ML20203N035

Regulatory and Technical Reports.Compilation for First Quarter 1986,January-March

ML20203N035
Person / Time
Issue date:	04/30/1986
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V11-N01, NUREG-304, NUREG-304-V11-N1, NUDOCS 8605020454
Download: ML20203N035 (73)

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NUREG-0304 Vol.11, No.1 1

Regulatory and Technical Reports (Abstract Index Journal?

Compilation for First Quarter 1986 January - March U.S. Nuclear Regulatory Commission Office of Administration

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NUREG-0304 Vol.11, No.1 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for First Quarter 1986 January - March Data Published: April 1980 Policy and Publications Management Branch Divi 2 ion of Technical Information and Document Control Office of Administration U.S. Nuclear Regulatory Commission W:thington, D.C. 20665 na mee, 1

r-CONTENTS Preface

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

Conference Proceedings....

Contractor R eports.............................................................

Contractor Report Number Index..................................................... 2 Personal Au thor Index............................................................... 3 S u bject index........................................................................ 4 NRC Originating Organization Index (Staff Reports)........................................ 5 NRC Contract Sponsor index (Contractor Reports)......................................... 6 Contractor i ndex...................................................................... 7 Licensed Facility index.................................................................. 8 til I

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

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 Technical Information and Document Control Policy and Publications Management Branch Publishing and Translations Section i

l Woodmont 501 U.S. Nuclear Regulatory Commission Washington, D.C. 20556 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, and NUREG/CR-XXXX. These precede the following indexes:

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

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

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

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author, (5) date report was published, (6) number of pages in the report, (7) the N RC Document Control System accession number, (8) the microfiche address (for internal NRC use).

Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND

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RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National l

Laboratory. May 1981.141 pp. 8105280299. ANL-813. 08632:070.

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

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

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

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

v

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

ADD - addendum i

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 i

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.

vi

Main Citations and Abstracts The report listings in this com?lation are ar-is an NRC contractor-prepared report. The i

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 i

NUREG/CP-XXXX is an NRC-sponsored report.

conference report, and NUREG/CR XXXX NUREG-0020 V09 N12: LICENSED OPERATING REACTORS ity. There were four abnormal occurrences at the other NRC li-STATUS

SUMMARY

REPORT. Data As Of November censees. Three of the events involved medical misadministra.

30,1985.(Gray Book l} ROSS,P.A.; BEEBE.M.R. Division of tions. two therapeutic and one diagnoste. The other event in-Budget & Analysis. January 1986. 443pp. 8602140176.

volved exposure of radiographic personnel due to management a

34611:074.

and procedure control defciencies. There were no abnormal oc-The OPERATING UNITS STATUS REPORT - LICENSED OP-currences reported by the Agreement States. The report also ERATING REACTORS provides data on the operation of nucle.

contains information updating some previously reported abnor-ar units as timely and accurately as possible. This information is mal occurrences.

collected by the Office of Resource Management from the NUREG-0304 V10 N04: REGULATORY AND TECHNICAL ent.

NR s Rego at O c s a for ties The hree REPORTS. Annual Compilation for 1985.

• Division of Technical sections of the report are: monthly highlights and statistics for Information & Document Control. February 1986. 239pp.

d ta a compi on f d ta!!

his I nc es il formal reports in the NUREG series n

on ea nit p ed by NRC's Regional Offices, IE Headquarteis and the utilitres; prepared by the NRC staff and contractors, as well as proceed-i and an aopendix for miscellaneous information such as spent ings of conferences and workshops. The entnes in the compila-fuel storage capability, reactor. years of exponence and non.

tion are indexed for access by title and abstract, contractor power reactors in the U.S. It is hoped the report is helpful to all report nt. nber, personal author, subject, NRC organization, con-agencies and individuals interested in maintaining an awareness tractor, and hcensed facility.

of the U.S. energy situat'on as a whole-NUREG-0325 R08: U.S. NUCLEAF, REGULATORY COMMISSION NUREG-0020 V10 N01: LICENSED OPERATING REACTORS FUNCTIONAL ORGANIZATION CHARTS.

• Offce of Resource STATUS

SUMMARY

REPORT. Data As Of December Management, Director. January 1986. 57pp. 8601280266.

31,1985.(Gray Book 1). ROSS.P.A.; BEEBE,M.R. Division of 34384:021.

t Budget & Analysis. February 1986. 428pp. 8603140523.

Functional organizaton charts for the NRC Commission Of-34968.259.

fices, Divisions, and Branches are presented.

See NUREG-0020,V09,N12 abstract.

NUREG-0386 DO4: UNITED STATES NUCLEAR REGULATORY NUREG-0020 V10 N02: LICENSED OPERATING REACTORS COMMISSION STAFF PRACTICE AND PROCEDURE STATUS

SUMMARY

REPORT. Data As Of January DIGEST. JULY 1972 - JUNE 1985.

• Office of the Executive 31,1936 (Gray Book I)

• Dnnsion of Budget & Analysis. March Legal Director. *,' spen Systems, Inc. January 1986. 742pp.

j 1986. 442pp. 8604040589. 35406:250, 8602180031. 34628.061.

j See NUREG-0020,V09,N12 abstract.

This edition of the NRC Staff Practice and Procedure Dioest contains a digest of a number of Commission, Atomic h NUREG-0040 V09 N04; LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. October 1985-De-and Licensing Appeal Board, and Atomic Safety and i cember 1985 (White Book)

• Division of OA, Vendor & Techni-Board decisions issued dunng the penod July 1,1972 to June 30,1985 interpreting the NRC's Rules of Practice in 10 CFR cal Training Center Programs (Post 850212). February 1986.

Part 2. This edition replaces earlier editions and supplements 255pp. 8603180521. 35137:180.

and includes appropriate changes reflecting the amendment to This penodical covers the results of inspections performed by the Rules of Practice effective June 30,1985.

the NRC's Vendor Program Branch that have been distnbuted to the inspected organizations dunng the penod from October NUREG-0430 V06 N01: LICENSED FUEL FACILITY STATUS 1985 through December 1985. Also included in this issue are REPORT. inventory Difference Data. January-June 1985.(Gray the results of certain inspectons performed prior to October Book 11)

• Director's Office, Office of Inspection and Enforce-1985 that were not included in previous issues of NUREG-0040.

ment. February 1986.15pp. 8603130320. 34947:117.

NUREG-0090 V08 NO3: REPORT TO CONGRESS ON ABNOR.

8

""9 89*"##

MAL OCCURRENCES. July. September 1995.

• AEOD, Direc.

the informaton and completion of' any related NRC investiga.

tor's Office. February 1986. 48pp. 8604030385. 35386-288 Section 208 of the Energy Reorganization Act of 1974 identh t ns. Informaton in this report includes inventory difference

• "9 fies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to

[r"a*niu u on

,ou nium-23 '

be significant from the standpoint of public health and safety and requres a quarterly report of such events to be made to NUREG-0525 R11: SAFEGUARDS

SUMMARY

EVENT LIST Corgress This report covers the penod July 1 to September (SSEL). SMITH,H. Facility Assessment & Standardizat on 30,1985. Dunng the report penod, there were three at, normal Branch. January 1986. 58pp 8602060007. 34487.339.

occurrences at the nuclear power plants beensed to operate.

The Safeguards Summary Event List provides bnef summa.

These events involved, respectrvely (1) management control nes of hundreds of safeguards-related events involving nuclear deficiencies, (?) inoperable steam generator low pressure trip, matenal of facilitres regulated by the U S. Nuclear Regulatory j

and (3) management defcencies at Tennessee Valley Author.

Commisson. Events are described under the categories: bomb-1

2 Main Citations and Abstracts l

I l

related, intrusion, missing / allegedly stolen, transportation <elat-This report provides the status and results of the NRC Th.

ed, tampenng/ vandalism, arson, firearms-related, radiological moluminescent Dos: meter (TLD) Direct Radiaton Monitoring t

sabotage, nonradiological sabotage, and miscellaneous. Infor-Network. It presents the rad:ation levels measured in the vicinq maton in the event descriptions was obtained from official NRC of NRC licensed facility sites throughout the Country for the thirti reports.

quarter of 1985.

NUREG-0540 V07 N11: TITLE LIST OF DOCUMENTS MADE NUREG-0853 S05: SAFETY EVALUATION REPORT RELATED PUBLICLY AVAILABLE. November 1-30,1985.

• Division of TO THE OPERATION OF CLINTON POWER STATION. UNIT Technical Information & Document Control. January 1986.

NO.1. Docket No. 50-461.(llhnois Power Company)

• Office of 415pp. 8601280345. 34390:250.

Nuclear Reactor Regulaton, Director (post 851125). January This document is a monthly publication containing descrip-1986. 257pp. 8601290240. 34420.056.

l tions of information received and generated by the U.S. NRC.

Supplement No. 5 to the Safety Evaluation Report on the ap-This information includes (1) docketed material associated with phcation fi;ed by llhnois Power Company, Soyland Power Ccro-t civilian nuclear power plants and others uses of radioactive ma-erative, Inc., and Western Illinois Power Cooperative, Inc. as ap-terials, and (2) nondocketed material recerved and generated by plicants and owners, for a license to operata the Chnten Pow 6, NRC pertinent to its role as a regulatory agency. The following Staten, Unit No.1, has been prepared by t'e Office of Nuclear indexes are included: Personal Author index, Corpof ate Source The facility is located in Harp TownsNp, %pikog Commiss Reactor Regulation of the U.S. Nuclear He Index, Report Number index, and Cross Reference to Pnncipal v% County, tilinois.

Documents index.

This supplement reports the stacs of items that have been re-NUREG-0540 V07 N12: TITLE LIST OF DOCUMdNTS MADE soi.ed by the staff since Supplement No. 4 was issued.

PUBLICLY AVAILABLE. December 1-31,1985.

Division of Technical Information & Document Control February 1980.

NUREG-0885105: U.S. NUCLEAR REGULATORY COMMISSION 362pp. 8603030048. 34773:001.

1986 POLICY AND PLANN:NG GUIDANCE.

• Commissmers.

l See NUREG-0540,V07,N11 abstract.

February 1986. 58pp. 8602280729. 34754197, i

Tne purposes of the Policy and Planning Guidance document NUREG-0540 V08 N01: TITLE LIST OF DOCUMENTS MADE are to set forth the regulatory approach of the Nuclear Regula.

PUBLICLY AVAILABLE. January 1-31,1986.

• Division of Techni-tory Commission and to provide the supporting principles to the cal Informaton & Document Control March 1986. 400pp.

approach; to state the major policies and planning objectives of 8603260371. 35226.123-the Commission; and to provide a common basis for the devel-See NUREG.0540,V07,N11 abstract.

opment of programs, for the establishment of prionties, and for NUREG-0748 V05 N11: OPERATING REACTORS LICENSING the allocation of resources.

ACTIONS

SUMMARY

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

• Management Support Branch. February 1986. 323pp.

NUREG 0887 S08: SAFETY EVALUATION REPORT RELATED 8602280505. 34726.283.

TO THE OPERATION OF PERRY NUCLEAR POWER The Operating Reactors Licensing Actions Summary is de.

PLANT, UNITS 1 AND 2. Docket Nos. 50-440 And 50-441.(Cleve-signed to provide the Management of the Nuclear Regulatory land Electric illuminating Cornpany)

• Divisen of Boihng Water Commission (NRC) with an overview of heensing actions dealing Reactor (BWR) Licensing. January 1986.123pp. 8601290263.

with operating power and nonpower reactors.

34419:290.

This report relates to the applicaton for licenses to operate NUREG-0750 V22102: INDEXES TO NUCLEAR REGULATORY the Perry Nuclear Power Plant, Units 1 and 2 filec ay the Cleve-COMMISSION ISSUANCES FOR JULY-DECEMBER 1985.

• Di-land Electnc illuminating Company on tahalf of itself and as vision of Technical Information & Document Control. March agent for the Duquesne Light Company, the Ohio Edison Com-l 1986.109pp. 8604040477. 35412:044.

pany, the Pennsylvania Power Company and the Toledo Edison Digests and indexes for issuances of the Commission, the Corrpany as apphcants and owners, The Perry Nuclear Power Atomic Safety and Licensing Appeal Panel, the Atomic Safety Plant facility is located in Lake County, Ohio, approximately 35 and Licensing Board Panel, the Administratrve Law Judge, the maes northeast of Cleveland, Ohio. This Supplement, No. 8, ad-Director's Decisions, and the Denials of Petitions for Rulemak*

dresses the staus of certain issues that had not been resolved ing are presented.

at the time of publication of the Safety Evaluation Report and NUREG-0750 V22 N05: NUCLEAR REGUL ATORY COMMISSION Supplements 1 through 7. This supplement addresses the items ISSUANCES FOR NOVEMBER 1985. Pages 771873.

• Division conceming the issuance of a low power heense (5%).

of Technical Information & Document Control. January 1986.

115pp. 8602030017. 34462:164.

NUREG-0887 509: SAFETY EVALUATION REPORT RELATED Legalissuances of the Commission, the Atomic Safety and Li.

TO THE OPERATION OF PERRY NUCLEAR POWER censing Appeal Panel, the Atomic Safety and Licer. sing Board PLANT, UNITS 1 AND 2. Docket Nos. 50-440 And 50-441.(Cleve-Panel, the Administrative Law Judge, and NRC Program Office.

land Electric illuminating Company)

• Division of Boiling Watar React 7r (BWR) Licensing. March 1986. 29pp. 8603180524.

NUREG-0750 V22 N06: NUCLEAR REGULATORY COMMISSION 35139.107, ISSUANCES FOR DECEMBER 1985. Pagos 875-982.

• Divis,on This report relates to the apphcation for hcenses to operate of Technical Informaton & Document Centrof. February 1968.

the Perry Nuclear Power Plant, Units 1 and 2 filed by the Cleve-113pp. 8603100047. 34871:250.

land Electnc liluminating Company on behalf of itself and as See NUREG-0750,V22,N05 abstract.

agent for the Duquesne Light Cor9pany, the Ohio Edison Com-NUREG-0750 V23 N01: NUCLEAR REGULATORY COMMISSION pany, the Pennsylvania Power Company and the Toledo Ectson ISSUANCES FOR JANUARY 1986. Pages 147.

• Division of Company as apphcants and owners. The Perry Nuclear Power Technical Information & Document Control. March 1986. Sepp.

Plant facility is located in Lake County, Ohio, approximately 35 8604030276. 35392.052.

miles northeast of Cleveland, Oho. This supplement ceports the See NUREG-0750,V22 N05 abstract.

staFs evatuaton findings pertaining to the earthr.,uake eyesit j

that occurred in the vicarwty of the Perry Nuclear Power Plant NUREG-0837 V05 N93: NRC TLD DIRECT RADIATION MONI-site on January 31, 1986, and is hmited to that evaluation.

TORING NETWORK. Progress Report, July-September 1985.

Future suppemental reports wdl continue reporting on the JANG,J.I RABATIN,K.: COHEN,L Region 1, Office of Director.

status of new or unresolved issues since Supplement No. 8 was

j January 1986. 219pp. 8601160529. 34293 219.

issued in January 1986.

{

t-l Main Citations and Abstracts 3

NU84EG-0933 M4: A PRIORITIZATION OF GENERIC SAFETY contractor has wntten a more complete and detailed annual I?WS. EMmT,rt.: MINNERS.W.; VANDERMOLEN.H.: et al.

report of their work which can be obtained by wnting to NRC; Non of Sat.<y Review & Oversight (post 851125). February however, we beheve it is useful to have a summary of each 1*gM 25G)p. 860324091. 35208:143.

contractor's efforts for the year combined into one volume.

&,s report presents the prionty rankings for generic safety ices relaaed to nuclear power plants. The purpose of these NUREG-1031 Sob: SAFETY EVALUATION REPORT RELATED rankings is to assist in the timely and efficient a.locatioa of NRC TO THE OPERATION OF MILLSTONE NUCLEAR POWER 4

msources for the tesolut're of those safety issues that have a STATION UNIT No.3. Docket No. 50-423. (Northeast Nuclear nq iiticant mtential for rJtV :ing nsk. The safety prionty rankings Energy Company)

• Division of Pressunzed Water Reactor Li-are HIGH, MEDIUM, LO4. and DROP and have been assigned censing - A (post 851125). January 1986. 76pp. 8603050091.

s on the basis of nsk signNanca estimates, the ratio of risk to 34801103.

costs and other impacts estimated to result if resolutens of the The Safety Evaluaton Report issued in August 1984 provided safety issues were implemented, and the considerabon of un.

the results of the NRC Staff review of Northeast Nuclear Energy certainties and other quarAtative or qualitative factors. To the Company's apphcation for a hcense to operate the Millstone Nu-4 extent practicel, estimatcs are quantitative. -

clear Power Staten, Unit No. 3. Supplement No.1 to that report, issued in March 1985 updated the informaticn contained NUREG-0936 V'J4 N04: NRC REGULATORY AGENDA.Ouarterfy in the Safety Evaluahon Report and addressed the ACRS 4

Report October-December 1985.

Division of Rules and Report issued on September 10,1984. Supplement Nos. 2 and Re:ordt Mrh 19M 209pp. 8603260362. 35225:026.

3 dated September 1985 and November 1985 respectrvely, up-The NRC Wegulawry Agonda is a compilaton of all rules on dated the information contained % the Safety L'valuaton Report which the NRC has proposed or is considenng action and all and Suoplement No.1 and adessed pnor unresolved items.

pe" ions for ru'emaking which have been received by the Com-Supplement No. 4, issued in f(vember 1985 addressed the

m. son and asi pending disposibon by the Commisson. The items conceming tne issuance of a Icw power Ecense (5%).

Regulatory Agenda is upualcd and issued each quarter. The The low power license, NPF-44 was issued on November 25 Ager.das for April and October are published in ther entirety en 1985. This Supplement No. 5 addresses the items relating to the FCDERtt REGISTER while a notice o'. avMabahty is pub-the issuance of a full power license (100%). The facility is locat-hs o n the FFDERAL FEGISTER for the January and July ed in Waterford Township, New London County, Connecticut.

NUREG 0940 V04 N04: ENFORCEMENT ACTIONS.SIGNIFICANT NUREG 1100 V02: FY 1987 BUDGET ESTIMATES.

• Divisio#, of ACTIONS RESOLVED.QuarSrfy Progress Report, October-De-Budget & Analysis. January 1986. 132pp. 6602140398.

cermer 1985.

• Director's Office Offee of Inspection and En-34612:159.

forcement. February 1986. 383pp. 8603140471. 34967:236.

This report contains the fiscal year budget justifications to This compilaton summsnzes signifcant enforcement actons Congress. The budget estimates for salanes and expenses for that have been resolved during one quarterly pered (October -

fiscal year 1987 provide for obligations of S405,000,000 to oe December 1985) and snelodes copies of letters, notees, and funded in total by a new appropnat on.

orders sent by the Nuclear Regulatory Commisson to licensees NUREG-1101 V01: ONSITE DISPOSAL OF RADIOACTIVE with respect to these enforcement actons and the licensees' WASTE. Guidance For Disposal By Subsurface Burial regenses. tt is antL:ipated that the information in this publica-NEUDER.S.M. Division of Waste Management. March 1986.

tion will be widely diysemrnated to managers and employees 26pp. 8604030412. 35401:287, eryaged in act vities icensed by the NRC, in the interest of pr*

Volume 1 of this NUREG provides guidance pnmarily for aca-moting public health and safety as well as common defense domic, medical, and industrial licensees seeking authorization to a t secuntA dispose of small quantibes of radioactive material by onsste sub-NUREG-0954 S05: SAFETY EVALUATION REPORT REtMED surface disposal. Licensee requests for such authon2ations are TO THE OPERATION OF CATAWBA NUCLEAR made pursuant to Section 20.302 of 10 CFR Part 20 " Standards STATION. UNITS 1 AND 2. Docket Nos. 50-413 And 50-for Protection Against Radiation." This guidance supplements 414.(Duke Power Company)

• Division of Pressunzed Water Re-Sect.on 20.302 to assure that appropnate information is provid-actor Licensing - A (pott 851125). February 1986. 193pp, ed by the hcensee so that an adequate evaluation of the appli-8603140274. 34962:040.

caten can be performed by NRC staff. In addition, this guidance l

This report supplements the Safety Evaluation Report provides a descnption of dispost! methods and techniques ac-(NUREG.0954) issued in February 1983 by the Office of Nucle-ceptable to the NRC staff in its evatuation of the application.

ar Reactor Regulation of the U.S. Nuclear Regulatory Commes.

This guidance also identifies categones of radionuclides with re-sion with respect to tha applicaton filed by Duke Power Compa-spect to total radioactivity, waste packaging, burial frequency ny, North Carotina Muncipal Power Agency Number 1, North and other conditions acceptable for subsurface disposal. The Carohna Membership Corporaten. Saluda River Electric Cooper-category hsts of radionuchdes and associated disposal condi-ative, Inc., and Piedmont Munic: pal Power Agency, as applicants tons and entena are the primary data by which the NRC staff and owners for licenses to opera'e the Catawba Nuclear Sta.

will make an irvtial evaluation of inicematon provided in the ap-ton, Un'ts 1 and 2 (Docket Nos 50 413 und 50-414, respective-plicaton. An application for a proposed disposal actvity that ty). The facihty is located in Ycrk County, Snuth Carolina, ap-does not fit any of the disposal categones will be evaluated proxirratcfy 9 6 km (6 m!) north of RocA Hill and adjacent to against more complete guidance also desenbed in this docu-Lake Wyle. This supplement provides additonal information ment. Volume 2 of this NUREG desenbes techncal methodolo-supportmg the hcense for initial entcality and power ascension gy used by NRC staff to evaluate requests by hcensees for ap-to full power cperation for Unit 2.

proval of onsste disposal by bunal in soit NUREG-0975 V04: COMPILATION OF CONTRACT RESEARCH NUREG 1109 DRFT FC: REGULATORY ANALYSIS FOR THE FOR THE MATERIALS ENGINEERING BRANCH. DIVISION OF RESOLUTION OF UNRESOLVED SAFETY ISSUE A-44,STA.

ENGINEERING TECHNOLOGY. Annual Report For FY 1985.

• TION BLACKOUT. RUBINAM. Division of Safety Renew A Matenals Engineonng Branch.

March 1986.

772pp.

Oversig5t (rost 851125). January 1966. 40pp. 8601290262.

t 8603280266.35322:098.

34419 153.

This report presents summanes of the research work per-

" Station Olackout" is the complete loss Cf attemating current formed during Fiscal Year 1985 by laboratories and organiza-(AC) electnc power to the essential and nonessential buses in a tions under contracts administered by the NRC's Matenais Engi-nuclear powy plarat it results when both offsite power and the neenng Branch. Offee of Nuclear Regulatory Research. Each onsite emergency AC power systems are unavailable. Because I

4 Main Citations and Abstracts many safety systems required for reactor core decay heat re-partment of Energy. On February 5,1986, the NRC submitted moval and containment heat removal depend on AC power, the its pnncipal comments to the Department of Energy.

consequences of a station blackout could be severe. Because of the concern about the frequency of loss of offsite power, the NUREG-1171 DRFT: DRAFT ENVIRONMENTAL STATEMENT number of failures of emergency diesel generators, and the po-RELATED TO OPERATION OF THE SOUTH TEXAS tentially severe consequences of a loss of all AC power, "Sta.

PROJECT, UNITS 1 AND 2. Docket Nos. 50-498 And 50-tion Blackout" was designated as Unresolved Safety issue (USI) 499.(Houston Lighting And Power Company)

• Division of Pres-A-44. This report presents the regulatory analysis for USl A-44.

sunzed Water Reactor Licensing - A (post 851125). March it includes: (1) a summary of the issue, (2) the proposed techni-1986. 248pp. 8603260369. 35225:235.

cal resolution. (3) a!temative resolutions considered by the Nu.

The information in this stamment is the second assessment clear Regulatory Commisson (NRC) staff, (4) an assessment of of the environmental irraact associated with the construction the benefits and costs of the recommended resolution, (5) the and operation of the South Texas Project. Units 1 and 2. locat-decisiori "onale, and (6) the relationship between USl A-44 ed in Matagorda County, Texas. The first assessment was the and oth, i programs and requirements.

Final Environmental Stateraent related to constraction, issued in March 1975 pnor to issuance of construction permits for South NUREG-lit AFETY EVALUATION REPORT RELATED TO Texas. As of December 1985, South Texas Unit 1 was 92.3%

THE REN L OF THE OPERATING LICENSE FOR THE RE.

complete and Unit 2 was 60.1% complete. The projected fuel SEARCH HEACTOR AT PENNSYLVANIA STATE load date for Unit 1 is June 1987. The present assessment is UNIVERSITY. Docket No.50-005,(Pennsylvania State. University) the result of the NRC staff review of the activities associated

• Dmsson of Pressunzed Water Reactor Licensing - B (post with the proposed operation of the plant.

851125). January 1986. 79pp. 8601280120. 34386 047.

This Safety Evaluaton Report for the apphcation filed by the NUREG-1176: TECHNICAL SPEt,lFICATIONS FOR MILLSTONE Pennsylvania State Unrversity for a renewal of Operating Li-NUCLEAR POWER STATION. UNIT NO. 3. Docket No. 50-cense R.2 to continue to operate the Penn State Breazeale Re-423.(Northeast Nuclear Energy Co)

• Dn,inon of Pressunzed actor (PSBR) has been prepared by the Office of Nuclear Reac-Water Reactor Licensing - A (post 851125). January 1986.

tor Regulation of the U S. Nuclear Regulatory Commission. The 490pp. 8602280674. 34759.142.

facihty is owned and operated by the Pennsylvania State Uni-The Millstone Nuclear Power Station. Unit No. 3 Technical versity and is located on the campus in University Park, Penn-Specifications were prepared by the U.S. Nuclear Agulatory sytvana. Based on its technical review, the staff concludes that Commission to set forth limits, operatng condituas, and otkr the reactor facihty can continue to be operated by the University requirements applicable to a nuclear reactor facihty as set forth without endangenng the health and safety of the public, or the in Section 50.36 of 10 CFR Part 50 for the protrction of the environmer:t.

health and safety of the public.

NUREG-1159: TRAIN:NG MANUAL FOR URANIUM MILL WORK-NUREG 1179 V01: RUPTURE OF MODEL doY UF6 CYLINDER f

ERS ON HEALTH PROTECTION FROM URANIUM.

AND RELEASE OF URANIUM HEXAFLUORIDE.Sequoyr's MCELROY.N L.; BRODSKY,A. Divison of Rad;ation Programs &

Fuels Facil4y, Gore, Oklahoma. January 4,1986. SMITH.R D.;

Earth Sciences (post 840429). January 1986. 33pp-CAIN.C.L; CHAPPELL,R. N7C No Detailed Affihation Given.

This e rt ve informaton for uranium mill workers to am on Jamay 4 M6, a M M M @@

help them understand the radiaton safety aspects of working wa w

was w;th u'anium as it is processed from ore to yellowceke at the mdis. The report is designed to suppiement the radiat on stfety 8

training provided by uranium mills to their workers. It is wntten in inhaled hydrogen fluonde fumes. a' reacton prodt,1 of UF6 and an easdy readable style so that new employees with no previ ahe Mn Mal m mum wm inW W N ous exper'ence working with uranium or radier on can obtairi a "O"

basic undeending of the nature of radiation and the parhcular safety reqwremt its of working with uranium. The report should gen Hu&te and a second reacton product, uranyl fluonde. The be helpful to mail operators by providing trainirig matenal to sup-interval of release was approximately 40 minutes. The cyhnder, port their rad.ation safety trasning programs' which had been overfilled, ruptured while it was being heated NUREG-1162: TECHNICAL SPECIFICATIONS FOR PERRY NU.

becausa of the expansion of UF6 as rt changed from the sohd CLEAR POWER PLANT. UNIT 1. Docket No. SS440 (Cleveland to the hquid phase. Tne maximum safe cacacity for the cyhnder Electnc illuminating Company).

• Dvision of Boh.ag Water Reac.

is 27,560 pounds of product. Evidence indicates that it was for (BWR) Licensing. March 1986. 200pp 86040303B3.

filled with an amount exceeding this hmit.

35385:169.

The Perry Nutear Power Plant. Unit No.1 Technecal Specifica.

NUREG-1182: TECHNICt.L S?ECIFICATIONG FOR CATAWBA tions were prepared by the U.S. Nuclear Regulatory Commis.

NUCLEAR STATION,UN;TS 1 AND 2 Docket Nos. 50-413 And sion to set forth hmts. opeqting conditions, and other require-50-414.(Duke Power Company)

• Drvision of Pressunred Wator toents applicable to a nuclear reactor facihty as set forth in Sec.

Reactor Licensing - A (post 851125). February 1986. %4pp.

tsor. 50.36 of 30 CFR Part 50 for the protucton of the health 8603130343. 34943.126.

and safety of the pubhc.

The Catawta Nuclear Staten. Units 1 and 2 Technicas Sneci-ficatons were prepared by the U S. Nuclear Regulatory Com-NUREG 1168: STAFF EVAt UATION OF U.S. DEPARTMENT OF mission to set for*.. the lirr.its, operating conditions, and other ENERGY PROPOSAL FOR MONITORED RETRIEVABLE requirements apphcable to a nuclear reactor facihty as set forth STORAGE

• Dvimon of Fuel Cycfe & Matenal Safety. March in Section 50.36 cf 10 CFR 50 for the protection of the he:Jth 1986.120pp. 8604040103. 35413 291.

and safety of the puthc.

As directed by the Nuclear Waste Policy Act of 1982, the U.S D partrnent of Energy has prepared a proposal for the U.S NUREG-1189 V01: ASSESSMENT OF THE PUBLIC HEALTH Cong<ess for a facihty (Nat can be used for the monitored re-IMPACT FROM THE ACCIDENTAL RELEASE OF UF6 AT THE trievable storage of spent fuel from comme = :ial users. This SEOUOYAH FUELS CORPORATION FACILITY AT report descroes the evaluaton performed by the staff of the GCRE Or'_AHOMA Repnnted March 26,1986.

• Ad Hoc Pubhc U.S Nuclear Regulatory Commisson of the design concepts for Hea.th Assessment Task Force March 1986. 99pp.

tile monttored retrievable storWe facshty proposed by the De-8603270332.35344 253.

Main Citations and Abstracts 5

i 1

i Following the accidental release of UF(6) from the Sequoyah Rancho Seco by the NRC Executive Drector for Operatens in Fuels Facilty on January 4,1986, an Ad Hoc interagency Public conformance with NRC's recently established incident investi.

Hea'th Assessment Task Force was established. The Task gation Program.

Force consists of technical staff members from vanous agen-cies who have prepared this assessment of the public health NUREG/CP-0068: PROCEEDINGS OF THE INTERNATIONAL impact associated wth the accidental release. The assessment NUCLEAR REACTOR DECOMMISSIONING PLANNING CON-is based on data from the accident available as of February 14, FERENCE. BAUMANN.B.L: EDWARDS,K.M. UNC Nuclear in-l 1986, and describes the chemical and radiological effects from dustnes. February 1986. 496pp. u60318000f. 35132:115.

the intake of uranium and fluonde. Volume 1 of the report de.

This report contains summaries of papers presented at the i

j scnbes the effects from the intake of uranium and fluoride and International Nuclear Reactor Decommessening Planning Con-summarizes the findings and recommendations of the Task ference held in Bethesda, Maryland, July 16-18, 1985. The pur-Force. Volume 2 of the report consists of Appendees which pose of this conference was to bring together nuclear scientists, provide more detailed information used in the assessment engineers, and government and industry professionals to ex-change expenence and technology informaton concerning nu-1 NUREG-1189 V02: ASSESSMENT OF THE PUBUC HEALTH clear reactor decommissioning planrung. Also included in this

)

IMPACT FROM THE ACCIDENTAL RELEASE OF UF6 AT THE report are transats of the keynote speech presented by G. A.

~

SEQUOYAH FUELS CORPORATION FACLITY AT Arlotto of the U.S. Nuclear Regulatory Commission and the GORE, OKLAHOMA.

• Ad Hoc Public Hea!!h Jssessment Task luncheon speech by Dr. Dixy Lee Ray, former governor of Force. March 1986. 424pp. 8603270335. 35312:036-Washington State and former chairman of the U.S. Atomic See NUREG-1189,V01 abstract.

Energy Commission.

NUREG-1190: LOSS OF POWER AND WATER HAMMER EVENT NUREG/CP-0072 V01: PROCEEDINGS OF THE THIRTEENTH AT SAN ONOFRE, UNIT 1 ON NOVEMBER 21,1985.

• Incident WATER REACTOR SAFETY RESEARCH INFORMATION nyestgation Team. Janua y 1986. 200pp. 8602120333.

MEETING. WEISSAJ. Brookhaven National Laboratory.

On November 21, 1985, Southem Cahfomia Edison's San Office of Nuclear Regulatory Research, Director. February 1986.

49 8603 7 508 93 Oncfre NucTar Gmeratrv-Staten, Unit 1, located south of San 9

ns 151 papers out of the 178 Clemente, Califomia, experienced a partG1 loss of innlant ac electncal power white the plant was operating at 60 percent that were presented at the Thirteenth Water Reactor Safety Re-power. Following a manual reactor tnp, the piar.t lost all oplant seana InfortuaEn Me6ong held at the National Bureau of ar, power for 4 mintes and expenenced a severe incidence of Standards, GaithersbuG Maryland, during the week of October water hammer in the feedwater system which caused a leak

• 22 25, 1985 The papers are printed in the order of their pres-

[

damaged plant equipment, and challenged the integrity of the entaton in each sessic, and describe progress and results of j

4 plant's heat senk. TPe most significant aspect of the event in-proy3ms in nuclear sa.fety research conducted in this country I

volved the fadure of five safety-related check vales in the feed-and abroad. Foreign participation in the meeting included thirty-water system whose failure occurred in less than a year, without one papers presented by researchert from Japan, Canada and detection, and jeopardized the integnty of safety systems. The eight European countries. The t;tles of tN papers and the event involved a number of equipment malfunctions, operator naines of the authors have been updated and enay differ from i

errors, and procedural deficiences. This report documents the those that appeared in che final program of the meeting.

findings and conclusions of an NAC incident Irwestgation Team NUREG/CP-0072 V02: PPOCEEDINGS OF THE THIRTEENTH sent to Se:1 Onofre by the NRC Executive Director for Oper-WATER REACTOR SAFETY RESEARCH INFORMATION ations in co.iformance with NRC's recently estabhshed incident MEETING. WEISS,A.J. Brookhaven National Laboratory.

Investigation Program.

Office of Nuclear Regulatory Research, Drector. February 1966.

i NUREG-1195: LOSS OF INTEGRATED CONTROL SYSTEM 581pp. 6603170473. 34991:288.

j POWER AND OVERCOOLING TRANSIENT AT RANCHO See NUREG/CP.0072,V01 atstract.

EECO ON DECEMBER 26.1985.

• Incident investigaton Team.

February 1986.150pp. 8603100574. 34892:221.

NUOEG/CP-0072 V03: PROCEEDINGS OF THE THIRTEdNTH On December 26, 1905, Rancho Seco Nuclear Generating WATER REACTOR SAFETY RESEARCH INFORMATION Station, located in Clay, California, about 25 miles southeast of MEETING. WEISS,A.J. Brookhaven Nat>onal Laboratory.

j Sacr &mento, exporienced a loss of dc power within the integrat.

Office of Nuclear Regulatory Research, Director. February 1986.

ed control system (ICS) whde the plant was operating at 76 per.

439pp. 6603140509. 34969:327.

=

I cent power. The plant is owned by the Sacramento Municipal See NUREG/CP-0072,V01 abstract.

Utahty Dstnct (SMUD). Following the ic2s of 6CS de power, the i

NUREG/CP-0072 V04: PROCEEDINGS OF THE THIRTEENTH reactor tripped on high reactor coolant system (RCS) pressure s

WATER REACTOR SAFETY RESEARCH INFORMATION foPwed by a rapid overcooting transient and automatic initiation j

MEETING. WEISS,A.J. Brookhaven National Laboratory.

l of the safety features actuation system on low RCS pressure.

j The overcoobng transient continued until ICS dc power was re-Office of Nuclear Regulatory Research, Drector. February 1986.

j stored 26 minutes after its loss. The furdamental causes for 471pp. 86031405' t, 34965:288.

this transient were design weaknesses and vulnerabdities in the See NUREG/CP-0072,V01 abstract.

r ICS and in the equipment controlled by that system. These NUREG/CP-0712 V05: PROCEEDiMGS OF THE ThlRTELNTH

[

weaknesses and vulnerabdities were not adequately compensat-WATER REACTOR SAFETY RESEARCH INFORMATION ed by other design features, plant procedures or operator train-MEETING. WEISS,A.J. Brookhaven National Latinatory.

ing. These weaknesses and vulnerabdities were largely known Office of Nuclear Regulatory Research, Drector. February 1986.

to SMUD and the NRC staff by virtue of a number of precursor 464pp. 8603170471. 34990:184.

events and through related ana'yses and s'udies. Yet, adequate See NJREG/CP 0072,V01 abstract.

plant modifications were not made so that this event would be v

improbable, or so that its course or consequences would be al-

. UREG/CP-0072 V06: PROCEEDINGS O WE THIRTEENTH tered signihcantly The information was available and known WATER PEACTOR SAFETY RESEM CH INFORMATION e

which could have prevented this overcooling transsent; but in MEETING. WEISS.A.J. Brookhaven National Laboratory.

1 the absence of adequate plant modifications, the incident Office of Nuclear Regulatory Rc9earch, Drector. February 1986.

should have been espected. The report includes findings and 454pp. e603180054. 35131:030.

conclussons of the NRC incident investigation Team sent to See NUREG/CP-0072,V01 abstract i

f i

A m.

6 Main Citations and Abstracts l

1 NUREG/CP-0073: PROCEEDINGS OF THE WORKSHCP ON tor Safety Experiments. Plant Analyzer, Code Assessment and LARGE 1RRADIATOR RADIATION SAFETY. LUBENAU,J.O.:

Apphcaton, Code Maintenance (RAMONA-38), Benchmarking NUSSBAUMER.D.A. Office of State Programs, Director. March and Verfhcaten of LWR Severe Accident Codes; Stress Corro-1986. 21pp. 8604100130. 33506 045.

son Cracking of PWR Steam Generator Tubing. Probability This document reports thc resuits of an NRC-sponsored Based Load Combinatens for Design of Category 1 Structures, workshop on regulatory csnsideratens for large (poobtype) irra*

Soil-Structure interachon Evaluations, Identifcabon of Age-Re-diator radiabon safety. The workshop focused on the followng lated Failure Modes; Application of HRA/PRA Results to Re-regulatory activities: 1. censing, constructon OA, source tcading solve Human Rehability and Human Factors Safety issues, PRA inspecten, preoperational inspection, and inital and routine in-Technology Transfer Program, Protective Action Decisionmak.

l spections Workshop participar.ts developed consensus views ing, and Operational Safety Rehabikty Research.

on most of the key elemects of each of these regulatory activi-i

'res. The workshop report hns been prepared to serve as a radi-NUREG/CR-2907 V03: RADIOACTIVE MATERIALS RELEASED 4

aton safety reference sourcebook in conlunction with available FROM NUCLEAR POWER PLANTS. Annual Report 1982.

national design standards, regulatory requirements and regula-TICHLER,J.; NORDEN K. Brookhaven National Laboratory. Feb-tory guides for irradiator designers and operators, regulatory staffs and for persons irwofved in developing safety standards ruary 1986. 224pp. 8602280775.

BNL.NUREG-51581.

for these facihties.

34725:237.

Releases of radioactive matenals in airbome and hqued ef.

NUREG/CR-2000 V04N12: LCENSEE EVENT REPORT (LER) fluents form commercial l,gh. water reactors during 1982 have COMPILATION For Month O' December 1985

• Oak Ridge Na-been compiled and reported. Data on schd waste shipments as tional Laboratory. Jarwary 1986.111pp. 6601280321. ORNL/

well as selected operating information have been included. This NSIC-200. 34390:139' report supplements earlier annual reports issued by the former This monthly report contains Licensee Event Report (LER) operational information that was processed into tne LER data Atomic Energy Commission and the Nuclear Regulatory Com-mission. The 1982 release data are summanzed in tabula

• form.

file of the Nuclear Safety Irdormation Center (NSIC) dunng the one month penod identified on the cover of the document. The Data ccwering specific radonuchdes are summanzed.

LERs, from which this information is derived, are submitted t NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFl.

the Nuclear Regulatory Commission (NRC) by nuclear power plant heerWees m accordance with federal regulations. Proce-CATION TESTS Semiannual Report Covering The Period April-dures for LER report.ng for revisions to those events occumng September 1985. SOO P.: ABRAHAM T.: ANDERSON,C.: et af.

pro to 1984 are desenbed ir' NRC Regulatory Guide 1.16 and Brookhaven National Laboratory. January 1986. 179pp.

NUREG-1061, Instructons for Preparaton of Data Entry Sheets 8602060390. BNL-NUREG.51630. 34494.001.

for Licensee Event Reports. For those events occumng on and Several studies were completed this penod to evaluate exper.

after January 1,1984, LERs are being subnutted in accordance 4mer'tal and anatytical methodologies being used in the DOE with the revised rule contained in Title 10 Part 50.73 of the waste package program. The first invofves a determination of Code of Feceraf Regulatons (10 CFR 50.73 Licensee Event the relevance of the test conditions being used by DOE to char-Report System) which was pubhshed in the Federal Register actenze waste package component behavior in a salt repository (Vol. 48, NO.144) on July 26,1983. NUREG 1022, Licensee system. Another study focuses on the testing conditions and Event Report System - Descrip'"on of Systems and Guidelines procedures used to measure radionuchde sokblity and colloid for Reporting, provides supporting guidarO and information on formation in repository groundwaters. An attempt was also the suised LER rule. The LER summaries in this report are ar-made to evaluate the adequacy of selected waste package per.

ranged alphabetically by facility name and then chronologically formance codes. However, the latter work was hmited by an in-by event date for each frAty. Component system, keyword, ability to obtain several codes from DOE. Nevertheless, it was and component vendor endexes follow the summanes. Vendors are those identified by the utility when the LER form is initiated; possible to comment bnefty on the strucsares and intents of the the keywords for the cornponent, system, and general keyword codes based on publications.in the open liters ure. The final Indexes are assigned by the computer using correlation tables study invoNed an expenraental program to determm6 the hkeh-from the Sequience Coding and Search System.

hood of stress.conosen cracking of etenttc stainless steels and incoloy 825 in simulated tuff repository environments. Tests NUREG/CR-1000 V05 N1: LICENSEE EVENT REPORT (LER) for six. month exposure penods in water and air-steam condi-l COMPILATION For Month Of January 1906

• Oak Ridge Na-tions are desenbed.

tional Laboratory. February 1986.144PP. 8003140520. ORNL/

+

NSIC 20C. 349G4:151.

NUREG/CR-3117: SEISMIC AND DYNAMIC OUALIFICATION OF See NUREG/CR-2000.V04.N12 abstract.

RELATED ELECTRICAL AND MECHANICAL EQUIPMENT.

NUREG/CR4000 V05 N2: LICTNSEE EVENT REPCAT (LER)

SUBUDHI,M.; CURRERI,J ; REICH,M. BrooMaven National Lab-COMPILATION For Month Of February 1986.

• Oak Ridge Na.

oratory. March 1986. 63pp. 8A04040468 BNL NUREG-51643.

tional Laboratar). March 1986.1 Fop. 8634070255. ORNL/

35405.051.

NSIC-200. 35447:1M This report presents a surnmary of rnethods and procedures

~

See NUREG/CR 2000,V?4,N12 abstratt.

that may be used for seismc quahficatons of nuclear power NUREG/CR 2331 V05 N2: SAFETY RESEARCH PROGRAuS plant (9ectlanical and electncal equement. Incorporated into SPONSCRED BY OFFICE OF NUCL EAR REGULATORY text are sectiorts that erplain and clanfy commonly used quahfi-RESEARCH.Ouarterff Progress Ar,/rt.Apnl bJune 30,1985.

cation terminoiogy and that dehneate methods used for dynamic WEISS.A.J Prookhaven National Laboratcry. December 1985.

environment simulations u ed for quahfcaten. The report also 1

107pp.8602240008. BNL-NUREG 51454. 34tsG4 007.

presents a scenano of what or: curs at a typical Seismc Quahfi-

~

This progress report will desenbe current activtics and techn cation Recew Tearr ESORT) sudit of a NTOL Nutcear Power cal progress in the program at Brookhaven Natonal Laboratory Station. One of the rnain purposes of th s report is to assure sponsored by the Divtsion of Accident Evaluaton, Divison of that industry raembers erwolved wth seismc quehfcations Le Engneenng Technology, and Dnnsen of Risk Anafysis & Oper-fuHy cogrw2 ant of all appiopnste informaton required during an atens of the 11S. Nuclear Regulatory Commesson, Offee of Nu-audit.

riear Regufatory Research. The protects reported are the foh lowing: High Tempe,rature Reactor Resaarch, SSC/MINET De.

NUREGICR 3365 DRF FC: REPORT TO THE NRC ON GUID-velopment, vahdaten and Apphcation Thermal-Hydrauhc Reac-L m..

m u.

m m

. m..-

.A i.-

i

Main Citations and Abstracts 7

ANCE FOR PREPARING SCENARIOS FOR EMERGENCY the problem by NRC, IE Bulletin 79-07 was issued April 14, PREPAREDNESS EXERCISES AT NUCLEAR GENERATING 1979 to all hcensees and permit holders in order to indentify STATIONS. Draft Report For Comment. MARTIN G.F.;

and reevaluate all safety-related piping systems which had been HICKEY,E.E.; MOELLER.M.P.; et al Batteile Memonal institute, analyzed with a computer program that used algebraic summa-Pacific Northwest Laboratones. March 1986.

124pp.

tion. Reevaluation was required to include estimating the intenm 8604040577. PNL-4758. 35410:018.

capability of the facility to withstand a seismic event, consider-j Scenano guidance handbook was prepared to assist emer*

ation of as-builts and planning and sched;hng reanalyses. Eval-gency planners in developing scenarios for emergency pre-uaton of utility responses and NRC documents shows that the paredness exercises at nuclear power plants. The handbook bulfetin can be closed out for til (90%) of the 124 current fa-provides guidarre for the development of the objectwns of an cilibes on the basis of specific cntena. Followup items for the exercise, the desenptions of scenano events and responses, remaining 13 facilities are proposed for use by NRC/IE. A final and the instructions to the partctprats. Information concerning check of licensees' actions will be made per later IE Bulletin 79-j implementabon of the scenano, entiques and findings, and gen

• 14 on seismic analyses for as-built safety-related piping sys-i eration and format of scenario data are also included. Finally, tems.

examples of manual calculational techniques for producing radi-ological data are included as an appendix.

NUREG/CR 3805 V03: ENGINEERING CHARACTERIZATION OF NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINA.

GROUND MOTION. Task 11: Observatonal Data On Spatial Vari-TIONS ON SOLIDIFICATION, WASTE DISPOSAL,AND ASSOCI.

atons Of Earthquake Ground Motion.

CHANG,C,Y.:

i ATED OCCUPATIONA*

EXPOSURE.

PICIULO,P.L; POWER.M.S.; IDRISS,1 M.; et al. Woodward-Clyde Consultants, i

ADAMS.J.W.; DAVIS.M.S. Brookhaven National Laboratory. Jan.

Inc. February 1986. 296pp. 8603180548. 35138:171.

uary 1986. 33pp. 8604010157. BNL-NUREG-51699. 35342:035.

This report presents the results of part of a two task study on Test results are reported for the degradation of simulated de-the engineering characterization of earthquake ground moton contaminaten wastes by wet-air oxidabon. The data indicate for nuclear power plant design. Task i of the study, which is that wet-air oxidabon can effectively degrade organc complex.

presented in NUREG/CR-3805, Vol.1, developed a basis for ing agents typically used in chemical decontaminations. Al-selecting design response spectra taking into account the char-though less than 90% of the available organic carbon was oxi-acteristics of free-field ground motion found to be signif' ant in c

dized, more than 95% of tne organic reagent was degraded.

causing structural damage. Task ll incorporates additional con.

Results are grven on the solidification test of NS-1 decontami-siderations of effects of spatial variations of ground motion and nation reagent in cement.

soil-structure interaction on foundabon motions and structural NUREG/CR-3517: RECOMMENDATIONS TO THE NRC ON response. The results of Task 11 are presented in four parts: (1)

HUMAN ENGINEERING GUIDELINES FOR NUCLEAR POWER effects of ground motion charactenstics on structural response j

PLANT MAINTAINABILITY. BADALAMENTE,R.; FECHT,B.A.;

of a typical PWR reactor building with localized nonlinearities BLAHNIK.D E.; et a!. Battelle Memorial Institute, Pacife North.

and soil-structure interaction effects; (2) observational data on west Laboratories. March 1986.144pp. 8604070246. PNL 48!5.

spatial variations of earthquake ground motion; (3) soil structure 35446:156.

interaction effects on structural response; and (4) summary Trds document contains human engineenng guidelines which based on Tasks I and il studies. This report presents the results

]

can enhance the maintainability of nuclear power plants. The of the second part of Task II.

I guidelines have been denved from general human engineenng design principles, criteria, sqd data. The guidelines may be ap.

NUREG/CR-3950 V02: FUEL PERFORMANCE ANNUAL plied to existing plants as well as to plants under construction.

REPORT FOR 1984. BAILEY,W.J. Battelle Memorial Institute, They appfy to nuclear power plant systems, quipment and fa.

Pacific Northwest Laboratones. DUNENFELO.M.S. NRC No cilibes. as well as to maintenance tools and equipment The Detailed Affiliation Given. March 1986.114pp. 8604040483.

guidehnes are grouped into seven categories: accessibility and PNL-5210. 35405:118.

workspace, physcal environment, loads and forces, mainte-This annual report, the seventh in a series. provides a brief nance facilities, maintenance tools and equipment, operating desenption of fuel performance during 1984 in commercial nu-equipment design, and information needs. Each chapter of the clear power plants. Brief summanes of fuel design changes, fuel document details specific maintainability problems encountered surveillance programs, fuel operating experience, fuel problems, i

at nuclear power plants, and the safety impact of these prob-high-burnup fuel expenence, and items of general significance i

lems, and the specific maintainability design guidelines whose are provided. References to additional, more detailed informa-l application can serve to avoid these problems in new existing tion and related NRC evaluations are included.

1 plants.

NUREG/CR 3958: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR 3760: A STUDY ON DUCTILE AND BRITTLE FAIL

• ON TRANSIENTS. ACCIDENTS AND CORE MELT FREQUEN-URE DESIGN CRITERIA FOR DUCTILE CAST IRON SPENT-CIES AT A COMBUSTION ENGINEERING PRESSURIZED i

FUEL SHIPPING CONTAINEP3. SCHWARTZ,RW. Lawrence WATER REACTOR. BICKFORD,W.E. Batteile Memonal Insti-j Livermore National Laboratory. January 1986. 58pp-tute, Pacific Northwest Laboratories. March 1986. 56pp.

8602200311 UCRL-53532. 34669:260.

8603240391 PNL-5767. 35202:276.

(

This report presents proposats for establishing design and ac-Pacific Northwest Laboratory (PNL) performed a probabilistic 4

ceptance entena for t!'c ductile cast iron to be used for fabricat-risk analysis to develop estimates of core-melt frequency and ing spent. fuel shipping casks. These proposals address desyn public risk associated with control system' failures in a Combus-

[

critena for presenting bnttle fracture, based upon drop test ng a tion Engineenng pressunzed water reactor. Valuehmpact anaty-flawed prototype cask' ses of ossible modificahons to prevent control system failures NUREG/CR 3790: CLOSEOUT OF IE BULLETIN 79-07. SEISMIC were also conruted. These analyses were based on a failure j

STRESS ANALYSIS OF SAFETY.RELATED PIPING.

modes and effects analysis previously conducted at Oak Ridge l

FOLEY,W.J.; DEAN.R.S.; HENNICK.A. Parameter, Inc. January Nabonal Laboratory. The control system failure modes fall into l

1986. 55pp. 86012902% IEB 79-07. 34419 212.

three main scenarios: two scenanos concern overfill of the IE Informabon Notce 79-06 was issued March 23,1979 to steam generators, progressing to spillover into the steam knes.

4 alert utility personnel that a certain computer program known to The third scenario deals with small. break loss-of-coolant acci-have been used for seismic stress anafysis of piping at five fa-dents that may require operator action to 'epressunze the reac-cilities had been found to provide nonconservative results Le-tor coolant system. The analyses desenbed in this report were cause of algebraic sumrt:11,on of loads. Aftc. further review of performed in support of the U S. Nuclear Regufatory Commes-j

)

8 Main Citation 3 cnd Abstracts sion's program for Unresolved Safety issue A-47, Safety impli-vated or deposited material, intemal dose commitment resulting catens of Control Systems.

from inhalabon, and total whole-body doses. Extemal doses NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR from akbune maMnal are calculated using senfinite and ENVIRONMENT -lMPLICATIONS FOR NRC OUALITY ASSUR-fmite cloud approximatong At each stage in model execution, ANCE PROGRAMS BASED ON NUCLEAR POWER INDUSTRY the appropriate approximation is selected after considenng the AND REGULATORY PROJECTIONS THROUGH 1995.

cloud dimensions. Atrr'osphenc processes are represented in CHRISTENSEN.J.; SCHULLER,C.R.; HARTY,H.; et al. Battelle MEDSORAD by a combinaten of Lagrangian puff and Gaussian Memorial Institute, Pacific Northwest Laboratories. March 1986.

plume dispersion models, a source depletion (depositen veloci-109pp. 8604070361. PNL-5769. 35446:310.

ty) dry deposition model, and wet deposition model using wash-This report develops projechons for nuclear power plant regu-out coefficients based on precipitaten rates, fatory needs in general, and those relating to quality assurance NUREG/CR-4059: EVALUATION OF THE IMPACT OF THE in parbcular, for the time period 1985 to 1995. This required an MC&A REFORM AMENDMENTS ON A REPROCESSING FA-assessment of future prospects for the nuclear power industry CfLITY. EHINGER.M.H.; KERR.H.T.; HEBBLE.T.L; et al. Oak -

and its primary segments. Electric power demand projections and their relatonship to estimated schedules for nuclear plant Ridge National Laboratory. January 1986. 92pp. 8602050511.

ORNL/TM-9719. 34484:352.

constructen and operations were evafuated, and estimates of anticipated butness volume and long term economic viabihty An assessment was completed on the potential for large re-were made for each of the major segments of the U.S. nuclear processing plants to meet the requirements of the Nuclear, Reg-industry (utilibes, NSSS vendors, AEs, constructors, component ufatory Commission's proposed Category i Matenal Contro. and supphers and service vendors). These estimates were made for Accounting (MC8A) Reform Amendment. Tne requirements on two, five, and ten year intervals through 1995. Other significant which this assessment was based are gived in the working draft factors that are not specific to any one industry segment were revision to the rule dated December 30,1982. The Barnwell Nu-also reviewed. These included: 1) the exparding foreign pres.

char Fuel Plant (BNFP) was chosen as a reference design for ence in U.S. markets; 2) pending legislation; 3) trends in per.

the assessment, but most considerations would be relevant to sonnel availabihty; 4) new institutenal arrangments for nuclear any large Purex reprocessing facihty. Spent hght water reactor power generabon; 5) nuclear plant aging, hfe extension, and de.

(LWR) fuels containing 1% Pu were the g*esumed feed to the commissioning; 6) reactvation of mothballed projects; 7) ad.

plant; the design feed rate is 5 MTU/d. The approach taken for vanced and standardized plant designs; and 8) hkely technologi, the assessment was to characterize the vocess equipment and cal developments in computer apphcations and inspection meth.

the nuclear material distnbuten throughout the plant, to identify j

ods. The trends revealed by these analyses imply a number of quantabes of material snat must be removed consistent with significant challenges for nuclear power regulation in the U.S.

loss-detecten capabihties were based oit these test results. No Many projected changes have imphcations for NRC OA pohcies atternpt was made to construct detailed removal scenanos or and practices. These issues were broken into functional ele.

integrated MC&A systeras throughout the plant. The assess-ments each of which was analyzed in terms of its significance ment address three general types of material removals or and kinetics of emergency for each of the overall industry pro-losses: 1, sirigle space, single time (abrupt) 2. multiple space, jectons. Finally, NRC optons for dealing with each OA related single time (abrupt with coffatoration), and 3. single space, mul-issue were assessed.

tiple time (recurnng). With few exceptons, the abrupt loss <le-NUREG/CR-3983: STEAM EXPLOSION EXPERIMENTS AT IN.

tecton r@ments of the Nam Amndment wm M acha-TERMEDIATE SCALE:FITSB SERIES.

MITCHELL.D.E..

ble with existing or shghtly improved capabihbes. Some equip.

ment gns a@ measweant kchnology Wments EVANS.NA Sandia Natonal Laboratones. February 1986}

90pp. 8604010177. SAND 83-1057. 35342:096.

will be needed. Recurring loss-detection capabihties will be Expenments in the FITS chamber have been performed in somewhat poorer than capabihbes for abrupt loss detecdon.

which 13.7 kg of molten iron-alumina core melt simulant was NUREG/CR-4078: PROGRAM FOR FIELD VAUDATION OF THE dehvered into water chambers in which ths water mass was 1.5 SYNTHETIC APERTURE FOCUSING TECHNIQUE FOR UL-to 15 times greater than the melt. Experiments in subcooled TRASONIC TESTING (SAFT UT). Final Report. HAMUN,0.h water showed that spontaneous explosons occt 'ed over the Southwest Research institute. November 1985. 165pp.

range of water / melt mass ratio and geometry used and that in 8602130528. 17-5023. 34602:201' certain expenments, double explosions occurred. With double explosions, the first explosion enchanced fuel-coolant mixing for The purpose of the project was to vahdate the effectiveness the second explosen. In one test in saturated water, multiple of the Synthetic Aperture Focusing Technique for Ultrasonic trigger sites were observed but no propagating explosion result.

Testing (SAFT UT) as a nondestructive examinaton technique.

ed. Two distinct, but additrve, energy converson rates were cal.

SAFT UT is an ultrasonic imaging method for accurate rneas.

culated from the test results. Based on pressure records and wment of me spahal locaten and extem of scashcaW reh debns velocites, a kinetic energy converson ratio, nKE, had M swfaces such as Haws contamed M canponents of Nclear calculated values between 0.3 and 1.6 percent. A conversion power reactor systems. The increased measurermnt accuracy ratio, nD, related to the work done in pressurizing the chamber offered by SAFT, when compared with that of methods now in air ranged between 0.2 percent and 8.6 percent. The total frac-use, will improve the reliability of flaw seventy assessment. This bon of the melt thermal energy converted, ntot = nKE + nD, rep rt presents a comprehensive discussion of the work accom-reached a value of 9.9 percent in an experiment involving a phshed in evaluatng the performance capabilities of the devel-double explosion, but in this case, the value of nKE was limited oped SAFT UT inspecten system. Inspecten results obta,ned to 1.3 percent.

using both 0 degree togttudinal and angle-beam operating modes are presented. These results include laboratory and nu.

NUREG/CR-4000 V01: THE MESORAD DOSE ASSESSMENT clear power plant field site examinaten on a variety of defect MODEL. Volume 1: Technical Basis.

SCHEAPELZ,R1; types contained within carbon and stainless steel flat plate and BANDER.T.J.; RAMSDELLJ.V.: et al. Batteile Memonal inst-cyhndncal test specimens or components. The SAFT UT flaw tute, Pacific Northwest Laboratones. March 1986. 99pp.

images are evatuated by companng them to results obtained 8004040564. 35409:226.

from destructive sectioning or by using flaw fabrication data MESORAD is a dose assessment model for emergency re-which predicted actual flaw depth, onentation and size. On the sponse applicatons. Using release data for as many as 50 ra-basis of these evaluations, conclustns are presented which I

dionuchdes, the model calculates: external doses resulting from summarize the performance capabilities of the SAFT UT inspec-I exposure to radiaton emetted by ra$onuclides cor tained in ele-ton techr%ue.

Main Citations and Abstracts 9

NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM - PHASE actors during accident and off-normal conditions. The TRAC ILSemiannual Report, April 1985 - September 1985.

code is being assessed at Sandia against fest data from various WILKOWSKI,G.M.; AHMAD,J.; BARNES,C.R.; et al. Battelle Me-integral and separate effects test facilities. As part of this as-monal Institute, Columbus Laboratories. March 1986. 175pp.

sessment matnx, a large-break transient performed at the LOBI 8604040064. BMI-2120. 35408:010.

facility has been analyzed. Our results show that TRAC-PF1/

The efforts in this report are broken into seven work pack.

MODI correctfy predicts the major phenomena occurring during ages related to pipe-fracture research efforts and two work a large break accident. Subcooled and saturated discharge co-packages that are supporting research efforts. The pipe-fracture efficients both equal to 1.0 give good agreement with data for efforts involve only circumferential crack orientations. Thirty-five break flows and primary system depressurization. Accumulator pipe expenments have been conducted to date, with all but two flow is calculated to begin within 1 s of the observed time, and at 550 fahrenheit (288 centigrade). Approximately 42 additional the predicted accumulator injection is generalty within 5 of the pipe experiments from other programs were also analyzed. In measured value. Both the peak clad temperature (788 K) and the analysis effort, a screening enterion was developed to show the overall rod temperatures are in acceptable agreement with when the net.section-collapse analysis is valid. This shows that data (823 K PCT). Sens.tivity studies were done on the cold leg even wrought stainless steel can fail at less than net-section-nodalization, the magnitude of the core bypass flow, the saturat-collapse loads if the pipe diameter is sufficiently large. Numer-ed break flow discharge coefficient and the accumulator surge ous predictive J-estimation schemes have been evaluated and line resistance. A coding error was discovered which resulted in modified. A finite length surface cracked pipe estimation the calculation of substantial and prolonged liquid superheat in scheme has also been developed. Finite-element analyses of the core; correcting this error did not, however, significantly specimens with welds suggest that the size of tf e weld relatrve change any of the global behavior calculated. These TRAC-to the specimen of structure size can affect the deformation J PF1/ MOD 1 results are generally comparable to our previous values. Supporting research efforts involve geometry effects on RELAP5/ MOD 1 results for this same LOBI test. Similar break J-R curves, as well as characterizing the material properties for flow and pnmary pressure behavior were calculated, but for dif-each pipe tested.

ferent discharge coefficiersts; similar accumulator injection was NUREG/CR-4113: FLOW AND DISPERSION NEAR CLUSTERS calculated, but for different user-input surge line resistance. The OF BUILDINGS. HOSKER,R.P.; PENDERGRASS,W. Com-PCT predicted by RELAPS (820 K) was in much better agree-merce, Dept. of National Oceanic & Atmospheric Administra-ment with data, but the late-time rod temperatures predicted by tion. March 1986.110pp. 8604030043. 35392:247.

TRAC are in better agreement with data than those from This report is intended as an information summary and, to a limited extent, as an interim and rather qualitative guide for NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVAL-those who must routinefy face the complex air quality problems UATION OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER associated with atmospheric flow and effluent dispersion near PLANT. BALL.D.G.; CHEVERTON,R.D.; FLANAGAN,G.F.; et al.

clusters of buildings. A bnef summary of the flow pattems ex.

Oak Ridge National Laboratory. September 1985. 611pp.

pected near isolated simple buildings serves as an introduction.

8601280327. ORNL/TM-9567/V1. 34388:248.

Flow patterns associated with varying densities of uniformly An evaluation of the risk of pressunzed thermal shock (PTS) sized buildings in an extensive array are then discussed. Previ-resulting in a through-the-wall crack in a reactor pressure vessel ous work on flow near simple, isolated building clusters is re-was performed for the H.B. Robinson Unit 2 nuclear power viewed, along with the concept (Beranek. 1979,1984) of a plant. The information presented in this report covers one of region of influence". A systematic study of near-ground-level three plant-specific studies performed for NRC. The other two flow patterns within simple building clusters is summarized. The studies, for Oconee Unit 1 and Calvert Cliffs Unit 1, are docu-suite of presently available models for estimating near-building mented in NUREG/CR-3770 and NUREG/CR-4022, respective-and wake concentrations is bnefly described.

ly The specific objectives of the H.B. Robinson study were (1)

NUREG/CR-4132: NUCLEAR POWER SAFETY REPORTING to further refine the methodology for evaluating the tisk of PTS, SYSTEM FINAL EVALUATION RESULTS. FINLAYSON,F.C.;

(2) to provide a best estimate of the frequency of a through-the-NEWTON,R.D. Aerospace Corp. February 1986. 106pp.

wall crack for the H.B. Robinson Unit 2 vessel, (3) to determine 8603180510. 35139:137.

the dominant PTS sequences for the unit, and (4) to evaluate This report is the second in a senes of three investigating the the effectiveness of potential corrective measures. The exami-feasibility of adapting a voluntary, anonymous, non-punitive, nation of tens of thousands of transients indicated that PTS was third party managed reporting system in a U.S. commercial nu.

not an imoortant core melt initiator for H.B. Robinson Unit 2.

clear utility industry environment. Such a system is intended for The dominant nsk sequences were determined to be failure of use in identifying and quantifying, in an uninhibiting manner, fac-steam relief vatves to close on multiple steam lines.

tors that contnbute to the occurrence of significant safety inci-dents which elicit either positive or negative responses from NUREG/CR-4t83 V02: PRESSURIZED THERMAL SHOCK EVAL-equipment and humans in nuclear power plants. This report pre.

UATION OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER sents the iesults of a limited evaluation of such a system, for its PLANT. BALL.D.G.; CHEVERTON,R.D.; FLANAGAN,G F.; et al.

practicality, acceptability and usefulness in a commercial nucle.

Oak Ridge National Laboratory. September 1985. 387pp.

g ar environment. Conclusions and recommendations resulting 8601280352. ORNL/TM-9567/V2. 34399.046.

from the limited evaluation are also presented. A companion See NUREG/CR-4183,V01 abstract.

report (NUREG/CR-4133, " Nuclear Power Safety Reporting NUREG/CR-4188 V01: NUCLEAR POWER PLANT SIMULAT'JN System: Implementation and Operational Specifications,') pre-FACILITY EVALUATION METHODOLOGY.Hardoook Per n

I d procedures required t est bls p9 a

'sy LAUGHERY,K.R.; CARTER,R.J.; HAAS.P.M. Oak Ridp Nationai Laboratory. January 1986. 175pp. 8602260 %. ORNL/TM-NUREG/CR-4171:

TRAC-PF1/ MOD 1 INDEPENDENT 9570/V1. 34704:120.

ASSESSMENT: LOBI LARGE BREAK TRANSIENT A1-04R.

This report is Volume 1 of a two-part document which de-KMETYK LN. Sandia National Laboratories. December 1985.

scribes a project conducted to develop a methodology to evalu-139pp. 8601280334. SAND 85-0442. 34388:115.

ate the acceptability of nuclear power plant (NPP) simulation fa-The TRAC-PF1/ MODI independent assessment project at cilities for use in the simulator-based portion of NRC's operator Sandia National Laboratories is part of an overall effort funded licensing examination. The propcsed methodology is to be uti-by the NRC to determine the ability of vanous system codes to lized dunng two phases of the simulation facility 1fe-cycle, initial predict the detailed thermal / hydraulic response of light water re-simulator acceptance and recurrent analysis. The first phase is

10 Main Citations and Abstracts aimed at ensuring that the simulator provides an accurate repre-loadings. An axisymmetric geometnc model :s used and the cir-sentation of the reference NPP. There are two components of cumferential variation of the loads is expressed in terms of Fou-initial simulator evaluation: fidelity assessment and a direct de-rier series components. The solution, i.e., disp!acement and termination of the simulation facihty's adequacy for operator stress resultants at any point of time, are obtained and also rep-testing. The second phase is aimed at ensunng that the simula-resented by Fourier series components. Values of these compo-tion facility continues to accuratefy represent the reference nents at discrete locations are evaluated using modal superpo-plant throughout the hfe of the simulator. Recurrent evaluation sition. Failure, i.e., leakage of the containment, is defined to is comprised of three components: monitoring reference plant occur when either the membrane strain in the shell reaches changes. monitoring the simulator's hardware, and examining twice the shell matenal yield strain or a bifuraction point occurs.

the data from actual plant transients as they occur. Volume 1 is The ASME Boiler and Pressure Vessel Codes Level C criteria a set of guidelines which details the steps involved in the two can also be applied as a failure condition. A computer program life-cycle phases, presents an overview of the methodology and called LOADS was wntten to evaluate the point-in-time displace-data collection requirements, and addresses the formation of ments and stress resultants. This program uses the results ob-the evaluation team and the preparation of the evaluation plan.

tained from a modified BOSOR4 modal analysis program. The Volume 2 desenbes the technical bases and the development results of the LOADS program are input into a modified of the methodology.

BOSORS computer code to evaluate the containment resist-ance which considers large deformation and material nonlinear NUREG/CR-4188 V02: NUCLEAR POWER PLANT SIMULATION behavior. A user's manual for the BOSOR4/ LOADS /BOSORS FACILITY EVALUATION METHODOLOGY. Technical Bases.

program set is presented.

LAUGHERY,K.R.; CARTER,R.J. Oak Ridge National Laboratory.

January 1986. 89pp. 8602270413. ORNL/TM-9570/V2.

NUREG/CR-4255 V02: AEROSOL RELEASE AND TRANSPORT 34715:176.

PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-This report is Volume 2 of a two-part document which de-SEPTEMBER 1985. ADAMS.R.E.; TOBIAS,M.L Oak Ridge Na-scribes a project conducted to develop a methodology to evalu-tional Laboratory. December 1995. 40pp. 8602270507. ORNL/

ate the acceptability of nuclear power plant (NPP) simulation fa.

TM-9632/V2. 34715:135.

cilities for use in the simulator-based portion of NRC's operator This report summarizes progress for the Aerosol Release and licensing examination. The proposed methodology is to be uti-Transport Program sponsored by the Nuclear Regulatory Com-lized dunng two phases of the simulation facility life-cycle, initial mission, Office of Nuclear Regulatory Research, Division of Ac-si.nulator acceptance and recurrent analysis. The first phase is cident Evaluation, for the period April 1985-September 1985.

aimed at ensuring that the simulator provides an accurate repre-Topics discussed include (1) a steam-only test performed in the sentation of the reference NPP. There are two components of NSPP vessel; (2) development tests to study thermal ir put and initial simulator evaluation: fidelity assessment and a direct de-generation ef'iciency; (3) Aerosol-Moisture Interaction Test termination of the simulation facility's adequacy for operator (AMIT) preliminary and development tests to check various fea-testing. The second phase is aimed at ensunng that the simula-tures of the AMIT facility; (4) data from the first two tests in the tion facility continues to accurately represent the reference AMIT programs; (5) an analysis of changes necessary in Ander-

)lant throughout the life of the simulator. Recurrent evaluation sen Mark-Ill impactor design for AMIT experiments; (6) the is compnsed of three components

monitonng reference plant theory of capillary condensation on aerosols at nominally under-changes, monitoring the simulator's hardware, and examining saturated humidity levels; (7) work in modifying data processing the data from actual plant transients as they occur. Volume 2 codes to accommodate data retrieval equipment changes; (8) descr;bes the development of and technical bases for the eval-correction of sampb volume calculations for NSPP experiments uation methodology, including a discussion of the major issues, on aerosols in steam-air environments; and (9) implementation research base, and judgements/ decisions made by the re-and application of the CONTAIN code.

search team.

NUREG/CR-4219 V02: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR-4276 V02: VIBR,ATION AND WEAR IN STEAM GEN-PROGRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-ERATOR TUBES FCLLOWING CHEMICAL CLEANING.

SEPTEMBER 1985. PUGH.C.E. Oak Ridge National Laboratory.

ENDERLIN,W.I.; FITZSIMMONS D. Battelle Memorial institute, January 1986. -229pp. 8602240066. ORNL/TM-9593/V2.

Pacific Northwest Laboratories. March 1986.52pp.8603270379.

34684:134.

PNL-5477 V02. 35311:346.

The Heavy-Section Steel Technology (HSST) Program is an Chemical cleaning has been proposed to remove magnetite engineering research activity conducted by the Oak Ridge Na, buildup in some pressurized water reactor steam generators.

tional Laboratory for the Nuclear Regulatory Commission. The The U.S. Nuclear Regulatory Commission (NRC) has expressed program comprises studies related to a!I areas of the technolo-concern that such cleaning would combine with the tube dent-gy of matenals fabncated into thick-section pnmary-coolant con.

ing caused by magnetite formaton to enlarge mbe/tuwsupport tainment systems of I;ght-water-cooled nuclear power reactors.

plate clearances, increasing the level of flow-duced vibrations in The investigation focuses on the behavior and structural integri-that could lead to unacceptably high tube wear and failure ty of steel pressure vessels containing cracklike flaws. Current rates. In support of NRC, the Pacific Northwest Laboratory in-work is organized into ten tasks: (1) program management, (2) vestigated whether such increased clearances would exacer-fracture-methodology and analysis, (3) material characterization bate tube fretting wear. Using a full-length scale model of a and properties, (4) environmentally assisted crack growth stud-steam generator tube bundle, flow tests were conducted at an ies, (5) crack arrest technology (6) irradiation effects studies, instrumented location through clearances representing as-built (7) cladding evaluations, (8) intermediate vessel tests and anal-and post-cleaned tube conditions. Test results indicated little ysis, (9) thermal-shock technology, and (10) pressurized ther-potential for increased tube wear as a result of chemical clean-mal-shock technology.

ing, under normal operating conditions at tube support locations similar to that tested.

NUREG/CR-4223: STEEL CONTAINMENT RESISTANCE UNDER GENERAL DYNAMIC PRESSURES.

GREIMANN.L; NUREG/CR-4279 V01: AGING AND SERVICE WEAR OF HY-FANOUS,F.; BLUHM,D. Ames Laboratory, Energy & Mineral Re-DRAULIC AND MECHANICAL SNUBBERS USED ON SAFETY-sources Research Institute. February 1986.142pp.8603030055.

RELATED PIPING AND COMPONENTS OF NUCLEAR POWER IS-4554. 34774:148.

PLANTS. Phase I Study.

BUSH.S.H.;

HEASLER,P.G.;

The objective of this work is to develop a methodology for DODGE R.E. Battelle Memorial Institute, Pacific Northwest Lab-calculating the ultimate Capacity of steel containments under oratories. February 1986. 142pp. 8603100569. PNL 5479.

very general global or local, uniform or spatially varying dynamic 34891:149.

I Main Citations and Abstracts 11 This report presents an overview of hydraulic and mechanical (94)Nb, (99)Tc, (129)l, (137)Cs, and alpha-emitting transuranic snubbers used on nuclear piping systems and components, radionuclides with half-lives greater than five years. The total re-based on information from the literature and other sources. The sidual radionuclide inventories (excluding the pressure vessel) functions and functional requirements of snubbers are dis-at the seven nucteer power plants examined in this study cussed. The real versus perceived need for snubbers is re-appear to be proportional to the product of the unit power level viewed, based primanly on studies conducted by a Pressure (megawattage) and the length of operations in years. Thus, ex-Vessel Research Committee. Tests conducted to q.slify snub-trapolations of the radionuclide inventories at other nuclear bers, to accept them on a case-by-case basis, and u establish plants may perhaps be made. For the most part, the radionu-their fitness for continued operation are reviewed. This report chde compositions and inventories measured in this program had two primary purposes. The first was to assess the effects were in reasonably good agreement with the limited data base of vrious aging mechanisms on snubber operation The second used in the earlier conceptual assessment studies of the tech-was to determine the efficacy of existing tests in determining nology, safety, and costs of decommissioning a reference PWR the effects of aging and degradation mechanisms. These tests (NUREG/CR-0130) and a reference BWR (NUREG/CR-0672).

include breakaway force, drag force, velocity / acceleration range Thus, the conclusions reached in these conceptual studies from for activation in tension or compression, release rates within a radiological standpoint will essentially remain unchanged.

specified tension / compression limits, and restricted thermal movement. The snubber operating experience was reviewed NUREG/CR-4293: RELIABILITY ANALYSIS OF SHEAR WALL using licensee event reports and other historical data for a STRUCTURES. WANG,P.C.; HWANG,H.; PIRES,J.; et al. Brook-period of more than 10 years. Data were statistically analyzed haven National Laboratory. January 1986. 39pp. 8603030319.

using artxtrary snubber populations. Value-impact was consid-BNL-NUREG-51900. 34779:170.

ered in terms of exposure to a radioactive environment for ex-This report descnbes a method for the assessment of the reli-aminabon/ testing and the influence of lost snubber function and ability of low-rise shear wall structures, which are often used in subsequent testing program expansion on the costs and oper-nuclear power plants. The shear walls are modeled by stick ation of a nuclear power plant. The implications of the observed models with bearn elements, and are subjected to dead load, trends were assessed; recommendations include modifying or live load and earthquake during their lifetimes. The earthquake improving examination and testing procedures to enhance snub-load is assumed to be a segment of a stationary Gaussian proc-ber reliability. Optimization of snubber populations by selective ess with a zero-mean and a Kanai-Tajimi power spectral density removal of unnec%sary snubbers was also considered.

function. The seisms hazard at a site, represented by a hazard NUREG/CR-4286: EVALUATION OF RADIOACTIVE LIQUID EF-curve, is also included in the reliability analysis. Both shear and FLUENT RELEASES FROM RANCHO SECO NUCLEAR flexure limit states are analybcally defined. The flexure limit POWER PLANT. MILLER C.W.; COTTRELL,W.D.; LOAR,J.M.; et state is defined according to the ACI strength design formula, al. Oak Ridge National Laboratory. March 1986. 150pp.

while the shear limit state is established from test data. The reh-8604030073. ORNL/TM-6183. 35392:359.

ability analysis methodology is desenbed in detail. Illustrative ex-A protect has been carried out by Oak Ridge National Labora, amples are given to demonstrate the method and the applica-tory (ORNL) to estimate the concentrations of radionuclides in tions. This reliability method can also be used to generate the the environment that have resulted from the release of radioac.

fragility curve of the shear wall structure.

tive matenals in the liquid waste effluents from the Ranch NUREG/CR-4299: PRELIMINARY EVALUATION OF EFFLUENT Seco Nuclear Power Plant (RSNPP) and to estimate possible RADIOACTIVITY MONITORING SYSTEMS FOR BWR PLANTS.

radiation doses to man resulting from current environmental EDSEN inc. (subs. of EG&G concentrations. To carry out the objectives of this project, two ARAVE'A.E/'y 1986. 3'J.L EGaG Idaho' EGG-2401. 35342:001' Inc.). Januar 5pp. 8604010145.

visits were made to the RSNPP site by scientists from ORNL dunng November and December 1984 to conduct an environ-The need for upgrading the effluent monitoring systems at mental sampling program around the site. Elevated levels of commerical nuclear power plants was recognized following the some radionuclides were found in the immediate environment of TMI-2 accident in 1979 (NUREG-0737 Ctanfcation of TMI the plant. This radioactive contamination occurs pnmarily along Action Plan Requirements). Improvements have been made to streams receiving effluent from the plant and in fields irrigated these systems since then, but not all problems deahng with the with water from these streams. The primary contaminants are measurement of radioactive releases during severe accident 137C(s) and 134C(s) with lesser amounts of 60Co and 58Co.

scenarios have been addressed. This report discusses some of Specific pathways of exposure and usage factors were not pre-the generic issues associated with the transport and subse-cisely known for the dose assessment of current and potential quent sampling of noble gases, particulates, and iodine species use of contaminated water and soil around the RSNPP. The in-that utilities must consider to ensure accurate reporting during gestion of fish is the single most important pathway identified in severe accident conditions. In light of these generic concems, a this analysis.

specific postaccident upgrade is discussed, major measurement uncertainties are identified, and recommendations are made.

NUREG/CR-4289: RESIDUAL RADIONUCLIDE CONTAMINA-The focus of these recommendations is the transport behavior TION WITHIN AND AROUND COMMERCIAL NUCLEAR of iodine; sample-line losses may result in an order of magli-POWER PLANTS, ORIGIN, DISTRIBUTION, INVENTORY AND tude error for near-real-time measurements. Finally, a recom-DECOMMISSIONING ASSESSMENT.

ABEL,K.H.;

mendation for a laboratory sample-line test program is made.

ROBERTSON,D.E.; THOMAS,C.W.; et al. Battelle Memorial in-The laboratory effort would better define the uncertainty of the stitute, Pacific Northwest Laboratories. February 1986. 261pp.

commercial measurements and also provide data for the im-8603140488. PNL-5429. 34965:027.

provement of line-loss algonthms.

The residual radionuclide concentrations, distributions and in-ventones at seven nuclear power plants (four shutdown and NUREG/CR-4300 V02: ACOUSTIC EMISSION / FLAW RELATION-three operating) have been investigated to provide a data base SHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRES-for use in formulating policies, strategies and guidelines for the SURE VESSELS. Progress Report, April-September 1985.

j eventual decommissioning of retired nuclear power plants. This HUTTON,P.H.; KURTZ,R.J. Battelle Memorial Institute, Pacific study has addressed radionuclides (both activation and fission Northwest Laboratories. January 1986. 18pp. 8601290191.

products) transported from the reactor pressure vessel and de-PNL-5511. 34419:194.

posited in all other contaminated systems of each nuclear plant.

This report discusses technical progress for the period April Emphasis has been placed on measuring the long-lived radionu.

1985 to September 1985 under the NRC-sponsored research clides which are of special concern from a low-level waste man-program concemed with " Acoustic Emission / Flaw Relationships agement standpoint, including (60)Co, (59)Ni, (63)Ni, (90)Sr, for Inservice Monitoring of Nuclear Reactor Pressure Bound-l 1

12 Main Citations and Abstracts anes." Topics discussed include testing AE monitoring on oper-NUREG/CR-4311: REVIEW OF THE SHEARON HARRIS UNIT 1 ating reactors, refinement of an AE signal identification relation.

AUXILIARY FEEDWATER SYSTEM RELIABILITY ANALYSIS.

ship, developing AE/lGSCC relationships, evaluating the effects FRESCO,A.; YOUNGBLOOD,R.; PAPAZOGLOU,1 A. Brookha-of slow crack growth rate on AE generation, and activities to ven National Laboratory. February 1986. 70pp. 8603100084.

produce an AE monitoring standard and acceptance of the BNL NUREG-51902. 34890:127.

technology by the ASME code.

This report presents the results of a review of the Auxiliary Feedwater System Reliability Analysis for the Shearon Hams NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK Nuclear Power Plant (SHNPP) Unit 1. The objective of this VALVES USED IN ENGINEERED SAFETY-FEATURE SYS.

report is to estimate the probability that the Auxiliary Feedwater TEMS OF NUCLEAR POVJER PLANTS. GREENSTREET,W.;

System will fail to perform its mission for each of three different MURPHY,G.A.; GALLAHER R.B.; et al. Oak Ridge National Lab.

initiators: (1) loss of main feedwater with offsite power available, oratory. December 1985. 79pp. 8601280340. ORNL-6193/V1.

(2) loss of effsite power (3) loss of all ac power except vital 34387:243 instrumentation and control 125-V dc/120-V ac power. The This is the first in a senes of three reports on check valves scope, meMogy, and faHum data am p sen M W (CVs) to be produced under the U.S. Nuclear Regulatory Com-NUREG-0611, Appendix IIL The results are compared with mission's Nuclear Plant Aging Research program. This program those obtained in NUREG-0611 for other Westinghouse plants.

addresses the evaluation and identification of practical and NUREG/CR-4323: THE PROTECTION OF URANIUM TAILINGS cost-effective methods for detecting. moratonng, and assessing IMPOUNDMENTS AGAINST OVERLAND EROSION.

the seventy of time-dependent degradation (aging and se'rvice WALTER,W.H.; SKAGGS,R.L Battelle Memorial Institute, Pacific wear) of CVs in nuclear plants. These methods are to provide Northwest Laboratones. January 1986. 61pp. 8601210636.

capabilities for establishing degradation trends prior to fMure PNL-5520. 34318:160.

and developing guidance for effective maintenance. This report This study investigates the problems involved in designing examines failure modes and causes resulting from aging and protection methods to prevent erosion of a uranium tailings im-service wear, manufacturer-recommended maintenance and poundment cover from rainfall and runoff (overland flow) proc-surveillance practices, and measurable parameters (including esses. The study addresses the side slopes and top surface as functional indicators) for use in assessing operational readiness, separate elements. The side slopes are more subject to gully establishing degradation trends, and detecting incipient failure.

erosion and require absolute protection such as that provided The results presented are based on information derived from by rock riprap. The flatter top surface needs much less protec-operating expenence records, nuclear industry reports, manu-tion (vegetation / rock combinations) but some estimate of ero-facturer-supplied information, and input from plant operators.

sion rates are needed to compare attematives. A literature review indicated that, currently, procedures are not available for NUREG/CR 4307 Vot: LWR PRESSURE VESSEL SURVEIL-the design of rock riprap to prevent gully erosion. Therefore, LANCE DOSIMETRY IMPROVEMENT PROGRAM. Progress rock protection on the side slope will have to be based upon Report - October 1984 - September 1985. MOELROY,W.N.:

engineenng judgment determined by the particular site condi.

LIPPINCOTT.E.P. Hanford Engineenng Development Laborato-tions. The Mannning-kinematic equations (velocity and depth of ry. January 1986. 273pp. 8602280759. HEDL-TME 8514.

runoff) were nvestigated as a possible aid to the design of gully 34754:254~

erosion protection. Guidelines are suggested for the use of rock This report desenbes progress made in the Light Water Reac-riprap to prevent gully erosion. Three mathematical models tor Pressure Vessel Surveillance Dosimetry improvement Pro-were used to compute erosion rates for the top surface of a phypothetical tailings impoundment. The results recommend gram (LWR.PV-SDLP) dunng FY84. The primary concem of this that one or possibly both of the regression models could be program is to improve, test, verify, and standardize the physics-used to evaluate preliminary protection design for the top sur-dosimetry. metallurgy and associated reactor and damage analy-face. A physical process simulation modet should be used for sis procedures and data used for predicting the integrated ef-the final design.

fects of neutron exposure to LWR-PVs and their support struc-tures. These procedures and data are being recommended in a NUREG/CR-4324: TESTING OF NUCLEAR QUALIFIED CABLES new and updated set of ASTM standards being prepared l AND PRESSURE TRANSMITTERS IN SIMULATED HYDRO-tested, and venfied by program participants. These standards GEN DEFLAGRATIONS TO DETERMINE SURVIVAL MARGINS together with parts of the US Code of Federal Regulations and AND SENSITIVITIES. DANDINI,V.J. Sandia National Laborato-ASME codes, are needed and used for the assessment and ries. December 1985. 105pp. 8602270391. SAND 85-1481.

control of the condition of LWR-PVs and their support structures 4

during the 30- to 60-year lifetime of a nuclear power plant.

of tests was conducted at Sandia National Laborato-ries Central Receiver Test Facility that addressed the margin for NUREG/CR-4310: INVESTIGATION OF POTENTIAL FIRE RE-survival of nuclear reactor safety equipment in hydrogen bums.

LATED DAMAGE TO SAFETY-RELATED EQUIPMENT IN NU-The tests exposed aged and unaged specimens of nuclear CLEAR POWER PLANTS. WANLESS.J. NUS Corp.

qualified electrical cable and pressure transmitters to simulated WANLESS,J. Sandia National Laboratories. November 1985.

hydrogen burns of increasing seventy. Starting with a base heat 66pp. 8603280269. SAND 85-7247. 35324:191.

flux pulse corresponding to that resulting from a 75 percent Based on a review of vendor information, fire damage reports, cm MahaW mam in a W % mntahn( N Mat equipment qualification and hydrogen bum test results, and ma-flux levels of each successive pulse increased in increments of tenal properties, thirty.three types of equipment found in nuclear 50 percent of the base pulse. The heat flux levels of the most severe pulse were 300 percent of the base pulse. All specimens power plants were ranked in terms of the? potential sensitmty functioned property during exposure to all of the simulated hy.

to fire environments. The ranking considered both the functional drogen bums. Cables maintained a constant applied potential; requirements and damage proneness of each component. A fur-no shorts or open circuits were detected. The pressure transmit-ther review of the seven top ranked components was per-ters delivered steady signals indicating no significant degrada-formed, considering the relative prevalence and potential safety tion in performance. Post exposure tests indicated slight differ-signincance of each. From this, relays and hand switches domi-ences between the aged and unaged cable insulation character-nate as first choices for fire damage testing with logic equip-istics. The insulation of one cable failed dunng the 2400 Vac ment, power supplies, transmitters, and motor control centers high potential withstand test. Transmitter calibration checks per-as future candidates.

formed between exposures and after completion of the test

I Main Citations and Abstracts 13 series indicated slight changes in cahbration for both the aged to predicting counter-current flow in multiple channels due to a and unaged transmitter.

pump-induced instability in the tests. However, TRAC calcula-NUREG/CR-4327: ORGANIC IODIDE FORMATION FOLLOWING tions were performed with a three-channel input model and a NUCLEAR REACTOR ACCIDENTS.

B EAHM.E.C.;

single equivalent-channel model in which the three channels SHOCKLEY,W.E.; CULBERSON.O.L Oak Ridge National Labo-were lumped together as one. These calculations demonstrated J

ratory. November 1985. 45pp. 8603100038. ORNL/TM-9627.

that using the equivalent-channel model results in about the 34872:048.

same response as using the three channel model with respect A wide vanety of organic matenals will be present in contain-to differential pressure and channel mass flow rate.

ment following a light-water reactor accident. Organic iodides NUREG/CR-4343: INTEGRATED SEVERE ACCIDENT CONTAIN-can be produced by the reaction of fission product iodine with these organic materiais. This report emphasizes the importance MENT ANALYSIS WITH THE CONTAIN COMPUTER CODE.

of free radicals and radiation in leading to the formation of or.

BERGERON,KD.; WILLIAMS,D.C.; REXROTH,P.E.; et al.

ganic iodide. The ultimate aim is a description of the rate of for.

Sandia National Laboratones. December 1985. 129pp.

mation and removal of organic iodide, within the time span of a 8603280255. SAND 85-1639. 35325:340.

core heatup event, that can be used in a model of iodine be-Analysis of physical and radiological conditions inside the havior in containment.

containment building during a severo (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly dis-NUREG/CR-4328: PROBABluTY BASED LOAD COMBINATION CRITERIA FOR DESIGN OF SHEAR WALL STRUCTURES.

parate yet coupled phenomenologies. These include two-phase HWANG,H.; NAKAl,K.; REICH,M.; et al. Brookhaven Na9onal thennodynamics and thermal-hydraubs, aerosol physics, fission Laboratory. January 1986. 45pp. 8604010174. BNL-NUREG_

product phenomena, core-concrete interactions, the formation 51905.35344:352.

and combustion of flammable gases, and performance of engi-This report descnbes the development of probabihty-based neered safety features. In the past, this complexity has meant load combination enteria for the design of reinforced concrete that a complete containment analysis would require application shear wall structures subjected to dead load, live load and of suites of separate computer codes each of which would treat earthquake. The proposed design enteria are in the load and re-only a narrower subset of these phenomena, e.g., a thermal-hy-sistance factor design (LRFD) format. The load and resistance draulics code, an aerosol code, a core concrete interaction factors are determined for flexure and shear limit states and code, etc. In this paper, we describe the development and target hmit state probabihties. The flexure limit state is defined some recent applications of the CONTAIN coJe, which offers according to the ACI ultimate strength formula. The shear hmit an integrated treatment of the dominant containment phenom-4 state is established from experimental results. In order to test ena and the interactions among them. We describe the results whether the proposed critena meet the reliability based perform

• of a series of containment phenomenology studies, based upon ance objectives, four representative structures are selectd using realistic accident sequence analyses in actual plants, which a Latin hypercube sampling technique. These representative highlight various phenomenological effects that have potentially structures are designed using trial load and resistance factors.

important implications for source term and/or containment load-Then, a reliabihty analysis method is employed to assess their ing issues, and which are difficult or impossible to treat using a rehabilities. An objective function is defined and a minimization less integrated code suite. The results described show that technique is developed to find theoptimum load factors. In this analyses with non-integrated, separate-effects codes can ne-study, the resistance factors for shear and flexure, and load fac-tors for dead and lise loads are preassigned to simplify the mini-glect interactions that are important to the source term and,

z mization. The load factor for SSE is determined for the target furthermore, it is impossible to generalize whether the errors in such treatments would be " conservative" or "non-conserva-hmit state probability of 1.0 x 10(-6) or 1.0 x 10(-5) with a hfe-time of 40 years.

tive". It is concluded that integrated phenomenological analysis NUREG/CR-4337:

TRAC-PF1/ MODI INDEPENDENT e accid n ana sis t ASSESSMENT;OARTMOUTH COLI.EGE AIR-WATER COUNTER-CURRENT FLOW TESTS. DOBRANICH,D. Sandia NUREG/CR-4348 V01: COMMIX-1B:A THREE-DIMENSIONAL National Laboratones. December 1985. 86pp. 8602260074.

TRANSIENT SINGLE-PHASE COMPUTER PROGRAM FOR SAND 85-1594. 34703:150.

THERMAL HYDRAULIC ANALYSIS OF SINGLE AND MULTI-The TRAC-PF1/ MOD 1 independent assessment project at COMPONENT SYSTEMS.Vol LEquations And Numerics.

• Ar-Sandia is part of an overall effort funded by the NRC to deter-gonne National Laboratory. September 1985. 175pp.

1 mine the ability cf vanous advanced best-estimate systems 8603100036. ANL-85-42. 34889:314" codes to predict the detailed thermal / hydraulic response of The COMMIX-18 computer program, an extended version of LWRs during accident and off-normal conditions. As part of this l

effort, calculations for some Dartmouth College single-channel COMMIX 1 A, is designed to analyze steady-sate / transient, and three-channel air water counter-current flow tests have single-phase, three-dimensional fluid with heat transfer in reac-been performed. The single-channel calculations indicate the for components and multicomponent systems. The concepts of TRAC was able to predict the quaktative aspects of the different volume porosity, directional surface permeability, distnbuted re-i i

regione associated with the annular-flow regime. (There are four sistance, and distnbuted heat source or sink is used to model a distinct regions in the annular flow regime in which different flow domain with stationary structures. The new porous-medium pressure-loss mechanisms dominate.) However, severa: r,iscrep.

formulation permits simulation of either a single component or a ancies were discovered involving the magnitudes of the differ.

multicomponent system. The conservation equations of mass, ential pressure and the air velocity at which transihons between momentum, and energy based on the new porous-medium for-the different regions occurred. Also, a problem with TRAC nu-mulation are solved as a boundary-value problem in space and merics involving wall friction was discovered. Parametnc calcu.

an initial-value problem in time. Volume i of this report, entitled lations indicate that counter-current flow predictions are sensi-

" Equations and Numerics " describes in detail, the basic equa-tive to the number of cells used to model the channel and to tions, formulations, solution procedures, rebalancing scheme for the choice of fnction factor options. An input model using a faster convergence, models to describe the auxiliary phenom-VESSEL component as opposed to PIPE and TEE components ena, etc. Volume it, entitled " Users Manual," describes in detail, gave somewhat better results. Also, the problem with the fnc-flow chart, available options, input instructions, sample prob-tion numerics did not occur in this calculation. The three-chan-lems, etc.

nel tests were not useful for assessment purposes with respect a

14 Main Citations and Abstracts NUREG/CR-4344 V02: COMMIX-1B:A THREE-DIMENSIONAL compressive strength and reinforcement yield strength are in-TRANSIENT SINGLE-PHASE COMPUTER PROGRAM FOR cluded in the reliability analysis by using the Latin hypercube THERMAL HYDRAULIC ANALYSIS OF SINGLE AND MULTI-sampling technique. Then, the reliability analysis results and fra-COMPONENT SYSTEMS.Vol ll: User's Manual.

• Argonne Na-gility curves for two containments are presented.

tional Laboratory. September 1985. 226pp. 8603100563. ANL-854 2.34891:291.

NUREG/CR-4369: OUALITY ASSURANCE (QA) PLAN FOR l

See NUREG/CR-4348 Vol 1.

COMPUTER SOFTWARE SUPPORTING THE U.S. NUCLEAR NUREG/CR-4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 REGULATORY COMMISSION'S HIGH-LEVEL WASTE PRO-(VERSION 7.0),RELAPS/ MOD 1(CYCLE 14),AND TRAC-BD1 GRAM. WILKINSON G.F.; RUNKLE,G.E. Sandia National Lab-(VERSION 12.0) CODES USING SEPARATE-EFFECTS EX.

oratories. January 1986. 31pp. 8602050307. SAND 851774.

PERIMENTS. SAHA,P.; JO.J H.; NEYMOTIN L; et at Brookha-34486:219.

ven National Laboratory. August 1985.164pp. 8603100042.

A quality assurance plan has been developed for computer BNL-NUREG-51919. 34888:207, software created and/or maintained by Sandia National Labora.

This report presents the results of independent code assess-tories, Division 6431, and subsequently transferred to the U.S.

ment conducted at BNL The TRAC.PF1(Version 7.0) and Nuclear Regulatory Commission in support of its high-level RELAP5 MOD 1 (Cycle 14) codes were assessed using the enti-waste program. The plan contains requirements for software cat flow tests, level swell temt, countercurrent flow limitation storage and documentation, as well as a bnef description of the (CCFL) tests, post-CHF test, steam generator thermal perform-program maintenance process. Division 6431 has established a ance tests, and natural circulation tests. TRAC-BD1 (Version Computer Maintenance System for implementing this quality as-12.0) was applied only to the CCFL and post-CHF tests.

surance plan.

NUREG/CR-4363: A STUDY ON FABRICATION CRITERIA FOR DUCTILE CAST IRON SPENT-FUEL SHIPPING CONTAINERS.

NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALY-SCHWARTZ,M.W. Lawrence Livermore National Laboratory.

SIS METHODOLOGY. Volume 1: Methodology Report.

February 1986. 31pp. 8603030537. UCRL-53S62. 34779:207.

OZTUNALI,0.L Envirosphere Co. ROLES,G.W. NRC - No De-This report presents a study on enteria for fabncating ductile tailed Affiliation Given. January 1986. 777pp. 8602050287.

cast iron shipping containers used for transporting radioactive 34482:188.

matenais. Emphasis is on providing a specification that will not Under contract to the U.S. Nuclear Regulatory Commission, only describe the mechanical properties of the ductile iron but the Envirosphere Company has expanded and updated the im-will, in addition, ensure that these properbes will be reliably re-pacts analysis methodology used during the development of the produced in production castings.

10 CFel Part 61 rule to allow improved consideration of the NUREG/CR-4364: MANAGEMENT PERCEPTION OF THE costs and impacts of treatment and disposal of low-level waste HEALTH PHYSICS TECHNICIAN JOB. MAROTTA,F.J. Brookha-that is close to or exceeds Class C concentrations. These modi-ven National Laboratory. MAZOUR.T.J. Analysis & Technology, fications prMcipally include: (1) an update of the low-level radio-Inc. December 1985. 225pp. 8601280016. BNL-NUREG-51912.

active waste source term, (2) consideration of additional alterna-34386:117.

tive disposal technologies, (3) expansion of the methodology in 1984, an industry-wide job analysis of nuclear power reac-used to calculate disposal costs, (4) consideration of an adde-tor Health Physics Technicians (HPTs) was completed. These tional exposure pathway involving direct human contact with dis-results provided the basis for job descriptions, industry-wide posed waste due to a hypothetical drilling scenario, and (5) use task listings and recommendations for task selections for further of updated health physics analysis procedures (ICRP-30).

analysis and formal training. A total of 389 tasks were identified ano reviewed by 850 HPTs representing 39 plants and 6 vendor Volume 1 of this report desenbes the calculational algonthms of companies. Constructive enticism of the PHT job analysis fo-the updated analysis methodology, while Volume 2 describes cused on the fact that HPT supervisors / managers might have a the computer codes wntten to implement the updated analysis divergent perception of the HPT job that could lead to different methodology plus provides some example problems.

conclusions than those that were onginally drawn. The onfy way NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALY-to confirm (or deny) this hypothesis was to venfy the results of the job analysis of HPTs by also surveying radiation protection SIS METHODOLOGY. Volume 2: Codes And Example Problems.

management personnel. A total of 19 HPT supervisors /manag-OZTUNAll,0.L; PON,W.D.; ENG,R.; et al. Envirosphere Co*

ers completed the same survey that had been used for HPT job January 1986. 358pp. 8602050454. 34485:221.

incumbents. They were asked to rate each task on the basis of See NUREG/CR-4370,V01 abstract.

their job incumbents

• performance of the task (i e., How difficult do your HPTs find the task to perform 7). Of the 1161 ratings NUREG/CR-4371: COMMIX 2:A THREE-DIMENSIONAL TRAN-(389 tasks, each with a rating for frequency, difficulty, and im.

SIENT COMPUTER PROGRAM FOR THERMAL-HYDRAULIC portance), only 37 of the ratings differed between the HPTs and ANALYSIS OF TWO-PHASE FLOWS.

• Argonne National Labo.

HPT supervisors / managers (at the 99. percent confidence level).

ratory. October 1985. 123pp. 8602260078. ANL-85-47.

Therefore, the pnncipal result of this project was that the HPT 34702:206.

supervisor / manager responses validated the HPT job incum.

The mathematical bases of the computer program COMMIX 2 bent responses. Said another way, the HPT supervisors /manag-are described in detail. COMMIX-2, which desenbes steady-ers view the HPT job in the same way as do HPT job incum-state and transient single and two-phase flow conditions, has bents.

been developed from the single-phase flow version of COMMIX, NUREG/CR-4366: RELIABILITY ASSESSMENT OF CONTAIN.

COMMIX-1 A, and relies on the same concepts of volume poros-MENT TANGENTIAL SHEAR FAILURE.

PEPPER,S.;

ity, directional surface porosity, distntsted resistance, and heat HWANG.H.; PIRES.J. Brookhaven National Laboratory. January source. Two models for two-phase flow are included in the pro.

1986.43pp.8603030056. BNL-NUREG-51913. 34775:204.

gram: one is a three-equation Slip Model (SM), which assumes This report presents the latest developments of the reliability either a constant slip ratio between the phase velocities, or a analysis method for concrete containments. In specific, the tan.

relative slip velocity normalized by the mixture velocity; the gential shear limit state for reinforced concrete containments is other is a five-equation unequal phase velocity, equal phase described in detail. Also, the flexure limit state has been modi.

temperature (UVET) model, which is referred to as the Separat-fied such that the strain of tensile reinforcement is limited. Fur.

ed Phases Model (SPM). This report documents the state of the thermore, the variations of matenal strength such as concrete program development as of October 1984.

Main Citations and Abstracts 15 NUREG/CR-4372: PROBABALISTIC RISK ASSESSMENT (PRA)

Waste-form studies are being directed toward investgating APPLICATIONS. HIGGINS,J.C. Brookhaven National Laborato-spent-fuel leaching / dissolution behavior. Experiments have ry January 1986. 40pp. 8602200466. BNL-NUREG-51914.

been started to generate data on UO(2) and spent-fuel leach 34670:097, rates in simulated anoxic groundwaters. Inebal data indicate that The report presents the methodology developed to utilize the uraqium concentrations in the groundwaters and distilled-water results of a plant specific Probabihstic Risk Assessment (PRA) leachants are very low. The influence of groundwater species for onsite plant review and inspection purposes. The report then on the susceptibiltty of cast steel to pitting corrosion and stress-presents the results performed using these techniques at the corrosion cracking is being studied by potentiodynamic polariza Limerick Generating Station. Additionally, system reliability in-tion techniques. Potential cracking agents are being investigated sights obtained from the PRA were compared with pertinent by slow strain rate expenments. The pitting-corrosion model NRC Inspection Procedures to determine if changes were war

• was further developed, taking into account cation dissolution at ranted. Finally, conclusions regarding the usefulness methods the pit base and chemically active pit walls. Groundwater-radiol-and results are presented.

ysis modeling has continued, with the descnphon being ex-NUREG/CR-4378: OBJECTIVE INDICATORS OF ORGANIZA-tended to include bicarbonate anions in groundwater. Simula-TIONAL PERFORMANCE AT NUCLEAR POWER PLANTS.

tions show that modifications to the reactions accounting for bi-

+

i OLSON,J.; OSBORN,R.N.; JACKSON,D.H.; et al. Battelle Me-carbonate should improve predicted pH values. Spent. fuel monal Institute, Pacific Northwest Laboratories. January 1986.

specimens are being used in integral tests with flowing simulat-56pp. 8601210643. PNL-5576. 34318:106, ed groundwater to study the role of cladding in radionuchde re-This report summanzes research conducted under the spon-lease and to identify possible combined-effects processes.

sorship of the NRC, on the development and validation of orga-nizational performance measures at operating nuclear power NUREG/CR-4380: EVALUATION OF THE MOTOR-OPERATED plants. Publicly available data, including measures from Licens-VALVE ANALYSIS AND TEST SYSTEM (MOVATS) TO ee Event Reports, operating and outage data, and violations DETECT DEGRADATION,1NCORRECT ADJUSTMENTS.AND data, are used to predict penultimate measures of plant safety.

OTHER ABNORMALITIES IN MOTOR-OPERATED VALVES.

Penuttimate measures of safety include potentially significant CROWLEY,J.L; EISSENBERG,D.M. Oak Ridge National Labo-events, overexposures and near overexposures, and several ra-ratory. January 1986. 325pp. 8602270383. ORNL-6226.

diological release measures. ThJ 1981 and 1982 performance 34714:003.

measures are used in correlation and regression analyses to As a part of the NRC Nuclear Plant Aging Research (NPAR) predict performance on the penultimate safety measures in program a field test program was carried out to evaluate a 1982 and 1983. Many of the plant performance measures are technique of valve signature analysis to detect and differentiate consistentty predictive of the frequency of potentially significant abnormahtes, including time-dependent degradation (aging) and events. No strong. consistent predictors emerge for exposures incorrect adjustments in motor-operated valves. The technique or liquid radiological releases. Several performance measures described in this report is the Motor-Operated Valve Analysis are consistent predictors of gaseous releases. The regression and Test System (MOVATS) which is commercially available analyses indicate that the predictors do not tend to combine in from MOVATS, Inc. In-situ signature traces were obtained on 36 consistent, multivariate pattems, and controls for plant age.

motor operated valves at 4 nuclear plant sites. Desenbed in this size, type, region, and fuel cycle stage do not substantially letter report are the test equipment package, the method of ob-affect the resuits. The analysis concludes that existing perform

• taining the signatures, and the determinations made as a result ance data do not appear to be predictive of scme aspects of of analyzing these signature +. Based on evaluations of the sig-plant safety performance. The report recommends that more re-nature analysis techtuque, and on the results obtained from the liable, summary performance measures be created by combin-field test program, the capabilities and hmitations of MOVATS ing several of the performance measures tested in the current are discussed.

analysis.

NUREG/CR-4379 V02: LONG-TERM PERFORMANCE OF MATE.

NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESS-RIALS USED FOR HIGH-LEVEL WASTE PACKAGING.Second MENT OF U.S.AND FOREIGN NUOLEAR POWER PLANT Quarterty Report, Year Four July. September 1985. STAHL.D.;

DOSE EXPERIENCE. BAUM,J.W.: HORAN,J.R. Brookhaven Na-MILLER,N.E. Battelle Memorial Institute, Columbus Laborato-tional Laboratory. October 1985. 23pp. 8602270500. BNL-ries. January 1986.102pp. 8602200432. 34669:356.

NUREG-51918. 34715:069.

Waste-form studies are being directed toward studying spent.

Data gathered at the 1984 BNL Workshop on Historical Dose fuel leaching / dissolution behavior. Glass crystallinity data were Experience and Dose Reduction (ALARA) at Nuclear Power derived in preparaFon for a devitnfication/ glass-lecching experi-Plants and from recently published hterature were reviewed and ment. Glass-dissolution-rate data were collected which venfied anatyzed. Large differences were noted, between countnes and the Battelle dissolution model, and data from an organic-acid between simslar plants, for collective dose (man-rem) per plant leach experiment were analyzed. The reactions of groundwater and per unit of electncity generated (MWe-yr). During the period species with steels are being studied to evaluate susceptability 1978-1982, for PWRs, the U.S. ranked highest in terms of col-i to pitting and stress-corrosion cracking. Potential crackin9 lective dose per MWe-yr (1.2), and France, Sweden, and Fin-agents are being investigated by slow strain rate experiments.

land were lowest (0.27-0.37). For BWRs, Japan, the U.S., and The general-corrosion model was further developed, based on the Federal Republic of Germany ranked highest (2.21.9),and known pnnciples of mass transport and radiotytic production-F nland and Sweden were lowest (0.08 0.32). Only a small por-Spent fuels are being used in integral tests with flowing simulat.

tion of the differences could be attnbuted to average plant age, ed groundwater to study the role of cladding in radionuclide re*

vintage, or rated capacity. Fifteen factors were identified (in ad-lease and certain combined-effects processes. The water-chem-dition to age) which contnbute to differences in estimated order istry model was expanded to include uranium species, and inter

• of importance, these were plant chemistry, water punfication, actions between %n and chloride species and water-radiofyses materials selection for low cobalt and nickel, special tools, de-species were examined.

contamination of primary systems, required multi-plant acaons, l

NUREG/CR-4379 V03: LONG-TERM PERFORMANCE OF MATE.

worker motivation and commitment, permanent work force, RIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Third management commitment to dose control, three or more reac-Ouarterfy Report, Year Four October -December 1985.

tors per site, design for reliabihty, passivation of pnmary sys-1 l

STAHL,0.; MILLER,N E. Batteile Memorial Institute, Columbus tems, quality assurance, standardized plant design, and shield.

Laboratories. March 1986. 99pp. 8603280281. 35326:108.

ing.

16 Main Citations and Abstracts NUREG/CR-4389: PRESSURE NOISE IN PRESSURIZED WATER NUREG/CR-4402 V02: HIGH-TEMPERATURE GAS-COOLED RE-REACTORS. MULLENS.J.A.; THIE.JA Oak Ridge National Lab.

ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT oratory. December 1985 131pp.8601280018. ORNL/TM-9773.

EVALUATION.Ouarterly Progress Report, April 1 - June 34385:279.

30,1985. BALL,S.J.; CLEVELAND.J.C.; HARRINGTON,R.M.; et A general study of pressure fluctuations (noise) in the pnmary al. Oak Ridge National Laboratory. February 1986. 18pp.

coolant loop of pressurized waters reactors (PWRs) wss per.

8604010169. ORNL/TM-9798/V2. 35343:278.

J formed. The study included noise sources, PWR pressure dy-Modehng, code development, and analyses of the modular namics, and pressure-noise measurements. A detailed model of High-Temperature Gas-Cooled Reactor (HTGR) continued with noise in the Loss-of-Fluid Test (LOFT) Facshty was developed work on the side-by-side design. Fission-product release and j

and compared against measurements. The potential of pres-transpon expenments we completed Sechons of an N safety handbook were written.

sure-noise measurements for reactor monitonng and diagnosis of problems was assessed, particularly collapse of the pressur.

NUREG/CR-4411: ASSESSMENT OF SPECIALIZED EDUCA.

izer steam bubble, loss of subcooling, and sensing-hne defects.

TIONAL PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS. MELBER,B.D.; SAARI.L.M.; WHITE,A.S.; et al.

NUREG/CR-4390: DCC-1/DCC-2 DEGRADED CORE COOLABl*

Battelle Human Affairs Research Centers. February 1986.88pp.

LITY ANALYSIS. REED.A.W.; BALDT,K.R.; GORHAM-BER-8603180516. PNL-5602. 35137:091, GERON; et al. Sandia National Laboratones. October 1985.

This report assesses the job-relatedness of specialized edu-113pp. 8602260121. SAND 85-1967, 34580:001.

cational programs for licensed nuclear reactor operators. The The DCC-1 and DCC2 experiments provided the first data on approach used involved systematically companng the curriculum dryout and quench behavior of internally heated UO(2) debris in of speciahzed educational programs for college credit, to aca-water. The pressure range of this data extends from 1 to 170 demic knowledge identified as necessary for carrying out the atmospheres. Both of the experiments used the Annular Core jobs of licensed reactor operators. A sample of eight programs, Research Reactor (ACRR) at Sandia National Laboratones to including A.S. degree, B.S. degree, and coursework programs simulate the effect of radioactive decay heating. Th debris in were studied. Subject matter experts in the field of nuclear op-DCC-1 ranged from 75 microns to 10 mm in diameter, with a erations curriculum and training determined the extent to which mean diamerer of 0.75 mm. The bed depth was 0.5 and the po.

individual program cumcula covered the identified job-relaiid rosity was 0.J45. Dryout heat fluxes ranged from about 41 kW/

academic knowledge. The major conclusions of the report are; m(2) (0.012 W/g) at a saturation temperature of 100 degrees (1) there is a great deal of vanation among individual programs, centigrade to about 69 kW/m(2) (0.021 W/g) at 340 degrees ranging from coverage of 15% to 65% of the job-related aca-centigrade. This measured pressure dependence is a factor of demic knowledge. Four schools cover at least half, and four two to three lower than predicted by the analytical models. This schools cover less than one-third of this knowledge coritent; (2) is beheved to be the breadth of the debris distnbution, but the there is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S.

evidence is inconclusive. Quenches of dried debns took hours e,a cwrseM @ Wahal M We to complete. Quench fronts progressed uniformly without the liquid fingers observed in large particle tests. The debns distri-pmgrams n n ar eginenng m as M WaM bution in DCC-2 was much narrower than in DCC 1, with the j

majonty of particles having diameters between 0.5 and 8 mm. A 3

small amount of " fines" was added to the mixture. In DCC-2, NUREG/CR-4419: BIOASSAY MEASUREMENTS FOR URANIUM thermally stable local dry zones were observed at bed powers USING SPUTTER INITIATED RESONANCE lONIZATION d

below the conventional dryout point. These are caused by the SPECTROSCOPY. PARKS.J.E.; TAYLOR,E.H.; BEEKMAN,0.W.;

concentration of fines creating a low permeability zone. Data on et a!. Atom Sciences, Inc. January 1986. 22pp. 8601290231.

global dryout, in which the bed bottom can dry out, agree well 34419:267.

with analytical predictions. Quenches of dry zones took about it was determined earlier in a feasebility study (Ref.1) that 10 minutes to complete. The quench fronts were not uniform, Sputter initiated Resonance Ionization Spectroscopy (SIRIS) having a liquid finger which penetrated to the bottom of the bed would be feasable as an ultrasensstive analytical method for the before the quench was complete.

detection of uranium, plutonium, and thonum in unne and other bioassay samples. SIRIS is a method which uses sputtering to l

NUREG/CR-4393:

SUMMARY

OF SEMISCALE SMALL BREAK atomize a sample, tunable dye lasers to selectively ionize the LOSS OF-COOLANT ACCIDENT EXPERIMENTS (1979 TO element of interest, and a mass spectrometer to detect the sep-1985). LOOMIS.G.G. Idaho National Engineering Laboratory.

arated isotopes of the selected element. We report here dem-September 1985. 83pp. 8604010147. EGG-2419. 35343:323.

onstration measurements for the analysis of uranium in simple Following the loss-of-coolant accident at TM; in 1979, a multi-aquems solubon and in synthebc wim to a detechon hmn of 1 tude of small break loss-of-coolant accident experiments were mg/1. Sample preparation procedures are discussed. The tech-performed in the vanous Mods of the Semiscale facility at the neue of We mon was used W caWabon. De resuus of idaho National Engineenng Laboratory (INEL). A summary of the measurements are reported and discussed. Steps to lower what experiments have been performed and a description of the elechon knW to ROS mgh am mported Cost esumates for vanous Semiscale Mods is given. The signature response of various kinds of small breaks are characterized. Small break NUREG/CR 4420: TURC11ARGE SCALE METALLIC MELT.

loss-of coolant accident issues addressed by Semiscale testing CONCENTRATE INTERACTION EXPERIMENTS AND ANALY-are discussed, including: effect of break location and break size, SIS. GRONAGER,J.E.; SUO-ANTTILLA A.; BRADLEY,0.R.; et effect of core bypass flow, preferred primary coolant pump op-al. Sandia National Laboratories. January 1986. 200pp.

eration, effectivenss of upper head emergency core cooling in-8603100099. SAND 85-0707. 34889:115.

i jection, and recovery procedures. Phenomena of interest to Two large-scale molten debris-concrete experiments, sman break loss-of-coolant accident analysis is presented in-TURCIT, a thermite-concrete interaction expenment, and ciuding core uncovery heat transfer and natural circulation. Rec.

TURCISS, a stainless steel-concrete experiment, are reported i

ommendations are given that can improve calculational capabili.

here. The experiments consisted of teeming molten debris

)

ties for future small break testing.

(>100 kg) onto hmestone/ common sand concrete. The molten debns was allowed to cool naturally. The concrete ablation rate, composition of evolved gases, and aerosol data are presented.

t

Main Citations and Abstracts 17 The experimental results have been compared to CORCON cal-robin study, and d) prepare a project plan for conducting the culations in order to validate the code. This comparison showed needed follow-on work identified during the feassbelity study.

that while some parts of the code performed well (chemical This report includes an assessment of UT applicabens in the equilibrium model), other sections require further model devel-nuclear industry, emphasizing the UT/ISI system (p sonnet, opment (melt-concrete heat transfer modef). An analysis of the equipment, and procedures). A man-mt. ane systems model is two experiments was performed using a new analysis model.

used to describe the UT/ISI process, and the Relative Operat.

The results of 'he anslysis seem to suggest that the heat trans-ing Characteristic (ROC) analysis approach for analyzing NDT fer mechanism of concrete ablation is similar to nucleate boiling performance is discussed. The five basic performance-shaping heat transfer, rather than gas film heat transfer.

factors (variables) are evaluated with respect to potentialimpact NUREG/CR-4431:

SUMMARY

REPORT ON THE SEISMIC on UT/ISI performance. Conclusions are drawn based on this SAFETY MARGINS RESEARCH PROGRAM. CUMMINGS,G.E.

limited scope study, and recommendations are made for Lawrence Livermore National Laboratory. January 1986. 78pp.

needed follow-on work toward improving UT/ISI reliabelity.

8601160405. UCID-20549. 34291:086.

NUREG/CR-4436 V02: HUMAN RELIABILITY IMPACT ON IN-The Seismic Safety Margins Research Program (SSMRP) was SERVICE INSPECTION Volume 2: Review And Anatysis Of a U.S. NRC-funded multi-year program conducted by Lawrence Human Performance in Nondestructive Testing (Emphasizing l

Livermore National Laboratory. Its goal was to develop a com-Ultrasonics). TRIGGS,T.J.; RANKIN,W.L Battelle Human Affairs plete, fully coupled analysis procedure (including methods and Research Centers. BADALAMENTE.R.; et al. Battelle Memorial computer) for estimating the risk of an earthquake-induced ra-Institute, Pacific Northwest Laboratories. March 1986. 131pp.

dioactive release from a commercial nuclear power plant. The 8604040596. PNL-5641. 35405:341.

SSMRP was the first effort to trace seismically induced failure This report documents a review of the research literature in modes in a reactor system down to the individual component the human factors, nondestructive tesbng (NDT), and related level, and to take into account common-cause earthquake.in-duced failures at the component level. This report summarizes areas that was conducted to develop an understanding of how methods and results generated by SSMRP.

humahperformance in NDT can be improved through the appli-cation of human factors principles relating to task, training, pro-NUREG/CR-4433: DOCUMENT REVIEW REGARDING HAZARD-cedural, individual difference, and environmental variables. A OUS CHEMICAL CHARACTE91STICS OF LOW-LEVEL WASTE.

secondary purpose of this project was to develop a satisfactory BOWERMAN.B.S.; DAVIS,RE.: SISKIND.B. Brookhaven Nation-measure of performance that can be applied in NDT. The report al Laboratory. March 1986. 99pp. 8603260354. BNL-NUREG-begins with an overview of NOT with special emphasis on ultra-51936. 35282:247.

sonic testing for inservice inspect >on (UT/ISI) regarding inter-A review of commercial disposal site records and literature granular stress corrosion cracking (IGSCC). Then, the strengths sources was conducted in order to determine whether low-level and weaknesses of typical measures of performance accuracy radioachve wastes (LLW) could be classified as hazardous are discussed. Signal detection theory measures of perform-under the definitions of 40 CFR Part 261. Two types of LLW ance accuracy are discussed. S,,nal detection theory is then from both fuel-cycle and non-fuel-cycle sources were identified presented, and it is recommended that the relative operating as potential concems: spent organic solvents and lead metal.

characteristic (ROC) analysis, which follows from signal detec-The first category includes liquid scinillation wastes, while lead tion theory, be adopted for specifying human performance in metal wastes may include discarded shielding and containers.

NDT, Human performance in NDT and NDT-related areas is Additiorul LLW categones may contain hazardous constituents then discussed with respect to the five types of variables out-listed in Appendix Vill of 40 CFR Part 261. The information lined above. This study strongfy suggests that NDT technician review was not suffsient to make a f:rm jud;; ment as to whether performance could be improved through the systematic applica-any LLW as shipped for disposal would be hazardous under 40 bon of human factors principles within the framework of ROC CFR Part 261.

analysis, especially in the areas of task, training, and, to a NUREG/CR-4434: ASSESSMENT OF MODELLING NEEDS FOR lesser extent, procedural variables.

SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS KROEGER,P.G.: VAN TUYLE.G.J. Brookhaven National Labora; NUREG/CR-4434: RESULTS OF SEMISCALE MOD-2C SMALL tory. December 1985. 72pp. 8604010132. BNL-NUREG-51937.

BREAK (5%) LOSS-OF-COOLANT ACCIDENT EXPERIVENTS S-LH-1 AND S-LH-2. LOOMIS,G.G.; STREIT,J.E. EG&G Idaho, 35345:201' f the recent shift in emphasis of the DOE / industry In view o Inc. (subs. of EG&G, Inc.). November 1985. 110pp.

8 24 2424 683 241.

HTGR development efforts to smaller modular designs it became necessary to review the modelhng needs and the codes available to assess the safety performance of these new accidents (5% SBLOCAs) were performed in the Semiscale designs. This report provides a final assessment of the most Mod-2C facility. These experimeats were identical except for urgent modetting needs, comparing these to the tools available, downcomer to upper-head bypass flow (0.9% in Experiment S-and outlining the most significant areas where further modelling LH-1 and 3.0% in Experiment S-LH-2) and were performed at I

is required. Plans to implement the required work are presented.

high pressure and temperature [15.6 MPa (2262 psia) system pressure; 37K (67 degrees fahrenheit) core differential tempera.

I NUREG/CR 4436 V01: HUMAN RELIABILITY IMPACT ON IN-ture; 595 K (610 degrees fahrenheit) hot leg fluid temperature).

SERVICE INSPECTION. Volume 1: Phase 1 Summary Report.

From the experimental resuits, the signature response and tran-SPANNER J C.: BADALAMENTE.R. Battelle Memonal Institute, sient-, ass distribution are determined for a 5% SBLOCA. The l

Pacific Northwest Laboratories. RANKIN,W.L; et al. Battelle core thermal-hydraulic response is characten2ed, including core 1

Human Affairs Research Centers. March 1986. 106pp.

void distribution maps, and the effect of core bypass flow on 8604040606. PNL 5641,35405:229.

transient seventy is assessed. Comparisons are made between This report documents a feasibihty study conducted to idents-postexperiment RELAPS calculations and the experimental re-i fy, charactenze, and evaluate the human reliability aspects of ul-Sults, and the capability of RELAP5 to calculate the phenomena trasonic testinghnservice inspection (UT/ISI). Dunng this study, is assessed.

4 j

the following four inter-dependent tasks were completed. a) per.

l form a literature survey to identify significant human perform-NUREG/CR-4442: TRAC USER'S GUIDE. BOYACK,B.E.;

l ance factors and develop a better measure of performance for STUMPF,H.: LIME,J.F. Los Alamos Scientific Laboratory. No.

the UT/Isl process, b) develop task element descr:ption for a vember 1985. 316pp 8603120077. LA 10590-M. 34925:164.

typical UT/ISI process, c) prepare a test plant to conduct a This guide has been prepared to assist users in applying the human performance evaluation in conjunction with a mini-round Transient Reactor Analysis Code (TRAC). TRAC is an advanced t

i i

I

i 18 Main Citations and Abstracts best-estimate systems code for analyzing transients in thermal-were compared to the RELAP5 results and the INEL extrapola-hydraulic systems. The code is very general. Beca'/se it is gen-tions. In general, the RELAPS and INEL results appear to be eral, efforts to model specific nuclear power plants or experi-reasonable.

mental facihbes often present a challenge to the, TRAC user.

NUREG/CR-4453 V01: LIGHT-WATER-REACTOR SAFETY FUEL This guide has been wntten to assist first-time or intermediate SYSTEMS RESEARCH PRO Quarterly Progress 1

Report, January-March 1985.. GRAMS.

users. It is specifically wntten for the TRAC version designated Argonne National Laboratory.

TRAC-PF1/ MODI. The TRAC User's Guide should be consid-ered a companion tocument to the TRAC Code Manual; the January 1986. 41pp. 8603110532. ANL-85-71 V01. 34914:136.

user will need both documants to use TRAC effectively.

This progress report summarizes work performed by the Ma-terials Science and Technology Division of Argonne National NUREG/CR-4446: THE NUCLEAR INDUSTRY AND ITS Laboratory during January, February, and March 1985 on water REGULATORS:A NEW COMPACT IS NEEDED.

• International reactor safety problems related to fuel and cladding. The re.

Energy Associates, Ltd. February 1986.103pp. 8603100039.

search and development areas covered are Transient Fuel Re-IEAL-R/85-70. 34d89:010.

sponse and Fission Product Release and Clad Properties for The problem of the lack of integration in the nuclear power Code Verification.

decision-making process in the United States is the subject of this study. The three institutions with the greatest influence on NUREG/CR-4459: LIGHT WATER REACTOR SAFETY RE-commercial nuclear power generation include the utilities, the SEARCH PROGRAM. Semiannual Report, October 1983 March Nuclear Regulatory Commission (NRC), and the state public util.

1984. BERMAN,M. Sandia National Laboratories. February ity commissons (PUCs). The diverse objectives of the three in.

1986.132pp. 8604040095. SAND 85-2500. 35406:119.

stitutions are difficult to satisfy without producing confhet. This This report desenbes the investigations and analyses con.

has contnbuted to inefficiences and delays in nuclear plant con.

ducted at Sandia National Laboratories, Albuquerque, in support struction and operation, gaps in quality assurance, and may also of the Light Water Reactor Safety Research Program from Oc-result in compromises to public health and safety. This report tober 1983 through March 1984. The Core Melt / Coolant Inter.

reviews the perspectives of each of these institutions and pro.

actions (CMCl) Study investigates the charactenstics of explo-vides recommendations for improvements. Particutar emphasis sive and nonexplosive interactions between molten core materi-is given to recommendations that NRC might consider to help als and water and the probabilities and consequences of such alleviate the potential for adverse impacts on public health and interactions. The objectrve of the Hydrogen Program is to quan-safety resulting from a disaggregated nuclear power decision.

tify tne threat to nuclear power plants (containment structure, making process.

safety equipment, and the pnmary system) posed by hydrogen combustion. Experiments have been performed investigating the NUREG/CR-4450 DRF FC: MANAGEMENT OF RADIOACTIVE various modes of combustion. Data is being used to develop MIXED WASTES IN COMMERCIAL LOW-LEVEL WASTES. Draft and assess analytical models. All activities are continuing.

Report For Comment.

KEMPF,C.R.;

MACKENZIE.DA; BOWERMAN.B.S. Brookhaven National Laboratory. January NUREG/CR-4460: UNCERTAINTY AND SENSITIVITY ANALYSIS 1986.147pp.8603030052. BNL-NUREG-51944. 34774:001.

OF AN UPPER PLENUM TEST PROBLEM FOR THE MAEROS Managment options for three generic categories of radioac-AEROSOL MODEL HELTON,J.C.; IMAN,R.L; JOHNSON.J.D.;

tive mixed waste in commercial low-level wastes (LLW) have et al. Sandia National Laboratories. January 1986. 41pp.

been identfied and evaluated. These wastes were characterized 8602270377. SAND 85-2196. 34715:266.

as part of a BNL study in which LLW generators were surveyed The MAEROS aerosol model is being incorporated into the for information on potential chemical hazards in their wastes.

MELCOR code system for the calculation of risk from severe re-The general management targets adopted for mixed wastes are actor accidents. To gain insight to assist in this incorporation, a destruction, immobilization, and reclamation. Solidification, ab-computational test problem involving a three component aerosol sorption, incineration, acid digestion, wet-air oxidation, distilla-in the upper plenum of a pressunzed water reactor was ana-tion, liquid-liquid solvent extraction, specific chemical destruction lyzed with MAEROS. The following topics were investigated (1) techniques, and substtution have been considered for organic the CRAY 1 CPU time requirements to implement and solve the liquid wastes. Containment, segregation, decontamination, and system of differential equations on which MAEROS is based (2) solidification or containment of residues, have been considered the effects on computational time and representational accuracy for lead metal wastes which have themselves been contaminat.

due to ttie use of different overall section boundaries and num-ed and are not used for purposes of waste disposal shielding, bers of sections and components, and (3) the behavior of the packaging, or containment. For chromium-containing wastes, aerosol and the variables which influence this behavior. Uncer-solidification, incineration, wet-air oxidation, acid digestion, con.

tainty and sensitnnty analysis techniques based on Latin hyper.

tainment, and substtution have been considered. For each of cube sampling and regression analysis were used in the investi-these wastes, the management option evaluation has included gation. Five sections and overall section boundanes from 1.1E-6 an assessment of testing appropriate to determine the effect of m to 50.E-6 m were found to be adequate for the pro)lem the option on both the radiological and potential chemical haz.

under consideration. Further, solution time was found to ')e at i

ards present.

least several hundred times faster than real time, which is ett to be adequate for MELCOR. Stepwise regression was used to in-NUREG/CR-4452: REVIEW OF RELAPS CALCULATIONS FOR aM N soumos of vaMon in copaknal W aM H.B. ROBINSON UNIT 2 PHESSURIZED THERVAL SHOCK suspended aerosol concentration.

)

STUDY. YUELYS-MIKSIS; ROHATGI,U.S.; JO.J.J. Brookhaven National Laboratory. December 1985. 6, p. 8604030081. BNL-NUREG/CR-4462: A RANKING OF SABOTAGE / TAMPERING NUREG-51946. 35392;158.

AVOIDANCE TECHNOLOGY ALTERNATIVES.

Idaho National Engineering Laboratory (INEL) used the ANDREWS,W B.; TABATABAl,A.S.; POWERS,T.B.; et al. Bat-RELAPS/ MOD 1.6 code to simulate a number of transient sce-telle Memorial institute Pacific Northwest Laboratories. January narios for the USNRC PTS study of the H.B. Robinson Unit 2 1986.138pp. 8602050458. PNL-5690. 34485.083.

PWR plant. Eleven of these scenarios were reviewed at BNL on Pacific Northwest Laboratory conducted a study to evaluate the basis of information recerved before September 30,1984.

attematives to the design and operation of nuclear power Six of these eleven scenarios were selected for an in-depth plants, emphasizing a reduction of their vulnerability to sabo.

quinttative analysis performed on the t' asis of a simple method tage. Estimates of core melt accident frequency during normal j

developed at BNL The simple method uses the mass and operations and from sabotage / tampering events were used to energy balance equations to predict the temperature and pres-rank the attematives. Core melt frequency for normal operations sure of the reactor system. The results of these calculabons was estimated using sensitivity analysis of results of probabilis.

J

Main Citations and Abstracts 19 tic risk assessments. Core melt frequency for sabotage / tamper.

NUREG/CR-4468: ADAPTATION OF OCA-P,A PROBABILISTIC ing was estimated by developing a model based on probabilistic FRACTURE-MECHANICS CODE.TO A PERSONAL COMPUT-risk analyses, historic data, engineering judgment, and safe.

ER. BALL.D.G.; CHEVERTON,R.D. Oak Ridge National Labora-guards analyses of plant locations where core melt events tory. January 1986. 28pp. 8603100046. ORNL/CSD/TM-233.

could be initiated. Results indicate the most effective alterna-34872.005.

tives focus on large areas of the plant, increase safety system The OCA-P probabilistic fracture-mechanics code can now be redundancy, and reduce reliance on single locations for mitiga.

executed on a personal computer with 512 kilobytes of memory, tion of transients. Less effective options focus on specific areas a math coprocessor, and a hard disk. A user's guide for the par-of the plant, reduce reliance on some plant areas for safe shut, tscular adaptation has been prepared, and additional importance down, and focus on less vulnerable targets.

sampling techniques for OCA-P have been developed that allow the sampling of only the tails of selected distributions. Features NUREG/CR-4464: PERFORMANCE DEMONSTRATION TESTS have also been added to OCA-P that permit RTNDT to be used FOR DETECTION OF INTERGRANULAR STRESS CORRO.

as an " independent" variable in the calculation of P(F/E).

SiON CRACKING. HEASLER,P.G.; BATES,D.J.; TAYLOR T.T.;

NUREG/CR-4469 V01: INTEGRATION OF NONDESTRUCTIVE et al. Battelle Memorial Institute, Pacific Northwest Laboratories.

EXAMINATION RELIABILITY AND FRACTURE February 1986. 61pp. 8603190017. PNL-5705. 35152:284.

MECHANICS: Semiannual Report.Apnl September 1984.

This report evaluates detection tests of inservice inspectors DOCTOR,S.R.; BATES,0.J.; CHARLOT,LA.; et al. Battelle Me-(ISI), procedures and equipment that are employed to find inter-morial Institute, Pacific Northwest Laboratories. January 1986.

granular stress corrosion cracks in nuclear power plant piping.

t,p. 8602210744. PNL-5711,34671:091.

Performance is described by two fundamental parameters: false Ine progress report summarizes work performed by Pacific call probability and.probabihty of detection. Acceptable inspec-Northwest Laboratory on the NDE and fracture mechanics of tion performance and detection tests are therefore defined in nuclear reactor pnmary circuit components during the six terms of these two parameters.

months from April 1984 through September 1984.

NUREG/CR-4465:

TRAC-PF1/ MOD 1 INDEPENDENT NUREG/CR-4471: LOS ALAMOS PWR DECAY-HEAT REMOVAL ASSESSMENT.SEMISCALE MOD-2A 4NTERMEDIATE BREAK STUDIES

SUMMARY

RESULTS AND CONCLUSIONS.

TEST S-IB-3. KMETYK,LN. Sandia National Laboratories. Feb.

BOYACK B.E.; HENNINGER,R.J.; HORLEY,E.; et al. Los ruary 1986.170pp. 8604010182. SAND 85-2563. 35345:035.

Alamos Scientific Laboratory. March 1986.176pp.8604070236.

LA 37 S 4 The TRAC-PFI/ MODI independent assessment project at of d wn-decay-heat removalin pressurized Sandia National Laboratones is part of an overall effort funded by t"s NRC to determine the ability of vanous system codes t water reactors (PWRs) is currently under investigation by the predict the detailed thermal / hydraulic response of light water re.

Nuclear Regulatory Commission. One part of this effort is the actors during accident and off-normal conditions. The TRAC review of feed-and-bited procedures that could be used if the code is being assessed at SNLA against test data from various normal cooling mode through the steam generators were un-integral and separate effects test facilities. As part of this as-available. Feed-and-bleed cooling is effected by manually acti-vating the high-pressure injection (HPI) system and opening the sessment matnx, an intermediate break test (S-1B-3), performed power-operated relief valves (PROVs) to release the core decay at the Semiscale Mod-2A facility, has been analyzed. Using an energy. The feasibility of the feed-and-bleed concept as a di-input model with a 3-D VESSEL component, the vessel and verse mode of heat removal has been evaluated at the Los downcomer inventones dunng S-lB-3 were generally well pre.

Alamos National Laboratory. The TRAC PF1 code has been dicted, but the core heatup was underpredicted compared to used to predict the expected performances of the Oconee-1 data. An equivalent calculation with an all 1-D input model ran and Calvert Cliffs-1 reactors of Babcock and Wilcox and Com-about twice as fast as our basecase analysis using a 3-D bustion Engineering, respectively, and the Zion-1 and H.B. Rob-VESSEL in the input modet, but the results of the two calcula-inson-2 plants of Westinghouse. Feed and bleed was success-tions diverged significantly for many parameters of interest, with fully applied in each of the four plants studied, provided it was the 3-D VESSEL model results in better agreement with data-initiated no later than the time of loss of secondary heat sink.

NUREG/CR-4466: STATION BLACKOUT TRANSIENTS IN THE NUREG/CR-4472: SIAMESE IMAGING TECHNIOUE FOR QUASI-SEMISCALE FACILITY. CHAPMAN.J.C. EGSG Idaho, Inc.

VERTICAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR (subs. of EG&G. Inc.). December 1985. 53pp. 8602260082.

PIPING. COLLINS,H.D.; GRIBBLE.R.P. Battelle Memorial Insti-EGG-2432. 34703:233.

tute, Pacific Northwest Laboratones. January 1986. 59pp.

The test results of station blackout transients conducted in 8602200472. PNL-5717. 34670:136.

the Semiscale facility are discussed in this report. The Semis.

This report presents work performed on the program, "The cale MOD-2B facility simulates a pressurized water reactor Integration of Nondestructive Examination Reliability and Frac-(PWR) power plant. The expenments were initiated from condi.

ture Mechanics," to address improvements that can be tions typical of PWR plant operating conditions [pnmary pres.

a,chieved with imaging techniques for more accurately locating, sure of 15.2 MPa (2205 psi) and cold leg fluid temperature of sizing and onenting quasi-vertical planar defects in nuclear reac.

55 K (530 degrees fahrenheit)]. Five station blackout experi.

tor piping.

ments were conducted: three tests in the Power Loss (PL) Test NUREG/CR-4473: A STUDY OF THE OPERATION AND MAIN.

Series and the two Primary Boil-off (PBO) Tests. The responses TENANCE OF COMPUTER SYSTEMS TO MEET THE RE-of these tests were analyzed and compared. However, only one OUIREMENTS OF 10 CFR 73.55.

LEWIS J.R.;

test response (S-PL 2) is presented and discussed in detail. The FLUCKlGER,J.D.; BYERS,K.R.; et al. Battelle Memorial Institute, S-PL-2 experiment is characterized by examining the responses Pacific Northwest Laboratories. January 1986. 86pp.

of the pnmary and secondary pressures and fluid temperatures.

860t160226. PNL 5680. 34285:188.

the pressurizer liquid level, the pnmary fWid distribution, and the The Pacific Northwest Laboratory has studied the operation core thermal behavior. The mechanisms driving the S-PL 2 re-and maintenance of computer managed systems that can help sponses, the main elements of the station blackout transient, nuclear power plant licensees to meet the physical secunty re-the influences of initial and boundary conditions, and other tran-quirements of 10 CFR 73.55 (for access controf, alarm monitor-sients that may appear similar to station blackout are also dis-ing, and alarm recording). This report of that study describes a cussed. Information pertinent to station blackout nuclear safety computer system quality assurance program that is based on a issues is also presented in the report.

system of related internal controls. A discussion of computer

20 Main Citations and Abstracts system evaluation includes verfication and validation mecha.

attributes of performance assessment, followed by discussions nesms for assunng that requirements are stated and that the of DOE methods, irobabilistic methods capable of predicting product fulfills these requirements. Finally, the report desenbes waste package lifetime and radio-nuclide releases, process operator and security awareness training and a computer modeling of waste package barriers, sufficiency of the neces-system preventive maintenance program.

sary input data, and the applicability of probability density func-NUREG/CR-4474: SCALE MODELING OF REINFORCED CON-tions. It is recommended that the initial NRC performance as-CRETE CATEGORY I STRUCTURES SUBJECTED TO SEIS-sessment (for the basalt conceptual waste package design)

MIC LOADING. DOVE.R.C.; BENNETTJ.G. Los Alamos Scien-apply modular simulation, using available process models and data, to demonstrate this assessment method.

tific Laboratory. January 1986. 29pp. 8604010162. LA-10624-MS. 35342:069.

The laws that govern the scate-model requirements for rem..

NUREG/CR-4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BEHAVIOR IN COMPLEX NUCLEAR POWER PLANT forced concrete Category i structures over a full range of seis.

ENCLOSURES.PRESENT CAPABILITIES AND FUTURE PROS-mic loading extending from the elastic through the inelastic PECTS. BOCCIO J.L; USHER,J.L Brookhaven National Labora-ranges of response are developed. Three types of scaling are tory. December 1985. 69pp. 8603310046. BNL-NUREG-51948.

then examined. The third type, called "O" scaling in this report, 35338;t79 is the most useful for tailonng structural models to existing seis.

rnic test facilites. Finally, the way in which the three types of This report provides a summary of the work conducted dunng commonly used damping (viscous, structural, and Coulomb)

FY84 by Brookhaven National Laboratory (BNL), under FIN A-scale in these models is denved.

3252 " Fire Protection Research Program." It was undertaken under the cognizance of the Electrical Engineering Branch in NUREG/CR-4475: ORMGEN PC.A MICROCOMPUTER PRO-the Division of Engineenng Technology within the Office of Nu-GRAM FOR AUTOMATIC MESH GENERATION OF 2-D clear Regulatory Research. The report describes a mathemats-CRACK GEOMETRIES. BRYSON.J.W.; BASS,B.R. Oak Ridge cal model for predicting the thermal environment within complex National Laboratory. March 1986. 94pp. 8604030261. ORNL-nuclear power plant enclosures. It demonstrates the capability 6250, 35403:183.

of the existing numerical code by direct comparisons with elec-ORMGEN PC (Oak Ridge Mesh Generation. Personal Com-trical cable fire /large enclosure tests performed by Factory puter) automabcally generates two-dimensional finite-element Mutual Research Corporabon (FMRC) for the Electnc Power models for either cracked or uncracked structures. Element Research insttute (EPRI). It further demonstrates the potential connectivities and nodal point coordinates are wntten in formats usefulness of the existing code in addressing fire-protection that are compatible for subsequent fracture anafysis using either issues. This is done through a parametric study of the thermal the ORVIRT.PC finite-elemer't microcomputer program or the environment resulting from a series of fires within cabinets in a ADINA/ORVIRT mainframe system. ORMGEN.PC emphass:es nuclear power control room (similar to LaSalle). Also, it presents generality in its design. Finite-element models can be generated an example of how the code can be ublized by addressing an for disks, plates, cyInders, and even geometnes with holes.

Appendix R exemption request which deals with the vulnerability I

such as compact tenston specimens. Either surface or embed-of containment fans to a fire emanating from a reactor coolant ded flaw geometries can be modeled. Detail user instructsons i

pump bay. Recommendations are also given as to how the desenbe both preparation of input data and program operation.

model/ code can be further enhanced and where current effort Sample problems are presented that demonstrate the flexibility is proceeding.

of the program. ORMGEN.PC executes on an IBM PC/AT or PC/XT microcomputer; typical runtimes on an IBM PC/AT are NUREG/CR-4482 DRF FC: RECOMMENDATIONS TO THE NU.

30 to 45 s.

CLEAR REGULATORY COMMISSION ON TRIAL GUIDELINES NUREG/OR-4476: HIGH TEMPERATURE OXIDATION OF ZlR-FOR SEISMIC MARGIN REVIEWS OF NUCLEAR POWER CALOY 4 IN STEAM AND STEAM-HYDROGEN ENVIRON-PLANTS. Draft Report For Comment. PRASSINOS,P.G.;

MENTS. PRATER.J.T.; COURTRIGHT,E.L Battelle Memonal in.

RAVINDRA.M.K.; SAVY,J.B. Lawrence Livermore National Labo-sbtute, Pacific Northwest Laboratones. February 1986. 34pp.

ratory. March 1986. 115pp. 8603250207. UCID-20579.

8603100557. PNL 5558. 34892:155.

35218.047.

The oxidation kinetics of Zircaloy-4 in steam have been ex.

This report is the third report of the Expert Panel on the tended to 2400 degrees centigrade. The ZrO(2) and alpha-Zr QuanUficadon of Seismic Margins. The objective of this report is l

layers display parabolic growth behavior over the entire temper-to present detailed guidelines for the performance of seismic ature rangs studied A disconhnuity in the oxidation kinetics at margin reviews of nuclear power plants. The guidelines present-1510 degrees centigrade causes rates to increase above those ed in this report are based on the Paners second report entitled previously established by the Baker Just relationship. This in.

"An Approach to the Quanbfication of Seismic Margins in Nu-

)

crease coincides with the tetragonal-to-cubic phase transforma, clear Power Plants." It is intended that these guidelines be used J

tion in ZrO(2)-x. No additional discontinusty in the oxide growth in at least one trial plant review to demonstrate whether the ap-rate was observed when the metal phase melted. The effects of proach and the quantificaton techniques are adequate. Based i

temperature gradients were taken into account, and corrected on lessons learned from these trial reviews, the Panel can then values representative of near-isothermal conditions were com, be more presenptive about defining guidelines for general use, puted. Oxide growth was also measured in various steam-hydro.

NUREG/CR-4485: THE IMPACT OF FUEL CLADDING FAILURE gen mixtures at 1565 degrees centigrade and 1815 degrees j

EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT centigrado. Hydrogen concentrations up to 90 mot % had no effect on oxidation kinetics. The rate-controlling factor appears NUCLEAR POWER PLANTS. Case Study:PWR Dunng Routine to be diffusion tt' rough the oxide layer, Operations. MOELLER.M.P.; MARTIN,G F.; HAGGARD,0.L Bat-telle Memonal Institute, Pacific Northwest Laboratories. January i

NUREG/CR-4477: METHODOLOGIES FOR ASSESSING LONG.

1986.107pp. 8602050297. PNL 5606. 34484.245.

TERM PERFORMANCE OF HIGH-LEVEL RADIOACTIVE This report presort3 data in support of evaluating the impact WASTE PACKAGES.

STEPHENS.K.;

JOHNSON.R.;

of fuel cladding facure events on occupational radiation expo.

ZAREMDA.L; et al. Aerospace Corp. January 1986. 18 t pp.

sure. To determine quantitatively whether fuel cladding failure 8602280741. 34726:102.

contnbutes significantly to occupational radiation exposure, radi-Several methods the Nuclear Regulatory Commission (NRC) a o exposure measurements were taken at comparable loca-can use to independently assess Department of Energy (DOE) ti in two mirror imcae pressttred. water reactors (PWRs) and waste package performance were identfied by The Aerospace th common auxiliary building. One reactor, Unit 8, was experi-Corporaton. The report includes an overview of the necessary encing degraded fuel charactenred as 0.125 percent fuel pin-

Main Citations and Abstracts 21 hole leakers and was operating at approximately 55 percent of verdale, CA tests of these data indicate that they represent the the reactor's licensed maximum core power, while the other re-same population used in calibration of the two models.

actor, Unit A, was operating under normal conditons with less than 0.01 percent fuel pin-hole leakers at 100 percent of the re.

NUREG/CR-4501: MODELING OF VAPOR GENERATION IN actor's licensed maximum core power. Measurements consisted FLASHING FLOW. RIZNIC,J.; ISHit,M. Argonne National Labo-of gamma spectral analyses, radiation exposure rates and air.

ratory. December 1985. 38pp. 8602260097, ANL-86-0.

bome radionuclide concentrations. In addition, data from pn-34702:332.

mary coolant sample results for the previous 20 months on both A phenomenon of flashing related to discharging of an initially reactor coolant systems were analyzed. The resu!h of the subcooled liquid from a high pressure cond bon into a low pres-measurements and coolant sample analyses suggest that a sure environment is very important in several industrial systems 3560-megawatt-thermal (1100 MWe) PWR operating et full such as nuclear reactors and chemical reactors. A new model power with 0.125 percent failed fuel can expenence an increase for the flashing process is proposed here based on the wali nu-of 540 percent in radiation exposure rates as compared to a cleation theory, bubble growth model and dnft-flux bubble trans-PWR operating with normal fuel. In specific plant areas, the de-port model. In order to calculate the bubble number density, the graded fuel may elevate radiation levels even more.

bubble number transport equation with a distributed source from the wall nucleation sites is used. The model predictions in terms NUREG/CR-4486: VISA II-A COMPUTER CODE FOR PREDICT-of the void fraction are compared to Reocreux's and BNL ex-ING THE PROBABILITY OF REACTOR PRESSURE VESSEL perimental data. It shows that satisfactory agreements could be FAILURE. SIMONEN,F.A.; JOHNSON,K.L; LIEBETRAU A.M.; et obtained from the present model without any floating parameter al. Battelle Memonal Institute, Pacific Northwest Laboratories.

to be adjusted with data. This ressult indicates that, at least for March 1986. 32pp. 8604030417. PNL-5775. 35402:001, the experimental conditions considered here, the mechanistic The VISA-II (Vessel Integrity Simulation Analysis) code was prediction of the flashing phenomenon is possible based on the onginally developed as part of the NRC staff evaluation of pres-present wall nucleation based model.

sunzed thermal shock. VISA-Il uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor NUREG/CR-4502:

VIRGINIA REGIONAL SEISMIC (PWR) vessel subjected to a pressure and thermal transient NETWORK. FINAL REPORT (1977 1985). BOLLINGER,G.A.;

specified by the user. Linear elastic fracture mechanics is used SNOKE,J.A.; SIBOL,M.S; et al. Virginia Polytechnic Institute &

to model crack initiation and growth. Parameters for crack size State Univ., Blacksburg, VA. February 1986.68pp.8602280757, and location, copper content, initial reference temperature of 34754:130.

the nell-ductility transition, fluence, crack-initiation fracture tough-Eight years of monstonng with a 20-station regional network ness, and arrest fracture toughness are treated as random vari-has produced epicenters (M less than or equal to 4), focal ables. This report documents an upgraded version of the origi-depths and mechanisms of adequate number and quality to nal VISA code as desenbed in NUREG/CR 3384. Improvements reveal considerable differences between the two seismically include a treatment of cladding effects, simulation of flaw size, active portions of Virginia. Those two areas (southwestern shape and location, a simulation of inservice inspection, and up-(Giles County) and central parts of the state) are separated by dated simulation of the reference temperature of the nill-ductility only some 200 km. Despite their proximity, the two zones exhib-transition, and treatment of vessels with multiple welds and ine-it remarkable differences in geometncal/ mechanical characteris-tial flaws. The code has been extensively tested and venfied tics, in Giles County, seismic energy is released by predomi-and is wntten in FORTRAN for ease of installation on different nantly stnke-slip faulting in a near vertical, tabular zone ( 40 km computers.

hng) that is below the Appalachian decollement in central Vir-NUREG/CR-4494: RADIOLOGICAL ASSESSMENT OF BWR RE.

ginia, the seismicity is derived from mixed dip-slip and stnke-stip CIRCULATORY PIPE REPLACEMENT. PAFiKhURST M.A.;

faulting in a large, coin shaped volume ( 100 km diameter; 10 HADLOCK,0 E.; HAATY,R.; et al. Battelle Memonal Institute, km vertical thickness), above the major detachment faulting.

Stress estirr a, as derived from single-and composite-focal Pacific Northwest Laboratones. February 1986. 111pp.

8603140263 PNL-5742. 34961:288.

mechanism tions P-axes, are NE to ENE in Giles County Replacement of pnmary recirculating coolant pipe in BWRs is and NW to NL in central Virginia. The causes for the observed a major effort that has teen camed out at a number of nuclear variability ere unknown. The two zones are in different tectonos-generating stations. This report reviews the planned or actual tratigraphic (suspect) terranes and that difference could be rele-pipe replacement projects at six sites: Nine Mile Point-1, Monti-vant. The recently proposed Hydroseismicity model (Costain cello, Cooper Peach Bottom-2, Vermont Yankee. and Browns and Bollinger,1985) asenbes the observed seismicity variations Ferry 1. It covers the radiological issues of the pipe replace, in Virginia and throughout the Southeast to different drainage ment, measures taken to reduce doses to ALARA, estimated basin hydrologic characteristics plus differences in upper crustal and actual occupational doses, and lessons learned dunng the fractunng.

vanous replacemei ts. The basis for the decisions to replace the NURF.G/CR-4504: LONG-TERM SURVEILLANCE AND MONI-pipes, the methods used for preparaton and decontamination, TORING OF DECOMMISSIONED URANIUM PROCESSING the removal of old pipe, and the installation of the new pipe are SITES AND TAILINGS PILES. YOUNG J.A.; CADWELL.L.L.;

bnefly descnbed. Methods for reducing occupational radiation FREEMAN,H.D.; et al. Battelle Mamonal Institute, Pacific Nutth-dose dunng pipe repairs / replacements are recommended.

west Laboratones. March 1986. 50pp. 8604040460. PNL-5755.

NUREG/CR-4496: A SYSTEM FOR GENERATING LONG 35405.001.

STREAMFLOW RECORDS FOR STUDY OF FLOODS OF Natural processes and human activities could expose the LONG RETURN PERIOD. LINSLEY R.K : KRAEGER,8.A.;

public to radioactive and nonradioactive toxic materials from FRANZ.D D. Linsley, Kraeger & Associates, Ltd February 1986.

uranium processing sites in the years following decommission-10 tpp. 8603180342. 35138.070.

ing. This report desenbes secunty, surveillance, and monitoring A stochastic model for the multi-station generation of hourly methods that can be used to prevent or detect the spread of rainfall has been developed and tested. The model uses a mul-these materials and to determine when cleanup or preventive tivariate-normal background process which is continuous. The maintenance is required. Visual observations carried out at least process is truncated to yield intermittent output as rainfait. The annually can be used to detect any rapsdfy developing condi-model results showed excellent agreement between a 1000-yr tions that could expose the public to toxic matenals. If no such senes and observed rainfall. The synthetic rainfall was used cond tions develop dunng the first several years following de-eth a hydrologic simutation model to develop a synthetic flow commissioning, then visual observations can be made at less record of 1000 yrs. For a portion of the Russian River near Clo-trequent intervals to detect more gradual changes. Gamma-radi-

22 Main Citations and Abstracts ation,226(Ra;, and 238(U) measurements at locations showing The reliability of NDE techniques is reviewed based on experi-i significant deterioration can be used to determine whether re-ence through the end of 1983. Acceptability of proposed repair sidual radcactive matenals that exceed exir i standards have techniques for cracked BWR piping and of mitigating measures been exposed. Measurements of contamini oncentrations in is discussed, including the proposed weld overlay technique, en-standing surface water and groundwater can utect the spread vironmental mitigahng measures, and other stress related reme-of water containing elevated contaminant concentrabons. Dura-dial measures.

ble signs and stone markers can be used to warn of the possi-ble dangers associated with these sites.

NUREG/CR-4546: LABOR PRODUCTMTY ADJUSTMENT FACTORS.A Method For Es mahng Labor Construction Costs o

NUREG/CR-4509: WASTE PACKAGE RELIABILITY. SASTRE,C.;

Associated With Physical Modificanons To Nuclear Power PESCATORE.C.; SULLIVAN.T, Brookhaven National Laboratory.

Plants. RIORDAN.B.J. Mathtech, Inc.

• Science & Engineering February 1986. 100pp. 8603140483. BNL-NUREG-51953.

Associates, Inc. March 1986. 42pp. 8603280264. 35324:150.

34964:298.

NRC regulatory impact analyses (RIA's) address the costs 3

Probabilistc Reliability Analysis is identified as the preferred and benefits associated with proposed fegulatory requirements.

method to identify, organize, and convey the necessary informa-Many of these requirements will result in physical modification tion to meet the NRC standard on reasonable assurance of to ex sting structures and systems at nuclear power plants. This waste packags performance according to regulatory require-report develops quanttative labor productivity adjustment fac-ments. The document addresses both the qualitative and quan-tors for the performance of RIA's. These factors will allow ana-titative aspects of the analysis, and suggests reliability analysis lysts to modity "new construction" labor costs to account for I

requirements by a pros, tctive license applicant as well as changes in labor productivity due to diffenng work environments review procedures by the regulatory agency. In p&rticular, a at operating reactors and at reactors with construction in method for the quanttative evaluation of a waste package relt-progress. The technique developed in this paper rehes on the ability is, demonstrated through a simphfied analysis. The Energy Economic Data Base (EEDB) for baseline estimates of method is based on the repetitive usage of a performance the direct labor hours and/or labor costs required to perform model for values of the model parameters that span their range s ecific tasks in a new construction environment. The labor pro-of uncertainty. Techniques for selecting values of the input pa-ductivity cost factors adjust for constraining conditions such as rameters, viewed as random variables, and for generating em-pincal correlatior's among expenmental data are also desenbed.

working in a radiation environment, poor access, congestion Ap which would eed to be covered in a more comprehen-and interference, etc., which typically occur on construction tasks at operating reactors and can occur under certain circum-stances at reactors under construction. While the results do not NUREG/CR-4514: CONTROLLING PRINCIPLES FOR PRIOR portray all aspects of fabor productnnty, they encompass the PROBABILITY ASSIGNMENTS IN NUCLEAR RISK ASSESS-major work place conditions generally discernable by the NRC MENT. COOK,f.; UNWIN,S D. Sandia National Laboratories.

analysts and assign values that appear to be reasonable within February 1986. 42pp. 8604020713. SAND 85-1323. 35374:225-the context of industry experience.

As performed conventionally, nuclear probabilistic nsk assess-ment (PRA) may be enticized as utilizing inscrutable and unjusti.

NUREG/CR 4547: CONTEMPT 4/ MOD 6:A MULTICOMPONENT fiably " precise" quantitative informed judgment or extrapolaton SYSTEM ANALYSIS PROGRAM. LIN,C.; ECONOMOS,C.;

from that judgment. To meet this enticism, we propose and LEHNER.J.R.; et al. Brookhaven National Laboratoy. March argue for controlling pnnciples that govern the formulation of 1986. 304pr. 8604040473. BNL.NUREG-51966. 35408.281.

probability densities given only the informed input that would be CONTEMPT 4/ MOD 6 is a digital computer program that de-required for a simple bounding analysis. These principles are scnbes the response of multicompartment containment systems founded urn information theoretic ideas of maximum uncer-subjected to postulated loss-of-coolant accident (LOCA) condi-tainty and

'er both cases in which there exists a stochastic tions. The program is written in FORTRAN IV and can accom-model of t... phenomenon of interest and cases in which there modate both pressunzed water reactor (PWR) and boiling water is no such model. In part, the pnnciples are conventional and reactor (BWR) containment systems, including those with inert-we justify such an approach by appeal to certain analogies in ed atmospheres. Also, both design basis accident (DBA) and accounting practice and judicial decision-making. Examples are degreded core type LOCA conditions can be analyzed. The pro.

given. We believe that, by appropriate employment of these gram calculates the time variation of compartment pressures, principles, substantial progress towards PRA scrutability and temperatures and mass and energy inventories due to intercom-transparr7 would be achieved.

partment mass and energy exchange, LOCA source terms, con.

tainment fans and pumps, cooling sprays, heat conducting NUREG/CR-4520: DEVELOPMENT AND VERIFICATION OF CONDITIONS FOR DUCTILE TEARING INSTABILITY AND structures, sump drair,s, PWR ice condensers, BWR pressure ARREST. JOYCE,J.A. U.S. Naval Academy, Annapolis, MD.

suppressen systems, hydrogen and carbon monoride combus-February 1986. 31pp. 8603100581. 34892:188.

ten within compartments and energy transfer due to gas radi-The objective of this work was taken to an in depth look at ation. Dynamic storage allocation (DBA) is used to hmit the the process of ductile tearing instability and especially to evalu-amount of computer core used for each problem and to provide ate experimentally the conditons for arrest of a ductile tearing the multicompartment capability of up to 999 flow when numeri-instability. A secondary objective was to evaluate the sensitivity cally induced flow oscillatons are encountered. This capabil.ty of a ductile teasing instability arrest to rate at which it occurred proydes significant reduction of computer run time relative, to and to the material rate sensitivity. Major conclusions are that previous codes in the CONTEMPT senes.

the ductile teanng instability initiates slowly but in a mechanical NUREG/CR 4555: GENERIC COST ESTIMATES FOR THE DIS-spnng apparatus it approaches drop tower growth rate condi-POSAL OF RAD lOACTIVE WASTES.

SCIACCA,F

• tions and hence the phenomena is affected by the material rate sensstavity. The conditions r.ecessary for arrest are completely SHAFFER,C.; SIMPKINS.B.; et al. Science & Engineering Asso ciates, Inc. March 1986.194pp. 8603280272. 35326:265.

s ed and de strated by experimental data on a 3 percent NRC regulatory impact analyses address the costs and bene-fits associated with proposed regulatory requirements. Many of NUREG/CR-4545: PIPE CRACK EVALUATION IN OPERATING these requirements will result in physical modifications to exist-BOILING WATER REACTORS.

AUERBACH,C.;

ing structures and systems at nuclear power plants. This report CZAJKOWSKlC.J.; SANBORN,Y.; et al. Brookhaven Natonal provides a methodology and data needed to estimate the ge-Laboratory. March 1986. 61pp. 8604030268. BNL NUREG-neric costs of disposing of radioactive wastes that may be gen-

$1965. 35401123.

ersted as a result of NRC regulatons requiring modifications or

Main Citations and Abstracts 23 repairs to nuclear facilities. Also presented are descriptions of By priontizing the various areas of possible interest for in-typical low-level redwastes generated at nuclear power plants spection and by better defining inspection needs, the NRC ex-and the various processes used to treat the wastes in prepara-pects to make more effective use of finite inspection resources tion for shipment and burial The waste disposal cost estimates by concentrahng on those potential areas most segrwficant to included in this report cover all of the major elements that con-safety. Through review and application of the Indian Point Unit 3 tr;bute to the overall costs. The k9y factors that influence the Probabilistic Safety Study's numerical data and event tree mod-costs are discussed. Pertinent ranges of values for the key vari-eling, and by utilizing related documents, a technical basis for ables are explored and important sensitivities identified. Occu-priontizing areas for NRC inspection has been developed. This patonal radiation exposure associated with in-plant handling of was then tested at the plant site for the NRC Operating Reactor r

the wastes is also discussed.

Inspection Program, l&E Manual Chapter 2515. Inspecten ac-NUREG/CR-4564: HEAT TRANSFER FROM A ROD BUNDLE tivmes ad&essed include nonnel opershons, syskm and e UNDER NATURAL CIRCULATION CONDITIONS.

ponent testing, maentenance and surveillance. A computer pro-HALLINAN K.P.; VISKANTA.R. Purdue Univ., West Lafayette, gram entitled NSPKTR, which was developed specifically for this program, modeled me,nternal plant staks to me systern i

IN. March 1986. 74pp. 8604070241, 35447:345.

)

Steady-state and transient expenments were performed in a level and performed the risk and importance calculatons.

rectangular natural circulaton loop. Based on the steady-state NUREG/CR-4545: INVESTIGATION OF THE STABILITY OF data obtained, empincal correlations for fiuid friction and heat CLAY / BASALT PACKING MATERIALS.

PEACOR,0.R.;

transfer of the circulating fluid flowing through the tube bundle ESSENE.E.J.; LEE J.H. Michigan, Univ. of, Ann Arbor, ML March j

were developed. The pressure drop in the loop was found to 1986.134pp. 8604040075. 35414:074.

l depend on the Reynolds number. Friction factor relations for Geological systems analogous to proposed packing materials larninar forced flow through tube bundles were found to accu-have been reviewed and investigated and approximate tempera-rately model fluid fncton of the circulating fluid through the test ture hmits for smectite, mixed layer smectite / layer smectite / illite bur' dies. Empirical correlatons for the average Nusselt number and illite have been found to be 100, 200, and 250 degrees were developed for both parraffel-flow and counter-flow arrange.

centigrade, respectively. All phases are metastable, with reac-ments of the test bundle was found to have little effect on the tion rates being controlled by factors influencing kinetics, princ6-total flow resistance of the circulating fluid, while enhancing the pally temperature and rock / water rato. Even over geological average heat transfer from 5% to 15%, depending on the ther-time periods the phases may exist metastably at low tempera-j mal and flow conditions. The dynamic response of the circulat-tures with low water / rock ratios. Experiments with bentonite /

ing fiuid and of the loop structural components was predicted basalt mixtures confirm that there is little change at 300 de-i from a one-dimensonal model. Good correspondence was ob-grees centigrade and below in simulated packing conditions for j

tained between the predicted and measured local temperatures ex erimental durations of up to a year and a half. Data are con-and the time to reach steady-state conditions.

sistent with a lack of equillibrium, such that experimental reac-1 NUREG/CR-4565: PROBABILISTIC SAFETY STUDY APPLICA-tions occur at much higher temperatures than in geologic sys-1 TIONS PROGRAM FOR INSPECTION OF INDIAN POINT UNIT tems, and cannot be used to reliably predict the state of such l

3 NUCLEAR POWER PLANT. TAYLOR,J.H.; FULLWOOD.R.;

systems over long time penods. Analysis of geologic systems l

FRESCO,A. Brookhaven National Laboratory. March 1986.

suggests that a suitable continuous clay matrix can minimize the 73pp.8604040089. BNL-NUREG-51973. 35411:327.

degree of transition of clays and clay / basalt mixtures.

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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 Systern acces-contractor reports in this cornpilation. Each sion number.

contractor code is cross-referenced to the i

SECONDARY REPORT NUM8ER REPORT NUMBER SECONOARY REPORT NUM8ER REPORT NUM8ER 17 5023 NUREG/CR-4078 ORNL/ TM-9567/V1 NUREG/CR-4183 V01 ANL 85-42 NUREG/CR4348 V01 ORNL/TM-9567/V2 NUREG/CR-4183 V02 ANL 85-42 NUREG/CR-4348 V02 ORNL/TM-9570/V1 NUREG/CR-4188 V01 ANL-85 47 NUREG/CR4371 ORNL/TM-9570/V2 NUREG/CR-4188 V02 g Q g [g g g / A g V02

/V2 ANL-8571 V01 NUREG/CR-4453 VO, ANL-864 NUREG/CR4501 ORNL/TM-9632/V2 NUREG/CR4255 V02 BMI-2120 NUREG/CR-4082 V03 ORNL/TM-9719 NUREG/CR4059 BNL NUREG-51454 NUREG/CR-2331 VOS N2 ORNL/TM-9773 NUREG/CR4389 BNL NUREG-51581 NUREG/CR-2907 V03 ORNL/TM-9798/V2 NUREG/CR-4402 V02 BNL-NUREG-51630 NUREG/CR-3091 V07 PNL-4758 NUREG/CR 3365 DRF FC BNL-NUREG-51643 NUREG/CR 3137 PNL 4865 NUREG/CR-3517 BNL-NUREG 51699 NUREG/CR-3444 V03 PNL 5210 NUREG/CR 3950 V02 BNL-NUREG-51900 NUREG/CR-4293 PNL 5429 NUREGICR4289 BNL-NUREG 51902 NUREG/CR-4311 PNL 5477 V02 NUREG/CR-4276 V02 BNL-NUREG-51905 NUREG/CR-4328 PNL 5479 NUREG/CR 4279 V01 l

BNL-NUREG-51912 NUREG/CR4364 PNL 5511 NUREG/CR4300 V02 4

BNL NUREG-51913 NUREG/CR4366 PNL 5520 NUREG/CR4323 BNL-NUREG-51914 NUREG/CR4372 PNL 5558 NUREG/CR4476 BNL-NUREG 51918 NUHtG/CR4381 PNL 5576 NUREG/CR-4378 BNL NUREG-51919 NUREG/CR4359 PNL 5602 NUREG/CR4411 BNL-NUREG-51936 NUREG/CR4433 PNL.5606 NUREG/CR4485 BNL-NUREG-51937 NUREG/CR-4434 PNL-5641 NU9EG/CR 4436 V02 g5 v01 BNL-NUREG-51944 NUREG/CR4450 DRF FC C

BNL-NUREG-51946 NUREG/CR4452 PNL 5690 NUREG/CR4462 BNL-NUREG-51948 NUREG/CR4479 PNL 5705 NUREG/CR4464 BNL NUREG-51953 NUREG/CR4509 PNL.5711 NUREG/CR-4469 V01 BNL-NUREG-51965 NUREG/CR4545 pNL.5717 NUREG/CR-4472 BNL-NUREG-51966 NUREG/CR-4547 PNL.5742 NUREGICR4494 BNL NUREG-51973 NUREG/CR4565 PNL 5755 NUAEG/CR4504 EGG-2401 NUREG/CR4299 PNL-5767 NUREG/CR-3958 EGG-2419 NUREG/CR4393 PNL 5769 NUREG/CR-3959 EGG-2424 NUREG/CR-4438 PNL 5775 NUREG/CR4486 EGG-2432 NUREG/CR4466 SAND 83-1057 NUREG/CR 3983 HEDL TME 8514 NUREG/CR4307 V01 SAND 850442 NUREG/CR4171 IEAL-R/8570 NUREG/CR4446 SAND 85-0707 NUREG/CR4420 IEB-79-07 NUREG/CR-3790 SAND 85-1323 NUREG/CR-4514 1S-4554 NUREG/CR4223 SAND 851481 NUREG/CR4324 LA 10590 M NUREG/CR4442 SAND 851594 NUREG/CR-4337 LA-10624 MS NUREG/CR4474 SAND 851639 NUREG/CR4343 LA-10637 MS NUREG/CR4471 SANO85-1774 NUREG/CR-4369 hh h

ORNL-6193/V1 NUREG/CR-4302 V01 S4 ORNL-6226 NUREG/CR-4380 SAND 852500 NUREG/CR4459 ORNL-6250 NUREG/CR-4475 SAND 852563 NUREG/CR 4465 ORNL/CSD/TM-233 NUREG/CR4468 SAND 857247 NUREG/CR4310 ORNL/NSIC 200 NUREG/CR-2000 V04N12 UCID 20549 NUREG/CR4431 ORNL/NSIC-200 NUREG/CR 2000 VOS N1 UCID 20579 NUREG/CRJ482 ORF FC ORNL/NSIC 200 NUREG/CR-2000 V05 N2 UCRL-53532 NUREG/CR 3760 ORNL/TM-6183 NUREG/CR-4286 UCAL 53662 NUREG/CR4363 I

i.

25 i

i

'W._.

--4A._

-~_

Personal Author Index This index lists the personal authors of NRC report (s) prepared by the author. If informa-staff and contractor reports in al3habetical tion is needed, refer to the main citation by order. Each name is followec by the the NUREG number.

NUREG number and the title of the l

ASEL,K.H.

SALL.D.G.

NUREG/CR-4289: RESIDUAL RADONUCUDE CONTAMINATON NUREG/CR-4183 V31: PRESSURIZED THERMAL SHOCK EVALUATON WITHIN AND AROUND COMMERCIAL NUCLEAR POWER OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

PLANTS. ORIGIN.DISTRIBUTON, INVENTORY AND DECOMMISSION-NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON ING ASSESSMENT.

OF THE H.B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR-4468: ADAPTATION OF OCA.P.A PROBABluSTIC FRAC-ASRAHAM,T.

TURE-MECHANICS CODE,TO A PERSONAL COMPUTER.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATON TESTS.Semannual Report Covenng The Penod Apni-September 1985.

BALL SJ.

NUREG/CR4402 V02. HIGH-TEMPERATURE GAS-COOLED REACTOR ADAMS.J.W.

SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS EVALUATON.Ouarterty Progress Report Apri t. June 30,1985.

ON SOUDIFICATON. WASTE DISPOSAL,AND ASSOCIATED OCCU-PATIONAL EXPOSURE.

8ANOER,TJ.

NUREG/CR4000 V01: THE MESORAD DOSE ASSESSMENT NURE CR-4255 V02: AEROSOL RELEASE AND TRANSPORT PRO-GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL SEPTEMBER BARNES,C.R.

1985.

NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM PHASE O'**"""*'

PO I'00

• b'E' AHMAD J.

NUREG/CR4082 V03: DEGRADED PIPING PROGRAM PHASE BASSB.R.

ft.Semannual Report.Apnl 1985. September 1985.

NUREG/CR4475; ORMGEN PC.A MICROCOMPUTER PROGRAM FOR AUTOMATIC MESH GENERATION OF 2 0 CRACK GEOMETRIES.

ANDERSON C.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION BATES,0.J TESTSSemannual Report Covenn0 he Penod Aprd-September 1985.

NUREG/CR4464; PERFORMANCE DEMONSTRATON TESTS FOR T

DETECTON OF INTERGRANULAR STRESS CORROSION CRACK.

ANDREWS,W.B.

^

N R G/CR-4489 V01: INTEGRATON OF NONDESTRUCTIVE EXAMI-A E L Y ALTERNAT V S-NATON REUABluTY AND FRACTURE MECHANICS. Semiannual ARAVE.A.E.

Report Apnl. September 1984.

NUREG/CR4299: PREUMINARY EVALUATION OF EFFLUENT RADIO-EA ACTIVITY MONITORING SYSTEMS FOR BWR PLANTS NUR G[CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF ARMENTO WJ.

U.S AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE.

NUREG/CR-4059: EVALUATION OF THE IMPACT OF THE MC&A REFORM AMENDMENTS ON A REPROCESSING FACIUTY.

U NURE 0068, PROCEEDINGS OF THE INTERNATONAL NUCLEAR ATHEY,0.F.

REACTOR DECOMMISSIONING PLANN!NG CONFERENCE.

NUREG/CR-4000 V01: THE MESO 9AD DOSE ASSESSMENT MODELVolume 1: Technical Bass.

SEAHM C A

p AUERB ACH C.

CLEAR REACTOR ACCIDENTS.

NUREG/CR4545: PIPE CRACK EVALUATION IN OPERATING BOluNG WATER REACTORS.

BEEBE.M.R.

NUREG-0020 V09 N12: UCENSED OPERATING REACTORS STATUS AUSTIN,P.N.

SUMMARY

REPORT. Data As Of November 30,1985(Gray Book t)

NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NUREG-0020 V10 N01: UCENSED OPERATING REACTORS STATUS OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

SUMMARY

REPORT. Data As Of December 31,1985 (Gray Book l).

NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

BEEKMAN.D.W.

NUREG/CR4419: BIOASSAY MEASUREMENTS FOR URANIUM USING 9ADALAMENTE.R.

SPUTTER INITIATED RESONANCE ONIZATON SPECTROSCOPY.

NUREG/CR-3517. RECOMMENDATONS TO THE NRC ON HUMAN EN-GINEERING GUIDEUNES FOR NUCLEAR POWER PLANT MAIN.

BENNETT,J.G.

TAINABluTY.

NUREG/CR4474. SCALE MODEUNG OF REINFORCED CONCRETE NUREG/CR 4436 V01: HUMAN REUABluTY IMPACT ON INSERVICE CATEGORY l STRUCTURES SUBJECTED TO SEISMIC LOADING.

INSPECTION Volume 1: Phase 1 Summary Report.

NUREG/CR.4438 V02. HUMAN RELIABiUTY IMPACT ON INSERVICE DERGERON K.D.

INSPECTON Volume 2: Revow And Ana*yss Of Human Performance NUREG/CR-4343. INTEGRATED SEVERE ACCIDENT CONTAINMENT in Nondestructive Testing (Emphasmng Ultrasorwcs).

ANALYSIS WITH THE CONTAIN COMPUTER CODE.

BAILEY,W.J.

BERMAN,M.

NUREG/CR-3950 V02. FUEL PERFORMANCE ANNUAL REPORT FOR NUREG/CR-4459: UGHT WATER REACTOR SAFETY RESEARCH 1984 PROGRAM Semannual Report October 1983. March 1984.

SALDT.K.R.

BETHK E,0.W.

NUREG/CR.4390- DCC 1/DCC 2 DEGRADED CORE COOLABruTY NUREG/CR-3365 DRF FC: REPORT TO THE NRC ON GUIDANCE FOR ANALYSIS.

PREPARING SCENARIOS FOR EMERGENCf PREPAREDNESS EX-27

n 9

28 Personal Author Index ERCISES AT NUCLEAR GENERATING STATIONS. Draft Report For CAIN.C.L Comment NUREG-1179 V01: RUPTURE OF MODEL 48Y UF8 CYLINDER AND RE-LEASE OF URANIUM HEXAFLUOROE.Sequoyah Fuels a@,heWahornaJanuay W8&

NA C - 958. EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS.ACCOENTS AND CORE-MELT FREQUENCIES AT A CARRICK,LC.

COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR.

NUREG/CR-4289: RESOUAL RADIONUCLOE CONTAMINATON BLAHNIK,D.E.

WITHIN AND AROUND COMMERCIAL NUCLEAR POWER NUREG/CR-3517. RECOMMENDATIONS TO THE NRC ON HUMAN EN.

PLANTS. ORIGIN. DISTRIBUTION, INVENTORY AND DECOMMISSION.

GINEERING GUIDELINES FOR NUCLEAR POWER PLANT MAIN.

ING ASSESSMENT.

TAINA8tLITY.

CARTER,R.J.

BLUHM D.

NUREG/CR4188 V01: NUCLEAR POWER PLANT SIMULATON FACILI-NUREG/CR4223: STEEL CONTAINMENI RESISTANCE UNDER GEN-TY EVALUATON METHODOLOGY. Handbook.

ERAL DYNAMIC PRESSURES.

NUREG/CR4188 V02: NUCLEAR POWER PLANT SIMULATION FACILI.

TY EVALUATON METHODOLOGY.Techmcal Bases.

NUREG/CR4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE.

CHANG,C.Y.

HAVIOR IN COMPLEX NUCLEAR POWER PLANT NUREG/CR-3805 V01 ENGINEERING CHARACTERIZATION OF ENCLOSURES.PRESENT CAPABILITIES AND FUTURE PROSPECTS GROUND MOTON Task It: Observanonal Data On Spahal Vanahons SOESCH L Of Earthquake Ground Mohort NUREG/CR 4477: METHODOLOG:ES FOR ASSESSING LONG-TERM PERF RMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK.

CHAPMAN CALE FACILITY, DOLLINGER,0.A.

NUREG/CR4502: VIRGINIA REGIONAL SEISMIC NETWORK. FINAL CHAPMAN,M.C.

REPORT (19771985).

NUREG/CR-4502: VIRGINIA REGIONAL SEISMIC NETWORK. FINAL SOWERMAN.B.S.

NUREG/CR4433: DOCUMENT REVIEW REGARDING HAZARDOUS CHAPPELLR.

NUREG 1179 V01: RUPTURE OF MODEL 48Y UF8 CYLINDER AND RE.

N RE /C 450 D 1A E T F D A TIVE MIXED LEASE OF URANIUM HEXAFLUORIDE.Sequoyah Fuels WASTES IN COMMERCIAL LOW-LEVEL WASTES Draft Report For Co w Facility, Gore, Oklahoma. January 4,1988.

BOYACK.8.E.

CHARLOT,LA-NUREG/CR4442 TRAC USER'S GUCE.

NUREG/CR-4489 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-NUREG/CR4471: LOS ALAMOS PWR DECAY HEAT-REMOVAL STUD-NATION RELIABILITY AND FRACTURE MECHANICS Semiannual IES

SUMMARY

RESULTS AND CONCLUSIONS.

Report,Aprd. September 1984.

80ZARTH,D.P.

CHEN,P.C.

NUREG/CR4183 V01; PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR 3805 Vol ENGINEERING CHARACTERIZATON OF OUND MOTOUask R Obsenrahonal Data On Spabal vananons NIREG R418 VO All HERM OC EV LUATON OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

Of Earthquake Ground Motion.

8HADLEY,D.R.

CHEVERTON.R.D.

NUREG/CR4420. TURC1 LARGE SCALE METALLIC MELT CONCEN.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON TRATE INTERACTION EXPERIMENTS AND ANALYSIS OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON BROCKMANN.J.E.

OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR4420 TURCt LARGE SCALE METALLIC MELT CONCEN-NUREG/CR4468: ADAPTATON OF OCA-P,A PROBABluSTIC FRAC.

TRATE INTERACTION EXPERIMENTS AND ANALYSIS TURE MECHANICS CODE.TO A PERSONAL COMPUTER.

BRODSKY,A-CHRISTENSEN,J.

NUREG 1159 TRAINING MANUAL FOR URANIUM MILL WORKERS ON NUREG/CR 3959. TRANSITON TO AN OPERATING REACTOR ENVI.

HEALTH PROTECTON FROM URAN 1UM RONMENT lMPLICATIONS FOR NRC OUALITY ASSURANCE PRO-OROEK D.

GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-PHASE TORY PROJECTIONS THROUGH 1995.

NUREG/CR4082 V03: DEGRADED PIPING PROGRAM Il Sermannual Report.Aprd 1985. September 1985.

CLEVELAND J.C.

DRUST,F, NUREG/CR4402 V02: HIGH TEMPERATURE GAS. COOLED REACTOR NUREG/CR4082 V03. DEGRADED PIPING PROGRAM PHASE SAFETY STUDIES FOR THE DIVISON OF ACCOENT ll. Semiannual Report /pr41985. September 1985 EVALUATON Ouarterly Progress Report, Apre 1 June 30,1985 DRYSON.J.W.

COMEN,L NUREG/CR4475. ORMGEN PC A MICROCOMPUTER PROGRAM FOR NUREG-0837 V05 N01 NRC TLD DIRECT RADIATION MONITORING AUTOMATIC MESH GENERATION OF 2-0 CRACK GEOMETRIES NETWORK.Progresa Report July September 1985.

SUSH S.H.

COHEN,S.

NUREG/CR4279 V01: AGING AND SERVICE WEAR OF HYDAAULIC AND MECHANICAL SNUBBERS USED ON SAFETY RELATED PIPING NUREG/CR.45n GENERIC COST ESilMATES FOR THE DISPOSAL OF RADIOACTINE WASTES' AND COMPONENTS OF NUCLEAR POWER PLANTS Phase 1 Study.

BYERS.K.R.

COLLINS.H.D.

NUREG/CR4473 A STUDY OF THE OPERATION AND MAINTENANCE NUREG/CR4489 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-l OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 NATION RELIABILITY AND FRACTURE MECHA:4tCS Sermannual CFR 73 55.

Report.Apnl. Sectember 1964.

NUREG/CR4472: SIAMESE IMAGING TECHNIOUE FOR QUASI-VERTI.

CADWELL,LL CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

NUREG/CR4504 LONG-TERM SURVEILLANCE AND MONITORING OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL.

COLMAR,R.

INGS PILES.

NUREG 0933 SO4. A PRIORITIZATON OF GENERIC SAFETY ISSUES.

Personal Author index 29 COOu,L DOVE.R.C.

NUREG/CR4514: CONTROLUNG PRINCIPLES FOR PROR PROBA-NUREG/CR-4474: SCALE MODEUNG OF REbNFORCED CONCRETE BluTY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

CATEGORY I STRUCTURES SUBJECTED TO SEISM'C LOADING.

COTTRELL,W.D.

DUNENFELD,M.S.

NUREG/CR4286: EVALUATON OF RADIOACTIVE UQUO EFFLUENT NUREG/CR-3950 V02: FUEL PERFORMANCE ANNUAL REPORT FOR RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

1984.

COURTR8GHT,E.L ECONOt00S.C.

NUREG/CR4476: HIGH-TEMPERATURE OXOATON OF ZlRCALOY4 NUREG/CR4547: CONTEMPT 4/MOC8.A MULTICOMPONENT PSTEM IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

ANAL SIS PROGRAM.

t CRANE.B.

EDSEN.J.L NUREG/CR4477: METHODOLOGIES FOR ASSESSING LONG-TERM NUREG/CR4299: PRELIMINARY EVALUATION OF EFFLUENT RADIO-PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK.

ACTIVITY MONITORING SYSTEMS FOR bwR PLANTS.

AGES.

EDWARDS,K.M.

CROWLEVAL NUREG/CP-0068: PROCEED:NGS OF THE INTERNATONAL NUCLEAR 5

NUREG/CR4380. EVALUATION OF THE MOTOR-OPERATED VALVE REACTOR DECOMM:SSIONING PLANNnNG CONFERENCE.

ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT DEGRADATIONINCORRECT ADJUSTMENTS.AND OTHER ABNOR.

EHINGER M.H.

i MAUTIES IN MOTOR-OPERATED VALVES.

NUREG/CR4059. EVALUATON OF THE IMFACT OF THE MC&A REFORM AMENDMENTS ON A REP 90 CESSING FACluTY.

NUREG/CR-4327: ORGANIC ODOE FOPMATION FOLLOWING NU.

EISSENGERG,D.M.

t CLEAR REACTOR ACCIDENTS.

NUREG/CR4302 V01: AGING AND SERVICE WEAR OF CHECK J

VALVES USED IN ENGINEERED SAMTY-FEATURE SYSTEMS OF CUMMINGS.G.E-NUCLEAR POWER PLANTS.

NUREG/CR-4431:

SUMMARY

REPORT ON THE SEISMIC SAFETY NUREG/CR-4380: EVALUATION OF THE MCTOR. OPERATED VALVE j

MARGINS RESEARCH PROGRAM.

ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT DEGRADATION. INCORRECT ADJUSTMENTS.ANO OTHER ABNOR-NURE CR.3137: SEISMIC AND DYNAMIC QUAUFICATION OF RE-LATED ELECTRICAL AND MECHANICAL EQUIPMENT.

EKLUND.J.D.

i NUREG/CR-3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN-NUR G CR 4545; PIPE CRACK EVALUATION IN OPERATING BOILING N 8U WATER REACTORS.

ELLINOWOOD.B.

DALING.P.M.

NUREG/CR4328 PROBABIUTY BASED LOAD COMBINATION CRITE-

{

NUREG/CR-4462. A RANKING OF SABOTAGE / TAMPER!NG AVOID-RlA FOR DESIGN OF SHEAR WALL STRUCTURES.

j ANCE TECHNOLOGY ALTERNATIVES.

EMBREY,D.

i DANDINI,V.J.

NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON l

NUREG/CR4324. TESTING OF NUCLEAR QUALIFIED CABLES AND OF THE H.B ROBINSON UNIT 2 NUCLEAR POWER PLANT, t

PRESSURE TRANSMITTERS IN SIMULATED HYDROGEN DEFLA.

NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION 1

GRATIONS TO DETERMINE SURVIVAL MARG 6NS AND SENSITIVI.

OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

TIES.

EMRIT,R.

DAVIS.C 8.

l NUREG-0933 SO4: A PRIORITIZATON OF GENERIC SAFETY iS$UES.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVM.UATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

ENDERLIN,W.I.

i NUREG/CR.4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON NUREG/CR4276 V02: VIBRATON AND WEAR IN STEAM GENERA-OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

TOR TUBES FOLLOWING CHEMICAL CLEANING.

DAVIS.M.S.

3 ENG.M.

NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAM: NATIONS NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS ON SOUDIFICATON. WASTE DIS,*0SALAND ASSOCIATED OCCU.

METHODOLOGY Volume 2: Codes And Example Problems.

f PATIONAL EXPOSURE.

1 ENGEL.D.W.

1 DAVIS.R.E.

NUREG/CR4486: VISA II. A COMPUTER CODE FOR PREDICTING f

NUREG/CR4433: DOCUMENT REVIEW REGARDING HAZARDOUS THE PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.

1 CHEMICAL CHARACTERISTICS OF LOW. LEVEL WASTE.

ESSENE.E.J.

DEAN.R.S.

NUREG/CR4585: INVESTIGATION OF THE STABIUTY OF CLAY /

NUREG/CR 3790- CLOSEOUT OF IE BULLETIN 79-07 SEISMIC BASALT PACKING MATERIALS.

I STRESS ANALYSIS OF SAFETY.RELATED PIPING l

EVANSJC, DOORANICH.D.

NUREG/CR4289: RESIDUAL RADIONUCLOE CONTAMINATION NUREG/CR4337:

TR AC.PF1/ MOD 1 INDEPENDENT WITHIN AND AROUND COMMERC'AL NUCLEAR POWER i

ASSESSMENT DARTMOUTH COLLEGE AIR WATER COUNTER-CUR-PLANTS. ORIGIN.DISTRIBUTON, INVENTOHY AND DECOMMIS$lON-RENT FLOW TESTS.

ING ASSESSMENT, l

DOCTOR.S.R.

EVANS.N A.

NUREG/CR 4464. PrRFORMANCE DEMONSTRATION TESTS FOR NUREG/CA.3961 STEAM EXPLOSON EXPERIMENTS AT INTERMEDI-DETECTION OF INTERGRANULAR STRESS CORROSION CRACK.

ATE SCALE.FITSB SERIES.

ING NUREG/CR4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-FANOUS.F.

NATON REUABluTY AND FRACTURE MECHANICS Semeannual NUREG/CR4223: STEEL CONTAINMENT RESISTANCE UNDER GEN-l ReportApnl. September 1984.

ERAL DYNAMO PRESSURES.

i DODGE.M.E.

F ECHT,5.A.

l NUREG/CR-4219 V01: AGfNG AND SERVICE WEAR OF HYDRAUUC NUREG/CR 3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN.

t AND MECHANICAL SNUBBERS USED ON SAFf4V.RELATED PIPING GINEERING GUIDEUNES FOR NUCLEAR POWER PLANT MAIN-AND COMPONENTS OF NUCLEAR POWER PLANTS Phase i Study.

TAINABluTY.

i l

I s

b m

N 30 Personal Author Index NUREG/CR446r A RANKih(NJJVES.

OF SABOTAGE / TAMPERING AVOO-GORE,5.F.

ANCE TECHNOLOGY ALTER NUREG/CR4462: A RANKING OF SABOTAGE / TAMPERING AVOID-

\\

ANCE TECHNOLOGY ALTERNATIVES.

p,ggygg NUREG/CR-4132: NUCLEAR POWER SENTY REPORTING SYSTEM GORHAM-BERGERON FINAL EVALUATON RESULTS.

NUREG/CR-4390: DCC 1/DCC-2 DEGRADED CORE COOLABILITY FITZSIMMONS D.

ANALYSIS.

NUREG/CR4276 V02-V.13 RATION AND VT2AR IN STUM GENERA.

TOR TUBES FOLLOWING CHEMICAL CLEANING.

NUREG/

-4 2 V01: AGING AND SERVICE WEAR OF CHECK FLANAGAN.G.F.

VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUALON NUCLEAR POWER PLANTS.

OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR-4183 V02: DRESSunlZED THERMAL SHOCK EVAt UATON GREIMANN,L OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR-4223: STEEL CONTAINMENT RESISTANCE UNDER GEN-ERAL DYNAMIC PRESSURES.

NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION GRIBdLE,R.P.

OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT-NUREG/CR4472: S AMESE IMAGING TECHNIOUE FOR QUASI-VERTI-NUREG/7R-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION CAL TYFF (CVT) DEFECTS IN NUCLEAR REACTOR PIPING.

OF THE H.B ROBINSON UN!T 2 NUCLEAR POWER PLANT.

MONAGEME.

FLUCKlGER.J D NUREG/CR4473 A STUDY OF THE OPERATON AND MAINTENANCE NUREG/CR4420: TURC11ARGE SCALE METALLIC MELT.CONCEN-OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 TRATE INTERACTION EXPERIMENTS AND ANALYSIS.

GUERRIERI,D.

FOLEY,W.J.

NUREG/CR4082 V03: DEGRACED PIPING PROGRAM - PHASE NUREG/CR-3790: CLOSEOUT OF IE BULLETIN 7947. SEISMIC ll.Semannual Report. April 1985 - September 1985.

STRESS ANALYSIS OF SAFETY-RELATED PIPING.

HAAS,P.M.

FRANZ.D.D.

NUREG/CR-4188 V01: NUCLEAR POWER PLANT SIMULATON FACILl-NUREG/CR4496: A SYSTEM FOR GENERATING LONG STREAM-TY EVALUATON METHODOLOGY.Handbooit J

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETUFsN PERIOD.

HADLOCK,D.L NUREG/CR4494: RADIOLOGICAL ASSESSMENT OF BWR RECIRCU-FREEMAN H.D NUREG/CRM04: LONG. TERM SilRVEILLANCE AND MONITORING OF DE MISSONED URANIUU PROCES$ LNG SITES AND TAIL.

HAGGARD.D.L.

NUREG/CR4485. THE (MPACT OF FUEL CLADDING FAILURE FRESCO.A.

EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE-NUREG/CR4311. REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL.

AR POWER PLANTS Case Study PWR Dunng Routme Operations.

lARY FEEDWATER SYSTEM RELIABILITY ANALYSIS.

NUREG/CR4565: PROBABILISTIC SAFETY STUDY APPLICATONS HALLINAN,K.P.

PROGRAM FOR INSPECTION OF INDIAN POINT UNIT 3 NUCLEAR NUREG/CR4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER POWER PLANT.

NATURAL CIRCULATON CONDITIONS.

FL'LL NOOD.R.

HAMLIN D.R.

NURt:G/CR4565 PROBABILISTIC SAFETY STUDY APPLICATONS NUREG/CA-4078: PROGRAM FOR FIELD VALIDAT.ON OF THE SYN-PROGRAM FOR INSPECTON OF INDIAN POINT UNIT 3 NUCLEAR THETIC APERTURE FOCUSING TECHNOUE FOR ULTRASONIC POWER PLANT.

TESTING (SAFT UT) Final Report GALLAHER.R 8.

HARRINGTON,R.M.

NUREG/CR.4302 V01: AGING AND SERVICE WEAR OF CHECK NUREG/CR4402 V02: HIGH-TEMPERATURE GAS-COOLED REACTOR VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF SAFETY STUDIES FOR THE DivislON OF ACCIDENT NUCLEAR POWER PLANTS.

EVALUATON Quarterty Progress Report, Apnl 1. June 30,1985 GAUSE,E.

NUREG/CR.3091 V07: REVIEW OF WASTE PACKAGE VERIFICATON HARTLEY,C S NUREG/CRI3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN-TESTS Semannual Report Covenng The Period Apnt-September 1985.

GINEERING GUCELINES FOR NUCLEAR POWER FLANT MAIN-GEISENDORFER,C.

TAINABill1 i.

NUREG/CR4411: ASSESSMENT OF SPECIAllZED EDUCATIONAL PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS.

HARTY,H.

NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR ENVI-GHERSON P.

RONMENT -lMPLICATIONS FOR NRC QUALITY ASSURANCE PRO.

NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA.

i OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

TORY PROJECTONS THROUGH 1995.

NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

HARTY,R, NUREG/CR-4494. RADIOLOGICAL ASSESSMENT OF BWR RECIRCU-LATORY PIPE REPLACEMENT.

UR G CR-4555 GENERIC COST ESilMATES FOR THE DISPOSAL OF RADIOACTIVE WASTES.

H ART 2OG H.R.

GOLDIN.D.

NUREG/CR 4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI.

NUREG/CR-4555-GENERIC COST ESTIMATES FOR THE DISPOSAL NATON RELIABillTY AND FRACTURE MECHANICS.Semannual OF RADIOACTIVE WASTES.

Report April. September 1984.

GOOD.M.S.

HAWLEY,K.A.

NUREG/CR4469 V01: INTEGRATON OF NONDESTRUCTIVE EXAMI-NUREG/CR 4504 LONG TERM SURVEILLANCE AND MONITORING NATON RELIABILITY AND FRACTURE MECHANICS $emannual OF DECOMMISSIONED UAANIUM PROCESSING SITES AND TAIL.

Report. April September 1984 (NGS PILES.

9I Personal Author Index 31 HEASLER,P.G.

ISHil,M.

NUREG/CR-4279 VOI: AGING AND SERVICE WEAR OF HYDRAUUC NUREG/CR-4501: MODELING OF VAPOR GENERATION IN FLASHING AND MECHANICAL SNUBBERS USE'10N SAFETY rFLATED PIPING FLOW.

AND COMPONENTS OF NUCLEAR POWER PLANio Phaw 1 Study NUREG/CR4464: PERFORMANCE DEVONSTRATON TESTS FOR BYER,K.

DETECTION OF INTERGRANULAR STRESS CORROSION CRACK-NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON ING.

OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR-4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION NATION REUABILITY AND FRACTURE MECHANICS. Semiannual OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

Report.Apnl. September 1984.

JACKSON,D.H.

HEBBLE.T.L NUREG/CR-4378. OBJECTIVE INDICATORS OF ORGANIZATONAL NUREG/CR-4059. EVALUATON OF THE IMPACT OF THE MC&A PERFORMANCE AT NUCLEAR POWER PLANTS.

REFORM AMENDMENTS ON A REPROCESSING FACluTY.

JAIN,H.

HELTON J.C.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR-4460 UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN TESTS. Semiannual Report Covering The Penod Apni-September 1985.

UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL E L-JANG J.

NUREG-0837 V05 NO3: NRC TLD DIRECT RADIATION MONITORING HENNICK,A.

NUREG/CR-379C. CLOSEOUT OF IE BULLETIN 79-07. SEISMIC NETWORK. Progress Report, July-September 1985.

STRESS ANALYSIS OF SAFETY-RELATED PIPING.

gg HENNINGER.R.J.

NUREG/CR-4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 (VER.

NUREG/CR4471: LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUD.

SiON 7.0).RELAP5/ MOD 1(CYCLE 14).AND TRAC-BD1 (VERSION lES

SUMMARY

RESULTS AND CONCLUSIONS.

12.0) CODES USING SEPARATE-EFFECTS EXPERIMENTS, HICK EY.E.E.

JO.J.J.

NUREG/CR 3365 DAF FC: REPORT TO THE NRC ON GUIDANCE FOR NUREG/CR-4452: REVIEW OF RELAP5 CALCULATIONS FOR H B.

PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX-ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

ERCISES AT NUCLEAR GENERATING STATIONS Draft Report For Comment.

JOHNSON,J.D.

NUREG/CR 4460: UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN HIGGINS.J.C.

UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL NUREG/CR4372; PROBABAUSTIC RISK ASSESSMENT (PRA) APhl-MODEL CATIONS.

JOHNSON,K.t.

HOR AN.J.R.

NUREG/CR-4486. VISA 11. A COMPUTER CODE FOR PREDICTING NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF THE PROBABluTY OF REACTOR PRESSURE VESSEL FAILURE.

U S AND FOREIGN NUCLEAR POWER PLANT DOSE EXPER!ENCE.

JOHNSON,R.

NU Ed CR4471: LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUD-NUREG/CR4477: WHMNS MR ASSESSM WWN PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK.

IES

SUMMARY

RESULTS AND CONCLUSIONS.

AGES.

HOSK ER.R.P.

NUR G/CR 113' FLOW AND DISPERSION NEAR CLUSTERS OF J

E qE CR4528: DEVELOPMENT AND VERIFICATION OF CONDI-

~

TlONS FOR DUCTILE TEARING INSTABILITY AND ARREST.

HUENEFELD.J.C.

NUREG/CR4411: ASSESSMENT OF SPECIALIZED EDUCATIONAL K E M PF,C.R.

PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS.

NUREG/CR4450 DAF FC: MANAGEMENT OF RADIOACTIVE MIXED WASTES IN COMMERCIAL LOW-LEVEL WASTES Draft Report For HUMPHREYS.P.

Comment.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

K E R R,H.T.

NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON NUREG/CR4059. EVALUATION OF THE IMPACT OF THE MC&A OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PL \\NT.

REFORM AMENDMENTS ON A REPROCESSING FACILITY.

HURRELL,5.J.

KIEFNER,J.

NUREG/CR.4059 EVALUATION OF THE IMPACT OF THE MC&A NUREG/CR4082 V03 DEGRADED P1 PING PROGRAM PHASE REFORM AMENDMENTS ON A REPROCESSING FACIUTY.

II. Semiannual Report.Arni 1985. September 1985.

HUTTON.P.H.

KMETYK,LN NUMEG/CR4300 V02. ACOUSTIC EMISSION / FLAW RELATIONSHIP NUREG/CR4171-TRAC-PF1/ MODI INDEPENDENT FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE ASSESSMENT LOBI LARGE BREAK TRANSIENT A104R VESSELS Progress Report. Apnt September 1985.

NUREG/CR-4465:

TRAC.PF1/ MOD t INDEPENDENT HWANG.H.

ASSESSMENT SEMISCALE MOO-2A INTERMEDIATE BREAK TEST NUREG/CR-4293. REUABluTY ANALYSIS OF SHEAR WALL STRUC.

S'30'3-TURES NUREG/CR4328 PROBABluTY BASED LOAD COMBINATION CRITE.

KOR NS.D.E.

AfA FOR DESIGN OF SHEAR WALL STRUCTURES NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR 4366. REUABluTY ASSESSMENT OF CONTAINMENT OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

TANGENTIAL SHEAR FAILURE.

NUREG/CR4183 V02: PRESSURIZED THERMAL SHOC EVALUATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

IDRISS.I.M.

NUREG/CR-3805 V03 ENGINEERING CHARACTERIZATION OF KRAEGER,BA GROUND MOTION Tasti 11-Observational Data On Spatial Venations NUREG/CR-4496-A SYSTEM FOR GENERATING LONG STREAM.

Of Earthquake Ground Motion.

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN PERIOD.

IM AN.R.L q

NUMEG/CR-4460 UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN KRAMER.G.

I UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL NUREG/CR 4082 V01 DEGRADED PIPING PROGRAM. PHASE MODEL l! Semiannual Report.Apnl 1985 September 1985

32 Personal Author index K ROEGER.P.G.

LOOMIS,G.G.

NUREG/CR-4434 ASSESSMENT OF MODELLING NEEDS FOR NUREG/CR.4393.

SUMMARY

OF SEMISCALE SMALL BREAK LOSS-SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

OF-COOLANT ACCIDENT EXPERIMENTS (1979 TO 1985).

NUREG/CR-4438. RESULTS OF SEMISCALE MOD 2C SMALL BREAK KURTZ,R.J.

($ ) togs.or. COOLANT ACCIDENT EXPERIMENTS S-LH-1 AND S-NUREG/CR-4300 V02: ACOUSTIC EMISSION / FLAW RELATIONSHIP LH-2.

FOR IN-SE RVICE MONITORING OF NUCLEAR PRESSURE VESSELS Progress Report, Aprd-September 1985.

LUBENAU,J.O.

NUREG/CP-0073. PROCEEDINGS OF THE WORKSHOP ON LARGE IR.

LAMONICA,LB.

RADIATOR RADIATION SAFETY.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT MACKENZIE.D.R.

NUREG/CR-4183 V02. PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR.4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED OF THE H B ROBlNSON UNIT 2 NUCLEAR POWER PLANT.

WASTES IN COMMERCIAL LOW-LEVEL WASTES Draft Report For Comment LANDOW.M.

NUREG/CR4082 V03. DEGRADED PIPING PROGRAM PHASE MAISE,G.

11 Semiant ual Report.Aprd 1985 September 1985.

NUREG/CR4547. CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM LAUGHERY,K.R.

ANALYSIS PROGRAM NUREG/C44188 V01. NUCLEAR POWER PLANT SIMULATION FACILi-MAROTTA,F.J NUREG/CR 4364. MANAGEMENT PERCEPTION OF THE HEALTH NU EG/CR 41 V02 N AR E

t NT SIMULATION FACILi-TV EVALUATION METHOOOLOGY Technical Bases.

PHYSICS TECHNICIAN JOB.

LE ALE.M.W.

M ARSCH ALL C.W.

NUREG/CR 4289 RESIDUAL R ADIONUCLIDE CONTAMrNATION NUREG/CR.4082 V03 DEGRADED PIPING PROGRAM PHASE WITHIN AND AROUND COMMERCIAL NUCLEAR POWER 11 Semiannual Report.Aprd 1985 September 1985.

PLANTS. ORIGIN. DISTRIBUTION, INVENTORY AND DECOMMISSION-ING ASSESSMENT.

M ART,G.A.

NUREG/CR 4469 VOI: INTEGRATION OF NONDESTRUCTIVE EXAMi-LEE.J.H.

NA TION RELIABILITY AND FRACTURE MECHANICS Semiannual NUREG/CR 4585 INVESTIGATION OF THE STABILITY OF CLAY /

Report.Apni September 1984 BASALT PACMNG MATERIALS M ARTIN,0.F.

LEHNER.J R.

NUREG/CR.3365 DAF FC. REPORT TO THE NRC JN GUIDANCE FOR NUREG/CR 4547: CONTEMPT 4/ MOD 6 A MULTICOVPONENT SYSTEM PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX.

AN ALYSIS PROGRAM.

ERCISES AT NUCLEAR GENERATING STATIONS Dra t Report For r

Comment.

LEIGH,C D.

NUREG/CR.4485 THE IMPACT OF FUEL CLADDING FAILURE NUhEG/CR 4460 UNCERTA;NTY AND SENSITIVITY ANALYSIS C F AN EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE-UPPER Pt ENUM TEST PROBLEM FOR THE MAEROS AERJSOL AR POWER PLANTS Case Study PWR Dunng Routne Operates MODEL MASON,C.L LE PE L.E. A.

NUREG/CR-4183 V01 PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR 4281 RE SIDU AL RADIONUCLIDE CONTAM; NATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

W. THIN AND AROUND COMMERCIAL NUCLEAR POWER NUREG/CR4183 V02 PRESSURIZED THERMAL SHOCK EVALUATION PLANTS. ORIGIN. DISTR:BUTION, INVENTORY AND DECOMMISSION-OF THE H D ROBINSON UNIT 2 NUCLEAR POWER PLANT.

ING ASSESSVENT.

M A X EY,W.

LEWIS.J R.

NUREG/CR-4082 V03-DEGRADED PIPING PROGRAM PHASE NUREG/CR 4473 A STUDY OF THE OPERATION AND MAINTENANCE il Semiannual Report.Aprd 1985 September 1985 OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55 MAZOUR,T.J.

NUREG/CH 4364 MANAGEMENT PERCEPTION OF THE HEALTH j

NU EG/ R 4 M VISA II A COVPUTER CODE FOR PREDICTING THL PROBABILITY OF REACTOR PRESSURE VESSEL FAILURE.

MCBRIDE.A.F.

LIME.J F.

NUREG/CR4183 V01. PRESSURIZED THERMAL SHOCK EVALUATION OF THE H B ROR!NSON UNIT 2 NUCLEAR POWER PLANT F G 7 OS A AM DECAY.HE AT.RE MOVAL STUD-REG /CR483 M RESMM WM Sm MMON lES

SUMMARY

RESULTS AND CONCLUSIONS OF THE H E ROBtNSON UNIT 2 NUCLEAR POWER PLANT.

LiN,C.

MCDRIDE,K.C.

NUREG/CR 4547. CONTEMPT 4/ MOD 6 A MULTICOVPONENT SYFTEM NUREG/CR 4473 A STUDY OF THE CPERATION AND MAINTEMNCE ANALYSIS PROGR3 M OF COMPUTER SYSTEMS TO MEET THE flEQUIREMENTS OF to CF A 73 55 LINSLEY,R K.

NUHEG/CR.44 M A SYSTEM FOR GENERATING LONG STREAM.

MCELROY,N L.

FLOW RECORDS F OR STUDY OF FLOODS OF LONG RETURN NUREG 1159 TRAINtNG MANUAL FOR URANIUM M LL WORMERS ON l'E RIOD HE ALTH PROTECTION FROM URANIUM LIPINSKI.R.J.

MCE LROY,W.H.

NUREG/CR4Mo DCC.1/DCC 2 DEGRADED CORF COOLABILIT Y NUREG/CR 4307 V01 LWR PRESSURE VESSEL SURVEILLANCE D0 AN ALY SIS SIMEIRY IMPROVEMENT PROGRAM Progress Report October LIPPINCOTT.E.P.

NUREG/CR 4307 Voi LWR PRESSURE VESSEL SURVElLLANCE 00 MELBER.B.D.

SIMETRY IMPROVEMENT PROGR AM Pro 0ress Report October NUREG/CR 441 t' ASSESSMENT Or SPECIALIZED EDUCATIONAL 1984 September t%5 PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS LCAR.J M.

MILLER.C.W.

NURE G/CH 4286 EVALUATION OF RADIOM!TIVE LIQUlO f FFLUENT NUREGICH 4286 EVALUATION OF RADIOACTNE LIQUID EFFLUENT f1 ELE ASES F ROM RANCHO SECO NUCL E AR POWE R PLANT, RELE ASES FROM RANCHO SECO NUCLEAR POWER PLANT.

Personal Author index 33 MILLE R.N.E.

OGDEN,0.M.

NUREG/CR4379 V02: LONG TERM PERFORMANCE OF MATERIALS NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION USED FOR HIGH-LEVEL WASTE PACKAGING Second Quarterly OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

Report Year Four July-September 1985 NUREG/CRat83 V02 PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR-4379 V03. LONG. TERM PERFORMANCE OF MATERIALS OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

USED FOR HIGH-LEVEL WASTE PACKAGING Third Quarterty Repc:1, Year Four October -December 1985.

OLSON,J.

NUREG/CR4378. OBJECTIVE INDICATORS OF ORGANIZATIONAL MILSTEAD,W.

PERFORMANCE AT NUCLEAR POWER PLANTS.

NUREG-0933 SJ4. A PRIORITIZATION OF GENERIC SAFETY ISSUES.

OLSONA MINARICK.J.W.

NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR4183 V01: PRESSURf2ED THERMAL SHOCK EVALUATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EV/LUATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

OSBORN,R.N.

NURE 50933 SO4: A PR!ORITIZATION OF GENERIC SAFETY ISSUES.

NUREG/CR4378 OBJECTIVE INDICATORS OF ORGANIZATIONAL PERFORMANCE AT NUCLEAR POWER PLANTS.

MITCHZLLD.E.

ATE AE IT'SB ER ES N EG/CRI 62. A RANKING OF SABOTAGE / TAMPERING AVOID-ANCE TECHNOLOGY ALTERNATIVES.

MOELLER,M.P.

NUREG/CR-3365 DAF FC. REPORT TO THE NRC ON GUIDANCE FOR OZTUNAlt,0.L PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX.

NUREG/CR4370 V01 UPDATE OF PART 61 IMPACTS ANALYSIS ERCISES AT NUCLEAR GENERATING STATIONS Draft Report For METHODOLOGY. Volume 1: Methodology Report Comment NUREG/CR4J70 V02; UPDATE OF PART 81 IMPACTS ANALYSIS NUREG/CR 4485 THE IMPACT OF FUEL CLADDING FAILURE METHODOLOGY. Volume 2: Codes And Example Problems.

EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE.

AR POWER PLANTS Case Study PW3 Dunng Routine Operatans.

M NURE /CR 082 V03 DEGRADED PIPING PROGRAM - PHASE MOLER.R.

Il Semiannual Report.Apnl 1985. September 1985.

NUREG/CR4477. METHODOLOGIES FOR ASSESSING LONG-TERM PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK.

PAPAZOGLOU LA.

AGES NUREG/CR4311. REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL.

lARY FEEDWATER SYSTEM RELIABILITY ANALYSIS.

MULLE NS.J.A.

NUREG/CR 4389. PRESSURE NOISE IN PRESSURIZED W ATLA RE.

PAPPIN,J.L ACTORS NUREG/CR4494 RADIOLOGICAL ASSESSMENT OF BWR RECIRCU.

I LATORY PIPE REPLACEMENT.

I pg NUREG/CR 4302 V01 AGING AND SERVICE WEAR OF CHECK PARKHURST,M.A.

I VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEM 3 OF NUREG/CR 4494 RADIOLOGICAL ASSESSMENT OF BWR RECIRCU-NUCLEAR POWER PLANTS LATORY PIPE REPLACEMENT.

N A K AG AKt.M.

l P ARKS,J.E.

NUREG/CR.4082 V03 DEGRADED PIPING PPOGRAM PHASE NUREG/CR 44t3 EIOASSAY MEASUREMENTS FOR URANIUM USING 11 Semiar r'ual Report. April f 985. September 1985 SPUTTER INITIATED RESON/NCE IONIZATION SPECTROSCOPY.

N AK Al.M.

PASUPATHI,V.

NURE J/CR4293 RELIABlUTY ANALYSIS OF SHEAR WALL STRUC-NUREG/CR 4082 V03 DEGRADED PlPING PROGRAM PHASE NU E CR4328 PROBABillTY BASED LOAD CCMBINATION CRITE.

RiA FOR DES!GN OF SHEAR WALL STRUCTURES.

PATRICK,M.G.

N ASSE RSHARIF.B.

NUREG/CR-3959 TRANSITION TO AN OPERATING REACTOR ENVI-NUREG/CR4471 LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUD-RONVENT lMPLICATIONS FOR NRC OUAllTY ASSURANCE PRO.

IES SUMVARY RESULTS AND CONCLUSIONS GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-TORY PROJECTIONS THROUGH 1995 NEUDER.S M.

NUREG-I t 01 V01 ONSITE DISPOSAL OF R ADIO ACTIV E PEACOR.D.R.

W ASTE Gv dance For Disposal Py Subsurface Buna, NUREG/CR 4585. INVESilGATION OF THE STABILITY OF CLAY /

BASALT PACklNG MATERIALS NEWTON.R D.

NUREG/CR-4132 NUCLEAR POWER SAFETY REPORTING SYSTEM PENDE R GR ASS,W.

FINAL EVALUATION RESULTS NUHEG/CR 4113 FLOW AND DISPERSION NEAR PLUSTERS OF BUILDINGS NEYMOTIN.L NUHEG/CR 4359 INDE PENDENT ASSESSMENT OF TR AC-PF1 (VER.

PE PPE R,S.

SiON 7 0LRELAPS/ MOD 1(CYCLE 14) AND TR AC Hot (VERSION NUREG/CR 4M6 RELIARILITY ASSESSMENT OF CONTAINMENT 12 0) CODES USING SEPARATE EFFECTS EXPER' VENTS TANGENTIAL SHEAR FAILURE.

NORDEN.K.

PE SC A TOR E,C.

NUREG/CR 2907 V03 RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR 4509 WASTE PACKAGE RELIADlLITY.

NUCLE AR POWER PLANTS Annual Report 1982.

PHILLIPS.L D.

NOUR B A KHSH.H.P.

NUREG/CR 4181 V01 PRESSURIFO THERMAL SHOCK EVALUATION NUREG/CR4183 V01 PRESSURIZED THERMAL SHOCK EVALUAtlON OF THE H B RORINSON UNIT 2 NUCLEAR POWER PLANT OF THE H B ROWNSON UNIT 2 NUCLE AR POWER PLANT.

NUREG/CR 4183 V02 PRESSURtlED THERMAL SHOCK EVALUATION NUHEG/CR 4183 V02 PRESSURilED THERMAL SHOCK EVALVATION OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

OF THE H B RODINSON UNIT 2 NUCLEAR POWER PLANT.

PICIULO,P.L NUSSB AUMER.D A.

NUHEG/CR 3444 V03 THE IMPACT OF LWR DECONTAMINATIOP:S HUREG!CP 0073 PROCEEDINGS OF THE WORKSHOP ON LARGE 1R-ON SOLIOlFICATION WASTE DISPOSAL,AND ASSOCIATED OCCV.

RADIATOR RADIAtlON SAF ETY PATIONAL EXPOSURE.

34 Personal Author index PIRES,J.

RIZNIC,J.

NUREG/CR4293. RELIABILITY ANALYSIS OF SHEAR WALL STRUC-NUREG/CR-4501: MODELING OF VAPOR GENERATION IN FLASHING TURES.

FLOW.

NUREG/CR4366: RELIABILITY ASSESSMENT OF CONTAINMENT TANGENTIAL SHEAR FAILURE.

ROBERTSON,D.E.

NUREG/CR-4289. RESIDUAL RADIONUCUDE CONTAMINATION G d933 SO4. A PRIORITIZATION OF GENERIC SAFETY ISSUES'WITH'N AND AROUND COMMERCIAL NUCLEAR POWER U

PLANTS. ORIGIN. DISTRIBUTION INVENTORY AND DECOMMISSION-poN,W.o.

ING ASSESSMENT, NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS METHODOLOGY. Volume 2. Codes And Example Problems.

RO A Gl. S p

POWER.M.S.

SION 7.0).RELAP5/ MODI (CYCLE 14).AND TRAC-BD1 (VERSION NUREG/CR-3805 V03. ENGINEERING CHARACTERIZATION OF 12.0) CODES USING SEPARATE-EFFECTS EXPERIMENTS.

GROUND MOTION. Task II. Obsentatonal Data On Spatial Vanat>ons NUREG/CR-4452: REVIEW OF RELAPS CALCULATIONS FOR H B.

Of Earthqualie Ground Moton.

ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

POWERS,T.B.

ROLES,G.W.

NUREG/CR4462: A RANKING OF SABOTAGE / TAMPERING AVOID-NUREG/CR4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS ANCE TECHNOLOGY ALTERNATIVES-METHODOLOGY. Volume 1: Methodology Report.

NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS NU G 82 DRF FC: RECOMMENDATIONS TO THE NUCLEAR METHODOLOGY. Volume 2: Codes And Example Problems.

REGULATORY COMMISSION ON TRIAL GUIDELINES FOR SEISMIC ROSS P.A.

MARGIN REVIEWS OF NUCLEAR POWER PLANTS Draft Report For Comment NUREG-0020 V09 N12: LICENSED OPERATING FsEACTORS STATUS

SUMMARY

REPORT Data As Of November 30,1985 (Gray Book I)

PRATE R.J.T.

NUREG4020 V10 N01: LICENSED OPERATING REACTORS STATUS NUREG/CR-447L HFiH-TEMPERATURE OXIDATION OF ZlRCALOY4

SUMMARY

REPORT Data As Of December 31.1985 (Gray Book 1).

IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

RMM PUGH.C.E.

NUREG 1109 DRFT FC; REGULATORY ANALYSIS FOR THE RESOLU-NUREG/CR-4219 V02: HEAVY-SECTION STEEL TECHNOLOGY PRO-TION OF UNRESOLVED SAFETY ISSUE A44. STATION BLACKOUT.

CRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL SEPTEMBER 1985.

RUNKLE.G.E.

RABATIN K.

NUREG/CR4369 OUAUTY ASSURANCE (OA) PLAN FOR COMPUTER NUREG 0837 V05 NO3 NRC TLD DIRECT RADIATION MONITOR!NG SOFTWARE SUPPORTING THE U S. NUCLEAR REGULATORY COM-NETWORK Progress Report. July. September 1985.

MISSION'S HIGH-LEVEL WASTE PROGRAM.

l R AMSDELL.J V SAM,W NUREG/CRdOOO V01.

THE MESORAD DOSE ASSESSMENT NUREG/N44 R ASSESSM & SMAM NGM MODELVolume 1.Techmcal Basis.

PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS.

RANKIN.W.L SAHA.P.

NUREG/CR4436 VL 1: HUMAN RELIABfLITY IMPACT ON INSERVICE NUREG/CA-4359. INDEPENDENT ASSESSMENT OF TRAC-PF1 (VER-INSPECTION Volume 1. Phase 1 SummaryReport.

SiON 7.0).RELAP5/ MOD 1(CYCLE 14),AND TRAC-BD1 (VERSION NUREG/CR4436 V02 HUMAN REUABILITY IMPACT ON INSERVICE 12.0) CODES USING SEPARATE-EFFECTS EXPERIMENTS.

INSPECTION Volume 2. Rewew And Analysis Of Human Performance in Nondestructrve Testing (Emphaseng Ultrasomcs)

SANDORN,Y.

R ANKIN,W.R.

NUREG/CR4545. PIPE CRACK EVALUATION IN OPERATING BOILING WATER REACTORS.

NUREG/CR-4462: A RANKING OF SABOTAGE / TAMPERING AVOID-ANCE TECHNOLOGY ALTERNATIVES SASTRE.C.

R AVINDRA.M.K.

NUREG/CR4482 DAF FC: RECOMMENDAtlONS TO THE NUCLEAR SAyy,J.g, REGULATORY COMMISSION ON TRIAL GUIDELINES FOR SEISMIC NUREGICR-4482 DRF FC: RECOMMENDATIONS TO THE NUCLE AR MARGIN REVfEWS OF NUCLEAR POWER PLANTS Draft Report For Comment REGULATORY COMMISSION ON TRIAL GUIDELINES FOR SEISMIC MARGIN REVIEWS OF NUCLEAR POWER PLANTS Draft Report For R E E D.A.W.

Comment NUREG/CR-4390 DCC 1/DCC 2 DEGRADED CORE COOLABILITY ANALYSIS SCHERPELZ,R.L NUREG/CR 4000 V01: THE MESORAD DOSE ASSESSMENT REICH.M.

MODELVolume 1. Technical Basis.

NUREG/CR-3137. SEISMIC AND DYNAMIC OVALIFICATION OF RE-j LATED ELECTRICAL AND MECHANICAL FOUiPMENT.

SCHMIDT,T.R.

NUREG/CR4293 RELIABILITY ANALYSIS OF SHEAR WALL STRUC.

NUREG/CR4390. DCC 1/DCC-2 DEGRADED COAE COOLABillTY TURES ANALYSIS.

NUREG/CR 4328 PROBABILITY BASED LOAD COMBINATION CRITE.

RIA FOR DESIGN OF SHEAR WALL STRUCTURES.

SCHRElBER.R.E.

RE X ROTH.P.E-NUREG/CR 4462 A RANKING OF SABOTAGE / TAMPERING AVOID-NUREG/CR-4343 INTEGRATED SEVERE ACCIDENT CONTAINMENT ANCE TECHNOLOGY ALTERNATIVES.

ANALYSIS WITH THE CONTAIN COMPUTER CODE-SCHULLER,C.R.

j NUREG/CA.3959 TRANSITION TO AN OPERATING PEACTOR ENVI.

U G-0933 SO4 A PRIORITIZATION OF GENERIC SAFETY ISSUES R

N A

S FOR NRC OVAW ASSWEE N GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-R)GGS,R.

TORY PROJECTIONS THROUGH 1995 NUREG 0933 SO4 A PRIORITIZATION OF GENERIC SAFETY ISSUES SCHULT2.D.H.

AlORDAN,0.J.

NUREG/CR-3365 DRF FC. PIEPORT TO THE NRC ON GUIDANCE FOR NUREG/CH 4546 LABOR PRODUCTIVITY ADJUSTMENT FACTOR $ A PREPARING SCENARIOS'FOR EMERGENCY PREPAREDNESS EX-Method for Esimatmg Labor Constructen Costs Associated with ERCISES AT NUCLEAR GENERATING STATIONS Draft Report For Physical Modifica' ions To Nuclear Power Plants Comment

Personal Author index 35 SCHWARTZ.M.W.

SMITH,S.

NUREG/CH-3760: A STUDY ON DUCTILE AND BRITTLE FAILURE NUREG/CR4477: METHODOLOGIES FOR ASSESSING LONG-TERM DESIGN CRITERIA FOR DUCTILE CAST IRON SPENT FUEL SHIP-PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-PING CONTAINERS-AGES.

NUREG/CR-4363: A STUDY ON FABRICATION CRITERIA FOR DUC-TILE CAST IRON SPENT-FUEL SHIPPING CONTAINERS-SNOKE,J.A.

SCIACCA F NUREG/CR-4502: VIRGINIA REGIONAL SEISMIC NETWORK. FINAL NUREG/dR4555: GENERIC COST ESTIMATES FOR THE DISPOSAL REPORT (19771985).

OF RADIOACTIVE WASTES.

SOMERVILLE.P.G.

SCOTT,P.

NUREG /CR-3805 V03: ENGINEERING CHARACTERl2ATION OF NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM. PHASE GROUND MOTON. Task 11: Observational Data On Spatial Vanations ll. Semiannual Report.Apnl 1985 September 1985.

Of Earthquake Ground Motion.

SEGE.G.

SOO,P.

NUREG-0933 SO4: A PRIORITIZATION OF GENERIC SAFETY ISSUES-NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION SEL8Y,D.L TESTS Semiannual Report Covenng The Penod Apr4-Septemoer 1985.

NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION SPAAR.M T NU E CR418 V02 R S IZ D HERM O K EVALUATION NUREG/CR4419: BIOASSAY MEASUREMENTS FOR URANIUM USING OF THE H 8. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

SPUTTER INITIATED RESONANCE IONIZATION SPECTROSCOPY.

SHAFFER,C.

SPANNER J.C.

NUREG/CR4555: GENERIC COST ESTIMATES FOR THE DISPOSAL NUREG/CR4436 V01: HUMAN REUABlWTY IMPACT ON INSERVICE OF RADIOACTIVE WASTES.

INSPECTION. Volume 1: Phase 1 Summary Report.

NUREG/CR-4436 V02: HUMAN REUA81UTY IMPACT ON INSERVICE SHIKIAR.R.

INSPECTION. Volume 2: Review And Analysis Of Human Performance NUREG/CR4378. OBJECTIVE INDICATORS OF ORGANIZATONAL in Nondestructive Testing (Emphasmng Ultrasonics).

PERFORMANCE AT NUCLEAR POWER PLANTS.

NUREG/CR4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-SHINOZUKA.M NATION REUABILITY AND FRACTURE MECHANICS Semiannual NUREG/CRd328: PROBABluTY BASED LOAD COM8INATION CRITE-Report,Apnl September 1984.

RIA FOR DESIGN OF SHEAR WALL STRUCTURES STAHL.D.

SHOCKLEY W.E.

NUREG/CR-4379 V02: LONG TERM PERFORMANCE OF MATERIALS NUREG/CR-4327: ORGANIC IODOE FORMATON FOLLOWING NU-USED FOR HIGH-LEVEL WASTE PACKAGING Second Quarterly CLEAR REACTOR ACCOENTS.

Report. Year Four July-September 1985.

NUREG/CR-4379 V03. LONG TERM PERFORMANCE OF MATERIALS SIBOL.M.S.

USED FOR HIGH. LEVEL WASTE PACKAGING Third Quarterty NUREG/CR-4502: VIRGINIA REGIONAL SEISMIC NETWORK FINAL Report, Year Four October -December 1985.

REPORT (19771985).

STEPHENS.K.

U Ed/CR 3805 V03-ENGINEERING CHARACTERl2ATION OF GROUND MOTION Task 11: Observational Data On Spatial Vanatens g[3-Of Earthquake Ground Motion.

STREIT,J.E.

SIMONEN.E.P'4486. VISA !!. A COMPUTER CODE FOR PREDICTING NUREG/CR NUREG/CR-4438. RESULTS OF SEMISCALE MOD-2C SMALL BREAK THE PROBA81UTY OF REACTOR PRESSURE VESSEL FAILURE.

($%) LOSS-OFOOOLANT ACCIDENT EXPERIMENTS S-LH 1 AND S-LH 2.

SIMONEN.F.A.

NUREG/CR-4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI.

STUMPF,H.

NATION REUABILITY AND FRACTURE MECHANICS Semiannual NUREG/CR 4442. TRAC USER'S GUOE.

Report.Apnl. September 1984 NUREG/CR-4486 VISA 11. A COMPUTER CODE FCR PREDICTING SUBUDHl,M.

THE PROBABruTY OF REACTOR PRESSURE VESSEL FAILURE.

NUREG/CR-3137: SEISMIC AND DYNAMIC OUAUFICATION OF RG O

^

SIMPKINS,B.

NUREG/CR-4555 GENERIC COST ESTIMATES FOR THE DISPOSAL SULLIVAN.T.

OF RADIOACTIVE WASTES.

NUREG/CR 3091 V07. REVIEW OF WASTE PACKAGE VERIF6CADON SISKIND.B.

TESTS Semiannual Report Covenng The Penod ApreSeptember 1995.

NUREG/CR-4433. DOCUMENT REVIEW REGARDING HAZARDOUS NUREG/CR 4509 WASTF PACKAGE REUABluTY.

CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

SUO ANTTILLA.A.

SKAGGS R.L NUREG/CR-4420 TURCf LARGE SCALE METALUC MELT CONCEN.

NUREG/CR.4323. THE PROTECTION OF URANIUM TAILINGS IM.

TRATE INTERACTION EXPERIMENTS AND ANALYSIS POUNUMENTS AGAINST OVERLAND EROslON.

T AB ATA B Al.A.S.

SLOVIK.G.C.

NUREG/CR 4462. A RANKING OF SABOTAGE / TAMPERING AVOID-NUREG/CR 4359 INDEPENDENT ASSESSMENT CF TRAC PFt (VER.

ANCE TECHNOLOGY ALTERNATIVES.

SiON 7 0).RELAPS/ MODI (CYCLE 14),AND TRAC 801 (N ERSION 12 0) CODES USING SEPARATE-EFFECTS EXPERIMENTS TAWIL.J.J.

NUREG/CR 4462: A RANKING OF SABOTAGE /T%MPERING AVOO-ANCE TECHNOl.OGY ALTERNATIVES.

N EG-0525 Rt1: SAFEGUAF10S

SUMMARY

EVLNT UST (SSEL).

SMITH.R.

TAYLOR E.H.

NUREG/CR 4471 LOS ALAMOS PWR DECAY HEAT REMOVAL STUD.

NUREG/CR 4419-DIOASSAY MEASUREMENTS FOR URANIUM USING IES

SUMMARY

RESULTS AND CONCLUSIONS SPUTTER INITIATED RESONANCE lON12ATION SPECTROSCOPY.

SMITH.R.D.

TAYLOR.J.H.

NUREG-1179 Vol. RUPTURE OF MODEL 48Y UF6 CYUNCER AND RE.

NUREG/CR 4565 PROBABluSTIC SAFETY STUDY APPLICATIONS LEASE OF URANIUM ' HEXAFLUORIDE Sequoyah Fuels PROGRAM FOR INSPECTION OF INDIAN POINT UNIT 3 NUCLEAR Facihty, Gore. Oklahoma. January 4.1986 POWER PLANT.

1 36 Personal Author index TAYLOR,T.T.

W A NG.P.C.

l NUREG/CR-4464 PERFORMANCE DEMONSTRATION TESTS FOR NUREG/CR-4293. RELIABILITY ANALYSIS OF SHEAR WALL STRUC-DETECTION OF INTERGRANULAR STRESS CORROSION CRACK.

TURES.

ING.

NUREG/CR 4469 V01. INTEGRATION OF NONDESTRUCTIVE EXAMI.

W ANLE SS,J.

NATION REllABILITY AND FRACTURE MECHAN'CS Semiannual NUREG/CR-4310 INVESTIGATION OF POTENTIAL FIRE-RELATED Report.Aprd September 1984.

DAMAGE TO SAFETY RELATED EQUIPMENT IN NUCLEAR POWER PLANTS TEUTONICO,LJ.

NUREG/CR-4545 PIPE CRACK EVALUATION IN OPERATING BOILING W E E K S,J.R.

WATER RE ACTORS.

NUREG/CR-4545 PIPE CRACK EVALUATION IN OPERATING DOILING WATER REACTORS THEOFANOUS,T.G.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION WEISS,A.J.

OF THE H B RODINSON UNIT 2 NUCLEAR POWER PLANT NUREG/CP-0072 V01: PROCEEDINGS OF THE THIRTEENTH WATER NUREG/CR-4t&3 V02: PRESSURIZED THERMAL SHOCK EVALUATION REACTOR SAFETY RESEARCH INFORMATON MEETING OF THE H.8 RO8iNSON UNIT 2 NUCLE AR POWER PLANT.

NI REG /CP.0072 V02. PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING THIE.J.A.

NUREG/CP-OO72 V03. PROCEEDINGS OF THE THIRTEENTH WATER NUREG/CR-4389 PRESSURE NOISE IN PRESSURIZED WATER RE.

REACTOR SAFETY RESEARCH INFORMATION MEET!NG ACTORS NUREG/CP 0072 V04. PROCEEDINGS OF THE THtRTEENTH WATER REACTOR SAFETY RFSEARCH INFORMATON MEETING THOMAS,C.W.

NUREG/CP.0072 V05 PHOCEEDINGS OF THE THIRTEENTH WATER NUREG/CR 4289 RESIDUAL RADIONUCLIDE CCNTAMINATION REACTOR SAFETY RESEARCH INFORMATION MEET;NG NUREG/CP-OO72 V06 PROCEEDINGS OF THE THIRTEENTH W ATER WITHIN AND AROUND COMME RCI AL NUCLEAR POWER PLANTS.ORl GIN.DISTPIBUTON. INVENTORY AND DECOMMISSION ^

NU E /CR 233 5

SA ESEAF P

R MS SPON.

ING ASSESSMENT.

SORED BY OFFICE OF NUCLEAR REGULATORY THOM AS,W.V.

RESEARCH Ouarterly Progress Report Aont 1-June 30,1985.

NUREG/CR-4289 RESIDUAL RADIONUCLIDE CONTAM: NATION WHITE,A.S WITHIN AND AROUND COMMERCIAL NUCLEAR POWER NUREG/CR-4411: ASSESSMENT OF SPECIAll?ED EDUCATIONAL PLANTS. ORIGIN DISTRIBUTION. INVENTORY AND DECOMMISSION-PROGR AMS FOR LICENSED NUCLEAR RE ACTOR OPERATORS ING ASSESSMENT.

WILKINSON G.F.

TICHLER,J.

NUREG/CR.4369 OUALITY ASSURANCE (OA) PLAN FOR COMPUTER NUREG/CR 2907 V03 RADIOACTIVE MATERIALS RELEASED FROM SOFTW ARE SUPPORTING THE U S. NUCLEAR REGULATORY COM-NUCLEAR POWER PLANTS. Annual Report 1982-MISSION'S HIGH LEVEL WASTE PROGRAM.

TILLS,J.L WILKOWSKl,G.M.

NUREG/CR-4541 INTEGRATED SEVERE ACCIDENT CONTAINMENT NUREG/CR-4082 V03 DEGRADED PIPING PROGRAM PHASE ANALYSIS WITH THE CONTAIN COMPUTER CODE-11 Semiannual Report.Apfd 1985 - September 1985 TOBIAS.M L WILLIAMS.D.C.

NUREG/CR 4255 V02 AEROSOL RELEASE AND TRANSPORT PRO-NUREG/CR 4343 INTEGRATED SEVERE ACCIDENT CONTAINMENT GRAM SEMlANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER ANALYSIS WITH THE CONTAIN COMPUTER CODE.

1985 WILSON,J H.

TRIGGS.T.J.

NUREG/CR.4402 V02 HIGH. TEMPERATURE GAS COOLED REACTOR NUREG/CR 4436 V01. HUMAN REUABILITY IMPACT ON INSEnvlCE SAFETY STUDIE S FOR THE DivisiOH OF ACCIDE NT INSPECTION Volume t Phann 1 Surnmary Report EVALUATION Ouarterty Progress Report, Aptd 1 June 30,1985 NUREG/CR 4436 V02 HUMAN REllAthLlTY IMPACT ON INSERvlCE INSPECTION Volume 2 Review And AnaWs Of Human Performance WITHERSPOON,J.

In Nondestructive Testag (Emphawing Ultrason(si NUREG/CR 4286 EVALUATION OF RAD'OACTIVE LtOVO EFFLUENT RELEASES FROM RANChiO SECO N) CLEAR POWER PLANT.

UNWIN,S D.

NUREG/CR 4514 CONTROLLING PRINCIPLES FOR PROR PRORA-Y OUNG,J. A.

OlLITY ASSIGNMENTS IN NUCLE AR RISK ASSESSME NT, NUREG/CR.4504 LONG-TERM SURVEILLANCE AND MONITORING OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL.

USHER,J.L INGS PitES NUREG/CR 4479 THE USE OF A FIELD MODEL TO ASSESS FIRE DE.

HAVOR IN COMPLE X NUCLEAR POWER PLAN T YOUNGBLOOD.R.

ENCLOSunES PRESENT CAPA31LlitES AND FUTURE PROSPECTS NUREG/CR.4311 REVIEW OF THE SHEARON HARRIS UNIT 1 AUXil-lARY FEEDWATER SYSTEM REUABILITY ANALYSIS VAN POOYEN,0.

NUREG/CR 4545 PtPE CRACK EVALUATON IN OPER ATING BOfLING YUELYS-MIKSl3 W ATER REACTORS NUREG'CH 4359 INDEPENDENT ASSESSMENT OF TRAC PF1 (VER-StON 7 0).RELAP5/ MODI (CYCLE 14 LAND TRAC 801 (VERSION VAN TUYLE,0.J.

12 0) CODES USING SEPARATE E FF ECTS E XPER MENTS NUREG/CR 4434 ASSF SSME N T OF MODELLING NEEDS FOR NUREG/CH 4452 REVIEW OF RELAPS CALCULATIONS FOR H 0 SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS RORINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

V ANDE R MOLE N.H.

ZAREMBA,L NUREG 003 SO4 A PRORlil2ATION OF GENERIC SAFETY ISSUES NUREGICR 4477 METHODOLOGIES FOR ASSESSING LONG TERM PERFORMANCE Or HIGH-LEVEL RADIOACTIVE WASTE PACM-VISKANTA.R.

AGES NUREG/CR 4556 HE AT TRANSFER FROM A F OD HUNDLE UNDE R NATURAL CIRCUL ATION CONDITIONS ZlGLER G.L NUREG/CR 4183 V01 PRESSUR!IED THEnMAL SHOCK EVALUATION CALTER W H.

OF THE H H ROB:NSON UNIT 2 NUCLEAR POWER PLANT NUREG/CR 432? THE PROTFCTION OF URANIUM TAILINGS IM NUREG/CR 4183 V02 PRES $URIZED THERMAL SHOCK EVALUATION POUNDMENTS AGAINST OVE ALAND EROSION OF THE H 0 ROBINSON UNIT 2 NUCLEAR POWER PLANT.

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.

10 CFR 60 Acoustic Emission NUREG/CR-4509: WASTE PACKAGE REUA81UTY.

NUREG-0975 V04: COMPILATON OF CONTRACT RESEARCH FOR THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER.

10 CFR Part 61 ING TECHNOLOGY. Annual Report For FY 1985.

NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS NUREG/CR4300 V02: ACOUSTIC EMISSION / FLAW RELATIONSHIP METHODOLC Y. Volume 1: Methodology Report FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR-437J VO2; UPDATE OF PART 61 IMPACTS ANALYSIS VESSELS Progress Report, Apni-September 1965.

METHODOLOGY. Volume 2: Codes And Example Problems.

Adnormal Occurrence fRNCE NURE 01 V01: ONSITE DISPOSAL OF RADIOACTIVE t

1 5 WASTE. Guidance For Disposal By Subsurface Burial AI 40 CFR 192 NUREG/CR4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT NUREG/GR-4504: LONG-TERM SURVEILLANCE AND MONITORING OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL-ANALYSIS WITH THE CONTAIN COMPUTER CODE.

INGS PILES' NUREG/CR4420: TURC1.LARGE SCALE METALUC MELT-CONCEN.

TRATE INTERACTION EXPERIMENTS AND ANALYSIS.

40 CFR Part 261 NUREG/CR-4433; DOCUMENT REVIEW REGARDING HAZARDOUS A' 8'h'VI*'

CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

NUREC/CR4460- UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL

)

ALARA MODEL NUREG/CP-0068: PROCEEDINGS OF THE INTERNATIONAL NUCLEAR REACTOR DECOMMISSIONING PLANNING CONFERENCE.

Aerosol Release And Trar sport Program NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF NUREG/CR4255 V02: AEROSOL RELEASE AND 1 MNSPORT PRO-U.S AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE.

GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER ATWS NUREG/CR-4434: ASSESSMENT OF MODELLING NEEDS FOR Aging SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

NUREG/CR-2331 V05 N2: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY Abstract RESEARCH.Ouarterty Progress Report.Apnl 1 June 30.1385.

NUREG-0304 V10 NC4: REGULATORY AND TECHNICAL NUREG/CR-4279 VOI: AGING AND SERVICE WEAR OF HYDRAUUC ncPORTS Annual Compilation for 1985-AND MECHANICAL SNUB 8ERS USED ON SAFETY-RELATED PIPING Access ConW AND COMPONENTS OF NUCLEAR POWER PLANTSJhate 1 Study.

NUREG/CR4302 V01: AGING AND SERVICE WEAR OF CHECK NUREG/CR4473: A STUDY OF THE OPERATION AND MAINTENANCE VALVEG USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 NUCLEAR POWER PLANTS.

CFR 73 55.

NUREG/CR-4380: EVALUATION OF THE MOTOR-OPERATEC VALVE Accident ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT NUREG-118J V01: ASSESSMENT OF THE PUBUC HEALTH IMPACT DEGRADATON,1NCORRECT ADJUSTMENTS.AND OTHE R ABNOR-FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH IS ES.

FUELS CORPORATION FACluTY AT GOREOKLAHOMA.Repnnted Air Quality NUR 119 V0'2: ASSESSMENT OF THE PUBL6C HEALTH IMPACT NUREG/CR-4113-FLOW AND DISPERSION NEAR CLUSTERE OF FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH BUILDINGS.

FUELS COR' ORATION FACluTY AT GORE. OKLAHOMA.

NUREG/CR-3958: EFFECTS OF CONTROL SYSTEM FAILURES ON Emuent TRANSIENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A NUREG/CR-2907 V03: RADIOACTIVE MATERIALS RELEASED FROM COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR.

NUCLEAR POWER PLANTS. Anrcal Report 1982.

NUREG/CR-4324: TESTING OF NUCLEAR OUALIFIED CABLES AND A

no Release PRESSURE TRANSMITTERS IN SIMULATED HYDROGEN DEFLA-GRATIONS TO DETERMINE SURV VAL MARGINS AND SENSITr/I-NUREG-1179 V01: RUPTURE OF VODEL 48Y UF6 CYUNDER AND RE-TIES.

LEASE OF URANIUM HEXAFLUORIDE.Sequoyah Fuels NUREG/CR-4327: ORGANIC IODIDE FORMATON FOLLO'MNG NU.

Facility, Gore, Oklahoma, January 4.1986.

CLEAR REACTOR ACCIDENTS.

NUREG/CR-4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT nmbient Radiation Level ANALYSIS WITH THE CONTAIN COMPUTER CODE.

NUREG-0837 VOS NO3: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4434: ASSESSMENT OF MODELUNG NEEDS FOR NETWORK. Progress Report, July September 1985.

4 SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

NUREG/CR-4434: ASSESSMENT OF MODELUNG NEEDS FOR Annual Report SAFETY ANALYSIS OF CURRENT HTGA CONCEPTS.

NUREG-0975 V04: COMPILATON OF CONTRACT RESEARCH FOR NUREG/CR4465:

TRAC PF1/ MOD 1 INDEPENDENT THE MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEER.

ASSESSMENT:SEMISCALE MOD 2A INTERMEDIATE BREAK TEST ING TECHNOLOGY. Annual Report For FY 1985.

S-IB-3 NUREG/CR-2907 V03: R ADICACTIVE MATERIALS RELEASED FROM NUREG/CR-4485: THE IMPACT OF FUEL CLADDING FAILURE NUCLEAR POWER PLANTS. Annual Report 1982.

EVENTS ON OCCUPATONAL RADIATION EXPOSURES AT NUCLE.

NUREG/CR-3950 V02: FUEL PERFORMANCE ANNUAL REPORT FOR AR POWER PLANTS. Case Rtudy-PWR Dunng Routine Operatons.

1984.

37

38 Subject Index Anonymity CONTEMPT NUREG/CR-4132-NUCLEAR POWER SAFETY REPORTING SYSTEM NUREG/CR4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM FINAL EVALUATON RESULTS.

ANALYSIS PROGRAM.

Assessment CONTEMPT 4/ MOO 6 NUREG-1189 V01: ASSESSMENT OF THE PUBUC HEALTH IMPACT NUREG/CR-4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM FROM THE #OCIDENTAL RELEASE OF UF6 AT THE SEQUOYA4 ANALYSIS PROGRAM.

FUELS CORPORATION FACiUTY AT GORE, OKLAHOMA.Repnnted March 26,1986.

CORCORN NUREG-1189 V02' ASSESSMENT OF THE PUBLIC HEALTH IMPACT NUREG/CR4420: TURC1:LARGE SCALE METALUC MELT CONCEN-FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH TRATE INTERACTON EXPERIMENTS AND ANALYSIS.

FUELS CORPORATION FACIUTY AT GORE. OKLAHOMA.

Cabinet Fire Austenttic Stainless Steel NUREG/CR-4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION HAVIOR IN COMPLEX NUCLEAR POWER PLANT TESTS. Semiannual Report Covenng The Penod Apni-September 1985.

ENCLOSURES PRESENT CAPABluTIES AND FUTURE PROSPECTS.

Auxiliary Feedwater System Cable NUREG/CR-4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL*

NUREG/CR-4324: TESTING OF NUCLEAR QUALIFIED CABLES AND lARY FEEDWATER SYSTEM REUABluTY ANALYSIS-PRESSURE TRANSMITTERS IN S.MULATED HYDROGEN DEFLA-GRATONS TO DETERMINE SURVIVAL MARGINS AND SENSITIVI-90SOR4 TIES.

NUREG/CR-4223: STEEL CONTAINMENT RESISTANCE UNDER GEN-ERAL DYNAMIC PRESSUAES.

Cable Fire 80 SORS NUREG/CR-4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-NUREG/CR4223: STEEL CONTAINMENT RESISTANCE UNDER GEN-HAVIOR IN COMPLEX NUCLEAR POWER PLANT ERAL DYNAMIC PRESSURES.

ENCLOSURES PRESENT CAPABluTIES AND FUTURE PROSPECTS.

Backfilt Capacity NUREG/CR-4585: INVESTIGATION OF THE STABILITY OF CLAY /

NUREG/CR-3959: TRANSITON TO AN OPERATING REACTOR ENVI-BASALT PACKING MATERIALS.

RONMENT -IMPLICATIONS FOR NRC QUALITY ASSURANCE PRO-GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-Barrier Degradation Event TORY PROJECTIONS THROUGH 1995.

NUREG/CR-4485: THE IMPACT OF FUEL CLADDING FAILURE EVENTS ON OCCUPATONAL RADIATION EXPOSURES AT NUCLE-Cast Stalniese Steel AR POWER PLANTS. Case StudyPWR Dunng Routine Operations.

NUREG/CR4469 VO1: INTEGRATON OF NONDESTRUCTIVE EXAMI-NATION REUABIUTY AND FRACTURE MECHANICS. Semiannual Bioassay Report Apnl September 1984.

NUREG/CR4419: BIOASSAY MEASUREMENTS FOR URANIUM USING SPUTTER INITIATED RESONANCE IONIZATON SPECTROSCOPY.

Check Valve NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK 4U EG CR-4501: MODEUNG OF VAPOR GENERATION IN FLASHING U EAR WRP S

FLOW.

Chemical Cleaning UR 525 R11: SAFEGUARDS

SUMMARY

EVENT UST (SSEL)-

TOR TUBES FOLLOWING CHEMICAL CLEANING.

Bubble Transport Equation NUR G/CR-4501: MODEUNG OF VAPOR GENERATION IN FLASHING

/

-344 HE PACT OF LWR DECONTAMINATIONS ON SOLIDIFICATON. WASTE DISPOSAL.AND ASSOCIATED OCCU-Budget PATIONAL EXPOSURE.

NUREG-1100 V02: FY 1987 BUDGET ESTIMATES.

CM Prp Building Wasts fiUREG/CR-4453 V01: UGHT WATER-REACTOR SAFETY FUEL SYS-NUREG/CR-4113: FLOW AND DISPERSION NEAR CLUSTERS OF TEMS RESEARCH PROGRAMS. Quarterly Progress Report. January.

BUIL3NGS.

March 1985.

Burial Cladd6ng NUREG-1101 V01: ONSITE DISPOSAL OF RADIOACTIVE NUREG/CR-4219 V02: HEAVY-SECTON STEEL TECHNOLOGY PRO-WASTE. Guidance For Disposal By Subsurface Bunal.

GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER 1985.

COMMIX-1E NUREG/CR-4485: THE IMPACT OF FUEL Cl. ADDING FAILURE NUREG/C34348 V01: COMMIX-1B.A THREE-DIMENSIONAL TRAN-EVENTS ON OCCUPATONAL RADIATON EXPOSURES AT NUCLE-SIENT S NGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-AR POWER PLANTS. Case Study PWR Dunng Routine Operations.

DRAULIC ANALYSIS OF SINGLE AND MULTICOMPONENT SYSTEMSVol1. Equations And Numencs.

Class C Concentration j

NUREG/CR-4348 V02: COMMIX-18:A THREE DIMENSIONAL TRAN-NURTG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-METHODOLOGY. Volume 1: Methodology Report.

DRAULIC ANALYSIS OF SINGLE AND MULTICOMPONENT NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYS!S SYSTEMS.Velll. User's Manual.

METHODOLOGY. Volume 2: Codes And Example Problems.

COMMIX-2 Closecut NUREG/CR-4371: COMMIX-2:A THREE-DIMENSIONAL TRANSIENT NUREG/CR-3790- CLOSEOUT OF IE BULLETIN 79-07; SEISMIC COMPUTER PROGRAM FOR THERMAL-HYDRAUUC ANALYSIS OF STRESS ANALYSIS OF SAFETY-RELATED PIPING.

TWO-PHASE FLOWS.

Code CONTAIN NUREG/CR4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 (VER-NUREG/CR-4255 V02: AEROSOL RELEASE AND TRANSPORT PRO-SiON 7.0).RELAP5/ MOD 1(CYCLE 14).AND TRAC-BD1 (VERSION GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 12.0) CODES USING SEPARATE-EFFECTS EXFERIMENTS.

1985.

NUREG/CR-4442: TRAC USER'S GUIDE.

NUREG/CR-4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT NUREG/CR-4452: REVIEW OF RELAPS CALCULATIONS FOR H.B.

ANALYSIS WITH THE CONTAIN COMPUTER CODE.

ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

Subject index 39 NUREG/CR-4465:

TRAC-PF1/ MOO 1 INDEPENDENT Containment ASSESSMENTSEMISCALE MOD-2A INTERMEDIATE BREAK TEST NUREG/CR-4223. STEEL CONTAINMENT RESISTANCE UNDER GEN-S-IB-3.

ERAL DYNAMIC PRESSURES.

NUREG/CR4468: ADAPTATION OF OCA-P,A PROBABluSTIC FRAC-NUREG/CR4366: REUABILITY ASSESSMENT OF CONTAINMENT TURE-MECHANICS CODE.TO A PERSONAL COMPUTER.

TANGENTIAL SHEAR FAILURE.

NUREG/CR-4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM NUREG/CR4547: CONTEMPT 4/ MOD 6 A MULTICOMPONENT SYSTEM ANALYSIS PROGRAM.

ANALYSIS PROGRAM.

Code Verification Contamination NUREG/CR-4453 V01: UGHT-WATER-REACTOR SAFETY FUEL SYS-NUREG-1101 V01: ONSITE DISPOSAL OF RADIOACTIVE TEMS RESEARCH PROGRAMS. Quarterty Progress Report. January-WASTE. Guidance For Disposal By Subsurface Bunal.

March 1985.

NUREG/CR-4286: EVALUATION OF RADIOACTIVE UQUID EFFLUENT RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

Combustible Gas Production NUREG/CR4420- TURC1:LARGE SCALE METALUC MELT CONCEN-Control Room TRATE INTERACTION EXPERIMENTS AND ANALYSIS.

NUREG/CR 4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-HAVIOR IN COMPLEX NUCLEAR POWER PLANT G

THE USE OF A FIELD MODEL TO ASSESS FIRE BE-HAVIOR IN COMPLEX NUCLEAR POWER PLANT Core ENCLOSURES.PRESENT CAPABluTIES AND FUTURE PROSPECTS.

NUREG/CR-4465:

TRAC PF1/ MOD 1 INDEPENDENT Cup h A hSSMENT:SEMISCALE MOD-2A INTERMEDIATE BRE NUREG/CR-4255 V02 AEROSOL RELEASE AND TRANSPORT PRO-GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER Core Damage 1985.

NUREG/CR-4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT Core Heatup ANALYSIS WITH THE CONTAIN COMPUTER CODE.

NUREG/CR4434: ASSESSMENT OF MODELUNG NEEDS FOR HUREG/CR-4?48 VOI: COMMIX-18:A THREE-DIMENSIONAL TRAN-SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY.

ORAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT Core Melt EYSTEMS.Vol LEquations And Numencs.

NUREG/CR 3958: EFFECTS OF CONTROL SYSTEM FAILURES ON NUREG/CR-4348 V02-COMMIX-18.A THREE-DIMENSIONAL TRAN-TRANSIENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A SIINT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY*

COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR.

DRAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT NUREG/CR-4459: LIGHT WATER REACTOR SAFETY RESEARCH SYETEMS.Voi tt: User's Manual PROGRAM. Semiannual Report. October 1983 - March 1984.

NUREG/CR4371: COMMIX-2:A THREE-DIMENSIONAL TRANSIENT NUREG/CR-4462 A RANKING OF SABOTAGE / TAMPERING AVOID-COMSUTER PROGRAM FOR THERMAL-HYDRAlJUC ANALYSIS OF TWO 3HASE FLOWS.

ANCE TECHNOLOGY ALTERNATIVES.

NUREGiCR-4459 LIGHT WATER REACTOR SAFETY RESEARCH Corrosion PROGPAM. Semiannual Report. October 1983 - March 1984.

NUREG/CR-4460 UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL TESTS. Semiannual Report Covenng The Penod April-September 1985.

MODEL Cost NUREG/CR-4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS HAVIOR IN COMPLEX NUCLEAR POWER PLANT ENCLOSURES.PRESENT CAPABluTIES AND FUTURE PROSPECTS.

METHODOLOGY Volume 1: Methodology Report.

NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS NUREG/CR 4486: VISA II - A COMPUTER CODE FOR PREDICTING METHODOLOGY. Volume 2-Codes And Example Problems.

NUREG/

-4 45 ROBAB USTIC SAF S D A l'ONS

^

PROGRAM FOR INSPECTION OF INDIAN POINT UNIT 3 NUCLEAR Method For Estimating Labor Construction Costs Associated With POWER PLAi4T.

Physical Modifications To Nuclear Power Plants.

NUREG/CR-4555: GENERIC COST ESTIMATES FOR THE DISPOSAL Computer Program OF RADIOACTIVE WASTES.

NUREG/CR-4222: STEEL CONTAINMENT RESISTANCE UNDER GEN.

ERAL DYNAMIC PRESSURES' Counter-Current Flow NUREG/CR-4337:

TRAC PF1/ MODI INDEPENDENT Computer Security ASSESSMENT DARTMOUTH COLLEGE AIR-WATER COUNTER-CUR-NUREG/CR-4473: 4 STUDY OF THE OPERATION AND MAINTENANCE RENT FLOW TESTS.

OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 CFR T3.55.

Crack NUREG4975 V04: COMPILAT*ON OF CONTRACT RESEARCH FOR Concrete THE MATERIALS ENGINEERING BRANCH, DIVISION OF ENGINEER-NUREG/CR-4328: PFOBABiLITY BASED LOAD COMBINATION CRITE.

ING TECHNOLOGY. Annual Report For FY 1985.

RIA FOR DESIGN CF SHEAR WALL STRUCTURES.

NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM - PHASE ll. Semiannual Report.Apri 1985 - September 1985.

Concrete Ablation NUREG/CR-4545: PIPE CRACK EVALUATION IN OPERATING BOILING NUREG/CR-4420: TURC1:LARGE SCALE METALUC MELT-CONCEN-WATER REACTORS.

TRATT INTERACTION EXPERIMENTS AND ANALYSIS.

Crack Growth Condensation NUREG/CR4219 V02-HEAVY-SECTION STEEL TECHNOLOGY PRO-NUREG/CR-4255 V02: AEROSOL RELEASE AND TRANSPORT PRO-GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1985.

1985.

NUREG/CR-4528: DEVELOPMENT AND VERIFICATION OF CONDI-TIONS FOR DUCTILE TEARING INSTABILITY AND ARREST.

Congress NUREG-0090 V08 NO3: REPORT TO CONGRESS ON ABNORMAL Cylinder Rupture OCCURRENCES.Jufy - September 1985.

NUREG-1179 V01: RUPTURE OF MODEL 48Y OF6 CYUNDER AND RE-LEASE OF URANIUM HEXAFLUORIDE.Sequoyah Fuels Construction Facihty, Gore, Oklahoma January 4,1986.

NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR ENVI-RONMENT -lMPLICATIONS FOR NRC QUAUTY ASSURANCE PRO-DCC-1/DCC-2 GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-NUREG/CR-4390- DCC-1/DCC-2 DEGRADED CORE COOLABluTY TORY PROJECTIONS THROUGH 1995.

ANALYSIS.

40 Subject Index Detris NUREG/CR-4289: RESIDUAL RADIONUCUDE CONTAMINATION NUREG/CR-3983: STEAM EXPLOSION EXPERIMENTS AT INTERMEDI-WITHIN AND AROUND COMMERCIAL NUCLEAR POWER ATE SCALE.FITSB SERIES.

PLANTS. ORIGIN. DISTRIBUTION, INVENTORY AND DECOMMISSION-ING ASSESSMENT.

Decay-Heat-Removal NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS MUREG/CR-4471: LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUD-METHODOLOGY. Volume 1: Methodology Report.

IES

SUMMARY

RESULTS AND CONCLUSIONS-NUREG/CR-4370 V02: UPDATE OF PART Pt 1MPACTS ANALYSIS Decision Mak6ng METHODOLOGY. Volume 2. Codes And Example Problems.

NUREG/CR4446; THE NUCLEAR INDUSTRY AND ITS NUREG/CR4433: DOCUMENT REVIEW REGARDING HAZARDOUS REGULATOF.S A NEW COMPACT IS NEEDED' CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE-NUREG/CR-4555: GENERIC COST ESTIMATES FOR THE DISPOSAL Decommission 6ng OF RADIOACTIVE WASTES.

MUREG/CP-0068. PROCEEDINGS OF THE INTERNATIONAL NUCLEAR REACTOR DECOMMISSIONING PLANNING CONFERENCE.

Dissolution

]

WUREG/CR-3959. TRANSeTION TO AN OPERATING REACTOR ENVI.

NUREG/CR-4379 V02: LONG-TERM PERFORMANCE OF MATERIALS l

RONMENT -lMPUCATIONS FOR NnC OUAUTY ASSURANCE PRO.

USED FOR HIGH-LEVEL WASTE PACKAGING Second Ouarterly GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-Report, Year Four July-September 1985.

TORY PROJECTIONS THRO)GH 1995.

NUREG/CR-4289: RESIDUAL RADIONUCUDE CONTAMINATION Dissolution Behavior WITH'N AND AROUND COMMERCIAL NUCLEAR POWER NUREG/CR4379 V03. LONG-TERM PERFORMANCE OF MATERIALS PLANTS. ORIGIN. DISTRIBUTION. INVENTORY AND DECOMMISSION-USED FOR HIGH-LEVEL WASTE PACKAGING. Third Quarterty ING ASSESSMENT.

Report, Year Four October -December 1985.

MUREG/CR4504: LONG-TERM SURVEILLANCE AND MONITORING OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL-Dose l

INGS PILES NUREG-1101 V01: ONSITE DISPOSAL OF RADIOACTIVE WASTE. Guidance For Dsposal By Subsurface Bunal.

Decontamination NUREG/CR4000 V01: THE MESORAD DOSE ASSESSMENT NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS MODELVolume 1: Technical Basis.

ON SOUDIFICATION. WASTE DISPOSAL,AND ASSOCIATED OCCU-NUREG/CR-4286; EVALUATION OF RADIOACTIVE LIQUID EFFLUENT PATIONAL EXPOSURE-RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF REG /CR 4472: SIAMESE IMAGING TECHNIQUE FOR QU ASI-VERTI-S.AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCL CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

Dosimetry Degradat:en NUREG/CR4307 VOI: LWR PRESSURE VESSEL SURVEILLANCE DO-MUREG/CR-4279 V01: AGING AND SERVICE WEAR OF HYDRAULIC SIMETRY IMPROVEMENT PROGRAM Progress Report - October AND MECHANICAL SNUBBERS USED ON SAFETY RELATED PIPING 1984 - September 1985.

AND COMPONENTS OF NUCLEAR POWER PLANTS Phase i Study NUREG/CR-4380- EVALUATION OF THE MOTOR-OPERATED VALVE Downcomer ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT NUREG/CR-4465:

TRAC-PF1/ MOD 1 INDEPENDENT DEGRADATION. INCORRECT ADJUSTMENTS.AND OTHER ABNOR.

ASSESSMENT.SEMISCALE MOD-2A INTERMEDIATE BREAK TEST MALITIES IN MOTOR-OPERATED VALVES.

S-lB 3.

Degraded Core Accident Dryout Heat Flux NUREG/CR-4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM NUREG/CR-4390- DCC-1/DCC-2 DEGRADED CORE COOLABluTY ANALYSIS FROGRAM.

ANALYSIS.

Degra'.1ed Core Coolability Ductile Cast Iron RUREG/CR43SO. DCC-1/DCC-2 DEGRADED CORE COOLABILITY NUREG/CR-3760- A STUDY ON DUCTILE AND BRITTLE FAILURE ANALYSIS DESIGN CRITERIA FOR DUCTILE CAST IRCN SPENT-FUEL SHIP-Descriptors PING CONTAINERS.

WUREG/CR-3950 V02-FUEL PERFORMANCE ANNUAL REPORT FOR Ductile Tearing Instability NUREG/CR-4528: DEVELOPMENT AND VERIFICATION OF CONDI-Design Criteria TIONS FOR DUCTILE TEARING INSTABiUTY AND ARREST.

NUREG/CR-3760: A STUDY ON DUCTILE AND BRITTLE FAILURE DES G ERIA FOR DUCTILE CAST 1RON SPENT-FUEL SHIP-NU

/CR-3805 V03: ENGINEERING CHARACTERIZATION OF NUREG/CR-4328: PROBABILITY BASED LOAD COMBINATION CRITE-GROUND MOTION. Task 11: Observational Data On Spatial Vanations RIA FOR DESIGN OF SHEAR WALL STRUCTURES.

Of Earthquake Ground Motson.

NUREG/CR-4293: REUABluTY ANALYSIS OF SHEAR WALL STRUC-Destruction TURES.

WUREG/CR-4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED NUREG/CR4328: PROBABluTY BASED LOAD COMBINATION CRITE.

WASTES IN COMMERCIAL LOW-LEVEL WASTES. Draft Report For RIA FOR DESIGN OF SHEAR WALL STRUCTURES.

Comment.

NUREG/CR-4366. RELIABluTY ASSESSMENT OF CONTAINMENT l

TANGENTIAL SHEAR FAILURE.

i D6 gest NUREG/CR-4502: VIRGINIA REGIONAL SEISMIC NETWORK. FINAL WUREG-0750 V22102: INDEXES TO NUCLEAR REGULATORY COM-REPORT (1977-1985).

I MISSION ISSUANCES FOR JULY-DECEMBER 1985.

Eddy Current D6 graph Matrix Analysis NUREG-0975 V04: COMPILATION OF CONTRACT R*: SEARCH FOR Direct Radiation Monitoring Network THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER-WUREG-0837 VOS NO3. NRC TLD DIRECT RADIATION MONITORING ING TECHNOLOGY. Annual Report For FY 1985.

s NETWORK. Progress Report, July-September 1985.

Education D6spers6on NUREG/CR-4411: ASSESSMENT OF SPECIALIZED EDUCATIONAL WUREG/CR-4113: FLCW AND DISPERSION NEAR CLUSTERS OF PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS.

BUILDINGS.

l Effectivu Peak Accoloration Disposal NUREG/CR-3805 V03: ENGINEERING CHARACTERIZATION OF WUREG 1101 V01: ONSITE DISPOSAL OF RADIOACTIVE GROUND MOTION. Task 11: Observational Data On Spatal Variations WASTE Guidance For Dsposal By Subsurface Bunal-Of Earthquake Ground Motion.

Subject index 41 Effluent Monitor NUREG/CR-4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL-NUREG/CR-4299: PRELIMINARY EVALUATION OF EFFLUENT RADIO-lARY FEECWATER SYSTEM RELIABiUTY ANALYSIS.

ACTMTY MONITORING SYSTEMS FOR BWR PLANTS.

NUREG/CR43C3: RELIABluTY ASSESSMENT OF CONTAINMENT TANGENTIAL SHEAR FAILURE.

Elastic-Plastic Fracture Mechanics NUREG/CR4380: EVALUATION OF THE MOTOR OPERATED VALVE NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM - PHASE ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT II.Senuannual Report.Apnt 1985 - September 1985.

DEGRADATON. INCORRECT ADJUSTMENTS.AND OTHER ABNOR-MALITIES IN MOTOR 4PERATED VALVES.

U

/CR-3365 C: REPORT TO THE NRC ON GUIDANCE FOR NUREG/CR-4485: THE IMPACT OF FUEL CLADDING FAILURE PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX-EVENTS ON OCCUPATIONAL RADIATON EXPOSURES AT NUCLE-ERCISES AT NUCLEAR GENERATING STATIONS. Draft Report For AR POWER PLANTS. Case Study PWR Dunng Roubne Operations.

Comment NUREG/CR4486: VISA II - A COMPUTER CODE FOR PREDICTING NUREG/CR-4000 V01: THE MESORAD DOSE ASSESSMENT THE PROBABluTY OF REACTOR PRESSURE VESSEL FAILURE.

MODELVolume 1: Technical Basis.

Enforcement Action NUREG/CR4446:

THE NUCLEAR INDUSTRY AND ITS NUREG-0940 V04 N04: ENFORCEMENT ACTIONS. SIGN!FICANT AC-REGULATORS:A NEW COMPACT IS NEEDED.

TIONS RESOLVED.Ouarterty Progress Report. October-December 1985.

Feed-And-Bleed Procedure NUREG/CR4471: LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUD-G CR-4 A SESSMENT OF CPECIALIZED EDUCATONAL PROGRAMS FOR UCENSED NUCLEAR REACTOR CPERATORS..

Ferrttic Steel Entropy NUREG/CR-4219 V02: HEAVY-SECTION STEEL TECHNOLOGY PRO.

I NUREG/CR-4514: CONTROLUNG PRINCIPLES FOR PRIOR PROBA.

GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER L

BluTY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

1985.

Environmental Measurement Fleid Model NUREG/CR-4286: EVALUATION OF RADIOACTIVE LOUID EFFLUENT NUREG/CR-4479: THE USE OF A FIELD MO9EL TO ASSESS FIRE BE.

RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

HAVIOR IN COMPLEX NUCLEAR POWER PLANT i

Environmental Statement j

NUREG-1171 DRFT: DRAFT ENVIRONMENTAL STATEMENT RELATED Finite-Difference Procedures TO OPERATION OF THE SOUTH TEXAS PROJECT, UNITS 1 AND NUREG/CR4348 V01: COMMIX 18:A THREE-DIMENSIONAL TRAN-

2. Docket Nos. 50-498 And 50499(Houston Ughting And Power Com-SIENT SINGLE PHASE COMPUTER PROGRAM FOR THERMAL HY-pany)

DRAULIC ANALYSIS OF SINGLE AND MULTICOMPONENT Equipment Qualification SYSTEMS.Vol 1: Equations And Numencs.

NUREG/CR-3137: SEISMIC AND DYNAMIC OUAUFICA'ON OF RE-NUREG/CR-4371: COMMIX-2:A THREE-DIMENSONAL TRANSIENT LATED ELECTRICAL AND MECHANICAL EQUIPMENT.

COMPUTER PR'XiRAM FOR THERMAL-HYDRAULIC ANALYSIS OF NUREG/CR-4310: INVESTIGATION OF POTENTIAL FIRE-RELATED TWO-PHASE FLOWS.

DAMAGE TO SAFETY-RELATED EOUfPMENT IN NUCLEAR POWER PLANTS.

Finite-Element Analysis NUREG/CR-4475: ORMGEN.PCA MICROCOMPUTER PROGRAM FOR Erosion AUTOMATIC MESH GENERATION OF 2-D CRACK GEOMETRIES.

NUREG/CR-4323: THE PROTECTION OF URANIUM TAIUNGS IM-POUNDMENTS AGAINST OVERLAND EROSION.

Fire Damage NUREG/CR-4310- INVESTIGATION OF POTENTIAL FIRE-RELATED N RE -1189 V01: ASSESSMENT OF THE PUB 4C HEALTH IMPACT TS' FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH FUELS CORPORATION FACluTY AT GORE. OKLAHOMA.Repnnted Fire Protection March 26.1986.

NUREG-1189 V02: ASSESSMENT OF THE PUBLIC HEALTH IMPACT NUREG/CR-4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-HAVIOR IN CCMPLEX NUCLEAR POWER PLANT FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH FUELS CORoO9ATION FACIUTY AT GORE. OKLAHOMA.

ENCLOSURES:PRESENT CAPABluTIES AND FUTURE PROSPECTS.

NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS ON FIC TON. WASTE DISPOSALAND ASSOCIATED OCCU-E/

343: INTEGRATED SEVERE ACCIDENT CONTAINMENT NUREG/CR-4485: THE IMPACT OF FUEL CLADDING FAILURE ANALYSIS WITH THE CONTAIN COMPUTER CODE.

EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE.

"""U N EG/

402 HIGH-TEMPERATURE GAS-COOLED REACTOR FITS 8 SAFETY STUDIES FOR THE OtVISON OF ACCIDENT NUREG/CR-3983: STEAM EXPLOslON EXPERIMENTS AT INTERMEDI-EVALUATION.Ouarterly Progress Report. Apnl 1

• June 30,1985.

ATE SCALE.FITSB SERIES.

NUREG/CR4453 V01: UGHT-WATER-REACTOR GAFETY FUEL SYS-TEMS RESEARCH PROGRAMS. Quarterly Progress Report. January-FY 1987 Budget March 1985.

NUREG-1100 V02: FY 1987 BUDGE ESTIMATES.

Fabrication Criteria

• • "I j

NUREG/CR-4363: A STUDY ON FABRICATON CRITERIA FOR DUC-NUREG/CR-4501: MODEUNG OF VAPOR GENERATION IN FLASHING FLOW.

TILE CAST LRON SPENT-FUEL SHIPPING CONTAINERS.

Failure Flaw NUREG/CR-3760- A STUDY ON DUCTILE AND BRITTLE FAILURE NUREG/CR-4078; PROGRAM FOR FIELD VAUDATION OF THE SYN-DESIGN CRITERIA FOR DUCTILE CAST 1RON SPENT. FUEL SHIP.

THETIC APERTURE FOCUSING TECHNIQUE FOR ULTRASONIC PING CONTAINERS.

TESTING (SAFT UT) Final Report.

NUREG/CR-3958: EFFECTS OF CONTROL SYSTEM FAILURES ON NUREG/CR-4219 V02: HEAVY-SECTION STEEL TECHNOLOGY PRO-TRANSIENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR.

1985.

NUREG/CR-4302 V01: AGtNG AND SERVICE WEAR OF CHECK NUREG/CR-4300 V02: ACOUSTIC EMISSION / FLAW RELATIONSHIP VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF FOR IN-SERVICE MONITORING CF NUCLEAR PRESSURE NUCLEAR POWER PLANTS.

VESSELS. Progress Report, Apnt-September 1985.

42 Subject Index Flood Groundmotion NUREG/CR-4496: A SYSTEM FOR GENERATING LONG STREAM-NUREG/CR-3805 V03. ENGINEERING CHARACTERl2ATION OF FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN GROUND MOTION. Task 11: Observational Data On Spatial Vanations PERIOD.

Of Earthquake Ground Moton.

Fluid Friction Guidance NUREG/CR4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER NUREG/CR-3365 DAF FC: REPORT TO THE NRC ON GUIDANCE FOR NATURAL CIRCULATION CONDITIONS.

PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX-ERCISES AT NUCLEAR GENERATING STATIONS Draft Report For Fluoride Comment i

NUREG-1189 V01: ASSESSMENT OF THE PUBUC HEALTH IMPACT FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH GuldeHnes FUELS CORPORATION FACluTY AT GORE OKLAHOMA.Repnnted NUREG/CR-3517: RECOMMENDATONS TO THE NRC ON HUMAN EN-March 26.1986.

GINEERING GUIDELINES FOR NUCLEAR POWER PLANT MAIN-NUREG-1189 V02: ASSESSMENT OF THE PUBL J HEALTH IMPACT TAINABlUTY.

FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH NUREG/CR-4482 DRF FC-RECOMMENDATIONS TO THE NUCLEAR FUELS CORPORATION FACluTY AT GORE. OKLAHOMA.

REGULATORY COMMISSION ON TRIAL GUIDELINES FOR SEISMIC MARGIN REVIEWS OF NUCLEAR POWER PLANTS. Draft Report For Forced Ventilation Comment NUREG/CR4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-HAVIOR IN COMPLEX NUCLEAR POWER PLANT Guity Erosion ENCLOSURES.PRESENT CAPABIUTIES AND FUTURE PROSPECTS.

NUREG/CR-4323: THE PROTECTION OF URANIUM TAluNGS IM-POUNDMENTS AGAINST OVERLAND EROSION.

NUREG/CR4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER HECTR NATURAL CIRCULATON CONDITIONS.

NUREG/CR-4459: UGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report, October 1983 March 1984.

p NUREG-0975 V04: COMPILATION OF CONTRACT RESEARCH FOR Handbook THE MATERIALS ENGINEERING BRANCH DIVISON OF ENGINEER-NUREG/CR-4188 V01: NUCLEAR POWER PLANT SIMULATION FAClu-ING TECHNOLOGY. Annual Report For FY 1985.

TY EVALUATION METHODOLOGY. Handbook.

NUREG/CR4082 V03: DEGRADED PIPING PROGRAM PHASE lt. Semiannual Report.Aptd 1985 - September 1985.

Hazard NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED i

OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

WASTES IN COMMERCIAL LOW-LEVEL WASTES. Draft Report For NUREG/CR-4183 V02. PRESSURIZE 0 THERMAL SHOCK EVALUATION Comment OF THE H B. ROB'NSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR-4219 V02: HEAVY-SECTION STEEL TECHNOLOGY PRO-Hazardous Weste GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER NUREG/CR-4433: DOCUMENT REVIEW REGARDING HAZARDOUS 1

^

N G/CR 4468: ADAPTATION OF OCA-P,A PROPABluSTIC FRAC-TURE-MECHANICS CODE.TO A PERSONAL COMPUTER.

Health Physics Technique NUREG/CR-4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-NUREG/CR4364: MANAGEMENT PERCEPTION OF THE HEALTH NATION REUABluTY AND FRACTURE MECHANICS: Semiannual PHYSICS TECHNICIAN JOB' Report.Apnl - September 1984.

NUREG/CR4475: ORMGEN PC:A MICROCOMPUTER PROGRAM FOR HeWth Physics Training AUTOMATIC MESH GENERATION OF 2-D CRACK GEOMETRIES.

NUREG-1159: TRAINING MANUAL FOR URANIUM MILL WORKERS ON NUREG/CR-4528: DEVELOPMENT AND VERIFICATION OF CONDI-

^

TIONS FOR DUCTILE TEARING INSTABlWTY AND ARREST, FragiHty Heat Transbr NUREG/CR-4293: REUABluTY ANALYSIS OF SHEAR WALL STRUC-NUREG/CR4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER TdAES NATURAL CIRCULATION CONDITONS.

NUREG/CR4 UABIU ASSESSMENT OF CONTAINMENT Heavy-Section Steel Technology Program NUREG/CR4219 V02: HEAVY-SECTON STEEL TECHNOLOGY PR4 Fug GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER NUREG/CR-3983: STEAM EXPLOSION EXPERIMENTS AT INTERMEDI-M85.

ATE SCALE-FITSB SERIES.

High Temperature Ou!dation Fuel Cycle NUREG/CR-4476: HIGH-TEMPERATURE OXIDATION OF ZlRCALOY-4 NUREG-0525 R11: SAFEGUARDS

SUMMARY

EVENT UST (SSEL).

IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

Fuel Element Failure High-Level Weete NUREG/CR4485: THE IMPACT OF FUEL CLADDING FAILURE NUREG-1168: STAFF EVALUATON OF U.S.

DEPARTMENT OF EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE.

ENERGY PROPOSAL FOR MONITORED RETRIEVABLE STORAGE.

AR POWER PLANTS. Case Study.PWR Dunng Routine Operatons.

NUREG/CR-4379 V02: LONG-TERM PERFCRMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.Second Quarterty Fuel Performance Report Year Four July-September 1985.

NUREG/CR3950 V02: FUEL PERFORMANCE ANNUAL REPORT FOR NUREG/CR-4477: METHODOLOGIES FOR ASSESSING LONG-TERM 1984.

PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-AGES.

Functional Indicator NUREG/CR-4585: INVESTIGATION OF THE STABluTY OF CLAY /

NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK BASALT PACKING MATERIALS.

VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF NUCLEAR POWER PLANTS.

Holography NUREG/CR-4472: SIAMESE IMAGING TECHNIQUE FOR QUASI-VERTI-Generic Safety leaue CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

NUREG-0933 SO4: A PRIORITl2ATON OF GENERIC SAFETY ISSUES.

Human Engineering Grahof Number NUREG/CR3517: RECOMMENDATONS TO THE NRC ON HUMAN EN-NUREG/CR4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER GINEERING GUIDELINES FOR NUCLEAR POWER PLANT MAIN-NATURAL CIRCULATION CONDITIONS.

TAINABluTY.

Subject Index 43 Human Error Incident Response NUREG/CR-2331 VOS N2: SAFETY RESEARCH PROGRAMS SPON-NUREG/CR4000 V01: THE MESORAD DOSE ASSESSMENT SORED BY OFFICE OF NUCLEAR REGULATORY MODELVolume 1: Technical Basis.

RESEARCH.Ouarterty Progress Report.Apnl 1, June 30,1985.

NUREG/CR4132 NUCLEAR POWER SAFETY REPORTING SYSTEM Index FINAL EVALUATION RESULTS.

NUREG-0304 V10 N04: REGULATORY AND TECHNICAL REPORTS. Annual Compilat on for 1985.

Human Factore NUREG-0750 V22102: INDEXES TO NUCLEAR REGULATORY COM-NUREG/CR-3517: RECOMMENDATONS TO THE NRC ON HUMAN EN-MISSION ISSUANCES FOR JULY-DECEMBER 1985 GINEERING GUIDEUNES FOR NUCLEAR POWER PLANT MAIN-TAINABluTY.

Ineervice Inspection NUREG/CR4436 V01: HUMAN REUABILITY IMPACT ON INSERVICE NUREG/CR-4436 V01: HUMAN REUABluTY IMPACT ON INSERVICE INSPECTON. Volume 1: Phase 1 Summary Report.

INSPECTION Volume 1: Phase t Summary Report NUREG/CR-4436 V02-HUMAN RELIABILITY IMPACT ON INSERVOE NUREG/CR4436 V02: HUMAN RELIABluTY IMPACT ON INSERVICE INSPECTON. Volume 2-Review And Anatyss Of Human Performance INSPECTION. Volume 2: Review And Anafyss Of Human Performance l

In Nondestructive Teshng (Emphasiang Ultrasonics).

In Nondestructive Tesbng (Emphasiang Ultrasonics).

l Human Performance Inservice Monitoring NUREG/CR-4132-NUCLEAR POWER SAFETY REPORTING SYSTEM NUREG/CR4300 V02: ACOUSTIC EMISSION / FLAW RELATONSHIP FINAL EVALUATION RESULTS.

FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Progress Report, April-September 1985.

NUREG/CR-4279 V01: AGING AND SERVICE WEAR OF HYDRAULIC Inspection AND MECHANICAL SNUBBERS USED ON SAFETY-RELATED PIPING NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK AND COMPONENTS OF NUCLEAR POWER PLANTS. Phase i Study.

VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF NUCLEAR POWER PLANTS.

Hydrogen Burn NUREG/CR-4372 PROBABAUSTIC RISK ASSESSMENT (PRA) APPLf-NUREG/CR-4310: INVESTIGATION OF POTENTIAL FIRE-RELATED CATIONS.

DAMAGE TO SAFETY-RELATED EQUIPMENT IN NUCLEAR POWER PLANTS.

Inspection Guidance NUREG/CR4324; TESTING OF NUCLEAR QUAUFIED CABLES AND NUREG/CR-4565 PROBABILISTIC SAFETY STUDY APPLICATONS PRESSURE TRANSMITTERS IN SIMULATED HYDROGEN DEFLA-PROGRAM FOR INSPECTON OF INDIAN POINT UNIT 3 NUCLEAR GRATONS TO DETERMINE SURVIVAL MARGINS AND SENSITIVI-POWER PLANT.

TIES.

NUREG/CR-4547: CONTEMPT 4/ MOD 6A MULTICOMPONENT SYSTEM installation Cost ANALYSIS PROGRAM.

NUREG/CR-4546: LABOR PRODUCTIVITY ADJUSTMENT FACTORS.A Method For Estmahng Labor Construction Costs Associated With NU G/CR 9 IGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report. October 1983 - March 1984.

Instrumentation NUREG/CR4324: TESTING OF NUCLEAR QUAUF;ED CABLES AND N

G1 1: RUPTURE OF MODEL 48Y OF6 CYUNDcR AND RE-LEASE OF URANIUM HEXAFLOO 'IDE.Sequoyah Fuels S'

Facility. Gore.Oldahoma, January 4,1986.

Hypocenter Bulletin Integrated Control System NUREG/CR-4502: VIRGINIA REGIONAL SEISMIC NETWORK. FINAL NUREG-1195: LOSS OF INTEGRATED CONTROL SYSTEM POWER REPORT (1977-1985).

AND OVERCOOUNG TRANSIEN~ AT RANCHO SECO ON DECEM-BER 26,1985.

IE Bulletin 79 07 NUREG/CR-3790: CLOSEOUT OF IE BULLETIN 79-07: SEISMIC InWgranular Skees Cmelon CMng STRESS ANALYSIS OF SAFETY-RELATED PlPING.

NUREG/CR-4436 V01: HUMAN RELIABIUTY IMPACT ON INSERVICE INSPECTION. Volume 1: Phase 1 Summary Report.

IGSCC NUREG/CR4464: PERFORMANCE DEMONSTRATON TESTS FOR NUREG/CR-4464: PERFORMANCE DEMONSTRATION TESTS FOR DETECTON OF INTERGRANULAR STRESS CORROSION CRACK, DETECTON OF INTERGRANULAR STRESS CORROSION CRACK-ING.

ING.

NUREG/CR-4494: RADIOLOGICAL ASSESSMENT OF BWR RECIRCU-LATORY PIPE REPLACEMENT.

Ice Condenser NUREG/CR-4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM Inventory ANALYSIS PROGRAM.

NUREG/CR-4289: RESIDUAL RADIONUCUDE CONTAMINATON WITHIN AND AROUND COMMERCIAL NUCLEAR POWER immobilization PLANTS. ORIGIN.DISTRIBUTON, INVENTORY AND DECOMMISSION-NUREG/CR-4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED ING ASSESSMENT.

WASTES IN COMMERCIAL LOW-LEVEL WASTES. Draft Report For Comment.

Inventory Difference Data NUREG-0430 V06 N01. UCENSED FUEL FACluTY STATUS Immunity REPORT. Inventory Difference Deta. January-June 1985.(Gray Book 11)

NUREG/CR-4132-NUCLEAR POWER SAFETY REPORTING SYSTEM FINAL EVALUATION RESULTS.

Irradiation NUREG/CR-4219 V02-HEAVY-SECTON ETEEL TECHNOLOGY PRO-Impact GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER NUREG-1189 V01: ASSESSMENT OF THE PUBLIC HEALTH IMPACT 1985.

FROM THE ACCOEl4TAL RELEASE OF UF6 AT THE SEOUOYAH FUELS CORPORATICtd FACluTY At GORE, OKLAHOMA. Reprinted J-Integral March 26.1986.

NUREG/CR4082 V03: DEGRADED PIPING PROGaAM - PHASE NUREG 1189 V02: ASSESSMENT OF THE PUBUC HEALTH IMPACT II. Semiannual Report.Apn1 1985. September 1985.

FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH NUREG/CR-4528: DEVELOPMENT AND VERIFICATON OF CONDI-FUELS CORPORATON FACILITY AT GORE. OKLAHOMA.

TIONS FOR DUCTILE TEARING INSTABluTY AND ARREST.

Impact Analysis Job Analysis NUREG/CR4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS NUREG/CR4364: MANAGEMENT PERCEPTION OF THE HEALTH METHODOLOGY. Volume 1: Methodology Report.

PHYSICS TECHNICIAN JOB.

44 Subject index Key Curve Method Load NUREG/CR-4528. DEVELOPMENT AND VERIFICATION OF CONDl-NUREG/CR-4293: RELIABluTY ANALYSIS OF SHEAR WALL STRUC-TIONS FOR DUCTILE TEARING INSTABluTY AND ARREST.

TURES.

NUREG/CR 4368: RELIABILITY ASSESSMENT OF CONTAINMENT l

N REG /CR-3950 V02: FUEL PERFORMANCE ANNJAL REPORT FOR 1984.

Load Combination NUREG/CR-2331 */05 N2: SAFETY RESEARCH PROGRAMS SPON-g SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-2000 V04N12: UCENSEE EVENT REPORT (LER)

COMPILATION For Month Of December 1985.

RESEARCH.Ouarterly Progress Report. April 1-June 30,1985.

f NUREG/CR-2000 V05 N1: UCENSEE EVENT REPORT (LER)

NU E C 5 2 U SEE EVENT REPORT (LER)

RE /

328: PROBABluTY BASED LOAD COMBINATION CRITE-COMPILATION For Month Of February 1986.

RlA FOR DESIGN OF SHEAR WALL STRUCTURES-LOADS Loop Blowdown Investigation NUREG/CR-4223: STEEL CONTAINMENT RESISTANCE UNDER GEN-NUREG/CR4171:

TRAC-PF1/ MOD 1 INDEPENDENT ERAL DYNAMIC PRESSURES.

ASSESSMENT: LOBI LARGE BREAK TRANSIENT A1-04R.

LO61 Large Break Transient A1-04R Loss Of Oc Power NUREG/CR-4171:

TRAC-PF1/ MOD 1 INDEPENDENT NUREG-1195: LOSS OF INTEGRATED CONTROL SYSTEM POWER ASSESSMENT. LOBI LARGE BREAK TRANSIENT A1-04R.

AND OVERCOOLING TRANSIENT AT RANOHO SECO ON DECEM-BER 26,1985.

NUREG/CR-4393:

SUMMARY

OF SEMISCALE SMALL BREAK LOSS-Loss Of Power OF-COOLANT ACCIDENT EXPERIMENTS (1979 TO 1985)

NUREG-1190: LOSS OF POWER AND WATER HAMMER EVENT AT NUREG/CR-4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM SAN ONOFRE. UNIT 1 ON NOVEMBER 21,1985.

ANALYSIS PROGRAM.

Lose-Of-Coolant Accident R

CR-4546: LABOR PRODUCTIVITY ADJUSTMENT FACTORS.A NUREG/CR-4393.

SUMMARY

OF SEMISCALE SMALL BREAK LOSS-Method For Estimanng Labor Construction Costs Associated With NURE 38 E TS M

LE 2

ALL BREAK Physical Modif> cations To Nuclear Power Plants.

(5%) LOSS-OF COOLANT ACCIDENT EXPERIMENTS S-LH-1 AND S-Large Irrad6ator Radiation Safety LH-2.

NUREG/CP-0073 PROCEEDINGS OF THE WORKSHOP ON LARGE IR-RADIATOR RADIATION SAFETY.

Low-Level Weste NUREG/CR4289: RESIDUAL RADIONUCUDE CONTAMINATION Leaching WITHIN AND AROUND COMMERCIAL NUCLEAR POWER NUREG/CR4379 V02: LONG-TERM PERFORMANCE OF MATERIALS PLANTS. ORIGIN. DISTRIBUTION. INVENTORY AND DECOMMISSION-USED FOR HIGH-LEVEL WASTE PACKAGING.Second Ouarterty ING ASSESSMENT.

Report, Year Four July-September 1985.

NUREG/CR4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS j

NUREG/CR-4379 V03: LONG-TERM PERFORMANCE OF MATERIALS METHODOLOGY. Volume 1: Methodology Report.

USED FOR HIGH-LEVEL WASTE PACKAGING. Third Quarterty NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS 1

l Report Year Four October -December 1985.

METHODOLOGY. Volume 2 Codes And Examp6e Problems.

NUREG/CR-4433: DOCUMENT REVIEW REGARD!NG HAZARDOUS 1

Lead l

CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

NUREG/CR 4433; DOCUMENT REVIEW REGARDING HAZARDOUS NUREG/CR-4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

WASTES IN COMMERCIAL LOW-LEVEL WASTES. Draft Report For Comment Legal lssuances NUREG 0750 V22 N05-NUCLEAR REGULATORY COMMISSION IS-MC&A NUREG/CR-4059: EVALUATION OF THE IMPACT OF THE MC&A NU EG0 2 06 LA E

OR COMMISSION IS-REFORM AMENDMENTSON A REPROCESSING FACluTY.

SUANCES FOR DECEMBER 1985 Pages 875-982.

NUREG4750 V23 N01: NUCLEAR REGULATORY COMMISSION IS.

SUANCES FOR JANUARY 1986 Pages 147.

NUR G/CR-4460- UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN Ucensed Operating Reactors UPPER PLENUM TEST PROBLEM FOR THE MAFROS AEROSOL NUREG-0020 V09 N12: UCENSED OPERATING REACTORS STATUS MODEL SUMMA' ~4EPORT. Data As Of November 30,1985.(Gray Book 1)

NUREG-C "O N01: UCENSED OPERATING REACTORS STATUS MESORAD SUMMAn

_ PORT Data As Of December 31,1985 (Gray Book 1).

NUREGICR-4000 V01: THE MESORAD DOSE ASSESSMENT NUREG-0020 V10 NO2-UCENSED OPERATING REACTORS STATUS MODELVolume 1:Techrncal Basis.

SUMMARY

REPORT. Data As Of January 31,19864 Gray Book f)

MINET Ucensee Contractor And Vendor inspection NUREG/CR-2331 V05 N2-SAFETY RESEARCH PROGRAMS SPON-NUREG-0040 V09 N04: UCENSEE CONTRACTOR AND VENDOR IN-SORED BY OFFICE OF NUCLEAR REGULATORY SPECTION STATUS REPORT. October 1985-December 1985.(White RESEARCH. Quarterly Progress Report. April 1-June 30,1985.

Book)

MOVATS Ucensee Event Report NUREG/CR-4380 EVALUATION OF THE MOTOR OPERATED VALVE NUREGICR-2000 V04N12-LICENSEE EVENT REPORT (LER)

ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT DEGRADATION,1NCORRECT ADJUSTMENTS,AND OTHER ABNOR.

NU E C VOS 1

E

/ENT REPORT (LER)

MAUTIES IN MOTOR 4PERATED VALVES.

d COMPILATION For Month Of January 1986.

NUREG/CR-200C V05 N2: UCENSEE EVENT REPORT (LER)

COMPILATION.For Month Of February 1986.

CR-3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN-Uquid Effluent GINEERING GUIDELINES FOR NUCLEAR POWER PLANT MAIN-N!! REG /CR 2907 V03: RADIOACTIVE MATERIALS RELEASED FROM TAINABILITY.

l NUCLEAR POWER PLANTS. Annual Report 1982.

Uguld Waste NUREG/CR-3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN-NUREG/CR4286: EVALUATION OF RADIOACTIVE LIQUID EFFLUENT GINEERING GUIDEUNES FOR NUCLEAR POWER PLANT MAIN-RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

TAINABILITY.

f

Subject index 45 7.;c ; -...t NUREG/CR-4504: LONG-TERM SURVEILLANCE AND MONITORING NUREG/CR-4364: MANAGEMENT PERCEPTION OF THE HEALTH OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL-PHYSICS TECHNICIAN JOB.

INGS PILES.

NUREG/CR-4378: OBJECTIVE INDICATORS OF ORGANIZATIONAL PERFORMANCE AT NUCLEAR POWER PLANTS.

Motor Operator NUREG/CR4380 EVALUATON OF THE MOTOR-OPERATED VALVE Manning-Kinet;c Equation ANALYSIS AND TEST SYSTEM (MOVATS) T3 DETECT NUREG/CR4323: THE PROTECTION OF URANIUM TAluNGS IM-DEGRADATION,1NCORRECT ADJUSTMENTS.AND OTHER ABNOR.

POUNDMENTS AGAINST OVERLAND EROSION-MALITIES IN MOTOR-OPERATED VALVES.

Mass Spectromtry Multiphoton lonization NUREG/CR-4419-BIOASSAY MEASUREMENTS FOR URANIUM USING NUREG/CR-4419: BIOASSAY MEASUREMENTS FOR URANIUM USING 4

SPUTTER INITIATED RESONANCE ONIZATION SPECTROSCOPY.

SPUTTER INITIATED RESONANCE ONIZATON SPECTROSCOPY.

MaterialControl And Accounting NRC Regulatory Agenda NUREG/CR-4059: EVALUATION OF THE IMPACT OF THE MC&A NUREG-0936 V04 N04: NRC REGULATORY AGENDA Quarterty Report, REFORM AMENDMENTS ON A REPROCESSING FACIUTY.

October-December 1985.

Mathematical Model NSPKTR NUREGICR4477: METHODOLOGIES FOR ASSESSING LONG-TERM NUREG/CR 4565: PROBABluSTIC SAFETY STUDY APPUCATONS PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-PROGRAM FOR INSPECTON OF INDIAN POINT UNIT 3 NUCLEAR POWER PLANT.

Mechanical Snubber Na NUREG/CR4279 V01: AGING AND SERVICE WEAR OF HYDRAUUC AND MECHANICAL SNUBBERS USED ON SAFETY-RELATED PIPING UR G/

56: HEAT TRANSFER FROM A ROD BUNDLE UNDER AND COMPONENTS OF NUCLEAR POWER PLANTS. Phase i Study.

NATURAL CIRCULATON CONDITIONS.

Melt Concrete Heat Transfer Noise NUREG/CR-4420- TURC1.LARGE SCALE METALUC MELT-CONCEN.

NUREG/CR-4389: PPESSURE NOISE IN PRESSURIZED WATER RE-TRATE INTERACTION EXPERIMENTS AND ANALYSIS.

ACTORS.

Meteorology Non-Equilibrium Phase Change NUREG/CR-4113: FLOW AND DISPERSION NEAR CLUSTERS OF NUREG/CR-4501: MODEUNG OF VAPOR GENERATON IN FLASHING BUILDINGS.

FLOW.

Methodology Nondestructive Examination NUREG/CR-4188 V02: NUCLEAR POWER PLANT SIMULATION FACIU-NUREG-0975 V04: COMPtLATON OF CONTRACT RESEARCH FOR TY EVALUATION METHODOLOGY.Techrucal Bases.

THE MATERIALS ENGINEERING BRANCH, DIVISION OF ENGINEER-NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS ING TECHNOLOGY. Annual Report For FY 1985.

METHODOLOGY. Volume 1: MethodolognReport NUREG/CR-4078: PROGRAM FOR FIELD VALIDATION OF THE SYN-NUREG/CR-4477: METHODOLOGIES Fw ASSESSING LONG-TERM THETIC APERTURE FOCUSING TECHNIOUE FOR ULTRASONIC PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-TESTING (SAFT UT) Final Report.

AGES.

NUREG/CR4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-NATON RELIABluTY AND FRACTURE MECHANICS. Semiannual Microcomputer Program Report,Apnl - September 1984.

NUREG/CR-4475: ORMGEN PC.A MICROCOMPUTER PROGRAM FOR AUTOMATIC MESH GENERATION OF 2-D CRACK GEOMETRIES.

Nondestructive Testing NUREG/CR-4436 V01: HUMAN RELIABfUTY IMPACT ON INSERVICE 80d'38i"9 INSPECTION. Volume 1: Phase 1 Summary Report.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR-4436 V02: HUMAN REUABluTY IMPACT ON INSERVICE TESTS Serrmannual Report Covenng The Penod Apni-September 1985-INSPECTION. Volume 2: Revew And Analyss Of Human Performance MoMcahon in Nondestructive Testing (Emphasang Ultrasorucs).

NUREG/CR-4464: PERFORlWANCE DEMONSTRATION TESTS FOR NUREG/CR-4546: LABOR PRODUCTIVITY ADJJSTMENT FACTORS.A DETECTION OF INTERGRANULAR STRESS CORROSION CRACK-Method For Estimating Labor Constructen Costs Assoctated With ING Phystal ModMcations To Nuclear Power Piarts-NURE' /CR 4472-SIAMESE IMAGING TECHNIOUE FOR OUASI-VERTI-G NUREG/CR-4555: GENERIC COST ESTIMATES FOR THE DISPOSAL OF RADIOACTIVE WASTES.

CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

Modular Reactor Design Nuclear Generation NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR ENVI-NUREG/CR-4434: ASSIS$ MENT OF MODELUNG NEEDS FOR SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS' RONMENT lMPLICATIONS FOR NRC OUAUTY ASSURANCE PRO-GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-Molten Core - Concrete Irderaction TORY PROJECTIONS THROUGH 1995.

NUREG/CR-4420: TURC11ARGE SCALE METALUC MELT CONCEN-TRATE INTERACTION EXPER!MENTS AND ANALYSIS.

Nuclear Steam Supply System Vendor NUREG/CR-3959: TRANSITON TO AN OPERATING REACTOR ENVI.

Monitored Retrievable Storage RONMENT -IMPUCATIONS FOR NRC OUAUTY ASSURANCE PRO-NUREG-1168-STAFF EVALUATON OF U.S.

DEPARTMENT OF GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-ENERGY PROPOSAL FOR MONITORED RETRIEVABLE STORAGE.

TORY PROJECTIONS THROUGH 1995.

Monitoring Numerical Analysis NUREG/CR.4299. PRELIMINARY EVALUATION OF EFFLUENT RADIO.

NUREG/CR-4348 V01: COMMIX 1B.A THREE-DIMENSIONAL TRAN-i ACTIVITY MONITORING SYSTEMS FOR BWR PLANTS SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-NUREG/CR4300 V02: ACCUSTIC EMISSION / FLAW RELATIONSHIP DRAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE SYSTEMS.Vol 1. Equations And Numercs.

VESSELS Progress Repcyt Apnt-September 1985.

NUREG/CR4348 V02: COMMIX-18-A THREE-DIMENSIONAL TRAN.

NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-VALVES USED IN EN31NEERED SAFETY-FEATURE SYSTEMS OF DRAULIC ANALYSIS OF SINGLE AND MULTICOMPONENT NUCLEAR POWER PLANTS.

SYSTEMS.Vol ILUser's Marmal.

NUREG/CR-4380 EVALUATION OF THE MOTOR-OPERATED VALVE ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT OCA-P DEGRADATION. INCORRECT ADJUSTMENTS.AND OTHER ABNOR-NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON MALITIES IN MOTOR-OPERATED VALVES.

OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

1 i

l

46 Subject index NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR 4452: REVIEW OF RELAPS CALCULATIONS FOR H B.

OF THE H.B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

NUREG/CR-4468. ADAPTATION OF OCA-P.A PROBABILISTIC FRAC-TURE MECHANICS CODE,TO A PERSONAL COMPUTER.

Packaging NUREG/CR 4477: METHODOLOGIES FOR ASSESSING LONG-TERM ORMGEN PC PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-NUREG/CR-4470 ORMGEN PC.A MICROCOMPUTER PROGRAM FOR AGES.

AUTOMATIC M'SH GENERATION OF 2-D CRACK GEOMETRIES.

Packing BAS P KlN NU G E M CT OF FUEL CLADDING FAILURE RA EVENTS ON OCCUi4TIONAL RADIATION EXPOSURES AT NUCLE-AR POWER PLANTStae Study.PWR Dunng Routine Operations.

Pathway Analysis NUREG-1101 V01: ONSITE DISPOSAL OF RADIOACTIVE Occupational SaMy NUREG/CR4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF WASTE.Gudance For Disposal By Subsurface Bunal.

U.S.AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE.

Pebble Bed Core Operating Experience NUREG/CR4434: ASSESSMENT OF MODELUNG NEEDS FOR WUREG/CR-4302 V01: AGING AND SERVICE WFAR OF CHECK SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF Performance Assessment NUCLEAR POWER PLANTS.

NUREG/CR4477; METHODOLOGIES FOR ASSESSING LONG-TERM Operating Reactors Licensing Action PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-NUREG-0748 V05 N11: OPERATING REACTORS UCENSING ACTIONS AGES.

SUMMARY

. Data As Of Noverr ber 30.1985 (Orange Book)

PM hmm%

Organic lodide NUREG/CR-4464: PERFORMANCE DEMONSTRATION TESTS FOR WUREG/CR-4327: ORGANIC IODIDE FORMATION FOLLOWING NU-DETECTION OF INTERGRANULAR STRESS CORROSION CRACK-CLEAR REACTOR ACCIDENTS.

ING.

Organic Liquid Personal Computer NUREG/CR4433: DOCUMENT REVIEW REGARDING HAZARDOUS NUREG/CR-4468: ADAPTATION OF OCA-P,A PROBABIUSTIC FRAC-CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

TURE-MECHANICS CODE,TO A PERSONAL COMPUTER.

Organization Chart Petitions For Rulemaking WUREG-0325 R08. U.S. NUCLEAR REGULATORY COMMISSION FUNC-NUREG-0936 V04 N04: NRC REGULATORY AGENDA.Quarterty Report.

TIONAL ORGANIZATION CHARTS.

October-December 1985.

Organization

pip, WUREGICR-4378. OBJECTIVE INDICATORS OF ORGANIZATIONAL NUREG-09/5 V04. COMPILATION OF CONTRACT RESEARCH FOR PERFORMANCE AT NUCLEAR POWER PLANTS-THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER-Overcooli ING TECHNOLOGY. Annual Report For FY 1985.

NUREG-195: LOSS OF INTEGRATED CONTROL SYSTEM POWER NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM - PHASE AND OVERCOOUNG TRANS!ENT AT RANCHO SECO ON DECEM-NI G

7 AGI A

E E EAR OF HYDRAULIC WU EG 83 V01: PRESSURIZED THERMAL SHOCK EVALUATION AND MECHANICAL SNUBBERS USED ON SAFETY-RELATED PIPING OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

AND COMPONENTS OF NUCLEAR POWER PLANTS Phase i Study.

WUREG/CR-4183 V02. PRESSURIZED THERMAL SHOCK EVALUATON NUREG/CR-4472: SIAMESE IMAGING TECHNIQUE FOR QUASI-VERTI-OF THE H B~ ROBINSON UNIT 2 NUCLEAR POWER PLANT

• CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

NUREG/CR-4545: PIPE CRACK EVALUATION IN OPERATING BOILING Overland Flow WATER REACTORS.

NUREG/CR4323: THE PROTECTON OF URANIUM TAluNGS IM-POUNDMENTS AGAINST OVERLAND EROS!ON' UR CR 94: RADIOLOGICAL ASSESSMENT OF BWR RECIRCU-

@xide Growth Rate LATORY PIPE REPLACEMENT.

NUREG/CR4476: HIGH-TEMPERATURE OXIDATION OF ZlRCALOY-4 IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

N N

/CR4528: DEVELOPMENT AND VERIFICATION OF CONDI-PISC ll TIONS FOR DUOTILE TEARING INSTABluTY AND ARREST.

NLREG/CR4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-NATION RELIABILITY AND FRACTURE MECHANICS Semiannual Plutonium Report.Apnl - September 1984.

NUREG/CR-4419: BIOASSAY MEASUREMENTS FOR URAN 10M USING SPUTTER INITIATED RESONANCE lON!ZATION SPECTROSCOPY.

PISC lli NUREG/CR-4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI-Policy And Planning Guidance NATION RELIABluTY AND FRACTURE MECHANICS. Semiannual NUREG-0885105: U.S. NUCLEAR REGULATORY COMMISSION 1986 Report Apnl - September 1984.

POUCY AND PLANNING GUIDANCE-PRA Postulated Accident NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR-4337:

TRAC-PF1/ MOD 1 INDEPENDENT OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

ASSESSMENT.DARTMOUTH COLLEGE AIR-WATER COUNTER-CUR-NUREG/CR-4372: PROBABAUSTIC RISK ASSESSMENT (PRA) APPU-RENT FLOW TESTS.

CATIONS.

(

Power Plant Life Extension PRA Guidance NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR ENVI-NUREG/CR-4565: PROBABluSTIC SAFETY STUDY APPLICATIONS RONMENT -lMPLICATIONS FOR NRC QUAUTY ASSURANCE PRO-PROGRAM FOR INSPECTION OF INDIAN POINT UNIT 3 NUCLEAR GRAMS BASED ON NUCLEAR POWER INDUSTRY ANJ REGULA-POWER PLANT.

TORY PROJECTONS THROUGH 1995.

PTS Practice And Procedures Digest NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATON OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

Prosaure C _.._ le NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR-4390: DCC 1/DCC-2 DEGRADED CORE COOLABluTY OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

ANALYSIS.

Sebject Index 47 Pressure Noise NUREG/CR-4446:

THE NUCLEAR INDUSTRY AND ITS NUREG/CR-4389: PRESSURE NOISE IN PRESSURIZED WATER RE-REGULATORS;A NEW COMPACT IS NEEDED.

ACTORS.

Quarterly Progrees Report Pressure Transmitter NUREG-0940 V04 N04: ENFORCEMENT ACTONS.SIGNIFICANT AC.

NUREG/CR-4324: TESTING OF NUCLEAR QUAUFIED CABLES AND TONS RESOLVED.Quarterty Progress Report,0ctober-December PRESSURE TRANSMITTERS IN SIMULATED HYDROGEN DEFLA-1985.

GRATONS TO DETERMINE SURVIVAL MARGINS AND SENSITIVI-NUREG/CR-2331 V05 N2: SAFETY RESEARCH PROGRAMS SPON-TIES.

SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Quarterty Progress Report.Apnl 14une 30,1985.

Pressure Vessel NUREG/CR-4402 V02: HIGH-TEMPERATURE GAS-COOLED REACTOR NUREG-0975 V04: COMPILATION OF CONTRACT RESEARCH FOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEEH-EVALUATON. Quarterly Progress Report, Apnl 1 - June 30,1985.

ING TECHNOLOGY. Annual R For FY 1985.

NUREG/CR-4453 V01: UGHT-WATER-REACTOR SAFETY FUEL SYS-NUREG/CR-4219 V02-HEAVY ECTON STEEL TECHNOLOGY PRO-l TEMS RESEARCH PROGRAMS. Quarterty Progress ReportJanuary-GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER March 1985.

1985.

NUREG/CR 4307 V01: LWR PRESSURE VESSEL SURVEILLANCE DO-QuartWy Repwt SIMETRY IMPROVEMENT PROGRAM. Progress Report - October NUREG4936 V04 N04 NRC REGULATORY AGENDA. Quarterly Report, NUREG'/C N6 VI II - A COMPUTER CODE FOR PREDICTING NU E R 37 0 LdNG-TERM PERFORMANCE OF MATERIALS THE PROBABluTY OF REACTOR PRESSURE VESSEL FAILURE-USED FOR HIGH-LEVEL WASTE PACKAGING. Third Quarterty Pressurized Thermal Shock Report, Year Four October -December 1985.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

Quench Behavior NUREG/CR-4183 V02-PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR-4390: DCC-1/DCC-2 DEGRADED CORE COOLABluTY OF THE H.B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

ANALYSIS.

NUREG/CR-4452: REVIEW OF RELAP5 CALCULATIONS FOR H.B.

ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

RAMONA-35 NUREG/CR-4468: ADAPTATION OF OCA-P.A FROBABILISTIC FRAC-NUREG/CR-2331 VOS N2-SAFETY RESEARCH PROGRAMS SPON-TURE-MECHANICS CODE.TO A PERSONAL COMPUTER.

SORED BY OcFICE OF NUCLEAR REGULATORY RESEARCH.Ouarterty Progress Report.Apnl 14une 30,1985.

p NUREG/CR-4462 A RANKING OF SABOTAGE / TAMPERING AVOID-RELAPS ANCE TECHNOLOGY ALTERNATIVES.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

Prismatic Core NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON NUREG/CR-4434: ASSESSMENT OF MODELLING NEEDS FOR OF THE H B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

NUREG/CR-4438: RESULTS OF SEMISCALE MOD-2C SMALL BREAK Probabilistic Fracture Mechanics (5%) LOSS-OF-COOLANT ACCOENT EXPERIMENTS S-LH 1 AND S.

NUREG/CR-4468: ADAPTATION OF OCA-P,A PROBABluSTIC FRAC-LH-2.

TURE-MECHANICS CODE.TO A PERSONAL COMPUTER' NUREG/CR-4452: REVIEW OF RELAPS CALCULATIONS FOR H.B.

ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

Probabilistic Model NUREG/CR-4293. REUABILITY ANALYSIS OF SHEAR WALL STRUC-RELAP5/ MOD 1 TURES.

NUREG/CR-4359: INDEFENDENT ASSESSMENT OF TRAC-PF1 (VER.

SION 7.0).RELAPS/ VOD1(CYCLE 14).AND TRAC-BD1 (VERSION Probabilistic Reliability Analysis 12.0) CODES USING SEPARATE-EFFECTS EXPERIMENTS.

NUREG/CR-4509: WASTE PACKAGE RELIABlUTY.

REMIX Probabilistic Risk Assessment NUREGsCR-4183 V01: PRESSURIZED THERM AL SHOCK EVALUATION NUREG/CR-4372: PROBABAUSTIC RISK ASSESSMENT (PRA) APPU-OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

CATIONS.

NUREG/CR-4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR-4462: A RANKING OF SABOTAGE / TAMPERING AVOID-OF THE H.B. ROB!NSON UNIT 2 NUCLEAR POWER PLANT.

ANCE TECHNOLOGY ALTERNATIVES.

NUREG/CR-4514: CONTROLLING PRINCIPLES FOR PRIOR PROBA-RIS BluTY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

NUREG/CR-4419: BIOASSAY MEASUREMENTS FOR URANIUM USING Probabilistic Safety Study SPUTTER INITIATED RESONANCE lONIZATION SPECTROSCOPY.

NUREG/CR-4565: PROBABILISTIC SAFETY STUDY APPLICATIONS Radiation PROGRAM FOR INSPECTON OF INDIAN POINT UNIT 3 NUCLEAR NUREG-1159: TRAINING MANUAL FOR URANIUM MILL WORKERS ON POWER PLANT.

HEALTH PROTECTION FROM URANIUM.

Probability Theory NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF NUREG/CR-4328; PROBABluTY BASED LOAD COMBINATION CRITE.

U.S.AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE.

RIA FOR DESIGN OF SHEAR WALL STRUCTURES.

Radiation Exposure Cost Productivity NUREG/CP-0068; PROCEEDINGS OF THE INTERNATIONAL NUCLEAR NUREG/CA-4546: LABOR PRODUCTIVITY A[MUSTMENT FACTORS.A REACTOR DECOMMISSIONING PLANNING CONFERENCE.

Method For Estimating Labor Construction Costs Associated With Physical Modfications To Nuclear Power Plants.

a u E / R 4485: THE IMPACT OF FUEL CLADDING FAILURE Pump EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE-4 NUREG/CR-4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL.

AR POWER PLANTS. Case Study PWR Dunng Routine Operations.

lARY FEEDWATER SYSTEM REUABluTY ANALYSIS.

Radiation Protection Quality Assurance NUREG/CR-4364: MANAGEMENT PERCEPTON OF THE HEALTH NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR ENVI-PHYSICS TECHNICIAN JOB.

RONMENT 4MPUCATIONS FOR NRC QUAUTY ASSURANCE PRO.

NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-U.S.AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE.

TORY PROJECTONS THROUGH 1995.

NUREG/CR-4369: QUALITY ASSURANCE (CA) PLAN FOR COMPUTER Radiation Safety SOFTWARE SUPPORTING THE U S. NUCLEAR REGULATORY COM.

NUREG/CP-0073: PROCEEDINGS OF THE WORKSHOP ON LARGE IR.

MISSION'S HIGH-LEVEL WASTE PROGRAM.

RADIATOR RADIATION SAFETY.

48 Subject Index Radiation Safety Training NUREG/CP-0072 VOS: PROCEEDINGS OF THE THIRTEENTH WATER NUREG 1159; TRAINING MANUAL FOR URANIUM MILL WORKERS ON REACTOR SAFETY RESEARCH INFORMATION MEETING.

HEALTH PROTECTION FROM URANIUM.

NUREG/CP0072 V06: PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCil INFORMATION MEETING.

Radioactive Decay Heating NUREG/CR-4390: DCC-1/DCC-2 DEGRADED CORE COOLABILITY Recirculating Coolant Pipe ANALYSIS.

NUREG/CR-4494: RADIOLOGICAL ASSESSMENT OF BWR RECIRCU-Radioactive Uquid Effluent Release NUREG/CR4286: EVALUATION OF RADIOACTIVE UQUID EFFLUENT Reclafiation RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT-NUREG/CR4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED Radinch Maw WASTES IN COMMERCIAL LOW-LEVE' '*IASTES. Draft Report For Comment NUREG/CR-2907 V03. RADIOACTIV MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1982.

Reform Amendment Radioactive Mixed Waste NUREG/CR-4059. EVALUATION OF THE IMPACT OF THE MC&A NUREG/CR-4450 DAF FC: MANAGEMENT OF RADIOACTNE MIXED REFORM AMENDMENTS ON A REPROCESSING FACIUTY.

W IN COMMERCIAL LOW-LEVEL WASTES.Oraft Report For Regulatory Activity NUREG/CP-0073 PROCEEDINGS OF THE WORKSHOP ON I ARGE IR-Radioactive Waste RADIATOR RADIATION SAFETY.

NUREG-1101 V01: ONSITE DISPOSAL OF RADIOACTIVE PA'C FL R ONTAMINATIONS U G 1109 FC: REGULATORY ANALYSIS FOR THE RESOLU-NU E /CR V03 H ON SOUDIFICATION. WASTE O!SPOSAL.AND ASSOCIATED OCCU-TlON OF UNRESOLVED SAFETY ISSUE A44. STATION BLACKOUT.

PATIONAL EXPOSURE NUREG/CR-4433; DOCUMENT REVIEW REGARDING HAZARDOUS Regulatory And Technical Report CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

NUREG-0304 V10 N04: REGULATORY AND TECHNICAL NUREG/CR-4555 GENERIC COST ESTIMATES FOR THE DISPOSAL REPORTS. Annual Compdabon for 1985.

OF HADIOACTIVE WASTES.

Regulatory Approach Radeologicae Assessment NUREG-0885 105: U.S. NUCLEAR REGULATORY COMMISSION 1986 NUREG/CR-4494: RADIOLOGICAL ASSESSMENT OF BWR RECIRCU.

POUCY AND PLANNING GUIDANCE.

LATORY PIPE REPLACEMENT.

Reinforced Concrete Radiological Data NUREG/CR4474: SCALE MODEUNG OF REINFORCED CONCRETE NUREG/CR-3365 DAF FC: REPORT TO THE NRC ON GUIDANCE FOR CATEGORY I STRUCT!!RES SUBJECTED TO SEISMIC LOADING.

PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX-ERCISES AT NUCLEAR GENERATING STATIONS Draft Repo't For Release Comment.

NUREG-1179 V01: RL'PTURE OF MODEL 48Y UF6 CYUNDER AND RE.

LEASE OF URANIUM HEXAFLUORIDE Sequoyah Fuels Radionuclide Facihty, Gore. Oklahoma. January 4,1986.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION NUREG-1189 V01: ASSESSMENT OF THE PUBUC HEALTH IMPACT TESTS. Semiannual Report Covenng The Penod Apnt-September 1985 FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEOUOYAH NUREG/CR4286. EVALUATION OF RADIOACTIVE LIOUlO EFFLUENT FUELS CORPORATION FACluTY AT GORE, OKLAHOMA Reprinted RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT..

NUREG 1189 Vr)2: ASSESSMENT OF THE PUBUC HEALTH IMPACT March 26,1986.

Rainfall FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH NUREG/CR4496: A SYSTEM FOR GENERATING LONG STREAM-FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN FUELS CORPORATON FACIUTY AT GORE. OKLAHOMA.

PERIOD-NUREG/CR-4286: EVALUATION OF RADIOACTIVE UQUID EFFLUENT RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

Reactor NUREG/CR-3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN-Rollety NUREG/CR-4477: METHODOLOGIES FOR ASSESSING LONG-TERM GINEERING GUIDELINES FOR NUCLEAR POWER PLANT MAIN-TAINABlUTY.

PERFORMANCE OF HIGH-LEVEL RADIOACTNE WASTE PACK.

AGES.

Reactor Operator NUREG/CR-4411: ASSESSMENT OF SPECIAUZED EDUCATIONAL Reliability Anaiysis PROGRAMS FOR UCENSED NUCLEAR REACTOR OPERATORS.

NUREG/CR-4293: REUABlWTY ANALYSIS OF SHEAR WALL STRUC.

TURES.

Reactor Pressure Baundary NUREG/CR-4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUXil-NUREG/CR-4300 V02: ACOUSTIC EMtSSION/ FLAW RELATIONSHIP IARY FEEDWATER SYSTEM REUABILITY ANALYSIS.

FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR-4328: PROBABluTY BASED LOAD COMBINATION CRITE-VESSELS Progress Report, Apnt-September 1985~

RIA FOR DESIGN OF SHEAR WALL STRUCTURES.

NUREG/CR-4366: RELIABluTY ASSESSMENT OF CONTAINMENT Reactor Safety TANGENTIAL SHEAR FAILURE.

NUREGICR-2331 V05 N2: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY Remote HandHng Equipment RESEARCH.Ouarterty Progress Report.Apnl 1 June 30,1985.

NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF 7

NUREG/CR4485. THE IMPACT OF FUEL CLATING FAILURE U.S.AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE.

EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE-AR POWER PLANTS Case Study PWR Dunng Routme Operations Report 6ng System NUREG/CR-4547: CONTEMPT 4/ MOD 6:A MULTICOMPONENT SYSTEM NUREG/CR-4132; NUCLEAR POWER SAFETY PEPORTING SYSTEM ANALYSIS PROGRAM.

FINAL EVALUATION RESULTS.

Reactor Safety Research hprocessing Plant NUREG/CP4072 V01 PROCEEDINGS OF THE THIRTEENTH WATER NUREG/CR-4059: EVALUATION OF THE IMPACT OF THE MC&A REACTOR SAFETY RESEARCH INFORMATION MEETING REFORM AMENDMENTS ON A REPROCESSING FACluTY.

NUREG/CP4072 V02-PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

Residual Radlonuclide inventory NUREG/CP-0072 V03. PROCEEDINGS OF THE THIRTEENTH WATER NUREG/CR4289: REhlDUAL RADIONUCUDE CONTAMINATION REACTOR SAFETY RESEARCH INFORMATION MEETING.

WITHIN AND ARC UND COMMERCIAL NUCLEAR POWER NUREG/CP-0072 V04 PROCEEDINGS OF THE THIRTEENTH WATER PLANTS. ORIGIN.DISTFIBUTION. INVENTORY AND DECOMMISSION-REACTOR SAFETY RESEARCH INFORMATION MEETING.

ING ASSESSMENT.

,.,..- i..

t, m

F.i.

Subject index 49 Resistance Factor Safety Evaluation Report NUREG/CR4328: PROBABILITY BASED LOAD COMBINATION CRITE-NUREG4853 SOS: SAFETY EVALUATION REPORT RELATED TO THE RfA FOR DESIGN OF SHEAR WALL STRUCTURES.

OPERATION OF CLINTON POWER STATION. UNIT NO.1. Docket No.

Resonance Ionization Spectroscopf 50-461.(lllinois Power Company)

NUREG4887 S08: SAFETY EVALUATION REPORT RELATED TO THE NuPEG/CR4419: BIOASSAY MEASUREMENTS FOR URANIUM USING OPERATION OF PERRY NUCLEAR POWER PLANTUNITS 1 AND SPUYTFR INITIATED RESONANCE lONIZATION SPECTROSCOPY-

2. Docket Nos. 50-440 And 50-441.(Cleveland Electne liluminating Com-q Reynolds Numt*

pany)

NUREG/CR-4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER NUREG4887 SO9: SAFETY EVALUATION REPORT RELATED TO THE NATURAL CIRCULATION CONDITIONS.

OPERATION OF PERRY NUCLEAR POWER PLANT, UNITS 1 AND

2. Docket Nos. 50-440 And 50-441.(Cleveland Electnc illuminating Com-Risk pany)

NUREG/CR-4183 V01: oRESSURIZED THFRMAL SHOCK EVALUATION NUREG-1158: SAFETY E\\ ALUATION REPORT RELATED TO THE RE-OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NEWAL OF THE OPERATING LICENSE FOR THE RESEARCH REAC-NUREG/CR-4183 V02: PRESSURIZED THERVAL SHOCK EVALUATION TOR AT PENNSYLVANIA STATE UN'VERSITY. Docket No. 50-OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

005 (Pennsylvansa State University)

Risk Analysis Safety Fsature System NUREG/CR-4514. CONTROLLING PRINCIPLES FOR PRIOR PROBA.

NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK BILITY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF NUCLEAR POWER PLANTS.

Rockriprap NUREG/CR4323. THE PROTECTION OF URANIUM TAILINGS IM-Safety Fuel Systems Research Program POUNDMENTS AGAINST OVERLAND EROSION.

NUREG/CR4453 V01: LIGHT WATER-REACTOR SAFETY FUEL SYS-Rod Bundle TEMS RESEARCH PROGRAMS. Quarterly Progress Report. January-March 1985.

NUREG/CR-4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.

Safety Ind'cator Rules NUREG/CR-4378: OBJECTIVE INDICATORS OF ORGANIZATIONAL PERFORMANCE AT NUCLEAR POWER PLANTS.

NUREG-0936 V04 N04: NRC REGULATORY AGENDA.Ouarterly Report.

October-Decernber 1985.

Safety Margire Rules Of Practice NUREG/CR-4431:

SUMMARY

REPORT ON THE SEISMIC SAFETY MARGINS RESEARCH PROGRAM.

Rupture L

E OF U AN UM HE AFLUOR DE a

F U EG/ R 13 : N EAR POWER SAFETY REPORTING SYSTEM Facey,G"e, Oklahoma. January 4,1986.

F!NAL EVALUATION RESULTS.

SAFT UT Safety Research NUREG/CR4078: PROGRAM FOR FIELD VALIDATION OF THE SYN.

NUREG/CP-0072 V01: PROCEEDINGS OF THE THIRTEENTH WATER THETIC APERTURE FOCUSING TECHNIQUE FOR ULTRASONIC REACTOR SAFETY RESEARCH INFORMATION MEETING.

TESTING (SAFT UT) Final Report.

NUREG/CP-0072 V02: PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

SIRIS NUREG/CP-0072 V03: PROCEEDINGS OF THE THIRTEENTH WATER HUREGICR4419: BIOASSAY MEASUREMENTS FOR URANIUM USING REACTOR SAFETY RESEARCH INFORM ATION MEETING.

SPUTTER INITIATED RESONANCE lONIZATION SPECTROSCOPY.

NUREGICP-0072 V04: PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

SQRT NUREG/CP-0072 V05. PROCEEDINGS OF THE THIRTEENTH WATER NUREG/CR 3137: SEISMIC AND DYNAMIC OUALIFICATION OF RE-REACTOR SAFETY RESEARCH INFORMATION MEETING.

LATED ELECTRICAL AND MECHANICAL EQUIPMENT.

NUREG/CP4072 V06: PROCEEDINGS OF THE THIRTEENTH WATER p

REACTOR SAFETY RESEARCH INFORMATION MEETING.

NUREG/CR4431:

SUMMARY

REPORT ON THE SEISM!C SAFETY Safety Research Program MARGINS RESEARCH PROGRAM.

a NUREG/CR-2331 V05 N2: SAFETY RESEARCH PROGRAMS SDON-Sabotage SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4462: A RANK'NG OF SABOTAGE / TAMPERING AVOID-RESEARCH Ouarterly Progress ReporLApnl 1 June 30,1985.

NUREG/CR-4459-LIGHT WATER REACTOR SAFETY RESEARCH ANCE TECHNOLOGY ALTERNATIVES.

PROGRAM Semiannual Report,0ctober 1983 - March 1984.

Safeguards Safety Study NUREG-0525 R11: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL).

NUREG/CR4462. A RANKING OF SABOTAGE / TAMPERING AVOID-NUREG/CR-4402 V02: HIGH TEMPERATURE GAS-COOLED REACTOR ANCE TECHNOLOGY ALTERNATIVES-SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION Ouartetty Progress Report, Apnl 1 - June 30,1985.

Safety NUREG/CR4293: RELIABILITY ANALYSIS OF SHEAR WALL STRUC-NU EG CR-4474: SCALE MODELING OF REINFORCED CONCRETE CATEGORY I STRUCTURES SUBJECTED TO SEISMIC LOADING.

Safety Analysis NUREG/CR4434 ASSESSMENT OF MODELLING NEEDS FOR Scale Nel

=

SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

NUREG/CR-4474 SCALE MODELING OF REINFORCED CONCRETE CATEGORY I STRUCTURES SUBJECTED TO SEISMIC LOADING.

Safety Evaluation NUREG-0954 SOS: SAFETY EVALUATION REPORT RELATED TO THE Securfty OPERATION OF CATAWBA NUCLEAR STA TION. UNITS 1 AND NUREG/CR-4473:A STUDY OF THE OPERATION AND MAINTENANCE 2 Docker Nos. 50-413 And 50-414 (Duke Power Company)

OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 NUREG 1031 SOS: SAFETY EVALUATION REPORT RELATED TO THE CFR 73 55.

OPERATION OF MILLSTONE NUCLEAR POWER STATION. UNIT No 3 Docket No. 50-423. (Northeast Nuclear Energy Company)

Seismic Loan 2ing NUREG-1168 STAFF EVALUATION OF U S.

DEPARTMENT OF NUREG/CR 4474: SCALE MODELING OF REINFORCED CONCRETE ENERGY PROPOSAL FOR MONITORED RETTilEVABLE STORAGE.

CATEGORY I STRUCTURES SUBJECTED TO SEISMIC LOADING.

1 1

50 Subject index Seismic Margin NUREG/CR-4188 V02: NUCLEAR POWER PLANT SIMUL ATON FACILl-NUREG/CR4482 DRF FC: RECOMMENDATIONS TO THE NUCLEAR TY EVALUATON METHODOLOGY. Technical Bases.

REGULATORY COMMISSION ON TRIAL GUIDELINES FOR SEISMIC MARGIN REVIEWS OF NUCLEAR POWER PLANTS. Draft Report For Small-8reak LOCA Comment.

NUREG/CR-4438: RESULTS OF SEMISCALE MOD-2C SMALL BREAK (5%) LOSS-OF-COOLANT ACCIDENT EXPERIMENTS S-LH-1 AND S-Setemic Qualification Team M2.

NUREG/CR-3137: SEldMIC AND DYNAMIC OUAllFICATON OF RE-LATED ELECTRICAL AND MECHANICAL EQUIPMENT.

Snm Seismic R6sk Analye6e NUREG/CR-4279 V01: AGING AND SERVICE WEAR OF HYDRAUUC NUREG/CR-4431:

SUMMARY

REPORT ON THE SEISMIC SAFETY AND MECHANICAL SNUBBERS USED ON SAFETY-RELATED PIPING MARGINS RESEARCH PROGRAM.

AND COMPONENTS OF NUCLEAR POWER PLANTS. Phase i Study.

Seismic Strese Analysis Software Quality Assurance NUREG/CR 3790- CLOSEOUT OF IE BULLETIN 79-07; SEISMIC NUREG/CR-4369: QUALITY ASSURANCE (OA) PLAN FOR COMPUTER OTRESS ANALYSIS OF SAFETY-RELATED PIPING.

SOFTWARE SUPPORTING THE U.S. NUCLEAR REGULATORY COM-MISSION'S HIGH-LEVEL WASTE PROGRAM.

[

NUREG/CR4502-VIRGINIA REGONAL SEISMIC NETWORK. FINAL Whh REPORT (19771985).

NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATONS Semiannual Progress Report ON SOUDIFICATON. WASTE DISPOSAL AND ASSOCIATED OCCU-NUREG/CR4219 V02: HEAVY-SECTlCN STEEL TECHMLOGY PRO.

PATONAL EXPOSURE.

GRAM SEMIANNUAL PROGRESS REPORT FOR ARRil-SEPTEMBER 1985 Solubility NUREG/CR4255 V02: AEROSOL RELEASE AND TRANSPORT PRO.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATON GRAM SEM1 ANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER TESTS. Semiannual Report Covenng The Period Apni-September 1985.

1985.

Source Term NUREG/CR4327: ORGANIC ODIDE FORMATON FOLLOWING NU-U G

-408 V03: DEGRADED PIPING PROGRAM - PHASE CLEAR REACTCR ACCIDENTS.

II. Semiannual Report.Apnl 1985 - Scotember 1985.

NUREG/CR4459: UGHT WATER REACTOR SAFETY RESEARCH PROGRAM. Semiannual Report. October 1983 - March 1984 Spatial Verlation NUREG/CR-4469 V01: INTEGRATION OF NONDESTRUCTIVE EXAMI.

NUREG/CR-3805 V03: ENGINEERING CHARACTERIZATON OF NATION REUABluTY AND FRACTURE MECHANICS Semiannual GROUND MOTON. Task 11: Observational Data On Spatial vanations Report,Apnl - Sep' ember 1984.

Of Earthquake Ground Motion.

Semiscale Spent Fuel Shipping Casks NUREG/CR4393:

SUMMARY

OF SEMISCALE SMALL BREAK LOSS-NUREG/CR-3760- A STUDY ON DUCTILE AND BRITTLE FAILURE DESIGN CRITERIA FOR DUCTILE CAST IRON SPENT-FUEL SHIP-NU E CP 8 E S F L

2 ALL BREAK PING CONTAINERS.

(5%) LOSS-OF-COOLANT ACCIDENT EXPERIMENTS S-LH-1 AND S-LH-2.

NUREG/CR-4465:

TRAC PF1/ MODI INDEPENDENT Spent-Fuel ASSESSMENT:SEMISCALE MOD-2A INTERMEDIATE BREAK TEST NUREG/CR4379 V02-LONG-TERM PERFORMANCE OF MATERIALS S.lB-3.

USED FOR HIGH-LEVEL WASTE PACKAGING Second Quarterly NUREG/CR-4466; STATON BLACKOUT TRANSIENTS IN THE SEMIS-Report, Year Four July-September 1985.

CALE FACluTY.

NUREG/CR4379 V03: LONG-TERM PERFORMANCE Oi MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.Thrd Quarterty

• E " "'

U G/

3: INTEGRATED SEVERE ACCIDENT CONTAINMENT ANALYSIS WITH THE CONTAIN COMPUTER CODE.

Spent-Fuel Shipping Container Service Wear NUREG/CR-4363: A STUDY ON FABRICATION CFitTERIA FOR DUC-NUREG/CR-4302 V01: AGING AND SERVICE WEAR OF CHECK TILE CAST IRON SPENT-FUEL SHIPPING CONTAINERS.

VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF NUCLEAR POWER PLANTS.

State Regulation NUREG/CR-4380 EVALUATON OF THE MOTOR OPERATED VALVE NUREG/CR-4446:

THE NUCLEAR INDUSTRY AND ITS ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT REGULATORS A NEW COMPACT IS NEEDED.

DEGRADATON. INCORRECT ADJUSTMENTS.AND OTHER ABNOR-MALITIES IN MOTOR-OPERATED VALVES.

Station Blackout NUREG-1109 DRFT FC: REGULATORY ANALYSIS FOR THE RESOLU-E B

OFF E OF N L AR REGULA R C

C RESEARCH.OuartertrProgress Report. April 1 June 30,1985.

NUREG/CR4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT ANALYSIS WITH THE CONTAIN COMPUTER CODE.

EG/CR-4476: HIGH-TEMPERATURE OXIDATION OF ZlRCALOY-4 Shear Wall Structure IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

NUREG/CR-4293: RELIABIUTY ANALYSIS OF SHEAR WALL STRUC-s TURES.

Steam Explosion NUREG/CR-4328: PROBABluTY BASED LOAD COMBINATON CRITE-NUREG/CR-3983. STEAM EXPLOSlON EXPERIMENTS AT INTERMEDI-RIA FOR DESIGN OF SHEAR WALL STRUCTURES.

ATE SCALE.FITSB SERIES.

i Signal Detecuon Theory Steam Generator NUREG/CR4436 V01: HUMAN REUABluTY IMPACT ON INSERVICE NUREG-0975 V04: COMPitATION OF CONTRACT RESEARCH FOR N

G/

-4 36 U AN RE I

i ACT ON INSERVICE THE MATERIALS ENGINEERING BRANCH, DIVISION OF ENGINEER-INSPECTON. Volume 2: Review And Analysis Of Human Performance ING TECHNOLOGY. Annual Report For FY 1985.

NUREG/CR 2331 V05 N2: SAFETY RESEARCH PROGRAMS SPON-In Nondestructrve Testing (Emphasizing Ultrasorucs).

SORED BY OFFICE OF NUCLEAR REGULATORY Simulator RESEARCH.Ouarterty Progress Report. April 1 June 30,1985.

NUREG/CR4188 V01: NUCLEAR POWER PLANT SIMULATON FAClu-NUREG/CR-4276 V02: VIBRATON AND WEAR IN STEAM GENERA-TY EVALUATON METHODOLOGY. Handbook-TOR TUBES FOLLOWING CHEMICAL CLEANING.

t

Subject Index 51 Stroemflow Record TURC1 NUREG/CR-4496: A SYSTEM FOR GENERATING LONG STREAM-NUREG/CR 4420- TURC1:LARGE SCALE METALUC MELT CONCEN-FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN TRATE INTERACTON EXPERIMENTS AND ANALYSIS.

PERIOD.

Tallinge Stroes NUREG/CR4504: LONG TERM SURVEILLANCE AND MONITORING NUREGICR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION CF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL-TESTS.Semannual Report Covenng The Penod Apni-September 1985.

INGS PlLES.

Stroes Analysis Tampering NUREG/CR4475: ORMGEN PC A MOROCOMPUTER PROGRAM FOR NUREG/CR-4482: A RANKING OF SABOTAGE / TAMPERING AVOID-AUTOMATIC MESH GENERATON OF 2-D CRACK GEOMETRIES ANCE TECHNOLOGY ALTERNATIVES.

Stress Corroelon Cracking NUREGICR-2331 V05 N2-SAFETY RESEARCH PROGRAMS SPON-Tengendal Sheer NUREG/CR-4366. RELIABluTY ASSESSMENT OF CONTAINMENT SORED BY OFFICE OF NUCLEAR REGULATORY TANGENTIAL SHEAR FAILURE.

RESEARCH.Quarterfy Progress ReportApnl 14une 30,1985.

Subeutunon Toering InstatWitty NUREG/CR-4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED NUREG/CR-4528: DEVELOPMENT AND VERIFICATION OF CONDI-WASTES IN COMMERCIAL LOW. LEVEL WASTES. Draft Report For TIONS FOR DUCTILE TEARING INSTABluTY AND ARREST.

Comment Teoring Modulue NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM. PHASE 1

V01: ONSITE DISPOSAL OF RADIOACTIVE Semannual Report,Apnl 1985 - September 485.

WASTE. Guidance For Disposal By Subsurface Bunal.

Technical Speci6 cation Supw Sy m Code NUREG-1162: TECHNICAL SPECIFICATIONS FOR PERRY NUCLEAR NUREG/CR-2331 VOS N2: SAFETY RESEARCH PROGRAMS SPON-POWER PLANT, UNIT 1. Docket No. 50 440 (Cleveland Electne lilumi-SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Quarterty Progress Report.Apnl 1 June 30,1985.

NU E -1 NICAL SPECIFICATIONS FOR MILLSTONE NUCLE-AR POWER STATON. UNIT NO. 3. Docket No. 50-423.(Northeast Nu-Surveillance clear Energy Co)

NUREG/CR4302 V01: AGING AND SERVICE WEAR OF CHECK NUREG-1182: TECHNICAL SPECIFICATIONS FOR CATAWBA NUCLE-VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF AR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-414 (Duke NUCLEAR POWER PLANTS.

Power Company)

NUREG/CR4307 V01: LWR PRESSURE VESSEL SURVEILLANCE DO.

RY I VEMENT PROGRAM. Progress Report - October T

UREG-0975 V04: COMPILATON OF CONTRACT RESEARCH FOR NUREG/CR 380: EVALUATION OF THE MOTOR 4PERATED VALVE THE MATERIALS ENGINEERING BRANCH.C' VISION OF ENGINEER-ANALYSIS AND TEST SYSTEM (MOVATSI TO DETECT ING TECHNOLOGY. Annual Report For FY 1985.

DEGRADATION. INCORRECT ADJUSTMENTS.AND OTHER ABNOR.

NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION MAlfTIES IN MOTOR-OPERATED VALVES.

TESTS.Semannual Report Covenng The Penod Apni-September 1985.

NUREG/CR4504: LCNG-TERM SURVEILLANCE AND MONITORING NUREG/CR-4279 V01: AGING AND SERVICE WEAR OF HYDRAUUC OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL.

AND MECHANICAL SNUBBERS USED ON SAFETY-RELATED PIPING INGS PILES.

AND COMPONENTS OF NUCLEAR POWER PLANTS Phase i Study.

NUREG/CR4324: TESTING OF NUCLEAR QUAUFIED CABLES ANb Survey PRESSURE TRANSMITTERS IN SIMULATED HYDROGEN DEFLA-NUREG/CR-4433. DOCUMENT REVIEW REGARDING HAZARDOUS GRATIONS TO DETERMINE SURV! VAL MARGINS AND SENSITIVI-CHEMICAL CHA7ACTERISTICS OF LOW-LEVEL WASTE.

TIES.

NUREG/CR-4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 (VER.

Synthetic Aperture Focusing Techniqu' NUREG/CR-4078: PROGRAM FOR FIELD VALIDATION OF THE SYN-SlON 7.0).RELAP5/ MOD 1(CYCLE 14).AND TRAC-801 (VERSION 12.0) CODES USING SEPARATE-EFFECTS EXPERIMENTS.

THETIC APERTURE FOCUSING TECHNIQUE FOR ULTRASONIC NUREG/CR-4464: PERFORMANCE DEMONSTRATION TESTS FOR TESTING (SAFT UT)Fmal Report CETECTON OF INTERGRANULAR STRESS CORROSON CRACK.

Systeme interaction ING.

NUREG/CR-4469 V01: INTEGRATON OF NONDESTRUCTIVE EXAMI-TLD NATON REUABluTY AND FRACTURE MECHANICS.Sermannual NUREG4837 V05 NO3: NRC TLD DIRECT RADIATION MONITORING eporW - September M NETWORK. Progress Report, July-September 1985.

Thwmal Shock TRAC NUREG/CR4219 V02: HEAVY-SECTON STEEL TECHNOLOGY PRO-NUREG/CR4442: TRAC USER'S GUIDE.

GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER 1985.

TRAC-8D1 NUREG/CR4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 (VER-Themal-Hydrounc Response SiON 7.0),RELAP5/ MOD 1(CYCLE 14) AND TRAC-BD1 (VERSON NUREG/CR-4171:

TRAC PF1/ MODI INDEFENDENT 12.01 CODES USING SEPARATE EFFECTS EXPERIMENTS.

ASSESSMENT: LOBI LARGE BREAK TRANSIENT A1-04R NUREG/CR-4465:

TRAC-PF1/ MODI INDEPENDENT TRAC-PF1 ASSESSMENT:SEMISCALE MOO-2A INTERMEDIATE BREAK TEST NUREGICR-4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 (VER-S-IB-3.

SiON 7 0).RELAP5/ MOD 1(CYCLE 14).AND TRAC-801 (VERSION 12.0) CODES USING SEPARATE-EFFECTS EXPERIMENTS.

T y

e VOS N2: SAFETY RESEARCH PROGRAMS SPON-TRAC-PF1/ MOO 1 SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4337:

TRAC-PF1/ MOD 1 INDEPENDENT RESEARCH.Ouarterty Progress Report.Apnl 14une 30,1985.

ASSESSMENT.DARTMOUTH COLLEGE AIR-WATER COUNTER-CUR.

NUREG/CR4337:

TRAC-PF1/ MODI INDEPENDENT RENT FLOW TESTS.

ASSESSMENT:DARTMOUTH COLLEGE AIR-WATER COUNTER CUR-NUREG/CR4465.

TRAC-PF1/ MOD 1 INDEPENDENT RENT FLOW TESTS.

ASSESSMENT.SEMISCALE MOD-2A INTERMEDIATE BREAK TEST NUREG/CR-4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT S-IB-3.

ANALYSIS WITH THE CONTAIN COMPUTER CODE.

NUREG/CR-4348 V01: COMMIX-18 A THREE-DIMENSIONAL TRAN-TRAC-PF1/Modt SIENT SINGLE-FHASE COMPUTER PROGRAM FOR THERMAL HY.

NUREG/CR-4171:

TRAC-PF1/ MODI INDEPENDENT ORAUUC ANALYSIS OF SINGLE AND MULTICOMPCNENT ASSESSMENT.LOBf LARGE BREAK TRANSIENT A104R.

SYSTEMS.Vol1: Equations And Numencs.

l

i 52 Subject Index NUREG/CR-4348 VC2: COMMIX 18.A THREE-DIMENSIONAL TRAN-Ultrasonic Imaging SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-NUREG/CR-4472: SIAMESE IMAGING TECHNIQUE FOR QUASI-VERTI-DRAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

SYSTEMS.Vol 11 Us#s Manual.

NUREG/CR 4371: COMMIX-2:A THREE-DIMENSIONAL TRANSIENT Ultrasonic Test 6ng COMPUTER PROGRAM FOR THERMAL-HYDRAUUC ANALYSIS OF NUREG/CR-4078; PROGRAM FOR FIELD VALIDATION OF THE SYN-TWO PHASE FLOWS.

THETIC APERTURE FOCUSING TECHN!OUE FOR ULTRASONIC NUREG/CR4442 TRAC USER 1 GUIDE.

TESTING (SAFT UT)Enal Report.

Thermoluminescent Doelmeter Uncertainty NUREG4837 V05 NO3: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NETWORK. Progress Report, July-September 1985.

OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EVALUATION Thorium OF THE H B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

NUREG/CR4419: BIOASSAY MEASUREMENTS FOR URANIUM USING SPUTTER INITIATED RESONANCE ONIZATION SPECTROSCOPY.

Uncertainty Analyets NUREG/CR4460- UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN Title List UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL NUREG4540 V07 N11: TITLE UST OF DOCUMENTS MADE PUBLICLY MODEL AVAILABLE. November 1-30,1985.

NUREG/CR-4514: CONTROLUNG PRINCIPLES FOR PRIOR PROBA-NUREG-0540 V07 N12-TITLE UST OF DOCUMENTS MADE PUBLICLY BluTY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

AVAILABLE. December 1 31.1985.

NUPEG-0540 V08 N01: TITLE UST OF DOCUMENTS MADE PUBLICLY Unresolved Safety leeue A-44 AVAILABLE. January 1-31.1986.

NUREG 1109 DRFT FC-REGULATORY ANALYSIS FOR THE RESOLU-TION OF UNRESOLVED SAFETY ISSUE A-64. STATION BLACKOUT.

Training NUREG-1159. TRAINING MANUAL FOR URANIUM MILL WORKERS ON Uranium HEALTH PROTECTION FROM URANIUM.

NUREG-1189 V01: ASSESSMENT OF THE PUBUC HEALTH IMPACT NUREG/CR-4364: MANAGEMENT PERCEPTON OF THE HEALTH FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH PHYSICS TECHNICIAN JOB.

FUELS CORPORATION FACluTY AT GORE. OKLAHOMA.Repnnted NUREG/CR4411: ASSESSMENT OF SPECIALIZED EDUCATIONAL March 26,1986.

PROGRAMS FOR UCENSED NUCLEAR REACTOR OPERATORS.

NUREG-1189 V02: ASSESSMENT OF THE PUBUC HEALTH IMPACT FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH Transient FUELS CORPORATION FACILITY AT GORE. OKLAHOMA.

NUREG-1195: LOSS OF INTEGRATED CONTROL SYSTEM POWER NUREG/CR4419: BIOASSAY MEA:,UREMENTS FOR URANIUM USING AND OVERCOOLING TRANSIENT AT RANCHO SECO ON DECEM-SPUTTER INITIATED RESONANCE ONIZATION SPECTROSCOPY.

BER 26.1985.

NUREG/CR 3958: EFFECTS OF CONTROL SYSTEM FAILURES ON Uranium Hesafluoride TRANSIENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A NUREG-1179 V01: RUPTURE OF MODEL 48Y UF6 CYUNDER AND RE-COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR.

LEASE OF URANIUM HEXAFLUORIDE.Sequoy.h Fuels NUREG/CR-4171:

TRAC-PF1/ MODI INDEPENDENT Facity, Gore.OWahoma. January 4,1 9456.

ASSESSMENTiOBI LARGE BREAK TRANSIENT A104R.

NUREG/CR-4348 V02: COMMIX-18.A THREE-DIMENSIONAL TRAN-Uranium Mill Worker SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-NUREG.1159: TRAINING MANUAL FOR URANIUM MILL WORKERS ON ORAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT HEALTH PROTECTION FROM URANIUM.

SYSTEMS.Volll: User's Manual.

NUREG/CR-4452 REVIEW OF RELAPS CALCULATONS FOR H.B.

Uranium E - - - - 4 ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

NUREG/CR-4504: LONG-TERM SURVEILLANCE AND MONITORING NUREG/CR-4453 V01: UGHT WATER-REACTOR SAFETY FUEL SYS-OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL.

TEMS RESEARCH PROGRAMS. Quarterty Progress Report, January-INGS PILES.

March 1985 NUREG/CR-4466: STATON BLACKOUT TRANSIENTS IN THE SEMIS-Uranium Tailings impoundment CALE FACluTY.

NUREG/CR-4323: THE PROTECTION OF URANlUM TAIUNGS IM-POUNDMENTS AGAINST QVERLAND EROSION.

NUREG/CR-4337:

TRAC-PF1/ MOD 1 INDEPENDENT User's Guide ASSESSMENT.DARTMOUTH COLLEGE AIR WATER COUNTERCUR.

NUREG/CR-4442 TRAC USER'S GUIDE-RENT FLOW TESTS.

User's Manual fransient Reactor Analysis Code NUREG/CR-4348 V02: COMMIX-1B A THREE-DIMENSONAL TRAN-NUREG/CR-4442: TRAC USER'S GUIDE.

SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY-DRAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT Transport SYSTEMS.Vol ll: User's Manual.

NUREGtCR-4402 V02 HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCOENT VISA-il EVALUATON Quarterly Progress Report, Apnl 1 - June 30,1985.

NUREG/CR-4486: VISA 11 - A COMPUTER CODE FOR PREDICTING THE PROBABluTY OF REACTOR PRESSURE VESSEL FAILURE.

NUREG/CR-2331 VOS N2: SAFETY RESEARCH PROGRAMS SPON-Value-impact t

SORED BY OFFICE OF NUCLEAR REGULATORY NUREG-1109 DRFT FC: REGULATORY ANALYSIS FOR THE RESOLU-RESEARCH Ouarterty Progress Report.Apnl 1-June 30.1985.

TON OF UNRESOLVED SAFETY ISSUE A-44.STATON BLACKOUT.

NUREG/CR-4276 V02: VIBRATION AND WEAR IN STEAM GENERA-NUREG/CR-4546: LABOR PRODUCTIVITY ADJUSTMENT FACTORS.A TOR TUBES FOLLOWING CHEMICAL CLEANING.

Method For Estanetmg Labor Construction Costs Associated Wdh Physical ModMcatens To Nuclear Power Plants.

Two-phase Flow NUREG/CR-4555: GENERIC COST ESTIMATES FOR THE DISPOSAL NUREG/CR-4371: COMMIX 2:A THREE-DIMENSIONAL TFIANSIENT OF RADIOACTIVE WASTES.

COMPUTER PROGRAM FOR THERMAL-HYDRAULIC ANALYSIS OF TWO-PHASE FLOWS.

Valve NUREG/CR-4501: MODEUNG OF VAPOR GENERATON IN FLASHING NUREG/CR4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL-FLOW.

lARY FEEDWATER SYSTEM RELIABluTY ANALYSIS.

NUREG/CR-4380 EVALUATON OF THE MOTOR OPERATED VALVE Ultimate Containment Capacity ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT NUREG/CR-4223; STEEL CONTAINMENT RESISTANCE UNDER GEN-DEGRADATION. INCORRECT ADJUSTMENTS,AND OTHER ABNOR-ERAL DYNAMIC PRESSURES.

MAUTIES IN MOTOH-OPERATED VALVES.

St.teci 'ndex 53 Vapor Generation NUHEGICR4585: INVEST)GATION OF THE STABluTY OF CLAY /

NUREGICR4501: MODELING OF VAPOR GENERATION IN FLASHING BASALT PACKING MATERIALS.

FLOW.

VeaselIntegrity Simulation Analys6s Weate Package Containment RolletWitty NUREG/CR-4486: VISA 11 - A COMPUTER CODE FOR PREDICTING NUREG/CR4509: WASTE PACKAGE RCU ABIL TY.

THE PROBABlWTY OF REACTOR PRESSURE VESSEL FAILURE.

Weate Storage NUREG-1168: STAFF EVALUATION OF U.S.

DEPARTMENT OF UR /CR-4276 V02-VIBRATION AND WEAR IN STEAM GENF.RA-ENERGY PROPOSAL FOR MONITORED RETRIEVABLE STORAGE.

TOR TUBES FOLLOWING CHEMICAL CLEANING.

Water Hammer VW Regional SM mork NUREG-1190 LOSS OF POWER AND WATER HAMMER EVENT AT NUREG/CR-4502: VIRGINIA REGIONAL SEISMIC NE1 WORK FINAL SAN ONOFRE UNIT 1 ON NOVEMBER 21,1985.

REPORT (1977-1985).

Wall Nucteet6on NUREG/CR4276 VC2: VIBRATION AND WEAR IN STEAM GENERA.

NUREG/CR4501: MODELING OF VAPOR GENERATION IN FLASHING TOR TUBES FOLLOWING CHEMICAL CLEANING.

FLOW.

Weed Overlay Weste 06apoest NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS NUREG/CR-4469 VOI: INTEGRATION-OF NONDESTRUCTIVE EXAMI-NATION RELIABluTY AND FRACTURE-MECHANICS Semannual ON SOUDIFICATION. WASTE DISPO-SAL,AND ASSOCIATED OCCU-PATIONAL EXPOSURE.

Report.Apnt - September 1984.

NUREG/CR-4477: METHODOLOGIES FOR ASSESSING LONG TERM Weidment PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK

• AGES.

NUREG/CR4219 V02-HEAVY-SECTION STEEL TECHNOLOGY PRO-GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL-SEPTEMBER Weste Form /Overpack 1985.

NUREG/CR-4379 V03: LONG-TERM PERFOAMANCE OF MATERIALS Wet-Air Oxidation USED FOR HIGH-LEVEL WASTE PACKAGING.Thrd Quarterty Report, Year Four October -December 1985.

NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS ON SOLIDIFICATION. WASTE DISPOSAL,AND ASSOCIATED OCCU-Waste Package PATIONAL EXPOSURE.

NUREG/CR-3091 V07; REVIEW OF WASTE PACKAGE VERIFICATION TESTS Semannual Report Covenng The Penod Apnl. September 1985.

Workshop NUREG/CR4379 V02: LONG-TERM PERFORMANCE OF MATERIALS NUREG/CP-0073: PROCEEDINGS OF THE WORKSHOP ON LARGE IR-USED FOR HIGH-LEVEL WASTE PACKAGING.Second Quarterly RADIATOR RADIATION SAFETY.

Report. Year Four July-September 1985 NUREG/CR4379 V03: LONG-TERM PERFORMANCE OF MATERIALS Zircoloy-4 USED FOR HIGH. LEVEL WASTE PACKAGING.Thrd Quarterly NUREG/CR-4476: HIGH-TEMPERATURE OXIDATION OF ZlRCALOY4 Report, Year Four October -December 1985.

IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

5 e

~ ~ _ _ _

NRC Originating Organization index (Staff Reports)

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

OFFICE OF EXECtTTIVE DIRECTOR FOR OPERATIONS (EDO)

NRC - NO DETAILED AFFILIATION GIVEN REGION 1. OFFICE OF DIRECTOR NUREG-1179 V01: F.UPTURE OF MODEL 48f UF6 CYLINDER AND NUREG-0837 V05 NO3: NRC TLD DIRECT RADIATION MONITORING RELEASE OF URANIUM HEXAFLUORIDE.Sequoyah Fuels NETWORK. Progress Report. July-September 1985.

Facety. Gore Oklahoma, January 4,1986.

EDO-OFFICE OF ADMINISTRATION FOR 984 DI O

ECH ICAL OR M

NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS O

REPORTS. Annual Compdahon for 1985.

METHODOLOGY. Volume 1: Methodology Report.

NUREG-0540 V0/ N11: TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG/CR-4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS LY AVAILABLE. November 1-30,1985.

METHODOLOGY. Volume 2: Codes And Example Prot.lems.

OFFit:E OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

LY AVA LABLE

-3 985 NUREG4540 V08 N01: TITLE LIST OF DOCUMENTS MADE PUBUC-OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR NUREG/CP4072 V01: PROCEEDINGS OF THE THIRTEENTH LY AVAILABLE. January 1-31,1986 NUREG-0750 V22102: INDEXES TO NUCLEAR REGULATORY COM.

WATER REACTOR SAFETY RESEARCH INFORMATON MEETING.

MISSON ISSUANCES FOR JULY-DECEMBER 1985 NUREG/CP 0072 V02: PROCEEDINGS OF THE THIRTEENTH NUREG-0750 V22 N05: NUCLEAR REGULATORY COMMISSION IS.

WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

SUANCES FOR NOVEMBER 1985. Pages 771-873.

NUREG/CP-0072 V03. PROCEEDINGS OF THE THIRTEENTH NUREG-0750 V22 N06: NUCLEAR REGULATORY COMMISSION IS.

WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

SUANCES FOR DECEMBER 1985. Pages 875-982.

NUREG/CP-0072 V04: PROCEEDINGS OF THE THIRTEENTH NUREG-0750 V23 NO1: NUCLEAR REGULATORY COMMISSION IS-WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

SUANCES FOR JANUARY 1986. Pages 1-47.

NUREG/CP-0072 V05: PROCEEDINGS OF THE THIRTEENTH DIVISION OF RULES AND RECORDS WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

NUREG4936 V04 N04: NRC REGULATORY AGENDA.Ouarterty NUREG/CP 0072 V06. PROCEEDINGS OF THE THIRTEENTH Report October-December 1985.

WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

EDO OFFeCE OF EXECUTIVE LEGAL DIRECTOR 8 0429 NU EG4386 D ST T N LEAR REGULATORY COM-O E PR O FR M R MISSION STAFF PRACTICE AND PROCEDURE DIGEST. JULY 1972 MATERIALS ENGINEERING BRANCH

- JUNE 1985-NUREG-0975 V04: COMPILATION OF CONTRACT RESEARCH FOR EDO - OFFICE OF STATE PROCRAMS THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGI-OFFICE OF STATE PROGRAMS, DIRECTOR NEERING TECHNOLOGY. Annual Report For FY 1985.

NUREG/CP-0073: PROCEEDINGS OF THE WORKSHOP ON LARGE IRRADIATOR RADIATON SAFETY.

INTRA AGENCY COMMITTEES, REVIEW GROUPS, ETC.

INCIDENT INVESTIGATION TEAM EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG-1190: LOSS OF POWER AND WATER HAMMER EVENT AT DATA SAN ONOFRE. UNIT 1 ON NOVEMBER 21,1985.

AEOD. DIRECTOR'S OFFICE NUREG 1195: LOSS OF INTEGRATED CONTROL SYSTEM POWER NUREG4090 V08 NO3: REPORT TO CONGRESS ON ABNORMAL AND OVERCOOUNG TRANSIENT AT RANCHO SECO ON DECEM-OCCURRENCES. July September 1985.

BER 26,1985.

OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80)

EDO-RESOURCE MANAGEMENT DIRECTOR'S OFFICE, OFFICE OF INSPECTION AND ENFORCEMENT OFFICE OF RESOURCE MANAGEMENT, DIRECTOP NUREG4430 V06 N01: UCENSED FUEL FACluTY STATUS NUREG 0325 R08: U.S. NUCLEAR REGULATORY COMMISSION REPORT. inventory Difference Data January-June 1985.(Gray Book 11)

FUNCTIONAL ORGANIZATON CHARTS.

NUREG-0940 V04 N04: ENFORCEMENT ACTIONS:SIGNIFICANT AC-DIVISION OF BUDGET & ANALYSIS TONS RESOLVED Ouarterty Progress Report, October-December NUREG4020 V09 N12: UCENSED OPERATING REACTORS STATUS 1985.

SUMMARY

REPORT. Data As Of November 30,1985.(Gray Book f)

DIVISION OF GA, VENDOR & TECHNICAL TRAINING CENTER PRO-NUREG4020 V10 Not: UCENSED OPERATING REACTORS STATUS GRAMS (POST 85021

SUMMARY

REPORT. Data As Of December 31.1985 (Gray Book 1).

NUREG4040 V09 N04: UCENSEE CONTRACTOR AND VENDOR IN-NUREG4020 V10 NO2: LICENSED OPERATING REACTORS STATUS SPECTION STATUS REPORT. October 1985-December 1985 (White

SUMMARY

REPORT. Data As Of January 31.1986.(Gray Book l}

Book)

NUREG-1100 V02: FY 1987 BUDGET ESTIMATES.

MANAGEMENT SUPPORT BRANCH OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG-0748 VOS N11: OPERATING REACTORS LICENSING AC.

N a a As Q hm M8@any M EG-68 T EVA V DEPARTMENT OF ENERGY PROPOSAL FOR MONITORED RETRIEVABLE STOR-OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

FACil ASSESSMENT & STANDARDIZATION BRANCH 515)

NUREG-0525 R11: SAFZGUARDS

SUMMARY

EVENT UST (SSEL).

NUREG4853 S05: SAFETY EVALUATION REPORT RELA'ED TO O

1 ONS TE OtSPOSAL OF RADIOACTIVE THE OPERATION OF CLINTON POWER STATON,Uhli NO.

WASTE. Guidance For Disposai By Subsurface Bunal.

No 4

P DIV pp W TE EA OR LICENSIN1 A

U.S. NUCLEAR REGULATORY COMMISSION (POST 851125)

COMMlSSIONERS NUREG-0954 S05: SAFETY EVALUAllON REPORT RELATEL' TO NUREG-0885105: U.S. NUCLEAR REGULATORY COMM:SSION 1986 THE OPERATION OF CATAWBA NUCLEAR STATION,UN113 1 POUCY AND PLANN;NG GUIDANCE.

AND 2. Docket Nos. 50 413 And 50-414 (Duke Power Company) 55

56 NRC Originating Organization index NUREG 1031 SOS: SAFETY EVALUATION REPORT RELATED TO REACTOR AT PENNSvlVANIA STATE UNIVERSITY. Docket No. 50-THE OPERATION OF MILLSTONE NUCLEAR POWER 005 (Pennsylvania $ tate Unrvers }

A R W

NS G OgG DI i O STATION, UNIT No.3. Docket No. 50-423. (Northeast Nuclear Energy THE OPERATION OF PERRY NUCLEAR POWER PLANT, UNITS 1 NU G 11 1 DRFT: DRAFT ENVIRONMENTAL STATEMENT RELAT.

A

2. Docket Nos. 60 AM W41.@velaM Mc hnab ED TO OPERATION OF THE SOUTH TEXAS PROJECT, UNITS 1 SAFETY EVALUATION REPORT RELATED TO AND 2. Docket Nos. 50-498 And 50-499.(Houston Lighting And Power NU E 887 Company)

THE OPERATION OF PERRY NUCLEAR POWER PLANT, UNITS 1

{

NUREG-1176. TECHNICAL SPECIFICATIONS FOR M LLSTONE NU-AND 2 Docket Nos. 50-440 And 50-441 (Cleveland Electric illutninat-CLEAR POWER STATION UNIT NO. 3 Docket No. 50423 (Northeast ing Company)

NUHEG-1162: TECHNICAL SPECIFICATIONS FOR PERRY NUCLEAR Nuclear Energy Co)

NUREG-1182: TECHNICAL SPECIFICATIONS FOR CATAWBA NU-POWER PLANT, UNIT 1 Docket No. 50-440 (Cleveland Electnc illumi.

CLEAR STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-nating Company).

DIVISION OF SAFETY REVIEW & OVERSIGHT (POST 851125) 414 (Duke Power Company)

NUREG-0933 SO4: A PRIORITIZATION OF GENERIC SAFETY DIVISION OF PRESSURIZED WATER REACTOR LICENSING - B

( ST 85M25)

N REG 109 DRFT FC. REGULATORY ANALYSIS FOR THE RESO-WUREG 1158 SAFETY EVALUATION REPORT RELATED TO THE LUTION OF UNRESOLVED SAFETY ISSUE A-44. STATION BLACK-RENEWAL OF THE OPERATING LICENSE FOR THE RESEARCH OUT.

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organizations that sponsor organization is followed by the sponsored the contractor reports listed in NUREG/CR number and title of the this compilation. It is arranged alphabetically report (s) prepared by that organizaticq. 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 numbei.

(o.g.,

divisions) where appropriate. The OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

NUREG/CR4171:

TRAC-PF1/ MOD 1 INDEPENJENT REGION 1. OFFICE OF DIRECTOR ASSESSMENT: LOBI LARGE BREAK TRANSIENT A1-04R.

NUREG/CR4372: PROBABAUSTIC RISK ASSESSMENT (PRA) AP-NUREG/CR4255 V02: AEROSOL RELEASE AND TRANSPORT 'RO-PUCATIONS.

GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEFEM-NUREG/CR4565: PROBABluSTIC SAFETY STUDY APPUCATIONS BER 1985.

PROGRAM FOR INSPECTION OF INDIAN POINT UNIT 3 NUCLEAR NUREG/CR-4327: ORGANIC IODIDE FORMATION FOLLOWING NU-FOWER PLANT.

CLEAR REACTOR ACCIDENTS.

NUREG/CR-4337:

TRAC-PF1/ MOD 1 INDEPU. TNT EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL ASSESSMENT.DARTMOUTH COLLEGE AIR-WATER COUNTCR-DATA I

CURRENT FLOW TESTS NUREG/CR-4343: INTEGRATED SEVERE ACCIDENT CONTA IN-NU E O

2-LICENSEE EVENT REPORT (LER)

COMPILATION For Month Of December 1985 MENT ANALYSIS WITH THE CONTAIN COMPUTER CODE.

NUREG/CR-2000 V05 N1: UCENSEE EVENT REPORT (LER)

NUREG/CR-4348 V01: COMMIX 18 A THREE-DIMENSIONAL TRA N-COMPILATON f or Month Of Janu 1986 SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERM AL NUREG/CR-2000 VOS N2-UCEN EE EVENT REPORT (LER)

HYDRAUUC ANALYSIS OF SINGLE AND MULTICOMPONEllT j

COMPILATION For Month Of February 1986.

SYSTEMS.Vol1. Equations And Numencs.

NUREG/CR-4348 V02: COMM;X-18:A THREE-DIMENSIONAL TRA 4-OFFICE OF INSPECTION & ENFORCEMENT (POST DIVISION OF EMERGENCY PREPAREDNESS & ENGINEE12/11/80) RING RE*

SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HYDRAUUC ANALYSIS OF SINGLE AND MULTICOMPONENT SPONSE (POST 830103)

SYSTEMS Vol II User's Manual.

NUREG/CR-3365 DRF FC: REPORT TO THE NRC ON GUIDANCE NUREG/CR-4359: INDEPENDENT ASSESSMENT OF TRAC-PF1 FOR PREPARING SCENARIOS FOR EMERGENCY PREPARED-(VERSION 7.0)RELAPS/ MOO 1(CYCLE 14).AND TRAC-BD1 (VER.

NESS EXERCISES AT NUCLEAR GENERATING STATIONS Draft SiON 12 0) CODES USING SEPARATE-EFFECTS EXPERIMENTS.

Report For Coument.

NUREG/CR-4371; COMMIX-2.A THREE-OlMENSIONAL TRANSIEN1 NUREG/CR-3790: CLOSEOUT OF IE BULLETIN 79-07. SEISMIC COMPUTER PROGRAM FOR THERMAL-HYDRAUUC ANALYSIS STRESS ANALYSIS OF SAFETY-RELATED PIPING.

OF TWO-PHASE FLOWS NUREG/CR-4000 V01: THE MESORAD DOSE ASSESSMENT NUREG/CR-4390: DCC-1/DCC-2 DEGRADED CORE COOLABluTY MODELVolume 1:Techrucal Basis.

DIVISION F VE R & TECHNICAL TRAINING CENTER PRO.

NURE 393.

SUMMARY

OF SEMISCALE SMALL BREAK LOSS-NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR EN_

OF-COOLANT ACCIDENT EXPERIMENTS (1979 TO 1985).

VIRONMENT -lMPL! CATIONS FOR NRC OUAUTY ASSURANCE NUREG/CR-4402 V02: HIGH TEMPERATURE GAS-COOLED REAC-PROGRAMS BASED ON NUCLEAR POWER INDUSTRY AND REG-TOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT ULATORY PROJECTIONS THROUGH 1995 EVALUATION.Ouarterty Progress Report. Apnl 1. June 30,1985.

NUREG/CR4446.

THE NUCLEAR INDUSTRY AND ITS NUREG/CR-4420- TURC11ARGE SCALE METALUC MELT-CONCEN-REGULATORS A NEW COMPACT IS NEEDED.

TRATE INTERACTION EXPERIMENTS AND ANALYSIS.

NUREG/CR-4434: ASSESSMENT OF MODELLING NEEDS FOR OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS.

DIVISION OF SAFEGUARDS NUREG/CR4438: RESULTS OF SEMISCALE MOOL2C SMALL NUREG/CR4059; EVALUATION OF THE IMPACT OF THE MC&A BREAK (5%) LOSS-OF-COOLANT ACCIDENT EXPERIMENTS S-REFORM AMENDMENTS ON A REPROCESSING FACruTY.

LH-1 AND S-LH-2.

DIVISION OF WASTE MANAGEMENT NUREG/CR-4442: TRAC USER'S GUIDE.

NUREGICR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICA-NUREG/CR4452: REVIEW OF RELAP5 CALCULATIONS FOR H.B.

TION TESTS.Sermannual Report Covenng The Penod Aprd-Septem-ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY.

ber 1985.

NUREG/CR-4453 V01: UGHT-WATER. REACTOR SAFETY FUEL NUREG/CR4369: QUALITY ASSURANCE (QA) PLAN FOR COMPUT-SYSTEMS RESEARCH PROGRAMS.

Quarterly Progress ER SOFTWARE SUPPORTING THE U.S. NUCLEAR REGULATORY Report. January-March 1985.

COMMISSION'S HIGH-LEVEL WASTE PROGRAM-NUREG/CR-4453: LIGHT WATER REACTOR SAFETY RESEARCH NUREG/CR4370 V01 UPDATE OF PART 61 IMPACTS ANALYSIS PROGRAM Semaannual Report. October 1983 March 1984 METHODOLOGY Volume 1: Methodology Report-NUREG/CR4465:

TRAC PF1/ MOD 1 INDEPENDENT NUREG/CR4370 V02: UPDATE OF PART 61 IMPACTS ANALYSIS ASSESSMENT:SEMISCALE MOD 2A INTERMEDIATE BREAK TEST METHODOLOGY Volume 2: Codes And Examp;e Problems.

S-4B-3 NUREG/CR4433: DOCUMENT REVIEW REGARDING HAZARDOUS NUREGICR-4466: STATION BLACKOUT TRANSIENTS IN THE SE-CHEMICAL CHARACTERISTICS OF LOW-LEVEL WASTE.

NUREGiCR-4450 DRF FC: MANAGEMENT OF RADtOACTIVE MIXED NU 6 H GH-TEMPERATURE OXIDATION OF ZlRCALOY.

IN COMMERCIAL LOW-LEVEL WASTES. Draft Report For 4 IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

NUREG/CR-4477: METHODOLOGIES FOR ASSESSING LONG TERM NUREG/CR-4501: MODEUNG OF VAPOR GENERATION IN FLASH-PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK-NR / R-4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER NUREG/CH-4509-WASTE PACKAGE RELIABIUTY.

DIVI N R K N LYS S ER NS (POST 840429)

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

NUREG/CR-3958: EFFECTS OF CONTROL SYSTEM FAILURES ON DIVISION OF ACCIDENT EVALUATION TRANSIENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A NUREG/CR-2331 V05 N2: SAFETY RESEARCH PROGRAMS SPON-COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR.

SORED BY OFFICE OF NUCLEAR REGULATORY NUREGICR-4132: NUCLEAR POWER SAFETY REPORTING SYSTEM RESEARCH Ouarterfy Progress Report.Apnl 1-June 30,1985.

FINAL EVALUATION RESULTS.

NUREG/CR-3983. STEAM EXPLOSION EXPERIMENTS AT INTERME.

NUREG/CR-4183 V01: PRESSURIZED THERMAL SHOCK EVALUA-DIATE SCALE.FITSB SERIES TION OF THE 1 B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

57

I 58 NRC Contract Sponsor Index NUREG/CR-4183 V02-PRESSURIZED THERMAL SHOCK EVALUA-NUREG/CR-4300 V02: ACOUSTIC EMISSON/ FLAW RELATONSHIP TON OF THE H.B. ROBINSON UNIT 2 NUCLEAR POWER PLANT.

FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR-4188 VO1: NUCLEAR POWER PLANT SIMULATON FA-VESSELS. Progress Report. Apnt-September 1985.

CluTY EVALUATON METHODOLOGY. Handbook.

NUREG/CR4302 V01: AGING AND SERVICE WEAR OF CHECK NUREG/CR-4188 V02: NUCLEAR POWER PLANT SIMULATON FA-VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTEMS OF CIUTY EVALUATON METHODOLOGY.Tectncal Bases.

NUCLEAR POWER PLANTS.

NUREG/CR-4378: OBJECTIVE INDICATORS OF ORGANIZATIONAL NUREG/CR4307 V01: LWR PRESSURE VESSEL SURVEILLANCE PERFORMANCE AT NUCLEAR POWER PLANTS.

DOSIMETRY IMPROVEMENT PROGRAM. Progress Report. October NUREG/CR-4438 V01: HUMAN REUABluTY IMPACT ON INSERVICE 1984 - September 1985.

INSPECTION. Volume 1: Phase 1 Summary Report.

NUREG/CR4310- INVESTIGATION OF POTENTIAL FIRE-RELATED NUREG/CR-4436 V02: HUMAN RELIABluTY IMPACT ON IfdERVICE DAMAGE TO SAFETY-RELATED EQUIPMENT IN NUCLEAR INSPECTON. Volume 2: Rewsw And Analysis Of Human Perform-POWER PLANTS.

ance in Nondestructive Testing (Emphasinng Ultrasonics).

NUREG/CR-4324. TESTING OF NUCLEAR OUAUFIED CABLES AND NUREG/CR4460: UNCERTAINTY AND SENSITMTY ANALYSIS OF PRESSURE TRANSMITTERS IN SIMULATED HYDROGEf1 DEFLA-AN UPPER PLENUM TEST PROBLEM FOR THE MAEROS AERO-GRATIONS TO DETERMINE SURVIVAL MARGINS ANr) SENSITIVI-SOL MODEL TIES.

NUREG/CR-4473: A STUDY OF THE OPERATION AND MAINTE-NUREG/CR4328: PROBABluTY BASED LOAD COMBINATON CRI-NANCE OF COMPUTER SYSTEMS TO MEET THE REQUIRE-TERIA FOR DESIGN OF SHEAR WALL STRUCTURES.

MENTS OF 10 CFR 73 55.

NUREG/CR4363; A STUDY ON FABRICATION CRITERIA FOR DUC-NUREG/CR-4496: A SYSTEM FOR GENERATING LONG STREAM-TILE CAST IRON SPENT FUEL SHIPPING CONTAINERS.

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN NUREG/CR4366: RELIABluTY ASSESSMENT OF CONTAINMENT PERIOD.

TANGENTIAL SHEAR FAILURE.

NUREG/CR-4514: CONTROLUNG PRINCIPLES FOR PROR PROBA-NUREG/CR4380: EVALUATION OF THE MOTOR-OPERATED VALVE BILITY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST DEGRADATON. INCORRECT ADJUSTMENTS.AND OTHER AB-840429)

NORMAUTIES IN MOTOR-OPERATED VALVES.

NUREG/CR4113: FLOW AND DISPERSION NEAR CLUSTERS OF NUREG/CR-4389-PRESSURE NOISE IN PRESSURIZED WA1EH RE-BUILDINGS.

ACTORS.

NUREG/CR-4323: THE PROTECTION OF URANIUM TAluNGS IM-NUREG/CR-4431:

SUMMARY

REPORT ON THE SEISMIC SAFETY POUNDMENTS AGAINST OVERLAND EROSlON.

MARGINS RESEARCH PROGRAM.

NUREG/CR-4364: MANAGEMENT PERCEPTON OF THE HEALTH NUREG/CR-4464: PERFORMANCE DEMONSTRATON TESTS FOR PHYSICS TECHNICIAN JOB.

DETECTON OF INTERGRANULAR STRESS CORROSON CRACK.

NUREG/CR4379 V02: WNG-TERM PERFORMANCE OF MATERI-ING.

ALS USED FOR HIGH-LEVEL WASTE PACKAGING.Second Quar-NUREG/CR4468: ADAPTATION OF OCA-P,A PROBABlUSTIC FRAC-terty Report, Year Four July-September 1985.

TURE-MECHANICS CODE.TO A PERSONAL COMPUTER.

NUREG/CR-4379 V03: LONG-TERM PERFORMANCE OF MATERI.

NUREG/CR-4469 VO1: INTEGRATION OF NONDESTRUCTIVE EXAM-ALS USED FOR HIGH-LEVEL WASTE PACKtGING. Third Quarterty INATON RELIABluTY AND FRACTURE MECHANICS: Semiannual Report, Year Four October -Decereber 1985.

Report Apnl - September 1984.

NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSVENT OF NUREG/CR4472-SIAMESE IMAGING TECHNIQUE FOR QUASI-VER-U.S.AND FOREIGN NUCLEAR POWER PLANT DOFE EXPER1-TICAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

ENCE.

flUREG/CR-4474: SCALE MODEUNG OF REINFORCED CONCRETE NUREG/CR-4419: BIOASSAY MEASUREMENTS FON URANIUM CATEGORY I STRUCTURES SUBJECTED TO SEISMIC LOADING.

USING SPUTTER INITIATED RESONANCE ONIZAflON SPEC-NUREG/CR4475: ORMGEN PC.A MICROCOMPUTER PROGRAM TROSCOPy.

FOR AUTOMATIC MESH GENERATON OF 2 D CRACK GEOME-NUREGICR4502: VIRGINIA REGIONAL SEISMIC NETWORK. FINAL TRIES.

REPORT (19771985).

NUREG/CR-4479: THE USE OF A FIELD MODEL TO ASSESS FIRE NUREG/CR4504: LONG-TERM SURVEILLANCE AND MONITORING BEHAVIOR IN COMPLEX NUCLEAR POWER PLANT OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL-ENCLOSURES PRESENT CAPABILITIES AND FUTURE PROS-INGS PILES.

PECTS.

NUREG/CR-4585: INVESTIGATION OF THE STABIUTY OF CLAY /

NUREG/CR-4482 DRF FC: RECOMMENDATIONS TO THE NUCLEAR BASALT PACKING MATERIALS.

REGULATORY COMMISSION ON TRIAL GUIDEUNES FOR SEIS-DIVISION OF ENGINEERING TECHNOLOGY MIC MARGIN REVIEWS OF NUCLEAR POWER PLANTS. Draft NURCO/CF,-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS Repor1 For Comment.

ON SOUDtFICATION. WASTE DISPOSAL,AND ASSOCIATED OCCU-NUREG/CR-4486: VISA 11 - A COMPUTER CODE FOR PREDICTING PATIONAL EXPOSURE.

THE PROBABluTY OF REACTOR PRESSURE VESSEL FAILURE.

NUREG/CR-3760 A STUDY ON DUCTILE AND BRITTLE FAILURE NUREG/CR45'.8: DEVELOPMENT AND VERIFICATION OF CONDI-DESIGN CRITERIA FOR DUCTILE CAST IRON SPENT-FUEL SHIP-TONS FOR DUCTILE TEARING INSTABILITY AND ARREST.

PING CONTAINERS.

NUREG/CR-3805 V03: ENGINEERING CHARACTERIZATION OF EDO-RESOURCE MANAGEMENT GROUND MOTION. Task 11: Observational Data On Spatial Vanations OFFICE OF RESOURCE MANAGEMENT, DIRECTOR Of Earthquake Ground Moton.

NUREG/CR-4546:

LABOR PRODUCTIVITY ADJUSTMENT NUREG/CR-4078: PROGRAM FOR FIELD VALIDATION OF THE SYN-FACTORS.A Method For Estimating Labor Construction Costs Asso-THETIC APERTURE FOCUSING TECHNIQUE FOR ULTRASON'C oated With Physical Modifications To Nuclear Power Plants.

TESTING (SAFT UT) Final Report.

NUREG/CR-4555: GENERIC COST ESTIMATES FOR THE DISPOSAL NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM PHASE OF RADIOACTIVE WASTES.

II. Semiannual Report.Apnl 1985 - September 1985 DIVISION OF BUDGET & ANALYSIS NUREG/CR4219 V02-HEAVY-SECTON STEEL TECHNOLOGY PRO-NUREG/CR-2907 V03: RADIOACTIVE MATERIALS RELEASED FROM GRAM SEMIANNUAL PROGRESS REPORT FOR ARRfL-SEPTEM.

NUCLEAR POWER PLANTS. Antiual Rooort 1982.

NUREG/CR 276 V02: VIBRATION AND WEAR IN STEAM GENERA.

OF OF R REA TO T

DIRECTO (POST TOR TUBES FOLLOWING CHEMICAL CLEANING.

NUREG/CR-4279 V01: AGim ANr) SERVICE WEAR OF HYDRAUUC 851125)

NUREG/CR-3137: SEISMIC AND DYNAMIC OUAUFICATION OF RE-AND MECHANICAL S'.v8BERS USED ON SAFETY RELATED LATED ELECTRICAL AND MECHANICAL EQUIPMENT.

PIPING AND COMPONENTS OF NUCLEAR POWER PLANTS. Phase NUREG/CR-4223: STEEL CONTAINMENT RESISTANCE UNDER l Study GENERAL DYNAMIC PRESSURES.

NUREG/CR-4289: RESIDUAL RADIONUCLIDE CCNTAMINATION NUREG/CR-4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUX-WITHIN AND AROUND COMMERCIAL NUCLEAR POWER IUARY FEEDWATER SYSTEM REUABluTY ANALYSIS PLANTS. ORIGIN. DISTRIBUTION, INVENTORY AND DECOMMIS-NUREG/CR-4411: ASSESSMENT OF SPECIAUZED EDUCATIONAL S!ONING ASSESSMENT.

PROGRAMS FOR UCENSED NUCLEAR REACTOR OPERATORS.

NUREG/CR-4293: REUABluTY ANALYSIS OF SHEAR WALL STRUC-NUREG/CR4462: A RANKING OF SABOTAGE / TAMPERING AVOC-TURES.

ANCE TECHNOLOGY ALTERNATIVES NUREG/CR-4299: PRELIMINARY EVALUATON OF EFFLUENT RA.

NUREG/CR-4547:

CONTEMPT 4/ MOD 8.A MULTICOMPONENT DIOACTIVITY MONITORING SYSTEMS FOR BWR PLANTS.

SYSTEM ANALYSIS PROGRAM.

i NRC Contract Sponsor index 59 DIVISION OF HUMAN FACTORS TECHNOLOGY (POST 851125)

DIVISION OF BOILING WATER REACTOR (BWR) LICENSING NUREG/CR-3517: RECOMMENDATIONS TO THE NRC ON HUMAN NUREG/CR-4494: RADIOLOGICAL ASSESSMENT OF BWR RECIR-ENGINEERING GUIDELINER FOR NUCLEAR - POWER PLANT CULATORY PIPE REPLACEMENT.

MAINTAINABILITY.

NUREG/CR-4545: PIPE CRACK EVALUATION IN OPERATING BOIL-DMSION OF PRESSURIZED WATER REACTOR LICENSING - A ING WATER REACTORS.

(POST 851125)

DMSION OF SAFETY REVIEW & OVERSIGHT (POST 851125)

NUREG/CR 3950 V02 FUEL PERFORMANCE ANNUAL REPORT NUREG/CR-4485:. THE IMPACT OF FUEL CLADDING FAILURE DIVI ION PRESSURIZED WATER REACTOR LICENSING - B EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NU.

(POST 851125g CLEAR POWER PLANTS. Case Study:PWR Dunng Routme Oper-NUREG/CR-4286: EVALUATON OF RADIOACTIVE LIQUID EFFLU-atens.

ENT DELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

NUREG/CR-4471: LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUDIES

SUMMARY

RESULTS AND CONCLUSIONS.

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Contractor index This index lists, in alphabetical order, the numbers and titles of their reports. If further contractors that prepared the NUREG/CR information is needed, refer to the main ci-reports listed in this compilation. Listed tation by the NUREG/CR number.

below each contractor are the NUREG/CR AD HOC PUBUC HEALTH ASSESSMENT TASK FORCE BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NUREG-1189 VOI: ASSESSMENT OF THE PUBUC HEALTH IMPACT LA8 ORATORIES FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH NUREG/CR-3365 DAF FC-REPORT TO THE NRC ON GUIDANCE FOR FUELS CORPORATION FACluTY AT GORE.OKLAHOMARepnnted PREPARING SCENARIOS FOR EMERGENCY PREPAREDNESS EX-March 26.1986.

ERCISES AT NUCLEAR GENERATING STATIONS. Draft Report For NUREG 1189 V02: ASSESSMENT OF THE PUBLIC HEALTH IMPACT Comment.

FROM THE ACCIDENTAL RELEASE OF UF6 AT THE SEQUOYAH NUREG/CR-3517: RECOMMENDATIONS TO THE NRC ON HUMAN EN-FUELS CORPORATON FACluTY AT GORE. OKLAHOMA-GINEERING GUIDEUNES FOR NUCLEAR POWER PLANT MAIN-TAINABILITY.

AEROSPACE CORP.

i NURIG/CR-3950 V02 FUEL PERFORMANCE ANNUAL REPORT FOR NUREG/CR-4132 NUCLEAR POWER SAFETY REPORTING SYSTEM 1984.

FINAL EVALUATON RESULTS NUREG/CR 3958: EFFECTS OF CONTROL SYSTEM FAILURES ON l

NUREG/CR4477: METHODOLOGIES FOR ASSESSING LONG-TERM TRANSIENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A l

PERFORMANCE OF HIGH-LEVEL RADIOACTIVE WASTE PACK.

COMBUSTON ENGINEERING PRESSURIZED WATER REACTOR.

AGES.

NUREG/CR-3959: TRANSITION TO AN OPERATING REACTOR ENVl-RONMENT -lMPUCATONS FOR NRC OUAUTY ASSURANCE PRO-l AMES LABORATORY, ENERGY & MINER AL RESOURCCS RESEARCH GRAMS BASED ON NUCLEAR POWER INDUSTRY AND REGULA-2 INSTITUTE NUREG/

4 STEEL NTAINMENT RESISTANCE UNDER GEN-NUR G/ R VO :

H ME O AD DOSE ASSESSMENT MODELVolume 1:Techrucal Basis.

ANALYSIS & TECHNOLOGY, INC.

NUREG/CR4276 V02: VIBRATION AND WEAR IN STEAM GENERA-i NUREG/CR4364: MANAGEMENT PERCEPTION OF THE HEALTH TOR TUBES FOLLOWING CHEMICAL CLEANING.

PHYSICS TECHNICIAN JOB.

NUREG/CR-4279 V01: AGING AND SERVICE WEAR OF HYDRAUUC AND MECHANICAL SNUBBERS USED ON SAFETY-RELATED PIPING ARGO4NE NATIONAL LABORATORY AND COMPONENTS OF NUCLEAR POWER PLANTS Phase i Study.

I NUREG/CR4348 V01: COMMIX-18 A THREE-DIMENSIONAL TRAN-NUREG/CR4289: RESIDUAL RADIONUCUDE CONTAMINATION SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY.

WITHIN AND AROUND COMMERCIAL NUCLEAR POWER DRAULIC \\NALYSIS OF SINGLE AND MULTICOMPONENT PLANTS. ORIGIN. DISTRIBUTION INVENTORY AND DECOMMISSION-SYSTEMS Vol LEquations And Numencs.

ING ASSESSVENT.

NUREG/CR4348 V02: COMM:X 18 A THREE-DIMENSIONAL TRAN-NUREG/C44300 V02: ACOUSTIC EMISSION / FLAW RELATIONSHIP SIENT SINGLE-PHASE COMPUTER PROGRAM FOR THERMAL HY.

FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE DRAULIC ANALYSIS OF SINGLE AND MULTICOMPONENT VESSELS. Progress Report. Aprd-September 1985.

SYS TEMS.Vol ll User's Manual.

NUREG/CR-4323: THE PROTECTION OF URANIUM TAlUNGS IM-NURE3/CR-4371: COMMIX 2A THREE-DIMENSIONAL TRANSIENT POUNDMENTS AGAINST OVERLAND EROSION.

COMPUTER PROGRAM FOR THERMAL-HY?RAULIC ANALYSIS OF NUREG/CR-4378: OBJECTIVF. INDICATORS OF ORGANIZATIONAL TWCLPHASE FLOWS.

PERFORMANCE AT NUCLEAR POWER PLANTS.

NUREG/CR4453 V01: UGHT WATER-REACTOR SAFETY FUEL SYS-NUREG/CR-4411: ASSESSMENT OF SPECIAUZED EDUCATIONAL TEM 3 RESEARCH PROGRAMS. Quarterty Progress Report. January-PROGRAMS FOR UCENSED NUCLEAH REACTOR OPERATORS.

Marcrt 1985.

NUREG/CR-4436 V01: HUMAN REUA81UTY IMPACT ON INSERVICE NUREG/CR-4501: MODELING OF VAPOR GENERATON IN FLASHING INSPECTON Volume 1: Phase 1 Summary Report.

FLOW.

NUREG/CR-4436 V02: HUMAN REUA8ILITY IMPACT ON INSERVICE INSPECTION. Volume 2: Review And Analysis Of Human Performance ASPEN SYSTEMS,INC.

In Nondestructive Testing (Emphaseng Ultrasorucs).

NUREG-0386 DO4: UNITED STATES NUCLEAR REGULATORY COM-NUREG/CR4462: A RANKING OF SABOTAGE / TAMPERING AVOID-MISSION STAFF PRACTICE AND PROCEDURE DIGEST. JULY 1972 -

ANCE TECHNOLOGY ALTERNATIVES.

JUNE 1985-j NUREG/CR-4464-PERFORMANCE DEM3NSTRATON TESTS FOR f

ATOM SCIENCES,1NC.

ETECTION OF INTERGRANULAR STHESS CORROSION CRACK.

NUREG/CR4419. BIOASSAY MEASUREMENTS FOR URANIUM USING NURE'G/CR4469 V01: INTEGRATON OF NONDESTRUCTIVE EXAMI-i j

SPUTTER INITIATED RESONANCE lONIZATION SPECTROSCOPY.

NATION REUA8tuTY AND FRACTURE MECHANICS.Semeannual I

EATTELLE HUMAN AFFAIRS RESEARCH CENTERS ReportA$ - September 1984.

NUREG/CR-4472: SIAMESE IMAGING TECHNOUE FOR QUASt-VERTI-NUREG/CR4411: ASSESSMENT OF SPECIAUZED EDUCATIONAL PROGRAMS FOR LICENSED NUCLEAR REACTOR OPERATORS CAL TYPE (OVT) DEFECTS IN NUCLEAR REACTOR PIPING.

NUREG/CR4436 VOI: HUMAN RELIA 81UTY IMPACT ON INSERVICE NUREG/CR-4473: A STUDY OF THE OPERATION AND MAINTENANCE INSPECTION. Volume 1: Phase 1 Summary Report.

OF COMPUTER SYSTEMS TO MEET THE REQUIREMENTS OF 10 NUREG/CR4436 V02: HUMAN REUABluTY IMPACT ON INSERVICE CFR 73 55.

INSPECTON Volume 2: Review And Analysis Of Human Performance NUREG/CR-4476: HIGH-TEMPERATURE OXIDATION OF ZlRCALOY4 i

in Nondestructnre Testm0 (Emphasizmg Ultrasorucs).

IN STEAM AND STEAM-HYDROGEN ENVIRONMENTS.

NUREG/CR4485: THE IMPACT OF FUEL CLADOING FA' LURE CATTELLE MEMORIAL INSTITUTE, COLUM8US LABORATORIES EVENTS ON OCCUPATIONAL RADIATON EXPOSURES AT NUCLE-NUREG/CR-4082 V03: DEGRADED PIPING PROGRAM PHASE AR POWER PLANTS. Case Study PWR During Routine Operations.

Il Semiannual Report Apnl 1985 - September 1985.

NUPEG/CR-4486: VISA 11. A COMPUTER CODE FOR PREDICTING NUREG/CR4379 V02: LONG-TERM PERFORMANCE OF MATERIALS THE PROBABlWTY OF REACTOR PRESSURE VESSEL FAILURE.

USED FOR HIGH LEVEL WASTE PACKAGING Second Quarterfy NUREG/CR4434 RADIOLOGICAL ASSESSMENT OF BWR RECIRCU.

Report, Year Four July-September 1985 LATORY PIPE REPLACEMENT.

NUREG/CR4379 V03. LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR45C4: LONG-TERM SURVEILLANCE AND MONITORING USED FOR HIGH-LEVEL WASTE PACKAGING. Third Quarterfy OF DECOMMISSIONED URANIUM PROCESSING SITES AND TAIL-Report. Year Four October -December 1985 INGS PtLES.

61

62 Contractor Index BROOKHAVEN NATIONAL LA80RATORY NUREG/CR-4370 VG2: UPDATE OF PART 61 IMPACTS ANALYSIS NUREG/CP-0072 V01: PROCEEDINGS OF THE THIRTEENTH WATER METHODOLOGY. Volume 2: Codes And Example Problems.

REACTOR SAFETY RESEARCH INFORMATION MEETING NUREG/CP-0072 V02: PROCEEDINGS OF THE THIRTEENTH WATER HANFORD ENGINEERING DEVELOPMENT LABORATORY REACTOR SAFETY RESEARCH INFORMATION MEETING NUREG/CR-4307 VO1: LWR PRESSURE VESSEL SURVEILLANCE DO-NUREG/CP-0072 V03: PROCEEDINGS OF THE THIRTEENTH WATER SIMETRY IMPROVEMENT PROGRAM. Progress Report October REACTOR SAFETY RESEARCH INFORMATION MEETING 1984 September 1985.

NUREG/CP-0072 V04: PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING IDAHO NATIONAL ENGINEERING LABORATORY NUREG/CP.0072 V05: PROCEEDINGS OF THE THIRTEENTH WATER NUREG/CR-4393.

SUMMARY

OF SEMISCALE SMALL BREAK LOSS-REACTOR SAFETY RESEARCH INFORMATION MEETING OF-COOLANT ACCIDENT EXPERIMENTS (1979 TO 1985).

NUREG/CP-0072 V06: PROCEEDINGS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING INTERNATIONAL ENERGY ASSOCIATES. LTD.

NUREGICR-2331 V05 N2: SAFETY RESEARCH PROGRAMS SPON-NUREG/CR4446.

THE NUCLEAR INDUSTRY AND ITS SORED BY OFFICE OF NUCLEAR REGULATORY REGULATORS A NEW COMPACT IS NEEDED.

RESEARCH Ouarterfy Proaress Report.Aprd 1 June 30.1985.

NUREG/CR 2907 V03: RAOIOACTIVE MATERIALS RELEASED FROM INTERPACIFIC TECHNOLOGY INC.

NUCLEAR POWER PLANTS. Annual Report 1982.

NUREG/CR-3805 V03. ENGINEERING CHARACTERIZATION OF NUREG/CR-3091 V07: REVIEW OF WASTE PACKAGE VERIFICATION GROUND MOTION Task 11: Observational Data On Spatial Vanatens TESTS.Sermannual Report Covenng The Penod Apnl. September 1985-Of Earthquake Ground Motion.

NUREG/CR-3137: SEISMIC AND DYNAMIC QUAUFICATION OF RE-LATED ELECTRICAL AND MECHANICAL EQUIPMENT.

LAWRENCE LIVERMORE NATIONAL LA80RATORY NUREG/CR-3444 V03: THE IMPACT OF LWR DECONTAMINATIONS NUREG/CR-3760: A STUDY ON DUCTILE AND BRITTLE FAILURE ON SOUOlFICATION. WASTE DtSPOSAL,AND ASSOCIATED OCCU-DESIGN CRITERIA FOR DUCTILE CAST IRON SPENT-FUEL SHIP-PATIONAL EXPOSURE.

PING CONTAINERS NUREG/CR4293: RELIABluTY ANALYSIS OF SHEAR WALL STRUC-NUREG/CR-4363: A STUDY ON FABRICATION CRITERIA FOR DUC-TURES-NUREG/CR4311: REVIEW OF THE SHEARON HARRIS UNIT 1 AUXIL-TILE CAST IRON SPENT-FUEL SHIPPING CONTAINERS.

NUREG/CR-4431:

SUMMARY

REPORT ON THE SEISMIC SAFETY lARY FEEDWATER SYSTEM RELIABILITY ANALYSIS-MARGINS RESEARCH PROGRAM'MENDATIONS TO THE NUCLEAR NUREG/CR4328: PROBA81UTY BASED LOAD COM81NATON CRITE' NUREG/CR-4482 DRF FC: RECOM NU R-43 9 N E EN N A SESS E T TRAC-PF1 (VER-REGULATORY COMMISSION ON TRIAL GUIDELINES FOR SEISMIC SON 7.0).RELAP5/ MOD 1(CYCLE 14).AND TRAC-801 (VERSION MARGIN REVIEWS OF NUCLEAR POWER PLANTS Draft Report For 12.0) ^., ODES USING SEPARATE-EFFECTS EXPERIMENTS.

Comn d NURE6/CR4364: MANAGEMENT PERCEPTION OF THE HEALTH LINSLEY, KRAEGER & ASSOCIATES, LTD.

NU E CR ELIA TY ASSESSMENT OF CONTAINMENT NUREG/CR-4490: A SYSTEM FOR GENERATING LONG STREAM-TANGENTIAL SHEAR FAILURE.

FLOW RECORDS FOR STUDY OF FLOODS OF LONG RETURN NUREG/CR-4372: PROBABALISTIC RISK ASSESSMENT (PRA) APPLI.

PERIOD.

CATIONS NUREG/CR-4381:

SUMMARY

OF COMPARATIVE ASSESSMENT OF LOS ALAMOS SCIENTIFIC LA80RATORY U S AND FOREIGN NUCLEAR POWER PLANT DOSE EXPERIENCE NUREG/CR-4442: TRAC USER'S GUIDE.

NUREG/CR-4433: DOCUMENT REVIEW REGARDING HAZARDOUS NUCEG/CR-4471. LOS ALAMOS PWR DECAY-HEAT REMOVAL STUD.

IES

SUMMARY

RESULTS AND CONCLUSONS.

CHEMICAL CHARACTERISTICS OF LOW-LEVEL WA4TE'EEDS NUREG/CR-4434 ASSESSMENT OF MODELUNJ N FOR NUREG/CR4474: SCALE MODEUNG OF REINFORCED CONCRETE SAFETY ANALYSIS OF CURRENT HTGR CONCEPTS CATEGORY I STRUCTURES SUBJECTED TO SEISMIC LOADING.

NUREG/CR4450 DRF FC: MANAGEMENT OF RADIOACTIVE MIXED W

IN COMMERCIAL LOW-LEVEL WASTES Draft Report For M

NUREG/C 546. LABOR PRODUCTIVITY ADJUSTMENT FACTORS A NUREG/CR4452. REVIEW OF RELAP5 CALCULATONS FOR H B.

Method For Estimating Labor Construction Costs Associated With ROBINSON UNIT 2 PRESSURIZED THEAVAL SHOCK STUDY.

Physmal Modifcations To Nuclear Power Plants.

NUREG/CR4479: THE USE OF A FIELD MODEL TO ASSESS FIRE BE-HAV10R IN COMPLEX NUCLEAR POWER PLANT MICHIGAN, UNIV. OF, ANN AR80R, MI ENCLOSURES PRESENT CAPABluTIES AND FUTURE PROSPECTS.

NUREG/CR4585: INVESTIGATION OF THE STABluTY OF CLAY /

NUREG/CR4509: WASTE PACKAGE RELIABluTY.

BASALT PACKING MATERIALS.

NUREG/CR-4545: PIPE CRACK EV ALUATION IN OPERAT;NG SOluNG WATER REACTORS, NUS CORP.

NUREG/Ch4547: CONTEMPT 4/ MOD 6.A MULTICOMPONENT SYSTEM NUREG/CR-4310: INVESTIGATON OF POTENTIAL FIRE-RELATED ANALYSIS PROGRAM DAMAGE TO SAFETY-RELATED EQUIPMENT IN NUCLEAR POWER NUREG/CR-4565: PROBABluSTIC SAFETY STUDY APPUCATIONS PLANTS.

PROGRAM FOR INSPECTION OF INDIAN POINT UNIT 3 NUCLEAR POWER PLANT.

OAK RIDGE NATIONAL LABORATORY NUREG/CR-2000 V04N12: LICENSEE EVENT REPORT (LER)

COMEX CORP.

COMPILATION For Month Of December 1985.

NUREG/CR-3365 DAF FC: REPORT TO THE NRC ON GUIDANCE FOR NUREG/CR-2000 V05 N1: LICENSEE EVENT REPORT. (LER)

PREPAHING SCENARIOS FOR EMERGENCY PREPAREDNESS EX-COMPILATON For Month Of January 1986.

ERCISES AT NUCLEAR GENERATING STATONS Draft Report For NUREG/CR-2000 VOS N2: LICENSEE EVENT REPORT FLER)

Cornment.

COMPILATION For Month Of February 1986.

NUREG/CR 4059: EVALUATON OF THE IMPACT OF THE MC&A COMMERCE. DEPT. OF, NATIONAL OCEANIC & ATMOSPHERIC REFORM AMENDMENTS ON A REPROCESSING FACILITY.

ADM!NISTRATION NUREG/CR4183 V01: PRESSURIZED THERMAL SHOCK EVALUATION NUREG/CR4113: FLOW AND DISPERSION NEAR CLUSTERS OF OF THE H B. ROslNSON UNIT 2 NUCLEAR POWER PLANT.

BUILDINGS.

NUREG/CR4183 V02: PRESSURIZED THERMAL SHOCK EVALUATON OF THE H.B ROBINSON UNIT 2 NUCLEAR POWER PLANT.

EG4G IDAHO,INC. (SUBS OF EG40, INC.)

NUREG/CR-4188 V01: NUCLEAR POWER PLANT SIMULATION FAClu-NUREG/CR4299. PREUMINARY EVALUATON OF EFFLUENT RADIO.

TY EVALUATION METHODOLOGY Handbook.

ACTIVITY MONITORING SYSTEMS FOR BWR PLANTS NUREG/CR-4188 V02: NUCLEAR POWER PLANT SIMULATON FAClu-NUREG/CR4438: RESULTS OF SEMISCALE MOO-2C SMALL BREAK TY EVALUATION METHODOLOGY.Tectncal Bases.

(5%) LOSS 4F-COOLANT ACCIDENT EXPERIMENTS S LH 1 AND S.

NUREG/CR4219 V02: HEAVY SECTON STEEL TECHNOLOGY PRO-LH-2 GRAM SEMIANNUAL PROGRESS REPORT FOR ARRIL SEPTEMBER NUREG/CR4466. STATION BLACKOUT TRANSIENTS IN THE SEMIS-1985.

CALE FACILITY, NUREG/CR-4255 V02: AEROSOL RELEASE AND TRANSPORT PRO.

GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER ENVIROSPHERE CO.

1985.

NUREG/CR-4370 V01: UPDATE OF PART 61 IMPACTS ANALYSIS NUREG/CR4286. EVALUATION OF RADIOACTIVE LIOUID EFFLUENT METHODOLOGY.Volurne 1: Methodology Report.

RELEASES FROM RANCHO SECO NUCLEAR POWER PLANT.

Contractor index 63 NUREG/CR4302 Vot: AGING AND SERVICE WEAR OF CHECK NUREG/CR-4369: OUALITY ASSURANCE (OA) PLAN FOR COMPUTER VALVES USED IN ENGINEERED SAFETY-FEATURE SYSTFMS OF EOFTWARE SUPPORTING THE U.S. NUCLEAR REGULATORY COM-NUCLEAR POWER PLANTS.

MISSION'S HIGH-LEVEL WASTE PROGRAM.

NUREG/CH-4027: ORGANIC IODIDE FORMATION FOLLOWING NU-NUREG/CR-4390: DCC 1/DCC-2 DEGRADED CORE COOLABILITY CLEAR REACTOR ACCIDENTS.

ANALYSIS.

NUREG/CR-4380: EVALUATION OF THE MOTOR-OPERATED VALVE NUREG/CR4420: TURC1:LARGE SCALE METALLIC MELT-CONCEN-ANALYSIS AND TEST SYSTEM (MOVATS) TO DETECT TRATE INTERACTION EXPERIMENTS AND ANALYSIS.

DEGRADATION,1NCORRECT ADJUSTMENTS.AND OTHEH ABNOR-NUREG/CR-4459: LIGHT WATER REACTOR SAFETY RESEARCH MALITIES IN MOTOR-OPERATED VALVES.

PROGRAM.Semannual Report October 1983 March 1984.

NUREG/CR4389: PRESSURE NOISE IN PRESSURIZED WATER RE-NUREG/CR-4460: UNCERTAINTY AND SENSITIVITY ANALYSIS OF AN ACTORS-UPPER PLENUM TEST PROBLEM FOR THE MAEROS AEROSOL NUREG/CR-4402 V02. HIGH-TEMPERATURE GAS-COOLED REACTOR MODEL SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NUREG/CR-4465:

TRAC-PF1/ MOD 1 INDEPENDENT EVALUATION Ouarterfy Progress Report, Apnl 1 June 30,1985.

ASSESSMENT:SEMISCALE MOD-2A INTERMEDIATE BREAK TEST

)

NUREG/CR-4468: ADAPTATION OF OCA-P.A PROBASILISTIC FRAC-S-18-3.

TURE-MECHANICS CODE.TO A PERSONAL COMPUTER-NUREG/CR-4514: CONTROLLING PRINCIPLES FOR PRIOR PROBA-NUREG/CR-4475: ORMGEN PC:A MICROCOMPUTER PROGRAM FOR BILITY ASSIGNMENTS IN NUCLEAR RISK ASSESSMENT.

AUTOMATIC MESH GENERATION OF 2-0 CRACK GEOMETRIES.

SCIENCE & ENGINEERING ASSOCIATES,INC.

N

/CR-4546: uBOR NMW MSWENT MCTORSA NUR G/C 90: CLOSEOUT OF IE BULLETIN 79-07: SEISMIC Method For Estimating Labor Construction Costs Assocated With STRESS ANALYSIS OF SAFETY-RELATED PIPING.

Physical Modifications To Nuclear Power Plants.

t NUREG/CR-4555: GENERIC COST ESTIMATES FOR THE DISPOSAL PURDUE UNIV WEST LAFAYETTE. lN OF RADIOACTIVE WASTES.

NUREG/CR-4556: HEAT TRANSFER FROM A ROD BUNDLE UNDER NATURAL CIRCULATION CONDITIONS.

SOUTHWEST RESEARCH INSTITUTE NUREG/CR-4078: PROGRAM FOR FIELD VALIDATION OF THE SYN-S. COHEN & ASSOCIATES. INC.

THETIC APERTURE FOCUSING TECHNIQUE FOR ULTRASONIC NUREG/CR-4555. GENERIC COST ESTIMATES FOR THE DISPOSAL OF RADIOACTIVE WASTES.

TESTING (SAFT UT). Final Report.

U.S. NAVAL ACADEMY, ANNAPOLIS, MD SANDIA NATIONAL LA80RATORIES NUREG/CR-3983: STEAM EXPLOSION EXPERIMENTS AT INTERMEDt-NUREG/CR-4528: DEVELOPMENT AND VERIFICATION OF CONDb r

ATE SCALE FITSB SERIES TiONS FOR DUCTILE TEARING lNSTABILITY AND ARREST.

NUREG/CR-4171:

TRAC-PF1/ MOD 1 INDEPENDENT f

ASSESSMENT. LOBI LARGE BREAK TRANSIENT A104R.

UNC NUCLEAR INOUSTRIES NUREG/CR4310: INVESTIGATION OF POTENTIAL FIRE-RELATED NUREG/CP-0068: PROCEEDINGS OF THE INTERNATIONAL NUCLEAR DAMAGE TO SAFETY-RELATED EQUIPMENT IN NUCLEAR POWER REACTOR DECOMMISSIONING PLANNING CONFERENCE.

PLANTS NUREG/CR-4324: TESTING OF NUCLEAR OUALIFIED CABLES AND VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., SLACKS 8URG, PRESSURE TRANSMITTERS IN SIMULATED HYDROGEN DEFLA.

VA GRATIONS TO DETERM!NE SURVIVAL MARGINS AND SENSITIVI.

NUREG/CR4502: VIRGINtA REGIONAL SEISMIC NETWORK. FINAL TIES.

REPORT (1977-1985).

NUREG/CR-4337:

TRAC-PF1/ MOD 1 INDEPENDENT l

ASSESSMENT.DARTMOUTH COLLEGE AIR-WATER COUNTERCUR.

WOOOWARD-CLYDE CONSULTANTS,INC.

]

RENT FLOW TESYS NUREG/CR 3805 V03: ENGINEERING CHARACTERIZATION OF NUREG/CH.4343: INTEGRATED SEVERE ACCIDENT CONTAINMENT GROUND MOTION. Task 11: Observational Data On Spatial Vanations ANALYSIS WITH THE CONTAIN COMNTER CODE.

Of Earthquake Ground Motion.

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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 ical order. They are preceded by their number.

S 413 Catawba Nudes Staten, Und 1, DAe Poww NUREG4954 S05 50 5 Pennsyvama State Unw. Reseach Reactor NUREG1158 Co.

50440 Perry Nudes Power Plant, UrW l. Oeveland NUREG-0887 SO8 50413 Catawba Nudear Staton, Une 1, DAe Power NUREG1182 Electne murmnatng C Co.

50440 Perry Nudear Power Plant, Unt 1 Develand NUREG4887 SO9 50 414 Catawba Nudes Staton, Und 2, Duke Power NUREG4954 S05 Electnc numnabng C Co.

50440 Perry Nudes Power Plant, Und 1. Develand NUREG 1162 50-414 Catawba Nudear Stabon, Und 2, Dde Power NUREG1182 Elecinc mumnanng C Co-50441 Perry Nudear Power Plant, Und 2. Oeveland NUREG4887 S06

%461 Onton Power Staton, Una 1, pros Power Co. NUREG4853 SOS Electne mumnann0 C r

[

% 261 KB. Robnson Plant, Una 2, Caotna Power & NUREG/CR4183.'01 50441 Perry Nudes Power Plant, Und 2. Oeveiand NdREG4867 SOS S 261 HB Plant, Und 2, Carchna Power & NUREGICR4t83 V02 50 312 Seco Generann0 taDon, NUREG1195 S

S261 HB Plant, Und 2, Crohna Power & NUREGICR4452 S 312 Nudear Generanng Staten, NUREG/CR428!

8 S 286 indan Stanon, Und 3, Power Authonty of NUREG!CR4565

% 206 bMm WLW NUREG 1190 40402 Nudear Okla NUR S 400 Shearon Power Plant, Urut 1, NUREG/CR4311 404027 Kerr-McGee Nudes Corp., Oklahoma Cdy, OK, NUREG1189 V02 Caro 6na Power & bght C S 423 Mdistone Nudear Power Staton, una 3, NUREG-1031 SOS STN 50498 South Texas Prgect, Und 1. Houston Ughnng & NUREG 1171 DRFT Northeast Nudear Energy Co Power Co.

4 423 Mdistone Nudes Power Staton, Und 3, NUREG1176 STN-50499 South Texas Prgect, Und 2. Houston Ughbng & NUREG1171 DRFT Northeast Nudes Energy Co Power Co.

65

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%"32 BIBLIOGRAPHIC DATA SHEET NUREG-0304, Vol 11, No. I

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This jou'rnal includes all formai.r orts in the t REG series orepared by the NRC staff and cont *: actors. as well as proceedi qs of conferences and workshoos. The entries in the compilation 'are indexed for acces by title /and abstract, contractor reoort number, personal author, subject, NRC organiza on, coritractor, and licensed facility.

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