ML20214D245

Regulatory and Technical Reports.Compilation for Third Quarter 1986,July-September

ML20214D245
Person / Time
Issue date:	11/30/1986
From:	NRC OFFICE OF ADMINISTRATION (ADM)
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References
NUREG-0304, NUREG-0304-V11-N03, NUREG-304, NUREG-304-V11-N3, NUDOCS 8611210431
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NUREG-0304 Vol.11, No. 3 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for Third Quarter 1986 -

July - September U.S. Nuclear Regulatory Commission Office of Administration ps* "'*%4 ,

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l Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 -

A year's subscription consists of 4 issues for this publication.

Single copies of this publication .

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NUREG-0304 Vol.11, No. 3 Regulatory and Technical Reports (Abstract Index Journal) i Com ailation for Thirc Quarter 1986 July - September Date Published: November 1986 Policy and Publications Management Branch Division of TechnicalInformation and Document Control Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555

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CONTENTS Preface . . . . . . . ...................................................................v 1

Index Tab Main Citation and Abstracts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Staff R eoo rts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Conference Proceedings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Contractor Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 Contractor Report Number index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 Personal Author index . . . . . . . . . . . . . . . .. ................................ ............ 3 1

S ubject i ndex . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 d

NRC Originating Organization Index (Staff Reports) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 NRC Contract Sponsor index (Contractor Reports) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Contractor !ndex . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . ... . .7 4 Licensed Facility index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 1

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PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical a 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:

Division of Technicalinformation and Document Control Policy and Publications Management Branch Publishing and Translations Section Woodmont 501 U.S. Nuclear Regulatory Commission Washington, D.C. 20666 i The main citations Lnd abstracts in this compilar co 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)

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

The bibliographic elements of the main citations are the following:

Staff Report NU8tEG-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 intemal NRC use).

, Conference Report 1

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

Laboratory. May 1981.141 pp. 810si?nn?eq. ANL-81-3. 00832

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 intomal use).

Contractor Report NUREG/CR-1566: 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 intemal use).

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

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

Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275-2171. Non-U.S. customers must make payment in advance either by 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. i l

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Main Citations and Abstracts The report listings in this compilation are ar- is an NRC contractor-prepared report. The ranged by report number, where NUREG- bibliographic information (see Preface 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.

conferenca report, and NUREG/CR-XXXX NUREG-0020 V10 N05: LICENSED OPERATING REACTORS granular stress corrosion cracking suscept.bility of BWR ASME STATUS

SUMMARY

REPORT. Data As Of April 30,1986.(Gray Code Class 1, 2, and 3 pressure boundary piping and safe Book) POSS.P.A.; BEEBE.M.R. Division of Budget & Analysis. ends. For piping that does not fully compfy with the material se-July 1986. 459pp. 8608080060. 37446:098.

lection, testing, and processing guideline combinations of this The OPERATING UNITS STATUS REPORT - LICENSED OP- docurnent, varying degrees of augmented inservice inspection ERATING REACTORS provides data on the operation of nucle-will be required, pursuant to 10 CFR 50.55a(g)(6)(ii). This revi-ar units as timely and accurately as possible. This information is sion also includes guidance regarding crack evaluation and weld collected by the Office of Resource Management from the Headquarters staff of NAC's Office of Inspection and Enforce- overlay repair methods for long term operation or for continuing ment, from NRC's Regional Offices, and from utilities. The three interim operation of plants until a more permanent solution is implemented.

sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provid- NUREG-0386 D04 R02: UNITED STATES NUCLEAR REGULA-ed by NRC's Regional Offices, IE Headquarters and the utilities, TORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST, JULY 1,1972 - DECEMBER 31,1986.

• Offee of the and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- General Counsel.

• Aspen Systems, Inc. August 1986 563pp.

power reactors in the U.S. It is hoped the report is helpful to all 8609170018. 37870:001.

agencies and individuals interested in maintaining an awareness This Revision 2 of the fourth edition of the NRC Staff Practice of the U.S. energy situation as a whole. and Procedure Digest contains a digest of a number of Com-mission, Atomic Safety and Licensing Appeal Board, and Atomic NUREG 0020 V10 N06: LICENSED OPERATING REACTORS Safety and Licensing Board decisions issued during the period STATUS

SUMMARY

REPORT. Data As Of May 31,1986.(Gray July 1,1972 to December 31, 1985, interpreting the NRC's Book) ROSS,P.A.; BEEBE.M.R. Division of Budget & Analysis. Rules of Practice in 10 CFR Part 2. This Revision 2 replaces in August 1986. 464pp. B608260411. 37649:091 part earlier editions and supplements and includes appropriate See NUREG-0020 V10 N05 abstract chances reflecting the amendment to the Rules of Practice ef-NUREG-0040 V10 NO2: LICENSEE CONTRACTOR AND fective Decernber 31,1985.

VENDOR INSPECTION STATUS REPORT. Quarterty Report. April-June 1986.(White Book)

• Division of OA, Vendor & NUREG-0430 V06 N02: LICENSED FUEL FACILITY STATUS P

REPORT. inventory Difference Data. July-December 1985.(Gray 06 . 25 10. 7 06. Book !!).

• Office of Inspection & Enforcement, Director (Post This periodical covers the results of inspections performed by 820201). August 1986.15pp. 8609120243. 37815:244.

the NRC's Vendor Program Branch that have been distributed NRC is committed to the periodic publication of licensed fuel to the incpected organizations during the period from April 1986 facilities inventory difference data, following agency review of thru June 1986. Also, included in this issue are the results of the information and completion of any related NRC investiga-certain inspections performed prior to April 1986 that were not tions. Information in this report includes inventory diffe ence included in previous issues of NUREG-0040. data for acthe fuel fabrication facilities possessing more than one effective kilogram of high ennched uranium, low enriched NUREG-0304 V11 N02: REGULATORY AND TECHNICAL uranium, plutonium, or uranium-233.

REPORTS. Compilation For Second Quarter 1986, April-June.

• Division of Technical information & Document Control July 1986. 75pp. 8608190550. 37564:001- NUREG 0540 V08 N05: TlTLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.MAY 1-31, 1986.

• Dnnsion of Technical This joumal includes all formal reports in the NUREG series by the NRC staff and contractors, as well as proceedings of information & Document Control July 1986. 644pp.

8607240291. 3724'L057.

conferences and workshops. The entries in the compilation are indexed for access by title and abstract, contractor report This hment is a monthly publication containing descno-number, personal author, subject, NRC organization, contractor, tions of information received and generated by the U.S. NRC.

and licensed facihty. This information includes (1) docketed rnatorial associated with civlian nuclear power plants and other uses of radioactive ma-NUREG-0313 R02 DRFT: TECHNICAL REPORT ON MATERIAL terials, and (2) nondocketed material received and generated by SELECTION AND PROCESSING GUIDELINES FOR BWR NRC pertinent to its role as a regulatory agency. The following COOLANT PRESSURE BOUNDARY PIPING. HAZELTON,W.S. Indexes are included: Personal Author Index, Corporate Source Division of Boiling Water Reactor (BWR) Licensing. June 1986. Index, Report Number index, and Cross Reference to Principal 60pp. 8607240368. 37241:297.

Documents Index.

This report updates and supersedes the technical positior's NRC established in NUREG-0313 " Technical Report on Materi. NUREG-0540 V00 N06: TITLE LIST OF DOCUMENTS MADE at Selection and Processing Guidelines for BWR Coolant Pres- PUBLICLY AVAILABLE. JUNE 1-30, 1986.

• Division of Techni-sure Boundary Piping," published in July 1977, and its subse- cal Information & Document Control August 1986. 501pp.

quent revision published in July 1980. This report sets forth the 8608190538. 375b8:105.

NRC staff's revised acceptable methods to control the inter- See NUREG-0540 V05 NOS abstract.

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2 Main Citations and Abstracts NUREG-0540 V08 N07: TITLE LIST OF DWJMENTS MADE County, Illinois. This supplement reports the status of items that PUBLICLY AVAILABLE JULY 131,1986.* Dmsion of Techni- have been resolved by the staff suice Supplement No. 5 was el Information & Document Control. September 1986. 490pp. issued.

6609180058. 37904:230.

See NUREG-0540,V08,N05 abstract. NUREG-0853 S07: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF CLINTON POWER STATION. UNIT NUREG-0750 V23 N05: NUCLEAR REGULATORY COMMISSION NO.1. Docket No. 50-461. (Illinois Power Company.et al)

• Divi-ISSUANCES FOR MAY 1986. Pa9es 465-575.

• Division of son of Boiling Water Reactor (BWR) Licensing. September Technical Information & Document Control. July 1986.113pp. 1986. 56pp. 8610030414. 37669:220.

8608130198. 37496:199. Legal issuances of the Commission, Supplement No. 7 to tt'e Safety Evaluation Report on the ap-ths Atomic Safety and Licensing Appeal Panel, the Atomic plicabon filed by filinois Power Company, Soyland Power Coop-Safety and Licensing Board Panel, the Administrative Law erative, Inc., and Western Illinois Power Cooperative, Inc., as Judge, and NRC Program Offices. applicants and owners, for a license to operate the Clinton Power Station, Unit No.1, has been prepared by the Office of NUREG-0750 V23 N06: NUCLEAR REGULATORY COMMISSION Nuclear Reactor Regulation of the U.S. Nuclear Regulatory ISSUANCES FOR JUNE 1986. Pages 577-883.

• Dtvision of Commission. The facility is located in Harp Township, DeWitt Technical Information & Document Control. August 1i86.

County, Illinois. This supplement repc3 'he status of items that 305pp. 8609180034737906:290.

have Doen resolved by the staff since Supplement No. 6 was See NUREG-0750,V23,N05 abstract.

issued.

NUREG-0781 S01: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPLICA.

AND 2. Docket Nos. 50-498 And 50-499. (Houston Lighting And TIONS TO BIOASSAY AT URAN 1UM MILLS. ALEXANDER,R.E.;

Power Company)

• Division of Pressurized Water Reactor Li- NEEL,R.B.; PUSKIN.J.S.; et al. Office of Nuclear Regulatory Re-censing -A (post 851125). September 1986. 137pp. search, Director (Pre 860720). July 1986.171pp. 8610060350.

8610070270. 38128:323 38112:047.

The Safety Evaluation Report issued in April 1986 provided The intemal dosimetry model developed in this study is useful the results of the NRC staff's review of the Houston Lighting for estimation of the urinary concentration of natural uranium And Power Company's application for licenses to operate the excreted at various times after an inhalation exposure to yellow-South Texas Project. The facility consists of two pressunzed cake or to ore dust. According to this intemal dosimetry model, wtter nuclear reactors located in Matagorda County, Texas. the time interval between collections of bioassay samples fol-Supplement No.1 updates the informabon conta,ned in the lowing an inhalation exposure to airbome yellowcake (dried at Safety Evaluation and addresses the ACRS Report issued June high or low temperature) or to ore dust should be less than 40 10,1986. days. In vivo thorax scanning techniques are shown to be suffi-ciently sensitive to confirm overexposures to natural uranium in.

NUREG-0827 S01: INTEGRATED PLANT SAFETY ASSESSMENT haled in yellowcake dried at high temperature, but not to yellow-SYSTEMATIC EVALUATION PROGRAM - LACROSSE BOIL. cake dried to low temperature or to ore dust. New experimental ING WATER REACTOR. Docket No. 50-409. (Dairyland Power Cooperatve)

• Divtsion of Boiling Water Reactor (BWR) Licens.

data, which illustrate the dependence of lung solubility on the thermal history of the inhaled yellowcake, are incorporated into ing. August 1986. 46pp. 8608250229. 37633:113.

the model. This study employs the Task Group Lung Model to The Nuclear Regulatory Commission (NRC) has published its Supplement No.1 to the integrated Plant Safety Assessment represent the deposition of uranium into the respiratory tract and its clearance into the blood. The dosimetnc model adopted Report (IPSAR) (NUREG-0827) under the scope of the System.

atic Evaluation Program (SEP), for Dairyland Power Cooperative in this study is that proposed by the Intemational Commission La Crosse Boiling Water Reactor (LACBWR) located in Genoa, for Radiological Protection in Publication 30.

Wisconsin. The SEP initiated by the NRC to review th? design of older operabon nuclear power plants to reconfirm and docu- NUREG-0887 S10: SAFETY EVALUATION REPORT RELATED ment their safety. This report documents the review completed TO THE OPERATION OF PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos.50-440 And 50-441.(Cleve-under the SEP for those issues that required refined engineer- land Electnc illuminating Company)

• Division of Boiling Water ing evaluations or the continuation of ongoing evaluations after Reactor (BWR) Licensing. September 1986. 192pp.

the Final IPSAR for the LACBWR was issued. The review has 7 3 38 21 8

provided for (1) an assessment of the significance of differ. t the SafeIY Evaluation Report (NUREG-ences between current technical positions on selected safety EP 0887) on the application filed by the Cleveland Electric illumi-issues and those that existed when LACBWR was licensed, (2) nating Company on behalf of itself and as agent for the Du-a basis for deciding on how these differences should be re- quesne Light Company, the Ohio Edison Company, the Pennsyl-solved in an integrated plant review, and (3) a documented vania Power Company, and the Toledo Edison Company (the evaLation of plant safety when the supplement to the Final Central Area Power Coordinaton Group or (CAPCO), as apph-IPSAR and the Safety Evaluaton Report for converting the li- cant's and owners for a license to operats the Perry Nuclear cense from a provisional to a full-term hcense have been Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441) issued. The IPSAR and its supplements will form part of the )

has been prepared by the Office of Nuclear Reactor Regulation I bases for considering the conversion of the license. of the U.S. Nuclear Regulatory Commission. The facility is locat-NUREG-0853 S06: SAFETY EVALUATION REPORT RELATED ed in Lake County, Ohio, approximately 35 miles northeast of TO THE OPERATION OF CLINTON POWER STATION, UNIT Cleveland, Ohio. This supplement reports the status of certain NO.1. Docket No. 50-461.(lllinois Power Company.et al)

• Divi- issues and action items that had not been resolved or complet-sion of Boiling Water Reactor (BWR) Licensing. July 1986. ed at the time of publication of the Safety Evaluation Report 280pp. 8608180061. 37546:012. and Supplement Nos.1 through 9 to the report.

Supplement No. 6 to the Safety Evaluation Report on the ap-plication filed by lilinois Power Company, Soyland Power Coop- NUREG-0396 S05: SAFETY EVALUATION REPORT RELATED erative, Inc., and Westem Illinois Power Cooperativo, Inc., as TO THE OPERATION OF SEABROOk STATION, UNITS 1 AND applicants and owners, for a license to operate the Clinton 2. DOCKET Nos. 50 443 And 50-444.(Public Service Company Power Station, Unit No.1, has been prepared by the Office of Of New Hampshire,Et AI)

• Division of Pressurized Water Reac-Nuclear Reactor Regulation of the U.S. Nuclear Regulatory tor Licensing - A (post 851125). July 1986. 303pp. 8608180067.

Commission. The facility is located in Harp Township, CeWitt 37546:292. (

Main Citations and Abstracts 3 This report is Supplement No. 5 to the Safety Evaluation aminers. These standards are not a substitute for the operator Report (NUREG-0896, March 1983) for the application filed by licensing regulations and are subject to revision or other intemal the Public Service Company of New Hampshire, et al., for li- operator examination licensing pohey changes. As appropriate, censes to operate Seabrook Station, Units 1 and 2 (Docket these standards will be revised periodically to accommodate Nos. STN 50-443 and STN 50-444). It has been prepared by comments and reflect new information or experience.

the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission and provides recent information on NUREG-1047 S03: SAFETY EVALUATION REPORT RELATED open items identified in the SER. The facility is located in Sea. TO THE OPERATION OF NINE MILE POINT NUCLEAR brook. New Hampshire. Subject to favorable resolution of the STATION, UNIT NO. 2. Docket No. 50-410. (Niagara Mohawk items discussed in this report, the staff concludes that the facili. Power Corporation)

• Division of Boiling Water Reactor (BWR) ty can be operated by the applicant without endangenng the Licensing. July 1986.167pp. 8607280207. 37273:036.

health and safety of the public. This reort supplements the Safety Evaluation Report EG@47,Febuary M80 W N appen M W Man NUREG-0936 VOS N01: NRC REGULATORY AGENDA.Ouarterty gara Mohawk Power Corporation, as applicant and co-owner, Report, January-March 1986.

• Division of Rules and Records. for a license to operate the Nine Mile Point Nuclear Station, July 1986.191pp. 8608070364. 3,7421:137. .

Unit No. 2 (Docket No. 50-410). It has been prepared by the

.he NRC Regulatory Agenda is a compilation of att rules on Office of Nuclear Reactor Regulation of the U.S> Nuclear Reg-which the NRC has proposed or is consxfenng action and afi utatory Commissson. The facility is located near Osego, New pe%ons for rulemaking which have been received by the Com- York. Supplement 1 to the Safety Evaluation Report was pub-mission and are pending disposition by the Commission. The fished in June 1985 and contained the report from the Advsory Regulatory Agenda is updated and issued each quarter. The Committee on Reactor Safeguards as well as the resolution to a Agendas for April and October are published in their entirety in ntmber of outstanding issues from the Safety Evaluation the Federal Register while a notice of availability is published in Report. Supplement 2 was published in November 1985 and the Federal Register for the January and July Agendas. contained the resolution to a number of outstanding and con-NUREG-0940 V05 NO2: ENFORCEMENT ACTIONS:SIGNIFICANT firmatory issues. Subject to favaorable resolutici of the issues ACTIONS RESOLVED.Ouarterty Progress Report. April 4une discussed in this report, the NRC staff conclueds that the facility 1986.

• Office of Inspection & Enforcement Director (Post can be operated by the applicant without endangering the 820201). August 1986. 451pp. 8609100017. 37786:001. health and safety of the public.

This compilation summanzes signficant enforcement actions that have been resolved during one quarterty period (Apil-June NUREG-1047 SO4: SAFETY EVALUATION REPORT RELATED 1986) and includes copies of leters, notices, and orders sent by TO THE OPERATION OF NINE MILE POlNT NUCLEAR the Nuclesr Reguattory Commission to licensees with respect to STATION, UNIT 2. Docket No. 50-410. (Niagara Mohawk Corpo-these enforcement actions and licencee's responses. It is ar:tci- ration)

• Division of Boiling Water Reactor (BWR) Licensing.

pated that the information in this publication will be widely dis- September 1986. 78pp. 8610030417. 37671:006.

seminated to managers and employees engaged in activities li- This report supplements the Safety Evaluation Report censed by the NRC, in the interest of promoting public health (NUREG-1047, February 1985) for the application filed by Niaga-and safety as well as common defense and secunty. ra Mohawk Power Corporation, as applicant and co-owner, for a license to operate the Nine Mile Point Nuclear Station, Unit No.

NUREG-0956: REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS Final Report. 2 (Docket No. 50-410). It has been prepared by the Office of SILBERBERG,M.; MITCHELL,J.A.; MEYER R.O.: et al. Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Nuclear Regulatory Research, Director (Pre 860720). July 1986.

Commission. The facility is located near Oswego, New York.

371pp. 8608080276. 37444:001. Supplement 1 to the Safety Evaluation Report was published in June 1985 and contained the report from the Advisory Commit-This document desenbes a major advance in the technology for calculating source terms from postulated accidents at U.S. tee on Reactor Safeguards as well as the resolution to a light. water reactors.1 he improved tochnology consists of (1) an number of outstanding, issues from the Safety Evaluation extensive data base from severe accident research programs Report. Supplement 2 was published in November 1985 and contained the resolution to a r, umber of outstanding and con-initiated following the TMI accident, (2) a set of coupled and in.

tegrated computer codes (the Source Term Code Package), firmatory issues. Supplement 3 was published in July 1986 and which models key aspects of fission product behavior under contained the resoluton to a numMr of outstanding and con-severe accident conditions, and (3) a number of detailed firmatory items, one new confirmatory item, the evaluation of mechanistic codes that bridge the gap between the data base the Engineering Assurance Program, and the evaluation of a and the Source Term Code Package. The improved unoerstand, number of exempion requests. Subject to favorable resolution of the issues discussed in this report, the NRC staff concludes ing of severe accident phenomena has also allowed an identifi-cation of significant sources of uncertainty, which should be that the facility can be operated by the applicant without endan-considered in estimating source terms. These sources of uncer. gering the health and safety of the public.

tainty are also desenbed in this document. The current technol-NUREG-1048 S06: SAFETY EVALUATION REPORT RELATED ogy provides a significant improvement ,n i evaluating source TO THE OPERATION OF HOPE CREEK GENERATING terms over that available at the time of the Reactor Safety STATION. Docket No. 50-354.(Public Service Electric And Gas Study (WASH-1400) and, because of this significance, the Nu-Company, Atlantic City Electric Company)

• Division of Boiling clear Regulatory Commission staff is recommending its use' Water Reactor (BWR) Licensing. July 1986.84pp.8608180366.

~ NUREG-1021 R03: OPERATOR LICENSING EXAMINER STAND- 37530:309.

ARDS. SZYMANSKI,T. Division of Human Factors Technology Supplement No. 6 to the Safety Evaluation Report on the ap-(post 851125). September 1986. 20pp. 8609250149. 37979:038. plication filed by Public Service Electric and Gas Company on

+ The Operator Licensing Examiner Standards provide policy its own behalf as co-owner and as agent for the other co-and guidance to NRC examiners and establish the procedures owner, the Atlantic City Electric Company, for a license to oper-and practices for examining and licensing of applicants for NRC ate Hope Creek Generating Station has been prepared by the operator licenses pursuant to Part 55 of Title 10 of the Code of Office of Nuclear Regulatiun of the U.S. Nuclear Regulatory Federal Regulations (10 CFR 55). They are intended to assist Commission. The facility is located in Lower Alloways Creek NRC examiners and facility licensees to understand the exami- Township in Salem County, New Jersey. This supplement re-nation process better and to provide for equitable and consist- ports the status of certa:n items that had not been resolved at ent administration of examinations to all applicants by NRC ex- the time of the publication of the Safety Evaluation Report. This

4 Main Citations and Abstracts supplement supports the issuance of a full-power hcense to op- NUREG-1143 S01: SAFETY EVALUATION REPORT RELATED erate the Hope Creek Generating Station. TO THE FULL-TERM OPERATING LICENSE FOR MILLSTONE NUCLEAR POWER STATION, UNIT NO.1. Docket No. 50-NUREG-1057 S02: SAFETY EVALUATION REPORT RELATED 245.(Northeast Nuclear Energy Company) SHEA.J. Division of TO THE OPERATION OF BEAVER VALLEY POWER Pressunzed Water Reactor Licensing - B (post 851125). August STATION,UN;T 2. Docket No. 50-412.(Duquesne Light 1986. 39pp. 8609170421. 37884:263.

Company.et a!)

• Division of Pressurized Water Reactor Licens-This report, prepared by the Office of Nuclear Regulation of ing - A (post 851125). August 1986. 43pp. 8609030113. the U.S. Nuclear Regulatory Commission, supplements the 37710:037.

Safety EvaluaNn Report (NUREG 1143, October 1985). It ful-Supplement No. 2 to the Safety' Evaluation Report for the ap fills a commitment to provide the Advisory Committee on Reac-plication filed by Duquesne Light Company, et al., for license to tor Safeguards report, identifies the changes that have occurred operate the Beaver Valley Power Station, Unit 2 (Docket No. since the Safety Evaluation Report was issued, and specifies 50-412), located in Beaver County, Pennsytvania has been pre- the effective lifetime for the ful-term operating heense, pared by the Office of Nuclear Reactor Regulation of the Nucte-ar Regulatory Commission. The purpose of this supplement is to NUREG-1145 V02: U.S. NUCLEAR REGULATORY COMMISSION update the Safety Evaluation of (1) additional information sub- 1985 ANNUAL REPORT.

• Office of Resource Management, Di-mitted by the applicants since Supplement No. I was issued, rector. July 1986. 244pp. 8607290229. 37309:206.

and (2) matters that the staff had under review when Supple- This report covers the major activities, events, decisions and ment No. I was issued. planning that took place dunng fiscal year 1985 within the NRC NUREG-1080 V03: LONG-RANGE RESEARCH PLAN FY 1987-FY 1931. Office of Nuclear Regulatory Research, Director (Pre NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC 860720). August 1986.115pp. 8609150249. 37835:246. ISSUE C-8, BOILING WATER REACTOR MAIN STEAM ISOLA-The Long-Range Research Plan (LRRP) was prepared by the TION VALVE LEAKAGE AND LEAKAGE TREATMENT METH-Office of Nuclear Regulatory Research to provide the Commis- ODS. RIDGELY,J.N.; WOHL.M.L Division of Boiling Water Re-

%n with a framework for planning research relevant to current actor (BWR) Licensing. August 1996. 352pp. 6610063341.

regulatory objectives or to future needs. The LRRP lays out pro- 38111:053.

grammatic apporaches for research to help resolve regulatory NUREG-1169 desenbes the key technical findings related to issues. The plan is updated annually. Generic Issue C-8, Main Steam Isolation Valve Leakage and

• 0 ** * "

NUREG-1123: KNOWLEDGE AND ABILITIES CATALOG FOR causes d esh M kakage a's assesW W h M NUCLEAR POWER PLANT OPERATOR. BOILING WATER F.E. ing Water Reactor Owners Group and the Staff, evaluates alter-ACTORS.

• Dnnsion of Human Factors Technologv (post native accident management strategies involving the use of 851125). September 1986. 250pp. 8610030409. 37671:185.

main steam isolation valves and leakage control methods. A re-This document provides the basis for the development of alistic fission product transport model was developed to assess content-valid licensing examinatons for reactor operators (ROs) and senior reactor operators (SROs). The examinatons devel- the offsite dose consequences of alternete means of treating opad using the BWR catalog will cover those topics listed unce' MSIV leakage using non-safety-grade systems that could be Title 10, Code of Federal Regulations, Part 55. The BWR cata^

available for service following a loss-of-coolant accident. Also included is a discussion of regulatory guidance in the form of log contains approximately 7,000 knowledge and ability (K/A) Regulations (10CFR50 and 10CFR100), Regulatory Guides (1.3, statements for ROs and SROs at boiling water reactors. Each 1.5, and 1.96), Standard Review Plan Sections (6.7,15.6.4, and K/A statement has been rated for its importance to the safe op-15.6.5) and standard Technical Specifications that are to be eration of the plant in a manner ensunng personnel and public considered in the regulatory implementaten phase of the reso.

health and safety. The BWR K/A catalog is organtzed into five lution of Genenc issue C-8.

major sections: Plant-wide Genenc Knowledge and Ability State-ments; Plant Systems grouped by safety function; Emergency NUREG-1171: FINAL ENVIRC,NMENTAL STATEMENT RELATED and Abnormal Plant Evolutions; Components; and Theory. The TO THE OPERATION OF SOUTH TEXAS PROJECT UNITS 1 BWR catalog represents a minor modification of the form and AND 2. Docket Nos. 50-498 And 50-499.(Houston Lighting And content of its companion document, NUREG-1122, Knowledge Power Company)

• Division of Pressurized Water Reactor Li-and Abilities Catalog for Nuclear Power Plant Operators: Pres- censing - A (post 85112d). August 1986. 385pp. 8609170027.

surized Water Reactors. 37859:073.

" * "

• 8' " "

• NUREG-1137 S03: SAFETY EVALUATION REPORT RELATED second assessment of the environmental impact associated TO THE OPERATION OF VOGTLE ELECTRIC GENERATING w consMon aM operaen of N M has RW, PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Geor-gia Power Company,et a!)

• Division of Pressurized Water Reac- "* "" * " * "9 * ****' **'

tor Licensing - A (post 851125). August 1986. 225pp-construction, issued in March 1975 pnor to issuance of con-8609180229. 37906:001 In June 1985, the staff of the Nuclear Regulato'YCommission Statement related to operation was issued in March 1986. The issued its Safety Evaluation Report (NUREG-1137) regarding projected fuel load date for Unit 1 is June 1987. The present the application of Georgia Power Company, Municipal Electric assessment is the result of the NRC staff review of tne activities Authonty of Georgia, Ogelthorpe Power Corporation, and City of associated with the proposed operation of the plant, and in-Dalton, Georgia, for a license to operate the Vogtle Electnc ciudes the staff responses to comments on the Draft Environ-Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 50-mental Statement.

425). Supplement 1 to NUREG-1137 was issued by the staff in October 1985, and Supplement 2 was issued in May 1986. The NUREG-1196: AN OVERVIEW OF ENVIRONMENTAL MATERI-facility is located in Burke County, Georgia, approximately 26 ALS DEGRADATION IN LIGHT-WATER REACTORS.

miiss south-southeast of Augusta, Georgia, and on the Savan- SHAABAN.H.L; WU,P. Division of inspection Programs (Post nah River. This third supplement to NUREG-1137 provides 850212). August 1986. 66pp. 8609120235. 37814:137.

recent information regarding resolution of some of the open and This report provides a brief overview of analyses and conclu-confirmatory items that remained unresolved at the time the sions reported in published literature regarding environmentally Safety Evaluation Report was issued. This supplement also dis- induced degradation of material in operating light-water reac-cusses some new open items. tors it is intended to provide a synopsis of subjects of concern

i l

Main Citations and Abstra:ts 5 rather than to address a licensing basis for any newly discov-ered problems related to reactor materials. The subjects dis- sis rather than providing, at this time, answers or recom,nended cussed include mater;als degradation in reactor intemals, reac- actions. Five taaks are addressed: a tabulation and discussion tor pressure boundary components, steam generators, steam of the status of all cancelled and deferred LWR units; an identi-ines, and conciensers. In each of these systems, the degra- fication of potential safety and environmental issues; an identifi-ation mechanums and the suggested reasons for each mech- cation of regulatory or policy issues and needed information to anism are reviewea. Possible remedies or methods for avoiding determine the desirability of revising certain rules and policies-such degrodation are also given. an identification of regulatory optons and decision crtieria; and an identification of decision considerations in determining staff NUREG-1198 S01: RELEASE OF UF6 FROM A RUPTUREDrequirements and organizational coordination of LWR reactiva-MODEL 48Y CYLINDER AT SEQUOYAH FUELS CORPORA- tion policy and implementation eforts.

TlON FACILITY. Lessons-Leamed Report. NRC Staff Responses To The Recommendations Made By The Lessons-LeamM NUREG-1206: ANALYSIS OF FRENCH (PALUEL) PRESSURIZED Group.

• NRC - No Detailed Affiliation Given. August 1986- WATER REACTOR DESIGN DIFFERENCFS COMPARED TO 66pp. 8609150243. 37835:178. CURRENT U.S. PWR DESIGNS.

• Dmsion of Safety Review &

The uranium hexafluoride (UF(6)) release of January 4,1986, Oversight (post 851125). June 1986. 114pp. 86C7180002.

at the Sequoyah Fuels Corporation facility has been reviewed 37076:242.

by a NRC Lessons-Leamed Group. A Model 48Y cylinder con- To understand better the regulatory approaches to reactor taining UF(6) ruptured upon being heated after it was grossly safety in foreign countries, the staff of the Ntclear Regulatory overfIled. The UF(6) released upon rupture of the cylinder re- Commission has reviewed design information on the Paluel nu-acted with airbome moisture to produce hydrofluoric acid (HF) clear power plant, one of the current standard 1300-MWe plants and uranyl fluoride (UO(2)F(2)). One individual died from expo. Operating in France. This report provides the staff's evaluation sure to airbome HF and several others were injured. There of major design differences between this standardized French were no significant immediate effects from exposure to uranyl plant and current U. S. pressunzed water reactor plants, as well fluorioe. This supplement report contains NRC's response to as insights conceming French regulatory practices. The staff the recommendations made in NUREG-1198 by the Lessons identified approximately 25 design differences, and an analysis Leamed Group. In developing a response to each of the recom- of the safeh signihcance of each of these design features is mendations, the staff considered actions that should be taken: presented, along with an assessment comparing the relative (1) for the restart of the Sequoyah Fuels Facihty; (2) to make safety beneht of each.

near erm improvement; and (3) to improve the regulatory NUREG-1212 V01: STATUS OF MAINTENANCE IN THE U.S. NU.

CLEAR POWER INDUSTRY 1985. Volume 1: Findings And Con-NUREG-1202: TECHNICAL SPECIFICATIONS FOR HOPE clusions. CWALINA,G.; GRENIER.B.; JANKOVICH,J.; et at Divi-CREEK GENERATING STATION. Docket No. 50-354. (Public sion of Human Factors Technology (post 851125). June 1986.

Sennce Electric & Gas Company Of New Jersey)

• Division of 54pp. 8607280194. 37273:203.

Boiling Water Reactor (BWR) Licensing. July 1986. 515pp. This report presents the results and conclusions derived from 8608180070.37547:241.

activities performed under Phase 1 of the NRC Maintenance and

, The Hope Creek Generating Station Technical Specifications Surveillance Program (MSP). Findings are based on trends and I

were prepared by the U.S. Nuclear Regulatory Commission to pattems derived from operational data c >mpiled by the NRC for set forth the limits, operating conditions, and other requirements the period 1980 through 1985, site surveys conducted at eight applicable to a nuclear r? actor facility as set forth in Section plants, and questionnaires administered to NRC Resident Ir, 50.36 of 10 CFR 50 for the protection of the health and safety spectore to characterize nuclear power plant maintenance pro-of the public.

grams and practices. These activities have shown that plant NUREG-1203: TECHNICAL SPECIFICATIONS FOR CLINTON maintenance programs and practices are highly variable from POWER STATION. UNIT 1. Docket No. 50-461.(lilinois Power plant.to-plant and are currently undergoing major changes.

Company,et al)

• DiVsion of Boiling Water Reactor (BWR) Li- While measured plant performance has improved overall since censing. September 1986. 500pp. 8610030421. 37673:001. 1980, the maintenance-related contribution to reportable events The Clinton Power Station, Unit No.1 Technical Specifica- and challenges to safety systems remains high and is increas-tions were prepared by the U.S. Nuclear Regulatory Commis- ing by some measures. The results of Phase 1 of the MSP con-sion to set forth limits, operating conditions, and other require- Snwd a number of proNems in Mear poww plam dnw ments applicable to a nuclear reactor facility as set forth in Sec- nance which warran' further NRC and industry attention.

tion 50.36 of 10 CFR 50 for the protection of the health and safety of the pubh,c. NUREG.1212 V02: STATUS OF MAINTENANCE IN U.S. NUCLE-AR INDUSTRY 1985. Volume 2: Descriptions Of Programs And NUREG 1205: REACTIVATION OF NUCLEAR POWER PLANT Practices. CWALINA,G.; GRENIER B. JANKOVICH,J.; et al. Di-CONSTRUCTION PROJECTS. Plant Status, Policy Issues,And vision of Human Factors Technology (post 851125). June 1986.

Regulatory Options. SPANGLER,M.B. Office of Nuclear Reactor 235pp. 8607290499. 37276.001.

Regulation. Director (post 851125). July 1986. 140pp. This report documents a review of the status of maintenance 8608120699. 37467:178. programs and practices in the U.S. commercial nuclear power Prior to the TMI-2 acident on March 28,1979, four nuclear industry. The purpose of this review is to establish a baseline plant units that had previously been issued a constuction permit reference for evaluating the effectiveness of future industry ac-were cancelled, principalty because of reduced projections of tivities in maintenance. Two methods were used to collect pro-regional power demand. Since that time, an additional 31 units grammatic data. First, a Maintenance Review Protocol was used with cps have been cancelled and eight units deferred. On De- during site visits to eight selected power plants tu collect in-cember 23,1985 one of the deferred units (Umerick-2) was re- depth maintenance program information. Second, a Mainte-activated and construction resumed. The primary objective of nance Ouestionnaire was filled out by NRC Resident inspectors this policy stMy is to identify the principal issues requiring regarding maintenance programs at their plant. The protocol office-level censideration in the event of reactivation of the con- and questionnaire contained items regarding five broad catego-struction of one or more of the nuclear power plants falling into ries of maintenance: (1) organization and administration, (2) fa-two categories: (1) LWR units issued a construction permit cilities and equipment, (3) procedures (4) personnel, and (5) whose construction has been cancelled, and (2) LWR units work control. The study found that there is a wide variability in whose constnJction has been deferred. The study scope is limit. industry maintenance programs. However, the industry is cur-ed to identifying regulatory issues or questions deserving anaty- rently undergoing changes in maintenance practices, and adher-

- 4 6 Main Citations and Abstracts NUREG format to enable large-scafe product >on and distnbution ence to INPO and NUMARC guidance in maintenance may for informahon purposes.

bring about a more systematically developed program practices.

NUREG/CP-0080 V01: PROCEEDINGS OF THE TOPICAL MEET-NUREG-1213: PLANS AND SCHEDULES FOR IMPLEMENTA- ING ON REACTOR PHYSICS AND SAFETY Sessions 110.

TION OF U.S. NUCLEAR REGULATORY COMMISSION RE- FEINER.F. Amencan Nuclear Society. August 1986. 656pp.

SPONSIBILITIES UNDER THE LOW-LEVEL RADIOACTIVE 8609110368. 37804:021.

WASTE POLICY AMENDMENTS ACT OF 1985 (PL 99-240). The following subjects are discussed. This volume includes DUNKELMAN.M.M.; VEARNEY,M.S.; MACDOUGALL R.D. Divi-papers presented at the ANS Topical Meting on Advances in sion of Waste Management July 1986. 85pp. 8608120696. Reactor Physics and Safety (September 1719,1986 at Sarato-37467:105. ga Springs, NY). The papers relate to the following areas: chal-The purpose of this document is to make available to the lenges in reactor physics, pressure vessel embnttlement, safety states and other interested parties, the plans and schedules for limits and core instrumentation, the establishment of the con-the U.S. Nuclear Regulatory Commission's (NRC's) implementa-tainment precedent (including the Manhattan Project exenence).

tion of its responsibilities under the Low-Level Radioactive safety aspects of core design (related to both fast and thermal Waste Policy Amendments Act of 1985 (P.L 99-240) reaction), reactor physics methods, improvement and validation (LLRWPAA). This document identifies the provisions of the LLRWPAA which affect the programs of the NRC, identfies of plant simulaton codes, validation of reactor physics methods what the NRC must do to fulfill each of its requirements under and data, and physics ar,d safety of advanced reactor concepts.

the LLRWPAA, and establishes schedules for carrying out these NUREG/CP-0080 V02: PROCEEDINGS OF THE TOPICAL MEET-requirements. The plans and schedules are current as of June ING ON REACTOR PHYSICS AND SAFETY. Sessions 1116.

1986. FEINER,F. American Nuclear Socie'y. August 1986. 567pp.

8609120240. 37816:213.

NUREG-1215: COMPILATION OF CONTRACT RESEARCH FOR The following subjects are discussed. This volume includes THE CHEMICAL ENGINEERING BRANCH. DIVISION OF ENGI-NEERING TECHNOLOGY. Annual Report For FY 1985.

• Divi- papers presented at the ANS Topical Report Meeting on Ad-vances in Reactor Physics and Safety (September 17-19, 1986, soon of Engineenng Technology (Pre 86C720). July 1986.113pp.

8607240273.37244.001.

at Saratoga Springs, N.Y.). The papers relate to the following Thrs report presents summa"es of the research work per- areas: point and space-time core models for transient analysis, formed dunng Fiscal Year 1985 by laboratones and organiza- modeling of degraded cores, physics and safefy of research re-tions under contracts administered by the NRC's Chemical En- actors, reactor physics methods, nuclear plant analysers, data gineenng Branch Office of Nuclear Regulatory Research. Each bases and code systems, reactor noise, and aaavances In reac-contractor will complete detailed reports of their work; however tor physics and data.

we believe it is useful to have a summary of each contractor's NUREG/CP-0083: ANS TOPICAL MEETING ON RADIOLOGICAL effort for the year combined into one volume.

ACCIDENTS - PERSPECTIVES AND EMERGENCY PLANNING NUREG-1216: SAFETY EVALUATION REPORT RELATED TO PROGRAM AND ABSTRACTS. POWELL,R. Oak Ridge National THE OPERABILITY AND RELIABILITY OF EMERGENCY Laboratory. September 1986. 27pp. 8609170039. 37859:046.

DIESEL GENERATORS MANUFACTURED BY TRANSAMERI- This report contains summaries of papers to be presented at CA DELAVALINC. BERLINGER C.; MURPHY,E. Office of Nu- the Topical Meeting on Radichgical Accidents - Perspectives clear Reactor Regulation, Director (post 851125). August 1986. and Emergency Planning held at the Hohdav inn, Bethesda, 44pp. 8609100233. 37785:171. Maryland, September 15-17,1996. TN meting will review what A broad pattem of desgn, manufactunng, and quakty-related has been leamed from the past accidents involving nuclear ma-deficiencies with emergency diesel generators manufactured by terial in order to foster effectue emergency planning. The sum-Transamenca Delaval, Inc. (TDI) became evident following a maries have been compiled in one report to provide a basis for crankshaft failure at the Shoreham Power Station in August meaningful discussion and information exchange during the 1983. In response to these problems, U.S. nuclear utility owners course of the meeting.

of these engines formed an Owners Group to address oper-ational and regulatory issues relatve to diesel generator sets NUREG/CR-2000 V05 N6: LICENSEE EVENT REPORT (LER)

COMPILATION:For Month Of June 1986.

• Oak Ridge National used for standby emergency power. The Owners Group per, Laboratory. July 1986.149pp. 8608070370. ORNL/NSIC-200.

formed extenswe design reviews of all key engine components and developed recommendations to be implemented by the in- 37420:348.

dividual owners conceming needed component replacements This monthly report contains Licensee Event Report (LER) and modifications, component inspections to validate the "as, operational information that was processed into the LER data file of the Nuclear Safety Information Center (NSIC) du ing the manufactured" and "as-assembled" quality of key engine com-one month period identified on the cover of _the document. The ponents, engine testing, and an enhanced maintenance and LERs, from which this information is derived, are submitted to surveillance program.

the Nuclear regulatory Commission (NRC) by nuclear power NUREG-1220: TRAINING REVIEW CRITERIA AND PROCE- plant licensees in accordance with federal regulations. Proce-DURES.

Dwision of Human Factors Technology (post dures for LER reporting for revisions to those events occumng 851125). July 1986.123pp. 8608080057. 37447:197. prior to 1984 are described in NRC Regulatory Guide 1.16 and This report provides a set of training review cnteria and pro- NUREG-1061, Instructions for Preparation of Data Entry Sheets cedures which constitute a systematic means of implementing for Licensee Event Reports. For those events occumng on and two monitonng functions identified in the " Commission Policy after January 1,1984, LERs are being submitted in accordance Statement on Training and Quahfications of Nuclear Power with the revised rule contained in Title 10 Part 50.73 of the Plant Personnel" (March 20,1984,50 FR 11147). These func. Code of Federal Register (Vol. 48, No.144) on July 26,1983.

tions are: 1. Continuing evaluation of industry-wide training and NUREG-1022, Licensee Event Report System - Description of quahfication program effectiveness, and 2. Monitoring plant and Systems and Guidelinas for Reporting, provides supporting guid-industry trends and events involving personnel errors. The pro- ance and information on the revised LER rule. The LER summa-cedures are organized around the five essential elements of ries in this report are arranged alphabetically by facility name performance-based training articulated in the Policy Statement. and then chronologically by event date for each facihty. Compo-The package was designed for use by NRC personnel engaged nent, system, keyword, and component vendor indexes follow in the review of performance-based training programs in nuclear the summaries. Vendors are those identified by the utihty when power plants. It has been published in a modified version in the LER form is initiated; the keywords for the component,

Main Citations and Abstracts 7 system, and general keyword indexes are assigned by the com- mission. The 1983 release data are summanzed in tabular form.

puter using correlabon tables from the Sequence Coding and Data covering specific r2dionuclides are summarized.

NUREG/CR-3145 V04: GEOPHYSICAL INVESTIGATIONS OF NUREG/CR-2000 V05 N7: LICENSEE EVENT REPORT (LER) WESTERN OHIO-INDIANA REGION. Annual Report (October COMPILATION.For Month Of July 1986.

• Oak Ridge National 1984 - September 1985). POLLACK,H.N.; CHRISTENSEN.D.;

Laboratory. August 1986.169pp. 8609150245. ORNL/NSIC- WELC.J. Michigan, Univ. of Ann Arbor, MI. June 1986. 43pp.

200. 37835:009. See NUREG/Cr-2000,V05.N06 abstract. 8608290189.37695:182.

Earthquake activity in the Western Ohio - Indiana region has NUREG/CR-2000 V05 N8: LICENSEE EVENT REPORT (LER) been monitored with a precision seismograph network consist-COMPILATION:For Month Of August 1986.

• Oak Ridge Nation- ing of mne stations located in west-central Ohio and four sta-al Laboratory. September 1986.173pp. 8609260232. ORNL/ tions sited in Indiana. The nine Ohio static , have been upgrad-NSIC-200. 37996:057. ed by replacing geophe,,es and VCO/ amplifiers with new, and See NUREG/CR-2000,V05,N06 abstract. more reliable components. Seven local and near-regional earth-NUREG/CR-2331 VOS N4: SAFETY RESEARCH PROGRAMS quakes have been recorded dunng this report period, ranging in SPONSORED BY OFFICE OF NUCLEAR REGULATORY magnitude from 1.4 to 3.0 m(big). Of the seven events, four oc-RESEARCH.Ouarterty Progress Report, October-December cuned in westa N, and mree wee located wen outsW N 1985. WEISS.A.J. Brookhaven National Laboratory. June 1986. anay, one near Ngo, Ms, one b Mwn Kentuch, aM 101pp.8607240304. BNL-NUREG-51454. 37243:262. me westwn %sWania. N event e mW near This progress report will desenbe current activibes and techni- Chicago on 9 September 1985 was felt; however, only minor cal progress in the programs at Brookhaven National Laboratory damage was reported. Teleseismic P-wave arrival and residual sponsored by the Division of Accident Evaluation, Division of tables have been updated to include newly acquired data. The Erigineering Technology, and Division of Risk Analysis & Oper- results are substantia!!y the same as in previous years. Prelimi-ations of the U.S. Nuclear Regulatory Commission, Office nf Nu- nary resuns from refmchm expennets in me surnmw of W clear Regulatory Research. The projects reported are the fol- indicate apparent veiocities of 6.67 km/sec for refracted P arriv-lowing: High Temperature Reactor Research, SSC Modeling for als from the basement,8.27 km/sec for Pn arrivals, and 3.57 Low Flow Conditions Thermal-Hydraulic Reactor Safety Experi- km/sec for Lg arrivals.

ments, Thermal Hydrauhcs of Core / Concrete interactions, Plant Analyzer, Code Assessment and Application Code Maintenance NUREG/CR42$2 V04: COBRA-NC:A THERMAL HYDRAULICS CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR (RAMONA-3B), MELCOR Venfication and Benchmarking. Pool COMPONENTS.Vol. 4: Users' Manual For Containment Analysis.

Version of the SSC Code, Source Term Code Package Venfica- WHEELER.C.L; THURGOOD,M.J.; GUIDOTTI,T.E.; et al. Bat-bon and Benchmarking, Uncertainty Analysis of the Source Term; Stress Corrosion Cracking of PWR Steam Generator telle Memorial Institute, Pacific Northwest Laboratories. August 1986. 261pp. 8609120171. PNL-5515. 37818.060. COBRA-NC Tubing, Prubability Based Load Combinations for Design of Cat-egory i Structures, Soil-Structure Interaction Evaluations, Seis- is a digital computer program written in FORTRAN IV that simu-lates the response of nuclear reactor components and systems mic Research Coordination and Technology Transfer - Transfer and Use of the SMACS Code at BNL, Combinational Proce- to thermal-hydraulic trarpants. The code solves the multicom-ponent, compressible, three-dimensional, two-fluid, three-fluid dures for Pipsng Response Spectra Analyses, Validation of Seis*

mic Calculational Methods, Identification of Age Related Failure equations for two-phase flow. The three velocity fields are the Modes; Application of HRA/PRA Results to Support Resolution vapor / gas field, the continuous liquid field, and the liquid drop of Genenc Safety issues involving Human Performance, Protec-field. This volume of the manual provides the user with an ex-planation of the input required to simulate the response of multi-tive Acton Decisionmaking, Rebaselining of Risk for Zion, and Operational Safety Reliability Research, cernpartment nuclear containment systems to postualted loss-of-coolant accidents that result in the release of steam, water, NUREG/CR-2800 SO4: GUIDELINES FOR NUCLEAR POWER and/or noncondensable gases into the containment.

PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DE-VELOPMENT, TABATABAl,A.S. BatteIIe Memorial Insttute, P&- NUREG/CR-3320 V01: LWR PRESSURE VESSEL SURVEIL-cdic Northwest Laboratones. July 1986. 345pp. 8608070339. LANCE DOSIMETRY IMPROVEMENT PROGRAM. PSF Expe.1-PNL-4297. 37426:115. ments Summary And Blind Test Results. MCELROY,W.N. Han-This is the fifth in a series of reports to document the use of ford Engineering Development Laboratory. July 1986. 521pp.

a methodology developed by the Pacific Northwest Laboratory 8608070328. HEDL-TME 86-8. 37424:145.

to calculate for pooritization purposes, the risk, dose and cost To test the current fracture toughness and embnttlement pre-impacts of implementing resolutions to reactor safety issues diction methodologies,a Metallurgical Blind Test was initiated at (NUREG/CR-2800, Andrews et al.1983). This report contains the ORR/ pef. The PSF Expsriment is the only expenment to results of issue-specific analyses for 23 issues. Each issue was date tint has realistically simulated the damage that occurs considered within the constraints of available information as of within the PV of operating commercial power plants. This exper-winter 1986, and two staff-weeks of labor. The results are refer- tment was charactenzed by a two-year irradiation with extensive enced, as one consideration in setting prionties for reactor dosimetry, very well-controlled temperatures and benchmarked safety issues, in NUREG-0933, A Prioritization of Generic Safety dosimetry, for which HEDL served as the point of reference.

Issues. Metallurgical test data were obtained by Material Engineenng Associates, Inc (MEA) and HEDL for representative PV steels NUREG/CR-2907 V04: RADIOACTIVE MATERIALS RELEASED and by Fracture Control Corporation (FCC) and Westinghouse FROM NUCLEAR POWER PLANTS. Annual Report 1983. Research Division (W-RD) for representative support structure TICHLER,J.; NORDEN,K. Brookhaven National Laboratory. steefs. For the Blind Test, the experimental PV steel results August 1986. 240pp. 8609030139. BNL-NUREG-51581. were withheld from the participants; only the type of information 37709:157. normally contained in surveillance reports was given. The goal Releases of radioactive materials in airborne and liquid ef- was to predict from this limited information the metallurgical test fluents from commercial light water reactors during 1983 have results in the SSC-2 and PV wall capsules. Participants were been compiled and reported. Data on solid waste shipments as askcd to predict both the damage exposure parameter values well as selected operating information have been included. This and the measured metallurgical test results in the Simulated i

report supplements earlier annual reports issued by the former Surveillance Capsule and simulated PV capsules at the surface, Atomic Energy Commission and the Nuclear Regulatory Com- 1/4 thickness, and 1/2 thickness locations. Results and conclu-

8 Main Citations and Abstracts sions of the companson of physics-dosimetry and metallurgical and soil-structure interaction effects; (2) observational data on predictions with measurements are presented and discussed. spatial vanations of earthquake ground motion; (3) soil-structure The very low neutron exposure (< <10(17)n/cm(2), E>1 MeV) interaction effects on structural responses; and (4) summary and observed small property changes prevented the use of the based on Tasks I and li studies. This report presents the results support structure steel results for the Blind Test. These latter of the fourth part of Task II.

results do, however, provide important bounding information for low flux and low exposure irradiated support structure steels. NUREG/CR-3858: TRAC-PF1/ MOD 1:AN ADVANCED BEST-ES-TIMATE COMPUTER PROGRAM FOR PRESSURIZED WATER NUREG/CR-3585 V02: DE MINIMIS WASTE IMPACTS ANALY. REACTOR THERMAL-HYDRAULIC ANALYSIS.

• Los Alamos SIS METHODOLOGY. Volume 2. Impacts-BRC Users Guide And Scientific Laboratory. July 1986. 848pp. 8609180039. LA-10157-Methodology For Radioactive Wastes Below Regulatory MS. 37907:241.

Concern.This is An NRC Staff Report Published As Volume 2 The Los Alamos National Laboratory is developing the Tran-Of Contractor Report. FORSTROM J.M.; GOODE D.J. Division sient Reactor Analysis Code (TRAC) to provide advanced best-of Waste Management. July 1986. 153pp. 8608200415. estimate predic* ions of postulated accidents in light water reac-37597:235. tors. The TRAC-PFI/ MOD 1 program provides this capability for This report desenbes the methodology and computer program pressurized water reactors and for many thermal-hydraulic test used by ilRC to evaluate radiological impacts associated with facilities. The code features either a one- or a three-dimensional petitions to have specific shghtly contaminated rad:oactive heatment of the pressure vessel and its associated internals, a waste streams designated as "below regulatory concern." two-fluid nonequihbrium hydrodynamics model with a noncon-These wastes could be treated and disposed of at facilities densable gas field and solute tracking, flow-regime-dependent which are not licensed for low-level radioactive waste manage- constitutive equation treatment, optional reflood tracking capa-ment. The IMPACTS-BRC computer program is implemented on bility for bottom-flood and falling-filrn quench fronts, and consist-IBM-PC microcomputers using the FORTRAN programming lan- ent treatment of entire accident sequences including the gen-guage Radiological impacts (deses) are estmated for several eration of consistent initial conditions. In additic,n to the compo-patlways including direct gamma radiation exposure, worker in- nents contained in previous TRAC versions, TRAC-PF1/ MOD 1 halaton and exposure, offsite atmosphenc and water releases, includes an improved steam-generator model, a turbine compo-and intruder exposures. Annual impacts are calculated for the nent, and a control system together with modified constitutive maximum individual, entical groups, and general population. The relations to model the balance of plant on the secondary code treatment and disposal options include onsite incineration, incin- and to extend the applications to transients without loss of cool-eration at municipal and hazardous waste facilities, and disposal ant. This report describes the thermal-hydraulic models and the ,

at san,tary landfills and hazardous waste landfills. Modifications numerical solution methods used in the code. Detailed program-to the program (Volume 1) are pnmanly for merocomputer com- ming and Jser information are provided.

patability and to provide information needed to evaluate the pc-NUREG/CR-3925: SWIFT ll SELF-TEACHING titions. Default environmental and facility parameters are devel-ored representing conservative assumptions about site selec. CURRICULUM. Illustrative Problems For Sandia Waste-Isolation Flow And Transport Model For Fractured Media. REEVES.M.;

tion and operational procedures. In particular, the parameters of the groundwater pathway model are moddied to represent more WARD,D.S.; DAVIS,P.A.; et al. Sandia National Laboratories.

codrvatae assumptions than the original model (Volume 1). August 1986.107pp. 8609050042. SAND 84-158J. 37735:004.

Several documents have been wntten desenbing SWIFT 11 NUREG/CR-3805 V04: ENGINEERING CHARACTERIZATION OF the most current version of the SWIFT (Sandia Waste Isolation GROUND MOTION. Task ll: Soil-Structure Interaction Effects On Flow and Transport) Model. Reeves et al. [1986a], desenbes Structural Response. LUCO.J.E.; WONG,H.L Structural & Earth- the theory and implementation, and Reeves et al. [1986b], de-quake Engineering Consultants. CHANG,C.Y.: et al. Woodward- scribes the required input of data and parameters. Ward et al.

Clyde Consultants, Inc. August 1986. 427pp. 8609040428. [1984a], and [1984b] desenbe the comparison of the results 37733:272. from the SWIFT code with field data and other existing codes.

A detailed parametric study of the soil-structure inter 1ction ef- This document is devoted to assisting the analyst who desires fects on the seismic response of nuclear power plants is pre- to use the SWIFT ll code. The analyst is referred to the User's sented. The study includes evaluation of the effects introduced Manual for SWIFT ll (Reeves et al. [1986b]) for detailed data l by the soil charactenstics, embedment of the foundation, kine- input instructions. Eight examples are presented to illustrate the )

matic interaction, nonvertcaily incident waves, soil nonlineanty use of SWIFT 11. The implementation of the numerical simulation and depth to bedrock. The parametric study is based on analy- of the physical problem is desenbed for each example. For )

sis of the seismic response of a reactor building and an auxiliary each problem, a listing of the input data and a microfiche listing building model. A total of nine soil profiles, two embedment of the output are prouded. ,

I depths, four three-component free-field ground motions and '

three combinations of apparent honzontal velocities for the exci- NUREG/CR-3967: AN ANALYSIS OF EXCURSIONS AT SELECT-tat:on were considered in the course of the study. ED IN SITU URAN!UM MINES IN WYOMING AND TEXAS.

STAUB,W.P.; HINKLE N.E. Oak Ridge National Laboratory.

NUREG/CR-3805 V05: ENGINEERING CHARACTERIZATION OF WILLIAMS.R.E.; et al. Idaho, Univ. of, Moscow, ID. July 1986.

GROUND MOTION. Task 11: Summary Report. POWER.M.S.; 294pp.8607240286. ORNL/TM-9956. 37245:127.

CHANG,C.Y.: IDRISS,l.M.; et al. Woodward-Clyde Consultants, See NUREG/CR-4125,V01,R01 abstract.

Inc. August 1986.160pp. 8609090334. 37759:243.

This report presents the results of part of a two-task study on NUREG/CR-3974: GEOMEMBRANE SELECTION CRITERIA FOR the engineenng characterization cf earthquake ground motion URANIUM MILL TAILINGS PONDS. MITCHELL.D.H.;

for a nuclear power plant design. Task I of the study, which is CUELLO,R. Battelle Memorial Institute, Pacifc Northwest Lab-presented in NUREG/CR-3805, Vol.1, developed a basis for se- oratories. September 1986. 64pp. 8609160436. PNL-5224.

lecting design response spectra taking into account the charac- 37848:001.

t:ristics of free-field ground motion found to be significant in The selection enteria, partcularly those involving chemical causing structural damage. Task 11 incorporates additional con- compatibility, of geomembranes to be used in ponds at uranium siderations of effects of spatial variations of ground medon and mill operations are discussed in this report. The principal func-soil-structure interaction on foundation motions and structural tional critena which a geomembrane must meet for this applica-response. The resultst of Task 11 are presented in four parts: (1) tion are 1) a specified service life and 2) low permeability.

effects of ground rtotion characteristics on structural response Chemical compatibility with the waste is essential in meeting of a typcal PWR reactor building with localized nonlinearities these functional cnteria. In two different types of aging tests i

1 1

Main Citations and Abstracts 9 using simulated acidic uranium mill waste, degradation of chemi- brittle, corroded positive bus material; and (2) excessivo sulpha-cal and physical properties were examined in geomembranes of tion of positive plate active materials causing hardening and ex-high density polyethylene, polyvinyl chloride, and chlorosulfonat- pansson of positive plates.

ed polyethylene. Compatibility tests according to the National Sanitation Foundation procedures are recommended to ascer. NUREG/CR-4102: AIR CURRENTS DRIVEN BY SPRAYS IN RE-tain the stabihty of certain physical properties of the proposed ACTOR CONTAINMENT BUILDINGS. MARX,K.D. Sandia Na-geomembrane. It is likely that many geomembranes have ac. tional Laboratories. May 1986. 71pp. 8607240299. SAND 84-ceptable chemical stability for typical uranium mill applications, 8256. 37248:238 therefore, additional factors in the selection processes will in- The flow of air in nuclear reactor containment buildings due to clude seaming charactenstics, mechanical properties, site char. the introduction of water sprays is investigated numencally and actenstics, and costs. theoretically. The containment geometries studied are reactors with ice condensers. The numencal approach utilizes a comput-NUREG/CR-4013: LADTAP 11 - TECHNICAL REFERENCE AND USER GUIDE. STRENGE,D.L; PELOQUIN,R.A.; WHELAN,G. er code which couples the Navier-Stokes equations with the equations of motion of water droplets in two dimensions. A the-Badelle Memorial Institute, Pacific Northwest Laboratories. Apri' oretical desenption of some aspects of the flow is derived. For 1986. 325pp. 8608070335. PNL-5270. 37429.085.

typical water spray fluxes it is found that peak flow velocities of The Nuclear Regulatory Commission's LADTAP ti computer program performs environmental dose analyses for releases of 12-14 m/s are possible in an empty containment. The depend-ence of the flow velocities on spray flux, spray header location, radioactive effluents from nuclear power plants apto surface droplet size, spray injection characteristics, air fans, and the tur-waters. The analyses estimate radiation dose to individuals, bulence modelis described. Simulations more representative of population groups, and biota from ingestion and extemal expo-sure pathways. The estimated doses provide information for Na- complete reactors have been carried out by including the ice condenser and steam generator enclosures in the model. In tional Environmental Policy Act evaluations and for determining that case, it is found that the peak flow velocities are sensitive complence with Appendix ! of 10 CFR 50. This report describes the mathematical models used in the LADTAP 11 computer pro-to the relative locations of the steam generator wall and the spray headers. For the case of two headers centered over the gram, instructs the user in preparing input to the program, and supplies detailed information on program structure and param- wall, the peak velocity is about the same as for an empty con-eters used to modify the program

• tainment. On the other hand, for six headers spread over the steam generators, the peak is reduced to 9-10 m/s. However, NUREG/CR-4082 V04: DEGRADED PIPING PROGRAM - PHASE the effect of intemal structures is to lower the flow velocities ll. Semiannual Report, October 1985 - March 1986. over large regions in all cases. Implications for igniter perform-WILKOWSKI G.M.; AHMAD.J.; BARNES C.R.; et al. Battelle Me- ance are briefly discussed.

monal Institute, Columbus Laboratories. September 1986.

350pp. 8610030416. BMI-2120. 37674:195. NUREG/CR-4125 V01 R1: GUIDELINES AND WORKBOOK FOR The efforts in this report are broken into nine work packages ASSESSMENT OF ORGANIZATION AND ADMINISTRATION related to pipe-fracture research efforts and three work pack. OF UTILITIES SEEKING OPERATING LICENSE FOR A NU-ages that are supporting research efforts. The pipe-fracture ef- CLEAR POWER PLANT. Volume 1: Guidelines For Utility Organi-forts involve only circumferential crack orientations. Forty-two zation And Administration Plan. THURBER,J.A.; OLSON,J.,

pipe expenments have been conducted to date, with all but two OSBORN,R.N.; et al. Battelle Human Affairs Research Centers.

at 550 F (288 C). Approximately 42 additional pipe experiments September 1986. 41pp. 8610030400. PNL-5374. 37680:202.

from other programs were also analyzed. In the analysis effort. This report is a partial responso to the requirements of item a screening enterion was developed to show when the net-sec. l.B.1.1 of the "NRC Action Plan Developed as a Result of the tiordcollapse analysis is valid. This shows that even wrought TMI-2 Accident," NUREG-0660, and is designed to serve as a stainless steel can fail at less than net-sectiorH:ollapse loads if basis for rnlacing the earlier NUREG-0731, " Guidelines for the pipe diameter is sufficiently large. Numerous predictive J-es. L'tility Management Structure and Technical Resources." The timation schemes have been evaluated and ,nodfied. A finite Guidelines are intended to provide guidance to the user in pre-length surface cracked pipe estimation scheme has also been paring a written plan for a proposed nuclear organization and developed. Finite e lement analyses of specimens with welds administration. The purpose of the Workbook is to guide the suggest that the s.:e of the weld relative to the specimen of NRC reviewer through a systematic review and assessment of a structure size can affect the deformation J values. Supporting proposed organization and administration. It is the NRC's inten-research efforts involve geometry effects on J-R curves, as well tion to incorporate these Guidelines and Workbook into a future as characterizing the material properties 'or each pipe tested. revision of the Standard Review Plan (SRP), NUREG-0800.

However, at this time the report is being published so that the NUREG/CR-4P99: AGE-RELATED DEGRADATION OF NATU- material may be used on a voluntary basis by industry to sys-RALLY-AGED CLASS 1E BATTERY CELLS. BONZON,LL tematiccity prepare or evaluate their organization or administra-Sandia National Laboratories. April 1986. 61pp. 8609090481. tion plans. Use of the report by the NRC would not occur until SAND 8&2632. 37765:056.

after it has been incorporated in the SRP.

The seismic-fragility response of naturalty-aged nuclear sta-tion safety-related batteries is of interest for two reasons: (1) to NUREG/CR-4125 V02 R1: GUIDELINES AND WORKBOOK FOR determine actual failure modes and thresholds and (2) to deter- ASSESSMENT OF ORGANIZATION AND ADMINISTRATION mine the validity of using the electrical capacrty of individual OF UTILITIES SEEKING OPERATING LICENSE FOR A NU-cells as an indicator of the potential survivability of a uattery CLEAR POWER PLANT. Volume 2: Workbook For Assessment given a seismic event. Prior reports in this series discusseo the Of Organization And Management. THURBER,J.A.; OLSON J.;

seismic-fragility tests and results for three speofic naturally- OSBORN,R.N.; et al. Battelle Human Aftairs Research Centers.

aged cell types: 12-year old NCX-2250,10-year old LCU-13 September 1986. 92pp. 8610030402. PNL-5374. 37680:115.

and 10-year old FHC-19. This report focuses on the postseismic This report is a partial response a the requirements of item evaluations of the internals of the cells. Two distinct failure iB.1.1 of the "NRC Action Plan Jeveloped as a Result of the modes were observed (but not in all cells or typesh complete TMI-2 AccP ' NUREG-0660, ..nd is designed to serve as a physical separation at the terminal post-plate hanger interfs.:e basis for re, g the earfier NUMG-0731, " Guidelines for and cell capacity of less than 80% of nominal following tho 3- Utility Manage.nent Structure and Technical Resource." The hour postseistrac discharge capacity test. The synificant age-re- Guidelines are intended to provide guidance to the user in pre-lated effects in terms of seismic survivability, in decreasing paring a written plan for a proposed nuclear organization and order of importance, were dete' mined to be: (1) formation of administration. The purpose of the Workbook is to guide the

l l

10 Main Citations and Abstracts NRC reviewer through a systematic review and assessment of a canisters in a repository in basalt, and (2) calculation of the sol-proposed organization and administraton. It is the NRC's inten- ubility and speciation of radionuclides for representative ground tion to incorporate these Guidelines and Workbook into a future waters from the Yucca Mountain site in Nevada.

remsion of the Standard Review Plan (SRP), NUREG-0800.

However, at this time the report is being published so that the NUREG/CR-4255 V03 N1: AEROSAL RELEASE AND TRANS-matenal may be used on a voluntary basis by industry to sys- PORT PROGRAM SEMIANNUAL PROGRESS REPORT FOR tematically prepare or evaluate their organizaton or administra- OCTOBER 1985 - MARCH 1986. ADAMS,R.E.; TOBIAS,M.L tion plans. Use of the report by the NRC would not occur until Oak Ridge National Laboratory. June 1986.46pp.8609110264.

after it has been incorporated in the SRP. ORNL/TM-9632 V3. 37802:183.

NUREG/CR-4162: SURVEY OF STATISTICAL AND SAMPLING As part of a major reassessment of the release of rad'oachve NEEDS FOR ENV!RONMENTAL MONITORING OF COMMER. matenals to the environment (source terms) in severe reactor CIAL LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILI- accidents, a group of state-of-the-art computer codes was uti-TIES. EBERHARDT,LL; THOMAS.J.M. Battelle Memonal insti- lized to perform extensive analyses. A major prodixt of this tute, Pacific Northwest Laboratories. Juty 30, 1986. 99pp. source term reassessment effort was a demonstrated methodol-8608010056. PNL-4804. 37342:004. ogy for analyzing specific accident situations to provide source This project was designed to develop guidance for imple- term predictons. The computer codes forming this methodology menting 10 CFR Part 61 and to determine the overall needs for have been named the Source Term Code Package (STCP) and sampling and statistical work in charactenzing, surveying, moni- is the subject of this user's guide. The guide is intended to pro-tonng, and closing commercial low-level waste sites. When vide an understanding of the STCP structure and to facilitate cost-effectiveness and statistical reliability are of prime impor- STCP use. The STCP was prepared for operabon on a CDC tance, then ocuble sampling, compositing, and stratfcation system but is wntten in FORTRAN-77 to permit transportability.

(with optimal allocation) are identified as key issues. If the pnn- In the current version (Mod 1) of the STCP, the various calcula-cipal concem is avoiding questionable statistical practce, then tional elements fall into four major categones represented by the applicability of knging (for assessing spatial pattern), meth- the codes MARCH 3, TRAP-MELT 3, VANESA, and NAUA/

ods for routne monitonng, and use of standard textbook formu- SPARC>lCEDF, The MARCH 3 code is a combination of the iae in reportng monitonng results should be reevaluated. Other MARCH 2,CORSOR-M, and CORCON Mod 2 codes. The TRAP.

important issues identified include sampiing for estimatng MELT 3 code is a combination of the TRAP-MELT 2.0 and model parameters and the use of data from left-censored (less MERGE codes.

than detectable limits) distnbutons.

NUREG/CR-4219 V03 N1: HEAVY-SECTION STEEL TECHNOL- NUREG/CR-4257 V02. INSPECTION, SURVEILLANCE AND MON-OGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR ITORING OF ELECTRICAL EQUIPMENT IN NUCLEAR POWER OCTOBER 1985 MARCH 1986. PUGH.C.E. Oak Ridge Nabon- PLANTS. Volume 2: Pressure Transmitters. TOMAN.G.T. ARVIN/

al Laboratory. June 1986. 277pp. 8607240278. ORNL/TM- CALSPAN Advanced Technology Center.

• Oak Ridge National 9593/V3. 37244:110- Laboratory. July 1986. 51pp. 8609290015. 38013:130.

The Heavy-Section Steel Technology (HSST) Program is an This report was prepared under the U.S. Nuclear Regulatory engineenng research activity conducted by the Oak Ridge Na- Commission's Nuclear Plant Aging Research Program. It de-tional Laboratory for the Nuclear Regulatory Commission. The scnbes the types of transmitters used in nulcear power plants.

program compnses studies related to all areas of the technolo- Stresses that cause detenoration and age-related effects in gy of matenals fabncated into thick-secton pnmary-coolant con- ressure transmitters are evaluated, and means of detectng tainment systems of light-water-coolant nuclear power reactors.

and evaluating detenoration are desenbed. Recommendations The investigabon focuses on the behavior and structural integri-for monitoring deterioration are given. Cntena for determing the ty of steel pressure vessels containing cracklike flaws. Carrent capability of pressure transmitters to withstand design basis work is organized into ten tasks: (1) program management, (2) fracture-methodology and analysis. (3) material charactenzation accdent (DBA) conditions are discussed. It is shown that a com-and propertes, (4) environmentally assisted crack growth stud. bination of operability monitonng and condition monitoring may ies, (5) crack arrest technology, (6) irradiation effects studies, result in an improved probability for withstanding DBA condi-(7) cladding evaluations, (8) intemlediate vessel tests and anal- tions.

ysis, (9) thermal-shock technology, and (10) pressunzed ther-mal-shock technology. NUREG/CR-4265 V02. AN ASSESSMENT OF THE SAFETY IM-PLICATIONS OF CONTROL AT THE CALVERT CLIFFS-1 NU-NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONU- CLEAR PLANT. Volume 2: Appendices. BALL,S.J. Oak Ridge Na-CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE tional Laboratory. July 1986. 364pp. 8610030411. ORNL/TM-HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE 9640/V2* 37675:178' PROJECTS Annual Report For October 1984 -September 1985. Th M/ esmined the consequences of possible control MEYER,R.E.; ARNOLD,W.D.; BLENCOE,J.G.; et al. Oak Ridge system malfunction at the Calvert Cliffs-1 nuclear power plant National Laboratory. July 1986. 62pp. 8607180003. ORNL/TM- as technical support for an NRC program to assess the safety 9614 V4. 37076:181_ implications of nuclear power plant control systems. Plant sys-Information pertaining to the potential geochemical behavior tems capable of initiating plant overcooling and undercooling of radionuclides at candidate sites for a high-level radioachve were identified, as well as those with potential for overfill events waste repository, which is being developed by projects within the Department of Energy (DOE), is being evaluated by Oak in the steam generators. Failure mode and effects analyses were conducted on these candidate plant systems, with com-Ridge National Laboratory for the Nuclear Regulatory Commis, sion (NRC). During this report period, emphasis was placed on puter analysis applied where appropriate. This later process uti-lized a detailed RETRAN model of the Ca! vert Cliffs plant using the evaluation of information pertinent to the Hanford site in southeastern Washington. Results on the sorption / solubility be- adaptations made as part of this program. Where failures with havior of technetium, neptunium, and uranium in the basalt / safety consequences were found, probabilites of the pertnent water geochemical system are summarized and compared to scenarios were developed. Several control system failures were the resutts of DOE. Also, summaries of results are reported identified as being of possible safety concern. Of these, two from two geochemical modeling studies: (1) an evaluation of the were selected as being of sufficient interest to warrant further information developed by DOE on the native copper deposits of study and followup using plant simulations. The Appendices Michigan as a natural analog for the emplacement of copper contain detailed documentation.

Main Citations and Abstracts 11 NUREG/CR-4300 V03 N1: ACOUSTIC EMISSION / FLAW RELA- fied by experiment, which can be used to predict reactor and TIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR safety systems performance and behavior under abnormal con-PRESSURE VESSEL # Progress Report October 1985 - March ditions. The objective of this work is to provide NRC requisite 1986. HUTTON,P.H.; KURTZ,R.J. Battelle Memonal Institute, data bases and analytical methods to (1) identify and define Pacific Northwest Laboratories. July 1986. 23pp. 8607240365, safety issues, (2) understand the progression of risk-dgnificant PNL-5511. 37242:339. accident sequences, and (3) conduct safety assessments. The This report discusses technical progress for the period Octo- collective NRC-sponsored effort at Sandra National Laboratories ber 1985 to March 1986 for the NRC-sponsored research pro- is directed at enhancing the technology base supporting licens-gram concemed with " Acoustic Emission / Flaw Relationships for ing decisions.

Inservice Monitonng of Nuclear Reactor Pressure Boundaries."

Included in the discussion are the tupics of AE monitonng of pri. NUREG/CR-4367 ROI: ORVIRT.PC (VERSION 2.0):A 2-D FINITE-mary pving dunng reactor operation, continued substantiation of ELEMENT FRACTURE ANALYSIS PROGRAM FOR A MICRO-the AE signal identification method, development of AE/lGSCC COMPUTER. BRYSON.J.W. Oak Ridge National Laboratory.

relationships, and progress in establishing an ASTM AE stand. August 1986. 70pp. 8610030418. ORNL-6208/RI. 37672:261.

ard and ASME code acceptance of AE. ORVIRT.PC (Oak Ridge virtual crack extension. Personal Computer) is a 2-D finite element fracture-analyses programs NUREG/CR-4308: VANESA:A MECHANISTIC MODEL OF RADI. for an IBM PC/AT microcomputer. The code is based to a large ONUCLIDE RELEASE AND AEROSOL GENERATION DURING extent on the technique

• used in the ORMGEN-ADINA-ORVIRT CORE DEBRIS INTERACTIONS WITH CONCRETE. fracture-anafysis system. ORVIRT.PC is a stand-alone program POWERS,D.A.; BROCKMAN.J.E.; SHIVER,A.W. Sandia National capable of performing 2-D linear elastic stress and fracture-me-Laboratones. July 1986. 471pp. 8610030404. SAND 85-1370. chanics analysem. Thermal loadings may be analyzed in addition 37679.001. to mechanical loadings. Crack-face tractions may also be con-This document desenbes a model, called VANESA, of the re- sidered. Eight-noded isoparametric elements which combine bom pedormance and ease of rnoddng are employed in N lease of radionuclides and generation of aerosol accompanying *

• reactor core melt interactions with structural concrete. The doc- E' 9'" * #' ""

square root variation en the near-tip stress and strain fields are ument also serves as a user's manual for an implementation of used at the crack tip. Detailed user instructions are provided the VANESA model as a computer code. The technical bases which desenbe both preparation of input data and program op-for the VANESA model are reviewed. This review includes a de- eration. Sample problems are presented which demonstrate senption of the thermodynamics and kinetics of vaporization good agreement with known solutions.

from melts sparged by gases evolving from concrete. The ther-mochemistries of 25 dements of interest in reactor accident NUREG/CR-4439: LEPRICON ANALYSIS OF PRESSURE anatyses are described. Limitations to the rate of vaponzation VESSEL SURVEILLANCE DOSIMETRY INSERTED INTO H.B.

caused by condensed phase mass transport, surface process- ROBINSON-2 DURING CYCLE 9. MAERKER,R.E. Oak Ridge es, and gas phase mass transport are discussed. Limitations on National Laboratory. September 1986. 70pp. 8610030412.

the extent of vaponzation caused by the behavior of bubbles ORNL/TM-10132. 37669:151.

rising in a melt are treated. Mechanical generation of aerosols A second example of appfying the LEPRICON methodology as bubbles burst at melt surfaces or as a result of liquid entrain- to an existing pressurized water reactor is described. The ment is considered. A description of these processes based on present application is an analysis of ad hoc dosimetry inserted data for gas-sparged water systems is included in the VANESA into the H.B. Robinson-2 reactor to tronitor the effects on pres-model. The document concludes with a description of computer sure vessel fluence produced by the introduction of a low-leak-code implementation of the VANESA model. This implementa. age fuel management scheme during cycle 9. The use of simul-tion of the model was used in recent assessments of the be- taneous dosimetry both at a downcomer location and in the re-havior of radionulcides during severe reactor accidents. Com- actor cavity allowed a quantitative evaluation to be made by the parisons of the predictions of radionuclide release during core LEPRICON procedure of the relative merits of each location.

debris / concrete interactions obtained with the VANESA mode, Unfolded results using the dosimetry indicate that the cumula-and with older models are presented. tive neutron fluence above 1 MeV originally calculated for the critical lower circumferertial weld in the pressure vessel during NUREG/CR-4336: REVIEW OF llT RESEARCH INSTITUTE RE- cycle 9, 7.2x10(17)n/cm(2 plus minus)15.9 percent, should be PORTS ON STRUCTURAL STUDIES OF REACTOR CONTAIN-adjusted upward by about one standard deviation to a value of MENT.

• Sandia National Laboratories. CHIAPETTA,R.L Chia- 8.4x1007)n/cm(2) we a reduced uncMaW of 103 pem petta. Welch & Associates, Ltd. June 1986. 83pp. 8609120004. NUREG/CR-4445: EFFECTIVENESS AND SAFETY ASPECTS OF SAND 85-7207. 37815:112. SELECTED DECONTAMINATION PROCESSES. DUCE S.W.;

The report reviews the work conducted at Armour Research CASE,J.T, Idaho National Engineering Laboratory. August 1986.

Foundation subsequently named ITT Research Institte (ITTRI). 229pp. 8608280269. EGG-2453. 37664:155.

on structural aspects of nuclear reactor containment. This work in October 1983 a study titled " Effectiveness and Safety As-was originaily performud in the time period from 1958 to 1966. pects of Selected Decontamination Processes" was initiated by The general objective of the review is to assess the result of the Nuclear Regulatory Commission. The thrust of the program the ITTRI research for its possible application to present-day was to review the current chemical decontamination processes steel containment problems. being used to remove oxide films in light water reactor (LWR) primary coolant systems. Four different process applications, NUREG/CR-4340 V02: REACTOR SAFETY RESEARCH SEMI- CAN-DECON, LOMI, DOW NS-1 and Dilute CITROX were ob-ANNUAL REPORT. July - December 1985.

• Sandia National served at nine different facihties, two pressunzed water and Laboratories. July 1986. 350pp. 8610030396. SAND 85-1606. seven boiling water reactors. in addition, information on chemi- '

37676:263. Sandia National Laboratories is conducting, under cal decontaminations at four additional facilities was obtained.

USNRC sponsorship, phenomenological research related to the This report presents the results of this study as they pertain to safety of commercial nuclear power reactors. The research in- the effectiveness of the chemical decontamination processes cludes experiments to simulate the phenomenology of the acci- observed. The general topic areas of this report are (a) decon-dent conditions and the development of analytical models, veri- tamination factors, (b) man-rem savings, (c) decontamination radwaste, and (d) problems observed and lessons leamed.

12 Main Citations and Abstf acts NUREG/CR-4470: SURVEY AND EVALUATION OF VITAL IN- grees centigrade. The viscosities of the molten liquids were STRUMENTATION AND CONTROL POWER SUPPLY EVENTS. nearty independent of UO(2) content and vaned between 10 MUHLHEIM.M.D.; CASADA,M.L.; THORNTON,R.L; et al. Oak and 17 centipoise. The apparent melting temperature where Ridge National Laboratory. August 1986. 314pp. 8608260339. fluid behavior first appeared varied with composition. For Zr-ORNL/NOAC-230. 37648:139. UO(2) mixtures with 15 mo1% or greater UO(2), the tempera-This report desenbes Task 15, Review of Experience Data for ture at which fluid behavior first appeared was lower upon initial Power Supply Adverse System Interaction Events, for the heating then on subsequent heating cycles. This result suggest Systern Interac%n project (FIN No. 80789) sponsored by the that the apparent melting temperature depended on the micros-Nuclear Regulatory Commission being perfsrmed at the Oak - tructure of the solid phase.

Ridge Nabonal Laboratory Nuclear Operations Anafysis Center.

The task included (1) assessing the nuclear power plant operat. NUREG/CR-4521: TURC2 AND 3.LARGE SCALE UO2/ZRO2 ing experience data sources, (2) developing search methods MELT-CONCRETE INTERACTION EXPERIMENTS AND ANAL-and event selection enteria for identifying losses of instrumenta- YSIS. GRONAGER J.E.; SUO-ANTTILA.A.; BROCKMANN J.E.

tion and control power supply ever,ts, (3) reviewing all of the Sandia National Laboratories. June 1986.179pp. 8609110311.

events selected, and (4) performing a final evaluation of the SAND 86-0318. 37803:190 events. This report outlines each of these steps and presents Two large scale UO2/ZrO2/Zr debris-concrete experiments the resu'!s of this review. TURC2 and TURC3 are reported here. The expenments con-

• ""O * '* * "" "

NUREG/CR-4480: EROSION PROTECTION OF URANIUM TAIL-INGS IMPOUNDMENTS. WALTERS,W.H.; SKAGGS R.L tures onto limestone-common sand conrete. The molten materi-FOLEY.M.G.: et al. Batte!!e Memorial Institute, Pacific Northwes"t al was allowed to cool naturally -no intemal heating was Laboratories. September 1986.105pp. 8610030408. PNL-5724. present. Data for conrete ablation, gas evolution including com.

37678d 80. position and flow rate, and aerosol generation are presented.

The experimental results indicate very rapid crustirm with no de-Pacific Northwest Laboratory (PNL) prepared this report to tectable concrete ablatio, Gas reduction of HD to H2 and CO assist in the design and review of erosion protecton works for was found to occur even with a purety oxide (UO2/ZrO2) melt.

decommissioned uranium ta!!ings impoundments. The major causes of erosion over the long-term decommissioning penod Aerosol concentrations vaned from 62 g/m 3 to less than 1 g/

m 3 n the experiments A thermal analysis of the expenments are from rainfall-runoff (overland flow) and stream channel was performed. The analysis is consistent with the result that flooding. The method of protection recommended for the im-rapid crusting with minimal concrete ablation occurs in both ex-poundment side slopes and site drainage channels is rock periments.

nprap. Mulch is recommended for the top surface combinations of vegetation and rock. The design methods were developed NUREG/CR-4529: PIPING DAMPING - EXPERIMENTAL RE-from currently available procedures supplemented by field, labo- SULTS FROM LABORATORY TESTS IN THE SEISMIC ratory, and mathematical model studies performed by PNL RANGE. WARE.A.G.; ARENDTS.J.G. Idaho National Engineer-Guidelines for the placement of riprap, inspection, and mainte- ing Laboratory. June 1986.102pp. 8607240282. EGG-2447.

nance are presented. Other subjects discussed are rock selec- 37245:027.

tion and testing, slope stability, and overland erosion modelin9- The Idaho National Engineering Laboratory (INEL) has been NUREG/CR-4487: UNCERTAINTY AND SENSITIVITY ANALYSIS conducting a research program to assist the United States Nu-OF A DRY CONTAINMENT TEST PROBLEM FOR THE clear Regulatory Commission (USNRC) in determining best-esti-MAEROS AEROSAL MODEL. HELTON.J.C.; IMAN,R.L.; mate damping values for the seismic analysis of nuclear piping JOHNSON,J.D.; et al. Sandia National Laboratories. June 1986. systems. As part of this program, a 5-in. piping system was 65pp. 8609110287. SAND 85-2795. 37802.248. tested by the INEL, and data from USNRC/EPRI piping vibra-The MAEROS aerosol model is being incorporated into the tion tests at the ANCO Engineers facility were evaluated. These MELCOR code system for the calculaticn of risk from severe re, systems were subjected to various types of excitation methods actor accidents. To gain insght to assist in this incorporation, a and magn!tudes, the support configuration were varied, and the computational test problem involving a five-component aerosol effects of pipe insulation and intemal pressure were investigat-in the containment of a pressunzed water reactor was analyzed ed on the INEL system. The INEL has used several different with MAEROS. The following topics were investigated: (1) the methods to reduce the data to determine the damping in both CRAY 1 CPU time requirements to implement and solve the these pip;ng systems under the various test conditions. It was system of different al equations on which MAEROS is based, (2) concluded that at representative seismic excitation levels, pres-tha effects on comoutatonal time and representational accuracy sure was not a contnbuting factor, but the supports, insulation, due to the use of different overall section boundanes and num_ and magnitude of response all were major influences contribut-bars of sections and components, and (3) the behavior of the ing to damping. These tests are part of the ongoing program to a:;rosol and the vanables which influence this behavior. Uncer. determine how various parameters and data reduction methods tainty and sensitivity analysis techniques based on Latin hyper- affect piping system damping. The evaluation of all relevant test cube sampling and regression analysis were used in the investi- results, including these two series, will potentially lead to re-gation. Ten sectons and overall section boundaries from 0.E-6 vised damping guidelines for the seismic analysis of nuclear m to 50.E-6 m were found to be adequate for the problem plants, making them safer, less costfy, and easier to inspect and under consideration. Further, solution time was generally found maintain. The test results as well as accompanying evaluations to be a thousand times or more faster than real time, which is and recommendations are presented in this report.

felt to be adequate for MELCOR. Stepwise regression, stand-ardized regression coefficients and partial correlation coeffi- NUREG/CR-4530 V01: U.S./ FRENCH JOINT FUSEARCH PRO-cients were used to investigate the sources of variation in com- GRAM REGARDING THE BEHAVIOR OF POLYMER BASE putational time and suspended aerosol concentration. MATERIALS SUBJECTED TO BETA RADIATION.Vol 1: Phase-1 Normalization Results. WYANT.F.J.; BUCKALEW,W.H. Sandia NUREG/CR-4495: VISCOSITY OF ZlRCONIUM-URANIUM OXIDE National Laboratories. May 1986. 62pp. 8607240369. SAND 86-(Zr UO2) MIXTURES AT 1800 TO 2100 C. BUNNELLL.R.; 0366. 37243:102.

PRATER.J.T. Battel le Memonal Institute, Pacific Northwest Lab- As part of the ongoing multi-year joint NRC/CEA intemational oratories. Seotember 1986. 58pp. 8610020282. PNL 5627. cooperative test program to investigate tt's dose-damage 38074:273. equivalence of garr,ma and beta radiation on polymer base ma-The mcosities of Zr-UO(2) mixtures with 0 to 30 mo1% terials, dosimetry and ethylene-propylene rubber (EPR) oci.

UO(2) have been measured at temperatures up to 2100 de- mens were exchanged, irradiated, and evaluated for p dy

Main Citations and Abstracts 13 changes at research facilities in the U.S. (Sandia National Lab- of EPR cables, (2) investigate differences between using super-oratories) and France (Compagnie ORIS Industrie). The purpose heated-steam or saturated-steam at the start of an accident of this Phase-1 test series was to normalize and cross-correlate simulabon, (3) determine whether the aging technique used in the results obtained by one research center to the other, in the saturated test induced arbfical degradation, and (4) identify terms of exposure (1.0 MeV accelerated electrons and Co(60) the constituents in EPR which affect moisture absorpbon. The gammas) and postirradiation testing (ultimate elongation and cable electrical degradaton was determined by insulaton resist-tensile strength, hardness, and density) techniques. The dosim- ance and AC leakage current measurements. One aged multi-etry and material specimen results indicate good agreement be- conductor cable porducthad electncal degradation. Therefore, tween the two countries regarding the exposure conditions and the current qualification practice of using single conductor posbrradiation evaluation techniques employed. cables to qualify multiconductor cables may not be a conserva-tive approach for all cables. Physical and tensile properties NUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN (measured after the accident) for insulaton specimens did not NUCLEAR POWER PLANT PERSONNELA Feasibility improve as the accelerated aging time was increased. There-Study. Volume 1: Summary Of Results. WOODS.D.D.;

fore, the aging technique did not induce artificial degradation. In ROTH,E.M.; HANES LF. Westinghouse Electric Corp. July addition, the constituents that appear to affect moisture absorp-1986. 27pp. 8608190398. 37602:299. tion and or produce other chemical changes are fire retardants, This report summarizes the results of a feas"oility study to de- nonsurface treated clay, or lack vinyt silane.

termire if the current state of mode's )f humaa cognitive active-ties can serve as the basis for imph ved techniques for predict- NUREG/CR-4540: AN EVALUATION OF J-R CURVE TESTING ing human error in nuclear power plai 's' emergency operations. OF NUCLEAR PIPING MATERIALS USING THE DIRECT CUR-Based on the answer to this question, 'wo subsequent phases RENT POTENTIAL DROP TECHNIOUE. HACKETT.E.M.;

of research are planned. Phase 11 ss to oevelop a model of cog- KIRK M.T.: HAYS,R.A. David W. Taylor Naval Research & De-nitive activities, and Phase ill is to test the model. The feas4biMY velopment Center. August 1986. 76pp. 8600120251. 37815:034.

study included an analysis oi the cognitive activities that occut A method is described for developing J-R curves for nuclear in emergency operations and an assessment of the modeling piping materials using the D C. Potential Drop (DCPD) tech-concepts / tools available to capture these cognitive activities- nsque. Experimental calibration curves were developed for both The results indicated that a symbolic processing (or artificial in- three point bend and compact bpecimen geome' ries us:ng teiligence) model of cognitive activitie. in nuclear power plants ASTM A106 steel, a type 304 stainless steel and a high is both desirable and feasible. This cognitive model can be built strength aluminum alloy. These curves were fit with a power law upon the computational framework provided by an existing artifi- expression over the range of crack extension encountered cial intelligence system for medical problem solving. called Ca- during J-R curve tests (0.6 a/W to 0.8 a/W). The calibration ducex The resulting cognitive model will increase the capabil- curves were bsensitwe to both material and sidegrooving and ity to capture the human contribution to risk in probabilistic risk depended solely on specimen geometry and lead attachment assessment studies. Volume 1 summanzes the major findings points. Crack initiation in J-R curve tests using DCPD was deter-and conclusions of the study. Volume 2 provides a complete mined by a deviation from a linear region on a plot of COD vs.

desenption of the methods and results, including a synthesis of CCPD. The validity of this criterion for ASTM A106 steel was the cognitive activities that occur during emergency operations, determined by a series of multispecimen tests that bracketed and a hterature review on cognitive rnodeling relevant to nuclear the initiation region. A statistical differential slope procedure for power plants- determination of the crack initiation point is presented and dis-NUREG/CR-4532 V02: MODELS OF COGNITIVE BEHAVIOR IN cussed. J-R curve tests were performed on ASTM A106 steel NUCLEAR POWER PLANT PERSONNEL A Feasibility and type 304 stainless steel using both the elastic compliance Study. Volume 2: Main Report. WOODS.D.D.; ROTH,E.M.; and DCPD technique to assess R-curve comparability. J-R HANES,LF. Westinghouse Eixtric Corp. July 1986. 130pp. curves determined using the two approaches were found to be 8608200031. 37572:055. See NUREG-CR-4532 Vol 1. in good aggreement for ASTM A106 steel. The applicability of the DCPD technique to type 304 stainless steel and high rate NUREG/CR-4533: PROGRAM TO ANALYZE THE FAILURE loading of ferromagnetic materials is discussed.

MODE OF LEAD-ACID BATTERIES. ZUCKERBROD,D. Wes-tinghouse Electric Corp.

• Sandia National Laboratories. April NUREG/CR-4553: FIELD TESTS OF A PORTABLE TISSUE 1986. 34pp. 8607240367. SAND 86-7080. 37243:163 EQUIVALENT SURVEY METER FOR MONITORING MIXED The electrical characte-istics of large lead-acid cells from nu- BETA / GAMMA RADIATION FIELDS. MARTZ,D.E.; RICH,B.L; clear power plants were studied. The overall goal was to devel- JOHNSON,LO.; et al. EG&G Idaho, Inc. (subs. of EG&G, Inc.).

op nondestructive tests to predict cell failure using this easily May 1986. 48pp. 8607240349. EGG-2448. 37243:215.

obtained information. Cell capacitance, intemal resistance, for A portable radiation survey meter that provides a tissue equiv-hydrogen evolution and cell capacity were measured on a lead- alent response to photons and beta particles has been de-calcium cell in pd condition. A aigh float voltage and low in- signed and field tested. The detector is a very thin plastic scin-temal resistam were found to correlate with good cell capacity tillator that closely simulates the actual geometry and scattenng in cells selected from a set of six lead-antimony cells in poor properties of the relevant skin tissues. The meter reads uut the condition. D(0.07) dose rate directly, and indicates the tissue dose rates at other depths with the use of tissue equivalent filters of appropri-NUREG/CR 4536: SUPERHEATED-STEAM TEST OF ETHYLENE ate thickness. Data are presented w,ich compare the D(0.07)

PROPYLENE RUBBER CABLES USING A SIMULTANEOUS and D(10) dose rates recorded by the Tissue Equivalent (TE)

AGING AND ACCIDENT ENVIRONMENT. BENNETT,P.R.; ST survey meter with dose rates recorded by two commercial icn CLAIR,S.D.; GILMORE,T.W. Sandia National Laboratories. June chamt;er meters for a number of, laboratory and field sources.

1986.140pp. 8609120170. SAND 86-0450. 37814:254. Most commercial ion chamber meters fail to respond adequate-The superheated-steam test exposed different ethylene ly to the extreme off-axis particles from extended beta sources, rubber (EPR) cables and insulation specimens to simultaneous and hence require the application of targe beta correction fac-againg and a 21 day simultaneous accident environment. In ad- tors to change the instrument reading to the true D(0.07) dose dition, some insulation specimens were exosed to five different rate. The tissue equivalent survey meter exhibits an angular re-againg conditions prior to the 21 day simultaneous accident sponse to beta particles that is very similar to the angular re-simulation. The purpose of this superheated-steam test (a sponse of an extrapolation chamber. Consequently, there is follow-on to the satuarated-steam tests NUREG/CR-3538) was close agreement between the TE meter and extrapolation to (1) examine electrical degradation of different configurations chamber readings for a wide variety of beta and mixed beta /

14 Maln Citations and Abstracts gamma radiation fields. D(0.07), D(3), and D(10) dose rates. The computer code COMPBRN lli deterministically models measured with the INEL TE meter at a number of typical work the behavior of compartment fires. This code is an improvement j stations, are presented. of the original COMPBRN codes. It employs a different air en. '

trainment mode and numerical scheme to estimate properties of l NURE2/CR-4558: INTERACTION OF HOT SOLID CORE DEBRIS WITH CONCRETE. COPUS.E.R.; BRADLEY,D.A. Sandia Nation. N ceding hot gas layer model Woover, NBM N Mcm porates a number of improvements in shape factor calculations l d Laboratories. June 1986. 73pp. S610030211, SAND 85-1739.

37676:189. and error checking, which distinguish it frcm the COMPBRN 11 l Two experiments were performed in order to compare hot code. This report presents the ceiling hot gas layer model ern sclid UO2 concrete and hot solid steel-concrete interactions. ployed by COMPBRN lit at wer as several other modifications.

The first expenment, HSS-1, ablated 6 cm of limestone-common Informat on necessary to rur. COMPBRN 111, including descrip-sand concrete in a little more than three hours using a 9 Kg tions of required input and resulting output, are also presented.

slug of 304 stainless steel at an average debris temperature of Simulation of experiments and a sample problem are included 1350 degrees C. The second experiment. HSS-3, ablated 6.5 to demonstrate the usage of the code.

cm of limestone-common sand concrete in four hours using a 10 Kg slug of 30% UO2-20% ZrO2 at an average debris tem. NUREG/CR-4570: DESCRIPTION AND TESTING OF AN APPA-pertture of 1GSO degrees C. The information from these tests RATUS FOR ELECTRICALLY INITIATED FIRES THROUGH wts used to evaluate a 1-D concrete ablation model named SIMULATION OF A FAULTY CONNECTION. SPLETZER,B.L; HOTROX. HORINE,F. Sandia National Laboratories. June 1986. 38pp.

NUREG/CR-4560: RISK ASSESSMENT APPLICATION TO NRC An apparatts has been developed that allows the simulation INSPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO JANUARY 1986. CAMPBELL.D.J.; GUTHRIE V.H.; of a fauW conmeton in an eWaW W pa&g a sman e KIRCHNER,J.R.; et al. Oak Ridge National Laboratory. June sistance heater at the screw of a terminal stnp. The spparatus 1986. 71pp. 8607240281. ORNL/TM-10001. 37248:168. and associated control system are descnbed in detail. Details of This report describes progress on a research project aimed at a typical fire produced with the apparatus are presented, along accomplishing two objectives: (1) to evaluate the use of prob- with results of electncal fire intiat'on atterrpts with both qualified Abilistic risk assessment (PRA) results for NRC inspection deci- and unqualified nuclear power plant cable. Repeated use of the sion-making purposes and (2) to supply inspection personnel apparatus has shwn that a self-sustaining fire can be reliably with risk-related information in a format that will help them initiated in unqualified cable with power levels to the apparatus decide where to focus their efforts to reduce plant risk. The first not exceeding 200 W. Such fires may be initiated in approxi- l' objective has been accomplished by identifying decisions in. mately 10 minutes. Large-scale fires have been initiated with spectors need to make and the PRA-related information that will the apparatus, indicating that propagation to any desired fire support these decisions. A review of the inspection procedures size is possible. Fires initiated in qualified cable to date have in the inspection and Enforcement Manual, interview with in- been localized and neither self-sustaining nor capable of propa-spectors, and a review of documented PRAs were what allowed gation, the identification of inspection decisions and useful risk-related information. The second project objective is being accomplished NUREG/CR-4574: AN EXPERIMENTAL AND ANALYTICAL AS-through the development of the Plant Risk Status Information SESSMENT OF CIRCUMFERENT!AL THROUGH-WALL Mirngement System (PRISIM). PRISIM is a menu-dnven pro- CRACKED PIPES UNDER DURE BENDING. SCOTT,P.;

grim for a personal computer that will allow inspectors in the -

BRUST,F. Battelle Memorial Institute, Columbus Laboratories.

field to quickly access PRA-related information they can use for September 1986.119pp. 8610070135. BMI-2136. 38128:143. I miking decisions. The first PRISIM prototype is being devel- This study performed to assess the validity of various tech-opad for the ANO-1 Plant and is nearty complete. This report niques to predict crack initiation loads and maximum loads for describes, in detail, the program's value as a decision aid, spe-cific types of information provided by the program, plans for circumferentially through-wall-cracked pipes under pure banding.

testing and evaluating the program, and future plans for the de- Experimental data were developed for both carbon steel &

vtlopment of PRISIM prototypes for other nuclear power plants. stainless steel pipes. Predictions of crack initiation and maxi- l mum loads were made using net-section-collapse method, three l NUREG/CR-4562: PIPE DAMPING-RESULTS OF VIBRATION different J-estimation schemes, and the British R6 method. The TESTS IN THE 33 TO 100 HERTZ FREQUENCY RANGE. net-section-collapse method gave good maximum-load predic-WARE.A.G. Idaho National Engineenng Laboratory. July 1986. tions for certain types of pipe; however, fc. large diameter and/

53pp. 8609120161. EGG-2450. 37814:202. or low toughness pipe this analysis method tended to overpre-R; search was performed for the U.S. Nuclear Regulatory dict the expuimental maximum load. A plastic-zone screening Commission (USNRC) by the Idaho National Engineering Labo- cnterion was developed to show when this method was valid ritory (INEL) to determine best-estimate representations of and when elastic-plastic fracture mechanics should be used. In damping values for nuclear piping systems excited in the 33 to the J-estimation scheme analyses, sensitivity studies were con-100 Hz frequency range. Vibrations in this frequency range are ducted to assess the fit of the Ramberg-Osgood coefficients, as typicd of fluid-induced tranients, for which no formal pipe damp-ing guidelines exist. The available data found in the literature well as the use of deformation J and modified J (JM) crack growth resistance curves. The results showed that the GE/EPRI (nd the USNRC/INEL pipe damping data bank were reviewed, md two series of tests on 3- and 5-in. laboratory piping systems estimation scheme underpredicted the experimental loads by supported with several typical nuclear piping supports were con- the greatest amount. The LBB.NRC and Paris method gave ducted as part of this research program. It was concluded that more accurate predictions. The GE/EPRI method was also a suit"ble representation of the test data can be made by ex- found to be more senshe to N 6t of N stress-strain cme tending the Presuure Vessel Research Committee (PVRC) unin- than theLBB.NRC method. The R6 method underpredicted the sul;ted piping curve at 3 % of entical damping for frequencies failure loads for all cases. For maximum load predictions, the

> 20 Hz. GE/EPRI method still underpredicted the experimental load when the JM resistance curve was used. The other methods NUREG/CR-4566: COMPBRN lit - A COMPUTER CODE FOR occasionally overpredicted the maximum load using JM-resist-MODELING COMPARTMENT FIRES. HO,V.; SIM,N.; ance curve.

APOSTOLAKIS,G.; et al. Oak Ridge National Laburatory. April 1986.211pp.8609120154. ORNL/TM-10005. 37813.065.

l l

l l

1 Main Citations and Abstracts 15 NUREG/CR-4575: PREDICTIONS OF J-R CURVES WITH LARGE - 2.6 + square root 18.8 (log E/M(o)) - (loa E/M(o))(2) -67.2 CRACK GROWTH FROM SMALL SPECIMEN DATA. PAPA- where E = source of energy, dyne-cm SPYROPOULOS; MARSCHALL,C.; LANDOW,M. Battelle Msmo. M(o)= weight of inert matenal, gram (E/M(o) is expressed in ria! Institute, Columbus Laboratories. September 1986. 87pp. dyne-cm/g).

8610030075 BMI-2137. 37670:197.

This repon examines the practice of extrapolting small-speci. NUREG/CR-4596: SCREENING TESTS OF REPRESENTATIVE men J-resistance curves for use in predictive analyses nf large NUCLEAR POWER PLANT COMPONENTS EXPOSED TO crack growth in nuclear pipes pror to instability. The study in. SECONDARY ENVIRONMENTS CREATED BY FIRES.

volved a combined expenmental and analytical effort. The ex. JACOBUS M.J. Sandia National Laboratories. June 1986. 76pp.

penmentai effort included tests of 1.0-inch (25.4mm)-thick com. 8609110270. SAND 86-0394. 'l7806.060.

pact (tension) specimens of 1T,3T, and 10T planar dimensions. This report presents results of screeri.ng tests to datermine Both side-grooved and nonside-grooved specimens of Type 304 component survivability in secondary environments created by stainless steel and A516 Grade 70 carbon steel were tested at fires, specifically increased temperatures, ;nceased humidity, 550 F (288 C) The data were analyzed using deformation-J and the resence of partcuates and corrosve vapors. Additional-(J(D)) and modified 4 (J(M)) estimation schemes to develop re- ly, chloride concentrations were measured in the exhaust from sistance curses. Also, elastic-plastic finite element analyses several of the tests used to provide fire environments. Results were performed of IT,3T, and 10T nonside-grooved specimen show actual failure or some indication of failure for strip chart data for Type 304 stainless steel. The results of these analyses recorders, electronic counters, an oscilloscope amplifier, and were then used to assess extrapolation procedures. The study switches and relays. The chart recorder falures resulted from suggests that extrapolations of J(D)-resistance curves to large accu,nulation of particulates in the pen slider mechanism. The amounts of crack growth result in large underestimates of the electroncs counter experienced leakage current failures n crcuit fracture resistance of the material. The extrapolations of J(M)- boards after the fire exnosure and exposure to hgh humidity.

resistance curves, e4 hough providing more accurate estimates The oscilloscope amplifist experieinced thermal-related dnft as of resistance to fracture, may result in overestimates. In gener- bgh as 20 % before thermal protective circuitry shut the unit al, the far-field finite element J-resistance curves were in better down. In sme cases, switches and relays experienced high agreement with J(M) than with J(D)-resistance curves. cntact resistances with the low voltage levels used for the measurements. Finally, relays tested to thermal failure experi-NUREG/CR-4567: SOURCE TERM CODE PACKAGE:A USER'S GUIDE (MOD 1). GRIESEKE.J.; CYBULSKIS,P.; JORDAN,H.; et enced various failures, all at tempera;ures ranging from 150 de-al. Battelle Me,aorial Institute, Columbus Laboratories. July grees centigrade to above 350 degrees centigrade. The chloride 1986. 279pp. 8608180483. BMI-2138. 37531:033. measurements show that most cf the hydrogen chloride gener.

As part of a major reassessment of the release of radioactive ated in the test fires is combined wrth particulate by the tinw it materials to the environment (source terms) in severe reactnr reaches the exhaust duct, indicating that hydrogen, chloride con-accidents, a group of state-of-the-art computer codes was uti- densation may be less likely than small scale data implies.

lized to perform extensive analyses. A major product of this NUREG/CR-4597 V01: AGING AND SERVICE WEAR OF AUXil-source term reassessment effort was a demonstrated methodol-ogy for analyzing specific accident situations to provide source lARY FEEDWATER PUMPS FOR PWR NUCLEAR POWER term predictions. The computer codes forming this methodology P.LANTS.Vol.1. Operating Experience And Failure identification.

have been named the Source Term Code Package (STCP) and Oak Ridge National Laboratory. ADAMS,M.L; MAKAY,E.

is the subject of this user's guide. The guide is intended to pro- Energy Research & Consultants Corp. July 1986. 100pp.

vide an understanding of the STCP structure and to facilitate 8609110306. ORNL-6282/V1. 37605:320~

STCP use. The STCP was prepared for operation on a CDC Typical auxiliary feedwater pump (AUXFP) configurations are described in terms of configuration details, materials of con-system but isversion In the current wntten in FORTRAN-77 (Mod 'o permit

1) of the STCP, the varioustransportability; calcula shn, omaW re@hnts, W Wes of opah AUXFP failure modes and causes due to aging and service tional elements fal1 into four major categones represented by the codes MARCH 3, TRAP-MELT 3, VANESA, and NAUA/ wear are identified and explained, and measurable parameters OncMng Wmal WWs) W potenbal use in asses @

SPARC>lCEDF. The MARCH 3 code is a combination of the MARCH 2,CORSOR-M, and CORCON-Mod 2 codes. The TRAP- perational readiness, establishing degradation trends, and de-ng M nt s are gun A sees of measm to cm MELT 3 code is a combination of the TRAP-MELT 2.0 and rect p. resent deficiences in surveillance, monitoring, and in-serv-MERGE codes' ice testing practices is discussed. The main body of the reprot NUREG/CR-4593: INITIAL CONCEPTS ON ENERGETICS & is supplemented by a number of relevant appendixes; in parth MASS RELEASES DURING NONNUCLEAR EXPLOStVE far, a mnjor appendix is included on engiraering and desgn in.

EVENTS IN FUEL CYCLE FACILITIES. HALVERSON,M.A.; formation useful to assess AUXFP operational readiness. This MISHIMA,J. Battelle Memorial Institute, Pacific Northwest Lab- report was produced under the U.S. Nuclear Regulatory Com-oratories. September 1986. 57pp. 8609160005. PNL-5839. mission's Nuclear Plant Aging Research Program.

37852:021.

Explosions are one of the initiating events (a ;cidents) consio- NUREG/CR-4598: A USER'S GUIDE FOR THF TOP EVENT ered in the U.S. Nuclear Regulatory Commisson study of formal MATRIX ANALYSIS CODE (TEMAC).11 Aft ; SHORTEN-methods for estimating the airborne release of radionuclides CARIER Sandia Nat%nal LaboraMes. August 10% 66pp.

from fuel cycle facilities. Methods currently available to estimate 8609290106. SAND 86-0960. 38012:218.

the energetics and mass airborne release trom the four types of This document has been designed for users of the comp 7 explosive events (fast and slow physical explosions and fast program, Top Event MAtarix Analysis Code (TEMAC), devel-and slow chemical explosions) are reviewed. The likelihood that opod by the authors at Sandia National Laboratory for estimat.

fast physical explosions will occur in fuel cycle facilities appears ing risk, and performing uncertainity and sensitivity analyses to be remote and this type of explosion is not considered. Moth. with a Boolean expression sucn as produced by the Set Equa-ods to estirr. ate the consequences of slow physical and fast tion Tranformation System (SETS) computer program (Worrell, chemical explosions are available. Methods to estimete the con- 1985). SETS produces a mathematical representation of a fault sequences of slow chemical explosions are less well defined. tree used to model system unavailability. In the terminology of An empirical fit to the fractbnal airborne release measured for the TEMAC program, such a mathematical representation is re-the vanous conditions provides a " bounding value" for the air. ferred to as the top event. This report contains a tutorial on the bome release and was found to be: matrix-based approach to analyzing top events along with de-log (wt% airbome) = tailed instructions on how to use the TEMAG program. The

16 Main Citations and Abstracts TEMAC program was wntten in FORTRAN 77 while attemptira GlESEKE,J.A.; CiBULSKid,P.; et al. Battelle Memonal Institute, i to make the code as machine-independent (i.e. portable) as Columbus Laboratories. July 198F 242pp. 8608070206.

possible within the confines of the ANSI standard for FORTRAN 37427:159. This report presents results of analyses of the envi-

77. In addition, much effort has been directed toward making ronmental releases of fission products (source terms) for severe the TEMAC program sery user f iendly, accident scenanos in a boiling water reactor of the Mark I con-tainment design. The analyses were perforrred to support the NUREG/CR-4600: HUMAN FACTORS STUDY CONDUCTED IN Severe Accident Risk Reduction / Risk Rebaselining Program CONJUNCTION WITH A MINI-ROUND ROBIN ASSESSMENT (SARRP) which is being undertaken for the U.S. Nuclear Regu-OF ULTRASONIC TECHNICIAN PERFORMANCE. latory Commission by Sandis National Laboratones. In the WHEELER,W.A.; RANKIN,W.L Battelle Human Affa rs Research SARRP program, nsk estimates are being generated for a Centers. BADALAMENTE.R.; et al. Battel'e Memorial institute, number of reference plant designs. The Peach Bottom 2 Piant Pacific Northwest Labora:ories. August 1986. 100pp. has been used in this study as an example of a Mark i BWR Plant.

8608280251. PNL-5757. 37665:033.

This report descabes the findings from a limited human fac-NUREG/CR-4624 V02: RADIONUCLIDE RELEASE CALCULA-tors (HF) study conducted in conjunction with a Mini-Round TIONS FOR SELECTED SEVERE ACCIDENT Robin (MRR) held at the Pacific Northwest Laboratory. The pur- SCENARIOS. Volume 2.PWR,1ce Condenser Design.

pose of the HF evaluation was to acqdre preliminary data on DENNING R.S.; GIESEKE,J.A.; CYBULSKIS,P.; et al. Battelle performance shaping factors that affect ultrasonic testing (UT) Memorial Institute Columbus Laboratories. July 1986. 224pp.

'eliability; to test tha efficacy of relative operating characteristic 8608070154. 37449:107. This report presents results of analy-analysis for representing UT accuracy; and to determine the di- ses of the environmental releases of fission products (source rection of future HF efforts in the NOT area. The purpose of the terms) for severe accident scenarios in a pressunzed water re-MRR was to evaluate the ability of nondestructive testing (NOT) actor with an ice-condenser containment. The ealyses were technicians to detect intergranular stress corrosion cracking performed to support the Severe Accident Risk Reduction Risk (IGSCC) using UT techniques. A further purpose was to meas- Reduction / Risk Rebaselining Program (SARRP) which is being ure improvements in technician performance following imple- undertaken for the U.S. Nuclear Regulatory Commission by mentation of l&E Bulletin 83-02 and qualification for IGSCC de- Sandia National Laboratories. In the SARRP program, risic esti-tection at the EPRI-NDE Center. Twelve UT technicians carned mates are being generated for a number of reference plant de-out UT for IGSCC on samnies of welded sections of piping that signs. The Sequoyah Plant has been used in this study as an either had IGSCC or : lid not have IGSCC. Each technicians also example of a PWR ice-condenser plant.

filled out a questonnaire about performance shaping factors that codd affect UT and participated in a entical incident ir ter. hbREG/CR-4624 V03: RADIONUCLIDE RELEASE CALCULA-view to provide information on relevant UT inserve:e inspaction TlONS FOR SELECTED SEVERE ACC,lDENT experience. In addition, the UT equipment used by the MRR SCENARIOS. Volume 3:PWR.Subatmospheric Containment teams was evaluated for conformance to HF design principles. Design. DENNING R.; GIESEKE,J.A.; CYBULSKIS P.; et al. Bat-telle Memonal Institute, Columbus Laboratories. July 1986.

NUREG/CR-4622: VALIDATION OF STOCHASTIC FLOW AND 92pp. 8608070215. 37428:267.

TRANSPORT MODELS FOR UNSATURATED SOILS.A Com- This report presents results of analyses of the environmental prehensive Field Study. WlERENGA P.J.; GELHAR,LW.; releases of fission products (source terms) for severe accident SIMMONS.C.S.; et al. Battelle Memorial Institute, Pacific North- scenarios in a pressurized water reactor with a subatmospheric containment design. The ana'yses were performed to support west Laboratories. August 1986. 69pp. 8609120257. PNL-5875.

the Severe Accident Risk Reduction / Risk Rebaselining Program 378:5:260 This document desenbes a comprehensive field study for s W u@@n b h M Mar W evaluating the effects of site variability on flow and transport of SARRP program,"nsk estimates are being generated for a contaminants in the unsaturated zone. The field study .is de- number of reference plant designs. The Surry Plant has been signed to measure the variebility of key hydraulic properties used in this study as the reference plant for a subatmospheric within the unsaturated zone of an undisturbed field soil. A large design.

trench, 5 m wide x 6.5 m deep x 25 m long, will be used to observe the transport of contaminants. Flow of water-coincident NUREG/CR-4624 V04: RADIONUCLIDE RELEASE CALCULA-contaminants through the unsaturated zone will be monitored, TIONS FOR SELECTED SEVERE ACCIDENT and compansons will be made between predicted and field. SCENARIOS. Volume 4:BWR, Mark lil Design. DENNING.R.S.;

measured values. In addition, a controlled tysimeter test has GIESEKE,J.A.; CYBULSKIS P.; et al. Battelle Memorial Institute, been initiated to look at the effects of well-defined layering on Columbus Laboratories. July 1986. 169pp. 8608110084.

transport processes. The stochastic model of Gelhar will be 37450:049.

tested using the field and lysimeter test data. In this model, the This report presents results of analyses of the environmental releases of fission products (source terms) for severe accident fluctuations of measured hydraulic parameters are assumed to scenanos M a boikg water reacts we a Mark m containrnm represent statistically homogeneous random fields. The mean The analyses were performed to support the Severe Acesdent values of these parametors and their covariance functions will aseheg Wam @% wM is e n provide sufficient stat'stical charactenzation to evaluate the flow '*""' '" * *O 8 #Y "" * "

9 and transport processes at the site solely on the basis of these by Sandia National Laboratories."In the SARRP program, risk measured hydraulic properties. The data set will alsc be useful estimates are being generated for a number of reference plant in testing other models that ate available for performance as- designs. The Grand Gulf Plant has been used in this study as sessments of low-level waste sites and in addressing the ef* an example of a Mark Ill plant.

fects of spatial variability on transport of contaminants in the un-satrated zone. NUREG/CR-4624 VOS: RADIONUCLIDE RELEASE CALCULA.

TlONS FOR SELECTED SEVERE ACCIDENT NUREG/CR-4624 V01: RADIONUCLIDE RELEASE CALCULA- SCENARIOS.Vr%ne 5'Large Dry Containment Design.

TIONS FOR SELECTED SEVERE ACCIDENT DENNING,R.S.; (VESEKE,J.A.; CYBULSKIS P.; et al. Battelle SCENARIOS. Volume 1:BWR, Mark i Design. DENNNG,R.S.; Memorial Institute, Columbus Laboratories. July 1986. 158pp.

8608080277,37430:050.

~ - - -

_ a

Main Citations and Abstracts 17 This report presents results of analyses of the environmental NUREG/CR-4629: INDEPENDENT VERIFICATION OF RADIONU-releases of f.ssion products (source terms) for severe accident CLIDE RELEASE CALCULATIONS FOR SELECTED ACCIDENT scenarios in a pressurized water reactor with a large dry con- SCENARIOS. KHATIB-RAHBAR Brookhaven National Laborato-tainment. The analyses were performed to support the Severe ry. July 1986. 259pp. 8608180287. BNL-NUREG-51998.

Accident Risir Reduction / Risk Rebaselining Program (SARRP) 37543:222. An independent venfication of the Source Term which is being undertaken for the U.S. Nuclear Regulatory Com- Code Package (STCP) calculated source terms resulting from mission by Sandia National Laboratyies. In the SARRP pro- severe nuclear reactor accidents is presented. This consists of gram, nsk estimates are being generated for a number of refer- assessing the modeling options, data transfer, and problem no-

] ence plant designs. The Zion Plant has been used in this study dalization for both Boiling and Pressurized Water Reactor appli-J as an example of a large dry containment PWR design. cations. Independent calculations are also performed for four l NUREG/CR-4626 V01: IMPROVING THE RELIABILITY OF speck Ecdem sceah h Me Nnt @ waW man containment designs. A detailed review and assessment of the OPEN-CYCLE WATER SYSTEMS >.n Evaluation Of Biofouling Surveillance And Control Techniques For Use At Nuclear Power Battelle Columbus Laboratones (BCL) calculated results using Plants. NEITZEL,D.A.; JOHNSON,K.l.; DALING,P.M. Battelle the STCP for four reference plants is also made. Comparison of Memorial Insttute, Pacific Northwest Laboratories. September the STCP calculated source terms with the suite of codes as 19% 53pp. 8609170241. PNL 5876. 37885:005. documented in BMI-2104 is also given when applicable.

Nine surveillance and 11 control techniques were reviewed 3 for their applicability to open-cycle water systems at nuclear NUREG/CR-4633: ORPLOT.PC. A GRAPHIC UTILITY FOR j power plants. Joumal articles and technical reports were re- ORMGEN.PC AND ORVIRT.PC. INVERSINI C.; BRYSON.J.W.

viewed to collect information about the surveillance and control Oak Ridge National Laboratory. July 1986. 54pp. 8609090496.

techniques. Power plant personnel that are using or have used ORNL-6291. 37767:344-some of the techniques were it'terviewed to get data on the ef. ORPLOT.PC is an interactive graphic utility for' ORMGEN.PC fyctiveness of the techniques. A computerized decision support and ORVIRT.PC. It executes on an IBM PC/XT or PC/AT Mwam was used by the review staff to evaluate the tech. equipped with hard disk, graphic card, and 512K mirumum Nes The review results indicate all the techniques are used momory. The program is capable of (1) displaying finite element

k. wcycle water systems to survey for or control biofouling meshes generated by ORMGEN.PC complete with node num-v6.rwM be appled to the open cycle water system at nuclear bers, element numbers, and boundary conditions and (2) gener-m 14nts. The surveil!ance techniques reviewed include ating deformed mesh plots, contour plots, line (X-Y) plots, and frmW Mter ano substrate samples outside the plant by developed surface plots of ORVIRT.PC output. A zooming fea-SCfM :N mspection, monitonng growth paneis, inspecting ture allows detailed inspection of any subregion. Since simplicity sampiu num Nide the plant, measunng pressure differentials, and ease of use were important objectves during program de-measurwg tehyerature d:fferentials, and measuring flow differ, velopment, all commands are entered interactively using free entials. The convol techniques reviewed include chlorinating the format. The option of automatic or user defined scaling for most water using the AMERTAP system (AMERTAP Corporation), plots is ancther convenience. Plot files may be created and wnt-using the MAN system (MAN Corporation), installing tube scrap. ten to hard disk for subsequent hardcopy to printer or plotter.

ers, hydroblasting, applying antifoulant coatings, thermal back.

washing, rsing oxygen scaveagers, usi99 screens and strainecs' NUREGICR-4643: EVALUATION OF CORE DAMAGE SE-using hand scrapers, and nonthermal backflushing. OUENCES INITIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOLING. MITRA S.; BARADARAN,R. Impell NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATION DOSE Corp. YOUNGBLOOD,R. Brookhaven National Laboratory.

COMMITMENT FACTORS FOR MAJOR RADIONUCLIDES RE- August 1986. 59pp. 8609050447. BNL-NUREG 52003.

LEASED FROM NUCLEAR FUEL FACILITIES. CRISTY,M.; 37736:256.

LEGGETT R.W.; DUNNING.D.E.; et al. Oak Ridge National Lab- This report is concemed with core damage accident so-oratory. August 1986.118pp. 8609150258. ORNL/TM-9890. quences initiated by loss of component cooling water, leading to 36187:008.

loss of reactor coolant pump seal cooling, subsequent primary During the licensing process for nuclear fuel facilities, commit- coolant leakage, and failure to make up the coolant loss. Three ted dose equivalents must be calculated for potential exposures plants are considered: Indian Point Unit 3. Midland Unit 2 and to people in the area around these facilitie;. These committed Calvert Cliffs Unit 1. It is shown that design differences in the dose equivalents are usually calculated from tabulated dose- overall seal cooling support systems are responsible for consid-conversion facto's that convert the quantity of radioactive mate- erable variation in the likehhood of, and plarit response to, loss rial potent:sily taken in by individuals through ingestion or inha- of component cooling.

lation. For esiculating committed dose equivalents to children, the Nuclear Regulatory Commission has in the past appealed to NUREG/CR-4644: GEOCHEMICAL STUDIES OF COMMERCIAL age-specific dose-conversion factors listed in NUREG-0172 LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITES. Topical (1977), which is based on a computational methodology found Report. DAYAL,R.; PIETRZAK,R.F.; CLINTON.J.H. Brookhaven in ICRP Publication 2 (1959). Since the publication of NUREG- National Laboratory. June 1986.106pp. 8609100206. BNL-0172 new models and new concepts of risk have been provided NUREG-52004. 37785:214.

in ICRP Publications 26 and 30 (1977,1979). These documents The results of source term characten2aton studies for the provide a detailed methodology for calculating dose-conversion commercially operated low-level wste (LLW) disposal sites lo-factors for the various radionuchdes for an adult reference man. cated in the eastern United States are used to provide an un-The report NUREG/CR-0150 (1981) provides dose- conversion derstanding of the importance of hydoogical and geochemical factors for children. In this report are tabulated age-specific factors in comrolling the mechanics of leachate formation, evo-dose-conversion factors, given as multiples of the adult values, luton of leachate compositions, microbial degradation of organ-for inhalation or ingestion of each of the following isotopes: U- ic waste and development of anoxia n the trenches, and the 234, U-235, U-238, Th-228, Th-23G, Th-232, Ra-226, Ra-228, nature and extent of leaching of waste materials. The varying Pt>210, or Po 210. Our methodology is consistent as far as degrees of the inensty of these processes. as aetermined by practical wM1 that of ICRP Publications 26 and 30, but we have the different site characteristics, are clearty reflected in the con-modmed and extended the ICRP methodology as necessary to trasting leachate geochemistries of Maxey Flats and West include age dependence and to include metabolic and dosime- Valley trer%es, as compared to those of Barnwe'l and Shef-tric in'ormation that has been developed since the issuance of field 'rench64. These se important geochemicra! considerations these ICRP documents.

which not only define LLW source terms but also shed light on

18 Main Citations and Abstracts the nature of geochemical changes that are likely to occur NUREG/CR-4656: VERIFICATION TEST CALCULATIONS FOR along a redox gradont outside of the trench environment. THE SOURCE TERM CODE PACKAGE. DENNING,R.S.;

WOOTON.R.O.; ALEXANDER,C.A.; et al. Battelle Memonal in-NUREG/CR-4647: CALCULATION OF THE POWER DISTRIBU- stitute, Columbus Laboratories. July 1986.169pp 8608070351.

T10N IN THE VENUS PWR MOCK-UP BENCHMARK USING BMI-2140. 37425.306.

TWO-GROUP DIFFUSION THEORY, WILLIAMS.M.L; The purpose of this report is to demonstrate the reasonable.

LANDESMAN,M.; KAM.F.B. Oak Ridge National Laboratory. ness of the Source Term Code Package (STCP) results. Hand August 1986.46pp.8609290062. ORNL/TM 10095. 38012:317. calculations have been performed spanning a wide vanety of Bochmark calculatons of the relative power distnbution in the phenomena within the context of a single accident stequence, a

] VENUS PWR Mock-Up Expenment were performeJ with both loss of all ac power with late containment failure, in the Peach discrete ordinated transport theory and few-group diffusion Bottom (BWR) plant, and compared with STCP results. The theory. The diffusion calculabons use codes from the Electnc report identifies some of the limitations of hand calculation Power Research Institute (EPRI) Advanced Recycle Methodolo- effort. The processes involved in a core meltdown accident are gy Program (ARMP) which is based on methods commonly complex and coupkd Hand calculatons by their nature must used by U.S. utilities and vendors. The diffusion results are deal with gross simplifcatons of these processes. Their great.

compared with transport theory results as well as with the est strength is as an indicator that a computer code contains an measured power distnbution obtamed by gamma scans of the error, for example that it doesn't satisfy basic conservation pin-wile power calculations near the core penphery, which is laws, rather than in showing the analysis accurately repsesents the region that produces nearly all the neutrons that leak from reality. Hand calculations are an important element of venfica-the core and damage the reactor pressure vessel Results show tion but they do not satisfy the need for code validation. The tnat the dew-group diffusion calculations produce results which code validation program for the STCP is a separate effort. In are nearty as accurate as the transport calculations. The largest general the hand calculation results show that models used in error in the computed pin powers near the penphery, as com- the STCP codes (e g., MARCH, TRAP-MELT, VANESA) obey pared to measured values, is approximately 7%. The average basic conservaton laws and produce reasonable results. The discrepancy throughout the entire core is 2%, which is compa- degree of agreement and significance of the comparisons differ rable to the expenTiental uncertainty. A conservative enor esti- among the models evaluated.

mate of 5% to 10% is, therefore, recommended for the pin-wise 3 source distnbution used in determining the reactor vessel NUREG/CR-4659 V01: SEISMIC FRAGILITY OF NUCLEAR j fluence for cores similar to the VENUS configuation. POWER PLANT COMPONENTS (PHASE 1). BANDYOPAD-HYAY; HOFMAYER,C.H. Brookhaven National Laboratory. June NUREG/CR-4648; A STUDY OF TYPICAL NUCLEAR CONTA.N- 1986.101pp.8607240366. BNL-NUREG-52007. 37243.001.

MENT PURGE VALVES IN AN ACCIDENT ENVIRONMENT. As part of the Component Fragility Research Program, spon-WATKINS,J C.; STEELE,R.; HILL,R.C.; et al. EG&G Idaho, Inc. sored by the U.S. Nuclear Regulatory Comnssion, BNL is in-(subs. of EG&G, Inc ) August 1986. 79pp. 8609290138. EGG- volved in establishing seismic fragility levels for various nuclear 2459. 38012.071. power plant equipment by identifying, collecting and analyzing This report presents the results of the containment purge and existing test data from varcus sources. In Phase I r4 Ws pro-vent valve test program, conducted under the sponsorship of gram, BNL has reviewed approximately seventy test reports to the United States Nuclear Regulatory Commission (NRC), Office collect fragility or high level test data for switchgears, motor of Nuclear Regulatory Research. The test program investigated control centers and similar electncal cabinets. valve actuators butterfly valve operabihty and leak integnty under light-water-te- and numerous electrical devices, (e g., switches, transmitters, actor design basis and severe accident conditions Three nucle- indicators, relays) of various manufacturers and modals. This ar-designed butterfly valves typical of those used in domestic report provides an assessment and evaluat.on of the data col-nuclear power plant containment purge and vent apphcations lected in Phase 1. The fragihty data for medium voltage and low were tested. For a companson of response, tyco valves of the voltage switchgears and motor control centers are anayred same size with differing internal designs were tested. For ex- using the test response spectra as a measure of the fragibty trapolation insights, a larger-sized valve similar to one of the level. High level test data for some other components are also smaller valves was also tested. Dynamic flow tests were per- included in the report. The fragibty levels have been corrpared formed over the range of design basis accident pressures. Leak with those used in the PRA and Seismic Margin Studies. It ap-integnty testing was also performed at both design basis and pears that the BNL data better correlate with the HCLPF level severe accident temperatures and pressures. The valve experi- used in Seismic Margin Studies and can irr.r' rove this level as ments were performed with vanous piping configuratons and high as 60% for certain applications. Specific recommendations valve onentations to the flow to simulate the vanous installation are provided for proper application of BNL fragility data to other options in field apphcations. Testing was also performed in a studies.

standard ANSI tast section.

NUREG/CR-4660: ORGANIC COMPLEXANT-ENHANCED MO- i NUREG/CR-4649: SCALING ANALYSIS OF THE COUPLED BILITY OF TOXIC ELEMENTS IN LOW 4EVEL WASTES Final i HEAT DANSFER PROCESS IN THE H:GH-TEMPERATURE Report. SWANSON.J.L. Battelle Memonal Inststate, Pacific GAS-COOLED REACTOR CORE. CONKLIN.J.C. Oak Ridge Na- Northwest Laboratones. July 1986. 60pp. 8608070212. PNL-tonal Laboratory. August 1986. 36pp. 8609300542. ORNL/TM- 4965-10.37427:100.

10099. 38048.290 This is the final report of a project whose ob;ective was to de- l The differential equatons representing the coupled heat termine how and to what extent organic complexants affect the transfer from the sohd nuclear core components to the helium mobility of toxic elements in subsurface groundwater at com-in the coolant channels are scaled in terms of representative mercial low level waste disposal sites. The complexants EDTA quantiti, rhis scahng process identifies the relative impor- and picohnate, both of which are used in reactor decontamina-tance of the various terms of the couped differential equations. tion operatons, were studied most extensivety. Hydrous femc The relative importance of these terms is then used to simphty oxide, Fe(2)O(3).xH(2)O, and kaohnste clay were the soil compo-the numencal soluton of the coupled heat transfer for two nonts most used. Four tonic elements were studmd, Ni, Am, Cd.

bounding cases of full power operation and depressurization and Pu hi. Am and Pu have radoactive isotopes that are com-from full system operating pressure for the Fort St. Vrain High- monly present in commencal low-level wastes, and Cd is an ex-Temperature Gas-Cooled Reactor. This analysis ngorously justa- ample of a nonradoactive toxic element that might riso be in fees the simplifed system of equations used in the nuclear such wastes. A wide dr<ers,ty of effects of organic complexants safety analysis effort at ORNL. on toxic element sorpton was observe 1 Some complexes are

Main Citations and Abetracts 19 sorbed by soil components at some pH values, but others are with experimental results are presented.The TRAP-MELT 2 code  ;

not. Important reactions are slow in some systems but rapid in simulates the transport and deposition of aerosol particles and others. There are two separate reactions in which slow kinet.cs certain vapors in the reactor coolant system of a hght water re-have been observed in some systems; one is the slow dissocia- actor under hypothetical accident conditons.The improvements ton of a preformed complex and the other is the slow desorp- reported here are pnncipally concemed with fluid tion by complexant solutons of a prevously sorbed uncom- properties, vapor pressures,and turbulent deposition of plexed element. particles. Comparisons are made of code predictions with select-1 NUREG/CR-4667 V01: ENVIRONMENTALLY ASSISTED CRACK- ed Marviken project experimental results and with other results

_j ING IN LIGHT WATER REACTORS. Semiannual Report. April- pertinent to turbulent disposition of parbcles.

] September 1985. SHACK,W.J.; KASSNER,T.F.; MAlYA,P.S.; et NUREG/CR-4678: A METHOD FOR USING PRA TO ESTABLISH al gonne N o Laboratory. July 1986. 66pp. 8609300455.

~

QUALITY PROGRAM APPLICABILITY. GALLUP,D.R.;

This progress report summarizes work performed by Argonne WHITEHEAD.D.W.; VANNONI.M.G. Sandia National Laborato.

Natonal Laboratory on environmentally assisted cracking in nes. September 1986. 87pp. 8609260222. SAND 86-1508.

light water reactors during the six months from Apnl through 38001:186.

September 1985. A method for ranking the safety segrvficance of equipment at nuclear power plants is outlined. The method, which is based HUREG/CR-4673: HEAVY-SECTION STEEL TECHNOLOGY on probabilistic risk assessment techniques, consists of two

} PROGRAM - FIVE YEAR PLAN FY1985-1989. PUGH.C.E. Oak Ridge National Laboratory. Jufy 1986.165pp. 8609100031. .. passes" over plant equipment. The first pass produces a list of ORNL/TM-10108. 37780.020.

plant equipment used to prevent or mitigate intemal events The third in an annual series of five-year program plan docu- ranked quantitatively from most significant to least significant.

ments is presented for the Heavy-Soction Steel Technology pro- The significance measure for each component is based on the gram. The program is camed out by the Oak Ridge National increased risk to the public if the cornponent were not available Laboratory for the Materials Engineenng Branch, Division of En- for operabon. The second pass considers the impacts of special gineenng Technology, Office of Nuclear Regulatory Re earch of emer9encies on the equipment identifed in the first pass. Some the U.S. Nuclear Regulatory Commisson. The program is aimed special emergencies, e.g., earthquakes, are handled with design at advancing the understanding and validation of materials and guidelines. The result of this type of anafysis is a list of plant

\ structures behavior as they related to light water reactor pres- systems and components that need to be designated against a sure-vessel integnty. The program has nine technical tasks and given special emergency. Other special emergencies, e.g., fires, a management functon. A background statement and a plan-of. are handled by ranking the equipment that is required for their action is given for each. The nine technical tasks address frac. rmtigation. The result of this type of analysis is a list of equip-ture methodology and anaytsis, matena!s charactenzation, crack ment ranked according to its significance in preventing or miti-growth, crack arrest, irradiation effects, cladding evaluadons, in. gating special emergencies, a list that can be directly compared termediate-vessel testing, thermal-shock testng, and pressunzed to the list produced in the first pass.

thermal-shock expenments.

NUREG/CR-4683: PBF SEVERE FUEL DAMAGE SCOPING NUREG/CR-4676: SEISMIC REFLECTION 'IEOMETRY OF THE TEST-TEST RESULTS REPORT, KNIPE,A.D.; PLOGER,S.A.;

NEWARK BASIN IN EASTERN PENNSYLVANIA. Evidence For OSETEK,0.J. EG&G Idaho, Inc. (subs. of EG&G, Inc.). August Extensional Reactivation Of Paleozoic Thrust Faults. 1986. 362pp. 8609300386. EGG-2413. 38047:288.

RATCLIFFE.N.M.; BURTON W.C. Intenor Dept. of, Geological Survey. C ANGELO.R.M.; et al. Virginia Polytechrsc Institute & This report represents a comprehensive evaluation of the S te U Blacksburg, VA. July 1986. 27pp. 8608070360. Severe Damage Scoping Test (SFD-ST) performed in the Power Burst Facility (PBF) at the k:aho National Engineering Test Lab-Low-angle dips have been determined for the border fault of oratory. The test is part of an intemationally sponsored light the Newa*. basin near Riegelsville, Pennsylvania, based on VI- waW reacts seme aent ma@ program, maw W h BROSEIS profile and two continuously-cored dnit holes. A group ear Wapbeh h emd was >

of planar reflections in a zone 0.5 km thick in gneiss and car. st of fw, We, rmMuel d tests pedW h W benate mcks of the footwall block coincides with the updip pro- The SFD-ST fuel bundle comprised 32,0.9-m long, tra

>ction of imbncate iault slices and mylonites associated with ated (91 mwd /T) fuel rods. The bundle was surrounded ty w the Musconetcong thrust system of Drake and others (1967). msulahng shrW, we me Won mammed a,t a pressure of 7

- Contrasts in acoustic impedance among mylonitic dolostone MPa. The experiment consisted of a transient in which the inlet and mylonitic gneiss and their protoliths, measured on core Wam now to h test We was WM to 16 g/s, aM me samples, are sufficientfy large to account for these reflectons. McMar power WeaW, unM me peak kmpwaMe ap Analysis of drill core and surface outcrops supports the conclu- proah Nel mew De 3 h transM was tennbaW W sion tnat low-angte extensional faulting in the Early Mesozoic scram of N reactor, with the inlet coolant reflooding and cool-was localized by reactivation of Paleozoic imbricate thrust faults ing N bundle to saturaw tempeaue mn 8 min. De over-in the basement rocks. Extension in the NW SE quadrant was all technical objective of the test was to contnbute to the under-approximately perpendicular to the strike of the ancient thrust standing of fuel bundle dynamics, and the related hydrogen and fautts in Eastern Pennsytvania suggesting a passive origin of the fission product behavior, during a high t6mperature tram %t.

Newark basin here. Data presented provide some of the most The report pcovides a description of the map obsurven pv-explicit three-dimensional information obtained thus far in the nomena. InterpretaM of N test was based upon ms re-Eastem Unced States, in support of the concept of fault reacti- sponse of on-line instruments, posttest fission product sample vation in controlling formation of Early Mesozoic extensional analysis, nondestructive and destructive postirradiation examina-basins. tion of the fuel bundle, and a calculational study using the Severe Core Damage Anafysis Package (SCDAP).

NUREG/CR-4677: TRAP-MELT 2 CODE: DEVELOPMENT AND IM-PROVEMENT OF TRANSPORT MODELING. KUHLMAM,M.R.; NUREG/CR-4644 V01: QUANTIFICATION AND UNCERTAINTY KOGAN,V.; SCHUMACHER P.M. Battelle Memonal Institute, Co- ANALYSIS OF SOUACE TERMS FOR SEVERE ACCIDENTS IN lumbus Laboratories. July 1986. 87pp. 8608070332. BMI-2141. LIGHT WATER REACTORS (OVASAR).Part 1: Methodology 37428:359. And Program Plan. KHATIB-RAHBAR; PARK.C. Brookhaven New developments and improvements to the TRAP-MELT 2 National Laboratory. June 1984. 43pp. 8609290087. BNL-code are desenbed and comparison of the code predictions NUREG-52008. 38012:177.

20 Main Citations and Abstracts The methodological framework and program plan for system- ing direct current output signal is complex and not repeatable ate quantification and propagation o' uncertainities in radiologi- and the component related to ; rack extension cannnot be sep-cal source farms for light water reactors are presented. The arated from the total response. The tests done here show cali-QUASAR methodology is based on detailed sens.tivity analysis brations of a 10kHz altemating current system on an A5338 of the Source Term Code Package (STCP), followed by s sys- material and application of the calibration to a static unloading temMic uncertainty analysis on the most sensitive parameter / compliance test and to rapid servohydraubc tests. A d.c. compo-vanables, and phenomenological issues related to prediction of nent oependent on stress induced magnetization is still present radiological releases to the environment. but can now be separated from the high frequency component NUREG/CR-4694: APPLICATION OF THERMIX-KONVEK CODE using her senes nmMs. W map mcluse is mat an TO ACCIDENT ANALYSES OF MODULAR PEBBLE BED HIGH alternating current technque with a properly chosen excitation frequency can be used to detect crack growth in rapidly loaded TEMPERATURE REACTORS (HTRS) CLEVELAND'J C .

s@ mens. Additional comments on developing an improved GREENE,S.R. Oak Ridge Nationd Laboratory. August 198fl 80pp.8610070262. ORNL/TM-990138131:055. system are presented.

The THERMIX-KONVEK code was used to model the steady state and dynamic thermal behavior of tne U.S. pebble bed NUREG/CR-4702: POST-EOCENE FAULT NEAR EAST EDGE OF REELFOOT RlFT IN LAUDERDALE modular HTR concept and tnal calculations for accident condi-tions were performed. Results of these tnal calculations are COUNTY, TENNESSEE.AS DISCOVERED BY GRAVITY, EARTH RESISTIVITY SURVEYS AND DRILLING.

compared with other predictions by ORNL and by industrial prc-ponents. The basic equations, assumptions, and calculational S TEARNS,R G.; WILSON S.L.; NAVA,S.J. Vanderbilt Univ.,

Nashville, TN. August 1986. 48pp. 8608260277. 37642:256.

technique employed in THERMIX-KNOVEK are presented. Code A search for faults was made along the east edge of Reelfoot validation efforts conducted at Kerntorschungsanlage-Julich Rift using techniques proved at Ree! foot Scarp. Gravity was the (KFA) are summanzed, and the applicability of the code tc the U.S. pebble bed modular HTR is assessed. It was concluded larger area screening device. Narrow elongated residual anoma-lies could mark faults, but whether faults extend to the surface that the code is appliceble to analyses of the safety behavior of the U.S. modular pebble bed HTR. Significant code validation would be unknown. Earth resisitivity profiling was used to has been conducted using out-of-reactor test loops at KFA. Pre- search for abrupt near surface lateral changes in electrical con-ductivity which could indicate a fault within about 200 feet of the dicted results agree well with measured data except for test land surface. This technique led to a precisely located fault conditions leading to asymmetnc behavior which cannot be pre-about 5 miles east of Henning in Lauderdale County, Tennes-dicted with the two-dimensional THERMIX-KONVEK model- sce. Dniting Venfied the fault location, and offset sense Suggestions are made for additional code development includ-ing comparison of code predictions with reactor operating data. (downthrown on the east side). The throw of the fauft was esti-mated to be roughly 100 feet or more from earth resistivity pro-NUREG/CR-4698: AIRBORNE LIDAR MAPPING OF SF6 COM- files, but dnlling indicated only about 40-70 feet of offset.

CENTRATION DISTRIBUTIONS FOR TPANSPORT AND DIF-FUS'ON STUDIES. UTHE,E Er NIELSEN N B.. NUREG/CR-4705: IDENTIFICATION OF NORTHWEST TREND-LIVINGSTON,J.M.; et al. SRI Internationa}. August 1986. b8pp ING SEISMOGENIC GRABEN NEAR CHARLESTON. SOUTH ,

8608260229. 37642:217. CAROLINA. LENNON,G. South Carolina, Univ. of. Columbia, '

This study was decigned to demonstate the capabilities of a SC. August 1986. 53pp. 8608250405. 37635:246.

.lew airborne lidar f ansor for mapping tracer gas distnbution as Shallow stratigraphic dnlling and gamma logging in the meizo-a means of obtaining the necessary model development and seismal area of the Charleston,1896 earthquake were used to validation data, and a real-time method for evaluating distnbu_ map three Tertiary unconformities. Results suggest the pres-tions of radioactive matenals following an accidental release, ence of a northwst extending graben between Charleston and The Airbore Lidar Agent Remote Monitor (ALARM) uses two Kiawah Island, S.C. The 40 km long and 18 km wide graben is downward-pointing, high-energy (1 J), fine-tunable CO2 lasers bounded on three sides by inferred, mappable faults, namely operating in the 9- to 11-um wavelength region. A field demon- the Woodstock, Ashley River and Charleston faults. The stration of the ALARM capabilities for measurement of SF6 was Charleston fault extends northwest from Goose Creek, near designed and conducted with geometnes and conditions similar Charleston through Woodstock and toward Ridgeville. It lies to those that might occur in the event of an accidental radioac- parallel to and a few kilometers to the northeast of a similar tive gas release. Data examples of cross-plume column-content trace developed on the base of the middle Eocene Santee lime-and altitude-resolved SF6 distnbutions are presented. The re, stone reported by Colquhoun and others. The Ashley River fault suits indicate a column content detection of 2 ppb-km, depend. extends northwest from Kiawah Island through Rantowles, S C.

ing on the uniformity of surface differential reflectivity between Sparse gamma-log control Goes not allow for precise location of ALARM wave-length. Examples of cross-p!ume two-dimensional t1e trace. Nevertneless, the zone of tectonic distortion has SF6 distnbutions indicate that an altitude-resolved detection ca- been limited to a line of 3 km width joining centra: Kiawah pability of about 20 ppb can be achieved, limited by the current Island and Rantowies. Releveling data and Fydrogeologic ALARM system nois factors and analysis method uncertainities. anomalies in the Charteston area support these find:ngs. The These values are encouraging, although substantial improve- Ashley River fault and the Charleston fault are projected to land ments are needed in detection capacity to make the technique surface using cross-sections of Pleistocene sediments and geo-valuable for most applications. Recommendations are given for detic data.

needed ALARM techniques improvements.

NUREG/CR-4706. A STUDY OF SEISMICITY AND EARTH-NUREG/CR-4699: APPLICATION OF ALTERNATING CURRENT OUAKE HAZARD IN NORTHERN ALABAMA AND ADJACENT POTENTIAL DIFFERENCE TO CRACK LENGTH MEASURE- PARTS OF TENNESSEE AND GEORGIA. LONG.LT.;

MENT DURING RAPID LOADING. JOYCE,J A.; LIOW,J.S. Georgia Inst:tute of Technology, Atlanta. GA.

• Ala-SCHNIEDER,C.S. U.S. Naval Academy, Annapolis, MD. August bama, Stato of. August 19P6. 71pp. 8608270037. 37655:248.

1986. 46pp. 8609120248. 37815:195. The report summarizes network operations and related re-The objectrve of this project has been to investigate the appli- search performed by the Geological Survey of Alabama and cability of an alternating current potential difference method of Georgia institute of Technology, and descnbes special studies crack length estimation to rapid loading fracture mechanics of crustal thickness in southeastern Tennessee and of the tests in ferromagnetic materials. The more commonly used North-Georgia earthauake of October 9,1984. Travel times direct current method has been demonstratea to be very senti from deep-focus events and refraction based on quarry blasts tive to induced mpeteation under rapid loadings. The result- were used to study custal velocity. Data indicate a coherent ve-

Main Citations and Abstracts 21 locity structure for the granitic crust in the southeastem U.S. of one, in boreholes in basalt blocks and in steel pipes. Expan-Refraction velocities and separate solutions for location and sion strains and cunng temperatures have been monitored on depth were used to improve travel-time calculations and hypo- cement plugs in boreholes in basalt blocks, in PVC and in steel center locations. Analysis of crusta! thickness places the cmst- pipes with diameters from 25.4 mm to 196.9 mm and length to mantle boundary at an average depth of 55 km. Comparsions of diameter ratios of one and two. Dunng permeability tests, basalt reflections with the direct S-wave and synthetic seismograms blocks have fractured, presumably due to water injection pres-show that the thickness of a transition zone would have to be sure, cement expansion and packer pressure. Falling head tests less than 1 km. The mb=3.5 earthquake of October 9,1984 in- have been performed on some block fractures to study the in-dicated a right-tateral slip on a near-vertical, north-trending fluence of the complicated interaction between a cement bore-plane. A stress drop of 75 bars was computed for a moment of hole plug (e.g. swelling and shrinkage alterations) and the rock, 3.5 x 10(23) dyne-cm. Only one significant aftershock was de- as well as the normal stress across the fracture, on the hydrau-tected. lic conductivity of a fracture intersecting a plugged borehole.

NUREG/CR-4707: A PREUMINARY GEOLOGIC EVALUATION The hydraulic conductivity of the cement plugs in the steel OF THE ALABAMA-TENNESSEE TRANSVERSE SEISMIC pipes varies between 3.57 x 10 (-11) cm/ min and 3.65 x 10 (-9)

ZONE IN ALABAMA. NEATHERY,T.L; OSBORNE W.E.; cm/ min. Cement swelling tests remain inconclusive about size SZABO,M.W. Alabama. State of. August 1986. 59pp. effects, pnmarily of instumentation problems. Cement cunng 8609050428. 37736:200. temperatures increase from small to large diameter cement Historically Alabama is a region of moderate seismic activity. plugs.

Since 1886 the state has experienced 65 recorded earthquakes NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER whose epicenters ure in or very near the state. Data obtained from a recentiy installed seven seismic-station network indicat- CANNON AND HDR EXPERIMENTAL DATA. NEUMANN U.

ed that many > or minus til Mercalh scale earthquakes are oc-Kraftwerk Union Aktiengesellschaft. August 1986. 223pp.

cumng in north Alabama. The quakes occur within two principal 8608270424.37659:138 ap tih m d N T o seismic zones. One zone, the Southern Appalachian Seismic Zone, is associated with the tectonic grain of the Appalachian sient Reactor Analysis Code (TRAC-PD2) using data from the Mountain system and has been recognized for years. Another SUPER CANON and HEISS DAMPF g REACTOR (HDR)g, expe,r pg somewhat smaller seismic zone that extends from Montgomery, Port p Alabama to Nashville, Tennessee (Alabama-Tennessee Seismic the TRAC-PD2 assessment using the SUPER CANON data.

Zone) was identified only through recently interpreted seismic Fart ll is the TRAC-PD2 assessment using HDR data. Part til data. No surface manifestation of the earthquake activity in the provides recommendations for the user using the combined as-Alabama-Tennessee Seismic Zone has been identified, but geo. sessment results. In general, it is shown that the TRAC-PD2 physical, lineament, and joint data suggest possibio structures predictions were in good agreement with the actual test pres-which may be related to previous carthquake activity. sures and mass flow rates for both these tests. TRAC-PD2 pro-vided considerably better results than TRAC-P1 A. This was par-NUREG/CR-4725: TWO-PHASE FLOW MEASUREMENT IN THE ticularly true with regard to sound velocity predictions which play UPPER PLENUM OF A PWR DURING REFLOOD. GAWLIK K.; a significant role whenever the speed of pressure relief waves GRIFFITH,P. Massachusetts Institute of Technology, Cam- must be determined.

bridge, MA. August 1986. 99pp. 8609150252. 37834:253.

In order to study the two-phase hydrodynamics of a pressur- NUREG/lA-0002: HEAT TRANSFER PROCESSES DURING IN-ized water reactor curing emergency reflood, a group of four in- TERMEDIATE AND LARGE BREAK LOCAS. VOJTEK,1. Reac-struments has been developed for installation at the core and tor Safety Co. (GRS), Inc. September 1986. 160pp.

upper plenum interface. These instruments are a drag th.1y, a 8610030403.37680:243.

cut-out portion of the end-box tie plate; a turbine meter located The general purpose of this project was the investigation of above one of the tie-piste holes; a differential pressure meas- the heat transfer regimes during the high pressure portion of urement across the tie plate; and a collapsed liquid level meas- blowdown. The main attention has been focussed on the eval-urement above the tie plate. A single-module air and water ap- uation of those phenomena which are most important in reactor paratus simulates the PWR during reflood. Tests are done with safety, such as maximum and minimum critical heat flux and the upper plenum empty, and with it containing simulated con- forced convection film boiling heat transfer. The experimental trol rod guide tubes and baffle plates. The fiow regimes studied results of the 25-rod bund'e blowndown heaa transfer tests, are flood and unflood countercurrent flow, up",ow, and com- which were performed at the KWU heat transfer test facility in bined injection. The upper plenum internals have a significant Karlstein, were used as a database for the verification of differ-effect on the behavior of the two-phase flow. An algorithm to ent correlations which are used or were developed for the anal-determine the individual mass flowrates of gas and liquid is pre- ysis of reactor safety problems. The computer code BRUDI-VA sented. This algorithm depends on the drag body force meas- was used for the ca!culation of local values of important ther-urement, the differential pressure measurement, the collapsed mohydraulic parameters in the bundles.

liquid level, and the flooding line.

NUREG/IA-0003: INFLUENCE OF THE WETTING STATE OF A NUREG/CR-4729: DROPLET ENTRAINMENT IN VERTICAL AN. HEATED SURFACE ON HEAT TRANSFER AND PRESSURE NULAR FLOW AND ITS CONTRIBUTION TO MOMENTUM LOSS IN AN EVAPORATOR TUBE. KOHLER.W.; HEIN D.

TRANSFER. LOPES J.C.; DUKLER,A.E. Houston, Univ. of, Kraftwerk Union Aktiengesellschaft. September 1986. 130pp.

Houston..TX. September 1986. 55pp. 8609300382. 38050:045.

86 0 030406. 37682:001 Simultaneous measurements were made of the size, axial and The influence of the wetting state of a heated surface on radial velocity of drops entrained by the gas in annular flow. A heat transfer and pressure loss in an evaporator tube was in-model es developed to use these data to compare the rate of vestigated for a parameter range occurring in fossil-fired steam deposrtion or entrainment and the pressure gradient due to drop generators. Included in the analysis are quantities which deter-interchange.

mne the wetting state in steady and transient flow. Based on "JUREG/CR-4738: SIZ2 INFLUENCE ON THE SEALING PER- the expuiments performed a method of predicting ChF for a FORMANCE OF CEMENTITIOUS BOREHOLE PLUGS. vertical upflow evaporator tube was developed. The heat trans-AKGUN H.; DAEMEN.J.J. Arizona, Univ of, Tucson, AZ. Sep- fer in the unwetted region was newly formulated taking into ac-tember 1986. 319pp. 8610030405. 37681:043. cour't thermal nonequilibrium between the water and steam Flow tests have been conducted on cement pleas with diam- phases. Wall temperature excursions during pressure and en-eters of 158.8 mm and 196.9 mm, and length to diameter ratios thalpy transients are interpreted with the help of the boiling l

22 Main Citations and Abstracts curve and the Leidenfrost phenomenon. A method is developed lations and the experiment have been quantr%d over intervals by means of which it is possible to determine the influencer of in real time for a number of variables available from the meas-the pipe orientation on the location of the boiling cnsis as well urements dunng the expenment. The core inventory expressed as on the heat transfer in the unwetted region. The influence of by the differential pressure over the core was ger,erally under-the wetting state of the heated surface on the two phase flow predicted. Dryout times were generally underpredicted, probably pressure loss is interpreted as " Wall effect" and is calculated due to differences in the used dryout correlation.

using a simphfied computer model.

NUREG/lA4005: ASSESSMENT OF RELAP/ MOD 2, CYCLE NUREG/lA-0007: ASSESSMENT OF RELAPS/ MOD 2 AGAINST

] CRITICAL FLOW DATA FROM TESTS JIT 11 AND CFT 21.

J 36 AGAINST FIX-Il SPLIT BREAK EXPERIMENT NO. 3027.

ERIKSSON.J. Sweden, Govt. of. September 1986. 103pp. ROSDAHL,0.; CARAHER,0. Sweden. Govt. of. September 8610070255. 38129:309. 1986. 62pp. 8610030413. 37678:289.

The FIX-Il split break experiment No. 3027 has been analyzed RELAPS/ MOD 2 simulations of the entcal flow of saturated using the RELAPS/ Mod 2 code. The code version used, Cycle steam are reported together with simulatx>ns of the critical flow 36, is a frozen version of the code. Four different prediction cal- of subcooled liquid and low-quahty two-phase mixture. The ex-culations were carried out to study the sensitmty on vanous pa- periments which were simulated used nozzle diameters of 0.3 m rameters to changes of break dr, charge, initial coolant mass, and 0.5. RELAPS overpredicted the expenmental flow rates by and passive heat structures. The differences between the calcu- 10 to 25 percent unless discharge coefficients were applied.

l 1

l l

1 l Contractor Report Number index 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 NUMSER REPORT NUMSER SECONDARY REPORT NUM8ER REPORT NUMBER ANL-86-31 NUREG/CR-4667 V01 ORNU TM-9593/V3 NUREG/CR-4219 V03 N1 BMI-2120 NUREG/CR-4082 V04 ORNL/TM-9614 V4 NUREG/CR-4236 BMi-2136 NUREG/CR-4574 ORNUTM-9632 V3 NUREG/CR4255 V03 N1 BMi-2137 BMI-2138 NUREG/CR4575 NUREG/CR-4587 $UT ORNL/TM-9905 9

NUREG/CR-4694 BMI-2140 NUREG/CR4658 BMI-2141 ORNUTM-9956 NUREG/CR 3967 NUREG/CR-4677 PNL-4297 NUREG/CR-2800 SO4 BNL-NUREG-51454 NUREG/CR-2331 V05 N4 PNL-4804 BNL-NUREG-51581 NUREG/CR-4162 NUREG/CR-2907 V04 PNL-4965-10 NUREG/CR-4860 BNL-NUREG-51998 NUREG/CR-462g PNL-5224 NUREG/CR 3974 BNL-NUREG-52003 NUREG/CR4643 PNL-5270 NUREG/CR4013 BNL-NUREG-52004 NUREG/CR-4644 PNL-5374 NURE G/CR4125 V01 R1 BNL-NUREG-52007 NUREG/CR-4659 V01 PNL-5374 NURE;;G/CR-4125 V02 R1 BNL NUREG-52008 EGG-2413 EGG-2447 NUREGICR-4888 V01 NUREG/CR-4683 NUREG/CR-4529 PNL 5627 hhhh03 M NUREG/CR4495 EGG-2448 NUREG/CR-4553 PNL-5724 NUREG/CR-4480 EGG-2450 NUREG/CR-4562 PNL-5757 NUREG/CR4600 EGG-2453 PNL-5839 NUREG/CR-4593 NUREG/CR4445 PNL-5875 EGG-2459 NUREG/CR-4648 NUREG/CR4622 HEDL TME 86-8 PNL-5878 NUREG/CR-4626 V01 NUREG/CR-3320 V01 SAND 84-1586 NUREG/CR 3925 LA-10157 MS NUREG/CR-3858 SAND 84-2632 ORNL-6208/R1 NUREG/CR-4099 NUREG/CR-4367 RO1 SAND 844258 NUREG/CR-4102 ORNL4282/V1 NUREG/CR-4597 VOI SAND 85-1370 NUREG/CR-4308 ORNL4291 NUREG/CR-4633 SAND 85-1606 NUREG/CR-4340 V02 ORNUNOAC-230 NUREG/CR-4470 SAND 851739 NUREG/CR4558 ORNUNSIC-200 NUREG/CR-2000 VOS N6 SAND 85-2795 NUREG/CR-4487 ORNUNSC200 NUREG/CR4000 VOS N7 SAND 85-7207 NUREG/CR4336 ORNUNSG200 S ORNUTM 10001 OANL/TM 10305 NUREG/CR-2000 VOS N8 NUREG/CR-4560 NUREG/CR-4566 S$A M SAND 86 0366 hhN$

NUPEG/CR-4530 V01 ORNUTM-10095 NUREG/CR-4647 SAND 864394 NUREG/CR-4596 ORNL/TM-10099 SAND 86 0450 NUREG/CR-4536 NUREG/CR-4649 SAND 86-0960 ORNUTM-10108 NUREG/CR-4598 NUREG/CR-4673 SAND 86-1508 NUREG/CR-4878 ORNUTM-10132 NUREG/CR-4439 SAND 66 7080 NUREG/CR4533 23

- - --- a -*..a -

a. .__e - m e * - .-

(

I I

i

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

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

NUREG number and the title of the ADAMS,M.L BEFDLOW,P.A.

NUREG/CR4597 V01: AGING AND SERVICE WEAR OF AUXILIARY NUREG/CR-4480: EROSION PROTECTON OF URANIUM TAluNGS IM-FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol. POUNDMENTS.

1. Operating Expenenw And Failure Identificahon.

BENNETT,P.R.

ADAMS,R E.

NUREG/CR-4536: SUFERHEATED STEAM TEST OF ETHYLENE PRO.

NUREG/CR4255 V03 N1: AEROSAL RELEASE AND TRANSPORT PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER ACCIDENT ENVIRONMENT.

1985 MARCH 1986.

AHMAD,J. BERLINGER,C.

NUREG/CR4082 V04: DEGRADED PIPING PROGRAM - PHASE NUREG.1216: SAFETY EVALUATION REPORT RELATED TO THE li Sermannual Report. October 1985 + March 1986. OPERABluTY AND RELIABluTY OF EMERGENCY DIESEL GENERA-TORC MANUFACTURED BY TRANSAMERICA DELAVAL,1NC.

AKGUN.H.

NUREG/CR4738: SIZE INFLUENCE ON THE SEAUNG PERFORM. BLENCOFJ.G.

ANCE OF CEMENTITIOUS BOREHOLE PLUGS. NUREG/CR 4236: PROGRESS IN EVALUATION OF RADIONUCLIDE ALEXANDER,C.A. GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPCS4 TORY SITE PROJECTS. Annual Report For NUREG/CR4656: VERIFICATION TEST CALCULATIONS FOR THE October 1984 -September 1985.

SOURCE TERM CODE PACKAGE.

ALEXANDER,R.E.

NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPUCATIONS WEG'R3925: Sm n SWEONG MQNAsWah TO BIOASSAY AT URANIUM MILLS. Probierns For Sandia Waste-Isolation Flow And Transport Model For Fractured Media.

ANASTASI F.

NUREG/CR-3967; AN ANALYSIS OF EXCURSIONS AT SELECTED IN BONZON,LL SITU URANIUM MINES IN WYOMING AND TEXAS. NUREG/CR-4099: AGE RELATED DEGRADATON OF NATURALLY-AGED CLASS 1E BATTERY CELLS.

APOSTOLAKIS.G.

NUREG/CR.4568: COMPORN lli - A COMPUTER CODE FOR MODEL. BRADLEY,D.R.

ING COMPARTMENT FIRES. NUREG/CR-4558: INTERACTION OF HOT SOLID CORE DEBRIS WITH ARENDTS,J.G. CONCRETE.

NUREG/CR-4529: PIPING DAMP!NG - EXPERIMENTAL RESULTS BROCKMAN,J.E.

FROM LABORATORY TESTS IN THE SEISMIC RANGE. NUREG/CR-4308: VANESA;A MECHANISTIC MODEL OF RADIONU-ARNOLD.W.D. CUDE RELEASE AND AEROSOL GENERATON DURING CORE NUREG/CRJ236: PROGRESS IN EVALUATION OF RADIONUCUDE DEBRIS INTERACTIONS WITH CONCRETE.

GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL BROCKMA%N,J.E.

UCLEAR WAS R PO Y SITE PROJECTS. Annual Report For

~

NUREG/CR4521: TURC2 AND 3tARGE SCALE UO2/ZRO2 MELT.

CONCRETE INTERACTION EXPERIMENTS AND ANALYSIS.

BADALAMENTE,R.

NUREG/CR-4600: HUMAN FACTORS STUDY CONr9CTED IN CCN- BRODSKY,A.

JUNCTION WITH A MINI-ROUND ROBIN ASSESS.eENT OF ULTFsA. NUREG.0874: INTERNAL DOSIMETRY MODEL FOR APPUCATIONS SONIC TECHNICIAN PERFORMANCE. TO BIOASSAY AT URANIUM MILLS.

BALL.S.J. BRUST,F.

NUREG/CR-4265 V02: AN ASSESSMENT OF THE SAFETY IMPUCA. NUREG/CR-4082 V04: DEGRADED PIPING PROGRAM - PHASE TIONS OF CONTROL AT THE CALVERT CUFFS 1 NUCLEAR ll.Sermannual Report October 1985 - March 1986.

PLANT. Volume 2: Appendices. NUREG/CR-4574: AN EXPERIMENTAL AND ANALYT! CAL ASSESS.

BANDYOPADHYAY MENT OF CIRCUMFERENT:AL THROUGH. WALL CRACKED PIPES UNDER PURE BENDING.

NUREG/CR-4659 V01: SEISMIC FRAGluTY OF NUCLEAR POWER PLANT COMPOrdNTS (PHASE 1). BRYSON,J.W.

BARADAR,AN,R* NUREG/CR4367 Rot: ORVIRT.PC (VERSION 2.0):A 2-D FINITE-ELE-NUREG CR4643: EVALUATION OF CORE DAMAGE SEQUENCES INI- MENT FRACTURE ANALYSIS PROGRAM FOR A MICROCOMPUTER TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOUNG. NUREC 'OR4633: ORPLOT.PC:A GRAPHIC UTIUTY FOR ORMGEN.PC AND ORVIRT.PC.

BARNES.C.R BUCKALEW,W.H.

NUREG/CN4082 V04: DEGRADED PIPING PROGRAM - PHASC fl.Sermannual Report, October 1985 March 1986. NUREG/CR4530 Vol: U.S/ FRENCH JOINT RESEARCH PROGRAM REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-BEEBE.M.R. JECTED TO EETA RADIATON.Vol 1: Phase-1 Normalizabon Results.

NUREG4020 V10 N05: UCENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of April 30.1986(Gray Book) BUNNELL.LR.

NUREC-0020 V10 N06: UCENSED OPERATING REACTORS STATUS NUREG/CR-4495 VISCOSITY OF ZlRCONIUM. URANIUM CXIDE (Zr.

SUMMARY

REPORT. Data As Of May 31,1986.(Gray Book) UO2) MIXTURES AT 1800 TO 2100 C.

25

26 Personal Author index l

i BURTON,W.C. NUREG/CR-4624 V05: RADIONUCUDE RELEASE CALCULATIOl43 WUREG/CR-4676: SEISMIC REFLECTON GEOMETRY OF THE FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5 Large NEWARK BASIN IN EASTERN PENNSYLVANIA. Evidence For Exten- Dry Containment Design.

saonal Reactivation Of Paleozoic Thrust Faults. NUREG/CR-4656: VERIFtATON TEST CALCULATONS FOR THE SOURCE TERM CODE PACKAGE.

NUREG/CR-4560- RISK ASSESSMENT APPLICATION TO NRC IN- CWAUNA,G.

SPECTOA' PROGRESS REPORT FO", PEROD JANUARY 1985 TO NUREG-121a -/01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR JANUARY 1986. POWER INDUSTRY 1985. Volume 1 Findings And Conclusons.

NUREG-1212 V02: STATUS OF MAINTENANCE IN U.S. NUCLEAR IN-G/ 007: ASSESSMENT OF RELAPS/ MOD 2 AGAINST CRITI-CAL FLOW DATA FROM TESTS JIT 11 AND CFT 21. CYBULSKIS P.

CASADA,M.L NUREG/CR-4587: SOURCE TERM CODE PACKAGE A USER'S GUIDE NUREG/CR-4470: SURVEY AND EVALUATON OF VITAL INSTRUMEN-N / -Ml24 VOI: RADONUCUDE RELEASE CALCUt ATIONS TATION AND CONTROL POWER SUPPLY EVENTS.

FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 1:BWR, Mark i Design.

CASE.J.T' NUREG/ CR-4445: EFFECTIVENESS AND SAFETY ASPECTS NUREG/CR-4624 OF V02- SE.RADIONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2.PWR,1ce LECTED DECONTAMINATON PROCESSES.

Condenser Design.

CHANG,C.Y. NUREG/CR 4624 V03: RADIONUCUDE RELEASE CALCULATONS NUREG/CR-3805 V04: ENGINEERING CHARACTERIZATON OF FOR SELECTED SEVERE ACCIDENT SCENAROS.Volumo GROUND MOTION. Task 11: So4-Structure Interaction Effects On Struc- 3,PWR.Subatmosphenc Containment Design.

tural Response. NUf1EG/CR-4624 V04: RADIONUCLIDE RELEASE CALCULATIONS NUREG/GR-3805 V05: ENGINEERING CHARACTERIZATION OF FOR SELECTED SEVERE ACCOENT SCENAROS. Volume GROUND MOTON. Task 11: Summary Report.

4:BWR. Mark lil NUREG/CR 4624  : RADONUCUDE RELEASE CALCULATONS CHIAPETTAAL FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5:Larir NUREG/CR-4336- REVIEW OF llT RESEARCH INSTITUTE REPORTS Dry Containment Design.

ON STRUCTURAL STUDIES OF REACTOR CONTAINMENT. NUREG/CR-4656: VERIFICATON TEST CALCULATONS FOR THE SOURCE TERM CODE PACKAGE.

CHRISTENSEN.D.

NUREG/CR-3145 V04: GEOPHYSICAL INVESTIGATIONS OF WEST- D'ANGELOAM ERN OHIO INDIANA REGON. Annual Report (October 1984 - Septem- NUREG/CR-4676: SEISMIC REFLECTON GEOMETRY OF THE ber 1985b NEWARK BASIN IN EASTERN PENNSYLVANIA. Evidence For Exten.

CLEVELAND,J.C. sional Reactivation Of Paleozoic Thrust Faults.

NUREG/CH-4694. APPUCATON OF THERMIX-KONVEK CODE TO AC-C DEN YSE OF MODULAR PEBBLE BED HIGH TEMPERA-NU EG C l-4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

CUNTON,J.H.

NUREG/CR-4644: GEOCHEMICAL STUDIES OF COMMERCIAL LOW- DAUNO.P.M.

LEVEL RADCACTIVE WASTE DISPOSAL SITES. Topical Report. NUREG/CR-4626 V01: IMPROVING THE RELIABluTY OF OPEN-CYCLC WATER SYSTEMS.An Evaluation Of Biofouling Surveillance CONKUN J.C. And Control Techniques For Use At Nuclear Power Plants.

NUREG/CR-4649: SCAUNG ANALYSIS OF THE COUPLED HEAT TRANSFER PROCESS IN THE HIGH TEMPERATURE GAS COOLED DANIEL.S.H.

REACTOR CORE- NUREG/CR-4553: FIELD TESTS OF A PORTABLE TISSUE LiOUIVA-LENT SURVEY METER FOR MONITORING MIXED BETA /3AMMA COPUS.E.R. RADIATON FIELDS.

NUREG/CR-4558: INTERACTION OF HOT SOUD CORE DEBRIS WITH CONCRETE. DAVIS.P.A.

NUREG/CR-3925: SWIFT 11 SELF TEACHING CURRICULUM.fllustrative COSTAIN.J.K. Problems For Sandia Waste-Isolation Flow And Transport Model For NUREG/CR-4676: SEISMIC REFLECTON GEOMETRY OF THE Fractured Media.

NEWARK BASIN IN EASTERN PENNSYLVANIA. Evidence For Exten-sional Reactivation Of Paleozoic Thrust Faults. DAYAL,R.

NUREG/CR-4644: GEOCHEMICAL STUDIES OF COMMERCIAL LOW-

^

NU E'GICR-4628: RELATIVE AGE-SPECIFIC RADIATION DOCE COM-MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED DEBELUS,0.E.

FROM NUCLEAR FUEL FACIUTIES. NUREG/CR-3262 V04: COBRA-NCA THERMAL HYDRAUUCS CODE CUELLO,R. FOR TRANSIENT ANALYSIS OF NUCLEAR PEACTOR NUREG/CR-3974; GEOMEMBRANE SELECTION CRITERIA FOR URA- COMPONENTS.Vol 4. Users' Manual For Containment Analysis.

NIUM MILL TAluNGS PONDS.

DEf#llNG A CURTIS,L NUREG/CR-4624 V03: RADIONUCUDE RELEASE CALCULATONS NUREG/CR-4587: SOURCE TERM CODE PACKAGE:A USER'S GUIDE FOR SELECTED SEVERE AOC1 DENT SCENARIOS. Volume (god j). 3PWR,Subatmosphenc Cow Design.

CURTIS.LA. DEMNINGAS.

NUREG/CR4624 V01: RADIONUCUDE RELEASE CALCULATIONS NCRJG/CR-4624 V01: RADIONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACCIDENT SCENARIOS.Vobme FOR SELECTED SEVERE ACCCENT SCENAROS. Volume 1:BWR.Mwk i Design.

1:BWR. Mark i Design.

NUREG/CR-4624 V02: RADIONUCUDE RELEASE CALCULATONS NUREG'CR4624 V02: RADIONUCLIDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCOENT SCENAROS. Volume IPWR.lce FOR SELECTED SEVEHE ACCIDENT SCENARIOS. Volume 2:PWR,lce Condenser Design. Condenser Design.

NUREG/CR-4624 V03: RADIONUCUDE RELEASE CALCULATIONS NUREG/CR-4624 V04: RADIONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 3 PWR,Subatmospheric Containment Design. 4 BWR. Mark lli Design.

NUREG/CR-4624 v04: RADIONUCUDE RELEASE CALCULATIONS NUREG/CR 4624 VOS: RADIONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume FOR SELECTED SEVERE ACCIDENT SCENAROS. Volume 5Large 4.BWR, Marx ill Design. Dry Contarvnent Design.

l Personal Author index 27 NUREG/CR-4656: VERIFICATON TEST CALCULATIONS FOR THE GELHAR,LW.

SOURCE TERM CODE PACKAGE.

NUREG/CR-4622: VAUDATICN OF STOCHASTIC FLOW AND TRANS-DEWALL,K.G. PORT MODELS FOR UNSATURATED SOILS.A Comprehensive Field Study.

NUREG/CR-4648: A STUDY OF TYPICAL NUCLEAR CONTAINMENT PURGE VALVES IN AN ACCIDENT ENVIRONMENT. GIESEKE J.A.

DUCE S.W NUREG/CR-4624 V01: RADICNUCUDE RELEASE CALCULATIONS NUREG/CR-4445: EFFl!CTIVENESS AND SAFETY ASPECTS OF SE- FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume LECTED DECONTAMINATION PROCESSES.

NU EG -4624 02$ RADIONUCUDE RELEASE CALCULATONS CUKLER,A.E.

FOR SELECTED SEVERE ACCOENT SCENAROS. Volume 2:PWR lce NUREG/CR-4729: DROPLET ENTRAINMENT IN VERTICAL ANNULAR Condenser Design.

FLOW AND ITS CONTRIBUTON TO MOMENTUM TRANSFER. NUREG/CH-4624 V03: RADIONUCUDE RELEASE CALCULATONS DUNKELMAN.M.M.

FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 3:PWR.Substmosphenc Containment Design.

NUREG-1213: PLANS AND SCHEDULES FOR IMPLEMENTATON OF NUREG/CR-4624 V04: RADIONUCUDE RELEASE CALCULATONS U.S. NUCLEAR REGULATORY COMMISSON RESPONSl81UTIES FOR SELECTED SEVERE ACCIDENT SCENAROS. Volume UNDER THE LOW-LEVEL RADIOACTIVE WASTE POUCY AMEND- 4DWA. Mark lli Desen.

MENTS ACT CF 1985 (PL 99-240). NUREG/CR-4624 V05: RADIONUCUDE RELEASE CALCULATONS DUNNING,DE FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 51arge Dry Contamment Desegn.

NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATION DOSE COM-j MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED NUREG/CR-4656: VERIFICATON TEST CALCULATONS FOR THE FROM NUCLEAR FUEL FACluTIES. SOURCE TERM CODE PACKAGE.

DYCUS.F.M. **U*

NUREG/CR 4560 RISK ASSESSMENT APPUCATION TO NRC IN- NUREG/CR-4536: SUPERHEATED-STEAM TEST OF ETHYLENE PRO-SPECTON PROGRESS REPORT FOR PERIOD JANUARY 1965 TO PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND JANUARY 1986. ACCIDENT ENVIRONMENT.

E8 ERHARDT,LL GME,aJ.

NUREG/CR-4162: SURVEY OF STATISTICAL AND SAMPUNG NEEDS NUREG/CR-3585 V02- DE MINIMIS WASTE IMPACTS ANALYSIS FOR ENVIRONMENTAL MONITORING OF COMMERCIAL LOW. METHODOLOGY. Volume 2. Impacts-BRC Users Guide And Methodol-LEVEL RADIOACTIVE WASTE DISPOSAL FACluTIES. ogy For Radioactive Wastes Below Regulatory Concern.This is An ECKERMAN,K.F. NRC Staff Report Pubhshed As Volume 2 Of Contractor Report.

NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATION DOSE COM. GREENE,S.H.

MITMENT FACTORS FOR MAJOR RADONUCUDES RELEASED NUREG/CR-4694: APPUCATON OF THERMIX-KONVEK CODE TO AC-FROM NUCLEAR FUEL FACIUTIES.

CIDENT ANALYSES OF MODULAR PEB8LE BED HIGH TEMPERA-TURE REACTORS (HTRS).

NUREG/CR-4560 RISK ASSESSMENT APPUCATION TO NRC IN- GRENIER,5.

SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO NUREG-1212 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR JANUARY 1986.

POWER INDUSTRY 1985. Volume 1Frengs And Conclusions.

NUREG.1212 V02- STATUS OF MAINTENANCE IN U.S. NUCLEAR IN-U l 0005: ASSESSMENT OF RELAP/ MOD 2 CYCLE 36,AGAINST "" "

FIX-l! SPUT BREAK EXPERIMENT NO. 3027. gn:ESEKE,J.

FEINER.F. NUREG/CR-4587: SOURCE TERM CODE PACKAGE:A USER'S GUIDE NUREG/CP0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON (MOD 4 REACTOR PHYSICS AND SAFETY. Sessions 1-10.

NUREG/CP-0080 V02: PROCEEDINGS OF THE TOPICAL MEETING ON GRAFFITH,P.

REACTOR PHYSICS AND SAFETY. Sessions 1116. NUREG/CR-4725: TWO-PHASE FLOW MEASUREMENT i.e THE UPPER FLANAGAN,G.F. PLENUM OF A PWR DURING REFLOOD.

NUREG/CR-4560: RISK ASSESSMENT APPLICATION TO NRC IN. GRONAGER,Ji SPECTON PROGRESS REPORT FOR PERIOD JANUARY 1985 TO NUREG/CR-4521: TURC2 AND 3:LARGE SCALE UO2/ZRO2 MELT.

JANUARY 1986.

CONCRETE INTERACTON EXPERIMENTS AND ANALYSIS.

NUREG/CR-4566: COMP 8RN lif - A COMPUTER CODE FOR MODEL.

ING COMPARTMENT FIRES. GUERRIERI D.

NUREG/CR-4082 V04: DEGRADED PIPING PROGRAM PHASE

' ~

REG CR-4480: EROSON PROTECTON OF URANIUM TAluNGS IM-POUNDMENTS. GUIDOTTI,T.E.

FORSTROMJ'M NUREG/CR-3262 V04: COBRA-NC-A THERMAL HYDRAUUCS CODE NUREG/CR-3585 V02: DE MINIMIS WASTE IMPACTS ANALYSIS FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR METHODOLOGY. Volume 2. Impacts-BRC Users Guide And Methodol- COMPONENTS.Vol. 4.Usus' Manuel For Conta nment Analyers.

ogy For Radioactrve Wastes Below Regulatory Concern.This is An GUTHRIE,V.H.

NRC Slaff Report Published As Volume 2 Of Contractor Report NUREG/CR-4560: RISK ASSESSMFNT APPUCATON TO NRC IN-GALLUP,D.R.

SPECTON PROGRESS REPORT FOR PEROD JANUARY 1965 TO NUREG/CR-4678. A METHOO FOR USING PRA TO ESTABUSH QUAL. JANUARY 1986.

(TY PROGRAM APPUCA81UTY.

HACKETT.E.M.

GAWLIK,K.

NUREG/CR-4540: AN EVALUATON OF J-R CURVE TESTING OF NU-NUREG/CR-4725: TWO. PHASE FLOW MEASUREMENT IN THE UPPn CLEAR PIPING MATERIALS USING THE DIRECT CURRENT POTEN-PLENUM OF A PWR DURING REFLOOD. TIAL DROP TECHNOUE.

GE E,0.W. HALVERSON,M A.

NUREG/CR-4622: VAUDATON OF STOCHASTIC FLOW AND TRANS- NUREG/CR-4593: INITIAL CONCEPTS ON ENERGETICS & MASS RE-PORT MODELS FOR UN3ATURATED SOILS.A Comprehensive Field Study. LEASES DURING NONNUCLEAR EXPLOSIVE EVENTS IN FUEL CYCLE FACIUTIES.

28 Personal Author Index HANES LF. NUREG-1212 V02: STATUS OF MAINTENANCE IN U.S. NUCLEAR IN-NUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE- DUSTRY 1985. Volume 2:Descnptons Of Programs And Practices.

AR POWER PLANT PERSONNELA Feashhty Study. Volume 1: Sum-mary Of Results. JOHNSON,J.D.

NUREG/CR-45321,02: MODELS OF COGNITIVE BEHAVIOR IN NUCLE- NUREG/CR-4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A AR POWER PLANT PERSONNEL-A Feasibility Study. Volume 2. Main DRY CONTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL Report. MODEL HAYS.R A. JOHNSON K L NUREG/CR-4540 AN EVALUATON OF J-R CURVE TESTING OF NU- NUREG/CR-4626 V01: IMPROVING THE REUABlUTY OF OPEN-CLEAR PIPING MATERIALS USING THE DIRECT CURRENT POTEN- CYCLE WATER SYSTEMS.An EvaluatK,n Of Biofouling Surveillance TIAL DROP TECHNIQUE. And Control Techniques For Use At Nuclear Power Plants.

HAZELTON,W.S.

JOHNSON.LO.

NUREG-0313 R02 DRFT: TECHNICAL REPORT ON MATERIAL SELEC- NUREG/CR-4553: FIELD TESTS OF A PORTABLE TISSUE EQUIVA-TION AND PROCESSING GUIDEUNES FOR BWR COOLANT PRES- LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA SURE BOUNDARY PIPING' RADIATION FIELDS.

SON,EB.

REG /IA-0J03. INFLUENCE OF THE WETTING STATE OF A HEATED NUREG/CR-4690: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA-SURFACE ON HEAT TRANSFER AND PRESSURE LOSS IN AN TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES.

EVAPORATOR TUBE.

HELTON J.C. JORDAN,H.

NUREG/CR-4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A NUREG/CR-4587: SOURCE TERM CODE PACKAGE:A USER'S GUIDE DR NTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL N / -4624 V01: RADIONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume HILL,R.C. 1:BWR. Mark i Design.

NUREG/CR-4648: A ST'JDY OF TYPICAL NUCLEAR CONTAINMENT NUREG/CR4624 V02: RADIONUCUDE RELEASE CALCULAT.ONS PURGE VALVES IN AN ACCIDENT ENVIRONMENT. FOR SELECTED SEVEhE ACCOENT SCENAROS. Volume 2:PWR,lce Condenser Design.

HINKLE,N.E- NUREG/CR-4624 V03: RADONUCUDE RELEASE CALCULATIONS NUREG/CR-3967: AN ANALYSIS OF EXCURSIONS AT SELECTED IN FOR SELECTED SEVERE ACCIDENT SCENARIOS; Volume SITU URANIUM MINES IN WYOMING AND TEXAS- 3.PWR.Subatmosphenc Containment Design.

NUREG/CR-4624 V04: RADIONUCUDE RELEASE CALCULATIONS N'UREG/CR-4566: COMP 8RN lli - A COMPUTER CODE FOR MODEL- BWR.M Design.

ING COMPARTMENT FIRES.

NUREG/CR-4624 V05: RADIONUCUDE RELEASE CALCULATIONS HOFMAYER.C.H. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5 Large NUREG/CR-4659 V01: SEISMIC FRAGluTY OF NUCLEAR POWER Dry Containment Design.

PLANT COMPONENTS (PHASE 1).

NUREG/CR-4656: VERIFICATION TEST CALCULATONS FOR THE SOURCE TERM CODE PACKAGE.

HORINE,F.

NUREG/CR4570 DESCRIPTION AND TES1:NG OF AN APPARATUS JOYCE,J.A.

FOR ELECTRICALLY INITIATED FIREC THROUGH SIMULATION OF NUREG/CR-4699: APPUCATON OF ALTERNATING CURRENT PO-A FAULTY CONNECTON. TENTIAL DIFFERENCE TO CRACK LENGTH MEASUREMENT DURING RAPO LOADING.

HUTTON.P.H.

NUREG/CR-4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION- KAM,F.B.

SHIP FOR IN-SERVICE MONITORING OF NUCLEAR P9 ESSURE NUREG/CR-4647; CALCULATION OF THE POWER DISTRIBUTON IN VESSELS Progress Report. OctoLer 1385 March 1986. THE VENUS PWR MOCK-UP BENCHMARK USING TWO-GROUP DIF-FUSION THEORY.

NUREG/CR-3805 V04: ENGINEERING CHARACTER 12ATON OF KASSNER,T.F.

GROUND MOTION. Task 11: Sod-Structure Interaction Effects on Struc- NUREG/CR-4667 V01: ENVIRONMENTALLY ASSISTED CRACKING IN NU G/hNS V05: ENGINEERING CHARACTERl2ATION OF GROUND MOTION. Task II: Summary Report KEADNEY,M.S.

NUREG 1213: FLANS AND SCHEDULES FOR IMPLEMENTATION OF IMAN,R.L.

NUREG/CR-4487; UNCERTAIP 7 # " SENSITIVITY ANALYSIS OF A U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBluTIES UNDER THE LOW-LEVEL RADIOACTIVE WASTE POUCY AMEND-DRY CONTAINMENT TEST P sLEM FOR THE MAEROS AEROSAL MODEL. MENTS ACT OF 1985 (PL 90-240).

NUREG/CR4598: A USER'S GUOE FOR THE TOP EVENT MATRIX KELLY,R.F.

ANALYSIS CODE (TEMAC). NUREG/CR 4624 V01: RADIONUCUDE RELEASE CALCULATIONS INVERSONI,C. FOR SELECTED SEVERE ACCOENT SCENARIOS. Volume NU EG CR ORPLOT.PC: A GRAPHIC UTIUTY FOR ORMGEN.PC NU E R-4624 2: RADIONUCLIDE RELEASE CALCULATIONS 1 FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2.PWR,1ce JACO8S,G.K. Condenser Design.

l NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCUDE NUREG/CR-4624 V03: RADONUCUDE RELEASE CALCULATIONS )

GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL FOR SELECTED SEVERE ACCIDENT SCENAROS Volume '

NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For 3 PWR.Subatmovheric Containment Design.

October 1984 -September 1985. NUREG/CR-4624 V04: RADIONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume JACOSUS M.J. 4.BWA. Mark til Design.

NUREG/CR-4596: SCREENING TESTS OF REPRESENTATIVE NUCLE. NUREG/CR4624 V05. RADIONUCUDE RELEASE CALCULATIONS AR POWER PLANT COMPONENTS EXPOSED TO SECONDARY EN. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5.Large VIRONMENTS CREATED BY FIRES. Dry Containment Design.

JANKtVICH,J. KELMERS,A.D.

NUREG-1212 VOI: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR NUREG/CR-4236: PROGRESS IN EVALUA10N OF RADONUCUDE POWER INDUSTRY 1Gd5. Volume 1. Findings And Conclusions. GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH. LEVEL I

l Personal Author Index 29 NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Repor* For LA 4DOW,M.

October 1984 -September 1985. AUREG/CR-4082 V04: DEGRADED PIPING PROGRAM - PHASE t

KENNEDY,R.P.

II. Semiannual Report October 1985 March 1986.

l NUREG/CH-4575: PREDICTIONS OF J-R CURVES WITH LARGE j NUHEG/CR-3805 V05: ENGINEERING CHARACTER 12ATON OF CRACK GROWTH FROM SMALL SPECIMEN DATA.

, GROUND MOTION. Task 11: Summary Report.

KHATie-RAH 8AR NUREG/CR-4629: INDEPENDENT VERIFICATION OF RADIONUCUDE NUREG-1212 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR RELEASE CALCULATONS FOR SELECTED ACCIDENT SCENARIOS. POWER INDUSTRY 1985.VOL 1. Findings And Conclusaons.

NUREG/CR4688 V01: OUANTIFICATION AND UNCERTAINTY ANALf- NUREG-1212 V02- STATUS OF MAINTENANCE IN U.S. NUCLEAR IN-SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT DUSTRY 1985. Volume 2.Descnptions of Programs And Practices.

{ W ER R1 ACTORS (OUASAR).Part 1: Methodology And Program NUREG/CR4587: SOURCE TERM CODE PACKAGEA USER'S GUIDE KIEFNER,J. (MOD 1).

I NUREG/CR4082 V04: DEGRADED PtPING PROGRAM - PHASE It.Samiannual Report, October 1985 - March 1986. LE NU G/CR-4624 V01: RADIONUCUDE RELEASE CALCULATIONS KIRCHNER.J.R. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume NUREG/CR-4560: RISK ASSESSMENT APPUCATON TO NRC IN- 1:8WRMark 1 Design.

SPECTON PROGRESS REr' ORT FOR PERIOD JANUARY 1985 TO NUREG/CR4624 V02- RADIONUCUDE RELEASE CALCULATIONS JANU/.RY 1986. FOR SELECTED SEVERE ACC1 DENT SCENARIOS. Volume 2.PWR,lce KIRK,M.T. Condenser Design.

NUREG/CR-4624 V03: RADIONUCUDE RELEASE CALCULATONS NUREG/CR-4540: AN EVALUATON OF J-R CURVE TESTING OF NU* FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume CLEAR PIPING MATERIALS USING THE DIRECT CURRENT POTEN' 3:PWR.Subatmosphenc Containment Dessgn.

TIAL DROP TECHN1QUE.

NUREG/CR4624 V04. RADIONUCUDE RELEASE CALCULATONS KNIPE.A.D. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume NUREG/CR4 BF SEVERE FUEL DAMAGE SCOPING TEST. TEST NU E -4 RADIONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume $1arge

~ KOGAN.V. Dry Containment Desagn.

NUREG/CR-4587: SOURCf TERM CODE PACKAGE:A USER'S GUIDE NUREG/CR-4656; VERIFICATON TEST CALCULATONS FOR THE (MOD 1). SOURCE TERM CODE PACKAGE.

NUREG/CR-4624 V01: RADIONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume LEGGETT,R.W.

1.8WR. Mark I Desegn. NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATON DOSE COM-NUREG/CR-4624 V02: RADIONtlOUDE RELEASE CALCULATIONS MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2.PWR,lce FROM NUCLEAR FUEL FACluTIES.

Condenser Design.

NUREG/CR4624 V03: RADONUCUDE RELEASE CALCULATIONS LEIGH C.D.

FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume NUREG/CR-4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A NR R-46 R ELEASE CALCULATONS MODE FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume 4 BWRMark lil Design.

LENNON,G.

NUREG/CR4624 V05: RADIONUCUDE RELEASE CALCULATONS NUREG/CR4705: IDENTIFICATION OF NORTHWEST TRENDING FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume $1arge SEISMOGENIC GRABEN NEAR CHARLESTON, SOUTH CAROLINA.

Dry Containment Design.

NUREG/CR-4677: TRAP-MELT 2 CODE DEVELOPMENT AND IMPROVE- yow,J,3, MENT OF TRANSPORT MODEUNG.

NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE KOHLER,W. HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF NUREG/lA4003: INFLUENCE OF THE WETTING STATE OF A HEATED TENNESSEE AND GEORGIA.

SURFACE ON HEAT TRANSFER AND PRESSURE LOSS IN AN EVAPORATOR TU8E. W NGST M AM.

NUREG/CR-4698: AIRBORNE UDAR MAPPING OF SF6 COPCENTRA-KOONTZ,J. TON DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES.

NUREG-1212 V01: STATUS OF MAINTENANCE IN THE u.S. NUCLEAR POWER INDUSTRY 1985.VOL 1.Findergs And Conclusions. LMGLT.

NUREG-1212 V02: STATUS OF MAINTt; NANCE IN U.S. NUCLEAR IN- NUReiG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE DUSTRY 1985. Volume 2:Desenptions of Programs And Practices. HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF TENNESSEE AND GEORGIA.

NUREG/CR4082 V04: DEGRADED PIPING PROGRAM PHASE LOPES,J.C.

II. Semiannual Report, October 1985 - March 1986.

NUREG/CR4729: DROPLET ENTRAINMENT IN VERTICAL ANNULAR KUHLMA N,M.R. FLOW AND ITS CONTRIBUTON TO MOMENTUM TRANSFER NUREGICRs677: TRAP-MELTi CODE: DEVELOPMENT / ND IMPROVE- LUCO,J.E.

MENT OF TRANSPORT MODELING.

NUREG/CR-3805 V04: ENGINEERING CHARACTERIZATON OF KULHOWVICK,G. GROUND MOTON. Task II: Soil-Structure Interaction Effects On Struc-NUREG/CR-4082 V04: DEGRADED PIPING PROGRAM - PHASE tural Response ILSerrgannual Report, October 1985 March 1986.

MACDOUGALL,R.D.

KURTZ,R.J. NUREG-1213: PLANS AND SCHEDULES FOR IMPLEMENTATON OF NUREG/CR4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION. U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBluTIES SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE UNDER THE LOW-LEVEL RADIOACTIVE WASTE POUCY AMEND-VESSELS. Progress Report. October 1985 March 1986. MENTS ACT OF 1985 (PL 99-240).

LANDESMAN,M. MAERKER,R.E.

NUREG/CR-4647: CALCULATICN OF THE POWER DISTRIBUTION IN NUREG/CR-4439: LEPRICON ANALYSIS OF PRESSURE VESSEL SUR-THE VENUS PWR MOCK-UP OENCHMARK USING TWO-GROUP DIF- VErLANCE DOSIMETRY INSERTED INTO H.B. ROBINSON-2 FUSION THEORY. DURING CYCLE 9.

30 Personal Author Index MALVA.P.S. NAKAGAKl,M.

NUREG/CR4667 V01: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4082 V04: DEGRADED PIPING PROGRAM - PHASE UGHT WATER REACTORS. Semsannual Report.Apnt-September 1985. it. Semiannual Report, October 1985 - March 1986.

MAKAY,E. NAVA,S.J.

NUREG/CR-4597 V01: AGING AND SERVICE WEAR OF AUX 1UARY NUREG/CR-4702: POST-EOCENE FAULT NEAR EAST EDGE OF FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol REELFOOT RIFT IN LAUDERDALE COUNTY, TENNESSEE AS DIS-

1. Operating Experience And Failure identification. COVERED BY GRAVITY, EARTH RESISTIVITY SURVEYS AND MARSCHALL,C.

NUREG/CR-4575: PREDICTIONS OF J-R CURVES WITH LARGE NEATHERY,T.L CRACK GROWTH FROM SMALL SPECIMEN DATA. NUREG/CR4707: A PRELIMINARY GEOLOGIC EVALUATION OF THE ALABAMA-TENNESSEE TRANSVERSE SEISMIC ZONE IN ALABAMA.

NUREG/CR-4082 V04: DEGRADED PIPING PROGRAM . PHASE NEEL,R.B.

II. Semiannual Report, October 1985. March 1986. NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPUCATIONS TO BIOASSAY AT URANIUM MILLS.

NUREG/CR4553: FIELD TESTS OF A FORTABLE TISSUE EQUIVA- NEITZEL D.A.

LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA NUREG/CR4626 V01: IMPROVING THE RELIABluTY OF OPEN-RADIATON FIELDS. CYCLE WATER SYSTEMS.An Evaluation Of Biofouling Surveillance And Control Techniques For Use At Nuclear Power Plants.

NUREG/CR-4102 AIR CURRENTS DRIVEN BY SPRAYS IN REACTOR NEUMANN,U.

CONTAINMENT BUILDINGS- NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER MAXEY,W.

CANNM AND HDR EEMER DM NUREG/CR4082 V04: DEGRADED PIPING PROGRAM . PHASE NICHOLSON,T.J.

ILSemiannual Report, October 1985 March 1986. PUREG/CR-4622: VAUDATION OF STOCHASTIC FLOW AND TRANS-PORT MODELS FOR UNSATURATED SOILS.A Comprehensive Field MCELROY,W.N.

NUREG/CR-3320 V01: LWR PRESSURE VESSEL SURVEILLANCE 00-SIMETRY IMPROVEMENT PROGRAM. PSF Expenments Summary And NICOLOSI,S.L Blind Test Results. NUREG/CR-4656: VERIFICATION TEST CALCULATIONS FOR THE MCLAUGHLIN,P. SOURCE TERM CODE PACKAGE.

NUREG-1212 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR NIELSEN,N.B.

POWER INDUSTRY 1985.VOL 1. Findings And Conclusions. NUREG/CR4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA.

NUREG-1212 V02- STATUS OF MAINTENANCE IN U.S. NUCLEAR IN- TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES.

DUSTRY 1985. Volume 2:Descnptions of Programs And Practices.

NMM,K.

MEYER,R.E.

NUREG/CR4236: PROGRESS IN EVALUATON OF RADIONUCUDE EG/@M M RADCAN MW END N GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL

^" '^"""*'

NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For OLSON.J.

C::tober 1984 -September 1985. NUREG/CR-4125 V01 RI: GUIDEUNES AND WORKBOOK FOR AS-MEYER,R.O SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTIU-TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG-0956: REASSESSMENT OF THE TECHNICAL BASES FOR ES- PLANT. Volume 1: Guidelines For Utility Organization And Administration TIMATING SOURCE TERMS. Final Report.

Plan.

giggag43, NUREG/CR-4125 V02 R1: GUIDEUNES AND WORKBOOK FOR AS.

NUREG/CR-4593: INITIAL CONCEPTS ON ENERGETICS & MASS RE- SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTIJ-LEASES DURING NONNUCLEAR EXPLO IVE EVENTS IN FUEL TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLANT. Volume 2-Workbook For Assessment Of Organization And Man-CYCLE FACIUTIES.

agement MITCHELL.D.H.

NU EG/CR-3974 GE M MBRANE SELECTION CRITERIA FO7 URA.

EG CR 4125 V01 R1: GUIDEUNES AND WORKBOOK FOR AS.

SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTILl-MITCHELL.J.A. TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG-0956: REASSESSMENT OF THE TECHNICAL BASES FOR ES. PLANT. Volume 1: Guidelines For Utility Organization And Administration TIMATING SOURCE TERMS Final Report Plan.

NUREG/CR-4125 V02 R1: GUIDEUNES AND WORKBOOK FOR AS-MITRA,S. SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTILl-NUREG/CR4643: EVALUATON OF CORE DAMAGE SEQUENCES INT- TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOUNG. PLANT. Volume 2: Workbook For Assessment Of Organization And Man-MORLEY,B.B.

NUREG/CR-4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA. OSSORNE,W.E.

TON DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES. NUREG/CR4707: A PREUMINARY GEOLOGIC EVALUATON OF THE ALABAMA TENNESSEE TRANSVERSE SEISMIC ZONE IN ALABAMA.

MUMLHEIM.M.D.

NUREG/CR-4470. SURVEY AND EVALUATON OF VITAL INSTRUMEN. OSETEK D.J.

TATON AND CONTROL POWER SUPPLY EVENTS. f4UREG/CR4683: PGF SEVERE FUEL DAMAGE SCOPING TEST. TEST RESULTS REPORT.

MURPHY,E.

NURECr1218 SAFETY EVALUATION REPORT RELATED TO THE CSIENSKY,J.

OPERABluTY AND REUABluTY OF EMERGENCY DIESEL GENERA- NUREG/CR-3967: AN ANALYSIS OF EXCURSONS AT SELECTED IN TORS MANUFACTURED BY TRANSAMERICA DELAVAL.INC. SITU URANIUM MINES IN WYOMING AND TEXAS.

M9RPHY,G.A. PAPASPYROPOULOS NUREG/CR-447*: SURVEY AND EVALUATON OF VITAL INSTRUMEN. NU9EG/CR-4082 VC4: DEGRADED PIPING PROGRAM - PHASE TRTION AND CONTROL POWER SUPPLY EVENTS. Il Semiannual Report October 1985 March 1986.

l Personal Author Index 31 NUREG/CR-4575: PREDICTONS OF J-R CURVES WITH LARGE RATCUFFE.N.M.

CRACK GROWTH FROM SMALL SSECIMEN DATA. NUREG/CR-4876: SEISMIC REFLECTION GEOMETRY OF THE PARK.C.

NEWARK BASIN IN EASTERN PENNSYLVANIA Evidence For Exten-sional Reactwation Of Paleozoic Thrust Faults.

NUREG/CR-4688 V01: QUANTIFICATION AND UNCERTAINTY ANALY-SIS OF SOURCE TERMS FOR SEVERE ACODENTS IN LIGHT REEVES,M.

j WATER REACTORS (OVASAR).Part 1: Methodology And Prograrn NUREG/CR-3925; SWIFT 11 SELF. TEACHING CURRICULUM.lliustratrve Plan. Probierns For Sandia Waste-Isolatxm Flow And Transport Model For PARK,J.Y.

NUREG/CR4667 Vot: ENVIRONMENTALLY ASSISTED CRACKING IN RICH,5.L UGHT WATER REACTORS. Serrmannual ReportApril-Septernber 1985. NUREG/CR-4553 FIELD TESTS OF A PORTABLE TISSUE EQUIVA-PASUPATHI,V. LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA RADIATION FIELDS.

NUREG/CR4082 V04: DEGRADED PIPING PROGRAM - PHASE II.Serrmannual Report. October 1965 - March 1986. RIDGELY,J.N.

PAULA,H.M. NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE i NUREG/CR-4560 RISK ASSESSMENT APPUCATON TO NRC IN- C-8.BOtuNG WATER REACTOR MAW STE/M ISOLATON VALVE LEAKAGE AND LEAKAGE TREATMENT METHODS.

SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO JANUARY 1986. ROGNESS,D.

NUREG/CR-3967: AN ANALYSIS OF EXCURSIONS AT SELECTED IN N G 013: LADTAP II - TECHNICAL REFERENCE AND USER NS W W AND TEM GUOE.

ROSDAHL.O.

NUREG/1A-0007: ASSESSMENT OF RELAP5/ MOD 2 AGAINST CRITI-N EG 1 2 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR CAL FLOW DATA FROM TESTS JIT 11 AND CFT 21.

POWER INOUSTRY 1985.VOL 1F s And Conclusions. ROSS.P.A.

NUREG-1212 V02: STATUS OF MAINT IN U.S. NUCLEAR IN-DUSTRY 1985. Volume 2.Descriptons of Programs And Practces. NUREG-0020 V10 N05: UCENSED OPERATING DEACTORS STATUS

SUMMARY

REPORT.Deta As Of April 30.1986.(Grey Book)

F NUREG-0020 V10 N06: UCENSED OPERATING REACTORS STATUS PIETRZAK.R.N644:

NUREG/CF GEOCHEMICAL STUDIES OF COMMERCIAL LOW-

SUMMARY

REPORT. Data As Of May 31,1968.(Gray Book)

LEVEL RADIOACTIVE WASTE DISPOSAL SITES.Topcal Report.

ROTH,E.M.

PLOGER,S.A. NUREG/CR4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-NUREG/CR4683: PBF SEVERE FUEL DAMAGE SCOPING TEST TEST AR POWER PLANT PERSONNELA Feasitality Study. Volume 1; Sum.

NU GC 32 V02: MODELS OF COGNITIVE BEHAVIOR IN NUCLE.

POLLACK.H.N. AR POWER PLANT PERSONNELcA Feasitulity Study. Volume 2 Mac NUREG/CR-3145 V04: GEOPHYSICAL INVESTIGATIONS OF WEST. Report.

ERN OHO INDIANA REGION. Annual Report (October 1984 - Septem-ber 1985), RUTHER,W.E.

NUREG/CR4667 V01: ENVIRONMENTALLY ASSISTED CRACKING IN POWELL,R. UGHT WATER REACTORS. Semiannual Report Apnt-September 1985.

NUREG/CP 0083: ANS TOPICAL MEETING ON RADIOLOGICAL ACCI-0 S PER CTIVES AND EMERGENCY PLANNING PROGRAM U N RE 56: REASSESSMENT OF THE TECHNICAL BASES FOR ES-TIMATING SOURCE TERMS. Final Report POWER,M.S.

NUREG/CR-3805 V04: ENGINEERING CHARACTERIZATION OF SCHNEDER.C.S.

GROUND MOTON. Task II: So4-Structure interaction Effects on Struc- NUREG/CR-4699: APPUCATION OF ALTERNATING CURRENT PO-tural Response. TENTIAL DIFFERENCE TO CRACK LENGTH MEASUREMENT NUREG/CR-3805 V05: ENGINEERING CHARACTERIZAYlON OF DURING RAPID LOADING.

GROUND MOTION. Task 11: Summary Report.

E CHE M POWERS.D.A. NUREG/CR4587: SOURCE TERM CODE PACKAGEA USER'S GUIDE NUREG/CR-4308: VANESA A MECHANISTIC MODEL OF RADIONU. (MOD 1).

CUDE RELEASE AND AEROSOL GENERATON DURING CORE DEBRIS INTERACTONS WITH CONCRETE. SC N REG V01: RADONUCUDE RELEASE CALCULATIONS PRATER J.T. FOR SELECTED SEVERE ACODENT SCENARIOS. Volume NUREG/CR-4495: VISCOSITY OF ZlRCONIUM-URANIUM OXIDE (Zr. 1.8WR. Mark i Design.

UO2) MIXTURES AT 1000 TO 2100 C. NUREG/CR4624 V02: RADONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACODENT SCENARIOS. Volume 2.PWR,lce PUGH.C.E Condenser Design.

NUREG/CR-4219 V03 N1: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR-4624. V03: RADIONUCUDE RELEASE CALCULATIONS PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER FOR SELECTED SEVERE ACODENT SCENARIOS. Volume 1985 MARCH 1986. 3 PWR.Subetmospheric Contamment Design.

NUREG/CR4673: HEAVY-SECTON STEEL TECHNOLOGY PROGRAM NUREG/CR4624 V04: RADIONUCUDE RELEASE CALCULATIONS

. FIVE YEAR PLAN FY19851989. FOR SELECTED SEVERE ACCIDENT SCENAROS. Volume 4.BWA. Mark lit Desagn.

PUSKIN,J.S.

NUREG/CR-4624 V05: RADIONUCUDE REtrASE CAsCULATONS NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPUCATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS.Voiume 5.Large TO BIOASSAY AT UAANIUM MILLS. Dry Containment Design.

NUREG/CR-4677: TRAP-MELT 2 CODE. DEVELOPMENT AND IMPROVE-QUAYLE,S.

MENT OF TRANSPORT MODELING.

NUREG/CR-4587: SOURCE TERM CODE PACKAGEA USER'S GUIDE (MOD 1). SCOTT,P.

NUREG/CR-4082 V04: DEGRADED PIPING PROGRAM . PHASE RANKIN W.L ll.Sem6 annual Report, October 1985 - March 1986.

NUREG/CR4600 HUMAN FACTORS STUDY CONDUCTED IN CON- NUREG/CR.4574: AN EXPERIMENTAL AND ANALYTICAL ASSESS.

JUNCTON WITH A MINI-ROUND ROBIN ASSESSMENT OF ULTRA- MENT OF ORCUMFERENTIAL THROUGH. WALL CRACKED PIPES SONIC TECHNICAN PERFORMANCE. UNDER PURE BENDING.

32 Personal Author index SEELEY,F.G. COVERED BY GRAVITY, EARTH REStSTIVITY SURVEYS AND NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCLIDE DRILUNG.

GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For STEELE,R.

October 1984 September 1985. NUREG/CR 4648. A STUDY OF TYPICAL NUCLEAR CONTAINMENT PURGE VALVES IN AN ACCIDENT ENVIRONMENT.

SHAA8AN,H.L NUREG-1196: AN OVERV!EW OF ENVIRONMENTAL MATERIALS DEG- STRENGE,D.L l RADATON IN UGHT-WATER REACTORS. NUREG/CR-4013: LADTAP ll - TECHNICAL REFERENCE AND USER l SHACK,W.J. '

NUREG/CR-4667 V01: ENVIRONMENTALLY ASSISTED CRACKING IN SUO-ANTTILA A.

UGHT WATER REACTORS. Senuannual Report.Apni-September 1985. NUREG/CR-4521: TURC2 AND 3:LARGE SCALE UO2/ZRO2 MELT-SHEAA CONCRETE INTERACTION EXPERIMENTS AND ANALYSIS.

NURFG-1143 S01: SAFETY EVALUATION REPORT RELATED TO THE SWANSON,J.L FULL-TERM OPERATING UCENSE FOR MILLSTONE NUCLEAR POWER NUREG/CR-4660 ORGANIC COMPLEXANT-ENHANCED MOBluTY OF T . UNIT NO.1. Docket No. 50-245.(Northeast Nuclear TOXIC ELEMENTS IN LOW-LEVEL WASTES. Final Report SHIVER,A.W. SZABO.M.W.

NUREG/CR-4308: VANESA A MECHAN lSTIC MODEL OF RADIONU. NUREG/CR-4707: A PREUMINARY GEOLOGIC EVALUATION OF THE CUDE RELEASE AND AEROSOL GENERATION DURING CORE ALABAMA-TENNESSEE TRANSVERSE SEISMIC ZONE IN ALABAMA.

DEBRIS INTERACTONS WITH CONCRETE.

SHORTENCARIER NUREG 1021 R03: OPERATOR UCENSING EXAMINER STANDARDS.

4 NUREG/CR 4598: A USER'S GUIDE FOR THE TOP EVENT MATRIX ANALYSIS CODE (TEMACL TABATABAl,A.S.

NUREG/CR-2800 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT SIL8ERBERG,M. SAFETY ISSUE PRiORITIZATION INFORMATION DEVELOPMENT.

NUREG-0956: REASSESSMENT OF THE TECHNICAL BASES FOR ES-TiMATING SOURCE TERMS. Final Report TAYLOR,T.T.

NUREG/CR-4600: HUMAN FACTORS STUDY CONDUCTED IN CON-UREG/CR4566: COMPORN 111 - A COMPUTER CODE FOR MODEL- N TECH CIA P FOR ING COMPARTMENT FIRES.

SIMMONS,C.S. THOMAS.J.M.

NUREG/CR-4162: SURVEY OF STATISTICAL AND SAMPUNG NEEDS NUREG/CR-4622- VALIDATION OF STOCHASTIC FLOW AND TRANS-PORT MODELS FOR UNSATURATED SOILS A Comprehensive Field FOR ENV!RONMENTAL MONITORING OF COMMERCIAL LOW-Study. LEVEL RADIOACTIVE WASTE DISPOSAL FACIUTIES.

SKAGGS.R.L THORNTON,R.L NUREG/CR-4480- EROSION PROTECTON OF UAANIUM TAILINGS IM. NUREG/CR-4470: SURVEY AND EVALUATION OF VITAL INSTRUMEN-POUNDMENTS. TATION AND CONTROL POWER SUPPLY EVENTS.

SOMMERS P. THURBER.J.A.

NUREG/CR4125 V01 R1: GUIDEUNES AND WORKBOOK FOR AS. NUREG/CR4125 V01 R1: GUIDEUNES AND WORKBOOK FOR AS.

SESSMENT OF ORGANIZATON AND ADMINISTRATION OF UTiu- SESSMENT OF ORGANIZATON AND ADMINISTRATION OF UTlu-TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLANT. Volume 1: Guide 4nes For Utilty Organization And Administration PLANT. Volume 1: Guidelines For Utihty Orgaruzation And Admirustration Pim Plan.

NUREG/CR4125 V02 Rt: GUIDEUNES AND WORKBOOK FOR AS. NUREG/CR-4125 V02 R1: GUIDEUNES AND WORKBOOK FOR AS-SESSMENT OF ORGANIZATON AND ADMINISTRATION OF UTlu- SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTiu-TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLANT. Volume 2. Workbook for Assessment of Organization And Man- PLANT. Volume 2: Workbook For Assessment Of Orgaruzation And Man-rgement agement.

SPANGLER,4.8. THURGOOD,M.J.

NUREG-1205: REACTIVATION OF NUCLEAR POWER PLANT CON- NUREG/ Cit-3262 V04: COBRA-NC:A THERMAL HYDRAUUCS CODE STRUCTION PROJECTS. Plant Status.Pohey issues,And Regulatory FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR l Options. COMPONENTS.Vol. 4. Users' Manual For Containmer:t Analysis.

SPANNER,J.C. TICHLER,J.

NUREG/CR-4600: HUMAN FACTORS STUDY CONDUCTED IN CON- NUREG/CR-2907 V04. RADIOACTIVE MATERIALS RELEASED FROM JUNCTION WITH A MIN 1-ROUND ROBIN ASSESSMENT OF ULTRA

• NUCLEAR POWER PLANTS. Annual Repor11983.

SONIC TECHNICIAN PERFORMANCE.

SPLETZER'8'L NUREG/CR-4255 V03 N1: AEROSAL RELEASE AND TRANSPORT NUREG/CR-4570: DESCRIPTON AND TESTING OF AN APPARATUS PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER FOR ELECTRICALLY INITIATED FIRES THROUGH SIMULATON OF 1985. MARCH 1986.

A FAULTY CONNECTON.

ST CLAIR,S.D. TOMAN,G.T.

NUREG/CR4536: SUPERHEATED-STEAM TEST OF ETHYLENE PRO, NUREG/CR-4257 V02- INSPECTION. SURVEILLANCE.AND MONITOR-PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND ING OF ELECTRICAL ECUIPMENT IN NUCLEAR POWER ACCIDENT ENVIRONMENT. PLANTS. Volume 2. Pressure Transmitters.

STAUS,W.P. UTHE,E.E.

NUREG/CR-3967: AN ANALYSIS OF EXCURSIONS AT SELECTED IN NUREG/CR4698. AIRBORNE UDAR MAPPING OF SF6 CONCENTRA.

SITU URANIUM MINES IN WYOMING AND TEXAS. TON DISTRIBUTIONS FOR TRANSPORT AND DIF FUSON STUDIES.

STEARMS,R.G. VANNONI,M.G.

NUREG/CR-4702: POST. EOCENE FAULT NEAR EAST EDGE OF NUREG/CR-4678: A METHOD FOR USING PAA TO ESTABUSH QUAL.

REELFOOT RIFT IN LAUDERDALE COUNTY, TENNESSEE AS DIS- ITY PROGRAM APPUCABILITY.

Personal Author Index 33 VOJTEK,L W1ERENGA.P.J.

NUREG/lA-0002- HEAT TRANSFER PROCESSES DURING INTERMEDI- NUREG/CR 4622: VAUDATON OF STOCHASTIC FLOW AND TRANS-ATE AND LARGE BREAK LOCAS. PORT MODELS FOR UNSATURATED SOILS.A Comprehensrve Field WALTERS,W.H.

NUREG/CR4480 EROSION PROTECTION OF URANIUM TAILINGS IM- WILKOWSKI,G.M.

POUNDMENTS. NUREG/CR-4082 V04. DEGRADED PIPING PROGRAM - PHASE H.Semannual Repat, October M85 - March 1986.

WARD.D.S.

NUREG/CR-3925: SWIFT ll SELF TEACHING CURRICULUM.lliustrative WILLIAMS,M.L Problems For Sandia Waste-Isolaton Flow And Transport Model For NUREG/CR4647; CALCULATION OF THE POWER DISTRIBUTON IN Fractured Media-THE VENUS PWR MOCK-UP BENCHMARK USING TWO-GROUP DIF-W ARE,A.G. FUSION THEORY.

NUREG/CR-4529: PIPING DAMPING - EXPE9l MENTAL RESULTS NU CR-45 P PE PN. SU BR ON TESTS IN N REG 3967: AN ANALYSIS OF EXCURSONS AT SELECTED IN THE 33 TO 100 HERTZ FREQUENCY RANGE. SITU URAN!UM MINES IN WYOMING ANC TEXAS.

WATKINS.J.C. WILSON,SL NUREG/CR-4648: A STUDY OF TYPICAL NUCLEAR CONTAINMENT NUREG/CR-4702: POST-EOCENE FAULT NEAR EAST EDGE OF PURGE VALVES IN AN ACCIDENT ENVIRONMENT. REELFOOT RIFT IN LAUDERDALE COUNTY, TENNESSEE.AS DIS-COVERED BY GRAVITY, EARTH RESISTMTY SURVEYS AND WEISS,A.J. DRILUNG.

NUREG/CR-2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY WOHL,ML RESEARCH.Ouarterly Progress Report, October December 1985. NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE C-8,BCiUNG WATER REACTOR MAIN STEAM ISOLATON VALVE U E'G/CR-3145 V04: GEOPHYSICAL INVESTIGATONS OF WEST-ERN OHIO-INDIANA REGION. Annual Report (October 1984 Septem- WONG,HL ber 1985). NUREG/CH-3805 V04: ENGINEEP.ING CHARACTERIZATION OF GROUND MOTON. Task 11: Soil-Structure Interacton Effects On Struc-WH W M wral Responsa NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL WOOOS,D.D.

NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For NUHEG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE.

October 1984 -September 1985.

AR POWER PLANT PERSONNELA Feasibihty Study. Volume 1: Sum-WHEELER,CL mary Of Results.

NUREG/CR-3262 V04: COBRA-NC:A THERMAL HYDRAUUCS CODE NUREG/CR 4532 V02: MODELS OF COGN!TIVE BEHAVIOR IN NUCLE-FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR AR POWER PLANT PERSONNEL:A Feasibihty Study. Volume 2. Man COMPONENTS.Vol. 4. Users' Manual For Contamment Analysis. Report WHEELER,W.A. WOOTON,R.

NUREG/CR-4600t HUMAN FACTORS STUDY CONDUCTED IN CON- NUREG/CR-4587. ,0URCE TERM CODE PACKAGEA USER'S GUIDE JUNCTON WITH A MIN 1-ROUND ROBIN ASSESSMENT OF ULTRA- (MOD 1).

SONIC TECHNICIAN PERFORMANCE.

WOOT 6 4 WHELAN,0. NUREG/CR-4656: VERIFICATION TEST CALCULATIONS FOR THE NUREG/CR-4013: LADTAP 11 TECHNICAL REFERENCE AND USER SOURCE TERM CODE PACKAGE.

GUIDE.

WU,P.

WHITEHEAD.D.W. NUREG-1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG-NUREG/CR-4678: A METHOD FOR USING PRA TO ESTABUSH OUAL- RADATION IN UGHT-WATER REACTORS.

ITY PROGRAM APPUCABluTY.

WYANT,F.J.

WIDRIO,R.D-NUREG/CR4530 V01: U.S./ FRENCH JOINT RESEARCH PROGRAM NUREG/CR-4125 V01 R1: GUIDELINES AND WORKBOOK FOR AS. REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTill- JECTED TO BETA RADIATON Vol 1: Phase-1 Normalizaton Results.

TIES SEEKING OPERAT;NG UCENSE FOR A NUCLEAR POWER PLANT. Volume 1:Guxsuhnes For Utihty Orgaruzaton And Admnistraton YOUNG 8LOOD,R.

Plan. NUREG/CR4643: EVALUATON OF CORE DAMAGE S. QUENCES INI-NUREG/CR4125 V02 R1: GUIDEUNES AND WORKBOOK FOR AS- TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOUNG.

SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTiu.

TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER ZUCKER8ROO,D.

PLANT. Volume 2. Workbook for Assessment of Orgaruzation And Man. NUREG/CR-4533: PROGRAM TO ANALYZE THE FAILURE MODE OF ogement. LEAD ACID BATTERIES.

_ _ . _ _ _. J

Subject index This index was developed from keywords moved later when a reasonable thesaurus and 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.

(HOR) American Oyster NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER NUREG/CR4626 V01: IMPROVING THE REUABluTY OF OPEN-CANNON AND HOR EXPERIMENTAL DATA.

CYCLE WATER SYSTEMS.An Evaluation Of Biofouhno Surveillance And Control Technsques For Use At Nuclear Power Plants.

NUREG/CP4080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON REACTOR PHYSICS AND SAFETY. Sessions 110, Annual Report NUREG-1145 V02- U.S. NUCLEAR REGULATORY COMMISSION 1985 Abstract ANNUAL REPORT.

NUREG-0304 VII NO2: REGULATORY AND TECHNICAL NUREG-1215: COMPILATON OF CONTRACT RESEARCH FOR THE REPORTS.Compilater. cor Second Quarter 1986,Apni-June. CHEMICAL ENGINEERING BRANCH,DMSON OF ENGINEERING Admt TECHNOLOGY. Annual Report For FY 1965.

NUREG4956: REASSESSMENT OF THE TECHNICAL BASES FOR ES- NUREG/CR-2907 V04: RADIOACTIVE MATERIALS RELEASED FROM TIMATING SOURCE TERMS Final Report. NUCLEAR POWER PLANTS. Annual Report 1983.

NUREG/CP 0083: ANS TOPICAL MEETING ON RADIOLOGICAL ACCI- NUREG/CR-3145 V04: GEOPHYSICAL INVESTIGATONS OF WEST.

DENTS - PERSPECTIVES AND EMERGENCY Pt.ANNING PROGRAM ERN OHIO-INDIANA REGON. Annual Report (October 1984 - Septem.

bw 1984 NUREG NDEPENDENT VERIFICATON OF RADONUCUDE NUREG/CR4236: PROGRESS IN EVALUATON OF RADONUCUDE RELEASE CALCULATONS FOR SELECTED ACCIDENT SCENARIOS. GEOCHMAL WNAM EMD W W HMN NUREG/CR 4643 EVALUATION OF CORE DAMAGE SEOUENCES INI- NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOUNG. October 1984 -September 1985.

Accident Environment Artificial inteillgence NUREG/CR4648: A STUDY OF TYPICAL NUCLEAR CONTAINMENT NUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVOR IN NUCLE-PURGE VALVES IN AN ACCIDENT ENVIRONMENT. AR POWER PLANT PERSONNEL.A Feasitsty Study. Volume 1: Sum-Acousuc h mary Of Results.

NUREG/CR-4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATON- NUREG/CR4532 V02 MODELS OF COGNITIVE BEHAVIOR IN NUCLE.

SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE AR POWER PLANT PERSONNEL:A Feasibihty Study. Volume 2:Maan VESSELS Progress Report October 1985 - March 1986.

Administration Atletic Cleme NUREGICR4125 V01 R1: GUIDEUNES AND WORKBOOK FOR AS. NUREG/CR-4626 V01: IMPROVING THE RELIABluTY OF OPEN.

SESSMENT OF ORGANIZATION AND ADMINISTRATON OF UTiu- CYCLE WATER SYSTEMS.An Evaluation Of Biofouhng Surveillance TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER And Control Techniques For Use At Nuclear Power Plants.

PLANT. Volume 1:Guidehnes For Utility Orgaruation And Administration Plan. Atmospheric Diepersion NUREG/CR-4125 V02 R1: GUIDEUNES AND WORKBOOK FOR AS. NUREG/CR4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA-SESSMENT OF ORGANtZATION AND ADMINISTRATION OF UTlu- TON DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES.

TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER NUREG/CR4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA-PLANT. Volume 2:Wrvkbook For Assessment Of Orgarnation And Man.

agement. TON DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES.

Advanced Reeeerch Reectore Auxillery Feedwater Purnpo NUREG/CP 0080 V02: PROCEEDINGS OF THE TOPICAL MEETING ON NUREG/CR4597 V01: AGING AND SERVICE WEAR OF AUX 1UARY REACTOR PHYSICS AND SAFETY. Sessions 11-16. FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol

1. Operating Expenence And Failure identification.

Aerosol Behavior NUREG/CR4487: UNCERTAINTY AND SENSITMTY ANALYSIS OF A "

D CONTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL NUREG-1123: KNOWLEDGE AND ABILITIES CATALOG FOR NUCLEAR POWER PLANT OPERATOR.BOluNG WATER REACTORS.

Aeroeot Generation 8WR Ownere Group (BWROG)

NUREG/CR-4521: TURC2 AND 3:LARGE SCALE UO2/2RO2 MELT. HUREG-1189- TECHNICAL FINDINGS RELATED TO GENERIC ISSUE CONCRETE INTERACTON EXPERIMENTS AND ANALYSIS. C8 BOILING WATER REACTOR MAIN STEAM ISOLATION VALVE LEAKAGE AND LEAKAGE TREATMENT METHODS.

NUREG/CR-2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON- Deseft SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-NUR G 536 S TE EA T T YL PRO- ANCE OF CEMENTITOUS BOREHOLE PLUGS.

PYLENE RUFFA3 CABLES USING A SIMULTANEOUS AGING AND ACCIDENT Of/.RONMENT* Bonett Wtunium Technenum NUREG/CR4236: PROGRESS IN EVALUATON OF HADIONUCUDE Air Flow Petterne GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUREG/CR-410'c AIR CURRENTS DRIVEN BY SPRAYE IN REACTOR NUCLEAR WASTE REPOSITORY SITE PROJECTS Annual Report For CONTAINMENT BUILDINGS. October 1964 Septembw 1965.

Airtiorne Efflurmt seemit Tenene Strength NUREG/CR 2907 V04: RADOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1983. NUREG/CR4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-ANCE OF CEMENTITOUS BOREHOLE PLUGS, 35

36 Subject index Basalt Waste isolation Project (BWIP) Coment Permeability NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCUDE NUREG/CR-4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For C:tober 1984 -Sept ,mber 1965. Cement Shrinkage NUREG/CR-4738: SIZE INFLUENCE ON THE SEALING PERFORM-Batter Cell Aging ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

NUREG/CR-4533: PROGRAM TO ANALYZE THE FAILURE MODE OF LEAD-ACID BATTERIES. Centrifugal Pumps NUREG/CR4597 V01: AGING AND SERVICE WEAR OF AUXILIARY Bata Radiation Behavior FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol.

NUREG/CR-4530 VO1: U.S1 FRENCH JOtNT RESEARCH PROGRAM 1. Operating Expenence And Failure identification.

REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-JECTED TO BETA RADIATION.Vol 1: Phase-1 Normalization Results. Channel Checks Blind Test Chetating Agents NUREG/CR-3320 V01: LWR PRESSURE VESSEL SURVEILLANCE DO- NUREG/CR-4660: ORGANIC COMPLEXANT-ENHANCED MOBluTY OF SIMETRY IMPROVEMENT PROGRAM. PSF Expenments Summary And TOXIC ELEMENTS IN LOW-LEVEL WASTES Final Report.

Bhnd Test Results.

Cladding Blue Mussel NUREG/CR-4219 V01 N1: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR4626 V01: IMPROVING THE REUAB!UTY OF OPEN- PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER CYCLE WATER SYSTEMS.An Evaluation Of Biofouhng Surveillance 1985 - MARCH 1986.

And Control Techniques For Use At Nuclear Power Plants. NUREG/CR-4873: HEAVY SECTION STEEL TECHNOLOGY PROGRAM

~

Bolling Water Reactor (BWR)

NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE Class IE Battery Cell C-8 BOILING WATER REACTOR MAIN STEAM ISOLATION VALVE NUREG/CR-4009: AGE-RELATED DEGRADATION OF NATURALLY-LEAKAGE AND LEAKAGE TREATMENT METHODS. AGED CLASS 1E BATTERY CELLS.

Boiling Water Reactor Cognitive Model NUREG/CP-0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON NUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-NURE CP ED G F E T PICAL MEETING ON REACTOR PHYSICS AND SAFETY. Sessions 1116. NUREG CR-45 V02: MODELS OF COGNITIVE BEHAVIOR IN NUCLE.

AR POWER PLANT PERSONNEL:A Feasibelity Study. Volume 2.Masn Borehole Plugging Report.

NUREG/CR4738: SIZE INFLUENCE ON THE SEALING PERFORM-CMCE OF CEMENTITIOUS BOREHOLE PLUGS. Compartment Firea Borehole Sealing NUREG/CR-4566: COMPBRN lli - A COMPUTER CODE FOR MODEL-NUREG/CR4738: S'ZE INFLUENCE ON THE SEAUNG PERFORM- ING COMPARTMENT FIRES.

ANCE OF CEMENTITIOUS BOREHOLE PLUGS. Compilation Of Rules CHF NUREG-0936 V05 N01: NRC REGULATORY AGENDA.Ouarterty NUREG/CR 4738. SIZE INFLUENCE ON THE SEALING PERFORM. Report, January-March 1986.

ANCE OF CEMENTITIOUS BOREHOLE PLUGS. Component Cooling Water (CCW)

COBRA-NC NUREG/CR4643: EVALUATION OF CORE DAMAGE SEQUENCES INi-NUREG/CR 3262 V04: COBRA-NC:A THERMAL HYDRAUUCS CODE TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOLING.

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR CO COMPONENTS.Vol. 4. Users' Manual For Containment Analysis.

NU E R SCREENING TESTS OF REPRESENTATIVE NUCLE-COMPBRN til AR POWER PLANT COMPONENTS EXPOSED TO SECONDARY EN-RUREG/CR4568: COMPBRN 111 - A COMPUTER CODE FOR MODEL. VIRONMENTS CREATED BY FIRES.

ING COMPARTMENT FIRES.

Components CORCON NUREG/CR4659 V01: SEISM;O FRAGILITY OF NUCLEAR POWER NUREG/CR-4308: VANESA.A MECHANISTIC MODEL OF RADIONU- PLANT COMPONENTS (PHASE 1).

CUDE RELEASE AND AEROSOL GENERATION DURING CORE DEBRIS INTERACTIONS WITH CONCRETE.

Compoemng NUREG/CR-4521: TURC2 AND 3-LARGE SCALE UO2/ZRO2 MELT. NUREG/CR-4162: SURVEY OF STATISTICAL AND SAMPLING NEEDS CONCRETE INTERACTION EXPERIMENTS AND ANALYSIS. FOR ENVIRONMENTAL MONITORING OF COMMERCIAL LOW-LEVEL RADIOACTIVE WAST'. DISPOSAL FACIUTIES.

CRAC 2 NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE Computer Code ]

C-8 BOlWNG WATER REACTOR MAIN STEAM ISOLATION VALVE NUREG/CR-3925: SWi"T 11 SELF TEACHING CURRICULUM.lliustrative LEAKAGE AND LEAKAGE TREATMENT METHODS. Problems For Sandia Waste-Isolation Flow And Transport Model For Fractured Media.

Cable Fire NUREG/CR-4656: VEF;lFICATION TEST CALCULATIONS FOR THE NUREG/CR-4570: DESCRIPTON AND TESTING OF AN APPARATUS SOURCE TERV CODE PACKAGE.

FOR ELECTRICALLY INITIATED FIRES THROUGH SIMULATION OF R FAULTY CONNECTON. Computer Program NUREG/CR-4013; LADTAP 11. TECHNICAL REFERENCE AND USER Cement Borehole Pluge GUIDE.

NUREG/CR 4738: SIZE INFLUENCE ON THE SE%1NG PERFORM-ANCE OF CEMENTITOUS BOREHOLE PLUGS. Concentration Measurements NUREG/CR4698: AIRBORNE UDAFI MAPPING Of SF6 CONCENTPA-Coment Curing Temperature TION DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES.

NUREG/CR-4738: SIZE INFLUENCE ON THE SEAUNG PERFORM- NUREG'CR-4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA-ANCE OF CEMENTITIOUS BOREHOLE PLUGS. TON DISTRIBUTONS FOR TRANSPORT AND DIFiUSION STUDIES.

Comerit Expanelon Condenser NUREG/CR-4738: SIZE INFLUENCE ON THE SEALING PERFORM- NUREG 1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG-ANCE OF CEMENTITIOUS BOREHOLE PLUGS. RADATION IN UGHT-WATER REACTORS.

l Subject index 37 Construction Cancellation Croche NUREG-1205: REACTIVATION OF NUCLEAR POWER PLANT CON-STRUCTON PROJECTS. Plant Status. Policy issues.And Regulatory NUREG/CR-4574: AN EXPERIMENMI WO MPiYTICAL ASSESS-Optons. MENT OF CIRCUMFERENTIAL THROUGH WALL CRACKED PIPES UNDER PURE BENDING.

. Contaimnent j Cr!tical Flow Of Saturated Stoom NUREG/CR4102. AIR CURRENTS DRIVEN BY SPRAYS IN REACTOR NUREG/lA-0007: ASSESSMENT OF RELAP5/ MOD 2 AGAINST CRITI-

} NU E / 4 4 V02 ONUCUDE RELEASE CALCULATIONS CAL FLOW DATA FROM TESTS JIT 11 AND CFT 21.

FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume 2.PWR,1ce Criticat Heat Flux NU 4 V03: RADIONUCUDE RELEASE CALCULATIONS NURh/CR-4738: SIZE INFLUENCE ON THE SEALING PERFORM-FOR SELECTED EEVERE ACCOENT SCENARIOS. Volume ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

3:PWR.Subatmosphenc Contamment Desigr.

NUREG/lA-0002- HEAT TRANSFER PROCESSES DURING INTERMEDI-NUREG/CR4624 V04: RADIONUCLIDE RELEASE CALCULATONS ATE AND LARGE BREAK LOCAS.

i FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume 4 BWR. Mark lil Desgn Crustal Tincknees NUREG/CR4624 V05: RADIONUCLIDE RELEASE CALCULATONS NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHQUAKE FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5.Large

, Dry Contamment Desen. HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF TENNESSEE AND GEORGIA.

NUREG/CR-4648: A STUDY OF TYPICAL NUCLEAR CONTAINMENT PURGE VALVES IN AN ACCIDENT ENVIRONMENT. NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF Con h W A TENNESSEE AND GEORGIA.

NUREG/CR-3262 V04: COBRA-NC:A THERMAL HYDRAUUCS CODE Cylinder Rupture FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR COMPONENTS Vol 4. Users' Manual For Conta nment Analysis' NUREG 1198 S01: RELEASE OF UF6 FROM A RUPTURED MODEL Containment Structur, 48Y CYLINDER AT SEQUOYAH FUELS CORPORATON FACILITY. Lessons-Learned Report. NRC Staff Responses To The Rec-NUREG/CR-4336: REV!EW OF flT RESEARCH INSTITUTE REPORTS ommendations Made By The Lessons-Leamed Group.

ON STRUCTURAL STUDIES OF REACTOR CONTAINMENT.

Damping Contamination NUREG/CR-4529: PIPING DAMP'NG - EXPERIMENTAL RESULTS NUREG/CR-2800 SO4. GUIDELINES FOR NUCLEAR POWER PLANT FROM LABORATORY TESTS IN THE SEISMIC RANGE.

SAFETY ISSUE PRIORITIZATON INFORMATION DEVELOPMENT.

Control Power Supply Evente Decontaminet;on NURLG/CR-4470 SURVEY AND EVALUATON OF VITAL INSTRUMEN- NUREG-1215: COMPILATION OF CONTRACT RESEARCH FOR THE TATON AND CONTROL POWER SUPPLY EVENTS. CHEMICAL ENGINEERING BRANCH.DIVISON OF ENGINEERING TATON A D TROL E SUP Y S NURE / 4 45 E SS D SAFETY ASPECTS OF SE-LECTED DECONTAMINATON PROCESSES.

Control System Malfunction NUREG/CR4265 V02: AN ASSESSMENT CF THE SAFETY IMPLICA. Deferral TIONS OF CONTROL AT THE CALVERT CUFFS-1 NUCLEAR NUREG-1205: REACTIVATON OF NUCLEAR POWER PLANT CON-PLANT. Volume 2: Appendices.

STRUCTION PROJECTS. Plant Status, Policy issues,And Regulatory Coolant Activity NUREG/CR-2800 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT Design CC- phi 4

SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.

NUREG-1206: ANALYSIS OF FRENCH (PALUEL) PRESSURIZED Coolant Pressure W Mping WATER REACTOR DESIGN DIFFERENCES COMPARED TO CUR-RENT U.S. PWR DESIGNS' NUREG-0313 R02 DRFT: TECHNICAL REPORT ON MATERIAL SELEC-TION AND PROCESSING GUIDEUNES FOR BWR COOLANT PRES- Diffusion Theory SURE BOUNDARY PIPtNG.

NUREG/CR4647: CALCULATION OF THE POWER DISTRIBUTION IN

' Core Damage Wee THE VENUS PWR MOCK-UP BENCHMARK USING TWO-GROUP DIF-NUREG/CR-4643. EVALUATON OF CORE DAMAGE SEQUENCES INI- FUSION THEOAY.

TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOUNG.

Core Meet Frequence NUREG/CR4656: VERIFICATON TEST CALCULATIONS FOR THE NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE SOURCE TERM CODE PACKAGE.

C-8.BOluNG WATER REACTOR MAIN STEAM ISOLATON VALVE LEAKAGE AND LEAKAGE TREATMENT METHODS. Direct Bioasesy Corroelon NUREG-0874: INTERNAL DOSIMcTRY MODEL FOR APPUCATIONS TO BIOASSAY AT URANIUM M LLS.

NUREG 1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG-l RADATION IN UGHT WATER REACTORS.

2 Direct Current Potential Drop Technique NUREG/CR-4626 V01: lMPROVLNG THE REUABiUTY OF OPEN-CYCLE WATER SYSTEMS.An Evaluation Of Befoulmg Surveillance NUREG/CR4540- AN EVALUATION OF J-R CURVF TESTING OF NU-And Control Techniques For Use At Nuclear Power Plants. CLEAR PIPING MATERIALS USING THE DIRECT CURRENT POTEN-TIAL DROP TECHNIQUE.

Crack Arrest Disposal NUREG/CR4219 V03 N1: HEAVY SECTON STEEL TECHNOLOGY RAM E NUAL PAOGRESS REPORT FOR OCTOBER NUREG 1213: PLANS AND SCHEDULES FOR IMPLEMENTATON OF I U.S. NUCLEAR REGULATORY COMMISSION RESPONSIEiUTIES NUREG/CR4673: HEAVY SECTON MEEL TECHNOLOGY PROGRAM UNDER THE LOW-LEVEL RADIOACTIVE WASTE POLICY AMEND-FIVE YEAR PLAN FY19851989~ MENTS ACT OF 1985 (PL 99 240).

, Crack Growth NUREG/CR4162: SURVEY OF STATISTICAL AND SAMPUNG NEEDS i FOR ENVIRONMENTAL MONITORING OF COMMERCIAL LOW-NUREG/CR4219 V03 N1: HEAVY SECTON STEEL TECHNOLOGY LEVEL RADIOACTIVE WASTE DISPOSAL FACeUTIES.

PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1985

• MARCr4 I986. Dispoeol Site NUREG/CR4673: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM NUREG/CR4644: GEOCHEMICAL STUD'ES OF COMMERCIAL LOW-

. FIVE YEAR PLAN FY19851989.

1 LEVEL r4ADIOACTIVE WASTE DISPOSAL SITES. Topical Report.

l i

I 38 Subject Index Dose Electrical Degradetion NUREG/CR-4553. FIELD TESTS OF A PORTABLE TISSUE EOUlVA- NUREG/CR-4530: SUPERHEATED-STEAM TEST OF ETHYLENE PRO-LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND RADIATION FIELDS. ACCIDENT ENVIRONMENT.

Deee Consequence Calculation Electrical Equipment NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE NUREG/CR-4596: SCREENING TESTS OF REPRESENTATIVE NUCLE.

C-8. BOILING WATER REACTOR MAIN STEAM ISOLATION VALVE AR POWER PLANT COMPONENTS EXPOSED TO SECONDARY EN-LEAKAGE AND LEAKAGE TEEATMENT METHODS.

VIRONMENTS CREATED BY FIRES.

Dosimetry NUREG/CR-3320 V01: LWR PRESSURE VESSEL SURVEILLANCE DO. Electrical Fire SIMETRY IMPROVEMENT PROGRAM. PSF Expenments Summary And NUREG/CR4570: DESCRIPTION AND TESTING OF AN APPARATUS Bimd Test Results. FOR ELECTRICALLY INITIATED FIRES THRCUGH S!MULATON OF NUREG/CR-4439: LEPRICON ANALYSIS OF PRESSURE VESSEL SUR- A FAULTY CONNECTON.

VEILLANCE DOSIMETRY INSERTED INTO H.B. ROBINSON-2 DURING CYCLE 9. Electrically Initiated Fire NUREG/CR4530 V01: U.S./ FRENCH JOINT RESEARCH PROGRAM NUREG/CR-4570: DESCRIPTON AND TESTING OF AN APPARATUS REGARDING THE BEHAVOR OF POLYMER BASE MATERIALS SUS- FOR ELECTRICALLY INITIATED FIRES THPOUGH SIMULATION OF JECTED TO BETA RADIATON.Vol 1: Phase-1 Normahzation Results- A FAULTY CONNECTION.

Drtft Flux Model E

NU -465 PRE ON OF POOL VOID FRACTION BY NEW

~

SIMETRY IMPROVEMENT PROGRAM PSF Expenments Summary And Drilling Blind Test Results.

NUREG/CR4705: IDENTIFICATON OF NORTHWEST TRENDING SEISMOGENIC GRABEN NEAR CHARLESTON. SOUTH CAROLINA. Emergency Diesel Generator NUREG-1216: SAFETY EVALUATON REPORT RELATED TO THE Dry Containment OPERABILITY AND RELIABILITY OF EMERGENCY OtESEL GENERA-NUREG/CR-4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A TORS MANUFACTURED BY TRANSAMERICA DELAVAllNC.

DRY CONTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL MODEL. Emergency Planr%g Dry Containment Design NUREG/CP-0083: ANS TOPICAL MEETING ON RADIOLOGICAL ACCI-NUREG/CR4624 V05: RADIONUCLIDE RELEASE CALCULATIONS DENTS PERSPFCTIVES AND EMERGENCY PLANNING PROGRAM FOR SELECTED SEVERE ACCIDENI' SCENARIOS. Volume 51arge AND ABSTRACTS.

Enforcement Action Dynamic Analysie NUREG 0940 V05 NO2: ENFORCEMENT ACTONS.SIGNIFICANT AC.

NUREG/CR 4694: APPLICATON OF THERMIX-KONVEK CODE TO AC- TIONS RESOLVED.Quarterty Progress Report,Apnt-June 1986.

CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA-VURE NACTORS (HTRS). Engineered Safety Feature NUREG 1215: COMPILATON OF CONTRACT RESEARCH FOR THE CHEMICAL ENCiNEERING BRANCH.DIVISON OF ENGINEERING RE / R 702 POST-EOCENE FAULT NEAR EAST EDGE OF TECHNON. Annual Repod 6 W 1985.

REELFOOT RIFT IN LAUDERDALE COUNTY, TENNESSEE.AS DIS-ERED BY GRAVITY, EARTH RESISTIVITY SURVEYS AND Environmentalleeues NUREG/CR-4702: POST EOCENE FAULT NEAR EAST EDGE OF NUREG-1205: REACTIVATION OF NUCLEAR POWER PLANT CON-REELFOOT RIFT IN LAUDEHDALE COUNTY, TENNESSEE,AS D!S. STRUCTON PROJECTS. Plant Status, Policy lasues,And Regulatory COVERED BY GRAVITY, EARTH RESISTIVITY SURVEYS AND Optons.

Ethyleng Propylene Rubter (EPR) Cable Earthquake NUREG/CR4536: SUF ERHEATED-STEAM TEST OF ETHYLENE PRO-NUREG/CR 2800 SO4. GUOELINES FOR NUCLEAR POWER PLANT PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND SAFETY IS3UE PRIORIT12ATON INFORMATION DEVELOPMENT. ACCOENT ENVIRONMENT, NUREG/CR-3145 V04: GEOPHYSICAL INVESTIGATIONS OF WEST.

ERN OHIO 4NDIANA REGON. Annual Report (October 1984. Septem- Examinor Standarde NU E C -3805 V04: ENGINEERING CHARACTER'ZATON OF 21 m W M M M N M M M S GROUND MOTION. Task 11: Soif-Structure Interaction Effects On Struc* Escursions tural Response NUREG/CR-3967: AN ANALYSIS OF EXCURSONS AT SELECTED IN Earthquake Engineering SITU URANIUM MINES IN WYOMING AND TEXAS.

NUREG/CR-3005 V05: ENGINEERING CHARACTERIZATON OF GROUND MOTON. Task 11: Summary Report. E S Earthquake Hazard SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO

! NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE JANUARY 1966.

HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF TENNESSEE AN3 GEORGIA. Exposure Pathway Analysie NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE NUREG/CR4013: LADTAP 11. TECHNICAL REFERENCE AND USER HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF GUIDE.

TENNESSEE AND GEORGIA.

Effective Peak Acceleration NUREG/lA-0005: ASSESSMENT OF RELAP/ MOD 2, CYCLE 36,AGAINST NUREG/CR-3805 V04: ENGINEERING CHARACTERIZATON OF FIX-il SPLIT BREAK EXPERIMENT NO. 3027.

GROUND MOTON. Task li: Soil-Structure Interaction Effects On Struc-NU G/N5 V05: ENGINEERING CHARACTERIZATON OF Failure GROUND MOTON. Task 11: Summary Report. NUREG/CR4533: PROGRAM TO ANALYZE THE FAILURE MODE OF LEAD. ACID BATTERIES.

Effluente NUREG/CR4597 V01: AGING AND SERVICE WEAR OF AUXILIARY NUREG/CR4013: LADTAP 11. TECHNICAL REFERENCE AND USER FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol.

GUIDE. 1. Operating Exponence And Failure identification.

Subject index 39 Failure Cause NUREG/Ch4308: VANESA A MECHANISTIC MODEL OF RADIONU-Failure Mechanism CUDE RELEASE AND AEROSOL GENERATON DURING CORE DEBRIS INTERACTIONS WITH CONCRETE Failure Mode NUREG/CR4677: TRAP-MELT 2 CODE. DEVELOPMENT AND IMPROVE-NUREG/CR4099: AGE-RELATED DEGRADATION OF NATURALLY- MENT OF TRANSPORT MODEUNG.

AGED CLASS 1E BATTERY CELLS.

FW Produck Feet And Slow Physical Emplosion NUREG/CR-4624 V01: RADIONUCUDE RELEASE CALCULATONS NUREG/CR4593: INITIAL CONCEPTS ON ENERGETICS & MASS RE. FOR SELECTED SEVERE ACCIDENT SCENAROS. Volume LEASES DURING NONNUCLEAR EXPLOSIVE EVENTS IN FUEL 1.BWR. Mark i Desgn.

CYCLE FACIUTIES. NUREG/CR-4624 V02: RADIONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCOENT SCENARIOS. Volume 2.PWR,1ce Fatigue Condenser Design.

NUREG-1196: AN OVERVIEW CF ENVIRONMENTAL MATERIALS DEG- NUREG/CR4624 V03. RADIONUCUDE RELEASE CALCULATONS RADATON IN UGHT-WATER REACTORS. FOR SELECTED SEVERE ACCOENT SCENARIOS. Volume 3.PWR.Subatmosphenc Containment Desgn.

Fault NUREG/CR4624 V04: RADIONUCUDE RELEASE CALCULATIONS NUREG/CR4702: POST-EOCENE FAULT NEAR EAST EDGE OF FOR SELECTED SEVERE ACCfDENT SCENARIOS. Volume REELFOOT RIFT 'N LAUDERDALE COUNTY, TENNESSEE.AS DIS-4:BWR Mark lil Des'9n.

COVERED BY GRAVITY EARTH RESISTIVITY SURVEYS AND NUREG/CR4624 V05: RADIONUCLIDE RELEASE CALCULATIONS NU -4702: POST-EOCENE FAULT NEAR EAST EDGE OF REELFOOT RIFT IN LAUDERDALE COUNTY, TENNESSEE AS DIS-NUREG/CR4629 INDEPENDENT VERIFICATION OF RADIONUCUDE COVERED BY GRAVITY. EARTH RESISTIVITY SURVEYS AND RELEASE CALCULATONS FOR SELECTED ACCIDENT SCENARIOS.

DRfLUNG.

NUREG/CR4703' ONE DIMENSIONAL GRAVITY CALCULATION AND Five-Year Research Plan NU 47 O I I AV L LA N dND NUREG-1000 V03: LONG-RANGE RESEARCH PLAN FY 1967-FY 1991.

PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP, pg,,

Fault Reactivation NUREG/CR-4219 V03 N1: HEAVY 4ECTON STEEL TECHNOLOGY NUREG/CR4676: SE!SMIC REFLU TON GEOMETRY OF THE PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER NEWARK BASIN IN EASTERN PENT SYLVANIA.Ewdence For Exten. 1985 MARCH 1986, sional Peactivation Of Paleozonc Thrust Faults. NUREG/CR-4300 V03 Nu ACOUSTIC EMISSION / FLAW RELATION-Ship FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE Faulting VESSELS. Progress Report October 1965 March 1986.

NUREG/CR-4705: IDENTIFICATON OF NORTHWEST TRENDING NUREG/CR-4673: HEAVY 4ECTION STEEL TECHNOLOGY PROGRAM SEISMOGENIC GRABEN NEAR CHARLESTON. SOUTH CAROUNA. . FIVE YEAR PLAN FY1985-1989.

Faulty Electrecal Connection iMxible Morntwane uner NUREG/CR4570 DESCRIPTION AND TESTING OF AN APPARATUS NUREG/CR-3974: GEOMEMBRANE SELECTON CRITERIA FOR URA-i 4

FOR ELECTRICALLY INITIATED FIRES THROUGH SIMULATON OF NIUM VILL TAIUNGS PONDS.

A FAULTY CONNECTON.

Flood Erosion E /CR-4673: HEAVY SECTION STEEL TECHNOLOGY PROGRAM D T Nb / 21 3 VY4ECTIOt. STEEL TECHNOLOGY pig ,gg cg g.

PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1985 MARCH 1986. NUREG/CR-2800 SO4: GUCEUNES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRORITIZATION INFORMATON DEVELOPMENT.

Fleid Teete i

NUREG/CR-4553: FIELD TESTS OF A PORTABLE TISSUE EQUIVA. Focal Mechanisme LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA NUREG/C43145 V04: GEOPHYSICAL INVESTIGATONS OF WEST-RADIATON FIELDS. ERN OHtO-INDIANA REGON. Annual Report (October 1984 - Septem-ber 1985). .

Flim Boiling NUREG/lA4002: HEAT TRANSFER PROCESSES DURING INTERMEDI. Fracture Mechenico ATE AND LARGE BREAK LOCAS. NUREG/CR-4219 V03 N1: HEAVY 4ECTON STEEL TECHNOLOGY PROGRAM SEMlANNUAL PROGRESS REPORT FOR OCTOBER Final Environmental Statement 1985 - MARCH 1986.

NUREG-1171: FINAL ENVIRONMENTAL STATEMENT RELATED TO NUREG/CR4574; AN EXPERIMENTAL AND ANALYTICAL ASSESS-THE OPERATON OF SOUTH TEXAS PROJECT UNITS 1 AND MENT OF CIRCUMFERENTIAL THROUGH-WALL CRACKED PIPES

2. Docket Nos. 50 498 And 50-499(Houston Ughting And Power Com- UNDER PURE BENDING.

pany) NUREG/CR-4673; HEAVY-SECTION STEEL TECHNOLOGY PROGRAM Fmal Report - FIVE YEAR PLAN FY1985-1969.

NUREG/CR4699: APPUCATION OF ALTERNATING CURRENT PO-NUREG4956 REASSESSMEP OF THE TECHNICAL BASES FOR ES- TENTIAL DIFFERENCE TO CRACK LENGTH MEASUREMENT TIMATING SOURCE TERMS 6 inal Report.

DURING RAPID LOADING.

4 4

HUREG/CR-4660 ORGANIC COMPLEXANT ENHANCED MOBlUTY OF TOXIC ELEMENTS IN LOW LEVEL W ASTES. Final Report- Fracture Toughnees Finik Dmrence NUREG/CR3320 V01: LWR PRESSURE VESSEL SURVEILLANCE DO-NUREG/CR3262 V04: COBRA-NC:A THERMAL HYDRAULICS CODE SIMETRY IMPROVEMENT PROGRAM. PSF Expenments Summary And FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR Blind Test Resu'ta.

COMPONENTS.Vol 4. Users' Manual For Containment Analysis' Fractured Media Fire Darnage NUREG/CR-3925: SWIFT 11 SELF TEACHING CURRICULUM litustratrve NUREG/CR-4596: SCREENING TESTS OF REPRESENTATIVE NUCLE. Problems For Sandia Wastadsolation Flow And Transport Model For AR POWER PLANT COMPONENTS EXPOSEU TO SECONDARY EN. Fractured Media.

VIRONMENTS CREATED BY FIRES.

Fleelon Product NUREG-1206: ANALYSIS OF FRENCH (PALUEL) PRESSURIZED NUREG4956: REASSESSMENT OF THE TECHNICAL BASES FOR ES- WATER REACTOR DESIGN DIFFERENCES COMPARED TO CUR-TIMATING SOURCE TERMS. Foal Report RENT U.S. PWR DESIGNS.

40 Subject index NUREG/CR-4530 V01: U.S./ FRENCH JOINT RESEARCH PROGRAM Hand Calculations REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB- NUREG/CR-4656: VERIFICATION TEST CALCULATIONS FOR THE JECTED TO BETA RADIATION.Vol 1: Phase 1 Normahzation Results. SOURCE TERM CODE PACKAGE.

French P4 Health Effects NUREG 1206: ANALYSIS OF FRENCH (PALUEL) PRESSURIZED NUREG/CR-4628: RELATIVE AGE SPECIFIC RADIATON DCSE COM-CATER REACTOR DESIGN DIFFERENCES COMPARED TO CUR.

MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED RENT U.S. PWR DESIGNS.

FROM NUCLEAR FUEL FACIUTIES.

Fuel Cladding NUREG-1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG. Heat Flux RADATION IN UGHT-WATER REACTORS. NUREG/CR4568: COMPBRN til - A COMPUTER CODE FOR MODEL- l ING COMPARTMENT FIRES. I Functional Testing Heat Transfer NUREG/CR-4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-U EG 05: IDENTIFICATION OF NORTHWEST TRENDING ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

SEISMOGENIC GRABEN NEAR CHARLESTON. SOUTH CAROLINA.

Gas Detection Heavy-Section Steel Technology Program NUREG/CR-4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA. NUREG/CR4673: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES. - FIVE YEAR PLAN FY1985-1989.

NUREG/CR-4698: AIRBORNE LIDAR MAPPING OF SF6 CONCENTRA-TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES. High Frequency Damping NUREG/CR 4562: PIPE DAMPlNG-RESULTS OF V' AATION TESTS IN NU E / -4308: VANESA A MECHANISTIC MODEL OF RADIONU-33 m @ MZ NWN NE.

CUDE RELEASE AND AEROSOL GENERATION DURING CORE High-Level Weste Repositories DEBRIS INTERACTIONS WITH CONCRETE-NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCUDE Gaussian Quadrature GEOCHEMICAL INFORMATION DEVELOPED DY DOE HIGH-LEVEL NUREG/CP-0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For REACTOR PHYSICS AND SAFETY. Sessions 110. October 1984 -September 1985.

Generic lasue C-8 Hot-Gas Layer Model NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE NUREG/CR-4566. COMPBAN lli - A COMPUTER CODE FOR MODEL-CABOlWNG WATER REACTOR MAIN STEAM ISOLATION VALVE ING COMPARTMENT FIRES.

LEAKAGE AND LEAAAGE TREATMENT METHODS.

Geochemical Conditions NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCUDE NUREG/CR 2331.V05 N4: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For RESEARCH.Ouarterty Progress Report. October-December 1985.

October 1984 -September 1985. NUREG/CR-4532 Vot: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-AR POWER PLANT PERSONNELA Feashhty Study. Volume 1:5um-Geochemistry mary Of Results.

NUREG/CR4644. GEOCHEMICAL STUDIES OF COMMERCIAL LOW. NUREG/CR-4532 V02: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-LEVEL RADIOACTIVE WASTE DISPOSAL SITES.Topica! Report. AR POWER PLANT PERSONNELA Feashhty Study. Volume 2 Main Geologic Evaluatlon NUREG/CR4707: A PRELIMINARY GEOLOGIC EVALUATION OF THE Human Factors ALABAMA-TENNESSEF TRANSVERSE SEISMIC ZONE IN ALABAMA. NUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-Geomembranes AR POWER PLANT PERSONNELA Feasblity Study. Volume 1. Sum-NUREG/CR-3974. GEOMEMBRANE SELECTION CRITERIA FOR UAA- mary Of Results.

NIUM MILL TAluNGS PONDS. NUREG/CR4532 V02: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-AR POWER PLANT PERSONNEL:A Feasbhty Study.Voksne 2: Main Gravity Modeling Report.

NUREG/CR-4703. ONE DIMENSIONAL GRAVITY CALCULATION AND NUREG/CR-4600: HUMAN FACTORS STUDY CONDUCTED IN CON-PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP. JUNCTION WITH A MINI-ROUND ROBIN ASSESSMENT OF ULTRA.

NUREG/CR-4703 ONE DIMENSIONAL GRAVITY CALCULATION AND SONIC TECHNICIAN PERFORMANCE.

PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP.

Human Rollability NUREG/CR 4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-N EG/CR 3805 V04: ENGINEERING CHARACTER 12ATION OF GROUND MOTION. Task II: Soil-Structure Interaction Effects On Struc- AR POWER MT MRSONNELA FeasW StudWolume N tural Re se mary Of Results.

NUREG/N805 V05: ENGINEERING CHARACTERIZATION OF NUREG/CR-4532 V02. MODELS OF COGNITIVE BEHAVIOR IN NUCLE-GROUND MOTION. Task 11: Summary Report. AR POWER PLANT PEF.SONNELA Feasbhty Study. Volume 2. Main Report.

Groundwater Flow NUREG/CR-3925: SWIFT 18 SELF TEACHING CURRICULUM lllustrative Hydrogen Combustion Problems For Sandia Waste-isolation Flow And Transport Model For NUREG 1215: COMPILATION OF CONTRACT RESEARCH FOR THE Fractured Media. CHEMICAL ENGINEERING ERANCH. DIVISION OF ENGINEERING TECHNOLOGY. Annual Report For FY 1985.

NUREG/CR 2800 SO4: GUIDELINES FOR NUCLEAR POWER PLANT ICAP Code Assessment SAFETY ISSUE PRCRITIZATION INFORMATION DEVELOPMENT. NUREG/IA-0005: ASSESSMENT OF RELAP/ MOO 2. CYCLE 36.AGAINST Gully Erosion FIX Il SPUT BREAK EXPERIMENT NO. 3027.

NUREG/CR-4480: EROSION PROTECTON OF URANIUM TAluNGS IM- NUREG/IA4007: ASSESSMENT OF RELAP5/ MOD 2 AGA:NST CRITI-POUNDMENTS. CAL FLOW DATA FROM TESTS JIT 11 AND CFT 21.

HTRs Ice Condenser NUREG/CR-4694: APPLICATION OF THERMIX-KONVEK CODE TO AC. NUREG/CR 4624 V02: RADIONUCUDE RELEASE CALCULATIONS CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2.PWR,lce TURE REACTORS (HTRS). Condenser Design.

Subject Index -41 Igniter Systems NUREG/CR-4530 V01: U.S3 FRENCH JOINT RESEARCH PROGRAM NUREG 1215: COMPILATON OF CONTRACT RFSEARCH FOR THE REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-CHEMICAL ENGINEERING BRANCH,DIVISILM OF ENGINTERING JECTED TO BETA RADIATON.Vol 1: Phase-1 Normaltzstion Resulta.

i TECHNOLOGY. Annual Report For FY 19415. NUREG/CR-4873: HEAVY SECT:ON STEEL TECHNOLOGY PROGRAM l . FIVE YEAR PLAN FY19851989.

Importance Analysis NUREG/CR 4688 V01:QUANTIFICATIGN AND UNCERTAINTY ANALY- J-R Curve SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT NUREG/CR-4540: AN EVALUATION OF J-R CURVE TESTING OF NU-WATER REACTORS (OUASAR).Part 1: Methodology And Program CLEAR PIPING MATERIALS USING THE DIRECT CURRENT POTEN-Plan. TIAL DROP TECHNIQUE.

In Situ Kinetico NUREG/CR-3967: AN ANALYSIS OF EXCURSIONS AT SELECTED IN NUREG/CR-4308: VANESA A MECHANISTIC MODEL OF RADIONU-SITU URANIUM MINES IN WYOMING AND TEXAS.

CUDE RELEASE AND AEROSOL GENERATON DURING CORE in Vivo Lung Cotnting DEBRIS INTERACTIONS WITH CONCRETE.

NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPUCATONS TO BOASSAY AT URANIUM MILLS.

'A ^

N REG CR-4013. LADTAP ll - TECHNICAL REFERENCE AND USER Index GUIDE.

i NUREG-0304 VII NO2: REGULATORY AND TECHNICAL NUREG/CR-4013: LADTAP 11 - TECHNICAL REFERENCE AND USE1 REPORTS.Compilabon For Sacond Quarter 1986,Apni-June. GUIDE.

Indirect 84oassay LEPRICCN NUREG4874: INTERNAL DOSIMETRY MODEL FOR APPUCATIONS NUREG/CR-4439: LEPPICON ANALYSIS OF PRESSURE VESSEL SUR-TO BIOASSAY AT URANIUM MILLS. VEILLANCE DOSIMETRY INSERTED INTO H.B. ROBINSON-2 DUR1NG CYCLE 9.

NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPUCATIONS LER TO BIOASSAY AT URANIUM MILLS. NUREG/CR-2000 VOS N6: UCENSEE EVENT REPORT (LER)

COMPILATON For Month Of June 1986.

UR CR 600 HUMAN FACTORS STUDY CONDUCTED IN CON-JUNCTON WITH A MINI-ROUND ROBIN ASSESSMENT OF ULTRA-MP F[Mo A s '

j SONIC TECHNICIAN PERFORMANCE- LIDAR inservice Monitoring NUREG/CR-4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA-i NUREG/CR4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION. TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES.

' SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA-VESSELS. Progress Report, October 1985 - March 1986. TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES.

Inspection LLRWPAA NUREG/CR-4597 V01: AGING AND SERVICE WEAR OF AUX 1UARY NUREG 1213: PLANS AND SCHEDULES FOR IMPLEMENTATON OF FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol. U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBILITIES 1.Operaung Expenence And Failure Identification. UNDER THE LOW-LEVEL RADIOACTIVE WASTE POUCY AMENLN MENTS ACT OF 1985 (PL 99-240). s inspection And Enforcement NUREG/CR-4560: RGK ASSESSMENT APPUCATION TO NRC IN. LOCAs e

SPECTON PROGRESS REPORT FOR PEROD JANUARY 1985 TO NUREG/lA-0002: HEAT TRANSFER PROCESSES DURING INTERMEDI-JANUARY 1986. ATE AND LARGE BREAK LOCAS.

Inspector Reliability Lameller Teoring NUREG/CR-4600: HUMAN FACTORS STUDY CONDUCTED IN CON- NUREG/CR-2800 SO4: GUIDELINES FOR NUCLEAR POWER PLANT JUNCTION WITH A MINI ROUND ROBIN ASSESSMENT OF ULTRA- SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.

SONIC TECHNICIAN PERFORMANCE.

Latin Hypercube Sampl6ng Instrument Air unos NUREG/CR-4688 V01: OUANTIFICATION AND UNCERTAINTY ANALY.

NUREG/CR 2800 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT SIS OF SOURCE TERMS FOR SEVERE ACCOENTS IN LIGHT SAFETY ISFUE PRORITIZATON INFORMATION DEVELOPMENT. WATER REACTORS (QUASAR).Part 1: Methodology And Program Integrated Plant Safety Assessment NUREG-0827 S01: INTEGRATED PLANT SAFETY ASSESSMENT SYS- Leachate TEMATIC EVALUATON PROGRAM - LACROSSE BOILING WATER NUREG/CR-4644: GEOCHEMICAL STUDIES OF COMMERCIAL LOW-REACTOR. Docket No. 50-409. (Daryland Power Cooperat ve) LEVEL RADCACTIVE WASTE DISPOSAL SITES. Topical Report.

, Interlocks Lead-Acid Betteries NUREG/CR-2800 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT NUREG/CR-4533: PROGRAM TO ANALYZE THE FAILORE MODE OF SAFETY ISSUE PRIORITIZATON INFORMATION DEVELOPMENT.

LEAD-ACID BATTERIES.

Internal Doolmetry NUREG-08 4 IN ERNAL S ETRY MODEL FOR APPUCATONS NU G16 TE L FINDINGS RELATED TO GENERIC ISSUE C-8.BOluNG WATER REACTOR MAIN STEAM ISOLATION VALVF International Cooperative Test Program LEAKAGE AND LEAKAGE TREATMENT METHODS.

NUREG/CR-4530 V01: U.S1 FRENCH JOINT RESEARCH PROGRAM REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB- Legalleeuences JECTED TO BETA RADIATON.Vol 1: Phase 1 Normalizabon Results. NUREG 0750 V23 NOS: NUCLEAR REGULATORY COMMISSON IS-SUANCES FOR MAY 1986. Pages 465-575.

Inventory Difference Data NUREG4430 V06 N02- UCENSED FUEL FACIUTY STATUS Ucensed Operetsng Reactore REPORT. inventory Difference Data. July-Uecember 1985 (Gray Book 11). NUREG-0020 V10 M05: UCENSED OPERATING REACTORS STATUS

SUMMARY

REPO4T. Data As Of April 30,1986.(Gray Book)

Irradiation NUREG4020 V101"06: LICENSED OPERATING REACTORS STATUS NUREG/CR-4219 V03 Nt: HEAVY-SECTION STEEL TECHNOLOGY

SUMMARY

REPORT. Data As Of May 31,1986.(Gray Booy

. PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER NUREG 0020 V10 N06 UCENSED OPERATING REACTORS STATUS l 1985 - MARCH 1966.

SUMMARY

REPORT. Data As Of May 31,1986 (Gray Book)

- - =.

42 Subject index Licensee Contractor And Vendor Inspection Maintenance And Surveillance ProJram (MSP)

NUREG-0040 V10 NO2: UCENSEE CONTRACTOR AND VENDOR IN- NUREG-12t2 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR SPECTON STATUS REPORT. Quarterfy Report.Apni-June 1986 (White POWER INDUSTRY 1985. Volume 1. Findings And Conclusions.

Book)

Mark i Design Licensee Event Report NUREG/CR-4624 V01: RADIONUCLIDE RELEASE CALCULATIONS MUREG/CR-2000 V05 N6: UCENSEt: EVENT REPORT (LER) FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume COMPILATION-For Month Of June 1986 1.BWR, Mark 1 Design.

NUREG/CR-2000 V05 N6: LICENSEE EVENT REPORT (LER)

COMPILATION For Month Of June 1986- Muk lli Deel n MU EG/CR VOS N8 UCE EE EVENT REFORT (LER) NUREGICR-4624 V04. RADIONUCLIDE RELEASE CALCULATIONS

^"9"* '

FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume Light-Water Reactors 4.BWR, Mark lli Design.

NUREG-1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG-RADATION IN UGHT WATER REACTORS. Marviken Test NUREG/C43858: TRAC-PF1/ MOD 1:AN ADVANCED BEST-ESTIMATE NUREG/lA-0007: ASSESSMENT OF RELAPS/ MOD 2 AGAINST CRITI-COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR CAL FLOW DATA FROM TESTS JIT 11 AND CFT 21.

THERMAL-HYDRAUUC ANALYSIS.

Mass Flow Rates Measurement Lineament Analysie NUREG/CR4725: TWO-PHASE FLOW MEASUREMENT IN THE UPPER NUREG/CR-4707: A PREUMINARY GEOLOGIC EVALUATION OF THE PLENUM OF A PWR DURING REFLOOD.

ALABAMA-TENNESSEE TRANSVERSE SEISMIC ZONE IN ALABAMA.

Neit-Concrete Heat Transfer Liquid Effluent NUREG/CR-4521: TURC2 AND 3.LARGE SCALE UO2/ZRO2 MELT-NUREG/CR2907 V04: RADIOACTIVE MATERIALS RELEASED FROM CONCRETE INTERACTION EXPERIMENTS AND ANALYSIS.

NUCLEAR POWER PLANTS. Annual Report 1983.

Mett-Concrete Interaction Emportment N G/ 396 AN ANALYSIS OF EXCURSIONS AT SELECTED IN 52t M2 AND 3NE SCAE W2@O2 W-SITU URAN!UM MINES IN WYOMING AND TEXAS. CONCRETE INTERACTON EWERWENTS AND ANAMS.

Long-Range Research Plan Mett/Cancrete Interactione NUREG-1080 V03: LONG-RANGE RESEARCH PLAN FY 1987-FY 1991. NUniG/CR-4308: VANESA:A MECHANISTIC MODEL OF RADIONU-CUDE RELEASE AND AEROSOL GENERATON DURING CORE Lose Of Feedwater DEBRIS INTERACTIONS WITH CONCRETE.

NUREG/CP 0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON REACTOR PHYSICS AND SAFETY. Sessions 110. Metabollem NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATON DOSE COM-Low Level Waste MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED WUREG/CR4162: SURVEY OF STATISTICAL AND SAMPUNG NEEDS FROM NUCLEAR FUEL FAC!UTIES.

FOR ENVIRONMENTAL MONITORING OF COMMERCIAL LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITIES. Metallurgical Blind Test Low-Level Radioactive Weste NUREG/CR-3320 V01: LWR PRESSURE VESSEL SURVEILLANCE DO-WUREG 1213: PLANS AND SCHEDULES FOR IMPLEMENTATION OF E*"'"*" "'"'"*T U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBluTIES hnd hst Resuh UNDER THE LOW-LEVEL RADOACTIVE WASTE POLICY AMEND- Mineralog6 cal Charactedsstion MENTS ACT OF 1985 (PL 99-240)

NUREG/CR-4236: PROGRESS IN EVALUATON OF RADIONUCUDE Low-Level waste GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUREG/CR4644: GEOCHEMICAL STUDIES OF COMMERCIAL LOW- NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For LEVEL RADIOACTIVE WASTE DISPOSAL SITES. Topical ReprA October 1984 -September 1985.

MAEROS Code Model Minimum Film Boiling Temperature NUREG/CR4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A NUREG/tA4002: HEAT TRANSFER PROCESSES DURING INTERMEDI-DRY CONTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL ATE AND LARGE BREAK LOCAS.

MODEL Model Evolustion MARCH NUREG/CR-4698: AIRBORNE UDAR MAPPING OF SF6 CONCENTRA.

NUREG/CR4656: VERIFICATION TEST CALCULATIONS FOR THE TION DISTRIBUTIONS FOR TRANSPORT AND DIFFUSON STUDIES.

SOURCE TERM CODE PACKAGE. NUREG/CR4698: AIRBORNE UDAR MAPPING DF SF6 CONCENTRA-TION DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES.

NUREG/CR 2331 VOS N4: SAFETY RESEARCH PROGRAMS SPON- Model Validation SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4677. TRAP-MELT 2 CODE: DEVELOPMENT AND IMPROVE-RESEARCH.Ouarterty Progress Report, October-December 1985.

MENT OF TRANSPORT MODELING.

MELCOR Computer Code bodeling NUREG/CR-4487: UNCERTAINTY AND SF.NSITIVITY ANALYSIS OF A NUREG/CR4677: TRAP-MELT 2 CODE. DEVELOPMENT AND IMPROVE-DRY CONTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL MODEL MENT OF TRANSPORT MODELING.

gggy g ,g,,, Modular High Temperature Gae Cooled Reactor NUREG 1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE NUREG/CR-4694: APPUCATION OF THERMIX KON%EK CODE TO M-C-8. BOILING WATER REACTOR MAIN STEAM ISOLATION VALVE CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA-LEAKAGE AND LEAKAGE TREATMENT METHODS. TURE REACTORS (HTRS).

Main Steam leoletion Valves (MSiv) Molten Core-Concrete Interaction NUREG 1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE NUREG/CR4521: TURC2 AND 3 LARGE SCALE UO2/ZRO2 MELT.

C-8,801UNG WATER REACTOR MAIN STEAM ISOLATION VALVE CONCRETE INTERACTION EXPERIMENTS AND ANALYSIS.

LEAKAGE AND LEAKAGE TREATMENT METHODS.

Maintenance NUREG/CR-4162: SURVEY OF STATISTICAL AND SAMPLING NEEDS NUREG 1212 V02 STATUS OF MA!NTENANCE IN U.S. NUCLEAR IN- FOR ENVIRON 64 ENTAL MONITOhlNG OF COMMERCIAL LOW.

DUSTRY 1985. Volume 2.Descnptions Of Prngrams And Practices. LEVEL RADIOACTIVE WASTE DISPOSAL FACluTIES.

l

Subject index 43 NUREG/CR-4553: FIELD TESTS OF A PORTABLE TISSUE EQUIVA- NUREG/CR-4125 V02 R1: GUIDELINES AND WORKBOOK FOR AS.

LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTiu-RADIATION FIELDS.

TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLA Volume 2. Workbook For Assessment Of Organizatson And Man.

NUREG/CR-4688 V01: OUANTIFICATION AND UNCERTAINTY ANALY-SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT Organizationel Assesernent WATER REACTORS (OUASAR).Part 1: Methodology And Program Plan. NUREG/CR-4125 V01 R1: GUIDELINES AND WORKBOOK FOR AS-SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTiu-Mothballing Practice TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG-1205: REACTIVATION OF NUCLEAR POWER PLANT CON. PLANT. Volume 1:Guidehnes For Utikty Orgaruzaton And Administraton STRUCTION PROJECTS. Ptant Status,Pohey issues,And Regulatory Plan.

Wns.

Overland Erosion Net-Section Collapse

' NUREG/CR-4480: EROSION PROTECTION OF URANIUM TAluNGS IM-NUREG/CR-457a: AN EXPERIMENTAL AND ANALYTICAL ASSESS. POUNDMENTS.

MENT OF CIRCUMFERENTIAL THROUGH-WALL CRACKED PIPES UNDER PURE BENDING. P-Wave Residuals Nondestructive Evaluation NUREG/CR-3145 V04: GEOPHYSICAL INVESTIGATIONS OF WEST.

ERN OHIO-INDIANA REGION. Annual Report (October 1984 - Septern.

NUREG/CR4600: HUMAN FACTORS STUDY CONDUCTED IN CON- ber 1985). -

JUNCTION WITH A MINI-ROUND ROBIN ASSESSMENT OF ULTRA.

SONIC TECHNICIAN PERFORMANCE. PSF Experiments NUREG/CR-3320 V01: LWR PRESSURE VESSEL SURVEILLANCE DO-R 12 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLEAR POWER INDUSTRY 1985 Volume 1 Findings And Conclusons. Ind Te t Re NUREG 1212 V02: STATUS OF MAINTENANCE IN U.S. NUCLEAR IN-DUSTRY 1985. Volume 2-Descnptions Of Programs And Practices. Paleozoic Thrust Faults NUREG/CR-4676. SEISMIC REFLECTION GEOMETRY OF THE Nuclear Power Plant Management NEWARK BASIN IN EASTERN PENNSYLVANIAEvidence For Exten-NUREG/CR-4125 V01 R1: GUIDELINES AND WORKBOOK FOR AS. sonal Reactivation Of Paleozoec Thrust Faults.

SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTiu-TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER Paluel PLANT. Volume 1:Guidehnes For Utikty Organizaton And Admnatration NUREG 1206: ANALYSIS OF FRENCH (PALUEL) PRESSURIZED Plan.

WATER REACTOR DESIGN DIFFERENCES COMPARED TO CUR-NUREG/CR-4125 V02 R1: GUIDELINES AND WORKBOOK FOR AS. RENT U.S. PWR DESIGNS.

SESSMENT OF ORGANI2ATION AND ADMINISTRATION OF UTILi-TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER Pebble Bed Reactor PLANT. Volume 2. Workbook For Assessment Of Organizaten And Man. NUREG/CR-4694. APPUCATION OF THERMIX.KONVEK CODE TO AC-agement.

CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA-TURE REACTORS (HTRS).

NUREG/CR-4308: VANESA:A MECHANISTIC MODEL OF RADIONU- Performance Evaluation CUDE RELEASE AND AEROSOL GENERATION DURING CORE NUREG 1220: TRAINING REVIEW CRITERIA AND PROCEDURES.

DEBRIS INTERACTIONS WITH CONCRETE.

ORMGEN Permeability Testing NUREG/CR4633 ORPLOT.PC: A GRAPHIC UTILITY FOR ORMGEN.PC NUREG/CR-4738. SIZE INFLUENCE ON THE SEALING PERFORM.

AND ORVIRT PC. ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

ORPLOT Petitions For Rulemaking NUREG/CR4633: ORPLOT.PC: A GRAPHIC UTILITY FOR ORMGEN.PC NUREG-0936 VOS N01: NRC REGULATORY AGENDA.Ouarterly AND ORVIRT.PC. Report. January-March 1986.

ORVlRT Pipe Damaging NUREG/CR-4633: ORPLOT.PC: A GRAPHIC UTILITY FOR ORMGEN.PC NUREG/CR-4562 PlPE DAMPING-RESULTS OF VIBRATION TESTS IN AND ORVIRT.PC. THE 33 TO 100 HERTZ FREQUENCY RANGE.

Oganizational Assessment Pipes NUREG/CR-4125 V02 R1: GUIDEUNES AND WORKBOOK FOR AS- NUREG-0313 R02 DRFT: TECHNICAL REPORT ON MATERIAL SELEC-SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTiu- TlON AND PROCESSING GUIDELINES FOR BWR COOLANT PRES-TIES SEErlNG OPERATING UCENSE FOR A NUCLEAR POWER SURE BOUNDARY FIPING.

PLANT. Volume 2: Workbook For Assessment Of Organization And Man-agement. NUREG/CR-4574: AN EXPERIMENTAL AND ANALYTICAL ASSESS-MENT OF CIRCUMFERENTIAL THROUGH-WALL CRACKED PIPES Open Cycle Water System Fouling NUREG/CR-4626 V01: IMPROVING THE RELIABILITY OF OPEN- Piping CYCLE WATER SYSTEMS.An Evaluation Of Bofouhng Survedlance And Control Techniques For Use At Nuclear Power Plants. NUREG-1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG-RADATION IN LIGHT. WATER REACTORS.

NUREG/CR-4529- PIPING DAMPING - EXPERIMENTAL RESULTS NU E 0874; INTERNAL DOSIMETRY MODEL FOR APPLICATIONS FROM LABORATORY TESTS IN THE SEISMIC RANGE.

TO BIOASSAY AT URANtUM MILLS Plans And Schedules Organic Complexants NUREG-1213: PLANG AND SCHEDULES FOR IMPLEMENTATION OF NUREG/CR-4660: ORGANIC COMPLEXANT-ENHANCED MOBluTY OF U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBILITIES TOX1C ELEMENTS IN LOW-LEVEL WASTES.Finet Report. UNDER THE LOW LEVEL RADIOACTIVE WASTE POUCY AMEND-Organization MENTS ACT OF 1985 (PL 99-240).

NUREG/CR-4125 V01 R1: GUIDEUNES AND WORKBOOK FOR AS. Plutons SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTILI- NUREG/CR-4703: ONE DIMENSIONAL GRAVITY CALCULATION AND TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP.

PLANT. Volume 1:Guidehnes For Utility Organizaton And Administraton Plan. NUREG/CR-4703: ONE DIMENSIONAL GRAVITY CALCULATION AND PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP,

44 Subject Index Polonium Purge Valves NUREG/CR4628: RELATIVE AGE-SPECIFIC RADIATION DOSE COM- NUREG/CR4648; A STUDY OF TYPICAL NUCLEAR CONTAINMENT MITMENT FACTORS FOR MAJOR RADIONUCLIDES RELEASED PURGE VALVES IN AN ACCIDENT ENVIRONMENT.

FROM NUCLEAR FUEL FACILITIES.

OUASAR .

Pool Void Fraction NUREG/CR 4688 V01: OUANTIFICATION AND UNCERTAINTY ANALY-NUREG/CR 4657. PREDICTION OF POOL VOID FRACTION BY NEW SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT DRIFT FLUX CORRELATION. WATER REACTORS (QUASAR).Part 1: Methodology And Program Plan.

Power Plant Operators NUREG-1123 KNOWLEDGE AND ABILITIES CATALOG FOR NUCLEAR POWER PLANT OPERATOR. BOILING WATER REACTORS. N EG 0 VOS NC ENFORCEMENT ACTIONS SIGNIFICANT AC-Practice And Procedures D6 gest TIONS RESOLVED aarterty Progress Report.Apni June 1986.

NUREG,0386 004 R02: UNITED STATES NUCLEAR REGULATORY NUREG/CR-4678. A ETHOD FOR USING PRA TO ESTABLISH OUAL-COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. JULY ITY PROGRAM AP. tCABlUTY.

1,1972 DECEMBER 31,t986.

P RW Pressure Transmitters NUREG/CR-2331 V05 N4: SAFETY RESEARC' 0F% RAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY Pressure Vessel RESEARCH Ovarterfy Progress Report, October-December 1985.

• NUREG/CR4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION-SH'P FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE Ouarterty Report VESSELS Proaress Report, October 1985 March 1986 NUREG-0936 VOS NOI: NRC REGULATORY AGENDA.Ouarterfy NUREG/CR-4673: HEAVY-SF.CTION STFEL TECHNOLOGY PROGRAM Report. January-March 1986.

- FIVE YEAR PLAN FY1985-1989 RAMONA-38 Pressure Vessel Surveillance NUREG/CR-2331 V05 N4 SAFETY RESEARCH PROGRAMS SPON-NUREG/CR-3320 V01. LWR PRFSSURE VESSEL SURVEfLLANCE DO- OFFICE OF NUCLEAR REGULATORY SIMETRY IMPROVEMENT PHO 3 RAM FSF Evenments Summary And SORED BY RESEARCH Ouarterty Progress Report. October December 1985.

Blind Tesi Results RELAP5/ MOD 2 -

PrTsure Vessels NUREG/CR4219 V03 N1. HEAVY-SECTION STEEL TECHNOLOGY NUREG/lA4005: ASSESSMENT OF RELAP/ MOD 2. CYCLE 36.AGAINST PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER FIX-Il SPUT BREAK EXPERIMENT NO. 3027.

1985 - MARCH 1986. NUREG/lA4007: ASSESSMENT OF RELAPS/ MOD 2 AGAINST CRITI-CAL FLOW DATA FROM TESTS JIT 11 AND CFT 21.

PrCasurtzed Thermal Shock NUREG/CP4080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON Radiation Dose REACTOR PHYSICS AND SAFETY. Sessions 110. NUREG/CR-4013: LADTAP 1t - TECHNICAL REFERENCE AND USER PrCasurized Water Reactor NUREG/CP-0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON Radiation Safety Program REACTOR PHYSICS AND SAFFTY Sessions 1 10- NUREG-0940 V05 NO2: ENFORCEMENT ACTIONS,SIGNtFICANT AC-NUREG/CP-0080 %02: PROCEEDINGS OF THE TOPICAL MEETING ON TIONS RESOLVED.Ouarterly Progress Report.Apnadune 1986.

REACTOR PHYSICS AND SAFETY.Sess.ons 1116.

NUREG/CR3858; TRAC-PFI/ MODI.AN ADVANCED BEST-ESTIMATE Radiation Survey Meter COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR NUREG/CR-4553: FIELD TESTS OF A PORTABLE TISSUE EQUIVA.

THERMAL MORAUUC ANALYSIS.

LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA Probabilistic Risk Assessment RADIATION FIELDS.

RUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-AR POWER PLANT PERSONNELA Feasibility StudyVolume 1: Sum- Radiological Safety mary Of Results. NUREG-1198 S01: RELEASE OF UF6 FROM A RUPTURED MODEL NUREG/CR-4532 V02: MODELS OF COGNITIVE BEHAVIOR IN NUCLE- 48Y CYUNDER AT SEQUOYAH FUELS CORPORATION AR POWEH PLANT PERSONNEL.:A Feasebikty Study. Volume 2. Main FACILITY Lessons-Learned Report. NRC Staff Responses To The Rec-Peport. ommendations Made By The Lessons-Learned Group.

NUREG/CR-4560: RISK ASSESSMENT APPUCATION TO NRC IN-SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO Radionuclide JANUARY 1986. NUREG/CR-4308: VANESA:A MECHANISTIC MODEL OF RADIONU-NUREG/CR-4678: A METHOD FOR USING PRA TO ESTABLISH OUAL- CLIDE RELEASE AND AEROSOL GENERATION DURING CORE ITY PROGRAM APPLICABluTY. DEBRLS INTERACTIONS WITH CONCRETE.

Probability Risk Assessment Radlonuclide Release NUREG/CR-2331 VOS N4: SAFETY RESEARCH PROGRAMS SPON- NUREG/CR-4624 V01: RADIONUCUDE RELEASE CALCULATIONS SORED BY OFFICE OF NUCLEAR REGULATORY FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume RESEARCH. Quarterly Progress Report. October-December 1985 1:BWR. Mark i Design.

Prol$lem Solving NUREG/CR-4624 V02: RADIONUCLIDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2.PWR,1ce NUREG/CR-4532 V01: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-POWER PLANT PERSONNEL % Feassbehty Study Volume 1 Sum- V03: RADIONUCUDE RELEASE CALCULATIONS NU /C 4 NUREG/CR-4532 V02: MODELS OF COGNITIVE BEHAVIOR IN t*UCLE. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 3.PWR.Subatmosphenc Containment Desagn.

AR POWER PLANT PERSONNELA Feasibility Study. Volume 2. Main Report.

NUREG/CR-4624 V04. RADIONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume Progress Report , 4.BWR. Mark til Design.

NUREG/CR-4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION- NUREG/CR-4624 V05: RADIONUCLIDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5.Large SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS Progress Report, October 1985 March 1986. Dry Containment Design.

Public Law 99-240 Radlonuclide Solubility NUREG-1213. PLANS AND SCHEDULES FOR IMPLEMENTATION OF NUREG/CR-4236: PROGRESS IN EVALUATION OF RADIONUCUDE U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBluTIES GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL UNDER THE LOW-LEVEL RADIOACTIVE WASTE POLICY AMEND- NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For MENTS ACT OF 1985 (PL 99-240). utotwr 1984 -September 1985.

Subject index 45

- Radionucilde Sorption Regulation HUREG/CR4236: PROGRESS IN EVALUATON OF RADIONUCUDE NUREG-1205: REACTIVATION OF NUCLEAR POWER PLANT CON-GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL STRUCTION PROJECTS. Plant Status. Policy Issues,And Regulatory NUCLEAR WASTE REPOSITOR @ SITE PROJECTS. Annual Report For Options.

October 1984 -September 1985.

Regulatory Agenda Radionuct6de Transport NUREG-0936 V05 N01: NRC REGULATORY AGENDA.Quarterty

! NUREG/CR-3925: SWIFT !! SELF-TEACHING CURRICULUSA.Illustratrve Report. January-March 1986.

t Problems For Sandia Waste-Isolation Flow And Transport Model For Fractured Media. Regulatory And Technical Repor*e NUREG-0304 Vit N02: REGULATORY AND TECHNICAL Radionuclides -

REPORTS. Compilation For Second Quarter 1986, April-June.

NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATION DOSE COM-MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED Regulatory Guide 1 FROM NUCLEAR FUEL FACILITIES. NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE .

C8;BOluNG WATER REACTOR MAIN STEAM ISOLATION VALVE Radium LEAKAGE AND LEAKAGE TREATMENT METHODS.

NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATION DOSE COM-MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED Release FROM NUCLEAR FUEL FACILITIES. NUREG/CR4629: INDEPENDENT VERIFICATON OF RADIONUCUDE

RELEASE CALCULATIONS FOR SELECTED ACCIDENT SCENARIOS.

Reactor Containment (NUREG/CR4336: REVIEW OF 117 RESEARCH INSTITUTE REPORTS , Rollability ON STRUCTURAL STUDIES'OF REACTOR CONTAINMEM. NUREG/CR 2800 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT.

Reactor Coolant Pump (RCP) NUREG/CR-4560: RISK ASSESSMENT APPUCATION TO NRC IN-NUREG/CR4643. EVALUATION OF CORE DAMAGE SEQUENCES INI- SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO TIATED BY LOSS OF REACTOR COOL. ANT PUMP SEAL COOUNG. JANUARY 1986.

Reactor Coolant System Remote Sensing NUREG/CR-4677. TRAP-MELT 2 CODE: DEVELOPMENT AND IMPROVE- NLMEG/CR-4698: AIRBORNE LlDAR MAPPING OF SF6 CONCENTRA-MENT OF TRANSPORT MODELING. TION DISTRIBUTIONS FOR TRANSFORT AND DIFFUSION STUDIES.

NUREG/CR-4698: AIRBORNE LIDAR MAPPING OF SF6 CONCENTRA-Reactor Physics TION DISTRIBUTONS FOR TRANSPORT AND DIFFUSION STUDIES.

NUREG/CP4080 V01. PROCEEDINGS OF THE TOPICAL MEETING ON i

REACTOR PHYSICS AND SAFETY. Sessions 1 10. Repooltory Sealing NUREG/CR4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-Reactor Physics And Safety ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

• NUREG/CP-0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON ,

REACTOR PHYSICS AND SAFETY. Sessions 110. Rosin Domineralizer System NUREG/CP-0080 V02- PROCEEDINGS OF THE TOPICAL MEETING ON NUREG/CR-2800 SO4: GUIDELINES FOR NUCLEAR POWER PLANT REACTOR PHYSICS AND SAFETY Sessions 11 16. SAFETY ISSUE PRIORITIZATON INFORMATON DEVELOPMENT.

Reactor Safety Risk NUREG 1206: ANALYSIS OF FRENCH (PALUEL) PRESSURIZED NUREG/CR-4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A WATER REACTOR DESIGN DIFFERENCES COMPARED TO CUR- DRY CONTAINMENT TEST PROBLEM FOR THE MAEROS AEROSAL RENT U S. PWR DESIGNS. MODEL Reactor Vessel Rock Mechanics NUREG 1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG- NUREG/CR-4738: St2E INFLUENCE ON THE SEAUNG PERFORM-RADATION IN UGHT WATER REACTORS. ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

Realistic Fission Product Transport Rock Riprep NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE NUREG/CR4480: EROSION PROTECTION OF URANIUM TAILINGS IM-C-8. BOILING WATER REACTOR MAIN STEAM ISOLATION VALVE POUNDMENTS.

LEAKAGE AND LEAKAGE TREATMENT METHODS.

Rules Of Practice Red Oil Explosion NUREG-0386 004 R02: UNITED STATES NUCLEAR REGULATOHY NUREG/CR-4593: INITIAL CONCEPTS ON ENERGETICS & MASS RE- COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST, JULY LEASES DURING NONNUCLEAR EXPLOSIVE EVENTS IN FUEL 1,1972 DECEMBER 31,1986.

CYCLE FACIUTIES.

SMACS Code Reelfoot Rift NUREG/CR 2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON.

NUREG/CR-4702: POST EOCENE FAULT NEAR EAST EDGE OF SORED BY OFFICE OF NUCLEAR REGULATORY REELFCOT AlFT IN LAUDERDALE COUNTY, TENNESSEE.AS DIS- RESEARCH. Quarter 9 Progress Report, October December 198S.

COVERED BY GRAVITY, EARTH RESISTMTY SURVEYS AND DRILLING. STCP ,

NUREG/CR-4702: POST-EOCENE FAUT T NEAR EAST EDGE OF NUREG/CR4656: VERIFICATON TEST CALCULATIONS FOR THE REELFOOT RIFT IN LAUDERDALE CCUNTY, TENNESSEE.AS DIS- SOURCE TEFM CODE PACKAGE.

COVERED BY GRAVITY, EARTH RESISTMTY SURVEYS AND '

DRILUNG. SUPER CANON NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER Reelfoot Scarp CANNON AND HDR EXPERIMENTAL DATA.

NUREG/CR4703: ONE DIMENSIONAL GRAVITY CALCULATIOd AND j

PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP. SWIFT 11 1

NUREG/CR-4703: ONE DIMENSIONAL GRAVITY CALCULATION AND NUREG/CR-3925: SWIFT 11 SELF TEACHING CURRICULUM.lliustrative PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP. Problems For Sandia Waste 4 solation Flow And Transport Model For Fractured Media.

NUREG/CR4706: A STUDY OF SEISMICITY AND EARTHOUAKE Safety HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF NUREG-1205: REACTIVATION OF NUCLEAR POWER PLANT CON-TENNESSEE AND GEORGIA. STRUCTION PROJECTS. Plant Status.Pohey lasues,And Regulatory NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE Options.

MZARD IN NORTHERN Al ABAMA AND ADJACENT PARTS OF NUREG/CR-2800 SO4. GUIDEUNES FOR NUCLEAR POWER PLANT TENNESSEE AND GEORGIA. SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOFMENT I

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

.46 Subject Index .

. NUREG/CR-4445. EFFECTIVENESS AND SAFETY ASPECTS OF SE. NUREG/CR-4707: A PRELIMINARY GEOLOGIC EVALUATON OF THE LLCTED DECONTAMINATON PROCESSES. ALABAMA TENNESSEE TRANSVERSE SEISMIC ZONE IN ALABAMA.

I Saf;ty Evaluation Solomogenic Structure NUREG4940 V05 NO2- ENFORCEMENT ACTIONS.SIGNIFICANT AC* NUREG/CR-4705: IDENTIFICATION OF NORTHWEST TRENDING SI N BEN MAR CHARESEN,SM NUM-MUR 1 43 SO SA TY V A lC fD T LAT TO THE FULLTERM OPERATING UCENSE FOR MILLSTONE NUCLEAR Self Teaching Curriculum POWER STATON. UNIT NO.1. Docket No. 50-245.(Northeast Nuclear NUREG/CR-3925: SWIFT 11 SELF-TEACHING CURRICULUM.lliustrative Energy CompanW Problems por Sandia Waste-Isolabon Flow And Transport Model For Safety Evaluation Report Fractured Media.

NUREG 0853 S06: SAFETY EVALUATION REPORT RELATED TO THE 3

OPERATION OF CUNTON POWER STATON. UNIT NO.1. Docket No.

Semiannual %grose Report 50-481 (Ilkr'ois Power Company.et al) NUREG/CR-4219 V03 N1: HEAVY-SECTION STEEL TECHNOLOGY NUREG4887 S10- SAFETY EVALUATION REPORT RELATED TO THE PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER OPERATON OF PERRY NUCLEAR POWER PLAN 1, UNITS 1 AND 1965 MARCH 1986.

2 Docket Nos 50 440 And 50441.(Cleveland Electric illuminatir.g Com-panj Sensittvity Analye6e NURts0896 SOS: SAFETY EVALUATON REPORT RELATED 10 THE NUREG/CR-4598: A USER'S GUIDE FOR THE TOP EVENT MATRIX OPERATON OF SEABROOk STATION UNITS 1 AND 2. DOCKET Nos. ANALYSIS CODE (TEMAC).

50443 And 50444 (Pul$c Service Company Of New Hampshire Et Al) NUREG/CR4688 V01: OUANTIFICATION AND UNCERTAINTY ANALY-NUREG 1047 S03: SAFETY EVALUATON REPORT RELATED T6 THE SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT OPERATION OF NINE MILE POINT NUCLEAR STATION, UNIT NO-

~

WATER REACTORS (OUASAR).Part 1: Methodology And Program ,

2 Cocket No. 50410. (Niagara Mohawk Power Corporation) Plan. - -

NUREG-1048 S06: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF HOPE CREEK GENERATING STATON. Docket No. Service Water System 50-354.(Pubhc Sennce Electnc And Gas Company, Atlantic City Electnc NUREG/CR-2800 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT NU NIN7 S02: SAFETY EVALUATION REPORT PELATED TO SAFETY THE ISSUE PRORITIZATION INFORMATION DEVELOPMENT.

OPERATION OF BEAVER VALLEY POWER STATION. UNIT 2. Docket NU EG 1 7 So EV b EPORT RELATED TO THE N REG REASSESSMENT OF THE TECHNICAL BASES FOR ES-OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 TIMATING SOURCE TERMS. Final Report AND 2 Docket Nos. 50-424 And 50-425 (Georaia Power Company.et al) NUREG/CR-2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON-NUREG-1143 S01: SAFETY EVALUATION REITORT RELATED TO THE SORED BY OFFICE OF NUCLEAR REGULATORY 1 FULL-TERM OPERATING LICENSE FOR MILLSTONE NUCLEAR RESEARCH.Ouarterty Progress Report, October-December 1985.

POWER STATION, UNIT NO.1. Docket No. 50-245(Northeast Nuclear NUREG/CR 4624 V01: RADIONUCLIDE RELEASE CALCULATIONS Eruray Company) FOR SELECTED SEVERE ACCIDENT SCENAROS. Volume NUREG;1216: SAFETY EVALUATION REPORT RELATED TO THE 1:BWR. Mark 1 Design.

OPERABILITY AND REUABluTY OF EMERGENCY DIESEL GENERA- NUREG/CR-4624 V03: RADIONUCUDE RELEASE CALCULATIONS TORS MANUFACTURED BY TRANSAMERICA DELAVAL,1NC. FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume NUREG/CR 4694: APPUCATON OF THERMIX-KONVEK CODE TO AC- 3.PWR.Subatmosphenc Containment Design.

CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA

• NUREG/CR-4624 V04: RADIONUCUDE RELEASE CALCULATIONS TURE REACTORS (HTRS). FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume Safety Research 4.BWR. Mark til Design.

NUREG/CR-2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON- NUREG/CR-4624 V05: RADIONUCLIDE RELEASE CALCULATIONS SORED BY OFFICE OF NUCLEAR REGULATORY FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5.Large RESEARCH,Ouarterly Progress Report. October-December 1985.

NU E 4 0 PENDENT VERIFICATION OF RADIONUCUDE Seismic RELEASE CALCULATIONS FOR SELECTED ACCIDENT SCENARIOS.

NUREG/CR-4703. ONE DIMENSONAL GRAVITY CALCULATION AND NUREG/CR4688 Vot: OUANTIFICATION AND UNCERTAINTY ANALY-PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP. SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT NUREG/CR4703: ONE DIMENSIONAL GRAVITY CALCULATION AND WATER REACTORS (QUASAR).Part 1: Metrodology And Program PALEOZOIC STRUCTURE AND PLUTONS AT REELFOOT SCARP. Plan.

Seismic Analyele Severe Accident Program

NL' REG /CR4529. PIPING DAMPING EXPERIMENTAL RESULTS NUREG 1080 V03

LONG-RANGE RESEARCH PLAN FY 1987-FY 199t.

FROM LABORATORY TESTS IN THE SEISMIC RANGE.

Severe Accidente 4 99: AGE-RELATED DEGRADATION OF NATURALLY- NUREG/CR-4624 V02: RADIONUCUDE RELEASE CALCULATIONS AGED CLASS 1E BATTERY CELLS. FOR SELECTED SEVERE ACCIDENT SCENARIOS. volume 2.PWR,1ce NUREG/GR-4659 VOI: SEISMIC FRAGIUTY OF NUCLEAR POWER Condenser Design.

PLANT COMPONENTS (PHASE 1).

Setem6c Reflection Geometry NUREG-1215: COMPILATON OF CONTRACT RESEARCH FOR THE i NUREG/CR-4676: SEISMIC REFLECTION GEOMETRY OF THE CHEMICAL ENGINEERING BRANCH.DIVISON OF ENGINEERING i NEWARK BASIN IN EASTERN PENNSYLVANIAEv6dence For Exten- TECHNOLOGY. Annual Report For FY 1985.

sonal Reactivation Of Paleoroic Thrust Faults.

Seit G/C -4  : A STUDY OF SEISMICITY AND EARTHOUAKE 626 M MWNG THE REUABW & N HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF An va a n n0 Wlance And Control Technques For Use At Nuclear Power Plants.

TENNESSEE AND GEORGIA.

NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE j EN SS E A D GE GIA UR CR 0  : ENGINEERING 'CHARACTERl2ATION OF GROUND MOTION. Task 11: Soil-Structure Interaction Effects On Struc-Setemicity tural Response.

i NUREG/CR4706: A STUDY OF SEISMICITY AND EARTHOUAKE NUREG/CR-3805 V05: ENGINEERING CHARACTERl2ATION OF i l HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF GROUND MOTION.Tesk II: Summary Report I TENNESSEE AND GEORGIA.

NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHQUAKE Solid Weete Oleposal HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF NUREG/CR-2907 V04: RADIOACTIVE MATERIALS RELEASED FROM TENNESSEE AND GEORGIA. NUCLEAR POWER PLANTS. Annual Report 1983.

l e

Subject index 47 Sound veiocity Prodletione Strese Corrosion NUREG/lA4001: ASSESSMENT OF TRAC-PD2 USING SUPER NUREG-0313 R02 DRFT: TECHNICAL REPORT ON MATERIAL SELEC.

CANNON AND HOR EXPERIMENTAL DATA. "ON AND PFsXESSING GUIDELINES FOR BWR COOLANT PRES-Surs NNDARY PtPING.

l NUREG-0956: REASSESSMENT OF THE TECHNICAL BASES FOR ES- Strees Corrosion Creding l TIMATING SOURCE TERMS Final Report I NUREG/CR-2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON-NUREGICR2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON. SORED BY OFFICE OF NUCLEAR REGULATORY SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.

RESEARCH.Ouarterly Progress Report, October-December 1965.

Pr ss R October-December 1985.

NUREG/CR-4624 VO R ION IDE RELEASE CALCULATONS Structural Response FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2.PWR,lce NUREG/CR-3805 V04: ENGINEERING CHARACTERl2ATION OF NU CR-4 V03: RADIONUCUDE RELEASE CALCULATONS GROUND MOTON. Task it Soil-Structure Interaction Effects On Strue.

tural Response, FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 3:PWR.SGw64-.4 Containment De . gg'" - --~'~ -

NUREG/CR-4629: INDEPENDENT VERIF TION OF RADIONUCUDE RELEASE CALCULATONS FOR SELECTED ACCIDENT SCrNARIOS. NUREG/CR-4624 V03: RADIONUCUDE RELEASE CALCULATONS NUREG/CR4644 GEOCHEMICAL STUDIES OF COMMERCu.L LOW. FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume LEVEL RADOACTIVE WASTE DISPOSAL SITES.T al Report. 3.PWR Subatmospheric Contarvnent Design.

NUREG/CR4688 V01: OUANTIFICATION AND UNO RTAINTY ANALY.

SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT Summary Report WATER REACTORS (OUASAR).Part 1: Methodology And Program NUREG/CR-3805 V05: . ENGINEERING CHARACTERIZATION OF Pan. GROUND MOTON. Task it Summary Report Source Term Code Package (STCP) Super System Code NUREG/CR4629: INDEPENDENT VERIFICATON OF RADIONUCUDE NUREG/CR2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON-RELEASE CALCULATIONS FOR SELECTED ACCOENT SCENARIOS. SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Ouarterly Progress Report, October-Decernber 1985.

NUREG/CR4656: VERIFICATION TEST CALCULATONS FOR THE Superheated - Steam Test SOURCE TERM CODE PACKAGE. NUREG/CR-4536: SUPERHEATED-STEAM TEST OF ETHYLENE PRO.

PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND REG R 4624 V01: RADIONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume Suppreselon Pool

^

NU E 624  : RADONUCUDE RELEASE CALCULATIONS FOR SELECTED SEVERE ACCOENT SCENARIOS. Volume CWE RESE AM E'^ M SNM MM M -

DEBRIS INTERACTONS WITH CONCRETE.

4 BWR. Mark lli NUREG/CR4624 V : RADIONUCUDE RELEASE CALCULATONS FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 5.Large r

ntaw Nm U -4626 V01: IMPROVING THE REUABluTY OF OPEN-CYCLE WATER SYSTEMS.An Evaluation Of Beofouling Surveillance Spettel Yartation And Control Techniques For Use At Nuclear Power Plants.

NUREG/CR-3805 V05: ENGINEERING CHARACTERIZATON OF GROUND MOTION. Task it Summary Report 8*

NUR /CR4659 V01: SEISMIC FRAGIUTY OF NUCLEAR POWER Steinlese Steel PLANT COMPONENTS (PHASE 1).

NUREG4313 R02 DRFT: TECHNICAL REPORT ON MATERIAL SELEC-TON AND PROCESSING GUIDEUNES FOR BWR COOLANT PRES- Synthetic Linere SURE BOUNDARY PIPING. NUREG/C43974: GEOMEMBRANE SELECTON CRITERIA FOR URA-NIUM MILL TAILINGS PONDS.

- Station Blackout NUREG/CP.0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON Systematic Evolustion Program REACTOR PHYSICS AND SAFETY. Sessions 1-10 NUREG-0827 S01: INTEGRATED PLANT SAFETY ASSESSMENT SYS-NUREG/CP-0080 V02: PROCEEDINGS OF THE TOPICAL MEETING ON . TEMATIC EVALUATON PROGRAM LACROSSE BOfLING WATER REACTOR FHYSICS AND SAFETY. Sessions 1116. REACTOR.Decket No. 50-409. (Daryland Power Cooperative)

Statistice Systems Approach To Training NUREG/CR4162 SURVEY OF STATISTICAL AND SAMPUNG NEEDS NUREG-1220- TRAINING REVIEW CRITERIA AND PROCEDURES.

FOR ENVIRONMENTAL MONITORING OF COMMERCIAL LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACIUTIES. TACT lil '

NUREG 1169- TECHNICAL FINDINGS RELATED TO GENERIC ISSUE E

AN OVERVIEW OF ENVIRONMENTAL MATER!ALS DEG-RADATION IN UGHT WATER REACTORS.

1EMAC N E CR 3 N4: SAFETY RESEARCH PROGRAMS SPON- NUREG/CR4590: A USER'S GUIDE FOR THE TOP EVENT MATRI'X 6 A SIS CODE UEM

• SORED BY OFFICE OF NUCLEAR REGULATORY

. RESEARCH Ouarterty Progress Report, October-December 1985.

THERMIX MONVEK Stochaotic Flow NUREG/CR 4694: APPUCATION OF THERMIX-KONVEK CODE TO AC-NUREG/CR4622: VALIDATION OF STOCHASTIC FLOW AND TRANS. CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA-PORT MODELS FOR UNSATURATED SOILS.A Comprehensive Feld TURE REACTORS (HTRS).

TNT-Equivalent Concept Stratigraphy NUREG/CR-45W3: INITIAL CONCEPTS ON ENERGETICS & MASS RE.

NUREG/CR4705: IDENTIFICAff0N OF NORTHWEST TRENDING LEASES DURING NONNUCLEAR EXPLOSIVE EVENTS IN FUEL SEISMOGENIC GRABEN NEAR CHARLESTON. SOUTH CAROUNA. CYCLE FACluTIES.

Stress Analyele

• TRAC-PD2 NUREG/CR4633: ORPL(7T.PC: A GRAPHIC UTILITY FOR ORMGEN PC NUREG/lA-0001: ASSESSMENT OF TRAC.PD2 USING SUPFR AND ORVIRT.PC. CANNON AND HDR EXPERIMENTAL DATA.

i 1

d 48 Subject Index TRAC-PF1/ MOO 1 Traneamerece Delevel,inc NUREG/CR-3858: TRAC-PF1/ MOD 1 AN ADVANCED BEST-ESTIMATE NUREG 1216: SAFETY EVALUATION REPORT RELATED TO THE COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR OPERABluTY AND REUABILITY OF EMERGENCY DIESEL GENERA.

THERMAL-HYDRAUUC ANALYSIS. TORS MANUFACTURED BY TRANSAMERICA DELAVAL.INC.

TRAP-MELT Trenaient Putes Teste NUREG/CR-4656. VERIFICATION TEST CALCULATIONS FOR THE NUREG/CR-4738: SIZE INFLUENCE ON THE SEAUNG PERFORM-SOURCE TERM CODE PACKAGE. ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

TRAP-MELT 2 Code Tranelent Reactor Analysis Code (TRAC)

NUREG/CR-4677: TRAP MELT 2 CODE DEVELOPMENT AND IMPROVE- NUREG/CR-3858: TRAC-PF1/MODtAN ADVANCED BEST-ESTIMATE MENT OF TRANSPORT MODELING.

COMPUTER PROGRAM FOR PRESSURIZED . WATER REACTOR

) TURC2 And 3 THERMAL-HYDRAUUC ANALYSIS.

NUREG/CR-4521: TURC2 AND 3.LARGE SCALE UO2/ZRO2 MELT

• Trenomittwo CONCRETE INTERACTION EXPERIMENTS AND ANALYSIS.

NUREG/CR-4659 V01: SEISMIC FRAGILITY OF NUCLEAR POWER Technical W PLANT COMPONENTS (PHASE 1). I i NUREG-0940 VOS N02: ENFORCEMENT ACTIONS:SIGNIFICANT AC- Trenepwt i NURE 1202 TE N E CA FO H CR EK GEN- NUREG/CR-4677: TRAP-MELT 2 CODE. DEVELOPMENT AND IMPROVE-  !

ERATING STATION. Docket No. 50-354. (Public Service E?actne & Gas MENT OF TRANSPORT MODELING.

Company Of New Jersey)

Trenoport Theory j Thermal Hydraulic Trenaient Code NUREG/CR4647: CALCULATON OF THE POWER DISTRIBUTION IN i

' NUREG/CR-3262 V04: COBRA-NC A THERMAL HYDRAUUCS CODE THE VENUS PWR MOCK-UP BENCHMARK USING TWO-GROUP DIF-FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR FUSION THEORY.

COMPONENTS.Vol 4. Users' Manual For Containment Analysis.

Tm 7

Thermal Hydraulice NUREG 1196: AN OVERVIEW OF ENVIRONMENTAL MATERIALS DEG-3 NUREG/CP4080 V02 PROCEEDINGS OF THE TOPICAL MEETING ON RADATiON IN UGHT WATER REACTORS.

REACTOR PHYSICS AND SAFETY. Sessions 1116.

Tm h Thermal Shocle NUREG/CR-4725: TWO PHASE FLOW MEASUREMENT IN THE UPPER NUREG/CR-4219 V03 N1: HEAVY SECTON STEEL TECHNOLOGY PLENUM OF A PWR DURING REFLOOD.

PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER i 1985 - MARCH 1986 UMresonic Testing 1

NUREG/CR-4673 HEAVY-SECTON STEEL TECHNOLOGY PROGRAM NUREG/CR-4C30: HUMAN FACTORS STUDY CONDUCTED IN CON-

- FIVE YEAR PLAN FY1985-1989. JUNCTION WITH A MINI-ROUND ROBIN ASSESSMENT OF ULTRA-1' SONIC TECHNICIAN PERFORMANCE-Thermel-Hydraulic Anr;yele i'

NUREG/CR-3858: TRAC-PF1/ MODI AN ADVANCED BEST-ESilMATE Uncertainty Analyele COMPUTER PROGRAM FOR PRESSURIZED WATER REACTOR NUREG/CR-4598: A USER'S GUIDE FOR THE TOP EVENT MATRIX THERMAL-HYDRAUUC ANALYSIS. ANALYSIS CODE (TEMAC).

NUREG/CR-4688 V01: QUANTIFICATON AND UNCERTAINTY ANALY- t Thermel-Hydraulic Reactor Safety SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT l NUREG/CR-2331 V05 N4: SAFETY RESEARCH PROGRAMS SPON- WATER REACTORS (OVASAR). Pert 1: Methodology And Program SonED BY OFFICE OF NUCLEAR REGULATORY Plan.

1 RESEARCH.Ouarterly Progress Report, October December 1985.

Uneaturated Zone ThermochenD NUREG/CR-4622: VAUDATION OF STOCHASTIC FLOW AND TRANS-NUREG/CR-4308. VANESA.A MECHANISTIC MODEL OF RADIONU- PORT MODELS FOR UNSATURATED SOfLS A Comprehensive Field CUDE RELEASE AND AEROSOL GENERATON DURING CORE Studt DEBRIS INTERACTONS WITH CONCRETE.

Uranium NUREG/CR-4628: RELATIVE AGE-SPECIFIC RADIATION DOSE COM-NUREG/CR4628. RELATIVE AGE-SPECIFIC RADIATION DOSE COM-MITMENT FACTORS FOR MAJOR RADONUCUDES RELEASED MITMENT FACTORS FOR MAJOR RADIONUCUDES RELEASED FROM NUCLEAR FUEL FAC!UTIES FROM NUCLEAR FUEL FACluTIES.

Uranium Biosemey 4 4 AN EXPERIMENTAL AND ANALYTICAL ASSESS-TO Bl S AY URANI L

• MENT OF CIRCUMFERENTIAL THROUGH-WALL CRACKED PIPES UNDER PURE BENDING-Uranium HeseHuoride Rolesse Title List NUREG-1198 SOU RELEASE OF UF6 FROM A RUPTURED MODEL NUREG4540 V08 N05: TITLE UST OF DOCUMENTS MADE PUBUCLY 48Y CYUNDER AT SEQUOYAH FUELS CORPORATON AVAILABLE.MAY 1-31,1986 FACluTY. Lessons-Leemed Report. NRC Staff Responses To The Roc-NUREG-0540 V08 N06: TITLE'UST OF DOCUMENTS MADE PUBLICLY ommendatons Made By The Lessons-Leemed Group.

AVAILABLE. JUNE 1-30, 1986.

NURE 540 7 I Ur UST OF DOCUMENTS MADE PUBUCLY NUR G R-3 GEOMEMBRANE SELECTION CRITERIA FOR URA-NIUM MIL 4 TAILINGS PONDS.

j Top Event Metrin Analreis Code j NUREG/CR 4598: A USERS GUIDE FOR THE TOP EVENT MATRIX Uranium Mines

/.NALYSIS CODE (TEMAC). NUREG/CR-396h AN ANALYSIS OF EXCURSONS AT SELECTED IN SITU URANIUM MINES IN WYOMING AND TEXAS.

Tinic Element Migration NUREG/CR-4660: ORGANIC COMPLEXANT-ENHANCED MOBluTY OF Uranium Tallinge : . aL.to c TOXIC ELEMENTS IN LOW LEVEL WASTES. Final Report. NUREG/CR-4480: EROSION PROTECTION OF URANIUM TAluNGS IM-

POUNDMENTS.

! Treening 1

NUREG 1123: KNOWLEDGE AND A81UTIES CATALOG FOR NUCLEAR User Guide POWER PLANT OPERATOR BOILING WATER REACTORS. NUREG/CA-4013; LADTAP 11 TECHNICAL REFERENCE AND USER NUREG 1220: TRAINING REVIEW CRITERIA AND PROCEDURES. GUIDE-1

- , . .--f , . * - - . . - c., - - - , . , - - - . - , , - - - ,- e .,,...-#-._--~ , - ,

Subject Index 49 VANESA Vital Instrumentation NUREG/CR-4308: VANESA.A MECHANISTIC MODEL OF RADIOMU- NUREG/CR-4470: SURVEY AND EVALUATION OF VITAL INSTRUMEN-CLIDE RELEASE AND AEROSOL GENERATON DURING CORE ' TATON AND CONTROL FOWER SUPPLY EVENTS.

DEBRIS INTERACTIONS WITH CONCRETE. NUREG/CR-4470: SURVEY AND EVALUATION OF VITAL INSTRUMEN-NUREG/CR-4656: VERIFICATON TEST CALCULATONS FOR THE TATION AND CONTAOL POWER SUPPLY EVENTS.

SOURCE TERM CODE PACKAGE.

VENUS PWR Emportment NUREG 1215: COMPILATON OF CONTRACT RESEARCH FOR THE NUREG/CR-4647: CALCULATON OF THE POWER DISTRIBUTON IN CHEMICAL ENGINEERING BRANCH. DIVISION OF ENGINEERING THE VENUS PWR MOCK-UP BENCHMARK USING TWO. GROUP DIF. TECHNOLOGY. Annual Report For FY 1985.

FUSION THEORY.

,,,g __

! Vaportantion NUREG/CR-4236: PROGRESS IN EVALUATON OF RADIONUCLIDE NUREG/CR-4308: VANESA:A MECHANISTIC MODEL OF RADONU- GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL CLfDE RELEASE AND AEROSOL GENERATON DURING CCRE NUCLEAR WASTE REPOSITORY SITE PROJECTS. Annual Report For DEBRIS INTERACTONS WITH CONCRETE. October 1984 -September 1985.

l G CR .A MECHANISTIC MODEL OF RADIONU- UR 4102: AIR CURRENTS DRIVEN BY SPRAYS IN REACTOR CLIDE RELEASE AND AEROSOL GENERATON DURING CORE CONTAINMENT BUILDINGS.

DEBRIS INTERACTIONS WITH CONCRETE- Water Cooled Reactor NUREG/CP 0080 V02: PROCEEDINGS OF THE TOPICAL MEETING ON R 80; EROSION PROTECTION OF URANIUM TAILINGS IM- '

POUNDMENTS. Weldment NUREG/CR 4219 V03 N1: HEAVY-SECTION STEEL TECHNOLOGY C/ -3 45 V04. GEOPHYSICAL INVESTIGATONS OF WEST.

ERN OHIO INDIANA REGION. Annual Report (October 1984 Septem-

. RCH NUREG/CR-4873; HEAVY-SECTION STEEL TECHNOLOGY PROGRAM ber 1985).

. FIVE YEAR PLAN FY19851989.

Verification Calculations Wetting State NUREG/CR-4656. VERtFICATION TEST CALCULATONS FOR THE NUREG/CR-4738: SIZE INFLUENCE ON THE SEALING PERFORM-SOL.,RCE TERM CODE PACKAGE. '

ANCE OF CEMENTITIOUS BOREHOLE PLUGS.

Vibration Test Yellowcake NUREG/CR4562: PIPE DAMPING-RESULTS C,F VIBRATION TESTS IN NUREG-0874: INTERNAL DOSIMETRY MODEL FOR APPLICATIONS THE 33 TO 100 HERTZ FREQUENCY RANGE. TO BIOASSAY AT URANIUM MILLS.

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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). II 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.

EDO OFFICE OF ADMINISTRATION EDO-RESOURCE MANAGEMENT DIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL OFFICE OF RESOURCC MANAGEMENT. DIRECTOR NUREG-0304 VII NO2: REGULATORY AND TECHNICAL NUREG-1145 V02 U.S. . NUCLEAR REGULATORY COMMISSION REPORTS Compdaten For Second Quarter 1986,Apnl-June. 1985 ANNUAL REPORT.

NUREG-0540 VOS N05: TITLE LIST OF DOCUMENTS MADE PUBLIC- DIVISION OF BUDGEY & ANALYSIS LY AVAl ABLE.MAY 1-31, 1986.

NUREG4540 V08 N06: TITLE UST DF DOCUMENTS MADE PU,?UC- NUREG-0020 V10 N05: UCENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of April 30,19861 Gray Book)

LY AVAILABLE. JUNE 1 30,1986-NUREG-0020 V10 N06: UCENSED OPERATING REACTORS STATUS NUREG-0540 V08 N07: TITLE UST OF DOCUMENTS MADE PUBLIC-

SUMMARY

REPORT. Data As Qf May 31,1086.(Gray Book)

LY AVAILABLE JULY 1-31, 1986.

NUREG-0750 V23 N05: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR MAY 1986. Pages 465-575- OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

NUREG4750 V23 N06; NUCLEAR REGULATORY COMMISSON IS- OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST SUANCES FOR JUNE 1986. Pa a 577-883. 851125)

NUREG-1205: REACTIVATON OF NUCLEAR POWER PLANT CON-N VO 01 N REGULATORY AGENDA.Ouarterly CTON NEN M Status % issuesM Wam ep anuaparch 1986^

NUR 216: SAFETY EVALUATION REPORT RELATED TO THE OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) OPERABILITY AND RELIABluTY OF EMERGENCY DIESEL GEN-OFFICE OF INSPECTION & ENFORCEMENT. DIRECTOR (POST FRATORS MANUFACTURED BY TRANSAMERICA DELAVAL.INC.

820201) DIVISION OF HUMAN FACTORS TECHNOLOGY (POST 851125)

NUREG4430 V06 NO2: LICENSED FUEL FACT! IV STATUS NUREG 1021 R03: OPERATOR UCENSING EXAMINER STAND.

REPORT. inventory Difference DataJuly-December 1985.(Gray Book ARDS.

11). NUREG-1123; KNOWLEDGE AND A81UTIES CATALOG FOR NUCLE-NUREG-0940 V05 NO2- ENFORCEMENT ACTIONS SIGNIFICANT AC- AR POWER PLANT OPERATOR. BOLLING WATER REACTORS.

TiONS RESOLVED Ouarterly Progress Report.Aptd June 1986. NUREG-1212 V01: STATUS OF MAINTENANCE IN THE U.S. NUCLE.

DIVISON OF OA, VENDOR & TECHNICAL TRAINING CENTER PRO- AR POWER INDUSTRY 1985. Volume 1: Findings And Conclusions.

GRAMS (POST 85021 NUREG 1212 V02: STATUS OF MAINTENANCE IN U.S. NUCLEAR IN-NUREG4040 V10 NO2. UCENSEE CONTRACTOR AND VENDOR IN- DUSTRY 1985. Volume 2. Descriptions Of Programs And Practices.

SPECTON STATUS REPORT. Quartedy Report Apni-June NUREG 1220: TRAINING REVIEW CRITERIA AND PROCEDURES.

1986 (White Dook)

DIVISION OF PRESSURIZED WATER REACTOR LICENSING . A DIVISION OF INSPECTON PROGRAMS (POST 850212) (POST 851125)

NUREG-1196- AN OVERVIEW OF ENVIRONMcNTAL MATERIALS NUREG-0781 S01: SAFETY EVAUJATION REPORT RELATED TO DEGRADATION IN LIGHT-WATER REACTORS- THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 AND OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS et Nos. W98 M M1 (Houston Ug% W h DIVISION Oc WASTE MANAGEMENT Company)

NUREG-1213: PLANS AND SCHEDULES FOR IMPLEMENTATION OF NUREG4896 S05: SAFETY EVALUATON REPORT RELATED TO U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBluTIES THE OPERATON OF SEABROCk STATOftVNITS 1 AND UNDER THE LOW LEVEL RADIOACTIVE WASTE POUCY AMEND. 2. DOCKET Nos. 50-443 And 50-444.(Public Serwce Company Of MENTS ACT OF 1985 (PL 99-240). New Hampsture.Et Al)

NUREG/CR-3585 V02; DE MiNiMlS WASTE IMPACTS ANALYSIS NUREG-1057 S02: SAFETY EVALUATON REPORT RELATED TO MET 10DOLOGY.Vulume 2. Impacts-BRC Users Guide And Method. THE OPERATON OF BEAVER VALLEY POWER STATON, UNIT ology For Radioactive Wastes Below Regulatory Concern.Thrs is An 2. Docket No. 50-412.(Duquesne Ught Company.et al)

NRC Staff Report Pubhshed As Volume 2 Of Contractor Report NUREG-1137 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF VOGTLE ELECTRIC GENERATING U.S. NUCLEAR REGULATORY COMMISSION PLANT UNITS 1 AND 2. Docket Nos. 50 424 And 50-425.(Georgia OFFICE OF THE GENERAL COUNSEL Power Company,et al)

NUREG-0386 004 R02: UNITED STATES NUCLEAR REGULATORY NUREG 1171: FINAL ENVIRONMENTAL STATEMENT RELATED TO COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. JULY THE OPERATION OF SOUTH TEX /.S PROJECT UNITS 1 AND NRC hE A LED F' N GIVEN NUREG 1198 S01: RELEASE OF UF6 FROM A RUPTURED MODEL a)

DIVISION OF PRESSURIZED WATER REACTOR LICENSING . B 48Y CYUNDER AT SEQUOYAH FUELS CORPORATION (POST 851125)

FACILITYlessons-Learned Report. NRC Staff Responses To The Ricommendates Made Dy The Lessons-Leamed Group. NUREG 1143 Sot: SAFETY EVALUATION REPORT RELATED TO THE FULL TERM OPERATING UCENSE FOR MILLSTONE NUCLE.

A WER STATONMNIT M t Docket M 92454Ndeast OFFICE OF NUCLEAR REGULATORY RESEARCH FOST 4/05/81 OFFIC NUCLEAR PrGULATORY RESEAR H. DIRECTO (PRE DIVISI F IN ER REACTOR (BWR) UCENSING NUREG-0874 INTERNAL DOSIMETRY MODEL FOR APPUCATONS NUREG-0313 RO2 DRFT: TECHNICAL REPORT ON MATERIAL SE.

TO BIOASSAY AT URANIUM MILLS LECTION AND PROCES$1NG GUIDELINES FOR BWR COOLANT NURE G-0956 REASSESSMENT OF THE TECHNICAL BASES FOR PRESSURE BOUNDARY PIPING.

ESTIMATING SOURCE TERMS Final Report. NUREG-0827 S01: INTEGRATED PLANT SAFETY ASSESSMENT NUREG 1080 V03: LONG-RANGE RESEARCH PLAN FY 1987-FY SYSTEMATIC EVALUATION PROGRAM . LACROSSE BOILING 1991 WATER REACTOR Docket No. 50-409. (Dairyland Power Coopers-DIVISION OF ENGINEERING TECHNOLOGY (PRE 860720) tnre)

NUREG 1215: COMPILATION OF CONTRACT RESEARCH FOR THE NUREG4853 S06: SAFETY EVALUATION REPORT RELATED TO CHEMICAL ENGINEERING BRANCH. DIVISION OF ENGINEERING THE OPERATON OF CUNTON POWER STATION, UNIT NO.

TECHNOLOGY Annual Report For FY 1985. 1. Docket No. 50-461.(llhnois Power Company,et al) 51 4

4 0

52 NRC Originating Organization Index NUREG4353 S07: SAFETY EVALUATION REPORT RELATED TO STATION Docket No. 50-354 (Puble Servce Electre And Gas THE OPERATION OF CLINTON POWER STATION UNIT NO. Company,Atlante City Electnc Cornpany) 1 Docket No. 50461. (llhnois Power Company.et al) NUREG-1169: TECHNICAL FINDINGS RELATED TO GENERIC ISSUE NUREG4887 S10: SAFETY EVALUATION REPORT RELATED TO C-8,80lLING WATER REACTOR MAIN STEAM ISOL/ TIAN VALVE THE OPERATION OF PERRY NUCLEAR POWER PLANT, UNITS 1 LEAKAGE AND LEAKAGE TREATMENT METHODS.

AND 2 Docket Nos 50 440 And 50441 (Cleveland Electnc illununat- NUREG 1202: TECHNICAL SPECIFICATIONS FOR HOPE GIEEK W Company) GENERATING STATION. Docket No. 50-354. (Pubic Service Electnc NUREG 1047 S03. SAFETY EVALUATION REPORT RELATED TO & Gas Company Of New Jersey)

THE OPERATION OF NINE MILE POINT NUCLEAR' 4TATION, UNIT NUREG 1203. TECHNICAL SPECIFICATIONS FOR CLINTON POWER NO. 2. Docket No. 50-410 (Niagara Mohawk Power Corporata) ST ATION UNIT 1. Docket No. 50-481 (Illwes Power Company.et af)

NUREG-1047 SO4. SAFETY EVALUATION REPORT RELATED TO OlVISION OF SAFETY REVIEW & OVERSIGHT (POST 851125)

THE OPERATION OF NINE MILE POINT NUCLEAR STATION, UNIT NUREG 1206- ANALYSIS OF FRENCH (PALUEL) PREbSURIZED 2 Docket No. 50-410. (Niagara Mohawk Corporation) WATER REACTOR DESIGN DIFFERENCES COMPARED TO CUR-NUREG-1048 506: SAFETY EVALUATION REPORT RELATED TO RENT U.S. PWR DESIGNS.

THE OPERATION OF HOPE CREEK GENERATING

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NRC Contract Sponsor index (Contractor Reports)

This index lists the'NRC organizations that sponsor organization is followed by the sponsored the contractor reaorts listed in NUREG/CR number and title of the this compilation. It is arrangec 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.

l (e.g., divisions) where appropriate. The I

l EDO. OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTON (PRE DATA 860720)

AEOD. DIRECTOR'S OFFICE NUREG/CR-3262 V04: COBRA-NC:A THERMAL HYDRAULICS CODE NUREG/CR-2000 V05 N6: UCENSEE EVENT REPORT (LER) FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR COMPILATON.For Month Of June 1986. COMPONENTS.Vol 4. Users' Manual For Containtnent Analysis.

NUREG/CR-2000 VOS N7: LICENSEE EVENT REPORT (LER) NUREG/CR-3858: TRAC-PF1/ MOD 1:AN ADVANCED BEST-ESTI.

COMPILATION For Month Of July 1986.

MATE COMPUTER PROGRAM FOR PRESSURIZED WATER REAC-NUREG/CR-2000 VOS N8: UCENSEE EVENT REPORT (LER) TOR THERMAL-HYDRAULIC ANALYSIS.

COMPILATON For Month Of August 1986.

NUREG/CR-4622: VAUDATON OF STOCHASTIC FLOW AND TRANSPORT MODELS FOR UNSATURATED SO.LS.A Comprehen-OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) sive Field Study.

DIVISION OF OA, VENDOR & TECHNICAL TRAINING CENTER PRO-GRAMS (POST B5021 NUREG/CR4629: INDEPENDENT VERIFICATION OF RADIONU.

CUDE RELEASE CALCULATIONS FOR SELECTED ACCICENT NUREG/CR-4678: A METHOD FOR USING PRA TO ESTA8USH SCENARIOS.

QUAUTY PROGRAM APPUCABluTY.

NUREG/CR4644: GEOCHEMICAL STUDIES OF COMMERCIAL LOW-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS LEVEL RADIOACTIVE WASTE DISPOSAL SITES. Topical Report.

OlVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG/CR4688 Vot: OVANTIFICATION AND UNCERTAINTY ANAL.

NUREG/CR4593: INITIAL CONCEPTS ON ENERGETICS & MASS YSIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN LIGHT RELEASES DURING NONNUCLEAR EXPLOSIVE EVENTS IN FUEL WATER REACTORS (OUASAR).Part 1: Methodology And Program CYCLE FACIUTIES.

NOREG/CR4628: RELATIVE AGE-SPECIFIC RADIATION DOSE NUREG/CR4699: APPLICATON OF ALTERNATING CJRRE'4T PO.

1 COMMITMENT FACTORS FOR MAJOR RADONUCUDES RE- TENTIAL DIFFERENCE TO CRACK LENGTH MEASUF:EMENT

" ^ FUEL FACluTIES.

DIVISION OF WASTE AGEu NT NURE R 725 T ASE FLOW MEASUREMENT IN THE NUREG/CR-3925' SWIFT 11 SELF-TEACHING UPPER PLENUM OF A PWR DURING REFLOOD.

CURRICULUM filustrative Problems For Sandia Waste-Isolation Flow NUREG/CR4729: DROPLET ENTRAINMENT IN VERTICAL ANNU-And Transport Model For Fractured Media. LAR FLOW AND ITS CONTRIBUTION TO MOMENTUM TRANS-NUREG/CR.3967: AN ANALYSIS OF EXCURSIONS AT SELECTED IN

• SITU URANIUM MINES IN WYOMING AND TEXAS OfVISION OF ACCIDENT EVALUATON (POST 840101)

NUREG/CR-3974. GEOMEMBRANE SELECTON CRITERIA FOR NUREG/CR-2331 V05 N4. SAFETY RESEARCH PROGRAMS SPON-URANIUM MILL TAILINGS PONDS SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-4236: PROGRESS IN EVALUATON OF RADIONUCUDE RESEARNuade@ Wogress Repod October @cember W5.

GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH LEVEL NUREG/CR-4255 W3 NO AEROSAL EEASE M TRANSW NUCLEAR WASTE REPOSITORY SITE PROJECTS Annual Report PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER For October 1984 September 1985. B ^ 18 EG R V NESA:A MECHANISTIC MODEL OF RADIONU-U.S. NUCLEAR REGULATORY COMMISSION CLIDE RELEASE AND AEROSOL GENERATION DURING CORE NRC - NO DETAILED AFFillATION GIVEN DEBR!S INTERACTIONS WITH CONCRETE.

NUREG/CR-4013: LADTAP 11 TECHNICAL REFERENCE AND USER NUREG/CR4340 V02: REACTOR SAFETY RESEARCH SEMIANNU- ,

GUIDE. AL REPORT. July December 1985.

NUREG/CR4521: TURC2 AND 3 LARGE SCALE UO2/ZRO2 MELT-OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/01) CONCRETE INTERACTON EXPERIMENTS AND ANALYSIS.

OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR (POST NUREG/CR-4558: INTERACTION OF HOT SOUD CORE DEGRIS 860720) WITH CONCRETE.

NUREG/CR-4367 Rot: ORVlRT.PC (VERSION 2.0) A 2 D FINITE. ELE- NUREG/CR4587: SOURCE TERM CODE PACKAGE.A USER'S MENT FRACTURE ANALYSIS PROGRAM FOR A MICROCOMPUT. GUIDE (MOD 1).

ER.

NUREG/CR4649: SCAUNG ANALYSIS OF THE COUPLED HEAT DIVISION OF ENGINdEA4G SAFETY (POST 860720) TRANSFER PROCESS IN THE HIGH TEMPERATURE GAS-NUREG/CR4062 V04: DEGRADED PIPING PROGHAM . PHASE COOLED REACTOR CORE.

II. Semiannual Report, October 1985

• March 1986.

NUREG/CR4656. VERIFICATION TEST CALCULATIONS FOR THE NUREG/CR4480- EROSION PROTECTION OF URANIUM TAluNGS SOURCE TERM CODE PACKAGE.

IMPOUNDMENTS.

NUREG/CR4677: TRAP-MELT 2 CODE DEVELOPMENT AND IM-NUREG/CR 4574: AN EXPERIMENTAL AND ANALYTICAL ASSESS- PROVEMENT OF TRANSPORT MODEUNG.

MENT OF CIRCUMFERENTIAL THROUGH-WALL CRACKED PIPES NUREG/CR-4683. P8F SEVERE FUEL DAMAGE SCOPING TEST-UNDER PURE BENDING TEST RESULTS REPORT.

NUREG/CR4575 PREDICTIONS OF J-R CURVES WITH LARGE DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429)

CRACK GROWTH FROM SMALL SPECIMEN DATA. NUREG/CR4487: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR4738: SIZE INFLUENCE ON THE SEAUNG PERFORM- A DRY CONTAINMENT TEST PROBLEM FOR THE MAEROS AER.

ANCE OF CEMENTITOUS BOREHOLE PLUGS. OSAL MODEL.

DIVISION OF REACTOR SYSTEM SAFETY (POST 860720) NUREG/CR4532 V01: MODELS OF COGNITIVE BE'HAVOR IN NU.

NUREG/CR4439 LEPRICON ANALYSIS OF PRESSURE VESSEL CLEAR POWER PLANT PERSONNELA Feasbbty Study. Volume SURVEILLANCE POSIMETRY INSERTED INTC H B. ROBINSON-2 1. Summary Of Results.

DURING CYCLE 9 NUREG/CR4532 V02: MODELS OF COGNITIVE BEHAVIOR IN NU-NUREG/CR4495: VISCOSITY OF ZIRCONIUM URANIUM OXIDE (Zr- CLEAR POWER PLANT PERSONNEL:A Feasblity Study volume UO2) MIXTURES AT 1800 TO 2100 C. 2. Main Report 53

54 NRC Contract Sponsor index NUREG,OR-4560: RISK ASSESSMENT APPUCATION TO NRC IN. NUREG/CR4336: REVIEW OF llT RESEARCH IfsSitTUTE REPORTS SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO ON STRUCTURAL STUDIES OF REACTOR CONTAINMENT.

JANUARY 1986. NUREG/CR4445. EFFECTIVENESS AND SAFETY ASPECTS OF SE.

NUREG/CR4Ses: COMPBRN lli- A COMPUTER CODE FOR MODEL. LECTED DECONTAMINATION PROCESSES.

ING COMPARTMENT FIRES. NUREG/CR-4529: PIPING DAMPlNG - EXPERIMENTAL RESULTS NUREG/CR-4598. A USER'S GUIDE FOR THE TOP EVENT MATRIX FROM LABORATORY TESTS IN THE SEISMIC RANGE.

ANALYSIS CODE (TEMAC) NUREG/CR4530 V01: U S1 FRENCH JOINT RESEARCH PROGRAM NUREG/CR4624 VO1: RADIONUCLIDE RELEASE CALCULATIONS REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume SUBJECTED TO BETA RADIATION.Vol 1: Phase-1 Normalization 1 BWR. Mark i Design. Results.

NUREG/CR4624 V02: RADIONUCUDE RELEASE CALCULATIONS NUREG/CR4533. PROGRAM TO ANALYZE THE FAILURE MODE OF FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume LEAD-ACID BATTERIES.

2 PWR,1ce Condenser Design. NUREG/CR4536: SUPERHEATED-STEAM TEST OF ETHYLENE NUREG/CR4624 V03. RADIONUCLIDE RELEASE CALCULATIONS PROPYLENE RUBBER CABLES US!NG A SIMULTANEOUS AGING FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume AND ACCIDENT ENVIRONMENT.

3 PWR.Subatmosphenc Containment Design, NUREG/CR-4540: AN EVALUATION OF J-R CURVE TESTING OF NUREG/CR4624 V04. RADIONUCLIDE RELEASE CALCULATIONS NUCLEAR PIPING MATERIALS USING THE DIRECT CURRENT PO-FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume TENTIAL DROP TECHNIQUE.

4 BWR. Mark til Design. NUREG/CR-4562. PlPE DAMPING-RESULTS OF VIBRATION TESTS NUREG/CR4624 V05: RADIONUCUDE RELEASE CALCULATIONS IN THE 33 TO 100 HERTZ FREQUENCY RANGE.

FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume 5 Large NUREG/CR-4570: DESCRIPTION AND TESTING OF AN APPARATUS Dry Containment Design. FOR ELECTRICALLY INITIATED FIRES THROUGH SIMULATION DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST OF A FAULTY CONNECTION.

840429) NUREG/CR-4596: SCREENING TESTS OF REPRESENTATIVE NU-NUREG/6R-3145 V04: GEOPHYSICAL INVESTIGATIONS OF WEST- CLEAR POWER PLANT COMPONENTS EXPOSED TO SECOND-ERN OHIO-INDIANA REGION. Annual Report (October 1984 Sep- ARY ENVIRONMCNTS CREATED BY FIRES. '

tember 1985). NUREG/CR4597 VOI: AGING AND SERVICE WEAR OF AUXIUARY NUREG/CR4162: SURVEY OF STATISTICAL AND SAMPtaNG FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol.

NEEDS FOR ENVIRONMENTAL MONITORING OF COMMERCIAL 1. Operating Expenance And Failure identification.

LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITIES. NUREG/CR4633: ORPLOT.PC: A GRAPHIC UTtuTY FOR NUREG/CR4553- FIELD TESTS OF A PORTABLE TISSUE EQU!VA- ORMGEN PC AND ORVIRT.PC.

LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA NUREG/CR4647: CALCULATION OF THE POWER DISTRIBUTION IN RADIATION FIELDS THE VENUS PWR MOCK-UP BENCHMARK USING TWO. GROUP NUREG/CR4660. ORGANIC COMPLEXANT-ENHANCED MOBlUTY DIFFUSION THEORY.

OF Toxic ELEMENTS IN LOW. LEVEL WASTES Final Report. NUREG/CR4648: A STUDY OF TYPICAL NUCLEAR CONTAINMENT NUREG/CR4676. SEISMIC REFLECTION GEOMETRY OF THE PURGE VALVES IN AN ACCIDENT ENVIRONMENT.

NEWARK BASIN IN EASTERN PENNSYLVANIA Evidence For Ex- NUREG/CR4659 V01: SEISMIC FRAGluTY OF NUCLEAR POWER tensional Reactivation Of Paleozoic Thrust Faults. PLANT COMPONENTS (PHASE 1).

NURE G/CR-4698: AlREORNE UDAR MAPPING OF SF6 CONCEN- NUREG/CR4667 VOI: ENVIRONMENTALLY ASSISTED CRACKING TRATION DISTRIBUTONS FOR TRANSPORT AND DIFFUSION IN UGHT WATER REACTORS. Semiannual Report.Apni-September STUDIES. ggg5 NUREG/CR4702: POST-EOCENE FAULT NEAR EAST EDGE OF NUREG/CR-4873 HEAVY-SECTION STEEL TECHNOLOGY PRO-REELFOOT RlFT IN LAUDERDALE COUNTY. TENNESSEE.AS DIS- GRAM FIVE YEAR PLAN FY1985-1989 COVERED BY GRAVITY, EARTH RESISTIVITY SURVEYS AND DRILUNG. EDO-RESOURCE MANAGEMENT NUREG/CR4705- IDENTIFICATION OF NORTHWEST TRENDING DIVISION OF BUDGET & ANALYSIS SEISMOGENIC GRABEN NEAR CHARLESTON. SOUTH CAROUNA NUREG/CR-2907 V04: RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR4706: A STUDY OF SEISMICITY AND EARTHOUAKE NUCLEAR POWER PLANTS. Annual Report 1963.

HAZARD IN NORTHERN ALABAMA AND ADJACENT PA9TS OF TENNESSEE AND GEORGIA. OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

NUREG/CR-4707. A PREUM1 NARY GEOLOGIC EVALUATION OF DIVISION OF HUMAN FACTORS TECHNOLOGY (POST 851125)

THE ALABAMA-TENNESSEE TRANSVERSE SFRMiG ZONE IN NUREG/CR 4125 V01 R1: GUIDELINES AND WORKDOOK FOR AS-ALABAMA. SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTlu-DIVISION OF ENGINEERING TECHNOLOGY (PRE 860720) TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR-3320 vot: LWH PRESSURE VESSEL SURVEILLANCE PLANT. Volume 1Gudelines For Utility Organization And Administra-DOSIMETRY IMPROVEMENT PROGRAM PSF Expenments Summa. tion Plan.

ry And Blind Test Results NUHEG/CH4125 V02 H1: GUIDELINES AND WOHn3OOK FOH AS-NUREG/CR-3805 V04. ENGINEERING CHARACTER!ZATION OF SESSMENT OF ORGANIZATION AND ADMINISTRATION OF UTILI-GROUND MOTION. Task II. Soil Structure Interaction Effects On TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER Structural Response PLANT. Volume 2 Workbook For Assessment Of Organuaton And NUREG/CR 3805 V05: ENGINEERING CHARACTERIZATION OF Management. #

GROUND MOTION Task 11. Summary Report. NUREG/CR4600: HUMAN FACTORS STUDY CONDUCTED IN CON-NUREG/CR4099- AGE-RELATED DEGRADATION OF NATURALLY- JUNCTION WITH A MINI ROUND ROBIN ASSESSMENT OF UL-AGED CLASS 1E BATTERY CELLS TRASONIC TECHNICIAN PERFORMANCE NUREG/CR4102. AIR CURRENTS DRIVEN BY SPRAYS 'N 1EAC- DIVISION OF SAFETY REVIEW & OVERSIGHT (POST 851125)

TOR CONTAINMENT BUILDINGS NUREG/CR-2800 SO4: GUIDELINES FOR NUCLEAR POWER PLANT NUREG/CR4219 V03 N1: HEAVY-SECTION STEEL TECHNOLOGY SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOFMENT.

I PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1985 MARCH 1986.

NUREG/CR 4257 V02. INSPECTION. SURVEILLANCE.AND MONITOR.

INd OF ELECTRICAL EQUIPMENT IN NUCLEAR POWER NUREG/CR4470: SURVEY AND EVALUATON OF VITAL INSTRU-MENTATION AND CONTROL POWER SUPPLY EVENTS.

NUREG/CR4626 V01: IMPROVING THE RELIABluTY OF OrEN-CYCLE WATER SYSTEMS An Evaluation Of Biofouling Surveillance PLANTS Volume 2 Pressure Transmitters. And Control Techniques For Use At Nuciaar Power Plants NUREG/CR-4265 V02: AN ASSESsuENT OF THE SAFETY IMPUCA- NUREG/CR4643- EVALUATION OF CORE DAMAGE SEQUENCES TIONS OF CONTROL AT THE CALVERT CUFFS-1 NUCLEAR INITIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOL-PLANT. Volume 2 Appendices ING.

NUREG/CR 4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION. NUREG/CR-4694. APPLICATION 08 THERM 1X-KONVEK CODE TO SHIP FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE ACCIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEM-VE SSE LS Progress Report, October 1985 March 1986. PERATURE REACTORS (HTRS)

Contractor index This index lists, in alphabetical order, the numbers ano titles of their reports. If further l

contractors that prepared the NUREG/CR information 's 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 -

ALABAMA, STATE OF NUREG/CR-4656: VERIFICATION TEST CALCULATIONS FOR THE NUREG/CR-4706: A STUDY OF SEISM! CITY AND EARTHOUAKE SOURCE TERM CODE PACAAGE.

HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF NUREG/CR-4677: TRAP-MELT 2 CODE: DEVELOPMENT AND IMPROVE-TENNESSEE AND GEORGIA. MENT OF TRANSPORT MODEUNG.

NUREG/CR-4707: A PRELIMINARY GEOLOGIC EVALUATION OF THE ALABAMA. TENNESSEE TRANSVERSE SEISMIC ZONE IN ALABAMA. BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST AMERICAN NUCLEAR SOCIETY NUREG/CR-2000 SO4: GUIDEUNES FOR NUCLEAR POWER PLANT NUREG/CP-0080 V01: PROCEEDINGS OF THE TOPICAL MEETING ON SAFETY ISSUE PRORITl2ATION INFORMATON DEVELOPMENT.

REACTOR PHYSICS AND SAFETY. Sessions 1-10.

NUREG/CP-0080 V02- PROCLEDINGS OF THE TOPICAL MEETING ON NUREG/CR-3262 V04: COBRA-NC.A THERMAL HYDRAUUCS CODE REACTOR PHYSICS AND SAFETY. Sessions 11-16. FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR COMPONENTS.Vol. 4. Users' Manual For Contamment Analysis.

ARGONNE NATIONAL LABORATORY NUREG/CR-3974: GEOMEMBRANE SELECTION CRITERIA FOR URA.

NUREG/CR4667 Vot: ENVIRCNMENTALLY ASSISTED CRACKING IN NIUM MILL TAluNGS PONDS.

UGHT WATER REACTORS. Serniannual Report,Apnl-September 1985- NUREG/CR-4013: LADTAP 11 TECHNICAL REFERENCE AND USER ARIZONA, UNIV.OF TUCSON, AZ OE.

NUREG/CR-4162: SURVEY OF STATISTICAL AND SAMPLING NEEDS NUREG/CR-4738: Sl?E INFLUENCE ON THE SEAUNG PERFORM-ANCE OF CEMENTITIOUS 803EHOLE PLUGS.

FOR ENVIRONMENTAL MONITORING OF COMMERCIAL LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITIES. ,

ARVIN/CALSPAN ADVANCED TECHNOLOGY CENTER NUREG/CR4300 V03 N1: ACOUSTIC EMISSION / FLAW RELATION.

NUREG/CR4257 V02: INSPECTON. SURVEILLANCE.AND MONITOR. SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE ING OF ELECTRICAL EQUIPMENT IN NUCLEAR POWER VESSELS. Progress Report, October 1985. March 1986.

PLANTS. Volume 2.Prets.re Transmitters. NUREG/CR4480: EROSION PROTECTION OF URANIUM TAluNGS IM-POUNDMENTS.

ASPEN SYSTEMS, INC.

4 NUREG-0386 D04 R02; UNTED STATES NUCLEAR REGULATORY NUREG/CR4495: VISCOSITY OF ZlRCONIUM-UBANIUM OXIDE (Zr-COMMISSION STAFF PR ACTICE AND PROCEDURE DIGEST ;ULY UO2) MIXTURES AT 1800 TO 2100 C.

NUREG/CR4593; INITIAL CONCEPTS ON ENERGETICS & MASS RE-1,1972 DECEMBER 31,1986.

LEASES DURING NONNUCLEAR EXPLOSIVE EVENTS IN FUEL BATTELLE HUMAN AFFAIRS RESEARCH CENTERS CYCLE FACIUTIES.

NUREG/CR-4125 V01 R1: GUIDEUNES AND WORKBOOK FOR AS. NUREG/CR-4600: HUMAN FACTORS STUDY CONDUCTED IN CON.

SESSMENT OF ORGANIZATON AND ADMINISTRATON OF UTiu. JUNCTION WITH A MINI.ROUND ROBIN ASSESSMENT OF ULTRA-TIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER SONIC TECHNICIAN PERFORMANCE.

PLANT. Volume 1: Guidelines For Utility Organization And Admitustration NUREG/CR4622: VAUDATON OF STOCHASTIC FLOW AND TRANS-Plan. PORT MODELS FOR UNSATURATED SOILS.A Comprehensive Field j NUREG/CR-4125 V02 R1; GUIDEUNES AND WORKBOOK FOR AS- Study.

SESSMENT OF ORGANIZATON AND ADMINISTRATON OF UTiu. NUREG/CR-4626 V01: IMPROVING THE REUABluTY OF OPEN-TIES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER CYCLE WATER SYSTEMS.An Evaluation Of Biofouhng Surveillance PLANT. Volume 2. Workbook For Assessment Of Organization And Man- And Control Techniques For Use At Nuclear Power Plants. ,

agement. NUREG/CR4660 ORGANIC COMPLEXANT. ENHANCED MOBluTY OF 4

NUREG/CR-4600. HUMAN FACTORS STUDY CONDUCTED IN CON- TOXIC ELEMENTS IN LOW-LEVEL WASTES. Final Report.

JUNCTION WITH A MINI ROUND RODIN ASSESSMENT OF ULTRA. ,

SONIC TECHNICIAN PERFORMANCE. BROOKHAVEN NATIONAL LABORATORY BATTELLE MEMORIAL INSTITUTE, COLUMSUS LABORATORIES

" ' "* ^ "

NUREG/CR-4062 V04: DEGRADED PIPING PROGRAM - PHASE SORED BY ME & N W AR REWW l

RESEARCH.Ouarterty Progress Report October-December 1985.

! 11 Senwannual AN NUREG/CR4574: Report, October 1985.

EXPERIMENTAL ANDMarch AN 1986'ALYTICAL ASSESS. NUREG/CR-2907 V04: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1983.

MENT OF CIRCUMFERENTIAL THROUGH-WALL CRACKED PIPES UNDER PURE BENDING. NUREG/CR4553. FIELD TESTS OF A PORTABLE TISSUE EQUIVA-NUREG/CR4575: PREDICTONS OF J-R CURVES WITH LARGE LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA I

CRACK GROWTH FROM SMALL SPECIMEN DATA. RADIATON FIELDS.

NUREG/CR4587: SOURCE TERM CODE PACKAGE.A USER'S GUOE NUREG/CR4629; INDEPENDENT VERIFICATION OF RADIONUCLOE (MOD 1) RELEASE CALCULATONS FOR SELECTED ACCIDENT SCENAROS.

NUREG/CR4624 V01: RADONUCUDE RELEASE CALCULATIONS NUREG/CR4643: EVALUATION OF CORE DAMAGE SEQUENCES INI-FOR SELECTED SEVEtu ACCIDENT SCENARIOS Volume TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOUNG.

1 BWR, Mark i Dessn. NUREG/CR.4644. GEOCHEMICAL STUDIES OF COMMERCIAL LOW.

NUREG/CR-4624 V02: RADONUCLOE RELEASE CALCULATIONS LEVEL RADIOACTIVE WASTE DISPOSAL SITES. Topical Report FOR SELECTED SEVERE ACCIDENT SCENARIOS. Volume 2 PWR.lce NUREG/CR4659 V01: SEISMIC FRAGluTY OF NUCLEAR POWER NU $R4 V03. RADIONUCUDE RELEASE CALCULATIONS PLANT COMPONENTS (PHASE 1).

NUREG/CR4688 V01: OUANTIFICATON AND UNCERTAINTY ANALY.

FOR SEL *.CTED SEVERE ACCIDENT SCENARIOS. Volume 3 PWRSubatmosphenc Contamment Design SIS OF SOURCE TERMS FOR SEVERE ACCIDENTS IN UGHT NUREG/CR4624 V04 RADIONUCUDE RELEASE CALCULATIONS WATER REACTORS (OVASAR)Part 1: Methodology And Program FOR SELECTED SEVERE ACCIDENT SCENARIOS Volume Plan.

4 BWR.Ma'k til Desegn.

NUREG/CR4624 V05- RADONUCUDE RELEASE CALCULATIONS CHIAPETTA, WELCH & ASSOCIATES, LTD.

4 FOR SELECTED SEVERE ACCIDENT SCENARIOSVolume $ Large NUREG/CR4336: REVIEW OF 11T RLSEARCH INSTITUTE REPORTS i Dry Containment Design. ON STRUCTURAL STUDIES OF REACTOR CONTAINMENT.

, 55

56 Contractor index DAVID W. TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER NUCLEAR WASTE REPOSITORY SITE PROJECTS.Armual Report For NUREG/CR-4540: AN EVALUATON OF J-R CURVE TESTING OF NU- October 1984 Septernber 1985.

CLEAR PIPING MATERIALS USING THE DIRECT CURRENT POTEN. NUREG/CR4255 '/03 N1: AEROSAL RELEASE AND TRANSPORT TIAL DROP TECHNIOUE. PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1985 - MARCH 1986.

EG&O IDAHO, INC. (SUBS. OF EG4G, INC.) NUREG/CR4257 V02: INSPECTON. SURVEILLANCE.AND MONITOR-

, NUREG/CH4553: FIELD TESTS OF A PORTABLE TISSUE EQUIVA- ING OF ELECTRICAL EQUIPMENT IN NUCLEAR POWER LENT SURVEY METER FOR MONITORING MIXED BETA / GAMMA PLANTS. Volume 2. Pressure Transtmtters.

RADIATION FIELDS. NUREG/CR4265 V02: AN ASSESSMENT OF THE SAFETY IMPUCA-NUREG/CR4648; A STUDY OF TYPICAL NUCLEAR CONTAINMENT TiONS OF CONTROL AT THE CALVERT . CUFFS-1 NUCLEAR PURGE VALVES IN AN ACCIDENT ENVIRONMENT. PLANT. Volume 2:Apperdces.

NUREG/CR-4683; PBF SEVERE FUEL DAMAGE SCOPING TEST TEST NUREG/CR-4367 Rot: ORVIRT.PC (VERSION 2.0):A 2-0 FINITE-ELE-RESULTS REPORT- MENT FRACTURE ANALYSIS PROGRAM FOR A MICROCOMPUTER.

NUREG/CR4439 LEPRICON ANALYSIS OF PRESSURE VESSEL SUR-ENERGY RESEARCH & CONSULTANTS CORP. VEILLANCE DOSIMETRY INSERTED INTO H.B. ROBINSON-2 NUREG/CR4597 V01: AGING AND SERVICE WEAR OF AUXILIARY DURING CYCLE 9 '

FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol. NUREG/CR-4470: SURVEY AND EVALUATON OF VITAL INSTRUMEN-l 1.Operahng Exponence And Failure Identificahon. TATION AND CONTROL POWER SUPPLY EVENTS.

GEORGIA INSTITUTE OF TECHNOLOGY, ATLANTA, GA NUREG/CR-456J: RISK ASSESSMENT APPUCATION TO NRC IN-SPECTION PROGRESS REPORT FOR PERIOD JANUARY 1985 TO +

NUREG/CR-4706: A STUDY OF SEISMICITY AND EARTHOUAKE

HAZARD IN NORTHERN ALABAMA AND ADJACENT PARTS OF NUR CR4566 COMPBAN lil A COMPUTER CODE FOR MODEL-TENNESSEE AND GEORGIA.

ING COMPARTMENT FIRES.

HANFORD ENGINEERING DEVELOPMENT LA90HATORY NUREG/CR4597 V01: AGING AND SERVICE WEAR OF AUXILIARY j NUREG/CR3320 V01: LWR PRESSURE VESSEL SURVEILLANCE DO. FEEDWATER PUMPS FOR PWR NUCLEAR POWER PLANTS.Vol. .

Si TRYIM VEMENT PROGRAM PSF Expenments Summary Ard NU 628 LAT V A P 1 IC ADIATON DOSE COM-MITMENT FACTORS FOR MAJOR RADIONUCLIDES RELEASED HOUSTON, UNIV. OF, HOUSTON, TX FROM NUCLEAR FUEL FACluTIES.

NUREG/CR4729: DROPLET ENTRAINMENT IN VERTICAL ANNULAR NUREG/CR-4633: ORPLOT.PC: A GRAPHIC UTIUTY FOR ORMGEN PC FLOW AND ITS CONTRIBUTION TO MOMENTUM TRANSFER, AND ORVIRT.PC.

NUREG/CR4647: CALCULATION OF THE POWER DISTRIBUTION IN

%TAHO NATIONAL ENGINEERING LABORATORY THE VENUS PWR MOCK UP BENCHMARK USING TWO GROUP DIF.

NUREG/CR-4445; EFFECTIVENESS AND SAFETY ASPECTS OF SE. FUSION THEORY.

LECTED DECONTAMINATON PROCESSES. NUREG/CR-4649: SCAUNG ANALYSIS OF THE COUPLED HEAT NUREG/CR4529: PIPING DAMPlNG EXPERIMENTAL RESULTS TRANSFER PROCESS IN THE HIGH-TEMPERATURE GASCOOLED FROM LABORATORY TESTS IN THE SEISMIC HANGE. REACTOR CORE.

s NUREG/CR4562- PIPE DAMPING-RESULTS OF VlBRATON TESTS IN NUREG/CR 4673: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM THE 33 TO 100 HERTZ FREQUENCY RANGE. . FIVE YEAR PLAN FY1985-1989.

NUREG/CR4694: APPUCATON OF THERMIX.KONVEK CODE TO AC-1*AHO, UNIV. OF, MOSCOW,10 CIDENT ANALYSES OF MODULAR PEBBLE BED HIGH TEMPERA.

NUREG/CR-3967. AN ANALYSIS OF EXCURSIOh3 AT SELECTED IN TURE REACTORS (HTRS).

SITU URANIUM MINES IN WYOMING AND TFXAS.

SANDIA NATIONAL LASORATORIIS WPER CN NUREG/C43925: SWIFT ll SELF TEACHING CURRICULUM.lliustrative NUREG/CR 4641 EVALUATION OF CORE DAMAGE SEQUENCES INI- Problems For Sarda Waste-Isolaton Flow And Transport Model For TIATED BY LOSS OF REACTOR COOLANT PUMP SEAL COOLING. Fractured Media.

INTEROM, DEPT. OF, GEOLOGICAL SURVEY NUREG/CR 4099: AGE-RELATED DEGRADATON OF NATURALLY.

NUREG/CR4676

SEISMIC REFLECTON GEOMETRY OF THE NUREGI R-4 02 A R CURRE S DRIVEN BY SPRAYS IN REACTOR i NEWARK BASIN IN EASTERN PENNSYLVANIA. Evidence For Exton' j sonal Reactivaton Of Paleozoc Thrust Faults. CONTAINMENT NUREG/CR4308: BUILDINGS' MECHANISTIC MODEL VANESA:A OF RAD I

LOS ALAMOS SCIENTIFIC LADORATORY CLOE RELEASE AND AEROSOL GENERATON DURING CORE NUREG/CR-3858. TRAC-PF1/ MOD 1 AN ADVANCED BEST-ESTIMATE NUREG 43 IEW F 11 SEA CH INSTITUTE REPORTS

^ ""^ N ON STRUCTURAL STUDIES OF REACTOR CONTAINMENT.

ER HY LIC ANAL SIS- NUREG/CR4340 V02: REACTOR SAFETY HESEARCH SEMIANNUAL j MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE MA REPORT. July . December 1985.

NUREG/CR4487; UNCERTAINTY AND SENSITIVITY ANALYSIS OF A I NUREG/C44725. TWO-PHASE FLOW ME ASUREMENT IN THE UPPER i

PLENUM OF A PWR DURING REFLOOD. DRY CONT AINMENT TEST PROBLEM FOR THE MAEROS AEROSAL l MODEL

, MICHIGAN, UNIV. OF, ANN ARSOR, MI NUREG/CR4521: TURC2 AND 3 LARGE SCALE UO2/ZRO2 MELT +

l NUREG/CR3145 V04. GEOPHYSICAL INVESilGATIONS OF WEST. CONCRETE INTERACTON EXPERIMENTS AND ANALYSIS.

I ERN OHIO INDIANA REGION. Annual Report (October 1984 Septem- NUREG/CR4530 V01: U.S1 FRENCH JOINT RESEARCH PROGRAM ber 1985) REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-l JECTED TO BETA RADIATION.Vol 1: Phase 1 Normansation Results.

CAK MlOGE NATIONAL LA80RATORY NUREG/CR-4533; PROGRAM TO ANALYZE THE FAILURE MODE OF NUREG/CP4083. ANS TOPICAL MEETING ON RADIOLOGICAL ACCI. LEAD-ACID BATTERIES.

• DENTS PERSPECTIVES AND EMERGENCY PLANNING PROGRAM NUREG/CR4536: SUPERHEATED STEAM TEST OF ETHYLENE PRO-AND ABSTRACTS PYLENE RUBBER CABLES USING A SIMULTANEOUS AGING AND NUREG/CR2000 V05 No. UCENSEE EVENT REPORT (LER) ACCIDENT ENVIRONMENT.

COMPtLATON For Month Of June 1986. NUREG/CR4558: INTERACTON OF HOT SOUD CORE DEBRi$ WITH NUREG/CR2000 VOS N1: LICENSEE EVENT REPORT (LER) CONCRETE.

COMPILATON For Month Of July 1986. NUREG/CR4570. DESCRIPTON AND TESTING OF AN APPARATUS NUREG/CR2000 V05 N8. LICENSEE EVENT REPOdi (LER) FOR ELECTRICALLY INITIATED FIRES THROUGH SIMULATION OF COMPfLATION For Month Of August 1986 A FAULTY CONNECTON.

NUREG/C43967: AN ANALYSIS OF EXCURSIONS AT SELECTED IN NUREG/CR4596: SCREENING TESTS OF REPRESENTATIVE NUCLE.

SITU URANIUM MINES IN WYOMING AND TEXAS AA POWER PLANT COMPONENTS EXPOSED TO SECONDARY EN-

] NUREG/CR4219 V03 N1: HEAVY SECTION STEEL TECHNOLOGY VIRONMENTS CREATED BY FIRES.

7 PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER NUREG/CR4598: A USER'S GUIDE FOR THE TOP EVENT MATRIX 1985 MARCH 198a ANALYSIS CODE (TEMAC)

NUREG/C44236: PROGRESS IN EVALUATION OF RADIONUCUDE NUREG/C44678: A METHOD FOR USING PRA TO ESTABUSH OVAL.

GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH LEVEL ITY PROGRAM APPLICABluTY.

i I

Contractor index 57 SOUTH CAROUNA, UNIV. OF, COLuteSIA. SC COVERED BY GRAVITY, EARTH RESISTIVITY SURVEYS AND NUREG/CR-4705: 'DENTIFICATION OF NORTHWEST TRENDING DRILLING.

SEISMO !ENIC GRA3EN NEAR CHARLESTON. SOUTH CAROLINA.

VIRGINIA POLYTECHNIC 190STITUTE & STATE UNIV, SLACKSSURG, SRIINTERNATIONAL VA NUREG/CR4698: AIRBORNE LIDAR MAPPING OF FF6 CONCENTRA. NUREG/CR4678: SEISMIC REFLECTION GEOMETRY OF THE TION DISTalBUTIONS FOR TRANSPORT AND DIFFUSION STUDIES. NEWARK BASIN IN EASTERN PENNSYLVANIA. Evidence For Exter>

sional Reactrvation Of Paleozoic Thrust Faults.

STRUCTURAL & EARTHQUAKE ENGINEERING CONSJLTANTS I

NUREG/CR-3805 V04: ENGINEERING CHARACTER 12ATION OF WESTINGHOUSE ELECTRIC CORP.

! GROUND MOTION. Task IL Sod-Structure interaction Effects On Strue. NUREG/CR4532 vot: MODELS OF COGNITIVE BEHAVIOR IN NUCLE-l tural Response. AR POWER PLANT PERSONNEL.A Feasedsty Study. Volume 1: Sum-mary Of Results.

! STRUCTURAL IAECHANICS ASSOCIATES NUREG/CR4532 V02 MODELS OF COGNITIVE BEHAVIOR IN NUCLE.

' NUREG/CR-3805 V05: ENGINEERING CHARACTER 12ATION OF AR POWER PLANT PERSONNEL:A Foestdity Study. Volume 2. Man GROUND MOTION. Task 11: Summary Report. Report.

U.S. NAVAL ACADEMY ANNAPOUS, MD NUREG/CR4533: PROGRAM TO ANALYZE THE FAILURE MODE OF LEAD-ACID BATTERIES.

NUREG/CR4699: APPLICATION OF ALTERNATING CURRENT PO.

TENTIAL DIFFERENCE TO CRACK LENGTH MEASUREMENT WOOOWAfNKLYDE CONSULTANTS, INC.

DURING RAPID LOADING. NUREG/CR-3805 V04: ENGINEERING CHARACTERIZATION OF VANDERelLT UNIV., NASHVILLE. TN GROUND MOTION. Task 11: Soil-Structure Interaction Effects On Strue.

tural Reeponse.

NUREG/CR4702 POST-EOCENE FAULT NEAR EAST EDGE OF NUREG/CR-3805 V05: ENGINEERING CHARACTERl2ATION OF REELFOOT RIFT IN LAUDERDALE COUNTY, TENNESSEE.AS DIS. GROUND MOTION. Task II: Summary Report.

I 2

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Licensed Facility Index This index lists the facilities that were the Docket number and followed' by the report 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 l

ical order. They are preceded by their number.

l l

S 424 AMn W. Vogne Nudeer Piert, Und 1, George NURE41137 S03 54409 La Crosse Bahn0 Waler Reactor, Darytend NURE40827 S01 Power Ca Power Cooperenve 50425 kvm W. Vogue Nudeer Plert Und 2, Georps NURE41137 S03 S 245 Mestone Nudeer Power $1ston, Urut 1, NURE41143 $01 Power Ca Normeest Nudeer Energy Co S 412 Beaver Vaney Power Staton, Unit 2, Duquesne NURE41057 $02 54410 Nme hele Post Nudeer Staton, Una 2, Negers NURE41047 S03 Co. Mohost Power Corp.

ST4W56 Siston, Une 1. Commonween Edson NURE41002 S01 50410 Nme Wee Pont Nudeer Staton, Una 2, Ptegers NURE41047 SO4 Ca Moheek Power Corp.

STN-W57 Bradwood Staton, Und 2 Commonweem Edson NURE41002 S01 S 440 Pony Nudeer Power Pint, Umt 1, Cleveland NUREG4087 Sto Ca Elecmc leummeeng C 54317 Calvert CWfs Nudest Power Plant, Und 1, NUREG/CR4265 V02 S 441 Pony Nudeer Power Plant, Une 2. Cleveiend NUREG4087 S10 Baltmore Gas & Elecmc Elecmc ulununeeng C S 461 Chnton Power StattA Und I, Ibnces Power C4 NUREG4653 $06 50443 Seetroot Nudeer Staton, Und 1, Putec Service % REG 4006 SOS S 461 Chnson Power Staton, Und 1, Ihnom Power Ca NUREG4653 $07 Co. of New S 461 Orson Power Staton, Und I, anos Power Ca NURE41203 54444 Sestrook Nudeer Umt 2, P2ic Service NUREG4006 S05 S 261 H.S. Rotunnon Plant, Und 2, Caronne Power & NUREG/CR4439 Ca of New Hampshr Ught Co ST4S496 South Texas Prgect, Unit 1, Houston Ughtng & NUREG4781 Sol B261 H.B. Rotunnon Plant, Una 2, Carahna Power & NUREG/CR4439 ERR Power Ca bght ca STN-54498 South Texas Prqsct, Urut 1. Houston Ughtng & NURE41171 S 354 Hope Crest Nudeer Staton, unt 1, Putic NURE41048 $06 Power Ca Senate Elodnc & Gas Co ST454499 South Texas Prgect, Umt 2 Houston Ughhng 4 NUREG4781 S01 54354 Hope Creek Nudeer Stat A Und 1, Putte NURE41202 Power Ca Service Dednc & Gas Co STN 50499 Soum Texas Prqect, Urut 2. Houston Lighwig 4 NURE41171 404027 Kerr McGee Nudest Corp, Oklahoma Cay. 0K, NURE41196 Sol Power Ca 59 l l

l m_

esRC Pomes 335 W 5. hucLE Am REGULaf 0RY COMeestslose e atronY Nuwstm #Ampose r#0C. ses vet to, er ears 12 Set E'," '52 ' Bl8LIOGRAPHIC DATA SHEET NUREG-0304 Vol .11, No. 3 ni .NirauC, ks oN r . .ive. . f 2 Vif LE AND SV8 TLE J LE Avt 9 LANE Regulator and Technical Reports (Abstract Index Journal)

Compilation or Third Quarter 1986 f . oA,...,0.,Co., Liv o July-Septembe =m- 'iAa l

& QUTHORISI j

f s oAre nseonf essuto f woNin vtaa

/ November 1986 7 PERFORM 4NG ORGAN 12 Af TON NaWE D WestsNG ADDat S5 f raicNrdele Ceest 8 Pa0JECTiT A5E. WORN uNet NuMGtR Division of Technical nformation and Document Control Office of Administrati " " " " ' " " " "

U.S. Nuclear Regulatory ommission Washington, DC 20555 10 SPON50 RING omGANal Af EON asAW( ANo MAsLING D#ti$ flarwer te Cosei ito f vet OF AEPORT Quarterly Same as 7, above. , ,, ,,c o c o, f ,, ,, ,,,,,,,,e ,,,,

July - Sentember 1986 12 SUPPLtwtNiamv NOf t3 93 css 5f ma(1 (200 norde er essaf This journal includes all formal reports the NUREG series. prepared by the NRC staff

'onferences and workshops. The entries in and contractors, the compilation are as well as indexed proceedings for access ty f [ti le and abstract, contractor report number,

personal author, subject, NRC organiza ion, c tractor, and licensed facility, f

V le DOCOM4 NT agatvStl e Elv*CAOS-DESCnsf f 1 it a v ai6 Asit T v STA?EWENT abstract index Unlimited

't 80Cymit? CL A55tFIC tTtON t rk4 popeo

. ,oi %,,..... o,i N .No.o riaws Unelassified_

a r. -8,,,

Unclassified

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-% -y w = - - -

eount ctass man UNITED STATES 2

• • • 'uSIN" "~ C NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 ,yy,H C, O

OF 6 et.m 6USINESS 6

l PENAttV FOR PRIVATE USE. $300 120555078877 1 1Ah1AC19LicG -

US ARC 9.

AOM-DIV 0F PUB SVCS E Main Citations snd Abstracts POLICY & FUB PGT BR-FCR NUREG .

a W-501 WASHINGTON CC 20555 'z 9

w E Contractor Number index Report x

Personal Author index s-4 0

x

$zC E. Subject Index 5 2 m Z en

$k E NRC Originating "g

Organization Index gg

• x 5

8 5

NRC Contractor -

EI Sponsorindex k 8

z Contractor Index ii<

w Licensed Facility b I0 Index $

e S

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