ML20236S361

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Third Quarter 1987,July-September

ML20236S361
Person / Time
Issue date:	11/30/1987
From:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
To:
References
NUREG-0304, NUREG-0304-V12-N03, NUREG-304, NUREG-304-V12-N3, NUDOCS 8711250102
Download: ML20236S361 (60)

Text

NUREG-0304  !

Vol.12, No. 3 l

Regulatory and Technical Reports

{ Abstract Index Journal?

Compilation for Third Quarter 1987 July - September U.S. Nuclear Regulatory Commission l Office of Administration and Resources Management pm acog I hYk.] ....

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NUREG-0304 , i' Vol.12, No. 3 i; i

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Com allation for > .l Thirc Quarter 1987 July - September Date Published: November 1987 Policy and Publications Management Branch Division of Publications Services Office of Administration and Resources Management i U.S. Nuclear Regulatory Commission Washington, DC 20555 , s

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, \ l 1 CONTENTS Preface.. . . . . . .. . .... .. ... .. . . . .... . .. . .. . .... v

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( Main Citations and Abstracts . . .. . . . ........ .. .. ... . ... . ... .......... 1 y

• Staff Reports

• Conference Proceedings t

• Contrrtetor Reports 3

• International Agreement Reports Secondary Report Number Index . . . . .. . . .. ... . . .. . . . 2 Personal Author Index ... . .. ... .. . ..... ... . . ... . . 3 SubjecMndex ...... ................ .... . . .. ........ . .... . . . 4 NRC Gifnating Organization index (Staff Reports) . .. . ..... . ... . . . . ...... 5 NRC Onginating Organization Index (International Agreements) . . . ......... . .. .. . . 6 NRC Contract Sponsor index (Contractor Reports) . . ... . ... ...... ... . .. . ... . . 7 Contractor index . .... ...... .. . . .... . ..... ... .. ....... ............... . . . 8 International Organizatk n inifex . . . ... .. ... .... .. . . . .... . . . . . 9 Licensed Facility index ? . .... .. .. .. . . . . ....... . . . . . . ... . . 10

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PREFACE p -

Thh mpilation consisks of bibliographic data and abstracts for the formal regulatory and technical repot issued by the ll.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 537 U.S. Nuclear Regulatory Commission

,/ Washington, D.C. 20555 T% main citationr, 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 e Licensed Facility Index A detailed explanation of the entries precedes each index.

The bibliographic elements of the main citations are the following:

i

< Staff Report l

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

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

I' r

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

, Euthor,%) drte rep 9rt was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) tbn microfiche address (for internal NRC use).

)

Conference Report _

NUREG/CP-Mi/i EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINEtiRING 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: 9TUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORC-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.: BENNETT, P.R.

Sandia Laboratories. ruv 1981. '00 pp. 8107010449. SAND 80-0929. 08912:242.

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

I v

1' intemational 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 Govemment 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 cf Documents, to the following address:

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

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

In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/lA is used for intemational 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 i

l

_ _ _ _ . - _ _ _ - - _ - - - - - . - - . - - - - J

1 Main Citations and Abstracts The report listings in this compilation are wranged by report number, 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.

NUREG-0020 V11 N01: LICENSED OPERATING REACTORS rences reported by the Agreement States. The report also con.

STATUS

SUMMARY

REPORT. Data As Of December tains information updating some previously reported abnormal 31,1986.(Gray Book 1) ROSS.P.A. Drvision of Computer & Tele- occurrences.

communications Services (Post 870413). September 1987.

450pp. 8710080382. 42994:112. NUREG-0304 V12 N02: REGULATORY AND TECHNICAL RE-The OPERATING UNITS STATUS REPORT . LICENSED OP. PORTS (ABSTRACT INDEX JOURNAL). Compilation For ERATING REACTORS provides data on the operation of nucle. Second Quarter 1987, April-June.

• Division of Publication Serv.

at units as timely and accurately as possible. This information is ices (Post 870413). August 1987. 60pp. 8709040390.

collected by the Office of Administration and Resources Man. 42522.207.

agement from the Headquarters staff of NRC's Office of En. This journal includes all formal reports in the NUREG senes forcement (OE), from NRC's Regional Offices, and from utilities. prepared by the NRC staff and contractors; proceedings of con.

The three sections of the report are: monthly highlights and sta. ferences and workshops; as well as intemational agreement re-tastics for commercial operating units, and errata from previously ports. The entries in this compilation are indexed for access by reported data; a compilation of detailed information on each title and abstract, secondary report number, personal author, unit, provided by NRC's Regional Offices: OE Headquarters and subject, NRC organization for staff and international agree-the utihties; and an appendix for miscellaneous information such ments, contractor, international organization, and licensed facili-as spent fuel storage capability, reactor-years of experience and ty.

non- power reactors in the U.S. It is hoped the report is helpful to all agencies and individuals interested in maintaining an NUREG-0327 R04: OWNERS OF NUCLEAR POWER PLANTS. Percentage Ownership Of Commercial Nuclear Power awareness of the U.S. energy situation as a whole. Plants Dy Utility Companies. WOOD,R.S. Office of Nuclear Re-NUREG-0040 Vit N02: LICENSEE CONTRACTOR AND actor Regulation, Director (Post 870411). August 1987. 38pp.

VENDOR !NSPECTION STATUS REPORT. Quarterly 8708240084. 42316:072 {

Report April-June 1987.(White Book)

• Division of Reactor in- The repor; indicates percentage ownership of commercial nu- j' spection & Safeguards (Post 870411). August 1987. 150pp. clear power plants by utility companies. The report includes all 8708270427. 42391:011. plants operating, under construction, docketed for NRC safety This penodical covers the results of inspections performed by and environmental reviews, or under NRC antitrust review, but the NRC's Vendor inspection Branch that have been distnbuted does not include those plants announced but not yet under to the inspected organizations dunng the period from April 1987 review or those plants formally cancelled. Part i of the report through June 1987. Also, included in this issue are the results lists plants alphabetically with their associateJ applicants or li-of certain inspections performed pnor to April 1987 that were censees and percentage ownership. Part il list applicants or h-not included in previous issues of NUREG-0040. censees alphabetically with theit associated plants and percent-age ownership Part I also indicates which plants have recerved NUREG-0090 V09 N04: REPORT TO CONGRESS ON ABNOR-operating heenses (OLs).

MAL OCCURRENCES. October-December 1986.

• Office for Analysis & Evaluation of Operational Data Director. July 1987. NUREG-0430 V07 NO2: LICENSED FUEL FACILITY STATUS 150pp. 8708170105. 42171:217. REPORT. inventory Difference Data. July December 1986(Gray Section 208 of the Energy Reorganization Act of 1974 identi. Book II)

• Office of Nuclear Matenal Safety &

fees an abnormal occurrence as an unscheduled incident or Safeguards, Director. August 1987. 14pp. 8709090409.

event which the Nuclear Regulatory Commission determines to 42566.252.

be significant from the standpoint of pubhc health and safety NRC is committed to the periodic publication of heensed fuel and requires a quarterly report ei such events to be made to facilities inventory difference data, following agency review of Congress. This report covers the period October 1 to December the information and completion of any related NRC investiga-31,1986. Dunng the report penod, there were three abnormal tions. Information in this report includes inventory difference occurrences at the nuclear power plants heensed to operate. data for active fuel fabrication facilities possessing more than The events were (1) the loss of low pressure service water sys- one effective kilogram of high enriched uranium, low ennched tems at Oconee, (2) degraded safety systems due to incorrect uranium, plutonium, or uranium 233. i torque switch settings on Rotork motor operators at Catawba and McGuire Nuclear Stations, and (3) a secondary system pipe NUREG-0525 R13: SAFEGUARDS

SUMMARY

EVENT LIST break resulting in the death of four persons at Surry Unit 2. (SSEL).

Division of Safeguards & Transportation (Post There were six abnormal occurrences at the other NRC heens- 870413) July 1987. 85pp. 8708240212. 42316.282.

ees. One involved release of amencium-241 inside a waste stor- The Safeguards Summary Event List provides bnet summa <

age building at Wright-Patterson Air Force Base; three involved nes of hundreds of safeguards related events involving nuclear medical misadministration, one therapeutic and two diagnostic; matenal of facilities regulated by the U.S. Nuclear Regulatory one involved a suspens'on of heense for servicing teletherapy Commission. Events are described urder the categones: bomb-and radiography units; and one involved an immediate effective related, intrusion, missing / allegedly stolen, transportal.on-relat-order modifying license and order to show cause issued to an ed, tampenng/ vandalism, arson, firearms related, radiological industrial radiography company. There were no abnormal occur- sabotage, nonradiological sabotage, and miscellaneous. Infor-1 .

2 Miln Cit:tlins cnd Abstracts mation in the event desenptions was obtained from official NRC outstanding issues remaining after issuance of the Safety Eval-reports. uation Report and Supplement No.1. Supplement No. 3, issued in May 1987 addressed and resolved some of the outstanding NUREG-0540 V09 N05: TITLE LIST OF DOCUMENTS MADE issues remaining after issuance of the Safety Evaluation Report PUBLICLY AVAILABLE. May 1 31,1987.

• Division of Publica-and Supplement Nos.1 and 2. Supplement No. 4 also address-tion Services (Post 870413). July 1987. 346pp. 8707210755.

es and resolves some of the outstanding issues remaining after This iss nce of the Safety Evaluation Report and Supplement Nos.

ument is a monthly publication containing desenp- j, tions of information recorved and generated by the U.S. NRC.

This information includes (1) cacketed material associated with NUREG-0837 V07 N01: NRC TLD DIRECT RADIATION MONI-Crvilian nuclear power plants and other uses of radioactive ma-TORING NETWORK. Progress Report, January-March 1987.

terials, and (2) nondocketed matenal received and generated by JANO,J.; MCNAMARA.M.; COHEN,L, Region 1, Office of Direc-NRC pertinent to its role as a regulatory agency. The following tor. July 1987. 226pp. 8708100398. 42095:325.

indexes are included: Personal Author Index, Corporate Source This report provides the status and results of the NRC Ther-Index, Report Number Index, and Cross Reference to Pnncipal moluminescent Dosimeter (TLD) Direct Radiation Monitonng Documents Index.

Network. It presents the radiation levels measured in the vicinity NUREG 0540 V09 N06: TITLE LIST OF DOCUMENTS MADE t NRC licensed facility sites throughout the country for the first PUBLICLY AVAILABLE. June 1 30,1987.

• Division of Publica. quarter of 1987, t Se es (Post 870413). August 1987. 505pp. 8709030196.

NUREG-0885106: U.S. NUCLEAR REGULATORY COMMISSION See NUREG.0540,V09,N05 abstract. POLICY AND PLANNING GUIDANCE 1987.

• Commissioners.

September 1987, 58pp. 8710010440. 42883:183.

NUREG-0750 V24102: INDEXES TO NUCLEAR REGULATORY The purposes of the Policy and Planning Guidance document COMMISSION ISSUANCES. July-December 1986.

• Division of are: to set forth the regulatory approach of the Nuclear Regula-Publication Services (Post 870413). July 1987. 100pp. tory Commission and provide supporting pnnciples to the ap-8708270324.42389:007. proach; to state the major policies and planning objectives of Digests and indexes for issuances of the Commission, the the Commission; and to provide a common basis for the devel-Atomic Safety and Licensing Appeal Panel, the Atomic Safety opment of programs, the establishment of prionties, and the al-and Licensing Board Panel, the Administrative Law Judge, the location of resources.

Directors' Decisions, and the Denials of Petitions for Rulemak-ing are presented. NUREG-0936 V06 N02: NRC REGULATORY AGENDA.Ouarterty

' Pn une 1987 Dv sion Rules & ecords (Post NUREG-0750 V25 N01: NUCLEAR REGULATORY COMMISSION 704 J ISSUANCES FOR JANUARY 1987.Pages 162.

• Division of Publication Services (Post 870413). June 1987. 71pp. The NRC Regulatory Agenda is a compilation of all rules on which the NRC has proposed or is considenng action and all pe om ru aWg wM haw Wn mM h h h L89 l i u n es of th Commission, the Atomic. Safety and Li- mission and are pending disposition by the Commission. The censing Appeal Panel, the Atomic Safety and Licensing Board Panel, the Administrative Law Judge, and NRC program offices regulatory agenda is updated and issued each quarter.

are presented.

NUREG 0940 V06 NO2: ENFORCEMENT ACTIONS:SIGNIFICANT NUREG 0750 V25 N02: NUCLEAR REGULATORY COMMISSION ACTIONS RESOLVED.Ouarterly Progress Report, April-June ISSUANCES FOR FEBRUARY 1987.Pages63-128.

• Division of 1987.

• Office of Enforcement (Post 870413). August 1987.

Publication Services (Post 870413). July 1987, 74pp. 399pp. 8709090464. 42564:282.

8708060402.42072:330. This comp 4 tion summarizes significant enforcement actions See NUREG-0750,V25,N01 abstract. that have been resolved dunng one quarterly penod (April -

June 1987) and includes copies of letters, Notices, and Orders NUREG 0750 V25 NO3: NUCLEAR REGULATORY COMMISSION sent by the Nuclear Regulatory Commission to licensees with ISSUANCES FOR MARCH 1987.Pages 129-266.

• Division of respect to these enforcement actions. It is anticipated that the Publication Services (Post 870413). August 1987. 148pp. information in this publication will be widely disseminated to B709110219. 42626:269. managers and employees engaged in activities licensed by the See NUREG-0750,V25,N01 abstract.

NRC, so that actions can be taken to improve safety by avoid-NUREG-0750 V25 N04: NUCLEAR REGULATORY COMMISSION ing future violations similar to thosa described in this publica-ISSUANCES FOR APRIL 1987.Pages 267-416.

• Division of tion.

8 08 359 990 0

• NUREG-1002 SO4: SAFETY EVALUATION REPORT RELATED See NUREG-0750,V25,N01 abstract' TO THE OPERATION OF BRAIDWOOD STATION. UNITS 1 AND 2. Docket Nos. 50-456 And 50 457.(Commonwealth Edison NUREG 0781 SO4: SAFETY EVALUATION REPORT RELATED Company)

• Division of Reactor Projects - til,lV,V & Special TO THE OPERATION OF SOUTH TEXAS PROJECT, UNITS 1 Projects (Post 870411). July 1987. 22pp. 8708040187, AND 2. Docket Nos. 50 498 And 50-499. (Houston Lighting And 42039:049.

Power Company)

• Division of Reactor Projects lil,lV V & Spe- In November 1983, the staff of the Nuclear Regulatory Com-cial Projects (Post 870411). July 1987.162pp. 8708240094. {

mission issued its Safety Evaluation Report (NUREG 1002) re- )

42323:002. garding the application filed by the Commonwealth Edison Com-The Safety Evaluation Report issued in Apnl 1986 provided pany, as applicant and owner, for a license to operate Braid-the results of the NRC staff's review of the Houston Lighting wood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457).

and Power Company's application for licenses to operate the The first supplement to NUREG 1002 was issued in September South Texas Project. The facility consists of two pressunzed 1986; the second supplement to NUREG.1002 was issued in water reactors located in Matagorda County, Texas. Supple. October 1986; the third supplement to NUREG-1002 was issued ment No.1, issued in September 1986 updated the information in May 1987. This fourth supplement to NUREG 1002 reports contained in the Safety Evaluation Report and addressed the the status of certain items that remained unresolved at the time ACRS Report issued on June 10, 1986. Supplement No. 2, Supplement 3 was published. The facility is located in Reed issued in January 1987 addressed and resolved some of the Township, Will County, Illinois.

l

Main Cit';tions End Abstrrets 3 NUREG-1047 S06: SAFETY EVALUATION REPORT RELATED under the pilot program for ISAP. This report indicates how 82 TO THE OPERATION OF NINE MILE POINT NUCLEAR topics selected for review were addressed and presents the STATION, UNIT 2. Docket No. 50-410.(Niagara Mohawk Power staffs recommendations regarding the corrective actions to re.

Corporation.et al)

• Division of Reactor Projects - t/II (Post solve the 82 topics and other actions to enhance plant safety.

870411). Jufy 1987. 65pp. 8707230283. 41879:171. The report is being issued in draft form to obtain comments This report supplements the Safety Evaluation Report from the licensee, nuclear safety experts, and the Advisory (NUREG 1047, February 1985) for the application filed by Niag- Committee for Reactor Safeguards. Once those comments have era Mohawk Power Corporation, as applicant and co-owner, for been resolved, the staff will present its positions, along with a the license to operate Nine Mile Point Nuclear Station, Unit 2 fong-term implementation schedule from the licensee, in the (Docket No. 50-410). It has been prepared by the Office of Nu- final version of this report.

clear Reactor Regulation of the U.S. Nuclear Regulatory Com-NUREG 1185 V02 DRFT: INTEGRATED SAFETY ASSESSMENT L mission. The facility is located near Oswego, New York. This REPORT. Integrated Safety Assessment Program.Haddam Neck report supports the issuance of the full-power license for Nine ,

Mile Point Nuclear Station, Unit No. 2. PlantDocket No 50-213. (Connecticut Yankee Atomic Power Company)

• Division of Reactor Projects Ill.IV,V & Special f NUREG 1057 S06: SAFETY EVALUATION REPORT RELATED Projects (Post 870411). Jufy 1987. 715pp. 6708260092. i TO THE OPERATION OF BEAVER VALLEY POWER 42386.128 STATION, UNIT 2. Docket ' No. 50-412.(Duquesne Light See NUREG 1185,V01 DRFT abstract.

Company,et af)

• Division of Reactor Projects 1/ll (Post 870411). August 1987. 83pp. 8709090412. 42566:356. NUREG-1194: CONSTRUCTION APPRAISAL TEAM INSPECTION Supplement No. 6 to the Safety Evaluation Report for the ap. RESULTS ON WELDING AND NONDESTRUCTIVE EXAMINA-TION ACTIVITIES. WU P.C.; SHAABAN H.l. Division of Reactor plication filed by Duquesne Light Company, et al., for license to operate the Beaver Valley Power Station, Unit 2 (Docket No. Inspection & Safeguards (Post 870411). September 1987.57pp.

50-412), located in Beaver County, Pennsylvania, has been pre. 8710050558. 42905:227.

pared by the Office of Nuclear Reactor Regulation of the Nucle. This report summanzes data and findings of deficiencies and discrepancies in welding and nondestructive examination (NDE) ar Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation of (1) additional information sub. activities identified by the U.S. Nuclear Regulatory Commission mitted by the licensees since Supplement No. 5 was issued, Construction Appraisal Team (CAT) during its inspection of 11 and (2) matters that the staff had under review when Supple. plants. The CAT reviewed selected welds and NDE packages in its inspection of the following plant areas: piping and pipe sup.

ment No. 5 was #ssued.

ports and/or restraints; modification and installation of reactor NUREG 1144 R01: NUCLEAR PLANT AGING RESEARCH intemals; electrical installations and electrical supports; instru-(NPAR) PROGRAM PLAN. VORA.J.P. Division of Engineerin9 mentation tubing and supports; heating, ventilation, and air con-(Post 870413). September 1987. 100pp. 8710090155 ditioning (HVAC) systems and supports; fabncation and erection 43010482. of structural steel; fabrication of refueling cavity and spent fuel The nuclear plant aging research desenbed in this plan is in- pool liner; containment liner and containment penetrations; and tended to resolve issues related to the aging and service wear fire protection systems. The CAT inspected both structural of equipment and systems and major components at commer- welds and pressure-retaining welds and reviewed welder qualifi-cial reactor facilities and their possible impact on plant safety. cation test records and welding procedure documents for code Emphasis has been placed on identification and chargetenza- compliance. The NDE activities that were evaluated included tion of the mechanisms of matenal and component degradation visual examination, magnetic particle examination, liquid pene-dunng service and evaluation of methods of inspection, surveil- trant examination, ultrasonic examination, and radiographic ex-lance, condition monitonng, and maintenance as means of miti- amination of welds.

gating such effects. Specifically, the goals of the program are as follows: (1) to identify and charactenze aging and service NUREG 1213 R01: PLANS AND SCHEDULES FOR IMPLEMEN-wear effects which, if unchecked, could cause degradation of TATION OF U.S. NUCLEAR REGULATORY COMMISSION RE-equipment, systems, and maior components and thereby impair SPONSIBILITIES UNDER THE LOW LEVEL RADIOACTIVE plant safety, (2) to identify methods of inspection, surveillance, WASTE POLICY AMENDMENTS ACT OF 1985 (P.L.99-240).

and monitonng. or of evaluating residual life of equipment, sys- DUNKELMAN.M M. Division of Low Level Waste Management tems, and major components, which will ensure timely detection & Decommissioning (Post 870413). August 1987. 100pp.

of significant aging effects poor to loss of safety function, and 8709090253. 42564:182.

(3) to evaluate the effectiveness of storage, maintenance, The purpose of this document is to make available to the repair, and replacement practices in mitigating the rate and States and other interested partses, the plans and schedules for extent of degradation caused by aging and service wear. the U.S. Nuclear Regulatory Commission's (NRC's) implementa-tion of its responsibilities under the Low-Level Radioactive NUREG-1185 V01 DRFT: INTEGRATED SAFETY ASSESSMENT ask M cy hem M d N M MM REPORT. integrated Safety Assessment Pr09 ram.Haddam Neck (LLRWPAA) This document identifies the provisions of the Plant. Docket No. 50-213. (Connecticut Yankee Atomic Power LLRWPAA which affect the programs of the NRC, identifies Company) Division of Reactor Protects - til,lV,V & Special what the NRC must do to fulfill each of its requirements under Pro e s (Post 870411). July 1987. 141pp. 8708260003 the LLRWPAA, and establishes schedules for carrying out these The Integrated Safety Assessment Program (ISAP) was initiat-ed in November 1984, by the U.S. Nuclear Regulatory Commis- 6' sion to conduct integrated assessments for operating nuclear power reactors The integrated assessment is conducted on a NUREG-1232 V01: SAFETY EVALUATION REPORT ON TEN.

plant specific basis to evaluate all licensing actions, licensee ini- NESSEE VALLEY AUTHORITY. Revised Corporate Nuclear 18ated plant improvements and selected unresolved genenc/ Performance Plan.

• Office of Special Projects. July 1987.64pp.

safety issues to establish implementation schedules for each 8710080392. 42995.225.

item. In addition, procedures will be establ8shed to allow for a This Safety Evaluation Report on the information submitted by penodic updating of the schedules to account for licensing the Tennessee Valley Authonty (TVA) in its revised Corporate tssues that anse in the future This report documents the review Nuclear Performance Plan, through Revision 4. has been pre-of Haddam Neck Plant, operated by Connecticut Yankee Atomic pared by the U,S Nuclear Regulatory Commission staff The Power Company, which is one of two plants being reviewed TVA Corporate Nuclear Performance Plan addresses those cor-l

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_-__mm.___ _ _ _ _

4 Main Cititlans and Abstracts porate concems identified by the staff in its letter to TVA dated September 17,1985. On the basis of its review, the staff finds NUREG 1274: REVIEW PROCESS FOR LOW-LEVEL RADIOAC.

TiVE WASTE DISPOSAL LICENSE APPLICATION UNDER TVA's revised Corporate Nuclear Performance Plan (Revision 4) acceptable. LOW LEVEL RADIOACTIVE WASTE POLICY AMENDMENTS ACT. PITTIGILO.C.L. Division of Low Level Waste Management NUREG-1251 DRFT FC: IMPLICATIONS OF THE ACCIDENT AT 70 240196. 4 319 18 CHERNOBYL FOR SAFETY REGULATION OF COMMERCIAL NUCLEAR POWER PLANTS IN THE UNITED STATES. Draft This document identifies and describes the U.S. Nuclear Reg-A i Given. Aug st ulatory Commission's (NRC's) process for licensing a low level 79 8 09040049. 250 7 radioactive waste disposal facility within the time required by the This draft report issued for comment was prepared by the Nu- Low-Level Radioactive Waste Policy Amendments Act of 1985.

clear Regulatory Commission (NRC) staff to assess the implica- This document also estimates the level of effort and expertise tions of the accident at the Chernobyl nuclear power plant as that is needed to review a license application within the required they relate to reactor safety regulation for commercial nuclear time. It is intended to be used by the NRC staff as well as power plants in the United States. The facts used in this as. States and interested parties to provide a better understanding sessment have been drawn from the U.S. fact. finding report of what the NRC envisions will be involved in licensing a low-(NUREG 1250) and its sources. level radioactive waste facility.

NUREG-1253: TECHNICAL SPECIFICATIONS FOR NINE MILE NUREG 1275: OPERATING EXPERIENCE FEEDBACK REPORT -

POINT NUCLEAR STATION, UNIT 2. Docket No. 50 410.(Niagara NEW PLANTS. Commercial Power Reactors. DENNIG,R.L.;

Mohawk Power Corporation,et al)

• Divisson of Reactor Projects O'REILLY,P.D. Office for Analysis & Evaluation of Operationa!

l/11 (Post 870411). July 1987. 526pp. 8707210703. 41832:032. Data, Director. July 1987. 285pp. 8709040358. 42522.267.

The Nine Mile Point Nuclear Station, Unit 2, Technical Specifi- This report documents a detailed review of the cause of un-cations were prepared by the U.S. Nuclear Regulatory Commis- planned events during the early months of licensed operation soon to set forth the limits, operating conditions, and other fe-for plants licensed between March 1983 and April 1986. The quirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health major lessons and corrective actions that appear to have the and safety of the public. greatest potential for improving the effectiveness of plant star-tups are provided for consideration through the operating expe- i NUREG 1255: TECHNICAL SPECIFICATIONS FOR SOUTH rience feedback programs and activities of the industry and the i TEXAS PROJECT, UNIT 1. Docket No. 50-498.(Houston Lighting star and Power Cornpany)

• Division of Reactor Projects lil,lV,V &

Special Projects (Post 870411). August 1987. 445pp. NUREG 1276: TECHNICAL SPECIFICATIONS FOR BRAIDWOOD 8709090260. 42577:044. STATION. UNITS 1 AND 2. Docket Nos. 50-456 And 50-The South Texas Protect, Unit No.1, Technical Specifications 457.(Commonwealth Edison Company)

• Division of Reactor were prepared by the U.S. Nuclear Regulatory Commission to Projects - lilIV,V & Special Projects (Post 870411). July 1987, set forth the limits, operating conditions, and other requirements 493pp. 8707290233. 41952:082.

applicable to a nuclear reactor facility as set forth in Section The Braidwood Station, Units 1 and 2, Technical Specifica-50.36 of 10 CFR 50 for the protection of the health and safety tions were prepared by the U.S. Nuclear Regulatory Commis-of the public. Sion to set forth the limits, operating conditions, and other re-quirements applicable to a nuclear reactor facility as set forth in NUREG-1264: CONTAINMENT INTEGRITY RESEARCH PRO. Section 50.36 of 10 CFR Part 50 for the protection of the health GRAM PLAN.

• Division of Engineenng (Post 870413). August and safety of the public.

1987. 38pp. 8709160104. 42693:130.

This report presents a plan for research on the question of NUREG-1278: VOGTLE UNIT 1 READINESS containment performance in postulated severe accident scenar- REVIEW. Assessment Of Georgia Power Company Readiness los. It focuses on the research being performed by the Structur- Review Pilot Program. LEWIS,G. Division of Licensee Perform-al and Seismic Engineering Branch, Division of Engineenng, ance & Quality Evaluation (Post 870411). September 1987, Office of Nuclear ' Regulatory Research. Summaries of the plans 29pp. 8710080317. 42990:320.

for this work have previously been published in the " Nuclear Georgia Power Company (GPC) performed a readiness review Power Plant Severe Accident Research Plan" (NUREG4900). at Vogtle Unit 1 as a pilot program. The pilot program was a This report provides an update to reflect current status. This plan provides a summary of results to date as well as an outline new and innovattve approach for the systematic and disciplined of planned activitses and milestones to the contemplated com- review, with senior management involvement, of GPC's imple- 4 pietion of the program in FY 1989. mentation of design, construction, and operational readiness processes. The program's principal objective was to increase {

NUREG-1266 V01: NRC SAFETY RESEARCH IN SUPPORT OF the level of assurance that quality programs at Vogtle Unit 1 )

REGULATION 1986.

• Office of Nuclear Regulatory Research, have been accomplished in accordance with regulatory require-Director (Post 860720). September 1987. 66pp. 8710080407. ments. This report assesses the effectiveness of the GPC's 42991:067. readmess review pilot program (RRPP) at Vogtle Unit 1. It in- j This report is the second in a senes of annual reports re- cludes (1) an overview of what was experienced dunng the pro- '

sponding to congressionalinquines as to the utilization of nucie, gram's implementation, (2) an assessment of how well program ar regulatory research. NUREG 1175, "NRC Safety Research in objectives were met, and (3) lessons leamed on the future use Support of Regulation," published in May 1986, reported major of the readiness review concept. Overall, GPC and the NRC research accomplishments between about FY 1980 & FY 1985. staff believe that the RRPP at Vogtle Unit 1 was a success and This report narrates the accomplishments of FY 1986 and does that the program provided significant added assurance that not restate earlier accomplishments. Earlier research results are Vogtle Unit 1 licensing commitments and NRC regulations have mentioned in the contert of current results in the interest of been adequately implemented Although altering the NRC 16-continuity. Both the direct contnbutions to scientific and techni- censing review process for the few plants still in the construc-cal knowledge and their regulatory applications, when there has tion pipeline may not be appropnate, licensees may benefit sig-been a definite regulatory outcome during FY 1986, have been nificantly by performing readinen reviews on their own initiative desenbed. as GPC did for Vogtle.

Miln Citati:ns cnd Abstracts 5 NUREG 1279: TECHNICAL SPECIFICATIONS FOR BEAVER in the Inconel 600 strain gage encapsulation and the subse-VALLEY POWER STATION, UNIT 2.0ocket No. 50 412.(Du. Quent burnout of the strain gage elements. For DDs the degra-quesne Light Company.et al)

• Dtvision of Reactor Projects t/11 dation or failure cause seems to be excessive loads. The degra-(Post 870411). August 1987, 424pp. 8709090526, 42575:340. dation cause for most of the FDGs and LLDs seems to be i The Beaver Valley Power Station, Unit 2, Technical Specifica- either steam / water erosion or mechanical abrasion of sensor tion were prepared by the U.S. Nuclear Regulatory Commission tips, which are made of sapphire. However, soma of the FDG to set forth the limits, operating conditions, and other require- tips were found to be cracked also. The corrective actions are ments applicable to a nuclear reactor facility as set forth in Sec. being directed toward finding out what are the pnmary causes tion 50.36 of 10 CFR 50 for the protection of the health and for the instrument degradation or failure and what should be safety of the pubhc. done 13 prevent these degradations or failures from continuing or recurnng.

NUREG-1281: EVALUATION OF THE QUALIFICATION OF SPERT FUEL FOR USE IN NON POWER REACTORS.

• Divi. NUREG-1285: NRC STAFF EVALUATION OF THE GENERAL sion of Reactor Projects ill,lV,V & Special Projects (Post ELECTRIC COMPANY NUCLEAR REACTOR STUDY (" REED 870411). August 1987. 89pp. 8708240112. 42319:001. REPORT"). VIRGILIO,M.J.; BEVAN,R. Office of Nuclear Reactor This report summan2es the U.S. Nuclear Regulatory Commis- Regulation, Director (Post 870411). July 1987. 63pp.

sion staff's evaluation of the qualification of the stainless steel- P708060372. 42071:199.

clad uranium / oxide (UO2) fuel pins for use in non-power reac- In 1975, the General Electnc Company (GE) published a Nu-tors. The fuel pins were onginally procured in the 1960's as part clear Reactor Study, also referred to as "the Reed Report," an of the Special Power Excursion Reactor Test (SPERT) program, intemal product improvement study. GE considered the docu-Argonne National Laboratory (ANL) examined 600 SPERT fuel ment "propnetary" and thus, under the regulations of the Nucle-pins to venty that the pins were produced according to specifi- ar Regulatory Commission (NRC), exempt from mandatory cation and to assess their present condrtion The pins were vis- public disclosure. Nonetheless, members of the NRC staff re-ually inspected under 6X magnification and by X radiographic, viewed the document in 1976 and determined that it did not destructive, and metallograph.c examinations. Spectrographic raise any significant new safety issues. The staff also reached and chemical analyses were performed on the UO2 fuel. The the same conclusion in subsequent re m J. However, in re-results of the qualification examinations indicated that the sponse to recent inquiries about the rep 3r' he staff re-evaluat-SPERT fuel pins meet the requirements of Phillips Specification ed the Reed Report from a 1987 pers,t dve. This re-evalua-No. F 1 SPT and have suffered no physical damage since fabn- tion, documented in this staff report, concluded that (1) there cation. Therefore, the qualification results give reasonable as- are no issues ra Sed in the Rend Report that support a need to surance that the SPERT fuel rods are suitable for use in non- curtail the operation of any GE boihng water reactor (BWR); (2) power reactors provided that the effects of thin-wall defects in there are no new safety issues raised in the Reed Report of the region of the upper end cap and low-density fuel pellets are which the staff was unaware; and (3) although certain issues evaluated for the intended operating conditions. addressed by the Reed Report are still being studied by the NUREG 1282: SAFETY EVALUATION REPORT ON HIGH-URA- NRC and the industry, there is no basis for suspending licensing NIUM CONTENT, LOW ENRICHED URANIUM.ZlRCONIUM HY. and operation of GE BWR plants while these issues are being DRIDE FUELS FOR TRIGA REACTORS. Docket No. 50163.(GA resolved.

Technologies, Incorporated)

• Division of Reactor Projects -

lil,lV,V & Special Projects (Post 870411). August 1987.17pp. NUREG/CP-0088: TRANSACTIONS OF THE 9TH INTERNATION.

AL CONFERENCE ON STRUCTURAL MECHANICS IN REAC-8708240190. 42319:207.

The properties and performance of the TRIGA higher weight TOR TECHNOLOGY. Panel Session JK: Structural And Mechani-cal Engineering Research At The U.S. Nuclear Regulatory Com-percent (w%), low-ennched uranium fuels are compared with mission. BROWZIN.B S. Division of Engineenng (Post 870413).

those of the currently hcensed 8.5 w% fue!s. Neutron physics considerations, matenals properties, irradiation pedormance, fis, July 1987. 265pp. 8707210681. 41830:271.

These transactions of the JK panel session include prepnnts sion product release, pulse heating, and limiting design basis of papers which are hsted in the Second Announcement for the were evaluated. The performance of uranium-zirconium hydnde fuel is substantially independent of uranium content up to 45, 9th International Conference on Structural Mechanics in Reac-w% uranium. The behavior of the proposed 20- and 30-w% ura. tor Technology. These papers represent the body of the JK panel session, " Structural and Mechanical Engineenng Re-nium fuels is indistinguishable from that of the currently ap. search at the U S. Nuclear Regulatory Commission."

proved 8.5-w% uranium fuel. Both the 20-20 and 30 20 urani-urwirconium hydnde fuels are acceptable for use in the GA NUREG/CR 20J0 V06 N6: LICENSEE EVENT REPORT (LER)

Mark F TRIGA reactor. and these two types of fuel are generi- COMPILATION For Month Of June 1987.

• Oak Ridge National cally acceptable for use in other heensed TRIGA reactors, with Laboratory. July 1987,138pp. 8708120301. ORNL/NSIC 200.

the provision that case-by case analyses discuss individual re- 42128.253 actor operating conditions in applications for authonzation to This monthly report contains Licensee Event Report (LER) use them. o erational information that was processed into the LER data NUREG 1284: PROGRAM PLAN FOR CORRECTION OF U.S. IN- file of the Nuclear Safety Information Center (NSIC) dunng the STRMNT DEGRADATION OR FAILURE IN THE UPPER one month penod identified on the cover of the document. The PLENLM TEST FACILITY (UPTF)IN THE FEDERAL REPUBLIC LERs. from which this information is denved, are submitted to OF GERMANY. RHEE.G S.; CHEN,Y.S.; SHOTKIN.L.M. Division the Nuclear Regulatory Commission (NRC) by nuclear power of Reactor & Plant Systems (Post 870413). July 1987. 254pp. plant heensees in accordance with federal regulations. Proce-8708040274.42060:142, dures for LER reporting for revisions to those events occumng j This report documents the investigation of the f ailure or deg- pnor to 1984 are desCnbed in NRC Regulatory Guide 1.16 and radation of some of the advanced two phase flow instruments NUREG 1061, " Instructions for Preparation of Data Entry supplied by the USNRC to the Upper Plenum Test Facility Sheets for Licensee Event Reports." For those events occumr.g (UPTF), including Tie Plate Drag Bodies, Breakthrough Detec- on and after January 1,1984 LERs are being submitted in ac- ,

tors, Loop Drag Disc paddles, Fluid Distnbution Gnd sensors, cordance with the revised rule contained in Title 10 Part 50.73 and Liquid Level Detector sensors. The exact causes for these of the Code of Federal Regulations (10 CFR 50.73 Licensee instrument degradations or failures are not known, but several Event Report System) which was published in the Federal Reg-potential causes have been identified. For DBs and BTDs, the ister (Vo! 48, No 144) on July 26,1983. NUREG.1022, "Li-pnmary mechanism for the degradation seems to be a leakage censee Event Report System Desenption of Systems and l

l

_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ ___ _ ]

6 M:In Citati:ns and Abstrccts Guidelines for Reporting," provides supporting guidance and in- been compiled and reported. Data on solid waste shipments as formaton on the revised LER rule. The LER summanes in this well as selected operating information have been included. This report are arranged alphabet,cally by facihty name and then report supplements earlier annual reports issued by the former chronologically by event date for each facility. Component, Atomic Energy Commission and the Nuclear Regulatory Com-system, keyword, and component vendor indexes follow the mission. The 1984 release data are summanzed in tabular form.

summanes. Vendors are those ident6ed by the utility when the Data covenng specific radionuchdes are summan2ed.

LER form is initiated; the keywords for the component, system, and general keyword indexes are assigned by the computer NUREG/CR 3024: SUSTAINED CONCRETE ATTACK BY LOW-using correlation tables from the Sequence Coding and Search TEMPERATURE. FRAGMENTED CORE DEBRIS.

System. TARBELL,W.W.; BRADLEY,0 R.; BLOSE.R E.; et al. Sandia Na-t I La r s. July 1987. 247pp. 8709090444. SANC82 NUREG/CR-2000 V06 N7: LICENSEE EVENT REPORT (LER)

COMPILATION.For Month Of July 1987.

• Oak Ridge National Laboratory. August 1987.151pp. 8709150268. ORNL/NSIC- Four expenments were performed to study the interactions 200.42679:007. between low temperature core debns and concretes typical of See NUREG/CR 2000,V06,N06 abstract. reactor structures. The tests addressed accident situations where the core debns is at elevated temperature, but not NUREG/CR-2000 V06 NB: LICENSEE EVENT REPORT (LER) molten. Concrete crucibles were formed in nght circular cyhn-COMPILATION.For Month Of August 1987

• Oak Ridge Nation, ders with 45 kg of steel spheres (3 mm' diameter) as the debris al Laboratory. Septcah 1967.125pp. 8710000189. ORNL/ eit.wkett T9 cebns was heated by an inductive power supply NSIC-200. 42987.075. to nominal tempactwes of 1473 A 21673 K. Two tests were See NUREG/CR 2000,V06,N06 abstract. performed on each of two concrete types using either basalt or NUREG/CR-2331 V06 N4: SAFETY RESEARCH PROGRAMS limestone aggregate. For each concrete, one test was per.

SPONSORED BY OFFICE OF NUCLEAR REGULATORY formed with water atop the debns while the second had no RESEARCH.Quarterfy Progress Report, October-December water added. The results show that low temperature core debris 1986. WEISS,A.J. Brookhaven National Laboratory. May 1987- will erode either basalt or limestone common sand concretes.

75pp.8710080214. BNL-NUREG 51454. 42990.350. Downward erosion rates of 3 to 4 cm/hr were recorded for both This progress report will describe current act vities and techni- concrete types. The Ismestone concrete produced a crust layer cal progress in the programs at Brookhaven Nat onal Laboratory within the debris bed that was effective in preventing the down-sponsored by the Division of Accident Evaluation, Division of ward intrusion of water. The basalt concrete crust was formed Engineenng Technology, and Division of Risk Analysis & Oper- above the debns and consisted of numerous, convoluted, thin ations of the U.S. Nuclear Regulatory Commission Office of Nu- layers. Carbon dioxide and water release from the decomposi-clear Regulatory Research. The projects reported are the fol- tion of concrete were partially reduced by the metallic debns to lowing: Advanced-Reactor Review, Thermal-Hydraulic Reactor yield carbon monoxide and hydrogen, respectively. The overfy-Safety Experiments Thermodynamic Core-Concrete Interaction ing water pool did not effect the reduction reactions.

Expenments and Analysis, Plant Analyzer, Assessment and Ap-plication of TRAC.BF1 Code, Code Maintenance (RAMONA- NUREG/CR 3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGION. Final Report, October 38), MELCOR Venfication and Benchmarking, Source Term 1981 September 1986. CHRISTENSEN,D.; POLLACK,H.N.;

Code Package Venfication and Benchmarking, Uncertainty Anal-ysis of the Source Term (OUASAR), Soil-Structure Interaction LAY,T.; et al. Michigan, Univ. of, Ann Arbor, MI. March 1987.

333pp. 8707310041. 41997.018.

Evaluation and Structural Benchmarks, Characterirt on and De*

tection of Age-Related Failures of Selected Components and Earthquake activity in the Western Ohio-Indiana region was Systems With Consideration for Aging / Seismic interactions, monitored with a precision seismograph network consisting of Protective Action Decisionmaking, Risk and Risk Reduction for nine stations located in west. central Ohio and four stations lo-Zion, and Operational Safety Reliabittty Research. cated in Indiana. Five local and near regional earthquakes were recorded during the fiscal 1985-86 report penod, ranging in NUREG/CR 2452: RISK METHODOLOGY FOR GEOLOGIC DIS- magnitude from 0.5 to 5.0m(b). The two largest events (January POSAL OF RADIOACTIVE WASTE. Final Report. 31,1988 near Cleveland, Ohio, 5.0m(b); and July 12,1986 near '

CRANWELL,R.M.; CAMPBELL.J.E.; HELTON.J.C.; et al. Sandia St. Marys, Ohio, 4.5m(b)) were felt with minor damage reported Nat onal Laboratories. August 1987. 110pp. 8710010316. in each case. Focal mechanisms and isoseismal maps for these l SAND 812573. 42883:078. events are included in this report. These events are the largest i

This report contains the description of a nsk assessment to have occurred in Ohio since the events of March 2 and methodology developed for the US Nuclear Regulatory Commis- March 9,1937 (magnitude = 4.5 and 4 9. respectively). The re-sion for use in assessing tiie nsk from the disposal of radioac- maining three earthquakes all occurred in Ohio, north of the tive wastes in deep geologic formations. This methodology con. array. A total of 42 local and near regional events, eleven of sists of (1) techniques for selecting and screening scenanos, (2) which were felt, have now been recorded by the Ohio-Indiana models for use in simulating the physical processes and esti- array since its initiation in 1976. Teleseismic P wave amval and mating the consequences associated with the occurrence of residual tables were updated to include newty acquired data.

these scenarios, (3) probabilistic and statistical techniques for The results are similar to those in previous years. The local ve-use in nsk estimates and sensitivity and uncertainty analysis, locity structure was investigated using data acquired during a re-and (4) a procedure for utilizing these models and techniques to fraction expenment in the summer of 1984, and travel times of assess compliance with regulatory standards. The use of this local and near-regional earthquakes.

methodology is demonstrated in this report by applying it in the analysis of a hypothetical disposal site containing a bedded Salt NUREG/CR 3228 V05: STRUCTURAL INTEGRITY OF WATER formation as the host medium for the waste. REACTOR PRESSURE BOUNDARY COMPONENTS. Annual I Report For 1986. LOSS.F.J. Materials Engineenng Associates, )

NUREG/CR 2907 V05: RADIOACTIVE MATERIALS RELEASED inc. July 1987. 237pp. 8708110406 MEA 2207. 42109 337.

FROM NUCLEAR POWER PLANTS. Annual Report 1984. This program is being conducted for the NRC to previos ana-TICHLER.J.; NORDEN.K.; CONGEMI,J. Brookhaven National ,

lytical and expenmental methods and data that are necessary to Laboratory. August 1987. 256pp. 8708260142. BNL NUREG. l 51581.42332:001, ensure the structural safety and reliability of pressure boundary '

components in U.S. commercial, hght water reactor power sys-Releases of radioactive matenals in airbome and liquid ef- tems. Emphasis is placed on charactenzation cf material prop-fluents from commercial light water reactors dunng 1984 have ertjes performance in a nuclear environment for the application i

Main Citsti:ns and Abstracts 7 l

to plant life extension and mitigation of the consequences of This study empirically develops frequencies of safety signifi-postulated accident scenanos. Current work is organized into cant pipe failures in commercial nuclear power plants (NPPs).

three major tasks: (1) fracture mechanics investigations (2) en- Its pnmary purpose is to update the pipe break frequencies re-vironmentally-assisted crack growth in high temperature, primary ported in the Reactor Safety Study, WASH-1400, which are reactor water, and (3) radiation sensitivity and postirradiation used in many nsk analyses. The study involved reviewing vari-properties recovery. Research progress in these three tasks for Ous data sources for actual piping failure events of significant 1986 is summanzed in this report. magnitude. When extant in the documentation reviewed, infor-mation was extracted conceming conditional factors such as NUREG/CR-4161 V03: CRITICAL PARAMETERS FOR A HIGH- the system in which the failure. occurred, operational mode of LEVEL WASTE REPOSITORY. Volume 3. Salt. DIDWALL,E.M.

the plant, and size vi the pipe involved to estimate conditional Lawrence Livermore National Laboratory. WOLLENBERG H.A.;

pipe break frequencies useful to nsk anijste Because of the BINNALL,E.P.; et al Lawrence Berkeley Laboratory. July 1987.

high quality piping used in NPPs, there have been few signife-41pp. 8708 f 20128. UClD-20092. 42128:057, cant pipe failures. An attempt was made to augment the analy-This is the third report of a three volume senes addressing sis with synthetic data from a Delphi approach, but the wide un.

entical parameters for an underground high-level nuclear waste certainty bounds on the resulting estimates rendered the results repository in basalt, tuff, or salt. This report covers the identifi.

unsuitable for combining data.

cation and pnontization of cntical parameters for a repository in badded salt. For purposes of this work, a parameter is a physi- g g cal property whose value helps determine the charactenstics or OF THE SKIN USING THE COMPUTER CODE VARSKIN.

behavior of a repository system. A parameter is considered to TRAUB,R.Ja REECE,W.D.; SCHERPELZ,R.I.; et al. Battelle Me.

be entical if a mistake in its measurement, or in the inability t measure it, could lead to a wrong conclusion about the adequa.

l PdMMWWu%MM 101pp. 8708280132. PNL-5610. 42399:289 cy of a repository. Consideration was given to the relative im- P 9 PO portance of cntical parameters in four specific discipline areas: dose to the skin from radioactive contamination on the skin.

geomechanics, geology, hydrology, and geochemistry. Relative The calculational method is based on the , tables of absorbed emportance of the parameters was independently considered for energy distnbutions around point sources in water that have each of the four major phases of repository activity: site charac- been developed by M.J. Berger and published in 1971 in Pam-terization, site construction, site operation, and site closure and phlet No. 7 of the " Journal of Nuclear Medicine." The method decommissioning. Important activities that cover all phases are has been implemented in a computer code, VARSKIN, that will determination of hydrologic charactenstics and salt dissolution compute the radiation dose at a specified depth in the skin from rates, and their long- term monitonng. a radiation source ranging in size from a point to a disc having a NUREG/CR-4216: EXPERIMENTAL RESULTS FOR A 1:8-SCALE 100-cm(2) area. This report contains a source code listing of STEEL MODEL NUCLEAR POWER PLANT CONTAINMENT the computer code and tables of dose factors that have been PRESSURIZED TO FAILURE. KOENIG L.N. Sandia National derived from the code, as well as sample runs and user instruc-Laboratories. December 1986. 600pp. 8710010431 SAND 85- tions. This report also discusses the contamination level that 0790.42878 040. corresponds to the recording level described by the internation.

A 1:8 scale steel model of a light water reactor containment al Commission on Radiological Protection.

building was tested as part of an NRC sponsorad program whose objective is to qualify methods for predicting the re- NUREG/CR-4457: AGING OF CLASS 1E BATTERIES IN sponse of containment buildings subjected to severe accidents. SAFETY SYSTEMS OF NUCLEAR POWER PLANTS.

The model was pressun2ed in steps to 195 psig, and its re- EDSON,J.L.; HARDIN.J.E. EG&G Idaho, Inc. (subs. of EG&G, sponse was recorded by strain and displacement transducers Inc.). July 1987. 43pp. 6709040461. EGG 2488. 42522.019. i located throughout the model. Periodic leak-rate measurements This report presents the results of a study of aging effects on were also made. An overview of the test is given, and the re- safety-related battenes in nuclear power plants. The purpose is .

sponse of the model and specific features (such as equipment to evaluate the aging effects caused by operation within a nu- I hatches, airlocks, and piping penetrations) are presented in this clear facility and to evaluate maintenance, testing, and rnonttor-report. ing practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the U.S. Nu-  ;

NUREG/CR-4219 V04 N1: HEAVY SECTION STEEL ThLhWOL- clear Regulatory Commission's (NRC's) Nuclear Plant-Aging Re-OGY PROGRAM. Semiannual Progress Report For October search approach and investigates the materials used in battery 1986 - March 1987. PUGH.C.E. Oak Ridge National Laboratory. construction, identifies stressors and aging mechanisms, pre-August 1987, 300pp. 8710060244. ORNL/TM-9593. 42922:293. sents operating and testing expenence with aging effects, ana-The Heavy.Section Steel Technology (HSST) Program is an lyzes battery failure events reported in vanous data bases, and engineering research activity conducted by the Oak Ridge Na. evaluates recommended maintenance practices. Data bases tional Laboratory for the Nuclear Regulatory Commission. The that were analyzed included the NRC's Licensee Event Report  ;

Program compnses studies related to all areas of the technolo, system, the institute for Nuclear Power Operations' Nuclear gy of matenals fabricated into thick-section primary coolant con- Plant Reliability Data System, the Oak Ridge National Laborato-tainment systems of light water cooled nuclear power reactors. ry's In-Plant Reliability Data System, and The S.M. Stoller Cor.

The investigation focuses on the behavior and structural integri- l poration's Nuclear Power Expenence data base.

ty of steel pressure vessels containing cracklike flaws Current Work is organized into twelve tasks: (1) program management NUREG/CR-4485 V02: THE IMPACT OF FUEL CLADDING FAIL-(2) fracture methodology and analysis, (3) matenal characterize' -

URE EVENTS ON OCCUPATIONAL RADIATION EXPOSURES tion and properties, (4) environmentally assisted crack-growth AT NUCLEAR POWER PLANTS Case Study:PWR Dunng An studies (5) crack arrest technology, (6) irradiation effects stud' Outage. MOELLER M.P.; MARTIN.G F,; KENOYER,J.L. Battelle ies, (7) cladding evaluations, (8) intermediate vessel tests and Memonal institute, Pacific Northwest Laboratories. August 1987.

analysis, (9) thermal shock technology, (10) pressurized ther, 150 B710080227. PNL-5606. 42993.327, mal-shock technology, (11) Pressure Vessel Research Users This report is the second in a senes of case studies designed Facility, and (12) shipping-cask matenal evaluations. to evaluate the magnitude of increase in occupational radiation NUREG/CR-4407: PIPE BREAK FREQUENCY ESTIMATION FOR exposures at commercial U S. nuclear power plants resulting NUCLEAR POWER PLANTS. WRIGHT,R E.; STEVERSON J.A.; from small incidents or abnormal events. TN event eve,twated is ZUROFF,W F. EG&G Idaho. Inc. (subs of EG&G, Inc.). May fuel cladding failure, which can result h elevated pnmary cool-1987. 255pp. 8708040304. EGG 2421. 42037.260 ant activity and increased radiation exposure rates within a l

8 Main Citati:ns and Abstracts

)

l Plant. For this case study, radiation measurements were made NUREG/CR 4590 V01: AGING OF NUCLEAR STATION DIESEL at a pressunzed-water reactor (PWR) dunng a maintenance and GENERATORS Evaluation Of Operating And Expert refueling outage. The PWR had been operating for 22 months Expenence. Phase i Study. HOOPINGARNER,K.; VAUSE,J.W.;

with fuel cladding failure characten2ed as 105 pin-hole leakers, OlNGEE,D. A.; et al. Battelle Memorial Institute, Pacific North-the equivalent of 0.2t2 percent failed fuel. Gamma spectrosco- west Laboratones. August 1987. 300pp. 8709150302, PNL-py measurements, radiation exposure rate determination, ther- 5832. 42679:158.

moluminescent dosimeter (TLD) assessments, and air sample Pacific Northwest Laboratory evaluated operational and analyses were made in the plant's radwaste, pipe penetration, expert expenence pertaining to the aging degradation of diesel and containment building 1 Based on the data collected, evalua- generators in nuclear service. The research, sponsored by the tions indicate that the rela'ive contnbutions of actfvation prod- U S. Nuclear Regulatory Commission, identified and character-ucts and hssion products to the total exposure rates were con- ized the contnbution of aging to emergency diesel generator stant over the duration of the outage. This constancy is due to failures. Volume 1 reviews diesel generator expenence to identi-the significant contnbution from the longer lived isotopes of fy the systems and components mcst subject to aging degrada-c6aium (a hr.sion product) and cobalt (an activation product). tion and isolates the major causes of failure that may affect For this reason, fuel cladding failure events remain as signifi- future operational readiness. Evaluations show that as plants 1

cant to occupational radiation exposure during an outage as l age, the percent of aging related failures increases and failure dunng routine operations. As documented in the previous case modes change. A compilation is presented of recommended study (NUREG/CR4485 Vol.1), fuel cladding failure events in-corrective actions for the failures identified. A review of current creased radiation exposure rates an estimated 540 percent in relevant industry programs, research, and standards is included.

some areas of the plant dunne routine operations. Consequent- Volume 2 reports results of an industry wide workshop.

ly, such events can result in significantly greater radiation expo-sure rates in many areas of the plant dunng the maintenance NUREG/C3-4590 V02: AGING OF NUCLEAR STATION DIESEL and refueling outages than would have been present under GENERATORS. Evaluation Of Operating And Expert i

I normal fuel conditions. Expenence Workshop. HOOPINGARNER K.; VAUSE,JW. Bat- l telle Memonal Institute, Pacific Northwest Laboratories. August )

NUREG/CR 4534: ANALYSIS OF DIFFUSION n AME TESTS. 1987. 69pp. 8709150331. PNL 5832. 42678:301. j SHEPHARD J E. Sandia National Laboratones. August 1987. Pacific Northwest Laboratory evaluated operational and 89pp. 8710010109. SAND 86-0419. 42881:333.

expert experience pertaining to the aging degradation of diesel This report discusses the results and analysis of hydrogen dif. generators in nuclear service. The research, sponsored by the fusson flame lests conducted at the Nevada Test Site by EPRI U.S. Nuclear Regulatory Commission, identified and character-and the USNRC. Those tests were designed to simulate the ef- ized the contnbution of aging to emergency diesel generator fects of hydrogen combustion inside a nuclear power plant con- failures. This report, Volume 2, reports the results of an indus-tainment following a degraded-core accident. Test initial condi- try wide workshop held on May 28 and 29,1986 to discuss the tions and sample data plots are given for 16 tests. Mixing and technical issues associated with aging of nuclear service emer-ignition phenomena are discussed in terms of the source pa- gency diesel generators. The technical issues discussed most rameters and igniter location. A simple model is developed for extensively were; man / machine interfaces, component inter-simulahng the heat transfer and computing convective heat faces, thermal gradients of startup and cooldown and the need transfer coefficients for expenmental pressure measurements. for an accurate industry database for trend analysis of the Convective heat transfer coefficients are reported for four tests. diesel generator system. Volume 1 reports the results of the The effect of stagnation point heat transfer is estimated. Phase i research.

NUREG/CR 4577: AUTOMATED LONG TERM SURVEILLANCE NUREG/CR-4640: HANDBOOK OF SOFTWARE OUALITY AS-OF A COMMERCLAL NUCLEAR POWER PLANT. SMITH C.M.;

SURANCE TECHNIQUES APPLICABLE TO THE NUCLEAR IN.

GONZALEZ,R.C. Oak Ridge National Laboratory. August 1987. DUSTRY. BRYANT,J.L.; WILBURN.N P. Battelle Memorial Insti-66pp.B710080198. ORNL/TM 10015,42995:290.

tute, Pacific Northwest Laboratones. August 1987. 113pp.

This report presents and desenbes a pattern recognition 8709090601. PNL 5784. 42594 001.

system for monstonng nuclear reactor signals. The system is Pacific Northwest Laboratory is conducting a researen project based on detecting deviations from baseline signatures identi- to recommend good engineenng practices in the apphcation of fled during normal plant operation. The capabihties and hmita- 10 CFR 50, Appendix B requirements to assure quality in the tions of this pattern recognition approach were investigated development and use of computer software for the design and dunng a 21/2 year senes of continuous online expenments at operation of nuclear power plants for NRC and industry. This the Sequoyah 1 nuclear power plant.

handbook defines the content of a software quality assurance program by enumerating the techniques applicable. Definitions, NUREG/CR 4583 V03: DEVELOPMENT AND VALIDATION OF A desenptions, and references where further information may be REAL-TIME SAFT UT SYSTEM FOR THE INSPECTION OF obtained are provided for each topic.

LIGHT WATER REACTOR COMPONENTS. Annual Report, October 1985 September 1986. DOCTOR.S.R.;

HALL,T.E.; REID.L.D.; et al. Battelle Memorial Institute, Pacific NUREG/CR-4654: RADIONUCLtDE TRANSPORT AS VAPOR THROUGH UNSATURATED FRACTURED ROCK. GREEN,R.T.:

Northwest Laboratones. July 1987. 73pp. 8709210489. PNL- EVANS.D D. Anzona, Univ. of, Tucson, AZ. July 1987.199pp.

5822.42738.004.

8708040263.42036:028.

The Pacific Northwest Laboratory is working to design, fabri- The objective of this study is to identify and examine potential cate, and evaluate a real-time flaw detection and charactenza- mechanisms of radionuclides transport as vapor at a high-level ton system based on the synthetic aperture focusing technique radioactive waste repository located in unsaturated fractured for ultrasonic testing (SAFT UT). The system is desegned to per- rock. Transport mechanisms and processes have been investi-form inservice inspection of hght water reactor components. In- gated near the repository and at larger distances. Transport cluded objectives of this program for the Nuclear Regulatory mechanisms potentially important ai larger distances include or.

Commission are to develop procedures for S; eem cakbration dinary diffusion, viscous flow and free convection. Ordinary dif-and field operation, to vahdate the system through laboratory fusion includes self and binary diffusion, Knudsen flow and sur-and field inspections, and to generate an engine 6ang data base face diffusion. Pressure flow and sirp flow compnse viscous to tapport ASME Code acceptance of the technology. This flow. Free convective flow results from a gas density contrast.

progress report covers the programmatic work from October Transport mechanisms or processes dominant near the reposi-1985 through September 1986.

tory include ordinary diffusion, viscous flow plus several mecha-

l I

l MrIn Citati ns and Abstracts 9 nisms whose driving forces arise from the non-isothermal, radio- consists of a survey of light water reactor component failures active nature of high-level waste. The additional mechanisms in- associated with selected safety and support systems. Tables l clude forced diffusion, aerosol transport, thermal diffusion and are presented, indicating the systems and the components j thermophoresis. Near a repository vapor transport mechanisms within those systems most affected by aging. Also provided are l and processes can provide a significant means of transport engineering insights drawn from the data. The second c@ysis i from a failed canister to the geologic medium from which other consists of identifying and categonzing the reported fa ure processes can trancport radionuclides to the accessible envi- causes of component failures. The systems used in the failure-  ;

ronment. These issues are believed to be important factors that cause analysis were service water systems and Class 1E elec-must be addressed in the assessment of specific engineering incal power distribution systems for Babcock & Wilcox Compa-designs and site selection of any proposed HLW repository. ny pressurized water reactors and service water systems for NUREG/CR 4710: SHUTDOWN DECAY HEAT REMOVAL ANAL- boiling water reactors.

YSIS OF A COMBUSTION ENGINEERING PRESSURIZED NUREG/CR-4760: TEST OF 6-IN. THICK PRESSURE WATER REACTOR. Case Study. CRAMOND,W.R.; VESSELS.Senes 3. Intermediate Test Vessel V 8A.Teanng Be-ERICSON,D.M.; SANDERS,G.A. Sandia National Laboratones. havior Of Low Upper Shelf Matenal. BRYAN,R.H.; BASS,B.R.;

July 1987. 775pp. 8708110412. SAND 86-1797. 42107:282. BOLT,S E.; et al. Oak Ridge National Laboratory. May 1987.

This is one of six case studies for USl A 45 Decay Heat Re- 412pp. 8710060100. ORNL 6187. 42920:003.

moval (DHR) Requirements. The purpose of this study is to Tests of several 152 mm-thick vessels have been performed identify any potential vulnerabihties in the DHR systems of a to study the behavior of flaws under stress states similar to typical Combustion Engineering PWR, to suggest possible modi- those in full-scale reactor pressure vessels. The objective of the fications to improve the DHR capability, and to assess the value latest test. V BA, was to provido accurate quantitative data con-and impact of the most promising alternatives to the existing ceming the growth by ductile teanng and final instability of a DHR systems. The systems analysis considered small LOCAs flaw in a low upper-shelf toughness weld located in a cylinder and transient intemal initiating events, and seismic, fire, extreme of reactor vessel steel. This test is important because there are wind, internal and external flood, and lightning external events, vessels n service that contain welds which, because of high A full scale systems analysis was performed with detailed fault copper content, may have their Charpy upper shelf energy trees and event trees including support system dependencies.

values reduced to relatively low levels by neutron irradiation.

The system analysis resu!ts were extrapolated into rebase cate-The results of the V-8A test are intended to provide an experi.

gones using applicable past PRA phenomenological results and improved containment failure mode probabilities. Pubhc conse-mental basis for judging the accuracy of vessel fracture safety quences were estimated using site specific CRAC2 calculations.

analysis procedures for low-upper shelf toughness conditions.

The objective of the V-8A test was attained, with a tearing insta-The Value-impact (VI) analysis of possible attematives consid-ered both onsite and offsite impacts amving at several risk bility observed at a Pressure of 139 MPa (2 x desi9n Pressure).

measures such as averted population dose out to a 50-mile The flaw, which was initially a fatigue-sharpened notch with an radius and dollars per person rem averted. Uncertainties in the approximately eihptical profile, grew in depth and length to  ;

VI analysis are discussed and the issues of feed and bleed and 101.4 mm and 453 mm, respectively. Pretest and postte,t fra- '

secondary blowdown are analyzed.

ture-mechanics and stress analyses were made by simplif 4 methods, convenient for investigating a wide range of param-NUREG/CR 4731 V01: RESIDUAL LIFE ASSESSMENT OF eters, and by tt?ree dimensional finite element methods, which MAJOR LIGHT WATER REACTOR COMPONENTS - modeled the matenal properties and geometry more precisely.

OVERVIEW. Volume 1. SHAH,V.N.; MACDONALD,P.E. EG&G Ductile flaw growth and instabihty predictions based upon meas-Idaho, Inc. (subs. of EG&G, Inc.). June 1987. 198pp. ured J-resistance and tensile properties were made. Results of 8708060423. EGG-2469. 42071:001. analyses based on J(R)- controlled crack growth agree reason.

This report presents an assessment of the aging (time-de- ably well with expenmental observations, pendent degradation) of selected major hght water reactor com-ponents and structures The stressors, possible degradation NUREG/CR-4767: SHUTDOWN DECAY HEAT REMOVAL ANAL-sites and mechanisms, potential failure modes and currently YSIS OF A GENERAL ELECTRIC BWR4/ MARK 1. Case Study.

used non destructive examinations, in-service inspection (ISI) HATCH,S W.; ERICSON,D.M.; SANDERS,G.A. Sandia National and hfe assessment methods are discussed for seven mayor Laboratories. July 1987. 650pp. 8710140105. SAND 86-2419.

hght water reactor components: pressunzed water resear 43051:001.

(PWR) and boihng water reactor (BWR) pressure vessels, PWR A General Electric Boiling Water Reactor (BWR4) with a Mark containment structures, PWR reactor coolant piping, PWR I containment has been evaluated as part of Task Action Plan steam generators, BWR recirculation piping, and reactor pres- A 45, " Decay Heat Removal Requirements." Probabikstic nsk sure vessel supports. Unresolved technical issues related to hfe assessment models were constructed to determine the domi-extension of these components, including requirements for ad- nant internal, randomly initiated accident sequences and special vanced ISI and hfe assessment methods, are also discussed. emergency sequences (e g., earthquakes). The dominant se-quen s wee mw me at caws mgm NUREG/CR-4744 V01 N2: LONG TERM EMBRITTLEMENT OF be made to enhance the plant s abikty to remove decay heat.

CAST DUPLEX STAINLESS STEELS IN LWR Modifications which held promise went through a prehminary SYSTEMS. Semiannual Report, April September 1986 #8 8 Y """ "'

CHOPRA,0.K.; CHUNG,H M. Argonne National Laboratory. May m ae n was es' adgm @ wa-1987. 54pp. 8709040241. ANL 8716. 42526:101 cah M N W sh W NA e p This progress report summanzes work performed by Argonne bened in a value-impact format according to NRC guidehnes National Laboratory du ing the six months from April September msWs ate M kas& McaWs to e 1986 on long-term embnttlement of cast duplex stainless steels cay Mat awal & W at M %d M McWa used in k9ht water reactors' estimates of the value-impact results tended. however, to show NUREG/CR-4747 V01: AN AGING FAILURE SURVEY OF LIGHT marginal cost effectiveness under current guidelines for most of WATER REACTOR SAFETY SYSTEMS AND COMPONENTS the modifications. Alternate assumptions involving source term M E ALE,0.M.: SATTERWHITE,0. EG&G Idaho, Inc. (subs. of magnitude were found to significantly affect the results The in.

EG&G, Inc ). July 1987. 295pp. 8709090266. EGG 2473. sights gained from this study will become part of an information 42578 129 base which will be used to develop genenc recommendations This report desenbes the methods, analyses, results, and regarding the adequacy of decay heat removal systems in light conclusions of two different aging studies The first analysis water reactors.

10 Main Cit:ti ns and Abstracts NUREG/CR-4768 V01: METHODOLOGY AND APPLICATION OF adiacent regions; (3) a core debns management system that SURROGATE PLANT PRA ANALYSIS TO THE RANCHO would prevent the core debns from a failed reactor vessel from SECO POWER PLANT. Task 1. Analysis Of ANO 1 And interacting with significant quantities of water and with the con.

Oconee PRAs. GORE.B.F. Battelle Memonal Institute, Pacific crete of the cavity; (4) an alternate containment spray system Northwest Laboratones. July 1987, 75pp. 8707310128. PNL- that would operate without relying on any existing plant electri.

60321. a; t 996:304. cal power supplies, and (5) a passive containment heat removal This two-volume report presents the development and first system that would remove heat from the containment atmos- I application of a methodology for using genenc PRA information phere by condensing steam. These systems were considered i to identify nsk important systems and components at a plant for installation in three types of LWR containments. Conceptual {'

lacking a PRA. The methodology requires the detailed analysis designs and associated cost estimates were developed for two of both sailant>es and differences between related plants. It is proposed combinations of these systems. This work provides an applied in e analysis of the Rancho Seco plant using informa. input to the cost benefit analyses conducted in the SARRP Pro-tion from PRAs for the ANO-1 and Oconee plants. It is genenc, gram in support of NRC's ase ,sment of severe accident nsks drawing upon the functional and design similanties of B&W to be published in NUREG 1150, plants, yet it incorporates considerable plant specificity through the detailed comparative analysis of systems. Volume 1 pre. NUREG/CR-4805 V01: REACTOR SAFETY RESEARCH SEMI.

sents the analysis of the surrogate plant PRAs. Dominant cut ANNUAL REPORT. January June 1986.

• Sandia National Lab-sets leading to core melt are identified and analyzed to deter. oratones. May 1987. 359pp. 8707300091. SAND 86-2752.

mine the Fussel-Vesely importance of plant systems. In Volume 41983:130.

2 the dominant cut sets are further analyzed to identify and cat- Sandia National Laboratones is conducting, under USNRC egonze important system failure modes. The Rancho Seco plant sponsorship, phenomenological research related to the safety is then studied to determine the plausibility and importance of of commercial nuclear power reactors. The research includes similar failure modes for High Pressure injection, Low Pressure expenments to simulate the phenomenology of the accident injection, Emergency Feedwater Service Water, Vital AC power conditions and the development of analytical models, venfied by and DC Power systems. Plant specific information is then pre. expenment, which can be used to predict reactor and safety sented identifying Rancho Seco components, power supplies, systems performance and behavior under abnormal conditions.

and operating enodes associated with the failure modes for the The objective of this work * ,i provide NRC requisite data first four of these systems. bases and analytical methods to (1) identify and define safety ssues. (2) understand the progression of nsk significant aces-NUREG/CR 4768 V02: METHODOLOGY AND APPLICATION OF #

SURROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER PLANT. Final Report GORE B.F -

at Sada hnal Wah 4 HUENEFELD.J.C. Battelle Memonal Institufe, Pacific Northwest Laboratones. July 1987. 87pp. 8707310413. PNL-6032 2. s s 41994 017.

NUREG/CR-4810: EVALUATION OF DIESEL UNAVAILABILITY See NUREG/CR 4768 V01 abstract. AND RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

NUREG/CR-4769: RISK EVALUATIONS OF AGING VESELY,W.E.; DEMOSS,G.; LOFGREN,E.; et al. Brookhaven PHENOMENA.The Linear Aging Reliability Model And its Exten. National Laboratory. Msv 1987. 83pp. 8708260121. BNL-sions. VESELY,W E. EG&G Idaho, Inc. (subs. of EG&G, Inc.). NUREG 52022. 42331:180.

Apnl 1987.176pp. 8708130348. EGG 2476 42154:110. In this report nsk and reliability approaches are presented A model for light water reactor safety system component fail- which allow nsk acceptable test intervals to be determined for ute rates due to aging mechanisms has bean developed from any diesel. Data required to apply the approaches are also de-basic phenomenologicat considerations. In the treatment, the scnbed. The approaches can be applied not only to diesels, but occurrences of detenoration are modeled as following a Pois, to any component with suitable data. Incorporation of the ap-son process. The seventy of damage is allowed to have any proaches in personal computer (PC) software is discussed distnbution, however, the damage is assumed to accumulate in- which can provide tools for the regulator or pisnt personnel for dependently. Finally, the failure rate is modeled as being pro- any plant specific or genene application. The FRANTIC lli com-portional to the accumulated damage. Using this treatment, the puter code was run to validate the approacNs and to evaluate Anear aging failure rate model is obtained. The applicability of specific issuer associated with determining nsk-etlective test in-the linear aging model to various mechanisms is discussed. The tervals for diesels. Using the approaches presented, diesel acci-model is also extended to cover nonlinear and dependent aging dent unavailability can be more effectively monitored and con.

phenomena. The implementation of the linear aging model is trolled on a plant specific or genene basis. Test intervals can be demonstrated by applying it to the aging data collected in the made more nak effective than they are now, producing more ac-U S. Nuclear Regulatory Commission's Nuclear Plant Aging Re- ceptable accident unavailabilities, and avoiding the possible del-search Program, etenous effects of Regulatory Guide 1.108. The methods pre-sen'ed are one step toward performance-based technical specs-NUREG/CR 4781 DRFT: STUDY OF SEVERE ACCIDENT MITI- ficauons, which more directly control nsks.

GATION SYSTEMS. CHERDACkR.; HESS.C.; LEE,K.; et al.

Sandia National Laboratones. May 1987. 702pp. 8707300021. NUREG/CR 4817: LODINE PARTITION COEFFICIENT MEAS.

SAND 87-7064. 41977:020. UREMENTS AT SIMULATED PWR STEAM GENERATOR This study has been performed as part of the Severe Aces CONDITIONS Intenrn Data Report. CLINTON.S.D.;

dent Risk Reduction Program (SARAP) to investigate the feass SIMMONS,C M, Oak Ridge National Laboratory. May 1987.

bility and costs of installing vanous systems deogned to mits- 28pp.8708060384. ORNL/TM 10330. 42075 002.

gate the effects of severe accidents in light water reactors lodine parttion coefficients (defined as the ratio of the con-(LWRs). The pnmary function of the systems is to maintain the centration of sodine species in the aqueous solution to the integnty of the reactor containment in the event of a severe ac- iodine concentration in the vapor phase) were measured 6t sim-cident one in which substantial core melting and reactor ulated PWR steam generator conditions (285 degrees C and 6 9 vessel failure occur. The systems evaluated include (1) a hydro. MPa), using camer free radioactive 1-131 in the form of sodium j gen ignition system with capabilities for detecting hydrogen con- iodide. The iodine tracer concentration was maintained at 6 x Centrations, initiating hydrogen burning, and avoiding the tngger- 10( 11) mol/L; bonc acid concentration was vaned from 0 to 0 4 rng of detonations,- (2) a reactor Cavity flooding system Capable mol/L; and the solution pH (measured at 25 degrees C) wn ad-Df introducing large quantities of water in the reactor cavity and justed frorn 4 to 9 by the addition of lithium hydroxide. lodine l

l

. _ _ . __ __-_-_-__________-____-_____-_________a

l l

l Main Citatirns and Abstracts 11 partition coefficients decrease with increasing bonc acid con- NUREG/CR 4846: HIGH-LEVEL WASTE PRECLOSURE SYS-centration: however, the iodine volatility is essentially independ- TEMS SAFETY ANALYSIS. Phase 2, Final Report. LIGON D.M.;

l l

ent of the solution pH for a given boric acid concentration. STAMATELATOS,M.; BARSELL,A.W.; et al. Sandia National Sparging the solutions with air at room temperature increases Laboratories. June 1987.163pp. 8708250176. SAND 87 7029.

the iodine volatility by an order of magnitude, compared to that 42330:277.

achieved with argon sparging. lodine partition coefficient meas. The major effort of this phase was the demonstration of the urements ranged from a low of 200 (in 0.2 molanty boric acid methodology developed in Phase 1. A sample problem consist-sparged with air) to 400,000 (in punfied water sparged with Ing of six scenarios was quantified. The purpose of this sample argont problem is to check the application of the assembled methods, NUREG/CR 4834 V01: RECOVERY ACTIONS IN PRA FOR THE particularly the importance ranking scheme. All of the features RISK METHODS INTEGRATION AND EVALUATION PRO. of the complete analytical technique are applied including un-GRAM (RMIEP). Volume 1: Development Of The Data Based certainty and sensitivity ar.alysis, common cause failure evalua- l Method. WESTON,L.M.; WHITEHEAD,0.W. Sandia National tion and human error contribution. Other topics addressed in-Laboratones. GRAVES,N.L. Energy, Inc. June 1987. 295pp. clude Mine Related Data Development Human Reliability Analy.

8709090367. SAND 87-0179. 42572:071, sis and Common Cause Failure Analysis.

In a probabihs;c nsk assessment (PRA) for a nuclear power plant, the analyst identifies a set of potential core damage NUREG/CR-4850: STEAM GENERATOR GROUP PROJECT. Task events consisting of equipment failures and human errors and 10 Secondary Side Examination.Finaf Report. SCHWENK,E.B.

their estimated probabilities of occurrence, if operator recovery Battelle Memonal Institute, Pacific Northwest Laboratories. June i from an event within some specified time is considered. the 1987.107pp. 8708240213. PNL-6045. 42316:155. l probability of this recovery can be included in the PRA. This The Steam Generator Group Project (SGGP) is using the re- j report provides PRA analysis wrth an improved methodology for tired from service Surry 2A pressurized water reactor steam  !

including recovery actions in a PRA. A recovery action can be generator as a test bed to investigate the rehability and effec.

divided into two distinct phases, a Diagnosis Phase (realizing tiveness of in- service nondestructive eddy current inspection .

that there is a problem with a entical parameter and deciding equipment and procedures. In addition, service degraded tubes  !

upon the correct course of action), and an Action Phase (phys- from the generator will be used to validate models of remaining I ically accomplishing the required action). In this methodology, tube integnty developed in Phase I of the program. This infor- l simulator data are used to estimate recovery probabilities for mation will provide the technical basis for revision of Regulatory I the diagnosis phase. Different time-reliability curves showing the Guides 1.83 and 1.121 goveming in-service inspection and tube probability of failure of diagnosis as a function of time from the compelhng cue for the event are presented. These curves are plugging cnteria, respectively. This report is the second of two based on simulator exercises, and the actions are grouped reports covenng examination and charactenzation of the sec-based upon their operational similanties. This is an improvement ondary side of the Surry unit. It includes a photographic survey over existing diagnosis models that rely greatly upon subjective of typical Sur'Y 9enerator matenal and structural conditions' judgment to obtain such estimates. The action phase is mod- metallurgical failure analysis of a ruptured U bend tube, a de-eled using estimates from available sources. The methodology scription on how secondary side access penetrations were also includes a recommendation on where and when to apply made and an experimental stress analysis of a bent Row 1 the recovery action in the PRA process. tube.

NUREG/CR-4844 DRFT: INTEGRATED RELIABILITY AND RISK NUREG/CR 4851: SEISMICITY 1886-89 IN THE SOUTHEAST-ANALYSIS SYSTEM (IRRAS) USER'S GUIDE VERSION 1.0 ERN UNITED STATES.The Aftershock Sequence Of The (DRAFT). RUSSELL,K.D.; SNIDER.D.M.; SATTISON.M.B.; et al. Charleston, South Carohna Earthquake. ARMBRUSTER.J.G.;

EG&G Idaho, Inc. (subs. of EG&G, Inc.). June 1987.149pp. SEEBER,L. Lamont-Doherty Geological Observatory. May 1987.

8707300200. EGG 2495. 41986:016. 202pp. 8709090380. 42569:334.

The Integrated Reliability and Risk Analysis System (IRRAS) A search of contemporary newspapers in the Carolinas, Geor-is an integrated PRA software tool which gives the user the abil- gia and Eastern Tennessee dunng the 18861889 (inclusive) ity to create and analyze fault trees using an IBM PC. This pro- aftershock sequence of the Charleston earthquake of 1886 has gram provides functions that range from graphical fault tree provided more than 3000 intensity reports for 522 earthquakes construction to cut set generation and quantification. Also pro-vided in the system is an integrated full-screen editor for use as compared to 144 previously known earthquakes for the same period. Of these 144 events,138 were felt in Charleston /

when interfacing with remote computer systems. IRRAS-PC is being developed at the INEL as the USNRC's state.of the-art Summerville and had been assigned epicenters in that area.

microcomputer based probabilistic risk assessment (PRA) The new data provide 112 well-constrained microseismic ept-model development and analysis tool to address key nuclear centers. The 1886-1889 seismicity is characterized by a linear plant safety issues. The INEL role in the IRRAS program is that relation between log frequency and magnitude with a slope b1, of software developer and interfacer to the user community, in- a temporal decay of earthquake frequency proportional to time ciuding training and technology transfer. To support this role, ( 1), and a low level of seismicity pnor to the main shock. These this user's manual was developed as a guideline for use of the are frequently observed charactenstics of aftershock se-IRRAS software. Presented in the manual is an explanation of quences. By 1889, the level of seismicity had decreased more the installation procedures, the hardware and software require- than 2 orders of magnitude, reaching approximately the current ments, the forms and menus used by the program, and detailed level in the same area. The 1886-1889 epicenters delineate a explanations of the programs's capabilities and functions. The large aftershock zone that extends northwest about 250 km level of detail presented in this manual is intended to guide the from the coast into the Piedmont and at least 100 km along the beginning or infrequent user. IRRAS-PC has all the capabilities Fall Line. The aftershock zone occupies the same area as a and functions required to create, modify, reduce, and analyze zone of recent seismicity with unique charactenstics The por-fault tree models used in the analysis of complex systems and tion of this zone in the Piedmont is the only area of So';theast-processes IRRAS-PC uses advanced graphic and analytical ern United States where reservoir-induced seismicrty is unam-techniques to achieve the greatest possible realization of the biguously recognized. Seismicity elsewhere in the Southeast, potential of the microcomputer and when the needs of the user such as in the Virginia seismic zone, is deeper and apparently exceed this potential IRRAS-PC can call upon the power of the unrelated to reservoir loading mainframe

l l

l 12 Main Citati:ns and Abstrtets j i

l NUREG/CR 4860: FLAW DENSITY EXAMINATIONS OF A CLAD position depending on their method of preparation and history. j BOILING WATER REACTOR PRESSURE VESSEL SEGMENT. Solubility products between 10( 32) and 10( 33) can account for l COOK,K.V.; MCCLUNG.R.W. Oak Ridge National Laboratory. most of the data. These data can be used to estimate solubili-Apnl 1987. 33pp. 8707310061. ORNL/TM 10364. 41996;179. ties for cases where solubility limits transport of technetium in As part of the Oak Ridge National Laboratory's Heavy.Sece reducing high level waste repository environments. ,

tion Steel Technology Program, studies have been conducted to  !

determine flaw density in a section of reactor pressure vessel NUREG/CR-4870: AN EVALUATION OF THE EFFECTS OF cut from the Hope Creek Unit 2 vessel. This boiling water reac. DESIGN DETAILS ON THE CAPACITY OF LWR STEEL CON-tor vessel was never in service. One objective was to evaluate TAINMENT BUILDINGS. GREIMANN.L.; FANOUS F.;

the approximate 0.7 by 3 m (2 by 10-ft) segment of the vessel ROGERS.J.; et al. Sandia National Laboratories. May 1987.

provided using ultrasonic flaw detection methods performed 96pp. 8709090402. SAND 87 7066. 42569.238.

with both ASME Code techniques and supplemental ultrasonic As part of their work with severe accident loadings, the Con. J methods. A second objective was to evaluate the inner surface tainment Integnty Division at Sandia National Laboratones has i stainless steel cladding for cracks with a high sensitivity pene- been conducting a program to evaluate the performance of con-trant examination. Both objective

• are successfully completed. tainment buildings with intemal pressure. Sandia has suggested Frve Code-recordable indications were detected ultrasonically; using a strain entenon for pectr1 rupture of steel contairk however, all were found to be anomalies associat+:1 with the ments; the strain limit used 9 correlate with failure must be cladding. One flaw was detected by supplemental ultrasonic consistent with the level of detal. included in the analytical tests, and it was analyzed destructively. This flaw was a pipelike model. If detailed 3D mocels are dem6 ped, the strain limit may 1 indication, about 20 mm (0.8 in.) long extending along the be as large as the matenal's ultimv3 strain. This enterion is ap-length of the longitudinal weld in which it was located and was plied to strains calculated with a shell theory whict. includes ma- j about 20 mm below the cladding surface. The flaw had a tenal and geometric nonlineanties and is applied to the actual I through wall dimension (or length) of about 6 mm (0.24 in.) for containment. In this work, drawings from nine steel contain- l an approximate 3 mm (0.1 in.) distance along the 20-mm major ments were studied and several significant strain concentration length. No flaws were detected by trae penetrar t examination of regions were identified and classified as: eccentncities in stiffen-the cladding surface. er pattems around penetrations, eccentncities in the shell NUREG/CR-4863: SURFACE SPECTROSCOPY OF PRESSURE middle surface, f;at plate covers used on spare penetrations, VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS containment base connection details, and ellipsoidal and tons-FORMED IN PWR ENVIRONMENTS. HANN!NEN,H.E.; phencal heads and covers. The 1.8 scale model was analyzed VULLI.M.; CULLEN,W.H. Matenals Engineenng Associates, Inc. Using the shell finite element in ANSYS, and the results agreed July 1987.100pp. 8708040253. MEA 2194. 42036:227. well with Sandia's analytical results and the expenmental re-The composition and structure of corrosion products formed suits. Examples of each classification were analyzed by finite on corrosion latigue fracture surfaces of pressure vessel steels element and/or simplified equations. In the case of middle sur-tested in PWR. water conditions have been analyzed by using X- face eccentricities, the strains are self limiting. Although flat Ray Photoelectron Spectroscopy (XPS) and Auger Electron plates have pnmary bending strains, they are typically designed Spectroscopy (AES) techniques. This was the first time these so as not to control. Bolts in the base connection have pnmary electron spectroscopic techniques have been applied to corro- strains and may control.

sion fatigue fracture surf ace studies. The oxide phase on the corrosion fatigue fracture surfaces was Fe(3)O(4) (magnetite) in NUREG/CR-4871: RESULTS FROM THE DCH 1 EXPERIMENT.

TARBELL.W.W.; ROSS J.W.; ARELLANO,F.E.; et al. Sandia Na-all the specimens. Small amounts of sulfur (typically about 3 tional Laboratones. June 1987. 67pp. 8707300035. SAND 86-atomic percent) were present in the oxide film mainly as FeS(2).

In specimens showing the highest crack growth rates, the 2483. 41983:063.

amount of sulfur was about doubled near the crack tip com- The DCH-1 (Direct Containment Heating) test was the first ex-pared to the value obtained further behind the crack tip in the penment performed in the Surtsey Direct Heating Test Facility.

middle of the fracture surface. It was possible to locate the The test involved 20 kg of molten core debns simulant ejected crack tip condition on the high-temperature Pourbaix diagram in into a 1:10 scale model of the Zion reactor cavity. The melt wa,s the area where both magnetite and FeS(2) are stable phases. produced by a metallothermic reaction of iron oxide and alumi-The anticipated crack tip conditions are that the corrosion po- num powers to yield molten iron and alumina. The cavity model tentialis about 500 mV(SHE) or less and the pH value is neu. was placed so that the emerging debns propagated directly up-tral or slightly acid. XPS and AES analyses of the corrosion fa. wards along the vertical centerline of the chamber. Results from tigue fracture surfaces were found to reveal the possible water the expenment showed that the molten matenal was ejected chemistry impuntses dunng the corrosion fatigue tests, like Cl, from the cavity as a cloud of particles and aerosol. The dis-Na Ca. etc. persed debns caused a rapid pressunzation of the 103-m(3) j charrber atmosphere. Peak pressure from the six transducers

)

NUREG/CR 4865: THE SOLUBILITY OF ELECTROCEPOSITED ranged from 0 09 to 0.13 MPa (13.4 to 19.4 psig) above the ini- l TC(IV) OXIDES. MEYER,R E.; ARNOLD.W D.; CASE.F l. Oak tial value in the c? amber. Posttest debris collection yielded 11.6 Ridge NaDonal Laboratory. July 1987. 31pp. 8710010154. kg of material outside the cavity, of which approximately 1.6 kg i ORNL 6374. 42882:060- was attnbuted to the uptake of citygen by the iron particles. Me- 1 Solubilities of electrodeposited Tc(IV) oxides have been de- chanical sieving of the recovered debns showed a lognormal termined in solutions of Nacl, HCI, and synthetic groundwater in size distnbution with a mass mean size of 0.55 mm. Aerosol the pH range 0 to 10. Oxides were electrodeposited onto plath measurements indicated a substantial portion (2 to 16%) of the num electrodes, and the oxide covered platinum was then ejected mass was in the size range less than 10 micrometer placed into a small stirred cell. Solubilities were determined by aerodynamic equivalent diameter, counting the beta radiation of (99)Tc in the solution in the stirred cell. The solubilities are approximately constant in the pH NUREG/CR 4876: SILVER INDIUM.CADMlUM CONTROL ROD range 3 to 10; the values in this region for deposits from acid BEHAVIOR AND AEROSOL FORMATION IN SEVERE REAC-solution average (1.32 plus or minus 0 68) x 10( 8) mol/L. For TOR ACCIDENTS. PETTI.D.A. EG&G Idaho, Inc. (subs. of ondes deposded in basic solutions the average is (2.56 plus or EG&G. Inc.). Apnl 1987. 185pp. 8708130364. EGG 2501.

minus 0 54) x 10(.9) modL The electrodeposited oxides are hy- 42154:286.

drated and e <penments show that they have the composition Ag-In-Cd control rod behavior and aerosol formation in severe TcO(2)nH(2)O where n has an average value of 1.63 plus or reactor accidents are examined. Control rod behavior in in. pile minus 0.28. The oxides appear to vary in structure and/or com- and out-of pile expenments is reviewed A mechanistic model 1

l l

i l

l Main Cititi:ns and Abstr2 cts 13 named VAPOR is developed that calculates downward reloca- first discrepancy, however, which is more important of the two tson and simultaneous vaporization behavior of Ag-In-Cd alloy in possibly affecting the interpretation of ex vessel, low thresh-expected after control rod failure. Although cadmium is found to old (100 kev) dosimetry.

be the most volatile constituent of the alloy, all calculations using VAPOR predict that rapid relocation of the alloy down to NUREG/CR-4891: PROPERTIES OF REACTOR FUEL ROD MA-cooler portions of the core results in a small release for all TERIALS AT HIGH TEMPERATURES. Final Summary Report -

three control rod alloy vapors. Potential aerosol formation Severe Core Damage Property Tests Program. PRATER.J.T.; I mechanisms in severe reactor accidents are reviewed. Models COURTRIGHT,E.L. Batteile Memorial Institute. Pacific North-for homogeneous, ion-induced, and heterogeneous nucleation west Laboratories. July 1987, 45pp. 8708240077. PNL-6164, are investigated. Examination of these models indicates that 42316:110.

aeroso! formation occurs in three stages: (1) ion-induced nucle- This report summarizes work sponsored by the U.S. Nuclear ation causes aerosol generabon; (2) son-induced and heteroge- Regulatory Commission Division of Accident Evaluation to in-neous nucleation operate simultaneously; and (3) heterogene. Vestigate those physical properties that are needed to predict ous nucleatson is the dominant mechanism of gas-to-particle the behavior of fuel-rod assemblies dunng a loss-of coolant ac-conversion until equilibrium is established. Preliminary results of cident. The results include a determination of the oxidation ki-the control rod and aerosol behavior observed in PBF Test SFD netics of Zircaloy and Zircaloy uranium oxide mixtures in steam l l 1-4 are presented. Control rod material release in the test is and steam-hydrogen gas mixtures at 1300 to 2400 degrees C, l

! compared to VAPOR predictions. Conclusions from this work viscosity measurements of zirconium-oxide mixtures at 1800 to are presented, and theP impact on source term estimation is as- 2100 degrees C, an estimate of the heat of reaction for the dis-sessed. solution of uranium oxide by molten zirconium at 2000 degrees l NUREG/CR-4884: INTERPRETATION OF BIOASSAY MEASURE- C, and thermal diffusivity measurements on prereacted Zircaloy.

uranium oxide mixtures at 800 to 1500 degrees C. l MENTS. LESSARD.E.T.; YlHUA,X.; SKRABLE,K.W.; et al.

4 Brookhaven National Laboratory. July 1987. 850pp.

8710080369. BNL-NUREG 52063. 42991:133. NUREG/CR 4892: A STUDY OF THE EFFECTS OF PENETRA- )

This is a comprehensive manual desenbing how to compute TION FRAMING ON STEEL CONTAINMENT BUCKLING CA-intakes from both in-vivo and in-vitro bioassay measurements. PACITY. BAKER,W.E.; BUTLER,T.A. Los Alamos National Labo-To date, interpretations of intake have been inconsistent, par- ratory. May 1987. 50pp. 8707300287. LA 10977 MS. 41981:309.

ticu!arly in the early phases after an accidental intake. This Polycarbonate cylinders modeling steel containment struc-manualis armed at completely describing a consistent approach tures were tested to study the effects of different framing de-and instructing others on how to compute intakes and commit. s gns around large penetrations on the static buckling capacity ted organ dose equivalents. Tables for the interpretation of bio- of containments. Two of the four models had equipment hatch assay results are compiled for several hundred radionuclides. penetrations and two had personnel airlock penetrations. Both Measurements which employ a whole-body counter, a thyroid types of models were tested with axial and shear loads as fram-counter, a lung counter, or measurements on excreta can be ing was incrementally added. Results indicate that, for the converted into estimates of intake, based on the tables present- models constructed of polycarbonate, buckling is influenced ed in the appendices. The values in the tables were determined minimally with added framing. Numerical results support the ex.

by using lung, gastrointestinal tract and systemic retention penmental results. Extrapolation of the results to containment models published by the International Commission on Radiologi- constructed under field conditions with prototypic steel matenals cal Protection (ICRP79). In a few cases, pseudo-retention func- is discussed and further testing is recommended.

tions, organ retention functions, and excretion functions were used to generate the tabulated values. The biological and radio- NUREG/CR-4896: CONTAINMENT LOADS DUE TO DIRECT logical input parameters are included in an appendix, and a de- CONTAINMENT HEATING AND ASSOCIATED HYDROGEN senption of the mathematical apprcach that was used to derive BEHAVIOR. Analysis And Calculations With The CONTAIN Code. WILLIAMS,D C.; BERGERON,K.D.; CARROLL D.E.; et al.

the tabulated data is included in the methods section. Calcula-tions for various particle sizes are addressed along wrth meth. Sandia National Laborcies May 1987,164pp. 8708060351, ods to interpret multiple or continuous exposures. ExampMs of SAND 87-0633. 42071:261.

use are based on actual bioassay measurements following acci- One of the most important unresolved issues governing nsk in dental intakes, including tntium, Mn-54, Co-60, Sr-90, Nb 95, ra- many nuclear power plants involves the phenomenon called dioiodines, Cs-137, Ce-141, Ce 144, U 233, U-Nat, and Am-241. direct containment heating (DCH). In which it is postulated that molten conum elected under high pressure from the reactor NUREG/CR 4886: ANALYSIS OF THE NESDIP2 AND NESDlP3 vessel is dispersed into the containment atmosphere, thereby RADIAL SHIELD AND CAVITY EXPERIMENTS. MAERKER,R.E. causing sufficient heating and pressunzation to threaten con-Oak Ridge National Laboratory. May 1987. 57pp. 8708060387. tainment integnty. Models for the calculation of potential DCH ORNL/TM.10389. 42072:274. loads have been developed and incorporated into the CONTAIN Discrete ordinates calculations were made of a series of code for severe accident analysis. Using CONTAIN, DCH sce-measurements performed using the NESTOR reactor at AEE narios in PWR plants having three different representative con-Wintnth. These rneasurements are part of the NESDlP experi- tainment types have been analyzed: Surry (subatmosphenc mental program designed to benchmark methods and data large dry containment), Sequoyah (ice condenser containment),

commonly used in interpreting pressure vessel surveillance do- and Bellefonte (atmosphenc large dry containment). A large simetry placed at either in-vessel or ex vessel locations. Results number of parameter vanation and phenomenological uncertain-obtained using the LEPRICON procedures and updated ELXSIR ty studies were performed. Response of DCH loads to these cross sections indicate agreement to within 10 percent with variations was found to be quite complex; often the results differ measured threshold dosimeter actrvities behind vanous compo. Substantially from what has been previously assumed concern- l nents of a radial shield containing up to 25 cm of water and 24 ing DCH. Containment compartmentalization offers the potential em of steel. Sigriificant discrepancies in the energy range be. of greatly mitigating DCH loads relative to what might be calcu-tween 0.1 and 2 MeV were found to exist with measurements lated using single-cell representations of containments, but the made on the centerline of a 29-cm wide cavity and at vertical actual degree of mrtigation to be expected is sensitive to many locations considerably above the centerline of a 21 cm-wide uncertainties. Dominant uncertainties include hydrogen combus-cavity, however. A logical explanation of the latter discrepancy tion phenomena in the extreme environments produced by DCH is the failure of two independent two-dimensional calculations to scenanos, and factors which affect the rate of transport of DCH property treat the three dimensional effects of vertical cavity energy to the upper Containment. The importance of hydroger streaming. No satisfactory explanation has yet surfaced for the behavior is partly due to the fact that most of the metallic con.

l

1 1

14 Miln Citaticns and Abstracts tent of the dispersed conum is calculated to react rapidly with See NUREG/CR-4000,V01 abstract.  !

steam in the oxygen starved lower containment; hence, the im- ,

mediate energy release is reduced (relative to orygen reactions) NUREG/CR-4904: INVESTIGATION OF STEEL CONTAINMENT BUCKLING FROM DYNAMIC LOADS. BUTLER,T. A.;

but large quantities of hycrogen are rapidly generated. In addi- ,

BAKER,W E.; BABCOCK,C.O. Los Alamos National Laboratory.

tion, DCH loads can be aggravated by rapid blowdown of the May 1987. 54pp. 8708170218. LA 10985 MS. 42171.076.

pnmary system, co- dispersal of moderate quantities of water with the oebns, and quenching of de entrained debns in water; Buckling of free-standing nuclear steel containment buildings these factors act by increasing steam flows which, in turn, ac- from dynamic base excitation was investigated in a combined expenmental/ numerical program. A polycarbonate scale model celerates energy transport. Assessment of the actual contnbu, tion of DCH scenanos to plant nsk would require substantial ad- of a containment building was excited with scaled earthquake ditional work, but it may be noted that containment-threatening transients and single-frequency harmonic transients to deter-loads were calculated for a substantial portion of the scenanos mine the peak base acceleration levels requwed to induce buck-treated for some of the plants considered. ling. Buckling was identified using recorded signals from strain gages and accelerometers, with hgh speed video records, and NUREG/CR 4897: LOW-LEVEL WASTE SOURCE TERM by audibility. Expenmental results are compared with numencal EVALUATION. Review of Published Modeling And Experimental results obtained using a freezing-in-time technique. Results indi-l Work,And Presentation Of Low Level Waste Source Term Mod- cate that, depending on the critenon used, predictions can

! eling Framework And Preliminary Model Development. range from being quite conservative to being unconservative. It SULLIVAN,T.; KEMPF.C.R. Brookhaven National Laboratory. was also found that excitation of higher steel harmonics may in-February 1987, 105pp. 8708060330. BNL-NUREG 52066. fluence the buckling capacity of cor'tainments. Results of this 42072:065. study are considered preliminary in that additional tests need to l This report contains information on the efforts performed to be performed using models fabncated from steel rather than po-date on the Low Level Waste Source Term Evaluation Project, lycarbonate.

the objective of which is development of a model to predict ra-dionuclide release rates from a low level waste cisposal unit. NUREG/CR-4905: DETONABILITY OF H2 AIR-DILUENT MIX.

The approach for model development has been based on a TURES. TIESZEN.S.R.; SHERMAN.M P.; BENEDICK.W.B.; et al.

compartmentalized scheme focused on the four major process. Sandia National Laboratones. June 1987. 215pp. 8707300370.

es of water flow, container degradation, waste leaching and SAND 851263. 41985:133.

waste radionuclides transport to the trench boundaries. This This report desenbes the Heated Detonation Tube (HDT).

stage of the project is focused pnmanly on modeling release Detonation cell width and velocity results are presented for rates from shallow land bunal as currently practiced. Research H(2) air mixtures, undiluted and diluted with CO(2) and H(2)O efforts to this point include charactenzation work (of bunal for a range of H(2) concentration, initial temperature and pres-trenches themselves, of soils and structural features, and of sure. The results show that the addition of either CO(2) or waste forms and containers), review of published modeling H(2)O significantly increases the detonation cell width and work, review of several waste package performance system hence reduces the detonability of the mixture. The results also models, and development of onginal container degradation and show that the detonation cell width is reduced (detonability is waste teaching models. Characterization of the wastes, contain. increased) for increased initial temperature and/or pressure.

ers, and of the site (trench soils and structure) has been based NUREG/CR 4910- RELAY CHATTER AND OPERATOR R E-

~

on the premise that NRC guidance has been put into effect. SPONSE AFTER A LARGE EARTHOUAKE.An Improved PRA NUREG/CR 4900 V01: COMPONENT FRAGILITY RESEARCH Methodology With Case Studies. BUDNITZ,R.J.;

PROGRAM.Phaw I Demonstration Tests. Summary Report. LAMBERT,H.E.; HILL,E.E. Future Resources Associates, Inc.

HOLMAN.G S.; CHOU.C.K. Lawrence Livermore National Labo. August 1987. 204pp. 8708270393. 42390:117.

ratory. SHIPWAY,G.O.; et al. Wyle Laboratories. August 1987. The report addresses methodological weaknesses in the PRA 210pp. 8709090542. UCID 21002. 42563.001. systems analysis used for studying post-earthquake relay chat.

This report desenbes tests performed in Phase I of the NRC ter and for quantifying human response under high stress. An Component Fragility Research Program. The purpose of these improved PRA methodology for relay-chatter analysis is devel-tests was to demonstrate procedures for charactenzing the oped, and its use is demonstrated through analysis of the Zion-seismic fragility of a selected component, investigating how van. 1 and LaSalle-2 reactors as case studies. This demonstration ous parameters affect fragility, and finally using test data to de- analysis is intended to show that the methodology can be ap-velop practical fragility descriptions suitable for application in plied in actual cases. The analysis relies on SSMRP-based probabilistic nsk assessments. A three column motor control methodologies and data bases. For both Zion 1 and LaSalle-2, center housrng motor controllers of vanous types and sizes as it is assumed that loss of offsite power (LOSP) occurs after a well as relays of different types and manufacturers was subject- targe earthquake and that there are no operator recovery ac. {

ed to seismic input motions up to 2 Sg zero penod acceleration. tions. The report also presents an improved PRA methodology To investigate the effect of base flexibility on the structural be- for quantifying operator error under high stress conditions such havior of the MCC and on the functional behavior of the electri- as after a large earthquake. Single-operator error rates are de-cal devises, multiple tests were performed on each of four veloped, and a case study involving an 8 step procedure (estab-mounting configurations: four bolts per column with internal di- lishing feed-and-bleed in a PWR after an earthquake initiated agonal bracing, and two bolts per column with no top or internal accident) is used to demonstrate the methodology.

bracing. Device fragility was charactenzed by contact chatter NUREG/CR 4913: ROUND-ROBIN FAETEST ANALYSES OF A correlated to local in cabinet response at the device location.

1.6-SCALE REINFORCED CONCRETE CONTAINMENT Seismic capacities were developed for each device on the basis MODEL SUBJECT TO STATIC INTERNAL PRESSURIZATION.

of local input motion required to cause chatter; these results CLAUSS.O.B. Sandia National Laboratones. May 1987. 600pp. l were then applied to develop probabilistic fragility curves for each type of device, including estimates of the "high-confidence 8710050162. SAND 87 0891. 42908 037. '

l Analyses of a 1:6-scale reinforced concrete containment low probability of failure' capacity of each.

model that will be tested to failure at Sandia National Laborato-NUREG/CR 4900 V02: COMPONENT FRAGILITY RESEARCH nes in the spnng of 1987 were conduced by the following orga.

PROGRAM. Phase i Demonstration Tests. Appendices. nizations in the United states and Europe: Sandia National Lab-HOLMAN.G S.; CHOU.C.K. Lawrence Livermore National Labo- oratones (USA), Argonne National Laboratory (USA), Electric ratory. SHIPWAY,G D.; et al. Wyle Laboratones. August 1987. Power Research Instrtute (USA), Commissanat a L'Energie Ato-316pp. 8709090300. UCID 21002. 42573 006. mique (France), HM Nuclear Installations inspectorate (U K.),

Mrin Citatlans and Abstracts 15 l

Comitato Nazionale per la ricera e per lo sviluppo deirEnergia fuel / fission product morphology, fuel and cladding oxidation Nucleare e delle Energie Alternative (Italy) U.K. Atomic Energy state, extent of fuel liquefaction and shattering. Tellurium re-Autnonty, Safety & Reliability Directorate (U.K.), Gesellschaft lease behavior is examined relative to the extent of zircaloy fuer Reaktorsicherheit (FRG), Brookhaven National Laboratory cladding oxidation.

(USA), Central Electricity Generating Board (U.K.). Each organi- 1 zation was supplied with a standard information package, which NUREG/CR-4931: RESPONSE OF CENTRIFUGAL AND AXI.

included construction drawings and actual material properties VANE BLOWERS TO LARGE PRESSURE TRANSIENTS.

for most of the materials used in the model. Each organization GREGORY,W.S.; SMITH,P.R. Los Alamos National Laboratory.

worked independently using their own analytical methods. This June 1987, 610pp. 8709090188. LA 11019-MS. 42561:102.

report includes desenptions of the various analytical approaches The effect of large pressure pulses on the operation of cen-and pretest predictions submitted by each organization. Signifi. _ infugal and axi-vane blowers of the types found in ventilation cant milestones that occur with increasing pressure, such as ' systems used in the mening and nuclear industries was investi-damage to the concrete (cracking and crushing) and yielding of gated using the Los Alamos National Laboratory /New Mexico the steel components, and the failure pressure (capacity) and State University fluid dynamics test facility. Three blowers were failure mechanism are described. Analytical predictions for pres- tested for both quasi steady and transient pressures. a 24 in.

sure histories of strain in the liner and rebar and displacements and a 12 in, centnfugal blower and a 33-in. axivane blower are compared at locations where experimental results will be were subjected to pressure pulses at their exhaust and inlet, available after the test. Thus, these predictions can be com- which caused backflow and outrunning flow, respectively. Per.

~

pared to one another and to expenmental results after the test. formance curves were obtained for the first, second, and fourth quadrants.

NUREG/CR-4919: FIELD TESTING OF BENTONITE AND CEMENT BOREHOLE PLUGS IN GRANITE.- KIMBRELL,A.F.; NUREG/CR-4937: INVESTIGATION OF THE MEERS FAULT IN AVERY,T.S.; CAEMEN,J.J. Arizona, Univ. of, Tucson, AZ. July SOUTHWESTERN OKLAHOMA. LUZA,K.V.; MADOLE,R.F.;

Ths techn'ica re rt de's r s n [tu flow tests on bentonite 5pp 8 95204280291 and cement borehole plugs installed in granite, and laboratory The Meers Fault is part of's major system of NW trending back up expenments on similar plugs. Pnor to sealing the hy- faults that form the boundary between the Wichita Mountains draulic conductivity of the boreholes is tested. These measure- and the Anadarko basin in southwestern Oklahoma. The Meers ments, together with core and borehole videologs, permit the Fault is exposed at the surface over a length of at least 26 km; selection of seal test intervals. Commercial expansive cement it strikes N 60 degrees W and offsets Permian conglomerate and standard waterwell sealing bentonite products snd em- and shale. The fault is consistently down to the south, with a placement procedures are used. Transient (short-term) and maximum relief of 5 m near the center of the fault trace. Qua-steady-state (long-term) tests determine the sealing perform- ternary stratigraphy and 10 (14)C age dates constrain the age ance of the plugs under various stress conditions. Laboratory of the last movement of the Meers Fault. The last movement and field expenments confirm the difficulty of obtaining even ostdates the Browns Creek Alluvium, late Pleistocene to early moderately accurate performance values for bentonite plugs as 100-800 yr installed here, due to the simultaneous saturation, swelling, and Holocene, and predates producedthe by the East lastCache Al!uvium' ment, buned fault move consolidation. Nevertheless, conventional installation of readily B.P.

a soil Fan alluvium'etween that dates b 1>400 and 1'100 yr B'P. Two trenches available seals can provide adequate borehole seals (i e, seals excavated across the scarp near Canyon Creek document the with a hydraulic conductivity of about the same magnitude as near surface deformation and provide some information on re-that of intact granite). Bentonite plugs of this type are heteroge- currence. Flexing and warping was the dominant mode of defor-neous and weak. Improved testing procedures and analyses are mation. The stratigraphy in both trenches indicates one surface-needed if actual conductivity values are to be obtained in a rea- faulting event, which implies a lengthy recurrence interval for sonable testing tirne (e.g., months). Laboratory investigations on surface faulting on this part of the fault. Two samples that post-failure mechanisms in cementitious and bentonitic plugs, not date the last fault movement yielded (14)C ages between 1,600 covered in this research study, are needed to identify and pre- and 1,300 yr B.P. These dates are in agreement with dates ob-vent plug failure. tained from the soll buried by fault-related fan alluvium.

NUREG/CR 4925: FISSION PRODUCT BEHAVIOR DURING THE PBF SEVERE FUEL DAMAGE TEST 11. HARTWELL.J.K.; NUREG/CR 4943: PREPARATION OF DESIGN SPECIFICA-PETTI D.A.; HAGRMAN,0.L; et al. EG&G Idaho, Inc. (subs. of TlONS AND DESIGN REPORTS FOR EG&G. Inc.). May 1987. 672pp. 8708140041. EGG 2462. PUMPS. VALVES. PIPING,AND PIPING SUPPORTS USED IN 42159:038. SAFETY RELATED PORTIONS OF NUCLEAR POWER in response to the accident at Three Mile island Unit 2, the PLANTS. RODABAUGH,E.C.; MOORE,S E. Oak Ridge Nat!onal United States Nuclear Regulatory Commission initiated a series Laboratory. June 1987. 78pp. 8710060234. ORNL/TM-10425.

of Severe Fuel Damage (SFD) tests that were performed in the 42922:217.

Power Burst Facility (PBF) at the Idaho National Engineenng Section 111 of the ASME Boiler and Pressure Vessel Code re.

Laboratory to obtain data necessary to understand (a) fission quires the preparation of Design Specifications and Design Re-product release, transport, and deposition; (b) hydrogen genera- ports as part of the design process leading to construction of a l

ten; and (c) fuel / cladding matenal behavior dunng Cegraded nuclear power plant, in Compliance with provisions of Title 10 of core accidents. During the second expenment of this senes, the the Code of Federal Regulations (10 CFR). Guidelines for pre.

SFD Test 1 1, real time release and transport data of certain fis- panng this documentation are contained in nonmandatory Ap-sion products were obtained from on-line gamma spectroscopy pendixes B and C of the ASME Code. This report gives an in-measurements. Liquid and gas effluent grab samples were col- depth review of the ASME Code requirements and guidance, lected at selected penods dunng the test transient. Additional beginning with the first edition of the Code in 1963 through the ,

information was obtained from steamline depositen analysis, 1983 editions Summer 1985 Addenda. Recommendations for From these and other data, fission product release rates and $1bstantial revisions to the Code are presented based on the J

total release fractions are estimated and compared with predict- authors' experience in conducting design documentation audits i l ed release behavior using current models; fissen product distri- of pumps, valves, piping, and piping supports for nuclear power butions and a mass balance are summanzed, and certain prob- plants undergoing NRC review for Operating Licenses. It ts con-able chemical forms are predicted for iodine, cesium, and tellun- cluded that adequate Design Specifications and Design Reports um, Analysis indicates that volatile release from fuel is strong y are absolutely necessary for the normal operating life of a plant influenced by parameters other than fuel temperature; these are and are vital af plant life extension is planned.

I e

l l

16 Miln Citations and Abstracts NUREG/CR-4945:

SUMMARY

OF THE SEMISCALE PROGRAM NUREG/CR-4952: EXPERIMENTAL STUDY OF FILLET WELD (1965 1986). LOOMIS,G G. EG&G Idaho, Inc. (subs. of EG&G, UNDERCUT EFFECTS ON WELDED TUBING STRUCTURES I inc.). July 1987. 273pp. 8703140083. EGG-2509. 42161:001. UNDER CENTRIC AND ECCENTRIC CYCLIC LOADINGS.

This document summanzes significant results from the Semis- OKAILY,AA Calspan Corp. (subs. Arvin Industries / Franklin Re-cale Program, which examined pressunzed water reactor (PWR) search Center). June 1987,85pp. 8708170121. 42171:131, safety issues from 1965 to 1986. Most of these issues were re- The effects of weld undercut defects on the structural behav-

, lated to plant response dunng loss of-coolant accidents and ior of twelve welded tubing test specimens subjected to centnc l operational transients. The Semiscale program utilized a senes and eccentnc cyclic loadings were evaluated. The test program

) of non nuclear, scaled, PWR plant simulators to provide ther, included the design, fabncation, and testing of one group of six mal-hydrauhc data at prototypical pressures and temperatures fillet welded tubing specimens and one group of six reinforced for a wide range of nuclear stfety issues. Presented are: a his, partial penetration, groove-welded tubing specimens under cen-toncal perspective of the Semiscale Program relative to reactor tric and eccentnc loading. Each group consisted of three centn-safety with a catalog of the Semiscale expenmental facilities cally loaded specimens with 0,1/32, and 1/16-in undercuts and data bases, the relationship of Semiscale results to 10 CFR and three eccentncally loaded specimens with 0,1/32, and 1/

Part 50 (Appendix K), the impact of Semiscale results on scal. 16.in undercuts Permanent deflection, d, was recorded after 1, ing expenmental results to full size operating plants, a summary 5, and 10 load cycles at 10% increments of the calculated yield of safety issues that were addressed in Semiscale testing as load, P(y), starting at 50% P(y), through the ultimate load, P(u).

they arouse throughout the operational lifetime of Semiscale, Permanent deflection ratios, d/do, of a defective specimen the contnbutions of the Semiscale Program to safety technolo- versus that of a similar, but sound specimen (i.e., no undercut-gy, a desenption of phenomena obearved dunng Semiscale ting) at the same load level and number of cycles were then testing, the impact of Semiscale data on code development and a aM AnaWs d W resWs W to smal concbssns e lated to the effects of fillet weld undercut on permanent deflec-assessment, and finally, major conclusions and accomplish

• ments of the Semiscale program tion ratio and ultimate load values of welded tubing structural members under centnc and eccentnc cyclic loading.

NUREG/CR 4946: DAVIS BESSE UNCERTAINTY STUDY.

DAVIS.C.B. EG&G Idaho, Inc. (subs. of EG&G, Inc.). August NUREG/CR-49G3: CORRELATION OF RADIOIODINE RESU.

1987,83pp 8709090483. EGG 2510. 42564:099. spENSION WITH TEMPERATURE AT TMI-2. DANIEL.J.A.; ,

DANIEL E.A.; SETH,E.L. Daniel & Associates, Inc. July 1987. I The uncertainties of calculations of loss of feedwater tran- 137pp. 8708040231. DA TR/8705. 42037:054.

sients at Davis-Besse Unit I were determined to address con- This report addresses the observed long term behavior of ra. l cerns of the U S. Nuclear Regulatory Commission relative to the diciodine in specific locations in the TMI-2 Auxiliary and Fuel l effectiveness of feed and bleed cooling. Davis-Besse Unit 1 is a Handling Buildings, and provides data on the behavior of I-131 i pressunzed water reactor of the raised-loop Babcock & Wilcox at relatively low temperatures, (50 degrees 85 degrees F) and design. A detaded, quakty assured RELAPS/ MOD 2 model of non- equihbrium conditions, since the building ventilation sys-Davis Besse was developed at the Idaho National Engineenng tems were in operation. This report also discusses the observe:1 Laboratory. The model was used to perform an analysis of the effect of changes in the daily concentration of radiciodine due loss of feedwater transient that occurred at Davis- Besse on to diurnal temperature cycles, and estabhsnes a numerical rela-June 9,1985. A loss-of feedwater transient followed by feed tionship between radioiodine concentration and ambient temper-and bleed cochng was elso calculated. The evaluation of uncer, ature.

tainty was based on the compansons of calculations and data, compansons of different calculations of the same transient, sen. NUREG/CR 4954: LONG TERM PERFORMANCE OF SPENT sitivity calculations, and the propagation of the estimated uncer- FUEL WASTE FORMS. MEANS.J L.; MARKWORTH,A.J.;

tainty in initial and boundary conditions to the final calculated MCCOY,J.K.; et al. Battelle Memonal Institute, Columbus Lab-results. oratones. September 1987. 168pp. 8710090161. BMI-2154. I 43009.058.

NUREG/CR 4951: NEPHROTOXICITY OF URANYL FLUORIDE This report desenbes the results of a 2 year expenmental pro-AND REVERSIBILITY OF RENAL INJURV IN THE RAT. gram to investigate the leaching / dissolution charactenstics of ,

DIAMOND.G.L.; GELEIN R.M.; MORROW,P E.; et al. Rochester, spent fuel in simulated groundwater. Expenments were con-l Univ. of, Rochester, NY. September 1987.102pp.8710060148. ducted to evaluate the leaching of unitradiated UO(2) in simulat.

42921d 37. ed groundwater both in the presence and absence of an exter- l The oblective of the study was to determine the seventy and nal radiation field. Other expenments charactenzed the dissolu-duration of renal injury produced in the rat from exposure to low tion behavior of spent fuel under the same conditions as for levels of uranyl fluonde (UO(2)F(2)). Rats received multiple i.p~ UO(2). Leach data were interpreted in terms of rates and mech-injections of UO(2)F(2) (cumutative dose: 0 66 or 1.32 mg U/kg anisms of radionuclides release, thus providing insight on leach body wt). Renal injury was charactenzed pnmanly by cellular behavior under realistic repository conditions. Uranium release and tubular necrosis of the pars recta of the proximal tubule from both unirradiated and irradiated UO(2) appeared to be sol-ubility controlled, with only minor leachate compositional effects (S(2) and S(3)). with less severe cellular injury to the thick as-on the uranium concentrations. When leached in the presonce cending limb of the loop of Henie and collecting tubule. The of a radiation field, unirradiated UO(2) produced elevated urani-iniury was apparent early in the dosing phase of the study, at a um concentrabons only in the bone teachates. The leaching of time when renal uranium levels were between 0.7 1.4 ug U/g, both cesium and iodine from spent fuel was rapid and hnear and was most severe when the renal uranium burden was be- with the surface-area to-volume (S/V) ratio. Similar to cesium tween 3 4 5 6 ug U/g. Repair of the injury was rapid, with and iodine, strontium release was strongfy affected by leachate complete restoration within 35 days after the exposure. Associ- composition, with greater release essociated with lower pH. A ated with the injury were numerous abnormalities in kidney func. model desenbing spent fuel /UO(2) dissolution was developed, tion, including impaired tubular reabsorption, proteinuna and en- complementing the expenmental studies.

zymuna, which appear to be temporalty related, to van &ble de-grees, to the progression of renal injury. Renal injuiy r eceded NUREG/CR-4956: SYSTEM PERFORMANCE OF HIGH-LEVEL and outlasted functional abnormabties as assessed by unnafysis WASTE PACKAGE COMPONENTS. NICOLOSI.S L.;

and clearance measurements. KURTH.R E.; OUAYLE,S F.; et al. Battelle Memonal Institute.

September 1987. 300pp. 8710060258. BMI-2156. 42921241.

Main Cit 2ti:ns and Abstracts 17 This report presents results of analytical and experimental the desired financial assurance is one or more newly-formed studies that can be integrated to describe the performance of captive insurance companies. Potential benefits of the rulemak.

nuclear waste packages at the system level. Several water ing include administrative time to secure such funds, and the chemistry modules were developed, and groundwater radiolysis possibihty of more rapid cleanup with correspondingly reduced studies were performed to examine the production of radiolytic occupational and public exposure to radioactive matenals.

species at the outside surface of waste packages. Uncertainty analysis studies with a UO(2) water chemistry model assessed NUREG/CR-4962: METHODS FOR THE ELICITATION AND USE the importance of input parameters on the variability of calculat- OF EXPERT OPINION IN RISK ASSESSMENT. Phase 1 A Cnti-ed solubihties. A shell for a computer code was developed that cal Evaluation And Direchons For Future Research.

provides an alternative method for examining the system per- MOSLEH,A.; BIER,V.M.; APOSTOLAKIS.G. Pickard, Lowe &

formance of high-level waste packages. Integral experiments Garrick, Inc. August 1987. 85pp. 8709090358. PLG 0533.

were conducted to examine interactions between components 42566 268.

of the waste package and repository, bnnging together ele- The purpose of this work is to entically review and evaluate ments of repository environments with components of spent-fuel the elicitation and use of expert opinion in probabilistic nsk as-waste forms. The system performance investigations showed sessment (PRA) in light of the available empirical and theoreti-that radiolysis effects are not significant when the gamma cal results on expert opinion use. PRA practice is represented source is shielded by the overpack. It was observed that the by five case studies selected to represent a variety of aspects production of oxalate is related to the bicarbonate concentration of the problem: (1) Assessments of component failure rates and and dose rate. Other modeling studies showed that the dissolu- maintenance data, (2) Recent assessments of seismic hazard tion of UO(2) may be influenced significantly by solution pe and rates, (3) Assessments of containment phenomenology, (4) As-chlonde concentration. sessments of human error rates, and (5) Accident precursor NUREG/CR 4957: SURVEY OF GEOPHYSICAL TECHNIQUES studies. The review has yielded mixed results. On the negative FOR SITE CHARACTERIZATION IN BASALT, SALT AND TUFF. side, there appears to be little reliance on normative expertise JONES,G.M.; BLACKEY,M.E.; RICE.J.E.; et al. Weston Geo, in structuring the process of expert opinion elicitation and use; physical Corp. July 1987,149pp. 8709110098. 42609:161. most applications instead rely pnmarily on the contmon sense This report surveys various geophysical techniques that would of the experts involved in the analysis, which is not always an have application dunng site charactenzation activities in connec- adequate guide. On the positive side, however, there is evi-tion with the high-level. waste storage project. Geophysical tech. dence that expert opinions can in fact be used well in practical niques may help determine the nature and extent of faulting in settings. Suggestions are grven for Phase 11 work to enhance the target areas, along with structural information which would the applicability and use of appropriate expert opinion methods.

be relevant to questions conceming the future integnty of a high-level-waste repository. Following an introductory chapter NUREG/CR-4972: TWO-PHASE FLOW REGIME TRANSITION which describes the vanety of geophysical techniques available, CRITERIA IN POST DRYOUT REGION BASED ON FLOW VIS- ,

subsequent chapters focus on particular geophysical applica- UAllZATION EXPERIMENTS. OBOT,N.T.; ISHil,M. Argonne Na-tions to four rock types basalt, bedded salt, domal salt and tuff tional Labcratory. June 1987, 56pp. 8709030565. ANL-87 27.

charactenstic of the sites originally proposed for site charac. 2 terization. No one geophysical method can adequately charac- us study of film boiling using photographic and high terize the geological structure beneath any site. A combination s eed motion-picture methods was carried out to determine the of techniques is required. The seismic reflection method, which flow regime transition critena in the post CHF region. An ideal-is generally considered to be the most incisive of the geophysi- tzed inverted annular flow was obtained by introducing a liquid cal techniques, has to date provided only marginal information jet of Freon 113 through a nozzle, precisionly centered with re-on structure at the depth of the proposed repository at the Han- spect to the internal diameter of the test section, with an annu-ford, Washington, site, and no useful results at all at the Yucca far gas flow. The respective ranges for hquid and gas exit ve-Mountain, Nevada, site. This result appears to be partially due locities were 0.05-0.5 and 0.03-7.9 m/s. Nitrogen and helium to geological complexity beneath these sites, but may also be were used in the study. For the present configuration, there are partially attributed to the use of inappropriate acquisition and four basic flow regimes. Beginning from the nozzle exit, there is processing parameters. It appears that to adequately character. smooth, inverted annular flow section with hquid in the core and ize a site using geophysics, modifications will have to be made annulus; followed by the rough wavy section with an to standard techniques to emphasize structural details at the gas intactin the,d core, the agitated and the dispersed flow hqui regimes depths of interest. For a given hqu,d i jet velocity, the axial extent of each flow regime decreases with increasing gas velocity through the annu.

NUREG/CR-4958: IMPACT OF PROPOSED FINANCIAL ASSUR. lus. Generalized transition entena and simplified correlations for ANCE REQUIREMENTS ON NUCLEAR MATERIALS UCENS- the axial hmits of the vanous flow regimes have been devel-EES. HENDRICKSON P.; SCOTT,M.J.; MULLEN,M F,; et al. Bat- oped, by extending the results of previous studies on adiabatic telle Memorial Institute, Pacific Northwest Laboratories. Septem- inverted annular tiow.

ber 1987.137pp. 8710070036. PNL-6233. 42965:018.

The NRC is considering a possible rulemaking that would re- NUREG/CR 4973: INTRAPLATE SEISMICITY OUTSIDE OF THE quire certain materials hcensees to demonstrate financial abihty UNITED STATES. SCHWEITZER.J.; GLOVER,L. Virginia Poly-to clean up accidental releases of radioactive matenals. The technic Institute & State Univ., Blacksburg. VA. July 1987.88pp.

rulemaking would potentially affect approximately 16,350 NRC 8708040243.42036:327, and Agreement State keensees. This report was prepared to A survey of intraplate seismicity in eastem Canada, Austraha, provide background information and analysis for the potential and western Europe was undertaken. Eastem Canada has been rulemaking. Specific topics examined in the report include: 1) divided into four seismic zones; namely, LaMalbare (Quebec),

charactenstics of potentially affected hcensees,2) the availabil- westem Quebec, New Brunswick, and Baffin Island. Austraha, a ity anc cost of vanous financial assurance mechanisms,3) the relatively aseismic continent surrounded by very seismically l financial impacts to hcensees (including heensees classified as active areas, has been divided into three seismic zones, west-small businesses) of providing $2M of assurance per heensee ern Austraha, south Austraha, and eastern Austraka. Intraplate I and a sliding amount of assurance tied to risk, 4) the cost of seismicity in westem Europe is concentrated in ceniral and administenng a financial assurance rule, and 5) overall benefits northem Europe. Based on this survey, several general obser-and costs. Tabular information on past matenal heensee acci. vations about intraplate seismicity can be made. Much of the dents and cleanup efforts is also included. The financial assur- seismic activity in these intraplate regions is relatively shallow ance mechanism that appears to be most suitable for providing and is frequently locahzed in basement rocks. Seismic events

18 Maln Citations and Abstrzcts I

are frequently attributed to movement along pre existing faults, sions that are matenally different from those of a previous study although surface faulting rarely accompanies these earth- that did not consider these effects.

quakes. Paloo-nft zones seem to be particularly susceptible to reactivation. Plate intoractsons and asthenosphere drag both NUREG/CR 4981: A SAFETY ASSESSMENT OF THE USE OF play a role in development of a regional honzonal maximum GRAPHITE IN NUCLEAR REACTORS LICENSED BY THE U.S.

pnncipal compressive stress, but the actual triggenng mecha. NRC. SCHWEITZER,0.G.; GURINSKY,D.H.; KAPLAN,E.; et al.

nism and localization cause for this fault reactivation in intra. Brookhaven National Laboratory. September 1987. 42pp.

plate regions is still only poorly understood. 8710060161, BNL NUREG 52092. 42921:096.

NUREG/CR 4974: INTRAPLATE SEISMICITY IN THE EASTERN grapt te b in n on UNITED STATES. BOLLINGER G.A.; rgy accumu a o a d re EHLERS.E.G.;

MOSES,M.J. Virginia Polytechnic Institute & State Univ., Blacks- leases in order to assess what role, if any, a stored energy re-burg, VA. July 1987, 28pp. 8708040340. 42036:001, lease can have in initiating or contributing to hypothetical graph-ste burning scenarios in research reactors. It also addresses the Causes for intraplate earthquakes in the eastern United question of graphite ignition and self sustained combustion in States are indeed complex. On a regional scale, earthquakes are spatially and temporally stationary, at least on a scale of the event of a loss.of coolant accident (LOCA). The conditions decades. Spatially, the seismicity occurs in distinct zones super- necessary to initiate and maintain graphite burning are summa-nzed and discussed. From analyses of existing information it is posed on a less active regional background level. Histoncally, concluded that only stored energy accumulations and releases moderate to large earthquakes have occurred within the seismic below the burning temperature (650 degrees C) are pertinent. It zones defined by microsutnicity. Recent paleoseismicity stud-les suggest that these larger earthquakes can reasonably be is shown that there is no evidence from the Chernobyl event expected to occur into the future. Earthquake focal depths in that stored energy releases played a role either initiating or con-the eastem U.S. tend to be within the upper crust with little rela- tnbuting to this accident. The conclusions from these analyses are that the potential to initiate or maintain a graphite buming tion to surficial features. Because most eastern U.S. seismic incident is essentially independent of the stored energy in the zcnes are buned in the subsurface, geophysical methods are graphite, and depends on other factors that are unique for each used to infer the presence of possible causal geologic struc-tures at depth. Gravity data are useful in locating density (htho- research reactor and for Fort St. Vrain. There is no new evL logic) changes within the crust and Seismic activity tends to dence associated with either the Windscale Accident or the concentrate within gravity saddles formed at the intersection of Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in northeast and northwest trending gravity anomalies. Seismically this review.

important plutons and Triassic basins which are associated with seismic activity have been located by their magnetic signatures. NUREG/CR-4982: SEVERE ACCIDENTS IN SPENT FUEL The eastern U.S. is interpreted to be composed of a mosaic of POOLS IN SUPPORT OF GENERIC SAFETY ISSUE 81 allochthonous suspect terranes. This hypothesis suggests that SAILOR,V.L.; PERKINS.K.R.; WEEKS J.R.; et al. Brookhaven block boundanes ere zones of crustal weakness. National Laboratory. July 1987. 156pp. 8708120084. BNL-NUREG/CR 4975: A REVIEW OF THE RESOLVING POWER OF NUREG-52093. 42128:097, REFLECTION SEISMOLOGY METHODS TO DETECT SUB. This investigation provides an assessment of the likelihood SURFACE FAULTS AND/OR CHANGES IN LAYER THICK- and consequences of a severe accident in a spent fuel storage NESS. COSTAIN.J.K.; CORUH C. Virginia Polytechnic Institute & pool the complete draining of the pool. Potential mechanisms State Univ., Blacksburg, VA. July 1987, 37pp. 8708050045. and conditions for failure of the spent fuel, and the subsequent 42062:110. release of the fission products are identified. Two oider PWR Aspects of the resolving power of seismic reflection data in and BWR spent fuel storage pool designs are considered based detecting subsurface faults are reviewed. Resolution by reflec- on a preliminary screening study which tned to identify vulnera-tion seismology of displacements across subsurface faults is bilities. Internal and extemal events and accidents are as-coupled with the detection and resolution of thin beds. Limits of sessed. Conditions which could lead to failure of the spent fuel detail that can be denved from seismic data are greatly affected Zircaloy cladding as a result of cladding n.pture or as a result of by data acquisition and processing parameters. Temporal and a self sustaining oxidation reaction are presented. Propagation spatial bandwidths as keys for seismic resolution are deter- of a cladding fire to older stored fuel assemblies is evaluated. ,

mined by source, recording system, absorption, patterns, tempo- Spent fuel pool fission product inventory is estimated and the j ral frequency range, group interval, temporal and spatial noise, releases and consequences for the vanous cladding failure sce-Optimum time wavelets are important for maximum resolution in narios are provided. Possible preventive or mitigative measures I time which also affects resolution in the spatial domain. Re- are qualitatively evaluated. The uncertainties in the risk estimate I sponse of vibrators and the use of force control units are re, are large, and areas where additional evaluations are needed to viewed along with the effects of source and receiver arrays reduce uncertainty are identified.

used dunng data acquisition. Phase changes due to decoupling and phase compensation are included in the discussion. The NUREG/CR-4986: RADIATION DOSE ESTIMATES AND importance of balancing the effects of intnnsic damping before HAZARD EVALUATIONS FOR INHALED AIRBORNE stack is discussed, as are the eflects of data windowing in time RADIONUCLIDES. Final Report. MEWHINNEY,J.A. Lovelace l and space. Parameters affecting resolution along time and spa- B omed & Environmental Research Institute. September 1987.

tial axes are given to show that the Fresnel zone introduces an 100pp. 8710050547, 42905:122.

mportant limitation on the detail that can be derived from seis- This is the final report for a project whose objective was to conduct confirmatory research on physical chemical characteris-tics of aerosols produced dunng manufacture of mixed plutoni-NUREG/CR 4978: THE COOLDOWN ASPECTS OF THE TMI-2 um and uranium oxide nuclear fuel, to determine the radiation ACCIDENT. THEOFANOUS T.G. California, Univ. of Santa Bar- dose distribution in tissues of animals after inhalation exposure bara, CA. August 1987. 63pp. 8709150313. 42678:239. to representative aerosols of these matenais, and to provide es-  ;

The cooldown of the TMI 2 reactor vessel due to high pres- timates of the relationship of radiation dose and biological re-sure injection that occurred at 200 minutes into the accident is sponse in animals after such inhalation exposure. This report is re- examined. Flow regimes and condensation heat transfer in divided into three chapters which summanze the results of the cold legs and downcomer are considered. The presence of these investigations. The first report chapter summanzes the noncondensibles (hydrogen) and a mecharusm lea %ng to its ac- physical chemical charactenzation of samples of aerosols col-cumulation around the condensation interfaces lead to conclu- lected from gloveboxes at industrial facilities dunng normal op-

Main Citations and Abstracts 19 erations. This chapter provides insights into key aerosol charac- This report presents equations and parametric curves to pre-4ristics which are of potential importance in determining the bi- dict erosive-corrosive wear rates in low carbon steel secondary

> logical fate of specific radionuclides contained in the particu- piping systems of nuclear power plants. Two mechanisms caus-late that would be inhaled by humans following accidental re- ing the wear are identified and analytic expressions to predict lease. The second chapter describes the spatial and temporal the wear are presented. One mechanism, wear due to oxide distribution of radiation dose in tissues of three species of ani- dissolution, is present in sing 6e and two-phase flow lines; the mais exposed to representative aerosols collected from the in' other, wear by droplet impact induced fatigue, is only applicable dustnal facilities. These inhalation studies provide a basis for for two-phase flows. Pararnetric curves are presented to predict companson of the influence of physical Chemical form of the in- the dissolution wear as a function of temperature, pH and char.

haled particulate and the variability among species of animalin acteristic (friction) velocity. An algebraic expression is used for the radiation dose to tissue. The third chapter details the rela-droplet impact wear predictions as a function of superficial gas tionship between radiation dose and biological response in rats velocity and thermodynamic quality. Four mitigation strategies ve e Ifesp o incorporated into the models (alloying, pH increase, moisture re-ary rd n. This u nduct moval and velocity decrease) are introduced and the effective-rats and assuming results to be applicable to humans, indicates ,

ness of each is discussed. The models wear predictions are not that the hazard to health due to inhalation of these industrial aerosols is not different than previously determined for laborato- expected to be quantitatively exact, but should yield accurate ry produced aerosol of PuO(2). qualitative rankings of systems susceptible to erosive-corrosive

• • • U NUREG/CR 5002: METHODS FOR RECURRING LOSS TESTS.

JOHNSTON,J.W.; LITTLEFIELD,J.; KINNISON,R.R.: et al. Bat- NUREG/CR 5008: DEVELOPMENT OF A TESTING AND ANALY-telle Memorial Institute, Pacific Northwest Laboratories. Septem. SIS METHODOLOGY TO DETERMINE THE FUNCTIONAL ber 1987. 58pp. 8710080312. PNL-6249. 42990:261. CONDITION OF SOLENOID OPERATED VALVES.

Development of a program to use material control and ac- MEININGER,R.D.; WEIR,T.J. Pentek, Inc. September 1987, counting data for detecting recurring losses of nuclear material 35pp. 8710060130. 42921:060.

is described and some potential statistical methods are dis- This report describes the Phase i SBIR research conducted cussed. A moving sam test, including a moving vanation for false alarm control, ,s to develop and test a technique to determine the functionality of i desenbed. Three non-parametric tests are pe P discussed because of simplicity of use and because an estima.

tion of the vanance of the process difference is not required. the measurement and analysis of inrush current. The technique Considerations for a computer system to collect and process re. developed uses a clip-on current probe, thus enabling all meas-curring loss test data are identified and discussed. urements to be made from outside the reactor building without 9* #*

NUREG/CR-5007: PREDICTION AND MITIGATION OF ERO- s len id perated valve is analyzed in real time using a person-SIVE-CORROSIVE WEAR IN SECONDARY PIPING SYSTEMS al computer and Fast Fourier Transform (FFT) techniques.

OF NUCLEAR POWER PLANTS. KECK,R.G.; GRIFFITH,P.

Massachusetts Institute of Technology, Cambndge, MA. Sep-tember 1987. 46pp. 8710050583. 42911:163.

l

l  !

( ,

1 Secondary Report Number index j

This index lists, in alphabetical order, the performing organization-issued report codes for the l NRC contractor and international agreement reports in this compilation. Each code is cross- ,

referenced to the NUREG number for the report and to the 10-digit NRC Document Control l System accession number. l l

I SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER ANL 8716 NUREG/CR-4744 V01 N2 ORNL/TM 10330 NUREG/CR4817 ANL47 27 NUREG/CR 4972 ORNL/TM-10364 NUREG/CR.4860 BMI 2154 NUREG/CR-4954 ORNL/TM 10389 NUREG/CR-4886 BMI 2156 NUREG/CR-4956 ORNL/TM 10425 NUREG/CR4943 ORNL/TM-9593 NUREG/CR 4219 V04 N1 BNL-NUREG 51454 NUREG/OR-2331 V06 N4 BNL NUREG-51581 BNL-NUREG-52022 NUREG/CR 2907 V05 NUREG/CR-4810

.h PNL-5810 hhhhfhhd$-44 V02 NUREG/CR-4418 BNL NUREG-52063 NUREG/CR 4884 PNL 5784 NUREG/CR-4640 BNL NUREG-52066 NUREG/CR-4897 PNL 5822 NUREG/CR 4583 V03 BNL NUREG 52092 NUREG/CR-4981 PNL-5832 NUREG/CR4590 V01 BNL-NUREG 52003 NUREG/CR-4982 PNL 5832 NUREG/CR 4590 V02 DA TR/8/05 NUREG/CR-4953 PNL 60321 NUREG/CR 4768 V01 EGG 2421 NUREG/CR 4407 PNL 6032 2 NUREG/CR4768 V02 EGG 2462 NUREG/CR-4925 PNL 6045 NUREG/CR4850 EGG 2469 NUREG/CR-4731 V01 PNL4164 NUREG/CR4891 PNL-6233 NUREG/CR-4958 EGG 2473 NUREG/CR-4747 V01 EGG-2476 EGG 2488 NUREG/CR-4769 NUREG/CR-4457 h8 2573 SAND 82-2478 NUREG/CR 3024 EGG-2495 NUREG/CR-4844 DAFT SAND 85-0790 NUREG/CR-4216 .

EGG 2501 NUREG/CR-4876 SAND 851263 NUREG/CR 4905 i EGG 2509 NUREG/CR4945 SAND 86-0419 NUREG/CR4534 '

EGG 2510 NUREG/CR 4946 SAND 86-1797 NUREG/CR 4710 LA 10977 MS NUREG/CR 4892 SAND 86 2419 NUREG /CR4767 LA 10985-MS NUREG/CR 4904 SAND 86 2483 NUREG/CR-4871 LA 11019-MS NUREG/CR4931 SAND 86-2752 NUREG/cR4805 V01 MEA 2194 NUREG/CR4863 SAND 87 0179 NUREG/CR-4834 V01 MEA 2207 ORNL4187 NUREG/CR-3228 VOS NUREG/CR-4760 hy hhhfhhM SAND 87 7029 NUREG/CR-4846 ORNL4374 NUREG/CR-4865 SAND 87 7064 NUREG/CR4781 DRFT ORNL/NSIC-200 NUREG/CR-2000 V06 N6 SAND 87-7066 NUREG/CR4870 I ORNL/NSIC 200 NUREG/CR-2000 V06 N7 UCID 20092 NUREG/CR-4161 V03 i ORNL/NSIC-200 NUREG/CR 2000 V06 N8 UCID-21002 NUREG/CR-4900 V02 ORNL/TM 10015 NUREG/CR-4577 0C10 21002 NUREG/CR-4900 V01 I

21

l l

k 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.

APOSTOLAKIS,G. BINNALL,E.P.

NUREG/CR4962: METHODS FOR THE ELICITATION AND USE OF NUREG/CR-4161 V03: CRITICAL PARAMETERS FOR A HIGH-LEVEL EXPERT OPINION IN RISK ASSESSMENT. Phase i . A Cntical Evalua- WASTE REPOSITORY. Volume 3. Salt.

tion A,d Directions For Future Research.

BLACKEY,M.E.

ARELLANO,F.E. NUREG/CR-4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR NUREG/CR4871: RESULTS FROM THE DCH 1 EXPERIMENT. SITE CHARACTERIZATION IN BASALT, SALT AND TUFF.

ARMBRUSTER.J G. BLOSE,R.E.

NUREG/CR-4851: SEISMICITY 1886-89 IN THE SOUTHEASTERN NUREG/CR.3024: SUSTAINED CONCRETE ATTACK BY LOW-UNITED STATES.The Aftershock Sequence Of The Charleston. South TEMPERATURE. FRAGMENTED CORE DEBRIS.

Caroline Eadhquake.

BLUHM.D.

ARNOLD,W.D. NUREG/CR 4870: AN EVALUATION OF THE EFFECTS OF DESIGN NUREG/CR4865 THE SOLUBILITY OF ELECTRODEPOSITED TC(IV) DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT OXIDES. BUILDINGS AVERY,T.S. BOCClo J.

NUREG/CR4919: FIELD TESTING OF BENTONITE AND CEMENT NUREG/CR 4810: EVALUATION OF DIESEL UNAVAILABILITY AND BOREHOLE PLUGS IN GRANITE. RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

BABCOCK.C.D. BOLIG,C.A.

NUREG/CR-4904: INVESTIGATION OF STEEL CONTAINMENT BUCK. NUREG/CR4846: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS LING FROM DYNAMIC LOADS. SAFETY ANALYSIS. Phase 2 Final Report.

BAGGS,R.B. BOLLINGER,G.A.

NUREG/CR-4951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE* NUFsEG/CR4974: INTRAPLATE SEISMICITY IN THE EASTFAN VERSIDILITY OF RENAL INJURY IN THE RAT

• UNITED STATES.

BAKER,W.E*

NUREG/CR4892: A STUDY OF THE EFFECTS OF PENETRATION BOLT NUdlEG/S.E' CR4760:

TEST OF 6-IN.. THICK PRESSURE VESSELS.Senes NU EG/C 4 iV T OF EL NTA ME BUCK.

nWme ate Test Wssel WGang hm O W WM LING FROM DYNAMIC LOADS. Matenal BARSELL,A.W. BRADLEY,0.R.

NUREG/CR4846: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS NUREG/CR-3024: SUSTAINED CONCRETE ATTACK BY LOW-TEMPERATURE, FRAGMENTED CORE DEBRIS.

SAFETY ANALYSIS. Phase 2. Final Report.

BASS,B.R. BROCKMANN.J.E.

NUREG/CA-4760. TEST OF 6-IN. THICK PRESSURE VESSELS Senes NUREG/CR4871: RESULTS FROM THE DCH-1 EXPERIMENT.

3: Intermediate Test Vessel V-8A.Teanng Behavior Of Low Upper Shelf BROMERY,R.W.

Matenat NUREG/CR4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR BELANGEft,R. SITE CHARACTERIZATION IN BASALT, SALT AND TUFF. .

NUREG/ta-4884. INTERPRETATION OF BIOASSAY MEASUREMENTS.

BROWZIN.B.S.

BENEDICK W.B. NUREG/CP.0088 TRANSACTIONS OF THE 9TH INTERNATIONAL NUREG/CR 4905: DETONABILITY OF H2 AIR %.DT MtXTURES CONFERENCE ON STRUCTURAL MECHANICS IN REACTOR TECHNOLOGY. Panel Session JK: Structural And Mechanecal Engineer.

BENSON,S.M. ~ ing Research At The U S. Nuclear Regulatory Commission.

l NUREG/CR-4161 V03. CRITICAL PARAMETERS FOR A HIGH-LEVE.

WASTE REPOSITORY. Volume 3 Salt. BRY AN.R.H.

NUREG/CR4760. TEST OF 6-IN. THICK P9 ESSURE VESSELS.Senes BERGERON,K.D. 3. Intermediate Test Vessel V-8A.Teanng Behavior Of Low Upper Shelf NUREG/CR-4896 CONTAINMENT LOADS DUE TO DIRECT CONTAIN- Matenal.

MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysis {

And Calculations With The CONTAIN Code. BRYANT,JL j NUREG/CR4640. HANDBOOK OF SOFTWARE QUALITY ASSURANCE ,

BERMAN.M. TECHNIQUES APPLICABLE TO THE NUCLEAR INDUSTRY.

NUREG/CR-4905: DETONABILITY OF H2. AIR DILUENT MIXTURES BRYSON.J.W.

BE V AN.R. NUREG/CR4760: TEST OF 6-IN. THICK PRESSURE VESSELS Senes i NUREG.1285 NRC STAFF EVALUATION OF THE GENERAL ELECTRIC 3. Intermediate Test Vessel V 8A Teanng Behavior Of Low Upper Shelf l COMPANY NUCLEAR REACTOR STUDY (" REED REPORT"). Matenal BIER.V.M. . BUDNITZ,R.J.

NUREG/CR4962. METHODS FOR THE ELICITATION AND USE OF NUREG/CR 4910 RELAY CHATTER AND OPERATOR RESPONSE EXPERT OPINION IN RISK ASSESSMENT. Phase I . A Cntcal Evatus- AFTER A LARGE EARTHOUAKE An improved PRA Methodology With ton And Duectons For Future Research. Case Studies.

23 l

l

24 P:rtenil Auth:r Ind:x  !

l l

BUTLER.T.A. COURTRIGHT,E.L NUREG/CR4892: . A STUDY OF THE EFFECTS OF PENETRATION NUREG/CR4891: PROPERTIES OF REACTOR FUEL ROO MATERIALS FRAMING ON STEEL CONTAINMENT BUCKLING CAPACITY. AT HIGH TEMPERATURES Final Summary ReNrt

• Severe Core NUREG/CR4904: INVESTIGATION OF STEEL CONTAINMENT BUCK. Damage Property Tests Program.

LlNG FROM DYNAMIC LOADS.

CRAMONO WA CAMPSELLJ.E.

NUREG/CR-2452: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL NUREG/CR4710: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF OF RADIOACTIVE WASTE. Fmal Report. A COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR. Case Study.

CA". ROLL.D.E.

NUREG/CR4896: CONTAINMENT LOADS DUE TO DIRECT CONTAIN. CRANWELL,R.E MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR.Analygas NUREG/CR 2452: AISK METHODOLOGY FOR GEOLOGIC DISFOSAL And Calculations With The CONTAIN Code. OF RADIOACTIVE WASTE. Final Report CASE.F.L CRONE.A.J.

NUREG/CR-4865: THE SOLUBILITY OF ELECTRODEPOSITED TC(IV) NUREG/CR4937: INVESTIGATION OF THE MEERS FAULT IN SOUTH-OXIDES. WESTERN OKLAHOMA.

CHA807,G.E. CRONENBERG,A.W.

NUREG/CR4884. INTERPRETATION OF BIOASSAY MEASUREMENTS. NUREG/CR.4925: FISSION PRODUCT BEHAVIOR DURIN3 THE PBF CHEN,Y.S. SEVERE FUEL DAMAGE TEST 14 NUREG.1284: PROGRAM PLAN FOR CORRECTION OF U.S. INSTRU.

CULLEN,W.H.

MENT DEGRADATION OR FAILURE IN THE UPPER PLENUM TEST FACILITY (UPTF) IN THE FEDERAL REPUBLIC OF GERMANY.

NUREG/CR4863: SURFACE SPECTROSCOPY OF PRESSURE VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMEC IN i CHERDACK,R. PWR ENVIRONMENTS. j NUREG/CR.4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATION SYSTEMS. DAEMEN.J.J.

NUREG/CR4919: FIELD TESTING OF BENTONITE AND CEMENT l CHOPRA,0.K.

~

BOREHOLE PLUGS IN GRANITE.

NUREGICR4744 VOI N2- LONG. TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Somiannual DANIEL,E.A.

Report,Apnl. September 1986. NUREG/CR-4953: CORRELATION OF RADICIODINE RESUSPENSION WITH TEMPERATURE AT TMI-2.  ;

CHOU.C.K.

NUREG/CR4900 V01: COMPONENT FRAGILITY RESEARCH DANIEL,J.A.

NUREG/CR.4953: CORRELATION OF RADIOIODINE RESUSPENSION NU E / 4 V bMPO N lll RESEARCH WITH TEMPERATURE AT TMI-2.

PROGRAM. Phase i Demonstrabon Tests. Appendices.  ;

CHRISTENSEN,D. DAVIS,C.S.

NUREG/CR-3145 V05. GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR4946. DAVIS-BESSE UNCERTAINTY STUDY. j WESTERN OHIO-INDIANA REGION Final Report. October 1981 Sep-tsmber 1986. DEMOSSd NUREG/CR4810: EVALUATION OF DIESEL UNAVAILABILITY AND CHUNG.H.M. RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

NUREG/CR-4744 V01 N2: LONG. TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual DENNIO,R.L Report,Apnl. September 1986. NUREG 1275; OPERATING EXPERIENCE FEEDdACK REPORT . NEW PLANTS. Commercial Power Reactors.

NUREG/CR4913: ROUNO ROBIN PRETEST ANALYSES OF A 1.6- DIAMOND,G.L  :

SCALE REINFORCED CONCRETE CONTAINMENT MODEL SUBJECT NUREG/CR 4951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE.

TO STATIC INTERNAL PRESSURIZATION- VERSIBILITY OF RENAL INJURY IN THE RAT. j CLINTON,S.D.

DIDWALLE.M NUREG/CR4817: LODINE PARTITION COEFFICIENT MEASUREMENTS NUREG/CR.4161 V03: CRITICAL PARAMETERS FOR A HIGH LEVEL 4

AT SIMULATED PWR STEAM GENERATOR CONDITIONS.intenm WASTE REPOSITORY. Vobme 3 Salt l Data Report COHEH.L DINGE E.D.A.

NUREG.0837 V07 N0u NRC TLD DIRECT RADIATION MONITORING NUREG/CA-4590 V01: AGING OF NUCLEAR STATION DIESEL  ;

NETWORK. Progress Report, January-March 1987, GENEhATORS Evaluation Of Operating And Expert Expenence. Phase i Study.

CONGEMI,J.

NUREG/CA.2907 V05: RADIOACTIVE MATERIALS RELEASED FROM DOCTOR,8A NUCLEAR POWER PLANTS. Annual Report 1984. NUREG/C114583 V03: DEVELOPMENT AND VALIDATION OF A REAL.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER

~

UREG [4982: SEVERE ACCIDENTS IN SPENY FUEL POOLS IN SUPPORT OF GENERIC SAFETY ISSUE 82.

MEWAN,KE COOK,K.V NUREG 1213 RO1: PLANS AND SCHEDULES FOR IMPLF'AENTATION NUREG/CR.4880: FLAW DENSITY EXAMINATIONS OF A CLAD BOIL.

OF U S NUCLEAR REGULATORY COMMISSION RESPONSIBILITIES ING WATER REACTOR PRESSURE VESSEL SEGMENT' UNDER THE LOW. LEVEL RADIOACTIVE WASTE POLICY AMEND.

COIUH.C. MENTS ACT OF 1985 (P.L.99-240).

NUREG/CR4975: A REVIEW OF THE RESOLVING POWER OF RE.

FLECTION SEISMOLOGY METHODS TO DETECT SUBSURFACE EDSONAL FAULTS AND/OR CHANGES IN LAYER THICKNESS. NUREG/CR-4457: AGING OF CLASS 1E BATTERIES IN SAFETY SYS-TEMS OF NUCLEAR POWER PLANTS.

COSTAIN.J.K.

NUREG/CR4975 A REVIEW OF THE RESOLVING POWER OF RE. EHLERS.E.G.

FLECTION SEISMOLOGY METHODS TO DETECT SUBSURFACE NUREG/CR.4974: INTRAPLATE SEISMICITY IN THE EASTERN FAULTS AND/OR CHANGES IN LAYER THICKNESS. UNITED STATES.

w .. _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ .

f P2rscnil Authsr ind2x 25 l I ERICSON.D.M. GREIMANN.L NUREG/CR 4767: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR4870 AN EVALUATION OF THE EFFECTS OF DESIGN A GENERAL ELECTRIC BWR4/ MARK 1. Case Study. DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT BMINGS.

ERICSON.D.M.

NUREG/CR4710. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF GRIFFITH.P.

A COMBUSTION ENGINEERING PRESSURIZED WATER NUREG/CR 5007: PREDICTION AND MITIGATION OF EROSIVE COR-REACTOR Case Study. ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR EVANS.D.D.

NUREG/CR4654:- RADIONUCLIDES TRANSPORT AS VAPOR QURINSKY,0.H.

THROUGH UNSATURATED FRACTURED ROCK. NUREG/CR4981: A SAFETY ASSESSMENT OF THE USE OF GRAPH-ITE IN NUCLEAR REACTORS LICENSED BY THE U.S. NRC.

l FAILEY.M.P.

NUREG/CR4954 LONG-TERM PERFORMANCE OF SPENT FUEL HAGRMAN.D.L .

l WASTE FORMS- NUREG/CR-4925; FISSION PRODUCT BEHAVIOR DURING THE PBF FANOUS,F. SEVERE FUEL DAMAGE TEST 11.

NUREG/CR-4870: AN EVALUATION OF THE EFFECTS OF DESIGN HALL.T'C DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT NUREG/CR-4583 V03: DEVELOPMENT AND VALIDATION OF A REAL.

BUILDINGS. TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER FISHER.D.R. REACTOR COMPONENTS. Annual Report. October 1985. September NUREG/CR-4884. INTERPRETATION OF BIOASSAY MEASUREMENTS. 1986. j FISK,P.S. HANNINEN.H.E.

NUREG/CM957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR NUREG/CR4863: SURFACE SPECTROSCOPY OF PRESSURE SITE CHARACTERIZATION IN BASALT. SALT AND TUFF. VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN PWR ENVIRONMENTS.

FRENCH.C.S.

NUREG/CR4884. INTERPRETATION OF BIOASSAY MEASUREMENTS. HARDIN.J.E.

NUREG/CR 4457: AGING OF CLASS 1E BATTERIES IN SAFETY SYS-GASSER.R.D. TEMS OF NUCLEAR POWER PLANTS NUREG/CR4896: CONTAINMENT LOADS DUE TO DIRECT CONTAIN.

MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysis HARTWELL.J.K.

l And Calculations With The CONTAIN Code. NUREG/CR 4925: FISSION PRODUCT BEHAVIOR DURING THE PBF i SEVERE FUEL DAMAGE TEST 11.

GELEIN,R.M.

NUREG/CR4951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE- HATCH.S.W.

VERSIBILITY OF RENAL INJURY IN THE RAT. NUREG/CR4767. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A GENERAL ELECTRIC BWR4/ MARK l. Case Study.

NUREG/CR 3024: SUSTAINED CONCRETE ATTACK BY LOW. HELTON.J.C.

TEMPERATURE. FRAGMENTED CORE DEBRIS. NUREG/CR-2452: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL GINZ8URG,T.

NUREG/CR4810: EVALUATION OF DIESEL UNAVAILABILITY AND HENDRICKSON.P.

RISK EFFECTIVE SURVEILLANCE TESilNTERVALS. NUREG/CR 4958: IMPACT OF PROPOSED FINANCIAL ASSURANCE GLOVER,L NUREG/CR4973: INTRAPLATE SEISMICITY OUTSIDE OF THE HESS.C.

UNITED STATES- NUREG/CR-4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATION SYSTEMS GLOZMAN.V. '

NUREG/CR 4900 V01: COMPONENT FRAGILITY RESEARCH HESS.D.J.

PROGRAM. Phase i Demonstrabon Tests Summary Report- NUREG/CR-4956. SYSTEM PERFORMANCE OF HIGH LEVEL WASTE NUREG/CR 4900 V02: COMPONENT FRAGILITY RESEARCH PACKAGE COMPONENTS' PROGRAM. Phase i Demonstrabon Tests. Appendices.

HILL.E.E.

GOMEZ.R.D.

NUREG/CR-4871: RESULTS FROM THE DCH 1 EXPERIMENT, NUREG/CR4910: RELAY CHATTER AND OPERATOR RESPONSE AFTER A LARGE EARTHOUAKE An improved PRA Methodology With GONZALEZ,R.C. Case Studies.

NUREG/CR 4577: AUTOMATED LONG TERM SURVEILLANCE OF A COMMERCIAL NUCLEAR POWER PLANT. HO M I REG' CR 4900 V01: COMPONENT FRAGILITY RESEARCH GORE.B.F. PROGRAM Phase l Demonstration Tests Summary Report.

NUREG/CR-4768 V01: METHODOLOGY AND APPLICATION OF SUR. NUREG/CR-4900 V02: COMPONENT FRAGILITY RESEARCH ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER PROGRAM. Phase i Demonstrabon Tests. Appendices.

PLANT. Task 1. Analysis Of ANO 1 And Oconee PRAs.

NUREG/CR4768 V02: METHODOLOGY AND APPLICATION OF SUR. HOOPINGARNER.K.

ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER NUREG/CR4590 V01: AGING OF NUCLEAR STAT!ON DIESEL ,

PLANT. Final Report. GENERATORS Evaluation Of Operahng And Expert Exponence Phase l 1 Study i GRAVES.N.L NUREG/CR4590 V02: AGING OF NUCLEAR STATION DIESEL i NUREG/CR4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK GENERATORS Evaluehon Of Operating And Expert /

METHODS INTEGRATION AND EVALUATION PROGRAM Exponence Workshop.

(RMIEP) Volume 1: Development Of The Data Based Method.

GREEN.R.T. NUREGICR4768 V02. METHODOLOGY AND APPLICATION OF SUR-NUREG/CR 4654 RADIONUCLIDES TRANSPCQ AS VAPOR ROGAR PLANT PAA ANAltSiS TO TvC MAC:0 $CO OF THROUGH UNSATURATED FRACTURED ROCK, PLANT. Final Report GREGORY.W.S. IMAN.R.L NUREG/CR.4931: RESPONSE OF CENTRIFUGAL AND AXI VANE NUREG/CR 2452: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL BLOWERS TO LARGE PRESSURE TRANSIENTS. OF RADIOACTIVE WASTE. Final Report.

l l

I

]

'26 P;rsonil Author index l ISHil,M. LEVINfsE.N.

NUREG/CR4972: TWO. PHASE FLOW REGIME TRANSITION CRITE- NUREG/CR.495" SURVEY OF GEOPHYSICAL TECHNIQUES FOR l RIA IN POST.ORYOUT REGION BASED ON FLOW VISUALIZATION SITE CHARAC1ERl2ATION IN BASALT, SALT AND TUFF. j EXPERIMENTS. l LEWIS,0. I JANGJ. NUREG 1278: VOGTLE UNIT 1 READINESS REVIEW. Assessment Of NUREG.0837 V07 Noi: NRC TLD DIRECT RADIATION MONITORING George Power Company Readiness Review Pilot Program.

]

NETWORK. Progress Report, January-March 1967.

LIGON.D.M.

JENSEN,$.M.

NUREG/CR4846: HIGH. LEVEL WASTE PRECLOSURE SYSTEMS l MUREG/CR4925: FISSION PRODUCT BEHAVIOR DURING THE PBF SAFETY ANALYSIS. Phase 2, Final Report.

SEVERE FUEL DAMAGE TEST 11.

LIPSZTEtNJ.L NUREG/CR4884: INTERPRETATION OF BCASSAY MEASUREMENTS. i

/CR 884: INTERPRETATION OF BIOASSAY MEASUREMENTS.

' JOHNSTON.J.W.

LITTLEFIELD).

NUREG/CR.5002: METHODS FOR RECURRING LOSS TEGTS. NUREG/CR-5002: METHODS FOR RECURRING LOSS TESTS.

JONES,0.M. LOFGREN,E.

NUREG/CR4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR NUREG/CR4810: EVALUATION OF DIESEL UNAVAILABILITY AND SITE CHARACTER!ZATION IN BASALT, SALT AND TUFF. RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

MAPLAN,L LONGSINE,0.E.

NUREG/CR4981: A SAFETY ASSESSMENT OF THE USE OF GRAPH. NUREG/CR.2452: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL ITE IN NUCLEAR REACTORS LICENSED BY THE U.S. NRC. OF RADIOACTIVE WASTE. Final Report.

KECK,R.G. LOOMIS,G.G.

NUREG/CR.5007; PREDICTON AND MlTIGATION OF EROSIVE-COR- NUREG/CR.4945:

SUMMARY

OF THE SEMISCALE PROGRAM (1965-ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR 1986).

POWER PLANTS.

LOSS,F.J.

KEMPF.C.R. NUREG/CR.3228 V05: STRUCTURAL INTEGRITY OF WATER REAC-NUREG/CR4897: LOW LEVEL WASTE SOURCE TERM TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For EVALUATION. Review of PubHshed Modehng And Expenmental 198e, i Work,And Presentation Of Low Level Waste Source Term Modehng i Framework And Prelmnary Model Development. LUCERO,0.A.

KENCYER,J.L NUREG/CR4485 V02: THE IMPACT OF FUEL CLADDING FAILURE LUzA,g y, EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE

• NUREG/CR-4937; INVESTIGATION OF THE MEERS FAULT IN SOUTH- 1 AR POWER PLANTS. Case Stuoy:PWR Dunng An Outage. WESTERN OKLAHOMA.

KEILEY,T.E MACDONALD P.E NUREG/CR4871: RESULTS FROM THE DC,41 EXPERIMENT.

NUREG/CR4731 V01: RESIDUAL LIFE ASSESSMENT OF MMOR KIMSRELL.A.F. LIGHT WATER REACTOR COMPONENTS OVERVIEW Volume 1.

NUREG/CR4919: FIELD TESTING OF BENTONITE AND CEMENT "#

BOREHOLE PLUGS IN GRANITE.

N EG CR4937: INVESTIGATION OF THE MEERS FAULT IN SOUTH.

KINNISON,R.R. WESTERN OKLAHOMA.

NUREG/CR 5002: METHODS FOR RECURRING LOSS TESTS.

MAERKER,R.E.

KOENIG.LN. NUREG/CR.4886: ANALYSIS OF THE NESOIP2 AND NESOIP3 RADIAL NUREG/CR-4216: EXPERIMENTAL RESULTS FOR A 1:8-SCALE STEEL SHIELD AND CAVITY EXPEREENTS.

MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED TO FAILURE. MARKWORTH,A.J.

NUREG/CR4954: LONG-TEFiM PERFORMANCE OF SPENT FUEL KOHLI,R. WASTE FORMS.

NUREG/CR-4954: LONG TERM PERFORMANCE OF SPENT FUEL NUREG/CR4956: SYSTEM PERFORMANCE OF H:GH-LEVEL WASTE i WASTE FORMS. PACKAGE COMPONENTS. l l

KURTH,R.E MART,0.A. l NUREG/CR4956: SYSTEM PERFORMANCE OF HIGH-LEVEL WASTE NUREG/CR-4583 V03 DEVELOPMENT AND VALIDATON OF A REAL-PACKAGE COMPONENTS. TIME SAFT-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER

^ "" '

LA80NE.T.R. 986' NUREG/CR4884. INTERPRETATION OF BCASSAY MEASUREMENTS.

MARTIN,G.F.

LAMBERT,H.E NUREG/CR4485 V02: THE IMPACT OF FUCL CLAODING FAILURE NUREG/CRb910: RELAY CHATTER AND OPERATOR RESPONSE AFTER A LARGE EARTHQUAKE.An improved PRA Methodology With EVENTS ON OCCUPATIONAL RADIATON EXPOSURES AT NUCLE-AR POWER PLANTS Case Study PWR Dunng An Outage.

Case Studies.

LAY,T. MATTHEWS,$.D.

NUREG/CR4844 DAFT; INTEGRATED RELIABILITY AND RISK ANALY.

NUREG/CR-3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO-INDIANA REGIONMnel Report. October 1981 Sep. SIS SYSTEM (IRRAS) USER'S GUIDE . VERSION 1.0 (DRAFT).

MCCLUNG,R.W.

LEE.K. NUREG/CR4860: FLAW DENSITY EXAM NATONS OF A CLAD BOIL.

NUREG/CR4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATON ING WATER REACTOR PRESSURE VESSEL SEGMENT.

SYSTEMS.

WCCOYJ.K.

LESSARD.E.T. NUREG/CA.4954: LONG-TERM PERFORMANCE OF SPENT FUEL NUREG/CR.4884: INTERPRETATION OF BCASSAY MEASUREMENTS. WASTE FORMS. l I

Pircend Auth r Ind2x ' 27 O'REILLY,P.C.

MCNAMARA,M.

NUREG-0837 V07 NO1: NRC TLD DIRECT RADIATION MONITORING NUREG.1275: OPERATING EXPERIENCE FEEDBACK REPORT NEW PLANTS. Commercial Power Reactors.

NETWORK. Progress Reinrt, January-March 1987 ,

MEALE.B.M.

0SDT,N.T.

NUREG/CR4747 V01: AN AGING FAILURE SURVEY OF LIGHT NUREG/CR-4972: TWO. PHASE FLOW REGIME TRANSITION CRITE-WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. RIA IN POST.DRYOUT REGION BASED ON FLOW VISUALIZATION EXPERIMENTS.

MEANS,J.L NUREG/CRJ954: LONG-TERM PERFORMANCE OF SPENT FUEL OKAILY,A.A.

WASTE FORMS, NUREG/CR-4952: EXPERIMENTAL STUDY OF FILLET WELD UNDER. ','

MEININGER,R.D. CUT EFFECTS ON WELDED TUBING STRUCTURES UNDER CEN.

TRIC AND ECCENTRIC CYCLIC LOADINGS.

NUREG/CR-5008: DEVELOPMENT OF A TESTING AND ANALYSIS METHODOLOGY TO DETERMINE THE FUNCTIONAL CONDITION OF OLIVER.M.S.

SOLENOID OPERATED VALVES. ,

NUREG/CR4871: RESULTS FROM THE DCH.1 EXPERIMENT.

MERKLE.J.G.

NUREG/CR4760: TEST OF 6IN THICK PRESSURE VESSELS.Senes ORTIZ,N.R.

3: Intermediate Test Vessel V-8A.Teanng Behaver Of Low Upper. Shelf NUREG/CR.2452: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL Matenal. OF RADIOACTIVE WASTE. Final Report.

MEWHINNEY,J.A.

DES R NUREb C 4951: NEPHROTOXICITY OF URANYL FLOORIDE AND RG.

UAT N FOR INHALE AIR ORNE R ION VERSIBILITY OF RENAL INJURY IN THE RAT.

ll I

MEYER.R.E. PERKINS,K.R.

NUREG/CR4865: THE SOLUBILITY OF ELECTRODEPOSITED TC(IV)NUREG/CR4982: SEVERE ACCIDENTS IN SPENT FUEL POOW /N i OXIDES. SUPPORT OF GENERIC SAFETY ISSUE 82. .

MILLER,0.

NUREG/CR4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATIONPETTI.D.A.

SYSTEMS. NUREG/CR4B76: SILVER-INDIUM.CAEMlVM CONTROL ROO BEHAV.

IOR AND AEROSOL FORMATION IN SEVERE REACTOR ACCI. ;

MOELLER.M.P. DENTS.

NUREG/CR4465 V02: THE IMPACT OF FUEL CLADDING FAILURE NUREG/CR4925: FISSION PRODUCT BEHAVIOR DURING TNC rT85 fs EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE. SEVERE FUEL DAMAGE TEST 11.

AR POWER PLANTS. Case Study:PWR Dunng An Outa9e.

PILCH,M.

MOORE.S.E. j NUREG/CR4871: RESULTS FROM THE DCH.1 EXPERIMEf R NUREG/CR4943: PREPARATION OF DESIGN SPECIFICATIONS AND '

DESIGN REPOHTS FOR PUMPS, VALVES. PIPING.AND PIPING SUP- j PITTIGILO,C.L  :

PORTS USED IN SAFETY-RELATED PORTIONS OF NUCLEAR NUREG 1274: REVIEW PROCESS FOR LOW. LEVEL RADIOACTIVE POWER PLANTS. WASTE DISPOSAL LICENSE APPLICATION UNDER LOW-LEVEL RA G/C 951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE.

POLLACK,H.N.

VERSIBILITY OF RENAL INJURY IN THE RAT.

NUREG/CR.3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE MOSES,M.J WESTERN OHIO-INDIANA REGION. Final Report. October 1381. Sep-NUREG/CR-4974. INTRAPLATE SEISMICITY 1N THE EASTERN tomber 1986.  ;

s UNITED STATES.

PRATER.J.T.

MOSLEH.A. . NUREG/CR4891: PROPERTIES OF REACTOR FUEL ROD MATERIA NUREG/CR4962: METHODS FOR THE ELICITATION AND USE OF AT HIGH TEMPERATURES Fanst Summary Report Severo Core EXPERT OPINION IN RISK ASSESSMENT. Phase I A Crttical Evalue. Damage Property Tests Program.

ton And Directons For Future Research.

PUGH CL MULLEN.M.F.

NUREG/CR4958 IMPACT OF PROPOSED FINANCIAL ASSURANCE NUREG/CR4219 V04 N1: HEAVY.SECTION STEEL TECHNOLOGY PROGRAM. Semiannual Progress RepMt For Octchor 1966 - March REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES.

1967.

MURPHY,V.J.

NUREG/CR.4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR OUAYLE.,$.F.

SITE CHARACTERIZATION IN BASALT. SALT AND TUFF. NUREG/CR-4956: SYSTEM PERFORMANCE OF HIGH. LEVEL WAS I PACKAGE COMPONENTS.

NANSTAD,R.K.

NUREG/CR4760. TEST OF 6 IN . THICK PRESSURE VESSELS $enes PEE f

3 tnt iate Test Vessel V.8A.Teanng Behaver Of Low Upper-Shelf URE /CR4418' DOSE CALCULATION FOR CONTAMINATION OF THE SKIN USING THE COMPUTER CODE VARSKIN.

NESSITT J.F. REID.LD.

NUREG/CR-4590 V01: AGING OF NUCLEAR STATION DIESEL GENERATORS Evoluaton Of Operating And Expert Exponence Phase NUREG/CR-4583 V03' DEVELOPMENT AND VALIDATION OF I Study.

TIME SAFT UT SYSTEM FOR THE INEDECTION ~

OF UGHT WA

^"""*'

NICHOLLS.A.K.

fg ~

NUREG/CR.4958. IMPACT OF PROPOSED FINANCIAL ASSURANCE RHEE,0.S.

REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES NUREG 1284 PROGRAM PLAN FOR CORRECTION OF U S. IN NICOLOSI,S.L MENT DEGRADATION OR FAILURE IN THE UPPER PLENUM; NUREG/CR4956' SYSTEM PERFORMANCE OF HIGH. LEVEL WASTE FACILITY (UPTF) IN THE FEDERAL REPUBUC OF GERMANY. i PACKAGE COMPONENTS RICE,J.E.

NORDEN,K.

NUREG/CR 2907 V05 RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR4957. SURVEY OF GEOPHYSICAL TECHNIOUES SITE CHARACTERIZATION IN BASALT, SALT AND TUFF.

NUCLEAR POWER PLANTS Annual Report 1964

y. ,

, y > '

, , > - 9 ;f'}!

28 Perstnal Author inrcx '

. . . '\

' t

. cosmSoN,0.C. L ) ) '

SEEBERL ' j. t NbREG/CR-4760: TEST OF SIN. THICK PPES@RE VESSELS $enes 3 tntoime6 ate Test vessel V.8A.Teenng dat vor Of Low UprGSheH NUREG/CR.4851: SSISMCTV 164ts9 ' IN THE SOUTHEASTERN Matonal

• UNITED STATES.The Aftershock Sequence Of The Gariesten. South Carohna Earthquake.

. , RODASADGH,E.C. 'i .

SETH.EL I WJREG/CR4943 PREPARArith CF DESIGN SPECIFICATIONS ANO. NUREG/CR4953: CORRELATON OF RADIOIODINE RESUSPENSON

? [)

f,.

DESIGN REPORTS FOR PUWS.VALVKS, PIP;NG.AND PIPING SUPE .'

PORTS USED IN PAFETY POWER PLANTS. ' I P.

RELATED NRTICNS OF NUCLEAR

) ;1

, NTH TEMPERATURE AT tmh 2.

l

' $HAASAN.H1 ROGERSJ.

% t

/

I

/ NUREG 1194: C&STRUCTIOA APPRAISAL TEAM INSPECTON RE-g l

JUREG/CR4870: AN EVALUATION OF THE &FECTS OF DESIGN SULTS ON MMIN3 AND NONDESTRUCTIVE EXAMINATION AC.

l TIVITIES. / i

.f- OETAIL P ON THE CAPACITY OF LWR STEEL CONTAINMEW.

BUILDNGS. SHAH,VM. '

- t 5 'i / \,3 i-NUREGICA-4731 V01: RESOUAL UFE ASSESSMENT OF MAJOR CR3024: SU ANED CONCPFTE fTACK BY LOW- -

i ,' / EAPERATURE. FRAGMENTED CORE DEORIS. '

/ NOHEG/CR4871: PESULTS FROM THE DCH-1 EXPERIMENT.

r SHEPHARD.J.E. k i t

NUREG/CR4534: ANALYSIS OF OlFFU@ FLAME TESTS.

WM5 SHERMAH.M.P.

()

NUREG4tN Vit N01: UCEN3ED OPERAT:NG REACTOPS STARJS SUMMAR K REPORTDta As Of December 31.1986 (Gray Book f) NUREG/CR.4905: DETONABluTV OF H2-AIR.DILUEN1 MIXTUFiES.

m

[ SHIPWA Y,G.O.

A NUREG/CR4900 ' v0 f: COWONU w 2452: RISr METHODOLOGY FOR GEO bGIC DISPOSAL FRAGfuFr RESEARCH OF RADOACTIVE WAJTE. Final pW PROGRAM Pnase i Demonstroton Teun Suminary Report.

s RDSSELULD. .

NUREG/CR 4900 V02: COMPONENT FRAGIUTY RESEARCH

/. PROGRAM. Phase i Demonstration Tests. Appendices.

NUREG/CR 4844 OR/* INTEGRATED RELIABILtW 40 AISK Af4ALY-SIS SYSTEM (IRRAS) USER'S GUCE VF.RSIOth ! 0 (DRAFT). bH SAILOR',y.L, I N REG /CR 2: RISK pf.THODOLOGY l GblOGO OISPOSAL

. N / OF RADOACTIVE WAST ( Final Report. ',

NUREG/CR4982: SEWRE ACCIDENTS IN SPENT FUCv POOLS IN k g SUPPORT OF OENERO SAFETY ISSUE 82.

'F / f.i F

ti4AaANTA.P. P 4: F*ROGRAM PLAP POR C ECTON OF .MTRU-

, i MENT DEGR ADATION OR FAILURE IN WE UPPER PENUM TUT

< I / NUREG/CR-4810: EVALUATON OF DIESEL NNf XURIUTY AND FACILITY (UPTF)IN THE FEDERAL REPWS.'C OF GERMANY.

) RISK EFFECTIVE SURVEIU,ANCE W WTEF OLS. ' '

($ i )

SIGALLA.L.A. is l:

SAN D;RS,G.A.

) '

NUREG/CA.4418: DOSE CALCULATION (JOR CONTAM

(

l .>

NU4EG/CRJ710: SHUTDOWN DECAY HEAT MMOVAL ANALYSIS 7:

A COMBUSTION ENGINEERING PRfhSURIZED REACTOR Case Study.

WK W THE SKIN USING THE COMPUTEP CO7s WASKIN.

\, j

! , SIMMONS,C.M.

NUREG/CR4767: SHUTOOWN DECAY HEAT RFMNAL ANALYSIS C,5 (

I\

NUREG/CR 4817. IODINE FARTITG CONCENT MEASMEMENTS A GENERAL ELECTRIC BWR4/ MARK LCase Stdr$ "p '/ AT SIMULATED PWR STEAM GENERATOR CONDITIC.NS. Inter.m SASTRE.C.

NUREG/CR4981: A SA/67Y ASSESSMENT OF TH USE Of GRAPH-ITE IN NUCLEAR REAf?OR$ UCLNSED BY TyS L .NRC.

SKRABLE.K.W. .#

NUREG/CR4884. INTERPRETATION OF BIOASSAY MEASUREMENTS.

SATTERWHITE.D.  ;/ SMITH,C.M. I NUREG/CR4747 V01: AN AGING FAILURE SU/EY OF UGHT WATER REACTOR SAFtTV SYSTEMT, AND COA FONENTS. NUREG/CR-4577: AUTOMATED LONG-TE M SURVEILLANCE OF A k'  ! COMMERCIAL NUCLEAR POWER PLANT SATTISON,M.B.

i l

SMITH,P.R.

NUREG/CR4844 DAFT: INTk iRATED AEUABluTh AND RISK ANALY. . .

,_ p SYSTEM (IRRAE USEHs GUIDE VERSION 1 'ORIFT). NUREG/CR-4931: RESPONSE OF CEV : FUGAL AND AXI-VANE BLOWERS TO LARGE PRESSURE TR/mENTS.

SCHENR2.RJ. /

NURJJ/CRs418: DOSn CALCULATION FOR CONTAMINATION OF THE SKIN USING THLf COMPUTER CODE VARSKIN. t SMITH NUREG S.A[CR4958: IMPACT OF PROPOSED FINANDAL ASSURANCE REQUIREMENTS ON NUCLEAR MATERIALS UCEUSEES SCHWARTZ,S.Y. t NUREG/CR 3145 1/05: GEOPHYSICAL INVESTIGANONS ,OF THli SNIDER.D.M.

WESTERN OHC-lW,MNA REGON Final Repo1,0ctotnGm1 Sep NUREG/CR-4844 DAFT: INTEGRATED RF.UABluTY AND RISK ANALY.

temtwr1 W8. 3 SIS SYSTEM (IRRAS) USER'S GUlOE a VERSION 00 (DRAFT).

SCHWEITZER,0.Q. STAMATELATOS,M. i l 5 NUREG/CR4981: A SAFETY ASSESSMENT OF THE USE OF GRAPH- NUREG/CR4848: HIGH LEVEL WASTE MRECLO:.h.RE SYSTEMS ITE IN NUCLEAR REACTORS UCENSED BY THE V.S NRC. SAFETY ANALYSIS Phase 2. Fina: Report.

SCHWEITZER,J. STENHOUSE.M.J. O NUREG/CR 4973- INTRAPLATE SEISMICITY OUTSOE OF THE NUREG/CR4954: LONG.T ERM PFRFORMANC1 OF SPENT FUEL UNU TO SDTES.

7

' WASTE FORMS.

SCHWENA.E.3. STEVERSONJ.A.

'y NUREQCR4850 STEAM GENERATOR GROUP PROJECT.Ta6k 10 NUREG/.%4407; PIPE BREAK FREOVENCY ESTIMArlON FOR NU-Sec#ary S4e Exawnshon. Final Report. CLEAR POWER PLANIS

, SCOTT.M.J. $TEWART,H.D.

NUREG/CR4958 WPACT OF PROPOSED FINANCIAL ASSURANCE NUREG/CRJ844 DRFT. INTEGRATED REUABlWTY A hC RISM ANALY.

REQU(REMENTS ON HUCLEAR MATERIALS UCENSEES SIS SYSTEM (IRRAS) USER'S GUOE VERSON 10 (O/ AFT).

3p

l Parsonal Author Ind3x 29 ftULLIVAN,T, WAGNER,K.L NUREG/CR4897: LOW LEVEL WASTE SOURCE TERM NUREGICR.4844 DRFT: INTEGRATED RELIABlWTY AND RISK ANALY.

EVALUATION. Review of Published Modeling And Expenmental SIS SYSTEM (IRRAS) USER'S GUIDE VERSION 1.0 (DRAFT).

b Work.And Presentation Of Low-Level Weste Source Term Modeling Framework And Preliminary Model Development. WASHINGTON .

TAR 8ELLW.W. MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR.AnaYsis And Calculations With The CONTAIN Code.

NUREG/CR.3024. SUSTAINED CONCRETE ATTACK BY LOW.

TEMPERATURE. FRAGMENTED CORE DEBRIS WEEKS, .R NUREG/CR-4871:RESULTS FROM THE DCH-1 EXPERIMENT.

TAYLOR,G.C. SUPPORT OF GENERIC SAFETY ISSUE 82.

NUREG/CR 4954 LONG-TERM PERFORMANCE OF SPENT FUEL

^ '

U EG'/CR 5008. DEVELOPMENT OF A TESTING AND ANALYSIS THEOFANOUS,T.G. METHODOLOGY TO DETERMINE THE FUNCTIONAL CONDITION OF SOLENOID OPERATED VALVES.

NUREG/CR4978 THE COOLDOWN ASPECTS OF THE TML2 ACCI.

DENT. WEISS.A.J.

NL9EG/CR 2331 V06 N4: SAFETY RESEARCH PROGRAMS SPON-TICHLER,J. OF NUCLEAR REGULATORY SORLD BY OFFICE NUREG/CR.2907 V05- RADIOACTIVE MATERIALS RELEASED FROM RESEARCH Ouarterty Progress Report October December 1986.

NUCLEAR POWER PLANTS Annual Report 1984.

WESTON,LM.

TIESZEN.S.R NUREG/CP4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK NUREG/CR4905. DETONABILITY OF H2 AIR-DILUENT MIXTURES. METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Volume 1. Development Of The Data-Based Method.

TILLS,J.L NUREG/CR4896 CONTA.NMENT LOADS DUE TO DIPECT CONTAIN- WHITEHEAD.D W MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR.Anaysis NUREG/CR.48$4 V01: RECOVERY ACTIONS IN PRA FOR THE RISK And Calculations With The CONTAIN Code METHODS INTEGRATION AND EVALUATION PROGRAM (RMlEP). Volume 1: Development Of The Data-Based Method.

TRAUB,R.J.

NUREG/CR4418: DOSE CALCULATION FOR CONTAMINATION OF P THE SKIN USING THE COMPUTER CODE VARSKIN WIL8 NUREG/CURN.N'R4640; HANDBOOK OF SOFTWARE QUAUTY ASSURANCE TECHNIQUES APPLICABLE TO THE NUCLEAR INDUSTRY.

TSAO.L NUREG/CR4161 V03. CRITICAL PARAMETERS FOR A HIGH LEVEL WILLIAMS D.C.

WASTE REPOSITORY. Volume 3. Salt. NUREG/CR4898: CONTAINMENT LOADS DUE TO DIRECT CONTAIN-MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysis VAUSE.J.W. And Calculations With The CONTAIN Code NUREG/CR4590 V01: AGING OF NUCLEAR STATION DIESEL GENERATORS Evaluation Of Operating And Expert Expenence Phase WOLLENBERG.H.A.

N G CR-4590 V02, AGING OF NUCLEAR STATION DIESEL STE R SlT Y 3 Sa GENERATORS Evaluation Of Operating And Expert Exponence Workshop. WOOD,R.S.

NUREG-0327 R04- OWNERS OF NUCLEAR POWER NURE Fl4769 RISK EVALUATIONS OF AGING PHENOMENAThe BU y n?es.

Linear Aging Reliability Model And its Extensions NUREG/CR-4810: EVALUATION OF DIESEL UNAVAILABILITY AND WRIGHT,R.E.

RISK EFFECTIVE SURVEILLANCE TEST INTERVALS- NUREG/CR4407; PlPE BREAK FREQUENCY ESTIMATION FOR NU.

CLEAR POWER PLANTS.

VIRGIUO M.J.

NUREG 1285: NRC STAFF EVALUATION OF THE GENERAL ELECTRIC WU,P.C.

COMPANY NUCLEAR REACTOR STUDY (" REED REPORT") NUREG 1194. CONSTRUCTION APPRAISAL TEAM INSPECTION RE.

SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC.

VOR A.J.P. TlVITIES NUREG 1144 R01: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-GRAM PLAN YlHUA,X.

NUREG/CRABB4 INTERPRETATION OF BIOASSAY MEASUREMENTS.

NURE G/CR-4863 SURFACE SPECTROSCOPY OF PRESSURE ZUROFF,W.F.

VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN NUREG/CR4407: PIPE BREAK FREQUENCY ESTIMATION FOR NU-PWR ENVIRONMENTS CLEAR POWER PLANTS.

l 1

l l

Subject index 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.

i te Scale NUREG/CR4769: AISK EVALUATIONS OF AGING PHENOMENA.The NUREG/CR4913. ROUND ROB!N PRETEST ANALYSES OF A 1:6- Lineer Agmg Rehabihty Model And its Extensions.

SCALE REINFORCED CONCRETE CONTAINMENT MODEL SUBJECT TO STATIC INTERNAL PRESSURIZATION. Altt>orne Rad 6onuclide NUREG/CR 4986: RADIATION DOSE ESTIMATES AND HAZARD EVAL.

UATIONG FOR INHALED AIRBORNE RADIONUCLIDES Final Report N G/CR4216: EXPERIMENTAL RESULTS FOR A 1:8-SCALE STEEL MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED Animale FA W E. NUREG/CR4986: RADIATION DOSE ESTIMATES AND HAZA1D EVAL-i UATIONS FOR INHALED AIRBORNE RADIONUCLIDES Final Report ASME Code NUREG/CR 4943: PREPARATION OF DESIGN SPECIFICATIONS AND DESIGN REPORTS FOR PUMPS. VALVES. PIPING,AND PIPING SUP. Ad Vane Blower PORTS USED IN SAFETY-RELATED PORTIONS OF NUCLEAR NUREG/CR 4931: RESPONSE OF CEN~tlFUGAL AND AXIVANE POWER PLANTS. BLOWERS TO LARGE PRESSURE TRANSIENTS.

Abnormal Occurrences BWR NUREG 0000 V09 N04: REPORT TO CONGRESS ON ABNORMAL NUREG/CR-4707: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF OCCURRENCES. October December 1986. A GENERAL ELECTRIC BWR4/ MARK l. Case Study.

NUREG/CR4860: FLAW DENSITY EXAMINATIONS OF A CLAD BOIL-ING WATER REACTOR PRESSURE VESSEL SEGMENT.

N R G4304 V12 NO2: REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL) Compilation For second Quarter ,,,n 1987,Apni June.

NUREG/CR4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR Acc6 dent SITE CHARACTERIZATION IN BASALT SALT AND TUFF, NUREG 1251 DRFT FC: IMPLICATIONS OF THE ACCIDENT AT CHER.

NOBYL FOR SAFETY REGULATION OF COMMERCIAL NUCLEAR Batterlee POWER PLANTS IN THE UNITED STATES Draft For Comment. NUREG/CR-4457: AGING OF CLASS 1E BATTERIES IN SAFETY SYS-NUREG/CR4485 V02: THE IMPACT OF FUEL CLADDING FAILURE TEMS OF NUCLEAR POWER PLANTS.

EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE-AA POWER PLANTS Case Study PWR Dunno An Outage, Bentonite NUREG/CR 4978. THE COOLDOWN ASPECTS OF THE TMI-2 ACCl- NUREG/CR 4919. FIELD TESTING OF BENTONITE AND CEMENT DENT. BOREHOLE PLUGS IN GRANITE.

Accident Condit6ons RE Tana 98 U G/CR 88 ERPRETATION OF BIOASSAY MEASUREMENTS.

Acc6 dental Release Bolling Water Reactor NUREG/CR4958. IMPACT OF PROPOSED FINANCIAL ASSURANCE NUREG/CR 4767: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES A GENERAL ELECTRIC BWR4/ MARK 1. Case Study.

NUREG/CR4860. FLAW DENSfTY EXAMINATIONS OF A CLAD BOIL-G/CR4876- SILVER INDIUM-CADDUM CONTROL ROD BEHAV.

IOR AND AEROSOL FORMATION IN SEVERE REACTOR ACCl- Borehole Plug NUREG/CR4919 FIELD TESTING OF BENTONITE AND CEMENT NURE /CR4986 RADIATION DOSE ESTIMATES AND HAZARD EVAL.

UATIONS FOR INHALED AIRBORNE RADIONUCLIDES Final Report. BOREHOLE PLUGS IN GRANITE.

Aftershock Sequence Buckling NUREG/CR4851: SEISMICITY 1886-89 IN THE SOUTHEASTERN NUREG/CR48a2: A STUDY OF THE EFFECTS OF PENETRATION UNITED STATES.The Aftershock Sequence Of The Charleston South FRAMING ON STEEL CONT AINMENT BUCKLING CAPACITY.

Carolina Earthquake. NUREG/CR 4904 INVESTIGATION OF STEEL CONTAINMENT BUCK-LING FROM DYNAMIC LOADS.

Aging ,

NUREG 1144 R01: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO- Burning GRAM PLAN NUREG/CR-4981: A SAFETY ASSESSMENT OF THE USE OF GRAPH-NUREG/CR4457. AGING OF CLASS 1E BATTERIES IN SAFETY SYS- ITE IN NUCLEAR REACTORS LICENSED BY THE U.S NRC.

TEMS OF NUCLEAR POWER PLANTS.

NUREG/CR-4590 V01: AGING OF NUCLEAR STATION DIESEL CONTAIN ENERATORS Evaluation Of Operating And Expert Expenence Phase NUREG/CR4896 CONTAINMENT LOADS DUE TO DIRECT CONTAIN.

MENT HEATING AND ASSOCIATED MDROGEN BEHAVIOR. Analysis NUREGfCR 4590 V02- AGING OF NUCLEAR STATION DIESEL Of Operstmg And Expert And Calculations with The CONTAIN Code.

GENERATORS Evaluation Exponence Workshop NUREG/CR 4731 Vot: RES: DUAL LIFE ASSESSMENT OF MAJOR Capacity LIGHT WATER REACTOR COMPONENTS OVERVIEW Volume 1 NUREG/CR 4870 AN EVALUATION OF THE EFFECTS OF DESIGN NUREG/CR4747 V01. AN AGING FAILURE SURVEY OF LIGHT DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. BUILDINGS 31

32 Subjset Indax Cart >on Steel Control Rod Behavior l NUREG/CR-5007: PREDICTION AND MITIGATION OF EROSIVE COR- NUREG/CR4876: SILVER INDIUM CADMlUM CONTROL ROD BEHAV- l ROSIVE WEAR IN SECONDARY PIPING SYSTN OF NUCLEAR IOR AND AEROSOL FORMATION IN SEVERE REACTOR ACCT-POWER PLANTS. DENTS.

Cast Stainlese Steel Cooldown NUREG/CR-4744 V01 N2: LONG TERM EMBRt*7dMENT OF CAST NUREG/CR4978: THE COOLDOWN ' ASPECTS OF THE TMi-2 ACCl-DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual DENT.

Report.Apnt September 1986.

Core Damage Cement NUREG/CR-4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK NUREG/CR 49tp: FIELD TESTING OF BENTONITE AND CEMENT METHODS INTEGRATION AND EVALUATION PROGRAM BOREHOLE PLUGS IN GRANITE. (RMIEP) Volume 1. Development Of The Data Based Method.

NUREG/CR4891: PROPERTIES OF AEACTOR FUEL ROD MATERIALS AT HIGH TEMPERATURES Final Summary Report Severo Core NUR /CR 4 3 : RESPONSE OF CENTRIFUGAL AND AXI. VANE BLOWERS TO LARGE PRESSURE TRANSIENTS. O'**9' P'@ " **

Chernoby1 Core We NUREG 1251 DAFT FC: IMPLICATIONS OF THE ACCIDENT AT CHER- NUREG/CR-3024: SUSTAINED CONCRETE ATTACK BY LOW-NOBYL FOR SAFETY REGULATION OF COMMERCIAL NUCLEAR POWER PLANTS IN THE UNITED STATES. Draft For Comment.

NU EG/CR 71 LTS F M H 1 EXPERIMENT.

N 1E Core-Concrete interaction NUREG/CR4457: AGING OF CLASS 1E BATTERIES IN SAFETY SYS- NUREG/CR 3024: SUSTAINED CONCRETE ATTACK BY LOW-TEMS OF NUCLEAR POWER PLANTS TEMPERATURE. FRAGMENTED CORE DEBRIS.

NUREG/CR 4747 V01: AN AGING FAILURE SURVEY OF LIGHT WATER REACTOR SAFETY SYSTEMS AND COMPONENTS.

C "' '

E 3 V S EVALUATION PEPORT ON TENNESSEE Combust 6cn VALLEY AUTHORITY. Revised Corporate Nuclear Performance Plan.

NUREG/CR-4534. ANALYSIS OF DIFFUSION FLAME TESTS.

• Component NUREG/CR-4863- SURFACE SPECTROSCOPY OF PRESSURE NUREG/CR 3228 V05. STRUCTUPAL INTEGRITY OF WATER REAC- VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN TOR PRESSURE BOUNDARY COMPONENTS Annual Report For PWR ENVIRONMENTS.

1986.

NUREG/CR4731 V01: RESIDUAL LIFE ASSESSMENT OF MAJOR Cyclic Loading LIGHT WATER REACTOR COMPONENTS OVERVIEW. Volume 1. NUREG/CR4952: EXPERIMENTAL STUDY OF FILLET WELD UNDER-NUREG/CR 4900 V01: COMPONENT FRAGILITY RESEARCH CUT EFFECTS ON WELDED TUBING STRUCTURES UNDER CEN-PROGRAM Phase i Demonstration Tests Summary Report. TRIC AND ECCENTRIC CYCLIC LOADINGS NUREG/CR4900 V02: COMPONENT FRAGILITY RESEARCH PROGRAM Phase i Demonstration Tests. Appendices. DCH 1 Experiment NUREG/CR 4956: SYSTEM PERFORMANCE OF HIGH-LEVEL WASTE NUREG/CR4871: RESULTS FROM THE DCH-1 EXPERIMENT.

PACKAGE COMPONENTS.

Decay Heat Removal UR /CR4962: METHODS FOR THE ELICITATION AND USE OF EXPERT OPINION IN RISK ASSESSMENT. Phase 1 A Cntical Evalua* REACTOR Ca t dv tion And Directons For Future Research. NUREG/CR4767 SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A GENERAL ELECTRIC BWR4/ MARK 1. Case Study.

Concrete NUREG/CR-3024. SUSTAINED CONCRETE ATTACK BY LOW- "

TEMPERATURE. FRAGMENTED CORE DEBRIS NUREG 1264 PROGRAM PLAN FOR CORRECTION OF U S. INSTRU-NUREG/CR-4913. ROUND ROBIN PRETEST ANALYSES OF A 16- MENT DEGRADATION OR FAILURE IN THE UPPER PLENUM TEST SCALE REINFORCED CONCRETE CONTAINMENT MODEL SUBJECT FACILITY (UPTF) IN THE FEDERAL REPUBLIC OF GERMANY.

TO STATIC INTERNAL PRESSURIZATION' NUREG/CR4590 Vot: AGING OF NUCLEAR STATION DIESEL Construction Appraisal Team GENERATORS Evaluation of Operating And Expert Expenence Phase I Study.

NUREG 1194 CONSTRUCTION APPRAISAL TEAM INSPECTION RE.

SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC, NUREG/CR4500 V02: AGING OF NUCLEAR STATION DIESEL GENERATORS.Evaluat on Of Operating And Expert TIVITIES.

Expenence Workshop.

Containment NUREG/CR4731 Vot RESIDUAL LIFE ASSESSMENT OF MAJOR NUREG/CR-4216 EXPERIMENTAL RESULTS FOR A 16-SCALE STEEL LIGHT WATER REACTOR COMPONENTS OVERVIEW Volume 1.

MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED Design

• 0 F AILURE_

NUREG/CR4870- AN EVALUATION OF THE EFFECTS OF DESIGN NUREG/CR-4870: AN EVALUATION OF THE EFFECTS OF DESIGN

• DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT BUILDINGS. BUP DINGS NUREG/CR-4892 A STUDY OF THE EFFECTS OF PENETRATION NUREG/CR-4943 PREPARATION OF DESIGN SPECIFICATIONS AND FRAMING ON STEEL CONTAINMENT BUCKLING CAPACITY. DESIGN REPORTS FOR PUMPS. VALVES. PIPING.AND PIPING SUP-NUREG/CR4904 INVESTIGATION OF STEEL CONTAINMENT BUCK. PORTS USED IN SAFETY RELATED PORTIONS OF NUCLEAR LING FROM DYNAMIC LOADS. POWER PLANTS.

NUREG/CR 4913. ROUND. ROBIN PRETEST ANALYSES OF A 16 SCALE REINFORCED CONCRElt: CONTAINMENT MODEL SUBJECT Detonation Cell TO STATIC INTERNAL PRESSURIZATION NUREG/CR4905: DETONABILITY OF H2-AIR-DILUENT MIXTURES.

Containment integrtty Diesel NUREG-1264. CONTAINMENT INTEGRITY RESEARCH PROGRAM NUREG/CR4810 EVALUATION OF DIESEL UNAVAILABILITY AND PLAN. RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

Containment Load Diesel Generator NUREG/CR-4896 CONTAINMENT LOADS DUE TO DIRECT CONTAIN. NUREG/CR4590 V01: AGING OF NUCLEAR STATION DIESEL MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR.Anatyses GENERATORS Evaluaton Of Operating And Expert Expenence Phase And Calculations With The CONTAIN Code i Study

I i

SubliCt ind3x 33 l l

1 NUREG/CR4590 V02. AGING OF NUCLEAR RTATION DIESEL Food And 86eed Cooling GENERATORS Evoluation Of Operating And Expert NUREGICR4946: DAVIS BESSE UNCERTAINTY STUDY. f l

Fillet Wold Undercut DWfusion Flame NUREG/CR4952: EXPERIMENTAL STUDY OF FILLET WELD UNDER-NUREG/CR-4534: ANALYSTS OF DIFFUSION FLAME TESTS.

CUT EFFECTS ON WELDED TUBING STRUCTURES UNDER CEN.

Direct Containment Heating ^ ^ '

NUREG/CR4871: RESULTS FROM THE DCH 1 EXPERIMENT. Film Bolling NUREG/CR4896. CONTAINMENT LOADS DUE TO DIRECT CONTAIN-MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysts NUREG/CR-4972: TWO-PHASE FLOW REGIME TRANSITION CRITE-And Calculations With The CONTAIN Code. RfA IN POST DRYOUT REGION BASED ON FLOW VISUALIZATION EXPERIMENTS.

D6ssolution NUREG/C 4 LONG TERM PERFORMANCE OF SPENT FUEL Fi sour Requ ement REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES.

Does Calculation NUREG/CR4418. DOSE CALCULATION FOR CONTAMINATION OF F6selon Product THE SKIN USING THE COMPUTER CODE VARSKIN. NUREG/CR-4925: FISSION PRODUCT BEHAVIOR DURING THE PBF SEVERE FUEL DAMAGE TEST 11.

Doolmetry NUREG/CR4418. DOSE CALCULATION FOR CONTAMINATION OF Flaw Denotty THE SKIN USING THE COMPUTER CODE VARSKIN NUREG/CR4860: FLAW DENSITY EXAMINATIONS OF A CLAD BOIL-NUREG/CR4886. ANALYSIS OF THE NESDlP2 AND NESDiP3 RADIAL ING WATER REACTOR PRESSURE VESSEL SEGMENT, SHIELD AND CAVITY EXPERIMENTS.

Flow Regime Transition Dynam6c Load NUREG/CR 4904. INVESTIGATION OF STEEL CONTAINMENT BUCK-NUREG/CA4972: TWO. PHASE FLOW REGIME TRANSITION CRITE-LING FROM DYNAMIC LOADS. RIA IN POST DRYOUT REGION BASED ON FLOW VISUALf2ATION EXPERIMENTS.

N /CR-3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE WESTERN OHIO INDIANA REGION Final Report, October 1981 Sep NU EG 32 8 V05: STRUCTURAL INTEGRITY OF WATER REAC-tember 1988, TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For NUREG/CR-4851: SEISMICITY 1886-09 IN THE SOUTHEASTERN UNITED STATES.The Aftershock Sequence Of The Charleston, South NU G/CR4219 V04 N1: HEAVY SECTION STEEL TECHNOLOGY Carohna Earthquake PROGRAM. Semiannual Progress Report For October 1986 March NUREG/CR 4910: RELAY CHATTER AND OPERATOR RESPONSE NgG/CR4760:

A ER A LARGE EARTHOUAKE.An improved PRA Methodology With TEST OF 6-IN THICK PRES $URE VESSELS. Sones gg gg ,g g NUREG/CR-4974. INTRAPLATE SEISMICITY IN THE EASTERN Matenal.

UNITED STATES.

Fractured Rock Effluent NUREG/CH4654 RADIONUCLIDES TRANSPORT AS VAPOR NUREG/CR-2907 V05: RADIOACTIVE MATERIALS RELEASED FROM THROUGH UNSATURATED FRACTURED ROCK.

NUCLEAR POWER PLANTS. Annual Report 1984.

Fractures Electrodeposit NUREG/CR4861 SURFACE SPECTROSCOPY OF PRESSURE NUREG/CR4865: THE SOLUBILITY OF ELECTRODEPOSITED TC(IV) VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN OX1 DES.

PWR ENVIRONMENTS.

Embrittlement Fuel l NUREG/CR 4744 V01 N2: LONG-TERM EMBRITTLEMENT OF CAST NUREG 1281: EVALUATION OF THE QUALIFICATION OF SPERT FUEL j DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual FOR USE IN NON POWER REACTORS.

Report,Apnl-September 1986.

NUREG 1282: SAFETY EVALUATION REPORT ON HIGH URANIUM 5 Enforcement Action CONTENT, LOW ENRICHED URANIUM ZlRCONIUM HYDRIDE  !

NUREG 0940 V06 NO2: ENFORCEMENT ACTIONS SIGNIFICANT AC.

FUELS FOR TRIGA REACTORS. Docket No. 50163.(GA 4 TlONS RESOLVED Ouarterty Progres6 Report,Apr61 June 1987. Technologies, incorporated)

Fuel Ctedding C 5 7 EDICTION AND M!TIGATION OF EROSIVE COR. NUREG/CR4485 V02: THE IMPACT OF FUEL CLADDING FAILURE ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE.

POWER PLANTS. AR POWER PLANTS Case Study PWR Dunng An Outage Event Fuel Rod Material NUREG/CR4891, PROPERTIES OF REACTOR FUEL ROD MATERIALS NUREG 0525 R13 SAFEGUARDS

SUMMARY

EVENT LIST (SSEL).

AT HIGH TEMPERATURES F#nal Summary Report Severe Core Expert Opinion . Damage Property Tests Program.

NUREG/CR-4962: METHODS FOR THE ELICITATION AND USE OF And tions For ure Re rh NU E R.5008. DEVELOPMENT OF A TESTING AND ANALYSIS METHODOLOGY TO DETERMINE THE FUNCT!ONAL CONDITION OF Fanure SOLENC. 9 OPERATED VALVES.

NU%G/CR-4747 VOI: AN AGING FAILURE SURVEY OF LIGHT WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. General Elect. i BWR4 s NUREG/CA 4767; SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF i Feet FourMr Transform Program A GENEEAL ELECTRIC BWR4/ MARK I Case Study.

NUREG, ' R-5008 DEVELOPMENT OF A TESTING AND ANALYCIS METHODOLOGY TO DETERMINE THE FUNCTIONAL CONDITION OF Generic Safety issue 82 SOLENOID OPERATED VALVES NUREG/CR4982. SEVERE ACCIDENTS IN SPENT FUEL POOLS IN s Fettgue SUPPORT OF GENERIC SAFETY ISSUE 82 i NUREG/CR4863 SURFACE SPECTROSCOPY OF PRESSURE Geologic D6sposal ,

VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN NUREG/CR-2452: AISK METHODOLOGY FOR GEOLOGIC DISPOSAL l PWR ENVIRONMENTS. OF RADIOACTIVE WASTE. Final Report.

1 l

q

)

34 Subject Ind:x l l

l Geophysical Investigetton Inhalation Exposure )

NUREG/CR-3145 V05. GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR-4986. RADIAT60N DOSE ESTIMATES AND HAZARD EVAL- -

WESTERN OHIO-INDIANA REGION Frial Report. October 1981 Sep- UATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Final Report.

I tomber 1986.

Inservice Inspection l Geophysical Techr*iue NUREG/CR4731 V01: RESIDUAL UFE ASSESSMENT OF MAJOR NUREG/CR-4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR UGHT WATER REACTOR COMPONENTS OVERVIEW Volume 1. .

SITE CHARACTER 12ATION IN BASALT, SALT AND TUFF.

inspection NUREG 0040 V11 NO2: LICENSEE CONTRACTOR AND VENDOR IN- l N EG/CR4919: FIELD TESTING OF BENTONITE AND CEMENT SPECTION STATUS REPORT. Quarterty Report.Apni June 1987.(White BOREHOLE PLUGS IN GRANITE.

Book)

Grant NUREG 1194: CONSTRUCTION APPRAISAL TEAM INSPECTION RE-NUREG/CR-5007; PREDICTION AND MITlGATION OF EROSIVE COR. SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC-ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR TiVITIES. 3 NUREG/CR 4583 V03: DEVELOPMENT AND VALIDATION OF A REAL- j POWER PUNTS.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER l Graphite REACTOR COMPONENTS. Annual Report. October 1985 September NUREG/CR4981: A SAFETY ASSESSMENT OF THE USE OF GRAPH- 1986.

ITE IN NUCLEAR REACTORS LICENSED BY THE U S. NRC.

instruments {

H(2) Air Mixture NUREG 1284: PROGRAM pun FOR CORRECTION OF U S. INSTRU- .

NUREG/CR 4905: DETONABILITY OF H2-AIR-DILUENT MIXTURES- MENT DEGRADATION OR FAILURE IN THE UPPER PLENUM TEST l FACILITY (UPTF)IN THE FEDERAL REPUBUC OF GERMANY. i Hested Detonatlon Tube NUREG/CR-4905: DETONABILITY OF H2 AIR-DILUENT MIXTURES. Integrated ReliabHity Heavy Section Steel Technology NUREG/CR 4844 DRFT: INTEGRATED RELIABluTY AND RISK ANALY-NUI.EG/CR4219 V04 N1: HEAVY SECTION STEEL TECHNOLOGY SIS SYSTEM (IRRAS) USER'S GUIDE VERSION 1.0 (DRAFT).

RAM Serniannual Progress Report For October 1986 March NUREG-1185 V01 DRFT: INTEGRATED SAFETY ASSESSMENT High Pressure in}ection REPORT. Integrated Safety Assessment Program.Haddam Neck NUREG/CR 4978: THE COOLDOWN ASPECTS OF THE TMI-2 ACCI- Plant. Docket No. 50 213. (Connectcut Yankee Atome Power Compa-DENT. ny)

NUREG 1185 V02 DRFT: INTEGRATED SAFETY ASSESSMENT High-Level Waste REPORT. Integrated Safety Assessment Program.Haddam Neck NUREG/CR4846. HIGH LEVEL WASTE PRECLOSURE SYSTEMS Plant. Docket No. 50-213. (Connecticut Yankee Atomic Power Compa-SAFETY ANALYSIS Phase 2, Final Report. nY)

NUREG/CR4954: LONG TERM PERFORMANCE OF SPENT FUEL N G 4 YST PERFORMANCE OF HIGH LEVEL WASTE '"('u E / 497 INTRAPLATE SEISMICITY OUTSIDE OF THE UNITED STATES.

High-Level Waste Repository NUREG/CR.4974: INTRAPLATE SEISMICITY IN THE EASTERN NUREG/CR 4161 V03: CRITICAL PARAMETERS FOR A HIGHo.EVEL UNITED STATES.

WASTE REPOSITORY. Volume 3 Salt. inventory Difference Data NUREG/CR-4654. RADIONUCLIDES TRANSPORT AS VAPOR THROUGH UNSATURATED FRACTURED ROCK. NUREG 0430 V07 NO2: LICENSED FUEL FACILITY STATUS REPORT. inventory Detterence Data. July December 1986.(Gray Book ll)

Human Reflability NUREG/CR-4846: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS lodine Partition Coefficient SAFETY ANALYSIS Phase 2, Fin +1 Report. NUREG/CR-4817. IODINE PARTITION COEFFICIENT MEASUREMENTS AT SIMULATED PWR STEAM GENERATOR CONDITIONS.Intenm Hydrogen Data Report.

NUREG/CR 4534. ANALYSIS OF DIFFUSION FLAME TESTS.

LER N EG/ 896 CONTAINMENT LOADS DUE TO DIRECT CONTAIN~ COMPILA O For ont Of June 1987 MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysis NUREG/CR-2000 V06 N7: LICENSEE EVENT REPORT (LER)

And Calculations With The CONTAIN Code. COMPILATION For Month Of July 1987.

NUREG/CR 2000 V06 N8: LICENSEE EVENT REPORT (LER)

ARRAS COMPILATION For Month Of August 1987.

NUREG'CR-4844 DRFT: INTEGRATED RELIABILITY AND RISK ANALY.

j SIS SYSTEM (IRRAS) USER'S GUIDE VERSION 1.0 (DRAFT). gtpwpag '

ISAp NUREG 1213 RO1: PLANS AND SCHEDULES FOR IMPLEMENTATION NUREG 1185 V01 DAFT: INTEGRATED SAFETY ASSESSMENT OF U S. NUCLEAR REGULATORY COMMISSION RESPONSIBILITIES REPORT. integrated Safety Assessment Program.Haddam Neck UNDER THE LOW LEVEL RADIOACTIVE WASTE POLICY AMEND-Plant.Dockrt No. 50 213. (Connecteut Yankee Atome Power Compa. MENTS ACT OF 1985 (P.L 99 240).

ny)

NUREG-1185 V02 DRFT: INTEGRATED SAFETY ASSESSMENT LWR REPORT. integrated Safety Assessment Program.Haddam Neck NUREG/CR 2907 V05: RADIOACTIVE MATERtALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1984.

Plant. Docket No. 50-213. (Connecteut Yankee Atome Power Compa.

ny) NUREG/CR 3228 V05: STRUCTURAL INTEGRITY OF WATER REAC-TOR PREEGURE BOUNDARY COMPONENTS Annual Report For .

Implementation 1986. l NUREG-1213 R01: PLANS AND SCHEDULES FOR IMPLEMENTATION NUREG/CR 4216. EXPERIMENTAL RESULTS FOR A 1.8-SCALE STEEL j OF U S. NUCLEAR REGULATORY COMMISSION RESPONSIBILITIES MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED 4 UNDER THE LOW-LEVEL RADIOACTIVE WASTE POLICY AMEND- TO FAILURE.

MENTS ACT OF 1985 (P.L.99-240). NUREG/CR4583 V03: DEVELOPMENT AND VAUDATION OF A REAL-TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER inder REACTOR COMPONENTS. Annual Report, October 1985 September NUREG 0304 V12 NO2: REGULATORY AND TECHNICAL REPORTS 1986.

(ABSTRACT INDEX JOURNAL). Compilation For Second Quarter NUREG/CR 4731 V01: RESIDUAL UFE ASSESSMENT CF MAJOR l 1987 Apni June. UGHT WATER REACTOR COMPONENTS - OVERVIEW Volume 1.

I 1

_ _ _ _ _ - _ _ _ _ . . - _ . I

Subj:ct IndIx 35 NUREG/CR4744 V01 N2. LONG TERM EMBRITTLEMENT OF CAST NUREG/CR4781 DAFT; STUDY OF SEVERE ACCIDENT MITIGATION DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual SYSTEMS.

Report, April-September 1986. NUREG/CR 4870: AN EVALUATION OF THE EFFECTS OF DESIGN NURE G/CR4747 V01: AN AGING FAILURE SURVEY OF LIGHT DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. BUILDINGS.

NUREG/CR4769. RISK EVALUATIONS OF AGING PHENOMENA.The Linear Aging Rehabi Loss-Of Feedwater Transient NUREG/CR4781 DR_hty Model H: STUDY OFAnd its Extensions SEVERE ACCIDENT MITIGATION NUREG/CR-4946: DAVIS-BESSE UNCERTAINTY STUDY.

SYSTEMS NUREG/CR4870: AN EVALUATION OF THE EFFECTS OF DESIGN Low Upper Shelf Toughness DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT NUREG/CR-4760: TEST OF 6-IN4 THICK PRESSURE VESSELS.Senes BUILDINGS, 3:Intermedete Test Vessel V 8A.Teanng Behavior Of Low Upper-Shelf Large Pressure Transient Matenal NUREG/CR4931; RESPONSE OF CENTRIFUGAL AND AXI VANE Low Enriched Uranium Fuel BLOWERS TO LARGE PRESSURE TRANSIENTS.

NUREG 1282: SAFETY EVALUATION REPORT ON HIGH-URANIUM Leaching CONT ENT, LOW-ENRICHED Ur' \NIUM-ZlRCONIUM HYDRIDE NUREG/CR-4954: LONG TERM PERFORMANCE OF SPENT FUEL FUELS FOR TRIGA REACTOR $. Docket No. 50163.(GA WASTE FORMS. Technologies, incorporated)

Leak Rate Low-Level Radioactive Waste Policy Act NUREG/CR-4216: EXPERIMENTAL RESULTS FOR A 1.8 SCALE STEEL NUREG-1213 ROI: PLANS AND SCHEDULES FOR IMPLEMENTATION MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED OF U.S. NUCLEAR REGULATORY COMMISSION RESPONSIBILITIES TO FAILURE. UNDER THE LOW-LEVEL RADIOACTIVE WASTE POLICY AMEND-MENTS ACT OF 1985 (P.L 99 240).

NUREG-0750 V24102- INDEXES TO NUCLEAR REGULATORY COM- Low Level Waste MISSION ISSUANCES. July December 1986 NUREG 1274 REVIEW PROCESS FOR LOW LEVEL RADIOACTIVE NUREG 0750 V251401: NUCLEAR REGULATORY COMMISSION IS- WASTE DISPOSAL LICENSE APPLICATION UNDER LOW LEVEL RA-SUANCES FOR JANUARY 1987.Pages 1-62. OlOACTIVE WASTE POLICY AMENDMENTS ACT.

NUREG-0750 V25 NO2: NUCLEAR REGULATORY COMMISSION IS- NUREG/CR-4897: LOW LEVEL WASTE SOURCE TERM SUANCES FOR FEBRUARY 1987.Pages63-128. FVALUATION. Review of Published Modeling And Expenmental NUREG-0750 V25 NO3: NUCLEAR REGULATORY COMM!SSION IS- Work,And Presentation Of Low Level Waste Source Term Modehng SUANCES FOR MARCH 1987.P es 129-266 Framework And Prehmirery Model Development.

NUREG-0750 V25 N04. LUCLEA REGULATORY COMMISSION IS-SUANCES FOR APRIL 1987.Pages 267416. Mark i License Application NUREG/CR 4767: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF A GENERAL ELECTRIC BWR4/ MARK I Case Study.

NUREG-1274 REVIEW PROCESS FOR LOW LEVEL RADIOACTIVE WASTE DISPOSAL LICENSE APPLICATION UNDER LOW LEVEL RA-DIOACTIVE WASTE POLICY AMENDMENTS ACT.

[UfEG/C 2 METHO OR RECURRING LOSS TESTS.

Ucensed Fuel Facility NUREG 0430 V07 NO2: LICENSED FUEL FACILITY STATUS Meers Fault REPORT. Inventory Difference Data. July-Decomser 1986.(Gray Book ll) NUREG/CR4937: INVESTIGATION OF THE MEERS FAULT IN SOUTH-WESTERN OKLAHOMA.

Licensed Operating Reactors NUREG-0020 V11 NOI: LICENSED OPERATING REACTORS STATUS Mitigation

SUMMARY

REPORT. Data As Of December 31,1986 (Gray Book 1) NUREG/CR-4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATION SYSTEMS.

Uconsee PlUREG/CR i 5007: PREDICTION AND MITIGATION OF EROSIVE COR.

NUREG/CR 4958 IMPACT OF PROPOSED CINANCIAL ASSURANCE ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES. POWER PLANTS.

Ucensee Event Report NDE NUREG/CR 2000 V06 N6. LICENSEE EVENT REPORT (LER) NUREG-1194: CONSTRUCTION APPRAISAL TEAM INSPECTION RE-COMPILATION For Month Of June 1987- SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC-NUREG/CR-2000 V06 N7. LICENSEE EVENT REPORT (LER) TlVITIES~

COMPILATION For Month Of July 1987 NUREG/CR 2000 V06 N8. LICENSEE EVENT REPORT (LER) NESDIP i COMPILATION For Month Of August 1987. NUREF OR-4886. ANALYSIS OF THE NESDlP2 AND NESDIP3 RADIAL I Ught Water Reactor SHIELD AND CAVITY EXPERIMENTS. j NUREG/CR-2907 V05 RADIOACTIVE MATERIALS RELEASED FROM NESTOR NUR?G/CR-4886 ANALYSIS OF THE NESDlP2 AND NESDlP3 RADIAL NURE /CR 3 V05 TR CT A GR OF WATER REAC.

TOR PRESSURE BOUNDARY COMPONENTS Annual Report For SHIELD AND CAVITY EXPERIMENTS.

1986 NUREG/CH 4216. EXPERIMENTAL RESULTS FOR A 1.8 SCALE STEEL N'Ph'''Y i MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURl2ED NUREG/CR 4951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE- 1 VERSIBILITY OF RENAL INJURY IN THE RAT, j TO F AILURE.

NUREG/CR 4583 V03 DEVELOPMENT AND VALIDATION OF A REAL.

Nondestructive Examination TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER REACTOR COMPONENTS Annual Report. October 1985 September NUREG 1194 CONSTRUCTION APPRAISAL TEAM INSPECTION RE-1986 SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC-NUREG/CR-4731 V01: RESIDUAL LIFE ASSESSMENT OF MAJOR TIVITIES i LIGHT WATER REACTOR COMPONENTS OVERVIEW. Volume 1. NUREGICR 4583 V03 DEVELOPMENT AND VAllDATION OF A REAL- l NUREG/CR-4744 V01 N2. LONG TERM EMBRITTLEMENT OF CAST TV IAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER  ;

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual RE ACTOR COMPONENTS. Annual Report,0ctober 1985 September i Report, April. September 1986 1986 l NUREG/CR 4747 V01: AN AGING FAILURE SURVEY OF LIGHT WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. Nuclear Reactor Study l NUREG/CR 4769 RISK EVALUATIONS OF AGING PHENOMENA The NUREG-1285. NRC STAFF EVALUATION OF THE GENERAL ELECTRIC Linear Aging Rehabihty Model And its Extensions COMi>ANY NUCLEAR REACTOR STUDY (" REED REPORT")

__ _ _____ _ .J

36 Subject Index Occupational Radiation Policy And Plaaning Guidance NUREG/CR 4485 V02' THE IMPACT OF FilEL CLADDING FAILURE NUREG 0085106: U.S. NUCLEAR REGULATORY COMMISSION POLICY EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE- AND PLANNING GUIDANCE 1987.

AR POWER PLANTS. Case Study PWR Dunng An Outage.

Proclosure System Operating Esperience Feedback NUREG/CR4846: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS NUREG 1275: OPERATING EXPERIENCE FEEDBACK REPORT NEW SAFETY ANALYSIS Phase 2. Final Report.

PtANTS. Commercial Power Reactors.

Pressure EG 4910: RELAY CHATTER AND OPERATOR RESPONSE NUREG/CR 4913: ROUND-ROBIN PRETEST ANALYSES OF A 1:6-AFTER A LARGE EARTHOUAKE.An improved PRA Methodology With SCALE REINFORCED CONCRETE CONTAINMENT MODEL SUBJECT Case Stude TO STATIC INTERNAL PRESSURIZATION.

Ownerehlp Pressure Boundary NUREG 0327 R04: OWNERS OF NUCLEAR POWER NUREG/CR-3228 VOS: STRUCTURAL INTEGRITY OF WATER REAC-PLANTS Percentage Ownership Of Commercial Nuclear Power Plants TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For By Otikty Companies. 1986.

Oxide Pressure Vessel NUREG/CR-4865: THE SOLUBPLITY OF ELECTRODEPOSITED TC(IV) NUREG/CR-4219 V04 N1: HEAVY SECTION STEEL TECHNOLOGY OXIDES. PROGRAM Semiannual Progress Report For October 1986 March 1987.

Oside D6esolution NUREG/CR-4760: TEST OF 6 IN. THICK PRESSURE VESSELS.Senes NUREG/CR-5001: PREDICTION AND MITIGATION OF EROSIVE COR' 3: Intermediate Test Vessel V 8A.Teanng Behavior Of Low Upper Shelf t ROSlVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR Malenal.

POWER PLANTS- NUREG/CR4860: FLAW DENSITY EXAMINATIONS OF A CLAD BOIL-ING WATER REACTOR PRESSURE VESSEL SEGMENT, PWR NUREG/CR4485 V02: THE IMPACT OF FUEL CLADDING FAILURE NUREG/CR4863: SURFACE SPECTROSCOPY OF PRESSURE EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE.

^ ^

hb6 g iphEL NM N NURE R47 TD$W C Y H AT M V dNALYSIS OF NUREG/CR-4886: ANALYSIS OF THE NESDIP2 AND NESDIP3 RADIAL A COMBUSTION ENGINEERING PRESSURIZED WATER SHIELD AND CAVITY EXPERIMENTS.

REACTOR Case Study NUREG/CR4817: IODINE PARTITION COEFFICIENT MEASUREMENTS Pressurized Containment AT SIMULATED PWR STEAM GENERATOR CONDITIONS.Intenm NUREG/CR4216: EXPERIMENTAL RESULTS FOR A 1.8 SCALE STEEL Data Report. MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED NUREG/CR4863. SURFACE SPECTROSCOPY OF PRESSURE TO FAILURE.

VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN PWR ENVIRONMENTS. Pressurized Water Reactor NUREG/CR-4945:

SUMMARY

OF THE SEMISCALE PROGRAM (1965- NUREG/CR 4485 V02: THE IMPACT OF FUEL CLADDING FAILURE 1986). EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE-AR POWER PLANTS Case Study-PWH Dunng An Outa9e.

P:ttern Recognition NUREG/CR4710: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR-4577: AUTOMATED LONG TERM SURVEILLANCE OF A A COMBUSTION ENGINEERING PRESSURIZED WATER COMMERCIAL NUCLEAR POWER PLANT. REACTOR Case Study.

NUREG/CR4817: LODINE PARTITION COEFFICIENT MEASUREMENTS  :

NA / 489 : A STUDY OF THE EFFECTS OF PENETRAT;ON FRAMING ON STEEL CONTAINMENT BUCKLING CAPACITY.

NUR G 863: SURFACE SPECTROSCOPY OF PRESSURE Percentage Ownership VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN NUREG 0327 R04: OWNERS OF NUCLEAR POWER PWR ENVIRONMENTS.

PLANTS Percentage Ownership Of Commercial Nuclear Puwer Plent6 NUREG/CR4945:

SUMMARY

OF THE SEMISCALE PROGRAM (1965-By Utihty Companies. 1986).

Performance Modeling Probabilletic Risk Assessment NUREG/CR-4956. SYSTEM PERFORMANCE OF HIGH LEVEL WASTE NUREG/CR 4710: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF PACKAGE COMPONENTS. A COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR. Case Study.

P:titions For Rulemakin9 NUREG/CR-4767: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG 0936 V06 NO2- NRC REGULATORY AGENDA.Ouarterly A GENERAL ELECTRIC BWR4/ MARK l. Case Study.

Report.Apni June 1987 NUREG/CR 4768 V01: METHODOLOGY AND APPLICATION OF SUR-Pipe Break ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER PLANT. Task 1 - Analysis Of ANO 1 And Oconee PRAs.

NUREG/CR4407: PIPE BREAK FREQUENCY ESTIMATION FOR NU- NUREG/CR-4768 V02: METHODOLOGY AND APPLICATION OF SUR.

CLEAR POWER PLANTS' ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER Piping PLANT. Final Report.

NUREG/CR4943: PREPARATION OF DESIGN SPECIFICATIONS AND NUREG/CR4810: EVALUATION OF DIESEL UNAVAILABILITY AND DESIGN REPORTS FOR PUMPS. VALVES. PIPING.AND PIPING SUP- RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

PORTS USED IN SAFETY RELATED PORTIONS OF NUCLEAR NUREG/CR4834 V01: RECOVERY ACTIONS IN PRA FOR THE R:SK POWER PLANTS' METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Volume 1 Development Of The Data-Based Method.

Piping Support NUREG/CR4900 V01: COMPONENT FRAGILITY RESEARCH NUREG/CR4943; PREPARATION OF DESIGN SPECIFICATIONS AND PROGRAM Phase i Demonstration Tents Summary Report.

DESIGN REPORTS FOR PUMPS. VALVES. PIPING.AND PIPING SUP. NUREG/CR4900 V02: COMPONENT FRAGILITY RESEARCH PORTS USED IN SAFETY RELATED PORTIONS OF NUCLEAR PROGRAM Phase i Demonstration Tests. Appendices.

POWER PLANTS. NUREG/CR-4910: RELAY CHATTER AND OPERATOR RESPONSE AFTER A LARGE EARTHOUAKE.An improved PRA Methodology With Piping System Case Studies.

NUREG/CR-5007: PREDICTION AND MITIGATION OF EROSIVE.COR. NUREG/CR4962: METHODS FOR THE ELICITATION AND USE OF ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR EXFERT OPINION IN RISK ASSESSMENT Phase 1 A Cntical Evalue-POWER PLANTS. tion And Directions For Future Research.

i SubjsCt Ind3x 37 Pump Recovery Actlon NUREG/CR4943. PREPARATION OF DESIGN SPECIFICATIONS AND NUREG/CR4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK DESIGN REPORTS FOR PUMPS. VALVES, PIPING,AND PIPING SUP- *AETHODS INTEGRATION AND EVALUATION PROGRAM PORTS USED IN SAFETY-RELATED PORTIONS OF NUCLEAR 4M:Er' 4 ane 1; Development Of The Data Based Method.

1 POWER PLANTS.

Recurring Lose Ouality Assurance NUREG/CR 5002. METHODS FOR RECURRING LOSS TESTS.

NUREG/CR4962 METHODS FOR THE ELICITATION AND USE OF EXPERT OPINION IN RISK ASSESSMENT. Phase I A Cntical Evalue- Reed Report tson And Directions For Future Research. NUREG-1285 NRC STAFF EVALUATION OF THE GENERAL ELECTRIC COMPANY NUCLEAR REACTOR STUDY (" REED REPORT").

NUREG/CR4946: DAVIS-BESSE UNCERTAINTY STUDY. Reflect 6on Seismology NUREG/CR 4975: A REVIEW OF THE RESOLVING POWER OF RE-RMIEP FLECTION SEISMOLOGY METHODS TO DETECT SUBSURFACE NUREGICR4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK FAULTS AND/OR CHANGES IN LAYER THICKNESS i l

METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Volume 1. Development Of The Data-Based Method. Regulation NUREG-1266 V01: NRC SAFETY RESEARCH IN SUPPORT OF REGV-Radial Sh6 eld LATION 1986.

NUREG/CR4886. ANALYSIS OF THE NESDIP2 AND NESDIP3 RADIAL SHIELD AND CAVITY EXPERIMENTS. Regulatory Agenda NUREG 0936 V06 NO2; NRC REGULATORY AGENDA Ouarterty Stadiation Dose Report, April-June 1987.

NUREG/CR4418: DOSE CALCULATION FOR CONTAMINATION OF THE SKIN USING THE COMPUTER CODE VARSKIN Regulatory And Technical Report NUREG/CR4986 RADIATION DOSE ESTIMATES AND HAZARD EVAL

• NUREG 0304 V12 NO2: REGULATORY AND TECHNICAL REPORTS UATIONS FOR INHALED AIRBORNE RADIONUCLIDES. Final Report- (ABSTRACT INDEX JOURNAL). Compilation For Second Quarter G87,ApWune.

Radiation Exposure NUREG/CR-4485 V02: THE IMPACT OF FUEL CLADDING FAILURE Relay Chatter EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE. NUREG/CR 4910: RELAY CHATTER AND OPERATOR RESPONSE

1. ^ ^ " '^"

• U NURE R 88 NTE RE O F I SSA ASUREMENTS. Case Studies.

Radiation Monitoring Network NUREG 0837 V07 N01: NRC TLD DIRECT RADIATION MONITORING R /CR-4810: EVALUATION OF DIESEL UNAVAILABILITY AND NETWORK. Progress Report, January March 1987. RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

Radioactive Material "

NUREG/CR.2907 V05: RADIOACTIVE MATERIALS RELEASED FROM N RE / R 4951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE-NUCLEAR POWER PLANTS Annual Report 1984.

VERSIBILITY OF RENAL INJURY IN THE RAT.

Radioactive Waste N REG V N04: REPORT TO CONGRESS ON ABNORMAL OF R D OA IVE AST ana Report OCCURRENCES. October December 1986.

Redlo6odine Resuspens6on Resolv ng wer NUR /C 4953 RR LAT OF RADIOIODINE RESUSPENSION

~

FLECTION SEISMOLOGY METHODS TO DETECT SUBSURFACE Radlonucinde FAULTS AND/OR CHANGES IN LAYER THICKNESS.

NUREG/CR 4654. RADIONUCLIDES TRANSPORT AS VAPOR Risk THROUGH UNSATURATED FRACTURED ROCK.

OF RADIOACTIVE WASTE. Final Report.

Radionucl6de Release Rate NUREG/CR4897: LOW LEVEL WASTE SOURCE TERM NUREG/CR4768 V01: METHODOLOGY AND APPLICATION OF SUR-EVALUATION. Review of Published Modeling And Expenmental ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER Work And Presentation Of Low Level waste Source Term Modeling PLANT. Task 1 Analysis Of ANO-1 And Oconee PRAs Framework And Preliminary Model Development. NUREG/CR4768 V02. METHODOLOGY AND APPLICATION OF SUR-ROGATE PLANT PRA ANALYS!S TO THE RANCHO SECO POWER Rat PLANT Final Report.

NUREG/CR4951: NEPHROTOXICITY OF URANYL FLUORIDE AND RE. NUREG/CR-4769. RISK EVALUATIONS OF AGING PHENOMENA.The VERSIBILITY CF RENAL INJURY IN THE RAT. Linear Aging Reliability Model And its Extensions NUREG/CR 4810. EVALUATION OF DIESEL UNAVAILABILITY AND Reactor Safety RISK EF5ECTIVE SURVEILLANCE TEST INTERVALS.

NUREG/CA 4805 V01: REACTOR SAFETY RESEARCH SEMlANNUAL NUREGICR4844 DRFT: INTEGRATED RELIABILITY AND RISK ANALY-REPORT. January June 1986 SIS SYSTEM (IRRAS) USER'S GUIDE VERSION 1.0 (DRAFT)

Reactor Technology Risk Assessment NUREG/CP 0088. TRANSACTIONS OF THE 9TH INTERNATIONAL NUREG/CR4962. METHODS FOR THE ELICITATION AND USE OF CONFERENCE ON STRUCTURAL MECHANICS IN REACTOR EXPERT OPINION IN RISK ASSESSMENT. Phase 1 A Cntical Evalva-TECHNOLOGY. Panel Session JK: Structural And Mechanical Engineer- tion And Directions For Future Research.

mg Research At The U S. Nuclear Regulatory Commission.

Rlak Reduction Reactor Vessel NUREG/CR-4781 DAFT: STUDY OF SEVERE ACCIDENT MITIGATION NUREG/CR4896: CONTAINMENT LOADS DUE TO DIRECT CONTAIN- SYSTEMr.

MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysis And Calculations With The CONTAIN Code Rulemaking NUREG/CR-4978 THE COOLDOWN ASPECTS OF THE TMi-2 ACCI- NUREG/CR-4958 IMPACT QF PROPOSED FINANCIAL ASSURANCE DENT. REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES Readinosa Review Rules NUREG-1278 VOGTLE UNIT 1 READINESS REVIEW.Assessraent Of NUREG 0936 V06 NO2: NRC REGULATORY AGENDA Ouarterty Georgia Power Company Readiness Review Pilot Program. Report. April June 1987.

38 Subject Ind:x SAFT UT NUREG/CR4851: SEtSMICITY 1886-89 IN THE SOUTHEASTERN NUREG/CR4583 V03- DEVELOPMENT AND VAUDATION OF A REAL- UN!TED STATES.The Aftershock Sequence Of The Charleston, South TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER Carohna Earthquake.

REACTOR COMPONENTS. Annual Report. October 1985 September NUREG/CR4973: INTRAPLATE SEISMICITY OUTSIDE OF THE 1986. UNITED STATES.

NUREG/CR4974: INTRAPLATE SEISMICITY IN THE EASTERN SPERT UNITED STATES.

HUREG 1281: EVALUATION OF THE QUALIFICATION OF SPERT FUEL FOR USE IN NON-POWER REACTORS. Seismology Safeguards Summary Event List NUREG/CR-4975: A REVIEW OF THE RESOLVING POWER OF RE-FLECTION SEISMOLOGY METHODS TO DETECT SUBSURFACE NUREG-0525 R13 SAFEGUARDS

SUMMARY

EVENT UST (SSEL).

FAULTS AND/OR CHANGES IN LAY'M THICKNESS.

Safety NUREG1 01: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-N RE /C 4

SUMMARY

OF THE SEMISCALE PROGRAM (1965 NUREG/CR.3228 V05: STRUCTURAL INTEGRITY OF WATER REAC. 1986).

TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For 3

NUREG/CR4747 V01: AN AGING FAILURE SURVEY OF LIGHT NUREG 1144 ROI: NUCLEAR PLANT AGING RESEARCH (NPAR) PRO- i WATER REACTOR SAFETY SYSTEMS AND COMPONENTS. GRAM PLAN.

NUREG/CR4769 RISK EVALUATIONS OF AGING PHENOMENA.The Linear Aging Rehabihty Model And its Extensions Severe Accident NUREG/CR4805 V01: REACTOR SAFETY RESEARCH SEM1 ANNUAL NUREG 1264: CONTAINMENT INTEGRITY RESEARCH PROGRAM REPORT. January June 1986. PLAN.

NUREG/CR4846. HIGH-LEVEL WASTE PRECLOSURE SYSTEMS NUREG/CR4216: EXPERIMENTAL RESULTS FOR A 1:8-SCALE STEEL SAFETY ANALYSIS Phase 2, Final Report. MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED NUREG/CR4945.

SUMMARY

OF THE SEMISCALE PROGRAM (1965- TO FAILURE.

1986). NUREG/CR 4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATION SYSTEMS.

U EG/ A 98 A SAFETY ASSESSMENT OF THE USE OF GRAPH.

ITE IN NUCLEAR REACTORS LICENSED BY THE U.S. NRC.

NGS Safety Evaluation Report NUREG/CR4876: SILVER INDIUM CADMlUM CONTPOL ROD BEHAV.

NUREG 0781 SO4: SAFETY EVALUATION REPORT RELATED TO THE LOR AND AEROSOL FORMATION IN SEVERE REACTOR ACCI.

OPERATION OF SOUTH TEXAS PROJECT. UNITS 1 AND 2. Docket DENTS.

Nos 50498 And 50499 (Houston Lighting And Power Company} NUREG/CR 4982: SEVERE ACCIDENTS IN SPENT FUEL POOLS IN NUREG 1047 S06. SAFETY EVALUATION REPORT RELATED IU THE SUPPORT OF GENERIC SAFETY ISSUE 82.

OPERATION OF NINE MILE POINT NUCLEAR STATION, UNIT NA 15 NAL 10 PR R E TO THE N REG / R 25 FISSION PRODUCT BEHAVIOR DURING THE PBF OPERATION OF BEAVER VALLEY POWER STATION, UNIT 2. Docket SEVERE FUEL DAMAGE TEST 11.

No. 50412 (Duquesne Light Company et al)

NUREG-1232 VO1: SAFETY EVALUATION REPORT ON TENNESSEE Shutdown VALLEY AUTHORITY. Revised Corporate Nuclear Performance Plan. NUREG/CR-4710: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG 1282: SAFETY EVALUATION REPORT ON HIGH URANIUM A COMBUSTION ENGINEERING PRESSURIZED WATER CONTENT, LOW-ENRICHED URANIUM-ZlRCONIUM HYDRIDE REACTOR. Case Study.

FUELS FOR TRIGA REACTORS. Docket No. 50163 (GA NUREG/CR 4767: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF Technologies. incorporated) A GENERAL ELECTRIC BWR4/ MARK l. Case Study.

NUREG 1284: PROGRAM PLAN FOR CORRECTION OF U S. INSTRU-MENT DEGRADATION OR FAILURE IN THE UPPER PLENUM TEST Signat FACILITY (UPTF) IN THE FEDERAL REPUBLIC OF GERMANY. NUREG/CR-4577: AUTOMATED LONG TERM SURVEILLANCE OF A COMMERCIAL NUCLEAR POWER PLANT, Safety Regulation NUREC-1251 DRFT FC: IMPLICATIONS OF THE ACCIDENT AT CHER. Site Characterization NOBYL FOR SAFETY REGULATION OF COMMERCIAL NUCLEAR NUREG/CR-4957: SURVEY OF GEOPHYSICAL TECHNIQUES FOR POWER PLANTS IN THE UNITED STATES Draft For Comment. SITE CHARACTERIZATION IN BASALT, SALT AND TUFF, Saf:ty Research Skin Contamination NUREG 1266 V01: NRC SAFETY RESEARCH IN SUPPORT OF REGU- NUREG/CR.4418: DOSE CALCULATION FOR CONTAMINATION OF LATION 1986. THE SKIN USING THE COMPUTER CODE VARSWIN.

NUREG/CR 2331 V06 N4 SAFETY RESEARCH PROGRAMS SPON.

SORED BY OFFICE OF NUCLEAR REGULATORY Software RESEARCH.Quarterty Pronress Report October December 1986. l NUREG/CR-4640: HANDBOOK OF SOFTWARE QUALITY ASSURANCE i

, TECHNIOUES APPLICABLE TO THE NUCLEAR INDUSTRY.

NUREG/CR4457: AGING OF CLASS 1E BATTERIES IN SAFETY SYS- Solenoid Operated Valve TEMS OF NUCLEAR POWER PLANTS.

NUREG/CR-5008: DEVELOPMENT OF A TESTING AND ANALYSIS gg METHODOLOGY TO DETERMINE THE FUNCTIONAL CONDITION OF NUREG/CR4161 V03: CRITICAL PARAMETERS FOR A HIGH-LEVFL SOLENOID OPERATED VALVES.

W ASTE RLPOSITORY. Volume 3 Salt.

NUREG/CR4057: SURVEY OF GEOPHYSICAL TECHNIQUES FCH Solid Weste dis osal SITE CHA9ACTERIZATION IN BASALT. SALT AND TUM NUREG/CR-2907 V05: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1984.

Seismic Fragility NUREG/CR4900 V01: COL 1PONENT FRAGILITY RESEARCH Solubility PROGRAM Phase i Demonstration Tests Summary Report. NUREG/CR4865: THE SCLUBlUTY OF ELECTRODEPOSITED TC(IV)

NUREG/CR 4900 V02: COMPONENT FRAGILITY RESEARCH OXIDES.

PROGRAM Phase i Demonstration Tests. Appendices.

Seismicity NUREG/CR4897: LOW LEVEL WASTE SOURCE TERM NUREG/CR 3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE EVALUATION. Review of Published Modeling And Expenmental WESTERN OHlO-INDIANA REGION. Final Report. October 1981 - Sep. Work,And Presentation Of Low Level Waste Source Term Modehn0 tembe" 1986. Framework And Prehmenary Model Development.

1

Subjsct Ind3x 39 Spectroscopy NUREG f 255: TECHNICAL SPECIFICATIONS FOR SOUTH TEXAS NUREG/CR4863. SURFACE SPECTROSCOPY OF PRESSURE PROJECT, UNIT 1. Docket No. 50 498(Houston Lighting and Power VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN Compsay)

PWR ENVIRONMENTS. NUREG-1276. TECHNICAL' SPECIFICATIONS FOR BRAIDWOOD STATION UNITS 1 AND .2. Docket Nos. 50 456 And 504574 Common.

Spent Fuel Pool wealth Edison Company)

NUREG/CR4982: SEVERE ACCIDENTS IN SPENT FUEL POOLS IN NUREG 1279: TECHNICAL SPECIFICATIONS FOR BEAVER VALLEY SUPPORT OF GENERIC SAFETY ISSUE 82. POWER STATION, UNIT 2. Docket No. 50412,(Duquesne Light Company et af)

Spent Fuel Weste Form .

NUREG/CR4954: LONG TERM PERFORMANCE OF SPENT FUEL Temperature WASTE FORMS.

NUREG/CR-3024. SUSTAINED CONCRETE ATTACK BY LOW.

Startup TEMPERATURE FRAGMENTED CORE DEBRIS.

NUREG-1275. OPERATING EXPERIENCE FEEDBACK REPORT NEW NUREG/CR4891: PROPERTIES OF REACTOR FUEL ROD MATERIALS PLANTS Commercial Power Reactors. AT HIGH TEMPERATURES. Final Summary Report . Severe Core Damage Property Tests Program.

Steam Generator NUREG/CR 4953: CORRELATION OF RADIOIODINE RESUSPENSION NUREG/CR 4817: LODINE PARTITION COEFFICIENT MEASUREMENTS WITH TEMPERATURE AT TMI 2.

AT SIMULATED PWR STEAM GENERATOR CONDITIONSIntenm Data Report. Test interval NUREG/CR 4850 STEAM GENERATOR GROUP PROJECT. Task 10 - NUREG/CR4810:' EVALUATION OF DIESEL UNAVAILABILITY AND Secondary Side Examination. Final Report. RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

REG /CR 4216: EXPERIMENTAL RESULTS FOR A 1:8-SCALE STEEL NUREG 0540 V09 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY MODEL NUCLEAR POWER PLANT CONTAINMENT PRESSURIZED AVAILABLE. May 1 31,1987.

NUREG-0540 V09 N06. TITLE LIST OF DOCUMENTS MADE PUBLICLY NU EG/ 4 3: SURFACE SPECTROSCOPY OF PRESSURE AVAILABLE, June 1 30,1987.

VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN PWR ENVIRONMENTS. Transpod NUREG/CR4870: AN EVALUATION OF THE EFFECTS OF DESIGN DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT NUREG/CR4654: RADIONUCLIDES TRANSPORT AS VAPOR THROUGH UNSATURATED FRACTURED ROCK.

BUILDINGS NUREG/CR-4892. A STUDY OF THE EFFECTS OF PENETRATION Tubing FRAMING ON STEEL CONTAINMENT BUCKLING CAPACITY.

NUREG/CR4904: INVESTIGATION OF STEEL CONTAINMEN1 BUCK. NUREG/CR4952: EXPERIMENTAL STUDY OF FILLET WELD UNDER.

LING FROM DYNAMIC LOADS. CUT EFFECTS ON WELDED TUBING STRUCTURES UNDER CEN-TRIC AND ECCENTRIC CYCLIC LOADINGS.

Strategic Special Nuclear Material NUREG/CR 5002: METHODS FOR RECURRING LOSS TESTS. Tuff NUREG/CR4957: SURVEY OF GEOPHYSICAL TECHNIOUES FOR b 0^ ^ ^ ^

NU EG/CR4910- RELAY CHATTER AND OPERATOR RESPONSE AFTER A LARGE EARTHOUAKE.An improved PRA Methodology With Two-Phase FW Case SWes NUREG-1284 PROGRAM PLAN FOR CORRECTION OF U S INSTRU-Structural integrity MENT DEGRADATION OR FAILURE IN THE UPPER PLENUM TEST NUREG/CR4219 V04 N1: HEAVY SECTION STEEL TECHNOLOGY FACILITY (UPTF) IN THE FEDERAL REPUBLIC OF GERMANY.

PR RAM Semiannual Progress Report For October 1986 March Ultrasonic Testing NUREG/CR4583 V03: DEVELOPMENT AND VALIDATION OF A REAL-Structural Mechanics TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CP 0088. TRANSACTIONS OF THE 9TH INTERNATIONAL REACTOR COMPONENTS. Annual Report. October 1985 + September CONFERENCE ON STRUCTURAL MECHANICS IN REACTOR 1986.

TECHNOLOGY. Panel Session JK: Structural And Mechenecal Engineer-ing Research At The U S. Nuclear Regulatory Commission. Unc A a ya e DAVIS BESSE UNCERTAINTY STUDY.

L /YR4577. AUTOMATED LONG TERM SURVEILLANCE OF AUnplanned Event

. COMMEMIAL NUCLE AR POWER PLANT, NUREG 1275: OPERATING EXPERIENCE FEEDBACK REPORT NEW PLANTS. Commercial Power Reactors.

NUREG/CR4710. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF Unsaturated Zone A COMBUSTION ENGINEERING PRESSURIZED WATER NUREG/CR-4654: RADIONUCLIDES TRANSPORT AS VAPOR REACTOP Case Study. THROUGH UNSATURATED FRACTURED ROCK.

TLD Oranyl Fluoride NUREG 0837 V07 N01: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4951: NEPHROTOxlCiTY OF URANYL FLUORIDE Atc RE, NETWORK Progress Report. January March 1987. VERSIBILITY OF RENAL INJURY IN THE RAT.

TRIGA Reactor NUREG 1282 SAFETY EVAlbATION REPORT ON HIGH URANIUM NUREG/CR-4876 SILVER-INOlVM.CADM!UM CONTROL ROD BEHAV.

i CONTENT. LOW ENRICHED URANIUM ZlROONIUM HYDRIDE IOR AND AEROSOL FORMATION IN SEVERE REACTOR ACCl-FUELS FOR TRIGA REACTORS Docket No 50163 (G A DENTS.

Technologies, incorporated)

Tearing Behaylor VAR $ KIN NUREG/CR4760 TEST OF 8,1N. THICK PRESSURE VESSELS Senes NUREG/CR4418 DOSE CALCULATION FOR CONTAMINATION OF 3 Intermediate Test Vessel V.8A.Teanng Behawor Of Low Upper Shelf THE SKIN USING THE COMPUTER CODE VARSKIN l l

Matenal Valve i Technical Specification NUREG/CR4943 PREPARATION OF DESIGN SPECIFICATIONS AND NUREG-1253 TECHNICAL SPECIFICATIONS FOR NINE MILE POINT DESIGN REPORTS FOR PUMPS. VALVES. PIPING,AND PlPING SUP- 1 NUCLEAR STATION.UNtT 2 Docket No 50410(Niagara Mohawk PORTS USED IN SAFETY RELATED PORTIONS OF NUCLEAR Power Corporation.et al) POWER PLANTS e

1

40 Subj^ct Ind;x j

ventii tion system weieing NUR E G /CR-4931 ' RESPONSE OF CENTRIFUGAL AND AXI VANE BLOWERS TO LARGE PRESSURE TRANSIENTS NUREG 1194 CONSTRUCTION APPRAISAL TEAM INSPECTION RE-SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC-Waste Disposal TIVITIES NUREG 1274- REVIEW PROCESS FOR LOW LEVEL RADIOACTIVE

• *'d*

WASTE DISPOSAL LICENSE APPLICATION UNDER LOW-LEVEL RA.

DiCACTIVE WASTE POLICY AMENDMENTS ACT. NUREG/CR-4952: EXPERIMENTAL STUDY OF FILLET WELD UNDER-NUREG/CR.2452 RISK METHODOLOGY FOR GEOLOGIC DISPOSAL CUT EFFECTS ON WELDED TUBING STRUCTURES UNDER CEN-OF RADIOACTIVE WASTE. Foal Report TRIC AND ECCENTRIC CYCLIC LOADINGS.

i i

l l

l 1

i l

l l

l I

l

NRC Originating Organization Index (Staff Reports)

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 by sub-sections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number, l

OFFICE OF EXECUTIVE DIRECTOR FOR OPERAflONS (EDO) NRC - NO DETAILED AFFILIATION GIVEN REGION 1, OFFICE OF DIRECTOR NUREG 1251 DRFT FC: IMPLICATIONS OF THE ACCIDENT AT NUREG 0837 V07 N01: NRC TLD DIRECT RADIATION MONITORING CHERNOBYL FOR SAFETY REGULATION OF COMMERCIAL NU-NETWORK Progress Report. January-March 1987. CLEAR POWER PLANTS IN THE UNITED STATES. Draft For Com- '

OFFICE OF ENFORCEMENT (POST 870413) ment' NUREG-0040 V06 NO2: ENFORCEMENT ACTIONS.SIGNIFICANT AC-TIONS RESOLVED.Ouarterly Progress Report,Apni-June 1967. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

OFFICE OF SPECIAL PROJECTS OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR gPOST NUREG-1232 V01: SAFETY EVALUATION REPORT ON TENNESSEE 860720)

VALLEY AUTHORITY, Revised Corporate Nuclear Performance Plan- NUREG-1266 V01: NRC SAFETY RESEARCH IN SUPPORT OF REG-EDO 0"FICE FOR ANALYSIS & EVALUATION OF OPERATIGVAL DIVI I N F EERING (POST 870413)

OFF E FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, DI- PR A P

^

NURE 0090 V09 N04: REPORT TO CONGRESS ON ABNORMAL p OCCURRENCES October December 1986. NUREG/CP 0088: TRANSACTIONS OF THE 9TH INTERNATIONAL NUREG 1275: OPERATING EXPERIENCE FEEDBACK REPORT - CONFERENCE ON STRUCTURAL MECHANICS IN REACTOR NEW PLANTS. Commercial Power Reactors. TECHNOLOGY. Panel Session JK: Structural And Mechanical Engh OFFICE OF INFORMATION RESOURCES MANAGEMENT neenng Research At The U S. Nuclear Regulatory Commission.

DIVISION OF PUBLICATION SERVICES (POST B73413) DIVISION OF REACTOR & PLANT SYSTEMS (POST 870413)

NUREG 0304 V12 NO2: REGULATORY AND TECHNICAL REPORTS NUREG 1204. PROGRAM PLAN FOR CORRECTION OF US IN.

(ABSTRACT INDEX JOURNAL). Compilation For Second Quarter STRUMENT DEGRADATION OR FAILURE IN THE UPPER PLENUM 1987. April-June. TEST FACILITY (UPTF) IN THE FEDERAL REPUBLIC OF GERMA-NUREG 0540 V09 N05: TITLE LIST OF DOCUMENTS MADE PUBLIC. NY.

LY AVAILABLE. May 1 31,1987.

NUREG-0540 V00 N06: TITLE LIST OF DOCUMENTS MADE PUBLIC. OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

LY AVAILABLE. June 1 30,1987.

OFFICE OF NUCLEAR REACTOR REGUL* TION, DIRECTOR (POST NVREG 0750 V24102. INDEXES TO NUCLEAR REGULATORY COM. 870411)

MISSION ISSUANCES July December 1986 NUREG-0327 R04: OWNERS OF NUCLEAR POWER NUREG-0750 V25 N01: NUCLEAR REGULATORY COMMISSION IS. PLANTS. Percentage Ownership Of Commercial Nuclear Power SUANCES FOR JANUARY 1967.Pages 162. Plants By Utihty Companies.

NUREG 1285; NRC STAFF EVALUATION OF THE GENERAL ELEC-NUREG 0750 V25 NO2. NUCLEAR REGULATORY COMMISSION 15 SUANCES FOR FEBRUARY 1987.Pages 6312a. TRIC COMPANY NUCLEAR REACTOR STUDY (" REED REPORT").

NUREG 0750 V25 NO3. NUCLEAR REGULATORY COMMISSION IS. OlVISION OF REACTOR PROJECTS - 1/11 (POST 870411)

SUANCES FOR MARCH 1987 Pages 129 266. NUREG 1047 S06. SAFETY EVALUATION REPORT RELATED TO NUREG 0750 V25 N04 NUCLEAR REGULATORY COMMISSION IS. THE OPERATION OF NINE MILE POINT NUCLEAR STATION, UNIT SUANCES FOR APRIL 1987 Pages 267 416 2. Docket No. 50 410.(Niagara Mohawk Power Corporation.et al)

DIVISION OF RULES & RECORDS (POST 870413) NUREG 1057 $06: SAFETY EVALUATION REPORT RELATED TO NUREG 0936 V06 NO2: NRC REGULATORY AGENDA Ouarwrly THE OPERATION OF BEAVER VALLEY POWER STATION, UNIT Report, April June 1987. 2. Docket No. 50 412.(Duquesne Light Company.et al)

DIVISION OF COMPUTER & TELECOMMUNICATIONS SERVICES NUREG 1253. TECHNICAL SPECIFICATIONS FOR NINE MILE POINT ,

(POST 870413) NUCLEAR STATION, UNIT 2. Docket No. 50-410(Niagara Mohawk <

NUREG 0020 V11 N01: LICENSED OPERATING REACTORS STATUS Power Corporation.et al)

SUMMARY

REPORT. Data As Of December 31,1986.(Gray Book 1) NUREG 1279 TECHNICAL SPECIFICATIONS FOR BEAVER VALLEY POWER STATION, UNIT 2. Docket No. 50 412.(Duquesne Light OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS Company,et al)

OFFICE OF NUCLEAR MATERIAL SAFETY & OIVISION OF REACTOR PROJECTS - Ill.IV.V & SPECIAL PROJECTS SAFEGUARDS, DIRECTOR (POST 870411 NUREG 0430 V07 N02: LICENSED FUEL FACILITY STATUS NUREG 0781 SO4 SAFETY EVALUATION REPORT RELATED TO REPORT. Inventory Ditlerence Data July-December 1986; Gray Book THE OPERATION OF SOUTH TEXAS PROJECT. UNITS 1 AND 8' "

DIVI ON OF SAFEGUARDS & TRANSPORTATION (POST 870413) )

Di NO LO LEVEL WAST MANAGEME & EO S ION- NUREh1002 SO4. SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF BRAIDWOOD STATION, UNITS 1 AND ING (POST 870413) 2 Docket Nos 50 466 And 50-4571 Commonwealth Edisor. Company)

NUREG 1213 RO1: PLANS AND SCHEDULES FOR IMPLEMEN1 A.

NUREG 1185 V01 DRFT: INTEGRATED SAFETY ASSESSMENT TlON OF U S. NUCLEAR REGULATORY COMMISSION RESPONSI- REPORT Integrated Safety Assessment Program.Haddam Neck BILITIES UNDER THE LOW LEVEL RADIOACTIVE WASTE POLICY Plant Docket No 50 213. (Connecticut Yankee Atomic Power Com-AMENDMENTS ACT OF 1985 (P.L. 99-240} pany)

NUREG 1274. REVIEW PROCESS FOR LOW LEVEL RADIOACTIVE NUREG-1185 V02 DRFT: INTEGRATED SAFETY ASSESSMENT WASTE DISPOSAL LICENSE APPLICATION UNDER LOW LEVEL PEPORT. integrated Safety Assessment Program Haddam Neck l RADIOACTIVE WASTE PFIICY AMENDMENTS ACT.

Piant. Docket No 50 213 (Connecticut Yankee Atomic Power Com-C l ON NU E 1255 TECHNICAL SPECIFICATIONS FOR SOUTH TEXAS NUREG-0885 10 6 U.S. NUCLEAR REGULATORY COMMISSION PROJECT. UNIT 1. Docket No 50 498 (Houston Lighting and Power POLICY AND PLANNING GUIDANCE 1987- Company) 41 i

i

42 NRC Origin: ting Organizatlan Ind:x (Staff Rip;rts) j l

NUREG-1276 TECHNICAL SPECIFICATIONS 'FOR BRAIDWOOD DIVISION OF REACTOR INSPECTION & SAFEGUARDS (POST 1 STATION. UNITS 1 AND 2. Docket Nos. 50-456 And 50-457.(Com- 870411) monwealth Edison Company) NUREGM40 V11 NO2: LICENSEE CONTRACTOR AND VENDOR IN. ]

NUREG 1281: EVALUATION OF THE QUALIFICATION OF SPERT SPECTION STATUS REPORT. Quarterty Report,Apnl June FUEL FOR USE IN NON POWER REACTORS. 1987 (White Book)

NUREG 1282: SAFETY EVALUATION REPORT ON HIGH URANtUM NUREG-1194: CONSTRUCTION APPRAISAL TEAM INSPECTION RE.

CONTENT, LOW. ENRICHED URANIUM.ZlRCONIUM HYDRIDE SULTS ON WELDING AND NONDESTRUCTIVE EXAMINATION AC- -

FUELS FOR TRIGA REACTORS. Docket No. 50-163.(GA DIVISl[N OF LICENSEE PERFORMANCE & QUALITY EVALUATION I Technologies. Incorporated) l (POST 870411)

NUREG-1278: VOGTLE UNIT 1 READINESS REVIEW. Assessment Of Georgia Power Company Readiness Remw Pdot Program.

Ti

l NRC Originating Organization Index (International Agreements)

This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each

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entry is followed by a NUREG number and title of the report (s), if further information is needed, refer to the main citation by NUREG number.

I There were no NUREG/lA reports for this quarter, 1

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I NRC Contract Sponsor index (Contractor Reports)

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This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program office)  :

and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) prepared by that organi-zation. If further information is needed, refer to the main citation by the NUREG/CR number.

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) NUREG/CR-4457: AGING OF CLASS 1E BATTERIES IN SAFETY REGION 5. OFFICE OF DIRECTOR SYSTEMS OF NUCLEAR POWER PLANTS.

NUREG/CR4768 VOI: METHODOLOGY AND APPLICATION OF SUR-NUREG/CR 4534: ANALYSIS OF DIFFUSION FLAME TESTS.

ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER NUREG/CR-4577: AUTOMATED LONG TERM SURVEILLANCE OF A PLANT Task 1. Ana eysis Of ANO-1 And Oconte PRAs COMMERCIAL NUCLEAR POWER PLANT, NUREG/CR-4768 V02 METHODOLOGY AND APPLICATION OF SUR-NUREG/CR4583 V03. DEVELOPMENT AND VALIDATION (# A ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER REAL TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT PLANT. Final Report WATER REACTOR COMPONENTS. Annual Report. October 1985 -

EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NU /C 590 V01: AGING OF NUCLEAR STATION DIESEL OFF E OR ANALYSIS & EVALUATION OF OPERATIONAL DATA, DI- N fx nence Pha t NUREG/CR-2000 V06 N6 LICENSEE EVENT REPORT (LER) NUREG/CR-4500 V02: AGING OF NUCLEAR STATION DIESEL COMPILATION.For Month Of June 1987. GENERATORS Evaluation Of Operating And Expert NUREG/CR 2000 V06 N7: LICENSEE EVENT REPORT (LER) Expenence Workshop.

COMPILATION For Month CA July 1987. NUREG/CR4654: RADIONUCLIDES TRANSPORT AS VAPOR NUREG/CR-2000 V06 N8. LICENSEE EVENT REPORT (LER) THROUGH UNSATURATED FRACTURED ROCK.

COMPILATION.For Month Of August 1987. NUREG/CR-4731 V01: RESIDUAL LIFE ASSESSMENT OF MAJOR LIGHT WATER REACTOR COMPONENTS OVERVIEW. Volume 1.

OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/60) NUREG/CR-4744 V01 N2: LONG TERM EMBRITTLEMENT OF CAST DIVISION OF EMERGENCY PREPAREDNESS & ENGINEERING RE. DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual SPONSE (850212 87041 Report. April-September 1986.

NUREG/CR-4952: EXPERIMENTAL STUDY OF FILLET WELD UN. NUREG/CR 4747 V01: AN AGING FAILURE SURVEY OF LIGHT DERCUT EFFECTS ON WELDED TUBING STRUCTURES UNDER WATER REACTOR SAFETY SYSTEMS AND COMPONENTS.

CENTRIC AND ECCENTRIC CYCLIC LOADINGS. NUREG/CR4760: TEST OF 6-IN, THICK PRESSURE VESSELS.Senes

3. Intermediate Test Vessel V-BA.Teanng Behavior Of Low Upper.

OFFICE OF INFORMATION RESOURCES MANAGEMENT Shelf Matenal.

OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT, DI- NUREG/CR4846: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS RECTOR (POST 870413) SAFETY ANALYSIS Phase 2. Final Report.

NUREG/CR 2907 V05 RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR 4850: STEAM GENERATOR GROUP PROJECT. Task 10 -

NUCLEAR POWER PLANTS. Annual Report 1984. Secondary Side Examination. Final Report.

NUREG/CR4851: SEISMICITY 1886-89 IN THE SOUTHEASTERN C O AR MA ER #

SAFETY &

Charleston. South Carolina Earthauake.

NURE CR 4 5 Y OF GEOPHYSICAL TECHNIQUES FOR

^ IN S AT BO GW ER R ACT P ESSU SS L SEGME .

NUkG iM Fp pin IAL ASSURANCE NUREG/CFs4863: SURFACE SPECTROSCOPY OF PRESSURE REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES. VESSEL STEEL FATlGUE FRACTURE SURFACE FILMS FORMED I

G/CR 500 i TH DS FOR EU L S EST N E C 5 THE O BILITY OF ELECTRODEFOSITED TC(IV)

OXIDES OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81) NUREG/CR4870: AN EVALUATION OF THE EFFECTS OF DESIGN OFFICE OF NUCLEAR REGULATOR) RESEARCH, DIRECTOR (POST DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT B60720) BUILDINGS NUREG/CR-2331 V06 N4 SAFETY RESEARCH PROGRAMS SPON- NUREG/CR4886 ANALYSIS OF THE NESDIP2 AND NESDIP3 SORED BY OFFICE OF NUCLEAR REGULATORY RADIAL SHIELD AND CAVITY EXPERIMENTS.

RESEARCH Ouarterly Progress Report. October December 1986. NUREG/CR4892 A STUDY OF THE EFFECTS OF PENETRATION NUREG/CR4216. EXPERIMENTAL RESULTS FOR A 1.8-SCALE FRAMING ON STEEL CONTAINMENT BUCKLING CAPACITY, STEEL MODEL NUCLEAR POWER PLANT CONTAINMENT PRES- -

NUREG/CR-4900 V01: COMPONENT FRAGILITY RESEARCH SURIZED TO F AILURE. PROGRAM Phase i Demonstration Tests Summary Report NUREG/CR-4805 VOI: REACTOR SAFETY RESEARCH SEMIANNU. NUREG/CR 4900 V02: COMPONENT FRAGluTY RESEARCH AL REPORT January June 1986 PROGRAM Phase i Demonstration Tests Appendices.

NUREG/CR4897: LOW.LEVE L WASTE SOURCE TERM NUREG/CR4904 INVESTIGATION OF STEEL CONTAINMENT EVALUATION Review of Published Modehng And Expenmental BUCKLING FROM DYNAMIC LOADS.

Work.And Presentation Of Low-Level Waste Source Term Modeling NUREG/CR4913. ROUND ROBIN PRETEST ANALYSES OF A 1:6-Framework And Prehminary Model Development. SCALE REINFORCED CONCRETE CONTAINMENT MODEL SUB-DIVISION OF ENGINEERING (POST 870413) JECT TO STATIC INTERNAL PRESSURIZATION.

NUREG/CR-2452: RISK METHODOLOGY FOR GEOLOGIC DISPOS. NUREG/CR-4919 FIELD TESTING OF BENTONITE AND CEMENT AL OF RADIOACTIVE WASTE Final Report BOREHOLE PLUGS IN GRANITE.

NUREG/CR 3220 V05. STRUCTURAL INTEGRITY OF WATER REAC. NUREG/CR 4937; INVESTIGATION OF THE MEERS FAULT IN TOR PRESSURE BOUNDARY COMPONENTS Annual Report For SOUTHWESTERN OKLAHOMA. i 1986 NUREG/CR 4943 PREPARATION OF DESIGN SPECIFICATION $ d NUREG/CR 4161 V03 CRITICAL PARAMETERS FOR A HIGH LEVEL AND DESIGN REPORTS FOR PUMPS, VALVES. PIPING.AND PIPING WASTE REPOSITORY. Volums 3 Satt SUPPORTS USED IN SAFETY RELATED PORTIONS OF NUCLEAR NUREG/CR 4219 V04 N1. HEAVY SECTION STEEL TECHNOLOGY POWER PLANTS.

PROGRAM Semiannual Progress Report For October 1986 March NUREG/CR4954 LONG-TERM PERFORMANCE OF SPENT FUEL 1987 WASTE FORMS 45 i

46 NRC Centrcct Spsnser ind2x

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1 NUREG/CR 4956: SYSTEM PERFORMANCE OF HIGH LEVEL NUREG/CR4787: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS l

WASTE PACKAGE COMPONENTS OF A GENERAL ELECTRIC BWR4/ MARK l. Case Study.

NUREG/CR 4973: INTRAPLATE SEISMICITY OUTSIDE OF THE NUREG/CR 4810: EVALUATION OF DIESEL UNAVAILABILITY AND UNITED ST ATES RISK EFFECTIVE SURVEILLANCE TEST INTERVALS.

NUREG/CR4974. INTRAPLATE SEISMICITY IN THE EASTERN NUREG/CR4817: IODINE PARTITION COEFFICIENT MEASURE-UNITED STATES. MENTS AT SIMULATED PWR STEAM GENERATOR NUREG/CR-4975: A REVIEW OF THE RESOLVING POWER OF RE' CONDITIONSintenm Data Report.

FLECTION SEISMOLOGY METHOOS TO DETECT SUBSURFACE NUREG/CR4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK FAULTS AND/OR CHANGES IN LAYER THICKNESS- METHODS INTEGRATION AND EVALUATION PROGRAM i NUREG/CR-5007: PREDICTION AND MITIGATION OF EROSIVE. (RMIEP) Volume 1 Development Of The Data Based Method.

CORROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NU. NUREG/CR4844 DRFT; INTEGRATED RELIABILITY AND RISK

^

NU C D VELOPMENT OF A TESTING AND ANALYSIS fop ^A METHODOLOGY TO DETERMINE THE FUNCTIONAL CONDITION NUREG/CR4931: RESPONSE OF CENTRIFUGAL AND AXIVANE OF SOLENOID OPERATED VALVES- BLOWERS TO LARGE PRESSURE TRANSIENTS

" SMAM & BE SWSCAM NRAM EG/CR 3 US ONCR T KB L W TEM- hCA9ge94E NUREG/CR.4948. DAVIS BESSE UNCERTAINTY STUDY.

NU E / 485 V TH MP C OF EL CLADDING FAILURE "" "^ ^" "

• EVENTS ON OCCUPATIONAL RADIATION EXPOSI/RES AT NU. ^

CLEAR POWER PLANTS Case S PWR Dunn An Out ' P R MEN NUREG/CR4978! THE COOLDOWN ASPECTS OF THE TMI 2 ACCl-h SK T NS G HEf MENA The

S DY ACCIDENT MITIGA, NUREG CR4982: SEVERE ACCIDENTS IN SPENT FUEL POOLS IN NU / 1 SUPPORT OF GENERIC SAFETY ISSUE 82.

TICN SYSTEMS.

NUREG/CR4871: RESULTS FROM THE DCH-1 EXPERIMENT, OlVISION OF REGUI.ATORY APPLICATIONS (POST 870H1)

NUREG/CR4878: SILVER INDIUM CADMlUM CONTROL ROD BE. NUREG/CR4418: DOSE CALCULATION FOR CONTAMINATION OF HAVIOR AND AEROSOL FORMATION IN SEVERE REACTOR AC- THE SKIN USING THE COMPUTER CODE VARSKIN.

CIDENTS, NUREG/CR4884: INTERPRETATION OF BIOASSAY MEASURE-NUREG/CR4891: PROPERTIES OF REACTOR FUEL ROD MATERI- MENTS.

ALS AT HIGH TEMPERATURES Firw! Summary Report Severs NUREG/CR4951: NEPHROTOxfCITY OF URANY'. FLUORIDE AND Core Damage Pr Tests Pr ram. REVERSIBILITY OF RENAL INJURY IN THE RAT.

NUREG/CR 4896: TA!NMEN LOADS DUE TO DIRECT CON. NUREG/CR4958: IMPACT OF PROPOSED FINANCIAL ASSURANCE )

TAINMENT HEATING AND ASSOCIATED HYDROGEN REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES. i BEHAVIOR. Analysis And Calculabons With The CONT AIN Code NUREG/CR4988: RADIATION DOSE ESTIMATES AND HAZARD l NUREG/CR 4905 DETONABILITY OF H2-AIR DILUENT MIXTURES EVALUATIONS FOR INHALED AIRBORNE RADIONUCLIDES Final j NUREG/CR4910: RELAY CHATTER AND OPERATOR RESPONSE Report.

AFTER A LARGE EARTHOUAKE.An improved PRA Methodology DIVISION OF ENGINEERING SAFETY (860720-870413)

With Case Studies NUREG/CA 3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR4925: FISSION PRODUCT BEHAVIOR DURING THE PBF WESTERN OHIO-INDIANA REGION Final Report,0ctober 1981 -

SEVERE FUEL DAMAGE TEST 11. September 1986.

NUREG/CR4953 CORRELATION OF RADIOIODINE RESUSPEN-SiON WITH TEMPERATURE AT TMI 2. OFFICE OF NUCLEAR REACTOR REQULATION (POST 4/28/80)

NUREG/CR4962 METHODS FOR THE ELICITATION AND USE OF OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST EXPERT OPINION IN RISK A3SESSMENT. Phase 1 - A Cntical Eval- 870411) uation And Directons For Future Research, NUREG/CR4981: A SAFETY ASSESSMENT OF THE USE OF GRAPHITE IN NUCLEAR REACTORS LICENSED BY THE U S j DIVISION OF REACTOR & PLANT SYSTEVS (POST 870413)

NUREG/CR4407: PIPE BREAK FREQUENCY ESTIMATION FOR NU- NRC. . {

CLEAR POWEP PLANTS. OlvlSION OF LICENSES PERFORMANCE & OVALITY EVALUATION i NUREG/CR4710. SHUTDOWN DECAY HEAT REMOVAL ANALYSIS (POST 870411) l OF A COMBUSTION ENGINEERING PRESSURIZED WATER NUREG/CR4640: HANDBOOK OF SOFTWARE OUALITY ASSUR- (

REACTOR Case Study. ANCE TECHNIOUES APPLICABLE TO THE NUCLEAR INDUSTRY. j i

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Contractor Index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and i titles of their reports. If further information is needed, refer to the main citation by the )

NUREG/CR number.

I AMES LABORATORY, ENERGY & MINERAL RESOURCES RESEARCH NUREG/CR-4884: INTERPRETATION OF BIOASSAY MEASUREMENTS.

INSTITUTE NUREG/CR-4897: LOW LEVEL WASTE SOURCE TERM NUREG/CR4870: AN EVALUATION OF THE EFFECTS OF DESIGN EVALUATION. Review of Pubhshed Modehng And Expenmental DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT Work,And Presentation Of Low-Level Waste Source Term Modehng BUILDINGS. Framework And Prehminary Model Development.

NUREG/CR-4981: A SAFETY ASSESSMENT OF THE USE OF GRAPH.

ARGONNE NATIONAL LABORATORY ITE IN NUCLEAR REACTORS LICENSED BY THE U.S. NRC.

NUREG/CR4744 V0t N2: LONG TERM EMBRITTLEMENT OF CAST NUREG/CR4982: SEVERE ACCIDENTS IN SPENT FUEL POOLS IN q DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual SUPPORT OF GENERIC SAFETY ISSUE 82.  !

Report.Apni-September 1986. I NUREG/CR4972: TWO-PHASE FLOW REGIME TRANSITION CRITE- BURNS & ROE CO. )

RIA IN POST DRYOUT REGION BASED ON FLOW VISUAll2ATION NUREG/CR4781 DRFT: STLM OF SEVERE ACCIDENT MITIGATION EXPERIMENTS. SYSTEMS.

ARIZONA, UNIV. OF, TUCSON, AZ CALIFORNIA, UNIV. OF, SANT A BA RSAR A, CA l NUREG/CR4654. RADIONUCLIDES TRANSPORT AS VAPOR NUREG/CR-4978: THE COOLDOWN ASPECTS OF THE TMi-2 ACCI- l THROUGH UNSATURATED FRACTURED ROCK- DENT l NUREG/CR 4919. FIELD TESTING OF BENTONITE AND CEMENT BOREHOLE PLUGS IN GRANITE. CALSPAN CORP,(SUBS. ARVIN INDUSTRIES / FRANKLIN RESEARCH BATTELLE MEMORIAL INSTITUTE CEN)

NUREG/CR 4956: SYSTEM PERFORMANCE OF HIGH-LEVEL WASTE NUREG/CR4952: EXPERIMENTAL STUDY OF FILLET WELD UNDER-

' PACKAGE COMPONENTS. CUT EFFECTS ON WELDED TUBING STRUCTURES UNDER CEN.

TRIC AND ECCENTRIC CYCLIC LOADINGS.

BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES NUREG/CR-4954. LONG TERM PERFORMANCE OF SPENT FUEL DANIEL & ASSOCIATES,INC. ,

WASTE FORMS. NUREG/CR-4953: CORRELATION OF RADIOIODINE RESUSPENSION j WITH TEMPERATURE AT TMI-2 SATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORIES EG4G IDAHO, INC. (SUBS. OF EG&G, INC.) ,

NUREG/CR 4418: DOSE CALCULATION FOR CONTAMINATION OF NUREG/CR 4407; PIPE BREAK FREQUENCY ESTIMATION FOR NU.

)

THE SKIN USING THE COMPUTER CODE VARSKIN CLEAR POWER PLANTS. I NUREG/CR 4485 V02: THE IMPACT OF FUEL CLADDING FAILURE NUREG/CR4457: AGING OF CLASS 1E BATTERIES IN SAFETY SYS- j EVENTS ON OCCUPATIONAL RADIATION EXPOSURES AT NUCLE- TEMS OF NUCLEAR POWER PLANTS. q AR POWER PLANTS Case Study PWR Dunng An Outage. NUREG/CR 4731 Vo t: RESIDUAL LIFE ASSESSMEU OF MAJOR NUREG/CR4583 V03: DEVELOPMENT AND VALIDATION OF A REAL. LIGHT WATER REACTOR COMPONENTS - OVERVIEW 3 aturne 1.

TIME SAFT UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CR 4747 V01: AN AGING FAILURE SURVEY OF LIGHT REACTOR COMPONENTS. Annual Report. October 1985 September WATER REACTOR SAFETY SYSTEMS AND COMPONENTS.

1986 NUREG/CR4760: RISK EVALUATIONS OF AGING PHENOMENA.The NUREG/CR 4590 V0t: AGING OF NUCLEAR STATION DIESEL Dnear Aging Rehabihty Model And its Extensions.

GENERATORS Evaluabon Of Operating And Expert Expenence. Phase NUREG/CR4844 DRFT: INTEGRATED RELIABILITY AND RISK ANALY-1 Study. SIS SYSTEM (IRRAS) USER'S GUlOE VERSION 1.0 (DRAFT)

NUREG/CR 4590 V02- AGING OF NUCLEAR STATION DIESEL NUREG/CR4876: SILVER INDIUM CADMlVM CONTROL ROD BEHAV-GENERATORS Evaluation Of Operating And Export IOR AND AEROSOL FORMATION IN SEVERE REACTOR ACCI-Eapenence Workshop DENTS.

NUREG/CR.4640: HANDBOOK OF SOFTWARE QUALITY ASSURANCE NUREG/CR4925: FISSION PRODUCT BEHAVIOR DURING THE PDF TECHNIOUES APPLICABLE TO THE NUCLEAR INDUSTRY- SEVERE FUEL DAMAGE TEST 11.

NUREG/CR-4768 V01: METHODOLOGY AND APPLICATION OF SUR. NUREG/CR4945.

SUMMARY

OF THE SEMISCALE PROGRAM (1965 ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER 1986).

NUR G 768 YE H DOL AP ICATION OF SUR-ROGATE PLANT PRA ANALYSIS TO THE RANCHO SECO POWER ENERGY, INC.

PLANT Final Report NUREG/CR4834 V01: RECOVERY ACTIONS IN PRA FOR THE RISK NUREG/CR4850: STEAM GENERATOR GROUP PROJECT. Task 10

• METHODS INTEGRATION AND EVALUATION PROGRAM E6 Wume helopment a ne DateSaw MeM NU Cke [hE $ES FR CTOR FUEL ROD MATERIALS ]

AT HIGH TEMPERATURES Final Summary Report . Severe Core FUTURE RESOURCES ASSOCIAFES lNC. I NUREG/CR4910: RELAY CHATTER AND CPERATOR RESPONSE j NU E R 8 I PAC F ROPOSED FINANCIAL ASSURANCE ER A NE EARmOWE.An Wprove N WWogy M REQUIREMENTS ON NUCLEAR MATERIALS LICENSEES NUREG/CR 5007. METHODS FOR RECURRING LOSS TESTS. Case Studies.

BROOKHAVE.N NATIONAL LA8 ORATORY CA TECHNOLOGIES,INC./ GENERAL ATOMIC CO.

NUREG/CR-2331 V06 N4: SAFETY RESEARCH PROGRAMS SPON. NUREG/CR4846: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS SORED BY OFFICE OF NUCLEAR REGULATORY bAFETY ANALYSIS Phase 2, Final Report.

RESEARCH Ouarterly Progress %eport. October. December 1986.

NUREG/CR-2907 V05. RADIOACTIVE MATERIALS RELEASED FROM LAMONT DOHERTY GEOLOGICAL OBSERVATORY NUCLEAR POWER PLANTS Annual Report 1984 NUREG/CR 4851: SEISMICITV 1886 89 IN THE SOUTHEASTERN NUREG/CR-4810. EVALUATION OF DIESEL UNAVAILABILITY AND UNITED STATESThe Aftersaock Sequence Of The Charleston. South RISK EFFECTIVE SURVEILLANCE TEST INTERVALS Carohna Earthquake ,

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48 Contrcctor ind3x LAWRENCE BERKELEY LABORATORY OKLAHOMA, UNIV. OF, NORM AN. OK NUREG/CR-4161 V03: CRITICAL PARAMETERS FOR A HIGH LEVEL NUREG/CR4937. INVESTIGATION OF THE MEERS FAULT IN SOUTH-WASTE REPOSITORY. Volume 3 Salt. WESTERN OKLAHOMA.

LAWRENCE LIVERMORE NATIONAL LABORATORY PENTEK. lNC.

NUREG/CR4161 V03: CRiflCAL PARAMETERS FOR A HIGH-LEVEL NUREG/CA.5008. DEVELOPMENT OF A TESTING AND ANALYSIS WASTE REPOSITORY. Volume 3 Salt. METHODOLOGY TO DETERMINE THE FUNCTIONAL CONDIT!ON OF NUREG/CR4900 V01: COMPONENT FRAGILITY RESEARCH SOLENOID OPERATED VALVES.

PROGRAM Phase i Demonstrat on Tests Summary Report.

NUREG/CR4900 V02: COMPONENT FRAGILITY RESEARCH PICKA , & GARRICK INC PROGRAM Phase i Demonstration Tests. Appendices.

EXPERT OPINION IN RISK ASSESSMENT. Phase i . A Cntical Evalue.

LOS ALAMOS NATIONAL LABORATORY ton Ant) Directions For Future Research.

NUREG/CA-4892: A STUDY OF THE EFFECTS OF PENETRATION NUREG/CA.4 4 i ES G OF EEL NTA NME BUCK.

A 5 E ^

p g p HE R LING FROM DYNAMIC LOADSe NUREG/CR4931: RESPONSE OF CENTRIFUGAL AND AXIVANE SANDIA NATIONAL LABORATORIES - I BLOWERS TO LARGE PRESSURE TRANSIENTS. NUREG/CR-2452. RISK METHODOLOGY FOR GEOLOGIC DISPOSAL LOVELACE BIOMED & ENVIRONMENTAL RESEARCH INSTITUTE NURE 3024 US A N C RETE ATTACK BY LOW.

NUREG/CR-4986: RADIATION DOSE ESTIMATES AND HAZARD EVAL- TEMPERATURE. FRAGMENTED CORE DEBRIS.

UATIONS FOR INHALED AIRBORNE RADIONUCLIDES Final Report. NUREG/CR4216: EXPERIMENTAL RESULTS FOR A 18-SCALE STEEL MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE, MA " LEM POWER MT MAMM NNZED NUREG/CR-5007: PREDICTION AND MITIGATION OF EROSIVE COR.

f0 URE NUREG/CR-45'34: ANALYSIS OF DIFFUSION FLAME TESTS.

ROSIVE WEAR IN SECONDARY PIPING SYSTEMS OF NUCLEAR NUREG/CR-4710: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF POWER PLANTS. A COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR Case Study MATERIALS ENGINEERING ASSOCIATES,INC. NUREG/CR-4767: SHUTDOWN DECAY HEAT REMOVAL ANALYSIS OF NUREG/CR-3228 V05: STRUCTURAL INTEGRITY OF WATER REAC- A GENERAL ELECTRIC BWR4/ MARK l. Case Study TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For NUREG/CR4781 DRFT: STUDY OF SEVERE ACCIDENT MITIGATION 1986. SYSTEMS.

NUREG/CR4863: SURFACE SPECTROSCOPY OF PRESSURE NUREG/CR4805 V01: REACTOR SAFETY RESEARCH SEMIANNUAL VESSEL STEEL FATIGUE FRACTURE SURFACE FILMS FORMED IN REPORT. January June 1986.

PWR ENVIRONMENTS. NUREG/CR 4834 v01: RECOVERY ACTIONS IN PRA FOR THE RISK METHODS INTEGRATION AND EVALUATION PROGRAM MICHIGAN, UNIV. OF, ANN ARDOR, MI (RMIEP) Volume 1. Development Of The Data-Based Method. l NUREG/CR-3145 V05: GEOPHYSICAL INVESTIGATIONS OF THE NUREG/CR4846. HIGH LEVEL W ASTE PRECLOSURE SYSTEMS l WESTERN OHIO INDIANA REGION Final Report,0ctober 1981 - Sep. SAFETY ANALYSIS.Phaso 2. Final Report.

tomber 1986. NUREG/CR-4870: AN EVALUATION OF THE EFFECTS OF DESIGN DETAILS ON THE CAPACITY OF LWR STEEL CONTAINMENT OAK RIDGE NATIONAL LABORATORY BUILDINGS.

NUREG/CR-2000 V06 N6; LICENSEE EVENT REPORT (LER) NUREG/CR 4871: RESULTS FROM THE DCH-1 EXPERIMENT.

COMPILATION For Month Of June 1987. NUREG/CR4896: CONTAINMENT LOADS DUE TO DIRECT CONTAIN4 MENT HEATING AND ASSOCIATED HYDROGEN BEHAVIOR. Analysis NUREGICA-2000 V06 N7: LICENSEE EVENT REPORT (LER)

COMPILATION For Month Of July 1987 And Calculations With The CONTAIN Code.

CO PILA O For onth Of Augus 87 U E S A4 NUREG/CR-4219 V04 N1: HEAVY SECTl'ON STEEL TECHNOLOGY SCALE REINFORCED CONCRETE CONTAINMENT MODEL SUBJECT PROGRAM $emiannual Progress Report For October 1986 March TO STATIC INTERNAL PRESSURIZATION.

VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSBURG, NU G/CR 4577: AUTOMATED LONG TERM SURVEILLANCE OF A NU EG R47 F 6 N THI K P'RESSURE VESSELS Senes NUREG/CR4973: INTRAPLATE SEISMICITY OUTSIDE OF THE 3 intermediate Test Ver W V 8A.Teanng Behavior Of Low Upper S5 elf INTRAPLATE SEISMICITY IN THE EASTERN NURF / 49 NUR G/ R 4817: IOP 4 PARTITION COEFFICIENT MEASUREMENTS NUR G/ -4  : A REVIEW OF THE RESOLVING POWER OF RE-AT SIMULATED WR STEAM GENERATOR CONDITIONS.Intenm FLECTION SEISMOLOGY METHODS TO DETECT SUBSURFACE NUR G A /OR CHANGES IN N MNESS-8F FLAW DENSITY EXAMINATIONS OF A CLAD BOIL-ING WATEF .EACTOR PRESSURE VESSEL SEGMENT- WESTON GEOPHYSICAL CORP.

NUREG/CR ,65: THE SOLUBILITY OF ELECTRODEPOSITED TC(IV) NUREG/CR4957: SURVEY OF GEOPHYSICAL TECHNIOUES FOR NURI 44886: ANALYSIS OF THE NESDIP2 AND NESDIP3 RADIAL SHi' J AND CAVITY EXPERIMENTS WYLE LABORATORIES NUFr a/CR4943 PREPARATION OF DESIGN SPECIFICATIONS AND NUREG/CR 4900 V01: COMPONENT FRAGILITY RESEARCH

' . SIGN REPORTS FOR PUMPS. VALVES. PIPING.AND PIPING SUP- PROGRAM Phase i Demonstration Tests Summary Report.

ORTS USED IN SAFETY RELATED PORTIONS OF NUCLEAR NUREG/CR 4900 V02. COMPONENT FRAGILITY RESEARCH POWER PLANTS. PROGRAM Phase i Demonstration Tests. Appendices.

e

International 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/lA number, 1

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There were no NUREG/lA reports for this quart 6r.

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i Licensed Facility index 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.

i S 424 AMn W Vogtie Nuclear Plant. Umt 1, George NUREG-1278  % 213 Haddam Neck Plant Connachcut Yankee Atorme NUREG1185 V02 DRFT PowerCo Power Co 50412 Beavet Valley Power Statrm. Umt 2. Duquesne NUREG1057 S06 50 410 Nme Me Pon11 Nuclear Stabon, Umt 2. Nagara NUREG.1047 S06 L@t Co. Mohawk Power Corp.

50412 Beaver Valley Power Stauon, Umt 2. Duquesne NUREG1279 50 410 Nine Mde Pomt Nuclear Statm Umt 2 Nagara NUREGt253 i Co Mohawk Power Corp STN 50 456 Br Stahon, Urut 1 Commonwealth Esson NUREG-1002 504 Oconee Nuclear Stabon, Umt 2, Duke Power Co NUREG/CR 4768 V01 54270 50 312 Rancho Seco Nuclear Generaung Stabon, NUREG/CR4768 V01 ST450456 Bradwood Stahon, Umt 1, Commonwealth Eeson NUREG1276 Sacramento Mancipal Ubbty Co NURF.G/CR.4768 V02 50 312 Rancho Seco Nuclear Generahng Stanon, STN 50457 Braufwood Staton, Umt 2, Commonweann Esson NUREG1002 SO4 Sacramento Muracipal Upty Ca STN S498 South Texas Profect, Urvt 1. Houston L@bng & NUREG478i SO4

$fN %457 Braufwood Stabon, Unit 2, Commonwealth Eeson NUREG 1276 Co Power Co .

Browns For'y Nuclear Power Stabon, Umt 1, NUREG 1232 V01 STN 50-496 South Texas Prolect Umt 1. Houston Lighting 8 NUREG1255

% 259 Tennessee Valley Authon Power Co.

50 260 Browns Ferry Nuclear Power Staban, Umt 2, NUREGt232 V01 STN 50499 South Texas Project Umt 2 Houston lighang & NUREG4781 SO4 l Tennessee Vaney Autnon Power Co.

50 346 DeeBesse Nuclear Power Stabon, Urut 1, NUREG/CR 4946 54320 Three Mne isiarx! Nuclear Station, Umt 2, NUREG/CR4953 To6edo Esson Co. General Public Utdmes a m Connecteut Yankee Atome 5 V01 DRFT g, [f (, '

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July-September 1987 12 SUPPLt ME NT AR T NOf t B t 3 AS5T R AC T fdfM ivores or 'sur This journal includes all formal reports e NUREG series prepared by the NRC staff and contractors; proceedings of conference. .nd workshops; as well as international agreement reports. The entries in this c ., ation are indexed for access by title and abstract, secondary report number, p sol. author, subject, NRC organization for staff and international agreements, co acto international organization, and licensed facility.

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