.

' USED FOR HIGH-LEVEL WASTE PACKAGING Annual Report.Apnl I 1984. April 1985 HPI Jet ,

l NUREG/CR4151: INTEGRATON OF EMERGENCY ACTON LEVELS NUREG/CR-3426 V01: THERMAL AND FLUID MIXING IN 1/2-SCALE WITH COMBUSTON ENGINEERING EMERGENCY OPERATING TEST FACluTY. Facdity And Test Design Report.

PROCEDURES By Use of Combusbon Engineenng Owners Group NUREG/CR 3426 V02: THERMAL AND FLUO MIXING IN 1/2 SCALE Emergency Operahng Procedure Technical Guidelines. TEST FACluTY. Data Report.

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

i 1

[ Subject Index 47 HgST Human Factore NUREG/CR4284: NEUTRON EXPOSURE PARAMETERS FOR THE NUREG/C43481 V02: NUCLEAR POWER PLANT PERSONNEL QUAll-FIFTH HEAVY SECTON STEEL TECHNOLOGY IRRADIATION FICATIONS AND TRAINING: TAPS - The Task Analysis Profiling SERIES. System.

NUREG/C44227: HUMAN ENGINEERING GUIDEUNES FOR THE HTGR EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS.

NUREG/CR-3885 V04: HIGH-TELPERAN9E GAS COOLEQ REACTOR NUREG/CR4248: RECOMMENDATIONS FOR NRC POUCY ON SHIFT SAFETY STUDIES FOR THE OfVISON OF ACCIDENT SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS.

EVALUATION.Ouarterty Proyess Report, October 1-December NUREG/CR-4272: RESPONSE TREE EVALUATON EXPERIMENTAL 31,1984. ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR OPERATORS.

NUREG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE. Hydrogeologic Site Cherectertzetton NUREG/CR-4251 V01: MITIGATIVE TECHNIOUES FOR GROUND-Heefth Effecto M WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR-4214. HEALTH EFFECTS MODEL FOR NUCLEAR POWER ACCIDENTS Volume 1: Analysis Of Genenc Site Cordtions.

PLANT ACCIDENT CONSEQUENCE ANALYSIS Part NUREG/CR-4251 V02

MITIGATIVE TECHNOUES FOR GROUND-

1. Introduction. Integration & Summary Part II.Soentific Basis For Health WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR Effects Models. ACCIDENTS. Volume 2. Case Study Analysis Of Hydrologic Character.

iration And Mitigatrve Schemes.

Hoenh %

NUREG/CR4239- ANALYSIS OF THE ABILITY OF CURRENT HEALTH Hydrology 1

PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID- NUREG/C43736: FIELD AND THEORETICAL INVESTIGATIONS OF j UALS. FRACTURED CRYSTALUNE ROCK NEAR ORACLE, ARIZONA.

! Heat Transfer ICEDF NUREG/CR-3633 V04. TRAC-BD1/ MOD 1.AN ADVANCED BEST ESTI-NUREG-0956 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES MATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR FOR ESTIMATING SOURCE TERMS (Draft Report For Commentt NUR / 6H T TR SF E A F BACK IN PWR STEAM GENERATORS DURING TRANSIENTS.

TURE NIC COM ARTME ~

Heen Section Steel Technology IE Bulletin 81-01 NUREG/CR4264. NEUTRON EXPOSURE PARAMETERS FOR THE NUREG/CR4006. CLOSEOUT OF IE BULLETIN 8101. SURVEILLANCE FIFTH HEAVY SECTON STEEL TECHNOLOGY IRRADIATON OF MECHANICAL SNUBBERS.

SERIES-Ice Condeneer High Confidence Of Low Probabil6ty Of Failure NUREG/CR4130: ICEDF:A CODE FOR AEROSOL PARTICLE CAP-NUREG/CR-4334. AN APPROACH TO THE QUANTIFICATON OF SEIS. TURE IN ICE COMPARTMENTS.

MIC MARGINS IN NUCLEAR POWER PLANTS. y,,,,,

High Pressure irgection NUREG/CR4333: STE. GENEVCVE FAULT ZONE. MISSOURI AND IL-NUREG/C43426 V02: THERMAL AN' FLUID MIXING IN 1/2-SCALE LINOIS.

TEST FACluTY Data Report High Temperature Combustion NUREG/C44357: THE FEASIBIUTY OF DE*ECTING THE IMPORT OF NUREG/C43638 HYDROGEN-STEAM JET FLAME FACluTY AND EX. UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED PERIMENTS. STATES.

i High Ooes Job , A EC.t Aseeeement NUREG/CR4254. OCCUPATONAL DOSE REDUCTON AND ALARA NUREG/C44252: INDEPENDENT ASSESSMENT OF TAAO-PD2/ MODI AT NUCLEAR POWER PLANTS Study On High-Dose Jobs.Radwaste CODE WITH BCL ECC BYPASS TESTS.

Handhng.And ALARA incontrves.  !

High-Level Weste NUREG-0304 V10 NO2: REGULATORY AND TECHNICAL NUREG/C43851 V04: EVALUATION OF RADONUCLOE GEOCHEMI- REPORTS Compelation For Second Quarter 1985 CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NUREG4750 V21 102: INDEXES TO NUCLEAR REGULATORY COM-WASTE REPOSITORY SITE PROJECTS Annual Progress Report For MISSION ISSUANCES. January 4une 1985.

October 1963-September 1984 4

NUREG/C43900 V04: LONG TERM PERFORMANCE OF MATERIALS Ind6ene

, USED FOR HIGH-LEVEL WA9TE PACKAGING Annual ReportAprg NUREG/CR-3145 V03: GEOPHYSICAL INVESTIGATIONS OF THE

{ 1984.p etp 1985. WESTERN OHIO-INDIANA REGION ANNUAL REPORT (October NUREG/CH4303: HOH-LEVEL WASTE PRECLOSURE SYSTEMS 1982. September 1963, Volume 3).

SAFETY ANALYSIS Phase 1, Final Report.

NUREG/CR-4379 V01: LONG TERM PERFORMANCE OF MATERIALS Informed Opinion USED FOR HIGH-LEVEL WASTE PACKAGING First Quarterty NUREG/CP-0063: PROCEEDINGS OF THE 1984 STATISTICAL SYMPO-Report, Year Four AprW4une 1985. SIUM ON NATIONAL ENERGY ISSUES.

l High Temperature Ineervice inepection q NUREG/CR4037. DATA SMMMARY REPORT FOR FISSION PRODUCT NUREG 1155 V02: RESEARCH PROGRAM PLAN Steam Generators RELEASE TEST Hi-S. NUREG 1155 V04: RESEARCH PROGRAM PLAN NorvDestructive Ex-1 emenation.

High Temperature GeoCooled Reactor NUREG/C43865 V04. HIGH TEMPERATURE GAS-COOLED REACTOR inspection SAFETY STUDIES FOR THE DIVISION OF ACCOENT NUREG-0040 V09 NO2: LICENSEE CONTRACTOR AND VENDOR IN-i EVALUATION Ouarterty Progress Report. October 1 December SPECTION STATUS REPORT. Quarterfy Report.Apnl4une 1985 31,1964. (White Book)

NUREG 1144. NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-Human Engineering GRAM PLAN NUREG/C44227: HUMAN ENGINEERING GUOELINES FOR THE NUREG-1155 V02: RESEARCH PROGRAM PLAN Steam Generators EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS NUREG/C43949 V02: EDDY CURRENT INSPECTION FOR STEAM GENERATOR TUB NG PROGRAM Annual Progress Report For Penod Human Error Ending December 11,1984.

NUREG/CR4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY FEA- 1 PERFORMANCE IN CONTROL ROOM SIMULATORS. TURE SYSTEMS OF NUCLEAR POWER PLANTS.

48 Subject index Instrumentation Legal 1seuances NUREG/CR4239: ANALYSIS OF THE ABILITY OF CURRENT HEALTH NUREG-0750 V21 N05 NUCLEAR REGULATORY COMMISSION IS-PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID- SUANCES FOR MAY 1985. Pages 1.043-1.567.

VALS. NUREG-0750 V21 N06: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR JUNE 1985. Pages 1.569-1,786 Interactions Testing NUREG-0750 V22 N01: NUCLEAR REGULATORY COMMISSION IS-NUREG/CR 3710: LABORATORY STUDIES OF A BREACHED NUCLE- SUANCES FOR JULY 1985. Pages 1 176.

AR WASTE REPOSITORY IN BASALT.

Licensed Operating Reactore Intergranular Strees Corrosion Cracking NUREG-0020 V09 N06. LICENSED OPERATING REACTORS STATUS NUREG/CR-3613 V03 Nt: EVALUATION OF WELDED AND REPAIR-

SUMMARY

REPORT. Data As Of May 31,1985 (Gray Book 1)

WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report NUREG4020 V09 N07: LICENSED OPERATING REACTORS STATUS For Ovober 1984 Through March 1985.

SUMMARY

REPORT Data As Of June 30.1985 (Gray Book 1)

NUREG-0020 V09 N08: LICENSED OPERATING REACTORS STATUS lodis" SJMMARY REPORT Data As Of July 31.1985 (Gray Book 1)

NUREG/CR4081: ABSORPTION OF GASEOUS IODINE BY WATER DROPLETS. Licensee Contractor And Vendor inspection NUREG-0040 V39 NO2; LICENSEE CONTRACTOR AND VENDOR IN-lon Enchange Reelne SPECTION STATUS REPORT. Quarterty Report,Aptd-June 1985.

NUREG/CR4150: EPICOR-il RESIN DEGRADATON RI JULTS FROM (White Book)

FIRST RESIN SAMPLES OF PF-8 AND PF 20.

Licensee Event Report Irradiated Zircoloy Cladding NUREG 1022 S02: LICENSEE EVENT REPORT SYSTEM Evaluation Of NUREG/CR-3980 V04. LIGHT-WATER-REeTOR 'AFETY FUEL SYS- First Year Results And Recommendehons For Improvements TEMS RESEARCH PROGRAMS. Quarterfy hgess Report, October- NUREG/CR-2000 V04 N6. LICENSEE EVENT REPORT (LER)

December 1984. COMPILATON For Month Of June 1985.

NUREG/CH.2000 V04 N 7. LICENSEE EVENT REPORT (LER)

Irradianon COMPILATON For Month Of July 1985 NUREG-1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels. NUREG/CR-2000 V04 N8: LICENSEE EVENT REPORT (LER)

NUREG/CR4219 Vol. HEAVY SECTION STEEL TECHNOLOGY PRO- COMPILATON,For Month Of August 1985' GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

MARCH 1985. Licensee Security NUREG/CR4239: ANALYSIS OF THE ABlUTY OF CURRENT HEALTH NUREG/CR4250: VEHICLE BARR ERS EMPHASIS ON NATURAL FEA.

PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID- TURES' UALS.

NUREG/CR-4275. HEAVY SECTON STEEL TECHNOLOGY PROGRAM Load Factor R / 3876; NABW BASO WAD MMM ND NbREGI 4284 E TRON E POSURE PARAMETERS FOR THE ^ "

FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATON SERIES. Long-Range Research Plan NUREG 1080 V02: LONG-RANGE RESEARCH PLAN FY 1986-FY 1990.

J-Integrel/Teering Modufue NUREG/CR408' V02: DEGRADED PtPING PROGRAM . PHASE Lose Of Main And Auxiliary Feedwater it Semianrir., Report, October 1984 - March 1985. NUREG 1154. LOSS OF MAIN AND AUXIUARY FEEDWATER EVENT AT THE DAVIS-BESSE PLANT ON JUNE 9,1985.

LER NUREG-1022 S02: UCENSEE EVENT REPORT SYSTEM Evaluate Of Lose-Of-Coolant Accident NUREG/ 2000 VO N V NT R-3426 W2: NEN AW M MNG IN U2 SW S P RT (LER) TEST FACILITY Data Report.

COMPILATION For Month Of June 1985 NUREG/CR 3633 V01 S1: TRAC-BD1/ MODI AN ADVANCED BEST ES-NUREG/CR 2000 V04 N T: LICENSEE EVENT REPORT (LER) TIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR COMPILATION For Month Of July 1985.

TRANSIENT ANALYSIS.

NUREG/CR-2000 V04 N8: LICENSEE EVENT REPORT (LER)

COMPILATION.For Month Of August 1985. g,,,g,,,, y,,,,

LOCA NUREG/CR-4150: EPICOR-ll RESIN DEGRADATION RESULTS FROM NUREG/CR 3633 V01 S1: TRAC-001/ MOD 1:AN ADVANCED BEST ES- FIRST RESIN SAMPLES OF PF-8 AND PF 20 NUREG/CR4268: RATIO METHODS FOR COST-EFFECTIVE FIELD T' MATE COMPUTER PROGRAM FOR BOILING W ATER REACTOR SAMPLING OF COMMERCIAL RADIOACTIVE LOW-LEVEL WASTES T1ANSIENT ANALYSIS' NUREG/CR-4352: SUGGESTED STATE REQUIREMENTS AND CRITE-LeachTeog RIA FOR A LOW LEVEL RADOACTIVE WASTE DISPOSAL SITE NUREG/CR4379 V01: LONG-TERM PERFORMANCE OF MATERIALS REGULATORY PROGRAM.

USED FOR HIGH-LEVEL WASTE PACKAGING First Quartedy Lumped Parameter Method Report, Year Four Apni-June 1985.

NUREG/CR4182: VERIFICATON OF SOIL STRUCTURE INTE9 ACTION Leachings METHODS, NUREG/CR4259 TAILINGS NEUTRAtl2ATION AND OTHER ALTER. E NA ES FOR IMMOOlu2lNG TOXIC MATERIALS IN TAILINGS Final NUREG4958 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

Leak NUREG 1155 V04. RESEARCH PROGRAM PLAN NorvDestructive Ex.

MELCOR NUREG/CR4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC-sminataon.

NUREG/CR-3952: SEOOOYAH EQUIPMENT HATCH SEAL LEAKAGE. CIDENTAL RADIONUCLIDE RELEASES FROM NUCLEAR REAC-NUREG/CR4294 LEAK RATE ANALYSIS OF THE WESTINGHOUSE TORS.

REACTOR COOLANT PUMP.

NUREG/CR4398. COST ANALYSIS OF REVISONS TO 10 CFR PART MERGE 50, APPENDIX J. LEAK TESTS rOR PRIMARY AND SECONDARY NUREG-0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES CONTAINMENTS OF UGHT.W TER400 LED NUCLEAR POWER FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment)

PLANTS.

MINTEO Leak Before-Broek NUREG/CR-3851 V04 EVALUATION OF RADIONUCLlDE GEOCHEMI-l NUREG 1155 V03. RESEARCH PROGRAM PLAN Ppng CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR l WASTE REPOSITORY SITE PROJECTS Annual Progress Report For NUREG/CR 4305. COMMENTS ON THE LEAK-BEFORE BREAK CON-CEPT FOR NUCLEAR POWER PLANT P1 PING SYSTEMS. October 1983 September 1984

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

l l

l Subject Index 49 Ihintenance NAUA NUREG/CR4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR NUREG-0958 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

Melfunc6on NTS Hydrogen Dewar NUREG-1154. LOSS OF MAIN AND AUXfuARY FEEDWATER EVENT NUREG/CR-4138: DATA ANALYSES FOR NEVADA TEST SITE (NTS)

AT THE DAV'.G 8 ESSE PLANT ON JUNE 9,1985. PREMIXED COMBUSTION TESTS.

• D NUFREO-NP NUREG/CR4260:TORAC USER'S MANUALA Computer Code For Ana-Tornado-induced Flow And Matenal Transport in Nuclear Facril- NUREG/CR4118: NUFEGO.NP-A DIGITAL COMPUTER CODE FOR

~

THE UNEAR STABlUTY ANALYSIS OF BOluNG WATER NUCLEAR REACTORS.

Meteriale Control Unit NUREG/CR4107: SEQUENTIAL TEST PROCEDURES FOR DETECT- Natural Convection Core Cooling ING PROTRACTED iwATERiALS LOSSES- NUREG/CR-4240 V01: PHYSICS OF REACTOR SAFETY.Ouarterly Report. January-March 1985.

Materiale Property Test NUREG/CR4318 V01: REACTOR SAFETY RESEARCH Neutrellaation PROGRAMS.Ouarterty Report. January-March 1985. NUREG/CR-4259: TAluNGS NEUTRAUZATION AND OTHER ALTER-I NATIVES FOR IMMOB!UZING TOXIC MATERIALS IN TAluNGS. Foal

! Moseurement nepo,,.

I NUREG/CR-3481 V02: NUCLEAR POWER PLANT PERSONNEL QUAU.

I FICATIONS AND TRAINING. TAPS - The Task Analysts Profilm0 Neutron Doeimetry System- NUREG-1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels.

NUREG/CR-3609 EVALUATON OF NEUTRON DOSIMETRY TECH- NUREG/CR-3609: EVALUATON OF NEUTRON DOSIMETRY TECH-NIQUES FOR WELL LOGGING OPERATONS.

NU 439 I TS RCE T MEASUREMENTS AT PRAIRIE ISLAND NUCLEAR GENERATING STATION. New Madrtd NUREG/CR4333: STE GENEVIEVE FAULT ZONE, MISSOURI AND IL-Mechanical Snutstiers NUREG/CR-4008: CLOSEOUT OF IE BULLETIN 81-01. SURVEILLANCE LINOIS.

OF MECHANICAL SNUBBERS.

l Nondestructive Evaluation Missourt NUREG 1155 V04. RESEARCH PROGRAM PLAN Non-Destructhre Ex-NUREG/CR-4333: STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL- ammabon.

UNOIS. NUREG/CR4385: DESIGN AND DEVELOPMENT OF A SPECIAL PUR.

POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU-Mmgeuon Techn6quee CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

NUREG/CR4251 V02: MITIGATIVE TECHNIQUES FOR GROUND-WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR OPTRAN 11 ACCIDENTS.Volurne 2 Case Study Analysis Of Hydrologic Character- NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATONAL trabon And Mitigative Schemes. TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST FACluTY.

Metigettve Technique NUREG/CR4251 V01: MITIGATIVE TECHNIOUES FOR GROUND

• OPTRAN 1-2 WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATONAL ACCOENTS. Volume t. Analysis Of Genenc Sete Conditions.

TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST FACIUTY, Modmcanon NUREG/CR4398. COST ANALYSIS OF REVISONS TO 10 CFR PART

50. APPENDIX J. LEAK YESTS FOR PRIMARY AND SECONDARY ORECA NUREG/CR-3885 V04: HIGH-TEMPERATURE GASCOOLED REACTOR CO A NMENTS OF UGHT WATER-COOLED NUCLEAR POWER

^

SAFETY STUDIES FOR THE OtVISON OF ACCIDENT EVALUATION.Quarterty Progress Asport. October 1-December Monitoring 31,1984.

NUREG 0837 V04 N04. NRC TLD DIRECT RADtATON MONITOFING RFPORT. Progress Report, October-December 1984. OMN i NUREG-0837 VOS N01: NRC TLD DIRECT RADtATON MONITOFING NUREG-0958 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES NETWORK. Progress Report, January-March 1985. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

RUREG-1144. NUGLEAR PLANT AGING RESEARCH (NPAR) PRO-GRAM PLAN. Ohio NUREG 1155 V04: RESEARCH PROGRAM PLAN.NonDestructrve Ex- NUREG/CA 3145 V03 GEOPHYSICAL INVESTIGATIONS OF THE emmation. WESTERN OHIO-INDIANA REGON ANNUAL REPORT.(October NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM 1982. September 1983, Volume 3).

TESTING THE Z81 INTERMEDIATE SCALE PRESSURE VESSEL I NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC Operating Esportonce I

MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY FEA* NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC TURE SYSTEMS OF NUCLEAR POWER PLANTS MOTOROPERATED VALVES USED IN ENGINEERED SAFETY FEA.

NUREG/CR4300 V01: ACOUSTIC EMISSON/ FLAW RELATONSHIP TURE SYSTEMS OF NUCLEAR POWER PLANTS' FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS Progress Repnr1 October March 1985. Opereung Procedures NUREG/CR4352: SUGuESIED STATE REQUIREMENTS AND CRITE.

NUREG-080013 5 2 Rt: STANDARD REVIEW PLAN FOR THE REVIEW RfA FOR A LOW-LEVEL RADCACTIVE WASTE DISPOSAL SITE OF SAFETY ANALYSIS REPORTS FOR m EAR POWER NUREG/CR 54 OF SEISMOITY AND TECTONICS IN NEW PLANTS LWR Editen. Revision 1 to Section 13 $ 2. "Operabng And ENGLAND Fmal Report. Mantenance Procedures," and Revision 0 of Appendir A to Section 13 5.2, " Review.. "

Monoonergetic Electron Sources NUREG/CR4288. STANDARD BETA. PARTICLE AND MONOENERGE. Operating Reactore Uconoing Actione Tic ELECTRON SOURCES FOR THE CAUBRATON OF BETA-RADI. NUREG4748 V05 N05: OPERATING REACTORS UCENSING ACTONS ATON PROTECTON INSTRUMENTATON.

SUMMARY

Data As Of May 31,1085 (Orange Book)

NUREG4748 V05 N05. OPERATING REACTORS UCENSING ACTIONS Moodue

SUMMARY

. Data As Of Miy 31,1985. (Orange Book)

NUREG/CR4354 A STUDY OF SEISMICITY AND TECTONICS IN NEW NUREG4748 V05 N07: OPERATING REACTORS UCENSING ACTIONS ENGLAND Final Report.

SUMMARY

.Deta As Of July 3f.1985.(Orange Book)

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

50 Subject Index Operator Error Phyeles-Dosametry NUREG-1154. LOSS OF MAIN AND AUXtLIARY FEEDWATER EVENT NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME-AT THE DAVIS-8 ESSE PLANT ON JUNE 9,1985. TRY IMPROVEMENT PROGRAM. LWR Power Reactor Survedlance Physacs-Dossmetry Data Base Compendium.

NUREG/CR4125 V01: GUIDELINES AND WORKBOOK FOR ASSESS. Pipe Failure MENT OF ORGANIZATON AND ADMINISTRATION OF UTILITIES NUREG/CR-3660 V01: PROBABILITY OF PtPE FAILURE IN THE REAC-SEEKING OPERATING LICENSE FOR A NUCLEAR POWER TOR COOLANT LOOPS OF WESilNGHOUSE PWR PLANTS. Volume PLANT. Volume 1: Guidelines For Utdaty Organization And Administration 1: Summary Report.

Plan.

NUREG/CR-3660 V04: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS Volume 4. Pipe NUREG/CR-4125 V02: GUIDELINES AND WORKBOOK FOR ASSESS.

MENT OF ORGANIZATON AND ADMINISTRATION OF UTILITIES Fadure induced By Crack Growth in West Coast Plants NUREG/CR-4290 V02: PROBABILITY OF PIPE FAILURE IN THE REAC.

SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PLANT. Volume 2.Wortibook For Assessment Of Orgarwation And Man. TOR COOLANT LOOPS OF BA8 COCK AND WILCOX PWR PLANTS. Volume 2.Gudiotine Break Ir:directry induced By Earthquakes.

agement.

Piping Overcooling NUREG/CA-3935: THERMAL HYDRAULIC ANALYSES OF OVERCOOL, NUREG 1155 V03: RESEARCH PROG 9AM PLAN -

NUREG/CR-1877 V02 PIPING BENCHMARK PR EMS, VOLUME 11 ING SEOVENCES FOR THE H B. ROBINSON UNIT 2 PRESSURIZE 0 DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTON RE-THERMAL SHOCK STUDY

• SPONSE SPECTRUM METHOD.

oy ,.p.cg NUREG/CR-3613 V03 Nt: EVALUATON OF WELDED AND REPAIR.

WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report NUREG/CR-3900 V04: LONG-TERM PERFORMANCE OF MATERIALS US D OR HIGH-LEVEL WASTE PACKAGING. Annual Report.Apnl NU E CR 08 VO : G ED PIPING PROGRAM . PHASE NUREG/CR-4379 vot: LONG TERM PERFORMANCE OF MATERIALS Nt CR 29 NCLU ON D SU Y REPORT ON PHYSI-USED FOR HIGH-LEVEL WASTE PACKAGING.First Quarterty CAL BENCHMARKING OF PIPING SYSTEMS.

Report Year Four Apnt June 1985. NUPEG/CR-4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSI-CAL BENCHMARKING OF PIPING SYSTEMS.

OverFesaurtzation NUREG/CR-4305: COMMENTS ON THE LEAK-BEFORE-BREAK CON-NUREG/CR-4137. PRETEST PREDICTONS FOR THE RESPONSE OF CEPT FOR NUCLEAR POWER PLANT PlPtNG SYSTEMS A 1.8-SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-STATIC OVERPRESSURIZATON. POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU-CLEAR REACTOR VESSELS AND PIPLNG COMPONE_NTS.

Ovwtime NUREG/CR-4248; RECOMMENDATIONS FOR NRC POLICY ON SHIFT Plant Safety SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS. NUREG/CH-4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY.

Oxygen Diffusion NUREG/CR4358; APPLICATONS OF DENSITY PROFILING TO EQUIP

• Plugging MENT OUAUFICATON ISSUES NURE3-1155 V02: RESEARCH PROGRAM PLAN Steam Generators.

P-V'we Residuale Plutonium NUREG/CR-3145 V03' GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-4355 V01: 238 PU(IV) IN MONKEYS Overview Of Metabo-WESTERN OHOINDIANA REGION . ANNUAL REPORT.(October ham.

1982 September 1983, Volume 3).

Policy PHREEOE NUREG-1070: NRC POLICY ON FUTURE REACTOR NUREG/CR-3851 V04: EVALUATION OF RADIONUCLIDE GEOCHEMI. DESIGNS.Deossons On Severe Accident lasues In Nuclear Pcwer CAL INFORMATON DEVELOPED BY DOE HIGH LEVEL NUCLEAR Plant Regulat on.

WASTE REPOSITORY SITE PROJECTS Annual Progress Report For October 1983-September 1984. Poetirradiation Examination NUREG/CH-4318 V01: REACTOR SAFETY RESEARCH PRA PROGRAMS Quarterty ReportJanuary-March 1985.

NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE-OUENCE INFORMATION Practice And Proceduree Olgoet NUREG/CR-3485: PRA REVIEW MANUAL NUREG4388 003. UNITED STATES NUCLEAR REGULATORY COM-NUREG/CR4331: SIMPUFIED SEISMIC PROBA01USTIC RISK MISSION STAFF PRACTICE AND PROCEDURE DIGESTJULY 1972 ASSESSMENT. Procedures And Umrtations. SEPTEMBER 1983.

NUREG/CR-4350 V02: PROBABiUSTIC RISK ASSESSMENT COURSE DOCUMENTATION Volume 2: Probabdity And Statistics For PRA Appl 6- Promised Combustion Esperiment catior:s. NUREG/CR-4138. DATA ANALYSES FOR NEVADA TEST SITE (NTS)

NUREG/CR-4377: EVALUATONS AND UTILIZATONS OF RISK IM- PREMIXED COMBUSTON TESTS.

PORTANCES.

Preseure Boundary PTS NUREG/CR-3915: ACOUSTO EMISSON RESULTS OBTA* LED FROM NUREG/CR4378. HEAT TRANSFER. CARRYOVER AND FALL BACK IN TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL.

PWR STEAM GENERATORS DURING TRANSIENTS. NUREG/CR4305: COMMENTS ON THE LEAK-BEFORE-BREAK CON-CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS.

Penetration NUREG/CR4119: INTEGRITY OF CONTAINMENT PENETRATIONS Pressure Probability Theory UNDER SEVERE ACCIDENT CONDITONS FY84 ANNUAL REPORT. NUREG/CR-3878: PROBA0luTY BASED LOAD COMBINATON CRITE-RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

Performance Indicatore NUREG/CR4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR Pressure Vesset POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY, NUREG 1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels.

NUREG/CH-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME-Phyelcat 8enchmark TRY IMPROVEMENT PROGRAM LWR Power Reactor Survedlance NUREG/CR-4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSI. Physics Dossenetry Data Base Compendium.

CAL BENCHMARKING OF PIPING SYSTEMS. , NUREG/CR 3915. ACOUSTIC EMISSION RESULTS 00TAINED FROM TESTING THE Z0-1 INTERMEDIATE SCALE PRESSURE VESSEL.

Phyencal Security NUREG/CR-4219 V01: HEAVY SECTON STEEL TECHNOLOGY PRO-NUREG/CR4298: DESIGN AND INSTALLATON OF COMPUTER SYS- GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984

! TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 $5. MARCH 1985.

Subject index 51 NUREG/CR4249 PRESSURE VESSEL FRACTURE STUDIES PENE. Protracted Meteriale Lossee TRATING TO THE PWR THERMAL-SHOCK ISSUE EXPER!MENTS NUREG/CR4107: SEQUENTIAL TEST PROCEDURES FOR DETECT.

. TSE-5,TSE-5A AND TSE ING PROTRACTED MATERIALS LOSSES.

NUREG/CR-4275; HEAVY SECTON STEEL TECHNOLOGY PROGRAM

- FIVE YEAR PLAN FY 1984-1988. Quality Assurance NUREG/CR4304: PRESSURE VESSEL FRACTURE STUDIES PER- NUREG4940 V04 NO2- ENFORCEMENT ACTIONSSIGNIFICANT AC-NU CR 2 A RA R DY ESSUIE VESSEL INTEGRITY DURING OVERCOOUNG RADTWO ACCIDENTS.CONSIDERING BOTH 2-D ANO 3-D FLAWS.

NUREG/CR-4J25; A PARAMETRIC STUDY OF PWR PRESSURE NUREG/CR-3442: RADTWOA COMPUTER CODE FOR SAMULATING VESSEL INTEGRITY FAST TRANSIENT, TWO-DIMENSIONAL,TWO-LAYER RADIONU.

DURING OVERCOOUNG CUDE CONCENTRATION CONDITONS IN NU G 4 284 ET EXPO RE A T S FOR THE LAKES. RESERVOIRS. RIVERS. ESTUARIES.AND COASTAL REGIONS-FIFTH HEAVY SECTON STEEL TECHNOLOGY IRRADIATON Mh SERIES.

NUREG4837 V05 N01: NRC TLD DIRECT RADIATON MONITORING Presourtzed Thermet Shock NETWORK. Progress Report, January March 1&85.

NUREG-1155 V01: RESEARCH PROGRAM PLAN. Reactor Vessels.

NUREG/CR-3935: THERMAL-HYDRAUUC ANALYSES OF OVERCOOL. Radletion MonMoring ING SEQUENCES FOR THE HB. ROBINSON UNIT 2 PRESSURIZED NUREG 0837 V05 NO2: NRC TLD DIRECT RADIATION MONITORING THERMAL SHOCK STUDY. NETWORK. Progress Report, April-June 1985 NUREG/CR4325: A PARAMETRIC STUDY OF PWR PRESSURE VESSEL INTEGRITY DURING OVERCOOUNG Radletion Protection ACCIDENTS.CONSIDERING BOTH 2-D AND 3-0 FLAWS. NUREG/CP-0066: PROCEEDINGS OF AN INTERNAIONAL WORK.

NUREG/CR4376: HEAT TRANSFER, CARRYOVER AND FALL BACK IN SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION PWR STEAM GENERATORS DURING TRANSIENTS. (ALARA) AT NUCLEAR POWER PLANTS.MAY 29-JUNE 1,1984.

Primary Coosent Circuit NUREG/CR-4254. OCCUPATONAL DOSE REDUCTION AND ALARA AT N'JCLEAR POWER PLANTS. Study On High-Dose Jobs.Radwaste NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE Handhng.And ALARA nncontrtes.

REACTOR COOLANT PUMP.

I Radlet6cn Safety Program l

G 51 EGRATION OF EMERGENCY ACTON LEVELS NUREG 0940 V04 N02: ENFORCEMENT ACTIONS SIGNIFICANT AC-WITH COMBUSTION ENGINEERING EMERGENCY OPERATING REMEaQuader9 Rogress ReportA@,M81 PROCEDURESBy Use of Combustion Engineering Owners Group gm g,g,,,g Enwgency Operating Roceh Technical Guidelina NUREG/CR4357: THE FEASIBluTY OF DETECTING THE IMPORT OF pygmates UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED NUREG/CR-4355 VO1: 238 PU(IV) IN MONKEYS Ovennew Of Metabo. STATES.

Radioactive Weste Prierttisetton NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL NUREG/CR-2800 S03: GUIDEUNES FOR NUCLEAR POWER PLANT ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON.

SAFETY ISSUE PRIORITl2ATON INFORMATON DEVELOPMENT. TAMINATON.

NUREG/CR-3710: LABORATORY STUDIES OF A BREACHED NUCLE-ProbabNietic Fracture Mechenece AR WASTE REPOSITORY IN BASALT.

NUREG/CR-3660 V01: PROBABiUTY OF PIPE FAILURE IN THE REAC-TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume Redloolomont Trenoport 1: Summary Report- NUREG/CR-3710; LABORATORY STUDIES OF A BREACHED NUCLE- l

,, g AR WASTE REPOSITORY IN BASALT.

I NUREG/CR-2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE. Medlographore NU EG 28 5' 11 OBAB SAFETY ANALYSIS PROCE.

DURES 1UIDE.Sectione 8-12. TONS RESOLVED Quarterty Progress Report,Apnt-June,1985.

NUREG/Ch 4213. SETS REFERENCE MANUAL NUREG/CR4377; EVALUATIONS AND UTlu2ATIONS OF RISK IM- Radionuc#de Trenopod PORTANCES. NUREG/CR-3442: RADTWO A COMPUTER CODE FOR SIMULATING FAST-TRANSIENT. TWO DIMENSIONAL,TWO-LAYER RADIONU-ProbabNietic Rees Asseeement CUDE CONCENTRATION CONDITONS IN NUREG-1148: NUCLEAR POWER PLANT FIRE PROTECTON RE. LAKES, RESERVOIRS RIVERS, ESTUARIES.AND COASTAL REGIONS.

SEARCH PROGRAM NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE. Medlonucildeo OUENCE INFORMATON. NUREG/CR 3851 V04. EVALUATON OF RADIONUCLIDE GEOCHEMI.

NUREG/CR-3485: PRA REVIEW MANUAL CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NUREG/CR-4143: REVIEW AND EVALUATON OF THE MILLSTONE WASTE REPOSITORY SITE PROJECTS Annual Progret,e Report For UNIT 3 PROBABluSTIC SAFETY STUDY. Containment Failure October 1983 September 1964.

Modes.Radiologscal Source Terme And Offsete Consequences' NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC-NUREG/CH-4331. S6MPUFIED SEISMIC PROBABluSTIC RISK CIDENTAL RADIONUCUDE RELEASES FROM NUCLEAR REAC-ASSESSMENT Procedures And Urrwtations.

" A A 38 TORS M MA GIN IN N AR PO RP S NUREGICA-4251 V01: MITIGATIVE TECHNIOUES FOR GROUND-NUREG/CR4350 V02: PROBABluSTIC RISK ASSESSMENT COURSE WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR DOCUMENTATON. Volume 2: Probatulity And Statistice For PRA Appl 6-NUR 42 2: M T G TIVE CH IOU O GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR Prah=hm-me setety Analyste ACCIDENTS. volume 2 Case Study Anarysis Of Hydrologic Character.

NUREG/CR 2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE. tration And Mit6gettve Schemes.

DURES GUIDE. Sections 17 And Appendices NUREG/CR4397: IN-PLANT SOURCE TERM MEASUREMENTS AT NUREG/CR-2815 V02 R1: PROBABluSTIC SAFETY AN ALYSIS PROCE. PRAIRIE ISLAND NUCLEAR GENERATING STATION.

DURES GUIDE.Secelons 8-12.

Radium Treatment Progreeseen NUREG/CR4259: TAluNGS NEUTRAU2ATON AND OTHER ALTER-NUREG/CR4318 V01: REACTOR SAFETY RESEARCH NATIVES FOR IMMOBlU21NG TOXIC MATERIALS IN TAluNGS.Fmal PROGRAMS.Quartetty Report January-March 1985. Report.

52 Subject Index Redweete Reliability NUREG/CR4254: OCCUPATIONAL DOSE REDUCTION AND ALARA NUREG 1144. NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-AT NUCLEAR POWER PLANTS. Study On High-Dose Jobs.Radwaste GRAM PLAN.

Handimg.And ALARA lncentives. NUREG 1155 V04: RESEARCH PROGRAM PLAN Non.Destructrue Ex-NUREG/CR4397: IN-PLANT SOURCE TERM MEASUREMENTS AT amenation.

PRAIRIE ISLAND NUCLEAR GENERATING STATON. NUREG/CR-3876. PROBA81UTY BASED LOAD COMBINATION CRITE-RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

Reactor Cooient Loop Repair Weided Stainnees Steel NUREG/C43660 V01: PROBA0luTY OF PIPE FAILURE IN THE REAC-TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume NUREG/C436t3 V03 Nf: EVALUATON OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report 1: Summary Report.

For October 1984 Through March 1985.

NUREG/C43660 V04. PROBABluTY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4. Pipe Failure induced By Crack Growth in West Coast Plants. Repoeltory NUREG/C44290 V02: PROBA81UTY OF PIPE FArLURE IN THE REAC- NUREG/C43710 LABORATORY STUDIES OF A BREACHED NUCLE-TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR AR WASTE REPOSITORY IN BASALT.

PLANTS. Volume 2: Guillotine Break Indrectly induced By Earthquakes.

Research Utilization Report Reactor Core Thermal Hydraulle NUREG 1080 V02: LONG-RANGE RESEARCH PLAN FY 1986-FY 1990.

NUREG/CR 4240 V01: PHYSICS OF REACTOR SAFETY.Ouarterty Report. January-March 1985. Residual Life NUREG 1144. NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-Reactor Deelen GRAM PLAN.

NUHEG-1070- NRC POUCY ON FUTURE REACTOR Reelduel Strees Measurement DESIGNS.Doosions On Severe Accident issues in Nuclear Power Plant Regulatiort NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS. Annual Report. October 1983. September Reactor Operatore Ucensin0 1984.

NUREG/CR 4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW Resistence Factor PERFORMANCE IN CONTROL ROOM SJMULATORS. NUREG/CR-3876: PROBA81UTY BASED LOAD COM8tNATON CRITE-RfA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

Reactor Proseure Boundary NUREG/CR4300 V01: ACOUSTIC EMISSION / FLAW RELATONSHIP Reeparator FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/C43537: EXPEDIENT METHODS OF RESPIRATORY VESSELS Progress Report, October-March 1985. PROTECTIONlit. SUBMICRON PARTICLE TESTS AND

SUMMARY

OF OVAUTY FACTORS.

Reactor Safety NUREG/CR-3485: PRA REVIEW MANUAL Response Tree NUREG/CR3816 V03: REACTOR SAFETY RESEARCH Ouarterty NUREG/CR4272; RESPONSE TREE EVALUATION EXPERIMENTAL Report. Jury-September 1984 ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR NUREG/CR-3816 V04. REACTOR SAFETY RESEARCH.Ouarterty OPERATORS.

Report. October December 1984.

NUREG/CR-3885 V04. HIGH TEMPERATURE GAS-COOLED REACTOR Rg Waves SAFETY STUDIES FOR THE DIVISON OF ACCIDENT NUREG/C44354. A STUDY OF SEISMICITY AND TECTONICS IN NEW EVALUATON Quarterty Progress Report, October 1 December ENGLAND. Final Report.

31,1964 NUREG/C44143: REVIEW AND EVALUATON OF THE MILLSTONE Riek NUREG/C43816 V04: REACTOR SAFETY RESEARCH Ouarterty UNIT 3 PROBA 3.JSTIC SAFETY STUDY.Contanment Failure ModesBadiologca1 ocurce Terms And Offsste con es Report. October-Decer%er 1984 NUREG/C44240 V01: PHYSICS OF REACTOR FETY.Quarterty NUREG/CR4377: EVALUATIONS AND UTlu2ATIONS OF RISK IM-Report. January. March 1985. PORTANCES.

NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH PROGRAMS Quarterty Report. January-March 1985. Rules Of Practice NUREG4386 003: UNITED STATES NUCLEAR REGULATORY COM-Reactor Scram MISSON STAFF PRACTICE AND PROCEDURE DIGEST. JULY 1972 NUREG/CR.3948: EXPERIMENTAL RESULTS OF THE OPERATONAL SEPTEMBER 1983 TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST SAFT FACIUTY.

NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUS Reactor Trip POSE SAFT SYSTEM FOR NONDES TRUCTIVE EVALUATION OF NU.

NUREG-1154: LOSS OF MAIN AND AUXIUARY FEEDWATER EVENT CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

AT THE DAVIS-BESSE PLANT ON JUNE 9,1985.

SETS Reactor Vessel NUREG/CR4213: SETS REFERENCE MANUAL NUREG 1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels.

NUREG/CR 4365: DESIGN AND DEVELOPMENT OF A SPECIAL PU4 SIMOUAKE POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU- NUREG/C44182: VERIFICATION OF SOIL STRUCTURE INTERACTION CLEAR REACTOR VESSELS AND PIPING COMPONENTS. METHODS.

Reelfoot Rift SIT SG NUREG/C44333: STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL. NUREG/C44376. HEAT TRANSFER. CARRYOVER AND FALL BACK IN LINOIS. PWR STEAM GENERATORS DURING TRANSIENTS.

Reference Manuel SPARC NUREG/CR42f 3. SETS REFERENCE MANUAL NUREG-0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

M*9uistion $$NM l NUREG-1070: NRC POUCY ON FUTURE REACTOR DESIGNS Deosions On Severe Accident losues in Nuclear Power NUREG/C44107: SEQUENTIAL TEST PROCEDURES FOR DETECT.

Plant Regulation. ING PROTRACTED MATERIALS LOSSES.

f Ro9uletory And Technical Report safety NUREG4304 V10 N02: REGULATORY AND TECHNICAL NUREG/C44281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY, REPORTS Compilation For Second Quarter 1985.

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

Subject index 53 Safety Assurance EVALUATION OF HYPOTHESES ON THE SOURCE OF THE 1888 NUREG/CR4377: EVALUATIONS AND UTIUZArlONS OF RISK IM- CHARLESTON. SOUTH CAROUNA EARTHOUAKE.

PORTANCES.

Setem6c Riek Aseeeement Sofety Evolustion Regat NUREG/CR4331: SIMPUFIED SEISMIC PROBABlUSTIC RISK NUREG4)S75 S32: SAFETY EVALUATON REPORT RELATED TO THE ASSESSMENT. Procedures And Umstations.

OPERATON OF D'ABLO CANYON NUCLEAR POWER PLANT. UNITS 1 AND 2. Docket Nos.50-275 And 50323(Pacrlic Gas and Electne Setem6 city Company) NUREG/CR-3145 V03: GEOPHYSICAL INVESTIGATONS OF THE NUREGd98 S08: SAFETY EVALUATON REPORT RELATED TO THE WESTERN OHIO-INDIANA REGON - ANNUAL REPORT.(October OPERATON OF FERMI-2. Docket No. 50-341.(Detroit Edison Compa- 1982. September 1983. Volume 3).

"Y) NUREG/CR-4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA NUREG-0898 S03: SAFETY EVALUATION REPORT RELATED TO THE AND EASTERN TENNESSEE EARTHOUAKUL (19781984).

OPERATION OF SEABROOK STATION. UNITS 1 AND 2. Docket Nos. NUREG/CR-4317 VOL CANADIAN SEISMU AGREEMENT.Techrucal 50443 And 50444(Public Service Company of New HampsNre.et af) Report Covenn01979-1985 NUREG0940 V04 NO2: ENFORCEMENT ACTONS.SIGNIFICANT AC- NUREG/CR4339 A REVIEW OF RECENT RESEARCH ON THE SEIS-MOTECTONICS OF THE SOUTHEASTERN SEABOARD AND AN NUR 9 SO4 SA TY LU TO h E TED TO EVALUATION OF HYPOTHESES ON THE SOURCE OF THE 1888 FINAL DESIGN APPROVAL OF THE GESSAR ILBWR/8 NUCLEAR ISLAND DESGN. Docket No. 50447 (General Electnc Company) CHARLESTON. SOUTH CAROUNA EARTHOUAKE.

NUREG/CR4354: A STUDY OF SEISMICITY AND TECTONICS IN NEW NUREG4)989 S02: SAFETY EVALUATION REPORT RELATED TO THE ENGLANDFenal Report.

OPERATON OF RIVER BEND STATON. Docket No. 50-458.(Gulf NURE SO3 bA YPO OPERATON OF RIVER BEND STATON. Docket No. 50458.(Gulf ED TO THE SWomograph NUREG/CR4317 V01: CANADIAN SEISMIC AGREEMENT.Techrucal States Utubes Company) Report Covering 1979 1985.

I NUREG-0989 SO4. SAFETY EVALUATON REPORT RELATED TO THE OPERATON OF RIVER BEND STATON. Docket No. 50-458(Gulf SenWttaat6on l

States Utilities Compa . Electric Power Cooperative) NUREG/CR4287: ENVIRONMENTALLY ASSISTED CRACKING IN l UGHT WATER REACTORS. Annual Report. October 1983 September l NUREG4)991 S05: SAFE ALUATON REPORT RELATED TO THE i OPERATON OF UMERICK GENERATING STATION. UNITS 1 AND 1984.

2. Docket Not 50-352 And 50 353. (Ptvladelptua Elactric Company)

NUREG-0991 S08: SAFETY EVALUATION REPORT RELATED TO THE Service Water OPERATON OF UMERICK GENERATING STATION. UNITS 1 AND NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-

2. Docket Nos. 50-352 And 50-353 (PNiadelpNa Electnc Company) GRAM PLAN.

NUREG-1031 SO2: SAFETY EVALUATON RELATED TO THE OPER-ATON OF MILLSTONE NUCLEAR POWER STATION. UNIT 3 Docket Service Weer No S0-423 (Northeast Nuclear Energy Compa ) NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-NUREG 1048 S02: SAFETY EVALUATON HE T RELATFD TO THE GRAM PLAN.

OPERATION OF HOPE CREEK GENERATING STATON. Docket No. NUREG/CR 3819: SURVEY OF AGED POWER PLANT FACluTIES.

50-354.(PutAc Sennce Electric and Gas Company) NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC NUREG 1138: SAFETY EVALUATON REPORT RELATED TO THE RE- MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY-FEA.

NEWAL OF THE OPERATING UCENSE FOR THE TRAINING AND TURE SYSTEMS OF NUCLEAR POWER PLANTS.

RESEARCH REACTOR AT THE UNIVERSITY OF MICHIGAN Docket No. 50-2 (Uruversity of MicNgan) Severe Accident NUREG 1070: NRC POUCY ON FUTURE REACTOR Safety leeue A-47 MSONShsions On Swere AccM leeues in Mer W NUREG/CR4328 V01: EFFECTS OF CONTROL SYSTEM FAILURES '

A P WESTINGHOUSE NURE 1 V : LONG-RANGE RESEARCH PLAN FY 1988-FY 1990.

E R E ATE EACT 'Ma n R NUREG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE.

Safety Reeeerch NUREG/CR4080' DETERMINATON OF THE AVAILABluTY OF CORE NUREG/CR-3818 V03; REACTOR SAFETY RESEARCH.Ouarterty EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATONS.

NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer Report. July-September 1964.

V01: Code for Severe Reactor Accident Containment Analysis.

NUREG/CR4318 REACTOR SAFETY RESEARCH PROGRAMS Quarterty Report. January-March 1985. NUREG/CR-4119: INTEGRITY OF CONTAINMENT PENETRATONS UNDER SEVERE ACCIDENT CONDITIONS FYS4 ANNUAL REPORT.

Safety-Reteted Equipment NUREG/CR4130- ICEDF:A CODE FOR AEROSOL PARTICLE CAP.

NUREG-1148: NUCLEAR POWER PLANT FIRE PROTECTION RE. TURE IN ICE COMPARTMENTS.

SEARCH PROGRAM. NUREG/CR-4137: PRETEST PREDICTONS FOR THE RESPONSE OF A 1.8-SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO Safety-RWeted Personnel STATIC OVERPRESSURIZATON.

NUREG/CR4248. RECOMMENDATONS FOR NRC POLICY ON SHIFT SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS. Severe Fuel Damese NUREG/CR4318 V01: REACTOR SAFETY RESEARCH E /CR 3 1 D AND THEORETICAL INVESTIGATONS OF FRACTURED CRYSTALLINE ROCK NEAR ORACLE, ARIZONA. Shift ScheduNng NUREG/CR-4248: RECOMMENDATIONS FOR NRC POUCY ON SHIFT i g,,4 gg '

NUREG/CR4298: DESIGN AND INSTALLATON OF COMPUTER SYS- l TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55. Shift Technical Adviser Sodomic Margin NUREG/CR4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND NUREG/CR4334: AN APPROACH TO THE QUANTIFICATION OF SEIS- ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) Of 4 CREW PERFORMANCE IN CONTROL ROOM SIMULATORS.

MIC MARGINS IN NUCLEAR POWER PLANTS.

Sotomic OueNficetten So#-Structure interaction NUREG 1030 DAFT: SEISMIC OUAUFICATON OF EOUIPMENT IN Op. NUREG/CR-4182: VERIFICATION OF SOtL STRUCTURE INTERACTION ERATING NUCLEAR POWER PLANTS.Unreso8ved Safety issue A. METHODS.

48 Droft Report For Comment.

Setemic Reflection NUREG/CR 3444 V02: THE IMPACT OF LWR DECONTAMINATIONS NUREG/CR4339: A REVIEW OF RECENT RESEARCH ON THE SEIS- ON SOUDIFICATON. WASTE DISPOSAL AND ASSOCIATED OCCU-MOTECTONICS OF THE SOUTHEASTERN SEABOARD AND AN PATIONAL EXPOSURE.

l 54 Subject index Source Term TLD NUREG/CR4143: REVIEW AND EVALUATON OF THE MILLSTONE NUREG-0837 V04 N04. NRC TLD DIRECT RADIATON MONITORING UNIT 3 PROBABluSTIC SAFETY STUDY Containment Failure REPORT Progress Report October-December 1984.

Modes.Radological Source-Terms And Offsite Consequences- NUREG4837 V05 N01: NRC TLD DIRECT RADIATION MONITORING I

NETWORK. Progress Report, January-March 1985.

NUREG-0837 V05 NO2: NRC TLD DIRECT RADIATION MONITORING NUREG/Cd-3609: EVALUATON OF NEUTRON DOSIMETRY TECH- NETWORK. Prog ess Report, Apni-June 1985.

NIQUES FOR WELL-LOGGING OPERATIONS.

TORAC RE CR4379 Vot: LONG-TERM PERFORMANCE OF MATERIALS j USED FOR HIGH-LEVEL WASTE PACKAGING.Frst Ouarterty ,h,ang

,,~ TmWM N N Mew hanW in har %

Report, Year Four Apm 1985.

Statellity Analyste AO NUREG/CR4116: NUFEGO.NP;A DIGITAL COMPUTER CODE FOR NUREG/CR-3706. TRAC ANALYSES OF SEVERE OVERCOOUNG TRANSIENTS FOR THE OCONEE 1 PWR.

THE UNEAR STABluTY ANALYSIS OF BOluNG WATER NUCLEAR REACTORS.

TRAC-SO1/ MOO 1 Steineses Steet NUREG/CR-3633 V01 SI: TRAC-BD1/MODt.AN ADVANCED BEST ES-NUREG/CR4060- THE DC 1 AND DC-2 DEBRIS COOLABILITY AND TIMATE COMPUTER PROGRAM FOR BOiUNG WATEH REACTOR MELT DYNAMICS EXPERIMENTS. TRANSIENT ANALYSIS.

NUREG/CR-3633 V04: TRAC BD1/MODt.AN ADVANCED BEST ESil-i State Reguistion MATE COMPUTER PROGRAM FOR BOtuNG WATER REACTOR NUREG/CR-4352: SUGGESTED STATE REQUIREMENTS AND CRITE. TRANSIENT ANALYSIS. Volume 4: Developmental Assessment RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE REGULATORY PROGRAM. TRAC-PD2/ MOO 1 3

NUREG/CR4252: INDEPENDENT ASSESSMENT OF TRAC PD2/ MOD 1 i Statistu Sympoelum CODE WITH BCL ECC BYPASS TESTS.

NUREG/CP-0063; PROCEEDINGS OF THE 1984 STATISTICAL SYMPO-SIUM ON NATIONAL ENERGY ISSUES. TRAC-PF1

{

NUREG/CR-3706: TRAC ANALYSES OF SEVERE OVERCOOUNG S TRANSIENTS FOR THE OCONEE 1 PWR.

NUREG/CR4333. STE. GENEVIEVE FAULT ZONE MISSOURI AND IL.

UNOIS- TRAP-MELT NUREG4956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES E 5 V02: RESEARCH PROGRAM PLAN Steam Generators NUREG/CR-3949 V02: EDDY CURRENT INSPECTON FOR STEAM Temnge Neutreimation GENERATOR TUBING PROGRAM. Annual Progress Report For Penod NUREG/CR4259: TAILINGS NEUTRAUZATION AND OTHER ALTER-NUR C H ANSFER, CARRYOVER AND FALL BACK IN NATIVES FOR IMMOBiUZING TOXIC MATERIALS IN TAluNGS Final PWR STEAM GENERATORS DURING TRANSIENTS.

Repe Strategic Special Nueteer Motortel Team Skille NUREG/CR4107: SEQUENTIAL TEST PROCEDURES FOR DETECT. NUREG/CR4258. AN APPROACH TO TEAM SKILLS TRAINING OF NU-ING PROTRACTED MATERIALS LOSSES. CLEAR POWER PLANT CONTROL ROOM CREWS.

Strese Corrosion Crocking Technical Specincetione i

NUREG 1155 V03: RESEARCH PROGRAM PLAN Piping NUREG-0940 V04 NO2: ENFORCEMENT ACTONS SIGNIFICANT AC.

TIONS RESOLVED.Ouarterty Progress Report Apni-June,1985.

Submicron Particle Teete . NUREG 1126. TECHNICAL SPECIFICATONS FOR SHOREHAM NU.

NUREG/CR 3537: EXPEDIENT METHODS OF RESPIRATORY CLEAR POWER STATION, UNIT NO.1. Docket No. 50-322.(Long

' PROTECTON m. SUBMICRON PARTICLE TESTS AND

SUMMARY

lsland Ughting Company)

OF OUAUTY FACTORS. NUREG-1126: TECHNICAL SPECIFICATIONS FOR SHOREHAM NU-CLEAR POWER STATON, UNIT NO.1 Docket No 50 322(Long Subetructure Method Island Ughting Company) j NUREG/CR4182: VERIFICATON OF SOIL STRUCTURE INTERACTON NUREG 1141: TECHNICAL SPECIFICATONS FOR FERMl 2 ME THODS. FACIUTY. Docket No. 50 341. (Detroit Edison Company)

NUREG 1142: TECHNICAL SPECIFICATONS FOR RIVER BEND 8"'V'"*"C' STATION Docket No. 50458. (Gulf States Utilities Company)

NUREG/CR 3319- LWR PRESSURE VESSEL SURVEILLANCE DOSIME- NUREG 1149; TECHNICAL SPECIFICATIONS FOR LIMERICK GENER-TRY IMPROVEMENT PROGRAM LWR Power Reactor Sunteeltance ATING STATION, UNIT 1. Docket No.60-352. (PNiedelpNa Electric Physics-Dosimetry Data Base Compendium. Company)

NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC NUREG 1151: TECHNICAL SPECIFICATONS FOR DIABLO CANYON MOTOROPERATED VALVES USED IN ENGINEERED SAFETY FEA* NUCLEAR POWER PLANT UNITS 1 AND 2 Docket Nos.60-275 And TURE SYSTEMS OF NUCLEAR POWER PLANTS. 50 323 (Pacific Gas And Electric Company)

Survey Statistice NUREG/CR4268. RATO METHODS FOR COST-EFFECTIVE FIELD NUREG/CR 3091 V06: R.1 VIEW OF WASTE PACKAGE VERIFICATION SAMPUNG OF COMMERCIAL RADIOACTIVE LOW-LEVEL WASTES.

TESTS.Semeannual Report Covenng The Period October 1964 Merch Symbolic Manipulation 1985.

NUREG/CR4213: SETS REFERENCE MANUAL NUREG/CR 3426 vot: THERMAL AND FLUID MIXING IN 1/2-SCALE TEST FACluTY. Facey And Test Design Report Synthetic Aperture Focusing Technique NUREG/CR-3426 V02: THERMAL AND FLUlO MIXING IN 1/2 SCALE NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR- TEST FACluTY Data Repnrt

, POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU- NUREG/CR 3736: FIELD AND THEORETICAL INVESilGATONS OF CLEAR REACTOR VESSELS AND PIPING COMPONENTS. FRACTURED CR(STALUNE ROCK NEAR ORACLE ARIZONA.

NUREG/CR-3948. EXPERIMENTAL RESULTS OF THE OPERATONAL TAPS TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST NUREG/CH-3481 V02: NUCLEAR POWER PLANT PERSONNEL QUAU- FACIUTY, FICATIONS AND TRAINING TAPS ~ The Tash Analysis Profiling NUREG/CR4107: SEQUENTIAL TEST PROCEDURES FOR DETECT.

( System. ING PROTRACTED MATERIALS LOSSES.

Subject index 55 Testing Transient Analysis NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH NUREG/CR-3633 V04: TRAC-BD1/ MODI AN ADVANCED BEST ESTa-PROGRAMS.Ouarterty Report, January March 1985. MATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS Volume 4. Developmental Assessment.

NUREG-1155 V03: RESEARCH PROGRAM PLAN Piping.

Transient Reactor Analysis Code Thermal Annealing NUREG/CR-3706: TRAC ANALYSES OF SEVERE OVERCOOUNG NUREG-1155 VO1:RESEARCH PROGRAM PLAN Reactor Vessels. TRANSIENTS FOR THE OCONEE 1 PWR.

Thermal Conductivity Transport NUREG/CR4060- THE DC-1 AND DC 2 DEBRIS COOLABluTY AND NUREG/CR4255 V01: AEROSAL RELEASE AND TRANSPORT PRO-MELT DYNAMICS EXPERIMENTS. GRAM SEMIANNUAL PROGRESS REPOGT FOR OCTOBER 1984 -

g MARCH 1985.

NUREG/CR-3426 V01: THERMAL AND FLUlO MIXING IN 1/2-SCALE Tube integrity NUl / HE M L AN O IXING IN 1/2 SCALE NUREG 1155 V02: RESEARCH PROGRAM PLAN Steam Generators.

TEST FACIUTY. Data RW Tubing Thermal Shock NUREG/CR-3949 V02: EDDY-CURRENT INSPECTION FOR STEAM NUREG-1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels- GENERATOR TUBING PROGRAM. Annual Progress Report For Penod NUREG/CR-4249: PRESSUPE VESSEL FRACTURE STUDIES PENE. Ending December 31.1984.

TRATING TO THE PWR THERMAL-SHOCK ISSUE. EXPERIMENTS TSE-5,TSE-SA AND TSE-6 Ultrasonica NUREG/CR-4275: HEAVY-SECTON STEEL TECHNOLOGY PROGRAM FIVE-YEAR PLAN FY 1984-1988- NUREG 1155 V04: RESEARCH PROGRAM PLAN Non-Destructive Ex-aminah Thermai-Hydraulic NUREG/CR-3935. THERMAL-HYDR /.UUC ANALYSES OF OVERCOOL, Unauthorized Radioactive Material ING SEQUENCES FOR THE H.B. ROBINSON UNIT 2 PRESSURIZED NUREG/CR-4357: THE FEASIBILITY OF DETECTING THE IMPORT OF THERMAL SHOCK STUDY. UNAUTHORIZED RADCACTIVE MATERIALS INTO THE UNITED NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Computer STATES.

Code for Severe Reactor Accident Containment Analysis.

Uncertainty Thermal-Shock NUREG/CR-4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE NUREG/CR-4304: PRESSURE VESSEL FRACTURE STUDIES PER* CRAC2 COMPUTER CODE.

TAINING TO THE PWR THERMAL-SHOCK ISSUE.Expenment 'SE 7.

Unified Transport Approach U G 080 DETERMINATION OF THE AVAILABluTY OF CORE NUREG/CR-3442: RADTWO.A COMPUTER CODE FOR SIMULATING EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATIONS.

FAST TRANSlENT, TWO-DIMENSIONAL.TWO-LAYER RADIONU-CLlDE CONCENTRATION CONDITIONS IN Thermoluminescent Dosimeter LAKES. RESERVOIRS. RIVERS.ESTUAR!ES.AND COASTAL REGIONS.

NUREG-0837 VO4 N04: NRC TLD DIRECT RADIATON MONITORING REPORT. Progress Roport October-Docember 1984. Unresolved Safety issue NUREG-0837 VOS NO2- NRC TLD DeRFCT RADIATION MONITORING NUREG-0606 V07 NO3: UNRESOLVED SAFETY ISSUES NETWORK. Progress Report, Apni-June 1985.

SUMMARY

. Data As Of August 18.1985. (Aqua Book)

Thermomechanical History Unresolved Safety issue A-46 NUREG/CR-3P13 V03 N1: EVALUATION OF WELDED A*lD REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Serniannual Report NUREG 1030 DRFT: SEISMIC QUAUFICATION OF EQUIPMENT IN OP.

For October 1984 Through March 1965- ERATING NUCLEAR POWER PLANTS. Unresolved Safety issue A-

46. Draft Report For Comment.

Title List NUREG-0540 V07 N05. TITLE LIST OF DOCUMENTS MADE PUBUCLY Unresolved Safety issue A-49 AVAILABLE May 1-31,1985. NUREG/CR-3935: THERMAL-HYDRAUUC ANALYSES OF OVERCOOL-NUREG4540 V01 N05: TITLE UST OF DOCUMENTS MADE PUBLICLY ING SEQUENCES FOR THE H.B. ROBINSON UNIT 2 PRESSURIZED AVAILABLE.May 1 31,1985. THERMAL SHOCK STUDY.

NUREG-0540 V01 N06: TITLE LIST OF DOCUMENTS MADE PUBLICL)

AVAILABLE. June 1-30 1985. Uranium Mill Tallings NUREG4540 V07 N07:' TITLE UST OF DOCUMENTS MADE PUPUCLY NU'EG/CR-4259: TAIUNGS NEUTRAll2ATION AND OTHER ALTER-NURE O ITLE UST OF DOCUMENTS MADE PUBLICLY AVAILABLE. August 1-31,1985.

Tornado Model User *a Guide NUREG/CR-4260: TORAC USER'S MANUAL.A Computer Code For Ana. NUREG/CR 3901: DOCUMENTATION AND USER'S GUIDE.GS2 & GS3 lyzmg Tomado induced Flow And Matenal Trar. sport in Nuclear Facili- - VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS.

ties. NUREG/CR-4122: A FORTRAN 77 PROGRAM AND USER'S GUIDE FOR THE CALCULATION OF PARTIAL CORRELATION AND STAND.

Tornado Transient ARDIZED REGRESSION COEFFICIENTS.

NUREG/CR-4232: THE RESPONSE OF VENTILATION DAMPERS TO NUREG/CR-4130- ICEDF A CODE FOR AEROSOL PARTICLE CAP.

LARGE AIRFLOW PULSES. TURE IN ICE COMPARTMENTS.

Trainin9 User's Manual NUREG 1122: KNOWLEDGES AND ABluTIES CATALOG FOR NUCLE- NUREG/CR 4085: USERS MANUAL FOR CONTAIN 1.0.A Computer l NU E CR V 2: N R LN R NNE'L QUALI- Code for Severe Reactor Accident Containment Analysis.

i FICATIONS AND TRAINING: TAPS - The Task Analvsis Profilmg VANESA NulYEG CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU-DRFT FC: REASSESSMENT OF THE TECHNICAL BASES NUREG-0956 CLEAR POWER PLANT CONTROL ROOM CREWS. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment).

Transient Valve Failure NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATIONAL NUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST PLANT VALVE FAILURE-RATE VARIABILITY-SOME PREUMINARY FACIUTY. RESULTS.

56 Subject Index F Vehicle Barriers WASTE REPOSITORY SITE PROJECTS Annual Progress Report For NUREG/CR-4250: VEHICLE BARRIERS EMPHASIS ON NATURAL FEA- October 1983-September 1984.

TURES. NUREG/CR-4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS SAFETY ANALYSIS Phase 1, Final Report Velocity Structure NUREG/CR 3145 V03. GEOPHYSICAL INVESTIGATIONS OF THE Water Chemistry WESTERN OHIO-INDIANA REGION - ANNUAL REPORT.(October NUREG/CR-4287. ENVIRONMENTALLY ASSISTED CRACKING IN 1982 - September 1983. Volume 3). LIGHT WATER REACTORS Annual ReportOctober 1983 - September Ventilation NUREG/CR4260 TORAC USER'S MANUALA Computer Code For Ana- Wear lyzing Tornado-Induced Flow And Matenal Transport in Nuclear Facili- NUREG 1155 V02: RESEARCH PROGRAM PLAN Steam Generators.

ties.

Weld Overlay Ventilation Damper NUREG/CR-4287. ENVIRONMENTALLY ASSISTED CRACK:NG ;N NUREG/CR4232; THE RESPONSE OF VENTILATION DAMPERS TO LIGHT WATER REACTORS. Annual Report. October 1983 - September LARGE A:RFLOW PULSES.

1984.

Vibration NUREG-1155 V02: RESEARCH PROGRAM PLAN Steam Generators. I' N RE 1 55 V03. RESEARCH PROGRAM PLAN Piping.

Visual Display Units NUREG/CR4227; HUMAN ENGINEERING GUIDELINES FOR THE Welded Stainless Steet EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS. NUREG/CR-3613 V03 N1. EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE Semannual Report WAPPA For October 1984 Through March 1985.

NUREG/CR-3091 V06: REVIEW OF WASTE PACKAGE VERIFICATION TESTS Serrnannual Report Covenng The Penod October 1984 March Weldment 1985. NUREG/CR-4219 V01: HEAVY SECTION STEEL TECHNOLOGY PRO-GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

Waste MARCH 1985.

NUREG/CR-4379 V01: LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR-4275. HEAVY SECTION STEEL TECHNOLOGY PROGRAM USED FOR HIGH-LEVEL WASTE PACKAGING First Quarterty FIVE-YEAR PLAN FY 1984-1988.

Report, Year Four Apnl-June 1985.

Well-Logging Waste Dispoeal NUREG/CR-3609. EVALUATION OF NEUTRON DOS! METRY TECH-NUREG/CR-3444 V02: THE IMPACT GF LWR DECONTAMINATIONS NIOUES FOR WELL-LOGGING OPERATIONS.

ON SOLIDIFICATION. WASTE EISPOSAL AND ASSOCIATED OCCU-PATiONAL EXPOSURE. Workbook NUREG/CR-4125 V02: GUIDELINES AND WORKBOOK FOR ASSESS-Waste isolation Prom NUREG/CR-3851 V04: EVALUATION OF RADIONUCLIDE GEOCHEMI- MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR SEEKING OPERATING LICENSE FOR A NUCLEAR POWER WASTE REPOSIT')RY SITE PROJECTS. Annual Progress Report For PLANT. Volume 2. Workbook For Assessment Of Orgaruzation And Man-October 1983-September 1984. agement Waste Package Workshcp NUREG/CR-3091 V08: REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CP-0066: PROCEEDINGS OF AN INTERNATIONAL WORK.

TESTS. Semiannual Report Covenng The Penod October 1984 - March SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION o 1985 (ALARA) AT NUCLEAR POWER PLANTS,MAY 29-JUNE 1,1984 NUREG/CR-3900 V04: LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR 4125 VOI: GUIDELINES AND WORKBOOK FOR ASSESS-USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report,Apnl MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES 1984. tpni 1985. SEEKING OPERATING LICENSE FOR A NUCLEAR POWER NUREG/CR-4379 V01: LONG TERM PERFORMANCE OF MATERIALS PLANT. Volume 1.Guidehnes For Utikty OrgaruzatKm And Admirnstration USED FOR HIGH-LEVEL WASTE PACKAGING First Quarterty Plan.

Report, Year Four Apni-June 1985.

Zircaloy Fracture Waste Repository NUREG/CR-3980 V04: LIGHT WATER-REACTOR SAFETY FUEL SYS-NUREG/CR-3851 V04: EVALUATION OF RADIONUCLIDE GEOCHEMI- TEMS RESEARCH PROGRAMS. Quarterty Progress Report, October.

CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR December 1984.

NRC Originating Organization Index (Staff Reports)

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

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) OFFICE OF NUOLEAR REGULATORY RESEARCH (POST 4/05/81)

OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR NUREG 1154: LOSS OF MAIN AND AUXIUARY FEEDWATER EVENT NUREG-1080 V02: LONG-RANGE RESEARCH PLAN FY 1986-FY AT THE DAVIS-BESSE PLANT ON JUNE 9,1985. 1990.

REGION 1. OFFICE OF DIRECTOR ACCIDENT SOURCE TERM PROGRAL1 OFFICE NUREG4837 V04 N04: NRC TLD DIRECT RAD;ATON MONITORING NUREG-0856 DAFT FC: REASSESSMENT OF THE TECHNICAL REPORT. Progress Report, October-December 1984. BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For NUREG-0837 V05 N01: NRC TLD DIRECT RADIATON MONITORING Comment).

NETWORK. Progress Report, January-March 1985. DIVISION OF ENGINEERING TECHNOLOGY NUREG4837 V05 NO2: NRC TLD DIRECT RADIATON MONITORING NUREG-1144: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-NETWORK. Progress Report, Aprd June 1985. GRAM PLAN.

NUREG-1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels. F EDO OFFICE OF ADMINISTRATION NUREG 1155 V02: RESEAFtCH PROGRAM PLAN Steam Generators. (

DIVISON OF TECHNICAL INFORMATON & DOCUMENT CONTROL NUREG-1155 V03: RESEARCH PROGRAM PLAN Piping NUREG-0304 V10 NO2: REGULATORY AND TECHNICAL NUREG-1155 V04: RESEARCH PROGRAM PLAN.NomDestructive Ex-REPORTS Compilation For Second Quarter 1985 amination.

NUREG4540 V07 N05: TITLE UST OF DOCUMENTS MADE PUBLIC-LY AVAILABLE May 1-31,1985. EDO-RESOURCE MANAGEMENT NUREG 0540 V07 N06: TITLE UST OF DOCUMENTS MADE PU3LIC. DIVISION OF BUDGET & ANALYSIS LY AVAILABLE. June 1 30,1985. NUREG-0020 V09 N06: LICENSED OPERATING REACTORS STATUS NUREG-0540 V07 N07: TITLE UST OF DOCUMENTS MADE PUBUO.

SUMMARY

REPORT. Data As Of May 31,1985.(Gray Book f)

LY AVAILABLE. July 1 31,1985. NUREG-0020 V09 N07: LICENSED OPERATING REACTORS STATUS NUREG-0540 V07 N08: TITLE UST OF DOCUMENTS MADE PUBLIC-

SUMMARY

REPORT. Data As Of June 30,1985.(Gray Book I)

LY AVAILABLE. August 1-31,1985. NUREG-0020 V09 N08: UCENSED OPERATING REACTORS STATUS NUREG4750 V21102: INDEXES TO NUCLEAR REGULATORY COM-

SUMMARY

REPORT. Data As Of Jufy 31,1985.(Gray Book 1)

MISSON ISSUANCES. January June 1985. MANAGEMENT SUPPORT BRANCH NUREG-0750 V21 N05: NUCLEAR REGULATORY COMMISSION IS, NUREG 0748 V05 N05: OPERATING REACTORS UCENSING AC-SUANCES FOR MAY 1985. Pa s 1.043-1,567. TIONS

SUMMARY

. Data As Of May 31,1985. (Orange Book)

NUREG-0750 V21 N06: N REGULATORY COMMISSION IS. NUREG4748 V05 N06: OPERATING REACTORS UCENSING AC-SUANCES FOR JUNE 1985. Paoes 1.5691786. TIONS

SUMMARY

. Data As Of June 30,1985. (Orange Book)

NUREG4750 V22 N01: NUCLEAR REGULA' TORY COMMISSION IS- NUREG-0748 V05 N07: OPERATING REACTORS UCENSING AC.

SUANCES FOR JULY 1985. Pages 1 176. TONS

SUMMARY

. Data As Of July 31,1985.(Orange Book)

DIVISION OF RULES AND RECORDS NUREG4936 V04 NO2: NRC REGULATORY AGENDA.Ouartedy OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

Report.Apnl June 1985' OFFICE OF NUCLEAR REACTOR REGULATON, DIRECTOR NUREG-0800 13.5.2 RI: STANDARD REVIEW PLAN FOR THE EDO . OFFICE OF EXECUTIVE LEGAL DIRECTOR REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR OFFICE OF THE EXECUTIVE LEGAL DIRECTOR POWER PLANTS. LWR Edition. Revision 1 To Section 13.5.2. "Oper-NUREG-0386 D03: UNITED STATES NUCLEAR REGULATORY COM- sting And Maintenance Procedures," and Revision 0 of Appendix A MISSON STAFF PRACTICE AND PROCEDURE DIGESTJULY 1972 to Section 13.5.2, " Review...."

. SEPTEMBER 1983. NUREG-1070: NRC POUCY ON FUTURE REACTOR DESIGPSDecisions On Severe Acadent issues in Nuclear Power EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL Plant Regulation.

DATA NUREG-1149: TECHNICAL SPECIFICATONS FOR UMERICK GEN-AEOD, DIRECTOR'S OFFICE ERATING STATON, UNIT 1. Docket No. 50-352. (Pfiladelphia Elec.

NUREG-0090 V08 N01: REPORT TO CONGRESS ON ABNORMAL tric Company) _

OCCURRENCES.Janurary-March 1985. DMSION OF HUMAN FACTORS SAFETY NUREG-1022 SO2- UCENSEE EVENT REPORT SYSTEM Evaluation NUREG-1122 KNOWLEDGES AND ABluTIES CATALOG FOR NU.

Of First Year Results And Recommendations For improvements. CLEAR POWER PLANT OPEFtATORS.Pressurtred Water Reactors.

DIVISION OF UCENSING OFFICE OF IPsSPECTION & ENFORCEMENT (POST 12/11/80) NUREG4675 S32: SAFETY EVALUATON REPORT RELATED TO DIRECTOR'S OFFICE. OFFICE OF INSPECTON AND ENFORCEMENT THE OPERATION OF DIABLO CANYON NUCLEAR POWER NUREG-0940 V04 NO2: ENFORCEMENT ACTIONS.SIGNIFICANT AC* PLANT, UNITS 1 AND 2. Docket Nos 50-275 And 50-323.(Pacific Gas TIONS RESOLVED.Quarterty Progress Report,Apnl-June,1985.

OlVISION OF GA, VENDOR & TECHNICAL TRAINING CENTER PRO- and Electnc Compa NUREG-0798 S06: S ETY EVALUATON REPORT RELATED TO NUREG V N UCENSEE CONTRACTOR AND VENDOR IN- N 2 e et No. @etroit Edson SPECTON STATUS REPORT, Quarterfy Report.Aprd-June 1985.

IW '

NU b S03. SAFETY EVALUATON REPORT RELATED TO THE OPERATON OF SEADROOK STATON, UNITS 1 AND 2. Docket OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS Nos. 50-443 And 50-444.(Public Service Compar y of New DIVISON OF FUEL CYCLE & MATERIAL SAFETY Hampshire,et ei)

NUREG-1157: ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF NUREG-0979 SO4: SAFETY EVALUATON REPORT RELATED TO SOURCE MATERIAL UCENSE NO. SUB 1010. Docket No. 404027. FINAL DESIGN APPROVAL OF THE GESSAR ll.BWR/6 NUCLEAR (Sequoyah Fuels Corporation) ISLAND DESIGN. Docket No. 50-447.(General Electric Companyj NUREG4989 SO2: SAFETY EVALUATON REPORT RELATw TO U.S. NUCLEAR REGULATORY COMMISSION THE OPERATION OF RIVER BEND STATION. Docket No. 50-NRC - NO DETAILED AFFILIATION GIVEN 458 (Gulf States Utilities Company, Cajun Electric Power Cooperative)

NUREG/CR-4143: REVIEW AND EVALUATON OF THE MILLSTONE NUREG-0989 S03: SAFETY EVALUATION REPORT RELATED TO i UNIT 3 PROBABluSTIC SAFETY STUDY. Containment Failure THE OPERATION OF RIVER BEND STATION. Docket No. 50-Modes,Radi/qpeal Source-Terms And Offsste Consequences. 458 (Guit States Utilities Company) 57

58 NRC Originating Organization Index NUREG4989 SO4. SAFETY EVALUATION REPORT RELATED TO NUREG-1094: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATON OF RIVER BEND STATION. Docket No. 50 THE OPERATION OF DEAVER VALLEY POWER STATION. UNIT 458 (Gulf States utsties Company. Cajun Electne Power Cooperatrve)

2. Docket No. 50-412. (Duquesne Light Company) /

NUREG-0991 S05: SAFETY EVALUATION REPORT RELATED TO NUREG-1141: TECHNICAL SPECIFICATONS FOR FERMI-2 THE OPERATION OF LIMERICK GENERAT!NG STATON. UNITS 1 FACILITY. Docket No. 50-341. (Detrost Edison Company)

AND 2. Docket Nos. 50-352 And 50-353. (PNiadelpha Electnc Com- NUREG 1142: TECHNICAL SPEGIFICATIONS FOR RIVER BEND pany) STATON Docket No. 50-458. (Gulf States Utstes Company)

NUREG4991 S06: SAFETY EVALUATION REPORT RELATED TO NUREG 1151: TECHNICAL SPECIFICATIONS FOR DIABLO CANYON THE OPERATON OF LIMERICK GENERATING STATION, UNITS 1 NUCLEAR POWER PLANT UNITS 1 AND 2. Docket Nos. 50-275 And AND 2 Docket Nos. 50-352 And 50-353.(PNtadelpha Electnc Com- 50-323 (Pacife Gas And Electnc Company) pany) DIVISION OF SAFETY TECHNOLOGY ,

NUREG-1031 SO2- SAFETY EVALUATON RELATED TO THE OPER- NUREG4606 V07 NO3: UNRESOLVED SAFETY ISSUES ATION OF MILLSTONE NUCLEAR POWER STATION. UNIT

SUMMARY

. Data As Of August 16.1985. (Aqua Book) 3 Docket No 50-423 (Northeast Nuclear Energy Company) NUREG4933 S03: A PRIORITIZATON OF GENERIC SAFETY l NUREG-1048 SO2 SAFETY EVALUATION REPORT RELATED TO ISSUES. 1 THE OPERATON OF HOPE CREEK GENERATING NOREG-1030 DRFT: SEISMIC OUALIFICATION OF EOU.PMENT IN STATON Docket No. 50-354 (Pubic Service Electnc and Gas Com- OPERATING NUCLEAR POWER PLANTS Unresolved Safety lasue pany) A-46. Draft Report For Comment.

ed

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organizations that sponsor organization is followed by the sponsored the contractor reports listed in NUREG/CR number and title of the this compilation. It is arranged alphabetically report (s) prepared by that organization. If by major NRC organization (e.g., program further information is needed, refer to the office) and then by subsections of these main citation by the NUREG/CR number.

(e.g., divisions) where appropriata. The EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR-3935: THERMAL-HYDRAULIC ANALYSES OF OVER.

DATA COOLING SEQUENCES FOR THE H.B. ROBINSON UNIT 2 PRES-AEOD. DIRECTOR'S OFFICE SURIZED THERMAL SHOCK STUDY.

HUREG/CR-2000 V04 N6: UCENSEE EVENT REPORT (LER) NUREG/C43948: EXPERIMENTAL RESULTS OF THE OPERATION-COMPILATION.For Month Of June 1985.

AL TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER NUREG/GR-2000 V04 N7: UCENSEE EVENT REPORT (LER) BURST FACILITY.

COMPILATION.For Month Of July 1985.

NUREG/C43980 V04: LIGHT WATER-REACTOR SAFETY FUEL NUREG/CR-2000 V04 N8: LICENSEE EVENT REPORT (LER) SYSTEMS RESEARCH PROGRAMS. Quarterty Progress COMPILATION,For Month Of August 1985.

Report. October December 1984.

OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) TR DIVISION OF EMERGENCY PREPAREDNESS & ENGINEERING RE. SE TE T H NUREG/CR-4060- THE DC I' AND DC 2 DEBRIS COOLABluTY AND NU 4 E BULLETIN 8141: SURVEIL.

NURE /C 08 A RPTl GASEOUS IODINE BY WATER NUREG/C44151: INTEGRATION OF EMERGENCY ACTION LEVELS WITH COMBUSTION ENGINEERING EMERGENCY OPERATING NI REG /C 085: USERS' MANUAL FOR CONTAIN 1.0 A Computer PROCEDURES.By Use of Combustion Engmeanng Owners Group W e Re h CWW An@

Ernergency Operatog Rocedure TN Mnn NUREG/CR4116: NUFEGO.NP:A DIGITAL COMPUTER CODC FOR THE LINEAR STABluTY ANALYSIS OF BOLLING WATER NUCLE-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS AR REACTORS.

DVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG/CR4138. DATA ANALYSES FOR NEVADA TEST SITE (NTS)

NUREG/CR-4357: THE FEASIBlWTY OF CETECTING THE IMPORT PREMIXED COMBUSTION TESTS.

OF UNAUTHORIZED RADOACTIVE MATERIALS IN10 THE NUREG/CR4240 V01: PHYSICS OF REACTOR SAFETY.Ouarterty UNITED STATES. Report. January-March 1985.

DIVISION OF SAFEGUARDS NUREG/CR4252- INDEPENDENT ASSESSMENT OF TRAC-PD2/

NUREG/CR4107: SEQUENTIAL TEST PROCEDURES FOR DETECT. MOD 1 CODE WITH BCL ECC BYPASS TESTS.

ING PROTRACTED MATERIALS LOSSES. NUREG/CR-4255 V01: AEROSAL RELEASE AND TRANSPORT PRO.

NUREG/C44250- VEHICLE BARRIERS. EMPHASIS ON NATURAL GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

FEATURES.

DVISON OF WASTE MANAGEMENT MARCH 1985.

NUREG/CR-4318 V01: REACTOR

• SAFETY RESEARCH NUREG/CR-3091 V06: REVIEW OF WASTE PACKAGE VERIFICA- PROGRAMS Ouarterty Report. January-March 1985.

TION TESTS.Sermannual Report Covenng The Period October 1984

- March 1985. NUREG/CR4376: HEAT TRANSFER, CARRYOVER AND FALL BACK NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEO- IN PWR STEAM GENERATORS DURING TRANSIENTS.

CHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NU- DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429)

CLEAR WASTE REPOSITORY SITE PROJECTS. Annual Progress NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE-Report For October 1983-September 1984. OUENCE INFORMATION.

NUREG/CR-3901: DOCUMENTATION AND USER'S GUIDE:GS2 &

NUREG/CR-3481 V02: NUCLEAR POWER PLANT PERSONNEL OUAUFICATIONS AND TRAINING: TAPS - The Task Analysis Pro-GS3 - VARIABLY SATURATED FLOW AND MASS TRANSPORT t hng System.

MODELS-NUREG/C44150: EPICOR-Il RESIN DEGRADATION RESULTS NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY FROM FIRST RESIN SAMPLES OF PF-8 AND PF-20. PROTECTION.llt. SU8 MICRON PARTICLE TESTS AND

SUMMARY

OF QUALITY FACTORS.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/01)

EG/M22; A NRAN 77 MRM M USNS ME DIVISION OF ACCIDENT EVALUATION FOR THE CALCULATION OF PARTIAL CORRELATION AND NUREG/CR-3426 V01: THERMAL AND FLUID MIXING IN 1/2-SCALE STANDARDIZED REGRESSION COEFFICIENTS.

TEST FACIUTY. Fa ' And Test De ~ RW NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA FOR NUREG/CR-3426 V02 ERMAL AND UtD MIXING IN 1/2-SCALE ACCIDENTAL RADIONUCUDE RELEASES FROM NUCLEAR REAC.

TEST FACIUTY. Data Report. TORS.

NUREG/CR-3633 V01 S1: TRAC-BD1/ MODI.AN ADVANCED BEST NUREG/CR-4213: SETS REFERENCE MANUAL ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REAC- NUREG/CR4214: HEALTH EFFECTS MODEL FOR NUCLEAR TOR TRANSIENT ANALYSIS. POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part NUREG/CR-3633 V04: TRAC-BD1/ MOD 1 AN ADVANCED BEST ESTI. Untroduction, Integration & Summary.Part it: Scientific Basis For MATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR Health Effects Models.

TRANSIENT ANALYSIS.Voeume 4: Developmental Assessment NUREG/CR4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER NUREG/CR-3638: HYDROGEN-STEAM JET-FLAME FACluTY AND PLANT VALVE FAILURE-RATE VARIA81UTY--SOME PRELIMINARY EXPERIMENTS. RESULTS.

NUREG/C43706: TRAC ANALYSES OF SEVERE OVERCOOUNG NUREG/C44227: HUMAN ENGINEERING GUIDEUNES FOR THE TRANSIENTS FOR THE OCONEE-1 PWR. EVALUATION AND ASSESSMENT OF VISUAL DSPLAY UNITS.

NUREG/CR-3816 VU3: REACTOR SAFETY RESEARCH.Quarterty NUREG/CR4232: THE RESPONSE OF VENTILATION DAMPERS TO Report. July-September 1984.

LARGE AIRFLOW PULSES.

NUREG/CR-3816 V04: REACTOR SAFETY RESEARCH.Ouarterty NUREG/CR-4260 TORAC USER'S MANUALA Computer Code For Report. October-December 1984.

Analymng Tornado-induced Flow And Material Transport in Nuclear NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS-COOLED REAC- Facilities.

TOR SAFETY STUDES FOR THE DIVISION OF ACCIDENT NUREG/CR4272: RESPONSE TREE EVALUATION. EXPERIMENTAL EVALUATION.Ouarterty Progress Report, October 1-December 31,1984. ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REAC-TOR OPERATORS.

59

60 NRC Contract Sponsor index NUREG/CR-4280 THE EFFECTS OF SUPERVISOR EXPERIENCE

. NUREG/CR-3319 LWR PRESSURE VESSEL SURVEILLANCE DO-AND ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON SIMETRY IMPROVEMENT PROGRAM LWR Power Reactor Survest-CREW PERFORMANCE IN CONTROL ROOM SIMULATORS. lance Physics-Dosmetry Data Base Compendium.

NUREG/CR-4298: DESIGN AND INSTALLATION OF COMPUTER NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS SYSTEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55. ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED OC-NUREG/CR4314: BRIEF SURVEY AND COMPARISON OF COMMON CUPATIONAL EXPOSURE.

CAUSE FAILURE ANALYSIS. NUREG/CR-3613 V03 N1: EVALUATON OF WELDED AND REPAIR-NUREG/CR-4350 V02: PROBABILISTIC RISK ASSESSMENT WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual COURSE DOCUMENTATION. Volume 2: Probatxhty And Statistics Report For October 1984 Through March 1985.

For PRA Appications. NUREG/CR-3638: HYDROGEN-STEAM JET-FLAME FACILITY AND l NUREG/CR-4377: EVALUATIONS AND UTILIZATONS OF RISK IM- EXPERIMENTS.

I PORTANCES. NUREG/CR-3660 V01: PROBABILITY OF PIPE FAILURE IN THE RE-i OlVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST ACTOR COOLANT LOOPS OF WESTINGHOUSE PWR l 840429) PLANTS. Volume 1: Summary Report.

l NUREG/CR-3145 V03. GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-3660 V04: PROBABILITY OF PIPE FAILURE IN REAC-l WESTERN OHIO.INDtANA REGION - ANNUAL REPORT.(October TOR COOLANT LOOPS OF WESTINGHOUSE PWR

! 1982 September 1983. Volume 3). PLANTS. Volume 4. Pipe Failure induced By Crack Growth in West I Cor.st Plants.

NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL ,

ACCIDENTS METHODS. COSTS AND SCHEDULES FOR DECON. NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACILITIES.

l TAMINATION NUREG/CR 3876: PROBABILITY BASED LOAD COMBINATION CRI-NUREG/CR-3442: RADTWO.A COMPUTER CODE FOR SIMULATING TERIA FOR DESIGN OF CONCRETE CONTAINMENT STRUC-AS TRANSIE M SiONAL,TW LAY R RADION NU E /CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED LAKES, RESERVOIRS. RIVERS. ESTUARIES,AND COASTAL R E. FROM TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE GIONS VESSEL NUREG/CR-3809: EVALUATON OF NEUTRON DOSIMETRY TECH- NUREG/CR-3949 V02: EDDY CURRENT INSPECTION FOR STEAM NIQUES FOR WELL-LOGGING OPERATONS. GENERATOR TUBING PROGRAM. Annual Progress Report For

^

WAS R POSI B LT NUR / 4080 0 RMI ON OF THE AVAILABluTY OF NUREG/CR-3736: FIELD AND THEORETICAL INVESTIGATIONS OF CORE EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SIT-FRACTURED CRYSTALLINE ROCK NEAR ORACLE. ARIZONA. UATIONS.

NUREG/CR-3900 V04: LONG-TERM PERFORMANCE OF MATERI- NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE ALS USED FOR HIGH-LEVEL WASTE PACKAGING Annual N G/$ 19 E C IN ENT PENETRATIONS NUR CR 4 9 YS S OF THE ABILITY OF CURRENT UNDER SEVERE ACCIDENT CONDITIONS FY84 ANNUAL H AL PHYS STRUMENTS TO PREDICT DOSE IN EX- -4130- pCEDFA CODE FOR AEROSOL PARTICLE CAP.

NURE /

NUREG/CR-4251 V01: MITIGATIVE TECHNIQUES FOR GROUND-NU E / 3 T PREDICTIONS FOR THE RESPONSE WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR ACCIDENTS Volume 1-Analysis Of Genenc Site Conditions. OF A,1:8-SCALE STEEL LWR CONTAINMENT BUILDING MODEL NUREG/CR4251 V02 MITIGATIVE TECHNIQUES FOR GROUND-NUREG/C 4138 NA SE OR NEVADA TEST SITE (NTS)

WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR PREMIXED COMBUSTON TESTS ACCtDENTS. Volume 2. Case Study Analysis Of Hydrologic Character.

NUREG/CR-4182: VERIFICATON OF SOIL STRUCTURE INTERAC-tration And Mstigative Schemes.

TON METHODS NUREG/CR-4254. OCCUPATIONAL DOSE REDUCTION AND ALARA NUREG/CR4219 VO1: HEAVY-SECTION STEEL TECHNOLOGY PRO.

AT NUCLEAR POWER PLANTS. Study On High-Dose Jobs,Radwaste GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 -

Handing.And /LARA incentrves.

MARCH 1985 NUREG/CR 4259. TAIUNGS NEUTRAUZATON AND OTHER ALTER- NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC NATIVES FOR IMMOBluZlNG TOXIC MATERIALS IN MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY-TAluNGSFmal Report FEATURE SYSTEMS OF NUCLEAR POWER PLANTS.

NUREG/CR4266: STANDARD BETA-PARTICLE AND MONOENER. NUREG/CR-4249 PRESSURE VESSEL FRACTURE STUDIES PENE-GETIC ELECTRON SOURCES FOR THE CALIBRATON OF BETA- TRATING TO THE PWR THERMAL SHOCK ISSUE. EXPERIMENTS RADIATON PROTECTION INSTRUMENTATION. TSE-5,TSE-5A AND TSE4.

NUREG/CR4268: RATIO METHODS FOR COST-EFFECTIVE FIELD NUREG/CR-4257: INSPECTION. SURVEILLANCE.AND MONITORING SAMPUNG OF COMMERCIAL RADIOACTIVE LOW-LEVEL OF ELECTRICAL EQUIPMENT INSIDE CONTAINMENT OF NUCLE-WASTES. AR POWER PLANTS-WITH APPLICATIONS TO ELECTRICAL NUREG/CR4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA CABLES.

AND EA; STERN TENNESSEE EAATHOUAKES (1978-1984). NUREG/CR-4275: HEAVY SECTION STEEL TECHNOLOGY PRO-NUREG/CR4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS GRAM - FIVE-YEAR PLAN FY1984-1988.

SAFETY ANALYSIS. Phase 1, Fnal Report NUREG/CR4284: NEU'RON EXPOSURE PARAMETERS FOR THE NUREG/CR4317 V01: CANADIAN SEISMIC AGREEMENT. Technical FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATION Report Covenng 1979-1985. SERIES.

NUREG/CR-4333: STE. GENT.VIEVE FAULT ZONE. MISSOURI AND NUREG/CR4287; ENVIRONMENTALLY ASSISTED CRACKING IN ILUNOIS. LIGHT WATER REACTORS. Annual Report. October 1983 - Septern.

NUREG/CR-4352- SUGGESTED STATE REQUIREMENTS AND CRI- ber 1984.

TERIA FOR A LOW LEVE' RADIOACTIVE WASTE DISPOSAL SITE NUREG/CR-4290 V02: PROBABIUTY OF PIPE FAILURE IN THE RE-REGULATORY PROGRAH. ACTOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR NUREG/CR-4354: A STUDY OF SEISMICITY AND TECTONICS IN PLANTS. Volume 2:Guillohne Break Indirectly induced By Earth-NEW ENGLAND. Final Roport quakes.

NUREG/CR4355 V01: 233 PU(IV) IN MONKEYS.Ovennew Of Metabo. NUREG/CR-4291: CONCLUSION AND

SUMMARY

REPORT ON lism. PHYSICAL BENCHMARKING OF PIPING SYSTEMS.

NUREG/CR4379 V01: LONG-TERM PERFORMANCE OF MATERI- NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE ALS USED FOR HIGF-iEVEL WASTE PACKAGING.First Quarterty REACTOR COOLANT PUMP.

i Report, Year Four AprikJune 1985. NUREG/CR4300 V01: ACOUSTIC EMISSION / FLAW RELATIONSHIP EARTH SCIENCES BRANCH (POST 840429) FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR-4339- A FEVIEW OF RECENT RESEARCH ON THE VESSELS Progress Report, October-March 1985.

SEISMOTECTONICS OF THE SOUTHEASTERN SEABOARD AND NUREG/CR4304: PRESSURE VESSEL FRACTURE STUDIES PER.

AN EVALUATON 0: HYPOTHESES ON THE SOURCE OF THE TAINING TO THE PWR THERMAL-SHOCK ISSUE Expenment TSE-1886 CHARLESTON.GDUTH CAROUNA EARTHOUAKE. 7.

OlVISON OF ENGINEERING TECHNOLOGY NUREG/CR-4305: COMMENTS ON THE LEAK-BEFORE-BREAK NUREG/CR 1677 V02: PIPING BENCHMARK PROBLEMS. VOLUME 11 CONCEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS.

DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION RE- NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH SFONSE SPECTRUM HETHOD. PROGRAMS.Quarterty Report. January. March 1985.

NRC Contract Sponsor Index 61 NUREG/CR 4325: A PARAMETRIC STUDY OF PWR PRESSURE NUREG/CR-4125 V02: GUIDEUNES AND WORKBOOK FOR AS-VESSEL INTEGRITY DURING OVERCOOUNG SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTlu-ACCOENTS.CONSIDERING BOTH 2-D AND 3-D FLAWS. TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR-4331: SIMPLIFIED SEISMIC PROSABluSTIC RISK PLANT. Volume 2 Workbook For Assessment Of Organizabon And ASSESSMENT Procedures And umitabons. Management NUREG/CR-4334. AN APPROACH TO THE QUANTIFICATON OF NUREG/CR-4248: RECOMMENDATIONS FOR NRC POUCY ON SEISMIC MARGINS IN NUCLEAR POWER PLANTS. SHIFT SCHEDOUNG AND OVERTIME AT NUCLEAR POWER NUREG/CR-4358. APPUCATONS OF DENSITY PROFIUNG TO Pt. ANTS.

EOUIPMENT OUALIFICATON ISSUES. NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL NUCLEAR POWER PLANT CONTROL ROOM CREWS.

PURPOSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON NUREG/CR-4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLE-OF NUCLEAR REACTOR VuSSELS AND PIPING COMPONENTS- AR POWER PLANT MAINTENANCE FACTORS AND PLANT NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT CAFETY PRAIRIE ISt.AND NUCLEAR GENERATING STATION.

DIVISION OF SYSTEMS INTEGRATION (POST 811005)

N G/ 68: SENmW AND MMAINW SMS &

EDO-RESOURCE MANAGEMENT OFFICE OF RESOURCE MANAGEMENT, DIRECTOR THE CRAC2 COMPUTER CODE.

NUREG/CR-4398: COST ANALYSIS OF REVtSIONS TO 10 CFR NUREG/CR-4143: REVIEW AND EVALUATON OF THE MILLSTONE PART 50, APPENDIX J. LEAK TESTS FOR PRIMARY AND SEC- UNIT 3 PROBABILISTIC SAFETY STUDY.Contamment Failure ONDARY CONTAINMENTS OF UGHT-WATER-COCLED NUCLEAR Modes.Radological Source-Terms And Offsite Consequences.

POWER PLANTS

• DIVISION OF SAFETY TECHNOLOGY NUREG/CR-2800 S03: GUIDEUNES FOR NUCLEAR POWER PLANT OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80) SAFETY ISSUE PRORITIZATON INFORMATON DEVELOPMENT.

OlVISION OF ENGINEERING NUREG/CR-2815 V01 RI: PROBABluSTIC SAFETY ANALYSIS PRO-NUREG/CR-3952 SEQUOYAH EQUIPMENT HATCl- SEAL LEAKAGE. CEDURES GUIDE.Sectons 1-7 And Appendices.

OlVISION OF HUMAN FACTORS SAFETY NUREG/CR-2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PRO.

NUREG/CR-4125 V01: GUIDEUNES AND WORKBOOK FOR AS- CEDURES GUIDE.Seebons 8-12.

SESSMENT OF ORGANIZATON AND ADMINISTRATION OF UTlu- NUREG/CR-3485: PRA REVIEW MANUAL TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER NUREG/CR-4326 VO1: EFFECTS OF CONTROL SYSTEM FAILURES PLANT. Volume 1: Guidelines For Ublity Organizaton And Administra. ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTING-bon Plan. HOUSE PRESSURIZED WATFR REACTOR. Man Report I

Contractor Index This index lists, in alphabetical order, the numbers and titles of their reports. If further contractors that prepared the NUREG/CR information is needed, refer to the main ci-reports listed in this compilation. Listed tation by the NUREG/CR number.

below each contractor are the NUREG/CR AEROSPACE CORP. NUREG/CR-3613 V03 N1: EVALUATION OF WELDED AND REPAIR-NUREG/CR 4357: THE FEASIBILITY OF DETECTING THE IMPORT OF WELDED STAINLESS STEEL FOR LWR SERVICE.Semannual Report UNAUTHORIZED RADIOACTIVE MATERLALS INTO THE UNITED For October 1984 Through March 1985.

STATES. NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL AMES LABORATORY, ENERGY & MINERAL RLSOURCES RESEARCH NUREG/CR4130- ICEDF A CODE FOR AEROSOL PARTICLE CAP.

INSTITUTE TURE IN ICE COMPARTMENTS.

NUREG/CR-3952: SEOUOYAH EQUIPMENT HATCH SEAL LEAKAGE. NUREG/CR4151: INTEGRATON OF EMERGENCY ACTION LEVELS WITH COUBUSTION ENGINEERING EMERGENCY OPERATING N 37 RA STUDIES OF A BREACHED NUCLE- ROCENS.By Use of htion Engineenng h Grmp AR WASTE REPOSITORY IN BASALT P Emergency Operatirt rocedure Techrucal Guidehnes.

NUREG/CR 3980 V04: UGHT-WATERJIEACTOR SAFETY FUEL SYS- NUREG/CR-4248: REwMMENDATIONS FOR NRC POLICY ON SHIFT TEMS RESEARCH PROGRAMS. Quarteny Progress Report,0ctober. SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS.

December 1984 NUREG/CR-4251 V01: MITIGATIVE TECHNIQUES FOR GROUND-NUREG/CR-4240' VO1: PHYSICS OF REACTOR SAFrTY.Ouarteriy WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR Report. January-March 1985. ACCIDENTS. Volume 1: Analysis Of Genenc Site Condrtions.

NUREG/CR4287: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4251 V02: MITIGATIVE TECHNIOUES FOR GROUND-UGHT WATER RFACTORS. Annual Report,0ctober 1983 - September WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR 1984. ACCIDENTS. Volume 2. Case Study Analysis Of Hydrolegic Character-tration And Metigatrve Schemes.

ARIZONA, UNIV. OF, TUCSON, AZ NUREG/CR-4259: TAILINGS NEUTRALIZATION AND OTHER ALTER-NUREG/CR-3736: FIELD AND THEORETICAL INVESTIGATIONS OF NATIVES FOR fMMOBluZING TOXIC MATERIALS IN TAILINGS Final FRACTURED CRYSTALLINE ROCK NEAR ORACLE ARIZONA- Report.

ARVIN/CALSPAN ADVANCED TECHNOLOGY CENTER NUREG/CR-4268. RATIO METHODS FOR COST-EFFECTIVE FIELD NUREG/CR4257: INSPECTION SURVEILLANCE,AND MONITORING SAMPLING OF COMMERCIAL RADIOACTIVE LOW-LEVEL WASTES.

OF ELECTRICAL EQUIPMENT INSIDE CONTAINMENT OF NUCLEAR NUREG/CR4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY.

POWER PLANTS-WITH APPLICATIONS TO ELECTRICAL CABLES.

NUREG/CR-4298: DESIGN AND INSTALLATION OF COMPUTER SYS-ASPEN SYSTEMS, INC. TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55.

NUREG-0386 003: UNITED STATES NUCLEAR REGULATORY COM. NUREG/CR4300 V01: ACOUSTIC EMISSION / FLAW RELATIONSHIP MISSION STAFF PRACTICE AND PROCEDURE DIGEST. JULY 1972, FOR IN-SERVICE MONITORING OF NUCLEAR PRESCURE SEPTEMBER 1983. VESSELS Progress Report, October-March 1985.

NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH BATTELLE HUMAN AFFAIRS RESEARCH CENTERS PROGRAMS. Quarterly Report, January-March 1985.

NUREG/CR-4125 V01: GUIDELINES AND WORKBOOK FOR ASSESS-MENT OF ORGANIZATION AND ADMINISTRATION OF UTIUTIES BOSTON COLLEGE, CHESTNUT HILL, MA SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR-4354: A STUDY OF SEISMICITY AND TECTONICS IN NEW PLANT. Volume 1:Guidehnes For Util.ty Orgaruzation And Adtrurustration ENGLAND. Final Report.

Plan.

NUREG/CR-4125 V02 GUIDELINES AND WORKBOOK FOR ASSESS. BROOKHAVEN NATIONAL LABORATORY MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES NUREG/CP-0066: PROCEEDINGS OF AN INTERNATION/L WORK-SEEKING OPERATING UCENSE FOR A NUCLEAR POWER SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION PLANT. volume 2: Workbook For Assessment Of Orgaruzation And Man- (ALARA) AT NUCLEAR POWER PLANTS.MAY 29-JUNE 1,1964.

agerrant. NUREGICR-1677 V02- PIPING BENCHMARK PROBLEMS.VO' UME 11 NUREG/CR-4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION RE-POWER PLANT mal #ENANCE FACTORS AND PLANT SAFETY. SPONSE SPECTRUM METHOD.

NUREG/CR-2815 V01 R1: PROBABILISTIC SAFETY ANALYSIS PVOCE-DATTELLE MEMORIAL INSTITUTE. COLUMBUS LABORATORIES DURES GUIDE. Sections 17 And Appendices.

NUREG/CR-3900 V04. LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR-2815 V02 Rl: PROBABILISTIC SAFETY ANALYSIS PROCE-USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report.Apnl DURES GUIDE.See2 8-12.

I e Oc M h 985 NU G/CR-4377: EVALUATIONS AND IJTILIZATIONS OF RISK iM- 95 NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS NUREG/CR4379 V01: LONG-TERM PERFORMANCE OF MATERIALS ON SOUDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU-US FOR HI LEVEL ASTE PACKAGING.First Quarterty NURE / R 3485 PRA EVIEW MANUAL NUREG/CR-3876: PROBABlWTY BASED LCAD COMBINATION CRITE-BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES.

LABORATORIES NUREG/CR 4143: REVIEW AND EVALUATION OF THE MILLSTONE NUREG/CP-0063: PROCEEDINGS OF THE 1984 STATISTICAL SYMPO. UNIT 3 PF,OBABILISTIC SAFETY STUDY. Containment Failure SIUM ON NATIONAL ENERGY ISSUES. Modes.Radiologcal Source-Terms And Offsite Consequences.

NUREG/CR-2800 S03. GUIDEUNES FOR NUCLEAR POWER PLANT NUREG/CR-4182: VERIFICATION OF SOIL STRUCTURE INTERACTION SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT. METHODS.

NUREG/CR-3413: OFF SITE CONSEQUENCES OF RADIOLOGICAL NUREG/CR-4252: INDEPENDENT ASSESSMENT OF TRAC PD2/ MOD 1 ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON. CODE WITH BCL ECC BYPASS TESTS.

g TAMINATION. NUREG/CR4254: OCCUPATIONAL DOSE REDUCTION AND ALAHA NUREG/CR 3G09: EVALUATION OF NEUTRON DOSIMETRY TECH- AT NUCLEAR POWER PLANTS. Study On High-Dose Jobs,Radweste NIQUES FOR WELL-LOGGING OPERATIONS. , Handhng.And ALARA Incentrves.

63 l

\

64 Contractor Index NUREG/CR-4291: CONCLUSION AND

SUMMARY

REPORT ON PHYSI- lLLINOIS, STATE OF CAL BENCHMARKING OF PIPING SYSTEMS. NUREG/CR4333: STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL-CANADA, GOYT. OF LINOIS.

NUREG/CR-4317 V01: CANADIAN SEISMIC AGREEMENT.Techrucal INTERIOR, DEPT OF, GEOLOGICAL SURVEY Report Covenng 1979-1985. NUREG/CR 4339: A REVIEW OF RECENT RESEARCH ON THE SEIS-COhMERCE, DEPT. OF, NATIONAL BUREAU OF STANDARDS MOTECTONICS OF THE SOUTHEASTERN SEABOARD AND AN EVALUATION OF HYPOTHESES ON THE SOURCE OF THE 1886 NUREG/CR4266: STANDARD BETA-PARTICLE AND MONOENERGE- CHARLESTON. SOUTH CAROUNA EARTHOUAKE.

TIC ELECTRON SOURCES FOR THE CAllBRATION OF BETA-RADI-ATION PROTECTION INSTRUMENTATION. LAWRENCE BERKELEY LABORATORY COMMERCE, DEPT. OF, NATL OCEANOGRAPHIC & ATMOSPHERIC NUREG/CR-4355 VOI: 238 PU(IV) IN MONKEYS. Overview Of Metabo-lism.

ADMINISTRATION NUREG/Ch 4038: SENSITIVITY AND UNCEPTAINTY STUDIES OF THE LAWRENCE LIVERMORE NATIONAL LABORATORY CRAC2 COMPUTER CODE. NUREG/CR-3660 VOI: PROBABILITY OF PIPE FAILURE IN THE REAC-CONFERENCE OF RADIATION CONTROL PROGRAM DIRECTORS,INC. TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 1 Summary Report.

NUREG/CR-4352 SUGGESTED STATE REQUIREMENTS AND CRITE- NUREG/CR-3660 V04: PROBABILITY OF PIPE FAILURE IN REACTOR RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4 Pipe REGULATORY PROGRAM. Failure induced By Crack Growth in West Coast Plants.

CREARE, INC. NUREG/CR-4239: ANALYSIS OF THE ABluTY OF CURRENT HEALTH PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDIVID-NUREG/CR-3426 VO1: THERMAL AND FLUID MIXING IN 1/2-SCALE UALS.

TEST FACluTY. Facility And Test Dese Report NUREG/CR42SO V02: PROBABILITY OF PIPE FAILURE IN THE REAC-NUREG/CR-3426 V02: THERMAL AND FLUO MIXING IN 1/2-SCALE TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR TEST FACluTY. Data Report.

PLANTS Volume 2:Gudiotine Break Indirectly induced By Earthquakes.

E.C. RODABAUGH ASSOCIATES,INC. NUREG/CR-4331: SIMPUFIED SEISMIC PROBABILISTIC RISK ASSESSMENT. Procedures And Urrwtations.

NUREG/CR4305: COMMENTS ON THE LEAK-BEFORE-BREAK CON- NUREG/CR-4334: AN APPROACH TO THE QUANTIFICATION OF SEIS-CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS. MIC MARGINS IN NUCLEAR POWER PLANTS.

EG4G IDAHO, INC. (SUBS. OF EG&G, INC.),

LOS ALAMOS SCIENTIFIC LABORATORY NUREG/CR-3301: CATALOG OF PRA DCMINANT ACCIDENT SE- NUREG/CR-3706: TRAC ANALYSES OF SEVERE OVERCOOLING OUENCE LNFORMATION.

TRANSIENTS FOR THE OCONEE 1 PWR.

NUREG/CR-3633 V01 S1: TRAC-BD1/ MOD 1:AN ADVANCED BEST ES- NUREG/CR-4107: SEQUENTIAL TEST PROCEDURES FOR DETECT-TIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR ING PROTRACTED MATERIALS LOSSES.

TRANSIENT ANALYSIS. NUREG/CR4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER j

NUREG/CR-3633 V04. TRAC-BD1/ MOD 1;AN ADVANCED BEST ESTI-PLANT VALVE FAILURE-RATE VARIABILITY--SOME PREUMiNARY MATE COMPUTER PROGRAM FOR BOILING WATER REACTOR 9ESULTS.

TRANSIENT ANALYSIS Volume 4: Developmental Assessment.

NUREG/CR 4232- THE RESPONSE OF VENTILATION DAMPERS TO NUREG/CR-3819: SURVEY OF AGED POWER PLANT FACluTIES. LARGE AIRFLOW PULSES.

NUREG/CR-3935: THERMAL-HYDRAULIC ANALYSES OF OVERCOOL- NUREG/CR-4260- TORAC USER'S MANUALA Computer Code For Ana-ING SEQUENCES FOR THE H.B. ROBINSON UNIT 2 PRESSURIZED lyzing Tornado-induced Flow And Matenal Transport in Nuclear FacdF THERMAL SHOCK STUDY. ties.

NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPERATIONAL NUREG/CR-4314 BRIEF SURVEY AND COMPARISON OF COMMON TRANSIENT (OPTRAN) TESTS 1-1 AND 12 IN THE POWER BURST 1AUSE FAILURE ANALYSIS.

FACluTY.

NUREG/CR4080- DETERMINATION OF THE AVAILABluTY OF CORE MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE, MA EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATIONS. NUREG/CR-4376: HEAT TRANSFER. CARRYOVER AND FALL BACK IN NUREG/CR-4150: EPICOR-il RESIN DEGRADATION RESULTS FROM PWR STEAM GENERATORS DURING TRANSIENTS.

FIRST FiESIN SAMPLES OF PF-8 AND PF-20.

NUREG/CR-4227: HUMAN ENGINEERING GUOEUNES FOR THE M ATHTECH, INC.

EVALUATION AND ASSESSMENT OF VISUAL DISPLAY UNITS. NUREG/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART NUREG/CR-4272: RESPONSE TREE EVALUATION. EXPERIMENTAL 50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR CONTAINMENTS OF UGHT WATER-COOLED NUCLEAR POWER OPERATORS. PLANTS.

NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSlENTS AND ACCOENTS AT A 3-LOOP WESTINGHOUSE MEMPHIS STATE UNIV, MEMPHIS, TN PRESSURIZED WATER REACTOR. Main Report. NUREG/CR-4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT AND EASTERN TENNESSEE EARTHOUAKES (1978-1984).

PRAIRIE ISLAND NUCLEAR GENERATING STATION.

MICHIGAN, UNIV. OF, ANN ARBOR, M1 GA TECHNOLOGIES. lNCJGENERAL ATOMIC CO. NUREG/CR-3145 V03: GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS WESTERN OHIO-INDIANA REGION - ANNUAL REPORT.(October SAFETY ANAL YSIS. Phase 1 Final Report. 1982 - September 1983, Volame 3).

GENERAL PHYSICS CORP. NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU-NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU- CLEAR REACTOR VESSELS AND PIPING COMPONENTS.

CLEAR POWER PLANT CONTROL ROOM CREWS.

NUREG/CR-4280 THE EFFECTS OF SUPERVISOR EXPERIENCE AND OAK RIDGE NATIONAL LABORATORY ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW NUREG/CR-2000 V04 N6: UCENSEE EVENT REPORT (LER)

PERFORMANCE IN CONTROL ROOM SIMULATORS. COMPILATION.For Month Of June 1985.

HANFORO ENGINEERING DEVELOPMENT LABORATORY NUREG/CR-2000 V04 N7: UCENSEE EVENT REPORT (LER)

COMPtLATION For Month Of July 1985.

NUREG/CR-3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME-NUREG/CR-2000 V04 N8: UCENSEE EVENT REPORT (LER)

TRY IMPROVEMENT PROGRAM. LWR Power Reactor Surveillance COMPILATION.For Month Of August 1985.

Physics-Dosametry Data Base Compendium.

NUREG/CR-3442: RADTWO:A wMPUTER CODE FOR SIMULATING HARVARD UNIV., CAMBRIDGE, M A FAST. TRANSIENT, TWO-DIMENSIONAL TWO-LAYER RADIONU-CLOE CONCENTRATION CONDITIONS IN NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER PLANT ACCLDENT LAKES. RESERVOIRS. RIVERS. ESTUARIES.AND COASTAL REGIONS.

CONSEQUENCE ANALYSIS Part NUREG/CR-3481 V02: NUCLEAR POWER PLANT PERSONNEL QUAU-1: Introduction, Integration & Summary.Part II.Scientinc Basis For Heafth Effects Models.

FICATIONS AND TRAINING: TAPS - The Task Analysis Profiling System.

g

Contractor Index 65 NUREG/CR-3851 V04: EVALUATION OF RAD 60NUCUDE GEOCHEMI- S. COHEN & ASSOCIATES, INC.

CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NUREG/CR4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART WASTE REPOSITORY SITE PROJECTS Annual Progress Report For 50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY October 1983-September 1984. CONTAINMENTS OF LIGHT.WATERC)OLED NUCLEAR POWER NUREGICR-3885 V04: HIGH-TEMPERATURE GASC)OLED REACTOR PLANTS.

SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION.Quarterty Progress Report, October 1-December SANDIA NATIONAL LABORATORIES 31.1984. NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY NUREG/CR-3949 V02: EDDY 4URRENT INSPECTION FOR STEAM PROTECTONllt SUBMICRON PARTICLE TESTS AND

SUMMARY

RA B PROGRAM. Annual Progress Report For Penod OF 1 gR YDROGEN-STEAM JET-FLAME FACluTY AND EX-NURE CR 7 TA

SUMMARY

REPORT FOR FISSION PRODUCT NUIE 18 V03: REACTOR SAFETY RESEARCH Ouarterty NUREG/CR-4038: SEN'SITIVITY AND UNCERTAINTY STUDIES OF THE CRAC2 COMPUTER CODE.

NUR CR 38 EACTOR SAFETY RESEARCH.Ouartetty NUREG/CR-4081: ABSORPTON OF GASEOUS LODINE BY WATER RWOctobw Decembw 1984 DROPLETS. NUREG/CR-4060: THE DC-1 AND DC-2 DEBRIS COOLABluTY AND MELT DYNAMICS EXPERIMENTS.

NUREG/CR4219 V01: HEAVY-SECTON STEEL TECHNOLOGY PRO- NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Computer GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 - Code for Severe Reactor Accident Contamment Analyses.

MARCH 1985. NUREG/CR-4119: INTEGRITY OF CONTAINMENT PENETRATIONS NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC UNDER SEVERE ACCOENT CONDITONS FY84 ANNUAL REPORT.

MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY FEA- NUREG/CR-4122 A FORTRAN 77 PROGRAM AND USER'S GUIDE TURE SYSTEMS OF NUCLEAR POWER PLANTS. FOR THE CALCULATION OF PARTIAL CORRELATION AND STAND-NUREG/CR-4249 PRESSURE VESSEL FRACTURE STUDIES PENE. ARD12ED REGRESSION COEFFICIENTS.

TRATING TO THE PWR THERMAL-SHOCK ISSUE; EXPERIMENTS NUREG/CR-4137: PRETEST PREDICTIONS FOR THE RESPONSE OF TSE-5.TSE-5A AND TSE4. A 1:8-SCALE STEEL LWR CONTAINMENT BUILDING MOOEL TO NUREG/CR-4255 V01: AEROSAL RELEASE AND TRANSPORT PRO. STATO OVERPRESSUR12ATON.

GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 NUREG/CR-4138: DATA ANALYSES FOR NEVADA TEST SITE (NTS)

MARCH 1985. PREMIXED COMBUSTON TESTS.

NUREG/CR-4257: INSPECTION. SURVEILLANCE,AND MONITORING NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA FOR AC.

I OF ELECTRICAL EQUIPMENT INSOE CONTAINMENT OF NUCLEAR CIDENTAL RADONUCUDE RELEASES FROM NUCLEAR REAC-POWER PMNTS--WITH APPLICATONS TO ELECTRICAL CABLES. TORS.

NUREG/CR-4275: HEAVY-SECTON STEEL TECHNOLOGY PROGRAM NUREG/CR-4213: SETS REFERENCE MANUAL FIVE-YEAR PMN FY 1984-1988 NUREG/CR 4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER NUREG/CR-4280 THE EFFECTS OF SUPERVISOR EXPERIENCE AND PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW Untroducbon, Integration & Summary.Part II.Soentific Basis For Health PERFORMANCE IN CONTROL ROOM SIMULATORS Effects Models.

NUREG/CR-4284: NEUTRON EXPOSURE PARAMETERS FOR THE NUREG/CR-4250- VEHICLE BARRIERS. EMPHASIS ON NATURAL FEA-F HEAVY SECTON STEEL TECHNOLOGY IRRADIATON NU E /CR-4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS NUREG/CR-4304: PRESSURE VESSEL FRACTURE STUDIES PER- NU G CR 3 TAINING TO THE PWR THERMAL-SHOCK ISSUE. Experiment TSE-7. 02 PROBA STIC SK ASSESSMENT COURSE DOCUMENTATION. Volume 2- Probabshty And Statistics For PRA AppH-NUREG/CR-4305: COMMENTS ON THE LEAK-BEFORE-BREAK CON-CEPT FOR NUCLEAR POWER PLANT PtPING SYSTEMS. NU WUREG/CR 4325: A PARAMETRIC STUDY OF PWR PRESSURE CR-4358: APPUCATONS OF DENSITY PROFIUNG TO EQUIP-VESSEL INTEGRITY DURING MENT OVAUMCATION ISSUES' OVERCOOUNG ACCIDENTS.CONSOERING BOTH 2-0 AND 3-D FLAWS. SCIENCE & ENGINEERING ASSOCIATES,INC.

DARAMETER INC. NUREG/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART 50, APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY NUREG/CR-4008: CLOSEOUT OF IE BULLET.N 81-01: SURVEILLANCE OF MECHANICAL SNUBBERS. CONTAINMENTS OF UGHT WATER-COOLED NUCLEAR POWER PLANTS.

RENSSELAER POLYTECHNIC INSTITUTE. TROY, NY VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV, SLACKS 8URG, NUREG/CR-4118: NUFEGO-NP:A DIGITAL COMPUTER CODE FOR VA THE UNEAR STABluTY ANALYSIS OF BOluNG WATER NUCLEAR NUREG/CR4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA REACTORS.

AND EASTERN TENNESSEE EARTHOUAKES (1978-1984).

ROCKWELL INTERNATIONAL CORP. WATER, WASTE & LAND, INC.

NUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE NUREG/CR-3901: DOCUMENTATON AND USER'S GUIDE:GS2 & GS3 REACTOR COOLANT PUMP. - VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS.

l l

Licensed Facility IndOx This index lists the facilities that were the Docket number and followed by the report f subject of NRC staff or contractor reports. number. If further information is needed, The facility names are arranged in alphabet- refer to the main citation by the NUREG ical order. They are preceded by their number.

50412 Beevw Valey Poew Stabcn Unt 2, D@esw NURE41094 S 353 Umanck Generetng Staton, Umt 2, PMedelpha NUREG4001 S05 Light Ca Elects Ca S 346 Deve-Besse Nuclear Power Staton, Und 1, NURE41154 S 353 Lanenct Generetng Staton, Unt 2 PMedelpha NUREG0001 S06 Toisdo Ectson Ca Elects Ca S 275 DetWo Canyon pkicieer Power Plant, Und 1, NUREG4675 S32 S 423 Westone Phadear Pomer Staton, Umt 3, NUREG1031 S02 Pa:Ac Gas & Elects Co Norteest Nuclear Co S 275 Detdo Canyon Ptadear Power Plert, Und 1, NURE41151 S 423 Mestone Nudeer Pcoor Unt3, NUREG/CR4143 Pacic Gas & Electe Co Northeast Nudeer Enongy Co S 323 Detso Canyon Nudeer Power Plant, Und 2. NUREG4675 S32 S 200 Oconee Nudeer Sueon, Urd 1, DAe Power Ca NUREG/CR-3706 Peoec Gas & Elects Co 50L282 P6t* hiand Nudeer Steson, Unt 1, Nortem NUREG/CR4397 S 323 Detdo Canyon Nudeer Poser Plant, Urd 2, NURE41151 Steins Near Ca Peate Gee & Electe Co 50 306 Prano leier.1 Nudeer Staton, Urut 2, Northam NUREG/CR4397 S 341 ems Atrec Power Plant, Und 2. Detait NUREG4796 S06 States Pmgr Ca Sidi g=.n.: Pows, P= unt t ostet maii4, ,,,, ,Ca_,,,,,,,,,,,,,,,,,, ,,,,,,,,,

su rent ~. U- o, % NuRE 113. Ca STN S447 GESSE238, General Electe Ca NUREG4979 SO4 S 458 River Band Staton, Und 1, gun States Uittes MJIEG4000 SO4 54261 H.S. Rotmoon Plant, Una 2, Cardine Power & NUREG/CR 3935 Co.

Oght Ca 50458 Arver Bend Staton, Unt 1. gun States Uttles NURE41142 S 354 Hope Creek Nudeer Staton, Umt 1 Putic NUREG1048 S02 Ca Sarnce Electe & Gee Co S 443 Sestrook Nucteer Staten, Unit 1, Piele Service NUREG4996 S03 40 8027 Kar4acGee Nudeer Corp., Oldehoma Clly. OK, NURE41157 Ca of New 54352 tsnenca Generetng Staton, Und 1 Phledelphe NUREG4991 SOS 50444 Sestrook Nudeer und 2, Putic Servios NUREG0006 S03 Electr CE Ca of New Hengehr S 352 Lsnench Generetng Stolen, wJ 1, Phledelpha NUREG4991 S06 S 322 Shoreham Nucteer Power Staton, Long leland NUREG1126 Elodnc Ca Ughtng Co.

67

l N QQM 336 U 8. NUCLE AR REGULATORv COMui1880N 1 REPORT NUM6ER (Ass papa ey TsDC. #de For 40, if edFi k', 3, - BIBLIOGRAPHIC DATA SHEET NUREG-0304 Vol. 10, No. 3 SEE INSTRL rONS ON TME REVER5E 2 TITLE ANO hm TITLE J LE AVE SLANE Regul at and Technical Reports (Abstract Index Journal)

Compflati for Third Quarter 1985 July - Sep mber aoArtRf-TCOMPuTED MO~ T ,, VEAa j

S AUTHORISI

[O AT E REPOR T IiSUED vEAR MO7 l Octob# 1985 7 PERFORMING ORGANi2 ATION NAME D MAILING ADDRE55 ffac4deld Coes a PROJECT A5E /WC RK LNif hvMSE R Division of Technica Information and Document Control I Office of Administrat n *"'"""""

U.S. Nuclear Regulator Comission Washington, DC 20555 to SPONSORING ORG ANilATeON NAME AN0 MAILING '3RE55 tfackde ld Codes tie TvPE OP REPORT Quarterly Same as 7, above. , ,E R,Oo CO, E R E o ,,,,s... d.,,s, July - September 1985 12 SUPPLEMENT ARv NOTES 13 ASSTR ACT (200 *eres or 'esst This journal lists all formal reports in NUREG series pmpared by the NRC staff and contractors, as well as proceedings o onferences and workshops. The entries in the compilation are indexed for access t le and abstract, contractor report number, personal author, subject, NRC organi;:a on, c tractor, and licensed facility.

14 oOCVVENT ANALvliS - e KEYWOROS CR,PTORS 15 AV Asta tiLif y STATEMENT abstract index Unlimi ted 26 SECURITY CLA5544 CATION IThe pares

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