Text

Regulatory and Technical Reports (Abstract Index Journal). Compilation for First Quarter 1987,January-March
	ML20214N796
Person / Time
Issue date:	05/31/1987
From:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
To:	;
References
	NUREG-0304, NUREG-0304-V12-N01, NUREG-304, NUREG-304-V12-N1, NUDOCS 8706030076
	Download: ML20214N796 (62)
	v • d • e

- - - - - -

NUREG-0304 Vol.12, No.1 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for First Quarter 1987 January - March U.S. Nuclear Regulatory Commission Office of Administration and Resources Management pa arouq gpe ;

Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.

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

( . _ _ _ _ _ _ _ _ _ _ .-

NUREG-0304 1

Vol.12, No.1 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for First Quarter 1987 January - March Date Published: May 1987 Policy and Publications Management Branch Division of Publications Services Office of Administration and Resources Management U.S. Nuclear Regulatory Commission Washington, DC 20555 y m.,

CONTENTS v

Pref ace .

Index Tab

. .1 Main Citations and Abstracts .

Staff Reports

Conference Proceedings

Contractor Reports

International Agreement Reports 2

Secondary Report Number Index .

3 Personal AuthorIndex 4 Subject index ... .... ......... ..

5 NRC Originating Organization Index (Staff Reports) . ...

NRC Originating Organization Index (International Agreements) . . 6 NRC Contract Sponsor index (Contractor Reports) . 7 8

Contractor index . .. .

.9 International Organization Index .

Licensed Facility index . . 10 iii

~ . _ . . .

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

Division of Publications Services Policy and Publications Management Branch Publishing and Translations Section Woodmont 637 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indexes:

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

NRC Originating Organization Index (International Agreements)

NRC Contract Sponsor Index (Contractor Reports)

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

The bibliographic elements of the main citations are the following:

Staff Report NUREG-0808: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

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

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

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

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

Sar.dia 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 accessic 1 number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use),

v

1 international Agreement Report NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA. NEUMANN, U. Kraftwerk Union. August 1986. 223 pp. 8608270424. 37659:138.

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

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

ADD - addendum APP - appendix DRFT - draft ERR - errata N - number R - revision S - supplement

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

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

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

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

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

Main Citations and Abstracts The report listings in this compilation are arranged by report nurnber, where NUREG-XXXX is an NRC staff-onginated report, NUREG/CP-XXXX is an NRC-sponsored conference report, NUREG/CR-XXXX is an NRC contractor-prepared report, and NUREG/lA-XXXX is an inter-national agreement report. The bibliographic information (see Preface for details) is followed by a brief abstract of this report.

abnormal occurrences reported by Agreement States. One in-NUREG-0020 V10 N09: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of August volved an uncontrolled release of krypton-85 to an unrestncted 31,1986(Gray Book 1) ROSS,P.A.; BEEBE,M.R. Division of area; the other involved a contaminated radiopharmaceutical Computer & Telecommunicatons Services (Pre 870413). March used in diagnostic administrations. The report also contains in-1987. 463pp. 8703170242. 40052.208. formation updating some previously reparted abnormal occur.

The OPERATING UNITS STATUS REPORT LICENSED OP- rences.

ERATING REACTORS provides data on the operaton of nucle-NUREG-0304 VII N04: REGULATORY AND TECHNICAL RE.

ar units as timety and accurately as possible. This information is PORTS (ABSTRACT INDEX JOURNAL). Annual Compilation collected by the Office of Resource Management from the For 1986.

Office of Administration (Pre 870413). March 1987.

Headquarters staff of NRC's Office of Inspection and Enforce. 200pp. 87040f 0175. 40324.254.

ment, from NRC's Regional Offices, and from utilities. The three This journal includes all formal reports in the NUREG series sections of the report are: monthly highlights and statistics for prepared by the NRC staff and contractors; proceedings of con-commercial operating units, and errata from prevously reported ferences and workshops; as well as international agreement re-data; a compilation of detailed informaton on each unit, provid.

ports. The entries in this compilation are indexed for access by ed by NRC's Regional Offices, IE Headquarters and the utilities; title and abstract, secondary report number, personal author, and an appendix for miscellaneous information such as spent subject, NRC organization for staff and international agree-fuel storage capability, reactor years of expenence and non- ments, contractor, international organization, and licensed facili-power reactors in the U S. It is hoped the report is helpful to all ty, agancies and individuals interested in maintaining an awareness of the U S. energy situation as a whole. NUREG-0325 RIO: U.S. NUCLEAH REGULATORY COMMISSION NUREG-0020 V10 N10: LICENSED OPERATING REACTORS FUNCTIONAL ORGANIZATION CHARTS.

Office of Adminis-STATUS

SUMMARY

REPORT. Data As Of September traton (Pre 870413). February 1987. 57pp. 8702240496.

30,1986 (Gray Book 1) ROSS,P.A.; BEEBE.M.R. Division of 39725.023.

Computer & Telecommunicatons Services (Pre 870413). March Functonal organizaton charts for the NRC Commission Of-1987. 447pp. 8704090032. 404C3 027, fices Divisions, and Branches are presented.

See NUREG-0020,V10,N09 abstract.

NUREG-0386 D04 R04: UNITED STATES NUCLEAR REGULA.

NUREG-0040 V10 N04: LICENSEE CONTRACTOR AND TORY COMMISSION STAFF PRACTICE AND PROCEDURE VENDOR INSPECTION STATUS REPORT. Quarterty DIGEST. July 1972 - June 1986.

Office of the General Coun-Report, October-December 1986.(White Book)

Divison of OA, sel. February 1987. 661pp. 8703170245. 40050 267.

Vendor & Technical Training Center Programs (850212- This Revision 4 of the fourth editon of the NRC Staff Practice 870411). February 1987. 274pp. 8703170192. 40047:139 and Procedure Digest contains a digest of a number of Com-This penodical covers the results of inspections performed by misson, Atomic Safety and Licensing Appeal Board, and Atomic the NRC's Vendor Program Branch that have been distnbuted Safety and Licensing Board decisions issued dunng the penod to the inspected organizations dunng the penod from October July 1,1972 to June 30,1986, interpreting the NRC Rules of 1986 thru December 1986. Also, included in this issue are the Practice in 10 CFR Part 2. This Revision 4 replaces in part earli-results of certain inspections performed poor to October 1986 er editions and supplements and includes appropnate changes that were not included in prevous issues of NUREG-0040. reflecting the amendment to the Rules of Practice effective June 30,1986.

NUREG-0090 V09 N02: REPORT TO CONGRESS ON ABNOR-MAL OCCURRENCES Apni-June 1986.

Office for Analysis & NUREG-0430 V07 N01: LICENSED FUEL FACILITY STATUS Evaluation of Operational Data, Director. January 1987. 71pp. REPORT.tnventory Difference Data. January-June 1986 (Gray 8703030835. 39866.152' Book ll)

Office of Inspection & Enforcement, Director (820201-Section 208 of the Energy Reorganization Act cf 1974 identi-870411). February 1987.15pp. 8703120210. 39984.016.

fies an abnormal occurrence as an unscheduled incident or NRC is committed to the penodic publication of licensed fuel event which the Nuclear Regulatory Commission determines to facilities inventory difference data, following agency review of be significant from the standpoint of public health and safety the information and completion of any related NRC investiga-and requires a quartorty report of such esents to be made to tions. Information in this report includes inventory difference Congress. This report covers the pered April 1 through June data for active fuel fabncation facilities possessing more than 30,1986 During the report penod, there were two abnormal oc- one effective kilogram of high ennched uranium, low enriched curronces at the nuclear power plants licensed to operate. One uranium, plutonium, or uranium 233.

involved an out of sequence control rod withdrawal and the other involved a boiling water reactor emergency core cooling NUREG-0525 R12: SAFEGUARDS

SUMMARY

EVENT LIST system design deficiency. There were five abnormal occur- (SSEL). GRAMANN.R H. Division of Safeguards (Pre 870413).

February 1987. 47pp. 8703130205. 40005.289.

rences at the other NRC bconsees. Two involved willful failure to report diagnostic medical misadministrations to the NRC; one The Safeguards Summary Event List provides bnet summa-involved a therapeutic medcal misadministration, and two in- nes of hundreds of safeguards related events involving nuclear volved diagnostic medcal misadministrations There were two matenal of facilities regulated by the U S. Nuclear Regulatory 1

2 Main Citations and Abotracts Comtrussen. Events are desenbed under the categones: bomb- issues remaining after issuance of the Safety Evaluation Report related, intrusion, missing / allegedly stolen, transportaton-relat- and Supplement No.1.

ed, tampenng/ vandalism, arson, firearms related, radologmal sabotage, nonradclogical sabotage, and miscellaneous. Infor. NUREG-0837 V06 NO3. NRC TLD DIRECT RADIATION MONI-mation in the event descnptons was obtained from official NRC TORING NETWORK Progress Report, July. September 1986.

reports. JANG,J.; RABATIN,K ; COHEN.L Region 1. Office of Drector, February 1987. 224pp. 8703120254. 39985:111.

NUREG-0540 V08 N10: TITLE LIST OF DOCUMENTS MADE This report provides the status and results of the h a Ther-PUBLICLY AVAILABLE. October 1-31,1986.

Duisen of Tech- moluminescent Dosimeter (TLD) Direct Radiaton Monitonng rwcal informaton & Document Control (Pre 870120). December Network. It presents the radiaton levels measured in the vicinity 1986. 479pp. 8701200421. 39353:106. of NRC hcensed facility sites throughout the country for the third This document is a monthly publicaton containing desenp- quarter of 1986.

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

This informabon includes (1) docketed rnatorial associated with NUREG-0853 S08: SAFETY EVALUATION REPORT RELATED civilian nuclear power plants and other uses of radcactue ma- TO THE OPERATION OF CLINTON POWER STATION, UNIT terials, and (2) nondocketed matenal recewed and generated by 1. Docket No. 50-481.(llianois Power Company et al)

Division of NRC pertinent to its role as a regulatory agench The following Boiling Water Reactor (BWR) Licensing (851125-870411).

Indexes are included: Personal Author Index, Corporate Source March 1987. 47pp. 8704090033. 40466:158.

Index, Report Number Index, and Cross Reference to Pnncipal Supplement No. 8 to the Safety Evaluation Repo: on the ap-Documents Index. plication filed by lilinois Power Company, Soyland Power Coop-eratue, Inc., and Western Illinois Power Cooperative, Inc., as NUREG-0540 V08 N11: TITLE LIST OF DOCUMENTS MADE applicants and owners, for a license to operate the Chnton PUBLICLY AVAILABLE. November 1-30,1986.

Office of Ad- Power Staten. Unit No.1, has been prepared by the Office of ministration (Pre 870413). January 1987. 377pp. 8701200372. Nuclear Reactor Regulation of the U.S. Nuclear Regulatory 39319.208. Commisson. The facility is located in Harp Township, DeWitt See NUREG-0540,V08,N10 abstract. County, Illinois. This supplement reports the status of items that NUREG-0540 V08 N12: TITLE LIST OF DOCUMENTS MADE have been resolved by the staff since Supplement No. 7 was PUBLICLY AVAILABLE. December 1-31,1986.

Office of Ad- lasued.

ministration (Pre 870413). February 1987. 439pp. 8703120259. NUREG-0857 S11: SAFETY EVALUATION REPORT RELATED 39984032. TO THE OPERATION OF PALO VERDE NUCLEAR GENERAT.

See NUREG-0540,V08,N10 abstract ING STATION, UNITS 1,2 AND 3. Docket Nos. 50-528,50-529 NUREG 0540 V09 N01: TITLE LIST OF DCCUMENTS MADE And 50-530 (Arizona Public Service Company.et al)

Division of PUBLICLY AVAILABLE. January 1-31,1987.* Office of Adminis- Pressunzed Water Reactor Licensing . B (851125-870411),

tration (Pro 870413). March 1987. 350pp. 8704060097. March 1987. 51pp. 8704080187. 40441:240.

40398.060. Supplement No.11 to the Safety Evaluation Report for the See NUREG4540,V08,N10 abstract. application filed by Arizona Public Service Company, et al, for hcenses to operate the Palo Verde Nuclear Generating Station, MUREG-0750 V24 N01: NUCLEAR REGULATORY COMMISSION Units 1,2 and 3 (Docket Nos. STN 50-528/529/530) located in ISSUANCES FOR JULY 1986 Pages 1195.

Office of Adminis- Maricopa County, Arizona, has been prepared by the Office of tration (Pre 870413). January 1987. 204pp. 8702170032. Nuclear Reactor Regulation of the U.S. Nuclear Regulatory 39659 214. Commission. The purpose of this supplement is to update the Legal issuances of the Commission, the Atomic Safety and Li- Safety Evaluation Report by providing an evaluation of (1) addi-censing Appeal Panel, the Atomic Safety and Licensing Board tional information submitted by the applicant since Supplement Panel, the Administrative Law Judge, and NRC Program Offices No.10 was issued and (2) other matters requinng staff review are presented, since Supplement No.10 was issued, specifically those issues HUREG-0750 V24 N02: NUCLEAR REGULATORY COMMISSION that required resolution before Unit 3 low. power licensing.

ISSUANCES FOR AUGUST 1986.Pages 197-396.

Office of NUREG 0878 S08: SAFETY EVALUATION REPORT RELATED Administration (Pre 870413). February 1987. 208pp. TO THE OPERATION OF BYRON STATION, UNITS 1 AND e RE .075d V2 'N01 abstract' cM Ws 506 M 50M @mmonweam Mson Company)

Division of Pressunzed Water Reactor Licensing -

NUREG-0750 V24 NO3: NUCLEAR REGULATORY COMMISSION A (851125-870411). March 1987. 24pp. 8704010187, ISSUANCES FOR SEPTEMBER 1986 Pages 397-488.

Office 40321:134.

of Administration (Pre 870413). March 1987. 102pp. Supplement No. 8 to the Safety Evaluation Report related to 8704010165.40319.061. Commonwealth Edison Company's apphcation for licenses to See NUREG-0750,V24,N01 abstract. operate the Byron Station, Units 1 and 2, located in Rockvale Township, Igle County, Ilknots, has been prepared by the Office NUREG-0781 S02: SAFETY EVALUATION REPORT RELATED of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory TO THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 Commission. This supplement provides recent information re-AND 2 Docket Nos. 50-498 And 50-499. (Houston Lighting And garding resoluton of the hcense conditions identified in the Power Company)

Dtvision of Pressurized Water Reactor Li- SER. Because of the favorable resolution of the items dis-censing - A (851125-870411). January 1987. 76pp. cussed in this report, the staff concludes that Byron Station, 8702190398. 39690.058- Unit 2 can be operated by the licensee at power levels greater The Safety Evaluation Report issued in Apnl 1986 provided than 5% without endangenng the health and safety of the the results of the NRC staff's review of the Houston Lighting public.

and Power Company's appbcation for hcenses to operate the South Texas Project. The facility consists of two pressunzed NUREG-0933 S06: A PRIORITIZATION OF GENERIC SAFETY water nuclear reactors located in Matagorda County, Texas. ISSUES. EMRIT,R.; VANDERMOLEN.H.; PITTMAN.J.; et al Di-Supplement No.1, issued in September 1986 updated the infor- vision of Safety Review & Oversight (851125 870411). March maton contained in the Safety Evaluation Report and ad- 1987. 325p. 8704090020. 40467.073.

dressed the ACAS Report issued on June 10, 1986. Supple- The report presents the pnonty rankings for genenc safety ment No. 2 addresses and resolves some of the outstanding issues related to nuclear power plants. The purpose of these I

1 - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

3 l Main Citations and Abstracts f

seismic expenence data, supplemented by existing seismc test rankings is to assist in the timely and eff cient allocation of NRC data, apphed in accordance with the guidelines developed, pro-resou'ces for the resolution of those safety issues that have a vides the most reasonable a'temative to current quahfication cri-segnificant potential for reducing nsk. The safety poonty raniungs teria to venty the seismic adequacy of equipment in operating are HIGH, MEDIUM, LOW, and DROP and have been assagned nuclear plants. Explicit seismic quahfication should be required on the basis of nsk gnificance estimates, the ratio of nsk to only if seismic expenence data or existing test data on similar costs and other impacts estimated to result if resolutions of the components cannot be shown to apply.

safety issues were implemented, and the consideration of un-certainties and other quantitatrve of quahtatue factors. To the NUREG 1057 SO4: SAFETY EVALUATION REPORT RELATED extent practical, estimates are quantitatue. TO THE OPERATION OF BEAVER 50-412VALLEY (Duquesne POWER Light STATION, UNIT 2. Docket No.

NUREG-0936 V05 Nb3: NRC REGULATORY AGENDA.Ouarterly Company.et al)

Dvision of Pressunzed Water Reactor Licens-Report, July-September 1906

Divison of Rules & Records A (851125-870411). March 1987. 9f pp. 8704010184.

January 1987.160pp. 8702060261. 39527.089. ing (Pre 870413). 40324.162.

The NRC Regulatory Agenda is a compilation of all rules on Supplement No. 4 to the Safety Evaluaton Report for the ap- 5 which the NRC has proposed or is considenng action and all phcation filed by Duquesne Light Company, et al., for license to petitons for rulemaking which have been received by the Com- operate the Beaver Valley Power Station, Unit 2 (Docket No.

mission and are pending disposition by the Commission. The 50-412), located in Beaver County, Pennsylvania, has been pre-Regulatory Agenda is updated and issued each quarter. The pared by the Office of Nuclear Reactor Regulaton of the Nucle-Agendas for April and October are pubhshed in their entirety in er Regulatory Commission. The purpose of this supplement is to the Federal Register while a notice of availability is published in the Federaf Register for the January and July Agendas.

update the Safety Evaluaton of (1) additional information sub-mitted by the applicants since Supplement No. 3 was issuec',

NUREG-0940 V05 N04: ENFORCEMENT ACTIONS SIGNIFICANT and (2) matters that the staff had under review when Supple-ACTIONS RESOLVED.Ouarterly Progress Report, October-De- ment No. 3 was issued.

cember 1986

Office of Inspection & Enforcement, Director (820 870411). February 1987. 579pp. 8703120167. NUREG-1100 V03: BUDGET ESTIMATES Fiscal Years 1988-1989.

Division of Budget & Analysis (Pre 870413). January This' co$ipilaton summanzes significant enforcement actions 1987, 72pp. 8702060083. 39526.087.

that have been resolved dunng one quarterly penod (October- This report contains the fiscal year budget }ustificatons to December 1986) and includes copies of letters, Notices, and Congress. The budget provides estimates for salanes and ex-Orders sent by the Nuclear Regulatory Commission to licensees penses for fiscal years 1988 1989 with respect to these enforcement actions and the hcensee's V02: ONSITE DISPOSAL OF RADIOACTIVE NUREG-1101 responses. It is anticipated that the information in this pubhca- WASTE Methodology For The Radiological Assessment Of Dis-bon wdl be widely disseminated to managers and employees posal By Subsurface Dunal. NEUDER S M.; KENNEDY,W.E. Di-engaged in activities licensed by the NRC,in the interest of pro- vision of Waste Management (Pre 870413). February 1987, moting public health and safety as well as common defense 50pp 8704010173 40340 309.

and secunty. Volume 1 of this NUREG provides guidance for academic, NUREG-0975 V05: COMPILATION OF CONTRACT RESEARCH medical, and industnal licensees seeking authorization to dis-FOR THE MATERIALS BRANCH, DIVISION OF ENGINEERING pose of small quantities of radioactwe matenal by onsite subsur-SAFETY. Annual Rept For FY 1986.

Divison of Engineenng face disposal. Licensee requests for such authonzations are Safety (860720 870413). March 1987. 411pp. 8704030183. made pursuant to Section 20.302 of 10 CFR Part 20 " Standards 40377:267. for Protection Against Radiation." This volume (Volume 2) de-This report presents summanes of the research work per- scnbes the entena and technical methodology used by NRC formed dunng Fiscal Year 1986 by laboratones and organiza. staff to evaluate requests by licensees for aporoval of onsite tons under contracts administered by the NRC's Matenals disposal by bunal in soil. The tochnical methodology includes Branch, Office of Nuclear Regulatory Research. Each contractor the ONSITE/MAXIt code for calculating radiological exposure has wntten a more complete and detailed annual report of their from vanous pathways, the MOMOD84 code, and analytical work which can be obtained by wnting to NRC; however, we be. methods for calculating contaminant transport and concentra-lieve it is useful to have a summary of each contractor's efforts ton of radenuclides in flowing groundwater. Radiological expo-for the year combined into one volume, sure analyses include the following pathways: (1) exposure to direct gamma fr m any surface c ntamination or buned waste, HUREG 1030: SEISMIC QUAUFICATION OF EQUlPMENT IN OP. *

ERATlNG NUCLEAR POWER PLANTS Unresolved Safety issue 9**

denuchdes. (3) mgestrng agncultural products denved from radu A-48. CHANG.T Y Dwision of Safety Review & Oversight 1987. 183pp. 8703130028 onuchde- contaminated soit, and (4) inhahng radonuchdes resu-(8511254 70411). February spended at the bunal site. Licensee-proposed disposal actwities 40007;196. are evaluated in terms of radological impact on pubhc health The margin of safety provided in existing nuclear power plant and safety and the environment. The estimated committed ef-equipment to resist seismically induced loads and perform their fectwo dose equivalent resulting from the technical evaluation intended safety functons may vary considerably, because of will usually be the determining factor in the authonzation of the signsficant changes in design entena and methods for the seis-mec quahfication of equipment over the years Therefore, the proposed disposal.

seismic quahfication of equipment in operating plants must be NUREG tt37 S05: SAFETY EVALUATION REPORT RELATED reassessed to determine whether requahficaten is necessary. TO THE OPERATION OF VOGTLE ELECTRIC GENERATING The obective t

of USl A-46 is to estabhsh an emphcit set of PLANT, UNITS 1 AND 2 Docket Nos. 50-424 And 50425 (Geor.

guedehnes and acceptance critena to judge the seismic adequa- gia Power Company,et al) ' Division of Pressurized Water Reac-cy of equipment at all operatng plants,in heu of requinng quah- tor Licensing A (851125-870411). January 1987, 121pp.

fication to the current cntena that are apphed to new plants. 8702060100. 39526 288 This report summanzes the work accomphshed on USI A 46 by in June 1985, the staff of the Nuclear Regulatory Commission the Nuclear Regulatory Commission staff and its contractors. In tssued its Safety Evaluation Report (NUREG 1137) regarding addition, the collection and review of seismic expenence data the appbcation of Georgia Power Company, Municipal Electnc and eusting seismic test data by the SOUG and EPRI respec- Authonty of Georgia, Ogoithorpe Power Corporation, and the twely, and the review and recommendations of the SSRAP are City of Datton, Georgia, for hcenses to operate the Vogtfe Elec.

presented The pnncipal technical finding of USl A-46 ts that

4 Main Citations and Abstracts tric Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 50-425). Supplement 1 to NUREG-1137 was issued by the staff severe core damage, the performance of containment struc-in October 1985, Supplement 2 was issued in May 1986, Sup-tures under tevere accident loadings, possible radioactive re-piement 3 was issued in August 1986, and Supplement 4 was leases into the environment if the containment were to fail, and issued in December 1986. The facility is located in Burke the oftsste consequences of such releases. Volume 3 of this County, Georgia, approximately 26 miles south-southeast of Au- document provides discussions of NRC staff analyses of specif-gusta, Georgia, and on the Savannah River. This fifth supple- ic technical and regulatory issues, compares present nsk results ment to NUREG-1137 provides recent informaton regardmg with those of other studies, and describes computer codes used resolution of some of the open and confirmatory items that re- in the risk analyses.

mained unresoNed at the time the Safety Evaluaten Report was issM NUREG-1163: COORDINATION OF SAFETY RESEARCH FOR THE BABCOCK AND WILCOX INTEGRAL SYSTEM TEST NUREG 1137 S06: SAFETY EVALUATION REPORT RELATED PROGRAM. YOUNG,M.W.; SURSOCK,J.P. Division of Reactor TO THE OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Geor- System Safety (860720-870413). March 1987, 250pp.

gia Power Company et al)

Drvision of Pressunzed Water Reac- 8704010410.40328:068.

tor Licensing - A (851125-870411). March 1987. 125pp. This report describes the MIST facihty and all the Integral 8704010183. 40341:294. System Test (IST) support projects sponsored by the USNRC in June 1985, the staff of the Nuclear Regulatory Commission and by EPRI. These support projects have been deemed to play lasued its Saft,ty Evaluation Report (NUREG-1137) regarding an essential role in helping resolve issues raised by MIST scal-the apphcation of Georgia Power Company, Municipal Electnc ing compromises. Each support project is desenbed in detail Authonty of Georgia Oglethorpe Power Corporation, and the and application of the expected data to resolution of issues is City of Dalton, Georgia, for hcenses to operate the Vogtle Elec- discussed. The combined effort of MIST and seven other sup-tric Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and port projects will resolve virtually all questions addressed by the 50-425). Supplement I to NUREG.1137 was issued by the staff IST program, in October 1985, Supplement 2 was issued in May 1988. Sup.

s radin Dece r NUREG 1199: STANDARD FORMAT AND CONTENT OF A Lt.

6a pl n't 5 was s n Jan- CENSE APPLICATION FOR A LOW-LEVEL RADIOACTIVE uary 1987. The facihty is located in Burke County, Georgia, ap- WASTE DISPOSAL FACILITY,

Division of Waste Management proximately 28 miles south-southeast of Augusta, Georgia, and (Pre 870413), January 1987.108pp. 8702170341, 39668:281.

on the Savannah Rrver. This sixth supplement to NUREG 1137 The Standard Format and Content of a License Application provides recent information regarding resolution of some of the for a Low-Level Radioactive Waste Disposal Facility, NUREG.

opnn and confirmatory items that remained unresolved at the time the Safety Evaluation Report was issued. 1199, discusses the information to be provided in the Safety Analysis Report and estabhshes a uniform format for presenting NUREG-1150 DRF V1 FC: REACTOR RISK REFERENCE the information required to meet the kensq requirements fw DOCUMENT. Main Report Draft For Comment. ERNST,M.L.; land disposal of radioactive waste as required by 10 CFR 61.

MURPHY,J.A.; CUNNINGHAM,M.A.; et al. Office of Nuclear The use of the Standard Format will (1) help ensure that the Regulatory Research, Director (Post 860720). February 1987. Safety Ana'ysis Report (SAR) contains the information required 364pp. 8703100040. 39934.011.

by 10 CFR 61, (2) aid the appiscant in ensunng that the informa-This document discusses the nsks of severe accidents in a tion is complete. (3) help persons reading the SAR to locate in-set of commercial nuclear power plants. This risk is character.

formation, and (4) contnbute to shortening the time required for ized by the types and frequencies of accidents leading to the review process. The Standard Format and Content severe core damage, the performance of containment struc-tures under severe accident loadings, possible radioactrve re- (NUREG-1199) ensures that the information required to perform the review is provided, and in a usable format while the Stand-leases into the environment if the containment were to fail, and the offsite consequences of such releases. Volume 1 of this ard Review Plan, NUREG 1200, defines the technical review document summarizes the principal results of the risk anatyses, process. These documents provide assurance that NRC can review and process a license application wrthin 15 months and and displays these results in the context of specific regulatory lasues (e g., safety goals). meet the requirements of Section 9(1) and (2) of P.L.99-240, the Low. Level Radioactive Waste Policy Amendments Act NUREG 1150 DRF V2 FC: REACTOR RISK REFERENCE (LLRWPAA) of 1985.

DOCUMENT. Appendices A l. Draft For Comment.

DENNING.R.S.; LEONARD,M.; WREATHALL J. Office of Nucle.

NUREG 1200: STANDARD REVIEW PLAN FOR THE REVIEW OF at Regulatory Research, Director (Post 860720). February 1987.

320pp. 8703170207. 40053:342. A LICENSE APPLICATION FOR A LOW LEVEL RADIOACTIVE WASTE DISPOSAL FACILITY,

Division of Waste Managemsnt This document discusses the nsks of severe accidents in a (Pre 870413). January 1987. 473pp. 8702170380. 39669:243.

set of commercial nuclear power plants. This risk is character-The Standard Review Plan (SRP) is prepared for the guid-12ed by the types and frequencies of accidents leading to severe core damage, the performance of containment struc- ance of staff reviewers in the Office of Nuclear Material Safety tures under severe accident loadings, possible radioactive re- and Safeguards in performing safety reviews of applications to leases into the environment if the containment were to fait, and construct and operate a low-level waste disposal facility. The the offsite consequences of such releases. Volume 2 of this pnncipal purpose of the SRP is to assure the quality and uni-document provides a discussion of the methods used to calcu- formity of staff reviews and to present a well-defined base from late nsk, and summanzes the pnncipal results of the analyses of which to evaluate proposed changes in the scope and require.

the studied plants. ments of reviews. It is also a purpose of the SRP to make infor.

mation about regulatory matters widely available and to improve NUREG-1150 DRF V3 FC: REACTOR RISK REFERENCE communication and understanding of the staff's review process DOCUMENT. Appendices J-0. Draft For Comment.

Office of by interested members of the public and the nuclear industry.

Nuclear Regulatory Research, Director (Post 860720). February NUREG 1200 consists of 11 Chapters containing approximately 1987. 517pp. 8703180080. 40072.142.

60 individual SRP sections. Each section identifies who per.

This document discusses the nska of severe accidents in a forms the review, the matters that are reviewed, the basis for set of commercial nuclear power plants. This risk is character-review, how the review is performed, and the conclusions that l2ed by the types and frequencies of accidents leading to are sought.

1 1 IAaln Citations and Abstracts 5

}

of USl A-46, (3) a Summary of A-48 Tasks, (4) a Proposed im-NUREG 1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR piementation Procedure, (5) a Value-impact Analysis, (6) Imple-INCIDENT RESPONSE TRAINING ACCIDENT mentation, (7) a Summary of A-46 Risk Analyses and (8) Oper-

[ MANUALOverview And Summary Of Major Points.

MCKENNA.T.J.; MARTIN,J A.; MILLER.C.W.; et al. Division of ating Plants To Be Reviewed to USI A-46 Requirements.

Emergency Preparedness & Engineering Response (850212-870411). February 1987.111pp. 8703090078. 39931:107. NUREG-1224: SAFETY EVALUATION REPORT RELATED TO This is one in a series of volumes that collectuely provide for THE RENEWAL OF THE OPERATING LICENSE FOR THE the U.S. Nuclear Regulato6y Commission (NRC) emergency re- UNIVERSITY OF NEW MEXICO RESEARCH REACTOR. Docket No. 50-252. (University Of New Mexico)

Dwison of Pressur-sponse personnel the necessary background information for an adequate response to severe reactor accidents. The volumes in ized Water Reactor Licensing - 0 (851125-870411). March the senes are: Volume 1, " Overview and Summary of Major 1987. 48pp. 8704090018. 40466:110.

Points," Volume 2. " Severe Reactor Overview," Volume 3. "Re- This Safety Evaluation Report for the application filed by the sponse of Licensee and State and Local Officials," Volume 4, Unwersity of New Mexico for renewal of operating license

,Pubhc Protective Actions - Predetermined Cnteria and initial number R 102 to continue to operate a research reactor has Actons,, and Volume 5, U S. Nuclear Regulatory Commission been prepared by the Office of Nuclear Reactor Regulation of Response. Each volume serves, respectuely, as the text for a the U.S. Nuclear Regulatory Commission. The facility is owned course of instruction in a senes of courses for NRC response and oP* rated by the University of New Mexico and is located on personnel. These matenals do not provide guidance or license the Unwersity's campus in Albuquerque, New Mexico. The staff requirements for NRC licensees or state or local response orga- concludes that the AGN-201M type reactor facility can continue nizations. Each volume is accompanied by an appendix of to be operated by the University of New Mexico without endan-slides that can be used to present this material. The slides are genng the health and safety of the pubhc.

called out in the text.

NUREG-1210 V02: PILOT PROGRAM.NRC SEVERE REACTORNUREG 1237: TECHNICAL SPECIFICATIONS FOR VOGTLE ELECTRIC GENERATING PLANT, UNIT 1. Docket No. 50-ACCfDENT INCIDENT RESPONSE TRAINING MANUALSevere 424.(Georgia Power Company)

Division of Pressurized Water Reactor Accident Overview. MCKENN A T.J.; MARTIN,J A.;

MILLER,C.W.; et al. Division of Emergency Freparedness & En- Reactor Licensing A (851125-870411). January 1987.460pp.

gineenng Response (850212-870411). February 1987. 132pp. 8702060231, 39538:194.

8703090131. 39931:218. The Vogtle Electric Generatng Plant, Unit No.1, Technical See NUREG 1210,V01 abstract. Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and NUREG 1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR TRAINING other requirements apphcable to a nuclear facihty as set forth in ACCIDENT INCIDENT RESPONSE Secton 50.36 of 10 CFR 50 for the protection of the health and MANUALResponse Of Licensee And State And Local Officials.

SAKENAS,C.A.; MCKENNA T.J.; MILLER,C.W.; et al. Division of safety of the public.

Emergency Preparedness & Engineering Response (850212-870411). February 1987.101pp. 8703090090. 39931:006. NUREG-1240: TECHNICAL SPECIFICATIONS FOR SHEARON See NUREG-1210,V01 abstract. HARRIS NUCLEAR POWER PLANT UNIT 1. Docket No. 50-400.(Carohna Power & Light Company)

Division of Pressurized NUREG-1210 V04: PILOT PROGRAM.NRC SEVERE REACTOR Water Reactor Licensing A (851125-870411). January 1987, ACCIDENT INCIDENT RESPONSE TRAINING MANUAL.Public 473pp. 8702040174. 39508:110.

Protective Actions - Predetermined Cntena And initial Actions. The Shearon Har'is Nuclear Power Plant, Unit 1 Technical MARTIN.J.A.; MCKENNA,T.J.; MILLER,C.W.; et al. Division of Specifications were prepared by the U.S. Nuclear Regulatory Emergency Preparedness & Engineenng Response (350212- Commission to set forth the hmits, operating conditions, and 870411). February 1987,117pp. 8703090147. 39922.001. other requirements applicable to a nuclear facility as set forth in See NUREG 1210,V01 abstract. Secton 50.36 of 10 CFR 50 for the protection of the health and NUREG 1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR safety of the public.

ACCIDENT INCIDENT RESPONSE TRA NING MANUAL.U.S.

Nuclear Regulatory Commission Response. SAKENAS,C A. NUREG-124t GROUND-WATER PROTECTION ACTIVITIES OF MCKENNA,T,J.; PERKINS K.; et al. Division of Emergency Pre THE U S. NUCLEAR REGULATORY COMMISSION.

Division paredness & Engineenng Response (850212-870411). February of Waste Management (Pre 870413). February 1987. 66pp.

1987.105pp. 8703090067. 39921:258. 8703130105.40007:134.

See NUREG 1210,V01 abstract. The U.S. Nuclear Regulatory Commission (NRC) provides for ground- water protection through regulations and licensing con-NUREG f 211: REGULA70RY ANALYSIS FOR RESOLUTION OFditions that require prevention, detection, and correction of UNRESOLVED SAFETY ISSUE A-46. SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS. CHANG,T.Y.; ground-water contamination. Prepared by the interoffice Division of bafety Review & Oversight Ground Water Protection Grcup, this report evaluates the inter-ANDERSON.N R.

February 1987. 73pp. 8703120330. nal consistency of NRC's ground- water protection programs.

(8511 5 870411).

These programs have evolved consistently with growing public The margin of safety provided in existing nuclear power plant concerns about the significance of ground water contamination equipment to resist seismically induced loads and perform re- and environmental impacts. Early NRC programs provided for quired safety functions may vary considerably, because of sig- the protection of public health and safety by minimizing releases nificant changes in design cntena and methods for the seismic of radionuchdes. More recent programs have included provi-qualificat:on of equipment over the years. Therefore, the seismic sions for minimizing releases of non-radiological constituents, quahfication of equipment in operating plants must be reas. mitigating environmental impacts, and correcting ground water sessed to determine whether requalification is necessary The contamination. NRC's ground-water protecton programs are objective of USI A-46 is to estabbsh an exphcit set of guideltnes categorized according to program areas, including nuclear mate-and acceptance cnteria to judge the seismic adequacy of equip- nals and waste management (NMSS), nuclear reactor oper-ment at all operating plants, in heu of requinng these pants to ations (NRR), confirmatory research and standards develop-meet the entena that are applied to new plants. This report pre- ment (RES). inspecton and enforcement (IE), and agreement sents the regulatory analysis for Unresolved Safety issue (USI) state programs (SP).

A 46. It includes (1) Statement of the Problem, (2) the Ob tective

6 Main Citations and Abstracts NUREG-1247: TECHNICAL SPECLFICATIONS FOR VOGTLE This report documents a standard format suggested by the ELECTRIC GENERATING PLANT, UNIT 1. Docket No. 50-NRC for use in prepanng fundamental nuclear matenal control 424 (Georgia Power Company)

Division of Pressurized Water Reactor Licensing - A (851125-870411). March 1987. 450pp.

plans as required by the Matenal Control and Accounting 8704020068. 40344 250. Reform Amendment (portions of 10 CFR Part 74). The report The Vogtle Electric Generating Plant, Unit No.1, Technical also desenbes the necessary contents of a comprehensive plan and provides example acceptance entena which are intended to Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and communicate acceptable rneans of achieving the performance other requirements applicable to a nuclear facility as set forth in capabilities of the Reform Amendment. By using the suggested Section 50.36 of 10 CFR 50 for the protection of the health and format, the license applicant will minimize administrative prob-safety of the public. lems associated with the submittal review and approval of the FNMC plan. Preparabon of the plan in accordance with this MUREG-1248: TECHNICAL SPECIFICATIONS FOR PALO VERDE format will assist the NRC in evaluating the plan and in stand-NUCLEAR GENEPATING STATION UNIT 3. Docket No. 50 ardizing the review and hcensing process. However, conform-530.(Artrona Pubhc Service Company)

Division of Pressunzed ance with this guidance is not required by the NRC. A license Water Reactor Licensing - B (851125-870411). March 1987. applicant who employs a format that provides an equivalent 300pp. 8704090024. 40462:114. level of completeness and detail may use their own format.

The Palo Verde, Urut 3 Technical Specifications were pre.

pared by the U.S. Nuclear Regulatory Comm;ssion to set forth NUREG/CP-0054: PROCEEDINGS OF THE WORKSHOP ON the hmits, operating conditions, and other requirements applica- SOIL-STRUCTURE INTERACTION. GRAVES.H.L.; PHILIPPA-ble to a nuclear reactor facility as set forth in Secton 50.36 of COPOULO Brookhaven Nabonal Laboratory. December 1986.

10 CFR Part 50 for the protection of the health and safety of 423pp.8703090227. BNL-NUREG-52011. 39932:006.

the pubk The Workshop on Soil-Structure Interaction provided an ex-change of informat on between regulators, pracht oners and re-NUREG-1250: REPORT ON THE ACCIDENT AT THE CHERNO- searchers for the purpose of examining SSI licensing enteria in BYL NUCLEAR POWER STATION.

NRC . No Detailed Affili- the light of recent analytical and expenmental development.

ation Given.

Energy, Dept. of. *; et al. Environmental Protec-ton Agency. January 1987. 214pp. 8702170005. 39657:142. These proceedings contain the papers presented by panelists This report presents the compilation of information obtained and summanes of the sessions along with recommendat.ons of the panel members for each session. Technical areas covered by various organizations regarding the accident (and the conse-quences of the accident) that occurred at Unit 4 of the nuclear by the panels were (1) definition of free field motion, (2) ground power station at Chernobyl in the USSR on April 26,1986. Each moton input needed for site specific SSI analysis. (3) SSI meth-odology, and (4) expenence and experimental observation. The organization has independently accepted responsibdity for one or more chapters. The vanous authors are identified in a foot- summanes were denved to identify areas in the licensing critena note to each chapter. Chapter 1 provides an overview of the which could be changed to improve the licensing process.

report. Very bnefly the other chapters cover

Chapter 2, the NUREG/CP-0082 V01: PROCEEDINGS OF THE FOURTEENTH design of the Chernobyt nuclear station Unit 4; Chapter 3, WATER REACTOR SAFETY INFORMATION MEETING.

safety analyses for Unit 4; Chapter 4, the accident scenano; WEISS,A.J. Brookhaven National Laboratory.

Office of Nucle-Chapter 5, the role of the operator; Chapter 6, an assessment ar Regulatory Research, Director (Post 860720). February 1987.

of the radioactive release, dispersion, and transport; Chapter 7, 521pp. 8702270064. 39764:109.

the activities associated with emergency actions; and Chapter 8, This six. volume report contains 156 papers out of the 175 infoimation on the hea!th and environmental consequences that were presented at the Fourteenth Water Reactor Safety in-from the accident. These subjects cover the major aspects of formation Meeting held at the National Bureau of Standards, the accident that have the potential to present new information Garthersburg. Maryland, during the week of October 27-31, and lessons for the nuclear industry in general. The task of 1986. The papers are pnnted in the order of their presentaten evaluating the information obtained in these vanous areas and in each sesson and desenbe progress and results of programs the assessment of the potential implications has been left to in nuclear safety research conducted in this country and each organ.zation to pursue according to the relovance of the abroad. Foreign participation in the meeting included thirty four subject to their organizations. Those findings will be issued sep- different papers presented by researchers from Canada, arately by the cognizant organizations. The basic purpose of Czechoslovakia, Finland, Germany, Itaty, Japan, Mexico, Spain, this report is to provide the information upon which such as. Sweden, Switzerland and the United Kingdom. The titles of the sessments can be made. papors and the names of the authors have been updated and NUREG 1260 V01: A REPORT TO CONGRESS ON NUCLEAR meet REGULATORY RESEARCH Project Descriptions For FY87.

Office of Nuclear Regulatory Research, Director (Post 860720).

February 1987. 574pp. 8703260021. 40244.001. NUREG/CP-0082 V02: PROCEEDINGS OF THE FOURTEENTH The report presents project desenptions of NRC research WATER REACTOR SAFETY INFORMATION MEETING.

WEISS,A.J. Brookhaven National Laboratory.

Office of Nucle-projects funded in Fiscal Year 1987. The individual project de-ar Regulatory Research, Director (Post 860720). February 1987, senptions presented in this report are divided into six major 444pp. 8703030039. 39837:163.

groups of related projects. These groups, called issues, are as See NUREG/CP-0082,V01 abstract.

follows: Severo Accident, Risk and Reliability, Thermal Hydraulic Transients. Plant Aging and Life Extensson, Seismec Research, NUREG/CP-0082 V03: PROCEEDINGS OF THE FOURTEENTH and Waste Management. Within each issue, the project desenp. WATER REACTOR SAFETY INFORMATION MEETING.

tions are further divided into subgroups, calied subissues. An WEISS,A.J. Brookhaven National Laboratory.

Office of Nucle-overview is provided pnor to each issue and subissue giving a er Regulatory Research, Director (Post 860720). February 1987.

statement of the problem being addressed and the research ob. 433pp. 8703030810. 39867.011.

)ectives. See NUREG/CP-0082,V01 abstract.

MUREG 1280: STANDARD FORMAT AND CONTENT ACCEPT-NUREG/CP-0082 V04: PROCEEDINGS OF THE FOURTEENTH ANCE CRITERIA FOR THE MATERIAL CONTROL AND AC.

WATER REACTOR SAFETY LNFORMATION MEETING.

COUNTING (MC&A) REFORM AMENDMENT.10 CFR Part 74 WEISS.A.J. Brookhaven National Laboratory.

Office of Nucle-Subpart E. EMEIGH,C.W. Division of Safeguards (Pre 870413).

March 1987.105pp. 8704080097. 40446.284. ar Regulatory Research, Director (Post 860720) February 1987.

528pp. 8702270057. 39760:303.

7 Main Citations and Abstracts r

NUREG 1061, " Instructions for Preparation of Data Entry See NUREG/CP-0082,V01 abstract. Sheets for Licensee Event Reports." For those events occurnng on and after January 1,1984, LERs are being submitted in ac-NUREG/CP-0082 V05: PROCEEDINGS OF THE FOURTEENTH WATER REACTOR SAFETY INFORMATION MEETING. cordance with the revised rule contained in Title 10 Part 50.73 WEISS,A.J. Brookhaven National Laboratory.

Office of Nucle, of the Code of Federal Regulations (10 CFR 50.73 Licensee at Regulatory Research, Director (Post 860720). February 1987. Event Report System) which was published in the Federal Reg-594pp. 8702270056. 39765 270. ister (Vol. 48, No.144) on July 26,1983. NUREG-1022, "U-See NUREG/CP-0082,V01 abstract. censee Event Report System Descnpton of Systems and Guidehnes for Reporting," provides supporting guidance and in-NUREG/CP-0082 V06: PROCEEDINGS OF THE FOURTEENTH formation on the revised LER rule. The LER summanes in this WATER REACTOR SAFETY INFORMATION MEETING. report are arranged a!phabetically by facility name and then WEISS.A.J. Brookhaven National Laboratory.

Office of Nucle-chronologically by event date for each facety. Component, at Regulatory Research. Director (Post 860720). February 1987-424pp. 8703030035. 39836.099.

system, keyword, and component vendor indexes follow the summaries. Vendors are those identified by the utikty when the See NUREG/CP 0082,V01 abstract.

LER form is initiated; the keywords for the component, system, NUREG/CP 0084: PROCEEDlNGS OF THE WORKSHOP ON and A general keyword indexes are assigned by the computer CONTAINMENT PERFORMANCE DESIGN OBJECTIVE.MAYusing correlation tables from the Sequence Coding and Search 1213,1986, HARPERS FERRY, WEST VIRGINIA.

Brookhaven System.

Nat onal Laboratory. November 1986. 77pp. 8704080102. BNL-NUREG-52044. 40447.026. NUREG/CR-2000 V06 N1: LICENSEE EVENT REPORT (LER)

The " Containment Performance Design Oblectue Workshop" COMPILATION For Month Of January 1987.

Oak Ridge Na-was designed to obtain a broad range of knowledgeable views tional Laboratory. March 1987. It7pp. 8703250557. ORNL/

concerning the issues in the development and emplementation NSIC-200. 40234.330.

of a containment performance design objectue (CPDO). It was See NUREG/CR 2000,V05,N12 abstract.

a discussion workshop, involving invited experts representing a broad range of viewpoints, and drawn from utilities, reactor ven- NUREG/CR 2331 V06 N2: SAFETY RESEARCH PROGRAMS dors, architect engineers, unwersities, national laboratones, and SPONSORED BY OFFICE OF NUCLEAR REGULATORY public interest groups. The participants were requested to RESEARCH Ouarterly Progress Report Apnt-June 1986.

review background information conceming the safety goals and WEISS,A.J. Brookhaven National Laboratory. November 1986.

their status, a descnption of CPDO options selected for evalua- 107pp.8704080236. BNL NUREG-51454. 40445.157.

tion, an outline of an implementation approach and recognized This progress report will describe current actmties and techni-issues of CPDO structure and implementaten. The general ob- cal progress in the programs at Brockhaven National Laboratory jectue of the workshop was to generate information that could sponsored by the Dmston of Accident Evaluation, Division of be used in the NRC's study and decision process concerning Engineenng Technology, and Division of Risk Analysis & Oper-the formulation of a containment performance design objectwe. ations of the U.S. Nuclear Regulatory Commission Office of Nu-The participants' views were obtained on specific CPDO options clear Regulatory Research. The projects reported are the foi-and issues that were presented to the Workshop and new ones gow,ng High Temperature Reactor Research, SSC Code im-that emerged dunng the discussion. An attempt was also made to identify e eas of consensus emerging t'om the discusson-provements, Thermal Hydraulic Reactor Safety Expenments, Thermodynamic Core-Concrete Interacton Expenments and NUREG/CP 0085: MEETING WITH STATES ON THE LOW- Analysis, Plant Analyzer, Code Assess, rant and Apphcation, LEVEL RADIOACTIVE WASTE POLICY AMENDMENTS ACT Code Maintenance (RAMONA 30), MELCOR Venficaton and (LLRWPAA) OF 1985. MAUPIN.C.; SCHNEIDER.K. Assistant De- Benchmarking, Source Term Code Package Venfication and rector for State Agreements Programs. February 1987.168pp. Benchmarking, Uncertainty Analysis of the Source Term; Stress 8703260292. 40238.111. Corrosion Cracking of PWR Steam Generator Tubing, Sod-The purpose of this meeting was to discuss with selected Structure interaction Evaluation and Structural Benchmarks, State officia s NRC responsibihties under the Low Level Rado- identificaton of Age Related Failure Modes; Application of actwe Waste Policy Amendments Act, including the approach HRA/PRA Results to Support Resolution of Genenc Safety being taken and progress being made in fulfilling NRC responsi- issues involving Human Per1ormance, Protectue Action Deci-bihties. The NRC staff oblectue was to obtain State views on sionmaking, Rebasehning of Risk for Zion, Containment Per-technical and institutional issues associated with NRC and State formance Design Objectwo, and Operational Safety Refiability implementation of the Act and to determine any additonal areas Research.

in which NRC can be of assistance in the development of dis-posal facihties. We beheve this objectwo was accomphshed. NUREG/CR 2478 V03: A STUDY OF TRENCH COVERS TO MIN-The transcnpt of the meeting is being pubbshed at this time to IMl2E INFILTRATION AT WASTE DISPOSAL SITES Final Rept.

make available nformaton discussed at the meeting to those CARTWRIGHT,K.; LARSON T.H.; HERZOG,B L.; et al. Illinois, 6ndmduals and groups that have responsibilities under the State of. February 1987,139pp. 8703120215. 39987;171.

LLRWPAA for developing disposal capacity and for regulating We have investigated methods to limit inNtration through Iow level waste disposal sites. trench covers by reviewing current practices, testing selected NUREG/CR 2000 V05N12: LICENSEE EVENT REPORT (LER) geologic matenals, simulating selected cover designs, and de-COMPILATION For Month Of December 1986.

Oak Ridge Na. signing, constructing, and monitonng four field-Scale expenmen.

tional Laboratory. January 1987.133pp. 8702170056. ORNL/ tal covers. Of the many designs considered, we conclude that NSIC-200. 39657.009. multilayered soil covers are supenor to single layered covers.

This monthly report contains Licensee Event Report (LER) Laboratory and computer simulations indicate that a wick effect operational information that was processed into the LER data can be estau,,6ed by placing a fine grained layer over a file of the Nuclear Safety Information Center (NSIC) dunng the coarse-grained layer, thereby delaying and possibly reducing one month penod identified on the cover of the document. The moisture infiltraton through the entire cover. Field expenments LERs, from which this information is derwed, are submittad to indicate that the wick effect does occur under certain circum-the Nuclear Regulatory Commission (NRC) by nuclear power stances, but a potentially more important feature of the layered plant licensees in aCCordance with federal regulations. Proce- cover design is the ability of the coarse-grained layer to remove dures for LER reporting for revisions to those events occurnng moisture from the system through drain tales prior to 1984 are desenbed in NRC Regulatory Guide 1.16 and

8 Maln Citations and Abstracts NUREG/CR-3232: DETAILED STUDIES OF SELECTED,WELL cate that the addition of steam reduces the normalized peak EXPOSED FRACTURE ZONES IN THE ADIRONDACK MOUN- combustion pressure (Pmax/Po) as compared to equivalent TAINS DOME,NEW YORK, WlENER,R.W.; ISACHSEN,Y.W. hydrogen. air burns. Turbulence was found to affect the extent New York, State Unrv. of, Albany, NY. January 1987. 94pp. of combustion and other combustion charactenstics of the lean 8702260644. 39759.123. hydrogen burns (te., less than or equal to 10% hydrogen by The Adirondack Mountains constitute a relatively young (Mes- volume) where buoyancy govems flame propagation. The ex-ozoic? Cenozoic 7) dome on the craton. The dome is undergoing penmentally measured pressure decays were used to infer the contemporary uplift, based on geodetic relevehng, and is seismi- " global" heat transfer charactenstics dunng the postcombustion cally actrve. The breached dome provides a very large window cooling phase. Convecton was found to dominate the time-inte-through Paleozoic cover and thus permits ground study of the grated heat transfer of the leaner (less than or equal to 10%)

fracture systems that characterize the seismogense basement hydrogen: air bums, accounting for 50 to 70% of the postcom-and influence the patterns of bnttle deformation that are found bustion heat transfer. Radiation was slightly more prevalent in overtying Paleozoic rocks of the platform The predominant than convection for the hydrogen. air burns near stoichiometry.

fracture zones are linear valleys that trend NNE to NE, parallel When moderate quantities of steam were addod to the environ-to the long axis of the dome. The 36 field studies of the linea-ment, radiation became the dominant postcombustion coohng ment segments discussed in this report suggest that the proms- mechanism due to the increase in bulk gas emittance. If richer nent NE to NNE fracture systems in the eastern Adirondacks steam concentrations were added to the environment, radiaton are dominantly high angle faults down. stepped to the east, and Convection appear to be equally important heat transfer whereas those in the central Adirondacks are dominantly zero- mechanisms.

displacement crackle zones The ongin of these features is re-lated to the rapid uplift of the Ad.rondack dome. Similar features NUREG/CR-3469 V03: OCCUPATIONAL DOSE REDUCTION AT can be expected to be found in other areas of domal uplift or NUCLEAR POWER PLANTS. Annotated Bibhography Of Select-rapid regional uphft.

ed Readings in Radiation Protection And ALARA. BAUM.J.W.;

NUREG/CR-3412 V02: CONTAINMENT INTEGRITY KHAN,T.A. Brookhaven National Laboratory. November 1986.

PROGRAM. Progress Report. April 1983 -December 1984. 124pp.8703100021. BNL-NUREG-51708. 39933,248.

DLEJWAS T.E.; HORSCHEL D S. Sandia Natonal Laboratones. This report is the third in a senes of bibhographies supporting December 1988. 58pp. 8702060139 SAND 83-1482. 39527.249. the efforts at Brookhaven National Laboratory on dose reduc.

This report contains a desenption of work performed between tion at nuclear power plants. Abstracts for this report were se-Apnl 1983 and December 1984 under the Containment Integqty lected from papers presented at recent technical meetings, jour.

Program. The program is one of three at Sandia National Lab. nals and research reports reviewed at the BNL ALARA Center, oratones in the general area of containment integnty. The over- and searches of the DOE / RECON data base on energy related all ob}ective of the three programs is the qualification of meth. publications. The references selected for inclusion in the bibli-ods for rehably predicting the capability of conta'r, ment struc. ography relate not only to operational health physics topics but tures for light water reactor nuclear power plants to function also to plant chemistry, stress corrosion cracking, and other as-under loadings caused by severe accidents and extreme envi- pects of plant operation which have important impacts on occu-ronments. In the subject program, models of entire contain- pational exposure. Also included are references to improved ments are tested for loadings beyond the design basb. Expen- design, planning, matenals selection and other topics related to ments completed dunng the reporting penod include a series of what might be called ALARA engineering. Thus, an attempt has internal-pressure tests of 1/32 and 1/8 size models of hybnd been made to cover a broad spectrum of topics related directly steel containments. Compansons with pretest analyses are de. or indirectly to occupational exposure reduction. This report scnbed. contains 252 abstracts and both author and subject indices.

NUREG/CR 3444 V04: THE IMPACT OF LWR DECONTAMINA.

TiONS ON SOLIDIFICATION WASTE DISPOSAL,AND ASSOCl- NUREG/CR 3620 S02: INTRUDER DOSE PATHWAY ANALYSIS ATED OCCUPATIONAL EXPOSURE. Annual Report, Fiscal Year FOR THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The 1986. PICIULO,P.L.; ADAMS,J W. Brookhaven National Labora- ONSITE/ MAX 11 Computer Program. KENNEDY,W E.;

tory. October 1986. 48pp. 8704130283. BNL NUREG-51699. PELOQUIN.R.A.; NAPIER,0.A.; et al. Battelle Memorial Institute, Pacific Northwest Laboratones. February 1987- 453pp.

Les h tests wero initiated in order to determine if organic rea-gents released from different siz9 waste forms can be repre- The document entitled ' Intruder Dose Pathway Analysis of sented by a diffusion controlled mechanism. Data for the re- N Onsde % sal d Qamam Wam W mm MAX 11 Computer Program' (1984) by Napier et al. summarizes lease of EDTA from cement forms showed that the CFR from the 5 cm diameter 10 cm long forms were smaller than that our initial off ris to develop human-intruson scenanos and a from the 15 cm diameter x 15 cm diameter forms. This would m dified version of the MAXI computer program for potential not be expected if the predominant mechanism of release was diffusion. Specimens containing Co-60 spiked caton exchange waste disposal. Supplement 1 of NUREG/CR 3620 (1986) by resins solidified in cement were leached with deionized water Kennedy et al. Summarited modifications and improvements to and leachants containing either formate or picolinate. Data from the ONSITE/ MAX 11 software package. This document summa-the fest 60 days of leaching indicate that the releases of Co-60 ms a MM we d N MMM We W-were similar with desonized water and formate as leachants Pic- gram. This modified version of the computer program operates ohnic acid present in the leachant caused an acceleration of the on a personal compuW aM ped me user M opMnah release of CM select radiation dose conversson factors published by the Inter-natonal Commission on Radiological Protecton OCRP) in their NUREG/CR 3468: HYDROGEN AIR STEAM FLAMMA0lLITY Publication No. 30 (ICAP 1979 1982) in place of those pub-LIMITS AND COMBUSTION CHARACTERISTICS IN THE FITS lished by the ICAP in their Publicat on No. 2 (ICRP 1959) (as VESSEL. MARSHALL,0 W. Sandia National Laboratones De- implemented in the prevous verssons of the ONSITE/MAXII cember 1986.149pp. 8704080363 SAND 84-0383. 40445 008. computer program). The pathway to human models used in the Expenmentally observed flammatality hmits of computer program have not been changed from those de-hydrogen air. steam mixtures in both turbulent and quiescent en- scnbed previously (Napier et al.1984, Kennedy et al 1986)

Wonments were measured and a correlation developed that de- Computer hstings of the ONSITE/MAXII computer program and scribes the three-component flammabikty hmit. The combustion supporting data bases are included en the appendices of this pressure data measured for the hydrogen air steam tests 6ndi- document

Mzin Citations and Abstracts 9 system operatons. Factors affectng replacement energy costs, NUREG/CR 3461: STRESSCORROSION CRACKING OF LOW- such as random unit failures, maintenance and refueling require-STRENGTH CARBON STEELS IN CANDIDATE HIGH LEVEL ments, and load variatons, are treated in the anatysis. Seasonal WASTE REPOSITORY ENVIRONMENTS. BEAVERS,JA: costs are presented for the five-year penod beginneng with 1987 THOMPSON,N.G ; PARKINS,R.N. Battelle Memonal institute, and ending with 1991. This informaton updates cost eshmates Columbus Laboratones. February 1987. 88pp. 8703120328. that were developed previously for the NRC and pubhshed in BMI 2147, 39981:234.

NRC Report NUREG/CR-4012, Vol.1. The updates were under-A survey of ttse hterature was performed to identify potential stress-corrosion cracking agents for low-strength carbon and taken to extend the time frame of cost estmates and to ac.

count for recent changes 6n utahty system conditions, such as Iow alloy steels in repository envronments. It was found that a fluctuations in fuel pnces, changes in construction and retire-number of potent cracking agents are present, but stress-corro- ment schedules, and adjustments to system demand projec-son cracking is relatively unhkely in the bulk repository envvork ment because of their ',ow concentraton. On the other hand, tions.

concentration of these speceos may occur by a number of mechanisms, and thus it is Concetvable that the waste package NUREG/CR-4016 V02: APPLICATION OF SLIM MAUD:A TEST OF AN INTERAChvE COMPUTER BASED METHOD FOR OR-could fail prematurely by stress corrosion. Accordingfy,it is rec- GANIZING EXPERT ASSESSMENT OF HUMAN PERFORM.

ommended that the lower concentration hmits for potential ANCE AND RELIABILITY. Volume II. Appendices.

cracking agents be identified under typical repository envworp SPETTELL.C.M.; ROSA,E.A.; HUMPHREYS,P.C.; et al. Brookha-ments, in conjunction with modeling studies to assess the filieH. von National Laboratory. October 1988. 219pp. 8702060128.

hood that the concentrating mechanisms witi operate and to BNL NUREG-51828. 39537:248.

bound the upper limits of concentration for each mechanism. The U S. Nuclear Regulatory Commisson (NRC) has been c ng a ear resea Nam t sbgate N NUREG/CR 3950 V03: FUEL PERFORMANCE ANNUAL ent methods for using expert judgments to eshmate human REPORT Pacific Northwest FOR 1985'WU,S.

Laboratories BAILEY'.W.J.

NRC . No DetailedBattelle Affili- Memonal Institute error probabilibes (HEPs)/ in nuclear power plants. One of the abon Given. February 1987,103pp. 8703120248. PNL 5210. methods investigated, dertved from multi attnbute ublity theory, 6s the Success Likelihood Index Methodology implemented 39985.335.

This annual report, the eighth in a series, provides a brief de- through Multi-Attnbute Utility Decomposition (SLIM MAUD). This scnption of fuel perforrmnce dunng 1985 in commercial nuclear report desenbes a systemate test application of the SLIM.

power plants. Onef summanes of fuel design changes, fuel sur. MAUD methodology. The test application is evaluated on the veillance programs, fuel operahng expermnce and trends, fuel basis of three entena: practicality, acceptability, and usefulness.

problems, high-burnup fuel experience, and 6tems of general sig- Volume I of this report presents an overview of SLIM MAUD, ruficance are provided. References to additional, rnore detailed desenbes the procedures followed in the test applicabon, and information and related NRC evaluabons are included. provides a summary of the results obtained. Volume !! consists of technical appendices to support in detail the materials cork NUREG/CR 3968: STUDY OF OPERATING PROCEDURES IN tained in Volume I, and the users' package of explicit proco-NUCLEAR POWER PLANTS. Practices And Problems. dures to be followed in implementing SLIM-MAUD. The results MORGENSTERN,M.; OARNES,V.E.; MCGUIRE,M.V ; et al. Ott- obtained in the test apphcation provide support for the applica-telle Human Affairs Research Centers. February 1987.158pp. tion of SLIM-MAUD to a wide variety of apphcabons requinng 8703090132. PNL 5648. 39923.340. estimates of human errors.

This report desenbos the project activities, findings, and rec-ommendations of a project enttled " Program Plan for Assess- NUREG/CR 4300 V03 N2: ACOUSTIC EMISSION / FLAW RELA.

ing and Upgrading Operating Procedures for Nuclear Power TIONSHIP FOR INSERVICE MONITORING OF NUCLEAR Plants? The project was performed by the Pacific Northwest PRESSURE VESSELS. Progress Rept,Apni-September 1986.

Laboratory and Batteile Human Affairs Research Centers for the HUTTON,P.H. Battelle Memortal Institute, Pacific Northwest Division of Human Facturs Technology, Office of Nuclear Reac. Laboratones. January 1987,16pp. 8702260636. PNL 5511.

tor Regulabon, U.S. Nuclear Regulatory Commission (NRC). The 39759 217.

project tearn analyzed and evaluated sarrples of normal and This report discusses technical progress for the period April abnormal operabng procedures from 31 commercial nuclear 1988. September 1988 for the NRC sponsored research pro-power plants operating in the United States. The protect team Gram concerned with "Acousbc Emission / Flaw Relationships for also visited nine nuclear power plants in the United States to inservice Monitonng of Nuclear Reactor Pressure Boundaries."

obtain 6nformation on the development, use, and control of op, included in the discussion are the topics of AE monitonng of pet-eratng procedures. A peer review group was convened to mary piping dunng reactor operation, substantiation of the AE advise the project team on the conduct of the project and to 8'gnal identification method, development of AE/lGSCC rela-review and comment on the project report. The report contains tonships, and progress in establishing an ASTM AE standard findings on the useability of operahng procedures and on prac. and an ASME appendu for orkhne AE monitoring taces concerning the development, use, and control of operating procedures in nuclear power plants. The report includes recom- NUREG/CR-4301: STATUS REPORT ON EQUIPMENT OUALIFI-mendations to the NRC on the need to upgrade the quality of CATION ISSUES RESEARCH AND RESOLUTION.

operat ng procedures. The report also discusses an approach t 00NZON,LL; WYANT,F.J.; DUSTARD LD.; et al. Sandia Na-a program plan to assess and upgrade operating procedures. tional Laboratones. January 1987, 582pp. 8703090128.

SAND 851309,39922:118.

NUREG/CR 4012 V02: REPLACEMENT ENERGY COSTS FOR Since its inception in 1975, the Qualification Teshng Evalua-NUCLEAR ELECTRICITY GENERATING UNITS IN THE tion (OTE) Program has produced numerous results pertinent to UNITED STATES.19871991. VANKUlKEN J C.; GUZIEL,K.A :

equipment quahficaton issues. Many have been incorporated DUEHRING W.A.; et al. Argonne National Laboratory. January into Regulatory Guides, Rules, and industry practices and stand-1987,253pp. 8702190449. ANL AA 30. 39690 235. ards. This report summarties the numerous reports and findings Seasonal replacement energy costs are estimated for poten- to date. Thirty separate lasues are discussed encompassing taal shor1. term shutdowns of 116 nuclear electncity.generahng three Generic areas: accident sirnulation methods; aging simuta-units. These estimates were developed to help the U.S. Nuclear tion methods; and, special top 6cs related to equipment qualifica-Regulatory Commission (NRC) estabhsh regulatory pohcies, par- ton. Each 6ssue-specific section contains: (1) a brief desenption ticularty those requiring safety modificabons that might necessi- of the issue; (2) a summary of the applicable research effort; tate temporary reactor shutdowns. Cost estimates were denved and (3) a summary of the findings to date.

from probabilistc production. cost simulatons of pooled ut hty-

10 Main Citations and Abstracts NUREG/CR-4320: THE RELATIONSHIP AND INFLUENCES OF This is one of six case studies for USI A 45 Decay Heat Re-FUEL AND COOLANT SYSTEM PROCESSES DURING LWR moval (DHR) Requirements. The purpose of this study is to SEVERE ACCIDENTS. RIVARD,J B. Sandia Natonal Laborato- identify any potential vulnerabilities in the DHR systems of a nes. December 1988. 59pp. 8704080124. SAND 851449. typical Westinghouse 2 ioop PWR, to suggest possible modifica.

40446:182. tions to improve the DHR capab hty, and to assess the value &

This report places the processes of fuel and core damage, re- impact of the most promising altematives to the existing DHR actor coolant system (RCS) flow and heat transfer, and pres- systems. The systems analysis considered small LOCAs and sure vessel breach and their respectrve fission product consid- transient internal initiating events, and seismic, fire, extreme orations . in the context of typical nsk dominant accident se- wind, internal and external flood, and lightning external events.

quences. Thus the enhanced perception of the relationship (and A full-scale systems analysis was performed with detailed fault thus irnportance) of these processes to the potential conse- trees and event trees including support system dependencies.

quences of a severe accident is prov6ded. This is accomphshed The system analysis results were extrapolated into release cate-using both genenc and plant-specific relational methods. It is gones using applicable past PRA pnenomenological results and found that the vessel and RCS processes pervastvely influence improved containment failure mode probabilities. Pubhc conse-consequences. The expertrrent programs designed to provide quences were estimated using site specific CRAC2 calculatons.

the in-vessel and RCS fuel damage data base are examined in The Value-impact (VI) analysis of possible alternatives consid-light of this concluson, and some suggestons are offered. ered both onsite and offsite impacts arriving at several nsk NUREO/CR-4409 V02: DATA BASE ON NUCLEAR POWER measures such as averted population dose out to a 50-mile PLANT DOSE REDUCTION RESEARCH PROJECTS. radius and dollars per person rom averted. Uncertainties 6n the KHAN,T.A.; BAUM.J W. Brookhaven National Laboratory. No. VI analysis are discussed and the issues of feed and bleed and vember 1966. 232pp. 8704080217. DNL.NUREG-51934. secondary blowdown are analyzed.

40444:134-This report desenbes 142 internatonal protects on dose re- NUREO/CR-4469 V04: NONDESTRUCTIVE EXAMINATION ducton research. It is the second report on a data base main' (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report. October 1985 March tained by Brookhaven National Laboratory as part of an NRC 1986. DOCTOR,S R; BATES.D.J.; DEFFENBAUGH,1; et at sponsored project on occupaSonal dose reduction. The first Battelle Memonal institute, Pacific Northwest Laboratones.

report desenbed 180 similar protects. A wide area of research is March 1987. 91pp. 8704060414. PNL 5711,40410:169.

covered, including plant chemistry, stress corrosion cracking, The Evaluation and improvement of NDE Rehabihty for in-steam generator repair and replacement, robotics, and decon-tamination. Analysis indicates that dose reduction research is service inspection of Light Water Reactors (NDE Rehabikty)

Program at the Pacific Northwest Laboratory was established by beginning to affect occupatonal radiaton exposure. There is a general diminution in exposures in countnes with dose reduction the Nuclear Regulatory Commission to determine the rehabikty of current inservice inspection (ISI) techniques and to develop research progtsms, such as Japan, The Federal Repubhc of Germany, Canada, Sweden, France, and the United States- recommendations that will ensure a suitably high inspection reli-Most of the present resear.h, however, is directed towards en- abhty. The objectrves of this program include determining the gineenng approaches to dose reduction. More attention in the rehabihty of ISI performed on the primary systems of commer-cial hght. water reactors (LWRs); using probab hatic fracture me-non-engineenng areas is called for.

chanics analysis to determine the irnpact of NDE unreliability on NUREG/CR-4448: SHUTDOWN DECAY HEAT REMOVAL ANAL. system safety; and evaluating reliabihty improvements that can YSIS OF A GENERAL ELECTRIC DWR3/ MARK ICase Study, be achieved with improved and advanced technology. A final HATCH.S W.; ERICSON.D M.; SANDERS,G A. Sandia National objective is to formulate recommended revisions to ASME Code Laboratones. March 1987. 739pp. 8704060465. SAND 85-2373. and Regulatory requirements, based on material properties, 40405 123. service conditions, and NDE uncertainties. The program scope A General Electnc Doihng Water Reactor (8WR3) with a Mark is hmited to ISI of the pnmary systems including the piping, I containment has been evaluated as part of Task Action Plan vessel, and other inspected components. This is a progress A-45, " Decay Heat Removal Requirements." Probabikstic nsk report covenng the programmatic work from October 1985 assessment models were constructed to determine the dome, through March 1986.

nant 6nternal, randomly initiated accident sequences and special emergency sequences (e g , earthquakes). The dominant se- NUREG/CR 4491: DEVELOPMENT OF MODELS FOR WARM quences were reviewed to determine what modshcations might PRESTRESSING. STONESIFER R B. Computational Mechanics.

be made to enhance the plant's abihty to remove decay heat. RYDICKI,E F. PROSIG, Inc.

Materials Engineenng Associates, Modifications which held promise went through a prehminary Inc. January 1987, 78pp. 8702060276. MEA 2122. 39538.107.

i cost and design analysis. Additionally, the impact on the proba- The objective of this project is to evaluate available mathe-bakty of core melt accidents was estimated given implementa' matical models and associated fracture criteria for predicting tion of modifications. In the final step, these results were com- warm prestress (WPS) effects. A venfied model of the WPS bened en a value4mpact format according to NRC guidehnes. phenomenon is required before credit for improved low temper.

The results 6ndicate that feasible modifications to enhance ature toughness can be taken in analysis of postulated accident decay heat removal do exist at the subject plant. The central scenanos such as pressurtzed thermal shock. The pnmary basis estimates of the value impact results tended, however, to show of evaluation is finite element analysis using a highly refined marginal cost effectiveness under current guidelines for most of mesh and work hardening, modeled by a piece-wise knear fit of the modihcations. Afternate assumptions 6nvolving source term stress-strain data. The entena being evaluated are J(e), (Chell, magnitude and enterdiction entena were found to significantly et al), entical stress (Curry). T*(p) (Atluri) and a cntenon 6ntro-affect the results. The insights Gained from this study will duced herein which is related to differential CTOD and denoted become part of an information base which will be used to devel- dCTOD* FLOW. The finite element model is used to simulate a op genenc recommendations regarding the adequacy of decay load-unload. cool-fracture (LUCF) type of WPS cycle for which heat removal systems in hght water reactors. experimental results are available. The vanous models and cri-tena are evaluated in terms of their agreement with the finite.

NUREG/CR-4458: SHUTDOWN DECAY HEAT REMOVAL ANAL- element results such as crack opening displacements, stresses YSIS OF A WESTINGHOUSE 2 LOOP PRESSURIZED WATER and plastic zone sizes, and in terms of their abilty to predict REACTOR Case Study. CHAMOND,W R ; ERICSON,0 M,; fracture load The nonfinite element based models of Chell and SANDERS,0A. Sandia National Laboratones March 1987. Curry are used to simulate 32 additional WPS espenments so 900pp. 8704060155. SAND 66 2496 40305 017. as to further assess the relative ments of the models and the

Main Citations and Abstracts 11 loading cycles, from 0 to 14 MPa (2000 psi). Grout is intected J(e) cntical s'ress, and dCTOD* FLOW fracture entena. While through an axial borehole, at a pressure of 1.2 to 8.3 MPa (180 K(Ic) scatter band behavior allows significant fatitude for manip- to 1200 psi), pressure selected to provide a likely groutable ulation of model predictions, which impedes entical evaluation fracture aperture, while the fracture is stressed at a constant of the models and critena, both models and all three fracture normal stress. The fracture permeability is measured after grout-cntena are found to predict WPS behavior which is quahtatively ing. Flow tests on the ungrouted samples confirm the inverse consistent with experimental data. reaton between normal stress and fracture permeabikty. ,The NUREG/CR-4524: CLOSEOUT OF IE BULLETIN 80 24 PREVEN. equivalent aperture determined by these tests is a reliable indi-TION OF DAMAGE DUE TO WATER LEAKAGE INSIDE CON- cator of groutabLty. Post-grouting permeabihty measurements TAINMENT (OCTOBER 17,1980 INDIAN POINT 2 EVENT). as performed here, and frequently 6n practice, can be mislead-FOLEY,Wl; DEAN.R S.; HENNICK,A. Parameter, Inc. March ing, since incompletc grouting of fractures can result in maior 1987, $1pp. 8704090025. IEB.80-24. 40464 058. apparent reductions in permeability. The apparent permeability On October 24,1980, IE Information Notice 80-37 was issued reducton is caused by grouting of a small area of a highly pref-by the NRC to desenbe reactor vessel pit flooding which had emntial flowpath directty adjacent to the hole used for grouting been discovered a week earlier at Indian Point 2. The lower and for permeability testing. Expenmental results confirm claims nine feet of the reactor vessel had been wetted while at operat- in the literature that ordinary portland cement inadequately pen-ing temperature, and a thermal stress condit>on of potential etraws hne fractures.

safety significance had been caused. IE Bulletin 80 24 was issued by the NRC on November 21,1980 because of concern NUREG/CR-4550 V03: ANALYSIS OF CORE DAMAGE FRE- 1, about this event. Licensees of operating power reactors were OUENCY FROM INTERNAL EVENTS.SURRY UNIT required to take short term actons to ensure continued intarim BERTUCIO.R.C.; OUILICI,M.D.; YOUNG.J.; et al. Sandia Nation-al Laboratones. November 1986. 450pp. 8704 t30170. SAND 86-operation without containment flooding The long term purpose of the bulletin was to obtain operating data on which to base 2084,40512:157.

future NRC requirements for genenc corrective actions. The bul- This document contains the accident sequence for Surry, Unit letin was issued to holders of cons.ruction permits for informa- 1: one of the reference plants being examined as part of the tion. Inspection requirements for reviewing licensee actions NUREG 1150 effort by the Nuclear Regulatory Commission were clanfeed by issuing Temporary Instruction 2515/47 on De* (NRC). NUREG.1150 will document the risk of a selected group cember 18,1980, and a special memorandum on February 19 of nuclear power plants. As part of that work, this report con-1981. Evaluatton of utahty responses, hcensee event reports, an tains the overall core damage frequency estimate for Surry, Unit NRC/IE memorandum, NRC/IE inspection reports and an NRC/ 1, and the accompanying pla%t damage state frequencies. Sen-IE letter shows that the bulletin can be closed per specific cnte- sitivity and uncertainty analyses provide additional insights re-na for all of the 69 facilities to which it was issued for action

garding the dominant contnbutors to the Surry core damage fre-and that no further act>on is necessary. quency estimate, The mean core damage frequency at Surry was calculated to be 2.6E.5 per year. Station blackout type ac-NUREG/CR 4531: AN INVESTIGATION OF INTEGRAL FACILITY cidents (loss of all AC power) were the largest contnbutors to SCALING AND DATA RELATION METHOOS (INTEGRAL core damage frequency, accounting for approximately 38% of SYSTEM TEST PROGRAM). LARSON.T.K. EG&G Idaho, Inc. the total. The next type of dominant contnbutors were transient (subs. of EG&G, Inct February 1987.129pp. 8704080137. induced LOCAs caused by loss of electncal bus instators.

EGG 2440. 40440.178, These sequences account for 19% of core damage frequency.

The Integral Systems Test Program was initiated in 1982 by No other type of sequence accounts for more than 10% of core government and industry in response to the Three Mile Island damage frequency. The numencal results are dnven to some accident. Three different integral test facihties, each scaled to a degree by modehng assumptions and data selection for issues Babcock and Wilcox demgn nuclear steam supply system, will such as reactor coolant pump seal LOCAs, common cause f ail-ultimately contnbute data to tne program. Each of the facilities ute probabilities, and plant response to station blackout and was designed using different scahng methodologies, and each loss of electrical bus initiators. The sensitivity studies explore has different operating capabihties. The overalt scaling of each the impact of alternate theones and data on these issues. The facility is examined in this report, and local scahng is analy;e; results of the uncertainty and senstrvity analyses should be to demonstrate potential mmilanties and dissimilanties in facshty considered before any future actions are taken based on this response relative to expected plant responses. The scaling rela. analysis.

tionships are used to show how local thermal-hydraulic phenom-ena in each facihty can be compared to each other of to ex- NUREG/CR 4550 V04, ANALYSIS OF CORE DAMAGE FRE-pected piant behavior. The concept of an equibbnum plot is OUENCY FROM INTERNAL EVENTS PEACH BOTTOM UNIT used to show how the global response of each facility can be 2. KOLACZKOWSKI.A.; LAM 8PtGHT,J.A.; FERRELL,W.L; et al.

related for a specific small break loss-of-coolant transient. Po- Sandia National Laboratones. October 1986. 663pp.

tential complications that may anse as a consequence of the fa- 8703090175. SAND 86-2084. 39919.315.

cility scaling or facihty hmitations are enumerated. The potental This document contains the intomal event initiated accident use of dimensionless groupings for relating and specifying nu. sequence analyses for Peach Bottom, Unit 2; one of the refer-penments is discussed. Finalty, some specific expenments And ence plants being examined as part of the NUREG 1150 effort conditions are proposed for the purpose of simpkfying inte%ci- by the Nuclear Regulatory Commission. NUREG 1150 will docu-lity companson of test results' ment the nsk of a selected group of nuclear power plants. As part of that work, this report contains the overaft core damage NUREG/CR-4541: EXPERIMENTAL ASSESSMENT OF THE frequency estimate for Peach Bottom, Unit 2, and the accompa-SEALING EFFECTIVENESS OF ROCK FRACTURE GROUT- nying plant damage state frequencies. Senstivity and uncertain.

ING. SCHAFFER,A.; DAEMEN.JJ Anzona, Univ. of, Tucson, ty analyses provide additional 6nsights regarding the dominant AZ. March 1987,193pp. 87040f 0149 40340.050. contnbutors to the Peach Bottom core damage frequency esti-The objective of this investigation is to determine the effec- mate. The mean core damage frequency at Peach Bottom was tiveness of cement grouts as sealants of fractures en rock. La5- calculated to be 8 2E-6. Station Blackout type accidents (loss of oratory expenments have been conducted on seven 15.cm all AC r ower) were found to dominate the overall results. Antici.

granite cubes containing saw cuts, three 23 cm diameter ande- pated Transients Without Scram accidents were a!so found to Ste cores containsng induced tension cracks, and one 15-cm di-be non negligible contnbutors. The numencal results are largely ameter marble core contain.ng a natural fracture. Pnor to grout-dnven by common mode failure probabikty estimates and to ing, the hydraubc conductivity of the fractures is determined some extent, human e'ror. Because of significant data and anal-under a range of normal stresses, applied in the loading and un-

12 Main Citations and Abstracts ysis uncertainties in these two areas (important, for instance, to the most dominant scenarc in this study), it is recommended NUREG/CR-4613: EVALUATION OF NUCLEAR POWER PLANT OPERATING PROCEDURES CLASSIFICATIONS AND that the results of the uncertainty and sensitivity analyses be INTERFACES Problems And Techniques For Improvement.

considered before any actons are taken based on this analysis. BARNES V.E.; RADFORD,LR. Battelle Hurnan Affairs Research Centers.

Battelle MemonalInstitute, Pacific Northwest Labora-NUREG/CR-4551 V1 DRF: EVALUATION OF SEVERE ACCl- tories. February 1987. 120pp. 8703120284. PNL 5852.

DENT RISKS AND THE POTENTIAL FOR RISK 39983:256.

REDUCTION:SURRY POWER STATION, UNIT 1. Draft For Com-ment. BENJAMIN,A.S ; BOYD G.J.; KUNSMAN.D.M ; et al. This report presents activities and findings of a project de-Sandia National Laboratories. February 1987. 720pp. signed to evaluate current practices and problems related to 8703170262, SAND 86-1309. 40044.071. procedure classification schemes and procedure interfaces in The Severe Accident Risk Reduct!on Program (SARRP) has commercial nuclear power plants. The phrase " procedure clas-completed a rebaselining of the nsks to the public from a par- sification scheme" refers to how plant operating procedures are ticular pressurized water reactor with a subatmosphenc contain- categonzed and indexed (e g., normal, abnormal, emergency ment (Surry, Unit 1). Emphasis was placed on determining the operating procedures). The term " procedure interface" refers to magnitude and character of the uncertainties, rather than focus- how reactor operators are instructed to transition within and be-ing on a point estimate. The nsk reduction potential of a set of tween procedures. The project consisted of four key tasks, in-proposed safety option backfits was also studied, and their cluding (1) a survey of literature regarding problems associated with procedure classifications and interfaces within and between costs and benefits were also evaluated. It wss found that the risks from internal events are generally lower than previously procedures, as well as techniques for overcoming them; (2) evaluated in the Reactor Safety Study (RSS). However, certain interviews with experts in the nuclear industry to discuss the ap-unresolved issues (such as direct containment heating) caused propriate scope of different classes of operating procedures and the top of the uncertainty band to appear at a level that is com- techniques for managing interfaces between them; (3) a reanal-parable with the RSS point estimate. None of the postulated ysis of data gathered about nuclear power plant normal operat-safety options appears to be cost effective for the Surry power ing and off-normal operating procedures in a related project, plant. This work supports the Nuclear Regulatory Commission's " Program Plan for Assessing and Upgrading Operating Proce-assessment of severe accidents in NUREG 1150. dures for Nuclear Power Plants"; and (4) solicitation of the com-ments and expert opinions of a peer review group on the draft NUREG/CR-4552: A REVIEW OF THE SEABROOK STATION project report and on proposed techniques for resolving classifi-PROBA0lLISTIC SAFETY ASSESSMENT. Containment Failure cation and interface issues in additon to desenbing these ac-Modes And Radiological Source Terms. KHATIB-RAHBAR; trvities and their results, recommendations for the NRC and utah.

AGRAWAL,A.K.; LUDEWIG,H.; et al. Brookhaven National Labo. ty actions to address procedure classification and interface ratory. March 1987. 75pp. 8704090039. ONL-NUREG-51961. problems are offered.

40463.244.

A technical review and evaluation of the Seabrook Station NUREG/CR-4616: ROOT CAUSES OF COMPONENT FAILURES Probabilistic Safety Assessment has been performed. It is deter- PROGRAM. Methods And Applications. SATTERWHITE,0.;

mined that (1) containment response to severe core melt acci- CADWALLADER.L.; MEALE,0 M.; et al. EG&G Idaho, Inc. (subs.

of EG&G, Inc ). December 1986. 60pp. 8702060177. EGG 2455.

dents is judged to be an important factor in mitigating the con- 39526.227.

sequences, (2) failure dunng the first few hours ar ter core melt This report contains information pertaining to definst>ons, is also unlikely and the tmng of overpressure fadure is very methodologies, and apphcations of root cause analysis. Of spe-long compared to WASH 1400, (3) the point-estimate radologb cal releases are comparable in magnitude to those used in cific interest, and h:ghlighted throughout the discussion, are ap.

WASH.1400, and (4) the energy of release is somewhat higher plications pertaining to current and future Nuclear Regulatory than for the previously reviewed studies. Commission (NRC) light water safety programs. These applica-tions are discussed in view of addressing specific program NUREG/CR 4610: EFFECTS OF LATERAL SEPARATION OF issues under NRC consideration and reflect current root cause analysis capabilities.

OXIDIC AND METALLIC CORE DEBRIS ON THE BWR MK l CONTAINMENT DRYWELL FLOOR. HYMAN.C.R ; WEBER,C F.

Oak Ridge National Laboratory. January 1987, 118pp. NUREG/CR-4626 V02: IMPROVING THE RELIABILITY OF OPEN-CYCLE WATER SYSTEMS. Application Of Biofouhng 8704080065. ORNL/TM-10057. 40444.007. Surveillance And Control Techniques To Sediment And Corro-In evaluating core debns/ concrete for a BWR MK I contain-ment design, it has been common practice to assume that at soon Fouling At Nuclear Power Plants. JOHNSON,K.I.;

NEITZEL,D,A. Battelle Memonal Institute, Pacific Northwest reactor vessel breach, the core debns is homogeneous and of Laboratones. March 1987, 55pp. 8703300148. PNL 5878.

low viscosity so that it is uniformly distnbuted radially on the 40280.224.

drywell floor. In a recent study performed by the NRC spon- Biofouhng surveellance and control techniques are eva:uated sored BWR Severe Accident Technology (DWRSAT) program at Oak Ridge Nat>onal Laboratory, calculations indicate that at re- for their applicability to sediment and corrosion fouling and sug-gest >ons are given to improve their effectiveness. Altemative actor vessel bottom head failure, the debns is such that the me-techn# ques to better detect and Control sedimentation and cor.

talhc components (Zr, Fe, Ni, Cr) are completely molten while the exec components (UO2, ZrO2, FeO) are completery frozen. rosen are also evaluated. Environmental conditions that allow beofouhng, sedimentation, and corrosion to occur are summa.

Thus, the frozen oxides are expected to remain within the reac.

tot pedestal while the molten metallic species radially separate rized. A correlation between sediment and corrosen is identified and the causes are described. Environmental regulatons, espe-from the frozen oxidic species, flow through the opening in the cially those in the Clean Water Act of 1977, are reviewed to reactor pedestal, and spread over the annular region of the identify those that may hmat or prevent the use of survedlance drywell floor between the pedestal and the containment shell.

This report assesses the empact on calculatud containment re- and control techniques desenbed in this report. Flow velocity is the mator design factor that determines whether or not biofoul-sponse and the production and release of fission product laden ing. sedimentat>on, and corrosion will occur. Morutonng flow aerosols for two different cases of debris distnbution Uniform conditions can provide earfy warning of conditions that will allow distnbution and the laterally separated case of 95% oxides 5%

fouling to occur. Visual inspection is the most common and metals inside the pedestal and 5% oxides-95% metals outside most effective technique for identify og the cause and extent of the pedestal. The computer codes used in this assessment are CORCON. MOD 2, MARCON 2.10, and VANESA. fouhng in the operbeycle water system. Most biofouhng control techniques in current use are not effective against sediment and

Main Citations and Abstracts 13 i

corrosion. Frequent, high-velocity flushing of cooling loops may allow for a more mechanistc treatment in calculating tempera-effectuely remove sediment and reduce under sediment corro- tures in the fluid and sohd phases in the debns bed, in deter-sion. Alternative biocide treatments such as targeted chlonna- mnng debns bed dryout, debns bed quenching from either top-tion or the use of ozone or 2,2-dibromo-3-natrito propionamide flooding or bottom-flooding, single and two-phase pressure (DBNPA) may also be effectrve in reducing under-sediment cor- drops across the debns bed, debns bed po'osity, and in finding rosion. the minimum fluidization mass velocity. The inclusion of these m dels in a debns bed computer module will permit a more ac-NUREG/CR-4672: ANALYSIS OF INSTRUMENT TUBE RUP- curate prediction of the coolability charactenstics of the debris TURES IN WESTINGHOUSE 4-LOOP PRESSURlZED WATER bed and therefore reduce some of the uncertainties in assess-REACTORS. FLETCHER,C.D.; BOLANDER,M.A. EG&G Idaho, ing the severe accident characteristcs for BWR appleation.

inc. (subs. of EG&G, Inc.). December 1986.51pp.8702060201. Some of the debns bed theoretical models have been used to EGG-2461, 39527;310 develop a FORTRAN 77 subroutine module called DEBRIS.

A recent safety concern for Westinghouse 4-loop pressunzed DEBRIS is a dnver program that calls other subroutines to ana-water reactors (PWRs) is that, because of a seismic event, in- lyze the thermal characteristics of a packed debns bed. FOR-strument tubes may be broken at the flux mapping seal table, TRAN 77 hstings of each subroutine are provided in the appen-resulting in an uncovenng and heatup of the reactor core. This dix.

study's purpose was to determine the effects upon findings of a similiar 1980 study if certain test vanables changed. A 1980 NUREG/CR-4695: MATERIAL CONTROL AND ACCOUNTING U.S. Nuclear Regulatory Commission (USNRC) analysis of PWR (MC&A) LOSS DETECTION DURING TRANSITION PERIODS hehave used the RELAP4/ MOD 7 computer code to determine AND PROCESS UPSET CONDITIONS. GRIFFIN,E.A.;

the effects of breaking instrument tubes at the reactor vessel YOUNG,1K.; SMITH,B.W.; et al. Battelle Memonal Inst:tJte, Pa-lower plenum walt. The 1986 study discussed here was per- cific Northwest Laboratones. March 1987, 62pp. 8704130187, formed using RELAPS/ MOO 2, an advanced best estimate com- PNL 5890. 40494:263.

puter code. Separate effects analyses investigated instrument The Nuclear Regulatory Commission has implemented regula-tube pressure loss, heat loss, and tube nodalization effects on tions that require licensees to perform tests to detect significant break flow. Systems effects analyses: (a) investigated the ef- losses of strategic special nuclear matenal on a timefy basis facts of changing the break location from the reactor vessel to and to resolve anomalies resulting from such tests. These capa-the seal table, (b) compared RELAP4/ MOD 7 and RELAPS/ bilities have been demonstrated for processes operating at MOD 2 results for an identical transient, (c) venfied a key finding equilibrium; however, the conditions that exist dunng transition from the 1980 analysis, and (d) investigated instrument tube penods, i.e., at startup and shutdown, and dunng process ruptures in th Zion-1 PWR using best-estimate boundary and ini- upsets will impact a licensee's ability to achieve the specified tial conditions. The outcome of these analyses permits adjust- levels of loss detection and alarm resolution. This report dis-ment of the 1980 analysis findings for instrument tube ruptures cusses the types of data available, potentially useful loss tests, at seal table and indicates the best-estimate response of a and techniques that can be used it' developing models for the Westinghouse PWR to the rupture of 25 small instrument tubes abnormal conditions.

at the seal table.

NUREG/CR 4696: CONTAINMENT VENTING ANALYSIS FOR NUREG/CR 4685: POST-PLIOCENE DISPLACEMENT ON THE PEACH DOTTOM ATOMIC POWER STATION.

FAULTS WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF HANSON.DJ, DLACKMAN H.S.; NELSON,W.R.; et al. EG&G EAST-CENTRAL KENTUCKY. VANARSDALE,R B.;

ldaho, Inc. (subs. of EG&G, Inc.). February 1987, 82pp.

SERGEANT,R E. Kentucky, Univ. of, Lexington, KY, Fetaruary 8704070476. EGG-2464. 40438 227.

1987. 47pp. 8703130 t93. 40007.088. The extent to which containment venting is an effective The Kentucky Ruer Fault System forms the northern bounda. means of preventing or mitigating the consequences of ry of the Rome Trough (a Paleozoic aulacogon) in east-central overpressunzation dunng severe accidents was evaluated for Kentucky. Paleozoic recurrent movement along this fauft system the Peach Bottom Atomic Power Station Units 2 and 3 (boiling has been documented by a number of previous worliers; howev. water reactors with Mark I containments). Detailed analyses er, recognition of Mesozoic and earfy Tertiary displacement has were conducted on operator performance, equipment perform-not been possible due to the absence of preserved post.Paleo. ance, and the physical phenomenology for three severe acci-zoic strata Numerous faults of the Kentucky Rwer Fault System dent sequences currently 6dentified as being important contnbu-are partially overlain by Pliocene- Pleistocene terrace deposits tors to nsk. The results indicate that containment venting can along the Kentucky Rwer. Results from preliminary onlling and be effectwe in reducing nsk for several classes of severe acci-electncal-resistwity surveys indicate that a number of these dents but, based on procedures in draft form and equipment in faults may have been actWe since deposition of terrace materi. place at the time of the anafyses, has limited potential for fur-als Dased on indications from the prehminary survey, four sites ther reducing the nsk for severe accidents currently identified as were selected and nine trenches were excavated. Of these nine being inertant contnbutors to the nsk for Peach Bottom.

trenches, four revealed faulted or folded terrace sediments.

Companson of the nine trenches suggests that the folding and NUREG/CH PO VI DRF: CONTAINMENT EVENT ANALYSIS faulting of the terrace deposits is tectonic in orgin and that the FOR PON ulATED SEVERE ACCIDENTS. Surry Power Kentucky Rwer Fault System has been actwo within the last 5 Station, Unit 1. Draft For Comment. DENJAMIN,A S.; BEHR,V.L.;

milhon years and probably within the last 1 million years. KUNSMAN.D M ; et al. Sandia National Laboratones. February NUREG/CR-4689: THERMAL HYDRAULIC AND CHARACTERIS. 1987.195pp. 8703170196. SAND 861135. 40054 302.

TIC MODELS FOR PACKED DEBRIS BEDS MUELLER.G E.; A study has been performed as part of the Severe Accident SOZER,A. Oak Ridge National Laboratory January 1987. Risk Reduction Program (SARRP) to investigate the response 108pp.8702060116. ORNL/TM 10117. 39537.131. of a particular pressurized water reactor with a subatmosphenc APRIL is a mechanistic core wide meltdown and debns relo- containment (Surry Unit 1) to postulated severe accidents. A de-cation computer code for Boiling Water Reactor (BWR) severe tailed containment event tree for the Surry pfant has been de-accident analyses The capabilities of the code continue to be vised to descnbe the vanous possible accident pathways that increased by the improvement of existing models. This report can lead to radioactwe releases from containment. Data and contains information on theory and models for degraded core analyses from a large number of NRC and industry enonsored packed debns beds. The models, when incorporated into programs have been reviewed and used as a basis for c.atify-APRIL, will provide new and improved capabihties in predicting ing the event tree, ie , determining the likelihood of each path-BWR debns bed cootabihty charactenstes These rnodels will way for a vanety of accident sequence initiators A generalized

14 Main Citations and Abstracts containment event tree code, called EVNTRE, has been devel- dustry due to irpplementation of Rev. 3; however, a reassess-oped to facilitate the quantfication. The uncertainty in the re- ment of requirements in other areas can produce an estimated suits has been examined by performing the quantification three net cost savings to industry in excess of $4,500,000 in terms of times, using a different set of input each time to represent the net discounted future cost impact when a discount rate of 5% is vanation of operwon 6n the reactor safety Commurwty. In the so- used. Also, although changes in the guide were determined to called '* central" estimate, the i.kelshood of earty containment produce an unquantifiable change in nsk, it is anticipated that failure (occurnng before or at the time of reactor vessel breach) the changes will have a positrve irnpact on safety and thus will was found to be very low for most accident sequence initiators. lower the nsk and should enhance containment availability.

However, uncertainties surrounding the issues of direct contairw ment pressure capacity could lead to much higher earty failure NUREG/CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANAL-likelihoods. This work supports NRC's assessment of severe YSIS OF A BABCOCK AND WILCOX PRESSURIZED WATER accident nsks to be published in NUREG-1150. REACTOR. Case Study. CRAMOND,W.R.; ERICSON,D.M.;

SANDERS,G.A. Sandia Nabonal Laboratones. March 1987, NUHEG/CR-4708 V01 N1: PROGRESS IN EVALUATION OF RA- 800pp. 8704010504, SAND 86-1832. 40325:245.

DIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY This is one of six case studies for USl A-45 Decay Heat Re.

DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Semiannual Report For October 1985 March moval (DHR) Requirements. The purpose of this study is to 1986. MEYER,R E.; ARNOLD,W.D.; BLENCOE,J.G.; et al. Oak identify any potential vulnerabilities in the DHR systems of a Ridge National Laboratory. January 1987. 63pp. 8702170050. typical Babcock and Wilcox PWR, to suggest possible modifica-ORNL/TM 10147,39655:245. tions to improve DHR capability, and to assess the value and Information that is being developed by projects within the De' impact of the most promising alternatues to the existing OHR systems. The systems anafysis considered small LOCAs and partment of Energy (DOE) pertinent to the potential geochemi-cal behavior of radionuclides at candidate sites for a high-level transient internal initiating events, and seismic, fire, extreme radioactNe waste repository is being evaluated by Oak Ridge wind, internal and external flood, and lightrung external events.

A full-scale systems analysis was performed with detailed fault National Laboratory (ORNL) for the Nuclear Regulatory Com-mission (NRC). Dunng this report period emphasis was placed trees and event trees including support system dependencies.

The system anatysis results were extrapolated into release cate-upon the Yucca Mountain, Nevada, site. Several samples of tuff were anatyzed and characterized by X-ray diffraction and petro- gories using applicable past PRA phenomenological results and graphic analysis. Instial sorption experiments with cesium and improved containment farture mode probabilities. Public conse-strontium demonstrated the necessity of preequilibration and quences were estimated using site-specific CRAC2 calculations.

control of the CO(2) partial pressure over the expenment. A The value-impact (VI) analysis of possible alternatives consid-smati particle size effect was observed for sorption of N(p) on ered both onsite and offsite impacts arriving at several risk various size fractions of tuff. Difficulties were experienced in measures such as averted population dose out to a 50-mile radius and dollars per person rem averted. Uncertainties in the preparing standard solutions of europium. Preliminary compari-son of our data for sorption of cesium and strontium with those VI analysis are discussed and the issues of food and bleed and secondary blowdown are anatyred.

of NNWSI showed that sorption ratios were similar for approxi-mately the same conditions. One of our pnncipal concerns with NUREG/CR 4718: EXPERIMENTAL SUPPORT AND DEVELOP.

the NNWSI data is that much of their data was taken without MENT OF SINGLE ROD FUEL CODES PROGRAM. Summary control of CO(2) partial pressure. Report. LANNING.D D. Battelle Memorial Institute, Pacific NUREG/CR-4711: LOW UPPER-SHELF TOUGHNESS.HIGH. Northwest Laboratories. January 1987. 53pp. 8702170067.

TRANSITION TEMPERATURE TEST NSERT IN HSST PTSE 2 PNL 5972. 39658.051.

VESSEL AND WIDE PLATE TEST SPECIMENS. Final Report. This report summarizes the activities and results of the 11 DOMIAN.H A. Babcock & Wilcox Co.

Oak Ridge National Lab- year Expenmental Support and Development of Single Rod oratory. February 1987. 85pp. 8704130637. 40495.015. Codes Program, sponsored at Pacific Northwest Laboratory by A piece of A387, Grade 22 Class 2 (21/4 Cr .1 Mo) steel the U.S. Nuclear Regulatory Commission. The program included plate specially heat treated to produce low upper-Shelf (LUS) irradiation of extensively instrumented test fuel assemblies at toughness and high transition temperature was Installed in the the Halden Reactor in Norway and postirradiation examination side wall of Heavy Section Steel Technology (HSST) vessel V.8. at Harwell Laboratones in the United Kingdom; ex-reactor stud.

This vesselis to be tested by the Oak Ridge National Laborato. les on gap conductance and fuel rod mechanics; and model de-ry (ORNL) in the Pressunzed Thermal Shock Expenment 2 velopment for the FRAPCON-2 fuel performance computer (PTSE-2) project of the HSST program. Comparable pieces of code. Significant results included long term in-reactor data on the plate were made into six wide-plate specimens and other fuel temperature, fisson gas release, and rod elongation; quan-samples for charactenzation testing to be performed by ORNL tification of the inherent scatter in observed fuel temperatures in replicate rods; and definition and modeling of the thermal and NUREG/CR 4712: REGULATORY ANALYSIS OF REGULATORY mechanical consequences of fuel pellet cracking and fragment GUIDE 1.35 (REVISION 3, DRAFT 2) IN-SERVICE INSPEC-outward relocation. The data obtained are used as benchmark TION OF UNGROUTED TENDONS IN PRESTRESSED CON

data for fuel performance codes and models throughout the CRETE CONTAINMENTS. NAUS.D J. Oak Ridge National Labo- world.

retory. February 1987.123pp. 8704t30262. ORNL/TM.10163.

40495 206. NUREG/CR 4724: FATIGUE CRACK GROWTH RATES IN PRES-The objectives of this study were to review all the changes 6n SURE VESSEL AND PIPING STEELS IN LWR the latest versson (Rev. 3. Draft 2) of " Regulatory Guide 1.35" ENVIRONMENTS Final Report. CULLEN,W.H. Matenals Engi-and to provide a regulatory analysis for all positions in the noenng Associates, Inc. March 1987. 68pp. 8704010118. MEA-guide. Three tasks were undertaken to meet these objectives: 2175. 40321:347.

(1) review of the evolution of prestressed concrete contain- The measurement of fatigue crack growth rates for pressure ments, their prostressing systems and the developrrent of and piping steels in high temperature, pressunted water has

" Regulatory Guide 1.35"; (2) development of a comparatrve reg. been carned out using compact fracture specimens. Over the ufatory analysis between Rev. 3 (Draft 2) and the current ver- last ten years, the prograrns sponsored by the NRC and camed sson, which la in effect (Rev. 2). and (3) conduction of a backfit out at the Naval Research Laboratory and Matenals Engineer-analysis in conformance with the requirements of the Backfitting ing Associates have provided data for over three hundred tests Rule (Section 50109). Results of the study indicate that there of these specimens, which have been published in a series of are certain areas where additional costs may be incurred by in. NUREG topical reports and annual reports. This is the final

Main Citations and Abstracts 15 report in this senes and desenbes bnefly the significant findings background information on the Matenals Characterizaton of the program, reports on the most recent data which have Center (MCC) as well as discussion on statistical consideratons been acquired, and indicates some directions for future re- in fitting leaching and corrosco models to measurements. The MCC was estabbshed to assess and characterize waste pack-search in this area. Further testinq of compact specimens have been nearty phased out, and the program has tumed more to- age matenals for reliable performance for DOE's nuclear waste wards applications- onented tasks, such as vanable cyclic am- needs.

pistude testing, part through crack geometry tests, and environ- NUREG/CR-4736: COMBUSTION AEROSOLS FORMED mental effects on the stress- life curves. DURING BURNING OF RADIOACTIVELY CONTAMINATED NUREG/CR-4730: EVALUATION OF POTENTIAL MIXED MATERIALS EXPER! MENTAL RESULTG HALVERSON,M.A.;

WASTES CONTAINING LEAD, CHROMlUM,USED OIL.OR OR- BALLINGER,M.Y.; DENNIS,G.W. Battelle Memorial Institute, Pa-LIOUIDS. SISKIND.B.; MACKENZIE,0.R.; cific Northwest Laboratones March 1987. 61pp. 8703250500.

GANIC DOWERMAN.B S.; et al. Brookhaven Natonal Laboratory. Janu- PNL 5999. 40234 279.

ary 1987.175pp. 8702270062 DNL NUREG-52019. 39759 249. Safety assessments and environmental 6mpact statements for This report presents the results of follow on stud es conduct- nuclear fuel cycle facilities require an estimate of potential air-ed by Brookhaven National Laboratory (BNL) for the Nuclear borne releases. Radcactrve aerosols generated by fires were Regulatory Commission (NRC) on certain kinds of low-level investigated in expenments in which combustible sohds and lig-waste (LLW) which coud also be classified as hazardous waste unds were contaminated with radioactive materials and bumed.

subject to regulation by the Environmental Protection Agency Uranium in powder and hquid form was used to contaminate five (EPA). Such LLW is termed " mixed waste". Additional data fuel types: polychloroprene, polystyrene, polymethytmethacry-have been collected and evaluated on two categones of poten- late, cellulose, and a mixture of 30% tnbutytphosphate (TOP)in tial mixed waste, namely LLW containing metallic lead and LLW kerosene. Heat flux, oxygen concentration, air flow, contaminant containing chromium. Additionally, LLW with organic hquids, es' concentration, and type of ignition were varied in the experi-pecially houid scintillation wastes, are reviewed. In hght of a pro- ments. The highest release (7.1 wt%) came from burning TBP/

posed EPA rule to list used oil as hazardous waste, the poten- kerosene over contaminated nitric acid. Burning cellulose con-taal mixed waste hazard of used oil contaminated with radionu- taminated with uranyl nitrate hexahydrate liquid gave the lowest clides is discussed. release (0.01 wt%). Rate of release and particle size distnbution of airborne radioactrve particles were highly dependent on the NUREG/CR-4734: SEISMIC TESTING OF TYPICAL CONTAIN- fype of fuel bumed.

MENT PlPING PENETRATION SYSTEMS. CLOSE J A.;

HILL R.C.; STEELE,R. Idaho National Engineering Laboratory. NUREG/CR-4737: INTERPRETATIVE ANALYSIS OF DATA FOR December 1986. 41pp. 8702060293. EGG 2470. 39527.049. SOLUTE 1RANSPORT IN THE UNSATURATED ZONE.

This report provides the results of seismic tests of three typi. FUENTES,RR.; POLZER,WL Los Alamos Scientific Laborato-cal hght water reactor containment penetration systems to pro. ry. January 1987. 242pp. 8702170076. LA 10817 MS.

vide a technical basis for tho support and development of 39655.003.

equipment qualification procedures. The three systems tested. in this report, the movement of 60dide, bromide, and lithium (a) An eightinch gate vane system modeling a containnient under unsaturated flow conditions is modeled using the comput-spray system; (b) An eight inch butterfly vane system modeling er code CFITIM. This code is a soluton of the one dimensional a purge and vent system, and (c) A two-inch globe vane system convective dispersive equation when steady-state flow exists modeling the numerous small bore piping systems that are and when 6nteractions between the solute and Bandelier tuff often characten2ed by high vane to pipe size ratio. The vaNo can be descnbed by the linear isotherm. The model predicts types, sizes, piping configurations, penetrations and supports well the transport of the solutes lodide, bromide, and lithium used for the tests are typical of those found 6n commercial U S. when flow conditons are near steady-state, When assuming av-nuclear power plants for containment isolaton applications. The etage steady-state flow conditions, the model predicts disper-three systems tested were mounted in a fixture and excited with 36on factors for unsteady flow within one to two orders of mag-simulated seismic loads. The loads imposed dunng the tests natude of the predictons at steady state flow; retardation fac-were equal or greater than those expected during U S. operat. lors, on the other hand, are predicted much better than the dis-ing-basis and safe-shutdown earthquakes. The test results indi, persion factors. Differences 6n the estimated dispersion coeffi-cate that adverse vaNo, penetration, or piping system behavior cients for solutes of two steady-state pulses Indicate that the 6n-during typical seismic events is very unlikefy tended replication of those steady. state flow pulses was not achieved dunng expenmentation. A companson of breakthrough NUREQ/CR 4735 V01: EVALUATION AND COMPILATION OF WASTE PACKAGE TEST DATA. Biannual curves of solutes from one depth to another in the 3-m x 6 m DOE Report December 1985 - July 1986. INTERRANTE,C.; field expenmental caisson indicates poor conservation of solute ESCALANTE.E.; FRAAER A ; et al. Commerce, Dept. of. Nation- mass dunng transport.

al Bureau of Standards. March 1987.120pp. 8704010532.

40325:106.

NUREG/CR 4741: FEEDWATER TRANSIENT AND SMALL This report summartres results to date of NOS esaluations of DREAK LOSS OF COOLANT ACCIDENT ANALYSES FOR THE DELLEFONTE NUCLEAR PLANT, DAYLESS.P D.; DODBE.C A ;

Department of Energy (DOE) activities in waste packages de- CHAM 0ERS,R. EGaG Idaho, Inc. (subs. of EGaG, Inc ). March signed for containment of radioactive high-level nuclear waste 1987,110pp. 8704130238. EGG 2471. 40496 015.

(HLW). The waste package is a proposed engineenng tarner Specific sequences that may lead to coro damage were ana-that to part of a permanent repository for HLW. Candidate re-lyzed for the Dellefonte nuclear plant as part of the U.S. Nucle-pository sites include three d,fferent media tuff, basalt, and salt.

at Regulatory Commission's Severe Accident Sequence Anaty-Metal alloys are the pnncipal barners for the proposed canisters and overpacks In adotion, borosihcate glass and vanous pack- ses Program. The RELAP5, SCDAP, and SCDAP/RELAP5 com-puter codes were used in the analyses. The two main initiating ing matonals have been proposed as components of this engi- events investigated were a foss of all feedwater to the steam neenng system Thus, the associated technical problems in-volve corrosion, leaching, dissolution and transport within the generators and a small cold leg break loss of coolant accident The transients of pnmary 6nterest within these categones were waste packages This report gives status reports on waste the TMLD' and S(2)D sequences. Vanations on systems avail-package activities related to each of the three host media Ap-ability were also investigated. Possible operator actions that pended to the report are NBS reviews of selected DOE techns-could prevent or delay core damage were ident,fied, and two cal reports and NBS trip reports of pertinent meetings, semi-nors, and workshops attended. Also presented in the report is were 6nvestigated for a small break transient. All of the tran.

18 Main Citations and Abstracts sients were analyzed until either core damage began or long- be maintained, conditons typical of earty stages in an SGTR/

term decay heat removal was estabhshed. The analyses SORV fault produced large transient releases.

showed that for the sequences considered the intection flow from one high-pressure injection pump was necessary and suffi. NUREG/CR 4762: SHUTDOWN DECAY HEAT REMOVAL ANAL.

cient to prevent core damage in the absence of operator ac. YSIS OF A WESTINGHCUSE 3-LOOP PRESSURIZED WATER tions. Operator actions were able to prevent core damage in the REACTOR Case Study. SANDERS,G A ; ERICSON.D.M.;

S(2)D sequence; no operator actions were available to prevent CRAMOND,W R. Sandia Natonal Laboratones. March 1987.

core damage in the TMLD' sequence. 350pp 870401015P SAND 86 2377,40322.055.

NUREG/CR 4742: MELPROG PWR/ MODI ANALYSIS OF A TMLD' ACCIDENT SEQUENCE. KELLY,J E.; HENNINGER,R.J : moval (DHR) Requirements. The purpose of this study is to DEARING.J F. Sandia Natonal Laboratones. January 1987, dentify any potential vulnerabilities in the DHR systems of a 83pp. 8704070501. SAND 86-2175. 40437:114. typical Westinghouse 3-loop PWR, to suggest possible modifica.

The first complete, coupled, and mechanistic analysis of a tions to improve the DHR capabihty, and to assess the value TMLB' (station blackout) core meltdown accident for the Surry and impact of tne most promising alternatives to the existing plant has been made with MELPROG PWR/MODt. This anaty- DHR systems. The systems analysis considered small LOCAs sis has provided the timing of the major events occurring in the and transient internal initiating events, and seismic, fire, extreme accdont, the amount and timing of hydrogen produced by oxi- wind, internal and external flood, and lightning external events.

dation of the cladding, and the condition and composition of the A full-scale systems analysis was performed with detailed fault disrupted matenal at the time of vessel failure Due to the pre

trees and event trees including support system dependencies.

hminary nature of this first calculation, a hmited number of auxil- The system analysts results were extrapolated into release cate-lary calculations have been performed. Compenson of these re' gones using applicable past PRA phenomenological results and sults with previous calculations have provided further insights improved containment failure modo probabilities. Public conse-6nto this accident. In particular, it is shown that natural convec* quences were estimated using site specific CRAC2 calculations.

tion reduces the rate of core heating. but 6ncreases the rate of The Value-Impact (VI) analysis of possibio alternatrves consid-heating of upper plenum structures. This increased heating can ered both onsite & offsite impacts amving at several nsk meas-inhibit fission product deposition and increase the amount of ures such as averted population dose out to a 50 mile radius and dollars per person rem averted. Uncertainties in the VI anal.

molten structural steel in the melt at vessel failure. It is also shown that couphng between vessel flow and pnmary system ysis are discussed and the issues of feed and bleed and sec-flow may lead to earty het, ting and failure of the pnmary system. ondary blowdown are analyzed.

Hence, natural circulation within the vessel with couphng to the pnmary system can completely change the course and timing of NUREG/CR 4776: RESPONSE OF SEISMIC CATEGORY l TANKS TO EARTHOUAKE EXCITATION. BUTLER,T.A.;

a meltdown sequence This underknes the importance of a DENNETT,J G ; DABCOCK,C D.; et al. Los Alamos Scientific mult4monssonal vessel flow capability as provided by MEL- Laboratory. February 1987. 69pp. 8703260013. LA 10871 MS.

PROG. In addition, the effect of the modehng of the initial fuel 40238.043.

rod melting and relocation has been studied. Variations in the The response of vertical, above-ground, fluid-filled tanks to assumptions were foand to strongly affect hydrogen producton seismic loads is reviewed and hcensing cntena are recommend-and the subsequent course and timing of the accident.

ed for use by the U S Nuclear Regulatory Commission in as-NUREG/CR 4744 V01 Nt: LONG TERM EMBRITTLEMENT OF sessing the safety of seismic Category I tanks. Analysis meth.

CAST DUPLEX STAINLESS STEELS IN LWR ods and relevant esponments are first reviewed to provide a SYSTEMS Semiannual Report,0ctober 1985 March 1988. basis for recommending analytical techniques that are useful for CHOPRA,0.K.; CHUNG,H M. Argonne National Laboratory. Jan, tank safety evaluation Next, field damage that has occurred uary 1987. 47pp. 8704080064. ANL 88-54. 40441290. dunng several earthquakes, starting with the 1964 Great Alaska This progress report summarites work performed by Argonne Earthquake, are reviewed and the damage is categonzed. This National Laboratory dunng the six months from October 1985 to informatron 6s then used, along with expertmental evidence, to March 1986 on long-term embnttlement of cast duplex stainless assess the adequacy of current formal design codes. Finally, a steels used in hght. water reactors procedure for Category I tank evaluaton 6s recommended and topics that need additional research are identified.

NUREG/CR 4752: COINCIDENT STEAM GENERATOR TUDE RUPTURE AND STUCK OPEN SAFETY RELIEF VALVE CAR- NUREG/CR 4787: CONFFRENCE OF RADIATION CONTROL 01-RYOVER TESTS MB 2 Steam Generator Transient Response RECTOR'S INFORMATION FOR LICENSING LOW LEVEL RA-Test Program GARBELT,K.; MENDLER,0.J ; GARDNER.G C ; DIOACTIVE WASTE INCINERATORS AND COMPACTORS.

ot al. Westinghouse Electnc Corp. March 1987, 630pp. Conference of Radiation Control Program Directors, Inc. Janu-8704010141. EPRI NP 4787,40319168 ary 1987.152pp. 8703030822. 39866 219.

In PWR steam generator tube rupture (SGTR) faults, a direct This guxiance was wntten to assist Agreement States and ap-pathway for the release of radioactive fission products can exist phcants addressing low level waste processing as regulators or if there is a coincident stuck-open safety rolief vatvo (SORV) or as hcensees. The Low level Radoactive Waste Management if the safety rehet valve is cycled The test program consisted of Committee (E.5) of the Conference of Radiation Control Pro-sixteen separate tests designed to cover a range of steady- gram Directors prepared this guidance document after evaluat.

state and transient fault conditions. The main conclusions from ing current waste compaction and 6ncineration practices with these tests were that moisture carryover was very low in the ab- consideraton of present apphcable regulatory requirements for sence of an SGTR, that there was no signihcant increase in hconsing incineration and compaction processes to reduce low-moisture carryover dunng an SGTR/SORV fault and that very level radioactive waste volume have been licensed by Agree-little or no pnmary coolant passed through the steam generator ment States and by the U S. Nuclear Regulatory Commission without having irst completely mined with the bulk secondary for over 20 years. Incineration volume reduction factors 6n the hquid (pnmary coolant bypassing). Short term perturbations to range from 10 to 100 have been achieved and compaction can steady. state conditions were found to produce transient re- reduce the waste volume by factors from 2 to 10 or more with leases, which could be mainly due to pnmary coolant bypassing accompanying reduction in the costs for waste disposal. In or these releases were the equivalent of steady state releases preparation of this guidance, the focus has been on keeping re-over tens of hours, and could be 6mportant factors in determin- diation esposure "as low as reasonably achievable " Compac-Ing the overall activity release in these types of fault At very tion and incineration in particular to produce a more stable low water levels, when recirculation within the boiler could not waste form with enhanced performance charactenstics after dis-

Main Citations and Abstracts 17 posal is a positrve step toward that end. This document does and the restart of a pnmary coolant pump when a condition of not specifically address incinerator or compactor installatons at snadequate core cooling was detected. The influence of the nuclear power plants.

break size on the transient seventy was also studied with the inclusion of the 2.1% break transient with secondary steam-NUREG/CR 4784: AN ANALYTICAL AND EXPERIMENTAL IN. and-feed initiation when condition of inadequate core coohng VESTIGATION OF NATURAL CIRCULATION TRANSIENTS IN was detecM AH W of the egenments were performed at A MODEL PRESSURIZED WATER REACTOR. MASSOUD.M. high pressure and temperature [15 6 MPa (2262 psia) pnmary Maryland, Urw. of, College Park, MD. January 1987, 2 71pp. system pressure. 37.5 K (67.5 degrees F) core differential tem-8702170061. 3 % 56 098. perature,587 K (597 degrees F) hot leg fluid temperature) and The Urwersity of MarylandCollege Park "2x4 Loop" scaled all four expenments had an initial bypass flow rate near 3%.

model facility was used to study natural circulation. This facifity Compansons are made between the four experiments and con-simulates a B&W lowered loop type PWR, The expenmental in. clusons drawn regarding what recovery procedure to use to vestigation included determination of system charactenstics as prevent heater rod temperature excursons.

well as system response to imposed transients under symmetnc and asymmetnc conditons Asymmetnc transients were im. NUREG/CR 4797: PROGRESS REVIEWS OF SIX SAFETY PA-posed to study flow oscillation and possible instability The ana- RAMETER DISPLAY SYSTEMS. LINER,R.T.; DEDOR.J. Sci-lytical investigaton encompassed development of mathematical ence Applications International Corp. (formerty Science Apphca-model for single phase, steady state, and transient natural circu- tions, Inc) March 1987, 78pp 8704010129. SAIC-86/3066.

lation as well as modification of existing model for two-phase 40340 241.

flow analysis of phenomena such as small break LOCA, high A pilot program of progress reviews of Safety Parameter Dis-pressure coolant 6ngection, and pump coast down. The modifica- play Systems (SPDSs) was camed out through information-gath-tion included additon of models for once through steam genera- enng visde to six plants in the pered June. November 1985. The tot and electnc heater rods. The development included coding purpose was to sample 6ndustry progress toward the SPOS re-of a computer program entitled Symmetnc and Asymmetnc quirements stated ln NUREG 0737 Supplement 1 and thereby Analysis of Single-Phase Flow!' Flow instability resulting in ces- to determine the need for a post 6mplementation audit program.

sation of circulation was not observed. Pnmary system average While three plants had, to varying degroes, demonstrated the v6-temperature rose dunng a symmetnc to asymmetnc transient abikty of the SPDS concept through effective implementation, while the total secondary side flow rate was maintained. three of the six plants, some having been declared operatonal for as long as two years, had encountered maior problems to NUREG/CR 4789: THE SIMULATION OF THERMOHYDRAULIC the extent that their SPDSs could confuse or mislead operators PHENOMENA IN A PRESSURIZED WATER REACTOR PRI. ln an emergency. The problems observed had not been appar.

MARY LOOP, POPP.M. Maryland, Univ. of, College Park, MD.

ent from poor reviews of Safety Analysis Reports on the sys-January 1987. 280pp. 8702170045. 39658 001.

Several fluid flow and heat transfer phenomena were 6nvesti. tems. Maior conclusions were that (1) a significant number of plants may be having probtems with their SPOSs. and (2) assur-gated. Scaling and modeling laws for PWRs are reviewed and a ance that any given SPDS meets the requirements cannot be new scaling approach focusing on the overall loop behaver is presented. Scahng cnterna for one and two phase natural circu. determined with reasonable confidence without an on-site audit and discussions with personnel responsible for developing, op-laton are developed. Reactor vessel vent vane effects are in.

ciuded in the anatysis. Two new dimensionless numbers, which erating, and maintaining the system. This report summanzes ob-servatons from the six plant visits and presents a plan, with unquely descnbe one-phase flow 6n natural circulaton loops, were deduced and are discussed. A scaled model of the pri. procedures and guidance, for conducting post implementation audits of additional SPOSs if they are undertaken.

mary loop of a typical Dabock and Wilcox reactor was designed, built, and tested The model operates at a maximum pressure of NUREG/CR 4798: IRON OXIDE AEROSOL EXPERIMENTS IN 300 psq and has a maximum heat loput of 188 kW. It is about 4 STEAM AIR ATMOSPHERES NSPP TESTS 501 505 AND times smaller 6n height than the real reactor, with a nominal 511, DATA RECORD REPORT, ADAMS.R E.; TOGIAS,M L Oak volume scale of 1:500. Experiment measurements included pri. Ridge National Laboratory. January 1987. 89pp. 8704130219 mary side temperatures, hot leg velocities, and other pnmary ORNL/TM 10301. 40497.056 and secondary loop performance data. All test data is com, This data record report summantes the results from five tests pared to the theoretically denved performance predictons and involving Fe(2)O(3) test acrosol in a steam-ser environment and scaling laws. The capability of the model to simulate reactor one test in a dry air environment. This research sponsored t'y vent vane effects and small break loss of coolant accidents is the U S. Nuclear Regulatory Commission was conducted 6n the discuswd Suggested changes to the model and its instrumen, Nuclear Safety Pilot Plant at the Oak Ridge Natonal Laboratory.

taten are recommended. General system scaling with complete The purpose of this project is to provide a data base on the be.

and simultaneuug two-phase flow and heat transfer "mulation in haver of sorosols in containment under conditions assumed to all components is not possible with a low pressure foop using occur in postulated LWR accident sequences; this data base water. will provide e ponmental vahdation of aerosol behavioral codes NUREG/CR 4793: RESULTS OF SEMISCALE MOD 2C SMALL* under development. In the report a bnef desenption is given of HREAK LOSS.OF COOLANT ACCIDENT WITHOUT HPl(S-NH) each test together with the results in the form of tables and EXPERIMENT SERIES STREIT.J E. EGAG Idaho, Inc. (subs. of graphs included are data on aerosol mass concentration, aero.

EGAG, Inc.) January 1987. 61pp. 8704080273 EGG-2482- sol fallout and plateout rates, total mass fallout and plateout.

40441:180. aerosol particle sire, vessel atmosphere pressure, vessel at.

Four empenments simulating small-break, loss.of-coolant acci- mosphere temperatures, temperature gradients near the vessel dents of 0 5% and 219. without high pressure injection wera wall, and steam condensation rates on the vessel waft.

performed in the Semisca'e Mod-2C facility These expenments differed 6n break site and recovery procedures. Three of the en- NUREGICR 4801: CLIMATOLOGY OF EXTREME WINDS IN SOUTHERN CALIFORNIA. R AMSDELL,J V ; HUDDE.J M.;

periments had the same break site (0 5%), but recovery proce-dures were vaned to determine what influence vanous proco- ELLIOT,0 L ; et al Battelle Memonal Institute. Pacif.c Northwest dures had on transient seventy lhe recovery procedures in. L aboratorms January 1987.104pp 8702170052. PNL 6085 cluded secondary steam-and feed initiation when a conditon of 39655 307.

inadequate core cooling was detected, secondary steam and' A climatology of annual entreme winds in southern Cahfornia feed initiation when the vessel level reached the top of the core has been prepared The climatology includes a descnption of (before a co'editon of 6nadequate Core Cooling was detected), entremo wind regions, defined on the basis of observed winds

18 Main Citations and Abstracts and topography. Extreme wind distnbution parameters have scobes a method being used to obtam J(Ic), J.R curves, and J been estimated for 46 locatsons using data obtained from the at cleavage for three point bend tests conducted at drop tower Natsonal Chmatic Data Center, Probabihties associated with ex- rates through the ductste to brittle transition regime of the femtsc treme winds have been estmated for these locations The re- A106 steel being testod. The major concluseon is that these suits of the analysis are generalty cons, stent with previous esti- tests can now be accomphshed, though a high degree of exper-mates of extreme winds in southern California. Although, an sev- tise and considerable practical exponence is necessary to oral 6nstances the current estimates are significantry higher than obtain good tost results. The steel tested here is quite rate de-previous esumates. The data examined do not indicate that pendent as shown both by tensile tests and fracture toughness there has been a significant change 6n the extreme wind chmate tests. A load elevation of 30 to 50% results in the drop tower of southern Cahfornia. 100 in/second tests on this matenal m compenson with static NURE0/CR-4803: THE POSS10lLITY OF LOCAL DETONATIONS tests when both tests are conducted on the ductile upper shelf.

DURING DEGRADED-CORE ACCIDENTS IN THE BELLE. Nonetheless, for this matenal J(Ic) and J-R curves are not ele-FONTE NUCLEAR POWER PLANT. SHERMAN.M P.; vated by the loading rate and this rather surpnsing result corte.

DERMAN.M. Sandia National Laboratones Jsnuary 1987.45pp. sponds to a tendency for crack eMiaton to occur at a senauw 8704080052. SAND 861180 40446 240 bend angte for the high rate tests than for the static tests and a It is possible to objectvely determine whether a detonation conospondM gmaW amount of crad extenson in the taped can propagate in a grven geometry (volume, shape and size, specimen at a pen W a@ WW crad maw man la obstacle configuration, degree of confinement) for a given mix. present in the static test.

ture composit>on (concentrations of hydrogen, air and steam).

this is done by conservatively equating the detonation propaga- NUREO/CR 4820: COMPARISON OF THE 1982 SEADEX DIS-taon entena with the entena for transition from deflagration to PERSION DATA WITH RESULTS FROM A NUMDER OF DIF.

FERENT MODELS. LEWELLEN W S.; SYKES,R l; detonation. This paper attempts to reduce the degree of con.

CERASOLl,C P.; et al. Aeronautical Research Associates of servatism in this procedure by constructing estimates of the probabihty of transition to detonation based on subl active e* Pnnesten. February 1967, 163pp. 8703090089. ARAP $75.

39919,152 Popolatons of empincal data. A methodology is introduced The results from simulations by 12 dispersion models are which quahtatively ranks mixtures and geometnes according to the degree to which they are conducsve to transition to detona* compared with observations from an extensive field expenment conducted by the Nuclear Regulatory Commission in a shorehne tort The methodology is then applied to anatyring the potential environment dunng the earty summer 1982. Ten of these for local detonations in the Bellefonte reactor containment for a models are the same as used in the earlier compansons with venety of accident scenanos. Based on code-calculated rates and quantities of hydrogen generation and ca'culated rates for the 1981 field tests at idaho National Engineenng Laboratory.

All the models performed better on this SEADEX expenment.

transport and mixing, this methodology 6ndicated a low potential for detonation except for one volume n a few cases. Little difference between the models is evident with hourty sur.

9 ace data, witn the models able to predict the maximum value m NUREQ/CR 4813: ASSESSMENT OF LEAK DETECTION SYS. the neighborhood of a sampler site within a factor of 2 approsi.

TEMS FOR LWRs October 1985 - September 1986. matety 25% of the time. This to raised to approximately 40%

KUPPERMAN D S Argonne National Laboratory January 1907, when the Companson is based on 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months integrated dose for 48pp. 8704130658. ANL 86-52. 40494 324 the longer time samples the more sophisticated models do it has become apparent that no currentfy available single show a distinct advantage when measured by correlation Coeffb lead. detection method for hght water reactors combines opts. cient and root mean square error, if the data file is assumed 6n.

mal leakage detection sensitivity, leak locating ability, and the complete, with higher data samples possible between the actual desired level of accuracy in leakage measurement. In this data samples, then the best results for the hourty samples show paper, NRC guidehnes for leak detection will be reviewed, cut. over 80% calculated within a factor of 2 when a 15 degree un-rent practices descnbed, potential safety-related problems dis. certainty 6n the plume position is permitted This is raised to cussed, and potential improvements in leak detection technolo. 90% for the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months dose on the same compenson basis.

gy (with emphasis on acoustic methods) evaluated Although in-formation presented here is believed to be vahd for most plants NUREQ/CR 4424: EVALUAflON OF INTEGRAL CONTINUING additional data are needed to 6dentify exceptions For example' EXPERIMENTAL CAPADILITY (CEC) CONCEPTS FOR LIGHT although quantitative leakage determination is possible with' WATER REACTOR RESCARCH PWR SCALING CONCEPTS condensate flow monitors, sump monitors, and pnmary coolant CONDIE,K G ; DAVIS C D ; LARSON,T K.; et al. EGAG Idaho, l inventory balance, these teethods do not provide adequate lo- Inc. (subs. of CGAG, Inc ). February 1987. I t3pp. 8704130652.

Cation Informaton, and are not necessanly sensitive enough to EGO 2494 40494 048.

meet U S. Nuclear Regulatory Commission Regulatory Guide The United States Nuclear Regulatory Commission (USNRC) 1.45 goals. Leak detection capability can be ernproved at spect- in assessing their future research heeds for both separate of.

find sites by use of acoustic monitonng of moisture sensitive facts and integral enpenments has requested that EG&G tape (MST) However, current acoustic monitonnq techniques Idaho,Inc., identify and technica!!y evaluate potential concepts provkfe no source discnmination (e Q , to distinguish between that will maintain the capability to conduct future integral, ther.

leaks from pipe cracks and valves) and no leak rste information mal hydraulic facihty esponments of interest to reactor safety. In (a small leak may saturate the system) MST provides neither this report reactor transients and thermal hydraulic phenomena quantitative leak rate information not specific location informa. of importance (based on probabihstic nok assessments and the ton other than the location of the tape, moreover, its usefulness International Code Assessment Program) to reactor safety were wtth " soft insulation needs to be demonstrated examined and kjentified Estabhshed scahng methodologies were used to develop potential concepts for integral thermabby-NUREO/CR 44te: TRANSITION RANGE DROP TOWER J-R drauhc testing facihties Advantages and d eadvantages of each CURVE TESTING Or A106 STEEL JOYCE.J A US Naval concept are evaluated Anafysis is conducted to esamine the Academy, Annapohs, MD HACKETT,E M David W Taytor scalng i of various phenomena in each of the selected concepts Navat Research & Development Center. Fetxuary 1987. 3tpp Results generalty suggest that a lACihty Capable of operating at 8703250513.40234 248 typical reactor operating conditions will scale most phenomena fracture toughness properties should be measured in the lab-reasonably well Although many phnnomena in facihties using oratory at loading rates and temperatures similar to those en- Froon or water at nontypical pressure will scale reasonably *cil, pected in the appbcation of intorest This is not usually the case those phenomena that are heavity dependent on quahty (heat because of the esperirnental difficulties involved Ihis repor1 de- transfer or Critical flow for example) can be datortnd Further-

M:In Citations and Abstracts 19 umes demonstrate how the seisme margins review guidance more, relation of data produced in facilities operating with nonty-(NUREG/CR-4482) of the NRC Seisme Design Margins Pro-pical fluids or at nontypical pressure to large plants will be a dif-fcult and time consuming process. gram can be applied. The overall objectues of the tnal review are to assess the seissc margins of a particular pressunzed NUREG/CR 4825: A PRELIMINARY EVALUATION OF THE ECO- water reactor, and to test the adequacy of this review approach, NOMIC RISK FOR CLEANUP OF NUCLEAR MATERIAL Ll- quantificaton techniques, and guidehnes for performing the CENSEE CONTAMINATION INCIDENT S. OSTMEYER.R M ; review. Results from the tnal review will be used to revise the SKINNER.D J. Sandia Natonal Laboratones. March 1987 g7pp. seisme margin methodology and guidelines so that the NRC 8703260083. SAND 86 2108. 40245 215. and industry can readily apply them to assess the inherent This report documents an analyts of the economic nska from quantitatue seisme capacity of nuclear power plants.

nuclear materiallicensee contaminston incidents The resulta of the analyses are intended to provide a technical basis for en NUREG/CR 4829 V01: SHIPPING CONTAINER RESPONSE TO SEVERE HIGHWAY AND RAILWAY ACCIDENT NRC rutomaking that would require nuclear materiallicensees to CHOU C.K.;

Report. FISCHER.L.E.;

demonstrate adequate financial means to cover the cleanup CONDITIONS Main GERHARD.M A ; et al. Lawrence LNormore National Laboratory.

costs for accidental or inadvertent telease of radioactive maten*

als. The 6mportant products of this effort include (1) a method of February 1987. 284pp. 8703120326. UCID-20733. 39082.032.

categonting licensees according to the potential cost and fro- This report desenbes a study performed by the Lawrence quency of contamination incidents, (2) a model for ranking the Livermore Natonal Laboratory to evaluate the level of safety (stegones of licensees according to potential incident costs, provided under severe accident conditions dunng the shipment and (3) estimates of contamination nsk for the hcensee catego- of spent fuel from nuclear power reactors. The evaluaton is nes- performed using data from real accident histonea and using rep.

resentatue truck and rail cash models that likely meet 10 CFR NUREG/CR 4826 V01: SEISMIC MARGIN REVIEW OF THE 71 regulations. The responses of the representatNo casks are MAINE YANKEE ATOMIC POWER STATION. Volume t. Summa- calculated for structural and thermal loads generated by severe ry Report. PRASSINOS.P G ; MURRAY,R C ; CUMMINGS.O E. highway and railway accident conditions. The cask responses Lawrence Uvermore National Laboratory. March 1987. 200pp. are compared with those responses calculated for the 10 CFR 8704090044. UCID 20948. 40484.128. 71 hypothetcal accident conditons. By companng the re-This Summary Report is the first of three volumes for the sponses 41 is determined that most hghway and rantway acc6-

"Seismc Margin Review of the Masne Yankee Atome Power dont conditions fall within the 10 CFR 71 hypothetical accident Station." Volume 2 is the Systems Anafysis of the hrst tnal seis- conditions. For those accidents that have higher responses, the mc margin review. Volume 3 documents the results of the fra- probabihties and potential radiation exposures of the accidents gihty screening for the review. The three volumes demonstrate are compared with those 6dentified by the assessments made ln how the seismic margin review guidance (NUREG/CR 4482) of the "Fenal Envronmental Statement on the Transportation of tlas Nuclear Regulatory Commission (NRC) Seisme Desgn Mar- Radioactus Matenal by Air and Other Modes," NUREG 0170.

gins Program can be apphed. The overall oblectwes of the tnal Based ort this compenson, it is concluded that the radiological review are to assess the seismic margins of a particular pres-nsks from spent fuel under severe hghway and railway accident sunted water reactor, and to taat the adequacy of this review conditions as derNed in this study are less than nske previously approach, quantification techniques, and guidehnes for perform-6ng the review. Results from the trial review will be used to estirnated 6n the NUREG O t 70 document.

revise the seismic margin methodology and guidelines so that NUREO/CR 4829 V02: SHIPPING CONTAINER RESPONSE TO the NRC and industry can readily apply them to assess the in- HIGHWAY AND RAILWAY ACCIDENT SEVERE herent quantitetue sesmc capacity on nuclear power plants. CONDITIONS Appendices. FISCHER LE.; CHOU.C.K.;

GERHARD,M A ; et al. Lawrence Lhrermore Natonal Laboratory.

NUREQ/CR 4826 V02: SEISMIC MARGIN REVIEW OF THE February 1987. 300pp. 8703 t 20299 UCID 20733. 39982 318.

MAINE YANKEE AT0MIC POWER STATION. Volume 2 Systems Analysis. MOORE,0 L; JONE S,0 M ; OUILICI.M D ; et al. See NUREG/CR 4829,V01 abstract.

Energy, Inc. March 1987. 201pp. 8704090075. UCID-20948. NUREQ/CR 4443 V01: UNIVERSITY OF MARYLAND AT COL.

40466 232. LEGE PARK (UMCP) 2X4 LOOP TEST FACILITY. Annual Repoit This System Anatysis is the second of three volumes for the For 1985. DiMAR20,M.; HSU,Y.Y.; LIN.W K.; et al Maryland, "Seismc Margin Review of the Maine Yanties Atomic Power Univ. of, College Park, MD. March 1987, 275pp. 8704090054.

Station." Volume 1 is the Summary Report of the first tnal seis. 40485 020.

mio margin review Volume 3. Fragility Anaryve, documents the The efforts for the year 1985 of the 6nvestigators of the Un6-results of the fragibty screening for the review. The three vol. versity of Marytand on the UMCP 2x4 Loop facihty are present-umes demonstrate how the seismic margins review guidance ed. These efforts include. additional work on the f acility, theorst-(NUREG/CH 4482) of the NRC Seismic Design Margins Pro. 6 cal investigations, and empenmental 6nvestgations. The 2x4 gram can be apphed. The overall obloctwes of the tnal review Loop facskty is a low pressure scaled representation of a low-are to assess the seismic margins of a particular pressunted ered loop D&W reactor and once through steam generator. The water reactor, and to test the adequacy of this revin approach. report is prepared in three chapters and seven appendices. A quantification techniques, and guldehnes for perferming the br6ef descnption of the facihty including the final design details review Results from the trial review will be used to revise the are presented in chapter one. Chapter two includes the theorou-seismic margin methodology and guidehnes so that the NRC Cal basis for the superimental 6nvestigations. Chapter three con-and lndustry can readily apply them to assess the inherent tains the details of expenments, test results, and final conclu-quantitative seismic capacity of nuclear power plants. **"'

gop,cs d seus n the chap NURE0/CR 4826 V03: SEISMIC MARGIN REVIEW OF THE MAINE YANKEE ATOMIC POWER ST ATION Volume 3 Fragibty Analysis RAVINDRA.M K; HARDY,0 S ; HASHIMOTO.P.S ; et NUREQ/CR 4847: CASE HISTORIES OF WEST VALLEY SPENT FUEL SHIPMENTS Final Report.

Aerospace Corp. January al EOE, Inc. March 1967, 217pp 8704090063. UCID 20948 1987. 239pp. 8702270048. WPR-86(6811) 1. 39760 064.

40481.13g. in 1983, NRC/FC initiated a study on institutional 6ssues relat-This Fragibty Anafysis is the thwd of three volumes for the ed to spent fuel shipments onginating at the former spent fuel

" Seismic Margin Review of the Mace Yankee Atomec Power reprocessing facility in West Valley, New York. FC staff viewed Station." Volume 1 is the Summary Report of the hrst Inal sois- the shipment campagne as a one time opportunity to document mic margin review Volume 2. Systems Analysis, documents the the snstitutional issues that may ante with a substantial 6ncrease results of the systems screening for the review The three voi

20 M:In Cit:tlens and Abstracts in spent fuel shipping actuity NHC subsequentry contracted determine the feasibility of developing data based protective with the Aerospace Corporaton for the West Valley Study This report containa a detailed desenption of the events which took measure decision levels for assunng uniformity of licensing re.

quirements for radiation protecten of workers. Eleven facihties place pnor to and during the spent fuel shipments. The report licensed to work with unencapsulated radioactrve matenal wore also contains a discussion of the shipment issues that arose, visited to collect data for this purpose. Sutticient data were ob.

and presents general findings Most of the institutional issues tained from six facihties to estimate release fractions (i e., frac.

discussed in the report do not fall under NHC's transportation tion of matenal in process released to the workplace environ.

authonty. The case histones provide a reference to agenciess ment) for tntium tube filhng operatens (about 10F5)), tritium and other institutions they may be myolved in future spent fuel shipping campaigns. compound preparation (about 1007)),1125 radopharmaceutical preparations (about 10(.7)), Am-241 source production (about NUREQ/C&4852: THE MEERS FAULT. TECTONIC ACTIVITY IN 10011)), and forge filing and incineration of natural uranium SOUTHWESTERN OKLAHOMA. HAMELLl,A H ; (about 10( 7)). The fraction of matenal in process that may be SLEMMONS,0 0 ; DROCOUM,SJ Nevada, Univ. of, Hono, NV. ta n into the bodes of workers envolved in these processes March 1987. 50pp. 870409004 7. 40463 318. was estimaW to be about 4s1008), ACOMO),2x6,6x%

The Meers Fault in Southwestern Oklahoma is capable of 14), and W5)s%7), respnctively. Five levels of radiation pro.

producing large, damaging earthquakes Oy compenson to his- tection programs were developed. The most feasible approach toncal events, a mamum of M - 6 3/4 to 7 f /4 could be en- t assunng uniformy of licensing rewirements for these pro.

pected The most recent surface ruptunng event occurred in the grams is a technical one based on mathematical relationships late Holocene, and at appears that one or more pre-Holocene using decision levels developed by a panot of experts. This sp.

eventa preceded it. Surface rupture longtn is at least 37 km pr sch, however, requires a consensus among regulators and Displacements compnsing the present day scarp have left later. licensees on fundamental values used to determine the need at and high angle reverse components Vertical separation of for specific protectwo measures the ground suriace reaches $ m, while lateral separation ca.

coeds the vertical by a rate of about 31 to 51, reaching about NUREG/CR.4859: SEISMIC FHAGILITY TEST OF A 61NCH DI.

20 m. Indtvidual events apparentty had maximum displacements AMETER PIPE SYSTEM CHEN.W P; ONESTO A.Ta DEVITA,V.

of several meters The Meers Fault may be part of a largo' Energy Technology Engineenng Center. February 198 7,174pp.

active tone Based on surface empressions, the Washita Valley, 8703260100. 40237 2M Oklahoma and Potter County, Texas Faults may also have rup- This report contains the test results and assessments of seis.

tures during the late Quaternary, although not as recentry as the mic fragility tests performed on a 6-inch diameter piping system.

Meets Fault. Low sun angle photography en Southwontorn Okla. The test was funded by the U S. Nuclear Acgulatory Commis.

home revealed no evidence of fault activity, other than that of sion (NHC) and conducted by ETEC. The obloctive of the test the Meets Fault, although actwity may be concealed by poof was to investigate the ability of a representative nuclear piping preservation or ductile surface deformation. This suggests that system to withstand high level dynamic seismic and other load.

additional areas of activity may be sparse and rupture infre. ings. Levels of loadings achieved dunng soismic testing were 20 quentry- to 30 times larger than normal elastic design evaluations to ASME level D hmits would permit. 04 sed on falute data ob.

NUREQ/C44853; APPROxlMATE METHOOS FOH FRACTUNE tained dunng seismic and other dynamic testing, it was conclud-ANALYSES OF THROUGH. WALL CHACKED PlPES ed that nuclear piping systems are inherent'y able to withstand BRUST,f'.W Datteile MemorialInstitute, Columbut Laboratone, much larger dynamic seismic loadings than permitted by current February 1967,122pp. 8703250491. DMI 2145 40234124 do*gn practice entona or predicted by the probabilistic risk as.

Cunent leak before break snaryses involve assessing the sessment (PHA) methods and several proposed nonhnear meth.

load. carrying capacity of through wall cracked pipe. Five pre. odt of failure anarysis diction techniques were evaluated in this report the tochnical basis for two analysis methods developed in the Degradod NUREQ/CH 4881: DEVELOPMENT OF SifE SPECIFIC HE.

Piping Program LDH HCL 1 and LUG HCL 2, are presented in this SMSE SPECTHA DEHNHEUTmDJ CKWC; SAVYJ D report Other methods evaluated are the GE/EPHl, NUHLG/ Lawrence Livermore Nahonal Laboratory March 1987. 14 tpp.

CH 3464 and L0n NHC analyses These methods erJ aff based 8704010140 0C10 20980 40321189 on the J integral /teanng modulus theory As such, they all fall for a number of years the USNHC has employed site specfic under the category of lettimation schemes These J. estima. spectra (SSSP) in their evaluation of the adequacy of the Safe tion schemes are all relatively simple to be compareif to finite Shutdown Earthquake (SSE) As the data set eat considerably element analysis Predicting the fracture performance can be it reased for Eastern North Amenca (ENA) and as more rete.

achieved very quickly, and is done through the use of an IBM vant data has become available from earthquakes occurnng 6n PC cornputer code called NHCPIPE The suassments of the other parts of the world (e g , Italy), togethor with the fact that five methods involved companng the esperimentAl data to the recent data indicated the importance of the vertical component, predicted load versus load- point displacement curves This was 11 became Clear that an update of the SSSP's for ENA was de.

done for both carbon steel and stainless stoni pipes with cracks serablo This stuity used actual earthquake ground motion data In the base metal, as well at stainless steel pipes with cracks in with magnitudos within a certain range and recorded at dis.

TIG or submerged arc weids in addition, both deformahon and f ances and at sitet similar to those that would the chosen for modified J H curves were usod in the assessments Finalty, a the dotinition of an SSE An sitension analysis of the origin and sensitivity study wat made to show that the reference strust site of the uncertainty is an important part of this study The re-used in the Hamberg Osgoat relahon could t,o based on c'that suits

,,n, ,of this analysis of the uncert&ntiet is used to develop Crt-the yield strength or the flow strnsa If the contfcont is property g, ,,,,c,,ng the parthquake records to be used in the dan-adl usted d 5%P's We concluded that the SSSPs wore not very sensitive to the distobution of the source to site distance of NUMEQ/CR 4058: FEAS!UtlliY STUDY ON A oat A HAM D the earthquaho records used in the ana'ysit That is, the vana-SYSTLM FOH OrCtSIONS HEGAHDiNG OCCUPAf TONAL HA- bihty (uncertinty) introduced by the range of distances was roi.

04ATION PHOffCilON MI:ASUHIS W A TSON 0 C ,

abvely small compared to the vanataty entroducod by other fac.

FISHEH D H Dettelle Memonal institute, Pacific Nntthwest L ab-orator 6es Fetwuery 1987 71pp 6103M06tn PNL 6131 toes We also conduded that the SSSP are enmowhat sensitive 40235 081. to the datr t>uhon of the magnitwins of thoto earthquAhas, par, ticularty at roth sites and, by interence, at shallow soil sites We in a study commissioned by the U S Nxinar Hogulatory found that one important cntonon in antacting records to genor.

Commission, Pacihc Northwest Laboratory corufucted a study to ato SSSP is the deptn of soil at the oito

M:In Cit:ti:ns cnd Ab:trccts 21 6

This report desenbos the test facshty and test program for NUMEG/CM-4468: METALLURGICAL EVALUATION OF AN 18 studying thermal rmxing of high-pressure injecte (HPI) water in INCH FEEDWATER LINE FAILURE AT THE SURRY UNIT 2 the two-fertha scale model of three cold legs, termannular down.

POWER STATION. CZAJKOWSKl.C.J. Brookhaven National comer and lower plenum of a pressunzed water reactor. This Laboratory. March 1967. 43pp. 8704090019 BNL NUREG. test senes has been camed out by mutual agreement on the

$2057, 40464.016. pressunzed thermal shock (PTS) information exchange between A metallurgical failure anatyses was performed on pieces from the U.S. Nuclear Regulatory Comrmsson and Imatran Voima a catastrophically faded 18ech diameter feedwater hne from Oy. The test factity was onginalty designed to model the Finish the Surry Urvt 2 Nuclear Power State, The failed pipe had sa pant but M was %siged W MM for We test been globalty thinned and had a scattoped appearance on the 7 gram. facil@ can opera at atmosm preswa inside surface. All fracture surfaces examined showed a ductile with loop and HPl flows from different cold legs in the area of failure mode. The matenals of construction met the appropnate 6nterest to PTS. Transparent materials were used to allow flow specification requirements (both mechanical and chemical). The visualization dunng the tests. The choice of transparent materi-report has as its final conctution that the pipe failed due to ex* als htmt the upper temperature to 75 degrees C. The full buoy-ceserve thinning by an erosion corrosson mechanism, ancy effect was induced by salt additx>n and the HPI tempera-ture was used as a tracer. The test matnx consists of 20 tests.

The verwki parameters were flow rates and the number and NUMEG/lA4004: THERMAL MIXING TESTS IN A SEMIANNU-confguration of cold legs with HPl and loop flows. Four tests -

LAH DOWNCOMER WITH INTERACTING FLOWS FROM were done with decreasing loop flow temperature to simulate COLD LEGS. TUOMISTO.H; MUSTONEN,P. Finland, Govt. of.

October 1986. 343pp. 8703090081. 39907.230. pnmary flows dunng steam hne breaks.

l 1

(

s" Secondary Report Number Index This index lists, in alphabetical order, the performing organization-issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross-referenced to the NUREG number for the report and to the 10-digit NRC Document Control System accession number.

SECONDARY MPORT NUtstER MPORT NUsstER SECONDARY REPORT NUsstER REPORT NUsesER NUREG/CR4796 NUREG/CR4813 ORNUTM 10301 ANL-86 52 PNL-4054 NUREG/CR-3620 SO2 ANL-86 54 NUREG/CR-4744 V01 N1 PNL 5210 NUREG/CR-3950 V03 ANL AA-30 NUREG/CR4012 VO2 PNL 5511 NUREG/CR4300 V03 N2 ARAP 575 NUREG/CR-4620 PNL-5648 NUREG/CR-3966 BMI-2145 NUREG/CR-4853 PNL 5711 NUREG/CR4469 V04 NUREG/CR-3861 PNL 5652 NUREG/CR-4613 BMI 2147 NUREG/CR 2331 V06 N2 W BNL NUREG-51454 U BNL-NUREG-51699 NUREG/CR-3444 V04 NUREG/CR 3469 V03 PNL-5972 NUREG/CR-4716 BNL-NOREG-51706 NUREG/CR4736 BNL-NUREG-51826 NUREG/CH-4016 V02 PNL-5999 NUREG/CR-4409 V02 PNL-6065 NUREG/CR4801 BNL-NUREG-51934 PNL 6137 NUREG/CR4856 RNL NUREG-51961 NUREG/CR 4552 NUREG/CR-4797 NUREG/CP4054 SAIC-86/3066 BWNUREG-52011 SAND 631482 NUREG/CR-3412 V02 BNL-NUREG-52019 NUREG/CR-4730 NUREG/CR-3466 NUREG/CP-0064 SAND 644383 BNL NUREG-52044 SAN 0651309 NUREG/CR4301 BNL-NUREG-52057 NUREG/CR4866 SAND 65-1449 NUREG/CR-4320 EGG-2440 NUREG/CR 4531 SAND 65 2373 NUREG/CR-4448 EGG-2455 NUREG/CR-4616 SAND 661135 NUREG/CH-4700 V1 DRF MG-2461 NUREG/CR-4672 SAND 66-1180 NUREG/CR4803 EGG-2464 NUREG/CR-4696 SAND 66-1309 NUREG/CR-4551 V1 DRF EGG-2470 NUREG/CR-4734 SAND 661832 NUREG/CR-4713 NUREG/CR-4741 SAND 66-2064 NUREG/CR4550 V04 EGG-2471 NUREG/CR-4550 V03 EGG-2462 NUREG/CR-4793 SAND 66 2064 EGG-2494 EPRI NP-4767 NUREG/CR4824 NUREG/CR4752 3 g

f h NUREG/CR-4524 SAND 66 2377 NUREG/CR-4762 IEB M 24 NUREG/CR4458 NUREG/CR4737 SAND 66 2496 LA 10017 MS TPRD/U3009/R66 NUREG/CR4752 LA 106714AS NUREG/CR-4776 NUREG/CR4829 V02 MEA-2122 NUREG/CR-4491 UCO20733 UCID-20733 NUREG/CR4829 V01 MIA 2175 NUREG/CR4724 NUREG/CR4826 V01 NUREG/CR 2000 V05Nt2 UCO20948 ORNUNSC200 UCO20948 NOREG/CR4826 V03 ORNUNSC200 NUREG/CR-2000 %U6 N1 NUREG/CR4826 V02 NUREG/CR4610 UCJD 20948 ORNL/TM 10057 UCO20960 NUREG/CR4861 ORNUTM-10117 NUREG/CR-4689 NUREG/CR-4752 NUREG/CR 4706 V01 N1 WCAP 11226 ORNUTn410147 WPR-66(6611)-1 NUREG/CR4847 ORNL/TM. to163 NUREG/CR-4712 23

)

1

l l

Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by the author. If further information is needed, refer to the main cita-tion by the NUREG number.

B E E B E,M.R.

ADAMS J.W.

NUREG/CR3444 V04 THE IMPACT OF LWR DECONTAMINATIONS NUREG-0020 V10 N09: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of August 31.1986 (Gray Book f)

ON SOLIDIFICATION. WASTE DISPOSAL.AND ASSOCIATED OCCU-PATIONAL EXPOSURE Annual Report, Fiscal Year 1986 NUREG4020 VIO N10. LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of September 30.1986(Gray Book 1)

NUREG/CR 4798. IRON OX1DE AEROSOL EXPER;MENTS IN STEAM- BEHR,V.L AIR ATMOSPHERES NSPP TESTS 501505 AND 511. DATA RECORD NUREG/CR4700 V1 DRF: CONTAINMENT EVENT ANALYSIS FOR REPORT. POSTULATED SEVERE ACCIDENTS. Surry Power Station, Unit 1. Draft For Comment AGRAWAL,A.K.

NUREG/CR4552 A REVIEW OF THE SEABROOK STATION PROBABI- BENJAMIN A.S

LISTIC SAFETY ASSESSMENT Containment Failure Modes And Radi- NUREG/CR-4551 V1 DRF: EVALUATION OF SEVERE ACCIDENT '

d gcal Source kms RISKS AND THE POTENTIAL FOR RISK REDUCTON.SURRY POWER STATION. UNIT 1 Draft For Comment .

ALBRECHT,K.A.

NUREG/CR 4700 V1 DRF: CONTAINMENT EVENT ANALYSIS FOR NUREG/CR 2478 V03 A STUDY OF TRENCH COVERS TO MINIMl2E POSTULATED SEVERE ACCIDENTS. Surry Power Stata,Urut 1. Draft INFILTRATION AT WASTE DISPOSAL SITES Final Rept For Comment ANDERSON,N.R.

BENNETT,J.G.

NUREG-1211: REGULATORY ANALYSIS FOR RESOLUTION OF UNRE-SOLVED SAFETY ISSUE A-46. SEISMIC OUALIFICATION OF EQUIP- NUREG/CR4776 RESPONSE OF SEISMIC CATEGORY l TANKS TO MENT IN OPERATING PLANTS. EARTHOUAKE EXCITATON.

ARNOLD,W.D. BERMAN,M.

NUREG/CR-4708 V01 N1: PROGRESS IN EVALUATION OF RADIONU- NUREG/CR4803. THE POSSIBlUTY OF LOCAL DETONATIONS CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH' DURING DEGRADED-CORE ACCIDENTS IN THE BELLEFONTE NU-LEVEL NUCLEAR WASTE REPOSITORY SITE CLEAR POWER PLANT.

PROJECTS Semiannual Report For October 1985 - March 1986.

BERNREUTER.D.J.

NURE NUR G[C 4776- RESPONSE OF SEISMIC CATEGORY I TANKS TO pECTR '

EARTHOUAKE EXCITATION BERTUCIO,R.C.

CAlLEY,W.J. NUREG-1150 DRF V1 FC: REACTOR RISK FtEFERENCE NUREG/CR 3950 V03. FUEL PERFORMANCE ANNUAL REPORT FOR DOCUMENT Main Report Draft For Comment 1985' NUREG/CR-4550 V03. ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS SURRY UNIT 1.

CALLINGER.M.Y.

NUREG/CR4736. COMBUSTION AEROSOLS FORMED DURING BLACKMAN.H.S.

BURNING OF RADIOACTIVELY CONT AMINATED MATERIALS EX. NUREG/CR-4696: CONTAINMENT VENTING ANALYSIS FOR THE PERIMENTAL RESULTS. PEACH BOTTOM ATOMIC POWER STATION.

BARNES,V.E.

BLEJW AS T.E.

NUREG/CR-3968. STUDY OF OPERATING PROCEDURES IN NUCLE- V02: CONTAINMENT INTEGRITY AR POWER Pt. ANTS Practces And Problems NUREG/CR-3412 NUREG/CR4613 EVALUATION OF NUCLEAR POWER PLANT OPER- PROGRAM. Progress Report,Apn! 1983 -December 1984.

ATING PROCEDURES CLASSIFICATIONS AND INTERFACES Problems And Techniques For Improvement BLENCOE J.G.

NUREG/CR4708 V01 N1: PROGRESS IN EVALUATON OF RADIONU-NUREG R-4469 V04 NONDESTRUCTIVE EXAMINATION (NDE) REll- E EL N C EAR WAS E E S T RY TE ABILITY FOR INSERVICE INSPECTION OF UGHT WATER PROJECTS Semiannual Report For October 1985 March 1986.

REACTORS Semiannual Report. October 1985 - March 1986.

BOLANDER,M.A.

NUREG/CR-4672: ANALYSIS OF INSTRUMENT TU3E RUPTURES IN NUR G/CR-3469 V03 OCCUPATONAL DOSE REDUCTION AT NU- WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS.

CLEAR POWER PLANTS. Annotated Bib 6cgraphy Of Selected Read-In Radiaton Protecton And ALARA' BONZON,LL

^^ ^ NUREG/CR 4301: STATUS REPORT ON EQUIPMENT QUALIFICATION DOSE TIO RESEARCH P ECTS ISSUES RESEARCH AND RESOLUTION.

BAYLESS,P.D. BOWERMAN,B.S.

NUREG/CR-4741- FEEDWATER TRANSIENT AND SMALL BREAK NUREG/CR4730: EVALUATION OF POTENTIAL MIXED WASTES CON-LOSS OF COOL. ANT ACCIDENT ANALYSES FOR THE BELLEFONTE TAINING LEAD. CHROMIUM.USED OIL OR ORGANIC LIQUIDS.

NUCLEAR PLANT.

BEAVERS,J.A. BOYD.G.J.

NUREG/CR-386t: STRESS-CORROSION CRACKING OF LOW- NUREG/CR-4551 V1 DRF: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION.SURRY STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE POWER STATION. UNIT 1. Draft For Comment.

REPOSITORY ENVIRONMENTS.

25 l

26 Personal Author Index mRoCouM.S.J. CHUNG,H.M.

NUREG/CR4852- THE MEERS FAULT: TECTONIC ACTIVITY IN SOUTHWESTERN OKLAHOMA. NUREG/CR-4744 V01 N1: LONG TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Senuannual Report, October 1985 - March 1986.

NUREUCR-4695: MATERIAL CONTROL AND ACCOUNTING (MC&A) CLOSEJ.A.

LOSS DETECTON DURING TRANSITION PERIODS AND PROCESS UPSET CONDITONS. NUREG/CR-4734: SEISMIC TESTING OF TYPICAL CONTAINMENT PIPING PENETRATION SYSTEMS.

CRUST,F.W.

COMEN,L.

NUREG/CR-4853: APPROxlMATE METHODS FOR FRACTURE ANALY-SES OF THROUGH-WALL CRACKED PIPES. NUREG-0837 V06 NO3: NRC TLD DIRECT RADIATON MONITORING NETWORK. Progress Report. July. September 1986.

SUDNITZ,RJ.

COND4E,K.G. -

NURE S1150 DRF V1 FC: REACTOR RtSK REFERENCE DOCUMENT.Mam Report. Draft For Comment NUREG/CR-4824: EVALUATION OF INTEGRAL CONTINUING EXPERI-MENTAL CAPABluTY (CEC) CONCEPTS FOR LIGHT WATER REAC- 1 TCR RESEARCH PWR SCALING CONCEPTS.

NUREG/CR4012 V02: REPLACEMENT ENERGY COSTS FOR NUCLE- ]

CRAMONO,W.R.

AR CLECTRICITY GENERATING UNITS IN THE UNITED STATES.19871991. NUREG/CR4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case BUSTARD.LD. Study.

NUREG/CR.4301: STATUS REPORT ON EQUIPMENT QUAUFICATION NUREG/CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF ISSUES RESEARCH AND RESOLUTON. A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case l Study. '

NUREG/CR-4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF EG/CR-4776; RESPONSE OF SEISMC CATEGORY I TANKS TO "O" ***

EARTHOUAKE EXCITATION, CADWALLADER,L- * "*

NUREG/CR-4616: ROOT CAUSES OF OOMPONENT FAILURES NUREG/CR-4724: FATIGUE CRACK GROWTH RATES IN PRESSURE PROGRAM.Methode And Appicatons. VESSEL AND PIPING STEELS IN LWR ENVIRONMENTS. Final Report CARTWRIGHTK CUMMINGS,GL NUREG/CR-2478 V03- A STUDY OF TRENCH COVERS TO MINIMlZE NUREG/CR-4826 V01: SEISMIC MARGIN REVIEW OF THE MAINE INFILTRATION AT WASTE DISPOSAL SITES Foal Rept YANKEE ATOMIC POWER STATON. Volume 1. Summary Report CASEp.L CUNNINGHAM,M.A.

NUREG/CR.4708 V01 N1: PROGRESS IN EVALUATON OF RADONU- NUREG-1150 DAF V1 FC: REACTOR RISK REFERENCE CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH- DOCUMENT. Main ReportDraft For Comment.

LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS.Sermannual Report For October 1985 - March 1986.

REG CR4 : METALLURGICAL EVALUATON OF AN 18-INCH CATHEY,N.G.

FEEDWATER LINE FAILURE AT THE SURRY UNIT 2 POWER STA-NUREG/CR4550 V04. ANALYSIS OF CORE DAMAGE FREQUENCY TON.

FROM INTERNAL EVENTS PEACH BOTTOM UNIT 2.

-DAEMEN,JJ.

CERASOU.C.P.

NUREG/CR-4541: EXPERIMENTAL ASSESSMENT OF THE SEALING NUREG/CR-4820 COMPARISON OF THE 1982 SEADEX DISPERSION EFFECTIVENESS OF ROCK FRACTURE GROUTING.

DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

gygg, CHAMBERS.R.

NUREG/CR4824: EVALUATON OF INTEGRAL CONTINUING EXPERI-NUREG/CR4741: FEEDWATER TRANSIENT AND SMALL BHEAK MENTAL CAPABluTY (CEC) CONCEPTS FOR UGHT WATER REAC-LOSS OF COOLANT ACCOENT ANALYSES FOR THE BELLEFONTE TOR RESEARCH - PWR SCALING CONCEPTS.

NUCLEAR PLANT.

DEAN,R.S.

CHANG,T.V.

NUREG/CR4524: CLOSEOUT OF IE BULLETIN 80 24. PREVENTION NUREG-1030: SEISMC QUAUFCATION OF EQUIPMENT IN OPERAT. OF DAMAGE DUE TO WATER LEAKAGE INSIDE CONTAINMENT ING NUCLEAR POWER PLANTS. Unresolved Safety issue A46. (OCTOBER 17,1980 INDIAN POINT 2 EVENT).

NUREZr1211: REGULATORY ANALYSIS FOR RESOLUTON OF UNRE-SOLVED SAFETY ISSUE A46.SEISMC QUAUFICATION OF EQUIP. DEARINGJ.F.

MENT IN OPERATING PLANTS.

NUREG/CR-4742: MELPROG-PWR/ MODI ANALYSIS OF A TMLB' AC.

CHEN,J.C. CIDENT SEQUENCE.

NUREG/CR-4861: DEVELOPMENT OF SITE SPECIFIC RESPONSE DEBORA SPECTRA.

NUREG/CR-4797: PROGRESS REVIEWS OF SIX SAFETY PARAME-CHEN,"".P. TER DISPLAY SYSTEMS.

NUREG/CR-4859. SEISMIC FRAGIUTY TEST OF A 6-INCH DIAMETER DEFFENSAUGH,J.

PIPE SYSTEM.

NUREG/CR4469 V04: NONDESTRUCTIVE EXAMINATON (NDE) REU-CH0PRA.O.K. ABluTY FOR INSERVCE INSPECTION OF LIGHT WATER REACTORS.Semannual Report, October 1985 - March 1986.

NUREG/CR-4744 V01 N1: LONG TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Sermannual DENNING,R.S.

Report. October 1985 March 1986.

NUREG-1150 DAF V1 FC: REACTOR RISK REFERENCE DOCUMENT. Man ReportDraft For Comment.

CHOU C.K.

NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO NUREG-1150 DRF V2 FC: REACTOR RISK REFERENCE DOCUMENT.Apperdces A-1.Oraft For Comment.

SEVERE HIGHWAY AND RAILWAY ACCOENT CONDITIONS Mam Report DENNIS,G.W.

NUREG/CR4829 V02- SHIPPING CONTAINER RESPONSE TO NUREG/CR4736: COMBUSTION AEROSOLS FORMED DURING SEVERE HIGHWAY AND RAILWAY ACCIDENT COND(TONS.Appendees. BURNING OF RADCACTIVELY CONTAMINATED MATERIALS - EX-PERIMENTAL RESULTS.

Personal Author index 27 DEVITA,V. FOLEY,WJ.

NUREG/CR-4859 SEISMIC FRAGIUTY TEST OF A 6-INCH DIAMETER NUREG/CR-4524: CLOSEOUT OF IE BULLETIN 80-24.PREVENTON PIPE SYSTEM. OF DAMAGE DUE TO WATER LEAKAGE INSIDE CONTA!NMENT (OCTOBER 17,1980 INDIAN POINT 2 EVENT).

DIMARZO,M.

NUREG/CR-4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE FRAKER,A.

PARK (UMCP) 2X4 LOOP TEST FACILITY. Annual Report For 1985- NUREG/CR-4735 V01: EVALUATON AND COMPtLATON OF DOE WASTE PACKAGE TEST DATA. Bennual ReportDecember 1985 -

DOOSE.CA July 1986.

NUREG/C44741: FEEDWATER TRANS!ENT AND SMALL BREAK LOSS OF COOLANT ACCIDENT ANALYSES FOR THE BELLEFONTE FUENTES,H.R.

NUCLEAR PLANT. NUREG/CR-4737: INTERPRETATIVE ANALYSIS OF DATA FOR SOLUTE TRANSPORT IN THE UNSATURATED ZONE.

DOCTOR,S.R.

NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) REU- GAR 8ELT,K.

ABILITY FOR INSERVICE INSPECTON OF UGHT WATER NUREG/CH-4752 COINCIDENT STEAM GENERATOR TUBE RUPTURE REACTORS.Serniannual Report. October 1985 - March 1986. AND STUCK-OPEN SAFETY RELIEF VALVE CARRYOVER TESTS.MB-2 Stearn Generator Transent Response Test Program.

DOMIAN,H.A.

NUREG/C44711: LOW UPPER-SHELF TOUGHNESS.HIGH-TRANSl- GARDNER G.C.

TION TEMPERATURE TEST INSERT IN HSST PTSE-2 VESSEL AND NUREG/CR-4752: ColNCIDENT STEAM GENERATOR TUBE RUPTURE WIDE-PLATE TEST SPECIMENS. Final Report AND STUCK-OPEN SAFETY REUEF VALVE CARRYOVER TESTS.MB-2 Stearn Generator Transient Response Test Program.

ELUOT,D.L.

NUREG/C44801: CLIMATOLOGY OF EXTREME WINDS IN SOUTH- GARNSTY,R.

ERN CAUFORNIA. NUREG/C44752 COINCOENT STEAM GENERATOR TUBE RUPTURE ANO STUCK OPEN SAFETY REUEF VALVE CARRYOVER tearn enerata Dansient Response Test Nam REG -4016 V02: APPUCATON OF SUM.MAUD:A TEST OF AN INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING GERHARD,M.A.

EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO REUABluTY. Volume ll:Appem$ ices. SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITONS Main Report EMEIGH.C.W. NUREG/CR-4829 V02- SHIPPING CONTAINER RESPONSE TO NUREG-1280 STANDARD FORMAT AND CONTENT ACCEPTANCE SEVERE HIGHWAY AND RAILWAY ACCIDENT CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNTING CONDITONS. Appendices.

(MC&A) REFORM AMENDMENT.10 CFR Part 74 Subpart E.

GilTTER,J.G.

EMmT,R. NUREG 1210 V01: PtLOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG 0933 S06: A PRIORITIZATON OF GENERIC SAFETY ISSUES. DENT INCIDENT RESPONSE TRAINING MANUALOverwew And Surn-rnary Of Map N ERICSON,D M.

NUREG-1210 V02: PILOT PROGRAM:NRC SEVERE REACTOR ACCI-NUREG/C44448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF DEM INCOEM RESPONSE MAlm WKhe Reacta 4 A GENERAL ELECTRIC BWR3/ MARK 1, Case Study.

NUREG/C44458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NU G-1 1 03: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR Case DENT INCIDENT RESPONSE TRAINING MANUALResponse of U-NUR /C44713: SHUTDOWN DECAY HEAT REMOVAL ANALVSIS OF NUR 12 PO PROG NRC SEVERE REACTOR ACCI-A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case DENT INCIDENT RESPONSE TRAINING MANUALPublic Protective Actions - Predetennned Critena And latial Actions.

NUR /C44762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG-1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-A WESTfNGHOUSE 3-LOOP PRESSURt2ED WATER REACTORCase DENT INCOENT RESPONSE TRAINING MANUALU.S. Nuclear Regu.

3 latory Comrmssion Response.

ERNST,M.L GILLEN,K.T.

NUREG-1150 DAF V1 FC: REACTOR RISK REFERENCE NUREG/CR-4301: STATUS REPORT ON EQUIPMENT OUAUFICATON DOCUMENT. Main ReportDraft For Comment.

ISSUES RESEARCH AND RESOLUTON.

ESCALANTE.E.

WASTE PAC G E ATA. B 198 NUREG-[150 DRF V1 FC: REACTOR RISK REFERENCE DOCUMENT. Main ReportDraft For Comment.

July 1986.

GOOD,M.S.

FERRELL,W.L.

NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) RELi-NUREG/C44550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY ABluTY FOR INSERVICE INSPECTION OF UGHT WATER FROM INTERNAL EVENTS. PEACH 80TTOM UNIT 2. REACTORS. Semiannual Repor1. October 1985 - March 1986.

FlSCHER,L.E.

GRAMANN,R.H.

NUREG/C44829 V01: SHIPPING CONTAINER RESPONSE TO NUREG-0525 R12: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL).

SEVERE HIGHWAY AND RAILWAY ACCOENT CONDITONS. Main Report GRAVES,H.L NUREG/CR 4829 Vot SHIPPlNG CONTAINER RESPONSE TO SEVERE HIGHWAY AND RAILWAY ACCIDENT NUREG/CP-0054: PROCEEDINGS OF THE WORKSHOP ON SOIL-CONDtTONS Appendices. STRUCTURE INTERACTON.

GRIFFIN,E,A.

FISHER 3 R.

NUF FCR-4856. FEAS18;UTY STUDY ON A DATA-BASED SYSTEM NUREG/CR-4695: MATERIAL CONTHOL AND ACCOUNTING (MC&A)

FOH DECISIONS REGARDING OCCUPATIONAL RADIATION PRO- LOSS DETECTON DURING TRANSITION PERIODS AND PROCESS TECTION MEASURES UPSET CONDITIONS.

FLETCHER,C.D. GRIFFIN,JJ.

NUREG/CR-467? ANALYSIS OF INSTRUMENT TUBE RUPTURES IN NUREG/CR-4826 V03: SEISMIC MARGIN REVIEW OF THE MAINE YANKEE ATOMIC POWER STATION. Volume 3.Fragitrty Analysts.

WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS.

28 Personal Author index HSU,Y.Y.

NUREGICR4012 V02: REPLACEMENT ENERGY COSTS FOR NUCLE-AR ELECTRICfTY-GENERATING STATES 19871991. UNITS IN THE UNITED NUREG/CR 4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE PARK (UMCP) 2X4 LOOP TEST FACluTYAnnual Report For 1985.

HACKETT,E.M. HUSSE,J.M.

NUREG/CR-4818: TRANSITON RANGE DROP TOWER J-R CURVE TESTING OF A106 STEEL NUREG/CR-4801: CUMATOLOGY OF EXTREME WINDS IN SOUTH-ERN CAUFORNIA.

HUMPHREYS,P.C.

NURE CRd736: COMBUSTION AEROSOLS FORMED DURING NUREG/CR-4016 V02: APPUCATON OF SUM-MAUD A TEST OF AN BURNING OF RADIOACTIVELY CONTAMINATED MATER!ALS + EX. INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING PERIMENTAL RESULTS' EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND REUABluTY. Volume it:Appendces.

HAMILTON.B.P.

EE R EN T G UN TS E N ED Rd STCTES.19871991. 4300 V03 N2: ACOUSTIC EMISSION / FLAW RELATION-HANSON,D.J. SH1P FOR INSERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Progress Rept.Apni-September 1986.

NUREG/CR4696: CONTAINMENT VENTING ANALYSIS FOR THE HYMAN.C.R.

PEACH BOTTOM ATOMIC POWER STATON.

HARDY,GA NUREG/CR 4610: EFFECTS OF LATERAL SEPARATON OF OXIDIC AND METALUC CORE DEBRIS ON THE BWR MK I CONTAINMENT NUREG/CR-4826 V03: SEISMIC MARGIN REVtEW OF THE MAINE DRYWELL FLOOR.

YANKEE ATOMIC POWER STATION. Volume 3 Fragdsty Analysis.

INTERRANTE,C.

NU EG NUREG/CR4735 V01: EVALUATON AND COMPILATION OF DOE

-4550 V03. ANALYSIS OF CORE DAMAGE FREQUENCY WASTE PACKAGE TEST DATA. Bennual ReportDecember 1985 -

FROM INTERNAL EVENTS: SURRY, UNIT 1. M 1986.

NUREG/CR4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS. PEACH BOTTOM UNIT 2. ISACHSEN,Y"W*

HASHIMOTO.P.S. NUREG/CR-3232 DETAILED STUDIES OF SELECTED,WELL EXPOSED FRACTURE 2ONES IN THE ADIRONDACK MOUNTAINS DOME,NEW NUREG/CR 4826 V03: SEISMIC MARGIN REVIEW OF THE MAINE YORK.

YANKEE ATOMIC POWER STATION. Volume 3 Fragelsty Analysts.

HATCH,S.W. I NUREG/CR-4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR4708 V01 N1: PROGRESS IN EVALUATION OF RADONU.

O GENERAL ELECTRIC BWR3/ MARK I Case Study. CUDE LEVEL GEOCHEMICAL NUCLEAR INFORMATION DEVELOPED BY DOE HIGH.

WASTE REPOSITORY SITE PROJECTS. Semiannual Report For October 1985 - March 1986.

NUREG/CR4469 V04: NONDESTRUCTIVE EXAMlNATON (NDE) REll- JANG,J.

ABluTY FOR INSERVICE INSPECTON OF UGHT WATER !

REACTORS Sermannual Report, October 1985 - March 1986. NUREG4837 V06 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1986.

HENNICK A.

NUREG/CR-4524: CLOSEOUT OF IE BULLETIN 80-24 PREVENTION OHNSON,u OF DAMAGE DUE TO WATER LEAKAGE INSIDE CONTAINMENT NUREG/CR-4626 V02- IMPROVING THE REUABluTY OF OPEN-(OCTOBER 17,1980 INDIAN POINT 2 EVENT)' CYCLE WATER SYSTEMS. Application Of Biofouhng Surveillance And HENNINGER.R.J. Control Technsques To Sedment And Corrosion Fouling At Nuclear Power Plants.

NUREG/CR-4742:

C1 DENT SEQUENCE,MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' AC-JOHNSON,T.M.

HERZOG,8.L. NUREG/CR-2478 V03: A STUDY OF TRENCH COVERS TO MINtMlZE NUREG/CR 2478 V03; A STUDY OF TRENCH COVERS TO MINIMlZE INFILTRATION AT WASTE DISPOSAL SITESFinal Rept.

INFILTRATON AT WASTE DISPOSAL SITES Final Rept. JONES D.M.

NUREG/CR-4826 V02: SEISMIC MARGIN REVIEW OF THE MAINE N E /CR4734: SEISMIC TESTING OF TYPICAL CONTAINMENT PlPING PENETRATON SYSTEMS.

JOYCE,J.A.

NUREG/CR-4818: TRANSITION RANGE DROP TOWER J.R CURVE NU EG 210 V01: PILOT PROGRAM NRC SEVERE REACTOR ACCI- TESTING OF A106 STEEL DENT INCIDENT RESPONSE TRAINING MANUAL. Overview And Sum. KAUFMAN,M.

ma_rrOf Mahr Points.

NUREG-1210 V02- PILOT PROGRAM.NRC SEVERE REACTOR ACCI' NUREG/CR4735 V01: EVALUATON AND COMPILATON OF DOE DENT INCOENT RESPONSE TRAINING MANUALSevero Reactor Ac- WASTE PACKAGE TEST DATA. Bennual ReportDecember 1985 -

cident Overview. July 1986.

NUREG-1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-DENT INCIDENT RESPONSE TRAINING MANUAL. Response Of U- KEEFER D.A.

NIREY12 PO NUREG/CR-2478 V03: A STUDY OF TRENCH COVERS TO MINIMlZE OGR INRC SEVERE REACTOR ACCl- INFILTRATION AT WASTE DISPOSAL SITES. Final Rept.

DENT INCIDENT RESPONSE TRAINING MANUAL.Public Protective KELLY,J.E.

NU 121 5 T ROGRA R SE E REACTOR ACCI- NUREG/CR-4742' MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' AC-DENT INCOENT RESPONSE TRAINING MANUALU.S. Nuclear Regu- CIDENT SEQUENCE.

latory Comenssion Response.

KEMPF,C.R.

HOLLADAY,C.G.

NUREG/CR4801: CUMATOLOGY OF EXTREME WINDS IN SOUTH. NUREG/CR-4730- EVALUATION OF POTENTIAL MIXED WASTES CON-ERN CAUFORNIA. TAINING LEAD, CHROMlUM,USED OILOR ORGANIC LOUIDS.

HORSCHEL,DS KENNEDY.W.E.

NUREG-1131 NUREG/CR3412 V02: CONTAINMENT INTEGRITY V02: ONSITE DISPOSAL OF RADIOACTIVE PRC3 RAM Progress Report,Apnl 1983 -December 1984. WASTE Methodology For Tr e Radiological Assessment Of Disposal By Subsurface Burial.

Personal Author index 29 LINER,R.T.

NUREG/CR-3820 SO2: INTRUDER DOSE PATHWAY ANALYSIS FOR NUREG/CR.4797: PROGRESS REVIEWS OF SIX SAFETY PARAME-THE ONSITE DISPOSAL OF RADCACTIVE WASTES.The ONSfTE/

MAX 11 Computer Program.

TER DISPLAY SYSTEMS.

LUDEWIG,H.

KHAN.T.A.

NUREG/CR,3489 V03: OCCUPATIONAL DOSE REDUCTION AT NU- NUREG/CR4552- A REVIEW OF THE SEABROOK STATON PROBASI-CLEAR POWER PLANTS. Annotated B6hography Of Se6ected Read- USTIC SAFETY ASSESSMENT.Contamment Failure Modes And Radi-

%d Sowce Tenns.

NU G 09 V : A AS NUCLEAR POWER PLANT MACKENZIE,D.R.

DOSE REDUCTON RESEARCH PROJECTS.

NUREG/CR4730- EVALUATON OF POTENTIAL MIXED WASTES CON KHATIS-MAHSAR TAINING LEAD, CHROMlUM,USED OIL OR ORGANIC LIQUIDS.

NUREG/CR-4552: A REVIEW OF THE SEABROOK STATION PROBABI-USTIC SAFETY ASSESSMENT.Contamment Failure Modes And Rads

- - N REG 1 50 DRF V1 FC: REACTOR RISK REFERENCE DOCUMENT. Man Report. Draft For Comment.

KIMURA,C.Y.

NUREG/CR4829 V01: SHIPPING CONTAINER RESPONSE TO MARSHALL.B.W.

SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITONS. Main NUREG/CR-3468: HYDROGEN. AIR. STEAM FLAMMABluTY UMITS Report. AND COMBUSTON CHARACTERISTICS IN THE FITS VESSEL NUREG/CR-4829 V02: SHIPPING CONTAINER RESPONSE ACC1 DENT TO SEVERE HIGHWAY AND RAILWAY MART,G.A.

CONDITONS.Appendees. NUREG/CR4469 V04: NONDESTRUCTIVE EXAMINATON (NDE) REU-KOLACZKOWSKI,A. ABluTY FOR INSERVICE INSPECTON OF UGHT WATER REACTORS.Senwannual Report, October 1985 - March 1988.

NUREG/CR-4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS PEACH BOTTOM UNIT 2. MARTIN,J.A.

NUREG-1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-E 820 COMPARISON OF THE 1982 SEADEX DISPERSION DENT INCIDENT RESPONSE TRAINING MANUALOvennew And Sum-DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. mary Of Major Posnts.

NUREG-1210 V02: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-KUNSMAN,D.M. DENT INCIDENT RESPONSE TRAINING MANUALSevers Reactor Ac-NUREG/CR-4551 VI DRF; EVALUATION OF SEVERE ACCIDENT cadent Overwew.

RISKS AND THE POTENTIAL FOR RISK REDUCTION:SURRY NUREG-1210 V04: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-POWER STATON. UNIT 1. Draft For Comment. DENT INCIDENT RESPONSE TRAINING MANUALPubic Protective NUREG/CR4700 V1 DRF: CONTAINMENT EVENT ANALYSIS FOR Actons . P edetermined Cntena And Irwtial Actions.

POSTULATED SEVERE ACCIDENTS. Surry Power Station,Urut 1. Draft MARTIN,R.W.

KUPPERMAN,D S.

NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITIONS. Man NUREG/CR-4813: ASSESSMENT OF LEAK DETECTION SYSTEMS Report.

FOR LWRELOctober 1985 September 1988.

NUREG/CR4829 V02: SHIPPING CONTAINER RESPONSE TO ACCIDENT SEVERE HIGHWAY AND RAILWAY LAmentGHT.J.A. CONDITONS. Appendices.

NUREG/CR4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS. PEACH BOTTOM UNIT 2. MASSOUD,M.

EWCR.4788: M ANEM N b NNEM ^ MESW-REG'/CR-4708 V01 N1: PROGRESS IN EVALUATON OF RADIONU.

CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-NUCLEAR WASTE REPOSITORY SITE NURE 4 3 V01 U VERSITY OF MARYLAND AT COLLEGE LEVEL PROJECTS.Serruannual Report For October 1985 - March 1988. PARK (UMCP) 2X4 LOOP TEST FACluTY. Annual Report For 1985.

LANN8NG,D.D. U "

NUREG/CR4718: EXPERIMENTAL SUPPORT AND DEVELOPMENT OF NUREG' /CP-0085: MEETING WITH STATES ON THE LOW-LEVEL RA-SINGLE-ROD FUEL CODES PROGRAM.Suinmary Report. DIOACTIVE WASTE POUCY AMENDMENTS ACT (LLRWPAA) OF 1985.

LARSON,T.H.

NUREG/CR-2478 V03: A STUDY OF TRENCH COVERS TO MINIMlZE MCCREER Y,0.E.

INFILTRATON AT WASTE DISPOSAL SITES. Foal Rept.

i NUREG/CR4824: EVALUATON OF INTEGRAL CONTINUING EXPERI-I MENTAL CAPABILITY (CEC) CONCEPTS FOR UGHT WATER REAC-LARSON.T.K. TOR RESEARCH - PWR SCAUNG CONCEPTS.

NUREG/CR4531: AN INVESTIGATION OF INTEGRAL FACluTY SCAL-ING AND DATA RELATON METHODS (INTEGRAL SYSTEM TEST MCGUIRE,M.V.

PROGRAM). NUREG/CR-3968: STUDY OF OPERATING PROCEDURES IN NUCLE NUREG/CRaa24: EVALUATION OF INTEGRAL CONTINUING EXPERI- AR POWER PLANTS. Practces And Problems.

MENTAL CAPABluTY (CEC) CONCEPTS FOR UGHT WATER REAC-TOR RESEARCH PWR SCAUNG CONCEPTS. MCKENNA T.J.

LEONARD,M. NUREG-1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG-1150 DRF V2 FC: REACTOR RISK REFERENCE DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum-DOCUMENT.Appendees A-l. Draft For Comment. mary Of Major Pomts.

NUREG-1210 V02: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-DENT INCOENT RESPONSE TRAINING MANUALSevere Reactor Ac-NR /CR 20- COMPARISON OF THE 1982 SEADEX DISPERSf0N DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. NUR -1 1 3: PILOT PROGRAM NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALResponse Of U-LIOGETT,W. consee And State And Local Offemfs.

NUREG/CR4735 V01: EVALUATON AND COMPILATION OF DOE NUREG-1210 V04: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-WASTE PACKAGE TEST DATA. Biannual Report. December 1985 DENT INCIDENT RESPONSE TRAINING MANUALPublic Protective July 1988. Actions . Predeterrruned Cntena And Irvtial Actions.

NUREG 1210 V05: PILOT PROGRAM NRC SEVERE REACTOR ACCl-LIN,W.K. DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu-NUREG/CR-4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE latory Comrrussion Response.

PARK (UMCP) 2X4 LOOP TEST FACluTY. Annual Report For 1985.

30 Personal Author Index assas e a M NUREl/CR-4616; ROOT CAUSES OF COMPONENT FAILURES NUREG/CR4700 VI DRF: CONTAINMENT EVENT ANALYSIS FOR PROGRAM.Methode And Appicatora POSTULATED SEVERE ACCIDENTS. Surry Power Staten,Urvt 1. Draft For Comment.

MENDLER,OA MURPHY,JA NUREG/CR4752: COINCOENT STEAM GENERATOR TUBE RUPTURE AND STUCK-OPEN SAFETY REUEF VALVE CARRYOVER NUREG-1150 DAF V1 FC: REACTOR RISK REFERENCE TESTS.MB-2 Sloam Generator Transient Response Test Program. DOCUMENT. Man ReportDraft For Comment MENetNG,R.W. MURRAY,R.C.

NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO NUREG/CR-4826 V01: SEISMIC MARGIN REVIEW OF THE MAINE SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITONSMan YANKEE ATOMC POWER STATON. Volume 1. Summary Report NU CR4829 V02: SHIPPING CONTAINER RESPONSE TO MUSTONEN,P.

SEVERE HCHWAY AND RAILWAY ACCIDENT NUREG/lA4004: THERMAL MIXING TESTS IN A SEMIANNULAR CONDTIONS. Appendices.

DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS.

MEYER,R.E.

NAJAF1,8.

NUREG/CR4706 V01 N1: PROGRESS IN EVALUATION OF RADONU-CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- NUREG/CR4550 V04: ANALYSIS OF CORE DAMAGE FREOUENCY LEVEL NUCLEAR WASTE REPOSITORY FROM INTERNAL EVENTS. PEACH BOTTOM UNIT 2.

SITE PROJECTS.Serrmannual Report For Octot:er 1985 - March 1986. NAPIER,8A MILLER.C.W. NUREG/CR 3620 S02- INTRUDER DOSE PATHWAY ANALYSIS FOR NUREG-1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCl- THE ONSITE DISPOSAL OF RADCACTIVE WASTES.The ONSITE/

MAXH Computer Progrm DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum.

Of Magor Points.

NATSIAVAS,S.

NUR 1210 V02 PILOT PROGRAM.NRC SEVERE REACTOR ACCI DENT INCIDENT RESPONSE TRAINING MANUALSevere Reactor Ac' NUREG/CR-4776 RESPONSE OF SEISMC CATEGORY I TANKS TO cident Overview. EARTHOUAKE EXCITATON.

NUREG-1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALResponse Of U- NAUSAA NUY12 P LO PROG NUREG/CR-4712- REGULATORY ANALYSIS OF REGULATORY GUIDE INRC SEVERE REACTOR ACCl- 1.35 (REVISION 3 DRAFT 2) - IN-SERVICE INSPECTON OF UN-DENT INCIDENT RESPONSE TRAINING MANUALPubhc Protectree GROUTED TENDONS IN PRESTRESSED CONCRETE CONTAIN-Actions . Predetermined Cntena And initial Actons. MENTS.

NUREG.1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu. NEITZEL.DA letory Commisson Response NUREG/CR-4626 V02- IMPROVING THE RELIA 8luTY OF OPEN-MILSTEAD,W. CYCLE WATER SYSTEMS. Applicaton Of Biofouling Sunreillance And Control Techniques To Sediment And Corrosen Fouling At Nuclear NUREG-0933 S06: A PRORITIZATION OF GENERIC SAFETY ISSUES. Power Plants.

l MITCNELL JA NELSON,W.R.

NUREG-1150 DRF V1 FC: REACTOR RISK REFERENCE DOCUMENT.Mac ReportDraft For Comment. NUREG/CR-4696: CONTAINMENT VENTING ANALYSIS FOR THE PEACH BOTTOM ATOMIC POWER STATON.

MDFFETT,D.L NUREG/CR-2478 V03: A STUDY OF TRENCH COVERS TO MINIMlZE NEUDER,S.M.

INFILTRATION AT WASTE DISPOSAL SITES Final Rept. NUREG-1101 V02: ONSITE DISPOSAL OF RADIOACTIVE MOORE,D.L WASTE. Methodology For The Radological Assessment Of Dsposal By Subsurface Burial NUREG/CR4826 V02: SEISMIC MARGIN REVIEW OF THE MAINE NUREG/CR-3620 S02: INTRUDER DOSE PATHWAY ANALYSIS FOR YANKEE ATOMC POWER STATON. Volume 2 Systems Analysis.

THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

MORGENSTERN.M.

NUREG/CR-3968: STUDY OF OPERATING PROCEDURES IN NUCLE- O'KELLEY,G.D.

AR POWER PLANTS. Practices And Problema.

NUREG/CR-4708 V01 N1: PROGRESS IN EVALUATON OF RADtONU-MORRIS,8.M' CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-NUREG-115 0 DRF V1 FC: REACTOR RISK REFERENCE LEVEL NUCLEAR WASTE REPOSITORY SITE DOCUMENT.Mac ReportDraft For Comment. PROJECTS. Semiannual Report For October 1985 - March 1986.

MOUNT.M.E. ONESTD,A.T.

NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO NUREG/CR-4859: SEISMIC FRAGluTY TEST OF A 6-INCH DIAMETER SEVERE HIGHWAY AND RAILWAY ACCIDENT COND! TONS. Man PIPE SYSTEM.

NU CR-4829 V02: SHIPPING CONTAINER RESPONSE TO OSTMEYER,R M.

SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITIONS. Appendices. NUREG/CR4825: A PREUMINARY EVALUATION OF THE ECONOMIC RISK FOR CLEANUP OF NUCLEAR MATERIAL UCENSEE CONTAMI-MUELLER.G.E. NATION INCIDENTS.

NUREG/CR4689: THERMAL HYDRAUUC AND CHARACTERISTIC PARKINS,R N.

MODELS FOR PACKED DEBRIS BEDS.

NUREG/CR-3861: STRESS-CORROSION CRACKING OF LOW-STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE

'G CR4843 V01: UNIVERSITY OF MARYLAND AT CT,LLEGE ORY ENWROWENTS.

PARK (UMCP) 2X4 LOOP TEST FACluTY. Annual Report For 1985-PELOOUIN.RA MUNNO,FA NUREG/CR-3620 S02: INTRUDER DOSE PATHWAY ANALYSIS FOR NUREG/CR4843 VOI: UNIVERSITY OF MARYLAND AT COLLEGE THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

PARK (UMCP) 2X4 LOOP TEST FACluTY. Annual Report For 1985. MAXH Computer Program.

MURFIN/2.8. PERKINS,K.

NUREG/CR4551 V1 DRF: EVALUATON OF SEVERE ACCIDENT NUREG 1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-RISKS AND THE POTENTIAL FOR RISK REDUCTON SURRY DENT (NCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu-POWER STATION. UNIT 1. Draft For Comment. latory Commission Response.

9

Personal Author Index 31 PERTMER,G. RYBICKl.E.F.

NUREG/CR-4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE NUREG/CR-4491: DEVELOPMENT OF MODELS FOR WARM PRES-PARK (UMCP) 2X4 LOOP TEST FACILITY. Annual Roport For 1985. TRESSING PHIUPPACOPOULO SAKENAS,C.A.

NUREG/CP4054. PROCEEDINGS OF THE WORKSHOP ON SOIL

NUREG-1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-STRUCTURE INTERACTION. DENT INCIDENT RESPONSE TRAINING MANUALResponse Of U-consee And State And Local Officials.

PICIULOAL. NUREG-1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG/CR-3444 V04: THE IMPACT OF LWR UcCONTAMINATIONS DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu-ON SOLIDIFICATON. WASTE DISPOSAL,AND ASSOCIATED OCCU- latory Conmssion Response.

PATONAL EXPOSURE. Annual Report. Fescal Year 1986.

NUREG/CR4730- EVALUATON OF POTENTIAL MIXED WASTES CON- SALLETT,D.W.

TAINING LEAD, CHROMlUM,USED Oll,OR ORGANIC UQUIDS. NUREG/CR-4843 VO1: UNIVERSITY OF MARYLAND AT COLLEGE PARK (UMCP) 2X4 LOOP TEST FACILITY. Annual Report For 1985.

PITTMAN J.

NUREG4933 S06: A PRIORITIZATON OF GENERIC SAFETY ISSUES. SANDERS,G.A.

NUREG/CR4448. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF POLZER,W.L. A GENERAL ELECTRIC BWR3/ MARK 1. Case Study.

NUREG/CR-4737: IN1ERPRETATIVE ANALYSIS OF DATA FOR NUREG/CR4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF SOLUTE TRANSPORT IN THE UNSATURATED ZONE. A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case Study.

POPP,M.

NUREG/CR4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR-4789- THE SIMULATION OF THERMOHYDRAUUC PHE- A BABCOCK AND WiLCOX PRESSURIZED WATER REACTOR. Case NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP. Study.

NUREG/CR-4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE PARK (UMCP) 2X4 LOOP TEST FACIUTY. Annual Report For 1985. NUREG/CR-4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTOR. Case PRASSWOS,P.G. Study.

NUREG/CR-4826 V01: SEISMIC MARGIN REVIEW OF THE MAINE SATTERWNITE,0.

YANKEE ATOMIC POWER STATON Volume 1. Summary Report.

NUREG/CR4616: ROOT CAUSES OF COMPONENT FAILURES U G'/CR-4552: A REVIEW OF THE SEABROOK STATON PROBABI-USTIC SAFETY ASSESSMENT. Containment Fanure Modes And Radi- SAVY,J.B.

ological Source Terms- NUREG/CR4861: DEVELOPMENT OF SITE SPECIFIC RESPONSE QUluCI,M.D.

NUREG/CR-4550 V03: ANALYSIS OF CORE DAMAGE FREQUENCY SCHAFFER A.

FROM INTERNAL EVENTS.SURRY UNIT 1. NUREG/CR4541: EXPERIMENTAL ASSESSMENT OF THE SEALING NUREG/CR-4826 V02: SEISMIC MARGIN REVIEW OF THE MAINE EFFECTIVENESS OF ROCK FRACTURE GROUTING.

YANKEE ATOMIC POWER STATON. Volume 2. Systems Analysis.

SCHNEIDER,K.

NUREG/CP-0085: MEETING WITH STATES ON THE LOW LEVEL RA-NUREG 37 V06 NO3; NRC TLD DIRECT RADIATION MONITORING OACTIVE WASTE POUCY AMENDMENTS ACT (LLRWPAA) OF NETWORK. Progress Report, July-September 1986. S.

RADFORD.LR. SERGEANTAE.

NUREG/CR 3968: STUDY OF OPERATING PROCEDURES IN NUCLE. NUREG/CR-4685: POST-PUOCENE DISPLACEMENT ON FAULTS AR POWER PLANTS. Practices And Problems. WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF EAST CEN-NUREG/CR4613; EVALUATON OF NUCLEAR POWER PtANT OPER.

CLASSIFICATONS AND TRAL KENTUCKY.

ATING PROCEDURES INTERFACES. Problems And Techniques for improvement SHARPE,R.W.

RAMELLI,A.R. NUREG-1210 VOI: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-NUREG/CR-4852- THE MEERS FAULT: TECTONIC ACTIVITY IN DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum-mary Of Major Points.

SOUTHWESTERN OKLAHOMA.

NUREG-1210 V02: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-RAMSDELL.J.V. DENT INCIDENT RESPONSE TRAINING MANUALSevere Reactor Ac.

NUREG/CR4801: CUMATOLOGY OF EXTREME WINDS IN SOUTH- cident Overwew.

ERN CAUFORNIA. NUREG-1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-DENT INCOENT RESPONSE TRAINING MANUALResponse Of U-RAVINDRA M.K. censee And State And Local Officials.

NUREG/CR.4826 V03: SEISMIC MARGIN REVIEW OF THE MAINE NUREG-1210 V04: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-YANKEE ATOMIC POWER STATION. Volume 3.Fragthty Analyses. DENT INCIDENT RESPONSE TRAINING MANUALPublic Protectrve Actions - Redetennned Cntes And W Actions.

RIGGS.R. NUREG-1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG-G933 S06: A PRIORITIZATION OF GENERIC SAFETY ISSUES. DENT INCOENT RESPONSE TRAINING MANUALU.S. Nuclear Regu-latory Conmsseon Response.

RIVARD J.B.

NUREG/CR-4320 THE RELATONSHIP AND INFLUENCES OF FUEL SHERMAN, P D LANT SYSTEM PROCESSES DURING LWR SEVERE ACCI-DURING DEGRADED-CORE ACCIDENTS IN THE BELLEFONTE NU-CLEAR POWER PLANT.

ROSA E.A.

NUREG/CR4016 V02 APPUCATION OF SLIM.MAUD A TEST OF AN SHULL,R.

INTERACTIVE COMPUTER. BASED METHOD FOR ORGANIZING NUREG/CR-4735 V01: EVALUATON AND COMPILATION OF DOE EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND WASTE PACKAGE TEST DATA. Biannual ReportDecember 1985 -

RELIABluTY. Volume ll. Appendices.

July 1986.

ROSS,P.A. SIMONEh,F.A.

NUREG4020 V10 N09: LICENSED OPERATING REACTORS STATUS NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) REU-

SUMMARY

REPORT Data As Of August 31.1986.(Gray Booli !)

NUREG-0020 V10 N10: UCENSED OPERATING REACTORS STATUS ABILITY FOR INSERVICE INSPECTION OF UGHT WATER

SUMMARY

REPORT Data As Of September 30,1986.(Gray Book f) REACTORS. Semiannual Report,Cctober 1985 - March 1986.

32 Personal Author index

$4SKIND,8.

VANDERMOLEN,H.

NURE1/CH-4730: EVALUATION OF POTENTIAL MIXED WASTES CON-TAINING LEAD, CHROMlUM.USED Oll OR ORGANC UQUIDS. NUREG-0933 SO6: A PRIORITIZATION OF GENERO SAFETY ISSUES.

VANKUlKENJ.C.

NUREG/CR4825: A PREUMINARY EVALUATION OF THE ECONOMC " "

RISK FOR CLEANUP OF NUCLEAR MATERIAL UCENSEE CONTAMI- E R E U TS E NATlON INCIDENTS. STATES.19871991-SLEMMONS DR. VESELY,W.E. '

1 NUREZ/CR4852- THE MEERS FAULT:TECTONC ACTIVITY IN NUREG/CR-4616: ROOT CAUSES OF COMPONENT FAILURES SOUTHWESTERN OKLAHOMA. PROGRAM. Methods And Applications.

SMITH,8."*.

WANG,Z.V.

NUREG/CR 4695: MATERIAL CONTROL AND ACCOUNTING (MC&A)

LOSS DETECTON DURING TRANSITION PERIODS AND PROCESS NUREG/CR4343 V01: UNIVERSITY OF MARYLAND AT COLLEGE UPSET CONDITONS. PARK (UMCP) 2X4 LOOP TEST FACluTY. Annual Report For 1985.

SOZER.A- WATKINS,R.M.

NUREG/CR4689- THERMAL-HYDRAUUC AND CHARACTERISTIC NUREG 1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-MODELS FOR PACKED DEBRIS BEDS. DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum-mary Of Map Points.

SPANNER).C.

NUREG-1210 V02: PtLOT PROGRAM.NRC SEVERE REACTOR ACCl-NUREG/CR4469 V04: NONDESTRUCTIVE EXAMINATON (NDE) REU- DENT INCIDENT RESPONSE TRAINING MANUALSevere Reactor Ac. j ABtUTY FOR INSERVICE INSPECTION OF UGHT WATER cident h.

REACTORS Semiannual Report,0ctober 1985 - March 1986. j NUREG 1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCl- 1 SPETTELL.C.M.

DENT INCIDENT RESPONSE TRAINING MANUALResponse of Li-consee And State And Local Officials. l NUREG/CR4016 V02: APPLICATON OF SUM-MAUD.A TEST OF AN NUREG-1210 V04: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-INTERACTIVE COMPUTFR-8ASED METHOD FOR ORGANIZING DENT INCOENT RESPONSE TRAINING MANUALPublic Protective EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND Actions - Predetermined Crtteria And Irvttel Actions.

REUABluTY. Volume ll: Appendices.

NUREG-1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-STEELE,R. DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu- s latory Corrmssion Response. '

NOREG/CR-4734: SEISMIC TESTING OF TYPICAL CONTAINMENT PIPING PENETRATON SYSTEMS. WATSON,E.C.

NUREG/CR-4856: FEASIBluTY STUDY ON A DATA-BASED SYSTEM N E CR-2478 V03: A STUDY OF TRENCH COVERS TO MINIMlZE INFILTRATION AT WASTE DISPOSAL SITES. Final Rept. CT M RE -

STONESIFER,R.B. WEBER.C.F.

NUREG/CR4491: DEVELOPMENT OF MODELS FOR WARM PRES. NUREG/CR4610- EFFECTS OF LATERAL SEPARATION OF OXfDIC TRESSING.

AND METALUC CORE DEBRIS ON THE BWR MK I CONTAINMENT DRYWELL FLOOR.

STREITJ.E.

NUREG/CR4793: RESULTS OF SEMISCALE MOD.2C SMALL-BREAK WEISS,A.J.

LOSS-OF COOLANT ACCIDENT WITHOUT HPl (S-NH) EXPERIMENT NUREG/CP-0082 V01: PROCEEDINGS OF THE FOURTEENTH WATER SERIES.

REACTOR SAFETY INFORMATION MEETING.

SURSOCKJ.P- NUREG/CP-0082 V02: PROCEEDINGS OF THE FOURTEENTH WATER REACTOR SAFETY INFORMATON MEETING.

NUREG 1163: COORDINATON OF SAFETY RESEARCH FOR THE BABCOCK AND WILCOX INTEGRAL SYSTEM TEST PROGRAM. NUREG/CP-0082 V03: PROCEEDINGS OF THE FOURTEENTH WATER REACTOR SAFETY INFORMATION MEETING.

NUREG/CP-0082 V04: PROCEEDINGS OF THE FOURTEENTH WATER NUREG/CR4820: COMPARISON OF THE 1982 SEADEX DISPERSION REACTOR SAFETY INFORMATION MEETING.

DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. NUREG/CP-0082 V05: PROCEEDINGS OF THE FOURTEENTH WATER REACTOR SAFETY INFORMATION MEETING.

TAYLOR,T.T.

NUREG/CP-0082 V06: PROCEEDINGS OF THE FOURTEENTH WATER NUREG/CR4469 V04. NONDESTRUCTIVE EXAMINATON (NDE) REU. REACTOR SAFETY INFORMATON MEETING.

ABluTY FOR INSERVICE INSPECTION OF UGHT WATEP NUREG/CR-2331 V06 N2- SAFETY RESEARCH PROGRAMS SPON-REACTORS.Semaannual Report, October 1985 - March 1986. SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Ouarterty Progress Report.Apriklune 1986.

THOMPSON N.G.

NUREG/CR-3861: STRESS CORROSION CRACKING OF LOW. WHEELER.W.A.

STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE NUREG/CR-3968: STUDY OF OPERATING PROCEDURES IN NUCLE-REPOSITORY ENVIRONMENTS.

AR POWER PLANTS. Practices And Problems-TOBIAS.M.L WlENER,R.W.

NUREG/CR-4798: IRON OXIDE AEROSOL EXPERIMENTS IN STEAM-AIR ATMOSPHERES NSPP TESTS 501505 AND 511, DATA RECORD NUREG/CR-3232 DETAILED STUDIES OF SELECTED.WELL EXPOSED REPORT. FRACTURE ZONES IN THE ADIRONDACK MOUNTAINS DOME,NEW YORK.

TUOMISTO H.

WILLIAMS,D.C.

NUREG/lA-0004: THERMAL MIXING TESTS IN A SEMIANNULAR COWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS. NUREG/CR4551 V1 DRF: EVALUATON OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTON;SURRY VAN FLEET,LG. POWER STATION, UNIT 1. Draft For Comment.

NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) REll-ABILITY FOR INSERVICE INSPECTION OF UGHT WATER WITTE.M.C.

REACTORS Semiannual Report. October 1985 - March 1986. NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO VANARSDALE,R.B. SEVERE HIGHWAY AND RAILWAY ACCOENT CONDITONS. Main Report.

NUREG/CR-4685: POST.PLOCENE DISPLACEMENT ON FAULTS NUREG/CR-4829 V02: SHIPPING CONTAINER RESPONSE TO WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF EAST CEN- SEVERE HIGHWAY AND RAILWAY ACCIDENT TRAL KENTUCKY. CONDITIONS. Appendices.

Persond Auth:r indsx 33 WREATHALL,J. NUREG/CR4826 V02- SEISMIC MARGIN REVIEW OF THE MAINE YANKEE ATOMIC POWER STATON. Volume 2. Systems Analysis.

NUREG-1150 DAF V2 FC: REACTOR RISK REFERENCE DOCUMENT.Appendcas A-1 Draft For Comment.

NUREG/CR4695: MATERIAL CONTROL AND ACCOUNTING (MC&A)

WUA LOSS DETECTON DURING TRANSITION PERIODS AND PROCESS NUREG/CR-3950 V03: FUEL PERFORMANCE ANNUAL REPORT FOR UPSET CONDITONS.

1985.

YOUhG.M.W.

WYANT,F.J. NUREG.1163: COORDINATON OF SAFETY RESEARCH FOR THE NUREG/CR4301: STATUS REPORT ON EQUIPMENT OUALIFICATON BABCOCK AND WILCOX INTEGRAL SYSTEM TEST PROGRAM.

ISSUES RESEARCH AND RESOLUTON.

YOUNG.M.Y.

YOUNG,J. NUREG/CR4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE NUREG/CR4550 V03: ANALYSIS OF CORE DAMAGE FREQUENCY AND STUCKOPEN SAFETY RELIEF VALVE CARRYOVER TESTS.MB-2 Steam Generator Transent Response Test Program.

FROM INTERNAL EVENTS SURRY UNIT 1.

I t

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

See Loop Test Feeluty NUREG/CR4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY NUREG/CR-4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE FROM INTERNAL EVENTS. PEACH BOTTOM UNIT 2.

PARK (UMCP) 2X4 LOOP TEST FACILITY. Annual Report For 1985. NUREG/CR-4742: MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' AC-CfDENT SEOUENCE.

Ates Stoef NUREG/CR-4818: TRANSITION RANGE DROP TOWER J-R CURVE Accidental Rotenes TESTING OF A106 STEEL NUREG/CR-4825: A PRELIMINARY EVALUATION OF THE ECONOM6C RISK FOR CLEANUP OF NUCLEAR MATERIAL LICENSEE CONTAMI.

EG/CR-3469 V03: OCCUPATIONAL DOSE REDUCTION AT NU-CLEAR POWER PLANTS. Annotated Beliography Of Selected Reed. Accuenc Emboelen/ Flow RM ings in Radston Protection And ALARA. NUREG/CR-4300 V03 N2: ACOUSTIC EMISSION / FLAW RELATION-SHIP FOR INSERVICE MONITORING OF NUCLEAR PRESSURE VESSELS.Progrees Rept.Apni-September 1986.

EG/CR-4689- THERMAL-HYDRAULIC AND CHARACTERISTIC MODELS FOR PACKED DEBRIS BEDS- Adirondock Mountaine NUREG/CR-3232- DETAILED STUDIES OF SELECTED.WELL EXPOSED ASME Code FRACTURE ZONES IN THE ADIRONDACK MOUNTAINS DOME,NEW NUREG/CR-4859: SEISMIC FRAGILITY TEST OF A 6. INCH DIAMETER YORK.

PIPE SYSTEM.

Aerosol Abnonnel Occurrences NURTG/CR-4736: COMBUSTION AEROSOLS FORMED DURING NUREG-0000 V09 N02: REPORT TO CONGRESS ON ABNORMAL BURNING OF RADIOACTIVELY CONTAMINATED MATERIALS - EX-OCCURRENCES.Apni-June 1986.

PERIMENTAL RESULTS.

Abstract Airborne Release NUREG-0304 VII N04: REGULATORY AND TECHNICAL REPORTS NUREG/CR-4736: COMBUSTION AEROSOLS FORMED DURING (ABSTRACT INDEX JOURNAL). Annual Compilation For 1986.

BURNING OF RADIOACTIVELY CONTAMINATED MATERIALS - EX-G-1250: REPORT ON THE ACCIDENT AT THE CHERNOBYL NU-CLEAR POWER STATION. Annual Report NUREG/CR 4689: THERMAL-HYDRAULIC AND CHARACTERISTIC NUREG 0975 V05: COMPILATION OF CONTRACT RESEARCH FOR MODELS FOR PACKED DEBRIS BEDS. THE MATERIALS BRANCH, DIVISION OF ENGINEERING SAFETY. Annual Rept For FY 1986.

w Atmospheric Diepersion NUREG 1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALOvennew And Sum- NUREG/CR-4820: COMPARISON OF THE 1982 SEADEX DISPERSION DATA WITH RESULTS FROM A NUMBER OF DiFFERENT MODELS.

NUEG2 02 OT PROGRAM.NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALSevere Reactor Ac- ggg NUR SH DOWN Y EA REMOVAL ANALYSIS OF NU -1 $: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALResponse Of Li- NUREG/CR-4610: EFFECTS OF LATER'AL SE TION OF OXIDIC AND METALLIC CORE DEBRIS ON THE BWR MK I CONTAINMENT NU 12 i OT OG d'NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALPublic Protective NURE THERMAL-HYDRAULIC AND CHARACTERISTIC E REACTOR ACCl, MODELS FOR PACKED DEBR:S BEDS.

NU 21 05 T PR A NUREG/CR-4696: CONTAINMENT VENTING ANALYSIS FOR THE DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu- PEACH BOTTOM ATOMIC POWER STATION.

Rem ad=* And Wilcou g NUREG/CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF i

i NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITIONS. Main Study.

gg NUREG/CR4829 V02- SHIPPING CONTAINER RESPONSE TO

^ CR-3469 V03: OCCUPATIONAL DOSE REDUCTION AT NU-TIONS. CLEAR POWER PLANTS. Annotated Beliography Of Selected Reed-ings in Radiation Protection And ALARA.

Accident Scenerlo NUREG/CR-4803: THE POSSIBILITY OF LOCAL DETONATIONS DU GRA E ACCIDENTS IN THE BELLEFONTE NU-CYCLE WATER SYSTEMS. Apphcation Of Biofouhng Surveillance And Control Techniques To Sediment And Corrosion Fouling At Nuclear Accident Sequence Power Plants.

NUREG/CR-4320: THE RELATIONSHIP AND INFLUENCES OF FUEL AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI- Boiling Water Reector DENTS. NUREG/CR-4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR-4550 V03: ANALYSIS OF CORE DAMAGE FREQUENCY A GENERAL ELECTRIC BWR3/ MARK LCase Study.

FROM INTERNAL EVENTS.SURRY UNIT 1.

35

36 $UNGCt lfHfGX NUREG/CR-4610 EFFECTS OF LATERAL SEPARATION OF OX1DIC Containment Feliure AND METALUC CORE DEBRIS ON THE BWR MK 4 CONTAINMENT DRYWELL FLOOR. NUREG/CR-4552- A REVIEW OF THE SEABROOK STATION PROBABI-NUREG/CR 4889- THERMAL-HYDRAUUC AND CHARACTERtSTIC USTIC SAFETY ASSESSMENT. Containment Fedure Modes And Radi-clogical Source Terms.

MODELS FOR PACKED DEBRIS BEDS.

NUREG/CR-4698: CONTAlNMENT VENTING ANALYSIS FOR THE Containment integrity PEACH BOTTOM ATOMIC POWER STATION. NUREG/CR-3412 V02: CONTAINMENT INTEGRITY PROGRAM. Progress Report, April 1983 -December 1984.

NUREG/CR-4793: RESULTS OF SEMISCALE MOD 2C SMALL-BREAK Containment Penetration System LOSS-OF COOLANT ACCIDENT WITHOUT HPl (S-NH) EXPERIMENT SERIES. NUREG/CR 4734: SEISMIC TESTING OF TYPICAL CONTAINMENT PIPING PENETRATION SYSTEMS-Budget Containment Perfornience Design Oedoctive NUREG-1100 V03: BUDGET ESTIMATESFescal Years 1988-1989.

NUREG/CP-0084: PROCEEDINGS OF THE WORKSHOP ON A CON-Burnin8 TAINMENT PERFORMANCE DESIGN OBJECTIVE,MAY 12-

% 13,1986, HARPERS FERRY, WEST VIRGINIA.

NUREG/CR-4736. COMBUSTON AEROSOLS FORMED DURING BURNING OF RADIOACTIVELY CONTAMINATED MATERIALS - EX. Containment Venting PERIMENTAL RESULTS.

NUREG/CR-4696: CONTAINMENT VENTING ANALYSIS FOR THE CPITiti PEACH BOTTOM ATOMIC POWER STATON.

NUREG/CR-4737: INTERPRETATIVE ANALYSIS OF DATA FOR Contamination SOLUTE TRANSPORT IN THE UNSATURATED ZONE.

NUREG-1243: GROUND WATER PROTECTION ACTIVITIES OF THE Certson Steele U.S. NUCLEAR REGULATORY COMMISSION.

NUREG/CR-3861: STRESS-CORROSION CRACKING OF LOW. Conteminetton incident STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE NUREG/CR4825: A PRELIMINARY EVALUATION OF THE ECONOMIC REPOSITORY ENVIRONMENTS.

RISK FOR CLEANUP OF NUCLEAR MATERIAL LICENSEE CONTAMI.

NATON INCIDENTS.

Cent Steinises Steel NUREG/CR4744 V01 N1: LONG-TERM EMBRITTLEMENT OF CAST Cooient System Process DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Sermannual NUREG/CR-4320 THE RELATONSHIP AND INFLUENCES OF FUEL Report, October 1985 March 1986.

AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCl-DENTS.

Chernespyl NU -2 RE T THE ACCIDENT AT THE CHERNOBYL NU-UR -4550 V03: ANALYSIS OF CORE DAMAGE FREOUENCY Cloenup Coog FROM INTERNAL EVENTS:SURRY UNIT 1.

NUREG/CR-4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY NUREG/CR-4825: A PREUMINARY EVALUATION OF THE ECONOMIC FROM INTERNAL EVENTS PEACH BOTTOM UNIT 2. )

RISK FOR CLEANUP OF NUCLEAR MATERIAL UCENSEE CONTAMI- NUREG/CR-4741: FEEDWATER TRANSIENT AND SMALL BREAK NATON INCIDENTS. LOSS OF COOLANT ACCIDENT ANALYSES FOR THE BELLEFONTE NUCLEAR PLANT.

Climatology NUREG CR I CUMATOLOGY OF EXTREME WINDS IN SOUTH.

UR CR-4610: EFFECTS OF LATERAL SEPARATION OF OXIDIC Closoout AND METALUC CORE DEBRIS ON THE BWR MK I CONTAINMENT DRYWELL FLOOR.

NUREG/CR 4524: CLOSEOUT OF IE BULLETIN 80-24. PREVENTION C7 DAMAGE DUE TO WATER LEAKAGE INSIDE CONTAINMENT Core MWt l (OCTOBER 17,1980 INDIAN POINT 2 EVENT). NUREG-1250: REPORT ON THE ACCOENT AT THE CHERNOBYL NU-CLEAR POWER STATION. '

Cold Lge NUREG/CR-4552: A REVIEW OF THE SEABROOK STATION PROBABI-NUREG/lA-0004: THERMAL MIXING TESTS IN A SEM! ANNULAR USTIC SAFETY ASSESSMENT. Containment Failure Modes And Radi-DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS. Mai Source Tenns.

Comtpuetion Core Meltdown NUREG/CR 3468: HYDROGEN.AIRSTEAM FLAMMABluTY LIMITS NUREG-1150 DRF VI FC: REACTOR RISK REFERENCE AND COMBUSTION CHARACTERISTICS IN THE FITS VESSEL NUREG/CR-4736: COMBUSTON AEROSOLS FORMED DURING NU -1 F FC R RISK REFERENCE BURNING OF RADIOACTIVELY CONTAMINATED MATERIALS - EX* DOCUMENT.Appendces A-l. Draft For Comment.

PERIMENTAL RESULTS. NUREG-1150 DRF V3 FC: REACTOR RISK REFERENCE DOCUMENT.Appereces J-0. Draft For Comment.

g_ . . . _ _

NUREG/CR4742: MELPROG.PWR/ MODI ANALYSIS OF A TMLB' AC.

NUREG2R-4787: CONFERENCE OF RADIATION CONTROL DIREC.

TOR'S INFORMATION FOR UCENSING LOW-LEVEL RADIOACTIVE Corrosion WASTE INCINERATORS AND COMPACTORS.

NUREG/CR-4626 V02: IMPROVING THE RELIABluTY OF OPEN-CYCLE WATER SYSTEMS Application Of Biofouling Surveillance And EG CR-4 1 EFFECTS OF LATERAL SEPARATION OF OXIDIC er Plan AND METALUC CORE DEBRIS ON THE BWR MK I CONTAINMENT DRYWELL FLOOR. Cracked Pipe i.

Conteinment NUREG/CR-e53: APPROXIMATE METHOOS FOR FRACTURE ANALY.

NUREG/CR4524: CLOSEOUT OF IE BULLETIN R0-24 PREVENTION SES OF THh00GH-WALL CRACKED PIPES.

OF DAMAGE DUE TO WATER LEAKAGE in JE CONTAINMENT Crttical Safety Function (OCTOBER 17,1980 INDIAN POINT 2 EVENT).

NUREG/CR-4797: PROGRESS REVIEWS OF SIX SAFETY PARAME-Containment Event Analyste TER DISPLAY SYSTEMS.

NUREG/CR-4700 V1 DRF: CONTAINMENT EVENT ANALYSIS FOR DEBRIS FOSTULATED SEVERE ACCIDENTS. Surry Power Station,Umt 1. Draft For Comment NUREG/CR-4689: THERMAL HYDRAUUC AND CHARACTERISTIC MODELS FOR PACKED DW nP t

Subject Index 37 Earthquake DOE NUREG/CR4735 VOI: EVALUATON AND COMPILATION OF DOE NUREG/CR-4776: RESPONSE OF SEISMIC CATEGORY I TANKS TO WASTE PACKAGE TEST DATA. B6 annual Report. December 1985 - EARTHOUAKE EXCITATON.

NUREG/CR4852: THE MEERS FAULT; TECTONIC ACTIVITY IN Juhr1988.

SOUTHWESTERN OKLAHOMA.

UREG/CH-3469 V03; OCCUPATIONAL DOSE REDUCTION AT NU-CLEAR POWER PLANTS. Anrotated Bebliography Of Selected Read-ings in Radiation Protecton And ALARA. Economic Risk NUREG/CR4825: A PRELIMINARY EVALUATION OF THE ECONOMIC Data Base RISK FOR CLEANUP OF NUCLEAR MATERIAL LICENSEE CONTAMI-NUREG/CR4409 V02: DATA BASE ON NUCLEAR POWER PLANT NATION INCIDENTS.

DOSE REDUCTION RESEARCH PROJECTS.

Embrtttlement Data Relation Methode NUREG/CR4744 V01 N1: LONG-TERM EMBRITTLEMENT OF CAST NUREG/CR-4531: AN INVESTIGATION OF INTEGRAL FACluTY SCAL- DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual ING AND DATA RELATION METHODS (INTEGRAL SYSTEM TEST Report. October 1985 March 1986.

PROGRAM).

Data-Based System Emergency NUREG/CR4856: FEASIBILITY STUDY ON A DATA-BASED SYSTEM NUREG/CR-4797: PROGRESS REVIEWS OF SIX SAFETY PARAME-FOR DECISIONS REGARDING OCCUPATIONAL RADtATION PRO- TER DISPLAY SYSTEMS.

TECTION MEASURES.

Emergency Response g NUREG-1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-NUREG/CR-4689; THERMAL-HYDRAULIC AND CHARACTERISTIC DENT INCIDENT RESPONSE TRAINING MANUALOvennew And Sum-MODELS FOR PACKED DEBRtS BEDS. mary Of Major Points.

NUREG-1210 V02: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-Decay Heat Removal DENT INCIDENT RESPONSE TRAINING MANUALSevere Reactor Ac-NUREG/CR4448. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF cadent Overview.

A GENERAL ELECTRIC BWR3/ MARK I Case Study.

NUREG 1210 V03: PILOT PROGRAM NRC SEVERE REACTOR ACCI-NUREG/CR-4458. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF DENT INCIDENT RESPONSE TRAINING MANUAL. Response Of Li-A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case consee And State And Local Othcials.

Study.

NUREG/CR 4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG-1210 V04: PILOT PROGRAM NRC SEVERE REACTOR ACCI-A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case DENT INCIDENT RESPONSE TRAINING MANUALPublic Protective Study Actions Predetermined Cntena And initial Actions.

NUREG/CR-4762; SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG 1210 V05: PILOT PROGRAM NRC SEVERE REACTOR ACCI-DENT INCIDENT RESPONSE TRAINING MANUALO.S. Nuclear Regu-A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTOR. Case Study. latory Commisson Response.

Decontamination Enhcoment AN NUREG/CR-3444 V04 THE IMPACT OF LWR DECONTAMINATIONS NUREG 0940 V05 N04: ENFORCEMENT ACTONS SIGNIFICANT AC-ON SOLIDIFICATON. WASTE DISPOSAL,AND ASSOCIATED OCCU- TIONS RESOLVED,Ouarterly Progress Report, October-December PATIONAL EXPOSURE. Annual Report, Fiscal Year 1986. 1986.

Degraded Core Accident Equipment Qualification NUREG/CR-4803 THE POSSIBILITY OF LOCAL DETONATIONS NUREG/CR-4301: STATUS REPORT ON EQUIPMENT OVALIFICATION DURING DEGRADED-CORE ACCIDENTS IN THE BELLEFONTE NU- ISSUES RESEARCH AND RESOLUTION.

CLEAR POWER PLANT, NUREG/CR-4734: SEISMIC TESTING OF TYPICAL CONTAINMENT Department of Energy PIPING PENETRATON SYSTEMS.

NUREG/CR4735 V01: EVALUATION AND COMPILATION OF DOE Erosion-Corrosion Mechanism WASTE PACKAGE TEST DATA. Biannual Report. December 1985 -

July 1985. NUREG/CR-4868: METALLURGICAL EVALUATION OF AN 18-INCH FEEDWATER UNE FAILURE AT THE SURRY UNIT 2 POWER STA-g '

NUREG/CR4803: THE POSSIBluTY OF LOCAL DETONATIONS DURING DEGRADED-CORE ACCIDENTS IN THE BELLEFONTE NU- Events CLEAR POWER PLANT. NUREG4525 R12: SAFEGUARDS

SUMMARY

EVENT UST (SSEL).

D6 gest EXPohtal Assessment NUREG4386 D04 R04. UNITED STATES NUCLEAR REGULATORY NUREG/CR-4541: EXPERIMENTAL ASSESSMENT OF THE SEAUNG COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July EFFECTIVENESS OF ROCK FRACTURE GROUTING.

1972 - June 1986.

Extreme Winde D6epersion Model NUREG/CR-4801: CUMATOLOGY OF EXTREME WINDS IN SOUTH-NUREG/CR4820: COMPARISON OF THE 1982 SEADEX DISPERSION ERN CAUFORNIA.

DATA W1TH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

FITS Vessel Displacement NUREG/CR4685: POST-PUCCENE DISPLACEMENT ON FAULTS NUREG/CR 3468: HYDROGEN AIR STEAM FLAMMABluTY UMITS WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF EAST CEN- AND COMBUSTION CHARACTERISTICS IN THE FITS VESSEL TRAL KENTUCKY.

FRAPCON-2 Dose Reduction Research NUREG/CR 4718. EXPERIMENTAL SUPPORT AND DEVELOPMENT OF NUREG/CR-4409 V02: DATA BASE ON NUCLEAR POWER PLANT SINGLE-ROD FUEL CODES PROGRAM. Summary Report.

DOSE REDUCTION RESEARCH PROJECTS.

Fatigue Crack Growth Rate Drop Tower NUREG/CR4724: FATIGUE CRACK GROWTH RATES IN PRESSURE NUREG/CR-4818: TRANSITON RANGE DROP TOWER J-R CURVE VESSEL AND PIPING STEELS IN LWR ENVIRONMENTS. Final Report.

TESTING OF A106 STEEL Faults EVNTRE NUREG/CR-4685: POST PLIOCENE DISPLACEMENT ON FAULTS NUREG/CR-4700 V1 DRF: CONTAINMENT EVENT ANALYSIS FOR WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF EAST-CEN-POSTULATED SEVEFe ACCIDENTS. Surry Power Station, Unit 1. Draft TRAL KENTUCKY.

For Comment.

38 Subject index Penalbility Ground Water Protection NUREG/CR-4856: FEASIBluTY STUDY ON A DATA-BASED SYSTEM FOR DECISIONS REGARDING OCCUPATIONAL RADIATON PRO- NUREG-1243: GROUND WATER PROTECTON ACTIVITIES OF THE TECTION MEASURES. U.S. NUCLEAR REGULATORY COMMISSLON.

g Grout Testing NUREG/CR-4888: METALLURGICAL EVALUATION OF AN 18-INCH NUREG/CR-4541: EXPERIMENTAL ASSESSMEN1 OF THE SEAUNG FEEDWATER UNE FAILURE AT THE SURRY UNIT 2 POWER STA- EFFECTIVENESS OF ROCK FRACTURE GROUTING.

TON.

Heaardous Weste Feedwater Trenaient NUREG/CR-4730 EVALUATION OF POTENTIAL MIXED WASTES CON-NUREG/CR4741: FEEDWATER TRANSIENT AND SMALL BREAK TAINING LEAD, CHROMlUM,USED OIL.OR ORGANIC LIQUIDS.

LOSS OF COOLANT ACCIDENT ANALYSES FOR THE BELLEFONTE NUCLEAR PLANT. Hoevy Section Steel Technology Veseel NUREG/CR-4711: LOW UPPER-SHELF TOUGHNESS HIGH-TRANSI-R 1100 V03. BUDGET ESTIMATESFescal Years 1988-1989, "

WIDE-PLATE TEST SPECIMENS. Final Report.

NUREG/CR-3468: HYDROGEN AIR. STEAM FLAMMABluTY LIMITS High Pressure injection AND COMBUSTION CHARACTERISTICS IN THE FITS VESSEL NUREG/lA4004: THERMAL MIXING TESTS IN A SEMIANNULAR Fracture Anotyees DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS.

NUREG/CR-4853: APPROXIMATE METHODS FOR FRACTURE ANALY- High Level Nucieer Weste SES OF THROUGH-WALL CRACKED PIPES-NUREG/CR-4735 VO1: EVALUATON AND COMPILATON OF DOE Fracture % WASTE PACKAGE TEST DATA. Biannual Report. December 1985 -

NUREG/CR 4541: EXPERIMENTAL ASSESSMENT OF THE SEALING July 1986.

EFFECTIVENESS OF ROCK FRACTURE GROUTING.

R eW RWM Fracture Toughnees NUREG/CR-4708 V01 N1: PROGRESS IN EVALUATION OF RADONU-NUREG/CR-4491: DEVELOPMENT OF MODELS FOR WARM PRES- CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-THESSING. LEVEL NUCLEAR WASTE REPOSITORY SITE NUREG/CR-4818: TRANSITON RANGE DROP TOWER J-R CURVE PROJECTS.Senwannual Report For October 1985. March 1986.

TESTING OF A106 STEEL Fracture Zone ' ' * * #" "*"**"I NOREG/CR-3232. DETAILED STUDIES OF SELECTED,WELL EXPOSED NUREG/CR-3861: STRESS-CORROSON CRACKING OF LOW-FRACTURE ZONES IN THE ADlRONDACK MOUNTAINS DOME,NEW STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE yong, REPOSITORY ENVIRONMENTS.

Fragety Highway And Railway Accident Conditions NUREG/CR4826 V01: SEISMIC MARGIN REVIEW OF THE MAINE NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO YANKEE ATOMO POWER STATON Volume t. Summary Report.

SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITONS. Main !

NUREG/CR-4826 V02: SEISMO MARGIN REVIEW OF THE MAINE Report.

YANKEE ATOMO POWER STATION Volume 2. Systems Analysis. i NUREG/CR-4826 V03: SELSMIC MARGIN REVICW OF THE MAINE NUREG/CR-4829 V02: SHIPAING CONTAINER RESPONSE TO l SEVERE HIGHWAY AND RAILWAY ACCOENT NtiE / -48 : SE F Cl TY T A CH ETER CONDITIONS. Appendices.

PIPE SYSTEM.

Human Error 7. 7, Fuel NUREG/CR-4018 V02: APPLICATON OF SUM-MAUD A TEST OF AN NUREG/CR-4320 THE RELATONSHIP AND INF6UENCES OF FUEL INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZLNG AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI- EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND DENTS. RELIA 81UTY. Volume ll: Appendices.

Fuel Damage Human Factore NUREG/CR-4320 THE RELATONSHIP AND INFLUENCES C* FUEL AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI. NUREG/CR 3968: STUDY OF OPERATING PROCEDURES IN NUCLE.

DENTS. AR POWER PLANTS. Practices And Problems.

Fuel Performance Hydrogen.Alr:Stoem Mixture NUREG/CR-3950 V03: FUEL PERFORMANCE ANNUAL REPORT FOR' NUREG/CR-3468: HYDROGEN AIR. STEAM FLAMMABluTY UMITS 1985. AND COMBUSTON CHARACTERISTICS IN THE FITS VESSEL NUREG/CR-4718: EXPERIMENTAL SUPPORT AND DEVELOPMENT OF SINGLE-ROD FUEL CODES PROGRAM. Summary Report. IE Sulletin 80-24 NUREG/CR-4524: CLOSEOUT OF IE BULLETIN 80w24. PREVENTION 280- AN F MAT AND CONTENT ACCEPTANCE DME NE M WATER WW WSOE NAMM CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNTING OBER @80 WM MT 2 WEW (MC&A) REFORM AMENDMENT.10 CFR Part 74 Subpart E.

Incident Reeponse General Electric BWR3 NUREG 1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-NUREG/CR-4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF DENT INCIDENT RESPONSE TRAINING MANUALOvennew And Sum.

A GENERAL ELECTRO BWR3/ MARK l. Case Study. mary Of Maior Pomts.

NUREG-1210 V02: PILOT PROGRAM NRC SEVERE REACTOR ACCI-Generic SeMy leeue DENT INCIDENT RESPONSE TRAINING MANUALSevero Reactor Ac-NUREG4933 S06: A PRORITIZATION OF GENERO SAFETY ISSUES. cident Owennew.

,,gg NUREG-1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG/CR 4708 V01 N1: PROGRESS IN EVALUATON OF RADIONU- DENT INCOENT RESPONSE TRAINING MANUALResponse Of U-CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH- consee And State And Local Omcials.

LEVEL NUCLEAR WASTE REPOSITORY SITE NUREG 1210 V04. PILOT PROGRAM.NRC SEVERE REACTOR ACCl-PROJECTS.Sermannual Report For October 1985 March 1986. DENT INCIDENT RESPONSE TRAINING MANUALPubhc Protectus Actions Predetermoed Cntena And Irvtial Actons.

Graphite Fire NUREG 1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG 1250 REPORT ON THE ACCIDENT AT THE CHERNOBYL NU- DENT INCOENT RESPONSE TRAINING MANUALU.S. Nuclear Regu-CLEAR POWER STATON. latory Comrmssion Response.

Subject index 39 LLRWPAA incinerat:on NUREG/CP4085: MEETING WITH STATES ON THE LOW-LEVEL RA-NUREG/CR-4787: CONFERENCE OF RADtATION CONTROL DIREC- DOACTNE WASTE POUCY AMENDMENTS ACT (LLRWPAA) OF TOR'S INFORMATON FOR UCENSING LOW-LEVEL RADICACTNE 1985.

WASTE INCINERATORS AND COMPACTORS.

LWR INTEGRITY Indes NUREG/CR-3412 V02: CONTAJNMENT NUREG-0304 V11 N04: REGULATORY AND TECHNICAL REPORTS PROGRAM. Progress Report.Apnl 1983 -December 1984.

(ABSTRACT INDEX JOURNAL). Annual Compilabon For 1986. NUREG/CR-3444 V04: THE IMPACT OF LWR DECONTAMINATONS ON SOUDIFICATION. WASTE DISPOSAL,AND ASSOCIATED OCCV-Inflitration PATIONAL EXPOSURE. Annual Report, Fiscal Year 1986.

NUREG/CR-2478 V03: A STUDY OF TRENCH COVERS TO MINIM 12E NUREG/CR-4320: THE RELATIONSHIP AND INFLUENCES OF FUEL INFILTRATION AT WASTE DISPOSAL SITES. Final Rept AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI-Inseerice inspection DENTS.

NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) RELI-NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) REU- ABluTY FOR INSERVICE INSPECTION OF UGHT WATER ABluTY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS.Semannual R October 1985 - March 1986.

REACTORS Sermannual Report,0ctober 1985 March 1986. USES OF COMPONENT FAILURES NUREG/CR 4616: ROOT NUREG/CR4712: REGULATORY ANALYSIS OF REGULATORY GUIDE PROGRAM Methods And Apphcations.

1.35 (REVISION 3. DRAFT 2) - IN-SERV)CE INSPECTION OF UN- NUREG/CR4724: FATIGUE CRACK GROWTH RATES IN PRESSURE GROUTED TENDONS IN PRESTRESSED CONCRETE CONTAIN- VESSEL AND PIPING STEELS IN LWR ENVIRONMENTS Final Report.

MENTS. NUREG/CR4734: SEISMIC TESTING OF TYPICAL CONTAINMENT PIPING PENETRATON SYSTEMS.

inservice Monitoring NUREG/CR4744 V01 N1: LONG TERM EMBRITTLEMENT OF CAST NUREG/CR 4300 V03 N2- ACOUSTIC EMISSION / FLAW RELATON- DUPLEX STAINLESS STEELS IN LWR SYSTEMS Sermannual SHIP FOR INSERVICE MONITORING OF NUCLEAR PRESSURE Report.Octobnr 1985 March 1986.

VESSELS Progress Rept Apni-September 1986. NUREG/CR-4813: ASSESSMENT OF LEAK DETECTION SYSTEMS FOR LWRs. October 1985 - September 1986.

Inspection NUREG/CR4824: EVALUATION OF INTEGRAL CONTINUING EXPERJ-NUREG4040 V10 N04: UCENSEE CONTRACTOR AND VENDOR IN. MENTAL CAPABluTY (CEC) CONCEPTS FOR UGHT WATER REAC-SPECTION STATUS REPORT. Quarterfy Report, October-December TOR RESEARCH PWR SCAUNG CONCEPTS.

1986 (White Book)

Lateral Separation Instrument Tube Rupture NUREG/CR4610: EFFECTS OF LATERAL SEPARATION OF OXtDIC NUREG/CR-4672: ANALYSIS OF INSTRUMENT TUBE RUPTURES IN AND METALUC CORE DEBRIS ON THE BWR MK I CONTAINMENT WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS. DRYWELL FLOOR.

Integral Facility Scaling Leak Detection NUREG/CR4531: AN INVESTIGATION OF INTEGRAL FACluTY SCAL- NUREG/CR4813: ASSESSMENT OF LEAK DETECTION SYSTEMS ING AND DATA RELATON METHODS (INTEGRAL SYSTEM TEST FOR LWRs. October 1985 - September 1986.

PROGRAM).

LegalIseuancee Integral System Test Program NUREG-0750 V24 N01: NUCLEAR REGULATORY COMMISSION IS-NUREG-1163: COORDINATION OF SAFETY RESEARCH FOR THE SUANCES FOR JULY 1986 Pages 1 195.

BABCOCK AND WILCOX INTEGRAL SYSTEM TEST PROGRAM. NUREG-0750 V24 NO2: NUCLEAR REGULATORY COMMISSION IS-NUREG/CR-4531: AN INVESTIGATON OF INTEGRAL FACluTY SCAL-SUANCES FOR AUGUST 1986.Pages 197-396.

ING AND DATA RELATION METHODS (INTEGRAL SYSTEM TEST NUREG-0750 V24 NO3: NUCLEAR REGULATORY COMMISSION IS-PROGRAM). SUANCES FOR SEPTEMBER 1986.Pages 397 488.

Ucense Application NUREG/CR4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY NUREG-1199: STANDARD FORMAT AND CONTENT OF A UCENSE FROM INTERNAL EVENTS PEACH BOTTOM UNIT 2. APPUCATION FOR A LOW-LEVEL P ADtOACTIVE WASTE DISPOSAL FACIUTY, intruder Dose Pathway Analysis NUREG-1200: STANDARD REVIEW PLAN FOR THE REVIEW OF A Ll-NUREG/CR-3620 S02: INTRUDER DOSE PATHWAY ANALYSIS FORCENSE APPUCATION FOR A LOW-LEVEL RADIOACTIVE WASTE THE ONSTE DISPOSAL OF RAOIOACTIVE WASTES.h ONSITE/ DISPOSAL FACluTY.

MAXII Computer Program.

Ucensed Fuel Facility Inventory Difference Data NURE04430 V07 N01: UCENSED FUEL FACluTY STATUS NUREG4430 V07 N01: UCENSED FUEL FACluTY STATUS REPORT. inventory Difference Data. January-June 1986.(Gray Book II)

REPORT. inventory Difference Data. January June 1986.(Gray Book it)

Licensed Operating Reactors tron Oxide Aerosol NUREG-0020 V10 N09: UCENSED OPERATING REACTORS STATUS NUREG/CR-4798: IRON OX10E AEROSOL EXPERIMENTS IN STEAM-

SUMMARY

REPORT. Data As Of August 31.1986.(Gray Book I)

AIR ATMOSPHERES NSPP TESTS 501505 AND 511, DATA RECORD NUREG4020 V10 N10: LICENSED OPERATING REACTORS STATUS REPORT.

SUMMARY

REPORT. Data As Of September 30,1986.(Gray Book f)

J-integral Ucensee Event Report NUREG/CR4853: APPROXIMATE METHODS FOR FRACTURE ANALY. NUREG/CR 2000 V05N12: UCENSEE EVENT REPORT (LER)

SES OF THROUGH WALL CRACKED PIPES. COMPILATION For Month Of December 1986.

NUREG/CR-2000 V06 N1: LICENSEE EVENT REPORT (LER)

J-R Curve Testing COMPILATION.For Month Of January 1987.

NUREG/CR 4818: TRANSITION RANGE DROP TOWER J-R CURVE TESTING OF A106 STEEL Ucenalng NUREG/CR-4787: CONFERENCE OF RAC1ATION CONTROL DIREC-Kentucky River Fault System TOR'S INFORMATION FOR UCENSING LOW-LEVEL RADIOACTIVE NUREG/CR.4685: POST-PUOCENE DISPLACEMENT ON FAULTS WASTE INCINERATORS AND COMPACTORS.

WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF EASTCEN-NUREG/CR-4856: FEASIBluTY STUDY ON A DATA. BASED SYSTEM TRAL KENTUCKY. FOR DECISIONS REGARDING OCCUPATIONAL RADfATION PRO-TECTON MEASURES.

LER NUREG/CR-2000 V05N12: UCENSEE EVENT REPORT (LER) U9ht Water Reactor INTEGRITY COMPILATON For Month Of December 1986. NUREG/CR-3412 V02: CONTAINMENT NUREG/CR 2000 V06 N1: UCENSEE EVENT REPORT (LER) PROGRAM. Progress Report. April 1983 -December 1984.

COMPILATON.For Month Of January 1987.

40 Subject Index NUREG/CR-3444 V04: THE IMPACT OF LWR DECONTAMINATIONS ON SOUDIFICATON, WASTE DISPOSAL,AND ASSOCIATED OCCU-i W Failure Analysie PATIONAL EXPOSURE. Annual Report. Flecal Year 1988. NUREG/CR 4868: METAU URGICAL EVALUATON OF AN 18-INCH NUREG/CR-4320 THE RELATIONSHIP AND INFLUENCES OF FUEL FEEDWATER LINE FAILURE AT THE SURRY UNIT 2 POWER STA.

TION.

AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI-DENTS.

Mixed Weste NUREG/CR4489 V04: NONDESTRUCTIVE EXAMINATION (NDE) REU-A81UTY FOR INSERVICE INSPECTION OF UGHT WATER NUREG/CR-4730 EVALUATON OF POTENTIAL MIXED WASTES CON-REACTORS Sermannual R October 1985 - March 1986. TAINING LE%D, CHROMlUM,USED OIL,OR ORGANIC UOUlDS.

l NUREG/CR4618. ROOT USES OF COMPONENT FAILURES PROGRAMMethods And ~ tions. Model tvolustion NUREG/CR-4724: FATIGUE K GROWTH RATES IN PRESSURE NUREG/CR-4820: COMPARISON OF THE 1982 SEADEX DISPERSION VESSEL AND PIPING STEELS IN LWR ENVIRONMENTS Final Report. DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

NURE 2/CR-4734: SEISMIC TESTING OF TYPtCAL CONTAINMENT PIPING PENETRATON SYSTEMS. Monitoring NUREG/CR4744 V01 N1: LONG-TERM EMBRITTLEMENT OF CAST NUREG-1243: GROUND-WATER PROTECTON ACTIVITIES OF THE DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Sermannual U.S. NUCLEAR REGULATORY COMMISSION.

Report. October 1985 - March 1988.

NUREG/CR-4813: ASSESSMENT OF LEAK DETECTION SYSTEMS Multi Attribute Utility C: n , . __

FOR LWRs. October 1985 September 1988. NUREG/CR-4018 V02: APPUCATON OF SUM-MAUD:A TEST OF AN NUREG/CR-4824: EVALUATON OF INTEGRAL CONTINUING EXPERI- INTERACTIVE COMPUTER-BASED METHOD FOR ORGANtZLNG MENTAL CAPABILITY (CEC) CONCEPTS FOR UGHT WATER REAC. EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND TOR RESEARCH PWR SCAUNG CONCEPTS. REUABluTY. Volume LLAppendices.

Loadinge NDE Rollability NUREG/CR-3412 V02: CONTAINMENT INTEGRITY PROGRAM. Progress Report.Apnt 1983 -December 1984. NUREG/CR4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) REll-ABlUTY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS.Sermannual Report. October 1985 - March 1988.

Local Detonetton NUREG/CR-4803: THE POSSIBluTY OF LOCAL DETONATIONS WCMPE DURING DEGRADED. CORE ACCIDENTS IN THE BELLEFONTE NU- NUREG/CR-4853: APPROXIMATE METHODS FOR FRACTURE ANALY.

CLEAR POWER PLANT. SES OF THROUGH-WALL CRACKED PIPES.

Loos Degoction Natural Circulation NUREG/CH-4695: MATERIAL CONTROL AND ACCOUNTING (MCAA) NUREG/CA-4742: MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' AC- j LOSS DETECTION DURING TRANSITON PERIODS AND PROCESS CIDENT SEQUENCE. '

UPSET CONDITIONS.

NUREG/CR-4788: AN ANALYTICAL AND EXPERIMENTAL INVESTIGA-Low Temperature Toughnees TION OF NATURAL CIRCULATON TRANSIENTS IN A MODEL PRES-SURIZED WATER REACTOR.

NUREG/CR4491: DEVELOPMENT OF MODELS FOR WARM PRES- NUREG/CR-4789 THE SIMULATON OF THERMOHYDRAUUC PHE.

TRESSING.

NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP.

Low Upper-Shelf Toughnees Mondestructive Examination NUREG/CR-4711: LOW UPPER 4HELF TOUGHNESS,HIGH-TRANSI.

NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMINATION (NDE) REll-TON TEMPERATURE TEST INSERT IN HSST PTSE 2 VESSEL AND WIDE-PLATE TEST SPECIMENS. Final Report. ABlUTY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS.Sermannual Report. October 1985 March 1986.

Low-Level Medloactive Weste Nuclear Reactor Preneure Boundary NUREG/CA-4787: CONFERENCE OF RADIATON CONTROL DIREC-TOR'S INFORMATION FOR UCENSING LOW-LEVEL RADIOACTIVE NUREG/CR-4300 V03 N2: ACOUSTIC EMISSION / FLAW RELATION-CASTE INCINERATORS AND COMPACTORS. SHIP FOR INSERVICE MONITORING OF NUCLEAR PRESSURE VESSELS.Progrese Rept, April-Septernber 1986 Low-Level Radioactive Weste Policy Act Nuclear Safety Pilot Plant NUREG/CP-0085: MEETING WITH STATES ON THE LOW LEVEL RA*

DOACTIVE WASTE POUCY AMENDMENTS ACT (LLRWPAA) OF NUREG/CR-4798: IRON OXIDE AEROSOL EXPERIMENTS IN STEAM-1985. AIR ATMOSPHERES.NSPP TESTS 501-505 AND 511, DATA RECORD REPORT.

Low-Level Weete Disposal Factitty ONSITE/ MAX 11 NUREG-1199: STANDARD FORMAT AND CONTENT OF A LICENSE APPUCATON FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL NUREG/CR-3820 S02: INTRUDER DOSE PATHWAY ANALYSIS FOR THE ONSITE DISPOSAL OF RADCACTIVE WASTES.The ONSITE/ -

NUREG 1200: STANDARD REVIEW PLAN FOR THE REVIEW OF A U- Computer Warn.

CENSE APPUCATION FOR A LOW-LEVEL RADIOACTIVE WASTE n~.:

DISPOSAL FACIUTY. M Does NUREG/CR-3469 V03: OCCUPATONAL DOSE REDUCTION AT NU.

MELPROG-PWR/ MODI CLEAR POWER PLANTS. Annotated Bibbography Of Selected Read-NU EG 2 MELPROG-PWR/ MODI ANALYSIS OF A TMLB' AC-NU /CR V02: A AS ON NUCLEAR POWER PLANT DOSE REDUCTON RESEARCH PROJECTS.

Mark i NUREG/CR4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF Occupational Exposure A GENERAL ELECTRIC BWR3/ MARK l. Case Study. NUREG/CR-3444 V04: THE IMPACT OF LWR DECONTAMINATIONS ON SOUDIFICATON. WASTE DISPOSAL,AND ASSOCIATED OCCU-MeterialControl And Accounting PATONAL EXPOSURE. Annual Report, Fiscal Year 1988.

NUREG 1280 STANDARD FORMAT AND CONTENT ACCEPTANCE CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNTING Occupational Radiation (MC&A) REFORM AMENDMENT.10 CFR Part 74 Subpart E. NUREG/CR4858. FEASIBILITY STUDY ON A DATA-BASED SYSTEM NUREG/CR-4695: MATERIAL CONTROL AND ACCOUNTING (MC&A) FOR DECISIONS REGARDING OCCUPATIONAL RADIATION PRO-LOSS DETECTON DURING TRANSITION PERIODS AND PROCESS TECTION MEASURES.

UPSET CONDITONS.

Oneite Disposal Meers Fault NUREG-1101 V02-NUREG/CR4852: THE MEERS FAULT. TECTONIC ACTIVITY IN ONSITE DISPOSAL OF RADIOACTIVE WASTE. Methodology For The Radsological Assessment Of Disposal By SOUTHWESTERN OKLAHOMA. Subsurface Burial.

Subject index 41 Onsite Radioactive Waste Disposal Practice And Procedure Digest NUREG/CR-3620 S02: tNTRUDER DOSE PATHWAY ANALYSIS FOR NUREG-0386 D04 RG4: UNITED STATES NUCLEAR REGULATORY THE ONSITE DISPOSAL OF RAOlOACTIVE WASTES The ONSITE/ COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. July MAX 11 Computer Program- 1972 - June 1986.

Opegycle Water System Pressure Vessel NUREG/CR-4626 V02. IMPROVING THE RELIABlUTY OF OPE N- NUREG/CR-4711: LOW UPPER-SHELF TOUGHNESS.HIGH-TRANSI-CYCLE WATER SYSTEMS. Apphcaton Of Bx)fouhng Surveillance And TlON TEMPERATURE TEST INSERT IN HSST PTSE-2 VESSEL AND Control Techniques To Sedunent And Corrosx)n Fouhng At Nuclear WIDE-PLATE TEST SPECIMENS. Final Report.

Power Plants.

Pressure Vessel And Piping Steel Operating Procedures NUREG/CR-4724: FATIGUE CRACK GROWTH RATES IN PRESSURE NUREG/CR-3968: STUDY OF OPERATING PROCEDURES IN NUCLE- VESSEL AND PIPING STEELS IN LWR ENVIRONMENTS. Final Report AR POWER PLANTS. Practices And Problems.

NUREG/CR4613. EVALUATON OF NUCLEAR POWER PLANT OPER- Pressurized Thermal Shock ATING PROCEDURES CLASSIFICATIONS AND NUREG/CR-4491: DEVELOPMENT OF MODELS FOR WARM PRES-INTERFACES Problems And Techruques For improvement' TRESS!NG.

NUREG/CR 4711: LOW UPPER-SHELF TOUGHNESS.HIGH-TRANSI-Organization Chart TION TEMPERATURE TEST INSERT IN HSST PTSE 2 VESSEL AND NUREG-0325 R10: U.S. NUCLEAR REGULATORY COMMISSION FUNC- WIDE-PLATE TEST SPECIMENS. Final Report TONAL ORGANIZATION CHARTS ~ NUREG/lA4004: THERMAL MIXING TESTS IN A SEMIANNULAR DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS.

PWR NUREG/CR4458. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF se ed a a A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR Case A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case NUR /CR4551 V1 DAF: EVALUATION OF SEVERE ACCIDENT Study.

RISKS AND THE POTENTIAL FOR RISK REDUCTON SURRy POWER STATION, UNIT 1. Draft For Comment.

NUREG/CR4551 V1 DRF: EVALUATION OF SEVERE ACCIDENT RISKS AND THE POTENTIAL FOR RISK REDUCTION.SURRY NUREG/CR-4672: ANALYSIS OF INSTRUMENT TUBE RUPTURES IN POWER STATION. UNIT f. Draft For Comment.

WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS. NUREG/CR-4672: ANALYSIS OF INSTRUMENT TUBE RUPTURES IN NUREG/CR4700 VI DRF: CONTAINMENT EVENT ANALYSIS FOR WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS.

POSTULATED SEVERE ACCIDENTS. Surry Power Station.Urvt 1 Draft NUREG/CR4700 V1 DRF: CONTAINMENT EVENT ANALYSib FOR For Comment. POSTULATED SEVERE ACCIDENTS. Surry Power Staton,Urut 1. Draft NUREG/CR4713 SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case For Comment.

NUREG/CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF Study.

NUREG/CR4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case Study.

AND STUCK-OPEN SAFETY RELIEF VALVE CARRYOVER NUREG/CR4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE TESTS M8-2 Steam Generator Transent Response Test Program.

NUREG/CR-4762; SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF AND STUCK-OPEN SAFETY REUEF VALVE CARRYOVER TESTS MB-2 Steam Generator Transumt Response Test Program.

A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTOR Case NUREG/CR-4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF Study NUREu/CR4788. AN ANALYTICAL AND EXPERIMENTAL INVESTIGA- A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTOR. Case TION OF NATURAL CIRCULATION TRANSIENTS IN A MODEL PRES- Study.

SURIZED WATER REACTOR. NUREG/CR4788: AN ANALYTICAL AND EXPERIMENTAL INVESTIGA-NUREG/CR-4789: THE SIMULATION OF THERMOHYDRAULIC PHE- TION OF NATURAL CIRCULATION TRANSIENTS IN A MODEL PRES-NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP. SURIZED WATER REACTOR.

NUREG/CR 4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE NUREG/CR4789. THE SIMULATION OF THERMOHYEaAUUC PHE.

PARK (UMCP) 2X4 LOOP TEST F ACILITY. Annual Report For 1985. NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP.

NUREG/lA4004: THERMAL MIXING TESTS IN A SEMIANNULAR NUREG/CR-4843 V01: UNIVERSITY OF MARYLAND AT COLLEGE DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS. PARK (UMCP) 2X4 LOOP TEST FACILITY. Annual Report For 1985.

NUREG/lA4004: THERMAL MIXING TESTS IN A SEMIANNULAR E R-4541: EXPERIMENTAL ASSESSMENT OF THE SEAUNG EFFECTIVENESS OF ROCK FRACTURE GROUTING. Prestressed Concrete Containment NUREG/CR4712: REGULATORY ANALYSIS OF REGULATORY GUIDE EG 93 V 03 NRC REGULATORY AGENDA.Ouarterly us WM 1 DM 2) - ENG MW2 & N ED TENDONS IN PRESTRESSED CONCRETE CONTAIN-Report,Juh-September 1986. y Pilot Program Priorttization NUREG-1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCI. NUREG-0933 S06: A PRIORITIZATION OF GENERIC SAFETY ISSUES.

DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum-Probabilistic Risk Assessment NUREG 2 0 02 PI OT PROGRAM NRC SEVERE REACTOR ACCI- NUREG/CR-4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF DENT INCIDENT RESPONSE TRAINING MANUALSeveie Reactor Ac- A GENERAL ELECTRIC BWR3/ MARK 1 Case Study.

NUREG/CR-4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NU 1 3: PILOT PROGRAM NRC SEVERE REACTOR ACCl- A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTORCase DENT INCIDENT RESPONSE TRAINING MANUALResponse Of U-NU /CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NURE 121 iO N NRC SEVERE REACTOR ACCI. A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case DENT INCIDENT RESPONSE TRAINING MANUALPubhc Protective R N S E REACTOR ACCI. NUR /CR4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NU 21 5 i A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTOR. Case DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu.

latory Comnusson Response NU /CR-4824: EVALUATION OF INTEGRAL CONTINUING EXPERI-MENTAL CAPABluTY (CEC) CONCEPTS FOR LIGHT WAT ER REAC-Piping System TOR RESEARCH PWR SCALING CONCEPTS.

NUREG/CR4859 SEISMIC FRAGluTY TEST OF A 6-INCH DIAMETER Probaba!!stic Safety Assessment NUREG/CR4552: A REVIEW OF THE SEABROOK STATIOff PROBABI-Ptsstic Zone LISTIC SAFETY ASSESSMENT. Containment Failure Modes And Radi-NUREG/CR-4491: DEVELOPMENT OF MODELS FOR WARM PRES-ologeal Source Terms.

TRESSING.

42 Subject index proe dw.Cinessacetion Remonuendes NUREG/CR-4613: EVALUATION OF NUCLEAR POWER PLANT OPER- NUREG/CR-4708 V01 N1: PROGRESS IN EVALUATION OF RADIONU-ATING PROCEDURES CLASSIFICATIONS AND CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-INTERFACES Problems And Techncues For Improvement. LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS.Serniennual Report For October 1985 - March 1986.

NUREG/CR-4613: EVALUATION OF NUCLEAR POWER PLANT OPER- Reactor Cooient System ATING PROCEDURES CLASSIF8 CATIONS AND NUREG/CR4320t THE RELATIONSHIP AND INFLUENCES OF FUEL INTERFACES Prot 2 ems And Techniques For Improvement AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI-Procedures DENTS.

NUREG/CR-3968: STUDY OF OPERATING PROCEDURES IN NUCLE- Reector Riek AA POWER PLANTS. Practicos And Problems. NUREG-1150 DRF VI FC: REACTOR RISK REFERENCE Prefect Description DOCUMENT. Main ReportDraft For Comment.

NUREG-1150 DRF V2 FC- REACTOR RISK REFERENCE NUREG 1280 V01: A REPORT TO CONGRESS ON NUCLEAR REGULA- DOCUMENT. Appendices A-1. Draft For Comment TORY RESEARCH.Prolect Descnptions For FY87.

NUREG-1150 DRF V3 FC: REACTOR RISK REFERENCE DOCUMENTAppendees J-0. Draft For Comment NUREG 1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCI. Reactor Safety DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum- NUREG/CP-0082 V01 PROCEEDINGS OF THE FOURTEENTH WATER mery Of Malor Points.

REACTOR SAFETY INFORMATION MEETING.

NUREG-1210 V02 PILOT PROGRAM.NRC SEVERE REACTOR ACCI- NUREG/CP-0082 V02: PROCEEDINGS OF THE FOURTEENTH WATER DENT INCIDENT RESPONSE TRAINING MANUALSevero Reactor Ac- REACTOR SAFETY INFORMATION MEETING.

cident Overview. NUREG/CP-0082 V03: PROCEEDINGS OF THE FOURTEENTH WATER NUREG 1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCl- REACTOR SAFETY INFORMATION MEETING.

DENT INCIDENT RESPONSE TRAINING MANUALResponse Of U- NUREG/CP-0082 V04: PROCEEDNGS OF THE FOURTEENTH WATER conese And State And Local Officials. REACTOR SAFETY INFORMATION MEETING.

i l

NUREG 1210 V04: PILOT PROGRAM.NRC SEVERE REACTOR ACCl. NUREG/CP-0082 V05: PROCEEDINGS OF THE FOURTEENTH WATER DENT INCIDENT RESPONSE TRAINING MANUALPubic Protective REACTOR SAFETY INFORMATION MEETING.

Actons . Predetermined Cntena And Instial Actons. NUREG/CP4082 V06. PROCEEDNGS OF THE FOURTEENTH WATER NUREG 1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCI- REACTOR SAFETY INFORMATION MEETING.

DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu- NUREG/CR 4803: THE POSSIBluTY OF LOCAL DETONATIONS letory Commesson Response. DURING DEGRADED-CORE ACCIDENTS IN THE BELLEFONTE NU.

CLEAR POWER PLANT.

RELAPG/ MOO 2 NUREG/CR-4672 ANALYSIS OF INSTRUMENT TUBE RUPTURES IN Reactor Shutdown WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS. NUREG/CR-4012 V02: REPLACEMENT ENERGY COSTS FOR NUCLE-Redletion Esposure AR ELECTRICITY-GENERATING UNITS IN THE UNITED STATES:19871991.

NURE7/CR-4409 V02 DATA BASE ON NUCLEAR POWER PLANT DOSE REDUCTION RESEARCH PROJECTS. Recowy Procedure NUREG/CR-4793: RESULTS OF SEM! SCALE MOD-2C SMALL-BREAK Redletion Monitoring Network LOSSOF-COOLANT ACCOENT WITHOUT HPI (S NH) EXPERIMENT NUREG.0837 V06 NO3: NRC TLD DIRECT RADIATION MONITORING SERIES.

NETWORK. Progress Report. July-September 1986.

Radletten Protection NUREG 1280: STANDARD FORMAT AND CONTENT ACCEPTANCE NOREG/CR-3469 V03: OCCUPATONAL DOSE REDUCTION AT NU- CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNTING CLEAR POWER PLANTS. Annotated Babhography Of Selected Road. (MC&A) REFORM AMENDMENT.10 CFR Part 74 Subpart E.

Inge in Radiation Protection And ALARA.

NUMEl/CR-4856: FEASl81UTY STUDY ON A DATA BASED SYSTEM Regulatory Agende FOR DECIS80NS REGARDING OCCUPATONAL RADIATION PRO. NUREG-0938 V05 NO3: NRC REGULATORY AGENDA.Ouarterly TECTION MEASURES. Report. July September 1986.

Redleective Meterial Reguietory Analysis NUREG/CR-4736: COMBUSTION AEROSOLS FORMED DURING NUREG-1211: REGULATORY ANALYSIS FOR RESOLUTION OF UNRE.

BURNING OF RADIOACTIVELY CONTAMINATED MATERIALS EX- SOLVED SAFETY ISSUE A-46. SEISMIC QUALIFICATION OF EOUIP-PERIMENTAL RESULTS. MENT IN OPERATING PLANTS.

NUREG/CR-4712: REGULATORY ANALYSIS OF REGULATORY GUIDE Redleective Reteese 1.35 (REVISION 3. DRAFT 2) . IN-SERVICE INSPECTON OF UN-NURE11150 DRF V1 FC: REACTOR RISK REFERENCE GROUTED TENDONS IN PRESTRESSED CONCRETE CONTAIN-DOCUMENT. Main Report. Draft For Comment MENTS.

NUREG-1150 DAF V2 FC: REACTOR RISK REFERENCE DOCUMENT. Appendices A-1. Draft For Comment. Reguietory And Techn6 cal Report NUREG-1150 DHF V3 FC: REACTOR RISK REFERENCE NUREG-0304 Vit N04: REGULATORY AND TECHNICAL REPORTS DOCUMENT. J-0. Draft For Comment (ABSTRACT INDEX JOURNAL). Annual Compilation For 1986.

NUREG 1250: R T ON THE ACCIDENT AT THE CHERNOBYL NU-CLEAR POWER STATION. Regulatory Guide 1 NUREG/CR-4712: REGULATORY ANALYSIS OF REGULATORY GUIDE Radioactive Weste NUREG-1101 1.35 (REVISON 3. DRAFT 2) IN-SERVICE INSPECTON OF UN-V02: ONSITE DISPOSAL OF RADCACTIVE GROUTED TENDONS IN PRESTRESSED CONCRETE CONTAIN-WASTE. Methodology For The Radiological Assessment Of Dsposal By MENTS.

Seeurface Bunat Rollability N Aseeeement NUREG/CR-4626 V02: IMPROVING THE REUABILITY OF OPEN-NUREG-1101 V02: ONSITE DSPOSAL OF RADOACTIVE CYCLE WATER SYSTEMS. Application Of Befouhng Surveillance And WASTE. Methodology For The Radiological Assessment Of Dsposal By Control Techniques To Sediment And Corrosen Fouhng At Nuclear Subourface Bunal Power Plants.

Redlotegical Reiseee RT . m..t Energy Coot NOREG/CR-4552: A REVIEW OF THE SEABROOK STATION PROBABI- NUREG/CR-4012 V02: REPLACEMENT ENERGY COSTS FOR NUCLE-LOTIC SAFETY ASSESSMENT.Contarvnent Failure Modes And Radi- AR ELECTRICITY-GENERATING UNITS IN THE UNITED ological Source Terms. STATES:19871991.

Subject Index 43 NUREG-1137 S06: SAFETY EVALUATON REPORT RELATED TO TH Report To Congrees OPERATON OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 NUREG4090 V09 NO2: REPORT TO CONGRESS ON ABNORMAL AND 2. Docket Nos. 50424 And 50425.(George Power Company et al)

OCCURRENCES Apnt-hne 1986.

NUREG 1224: SAFETY EVALUATON REPORT RELATED TO THE RE NUREG 1260 Vvt: A REPORT TO CONGRESS ON NUCLEAR REGULA- NEWAL OF THE OPERATING UCENSE FOR THE UNIVERSITY OF TORY RESEARCH Project Desenphons For FYS7. NEW MEXICO RESEARCH REACTOR. Docket No. 50-252. (Urwersty 4975 V05: COMPILATON OF CONTRACT RESEARCH FOR THE MATERIALS BRANCH, DIVISION OF ENGINEERING gegetygoei SAFETY. Annual Rept For FY 1986- NUREG/CP-0064: PROCEEDINGS OF THE WORKSHOP ON 12- A CON.

NUREG-1260 V01: A REPORT TO CONGRESS ON NUCLEAR REGULA-TAINMENT PERFORMANCE DESIGN OBJECTIVE.MAY 13,1986. HARPERS FERRY, WEST VIRGINIA.

TORY RESEARCH Protect Descnptions For FY87.

^ ^ I " NU EG/ 97 OG REVIEWS OF SIX SAFETY PARAME-BO OM R A L TER DISPLAY SYSTEMS.

Riek Reduction Sefey hogram NUREG/CR-4551 V1 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR4616: ROOT CAUSES OF COMPONENT FAILURES RISKS AND THE POTENTIAL FOR RISK REDUCTON.SURRY PROGRAM Methods And Applications.

POWER STATON, UNIT 1. Draft For Comment.

Safety Reeeerch Rock Fracture Grouting NUREG-1163: COORDINATON OF SAFETY RESEARCH FOR THE NUREG/CR4541: EXPERIMENTAL ASSESSMENT OI: THE SEAUNG BABCOCK AND WILCOX INTEGRAL SYSTEM TEST PROGRAM.

EFFECTIVENESS OF ROCK FRACTURE GROUTING. NOREG/CR-2331 V06 N2: SAFETY RESEARCH PROGRAMS SPON-BY OFFICE OF NUCLEAR REGULATORY SORED Root Cause Analyste RESEARCH.Ouarterty Progress Report.Apnl4une 1966.

NUREG/CR4616: ROOT CAUSES OF COMPONENT FAILURES PROGRAM Methode And Applications. Scaling NUREG/CR-4531: AN INVESTIGATON OF INTEGRAL FACluTY SCAL-Rules ING AND DATA RELATION METHODS (INTEGRAL SYSTEM TEST NUREG4936 V05 NO3: NRC REGULATORY AGENDA.Quarterty PROGRAM).

ReportJuly-September 1986. NUREG/CR4824: EVALUATION OF INTEGRAL CONTINUING EXPER MENTAL CAPABlWTY (CEC) CONCEPTS FOR UGHT WATER REAC-Rulee of Meetico TOR RESEARCH PWR SCAUNG CONCEPTS.

NUREG4386 004 R04: UNITED STATES NUCLEAR REGULATORY COMMISSON STAFF PRACTICE AND PROC 9URE DIGESTJuly Seeling Effectivenees 1972 June 1986. NUREG/CR4541: EXPERIMENTAL ASSESSMENT OF THE SEAUN RE /C 93: RESULTS OF SEMISCALE MOD 2C SMALL-BREAK Sediment LOSS OF COOLANT ACCOENT WITHOUT HPl (S-NH) EXPERIMENT NUREG/CR-4626 V02: IMPROVING THE REUABluTY OF OPEN.

SERIES. CYCLE WATER SYSTEMS. Apphcation Of Biofouling Survedlance And S;D Sequence Control Techniques To Sediment And Corroson Fouling At Nuclear NUREG/CR 4741: FEEDWATER TRANSIENT AND SMALL BREAK Power Plants.

LOSS OF COOLANT ACCIDENT ANALYSES FOR THE BELLEFONTE Seemic CWegory i Tank NUCLEAR PLANT.

NUREG/CR4776: REFPONSE OF SEISMIC CATEGORY l TANK SEADEX D6epersion Data EARTHOUAKE EXCITATION.

NUREG/CH-4820: COMPARISON OF THE 1982 SEADEX DISPERSION DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. SWeic Doden NUREG/CR-4859: SEISMIC FRAGluTY TEST OF A 6-INCH DIA SLIM-MAUD PIPE SYSTEM.

NUREG/CR 4016 V02: APPUCATON OF SUM-MAUD:A TEST OF AN INTERACTIVE COMPUTER-BASED METHOO FOR ORGANIZING G/CR-4672- ANALYSIS OF INSTRUMENT TUBE RUPTURES E B UTY I: WESTINGHOUSE 4-LOOP PRESSURf2ED WATER REACTORS.

J Safe Shutdown Seiemic Escitetlon l

NUREG/CR4861: DEVELOPMENT OF SITE SPECIFIC RESPONSE NUREG/CR4776: RESPONSE OF SEISMO CATEGORY I TANK SPECTRA. EARTHOUAKE EXCITATION.

Sofeguardo Summary Event ust Setemic Hazard l NUREG-0525 R12: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL). NUREG/CR4852: THE MEERS FAULT: TECTONIC ACTIVITY IN SOUTHWESTERN OKLAHOMA.

Safety Evolustion Report NUREG4781 S02: SAFETY EVALUATON REPORT RELATEDSeismic TO THE Mergin OPERATON OF SOUTH TEXAS PROJECT, UNITS 1 AND 2. Docket NUREG/CR4026 V01: SEISMIC MARGIN REVIEW OF THE MA And Power Com YANKEE ATOMIC POWER STATON. Volume 1. Summary Report.

Nos. 50-498 And 50499. (Houston L THE NUREG4853 S06: SAFETY EVALUAT N EPORT RELAT D NUREG/CR4826 V02: SEISMIC MARGIN REVIEW OF THE MA OPERATON OF CUNTON POWER STATON, UNIT t. Docket No. 50- YANKEE ATOMIC POWER STATON. Volume 2. Systems Anafyss.

461 (Illino6s Power a NUREG/CH-4826 V03: SEISMIC MARGIN REVIEW OF THE MAI NUREG4857 S11: SAF .et VALUal) ATION REPORT RELATED TO THE YANKEE ATOMIC POWER STATION. Volume 3.Fragibty Analyss.

OPERATON OF PALO VERDE NUCLEAR GENERATING STATON, UNITS 1,2 AND 3. Docket Nos. 50-528.50-529 And 50- Seiemic Qualification Of Equipment 530 (Artzona Public Service Companv.et af) NUREG 1030 SEISMIC OUALIFICATION OF EQUIPMENT IN O NUREG-0876 SOS: SAFETY EVALUAflON REPORT RELATED INGTONUCLEARTHE POWER PLANTS. Unresolved Safety issue A-46.

OPERATION OF BYRON STATION, UNITS 1 AND 2. Docket Nos. 50- NUREG 1211: REGULATORY ANALYSIS FOR RESOLUTON O Commonwealth Edison Company) SOLVED SAFETY ISSUE A46. SEISMIC QUAUFICATON OF E 454 And NUREG 1057 50-455.

SO4: (SAFETY EVALUATION REPORT RELATED TO MENT THE IN OPERATING PLANTS.

OPERATION OF BEAVER VALLEY POWER STATON,UN!T 2. Docket No. 50412 (Duquesne Light Company.et of) Setemic Test NUREG-1137 S05: SAFETY EVALUATlON REPORT RELATED TO THE SEISMIC TESTING OF TYPICAL CONTAINM NUREG/CR-4734:

OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 PIPING PENETRATON SYSTEMS.

AND 2. Docket Nos. 50424 And 50-425.(Georgia Power Company,et al)

44 Subject index NUREG/CR-4859- SEISMIC FRAGILITY TEST OF A 6-INCH DIAMETER Small Break LOCA PlPE SYSTEM.

NUREG/CR4711: LOW UPPER-SHELF TOUGHNESS HIGH-TRANSl-Downcomer TON TEMPERATURE TEST INSERT IN HSST PTSE 2 VESSEL AND WOE PLATE TEST SPECIMENS. Final Report.

NUREG/lA 0004: THERMAL MIXING TESTS IN A SEMIANNULAR DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS. NUREG/CR4788 AN ANALYTICAL AND EXPERIMENTAL INVESTIGA-TON OF NATlHAL CIRCULATi1N TRANSIENTS IN A MODEL PRES-Semlocale Mod-2C FactIW SURIZED WATER REACTOR.

NUREG/CR-4793: RESULTS OF SEMISCALE MOO-2C SMALL BREAK NUREG/CR-4789: THE SIMULATION OF THERMOHYDRAULIC PHE-LOSS-OF400LANT ACCIDENT WITHOUT HPI(S-NH) EXPERIMENT NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP.

SERIES. NUREG/CR4793. RESULTS OF SEMISCALE MOD 2C SMALL-BREAK LOSSCF COOLANT ACCIDENT WITHOUT HPI (S-NH) EXPERIMENT Severe Accident '

NURE].1150 DRF VI FC: REACTOR RISK REFERENCE Soo-Structure Interaction DOCUMENT. Main ReportDraft For Comment.

NUREG 1150 DRF V2 FC: REACTOR RISK REFERENCE NUREG/CP-0054: PROCEEDINGS OF THE WORKSHOP ON SOIL.

DOCUMENT.Appendees A-l. Draft For Comment. STRUCTURE INTERACTON.

NUREG 1150 DRF V3 FC: REACTOR RISK REFERENCE Sondification DOCUMENT.Appendees J-O Draft For Comment.

NUREG-1210 V01: PILOT PROGRAM.NRC SEVERE REACTOR ACCI- NUREG/CR-3444 V04. THE IMPACT OF LWR DECONTAMINATONS DENT INCIDENT RESPONSE TRAINING MANUALOverview And Sum- ON SOLIDIFICATIONWASTE DISPOSAL,AND ASSOCIATED OCCU-mary Of Major Points. PATIONAL EXPOSURE. Annual Report, Fiscal Year 1986.

NUREA1210 V02: PtLOT PROGRAM.NRC SEVERE REACTOR ACCl* Solute Transport DENT INCIDENT RESPONSE TRAINING MANUALSevere Reactor Ac' NU 1 NUREG/CR-4737: INTERPRETATIVE ANALYSIS OF DATA FOR

03. PILOT PROGRAM.NRC SEVERE REACTOR ACCl- SOLUTE TRANSPORT IN THE UNSATURATED ZONE.

DENT INCIDENT RESPONSE TRAINING MANUALResponse Of Li.

Source Term NU 12 P PROGRA NRC SEVERE REACTOR ACCl- ^ ^ " ^O DENT INCIDENT RESPONSE TRAINING MANUALPublic Protective Actions . Predetermined Cnteria And Irvtial Actiora I c Terms NUREG 1210 V05. PILOT PROGRAM.NRC SEVERE REACTOR ACCl-DENT INCOENT RESPONSE TRAINING MANUALU.S. Nuclear Regu- Southern California NUR / NUREG/CR4801: CLIMATOLOGY OF EXTREME WtNDS LN SOUTH-20 TH TIONSHIP AND INFLUENCES OF FUEL " ^

AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCl- Spent Fuel Cask DENTS.

NUREl/CR4551 V1 DRF: EVALUATION OF SEVERE ACCOENT A E M RISKS AND THE POTENTIAL FOR RISK REDUCTON.SURRY H H A sin NU / 2A IE TE OOK STATION PROBABI- 0 S HIG AY AD R LWAY LCTIC SAFETY ASSESSMENT. Containment Fadure Modes And Rad-ological Source Terms. CONDITIONS. Appendices.

A DEN NUREJ/CR-4610: EFFECTS OF LATERAL SEPARATION OF OXOIC AND METALLIC CORE DEBRIS ON THE BWR MK I CONTAINMENT Spent Fuel Shipment NURE NUREG/CR-4847: CASE HISTORIES OF WEST VALLEY SPENT FUEL R4696 CONTAJNMENT VENTING ANALYSIS FOR THE SHIPMENTS Final Report.

PEACH BOTTOM ATOMIC POWER STATON.

NUREG/CR4700 V1 DRF: CONTAINMENT EVENT ANALYSTS FOR Standard Format And Content LA SEVERE ACCIDENTS Surry Power Station,Urut 1. Draft NUREG-1280: STANDARD FORMAT AND CONTENT ACCEPTANCE CRITERIA FOR THE MATERtAL CONTROL AND ACCOUNTING NUREG/CR 4741: FEEDWATER TRANSIENT AND SMALL BREAK (MC&A) REFORM AMENDMENT.10 CFR Part 74 Subpart E. I LOSS OF COOLANT ACCOENT ANALYSES FOR THE BELLEFONTE NUCLEAR PLANT. Standard Review Plan NUREG 1200; STANDARD REVIEW PLAN FOR THE REVIEW OF A L1-Shipping Container Response CENSE APPLICATION FOR A LOW-LEVEL RADIOACTIVE WASTE NUREG/CR4829 V01: SHIPPING CONTAINER RESPONSE TO DISPOSAL FACILITY.

SEVERE p HIGHWAY AND RAILWAY ACCIDENT CONDITIONS. Main States NUREG/CR4829 V02: SHIPPING CONTAINER RESPONSE TO NUREG/CP 0085: MEETING WITH STATES ON THE LOW-LEVEL RA-SEVERE HIGHWAY AND RA!LWAY ACCIDENT CONDITIONS. Appendices. DIOACTIVE 1985.

WASTE POLICY AMENDMENTS ACT (LLRWPAA) OF Shutdown Station Blackout NUREG/CR4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR4550 V03: ANALYSIS OF CORE DAMAGE FREQUENCY A GENERAL ELECTRIC BWR3/ MARK ICase Study FROM INTERNAL EVENTS.SURRY UNIT 1.

NUREG/CR4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR-4742: MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' AC-A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case COENT SEQUENCE.

NUR /CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF Steam Air Environment A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. Case NUREG/CR-4798: IRON OXtDE AEROSOL EXPERIMENTS IN STEAM-Study.

AIR ATMOSPHERES NSPP TESTS 501-505 AND 511 DATA RECORD NUREG/CR4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF REPORT.

A WESTINGHOUSE 3 LOOP PRESSURIZED WATER REACTOR. Case 34 Steam Generator Single-Rod Fuel Codes Program NUREG/CR-4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE NUREG/CR47f 8: EXPERIMENTAL SUPPORT AND DEVELOPMENT OF AND STUCKOPEN SAFETY REllEF VALVE CARRYOVER SINGLE-ROO FUEL CODES PROGRAM Summary Report.

TESTSV b2 Steam Generator Transient Response Test Program.

Stratepc Special Nuclear Material Site Specific Response Spectra NUREG/CR4695:

NUREG/CR-4861: DEVELOPMENT OF SITE SPECIFIC RESPONSE MATERIAL CONTROL AND ACCOUNTING (MC&A)

SPECTRA. LOSS DETECTON DURING TRANSITION PERIODS AND PROCESS UPSET CONDITONS.

Subject Index 45 Streehoseon Cracking Thermohydraulic Phenomena NUREG/CR-3861: STRESS-CORROSON CRACklNG OF LOW- NUREG/CR4789: THE SIMULATON OF THERMOHYDRAULO PHE-STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP.

REPOSITORY ENVIRONMENTS.

Thermolumineecent Doolmeter NUR G0837 V06 NO3 TLD E TON MONITORING NUR G R-475 : T STEAM GENERATOR TUBE RUPTURE AND STUCK.OPEN SAFETY REUEF VALVE CARRYOVER TESTS MB 2 Steam Generator Transaent Response Test Program. Through Well Crochd Pipe NUREG/CR4853: APPROXIMATE METHODS FOR FRACTURE ANALY.

Subourface Burtag SES OF THROUGH-WALL CRACKED PIPES.

NUREG 1101 V02: ONSITE DISPOSAL OF RADIOACTIVE WASTE. Methodology For The Radiologeal Assessment Of Deposal By U -0540 V08 N10 TITLE LIST OF DOCUMENTS MADE PUSUCLY Success ukelnaood inden Methodology AVAILABLE. October 1-31,1986.

NUREG/CR-4016 V02: APPLICATON OF SUM-MAUD:A TEST OF AN NUREG4540 V06 N11: TITLE UST OF DOCUMENTS MADE PUBUCLY IN7ERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING AVAILABLE. November 1-30,1986.

EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND NUREG-0540 V08 N12: TITLE UST OF DOCUMENTS MADE PUBUCLY REUA8ILITY. Volume it. Appendices AVAILABLE. December 1-31,1986.

NUREG4540 V09 N01: TITLE UST OF DOCUMENTS MADE PUBLICLY

^ """" '

U G 4 26 . IMPROVING THE REUA81UTY OF OPEN-CYCLE WATER SYSTEMS. Application Of 8.ofouling Surveellance And Training Control Techniques To Sediment And Corrosson Fouling At Nuclear NUREG 1210 V01: PtLOT PROGRAM NRC SEVERE REACTOR ACCI-Power Plants. DENT INCOENT RESPONSE TRAINING MANUALOverview And Sum-mary Of Major Points.

System Analyste NUREG 1210 V02: PtLOT PROGRAM.NRC SEVERE REACTOR ACCI-NUREG/CR-4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF DENT INCOENT RESPONSE TRAINING MANUALSevere Reactor Ac-A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case cedent OverSew.

S NUREG-1210 V03: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-NUR /CR-4713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF DENT INCIDENT RESPONSE TRAINING MANUALResponse Of U-A BABCOCK AND WILCOX PRESSURIZED WATER REACTOR Case consee And State And Local Offcials.

Study.

NUREG/CR4782: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG-1210 VO4: P! LOT PROGRAM.NRC SEVERE REACTOR ACCI-A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTORCase DENT INCOENT RESPONSE TRAINING MANUALPublic Protective Study. Actions Predetermined Criteria And Irvtial Actions.

NUREG/CR4826 V01: SEISMIC MARGIN REVIEW OF THE MAINE NUREG-1210 V05: PILOT PROGRAM.NRC SEVERE REACTOR ACCl-YANKEE ATOMIC POWER STATON Volume 1 Summary Report. DENT INCIDENT RESPONSE TRAINING MANUALU.S. Nuclear Regu-NUREG/CR4826 V02 SEISMIC MARGIN REVIEW OF THE MAINE latory Commission Response YANKEE ATOMIC POWER STATION. Volume 2 Systems Anahsis.

NUREG/CR4826 V03: SElSMIC MARGIN REVIEW OF THE MAINE Trenaient YANKEE ATOMIC POWER STATON Volume 3.Fragihty Analysas. NUREG/CR-4788: AN ANALYTICAL AND EXPERIMENTAL INVESTIGA-TION OF NATURAL CIRCULATION TRANSIENTS IN A MODEL PRES-TLD SURIZED WATER REACTOR.

NUREG 0837 V06 NO3: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4824: EVALUATON OF INTEGRAL CONTINUING EXPERI-NETWORK Progress Report, July. September 1986.

MENTAL CAPA81UTY (CEC) CONCEPTS FOR UGHT WATER REAC-TML8' TOR RESEARCH - PWR SCAUNG CONCEPTS.

NUREG/CR 4711: LOW UPPER-SHELF TOUGHNESS.HIGH-TRANSI.

Transient Reeponse Toot Program TON TEMPERATURE TEST INSERT IN HSST PTSE-2 VESSEL AND WOE-PLATE TEST SPECIMENS Final Report. NUREG/CR-4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE NUREG/CR-4742 MELPROG-PWR/ MOD 1 ANALYSIS OF A TML8' AC- AND STUCK-OPEN SAFETY REUEF VALVE CARRYOVER CIDENT SEQUENCE. TESTS MB-2 Steam Generator Transaent Response Test Program.

l Technical Specification Trane4 tion Range NUREG 1237: TECHNICAL SPECIFICATONS FOR VOGTLE ELECTRIC NUREG/CR-4818: TRANSITON RANGE DROP TOWER J-R CURVE GENERATING PLANT, UNIT 1 Docket No. 50-424.(George Power TESTING OF A106 STEEL NU 1740: TECHNICAL SPECIFICATIONS FOR SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1. Docket No. 50400 (Carohna Power TRESSING.

[ ELOPMENT OF MODELS FOR WARM PRES.

NUR 12 TE NICAL SPECIFICATONS FOR VOGTLE ELECTRIC GENERATING PLANT. UNIT 1. Docket No. 50424(Georgia Power TW h NU 1748. TECHNICAL SPECIFICATONS FOR PALO VERDE NU- NUREG/CR 2478 V03: A STUDY OF TRENCH COVERS TO MINIMlZE INFILTRATON AT WASTE DISPOSAL SITES.Fw1al Rept.

CLEAR GENERATING STATION, UNIT 3 Docket No. 50-530.(Anzona Public Service Company)

Tectonic Activtfy NUREG/CR4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE NUREG/CR4852: THE MEERS FAULT; TECTONIC ACTMTY IN AND STUCK-OPEN SAFETY REUEF VALVE CARRYOVER SOUTHWESTERN OKLAHOMA.

TESTS.MB-2 Steam Generator Transient Response Test Program.

Temperature Ungrouted Tendon NUREG/CR-4711: LOW UPPER. SHELF TOUGHNESS.HIGH TRANSI- NUREG/CR-4712- REGULATORY ANALYSIS OF REGULATORY GUIDE TON TEMPERATURE TEST INSERT IN HSST PTSE 2 VESSEL AND 1.35 (REVISION 3. DRAFT 2) - IN-SERVICE INSPECTION OF UN-W10E-PLATE TEST SPECIMENS. Final Ram GROUTED TENDONS IN PRESTRESSED CONCRETE CONTAIN-MENTS.

Thermal Hydraulic Phenomena NUREG/CR 4824: EVALUATION OF INTEGRAL CONTINUING EXPERI- Unresolved Safety leeue A 48 MENTAL CAPABlUTY (CEC) CONCEPTS FOR UGHT WATER REAC, NUREG-1030: SEISMIC OUAUFICATON OF EQUIPMENT IN OPERAT-TOR RESEARCH PWR SCAUNG CONCEPTS. ING NUCLEAR POWER PLANTS. Unresolved Safety issue A46.

Thermal Mixing NUREG-1211: REGULATORY ANALYSIS FOR RESOLUTION OF UNRE-NUREG/lA4004 THERMAL MIXING TESTS IN A SEMIANNULAR SOLVED SAFETY ISSUE A46 SEISMIC OUAUFICATION OF EQUIP-DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS. MENT IN OPERATING PLANTS.

46 Subject Index Unsaturated Zone Water Quality NUREG/CR-4737: INTERPRETATIVE ANALYSIS OF DATA FOR NUREG 1243; GROUNDWATER PROTECTION ACTIvlTIES OF THE SOLUTE TRANSPORT IN THE UNSATURATED ZONE. U.S. NUCLEAR REGULATORY COMMISSION.

Y Nuction y,,g y,,,,y NUREG/CR-4787: CONFERENCE OF RADIATION CONTROL DIREC.

TOR'S INFORMATION FOR UCENSING LOW LEVEL RADIOACTIVE NUREG/CR-4847: CASE HISTORIES OF WEST VALLEY SPENT FUEL WASTE INCINERATORS AND COMPACTORS. SHIPMENTS.Fnal Report.

Carm Prostressing Westinghouse NUREG/CR-4491: DEVELOPMENT OF MODELS FOR WARM PRES- NUREG/CR-4672- ANALYSIS OF INSTRUMENT TUBE RUPTURES IN TRESSING. WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS.

Caste Disposal Weetinghouse 2-Loop NUREG/CR-3444 V04. THE IMPACT OF LWR DECONTAMINATIONS ON SOUDiFICATION. WASTE DISPOSAL,AND ASSOCIATED OCCU- NUREG/CR-4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF PATIONAL EXPOSURE. Annual Report, Fiscal Year 1986. A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case Waste Disposal Site NUREG/CR-2478 V03. A STUD'/ OF TRENCH COVERS TO MINIMlZE Westinghouse M. cop lNFILTRATION AT WASTE DISPOSAL SITES Foal Rept. NUREG/CR-4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A WESTINGHOUSE 3-LOOP PRESSURIZED WATER REACTOR. Case Waste Package Test Data NUREG/CR-4735 V01: EVALUATON AND COMPILATION OF DOE SW.

WASTE PACKAGE TEST DATA. Bennual ReportDecember 1985 - Workshop 4 1986. NUREG/CP 0054: PROCEEDINGS OF THE WORKSHOP ON SOIL.

Water Leakage STRUCTURE INTERACTION.

NUREG/CR-4524 CLOSEOUT OF IE BULLETIN 80-24 PREVENTION NUREG/CP-0084: PROCEEDINGS OF THE WORKSHOP ON A CON-C7 DAMAGE DUE TO WATER LEAKAGE INSIDE CONTAINMENT TAINMENT PERFORMANCE DESIGN OBJECTIVE.MAY 12-(CCTOBER 17,1980 INDIAN POINT 2 EVENT). 13,1986, HARPERS FERRY, WEST VIRGINIA.

3 l

l

NRC Originating Org:nizcti n Ind3x (Staff R::p3rts)

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

OtVISION OF OA, VENDOR & TECHNICAL TRAINING CENTER PRO-OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) GRAMS (850212-870 REGION 1, OFFICE OF DIRECTOR NUREG4040 V10 N04: LICENSEE CONTRACTOR AND VENDOR IN-NUREG 0837 V06 NO3 NRC TLD DIRECT RADIATION MONITORING SPECTION STATUS REPORT. Quarterty Report,0ctober-December NETWORK Progress Report, July September 1986. *U F00 OFFICE OF ADMINISTRATION 870413) OFFICE OF INFORMATION RESOURCES MANAGEMENT ?

OFFICE OF ADMINISTRATION (PRE DIVISION OF COMPUTER & TELECOMMUNICATIONS SERVICES (Pt NUREG-0304 V11 N04. REGULATORY AND TECHNICAL REPORTS 870413)

(ABSTRACT INDEX JOURNAL) Annual Compilation For 1986.

NUREG-0325 R10: U S. NUCLEAR REGULATORY COMMISSION NUREG-0020 V10 N09 LICENSED OPERATING REACTORS MATL >

UMENTS MADE PUBLIC- NUREG 0 0L ENS OkRATI hO a- 2 NURE O VO E S F ?)

SUMMARY

REPORT.Deta As Of Septemter 30,1986.(Gr ,

NUREG-0540 V08 N12. TITLE LIST OF DOCUMENTS MADE OF PUBUC. L^

0 09 E LI O DOCUMENTS MADE PUBLIC- DV SAFE AR (PR 8704 3 NU EG 1 NUREG-0525 R12. SAFEGUARDS

SUMMARY

EVENT LIST (SS' .

LY AV AILABLE. January 1 31.1987 STANDARD FORMAT AND CONTENT ACCEP144E NUREG 1280:

NUREG4750 V24 N01. NUCLEAR REGULATORY COMMISSION IS-CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNTING V24 N 2 N LE E LATORY COMMISSION IS- (MC&A) REFORM AMENDMENT. to CFR Part 74 Subpart E.

NUREG O DIV OF RADIOACTIVE 24 NO3 U EAR EGU T RY COMMISSION IS- N E o V2 ONS TE DISPOSA NURE WASTE Methodology For The Rachological Assessment Of Disposal L R AT O & UMENT CONTROL 8 Subsurface Bunat DIV1 ON OF ECH NU EG 1199 STANDARD FORMAT AND CONTENT OF A LICENSE N REG 540 V08 N10. TITLE LIST OF DOCUMENTS MADE PUBUC- APPLICATION FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOS-LY AVAILABLE October f.31,1988 AL FACluTY.

OlvlSION OF RULES & RECORDS (PRE 870413) NUREG-1200: STANDARD REVIEW PLAN FOR THE REVIEW OF A NUREG 0936 V05 NO3. NRC REGULATORY AGENDA Quarterty LICENSE APPUCATION FOR A LOW-LEVEL RADIOACTIVE WAS Report. July September 1986. DISPOSAL FACILITY.

NUREG 1243. GROUNO-WATER PROTECTION ACTIVITIES OF T EDO . OF FICE OF STATE PROGRAMS U S. NUCLEAR HEGULATORY COMMISSION.

ASSISTANT DIRECTOR FOR ST ATE AGREEMENTS PROGRAMS NUREG/CP4085 MEETING WITH STATES ON THE LOW-LEVEL U.S.RA- NUCLEAR REGULATORY COMMISSION DIOACTIVE WASTE POLICY AMENDMENTS ACT (LLRWPAA) OFOF THE GENERAL COUNSEL OFFICE 1985 NUREG-0388 004 R04. UNITED STATES NUCLEAR REGULATORY COMMISS)ON STAFF PRACTICE AND PROCEDURE OtGEST.Juiy EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL 1972 June 1986.

DATA NRC - NO DETAILED AFFILIATION GfvEN OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, Di- NUREG 1250: REPORT ON THE ACCIDENT AT THE CHERNO8YL RECTOR NUCLEAR POWER STATION NUREG 0000 V09 NO2. REPORT TO CONGRESS ON ABNORMAL NUREG/CR-3620 502: INTRUDER DOSE PATHWAY ANALYSIS FO OCCURRENCES Apnl4une 1986 THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE MAxit Computer Program 12/11/80)

OFFICE OF INSPECTION & ENFORCEMENT (POSTINSPECTION NUREG/CR-3950

& ENFORCEMENT, V03. DIRECTOR FUEL PERFORMANCE(820201- ANNUAL REPORT Off fCE OF NUI G 430 V07 N01. LICENSED FUEL FACIUTY STATUS OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

REPORT Inventory Ditterence Data January 4une 1986 (Gray Bor k ll) OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR (POST NUREG 0940 V05 N04 ENFORCEMENT ACTIONS SIGNIFICANT AC-860720)

TIONS RESOLVEDOuarterty Progress Report. October December NUREG 1150 DAF V1 FC: REACTOR RISK REFERENCE 1986 DOCUMENT. Main ReportDraft For Comment DIVISION OF EMERGENCY PREPAREDNESS & ENGINEERING RE-NUREG-1150 DRF V2 FC. REACTOR RISK REFERENCE SPONSE (8502t 2 87041 DOCUMENT. Append >ces A-l Draft For Comment NUREG 1210 V01: PfLOT PROGRAM NRC SEVERE REACTOR NUREG-1150 ACCI- DRF V3 FC: REACTOR RISK REFERENCE DENT INCIDENT RESPONSE TRAINtNG MANUALOvennow And DOCUMENT.Appenchces J-0. Draft For Comment Summary Of Major Points NUREG 1260 V01. A REPORT TO CONGRESS ON NUCLEAR REG NUREG 1210 V02 PILOT PROGRAM NRC SEVERE REACTOR LATORY ACCI- RESEARCH Proyect Descnotions For FY87.

DENT INCIDENT RESPONSE TRAIN!NG MANUALSevers Reactor NUREG/CP4082 V01: PROCEEDINGS OF THE FOURTEENTH Accu 1ent Overv ew WATER REACTOR SAFETY INFORMATION MEETING NUREG 1210 V03 PILOT PROGRAM NRC SEVERE REACTOR ACCl-NUREG/CP4082 V02: PROCEEDINGS OF THE FOURTEENTH DENT INCIDENT RESPONSE TRAINtNG MANUALResponse Of Li- WATER REACTOR SAFETY INFORMATION MEETING censee And State And Local Offcats NUREG/CP-0082 V03: PROCEEDINGS OF THE FOURTEENTH NUREG-1210 V04 PILOT PROGRAM NRC SEVERE REACTORWATER ACCl-REACTOR SAFETY INFORMATION MEETING DENT INCIDENT RESPONSE TRAlN!NG MANUAL Public Protective NUREG/CP 0082 V04 PROCEEDINGS OF THE FOURTEENTH Actions Predetermined Cntena And init,al Actions WATER REACTOR SAFETY INFORMATION MEETING.

NUREG 1210 V05 PILOT PROGRAM NRC SEVERE Nuclear REACTOR ACCl-NUREG/CP-0082 VOS. PROCEEDINGS OF THE FOURTEENTH DENT INCIDENT RESPONSE TRAINING MANUALU S W ATER REACTOR SAFETY INFORMAtlON MEETING.

Regulatory Commission Response 47 1

48 NRC Originating Organization Index (Staff Reports)

NOREG/CP4082 V06: PROCEEDtNGS OF THE FOURTEENTH WATER REACTOR SAFETY INFORMATON MEETING. NUREG-1237: TECHNICAL SPECIFICATIONS FOR VOGTLE ELEC-DIVISION OF ENGINEERING SAFETY (860720 870413) TRIC GENERATING PLANT, UNIT 1. Docket No. 50-424 (Georgia Power Company)

NUREG-0975 V05: COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS BRANCH, DIVISION OF ENGINEERING NUREG 1240. TECHNICAL SPECIFICATIONS FOR SHEARON SAFETY. Annual Rept For FY 1986. HARRIS NUCLEAR POWER PLANT UNIT 1. Docket No. 50-400 (Carohne Power & bght Company)

DIVISON OF REACTOR SYSTEM SAFETY (860720-870413)

NUREG-1163: COORDINATION OF SAFETY RESEARCH FOR THE NUREG 1247: TECHNICAL SPECIFICATIONS FOR VOGTLE ELEC.

BABCOCK AND WILCOX INTEGRAL SYSTEM TEST PROGRAM. TRIC GENERATING PLANT, UNIT 1. Docket No. 50 424(Georgia Power Company)

EDO-MESOURCE MANAGEMENT DIVISON OF PRESSURIZED WATER REACTOR LICENSING - B DIVISION OF BUDGET & ANALYSIS (PRE 870413) (851125-870411)

NUREG-1100 V03: BUDGET ESTIMATES Fiocal Years 1988-1989. NUREG4857 S11: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PALO VERDE NUCLEAR GENERATING OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80) STATON, UNITS 1.2 AND 3. Docket Nos. 50-528,50-529 And 50-DIVISON OF PRESSURIZED WATER REACTOR LICENSING A NURE TY SL O RT RELATED TO THE N RE 4 8 S 2: SAFETY EVALUATON REPORT RELATED TO RENEWAL OF THE OPERATING LICENSE FOR THE UNIVERSITY THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 AND 2 Cocket Nos. 50 498 And 50-499. (Houston Ughting And Pows OF NEW MEXICO RESEARCH REACTOR Docket No. 50-252. (Uni-N$E 14 ECkN L SPECIFICATONS FOR PALO VERDE N NU 6 S08. SAFETY EVALUATON REPORT RELATED TO CLEAR GENERATING STATION. UNIT 3. Docket No. 50-530(Atuone THE OPERATION OF BVRON STATON, UNITS 1 AND 2 Docket Noe. 50454 And 50-455. (Commonwealth Edison Company) DIVIS O BOL ER REACTOR (BWR) LICENSING (851125-NUREG-1057 SO4: SAFETY EVALUATION REPORT RELATED TO 870411)

THE OPERATION OF BEAVER VALLEY POWER STATION, UNIT NUREG4853 S08: SAFETY EVALUATION REPORT RELATED TO 2 Docket No 50-412.(Duquesne Light Company et al) THE OPERATION OF CUNTON POWER STATON. UNIT 1. Docket NUREG-1137 505: SAFETY EVALUATION REPORT RELATED TO No. 50-461.(Illinois Power Company,et al)

THE OPERATION OF VOGTLE ELECTRIC GENERATING Divis)ON OF SAFETY REVIEW & OVERSIGHT (851125-870411)

PLANT. UNITS 1 AND 2.Dceket Nos. 50-424 And 50-425(Georgia NUREG ISSUES.

0933 S06: A PRIORITIZATION OF GENERIC SAFETY Power Company et al)

NUREG 1137 S06: SAFETY EVALUATION REPORT RELATED TO NUREG 1000: SEISMIC QUALIFICATION OF EQUIPMENT IN OPER-THE OPERATION OF VOGTLE ELECTRIC GENERATlNG ATING NUCLEAR POWER PLANTS Unresolved Safety lasue A 46.

PLANT UNITS 1 AND 2. Docket Nos. 50-424 And 50-425(Georgia NUREG 1211: REGULATORY ANALYSIS FOR RESOLUTION OF UN.

Power Company.et al) RESOLVED SAFETY ISSUE A46, SEISMIC QUAllFICATION OF EQUtPMENT IN OPERATING PLANTS.

l l

}

NRC Originating Organization index (International Agreem:nts)

This index lists those NRC organizations that have published intei. tional agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-and then by subsections of these (e.g., divisions, branches) where appr,opriate. Each fices)is entry followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

"ds'cf0POEEESE$tffdR"'AEEf#C7 ERA'8/$NSr NU EG M4 THERMAL MIXING TESTS IN A SEMIANNULAR DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS.

l l

49

A .______. __. -_ ;2_.

1 l

1

)

1 l

l 1

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by rnajor NRC organization (e.g., program offi,ce) and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) prepared by that organi-zation. If further information is needed, refer to the main citation by the NUREG/CR number.

NUREG/CR-4531: AN INVESTIGATION OF INTEGRAL FACILITY OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

REGION 2. OFFICE OF DIRECTOR SCALING AND DATA RELATION METHODS (INTEGRAL SYSTEM NUREG/CR4B68 METALLURGICAL EVALUATION OF AN 18-INCH TEST PROGRAM)

FEEDWATER LINE FAILURE AT THE SURRY UNIT 2 POWER STA- NUREG/CR4741: FEEDWATER TRANSIENT AND SMALL BREAK TION LOSS OF COOLANT ACCIDENT ANALYSES FOR THE BELLE-EDO - OFFICE FOR ANALYSIS & EV ALUATION OF OPERATIONAL NU EG/C 4 TH SSIBILITY OF LOCAL DETONATIONS DURING DEGRADED-CORE ACCIDENTS IN THE 8ELLEFONTE OFF FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA. D4- NUCLEAR POWER PLANT.

RECTOR DIVISION OF ENGINEERING SAFETY (860720-870413)

NUREG/CR-2000 V05N'2' LICENSEE EVENT REPORT (LER) NUREG/CR 3232- DETAILED STUDIES OF SELECTED.WELL EX-COMPILATION For Month Of December 1986 POSED FRACTURE ZONES IN THE ADIRONDACK MOUNTAINS NUREG/CR-2000 V06 N1 LICENSEE EVENT REPORT (LER) E NE ORK.

COMPILATION For Month Of January 1987. R PROGR AM Progress Report.Apnl 1983 -December 1984.

OFFICE NUREG/CR RE- 3444 V04. THE IMPACT OF LWR DECONTAMINATIONS OlVISIONOF OFINSPECTION & ENFORCEMENT EMERGENCY PREPAREDNESS (POST 12/11/80)

& ENGINEE RING ON SOLIDIFICATION. WASTE DISPOSALAND ASSOCIATED OCCU-SPONSE (850212-87041 PATIONAL EXPOSURE Annual Report. Fiscal Year 1986-NUREG/CR-4524. CLOSEOUT OF IE OULLETIN 80-24 PREVENTION NUREG/CR 3861: STRESS-CORROSlON CRACKING OF LOW-OF DAMAGE DUE TO WATER LEAKAGE INSIDE CONTAINMENT STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL (OCTOBER 17.1980 INDIAN POINT 2 EVENT). WASTE REPOSITORY ENVIRONMENTS.

NUREG/CR4300 V03 N2 ACOUSTIC EMISSION / FLAW RELATION-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS DIVISION OF FUEL CYCLE & MATERIAL SAFETY (PRE 870413) SHIP FOR INSERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR4736- COMBUSTION AEROSOLS FORMEO DURtNG VESSELS Progress Rept.Apni-September 1986.

DURNING OF RADIOACTIVELY CONTAMINATED MATERIALS NUREG/CR-4469 V04. NONDESTRUCTIVE EXAMINATION (NDE) RE.

BW M WS N E WSMCW & M WATER NUR 484 H STORIES OF WEST VALLEY SPENT REACTORS Semiannual Report. October 1985 March 1986.

O #

DIVI ON F AFEG ARD R 870413)

NUREG/CR-4695: MATERIAL CONTROL AND ACCOUNTING (MC&A) NURE R 4541: EXPERlMENTAL ASSESSMENT OF THE SEALING LOSS DETECTION DURING TRANSITION PERIODS AND PROC- EFFECTIVENESS OF ROCK FRACTURE GROUTING.

NUREG/CR-4685: POST-PLIOCENE DISPLACEMENT ON FAULTS DIVt ION F WA A EMENT (PRE 870413) WITHIN THE KENTUCKY RIVER FAULT SYSTEM 08 EAST-CEN-NUREG/CR 2478 V03 A STUDY OF TRENCH COVERS TO MINIMl2E TRAL KENTUCKY.

INFILTRATION AT WASTE DISPOSAL SITES Foal Rept. NUREG/CR4711: LOW UPPER-SHELF TOUGHNESS.HIGH TRANSI-NUMG/CR-3620 SO2: INTRUDER DOSE PATHWAY ANALYSIS FOR TION TEMPERATCRE TEST INSERT IN HSST PTSE-2 VESSEL THE ONSITE DISPOSAL OF RADIOACTIVE W'ASTES The ONSITE/ AND WIDE-PLATE 1EST SPECIMENS. Final Report.

Wit Computer Program. NUREG/CR-4712: REGULATORY ANALYSIS OF REGULATORY NUREG/CR4708 VOI N1: PROGRESS IN EVALUATION OF RADIO- GUIDE 1.35 (REVISION 3. DRAFT 2) - IN-SERVICE INSPECTION NUCLlDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CON-HIGH LEVEL NUCLEAR WASTE REPOSITORY SITE TAINMENTS.

PROJECTS Semiannual Report For October 1985 - March 1986 NUREG/CR-4724. FATIGUE CR ACK GROWTH RATES IN PRESSURE NUREG/CR4730. EVALUATION OF POTENTIAL MIXED WASTES VESSEL AND P! PING STEELS IN LWR ENVIRONMENTS Final CONTA:N!NG LEAD, CHROMlUM.USED OILOR ORGANtC LIO- rep 0't.

UlDS NUREG/CR-4734: SEISMIC TESTING OF TYPICAL CONTAINMENT NUREG/CR-4735 V01: EVALUATION AND COMPILATION OF DOE PlPING PENETRATION SYSTEMS.

W ASTE PACKAGE TEST DATA Biannual Report December 1985 NUREG/CR4744 V01 N1: LONG-TERM EMBRITTLEMENT OF CAST Juh 1986. DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUR EG/CR-4 737- INTERPRETAT!VE ANALYSIS OF DATA FOR Report. October 1985 - March 1986 SOLUTE TRANSPORT IN THE UNSATURATED ZONE. NUREGICR4787: CONFERENCE OF RADIATION CONTROL DIREC-NUREG/CR4825- A PRELIMINARY EVALUATION OF THE ECONOM- TORS INFORMATION FOR LICENSING LOW-LEVEL RADIOAC-IC RISK FOR CLEANUP OF NUCLEAR MATERIAL LICENSEE CON. TIVE WASTE INCINERATORS AND COMPACTORS.

TAMINATION INCIDENTS. NUREG/CR-4813. ASSESSMENT OF LEAK DETECTION SYSTEMS FOR LWRs. October 1985 September 1986.

OFFICE OFHCE OF OFNUCLEAR NUCLEAR REGULATORY REGULATORY RESEARCH, RESEARCH OtRECTOR (POST (P 4/05/81)NUREG/CR4818:

OST TRANSITION RANGE DROP TOWER J-R CURVE TESTING OF A106 STEEL 860720) NUREG/CR-4820: COMPARISON OF THE 1982 SEADEX DISPER-NUREG/CR-2331 V06 N2- SAFETY RESEARCH PROGRAMS SPON- SION DATA WITH RESULTS FROM A NUMBER OF DIFFERENT SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH Ouarter'y Progress Report.Apnt-June 1986 MODELS.

NUREG/CR4826 V01 SEISMIC MARGIN REVIEW OF THE MAINE NUREG/CR-3469 V03. OCCUPATIONAL DOSE REDUCTION AT NU- YANKEE ATOMfC POWER STATION Volume 1.Semmary Report.

CLEAR POWER PLANTS Annotated B.bliography Of Selected Read-NUREG/CR4826 V02: SEISMIC MARGIN REVIEW OF THE MAINE i s In Radiation Protection And ALARA. YANKEE ATOMIC POWER STATION Volume 2 Systems Analysis.

NUF EG/CR4301: STATUS RErORT ON EQUIPMENT OUALIFICA- NUREG/CR4826 V03 SEISMIC MARGlN REVIEW OF THE MAINE TION ISSUES RESEARCH AND RESOLUTION Y ANKEE ATOMIC POWER STATION Volume 3 Fragility Analysis.

NUREG/CR-4320 THE RELATIONSH:P AND INFLUENCES OF FUEL NUREG/CR4852 THE MEERS FAULT: TECTONIC ACTIVITY IN AND COOLANT SYSTEM PROCESSES DUR:NG LWR SEVERE AC-SOUTHWESTERN OKLAHOMA.

CIDENTS 51

52 NRC Contract Sponsor Index NUREG/CR-4853; APPROylMATE METHODS FOR FRACTURE NUREG/CR-4829 V02: SHIPPING CONTAINER RESPONSE TO ANALYSES OF THROUGH-WALL CRACKED PIPES. SEVERE H!GHWAY AND RAILWAY ACCOENT NUREG/CR4559- SEISMIC FRAGIUTY TEST OF A 6-INCH DIAME- CONDITIONS. Appendices.

TER PlPE SYSTEM. NUREG/CR4843 VG1: UNIVERSITY OF MARYLAND AT COLLEGE NUREG/CR-4861: DEVELOPMENT OF SITE SPECIFIC RESPONSE PARK (UMCP) 2X4 LOCP TEST FACIUTY. Annual Report For 1985.

SPECTRA. DIVISION OF ACCIDENT EVALUATON (POST 840101)

DIVISION OF REGULATORY APPLICATIONS (860720-870413) NUREG/CR-4793: RESULTS OF SEMISCALE MOD 2C SMALL-NUREG/CR4409 V02: DATA BASE ON NUCLEAR POWER PLANT BREAK LOSSOF-COOLANT ACCOENT WITHOUT HPI (S NH) EX-DOSE REDUCTON RESEARCH PROJECTS. PERIMENT SERIES.

I NUREG/CR4856: FEASIBluTY STUDY ON A DATA-BASED SYSTEM FOR DECISONS REGARDING OCCUPATIONAL RADIATION PRO. EDO-RESOURCE MANAGEMENT TECTON MEASURES COST & STATISTICAL ANALYSIS STAFF (861123-870413)

OfVISION OF REACTOR SYSTEM SAFETY (860720870413) NUREG/CR-4012 V02: REPLACEMENT ENERGY COSTS FOR NU-NUREG/CR-3468: HYDROGEN. AIR. STEAM FLAMMABluTY UMITS CLEAR ELECTRICITY GENERATING UNITS IN THE UNITED AND COMBUSTON CHARACTERISTICS IN THE FITS VESSEL STATES:19871991.

NUREG/CR 4016 V02: APPLICATION OF SUM MAUD'A TEST OF AN INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING OF U CTOR E N( / )

EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND N G/ -3469 V03: OCCUPATONAL DOSE REDUCTION AT NU-NURE / 50 0 N Y F CORE DAMAGE FREQUENCY CLEAR POWER PLANTS. Annotated Bibliography Of Selected Read-FROM TNT 0RNAL EVENTS.SURRY UNIT 1.

NUREG/CR4550 V04: ANALYSIS OF CORE DAMAGE FREQUENCY yygg%In[a AN S ECHNO OGY 51125-870411 FROM INTERNAL EVENTS PEACH BOTTOM UNIT 2 NUREG/CR-3968: STUDY OF OPERATING P OCEDURES lb NU-NVAEG/CR-4551 V1 DRF: EVALUATON OF SEVERE ACCIDENT CLEAR POWER PLANTS Practices And Problems RISKS AND THE POTENTIAL FOR RISK REDUCTIONSURRY NUREG/CR-4613: EVALUATION OF NUCLEAR POWER PLANT OP.

POWER STATON. UNIT 1. Draft For Comment. ERATING PROCEDURES CLASSIFICATIONS AND NJF.EG/CR4610- EFFECTS OF LATERAL SEPARATION OF OXIDIC INTERFACES. Problems And Tec ' ues For im ment.

ANU METALUC CORE DEBRIS ON THE BWR MK I CONTAIN- Division OF PRESSURIZED WA REACTO CENSING - A MENT ORYWELL FLOOR. (851125-870411)

NUREG/CR-4616: ROOT CAUSES OF COMPONENT FAILURES NUREG/CR-3950 V03: FUEL PERFORMANCE ANNUAL REPORT PROGRAM. Methods And Apphcations. FOR 1985 NUREG/CR-4672: ANALYSIS OF INSTRUMENT TUBE RUPTURES IN NUREG/CR'4552: A REVIEW OF THE SEABROOK STATION PROB-WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS. ABILISTIC SAFETY ASSESSMENT. Containment Failure Modes And NUREG/CR 4689: THERMAL-HYDRAULIC AND CHARACTERISTIC Radiological Source Terms.

MODELS FOR PACKED DEBRIS BEDS- NUREG/CR-4797: PROGRESS REVIEWS OF SIX SAFETY PARAME-NUREG/CR4696: CONTAINMENT VENTING ANALYSIS FOR THE TER DISPLAY SYSTEMS.

PEACH BOTTOM ATOMIC POWER STATION. NUREG/CR4868: METALLURGICAL EVALUATION OF AN 18-INCH NUREG/CR-4700 V1 DRF: CONTAINMENT EVENT ANALYSIS FOR FEEDWATER UNE FAILURE AT THE SURRY UNIT 2 POWER SFA-POSTULATED SEVERE ACCIDENTS. Suny Power Station,Urut TlON.

1. Draft For Comment. OlVISION OF PRESSURIZED WATER REACTOR UCENSING - B NUREG/CR-4718: EXPERIMENTAL SUPPORT AND DEVELOPMENT (851125-870411)

OF SINGLE-ROD FUEL CODES PROGRAM Summary Report NUREG/CR-4801: CUMATOLOGY OF EXTREME WINDS IN SOUTH-NUREG/CR-4742: MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' ERN CAllFORNIA.

ACCOENT SEQUENCE. DIVISION OF SAFETY REVIEW & OVERSIGHT (851125-870411)

NUREG/CR-4752: COINCIDENT STEAM GENERATOR TUBE RUP- NUREG/CR-4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS TURE AND STUCK OPEN SAFETY REUEF VALVE CARRYOVER OF A GENERAL ELECTRIC BWR3/ MARK 1. Case Study.

VESTS MB-2 Steam Generator Transient Response Test Program. NUREG/CR-4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS NUREG/CR-4788: AN ANALYTICAL AND EXPERIMENTAL INVESTI- OF A WESTINGHOUSE 2-LOOD PRESSURIZED WATER GATION OF NATURAL CIRCULATION TRANSIENTS IN A MODEL REACTORCase Study.

PRESSURIZED WATER REACTOR. NUREG/CR-4626 V02: IMPROVING THE REUABluTY OF OPEN-NUREG/CR-4789: THE SIMULATON OF THERMOHYDRAULIC PHE- CYCLE WATER SYSTEMS. Application Of Biofouhng Surveillance NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP. And Control Techniques To Sediment And Corrosion Fouhng At Nu-NUREG/CR-4798: IRON OXIDE AEROSOL EXPERIMENTS IN clear Power P' ants.

STEAM-AIR ATMOSPHERES NSPP TESTS 501-505 AND 511, DATA NUREG/C44713: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS RECORD REPORT. OF A BABCOCK AND WILCOX PRESSURIZED WATER NUREG/CR-4824: EVAL.UATON OF INTEGRAL CONTINUING EX- REACTOR Case Study.

PERIMENTAL CAPABiUTV (CEC) CONCEPTS FOR UGHT WATER NUREG/CH-4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS REACTOR RESEARCH - PWFt SCALING CONCEPTS. OF A WESTINGHOUSE 3-LOOP PRESSURIZED WATER NUREG/CR4829 V01: SH:PPING CONTAINER RESPONSE TO SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITIONS. Main REACTOR. RES NUREG/CR-4776: Case S . NSE OF SEISMIC CATEGORY I TANKS TO

, Report. EARTHOUAKE EXCITATION.

1

i Crntrcct:r Ind]x This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports. If further information is needed, refer to the main citation by the NUREG/CR number.

AERONAUTICAL RESEARCH ASSOCIATES OF PRINCETON NUREG/CR-4736: COMBUSTION AEROSOLS FORMED DURING NUREG/CR-4820: COMPARISON OF THE 1982 SEADEX DISPERSION BURNING OF RADIOACTIVELY CONTAMINATED MATERIALS - EX-DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. PERIMENTAL RESULTS.

NUREG/CR-4801: CLIMATOLOGY OF EXTREME WINDS IN SOUTH-AEROSPACE CORP. ERN CAUFORNIA NUREG/CR-4847: CASE HISTORIES OF WEST VALLEY SPENT FUEL NUREG/CR-4856. FEASIBluTY STUDY ON A DATA-BASED SYSTEM SHIPMENTS Final Report FOR DECISIONS REGARDING OCCUPATONAL RADIATION PRO-TECTION MEASURES.

ARGONNE NATIONAL LABORATORY NUREG/CR-4012 V02: REPLACEMENT ENERGY COSTS FOR NUCLE-BROOKHAVEN NATIONAL LABORATORY AR ELECTRICITY GENERATING UNITS IN THE UNITED STATES.1987-199 f. NUREG/CP-0054: PROCEEDINGS OF THE WORKSHOP ON SOIL-NUREG/CR-4744 V01 N1: LONG-TERM EMBRITTLEMENT OF CAST STRUCTURE INTERACTION.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUREG/CP 0082 V01: PROCEEDINGS OF THE FOURTEENTH WATER Report. October 1985 - March 1986. REACTOR SAFETY INFORMATION MEETING.

NUREG/CR-4813: ASSESSMENT OF LEAK DETECTION SYSTEMS NUREG/CP-0082 V02: PROCEEDINGS OF THE FOURTEENTH WATER FOR LWRs. October 1985 - September 1986. REACTOR SAFETY INFORMATION MEETING.

NUREG/CP 0082 V03: PROCEEDINGS OF THE FOURTEENTH WATER

^ NUR G/CP 2 V04: PR E GS H FOURTEENTH WATER EFF E K R R NG REACTOR SAFETY INFORMATION MEETING.

8ABCOCK & WILCOX CO. NUREG/CP-0082 V05: PROCEEDINGS OF THE FOURTEENTH WATER NUREG/CR-4711: LOW UPPER-SHELF TOUGHNESS.HIGH-TRANSl- REACTOR SAFETY INFORMATION MEETING.

TION TEMPERATURE TEST INSERT IN HSST PTSE 2 VESSEL AND NUREG/CP-0082 V06: PROCEEDINGS OF THE FOURTEENTH WATER WIDE-PLATE TEST SPECIMENS. Final Report REACTOR SAFETY INFORMATION MEETING.

NUREG/CP-0084: PROCEEDINGS OF THE WORKSHOP ON A CON-BATTELLE HUMAN AFFAIRS RESEARCH CENTERS TAINMENT PERFORMANCE DESIGN OBJECTIVE,MAY 12-NUREG/CR 3968: STUDY OF OPERATING PROCEDURES IN NUCLE- 13.1986, HARPERS FERRY, WEST VIRGINTA.

NUREG/CR-2331 V06 N2: SAFETY RESEARCH PROGRAMS SPON-NURE 6 EV L F NUC k POWER PLANT OPER- D W NM & EN EMN ATING PROCEDURES CLASSIFICATIONS AND RESEARCH.Ouarterly Progress Report.Apnt-June 1986.

INTERFACES Problems And Techruques For Improvement NUREG/CR-3444 V04: THE IMPACT OF LWR DECONTAMINATIONS BATTELLE WEMORIAL INSTITUTE, COLUM8US LABORATORIES ON SOUDIFICATION, WASTE DISPOSAL.AND ASSOCIATED OCCU-PATIONAL EXPOSURE. Annual Report, F'r scal Year 1986.

NUREG-1150 DRF V1 FC- REACTOR RISK REFERENCE DOCUMENT Main Report Draft For Comment NUREG/CR-3469 V03: OCCUPATIONAL DOSE REDUCTION AT NU-NUREG/CR-3861: STRESS-CORROSION CRACKING OF LOW- CLEAR POWER PLANTS. Annotated Bibhography Of Selected Read-STRENGTH CARBON STEELS IN CANDIDATE HIGH-LEVEL WASTE ings in Radiation Protection And ALARA.

REPOSITORY ENVIRONMENTS. NUREG/CR 4016 V02: APPUCATION OF SUM.MAUD:A TEST OF AN NUREG/CR-4853: APPROXIMATE METHODS FOR FRACTURE ANALY- INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING SES OF THROUGH-WALL CRACKED PIPES. EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST BASE ON NUCLEAR POWER PLANT NURE / -4 D DOSE REDUCTION RESEARCH PROJECTS.

NU E 36 SO2: INTRUDER DOSE PATHWAY ANALYSIS FOR NUREG/CR-4552: A REVIEW OF THE SEABROOK STATION PROBABI-THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/ LISTIC SAFETY ASSESSMENT. Containment Failure Modes And Radi-MAX 11 Computer Program. ological Source Terms.

NUREG/CR-3950 V03: FUEL PERFORMANCE ANNUAL REPORT FOR 1985.

NUREG/CR 4730: EVALUATION OF POTENTIAL MIXED WASTES CON-TAINING LEAD, CHROMlUM.USED OILOR ORGANIC LIQUIDS.

NUREG/CR 3968: STUDY OF OPERATING PROCEDURES IN NUCLE-AR POWER PLANTS. Practices And Prob 6 ems. NUREG/CR-4868: METALLURGICAL EVALUATION OF AN 18-INCH NUREG/CR-4300 V03 N2: ACOUSTIC EMISSION / FLAW RELATON- FEEDWATER LINE FAILURE AT THE SURRY UNIT 2 POWER STA-SHIP FOR INSERVICE MONITORING OF NUCLEAR PRESSURE TON.

VESSELS Progress ReptApr4-September 1986.

NUREG/CR-4469 V04: NONDESTRUCTIVE EXAMIN ATION (NDE) REU- CENTRAL ELECTRICITY GENERATING BOARD, BERKELEY NUCLEAR ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER LABS REACTORS. Semiannual Report. October 1985 March 1986- NUREG/CR-4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE NUREG/CR 4613: EVALUATION OF NUCLEAR POWER PLANT OPER. AND STUCK OPEN SAFETY RELIEF VALVE CARRYOVER ATING PROCEDURES CLASSIFICATONS AND TESTS.MB.2 Steam Generator Transient Response Test Program.

INTERFACES Problems And Tochtwques For improvement.

NUREG/CR-462C V02: IMPRC,iNG THE RELIABILITY OF OPEN- COMMERCE, DEPT, OF, NATIONAL BUREAU OF STANDARDS CYCLE WATER SYSTEMS. Apphcation Of Biofouhng Surveillance And Control Techniques To Sedrnent And Corrosson Fouhng At Nuclear NUREG/CR-4735 VOI: EVALUATION AND COMPILATION OF DOE WASTE PACKAGE TEST DATA. Biannual Report December 1985 -

Power Plants. July 1986.

NUREG/CR-4695: MATERIAL CONTROL AND ACCOUNTING (MC&A)

LOSS DETECTION DURING TRANSITION PERIODS AND PROCESS COMPUTATIONAL MECHANICS UPSET CONDITIONS.

NUREG/CR-4718: EXPERIMENTAL SUPPORT AND DEVELOPMENT OF NUREG/CR-4491: DEVELOPMENT OF MODELS FOR WARM PRES-SINGLE-ROD FUEL CODES PROGRAM. Summary Report TRESSING.

53 1

54 Contractor index CONFERENCE OF RADIATION CONTROL PROGRAM DIRECTORS. INC. NUREG/CR-4829 V01: SHIPPING CONTAINER RESPONSE TO NUREG/CR4787: CONFERENCE OF RADIATON CONTROL DIREC- SEVERE HIGHWAY AND RAILWAY ACCIDENT CONDITONS. Main TOR'S INFORMATION FOR UCENSING LOW-LEVEL RADOACTIVE Report WASTE INCINERATORS AND COMPACTORS.

NUREG/CR4829 V02: SHIPPING CONTAINER RESPONSE TO SEVERE HIGHWAY AND RAILWAY ACCIDENT DAVID W. TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER CONDITONS. Appendices.

NUREG/CR4818: TRANSITON RANGE DROP TOWER J-R CURVE TESTING OF A106 STEEL NUREG/CR-4861: DEVELOPMENT OF SITE SPECIFIC RESPONSE SPECTRA.

EG4G IDANO, INC. (SUBS. OF EG4G, INC.)

NUREG/CR4531: AN INVESTIGATION OF INTEGRAL FACluTY SCAL- LOS ALAMOS SCIENTIFIC LA80RATORY N G/CR4737: INTERPRETAME ANAWSIS OF DATA MR ING AND DATA RELATION METHODS (INTEGRAL SYSTEM TEST PROGRW SOLUTE TRANSPORT IN THE UNSATURATED ZONE.

NUREG/CR1616. ROOT CAUSES OF COMPONENT FAILURES NUREG/CR4776: RESPONSE OF SEISMIC CATEGORY I TANKS TO PROGRAM. Methods And Apphcations. EARTHOUAKE EXCITATION.

NUREG/CR4672: ANALYSIS OF INSTRUMENT TUBE RUPTURES IN WESTINGHOUSE 4-LOOP PRESSURIZED WATER REACTORS. MARYLAND, UNIV. OF, COLLEGE PARK, MD NUREG/CR-4696; CONTAINMENT VENTING ANALYSIS FOR THE NUREG/CR4788: AN ANALYTICAL AND EXPERIMENTAL INVESTIGA-PEACH BOTTOM ATOMIC POWER STATION. TON OF NATURAL CIRCULATON TRANSIENTS IN A MODEL PRES-NUREG/CR-4741: FEEDWATER TRANSIENT AND SMALL BREAK SURIZED WATER REACTOR.

LOSS OF COOLANT ACCIDENT ANALYSES FOR THE BELLEFONTE NUREG/CR4789: THE SIMULATION OF THERMOHYDRAUUC PHE-NUCLEAR PLANT.

NOMENA IN A PRESSURIZED WATER REACTOR PRIMARY LOOP.

NUREG/CR-4793: RESULTS OF SEMISCALE MOD 2C SMALL-BREAK NUREG/CR4843 V01: UNIVERSITY OF MARYLANO AT COLLEGE LOSS-OF-COOLANT ACCIDENT WITHOUT HPl (S-NH) EXPERIMENT PARK (UMCP) 2X4 LOOP TEST FACluTY. Annual Report For 1985.

SERIES.

NUREG/CR4824: EVALUATON OF INTEGRAL CONTINUING EXPERI- MATERIALS ENGINEERING ASSOCIATES,INC.

MENTAL CAPABluTY (CEC) CONCEPTS FOR UGHT WATER REAC. NUREG/CR4491: DEVELOPMENT OF MODELS FOR WARM PRES-TOR RESEARCH PWR SCAUNG CONCEPTS. TRESSING.

ELECTRIC POWER RESEARCH INSTITUTE NUREG/CR-4724: FATIGUE CRACK GROWTH RATES IN PRESSURE VESSEL AND PIPING STEELS IN LWR ENVIRONMENTS. Final Report.

NUREG-1250: REPORT ON THE ACCIDENT AT THE CHERNOBYL NU-CLEAR POWER STATION. NEVADA, UNIV. OF, RENO, NV ENERGY TECHNOLOGY ENGINEERING CENTER NUREG/CR 4859: SEISMIC FRAGIUTY TEST OF A 6-INCH DIAMETER HWESTE N OKLA PIPE SYSTEM.

NEW YORK, STATE UNIV. OF, ALBANY, NY OF N'JREG/CR-3232 DETAILED STUDIES OF SELECTED,WELL EXPOSED ENERGY NUREG-1250: DEPT' REPORT ON THE ACCIDENT AT THE CHERNOBYL NU- FRACTURE ZONES IN THE ADIRONDACK MOUNTAINS DOME.NEW YORK.

CLEAR POWER STATON.

ENERGY, INC. OAK RIDGE NATIONAL LABORATORY NUREG 1150 DRF V1 FC: REACTOR RISK REFERENCE NUREG/CR-2000 V05N12: UCENSEE EVENT REPORT (LER)

DOCUMENT Main Report Draft For Comment. COMPILATION For Month Of Decernber 1986, NUREG/CR-4826 V02: SEISMIC MARGIN REVIEW OF THE MAINE NUREG/CR-2000 V06 N1: UCENSEE EVENT REPORT (LER)

YANKEE ATOMIC POWER STATON. Volume 2. Systems Analysis. COMPILATON.For Month Of January 1987.

EOVlRONMENTAL PROTECTION AGENCY NUREG/CR4610- EFFECTS OF LATERAL SEPARATON OF OXIDIC AND METALLIC CORE DEBRIS ON THE BWR MK i CONTAINMENT NUREG 1250: REPORT ON THE ACCOENT AT THE CHERNOBYL NU- DRYWELL FLOOR.

i I

CLEAR POWER STATION. '

NUREG/CR-4689: THERMAL-HYDRAUUC AND CHARACTERISTIC EQE, INC. MODELS FOR PACKED DEBRIS BEDS. l NUREG/CR-4826 V03. SEISMIC MARGIN REVIEW OF THE MAINE NUREG/CR-4708 V01 N1: PROGRESS IN EVALUATON OF RADIONU.

YANKEE ATOMIC POWER STATON. Volume 3 Frag *ty Analysis. CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE FEDERAL EMERGENCY MANAGEMENT AGENCY PROJECTS.Serrmannual Report For October 1985 March 1988.

NUREG-1250: REPORT ON THE ACCIDENT AT THE CHERNOBYL NU. NUREG/CR-4711: LOW UPPER-SHELF TOUGHNESS,HIGH-TRANSI.

CLEAR POWER STATON. TON TEMPERATURE TEST INSERT IN HSST PTSE-2 VESSEL AND WIDE-PLATE TEST SPECIMENS. Final Report.

FUTURE RESOURCES ASSOCIATES,INC. NUHEG/CR-4712: REGULATORY ANALYSIS OF REGULATORY GUIDE NUREG-1150 DRF V1 FC: REACTOR RISK REFERENCE 1.35 (REVISON 3, DRAFT 2) - IN-SERVICE INSPECTION OF UN-DOCUMENT. Main ReportDraft For Comtnent. GROUTED TENDONS IN PRESTRESSED CONCRETE CONTAIN-MENTS.

IOAHO NATIONAL ENGINEERING LABORATORY NUREG/CR-4798: IRON OXIDE AEROSOL EXPERIMENTS IN STEAM.

NUREG/CR4734: SEISMIC TESTING OF TYPICAL CONTAINMENT AIR ATMOSPHERES.NSPP TESTS 501-505 AND 511. DATA RECORD PIPING PENETRATON SYSTEMS. REPORT.

ILLINOls, STATE OF PARAMETER, INC.

NUREG/CR-2478 V03: A STUDY OF TRENCH COVERS TO MINIMl2E INFILTRATION AT WASTE DISPOSAL SITES.Frial Rept NUREGICR4524: CLOSEOUT OF lE BULLETIN 80-24. PREVENTION OF DAMAGE DUE TO WATER LEAKAGE INSIDE CONTAINMENT INSTU1E OF NUCLEAR POWER OPERATIONS (OCTOBER 17,1980 INDIAN POINT 2 EVENT).

NOREG-1250: REPORT ON THE ACCIDENT AT THE CHERNOBYL NU- PROSIG, INC.

CLE W e0WER STATON.

NUREG/CR4491: DEVELOPMENT OF MODELS FOR WARM PRES-RE N* 1CKY, UNIV. OF, LEXINGTON, KY TRESSING.

NUREG/CR4685: POST-PLOCENE DISPLACEMENT ON FAULTS WITHIN THE KENTUCKY RIVER FAULT SYSTEM OF EAST-CEN. SANDIA NATIONAL LABORATORIES TRAL KENTUCKY. NUREG/CR 3412 V02: CONTAINMENT INTEGRITY PROGRAM Progress Report.Apnl 1983 -December 1984.

LAWRENCE LIVERMGRE NATIONAL LABORATORY NUREG/CR-3468: HYDROGEN AIR STEAM FLAMMABluTY LIMITS NUREG/CR4826 V01: SEISMIC MARGIN REVIEW OF THE MAINE AND COMBUSTON CHARACTERISTICS IN THE FITS VESSEL YANKEE ATOMO POWER STATION Volume 1 Summary Report. NUREG/CR-4301: STATUS REPORT ON EQUIPMENT QUAUFICATON NUREG/CR4826 V02: SEISMC MARGIN REVIEW OF THE MAINE ISSUES RESEARCH AND RESOLUTION.

YANKEE ATOMO POWER STATION Volume 2 Systems Analysis. NUREG/CR-4320: THE RELATIONSHIP AND INFLUENCES OF FUEL NUREG/CR-4826 V03. SEISMIC MARGIN REVIEW OF THE MAINE AND COOLANT SYSTEM PROCESSES DURING LWR SEVERE ACCI-YANKEE ATOMIC POWER STATON. Volume 3 Fragility Ana!ysis. DENTS.

Contractor Index 55 NUREG/CR4448: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR-4803: THE POSSIBILITY OF LOCAL DETONATIONS

- A GENERAL ELECTRIC BWR3/ MARK 1. Case Study. DURING DEGRADED-CORE ACCIDENTS IN THE BELLEFONTE NU-NUREG/CR4458: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF CLEAR POWER PLANT.

A WESTINGHOUSE 2-LOOP PRESSURIZED WATER REACTOR. Case NOREG/CR-4825: A PRELIMINARY EVALUATION OF THE ECONOMIC NUR /CR4550 V03: ANALYSIS OF CORE DAMAGE FREQUENCY Tl DENTS

, FROM INTERNAL EVENTS SURRY UNtf 1.

NOREG/CRAS5O V04: ANALYSIS OF CORE DAMAGE FREQUENCY SCIENCE APPLICATIONS INTERNATION AL CORP. (FORMERLY FROM INTERNAL EVENTS. PEACH BOTTOM UNIT 2. SCIENCE NM NUREG/CR4551 V1 DRF: EVALUATION OF SEVERE ACCIDENT NUREG/CR-4616: ROOT CAUSES OF COMPONENT FAILURES RISKS AND THE POTENTIAL FOR HISK REDUCTION-SURRY PROGRAM. Methods And Apphcabons.

POWER STATON. UNIT 1. Draft For Comment NUREG/CR4797: PROGRESS REVIEWS OF SIX SAFETY PARAME-NUREG/CR4700 V1 DRF: CONTAINMENT VENT ANALYSIS FOR TER DISPLAY SYSTEMS.

POSTULATED SEVERE ACCIDENTS. Suny Power Station,Urut 1. Draft UA NAVAL' ACADEMY, ANNAPOLIS, MO NU EG f3: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR4818: TRANSITION RANGE DROP TOWER J-R CURVE A BABCOCK AND WILCOX PRESSURIZED WATER REACTORCase TESTING OF A106 STEEL Study NUREG/CR-4742: MELPROG-PWR/ MOD 1 ANALYSIS OF A TMLB' AC. WESTINGHOUSE ELECTRIC CORP.

CIDENT SEQUENCE. NUREG/CR-4752: COINCIDENT STEAM GENERATOR TUBE RUPTURE NUREG/CR4762: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A WESTINGHOUSE 3 LOOP PRESSURIZED WATER REACTOR. Case AND STUCK-OPEN SAFETY RELIEF VALVE CARRYOVER TESTS.MB-2 Steam Generator Transient Response Test Program.

Study.

1 l

J B

g Intcrnationel Organization index This index lists, in alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed below each country and per-forming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/IA number.

R /A 4 THERMAL MixlNG TESTS IN A SEM1 ANNULAR DOWNCOMER WITH INTERACTING FLOWS FROM COLD LEGS 57

1 l

l l

[

l

Lirnsed Fallity Ind:x This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

AMn W. Vogue Nudear Plant, Urut 1, Georga NUREG1137 SOS STNS$29 Palo Verde Nuclear Stanon, Una 2, Anzona NUREGa57 S11 S 424 Power Co. P@c Serwce Co.

B424 AMn W Vogue Nudear Piart Una 1, George NUREG1137 S06 STN&S30 Palo Verde Nuclear Stanon, Und 3, Atuons NUREGa57 $11 Power Co. P@c Sarwce Co.

9 424 AMn W Vogne Nudeer Piant, Und 1. Georga NUREG1237 STNM530 Palo Verde Nuclear Staton, Una 3, Anzone NUREG1248 Power Co. P@c Senace Cct S 424 AMn W. Vogue Nudeer Plant Und 1, Georga NUREG1247 S 277 Peach Bonom Atome Power Staton, Una 2, NUREG/CR4550 V04 S 425 AMn Vogne Nudear Plant, Una 2. Georga NUREG1137 SOS S 277 Peach Bottom Stanon, Und 2, NUREG/CR4696 AMn Vogne Nuclear Plant, Und 2 Georga NUREG-1137 S06 M&M 50425 2 278 % WWL Beever Vasey Power Stanon Una 2. Du@esne NUREG1057 SO4 N k 54412 Benetonte Nudear Plant, Una 1, Tennessee Co. of New Hampshr 50438 NUREG/CR4741 50444 Seabrook Nudeer Stanon, Und 2, Pec Sennce NUREG/CR4552 Vaney Authonty Beestonte Nuclear Plant, Una 1, Tennessee NUREG/CR4803 Co. cl New Hampshr S 438 S400 Sheeton Hams NJclear Power Plant, Und 1 NUREG1240 Vasey Authonty W439 Benefonte Nuclear Plant Una 2 Tennessee NUREG/CR4741 Can$ne Power & Ught C Vasey Authonty ST450498 South Texas Protect, Unn 1, Houelon Ughtn0 & NUREG4781 S02 S 439 BeRefonte Nuclear Plant, Und 2, Tennessee NUREG/CR4803 Power Cct

~

STN#454 B ta und 1, Commonweanh Edson Co. NUREGM78 S08 p $

C co jS8 08 54280 Suny Power Stanon, Und 1, Viryne Electric & NUREG/CR4550 V03 54247 Point Sta tput 2, Consondated Edson NUREG/CR4524 280 Surry Power Stanon, Urut 1. Vrgne Elecinc & NUREG/CR4551 V1 ORF 54280 Suny Power Stanon, Una 1, Vrgrus Electnc & NUREG/CR4700 V1 DAF ank A Power Mane Yankee Akmc Power Plant, Mare NUREG/CR4826 V02 Power Co.

54309 S 281 Siry Power Stanon, Una 2, Vrpna Elecinc & NUREG/CR4868 Yankee Atomic Power Co.

54309 Mane Yankee Atomc Power Plant, Mano NUREG/CR4826 V03 Power Co.

54320 Three lAie Island Nudeer Stanon, Unit 2, NUREG4683 S02DRF ER Yankee Atome Power Co.

STNW528 Pelo Verde Nudear Stanon, Und 1, Atuona NUREGG57 S11 General P@c Unhoe Putec Semte Co. S 252 Urw. of New Mexco Research Reactor NUREG1224 59

U, S. C0VCoh%%f 9eInfInc crrItg i1997 181 68 2 s 60096

I kE*O97 NUMsE,e astesgaer 67 I/OC. edW Vdf Afo . 8f e8F)

NRC 60RM 336 U S. NUCLEE3 KE tuLATOR T COMMIS$sO4 h3x BIBUOGRAPHIC DATA SHEET NUREG-0304,Vol.12,f No. 1 ut N.TRu ON ON TRE REvERit 3 LE AVE BLANE 3 TITLE AND TITLE Regula ry and Technical Reports (Abstract Index Journal)

Compilat n for First Quarter 1987 . O.,ER,,0,, p M,LETED January-M ch MONTR , EAR S AUTRORi$1 .

6 @ REPORT ISSUED vEAR MO TRy l May g 1987 7 PEhFORM6NG ORGAN 12 ATION E AND MAILING ADDRE$$ tincieselg Cadet 8 PROJECTsigmORE UNii NUMSER Division of Publ ations Services /

Office of Adminis ation and Resources Management **'~o a^~' au"" a U.S. Nuclear Regul ory Comission .!

Washington, DC 20 /

' LING ADORESS trace wdele Copel to YvPE OF REPORT 10 $4N50R6NG ORGAN 62 ATION NAME AND

/ Reference Same as 7, above. / b PERIOO COVEREO flacsys,,, denys) j January - March 1987 12 SUPPLEMENT AR Y NOT ES tJ ASSTR ACT (200 weres er 'esat This journal includes all form reports i he NUREG series prepared by the NRC staff and contractors; proceedings of nferen and workshops; as well as international agreement reports. The entries this pompilation are indexed for access by title and abstract, secondary report n . er, ersonal author, subject, NRC organization for staff and international agreements . tractor, international organization, and licensed facility.

\

/

\

to DOCUMENT AN ALvliS - a RE vWe DS DESCR+PTORS t t A v alL A8t LIT Y STATEMENT compilation abstract inde l'nlimited 16 SECURITY CL A$$1FICATION Iin,e pages

. iOENTI,lE Rs.OPEN aEOTEaMi -

Unclassified iTnos reperto Unclassified 17 NUMsE R OF P AGES 16 PRICE

l NUCLEA2 KE LAT 3Y OMMISOON T E JD j WASHINGTON, D.C. 20566 N.

n,.=%%"'Jn?'!1.

1 Main Citations end Abstracts 12 0 5% 0 7 8 R 7 7 1 1ANIAC19L190 US NRC-GARM-ADM q, DIV 0F PUB SVCS 2 Secondary Report POLICY & PUB MGT BR-PDR NUREG Number index yf}NGTON DC 20555 3 Personal Author index

~4 Subject index 5 NRC Originating Organization Index (Staff Reports) 6 NRC Originating Organization Index (International Agreements) 7 NRC Contractor Sponsorindex j 8 Contractor index g International Organization Index E Licensed Index Facility - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

ML20214N796

Contents

Text

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu