ML20065Q675

Regulatory and Technical Reports (Abstract Index Journal). Annual Compilation for 1993

ML20065Q675
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Issue date:	03/31/1994
From:	NRC OFFICE OF ADMINISTRATION (ADM)
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References
NUREG-0304, NUREG-0304-V18-N04, NUREG-304, NUREG-304-V18-N4, NUDOCS 9405050332
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l NUREG-0304 Vol.18,.No. 4 l

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Reguia:ory anc Tecanical Reports (Aastract Incex J~ournaj

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Annual Compilation for 1993 U.S. Nuclear Regulatory Commission Office of Administration 3 9 REC s 0

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9405050332 940331 PDR NUREO PDR 0304 R

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Available from Superintendent of Documents U.S. Government Printing Office i

j Mail Stop SSOP Washington, DC 20402-9328 A year's subscription consists of 4 issues for this publication.

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

NUREG-0304 Vol.18, No. 4 Regulatory and Technical Reports (Abstract Index Journal)

Annual Compilation for 1993 Date Published: March FM Regulatory Publications Ilranch Division of Freedom ofInfornmtion and Publications Services Ollice of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 e*

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l international Agreement Report NUREG /l A-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 numb 0r of the originating organization (if given), and (9) the microfiche i

address (for NRC internal use).

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

ADD

- addendum APP

- appendix DR FT - draft ERR

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

Superintendent of Documents U.S. Govemment 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.

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

Technical Publications Section Regulatory Publications Branch Division of Freedom of Information and Publications Services P-223 U.S. Nuclear Regulatory Commission Washington, D.C. 2055E

-The main citations and abstracts in this compiiation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the followir'g indexes:

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

NRC Originating Organization Index (International Agreements)

NRC Contract Sponsor Index (Contractor Reports)

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

i The bibliographic elements of the main citations are the following:

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

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

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

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

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

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

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

Where the entries are (1) report nur'nber, (2) report title, (3) report authors, (4) organizatic.1al 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),

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CONTENTS Preface.

v Index Tab Main Citations and Abstracts

.... 1

• Staff Reports

• Conference Proceedings

• Contractor Reports

• international Agreement Reports Secondary Report Numbor index.

......2 Personal Author Index 3

Subject index

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

......5 NRC Originating Organization Index (International Agreements).

6 NRC Contract Sponsor index (Contractor Reports).

7 Contractor index.........

8 Internationat organization index.

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l Licensed Facility index..

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Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is.

an NRC staff-originated 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 re3 ort. The bibliographic information (see Preface for details) is followed by a brief abstract of tais report.

NUREG-0020 V17: LICENSED OPERATING REACTORS STATUS to the inspected organizations dunng the period from July

SUMMARY

REPORT. Data As Of December 31, 1992.(Gray through September 1993.

Book I) HARTFIELD.R.A. Division of Computer & Telecommuni-NUREG-0090 V15 N04: REPORT TO CONGRESS ON ABNOR-cations Services (Post 890205). March 1993. 335pp.

9304080036. 74493.001.

MAL OCCURRENCES. October-December 1992.

• Office for The Nuclear Regulatory Commission's annual summary of h.

Analysis & Evaluaton of Operational Data. Director. March censed nuclear power reactor data is based primarily on the 1993. 32pp. 9305250129. 75004:301, report of operating data submitted by licensees for each unit for Section 208 of the Energy Reorganization Act of 1974 ldenti-fees an abnormal occurrence as an unscheduled incident or the month of December because that report contains data for the month of December, the year to date (in this case calendar event that the Nuclear Regulatory Commission determines to be year 1992) and cumulative data, usually from the date of com-8'gnificant from the standpoint of public health and safety and rnercial operation. The data is not independently venfied, but requires a quarterly report of such events to be made to Con-various computer checks are made. The report is divided into gress This report covers the pered October through December 1992. There were two abnormal occurrences at nuclear power two sections. The first contains summary highlights and the second contains data on each individual unit in commercial op.

plants. Six abnormal occurrences involving medical midadminis-ersten Secton 1 capacity and avadabihty factors are simple trations (all therapeutic) at NRC-licensed facihties are discussed anthmetic averages Section 2 stems in the cumulativo column in this report. No abnormal occurrences were reported by are generally as reported by the hcensee and notes as to the NRC's Agreement States. The report also contains information use of weighted averages and starting dates other than com-updating previously reported abnormal occurrences.

mercial operation are provided NUREG-0090 V16 N01: REPORT TO CONGRESS ON ABNOR.

I MAL OCCURRENCES. January-March 1993.

• Office for Analy-NUREG-0040 V16 N04: LICENSEE CONTRACTOR AND sis & Evaluation of Operational Data. Director. June 1993.30pp.

VENDOR INSPECTION STATUS REPORT. Quarterly 9307220177. 75743:332.

Report, October December 1992 (White Book)

• Division of Re-Section 208 of the Energy Reorganizaton Act of 1974 identi-actor inspecton & Licensee Performance (Post 921004).Janu.

fies an abnormal occurrence as an unscheduled incident or ary 1993.198pp 9302230396. 64961:001.

event that the Nuclear Regulatory Commission determines to be I

This penodical covers the results of inspectens performed by significant from the standpoint of public health and safety and the NRC's Vendor inspection Branch that have been distributed requires a quarterly report of such events to be made to Con-to the inspected organizations dunng the penod from October gress. This report covers the penod January through March through December 1992.

1993. There is one abnormal occurrence at a nuclear power NUREG-0040 V17 N01: LICENSEE CONTRACTOR AND plant discussed in this report that involved a steam generator VENDOR INSPECTION STATUS REPORT. Quarterly tube rupture at Palo Verde Unit 2, and none for fuel cycle facih-Report, January-March 1993 (White Book)

• Division of Reactor ties. Three abnormal occurrences involving medical misadminis-Inspection & Licensee Performance (Post 921004). May 1993, tratons (two therapeutic and one diagnostic) at NRC-heensed 226pp. 9306180278. 75388:204.

facihties are also discussed in this report. No abnormal occur.

This penodical covers the resuits of inspections performed by rences were reported by NRC's Agreement States. The report the NRC's Vendor inspection Branch that have been distnbuted also contains information updating previously reported abnormal to the inspected organnations during the penod from January occurrences.

through March 1993' NUREG-0090 V16 NO2: REPORT TO CONGRESS ON ABNOR-NUREG-0040 V17 NO2: LICENSEE CONTRACTOR AND MAL OCCURRENCES. April June 1993.

• Office for Anatysis &

VENDOR INSPECTION STATUS REPORT. Quarterly Evaluation of Operational Data, Director. September 1993.

Report. April-June 1993.(White Book)

• Divison of Reactor in.

31pp. 9311010015. 76982.001.

spection & Licensee Performance (Post 921004). August 1993.

Section 208 of the Energy Reorganization Act of 1974 identi.

I'os an abnormal occurrence as an unscheduled incident or 109pp. 9309210034. 76483 049.

This penodical covers the results of inspections performed by event that the Nuclear Regulatory Commission determines to be the NRC's Vendor inspection Branch that have been distnbuted significant from the standpoint of public health and safety and to the irispected organizations during the perod from Apnl requires a quarterly report of such events to be made to Con-through June 1993.

gress. This report covers the period Apnl through June 1993, and discusses four abnormal occurrences at NRC-licensed fa-NUREG-0040 V17 NO3: LICENSEE CONTRACTOR.AND cilities, three involving medical brachytherapy misadministrations VENDOR INSPECTION STATUS REPORT. Quarterly and one involving a research reactor that operated without a Report. July September 1993.(White Book)

• Divison of Reactor safety system. One pool irradiation facihty contaminaton event, inspection & Licensee Performance (Post 921004). November two medical misadministrations (one " sodium iodide" and one 1993. 57pp 9312220138. 77543:150.

brachytherapy), and one industrial radiographer overexposure This perodical covers the results of inspectons performed by event that were reported by NRC Agreement States are also the NRC's Vendor Inspecton Branch that have been distnbuted discussed. The report also contains information updating one 1

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2 Main Citations and Abstracts previously reported abnormal occurrence and information on NUREG-0383 V03 R13: DIRECTORY OF CERTIFICATES OF three other events of interest COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC Approved Quality Assurance Pro-NUREG 0304 V17 N04: REGULATORY AND TECHNICAL RE-grams For Radioactive Materials Packages.

• Division of Indus-PORTS (ABSTRACT INDEX JOURNAL) Annual Compilation tnal & Medical Nuclear Safety (Post 870729). October 1993.

For 1992.

• Division of Freedom of Information & Pubhcations 146pp. 9311190096. 77266:246.

Services (Post 890205). February 1993.138pp. 9303240099.

See NUREG-0383 V01,R16 abstract.

74344:150.

NUREG-0386 D06 ROS: UNITED STATES NUCLEAR REGULA-This journal includes all formal reports in the NUREG series TORY COMMISSION STAFF PRACTICE AND PROCEDURE prepared by the NRC staff and contractors, proceedings of con-forences and workshops, grants, and international agreement DIGEST. Commission, Appeal Board And Licensing Board reports. The entnes in this compilation are indexed for access Decisions. July 1972 March 1992.

• Office of the General by title and abstract, secondary report number, personal author, Counsel (Post 860701). February 1993. 602pp. 9303110316.

74215 0 subject, NRC organization for staff and international agree-T ments, contractor, international organization, and licensed facili-Procedure Digest contains a digest of a number of Commission, Atomic Safety and Licensing Appeal Board and Atomic Safety NUREG-0304 V18 N01: REGULATORY AND TECHNICAL RE-and Licensing Board decisions issued dunng the period of July PORTS (ABSTRACT INDEX JOURNAL). Compilation For First 1,1972 to March 31, 1992, interpreting the NRC's Rules of Quarter 1993, January-March.

• Division of Freedom of Informa-Practice in 10 CFR Part 2.

tion & Publications Services (Post 890205). May 1993. 48pp.

NUREG-0386 D06 R06: UNITED STATES NUCLEAR REGULA-9306110059. 75338:252-TORY COMMISSION STAFF PRACTICE AND PROCEDURE See NUREG-0304,V17,N04 abstract.

DIGEST, Commission Appeal Board And Licensing Board Decisions. July 1972 - June 1992.

• Office of the General Coun-NUREG-C304 V18 NO2: REGULATORY AND TECHNICAL RE-sel (Post 860701). May 1993. 600pp. 9306010258. 75055:001.

PORTS (ABSTRACT INDEX JOURNAL). Compilation For This 6th revision of the sixth edition of the NRC Practice and Second Quarter 1993,Apnt-June.

• Division of Freedom of Infor-Procedure Digest contains a digest of a number of Commission, mation & Publications Services (Post 890205). August 1993-Atomic Safety and Licensing Appeal Board, and Atomic Safety 48pp. 9309210040. 76483:158.

and Licensing Board decisions issued dunng the penod of July See NUREG-0304,V17,N04 abstract.

1,1972 to June 30,1992, interpreting the NRC's Rules of Prac-i NUREG-0304 V18 NO3: REGULATORY AND TECHNICAL RE-PORTS (ABSTRACT INDEX JOURNAL). Compilation For Third NUREG 0386 D06 R07: UNITED STATES NUCLEAR REGULA-Quarter 1993, July-September.

• Division of Freedom of informa-TORY COMMISSION STAFF PRACTICE AND PROCEDURE tion & Publications Services (Post 890205). November 1993.

DIGEST. Commission. Appeal Board And Licensing Board 49pp. 9312160335, 77511:249.

Decisions. July 1972 - September 1992.

• Office of the General See NUREG-0304,V17,N04 abstract.

Couneel (Post 860701). August 1993. 500pp. 9309090041.

76379:138.

NUREG-0325 R16: U S. NUCLEAR REGULATORY COMMISSION This 7th revision of the sixth edition of the NRC Practice and FUNCTIONAL ORGANIZATION CHARTS. March 15, 1993.

• Procedure Digest contains a digest of a number of Commission, Ofc of Personnel (Post 870413). March 1993. 66pp.

Atomic Safety and Licensing Appeal Board, and Atomic Safety 9304060302. 74467:001.

and Licensing Board decisions issued dunng the period of July Functional organization charts for the U.S. Nuclear Regulatory 1.1972 to September 30,1992, Interpreting the NRC's Rules of Commission offices. divisions, and branches are presented.

Practice in 10 CFR Pet 2.

NUREG-0430 V12: LICENSED FUEL FACILITY STATUS i

NUREG-0383 V01 R16: DIRECTORY OF CERTIFICATES OF REPORT. inventory Difference Data. July 1,1991 June 30, COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC Approved Packages.

• Division of Wpay Bod m RN WW Mce d Nu&ar Mate-hf85 29 Industrial & Medical Nuclear Safety (Post 870729). October 1993. 491pp. 9311100081,77265:114.

NRC is committed to the periodic publication of licensed fuel This directory contains a Report of NRC Approved Packages facilities inventory difference data, following agency review of (Volume si), Certacates of Compliance (Volume 2), and a the information and completion of any related NRC investiga-Report of NRC Approved Quality Assurance Programs for Ra' tions, information in this report includes inventory d:fference dioactive Materials Packages (Volume 3). The purpose of this data for active fuel fabrication fact! sties possessing more than directory is to make available a convenient source of informa-one effective kilogram of high enriched uranium, low ennched tion on Quality Assurance Programs and Packagings which uranium, plutonium, or uranium-233.

have been approved by the U.S. Nuclear Regulatory Commis-sion. Shipments of radioactive material utiltzing these packag-NUREG-0525 V02 R01: SAFEGUARDS

SUMMARY

EVENT LIST ings must be in accordance with the proWsions of 49 CFR Part (SSEL). January 1,

1990 Through December 31, 1992.

173.471 and 10 CFR Part 71, as applicable. In satisfying the re-FADDEN.M.; YARDUMIAN.J. Operations Branch.. July 1993.

quirements of Section 71.12, it is the responsibility of the licens-250pp. 9308160105. 76142:001.

ees to insure themselves that they have a copy of the current The Safeguards Summary Event List provides brief summa-approval and conduct their transportation activities in accord-ries of hundreds of safeguards-related events involving nuclear ance with an NRC approved quality assurance program.

material or facilities regulated by the U.S. Nuclear Regulatory Commission. Events are described under the categories: Bomb.

NUREG-0383 V02 R16: DIRECTORY OF CERTIFICATES OF related, Intrusion, Missing / Allegedly Stolen, Transportation-relat-COMPLIANCE FOR RADIOACTIVE MATERIALS ed, Tampering / Vandalism, Arson, Firearms-reiated. Radiological PACKAGES Certdicates Of Compliance.

• Division of industnal Sabotage, Non-radiological Sabotage, and Miscellaneous. 8%

& Medical Nucioar Safety (Post 870729). October 1993.590pp.

cause of the public interest, the Miscellaneous category also in-9311190087. 77267:032-cludes events reported involving source matenal, byprcduct ma-See NUREG-0383,V01.R16 abstract.

terial, and natural uranium, which are exempt from safeguards

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Main Citations and Abstracts 3

requirements. Information in the event desenptions was ob-NUREG-0540 V15 NIO: TITLE LIST OF DOCUMENTS MADE tained from official NRC sourcos.

PUBLICLY AVAILABLE. October 1 31, 1993. ' Division of Freo.

NUREG-0540 V14 N11: TITLE LIST OF DOCUMENTS MADE dom of inf rmaton & Pubhcatons Sorvices (Post 890205). Do-cember 1993. 274pp. 9401060244. 77687.001.

PUBLICLY AVAILABLE. November 1 30, 1992.

• Division of Freedom of Informaton & Publicatons Services (Post 890205)

Seu NUREG 0540,V14,N11 abstract.

January 1993. 333pp. 9302230138. 64980 001.

This document is a monthly publication containing desenp-NUREG-0713 V12: OCCUPATIONAL RADIATION EXPOSURE AT tions of information received and generated by the U.S. Nuclear COMMERCIAL NUCLEAR POWER REACTORS AND OTHER Regulatory Commission (NRC) This information includos (1)

FACILITIES,1990. Twenty Third Annual Report. RADDATZ,C,T.

docketod material associated with civihan nuclear power plants Division of Regulatory Applications (Post 870413).

and other uses of radioactive matenals, and (2) nondocketed HAGEMEYER,0. Science Apphcatens !ntornational Corp. (for-matonal recorvod and generated by NRC portinont to its role as merly Scienco Apphcations, Inc ).

Jacary 1993. 294pp.

a regulatory agency. The following indexes are included: Por.

9302020464 64729 203.

I sonal Author, Corporate Source, Report Number, and Cross This report summan2es the occupational radiatica exposure Reference of Enclosures to Poncipal Documents.

Information that has boon reported to the NRC's Radiaton Ex-posur Int rmation Reporting System (REIRS) by nuclear power NUREG-0540 V14 N12: TITLE LIST OF DOCUMENTS MADE fach and cMain oh cage d E bnms Mng PUBLICLY AVAILABLE December 1-31, 1992.

• Division of the years 1969 through 1990. The bulk of tho data presented in Freedom of information & Pubhcations Services (Post 890205).

February 1993. 343pp 9303110330. 74213,085.

the report was obtainod from annual radiation exposure reports

]

Soo NUREG 0540,V14,N11 abstract.

submittod in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants.

NUREG4540 V15 NO1: TITLE LIST OF DOCUMENTS MADE Data on workers terminating their employment at certain NRC PUBLICLY AVAILABLE. January 1-31,1993

• Division of Freo-licensed facihties were obtained from reports submitted pursu-dom of Information & Pubhcations Services (Post 890205).

ant to 10 CFR 20.408. The 1990 annual reports subm;tted by March 1993. 320pp. 9304020324 '74448.001-about 443 licensees indicated that approximately 214,568 indi-Seo NUREG-0540,V14.N11 abstract.

viduals were monitored, 110.204 of whom were monitored by NUREG-0540 V15 NO2: TITLE LIST OF DOCUMENTS MADE nuclear power facilities They incuned an averago individual PUBLICLY AVAILABLE February 1 28, 1993.

• Division of Free.

dose of 019 rom (cSv) and an average measurable dose of dom of information & Pubhcations Services (Post 890205). Apnl about 0.36 (cSv). Termination radiation exposure reports wore 1993. 340pp. 9304300306. 74773:005.

analyzed to revoat that about 113,361 individuals completed See NUREG-0540,V14,N11 abstract.

their employment with one or more of the 443 covered bcens-l oes dunng 1990. Some 77,633 of these individuals terminated NUREG-0540 V15 NO3: TITLE LIST OF DOCUMENTS MADE from power reactor facilities, and about 11,083 of them were PUBilCLY AVAILABLE March 1-31, 1993.

• Davmion of Free-considered to be transient workers who received an average dom of Information & Publications Services (Post 890205) May doso of 0~67 rem (cSv)'

1993. 400pp 9306010264. 75057.001.

See NUREG 0540,V14,N11 abstract.

NUREG-0713 V13: OCCUPATIONAL RADIATION EXPOSURE AT NUREG-0540 V15 N04: TITLE LIST OF DOCUMENTS MADE COMMERCIAL NUCLEAR POWER REACTORS AND OTHER PUBLICLY AVAILABLE.Apnl 1 30, 1993.

• Division of Freedom FACILIT ES,1991,Twonty-Fourth Annual Report. RADDATZ,C.T.

of Inforrpation & Pubhcations Services (Post 890205) June Division of Rogulatory Applications (Post 870413).

1993 350pp. 9306290175. 75499 053 HAGEMEYER.D. Science Apphcations international Corp. (for-See NUREG 0540,V14, Nil abstract merly Science Applications, Inc ).

July 1993. 300pp.

9 08160% 761 M77.

NUREG-0540 V15 N05: TITLE LIST OF DOCUMENTS MADE This report summarizos the occupational radiation exposure PUBLICLY AVAILABLE.May 1-31, 1993.

• Division of Freedorn of informat on & Pubhcations Services (Post 890205) July 1993.

information that has been reported to the NRC's Radiation Ex-350pp. 9308160099. 76115 010.

posure Informaton Roporting System (REIRS) by nuclear power See NUREG-0540,V14,N11 abstract.

facihties and certain other categones of NRC licensees dunng the years 1969 through 1991, The bulk of the data presentod in NUREG-0540 V15 N06: TITLE LIST OF DOCUMENTS MADE the report was obtained from annual radiation exposuro reports i

PUBLICLY AVAILABLE June 1 30, 1993.

• Division of Freedom submitted in accordance with the requirements of 10CFR20.407 of Informaton & Pubhcatons Services (Post 890205). August and the technical specifications of nuclear power plants. Data 1993. 421pp. 9309030160. 76325:158-on workers terminating their employment at cortain NRC li-See NUREG-0540,V14,N11 abstract consed facihties wef e obtained from reports submitted pursuant NUREG-0540 V15 N07: TITLE LIST OF DOCUMENTS MADE to 10CFR20.408. The 1991 annual reports submittod by about PUBLICLY AVAILADLE. July 1-31, 1993.

• Division of Froodom 436 licenseos indicated that approirimately 206,732 individuals of Information & Pubhcations Services (Post 890205). Septem.

were monitored, 182,334 of whom were monitored by nuclear bor 1993. 350pp. 9309210229 76485 077.

power facilities. They incurred an averago individual dose of See NUREG-0540,V14.N11 abstract.

0.15 rom (cSv) and an average measurable dose of about 0.31 I' ^

'"'"*'*" '" " *" " P** " N * * " " " "

NUREG-0540 Vib N08: TITLE LIST OF DOCUMENTS MADE reveal that about 96.231 individuals completed their employ-PUBLICLY AVAILABLE August 1-31, 1993

• Division of Free-dom of Information & Pubhcations Services (Post fi90205). Oc-ment with one or more of the 436 covered licensees dunng tober 1993. 400pp. 9311080070. 77068 001.

1991. Some 68.115 of those individuals terminated from power See NUREG.0540,V14.N11 abstract-reactor facihties, and at,out 7,763 of them were considered to be transient workers who received an average doso of 0.52 rom NUREG-0540 V15 N09: TITLE UST OF DOCUMENTS MADE (cSv)

PUBLICLY AVAILABLE.Soptember 1-30, 1993.

• Division of Freedom of Information & Pubhcatons Services (Post 890205).

NUREG-0713 V14: OCCUPATIONAL RADIATION EXPOSURE AT November 1993. 365pp. 9312160329 77510.001.

COMMERCIAL NUCLEAR POWER REACTORS AND OTHER Soo NUREG 0540N14.N11 abstract

4 Main Citations and Abstracts FACILITIES 1992. Twenty-Fifth Annual Report. RADDATZ,C.T.

See NUREG 0750,V36,N01 abstract.

Dwision of Regulatory Apphcations (Post 870413).

HAGEMEYER D. Science Applications International Corp. (for.

NUREG-0750 V36 N04: NUCLEAR FIEGULATORY COMMISSION morfy Science Applications, Inc) December 1993. 300pp.

ISSUANCES FOR OCTOBER 1992. Pages 221249.

• Dvision l

9401120298. 77769 001.

of Freedom of Information & Publications Services (Post l

This report summan7es the occupational radiation exposure 890205). February 1993. 34pp. 9303120069. 74214:298.

information that has been reported to the NRC's Radiation Ex-See NUREG-0750,V36,N01 abstract.

posure Information Reporting System (REIRS) by nuclear power a

facilities and certain other categories of NRC licensees dunng NUREG-0750 V36 N05: NUCLEAR REGULATORY COMMISSION the years 1969 through 1992. The bulk of the data presented in ISSUANCES FOR NOVEMBER 1992. Pages 251-350.

• Dwision the report was obtained from annual radiation exposure reports of Freedom of Information & Publications Services (Post submittod in accordance with the requirements of 10 CFR 890205). March 1993.109pp. 9303300160. 74408.001, 20.407 and the technical specifications of nuclear power plants.

See NUREG-0750,V36,N01 abstract.

Data on workers terminating their employment at certain NRC licersed facilities were obtained from reporis submitted pursu-NUREG-0750 V36 N06: NUCLEAR REGULATORY COMMISSION i

ant to 10 CFR 20.408. Tho 1992 annual reports submitted by ISSUANCES FOR DECEMBER 1992. Pages 351-396.

• Division about 364 licensees indicated that approximately 204,365 indi.

of Freedom of information & Publications Services (Post viduals were monitored, 183.927 of whom were monitored by 890205) March 1993. 53pp. 9304060309. 74467.069.

nuclear power facilities. They incurred an average individual See NUREG.0750,V36,N01 abstract.

dose of 0.16 rem (cSv) and an average measurable dose of NUREG-0750 V37101: INDEXES TO NUCLEAR REGULATORY about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 indwiduals completed their COMMISSION ISSUANCES. January-March 1993.

• Division of employment with one or more of the 364 covered hcensees Freedom of Information & Publications Services (Post 890205).

dunng 1992. Some 71,846 of these individuals terminated from July 1993. 55pp. 9308160096. 76122.200.

power reactor facihties, and about 9,724 of them were consid-See NUREG.0750,V36.101 abstract, e

to ansient workers who received an average dose of NUREG-0750 V37102: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES. January 4une 1993.

• Division of NUREG-0725 R09: PUBLIC INFORMATION CIRCULAR FOR Freedom of Information & Publications Services (Post 890205).

SHIPMENTS OF IRRADIATED REACTOR FUEL.

• Dwision of November 1993. 77pp. 9312160325. 77505:001.

Safeguards & Transportation (870413-930206). March 1993.

See NUREG-0750.V36,101 abstract.

37pp. 93d4190142. 74624:305.

This circular has been prepared to provide information on the NUREG 0750 V37 N01: NUCLEAR REGULATORY COMMISSION l

shipment of irradiated reactor fuel (spent fuel) subject to regula-ISSUANCES FOR JANUARY 1993. Pages 154.

• Division of tion by the Nuclear Regulatory Commission (NRC), and to meet Freedom of information & Publications Services (Post 800205).

the requirennents of Public Law 96-295 The report provides a March 1993. 62pp. 9304080072. 74499.001.

bnef desenption of NRC authonty for certain aspects of trans-See NUREG-0750,V36,N01 abstract.

porting spent fuel. It provides descriptwo statistics on spent fuel shipments regulated by the NRC from 1979 to 1992. It also lists NUREG-0750 V37 NO2: NUCLEAR REGULATORY COMMISSION detailed highway and railway segments used within each state ISSUANCES FOR FEBRUARY 1993. Pages 55-134.

• Dwision trom October 1,1987 through December 31,1992.

of Freedom of Information & Publications Services (Post 89 205) April 1993. 85pp. 9305250140. 75004:213.

NUREG-0750 V36101: INDEXES TO NUCLEAR REGULATORY ee 50,W W abskan COMMISSION ISSUANCES. July September 1992.

• Dwision of Freedom of Information & Publications Services (Post 890205)-

NUREG-0750 V37 NO3: NUCLEAR REGULATORY COMMISSION Apnl 1993 52pp. 930525014 7. 75005.052.

ISSUANCES FOR MARCH 1993. Pages 135-249.

• Division of Digests and indexes for issuances of the Commission, the Freedom of information & Publications Services (Post 890205).

Atomic Safety and Licensing Board Panel, the Administrative May 1993.150pp. 9306210229. 75403.016.

Law Judges, the Directors' Decisions. and the Denials of Pet" See NUREG-07a0,V36.N01 abstract.

tions for Rulemaking are presented.

NUREG-0750 V36102: INDEXES TO NUCLEAR REGULATORY NUREG-0750 V37 N04: NUCLEAR REGULATORY COMMISSION COMMISSION ISSUANCES. July-December 1992.

• Division of ISSUANCES FOR APRIL 1993.Pages 251-354.

• Division of Freedom of information & Pubhcations Services (Post 890205).

Freedom of information & Publications Services (Post 890205).

June 1993. 73pp. 9307060044. 75563 001, July 1993.110pp. 9308160091. 76119:007.

See NUREG-0750,V36,101 abstract.

See NUREG-0750,V36.N01 abstract.

NUREG-0750 V36 N01: NUCLEAR REGULATORY COMMISSION NUREG-0750 V37 N05: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JULY 1992 Pages 1-45.

• Dwision of Free-ISSUANCES FOR MAY 1993.Pages 355-418,

• Division of dom of information & Pubhcations Services (Post 890205) Jan-Freedom of information & Publications Services (Post 890205).

uary 1993 45pp. 9302230178. 64977:308-August 1993. 69pp. 9308190001. 76152.038.

Legalissuances of the Commission, the Atomic Safety and Li-See NUREG-0750,V36,N01 abstract.

consing Board Panel, the Administratue Law Judges, and NRC Program Offices are presented, NUREG 0750 V37 N06: NUCLEAR REGULATORY COMMISSION NUREG-0750 V36 NO2: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JUNE 1993.Pages 419-515.

• Division of Freedom of information & PubHeations Services (Post 890205).

ISSUANCES FOR AUGUST 1992. Pages 47148.

• Division of Freedom of information & Pubhcatons Services (Post 890205).

August 1993.103pp. 9309210012. 76481:034 See NUREG-0750,V36,N01 abstract.

February 1993.110pp. 9302230128. 64978 220.

See NUREG-0750,V36,N01 abstract.

NUREG 0750 V38 N01: NUCLEAR REGULATORY COMMISS!ON NUREG-0750 V36 NO3: NUCLEAR REGULATOR'r COMMISSION ISSUANCES FOR JULY 1993 Pages 124.

• Dwision of Free-ISSUANCES FOR SEPTEMBER 1992. Pages 149-220.

• Dwi-dom of Information & Pubhcations Services (Post 890205). Sep-sion of Freedom of Information & Pubhcations Services (Post tomber 1993. 32pp. 9310120315. 76741:114.

890205). February 1993. 77pp. 9303090074. 74138:142.

See NUREG-0750,V36,N01 abstract.

l

_-,m.

4 1

Main Citations and Abstracts 5

NUREG-0750 V38 N02: NUCLEAR REGULATORY COMMISSION NUREG-0837 V13 N02: NRC TLD DIRECT RADIATION MONI-ISSUANCES FOR AUGUST 1993. Pages 25-79.

• Division of TORING NETWORK. Progress Report. April-June 1993.

Freedorn of Information & Publications Services (Post 890205).

STRUCKMEYER,R.; MCNAMARA,N. Region 1 (Post 820201).

November 1993.62pp.9311180066.77231:199.

August 1993. 250pp. 9309210028. 76482:103.

See NUREG-0750,V36,N01 abstract.

This report provides the status and results of the NRC Ther-moluminescent Dosimeter (TLD) Direct Radiation Monitoring NUREG-0750 V38 N03: NUCLEAR REGULATORY COMMISSION Network. It presents the radiation levels measured in the vicinity ISSUANCES FOR SEPTEMBER 1993. Pages 81168,

• Division of NRC hcensed faci lities throughout the country for the second of Freedom of information & Publications Services (Post quarter of 1993.

890205). November 1993. 98pp. 9312160311, 77505:100.

See NUREG 0750,V36,N01 abstract.

NUREG-0837 V13 NO3: NRC TLD DIRECT RADIATION MON!-

l TORING NETWORK. Progress Report. July-September 1993.

NUREG-0750 V38 N04: NUCLEAR REGULATORY COMMISSION STRUCKMEYER,R. Region 1 (Post 820201). November 1993.

ISSUANCES FOR OCTOBER 1993. Pages 169-186.

• Drvision 234pp. 9312160321. 77511:009, of Freedom of Information & Publications Services (Post This report provides the status and results of the NRC Ther-890205). December 1993. 24pp 9401030180. 77642:309.

moluminescent Dosimeter (TLD) Direct Radiation Monitonng Soo NUREG-0750,V36,N01 abtract.

Network, it presents the radiation levels measured in the vicinity of NRC licensed facihties throughout the country for the third NUREG 0797 S26: SAFETY EVALUATION REPORT RELATED quarter of 1993.

TO THEl OPERATION OF COMANCHE PEAK STEAM ELEC-TRIC STATION, UNIT 2. Docket No. 50-446.(Texas Utilities Elec-NUREG-0847 S11: SAFETY EVALUATION REPORT RELATED tric Company et al)

• Division of Reactor Projects - lit.IV,V (Post TO THE OPERATION OF WATTS BAR NUCLEAR 901216). February 1993. 230pp. 9303110338. 74214:068.

PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Ten-Supplement 26 to the Safety Evaluation Report related to the noseo Valley Authonty)

• Division of Reactor Projects - 1/11 (Post operation of the Comanche Peak Steam Electric Station 870411). April 1993. 50pp. 9305250123. 75005.001.

(CPSES), Unit 2, has been prepared by the Office of Nuclear Supplement No.11 to the Safety Evaluation Report for the Reactor Regulation of the U.S. Nuclear Regulatory Commission application filed by the Tennessee Valley Authority for license to (NRC). The facility is located in Somervell County, Texas, ap.

operate Watts Bar Nuclear Plant, Units 1 and 2. Docket Nos.

proximately 40 miles southwest of Fort Worth, Texas. This sup_

50 390 and 50-391, located in Rhea County, Tennessee, has piement reports the status of certain issues that had not been been prepared by the Office of Nuclear Reactor Regulation of resolved when the Safety Evaluation Report and Supplements the Nuclear Regulatory Commission. The purpose of this sup.

1, 2, 3, 4, 6,12, 21, 22, 23, 24, and 25 to that report were pub-plement is to update the Safety Evaluation of: (1) additional in-hshed This supplement deals primarily with Unit 2 issues; how, formation submitted by the applicant since Supplement No.10 ever, it also references evaluations for several licensing issues was issued; and (2) matters,that the staff had under review that relate to Unit 1, which have been resolved since Supple-when Supplement No.10 was issued.

ment 25 was issued.

NUREG-0847 S12: SAFETY EVALUATION REPORT RELATED NUREG-0797 $27: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-I 8'

nesee a M

sen of eacW Rom -

TRIC STATION, UNil 2. Docket No. 50-446.(Texas Utilities Elec-( $t 870411). October 1993. 60pp. 9311150028.

Inc Company et al.)

• Division of Reactor Projects - lit,lV,V (Post 7 221 1'

001216). Apn11993. 49pp. 9305100006. 74859 074.

Supplement No. 27 to the Safety Evaluation Report related to Supplemont No.12 to the Safety Evaluation Report for the the operation of the Comancho Peak Steam Electric Staton' application filed by the Tennessee Valley Authonty for license to Unit 2, has been prepared by the Offe e of Nuclear Reactor operate Watts Bar Nuclear Plant, Units 1 and 2 Docket Nos.

Regulation of the U.S Nuclear Regulatory Commission. The fa-50-390 and 50-391, located in Rhea County Tennessee, has cihty is located in Somervell County, Texas, approximafely 40 been prepared by the Office of Nuclear Reactor Regulation of miles southwest of Fort Worth, Texas. This supplement reports the Nuclear Regulatory Commission. The purpose of this sup-the status of certain issues that had not been resolved when plement is to update the Safety Evaluation of (1) additional in-formation submitted by the applicant since Supplement No.11 the Safety Evaluation Report and Supplements 1,2,3,4,6,12, 21, 22, 23, 24, 25, and 26 to that report were published This was issued, and (2) matters that the staff had under review supplement deals primarily with Unit 2 issues.

when Supplement No.11 was issued.

NUREG-0910 R02 S01: NRC COMPREHENSIVE RECORDS DIS-NUREG-0837 V12 N04: NRC TLD DIRECT RADIATION MONI-POSITION SCHEDULE.

• Division of information Support Serv-TORING NETNORK. Progress Report. October December 1992-ices (Post 890205). September 1993. 89pp. 9310120062.

STRUCKMEYER,H ; MCNAMARA,N. Region 1 (Post 820201).

76738:121.

March 1993. 326pp. 9304060314. 74466.001.

The approved records disposition schedules specify the ap.

This report provides the status and results of the NRC Thor.

propriate duration of retenten and the final disposition for rnoluminescent Dosimeter (TLD) Direct Radiation Monitorin9 records created or maintained by the NRC. NUREG-0910, Revi-Network. It presents the radiation levels measured in the vicinity soon 2, Supplement 1 makes editorial and administrative of NRC licensed facilities throughout the country for the fourth changes to the NRC Schedule and forwards 3 sets of changes quarter 1992.

to the National Archives and Records Administrations's General NUREG-0837 V13 N01: NRC TLD DIRECT RADIATION MONI-TORING NETWORK. Progress Report. January. March 1993.

NUREG-0933 S15: A PRIORITIZATION OF GENERIC SAFETY STRUCKMEYER,R.; MCNAMARA,N. Region 1 (Post 820201).

ISSUES. EMRIT R. Division of Safety issue Resolution (Post May 1993. 231pp. 9306180289. 75389 073.

880717). May 1993.186pp. 9306110046. 75338:007.

This report provides the status and results of the NRC Ther.

The report presents the prionty rankings for generic safety moluminiscent Dosimeter (TLD) Direct Radiation Monitoring Net.

Issues related to nuclear power plants. The purpose of these I

work, it presents the radiation levels measured in the vicinity of rankings is to assist in the timely and efficient allocaten of NRC NRC heensed facilitics throughout the country for the first quar.

resources for tne resolution of those safety issues that have a ter of 1993 significant potential for reducing nsk. The safety pnority rankings r

E

.. ~

.. _ ~ --

l

\\

i I

6 Main Citations and Abstracts are HIGH, MEDIUM, LOW, and DROP and have been assigned information in this publication will be widely disseminated to on the basis of nsk significance estimates, the ratio of nsk to managers and employees engaged in activities hcensed by the costs and other impacts estimated to result if resolutions of the NRC, so that actions can be taken to improve safety by avord-safety issues were implemented. and the consideration of un-ing future violations similar to those described in this publica-certainties and other quant!tative or qualitatrve factors To the tion.

extent practical, estimates are quantitativ '

NUREG-0940 V12 NO2: ENFORCEMENT ACTIONS: SIGNIFi-NUREG-0933 S16: A PRIORITIZATION OF GENERIC SAFETY CANT ACTIONS RESOLVED.Ouarterfy Progress Report April-i ISSUES. EMRIT,R. Division of Safety issue Resolution (Post June 1993.

• Ofc of Enforcement (Post 870413). September 880717). November 1993. 250pp, 9401030183. 77642.001-1993. 395pp. 9310120039. 76737:001.

The report presents the priority rankings for genonc safety This compilahon summarizes signihcant enforcement actions issues related to nuclear power plants. The purpose of these that have been resolved dunng one quarterly period (April -

rankings is to assist in the timely and efficient allocation of NRC June 1993) and includes copies of letters, Notices, and Orders resources for the resolution of those safety issues that have a sent by the Nuclear Regulatory Commission to ficensees with significant potential for reducing nsk. The safety prionty rankings respect to these enforcement actions. It is anticipated that the are HIGH, MEDIUM. LOW, and DROP and have been assigned information in this publication wil! be widely disseminated to on the basis of risk significance estimates, the ratio of nsk to managers and employees engaged in actnnties licensed by the costs and other impacts estimated to result if resolutions of the NRC, so that actions can be taken to improve safety by avoid-safety issues were implemented, and the consideration of un-ing future violations similar to those desenbed in this pubhca-certainties and other quantitative or qualitative factors. To the lion.

extent pr actical, estimates are quantitative.

NUREG-0980 V01 N02:

NUCLEAR REGULATORY NUREG-0936 V11 N04: NRC REGULATORY AGENDA.Ouarterly LEGISLATION.102d Congress.

• Office of the General Counsel Report. October December 1992.

• Division of Freedom of infor-(Post 860701). October 1993. 522pp. 9401100130. 77736.001.

mation & Publications Services (Post 890205). February 1993.

This document is a comrnation of nuclear regulatory legisla-147pp. 9303110325. 74212:297.

tion and other relevant matenal through the 102d Congress,2d The NRC Regulatory Agenda is a compilation of all rules on Session. Thea compilation has been prepared for use as a re-which the NRC has recently completed action, or has proposed source document, which the NRC intends to update at the end action, or is considenng action, and all petitions for rulemaking of every Congress. The contents of NUREG-0980 include The whien have been received by the Commission and are pending Atomic Energy Act of 1954, as amended; Energy Reorganiza-i disposition by the Commission. The Regulatory Acjenda is up-tion Act of 1974, as amended, Uranium Mill Tailings Radiation dated and issued each quarter' Control Act of 1978; Low-Level Radioactive Waste Policy Act; i

NUREG 0936 V12 N01: NRC REGULATORY AGENDA.Ouarterly Nuclear Waste Policy Act of 1982; and NRC Authorization and -

Report. January-March 1993.

• Division of Freedom of Informa.

Appropriations Acts. Other materials included are statutes and i

tion & Publications Services (Post 890205). April 1993.135pp.

treaties on export licensing, nuclear non-proliferation, and envi-9305250050. 75004-072.

ronmental protection.

See NUREG-0936,V11,N04 abstract' NUREG-0980 V02 N02:

NUCLEAR REGULATORY j

NUREG-0936 V12 NO2: NRC REGULATORY AGENDAOuarterly LEGISLATION,102d Congress.

• Office of the General Counsel Report,Apni-June 1993

• Division of Freedom of information &

(Post 860701). October 1993. 463pp. 9401100134. 77734:001.

l Publications Services (Post 890205) July 1993. 138pp.

See NUREG-0980,V01,N02 abstract.

[

9308160147. 76121:001.

l See NUREG'0936,V11,N04 abstract NUREG-1021 R07: OPERATOR LICENSING EXAMINER STAND-l ARDS.

• Division of Reactor Controls & Human Factors (Post j

NUREG-0936 V12 NO3: NRC REGULATORY AGENDA.Ouarterly 921004). January 1993. 352pp. 9302230189. 64962:001.

Report. Jury-September 1993.

• Division of Freedom of informa~

The Operator Licensing Examiner Standards provide policy 1

tion & Publications Services (Post 890205). October 1993-and guidance to NRC examiners and establish the procedures 139pp. 9312070041. 77353 279 and practices for examining licensees and applicants for reactor See NUREG-0936,V11,N04 abstract.

operator and senior reactor operator licenses at power reactor NUREG-0940 V11 N04: ENFORCEMENT ACTIONS: SIGNIFl.

facilities pursuant to Part 55 of Title 10 of the Code of Federal CANT ACTIONS RESOLVED Ouarterly Progress Regulations (10 CFR 55). The Examiner Standards are intended Report. October-December 1992.

• Ofc of Enforcement (Post to assist NRC examiners and facility licensees to better under-870413). March 1993. 321pp 9303300180. 74407.001.

stand the initial and requalification examination processes and This compilation summanzes significant enforcement actions to ensure the equitable and consistent administration of exami-that have been resolved dunng one quarterly penod (October.

nations to all applicants. These standards are not a substitute December 1992) and includes copies of letters, Notices, and for the operator licensing regulations and are subject to revision Orders sent by the Nuclear Regulatory Commission to licensees or other internal operator licensing policy changes. This revision with respect to these enforcement actions. It is anticipated that will officially become effective 90 days after its publication is no-the information in this publication will be widely disseminated to ticed in the Federal Register. The revised dynamic simulator re-managers and employees engaged in activities licensed by the qualification examination procedure (ES-604) may be used im-NRC, so that actions can be taken to improve safety by avoid.

mediately, if requested by the facility licensee. The corporate ncd fication lettors issued after the effective date will provide fa-ing future violations similar to those desc9 bed in this publica-i tion.

cility licensees with at least 90 days notice that the examina-l NUREG-0940 V12 N01: ENFORCEMENT ACTIONS:SIGNIFICANT u es ACTIONS RESOLVED.Ouarterly Progress Report, January-March i

1993

• Ofc of Enforcement (Post 870413). June 1093. 250pp.

NUREG-1100 V09: BUDGET ESTIMATES. Fiscal Years 1994-9306210211. 75406 001.

1995.

• Division of Budget & Analysis (Post 890205). April 1993.

This compilation Summanzes sign:ficant enforcement actions 213pp. 9304160046. 74643:208.

that have been resolved dunng one quarterly period (January -

This report contains the fiscal year budget justification to Corb March 1993) and includes copies of letters, Notices, and Orders gress. The budget provides estimates for salaries and expenses sent by the Nuclear Regulatory Commission to licensees with and for the Office of the inspector General for fiscal years 1994 respect to these enforcement actions. It is anticipated that the and 1995.

l

Main Citations and Abstracts 7

NUREG 1125 V14: A COMPILATION OF REPORTS OF THE AD-power plant aging reserh reflects recognition that a number of VISORY COMMITTEE ON REACTOR SAFEGUARDS.1992 plants are entering tie final portion of their original 40-year op-Annual

• ACRS Advisory Committee on Reactor Safeguards.

erating licenses :.ad that, in addition to current aging effects a April 1993. 212pp. 9304160042. 74612:001.

focus on sarety considerations for hcense renewal becomes This compilation contains 50 ACRS reports submitted to the timely. The pnmary purpose of performing regulatory research is Commission. Executrve Director for Operations, or to the Office to develop and provide the Commission and its staff with the d Nuclear Regulatory Research, dunng calendar year 1992.11 technical bases for regulatory riecisions on the safe operation also includes a report to the Congress on the NRC Safety Re-of licensed nuclear reactors and facilities, to find unknown or search Program. All reports have been made available to the unexpected safety problems, and to develop data and related public through the NRC Pubhc Document Room and the U.S. Li.

information for the purpose of revising the Commission s rules, brary of Congress. The reports are divided into two groups: Part regulatory guides, or other guidance.

1: ACRS Reports on Project Reviews, and Part 2: ACRS Re-I ports on Generic Subjects. Part I contains ACRS reports alpha.

NUREG-1272 V07 N01: OFFICE FOR ANALYSIS AND EVALUA.

betized by project name and by chronological order within TION OF OPERATIONAL DATA.1992 Annual Report - Power project name. Part 2 categorizes the reports by the most appro.

Reactors.

• Office for Analysis & Evaluation of Operational pnate genene subject area and by chronological ordor within Data, Director. July 1993. 300pp. 9309210044. 76483:206.

subject area.

The annual report of the U.S. Nuclear Regulatory Commis-s n's Office for Analysis and Evaluat;on of Operational Data NUREG-1145 V09: U.S. NUCLEAR REGULATORY COMMISSION

( EOD) is devoted to the activities performed during 1992. The 1992 ANNUAL REPORT.

• Office of Administration (Post rep rt is published in two separate parts. NUREG-1272, Vol. 7 890205). July 1993. 287pp. 9309090029. 76391:001 N.1, c v rs p w r reactors and presents an overview of the This report covers the major activities, events. decisions and p rating expenence of the nuclear power industry from the planning that took place during fiscal year 1992 within the U.S-NRC perspective, including comments about the trends of some Nuclear Regulatory Commission (NRC) or involving the NRC-key performance measures. The report also includes the princi-NUREG-12f 4 R11: HISTORICAL DATA

SUMMARY

OF THE SYS.

pal findings and issues identified in AEOD studies over the past TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE.

year and summarizes information from such sources as hcensee ALLENSPACH.F. Division of Reactor Inspection & Licensee event reports, diagnostic evaluations, and reports to the NRC's Performance (Post 921004). February 1993.

129pp.

Operations Center. The reports contain a discussion of the inci-9303250054. 74363:155.

dent investigation Team program and summarize the Incident i

The Historical Data Summary of the Systematic Assessment investigation Team and Augmented Inspection Team reports for of Licensee Perforrrance (SALP) is produced periodically by the that group of heensees. NUREG-1272, Vol 7, No. 2. covers U.S. Nuclear Regulatory Commission. This summary.provides nonreactors and presents a review of the events and concerns the results of the assessment for each facihty by NRC region during 1992 associated with the use of licensed material in non-and is further divided into the following sections: Section 1 pre.

reactor applications, such as personnel overexposures and sentt the most recent SALP report ratings for facihties in opor.

medical nilsadministrations. Each volume contains a hst of the ation and under construction; Section 2 presents a chronologi.

AEOD reports issued for 1984 1992.

cal hsting of all SALP report ratings for each operating facility; Section 3 presents a chronological hsting of all SALP report rat-NUREG-1272 V07 NO2: OFFICE FOR ANALYSIS AND EVALUA-ings for each facikty under construction. For histoncal purposes.

TION OF OPERATIONAL DATA.1992 Annual Report - Non.

past construction ratings for facihties that recently have been h.

reactors.

• Office for Analysis & Evaluation of Operational Data, l

consed also are listed in Section 3.

Director. October 1993.164pp. 9312070235. 77350 001.

The annual report of tne U S. Nuclear Regulatory Commis.

J NUREG-1214 R12: HISTORICAL DATA

SUMMARY

OF THE SYS-sion's Office for Analysis and Evaluation of Operational Data TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE.

(AEOD) is devoted to the activities performed dunng 1992. The ALLENSPACH.F. Division of Reactor inspection & Licensee report is pubhshed in two separate parts. NUREG-1272. Vol. 7, j

Performance (Post 921004). August 1993.134pp. 9309210018.

No.1, covers power reactors and presents an overv'ew of the 76500:101.

operating experience of the nuclear power industry from the See NUREG 1214,R11 abstract.

NRC perspective, including comments about the trends of some NUREG 1220 R01: TRAINING REVIEW CRITERIA AND PROCE-key performance measures. The report also includes the pnnc6-DURES.

• Division of Licensee Performance & Quality Evatua-pal findings and issues identified in AEOD studies over the past tion (870411-921003). January 1993. 104pp. 9303150141.

, year and summarizes information frorn such sources as licensee 74242:188.

event reports, diagnostic evaluations, and reports to the NRC's This document provides direction to NRC personnel for re.

Operations Center. NUREG 1272, Vol. 7, No. 2, covers non-viewing training programs at nuclear power plants to venfy com-reactors and presents a review of the events and concerns phance with the requirements of 10 CFR 50.120 and 10 CFR 55 dunng 1992 associated with the use of licensed material in non-as applicable. It describes the process for evaluating the effec.

reactor applications, such as personnel overexposures and tiveness of training programs, provides aids for collection of in-medical misadministrations. Both reports also contain a discus-formation dunng interviews and observations, and provides cri-sion of the incident investigation Team program and summarize teria for evaluating the implementation of a systems approach both the incident Investigation Team and Augmented inspection to training. This document is not intended to have the effect of Team reports. Each volume contains a hst of the AEOD reports Issued for 1981 1992.

a regulation, it establishes no binding requirements or interpre.

tations of NRC regulations. It is intended as guidance only.

NUREG-1275 V09: OPERATING EXPERIENCE FEEDBACK NUREG 1266 V07: NRC SAFETY RESEARCH IN SUPPORT OF REPORT - PRESSURE LOCKING AND THERMAL BINDING OF REGULATION FY 1992.

• Office of Nuclear Regulatory Re-GATE VALVES Commercial Power Reactors HSU.C. Division of search (Post 860720). May 1993. 76pp. 9306210373.

Safety Programs (Post 870413). March 1993. 30pp.

75401:319.

9304020327. AEOD/S92-07. 74435.275.

This report, the eighth in a series of annual reports, was pre-The potential for valve enoperability caused by pressure lock-j pared in response to congressional inquiries concoming how ing and thermal binding has been known for many years in the i

nuclear regulatory research is used, it summartzes the accom-nuclear industry. In spite of numerous generic communications plishments of the Office of Nuclear Regulatory Research dunng issued in the past by the Nuclear Regulatory Commission (NRC)

FY 1992. A special emphasis on accomplishments in nuclear and industry, pressure locking and thermal binding continues to

8 Main Citations and Abstracts occur to gate valves installed in safety-related systems of both NUREG 1364: REGULATORY ANALYSIS FOR THE RESOLU-boiling water reactors (BWRs) and pressurized water reactors TION OF GENERIC SAFETY ISSUE 106: PIPING AND THE (PWRs). The generic communications to date have not led to USE OF HIGHLY COMBUSTIBLE GASES IN VITAL AREAS.

effective industry action to fally identify, evaluate, and correct GRAVES,C.C. Division of Safety issue Resolution (Post the problem This report identifies: (1) conditions when the fail.

880717). June 1993. 48pp. 9307130111. 75653:153.

(

ure mechanisms have occurred, (2) the spectrum of safety sys-Highly combustible gases such as hydrogen, propane, and tems that have been subiected to the failure mechanisms; and acetylene are used at all nuclear power plants. Hydrogen is of I

(3) conditions that may introduce the failure mechanisms under particular importance because it is stored in large quantities and both normal and accident conditions. On the basis of the eval-is distributed and used continuously in buildings containing untion of the operating events, the Office for Analysis and Eval-safety-related equipment. Large hydrogen releases at the hydro-uation of Operatonal Data (AEOD) of the NRC concludes that gen storage facilities or in these buildings could lead to fires or the binding problems with gate vatves are an important safety explosions that might result in loss of safety-related equipment, issue that needs pnonty NRC and industry attention. This report This report gives the regulatory analysis for the resolution of also provides AEOD's recommendation for actions to effectively Generic Safety issue 106, " Piping and the Use of Highty Com-prevent the occurrence of valve binding failures.

bustible Gases in Vital Areas." Scoping analyses showed that the risk associated with the storage and distribution of hydrogen NUREG-1307 R03:

REPORT ON WASTE BURIAL for cooling electric generators at boiling water reactors (BWRS),

CHARGES. Escalation Of Decommissioning Waste Disposal the off-gas system at BWRs, the waste gas system at pressur-Costs At Low-Level Waste Bunal Facilities.

• Division of Reguta-ized-water reactors (PWRs), and station battery rooms and port-tory Applicatons (Post 870413). May 1993. 59pp. 9303110042.

able bottles of combustible gas used for maintenance at PWRs 75338:203.

and BWRs is small. On the basis of generic evaluations, the One of the requirements placed upon nuclear power reactor NRC staff has concluded that several possible methodt to licensees by the U.S. Nuclear Regulatory Commisson (NRC) is reduce nsk could provide cost-effective safety benefits at some for the hcensees to penodically adjust the estimate of the cost plants. However, in view of the observed large differences in of decommissioning their plants, in dollars of the current year, plant specific characteristics affecting the nsk associated with as part of the process to provide reasonable assurance that the use of hydrogen, and the marginal generic safety benefits adequate funds for decommissioning will be available when that can be achieved in a cost-effective manner, it is recom-needed. This report, which is scheduled to be revised annually, mended that this generic issue be resolved simply by making contains the development of a formula for escalating decom-these results available in a genenc letter. This informaton may missioning cost estimates that is acceptable to the NRC, and help licensees in their plant evaluations recommended by Ge-contains values for the escalation of radioactive waste burial neric Letter 88 20, Supplement 4, " Individual Plant Examination costs, by site and by year. The licensees may use the formula.

of External Events for severe Accident Vulnerabilities," June i

the coefficients, and the burial escalaton from this report in 28, Igg 1, their escalaton analyses, or they may use an escalation rate at least equal to the escalation approach presented herein. Revi.

NUREG-1366: IMPROVEMENTS TO TECHNICAL SPECIFICA-sion 3 of this report corrects several errors in the calculations TiONS SURVEILLANCE REQUIREMENTS.

LOBEL,R.t and disposal costs for the reference PWR and the reference TJADER T.R.

Division of Operational Events Assessment

BWR, (870411-921003). December 1992. 93pp. 9301220193.

64652:313.

NUREG-1350 V05: NUCLEAR REGULATORY COMMISSION IN-In August 1983 an NRC task group was formed to investigate FORMATION DIGEST.1993 Edition. OLIVE,K.L Division of problems with surveillance testing required by Technical Specifi-Budget & Analysis (Post 890205). March 1993. 127pp.

cations, and to recommend approaches to eHect improvements.

9305250029. 75005:098.

NUREG-1024

(" Technical Specifications-Enhancing Safety The Nuclear Regulatory Commission Information Digest impact") resulted, and it contained recommendations to review (digest) provides a summary of information about the U.S. Nu-the basis for test frequencies; to ensure that the tests promote clear Regulatory Commission (NRC), NRC's regulatory responsi-safety and do not degrade equipment; and to review surveil-bihtes, the activities NRC licenses, and general information on lance tests so that they do not unnecessarily burden personnel.

domesbc and worldwide nuclear energy. The digest, published The Technical Specifications improvement Program (TSIP) was annually, is a compilaton of nuclear and NRC-related data and established in December 1984 to provide the framework for re-is designed to provide a quick reference to major facts about writing and improving the Technical Specifications. As an ele-the agency and the industry it regulates. In general, the data ment of the TSIP, all Technical Specifications surveillance re-cover 1975 through 1992, with exceptons noted. Information on quirements were comprehensively examined as recommended generating capacity and average capacity factor for operating in NUREG-1024. The results of that effort are presented in this U.S. commercial nuclear power reactors is obtained from report. The study found that while some testing at power is es.

rnonthly operat ng reports that are submitted directly to the NRC sential, safety can be improved, equipment degradaton de-by the heenseo. This information is reviewed by the NRC for creased, and unnecessary personnel burden relaxed by reduc-consistency only and no independent vahdation and/or venfica-ing the amount of testing at power, ton is performed.

NUREG-1377 R04: NRC RESEARCH PROG" 4 PLANT NUREG-1363 V05: ATOMIC SAFETY AND LICENSING BOARD AGING. LISTING AND SUMMARIES OF R, s ISSUED PANEL ANNUAL REPORT. Fiscal Year 1992. COTTER.B.P.

THROUGH SEPTEMBER 1993. VORA.J.P. Dnnt w. of Engineer.

Atomic Safety & Licensing Board Panel. September 1993.43pp.

ing (Post 870413). December 1993. 118pp. 9401120296.

9310120269. 7676t:001.

77770:172.

In Fiscal Year 1992, the Atomic Safety and Licensing Board The U.S. Nuclear Regulatory Commission is conducting the Panel ("the Panel") handled 38 proceedings. The cases ad-Nuclear Plant Aging Research (NPAR) Program. This is a com-dressed issues in the construction, operation, and maintenance prehensive hardware-oriented engineenng research program fo-of commercial nuclear power reactors and other activites re-cused on understanding the aging mechanisms of components quiring a license from the Nuclear Regulatory Commission. This and systems in nuclear power plants. The NPAR program also report sets out the Panel's caseload dunng the year and sum-focuses on methods for simulating and monitoring the aging-re-marizes. highlights, and analyzes how the wide-ranging issues lated degradation of these components and systems. In addi-raised in those proceedings were addiessed by the Pane!'s tion, it provides recommendations for effective maintenance to judges and hcensing boards.

manage aging and for the implementation of the research re-6

~ i

I Main Citations and Abstracts 9

sults in the regulatory process. This document contains a listing NUREG 1427: REGULATORY ANALYSIS FOR THE RESOLU.

and index of reports generated in the NPAR program that were TION OF GENERIC ISSUE 143: AVAILABILITY OF CHILLED issued through September 1933 and summanes of those re-WATER SYSTEM AND ROOM COOLING. LEUNG.V.T. Division ports. Each summary desenbes the elements of the research of Scfety issue Resolution (Post 880717). December 1993.

covered in the report and outlines the significant results. For the 79pp. 9401140029. 77797:001.

convenience of the user, the reports are indexed by personal This report presents the regulatory analysis for Genenc issue author, corporate author, and subject.

(GI 143), "Availabahty of Chilled Water Svetem and Room Cool-ing." The heating, ventilating, and aa Mditioning (HVAC) sys-NUREG 1400: AIR SAMPLING IN THE WORKPLACE Final Report. HICKEY,E E.; STOETZEL,G A.; STROM,D J.; et al. Bat-tems and related auxiliaries are required to provide control in telle Memorial instrtute, Pacific Northwest Laboratory. Septem-environmental conditions in areas in light water reactor (LWR) ber 1993.104pp. 9310120325. 76742:181, plants that contain safety-related equipment. In some plants, This report provides technical information on air sampling that the HVAC and chilled water systems serve to maintain a suita.

will be useful for facilities following the recommendations in the ble environment for both safety and non-safety related areas.

NRC's Regulatory Guide 8.25 Revision 1, " Air Sampling in the Although some plants have an independent chilled water Workplace." That guide addresses air sampling to meet the re-system for the safety-related areas, the heat removal capabiltty often depends on the operability of other supporting systems Quirements in NRC's regulations on radiation protection, g

10CFR20. This report desenbes how to determine the need for such as the service water system or the component cooling air sampling based on the amount of material in process modF water system. The operability of safety related components de-fled by the type of material, release potential, and confinement pends upon operation of the HVAC and chilled water systems of the matenal. The purposes of air sampling and how the pur-to remove heat from areas containing the equipment. If cooling poses affect the types of air sampling provided are discussed.

to dissipate the heat generated is unavailable, the ability of the The report discusses how to locate air samplers to accurately Safety-related equipment to operate as intended cannot be as.

determine the concentrations of iarborne radioactive materials sured. Typical components or areas in the nuclear power plant that workers will be exposed to. The need for and the methods that could be affected by the failure of cooling from HVAC or of performing airflow pattern studies to improve the eccuracy of chilled water systems include the (1) emergency switchgear and air sampling results are included. The report presents and give battery rooms, (2) emergency diesel generator room, (3) pump examplos of several techniques that can be used to evaluate rooms for residual heat removal, reactor core isolation cooling, whether the airborne concentrations of matenal are representa-high-prossure core spray, and low-pressure core spray, and (4) tive of the air inhaled by workers. Methods to ad ust denved air comrol room. The unavailability of such safety-related equip-i concentrations for particle size are descn, bed. Methods to cali-ment or areas could cause the core damage frequency (CDF) to brate for volume of air sampled and estimate the uncertainty in increase significantly.

the volume of air sampled are described. Statistical tests for de-termining minimum detectable concentrations are presented.

NUREG-1444: SITE DECOMMISSIONING MANAGEMENT PLAN.

How to perform an annual evaluation of the adequacy of the air FAUVER.D.N.; AUSTIN J.H.; JOHNSON T.C.; et al. Division of Low-Level Waste Management & Decommissioning (Post sampling is also discussed-870413), October 1993. 200pp. 9311080087. 77069.099.

MUREG 1415 V05 NO2: OFFICE OF THE INSPECTOR The Nuclear Regu!atory Commission (NRC) staff has iderti-GENERALSemiannual Report, October 1, 1992 March 31, tied 48 sites contaminated with radioactive material that require 1993.

• Office of the inspector General (Post 890417). April special attention to ensure timely decommissioning. While none 1993. 38pp. 9306210379. 75401:278' of these sites represent an immediate threat to public health The Inspector General is required by statute to prepare a and safety, they have contamination that exceeds existing NRC semiannual report to Congress which summanzes the significant enteria for unrestncted use. All of these sites require some investigative and aud t activities of the office. The 6-month re-degree of remediation, and several involve regulatory issues porting penod ends March 31 and September 30. The report is that must be addressed by the Commission before they can be submitted to the Chairman not later than Apnl 30 and October released for unrestricted use and the applicable licenses termi-31, respectively, of each year. The Chairman prepares com-nated. This report contains the NRC staff's strategy for address-ments and his own report and submits both reports to Con-ng the technicat, legal, and policy issues affecting the timely O

decommissioning of the 48 sites and describes the status of de-commissioning activities at the sites.

MUREG-1415 V06 N01: OFFICE OF THE INSPECTOR GENERAL. Semiannual Report, April 1,1993 September 30, NUREG 1449: SHUTDOWN AND LOW-POWER OPERATION AT 1993. NORTON.LJ.; BARCHI,T.J ; FREDERICK,L.; et al. Office NUCLEAR POWER PLANTS IN THE UNITED STATES. Final of the Inspector General (Post 890417). October 1993. 46pp Report.

• Division of Systems Safety & Analysis (Post 921004).

9312160316. 77509:285 Septemt

' 200pp. 9310130052. 76743:001.

See NUREG-1415,V05,N02 abstract.

Therr

.sntains the results of the NRC staff's evaluation of shs and low power operations at U.S. commercial nu-NUREG-1423 V04: A COMPILATION OF REPORTS OF THE AD.

Cle, plants. The report describes studies conducted by VISORY COMMITTEE ON NUCLEAR WASTE.Juh 1992 - June the star. :n the following areas: operating experience related to 1993.

• Advisory Committee on Nuclear Waste _ August 1993.

shutdown and low power operations, probabilistic risk assess-81pp. 9309210031, 76482:337.

ment of shutdown and low-power conditions and utility pro-This compilation contains 17 reports issued by the Advisory grams for planning and conducting activities during periods the Committee on Nuclear Waste (ACNW) dunng the fifth year of its plant is shut down. The report also documents evaluations of a operation. The reports were submitted to the Chairman and number of technical issues regarding shutdown and low-power Commissioners of the U.S. Nuclear Regulatory Commission, the operations performed by the staff, including the principal find-Executive Director for Operations, the Director. Office of Nucle-ings and conclusions. Potential new regulatory requirements are ar Material Safety and Safeguards, or to the Director, Division of discussed, as well as potential changes in NRC programs. A High level Waste Management. Office of Nuclear Matenal draft report was issued for comment in February 1992. This Safety and Safeguards. All reports prepared by the Committee report is the final version and includes the responses to the have been made available to the public through the NRC Public comments along with the staff regulatory analysis of potential Document Room and the U S Lbrary of Congress.

new reautrements.

a

10 Main Citations and Abstracts NUREG-1453: REGULATORY ANALYSIS FOR THE RESOLU-action to resolve Generic issue 105 is licensee participation in TION OF GENERIC ISSUE 142. LEAKAGE THROUGH ELEC-individual plant examinations (IPEs).

TRICAL ISOLATORS IN INSTRUMENTATION CIRCUITS.

NUREG-1467: FEDERAL GUIDE FOR A RADIOLOGICAL ROURK,C.J. Division of Safety Issuo Resolution (Post 880717).

RESPONSE. Supporting The Nuclear Regulatory Commission September 1993. 21pp. 9310130042. 76744.266.

During The Initial Hours Of A Serious Accident. HOGAN,R.T. Di-Genoric issue (GI) 142 deals with staff concerns about the vision of Operational Assessment (Post 870413). November design of isolation devices used to ensure separation between 1993. 20pp. 9401030173. 77642.290.

Class 1E and non Class 1E electrical control and instrumenta-This document is a planning guide for those Federal agencies tion circuits This issue was initiated in June 1987. Staff reviews that work with the Nuclear Regulatory Commission (NRC) dunng of the implernentation of the Safety Parameter Display System the initial hours of response to a senous radiological emergency (SPDS) requirement indicated that some isolation devices used in which the NRC is the Lead Federal Agency (LFA). These to provide an inter' ace between the non-Class 1E SPOS and Federal agencies are: DOE, EPA, USDA, HHS, NOAA, and the Class 1E safety systems would allow signal leakage if elec-FEMA. This guide is intended to help these agencies prepare tocally challenged it was unknown if the amount of leakage for a prompt response. Instructions are provided on receiving posed a hazard to safe operation of the Class 1E system. A the initial notification, the type of person to send to the scene, review of failure records does not reveal any incidents of the facility at which people are needed, how to get them to that system damage caused by isolation device challende Furthor-and what they should do when they arrive. Federal more, a review of existing PRA data indicates that the safety facility'es not specifically mentioned in this guide may also be agenci significance of ID challenge is low, at the expected challenge asked to support the NRC-event frequency. However, based upon the potential design vanations in future control systems resulting from application of NUREG 1470 V02: CHIEF FINANCIAL OFFICER'S ANNUAL computer technology, addctional design and qualification test re" REPORT - 1993.

• Office of the Controller (Post 890205). Sep-quirements for future plants are recommended.

tember 1993.116pp. 9311030220. 77029.003.

The Chief Financial Officers Act of 1990 requires the NRC NUREG-1461: REGULATORY ANALYSIS FOR THE RESOLU-Chief Financial Officer to prepare and submit an annual report TION OF GENERIC ISSUE 153: LOSS OF ESSENTIAL SERV, to the agency head ou ins wtor of the Office of Manage-ICE WATER IN LWRS SU.T. M. Division of Safety issue Reso-ment and B4et. This 1993 repon is the second annual report lution (Post 860717). August 1993. 32pp. 9309030241.

for the NRC and includes a desenption and analysis of the status of Unandal managemed for Mcal War M, an aM in s eport, the staff of the U S. Nuclear Regulatory Com-financial statement and audit reports foi Fiscal Year 1992, and mission provides a regulatory analysis for the proposed resolu-a summary of the reports on internal accounting and administra-tion of Genenc issue 153 (GI 153) " Loss of Essential Service e conM sysms b N Water in LWRs " GI 153 deals with the concerns pertaining to the rehability of essential service water (ESW) system and relat-NUREG-1472: REGULATORY ANALYSIS FOR THE RESOLU-ed problems for all hght water reactors except the seven multi-TION OF GENERIC ISSUE 57. Effects Of Fire Protection unit sites addressed by GI-130 " Essential Service Water Pump System Actuation On Safety-Related Equipment. WOODS,H.W.

failures at MuitFUnit S4tes." On the basis of the technical find-Division of Safety issue Resolution (Post 880717). October ings of a scoping study for GI 153, the staff recommends that 1993. 40pp. 9311080077. 77068.321.

the insights gained from the study serve as a complement to Actuation of Fire Protection Systems (FPS) in Nuclear Power the on-going ESW performance inspection program. The staff Plants have resulted in adverse interactions with equipment im-also concludes that ESW system reliabihty is being addressed portant to safety. Precursor operational experience has shown by vanous on-going regulatory programs. Therefore, the staff that 37% of all FPS actuations damaged some equipment, and recommends that GI 153 should be considered " RESOLVED."

20% of all FPS actuations have resulted in a plant transient and The need for future action (s) on ESW rehability is expected to reactor inp. On an average, 0.17 FPS actuations per reactor be deterrnined from these on-going programs.

year have been expenenced in nuclear power plants in this country. This report presents the regulatory analysis for Gl-57, NUR EG-1463: REGULATORY ANALYSIS FOR THE RESOLU-

" Effects of Fire Protection System Actuation on Safety-Related TION OF GENERIC SAFETY ISSUE 105: INTERFACING Equipment". The nsk reduction estimates, cost / benefit analy-SYSTEM LOSS-OF-COOLANT ACCIDENT IN LIGHT WATER ses, and other insights gained during this effort have snown that REACTORS.

Dnnsion of Safety issue Resolution (Post implementation of the recommendations contained in this report 880717). July 1993. 77pp. 9308160153. 76114 001.

can significantly reduce nsk, and that these improvements can An interfacing systems loss of coolant accident (ISLOCA) in' be warranted in accordance with the backfit rule,10 CFR volves fa: lure or improper operabon of pressure isolation valves 50.109(a)(3). However, plant specific analyses are required in (PlVs) that compose the boundary between the reactor coolant order to identify such improvements. Generic analyses can not system and low-pressure rated systems. Some ISLOCAs can serve to identify improvements that could be warranted for indi-bypass containment and result in direct release of fission prod-vidual, specific plants. Plant specific analyses of the type ucts to the environment. A cost / benefit evaluation, using three needed for this purpose are underway as part of the Individual PWR analyses, calculated the benefit of two potential modifica-Plant Examination of External Events (IPEEE) program.

tions to the plants. Alternative I is improved p! ant operations to optimize the operator's performance and reducg human error NUREG-1473: ELECTRICAL DISTRIBUTION SYSTEM FUNC-probabihties Alternative 11 adds pressure sensing devices, ca-TIONAL INSPECTION (EDSFI) DATA BASE PROGRAM.

bbng, and instrumentation between two PlVs to provide opera-GAUTAM,A.S. Division of Reactor inspection & Ucensee Per-tors with continuous monitoring of the first PlV. These two alter-formance (Post 921004). January 1993. 45pp. 9303120055.

natives were evaluated for the base case plants (Case 1) and 74237:224.

for each p! ant, assuming the plants had a particular auxihary This document describes the organization, installation proce-building design in which severe flooding would be a problem if dures, end operating instructions for the database computer an ISLOCA occuned. The auxitiary building design (Case 2) was program containing inspection findings from the U.S. Nuclear selected from a survey that revealed a number of designs with Regulatory Commission's (NRC's) Electrical Distnbution System features that provided less than optimal resistance to ECCS Functional Inspections (EDSFis). The program enables the user equipment loss caused by a ISLOCA-induced environment. The to search and sort findings, ascertain trends, and obtain printed results were judged not to provido sufficient basis for generic reports of the findings. The findings include observations, unre-requnements it was concluded that the most viable course of solved issues, or possible deficiencies in the design and imple-l i

Main Citations and Abstracts 11 l

mentation of electncal distnbution systems in nuclear plants.

outstanding technical issues and concerns that had previously This database will assist those propanng for electncal inspec-been raised regarding voltage-based plugging criteria for tions, searching for deficiencies in a plant, and determining the ODSCC. Most of these issues are relevant to the long-term ap-corrective actions previously taken for similar deficiencies. This proval of voltage-based pluggin0 cnterie This report desenbes database will be updated as now EDSFis are completed.

the results of the task group's review 4 a evaluation of: (1) the NUREG-1474: EFFECT OF HURRICANE ANDREW ON THE issues related to tube integnty, including the potential for tube TURKEY POINT NUCLEAR GENERATING STATION FROM rupture or leakage under postulated-accident conditions and the AUGUST 20,30, 1992. HEBDON,F.J. Office for Analysis & Eval, safety imphcatsons of these issues; (2) the radiological doses uation of Operational Data, Director.

• Institute of Nuclear and the potential for core damage associated with a range of Power Operations. March 1993. 80pp. 9307060041. 75585.001.

assumed primary-to-secondary leak rates; and (3) the safety On August 24, 1992, Hurricane Andrew, a Category 4 hurrl-significance of ODSCC of steam generator tubes.

cane, struck the Turkey Pnint Electncal Generating Station with NUREG 1479; RESULTS FROM TWO WORKSHOPS: STATE RA.

sustained winds of 145 mph (233 km/h) This is the report of DIATION CONTROL PROGRAM 3 DEVELOPING AND AMEND-the team that the U.S. Nuclear Regulatory Commission and the ING REGULATIONS AND FUNDING. PARKER.G. Office of Institute of Nuclear Power Operations jointly sponsored: (1) to State Programs (Post 911117). September 1993. 34pp.

review the damage that the hurricane caused the nuclear units 9310130049. 76743.190.

and the utility's actions to prepare for the storm and recover from it; and (2) to compile lessons that might benefit other no-The first section of this document presents the results of a clear reactor facilities ~

technical workshop on the process of regulations development and amendment sponsored by the Nuclear Regulatory Commis-NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT sion (NRC). This workshop focused on methods for reducing TO CONSTRUCT AND OPERATE A FACILITY TO the time it takes to promulgate regulations to help those States RECEIVE STORE, AND OtSPOSE OF 11E.(2) BYPRODUCT that are having difficulty meeting the three year deadline for MATERIAL NEAR CLIVE. UTAH Docket No. 40-8989. Envirocare adopting new NRC regulations. Workshop participants respond-Of Utah,Inc. BRUMMETT,E.; ABU-EID,Ra MULLINS,A.; et al. Di-ed to six questions, reviewed the procedures used by the vari-vision of Low-Level Waste Management & Decommissioning ous States for revising and adopting changes to their regula-(Post 870413). August 1993. 206pp. 9309210022. 76481:269, tions, and reviewed the time-flow charts used by various States.

A Final Environmental Impact Statement (FEIS) related to the This workshop was designed to provide guidance to States that hcensing of Envirocare of Utah, Inc.'s proposed disposal facility are promulgating and revising regulations. The second section in Tooele County, Utah (Docket No. 40-8989) for byproduct ma-of this document summarites the proceedings of a technical tenal as defined in Section 11e.(2) of the Atomic Energy Act, as workshop, also sponsored by the NRC, on funding radiation amended, has been prepared by the Office of Nuclear Matenal control programs that emphasized fee schedules and effective Safety and Safeguards. This statement desenbes and evalu-strategies for the 1990s. This workshop focused on determining ates' (1) the purpose of and need for the proposed action; (2) the true costs of running a program, on setting realistic fees for the alternatives considered, and (3) tho environmental conse-the various categories of licenses, and on the most efficient quences of the proposed action. The NRC has concluded that methods for sending invoices, recording receipts, depositing the proposed action evaluated under the National Environmen-money received and issuing licenses. Workshop participants re-tal Policy Act of 1969 (NEPA) and 10 CFR Part 51,is to permit sponded to seven questions; reviewed the methods various the apphcant to proceed with the protect as described in this States use to determine true costs; reviewed the procedure that Statement-the vanous States use to produce invoices and licenses; re-i NUREG-1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATE, viewed the procedures that the States are required to abide by MENT TO CONSTRUCT AND OPERATE A FACILITY TO RE-when they receive money; and reviewed the method used by CEIVE, STORE, AND DISPOSE OF t tE (2) BYPRODUCT MA-the NRC to determine the cost of its vanous programs.

TERIAL NEAR CLIVE, UT AH. Docket No. 40-8989, Envirocare NUREG-1480: LOSS OF AN IRIDIUM 192 SOURCE AND THER.

Of Utah, Inc. BRUMMETT.E.; ABU EID,R.; MULLINS,A.; et al.

APY MISADMINISTRATION AT INDIANA REGIONAL CANCER Division of Low-Level Waste Management & Decommissioning (Post 870413) February 1993. 224pp. 9303120020. 74238 001.

C. ENTER,1NDIANA. PENNSYLVANIA,0N NOVEMBER 16, 1992.

A Draft Environmental Statement (DEIS) related to the licens-NRC - No Detailed Affiliation Given. February 1993. 223pp.

9303120040.74263:047.

ing of Envirocare of Utah, Inc.'s proposed disposal facility in On December 1,1992, the Indiana Regional Cancer Center Tooele County, Utah, (Docket No 40-8989) for byproduct mate

• rial as defined in Section 11e(2) of the Atomic Energy Act, as reported to the t).S. Nuclear Regulatory Commission's (NRC)

Region i that they believed a 1.37 E + 11 becquerel (3.7-cune) amended, has been prepared by the Office of Nuclear Material iridium 192 source from their Omnitron 2000 high dose rate Safety and Safeguards. This statement desenbes and evaluates remote brachytherapy afterloader had been found at a bioha-(1) the purpose of and need for the proposed action, (2) the al-2ard waste transfer station in Camogie, Pennsylvania. After noti-tornahves considered, and (3) the environmental consequences of the proposed action. The NRC has concluded that the pro-fying the NRC, this cancer center, one of several operated by the licensee. Oncology Services Corporation, retrieved the posed action evaluated under the National Environmental Policy source, and Region I dispatched an inspector and a supervisor Act of 1969 (NEPA) and 10 CFR Part 51,is to permit the apph.

to investigate the event. The source was first detected when it cant to proceed with the project as described in this Statement.

triggered. radiation alarms at a waste incinerator facility in NUREG 1477 DRFT FC: VOLTAGE-BASED INTERIM PLUGGING Warren, Ohio. The licensee informed the NRC that the source CRITERIA FOR STEAM GENERATOR TUBES Draft Report For wire had apparently broken during treatment of a patient on No-Comment.

• IPC Task Group June 1993.120pp. 9307060061.

vember 16, 1992, leaving the source in the patient. On the 75563 068.

basis of the seriousness of the incident, the NRC elevated its This report presents the preliminary results of a special U.S.

response to an Incident investigation. The incident investigation Nuclear Regulatory Commission (NRC) task group established:

Team initiated its investigation on December 3,1992. The in-(1) to review the technical bases for and outstanding issues re.

vestigation team concluded that the patient received a serious lated to intenm approval of voltage based intenm plugging cnte-misadministration and died on November 21, 1992, and that ria for outside-diameter stress corrosion cracking (ODSCC) of over 90 individuals were exposed to radiation from November steam generator tubes; and (2) to prepare conclusions and rec-16 to December 1,1992. In a press release dated January 26, ommendations concerning implementation of these enteria The 1993, the, Indiana County Coroner stated that the cause of task group activities included identification and assessment of death hsted in the official autopsy reoort was ' Acute Radiational

. -. - - ~.. -. -.. - -

---

~.

12 Main Citations and Abstracts Exposure and Consequences Thereof.' An almost identical roll-up door on the Turbine Building TMi Secunty reported this source-wire failure occurred with an afterloader in Pittsburgh, event to the U.S. Nuclear Regulatory Commission's (NRC's)

Pennsylvania, on December 7,1992, but with minimal radiologi-Headquarters operations officer and declared a Security Emer.

cal consequences. This incident was included in the investiga.

gency upon determining that the protected area of the plant had tion. This report discusses the Omnitron 2000 high dose rate af-been compromised. At 7:23 a m., the TMI 1 shift supervisor offi-terloader source-wire failure, the reasons why the failure was cially notified the NRC Headquarters operations officer that he not detected by Indiana Regional Cancer Center, the potential had declared a Site Area Emergency effective at 7:05 a.m.

consequences to the patient, the estimated radiological doses Upon considering the possible significance to physical security to workers and the public, and regulatory aspects associated and the regulatory questions that could result from the event, with this incident.

the NRC Executive Director for Operations established an inci-dent innstigation team to determine what happened and make NUREG-1482 DRFT FC: GUIDELINES FOR INSERVICE TESTING appropriate findings and conclusions. In this report the team de-AT NUCLEAR POWER PLANTS. Draft Repori For Comment.

scnbed the event and the response to the event evaluated the CAMPBELL,P. Division of Engineenng (Post 921004). November regulatory requirements, and presented the team,a findings and 1993. 200pp. 9312220172. 77544:159.

concNssns.

in this report, the staff gives guidelines for developing and im-piementing programs for the inservice testing of pumps and NUREG-1487 V01: FISCAL YEAR 1994 1998 INFORMATION valves at commercial nuclear power plants. The report includes TECHNOLOGY STRATEGIC PLAN

• Office of Information Re-U.S Nuclear Regulatory Commission (NRC) guidance and rec-sources Management (Post 890205). November 1993. 28pp.

ommendations on inservice lesting issues. The staff discusses 9312160341. 77509.332.

the regulations, the components to be included in an inservice A team of senior managers from across the U.S. Nuclear testing program, and the preparation and content of cold shut-Regulatory Commission (NRC), working with the Office of Infor.

down and refueling outage justifications and requests for relier mation Resources Management (IRM), has completed an NRC from the Amencan Society of Mechan $ cal Engineers Code re' Strategic information Technology (IT) Plan. The Plan addresses quirements. The statt also gives specific guidance on relief ac-three major areas: (1) IT Program Management, (2) IT Infra.

ceptable to the NRC and advises licensees in the use of this structure, and (3) Information ano Applications Management.

Information for application at their facilities. The staff discusses Key recommendations call for acceleraMg the replacement of the revised standard technical specifications for the inservice Agency workstations, implementing a new *locument manage-testing program requirements and gives guidance on the proc-ment system, applying business process roeinineering to se-ess a licensee may follow upon finding an instance of noncom-lected Agency work processes, and establishing an information pliance with the Code' Technology Council to advise the Director of IRM.

NUREG 1484 DRFT: DRAFT ENVIRONMENTAL IMPACT STATE-NUREG-1488 DRFT FC: REVISED LIVERMORE SEISMIC MENT FOR THE CONSTRUCTION AND OPERATION OF CLAl-HAZARD ESTIMATES FOR 69 NUCLEAR POWER PLANT BORNE ENRICHMENT CENTER HOMER, LOUISIANA. Docket SITES EAST OF THE ROCKY MOUNTAINS. Draft Report For No. 70-3070, Louisiana Energy Services,L.P.

• Division of Fuel Comment. SOBEL,P. Division of Engineering (Post 921004). Oc-Cycle Safety & Safeguards (Post 930207). November 1993.

tober 1993. 98pp. 9311080066. 77069:001.

361pp. 9312070063. 77354:061.

This report presents updatad Lawrence Livermore National This Draft Environmental impact Statement (DEIS) was pre.

Laboratory (LLNL) probabilistic seismic hazard analysis esti-pared by the Nuclear Regulatory Commission in accordance mates for 69 nuclear power plant sites in the region of the with NRC regulation 10 CFR Part 51, which implements the Na, United States east of the Rocky Mountains. LLNL performed a tional Environmental Policy Act (NEPA), to assess the potential re-elicitation of seismicity and ground motion experts to improve environmental impacts of the construction and operation of a their estimates of uncertainty in seismicity parameters and proposed gaseous centrifuge enrichment facility to be buift in ground motion models. Using these revised inputs, LLNL updat-Claiborne Parish, LA The proposed facihty will have a produc.

ed the seismic hazard estimates documented in NUREG/CR-tion capacity of about 866 tonnes annually of up to 5 percent 5250 (1989). These updated hazard estimates will be used in enriched UF(6), using a proven centrifuge technology, included future NRC actions. A detailed summary of the revised probabi-in the assessment are construction, both normal operations and hstic hazard methodology, the seismicity and ground motion potential accidents (internal and external events), and the even-inputs, and sensitivity studies is presented in a report by J. Savy tual decontamination and decommissioning of the site. In order to help assure that releases from the operation of the facility (1993, LLNL Report UCRL-ID-115111). This report summarizes the Savy (1993) report and documents the 1993 LLNL hazard and potential impacts on the public are as low as reasonably results for the 69 sites. For the purpose of companng these achievable, an environmental monitoring program was devel, probabilistic seismic hazard estimates to the seismic design of oped to detect significant changes in the background levels of the nuclear power plants, a tabla of safe-shutdown earthquake uranium around the site. Other issues addressed include the spectral values is included.

purpose and need for the facility, the alternatives to the pro-posed action, and the site selection process. The NRC con-NUREG/CP-0040: PROCEEDINGS OF WORKSHOP V-FLOW ciudes that the facility can be constrvCted and operated with AND TRANSPORT THROUGH UNSATURATED FRACTURED small and acceptable impacts on the pubhc and the environ.

ROCK - RELATED TO HIGH LEVEL RADIOACTIVE WASTE ment and proposes to issue a license to the applicant, Louist-DISPOSAL. Held At Radisson Suite Hotel Tucson' ana Energy Services, to authorire constructeon ard operation of Arizona, January 7 10,1991. EVANS.D.D.; NICHOLS'ON,T,J. Ari the proposed facshty. This DEIS provides the public with the op-zona, Univ. of, Tucson, AZ. June 1993. 250pp. 9307220313.

portunity to comment on the proposed action and the treatment 75744:119 of potential environmental impacts' The " Workshop on Flow and Transport Through Unsaturated NUREG-1485: UNAUTHORIZED FORCED ENTRY INTO THE Fractured Rock Related to High-Levbl Radioactive Waste Dis.

PROTECTED AREA AT THREE MILE ISLAND UNIT 1 ON FEB-posal" was cosponsored by the NRC, the Center for Nuclear RUARY 7,1993.

• Ofc of the Executive Director for Operations.

Waste Regulatory Analyses, and the Uscesity of Arizona (UAZ)

April 1993.142pp. 9304210263. 74677.061.

and was held in Tucson, Arizona, on January 710,1991. The On February 7,1993, at 6 53 a.m. Eastern Standard Time focus of this workshop, similar to the earher four (the f'rst being i

(EST) an intruder drove into the site owner-controlled area, in 1982), related to hydrogeologic technical issues associated through a gate into the protected area of Three Mile Island Nu-with possible disposal of commercial high level nuclear wasto clear Generating Station, Unit 1 (TMI-1) and crashed through a (HLW) in a geologic repository within an unsaturated fractured

.Y

Main Citations and Abstracts 13 rock system which coincides with the UAZ field studios on HLW to discussion of the framework for a performance-based regula.

disposal. The presentations and discussons centered on flow tory approach, in addition, panelists and attendees discussed

)

and transport processes and conditions, relevant paramotors, scope, schedules, and status of specific regulatory items: con-as well as state-of-tho ari measurement techniques, and modol-tainment leakage losting requirements, fire protection require-ing capabihtios. Tho workshop consisted of four half-day techni-ments, requirements for environmental qualification of electncal cal meetings; a one day field visit to the Apache leap test site equipment, roquests for information under 10CFR50 54(f), re-to review ongoing field studies that are examining site charac-quirements for combustible gas control systems, and quality as-l tentation techniques and developing data sets for model valida-surance requirements.

tion studies; and a final half-day session devoted to examining research needs related to modchng groundwater flow and radio.

NUREG/CP-0130 V01: PROCEED NGS OF THE 22ND DOE /NRC nuchde transport in unsaturated, fractured rock. These proceed-NUCLEAR AG CLEAN!NG CONFERENCE. Sessions 18. Hold in ings prov'do extended abstracts of the technical presentations Denver, Colorado, August 24 27,1992, FIRST,M.W. Harvard i

and short summanes of the research group reports.

School of Public Health, Boston, MA. July 1993. 500pp.

9307270008. CONF-9020823. 75801:00t, NUREG/CP-0126 V01: PROCEEDINGS OF THE TWENTIETH This documeni contains the papers and the associated dis-WATER REACTOR SAFETY INFORMATION MEETING.

cussions of the 22nd DOE /NRC Nuclear Asr Cleaning Confer-WEISS,A.J. Brookhaven National Laboratory. March 1993.

ence. Major topics are: (1) advancod reactois; (2) reprocessing; 535pp. 9304260111 74713 001-(3) filter testing; (4) waste managemon', (5) instruments and This three volume report contains 93 papers out of the 100 samphng; (6) reactor accidents; (7) filters and filter performance; that were presented at the Twentieth Water Reactor Safety in-(8) adsorber testing and performance; (9) carbon testing; and formation Meeting held at the Bethosda Mamott Hotel, Bo*hes-(10) ventilation systems.

l da, Maryland, dunng the week of October 21-23,1992. The l

papers are panted in the order of their presentation in each sos-NUREG/CP-0130 V02: PROCEEDINGS OF THE 22ND DOE /NRC sion and describe progress and results of programs in nuclear NUCLEAR AIR CLEANING CONFERENCE. Sessions 9-16. Held I

safety research conducted in this country and abroad Foreign in Denver, Colorado, August 24 27, 1992. FIRST,M.W. Harvard j

participation in the meeting included 10 different papers pre.

School of Public Health, Boston, MA. July 1993. 500pp.

sented by researchers from CEC, China, Finland, Franco, Ger.

9307270050 CONF-9020823 75803.070.

many, Japan, Spain and Taiwan, The titles of the papers and See NUREG/CP-0130,V01 abstract, the names of the authors have been updated and may diffor NUREG/CP-0131: PROCEEDINGS OF THE JOINT lAEA/CSNI from those that appeared in the final program of the moots.ng SPECIALISTS

• MEETING ON FRACTURE MECHANICS VERI-l NUREG/CP-0126 V02: PROCEEDINGS OF THE TWENTIETH FICATION BY LARGE SCALE TESTING. Hold At Pollard WATER REACTOR SAFETY INFORMATION MEETING.

Auditorium, Oak Ridge,Tennesseo. PUGH,C.E_; BASS,8R.:

WEISS.A J-Brookhaven National Laboratory. March 1993.

KEENEY.J.A. Oak Ridge National Laboratory. October 1993.

555pp. 9304190226. 74621:001, 1,000pp. 9311080091. ORNL/TM 12413. 77070 001, Soo NUREG/CP.-0120,V01 abstract This report contains 40 papers that were presented at the Joint lAEA/CSNI Specialists' Meeting - Fracture Mechanica Ver.

MUREG/CP-0126 V03: PROCEEDINGS OF THE TWENTIETH "O

WATER REACTOR SAFETY INFORMATION MEETING WEISS.A J.

Brookhaven National Laboratory. March 1993 1992. The papers are pnnted in the order of their presentation 585pp. 9304190227. 74623 001 See NUREG/CP-0126,V01 at$stract-and/or ductile) expenmonts, " analyses of those experiments, NUREG/CP-0128: PROCEEDINGS OF THE INTERNATIONAL and compansons betwoon predictions and expenmental results.

WORKSHOP ON THE CONDUCT OF INSPECTIONS AND IN.

The goal of the meeting was to allow international experts to SPECTOR OUAllFICATION AND TRAINING. GRIMES,0 K.

examine the fracture behavior of various materials and struc.

i Office of Nuclear Reactor Regulation, Director (Post 870411).

tures under conditions relevant to nuclear reactor components Fobruary 1993 231pp. 9303160283. NEA/CNRA/R(92)3.

and operating environmonts. The emphasis was on the ability of 74254 069 various fracture models and analysis methods to predict the The results of an international workshop on nuclear reactor wide range of experimental data now available. The internation-inspection are presented Topics include typos of inspection al nature of the meeting is illustrated by the fact that papers programs (residfint, teams, contrallred), methods of inspect on, were presented by researchers from CSFR, Finland, France, investigdtion of incidents / accidents, achieving correction of defi.

Germany, Japan, Russia, U.S., and the U.K. There were experts ciencies found dunng inspections, training and quahfications of present from several other countrios who participated in dis-inspectors, and inspections of shutdown activities and low cussing the results presented. The titles for some of the final power operations, Represented at the conference were Bol-papers and the names of the authors have boon updated in this gium, Bulgana, Canada, CFSR, France, Finland, Germany, Hun-report and may differ shghtly from those that appeared in the gary, IAEA, The Netherlands, Norway, OECD, Russia, Spain, final program of the meeting.

i Sweden, Switzerland, U K., U.S, and the Ukraine.

NUREG/CP-0132: TRANSACTIONS OF THE TWENTY-FIRST 4

NUREG/CP-O t29: PROCEEDINGS OF THE WORKSHOP ON WATER REACTOR SAFETY INFORMATION MEETING.

)

PROGRAM FOR ELIMINATION OF REQUIREMENTS MARGIN.

MONTELEONE,S. Office of Nuclear Regulatory Research (Post AL TO SAFETL DEY,M. Advanced Reactors Branch (Post 860720). October 1993. 200pp. 9311010020. 76982.042.

910830). ARSENAULT,F.; PATTERSON.M.; et at SCIENTECH, This report contains Summaries of papers on reactor safety Inc. September 1993.194pp 9310120259. 76741:245.

research to be prosented at the 21st Water Reactor Safety in-Those are the proceedings of the Public Workshop on the formation Meeting at the Bethesda Mamott Hotel, Bethesda, U.S. Nuclear Regulatory Commission's Program for Elimination Maryland, October 25-27, 1993. The summaries briefly describe of Requirements Marginal to Safety. The workshop was hold at the programs and results of nuclear safoty rosearch sponsored the Hohday Inn, Bethesda, on April 27 and 28,1993. The pur-by the Office of Nucioar Regulatory Research, U.S. NRC, Sum-pose of the workshop was to provido an opportunity for pubhc marios of invited papers concerning nuclear safety l*, sues from and industry enput to the program The workshop addressed the U S. government laboratones, the electric utihties, the Electric institutionalizaton of the program to review regulations with the Power Research Institute (EPRI), the nuclear industry, and from purpose of ehminating those that are marginal. The obgective is foreign governments and industry are also includod. The sum-to avoid the dilution of safety efforts. One session was devoted maries have been compiled 6n one report to provide a basis for

14 Main Citations and Abstracts rneaningful discussion and information exchange during the this bibliography were selected from proceedings of technical course of the n.eeting and are given in the order of their pres-meetings and conferences, journals, research reports, and entation in each session.

searches of the Energy Science and Technology database of the U S. Department of Energy. The subject materiaf of these NUREG/CP-0134: INTERNATIONAL ATOMIC ENERGY AGENCY abstracts relates to radiation protection and dose reduction, and SPECIAllSTS MEETING ON EXPERIENCE IN ranges from use of robotics to operational health physics, to AGING, MAINTENANCE, AND MODERNIZATION OF INSTRU-water chemistry. Matenal on the design, planning, and manage-MENTATION AND CONTROL SYSTEMS FOR IMPROVING ment of nudear power stations is included, as well as informa-Ni tCLEAR POWER PLANT AVAILABILITY. Held At tion on decommissioning and safe storage offorts. Volume 7 Rockville,MD May 5-7,1993.

International Atomic Energy contains 29,3 abstracts, an author index, and a subject index.

Agency.

• Oak Ridge National Laboratory. October 1993 The author index is specific for this volume. The subject index is 580pp. 9312070237, 77350:147.

cumulative and lists all abstract numbers from volumes 1 to 7.

This report presents the proceedings of the Specialist's Meet, The numbers in boldface indicate the abstracts in this volume; ing on Experience in Aging. Maintenance and Modernization of the numbers not in boldface represent abstracts in previous vol-instrumentation and Control Systems for improving Nuclear umes.

Power Plant Availability that was held at the Ramada inn in Rockville, Maryland on May 5-7, 1993. The Meeting was pre.

NUREG/CR-3950 V08: FUEL PERFORMANCE ANNUAL sented in cooperation with the Electric Power Research insti-REPORT FOR 1990. PREBLE E.A.; PAINTER,C.L; ALVIS,J.A.;

tute, Oak Ridge National Laboratory and the International et al. Battelle Memorial Institute, Pacific Northwest Laboutory.

Atomic Energy Agency. There were approximately 65 partica-November 1993.138pp. 9312220161. PNL-5210. 77543.243 1

pants from 13 countries at the Meeting. The program chairman This annual report, the thirteenth in a senes, provides a brief was Jerry L. Mauck of the U S. Nuclear Regulatory Commission.

description of fuel performance dunng 1990 in commercial nu-NUREG/CR-2850 V11: DOSE COMMITMENTS DUE TO RADIO-clear power plants. Brief summanes of fuel design changes, fuel ACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES surveillance programs, fuel operating experience and trends, IN 1989 BAKER,D.A. Battelle Memorial Institute, Pacific North-fuel problems high-burnup fuel experience, and stems of general west Laboratory. February 1993.192pp. 9303120086. PNL-significance are provided. References to additionaf, more de-4221. 74239 001.

tailed information, and related NRC evaluations are included Population and individual radiation dose commitments have where appropriate.

been estimated from reported radionuchde releases from com-mercial power reactors operating dunng 1989. Fifty-year dose NUREG/CR-4214 R1P2A2: HEALTH EFFECTS MODELS FOR cornmitments for a one-year exposure from both liquid and at-NUCLEAR POWER PLANT ACCIDENT CONSEQUENCE mosphenc releases were calculated for four population groups ANALYSIS. Modification Of Models Resulting From Addition Of (infant, child, teen-ager and adutt) residing between 2 and 80 Effects Of Exposure To Alpha-Emitting Radionuclides.Part 11:

km from each of 72 reactor sites. This report tabulates the re-Scientific Bases For Health.m ABRAHAMSON,S. Wisconsin, suits of these calculation W~rng the dose commitments for Univ. of, Madison, WI. BENDER,M.A. Brookhaven National Lab-both water and airborne pathways for each age group and oratory. BOECKER.B.B.; et al. Inhalation Toxicology Research organ. Also included for each of the sites is an estimate of indi-Institute. May 1993. 87pp. 9306020013. LMF-136. 75082:159.

vidual doses which are compared with 10 CFR Part 50, Appen-Several studies designed to identify and quantify the potential dix I design objectives. The total collective dose commitments health effects of accidental releases of radionuclides from nu-(from both liquid and airborne pathways) for each site ranged clear power plants have been sponsored by the Nuclear Regu-from a high of 14 person-rem to a low of 0.005 person-rem for latory Commission. Report NUREG/CR-4214, Rev.1. Part 11 the sites with plants in operation and producing power during (NRC,1989a) desenbes in detail the most recent health effects the year. The anthmetic mean was 1.2 person-rem. The total models that have evolved from these efforts. Since the Part li population dose for all sites was estimated at 84 person-rem for report was published in 1989, two addenda to that report have the 140 milhon people considered at risk. The individual dose been prepared to 1) incorporate other scientific information re-commitments estimated for all sites were below the Appendix I lated to low-LET health efiscts models and 2) extend the models to consider the possible health consequences of includ-design objectives.

ing alpha-emitting actinide radionuclides in the exposure source NUREG/CR-2907 V11: RADIOACTIVE MATERIALS RELEASED term. The first addendum was published as NUREG/CR-4214, FROM NUCLEAR POWER PLANTS. Annual Report 1990.

Rev.1. Part 11, Addendum 1 (NRC,1991). This report,. the TICHLER.J.; DOTY,K.; CONGEMI.J. Brookhaven National Labo-second addendum to the Part 11 report, extends the health ef-ratory. October 1993.360pp.9311080278. BNL NUREG-51581.

fects models to consider chronic irradiation from alpha-cmitting 77076:113 radionuclides as well as low-LET sources. Consistent with the Releases of radioactive matenals in airborne and hquid ef-organization of past reports, this report has three main sections fluents frorn commercial light water reactors dunng 1990 have that address earty-occurring and continuing effects, late somatic been compiled and reported. Data on solid waste shipments as effects, and genetic effects. These results should be used with well as selec'.ed operating information have been included. This the basic NUREG/CR-4214 report and Addendum 1 to obtain report supplements earlier annual reports issued by the former current views on potential health effects models for radionu-Atomic Energy Commission and the Nuclear Regulatory Com-clides released accidentally from nuclear power plants.

mission. The 1990 release data are summanzed in tabular form.

Data covering specife radionuclides are summanzed.

NUREG/CR-4214 R2 PT1: HEALTH EFFECTS MODEL FOR NU-NUREG/CR 3469 V07: OCCUPATIONAL DnGE REDUCTION AT CLEAR POWER PLANT ACCIDENT CONSEQUENCE NUCLEAR POWER PLANTS: ANNM ATED BIBLIOGRAPHY ANALYSIS.Part I: Introduction, Integration,And Summary.

OF SELECTED READINGS IN RAD ATION PROTECTION AND SCOTT,B.R.

Inhalation Toxicology Research Institute.

ALARA. KAURIN.D.G.; KHAN,T1,; SULLIVAN.S.G.; et al.

EVANS,J.S. Harvard School of Public Health, Boston, MA.

Brookhaven National Laborat-ry.

Jufy 1993.

106pp.

ABRAllAMSON.S.; et al. Wisconsin, Univ. of, Madison, WI. Oc-9308160137. BNL-NUREG 51708 '5120:001.

tober 1993. 200pp. 9311080251. ITRI-141, 77078:027.

The ALARA Center at Brookha n Nationat Laboratory pub-This report is a revision of NUREG/CR-4214, Rev.1, Part I hshes a series of bibliographies o! llected readings in radiation (1990), "Heafth Effects Models for Nuclear Power Plant Acci-protection and ALARA in the cont Ang effort to collect and dis-dent Consequence Analysis." This revision has been made to seminate information on radiatio a dose reduction at nuclear incorporate et anges to the Health Effects Models recommend-power plants. Th:s is volume 7 of the senes. The abstracts in ed in two addenda to the NUREG/CR-4214, Rev.1, Part it.

I i

)

l Main Citations and Abstracts 15 1989 report. The first of these addenda provided recommended leakage began. Using the virtual crack extension method, two changes to the health effects models for low-LET radiations through cracks with different lengths were found to be unstable based on recent reports from UNSCEAR, ICRP and NAS/NRC at this pressure which would allow almost instantaneous release (BEIR V) The second addendum presented changes needed to of the vessel contents.

incorporate alpha-emitting radionuclides into the accident expo-sure source term. As in the earlier version of this report, models NUREG/CR-4469 V15: NONDESTRUCTIVE EXAMINATION are provided for early and continuing effects, cancers and thy-(NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT road nodules, and genetic effects. Weibull dose-response func.

WATER REACTORS. Semiannual Report, October 1991 March tions are recommended for evaluating the risks of early and 1992. DOCTOR,S R.; DIAZ,A.A.; FRILEY,J.R.; et al. Battelle Me-continuing health effects. Three potentially lethal early ettocts-.

monal Institute, Pacific Northwest Laboratory. September 1993.

the homatopoietic, pulmonary, and gastrointestinal syndromes--

46pp. 9311010023. PNL 5711,76982;227.

are considered. Linear and linear-quadratic models are recom.

The Evaluation and Improvement of NDE Rehabikty for In-mended for estimating the risks of seven types of cancer in service inspection of Light Water Reactors (NDE Reliability) adults-leukemia, bone, lung, breast, gastrointestinal, thyroid, Program at the Pacific Northwest Laboratory was established by and "other." For most cancers, both incidence and mortality are the Nuclear Regulatory Commission to determine the reliability addressed Five classes of genetic disease-dominant. x hnked, of current inservice inspection (ISI) techniques and to develop aneuploidy, unbalanced translocations, and rnuftifactorial dis-recommendations that will ensure a suitably high inspection reli-eases-are also considered. Data are provided that should ability. The objectives of this program include determining the enable analysts to consider the timing and severity of each type reliability of ISI performed on the pnmary systems of commer-of health nsk.

cial light-water reactors (LWRs); using probabikstic fracture me-NUREG/CR-4219 V09 N2: HEAVY SECTION STEEL TECHNOL-chanics analysis to determine the impact of NDE unreliability on OGY PROGRAM. Semiannual Progress Report For April-Sep-system safety; and evaluating reliability improvements that can tember 1992. PENNELL.W.E. Oak Ridge National Laboratory.

be achieved with improvod and advanced technology. A final November 1993. 132pp. 9312070165. ORNL/TM-9593.

objectsve is to formulate recommended revisions to ASME Code 77355 061 and Regulatory requirements, based on material properties, The Heavy-Section Steel Technology (HSST) Program is con-service conditions, and NDE uncertainties. The program scope ducted for the Nuclear Regulatory Commission (NRC) by Oak is kmited to ISI of the primary systems including the piping, Ridge National Laboratory (ORNL). The program focus is on the vessel, and other components inspected in accordance with development and validation of technology for the assessment of Section XI of the ASME Code. This is a progress report cover-fracture 15revention margins in commercial nuclear reactor pres-ing the programmatic work from October 1991 through March sure vessels. The HSST Program is organized in 11 tasks: (1) 1992-program management, (2) fracture methodology and analysis' NUREG/CR-4469 V16: NONDESTRUCTIVE EXAMINATION (3) matenal charactenzation and properties, (4) special technical (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT assistance,'(5) fracture analysis computer programs, (6) cleav-WATER REACTORS Semiannual Report. April 1992-September age-crack initiation, (7) cladding evaluations, (8) pressurized-1992. DOCTOR,S.R.; DIAZ,A.A.; FRILEYAR ; et al. Battolle Me-thermal-shock technology, (9) analysis methods validation (10) fracture evaluation tests and (11) warm prestressing. The pro-morial Institute, Pacific Northwest Laboratory, November 1993.

gram tasks have been structured to place emphasis on the res-45pp. 9312170058. PNL-5711,77516:029, olution fracture issues with near-term licensing significance. Re-The Evaluation and improvement of NDE Reliability for In-sources to execute the research tasks are drawn from ORNL service inspection of Light Water Reactors (NDE Reliability) with subcontract support from universities and other research Program at the Pacific Northwest Laboratory was established by laboratones. Close contact is maintained with the sister Heavy-the Nuclear Regulatory Commission to determine the reliabihty Section Steel irradiation (HSSI) Program at ORNL and with re-of current inservice inspection (ISI) techniques and to develop lated research programs both in the United States and abroad.

recommendations that will ensure a suitably high inspection reli-This report providos an overview of pnncipal developments in ability. The objectives of this program include determining the l

cach of the eleven program tasks from Apnl 1992, to Septem-reliabihty of ISI performed on the pnmary systems of commer-ber 1992' cial light-water reactors (LWRs); using probabilistic fracture me-chanics analysis to determine the impact of NDE unreliabitity on NUREG/CR-4273: CRACK PROPAGATION IN HIGH STRAIN RE-system safety; and evaluating reliability improvements that can GtONS OF SEQUOYAH CONTAINMENT. GREIMANN,L.:

be achieved with improved and advanced technology. A final FANOUSE: BLUHM.D. Iowa State Univ Ames, IA. March objective is to formulate recommended rovisions to ASME Code 1993. 48pp 9304080044. IS-4878. 74494:243.

and Regulatory requirements, based on material proper-ties, The rate of release of radioactive matenals from a contain-service conditions, and NDE uncertainties. The program scope ment dunng a severe accident has a significant impact on the is limited to ISI of the pnmary systems including the piping, consequences of the accident. One hypothosis for a contain-vessel, and other components inspected in accordance with mont leakage model states that the containment will develop a Section XI of the ASME Code. This is a progress report cover-controlled relatively small leak before the pressure reaches the ing the programmatic work from April 1992 through September point where a general rupture of the shell occurs Another 1992.

states that overall failure will occur with total release of the i

wessel contents almost instantaneously. The Sequoyah ice con-NUREG/CR 4551 V7RIPI: EVALUATION OF SEVERE ACCI-dwer containment vessel has been studied for some time to DENT RISKS: ZION UNIT

1. Main Report. PARK,C.K.;

predic the possible location and extent of leakage which could CAZZOLi,E.; GRIMSHAW,C.; et al. Brookhaven National Labo, ocent dunng a severe accident. In this work, three cr't' cal high ratory. March 1993. 200pp. 9304080019. BNL/NUREG-52029.

i strain locations were studied to predict crack propagation from 74498:023.

an ntially small defect. The 1/2-inch plate near the Sequoyah in support of the Nuclear Regulatory Commission's (NRC's) spnngline was selected for further study. A detailed into ele-assessment of the nsk from severe accidents at commercial nu-ment model of the region was prepared and a virtual crack ed clear power plants in the U.S. reported in NUREG 1150, revised tension method for calcGating the J integral was developed for calculation of the risk to the general public from severe acci-use with the general purpose finito element program. The pres-dents at the Zion Power Station, Unit 1 has been completed, sure in the model was increased to 78 psi which produced a This power plant, located on the western shore of Lake Michi-maomum membrane strain of 6.5 percent. At this point the sur-gan on the outskirts of Zion,is operated by the Commonwealth face crack was assumed to propagate through the plate and Edison Company. The emphasis in this r!sk analysis was not on a.

l 16 Main Citations and Abstracts determining a "so<alled" point estimate of risk. Rather, it was evaluation of " Validity Limits on J-H Curve Determination" to determine the distribution of nsk, and to discover the uncer-(being conducted at Brown University) (5) Finally, technical of-tainties that account for the breadth of this distnbution. Off site forts reisted to the ASME Section XI pipe flaw evaluation efforts nsk initiation by events, both internal to the power station and are summanzed extemal to the power station, was assessed.

NUREG/CR 4667 V15: ENVIRONMENTALLY ASSISTED CRACK-NUREG/CR 4551V7 RIP 2A: EVALUATION OF SEVERE ACCl, ING IN LIGHT WATER REACTORS Somiannual Report,Apnt-DENT RISKS-ZION UNIT

1. Appendix A PARK,C K CAZZOLI,E,; GRIMSHAW,C.; et al. Brookhaven National Labo' September 1992. RUTHER W.E,; CHUNG,H M.; CHOPRA,0.K.;

ratsy M ch 1993. 600pp. 9304080029. BNL/NUREG 52029.

06 900 NL-9 5 96 07.

See NUREG/CR-4551,V07,R01,PT1 abstract This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking NUREG/CR-4551V7 RIP 20: EVALUATION OF SEVERE ACCI-(EAC) in light water reactors (LWRs) during the six months from DENT RISKS: ZION UNIT 1. Appendices B, C, D, And E.

Apnl 1992 to September 1992. Topics that have been investi-PARK C K.; CAZZOLt.E.; GRIMSHAW,C ; et al. Brookhaven Na-gated include: (1) fatiguo and stress corrosion cracking (SCC) of tional Laboratory. March 1993. 300pp. 9304080012. BNL/

low-alloy steel used in piping, steam generators, and reactor NUREG 52029. 74497:001-pressure vossols; (2) EAC of cast stainless steels (SSs); and (3)

See Nt.l REG /CR 4551,V07,R01,PT1 abstract.

radiation-induced segregation and irradiation assisted SCC of NUREG/CIj 4599 V02 N2: SHORT CRACKS IN PIPING AND Type 304 SS after accumulation of relatively high fluence. Data PIPING WELDS. Semiannual Report, October 1991 - March on fatigue of low alloy steel in LWR environments have been re-1092. WILKOWSKl,G.M.; BRUST F.; FRANCINI.R.; et al. Bat-viewed. Based on fracture-mechanics models and engineering telle Memorial Institute, Columbus Laboratones. May 1992.

judgement, interim fatigue design curves were developed that 53pp. 9306180295. BMI-2173. 75389:304.

are consistent with available fatigue. life data. Crack growth data This is thd fourth semiannual report of the U S. Nuclear Regu-were obtained on fracture-mechanics specimens of A533-Gr B latory Commission's Short Cracks in Piping and Piping Welds and A106-Gr B ferritic steels and on cast austenitic SSS in the research program. This 4-year program began in March 1990.

as-received and thermally aged conditions in simulated BWR The overall objective of this program is to venfy and improve water at 289 degrees C. The data were compared with predic-fracture analyses for circumferentially cracked large diameter tions based on crack growth correlations for ferritic steels in ox-nuclear piping with crack sizes typically used in leak-before-ygenated water and correlations for wrought austenitic SS in ox-break analyses or in-service flaw evaluations. Progress durin9 ygenated water developed at ANL and rates in air from Section this reporting penod involved- (1) completing two through-wall-XI of the ASME Code Microchemical and microstructural cracked pipe orperiments and supplementary matenal property changes in high-and commercial-purity Type 304 SS specimens data, (2) an internal circumferential surface cracked pipe experi' from control-blade absorber tubes and a control-bhde sheath ment was completed which showed that the R/t effects on the from operating BWRs were studied by Auger electron spectros-Net-Section-Collapse predicted loads for surface-cracked pipe copy and scanning electron microscopy. Slow-strain-rate-tensile to be independent of crack size, (3) the anisotropy investigation tests were conducted on irradiated specimens in air and simu-showed that pipe dimensions may be as important in determin' lated BWR water.

ing the out-of plane crack growtn anglo as the anisotropy of the toughness, (4) we initiated a probabilistic analysis of LBB to NUREG/CH-4667 V16: ENVIRONMENTALLY ASSISTED CRACK-assess the potential changes in the leakage detection enteria in ING IN LIGHT WATER REACTORS.

Semiannual NRC Reg Guide 1 AS, and (5) other efforts involved a sensitivity Report, October 1992 March 1993.

CHUNG H.M.;

study on the effect of thermal aging of cast stainless steel on CHOPRA,0.K.; RUTHER,W.E.; et al. Argonne National Labora-the moment-carrying capacity of the pipe as a function of time.

tory. September 1993. 67pp. 9310120274. ANL 93/27.

NUREG/CR-4599 V03 N1: SHORT CRACKS IN PIPING AND 76742.069.

PtPING WELDS Semiannual Report. April-Septomber 1992.

This report summarizes work performed by Argonne National WILKOWSKl,G M ; ORUST,F.; FRANCINI.R.; et al. Battelle Me-Laboratory on fatigue and environmentally assisted cracking i

morial institute, Columbus Laboratories. October 1993. 200pp.

(EAC) in light water reactors (LWRs) during the six months from j

9311080268. BMI.2173. 77077.218.

October 1992 to March 1993. Fatigue and EAC of piping, pres.

This is the fifth semiannual report of the U.S. Nuclear Regula-sure vessels, and core components in LWRs are irnportant con-tory Commission's Short Cracks in Piping and Piping Welds re-corns as extended reactor lifetimes are enusaged. Topics that soarch program. This 4 year program began in March 1990. The have been investigated include (1) fatigue of low-alloy steel overall objective of this program is to vonfy and improve frac.

used in piping, steam generators, and reactor pressure vessels, ture analyses for circumferentially cracked large-diameter nucle-(2) EAC of cast stainless steels (SSs), (3) radiation-induced ar piping with crack stres typically used in leak-before-break segregatson and irradiation-assisted stress corrosion cracking of analyses or in-service flaw evaluations. During this reporting Type 304 SS after accumulation of relatively high fluence, and period, the overall program and results were entically reviewed (4) EAC of low alloy steels. Fatigue tests were conducted on and consequently several changes to the current program were medium-Sulfur-content A106~Gr B piping and A53SGr B pibs-made to meet the final program objectives. Progress dunng this sure vessel steels in simuleted PWR water and in air. Additional reporting period involved: (t) for the surface-cracked pipe eval-crack growth data were obtained on fracture-mechanics speci-l l

uations, the tensile and Charpy V notch data for a carbon-man-mens of cast austenitic SSs in the as-received and thermally l.

ganese submerged arc weld metal were completed, and 3D aged conditions and chromiurn-nickel-plated A533-Gr B steel in finite-element (FE) analyses of uncracked stainless steel pipe simulated boiling-water reactor (BWR) water at 289 degrees C.

experiments to resolve the discrepancies between experimental The data were compared with predictions based on crack data and FE predictions were completed. (2) Significant leak-growth correlations for ferritic steels in oxygenated water and I

rate analyses for cracked pipe using advanced probabilistic correlations for wrought austenitic SS in oxygenated water de-j analysis was conducted to provide a technical basis for veloped at ANL and rates in air from Section XI of the ASME changes to NRC Reg Guide 145. (3) A new PC-based circum-Code Microchemical and microstructural changes in high-and forential surface-cracked pipe codo, NRCPIPES Version 1.0, commercial punty Type 304 SS specimens from control-blade was completed. (4) Subcontracted efforts include; numerical absorber tubes and control-blade sheath from operating BWRn analysis of residual stresses on elastic-plastic fracture of cracks were studied by Auger electron spectroscopy and scanning in welds (being conducted at the University of Michigan), and electron microscopy.

i l

l

+ m.

m

,-,mm., -.

.-r,i%

....y--+

--r w-_-.,,.---or mm-,

.m

g. _.,

H Main Citations and Abstracts 17 NUREG/CR-4735 V08: EVALUATION AND COMPILATION OF NUREG/CG-4832 V05: ANALYSIS OF THE LASALLE UNIT 2 NU-DOE WASTE PACKAGE TEST DATA Biannual Report August CLEAR POWER PLANT: RISK METHODS INTEGRATION AND 1989 - January 1990. INTERRANTE.C.G. Geology & Engineer-EVALUATION PROGRAM Parametor Estimation Analysis And ing Branch (Post 910506). FRAKER,A C; ESCALANTE,E, Na-Screening Human Reliability Andlysis.

WHEELER,T.A.;

tional Institute of Standards & Technology (formerly National SWAIN,A.D4 LAMBRIGHT.J.Aa et al. Sandia National Laborato-Bureau of Standa. June 1993.114pp. 9306290106. 75496:183.

ries. March 1993. 208pp. 9304190161. SAND 92-0537.

This report summarizes *, 'ons by the National institute 74643:007.

of Standards and Techno( %ST) of some of the Depart-This volume describes the methodologies used in the data ment of' Energy (DOE) activities on waste packages designed analysis, the screening human enor analysis, and the common for containment of radioactive high-level nuclear waste (HLW) rnode human error analysis performed in support of the LaSalle for the six-month penod, August 1989 January 1990. This in-PRA. Selected results are presented in this 91ume. The remain-

~

cludes reviews of related materials research and plans, informa-der of the results are presented in other v.mes of this report tion on the Yucca Mountain, Nevada disposal site activities, and where they are actually used. The data rs aw process used in other information regarding supporting research anti special as-the determination of the data used for FN initial screening anal-sistance. Short discussions are given relating to the pubhcations ysis is desenbed and the final screening data base is given. The reviewed and complete reviews and evaluationc are included.

final data selection process is described and the final data dis-3eports of other work are included in the Appendices.

tributions are presented. The actual implementation of the data base for the integrated accident sequence quantification is de-NUREG/CR-4744 V07 Nt: LONG-TERM EMBRITTLEMENT OF scnbed in Volume 2 of this report on Integrated Quantification CAST DUPLEX STAINLESS STEELS IN LWR and Uncertainty Analysis, Several new methods developed for SYSTEMS. Semiannual Report, October 1991 - March 1992.

use in analyzing both pre-and post-accident human errors for CHOPRA,0.K. Argonne National Laboratory. May 1993,152pp.

the initial screening analysis are desenbed. Most of the actual 9306180315. ANL-92/42. 75387 001.

results are given in other volumes of this report under the ap-This progress report summarizes work performed by Argonne propriate sub-analysis descriptions. A method for determining National Laboratory on long-term thermal embnttlemerit of cast procedural common mode analysis is desenbed and the results duplex stainless steels in LWR systems dunng the six months presented.

from October 1991 to March 1992. Charpy-impact, tensile, and fracture toughness J-R curve data are presented for several NUREG/CR 4832 VOS: ANALYSIS OF THE LASALLE UNIT 2 NU-heats of cast stainless steel that were aged 10,000-58,000 h at CLEAR POWER PLANT: RISK METHODS INTEGRATION AND 290, 320, and 350 degrees C. The results indicate that thermal EVALUATION PROGRAM (RMIEP). Seismic Analysis.

aging decreases the fracture toughaess of cast stainless steels.

WELLS.J E.; LAPPA.D A.; BERNREUTER.D.L; et al. Lawrence in general, CF-3 steels are the least sensitive to thermal aging Liverrnore Nationa! Laboratory. November 1993. 290pp.

and CF-8M steels are the most sensitive. The values of fracture 9312070214. UCID-212d5. 77349:002.

toughness J(IC) and teanng modulus for CF-8M steels can be This report desenbes the methodology used and the results as low as = 90 kJ/m(2) and = 60, respectively. The fracture obtained from the application of a simplified seismic nsk meth-toughness data are consistent with the Charpy-impact results, i e,

odology to the LaSalh County Nuclear Generating Station Unit unaged and aged steels that show low impact energy also

2. This study is part of the Level I analysis being performed by exhibit lower fracture toughness. All steels reach a minimum the Risk Methods integration and Evaluation Program (RMIEP).

saturation fracture toughness after thermal aging, the time to Using the RMIEP developed event and fault trees, the analysis I

reach saturation depends on the aging temperature. The results resulted in a seismically induced core damage frequency point also indicate that low-strength cast stainless steels are general-estimate of 6.0E-7/yr, This result, combined with the compo-fy insensitive to thermal aging.

nent importance analysis, indicated that system failures were dominated by random events. The dominant components in-NUREG/CR-4744 V07 N2: LONG TERM EMBRITTLEMENT OF cluded diesel generator failures (failure to swing, failure to start.

CAST DUPLEX STAINLESS STEELS IN LWR failure to run after started), and condensate storage tank.

SYSTEMS. Semiannual Report. April-September 1992.

CHOPRA,0.K. Argonne National Laboratory. July 1993. 54pp.

NUREG/CR-4832 V09: ANALYSIS OF THE LASALLE UNIT 2 NU-9308160133. ANL-93/11. 76120:301.

CLEAR POWER PLANT: RISK METHODS ;NTEGRATION AND This progress report summarizes work performed by Argonne EVALUATION PROGRAM (RMIEP). Internal Fire Analysis.

National Laboratory on long-term thermal embrittlement of cast LAMBRIGHT.J.A.; BROSSEAU,D.A.; PAYNE,A.C,; et al Sandia duplex stainless steels in LWR systems dunng the six months National Laboratones. March 1993. 600pp. 9304260106.

from April. September 1992. A procedure and correlations are SAND 92-0537, 74711.001.

presented for predicting Charpy4mpact energy tensile flow This report is a desenption of the intemal fire analysis per-stress, fracture toughness J-R cerves, teanng modulus and J(Ic) formed on the LaSalle County Nuclear Generating Station, Unit of aged cast stainless steels from known matenal information.

2. As part of this effort, a new data base for fires was construct-The " saturation *' impact strength and fracture toughness of a ed (NUREG/CR-4586). This data base aided in quantification of specific cast stainless steel, ie., the minimum value that would fire initiating event frequencies. The most detailed integration be achieved for the material after long. term service, is estimat.

between fire nsk assessment and internal events analysis, to ed frorn the chemical composition of the steel. Mechanical date, was also accomplished. The same system fault trees used a

properties as a function of time and temperature of teactor for snternal events were utilized for the fire analysis, which ir*-

service are estimated from impact energy and flow stress of the cluded modeling of components down to the contact pair level.

unaged material and the kinetics of embntilement, which are Subsidiary equations were created to map the effects of cable also determined from chemical composition. The J(tc) values failures and spurious actuations. All component and associated are determined from the estimated J-R curve and flow stress-cablo locations were traced and mapped into the fault trees. A Examples of estimating mechanical properties of cast stainless detated screening analysis was performed which showed most steel components dunng reactor service are presented. A plant areas had a negligible contribution to fire-induced core common " lower-bound" J-R curve for cast stainless steels of damage frequency. A detailed analysis of the fire risk resulted in unknown chemical composition is also defined for a given grade a total (mean) core damage frequency of 3.21E-5 per year.

of steel femte content, and temperature.

- - - + -

y,.p

~

,.-ew

-gi.-.y s

18 Main Citations and Abstracts NUREG/CR-5229 V05: FIELD LYSIMETER INVESTIGATIONS:

ed and the relationship between the codes and the needed LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM input and output data is discussed. Code listings for codes not FOR FISCAL YEAR 1992. Annual Report. MCCONNELLJW.;

documented elsewhere and complete or sample listings of the ROGERS,R D.; JASTROW.J.D.; et al. EG&G Idaho, Inc Febru-input and output files are also presented.

ary 1993. 68pp. 9303120065. EGG-2577. 74235 255.

NUREG/CR-5105 V02 P2: INTEGRATED RISK ASSESSMENT The Field Lysimeter investigattens: Low-Level Waste Data FOR THE LASALLE UNIT 2 NUCLEAR POWER Base Development Program, funded by the U.S. Nuclear Regu-latory Commission, is: (a) studying the degradation effects in PLANT:Phenomenology And Risk Uncertainty Evaluation Pro-EPICOR-il organic ton-exchange resins caused by radiation; (b) gram (PRUEP). Appendices D-G. BROWN T.D.; PAYNE,A.C.;

MILLER.L.A.; et al. Sandia National Laboratones. May 1993.

examining the adequacy of test procedures recommended in the Branch Technical Positon on Waste Form to moet the re-398pp 9306210303. SAND 92 2765. 75404:246.

See NUREG/CR 5305'V02,P1 abstract ~

quirements of 10 CFR 61 using solidified EPICOR-il resins; (c) obtaining performance informaton on solidified EPICOR-ll ion-NUREG/CR-5358: REVIEW OF ASME CODE CRITERIA FOR exchange resins in a disposal environment; and (d) determining CONTROL OF PRIMARY LOADS ON NUCLEAR PIPING the condit on of EPICOR-il kners. Results of the seventh year of SYSTEM BRANCH CONNECTIONS AND RECOMMENDA-data acquisition from the field testing are presented and dis

• TIONS FOR ADDITIOtal DEVELOPMENT WORK.

cussed. During the continuing field testing, both Portland type l-RODABAUGH.E.C; GWALTNiiY,R.C.; MOORE,S.E. Oak Ridge il cement and Dow vinyt ester styrene wasto forms are being National Laboratory. Nover' ber 1993. 52pp, 9312070270.

tested in lysimeter anays located at Argonne National Laborato-ORNL/TMel1572. 77352:23L ry East in Illinois and at Oak Ridge National Laboratory. The This report collects and us es available data to reexamine the study is designed to provide continuous data on nuclide release entena for controlhng pnrrdry loads in nuclear piping branch and movement, as well as environmental conditions, over a 20-connections as expresis in Section lit of the ASME Boiler and year penod.

Pressure Vessel Code. In particular, the pnmary load stress indi-NUREG/CR-5247 V01 R1: RASCAL VERSION 2.0 USER'S ces gis in NB-3650 and NB 3683 are reexamined. The report GUIDE. ATHEY.G.F. Phoenix Associates, Inc. SJOREEN A.L.

Concludes that the present usage of the stress indices in the Oak Ridge National Laboratory. RAMSDELLJ.V.; et al. Battelle entena equations should be continued. However, the complex Memorial Institute, Pacific Northwest Laboratory February 1993, treatment of combined branch and run moments is not support-187pp 9303150148 PNL 8454. 74242.001.

ed by available information. Therefore, it is recommended that The Radiological Assessment System for Consequence this combined loading evaluation procedure be replaced for pri-Analysis; Version 2.0 (RASCAL 2.0) has been developed for mary loads by the separate log evaluation procedure specified use by the NRC personnel who report to the sto of a nuclear in NC/ND-3653.3(c) and NC/ND-3653.3(d). No recommendation accident at the time of radiological emergencies. It supplements is made for fatigue or secondary load evaluations for C: ass 1 assessments based on plant conditions and quick estimates P' ping Further work should be done on the development,of based on paper methods and providen rough compansons to better enteria for treatment of combined branch and run EPA Protective Action Guidance and thresholds for acute health moment effects.

effects The system, which can be run on any DOS system. was NUREG/CR-5360: XSOR CODES USERS MANUAL. JOW.H-N,;

developed to allow consideration of the dominant aspects of MURFIN,W.B.; JOHNSONJD. Sandia National Laboratones.

source term, transport dose, and consequences. The model November ' 1993.

31Bpp, 9401030178. SAND 89-0943.

that was previously the whole of RASCAL has been renamed 77651:001 i

ST-DOSE. Two new models have been added to RASCAL 2.0-This report describes the source term estimation codes, The first, FM-DOSE. computes doses from environmental cofb XSORs. The codes are written for three pressurized water reac-centrations. The second. DECAY, computes radiologic decay tors (Surry, Sequoyah, and Zion) and two boiling water reactors and ingrowth over a selected time penod. This volume includes (Peach Bottom and Grand Gulf), The ensemble of codes has complete instructions for system use been named "XSOR". The purpose of XSOR codes is to ests NUREG/CH 5247 V02: RASCAL VERSION 2.0 WORKBOOK, mate the source terms which would be released to the atmos.

ATHEY,G F, Athey Consulting. MCKENNA.T J. Incident Re-phere in severe accidents. A source term includes the release sponse Branch. May 1993 105pp. 9306110031. 75321:069.

fractions of severat radionuclide groups, the timing and duration The Radiological Assessment System for Consequence Anal-of releases, the rates of energy release, and the elevation of re-ysis, Version 2.0 (RASCAL 2 0) has been developed for "se by leases The codes have been developed by Sandia National the NRC personnel who respond to radiological emergencies.

Laboratories for the U.S Nuclear Regulato y Commission (NRC)

This workbook is intended to complement the RASCAL 2.0 in support of the NUREG 1150 program. The XSOR codes are User's Guide (NUREG/CR-5247 Vol 1). The workbook con.

fast running parametric codos and are used as surrogates for tains exercises designed to familianze the user with the comput-detailed mechanistic codes. The XSOR codes also provide the er based tools of RASCAL through hands-on problem solving, capability to explore the phenomena and their uncertainty which The workbook is composed of four major sections. The first part are not currently modeled by the mechanistic codes. The uncer-is a RASCAL familiarization exercise to acquaint the user with tainty distnbutions of input parmeters may be used by an XSOR the operation of the forms, menus, on line ht'p, and documen.

code to estimate the uncertainty of source terms.

tation. The latter three parts contain exercises n using the three NUREG/CR-5404 V02: AUXILIARY FEEDWATER SYSTEM tools of RASCAL Version 2.0: DECAY, FM-DOSE, and ST-DOSE. Each section of exercises is followed by discussion on AGING STUDY. Phase i Follow-On Study. KUECK.J.D. Oak Ridge ' National Laboratory, July 1993. 39pp. 9308160268.

how the tools could be used to solve the problem ~

ORNL-6566/V1. 76118.327.

NUREG/CR-5305 V02 P1: INTEGRATED RISK ASSESSMENT This report documents the results of a Phase i follow on FOR THE LASALLE UNIT 2 NUCLEAR POWER study of the Auxiliary Feedwater (AFW) System that has been PLANT.Phonomenology And Risk Uncertainty Evaluation Pro-conducted for the U.S. Regulatory Commission's Nuclear Plant gram (PRUEP) Appendices A-C. BROWN.T.D.; PAYNE.A.C.;

Aging Research Program. The Phase I study found a number of

(

MILLER LA.: et at Sand:a National Laboratories. May 1993.

significant AFW System functions that are not being adequately 200pp. 9306210243 SAND 92 2765 75M8 056.

tested by conventional test methods and some that are actually This volume contains a descnption c the codes and input /

being degraded by conventional testing. Thus, it was decided output files used to perform the LaSalle Level ll/ill Probabilistic that this 10!!ow on study would focus on these testing omissions Risk Assessment. A chart showing the process flow is present-and equipment degradation. The deficiencies in current monitor-i

I l

4 Main Citationa and Abstracts 19 Ing and operating practice are catogonzod and ovaluated Areas The thron volumes of this report detail a standard invostiga-of component degradation caused by cunent practico are dis-tion process for use by Nuclear Regulatory Commission (NRC) cussed Recommendations are made for improved diagnostic personnel when investigating human performanco related methods and test procedures events at nuclear power plants. The process, called the Human NUREG/CR-5410: STATISTICALLY BASED REEVALUATION OF ormanco investigabon Rocess M% was hW M PISC Il ROUND ROBIN TEST DATA. HEASL ER,P.C ;

meet the special noods of NRC personnot especiaHy NRC resh TAYLOR,T T,; DOCTOR,S R. Battelle Memorial Instituto Pacific dont and regional inspectors HPIP is a systematic investigation Northwest Laboratory. May 1993.150pp. 3306020025. PNL-proctas combining current proceduros and field practicos, gr 77, 75082 001.

expert experience, NRC human performance research, and ap-i This report presents a ro-analysis of an intemational PISC-Il plicable investigation techniques The process la casy to loarn round-robin inspection tesults using formal statistical techniques and helps NRC personnel perform better field investigations of to account for exponmental error The analysis examinos. U S, the root causos of human performanco problems. lho human team performanco vs. Other participants performance; flaw performance data gathorod through such investigations provides sinng portormance and errors associated with flaw siang; fac.

a bottor understandmg of the human performance issues that tors influencing flaw detection probability; and performance of cause events at nuclear power plants. Volume lit is a detailed all participants with respect to recently developed ASME Soc.

documentation of the dovelopmont effort and the pilot training tion XI flaw detection performanco demonstration requiromonts, program.

and evelops conclusions concerning ultrasonic inspGetion ca-g AND REPORTING OF SINGLE AND MULTIPLE FAILURE NUREG/CR 5455 V01: DEVELOPMENT OF THE NRC'S HUMAN EVENTS WHITEHEAD.D.W Sandia National Laboratonos.

PERFORMANCE INVESTIGAllON PROCESS (HPlP)

PAULA.H M JDF Associatos, Inc. PARRY,G,W.; et al. NUS PARADIES,M; UNGER,L System improvements, Inc. October Corp. March 1993, 149pp. 9304190151, SAND 89-2562.

1993. 25pp. 9312070146. SI 92-101. 77355.191.

74622 204.

The three volumos of this report dofail a standard investiga-This document prosents recommendations on how the colloc-tion process for use by Nuclear Rogulatory Commission (NRC) tron and documentation of failure events at nuclear power pornonnel when investigating human performanco related ovents at nuclear power plants. The process, called the Human plants can bo improved These recommendations, if adopted.

Performanco lovestigation Process (HPIP), was developed to should enhance the robabibty improvemont and nsk assessment meet tho special noods of NRC personnot, especially NRC resi~

programs that are dependent on such information. The report dont and regional inspoctors. HPIP is a syt!omatic investigation concentrates on how the recommendations should provido the process combining curront proceduros and field practicos, information nocessary to improve the paramotor estimations for export expanonco, NRC human performanco research, and ap-both independent and dependent events in a probabilistic nsk plicablo investigation techniques The process is easy to learn assessment and alludos to the fact that this sanw information and helps NRC personnel perform bottor field investigations of can be used to enhance other nucloor power plant activities.

thn root causes of human performance problems.,The human Soveral existing data basos are reviewod and areas where infor.

performanco data gathered through such investigations providos mation is lacking, orther because certain information is not re-a better understanding of the human performance issues that quired to be reported or because required information was couw ovents at nuclear power plants. Volume i is a conciso de-simply not reported, are identified. Finally, data noods identified senption of the nood for tho human porformance investigation from recent PRAs are discussed, process, the process' components, the methods usod to devot' op the process, the methods proposed to test the process, and NUREG/CR-5488: RISK-BASED INSPECTION GUIDE FOR conclusions on tho process usefulness' THREE MILE ISLAND NUCLEAR STATION UNIT 1.

HARRISON,0.G ; GORE,B F.; VO.T.V.; et at Battelle Memonal NUREG/CH-5455 V02: DEVELOPMENT OF THE NRC'S HUMAN Instituto, Pacific Northwest Laboratory. February 1993. 132pp.

PE RF ORMANCE INVESTIGATION PROCESS (HPIP) 9303120073. PNL-7187. 74239:193 PARADIES.M ; UNGER.L. System improvements, Inc. October The level one probabilistic risk assessment (PRA) for Three 1993 32ppp 9312070222. Sl.02101. 77360.001.

Melo Island Nuclear Station Unit 1 (TMI.1) has boon analytod to The three volumes of this report detail a standard investiga-identify plant systems and components important to minimiting tion process for use by U.S Nuclear f gulatory Commission public risk, as measured by system contnbutions to the annual (NRC) personnel when investigating hur. n performance related probability of coro damage, and to identify the pnmary failure events at nuclear power plants The process, called tho Human modos of thoso components This report presents a monos of Ponormanco Investigation Process (HPIP), was developed to tables, organized by system and priontitod by risk importance, moet the special noods of NRC personnel, especially NRC resi-which identify components associated with over 95*4 of the ird dont and regional inspectors HPIP la a systematte investigation spectabla risk due to plant operation The systems addiossed, process combining current procodutos and field practicos.

onport expononco, NRC human performanco research, and ap' In desconding order of importance, are: tho Docay Hoat Homov-plicahlo investigation techniques, The process is easy to learn al, High Pressure injection Decay Heat Cooling Watar, AC and helps NRC pnrsonnel perform bottor field investigations of Power, Nuclear Servicos Cooling Water, Main Steam Emergon-the root causes of human porformance probioms The human cy Feodwater, Reactor Coolant System Pressuro Control, inter-performanco data gathored through such investigations providos mediato Closed Cooling Water, instrument Air, DC Power, and a botter vndorstanding of the human performanco issues that Engincored Safeguards Actuation Systems. This ranking is cause ovents at nuclear power plants. Volumo 11 is a field beod on the Fussol-Vosely importance Measure of nsk impor-tanco, ie., the traction of the 7tal annual probability of core manual for use by investigators whon performing event investi-pations Volumo il includes the HPIP Procedure, the HPIP Mod-damage which involves failurus of the system of interost.

ulos, and Appendices that provido extensive documentation of Though not involved in the provention of core damage and thus each imostigation technique' not ranked, containment protection systems are of fundamental importanco in proventing and minimizing public risk due to a re-NUREG/CR 5455 V03: DEVELOPMENT OF THE NRC'S HUMAN losso of radionuclidos, should core damage occur, Thorefore, PERFORMANCE INVESTIGATION PROCESS (HPIP).

containment protection systems are includod in this report, con-PARADIES.M ; UNGERL System improvements, Inc October sisting of: the Reactor Building isolation, Reactor Building 1993 110pp 9312070101. SI-92-101, 77360 251.

Spray, and Reactor Building Emergoney Cooling Systoms.

=

20.

Main Citations and Abstracts NUREG/CR-5591 V01 N2: HEAVY-SECTION STEEL IRRADIA-and accident conditions. Failure mechanisms, such as tube TION PROGRAM. Semiannual Progress Report For April-Sep-ejecton, tube rupture, global vessel failure, and localized vesel tember 1990. CORWIN,W.R. Oak Ridge Natonal Laboratoy.

creep rupture, were studied. Newly developed models and exist-November 1993. 55pp 9312070266. ORNL/TM.11568.

ing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a 77352.175-The pnmary goal of the Heavy Section Steel Irradiation broader range of conditions could be considered simultaneous-(HSSI) program is to provide a thorough, quantitative assess-ly, calculations relied heavity on models with closed-form or sim-mont of the effects of neutron irrad:ation on the matenal behav-phfied numerical solution techniques. Finite element techniques inr, and in particular the fracture toughness properties, of typical were employed for analytical model venfication and examanjng pressure vessel steels as they relate to hght-water reactor pres-more detailed phenomena. High temperature creep and tensile sure-vessel integnty. The program includes the direct continu-data were obtained for predicting vessel and penetratiori struc-ation of irradiation studies previously conducted within the tural response.

Heavy-Section Steel Technology program augmented by en-NUREG/CR-5672 V03: CHARACTERISTICS OF LOW-LEVEL RA-hanced examinatons of the accompanying microstructural DiOACTIVE DECONTAMINATION WASTE. Annual Report For changes Effects of specimen size, matenal chemistry, product form and microstructure, irradiation fluence, flux, temperature Fiscal Year 1992.

AKERS D.W.;

MCCONNELL,J.W.:

MORCOS.N. EG&G Idaho, Inc. February 1993. 68pp.

and spectrum, and post-irradiation annealing are being exam-ined on b wide rango of fracture properties. During this period 9303120059. EGG-2635. 74241:207.

detailed statistical analyses of the fracture data on K(Ic) shift of This document addresses the work performed dunng fiscal high-copper welds revealed greater shifts in fracture toughness year 1992 at the Idaho National Engineenng Laboratory by the than in Charpy transition temperatures. Testing of the duplex Low-Level Radioactive Waste-Decontamination Waste Program specimens from the second phase of the irradiated crack arrest (FIN A6359), which is funded by the U.S. Nuclear Regulatory testing on high-copper welds was initiated. Short-term aging Commission. The program evaluates the physical stabihty and studies were conducted on stainless steel weld-overlay clad-leachability of sohdified waste streams generated in the decon-ding. Additional determinations were made of chemistry ano un-tamination process of primary coolant systems in operating nu.

irradiated RT(NDT)s of the low upper-shelf weld metpl from the clear power stations. The data in this document include the Midland reactor and fracture toughness testing begun. An initial chemical composition and characterizabon of waste streams model desenbing the evolution of radiation-induced self-defect /

from Peach Bottom Atomic Power Station Unit 3 and from Nine solute clusters and other microstructures was developed and Mile Point Nuclear Plant Unit 1. The results of compressive expenments initiated to examine the effects of low-energy, low-strength testing on immersed and unimmersed solidified waste-lemperature neutron irradiations.

form specimens from Peach Bottom, and the results of leachate analysis are addressed. Cumulative fractional release rates and NUREG/CR-5631 R1 ADD-CONTRIBUTION OF MATERNAL RA.

leachability indexes of those specimens were calculated and D10NUCLIDE BURDENS TO PRENATAL RADIATION are included in this report.

DOSES Relationships Between Annual Limits On intake And Prenatal Doses. SIKOV,M R.; HUI.T.E. Battelle Memorial Insti-NUREG/CR-5699 V01: AGING AND SERVICE WEAR OF CON-P TROL ROD DRIVE MECHANISMS FOR BWR NUCLEAR 9 1080 97, PN 7745 7707 PLANTS. GREENE,R.H. Oak Ridge National Laboratory. No.

This addendum desenbes approaches for calculating and ex-vember 1992.149pp. 93031103'70. ORNL-6666. 74212:090.

pressing radiation doses to the embryo / fetus from matemal in, This Phcse i Nuclear Plant Aging Research (NPAR) study ex-takes of radionuclides at levels corresponding to fractions or amines the aging phenomena associated with BWR control rod multiples of the Annual Lim:ts on intake (All) Information con.

drive mechanisms (CRDMs) and assesses the ments of various cerning metabolic or dosimetnc charactenstics and the placen, methods of " managing" this aging. Informatio,n for this study tal transfer of selected, occupationally signahennt radionuclides was acquired from: (1) the results of a special CROM aging was presented in NUREG/CR-5631, Revision 1. That informa-questionnaire distnbuted to each U.S. BWR utility; L2) a first-of-von was used to estimate levels of radioactivity in the embryo /

its4:nd workshop held to discuss CADM aging and mainte-fetus as a function of stage of pregnancy ar.o time after entry.

nance concems; (3) an analysis of the Nuclear Plant Reliability Extension of MIRD methodology to accommodate gestationat-Data System (NPRDS) failure cases attr!buted to the control rod stage-dependent characteristics allowed dose calculations for dnve (CRD) system; and (4) personal information exchange with the simplified situation based on introduction of 1 yCi into the nuclear industry CRDM maintenance experts. Nearly 23% of the woman's transfer compartment (blood). The expanded scenar-NPRDS CRD system component failure reports were attributed los in this addendum include repeated or chronic ingestion or to the CRDM. The CRDM components most often requiring re-inhalation intakes by a woman during pregnancy and body bur-placement due to normal wear and aging are the Graphitar dens at the beginning of pregnancy. Tables present dose equiv-seals. The predominant cause,s of aging for those seals are me-alent to the embryo / fetus relative to intakes of these radionu, chanical wear and thermally induced embnttlement. More than clides in vanous chemical or physical forms and from pre-exist, 59% of the NPRDS CRD system failure reports were attributed ing matemal burdens corresponding to All, complementary t

components that compnse the hydraulic control unit (HCU).

intake values (fraction of an All and yCi) that yield a dose The predominant HCU components experiencing the effects of equivalent of 0 05 rem are included. Similar tables give those service wear and aging are valve seals, discs, seats, stems, measures of dose equivalency to the uterus from intakes of ra-packing, and diaphragms.

dionuclides for use as surrogates for embryo / fetus dose when biokinetic information is not availabie.

NUREG/CR 5747: ESTIMATE OF RADIONUCLIDE RELEASE NUREG/CR-5642 LIGHT WATER REACTOR LOWE9 HEAD CHARACTERISTICS INTO CONTAINMENT UNDER SEVERE FAILURE ANALYSIS.

REMPE,1L; CHAVEZ,S.A.;

ACCOENT CONDITIONS Final Report. NOURBAKHSH H.P.

THINNES,G.L.; et al. EG&G Idaho, Inc. October 1993. 450pp.

Brockhaven National Laboratory. November 1993, 142pp.

9401030188. BNL-NUREG-52289. 77652:001.

9311080148. EGG-26f 8. 77082:018.

A detailed review of the available light water reactor source This document presents the results from a U.S. Nuclear Reg-term information is presented as a technical basis for develop-ulatory Commission-sponsored research program to investigate ment of updated source terms into the containment under the mode and timing of vessel lower head failure. Major objec-severe accident conditions. Simplified estimates of radionuclide tives of the analysis were to identify plausible failure mecha-release and transport characteristics are specsfied for each nisms and to develop a method for datermining which failure mode would occur first in different light water reactor designs unique combination of the reactor coolant and containment I

l Main Citations and Abstracts 21 system combinations. A quantitative uncertainty analysis in the Overall, short-term and long term contractor personnel had the release to the containment using NUREG-1150 methodology is highest rates of positive tests. Licensee employees had fower also presented.

rates of positive test results.

NUREG/CR-5754: BOLLING-WATER REACTOR INTERNALS AGING DEGRADATION STUDY, Phase 1. LUK,K.H. Oak Ridge NUREG/CR 5759: RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN National Laboratory. September 1993. 56pp. 9310120363.

ORNL/TM-I tS76. 76740:333-PRESSURIZED WATER REACTOR PLANTS. SIMION.G.P. Sci-This report documents the results of an aging assessment ence Applications international Corp. (formerly Science Applica-study for boiling water reactor (BWR) internals. Mator stressors tions, Inc.). VANHORN.R.L; SMITH,C.L; et al. EG&G Idaho, Inc. June 1993. 250pp. 9306290067. EGG-2640. 75495:243.

for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel.

This report presents tho evaluation of the potential safety Welding and cold-working, dissolved oxygen and impurities in concerns for pressurized water reactors (PWRs) identified in the coolant, applied loads and exposures to fast neutron fluxes Generic Safety issue 106, Piping and the Use of Highly Com-aro other important stressors. Based on results of a component bustible Gases in Vital Areas. A Westinghouse fou:-loop PWR plant was analyzed for the risk due to the use of combustible failure information survey, stress corrosion cracking,legradation (SCC) and fatigue are identified as the two rnator aging-related gases (predominantly hydrogen) within the plant. The analytis mechanisms for BWR internals. Significant reported failures in-evaluated an actual hydrogen distnbution configuration and con-ciude SCC in jet. pump holddown beams, in-core neutron flux ducted several sensitivity studies to determine the potential vari-monstor dry tubes and core spray spargers. Fatigue failures ability among PWRs. The sensitivity studies were based on hy-were detected in feedwater spargers. The implementation of a drogen and safety-related equipment configurations observed at plant Hydrogen Water Chemistry (HWC) program is considered other PWRs within the United States. Several options for im-as a promising method for controlhng SCC problems in BWR.

proving the hydrogen distribution system design were identified More operating data are needed to evaluate its effectiveness and evaluated for their effect on risk and core darnage frequen-for intemal components. Long-term fast neutron irradiation ef-cy. A cost / benefit analysis was performed to determine whether fects and high-cycle fatigue in a corrosive environment are un-attematives considered were justifiable based on the safety im-certainty factors in the aging assessment process. BWR inter-provement and economics of each possible improvement.

nals are examined by vtsual inspections and the method is NUREG/CR-5766: AUXILIARY FEEDWATER SYSTEM RISK-access hmited. The presence of a large water gap and an ab-BASED INSPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 sonce of ex. core neutron flux monitors may handicap the use of NUCLEAR POWER PLANT. PUGH R.; GORE,B.F.; VO,T.V.; et advanced inspection methods, such as neutron noise vibration al. Batteile Memurial Institute, Pacific Northwest Laboratory.

measurements, for BWR.

February 1993. 36pp. 9302230239. PNL 7609. 64963:001, NUREG/CR-5755: STIFFNESS OF LOW. ASPECT RATIO, REIN, in a study sponsored by the U.S. Nuclear Regulatory Com-FORCED CONCRETE SHEAR WALLS. FARRAR,C.R. Los mission (NRC). Pacific Northwest Laboratory has developed and Alamos National Laboratory. BAKER W E. Now Mexico, Univ. of, apphed a methodology for deriving plant-specific nsk-based in-Albuquerque, NM. January 1993. 149pp. 9302020466. LA-spection guidance for the auxiliary feedwater (AFW) system at 12181-MS. 64729 054.

pressunzed water reactors that have not undergone probabilistic This report summarizes the information relating to stiffness of risk assessment (PRA). This methodology uses existing PRA re-low-aspect-ratio, reinforced concrete shear walls that has been sults and plant operating experience information. Existing PRA-obtained from stabc and dynamic tests of scale-model Seismic based inspection guidance information recently developed for Category 1 structures (exclusive of containment) and structural the NRC for various plants was used to identify generic compo-elements. Although numerous static and dynamic tests of shear nont failure modes. This information was then combined with wall elements are reported in the technical hterature, most of plant-specific and industry-wide component information and fail-these wero ultimate strength tests When these tests are eram-ure data to identify failure modes and failure rnechanisms for ined to determine stiffness values, there is a considerable range the AFW system at the selected plants. San Onofre-2 was se-in the results obtained. The types of structures and structural lected,as one of a series of plants for study. The product of this elements tested. the test procedures, the methods used to effort is a prioritized listing of AFW failures which have occurred measure stiffness (both directly and indirectly), and a summary at the plant and at other PWRs. This hsting is intended for use of the results are discussed. This report concludes by showing by NRC inspectors in the preparation of inspection plans ad-the changes in stiffness of shear walls as a function of the peak dressing AFW risk-important components at the San Onofre-2 nominal base shear stress that the structure expenences dunng plant.

a seismic event NUREG/CR 5776:

DAMPING IN LOW. ASPECT-NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR RATIO. REINFORCED CONCRETE SHEAR WALLS.

POWER INDUSTRY. Annual Summary Of Program Performance FARRAR,C.R. Los Alamos National Laboratory. BAKER W.E.

Reports,CY 1992. FLEMING,T.; WESTRA.C.; FIELD,l.; et al.

New Mexico, Univ. of, Albuquerque, NM. May 1993. 83pp.

Battelle Human Affairs Research Centers. July 1993. 140pp.

9306210337. LA 12201 MS. 75403:242.

9308100128. PNL-8688. 76142.200.

This report summarizes the information obtained from static This report summarizes the data from the semiannual reports ind dynamic tests of scale-model Seismic Category 1 structures on fitness-for duty programs submitted to the NRC by 54 utihties

. sxclusive of containment) on the damping of low-aspect ratio, for two reporting penods: January 1,1992 to June 30, 1992.

unforced concrete shear walls. The report reviews experimen-and from July 1,1992 to December 31,1992. During CY 1992, 14 i assessments of damping in low-aspect-ratio shear walls that hcensees reported that they conducted 266,551 tests for the h ve been reported in the hterature and presents a summary of presence of illegal drugs and alcohol. Of those tests,1,818 the types of structures and structural elements tested. It dis-(68%) were confirmed positive. Positive test results varied by cusses the testing methods and the methods used to determine category of test and category of worker. The ma}onty of positive equivalent viscous damping ration (both directly and indirectly),

test results (1,110) were obtained through pre-access testing, a numerical study that examines the accuracy of various meth.

Of tests conducted on workers having access to the protected ods for estimating damping from measured acceleration input area, there were 461 positive tests from random testing, and and responso data, and tabulates the damping results. The 178 positive tests from for-cause testing. Followup testing of report concludes by graphically showing the changes in the workers who had previously tested positive resulted in 69 posi-damping of the shear walls as a function of the peak nominal tive tests Posrtive test results also vaned by category of worker.

base shear stress experienced by the structure during simulated

)

e

22 Main Citations and Abstracts soismic events. Also included are comparisons of the damping failure modos, causes, and effects. The results af this evalua-results obtained in this program with those obtained by other in-tion, along with an assessment of component rt.storial and op-orating environment, lead to the conclusion that both the B&W vestigators.

and CE CRD systems are susceptible to age degradation. Fail-NUREG/CR-5778 V03: NEW YORK /NEW JERSEY REGIONAL ures of the CRD system have resulted in significant plant ef-SEISMIC NETWORK. Final Report For Apnl 1985 September tects including power reductions, plant shutdowns, scrams, and 1992. SEEBER,L; JOHNSON.D.; ARMBRUSTER,J. Lamont-Do-ESF actuations. Information on current plant system inspection herty Geological Observatory. July 1993. 74pp. 9307270035.

and maintenance practices were obtained from two B&W P ants, and four CE plants through an industry survey. The re-75802.317.

l For almost 20 years, Lamont-Doherty Earth Observatory has suits of this survey indicate that some plants have modified the operated the pnmary network for monitoring earthquake activity system, replaced comporients, and established preventivo main-in the New York State, northern New Jersey, and northwestern tenance programs, some of which effectively address the aging Vermont area, with support by both NRC and the USGS. The issue, while others do not. The potential application of some ad-pnmay purpose of this research is directed toward the determi-vanced monitonng inspection techniques are discussed.

nation of local seismicity and the possible identification of asso-ciated geologic and tectonic features, From April 1985 to Sep-NUREG/CR-5791: RISK EVALUATION FOR A GENERAL ELEC-tomber 1992, the notwork recorded and located 346 regional TRIC BWR, EFFECTS OF FIRE PROTECTION SYSTEM ACTU-earthquakes. Scientific activity, pnmanly in the form of after.

ATION ON SAFETY RELATED EQUIPMENT. Evaluation Of Ge-shock monitoring, was concentrated upon a number of signif'-

notic issue 57. LAMBRIGHT J.A.; ROSS.S. Sandia National cant earthquakes: The Ardsley, NY; the Chardon, OH; the Ash" Laboratories, KLAMERUS.E.; et al. Science & Engineering As-tabula, OH; and the Saguenay, Canada earthquakes; in addition sociates, Inc. December 1992. 312pp. 9301220181. SAND 91-to the Summit, NY event. These studies involved the deploy' 1536. 64652.001.

ment of portable seismographs in the epicentral areas. Many of Nuclear power plants have experienced actuations of fire pro-those sequences were in northeastern North Amenca, but out-taction systems (FPS) under conditions for which those systems side the L-DEO soismic network and were not covered by other were not intended to actuate, and also have experienced actu-permanent networks. Spatial correlations between structures ations with the presence of a fire. These actuations have often and earthquakes were found at a wide range of scalos, and sys-damaged nearby plant equipment. A review of past occurronces tomatic searches of archival rnatorial were used to improve con-of both types of such events on nuclear power plant safety, and straints on histonc sources' a cost-benefit analysis of potential corrective measures has NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABI-been performed. Thirteen different scenarios leading to actu-ation of fire protection systems due to a variety of causes were LISTIC FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR identified. A quantification of these thirteen sconanos, where ap-YANKEE ROWE REACTOR PRESSURE VESSEL DICKSON,T.L.; CHEVERTON,R D.; BRYSON,J.W.; et al. Oak plicable, was performed on a BWR4/MKl. This report estimates the Contribution of FPS actuations to core damage frequency, Ridge National Laboratory. August 1993.114pp, 9309210205.

Proposes physical modifications to reduce the risk from the ORNUTM-11945. 76484.336.

dominant contributors, and estimates the values and impacts of The Nuclear Regulatory Commission (NRC) roquested Oak Ridge National Laboratory (ORNL) to oerform a pressurized.

the proposed modifications.

therm =1 shock (PTS) probabilistic fracture mechanics (PFM)

NUREG/CR-5801: PROCEDURE FOR ANALYSIS OF COMMON-sensitivity ar.alysis for the Yankee Rowe raactor pressure CAUSE FAILURES IN PROBABILISTIC SAFETY ANALYSIS.

vessel, for the fluences corresponding to the end of operating MOSLEH,A. Maryland, Univ. of College Park, MD.

• Sandia Na-cycle 22, using a specific small-break loss of-coofant transient tional Laboratories. April 1993. 51pp. 9306010342, SAND 91-as the loading condition. Regions of the vessel with distinguish-7087.75062:001.

ing features were to be %ated individually--upper axial weld, This report provides practical guidelines for treatment of lower axial weld, circurr,erential weld, upper plate spot welds, common-cause f ailures (CCF) in nsk and reliability studies. The upoor plate regions berwoen the spot welds, iower plate spot procedures outlined in this report are organized according to welds, and the lower plate regions between the spot welds. The three phases of analysis, screening analysis, detailed qualitative fracture analysis methods us'ed in the analysis of through-clad analysis, and detailed quantitative analysis. The results of the surface flaws were those contained in the established OCA-P screoning analysis phase include conservatrve identification of computer code, which was developed dunng the Integrated potetial common-cause vulnerabilities and determination of the Pressurized Thormal Shock (IPTS) Program. The NRC request scope and focus for more detailed analysis in Phases 11 and 111.

specified that the OCA-P code be enhanced for this study to Phase !!, the detailed qualitative analysis, provides a better un-also calculate the conditional probabilities of failuro for subclad derstanding of the plant-specific susceptibilities of the systems flaws and ombedded flaws. The results of this sensitivity analy.

and components to causes and coupling mechanisms of CCF.

sis provide the NRC with (1) data that could be used to assess The information from this phase can then be used as a basis the relative influence of a number of key input parameters in the for a plant-specific quantitative assessment of CCF frequencies.

Yankee Rowe PTS analysis and (2) data that can be used for Detailed guidelines are provided for Phase ill to aid the analyst readily determining the probability of vessel failure once a more in using this qualitativo information and generic data in develop-accurato rndication of vessel embnttlement becomes available.

ing a plant-specific CCF base. Depending on the overall objec-This report is designated as HSST report No.117.

tive of the study, CCF analysis can stop at the end of any of the NUREG/CR-5783: AGING ASSESSMENT OF THE COMBUS-three phases.

TION ENGINEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES. GROVE.Ea GUNTHER.W. Drookhaven National NUREG/CR-58i7 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE Laboratory. January 1993. 219pp. 9302230172. BNL-NUREG-RESEARCH AT CNWRA. Calendar Year 1991 ABABOU R.;

BAGTZOGLOU,A,C.; CHOWDHURY,A.H.; et al. Contor for Nu-52299. 64978.001.

The offects of aging upon the Babcock & Wilcox'(B&W) and clear Waste Regulatory Analyses. May 1993. 500pp.

Combustion Engineenng (CE) control rod drive systems have 9306290022. CNWRA 91-01 A. 75494:144, been ovaluaiad For this study, the CRD systern boundary in-This is an annual status report on the results of research con-ciudod the control rod assembhos, guide tubes, control rod dove ducted on behalf of the U.S. NRC by the Center for Nuclear rnechanism, control system components, rod position indication Waste Regulatory Analyses in support of activitses under the components, and cooling system. Detailed operation expenonce Nuclear Waste Policy Act, as amended. Nino specific projects data for 1080 to 1990 was evaluated to identify the predorninant are underway; eight of which are reported here. The Goochem-l

- - _~ -.

~.

Main Citations and Abstracts 23 istry project is using laboratory methods and computer calcula-eling are described in the Unsaturated Mass Transport (Geo-tions to assess key geochemical constraints and to evaluate chemistry) project. Numerical simulation of a laboratory scale

. sorptrve properties of zeolites present at the proposed reposi-non-isothermal 1wo-phase flow is discussed in the Thermohy-tory sito. The Thermohydrology project has as its focus im.

drology chapter Methods for estimating rock joint roughness proved understanding of heat and fluid flow in unsaturated coefficient are the focus of the Seismic Rock Mechanics project media. Laboratory, field, and calculational studies are combined for which the Tilt Test, Tse and Crucen's equations, and fractal-in the Seismic Rock Mechanics project to examine the effects based equations were tested and found to be unsatisfactory. In of repeated seismic loadings on the rock-mechanical and hydro-the Integrated Waste Package Experiments chaptor, investiga-i logical responses of rock masses. The integrated Waste Pack-tions of pit initiation and repassivation potential for alloys 825 l

age Expenments have been initiated to evaluate degradation and C-22 and stainless steel 304L and 316L are described.

modes of candidate waste container alloys. Three-dimensional Testing of the BIGFLOW computer code and visualization of computer analysis techniques are being used to investigate spa-fracture topology is the theme of the Stochastic Hydrology tial variability of flow and transport in variably saturated frac-project. Preliminary analysis of field data from the Akrotiri site in tured porous media in the Stochastic Flow and Transport Greece is developed in the Geochemical Analogs project.

project. The recently initiated Geochemical Analogs project Mechanistic modeling of sorption using the MINTEOA2 code is seeks to investigate the role of such analogs in the licenssng investigated as part of the Sorption project. Adaptive gridding process, and is currently focused on characterizing and evaluat-and " modified equations" methods for solving the flow and ing a potential site for investigation. The Sorption Modeling transport equations are desenbed in the Performance Assess-Project has as its objective the evaluation and eventual selec-ment chapter. Finally, the Volcanism chapter focuses on using tion of model(s) of sorption processes which are deemed tech-nonhomogeneous Poisson processes for estimating probability nically acceptable in the context of repository licensing. Finally, of volcanic events at the potential repository site.

the Performance Assessment project is directed toward devel.

oping and evaluating methodolog es for evaluation of the long.

NUREG/CR-5818: UNCERTAINTY ANALYSIS OF MINIMUM term performance of the proposed repository.

VESSEL LIOUlO INVENTORY DURING A SMALL-BREAK NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE A B&W MMN ANAM & WE N WASTE RESEARCH AT CNWRA. January-June 1992.

a% k NW ABABOU,R.; AHOLA,M.P.; BACA,R.G.; et al. Center for Nuclear Waste Regulatory Analyses. May 1993.105pp. 9306210312.

1992. 89pp. 9302230168. EGG 2665. 64979:001.

CNWRA 92-01S. 75404:022.

The Nuclear Regulatory Commission (NRC) revised the emer-This is a semi annual status report on the results of research gency core cooling system licensing rule to allow the use of conducted c,a behalf of the U.S. Nuclear Regulatory Commis-best eshmate computer codes, provided the uncertainty of the sion by the Center for Nuclear Waste Regulatory Analyses in calculations are quantified and used in the licensing and regula-support of activities under the Nuclear Waste Policy Act, as tion process. The NRC developed a genenc methodology called amended. Nine specific projects are under way as reported Code Scaling, Applicabiirty, and Uncertainty (CSAU) to evaluate here. The Geochemistry Project staff is using laboratory meth_

best estimate code uncertainties. The objective of this work was oos and computer calculations to assess key geochemical con-to adapt and demonstrate the CSAU methodology for a small-straints and to evaluata sorptive proporties of zeohtes phasent break loss-of-coolant accident (SBLOCA) in a Pressurized at the proposed repository site. The Thermohydrology Project Water Reactor of Babcock & Wilcox Company lowered loop has as its focus improved understanding of heat and fluid flow design using RELAP5/ MOD 3 as the simulation tool. The CSAU in unsaturated media. Laboratory, field, and calculational studies methodology was, successfully demonstrated for the new set of are combined in the Seismic Rock Mechanics Project to exam-variants defined in this project (scenario, plant design, code).

ine the effects of repeated seismic loadinge on the rock-me-However, the robustness of the reactor design to this SBLOCA chanical and hydrological responses of rock masses. The Inte-scenano hmits the applicability of the specific results to other grated Waste Package Experiments have been initiated to plants or scenanos. Several aspects of the code were not exer-evaluate degradat on modes of candidate waste container cised because the conditions of the transient never reached alloys. Three-dimensional computer analyses techniques are enough severity. The plant operator proved to be a determining being used to investigate spatial variability of flow and transport factor in the course of the transient scenario, and steps were in ianably saturated fractured porous media in the Stochastic taken to include the operator in the model, simulation, and anal-Flow and Transport Project. The Geochemical Analogs Project yses.

staff s, eks to investigate the role of such analogs in the licens-NUREG/CR-5822: ANALYSIS OF THERMAL MIXING AND eng process, and is currently focusod on characterizing and evaluating two potential sites for investigation. The Sorption BORON DILUTION IN A PWR. SUN J.G.; SHA,W.T. Argonne Modeling Project has as its objectivo the evaluation and eventu-National Laboratory. February 1993, 94pp. 9303250058. ANL-91/43.74358:154.

al selection of model(s) of sorption processes which are Thermal mixing and boron dilution in a pressurized water re-deemed technically acceptable in the context of reposrtory I" actor were analyzed with COMMIX codes. The reactor system censing., The Performance Assessment Project is directed toward developing and evaluating methodologies for evaluation was a four-loop Zion reactor that was initially filled with hot boron-rich water. It was assumed that the reactor coolant of the long-term performance of the proposed repository. Final-pumps are tnpped. Following the tnp, cold unborated water from ly, in the recently initiated Volcanism Project, a comprehensive seal inlection or other sources continuously flows into the reac-evaluation of the state of knowledge regarding volcanism in the tor coolant system and dilution takes place first in the pump basin and range province has been completed, suction line and then in the reactor vessel. The thermal faixing NUREG/CH-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE and boron dilution under these conditions were analyzed. For WASTE RESEARCH AT CNWRA. July-December 1992, the analysis of thermal mixing, water at room temperature (re-

.ABABOU,R.; AHOLA M.P.: BACA.R.G.; et al. Center for Nuclear ferred to as cold water) was fed into the cold leg of the reactor Waste Regulatory Analyses. July 1993. 200pp. 9309090005, system at vanous flow rates. For the analysis of boron dilutio1, CNWRA 92-02S. 76393.001.

cold and hot unborated water was fed into the cold leg at a Progress frorn July 1 to December 31,1992 on the nine NRC.

high flow rate. The subsequent transient thermal mixing anc' sponsored research projects conducted at the Center for Nucle-boron dilution that would occur in the reactor system were simu-at Waste Regulatory Analyses is described. lon exchange ex-lated for 12 h, depending on the flow rate. A third analysis was penments between clinoptitolite and aqueous solutions of performed for the boron dilution after the start of the reactor Na( +) and Sr(2 +) aad three applications of reaction-path mod-coolant pump, which forces a slug of cold unborated water from

--=w

24 Main Citations and Abstracts the pump suction line into the reactor vesset This transient was plant-specifc and industry wide component information and fail-simulated for 20 sec, by which time the slug has been pushed ure data to identify failure modes and failure mechanisms for out of the reactor core. The rates of reactivity insertion were the AFW system at the selected plants. Fort Calhoun was so-evaluated for these analyses, lected as the sixth plant for study. The product of this effort is a pnon ed hsung of AFW fahes e have occM at N NUREG/CR-5829: AUXILIARY FEEDWATER SYSTEM RISK.

plant and at other PWRs. This listing is intended for use by BASED INSPECTION GUIDANCE FOR THE DAVIS-BESSE NU-NRC inspectors m the preparation of inspection plans address-CLEAR POWER PLANT. NICKOLAUS,J R.; MOFFITT,N.E.;

ing AFW risk-emportant components at the Fort Ctlhoun plant GORE,0 F.; et al. Battelle Memorial Institute, Pacific Northwest Laboratory. September 1993. 38pp. 9310120332. PNL-7905.

NUREG/CR-5835: AUXILIARY FEEDWATER SYSTEM RISK.

76742:139.

BASED INSPECTION GUIDE FOR THE BEAVER in a study sponsored by the U.S. Nuclear Regulatory Com-VALLEY. UNITS 1 AND 2 NUCLEAR POWER PLANTS.

mission (NRC), Pacific Northwest Laboratory has developed and LLOYD,R C.; VEHEC.T.A.; MOFFITT,N.E.; et al. Battelle Memo-applied a methodology for dertving plant-specific risk based in' flal institute, Pacific Northwest Laboratory. February 1993.40pp.

spection guidance for the auxiliary feedwater (AFW) system at 9303120101, PNL 7925 74236:271.

pressurtred water reactors that have not undergone probabilistic in a study sponsored by the U.S. Nuclear Regulatory Com-nsk assessment (PRA). This methodology uses exist ng PRA re-mission (NRC), Pacife Northwest Laboratory has developed and suits and plant operating experience information. Existing PRA-applied a methodology for deriving plant-specife risk-based in-based inspection guidance information recently developed for spection guidance for the auxiliary feedwater (AFW) system at the NRC for various plants was used to identify genonc compo-pressunred water reactors that have not undergone probabilistic nent failure modes. This information was then combined with risk assessment (PRA). This methodology uses existing PRA re-plant-specific and industry-wide component information and fail-suits and plant operating experience information. Existing PRA.

ure data to identify failure modes and failure mechanisms for based inspection guidance information recently developed for the AFW system at the selected plants. Davis-Besse was se-the NRC for various plants was used to identify generic compo-lected as one of a senes of plants for study. The product of this nent failure modes. This information was then combined with effort is a prioritized listing of AFW failures which have occurred plant-specific and industry wide component information and fail-i at the plant and at other PWRs This listing is intended for use ure data to identify failure modes and failure mechanisms for by NRC inspectors in the preparation of inspection plans ad-the AFW system at the selected plants. Beaver Valley Units 1 dressing AFW nsk important components at the Davis Besse and 2 were selected as two of a series of plants for study. The plant.

product of this effort is a priontized listing of AFW failures which NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-have occurrod at the plant and at other PWRs. This testing is BASED INSPECTION GUIDE FOR THE H B. ROB!NSON NU.

intended for use by NRC inspectors in the preparation of in-CLEAR POWER PLANT. MOFFITT,N E.;

LLOYD,R.C.;

spection plans addressing AFW risk-important components at GORE,0 F.; et al Battelle Memorial Institute. Pacife Nonhwest Beaver Va!!oy Units 1 and 2,

'y.

August 1993. 37pp. 9308160303. PNL 7907.

NUREG/CR-5836: AUXILIARY FEEDWATER SYSTEM RISK-76 9 93 BASED INSPECTION GUIDE FOR THE PALO VERDE NUCLE-In a study sponsored by the U.S. Nuclear Regulatory Com.

AR POWER PLANT. BUMGARDNER.J.D4 MOFFITT,N.E.;

mission (NRC), Pacific Northwest Laboratory has developed and GORE,B.F.; et al. Battelle Memonal Institute, Pacife Northwest apphod a methodology for denying plant-specife nsk-based in-Laboratory. February 1993. 34pp. 9302230328. PNL 7908.

spection guidance for tt e auxiliary feedwater (AFW) system at 000d 20*

pressunzed water reactors that have not undergone probabdistic In a study sponsored by the U.S. Nuclear Regulatory Com-nsk assessment (PRA). This methodology uses existing PRA re-mission (NRC), Pacific Northwest Laboratory has developed and sults and plant operating exponence information. Existing PRA.

based inspection guidance information recontly de'veloped for applied a methodology for deriving plant specific risk-based in-spection guidance for the auxifiary feedwater (AFW) system at the NRC for various plants was used to identify generic compo-pressurized water reactors that have not undergone probabilistic nent faikke modes. This information was then combined with risk assessment (PRA). This methodology uses existing PRA re-plant-specific and industry-wide component information and fail-suits and plant operating experience information. Existing PRA-ure dhta to identify failure modes and fai!ure mechanisms for based inspoction guidance information recently developed for the AFW system at the selected plants. H. B. Robinson was se.

the NRC for various plants was used to identify generic compo-lected as orte of a series of plants for study. The product of this nent failure modes. This information was then combined with offort is a priontized listing of AFW failures whch have occurred P ant-specife and industry-wide component information and fail-l at the plant and at other PWRs. This hsting is intended for use ure data to identify failure modes and failure mechanisms for by NRC inspectors in the preparation of inspection plans ad-the AFW system at tho selected plants. Palo Verde was select-dressing AFW risk.important components at the H. B. Robinson ed as one of a series of plants for study. The product of this plant.

ofiort is a prioritized listing of AFW failures which have occurred NUREG/CR-5834: AUXILIARY FEEDWATER SYSTEM RISK' at the plant and at other PWRs. This listing is intended for use BASED INSPECTION GUIDE FOR THE FORT CALHOUN NU-by NRC inspectors in the preparation of inspection plans ad-CLEAR POWER PLANT.

MOFFITT.N E.;

GORE,B.F.;

dressing AFW risk 4mportant components at the Palo Verde j

VEHEC,T.A.; et al. Battelle Memorial institute, Pacific Northwest plants.

Laboratory. February 1993. 34pp. 9303120001, PNL-7906, I

74235:324.

NUREG/CR-5843: CORCON-MOD 3;AN INTEGRATED COMPUT.

l l

In a study sponsored by the U.S. Nuclear Regulatory Com-ER MODEL FOR ANALYSIS OF MOLTEN CORE-CONCRETE l

mission (NRC), Pacifc Northwest Laboratory has developed and INTERACTIONS. User's Manual.

BRADt.EY,D.R.;

(

apphed a methodology for deriving plant-specihc risk-based in-GARDNER,D.R.; BROCKMANN,J.E.; et al. Sandia National Lab-l spection guidance for the auxihary feedwater (AFW) systom at oratorios. October 1993. 278pp. 9312070049. SAND 92-0167.

l pressurized water reactors that have not undergone probabilistic 77353 001.

nsk assessment (PRA). This methodology uses existing PRA re-The CORCON Mod 3 computer code was developed to suits and plant operating experience information. Existing PRA-mechanistically model the important core-concrete interact #on based inspection guidance information recently developed for pheaomena, including those phenomena relevant to the assess-the NRC for vanous plants was used to identify generic compo-ment of containment failure and radionuchde release. The codo nent failure modes This information was then combined with can be applied to a wide range of severe accident scenanos

Main Citations and Abstracts 25 and reactor plants. The code represents the current state of the to be no evidence to suspect that isolation device failure is an art for simulating core debns interactions with concrete. This issue which should be studied further.

document composes the user's manual and gives a bnef de-scnption of the models and the assumptsons and limitations in NUREG/CR-5882: TRAC-B THERMAL-HYDRAULIC ANALYSIS the code Also discussed are the input parameters and the OF THE BLACK FOX BOILING WATER REACTOR.

code output. Two sample problems are also given.

MARTIN.R P. EG&G Idaho, Inc. May 1993. 68pp. 9306180322.

NUREG/CR-5844: AGING ASSESSMENT OF BISTABLES AND EGG-2677. 75387:169.

SWITCHES IN NUCLEAR POWER PLANTS LEE.8 S.;

Thermal-hydraulic anatyses of six hypothetical accident sce-VILLARAN.M.. SUBUDHl.M. Brookhaven National Laboratory.

nanos for tre General Electnc Black Fox Nuclear Project boiling January 1993. 213pp. 9302230104. BNL-NUREG 52318.

water reactor were performed using the TRAC-BF1 computer 64979:119.

code. This work is sponsored by the U.S Nuclear Regulatory Distables and process switches play vital roles in the instru-Commission and is being done in conjunction with future analy.

mentation and cornrol logic of a nuclear power station. To un.

sis work at the U.S. Nuclear Regulatory Commission Technical derstand the aging characteristics of these components, moro Training Center in Chattanooga, Tennessee. These accident than 5,000 NPRDS events and more than 1200 LERS were re-scenanos were chosen to assess and benchmark the thermal-viewed and analyted Telephone surveys were conducted and hydeauhc capabilities of the Black Fox Nuclear Project simulator nuclear plant site visits were made to collect information on op-at the Technical Training Center to model abnormal transient erating experience and the current status of these devices.

conditions, i

interaction with the equipment manufacturers provided further detai!s on the designs of histables and switches. The aging NUREG/CR 5883: HEALTH RISK ASSESSMENT OF IRRADIAT-charactenstics studied included the effects of aging on failure ED TOPAZ. NELSON.K.; BAUM,J.W. Brookhaven National Lab-frequency, failure mode, and fa:Iure cause. This study found that oratory, January 1993. 156pp. 9302220410. BNL-NUREG-two groups of bistables are in operation. The first group which 52330. 64901.001.

consists of older bastables, needs attention in the near future Irradiated topaz gemstones are currently processed for color concerning the decision on replacement or refurbishment; how-Improvernent by subjecting clear stones to neutron or high-

}

ever, the second group, which is' newer, still has time to energy electron irradiations, which leads to activation of trace manage aging concerns. Most of the original switches employed elements in the stones. Assessment of the nsk to consumers in reactor protection systems have been replaced with transmit-required the identification and quantification of the resultant ra-tors and bistables, and the trend shows that the remaining ones dionuclides and the attendant exposure. Representative stones will be replaced in the near future, Based on the results of the from Brazil, India, Nigeria, and Sri Lanka were irradiated and analyses, recommendations for better aging management are anatyred for gamma ray and beta particle emissions, using made, and the areas for future studies are identified sodium lodide and germanium spectrometers, and Geiger-I NUREG/CR-5851: LONG TERM PERFORMANCE AND AGING Muller, plastic and liquid scintillation, autoradiography, and ther-

)

CHARAGTERISTICS OF NUCLEAR PLANT PRESSURE moluminescent-dosimetry measurement techniques. Based on TRANSMITTERS.

HASHEMIAN,H.M ;

MITCHEL L.D.W.;

these studies and other information derived from published litet.

~

FAIN.R E.; et al. Analysis & Measurement Services Corp. March ature, dose and related risk estimates were made for typical 9

t 993 375pp. 9304020285. 74450 001.

user conditions. New cntena and methods for routine assays for This report presents the results of a comprehensive research acceptable release, based on gross beta and gross photon and development project conducted for the NRC to study the emasions from the stones, were also developed.

effects of normal aging on calibration and response time of nu-

};

clear plant pressure, level, and flow transmitters and to develop NUREG/CR-5884 V1 DRF: REVISED ANALYSES OF DECOM-l and validate new methods for testing the performance of the MISSIONING FOR THE REFERENCE PRESSURIZED WATER 4

j transmitters as installed in nuclear power plants. The project in.

REACTOR POWER STATION Effects Of Current Regulatory j

volved research in seven areas as follows: 1) Aging tests of And Other Considerations On The Financial Assurance.... Main

]

complete transmitter assemblies; 2) Aging tests of critical com.

Report. Draft Report For Comment. KONZEK,G.J.: SMITH,R.I.;

ponents of transmitters, 3) Testing the effects of sensing line BIERSCHBACH.M.; et al. Battelle Memonal Institute, Pacific f

longth, blockages, and voids on the response time of prossure Northwest Laboratory. October 1993. 200pp. 9311080246. PNL-t sensing systems; 4) Oil loss phenomenon in Rosemount and 8742.77078.182.

i otMr transmitters; 5) Validation of new methods for on-line test-With the issuance of the Decommissioning Rulo (June 27, ing of response times of pressure transmitters; 6) On-line detec-1988), nuclear power plant licensees are required to submit to tinn of oil loss in Rosemount transmitters; and 7) Analysis of Li-the U S. Nuclear Regulatory Commission (NRC) for review, de-j consee Event Report (LER) and Nuclear Plant Reliability Data commissioning plans and cost estimates. This reevaluation System (NPRDS) databases for failures of pressure sensing study provides somo of the needed bases documentation to tho systems in nuclear power plants.

NRC staff that will assist them in assessing the adequacy of the NUREG/CR-5863; RISK ASSESSMENT OF ISOLATION DEVICES licensee submittals This report presents the results of a review IN SAFETY ' SYSTEMS. CRAMOND,W R ; MITCHELLO D and reevaluation of the PNL 1978 decommissioning study of the i

Sandia National Laboratones MILLER S P.; et al Science Appl,_-

Trojan rulear power plant for the DECON, SAFSTOR, and cations International Corp. (formerly Science Applications, Inc ).

ENTOMB decommissioning alternatives. These alternatives now January 1993 196pp. 9302230219. SAND 92 0538. 64963 037.

include an initial 5-7 year period dunny which the spent fuel is "lectrod 91ators are used to ma.ntain electncal separation stored in the spent fuel pool, pnor to beginning major disassem-etwee 5

,ety and non-safety systems in nuclear power plants bly or extended safe storage of the plant. This report also in-

.ern is that these devices may fail allowing unwanted cludes consideration of the NRC requirement that decommis-

,,unals or energy to act upon safety systems, or preventing de, sioning actrvitien leading to termination of the nuclear license be sired signals from performing their intended function. While completed within 60 years of final reactor shutdown, consider-operational history shows many isolation device problems re.

ation of pacl< aging and disposal requaiements for Greater Than-quinng adiustments and maintenance, we could not find inci-Class C low-level waste, and reflects all costs in 1993 dollars, dents where there was a safety implication. Even hypothesizing Sensitivity of the total license termination cost to the disposal multiple smultaneous failures did not lead to ssgnif cant contn-costs at different low-tovel radioactive waste disposal sites, and hutions to core damage frequency. Although the analyses per-to different depths of contaminated concrete surface removal formed in this study were not extensive or detailed, there seems within the facihties are also examined I

L

m 26 Main Citations and Abstracts NUREG/CR-5884 V2 DRF: REVISED ANALYSES OF DECOM-apphed a methodology for deriving plant specific nsk-based in-MISSIONING FOR THE REFERENCE PRESSURIZED WATER spection guidance for the auxihary feedwater (AFW) system at REACTOR POWER STATION Effects Of Current Regulatory pressurized water reactors that have not undergone probabilistic And Other Considerations On The Financial nsk assessment (PRA). This methodology uses existing PRA re-Assurance.... Appendices. Draft Report For Comment.

suits and piant operating experionce information. Existing PRA+

KONZEK G.J.; SMITH,R1; BIERSCHBACH M; et al. Battelle based Inspection guidance information recently developed for Memorial Institute, Pacific Northwest Laboratory October 1993-the NRC for various plants was used to identify generic compo-400pp. 9311080238. PNL 8742. 77081:001-nent failure modes. This information was then combined with See NUREG/CR-5884,V01.DRF abstract-plant-specific and industry. wide component information and fail-NUREG/CR-5894: RADIONUCLIDE CHARACTERIZATION OF ure data to identify failure modes and failure mechanisms for REACTOR DECOMMISSIONING WASTE AND NEUTRON AC.

the AFW system at the selected plants. Point Beach was so-TlVATED METALS ROBERT SON.D E.;

THOMAS,C.W,;

lected as one of a series of plants for study. The product of this WYNHOFF,N L: et al. Battelle Memorial Institute, Pacific North-effort is a prioritized listing of AFW failures which have occurred west Laboratory. June 1993. 80pp. 9307060129. PNL 8106.

at the plant and at other PWRs. This listing is intended for use 75572:213.

by NRC inspectors in the preparation of inspection plans ad-This study is providing the NRC and licensees with a more dressing AFW nsk important components at the Point Beach comprehensive and defensible data base and regulatory as-plant.

sessment of the radiolog! cal factors associated with reactor de-commissioning and disposal of wastes generated during these NUREG/CR 5901: A SIMPLtFIED MODEL OF AEROSOL SCRUB-activities. The ob ectives of this study are being accomplished BING BY A WAT ER POOL OVERLYING CORE DEBRIS INTER.

l dunng a two-phase sampling, measurement, and ' assessment ACTING WITH CONCRETE. Final Report. POWERS.D.A.;

program involving the actual decommissioning of Shippingport SPRUNG,J.L. Sandia National Laboratones. November 1993 Station and the detailed analysis of neutron-activated materials 126pp. 9312160284. SAND 92-1422, 77509.158. -

from commercial reactors. The radiological charactenzation A classic model of aerosol scrubbing from bubbles rising studies of Shippingport docommissioning matenats have now through water is applied to the decontamination of gases pro-been completed, and analyses of dismantled piping and scab-duced dunng core debris interactions with concrete. The model, blod concrete have shown that neutron activation products' onginally developed by Fuchs, uesenbes acrosol capture ey dif-dominated by (60)Co, compnsod the residual radionuclide inven-fusion, sedimentation, and 6nertial impaction. This original model tory. Fission products and transuranic radionuclides were essen-for spherical bubbles is modified to account for ellipsoidal dis-tially absent. Waste classihcation assessments have shown that all decommissioning matenals (except reactor pressure vessel intomals) could be disposed of as Class A waste. Spent fuel fied in the application of the model to the decontamination of disassembly hardware from the Shippingport Core-3 was ana-aerosols produced during core debris interactions with concrete lyzed for long-lived activation products specihed in 10 CFR 61, by a water pool of specihed depth and subcooling. These un-and the hardware was classified with respect to 10 CFR 61 cortain variables include properties of the aerosols, the bubblos.

waste disposal rules. Niobium.94 and (63)Ni concentrations in the water and the ambient pressure. Ranges for the values of Inconel X750 and stainless steel components exceeded their the uncertain variables are defined based on the hierature and Class C hmits.

expenence. Probabihty density functions for values of these ure certain variables are hypothesized. The model of decontamina-NUREG/CR-5897: /UXlLIARY FEEDWATER SYSTEM RISK-tion is applied in a Monte Carlo sampling of the decontamina-BASED INSPECTION GUIDE FOR THE SOUTH TEXAS tion by pools of specified depth and subcooling. Results are PROJECT NUCLEAR POWER PLANT. BUMGARDNER,J.D.;

analyzed using a nonparametric, order statistical analysis that NICHOLAUS,J R ; MOFFITT.N E.; et al. Battelle Memonal Insti-tute, Pacific Northwest Laboratory. December 1993. 36pp.

allows quantitative differentiation of stochastic and phenomeno-9401140022, PNL 8104. 77797:301.

logical uncertainty. The sampled values of the decontamination 10 a study sponsorod by the U.S. Nuclear Regulatory Com.

factors are used to construct estimated probability density func.

mission (NRC), Pacihc Northwest Laboratory has developed and tions for the decontamination factor at confidence levels of apphed a methodology for deriving plant. specific nsk-based in-50%,90% and 95% The decontamination factors for pools 30 spection guidance for the auxihary foodwater (AFW) system at 50,100,200,300, and 500 cm deep and subcoohng levels of 0, pressunzed water reactors that have not undergone probabikstic 2, 5,10. 20, 30, 50, 70 degrees C are correlated by simple poly <

nsk assessment (PRA). This methodology uses existing PRA re-nomial regression. These polynomial equations can be used to suits and plant operating experience information. Existing PRA-estimate decontamination factors at prescribed confiderice based inspection guidance information recently devoloped for levels.

l the NRC for various plants was used to identify generic compo-l nent failure modes. This information was then combined with NUREG/CR-5903: VALIDATION OF SMART SENSOR TECHNOL-plant specihc and industry wide component information and fail-OGIES FOR INSTRUMENT CAL.lBRATION REDUCTION IN NU-l ure data 10 identify failure modes and failure mechanisms for CLEAR POWER PLANTS. HASHEMIAN,H.M.; MITCHELL.D.W.;

j the AFW system at the selected plants. South Texas Project PETERSEN,K.M.; et al. Analysis & Measurement Services Corp.

was selected as one of a senes of plants for study. The product January 1993.160pp. 9302230183. 64964:004.

of this effort is a poorit:2nd hsting of AFW failures which have This report presents the prehminary results of a research and occurred at the plant and at other PWRs This hsting is intended development project on the validation of new techniques for ork for use by NRC inspectors in the preparation of inspection plans line testing of calibration drift of process instrumentation chan-addressing AFW nsk-important components at the South Texas nels in nuclear power plants. These techniques generally in-Project plant.

volve a computer based data acquisition and data analysis NUREG/CR 5898: AUXILIARY FEEDWATER SYSTEM RISK.

system to trend the output of a large number of instrument l

DASED INSPECTION GUIDE FOR THE POINT BEACH NUCLE-channels and identify the channels that have dofted out of toler.

AR POWER PLANT. LLOYD R.C.; MOFFITT,N E.; GORE.B F ;

ance. This helps limit the calibration effort to those channels et al. Battetle Memorial Institute, Pacihc Northwest Laboratory.

which need the cahbration, as opposed to the cunent nuclear February 1993. 32pp. 9303120004. PNL-8105. 74240:225 industry practice of calibrating essentialty all the safety-related in a study sponsored by the U.S. Nuclear Regulatory Com-instrument channels at overy refuehng outage.

mission (NAC), Pacehe Northwest Laboratory has developed and

. - - - - - - ~ --

i Main Citations and Abstracts 27 i

NUREG/CR-5907: CORE-CONCRETE INTERACTIONS WITH croscope analyses indicated substantial depletion of copper in OVERLYING WATER POOLS.T he WETCOR-1 Test.

the matrix but no evidence of copper clustering. Statistical anal-BLOSE.R E. Ktech Corp. POWERS,D.A.; COPL'S.E.R ; et al.

yses of the Charpy and chemical composition results as well as Sandia ; National Laboratones. November 1993. 171pp.

interpretahon of the ASME procedures for RT(NDT) determina-9312070274. SAND 921563. 77352.001, tion are discussed.

The WETCOR 1 test of simuttaneous interactions of a high-temperature melt with water and a limestone / common-sand NUREG/CR-5917 V01: SENSITMTY AND UNCERTAINTY ANAL-i concrete is desenbed The test used a 34.1-kg melt of 76.8 w/o YSES APPLIED TO ONE DIMENSIONAL RADIONUCLIDE Al(2)O(3),16.9 w/o CaO, and 4.0 w/o SiO(2) heated by induc.

TRANSPORT IN A LAYERED FRACTURED ROCK.MULTFRAC tion using tungston susceptors. Once quasi-steady attack on

- Analytic Solutions And Local Sensitivities. GUREGHIAN,A.B.;

1 concrete by the melt was established, an attempt was made to WU,Y.-T.; SAGAR.B.; et al. Center for Nuclear Waste Regula-

]

quench the melt at 1850 K with 295 K water flowing at 57 liters tory Analyses. December 1992.

139pp. 9302230456.

per minute. Net power into the melt at the time of water addition CNWR/ 91010. 64959 001.

was 0.61 t 0.19 W/cm(3). The test configuration used in the This report documents the denvation and venfication of the WETCOR-1 test was designed to delay melt freezing to the closed form analytical solutions of the one-dimensional nond:s-walls of the test fixture This was done to test hypotheses con-persive and isothermal transport of a radionuclide in a layered cerning the inherent stability of crust formation when high-tem-system of saturated planar fractures coupled with diffusion into perature melts are exposed to water. No instability in crust for-the ad acent saturated rock matrix. The analytical solutions are i

mation was observed The flux of heat through the crust to the based on the Laplace transform method where the domains of water pool maintained over the melt in the test was found to be radionuclide migration in both fractures and rocit layers are one-0 5210.13 MW/m(2). Solidified crusts were found to attenuate dimensional and of the semi-infinite type, implying in this in-aerosol emissions dunng the melt concrete interactions by fac-stance that radionuclide diffusion from the fractures wall to the tors of 1.3 to 3 5. The combination of a solidified crust and a rock matnx may extend to infinity. The sorption phenomena in 30-cm deep subcooled water pool was found to attenuate aero-both fracture and rock matrix layers are desenbod by a linear sol emissions by factors of 3 to 15.

equ:libnum sorption isotherm. Two types of radionuclide release NUREG/CR-5911: SOURCE TERM EVALUATION FOR RADIO-modes are considered. the continuously decaying, and the peri.

ACTIVE LOW-LEVEL WASTE OtSPOSAL PERFORMANCE AS_

odicaHy fluctuating decaying source, which may, in turn, be sub-SESSMENT. COWGILLM.G.; SULLIVAN,T.M. Brookhaven Na.

ject to step and band release modes The initial concentrations tional Laboratory. January 1993. 97pp. 9302230094. BNL.

in the fracture and rock matrix layers may be assigned spatially NUREG-52334. 64968.001.

varying values in the case of the first, whereas uniform ones information compiled on the low 4evel radioactive waste dis-may be implemented in both cases.

posed at the three currently operating commercial disposal sites NUREG/CR-5917 V02: SENSITIVITY AND UNCERTAINTY ANAL-during the period 1987 1989 have been reviewed and proc.

YSES APPLIED TO ONE-DIMENSIONAL RADIONUCllDC essed in order to determine the total activity distnbution in terms of waste stream, waste classification and waste form. The TRANSPORT IN A LAYERED FRACTU3ED ROCK. Evaluation review identified deficiencies in the information currently being Of The Lsmit State Approach. WU,Y..T.; GUREGHIAN,A.B.;

I recorded on shipping manifests and the developmont of a uni-SAGAR.B.; et at Center for Nuclear Waste Regulatory Analy-ses. December 1992. 71pp. 9301220115. CNWRA9 t-010.

torm manifest is recommended. The data from waste disposed 64 during 1989 at one of the sites (Richiand, WA) were more de-l The t State approach is based on partitioning the parame-tailed than the data available dunng other years and at other i

8"**'.and thus were amenable to a more in depth treatment.

This included determination of the distribution of activity for is smaller than a chosen value (called the limit state), and the each radionuclide by waste form, and thus enabled these data other in which it is larger. Through a Taylor expansion at a suit-e M N W me We) to be evaluated in terms of the specific needs for improved 1

.g, n

ng mleases from waste packages. From the results, preliminary lists have been prepared of the isotopos which success and efficiency of the limit state method depends upon might be the most significant from the aspect of the develop-choosing an optimum point for the Taylor expansion. The point ment of a source term model~

in the parameter space that has the highest probability of pro-ducing the value chosen as the limit state is optimal for expan.

NUREG/CR-5914: CHEMICAL COMPOSITION AND RT(NDT) DE-sion. When the parameter space is transformed into a standard TERMINATIONS FOR MIDLAND WELD WF-70.

Gaussian space, the optimal expansson point, known as the NANST AD.R.K.; MCCABE.D E.; SWAIN.R.L; et aliOak Ridge Most Probable Point (MPP), has the property that its location on National Laboratory Docomber 1992. 379pp. 9302010102.

the Limit State surface is closest to the origin. Additionally, the ORNL 6740. 64728.001.

projections onto the parameter axes of the vector from the The Heavy-Section Steel Irradiation Program Tenth irradiation origin to the MPP are the sensitivity coefficients. Once the MPP Series has the objective to investigate the effects of radiation is determined and the Limit State surface approximated, formu-on the fracture toughness of the low upper-shelf submerged arc las (see Equations 4 7 and 4-8) are available for determining welds (B&W designation WF-70) in the reactor pressure vessel the probability of the performance measure being less than the of the canceled Midland Unit I nuclear plant. This report dis-limit state. By choosing a succession of limit states, the entire i

cusses determination of variations in chemical composition and cumulative distribution of the performance measure can be de-

)

reference temperature (RT(NDT)) throughout the welds. Speci-termined. Methods for determining the MPP and also for irn.

mens were machined from different sections and through thick-proving the estimate of the probability are discussed in this ness locations in both the beltline and nozzle course wolds. The report.

nil-ductdity transition temperatures ranged from -40 to -60 de-grees C (-40 and 76 degrees F) while the RT(NOT)s, controlled NUREG/CR 5922: MODULAR HIGH TEMPERATURE GAS-by the Charpy behavior, varied from 40 to 37 degrees C (-4 to COOLED REACTOR SHORT TERM THERMAL RESPONSE TO 99 degrees F). The upper shelf energies varied from 77 to 108 FLOW AND REACTMTY TRANSIENTS. CLEVELAND.J.C. Oak J (57 to 80 ft4b). The combined data revealed a mean 41 J (30, Ridge National Laboratory. Fobruary 1993. 55pp 9303110375.

ft lb) temperature of -8 degrees C (17 degrees F) with a mean ORNL/TM 12179. 74212.239.

upper. shelf energy of 88 J (65 ft4b) The copper contents range The research reported here has been conducted at the Oak from 0.21 to 0 34 wt % in the beltline weld and from 0.37 to Ridge National Laboratory for the Nuclear Regulatory Commis-0.46 wt % in the nozzle course weld Atom probe f+1d son mi-sion's Division of Regulatory Applications of the Office of Nucle-

m.

28 Main Citations and Abstracts ar Regulatory Research. The short-term thermal response of This report contains a compilation of information generated the Modular High Temperature Gas-Cooled Reactor (MHTGR) during the ISLOCA research program. Presented is a screening is analyzed for a range of flow and reactivity transients. These analysis and a procedures guide for performing an ISLOCA transients include loss of forced circulation without scram, spuri-evaluation. This methodology has been distilled from past analy-ous wtthdrawal of a control rod group, moisture ingress, control ses performed for the U.S. Nuclear Regulatory Commission and rod and control rod group ejections, and a rapid core coohng documented in a series of NUREG/CR reports. The methodolo-event. For each event analyzed, an event description, a discus-gy compnses five distinct steps: (a) containment penetration son of the analysis approach and assumptions, and results are screening; (b) interfaces for ISLOCA analysis; (c) mechanisms presented. When possible, results of these analyses are com-for failing the pressure boundary; (d) construction of event trees I

pared with those presented by the designers in the MHTGR and estimation of rupture probabilities; and (e) quantification of Probabilistic Risk Assessment. The importance of inherent the event tree. Included in the methodology are steps required safety features is illustrated, and conclusions are presented re-for a detailed human reliability analysis. In addition, this report garding the safety performar ce of the MHTGR. Recommenda-presents a BWR ISLOCA evaluation, a survey of PWR auxiliary tions are made for a more in depth examination of MHTGR re-building designs and identification of one design deemed most sponse for some of the analyzed transients. The coupled heat disadvantageous with respect to ISLOCA risk, and a PWR transfer-neutron kinetics medel is described in detail in Appen-ISLOCA cost / benefit analysis.

dix A.

NUREG/CR-5933: HIGH PRESSURE COOLANT INJECTION NUREG/CR-5926: SANS INVESTIGATION OF LOW ALLOY (HPCI) SYST EM RISK BASED INSPECTION GUIDE FOR STEELS IN NEUTRON IRRADIATED, ANNEALED, AND REIR-DRESDEN NUCLEAR POWER STATION UNITS 2 AND 3.

RADIATED CONDITIONS.

KAMPMANN.R.;

F RISIUS.F.;

SHIER,W.; VILLARAN,M.; GUNTHER,W. Brookhaven National HACKBARTH,H.; et at Institute for Matenals Research. Febru-Laboratory. February 1993. 48pp. 9303120079. BNL-NUREG-ary 1993. 44pp. 9303120100. MEA-2490. 74240:262-52343. 74241:280.

Small Angle Neutron Scattenng (SANS) expenments were A review of the operating experience for the High Pressure made on several low alloy steels and submerged-arc welds pro-Coolant injection (HPCI) system at the Dresden Nuclear Power totypic of nuclear reactor vessel construction. The objective was Station Units 2 and 3 is described in this report. The information the charactenzation of radiation enhanced and/or radiation-in-for this review was obtained from Dresden Licensee Event Re-duced precipitation contnbuting to mechanical property changes ports (LERs) that were generated between 1980 and 1989.

observed in tensile and notch ductility tests of the matenals.

These LERs have been categonzed into 23 failure modes that The matenals were irradiated in the UBR Test Reactor under have been pnontized based on probabilistic risk assessment closely controlled conditions. A portion of the samples were ex-considerations. In addition, the results of the Dresden operating amined in the 288 degrees C irradiated (I) condition; others expenence review havo been compared with the results of a were examined in the postirradiation annealed (IA) condition similar, industry wide operating experience review. This compari-and in the 288 degrees C reirradiated (IAR) condition. E' pen-son provides an indication of areas in the Dresden HPCI system x

mental vanables included matenal composition (pnmanly %Cu, that should be given increased attention in the priontization of

%P. %N content), postirradiation annealing temperature (454 inspection resources degrees C and 399 degrees C) rearradiation fluence level, and NUREG/CR 5934: HIGH PRESSURE COOLANT INJECTION I

neutron-fluence rate (~0.08, 0.7, and 9 x 10(12) n/cm(2)-s( 1),

I E > 1 MeV). The apparent influcnce of the described vanables (HPCI) SYSTEM RISK-BASED INSPECTION GUIDE FOR OUAD-CITIES STATION, UNITS 1 AND 2.

VILLARAN.M.;

on the size, number density, and composition of copper-nch TRAVIS,Ra GUNTHER,W. Brookhaven National Laboratory, precipitates was the pnmary focus of the SANS analyses SANS January 1993. 55pp. 9303190108. BNL-NUREG-52344.

l observations are related to measured notch ductility'and tensile I

property changes, with a view toward mechantstic explanation T e H g!k Pressure Coolant injection (HPCI) system has beer, he observed mechanical property trends for 1, IA, and lAR examined from a nsk perspective. A '3ystem Risk-Based Inspec-tion Guide (S-RIG) has been develuped as an aid to HPCI NUREG/CR 5927 V01: EVALUATION OF A PERFORMANCE AS-system inspections at Quad Cities Included in this S-RIG is a SESSMENT METHODOLOGY FOR LOW-LEVEL RADIOAC-discussion of the role of HPCI in mitigating accidents and a TiVE WASTE DISPOSAL FACILITIES Evaluation Of Modeling presentation of PRA-based failure modes which could prevent Approaches. KOZAK.M.W4 OLAGUE.N E.; RAO,R.R.; et al.

proper operation of the system. The S-RIG uses industry oper.

Sandia National Laboratories August 1993.87pp 9309210242.

ating expenence, including plant-specific illustrative examples to SAND 912802. 76486.118.

augment the basic PRA failure modes. It is designed to be used This report represents an update to our earlier reports on low-as a reference for both routine inspections and the evaluation level waste performance assessment. This update addresses of the significance of component failures, needed improvements and recommended approaches to the existing state of the art in modeling, treatment of uncertainty, NUREG/CR-5936: ENHANCEMENTS TO THE ACCIDENT PRE-and use of data Greater attention is paid to developing an inte.

CURSOR METHODOLOGY. BOHNHOFF,W.J.; DINGMAN,S E.;

grated approach to performance assessment than was done in CAMP,A.L. Sandia National Laboratones. February 1993.138pp.

earlier developments of the methodology. Furthermore, insights 9303120072. SAND 92-2109, 74241:001.

are being developed by participating in validation exercises, and A feasibility study for developing an improved tool and im-by evaluating which validatson data are needed to improve con.

proved models for performing event assessments is described.

fidence in the methodology. It is emphasized that the perform.

The study indicates that the IRRAS code should become the l

ance assessment methodology update is a work in progress; base tool for performing event assessments, but that modifica-the recommendations given here will form the general directions tions would be needed to make it more suitable for routine.use.

toward which the methodology is heading, but some of the spe.

Alternative system modeling approaches are explored and an cific approaches may continue to evolve as the research pro, approach is recommended that is based on improved train-level models. These models are demonstrated for Giand Gulf and gresses Sequoyah. The insights that can be gained from importance NUREG/CR-5928. ISLOCA RESEARCH PROGRAM Final Report.

measures are also demonstrated. The feasibility of using Individ-GALYEAN,W.J.; KELLY,D L; SCHROEDER.J Aa et at EG&G ual Plant Examination (IPE) submittals as the basis for train-Idaho, Inc. July 1993. 107pp. 9308160131. EGG-2685.

level models for precursor studies was also examined. The level 76120.108.

of reported detail was found to vary widely, but in general, the wn.

Main Citations and Abstracts 29 submittals would not provide sufficient information to fully define two short-term depressurized station blackout sequences (with the model. The feasibility of developing an industry nsk profile a dry cavity and with a flooded cavrty) and a Loss-Of-Coolant from Accident Sequence Precursor results and of trending pre-Accident (LOCA) concurrent with complete loss of the Emer-cursor results for individual plants la considered but not recom-gency Core Cooling System (ECCS) were analyzed for the mended because of data sparsity.

Peach Bottom Atomic Power Station (a BWR 4 with a Mark i NUREG/CR-5937: INTENTIONAL DEPRESSURlZATION ACCI-containment). The results indicate that for the sequences ana-DENT MANAGEMENT STRATEGY FOR PRESSURIZED lyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the y

WATER REACTORS.

BROWNSON,D.A/

HANEY,L.N.,

CHIEN,N D. EG&G Idaho, Inc. April 1993.165pp. 9305100005' MELCOR/CORBH Package predicts significantly longer time for EGG-2688. 74858:270.

vessel failure and reduced energy of the released material for in a previous investigation of the Surry nuclear power station, the station blackout sequences (when compared to the 3TCP it was concluded that intentional depressurization of the reactor results). MELCOR also calculated smaller releases into the en-coolant system (RCS) could prevent or mitigate the effects of vironment than STCP for the station blackout sequences.

direct containment heating (DCH) dunng a station blackout tran-NUREG/CR 5943: SENSITIVITY ANALYSIS AND BENCHMARK.

sient. Two strategies, early and late depressunzation, were in-ING OF THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

vestigated as methods to mitigate DCH. The investigation con-SUEN C.1; SULLIVAN,T.M. Brookhaven National Laboratory.

a cluded that since there are greater opportunities to recover July 1993. 8tpp. 9307270012. BNL-NUREG-52346. 75802:236.

plant functions before core damage occurs and operator re' To evaluate the tource term for low-level waste disposal, a sponse uncertainties are lessened, the strategy of late depres-comprehensive model had been developed and incorporated sunzation is preferred over earfy depressurization. The results of into a computer code, called BLT (Breach-Leach-Transport).

the Surry analysis were extended to other U.S. pressunzed Since the release of the original version, many new features water reactors (PWRs) in order to evaluate their capability to and improvements had also been added to the Leach model of successfully employ the late depressurization strategy to pre.

the code. This report consists of two different studies based on vent or mitigate DCH By applying appropriate scaling factors to the new version of the BLT code: 1) a sones of verification /sen-the selected key parameters, this evaluation resulted in the cat-sitivity tests; and 2) benchmarking of the BLT code using field egonzation of four PWR groups based upon their perceived late data. Based on the results of the venfication/ sensitivity tests, depressunzation capability. In this report, a PWR representabve we concluded that the new version represents a significant im-of eacn of the four PWR groups was chosen for detailed analy-provement and it is capable of providing more realistic simula-sis of its capabel:ty to intentionally depressurize employing the tions of the teaching process. Benchmarking work was carried late depressurization strategy. The phenomenological behavior, out to provide a reasonable level of confidence in the model hardware performance, and operational performance of these predictions. In this study, the experimentally measured release PWRs dunng the intentional depressurization strategy were con-curves for nitrate, technetium-99 and tritium from the saltstone siderett The phenomenological behavior was analyzed using lysimeters operated by Savannah River Laboratory were used.

the SCDAP/RELAPS/ MOD 3 severe accident analysis code, The The model results are observed to be in general agreement results of these evaluations were then extended to the remain-w th the experimental data, within the acceptable limits of un-ing PWRs comprising each PWR group' certainty.

NUREG/CR-5938: NATIONAL PROFILE ON COMMERCIALLY GENERATED LOW LEVEL RADIOACTIVE MIXED WASTE, NUREG/CR 5944: A CHARACTER 12ATION OF CHECK VALVE DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLE-KLEIN,1A.; MROCHEK,J E.; JOLLEY,R L.; et al. Oak Ridge Na-tional Laboratory. December 1992. 460pp. 9301220128. ORNL, AR POWER INDUSTRY, CASADA,D.A.; TODD,M.D.. Oak Ridge 6731. 64650 094.

National Laboratory. September 1993. 187pp. 9310120234.

This report details the findings and conclusions drawn from a ORNL-6734. 76739:342.

j survey undortaken as part of a joint U S. Nuclear Regulatory Check valve operating problems in recent years have resulted i

in s,gnificant operating transients, increased cost and decreased i

Commission and U S. Environmentat Protection Agency spon-sored project entitled " National Profile on Commercially Gener, system availability. As a result, additional attention has been ated Low-Levet Radioactive Mixed Waste." The overall objec-given to check valves by utilities (resulting in the formation of tive of the work was to compile a national profile on the vol, the Nuclear Industry Check Valve Group), as well as the U.S.

umes, charactenstics, and treatability of commercially generated Nuclear Regulatory Commission and the Amulcan Society of low-level mixed wasto for 1990 by five major facihty categones-Mechanical Engineers Operation and Maintenance Committee, academic, industnal, medical, and NRC/ Agreement State-li-All these organizations have the fundamental goal of ensunng consed government facilities and nuclear ublities. Included in reliable operation of check valves. A key ingredient to an engi-this report are descnptions of the methodology used to collect neenng oriented reliability improvement effort is a thorough un-and collate the data, the procedures used to estimate the mixed derstanding of relevant historical experience, A detailed review waste generation rate for commercial facilities in the United of historical failure data, available through the institute of Nucle-States in 1990, and the identification of available treatment er Power Operation's Nuclear Plant Rehability Data System, has boon conducted. The focus of the review is on check valve fall-technologies to meet apphcable EPA treatment standards (40 CFR Part 268) and, if possible, to render the hazardous compo-ures that have involved significant degradation of the valve in-nont of specific mixed waste streams nonhazardous. The report ternal parts. A variety of parameters are considered, including also contains information on existing and potential commercial size, age, system of service, method of failure discovery, the af-waste treatment facilities that may provide treatment for specific facted valvo parts, attributed causes, and corrective actions.

waste streams identified in the national survey, NUREG/CR-5949: ASSESSMENT OF THE POTENTIAL FOR NUREG/CR 5942: SEVERE ACCIDENT SOURCE TERM CHAR.

HIGH PRESSURE MELT EJECTION RESULTING FROM A ACTERISTICS FOR SELECTED PEACH DOTTOM SE-SURRY STATION BLACKOUT TRANSIENT, KNUDSON,DL; OUENCES PREDICTED BY THE MELCOR CODE.

DOBBE.C.A. EG&G Idaho, Inc. November 1993. 194pp.

CARBAJO,J.J Oak Ridge National Laboratory. September 9312170064. EGG-2689. 77515:039.

1993 345pp 9310120226. ORNL/TM 12229 76739.001.

Containment integnty could be challenged by direct heating The purpose of this report is to comparo in.containtnent associated with a high pressure melt ejection (HPME) of core source terms developed for NUREG 1159, which used the materials fol!owing reactor vessel breach during certain severe Source Term Code Package (STCP), with those generated by accidents. Intentional reactor coolant system (RCS) depressuri-MELCOR to identify significant differences For this companson, zation, where operators latch pressurizer relief valves open, has

30 Main Citations and Abstracts been proposed as an accident management strategy to reduce selected events was published in May 1991 Causal factors for nsks by mitigating the seventy of HPME. However, decay heat operator performance were identified and categorized in the in-levels, valve capacities, and other plant-specific charactonstics terim report. Subsequently 10 more onsite analyses have been determine whether the required operator action will be effective.

conducted. This is a report on the analysis of the 16 events.

Without operator action, natural circulation flows could heat se Summanos of the 16 operating events analyzed can be found in vessel RCS pressure boundanes (surge line and hot leg piping.

Appendix A.

steam generator tubes, etc.) to the point of failure before vessel NUREG/CR-5955: MATERIALS AND DESIGN BASES ISSUES IN breach, providing an alternate mechanism for RCS depressun.

ASME CODE CASE N4 7, HUDDLESTON.RL; ration and HPME mitigation This report contains an assess.

SWINDEMAN.R.W< Oan Ridge National Laboratory. Apnl 1993 ment of the potential for HPME dunng a Surry station blackout transient without operator action and without recovery. The as.

42pp 9306010335. ORNL/TM-12266. 75062:284.

A preliminary evaluation of the design bases (poncipally sessment included a detailed transient analysis using the ASME Code Case N47) was conducted for design and oper.

SCDAP/RELAPS/ MOD 3 computer code to calculate the plant ation of reactors at elevated temperatures where the timo de-response with and without hot leg countercurrent natural circu.

tation, with and without reactor coolant pump seal leakage, and pendent effects of creep, creep-fatigue, and creep ratcheting 1

with vanations on selected core damage progression param.

are significant. Areas where Code rules or regulatory guides eters. RCS depressurization-related probabihties were also eval.

may be lacking or inadequate to ensure the operation over the expected life cycles for the next-generation advanced high-tem-usted. primanly based on the code results.

0"*

U.S. " Nuclear Regulatory ' ommission, have been identified as NUREG/CR-5951: THE MANAGEMENT OF ATWS BY BORON

+

C INJECTION. DIAS,M P. Moratuwa, Univ. of Sri Lanka. YAN H.;

unresolved issues. Twenty two unresolved issues were identi-THEOFANOUS,T.G Cahfomia, Univ. of, Santa Barbara, CA.

fied and brief scoping plans developed for resolving these March 1993. 58pp 9304020304. 74451:037.

issues.

Expenmental simulations of the multidimensional raixing/strat-ification phenomena in the lower plenum of a Boiling Water Re-NUREG/CR-5956: CONSIDERATION OF UNCERTAINTIES IN actor during operation of the Standby Liquid Control System SOIL STRUCTURE INTERACTION COMPUTATIONS.

(SLCS) are reported. The simulations both at full-and 1/2-scale COSTANTINO,C.J ; MILLER.C.A. City College of New York, allow the demarcation of the fully entraining regime, which is New York, NY.

• Viking Systems international December 1992.

also interpreted in terms of an approximate consideration of 140pp. 9301220172. CEERC 91 105. 64680:046.

flow stability cntena. based on the local Froude number. These This report presents a summary of the results obtained in a results are combined with analyses of the subsequent dispor-study conducted to evaluate and quantify some important ef-sion (of entrained boron) throughout the primary system and in fects of soil-structure interaction (SSI) on the seismsc response combination with neutron diffusion and natural convection of Category I facilities. The current procedures utilized in SSI (power-flow-void coupling) predictions of reactor kinetic behav-evaluations typically use complex computer analyses to treat ior are made. On this basis the performance of SLCS dunng the various important aspects of the problem. The purpose of ATWS is assessed and a discussion on current Emergency Op-this study has been to provide input to the Staff for developing erating Procedures is offered.

consistent guidelines for assessing adequacy of these SSI com-putations as well as clarification of some portions of the SRP NUREG/CR 5952: EVALUATION OF CRACK POP-INS AND THE mlating to SSI calculations. The specific areas addressed in this DETERMINATION OF THEIR RELEVANCE TO DESIGN CON-study am (a) summaMng the enteria needed when using the SIDERATIONS. MCCABE.D E. Oak Ridge National Laboratory.

large computer codes used in SSI studies; (b) providing roQom-Fetsruary 1993.

30pp.

9303110379. ORNL/TM-12247.

mondations for specification of control point location for soil 74212:057.

sites; (c) providing specific critoria to allow the Staff to judge the The issue with regard to crack pop ins is to deterrnine if such adequacy of fixed base structural analyses, (d) the development events are significant to design considerations. The literature of expanded guidelines for inclusion of variability in soit proper-contains ample evidence of pop-in occurrences, but scant infor.

ties in the SSI calculations, and (e) the development of esti-mation is offered on how pop-ins should be handled as an issue mates of radiation damping inherent in the detailed numoncal for dosign problems. Because there are two types of cleavage analyses performed or SSI evaluations of typical Category I crack ongins the probiern was subdivided into two classes of structures.

matenals monolithic and weldments with bnttle zones. The weldment situation can be ar,alyzed as a crack-arrest toughness NUREG/CR 5957: SYSTEM 80 + (TM) CONT AINMENT - STRUC-capability problem, following the recommendations of Sumpter TURAL DESIGN REVIEW. GREIMANN,L.; FANOUS,F.;

et al. For monolithic materials, pop-ins are more dangerous, CHALLA.R.; et af. Iowa State Univ., Ames, lA. May 19934 112pp.

since they appear to be a part of the more commonly encoun-9306010330.IS 5083. 75063.001.

tered full cleavage K(Jc) instability distributions. A recommenda-A review of the structural design of the Combustion Engineer-tion is made on how to determine if pop-in events he outside of ing (CE) System 80 +(TM) steel containment was completed.

the larger body of K(Jc) instabihties. The evaluation procedum The stress analysis and the evaluation of the structure against recommended by the Amencan Socloty for Testing and Mater" buckling were performed by using DOSOR4 and BOSORS finite als for pop-ins seems to dismiss the possibihty that small crack difference software, respectively. The CE System 80 4 (TM) con-iumps can be a safety-related issue The present work suggests tainment was modelled as an axisymmetric shell consisting of that nearly all pop-in events, regardless of the magnitude of different segments and mesh points with the additional mass of crack jump, are relevant to safety issues.

the penet ations and appurtenance being smeared around the NUREG/CH 5953: STUDIES OF HUMAN PERFORMANCE circumference. The transition region was modelled using elastic DURING OPERATING EVE NTS 1990-1992. MEYER,0R; spnngs with a foundation modulus of 180 lbs/in(3). The HILL,S G ; STEINKE W F EG&G Idaho. Inc. January 1993.

stresses due to the individual loads (dead loads, internal and 86pp 0302010090. EGG-2690. 64727;001.

external pressures and temperatures) were computed using the I

in order to better evaluate the human factors influencing op-stross analysis option in the BOSOR4 program. The stresses I

erator performance dunng operating events at nuclear power from individual loads were combinnd according to ASME Code stations, the Office for Analysis and Evaluation of Operational into stress intonsities. Service Level B loadings produced a 20 Data (AEOD) of the U S. Nuclear Regulatory Commission (NRC) percent ovepstress in a small zone just above the transition initiated a project to perform onsite analysos of selected events region. All other stress intensities were within allowable limits.

An intenm report on the results of the analysis of the first six For the Systom 80 +(TM), the perfect shell with an elastic mate-

l l

Main Citations and Abstracts 31 rial was initially analyzeit The caiculated factor of safety values This report charactentes discrete sources of naturally-occur-were 2.3 (Level B) and 1.59 (Levels C and D). Finally, sensitivity nng and accelerator produced radioactive material (NARM) and studies were conducted to investigate the effects of mesh size estimates risks posed by the possession, use and disposal of and transition zone stiffness on the controlhng buckling load them. A distinction between discrete and diffuse NARM sources NUREG/CR-5958: TWO-PARAMETER FRACTURE MECHANICS',

s made with discrete sources being high actrvity, low volume and diffuse sources being low activity, high volume. Two nano-THEORY AND APPLICATIONS. O'DOWD.N P. Impenal College' cunes per gram is used as a separation guido between high and London, UK. SHlH C.F,. Brown Univ., Providence, RI.

• Navy' I w activity, although use of this value does not impact the re-Dept.

of.

February 1993.

45pp 9303120046-CDNSWCSMECR1692. 74238 225 port's conclusions. Most NARM is under regulatory control of O

A family of self.similar fields provides the two parameters re-ments are not uniform. Use in consumer products has dochnod quired to charactenze the full range of high and low-triaxiahty with virtually no production today; however, lack of information crack tip states. The two parameters, J and O, have distinct available conceming radiation exposures resulting form posses-roles: J sets the size scale of the process zone over which "O

large stresses and strains develop, wtWe O scales'the near tip stress distnbution relative to a high triawiakty reference stress a n p

u an asussh %w ng acd state. An immediate consequence of the theory is this: it is the toughness values over a range of crack tip constraint that fuity l

in nuclear medicine programs has recently increased. Available j

characterize the material's fracture resistance. It is shown that ra a n apose data regWng M hadng aM m in&

O provides a common scale for interpreting cleavage fracture w

sWhWsWMMh and ductde teanng data thus allowing both failure modes to be incorporated in a single toughness locus. The evolution of O, as NUREG/CR-5964:

SAPHIRE TECHNICAL REFERENCE plasticity progresses from small scale yielding to fully yielded MANUAL:lRRAS/ SARA VERSION 4,0.

RUSSELL,K.D.;

conditions, has been quantified for several crack geometries ATWOOD,C.L.: SATTISON M.B.; et al. EG&G Idaho, Inc. Janu-and for a wide range of material strain hardening properties. An ary i993. 75pp. 9302230366. EGG-2692. 64998:199.

indicator of the robustness of the J-O fields is introduced. O as This report provides information on the pnnciples used in the a field parameter and as a pointwise measure of stress level is construction and operation of Version 4.0 of the Integrated Reli-i discussed.

abihty and Risk Analysss System (IRRAS) and the System Anal-ysis and Risk Assessment (SARA) system. It summartzes the NUREG/CR-5959: HIGH PRESSURE COOLANT INJECTION fundamental mathematical concepts of sets and logic, fault (HPCI) SYSTEM RISK-BASED INSPECTION GUIDE FOR trees, and probability. The report then descnbes the algorithms ENRICO FERMI ATOMIC POWER PLANT, UNIT 2.

VILLARAN.Ma TRAVIS,R.; GUNTHER.W. Brookhaven National that these programs use to construct a fault tree and to obtain the minimal cut sets. It gives the formulas used to obtain the l

Laboratory. January 1993. 55pp 9303190116. BNL-NUREG.

52352. 74314 059 pr bability of the top event from the minimal cut sets, and the The High Pressure Coolant injection (HPCI) system has been formulas for probabilities that are 9ppropriate under various as-sumptions concoming repairability and mission time. It defines examined from a risk perspective. A System Risk-Based inspec-tion Guide (S. RIG) has been developed as an aid to HPCI the measures of basic event importance that these programs system inspections at the Enrico Fermi Unit 2 Nuclear Power can calculate. The report gives an overview of uncertainty anal-Plant. Included in this S-RIG is a discussion of the role of HPCI ysis ussg simple Monte Carlo samphng or Latin Hypercube sampling, and states the algorithms used by these programs to in mitigating accidents and a presentation of PRA-based failure generate random basic event probabilitios from various distnbu-modes which could prevent proper operation of the system. The ti ns. Further references are given, and a detailed example of S HIG uses industry operating expenonce, including plant-spo-1 cific illustrative examples to augment the basic PRA failure the reduction and quantification of a simple fault tree is provid-i ed in an appendix.

modes It is designed to be used as a reference for both routine Inspections and the evaluation of the significance of component NUREG/CR-5966: A SIMPLIFIED MODEL OF AEROSOL RE-failures.

MOVAL BY CONTAINMENT SPRAYS. POWERS.D.A. Sandia MUREG/CR 5961: POSTTEST DESTRUCTIVE EXAMINATION OF National Laboratories. BURSON,S.B. Severe Accident issues THE STEEL LINER IN A 1:6 SCALE REACTOR CONTAIN-Branch. June 1993. 180pp. 9306210236. SAND 92-2689.

75407:095.

MENT MODEL. LAMBERT.LD. Sandia National Laboratones.

February 1993. 40pp. 9302220439. SAND 921721. 64899.228.

Spray system in nuclear reactor containments are described.

A 1+ scale model of a nuclear reactor containment model lhe scrubbing of aerosols from containment atmospheres by was built and tested at Sandia National Laboratones as part of spray droplets is discussed. Uncertainties are identified in the a research program sponsored by the Nuclear Regulatory Com.

prediction of spray performance when the sprays are used as a means for decontaminating containment atmospheres. A mission to investigate containment integnty. The overpressure test was terminated due to leakage from a large tear in the mechanistic model based on current knowledge of the physical steel liner A limited destructive examination of the liner and an-phenomena involved in spray performance is developed. With chorage system was conducted to gain information about the this modol, a quantitative uncertainty analysis of spray perform-failure mechanism and is descnbed. Sections of hner were re_

ance is conducted using a Monte Carlo method to sample 20 moved in areas where liner distress was evident or where large uncertain quantities related to phenomena of spray droplet be-strains were indicated by instrumentation dunng the test. The havior as well as the initial and boundary conditions expected to be associated with severe reactor accidents. Results of the un-condition of the hner, anchorage system, and concrete for each of the regions that were investigated aro described The proba-cedaWy analysis are used to construct simphfied expressions ble cause of the observed posttest condition of the liner is dis-for spray decontamination coefficients. Two variables that affect cussed' aerosol capture by water droplets are not treated as uncertain; they are (1) 'O', spray water flux into the containment, and (2)

NUREG/CR-5962: HEALTH AND SAFETY IMPACTS FROM DIS-

'H'. the total fall distance of spray droplets. The choice of CRETE SOURCES OF NATURALLY OCCURRING AND AC-values of these variables is left to the user since they are plant CELERATOR PRODUCED RADIOACTIVE MATERIALS and accident specific. Also, they can usually be ascertained with (NARM) NUSSDAUMER.D. WIBLIN.C ; WELCH,L Advanced some degree of certainty. The spray decontamination coeffi-Systems Technology, Inc, February 1993 54pp. 9303120051.

cients are found to be sufficiently dependent on the extent of 74238.274.

decontamination that the fraction of the snitial aerosol remaining

32 Main Citations and Abstracts in the atmosphere, m(f), is exphcitly treated in the simplified ex-stram hardening material (strain hardening exponent (g) of 10).

These analyses suggest that d is an effective indicator of pressions.

both the accuracy of T-MBL estimates of J(o) and of apple,cabil-NUREG/CR-5968: POTENTIAL CHANGE IN FLAW GEOMETRY ity limits on evolving fracture analysis metnodologies (i.e. T-OF AN INITIALLY SHALLOW FINITE LENGTH, SURFACE MBL, J-0, and J/J(o)). Specifically, when 1#1 >0.4 these FLAW DURING A PRESSURIZED-THERMAL-SHOCK TRAN-analyses show that the T-MBL approximation of J(o)is accurate SIENT, SHUM,D.K.; BRYSON J.W.; MERKLE.J.G. Oak Ridge to within 20% of a detailed finite-element analysis. As "structur-National Laboratory. September 1993. 31pp. 9311010029 al type" configurations, i.e. shallow cracks in tension, generally ORNL/TM 12279. 76982:265.

have 101 >0.4, it appears that only an elastic analysis may This study presents preliminary estimates on whether an ini-be needed to determine reasonably accurate J(o) values for tially shallow, axially onented, inner-surf ace finite-length flaw in structural conditions.

a PWR-RPV would tend to elongate in the axial direction and/or deepen into the wall of the vessel during a postulated PTS ban.

NUREG/CR-5971: CONTINUUM AND MICROMECHANICS sient. Analysis results obtained based on the assumptions of (1)

TREATMENT OF CONSTRAINT IN FRACTURE. DODDS,R.H.

hnear-elastic material response, and (2) cladding with the same IWinois, Univ. of, Urbana, IL. SHIH,CI. Brown Univ., Providence,

/

toughness as the base metal, indicate that a nearly semicircular Rt. ANDERSON,T.L Texas A&M Univ., Cv;5ne Station. TX. July flaw would hkely propagate in the axial direction followed bY 1993.47pp.9308160294. UILU-ENG92-2014I 76 W 21 propagation into the wall of the vessel. Note that those results Two complementary methodologies are described to quantify correspond to initiation within the lower-shelf fracture toughness the effects of crack-tip stress tnaxiality (constraint) on the mac-temperature range, and that their general valid,ty within the roscopic measures of elastic-plastic fracture toughness, J and lower-transition temperature range remains to be determined.

CTOD. In the continuum mechanics methodology, two param-The sensitivity of the numerical results and conclusions to the eters, J and O, suffice to charactenze the full range of near-to following analysis assumptions are evaluated: (1) reference flaw environments at the onset of fracture. A micromechanics meth-geometry along the entwo crack front and especially within the odology is described which predicts the toughness locus using cladding region; (2) linear-elastic vs elastic plastic desenption of crack tip stress fields and critical J-values from a few fracture matenal response, and (3) base matenal-only vs bimaterial clad-toughness tests. A robust micromechanics model for cleavage ding base vessel-model assumption. The sensitivity evaluation fracture has evolved from tho observations of a strong, spatial 6ndicates that the analysis results are very sensitue to the self-similarity of crack-tip pnncipal stresses under increased above assumptions, This report is designated HSST Report No.

loading and across different fracture specimens. This report ex-139-piores the fundamental concepts of the J O description of NUREG/CR 5969: J AND CTOD ESTIMATION EQUATIONS FOR crack-tip fields, the fracture toughness locus and micromechan-ics approaches to predict the variability of macroscopic fracture SHALLOW CRACKS IN SINGLE EDGE NOTCH BEND SPECI.

toughness with constraint under elastic-plastic conditions. Com-MENS. KIRK,M.T.; DODDS.R H. lilinois, Univ. of, Urbana, IL.

• putational resuits are presented for a surface cracked plate con-Navy, Dept. of. July 1993. 28pp. 9308160121. UILU-ENG91 taining a 6:1 sembelliptical, a=t/4 flaw subjected to remote un-20f 3. 76121:333, iaxial and biaxial tension.

Fracture toughness values determined using shallow cracked single edge notch bend SE(B), specimens of structural thick-NUREG/CR-5972: EFFECTS OF NONSTANDARD HEAT TREAT-ness are useful for structural integrity assessments. Results MENT TEMPERATURES ON TENSILE AND CHARPY IMPACT from two dimensional plane strain finite-element analyses a e PROPERTIES OF CARBON STEEL CASTING REPAIR WELDS.

used to develop J and CTOD estimation strategies appropriate NANSTAD,R.K.; GOODWIN.G.M.; SWINDEMAN.M.J. Oak Ridge for application to both shallow and deep crack SE(B) speci.

National Laboratory. April 1993.118pp. 9304210258. ORNL/

mens. Crack depth to specimen width (a/W) ratios between TM-12280. 74677;192 0.05 and 0.70 are modelled using Ramberg.Osgood strain hard-Carbon steel castings are used for a number of different com-ening exponents (n) between 4 and 50. The estimation formulas pononts in nuclear power plants, including valve bodies and divido J and CTOD into small scale yielding (SSY) and large bonnets. Components are often repaired by welding processes, scale yielding (LSY) components. For each case, the SSY com' and both welded components and the repair welds are subject-ponent is determined by the linear ofastic stress intensity factor',ed to a varigty of postweld heat treatments (PWHT) with tem-K(1). The formulas differ in evaluation of the LSY component eratures as, high as 899 degrees C (1650 degrees F), well The techniques considered include: estimating J or CTOD from above the normal 593 to 677 degrees C (1100 to 1250 degrees plastic work based on load hne displacement (A(p1)/LLD), from F) temperature range. The temperatures noted are above the plastic work based on crack mouth opening $splacemont A1 transformation temperature for the matenals used for these (A(p1)/CMOD), and from the plastic component of crack mouth components. A test program was conducted to investigate the opening displacement (CMOD(pt)), A(p1)/CMOD provides the potential effects of such " nonstandard" PWHTs on mechanical most accurate J estimation possible-properties of carbon steel casting welds. Four weldments were fabricated, two each with the shielded-metal arc (SMA) and flux-NUREG/CR-5970: APPROXIMATE TECHNIQUES FOR PREDICT-corod-arc (FCA) processes, with a high-carbon and low-carbon ING SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS filler metal in each case. All four welds were sectioned and (JC). KIRK M.T,; DOODS.R H. Ilknois, Univ. of, Urbana, IL..

given simulated PWHTs at temperatures from 621 to 899 de-Navy, Dept of. July 1993. 38pp. 9308160116. UILU-ENG92, grees C (1150 to 1650 degrees F) in increments of 56 degrees 20!6. 76122.00f.

C (100 degrees F) and for times of 5,10,20, and 40 h at each This investigation examines the ability of an elastic T-stress temperature. Hardness, tensile, and Charpy V notch (CVN) analysis coupled with a modified boundary layer (MBL) solution impact tests were conducted for the as-welded and heat-treated to pred'ct strenses ahead of a crack tip in a variety of planar conditions. Results were plotted versus a time temperature rela-geometries The approximate stresses are used as input to estp tionship (tempenng parameter) to enable a more direct compari-mate the effective dnving force for cloavage fracture (J(o)) using son of the effects of the various PWHT conditions. Heat treat-the micromechanically based approach introduced by Dodds ments at 621 and 677 degrees C (1150 and 1250 degrees F) and Anttorsnn Finite element analyses for a wide variety of gave results amenable to prediction, and regression analyses planar cracked geometries are conducted which have elastic are presented for those conditions. Heat treatments at 732 to biaxiafily parameters (d) ranging from -0 99 (very low con.

899 degrees C (1350 to 1650 degrees F), however, resulted in straint) to +2.96 (very high constraint). The magnitude and sign substantial changes in mechanical properus of these SMA and e' $ mdicate the rate at which crack-tip constraint changes FCA welds, with the changes not amenable to prediction and with encreasing apphed load All results pertain to a moderately

Main Citations and Abstracts 33 l

highly dependent on the weld rnetal Heat treatments in that train-level databases were developed using the Peach Bottom temperature range should not be apphed to these matenals Unit 2 PRA and one train-level database was developed using without prior qualification for the intended use.

the Beaver Valley Unit 2 IPE. The development, use, limitations, NUREG/CR-5973: CODES AND STANDARDS AND OTHER and results of these train-level databases are discussed.

GUIDANCE CITED IN REGULATORY DOCUME NTS.

NUREG/CR-5977: A PERFORMANCE INDICATOR OF THE EF.

NICKOLAUS J.R.; VINTHER.R.W.; MAGUIRE MOFFITT; et al.

FECTIVENESS OF HUMAN. MACHINE INTERFACES FOR NU-Battelle Memonal Institute, Pacific Northwest Laboratory, Janu-ary 1993. 73pp. 9302230373. PNL 8462. 64960.001.

CLEAR POWER PLANTS.

MORAY,N.;

JONES,B.J.;

As part of the U.S. Nucicar Regulatory Commission (NRC)

RASMUSSEN,J.; et al. Illinois, Univ. of, Urbana. IL, January Standard Review Plan Update and Development Program, Pa.

1993.88pp.9302010095. UILU ENG92-4007,64727;091.

cific Northwest Laboratory developed a listing of industry con-Effectwe interfaces must call up operators' deep understand.

sensus codes and standards and other government and indus-ing of plant operation if operators are to deal effectively with try guidance referred to in regulatory documents, This hstmg normal operation and diagnosis of transients. The prese71 re-identifies the version of the code or standard cited in the regula-ser,rch examines the ability of a memory recall task to Micate tory document, the regulatory document, and the current ver-the ability of an interface to couple plant state to y ator soon of the code or standard 11 also provides a summary char-knowledge Novices, people with intermediate experience, and acterization of the nature of the citation. This listing was devel-experienced nuclear power plant operators viewed three kinds oped from electronc searches of the Code of Federal Reguta-of displays. They watched nine simulated transients and tried to tions and NRC's Buttotins, Information Noticos, Circulars, Ge-recall the values of vanables, or the states through which the neric Letters, Policy Staternents, Regulatory Guides. and Stand-plant passed, and to detect and diagnose the nature of the tran-ard Review Plan (NUREG-0800)-

sients The displays were simulated analog instruments, simulat-NUREG/CR-5975: INCENTIVE REGULATION OF INVESTOR.

ed analog with pressure-temperature graphics, and an animated OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY representation of the Rankine cycle. The recall tasks did not REGULATORS. MCKINNEY,M D.; ELLIOTT,D.B. Battelle Memo.

show promise as indirect performance indicators of the quality nal Institute, Pacific Northwest Laboratory. January 1993. 80pp.

of the interfaces, but the diagnosis test detected differences in 9302220421. PNL-8466. 64899.001.

the quality of the displays and the levels of expertise.

The U.S. Nuclear Regulatory Commission (NRC) penodically surveys the Federal Energy Regulatory Commission (FERC) and NUREG/CR-5978: SOURCE TERM ATTENUATION BY WATER state regulatory commissions that tegulate utihty owners of nu-IN THE MARK i BOILING WATER REACTOR DRYWELL.

clear power plants. The NRC is interested in 6dentifying states POWERS,0.A Sandia National Laboratories. September 1993.

that have established economic or performance incentwe pro-200pp. 93 t 1010038. SAND 92 2688. 76983:001, grams applicable to nuclear power plants, including states with Mechanistic modets of aerosol decontamination by an overty-new programs, how the programs are being implemented, and ing water pool dunng core debns/ concrete interactions and in determining the f,nancial impact of the programs on the utili.

spray removal of aerosols from a Mark i drywell atmosphere are i

ties. The NRC interest stems from the fact that such programs developed. Eighteen uncertain features of the pool decontami-have the potential to adversety affect the safety of nuclear nation model and 19 uncertain features of the model for the power plants. The cuirent report is an update of,NUREG/CH-rate coefficient of spray removal of aerosols are identified.

4911, Incentive Regulation of Nuclear Power Plants by State Ranges for values of perameters that charactenze these uncer-Regula! ors, pubbshed in February 1991. The information in this tain features of the models are established. Probability density report was obtained from interviews conducted with each state functions for values within these ranges are assigned according regulatory agency that administers an incentive program and to a set of rules. A Monte Carlo uncertainty analysis of the de-each utility that owns at least 10% of an affected nuclear power contamination factor produced by water pools 30 and 50 cm 4

plant, The agreements, orders, and settlements that form th basis for each incentive program were reviewed as required.

doeP and subcooled 0-70 K is perfo.med. An uncertainty anafy-l The interviews and supporting documontation form the basis for sis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0 001 cm(3) H(2)O/cm(2)-S 1

the individual stato reports desenbang the structure and financial impact of each incentive program.

and decontamination factors of 1.1, 2, 3.3,10,100, and 1000, MUREG/CR 5976: DEVELOPMENT AND USE Tr A TRAIN-NUREG/CR-5980: THREE DIMENSIONAL REDISTRIBUTION OF LEVEL PROBABILISTIC RISK ASSESSMEN; SMITH,C L; TRITIUM FROM A POINT OF RELEASE INTO A UNIFORM FOWLER.R.D.; WOLFRAM.LM. EG&G Idaho, Inc. Apnl 1993-UNSATURATED SOll.A Deterministic Model For Tntium Migra-74pp. 9306010327. EGG-2694. 75062.209-tion in An Arid Disposal Site. SMILES.D E.; GARDNER,W R ;

The Idaho National Engineenng Laboratory examined the po-SCHULZ,R.K. Cakfomia, Univ. of, Berkeley, CA, January 1993.

tential for the development of train-level probabilistic nsk as-30pp. 9302020469. 64 729-020.

sessment (PRA) databases. These train-level databases will This report presents a three dimensional model for intium mi-allow the Nuclear Regulatory Commission to investigate effects gration in an arid waste disposal site. When tritiated water is re-on plant core damage frequency (CDF) given a train is failed or leased at a point in a uniform and relatwely dry soil it redistrib-taken out of service. The intent of this task was to develop utes in both the liquid and vapor phases. The flux density of tnt-usor-fnendly databases that required a minimal amount of per-sum in each phase is of the same order of magnitudo however Sonnel involvement to be usable. It was onginally nntended that so tntium redistnbution is modeled as if transfer occurs "in par-the train-level models would not be expanded to include basic evonts below tho top gate of a train, with the possible exception alle!" in the liquid and vapor phases. The approach we descnbe of including some of the major train-related components (e g.,

uses the d!ffusion equation cast in radial (spherical) coordinates important pumps and motor-operated valves), it was found that and takes into account radioactive decay. It permits calculation a databaso $1milar to the ongsnal plant PRA provided the accu-of radial profiles of tntium concentration, within and external to racy needed to measure the changes in plant CDF. The Peach a sphere of released solution. We assume the concentration Bottom Unit 2 NUREG 1150 PRA la large fault tree model) and within this sphere initially to be uniform. The solution also pre-the Deaver Valley Unit 2 IPE (a large event tree model) were dicts attenuation and rate of advance of the maximum of tntium selected to demonstrate the feasibility of developing train leve!

concentration as rt advances in the soil. With doop disposal'in a databases. Five ditterent methods for developing train-level da-desert soil, the modul predicts that tritium migration will be very tabases were hypothesized and are examined. Uitsmately, two short range, with a maximum of a few meters

34 Main Citations and Abstracts NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE NUREG/CR 5984: CODE AND MODEL EXTENSIONS OF THE MACHINED NOTCHES ON THE FRACTURE TOUGHNESS OF THATCH CODE FOR MODULAR HIGH TEMPERATURE GAS.

SEVERAL STRUCTURAL ALLOYS. JOYCE,JA U.S. Naval COOLED REACTORS. KROEGER.P.G.; KENNETT,RJ. Brook-Academy, Annapolis. MD. LINK R.E. Navy, Dept. of. September haven National Laboratory. May 1993. 46pp. 9306010318. BNL-1993. 90pp. 9310130032. 76744:290.

NUREG 52356. 75062:166.

1 Recent computational studies of the stress and strain fields at This report documents several model extensions and im-the tip of very sharp notches have shown that the stress and provements of the THATCH code, a code to modet thermal and strain fields are very weakly dependent on the initial geometry fluid flow transients in High Temperature Gas-Cooled Reactors.

of the notch once the notch has been blunted to a radius that is A heat exchanger model was added. which can be used to rep.

6 to 10 times the initial root radius, it follows that if the fracture resent the steam generator of the main Heat Transport System toughness of a material is sufhciently high so that fracture initi-of the auxiliary Shutdown Cooling System. This addition permits ation does not occur in a specimen until the crack-tip opening the modeling of forced flow cooldown transients with the displacement (CTOD) reaches a value from 6 to 10 times the THATCH code. An enhanced upper head model, considering size of the initial notch tip diameter, then the fracture toughness the actual conical and spherical shape of the upper plenum and will be independent of whether a fatigue crack or a machtned reactor uppor head was added, permitting more accurate (nod-j notch served as the initial crack. In this experimental program elling of the heat transfer in this region. The revised models are l

the fracture toughness (J(Ic) and J resistance (J-R) curve, and described, and the changes and addition to the input records CTOD) for several structural alloys was measured using speci-are documentod.

mens with conventional fatigue cracks and with EDM machined notches. The results of this program have shown, in fact, that NUREG/CR-5987: MICROBIAL-INFLUENCED CEMENT DEGRA-most structural materials do not achieve inrtsation CTOD values DATION LITERATURE REVIEW.

ROGERS,P,.D.;

on the order of 6 to 10 times the radius of even the smallest HAMILTON,M.A.; MCCONNELL.J.W. EG&G Idaho, Inc. March EDM notch tip presently achievable. It is found furthermore that 1993. 34pp. 9303300171. EGG-2695. 74407:322.

tougher matenals do not seem to be less dependent on the The Nuclear Regulatory Commission stipulates that disposed type of notch tip present. Some matenals are shown to be low level radioactive waste (LLW) be stabihred. Because of ap-much more dependent on the type of initial notch tip used, but parent ease of use and normal structural integnty, cement has no simple pattern is found that relates this observed depend, been widely used as a binder to sohdify LLW. However, the re-ence to the material strength, toughness, of strain hardening sulting waste forms are sometimes susceptible to failure due to rate.

the actions of waste constituents, stress, and environment. This report reviews literature which addresses the effects of micro-NUREG/CR-5982: EFFECTIVENESS OF CONTAINMENT biologically influenced chemical attack on cement-solidified SPRAYS IN CONTAINMENT MANAGEMENT.

LLW Groups of microorganisms are identified, which are capa.

NOURBAKHbH.H.P.; PEREZ,S E.; LEHNER,J.R. Brookhaven ble of metabolically converting organic and inorganic substrates National Laboratory. May 1993. 54pp. 9306180261. BNL.

into organic and mineral acids. Such acids aggressively react NUREG 52354. 75427;109.

with concrete and can ultimately lead to structural failure. Mech-A hmited study has been performed assessing the effective-anisms inherent in microbial-influenced degradation of cement-ness of containment sprays to mitigate particular challenges based material are the focus of this report. This report provides which may occur dunng a severe accident Certain aspects of sufficient evidence of the potential for microbial-influenced dete-F three specific topics related to using sprays under severe accL noration of cemerit-sohdified LLW to justify the enumerat6on of dent conditions wore investigated The first was the effective.

the conditions necessary to support the microbiological growth ness of sprays connected to an alternate water supply and and population expansion, as well as the development of appro-pumping source because the actual containment spray pumps priate tests necessary to determine the resistance of cement-are inoperable. This situation could occur dunng a station blach-solidified LLW to microbiological-induced degradation that could out The second topic concerned the adverse as well as benefi-impact the stabihty of the waste form.

j cial effects of using containment sprays dunng severe accident NUREG/CR-5488: SOIL CHARACTERIZATION METHODS FOR scenanos where the containment atmosphere contains substan-UNSATURATED LOW-LEVEL WASTE SITES. WlERENGA.P.J.;

I 1:a! quantities of hydrogen along with steam. The third topic was YOUNG.M.H. Arizona, Univ. of. Tucson, AZ. GEE,G.W.; et al.

the feasibility of using containment sprays to moderate the con-Battelle Memorial Institute, Pacific Northwest Laboratory. Febru-sequences of DCH.

ary 1993.150pp. 9303120106. PNL-8480. 74236:120.

To Tupport a license application for the disposal of low-level NUREG/C$1-5983: SAFETY ASPECTS OF FORCED FLOW COOLDOWN TRANSIENTS IN MODULAR HIGH TEMPERA.

raduactive waste (LLW), applicants must characterize the un-TURE GAS-COOLED REACTORS. KROEGER.P.G. Brookhaven saturated zone. This requires an integrated plan to be devel-ped for sampling and analyzing the soil horizons for physical National Lpboratory. May 1993. 24pp. 9306010323. BNL-and hydrauhc properties. This document provides a strategy for NUREG-52355. 75058.297.

developing this characterization plan. It describes principle of Dunng some of the design basis accidents in Modular High contaminant flow and transport, site characterization and ni-Temperature Gas Cooled Reactors (MHTGRs), the main Heat tonng strategies, and data management, it also discusses meth-Transport System (HTS) and the Shutdown Cooling System ods and practices that are currently used to monitor properties (SCS) are assumed tc have failed Decay heat is then removed and conditions in the soil profile, how these properties influence by the passive Reactor Cavity Cooling System (RCCS) only. If water and waste migration, and why they are important to the either forced flow cooling system becomes available during license application. The methods part of the document is divid-such a transient, its restart could significantly reduce the down-ed into sections on laboratory and field-based properties, then time. This report used the THATCH code to examine whether further subdivided into the description of methods for determin-such restart, during a period of elevated core temperatures, can ing 18 physical, flow, and transport properties. Because of the be accomplished within safe limits for fuel and metal compo-availabikty of detailed procedures in many texts and journal arts-nent temperatures. If the reactor is scrammed, either system cles, the reader is often directed for details to the available liter-can apparently be restarted at any time, without exceeding any ature. References are made to experiments performed at the safe limits. However, under unscrammed conditions a restart of Las Cruces Trench site, New Mexico, that support LLW site forced coohng can lead to reenticahty, with fuel and metal tem-characterization activities A major contribution from the Las peratures significantly exceeding the safety limits.

Cruces study is the experience gained in handling data sets for

~

. - - ~

Main Citations and Abstracts 35 site r,haractenration and the subsequent use of these data sots ing analysis of failure data to estimate the fraction of dependont in modekng studies.

failures among the failures. In addition, the proposed method NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMITY can evaluate the impact of the observed dopendency on system t

DOSIMETERS-PILOT

TEST, FOX,R.A.;

HARTY,R ;

unavailability and plant nsk. The formulations denvN! in this MCDONALD,J.C. Battelle Memorial Institute, Pacific Northwest rep rt have undergone vanous levels of validations through Laboratory. July 1993. 51pp 9308160263. PNL-8467.

computer simulation studies and pilot applicatons. The pilot ap-76119:169, plications of these methodologies showed that the contnbution A working group of the Hoalth Physics Society Standards of dependent failures of diesel generators in one plant was nog-Committeo (HPSSC) has issued a dratt standard for extremity ligible, while in another plant, it was quite significant. It also

]

dosimotors. To determine the appropriateness of tne proposed showed that in the plant with significant contribution of depend-i standard, Pacific Northwest Laboratory (PNL) has conducted ency to Emergency Power System (EPS) unavailability, the con-three separata evaluations of the performance by processors of inbution changed with time. Similar findings were reported for estromity dosimeters. The dosimeters were tested in each of the Containment Fan Cooler breakers. Drawing such conclu-the arradiaton categories specihed in the draft standard: high-sions about system performance would not have boon possible energy photons (general and accident dosimetry), low-energy with any other reported dependency methodologies.

photons (general and accident dosimetry), beta particles, neu-NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMA-trons (first and riecond evaluations only), and a mixture catego-TION AND SYSTEM UNAVAILABILITY EVALUATION BASED ry, in the first evaluation only about 60% of the processors met ON FAILURE DATA STATISTICS. Detailed Desenpton And Ap-the draft standard's performance enteria for accuracy and preci-phcatons. AZARM.M.AJ HSU F ; MARTINEZ-GURIDl; et al sion. The second evaluation showed an overall improvement of Brookhaven National Laboratory. July 1993.70pp.9307270044.

15% to 18%, but most processors were still unable to meet the BNL NUREG.52362. 75802:143.

performance cntena consistently in all irradiation categories-This report introduces a new perspective on the basic con-After these evaluations, PNL suggested sov9ral changes to the copt of dependent failures where the definition of dependency draft standard, including rodofining some of the test categones is based on clustering in failure ilmes of similar components, and making the tolerance levels of the entena more Consistent This perspective has two significant imphcatons. firstly, it re-with those in the standard for whole-body dosimetry. This report luxes the conventional assumption that dependent failures must summantes the third evaluation, which yielded an overall pass be simultaneous and result from a severe shock; secondly, it rate of 67% Suggestions are grven to render the draft standard allows the analyst to use all the failures in a time continuum to generally more consistent with the entena for whole-body do-estimate the potential for multiple failures in a window of time

• • N (e g, a test interval), therefore arnving at a more accurate value NUREG/CR-599 I: PORFLOW: A MULilFLUID MULTIPHASE for system unavailability. In addition, the models developed here MODEL FOR SIMULATING FLOW, HEAT TRANSFER, AND Pf0V'de a method for plant-specific analysis of dependency, re-MASS TRANSPORT IN FRACTURED POROUS MEDIA. User's flecting the plant specific ma'ntenance practicos that reduce or Manual - Version 2.41. RUNCHAL,A.K. Analytic & Computional increase the contnbution of dependent failures to system un-Resoarch, Inc. SAGAR,0. Contor for Nuclear Waste Regulatory availabihty. The proposed methodology can be used for screen-Analyses February 1993 221pp. 9303120062. CNWRA 92-003.

ing analysis of failure data to estimate the fraction of dependent 74240.004.

failures among the failures. In addition, the proposed method The PORFLOW software package is designed to simulato can evaluate the impact of the observed dependency on system flow, heat transfer, and mass transport in throu-dimensional het.

unavailabihty and plant risk. The formulations denved in this erogenous porous and fractured modia. Phase change and gas report have undergone vanous levels of vahdations through i

phase fibw is included. Radionuchde decay chains of up to four computer simulabon studios and pilot apphcations. The pilot ap-I members can be included in transport analysos. The mathemats-plications of these methodologies showed that the contnbution cal basis of the model is desenbed in Chapters 2 and 3, the of dependent failures of diesel generators in one plant was neg-code structure is discussed in Chapters 4 and 5, detailed in, hgible, while in another plant, it was quite significant. It hiso structons for the user are in Chapter 6, and a few test prob.

showed that in the plant with significant contribution of depend-loms are in Chapter 7. The PORFLOW is a general purposo ency to Emergency Power System (EPS) unavailabihty, the con-Mitware that can be adapted to many different problems. Ana.

inbution changed with time. Similar findings were reported for lytic and Computational Research, incorporated of Los Angoles, the Containment Fan Cooler breakers. Drawing such conclu-CA owns the copynght to the software, howevor, the U S. Gov.

sions about system performance would not have been possible ernmont has retained limited nghts on its use.

with any other reportud dependency methodologies.

NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMA-NUREG/CR-5995: TECHNICAL SPECIFICATION ACTION STATE-TION AND SYSTEM UNAVAILABILITY EVALUATION BASED MENTS REQUIRING SHUTDOWN.A Risk Perspective With Ap-ON FAILURE DATA STATIST;CS. Summary Report.

plication To The RHR/SSW Systems Of A BWR. MANKAMO,T.

AZARM.M Aa HSO F,; MARTINEZ-GURIDI, et al Brookhaven Avaplan Oy (Finland). KIM,l.S.; SAMANTA P K. Brookhaven Na-j National Laboratory. July 1993 37pp 9307270039 BNL-tional Laboratory. November 1993.186pp. 9312170073. BNL-4 NUREG 523G2, 75803 032.

NUREG-52364. 77515:215 This report introduces a new perspective on the basic con-When safety systems fail during power operation, the limiting cept of dopondent failures where the definition of dependency conditions for operation (LCOs) and associatod action state-is based on clustering in failure times of similar components, monts of technical specifications typically require that the plant This perspective has two significant imphcations: firstly, it re-be shut down within the kmats of allowed outage time (AOT).

laxes the conventional assumption that dependent failures must However, when a system needed to remove decay heat, such l

be simultaneous and result from a severe shock; secondly, it as the residual heat removal (RHR) system, is inoperable or de-

)

allows the analyst to use all the failures in a time continuum to graded, shutting down the plant may not nocessarily be prefora-estimate the potential for multiple failures in a window of time ble, from a nsk perspective, to continuing power operation over -

(e g, a test interval). therefore arriving at a more accurate value a usual repair time, giving pnonty to the repairs. The r%k impact for system unavailabihty In addition, the models developed hero of the basic operational afternatives, Lo4 continued operation or provide a method for plant-specific analysis of dependency, re-shutdown, was ovaluated for failures in the RHR and standby flecting the plant specific maintenance practices that reduce or service water (SSW) systems of a boihng water reactor (BWR) increase the contribution of dependent failures to system un-nuclear power plant. A complete or partial failure of the SSW availabihty. The proposed methodology can be ut,ed for screen-system fails or degrades not only the RHR sys'am but other

4 a

36 Main Citations and Abstracts front-line safety systems supported by the SSW system. This ed as part of a " blind" modeling exercise to demonstrate the report presents: (a) the methodology to evaluato the risk tmpac'.

ability or inability of uncalibrated models to predict unsaturated j

of LCOs and associated AOT: (b) the results of nsk evaluation flow and solute transport in spat: ally vanable porous media.

from its application to the RHR and SSW systems of a BWR; (c)

Simulations were conducted using a recently developed multi-the findings from the nsk sensitivity analyses to identify alterna-phase flow and transport simulator. Un! form and he%rogeneous tive operational policies; and (d) the major insights and recom-sod models were tested, and data from a previo expenment mendations to emprove the technical specifications action state.

at the site were used with an inverse procedure to estimate ments.

water retention parameters. A spatial moment analysis was used to provide a quantitative basis for comparing the mean ob-i NUREG/CR-5996: SUBSURFACE INJECTION OF RADIOACTIVE served and simulated flow and transport behavior. The results TRACERS Field Expenment For Model Validation Testing.

sl FAYER,M.J.; SISSON.J B.; JORDAN,W.A., et al. Battelle Memo-of this study suggest that defensible predictions of waste migra-ton aM fate at toe levd waste sites win Anat@ reque so j

I nel Institute, Pacific Northwest Laboratory. February 1993.50pp.

specific data for model calibration.

9303120155. PNL-8499 74234:302.

Accurate predictions of the movement of radinactrve contami-NUREG/CR-5999: INTERIM FATICUE DESIGN CURVES FOR l

nants from disposal facilities are required to evaluate effects-CARBON, LOW-ALLOY, AND AUSTENITIC STAINLESS optimize data collectio1, esign remediation strategies, and pre-STEELS IN ' LWR ENVIRONMENTS. MAJUMDAR S.:

dict the longterm reNits of such strategies. A held expenment CHOPRA,0.K.; SHACK,W.J. Argonne National Laboratory April was undertaken i,- 1980 and 1981 to provide data to test the 1993. 34pp. 9305100010. ANL-93/3. 74858.189.

hmits of model ps adictions The purpose of this report is to pro-Existing data in the hterature on fatigue of carbon, low alloy, vsde a complete record of data generated dunng that field ex-and austenstic stainless steels in LWR environments are re-periment fnr use as a model vahdation test case. The report viewed. It is found that both temperature and dissolved-oxygen I

combines the information in Sisson and Lu (1984) with unpub-concentration in water significantly affect fatigue life. At the very lished laboratory and field data on the hydraulic properties of low dissolved-oxygen levels characteristic of pressurized water the sediments and core data collected at the and of the experi-reactors and boiling water reactors with hydrogen-water chemis-ment. The unique features of this expenment were the docu-try, environmental effects on fatigue life are modest. However, mented control of the inputs, the three-dimensional nature of at higher dissolved-oxygen levels (2100 ppb), significant reduc-the expenment, the measurement of radioactive tracers in situ.

tions in fatigue hfe can occur. The susceptibahty of carbon and low-and the use of multiple injections. The in situ monitonng meth-alloy steels to reduced fatigue life is strongly related to ods were neutron moderation for water content and gamma sulfur concentration. Although the fatigue lives of austenstic onergy analysis for tracer concentration. The data are providud stainless steels may be reduced, the reductions are much on 3.5-in, diskettes. The data include observation and injection smaller than those observed in high-sulfur carbon and low-alloy well construction details, in}ection solution concenLations, radio-steels in oxygenated water, fatigue life depends strongly on active tracer and water content distnbutions in space and time.

strain rate. Interim fatigue design curves are proposed that take neutron probe calibration information and sediment properties into account temperature, dissolved. oxygen level in the water, determined in both the laboratory and field-the sulfur level in the steel, and strain rate. Design curves for NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES carbon and low-alloy steels for lives up to 10(8) cycles are also OF LARGE SCALE INTERNATIONAL REFERENCE EXPERi-proposed MENTS (PROJECT FALSIRE) B ASS.B.R.;

PUGH C.E.;

NUREG/CR-6007: STRESS ANALYSIS OF CLOSURE BOLTS KEENEY WAL KER,J; et al. Oak Ridge National Laboratory.

June 1993.150pp. 9?O7020010. ORNL/TM-12307. 75585:204 FOR SHIPPING CASKS. MOK,G.C.; FISCHER.LE. Lawrence This report summt.zes the recently completed Phase I of the Livermore National Laboratory. HSU,S.T. Kaiser Engineering Project for Fracture Analysis of Large-Scale International Refer-(formerly Kaiser Engineers). January 1993,139pp.9302010084 UCRL-ID-110637. 64726:111, ence Expenments (Proloct FALSIREL Project FALSIRE was cre, ated by the Fracture Assessment Group (FAG) of Principal This report specifies the requirements and enteria for stress i

Working Group No. 3 (PWG/3) of the Organization for Econom.

analysis of closure bolts for shipping casks containing nuclear ic Cooperation and Development (OECD)/Nucioar Energy Agen.

spent fuels or high level radioactive materials. The specification cy's (NEA's) Committee on the Safety of Nuclear installations is based on existing information concerning the structural be-(CSNI). Motivation for the project was derived from recognition havior, analysis, and design of bolted joints. The approach taken was to extend the ASME Boiler and Pressure Vessel by the CSNI-PWG/3 that inconsistencies were being revealed in predictive capabilities of a vanety of fracture assessment meth-Code requirements and enteria for bolting analysis of nucl6ar ods, especially in ductrio fracture applicat'ons As a conse-p' Ping and pressure vessels to include the appropriate design i

quence, the CSNI/ FAG was formed to evaluate fracture pred.c.

and toad characteristics of the shipping cask. The charactens-tion capabilities currently used in safety assessments of nuclear tics considered are large, flat, closure hds with metal-to-metal components. Members are from laboratones and research orga.

contact within the bolted joint, significant temperature and nizations in Western Europe, Japan, and the United States of impact loads; and possible prying and bending effects. Specific Amenca (USA) On behalf of the CSNI/ FAG, the U S. Nuclear formulas and procedures developed apply to the bolt stress Regulatory Commission's (NRC's) Heavy-Section Stoel Technol.

analysis of a circular, flat, bolted closure. The report also in-ogy (HSST) Program at the Oak Ridge National Laboratory cludes critical load cases and desirable design practices for the (ORNL) and the Gesellschaft fur Anlagen--und Reaktorsicher.

bolted closure, an in-depth review of the structural behavior of heit (GRS), Koln, Federal Republic of Germany (FRG) had re, botted joints, and a comprehensive bibliography of current infor-sponsibility for organization arrangements related to Project mation on bolted joints.

FALSIRE. The group is chaired by H. Schulz from GRS, Koln, NUREG/CR-6011: REVIEW OF STRUCTURE DAMPfNG VALUES FRG' FOR ELASTIC SEISMIC ANALYSIS OF NUCLEAR PCWER NUREG/CR-5998: SIMULATION OF UNSATURATED FLOW AND PLANTS. HASHIMOTO,P.S.; STEELE.L.K.; JOHNSON,J.J.; et at NONREACTIVE SOLUTE TRANSPORT IN A HETEROGENE.

EOE Engineenng Consultants (formorty EOE Engineering, Inc.).

OUS SOIL AT THE FIELD SCALE. ROCKHOLD.M L. Battelle March 1993. 715pp 9303300197. 74405:001, Memonal Institute, Pacific Northwest Laboratory. February 1993.

Current U.S Nuclear Regulatory Commission guidance on 65pp 9303120158 PNL-8496. 7424t139.

structure damping values for elastic seismic design analysis of A field scale unsaturated flow and solute transport expen-nuclear power plants are contaired in Regulatory Guide 161 ment at the Las Cruces trench site in New Monico was simulat-(R G.1.61). The objectsvos of the study described in this report 1

Main Citations and Abstracts 37 are to investigate the adequacy of R G.1.61 structure damping Reports (LERs) that were generated between 1980 and 1992.

values based on currently available data, and to recorrimond re-These LERs have been categorized into 21 failure modes that visions to R.G.1.61 as appropriate. Measured structure damp-have been prioritized based on probabilistic risk assessment ing values, and associated structure, foundation, excitation, and considerations, in addition, the results of the Hatch operating input / response parameters, were collected and compiled.

experience review have been compared with the results of a These data were analyzed to identify the parameters that signife similar, industry wide operating experience review. This compari-cantly influence structure damping and to quantify structure son provides an indication of areas in the Hatch HPCI system damping in terms of these parameters. Based on this study, cur.

that should be given increased attention in the priontrzation of rent R.G.1.61 damping values for structure design are erther inspection resources.

adequate, or require only minor revision, depending on the structure material. More explicit guidance on structure damping NUREG/CR-6015: STRUCTURAL AGING PROGRAM TECHNI.

values for seismic ana!ysis to determine input to equipment has CAL PROGRESS FOR PERIOD JANUARY DECEMBER 1992.

been prepared, along with other recommendations to improve NAUS.D.J.; OLAND.C.B. Oak Ridge National Laboratory. July the applicability of R G.1.61.

1993.164pp.9309030224. ORNL/T M 12342, 76328:021.

NUREG/CR4012: STIFFNESS AND DAMPING PROPERTIES OF The Structural Aging (SAG) Program is conducted for the Nu-A LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON clear Regulatory Commission (NRC) by the Oak Ridge National RECORDED EARTHOUAKE RESPONSES. HASHIMOTO.P S.;

Laboratory (ORNL). The program has the overall objectrve of TlONG L.W.: STEELE,L K.; et al. EOE Engineen,ng Consultants preparing an expandable handbook or report which will provide potential structural safety issues and acceptance enteria for use (formerfy EOE Engineering, Inc.). March 1993. 250pp.

9304020300. 74449.001.

by the NRC in nuclear power plant evaluations of continued An investigation into the structural properties and seismic re-service. Initial focus of the program is on concrete and pon.

sponses, of a low aspect ratio shear wall building, which has crete-related materials which comprise safety-related (Catebory construction similanty to typical nuclear plant structures, has I) structures in light-water reactor facilities. The SAG Program is been performed using actual recorded earthquake motions. This organized into four tasks: Task S.1-Program Management, Task effort used a combination of mooalidentification to obtain struc-S.2-Materials Property Data Base, Task S.3-Structural Compo-ture modal parameters directfy from the recorded motions, and nent Assessment / Repair Technology, and Task S.4-Ouantitative elastic structural analysis using methods and entena frequently Methodology for Continued Service Determinations. In meeting the individual objectives of these tasks resources are drawn employed by the nuclear industry. Modal parameters determined from ORNL with subcontract support from universities and other by modal identification provide excellent fits to the building mo-tions recorded during the 1984 Morgan Hill earthquake. Modal research laboratories. This report providos an overv'ew of prin-parameters identified for the 1989 Loma Prieta earthquake are espal developmenia in each of the four program tasks from Jan-uary 1,1992 to December 31, 1992. Planned activities under more uncertain. Investigation of building stJtnesses generath confirms the adequacy of bounding estimates currently recom-each of these tasks aro also presented.

mended for nuclear plant structure seismic analysis. Damping NUREG/CR4018: SURVEY AND ASSESSMENT OF CONVEN-values identified for this building supplemont the database being TIONAL SOFTWARE VERIFICATION AND VAllDATION METH-compiled to investigate current nuclear plant structure damping ODS. MILLER,L.A.; GROUNDWATER,E.; MIRSKY,S M.: et al.

criteria.

Science Applications Intemational Corp. (formerly Science Ap-NUREG/CR4013: METHODS USED FOR THE TREATMENT OF plications, Inc.). April 1993.186pp. 9305100042. EPRI TR-NON-PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

102106. 74858:001.

CONOSCENTE,J P.: MASLENIKOV,0.R.; JOHNSON.J.J. EOE This report documents the results of the first (of ton) tasks Engineering Consultants (formerly EOE Engineering, Inc.). May being performed under a contract jointly funded by the USNRC 1993. 62pp. 9306210371. 75402:286.

and EPRI to develop and document guidelines for the verifica-Non-proportional or non-classical damping is defined as a tion and validation of expert systems in the nuclear industry.

form of viscous damping that introduces coupling between the This task conducted an extensive survey of conventional soft-undamped modal tAordinates of motion. Such problems have ware venfication and validation (V&V) methods. A total of 134 practical applications in the dynamsc analysis of soiLstructure different methods were identified which can be applied to either systems, structure-equipment systems, and structural systems the requirements design of implementation phases of software.

made of materials with different energy dissipation capacities.

These methods were classified by a sequential hfecycle model, Presented in this report is a review of the methods most com_

characterized by factors of power and ease-of-use, and as.

monly used in structural analysis for the solution of the dynamic sessed according to their applicability to expert systems. Expert response of systems with non-proportional damping Both rigor-systems were decomposed into four components: knowledge ous and approximate methods are described. Since ngorous base, inference engine, interfaces, and tools / utilities. The con-methods usually require large computational efforts, approxi-ventional software V&V methods were found to be directly or, mate methods using undamped mode shapes are often pre.

by extension, applicable to all of the expert system techniques fened. In the study described here, the accuracy of three ap.

except the knowledge base, proximate rnethods was evaluated for three benchmark prob-NUREG/CR4021: A LITERATURE REVIEW OF COUPLED tems, with various parametnc vanations. Results were compared with the exact solution for different combinations of structural THERMAL-HYDROLOGIC-MECHAN ICAL -CHEMICAL PROC-properties. Based on these results, conclusions and recommen-ESSES PERTINENT TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA MOUNTAIN.

dations are presented for the use of the selected approximate methods' MANTEUFEL,R.D.; AHOLA,M P.; TURNER.D R.; et al. Center for Nuclear Waste Regulatory Anatyses. July 1993. 240pp.

NUREG/CR4014: HIGH PRESSURE COOLANT INJECTION 9308160184. CNWR A92-011. 76117:040.

SYSTEM RISK-DASED INSPECTION GUIDE FOR HATCH NU.

A literature review has been conducted to determine the state CLEAR POWER STATION. DIBIASIO,A.M. Brookhaven National of knowledge available in the modeling of coupled thermal (T),

Laboratory. May 1993. 57pp. 9306110025. BNL-NUREG 52367.

hydrologic (H), mechanical (M), and chemical (C) processes rel-75336:303.

event to the design and/or performance of the proposed high-A review of the operating expenence for the High Pressure lovel waste (HLW) repository at Yucca Mountain, Nevada. The j

Coolant injection (HPCI) system at the Hatch Nuclear Power review focuses on identifying coupling mechanisms between in-Station, Units 1 and 2, is desented in this report. The informa-dividual processes and assessing their importance (ie., if the tion for this review was obtained from Hatch Licensoe Event coupling is either important, potentially important, or negligible).

}

~

38 Main Citations and Abstracts The significance of considonng THMC.couplod procones hos in nos, Sandia National laboratones and ANATECH, Inc. on the whether or not ma processes impact the design and/or por.

vanous portions of the phnnomenology involved Those indo-formanco otyoctavos of the repostory. A review, such as report-pondent evaluations are includnd here as Parta 11 through Vi ed here, is unoful m identifying which coupled offects will be im-The resulta, and their intogtation in Part t, demonstrate the nub-portant, hence which coupled offects will nood to be envcatigat-stantial synorgism and convergence necessary to recognito that ud by the U S. Nucluar flogulatory Commission in ordur to the hasua has buon resolved.

assess the assumptions, data, analyus, and conclusions in the design arn! performance assessment of a geologic repository.

NUREG/CR4026: TitEOHETICAL AND EXPEf11 MENTAL INVES-Although this work stems frorn fogulatory interest in the design TIGATION OF THERMOHYDHOLOGIC PROCESSES IN A PARTIALLY SATURAIED, FRACTURED POROUS MED%M.

of tho goologec repository, it should be emphasized that the ro.

pository design Imphettiy cons &rs all of the repository porform.

GREEN.H T.; MANTEUF EL.R D, Centor for Nuclear Wasto Hog-i

~

ance objectivos, inciuding thoso associated with the limo ano, utatory Analyson. DODGE,ET.; et al. Southwest Rosaarch Insti-

.j y

permanent closure lho scopo of this tuview sa considorod tuto. July 1993. 225pp. 9300000010. CNWRA 02 006.

beyorni previous assessments in that it attumpts (with the cut.

76300 064 runt state-of knowledge) to dolormine which couphngs are im.

The performance of a geologic repository for high-lovut nucle-j portant, and identify which computer codos are currently avail.

at waste will bo influenced to a large degroo by thermohp'Ao-

)

ablo to modul coupled processos, gic phenomena created by the emplacement of heat generating NUREG/CH4022: HIGH PRESSURE COOL ANT INJECTION (HPCI) SYSTEM RISK OASED INSPECTION GUIDE FOft "U

DROWNS FERHY NUCLEAR POWER ST ATION WONG.S *-

Thus, these phenomena must be well understood prior to a de-DtBIAsiO,A.M.; GUNIHF R,W. Diookhavon National laboratory.

finittvo assessment of a potential repository sito, An investiga-Septomtwr 1993 60pp 9311010051 BNL-NUHEG-523 70.

tion has been undertaken along three separate avtinuets of UM analysis: (1) laboratory exp9nments; (2) mathematical models; The tbgh Prosauro Coolant injuction (HPCI) system has boon and (3) bimilitudo analysis. A summary of accomphahments to examined from a nsk porspechvo A Systum Ibsk-Based lospoc-dato is as follows: (1) A reviow of the htorature on the theory of lion Guido (S RIG) has buon developod as an aid to HPCI heat and mass transfor in partially saturated porous medium; (2) system trwpoctions at the Browns Forry Nucioar Power Plant, A dovolopment of the governing conservation and constitutive Units I,2 and 3 T he role of the HPCI system in mitigating acci oquations; (3) A development of a dimensionless form of the dents is discussed in this S HIG, along with insghts on identi~

tioverning equation; (4) A numerical study of the importance and food nak bawd fatture modos which could provent proper opur-sonsitivity of flow to a bot of dimensionions Droups; (5) A survey ation of tho system. The S-RIG providos a rev6ew of industry' and evaluation of experimental measuromont techniques; (6) wide operating empononca, includag plant-t pocific illushativo Exocution of faboratory exportments of nonssothormal flow tri a examples to augment the PRA and operational considorations porous medium with a simulated fracture in idnntifying a catakxjuo of basic PHA failure modos for the HPCI system. It is designed to be uwd as a roturence for rou-NUREG/CH-6027: PRELIMINARY EVALUATION OF SNUDDER tenu inspections, self 4nitiated safety system functional inspec-SINGLE FAILURES.

WARE,A Ga DI ANDFORD.H K.;

160ns (SSFish and the ovaluaton of nsk signihcance of compo-KELLY,0 L; ot al EGAG Idaho, loc. Apnl 1993 49pp.

nont failuros at the nuclear power plant-9305100059. EGG 2007,74857;316.

The United States Nuclear Regulatory Commension developed NUREG/CH-6023: GENEntC ANALYSIS FOR EVALUAllON Of-HN L2430, Prohminary Evaluation of Snubber Single failuros, LOW CHARPY UPPER SHELT ENERGY EFFECTS ON SAFETY MARGINS AGAINST FRACTURE OF HEAC10R wnh the objecM of perforrning a prohmanary evaluation of the PRESSURE VESSFL MATERIALS DICKSON,T L Oak Ridge safety imphcation of a potential single failure of a snubber usod National Laboratory July 1993 8?pp. 9308160110. ORNt/ I M.

to support safety-related piping or equipment. The Idaho Nation-al En0inocring Laboratory staff conducted a quahtative review of 12340. 70f 20 215.

Appendix G to 10 CF H Part 50 requires that rtiactor prusnure a large number of hght water reactor systems, and a quanista-veuel bolthno matonals rnaintain Cha py uppor. shelf energios tivo stress analysa of four systems. A candidato hst was devol-of no less than 50 ft Ib during the plant operating hfe, unions it opod that ranked the systems as having a high, medium, or low is demonstrated in a manner approved by the Nuclear Hegula-P'obabihty of causing a signihcant increase in the core damage tory Commismon (NRC), that lower values of Charpy upper 6holl imquency should a single snubber fall to function. Two systems energy provide margins of safety against fracture oquivalent to woro ranked high, a PWR ice condenser main steam contain-those in Appendm G to Section XI of the ASME Codo. Analysos nient penetration and a BWR Mark i torus, Six systems wore bawd on acceptanco entoria and analysis methods adopted in ranked medium and the romaining 30 were rankad as low or thn ASME Cndo Case N 512 are descobod boroin. Additionalin.

Iowlamedium. The two systems ranked high and two systems formation on matenal proporties was providud by the NHC, ranked medium a PWH ice condonsor auxihary foodwater kno Office of Nuclear Augulatory Howatch Materials Enginoonng and a PWR reactor coolant system loop drain kne, were chosen Dranch Thow conos. npecified by the NHC, represent Dunene for a quantitativo stress analysis Of the four systems analy7ed.

i l

apphcations to bothng water reactor and piussurited water reac.

only the PWR ice condanser main steam containment penetra-l tor vussels. This report is designalud as HSST Report No.140 tion is judged to bo significantly susceptible to the failure of a singlo snubbor, NUREG/CR-6025: THE PROOAbillTY ()F MARK-l CONT AIN-MENT FAILURE OY MELT ATTACK OF THE LINLH.

NUHEG/CR-6028: BIGFLOW. A NUMERICAL CODE FOR SIMU-THEOf ANOUS,T.G ; YAN H Cahfornia. Univ of, Sarita Barbara, LATING FLOW IN VARIABLY SATURATED, HETEROGENE-CA. PODOWSKI,M.2a et al. Rensselaer Polytechnic Instituto.

OUS GEOLOGIC MEDIA. Theory And User's Manual - Vorsion Troy, NY, November 1993. 350pp. 9312160292. 77508.150 1.1 ABABOU,H. Franco. BAGTIOGLOU.A C. Center for Nuclo-

'this report is a followup to the work prownted in NUHEG/

ar Waste Regulatory Analyses. Juno 1993.160pp 9307060141, CH.5423 addressing oarly failure of a BWH Mark I containrnant CNWHA 02 026. 75572:200.

by molt attn Ao kner, and it constitutos a part of the imple-This report documents DIGFLOW L1, a numorical code for montahos ;

1 4 %A-Oriented Accidont Analysis Mothodology simulating flow in variably saturated heterogeneous geologic

' therein In particular, it owpands the quantF media It contains the undorlying inathematical and numoncal (HOAAMI(

fication to inclus

)ur independant ovaluations carrlud out at models, test problems, benchmarks, and apphcations of the j

Renssolaer Polytechnic Instituto Ar00nne National Laborato-BIGFLOW codo. The DIGFLOW software package is composod 1

f v---,-

n

,. na

,,..a

. - - ~

-r-,

n --,

Main Citations and Abstracts 39 of a simulation and an interactive data processing code (DATA-for four types of concreto (limestone, hmestone sand, basalt.

FLOW). The simulation code solves linear and nonlinear porous and sihceous) and for their mixtures with urania and zirconia.

media flow equations based on Darcy's law, appropriately gon-The measured solidus temperatures for the urania zirconia con-erahzed to account fot 30. deterministic, or random heterogene-crete mixtures were significantly lower (hundreds of degrees)

I ity A modified Picard Scheme is used for linearizing unsaturated than those employed in the CORCON. Mod 2 thermal hydraulic flow equations, and preconditioned iterative methods are used for solving the resulting matrix systems. The data processor code, and the measured liquidus temperatures were significantly (DATAFLOW) allows interactive data entry, manipulation, and higher (also hundreds of degrees). The liquidus temperatures for urania zirconia-concrete mixtures contajning hmestone or analysss of 3D datasets. The report contains analyses of com-7 I

putational performance carr ed out using Cray 2 and Cray Y/

Imostone-sand concrete wero generally above 2850 K, which MP8 supercomputers. f3enchmark tests include compansons was the upper temperature limit of our experiments. The revised with other independently oeveloped codes, such as PORFLOW sow and liquidus temperatures are to be incorporated in the and CMVSFS, and with analytical or semi-analytical solutions.

CORCON-Mod 3 thermal hydraulic code which is an integral art of the U.S. Nuclear Regulatory Commission's MELCOR NUREG/CR-6029 V01: AGING ASSESSMENT OF NUCLEAR Code. DTA was also employed to redetermine the calcia-urania AtR TREATMENT SYSTEM HEPA FILTERS AND (CaO-UO(2)) phase diagrarn which is required in computer pro-ADSORBERS Phase 1. WINEGARDNER,W. Baltelle Memorial grams that calculate the phase diagrams (and solidus and liqui-Institute, Pacific Northwest Laboratory. August 1993. 48pp-dus temperatures) of urania-zirconia. concrete systems from the 9309210238 PNL-8594. 76486.075-phase diagrams of simpler systems.

A Phase I aging assessment of high-efficiency particulate air (HEPA) filters and activated carbon gas adsorption units (ad~

NUREG/CR 6034: OKLAHOMA SEISMIC NETWORK. Final sorbers) was performed by the Pacific Northwest Laboratory Report. LUZA,K.VJ LAWSON.J.E. Oklahoma, Univ. of. Norman, (PNL) as part of the U.S. Nuclear Regulatory Commission's OK. July 1993. 42pp. 9308160178. 76122:314.

(NRC) Nuclear Plant Aging Research (NPAR) Program. Informa-The Nemaha uplift is composed of a number of crustal blocks tion concerning design features; failure expenence; aging mech-anisms, effects, and stressors; and surveillance and monitonng typically 3 to 5 miles (5 to 8 km) wide and 5 to 20 miles (8 to 32 methods for these key air-treatmont system components was km) long. In Oklahoma, several discontinuous upidts, such as compiled Over 1100 failures, or 12 porcent of the filter installa' the Oklahoma City, Lovell, Garber, and Crescent uphfts, occur tions, were reported as part of a Department of Energy (DOE) along the main axis of the Nemaha uplift. A statewide network survey. Investigators from other national laboratories have sug-of 12 stations records seismol09ical data in Oklahoma. Six se-gested that aging effects could have contributed to over 80 per mipermanent seismograph stations, four radiotelernetry stations, cent of these failures. Tensito strength tests on aged fittor the Oklahoma Geophysical Observatory's seismograph station, media specimens indicated a decrease in strength. Filter aging and a borehole seismograph station at the Observatory com-mechanisms range from those associated with particle loading pnse the Oklahoma Geological Survey's seismic network. From to reactions that alter properties of sealants and gaskets. Low January 1,1987, through December 31,1992,373 earthquakes radioiodine decontamination factors associated with the Three were located by the Oklahoma seismic network. The distribution Mile Island (TMI) accident were attributed to the premature of the earthquakes by state is as follows: 315 in Oklahoma,28 aging of the carbon in the adsorbers. Mechanisms that can lead in Texas,14 in Kansas, 8 in Arkansas, 7 in Missouri, and 1 in to impaired adsorber performance include oxidation as well as Nebraska. Of the 352 earthquakes, 23 were reported felt. The the loss of potentially available active sites as a result of the earthquake epicentral data produces at least three seismic adsorption of pollutants. Stressors include heat, moisture, radi-trends. These trends are located in north-central Oklahoma, at ation, and airborne particles and contaminants.

the eastern marDin of the Anadarko basin, and in the Arkoma NUREG/CR-603 t CAVITATION GUIDE FOR CONTROL basen-Ouachita Mountains area.

993 1 9pp 93 6'010310 062 NUREG/CR-6035: FEASIBILITY STUDY FOR IMPROVED This guide teaches the basic fundamentals of cavitation to STEADY-STATE IN!TIAll2ATION ALGORITHMS FOR THE provide the reader with an understanding of what causes cavita-RELAPS COMPUTER CODE. PAULSEN.M.P.; PETERSON,C E.;

tion whe,i et occurs and the potential problems cavitation can KATSMA,K.R. Computer Simulation & Analysis, Inc. April 1993.

cause to a valve and piping system. The document provides 75pp. 9306010277. 75058:225.

guidelines for understanding how to reduce the cavitation and/

A design for a new steady-state initialization method is pre-or select control valves for a cavitating system The guide pro-sented that represents an improvement over the current method vides a method for predicting the intensity of control valves and used in RELAPS. Current initialization methods for RELAPS how the effect of cavitation on a system will vary with valve solve the transient fluid flow balance equations simulating a type, valve sire, valve function, operating pressure, duration of transiont to achieve steady-state conditions. Because the tran-operation and details of the piping installation. The guide de-sient solution is used, the initial conditions may change from the fines six cavitation hmsts identifying cavitation intensities ranging desired values requinng the use of controllers and long trark from inception to the maximum intensity possible. The intensity sient running times to obtain steady-state conditions for system of the cavitation at each limit is described including a bnef dis-problems. The new initialization method allows the user to fix cust> ion of how each level of cavitation influences the valve and thermal-hydraulic values in volumes and junctions where the system. Examples are included to demonstrate how to apply the conditions are best known and have the code compute the ini-method, including making both size and pressure scale effects taal conditions in other areas of the system. The steady state corrections. Methods of controlling cavitation are discussed pro-balance equations and solution methods are presented. The viding information on various techniques which can be used to constitutsve, component, and special purpose models are re-design a new system or modify an existing one so it can oper-viewed with respect to modifications required for the new ate at a desired level of cavitation.

steady state initialization method. The requirements for user NUREG/CR-6032: SOLIDUS AND LIQUIDUS TEMPERATURES input are defined and the feasibility of the method is demon-OF CORE-CONCRETE MIXTURES.

ROCHE,M F.;

strated with a testbed code by initializing some simple channel LElBOWITZ,L.; FINK,J K.: et al. Argonne National Laboratory.

problems. The initialization of the sample problems using the June 1993 55pp. 9306290008, ANL-93/9. 75494:001.

old and the new methods are compared.

Sohdus and liquidus temperatures were measured by a Com-bination of ditterential thermal analysis and rotational i scometry

=

40 Main Citations and Abstracts NUREQ/CH4036: INITIAL RESULTS OF THE INFLUENCE OF Aging is accelerated by moisture, non-condensable gases (e g-,

BIAXIAL LOADING ON FRACTURE TOUGHNESS. THEISS,T.J.;

air), dirt, and other contamination within the refrigerant contain-BASS,0.R.: BRYSON,J W.; et al. Oak Ridge National Laborato-ment system, excessive start /stop cycling, and operating below ty.

June 1993. 94pp. 9306290012. ORNL/TM-12349.

the rated capacity. Aging is also accelerated by corrosion and 75494.056.

fouling of the condenser and evaporator tubes. The principal A testing program to examine the influence of biaxial loads on cause of chiller failures is lack of adequate monitoring. Lack of the fracture toughness of shallow-flaw specimens under condi-performing schedulod maintenance and human errors also cord tions prototypic of a reactor pressure vessel was begun. Exist-tnbute to failures.

ing data suggest that shallow-flaw specimens under biaxialload-NUREG/CR-6047: CONTINUOUS SPECTROSCOPIC ANALYSIS ing will exhibit a toughness reduction compared to comparable uniarlal specimens. Quantification of this toughness reduction is OF VANADOUS AND VANADIC IONS. BISHOP,J.V ;

the main goal of the biaxial fracture toughness program. A cru-DUTCHER.R.A.; FISHER.M.S.: et al. Omni Tech International, ciform specimen with a two-dimensional shallow through-thick-Ltd. October 1993. 28pp. 9311080119. 77072.304.

ness flaw under a biaxial load ratio of 0.6:1 was used for biaxial Spectroscopic methods were investigated for the determina-fracture toughness testing. The entical fracture load for each tion of vanadium ions in aqueous solutions ansing in the pro-specimen was approximately the same, but the uniarial speci-duction of vanad:um (11) formate and its use in the LOMI (Low men withstood substantially more deformation at failure than did Oxidation-state Metal lon) process for the chemical decontami-the biaxial specimens. Three-dimensional, elastic-plastic, finito-nation of systems in nuclear power plants. In the LOMI process, element posttest analyses were necessary to estimate fracture a dilute solution of vanadous formate and picotinic acid is used.

toughness. In all cases, agreement between the measured and The vanadous formate reduces metal oxides in the scale on the computed load vs deformation responses was excellent. Tough-equipment, causing the scale to break up and become suspend-ness values for the cruciform specimens were compared with ed. The picolinic acid chelates these materials and makes them data from previously tested, deep-and shallow-crack speci-soluble. During the decontamination the progress is followed by mens. Results from these tests indicate that the shallow-crack analyses of the metal ions and of the radioactivity. When the toughness increase is partally, but not totally, removed by the values stop increasing. the decontamination is terminated. At apphcation of biaxial loading However, additional data are re-present, it cannot be determined if the values are no longer quired to solidity these conclusions. A proposed test matrix for changing due to all the scale being removed or due to the van-additional uniaxial and biaxial testing is described. This report adous ion being spent. Infrared and ultraviolet-visible analysis has been designated HSST Report No.138.

were investigated as the means of analyzing for vanadium spe-cies. It was found that the complex formed by V(ll) with picolinic NUREG/CR-6041: DISPOSAL UNIT SOURCE TERM (DUST) acid could be used for colorimetnc analysis for V(ll) in the range DATA INPtJT GUIDE. SULLIVAN,T.M. Brookhaven a.bnal of 0 - 0.011 moles / liter, which encompasses the concentration Laboratory. May 1993.92pp.9306180317. BNL-NUREu42375.

range used in the LOMI process. The findings will be used to 75388:001.

develop an on-line instrument for continuously monstoring V(ll)

Performance assessment of a low. level waste (LLW) disposal during decontamination.

facility begins with an estimation of the rate at which radionu-clides migrate out of the facility (i.e., the source term). The NUREG/CR 6048: PRESSURIZED-WATER REACTOR INTER.

focus of this work is to develop a methodology for calculating NALS AGING DEGRADATION STUDY. Phase 1. LUK,K.H. Oak the source term. In general, the source term is influenced by Ridge National Laboratory. September 1993.

66pp.

the radionuclide inventory, the wasteforms and containers used 9310120328. ORNL/TM.12371. 76740:266.

to dispose of the inventory, and the physical processes that g p g

g Mad to release from the facility (fluid flow, container degrada-effects of aging degradations on pressunzed-water reactor on, wasteform leaching, and radionuclide transport). The com-(PWR) intemals. Primary stressors for internals are generated puter code DUST (Disposal Unit Source Term) has tpen dever-by the pnmary coolant flow in the reactor vessel, and they in-opod to model these processes. This document presents the d

models used to calculate release frorn a disposal facihty, venfi-ressure pulsations. Other stressors are apphed 'oads, manu-cation of the model, and instructions on the use of the DUST facturing processes, impurities in the coolant and exposures to code. In addition to DUST, a preprocessor, DUSTIN, which helps the code user create input decks for DUST and a post-formation indicates that fatigue, stress corrosion cracking (SCC) processor, GRAFXT, which takes selected output files and plots and mechanical wear are the three major aging-related degra.

them o com ter terrninal havo been written. Use of thes dation mechanisms for PWR internals. Significant reported fail-ures include thermal shield flow-induced vibration problems, NUREG/CR 6043 V01: AGING ASSESSMENT OF ESSENTIAL SCC in guide tube support pins and core support structure HVAC CHILLERS USED IN NUCLEAR POWER PLANTS Phase bolts, fatigue-induced core baffle water jet impingement prob-

1. BLAHNIK.D.E.; KLEIN,R F. Battelle Memorial Institute, Pacific tems and excess wear in flux thimbles. Many of the reported Northwest Laboratory. September 1993.101pp. 9310120252.

problems have been resolved by accepted engineering prac-PNL-8614. 76741;146.

tices. Uncertainties remain in the assessment of long-term nou-The Pacific Northwest Laboratory conducted a Phase I aging tron irradiation effects and environmental factors in high-cycle assessment of chillers used in the essential safety air condition-fatigue failures. Reactor internals are examined by visual in-ing systems of nuclear power plants. Centnfugal chillers in the spections and the technique is access limited. Improved inspec-75-to 750 ton refrigeration capacity range are the predominant tson methods, especially one with an early failure detection ca-type used. The chillers used, and air-conditioning systems pability, can enhance the safety and efficiency of reactor oper-served, vary in design from plant.to-plant, it is crucial to keep ations.

chiller internals very clean and to prevent the leakage of water, NUREG/CR-6049: PIPING BENCHMARK PROBLEMS FOR THE air, and other contaminants into the refrigerant containment system. Periodic operation on a weekly or monthly basis is nec.

GENERAL ELECTRIC ADVANCED BOILING WATER REAC-essary to remove moisture and non-condensable gases that TOR. BEZLER,P.; DEGRASSi,G.; DRAVERMAN.J.; et al. Brook-gradually build up inside the chiller. This is especially desirable if haven National Laboratory. August 1993.17Bpp. 9309030213.

a chiller is required to operate only as an emergency standby BNL NUREG 52377. 76327:127.

unit. The pnmary stressors and aging mechanisms that affect To satisfy the need for verification of the computer programs chillers include vibration, excessive temperatures and pressures, and modehng techniques that will be used to perform the final thermal cychng. chemical attack, and poor quality cooling water.

peping analyses for an advanced boiling water reactor standard

r Main Citations and Abstracts 41 design, three benchmark problems were developod. The prob-NRC staff that will assist them in assessing the adequacy of the lems are representative piping systems subjected to represonta.

hcensee submittals. The CECP, designed to be used on a por, tsve dynamic loads with solutions developed using the methods sonal computer, provides estimatos for the cost of decommis-being proposed for analysis for the advanced reactor standard sioning PWR power stations to the point of license termination.

design. It will be required that the combined hcense holdats Such cost estimates include component, piping, and equiprnent demonstrate that their solutions to those problems are in agroe.

removal costs; packaging costs; decontamination costs; trans-ment with the benchmark problem set.

portation costs; bunal costs; and manpower costs. In addition to NUREG/CR-6050: RADIATION EXPOSURE MONITORING AND costs, h M abo cakulaMs Wnal dum psonh, INFORMATION TRANSMITTAL (REMIT) SYSTEM User's crew-hours. and exposure person-hours associated with decom-Manual. CALE.R.; CLARK,T ; DIXSON R.; et at Science Applica-

• • • "'"E hons international Corp. (formerly Science Applications, Inc)-

NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF June 1993 300pp. 9307220201. SAIC-93/1310-01. 75742.001.

SEVERE ACCIDENT MANAGEMENT STRATEGIES The Radiation Exposure Monitoring and information Transmit-KASTENBERG W E.; APOSTOLAKIS,G.; DHIR,V.K.; et al Calf tal (REMIT) system is designed to assist U.S Nuclear Regula-fornia, Univ. of, Los Angeles, CA. September 1993. 350pp.

tory Commission (NRC) 16consees in meeting the reporting re-9310130038. 76743 231.

quirements of the revised 10CFR20 and in agreement with the Severe accident management can be defined as the use of guidance contained in R.G. 8.7, Rev.1, " Instructions for Ro*

existing and/or alternatrvo resources, systems and actors to cording and Reporting Occupational Exposure Data." REMIT is prevent or mitigate a core-melt accident. For each accident 60-a personal computer (PC) based menu driven system that facili-quence and each combination of severe accident management tales the manipulation of data base files to record and report strategies, there may be several options available to the opora-radiation exposure information. REMIT is designed to be usor-tor, and each involves phenomenological and operational con-fnnndly and contains the fuit text of R.G. 8.7, Rev.1, on-hne as sidorations regarding uncertainty. Operational uncertaintios in-well as context sensitive help throughout the program. The user clude operator, system and instrumentation behavior dunng an can enter data directly from NRC Forms 4 or 5, REMIT allows accident. A framework based on decision troos and influence the user to v6ew tho individuars exposure in relation to rogula-diagrams has been developed which incorporates such critena tory or administrative limits and alerts the user to exposures in as feasibility, effectiveness, and adverse effects, for evaluating excess of those limits The system also provides for the calcuta-potential severe accident management strategies The frame-tion and summation of dose from intakes and the determination work is also capable of propagating both data and modof uncer-of the dose to the maximally exposed entremity for the morntor-tainty,11 is appiwd to soveral potential strategies including PWR ing year REMIT can produce NRC Forms 4 and 5 in paper and cavity flooding BWR drywell flooding, PWR depressuntation, electronic format and can import / export data from ASCll and and PWR food and blood, data base files.

NUREG/CR-6052: METHODOLOGY FOR RELIABILITY BASED NUREG/CR-6058:

VIRGINIA REGIONAL SEISMIC CONDITION ASSESSMENT, Application To Concrete Structures NETWORK. Final Roport (1986 -199?) BOLLINGE R,G. A.;

in Nuclear Plants MORl,Y.; ELLINGWOOD.B. Jotins Hopkins SIBOL,M.S ; CHAPMAN.M C.; et al. Virginia Polytechnic Instituto Univ., Baltimore, MD.

• Oak Ridge National laboratory. August

& State Univ., Blacksburg, VA. July 1993.115pp. 9308160175.

1993.164pp.9308200285. ORNLSUB93.SD684. 76159 048.

76t 22 084 Structuros in nuclear power plants may be exposed to ag, in 1986, the Virginia Regional Seismic Network was one of l

grossive environmental offects that cause their strength to de.

the few fully calibrated digital seismic networks in the United crease over an extended period of service. A major concern in States. Continued operation has resulted in the archival of sig-i evaluating the continued service of such structures is to ensure nals from 2000 + local regional and teleseismic sources. Seis.

that in their current condit6on they are ablo to withstand future motectonic studios of the central Virginia seismic zone showed q

extremo load events dunng the intended service life with a lovel the activity in the western part to be related to a large antifor.

of reliability sufficient for public safety. This report describes a mal structure while seismicity in the eastern portion is associat-methodology to facilitate quantitative assessments of current ed spatially with dike swarms. The eastern Tennessee seismic and future structural reliability and performance of structures in zone extends over a 3OOx50 km area and is the result of a nuclear power plants. This methodology takes into account the compressive stress field acting at the intersoction between two nature of past and future loads, and randomness in strength large crustal blocks. Hydroseismicity, which proposes a signifi-and in degradation resulting from environmental factors. An cant role for meteoric water in intraplate seismogenesis, found adaptive Monte Carlo simulation procedure is used to evaluato support in the observation of comrnon cyclicities between timo-dependent system reliability. The time dependent reliability streamflow and earthquake strain data. Seismic hazard studies is sensitive to the time-varying load characteristics and to the have provided the following results: 1) Damage areas in the chedce of initial strength and strength degradation models but eastern United States are three to five times larger than those not to correlation in component strengths within a system In_

observed in the west; 2) Judged solely on the basis of cata-spection/ maintenance strategios are identified that minimize the toged earthquake recurrence rates, the next major shock in the expected future costs of keeping the failure probability of a southeast region will probably occur outside the Charleston, structure at or below an established target failure probability South Carolina, area, and 3) Investigations yielded necessary during its anticipated service period.

hazard parameters (for example, maximum rnagnitudes) for sev.

oral sites in the southeast. Basic to these investigations was the NUREG/CR-6054 DRF FC: ESTIMATING PRESSURIZED WATER development and maintenance of several seismological data J

REACTOR DECOMMISSIONING COSTS A User's Manual For bases.

The PWR Cost Estimating Computer Program (CECP)

Software Deatt Report For Comment. BIERSCHBACH.M. Bat.

NUREG/CR-6059: MACCS VERSION 1.5.11.1: A MAINTENANCE telle Memonal Institute, Pacibc Northwest Laboratory October REL EASE OF THE CODE. CHANIN.D.; FOSTER,Jc, et al. Tech-1993 150pp. 0311080114. PNL-8497, 77072.163.

nadyne Engineenng Consultants, Inc. ROLLSTIN,J. GRAM, Inc.

With the issuance of the Decommissioning Rule,(July 27.

October 1993. 46pp 9311150030. SAND 92-2146. 77210:190.

1988), nuclear power plant licensees are required to submit to A new version of the MACCS code (version 1.511.1) has the il S. Nuclear Regulatory Commission (NRC) fur review, de-been developed by Sandia National Laboratories under spon-commissioning plans and cost estimates This user's manual sorship of the U.S Nuclear Regulatory Cornmission. MACCS and the accompanying Cost Estimating Computer Program was developed to support evaluations of the off-site conse-(CECP) software provide a cost calculating mettiodology to the quences from hypothetical severe accidents at commercial 1

4 e

-.e

--rw-c---

=ww w

w

--mv v

42 Main Citations and Abstracts for LOFT LP-OL6 is 1104 8 K. The best-estimate peak cladding power plants. MACCS is the only current public domain code in temperature for LOFT LP-LB-1 is 1284 0 K.

the U S which embodies all of the following modeling capabili-ties: (1) weather sampling using a year of recorded weather NUREG/CR-6062: PERFORMANCE OF PORTABLE RADIATION data (2) mitigative actions such as evacuation, sheftenng, relo-SURVEY INSTRUMENTS. EISENHOWER,E.;

WELCH.L.;

cation, decontamination, and interdiction; (3) economic costs of WlBLIN,C. Advanced Systems Technology. Inc. December mitigative actions; (4) cloudshine, groundshine, and inhalation 1993. 35pp. 9401060236. 77686 271.

pathways as well as food and water ingestion; (5) calculation of This report examines alleged and documented deficiencies in both individual and societal doses to vanous organs; and (6) the periormance and the calibration of existing portable radi-calculation of both acute (non stochastic) and latent (stochastic) health effects and nsks of health effects. All of the conse-ation survey instruments. This report also examines a limited quence measures may be generated in the form of a comple-number of reported overexposures and excessive exposures at-mentary cumulative d.stobution function (CCDF) The current tnbuted to instrumentation or calibration problems. The high fail-version implements a revised cancer model consistent with ute rates in performance testing of a limited number of instru-recent reports such as BEIR V and ICRP 60. In addition, a ments indicate further testing is needed to demonstrate which number of error corrections and portability enhancements have instruments are acceptable and for what applicatiort Further, been implemented This report descnbes only the changes the adequacy of calibration is not demonstrated at this time as made in creating the new version. Users of the codo will need many calibrations are performed by nonaccredited calibration to obtain the code's onginal documentation, NUREG/CR-4691.

laborator.es A review of the regulatory requirements and prac-tices of the NRC and Agreement States regarding the use of NUREG/CR-6060: HYDROGEN MIXING STUDIES tHMS) AS-existing performance standards such as ANSI N42-17A-1988 SESSMENT MANUAL LAM.KL; WILSON,T.L. Los Afamos Na.

and the use of accredited calibration laboratones demonstrates tional Laboratory. T R AVIS.J.R Science Applications Internation-that (1) the regulatory programs do not require compliance with al Corp. (formerly Science Applications, Inc ). June 1993.94pp existing industry standards; and (2) instruments are generally 9308160049. LA-12593-M. 76121:148.

not required to be calibrated by accredited laboratories Options This report documents some calculations performed to are recommended that might encourage the use of industry por-assess the Hydrogen Mixing Studres (HMS) code. Results are formance standards and calibration techniques.

presented first for some anaytical test problems, including lam-inar flow and mass defusiort The von Karman vortex street NUREG/CR-6065: SYSTEMS ANALYSIS OF THE CANDU 3 RE-prot lem and the Sandia FLAME Facility and Heis Dampf Reak-ACTOR. WOLFGONG,) R.; LINN.M Aa WRIGHT,A.L.; et at Oak tor (HDR) containment facility test problems are then discussed For the ana!ytical problems, the code gave results that agree Ridge National Laboratory. July 1993. 334pp. 9308160298.

ORNL/TM-12396. 76118.001.

exceptionally well with the analytical solutions Calculations for This r(port presents the results of a systems failure analysis e

the von Karman vortex street problem were performed at so-study of the CANDU 3 reactor design; the study was performed lected Reynolds numbers for several obstacle types The com-for the U S. Nuclear Regulatory Commission. As part of the puted flow patterns agree well with expenmental observations-study a review of the CANDU 3 design documentation was per-specifically the occurrerice of a vortex street (double row of vor' formed, a plant assessment methodology was developed, repro-tices) above a cntical Reynolds number Calculations for the sentative plant initiating events were identified for detailed anal-von Karman vortex street problem were p(rformed at selected Reynolds numbers for several obstacle types. The computed ysis, and a plant assessment was performed The results of the flow patterns agree well with expenmental observations-specif'~

plant assessment included classification of the CANDU 3 ovent cally the occurrence of a vortex street (double row of vortices) sequonces that were analyzed, determination of CANDU 3 sys-above a cntical Reynolds number The last assessment problem tems that are "significant to safety," and identification of key involves modeling the experiment T315. The expenment was operator actions for the analyzed events camed out in the HOR containrnent building which is a large, NUREG/CR-6070: MODELING APPROACHES FOR CONCRETE multi-compartment facility (11300 m(3) free volumq in 72 com partments) in the expenrnent, a steam-water mature was first BARRIERS USED IN LOW 4 EVEL WASTE DISPOSAL.

iniected into the containment to simulate a large-break blow.

SEITZ,R.R.; WALTON.J.C. EG&G Idaho, Inc November 1993.

down of a pressure vessel, and then superheated steam was 35pp. 9312220140. EGG-2701. 77543:208 injected that was followed by a release of helium-hydrogen light A senes of three NUREGs and several papers addressing dif-The calculated results (pressure, temperature, and gas ferent aspects of modeling performance of concrete bamers for gas concentr6tions) agree reasonably well with the expenmental low-level radioactive waste disposal have been prepared prevt-ously for the Concrete Bamers Research Project. This docu-data NUREG/CR-6061: DETERMtNATION OF THE BtAS IN LOFT mont entegrates the information from the previous documents into a general summary of mode!s and approaches that can De FUEL PEAK CLADDING TEMPERATURE DATA FROM THE used in performance assessments of concrete bamers Models DLOWDOWN PHASE OF LARGE-BRE AK LOCA EXPERh for concrete degradation, flow, and transport through cracked MENTS. BERT A,V T.; HANSON,R G ; JOHNSEN,G W et al concrete bamers are discussed. The models for flow and trans-Idaho National Engineonng Laboratory May 1993 82pp port assume that cracks have occurred and thus should only be 9306210362. EGG 2610. 75403140 used for later times in simulations after fully penetrating cracks Data from the Loss of Fluid Test (LOFT) Program help quanti-are formed. Most of the models have been implemented in a fy the margin of safety inherent in pressunted water reactors computer code, CEMENT, that was developed concurrent'y with dunng postulated loss-of-coolant accidents (LOCAs). This report this document. User documentation for CEMENT is provided anahzes how well externally-mounted fuel rod cladd;ng surface separate from this report. To avoid duplication, the reader is re-thermocouples in LOFT accurately reflected actual cladding ferred to the three previous NUREGs for detailed discussions of temperatute dunng large break LOCA expenments. The anafysis each of the mathematical models. Some additional information shows that there can be a significant difference treferred to as that was not presented in the previous documents is also in-bias) between the surface-rnounted thermocouple reading and cluded Sections discussing lessons teamed from applications the actual cladding temperature, and that the magnitude of this to actual performance assessments of low-level waste disposal bias depends on tne rate of hcat transfer between the fue! rod f acilities are provided Sensitive design parameters are empha-cladding and coolant. Further, it is shown that, in terms of peak sized to icentify critical areas of performance for concrete bar-cladding temperature recorded dunng LOFT large-break LOCA nors, and potential problems in performance assessments are emperiments, the mean bras is 1142 16 2 K (20 5 1 29 2 de-also identified and discussed.

grees F) The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1104 8 K 1he best-estimate peak cladd ng temperature for LOM LP-LR1 is 1284 0 K

l Main Citations and Abstracts 43 NUREG/CR-6071: IMPACT OF ENDF/B.VI CROSS-SECTION pipe elbow under complex, large amplitude loading. The data DATA ON H.B. ROBINSON CYCLE 9 DOS! METRY CALCULA.

were obtained by testing at room temperature a large, scale TIONS. WILLIAMS,M L.; ASGARI.M. Louisiana State Univ.,

modified model of one loop of a PWR pnmary coolant system at Baton Rouge, LA. KAM,F.B Oak Ridge National Laboratory.

the Tadotsu Engineeiing Laboratory in Japan. Fatigue crack ini.

October 1993.

3tpp.

9311080105. ORNL/TM 12406.

tiation time is reasonably predicted by applying a modified local 77069.275.

strain approach (Coffin-Mason-Goodman equation) in conjunc-Dosimeters that were removed from the H.B. Robsnson reac-tion with Miner's rule of cumulative damage. Three fracture me-i tor following Cycle 9 were analyzed and compared with calculat-chanics methodologies are applied to investigate the crack ed results in an earlier study. This work updates the calculation growth behavior observed in the hot leg of the model. These useng recently available ENDF/B-Vi data sn order to assess ad-are: the AK methodology (Paris law), AJ concepts and a re-vantagos to using the newer cross sections in reactor pressure cently developed firnit load stress-range criterion. The report in-vessel fluence calculations. A companson is also made to de-cludes a discussion on the pros and cons of the analysis in-termine the impact of various cross section hbraries on comput-volved in each of the methods, the role played by the key pa-ed dosimeter activities Significant improvements are obtained rameters influencing the formutation and a comparison of the with the ENDF/B-VI cross sections. Other factors, such as dif-results with the actual crack growth behavior observed in the vi-forences in group structures of multsgroup libraries, may also bration test program, Some conclusions and recommendations affect the calculated dosimetor activities for improvement of the methodologies are also provided.

NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUS-NUREGICR-6079: SElSMOLOGICAL INVESTIGATION OF TION BEHAVIOR OF HYDROGEN. AIR MIXTURES WITH TUH BULENT JET IGNITION AT LARGE SCALE. DOROFEEV,S.B.';

EARTHOUAKES IN THE NEW MADRID SEISMIC ZONE. Final BEZMELNITSiN,A.; EFIMENKO.A.A.; et al. Russian Research Report.Soptember 1986 - December 1992. HERRMANN.R.B.;

NGUYEN.B St. Louis Univ., St. Louis, MO. August 1993. 75pp.

Center (Kurchatov institute) June 1993. 83pp. 9308160272.

9309030148. 76325:084.

RRCKl 80-05/3. 76117:272.

This report desenbes research carned out in the KOPER facil' Earthquake activity in the New Madnd Seismic Zone had been monitored by regional seismic networks since 1975.

ity on spontaneous detonation ignition in hydrogen-air mixtures by turbulent jet ignition. The KOPER facihty is a sembconfined During this time period, over 3700 earthquakes have been 10-volume of 47.7 m(3) and consists of a steel canyon and a cated within the region bounded by latitudes 35 degrees 39 de-j robust frame placed above it. The frame sides are sealod with grees N and longitudes 87 degrees-92 degrees W. Most of thin polyethylene sheet A " Jet" chamber of 0.55 m(3), located these earthquakes occur within a 1,5 degree x 2 degree zone i

on the bottom of the canyon was used to produce a jet of hot centered on the Missouri Boothoel Source parameters of larger earthquakes'in the zone and in eastern North America are de-gases, which was vented into the hydrogen-air mixture. The ef-fects of three vanables were investigated: hydrogen concentra-termined using surface-wave spectral amplitudes and broad-band waveforms for the purpose of det)rmining the focal mech-tion (18-30% vol); jet onfice diameter (100 400 mm); and the composition of combustion products in the turbulent jet (by anism, source depth and seismic moment. Waveform modeling varying the hydrogen molo fraction in the "let"-chamber from 25 of broadband data is shown to be a powerful tool in defining to 50% vol). The possibility of init ation of turbulent combustion these source parameters when used complementary with re-l and local detonation was demonstrated Local detonation devel-gional seismic network data, and iri addition, in venfying the cor-ops after a deiay of 10-25 ms from ignition. For spontaneous rectness of previously published focal mechanism solutions.

detonation initiation, the minimum hydrogen concentration is NUREG/CR-6080: REPLACEMENT ENERGY, CAPACITY, AND within the range of 20 to 25% vol., and the minimum jet orifice RELIABILITY COSTS FOR PERMANENT NUCLEAR REACTOR diametor lies in the rango of 100 to 200 mm for the KOPER fa-SHUTDOWNS.

VANKUlKEN,J C.;

BUEHRING.W.A.;

cihty A minimum rr.tio of turbutent jet size L and mixture deto-HAMILTON,S.; et al. Argonne National Laboratory. October nation cell A, L/A. 1213 is required for detonation initiation 1993. 37pp. 9311180062. ANL-03/19. 77231:267.

which is supported by other type of turbulent jet initiation experi-Average replacement power costs are estimated for potential ments (ctosed votume and continuous venting) and by theorets-permanent shutdowns of nuclear electricity. generating units. Re-cal analysis.

placerm n p.Ar costs are considered to include replacement NUREG/CR-6073:

LYSIMETER LITERATURE REVIEW enor';y, capacity, and reliability cost components. These esti-ROGERS,R.D ; MCCONNELL J W. EG&G Idaho, Inc. August mares were developtd to assist the U.S. Nuclear Regulatory 1993. 75pp. 9309210201, EGG-2706. 76484 257.

Commission in evaluating regulatory issues that potentially Many reports have been pubbshed concerning the use of lysi, affect changes in senous reactor accident frequencies. Cost os-metars to obtain data on the performance of buned radioactive timates were derived from long-term production-cost and capac-waste. This document presents a review of some of those re, sty expansion simulations of pooled utihty-system operations.

ports This review includes fysimeter studies using radioactive Factors that affect replacement power cost, such as load waste forms at Savannah River Site, Hanford Site, Argonne Na.

growth, replacement sources of generation, and capital costs t onal Laboratory, and Oak Ridge National Laboratory; radionu.

for replacement capacity, were treated in the analysis, Costs chde tracer studies at Whiteshell Nuclear Research Estabbsh.

are presented for a representative reactor and for selected sub+

ment and Los Alamos National Laboratory, and water move.

Categories of reactors, based on estimates for 112 individual re-rnent studies at the Nuclear Regulatory Commission's Beltsville, actors.

Marytand site, at the Hanford Site, and at New Mexico State NUREG/CR-6081: ENHANCED REMOVAL OF RADIOACTIVE University. The tests. results, and conclusions of each report PARTICLES BY FLUOROCARBON SURFACTANT SOLU-are summanted and conclusions concerning fysimeter technol-TIONS. KAISER,R ; HARLING,0 K.

Entropic Systems, Inc.

ogy are presented from an overall analysis of the literatu o' August 1993. 97pp. 9309210192. 76484:156.

NUREG/CR-6078: ANALYSIS OF CRACK INITIATION AND The proposed research addressed the apphcation of ESI's GROWTH IN THE HIGH LEVEL VIBRATION TEST AT TA-particle removat process to the nondestructive decontamination DOTSU. KASSIR.M.K.; PARK,YJ; HOFMAYER C.H.; et al.

of nuclear equipment. The cleaning medium used in this proc.

Drookhaven National Latx>ratory. August 1993

82pp, ess is a solution of a high molecular weight fluorocarbon surfac-l 9309210185. BNL-NUREG-52383. 76484 075 tant in an inert perfluonnated hauid which results in enhanced I

The High level Vibration Test data are used to assess the particle removal, The perfluorinated hquids of interest, which are

(

accuracy and usefulness of current engineenng methodologies recycled in the process, are nontoxic, nonflammable, and envi-for predichng crack initiabon and growth in a cast stainless steel ronmentally compatible. and do not present a hazard to the t

i

44 Main Citations and Abstracts ozone layer. The information obtained in the Phase i program damago frequencies and pubhc nsks associated with failures of lndicated that the proposed ESI process is technically effective those systems, develops three proposed resolution strategies to and economically attractive. The fluorocarbon surfactant solu-this generic issue, and porforms a value/ impact analysis of the tions used as working media in the ESI process survived expo-proposed resolutions. Existing probabikstic risk assessments sure of up to 10 Mrad doses of gamma rays, and are consid-(PRAs) for four representativo plants, including one plant from cred sufficiently radiation resistant for the proposed process. UL each vendor, form the basis for the core damage frequency and trasonic cleaning in perfluonnated surfactant solutions was pubhc nsk calculations. Both internal and extomal events were found to be an effective method of removing radioactive iron considered. It was concludod that all throo proposed resolution J

(Fe 59) oxide particles from contaminated test pmces. Radioac-strategies excood the $1.,000/porson rem cost-offectiveness tive particles suspended in the process hquids could be quanti-ratio. Additional evaluations were performed to develop " geno'-

tatively removed by filtration through a 0.1 um membrane filtur.

ic" insights on potential design-and configuration-rotated vul-Projected economics indicato a pre' tax pay back time of I norabilities and potential high-Iroquency accident sequences, it month for a commercial scale system.

was concludod that some high-frequency toquences exist but are plant-specific in nature or have been resolved through hard-NUREG/CR4082: DATA COMMUNICATIONS PRECKSHOT,G G ware and/w operational changes. The plant specific Individual Lawrer'ce Uvermore Nabonal Laboratory August 1993 06pp Plant Examinations (tPEs) are an effective vehicle for identifica-9309030171. UCRL-ID-114567. 76320 219 tion and resolution of these plant-specific anomalies and hard-The purpose of this papor is to recommend regulatory guid..

ware conhgurations.

ance for reviewers examining computer commun!catit $ systems used in nuclear power plants. The recommendatic.s cover NUREG/CR-6085: UNITED STATES SEISMOGRAPHIC NET.

three areas important to these communications systems' WORK. BULAND,R. Interior, Dept. of, Geol 09ical Survey. Sop-system design, communication protocols, and communication tembm IM 83pp. 93W20339. 767R02R mods The hrst area, system design, considers throo aspects of The concept of a United Statos National Seismograph Net.

system design-queshons about architecture, specific nsky design einments or omissions to k>ok f <r in design $ being re.

work (USNSN) dates back nearly 30 years. The kloa was re-vived several timos over the decades, but never fundod. For ex-viewed, and recommendations for muhiplexed data communica hon systems used in r,atety systems The socrad area revnws amplo, a national network was proposed and discussed at great portinnnt aspects of communication protocol design and makes longth in the so called Bolt Report (U S. Earthquake Observa-tones: Rocommendations for a New National Network, National recommendations for newly designed protocols or the selection of existing protocols for safety system, information d6 play, and Academy Press, Washington, D C. 1980,122 pp). From the be-non-safety control system use. The third area covers communi.

ginning, a national network was viewed as augmenting and cation media selection, which differs significantly from traditional complomontin0 the rotatively densa, predominantly short period waro and cable Tho recommendations for communication rnedia vertical coverage of selected areas provided by the Regional extend or enhance the concerns of pubhshed IEEE standards Soismograph Networks (RSN's) with a sparse, well-distnbuted about three subjects: data rato,impoited hazards and maintain-network of throo. component, observatory quality, permanent stations The opportunity finally to bo0in developing a national ability NUREG/CR-6083: REVIEWING REAL, TIME PERFORMANCE OF NUCLEAR REACTOR SAFETY SYSTEMS PRECKSHOT,G G-(NRC). Under the agrooment signed in 1987, the NRC has pro-l Lawrence Uvermore National Laboratory. August 1993. 88pp' vided $5 M in new funding for capital equipment (over the 930903011?. UCHL ID 114565 7G325 001.

ponod 19871992) and the USGS has provided personnel and The purpose os this paper is to recommand regulatory guid.

facihbos to develop, doploy, and operate the network. Docauso ance for reviewers examining real-bmo portormance of comput-the NRC funding was oarmarked for the eastern United Statos, nr based safety systems used in nuclear power plants. Three now USNSN station deployments are mostly east of 105 de-areas of guidanco are covered in this report. The hrst area grees West longitudo while the network in the western United covers how to determino if. when, and what prototypes should States is mostly made up of cooporating stations (stations be required of developers to make a convincing demonstration meeting USNSN design goals, but deployed and operated by i

that spec!he problems have been solved or that performanco other institutions which provido a logical extension to the j

goals have boon mat. Tho second area has recommendations USNSN

(

for timing analysos that will prove that the real-timo system will j

moet its safetyamposed deadlines The third area has descnp-HUREG/CR-6090: THE PROGRAMMA0LE LOGIC CONTROLLER tions of means for assessing expected or actual real timo por-AND ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS.

formance before, dunng, and after development is completod-PALOMAR,Ja WYMAN,R. Lawrence Livermore Nabonal Labora.

To ensure that the dobverod real bmo software product moots tory. September 1993 96pp. 9310120250. UCnL40112900.

performance goals. the paper recommonds certain types of 76740 170.

code exncubon and communications scheduhng Technical This document providos recommendations to guido revlowers background is provided in the appondix on methods of timing in the application of Programmable Logic Controllers (PLCs) to analysis, scheduhng real time computahons, prototyping. real-the control, monitoring and protection of nuclear reactors The bme software development approachos, morkhng and monsuro^

first topics addressed are system-levol design issues, specificah mont, and real t me operating systems ly including safety. The document then descussos concerns NUREG/CH-6084: VALUE-lMPACT ANALYSIS OF GENERIC about the PLC manufactunng organization and the protection ISSUE 143, "AVAILADIUTY OF HEATING, VENTILATION, AIR system engineoeing organization. Supplernenting this document CONDITIONING (HVAC) AND CHILLED WATER SYSTEMS."

are two appendices. Appendix A summantes PLC charactens-DAUNG.P M ; M ARLER.J Ea VO.T V. Battelle Momonal Instituto, tics. Specif cally uddressed are those charactoristics that make Pacihc Northwest L aboratory. November 1993 700pp.

the PLC more suitable for emergency shutdown systems than 0312170083 PNL-8750 77513.001.

other electrical / electronic based systems, as well as character.

This study evaluates the values and impacts associated with istics that improve rehabihty of a system. Also covered are PLC potenhal resolubons to Gonoric Innue 143, " Availability of HVAC charactonstics that may creato an unsafe operating environ-and Challed water Systems." The study identifies vulnorabibtics mont. Appendix 0 provides an overview of the use o' program-rotated to f ailures of HVAC, chillod water, and room cooling sys-mable logic controllers in emergency shutdown systems. The tems, develops estimatos of room heatup rates and sa ety-relat.

intent is to famshante the reador with the dosign, development, r

I ed equeprnent vulnerabihties, dovolops estimaten of the core test, and maintenance phases of applying a PLC to en CSD l

l

-,,-_-----_-_.u a

Main Citations and Abstracts 45 system Estn phase is desenbod in detail and information porti-desert soil, the model preMts that intium migration will be short nent to the appbcation of a PLC is pointed out range, with a maximum o, a few meters.

NUREG/CR-6098: LOADING HATE EFFECTS ON STRENGTH NUREG/CR-6111: INTEGRATED SYSTEMS ANALYSIS OF THE AND FRACTURE TOUGHNESS OF PIPE STEELS USED IN PlUS REAC.'OR. FULLWOOD,F.; KROEGER.P.G.; HIGGINS,J.;

TASK 1 OF THE IPIRG PROGRAM. MARSCHALL,C.W.;

et af. Brookhaven National Laboratory. November 1993.450pp.

LANDOW,M P.; WILKOWSki,G M Battelle Momorial Institute, 9312160278. BNL-NUREG-52393. 77505:198.

Columbus Laboratories October 1993.129pp 9311080142-Results are presented of a systems failure analysis of the DMI-2175 77084'041.

PlUS plant systems that are used dunng normal reactor oper.

Matena! characterization tests were conducted on laboratory ation and postulated accidents. This study was performed to specimens machined from pipes to determine the effect of dy-provide the NRC with an understanding of the behavior of the namic loading (ie., rates comparable to those for high amplF plant The study apphed two diverse farture identification moth-tude seismic evente ^n tensile properties and fracture resist-ods. Failure Modes Effects & Cnticall'y Analysis (FMECA) and ance at 288 dagrees J (550 degrees F) Specimens were fabrL Hazards & Operability (HAZOP) to the plant systems, supported cated from seven difforont pipes, including carbon steels and by several deterministic analyses. Conventional PRA methods stainless steels (both base metal and weld metal), which were were also used along with a scheme for classifying events by to be subjected to full-scale pipe tests in IPIRG Task 1.0. For initiator frequency and combinations of f ailures. Pnncipal results the stainless steels tested at 288 degrees C (550 degrees F;-

of this study are: (a) an extenssve listing of potential event se-tensile strength was unchanged, whiie yield strength and frac' quences, grouped in categones that can bo usod by the NRC; ture resistance were increased. The increase in fracture resist-(b) identification of support systems that are important to safety; ance was modest for the wrought base metals and substantW and (c) identrhcation of key oporator actions.

for the weld metal and the cast base metal. The carbon ttea tested were sensitive to dynamic strain aging, and hence the NUREG/CR-6113: CLASS 1E DIGITAL SYSTEMS STUDIES.

strength and toughness was affected by both temperature and HECHT,H.; TAl.A.T.: TSO,K.S. SoHaR, Inc. October 1993, strain rate effects. The cart on steel base metal and wolds ex-220pp. 9311080138. 77083.139.

hibited ultimate tensile strength values at 288 degrees C (550 This document is fumished as part of the effort to develop degrees F) that were greater than at room temperature Further-NRC Class IE Digital Computer Systems Guidelines which is more, the ultimate tonsile strength at 288 degrees C (550 de-Task 8 of USAF Rome Laboratories Contract F30602-89-D-groos F) was lowered s6gnificantly by increased strain rate and.

0100. The report addresses four rnajor topics, namely, computer in the carbon steel base metals, increased strain rate also low-programming languages, software design and development, ered the fracture resistance, substantially in the base metal of software testing and fault tolerance and fault avoidance. The one pipe topics are intended as stepping stones leading to a Draft Regu-NUREG/CR-6101: SOFTWARE RELIABILITY AND SAFETY IN latory Guide document. As part of this task a small scale survey

]

NUCLEAR REACTOR PROTECTION SYSTEMS.

of software fault avoidance and fault tolerance practices was LAWRENCE,J D. Lawrence Livermore National Laboratory. No.

conducted among vendors of nuclear safety related systems j

vember 1993.1Sopp 9312220167. 77544:011.

and among agencies that develop software for other applica-j Planning the development, use and regulation of computer tions demanding very high rehabibty. The findings of the present systems in nuclear teactor protection systems in such a way as report are in part basad on the survey and in part on review of to enhance tellatnhty and safety is a complex issue. This report software literature relating to nuclear and other entical installa-is one of a senes of reports from the Computer Salety and Role tons, as well as on the authors' experience in those areas.

ability Group. Lawrence Livermore National Laboratory, that in' NUREG/CR-6114 V01: APPLICATION OF AN INFILTRATION vestigates different aspects of computer software in reacior pro-EVALUATION METHODOLOGY TO A HYPOTHETICAL LOW-tectico systems. There are two central themes in the report LEVEL WASTE DISPOSAL FACILITY. MEYER.P.D. Battelle Me-First software considerahons cannot be fully understood in iso-lation from computer ha;dware and application considerations.

morial Institute, Pacific Northwest Laboratory. December 1993.

42pp. 9401140021. PNL-8842. 77797;138.

Second, the process of engineenng rehability and safety into a An analysis of infiltration and percolation at a hypothetical computnr system requaes activities to be carned out throughout low-level waste (LLW) disposal facility was carried out. The the software life cycle. The report discusses the many actrvaties that can he carned out dunng the software life cycle to improve analysis was intended to illustrate general issues of concern as-the safety and reliatulity of the resulting product The viewpoint sessing the performance of LLW disposal facilities. Among the processes considered in the analysis were precipitation, runoff, is pnmarily that of the assessor, or auditor.

infiltration, evaporation, transpiration, and redistobuton. The hy-NURLG/CR-6108: SPHERICAL DIFFUSION OF TRITIUM f ROM pothetical facihty was located in a humid environment character-I A POINT OF RELEASE IN A UNIFORM l iSATURATED ired by frequent and often intense precipstation events. The fa-SOIL A Deterministic Model For Tntium Migration in An And Dis-cikty consisteri of a series of concrete vaults topped by a multi-posal Site. SMILES.D E.;

ARDNER,W R.; SCHUL 2,R K. Cah-layer cover. Cover features included a sloping soil surface to for nia.

Univ. of, Berkeley, CA.

October 1993 38pp.

promote runoff, plant growth to minirnize erosion and promote 93 t 1080128. 7 7034 001, transpirataon, a sloping clay layer, and a sloping capillary bamer, This report presents a three-dimensional modet for tntium me The analysis within the root zone was carried out using a one-

)

gration in an arid waste disposal site. When tntiated water is re.

darnensional, transient simulation of water flow. Below the foot i

leased at a point in a uniform and relatively dry soll it redistnb-zone, the Snalysis was pnmarily two-dimensional and steady-utes in both the liquid and vapor phases. The flux density of tnt-state. Results of the situations illustrated the hmited value of rum en each phase es of tho samo order of magnitude however daily precipitation data. For the humid site studied, hourly rain-so tntium redistnbution is modeled as if transfer occurs "in par-fall data provided significantly better estimates of the water bal-allel" in the hquid and vapor phases. The approach we describe ance. Results also demonstrated the importance of transpiration uses the diffusion equation cast in radial (sphenca!) coordinates in removing water from the soil column, implying a need for ac-and talies into account radioactivo decay. It permits calculation curata models of plant growth and water utilization. In add. tion, of radial profiles of intium concentration, within and extemal to the amount of water prodicted to percolate below the root Zone a sphere of released solution. We assume the concentration was often less than the amount roquered to keep the clay barrier within this sphere mitially to be uniform. The solution also pre.

Iayer fully saturated, even in the relatwely wet environment dicts attenuation and rate of advance of the maximum of tntium studied. This could be a concom if the clay were subject to concentratinn as it advances in the soil. With deep disposal in a shnnhng undnr unsaturated conditions The two-dimensional

46 Main Citations and Abstracts simulations showed that the sloping clay barrier diverted 75 per.

volve five major steps: defining the system; evaluating qualita-cent of the water reaching it. The sloping capillary bamer, in tive risk assessment results; using this and information from contrast, diverted more than 99.99 percent of the water reach-plant probabilistic risk assessments to perform a quantitative ing it. Performance of the capillary bamer, however, was shown nsk analysis; selecting target failure probabilities; and develop-to wary significantly with the hydraulic properties of the two ma-ing an inspection program for components using economic deci-terials of which it is composed. Predicting performance simply sion analysis and structural reliability assessment rnethods, in-by inspecting the water retention and hydraulic conductivity cluded: extensive bibliography. Companion Volume 2. Part 2 functions was difficult. An analytical expression was presented document will recommend risk-based inspection program for i

that can be used to estimate capillary barrier performance and consideration by Section XI of the ASME Boiler and Pressure to determine appropnate matenals for construction.

Vessel Code.

NUREG/CR-6117: NEUTRON SPECTRA AT DIFFERENT HIGH NUREG/GR-0006: DEPOSITION: SOFTWARE TO CALCULATE FLUX ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SUR-PARTICLE PENETRATION THROUGH AEROSOL TRANS-1 VEILLANCE LOCATIONS. REMEC,l. Josef Sefan institute.

PORT SYSTEMS Final Report.

ANAND,N.K.;

j KAM.F B. Oak Ridge National Laboratory. December 1993.

MCFARLAND,A.R.; WONG,F.S.; et al. Texas A&M Univ., Col-137pp.9401140025. ORNL/TM-12484. 77797.177, lege Station, TX. Apnl 1993. 45pp. 9305100008. 74858:223.

This project addresses the potential problem of radiation em-User-friendly software (DEPOSITION 2.0) has been devel-brittlement of reactor pressure vessel (RPV) supports, Surveil-oped which permits characterization of aerosol particle losses in lance specimens irradiated at the High Flux isotope Reactor transport systems. The sub-models which comprise the DEPO-i (HFIR) at relatively low neutron flux levels (about 1.5E+ 8 cm(-

SITION code are presented and the limitations of these sub-

2) s( 1)) and low temperature (about 50 degrees C) showed em-models are noted. These sub models have all been previously britt!ement more rapidly than expected, Commercial power reac" published in the peer-reviewed literature. The software can be tors have similar flux levels and temperatures at the vessel sup-used to determine the penetration of aerosol through existing port structures. The purposes of this work are to provide the transport systems; it will provide the optimal tube diameter for a neutron fluence spectra data that are needed to evaluate previ-transport system operated at a given flow rate and at a given

[

ously measured mechanical property changes in the HFIR, to particle size; it will provide a value for the rnaximum penetration explain the discrepancies in neutron flux levels between the for a transport system that would connect two points in three.

nickel dosimeters and two other dosimeters, neptunium and be' dimensional space; and, it will provide tables of data and Create I

ryllium, and to address any questions or peculianties of the output files for parametric studies on the effects of varying parti-HFIR reactor environment.

cle size, flow rate and tube diameter. Use of this software for NUREG/CR-6118: ASSESSMENT OF THE EFFECTIVENESS OF specific examples is given herewith in an Appendix. Reference THE LEU REFORM RULE AND ITS IMPLEMENTATION.

to this software is included in NRC Regulatev Guide 8.25 MORAN.B W.; NATIONS.J.O. Martin Marietta Energy Systems, (1992) where it is considered to be an acceptable method for Inc. HAMMOND.G.A. 21st Century industnes. Inc. November calculating the penetration of particles through sampling sys-l 1993. 34pp. 9312070227. K/NSP-117. 77349.295.

tems.

The U S. Nuclear Regulatory Commission (NRC) amended its material control and accounting (MC&A) requirements in 1985 AND ITS APPLICATION TO SEVERE ACCtDENT PHENOM-for licensees possessing and using special nuclear material ENA. ISHH,M.; ZHANG,G. Purdue Univ., West Lafayette, IN.

(SNM) of low strategic significance in quantities larger than one NO,H.C. Korea Advanced Institute of Science and Technology.

effective lulogram (kg). The goal of the Low-Ennched Uranium October 1993.109pp. 9311080286. 77077:109.

(LEU) Reform Rule (i E.,10CFR 74.31) was to estab!ish MC&A Severe accidents in hght water reactors are characterized by requirements for the LEU bcensees at a level consistent with an occurrence of multiphase flow with complicated phase the safeguards risk associated with the relatively low strategic importance of such matenal. The amended requirements were changes, chemical reaction and vanous bifurcation phenomena.

Because of the inherent difficulties associated with fuliscale wntten in a performance-onented manner, rather than a pre.

senptive one, in an effort to allow the hcensees the opportunity testing, scaled down and simulation experiments are an essen-tial part of the severe accident analyses. However, one of the i

to choose the most cost effactive means of satisfying the re-quirements. The LEU Reform Rule was implemented in January most significant shortcomings in the area,s the lack of well-es-i tablished and reliable scaling method and scaling enteria. In

{

1988 and the fuel cycle facihties have had sufficient experience in implementing the rule to allow a moaningful review of its ef-view of this, the stepwise integral scaling method is developed i

for severe acckfent anafyses. This new scaling method is quite fectiveness. This document provides technical analysis and rec.

different from the conventional approach. However, its focus on i

ommendations to assist the NRC in making a determination if the rule is achieving its intended purpose, and if not, to make dominant transport mechanisms and use of the integral re-sponse of ine system make this method relatrvely simple to the necessary changes to accomplish this.

apply to very complicated multi-phase flow problems. In order to NUREG/GR-0005 V02 P1: RISK-BASED INSPECTION-DEVELOP-demonstrate its applicability and usefulness, three case studies MENT OF GUIDELINES Light Water Reactor (LWR) Nuclear have been made. The phenomena considered are: 1) conum i

Power Plant Components.

• Amencan Society of Mechanical dispersion in DCH; 2) corium spreading in BWR MARK l con-Engineers. July 1993.173pp. 9308200281. CRTD-VOL.20-2.

tainment, and 3) incore boit-off and heating process. The results 76159.209.

of these studies clearty indicate the effectiveness of their step.

Effective inservice inspection programs can play a significant wise integMI scaling method. Such a simple and systematic role in minirnizing equipment and structural failures Most of the scahng method has not been previously available to severe ac-i current inservice trupection programs for light water reactor cident analysis.

I (LWR) nuclear power plant components are based on experi, ence and ongineers' qualitative judgment. These programs in-NUREG/GR-0010: HYBRID DIGITAL SIGNAL PROCESSING AND ciude only an imphcit consideration of risk, which combines the NEURAL NETWORKS FOR AUTOMATED DIAGNOSTICS probability of failure of a component under its operation and USING NDE METHODS. UPADHYAYA,B.R.; YAN,W. Tennes-i loading conditions and the consequences of such failure, if it see, Univ. of, Knoxville, TN. November 1993. 150pp.

occurs. This document recommends appropnate methods for 9312220131. 77543:001.

[

establishing a risk-based inspection program for LWR nuclear The pnmary purpose of the current research was to develop power plant components. The process has been built from a an integrated approach by combining information compression general methodology fVolume 1) and has been expanded to in-methods and artificial neural networks for the monitoring of

'i 4

l I

k Main Citations and Abstracts 47 plant components using nondestructive examination data. Spe.

NUREG/lA-0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUT.

crfically, data from eddy current inspection of heat exchanger ER CODE AGAINST THE NET LOAD TRIP TEST DATA FROM tubing were utilized to evaluate this technology. The focus of YONG GWANG, UNIT 2. ARNE,N.; CHO,S. Korea Electric Fower the research was to develop and test vanous data compression Corp. LEE,S.H. Korea InstifA nf Noc! ear Safety. June 1993.

methods (for eddy current data) and the performance of differ-75pp, 0306290172. 754981.39.

ent neural network paradigms for defect classification and The results of the RELAP5/ MOD 2 computer coc. simulation defect parameter estimatiott Feedforward, fully-connected for the 100% Net Load Trip Test in Yong Gwang Unit 2 are neural networks, that uso the back-propagation algonthm for analyzed here and compared with the plant operation data. The network training, were implomented for defect classification and control systems for the control rod, feedwater, steam generator defect parameter estimation using a modular network architec^

level, steam dump, pressurrzer level and pressure are modeled ture. A large eddy current tube inspection database was ac-to be functioned automatically until the power level decreases quired from the Metals and Ceramics Division of ORNL. These below 30% nuclear power. A sensitivity study on control rod data were used to study the performance of artificial neural net' worth was carried out and it was found that variable rod worth works for defect type classification and for estimating defect pa-should be used to achieve good prediction of neutron power.

rameters. A PC-based data preprocessing and display program was also developed as part of an expert system for data man-The results obtained from RELAPS/ MOD 2 simulation agree well agement and decision making. The results of the analysis with the plant operating data and it can be concluded that this showed that for effective (low-error) defect classification and code has the capability in analyzing the transient of this type in -

a best estimate means.

estimation of parameters, it is necessary to identify proper fea-ture vectors using different data representation methods. The NUREG/lA-0094: ASSESSMENT OF RELAPS/ MOD 3 AGAINST integration of data compression and artificial neural networks for TWENTY-FIVE POST-ORYOUT EXPERIMENTS PERFORMED information processing was established as an effective tech-AT THE ROYAL INSTITUTE OF TECHNOLOGY. N1LSSON,L nique for automation of diagnostics using nondestructive exami-nation methods Swedish Nuclear Power inspectorate (Statens Karnkraftinspek.

tion). May 1993. 92pp. 9306110035. STUDSVIKNS90/93.

MUREG/lA-0085: ASSESSMENT OF FULL POWER TURBINE 75377:080.

TRIP START-UP TEST FOR C. TRILLO I WITH RELAPS/

Assessment of RELAPS/ MOD 2 has been made against vari-MOD 2. LOZANO,M.F.; MORENO,P.; DE LA CAL,C.; et al. Spain, ous experimental data, among other data from twenty-five post-Govt. of. July 1993. 73pp. 9308160158. ICSP TR.TTRIP-R.

dryout experiments conducted at the Royal Institute of Technol-76119.218.

ogy (RIT) in Stockholm. As the MOD 3 version of RELAPS has C. Tnilo I has developed a model of the plant with RELAPS/

now been released, incorporatiag a different method of calculat.

MOD 2/36.04 This model will be validated against a selected ing entical heat flux compaied to RELAPS/ MOD 2, it seemed set of start-up tests. One of the transients selected to that aim just2fied to make another assessment against the same RIT.

is the turbine inp, which presents very specific charactenstics data. The results show that the exial dryout position is generally that make it significantly different from the same trarissent in better predicted by the MOD 3 than by the MOD 2 version. The other PWRs of different design, the mon difference being that prediction is, however, still nonconservative, i.e. the calcu!ated the reactor is not tnpped: a reduction in pnmary power is carried dryout position falls in most cases downstrearn the actual meas-out instead. Pre. test calculations were done of the Turbine Tnp ured point. While the pre-dryout heat transfer seems to be equal Test and compared against the actual test. Minor problems in for MOD 2 and MOD 3, both versions giving slightly higher wall the first model, specially in the Control and Limitation Systems, temperatures than the expenments, there is a considerable dil-were identified and post-test calculations had been carned out-ference in the post-dryout heat transfer. The results of the RIT The results show a good agreement with data for all the com-data comparison indicate that MOD 3 underpredicts the post-pared vanables' dryout wall temperatures remarkably while MOD 2 gave reasona-1.

NUREG/lA-0090: ASSESSMENT OF RELAP5/ MOD 2 USING THE ble agreement. In this respect RELAP/ MOD 3 shows no im-l TEST DATA OF REWET-il REFLOODING EXPERIMENT SGI/

provenent over OD2.

R. HAMALAINEN.A. Technical Research Centre of Finland NUREG/lA-00P 4ELAP5 ASSESSMENT USIfJG LSTF TEST (VTT). May 1993. 44pp. 9306230098. 75462:278 DATA SB-CL 16. LEE,S.H.; CHUNG,B-D.; KIM,H-J. Korea Insti-An analyses of a reflooding experiment wrth RELAP5/ MOD 2 tute of Nuclear Safety. May 1993. 100pp. 9306210225.

cycle 36 04 is presented. The expenment had been camed out t;

in the REWET-Il facil:ty simulating the reactor core with a 75402:065

[

bundle of 1,9 electrically heated rods. On the basis of the results 5% cold log break test, run SB-CL-18, conducted at the of two calculations recommendations for the core nodalization Large Scale Test Facility (LSTF) was analyzed using RELAP5/

are presented, and a modification to the code is proposed.

MOD 2 Cycle 36.04 and RELAP5/ MOD 3 Version Sm5 codes.

t The test was conducted with the main objective being the in-NUREG/lA-0091: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A vestigation of thermal-hydraulic mechanisms responsible for NATURAL CIRCULATION EXPERIMENT IN NUCLEAR POWER early core uncovery, including manometric effect due to an PLANT BORSSELE. WINTERS L Netherlands Energy Research asymmetric coolant holdup in the steam generator upflow and Foundation ECN. July 1993. 66pp. 9308160307. ECN 89-91, downilow side. The present analysis, carried out with RELAPS/

76122-247.

MOD 2 and MOD 3 codes, demonstrates the code's capability to As part of the ICAP (International Code Assessment and Ap-predict, with sufficient accuracy, the main phenomena occurring i

(

plications Program) agreement between ECN (Netherlands in the depressurization transient, both from a qualitative and h

Energy Research Foundation) and USNRC, ECN has performed quantitative point of view. Nevertheless, several differences re-l a number of assessment calculations for the thermohydraulic garding the evolution of phenomena and affecting the timing

(

system analysis code RELAP5/ MOD 2/36 05 This document order have to be pointed out in the base calculations. The sen-describes the assessment of this computer program versus a sitivity study on the break flow and the nodalization study in the natural circulation expenment as conducted at the Borssele Nu-components of the steam generator U-tubes and the cross,over clear Power Plar1. The results of this companson show that the legs were also carried out. The RELAP5/ MOD 3 calculation with code RELAPS/ MOD 2 predicts well the natural circulation be-the nodal:Iation change resulted in good predictions of the havior of Nuclear Power Plant Borssole. The work has been major thermaLhydraulic phenomena and their timing order.

sponsored by the Dutch Licensing Authonty and ECN.

l.

f i.

Iu a

i.

48 Main Citations and Abstracts NUREG/lA 0096: NUMERICS AND IMPLEMENTATION OF THE cy, the main phenomena occurnng in the depressunzation tran.

UK HORIZONTAL STRATIFICATION ENTRAINMENT OFF-sierA both from a quahtatue and quantitative point of view.

TAKE MODEL INTO RELAPS/ MOD 3 BRYCE.W.M. Winfnth Nevertheless, several differences regarding the evolubon of Technology Centre (United Kingdom). June 1993, 47pp, phernmena axi affecting the timing order have to be pointed 9307220307, AEA-TRS-1050. 75743.253.

out in the bats calculation. Three calculations were carried out This report presents the numencs and implementation details to study the sensitivity to change of the nodalization in the com-to add the same improved discharge quality correlations into ponents of the loop seal cross-over legs, and of the auxihary RELAP5/ MOD 3. Iri the light of experience with the modified feedwater control logics, and of the break discharge coefficient.

RELAPS/ MOD 2 code, some of the numencs have been sEghtly changed for RELAP5/ MOD 3. The desenption is quite detailed in NUREG/iA-0104: RELAPS/ MOO 3 ASSESSMENT USING THE order to facilitate change by some future code developer. A SEMISCALE 50% FEED LINE BREAK TEST S-FS.11. LEE,E J.;

simple test calculation was performed to confirm the coding of CHUNG,B-D.; KIM.H-J. Korea Institute of Nuclear Safety. June the correlations implemented in RELAPS/ MOD 3.

1993. 200pp. 9306290153. 75497:272.

NUREG/lA-0099: RELAPS ASSESSMENT USING SEMISCALE The RELAPS/ MOD 3 Sm5 code was assessed using the 1/

SBLOCA TEST S-NH-1. LEE.E-J.; CHUNG,B-D.; KIM.H-J. Korea 1705 volume scaled Semiscale 50% Feed Line Break (FLB) institute of Nuclear Safety. June 1993. '50pp. 9306290162.

test S-FS-11. Test S FS-11 was designed in three phases: (a) 75498:090 blowdown phase, (b) stabilization phase, and (c) refill phase.

2anch cold leg break test S-NH-1, conductod at the 1/1705 The fist objectwe was to assess the code applicability to 50%

volume scaled facility Semiscale, was analyzed using RELAP5/

FLB situation, the second was to evaluate the FSAR conserv-MOD 2 Cycle 36.04 and MOD 3 Version SmS. Loss of HPIS was atisms regarding SG heat transfer degradation, steam line assumed, and reactor trip occurred on a low PZR pressure check valve failure, break flow state, and peak pnmary system signal (13.1 MPa), and pur ps began an unpowered coastdown pressure, and the third was to validate the EOP effectiveness.

on Si signal (12.$ MPa). The system was recovered by opening The code was able to simulate the major T/H parameters ADVs when the PCT becamo higher than 811 K. Accumulator except for the two-phase break flow and the secondary convec-was finalty injected into the system when the primary system ttve heat transfer rate. The two-phase break flow had still defi-pressure was less than 4.0 MPa. The expenment was terminat.

ciencies. The current boiling heat transfer rate was developed ed when the pressure reached the LPIS actuation set point.

from the data for flow inside of a heated tube, not for flow RELAPS/ MOD 2 analyses demonstrated its capability to predict, around heated tubes in a tube bundle. Results indicated that with a sufficient accuracy, the main phenomena occumng in the the assumption of 100% heat transfer until the hquid inventory depressunzation transient, both from a quahtatue and quantita.

depletion was not conservatwe, the failed affected steam gener-two points of view. Nevertheless, several differences were ator main steam line check valve assumption was not either noted regarding the break flow rate and inventory distribution conservative, the measured break flow experienced all types of due to deficiencies in two. phase choked flow model, horizontal flow conditions, the relative proximity to the 110% design pres-stratification interfacial drag, and a CCFL model The main sure limit was conservative. The automatic actions during the reason for the core to remain nearly fully covered with the liquid blowdown phase were effectwo in mstigating the consequences.

was the under-prediction of the break flow by the code. Several The stabilization operation performed by operator actions were sensitwity calculatsons were tried using the MOD 2 to improve effectue to permit natural circulation cooldown and depressuri-l the results by using the dJferent opbons of break flow modeling zation. The voided secondary refill operations also verified the r

(downward, homogeneous, and area increase). The break area effectueness of the operations while recovenng the inventory in compensating concept based on "the integrated break flow a voided steam generator..

matching" gave the best results than downward junction and l

homogeneous options. And the MOD 3 showed improvement in NUREG/lA-0105: ASSESSMENT OF RELAP5/ MOD 3 VERSION predicting a CCFL iri SG and a heatup in the core SMS USING INADVERTENT SAFETY INJECTION INCfDENT NUREG/lA-0100- ASSESSMENT OF CCFL MODEL OF RELAPS/

l MOD 3 AGAINST SIMPLE VERTICAL TUBES AND ROD 6

.9 0618 330 7 3 7 32 BUNDLE TESTS. CHO.S.; ARNE,N. Korea Electnc Power Corp.

An inadvertent safety injectihin incident occurred at Kori Unit 3 i

CHUNG,B-D.; et al. Korea institute of Nuclear Safety. June 1993 122pp. 9307220299 75744:001.

in September 6,1990. It was analyzed using the RELAP5/

The CCFL model used in RELAPS/ MOD 3 version Sms has MOD 3 code. The event was initiated by a closure of main feed-been assessed against simple vertical tubes and rod bundle water control valve of one of three steam generators. High tests performed at a facility of Korea Atomic Energy Research pressure safety Wection system was actuated by the low pres-sure signal of ma,n steam line. The actual sequerte of plant i

Institute. The effect of changes in tubo diameter and nodaliza-tion of tube section were investigated. The roles of interfacial transient with the proper estimations of operator actions was in-drags on the flooding characteristics are discussed. Differences vestigated in the present calculation. The asymmetnc loop be-between the calculation and the experiment are also discussed.

haviors of the plant were also considered by nodalizing the A comparison between model assessment results and the test loops of the plant into three. The calculational results are com-data showed that the calculated value lay well on the expen-pared with the plant transient data. It is shown that the overall j

montal flooding curve specified by user, but the pressure jump plant transient depends strongly on the auxiliary feedwater flow-l before onset of flooding was not calculated.

rate controlled by the operator and that the code gwes an ac-ceptable prediction of the plant behavior with the proper as-NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 9.1.8 USING sumptions of the operator actions. The results also show that RELAPS/ MOD 3. LEE,S.H.; CHUNG.B-D.; KIM.H-J. Korea insti-the solidification of pressurizer does not occur and the liquid-.

tute of Nuclear Safety. June 1993. 250pp. 9307220272.

vapor mixture does not flow out through pressurizer PORV. The 75743 001, behavior of primary pressure during pressunzer POHV actuation 2" cold leg break test 91.b, conducted at the BETHSY facility is poorly predicted because the actual behavior of pressurizer was anatyred using the RELAP5/ MOD 3 Version Sm5 code. The PORV could not be modelled in the present simulation.

test 9.1.b was conducted with the main objectwe being the in-vestigation of the thermaLhydraulic mechanisms responsible for NUREG/lA-0106: ASSESSMENT OF PWR STEAM GENERATOR l

the large core uncovery and fuel heat-up, requiring the imple.

MODELLING IN REI APS/ MOD 2. PUTNEY.J.M.; PREECE R.J.

montation of an ultimate procedure. The present anatysis dem-National Power (United Kingdom). June 1993.' 123pp.

onstrates the code's capabihty to predet, with sufficient accura-9307120157. TEC/L/001/R91. 75623:025.

l

Main Citations and Abstracts 49 An assessment of Steam Generator (SG) modelling in the ll Nuclear Power Plant. The ANV collaboration consisted in the PWR thermal-hydraulic code RELAPS/ MOD 2 is presented. The supply of design and actual data, the cooperation in the simula-assessment is based on a review of code assessment calcula-tion of the control systems and other model components, as tions performed in the UK and elsewhere, detailed calculations well as in the results analysis. The obtained model has been as-against a series of commissioning tests carried out on the Wolf sessed against the following transients occurred in plant: Atnp Creek PWR and analytical investigations of the phenomena in-from the 100% power level (CSN); a load rejection from 100%

volved in normal and abnormal SG operation. A number of mod-to 50% (CSN): a load rejection from 75% to 65% (ANV); a elling deficiencies are identified and their implications for PWR feedwater turbopump tnp (ANV). This copy is a report of the safety analysis are discussed - including methods for compen-feedwater turbopump inp transient simulation. This transient oc-sating for the deficiencies through changes to the input deck.

Currod actually in plant on June 19,1989.

Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP/ MOD 3.

NUREG/lA-0112: ASSESSMENT OF RELAP5/ MOD 2 AGAINST ECN-REFLOOD EXPERIMENTS.

WOUDSTRA.A.;

NUREG/lA-0107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A VANDEBOGAARD,J.; STOOP P.M.

Netherlands Energy Re-LOAD REJECTION FROM 100% TO 50% POWER IN THE search Foundation ECN. July 1993. 87pp. 9308160054. ECN.C-VANDELLOS 11 NUCLEAR POWER PLANT. LLOPIS.C.;92-008. 76121:243.

MENDIZADAL.R.; PEREZ,J. Spain, GovL of. June 1993. 53pp-As part of the ICAP (International Code Assessment and Ap-3307220264. ICSP-V2-R50-R. 75742:241.

plications Program) agreement between ECN (Netherlands An assessment of RELAPS/ MOD 2 cycle 36.04 against a load Energy Research Foundation) and USNRC, ECN has performed rejection from 100% to 50% power in Vandellos 11 NPP (Spain) a number of assessment calculations with the computer pro-is presented. The work is inscribed in the framework of the grarn RELAP5. This report describes the results as obtained by Spanish contnbution to ICAP Project. The model used in the ECN from the assessment of the thermohydraulic computer pro-simulation consists of a single loop, a steam generator and a gram RELAP5/ MOD 2/CY 36.05 versus a series of reflood ev steam line up to the steam header, all of them enlarged on a periments in a bundle geometry. A total number of seven se.

e scale of 3:1; and full-scaled reactor vessel and pressunzer. The lected experiments have been analyzed, from the reflood exper.

results of the calculations have been in reasonable agreement imental program as previously conducted by ECN under con-with plant measurements.

tract of the Commission of the European Communities (CEC). In NUREG/lA-0108: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A this document, the results of the analyses are presented and a TURBINE TRIP FROM 100*4 POWER IN THE VANDELLOS 11 comparison with the expenmental data is provided.

NUCLEAR POWER PLANT.

LLOPIS.C -

PEREZ,J[-

NUREG/lA-0113: PRELIMINARY ASSESSMENT OF PWR MENDl4ABAL R.

Spain. Govt.

of.

June 1993.

58p 9306210328. ICSP-V2-R100-R. 75403.325.

STEAM GENERATOR MODELLING IN RELAPS/ MOD 3.

An assessment of RELAPS/ MOD 2 cycle 36.04 against a tur-PREECE R.J.; PUTNEY,J.M. National Power (United Kingdom).

bine trip from 100% power in Vandellos 11 NPP (Spain) is pre.

July 1993. 26pp. 9309090021. 76381:154.

sented. The work is insenbed in the framework of the Spanish A preliminary assessment of Steam Generator (SG) modelling contnbution to ICAP Project. The model used in the simulation in the PWR thermal-hydraulic code RELAPS/ MOD 3 is present-consists of a single loop, a steam generator and a steam line ed. The study is based on calculations against a series of up to the steam header all of them enlarged on a scale of 3:1; steady-state commissioning tests carried out on the Woff Creek and full scaled reactor vessel and pressunzer. The results of the PWR over a range of load conditions. Data from the tests are calculations have been in reasonable agreement with plant used to assess the modelling of primary to secondary side heat measurements. An additional study has been performed, to transfer and, in particular, to examine the effect of reverting to check the ability of a model in which all the plant components the standard form of the Chen heat transfer correlation in place are full-scaled to reproduce the transient. A second study has of the modified form applied in RELAP5/ MOD 2. Comparisons between the two versions of the code are also used to show been performed using the Homogeneous Equilibnum Model in the pressunzer trying to elucedate the influence of the velocity how the new interphase drag model in RELAP5/ MOD 3 affects slip in the pnmary depressurization rate.

the calculation of SG liquid inventory and tho void fraction pro-file in the riser.

NUREG/lA-0109: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10% LOAD REJECTION TRANS!ENT FROM 75% STEADY NUREG/'A-0116: ASSESSMENT OF RELAP5/ MOD 3/V5MS STATE IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

AGAINST THE UPTF TEST NUMBER 11 (COUNTERCURRENT LLOPIS.C.; PEREZ,J.; CASALS,A.; et al. Spain, Govt. of. May FLOW IN PWR HOT LEG). CURCA TIVIG,F. Siemens AG -

1993. 59pp. 9306210175. UNID-91-08. 75401:065.

KWU Group (formerly Siemens AG - Bereich Energieerzeugung The Consejo de Segundad Nuclear (CSN) and the Asociacion (KWU)). May 1993.100pp. 9306210180. KWU E412/91/E10.

Nuclear Vandellos (ANV) have developed a model of Vandellos 75401:122.

11 Nuclear Power Plant. The ANV collaboration consisted in the Analysis of the UPTF Test No.11 using the "best-estimate" supply of design and actual data, the cooperation in the simula-computer code RELAPS/ MOD 3/ Version Sm5 is presented. Test tson of the control systems and other model components, as No.11 was a quasi-steady state, separate effect test designed well as in the results analysis. The obtained moval has been as.

to investigate the conditions for countorcurrent flow of steam sessed against the following transients occurring in the plant: a and saturated water in the hot leg of a PWR. An unphysical trip from the 100% power level (CSN); a load rejection from result was received using a CCFL correlation of the Wallis type 100% to 50% (CSN), a load rejection from 75% to 65% (ANV),

with the intercept C = 0.644 and the slope m = 0.8. The un-and a feedwater turbopump trip (ANV). This copy is a report of Physical prediction is an indication of possible programming the load rejection from 75% to 65% transient simulation. This errors in the CCFL model of the RELAP5/ MOD 3/V5m5 comput-transient was one of the tests carried out in Vandellos 11 NPP er code.

dunng the startup tests.

NUREG/lA-0118: ANALYSIS OF LOFT TEST L5-1 USING NUREG/lA-0110: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A RELAP5/ MOD 2. COOPER,S. United Kingdom. May 1993.43pp.

MAIN FEEDWATER TURBOPUMP TRIP TRANSIENT IN THE 9306210192. TD/SPB/ REP /0130. 75401:234.

VANDELLOS 11 NUCLEAR POWER PLANT, LLOPIS,C.;

The RELAPS/ MOD 2 code. Reference 1, is being used by No-PEREZ,J.; CASALS.A.; et al. Spain, Govt. of. December 1993.

clear Electnc for the calculation of Small Break Loss of Coolant 54pp. 9401060229. ICAP-00219. 77686:306.

Accidents (SBLOCA) and pressurized transient sequences in The Conscio de Segundad Nuclear (CSN) and the Asociacion the Sizewell "B" PWR. To validate the code for this purpose, it Nuclear Vandellos (ANV) have developed a model of Vandellos eas been used to model expenments of this type of transient

50 Maln Citations and Abstracts carried out in various integral test facilities. A number of those Garona Nuclear Powor Plant using TRAC-BF1 code is present.

studios have been for expenments carried out in the LOFT ex-ed Reasonablo and realistic adjustments have been made in penmontal reactor, Reference 2, and are desenbod in Refer-the model to improve its performance. This work is part of the ences 3,4,5,6, and 7. To assist in assessing the capability of validation sot for the TRAC model that is being developod for RELAPS/ MOD 2, the LOFT test L51 has been selected for wider use and to test the code capabilities. As a result of the analysis. This test was designed to simulate the rupture of a analysis, st is felt that TRAC-BF1 is capable of reproducing the single 14 inch diameter accumulator injection line in a commor-plant behavior with an acceptable degree of accuracy although cial PWR, equivalent to a 25% area bteak in the broken loop bottor models are clearly needed, in addition

%cr noding cold log Early in the transient the pumps were tripped and the work and code improvements The code took almost 14000 HPIS ingoction initiated, towards the end of the transient, accu-sec. which makes a 1/230 calculation time to real timo ratio, mulator and LPIS injection began. It should be noted that for For this transient a mechanistic separator model is noodod. It Sizowell "B" analyses a 25% break is classified as largo.

will also help to cut down running costs if the vessol noding whereas in this report, as in the extemal literature, this break could have a different number of cells at different heights.

size is referred to as intermediate Though not very important for this transient, the entical flow NUREG/lA-0119: ASSESSMENT AND APPLICATION OF BLACK, model will allow for realistic RV flow assumptions. There are not OUT TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH guidelines available for separator modelling in transionts, it has RELAP5/ MOD 2 REVENTOS,F4 BAPTISTA.J.S.; NAVAS.A P.;

boon found that a detailed noding in the separator region may et al. Spain, Govt. of. June 1993. 63pp. 9306290137. ICSP-AS-be needed to represent stoam-water interaction.

BOUT R. 75497;140' cioar Asco has prepared a model of Asc The Asociacion Nu NUREG/lA-0123: APPLICATION OF FULL POWER BLACKOUT FOR C.N.

ALMARAZ WITH RELAPS/ MOD 2. LECHAS.A L NPP using RELAPS/ MOD 2. This model, which include thermal-hydraulics, kinetics and protection and controls, has been quali-S ain, Govt. of. June 1993. 100pp. 9306290128. ICSP-AL-80 fied in stevious calculations of severa' actual plant transients.

The first part of the transient presented in this report is an roup of Almaraz Nuclear Power Plant has de-actual black-out and one of the transionts of the quahfication velo ed a model of the plant with RELAP5/ MOD 2/36.04. This model is the result of the work experience on the code process. The results are in agreement with plant data. The second par) of the transient is a hypothetical case. It consists in RELAPS/ MODI that was the standard code donng the penod ro-starting a pnmary pump and assume a new black out. The 1984/1989. Different solutions were adopted in the network to adoquate the model to RELAPS/ MOD 2 Computer Code. This transient was selected for ICAP because it presents an experi-r m he on a sa o nt o w

ence with the same transient calculated with RELAP5/ MOD 1/

NUREG/lA-0120: ASSESSMENT OF THE TURBINE TRIP TRAN-CY 29 Computer Code. The companson between both analysis SIENT IN COFRENTES NPP WITH TRAC-BF1, CASTRILLO,F.

will be interesting.

Hidroelectrica Espanota GOMEZ,Aa GALLEGO,l.; et al. Union Iberoamencana Do Tecnologia June 1993. 74pp. 9306290147.

NUREG/lA-0124: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A EST SIAN-22. 75497:203.

PRESSURIZER SPRAY VALVE INADVERTED FULLY OPEN-This report presents the results of the assessment of TRAC-ING TRANSIENT AND RECOVERY BY NATURAL CIRCULA.

BF1 (GI-J1) code with the model of C. N Cofrontes for simula-TION IN JOSE CABRERA NUCLEAR STATION ARROYO.R.;

(

tion of the transient originated by the manual tnp of the rnain REBOLLO.L Union Electrica Fonosa, S.A. June 1993. 126pp.

l turbine. C. N. Cofrentes is a General Electric designed BWR/6 9307060121, ICSP-JC-SPR-R. 75572:049.

t olant, with a nominal core thermal power of 2804 Mwt, in com-This document presents the companson between the simula-i mercial operation since 1985, owned and operated by Hidroo-tion results and the plant measurements of a real event that lectnca Espanola, S. A. The plant incorporates all the characle'-

took place in Jose Cabrera nuclear power plant in August 30, istics of BWR/6 reactors, with two turbine dnven FW pumps. As 1984. The event was onginated by the local, continuous and in-a result of this assessment a model of C. N. Cofrentes has adverted opening of the pressurizer spray valve PCV-400A.

been developed for TRAC-BF1 that fairly reproduces operation-Jose Cabrera power plant is a single loop Westinghouse PWR at transient behavior of the plant. A special purpose codo was belonging to UNION ELECTRICA FENOSA, S.A. (UNION generated to obtain reactivity coefficients, as required by TRAC-FENOSA), a Spanish utility which participates in the Internation.

BF1, from the 3D simulator.

al Code Assessment and Applications Program (ICAP) as a -

NUREG/lA-0121: ASSESSMENT OF A PRESSURIZER SPRAY member of UNIDAD ELECTRICA, S.A. (UNESA) This is the VALVE FAULTY OPENING TRANSIENT AT ASCO NUCLEAR second of its two contnbutions to the Program: The first one POWER PLANT WITH RELAPS/ MOD 2.

REVENTOS F.;

was an application case and this is an assessment one. The BAPTISTA.J S ; NAVAS,A P.; et al. Spain, Govt. of. December simulation has been performed using the RELAPS/ MOD 2 cycle 1993 58pp.9401140031. ICSP-AS-SPR-R. 77797:080.

36.04 codo, running on a CDC CYBER 180/830 computer The Association Nuclear Asco has prepared a model of Asco under NOS 2.5. operating system. The main phenomena have Nuclear Power Plant using RELAP5/ MOD 2. This model, which been calculated correctly and some conclusions about the 3D includes thermalhydraulics, kinetics and protection and controls, charactonstics of the condensation due to the spray and its sim-has been qualified in previous calculations of several actual ulation with a ID tool have been reached.

plant transients. One of the transients of the qualification proc-NUREG/lA-0125: ASSESSMENT OF RELAPS/ MOD 2 COMPUT-ess is a " pressurized spray valve faulty opening" presented in ER CODE AGAINST THE NATURAL CIRCULATION TEST this report. It consists in a primary coolant depressunration that DATA FROM YONG-GWANG UNIT 2. ARNE.Na CHO,S. Korea causes the reactor trip by overtemperature and later on the ac-tuation of the safety injection. The results are in close agree-Electnc Power Corp. KIM,H-J. Korea institute of Nuclear Safety; June 1993.110pp. 9306290132. 75497.032.

ment with plant data' The results of the RELAPS/ MOD 2 computer code simulation NUREG/lA 0122: ASSESSMENT OF MSIV FULL CLOSURE FOR for the Natural Circulation Test in Yong-Gwang Unit 2 are ana-SANTA MARIA DE GARONA NUCLEAR POWER PLANT lyzed here and compared with the plant operation data. The USING TRAC-BF1 (G1JI) ' CRESPO.) L.; FERNANDEZ,R.A.

result of companson reveals that the code calculation does Cantabna, Univ. of, Spain June 1993. 46pp. 9307060112. ICSP-present well the overall macroscopic behaviors of thermalhy-GA MSIV T. 75572 167, draube parameters in primary and secondary system compared An assessment of the first 60 seconds of a spunous Main with the plant operating data. The sensitivity study is performed l

Steam Insolation Valve (MSIV's) closure for Santa Maria to find out the effect of steam dump flow rate on the primary

l Main Citations and Abstracts 51 temperatures and it is found that the primary temperatures are (LBLOCA), and dunng selected small-break LOCA (SBLOCA) very sensitue to the steam dump flow rate dunng the Natural transients. The program included tests at the Cylindrical Core Circulation. Because of the inherent uncertainties in the plant Test Facility (CCTF), the Slab Core Test Facility (SCTF), and data, the assessment work is focussed on phenomena whereby the Upper Plenum Test Facility (UPTF), and computer analyses the companson between plant data and calculated data is using TRAC. Tests at CCTF investigated core thermal-hydrau-based more on trends than on absolute values.

lies and overall system behavior while tests at SCTF concen-trated on multidimensional core thermal-hydraulics. The UPTF NUREG/lA-0126: -2D/3D PROGRAM WORK

SUMMARY

tests investigated two-phase flow behavior in the downcomer, REPORT. DAMERELL.P.S.; SIMONS.J.W. MPR Associates, Inc.

upper plenum, tie plate region, and pnmary loops. TRAC anaty-

• et al. Japan Atomic Energy Research Institute. June 1993-400pp. 9307220220. GRS 100. 75745

  042-ses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the The 2D/3D Program was carried out by Germany, Japan and test and analysis results in each of the main areas where im-the United States to investigate the thermal-hydraulics of a proved information was obtained in the 2D/3D Program. The PWR large-break LOCA. A contributory approach was utihred in discussion is organized in terms of the reactor safety issues in-which each country contnbuted significant effort to the program vestigated.

and all'three countnes shared the research results. Germany 1

constructed and operated the Upper Plenum Test Facility NUREG/lA-0128: INTERNATIONAL CODE ASSESSMENT AND (UPTF), and Japan constructed and operated the Cyhndrical APPLICATIONS PROGRAM:

SUMMARY

OF CODE ASSESS-Core Test Facihty (CCTF) and the Slab Core Test Facihty MENT STUDIES CONCERNING RELAPS/ MOD 2, RELAPS/

(SCTF), The US contribution consisted of provision of advanced MOD 3, AND TRAC-B. SCHULTZ,R.R. EG&G Idaho. Inc. De-instrumentation to each of the three test facilities, and assess, cember 1993.

268pp.

9401060241. EGG-EAST 8719.

{

ment of the TRAC computer code against the test results. Eval-77686.001.

J uations of the test results were carried out in all three countries.

Members of the International Code Assessment Program This report summanzes the 2D/3D Program in terms of the con-(ICAP) have assessed the U. S. Nuclear Regulatory Commission inbuting efforts of the participants.

(USNRC) advanced thermal-. hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define i

NUREG/lA-0127: REACTOR SAFETY ISSUES RESOLVED BY user guidehnes, and to determine the state of each code. The THE 2D/3D PROGRAM. DAMERELL,P.S ; SIMONS.Jav. MPR results of sixty-two code assessment reviews, conducted at Associates, Inc.

• Japan Atomic Energy Research institute. July INEL, are summarized. Code deficiencies are discussed and 1993. 400pp. 9308160173. GRS-101. 76116.001.

user recommended nodalizations investigated dunng the course The 2D/3D Program studied multidimensional therma!-hydrau-of conducting the assessment studies and reviews are listed. All bcs in a PWR core and primary system during the end-of-blow-the work that is summarized was done using the RELAP5/

down and post blowdown phases of a large-break LOCA MOD 2, RELAPS/ MOD 3, and TRAC-8 codes.

l l

--.. ~

u6.Jun.A

&=

4

.a~.4 u.-a

_J*s a

a s

A 1

I I

I i

l l

l l

l 1

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

SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER AEA-TRS-1050 NUREG/lA 0096 EGG 2697 NUREG/CR4027 AEEW-R2501 NUREG/lA4096 EGG-2701 NUREG/CR 6070 AEOD/S92 07 NUREG 1275 V09 EGG-2706 NUREG/CR4073 ANL 91/43 NUREG/CR 5822 EGG-EAST-8719 NUREG/lA-0128 ANL 92/42 NUREG/CR 4744 V07 N1 EPHI TR 102106 NUREG/CR 6018 ANL-93/11 NUREG/CR-4744 V07 N2 E ST.SIAN-22 NUREG/lA 0120 ANL 93/19 NUREG/CR-6080 GRS-100 NUREG/lA 0126 ANL 93/2 NUREGICR-4667 V15 GRS.101 NUREG/lA 0127 ANL 93/27 NUREG/CR 4667 V16 ICAP-00219 NUREG/iA 0110 ANL 93/3 NUREG/CR 5999 ICSP AL-BOUT R NUREG/lA 0123 ANL-93/9 NUREG/CR-6032 iCSP AS-BOUT R NUREG/lA-0119 BHARC700/93/029 NUREG/CR4758 V03 tCSP-AS-SPR-R NUREG/lA-0121 BMl-2173 NUREG/CR-4599 V02 N2 ICSP.GA-MSIV T NUREQ/lA 0122 BMb2173 NUREG/CR4599 V03 N1 ICSP-JC-SPR R NUREG/lA-0124 BML2175 NUREG/CR-6098 ICSP TR-TTRIP-R NUREG/lA 0085 DNL-NUREG-51581 NUREG/CR-2907 VII ICSP-V2-R100-R NUREG/lA 0108 BNL-NUREG-51708 NUREG/CR-3469 V07 ICSP-V2-R50-R NUREG/tA-0107 BNL NUREG-52289 NUREG/CR-5747 IS4878 NUREG/CR-4273 BNL-NUREG-52299 NUREG/CR-5783 IS-5083 NUREG/CR-5957 BNL-NUREG 52318 NUREG/CR4844 ITRI-t 41 NUREG/CR-4214 R2 PT1 DNL-NUREG-52330 NUREG/CR-5883 K/NSP 117 NUREG/CR4 t 18 BNL-NUREG42334 NUREG/CR-5911 KWU E412/91/E10 NUREG/tA 0116 BNL-NUREG 52343 NUREG/CR4933 LA-12181-MS NUREG/CR 5755 BNL-NUREG 52344 NUREG/CR-5934 LA 12201 MS NUREG/CR-5776 BNL-NUREG 52346 NUREG/CR-5943 LA-12593-M NUREG/CR-6060 DNL-NUREG-52352 NUREG/CR-5959 LMF-136 NUREG/CR-4214 R1P2A2 BNL-NUREG-52354 NUREG/CR-5982 MEA-2490 NUREG/CR-5926 DNL-NUREG-52355 NUREG/CR-5983 MPR-1345 NUREG/lA 0126 DNL NUREG42356 NUREG/CR-5984 MPR-1346 NUREG/IA-0127 BNL NUREG-52362 NUREG/CR-5993 VOI NEA/CNRA/R(92)3 NUREG/CP 0128 BNL NUREG-52362 NUREG/CR 5993 V02 ORNL-6566/V1 NUREG/CR-5404 V02 BNL NOREG-52364 NUREG/CR 5995 ORNL-6666 NUREG/CR-5699 V01 DNL NUREG42367 NUREG/CR-6014 ORNL-6731 NUREG/CR4938 BNL-NUREG-52370 NUREG/CR-6022 ORNL4734 NUREG/CR4944 BNL-NUREG-52375 NUREG/CR4041 ORNL4740.

NUREG/CR 5914 DNL-NUREG 52377 NUREG/CR-6049 ORNL/TM-11568 NUREG/CR-5591 V01 N2 BNL NUREG 52383 NUREG/CR-6078 ORNL/TM 11572 NUREG/CR 5358 BNL NUREG-52393 NUREG/CR 6111 ORNL/TM 11876 NUREG/CR-5754 BNL/NUREG42029 NUREG/CR-4551V7 RIP 2D ORNL/TM-11945 NUREG/CR 5782 DNL/NUREG $2029 NUREG/CR4551 V7 RIP 1 ORNL/TM-12179 NUREG/CR-5922 BNL/NUREG-52029 NUREG/CR-4551V7R1P2A ORNL/TM-12229 NUREG/CR-5942 CONSWCSMECR1692 NUREG/CR-5958 ORNL/TM-12247 NUREG/CR-5952 CDNSWCSMECR1792 NUREG/CR 5969 ORNL/TM 12266 NUREG/CR4955 CONSWCSMECR1892 NUREG/CR 5970 ORNL/TM-12279 NUREG/CR 5968 CEERC-91 105 NUREG/CR-5956 ORNL/TM-12200 NUREG/CR-5972 CNWRA 9101 A NUREG/CR-5817 V02 ORNL/TM-12307 NUREG/CR-5997 CNWRA 92-003 NUREG/CR-5991 ORNL/TM-12340 NUREG/CR-6023 CNWRA 92-006 NUREG/CR 6026 OPNL/TM-12342 NUREG/CR-6015 CNWRA 92 01S NUREG/CR-5817 V03 N1 ORNL/TM-12349 NUREG/CR-6036 CNWRA 92-026 NUREG/CR-6028 ORNL/TM 12371 NUREG/CR-6048 CNWRA 92-02S NUREG/CR-5817 V03 N2 ORNL/TM-12396 NUREG/CR4065 CNWRA91-010 NUREG/CR 5917 V02 ORNL/TM 12406 NUREG/CR-6071 CNWRA91010 NUREG/CR4917 V01 ORNL/TM-12413 NUREG/CP 0131 CNWRA92-011 NUREG/CR-6021 ORNL/TM 12484 NUREG/CR4117 CONF-9020823 NUREG/CP 0130 V01 ORNL/TM 9593 NUREG/CR4219 V09 N2 CONF-9020823 NUREG/CP 0130 V02 ORNLSUD93-SD084 NUREG/CR-6052 CRTD VOL.20-2 NUREG/GR 0005 V02 P1 PNL-4221 NUREG/CR-2850 VII ECN49 91 NUREG/V.009I PNL-5210 NUREG/CR-3950 V08 ECN C-92-008 NUREGi A-0112 PNL-5711 NUREG/CR 4469 V15 EGG-2577 NUREG/CR 5229 V05 PNL-5711 NUREG/CR 4469 V16 EGG-2610 NUREG/CR 6061 PNL 7187 NUREG/CR-5488 EGO 2618 NUREGICR 5642 PNL 7809 NUREGICR-5766 EGG-2635 NUREG/CR 5672 V03 PNL 7745 NUREG/CR-5631 R1 ADD EGG-2640 NUREG/CR-5759 PNL-7905 NUREG/CR4829 E GG-2665 NUREG/CR4818 PNL 7906 NUREG/CR 5834 EGG-2677 NUREG/CR 5882 PNL 7907 NUREG/CR-5833 EGG-2685 NUREG/CR 5928 PNL-7908 NUREG/CR4836 IGG-2688 NUREG/CR-5937 PNL 7925 NUREG/CR 5835 E GG-2689 NUREG/CR-5949 PNL 8104 NUREG/CR-5897 EGG-2690 NUREG/CR-5953 PNL-8105 NUREG/CR-5898 EGO-2692 NUREG/CR-5964 PNL-8106 NUREG/CR-5894 EGG-2694 NUREG/CR-5978 PNL-8454 NUREG/CPA247 V01 R1 E GG-2695 NUREG/CR4987 PNL-8462 NUREG/CR-5973 53

54 Secondary Report Number Index SECONDARY REPORT NUMBER REFORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER PNL f 466 NUREG/CR-5975 SAND 92 0538 NURE G/CR 4663 PNL -8467 NUREG/CR 5989 S AND92-1422 NUREG/CR 5901 PNL 8430 NUREG/CR 59Ao SAND 92-1563 NURFG/CR-5907 S A ND92-1721 NUREG/CR-5961 PNL 8496 NURf G/CR-5998 S AND92-2109 NUREG/CR 5936 PNL 8497 NUREG/CR 6054 DRr rc SAND 92 2146 NUREG/CR 6059 PNL-8499 NURE G/CR 5996 5 A ND92-268H NUREG/CR 5978 PNL 8577 NUREG,,CR 54 t o SAND 92 2689 NURE G/CR-5966 PNL-8594 NUREG/CR 6029 V01 SAND 92 2 765 NUREG/CR 5305 V02 P1 PNL 8614 NURE G/CR 6043 VD1 SAND 92 2765 NUREG/CR-5305 V02 P2 PNL 8088 NUREG/CR 5758 V03 SI 92-101 NUREG/CR-5455 V03 PNL B742 NUREG/CR 5664 V2 DRF SI-92101 NUREG/CR-5455 VOI PNL 8742 NURE G/CR 5884 V1 UHF SI 101 NUREG/CR 5455 V02 PNL 8750 NUR E G / CR-6084 ST UDSvikN590/ 93 NURE G/lA-0094 PNL 8842 NUREG/CR 6114 V0 t TD/ SPD/ RE P/0130 NUR E G/l A-0118 RRCM 80 05/3 NUREG /CH-60 72 T E C / t / 0471/ R91 NUREG/lA 0106 UCID-2124 5 NUREG/CR-4832 VOS SAIC 91/6660 NURE G/CR-6018 UCRL-lD 110637 NUREG/CR 6007 SAIC 93/1310 01 NUREG/CR 6050 UCRLJD-112900 NUREG/CR-6090 SAND 89 0943 NUREG/CR 5360 UCRL lD-114565 NUREG/CFM083 SAND 89 2562 NURE G /CR-54 71 UCRL 1D-114567 NUREG/CR 6082

• g SAND 91-1536 NURL-G/CR 5791 UiL U-E NG01 2013 NUREG/CR-5969 SAND 91 -2802 NURf G/CR 5927 V01 Ult U E NG92 2014 NUREG/CR 5971 SAND 91 70H 1 NUREG/CR 5801 U1LU ENG92 2016 NOREG/CR-5970 SAND 92-0167 NUREG/CR 5643 UtLU-E NG92-400 7 NUREG/CR 5977 SAND 92 053 7 NUREG/C;R 4832 VOS UNLD-9108 NUREG/1 A-0109 SAND 92 0537 NUREG/CR 4832 V09 VARGOS 93/1 NUREG/CFM072 l

\\

l i

l

Personal Author Index.

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

APOSTOLAKIS,G.

NUREG/CR.5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREG/CR-6058. A FRAMEWORK FOR THE ASSE SSMENT OF SEARCH AT CNWRA Calendar Year 1991.

SEVERE ACCIDENT MANAGEMENT STRATEGIES NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. January June 1992.

ARMBRUSTER.J.

NUREG/CR-5817 V03 N2: NRC HIGH LEVEL RADIOACTIVE WASTE NUREG/CR 5778 V03. NEW YORK /NEW JERSEY REGIONAL SEISMIC RESEARCH AT CNWRA. July December 1992.

NETWORK. Final Report For Apnl 1985 Septemtwr 1992.

NUREG/CR 6028: BIGFLOW-A NUMERICAL CODE FOR SIMULATING FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC ARNE,N.

MEDIA. Theory And User's Manual - Version 1.1.

NUREG/lA-0092: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE AGAINST THE NET LOAD TRIP TEST DATA FROM YONG-ABRAHAMSON.S.

GWANG. UNIT 2.

NUREG/CR-4214 RIP 2A2: HEALTH EFFECTS MODELS FOR NUCLE.

NUREG/lA-Of 00: ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 AR POWER PLANT ACCIDENT CONSEQUENCE AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS.

ANALYSIS Modification Of Models Result #ng From Addihon Of Effects NUREG/lA 0125-ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE Of Exposure to Alpha-Emitting Radionuchdes Part II: Scientshc Dases AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG.

For Health....

GWANG UNIT 2.

NUREG/CR 4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part 1:

ARROYO.R Introduchon.Integranon,And Summary-NUREG/lA 0124. ASSESSMENT OF RELAP5/ MOD 2 AGAINST A PRES.

SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN.

ABU-EIO,R.

SIENT AND RECOVERY DY NATURAL CIRCULATION IN JOSE CA.

NUREG-1476-FINAL ENVIRONMENTAL IMPACT STATEME.NT TO BRERA NUCLEAR STATION.

CONSTRUCT AND OPERATE A FACILITY TO RECEIV. STORE, AND DISPOSE OF 11E-(2)

BYPRODUCT MATERIAL NEAR ARSE N AULT,F, CLIVE. UTAH Docket No. 40-8989. Enynocare Of Utah.Inc.

NUREG/CP-0129 PROCEEDINGS OF THE WORKSHOP ON PROGRAM NUREG 1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT FOR ELIMINATION OF REQUIREMENTS M ARGINAL TO SAFETY.

TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE-ASGARl,M.

AND DISPOSE OF 11E42) BYPRODUCT MATERIAL NEAR CtlVE.

NUREG/CR-6071 IMPACT OF ENDF/B-VI CROSS-SECTION DATA ON UTAH Docket No, 40-8989, Envirocare Of litah. Inc-H B ROBtNSON CYCLE 9 DOSIMETRY CALCULATIONS.

AHOLA M.P.

ATHEY,G.F.

NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/CR-5247 VOI RI: RASCAL VER$lON 2.0 USER'S GUIDE.

RESEARCH AT CNWRA January-June 1992-NUREG/CR-5247 V02. RASCAL VERSION 2 0 WORKBOOK.

NUREG/CR-5817 V03 N2. NRC HIGH LEVEL RADIOACTIVE WASTE hESEARCH AT CNWRA. July December 1992.

ATWOOD,C.L NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL-NUREG/CR.5964:

SAPHIRE TECHNICAL REFERENCE I

HYDROLOGIC-MECHAN ICAL -CHEMICAL PROCESSES PERTINENT MANUALIRRAS/ SARA VERSION 4 0.

I TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA

)

MOUNTAIN AUSTIN,J.H.

i NUREG-1444 SITE DECOMMISSIONING MANAGEMENT PLAN.

l AK E R S,D.W.

NUREG/CR 5672 V03: CHARACTER:STICS OF LOW 4EVEL RADIOAC-AZARM,M.A.

TlVE DECONTAMINATION WASTE Annual Report For Fiscal Year NUREG/CR-5993 VOI: METHODS FOR DEPENDENCY ESTIMATION 1992.

AND SYSTEM UNAVAILABluTY EVALUATION BASED ON FAILURE DATA STATISTICS. Summary Report.

ALLENSPACH,F.

NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATION NUREG-1214 R11: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAlu.lRE IC ASSESSMENT OF LICENSEE PERFORMANCE.

DATA STATISTICS Detailed Descnption And Applicatons.

NUREG 1214 R12. HISTORICAL DATA

SUMMARY

OF THE SiOT EIVAT.

IC ASSESSMENT OF LICENSEE PERFORMANCE.

B AC A,R.G.

NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE ALLISON,C.M.

RESEARCH AT CNWRA January. June 1992.

NUREG/CR-5642 LIGHT WATER REACTOR LOWER HEAD FAILURE NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE ANALYSIS.

RESEARCH AT CNWRA July-December 1992 ALVIS.J.A.

BAGTZOGLOU.A.C.

NUREG/CR 3950 V08. FUEL PERFORMANCE ANNUAL REPORT FOR NUREG/CR 5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE.

1990 SE ARCH AT CNWRA Calendar Year 1991.

NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE ANAND,N.K.

RESEARCH AT CNWRA July-December 1992.

I NUREG/GR-0006 DEPOSITION SOFTWARE TO CALCULATE PARTI-NUREG/CR 6028: BIGFLOW: A NUMERICAL CODE FOR SIMULATING CLE PENETRATION THROUGH AEROSOL TRANSPORT FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC SYSTE MS Final Report.

MEDIA. Theory And User's Manual-Version 1.1.

ANDERSON,T.L B AKE R.D.A.

NUREG/CR-5971. CONTINUUM AND MICROMECHANICS TREATMENT NUREG/CR-2850 VII: DOSE COMMITMENTS DUE TO RADIOACTIVE OF CONSTRAINT IN FRACTURE.

RELEASES FROM NUCLEAR POWER PLANT SITES IN 1989 55

1 1

56 Personal Author Index BAKER.L BEZLER P.

NUREG/CR.6032: SOLIDUS AND LIQUIDUS TEMPERATURES OF NUREG/CR.6049: PlPING BENCHMARK PROBLEMS FOR THE GEN-CORE. CONCRETE MIXTURES.

ERAL ELECTRIC ADVANCED BOILING WATER REACTOR l

BAKER,WL BEZMELNITSIN,A.

i NUREG/CR.5155 STIFFNEST OF LOW ASPECT RATIO, REINFORCED NUREG/CR 6072 EXPERIMENTAL STUDY ON THE COMBUSTION BE-j CONCRETE SHEAR WALLS.

HAVIOR OF HYDROGEN-AIR MIXTURES WITH TUR3ULENT JET IG-1 NUREG/CR.5776 DAMPING IN LOW-ASPECT-RATIO. REINFORCED NITION AT LARGE SCALE.

CONCRETE SHEAR WALLS.

l DICKEL,J.H.

BANDYOPADHYAY,K NUREG/CR 5759: RISK ANALYSIS OF HIGHLY COMDUSTlBLE GAS j

NUREG/CR-6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN STORAGE, SUPPLY. AND DISTRIBUTION SYSTEMS IN PRESSUR-j THE HIGH LEVEL VIBRATION TEST AT TADOTSU IZED WATER REACTOR PLANTS.

i BAPTISTA.J.S.

DIERSCHBACH,M.

NUREG/lA-0119 ASSESSMENT AND APPLICATION OF BLACKOUT NUREG/CR-5884 VI DAF: REVISED ANALYSES OF DECOMMISSION-i TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

ING FOR THE REFERENCE PRESSURIZED WATER REACTOR R

a ent Wam W h bs&o j

N RE /IA0121: ASSESSMENT OF A PRESSURIZER SPRAY VALVE F AULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT

~

C t

WITH RELAP5/ MOD 2 NUREG/CR-5884 V2 DRF: REVISED ANALYSES OF DECOMMISSION-ING FOR THE REFERENCE PRESSURIZED WATER REACTOR BARCHI T.J POWER STATION Effects Of Current Regulatory And Other Consider-NUREG-1415 V06 No t:

OFFICE OF THE INSPECTOR atons On W Fmal Ansance,~ApnkesM Re# 6 GENERAL Semiannual Report. April 1,1993 September 30,1993 Comment NUREG/CR 6054 DAF FC: ESTIMATING PRESSURIZED WATER RE-B ASS,0.R.

ACTOR DECOMMISSIONING COSTS.A Users Manual For The PWR NUREG/CP-0131: PROCEEDINGS OF THE JOINT lAEA/CSNI SPECIAL.

ISTS' MEETING ON FRACTURE MECHANICS VERIFICATION BY Cost Est rnating Computer Program (CECP) Software. Draft Report For

' LARGE-SCALE TESTING Hold At Pollard Auditonum. Oak Comment Anige,Tenneswe.

NUREG/CR-5782: PRESSURIZED THERMAL SHOCK Pf00ADILISTIC EISHOP,J.V.

FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CR-6047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF

• /ANADOUS ANU VANADIC IONS.

ROWE REACTOR PRESSURE VESSEL NUREG/CR.5997: CSNI PROJECT FOR FRACTURE ANALYSES OF E

L TERNATIONAL REFERENCE EXPERIMENTS

/ 6043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC NUREG/CR.6036 INITIAL RESULTS OF THE INFLUENCE OF BIAX1AL CHILLERS USED IN NUCLEAR POWER PLANTS Phase 1.

LOADING ON FRACTURE TOUGHNESS.

BLANDFORD.R K-B AU M.J.W.

NUREG/CR-6027: PRELIMINARY EVALUATION OF SNUBBER SINGLE NUREG/CR.3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU.

FAILURES.

CLEAR POWER PLANTS ANNOTATED DIBLIOGRAPHY OF set.ECT.

DLOSE. R ED READINGS IN RADIATION PROTECTION AND ALARA NUREG/CR W07. CORE CONCRETE INTERACTIONS WITH OVERLY-I NUREG/CR-5883s HEALTH RISK ASSESSMENT OF 1RRADIATED ING WATI H POOLS The WETCOR-1 Test I

TOPAZ, l

BEA VE N,P.A.

DLUHM.D, j

NUREG/CR 5928 SANS INVESTIGATION OF LOW ALLOY STEELS IN NUREGICR-4273. CRACK PROPAGATION IN HIGH STRAIN REGIONS NEUTRON 1RRADIATED. ANNEALED, AND REIRRADIATED CONDI.

OF SEQUOYAH CONTAINMENT.

NUREG/CR 5957: SYSTEM B0 + (TM) CONTAINMENT - STRUCTURAL TlONS.

DESIGN REVIEW.

DLCKJL I

NUREG/CR-6012. STlFFNESS AND DAMPING PROPERTIES OF A BOECKER,0.D.

LOW ASPECT RATIO SHEAR WALL BUILDING DASED ON RECORD.

NUREG/CR-4214 RIP 2A2: HEALTH EFFECTS MODELS FOR NLK;LE.

AR POWER PLANT ACCIDENT CONSEQUENCE ED EARTHOUAKE RESPONSES ANALYSIS Modshcaton Of Models Resulting From Additon Of Effects BENDER,M.A.

Of Exposure To Alpha Emitting Radionuchdes Part II: Scientific Bases NUREG/CR-4214 RIP 2A2: HEAL TH EFFECTS MODELS FOR N'JCLE.

For Hoafth....

AR POWER PLAM ACCIDENT CONSEQUENCE NUREG/CR 4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR I

ANALYSIS.Modificaton Of Models Resulting From Additon Of Etiects POWER PLANT ACCIDENT CONSEOUENCE ANALYSIS Part 1.

Of Exposure To Alpha Ermtting Radonucl&s Part il Scecntife Bases introducion Integration.And Eummary.

For Ho#th..-

NUREG/CR-4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR BOHNHOFF,W.J.

POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part 1.

NUREG/CR 5936-ENHANCEMENTS TO THE ACCIDENT PRECURSOR METHODOLOGY.

Introduction.Intog ation,And Summary.

BERNREUTER,0L.

DOLLINGER,0.A.

NUREG/CR 4832 V08: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREGICR-6058-VIRGINIA REGIONAL SEISMIC NETWORK hnal l

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION Report (1986 1992).

PROGRAM (RMIEP) Smsrnic Analysis.

DERTA V.T.

NUREG/CR-5843; CORCON MOD 3:AN INTEGRATED COMPUTER NUREG/CR-6001: DETERMINATION OF THE DIAS IN LOFT FUEL MODEL FOR ANALYSIS OF MOLTEN CORE CONCRETE PEAK CLADDING TEMPERATURE DATA FROM THE BLOWOOWN INTERACTIONS. User's Manual.

PHASE OF LARCE-BREAK LOCA EXPERIMENTS.

DR AVE RM AN,J.

BE RTING.F.M.

NUREG/CR 6049: PIPING DENCHMARK PROBLEMS FOR THE GEN-NUREG/CR 3il50 VDa. FUEL PERFORMANCE ANNUAL REPORT FOR LRAL ELECTRIC ADVANCED DOILING WATER REACTOR.

1990 BROCK,R, BE Y E R.C.E.

NUREGICR-5977. A PERFORMANCE INDICATOR OF THE EFFECTIVE-NUREG/CR 3950 V08 FUEL PERFORMANCE ANNUAL REPORT FOR NESS OF HUMAN-MACHlfE INTERFACES FOR NUCLEAR POWER 1990 PLANTS.

l' l

l u ~-

l l

l Personal Author index 57 DROCKMANNJL BURSON.S B NUREG/CR SM3 COFtCON MOD 3 AN INTEGRATED COMPUT E R NUREG/CR Sie66 A SIMPilFIED MODI L OF AEROSOL RE MOVAL BY MODE L FOR ANALYSIS OF MOLT E N CORE -CONCRE1E CONTAINMENT SPRAYS INTE RACTIONS User's Manual NURE G/CR 5900 CORE CONCRETE INTERACTIONS WITH OVERLY.

CADY,R E.

ING WATE R POOLS The WETCOR 1 Test NUREG/CR 59M Soil CHARACILT112ATION METHODS f OR UN SATURATED LOW LEVEL WASTE SITES NUREG/CR 4832 V09 ANAL YSIS OF THE LASALLE UNIT 2 NUCLEAR C A L E,R.

POWER PLANT: RISK METHODS INTE GRATION AND EVALUATION NUREG/CR 6050 RADIATION E XPOSURE MONITORING AND INFOR-PROGRAM (RMIE P) internal Fwe AnaWs MATION 1RANSMIT T AL (RE MIT) SYSTE M User's Manual DROWN.C.

C A M P,A.L NUREG O430 Vi?.

LICE NSED FUEL FACillTY STATUS NUREG/CH-5936 ENHANCEMENTS TO THE ACCIDENT PRECURSOR Al: TORT inventory Ditteronce Data. July 1,1991. June 30.1992 (Gray METHODOLOGY Dook 11)

CAMPBE LL,P.

DROWN.T.D.

NUREG-1482 DRFT FC. GUIDELINES FOR INSERVICE TESTING AT NUREG/CR 5305 V02 Pt INTEGRATED RISK ASSESSMENT FOR THE NUCLEAR POWER PLANTS. Draft Report For Comment-LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenonwnology Ard RA Uncert*rity Evaluation Program (PRUEP) Appar*es A C CAMPDELL R D.

NUREG/CR 5305 V02 P2 INTEGRA1ED RISK ASSESSMENT FOR THE NUREG/CR 4832 VD0 ANALYSIS OF THE L ASALLE UNIT 2 NUCLEAR LASALLE UNIT 2 NUCLE AR POWER PLANT Phonokonology And POWER PLANT; RISK METHODS INTEGRATION AND EVALUATION Rsk Uncertainty Evalumbon F% gram (PRUEP) ApporKhce4 D G PROGRAM (RMIEP) Sesmec Ar,alyms DROWNSON.D.A.

C ANI ANO,R.J.

NURE G/CR 5937 INT E N TIONAL DE PRESSURIZA TION ACCIDI N T NUREG 1444 SITE DECOMMISSIONING MANAGEMENT PLAN MANAGEMINT STRATE GY FOR PRESSUni2ED WATER REAC.

CARBAJO,J.J.

NUREG/CR-5942 SEVERE ACCIDENT SOURCE TERM CHARACTER.

BRUM METT.E.

ISTICS FOR SELECTED PEACH UOTTOM SEQUENCES PHEDICTED NUREG-14 76. FINAL E NVIRONMENT AL IMPACT STATEMENT TO BY THE MELCOR CODE CONSTRUCT AND OPERATE A F ACILITY TO RECEIVE. STORE, AND C ARDILE.F.P DISPOSE OF t t E (2)

BYPRODUCT M A T ERIAL NEAR ggggg 34j4 SITE DECOMMISSIONING MANAGEMENT PLAN.

CLIVE.U T AH Docket No 404989. Envvocare Of Utah.Inc.

NUREG-1476 DRFT: DRAFT E NVIRONME NTAL IMPACT STATEMENT CASADA,D.A.

TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, NUREG/CR 5044 A CHARACTEnt2ATION OF CHECK VALVE DEGRA.

AND OtSPOSE OF 110 (2) DYPRODUCT MATERIAL NEAR CLIVE-DATION AND F AILURE DPERIENCE IN 1HE NUCLEAR POWER IN-UTAH Docket No. 40 8983 Envirocare Of Utah, loc-DUSTRY.

DRUST,F.

C AS AL9.A.

NUREGICR-4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING WFLDS Sornmnoual Report, October 1991. March 1992.

NUREG/lA4)t09 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10%

NUREG/CR 4599 V03 N1. SHORT CFIACKS IN PlPING AND PirlNG l.OAD REJECTION TRANSIENT FROM 75** STEADY STATE IN THE WELDS Som> annual Report, Apnl Septomter 1992 V ANDE LLOS il NUCt E AR POWI R PLANT.

NUREG/lA-0110. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A MAIN F f EDWATER TURBOPUMP ' RIP TRANSIENT IN THE VANDELLOS ll L E A 0096 NUMERICS AND IMPLEMENTATION OF THE UK HOR 110NTAL STRATIFICATION ENTRAINMf NT OF F.TAKE MODEL CASTRILLO,F.

IN10 REL AP5/ MOD 3 NUREG/lA-Ot20 ASSESSMENT OF THE TURBINE TRIP TRANSIENT BR YSON.J.W.

NUREG/CR 5782 PRESSURIZED THERMAL SHOCK PRORABillSTIC CASTRO,J.C.

f RACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CH 6025 THE PROBADILITY OF MARK.1 CONTAINMENT ROWE REACTOR PRESSURE VESSEL NUREG/CR 5968 POTENTIAL CHANGE IN FLAW GEOMETRY OF AN FAILURE Bf MELT. ATTACK OF THE LINER INtilALLY SHALLOW FINITE LENGTH SURF ACE FLAW DURING A C AVALLO.J D.

PRE SSURIZED. THERMAL-SHOCK TRANSfE NT NURE G/CR 6080. RE PLACE ME NT E NE RGY. CAPACITY, AND REll.

NUREGrCR 6036 INITIAL RESUI.1S OF THE INFLUENCE OF O! AXIAL ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT-LOADING ON F RACTURE TOUGHNESS-DOWNS BUEHRING.W. A.

CA2ZOLI.E.

NURF G/CR 6000 REPL ACEMENT ENERGY CAPACITY. AND REll-NUREG/CR.4551 V 7R IP l, EVAL.UATION OF SEVERE ACCIDE NT ABil.lTY COSTS F OR PE RMANENT NUCLE AR REACTOR SHUT-RISKS ZION UNIT i Main Report DOWNS NUREG/CR 4551V7 RIP 2A-EVALUATION OF SEVE RE ACCIDENT BULAND,R.

TU$KS ZION UNIT 1 Appendix A NUREG/CR 4551V7 RIP 20 EVALUATION OF SEVERE ACCIDENT NURf G/CR6085 UNITED ST ATES SE1SMOGR APHIC NETWORK RISKS ZION UNIT 1 AppendK.es B. C, D. An1 E.

BUL M AH N.K.D.

CHALLA.R.

NUREG/CR 5759. RISK ANAL YSis OF HIGHLY COMBUSTlHLE GAS NUREG/CR 5957: SYSTf M 80 + (TM) CONTAINMENT - STRUCTURAL STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRE SSUR-DESIGN REVIEW IIED WA1ER REACTOR PLANTS CHANIN.D.

DUMGARDNER.J.D.

NURE G/CR 6059 MACCS VERSION 1.5111: A MAINTENANCE RE-NUREG/CR 5766 AUXILIARY F ELDWATE R SYSTEM RISK-BASED IN-LEASE OF THE CODE SPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR POWI R PLANT.

CHANIN.D.L NUREG/CR 5830 AUXILLARY F EEDWATER SYSTEM RISK BASED IN.

NUREG/CR-5305 V02 Pl. INTE GRATE D RISK ASSESSMENT FOR THE SPECTION GUIDE FOR THE PALO VERDE NUCLE AR POWER LASALLE UNIT 2 NUCLEAR POWER PLANT.Phanomenology And PLAN T.

Rmk Urvstaenty Evaluabon Program (PRUEP) A spenrhces A C.

NUREG/CR 5N7 AUtlLIARY FEEDWATER SYSTEM RISK DASED IN-NUREG/CR 5305 V02 P2. INTEGRATED RISK A SESSMENT FOR THE SPECTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR LASALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And POWER PL ANT.

Ad Uncertenty Evaluation Program (PRUEP) Apperxbces D-G E

58 Personal Author index l

CHAPMAN.M.C.

NUREG/lA-0103. ASSESSMENT OF DETHSY TEST 91D USING NUREG/CR.6058 VIRGINIA REGIONAL SEISMIC NETWORK F inal RELAPS/ MOD 3.

Report (19PS,1992)

NUREG/IA-0104: RELAPS/ MOD 3 ASSESSMENT USING THE SEMIS-CALE 50% FEED LINE BREAK TEST S-FS 11 CHAVEZ,S.A.

NUREG/tA Ot05. ASSESSMENT OF RELAP5/ MOO 3 VERSION SM5 NUREG/CR.5642 LIGHT WATER REACTOR LOWER HEAD FAILURE USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF ANALYSIS.

KORI UNIT 3 PLANT.

CHEN,J C.

CHUNO.H.M.

NUREG/CR4832 V08 ANALYSIS OF THE LASALLE UNIT 2 NUCL EAR NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN POWER PLANT-RISK METHODS INTEGRATION AND EVALUATION LIGHT WALER REACTORS. Semiannual ReportApril-September 1992.

PROGRAM (RM!EP) Seismic Analysis.

NUREG/CR4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN fg CHENGFL NUREG/CR-6111. INTEGRATED SYSTEMS ANALYSIS QF THE PIUS REACTOR CICOTTE,0.R.

^ ^

CHEVERTON.R D.

NUREG/CR 5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC CLARK,T.

FaACTUR" MECHANICS SENSITIVITY ANAL.YSIS FOR YANKEE NUREG/CR-6050: RADIATION EXPOSURE MONITORING AND INFOR.

RGWE REACTOR PRESSURE VESSEL.

MATION TRANSMITTAL (REMIT) SYSTEM User's Manual.

CHtEN.N.O CLEVELAND,J C.

NUREG/CR 5937 INTENTIONAL DEPRESSURl2ATION ACCIDENT NUREG/CR-5922 MODULAR HIGH-TEMPERAlURE GAS-COOLED RE-MANAGEMENT STRATEGY F09 PRESSURIZED WATER REAC-ACTOR SHORT-TERM THERMAL RESPONSE TO FLOW AND REAC-TORS.

TIVITY TRANSIENTS.

CHO,C.S.

URE C -5917 VOI: SENSITIVITY AND UNCERTAINTY ANALYSES F LU FB MELT ATTAC OF E LINER APPLIED TO ONE. DIMENSIONAL RADIONUCLIDE TRANSPORT IN A LAYERED FRACTURED ROCK.MULTFRAC - Ar.dlytic Solutions And CHO,S.

NUREG/lA-0092: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE Local Sensnivmes.

NUREG/CR-5917 V02. SENSITIVITY AND UNCERTAINTY ANALYSES AGAINST THE NET LOAD TRIP TEST DATA FROM YONG.

APPLIED TO ONE-DIMENSIONAL RADIONUCLIOE TRANSPORT IN A GWANG, UNIT 2.

NUREG/lA 0100 ASSESSMENT OF CCFL MODEL OF RELAP5/ MOD 3 LAYERED FRACTURED ROCK Evaluation Of The Limit State Ap-AGAINST SIMPLE VERTICAL TURES AND ROD BUNDLE TESTS.

proach.

NUREG/lA.0125 ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG.

CONGEMI,J.

NUREG/CR-2907 VII, RADIOACTIVE MATERIALS RELEASED FROM GWANG UNIT 2.

NUCLEAR POWER PLANTS Annual Report 1990.

CHUPR A.O.K.

NUREG/CR4fi67 V15: ENVIRONMENTALLY ASSISTED CRACKING IN CONNOR,C B.

LIGHT WATER RE ACTORS. Semiannual ReportApni-September 1992.

NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE 3

NUREG/CR 4667 V16. ENVIRONMENT ALLY ASSidTED CRACKING IN RESEARCH AT CNWRA. July-December 1992.

LIGHT WATER REACTORS Semiannual Report,0ctober 1992 - March CONOSCENTE,J.P.

1993.

NUREG/CR 4744 V07 N1: LONG. TERM EMBRITTLEMENT OF CAST NUREG/CR-6013. METHODS USED FOR THE TREATMENT OF NON-DUPL EX ST AINL ESS STEELS IN LWR SYSTEMS Semiannual PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

Report. October 1991 - March 1992, NUREG/CR 4744 V07 N2' LONG-TERM EMBRITTLEMENT OF CAST COOPER,S.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual NUREG/lA 0118; ANALYSIS OF LOFT TEST L51 USING RELAPS/

i MOD 2.

l Report. April-September 1992.

NURLG/CR 5999 INTERIM FATIOUE DESIGN CURVES FOR CARBON,

~

LOW-alloy, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI.

COPUS.E.R.

NUREG/CR 5907: CORE-CONCRETE INTERACTIONS WITH OVERLY.

RONMENT S.

ING WATER POOLS.The WETCOR-1 Test CHOWDHURY,A H.

NUREu/CR 5217 V07. NRC HIGH-LEVEL RADIOACTIVE WASTE RE.

CORWIN,W.R.,

t' SEARCH AT (,NWRA Calendar Year 1991, NUREG/CR-5s91 V01 N2: HEAVY SECTION STEEL IRRADIATION NUREG/CR.58:7 V03 N1. NRC HIGH LEVEL RADIOACTIVE WASTE PROGRAM. Semiannual Progress Roport for Apnl-September 1990.

t RE SEARCH AT CNWRA January June 1992.

NUREG/CR 58I7 V03 N2 NRC HIGH LEVEL RADIOACTIVE WASTE COSTANTINO,C.J.

RESEARCH AT CNWRA July-Decemt>er 1992 NUREG/CR 5956. CONSIDERATION OF UNCERTAINTIES IN SOIL-NUREG/CR 6021. A LITERATURE REVIEW OF COUPLED THERMAL.

STRUCTURE INTERACTION COMPUTATIONS HYDROLOGIC MECHAN ICAL -CHEMICAL PROCESSES PERT!NENT TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA COTTER B.P.

NUREG-1363 V05: ATOMIC SMETY AND LICENSING BOARD PANEL MOUNT AIN, ANNUAL REPORT. Fiscal Yoer 1992.

I CHU.C.C.

NUREG/CR 6025 THE PROBABILITY OF MARK 1 CONTAINMENT COWGILL.M.G.

FAILURE BY MELT ATTACK OF THE LINER NUREG/CR-5911: SOURCE 1ERM EVALUATION FOR RADIOACTIVE LOW-LEVEL WASTE DISPOSAL PERFORMANCE ASSESSMENT, CHU ANG T.Y.

NUREG/CR 4832 V08: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR CRAGNOLINO,0.

POWER PLANT. RISK METHODS INTEGRATION AND EVALUATION NUREG/CR-E*U W2; NRC HIGH-LEVEL RADIOACTIVE WASTE RE-PROGRAM (RMIEPl Se srrac Analyses.

SEARCH AT CNWRA. Calendar Year 1991.

NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE CHUNG,B-D, RESE ARCH AT CNWRA. January June 1992.

NUREG/lA-0095: RELAPS ASSESSMENT USING LSTF TEST DATA SB-NUREG/CR-5817 V03 N2; NRC HIGH-LEVEL RADIOACTIVE WASTE CL 18 RESEARCH AT CNWRA. July-December 1992.

NUREG/lA-0099-RELAP5 ASSESSMENT USING SEMISCALE SBLOCA TEST S-NH-1 CHAMOND W.R.

NUREGriA 0100 ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 NUREG/CR 5863. RISK ASSESSMENT OF ISOLATION DEVICES IN AGAINST SIMPLE VERTICAL TUDES AND ROD BUNDLE TESTS SAFETY SYSTEMS.

l l

l l

l

. mm

.4 y

Personal Author index 59 CRESPO.J L DJE MIL.T.

NUREG/lA-0122 ASSESSMENT OF MS!V FULL CLOSURE FOR NUREG/CR 5977 A PERFORMANCE INDICATOR OF THE EFFECTIVE.

SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING NESS OF HUMAN-MACHINE INTERFACES FOR NUCLEAR POWER TRAC-BF1 (01J1)

PLAN TS CURCA TIVIGJ, DOBBE.C.A.

NUMEG/lA-0116. ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST NUREG/CR 5949 ASSESSMENT OF THE POTENTIAL FOR HIGH THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWP PRESSURE MELT EJECTION RESULTING FROM A SURRY STATION j

HOT LEG)

BLACKOUT TRANS!ENT.

DALING.P.M.

DOCTOR,S.R.

NUREG/CR 6084 VALUE-lMPACT ANALYSIS OF GENERIC ISSUE 143, NUREG/CR-4469 VIS: NONDESTRUCTIVE EXAMINATION (NDE) RELl-

" AVAILABILITY OF HEATING, VENTILATION, AIR CONDITIONING ABILITY FOR INSERVICE INSPEC TION OF LIGHT WATER (HVAC) AND CHILLED WATER SYSTEMS "

REACTORS Semaannual Report. October 1991 March 1992 NUREG/CR-4469 V16 NONDESTRUCTIVE EXAMINATION (NDE) RELI-DAME RE LL,P.S.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/lA.012fi 2D/30 PROGRAM WORK

SUMMARY

REPOrli REACTORS Semiannual ReportApril 1992. September 1992.

NUREG/lA 0127. REACTOR SAFETY ISSUES RESOLVED BY THE 2D/

NUREG/CR-5410: STATISTICALLY BASED REEVALUATION OF PISC-ll 3D PROGRAM ROUND ROBIN TEST DATA.

DAMERON.R.A.

DODDS,R.H.

NUREG/CR 6025-THE PROBABILITY OF M ARK-1 CONT AINME N T NUREG/CR-5969. J AND CTOD ESTIMATION EQUATIONS FOR SHAL-FAILURE BY MLLT. ATTACK OF THE LINER LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS.

NUREG/CR 5970: APPROXIMATE TECHNIQUES FOR PREDICTING DANIEL.S L SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNE SS (JC)

NUREG/CR.4832 V09' ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION OF CONSTRAINT IN FRACTURE.

PROGRAM (RMIEP) Internal F ne Analy9s NUREGICR-5791: RISK EVALUATION FOR A GENERAL ELECTRIC DODGE.F.T.

DWR, EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON NUREG/CR-5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE.

SAF ETY RELATED EQUIPMENT. Evaluaten Of Genenc issuo 57.

SEARCH AT CNWRA Calendar Year 1991.

NUREG/CR 5817 V03 N2: NRC HIGH-LEVEL RADIO /CTIVE WASTE DE LA CAL,C.

RESEARCH AT CNWRA Jufy-December 1992.

NUREG/1A-0085-ASSESSMENT OF FULL POWER TURBANE TRtP NUREG/CR 6026^ THEORETICAL AND EXPERIMENTAL INVESTIGA.

START UP TEST FOR C TRILLO I WITH RELAPS/ MOD 2 TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY SATURATED, FRACT URED POROUS MEDIUM l

NUMEG/CR.6049-PIPING BENCHMARK PROBLEMS FOR THE GEN-DOROF EEY,S.B.

ERAL ELECTRIC ADVANCED DOluNG WATER REACTOR.

NUREG/CR 6072 EXPERIMENTAL STUDY ON THE COMOUSTION DE-HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-DE Y,M-NITION AT LARGE SCALE NUREG/CP-0129 PROCEEDINGS OF THE WORKSHOP ON PROGPAM FOR EUMINATION OF REQUIREMENTS MARGINAL TO SAF ETY.

DOTY.K.

NUREG/CH-2907 V11: RADIOACTIVE MATERIALS RELEASED FROM DHIR,V.M.,

NUCLEAR POWER PLANTS. Annual Report 1990 NUREG/CR 6056 A FRAMEWORK FOR THE ASSESSMENT OF SLVERE ACCIDCNT MANAGEMENT STRATEGIES DURBIN,N.

NUREG/CR-5758 V03 FITNESS FOR DUTY IN THE NUCLEAR POWER DI AS,M.P.

INDUSTRY. Annual Summary Of Program Performance Reports,CY NUREG/CR 5951 THE MANAGEMENT OF ATWS BY DORON INJEC-1992-TION I

DUTCHE R,R.A.

DI AZ,A.A.

NUREG/CR.6047: CONTINUOUS SPE CTROSCOPIC ANALYSIS OF NUREG/CR-4469 V15. NONDESTRUCTIVE EXAMINATION (NDE) REll-VANADOUS AND VANADIC IONS.

ADILITY FOR INSERVICE INSPECTION OF LIGHT WATER RE ACTORS Semiannual Report. October 1991 March 1992.

EFIMENKO.A.A.

NUREG/CR-4469 V16. NONDESlRUCTIVE EXAMINATION (NDE) REll-NUREG/CR 6072. EXPERIMENTAL STUDY ON THE COMBUSTION BE.

ABILITY FOR INSERVICE INSPECTION Of LIGHT W AT E R HAvlOR OF HfDROGEN-AIR MIXTURES WITH TURBULENT JET IG-RE ACTORS.Semsannual Report Apnl 1992-September 1992.

NITION AT LARGE SCALE.

DIDIASIO,A M-EISE NHOWER,E, NUREG/CR 6014. HIGH PRESSURE COOLANT INJECTION SYSTEM NUREG/CR-6062: PERFORMANCE OF PORTABLE RADIATION RISK RASED INSPECTION GUIDE FOR HATCH NUCLE AR POWER SURVEY INSTRUMENTS.

STATION NUREG/CR 6022 HIGH PRESSURE COOLANT INJECTION iHPCI)

ELLINGWOOD.B.

SYSTE M RISK-BASED INSPECTION GUIDE FOR BROWNS FERRY NUREG/CR 6052: METHODOLOGY FOR REllABILITY BASED CONDI-NUCLEAR POWER STATION TION ASSESSMENT. Apphcabon To Concrete Structures in Nuclear Plants.

1 DICKSON.T.L.

NUREG/CR 5782 PRESSURIZED THERMAL SHOCK PROBABitfSTIC E LLIOTT,0.B.

FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CR 5975: INCEN1tVE REGULATION OF INVESTOR-OWNED ROWE. REACTOR PRESSURE VESSEL NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULATORS.

NUREG/CH 6023 GENERIC ANALYSIS FOR EVALUATION OF LOW CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS ELLISON,P C.

AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERI-NUREG/CR-5928 ISLOCA RESEARCH PROGRAM Fina! Report.

ALS E MRIT,R.

DINGMAN.S.E NUREG 0933 S15-A PRIORITIZATION OF GENERIC SAFETY ISSUES.

]

NUREG/CR-5936 ENHANCEMENTS TO THE ACCIDENT PRECURSOR NUREG-0933 SI6. A PRIORITIZATION OF GENERIC SAFETY ISSUES METHODOLOGY I

ESCALANTE,E.

DIX SON,R.

NUREG/CR-4735 V08. EVALUATION AND COMPILATION OF DOE 1

NUREG/CR-6030 R ADIATION EXPOSURE MONITOH.NG AND INFOR-WASTE PACKAGE TEST DATA. Brannual Report August 1989 - Janu.

M AT!ON 1R ANSMITT AL (Rf M:T) SYSTEM user's Manual ary 1990.

l

.~. _

60 Personal Author index EVANS,D,0-FOWLER,R.D.

NUREG/CP-0040 PROCEEDINGS OF WORKSHOP V, FLOW AND NUREG/CR4976: DEVELOPMENT AND USE OF A TRAIN LEVEL TRANSPORT THROUGH UNSATURATED FRACTURED ROCK.. RE PROBADILISTIC RISK ASSESSMENT.

LATED TO HIGH LEVEL RAD'OACTIVE WASTE DISPOSAL Hold Al Radmon Suite Hotet. Tucson, Anzona. January 7 10,1991 FO X,R.A.

NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMITY DOSI.

E V A N S,J.S.

METERS -PILOT TES7 NUREG/CR-4214 R2 PTl: HE ALTH EFFECTS MODEL FOR NUCLEAR F R AK ER.A.C, POWER PLANT ACC; DENT CONSEQUE NCE ANALYSIS Part 1.

NUREG/CR 4735 V08 EVALUATION AND COMPILATION OF DOE introductinn, integration.And Summary WASTE PACKAGE TEST DATA. Biannual Report. August 1989. Janu-F ADDE N.M.

ary 1990, NUREG-0525 V02 R01-SAFEGUARDS

SUMMARY

EVE NT LIST FRANC (SSEL) January 1,1990 Through December 31.1992 g / 4599 V02 N2' SHORT CRACKS IN PIPING AND PlP NG WELDS Semiannual Report. October 1991. March 1992.

FAIN,RI.

NUREG/CR-4599 V03 N1; SHORT CRACKS IN PIPING AND PIPING NUREG/CR 5851 LONG TERM PERFORMANCE AND AGING CHAR-WELDS Somiannual Report. Apnt-Septemter 1992.

ACTERiSTICS OF NUCLEAR PLANT PRESSURE TRANSMIT 1ERS FRANCIS,A.A.

F ANOUS.F.

NUREG/CR.5938: NATIONAL PROFILE ON COMMERCIALLY GENER.

NUREG/CR 4273. CRACK PROPAGATION IN HIGH STRAIN REGIONS ATED LOW-LEVEL RADIOACTIVE MIXED WASTE.

OF SEOUOYAH CONTAINMENT.

NUREG/CR 5957. SYSTEM BO +(TM) CONTAINMENT ~ STRUCTURAL FREDERICK,L DESIGN REVIEW NUREG-1415 V06 N01-OFFICE OF THE INSPECTOR GENER ALSemiannual Report.Apol 1.1993. September 30.1993.

FARRAR.C.R.

NUREG/CR 5755 STIFFNESS OF LOW. ASPECT RATIO. REINFORCED FRILEY,J.R.

CONCRE~TE SHEAR WALLS NUREG/CR-4469 V15 NONDESTRUCTIVE EXAM! NATION (NDE) REll-NUREG/CH 5776: DAMPlNG IN LOW ASPECT RATIO. REINFORCED ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER CONCRETF SHEAR WALLS RE ACTORS Sem, annual neoort. October 1991 March 1992 NUREG/CR-4469 V16 NONDESTRUCTIVE EXAMINATION (NDE) REll.

FAUVER,0N.

ABILITY FOR INSERVICE INSPECTION OF LIGH T WATER NUREG 1444 SlYE DECOMMISSIONING MANAGEMENT PLAN RE ACTORS Semiannual Report.Apnl 1992-September 1992.

F AYE R.M.J.

NUREG/CR-5996 SUBSURFACE INJECTION OF RADIOACTIVE UAE CR 5926: SANS INVESTIGATION OF LOW ALLOY STEELS IN TRACE RS Fmid Expenment For Model Vawlation Testing NEUTRON IRRAQlATED, ANNEALED, AND REIRRADIATED CONDI-TIONS F ERNANDEZ,RA.

NUREG/lA 0122. ASSESSMENT OF MSIV FULL CLOSURE FOR f '6111: INTEGRATED SYSTEMS ANALYSIS OF THE PI SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING g

TRAC-DF 1 (GlJ11' REACTOR.

f FIELD,l.

GAAL,M.

I NUREG/CR-5758 V01 FITNESS f OR DUTY IN THE NUCLEAR POWER NUREG/CP-0129: PROCEEDINGS OF THE WORKSHOP ON PROGRAM INDUST RY_ Annual Summary Of Program Performance Reports.CY FOR EllMINATION OF REOUIREJENTS MARGINAL TO SAFETY.

1992 GALLEGO,1.

FINK,J K.

NUREG/lA.0120: ASSESSMENT OF THE TUR81NE TRIP TRANS!ENT NUREG/CR 6032-SOLIDUS AND LIOUlDUS TEMPERATURES OF IN COFRENTES NPP WITH TRAC-8F1.

1 CORF. CONCRETE MIXTURES GALY E AN,W.J.

FIRST,M.W.

NUREG/CR 5928. ISLOCA RESEARCH PROGRAM. Final Report.

NUREG/CP 0130 V01. PROCEEDINGS OF THE 22ND DOE /NRC NU.

CLE AR Ain CLEANING CONF ERENCE Sessions 18 Held in GARDNER,D R.

Donver.Colmado. August 24 27,1992.

NUREG/CR-5843-CORCON MOD 3.AN INT EGRATED COMPUTER NUREG/CP 0130 V02: PROCEEDINGS OF THE 22ND DOE /NRC NU.

MODEL FOR ANALYSIS OF MOLTEN CORE CONCRETE CLEAR AIR CL E ANtNG CONFf RENCE Sessions 916 Held in INTERACTIONS. User's Manual Donver, Color ado. August 24-27,1992.

GARDNER,WA FISCHE R.L.E.

NUREG!CR-5980: THREE DtMENSIONAL REDISTRIGUTION OF TRITI.

NUREG/CR-6007 STRESS ANALYSIS OF CLOSURE DOLTS FOR UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED SHIPPING CASKS.

SOILA Determinishc Model For Tntium Mgration in An And Dmposal i

Site.

FISHER M S.

NUREG/CR 6108: SPHERICAL DIFFUSION OF TRITIUM FROM A NUREGICR 6047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF POINT OF RELEASE IN A UNIFORM UNSATURATED SOll.A Deter.

VANADOUS AND V ANAD3C IONS minatic Model For intium M.gration in An And Disposal Site.

l FLEMING,T GARNER.LW.

NUREG/CR4758 V03 FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/CR.5833 AUXILIARY F EEDWATER SYSTEM RISK BASED IN-INDUST RY. Annual Summary Of Pregram Performance Reports.CY SPECTION GU6DE FOR THE H.B ROBINSON NUCLEAR POWER 1992.

PLANT FONTAN A.M.H G AUT AM,A.S.

1 NUREG/CFM065 SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR NUREG-1473 ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL IN.

l SPECTION (EDSFI) DATA BASE PROGRAM.

l NUREG rCR4 v03. FITNESS FOR DUTY IN THE NUCLEAR POWER G E E,G.W.

INDUSTRY A vual Summary Of Prepam Performance Reports,CY NUREG/CR-5988. SOIL CHARACTERl2AllON METHOOS FOR UN.

1992 SATURATED LOW-LEVEL WASTE Sif ES FOSTER,J.

GHADIAli,N, NUREG/CR-6059 MACCS VERSION 15111. A MAINTENANCE RE-NUREG/CR-4599 V02 N2' SHORT CRACKS IN MPING AND PIPING l

LEASE OF THE CODE.

WELDS Semiannual Report, October 1991. March 199?.

1 1

i Personal Author index 61 NUREG/CR 4599 V03 N1 SHORT CRACKS IN PIPING AND P1PlNG GREENE R.H.

WELDS Semiannuat Report, Aprd Septemtmr 1992.

NUREG/CR-5699 V01: AGING AND SERVICE WEAR OF CONTROL GHAN.LS.

ROD DRIVE MECHANISMS FOR BWR NUCLE AR PLANTS.

NUREG/CR-5818: UNCERTAINTY ANALYSIS OF MINIMUM VESSEL GREENWOOD'M'S.

LIOulD INVENTORY DURING A SMALL. BREAK LOCA IN A B&W PLANT-AN APPLICATION OF THE CSAU METHODOLOGY USING NUREG/CR-4469 V15; NONDESTRUCTIVE EXAMINATION (NDE) RELL THE REL AP5/ MOD 3 COMPUTER CODE' ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report, October 1991 March 1992.

GHOSH.A.

NUREG/CR4469 V16: NONDESTRUCTIVE EXAMINATION (NDE) RELI-NUREG/CR-SR17 V03 N2. NRC HIGH. LEVEL RADIOACTIVE WASTE ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER RESEARCH AT CNWRA. July. December 1992.

RE ACTORS. Semiannual Report.Apnl 1992-September 1992, GILBFRT E.S.

GREIMANN.L.

NUREG/CR-4214 rip 2A2: HEALTH EFFECTS MODELS FOR NUCLE-NUREG/CR.4273' CRACK PROPAGATION IN HIGH STRAIN REGIONS AR POWER PLANT ACCIDENT CONSEQUENCE OF SEQUOYAH CONTAINMENT.

ANALYSIS Modificahon Of Models Resulting From Addition Of Effects NUREG/CR 5957: SYSTEM 80 +(TM) CONTAINMENT - STRUCTURAL Of Exposure To Alpha Ernitung Radonuclides Part II: Scientif c Bases DESIGN REVIEW.

For Health...

NUREG/CR 4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR GRIFFITH,R.O.

POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Pa1 1:

NUREG/CR-5843: CORCON-MOD 1AN INTEGRATED COMPUTER Int <oducton.Integrabon,And Summary MODEL FOR ANALYSIS OF MOLTEN CORE CONCRETE COMEZ,A.

INTERACTIONS. User's Manual.

NUREG/lA-0120. ASSESSMENT OF THE TURBINE TRIP TRANSIENT IN COFRENTES NPP WITH TRAC-BF1.

NUI E /C 0128. PROCEEDINGS OF THE INTERNATIONAL WORK.

i GOOD,M.S.

SHOP ON THE CONDUCT OF INSPECTIONS AND INSPECTOR NUREG/CR 4469 V15' NONDESTRUCTIVE EXAMINATION (NDE) RELi-QUAllFICATION AND TRAINING.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Sermannual Report, October 1991 March 1992, GRIMSHAW,C.

NUREG/CR-4551 V7R1Pt: EVALUATION OF SEVERE ACCIDENT GOODWIN,G M.

RISKS: ZION UNIT 1. Main Roport NUREG/CR 5972. EFFECTS OF NONSTANDARD HEAT TREATMENT NUREG/CR-455 tV7R1P2A-EVALUATION OF SEVERE ACCIDENT TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES RISKS: ZtON UN(T 1. Append x A.

OF CARBON-STEEL CASTING REPAIR WELDS.

NUREG/CR-4551V7 RIP 20: EVALUATION OF SEVERE ACCIDENT GORE,B.F.

RISKS. ZION UNIT 1.Appendees B, C, D And E.

NUREG/CR 5438 RISK BASE D INSPECTION GUIDE FOR THREE MILE GROUNDWATER,E'. SURVEY AND ASSESSMENT OF CONVENTIONAL ISLAND NUCLEAR STATION UNIT 1~

NUREG/CR-6018 NUREG/CR 5760. AUXILIARY FEEDWATER SYSTEM RISK. BASED IN.

SPECTICN GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR SOF TWARE VERIFICATION AND VALIDATION METHODS.

POWER PLANT, NUREG/CR 5629 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

GROVE,E.

SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER NUREG/CR-5783: AGING ASSESSMENT OF THE COMBUSTION ENGI-PL AN T, NEERING AND BABCOCK & WILCOX CONTROL ROO DRIVES-NUREG/CR 5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER GUNTHER,W.

l PL ANT.

NUREG/CR-5783: AGING ASSESSMENT OF THE COMDUSTION ENGL NUHEG/CR-5834 AUXILIARv' FEEDWATER SYSTEM RISK-BASED IN-NEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES.

SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER NUREG/CR-5933: HIGH PRESSURE COOLANT INJECTION (HPCf)

PLANT-SYSTEM RISK.DASED INSPECTION GUIDE FOR DRESDEN NUCLE-NURLG/CR-5835 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-AR POWER STATION UNITS 2 AND 3 SPECTION GUIDE FOR THE BEAVER VALLEY. UNITS 1 AND 2 NU' NUREG/CR-5934. HIGH PRESSURE COOLANT INJECTION (HPCI)

NL C 5 6 UX LI RY FEEDWATER SYSTEM RISK BASED IN-KU T D

ECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER NUREG/CR-5959: HIGH PRESSURE COOLANT INJECTION (HPCI)

NUREG/CR 5897. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SYSTEM RISK-BASED INSPECTION GUIDE FOR ENRICO FERMI A

PO A T

2 SPECTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR REG R SVRE COOLANT INJECTION (HPCI)

NUREG/CR-5896.' AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

SYSTEM RISK BASED INSPECTION GUIDE FOR BROWNS FERRY SPECTION GUIDE FOR THE POINT BEACH NUCLEAR POWER NUCLEAR POWER STATION OUREGHIAN,A.B.

GR AN T.T.

NUREG/CR-5917 V01: SENSITIVITY AND UNCERTAINTY ANALYSES NUREG/CR-5758 V01 FITNESS FOR DUTY IN THE NUCLEAR POWER APPLIED TO ONE-DIMENSIONAL RADIONUCLIDE TRANSPORT IN A INDUSTRY Annual Summary Of Program Performance Reports.CY LAYERED FRACTURED ROCK MULTFRAC Analytic Solutions And 1992.

Local Sensitivthes.

NUREG/CR 5917 V02: SENSITIVITY AND UNCERTAINTY ANALYSES GRAVES,C.C.

APPLIED TO ONE-DIMENSIONAL RADIONUCLIDE TRANSPORT IN A NUREG-1364 REGUL ATORY ANALYSIS FOR THE RESOLUTION OF LAYERED FRACTURED ROCK.Evaluahon Of The Limit State Ap-GENERIC SAFETY ISSUE 106. PIPING AND THE USE OF HIGHLY proach.

COMBUSTIBLE GASES IN VITAL AREAS O'

GREEN R.T NUREGICR-5817 V02. NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREGICR 5358: REVIEW OF ASME CODE CRITERIA FOR CONTROL SEARCH AT CNWRA Calendar Year 1991 OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM DRANCH CON-NUREGICR 5817 V03 N1. NRC H!GH-LEVEL RADIOACTIVE WASTE NECTIONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP.

RESEARCH AT CNWRA. January June 1992.

MENT WORK.

NUREG/CR 5817 V03 N2: NRC HIGH LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA Juty-Decemter 1992 HACK B ARTH.H.

NUREG/CR 4026 THEORETICAL AND EXPERIMENTAL INVESTIGA-NUREG/CR 5926; SANS INVESTIGATION OF LOW Alt.OY STEELS IN TION OF THERMOHYDROtOGIC PROCESSES IN A PARTIALLY NEUTRON IRRADIATED. ANNEALED, AND REIRRADIATED CONDI-SATURATED f RACTURED POROUS MEDIUM TiONS.

62 Personal Author Index HAGEMEYER,0.

NUREG/CR-6011: REVIEW OF STRUCTURE DAMPlNG VALUES FOR NUREG-0713 V12: OCCUPATIONAL RADIATION EXPOSURE AT COM-ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PL ANTS.

MERCIAL NUCLEAR POWER REACTORS AND OTHER NUREG/CR4012: STIFFNESS AND DAMPlNG PROPERTIES OF A F ACILITIES 1990.TwentrTNrd Annual Report.

LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD-NUREG 0713 V13. OCCUPATIONAL RADIATION EXPOSURE AT COM-ED EARTHOUAKE RESPONSES MERCIAL NUCLEAR POWER REACTORS AND OTHER F ACILITIES,1991 Twenty Fourth Annual Report.

HAWTHOR6.J.

NUREG-0713 V14' OCCUPATIONAL RADIAllON EXPOSURE AT COM-NUREG/CR-5920: SANS INVESTIGATION OF LOW ALLOY STEELS IN MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI-1992 Twenty Fdth Annual Report TIONS.

NUREGICR 60*>0: RADIATION EXPOSURE MONITORING AND INFOR.

MATION TRANSMITT AL (REMIT) SYST EM User's Manual.

HEAMES,T.J.

NUREG/CR-6025: THE PROBABILITY OF MARK 4 CONTAINMENT HAGGARD.D.L F AILURE BY MELT-ATTACK OF THE LINER.

NUREG/CR-5894 RADIONUCLIDE CHARACTERIZATION OF REAC.

TOR DE COMMISSIONING WASTE AND NEUTRON ACTIVATED HE ASLER,P.G.

MET AL S NUREG/CR-4469 V15: NONDESTRUCTIVE EXAMINATION (NDE) RELI-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER RE ACTORS Semiannual Report. October 1991. March 1992.

HAMAL AINEN.A'0. ASSESSMENT OF RELAPS/ MOD 2 USING THE TEST NUREG/CR-4469 V16: NONDESTRUCTIVE EXAMINATION (NDE) REll-NUREG/iA 009 DATA Or REWET II REFLOODING EXPERIMENT SGI/R ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER REACTORS Semiannual Report. April 1992. September 1992, HAMDAte'L

^

^

NUREG 1476 FINAL ENVIRONMENTAL IMPACT STATEMENT TO OUNO B TEST CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE. AND OfSPOSE OF 11E (2)

BYPRODUCT MATERIAL NEAR HEATH,C.H.

CLIVE, UTAH Docket No 40 0989.Enwrocare Of Utah.Inc-NUREG/CR-5642. LIGHT WATER REACTOR LOWER HEAD FAILURE NUREG-1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT ANALYSIS.

TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE.

AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR CLfvE.

HEBOON.F.J.

UTAH Docket No. 40 89B9, Enwrocare Ol Utah tnc.

NUREG-1474. EFFECT OF HURRICANE ANDREW ON THE TURKEY HAMILTON M.A.

NUREG/CR 5987 MICROBIAL. INFLUENCED CEMENT DEGRADATION

- LITERATURL REVIEW-HECHT,H.

NUREG/CR 6113' CLASS 1E DIGIT AL SYSTEMS STUDIES.

HAMILTON,S.

NURE G/CR4080 REPLACEMENT ENERGY. CAPACITY, AND RELi-HELLER.P.R.

ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT-NUREG/CR-5996: SUBSURFACE INJECTION OF RADIOACTIVE DOW %

TRACERS Fueld Experiment For Model Valedation Testn.g H AM ME R,J. A.

HERN AN,R.W.

NUREG/CR-5973. CODES AND STANDARDS AND OTHER GUIDANCE NUREG/CR 5488: RISK BASED INSPECTION GUIDE FOR THREE MILE CITED IN REGULA10RY DOCUMENTS ISLAND NUCLEAR STATION UNIT 1.

HAMMOND,0 A.

R

79. SEISMOLOGICAL INVESTIGATION OF EARTH-U RE OR RULE AND TS M LE ENTATI OVAKES IN THE NEW MADRID SEISMIC ZONE. Final j

Report. September 1986 December 1992.

H A NE Y,LN.i NUREG/CR-5937: INTENTIONAL, DEPRESSUnl2ATION ACCIDENT Hi

,E AGEMENT STRATEGY FOR PRESSURf7ED WATER REAC-G-400: AIR SAMPLING IN THE WORKPLACE Final Report t

HIGGINS,J.

HANSON,R G.

NUREG/CR-6111: INTEGRATED SYSTEMS ANAlfSIS OF THE PIUS NUREG/CR 6061: DETERMINATION OF THE BIAS IN LOFT FUEL REACTOR PEAK CLADDfNG TEMPERATURE DATA FROM THE BLOWDOWN PHASE OF LARGE BRE AK LOCA EXPEniMENTS HIGGINS.S.J.

NUREG/CR-5305 V02 P1; INTEGRATED RISK ASSESSMENT FOR THE HARLING,0.K.

LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NUREG/CR4001. ENHANCED REMOVAL OF RADIOACTIVE PARTI, Rs U n Eaa R P) A Qs

.OR THE CLES BY FLUOROCARBON SURFACTANT SOLUTIONg g

G LASALLE UNIT 2 NUCLEAR POWER PLANT:Phenomenology And HARRISON.D.G.

NUREG/CR 5488. RISK BASED INSPECTION GUIDE FOR T HREE MILE Risk Uncertainty Evaluation Pro 9fam (PRUEP) Appendices D-G.

ISLAND NUCLEAR STATION UNIT 1.

HARTFIELD.R A.

NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE NUREGo020 V17: LICENSED OPERATING REACTORS ST ATUS SUM-RESEARCH AT CNWRA July-December 1992 MARY RE PORT. Data A6 Of December 31,1992.(Grey Book 1)

HILL,S.G.

NUREG/CR-5953: STUDIES OF HUMAN PERFORMANCE DURING OP.

HARTY,R-NUREG/CR 5989 PERFORMANCE TESTING OF EXTREMITY DOSI-E RATING EVENTS.19901992.

METE RS -PILOT TEST.

HILLS,R.G.

HASHE MI AN,H.M.

NUREG/CR-5988. SOIL CHARACTERIZATION METHODS FOR UN-NURE G/CR-5851 LONG TERM PERFORMANCE AND AGING CHAR-SATURATED LOW LEVEL WASTE SITES.

ACTE RISTICS OF NUCLE AR PLANT PRESSURE TRANSMITTERS NUREG/CR 5903 VALIDATION OF SMART SENSOR TECHNOLOGIES HINS.A.G.

FOR INSTRUMENT CAllBRATION REDUCTION IN NUCLEAR NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual Report, April-September 1992.

POWER PLANTS.

HASHIMOTO.P.S.

HOCK E Y,R.L NUREG/CR 4832 VDS ANALYSIS OF THE LASALLE UNfT 2 NUCLEAR NUREG/CR 4469 V15: NONDESTRUCTIVE EXAMINATION (NDE) REll-POWER PLANI. RISK METHODS INTEGRATION AND EVAL UATION ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER PROGRAM (RMIEP) Smmac Ana!ys's.

REACTORS Samtannual Report. October 1991 March 1992.

l 1

Personal Author index 63 HOFMAYER,C.H.

NUREGICR-5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR 1HE NUREG/CR 6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN LASALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And THE HIGH LEVEL VfBRATION TEST AT TADOTSU.

Risk Uncertainty Evaluation Program (PRUEP). Appendices D-G.

NUREG/CR 5360: XSOR CODES USERS MANUAL NUREG 1467.

FEDERAL GUIDE FOR A RADIOLOGICAL JOHNSON.J.J.

RESPONSE Supporting The Nuclear Regulatory Commission Dunng NUREG/CR-4832 V08. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR The initial Hours Of A Senous Accident POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Seesmc Analysis.

HOPKINS,J.B.

NUREG/CR 6011. REVIEW OF STRUCTURE DAMPING VALUES FOR NUREG/CR 5829. AUXlllARY FEEDWATER SYSTEM RISK BASED IN-ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PLANTS.

SPECTION GUIDANCE FOR THE DAVIS BESSE NUCLEAR POWER NUREG/CR 6012. STIFFNESS AND DAMPING PROPERTIES OF A

PLANT, LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD-ED EARTHOUAKE RESPONSES HSIUNG,S.H.

NUREG/CR 6013 METHODS USED FOR THE TREATMENT DF NON-NUREG/CR 5817 V02 NRC HIGH4EVEL HADICACTfvE WASTE RE-PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS SFARCH AT CNWRA Calendar Year 1991.

NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE JOHNSON,T.

RESEARCH AT CNWRA. January June 1992.

NUREG-1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE.

HSIUNG,S.M.

AND DISPOSE OF 11E (2) BYPRODUCT MATERIAL NEAR CLIVE, NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE UTAH Docket No. 40-8989 Envrocare Of Utah, Inc.

RESE ARCH AT CNWRAJuly-December 1992.

JOHNSON,T.C.

N REG 1275 V09 OPERATING EXPERIENCE FEEDBACK REPORT -

PRESSURE LOCKING AND THERMAL BINDING OF GATE JOHNSON,T.L VALVES Commercial Power Reactors.

NUREG-1476 FINAL ENVIRONMENTAL IMPACT ST ATE MENT TO HSU,F CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, AND NUFkEG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION DISPOSE OF 11E (2)

BYPRODUCT MATERIAL NEAR AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE CLIVE. UTAH Docket No. 404 989, Envuocare Of Utah.Inc.

DAT A STATISTICS Summa'y Report JOLMY,R L NUREG/CR 5993 V02: METHODS FOR DEPENDENCY ESTIMATION NUREG/CR 5938. NATIONAL PROFILE ON COMMERCIAll.Y GENER-AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE ATED LOW-LEVEL RADIOACTIVE MIXED WASTE.

DAT A STATISTICS Detailed Description And Applications.

HSU.S.T JONES.B.J.

NUREG/CR 6007. STRESS ANALYSIS OF CLOSURE BOLTS FCWt NUREG/CR 5977: A PERFORMANCE INDICATOR OF THE EFFECTIVE.

SHIPPlfNG CASKS.

NESS OF HUMAN-MACHINE INTERFACES FOR NUCLEAR POWER PLAN T S.

H UB E R,D.S.

JORDAN A.

NUREG-1415 V06 N01: OFFICE OF THE INSPECTOR GENERALSemiannual Report.Apnl 1,1993 - September 30,1993.

TRACERS Field Expenment For Model Validation Testing.

HUDDLESTON,RL JOW N NU

/t 5955 MATERIALS AND DESIGN BASES ISSUES IN ASME HU1,T.E.

JOY,D.

NUREG/CR-5631 R1 ADD-CONTRIBUTION OF MATERNAL RADIONU.

NUREG 0430 V12:

LICENSED FUEL FACILITY STATUS CLIDE BURDENS TO PRENATAL RADIATION DOSES RelatonsNps REPORT. Inventory Difference Data. July 1,1991 - June 30,1992 (Gray Between Annual Limits On intabe And Prenatal Doses Book li)

INTERR ANTE,C.G.

JOYCE,J.A.

NUREGICR-4735 V08 EVALUATION AND COMPILATION OF DOE NUREG/CR-5961: THE EFFECT OF ELECTRIC DISCHARGE MA.

WASTE PACKAGE TEST DATA. Biannual Report. August 1989 Janu.

CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL ary 1990.

STRUCTURAL ALLOYS ISHil,M.

KAISER,R.

NUREG/GROOO9 STEPWISE INTEGRAL SCAllNG METHOD AND ITS NUREG/CR-6081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-APPLICATION TO SEVERE ACCIDENT PHENOMENA CLES BY FLUOROCARBON SURFACTANT SOLUTIONS JAE,M.

KAM.F.B.

NURE G/CR 4056. A FRAMEWORK FOR THE ASSESSMENT OF NUREGICR 6071: IMPACT OF ENDF/B-VI CROSS-SECTION DATA ON SEVERE ACCIDENT MANAGEMENT STRATEGIES.

H B. ROBINSON CYCLE 9 DOSIMETRY CALCULATIONS.

NUREG/CR-6117: NEUTRON SPECTRA AT DIFFERENT HIGH FLUX J ASTROW J D.

ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEILLANCE NUREG/CR 52?9 V05 FIELD LYSIMETER INVESTIGATIONS. LOW-LOCATIONS.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FISCAL YEAR 1992 Annual Report K AMPM ANN,R.

NUREG/CRT9?6: SANS INVESTIGATION OF LOW ALLOY STEELS IN JOHN SE N.G.W.

NFUTPON IRRADIATED. ANNEALED, AND REIRRADIATED CONDI-NUREG/CR-6061: DETERMINAllON OF THE BIAS IN LOFT FUEL TIONS PEAK CLADOING TEMPERATURE DATA FROM THE BLOWDOWN PHASE OF LARGE-BREAK LOCA EXPERIMENTS.

KASSIR.M.K.

NUREG/CR 6078. ANALYSIS OF CRACK INITIATION AND GROW 1H IN JOHNSON,D.

THE HIGH LEVEL VlBRATION TEST AT TADOTSU.

NUREG/CR 5771) V03 NEW YORK /NEW JERSEY REGIONAL SEISM C NETWORK Final Report For Apdl 1985. September 1992.

K ASSNE R,T.F.

NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN JOHNSON,J.D.

LIGHT WATER REACTORS. Semssnnual Report.Apnt-September 1992.

NUREG/CR 5305 V02 P1 INTEGRATED RISK ASSESSMENT FOR THE NUREG/CR 4667 V16 ENVIRONMENTALLY ASSISTED CRACKING IN LASALLE UNIT 2 NUCLEAR POWER PLANTPhenomenology And LIGHT WATER REACTORS Gemiannual Report. October 1992 March Rmk Uncertaanty Evaluation Program (PRUEP) Appendices A-C.

1993 a

64 Personal Author index KASTENDERG W.E.

KINCAID,C.T.

NUREG/CR4056. A FRAMEWOR4 FOR THE ASSESSMENT OF NUREG/CR-5988: SOIL CHARACTERilATION METHODS FOR UN-SEVERE ACCIDENT MANAGEMENT STRATEGIES, SATURATED LOW-LEVEL WASTE SITES.

KATSMA,K.R.

KINCAID,R.H.

NUREG/CR4035: FE ASIBILITY STUDY FOR IMPROVED STEADY-NUREG/CR 4832 V08-ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR STATE INITIAllZATION ALGORITHMS FOR THE RELAP5 COMPUT-POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION ER CODE.

PROGRAM (RMIEP) Seismg: Analysis.

K AURIN,D.G.

KlHNEMAN,J.D.

I NUREG/CR.3469 V07: OCCUPATIONAL DOSE REDUCT'ON AT NU-NUREG-1444; SITE DECOMMISSIONING MANAGEMENT PLAN.

CLEAR POWER PLANTS. ANNOTATED DIBLIOGRAPHY OF SELECT-ED READINGS IN RADIATION PROTECTION AND ALARA.

KIRK,M.T.

NUREG/CR-5969-J AND CTOD ESTIMATION EOUATIONS FOR SHAl-KAVLCKY A NUREG/CR'6080: REPLACEMENT ENERGY, CAPACITY, AND REll-LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS NUREG/CR 5970: APPROXIMATE TECHNIQUES FOR PREDICTING ABillTY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT-SIZE EFTECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC).

DOWNS.

t KLAMERUS E.

KE NU l'CP-0131: PROCEEDINGS OF THE JOINT IAEA/CSNI SPECIAL-NUREG/CR-5791: RISK EVALUATION FOR A GENERAL ELECTRIC ISTS' MEETING ON FRACTURE MECHANICS VERIFICATION OY b^

8 "" "

LARGE-SCALE TESTING. Held Al Pollard Auditonum, Oak FR UEM AN S NSIT ITY ANAL S FO Y K

^

^

ROWE REACTOR PRESSURE VESSEL KEENEY-WALKER,J KLEIN,R.F.

NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF NUREG/CR-6043 VOI: AGING ASSESSMENT OF ESSENTIAL HVAC LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS CHILLERS USED IN NUCLEAR POWER PLANTS Phase I.

(PROJECT F ALSIRE)

KNUDSON,0L.

KELLY,D.L NUREG/CR 5949: ASSESSMENT OF THE POTENTIAL FOR HIGH NUREG/CR 5928. ISLOCA RESE ARCH PROGRAM Finni Report.

PRESSURE MELT EJECTION RESULTING FROM A SURRY STATION NUREG/CR4027. PRELIMINARY EVALUATION OF SNUD6ER SINGLE BLACKOUT TRANSIENT.

FAILURES.

KOCHURKO.A.S.

KENNETT RJ '

NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE.

NUREG/CR-5984: CODE AND MODEL EXTENSIONS OF THE THATCH HAVIOR OF HYDROGEN. AIR MIXTURES WITH TURBULENT JET 10-CODE FOR MODULAR HIGH TEMPERATURE GASCOOLED REAC.

NITION AT LARGE SCALE.

TORS.

KOCMOUD,C.J.

K H AN,T.A.

NUREG/GR-0006. DEPOSITION SOFTWARE TO CALCULATE PARTI-NUREG/CR 3409 VOT. OCCUPATIONAL DOSE REDUCTION AT NU-CLE PENETRATION THROUGH AEROSOL TRANSPORT l

CLEAR POWER PLANTS ANNOTATED DIBLIOGRAPHY Or 1ECT-SYSTEMS Final Report.

ED READINGS IN RADIATION PROTECTION AND ALARA.

KONZE K,G.J.

KILINSKl,T.

NUREG/CR-5884 V1 DRF: REVISED ANALYSES OF DECOMMISSION-NUREGICR 4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING ING FOR THE REFERENCE PRESSURilED WATER REACTOR WEL DS Semiannual Report. October 1991 March 1992.

POWER STATION Ettects Of Current Regulatory And Other Consider.

NUREG/CR 4599 V03 NI: SHORT CRACKS IN PIPING AND PIPING ations On The financial Assurance.... Main Report. Draft Report For WELDS Semiannual Report, April September 1992-Comment.

I KIM,WJ.

NUREG/CR-5884 V2 DRF: REVtSED ANALYSES OF DECOMMISSION-NUREG/lA-0095 RELAP5 ASSESSMENT USING LSTF TEST DATA SO.

ING FOR THE REFERENCE PRESSURIZED WATER. REACTOR J

CL 18 POWER STATION Effects Of Current Regulatory And Other Consider-NUREG/lA 0099 RELAPS ASSESSMENT USING SEMISCALE SOLOCA abons On The Dnancial Assurance-. Appendices Draft Report For I

TEST S-NH4 Comment NUREG/lA 010& ASSESSMENT OF GCFL MODEL OF RELAPS/ MOD 3 AGAINST SIMPLE VERTICAL TUDES AND ROD DUNDLE TESTS KORTH G.E.

NUREG/lA4101 ASSESSMENT OF DETHSY TEST 91.8 USING NUREG/CR-5642: LIGHT WATER REACTOR LOWER HEAD FA! LURE RELAP5/ MOD 3 ANALYSIS.

NUREG/lA-0104. RELAF5/ MOD 3 ASSESSMENT USING THE SEMIS-CALE 50% FEED tlNE BREAK TEST S-FS-11.

KOTTLE S.

NUREG/lA-0105. ASSESSMENT OF RELAP5/ MOD 3 VERSION SM5 NUREG/CR-6047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF VANADOUS AND VANADIC IONS.

KORI UNIT 3 PLANT.

NUREG/lA 0125. ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE K OZ AK,M.W.

AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG_

NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS-GWANG UNIT 2.

MENT METHODOLOGY FOR LOW 4EVEL RADIOACTIVE WASTE DISPOSAL FACILITIES Evaluabon Of Modehng Approaches.

NUREG/1A-0105. ASSESSMENT OF RELAP5/ MOD 3 VERSION SMS KRISHNASWAMY,P, USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF NUREG/CA-4509 V02 N2: SHORT CRACKS IN PIPING AND PIPING KORI UNIT 3 PLANT.

WELDS Semiannual Report, October 1991 - March 1992.

NUREG/CR-4599 V03 N1: SHORT CRACKS IN PIPING AND PIPING KIM,LS.

WELDS. Semiannual Report, April. September 1992.

NUREG/CR-5995: TECHNICAL SPECIFICATION ACilON STATEMENTS REQUIRING SHU1DOWN A Rit,k Perspectsve Wtth Appitcation To The KROEGER,P.G.

RHR/SSW Systems Of A BWR NUREG/CR-5983: SAFETY ASPECTS OF FORCED FLOW COOLDOWN TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED KIM,K.T.

REACTORS.

NUREG/lA 0105. ASSESSMENT OF RELAP5/ MOD 3 VERSION SMS NUREG/CR.5984: CODE AND MODEL EXTENSIONS OF THE THATCH USING IN ADVERTENT SAFETY INJECTION INCIDENT DATA OF CODE FOR MODULAR HIGH TEMPERATURE GAS. COOLED REAC-KORI UNIT 3 PLANT.

TORS.

l er

.m.+

yi-+-m m

wm y.-.sn--ww-u-r e--

Personal Author index 65 NURE.G/CR4111-INTEGRATED SYSTEMS ANALYSIS OF THE PlUS LEE,S.H.

REACTOR NUREG/iA-0092. ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE AGAINST THE NET LOAD TRlP TEST DATA FROM YONG-KUECMAD-GWANG. UNIT 2 NUREG/CR8A04 VO2-AUXILtARY F EE DWATE R SYSTEM AGING NUREG/IA-0095 RELAPS ASSESSMENT USING LSTF TEST DATA SB-STUDY Phase I Follow-On Study CL-18.

^

MUHTZ,RJ RELAP5/ MOD 3.

NUREG/CR4469 V15. NONDESTRUCTIVE EX AMINATION (NDE) REL.1, ABILITY FOR INSE RVICE INSPECTION OF LIGHT W ATE R LEHNER,J.R.

NUREG/CR 5982: EFF ECTIVENESS OF CONTAINMENT SPRAYS IN N REG /C 44 N DE U VE E M ATI N NDE) REL.I.

ABillTY FOR INSERVICE INSPECTION OF LIGHT WATER CONTAINMENT MANAGEMENT.

REACTORS Semiannual Hoport, April 1992-September 1992 LElBOWITZ,L LAM.K.L NUREG/CR 6032 SOLIDUS AND LIQUIDUS TEMPERATURES OF NUREG/CA 6000. HYDROGEN MIXING STUDIES (HMS) ASSESSMENT CORE-CONCRETE MIXTURES.

LESLIE,0.W.

LAMBE R T,L.D.

NUREG/CR-5817 V02. NRC HIGH-t EVEL RADIOACTIVE WASTE RE-NUREG/CR-596t: POSTTEST DESTRUCTIVE EXAMINATION OF THE SEARCH AT CNWRA Calendar Year 1991.

STEEL LINER IN A 1.6 SCALE REACTOR CONTAINMENT MODEL NUREG/CR 5H17 V03 N1: NRC HIGH LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. January June 1992 LAMBRIGHT.J.A-NUHEG/CR 5817 V03 N2, NRC HIGH LEVEL RADIOACTIVE WASTE i

NUREG/CR4832 VOS. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR RESEARCH AT CNWRA July-December 1992 POWER PLANT: HISK METHODS INTEGRATION AND EVALUATION PROGRAM Paramelor Estimaton Analysis And Screening Human Reli-LEUNG,V.T.

absty Analysis-NUREG-1427. REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG/CH.4832 V09: ANALYSIS OF THE ( ASALLE UNIT 2 NUCLEAR GENERIC ISSUE 143 AVAIL ABILITY OF CHILLED WATER SYSTEM POWER PLANT: RISK METHODS INTEGRATION AND CVALUATION AND ROOM COOLING.

PROGRAM (RMIEP)lnternal Fire Analyms.

NURL'G/CR 5791 RISK EVALUATION FOR A GENERAL ELECTRIC LIM H.

BWR, EFF ECTS OF FIRE PROTECTION SYSTEM ACTUATION ON NUREG/CR 6056-A FRAMEWORK FOR THE ASSESSME NT OF I

SAFETY.RELATED EOUiPMENT. Evaluation Of Genenc issuo 57' SEVERE ACCIDENT MANAGEMENT STRATEGIES LANDOW.M,P.

NUREG/CR-4599 V02 NZ SHORT CRACKS IN PIPING AND PIPING NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE MA-WELDS Sermannual Report. October 1991 March 1992 NUREG/CR 4509 V03 N1 SHORT CRACKS IN PIP 1NG AND PIPING CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL WELDS Somiannual Report. April. september 1992.

STRUCTURAL ALLOYS.

J NUREG/CR -6098 LOADING RATE EFFECTS ON STRENGTH AND FRACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF LIN N.M. A.

THE iPIRG PROGRAM NUREG/CR 6065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR.

]

LAPP A,D.A.

LLOPIS,C.

NUREG/CR-4832 VOR ANALYSIS OF THE LASALLE UNIT 2 NUCLF AR NUREG/lA 0107-ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD POWER PLANT RISK METHODS INTEGRATION AND EVALUATION REJECTION F ROM 100% TO 50% POWER IN THE VANDELLOS 11 PROGRAM (HMIE P) Smsmsc Analyvs.

NUCLEAR POWER PLANT.

NUREG/lA.0108 ASSESSMENT OF RELAP5/ MOD 2 AGAINST A 1UR-LARRE A.E.

DINE TRIP FROM 100% POWER IN THE VANDELLOS ll NUCLEAR NUREG/lA-0085 ASSESSMENT OF FULL POWER TURBINE TRIP POWER PLANT.

ST ART-UP TEST FOR C TRILLO l WITH RE LAP 5/ MOD 2.

NUREG/lA-0109. ASSESSMENT OF RELAP5/ MOD 2 AGAINST A 10%

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE LAWRE NCE,J D.

VANDELLOS 11 NUCLEAR POWE R PLANT NUREG/CR 6601: SOFTWARE RELIABILITY AND SAFETY IN NUCL E.

NUREG/lA 0110: ASSESSMENT OF RELAP'5/ MOD 2 AGAINST A MA!N AR REACTOR PROTECTION SYSTEMS FE EDWATER TURDOPUMP TRIP TRANSIENT IN THE VANDELLOS ll NUCLEAR POWER PLANT,

[

LAWSON.J.E, NUREG/CR 6034' Okt.AHOMA SEISMIC NETWORK Final Report.

LLOYD.R.C.

LECHAS,A.L NUREG/CR-5833. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/lA 0123 APPLICATION OF FULL POWER OLACKOUT FOR SPECTION GUIDE FOR THE H B ROBINSON NUCLEAR POWER C N ALMARAZ WITH RELAPS/ MOD 2 PLANT.

NOREG/CR 5835 AUXILIARY FEEDWATER SYSTEM RISK BASED IN-i LE E.B.B-SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NU-

~

NUREG/CR 5844 AGING ASSESSMENT OF BIST ARL E S AND CLEAR POWER PLANTS SWITCHES IN NUCLEAR POWLR PLANTS.

NUREG/CR 5898. AUX 1LIARY F EEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE POINT BE ACH NUCLE AR POWER LEE,E J.

PLANT.

NUREG/lA 0099 RELAPS ASSESSMENT USING SEMISCALE SRLOCA TESTS NH.1-LOUEL R.

NUREG/lA Ot04 RELAP5/ MOD 3 ASSESSMENT USiNG THE SEMIS' NUREG 1366 IMPROVEMENTS TO TECHNICAL SPECIFICATIONS CALE 50% FEED LINE BRE AK TEST S-FS-11 SURVEILLANCE REQUIREMENTS LEE,J.D'G/CR 5977 A PERFORMANCE INDICATOR OF THE EFFECTIVE' LOPEZ,A NURE NUREG /lA-0085 ASSESSMENT OF FULL POWER TURHINE TRIP SS OF HUMAN-MACHINE INTERF ACES FOR NUCLEAR POWEFt ST ART.UP TF St FOR C. THILLO I WITH RELAPS/ MOD 2 LE E.M.

LOPEZ,E.

NUREG/CR4551 V7A1P1: E V AL U ATION OF SEVERE ACCIDE N t NUREG/lA4065. ASSESSMENT OF FULL POWER TURBINE TRIP RtSkS ZION UNIT 1 Main Report START.UP TEST FOR C. Trit LO I WtTH RELAP5/ MOD 2.

NUREG/CR 4551V7RtP2A. E V ALU ATION OF SEVERE ACCIDF NT LO2ANO.M.F.

RISAS ZION UNIT 1 Appenda A.

NUREG/CR 4551V7R t P2B EVALUATION OF SEVERE ACCIDF N T NUREG/lA 0085-ASSESSMENT OF FULL POWER TURDINE TRIP RISKS ZION UNIT 1. Appendices B, C, D. And E.

START-UP TEST FOR C. TRILLO I WITH RELAPL/ MOD 2.

n,n

~.

~

n 66 Personal Author Index LU.A,H.

MATSUKOV,1.D.

NUREG/CR-5996 SUBSURFACE INJECTION OF RADIOACTIVE NUREG/CR.6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-TRACERS F eld Expenment For Model Vahdahon Testing HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-NITION AT LARGE SCALE.

LUCERO.D A.

NUREG/CR 5901: CORE. CONCRETE INTERACTIONS WITH OVERLY-M A X WELL,J.S.

ING W ATER POOLS The WETCOR 1 Test NUREG/CA.6025. THE PROBABILITY OF MARK.I CONTAINMENT

^

LUK,K.H.

NUREG/CR 5754-DOILING WATER REACTOR INTERNALS AGING MC AF EE.W.J.

DEGRADATION STUDY. Phase 1.

NUREG/CR 6030 INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL NUREG/CR-6048: PRESSURIZED-WATER REACTOR INTERNALS LOADING ON FRACTURE TOUGHNESS.

AGING DEGRADATION STUDY. Phase 1.

MCCABE,0.E.

LUZA,K.V.

NUREG/CR 5914-CHEMICAL COMPOSITION AND RT(NDT) DETERMI.

NUREG/CR'6034: OKLAHOMA SEISMIC NETWORK Final Report NAT!ONS FOR MIDLAND WELD WF 70.

NUREG/CR 5952: EVALUATION OF CRACK POP.lNS AND THE DE-MAGUfRE MOFFITT TERMINATION OF THEIR RELEVANCE TO DESIGN CONSIDER.

NUREG/CR-5973 CODES AND STANDARDS AND OTHER GUIDANCE

ATIONS, CITED IN REGULATORY DOCUMENTS.

MCCONNELL J.W.

M AJUM D AR.S.'

NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5229 V05: FIELD LYSIMETER INVESTIGATIONS. LOW-LIGHT WATER REACTORS. Semsannual Report.Apol-September 1992.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR-5999. INTERIM FATIOUE DESIGN CURVES FOR CARBON, FISCAL YEAR 1992 Annual Report NUREG/CR-5672 V03' CHARACTERISTICS OF LOW-LEVEL RADIOAC.

LOW. ALLOY, AND AUSTENiilC STAINLESS STEELS IN LWR ENVI.

TIVE DECONTAMINATION WASTE Annual Report For Fiscal Year RONMENTS.

1992.

MANKAMO T.

NUREG/CR-5987: MICROBIAL. INFLUENCED CEMENT DEGRADATION NUREG/CR-5995: TECHNICAL SPECIFICATION ACTION STATEMENTS

- LITERATURE REVIEW.

nEQUIRING SHUTDOWN A Ra,k Perspective With Apphcahon To The NUREG/CR 6073: LYSIMETER LITERATURE REVIEW.

RHR/SSW Systems Of A BWR.

MCCORD.J.T.

M ANTE UFEL,R.D.

NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS-NUREG/CR-5817 V02: NRC HIGH LEVEL RADIOACTIVE WASTE RE.

MENT METHOOOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE SEARCH AT CNWRA Calendu Year 1991 DISPOSAL F ACILITIES Evaluahon Of Modehng Approaches NUREG/CR-5817 V03 N1 NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. January. June 1992.

MCDONA M,J.C.

NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/CR-5989 PERFORMANCE TESTING OF EXTREMITY DOSI-RESEARCH AT CNWRA. July-December 1992.

METFRS. Pk.OT TEST.

NUREG/CR 6021: A LITERATURE REVIEW OF COUPLED THERMAL.

HYDROLOGIC-MECHAN ICAL CHEMICAL PROCESSES PERTINENT MCDUFFIE,P.N.

TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA NURFG/CR-5884 V1 DRF: REVISED ANALYSES OF DECOMMISSION.

MOUNTAIN IN3 FOR THE REFERENCE PRESSURIZED WATER REACTOR NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-POWER STATION Effects Of Cerent Regulatory And Other Consider.

TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIAll.Y a.t.ons On The Financial Assurance.... Main Report Draft Report For SATURATED, FRACTURED POROUS MEDIUM Comment NUREG/CR-5884 V2 DAF: REVISED ANALYSES OF DECOMMISSION.

MARLER,J E.

ING FOR TF REFERENCE PRESSURIZED WATER REACTOR NUREG/CR 6084 VALUE-lMPACT ANALYSIS OF GE NERIC ISSUE 143' POWER STADJN Effects Of Current Regulatory And Other Consider.

" AVAILABILITY OF HEATING, VENTILATION, AIR CONDITIONING ations On 1he Financial Assurance.... Appendices Draft Report For (HVAC) AND CHILLED WATER SYSTEMS "

Comment M ARSCH ALL C.W.

MCFARLAND A.R NUREG/CR-4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING NUREG/G50006: DEPOSITION SOFTWARE TO CALCULATE PARTI-CLE PENETRATION THROUGH AEROSOL TRANSPORT NU EG/

4 9 3

S OR C ACKSI Pt AND PIP'NG SYMF MS. Final Report WELDS Semiannual Report, April September 1992.

NUREG/CR-6098. LOADING RATE EFFECTS ON STRENGTH AND A

FR U E TOUGHNESS OF P!PE STEELS USED IN TASK 1 OF Ug G 400. AIR SAMPLING IN THE WORKPLACE. Final Report MCKE.NNA,T).

MARTIN,0 E.

NUREG/CR 5247 V01 R1: RASCAL VERSION 2.0 USER'S GUIDE.

NUREG,1444: SITE DECOMMISSIONING MANAGEMENT PLAN.

NUREG/CR-5247 V02. RASCAL VERSION 2.0 WORKBOOK.

MARTIN,R.P.

MCKINNEY,M.D.

NUREG/CR-5882' TRAC-8 THERMAL-HYDRAULIC ANALYSIS OF THE NUREG/CR-5975: INCENTIVE REGULATION OF INVESTOR-OWNED BLACK FOX BOILING WATER REACTOR.

NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULATORS MARTINEZ GURIDI NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION MCN AM ARA,N.

AND SYSTEM UNAVAILABILfTY EVALUATION BASED ON FAILURE NUREG-0837 V12 N04. NRC TLD DIRECT RADIATION MONITORING DATA STATISTICS Summary Report.

NETWORK Progress Report. October-Docember 1992.

NUREG/CR5993 V02. METHODS FOR DEPENDENCY ESTIMATION NUREG 0837 V13 N01: NRC TLD DIRECT RADIATION MONITORING AND SYSTEM UNAVAILABILITY EVALUATION BASED ON F AILURE NETWORK. Progress Report January March 1993 NUREG 0837 V13 NO2: NRC TLD DIRECT RADIATION MONITORING DATA STATISTICS Detaded Descnpton And Apphcahons,

NURE G/CR 6111: INTEGRATED SYSTEMS ANALYSIS OF f HE PIUS NETWORK. Progress Report. Apnl4une 1993.

REACTOR.

MEND 12ABAL,R.

M ASLE NIKOV,0.R.

NUREG/lA-0107: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD NUREG/CR 4832 V00. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR REJECTION FROM 100% TO 50*4 POWER IN THE VANDELLOS 11 POWER PLANT; RISK METHODS INTEGRATION AND EVALUATION NUCLEAR POWER PLANT.

PROGRAM (RMEP) Seisme Analysis.

NUREG/lA 0108: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A TUR-NUREG/CR-6013 METHODS USED FOR THE TRE ATMENT OF NON-BiNE TRIP FROM 100% POWER IN THE VANDELLOS ll NUCLEAR PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

POWER PLANT.

4

Personal Author index 67 i

1 NUREG/lA 0109. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10%

NUREG/CR4834 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER VANDELLOS ll NUCLEAR POWFR PLANT.

PLANT.

NUREG/lA-0110: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A MAIN NUREG/CR 5835: AUXILIARY FCEDWATER SYSTEM RISK BASED IN-FEEDWATER TURBOPUMP TRIP TRANSIENT IN THE VANDE LLOS 11 SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NU.

NUCLEAR POWER PLANT.

CLEAR POWER PLANTS.

NUREG/CR-5836: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

MENSING,R.W'011. REVIEW Or STRUCTURE DAMPING VALUES FOR NUREG/CR 6 SPECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER PLANT.

ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PLANTS NUREG/CR-5897: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-G SPECTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR MERKLE.J[CR,5968. POTENTIAL CHANGE IN FLAW GEOMETRY OF AN NUREG NU EG 58 8 AUXILIARY FEEDWATER SYSTEM RISK-DASED IN-INITIALLY SHALLOW FINITE ifNGTH SURFACE FLAW DURING A PRESSURIZ[D.lHERMAL-SHOCK TRANSIENT.

SPECTION GUIDE FOR THE POINT BEACH NUCLEAR POWER PLANT.

M E Y E R,0.R.

ERA G EV NT 9 99 U

/CR 5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTR Y. Annual Summary Of Program Performance Reports.CY MEY E R.P.D.

1992-NUREG/CR 6114 V01: APPLICATION OF AN INFILTRATION EVALUA.

I N TF Y TO A HYPOTHETICAL LOW. LEVEL WASTE MOK,G NunEG/CR 6007. STRESS ANALYSIS OF CLOSURE BOLTS FOR SHIPPING CASKS.

MICHAUD,W.F.

NUREG/CR-4067 V16-ENVIRONMENTALLY ASSISTED CRACKING IN MONTELEONE,S.

LIGHT WATER REACTORS Swannual ReportOctobet 1992 - March NUREG/CP 0132. TRANSACTIONS OF THE TWENTY-FIRST WATER 1993.

REACTOR SAFETY INFORMATION MEETING.

MILICl,T, MOORE S.E.

NUREG/CR-6056 A FRAMEWORK FOR THE ASSESSMENT OF NUREG/CR 5358 REVIEW OF ASME CODE CRITERIA FOR CONTROL SEVERf ACCIDENT MANAGEMENT STRATEGIES OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM BRANCH CON-NECTIONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP-MILLE R.C. A.

MENT WORK.

NUREG/CR-5956 CONSIDERATION OF UNCERTAINTIES IN SOIL-STRUCTURE INTERACTION COMPUTAT6ONS.

MOR AN,8.W.

NUREG/CR-6118: ASSESSMENT OF THE EFFECTIVENESS OF THE MILLER.LA.

LEU RFFORM RULE AND ITS IMPL EMENTATION NUREG/CR 530$ V02 PI: INTEGRATED RISK ASSESSMENT FOR THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And MOR ANTE,R.

Risk Uncertainty Evaluation Program (PRUEP) Appendices A-C.

NUREG/CR-6111: INTEGRATED SYSTEMS ANALYSIS OF THE PIUS NUREG/CR 5305' V02 P2. INTEGRATED RISK ASSESSMENT FOR THE REACTOR LASALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And Risk Uncertainty E valuation Prooram (PRUEP) Appendices D.G MOR A Y,N.

NUREG/CR 6018. SURVEY ANET ASSESSMENI OF CONVENTIONAL NUREG/CR-5977: A PERFORMANCE INDICATOR OF THE EFFECTIVE.

SOFTWARE VEHiFICATION AND VALIDATION METHODS NESS OF HUMAN MACHINE INTERFACES FOR NUCLEAR POWER NUREG/CR 6059 MACCS VERSION 1.511.1. A MAINTENANCE RE.

PLANTS.

LEASE OF THE CODE.

MORCOS.N.

MILLE R.M.K*

NUREG/CR-5672 V03 CHARACTERISTICS OF LOW-LEVEL RADIOAC-NUREGICR 5914 CHEMICAL COMPOSITION AND RT(NOT) DETERMI.

NATIONS FOR MIDLAND WELD WF-70.

TiVE DECONTAMINATION WASTE. Annual Report for Fiscal Year ggg7 MILLER,S.P'R 58 3 MOR M P NURE /C F ISK ASSESSMENT OF ISOLATION DEVICES IN NUREG/th-0085: ASSESSMENT OF FULL POWER TURBINE TRIP START UP TEST FOR C. TRILLO I WITH RELAP5/ MOD 2 MIRSKY,S.M.

NUREG/lA-0119' ASSESSMENT AND APPLICATION OF BLACKOUT NUREG/CR-6018 SURVEY AND ASSESSMENT OF CONVENTIONAL TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/

SOFTWARE VERIFICATION AND VAllDATION METHODS NUREG/IA 0121: ASSESSMENT OF A PRESSURIZER SPRAY VALVE MITCHELL.D.D.

FAULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT NUREG/CR-5t163. RISK ASSESSMENT OF ISOLATION DEVICES IN WITH RELAP5/ MOD 2.

SAFETY SYSTEMS MITCHELLD.W.

NUREG/CR-6052 METHODOLOGY FOR RELIABILITY BASED CONDI-NUREG/CR 5851. LONG TERM PERFORMANCE AND AGING CHAR-TION ASSESSMENT. Application To Concrete Structures in Nuclear ACTERISTICS OF NUCLEAR PLANT PRESSURE TRANSMITTERS.

Plants.

NUREGICR-5903. VALIDATION OF SMART SENSOR TECHNOLOGIES FOR INSTRUMENT Call 8 RATION REDUCTION IN NUCLEAR HORTON,D.K.

POWER PLANTS NUREG/CR 6027. PRELIMINARY EVALUATION OF SNUBBER SINGLE FAILURES.

MOFFITT,N E.

NUREG/CR 5488. RISK-BASED INSPECTION GUIDE FOR THREE MILE MOSLEH,A.

ISLAND NUCLEAR STATION UNIT 1 NUREGrCR-5801; PROCEDURE FOR ANALYSIS OF COMMON-CAUSE NUREGICR 5766. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-FAILURES IN PROBABILISTIC SAFETY ANALYSIS.

SPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR POWER PLANT, MROCHEK,JL NUREGrCR 5829. AUXlLIARY F EEDWATER SYSTEM RISK-BASED IN-NUREG/CR-5938 NATIONAL PROFILE ON COMMERCIALLY GENER-SPECTION GUIDANCE FOR THE DAVIS-DESSE NUCLEAR POWER ATED LOW-LEVEL RADIOACTIVE MlXED WASTE.

PLANT.

NUREG/CR 5833 AUXlLIARY FEEDWATER SYSTEM RISK-BASED IN-MULLINS,A.

SPECTION GUIDE FOR THE H EL ROOiNSON NUCLEAR POWER NUREG-1476 FINAL ENVIRONMENTAL IMPACT STATEMENT TO PLANT CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE, AND

68 Personal Author index DISPOSE OF 11E (2)

OYPRODUCT MATERIAL NEAR NORTON.LJ.

CLIVE. UTAH Docket No. 40-8989, Enwocare Of UtahJnc.

NUREG 1415 V06 Noi: OFFICE OF THE INSPECTOR NUREG-1476 DRFT: DRAFT ENV!RONMENTAL IMPACT STATEMENT GENERALSerniannual Report.Aprd 1,1993 September 30.1993.

TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE, AND DISPOSE OF 11E(2) DYPRODUCT MATERIAL NEAR CLIVE, NOURBANHSH.H P.

UTAH.Dodet No 40 8989, Envrocare Of Utah. Inc.

NUREG/CR-5747; ESilMATE OF RADIONUCLIDE RELEASE CHARAC-TERISTICS INTO CONTAINMENT UNDER SEVERE ACCIDENT MURFIN,W.B.

CONDITIONS Final Report.

NUREG/CR 5360: XSOR CODES USERS MANUAL NUREG/CR 5982. EFFECTIVENESS OF CONTAINMENT SPRAYS IN CONTAINMENT MANAGEMENT.

MURPHY,W M.

NUREG/CR 5817 V02: NRC HiGH LEVEL RADIOACTIVE WASTE RE.

NOVO.M.

SEARCH AT CNWRA Calendar Year 1991.

NUREG/lA-0085 ASSESSMENT OF FULL POWER TURBINE TRIP NUREG/CR 56t7 V03 NI. NRC HIGH-LEVEL RADIOACTIVE WASTE START-UP TEST FOR C. TRILLO I WITH RELAPS/ MOD 2 RESEARCH AT CNWRA January-June 1992.

NUREG/CR-5817 V03 N2. NRC HIGH LEVEL RADIOACTIVE WASTE NUSSBAUMER,0.

RESEARCH AT CNWRA. July-Der, ember 1992.

NUREG/CR-5962: HEALTH AND SAFETY IMPACTS FROM DISCRETE SOURCES OF NATURALLY OCCURRING AND ACCELERATOR-PRO-NANSTAD.R.K-DUCED RADIOACTIVE MATERIALS (NARM)

NUREG/CR-5914. CHEMICAL COMPOSITION AND HT(NDT) DETERM1 NATIONS FOR MIDLAND WELD WF-70.

O'DOWD.N.P.

NUREG/CR 5972' EFFECTS OF NONSTANDARD HEAT TREATMENT NUREG/CR-5958.

1VIO-PARAMETER FRACTURE MECHANICS:

TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES THEORY AND APPLICATIONS.

OF CARBON STE EL CASTING REPAIR WELDS.

NUREG/CR 6036 INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL OKRENT,0.

LOADING ON FRACTURE TOUGHNESS.

NUMEG/CR-6056' A FRAMEWORK FOR THE ASSESSMENT OF SEVERE ACCIDENT MANAGEMENT STRATEGIES NATIONS,J.O.

NUREG/CR 6110 ASSESSMENT OF THE EFFECTIVENESS OF THE OLAGUE.N.E.

LEU REFORM RULE AND ITS IMPLEMENTATION NUREG/CR-5927 VOI: EVALUATION OF A PERFORMANCE ASSESS.

MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE N AUS.D.J.

DISPOSAL FACILITIES Evaluation Of Modeling Approaches NUREG/CR 6015. STRUCTURAL AGING PROGRAM T ECHNiCAL PROGRESS FOR PERIOD JANUARY - DECEMBER 1992.

OL AND,C.B.

NUREG/CR-6015: S TRUCTURAL AGING PROGRAM TECHNICAL NAV AS.A.P.

PROGRESS FOR PERIOD JANUARY - DECEMBER 1992.

NUREG/lA4119. ASSESSMENT AND APPLICATON OF BLACKOUT TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/

OUVE,K L NUREG 1350 V05. NUCLEAR REGULATORY COMMISSION INFORMA-MOD 2.

NUREG/lA4121. ASSESSMENT OF A PRESSURIZER SPRAY VALVE TION DIGEST.1993 Edition.

FAULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT OLSZEWSKI.M.

WITH RELAPS/ MOD 2 NUREG/CR 6065. SYSTEMS ANALYS!S OF THE CANDU 3 REACTOR.

NELSON.K, ORTIZ,M.G.

NUREG/CR-5883 HEALTH RISK ASSESSMENI OF tRRADIATED NUREG/CR-5818: UNCERTAINTY ANALYSIS OF MINIMUM VESSEL T OP AZ.

LIQUID INVENTORY DURING A SMALL-BREAK LOCA IN A B&W PLANT AN APPLICATION OF THE CSAU METHODOLOGY USING NGUY EN.B.

NUREG/CR-6079 SEISMOLOGICAL INVESTIGATION OF E ARTH-THE RELAP5/ MOD 3 COMPUTER CODE.

OUAKES IN THE NEW MADRID SEISMIC ZONE Final OSB RN L ReportSeptember 1988 December 1992 ATED LOW-LEVEL RADIOACTIVE MIXED WASTE.

NICHOLAUS.J.R.

NUREG/CR-5897: AUXILIARY FEEDWATER SYSTEM RISK BASED IN-PABA NRT, ECT ON GunDE FOR THE SOUTH TEXAS PROJECT NUCLEAR SEARCH AT CNWRA Calendar Year 1991.

NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOAC11VE WASTE NtCHOLSON,T.J.

PROCEEDINGS OF WORKSHOP V: ' FLOW AND NUREGICP-0040;HROUGH UNSATURATED FRACTURED ROCK - RE-NR CR 8 VO N H H LEVE RADIOACTIVE WASTE TRANSPORTT RESEARCH AT CNWRA July December 1992.

L ATED TO HIGH LEVEL RADIOACTIVE WASTE DISPOSAL. Held At I

Radisson Sute Hotel. Tucson. Anzona January 7-10.1991-PAINTER,C.L NUREG/CR 5988 SOfL CHARACTERIZATION MrTHODS FOR UN-NUREG/CR-3950 V08 FUEL PERFORMANCE ANNUAL REPORT FOR SATURATE D LOW 4EVEL WASTE SITES

3990, l

NICKOLAUS,J.R.

P ALOMAR.J.

NUREG/CR,5829. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR-6090 THE PROGRAMMABLE LOGIC CONTROLLER AND I

SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS.

PLANT.

NUREG/CR 5973 CODES AND STANDARDS AND OTHE9 GUIDANCE PARADIES.M.

CITED IN REGULATORY DOCUMENTS.

NUREG/CR 5455 V01: DEVELOPMENT OF THE NRC'S HUMAN PER-FORMANCE INVESTIGATION PROCESS (HPIP).

NILSSON,L NUREG/CR-5455 V02: DEVELOPMENT OF THE NRC'S HUMAN PER-NUREG/tA 0094 ASSESSMENT OF RELAP5/ MOD 3 AGAINST FORMANCE INVESTIGATION PROCESS (HPIP).

TWENTY-FIVE POST-DRYOUT EXPERIMENTS PERFORMED AT THE NUREG/CR-5455 V03: DEVELOPMENT OF THE NRC'S HUMAN PER-ROYAL INSTITUTE OF TECHNOLOGY.

FORMANCE INVESTIGATION PROCESS (HPIP).

NITZEL.M.E.

P ARK,C.K.

NUREG/CR 6027; PREllMINARY EV ALUATION OF SNUBBER SINGLE NUREG/CR-4551 V7R1P1 EVALUATION OF SEVERE ACCIDENT l

F ALLURES RISKS: ZION UNIT 1 Main Report.

NUREGICR 4551V7AIP2A: EVALUATION OF SEVERE ACCIDENT NO,H C.

RISKS ZION UNIT 1 Appendu A.

NUREG/GR 0009 STEPWISE INTEGRAL SCALING METHOD AND ITS NUREG/CR-4551V7R1P2B EVALUATION OF SEVERE ACCIDENT APPLICATON TO SEVERE ACCIDENT PHENOMENA RISKS ZION UNIT 1. Appendices B, C, D, And E,

Personal Author index 69 PARK.H.

PETERSEN,K.M.

NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF NUREG/CR-5851: LONG TERM PERFORMANCE AND AGING CHAR-SEVERE ACCIDENT MANAGEMENT STRATEGIES.

ACTERISTICS OF NUCLEAR PLANT PRESSURE TRANSMITTERS.

NUREG/CR-5903: VALIDATION OF SMART SENSOR TECHNOLOGIES NURE /CR-4687 V15: ENVIRONlwENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS Serniennual Report E P TS^

NUREG/CR 4867 V16: ENVIRONMENTALLY AS$,Apnt-September 1992.

1STED CHACKING IN l,

T WATER REACTORS. Somiannual Report,0ctober 1992 - Maren PETERSON,Ci l

NUREG/CR 035: FEASIBILITY STUDY FOR IMPROVED STEADY-STATE INITIALIZATION ALGORITHMS FOR THE RELAPS COMPUT.

l PARK Y.J.

ER CODE.

NUREG/CR-6078 ANALYSIS OF CHACK INITIATION AND GROWTH IN THE HIGH LEVEL VlBRATION TEST AT TADOTSU.

PODOWSKI,M.Z.

NUREG/CR4025: THE PROBABILITY OF MARK-1 CONTAINMENT PARKER.G.

FAILURE BY MELT-ATTACK OF THE LINER NOREG 1479: RESULTS FROM TWO WORKSHOPS: STATE RADb ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG-POWERS,0.A-ULATIONS AND FUNDING.

NUREG/CR-5901: A SIMPLIFIED MODEL OF AEROSOL SCRUBBING BY A WATER POOL OVERLYING CORE DEBRIS INTERACTING PARRY,G.W' 4471: ENHANCEMENTS TO DATA COLLECTION AND RE.

NUREG/CR NC E

l PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

NUR ING WATER POOLS.The WETCOR-1 Test f

PATTERSON.M.

NUREG/CR-5966: A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY NUREG/CP 0129. PROCEEDINGS OF THE WORKSHOP ON PROGRAM CONTAINMENT SPRAYS.

FOR ELIMINATION OF REQUIREMENTS MARGINAL TO SAFETY.

NUREG/CR 5978: SOURCE TERM ATTENUATION BY WATER IN THE MARK i BOILING WATER REACTOR DRYWELL PAULA,H.M.

NUREG/CR-6025: THE PROBABILITY OF MARK-l CONTAINMENT NUREG/CR-5471 ENHANCEMENTS TO DATA COLLECTION AND RE-FAILURE BY MELT-ATTACK OF THE LINER.

PORTING OF $1NGLE AND MULTIPLE FAILURE EVENTS.

PRATT,W.T.

NU G 6035: FEASIBILITY STUDY FOR IMPROVED STEADY ~

RSKS N1 1 a R STATE INITIAll2ATION ALGORITHMS FOR THE RELAP5 COMPUT.

ER CODE.

NUREG/CR 4551V7R1P2A: EVALUATION OF SEVERE ACCIDENT RISKS. ZION UNIT 1. Appendix A.

I PAYNE,A.C.

NUREG/CR-4551V7R1P2B: EVALUATION OF SEVERE ACCIDENT NUREG/CR 4832 V05. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR RISKS. ZION UNIT 1. Appendices B C, D. And E.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION P

RA arameter Estimation Analysis And Screening Human Reh-NURE [CR-3950 V08: FUEL PERFORMANCE ANNUAL REPORT FOR NURE /CR-4832 V09: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR 1990.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Internal Fue Analysis.

PRECKSHOT,G.G.

NUREG/CR-5305 V02 PI: INTEGRATED HISK ASSESSMENT FOR THE NUREG/CR-6082: DATA COMMUNICATIONS.

LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NUREG/CR-6083: REVIEWING REAL-TIME PERFORMANCE OF NU.

Risk Uncertainty Evaluation Program PRUEP) Appendices A C.

CLEAR REACTOR SAFETY SYSTEMS.

NUREG/CR-5305 V02 P2-INTEGRATE (D RISK ASSESSMENT FOR THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And PREECE.R.J.

Risk Uncertainty Evaluation Program (PRUEP) Appendices D-G NUREG/lA-0106: ASSESSMENT OF PWR STEAM GENERATOR MOD-ELLING IN RELAPS/ MOD 2.

PAYNE,G.A.

NUREG/lA-0113: PRELIMINARY ASSESSMENT OF PWR STEAM GEN-NUREG/CR-3950 V08 FUEL PERFORMANCE ANNUAL REPORT Fort ERATOR MODELLING IN RELAPS/ MOD 3.

1990.

PRIKR YL,J.D.

NUREG

-5817 V02: NRC HIGH LEVEL RADIOACTIVE WASTE RE-SE A A CN A Ca endar Ye 991.

NU E /

58 1 03

F G L /EL RADIOACTIVE WASTE NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA January-June 1992 RESEARCH AT CNWRA. January 4une 1992, NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. July December 1992.

E p

p PENNELL,W.E.

ISTS* MEETING ON FRACTURE MECHANICS VERIFICATION BY NUREG/CR-4219 V09 N2. HEAVY SECTION STEEL TECHNOLOGY LARGE-SCALE TESTING. Held At Pollard Auditorium.0ak PROGRAM Semiannual Progress Report For Apr$ September 1992.

Ridge, Tennessee.

NUREG/CR-6038: INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL NUREG/CR 5997: CSNI PROJECT FOR FRACTURE ANALYSES OF LOADING ON FRACTURE TOUGHNESS LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS PE REZ,J, NUREG/lA-0107: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD PUGH,R.

REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS Il NUREG/CR-5766. AUXILIARY FEEDWATER SYSTEM RISK BASED IN- '

NU

/t 01 SE S NT OF RELAPS/ MOD 2 AGAINST A TUR-SPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR BlNE TRIP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

NU E / A ASSESSMENT OF RELAP5/ MOD 2 AGAINST A 10%

HAAS.

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE NUREG/CR-4832 V08: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR VANDELLOS 11 NUCLEAR POWER PLANT.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION NUREG/lA-0110: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A MAIN PROGRAM (RMIEP).Seisrnic Analysis.

FEEDWATER TURBOPUMP TRIP TRANS!ENT IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

PUTNEY,J.M.

NUREG/lA 0106: ASSESSMENT OF PWR STEAM GENERATOR MOD-f PEREZ,S.E.

ELLING IN RELAP5/ MOD 2.

l NUREG/CR-5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN NUREG/lA-0113: PRELIMINARY ASSESSMENT OF PWR STEAM GEN-j CONTAINMENT MANAGEMENT ERATOR MODELLING IN RELAP5/ MOD 3.

l e

70 Personal Author Index RODABAUGH,E.C.

RADDATZ,C.T.

NUREG 0713 V12. OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/C4 5358: REVIEW OF ASME CODE CRITERIA FOR CONTROL VERCIAL NUCLEAR POWER REACTORS AND OTHER OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM BRANCH CON-NECTIONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP.

F ACluTIES.1990 Twenty. Third Annual Report.

NUREG-0713 V13. OCCUPATIONAL RADIATION EXPOSURE AT COM-MENT WORK.

MERCIAL NUCLEAR POWER REACTORS AND OTHER F ACILITIES.1991. Twenty. Fourth Annual Report ROGER 31,R.D.

NUREG 0713 V14: OCCUPATIONAL RADIATION EXPOSURE AT COM.

NUREG/CR-5229 VOS; FIELD LYSIMETER INVESTIGATIONS: LOW.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR 1992 Twenty-Fifth Annual Report FISCAL YEAR 1992. Annual Report.

NUREG/CR-5987: MICROBIAL-INFLUENCED CEMENT DEGRADAllON RAHMAN,$.

- LITERATURE REVIEW.

NUREG/CR-4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING NUREG/CR-6073: LYSIMETER UTERATURE REVIEW.

WEL DS Semiannual Report, October 1991 - March 1992.

NUREG/CR.4*>99 V03 N1: SHORT CRACKS IN PlPING AND PIPING ROLLSTIN,J.

WELDS Semiannual Report, Apnl September 1992.

NUREG/CR.6059 MACCS VERSION 1.5.11.1: A MAINTENANCE RE-RAMSDELLJ.V.

NUREG/CR 5247 V01 RI: RASCAL VERSION 2 0 USER'S GUIDE.

ROSS,$.

NUREG/CR-5791: RISK EVALUATION FOR A GENERAL ELECTRIC RAO M C BWR. EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON NUREd/CR 6036-INITIAL RESULTS OF THE INFLUENCE OF BIAxlAL SAFETY-RELATED EQUIPMENT. Evaluation Of Genenc issue 57.

LOADING ON FRACTURE TOUGHNESS ROSSBACH,LW.

R AO.R.R.

NUREG/CR-5835: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR4927 V01 EVALUATION OF A PERFORMANCE ASSESS.

SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NU-MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE CLEAR POWER PLANTS.

DISPOSAL F ACILITIES Evaluation Of Motieling Approaches.

ROURK,C.J.

R ASHID,Y.R.

NUREG/CR-6025: THE PROBABluTY OF MARK-1 CONTAINMENT NUREG-1453 REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 142; LEAKAGE THROUGH ELECTRICAL ISOLA-FAfLURE BY MELT. ATTACK OF THE UNER.

TORS IN INSTRUMENTATION CIRCUlTS.

RASMUSON.D.M.

NUREG/CR-5471. ENHANCEMENTS TO DAT A COLLECTION AND RE-RUNCHAL.A.K.

PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS NUREG/CR-5991: PORF LOW: A MULTIFLUID MULTIPHASE MODEL NUREG/CR-5964 SAPHiRE TECHNICAL REFERENCE FOR SIMULATING FLOW, HEAT TRANSFER, AND MASS TRANS-MANUAL:lRRAS/SAR.A VERSION 4 0.

PORT IN FRACTURED POROUS MEDIA. User's Manual - Version 2.41.

RASMUSSEN J.

NUREG/CR-5977; A PERFORMANCE INDICATOR OF THE EFFECTIVE-RUSSELL,K.D.

NESS OF HUMAN. MACHINE INTERFACES FOR NUCLEAR POWER NUREG/CR-5964.

SAPHIRE TECHNICAL REFERENCE PLANTS.

MANUAL:lRRAS/ SARA VERSION 4.0.

R AVINDRA.M.K.

RUTHER.W.E.

NUREG/CR-4B32 V08 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR 4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION UGHT WATER REACTORS. Semiannual Report.ApriLSeptember 1992.

PROGRAM (RMIEFrSeismic Analys s NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN gf' ^

REBOLLO,L NUREGilA 0124 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRE S-SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN-SAGAR,B.

SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA' NUREG/CR-5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE-BRERA NUCLEAR STATION, SEARCH AT CNWRA. Calendar Year 1991.

NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE WC.L RESEARCH AT CNWRA. January-June 1992 NUREG/CR 6117. NEUTRON SPECTRA AT DIFFERENT H6GH FLUX NUREG/CR-5917 V01: SENSITIVITY AND UNCERTAINTY ANALYSES ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEILLANCE APPLIED TO ONE-DIMENSIONAL RADIONUCUDE TRANSPORT IN A LOCATIONS, LAYERED FRACTURED ROCK MULTFRAC Analytic Solutions And Local Sensstivities.

REMPEJ L NUREG/CR.5917 V02: SENSITIVITY AND UNCERTAINTY ANALYSES NUREd/CR-5642: UGHT WATER REACTOR LOWER HEAD F AILURE APPLIED TO ONE-DIMENSIONAL RADIONUCUDE TRANSPORT IN A ANALYSIS' LAYERED FRACTURED ROCK. Evaluation Of The Umit State Ap-proach.

REVENTOS.F.

NUREG/CR 5991: PORFLOW: A MULTIFLUID MULTIPHASE MODEL -

NUREG/lA-0119. ASSESSMENT AND APPLICATION OF BLACKOUT FCG SIMULATING FLOW, HEAT TRANSFER, AND MASS TRANS-TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

PORT IN FRACTURED POROUS MEDIA. User's Manual - Version MOD 2 NUREG/lA-0121: ASSESSMENT OF A PRESSURIZER SPRAY VALVE 2.41.

FAULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT SAMANTA.P.K.

WITH RELAPS/ MOD 2 NUhEG/CR-5995; TECHNICAL SPECIFICATION ACTION STATEMENTS REOuiRING SHUTDOWN.A Risk Perspective With Apphcation To The ROBERTSON,D.E.

NUREG/CR-5894. RADIONUCUDE CHARACTER!ZATION OF REAC.

RHR/SSW Systems Of A BWR, TOR DECOMM)S$10NING WASTE AND NEUTRON. ACTIVATED SANECKI,J.E.

MET ALS_

NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS. Semiannual Report. April-September 1992.

ROCHE.M.F.

NUREG/C44667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-603? SOLIDUS AND LIQUIDUS TEMPERATURES OF UGHT WATER REACTORS. Semiannual Report,0ctober 1992. March CORE CONCRETE MIXTURES.

1993.

ROCKHOLD.M.L NUREGICR 5998 SIMULATION OF UNSATURA1ED FLOW AND NON-SANTAMARIA,J.G.

REACTIVE SOLUTE TRANSPORT IN A HETEROGENEOUS SOIL AT NUREG/lA-0085: ASSESSMENT OF FULL POWER TURBINE TRIP START-UP TEST FOR C. TRILLO i WITH RELAP5/ MOD 2.

THE FIELD SCALE

i Personal Author index 71 SQTTISON,M.S.

SHIE R,W.

NUREG/CR-5759. RISK ANALYSIS OF HIGHLY COMDUSTIBLE GAS NUREG/CR 5933: HIGH PRESSURE COOLANT INJECTION (HPCI)

STORACE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-SYSTEM RISK-BASED INSPECTION GUIDE FOR DRESDEN NUCLE-12ED WATER REACTOR PLANTS AR POWER STATION UNITS 2 AND 3.

NUREG/CR 5964 SAPHIRE TECHNICAL REFERENCE NUREG/CR-6111 INTEGRATED SYSTEMS ANALYSIS OF THE PIUS MANUAL IRRAS/ SARA VERSION 4.0.

REACTOR.

SCHROEDER.J.A.

SHIH.C.F.

NUREG/CR4928 ISLOCA RESEARCH PROGRAM Final Report NUREG/CR-5958 TWO-PARAMETER FRACTURE MECHANICS:

THEORY AND APPLICATIONS.

SCHULTZ,R.R.

NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT NUREG/CH 606): DETEHMINATION OF THE BIAS IN LOFT FUEL OF CONSTRAINT IN F RACTURE, PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN PHASE OF LARGE-OREAk LOCA EXPEnlMENTS.

SHIVER,A.W.

NUREG/lA-0128 INTERNATIONAL CODE ASSESSMENT AND APPLI-NUREG/CR-5305 V02 P1 INTEGRATED RISK ASSESSMENT FOR THE CATIONS PROGRAM

SUMMARY

OF CODE ASSESSMENT STUDIES LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And CONCERNING HELAPS/ MOD 2, RELAP5/ MOD 3, AND TRAC.B Rish Uncertainty Evaluation Program (PRUEP) Appendices A-C.

NUREG/CR 5305 V02 P2. INTEGRATED RISK ASSESSMENT FOR THE SCHUL 2,H.

LASALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALW S Of Risk Uncertainty Evaluabon Program (PRUEP) Appendices D-G LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS (PROJECT FALSIRE)

SHTEYNGAFIT,S.

NUREG/CR-6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN SCHULZ,R.K.

THE HIGH LEVEL VfDRATION TEST AT TADOTSU.

NUREG/CR 5980: THREE DIMENSIONAL REDISTRIBUTION OF TRITI-UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED SHUM,D.K.

SOIL. A Determinishc Modet for Tntium M grat on in An And Dmposai NUREG/CR 5782. PRESSURIZED THERMAL SHOCK PROBABILISTIC Site FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CR 6108 SPHERICAL DIFFUSION OF TRITIUM FROM A ROWE REACTOR PRESSURE VESSEL.

POINT OF RELEASE IN A UNIFORM UNSATURATED SOILA Deter.

NUREG/CR-5968 POTENTIAL CHANGE IN FLAW GEOMETRY OF AN mirushe Model For inbum Migrahon in An And Dmposal Sita INITIALLY SHALLOW FINITE-LENGTH SURFACE FLAW DURING A PRESSURIZED THERMAL SHOCK TRANSIENT.

SCOTT.B FL NUREG/CR4214 RIP 2A2. HEALTH EFFECTS MODELS FOR NUCLE.

SIBOL,M.S.

AR POWER PLANT ACCIDENT CONSEQUENCE NUREG/CR4058. VIRGINIA REGIONAL SEISMIC NETWORKJinal ANALYSIS Moddicahon Of Mcdois Resulting From Addshon Of Ettects Report (1988 1992)

Of Exposure To Arpha Enutt ng Radionuchdes Part II. Scenhhc Bases For Health._

SIDOROV,V.P.

NUREG/CR4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR NUREG/CR4072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part 1.

HAVIOR OF HYDROGEN AIR MIXTURES WlIH TURBULENT JET IG-Introduction,Integrat on,And Summary NITION AT LARGE SCALE.

SCOTT,P.

SIENICKI,J.J.

NUREG/CR4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING NUREG/CR-5642: LIGHT WATER REACTOR LOWFR HEAD FAILURE WELDS Semiannual Report. October 1991 - March 1992.

ANALYSIS 1

NUREG/CR4599 V03 N1: SHORT CRACKS IN PIPING AND PIPING NUREG/CR4025-THE PROBADILITY OF MARX l CONTAINMENT l

WELDS Semiannual Report, Apnl September 1992.

FAILURE BY MELT ATT ACK OF THE LINER.

SEESER,L SIEVERS,J.

NUREGICR-5778 V03 NEW YORK /NEW JERSEY REGIONAL SElSMIC NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF NETWORK Final Report f or Apnl 1985 - September 1992 LARGE-SCALE INTE RNATIONAL REFERENCE EXPERIMENTS (PROJECT FALSIRE)

SEITZ R.R.

NUREG/CR 6070 MODELING APPROACHES FOR CONCRETE BAR.

SlKOV,MA AlERS USED IN LOW-LEVEL WASTE DISPOSAL NUREG/CR 5631 R1 ADO: CONTRIBUTION OF MATERNAL RADIONU-CLIDE BURDENS TO PRENATAL RADIATION DOSES Relabonships SE N A.P.P.

Detween Annual Limsts On intake And Frenatal Dosos.

NUREG/CR 5835 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN SPECTION GUIDE FOR THE DEAVER VALLEY UNITS 1 AND 2 NU-SIMlON,G.P.

CLEAR POWER PLANTS NUREG/CR 5759 RISK ANALYSIS OF HIGHLY COMBUSilRLE GAS STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-SHA W.T.

IZED WATER REACTOR PLANTS.

NUREG/CR 5822, ANALYSIS OF THERMAL MIXING AND BORON Dl-LUTION IN A PWR.

SIMONEN.F.A.

NUREG/CR 4469 V15. NONDESTRUCTIVE EXAMINATION (NDE) REU-i SH AC K.W.J.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WA TE R

]

NUREG/CRJ687 V15. ENVIRONMENTALLY AS$1STED CRACKING IN REACTORS Semiannual Report October 1991. March 1992 UGHT WATER REACTORS Semenual Roport. April-September 1992-NUREG/CR-4469 V16: NONDESTRUCTIVE EXAMINATION (NDE) REU-j NUREG/CH4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN ADlLITY FOR INSERVICE INSPECTION OF LIGHT WATER l

LIGHT WATER REACTORS. Semiannual Report October 1992 - March REAClORS Semiannual Report, April 1992. September 1992 1993 NUREG/CR 5999. INTERIM FATIOUE DESIGN CURVES FOR CARBON, SIMONS,J.W.

LOW ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVl-NUREG/lA-0126 20/3D PROGRAM WORK

SUMMARY

REPORT.

RONMENTS NUREG/lA-0127. REACTOR SAFETY ISSUES RESOLVED BY THE 2D/

3D PROGRAM.

SHELL,C.S.

NUREG/CR.5903 VALIDATION OF SMART SENSOR TECHNOLOGIES SIMPSON,R.B.

FOR INSTRUMENT CAllBRATION REDUCTION IN NUCLEAR NUREG/CR.5907: CORE 40NCRETE INTERACTIONS WITH OVERLY.

POWER PLANTS ING WATER POOLS.The WETCOR 1 Test SHERFEY,LL SISSON.J.B.

NUREGrCR5973 CODES AND STANDARDS AND OTHER GUIDANCE NUREG/CR 5996. SUB$i>RFACE INJECTION OF RADIOACTIVE CITED IN REGULATORY DOCUMENTS TRACERS Field Expenment For Model Validabon Teshng i

3.,

.. ~.. -.

.c-.

~~.,-.,,y

..m

-r-v

. - - -i

72 Personal Author index SJOREEN,A.L STEINKE,W.F.

NUREG/CR-5247 V01 R1: RASCAL VERSION 2.0 USER'S GUIDE.

NUREG/CR-5953. STUDIES OF HUMAN PERFORMANCE DURING OP-ERATING EVENTS.1990-1992.

NUREG/CR 5836. AUXILIARY FEEDWATER SYSTEM RISK-DASED IN-STICKLER,LA.

SPECTION GUIDE FOR THE PALO VERDE NilCLEAR POWER NUREG/CA.5642. LIGHT WATER REACTOR LOWER HEAD FAILURE PLANT.

ANALYSIS.

SMILES,0.E.

SilRE WALT,0.L NUREG/CR-5980: THREE DIMENSIONAL REDISTRIBUTION OF 1RITI-NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED RESEARCH AT CNWRA. January-June 1992 l

SOILA Determmiste Model For Tntium Migration in An Ard Disposal NUREG/CR 5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE Site-RESEARCH AT CNWRA. July-December 1992.

NUREG/CR4108 SPHERICAL DIFFUSION OF TRITIUM FROM A POINT OF RELEASE IN A UNIFORM UNSATURATED SOll.A Deter-STOETZEL,G.A.

ministe Model For Tntium Migration in An Arid Disposal Sete.

NUREG-1400: AIR SAMPLING IN THE WORKPLACE.Finst Report I

"O STOOP,P.M NUR'EG' /CR-5759 RISK ANALYSIS OF HIGHLY COMBUSilBLE GAS NUREG/lA 0112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN-RE-STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-FLOOD EXPERIMENTS.

IZED WATER RE ACTOR PLANTS.

NUREG/CR 5976: DEVELOPMENT AND USE OF A TRAIN-LEVEL STOWE"G/CR.6047; R.A PRODADILISTIC RISK ASSESSMENT' NURE CONTINUOUS SPECTROSCOPIC ANALYSIS OF SMITH,R.I.

VANADOUS AND V ANADiC IONS.

NUREG/CR-5884 VI DRF; REVISED ANALYSES OF DECOMMISSION-8 ING FOR THE REFERENCE PRESSURIZED WATER REACTOR NURE'G-1400: AIR SAMPUNG IN THE WORKPLACE Final Report POWER STATION. Effects Of Current Regulatory And Other Consdor-on The Financial Assurance _Mam Report Draft Report For STRUChMEYER,R.

NUPLG 0837 V12 N04. NRC TLD DIRECT RADIATION MONITORING NUREG/CR5684 V2 DRF: REVISED ANALYSES OF DECOMMISSION.

ING FOR THE REFERENCE PRESSURIZED WATER REACTOR NL*WORKProgress Report October-Occomber 1992.

NUREG4837 V13 N01: NRC TLD DIRECT RADIATION MONITORING POWER STATION Effects Of Curront Regulatory And Other Consider.

NETWOFKProgress Report January. March 1993.

ations On The Financial Assurance.. Appendices Draft Report For NUREG-0837 V13 NO2: NRC TLD DIRECT HADIATION MONITORING Comment NETWORK. Progress Report April-June 1993.

SNOKE,J.A.

NUREG-0837 V13 NO3: NRC TLD DIRECT RADIATION MONITORING NUREG/CR 6058 VIRGINIA REGIONAL SEISMIC NETWORK Final NETWORK Progress Report July September 1993.

Report (1986 -1992).

SNOW.S.D.

NUREG 1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG/CR-5642 LIGHT WATER REACTOR LOWER HEAD FAILURE GENERIC ISSUE 153; LOSS OF ESSENTIAL SERVICE WATER IN ANALYSIS.

LWRS.

SODEL,P.

SUBUDHI,M.

NUREG-1488 DRFT FC; REVISED LIVERMORE SELSMiC HAZARD ES-NUREG/CR 5844: AGING ASSESSMENT OF BISTABLES AND TlMATES FOR 69 NUCLEAR POWER PLANT SITES EAST OF THE SWITCHES IN NUCLEAR POWER PLANTS ROCKY MOUNTAINS.Oraft Report For Comment SOO.P.

NUREG/CR-5943. SENSITIVITY ANALYSIS AND BENCHMARKING OF NUREG/CR-6111: INTEGRATED SYSTEMS ANALYSIS OF THE PlUS THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

REACTOR.

SUE S,R.H.

SPANNE R,J.C.

NUREG/CR-4832 V08: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR-4469 V15-NONDESTRUCTIVE EXAMINATION (NDE) REU-POWER PLANT; RISK METHOOS INTEGRATION AND EVALUATION ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER PROGRAM (RMIEP) Seismic Analysis.

REACTORS Som annual ReportOctober 1991 March 1992.

NUREG/CR-4469 V16. NONDESTRUCTIVE EXAMINATION (NDE) REll-SULLIVAN,S.G.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-REACTORS Semaannual Report.Aprd 1992-September 1992.

CLEAR POWER PLANTS: ANNOTATED BIBUOGRAPHY OF SELECT-SPENCER.B.W.

i NUREG/C46025-THE PROBABluTY OF MARK.I CONTA!NMENT SULLIVAN,T.M.

l FAILURE BY MELT-ATTACK OF THk; UNER NUREGICR-5911: SOURCE TERM EVALUATION FOR RADIOACTIVE l

LOW-LEVEL WASTE DISPOSAL PERFORMANCE ASSESSMENT.

i SPRUNG,J.L NUREG/CR-5943 SENSITIVITY ANALYSIS AND BENCHMARKING OF NUREG/CR-5901: A SIMPUFfED MODEL OF AEROSOL SCRUBB;NG THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

BY A WATER POOL OVERLYING CORE DEBRIS INTERACTING NUREG/CR-6041; DISPOSAL UNIT SOURCE TERM (DUST) DATA WITH CONCRETE. Final Report INPUT GUIDE.

SRIDHAR,N.

^

^

NU G/CR-5822: ANALYSIS OF THERMAL MIXING AND BORON DI-SE R A N R endar 91.

NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE LUTION IN A PWR.

RESE ARCH AT CNWRA January-June 1992.

NUREG/CR-5817 V03 N2 NRC HIGH-LEVEL RADIOACTIVE WASTE SURMEIER,J.

RESEARCH AT CNWRA.JuYDecember 1992.

NUREG-1476; FINAL ENVIRONMENTAL IMPACT STATEMENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, AND STE ELE,L.K.

DISPOSE OF 11E (2)

BYFRODUCT MATERIAL NEAR NUREG/CR-6011: REVIEW OF STRUCTURE DAMPING VALUES FOR CUVE. UTAH Docket No. 40-8989. Envirocare Of Utah.Inc.

ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PLANTS.

NvREG 1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT NUREG/CR 6012 STIFFNESS AND DAMPlNG PROPERTIES OF A TO CONSTRUCT AND OPERATE A FACILITY TO RECElVE, STORE, LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD.

AND DISPOSE OF 11EJ2) BfPRODUCT MATERIAL NEAR CUVE, ED EARTHOUAKE RESPONSES.

UTAH Docket No.40-8989 Envirocare Of Utah, Inc.

l l

Personal Author index 73 SVEDE M AN,S.J.

NUREG/CR-6012: STIFFNESS AND DAMPING PROPERTIES OF A NUREG/CR 4026 THEORETICAL AND EXPERIMENTAL INVESTIGA-LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY ED EARTHOUAKE RESPONSES.

SATURATED, FRACTURED POROUS MEDIUM.

TJADER T.R.

SW AIN.A.D.

NUREG-1366: IMPROVEMENTS TO TECHNICAL SPECIFICATIONS NUREG/CR-4832 V05 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR SURVEILLANCE REQUIREMENTS.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM Parameter Estimation Analysis And Screening Human Reh-TODD,M.D.

ability Analysis NUREG/CR-5944 A CHARACTERIZATION OF CHECK VALVE DEGRA-SWAIN,R.L DATlON AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-DUSTRY.

NUREG/C45914. CHEMICAL COMPOSITON AND RT(NOT) DETERMI-NATIONS FOR MIDLAND WELD WF TR AVIS.J.R.

SWIDER J.

NUREG/CR-6060: HYDROGEN MNNG STUDIES (HMS) ASSESSMENT MANUAL' NUREG/CR4056. A FRAMEWORK FOR THE ASSE,SSMENT OF SEVERE ACCIDENT MANAGEMENT S9ATEGIES TRAVIS.R.

t SWINDEM AN,M.J.

NUREG/CR-5934-HIGH PRESSURE COOLANT INJECTION (HPCI)

NUREG/CR-5972: EFFECTS OF NONSTANDARD HEAT TREATMENT SYSTEM RISK-BASED INSPECTION GUIDE FCR OUAD-CITIES TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES OF CARBON. STEEL CASTING REPAIR WELDS.

NURE /C 5 9 H GH PRESSURE COOLANT INJECTION (HPCI)

SYSTEM RISK-BASED INSPECTION GUIDE FOR ENRICO FERMI SWINDE M AN,R.W.

ATOMIC POWER PLANT, UNIT 2.

NUREG/CR-5955: MATERIALS AND DESIGN BASES ISSUES IN ASME CODE CASE N-47.

TSO'K'S' NUREG/CR-6113. CLASS 1E DIGITAL SYSTEMS STUDIES.

gyp NUREG/CR 5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE N

U G R-6031; CAVITATION GUIDE FOR CONTROL VALVES.

LASALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And Rmk Uncertainty Evaluaton Program (PRUEP) Appen6ces A-C NURE G'CR-5305 V02 P2. INTEGRATED RISK A$$ESSMENT FOR THE TURNER D.R LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NUREG/CR 5817 V02: NRC HIGH LEVEL RADIOACTIVE WASTE RE-Risk Unem1ainty Evaluahon Program (PRUEP) Appendices D-G

^

NUREG CR 58 03 L EL RADIOACTIVE WASTE TAl,A.T[G/CR 6113. CLASS 1E DIGITAL SYSTEMS STUDIES RESEARCH AT CNWRA. January-June 1992.

NUR NUREG/CR.5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. July-December 1992-TAM,P.S.

NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL.

NUREG-0847 S12: SAFETY EVALUATION REPORT RELATED TO THE HYDROLOGIC-M9CHAN ICAL 4HEMICAL PROCESSES PERT.NENT OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA 2 Docket Nos 50-390 And 50-391 (Tennesoe Valley Authonty)

MOUNTAIN TAYLOR,T.T.

UNGER,L NUREG/CR 4469 VIS: NONDESTRUCTIVE EXAMINATION (NDE) REL)

NUREG/CR-5455 V01: DEVELOPMENT OF THE NRC'S HUMAN PER-ABluTY FOR INSERVICE INSPECTION OF LIGHT WATER FORMANCE INVESTIGATON PROCESS (HPIP)

REACTORS Semiannual Report,0ctober 1991 - March 1992.

NUREG/CR-5455 V02: DEVELOPMENT OF THE NRC'S HUMAN PER-NUREG/CR-5410 STATISTICALLY BASED REEVALUATION OF PISC-Il FORMANCE INVESTIGATION PROCESS (HPIP).

ROUND ROBIN TEST DATA.

NUREG/CR-5455 V03: DEVELOPMENT OF THE NRC'S HUMAN PER.

FORMANCE INVESTIGATION PROCESS (HPIP).

NUREG/CR-6036: INITIAL RESULTS OF THE INFLUENCE OF BtAXIAL UP ADHYA YA,B.R.

LOADING ON FRACTURE TOUGHNESS NUREG/GR 0010: HYBR3D DIGITAL SIGNAL PROCESSING AND THEOF ANOUS,T.G.

NEURAL NETWORKS FOR AUTOMATED DIAGNOSTICS USING NDE METHODS.

NUREG/CR 5951: THE MANAGEMENT OF ATWS BY BORON INJEC.

TION.

VAN TUYLE,G-NUREG/CR-6025' THE PROBABILITY OF MAR K-1 CONTAINMENT NUREG/CR 6111: INTEGRATED SYSTEMS ANALYSIS OF THF PtUS FAILURE BY MELT. ATTACK OF THE LINER REACTOR.

THINNES,0.L V ANDE BOGAARD.J.

NUREG/CR-5642-LIGHT WATER REACTOR LOWER HEAD F AILURE ANALYSIS NUREG/lA-0112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN RE-FLOOD EXPERIMENTS.

THOMAS.C.W.

VANHORN,R.L NUREG/CR-5894: RADIONUCLIDE CHARACTER!ZATION OF REAC-NUREG/CR-5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS TOR DECOMMISSIONING WASTE ANC NEUTRON'ACTIVAT E D Al8' STORAGE, SUPPLY, AND DISTRIBUTON SYSTEMS IN PRESSUR-IZED WATER REACTOR PLANTS TICHLER,J N CLE R P E PLAN S An alRe NURE CR 6080: REPLACEMENT ENERGY, CAPACITY, AND RELi-ABluTY CCSTS FOR PERMANENT NUCLEAR REACTOR SHUT-TINGLE,A.

DOWNS.

NUREG/CA-4551 V7R1P1: EVALUATION OF SEVERE ACCIDENT RISKS ZION UNIT 1 Main Report.

VEHEC.T.A.

NURE G/CR-4551V7R1P2A. EVALUATION OF SEVERE ACCOE NT NUREG/CR 5834. AUxlLIARY FEEDWATER SYSTEM RISK-BASED IN.

RISKS ZON UNIT 1.

n6x A.

SPECTON GUIDE FOR THE FORT CALHOUN NUCLEAR POWER NURE G /CR-4551V7R IP.. EVALUATION OF SEVERE ACCIDENT PLANT.

RISKS ZION UNIT 1.Appen6ces B, C, D And E.

NUREG/CR-5835 AUxluARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE BEAVER VALLEY. UNITS 1 AND 2 NU-TtONG.LW.

I CLEAR POWER PLANTS.

NUREG/CR-4BJ2 VOS ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR-5898 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-POWER PLANT: RISK METHODS INTEGR6 TON AND EVALUATON SPECTION GUIDE FOR THE PO!NT BEACH NUCLEAR POWER PROGRAM lRM)EP) Seismic Analysis.

PLANT.

l l

~

t L

==

74 Personal Author Index WANG.S.K.

VESE LKA,T.D.

NUREG/CR4080- REPLACEMENT ENERGY. CAPACITY, AND REll-NOREG/CR-5642. UGHT WATER REACTOR LOWER HEAD FAILURE ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT-ANALYSIS.

DOWNS W ANG,Y.K.

NURE lb5993 V01: METHODS FOR DEPENDENCY ESTIMATION AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE DATA STATISTICS Summary Report W ARE*A'G' NUREG/CR 59?3 V02: METHODS FOR DEPENDENCY ESTIMATION NUREG/CR 6027: PRELIMINARY EVALUATION OF SNUBBER SINGLE AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE FAILURES DAT A STATISTICS Detailed Desenption And Applations.

WARNER,R.D.

VICENTE.K.J.

NUREG/CR-5973. CODES AND STANDARDS AND OTHER GUIDANCE NUREG/CR4977 A PERFORMANCE INDICATOR OF THE EFFECTIVE.

CITED IN HEGULATORY DOCUMENTS.

NESS OF HUMAN MACHINE INTERFACES FOR NUCLEAR POWER PLANTS.

WASTLER,$.

VILLARAN M.

NUREG-1476: FINAL ENVIRONMENTAL IMPACT ST ATEMENT TO NURE G/CR-5844. AGING ASSESSMENT OF BISTABLES AND CONSTRUCT AND OPERATE A FACILITY TO RECEIVE STORE, AND SWITCHES IN NUCLF AR POWER PLANTS-OfSPOSE OF t t E.(2)

BYPRODUCT MATERIAL NEAR NUREG/CR 5933, HIGH PRESSURE COOLANT INJECTION (HPCI)

CLIVE.UT AH. Docket No. 40-8989,Envirocare Of Utah Inc.

SYSTEM RISK. BASED INSPECTION GUIDE FOR DRESDEN NUCLE-NUREG 1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT AR POWER STAllON UNITS 2 AND 3-TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE, NUREG/CR-5934: HIGH PRESSURE COOLANT INJECTION (HPCI)

AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NEAR CUVE.

SYSTEM RISK BASED INSPECTION GUIDE FOR OUAD CITIES UTAH Docket No. 40-8989,Ermrocaro Of Utah, Inc.

STATION. UNITS 1 AND 2 NUREG/CR-5959-HIGH PRESSURE COOLANT INJECTION (HPCI)

WEBER M.F SYSTEM RISK-BASED INSPECTION GUIDE FOR ENRICO FERMI NUREG 1444 SITE DECOMMISSIONING MANAGEMENT PLAN.

ATOMIC POWER PLANT, UNIT 2.

NUREG 1476 FINAL ENVIRONMENTAL lMPACT STATEMENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, AND VINTHER,R.W.

CISPOSE OF 11E (2)

BYPRODUCT MATERIAL NEAR NUREG/CR-5973. CODES AND STANDARDS AND OTHER GUIDANCE CLIVE. UTAH Docket No. 40-8989.Envirocare Of Utah,Inc.

CITED IN REGULATORY DOCUMENTS NUREG-1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT TO CONSTRUCT AND OPERATE A FACluTY TO RECEIVE, STORE, VO,T.V.

AND DISPOSE OF 11E (2) BYPRODUCT MATERIAL NEAR CUVE, NUREGICR 4469 V15: NONDESTRUCTIVE EXAMINATION (NDE) REU-ABILITY FOR INSE RVICE INSPECTION OF LIGHT WAT ER UT AH Docket No. 40-8989 Envirocar10f utah. Inc.

REACTORS Semiannual Report Octoter 1991. March 1992.

NUREG/GR 4469 V16: NONDESTRUCTIVE EXAMINATION (NDE) RELf-WEtSS.A.J.

ABluTY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CP 0126 VOI: PROCEEDINGS OF THE TWENTIETH WATER RE ACTORS Semiannual Report.Aptd 1992-September 1992.

REACTOR SAFETY INFORMATION MEETING.

NUREG/CR 5488. RISK-DASE.D INSPECTION GUIDE FOR THREE MILE NUREG/CP 0126 V02: PROCEEDINGS OF THE TWENTIETH WATER ISLAND NUCLE AR STATION UNIT 1-REACTOR SAFETY INFORMATION MEETING.

NUREG/CR-5766 AUxlLIARY F EEDWATER SYSTEM RISK-BASED IN-NUREG/CP-0126 V03: PROCEEDINGS OF THE TWENTIETH WATER SPECTION guide FOR THE SAN ONOFRE UNIT 2 NUCLEAR REACTOR SAFETY INFORMATION MEETING.

POWER PLANT.

NUREGICH 5829 AUXILIARY FEEDWATER SYSTEM RISK BASED IN-WELCH,L SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER NUREG/CR 5962: HEALTH AND SAFETY IMPACTS FROM DISCRETE SOURCES OF NATURALLY-OCCURRING AND ACCELERATOR-PRO.

NUI CR 5833 AUXlUARY FEEDWATER SYSTEM RISK-BASED IN, A OACT E MAT LS (NARM)

CTION GUIDE FOR 1HE H B. ROBINSCN NUCLEAR POWER CE,0 g

SURVEY INSTRUMENTS.

NUREGICR 5834 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER WELLS,J.E.

PLANT.

NUREG/CR-4832 V0B: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR-5835: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NU-PROGRAM (RMiEP) Seismic Analysis.

CLEAR POWER PLANTS NUREG/CR-5836. AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-SPECTnON GUIDE FOR THE PALO VERDE NUCLEAR POWER WESTRA,C.

NUREG/CA-5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/CR-5897: AUXILIAR'Y FEEDWATER SYSTEM RISK. BASED IN-INDUSTRY. Annual Summary Of Program Performance Reports,CY PL ANT.

SPECTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR 1992 POWER PLANT.

NUREG/CR 5898 AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-WHEELER,T.A.

SPECTION GUIDE FOR THE POINT BEACH NUCLEAR POWER NUREG/CR-4832 V05: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR POWER PLANT: AlLM METHODS INTEGRATION AND EVALUATION PLANT-NUREG/CR4i004. V ALUE/MPACT ANALYSIS OF GENERIC ISSUE 143.

PROGRAM. Parameter Estimation Analysis And Screening Human Reli.

" AVAILABILITY OF HEATING. VENTILATION, AIR CONDITIONING atdty Analysis (HVAC) AND CHILLED WATER SYSTEMS" WMITEHEAD,0.W.

VOR A.J.P.

NUREG/CR-6471: ENHANCEMENTS TO DATA COLLECTION AND RE-NUREG 13M R04 NRC RESEARCH PROGRAM ON PLANT AGING.

PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

LISTING AND SUMMARIES OF REPORTS ISSUED THROUGH SEP, TEMBER 1993 gg NUREG/CR-5962 HEALTH AND SAFETY IMPACTS FROM DISCRETE W AGNE R,R.

SOURCES OF NATURALLY OCCURRING AND ACCELERATOR-PRO-NUREG/CR 5926' SANS INVESTIGATION OF LOW ALLOY STEELS IN DUCED RADIOACTIVE MATERIALS (NARM).

NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI.

NUREG/CR 6062: PERFORMANCE OF PORTABLE RADIATION TIONS SURVEY INSTRUMENT S.

WALTON,J.C.

WlBLIN C.M.

NURE G/GR-6070-. MODEUNG AFPROACHES FOR CONCRETE BAR.

NUREG-1400 AIR SAMPLING IN THE WORKPLACE. Final Report.

RfERS USED IN LOW 4EVEL WASTE DISPOSAL

!l Personal Author Index 75 WICKLIFF,0.S.

WRIGHT,T.

NUREG/CR-5229 V05. FIELD LYSIMETER INVESTIGATIONS. LOW.

NUREG/CR-5938: NATIONAL PROFILE ON COMMERCIALLY GENER-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR ATED LOW-LEVEL RADIOACTIVE MIXED WASTE.

FISCAL YEAR 1992 Annual Report-WU.S.L WlERENGA,P.J.

NUREG/CR-3950 V08. FUEL PERFORMANCE ANNUAL REPORT FOR NUREG/CR-5988-SOIL CHARACTERIZATION METHODS FOR UN.

1990.

[

SATURATED LOW-LEVEL WASTE Si1ES.

WU,Y.-T.

h WILKOWSKl,G.M.

NUREG/CR-5917 V01: SENSITIVITY AND UNCERTAINTY ANALYSES NUREG/CR4599 V02 N2. SHORT CRACKS IN PIPING AND P1 PING APPLIED TO ONE-DIMENSIONAL RADIONUCLIDE TRANSPORT IN A WELDS Semsannual Report. October 1991. March 1992 LAYERED FRACTURED ROCK MULTFRAC Analytic Solutions And NUREG/CR 4599 V03 N1: SHORT CRACKS IN PtPING AND PIPING Local Sensitmtses.

WELDS Semiannual Repon. Apni-September 1992.

NUREG/CR-5917 V02: SENSITIVITY AND UNCERTAINTY ANALYSES NUREG/CH 6098 LOADING RATE EFFttTS ON STRENGTH AND APPLIED TO ONE DIMENSIONAL RADIONUCLIDE TRANSPORT IN A FRACTURE TOUGHNESS OF PlPE STEELS USED IN TASK 1 OF LAYERED FRACTURED ROCK Evaluabon Of The Limit State Ap.

THE IPIRG PROGRAM.

proach.

WILLIAMS.M.L WY MAN,R.

NUREGICR6071 IMPACT OF ENDF/B VI CROSS-SECTION DATA ON NUREG/CR 6090: THE PROGRAMMABLE LOGIC CONTROLLER AND H B. ROBINSON CYCLE 9 DOSIMETRY CALCULATIONS.

ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS.

WILLING,D.L WYNHOFF,N.L NUREG/CR-6080. REPLACE MENT ENERGY. CAPACITY, AND RELi-NUREG/CR-5894. RADIONUCLIDE CHARACTERIZATION OF REAC-ABILITY COS'1S FOR PERMANENT NUCLEAR REACTOR SHUT.

TOR DECOMMISSIONING WASTE AND NEUTRON-ACTIVATED DOWNS.

METALS.

WILSON,T.L XING,L NUREG/CR 6060. HYDROGEN MIXING STUDIES (HMS) ASSESSMENT NUREG/CR 6056. A FRAMEWORK FOR THE ASSESSMENT OF M ANUAL.

SEVERE ACCIDENT MANAGEMENT STRATEGIES.

WINE G ARDNE R.W.

Y AK LE,J.L NUREGICR-6029 V01 AGING ASSESSMENT OF NUCLEAR AIR-NUREG/CR.5863 RISK ASSESSMENT OF ISOLATION DEVICES IN TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS Phase 1.

SAFETY SYSTEMS.

WINTERS L YAN.H.

NUREG/lA4091-ASSESSMENT OF RELAPS/ MOD 2 AGAINST A NATU.

NUREG/CR-5951: THE MANAGEMENT OF ATWS BY BORON INJEC-RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT TION BORSSEL E.

NUREG/CR 6025, THE PROBADlLITY OF MARK-1 CONTAINMENT FAILURE DY MELT-ATTACK OF THE LINER WITT.R.J.

NUREG/CR-5642 LIGHT WATER REACTOR LOWER HEAD FAILURE YAN,W.

ANALYSIS.

NUREG/GR-0010: HYBRID DIGITAL SIGNAL PROCESSING AND NEURAL NETWORKS FOR AUTOMATED DIAGNOSTICS USING NDE WITTMEYER.G.W.

METHODS.

NUREG/CR-5817 V02. NRC HIGH.LEVE L RADIOACTIVE WASTE RE-SCARCH AT CNWRA Calendar Year 1991 Y ANKIN,J.G.

NUREG/CR-5817 V03 N1: NRC HiGH= LEVEL RADIOACTIVE WASTE NUREG/CR-6072. EXPERIMENTAL STUDY ON THE COMBUSTION BE-RESEARCH AT CNWRA January June 1992 HAVIOR OF HYDROGEN-AIR M:XTURES WITH TURBULENT JET IG-NUREG/C45817 V03 N2. NRC HIGH LEVEL RADIOACTIVE WASTE NITION AT LARGE SCALE.

RESEARCH AT CNWRA. July-Decemoer 1992.

YARDUMIAN,J.

WOLFGONG.J.R.

NUREG-0525 V02 Ro t-SAFEGUARDS

SUMMARY

EVENT LIST NUREG/CR 6065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR-(SSEL) January 1,1990 Through December 31,1992.

WOLF R A M,LM.

YOUNG.F.I.

NUREG/CR 5976 DEVELOPMENT AND USE OF A TRAIN LEVEL NUREG/CR-5488: RISK-BASED INSPECTION GUIDE FOR THREE MILE PRODAD'LISTIC RISK ASSESSMENT.

ISLAND NUCLEAR STATION UNIT 1.

WONG,F.S.

YOUNG,M.H.

NUREG/GROOO6 DE POSITION SOFTWARE TO CALCULATE PARTI-NUREG /CR-5988. SOlt CHARACTER 12ATION METHODS FOR UN-CLE PENETRATION THROUGH AEROSOL TRANSPORT SATURATED LOW-LEVEL WASTE SITES SYSTE MS Fen.il Report YOUNG.S.R.

WONG.S.

NUREG/CR-5817 V03 N1, NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/CR 6022 HIGH PRESSURE COOLANT INJECTION (HPCI)

RESEARCH AT CNWRA. January. June 1992 SYSTEM RISK-BASED INSPECTION GUIDE F OR BROWNS FERRY NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE NUCLEAR POWER STATION RESEARCH AT CNWRAJuly-December 1992.

WOODS.H.W.

YOUNGBLOOD.R.

NUREG-1472 REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG/CR-6111; INTEGRATED SYSTEMS ANALYSIS OF THE PIUS GENERIC ISSUE 57. Ettects Of Fre Protechon System Actuahon On REACTOR.

Safety-Related Equipment Y U,D.

WOUDSTR 4,A.

NUREG/CR 6056-A FRAMEWORK FOR THE ASSESSMENT OF NUREG/fA.0112. ASSESSMENT OF RELAP5/ MOD 2 AGAINST ECN-RE-SEVERE ACCIDENT MANAGEMENT STRATEGIES.

FLOOD EXPERIME NTS ZHANG,G.

WRIGHT,A.L NUREG/GR 0009. STEPWISE INTEGRAL SCALING METHOD AND ITS NUREG/CR 6065 SYSTEMS ANALYSIS OF THE CANDU 3 RE ACTOR APPLICATION TO SEVERE ACCIDENT PHENOMENA.

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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.

2D/30 Program Acc6 dent Mitigation NUREG/lA-0126 20/3D PROGRAM WORK

SUMMARY

RE POR T NUREG/CR 5907: CORE CONCRETE INTERACTIONS WITH OVERLY.

NUREG/lA 0121 RE ACTOR SAFETY ISSUES RESOLVED BY THE 2D/

ING WATER POOLS.The WETCOR 1 Test 30 PROGRAM A533 B Steel Accident Progression NUREG/CR-5360: XSOR CODES USERS MANUAL NUREG/CH+03& INiilAL RESULTS OF 1HE INFLUENCE OF DIAXIAL LOADING ON FRACTURE TOUGHNESS Accident Sequence Precursor ACRS Report NUREG/CR-5936 ENHANCEMENTS TO THE ACCIDENT PRECURSOR MET HODOLOGY-NUREG 1125 Vt4 A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.1992 Annual Adeorber AEOD NUREG/CR 6029 V01. AGING ASSESSMENT OF NUCLEAR AIR-NUREG 1272 V07 NOI: OF FICE FOR ANALYSIS AND EVALUATION OF TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS Phase L OPERATIONAL DATA.1992 Annual Report - Power Reactors.

Advanced Reactor NUREG-1272 V07 NO2 OF FICE FOR ANALYSIS AND EVALUATION OF OPERATFDNAL DATA.1992 Annual Report - Nonreactors NUREG/CR 5955: MATERIALS AND DESIGN DASES ISSUES IN ASME CODE CASE N 47.

ALARA NUREG/CM1469 V07 OCCUPATIONAL DOSE REDUCTION AT NU-Advisory Committee On Nuclear Waste CLEAR POWER PtANTS ANNOTATED BIBLIOGRAPHY OF SELECT-NUREG-1423 V04 A COMPILATION OF REPORTS OF THE ADVISORY E D READINGS IN RADIATION PROTECTION AND ALARA COMMITTEE ON NUCLEAR WASTEJuly 1992. June 1993 AS W Code Aerosol NUREG/CR 5901: A SIMPLIFIED MODEL OF AEROSOL SCRUDDING NUREG 1482 DRF.T I.C: GUIDELINES FOR INSERVICE TESTING AT NUCLEAR POWER PLANTS Draft Report for Comrrent.

SY A WATER POOL OVERLYING CORE DEBRIS INTERACTING NUREG/CR 5350 REVIEW OF ASME CODE CRITERIA FOR CONTROL WITH CONCRETE Final Report OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM DRANCH CON NUREG/CR-5907: CORE CONCbETE INTERACTIONS WITH OVERLY-NECTIONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP.

ING WATER POOLS The WE TCOR-1 Test Mf NT WORK NUREG/CH-5966. A SIMPLIFIED MODEL OF AEROSOL REMOVAL DY CONTAINMENT SPRAYS.

ASME Code Case N-47 NL C/CF 59 MATERIALS AND DESIGN DASES ISSUES IN ASME UR O R 3.

CORCON MOD 3.AN INTEGRATED COMPUTER MODEL FOR ANALYSIS OF MOLT E N CORE-CONCRETE ATWS INTERACTIONS User's Manual.

NU G/CH-5951: THE MANAGEME NT OF ATWS BY DORON INJEC.

Aerosol Transport System NUREG/GR-0006: DEPOSITION SOFTWARE TO CALCULATE PARTI-Abnormal Occurrence CLE PENETRATION THROUGH AEROSOL TRANSPORT NURE G 00u0 V15 N04 REPORT TO CONGRESS ON ABNORMAL SYSTEML Nal Report OCCURRENCES Octotmr De CernDer 1992 NUREG4090 Vl6 Nol: RfPORT TO CONGRESS ON ABNORMAL A94 OCCURRENCES January March 1993 NUREG/CP4134. INTERNATIONAL ATOMIC ENERGY AGE NCY SPE-NUREG-0090 Vt6 N02-REPORT TO CONGRESS ON ADNORMAL CIALISTS MEETING ON EXPERIENCE IN AGING. MAINTENANCE, OCCURRENCES Apni-June 1993 AND MODERNIZATION OF INSTRUMENTATION AND CONTROL SYSTEMS FOR IMPROVING NUCLEAR POWER PLANT Abnormat Transbent Condrtion AVAILABILITY. Held At Roc 6 mile.MD,May5-7,1993.

NUREG/CR-5862. TRAC.B THERMAL-HYDRAULIC ANALYSIS OF THE NUREG/CR 5404 V02: AUXILIARY FE EDWATER SYSTEM AGING OLACK FOX DOILING WATE R RE ACTOR STUDY. Phase I Follow On Study NUREG/CR 5699 V01: AGING AND SERVICE WEAR OF CONTROL Accelerator Produced ROD DRIVE MECHANISMS FOR OWR NUCLEAR Pt ANTS NURE G/CR 5962. HE ALTH AND SAFETY IMPACTS F ROM DISCRETE NUREG/CR-5754. BOILING-WATER REACTOR INTERNALS AGING SOURCES OF NATURALLY OCCURRING AND ACCELERATOR-PRO-DEGRADATION STUDY. Phase 1.

DUCED HADIOACTIVE MATERIALS (NARM)

NUREG/CH-5851: LONG TERM PERFORMANCE AND AGING CHAR-ACTERISTICS OF NUCLEAR PLANT PRESSURF TRANSMITTEPS Accident Consequence Analysie NUREG/CR-6015: STRUCTURAL AGING P740GF AM TECHNICAL NUMEG/CR-4214 R1P2A2. HEALTH EFFECTS MODELS FOR NUCLE-PROGRESS FOR PERIOD JANUARY, DECEMBER 1992.

AR POWER PLANT ACCIDENT CONSCOVENCE NUREG/CR-f420 V01: AGING ASSESSMENT Of NUCLEAR AIR.

ANALYSTS Mo6hcaton Of Morlots Resulting From AMtion Of Effects TREATMENT SYSTEM HEPA FILTERS AND ADSCRBERS Phase L Of Exposure To Alpha Emnung Adonucintes Part II Scientshc Bases NUMEG/CR-6043 V01: AGING ASSESSMENT OF JSSENTIAL HVAC For Healttt..

CHILLERS USED IN NUCLEAR POWER Pt ANTS Ptone L NUREG/CR4048: PRE SSURIZED-WATER REACTEC INTERNALS Accident Management AGING DEGRADATION STUDY, Phase 1.

NUREG/CR 5937. IN T ENTION AL DLPRE SSUR!Z A T!ON ACCIDENT NUREG/CR-6052 METHODOLOGY FOR RELIABILITY L4ASED CONDI-M ANAGEME NT STR AT E GY FOR PRESSURIZED WATER REAC.

TION ASSESSMENT. Apphcahon To Concrete Structures in Nuclear TORS Plants 77 s

h I

=

78 Subject index NUREG/CR-5897. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-Aging Assessment SPECTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR NUREGICR-5783-AGING ASSESSMENT OF THE COMBUSTION ENGL NEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES.

POWER PLANT.

NUREG/C45844. AGING ASSESSMENT OF BISTABL ES AND NUREG/C45898. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SWITCHES IN NUCLEAR POWER PLANTS.

SPECTION GUIDE FOR THE POINT BEACH NUCLEAR POWER PLANT.

Agreement States NUREG-1479: RESULTS FROM TWO WORKSHOPS: STATE RADI-Average Dose ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG-NUREG-0713 V12: OCCUPATIONAL RADIATION EXPOSURE AT COM-ULATIONS AND FUNDING.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACluTIES 1990 Twenty TNrd Annual Report.

Air Samphng NUREG-1400: AIR SAMPLING IN THE WORKPLACE Final Report.

BETHSY Test 9 NUREG/GR 0006: DEPOSITION. SOFTWARE TO CALCULATE PARTI-NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 9.1.8 USING CLE PENETRATION THROUGH AEROSOL TRANSPORT RELAPS/ MOD 3.

SYSTEMS Final Report.

BIGFLOW Air Treatment System NUREG/CR-6028; B!GFLOW: A NUMERICAL CODE FOR SIMULATING NUREG/CR4029 VO t; AGING ASSESSMENT OF NUCLEAR AIR.

FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC TREA1 MENT SYSTEM HEPA FILTERS AND ADSORBERS. Phase 1.

MEDIA. Theory And User's Manual - Version 1,1.

Airborne Effluent NUREG/CR-2907 VII: RADIOACTIVE MATERIALS RELEASED FROM BWR NUCLEAR POWER PLANTS Annual Report 1990 NUREG/CR-5699 V01: AG!NG AND SERVICE WEAR OF CONTROL ROD DRIVE MECHANISMS FOR BWR NUCLEAR PLANTS.

Airborne Radioactive Material NUREG/CR-5754: BOluNG-WATER REACTOR INTERNALS AGING NUREG 1400 AIR SAMPUNG IN THE WORKPLACE Fenal Report.

OEGRADATION STUDY. Phase 1.

NUREG/C45791: RISK EVALUATION FOR A GENERAL ELECTRIC Alpha-Emit 16n0 Radionuclide BWR, EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON NUREG/CR-4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR SAFETY-RELATED EQUIPMENT. Evaluahon Of Generic issue 57.

POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part 1:

NUREG/CR 5882. TRAC-B THERMAL-HYDRAULIC ANALYSIS OF THE Introduction, integration,And Summary BLACK FOX BOluNG WATER REACTOR NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARACTER-S CM EACH BON NNS NIN NU LG S V09 U S. NUCLEAR REGULATORY COMMISSION 1992 BY THE MELCOR CODE.

ANNUAL REPORT NUREG/CR-5951: THE MANAGEMENT OF ATWS BY BORON INJEC-NUREG-1470 V02: CHIEF FINANCIAL OFFICER'S ANNUAL REPORT -

TlON.

1993 NUREG/CR-5978: SOURCE TERM ATTENUATION BY WATER IN THE NUREG/CR-3950 V08: FUEL PERFORMANCE ANNUAL REPORT FOR MARK i BOILING WATER REACTOR DRYWELL g 999.

NUREG/CR-5995: TECHNICAL SPECIFICATION ACTION STATEMENTS REOUIRING SHUTDOWN A Risk Perspective With Apphcation To The Arid Disposal Site NUREG/CR-5980: THREE DIMENSIONAL REDISTRIBUTION OF TRITI-RHR/SSW Systems Of A BWR.

NUREG/CR-6022: HIGH PRESSURE COOLANT INJECTION (HPCl)

UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED SYSTEM RISK-BASED INSPECTION GUIDE FOR BROWNS FERRY SOIL.A Deterministic Model For Tnhum Migrahon in An And Disposal NUCLEAR POWER ST AilON Site.

NUREG/CR 6049. PIPING BENCHMARK PROBLEMS FOR THE GEN-NUREG /CR-6108-SPHERICAL DIFFUSION OF TRITIUM FROM A POINT OF RELEASE IN A UNIFORM UNSATURATED SOIL A Deter.

ERAL ELECTRIC ADVANCED BOfuNG WATER REACTOR.

minist c Model For Tntium M>grabon in An And Dmposal Site.

Blaximi Loading NUREG/CR-6036: INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL Atmospheric Dispersion NUREG/CR-5247 V01 Rt: RASCAL VER$10N 2.0 USER'S GUIDE.

LOADING ON FRACTURE TOUGHNESS.

NUREG/CR-5247 V02' RASCAL VERSION 2 0 WORKBOOK.

NUREG/CR 6059: MACCS VERSION 1.5.11.1: A MAINTENANCE RE-Blaxial Tension LEASE OF,THE CODE.

NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT Atomic Safety And Licensing Board Panel NUREG-1363 V05 ATOMIC SAFETY AND LICENSING POARD PANEL Blatable ANNUAL REPORT Fmcal Year 1992.

NUREG/CR-5844: AGING ASSESSMENT OF B! STABLES AND SWITCHES IN NUCLEAR POWER PLANTS.

Austenitic Stainless Steel NUREG/CR 5099: INTERIM FATIQUE DESIGN CURVES FOR CARBON.

Blackout LO L

Y, AND AUSTENiTIC STAINLESS STEELS IN LWR ENVI.

NUREG/lA-0119: ASSESSMENT AND APPLICATION OF BLACKOUT TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/

MOD 2.

Auxillary Feedwater System NUREG/lA-0123: APPLICATION OF FULL POWER BLACKOUT FOR NUREG/CR-5404 V02. AUXILIARY FEEDWATER SYSTEM AGING C N. ALMARAZ WITH RELAPS/ MOD 2.

STUDY Phase I Follow On Study NUREG/CR-5766. AuxlLIARY FEEDWATER SYSTEM RISK-BASED IN-Bomng Water Reactor SPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR NUREG/CR-5699 VOI: AGING AND SERVICE WEAR OF CONTROL POWER PLANT.

ROD DRIVE MECHANISMS FOR BWR NUCLEAR PLANTS NUREG/CR 58?9 AUXlLIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR-5754. BOILING-WATER REACTOR INTERNALS AGING SPECTION GUtDANCE FOR THE DAVfS-BESSE NUCLEAR POWER DEGRADATION STUDY. Phase 1.

PLANT.

NUREG/CR-5791: RISK EVALUATION FOR A GENERAL ELECTRIC NUREG/CR 5833. AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-BWR. EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON SPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER SAFETY RELATED EOUIPMENT. Evaluation Of Genenc issue 57.

NUREGICR-5834 AUXILIARY FEEDWATER SYSTEM RISK BASED IN.

NUREG/CR-5882: TRAC-B THERMAL-HYDRAUUC ANALYSIS OF THE PLANT.

BLACK FOX BOluNG WATER REACTOR.

SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARACTER-PLANT.

NUREG/CR 5835. AUXiuARY FEEDWATER SYSTEM RISK-BASED IN-ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED SPECTION GUIDE FOR THE BEAVER VALLEY. UNITS 1 AND 2 NO-BY THE MELCOR CODE.

NUREG/CR-5951: THE MANAGEMENT OF ATWS BY BORON INJEC.

CLEAR POWER PLANTS NUREG/M5836. AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-TION.

SPECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER NUREG/CR-5978 SOURCE TERM ATTENUATION BY WATER IN THE MARK f BOluNG WATER REACTOR DRYWELL PLANT.

i

Subject Index 79 NUREG/CR 5995 TECHNICAL SPECIFICATION ACTION STATEMENTS Charpy REOUtRING SHUTDOWN.A Risk Perspechve With Application To The NUREG/CR 6023: GENERIC ANALYSIS FOR EVALUATION OF LOW RHR/SSW Systems Of A BWR-CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS NUREG/CR.6022: HIGH PRESSURE COOLANT INJECTION (HPCI)

AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATfRl-SYSTEM RISK-BASED INSPECTION GUIDE FOR BROWNS FERRY A(3' NUCLEAR POWER STATION NUREG/CR-6049 PIPING BENCHMARK PROBLEMS FOR THE GEN-Charpy impact Toughness ERAL ELECTRIC ADVANCED BOILING WATER REACTOR NUREG/CR 6972: EFFECTS OF NONSTANDARD HEAT TREATMENT Boron Dilution TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES NUREG/CR-5822: ANALYSIS OF THERMAL MIXING AND BORON 01 OF CARBON-STEEL CASTING REPAIR WELDS.

LUTION IN A PWR.

Charpy V Notch Impact Boron piection NUREG/CR 5914. CHEMICAL COMPOSITION AND RT(NOT) DETERMi-WREGiCM951 THE MANAGEMENT OF ATWS BY BORON INJEC-NATIONS FOR MIDLAND WELD WF 70.

TtON.

Check Valve Branch Connection NUREG/CR-5944: A CHARACTERl2ATION OF CHECK VALVE DEGRA.

NUREG/CR-5358 REVIEW OF ASME CODE CRITERIA FOR CONTROL DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM BRANCH CON-DUSTRY.

1 NECTIONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP-

)

MENT WORK.

Chemical Composition NUREG/CR-5914: CHEMICAL COMPOSITION AND RT(NDT) DETERMI-NU EG C 9 SEISMOLOGICAL INVESTIGATION OF E ART H-OUAKES IN THE NEW MADRID SEISMIC ZONE Final Chilled Water System Roport September 1986 - December 1992 NUREG 1427. REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 143: AVAILABILITY OF CHILLED WATER SYSTEM Budget AND ROOM COOUNG NUREG 1100 V09 BUDGET ESTIMATES Fiscal Years 1994-1995.

Byproduct Material Chiller NUREG 1476 FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR 6043 VOI: AGING ASSESSMENT OF ESSENTIAL HVAC CONSTRUCT AND OPERATE A FACILITY TO RECElvE. STORE, AND CHILLERS USED IN NUCLEAR POWER PLANTS Phase 1.

DISPOSE OF I t E (2)

BYPRODUCT MATERIAL NEAR CLIVE.UT AH Docket No 40 8989. Envirocare Of Utah.Inc Clait>orne Enrichment Center NUREG-1476 DRFT: ORAFT ENVIRONMENT AL lMPACT STATEMENT NUREG-1484 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE, FOR THE CONSTRUCTION AND OPERATION OF CLAIBORNE EN-AND DISPOSE OF 11E (2) BYPRODUCT MATER;AL NEAR CLIVE.

RICHMENT CENTER, HOMER, LOulSIANA Docket No. 70-3070, Louise UTAH Docket No 40-8989 Enwocare Of Utah, Inc.

ana Energy Services,L.P, CANDU 3 Reactor Class 1E Digital System NUREG/CR 6065. SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR-NUREG/CR 6113 CLASS 1E DIGITAL SYSTEMS STUDIES.

CFO's Act Closure Bolt NUREG-1470 V02 CHIEF FINANCIAL OFFICER'S ANNUAL REPORT -

NUREG/CR-6007: STRESS ANALYS!S OF CLOSURE BOLTS FOR E3 SHIPPING CASKS CORCON-Mod 3 e

ConecM Dose NUREG/CR 5843 CORCON MOD 3 AN INTEGRATED COMPUTER NUREG-0713 V14 OCCUPATIONAL RADIATION EXPOSURE AT COM-MODE L FOR ANALYSIS OF MOLTEN CORE-CONCRETE MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES INTERACTIONS. user's Manual.

1992 Twenty-Fittn Annual Report.

Calibration NUREG/CR.5903: VALIDATION OF SMART SENSOR TECHNOLOG!ES Combustible Gas FOR INST RUMENT CAllBRATION RE DUCTION IN NUCLF.AR NUREG-1364 REGULATORY ANALYSIS FOR THE RESOLUTION OF POWER PLANTS.

GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY COMBUSTIBLE GASES IN VITAL AREAS.

Carbon Content NUREG/CR 5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS NUREG/CR-5972: EFFECTS OF NONST ANDARD HEAT TREATMENT STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES 17ED WATER REACTOR PLANTS OF CARBON STEEL CASTING REPAIR WELDS Combustion Dehavior Cavitation NUREG/CR 6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE.

NUREGICR.6031-CAVIT ATION GUIDE FOR CONTROL VALVES HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-NITION AT LARGE SCALE Cavity Flood 6ng NURE G /CR-6056. A FRAMEWORK FOR THE ASSESSMENT OF Common Cause failure SEVERE ACCIDENT MANAGEMENT STRATECfES NUREG/CR-5801: PROCEDURE FOR ANALYSIS OF COMMON CAUSE FAILURES IN PROBABILISTIC SAFETY ANALYSIS.

Cement Degradation NUREG/CH-5987; MICROBIAL-INFLUENCED CEMENT DEGRADATION Communication Media

~

NUREG/CR 6082 DATA COMMUNICATIONS.

Certincates Of Compilance Concrete NUREG 0383 V01 R16 DIRECTORY OF CERTIFICATES OF COMPLI.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC NUREG/CR 5901: A SIMPLIFIED MODEL OF AEROSOL SCRUBBING

)

BY A WATER POOL OVERLYING CORE DEBRIS INTERACTING Approved Packages NURE G.OJ83 V02 R16 DIRECTORY OF CERTIFICATES OF COMPLl-WITH CONCRETE.Fsnal Floport.

ANCE FOR RAD 60 ACTIVE MATERIALS PACKAGES Certificates Of NUREG/CR-5907: MICROBIAL-INFLUENCED CEMENT DEGRADATION

- LITERATURE REVIEW.

Compliance NUREG-0383 V03 R13. DIRECTOnY OF CERTIFICATES OF COMPU-NUREG/CR 6015-STRUCTURAL AGING PROGRAM TECHNICAL ANCE FOR RADIOACTIVE MATERIALS PACKAGESReport Of NRC PROGRESS FOR PERIOD JANUARY DECEMBER 1992.

Approwd Quahty Assurance Prograrns For Radscactive Matenals Pack-NURE G/CR-6032: SOLIDUS AND UOUIDUS TEMPERATURES OF CORE-CONCRETE MIXTURES.

ages

80 Subject index f

Concrete Barrier NUREG/CR 4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4070: MODELING APPROACHES FOR CONCRETE BAR-LIGHT WATER REACTORS. Semiannual Report. October 1992 March AlERS USED IN LOW-LEVEL WASTE DISPOSAL 1993.

Concrete Structure Crack NUREG/CR4052: METHODOLOGY FOR RELIABluTY BASED CONDI-NUREG/CR 4599 V02 N2: SHORT CRACKS IN PlPING AND PIPING TION ASSESSMENT. Apphcation To Concrete Structures in Nuclear WELDS Semiannual Report. October 1991 March 1992.

Plants.

NUREG/CR4599 V03 N1: SHORT CRACKS IN PIPING AND PIPING WELDS Semannual Report, April-September 1992, Configurable System NUREG/CR4078: ANALYSIS OF CRACK INITIATION AND GROWTH IN NUREG/CR.6090: THE PROGRAMMABLE LOGtC CONTROLLER AND THE HIGH LEVEL VIBRATION TEST AT TADOTSO.

ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS.

Crack Pop-in NUREG/CR-5952: EVALUATION OF CRACK POP.lNS AND THE DE-1)RE CR5971; CONTINUUM AND MICROMECHANICS TREATMENT TERMINATION OF THEIR RELEVANCE TO DESIGN CUNSIDER-OF CONSTRAINT IN FRACTURE.

ATIONS.

Containment Crack Propagation NUREG/CR-4273: CRACK PROPAGATON IN HIGH STRAIN REGIONS NUREG/CR-4273: CRACK PROPAGATON IN HIGH STRAIN REGIONS OF SEQUOYAH CONTAINMENT.

NUREG/CR5747. ESTIMATE OF RADIONUCUDE RELEASE CHARAC.

OF SEOUOYAH CONTAINMENT.

TERISTICS INTO CONTAINMENT UNDER SEVERE ACCIDENT Crack-Arrest CONDIT!ONS Final Roport.

NUREG/CR 5952; EVALUATON OF CRACK POP-INS AND THE DE-NUREG/GR0009: STEPWISE INTEGRAL SCALING METHOD AND ITS TERMINATION OF THEIR RELEVANCE TO DESIGN CONSIDER-APPLICATION TO SEVERE ACCIDENT PHENOMENA.

ATIONS.

Containment Failure NUREG/CR-6025: THE PROBABILITY OF MARK-1 CONTAINMENT Cracking FAILURE BY MELT ATTACK OF THE LINER NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS. Semiannual Report. April-Soptember 1992.

Containment Heating NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5949: ASSESSMENT OF THE POTENTIAL FOR HIGH LIGHT WATER REACTORS. Semiannual Report. October 1992 - March PRESSURE MELT EJECTION RESULTING FROM A SURRY ST ATION 1993.

BLACKOUT TRANSIENT.

DEPOSITION Computer Code Containment Penetration NUREG/GR 0006. DEPOSITION SOFTWARE TO CALCULATE PARTI-NUREG/CR-5978 ISLOCA RESEARCH PROGRAM Fmal Report.

CLE PENETRATION THROUGH AEROSOL TRANSPORT NUREG/CR 6027: PREUM NARY EVALUATION OF SNUBBER SINGLE SYSTEMS. Final Report.

F AILURES.

DOE Waste Package AND W N N & N RE 9

A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY WASTE PACKAGE TEST DATA. Biannual Report. August 1989 Janu-CONTAINMENT SPRAYS.

NUREG/CR 5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN ary im CONTAINMENT MANAGEMENT.

Damping NUREG/CR 5776: DAMPING IN LOW. ASPECT-RATIO. REINFORCED Containment System NUREG/CR 6060: HYDROGEN MlXING STUDIES (HMS) ASSESSMENT CONCRETE SHEAR WALLS.

Data Collection NUREG/CR-5471: ENHANCEMENTS TO DATA COLLECTION AND RE-Control Rod Drive NUREG/CR-5699 V01: AGING AND SERVICE WEAR OF CONTROL PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS ROO DRIVE MECHANISMS FOR BWR NUCLEAR PLANTS NUREG/CR-5783: AGING ASSESSMENT OF THE COMBUSTION ENGi-Data Communicat6on NEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES, NUREG/CR-6082. DATA COMMUNICATIONS Control Valve Decommissioning NUREG/CR4031: CAVITATION GUIDE FOR CONTROL VALVES.

NUREG 1307 R03: REPORT ON WASTE BURIAL CHARGES Escalation Of Decommissiorung Waste Disposal Costs At Low-Level Waste Bunal Cooldown Transient Facilities.

NUREG/CR$983. SAFETY ASPECTS OF FORCED FLOW COOLDOWN NUREG-1444 SITE DECOMMISSIONING MANAGEMENT PLAN.

TRANSIENTS IN MODULAR H!GH TEMPERATURE GAS-COOLED NUREG/CR 5894: RADIONUCUDE CHARACTERIZATION OF REAC-REACTORS.

TOR DECOMMIS$10NING WASTE AND NEUTRON-ACTIVATED METALS.

Core Debris NUREG/CR 6054 DRF FC; ESTIMATING PRESSURIZED WATER RE-NUREG/CR5901: A SIMPLIFIED MODEL OF 4EROSOL SCRUBBING ACTOR DECOMMISSONING COSTS.A User's Manual For The PWR BY A WATER POOL OVERLYING CORE DEBRIS INTERACTING Cost Estimating Computer Program (CECP) Software. Draft Report For WITH CONCRETE. Final Report Comment Core Melt Progression NUREG/CR-6025. THE PROBABluTY OF MARK-1 CONTAINMENT UREG/CR 5966: A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY FAILURE BY MELT ATTACK OF THE LINER.

CONTAINMENT SPRAYS NUREG/CR C.047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF Core-Concrete Interaction VANADOUS AND VANADIC IONS.

NUREG/CR5843' CORLON-MOD 3 AN INTEGRATED COMFUTER NUREG/CR46081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-MODEL FOR ANALYSIS OF MOLTEN CORE CONCRETE CLES BY FLOOROCARBON SURFACTANT SOLUTIONS.

INTERACTIONS, User's Manual.

NUREG/CR-5907 CORE CONCRETE INTERACTIONS WITH OVERLY-Decontamination Weste Program lNG WATER POOLS.The WETCOR 1 Test' NUREG/CR5672 V03: CHARACTERISTICS OF LOW-LEVEL RADIOAC.

Corium D6spersion TlVE DECONTAMINATION WASTE. Annual Report For Fiscal Year NUREG/GR-0009. STEPWISE INTEGRAL SCAUNG METHOD AND ITS 1992.

APPLICATON TO SEVERE ACCIDENT PHENOMENA.

Dependency Estimation Corrosion Fatague NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION NUREGICR-4667 Vt5 ENVIRONMENTAu.Y ASSISTED CRACKING IN AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE UGHT WATER REACTORS. Semiannual Report,AprirSeptember 1992.

DATA STATISTICS Summary Report.

l.

Subject index 81 NUREG/CR-5993 V02. METHODS FOR DEPENDENCY ESTIMATION PLANT-AN APPLICATION OF THE CSAU METHODOLOGY USING AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE THE RELAPS/ MOD 3 COMPUTER CODE.

DATA STATISTICS Detaded Descnpton And Applicatons.

EDSFl Depressurtrat6on NUREG 1473 ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL IN-NUREG/CR 5937. INTENTIONAL DEPRESSURtZATION ACCIDENT SPECTION (EDSFI) DATA BASE PROGRAM.

MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC.

TORS.

ENDF/D-VI Cross-Section Data NUREG/CR-6071: IMPACT OF ENDF/B VI CROSS-SECTION DATA ON UF EG/CR 5955. MATERIALS AND DESIGN BASES ISSUES IN ASME COOE CASE N-47.

EPICOR-il Design Basis Event NUREG/CR 5229 VOS. FIELD LYSIMETER INVESTIGATIONS-LOW-NUREG/CR-6027: PRELIMINARY EVALUATION OF SNUBBER SINGL.E LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FAILURES.

FISCAL YEAR 1992. Annual Report.

Design Consideration Earthquake NUREG/CR 5952; EVALUATION OF CRACK POP-lNS AND THE DE.

NUREG/CR-6034: OKLAHOMA SEISMIC NETWORK Final Report TERMINATION OF THEIR RELEVANCE TO DESIGN CONSIDER-NUREG/CR-6079: SEISMOLOGICAL INVESTIGATION OF EARTH-ATIONS OUAKES IN THE NEW MADRID SEISMIC ZONE Final Report September 1986 - December 1992 Destructlye Examination NU'EG/CR 6085-UNilED STATES SEISMOGRAPHIC NETWORK.

NUREG/CR-5961; POSTTEST DESTRUCTIVE EXAMINATION OF THE STEEL LINER IN A 16-SCALE REACTOR CONTAINMENT MODEL Earthquake Response NUREG/CR 6012: STIFFNESS AND DAMPlNG PROPERTIES OF A Detonation LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD-NUREG/CR 6072. EXPERIMENTAL STUDY ON THE COMBUSTION BE-ED EARTHOUAKE RESPONSES.

HAV)OR bF HYDROGEN AIR MIXTURES WITH TURBULENT JET IG-NITION AT LARGE SCALE.

Economic Regulat6on NUREG/CR 5975: INCENTIVE REGULATION OF INVESTOR OWNED D6sposal Unit Source Term NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULATORS NUREG/CR 6041: DISPOSAL UNIT SOURCE TERM (DUST) DATA INPUT GUIDE.

Electric Discharge NUREG/CR-5981. THE EFFECT OF ELECTRIC DISCHARGE MA-UR 90 02 Sot: NRC COMPREHENSIVE RECORDS DISPOSI-

^

TION SCHEDULE.

Dose Assessment Electrical Distribution System NUREG/CR-5247 V01 R1: RASCAL VERSION 2.0 USER'S GUIDE NUREG-1473: ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL IN-NUREG/CR 5247 V02: RASCAL VERSION 2.0 WORKBOOK.

SPECTION (EDSFI) DATA BASE PROGRAM Dose Commitment Electrical isolator NUREG/CR 2850 V11: DOSE COMMITMENTS DUE TO RADIOACTIVE NUREG-1451 REGULATORY ANALYSIS FOR THE RESOLUTION OF RELEASES FROM NUCLEAR POWER PLANT SITES IN 1989 GENERIC ISSUE 142: LEAKAGE THROUGH ELECTRICAL ISOLA-TORS IN INSTRUMENTATION CIRCUITS Dosimetry NUREG/CR-6059 MACCS VERSION t 5.111: A MAINTENANCE RE.

Embrlttlement LEASE OF THE CODE NUREG/CR 4744 V07 N1: LONG-TERM EMBRITTLEMENT OF CAST NUREG/CR-6071. IMPACT OF ENDF/B-VI CROSS-SECTION DATA ON DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual H.B. ROBINSON CYCLE 9 DOSIMETRY CALCULATIONS Report October 1991 March 1992.

NUREG/CR 4744 V07 N2: LONG TERM EMBRITTLEMENT OF CAST Draft Environmental lmpact Statement DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semaannual NUREG 1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATEMENT Report. April September 1992.

TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, NUREG/CR 5591 V01 N2: HEAVY SECTION STEEL IRRADIATION AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NEAR CLIVE-PROGRAM Semiannual Progress Report For Aprd-September 1990.

UTAH Docliet No 40-8989. Envwocare Of Utah, Inc.

NUREG-1484 DRFT; DRAFT ENVIRONMENTAL IMPACT STATEMENT Emergency Preparedness FOR THE CONSTRUCTION AND OPERATION OF CLAIDORNE EN~

NUREG-1474: E 'ECT OF HURRICANE ANDREW ON THE TURKEY RICHMENT CENTER. HOMER. LOUISLANA. Docket No 70-3070.Louiso POINT NUCL LENERATING STATION FROM AUGUST 20-30 ana Energy Services L P, j g3p.

NUREG-1485: L. AUTHORIZED FORCED ENTRY INTO THE PROTECT-

^

I E

N G/CR-5978 SOURCE TERM ATTENUATION BY WATER IN THE MARK 1 DOILING WATER REACTOR DRYWELL Emergency Shutdown System Ductate Fracture NUREG/CR-6090: THE PROGRAMMABLE LOGIC CONTROLLER AND NUREG/CR-6021 GENERIC ANALYSIS FOR EVALUATION OF LOW ITS APPLICATION IN NUCLEAR AFACTOR SYSTEMS CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS INST FRACTURE OF REACTOR PRESSURE VESSEL MATERI.

UY REG C CAVITATION GUIDE FOR CONTROL VALVES.

DuctHe-Brittle Enforcement Action NUREG/CR 5952. EVALUATION OF CRACK POP INS AND THE DE.

NUREG 0940 VII N04. ENFORCEMENT ACTIONS. SIGNIFICANT AC-TERMINAllON OF THEIR RELEVANCE TO DESIGN CONSIDER.

TIONS RESOLVED.Ouarterly Progress Report, October-December ATIONS 1992-NUREG-0940 V12 N01: ENFORCEMENT ACTIONS SIGN!FICANT AC-Dynamic Load TIONS RESOLVED.Ouarterly Progress Repor1. January-March 1993 NUREG/CR 6049 PIPING BENCHMARK PROBLEMS FOR THE GEN-NUREG-0940 V12 NO2. ENFORCEMENT ACTIONS. SIGNIFICANT AC.

ERAL ELECTRIC ADVANCED BOILING WATER REACTOR TIONS RESOLVED.Ouarterly Progress Report.Apni-June 1993.

ECCS Engineered Safety System NUREG/CR 5818-UNCERT AINTY ANALYSIS OF MINIMUM VESSEL NUREG/CR-6111: INTEGRATED SYSTEMS ANALYSIS OF THE PIUS LIQUlO INVENTORY DURING A SMAL L-BREAK LOCA IN A B&W REACTOR.

e

rm 82 Subject Index Envirocare Fire Protection System NUREG-1476: FINAL ENVIRONMENTAL-IMPACT STAT EMENT TO NUREG-1472: REGULATORY ANALYSIS FOR THE RESOLUTION OF CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, AND GENERIC ISSUE 57. Effects Of Fre Protection System Actuation On DISPOSE OF 11E f 2)

BYPRODUCT MAT ERIAL NEAR Safety-Related Equipment CLIVE UTAH Docket No.40-8989.Enwocare Of Utah.Inc.

NUREG/CR 5791: RISK EVALUATION FOR A GENERAL ELECTRIC NUREG 1476 DRFT; DRAFT ENVIRONMENTAL IMPACT STATEMENT BWR, EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE, STORE, SAFETY-RELATED EQUIPMENT. Evaluation Of Genenc issue 5'/

AND DISPOSE OF 11E(2) BYPRODUCT MATERIAL NEAR CLIVE, UTAH Docket No. 40-8989 Enwocare Of Utah,Inc.

Fiscal Year NUREG-1100 V09: BUDGET ESTlMATES Fiscal Years 1994 1995.

Essential Service Water NUREG 1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF Fttness For Duty GENERIC ISSUE 153-LOSS OF ESSENTIAL SERVICE WATER IN NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER LWRS4 (NDUSTRY. Annual Summary Of Program Performance Reports,CY m2.

Event Analysis NUREG/CR 5976-DEVELOPMENT AND USE OF A TRAIN-LEVEL Fluorocarbon PROBABILISTIC RISK ASSESSMENT.

NUREG/CR-6081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-CLES BY FLOOROCARBON SURFACTANT SOLUTIONS.

Event Tree Analysis NUREG/CR-5964 SAPHIRE TECHNICAL REFERENCE MANUAL.lRRAS/ SARA VERSION 4 0.

UREG 485 UNAUTHORIZED FORCED ENTRY INTO THE PROTECT-ED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7,1993.

Examiner Standards NUREG-1021 R07: OPERATOR LICENSING EXAMfNER STANDARDS.

Fracture Analyses Empert System NUREG/CR 5997; CSNI PROJECT FOR FRACTURE ANALYSES OF NUREG/CR4018. SURVEY AND ASSESSMENT OF CONVENTIONAL l.ARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS SOFTWARE VERIFICATION AND VAllDATION METHODS.

(PRCUECT FALSIRE).

External Event Fracture Mechanics NUREG/CR-4832 V08 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CP 0131: PROCEEDINGS OF THE JOINT 1AEA/CSNI SPECIAL-POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION ISTS' MEETING ON FRACTURE MECHANICS VERIFICATION BY PROGRAM (RMIEP) Seismic Analysis.

LARGE-SCALE TESTING. Held At Pollard Auditorium, Oak Ridge, Tennessee.

Entremity Dosimetry NUREG/CR-4219 V09 N2: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR-5989 PERFORMANCE TESTING OF EXTREMITY DOSI-PROGRAM. Semiannual Progress Report For ApriLSeptember 1992.

METERS-PILOT TEST.

NUREG/CR-4469 V15: NONDESTRUCTIVE EXAMINATION (NDE) RELi-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Failure Data REACTORS Semiannual Report,0ctober 1991. March 1992.

NUREG/CR-6944: A CHARACTERl2ATION OF CHECK VALVE DEGRA.

NUREG/CR-4469 V16: NONDESTRUCTIVE EXAMINATION (NDE) RELI-DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER DUSTRY' REACTORS. Semiannual Report. April 1992. September 1992..

NUREG/CR-4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING Fallure Event WELDS. Semiannual Report October 1991. March 1992.

NUREG/CR-5471 ENHANCEMENTS TO DATA COLLECTION AND RE.

NUREG/CR-4599 V03 N1: SHORT CRACKS IN PIP 1NG AND PIPING PORTING OF SINGLE AND MULTIPLE FA! LURE EVENTS.

WELDS. Semiannual Report, April-September 1992.

NUREG/CR-5410: STATISTICALLY BASED REEVALUATION OF PISC-Il Fatique Design Curve NUREG/CR 5999: INTERIM FATIQUE DESIGN CURVES FOR CARBON, ROUND ROBIN TEST DATA.

LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI..

NUREG/CR 5591 V01 N2: HEAVY SECTION STEEL IRRADIATION PROGRAM. Semiannual Progress Report For Apni-September 1990.

RONMENTS.

NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC Fault Tolerance FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CR-6101: SOFTWARE RELIABILITY AND SAFETY IN NUCLE-ROWE REACTOR PRESSURE VESSEL AR REACTOR PROTECTION SYSTEMS NUREG/CR 5958: TWO-PARAMETER FRACTURE MECHANICS:

NUREG/CR-6113 CLASS 1E DIGITAL SYSTEMS STUDIES.

THEORY AND APPLICATIONS.

NUREG/CR-5970 APPROXIMATE TECHNIOUES FOR PREDICTING Federal Guide SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC).

NUREG 1467:

FEDERAL GUIDE FOR A RADIOLOGICAL NUREG/CR 5971; CONTINUUM AND MICROMECHANICS TREATMENT RESPONSE Supporting The Nuclear Regulatory Commission During OF CONSTRAINT IN FRACTURE.

The initial Hours Of A Sonous Accident-NUREG/CR-5999. INTERIM FATIQUE DESIGN CURVES FOR CARBON, LOW ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI.

Feed Line Break RONMENTS NUREG/lA-0104 RELAP5/ MOD 3 ASSESSMENT USING THE SEMIS-NUREG/CR 6d98: LOADING RATE EFFECTS ON STRENGTH AND CALE 50% FEED LINE BREAK TEST S-FS 11' FRACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF THE IPIRG PROGRAM.

Final Environmental impact Statement NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO N EG CP 0131 PROCEEDINGS OF THE JOINT (AEA/CSNI SPECIAL+

S SE OF 11E )

BYPR D CT M/ E AL N R ISTS* MEETING ON FRACTURE MECHANICS VERIFICATION BY CLIVE.UT AH. Docket No. 40 8989,Envirocare Of Utah,lnc.

LARGE-SCALE TESTING.

Held At Pollard Auditorium, Oak Financial Management Ridge, Tennessee.

NUREG-1470 V02: CHIEF FINANCIAL OFFICER'S ANNUAL REPORT -

NUREC/CR-4744 V07 N2: LONG-TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual 1993 Report,Apni-September 1992.

Finger Dostmeter NUREG/CT5914: CHEMICAL COMPOSITION AND RTiNDT) DETERMI-NUREG/CR 5989 PERFORMANCE TESTING OF EXTREMITY DOSI.

NATIONS FOR MIDLAND WELD WF-70.

METERS-PILOT TEST.

NUREG/CR-5958. TWO-PARAMETER FRACTURE MECHANICS:

THEORY AND APPLICATIONS.

Finite 4enght Flaw NUREG/CR 5969. J AND CTOD ESTIMATION EQUATIONS FOR SHAL.

NUREG/CR 5968. POTENTIAL CHANGE IN FLAW GEOMETRY OF AN LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS.

!NITIALLY SHALLOW FINITE LENGTH SURFACE Fl.AW DURING A NUREG/CR-5970: APPROXIMATE TECHNIOUES FOR PREDICTING PRESSURIZED THERMAL SHOCK TRANSIENT.

SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC).

1

4 c.w..4+

.4-4-

.L g.h-e a

J

.e

.4 24 Ar'E-4

&+L--**-A-o 4

e-

+-

w--

Subject index 83 NUREG/CR4981: THE EFFECT OF ELECTRIC DISCHARGE MA-Generic Safety lasue CHINED NOTCHES ON THE FRACTI)RE TOUGHNESS OF SEVERAL NUREG-0933 S15. A PRIORITIZATION OF GENERIC SAFETY ISSUES STRUCTURAL ALLOYS.

NUREG-0933 S16; A PRIORlilZATION OF GENERIC SAFETY ISSUES.

NUREG/CR4997: CSNI PROJECT FOR FRACTURE ANALYSES OF LARGE-SCAL E INTERNAitONAL REFERENCE EXPERIMENTS Generic Safety issue 105 (PROJECT FALSIRE)

NUREG-1463: REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG/CR 6036: INITIAL RESULTS OF THE INFLUENCE OF BIAX1AL GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS OF-LOADING ON FRACTURE TOUGHNESS COOLANT ACCIDENT IN LIGHT. WATER REACTORS.

NUREG/CR-6098 LOADING RATE EFFECTS ON STRENGTH AND FRACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF Generic Safety lasue 106 THE IP1R'G PROGRAM NUREG-1364. REGULATORY ANALYSIS FOR THE RESOLUTION OF Fractured Media GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY NUREG/CR-5817 V02. NRC HIGH-LEVEL RADIOACTIVE WASTE RE-COMBUSilBLE GASES IN VIT AL AREAS.

SEARCH AT CNWRA Calendar Year 1991.

N G CR 6028 BIGFLOW: A NUMERICAL CODE FOR SIMULATING RE A CN RA Ja ary n 1992 NUREG/CR4817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE F'.OW IN VARIABLY SATURATED. HETEROGENEOUS GEOLOGIC RESEARCH AT CNWRA. July December 1992 MEDIA. Theory And User's Manual.- Version 1.1.

NUREG/CR4026 THEORETICAL AND EXPERIMENTAL INVESilGA-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY Ground Motion SATURATED, FRACTURED POROUS MEDIUM NUREG.1488 DRFT FC: REVISED LIVERMORE SFISMIC HAZARD ES-TIMATES FOR 69 NUCLEAR POWER PLANT SITES EAST OF THE Fractured Porous Media ROCKY MOUNTAINS Draft Report For Comrnent.

NUREG/CR 5991-PORFLOW. A MULTIFLUID MULTIPHASE MODEL FOR SIMULATING FLOW. HEAT TRANSFER. AND MASS TRANS-HEPA Filter DORT IN FRACTURED POROUS MEDIA. User's Manual Version NUREG/CR 6029 VOI: AGING ASSESSMENT OF NUCLEAR AIR-241.

TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS Phase L Fractured Rock HTGR Type Reactor NUREG/CP-0040 PROCEEDINGS OF WORKSHOP V: FLOW AND NUREG/CR.5983 SAFETY ASPECTS OF FORCED FLOW COOLDOWN TRANSPORT THROUGH UNSATURATED FRACTURED ROCK -- RE-TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSAL. Held At REACTORS Radisson Suite Hotel. Tucson, Anzona. January 7 10.1991 NUREG/CR-5984. CODE AND MODEL EXTENSIONS OF THE THATCH NUREG/CR 5917 V01. SENSITIVITY AND UNCERTAINTY ANALYSES CODE FOR MODUL AR HIGH TEMPERATURE GAS-COOLED REAC-APPLIED TO ONE-DIMENSIONAL RADIONUCLIDE TRANSPORT IN A TORS.

LAYERED FRACTURED ROCK MULTFRAC Analytic Solutions And Local Sensitmtes.

HVAC NUREG/CR4917 V02: SENSITIVITY AND UNCERTAINTY ANALYSES NUREG 1427. REGULATORY ANALYSIS FOR THE RESOLUTION OF APPLIED TO ONE-DIMENSIONAL RADIONUCLIDE TRANSPORT IN A GENERIC ISSUE 143. AVAILABILITY OF CHILLED WATER SYSTEM L AYERED FRACTURED RUCK. Evaluation Of The Limit State Ap-AND ROOM COOLING.

proach.

NUREG/CR4043 VOI: AGING ASSESSMENT OF ESSENTIAL HVAC Fuel Peak Cladding NUREG /CR4061' DETERMINATION OF THE BIAS IN LOFT FUEL HVAC Systern PEAK CLADDING TEMPERATURE DATA FROM THb BLOWDOWN NUREG/CR40fl4. VALUE-lMPACT ANALYSIS OF GENERIC ISSUE 143.

PHASE OF LARGE-BREAK LOCA EXPERtMENTS

" AVAILABILITY OF HEATING. VENTILATION, AIR CONDITIONING Fuel Performance (HVAC) AND CHILLED WATER SYSTEMS."

N G/CR-3950 V09 FUEL PEHFORMANCE ANNUAL REPORT FOR Health Effect NUREG/CR4214 RIP 2A2 HEALTH EFFECTS MODELS FOR NUCLE-Functional inspection AR POWER PLANT ACCIDENT CONSEOUENCE NUREG-1473: ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL IN.

ANALYSIS Modification Of Models Resulting From Addition Of Effects SPECTION (EDSFl) DATA BASE PROGRAM Of Exposure To Alpha Emitting RadionuclKies Part 11. Scientific Bases For Health..-.

Fundamental Nuclear Material Control NUREG/CR.4214 R2 PTI: HEALTH EFFECTS MODEL FOR NUCLEAR NUREGICR6118. ASSESSMENT OF THE EFFECTIVENESS OF THE POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part i LEU REFORM RULE AND ITS IMPLEMENTATION Introductioruntegralson,And Summary.

Gate Valve Health R6sk NUREG 1275 V09. OPERATING EXPERIENCE FEEDRACK REPORT -

NUREGICR 4883 HEALTH RlSK ASSESSMENT OF IRRADIATED PRESSURE LOCislNG AND THERMAL BINDING OF GATE

yopAz, VALVES. Commercial Power Reactors Heat Transfer nut 2 EGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 57. Ettects Of Fire Protection System Actuson On NUF EG/ R 5983 AFET AS CTS OF FORCED FLOW COOLDOWN Safety Related Equipment TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED Generic lasue 142 REACTORS' NUREG-1453 REGULATORY ANAL YSIS FOR THE RESOLUTION Of NUREG/CR 5984. CODE AND MODEL EXTE NSIONS OF THE THATCH GFNERIC ISSUE 147 LEAAAGE THROUGH ELECTRICAL ISOLA-CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLED REAC.

TORS IN INSTRUMENTATION CIRCUITS NUREG/CR4991: PORFLOW. A MULTIFLUID MULTIPHASE MODEL Generic issue 143 FOR SIMULATING FLOW. HEAT TRANSFER, AND MASS TRANS.

NUREG 1427. REGULATORY ANALYSIS FOR THE RESOLUTION OF PORT IN FRACTURED POROUS MEDIA. User's Manual - Version GENERIC 15SUF 143. AVAIL ABILITY Or Chile.ED WATE R SYSTEM 2 41-AND ROOM COOLING NUREG/CR t084 V ALUE.lMPACT ANALYSIS OF GENERIC ISSUE 143.

Heat Treatment

" AVAILABILITY OF HEATING. VENTIL ATION. AIR CONDITIONING NUREGICR 5972 EFFECTS OF NONSTANDARD HEAT TREATMENT -

(HVAC) AND CHILLED WATER SYSTEMS "

TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES OF CARDON STEEL CASTING REPAIR W'ELOS.

NUREG 1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF Heavy-Section steellrradiation Program GENERIC ISSUE 153 LOSS OF ESSENTIAL SERVKT WATER IN NUREG/CR4591 V01 N2: HEAVY-SECTION STEEL IRRADIATION LWHS PROGRAM Sem4 annual Progress Report for Apr0Septomber 1990.

84 Subject index Heavy SecHon Steel Technology Program Hydraulic Transport NUREG/CR-4219 V09 N2 HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR 5943: SENSITIVITY ANALYSIS AND BENCHMARKING OF PROGRAM Semiannual Progress Report For Apnt-September 1992 THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

Heterogeneous Soll Hydro 9en NUREG/CR-5994 SIMULATION OF UNSATURATED FLOW AND NON-NUREG/CR-6060. HYDROGEN MIXING STUDIES (HMS) ASSESSMENT REACTIVE SOLUTE TRANSPORT IN A HETEROGENEOUS SOIL AT MANUAL THE FIELD SCALE.

Hydrogen-Air UE/

7 N UTRON SPECTRA AT DIFFERENT HIGH FLUX ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEtLLANCE N AT LM SCAE. "

N LOCATIONS.

Hydrogeology High Pressure Coolant injection System NUREG/CR-5817 V02; NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREG/CR-5933: HIGH PRESSURE COOLANT INJECTION (HPCI)

SEARCH AT CNWRA Calendar Year 1991.

SYSTEM RISK-BASED INSPECTION GUIDE FOR DRESDEN NUCLE.

NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE AR POWER STATION UNITS 2 ANO 3.

NUREG/CR-5934: HIGH PRESSURE COOLANT INJECTION (HPCI)

RESEARCH AT CNWRA. January June 1992.

NUREG/CR-5817 V03 N2; NRC HIGH-LEVEL RADIOACTIVE WASTE SYSTEM RISK-BASED INSPECTION GUIDE FOR QUAD-CITIES RESEARCH AT CNWRA. July-December 1992.

ST ATION, UNITS 1 AND 2 NUREG/CR 5959-HIGH PRESSURE CDOLANT INJECTION (HPCI)

SYSTEM RISK BASED INSPECTION GUIDE FOR ENRICO FERMI ICAP Program NUREG/lA-0090: ASSESSMENT OF RELAP5/ MOO 2 USING THE TEST ATOMIC POWER PLANT, UNIT 2 NUREG/CR4014 HIGH PRESSURE COOLANT INJECTION SYSTEM DATA OF REWET il REFLOODfNG EXPERfMENT SGl/R.

RISK-BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER NUREGAA 0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NATU-STATION RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT NUREG/CR 4022: HIGH PRESSURE COOLANT INJECTION (HPCI)

DORSSELE-SYSTEM RISK BASED INSPECTION GUIDE FOR BROWNS FERRY NUREG/lA-0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE NUCLEAR POWER STATION AGAINST THE NET LOAD TRIP TEST DATA FROM YONG-GWANG. UNIT 2.

High Pressure Melt E}ection NUREG/lA-0094: ASSESSMENT OF RELAPS/ MOD 3 AGAINST NUREG/CR-5949-ASSESSMENT OF THE POTENTIAL FOR HIGH TWENTY-FIVE POST DRYOUT EXPERIMENTS PERFORMED AT THE PRESSURE MELT EJECTION RESULTING FROM A SURRY STATION ROYAL INSTITUTE OF TECHNOLOGY.

BLACKOUT TRANSIENT.

NUREGAA-0095: RELAPS ASSESSMENT USING LSTF TEST DATA SB-CL 18 R

Am N S AND WMMAMN & M W G 1 23 V 4 COM LATION OF REPORTS OF THE ADVISORY COMMITTEE ON NUCLEAR WASTE.Jul{1992 June 1993 RADIOACTIVE WASTE RE-SffRCH AT N E/

-0 9

L 5 ASSESSMENT USING SEMISCALE SBLOCA A

Y 9R TEST S-NH 1.

NUREG/CR-5817 V03 N1; NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/lA 0100; ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 RESFARCH AT CNWRA. Janua June 1992.

AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS.

NUREG/CR4817 V03 N2: NRC llGH-LEVEL RADIOACTIVE WASTE NUREGAA-0103: ASSESSMENT OF DETHSY TEST 91.8 USING RESEARCH AT CNWRA July-December 1992.

RELAP5/ MOD 3.

H6gh-Level Radloactive Waste Disposal NUREGAA-0104: RELAP5/ MOD 3 ASSESSMENT USiNG THE SEMIS.

NUREG/CP 0040 PROCEEDINGS OF WORKSHOP V: FLOW AND CALE 50% FEED LINE BREAK TEST S-FS-11.

TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE-NUREGAA-0105: ASSESSMENT OF RELAPS/ MOD 3 VERSION SMS LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSAL. Held At USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF Radsson Susie Fbtel. Tucson Anzona, January 7 10,1991.

KORI UNIT 3 PLANT.

NUREG/CR.6026: THEORETidAL AND EXPLRIMENTAL INVESTIGA.

NUREG/lA4106: ASSESSMENT OF PWR STEAM GENERATOR MOD-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY ELLING IN RELAPS/ MOD 2.

NUREGAA-0107: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD SATURATED, FRACTURED POROUS MEDIUM.

REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS 11 High-Level Waste Repository NUCLEAR POWER PLANT, NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL-NUREG/lA-0110: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A MAIN HYDROLOGIC-MECHAN ICAL CHEMICAL PROCESSES PERTINENT FEEDWATER TURBOPUMP TRIP TRANSIENT IN THE VANDELLOS 11 TO THE PROPOSED HtGH~ LEVEL WASTE REPOSITORY AT YUCCA NUCLEAR POWER PLANT.

MOUNT AIN.

NUREG/lA-0112: ASSESSMENT OF RELAP5/ MOD 2 AGAINST ECN-RE-FLOOD EXPERIMENTS.

Human Performance NUREG/lA 0113; PRELIMINARY ASSESSMENT OF PWR STEAM GEN-NUREG/CR 5455 V01: DEVELOPMENT OF THE NRC'S HUMAN PER-ERATOR MODELLING IN RELAP5/ MOD 3.

FORM ANCE INVESTIGATION PROCESS (HPIP)

NUREGAA4116: ASSESSMENT OF RELAP5/ MOD 3/V5M5 AGAINST NUREG/CR 5455 V02' DEVELOPMENT OF THE NRC'S HUMAN PER-THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWR FORMANCE INVESTIGATION PROCESS (HPIP)

HOT LEG)

NUREG/CR-5455 VOi DEVELOPMENT OF THE NRC'S HUMAN PER-NUREG AA-0118: ANALYSIS OF LOFT TEST LS 1 USING RELAPS/

FORMANCE INVESilGATION PROCESS HPIP NUREG 1A4119' ASSESSMENT AND APPLICATION OF CLACKOUT E ATING f VENTS 9 9 '

TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/

MOD 2.

Human Reliability Analyslo NUREGAA-0120: ASSESSMENT OF THE TURBINE TRIP TRANSIENT NUREG/CH-4832 VOS ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR IN COFRENTES NPP WITH TRAC-BFf.

POWER PLANT; RISK METHODS INTEGRATION AND EVALUATION NUF;EG AA-0121: ASSESSMENT OF A PRESSURIZER SPRAY VALVE PROGRAM. Parameter Eshmahon Analyses And Screening Human Reh.

FAULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT atnhty Analysis WITH RELAPS/ MOD 2.

NUREGAA 0122: ASSESSMENT OF MSIV FULL CLOSURE FOR Human-Machine Interface NUREG/CR $977: A PERFORMANCE INDICATOR OF THE EFFECTIVE-SANT A MARIA DE GARONA NUCLEAR POWER PLANT USING NESS OF HUMAN-MACHINE INTERFACES FOR NUCLEAR POWER TRAC-BF1 (G1J1).

NUREG AA4123: APPLICATION OF FULL POWER BLACKOUT FOR PLANTS C N. ALM ARAZ WITH RELAPG/ MOD 2.

NUREGAA 0124: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A PRES-Hurricane Andrew NUREG 1474 EFFECT OF HURRICANE ANDREW ON THE TURKEY SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN-PONT NUCLE AR GENERATING STATION FROM AUGUST 20-30.

SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA-1992.

BRERA NUCLEAR STATION.

Subject Index 85 NUREG/lA4125; ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE NUREG/CR-5836: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG-SPECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER GWANG UNIT 2.

PLANT.

NUREG/lA-0128: INTERNATIONAL CODE ASSESSMENT AND APPLl-NUprG/CR-5897: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

CATIONS PROGRAM:

SUMMARY

OF CODE ASSESSMENT STUDIES S '_CTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR CONCERNING RELAP5/ MOD 2, RELAPS/ MOD 3, AND TRAC-D.

POWER PLANT.

NUREG/CR-5898: AUXILIARY FEEDWATER SYSTEM R:SK BASED IN.

IPIRG Program SPECTION GUIDE FOR THE POINT BEACH NUCLEAR POWER NUREG/CR4098: LOADING RATE EFFECTS ON STRENGTH AND PLANT.

FRACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF NUREG/CR-5933; HIGH PRESSURE COOLANT INJECTION (HPCI)

THE iPIRG PROGRAM.

SYSTEM RISK-BASED INSPECTION GUIDE FOR DRESDEN NUCLE-IRRAS AR POWER STATION UNITS 2 AND 3.

NUREG/CR-5934: HIGH PRESSURE COOLANT INJECTION (HPCI)

NUREG/CR-5964:

SAPHIRE TECHNICAL REFERENCE SYSTEM RISK-BASED INSPECTION GUIDE FOR QUAD-CITIES MANUAL:lRRAS/ SARA VERSION 4.0.

STATION. UNITS 1 AND 2.

ISLOCA NUREG/CR 5959: HIGH PRESSURE COOLANT INJECTION (HPCI)

NUREG-1463: REGULATORY ANALYSIS FOR THE RESOLUTION OF SYSTEM RISK BASED INSPECTON GUIDE FOR ENRICO FERMI

^

NUREG/ R40 : H E U'RE COOLANT INJECTION SYSTEM ATA DENT N LIGHT T R RF O S' RISK-BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER ISLOCA Research Program STATION NUREG/CR 5928: ISLOCA RESEARCH PROGRAM Final Report.

NUREG/CR 6022: HIGH PRESSURE COOLANT INJECTION (HPCI)

SYSTEM RISK-BASED INSPECTION GUIDE FOR BROWNS FERRY Inadvertent Safety injection NUCLEAR POWER STATION.

NUREG/lA-0105: ASSESSMENT OF RELAPS/ MOO 3 VERSION 5MS USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF inspector Qualification And Training KORI UNIT 3 PLANT.

NUREG/CP-0128: PROCEEDINGS OF THE INTERNATIONAL WORK.

SHOP ON THE CONDUCT OF INSPECTIONS AND INSPECTOR Industrial Radiography OVALIFICATION AND TRAINING.

NUREG4713 V12-OCCUPATIONAL RADIATION EXPOSURE AT COM-MERCIAL NUCLEAR POWER REACTORS AND OTHER instrument FACILITIES,1990 Twenty-Thad Annual Report.

NUREG/CFI-6062. PERFORMANCE OF PORTABLE RADIATON SURVEY INSTRUMENTS.

Infiltration NUREGICR 6114 V01: APPLICATION OF AN INFILTHATION EVALUA.

Instrumentation And Control System TION METHODOLOGY TO A HYPOTHETICAL LOW-LEVEL WASTE NUREG/CP-0134: INTERNATIONAL ATOMIC ENERGY AGENCY SPE.

DISPOSAL FACILITY.

CIALISTS MEETING ON EXPERIENCE IN AGING. MAINTENANCE, AND MODERNIZATION OF INSTRUMENTATION AND CONTROL Informst!on Circular SYSTEMS FOR IMPROVING NUCLEAR POWER PLANT NUREG-0725 R09: PUBLIC INFORMATION CIRCUI.AR FOR SHIP-MEN 1S OF IRRADIATED REACTOR FUEL AVAILABILITY. Held At Rockville,MD.May 5-7,1993.

Instrumentation Chanrd Information Digest NUREG-1350 VOS: NUCLEAR REGULATORY COMMISSION INFORMA-NUREG/CR-5903: VALIDATION OF SMART SENSOR TECHNOLOGIES TON DIGEST.1993 Edihon.

FOR INSTRUMENT CALIBRATION REDUCTION IN NUCLEAR POWER PLANTS.

Informstlon Technology Strategic Plan Instrumentation Circuit NUREG-1487 V01: FISCAL YEAR 1994-1998 INFORMATION TECHNOL-OGY STRATEGIC PLAN.

NUREG-1453. REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 142. LEAKAGE THROUGH ELECTRICAL ISOLA-Initial Notmcation TORS IN INSTRUMENTATION CIRCulTS.

NURE G-1467:

FEDERAL GUIDE FOR A

RADIOLOGICAL

" 9" N

G/C 5

Pt NTEGRATED RISK ASSESSMENT FOR THE ru I Of A aA ent.

LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And inservice Testing Risk Uncertainty Evaluaton Program (PRUEP) Appendices A.C.

NUREG-1482 DRFT FC: GUIDELINES FOR INSERVICE TESTING AT NUREG/CR-5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR THE NUCLE AR POWER PLANTS. Draft Report For Comment.

LASALLE UNIT 2 NUCLEAR POWER PLANT:Phenomenology And Risk Uncertatnty Evaluston Program (PRUEP). Appendices D-G.

Inspection NUREG/CP-Ot?8-PROCEEDINGS OF THE INTERNATIONAL WORK.

Integrated System SHOP ON THE CONDUCT OF INSPECTIONS AND INSPECTOR NUREG/CR-6111: INTEGRATED SYSTEMS ANALYSIS OF THE PlVS OUALIFICATION AND TR.AINING.

REACTOR.

Inspection Guidance Interfacing System NUREG/CR-5829 AUXlLIARY FEEDWATER SYSTEM RISK-BASED IN.

NUREG/CR-5928 ISLOCA RESEARCH PROGRAM Final Report SPECTION GUIDANCE FOR THE DAVIS BESSE NUCLEAR POWER

PLANT, internal Fire Analysis NUREG/CR-4832 V09: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR inspection Guide POWER PLANT: RISK METHODS INTEGRATION AND EVALUATON NUREG/CR-5488. RISK BASED INSPECTION GUIDE FOR THREE MILE PROGRAM (RMIEP). Internal Fwe Analysis.

ISLAND NUCLE AR STATION UNIT 1.

NUREG/CR-5766: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-International Atomic Energy Agency SPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR NUREG/CP 0134: INTERNATIONAL ATOMIC ENERGY AGENCY SPE-POWER PLANT.

CIALISTS MEETING ON EXPERIENCE IN AGING. MAINTENANCE.

NUREG/CR-5833. AUXILIARY FEEDWATER SYSTEM RISK-DASED IN.

AND MODERNIZATION OF INSTRUMENTATION AND CONTROL SPECTION GUIDE FOR THE H 8. ROBINSON NUCLEAR POWER SYSTEMS FOR IMPROVING NUCLEAR POWER PLANT PLANT.

AVAILABILITY. Held At Rockville MO,May5-7,1993.

NUREG/CR-5834. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER Investigation Process Pt. ANT NUREG/CR-5455 V01: DEVELOPMENT OF THE NRC'S HUMAN PER-NUREG/CR-5835: AUXILIARY FEEDWATER SYSTEM RISK. BASED IN-FORMANCE INVESTIGATION PROCESS (HPIP).

SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NU-NUREG/CR-5455 V02. DEVELOPMENT OF THE NRC*S HUMAN PER.

CLEAR POWER PLANTS.

FORMANCE INVESTIGATION PROCESS (HPIP)

)

86 Subject Index NUREG/CR-5455 V03 DEVELOPMENT OF THE NRC'S HUMAN PER.

NUREG-0750 V36 NO1: NUCLEAR REGULATORY COMMISSION IS-FORMANCE INVESTIGATION PROCESS (HPlP).

SUANCES FOR JULY 1992 Pages 1-45.

NUREG 0750 V36 NO2: NUCLEAR REGULATORY COMMISSION IS-tridium-192 SUANCES FOR AUGUST 1992. Pages 47148.

NUREG-1480: LOSS OF AN IRIDIUM-192 SOURCE AND THERAPY MIS-NUREG-0750 V36 NO3: NUCLEAR REGULATORY COMMISSION IS-ADMIN!STRATION AT INDIANA REGIONAL CANCER SUANCES FOR SEPTEMBER 1992. Pages 149-220.

CENTER, INDIANA. PENNSYLVANIA.ON NOVEMBER 16,1992-NUREG-0750 V36 N04: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR OCTOBER 1992. Pages 221249.

NUREG-0750 V36 N05: NUCLEAR REGULATORY COMMISSION IS-UF EG/CR 83 HEALTH RISK M ESSMENT OF IRRADIATED SUANCES FOR NOVEMBER 1992. Pages 251-350.

TOPAZ' NUREG-0750 V36 N06: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR DECEMBER '992. Pages 351396.

Isolation Device NUREG/CR-5863 RISK ASSESSMENT OF ISOLATION DEVICES IN NUREG-0750 V37101: INDEAdS TO NUCLEAR REGULATORY COM-SAFETY SYSTEMS MISSION ISSUANCES January-March 1993.

NUREG-0750 V37102: INDEXES TO NUCLEAR REGULATORY COM-LOCA MISSION ISSUANCES Januarydune 1993.

NUREG/CR 5818 UNCERTAINTY ANALYSIS OF MINIMUM VESSEL NUREG-0750 V37 Not: NUCLEAR REGULATORY COMMISSION IS-LIQUID INVENTORY DURING A SMALL-BREAK LOCA IN A B&W SUANCES FOR JANUARY 1993. Pages 1-54.

PLANT-AN APPLICATION OF THE CSAU METHODOLOGY USING NUREG-0750 V37 NO2: NUCLEAR REGULATORY COMMISSION IS-THE RELAPS/ MOD 3 COMPUTER CODE SUANCES FOR FEBRUARY 1993. Pages 55-134, NUREG/lA 0126. 2D/3D PROGRAM WORK

SUMMARY

REPORT NUREG-0750 V37 NO3: NUCLEAR REGULATORY COMMISSION IS-NUREG/lA-012?: REACTOR SAFETY ISSUES RESOLVED BY THE 2D/

SUANCES FOR MARCH 1993. Pages 135 249.

3D PROGRAM.

NUREG-0750 V37 N04. NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR APRll 1993.Pages 251-354.

LOFT 50 V37 NE WCM MWM @MM G NUREG/CR 6061 DETERMINATION OF THE BIAS IN LOFT FUEL PEAK CLADDING TEMPER TURE DATA FROM THE BLOWOOWN NUREG 50 7N AR REGULATORY COMMISSION IS-PHASE OF LARGE BREAM LOCA EXPERIMENTS.

SUANCES FOR JUNE 1993.Pages 419-515.

NUREG 0750 V38 NO1: NUCLEAR REGULATORY COMMISSION IS-LOFT Test L5-g NUREG/lA-0118 ANALYSIS OF LOFT TEST LS-1 USING RELAP5/

SUANCES FOR JULY 1993.Pages 124.

NUREG-0750 V38 NO2: NUCLEAR REGULATORY COMMISSION IS-MOD 2 SUANCES FOR AUGUST 1993. Pages 25-79.

LSTF Data $8-CL-18 NUREG-0750 V38 NO3; NUCLEAR REGULATORY COMMISSION IS-NUREG/lA-0095. RELAPS ASSESSMENT USING LSTF TEST DATA SB-SUANCES FOR SEPTEMBER 1993. Pages 81 168.

CL 18 NUREG 0750 V38 N04: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR OCTOBER 1993. Pages 169-186.

5 LWR NUREG-1461 REGULATORY ANALYSIS FOR THE RESOLUTION OF Licensed Fuel facility Status Report GENERIC ISSUE 153. LOSS OF ESSENTIAL SERVICE WATER IN NUREG 0430 V12:. LICENSED FUEL FACILITY STATUS

~

NUF G 1463 REGULATORY ANALYSIS FOR THE RESOLUTION OF i

GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF-COOLANT ACCIDENT IN LIGHT. WATER REACTORS.

Licensed Operating Reactor NUREG/CR-4469 V15' NONDESTRUCTIVE EXAM 4 NATION (NDE) RELi-NUREG 0020 V17: LICENSED OPERATING REACTORS STATUS SUM-ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER MARY REPORT. Data As Of December 31,1992,(Gray Book 1) -

REACTORS Semiannual Report. October 1991 March 1992 NUREG/CR-4469 V13. NONDESlKUCTIVE EXAMINATION (NDE) REll-Light Water Reactor ADILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG-1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 153: LOSS OF ESSENTIAL SERVICE WATER IN NUREG CR 466 1 L ON T

YA S ACKING IN NUREG 1463: REGULATORY ANALYSIS FOR THE RESOLUTION OF NU EG C 4 67 V1 E RO E LY S RA N i GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF-LIGHT WATER REACTORS Semiannual Report. October 1992 March COOLANT ACCIDENT IN LIGHT WATER REACTORS.

1993 NUREG/CR-4744 V07 N1: LONG-TERM EMBRITTLEMENT OF CAST NUREG/CR-4469 V15 NONDESTRUCTIVE EXAMINATION (NDE) REll-DUPL EX STAINLESS STEELS IN LWR SYSTEMS Semiannual ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Report. October 1991 - March 1992 REACTORS Semiannual Report,0ctober 1991 March 1992.

NUREG/CR-4744 V07 N2. LONG-TERM EMBRITTLEMENT OF CAST NUREG/CR-4469 V16. NONDESTRUCTIVE EXAMINATION (NDE) RELi-DUPLEX STAINLESS STEELS IN LWR SYSTE MS Semiannual ABILITY FOR INSERVICE INSPECTION' OF LIGHT WATER Report,Aptil September 1992.

REACTORS Semiannual Report, April 1992-September 1992.

NUREG/CR 5642 LIGHT WATER REACTOR LOWER HEAD FAILURE NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN ANALYSIS LIGHT WATER REACTORS. Semeannual Report.Apni-September 1992-NUREG/CR-5999. INTERIM FATIOUE DESfGN CURVES FOR CARBON, NUREG/CR-4667 V16. ENVIRONMENTALLY ASSISTED CRACKING IN LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVb LIGHT WATER REACTORS Semlannual Report,0ctober 1992 March RONMENTS-1993 NUREG/GR-0005 V02 P1 RISK-BASED INSPECTION-DEVELOPMENT NUREG/CR-4744 V07 N1: LONG TERM EMBRITTLEMENT OF CAST OF GUIDELINES Light Water Reactor (LWR) Nuclear Power Plant DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual Components Report. October 1991 March 1992.

NUREG/CR-4744 V07 N2: LONG TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Serniennual U G/CR

. DETERM! NATION OF THE BIAS IN LOFT FUEL PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN NURE 2 LIG A ER REACTOR LOWER HEAD FAILURE PHASE OF LARGE BRE AK LOCA EXPERIMENTS ANALYSIS.

NUREG/CR-5999' INTERIM FATIOUE DESIGN CURVES FOR CARBON.

g,,g,g, NUREG 1453 REGULATORY ANALYSIS FOR THE RESOLUTION OF LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI-GENERIC ISSUE 142: LEAKAGE THROUGH ELECTRICAL ISOLA.

RONMENTS.

NUREG/GR4005 V02 P1: RISK BASED INSPECTION-DEVELOPMENT TORS IN INSTRUMENTATION CIRCUlTS.

OF GUIDELINESbght Water Reactor (LWR) Nuclear Power Plant legal lesuances Components NUREG 0750 V36101 INDEXES TO NUCLEAR REGULATORY COM-M!SSION ISSUANCES Juiv-September 1992 Liner NUREG-0750 V36102, INDEXES TO NUCLEAR REGULATORY COM-NUREG/CR-6025. THE PROBABILITY OF MARKl CONTAINMENT MISSION ISSUANCES. Jury-December 1992 FAILURE BY MELT ATTACK OF THE LINER.

j l

l o

-n-

s.

Subject Index 87 LJguld Control System NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS-NUREG/CR-5951 THE MANAGEMENT OF ATWS Bf DORON INJEC-MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE TION DISPOSAL FACILITIES Evaluation Of Modehng Approachos.

Liquidus Temperature NUREG/CR 6070; MODELING APPROACHES FOR CONCRETE BAR+

Rif RS USED IN LOW-LEVEL WASTE DISPOSAL-NUREG/CR 6032. SOLIDUS AND LIQUIDUS TEMPERATURES OF CORE-CONCRETE MIXTURES NUREG/CR4114 V01: APPLICATION OF AN INFILTRATION EVAL UA-TlON METHODOLOGY TO A HYPOTHETICAL LOW-LEVEL WASTE l

Load Fle}ection DISPOSAL FACILITY-NUREG/lA-0107-ASSESSMENT OF RELAP5/ MOD 2 AGAINST #1 LOAD l

i REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS 11 Low-Level Waste Data Base i

NUCLE AR POWER PLANT.

NUREG/CR 5229 VOS. FIELD LYSIMETER INVESTIGATIONS: LOW.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR Load Rejection Tranelent FISCAL YE AR 1992 Annual Reprt.

NUREG/lA 0109 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10%

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE Low-power

?

VANDELLOS il NUCLEAR POWER PLANT NUREG 1449. SHUTDOWN AND LOW-POWER OPERAllON AT NU-Load Tdp CLEAR POWER PLANTS IN THE UNITED STATES Final Report.

NUREG/lA-0092. ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE U 8 '**

AGAINST THE NET LOAD TRIP TEST DATA FROM YONG.

GW ANG. UNIT 2.

NUREG/CR-5229 V05 FIELD LYSIMETER INVESTIGATIONS. LOW-LE VEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR Loading Rate FISCAL YEAR 1992 Annual Reprt.

NUREG/CR-6098 LOADING RATE EFFECTS ON STRENGTH AND NUREG/CR 6073 LYSIMETER LITERATURE REVIEWS FRAC 1URE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF THE IPIRG PROGRAM M ACCS Version 1 NUREG/CR-6059 MACCS VERSION 1511.1 A MAINTENANCE RE-N R G/

6 90 THE PROGRAMMABLE LOGIC CONTROLLER AND ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS '

MARK l Long Term Performance NUREG/CR-6025: THE PROBASILITY OF MARK 1 CONTAINMENT NUREG/CR-5851. LONG TERM PERFORMANCE AND AGING CHAR.

FAILURE BY MELT-ATTACK OF THE LINER ACTERISTICS OF NUCLEAR PLANT PRESSURE TRANSMITTERS.

MELCOR Code Losa-Of-Coolant NUREG/CR-5942' SEVERE ACCIDENT SOURCE TERM CHARACTER-NUREG/CP-0126 V01: PROCEEDINGS OF THE TWENTIETH WATER ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED REACTOR SAFETY INFORMATION MEETING DY THE MELCOR CODE NUREG/CP 0126 V02. PROCEEDINGS OF THE TWENTIETH WATER REACTOR SAFETY INFORMATION MEETING MHTGR NUREG/CP-Ot26 V03: PROCEEDINGS OF THE TWENTIETH WATER REACTOR SAFETY INFORMATION MEETING.

NUREG/CR 5922 MODULAR HIGH TEMPERATURE GAS COOLED RE.

ACTOR SHORT TERM THERMAL RESPONSE TO FLOW AND REAC.

Loss-Of-Coolant Accidant TIVITY TRANSIENTS NUREG 1463. REGULATORY ANALYSIS FOR THE RES' LUTION OF O

GENERIC SAFETY ISSUE 105 INTERFACING SYSTEM LOSS OF.

MSIV COOLANT ACCIDENT IN LIGHT-WATER REACTORS.

NURE G/l A-0122<. ASSESSMENT OF MSIV FULL CLOSURE FOR NUREG/CR 5942: SEVERE ACCIDENT SOURCE TERM CHARACTER-SANTA MARIA DE GARONA NUCLE AR POWER PLANT USING ISTICS FOR SELECTED PE ACH DOTTOM SEQUENCES PREDICTED TRAC DF 1 (GIJt t BY THE MELCOR CODE.

Minimum Vessel Liquid Low May Sted NUREG/CR 5818 UNCERTAINTY ANALYSIS OF M!N! MUM VESSEL NUREG/CR 5926. SANS INVESTIGATION OF LOW ALLOY STEELS IN LIQUlO INVENTORY DURING A SMALL.RREAK tOCA IN A D&W NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI-PLANT-AN APPLICATION OF THE CSAU METHODOLOGY USING TiONS THE RELAP5/ MOD 3 COMPUTER CODE Low Upper Shelf NUREG/CR 6023' GENERIC ANALYSIS FOR EVALUATION OF LOW Multiphase Transport CHARPY UPPERSHELF ENERGY EFFECTS ON SAFETY MARGINS NUREG/CR 5991: PORf LOW A MULTIFLUID MULTIPHASE MODEL AGAINST F RACTURE OF REACTOR PRESSURE VESSEL' MATERI.

FOR SIMULATING FLOW HEAT TRANSFER. AND MASS TRANS-ALS PORT IN FRACTURED POROUS MEDIA User's Manual - Vmsion 2 41.

Low 4nriched Uranium NUREG/CR 6118 ASSESSMENT OF THE EFFECTIVENESS OF THE NRC Regulation LEU REFORM RULE AND ITS IMPLEMENTATION.

NUREG 1479-RESULTS FROM TWO WORKSHOPS. STATE RADI-ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG.

Low-Level Radioactive Waste ULATIONS AND IUNDING NUREG-1423 V04 A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON NUCLEAR WASTE Jul 1992 June 1993 NUHEG/CR 5672 V03 CHARACTERISTdS OF LOW-LEVEL RADIOAC-NRC Research Program TlVE DECONTAMINATION' WASTE Annual Report For Fmcal Yea, NUREG-1377 R04 NRC RESEARCH PROGRAM ON PLANT AGING' 3997 LISTING AND SUMMARIES OF REPORTS ISSUED THROUGH SEP-NUHEG/CR-5938 NATIONAL PROFILE ON COMMERCIALLY GENER-TEMBER 1993 ATED LOW LEVEL RADIOACTIVE MIXEO WASTE NUREG/CR 5943. SENSITIVITY ANALYSIS AND BENCHMARKING OF Natural Circulation THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

NUREG/lA-0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NATU-NUREG/CR 598L MICROBIAL,1NFLUENCED CEMENT DEGRADATION RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT

- LITERATURE REVIEW BORSSEL E NUREGICA 5988 SOIL CHARACTERIZATION METHODS FOR UN-NUREG/lA-0124 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES.

NUPE 41 i

AL T

RCE TERM (DUST) DATA SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN.

INPUT GUIDE' SIENT AND RECOVERY BY NA1 URAL CIRCULATION IN JOSE CA-BRERA NUCLEAR STATION Low Level Rad 6oartive Waste Disposal NUREG/lA-0125 ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE NUREG/CR 5911-SOURCE TERM EVALUATION FOR RADIOACTIVE AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG-LOW-LEVEL WASTE DISPOSAL PERrORMANCE ASSESSMF NT GW ANG UNIT 2

=

88 Subject index Nuclear Waste Naturnity Occurring NUREG/CR 5962: HEALTH AND SAFETY IMPACTS FROM DISCRETE NUREG/CR-4735 V08: EVALUATION AND COMPILATION OF DOE SOURCES OF NATURALLY-OCCURRING AND ACCELERATOR-PRO-WASTE PACKAGE TEST DATA. Biannual Report,Aegust 1989 Janu-DUCED RADtOACTIVE MATERIALS (NARM).

ary 1990.

Nemaha Uplift Occupational Dose Reduction NUREG/CR 6034 OKLAHOMA SEISMIC NETWORK. Final Report NUREG/CR 3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-CLEAR POWER PLANTS: ANNOTATED DIDLIOGRAPHY OF SELECT-Neural Network ED READINGS IN RADIATION PROTECTION AND ALARA.

NUREG/GR4010; HYDRID DIGITAL SIGNAL PROCESSING AND NEURAL NETWORKS FOR AUTOMATED DIAGNOSTICS USING NDE Occupational Radiation Exposure NUREG-0713 V13: OCCUPATIONAL RADIATION EXPOSURE AT COM-METHODS MERCIAL NUCLEAR POWER REACTORS AND OTHER Neutron Kinetic FACILITIES,1991. Twenty-Fourth Annual Report.

NUREGICR-5922 MODULAR HIGH TEMPERATURE GAS-COOLED RE-NUREG-0713 V14: OCCUPATIONAL RADIATION EXPOSURE AT COM-ACTOR SHORT TERM THERMAL RESPONSE TO FLOW AND REAC.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES TiVITY TRANSIENTS-1992. Twenty-Fifth Annual Report NUREG/CR-6050: RADIATION EXPOSURE MONITORING AND INFOR.

Nondestructive Evaluation MATION TRANSMITTAL (REMIT) SYSTEM. User's Manual.

NUREG/CR-5410 STATISTICALLY BASED REEVALUATION OF PISC-Il ROUND ROBIN TEST DATA.

Office Of The inspector General NUREG/CR 6052: METHOOOLOGY FOR RELIABILITY BASED CONDI-NUREG-1415 V05 NO2: OFFICE OF THE INSPECTOR TION ASSESSMENT. Apphcaton To Concrete Structures in Nuclear GENERALSemiannual ReportOctober 1,1992 March 31,1993.

Planta.

NUREG-1415 V06 N01: OFFICE OF THE INSPECTOR GENE RAL. Semiannual Report,Aprd 1,1993 - September 30,1993.

Nondestructive Examination NUREG/CR-4469 V15, NONDESTRUCTIVE EXAMINATION (NDE) REll-Official Record ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG4910 R02 S01: NRC COMPREHENSIVE RECORDS DISPCsi-REACTORS Semiannual Report,0ctober 1991 March 1992.

TION SCHEDULE.

NUREG/CR-4460 V16 NONDESTRUCTIVE EXAMINATION (NDE) REll-ABILITY FOR 6NSERVICE INSPECTION OF LIGHT WATER Operating Events REACTORS Semiannual Report.Aprd 1992 September 1992-NUREG/CR49S3; STUDIES OF HUMAN PERFORMANCE DURING OP-NUREG GR4010: HYDRID DIGITAL SIGNAL PROCESSING AND ERATING EVENTS.1990-1992, NEURAL NETWORKS FOR AUTOMATED DIAGNOSTICS USING NDE METHODS.

Operating Experience NUREG 1272 V07 N01: OFF6CE FOR ANALYSIS AND EVALUATION OF Nonisothermal Flow OPERATIONAL DATA.1992 Annual Report - Power Reactors.

NUREG/CP 0040: PROCEEDINGS Or WORKSHOP V: FLOW AND NUREG 1272 V07 NO2: OFFICE FOR ANALYSIS AND EVALUATION OF TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE-OPERATIONAL DAT A.1992 Annual Report. Nonreactors.

LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSALHold Al NUREG/CR 5404 V02: AUXILIARY FEEDWATER SYSTEM AGING Radisson Suite Hotal. Tucson. Anzona, January 7 10,1991.

STUDY. Phase i Foflow-On Study.

NUREG/CR-6043 VOI: AGING ASSESSMENT OF ESSENTIAL HVAC CHILLERS USED IN NUCLEAR POWER PLANTS. Phase L Notch Bend NUREG/CR-5969 J AND CTOO ESTIMATION EQUATIONS FOR SHAL-LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS, Operating Experience Feedback Report NUREG-1275 V09: OPERATING EXPERIENCE FEEDBACK REPORT.

Nuclear Air Cleaning PRESSURE LOCKING AND THERMAL BINDING OF GATE NUREG/CP-0130 VO1: PROCEEDINGS OF THE 22ND DOE /NRC NU-VALVES. Commercial Power Reactors.

CLEAR AIR CLEANING CONFERENCE. Sessions 1-8tiold in Denver. Colorado, August 24 27,1992.

Operational Event NUREG/CP 0130 V02: PROCEEDINGS OF THE 22ND DOE /NRC NU-NUREG/CR-5936. ENHANCEMENTS TO THE ACCIDENT PRECURSOR CLEAR AIR Ct EANING CONF ERENCE.Sessms 9-16, Held in METHODOLOGY.

Denver. Colorado,A gust 24 27,1992.

Operator Action NUREG/CR-6065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR.

Nuclear Piping System NUREG/CR 5358 REVIEW OF ASME CODc CRITERIA FOR CON 1ROL OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM DRANCH CON.

Operator Licensing NUREG-1021 R07 OPERATOR LICENSING EXAMINER STANDARDS.

NECTIONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP.

MENT WORK.

Organization Chart NUREG-0325 R16: U.S NUCLEAR REGULATORY COMMISSION FUNC.

Nuclear Power Pfant Accident NUREG/CR-4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR TIONAL ORG ANtZATION CHARTS. March 15,1993.

POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part 1:

P!SC-Il Introduction,Integraton,And Surnmary.

NUREG/CR 5410: STATISTICALLY BASED REEVALUATION OF PISC-Il ROUND ROBtN TEST DATA.

Nuclear Reactor Shutdown NUREG/CR4080. REPLACEMENT ENERGY, CAPACITY. AND REll-PlVS Reactor ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT, NUREG/CR-6111: INTEGRATED SYSTEMS ANALYSIS OF THE PlVS DOWNS REACTOR.

Nuclear Regulatory Legislation PORFLOW NUREG-0900 V01 N02: NUCLEAR REGULATORY LEGISLATION.102d NUREG/CR.5991 PORFLOW: A MULTIFLUID MULTIPHASE MODEL Congress.

FOR SIMULATING FLOW, HEAT TRANSFER, AND MASS TRANS-NUREG4960 V02 NO2 NUCLEAR REGULA10RY LEGISLATION.102d PORT IN FRACTURED POROUS MEDIA. User's Manual - Version Congress 2 41.

Nuclear Regulatory Research NUREG 1266 V07. NRC SAFETY RESEARCH IN SUPPORT OF REGU.

FRA NUREG/CR-5488 RISK BASED INSPECTION GUIDE FOR THREE MILE

' ATf0N - FY 1992.

ISLAND NUCLEAR ST ATION UNIT 1.

NUREG/CR-5766. AUXlLIARY FEEDWATER SYSTEM RISK-BASED IN-Nuclear Safety Research NUREG/CP-0132. TRANSACTIONS OF THE TWENTY FIRST WATER SPECTION GUIDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR REACTOR SAFETY INFORMATION MEETING.

POWER PLANT, j

1

-m Subject index 89 NUREG/CR-5829. AUXIUARY FEEDWATER SYSTEM AlSK-BASED IN.

NUREG4936 V12 NO2: NRC REGULATORY AGENDA.Ouarterly SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER Report.Apni-June 1993.

PLANT.

NUREG-0936 V12 NO3: NRC REGULATORY AGENDA.Ouarteri)

NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-Report July September 1993 SPECTION GUIDE FOR THE H B. ROBINSON NUCLEAR POWER Pu NT.

Physical Inventory N'UREG/CR-5834: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR-6118: ASSESSMENT OF THE EFFECTIVENESS OF THE SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER PLANT.

LEU REFORM RULE AND ITS IMPLEMENTATION.

NUREG/CR-5835. AUXlLIARY FEEDWATER SYSTEM RISK-BASED IN-Ptpe SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NU-NUREG/CR-4599 V03 N1: SHORT CRACKS IN PIP 6NG AND PIPING CLEAR POWER PLANTS NUREG/CR-5836. AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-WEL.DS. Semiannual Report, April-September 1992.

SPECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER NUREG/CR-6098. LOADING RATE EFFECTS ON STRENGTH AND FRACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF NU CR 5898: AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE POINT BEACH NUCLEAR POWER

PLANT, piping NUREG/GR-5964 SAPHIRE TECHNICAL REFERENCE NUREG-1364: REGULATORY ANALYSIS FOR THE RESOLUTION OF MANUAL:lRRAS/ SARA VERSION 4 0' GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY t

COMBUSTIBLE GASES IN VITAL AREAS.

)

PRUEP NUREG/CR 4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING l

NUREG/CR 5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE WELDS Semiannual Report. October 1991

• March 1992.

LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NUREG/CR-6049: PIPING BENCHMARK PROBLEMS FOR THE GEN-Ask Uncertainty Evaluation Program (PRUEP) Appendices A-C.

ERAL ELECTRIC ADVANCED BOluNG WATER REACTOR.

NUREG/CR-5305 V02 P2. INTEGRATED RISK ASSESSMENT FOR THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenorpenology And Plant Aging Risk Uncertainty Evaluahon Program (PRULP) Appendices D-G NUREG-1377 R04: NRC RESEARCH PROGRAM ON PLANT AGtNG:

USTING AND SUMMARIES OF REPORTS ISSUED THROUGH SEP.

PWR TEMBER 1993.

NUREG/CR-5759. RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS STORAGE, SUPPLY. AND DISTRIBUTION SYSTEMS IN PRESSUR.

Plant Transient 12ED WATER REACTOR PLANTS, NUREG/lA-0085: ASSESSMENT OF FULL POWER TURBINE TRIP NUREG/CH 5822. ANALY SIS OF THERMAL MIXING AND BORON DI-START-UP TEST FOR C TRILLO I WITH RELAP5/ MOD 2.

LUTION IN A PWR NUREG rCR-5937: INTENTIONAL DEPRESSURl2ATION ACCIDENT Plugging Crtterta MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC-TORS.

NUREG-1477 DRFT FC: VOLTAGE-BASED INTERIM PLUGGING CRITE-RlA FOR STEAM GENERATOR TUBES. Draft Repor1 For Comment.

NUREG/CR 6048-PRESSURIZED-WATER REACTOR INTERNALS AGING DEGRADATION STUDY. Phase 1.

Power Reactor NUREG/CR 6054 DAF FC: ESTIMATING PRESSURIZED WATER RE-ACTOR DECOMMISSIONING COSTS A User's Manual For The PWR NUREG-0713 V14' OCCUPATIONAL RADIATION EXPOSURE AT COM-Cost Estimating Computer Program (CECP) Software Draft Report Fo' MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES Comment.

1992 Twenty-Fifth Annual Report.

Practice And Procedure Digest NUREG-0386 DO6 R05: UN!TED STATES NUCLEAR REGULATORY JE /C 5A LYS:S OF THE LASALLE UNIT 2 NUCLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION COMMISSION STAFF PRACTICE AND PROCEDURE PROGRAM. Parameter Eshmation Analysis And Screening Human Reh-DIGEST. Commission, Appeal Board And Licensing Board Decisions. July abAty Analysis 1972 - March 1992.

NUREG-0386 D06 R06. UNITED STATES NUCLEAR REGULATORY Particle Penetration COMMISSION STAFF PRACTICE AND PROCEDURE NUREG/GR-0006. DEPOSITION SOFTWARE TO CALCULATE PARTI.

DIGEST Commission, Appeal Board And Ucensing Board Decisions. July CLE PENETRATION THROUGH AEROSOL TRANSPORT 1972. June 1992.

SYSTEMS Final Report.

NUREG 0386 D06 R07: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE Performance Assessment DIGEST. Commission, Appeal Board And Ucensing Beard NUREG/CR 5927 V01; EVALUATION OF A PERFORMANCE ASSESS-Decessons. July 1972 - September 1992.

MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITIES Evaluation Of Modeling Approaches Prenatal Radiation Dose NUREG/CR 5631 R1 ADO: CONTRIBUTION OF MATERNAL RADIONU-Performance History CUDE BURDENS TO PRENATAL RADIATION DOSES. Relationships NUREG 1214 R11' HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-Between Annual Umsts On intake And Prenatal Doses.

IC ASSESSMENT OF LICENSFE PERFORMANCE NUREG-1214 R12 HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-Pressure Boundary IC ASSESSMENT OF UCENSEE PERFORMANCE.

NUREG/CR-5928' ISLOCA RESEARCH PROGR AM Final Report.

Performance incentive Pressure Transmitter NUREG/CH 5975. INCENTIVE REGULATION OF INVESTOR-OWNED NUREG/CR-5851: LONG TERM PERFORMANCE AND AGING CHAR.

NUCLEAR POWER PLANTS BY PUBUC UTILITY REGULATORS-ACTERISTICS OF NUCLEAR PLANT PRESSURE TRANSMITTERS.

Performance Indicator NUREG/CR-5977. A PERFORMANCE INDICATOR OF THE EFFECTIVE-GfC 5 91 V01 N2: HEAVY SECTION STEEL IRRADIATION NESS HUMAN-MACH (NE INTERFACES FOR NUCLEAR POWER PROGRAM Semiannual Progress Report For Apni-september 1990.

NUREGICR-5926: SANS INVESTIGATION OF LOW ALLOY STEELS IN Performance Standard NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI-NUREG/CR 6062: PERFORMANCE OF PORTABLE RADIATION-TIONS.

SURVEY INSTRUMENTS NUREG/CR-6117: NEUTRON SPECTRA AT DIFFERENT HIGH FLUX ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEllLANCE Petitione For Rulemak6ng LOCATIONS NUREG-0936 VII N04 NRC REGULATORY AGENDA Ouarterly Repo.rt October-Decemter 1992.

Pressurized Thermal Shock NUREG 0936 V12 N01: NRC REGULATORY AGE NDA.Ouarterly NUREG/CR 4219 V09 N2: HEAVY-SECTION S1 EEL TECHNOLOGY Report. January March 1993.

PROGRAM Senuannual Progress Report For April-September 1992.

. ~ _

sh.

90 Subject Index NUREG/CR-5702. PRESSURIZED THERMAL SHOCK PROBABillSTIC Public information FRACTURE MECHANICS SENSITMTY ANALYSIS FOR YANKEE NUREG-1467.

FEDERAL GUIDE FOR A RADIOLOGICAL RESPONSE.Supporung The Nuclear Regulatory Commission Dunng ROWE REACTOR PRESSURE VESSEL NUREG/CR-5968: POTENTIAL CHANGE IN FLAW GEOMETRY OF AN The initial Hours Of A Senous Accident.

INITIALLY SHALLOW FINITE-LENGTH SURFACE FLAW DURING A Public Ut#lty Regulator PRESSURIZED-THERMAL SHOCK TRANSIENT, NUREG/CR-5975: INCENTIVE REGULATION OF INVESTOR-OWNED NUCLEAR POWER PLANTS DY PUBLIC UTIUTY REGULATORS.

Pressurtred Water Reactor NUREG/CR-5759 RISK ANALYSIS OF HIGHLY COMBUSTIDLE GAS STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR.

Pump NUREG 1482 DRFT FC: GUIDELINES FOR INSERVICE TESTING AT IZED WATER REACTOR PLANTS, NUCLEAR POWER PLANTS. Draft Report For Comment.

NUREG/CR-5822: ANALYSIS OF THERMAL MIXING AND BORON DI-LUTlON IN A PWR NUREG/CR-5937. INTENTIONAL DEPRESSURIZATION ACCIDENT RASCAL NUREG/CR-5247 V01 R1 RASCAL VERSION 2.0 USER'S GUIDE.

MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC.

NUREG/CR-5247 V02: RASCAL VE RSION 2 0 WORKBOOK.

TORS NUREG/CR4048 PRESSURIZED-WATER REACTOR INTERNALS AGING DEGRADATION STUDY. Phase 1 NUREG/lA-0095: RELAPS ASSESSMENT USING LSTF TEST DATA SB-NUREG/CR4054 DRF FC: ESTIMATING PRESSURIZED WATER RE-ACTOR DECOMMISSIONING COSTS A User's Manual For The PWR NU E /lA-0099 RELAPS ASSESSMENT USING SEMISCALE SBLOCA Cost Estimahng Computer Program (CECP) Software. Draft Report For TEST S-NH 1.

Comment.

RELAP5 Computer Code Pressurtzer Spray Valve NUREG/CR-6035; FEASIBILITY STUDY FOR IMPROVED STEADY-NUREG/lA 0121: ASSESSMENT OF A PRESSURIZER SPRAY VALVE STATE INITIAllZATION ALGORITHMS FOR THE RELAP5 COMPUT-F AULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT ER CODE' WITH RELAP5/ MOO 2.

NUREG/lA-0124 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES-RELAP5/ MOD 2 SURIZER SPRAY V ALVE INADVERTED FULLY OPENING TRAN-NUREG/lA 0085: ASSESSMENT OF FULL POWER TURBINE TRIP START-UP TEST FOR C. TRIL LO I WITH RELAPS/ MOO 2 SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA-NUREG/lA-0090: ASSESSMENT OF RELAPS/ MOD 2 USING THE TEST BRERA NUCLEAR STATION.

DATA OF REWET-il REFLOODING EXPERIMENT SGl/R.

NUREG/lA4091: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A NATU-Primary Coolant Circuit NUREG/CR4078 ANALYSIS OF CRACK INITIATION AND GROWTH IN RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT BORSSELE-THE HIGH LEVEL VIBRATION TEST AT TADOTSU NUREG/lA 0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE Primary Load AGAINST THE NET LOAD TR!P TEST DATA FROM YONG-NUREG/CR 5358 REVIEW OF ASME CODE CRITERIA FOR CONTROL NUREG/lA-0106. ASSESSMENT OF PWR STEAM GENERATOR MOD-GWANG. UNIT 2.

OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM BRANCH CON-ELLING IN RELAPS/ MOD 2.

NECTIONS AND RECOMMENDATlONS FOR ADDITIONAL DEVELOP-NUREG/lA 0107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A LOAD MENT WORK.

REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

Probabil6stic Risk Assessment NUREGAA 0108. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A TUR-NUREG/CR-4551 V7R1P1: EVALUATION OF SEVERE ACCIDENT BINE TRIP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR HISKS ZION UNIT 1. Main Report-NUREG/CR-4551V7 RIP 2 A: EVALUATION OF SEVERE ACCIDENT POWER PLANT.

NUREGAA-0109. ASSESSMENT OF RELAP5/ MOD 2 AGAINST A 10%

RISKS ZtON UNIT 1 Appenda A.

NUREG/CR-4551V7 RIP 2B. EVALUATION OF SEVERE ACCIDENT LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE RISKS. 210N UNIT 1 Appendices B. C, D. And F.

VANDELLOS !! NUCLEAR POWER PLANT.

NUREG/CR-4832 VOS ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/lA-0110: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A MAIN POWER PLANT: RISK METHODS INTEGRATION AND EV ALUATION FEEDWATER TURBOPUMP TRIP TRANSIENT IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

PROGRAM Parameter Esbmation Anatyws And Screening Human Reh-NUREG/lA-0112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN-RE-arnhty Anafyws FLOOD EXPERIMENTS.

NUREG/CR 4832 V09 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/lA-0118. ANALYSIS OF LOFT TEST LS-1 USING RELAP5/

POWER Pt. ANT: RISK METHODS INTEGRATION AND EVALUATION MOD 2.

NUREG/CR-5471: ENHANCEMENTS TO DATA COLT ECTION AND RE-NUREGMA 0119: ASSESSMENT AND APPLICAT'ON OF BLACKOUT PROGRAM (RMtEP) Internal Fue Ana yms.

TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

PORTING OF SINGLE AND MULTIPLE F ALLURE EVENTS.

MOD 2.

NUREG/CR 5759: RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS NUREGMA-0121: ASSESSMENT OF A PRESSURIZER SPRAY VALVE STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR.

FAULTY OPENING TRANSIENT AT ASCO NUCLEAR POWER PLANT IZED WATER REACTOR PLANTS, NUREGICR 5791: RISK EVALUATION FOR A GENERAL ELECTRIC WITH RELAPS/ MOD 2, NUREG/lA 0123: APPLICATION OF FULL POWER BLACKOUT FOR BWR, EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON C.N ALMARAZ WITH RELAPS/ MOD 2.

SAFETY-RELATED E'OUIPMENT. Evaluation Of Genenc issun 57 NUREG/lA-0124: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES-NUREG/CR 5863 RISK ASSESSMENT OF ISOLATION DEVICES IN SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN-SAFETY SYSTEMS SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA-NUREG/CR-5936 ENHANCEMENTS TO THE ACCIDENT PRECURSOR BRERA NUCLEAR STATION METHODOL.OGY.

NUREGAA-0125: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE NUREG/CR-5976. DEVELOPMENT AND USE OF A TRAIN LEVEL AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG-PROBABILISTIC RISK ASSLSSMENT.

GWANG UNIT 2.

NUREGAA 0128: INTERNATIONAL CODE ASSESSMENT AND APPLl-ProbabHistic Safety Analysis CATIONS PROGRAM:

SUMMARY

OF CODE ASSESSMENT STUDIES NUREG/CR 5801: PROCEDURE FOR ANALYSIS OF COMMON-CAUSE CONCERNING FIELAPS/ MOD 2, RELAPS/ MOD 3. AND TRAC-B.

F AILURES IN PROBABILISTIC SAFETY ANALYSIS RELAP5/ MOD 3 Program Performance Report NUREG/CR-5818. UNCERTAINTY ANALYSIS OF MINIMUM VESSEL NUREG/CR 5758 V03 FITNESS FOR DUTY IN THE NUCLEAR POWER LIOUID INVENTORY DUR!NG A SMALL-BREAK LOCA IN A B&W INDUST RY. Annual Summary Of Program Performarce Reports.CY PLANT-AN APPLICATION OF THE CSAU METHODOLOGY USING 1992.

THE RELAP5/ MOD 3 COMPUTER CODE.

NUREG/CR-6061: DETERMINATION OF THE BIAS IN LOFT FUEL Protected Area PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN NUREG 1485 UNAUTHORIZED FORCED ENTRY INTO THE PROTECT-PHASE OF LARGE BREAK LOCA EXPERIMENTS.

ED AREA AT THREE M:LE ISLAND UNIT 1 ON FEBRUARY 7,1993.

Subject index 91 NUREG/IA-0094: ASSESSMENT OF RELAPS/ MOD 3 AGAINST Radiat>on Transport TWENTY FIVE POST.DRYOUT EXPERIMENTS PERFORMED AT THE NUREG/CR-5247 V01 R1: RASCAL VERSION 2.0 USER'S GUIDE.

ROYAL INSTITUTE OF TECHNOLOGY.

NUREG/CR-5247 V02. RASCAL VERSION 2 0 WORKBOOK.

NUREG/lA-0096: NUMERICS AND IMPLEMENTATION OF THE UK HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE MODEL Radioactive Material INTO RELAPS/ MOD 3 NUREG4383 V01 R16: DIRECTORY OF CERTIFICATES OF COMPLI-NUREG/lA 0100- ASSESSMENT OF CCFL MODEL OF RELAP5/ MOD 3 ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS-Approved Packages NUREG/tA 0103. ASSESSMENT OF BETHSY TEST 9.1.B USING NUREG 0383 V02 R16: DIRECTORY OF CERTIFICATES OF COMPLl-NURE lA RELAPS/ MOD 3 ASSESSMENT USING THE SEMIS ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of CALE 50% FEED LINE BREAK TEST S-FS-11 Comphance.

NUFiEG/lA4105: ASSESSMENT OF RELAP5/ MOD 3 VERSION EMS NUREG-0383 V03 R13: DIRECTORY OF CERTIFICATES OF COMPLI-USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC KORI UNIT 3 PLANT.

Approved Quality Assurance Programs For Radioactwe Matenals Pack-NUREG/lA-0113. PRELIMINARY ASSESSMENT OF PWR STEAM GEN.

ages ERATOR MODELLING IN RELAPS/ MOD 3.

NUREG/CR-2907 V11: RADICACTIVE MATERIALS RELEASED FROM NUREG/lA4128 INTERNATIONAL CODE ASSESSMENT AND APPLi-NUCLEAR POWER PLANTS. Annual Report 1990.

CATIONS PROGRAM.

SUMMARY

OF CODE ASSESSMENT STUDIES NUREG/CR-5962: HEALTH AND SAFETY IMPACTS FROM DISCRETE CONCERNING RELAPS/ MOD 2, RELAPS/ MOD 3. AND TRAC B.

SOURCES OF NATURALLY OCCURRING AND ACCELERATOR PRO-RELAP5/ MOD 3/V5m5 NUREG/lA-0116. ASSESSMENT OF RELAPS/ MOD 3/V5M5 AGAINST Radioactive Particle THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW'lN PWR NUREG/CR 6081: ENHANCED REMOVAL OF RADIOACTIVE PARTI.

W LEG)

CLES BY FLUOROCARBON SURFACTANT SOLUTIONS.

NU EG/CR 6050 RADIATION EXPOSURE MONITORING AND INFOR-NUREG/CR-2850 V11: DOSE COMMITMENTS DUE TO RADIOACTIVE MATION TRANSMITTAL (REMIT) SYSTEM User's Manual ~

RELEASES FROM NUCLEAR POWER PLANT SITES IN 1989, REWET il Reflooding Emperiment NUREG/lA-0090 ASSESSMENT OF RELAPS/ MOD 2 USING THE TEST Radioactive Tracer DATA OF REWET-il REFLOODING EXPERIMENT SGI/R.

NUREG/CR 5990: SUBSURFACE INJECTION OF RADIOACTIVE TRACERS. Field Expenment For Model Validation Testing RHR System NUREG/CB 5995: TECHNICAL SPECIFICATION ACTION STATEMENTS Radioactive Waste REOViniNG SHUTDOWN A Risk Perspectwe With Application To The NUREG/CR4073: LYSIMETER LITERATURE REVIEW.

RHR/SSW Systems Of A BWR, Radiological Response RMIEP NUREG 1467.

FEDERAL GUIDE FOR A RADIOLOGICAL NUREG/CR-4832 V08. ANALYSi$ OF THE LASALLE UNIT 2 NUCLEAR RESPONSE Supporting The Nuctaar Regulatory Commission Dunng POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION The initial Hours Of A Senous Accident.

PROGRAM (RMiEP).Soismic Analysa Radionuclide N EG/CR-4214 R1P2A2. HEALTH EFFECTS MODELS FOR NUCLE-DE B DE S TO R T L R D AT SF Ra h

AR POWER PLANT ACCIDENT CONSEQUENCE Between Annual Limits On intake And Prenatal Doses.

i ANALYSIS Modifcation Of Models Resulting From Addition Of Effects NUREG/CR-5894: RADIONUCLIDE CHARACTERi2ATION OF REAC-Of Exposuie To Alpha-Emitting Radionuclides Part II. Scientific Bases TOR DECOMMISSIONING WASTE AND NEUTRON-ACTIVATED l

NU E / f 14 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part i Radionuclide Migration Introduction, Integration.And Summary.

NUREG/CR-5943-SENSITIVITY ANALYSIS AND BENCHMARKING OF Radiation Control Program THE BLT LOW. LEVEL WASTE SOURCE TERM CODE.

NUREG-1479 RESULTS FROM TWO WORKSHOPS: STATE RADI, NUREG/CR-6041: O!SPOSAL UNIT SOURCE TERM (DUST) DATA P

AT O CONTROL PR AMS DEVELOPING AND AMENDING REG-N gE /CR 6 08. SPHERICAL DIFFUSION OF TRITlUM FROM A POINT OF RELEASE IN A UNIFORM UNSATURATED SOILA Deter.

Radiation Embrlttlement ministic Model For Tntium Migret on in An And Disposal Site.

NUREG/CR-5926-SANS INVESTIGATION OF LOW ALLOY STEELS IN NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI.

Radionuclide Release TIONS..

NUREG/CR-5747; ESTIMATE OF RADIONUCLIDE RELEASE CHARAC-NUREG/CR 6117: NEUTRON SPECTRA AT DIFFERENT HIGH FLUX TERISTICS INTO CONTAINMENT UNDER SEVERE ACCIDENT ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEILLANCE CONDIT!ONS Final Report.

LOCATIONS.

Radionuclide Transport Radiation Exposure NUREG/CR 5917 V01; SENSITIVITY AND UNCERTAINTY ANALYSES NUREG-0713'V12 OCCUPATIONAL RADIATION EXPOSURE AT COM-APPLIED TO ONE DIMENSIONAL RADIONUCLIDE TRANSPORT IN A MERCIAL NUCLEAR POWER REACTORS AND OTHER LAYERED FRACTURED ROCK:MULTFRAC - Analytic Solutions And FACILITIES,1990 Twenty Third Annual Report.

Local Sensit;vities.

NUREG 1480 LOSS OF AN IRIDIUM 192 SOURCE AND THERAPY MIS' NUREG/CR-5917 V02: SENSITIVfTY AND UNCERTAINTY ANALYSES ADMINISTRATION AT IND!ANA RE GIONAL CANCER APPUED TO ONE-OlMENSIONAL RADIONUCLIDE TRANSPORT IN A CE.NTER,1NDIANA. PENNSYLVANIA.ON NOVEMBER 16,1992-LAYERED FRACTURED ROCK. Evaluation Of The Limit State Ap-Radiat. ton Hazard proach.

NUREG/CR-5883 HEALTH RISK ASSESSMENT OF 1RRADIATED Reactivity Transient NUREG/CR.5922: MODULAR HIGH TEMPERATURE GAS COOLFD RE-Radiation Protection ACTOR SHORT TERM THERMAL RESPONSE TO FLOW AND REAC.

NUREG/CR-3469 V07. OCCUPATIONAL DOSE REDUCTION AT NU-TIVITY TRANSIENTS.

CLEAR POWER PLANTS. ANNOTATED BIBLIOGRAPHY OF SELECT.

ED READINGS IN RADIATION PROTECTION AND ALARA.

Reactor Accident NUREG/CP-0126 VOI: PROCEEDINGS OF THE TWENTIETH WATER Radiation Survey REACTOR SAFETY INFORMATION MEETING.

NUREG/CR 6062. PERFORMANCE OF PORTABLE RADIATION NUREG/CP 0126 V02. PROCEEDINGS OF THE TWENTIETH WATER SURVEY INSTRUMENTS REACTOR SAFETY INFORMATION MEETING.

i

't 92' Subject Index NUMEG/CP-Cl?6 V03: PROCEEDINGS OF THE TWENTIETH WATER Reactor Safety REACTOR SAFETY INFORMATION MEETING.

NUREG/CP-0126 V01: PROCEEDINGS OF THE TWENTIETH WATER NUREG/CR-5933: HIGH PRESSURE COOLANT INJECTION (HPCJ)

REACTOR SAFETY INFORMATION MEETING.

SYSTEM RISK-BASED INSPECTION GUIDE FOR DRESDEN NUCLE-NUREG/CP-0126 V02: PROCEEDINGS OF THE TWENTIETH WATER AR POWER STATION UNITS 2 AND 3.

REACTOR SAFETY INFORMATION MEETING.

NUREG/CR 5934. HIGH PRESSURE COOLANT INJECTION (HPCI)

NUREG/CP 0126 V03: PROCEEDINGS OF THE TWENTIETH WATER SYST EM RISK-BASED INSPECTION GUIDE FOR OUAD-CITIES REACTOR SAFETY INFORMATION MEETING.

STATIONzUNITS 1 AND 2 NUREG/CR 4551 V7 RIP 1: EVALUATION OF SEVERE ACCIDENT NUREG/CR-5959-HIGH PRESSURE COOLANT INJECTION (HPC0 RISKS: ZION UNIT 1. Main Report.

SYSTEM RISK-BASED INSPECTION GUIDE FOR ENRlCO FERMI NUREG/CR 4551V7 RIP 2A: EVALUATION OF SEVERE ACCIDENT ATOMIC POWER PLANT. UNIT 2.

RISKS. ZION UNIT 1.Appendm A.

NUREG/CR 5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN NUREG/CR 4551V7R1P2B: EVALUATION OF SEVERE ACCIDENT CONTAINMENT MANAGEMENT.

RISKS: ZION UNIT 1 Appendices B. C, D. And E.

NUREG/CR-6072: HIGH PRESSURE COOLANT INJECTION (HPC0 NUREG/CR-6049: PIPING BENCHMARK PROBLEMS FOR THE GEN-SYSTEM RISK-BASED INSPECTION GUIDE FOR BROWNS FERRY ERAL ELECTRIC ADVANCED BOILING WATER REACTOR.

NUCLEAR POWER STATION.

NUREG/GR-6059 MACCS VERSION 15.11.1: A MAINTENANCE RE-Reactor Safety Research LEASE OF THE CODE.

NUREG/CP-0132: TRANSACTIONS OF THE TWENTY FIRST WATER REACTOR SAFETY INFORMATION MEETING.

. NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION Reactor Safety System AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE NUREG/CR-6083: REVIEWING REAL-TIME PERFORMANCE OF NU-DAT A STATISTICS Summary Report CLEAR REACTOR SAFETY SYSTEMS NUREG/CR-5993 V02; METHOOS FOR DEPENDENCY ESTIMATION AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE Real-Time Performance

~

DATA STATISTICS Detailed Descnption And Apphcations.

NUREG/CR-6083: REVIEWING REAL TIME PERFORMANCE OF NU-Reactor Containment NUREG/CR4961: POSTTEST DESTRUCTIVE EXAM: NATION OF THE Recall Technique STEEL LINER IN A 16-SCALE REAC10R CONTAINMENT MODEL NUREG/CR-5977: A PERFORMANCE INDICATOR OF THE EFFECTIVE-S OF HUMAN-MACHINE INTERFACES FOR NUCLEAR POWER Ru CWSW NUREG/CR-5783 AGING ASSESSMENT OF THE COMBUSTION ENGI-NEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES-Reference Experiment NUREG/CR-5844 AGING ASSESSMENT OF BISTABLES AND NUREG/CR-5997; CSNI PROJECT FOR FRACTURE ANALYSES OF SWITCHES IN NUCLE AR POWER PLANTS.

LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS Reactor Coolant System (PROJECT FALSIRE).

NUREG/CR-5360: XSOR CODES USERS MANUAL Rh RWe Reactor Cooling System NUREG/CR-6110: ASSESSMENT OF THE EFFECTIVENESS OF THE NUREG/CR-5933: HIGH PRESSURE COOLANT INJECTION (HPCI)

LEU REFORM RULE AND ITS IMPLEMENTATION.

SYSTEM RISK-BASED INSPECTION GUIDE FOR DRdSDEN NUCLE-RegWaW Agenda AR POWER STATION UNITS 2 AND 3' COOLANT INJECTION (HPCI)

NUREG/CR-5934. HIGH PRESSURE NUREG 0936 VII N04; NRC REGULATORY AGENDA.Ouarterly SYSTE M RISK-BASED INSPECTION GUIDE FOR QUAD-CITIES Report. October-December 1992.

NUREG-0936 V12 N01: NRC REGULATORY AGENDA.Ouarterly STATION. UNITS 1 AND 2 NUREG/CR 5959 HIGH PRESSURE COOLANT INJECTION (HPCI)

Report. January-March 1993.

NUREG-0936 V12 NO2: NRC REGULATORY AGENDA.Ouarterly SYSTEM RISK-BASED INSPECTION GUIDE FOR ENRICO FERMI ATOMIC POWER PLANT. UNIT 2 Report,ApribJune 1993.

NUREG/CR 5983: SAFETY ASPECTS OF FORCED FLOW COOLDOWN NUREG-0936 V12 NO3; NRC REGULATORY AGENDA.Ouarterly TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED Report. July-September 1993.

REACTORS NUREG/CR-5984 CODE AND MODEL EXTENSIONS OF THE THATCH Regulatory And Technical Report CODE FOR MODULAR HIGH TEMPERATURE GAS COOLED REAC-NUREG-0304 V17 N04: REGULATORY AND TECHNICAL REPORTS l

TORS' (ABSTRACT INDEX JOURNAL). Annual Compdation For 1992.

NUREG 0304 V18 N01: REGULATORY AND TECHNICAL REPORTS l

l Reactor Maintenance l ABSTRACT INDEX JOURNAL). Compilation For First Quarter i

NUREGICR 5783 AGING ASSESSMENT OF THE COMBUSTION ENGl-1993. January-March.

NEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES NUREG-0304 V18 NO2-REGULATORY AND TECHNICAL REPORTS

(

(ABSTRACT INDEX JOURNAL). Compilation For Second Ouarter I

Reactor Pressure Vessel 1993.Aprildune NUREG/CP 0131: PROCEEDINGS OF THE JOINT IAE A/CSNI SPECIAL.

NUREG-0304 V18 NO3: REGULATORY AND TECHNICAL REPORTS ISTS' MEETING ON FRACTURE MECHANICS VERIFICATION BY (ABSTRACT INDEX JOURNAL). Compdation For Third Quarter LARGE-SCALE TESTING Held Al Pollard Auditonum. Oak 1993. July-September.

Ridge. Tennessee NURLG/CR 5410: STATISTICALLY BASED REEVALUATION OF PtSC-Il ReguW/ Document ROUND ROBIN TEST DATA.

NUREG/CR-3973: CODES AND STANDARDS AND OTHER GUIDANCE NUREG/CR 5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC CITED IN REGULATORY DOCUMENTS.

FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE ROWE REACTOR PRESSURE VESSEL.

Regulatory Review And improvement NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF NUREG/CP-0129: PROCEEDINGS OF THE WORKSHOP ON PROGRAM LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS FOR ELIMINATION OF REQUIREMENTS MARGINAL TO SAFETY, (PROJECT FALSlRE)

NUREG/CR 6023. GENERIC ANALYSIS FOR EVALUATION OF LOW Reinforced Concrete CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS NUREG/CR-5755: STIFFNESS OF LOW-ASPECT RATIO. REINFORCED AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERb CONCRETE SHEAR WALLS.

ALS NUREG/CR-5776: DAMPING IN LOW-ASPECT-RATIO. REINFORCED NUREG/CR 6071: IMPACT OF ENDF/B-VI CROSS-SECT!ON DATA ON CONCRETE SHEAR WALLS.

H B ROBINSON CYCLE 9 DOSIMETRY CALCULATIONS Rehability Reactor Protection System NUREG/CR-5944 A CHARACTERIZATION OF CHECK VALVE DEGRA-NUREG'CR 6101: SOFTW ARE REllABILITY AND SAFETY IN NUCLE-DATION AND FAILURS EXPERIENCE IN THE NUCLEAR POWER IN-AR REACTOR PROTECTION SYSTEMS.

DUSTRY.

4 i

Subject index 93 Replacement Capacity Cost NUREG4386 006 R07; UNITED STATES NUCLEAR REGULATORY

- NUREG/CR-6080 REPLACEMENT ENERGY, CAPACITY, AND RELi-COMMISSION STAFF PRACTICE AND PROCEDURE ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT.

DIGEST. Commission, Appeal Board And Licensing Board DOWNS Decisions. July 1972 September 1992.

Repiecement Energy Cost SANS NUREG/CR 6080 REPLACEMENT ENERGY. CAPACITY, AND REU-NUREG/CR-5926: SANS INVESTIGATION OF LOW ALLOY STEELS IN ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT.

NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI-DOWNS.

TIONS.

Report To Congress SAPHIRE NUREG-0090 Vt5 N04. REPORT TO CONGRESS ON ABNORMAL NUREG/CR-5964:

P.APHIRE TECHNICAL REFERENCE OCCURRENCES October.Docember 1992.

MANUAL 1RRAS/ SARA VERSION 4 0' NUREG-0000 V16 NOI: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES January March 1993.

Safeguards Summary Event Ust i

NUREG-0090 V16 NO2-REPORT TO CONGRESS ON ABNORMAL NUREG 0525 V02 R01: SAFEGUARDS

SUMMARY

EVENT LIST OCCURRENCES Apni June 1991 (SSEL). January 1,1990 Through December 31,1992.

Residual Heat Removal NUREG-1275 V09 OPERATING EXPERIENCE FEEDBACK REPORT -

EG 1 OVEMENTS TO TECHNICAL SPECIFICATIONS PRESSURE LOCKING AND THERMAL BINDING OF GATE SURVEILLANCE REQUIREMENTS.

VALVES Commercial Power Reactors.

]

Safety Evaluation Report p[_...

R A

A ON REN REMD M THE UREG/CR 5229 V05. FIELD LYSiMETER INVESTIGATIONS: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR o et M h@as Me EW FISCAL YEAR 1992 Annual Report.

i Response Time Testing NU 9

7: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-5901 VALIDATION OF SMART SENSOR TECHNOLOGIES OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION.

U T FOR NSTRUMENT CAllBRATION REDUCTION IN NUCLEAR VN'U NUR G E

N T

THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND Ring Dosimeter

2. Docket Nos 50 390 And 50-391.(Tenneses Valley Authorcy)

NUREGICR-5989: PERFORMANCE TESTING OF EXTREMITY DOSI-NUREG-0847 S12: SAFETY EVALUATION REPORT RELATED TO THE METERS-PILOT TESy OPERATION OF WATTS BAR NUCLEAR PLANT UNITS 1 ApD

2. Docket Nos 50-390 And 50-391.(Tennesee Valley Authority)

Risk NUREGd 449: SHUTDOWN AND LOW-POWER OPERATION AT NU.

Safety Information CLEAR POWER PLANTS IN THE UNITED STATES Final Report.

NUREG/CP4126 V01: PROCEEDINGS OF THE TWENTIETH WATER REACTOR SAFETY INFORMATION MEETING Risk Analysis NUREG/CP 0126 V02: PROCEEDINGS OF THE TWENTIETH WATER NUREG/CR 6056 A FRAMEWORK FOR THE ASSESSMENT OF REACTOR SAFETY INFORMATION MEETING.

SEVERE ACCIDENT MANAGEMENT STRATEGIES.

Safety Researcli Risk Management NUREG-1266 V07 NRC SAFETY RESEARCH IN SUPPORT OF REGU-NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF LATION, FY 1992.

SEVERE ACCIDENT MANAGEMENT STRATEGIES Safety System Risk Methode Jntegration NUREG/CR-5663: RISK ASSESSMENT OF ISOLATION DEVICES IN NUREG/CR-4832 V09 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR SAFETY SYSTEMS.

POWER PLANT: RtSK METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP). internal Fire Analysis.

Safety information NUREG/CP 0126 V03: PROCEEDINGS OF THE TWENTIETH WATER Risk-Based inspection REACTOR SAFETY INFORMATION MEETING.

NUREG/GR-0005 V02 PI: RISK-BASED INSPECTION DEVELOPMENT OF GUIDELINESlight Water Reactor (LWR) Nuclear Power Plant Safety-Related Equipment Components NUREG-1472: REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 57. Effects Of Fire Protection System Actuation On Risk-Based Regulation Safety-Related Equipment.

i NUREG/CP-0129 PROCEEDINGS OF THE WORKSHOP ON PROGRAM FOR EUMINATION OF REQUIREMENTS MARGNAL TO SAFETY.

Satellite Telemetry NUREG/CR-8085 UNITED STATES SEISMOGRAPHIC NETWORK.

Rod Bundle NUREG/lA-0100 ASSESSMENT OF CCFL MODEL OF RELAP5/ MOD 3 Scaling AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS, NUREG/GR 0009. STEPWISE INTEGRAL SCALING METHOD AND ITS APPLICATION TO SEVERE ACCIDENT PHENOMENA.

Rules NURE G 0936 V11 N04: NRC REGULATORY AGENDA Ouanerty security Event Raport,0ctober-December 1992 NUREG-1485-UNAUTHORIZED FORCED ENTRY INTO THE PROTECT-NUREG-0936 V12 No t - NRC REGULATORY AGENDA Ouarterly ED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7,1993.

Report. January March 1993 NURtG-0936 V12 NO2. NRC REGULATORY AGENDA.Quartetty Seismic Roport. April June 1991 NUREG/CR-5755: STIFFNESS OF LOW-ASPECT RATIO REINFORCED NUREG 0936 V12 NO3 NRC REGULATORY AGENDA.Ouarterty CONCRETE SHEAR WALLS.

Report. July-September 1993 NUREG/CR-5956: CONSIDERATION OF UNCERTAINTIES IN SOIL.

STRUCTURE INTERACTION COMPUTATIONS.

Rv6es Of Practice NUREG/CR-6011: REVIEW OF STRUCTURE DAMPING VALUES FOR NUREG-0386 D06 ROS. UNITED STATES NUCLEAR REGULATORY ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PLANTS.

COMMtSSION STAFF PRACTICE AND PROCEDURE DIGEST Commission.Appeat Doard And Licensing Board Decssions. July Selsmic Analysis 1972 - March 1992.

NUREG/CR-4B32 VO8, ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUHEG4386 006 R06, UNITED STATES NUCLEAR REGULATORY POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION COMMISSPON STAFF PRACTICE AND PROCEDURE PROGRAM (RMIEP) Seismic Analysis.

DICEST Commession. Appeal Board And Leconsing Board Dectstana. July NUREG/CR-6013: METHOOS USED FOR THE TREATMENT OF NON-1972 - June 1992.

PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

l i

y

,,,---er.-+

g----

w

-a-

,r-

94 Subject Index Seismic Design Shipping Cask NUREG/CR 6011: REVIEW OF STRUCTURE DAMPING VALUES FOR NUREG/CR 6007: STRESS ANALYSIS OF CLOSURE BOLTS FOR ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PLANTS.

SHIPPING CASKS.

Setemic Effect Shutdown NUREG/CR-6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN NUREG-1449. SHUTDOWN AND LOW-POWER OPERATION AT NU.

THE HIGH LEVEL VlBRATION TEST AT TADOTSU.

CLEAR POWER PLANTS IN THE UNITED STATES. Final Report.

Seismic Event i

NUR G/ R 6027: PRELIMINARY EVALUATION OF SNUBBER SINGLE

[gjGR ce 0; HYBRID DIGITAL SIGNAL PROCESSING AND NEURAL NETWORKS FOR AUTOMATED DIAGNOSTICC USING NDE Selsmic Hazard MET HODS.

NUREG 1488 DRFT FC: REVISED LIVERMORE SEISMIC HAZARD ES.

s TIMATES FOR 69 NUCLEAR POWER PLANT SITES EAST OF THE Smart Sensor ROCKY MOUNTAINS Draft Report For Comment.

NUREG/CR 5903 VALIDATION OF SMART SENSOR TECHNOLOGIES FOR INSTRUMENT CALIBRATION REDUCTION IN NUCLEAR Seismic Network POWER PLANT S.

NUREG/CR 5778 V03. NEW YORK /NEW JERSEY REGIONAL SEISMIC NETWORK Final Re For A{il 1985 - Segember 1992 Snubber NUREG/CR-6027: PRELIMINARY EVALUATION OF SNUBBER SINGLE

/

VIA A R 10 AL SE SMI E

RK Final FAILURES.

Report (1986 -1992)

NUREG/CR-6079. SEISMOLOGICAL' INVESTIGATION OF EARTH-N Al G/CR

13. CLASS 1E DIGITAL SYSTEMS STUDIES.

pte r 86 -

m 992 Setemographic Network Software Reliability NUREG/CR 6085: URTED STATES SEISMOGRAPHIC NETWORK.

NUREG/CR-6101: SOFTWARE RELIABILITY AND SAFETY IN NUCLE-AR REACTOR PROTECTION SYSTEMS Semiscale S-FS-11 NUREG/lA-0104 RELAP5/MO93 ASSESGMENT USING THE SEMIS-Software Verification CALE 50% FEED LINE DREAK TEST S FS-11.

NUREG/CR-6018. SURVEY AND ASSESSMENT OF CONVENTIONAL Semiscale SBLOCA NUREG4A-0099 RELAPS ASSESSMENT USING SEMISCALE SBLOCA Soit TEST S NH-1, NUREG/CR-59L8; SOIL CHARACTERIZATION METHODS FOR UN-Service Water System SATURATED LOW LEVEL WASTE SITES.

NUREG/CR-5995: TECHNICAL SPECIFICATION ACTION STATEMENTS Soll-Structure interaction hR S UTDOWN A R sk Perspective With Apphcation To The NUREG/CR-5956: CONSIDERATION OF UNCERTAINTIES IN SOIL-STRUCTURE INTERACTION COMPUTATIONS.

Severe Accident NUREG/CR 4273 CRACK PROPAGATION IN HIGH STRAIN REGIONS Solidus Temperature OF SEOUOYAH CONTA!NMENT.

NUREG/CR-6032: SOLIDUS AND LIOUIDUS TEMPERATURES OF NUREG/CR-4551 V7RIPI: EVALUATION OF SEVERE ACCIDENT CORE-CONCRETE M!XTURES.

RISKS ZION UNIT 1 Main Report NUREG/CR-4551V7R1P2A: EVALUATION OF SEVERE ACCIDENT Solute Transport RISKS' 710N UNIT 1. Appendix A.

NUREG/CR-5998: SIMULATION OF UNSATURATED FLOW AND NON-NUREG/CR 4551V7RtP20: EVALUATION OF SEVERE ACCIDENT REACTIVE SOLUTE TRANSPORT IN A HETEROGENEC JS SOIL AT NUREG/CR5747: ESTI ppendices B. C, D, And E.

RtSKS: ZION UNIT 1.A THE FIELD SCALE.

MATE OF RADIONUCLIDE RELEASE CHARAC-TERISilCS INTO CONTAINMENT UNDER SEVERE ACCIDENT Source Term CONDITIONS Final Report NUREG/CR-5911: SOURCE TERM EVALUATION FOR RADIOACTIVE NUREG/CR-5942; SEVERE ACCIDENT SOURCE TERM CHARACTER-LOW-LEVEL WASTE DISPOSAL PERFORMANCE ASSESSMENT-ISTICS FOR SELECTED PEACH BOTTOM SEOUENCES PREDICTED NUREG/CR5943: SENSITIVITY ANALYSIS AND BENCHMARKING OF E C 57 O

E E 8 WAH W WE V E AC ENT AN MENT STRATEG ES NUREG/GR-0009 STEPWISE INTEGRAL SCALING METHOD AND ITS MARK l BOILING WATER REACTOR ORYWELL APPLICATION TO SEVERE ACCIDENT PHENOMENA.

Spectroscopic Analysta Severe Reactor Accident NUREG/CR 6047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF NUREG/CR-5843: CORCON-MOD 3 AN INTEGRATED COMPUTER VANADOUS AND VAN ADIC IONS.

MODEL FOR ANALYSIS OF MOLTEN CORE-CONCRETE INTERACTIONS User's Manual Spent Fuel NUREG/CR-5966 A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY NUREG 0725 R09-PUBLIC INFORMATION CIRCULAR FOR SHIP-CONTAINMENT SPRAYS.

MENTS OF IRRADIATED REACTOR FUEL NUREG/CR-5894. RADIONUCLIDE CHARACTERIZATION OF REAC-Shallow Crack TOR DECOMMISSIONING WASTE AND NEUTRON-ACTIVATED NUREG/CR-5969. J AND CTOD ESTIMATION EQUATIONS FOR SHAL-METALS.

LOW CRACKS W TNGLE EDGE NOTCH DEND SPECIMEN 3 Stainless Steel Shear wha NUREG/CR-4469 V15: NONDESTRUCTIVE EXAMINATION (NDE) REll-NUR(0'CR-5755: STIFFNESS OF LOW-ASPECT RATIO, REINFORCED ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER NL

/C 7 DA IN N LOW-ASPECT-RATIO. REINFORCED REACTORS. Semiannual Report, October 1991 March 1992.

CONCRETE SHEAR WALLS NUREG/CR-4744 V07 N1 LONG TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual Shear Walt Building Report, October 1991 March 1992.

NUREG/CR 6012: STIFFNES6 AND DAMPING PROPERTIES OF A NUREGICR-4744 V07 N2: LONG TERM EMBRITTLEMENT OF CAST LOW ASPECT RATIO SHEAF WALL BUILDING BASED ON RECORD.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual ED EADtHOUAKE RESP MES ReportApni-September 1992 Shipment Standard Review Plan NUREG 0725 R09 PUBLIC INFORMATION CIRCULAR FOR SHIP-NUREG/CR-5973: CODES AND STANDARDS AND OTHER GUIDANCE MENTS OF IRRADIATED REACTOR FUEL CITED IN REGULATORY DOCUMENTS.

Subject Index 95 State Regulation Subsurf ace injection NUREG 1479 RESULTS FROM TWO WORKSHOPS STATE RADI-NUREG/CR-5996: SUBSURFACE INJECTION OF R ADIOAC16VE ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG-TRACERS Field Experiment For Model Validahon Testing ULATIONS AND FUNDING Surface-Crack Station Dieckout NUREG/CR 5971: CONTINUUM AND MICROMECHANICS TREATMENT NUREG/CR 5949-ASSESSMENT OF THE POTENTIAL FOR HIGH OF CONSTRAINT IN FRACTURE-PRESSURE MELT EJECilON RESULTING FROM A SURRY STATION BLACKOUT TRANSIENT.

Surfactant Waste Reduction NUREG/CR4081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-Steady-State Algorithms CLES BY FLUOROCARBON SURFACTANT SOLUTIONS NUREG/CR 6035: FEASIBlUTY STUDY FOR IMPROVED STEADY.

STATE INITIALIZATION ALGORITHMS FOR THE RELAP5 COMPUT-Surveillance Requ6rement ER CODE.

NUREG-1368-IMPROVEMENTS TO TECHNICAL SPECIFICATIONS SURVEILLANCE REOUIREMENTS.

Steam Generator NUREG-1477 DRFT FC: VOLTAGE BASED INTERIM PLUGGING CRITE-Switch AtA FOR STEAM GENERATOR TUBES Draft Report For Comment.

NUREG/CR-5844. AGING ASSESSMENT OF BISTABLES AND NUREG/lA 0106 ASSESSMENT OF PWR STEAM GENERATOR MOD-SWITCHES IN NUCLEAR POWER PLANTS ELLING IN RELAP5/ MOD 2 NUREG/lA-0113 PRELIMINARY ASSESSMENT OF PWR STEAM GEN-System 80 &

ERATOR MODELLING IN RELAP5/ MOD 3.

NUREG/CR-5957; SYSTEM 80 + (TM) CONTAINMENT - STRUCTURAL Steel Containment NUREG/CR-5957; SYSTEM 80 +(TM) CONTAINMENT - STRUCTURAL System Analysis DESIGN REVIEW NUREG/CR 6065-SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR 8'"I U"

Syatem Design NUREG/CR.5961: POST 1EST DESTRUCTIVE EXAMINATION OF THE NUREG/CR-6082: DATA COMMUNICATIONS.

STEEL LINER IN A 1-6-SCALE REAClOR CONTAINMENT MODEL.

System Failure

. tress Analysis NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION s

NUREG/CR6007: STRESS ANALYSIS OF CLOSURE BOLTS FOR AND SYSTEM UNAVAILABluTY EVALUATION BASED ON FAILURE SHIPPING CASKS DATA STATISTICS. Summary Report.

NUREG/CR 5993 V02: METHODS FOR DEPENDENCY ESTIMATION Stress Corrosion AND SYSTEM UNAVAILADILITY EVALUATION BASED ON FAILURE NUREG/CR 4735 V08. EVALUATION AND COMPILATION OF DOE DATA STATISES.DetaM Descnpton AM Applicatons WASTE PACKAGE TEST DATA. Brannual Report, August 1989 Janu.

ary 1990 System Train NUREG/CR 5976: DEVELOPMENT AND USE OF A TRAIN LEVEL Strees Corrooton Cracking PROBABILISTIC RISK ASSESSMENT NUREG-1477 DRFT FC: VOLTAGE-BASED INTERIM PLUGGING CRITE-RlA FOR STE AM GENERATOR TUBES Draft Report For Comment NUREGICR 5754 BOILING-WATER REACTOR INTERNALS AGING Systematic Assessment Of Ucensee Performance NUREG-1214 R11: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT.

DEGRADATION STUDY. Phase 1.

NUREG/GR6048. PRESSURIZED-WATER REACTOR INTERNALS IC ASSESSMENT OF LICENSEE PERFORMANCE.

NUREG-1714 R12: HISTORICAL DATA

SUMMARY

OF THE SfSTEMAT-AGING DEGRADATION STUDY. Phase 1.

IC ASSESSMENT OF LICENSEE PERFORMANCE.

j Stress Triantatity THATCH Code NUREG/CR-5958.

TWO-PARAMETER FRACTURE MECHANICS NUREG/CR-5%4. CODE AND MODEL EXTENSIONS OF THE THATCH THEORY AND APPLICATIONS ~

CODE FOR MODULAR HIGH TEMPERATURE GAS COOLED REAC-Structural Alloy TORS.

NUREG/CR 5981: THE EFFECT OF ELE CTRIC DISCHARGE MA.

CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL TLD NUREG-0837 V12 N04 NRC TLD DIRECT RADIATION MONITORING STRUCTURAL ALLOYS.

NETWORK. Progress Report Octobor-December 1992, Structural Assessment NUREG-0837 V13 N01: NRC TLD DIRECT RADIATION MONITORING NUREG/CR 5970: APPROXIMATE TECHNIOUES FOR PREDICTING NETWORK Progress Report. January-March 1993 NUREG-0837 V13 NO2: NRC TLD DIRECT RADIATION MONITORING SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JCI NETWORK. Progress Report A it-June 1993.

Structural DeWon NUREG-0837 V13 NO3 NRC T D DIRECT RADIATION MONITORING NUREG/CR 5057: SYSTEM 80 + (TM) CONTAINMENT STRUCTURAL NETWORK. Progress Report July-September 1903.

DESIGN REVIEW Structural Rellabdity NUREG/lA 0128 INTERNATIONAL CODE ASSESSMENT AND APPLI-NUREG/CR 6015 ST RUCTURAL AGING PROGRAM TECHNICAL CATIONS PROGRAM

SUMMARY

OF CODE ASSESSMENT STUDIES PROGRESS FOR PERIOD JANUARY. DECEMBER 1992.

CONCERNING RELAP5/ MOD 2, RELAP5/ MOD 3, AND TRAC-B.

Structural System TR AC-BF1 NUREG/CR 6013. METHODS USED FOR THE TREATMENT OF NOR NUREG/CR 5882: TRAC-B THERMAL-HYDRAUllC ANALYSIS OF THE PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

EILACK FOX BOILING WATER REACTOR.

NUREG/l A-0120: ASSESSMENT OF THE TURBINE TRIP TRANSIENT Structure Damping IN COFRENTES NPP WITH TRAC-BFt.

NUREG/CR 6011: REVIEW OF STRUCTURE DAMPlNG VALUES FOR NUREG/lA 0122. ASSESSMENT OF MSIV FULL C.

RE FOR ELASTIC SEtSMIC ANALYSIS OF NUCLEAR POWER PLANTS SANTA MARIA DE GARONA NUCLEAR POWER 6

• iT USING NUREG/CR 6012: SilFFNESS AND DAMPING PROPERTIES OF A TR AC-BF1 (G1J1).

LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD-ED EARTHOUAKE RESPONSES Technical Specifications NUREG/CR-6013. METHODS USED FOR THE TREATMENT OF NON-NUREG 1366: IMPROVEMENTS TO TECHNICAL SPECIFICATIONS PROPORilONALLY DAMPED STRUCTURAL SYSTEMS.

SURVEILLANCE REOUIREMENTS.

Substance Abuse Therapy Misadministration NUREG/CR 5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG-1480: LOSS OF AN IRIDIUM-192 SOURCE AND THERAPY MIS-INDUSTRY Annual Summa 7 Of Program Pedormance Reports.CY ADMINISTRATION AT INDIANA REGIONAL CANCER 1992 CENTER,1NDIANA. PENNSYLVANIA.ON NOVEMBER 16,1992.

= - -

= -. -.

96 Subject index Thermal Mixing NUREG-0383 V02 R16: DIRECTORY OF CERTIFICATES OF COMPL1-NUREG/CR-5822: ANALYSIS OF THERMAL MIX 1t;G AND BORON Dl-ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certdicates Of LUTION IN A PWR.

Compliance.

NUREG-0383 V03 R13: DIRECTORY OF CERTIFICATES OF COMPLI-EG/

MODULAR HIGH TEMPERATURE GAS-COOLED RE-

• E

• ^*

ACTOR SHORT TERM THERMAL RESPONSE TO FLOW AND REAC.

TlVITY TRANSIENTS.

ThermahHydraulle TMum NUREG/CR-5882: TRAC 8 THERMAL-HYDRAULIC ANALYSIS OF THE NUREG/CR-5980: THREE DIMENSIONAL REDISTRIBUTION OF TRITI-BLACK FOX BOLLING WATER REACTOR.

UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED NUREG/lA 0126: 2D/3D PROGRAM WORK

SUMMARY

REPORT.

SOILA Deterministic Model For Tritrum Migraton in An And Disposal NUREG/lA-0127: REACTOR SAFETY ISSUES RESOLVED BY THE 2D/

Site.

3D PROGRAM.

NUREG/CR-6106: SPHERICAL DIFFUSION OF TRITIUM FROM A POINT OF RELEASE IN A UNIFORM UNSATURATED SOILA Deter-mnis u

a in An And % sal h UE 60 1 A LITERATURE REVIEW OF COUPLED THERMAL-HYDROLOGCMECHAN ICAL -CHEMICAL PROCESSES PERTINENT Tritium Migration TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA MOUNTAIN' NUREG/CR-5900: THREE DIMENSIONAL REDISTRIBUTION OF TRITI-UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED Thermoelastic Property SOILA Determnstic Model For Tritsum Migration in An And Disposal NUREG/CR-5968: POTENTIAL CHANGE IN FLAW GEOMETRY OF AN Site INITIALLY SHALLOW F) NITE-LENGTH SURFACE FLAW DURING A PRESSURIZED. THERMAL-SHOCK TRANSIENT, Tube Degradation NUREG 1477 DAFT FC: VOLTAGE-BASED INTERIM PLUGGING CRITE-Thermohydraulic RIA FOR STEAM GENERATOR TUBES. Draft Report For Comment NUREG/lA4112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN-RE-FLOOD EXPERIMENTS Turbine Trip NUREG/lA4085: ASSESSMENT OF FULL POWER TURBINE TRIP Thermohydrologic START-UP TEST FOR C. TRILLO I WITH RELAPS/ MOD 2 NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA.

NUREG/lA-0108: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A TUR.

TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY BINE TRIP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR t

SATURATED, FRACTURED POROUS MEDIUM.

PO"!ER PLANT.

Thermoluminescent Doelmeter NUREG 0837 V12 N04: NRC TLD DIRECT RArHArlON MONITORING NETWORK Progress Report. Octoie-December 1992.

E A 120 A ESSMENT OF TNE TURBINE TRIP TRANSIENT NUREG4837 V13 N01: NRC ILD DIRECT RADIATION MONITORING IN CGFRENTES NPP WITH TRAC-BF1.

NETWORK Progress Report, January-March 1993.

NUREG-0837 V13 NO2: NRC TLD OlRECT RADIATION MONITORING Turt>opump Trip NETWORK Progress Report. April-June 1993.

NUREG/lA-0110: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A MAIN NUREG 0837 V13 NOT NRC TLD DIRECT RADIATION MONITORING FEEDWATER TURBOPUMP TRIP TRANSIENT IN THE VANDELLOS ll NETWORK. Progress Report July September 1993.

NUCLEAR POWER PLANT.

Title List Turbulent Jet ignition NUREG4540 V141411: TITLE LIST OF DOCUMENTS MADE PUBUCLY NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-V1N T E' L OF DOCUMENTS MADE PUBLICLY NURE 4

AVAILABLE. December 131,1992.

. T NA LARGE A NUREG.0540 V15 N01. TITLE LIST OF DOCUMENTS MADE PUBLICLY UK Numerics And implementation NU F a 2 i LIST OF DOCUMENTS MADE PUBUCLY NUREG/lA 0096: NUMERICS AND IMPLEMENTATION OF THE UK AVAILABLE Fetwuary 1 28,1993, HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE MODEL NUREG 0540 V15 NO3: TITLE UST OF DOCUMENTS MADE PUBLICLY INTO RELAP5/ MOD 3.

AVAILABLE. March 1 31,1993.

NUREG-0540 V15 N04: TITLE LIST OF DOCUMENTS MADE PUBLICLY UPTF Test 11 AVAILABLE. April 1-30,1993.

NUREG/lA-0116. ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST NUREG-0540 V15 N05. lITLE UST OF DOCUMENTS MADE PUBUCLY THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWR AVAILABLE May 1 31,1993 HOT LEG).

NUREG-0540 V15 N06. llTLE UST OF DOCUMENTS MADE PUBUCLY AVAILABLE. June 1 30,1993.

Unsaturated Flow NUREG-0540 V15 N07: TITLE LIST OF DOCUMENTS MADE PUBLICLY NUREG/CR-5998: SIMULATION OF UNSATURATED FLOW AND NON.

N TirLE UST OF DOCUMENTS MADE PUBUCLY REACTIVE SOLUTE TRANSPORT IN A HETEROGENEOUS SOfL AT NU E 540 AVAILABLE.A ust 131' 1993 THE FIELD SCALE.

NUREG-0540 V NO9 TITLE UST OF DOCUMENTS MADE PUDUCLY NUREG/CR4028: BIGFLOW: A NUMERICAL CODE FOR SIMULATING AVAILABL E. September 1 30, 1993.

FLOW IN VARIABLY SATURATED. HETEROGENEOUS GEOLOGIC NUREG-0540 V15 N10: TITLE UST OF DOCUMENTS MADE PUDLICLY MEDIA. Theory And User's Manual - Version 1.1.

AVAILADLE. October 1 31,1993 NUREG/CR-6114 V01: APPUCATION OF AN INFILTRATION EVALUA-TION METHODOLOGY TO A HYPOTHETICAL LOW-LEVEL WASTE Training Program DISPOSAL FACILITY.

NUREG 1220 RO1: TRAINING REVIEW CRITERIA AND PROCEDURES.

Unsaturated Soll RG2 RO1. TRAINING REVIEW CRITERIA AND PROCEDURES Uh h A [N OF R E NT U F M ONSAT R T Trans6ent Wortcer SOILA Deterministic Model For Tntsum Migration in An Arid Disposal NUREG-0713 Vt4. OCCUPATIONAL RADIATION EXPOSURE AT COM-S'te.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES NUREG/CR-6108; SPHERICAL DIFFUSION OF TRITIUM FROM A 1992. Twenty-Fifth Annual Report.

POINT OF RELEASE IN A UNIFORM UNSATURATED SOILA Deter-ministic Model For Tntium Migration in An And Disposal Site.

Transportation NUREG-0383 V01 R18: DIRECTORY OF CERTIFICATES OF COMPU.

Untaturated Waste Site ANCE FOR RADIOACTIVE MATERtALS PACKAGES Report Of NRC NUREG/CR-5988: Soll CHARACTERIZATION METHODS FOR UN-Approved Packages SATURATED LOW-LEVEL WASTE SITES.

Subject index 97 Unsatursted Zone Waste Burial NUREG/CH-5996-SUBSURFACE INJECT 60N OF RADIOACTIVE NUREG-1307 R03. REPORT ON WASTE BURIAL CHARGES Escalation TRACERS Field Expenment For Model Validation Testing.

Of Decommesmoning Weste Disposat Costs At Low-Level Waste Bunal Facilities NUREG 1482 DRFT FC. GUIDELINES FOR INSERVICE TESTING AT Waste Disposal NUCLEAR POWER PLANTS Draft Report f or Comment.

NUREG/CR-2907 VII. RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Arinual Report 1990.

NUREG4040 V16 N04. LICENSEE CONTRACTOR AND VENDOR IN-Waste Treatment SPECTION STATUS REPORT. Quarterly Report,0ctober-December NUREG/CR-5938. NATIONAL PROFILE ON COMMERCIALLY GENER-1992 (White Book)

ATED LOW LEVEL RADIOACTIVE MIXED WASTE.

NUREG4040 V17 N01; LICENSEE CONTRACTOR AND VENDOR IN-SPECTION STATUS REPORT. Quar erly Report, January-March Water Pool 1993 (White Book)

NUREG/CR-5901: A SIMPLIFIED MODEL OF AEROSOL SCRUBBING NUREGOO40 V17 NO2: LICENSEE CONTRACTOR AND VENDOR IN-BY A WATER POOL OVERLYING CORE DEBRIS INTERACTING SPECTION STATUS REPORT Ouarter'y Report,Apnt-June 1993 (White WITH CONCRETE. Final Report NUREG/CR-5907: CORE CONCRETE INTERACTIONS WITH OVERLY.

Booiq NUREG-0040 V17 NO3: LICENSEE CONTRACTOR AND VENDOR IN.

ING WATER POOLS The WETCOR 1 Test.

SPECTION STATUS REPORT. Quarterty Report, July-September Weld 1993 (White Book)

NUREG/CR-4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING Vertical Tube WELDS Semiannual Report, October 1991 - March 1992.

NUREG/lA4100 ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 NUREG/CR-4599 V03 N1: SHORT CRACKS IN PIPING AND PlPING AGAINST SIMPLE VERTICAL TUBES AND ROD DUNDLE TESTS.

WELDS Semiannual Report, Apni-September 1992.

NUREG/CR-5914. CHEMICAL COMPOSITION AND RT(NOT) DETERMi-Vessel Falture Mechanism NATIONS FOR M:DLAND WELD WF 70.

NUREG/CR-5642 LIGHT WATER REACTOR LOWER HEAD FAILURE NUREG/CR-5972. EFFECTS OF NONSTANDARD HEAT TREATMENT TEMPERATURES ON TEr4SILE AND CHARPY IMPACT PROPERTIES ANALYSIS.

' OF CARDON-STEEL CASTING REPAlR WELDS.

Vibration NURE(i/CR 6078. ANALYSIS OF CRACK INITIATION AND GROWTH IN XSOR Codes THE HIGH LEVEL VIBRATION TEST AT TADOTSU NUREG/CR-5360. XSOR CODES USERS MANUAL Yucca Mountain Viscometry NUREG/CR 6032. SOLIDUS AND LIQUIDUS TEMPERATURES OF NUREG/CR 4735 V08. EVALUATION AND COMPILATION OF DOE CORE-CONCRETE MIXTURES.

WASTE PACKAGE TEST DATA Biannual Report, August 1989 - Janu-ary 1990.

Vital Gas NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL-NUREG 1364 REGULATORY ANALYSIS FOR THE RESOLUTION OF HYDROLOGIC MECHAN ICAL CHEMICAL PROCESSES PERTINENT GENERIC SAFETY ISSUE 100: PIPING AND THE USE OF HIGHLY TO THE PROPOSED HIGH LEVEL WASTE REPOSITORY AT YUCCA COMBUSilBLE GASES IN VITAL AREAS.

MOUNTAIN.

l l

1

M

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.

ADV1SORY COMMITTEE (S)

NUREG-0540 Vid N11 ilTLE LIST OF DOCUMENTS MADE PUBLIC.

ADVISORY COMMITTEE ON NUCLEAR WASTE LY AVAILABLE November 1-30,1992.

NUREG-1423 V04 A COMPILATION OF REPORTS OF THE ADVISO-NUREG 0540 V14 N12. TITLE LIST OF DOCUMENTS MADE PUBLIC-RY COMMITTEE ON NUCL EAR WASTE July 1992 - June 1993.

LY AVAILABLE. December 1-31,1992.

ACRS - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUREG 0540 V15 NOI: TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG-1125 V14. A COMPILATION OF REPORTS OF THE ADVISO' LY AVAILABLE January 1 31,1993.

RY COMMITTEE ON REACTOR SAFEGUARDS.1992 Annual.

NUREG-0540 V15 NO2. TITLE LIST OF DOCUMENTS MADE PUBLIC-ATOMIC SAFETY BOARD (S) & PANEL (S)

LY AVAILABLE February 1-28,1993 NUREG-0540 V15 NO3: TITLE LIST OF DOCUMENTS MADE PUBLIC.

ATOMIC SAFETY & LICENSING BOARD PANEL LY AVAILABLE March 1-31 1993 NUREG-1363 V05 ATOMIC SAFETY AND LICENSING BOARD PANEL ANNUAL REPORT. Fiscal Year 1992.

NUREG 0540 V15 N04. TITLE LIST OF DOCUMENTS MADE PUBLIC-LY AVAILABLE.Apnl 1-30.1993 OFFICF OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

NUREG 0540 V15 N05: TITLE LIST OF DOCUMENTS MADE PUBLIC-OFC OF THE EXECUTIVE DIRECTOR FOR OPERATIONS LY AVAILABLE May 1 31.1993 NUREG-1485: UNAUTHORIZED FORCEO ENTRY INTO THE PRO-NUREG 0540 V15 N06. TITLE LIST OF DOCUMENTS MADE PUBLIC-CTED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7, N EG O 15 07:

LIST OF DOCUMENTS MADE PUBLIC-REGION 1 (POST 820201)

LY AVAILABLE. July 1.31. 1993.

NUREG4837 V12 N04 NRC TLD DIRECT RADIATION MONITORING NUREG 0540 V15 N08: TITLE LIST OF DOCUMENTS MADE PUBLIC-NETWORK Progress Report October-December 1992 LY AVAILABLE. August 1 31,1993 NUREG 0837 V13 N01: NRC TLD DIRECT RADIATION' MONITORING NUREG-0540 V15 NC9 TITLE LIST OF DOCUMENTS MADE PUBLIC-NETWORK Progress Report January March 1993 LY AVAILABLE. September 1 30.1993.

NUREG-0837 V13 NO2. NRC TLD DIRECT RADIANON MONITORING NUREG 0540 V15 N10: TITLE LIST OF DOCUMENTS MADE PUBLIC-NETWORK Progress Report Apnt-June 1993 LY AVARABLE. October 1-31, 1993.

NUREG 0837 V13 NO3. NRC TLD DIRECT RADIATION MONITORING NUREG-0750 V36101: INDEXES TO NUCLEAR REGULATORY COM-NETWORK Progress Repori July-September 1993 MISSION ISSUANCES. July September 1992.

NUREG/CR 5488. RISK BASED INSPECTION GUIDE FOR THREE NUREG4750 V36102. INDEXES TO NUCLEAR REGULATORY COM-MILE ISLAND NUCLEAR STATION UNIT 1.

MISSION ISSUANCES. July-December 1992.

NUREG/CR-5835. AUXILIARY FEEDWATER SYSTEM RISK-BASED NUREG-0750 V36 N01: NUCLEAR REGULATORY COMMISSION IS-INSPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 SUANCES FOR JULY 1992 Pages 1-45 NUCLEAR POWER PLANTS.

NUREG-0750 V36 NO2: NUCLEAR REGULATORY COMMISSION IS.

REGION 2 (POST 820201)

SUANCES FOR AUGUST 1992. Pages 47148.

NUREG/CR-5833. AUXILIARY F EEDWATER SYSTE M RISK-BASED NUREG-0750 V36 NO3: NUCLEAR REGULATORY COMMIS$10N IS-INSPECTION GUIDE FOR THE H B ROBINSON NUCLEAR POWER SUANCES FOR SEPTEMBER 1992. Pages 149-220.

PLANT.

NUREG-0750 V36 N04 NUCLEAR REGULATORY COMMISSION IS-REGION 5 (POST 820201)

SUANCES FOR OCTOBER 1992. Pages 221249 NUREG/CR 5836 AUXILIARY FEEDWATER SYSTEM RISK BASED NUREG 0750 V36 N05. NUCLEAR REGULATORY COMMISSION IS INSPECTION GUIDE FOR THE PALO VERDE NdCLEAR POWER SUANCES FOR NOVEMBER 1992 Pages 251350

PLANT, NUREG 0750 V36 N06 NUCLEAR REGULATORY COMMISSION IS-OFC OF ENFORCE MENT (POST 870413)

SUANCES FOR DECEMBER 1992. Pages 351396.

NUREG4940 VII N04. ENFORCEMENT ACTIONS. SIGNIFICANT AC.

NUREG-0750 V37101: INDEXES TO NUCLEAR REGULATORY COM-TIONS RESOLVED Quarterly Progress Report,0ciober-December MISSION ISSUANCES January-March 1993.

1992, NUREG-0750 V37102 INDEXES TO NUCLEAR REGULATORY COM-NUREG CD40 V12 N01. ENFORCEMENT ACTIONS SIGNIFICANT AC.

MISSION ISSUANCES January June 1993.

TIONS RESOLVED Quarter'v Progress Report, January.-March 1993 NUREG-0750 V37 N01 NUCLEAR REGULATORY COMMISSION IS-NUREG-0940 V12 NO2 ENFORCEMENT ACTIONS SIGNIFICANT AC SUANCES FOR JANUARY 1993. Pages 1-54 TIONS RESOtVED.Ouarterty Progress Repor1.Apnt-June 1993 NUREG 0750 V37 NO2; NUCLEAR REGULATORY COMMISSION IS OFC OF PERSONNEL (POST 87041h SUANCES FOR FEBRUARY 1993. Pages 55-134.

NUREG 0325 R16. U.S. NUCLEAR REGULATORY COMMISSION NUREG-0750 V37 NO3. NUCLEAR REGULATORY COMMISSION IS FUNCTIONAL ORGANIZATION CHARTS March 15,1993.

SUANCES FOR MARCH 1993. Pages 135-249.

NUREG 0750 V37 NOC NUCLEAR REGULATORY COMMISSION IS-EDO - OFFICE OF ADMINISTRATION (_ PRE 870413 & POST 890205)

SUANCES FOR APRIL 1993 Pages 251354 OF FICE OF ADMINISTR ATION (POST 890205)

NUREG 0750 V37 N05; NUCLEAR REGULATORY COMMISSION IS NUREG 1145 V09 US NUCLE AR REGULATORY COMM!SSION SUANCES FOR MAY 1993 Pages 355-418.

1992 ANNUAL REPORT NUREG-0750 V37 N06. NUCLEAR REGULATORY COMMISSION IS-DIVISION OF FREEDOM OF INFORMATION & PUBLICATIONS SERV-SUANCES FOR JUNE 1993 Pages 419 515 ICE S (POST 890205 NUREG-0750 V38 N01: NUCLEAR REGULATORY COMMISSION IS-NUREG-0304 V17 N04. REGULATORY AND TECHNICAL REPORTS SUANCES FOR JULY 1993 Pages 1-24.

(ABSTRACT INDEX JOURNALL Annual Compilation For 1992.

NUREG 0750 V38 N02. NUCLEAR REGULATORY COMMISSION IS NUREG 0304 V18 NOI: REGULATORY AND TECHNICAL REPORTS SUANCES FOR AUGUST 1993. Pages 25-79.

( ARSTR ACT INDEX JOURNAL). Compilation for First Quarter NUREG4750 V38 NO3 NUCLEAR REGULATORY CC MMISSION IS-1993.JanuarpMarch SUANCES FOR SEPTEMBER 1993 Pages 81168.

NUREG-0304 v18 N02 REGULATORY AND TECHNICAL REPORTS NUREG 0750 V38 N04. NUCLEAR REGULATORY COMMISSION IS-(ABSTRACT INDEX JOURNAL) Compdation For Secnnd Quarter SUANCES FOR OCTOBER 1993. Pages 169-186.

1993. Aprd-June NUREG 0936 VII N04. NRC REGULATORY AGENDA.Ouarterty NUREG-0304 V18 NO3 REGULATORY AND TECHNICAL REPORTS Report, October-December 1992.

(ABSTRACT INDEX JOURNAR Compilation For Third Quarter NUREG-0936 V12 N01: NRC REGULATORY AGENDA Ouarterty I D93.JulpSeptember Report. January-March 1993.

99 e

l 100 NRC Originating Organization Index (Staff Reports) l NUREG-0936 V12 NO2. NRC REGULATORY AGENDA Quaderly DIVISION OF HIGH-LEVEL WASTE MANAGEMENT (POST 870413)

Report.ApriLJune 1993 NUREG/CR-5917 V01: SENSITIVITY AND UNCERTAINTY ANALYST'S NUREG 0935 V12 NO3: NRC REGULATORY AGENDA Quarterfy APPLIED TO ONE-DIMENSIONAL RADIONUCLIDE TRANSPORT IN i

l Report, July-September 1991 A LAYERED FRACTURED ROCK MULTFRAC - Analytic Solutions And Local SensitMbes.

EDO OFFICE OF THE CONTROLLER (PRE 820418 & POST 890205)

GEOLOGY & ENGINEERING BRANCH (POST 910506)

OFFICE OF THE CONTROLLER (POST 890205)

NUREG/CR-4735 V08-EVALUATION AND COMPILATION OF DOE i

t NUREG-1470 V02. CHIEF FINANCIAL OFFICER'S ANNUAL REPORT WASTE PACKAGE TEST DATA. Biannual Report, August 1989 - Jan-i DIVI ON LOW-LEVEL WASTE MANAGEMENT & DECOMMISSION-L E-0 V BUD T STIMA ES F sca Years 1994 1995 0

UCLEAR REGULATORY COMMISSION INFOR-S 8 04t@ECOMMISSIONiNG MANAGEMENT PLAN.

[p p

y3 g

99 NUREG 1476 FINAL ENVIRONMENTAL IMPACT STATEMENT TO OFFICE OF STATE PROGRAMS (PRE 870413 & POST 911117)

CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE, OFFICE OF STATE PROGRAMS (POST 911117)

AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR NUREG-1479. RESULTS FROM TWO WORKSHOPS-STATE RADI CLIVE, UTAH Docket No 40-8989. Ermrocare Of Utah.Inc.

ATION CONTROL PROGRAMS DEVELOPING AND AMENDING NUREG-1476 DRFT: DRAFT ENVIRONMENTAL IMPACT STATE-l REGULATIONS AND FUNDING MENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE.

i STORE, AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL CUVE, UTAH Docket No. 40-8989, Envirocare Of Utah. Inc.

DATA DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207)

OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA DI-NUREG 1484 DRFT: DRAFT ENVIRONMENTAL IMPACT STATE-RECTOR MENT F OR THE CONSTRUCTION AND OPERATION OF CLAl-NUREG-0090 V15 N04' REPORT TO CONGRESS ON ABNORMAL BORNE ENRICHMENT CENTER. HOMER, LOUISlANA Docliet No.

n y s2 NU G0 8 01 EPO NGRESS ON ABNORMAL OPER N

525 W2 m SAMES WWW MM W NU NO E

T CONGRESS ON ABNORMAL anuan 1, W W @ b h E W OCCURRENCES April-June 1993.

NUREG 1272 V07 N01: OFFICE FOR ANALYSIS AND EVALUATION U.S. NUCLEAR REGULATORY COMMISSION OF OPERATIONAL DATA 1992 Annual Report - Power Reactors OF FICE OF THE GENERAL COUNSEL (POST 860701)

NUREG 1172 V07 NO2 OFFICE FOR ANALYSIS AND EVALUATION NUREG-0386 D00 ROS. UNITED STATES NUCLEAR REGULATORY i

OF OPE RATIOP AL DAT A.1992 Annual Report - Nonructors COMMISSION STAFF PRACTICE AND PROCEDURE NUREG-1474. EFFECT OF HURRICANE ANDREW ON THE TURnEY DIGEST Commission Appeal Board And Licensing Board POINT NUCLEAR GENERATING STATION FROM AUGUST 20-30.

Decisions July 1972 2 March 1992.

NUREG 0386 006 R06: UNITED STATES, NUCLEAR REGULATORY DIVISIC iOF OPERATIONAL ASSESSMENT (POST 870415)

NUREG-1467.

FEDERAL GUIDE FOR A RAD:OLOGICAL COMMISSION STAFF PRACTICE AND PROCEDURE RESPONSE. Supporting The Nuclear Regulatory Commission Dunng DIGEST. Commission, Appeal Board And Licensing Board The In<tial Hours 01 A Senous Accident' Decisions. July 1972 - June 1992.

INCIDENT RESPONSE BRANCH NUREG-0388 D06 R07: UNITED STATES NUCLEAR REGULATORY NUREG/CR 5247 V01 R1: RASCAL VERSION 2.0 USER S GUIDE.

COMMISSION STAFF PRACTICE AND PROCEDURE NUREG/CR 5247 V02 RASCAL VERSION 2 0 WORMBOOK DIGEST Commssion, Appeal Board And Licensing Board DIVISION OF SAFEif PROGRAMS (POST B70413)

Decisions. July 1972 - September 1992.

NUREG-1275 V09. OPERATING EXPERIENCE FEEDGACK REPORT

• NUREG-0960 V01 NO2: NUCLEAR REGULATORY LEGl' ATION 102d PRESSURE LOCxlNG AND THERMAL DINDING OF GATE Congress.

VAL VES Cornmercial Power Reactors NUREG-0980 V02 NO2: NUCLEAR REGULATORY LEGISLATION.102d TRENDS & PATTERNS ANALYSIS BRANCH Cor ess'HE INSPECTOR GENERAL (POST 890417)

NUREG/CR-5171: ENHANCEMENTS TO DATA COI.LECTION AND OFFICE FT REPORTING OF SINGLE AND MULTIPLE F AILURE EVENTS NUREG-1415 VOS NO2; OFFICE OF THE INSPECTOR NUREG/CR.5964.

SAPHIRE TECHNICAL REFERENCE GENERALSermannual Report. October 1,1992 - March 31,1993 MANUALIRRAS/ SARA VERSION 4 0 NUREG 1415 V06 NOI: OFFICE OF THE INSPECTOR EDO OFFICE OF INFORMATION RESOURCES $4ANAGEMENT & ARM NRC N DE E

Flll N IVb

. (POST 861109)

NUREG-It80. LOSS OF AN IRIDIUM-192 SOURCE AND THERAPY OFFICE OF INFORMATION RESOURCES MANAGEMENT (POST MISADMINISTRATION AT INDIANA REGIONAL CANCER CE E I

j AINTY'AN L'YSES

^

NUF EG 487 V11 FISCAL YEAR 1994-1998 INFORMATION TECH-NUgE 2

TV APPLIED TO ONE DIMENSIONAL RADIONUCLIDE TRANSPORT IN DIVI I N OF

& TE LECOMMUNICATIONS SERVICES (POST 890205)

A LAYERED FRACTURED ROCK. Evaluation Of The Limit State Ap.

NUREG-0020 V17. LICENSED OPERATING REACTORS STATUS proach.

0 DIVI i OF N ( R ATI S

F SE V C

(

8 EDO - OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

NUREGO910 R02 601; NRC COMPREHENSIVE RECORDS DISPOSI.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST e60720)

TION SCHEDULE NUREG-1266 V07; NRC SAFETY RESEARCH IN SUPPOR T OF REG-ULATION - FY 1992.

EDO-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CP-0132; TRANSACTIONS OF THE TWENTY.FIRST WATER OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS REACTOR SAFETY INFORMATION MEETING.

NUREG-0430 V12.

LICE NSED FUEL FACILITY STATUS' DNISION OF ENGINEERING (POST 870413)

REPORT. inventory Difference Data. July 1,

1991 June 30.

NUREG-1377 Rot NRC RESEARCH PROGRAM ON PLANT AGING:

1992 iGray Dook il)

LISTING AND SUMMARIES OF REPORTS ISSUED THROUGH DIVISION OF SAFEGUARDS & TRANSPORTATION (870413-930206)

SEPTEMBER 1993.

NUREG-0725 R09 PUBLIC INFORMAflON CIRCULAR FOR SHIP.

DIVISION OF REGULATORY APPLICATIONS (POST 870413)SURE AT MENTS OF IRRADIATED REACTOR FUEL NUREG 0713 V12. OCCUPATIONAL RADIATION EXPO DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST COMMERCIAL NUCLEAR POWER REACTORS AND OTHER 870729)

FACILITIES 1990. Twenty-Third Annual Report.

NUREG-0383 V01 R16: DIRECTORY OF CERTIFICAYES OF COMPLl-NUREG-0713 V13: OCCUPATIONAL RADIATION EXPOSURE AT ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC COMMERCIAL NUCLEAR POWER REACTORS AND OTHER Approved Packaps FACILITIES 1091. Twenty-Fourth Annual Report.

NUALG 0383 V02 H16. DIRECTORY OF CERTIFICATES OF COMPLl--

NUREG 0713 V14: OCCUPATIONAL RADIATION EXPOSURE AT ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certihcales D' COMMERCIAL NUCLEAR POWER REACTORS AND OTHER FA-Comphance.

CluTIES 1992. Twenty fifth Annual Report.

NUREG-0383 V03 Rt3 DIRECTORY OF CERTIFICATES OF COMPL i-NUREG.1307 R03.

REPORT ON WASTE BURIAL ANCE FOR HADIOACTIVE MATERIALS PACKAGES Report Of NRC CHARGES Escalation Of Decommissioning Waste Disposal Costs At Approved Quality Assurance Programs For Radioactive Matenals Low-level Weste Dunal Facihties Packages.

NUREG-1400' AIR SAMPLING IN THE WORKPLACE. Final Report.

NRC Originating Organization index (Staff Reports) 101 WASTE MANAGFMENT BRANCH (POST 910830)

DIVISION OF REACTOR PROJECTS tiljV,V (POST 901216)

NUREG/CR-5988: SOfL CHARACTERIZATION METHODS FOR UN.

NUREG-0797 S26: SAFETY EVALUATION REPORT RELATED TO SATURATED LOW-LEVEt WASTE SITES.

THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC DlVISION OF SAFETY ISSU L RESOLUTION (POST 880717)

STATION, UNIT 2 Docket No. 50-446.(Texas Utilities Electne NUREG4933 S15: A FRIORITIZATION OF GENERIC SAFETY Company,et al)

ISSUES.

NUREG-0197 S27: SAFETY EVALUATION REPORT RELATED TO NUREG 0933 S16: A PRIORITIZATION OF GENERIC SAFETY THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC STA-

)

N EG 164; REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY PROJ CT DIRECTORATE 111-3 COMBUSTIBLE GASES IN VITAL AREAS.

NUREG/CR-5829: AUXILIARY FEEDWATER SYSTEM RISK-BASED NUREG-1427: REGULATORY ANALYSIS FOR THE RESOLUTION OF INSPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR GENERIC ISSUE 143: AVAILABlUTY OF CH!LLED WATER POWER PLANT.

SYSTEM AND ROOM COOLING.

DIVISION OF OPERATIONAL EVENTS ASSESSMENT (870411-921003)

NUREG 1453. REGULATORY ANALYSl$ FOR THE RESOLUTION OF NUREG-1366: IMPROVEMENTS TO TECHNICAL SPECIFICATIONS GENERIC ISSUE 142 LEAKAGE THROUGH ELECTRICAL ISOLA.

SURVEILLANCE REQUIREMENTS.

TORS IN INSTRUMENTATION CIRCUITS.

DIVISION OF UCENSEE PERFORMANCE ' QUAUTY EVALUATION NUREG-1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF (870411 921003)

GENERIC ISSUE 153: LOSS OF ESSENTIAL SERVICE WATER IN NUREG-1220 H01: TRAINING REVIEh CRITERIA AND PROCE.

LWRS^

DURES.

NUREG-1483: REGULATORY ANALYSIS FOR THE RESOLUTION OF DIVISION OF REACTOR CONTROLS & HUMAN FACTORS (POST GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF.

921004)

COOLANT ACCIDENT IN UGHT WATER REACTORS.

NUREG-1021 R07: OPERATOR UCENSING EXAMINER STAND-NUREG-1472. REGULATORY ANALYSIS FOR THE RESOLUTION OF ARDS.

GENERIC ISSUE 57 Effects Of Fire Protection System Actuation On D

O EG 4 HU DOWN L

ER O AT f4 AT NU.

SEVE M

BRANCH CLEAR POWER PLANTS IN THE UNITED STATESFinal Report.

NUREG/CR 5968: A SIMPUFIED MODEL OF AEROSOL REMOVAL REACT ST MS BRA BY CONTAINMENT SPRAYS.

FOR 1990.

ADVANCED REACTORS BRANCH (POST 910830)

NUREG/CP4129: PROCEEDINGS OF THE WORKSHOP ON PRO.

DIVISION OF REACTOR INSPECTION & UCENSEE PERFORMANCE GRAM FOR ELIMINATION OF REQUIREMENTTg MARGINAL TO Nt E O

N04: UCENSEE CONTR/. 'OR AND VENDOR IN-SAFETY.

SPECTION STATUS REPORT. Quarterty Report. October December INT A NCY C MMITTEES, REVIEW GROUPS, ETC.

N1 G 0 1 f401: LICENSEE CONTRACTOR AND VENDOR IN.

NUREG 1477 DAFT FC: VOLTAGE-BASED INTERIM PLUGGING CRI.

SPECTION STATUS REPORT. Quarterty Report. January-March TERIA FOR STEAM GENERATOR TUBES. Draft Report For Com.

1993.(White Book) ment.

NUREG-0040 V17 NO2: UCENSEE CONTRACTOR AND VENDOR IN-SPECTION STATUS REPORT. Quarterly Report, April-June EDO. OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428) 1993 (White Book)

OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST NUREG-0040 V17 NO3: UCENSEE CONTRACTOR AND VENDOR IN-670411)

SPECTION STATUS REPORT Quarterly Report, July-September NUREG/CP 0128: PROCEEDINGS OF THE INTERNATIONAL WORK

• 1993(White Book)

SHOP ON THE CONDUCT OF INSPECTIONS AND INSPECTOR NUREG-1214 R11: HISTORICAL DATA

SUMMARY

OF THE SYSTEM-OUAUFICATION AND TRAINING.

ATIC ASSESSMENT OF UCENSEE PERFORMANCE.

DIVISION OF REACTOR PROJECTS - 1/11 (POST 670411)

NUREG-1214 R12: HISTORICAL DATA

SUMMARY

OF THE SYSTEM-NUREG-0847 S11: SAFETY EVALUATION REPORT RELATED TO ATIC ASSESSMENT OF UCENSEE PERFORMANCE.

THE OPERATION OF WATTS BAR NUCLEAR PLANT. UNITS 1 AND NUREG-1473: ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL

2. Docket Nos. 50-390 And 50-391.(Tennesee Valley Authorny)

INSPECTION (EDSFI) DATA BASE PROGRAM.

NUREG-0847 S12: SAFETY EVALUATION REPORT RELATED TO DIVISION OF ENGINEERING (POST 921004)

THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND NUREG-1482 DRFT FC: GUIDELINES FOR INSERVICE TESTING AT

2. Docket Nos 50-390 And 50 391.(Tennesee Valley Authonty)

NUCLEAR POWER PLANTS. Draft Report For Comment PROJECT DIRECTORATE l-4 NUREG-1488 DRFT FC: REVISED UVERMORE SEISMIC HAZARD NUREG/CR 5488: RISK-BASED INSPECTION GUIDE FOR THREE ESTIMATES FOR 69 NUCLEAR POWER PLANT SITES EAST OF MILE ISLAND NUCLEAR STATION UNIT 1.

THE ROCKY MOUNTAINS. Draft Report For Comment.

m

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 I

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 EDO - OFFICE OF NUCt.E AR REGULATORY RESEARCH (POST 820405)

NURE G /l A >)110 ASSESSMENT OF RELAP5/ MOD 2 AGAINST A OF flCE OF NUCll AR RE GULAIORY RE SE ARCH (POST 060720)

MAIN F EEDWATER TUABOPUMP TRIP TRANS!ENT IN THE VAN-NUH E G /l A-0090- ASSESSMENT OF RE LAP 5/ MOD 2 USING THE TE ST DAT A OF REWET-li HF Fl OODING EXPER: MENT SGl/R DELLOS 4 NUCLEAR POWER PLANT.

NUREG/lA-0112 ASSESSMENT OF RELAP5/ MOD 2 AGAINST ECN-NUHEG/lA-0091 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A F AN ORCS L NURE 011 F

NARY ASSESSMENT OF PWR STEAM NURE G/lA 0092 ASSESSMENT ()F RELAP5/ MOD 2 COMPUTE R GENERATOR MODELLING IN RELAP5/ MOD 3 CODE AGAINST 1HE NET LOAD THIP TEST DATA FROM YONG-NUREG/lA-0116 ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST GW ANG UNIT 2 THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN NUR E G /l A.0094 ASSESSMENT OF RE LAPS / MOD 3 AGAINST PWR HOT LEG)

TWE NTY FivE POSI DRYOUT EXPER;MENTS PERFORMED AT NUREG/lA-0118. ANALYS!S OF LOFT TEST LS-1 USING RELAP5/

g THE ROYAL lNSTITUTE OF TECHNOLOGY MOD 2 NUHEG/IA 0095 HELAPS ASSESSME NT USING LSTF TEST DAT A NUREG/lA-0119 ASSESSMENT AND APPLICA RON OF BLACKOUT SB CL 10.

TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

NURE G /l A-0096 NUMERICS AND IMPLE ME NTAllON OF THE UK MOD 2 HOhl2ONTAL S T RA TIF ICAllON ENTRAINMENT OFF TAKE MODE L INTO RELAPS/ MOD 3 NUREG/lA-0120- ASSESSMENT OF THE TURD lNE TRfP TRANSIENT

^

IN COFRE NTES NPP WITH TRAC-BFI L

T S-NUREG/lA 0122 ASSESSMENT OF MSN T ULL CLOSURE FOR NURE G!!A-0100 ASSESSMENT OF CCFL MODEL Or RE LAP 5/

SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING MOD 3 AGAINST SlMPLE VERTICAL TUBES AND ROD DUNDLE TRAC BF1 (G1J1)

TESTS NUREG/l+0123 APPLICATION OF FULL POWER BL ACkOUT FOR NURf G/IA-0103 ASSE SSME NT OF DETHSY TE ST 918 USING C N ALMARAZ WITH RELAPS/ MOD 2 RE LAPS / MOD 3 NUREG/lA-0124 ASSESSME NT OF RELAP5/ MOD 2 AGAINST A CHE G/l A 0104 RE LAP 5/ MOD 3 ASSESSMENT USING THE SE MtS-PRE SSURf 2E R SPRAY UALVE INADVERTED F ULLY OPE NING cat E 50% F EED tlNE BREAK TE ST S FS 11 TRANSIENT AND RECOVERY BY NATURAL CIRCULATION IN NURE G/tA 0105 ASSE SSMENT OF HE L AP5/ MOD 3 VERSION SMS JOSE CABRERA NUCLEAR STATION USING INADVERTENT SAFETY INJf CTION INCIDENT DATA OF NURE G/l A 0125 ASSESSMENT OF RE LAPS / MOD 2 COMPUTEH CODE AGAINST THE NATURAL CIRCULATION TEST DATA FROM NLf I 01 A SESSMENT OF PWR STEAM GENERATOR MODEL L ING IN RE L APS/ MOD 2 NUREG/lA 0107 ASSESSMENT Or RE LAPS / MOD 2 AGAINST A NUREG/lA-0126 2D/3D PROGRAM WORK

SUMMARY

REPORT LOAD HEJF CTION FROM 100% TO 50% POWE R IN 'HE VAN.

NUREG/iA 0127 REACTOR SAFETY ISSUES RESOLVED DY THL_

DE LLOS ll NUCLE AR POWER PLANT.

2D/3D PROGRAM NUHEG/lA 0108 ASSE SSf 'E NT OF RELAP5/ MOD 2 AGAINS T A NUREG/tA0128. lNTERNATIONAL CODE ASSESSMENT AND APPLl-TURDINE THtP FROM 100% POWER IN THE VANDELLOS 11 NU.

CAllONS PROGRAM

SUMMARY

OF CODE ASSE SSMENT STUD-CLI AR POWE R PL ANT.

IES CONCERNING RE LAPS / MOD 2. RELAPS/ MOD 3. AND TRAC-0.

NUHE G/ A-0109 ASSESSMENT OF RELAP5/ MOD 2 AGAINST A 10%

DIVISION OF ENGINE E AlNG (POST 870413)

L OAD REJEC'iiON TRANSlE NT FROM 75% STE ADY ST ATE IN NURE G/l A 0085 ASSESSMENT OF FULL POWER TURBINE TRIP THE V ANDE L L OS 11 NUCL E AR POWL R PLANT START-UP TEST F OR C. TRlLLO 1 WITH REL APS/ MOD 2 l

I i

103 I

M

NRC Contract Sponsor Index (Contractor Reports)

This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by 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. Il further information is needed, refer to the main citation by the NUREG/CR number.

(DO - Of'FICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTtO CRACKING DATA IN LIGHT WATER REACTORS Semiannual Report,Apnl September DIVISION Or OPF RAflONAL ASSESSMENT (POST 870413) 199p NURE G/CR 5247 V01 RI RASCAL VE RSiON 2 0 USE R S GUIDE NUREG/CR-4667 V16. ENVIRONMENTALLY ASSISTED CRACKING Di P

lTY R A

B 04 IN LIGHT WATE R REACTORS Somiannual Report October 1992 -

Lt1 t

NUR G 4744 V07 NL LONG. TERM EMBRITTLEMENT OF CAST NUREG/CR 5953 STUDIES OF HUMAN PERFORMANCE DURING DUPLE X STAINLESS STEELS IN LWR SYSTEMS Semiannual OPERA RNG EVENTS 1990 1992 Report. October 1991 March 1992 NUREG/CR 4744 V07 N2. LONG-TERM EMBRITTLEMENT OF CAST EDO OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM DUPLE X STAINLE SS STEELS IN LWR SYSTEMS Semiannual POST 861109)

Report.Apnt-September 1992 Of f ICE OF INF ORMATiON RESOURCES MANAGE MENT (POST NUREG/CR 4832 V08 ANALYSIS OF THE LASALLE UNIT 2 NUCLE-890205)

AH POWER PLANT RISK METHODS INTEGRATION AND EVALUA.

NUREG/CR 2907 VII RADIOACTIVE MATER!ALS RELEASED F ROM TlON PROGRAM (RMIEPLSrnsmic Analysis NUCLEAR POWE R PLANTS Annual Roport 1940 NUREG/CR-5358 REVIEW OF ASME CODE CRITERIA FOR CON-OlVISION OF COMPU TE R & T E L E COMMUN1 CATIONS SERVICES TROL OF PRIMARv LOADS ON NUCL EAR PIPING SYSTEM NL 25 V 11. DOSE COMMITMENTS DUE TO RADIOAC-T D

0 N

ORK TIVE RE LE ASES F ROM NUCLEAR POWER PLANT SITES IN 1989 NURE G /CR-5404 V02 AUXILIARY FEEDWATER SYSTEM AGING EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS aw l Mown %

OtVISION OF SAF EGUARDS & TRANSPORTATION (870413 9302061 NUREG/CR 5410. STATISTICALLY BASE D REEVALUATION OF NUREG/CR 6007, STRESS ANALYSIS OF CLOSURE BOLTS FOR PISC Il ROUND ROBIN TEST DATA SHlPPiNG CASKS NUREG/GR 5591 VOI N2: HEAVY-SECTION STEEL IRRADIATION DIVISION OF HIGH LE VE L WASTE MANAGEMENT (POST 870413)

PROGRAM Semiannual Progress Report For Apni-Septemtier 1990.

NUPEG/CH 4735 VW EVAL UATION AND COMP!LATION OF DOE NUREG/CR 5699 VO1 AGINu AND SERVICE WEAR OF CONTROL W AS1E PACK AGE TE ST DAT A Biannual Report. August 1989 Jan.

ROD DRIVE MECHANISMS FOR BWR NUCLEAR PLANTS uary 19un NUREG/CR 5754 BOILING-WATE R REACTOR INTERNALS AGING NUREG/CH 5917 Voi SE NSITIVITY AND UNCE RTAINTY ANALYSES DEGRADATION STUDY. Phase 1 APPLIED TO ONE DIME NSIONAL RADIONL.CLIDE TRANSPORT IN NUREG/CR 5755. STIFFNESS OF LOW-ASPECT RATIO. REIN-A L AYF Rf D F RACTURED ROCK MULT F RAC - Analytc Solutions FORCED CONCRETE SHEAR WALLS An.1 L ocal Sonsitetes NUREG/CR-5776 DAMPlNG IN LOW-ASPECT RATIO. REINFORCED NUREG/CR 5917 V02 SENSITIVITY AND UNCERTA!NTY' ANALYSES CONCRETE SHEAR WALLS APPLII D TO ONE-DIMENSIONAL RADIONUCliDE TRANSPORT IN NUREG/CR 5778 V03 NEW YORK /NEW JERSEN REGIONAL SEIS-A L AYE Rf D FRACTURE.D ROCK Evaluation Of The Limit State Ap.

MIC NETWORK. Final Repo 1 For Apnl 1985 - September 1992.

proach NUREG/CR 5782. PRESSURIZED THERMAL SHOCK PROBABILIS-NURE G/CR 6021 A LITE RATURE REVIEW OF COUPL EU THERMAL.

TIC FRACTURE MECHANICS SE NSITIVITY ANALYSIS FOR HYDROLOGIC MECH ANICAL -CHE MICAL PROCESSE S PE RTI-YANKE E ROWE REACTOR PRESSURE VESSEL NE NT TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT NUREG/CR-5783 AG!NG A9SESSMENT OF THE COMBUSTION EN-YUCCA MOUNTAIN G!NEERING AND BABCOCK & WILCOX CONTROL ROD DRIVES.

DIVISION OF LOW LEVEL WASTE MANAGEME NT A DECOMMISSION-NUREG/CR 5844 AGING ASSESSMENT OF Bl STABLES AND ING (POST fl70413)

SWITCHES IN NUCLEAR POWER PL ANTS NURE G/CR 5911 SOURCE TERM EVALUATION FOR RADIO ACTIVE NUREG/CR-5851. LONG TERM PERFORMANCE AND AGING CHAR-LOW t E VEL WASTE DISPOSAL PERFORMANCE ASSESSME NT AC TERISTICS OF NUCLEAR PLANT PRESSURE TRANSMITTERS.

NUHiG'CR 5938 NATIONAL PROFILE ON COMMERCIAL LY GEN-

, / CR -5903. VALfDATION OF SMART SENSOR TECHNOL-i RATED LOW LEVEL RADIOACTIVE MIXED W ASTE F OR INSTRUMENT CAllBr..\\ TION REDUCTION IN NUCLE-NURE G/CH 6041 DISPOSAL UNIT SOURCE TERM (DUST) DATA M R PLANTS INPUT GutDE DIVISION OF F UE L CYCt E SAF ETY & SAFEGUARDS (POST 910?O7 NUni 0 ( H 5914 CHEMICAL COMPOSITION AND RT(NDT) DETER-NUREG/CR-6118 ASSESSMENT OF THE EF F ECTIVE NESS OF TH MlNATIONS FOR MIDI AND WEL D WF-70.

L E U REf 0RM RULE AND ITS IMPLE MENT ATION NUREG/CH 5926 SANS INVESTIGATION OF LOW ALLOY STEELS IN NEUT RON IRRADIATED, ANNEALED, AND REIRRADIATED O

L RESE ARCH (POST 820405)

N F G ST NU E /

44 A CHARACTERi2ATION OF CHECK VALVE DEG-NUREG/CH 4219 V09 N2-HEAVY-SECTION STEEL TECHNOLOGY RADATION AND F AILURE EXPERIENCE IN THE NUCLEAR PROGRAM Semennual Pro.yess Report for Apoi september 1992 POWER INDUSTRY.

NUREG /CR 4273-CRACK PROPAGATION IN HIGH S T R AIN R E.

NUREG/CR-5952. EVALUATION OF CRACK POP-INS AND THE DE-GIONS OF SEQUOYAH CONT A6NME NT TERMINATION OF THElR RELEVANCE TO DESIGN CONSIDER.

NUREG/CR 4469 V15 NONDESTRUCTIVE EXAMINATION (NDE) RE.

ATIONS LIABIL ITY FOR INSE RVICE INSPE CTION OF LIGHT WATER NUREG/CR 5955 MATERIALS AND DESIGN BASES ISSUES IN RE ACTORS Semiannual Repoe1. October 1991 March 1992 ASME CODE CASE N 47 NUREG /CR 4469 V16 NONDE STRUCTIVE EXAM:N ATION (NDE) RE-NURE G/CR 5958 TWO. PARAMETER FRACTURE MECHANICS L I ABillTY FOR INSE RVICE INSPE CTION OF LIGHT WATER THEORY AND APPLICAT!ONS.

RE ACTORS Sem: annual Report. Apol 1992-September 1992 NUREG/CR 5961 POSTTE ST DESTRUCTIVE EXAMINATION OF THE NUREG>CR~4594 V02 N2 SHORT CRACKS IN PIPING #ND P PIN 7 STEEL LINER IN A I & SCALE REACTOR CONTAINMENT MODEL.

wt L DS Semiannual Ronori. October 1991 - March 1992 NURE G/CR 5968 POTENTIAL CHANGE IN F LAW GEOMET RY OF

(

WUni G. CR 4599 VO3 N1 SHORT CRACKS IN PIPING AND P! PING AN INITI AL LY SHALLOW FINITE-LENGTH SURFACE FLAW AEi DS SemiannuW Heport Apnl Septemte 1992 DURING A PRESSURIZED-THERMAL-SHO CK TRANSIENT.

105

m 106 NRC Contract Sponsor index NUREG/CR-5969: J AND CTOD ESTlMATION EQUATIONS FOR NUREG/CR-5817 V03 N2; NRC HIGH-LEVEL RADIOACTIVE WASTE SHALLOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS RESEARCH AT CNWRA. July December 1992.

NUREG/CR-5970: APPROXIMATE TECHNIQUES FOR PREDICTING NUREG/CR-5883; HEALTH RISK ASSESSMENT OF IRRADIATED SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC)

TOPAZ.

NUREG/CR 5971: CONTINUUM AND MICROMECHANICS TAEAT-NUREG/CR 5884 VI DRF: REVISED ANALYSES OF DECOMMIS-MENT OF CONSTRAINT IN FRACTURE.

SiONING FOR THE REFERENCE PRESSURIZED WATER REAC-NUREG/CR5972: EFFECTS OF NONSTANDARD HEAT TREATMENT TOR POWER STATION Effects Of Current Regulatory And Other TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPER-Considerations On The Financial Assurance.. Main Report. Draft TIES OF CARDON. STEEL CASTING REPAIR WELDS.

Report For Comment NUREG/CR5981: THE EFFECT OF ELECTRIC DISCHARLI.; MA' NUREG/CR-5884 V2 'DRF: REVISED ANALYSES OF DECOMMIS-CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVER-SiONING FOR THE REFERENCE PRESSURIZED WATER REAC.

ons ahons On he Mandal Assurance-Appendices man NU EG/C 988 SO L CHARACTERIZATION METHODS FOR UN.

SATURATED LOW-LEVEL WASTE SITES Report For Comrnent NUREG/CR5997: CSNI PROJECT FOR FRACTURE ANALYSES OF NWEWNBR RANN%W CHAMCMAM & N LARGE-SCALE INTERNATIONAL REFERENCE EXPERtMENTS TOR DECOMMISSIONING WASTE AND NEUTRON ACTIVATED (PROJECT F ALSIRE).

METALS.

NUREG/CR-5999: INTERIM FATIQUE DESIGN CURVES FOR NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE AS-CARBON. LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN SESSMENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE LWR ENVIRONMENTS WASTE DISPOSAL FACILITIES Evaluaten Of Modeling Approaches.

NUREG/CR 6011: REVIEW OF SLUCTURE DAMPING VALUES FOR NUREG/CR-5943: SENSITIVITY ANALYSIS AND BENCHMARKING ELASTIC SEISMIC ANALYSG OF NUCLEAR POWER PLANTS.

OF THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

NUREG/CR-6012: STIFFQS AND DAMPING PROPERTIES OF A NUREG/CR-5962: HEALTH AND SAFETY IMPACTS FROM DIS-LOW ASPECT RA% SHEAR WALL BUILDING BASED ON RE.

CRETE SOURCES OF NATURALLY OCCURRING AND ACCELERA-CORDED FF.ir1 QUAKE RESPONSES.

TOR-PRODUCED RADIOACTIVE MATERIALS (NARM).

NURF%6013: METHODS USED FOR THE TREATMENT OF NUREG/CR-5980: THREE DIMENSIONAL REDISTRIBUTION OF NON PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS TRITIUM FROM A POINT OF RELEASE INTO A UNIFORM UN-NUREG/CR4015: STRUCTURAL AGING PROGRAM TECHNICAL SATURATED SOIL.A Deterministic Model For Tntium M gration in An PROGRESS FOR PERIOD JANUARY - CECEMBER 1992.

NUREG/CR-6023; GENERIC ANALYSIS FOR EVALUATION OF LOW And Disposal Site.

CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MAR-NUREG/CR 5987: MICROBIAL-INFLUENCED CEMENT DEGRADA-TION - LITERATURE REVIEW.

GINS AGAINST FRACTURE OF REACTOR PRESSURE VESSEL NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMITY DOSI-MATERIALS.

NUREG/CR-6029 V01: AGING ASSESSMENT OF NUCLEAR AIR-METERS-PILOT, TEST.

NUREG/CR 5991: PORFLOW: A MULTIFLUID MULTIPHASE MODEL TREA1 MENT SYSTEM HEPA FILTERS AND ADSORBERS. Phase 1.

NUREG/CR4031: CAVITATION GUIDE FOR CONTROL VALVES FOR SIMULATING FLOW. HEAT TRANSFER, AND MASS TRANS-NUREG/CR 6034' OKLAHOMA SEISMIC NETWORK. Final Report-PORT IN FRACTURED POROUS MEDIA User's Manual, Version NUREG/CR-6036: INITIAL RESULTS OF THE INFLUENCE OF BIAX-2.41.

IAL LOADING ON FRACTURE TOUGHNESS NUREG/CR-5996: SUBSURFACE INJECTION OF RADIOACTIVE NUREG/CR4043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC TRACERS Field Expenment For Model Validation Testing.

CHILLERS USED IN NUCLEAR POWER PLANTS Phase l-NUREG/CR-5998: SIMULATION OF UNSATURATED FLOW AND NUREG/CR-6048: PRESSURIZED-WATER REACTOR INTERNALS NONREACTIVE SOLUTE TRANSPORT IN A HETEROGENEOUS AGING DEGRADATION STUDY. Phase 1-SOIL AT THE FIELD SCALE NUREG/CR 6052. METHODOLOGY FOR RELIABILITY BASED CON-NUREG/CR 6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-DITION ASSESSMENT. Application To Concrete Structures in Nucle.

TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY

^

^

FRACT R R

DU N

R 6058. VIRGINIA REGIONAL SEISMIC NETWORKFinal NURE R N ING FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEO-NUR R f1 PACT OF ENDF/B VI CROSS-SECTION DATA LOGC MEDIA. Theory And User's Manual - Version 1.1.

ON H B. ROBINSON CYCLE 9 DOSIMETRY CALCULATIONS NUREG/CR 6047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF NUREGICR4078 ANALYSIS OF CRACK INITIATION AND GROWTH VANADOUS AND VANADIC lONS.

IN THE HIGH LEVEL VfBRATION TEST AT TADOTSU' OF EARTH.

NUREG/CR-6050; RADIATION EXPOSURE MONITORING AND IN-NUREG/CR4079: SEISMOLOGICAL INVESTIGATION FORMATION TRANSMITTAL (REMIT) SYSTEM User's Manual.

OUAKES IN THE NEW MADRID SEISMIC ZONE. Final NUREG/CR4054 DRF FC: ESTIMATING PRESSURllED WATER RE-ReportSepternber 1986 December 1992 ACTOR DECOMMISSIONING COSTS.A User's Manual For The NURtG/CR-6085 UNITED STATES SEISMOGRAPHIC NETWORK.

PWR Cost Estimatng Computer Program (CECP) Software. Draft NUREG/CR4096 LOADING RATE EFFECTS ON STRENGTH AND FRACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF Report For Comment NUREG/CR-6062. PERFORMANCE OF PORTABLE RADIATION THE IPIRG PROGRAM.

SURVEY INSTRUMENTS.

DIVistON OF REGULATORY APPLICATIONS (POST 870413)

NUREG/CR-6070; MODELING APPROACHES FOR CONCRETE BAR-NUREG/CR 3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-RIERS USED IN LOW-LEVEL WASTE DISPOSAL CLEAR POWER PLANTS. ANNOTATED BIBLIOGRAPHY OF SE-NUREG/CR4073; LYSIMETER LITERATURE REVIEW.

LECTED READINGS IN RADIATION PROTECTION AND ALARA.

NUREG/CR-4214 RIP 2A2: HEALTH EFFECTS MODELS FOR NU-NUREG/CR-6080: REPLACEMENT ENERGY, CAPACITY, AND RELi-CLEAR POWER PLANT ACCIDENT CONSEQUENCE ABILITY COSTS FOR PERMANENT NUCLEAR REACTOR SHUT-ANALYSIS Modif6 cation Of Models Resulting From Addition Of El-DOWNS.

NUREG/CR-6081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-tects CM 1 wposure To Alpha.Emiting Radionuctedes.Part II. Scientific CLES BY FLOOROCARBON SURF ACTANT SOLUTIONS.

NUREG

?t4 R2 PTt: HEALTH EFFECTS MODEL FOR NUCLE-NUREG/CR-6108: SPHERICAL DIFFUSION OF TRITIUM FROM A 2

' Matth....

Bases AR PC PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part 1:

POINT OF RELEASE IN A UNIFORM UNSATURATED SOIL A De-term #nistic Model For Tntium Migration in An Arid Disposal Site.

introdC degraten.And Summart NUREG/L H :1 19 VOS: FIELD LYSIMLTER INVESTIGATIONS: LOW-NUREGICR4114 VO1: APPLICATION OF AN INFILTRATION EVAL-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR UATION METHODOLOGY TO A HYPOTHETICAL LOW-LEVEL WASTE DISPOSAL FACILITY.

FISCAL YEAR 1992. Annual Report NUREG/CR-5631 R1 ADD CONTRIBUTION OF MATERNAL RADIO-DIVISION OF SAFETY ISSUE RESOLUTION (POST 880717) '

NUCLIDE BURDENS TO PRENATAL RADIATION NUREG/CR-4551 V7R1P1: EVALUATION OF SEVERE ACCIDENT RISKS: ZION UNIT 1. Main Report DOSES. Relationships Between Annual Limits On intake And Prenatal NUREG/CR-4551V7R IP2A: EVALUATION OF SEVERE ACCIDENT Doses.

RISKS: ZION UNIT 1. Appendix A.

NUREG/CR-5672 V03: CHARACTERISTICS OF LOW LEVEL RADIO-ACTIVE DECONTAMINATION WASTE. Annual Report For Fiscat NUREG/CR 4551V7R1P20 EVALUATION OF SEVERE ACCIDENT RISKS: 2 TON UNIT 1.Appondlces B, C, D, And E.

Year 1992.

NUREG/CR-4832 V05: ANALYSIS OF THE LASA'.LE UNIT 2 NUCLE-NUREG/CR 5817 V02. NRC HIGH-LEVEL RADIOACTIVE WASTE RE-AR POWER PLANT: RISK METHODS INTEGRATION AND EVALUA-SEARCH AT CNWRA. Calendar Year 1991 NUREG/CR-5817 V03 N1; NRC HIGH-LEVEL RADIOACTIVE WASTE TlON PROGRAM Parameter Estimation Analysis And Screening RESEARCH AT CNWR A January-June 1992.

Human Reliability Analysis.

.q

NRC Contract Sponsor index 107 NUREG/CR-4832 V09 ANALYSIS OF THE LASALLE UNIT 2 NUCLE-NUREG/C45982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN AH POWER PLANT: RISK METHODS INTEGRATION AND EVALUA-CONTAINMENT MANAGEMENT.

T10N PROGRAM (RMIEP) Internal Fra Analvssa NUREG/CR-5983: SAFETY ASPECTS OF FORCED FLOW COOL-NUREG/CR-5305 V02 P1: INTEGRATED Al5K ASSESSMENT FOR DOWN TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology COOLED REACTORS.

And Risk Uncertainty Evaluation Program (PRUEP) Appendices A.C.

NUREG/CR-5984 CODE AND MODEL EXTENSIONS OF THE NUREG/CR-5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR THATCH CODE FOR MODULAR HIGH TEMPERATURE GAS.

THE LASALLE UNIT 2 NUCLEAR POWER PLANT:Phenomenology COOLED REACTORS.

Antt Risk Uncertain Evaluation Pr am (PRUEP) Appendices D G-NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION

/R ENHA NTS TO AA OLLECTION AND 0

REPORTING OF SINGLE AND MULTIPLE FAILURE EVENTS UAE DATA STATISTICS Summary Report.

NUREG/CR-5747: ESTIMATE OF RADIONUCLIDE RELEASE CHAR.

NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATION ACTERISTICS INTO CONTAINMENT UNDER SEVERE ACCIDENT AND SYSTEM UNAVAILABluTY EVALUATION BASED ON FAIL.

CONDITIONS Final Report UAE DATA STATISTICS. Detailed Descnption And Applications.

NUREG/CR-5759. RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS NUREG/CR-SM5: TECHNICAL SPECIFICATION ACTION STATE-STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR.

MENTS REQUIRING SHUTDOWN.A Risk Perspectrve With Applica-IZED WATER REACTOR PLANTS tion To The RHR/SSW Sysoms Of A BWR.

NUREG/CR 5791: RtSK EVALUATION FOR A GENERAL ELECTRIC NUREG/CR-6018: SURVEY ANU ASSESSMENT OF CONVENTIONAL BWR. EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON SJFTWARE VERIFICATION AND VAllDATION METHODS.

SAFETY RELATED EOUiPMENT. Evaluatx>n Of Genene issue 57.

NUREG/CR6025: THE PROBABluif N MM4K-i CONTAINMENT NUREG/CR 5801: PROCEDURE FOR ANALYSIS OF COMMON.

FAILURE BY MELT ATTACK OF THE LINER, CAUSE F AILURES IN PROBABILISTIC SAFETY ANALYSIS NUREG/CR-6032: SOUDUS AND LIOUIDUS TEMPERATURES OF NUREG/CR 5863. RISK ASSESSMENT OF ISOLATION DEVICES IN CORE-CONCRETE MIXTURES.

SAFETY SYSTEMS NUREG/CR4035: FEASIBILITY STUDY FOR IMPROVED STEADY-NUREG/CR-5901: A SIMPLIFIED MODEL OF AEROSOL SCRUBBING STATE INITIALIZATION ALGORITHMS FOR THE RELAPS COM-BY A WATER POOL OVERLYING CORE DEDRIS INTERACTING PUTER CODE.

WITH CONCRETE Final Report.

NUREG/CR 6056 A FRAMEWORK FOR THE ASSESSMENT OF NUREG/CA-5928 ISLOCA RtSEARCH PROGRAM Final Report.

NUREG/CR 5936: ENHANCEMENTS TO THE ACCIDENT PRECUR-SEVERE ACCIDENT MANAGEMENT STRATEGIES.

SOR METHODOLOGY.

NUREG/CR-6060: HYDROGEN MIXING STUDIES (HMS) ASSESS-MENT MANUAL' DETERMINATION OF THE BIAS IN LOFT FUE NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARAC-NUREG/CR 6061:

TERISTICS FOR SELECTED PEACH DOTTOM SEQUENCES PRE-PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN Nt G/ 5 H E TECHNICAL REFERENCE PHASE OF LARGE-BREAK LOCA EXPERIMENTS MANUALIRRAS/ SARA VERSION 4.0.

NUREG/CR4065: SYSTEMS ANALYSIS OF THE CANDU 3 REAC.

NURE A SIMPL ED MODEL OF AEROSOL REMOVAL NUREG/CR4072: EXPERIMENTAL STUDY ON THE COMBUS110N NUREG/CR-5976: DEVELOPMENT AND USE OF A TRAIN-LEVEL DEHAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT PROBABILISTIC RISK ASSESSMENT.

JET IGNITION AT LARGE SCALE.

NUREG /CR4027. PRELIMINARY EVALUATION OF SNUBBER NUREG/CR4111: INTEGRATED SYSTEMS ANALYSIS OF THE PIUS SINGLE F AILURES REACTOR.

NUREG/CR-6059: MACCS VERSION 1.5.11.1 A MAINTENANCE RE.

NUREG/CR 6113. CLASS 1E DIGITAL SYSTEMS STUDIES.

LEASE OF THE CODE.

NUREG/CR4084 VALUE.lMPACT ANALYSIS OF GENERIC ISSUE EDO OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428) 143 ' AVAILABILITY OF HEATING, VENTILATION. AIR CONDt.

OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST TIONING (HVAC) AND CHILLED WATER SYSTEMS "

870411)

NUREG/CR-6117: NEUTRON SPECTRA AT DIFFERENT HIGH FLUX NUREG/CR-6101: SOFTWARE REUABIUTY AND SAFETY IN NU-ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEILLANCE CLEAR REACTOR PROTECTION SYSTEMS.

LOCATIONS.

PROGRAM MANAGEMENT, POUCY DEVELOPMENT & ANALYSIS DIVISION OF SYSTEMS RESEARCH (POST 880717)

STAFF (POST 870411)

NUREG/CR 5455 V01: DEVELOPMENT OF THE NRC'S HUMAN PER-NUREG/CR4973. CODES AND STANDARDS AND OTHER GUID.

FORMANCE INVESTIGATION PROCESS (HP1P)

ANCE CITED IN REGULATORY DOCUMENTS.

NUREG/CR 5455 V02: DEVELOPMENT OF THE NRC'S HUMAN PER-NUREG/CR-5975: INCENTIVE REGULATION OF INVESTOR-OWNED F ORMANCE INVESTIGATION PROCESS (HP:P).

NUREG/CR-5455 VU3. DEVELOPMENT OF THE NRC'S HUMAN PER-NUCLEAR POWER PLANTS BY PUBUC UTluTY REGULATORS.

FORMANCE INVESTIGATION PROCESS (HPIP)

DIVISION OF REACTOR CONTROLS & HUMAN FACTORS (POST NUREG/CR 5642: UGHT WATER REACTOR LOWER HEAD FAILURE 921004)

NUREG/CR-6082. DATA COMMUNICATIONS NU E /

5818: UNCERTAIN 1Y ANALYSIS OF MINIMUM VESSEL

^

CE R CT S SSE LIQUlO INVENTORY DURING A SMALL-BREAK LOCA IN A B&W NUREG/CR4090: THE PROGRAMMABLE LOGIC CONTROLLER PLANT-AN APPLICATION OF THE CSAU METHODOLOGY USING THE RELAPS/ MOD 3 COMPUTER CODE.

AND ITS APPUCATION IN NUCLEAR REACTOR SYSTEMS.

NUREG/CR 5843 CORCON MOD 3.AN INTEGRATED COMPUTER DIVIS!ON OF SYSTEMS SAFETY & ANALYSIS (POST 921004)

NUREG/CR-3950 V08. FUEL PERFORMANCE ANNUAL REPORT MODEL FOR ANALYSIS OF MOLTEN CORE-CONCRETE FOR m INTERACTIONS UWs Manual NUREG/CR 5882 TRAC-B THERMAL HYDRAUUC ANALYSIS OF NUREG/CR 5488. RISK-BASED INSPECTION GUIDE FOR THREE MILE ISLAND NUCLEAR STATION UNIT 1.

^

0 NL REG /CR 590 E

N RETE INT R TIONS WITH OVER.

LYING W ATER POOLS The WETCOR-1 Test ON GwDE M mE SAN ONW W 2 WMAR NUREG/CA 5922. MODULAR HIGH-TEMPERATURE GAS-COOLED AL ESOSE M MW AND NU E /CR AUXIUARY FEEDWATER SYSTEM RISK-BASED T ITY A SIE N S INSPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR NUREGICR-5937: INTENTIONAL DEPRESSURl2ATION ACCIDENT POWER PLANT.

MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC-NUREG/C45833: AUXIUARY FEEDWATER SYSTEM RISK-BASED yong.

INSPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER NUREG/CR 5949; ASSESSMENT OF THE POTENTIAL FOR HiGH PLANT-PRESSURE MELT EJECTION RESUL11NG FROM A SURRY STA.

NUREG/CR-5834: AUXILIARY FEEDWATER SYSTEM RISK-BASED TION BLACh0VT TRANStENT.

INSPECTION GUlOE FOR THE FORT CALHOUN NUCLEAR NUFrG/CR 5951: THE MANAGEMENT OF ATWS BY BORON INJEC.

POWER PLANT.

itoN NUREG/C45835: AUXILIARY FEEDWATER SYSTEM RISK-BASED NURE G/CR4977 A PERFORMANCE INDICATOR OF THE EFFEC.

INSPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 TlVENESS OF HUMAN-MACHINE INTERFACES FOR NUCLEAR NUCLEAR POWER PLANTS.

POWER PLANTS NUREG/CR 5836: AUXIUARY FEEDWATER SYSTEM RISK-BASED NURE G/CR-5978 SOURCE TERM ATTENUATION BY WATER IN INSPECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER THE MARK 1 BOluNG WATER RE ACTOR DRYWELL PLANT.

ss

108 NRC Contract Sponsor Index NUHl G/CR 5047 AUptt TAR Y f EE (JWATE R SYSil M FIISK OASED NUREG/CR-6014 H:GH PRESSURE' COOLANT INJECTION SYSTE M RISK BASED INSPECTION GUOE FOH HATCH NUCLE'AH POWER INSPf CTION GlHOF FOR THE SOUTH TEXAS PROJECT NUCLL.

STAftON AH POWE Il Pl ANT.

NURE G/CR f3022 HIGH PRESSURE COOLANT INJF CTION (HPCI)

NURE G/Ch MW8 AU71t lARV f f E DW ATER SYSTF M RISK BASED SYSTEM A;SK-BASED INSPE CTION GUIDE FOR DROWNS F ERRY INSPE CTION GUIDE F On THf POINI BF ACH NUCLE AR POWE R NUCLE AR POWE R ST ATION P1 ANT DIVISION OF RADIATION SAFETY & SAFE GUARDS (POST 02I004 NUREGICH 5'm HIGH PRE SSURE COOL ANT INJECTION (HPCI)

NUREG/CR 5758 V03 F IT NESS F OR DUTY IN THE NUCLLAR SYSTEM RISK.OASE D INSPECilON GUIDE F OR DRf SDE N NU POWER INDUSTRY Annual Summary Of Program Perlormance CL E AH POWL R STATlON UNITS 2 AND 3 Heports.CY 1992 NUHE G/CD 5934 HIGH PRf SSURE COOLANT !NJi CTION (HPCI)

DIVlSION OF ENGINEE RING (POST 921004)

NUREG/GR-5956 CONSIDf RATION OF UNCERT AINTIES IN SOll.

SYS T E M HGK HASED INSPECTION GUIDE FOR OUAD C! TIES STRUCTURE INTERACTION COMPUTATIONS ST ATION UNITS 1 AND 2 NURE G/CR 5957: SYSTEM A., + (TM) CONT AINMENT - STRUCTUR-NUni_G/CR 5959 UfGH PHESGURL COOL ANT INJE CTION (HPCI)

^

SIGN EV Sv51EM hl5k OASED IN53'ECTION GOOF F OH F NRICO F ERMI Uf C

ATOMIC POWE h Pt ANT, UNIT 2 f RAL ele CTRIC ADVANCED BOILING WATER RE ACTOR-

i 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 titles of their reports. If further information is needed, refer to the main citation by the NUREG/CR number.

21ST CENTURY INDUSTRIES, INC.

AVAPLAN OY (FINLAND)

NURE G/CR 6116 ASSESSMENT OF THE EFFECTIVENESS OF THE NUREG/CR 5995 TECHNICAL SPECIFICATION ACTION STATEMENTS LEU REFORM RULE AND ITS IMPLEMENT ATION REQUIRING SHU1DOWN A Risk Pwspective With Applicahon To The

"""# b bN'**

A UW" ADVANCED SYSTEMS TECHNOLOGY, INC.

BATTELLE HUMAN AFFAIRS RESEARCH CENTERS C

Li SAFET M ACTS F M D SCRETE NUREG/CR 5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER SOURCES OF NATURALLY. OCCURRING AND ACCELERATORPRO.

INDUST RY. Annual Summary Of Progvam Performance Reports,CY DUCED RADIOACTIVF MATERIALS (NARM)

NUREG/CFI-6062; PERFORMANCE OF PORTABLE RADIATION 1992 SURVEY INSTRUME NTS.

BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES AMERICAN SOCIETY OF MECHANICAL ENGINEERS NUREG/CR 4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING NUREG/GR-0005 V02 P1: RISK-BASED INSPECTION-DEVELOPMENT WELDS Swniannual Report. October 1991 - March 1992.

OF GUIDELINES bght Water Reactor (LWR) Nuclear Power Plant NUREG/CR 4599 V03 N1. SHORT CRACKS IN PIPING AND PtPING Component" WELDS Semlannual Report. Apni-September 1992.

N

/R A

RA EGS ON S m W AND ANALYSIS & MEASUREMENT SERVICES COFtP FR ACTURE TOUGHNESS OF PIPE STEELS USED IN TASK 1 OF NURE G/CR-5851: LONG TERM PERFORMANCE AND AGING CHAR.

THE IPIRG PROGRAM.

ACTERISTICS OF NUCLEAR PLANT PRESSURE TRANSMITTERS NUREG/CR 5903. VALIDAllON OF SMART SENSOR TECHNOLOGIES BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST FOR INSTRUMENT CAllDRAilON REDUCTION IN NUCLEAR LABORATORY POWER PLANTS.

NUREG-1400 AIR SAMPLING IN THE WORKPLACE Final Report.

ANALYTIC & COMPUTIONAL RESEARCH,INC.

NUREG/CR 2850 V11 DOSE COMMITMENTS DUE TO RADIOACTIVE NUREG/CR 5991: PORFLOW: A MULTIFLUID MULTIPHASE MODEL RELF ASES FROM NUCLEAR POWER PLANT SITES IN 1989.

FOR SIMULATING FLOW. HEAT TRANSF ER, AND MASS TRANS-NUREG/CR-3950 V00 FUEL PERFORMANCE ANNUAL REPORT FOR PORT IN FRACTUREO POROUS MEDIA User's Manual - Vernon 1990.

2 41.

NUREG/CR-4214 RIPPA2: HEALTH EFf ECTS MODELS FOR NUCLE-AR POWER PL ANT ACCIDENT CONSEOUENCE NUREG/CR S

E bO0 ABILITY OF MARK 1 CONTAINMENT FAILURE BY MELT. ATTACK OF THE LINER.

NUREG/CR 4214 R2 PT1. HEALTH EFFECTS MODEL FOR NUCLEAR ARGONNE NATIONAL LABORATORY POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS.Part L NUREG/CR4667 VtS: ENVIRONMENTALLY ASSISTED CRACKING IN Introducton, Integration,And Summary.

LIGHT W ATE R REACTORS. Senuannual Report Aynt Septeerter 1992 NUPEG/CR4667 V16. ENVIRONMC NTALLY AS$laiED CRACKING IN NUREG/CR4469 V15. NONDESTRUCTIVE EXAMINATION (NDE) REll-1 LIGHT WATER REACTORS Sermannual Report, October 1992 March ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER DEACTORS.Somennual Report, October 1991. March 1992.

I 1993 NUREG/CR 4744 V07 NI: LONG-TERM EMBRITTLEMENT OF CAST NUREG/CR 4460 V16. NONDESTRUCTIVE EXAMINATION (NDE) RELi-DUPL EX STAINLESS SIEELS IN LWR SYSTEMS Senknnual ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Report, October 1991. March 1992.

REACTORS Semiannual Report.Apnl 1992 September 1992.

NUREG/CR-4744 V07 N2: LONG 1ERM EMBRITTLEMENT OF CAST NUREG/CR-5247 V01 RI: RASCAL VERSION 2.0 USEfrS GUtDE.

DUPLEX STAINl ESS STEELS IN LWR SYST E MS Semiannual NUREG/CR 5410. STATISTICALLY BASED REEVALUATION OF PISCit Report. Anni-September 1992 ROUND ROBIN TEST DATA.

NUREG/CA 5647. LIGHT WATER REACTOR LOWER HEAD FAILURE NUREG/CR$488: RISK-DASED INSPECTION GUIDE FOR THREE MILE NL L 2 ANALYSIS OF THERMAL MIXING AND BORON DI-NL EG/CR R1 A T B TlON OF MATERNAL RADIONI)-

CLIDE BURDENS TO PRENATAL RADIATION DOSES RelahonsNps NUREG/CR-6999 INTERIM FATIOUE DESIGN CURVES FOR CAR 00N, Between Annuae Limits On intake And Prenatal Doses.

LOW ALLOY, AND AUSTENITIC ST AINLESS STEELS IN LWR ENVI.

NUREG/CR-5758 V03 FITNESS FOR DUTY IN THE NUCLEAR POWER RONMENTS NUREG/GR 6025 THE PROBARILITY OF MARK.I CONTAINMENT INDUSTRY. Annual Summary Of Program Psiornence RefortsCY FAILURE BY MELT-ATTACK OF THE LINER 1992 NURE G/CR-6032: SOLIDUS AND LIQUIDUS TEMPERATURES OF NUREG/CR-57f4 AUXtLIARY FEEDWATER SYSTEM RISK BASED IN-CORE CONCAETE MIXTURES SPECTION GutDE FOR THE SAN ONOFRE UNIT 2 NUCLEAR NUREG/CR-6080- REPt. ACE ME NT E NERGY, CAPACITY, AND REll-POWER PLANT.

ABILITY CCSIS FOR PERMANENT NUCLEAR RE ACTOR SHUT

• NUREG/CR-5829' AUXlLIARY FEEDWATER SYSTEM RISK BASED IN-DOWNS.

SPECTION OUIDANCE FOR THE DAVIS BESSE NUCLEAR POWER PLANT ORIZONA, UNIV. OF, TUCSON, AZ NUREG/CR.5833 AUXILIARY FEEDWATER SYSTEM RISK.8ASED IN-NUREG/CP.0040- PROCEEDINGS OF WORKSHOP V: FLOW AND SPECTION GUIDE FOR THE H B. ROBINSON NUCLEAR POWER TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE-Pu NT LAIED TO HIGH LEVEL RAD!OACTIVE WASTE DISPOSAL Held At NUREG/CR 5834 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE FORT CALHOUN NUCLEAR POWER b

DS FOR UN.

NUI A

8 L

kA i SATURATED LOW. LEVEL WASTE SITES NUR CR 5835 AuxlLIARY FEEDWATER SYSTEM RISK-BASED IN.

SPECTION GUIDE FOR THE BEAVER VALLEY, UNITS 1 AND 2 NO-ATHEY CONSULTING NUREG/CR 5247 V02 RASCAL VERSION 2 0 WORKBOOK.

CLEAR POWER PLANTS 109

110 Contractor index NUREG/CR 5836. auxillary FEEDWATER SYSTEM RISK DASED IN-NUREG/CR-5911. SOURCE TERM EVALUATION FOR RADIOACTIVE SPECTION GUIDE FOR THE PALO VERDE NUCLEAR POWER LOW-LEVEL WASTE DISPOSAL PERFORMANCE ASSESSMENT.

PLANT NUREG/CR 5933: HIGH PRESSURE COOLANT INJECTION (HPCI) i HUREG/CR-5884 VI DRF: REVISED ANALYSES OF DECOMMISSION-SYSTEM RISK-DASED INSPECTION GUIDE FOR DRESDEN NUCLE ING FOR THE REFERENCE PRESSURIZED WATER REACTOR AR POWER STATION UNITS 2 AND 3.

POWER STATION f ffects Of Current Regulatory And Other Con $ dor-NUREGICR 5934. HIGH PRESSURE COOLANT INJECTION (HPCI) atmins On The Financal Assurancem.. Main Report Draft Report for SYSTEM RISK DASED INSPECTION GUIDE FOR OU AD-CITIES Comment STATION. UNITS 1 AND 2.

i NURE G/CH-5884 V2 DRF: REVISED ANALYSES OF DECOMMISSION-NUREG/CR 5943 SENSITIVITY ANALYSIS AND BENCHMARKING OF ING fOR THE REFTRENCE PRESSunilED WATER REACTOR THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

l POWER ST ATION Effects Of Current Regulatory And Other Consdor-NUREG/CR 5959 HIGH PRESSURE COOLANT INJECTION (HPCI) abons On The Financial Assurance.mAppendices Diaft Report For SYSTEM RISK-BASED INSPECTION GUIDE FOR ENRICO FERMI Comment ATOMIC POWER PLANT, UNIT 2.

NUREG/CR 5894. RADIONUCLIDE CHARACTER 12ATION OF REAC.

NUREG/CR 5982. EFFECTIVENESS OF CONTAINMENT SPRAYS IN CONT. !NMENT MANAGEMENT.

1 A

TOR DECOMM16SIONING WASTE AND NEUTRON ACTIVATED NUREG/CR 5983: SAFETY ASPECTS OF FORCED FLOW COOtDOWN j

ME T AL S.

NUREG/CR 5897. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS COOLED j

SPECTION GUIDE FOR THE SOUTH TEXAS PROJECT NUCLEAR REACTORS.

POWE R P1 ANT.

NUREG/CR 5984 CODE AND MODEL EXTENSIONS OF THE THATCH NUREG/CR48% AUXILIARY FF EDWATER SYSTEM RISK-DASED IN-CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLED REAC-SPECllON GUIDE FOR THE POINT DEACH NUCLEAR POWER TORS 1

PL ANT.

NUREG/CR-5993 V01 METHC S FOR DEPENDENCY ESTIMATION l

NURE G/CR 5973 CODES AND STANDARDS AND OTHE R GUIDANCE AND SYSTEM UNAVAILABILI ( EVALUATION DASED ON F AILURE CITED IN REGULATORY DOCUMENTS-DATA STATISTICS Summary ' sport.

NUREG/CR49/S INCENTIVE REGULATION OF INVESTOR. OWNED NUREG/CR-5993 V02: METHf OS FOR DEPENDENCY ESTIMATION NUCLEAR POWER PLANIS DY PUBLIC UTillTY RFGUL ATORS AND SYSTEM UNAVAILABI'..TY EVALUATION BASED ON FAILURE NUHEG/GR 5988 SOIL CHARACTERIZATION METHODS FOR UN' DATA STATISTICS Detailed Descnpton And Applicabons.

SATURATED LOW LEVEL WASTE SITES.

NUREG/CR 5995, TECHNICAL SPECIFICATION ACTION STATEMENTS NUREG/CR 5989 PERFORMANCE lESTING OF EXTREMITY DOSI-REQUIRING SHUTDOWN A Risk Perspective With Applicahon To The NU

/ F5 SURFACE INJECTION OF RADIOACTIVE NUR CR 14 HG P ESSURE COOLANT INJECTION SYSTEM RISK BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER SYM$lA ION O f SA Er F W AND NON NU E /F FC V

E TRANSPORT IN A HETEROGENEOUS SOlt AT yghA[fCR-6022:

0 HIGH PRESSURE COOLANT INJECTION (HPCI)

SYSTEM RISK DASED INSPECTION GUIDE FOR BROWNS FERRY NUREG/CR 4029 V01: AGING ASSESSMENT OF NUCLEAR AIR.

NUCLEAR POWER STATION TREATMENT SYSTEM HEPA FILTERS AND ADSORDERS Phaw l.

NUREG/CR 6041: DISPOSAL UNIT SOURCE TERM (DUST) DATA NUREG/CR 6043 V01: AGING ASSESSMENT OF ESSENTIAL. HVAC (NPUT GUIDE.

j CHIL LERS USED IN NUCLEAR POWER PLANTS Phase I NUREG/CR-6049. PIPING BENCHMARK PROBLEMS FOR THE GEN-NUREGICR-6054 DRF FC: EST4 MATING PRESSURIZED WATER RE.

ERAL ELECTRIC ADVANCED DOILING WATER REACTOR

)

ACTOR DECOMMISSIONING COSTS A User's Manual For The PWR NUREG/CR-6078 ANALYSIS OF CRACK INITIAT;ON AND GROWTH IN j

Cost Eshmahng Computer Program (CECP) Software Draft Report For THE HIGH LEVEL VlBRATION TEST AT TADOTSU.

Comment NUREG/CR-6111; INTEGRATED SYSTEMS ANALYSIS OF THE PIUS j

NUREG/CR.6084 VALUE-lMPACT ANALYSIS OF GENERIC ISSUE 143.

REACTOR.

"AVAILA0ftlTY OF HEATING, VENTILATION, AIR CONJITIONING 1HVAC) AND CHILL E D WATER SYSTEMS "

BROWN UNIV., PROVIDENCE, Rt NUREG/CR 6114 VO1: APPLICATION OF AN INFILTRATION EVALUA.

NUREG/CR-5958: TWO PARAMETER FRACTURE MECHANICS.

TION METHODOLOGY TO A HYPOlHETICAL LOW LEVEL WASTE THEORY AND APPLICATIONS DISPOSAL FACluTY' NUREG/CR.5971' CONTINUUM AND MICROMECHANICS TREATMENT OF CONSTRAINT IN FRACTURE.

i BROOKHAVEN NATIONAL LABORATORY l

NUREG/CP 0126 V01: PROCEEDINGS OF THE TWENTIETH WATER CALIFORNIA INSTITUTE OF TECHNOLOGY, PASADENA, CA RFACTOR SAF CTY INFORM ATION MEETING NUREG/CR-6012: SilFFNESS AND DAMPING PROPERTIES OF A NUREG/CP 0126 V02 PROCEEDINGS OF THE TWENTIETH WATER LOW ASPECT RATIO SHEAR WALL DUILDING BASED ON RECORD-RE ACTOR SAF ETY INF ORMATION MEETING NUHEG/CP4126 V03. PROCEEDINGS OF THE TWENTIETH WATER ED EARTHOUAKE RESPONSES.

AF ACTOR SAFETY INFORM ATION MEE TING NUREG/CR-2907 VII: RADIOACTIVE MATERVAS RELEASED FROM CALIFORNIA. UNfV. 0F, BERKELEY CA NUREG/CR-5980: THREE DIMENSIONAL REDISTRIBUTION OF TRITP NUCLEAR POWER PLANTS. Annual Report 1990 NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU.

UM FROM A POINT OF RELEASE INTO A UNIFORM UNSATURATED CLE AR POWER PLANTS. ANNOTATED BIBLIOGRAPHY OF SELECT.

SOILA Determinishc Model For Tritium Mgrahon in An And Dsposal ED RE ADINGS IN RADIATION PROTECTION AND ALARA.

S<te.

NUREG!CR 4214 RIP 2A2 HEALTH EF FECTS MODELS FOR NUCLE.

NUREG/CR 8108 SPHERICAL DIFFUSION OF TRITIUM FROM A AR POWER PL AN T ACCIDENT CONSEQUENCE POINT OF RELEASF IN A UNIFORM UNSATURATED SOIL.A Deter.

ANALYSIS Modihcahon Of Modets Resulting From Addition Of Effects min #stic Model For Tntium Mgrahon in An Arid Dsposal Site Of E posure To Alpha-Emitting Ruonuchdeo Part II: Sc entihc Bases CALIFORNIA, UNIV. OF, LOS ANGELES, CA f or Health.~.

NUMEG/CR 4214 R2 FT1: HEALTH EFFECTS MODEL FOR NUCLEAR NUREG/CR 6056: A FRAMEWORK FOR THE ASSESSMENT OF POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part I-SEVERE ACCIDENT MANAGEMENT STRATEGIES.

Introduction.Integrahon,And Summary.

NUREG/CR.4551 V7 RIP 1: EVALUATION OF SEVERE ACCIDENT

. CALIFORNIA, UNIV. OF, SANTA BARBARA, CA RISKS' ZION UNti 1 Main Report NUREG/CR 5951: THE MANAGEMENT OF ATWS BY BORON INJEC-NUREG/CR-4551V7R1P2A: EVALUATION OF SEVERE ACCIDENT TION.

I NUREG/CR-6025: THE PROBABluTY OF MAR 01 CONTAINMENT RISKS 210N UNIT 1 Apper* A.

NUREG/CR-4551V7RIPm EVALUATION OF SEVERE ACCIDENT FAILURE BY MELT ATTACK OF THE LINER.

RISKS 710N UNIT 1 Appendices B, C. D. And E.

NUREG/CR-5747: ESilMA1E OF RADIONUCLlDE RELEASE CHARAC CENTER FOR NUCLEAR WASTE REGULATORY ANALYSES TE RISTICS INTO CONT AINMENT UNDE R SEVERE ACCIDENT NUREG/CR-5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE.

CONDITIONS Fanal Report SEARCH AT CNWRA. Calendar Year 1991.

NUREGICP,5783. AGING ASSESSMENT OF THE COMBUST 60N ENGl.

NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE NEERING AND DABCOCK & WILCOX CONTROL ROD DRIVES RESEARCH AT CNWRA. January June 1992.

NUREG/CR 5844 AGING ASSESSVENT OF O! STABLES AND NUREG/CR 5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE SWITCHES IN NUCLE AR POWER PLANTS RESEARCH AT CNWRA. July Docember 1992.

NUREG/CR 5883. HEALTH RISK ASSESSMENi OF IRRADIATED NUREG/CR 5917 V01: SENSITIVITY AND UNCERTAINTV ANALYSES.

TOPAZ APPLIED TO ONE DIMENSIONAL RADIONUCLIDE TRANSPORT IN A l

l i

~ - -.

Contractor index 111 LAffRED FRACTURED ROCK MULTFRAC - Ana%C Solutions And EOE ENGINEERING CONSULTANTS (FORMERLY EOE ENGINEERING, Local Sensitivmes INC.)

NUREG/GR-5917 V02 SENSITIVITY AND UNCERTAINTY ANALYSES NUREG/CR-6011' REVIEW OF STRUCTURE DAMPING VALUES FOR APPUED TO ONE DIMENSIONAL RADIONUCLIDE TRANSPORT IN A ELASTIC SEISMIC ANALYSIS OF NUCLEAR POWER PLANTS.

LAYE RED F RACTURED ROCK Evaluation Of The Lmt State Ap-NUREG/CR 6012: STlFFNESS AND DAMPING PROPERTIES OF A pmach LOW ASPECT RATIO SHEAR WALL BUILDING BASED ON RECORD.

NUREG/GR %91. PORFLOW. A MULTIFLUID MULTIPHASE MODEL ED EARTHOUAKE RESPONSES.

FOR SiMULAT!NG FLOW, HEAT TRANSFER. AND MASS TRANS-NUREG/CR-6013 METHODS USED FOR THE TREATMENT OF NON-PORT IN FRACTURED POROUS MEDIA. User's Manual - Veruon PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

24L NURE G /CR -6021 A LITERATURE REVIEW OF COUPLEO THERMAL-EOE,1NC.

HYDROLOGIC MECHAN ICAL -CHEMICAL PROCESSES PERTINENT NUREG/CR-4832 V08 ANALYSIS OF THE LASALLE UNIT 2 NUCL EAR TO THE PHOPOSED HIGH LEVEL WASTE REPOSITORY AT YUCCA POWER PLANT: AISK METHODS INTEGRATION AND EVALUATION MOUN T AIN PROGR AM (RM!EP) Seismic Analysis NUHE G/CR 6026 THEORETICAL AND EXPERIMENTAL INVESTIGA-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY FRANCE SATURATED, FRACTURED POROUS MEDIUM NUREG/CR-6028: BiGFLOW A NUMERICAL CODE FOR SIMULATING NUREG/CR-6028 BIGFLOW. A NUMERICAL CODE FOR SIMULATING FLOW IN VARIABLY SATURATED. HETEROGENEOUS GEOLOGIC FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC MEDIA Theory And User's Manual - Version 1.1.

MEDIA Theory And User's Manual - Vers on 1.1 GESELLSCHAFT FUR REAKTORSICHERHEIT CITY COLLEGE OF NEW YORK, NEW YORK, NY NUREG/CR-5997. CSNI PROJECT FOR FRACTURE ANALYSES OF NUREG/GR-5956 CONSIDERATION OF UNCERTAINTIES IN SOIL-LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS STRUCTURE INTER ACTION COMPUTATIONS, (PROJECT FALSIRE)

COMMONWE ALTH EDISON CO-GRAM, INC.

NUR EG /CR-5642 LIGHT WATER REACTOR LOWER HEAD FAILURE NUREG/CR-6059. MACCS VERSION 1511.1. A MA:NTENANCE RE-ANALYSIS LEASE OF THE CODE.

COMPUTER SIMULATION & ANALYSIS, INC-HARVARD SCHOOL OF PUBLIC HEALTH, BOSTON, M A NUREG/CR 6035. FEAS:BILITY STUDY FOR IMPROVED STE ADY.

NUREGICP-0130 V01 PROCEEDINGS OF THE 22ND DOE /NRC NU-STA TE INITIALIZATION ALGORITHMS FOR THE RELAPS COMPUT-CLEAR AIR CLEANING CONFE RENCE. Sessions 1-8 Held in ER CODE Denver. Colorado. August 24-27,1992.

NUREG/CP-0130 V02. PROCEEDINGS OF THE 22ND DOE /NRC NU-NU

/

29 V95. F: ELD LYSIMETER INVESTIGATIONS LOW-De ver Co ado A us 2 27 92 LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR-4214 R2 PT1: HEALTH EFFECTS MODEL FOR NUCLEAR NU C

2 L G T W AT EACTOR LOWER HEAD F AILURE AN ALY SIS NUREG'CR 5672 V03 CHARACTERISTICS OF LOW-LEVEL RADICAC-IDAHO NATIONAL ENGINEERING LABORATORY TIVE DECONTAMl NATION WASTE. Annual Heport For Fiscal Year NUREG/CR 6061: DETERMINATION OF THE BIAS IN LOFT FUEL m2.

PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN NUHEG/CR-5759 HISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS PHASE OF LARGE. BREAK LOCA EXPERIMENTS STORAGE. SUPPLY, AND DISTRidVTION SYSTEMS IN PRESSUR-IIED WATER REACTOR PLANTS ILLINOIS, UNIV. OF. URB AN A, IL NUREG/CR 5818 UNCERTAINTY ANALYSIS OF M:NLMUM VESSEL NUREG/CR-5969: J AND CTOO ESTIMATION EQUATIONS FOR SHAL-LIQU:D INVENTORY CURING A SMALL-BREAK LOCA IN A BaW LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS.

PLANT.AN APPLICATION OF THE CSAU METHODOLOGY USING NUREG/CR-5970: APPROxlMATE TECHNIQUES FOR PREDICTING THE REL APS/ MOD 3 COMPUTER CODE.

SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC)

NUREG/CR-SM2' TR AC-B THERMAL-HVDRAULIC ANALYSIS OF THE NUREG/CR 5971: CONTINUUM AND MICROMECHANICS TREATMENT ULACK FOX BOILING WATER REACTOR OF CONSTRAINT IN FRACTURE NURE G/CR 5928 ISLOCA RESEARCH PROGRAM Final Report NUREG/CR 5977. A PERFORMANCE INDICATOR OF THE EFF ECTIVE.

NUREG/CR 5937. INT E NTIONAL DEPHESSUR12ATION ACCIDENT NESS OF HUMAN MACHINE INTERFACES FOR NUCLEAR POWER MANAGEMENT STRATEGY FOR PRESSURIZED WATER REA<

pgg TORS NUREG/CR 5949 ASSESSMENT OF THE POTENTIAL FOR HIGH IMPERIAL COLLEGE, LONDON. UK PRE SSURE MELT EJECTION RESULTING FROM A SURRY ST ATION NUREG/CR 5958.

TWO PARAMETER FRACTURE MECHANICS THEORY AND APPLICATIONS NL 95 ES OF HUMAN PERFORMANCE DURfNG OP.

WHALA COM RESEARCH WSWE NU f /

4 AP RC TECHNICAL REFERENCE NUREG/CR 4214 RtP2A2 HEALTH EFFECTS MODELS FOR NUCLE-MANUAL 1RRAS/ SARA VERSION 4 0 AR POWER PLANT ACCIDE NT CONSEQUENCE NUREG/CR 5976 DEVEL OPMENT AND USE OF A TRAIN-LEVEL ANALYSIS Mod 6 cation Of Models Resulting Fro.n Addition Of Effects PROGABILISTIC RISK ASSESSMENT.

NUREG/CR5987-MICROBIAL-INFLUENCED CEMENT DEGRADATION Of Exposuse To Alpha Emitting Radionuchdes Part II: Scient6c Bases llTE RATURE REVIEW.

For Heafth.-

NUREG/CR-6027 PREllMINARY EVALUATION OF SNUBBER SINGLE NUREGiCR-4214 R2 PT1. HEALTH EFFECTS MODEL FOR NUCLEAR F AILURES POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part 1:

NURE G 'CR4061 DETERMINATION OF THE BIAS IN LOFT FUEL introducton, integration And Summary.

PE AK CL ADDING TEMPERATURE DATA FROM THE BLOWDOWN PHASE OF LARGE-BREAK LOCA EXPER!MENTS INSTITUTE FOR MATERIALS RESEARCH NUREG/CR 6070' MODEllNG APPROACHES FOR CONCRETE BAR.

NUREG/CR 5926 SANS INVESTIGATION OF LOW ALLOY STEELS IN RIERS USED IN LOW-LEVEL W ASTE DISPOSAL NEUTRON IRRADIATED, ANNEALED, AND REIRRADIATED CONDI-NUREG/CR4073 LYSIMETER LITERATURE REVIEW TIONS.

ELECTRIC POWER RESEARCH INSTITUTE INSTITUTE OF NUCLEAR POWER OPERATIONS NUREG'CR 6018 SURVEY AND ASSESSMENT OF CONVENTIONAL NUR EG-1474 EFFECT OF HURRICANE ANDREW ON THE TURKEY SOFTW ARE VEROCATION AND VALIDATION METHODS.

POINT NUCLEAR GENERATING STATION FROM AUGUST 20-30, 1992 ENTROPIC SYSTEMS, INC-NUREG'C'R 6081 ENHANCED REMOVAL OF RADIOACTIVE PARTI-INTERIOR, DEPT OF, GEOLOGICAL SURVEY CLES BY F LUOROCARBON SL!RF ACTANT SOLUTIONS NUREG/CR-6085. UNITED STATES SE!SMOGRAPHIC NETWORK.

112 Contractor Index INTERNATIONAL ATOMIC ENERGY AGENCY MOR ATUW A, UNIV. OF, SRI LANKA NUREG/CP.0134 INTERNATIONAL ATOMIC ENERGY AGENCY SPE-NUREG/CR 5951: THE MANAGEMENT OF ATWS DY BORON INJEC-CIALISTS MEETING ON EXPERIENCE IN AGING, MAINTENANCE.

TION.

AND MODERNIZATION OF INSTRUMENTATON AND CONTROL SYSTEMS FOR IMPROVING NUCLEAR POWER PLAN T NATION AL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERLY AVAILADILITY Held At Rockville,MD.May 5 7,1993.

NATIONAL DUREAU OF NUREG/CR 4735 V08. EVALUATION AND COMPtLATION OF DOE IOW A ST ATE UNIV, AMES, IA WASTE PACKAGE TEST DAT A. Biannual Report. August 1989. Janu-NUREG/CR 4273-CRACK PROPAGATION IN HIGH S1HAIN REGIONS ary 1990 OF SLOUOYAH CONTAINMENT.

NUREG/CR 5957: SYSTEM BO + (TM) CONTAINMENT - STRUCTURAL N AVY, DEPT, OF DESIGN REVIEW NUREG/CR 5958.

TWO. PARAMETER FRACTURE MECHANICS:

THEORY AND APPLICATONS.

JBF ASSOCIATES,INC.

NUREG/CR 5909 J AND CTOD ESTIMATION EQUATIONS FOR SHAL-NUREG/CR 5471: ENHANCEMENTS TO DATA COLLECTION AND RE' LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS.

PORTING OF SINGLE AND MUL11PLE FAILURE EVE NTS-NUREG/CR 5970: APPROXIMATE TECHNIQUES FOR PREDICTING SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC).

JOHNS HOPKINS UNIV., BALTIMORE, MD NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE MA-NUREG/CR4052: METHODOtOGY FOR REllAutuTY BASED CONDl-CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL TION ASSESSMENT ApplK;ahOn IO Concrete StrV(tures in Nucicar STRUCTURAL ALLOYS.

Ptents.

NEW MEXICO STATE UNIV., LAS CRUCES, NM NUREG/CR 117 EUTRON SPECTRA AT DIFFEHENT HIGH F LUX STRT VE S S ES' ISOTOPE REACTOR ;HFIR) PRESSURE VESSEL SURVEILLANCE LOCATONS-NEW MEXICO, UNIV. OF, ALBUOUEROUE, NM KAISER ENGlHEERING (FORMERLY KAISER ENGlHEERS)

NUREG/CR 6007; STRESS ANALYSIS OF CLOSURE DOLTS F OR NU G/

76 DA IN IN LOW. ASPECT-RATIO, REINFORCED SHIPPING CASKS' CONCRETE SHEAR WALLS.

KORE A ADVANCED INSTITUTE OF SCIENCE AND TECHNOLOGY^ "

NUREG F 4832 V08 ANALYS S OF HE LASALLE UNIT 2 NUCLEAR AP TI TO EV EA DENTF N M NA POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP) Seismic Analysis.

KTECH CORP.

NUREGICR 5907. CORE-CONCHETE INTERACTIONS WITH OVERLY.

NU COR ING WATER POOLS.The WETCOR 1 Test.

PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

LAMONT-DOHERTY GEOLOGICAL OBSERVATORY NUREG/CR 5778 V03. NEW YORK /NEW JERSEY REG 6CNAL SEISMIC OAK RIDGE NATIONAL LABORATORY NETWORK. Final Report For Apnl 1985. September 1992.

NUREG/CP4131: PROCEEDINGS OF THE JOINT lAEA/CSNI SPECIAL.

LAWRENCE LIVERMORE NATIONAL LABORATORY ISTS' MEETING ON F RACTURE MECHANICS VERIFICATION DY NUREG/CR 4832 V08. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR LARGE SCALE TESTING. Held At Pollard Auditonum, Oak POWER PLANT: RISK METHODS INTEGRATON AND EVAL,UATION Ridge, Tennessee.

PROGRAM (RMIEP) Seismec Analysis NUhtG/CP 0134 INTERNATIONAL ATOMIC ENERGY AGENCY SPE-NUREG/CR 60.01: STHESS ANALYSIS OF CLOSURE BOLTS FOR CIALISTS MEETING ON EXPERIENCE IN AGING. MAINTENANCE, SHIPPING CASKS AND MODERNIZATION OF INSTRUMENTATION AND CONTROL NURFG/CR 6082 DATA COMMUNICATIONS I

SYSTEMS FOR IMPROVING NUCLEAR POWER PLANT NUREG/CR 6083: REVIEWING REAL-TIME PERFORMANCE OF NU-AVAILABillTY. Held At Rockville.MD,May 5 7,1993 CLE AR REACTOR SAFETY SYSTEMS NUREG/C44219 V09 N2-. HEAVY SECTION STEEL TECHNOLOGY NUREG/CR-6090. THE PROGRAMMABLE LOGIC CONTROLLER AND PROGRAM Semiannual Progress Report For April-September 1992.

ITS APPLICATION IN NUCL EAR REACTOR SYS1 EMS NUREG/C45247 V01 R1: RASCAL VERSION 2.0 USER S GUIDE.

NUREG/CR4101. SOFTWARE RELIADIUTY AND SAF ETY IN NUCLE

• NUREG/CR-5358: REVIEW OF ASME CODE CRITERIA FOR CONTROL AR REACTOR PROTECTION SYSTEMS OF PRIMARY LOADS ON NUCLEAR PIPING SYSTEM BRANCH CON-NECTONS AND RECOMMENDATIONS FOR ADDITIONAL DEVELOP.

LOS ALAMOS NATIONAL LABORATORY N

G/CR 57 TIFF NE OF LOW ASPECT RATIO, REINFORCED NLRLG/CR5d4 V02: AUXILIARY FEEDWATER SYSTEM AGING IN A

R R ED

$1

^

A(N NUF

/C 0

AVY SECTION STEEL IRRADIATION C NCRET EAR NUREG/CR 6060. HYDROGEN MIXING STUDIES (HMS) ASSESSMENT NURE / R 6 VO GI AND R CE ER F N OL ROD DRIVE MECHANISMS FOR BWR NUCLEAR PLANTS.

NUREG/CR-5754: BOILING-WATER REACTOR INTERNALS AGING LOUISlANA STATE UNIV BATON ROUGE, LA DEGRADATION STUDY. Phase 1 NUREG/CR6071 IMPACT OF ENOFIB-VI CROSS-SECTION DATA ON NUREG/CR 5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC H B ROBINSON CYCLE 9 OOS1 METRY CALCULATIONS FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE LOVELACE BIOMED & E NVIRONMENTAL RESEARCH INSTITUTE ROWE REACTOR PRESSURE VESSEL.

NUREG/CR-5914: CHEMICAL COMPOS! TION AND AT(NDT) DETERMI.

NUREG/CR-4214 R2 Pit: HEALTH EF FECTS MODEL FOR NUCLEAR NAflONS FOR MOLAND WELD WF 70 POWER PLANT ACCIDENT CONSEOUE NCE ANAL.YSIS Part 1-NUREG/CR-5922: MODULAR HIGH TEMPERATURE GAS-COOLED RE.

j introducton.Integrahon And Summary.

ACTOR SHORT TERM THERMAL RESPONSE TO FLOW ANO REAC-MARTIN MARIETTA ENERGY SYSTf MS, NC.

TlVITY TRANSIENTS.

NUREG/C46118: ASSESSMENT OF Li EFFECTIVENESS OF THE NUREG/CR-5938: NATONAL PROFILE ON COMMERCIALLY GENER-ATED LOW-LEVEL RADIOACTIVE mixed WASTE.

LEO REFORM RULE AND ITS IMPLEMENTATION NUREG/C45942: SEVERE ACCIDENT SOURCE TERM CHARACTER-MARYLAND, UNIV. OF, COLLEGE PARK, MD ISTICS FOR SELECTEO PEACH BOTTOM SEQUENCES PREDICTED NUREG/CR 5801: PROCEDURE FOR ANALYSIS OF COMMON.CAUSE BY TK i MELCOR CODE.

F AILURES IN PROBABlWSTIC SAF ETY ANALYS:S NUREG/C45944. A CHARACTER 12ATION OF CHECK VALVE DEGRA-DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-MATERIALS ENGINEEFllNG A6SOCIATES,INC, DUSTRY.

NUREG/CR 5926: SANS INVESilGATIOR OF LOW ALLOY STEELS IN NUREG/CR 5952. EVALUATION OF CRACK POP INS AND THE DE-

. NEUTRON 1RRADIATED, ANNEALED. AND REIRRADIATED CONDI-TERMINATION OF THEIR RELEVANCE TO DESIGN CONSIDER.

TIONS.

AT60NS

Contractor index 113 NUREG/CR 5955: MATERIALS AND DE$iGN BASES ISSUES IN ASME NUREG/CR 5907: CORE CONCRETE INTERACTIONS WITH OVERLY.

CODE CASE N 47 ING WATER POOLS.The WETCOR-1 Test NUREG/CR-5968. POTENTIAL CHANGE IN FLAW GEOMETPY OF AN NUREG/CR-5227 V01: EVALUATION OF A PERFORMANCE ASSESS-INITIALLY SHALLOW FINITE-LENGTH SURFACE FLAW DURING A MENT METHOOOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE PRESSURIZFD THERMAL. SHOCK TRANSIENT.

DISPOSAL FACILITIES Evaluation Of Modeling Approaches.

NUREG/CR 5972-EFFECTS OF NONSTANDARD HEAT TREATMENT NUREG/CR 5936: ENHANCEMENTS TO THE ACCIDENT PRECURSOR j

TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES METHODOLOGY, OF CARDON STEEL CASTING REPAIR WELDS NUREG/CR 5901: POSTTEST DESTRUCTIVE EXAMINATION OF THE NUREG/CR 5997: CSNI PRCMECT FOR FRACTURE ANALYSES OF STEEL LINER IN A 16-SCALE REACTOR CONTAINMENT MODEL LARGE. SCALE INTERNATIONAL REFERENCE EXPERIMENTS NUREG/CP $966: A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY

_(PROJECT FALSIRE)

CONTA. sENT SPRAYS NURE G/CR 6015. STRUCTURAL AGING PROGRAM TECHNICAL NUREG/CFb5976. SOURCE TERM ATTENUATION BY WATER IN THE PROGRESS FOR PERIOD JANUARY DECEMBER 1992 M^ K

^ E" E^ T-NUREG/CH 6023 GENERIC ANALYSIS FOR EVALUATION OF LOW NUR /

HE PRO 8AB Y O RK1 CONTAINMENT CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS U"E M

^K HE L E NST FRACTURE OF REACTOR PRESSURE VESSEL MATERI-

^fG/CR S9 SV NUREG/CR 6036 INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL LEASE OF THE CODE.

LOADING ON FRACTURE TOUGHNESS SCIENCE & ENGINEERING ASSOCIATES,INC.

NUREG/CR 6048 PRESSURIZED-WATE A REACTOR INTERNALS AGING DEGRADATION STUDY, Phasc i NUREG/CR 5791: HISK EVALUATION FOR A GENERAL ELECTRIC NUREG/CR 6052 METHODOLOGY FOR RELIABILITY BASED CONDi.

BWR EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON TION ASSESSMENT. Apphcahon To Concrete Structures in Nuclear SAFETY.RELATED EQUIPMENT. EvalualKn Of Gononc issue 57.

Plants NUREG/CR-6065. SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR SCIENCE APPLICATIONS INTERNATIONAL CORP,(FORMERLY NUREG/CR 6071: IMPACT OF ENDF/D-VI CROSS-SECTION DATA ON SCIENCE APPLICATIONS, H B. ROBINSON CYCL E 9 DOSIMETRY CALCULATtONS.

NUREG-0713 V12: OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR 6117: NEUTRON SPECTRA AT DIFFERENT HIGH FLUX MERCIAL NUCLEAR POWER REACTORS AND OTHER ISOTOPE REACTOR (HFIR) PRESSURE VESSEL SURVEILLANCE F ACILITIES,1990. Twenty-Third Annual Report.

LOCATIONS-NUREG-0713 V13. OCCUPATIONAL RADIATION EXPOSURE AT COM-AL N W AR NR WNS M M6 OCCLAHOM A, UNIV. OF, NORM AN. OK NUREG/CR 6034. OKLAHOMA SEISMIC NETWORK Final Report NURE - 3 14 AL AD AT N EXPOSURE AT COM-MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES OMNI TECH INTERN ATIONAL, LTD, NUREG/CR 6047: CONTINUOUS SPECTROSCOPIC ANALYSIS OF 1992. Twenty-Fifth Annual Report.

VANADOUS ANO VANADIC IONS _

NUREG/CR 5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR.

PHOENIX ASSOCIATES,INC.

IZED WATER REACTOR PLANTS NUREG/CR-524 7 V01 R1: RASCAL VERSION 2 0 USER'S GUIDE.

NUREG/CR 5863: RISK ASSESSMENT OF ISOLATION DEVICES IN SAFETY SYSTEMS.

PURDUE UNIV., WEST LAFAYETTE,1N NUREG/CR 5993 V01: METHODS FOR DEPENDENCY ESTIMATION NUREG/GR OOO9. STEPWISE INTEGRAL SCALING METHOD AND ITS AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE APPLICATION TO SEVERE ACCIDENT PHENOMENA DATA ST ATISTICS Summary Report NUREG/CR-5993 V02. METHODS FOR DEPENDENCY ESTIMATION AND SYSTEM UNAVAILABillTY EVALUATION BASED ON FAILURE NUF EG F TE BA TY F Af K-1 CONTAIN; TNT DATA STATISTICS. Detailed Desenptson And Apphcabons FAILURE BY MELT' ATTACK OF THE LINER NUREG/CR-6018. SURVEY AND ASSESSMENT OF CONVENTIONAL SOFTWARE VERIFICATION AND VALIDATION METHODS.

ROM LABORATORIES NUREG/CR 6113 CLASS 1E DIGITAL SYSTEMS STUDIES.

NUREG/CH-6050: RADIATION EXPCSURE MONITORING AND INFOR4 MATION TRANSMITTAL (REMIT) SYSTEM User's Manual.

RUSSIAN RESEARCH CENTER (KURCHATOV INSTITUTE)

NUREG/CR@60. HYDROGEN MIXING STUDfES (HMS) ASSESSMENT NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-MANUAL.

HAVIOR OF HYDROGEN AIR MIXTURES WITH TURBULENT JET IG-NITION AT LARGE SCALE.

Sv ENTECH, INC.

NUREG/CP-0129 PROCEE DINGS OF THE WORKSHOP ON PROGRAM SANDIA NATION AL LABORATORIES FOR ELIMINATION OF REQUIREMENTS MARGINAL TO SAFETY.

NUREGICR4832 V05: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR 5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-PROGRAM Parameter Estimation Analysis And Scir,ening Human Reh-IZED WATER REACTOR PLANTS.

atalq Analysis.

NUREb/CR4832 V09. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR SOHAR, INC.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION NUREGICR-6113. CLASS 1E DIGITAL SYSTEMS STUDIES PROGRAM (RMlf P) Internal Fire Analysis NUREG/CR 5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE SOUTHWEST RESEARCH INSTITUTE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NUREG/CR 6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-NUREG/CR 5305 V02 P2. INTEGRATEU RISK A%gendices A-C.

TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY Risk Uncertainty Evaluation Program LPRUEP) A ESSMENT FOR THE SATURATED, FRACTURED POROUS MEDIUM.

LASALLE UNIT 2 NUCLEAR POWER PLANT:Phenomenology And Risk Uncertainty Evaluation Pr m PRUEP) Appendices D-G ST. LOUIS UNIV., ST. LOUIS, MO ATA CblLECTION AND RE-NUREG/CR 6079. SEISMOLOGICAL INVESilGATION OF EARTH-F E

ENT TO QUAKES IN THE NEW MADRID SEISMIC ZONE Final PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

NUREG/CH 5791. RISK EVALUATION FOR A GENERAL ELECTRIC Repor1 September 1986 December 1992.

BWR, EFFECTS OF FIRE PROTECTION SYSTEM ACTUATION ON NUFE R 801 PR,

RE OR YS FC t AUSE NU EG/CR 545 VO b OPMENT OF THE NRC'S HUMAN PER-FAILURES IN PROBABilISTIC SAFUY ANALYSIS FORMANCE INVESTIGATION PROCESS (HPIP).

NUREG/CR 5843. CORCON MOD 3 AN INTEGRATED COMPUTE R NUREG/CR 5455 V02: DEVELOPMENT OF THE NRC'S HUMAN PER-MODEL FOR AN AL YSIS OF MOLTEN CORE CONCRETE FORMANCE lNVESTIGATION PROCESS (HPIP).

INTER ACTIONS user's Manual NUREG/CR 5455 V03: DEVELOPMENT OF THE NRC'S HUMAN PER.

NUREG/CR-5863 RISK ASSESSMENT OF ISOLATION DEVICES IN FORMANCE INVESTIGATION PROCESS (HPIP)

SAFETY SYSTEMS NUREG/CR 5901: A SIMPLIFIED MODEL OF AEROSOL SCRUDBING TECHNADYNE ENGINEERING CONSULTANTS,lNC, BY A WATE R POOL OVERLYING CORE DEDR!S INTERACTING NUREG/CR-6059 MACCS VERSION 1,511.1: A MAINTENANCE RE.

WITH CONCRETE rinal Report LEASE OF THE CODE.

9

=

114 Contractor index TENNE SSEE, UNIV. OF, KNOX VILLE, TN VIKING SYSTEMS INTERNATION AL NUREG/GR 0010- HYDRiD DIGITAL SIGNAL PROCE SSING AND NUREG/CR-5956 CONSIDERATION OF UNCERTAINTIES IN SOIL.

NE URAL NETWORKS FOR AUTOMATED DIAGNOSTICS USING NDE STRUCTURE INTERACTION COMPU TAItONS VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., SLACKSOURG, VA TEX AS A&M UNIV., COLLEGE ST ATION, TX NURE G/CR-6058 VIRG NIA REGIONAL SFgSMIC NETWORK. Foal NUREG/CR 5971. CONTINUUM AND MICROMECHANICS TRE ATMENT Rep rt (1986 1992)

OF CONSTRAINT IN FRACTURE.

NUREG/GR.0006 DEPOSITION SOFTWARE TO CALCULATE PARTI-WISCONSIN, UNIV. OF, M ADISON, WI Cl f PENE T RAT ION THROUGH AEROSOL TRANSPORT NUREG/CR.4214 RIP 2A2: HE ALTH EFFECTS MODELS FOR NUCLE.

SYSTEMS Fmal Report.

AR POWER PLAN T ACCIDENT CONSEQUENCE ANALYSIS Modificaten Of Models Resulting From Addition Of Effects Nt I. CFI )J CAVI ATI C IDF FOR CONTROL VALVES.

,r cal.

NUREG/CR-4214 R2 PTI: HEALTH EFFECTS MODEL FOR NUCLEAR U S. NAVAL ACADEMY, ANNAPOLIS, MD POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Part i NUREG/CR $981 THE EFFECT OF ELECTRIC DISCHARGE MA-Introduchon.Integraten.And Summary CHINEO NOTCHES ON THE FFIACTURE TOUGHNESS OF SEVERAL NUREG/CH.5642. LIGHT WATER Rt ACTOR LOWER HEAD FAILURE STRUCTURAL AtLOYS ANALYSIS.

International Organization index This index lists, in alphabetical order, the countries and performing or forming organization are the NUREG/lA numbers an l

untry and per-mation is needed, refer to the main citation by the NUREG/lA number orts. If further infor-FE DEHAL Rf PUDUC OF GERMANY

% VE NS AG kWU GROUP NUHEG/IA 0109. ASSESSMENT OF RELAPS/ MOD 2 AG NUnl uaA 0116 ASSE SSMENT OF HE L AP5/ MOD 1'V5MS AGAINST LOAD REJECTION TRANSIENT F ROM 75% STEADY STA THF UPTF fl ST NO 11 (COUNilHCURRE NT F TOW IN PWH HOT THE VANDEt L OS 11 NUCL EAR POWER PLANT, L E G)

NUREGilA-0110 ASSESSMENT OF ret.APS/ MOD 2 AGAINST A FINLAND DEtt OS Il NUCL EAR POWER PLANT. MAIN FEE TC CHNCAL HLSE ANCH CEN1RE OF FINL AND C N Al MAHAZ l Y ll NUHEG/lA 00W) ASSESSMENT OF HEL APS/ MOD 2 USING THE NUREGAA 0123 APPLICATION OF FULL POWER HLACK TEST DAI A OF HE WE T T ll REF LOODING EXPERIM( N T SGI/n C N ALMAHAZ WITH RELAPS/ MOD 2 JAPAN CONSEJO JO DE SEGURIDAD NUCtE AR JAPAN ATOMIC ENE RG / RESFADCH INSHTUTE NURE G/lA 0085. ASSESSMENT OF FULL POWER TURDIN UNIDAD ELECTRICA, 5 ASTART UP TEST FOR C. TRtLLO i W)TH HELAP5/ MO NUHL G'tA.012ti 20!3D FHOGHAM WOHk SUMMAHY REPOR1.

NUHLGnA 012/

2D/3D PHCLHAMREACTOH SAFETY ISSUES RESOLVED OY THE NURE G/lA 0120 ASSLSSMENT OF THE TURBINE THIP T UNION FLE CTRICA FENOSAIN COFRENTES NPP WITH TRAC-DF1, RFPUUt iC OF MORE A NUREGaA-0124 ASSESSMENT OF F OHE A ( LEClHC POWEH COHPOR ATION PRESSURIZER SPHAY RELAP5/ MOD 2 AGAINST A NUHEGilA 0092 ASSESSMt NT OF RELAPS/ MOD 2 COMPUILH TRANSIENT AND RECOVERY OY NATURAL C CODF AGAINS T THE NF T t.OAD 'iniP TES T DA TA FROM YONG.

JOSE CARRENA NUCt E An STATION GWANG UNII 2 UNtVERSITY OF CANTABalA NUHl GAA 0100 ASSESSMENT OF GCfL MODEL OF RE L AP5/

NUREGnA 0122 ASSESSMENT OF MSiv FULL MOD 3 AGAINST SIMPLE VE HIICAL TUDES AND ROD OUNDLE SANTA MARIA DE GARONA NUCLEAR POWER Pt. ANT TRAC-DFt (G1Jf)

NL 1 (A 0125. ASSE SSMEN T OF HE L AP5/ MOD 2 COMPUTfH 0#CUF COOf. AGAINST THE NATUHAL CtHCUt AllON TEST DATA I HOM N OtKi GW ANG UNIT 2 WE kONE A INNiiiUTE OF NUCLE AR SAf E TY NUREGnA-0094WWWWWWW ASSLSSME N T OF RE LAPS / MOD 3 AGAINST NUHL Gila 0095 THE ROYAL INSTITUTE' OF TECHNOLOGY'TW SD Cl-18 Hil APS ASSESSMENT USING LSU TEST DATA tolHLGnA004 HELAPS ASSE SSME N T USING SE MP3 CAL E THE NETHERLANDS Sul OCA TEST S NH~ t NE1HERLANDS ENERGY PESE ARCH FOUNDATION ECN NUHt G nA-O t03 ASS ( SSME NT Of DFTHSY TL ST 91 B USING NUHEG/lA OU91 ASSESSMENT OF HLl APS/ MOD 3 RELAP5/ MOD 2 AGAINST A NATUHAL CIRCULATION E XPE RIMENT IN NUCLEAR PO NUHt G AA -0!O4 cal f 50% f f ( O llNE BRE AK TE ST S F S-11HEL AP5rMOD3 ASSESSMENT USING THE SEM PL ANT DORSSELE.

NUHEGAA-0112 ASSESSMENT OF REL AP5/ MOD 2 AGAIN NUHF Gn A-0105 U9NG INADVFHIL NT SAF ETY IN I(CTION !NCIDE NT DATA OFASSEbSMEN HEFLOOD FXPEHiMENTS kOHI UNIT 1 PL ANy

{

UNITED KINGDOM SPAIN NAllONAL POWE R NUREG AA-0106 ASSESSMENT OF PWR STEAM GENER

{

X'lACION NUCLE AH ASCO MODE t UNG IN RE LAP 5/ MOD 2.

AUH( G 'IA-O t t )

ASSf SSML NT AND APPUCATION OF Dl ACKOUT NUREGnA Ot t3. PHEUMINARY MOD.kSif NTS AT ASCO NUCLEAH POWED PLANT WlHf RELAPS/

GENERATOR MODEl. LING IN RELAPS/ MOO 3 ASSESSMENT i

NL L NUHL GHA O t2t ASSE SSVf NT OF A PRESSURi?ER SPRAY VALVE NALYSIS OF LOFT TEST L5-1 USING HELAPSI H

8 F AUL1Y OPI N'NG TR ANSIE NT AT ASCO NUCLEAR POWE H WiNFF PL ANT Wif H HEL AP5/ MOD 2 TECHNOLOGY CENTR AKUACf0N NUCL E AH VANDEL L OS NUREG/tA-0096 NUMERICS AND IMPLEMENTATION OF THE UK NUREu q A 0107 ASSESSMt NT OF RL L APS/ MOD 2 AGAINST A MODEL INTO RELAP5/ MOD 3 ENTRAINMENT OFF TAKB HORIZONTAL STRATIFICATION L OAD HEJECTION FHOM 100%

Diit OS H NUCLEAR POWF h Pt ANT.TO 50% POWER IN THE VAN NUHf GHA 0i06 UNITED STATES ASELSSMENT OF HE LAP 5/ MOD 2 AGAiNS T A IDAHO NATIONAL ENGINEERING LABORATORY Cl F AH POWER PL ANTTUHFt:NL THip F HOM t00% POWEH IN THE VANDEL LOS 41 NU-NUREGAA Ot28 INTERNATIONAL CODE ASSESSMENT AN CATIONS PROGRAM

SUMMARY

OF CODE ASSESSMENT ST Il S CONCERNING RELAP5/ MOD 2, HEL AP5/ MOD 3 AND TRAC 0 115

1 1

n.

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 prece.ded by their Docket number I

and followed by the report number. If further information is nee' ed, refer to the main citation d

by the NUREG number.

% 334 Beavw Vahey Powa Statm Und 1, Duresne N'JREG/CR 5835

% 171 Peacn Bottom Atorrsc Pows Stabon, Und 1, NORFG/Cw5942 L@t Co PMademrua Electnc Co.

% 412 Ewave vaney Poww Statm Urvt 2,(bauesne NUREGICA 5835 50-277 Peach Dottom Atome Poww Statm Urvt 2, NUREG/CR 5942 Lt.t Co PNadema Electnc Co, Sf4%556 Diack Fon Statm UM 1, Put4c Sarwce of NUREG/CR 5882

% 278 Peach Bottom Atome Power Statm Urvt 3.

NUREG/CR-5942 Oktanoma Ruladelphia Electnc Co.

SIN %557 Glack Fox Statm Und 2, Pubic Sme of NUREG/CR 5882 S 266 Pont Beach Nuchar Plant, Urvt 1, Waconsm NUREG/CR-5898 Ok* ma Esectnc Power Co, B259 Bmwns Fern Nucles Poww Statm UM 1 NUREG/CR4022 S 301 Pont Beach Nuclear Rard, Unit 2. Wisconso NUREG/CR-5896 Tenessas valley Author' Electnc Poww Co.

W200 Drowns Fay Nuclear Power Sta'm UM 2.

NUREG/CR 6022 9 254 Quad Cees Staten, Orvt 1, Commonwearm NUREG/CR 5934 Tenrwseoe Vanay Autton Esson Co 50 296 Browns Fer7 Nuckier Poww Staton, Urst 3, NUREG/CR-6022

% 265 Quad.Caes Stanon. Und 2, Commonweam NUREG/CR-5934 Tenressee Vaney Authon Esson Co

% 446 Comanche Peak Steam Dectnc Staten, Urut 2, NUREG4797 S26 50 361 San Onofre NurJear Staton, Urvt 2, Southem NUREG/CR-5766 Texas UtAts:Electr Cah Edson Co

% 448 Comanche Peak Steam Electnc Statm und 2.

NUREG4797 S27

% 327 Sequoyah Nuclear Plant, Urvt 1 Tennessee NUREG/CR 4273 Texas utstes Dectr Valley Authonty 50 346 Daws Besse Nuclear Power Stahon. UM 1, NUREG/CH 5829 B326 Sequoyah Nuclear Rant, Urut 2 Tennessee NUREG/CR-4273 Toledo E$ son Co V

Authonty

% 237 Oreeden Nuclear Poww Statm Un;t 2.

NUREG/CR 5933 STN S498 South exas Protect, Unit 1, Houston Lghnng & NUREG/CR-5897 Commonwealth Edson Co Poww Co S 249 Oresden Nuclear Power Stauon, UM 3, NUREG/CR 5933 STN M499 South Texas Propsct, Urut 2, Houston Ughtmg & NUREG/CR-5897 Commonweam Eason Co g

% 321 Edwm He n Nuclear Rant, Und 1. Geonya NUREG/CR-6014 S 280 ower Stabon, Urut 1 Yrgnia Electnc & NUREG/CR 5949

% 366 Edom hatch Nuclear Pant, Und 2, Guer@a NUREG!CR 6014 ggg3 P

Stabon, Urst 2, Vrgria Electre & NUREG/CR-5949

% 341 Enrco Fome Atome Power Plant, Und 2 Detrort NUREG/CR-5950

$2 002 Sys 80 Standarted Nuclear Power Plant NUREGrCR 5907 404 989 E oc e of Utait Inc., Salt Lake C4. UT NUREG-1476 s., Combuston Engmee

% 289 Three Mie island Nuclear Station, Und 1, NUREG 1485 404989 Envrocare of Utah, Inc., Saft Lake Citv. UT,'

NUREG-1476 DRFT S 285 F

no Statm Und 1, Omaha futAc NUREG/CR 5834 Generd ute stand Nuclear Slaton, Und 1, NUREG/CR 5488

% 261 H

ot m Plant, UM 2, Carchna Pows & NUREG/CR 5833 Pubic

% 261 HB R Plant Urut 2, Carcima Power & NUREG/CR 6071 Co 50 374 LaSa% County Stanon, Urut 2, Commonweafth NUREG/CR 4832 V05 Ught Co.

Em Co

% 390 Watts Bar Nuclear Plant, Und 1, Tennessee NUREG4847 $11 M 374 LaSano County Stabon, Urst 2, Commonweam NUREG/CR4832 v08 Vaney Authonty Esci Co 50 390 Ws'ts Bu Nuclear Plant, Unit 1 Tennessee NUREG46d7 S12

% 374 LaSane County Staton, UM 2. Commonweam NUREG/CR4832 V09 Valley Authanty Esson Co

% 391 Watts Bar Nuclear Rant, Und 2, Tennessee NUREG-0847 S11

% 374 LaSane County Statm UM 2. Commonweam NUREG/CR 5305 V02 P1 Valley Authonty Edson Co

% 391 Watts Bar Nuclear Rant, Und 2, Tennessee NUREG4847 St2 W374 ls5aile County Staten, UM 2. Commonwealth NUHEG/CR 5305 V02 F2 Vaney Authonty Edson Co.

% 29 Yankee Rows Nucker Power Staban, Yankee NUREG/CR-5782 70 3070 Loumana Energy Sorwces. Wasangton. DC, NUREG 1484 DRFT Atome Electre Co.

STN S$20 Palo verde Nuclear Stabon, Urdt 1, Anzona NUREG/CitS836 W295 Zon Nuc4ar Power Station, Urvt 1, NUREG/CR-4551 V7 RIP 1 PutAc Serwce Co Commonween Edson Co.

STN %529 Palo verde Nuciear Sta'.on, U'ut 2, Anzona NUREG/CR 5836

% 295 Zen Nuclear Power Staton, Und 1, NUREG/CR 455tV7R1P2A Pd,hc Sarwce Co Commonweam Edson Co.

STN %530 Pao Verde Nockw Statm Urvt 3, Arena NUREG/CR 5836

% 295 Zon Nuclear Power Staton, Urvt 1, NUREG/CR4551V7R1928 PWc Sece Co.

Commonweam Edson Co l

l l

117

N l

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t5 J

t#4C FORM 335 U.S. NUCLEAR REGULATORY COMMISSION

1. REPORT NUMBER (2-89)

(Assigned by NRC, Add Vol.,

NRCM 1102, Supp., Rav,, and Addendurn Num.

3 m,3202 BIBLIOGRAPHIC DATA SHEET

• ' * - d *avd (S instructkins on ine rev.rs.)

NUREG-0304 V 1,18, No. 4 a nTLe ANo suuTm.E

3. DATE REPORT PUBLISHED Regulatory and Technical Reports (Abstract Index Journal)

MONTH YEAR Annual Compilation for 1993 March 1994

4. FIN OR GRANT NUMBER 6,AufMORib)

6. TYPE OF REPORT Reference

7. PERIOD COVERED pnclusive Dates)

8. Pf. HF ORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Divtston, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address: if contractor, provide name and malling address.)

Division of Freedom of Information and Publications Services Office of Adrninistration U.S. Nuclear Regulatory Commission Washington, DC 20555

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above"; if contractor, provide NRC Division Office or Region, U.S. Nuclear Regulatory Commission, arx! malling address. )

Same as 8, atmve.

10 SUPPLEMENT ARY NOTES

11. ABSTRACT (200 words or tees)

This journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors, proceed-ings of conferences and workshops, grants, and international agreement reports The entries in this compilation are indexed for access by title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agreements, contractor, international organization, and licensed facility,

12. KEY WORDS/DEEORtPTORS (Ust words or phrases that will assist researchers in locateng the report.)

13. AVAILABILITY STATEMENT Unlimited

14. SECURITY CLASSIFICATION compilation

,, p,g abstract mdex Unclassified (This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE NAC FORM 33s (2-89)

i n

Printed on recycled paper 1

Federal Recycling. Program i

z Co 3

h Main Citations R

Z*

and Abstracts j

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@CM h

V S@o<$,

hb E

Se ndarY Report 2

g Number Index inM

@M E

E E

"m E!

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Personal Author index Z

L T" AMC, t

~~n dC Od$

4 Subject index ji 2S?;n d A"f?s a

cp >cg s' 2 "

NRC Originating Organization r"

5 index (Star, Report )

l:

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NRC Originating Organization b.

index (International Agreements)

OS o

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NRC Contractor 7

Sponsor index

]

Contractor index l

^

International Organization 9,

gj index m I?

n o,

6&j!

Licensed Facility E!j!

$j Index eg i

l I