ML20058L420

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Third Quarter 1993,July-September

ML20058L420
Person / Time
Issue date:	11/30/1993
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V18-N03, NUREG-304, NUREG-304-V18-N3, NUDOCS 9312160335
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{{#Wiki_filter:i ) NUREG-0304 Vol.18, No. 3 Regulatory anc Technical Reports < Abstract Index Journall ( J t i Compilation for Third Quarter 1993 July - September U.S. Nuclear Regulatory Commission Office of Administration 1 4p Ih.[y#! .c..... 0304 R

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i ? .f.[ i i. Available frem r Superintendent of Documents U.S. Govemment Printing Offee Post Office Box 37082 - Washington, D.C. 20013-7082 -{ A year's subscription consists of 4 issues for this publication. Single copies of this publication - I are available from National Techn' cal-i information Service, Springfield, VA 22161 l i i i a p i 1 .l a 1 I i i I i i l t i I

i NUREG-0304 Vol.18, No. 3 1 i t = Regulatory and Technical Reports i (Abstract Index Journal) i 1 1 l Compilation for l Third Quarter 1993 i l July - September i l l l Date Published: November 1993 l i i i l .t Regulatory Publications Branch Division of Freedom ofInformation and Publications Services l Office of Administration i l U.S. Nuclear Regulatory Commission i W shington, DC 20555 .r s,,, l i a i.

) i l l CONTENTS Preface. ...v index Tab Main Citations and Abstracts 1

• Staff Reports l
• Conference Proceedings
• Contractor Reports

= international Agreement Reports .... 2 I Secondary Report Number index. 3 Personal Author index.. Subject index...... 4 NRC Originating Organization index (Staff Reports). 5 NRC Originating Organization index (International Agreements)... ....... 6 NRC Contract Sponsor index (Contractor Reports). 7 i Contractor index. 8 International Organization Index.. 9 Licensed Facility index. .... 10 1 l l iii

i l PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and tecnnical reports issued by the U.S. Naclear 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 f and Publications Services P-223 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i l 1 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXh, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indexes: Secondary Report Number index Personal Author index Subject index NRC Originating Organization Indax (Staff Reports) NRC Originating Organization Index (international Agreements) NRC Contract Sponsor index (Contractor Reports) l Contractor Index international Organization index l Licensed Facility Index ( A detailed explanation of the entries precedes each iridex. The bibliographic elements of the main citations are the following: l Staff Report NUREG-0808: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048 09570:200. Where the entries are (1) aport 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 l 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. l Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled l the proceedim e, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Controi System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use). Contractor Report NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R. Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242. Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use). l V

t i i PREFACE i i This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to: i i Technical Publications Section - l Regulatory Publications Branch Division of Freedom of Information and Publications Services l P-223 i U.S. Nuclear Regulatory Commission Washington, D.C. 20555 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, l NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indexes: 3 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) i Contractor index International Organization index l Licensed Facility index A detailed explanation of the entries precedes each iridex. The bibliographic elements of the main citations are the following: Staff Report NUREG-0808: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048 09570:200. Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession nurnber, (8) the microfiche address (for internal NRC use). Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINF9lNG IN NUCLEAR REGULATION. JANERP, J.S. Argonne National i i Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070. I Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled I 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 intemal use). i Contractor Report I NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER i REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R. Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242. Where the entries are (1) report number, (2) report title,13) report authors, (4) organizational unit of l authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use). v

i l International Aareement Report NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA. NEUMANN, U. Kraftwerk Union. August 1986. 223 pp. 8608270424. 37659:138. Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report veas published, (6) number of pages in the report, (7) the NRC Document Control System sccession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC internal use). The following abbreviations are used to identify the document status of a report: ADD - addendum APP - appendix DRFT - draft ERR - errata N - number R - revision S - supplement V - volume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the National Technical information Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the following address: Superintendent of Documents e U.S. Government Printing Office Post Office Box 37082 Washington, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275-2171. Non-U.S. customers must make payment in advance either by international Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents. ] NRC Report Codes 1 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 I identification replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-1 NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reporte,d. in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/lA is used for international agreement reports. All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Publications Services. vi

Main Citotions 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 report. The bibliographic inforrnation (see Preface for details) is followed by a brief abstract of this report. I l l NUREG-0040 V17 NO2: LICENSEE CONTRACTOR AND NUREG-0525 V02 RO1: SAFEGUARDS

SUMMARY

EVENT LIST VENDOR INSPECTION STATUS REPORT. Quarterly (SSEL). January 1, 1990 Through December 31, 1992. Report.Apni-June 1993 FAT.ite Dook)

• Division of Reactor in-FADDEN,M.; YARDUMIAN.J. Operations Branch. July 1993 spection & Lcensee Performance (Post 921004). August 1993.

250pp. 9308160105. 76142:001. 109pp. 9309210034. 76483:049. The Safeguards Summary Event List provides bnet summa-This penodical covers the results of inspections pe formed by nes of hundreds of safeguards-related events involving nuclear tne NRC's Vendr,r inspecton Branch that have been distributed matenal or facilities regulated by the U.S. Nuclear Regulatory to the inspected organizations during the penod from Apnl Commission. Events are desenbed under the categories: Bomb-through June 1993. related, intrusion, Missing /Allegedty Stolen, Transportation-relat-ed Tampenng/ Vandalism, Arson, Firearms related, Radiological NUREG-0090 V ;6 N01: REPORT TO CONGRESS ON ABNOR-Sabotage, Non-radiological Sabotage, and Miscellaneous. Be-MAL OCCUPNENCES. January-March 1993.

• Office for Analy-cause of the public interest, the Miscellaneous category also in-sis & Evalueson of Operatonal Data. Director. June 1993.30pp.

cludes events reported involving source matenal, byproduct ma-9307220177.75743:332. tenal, and natural uranium, which are exempt from safeguar<1s Section 208 of the Energy Roorgaruzation Act of 1974 ident,. requirements. Information in the event desenptions was ob-fies an abnorma. occurrence as an unscheduied incident or tained from official NRC sources. event that the Nuc sar Regulatory Commission determines to be NUREG-0540 V15 N05: TITLE LIST OF DOCUMENTS MADE significant from tht standpoint of public health and safety and PUBLICLY AVAILABLE.May 1-31, 1993.

• Divison of Freedom requires a quarterly report of such events to be made to Con of informaton & Publications Services (Post 890205). July 1993.

gress. This report covers the penod January through March 350pp. 9308160099. 76115:010. 1993. There is one abnormal occurrence ret a nuclear power This document is a monthly publicaton containing descnp-plant discussed in this report that involved a steam generator tions of toformation riceived and generated by the U.S Nuclear tube rupture at Palo Verde Unit 2, and none for fuel cycle facin-Regulatory Commisson (NRC). This information includes (1) ties. Three abnormal occurrences involving medical misadmints-docketed matenal associated with crvilian nuclear power plants tratons (two therapeutic and one d, agnostic) at NRC-licensed and othe uses of radioactive matenals, and (2) nondocketed facilities are also discussed in this report. No abnormal occur-material received and generated by NRC pertinent to its role as rences were reported by NRC's Agreement States. The report a regulatory agency. The following indexes are included: Per-also contains iniurmuisun updating previously reported abnormal sonal Author, Corporate Source, Report Nurnber, and Cross occurrences. Reference of Enclosures to Pnneipal Documents. NUREG-0304 V18 NO2: REGULATORY AND TECHNICAL RE-NUREG-0540 V15 N06: TITLE LIST OF DOCUMENTS MADE PORTS (ABSTRACT INDEX JOURNAL). Compilation For PUBLICLY AVAILABLE. June 1-30, 1993.* Division of Freedom Second Quarter 1993. April-June.

• Divison of Freedom of infor-of information & Publications Services (Post 890205). August mation & Publicatons Services (Post 890205). August 1993.

1993. 421pp. 9309030160. 76325:158. l 48pp. 9309210040. 764B3:158. See NUREG-0540,V15.N05 abstract. This Joumal includes all formal reports in the NUREG senes NUREG-0540 V15 N07: TITLE LIST OF DOCUMENTS MADE prepared by the NRC staff and contractors; proceedings of con-PUBLICLY AVAILABLE. July 1-31, 1993.

• Drvision of Freedom ferences and workshops; as well as international agreement re-of Information & Pub!scations Services (Post 890205). Septem-ports. The entries in this Compilaton are indexed for access by ber '993. 350pp. 9309210229. 76485:077.

t!tle and abstract, secondary report numver, personal author' Se6 NUREG-0540,V15,N05 abstract. subject. NRC organizaton for staff and intemational agree-ments, contractor, internatonal organization, and licensed facili-NUREG-0713 V13: OCCUPATIONAL RADIATION EXPOSURE AT ty. COMMERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES.1991. Twenty-Fourth Annual Report RADDATZ,C.T. NUREG-0386 DOS R07: UNITED STATES NUCLEAR REGULA Division of Regulatory Applications (Post 870413). TORY COMMISSION STAFF PRACTICE AND PROCEDURE HAGEMEYER.D. Science Applications Internatonal Corp. (for-DIGEST. Commission, Appeal Board And Licensing Board rr.erly Science Applications, Inc.). July 1993. 300pp. Decisions. July 1972 - September 1992.

• Office of the General 9308160144. 76114:077.

Counsel (Post 860701). August 1993. 600pp. 9309090041. This report summarizes the occupatonal radiation exposure 76379:138. information that has been reported to the NRC's Radiation Ex-This 7th revision of the sixth edition of the NRC Practice and posure information Reporting System (REIRS) by nuclear power Procedure Digest contains a digest of a number of Commission, facilities and certain other categones of NRC licensees during Atomic Safety and Licensing Appeal Board, and Atomic Safety the years 1969 through 1991. The bulk of the data presented in and Licensing Board decisions issued dunng the penod of July the report was obtained from annual radiaton exposure reports 1,1972 to September 30,1992, rnterpreting the NRC's Rules of submitted in accordance with the requirements of 10CFR20.407 Practice in 10 CFR Part 2. and the techniera specifications of nuclear power plants. Data 1

i 2 Main Citations and Abstracts ) on workers terminating their employment at certain NRC ti-The NRC Regulatory Agenda is a compilation of all rutes on censed faciltties were obtained from reports submitted pursuant which the NRC has recently completed action, or has proposed to 10CFR20.408. The 1991 annual reports submitted by about acton, or is considenng action, and all petitions for rulemak.ing 436 bcensees andcated that approximately 206,732 individuais which have been received by the Commission and are pending were monitored. 182,334 of whom were monitored by nuclear disposition by the Commissson. The Regulatory Agenda is up-power facihties. They incurred an average individual dose of dated and issued each quarter. 0.15 rem (cSv) and an average measurable dose of about 0 31 (cSv). Termination radiation exposure reports were analyzed to NUREG-0940 V12 N02: ENFORCEMENT ACTIONS: SIGNIFI-reveal that about 96.231 individuals completed their employ-CANT ACTIONS RESOLVED. Quarterly Progress Report,Apnt-ment w:th one or more of the 436 covered licensees danng June 1993.

• Ofc of Enforcement (Post 870413). September 1991. Some 68,115 of these individuals terminated from power 1993. 395pp. 9310120039. 76737:001.

reactor facilites, and about 7,763 of them were considered to This compilation summanzes signifcant enforcement actions be transent workers who received an average dose of 0.52 rem that have been resolved dunng one quarterly penod (April - (cSv) June 1993) and includes copies of letters, Notices, and Orders NUREG-0750 V37101: INDEXES TO NUCLEAR REGULATORY sent by the Nuclear Regulatory Commisson to beensees with COMMISSION ISSUANCES. January-March 1993.

• Drvision of respect to these enforcement actions. It is anticipated that the Freedom of informanon & Pubications Servces (Post 890205).

information in this pubication will be widely disseminated to July 1993. 55pp. 9308560096. 76122.200. managers and employees engaged in activrtes licensed by the Digests and indexes for issuances of the Commission, the NRC, so that actons can be taken to improve safety by avoid-Atomic Safety and Licensing Board Panel, the Administrative ing future violations similar to those desenbed in this publsca-Law Judges, the Directors' Decisions, and the Denials of Peti-tion. tions for Rulemaking are presented. NUREG-1145 V09: U.S. NUCLEAR REGULATORY COMMISSION NUREG-0750 V37 N04: NUCLEAR REGULATORY COMMISSION 1992 ANNUAL REPORT,

• Office of Administraton (Post ISSUANCES FOR APRIL 1993.Pages 251-354.
• Division of 890205). July 1993. 287pp. 9309090029. 76391:001.

Freedom of information & Publications Services (Post 890205). This report covers the major activites, events, decisions and July 1993.110pp. 9308160091. 76119:007. planning that took place dunng fiscal year 1992 within the U.S. Legalissuances of the Commisson, the Atomic Safety and Li-Nuclear Regulatory Commission (NRC) or involving the NRC. censing Board Panel, the Administrative Law Judges, and NRC Program Offces are presented. NUREG-1214 R12: HISTORICAL DATA

SUMMARY

OF THE SYS-NUREG-0750 V37 N05: NUCLEAR REGULATORY COMMISSION TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE. ISSUANCES FOR MAY 1993.Pages 355-418.

• Division of ALLENSPACH F. Divison of Reactor inspecten & Licensee Freedom of information & Publications Services (Post 890205).

Performance (Post 921004). August 1993.134pp. 9309210018. August 1993. 69pp. 9308190001. 76152:038. 76500:101. See NUREG-0750,V37,N04 abstract. The Histoncal Data Summary of the Systematic Assessment NUREG-0750 V37 N06: NUCLEAR REGULATORY COMMISSION of Licensee Periormance (SALP)is produced penodically by the ISSUANCES FOR JUNE 1993.Pages 419-515.

• Divison of U.S. Nuclear Regulatory Commisson. This summary provides i

Freedom of Information & Pubications Services (Post 890205). the resuits of the assessment for each facility by NRC region August 1993.103pp. 9309210012. 76481:034. and is further divided into the following sections: Section 1 pre-See NUREG-0750,V37,N04 abstract. sents the most recent SALP report ratings for facilities in oper-ation and under constructon; Section 2 presents a chronologi-NUREG-0750 V38 N01: NUCLEAR REGULATORY COMMISSION ca! listing of all SALP report ratings for each operating facility; ISSUANCES FOR JULY 1993.Pages 1-24.

• Division of Free-Section 3 presents a chronological listing of all SALP report rat-dom of information & Publicatons Services (Post 890205). Sep-ings for each facility under construction. For histoncal purposes, tember 1993. 32pp. 9310120315. 76741:114-past construction ratings for facilit,es that recently have been li-See NUREG-0750,V37,N04 abstract.

censed also are listed in Secton 3. NUREG-0837 V13 NO2: NRC TLD DIRECT RAD!ATION MONI-TORING NETWORK. Progress Report. April-June 1993_ NUREG-1272 V07 N01: OFFICE FOR ANALYSIS AND EVALUA-STRUCKMEYER.R.; MCNAMARA,N. Region 1 (Post 820201). TION OF OPERATIONAL DATA.1992 Annual Report - Power August 1993. 250pp. 9309210028. 76482:103. Reactors.

• Offce for Analysis & Evaluation of Operatonal i

This report provides the status and results of the NRC Ther. Data, Director. July 1993. 300pp. 9309210044. 76483:206. l moluminescent Dosimeter (TLD) Direct Radiation Monitonng The annual report of the U.S. Nuclear Regulatory Commis-1 Network. ft presents the radiaton levels measured in the vcinity sion's Offee for Analysis and Evaluaton of Operatonal Data of NRC Icensed facihties throughout the country for the second (AEOD) is devoted to the actrvities performed dunng 1992. The quarter of 1993. report is pubhshed in two separate parts. NUREG 1272, Vol. 7, NUREG-0910 R02 S01: NRC COMPREHENSIVE RECORDS DIS. No.1, covers power reactors and presents an overview of the POSITION SCHEDULE.

• Division of informaton Support Serv-perating expenence of the nuclear power industry from the NRC perspective, including comments about the trends of some ices (Post 890205). 5 % w er 1993. 89pp. 9310120062.

76738 121 key perf rmance measures. The report also encludes the pnnei-pal find:ngs and issues identified in AEOD studies over the past The approved r.c 2 4m on schedules speedy the ap-propnate duratior, 01 resnth $ and the final disposition for year ad swnmams inWmahm kun sd somes as kense event reports, diagnostic evaluatons, and reports to the NRC's records created oi wintsne by the NRC. NUREG-0910, Revi-aWs Cmtet h repyts cone a dswssa N N M sion 2. Suppleme 1 m.ies editonal and administrative a a a s 86NWM changes to the No. Tched se and forwarc's 3 sets of changes in n eam a Augmented Inspecton Team reports tw to the National Arce w and Records Administratons's General Record Schedule. hat g7 of Iceses. MEM272, M 7, Na 2, cows noi setors and presents a review of the events and concerns NUREG-0936 V12 N02: NRC REGULATORY AGENDA Ouarterly dunng ?992 associated with the use of Icensed material in non-Report.Apnl-June 1993.

• Division of Freedom of informaton &

reactor applications, such as personnel overexposures and Pubicatons Servees (Post 890205). July 1993. 138pp. medical misadministrations. Each volume contains a hst of the 9308160147. 76121 001. AEOD reports issued for 1984-1992.

Main Citations and Abstracts 3 i NUREG-1363 V05: ATOMIC SAFETY AND LICENSING BOARD discussed, as well as potential changes in NRC programs. A l PANEL ANNUAL REPORT. Fiscal Year 1992. CO1TER,B.P. draft report was issued for comment en February 1992. This Atomic Safety & Licensing Board Panel. September 1993.43pp. report is the final verson and includes the responses to the 9310120269. 76761:001. comments along with the staff regulatory analysis of potential in Fiscal Year 1992, the Atomic Safety and Licensing Board new requirements. Panel ("the Pane!") handled 38 proceedings. The cases ad-dressed issues in the construction, operahon, and maintenance NUREG-1453: REGULATORY ANALYSIS FOR THE RESOLU-of commercial nuclear power reactors and other activities re. TlON OF GENERIC ISSUE 142: LEAKAGE THROUGH ELEC-quinng a license from the Nuclear Regulatory Commission. This TRICAL ISOLATORS IN INSTRUMENTATION CIRCUITS. report sets out the Paners caseload during the year and sum-ROURK,C.J. Division of Safety issue Resolution (Post B80717). manzes, highlights, and analyzes how the wide-ranging issues September 1993. 21pp. 9310130042. 76744266. raised in those proceedings were addressed by the Paners Generic issue (Gl) 142 deals with staff concems about the judges and bcensing boards. design of isolation devices used to ensure separation between Class 1E and non-Class 1E electncal control and instrumenta-NUREG-1400: AIR SAMPLING IN THE WORKPLACE. Final tion circuits. This issue was initiated in June 1987. Staff reviews Report. HICKEY.E.E.; STOETZELG.A.; STROM.D.J.; et al. Bat-of the implementaton of the Safety Parameter Display System telle Memonal Institute. Pacific Northwest Laboratory. Septem-(SPDS) requirement indicated that some isolation devces used ber 1993.104pp. 9310123325. 76742:181. to provide an interface between the non-Class 1E SPDS and This report provides technical informaton on air sampling that the Class 1E safety systems would allow signal leakage if elec-sll be useful for facihties following the recommendations an the incally challenged. It was unknown if the amount of leakage NRC's Regulatory Guide 8.25, Revision 1, " Air Sampling in the posed a hazard to safe operabon of the Class 1E system. A Workplace." That guide addresses air sampling to meet the re-review of failure records does not reveal any incidents of quirements in NRC's regulations on radiaton protection, system damage caused by isolation device challenge. Further-10CFR20. This report desenbes how to determine the need for more, a review of existing PRA data indicates that the safety air sampling based on the amount of matenal in process modi-significance of ID challenge is low, at the expected challenge fied by the type of matenal, release potential, and confinement event frequency. However, based upon the potential design of the matenal. The purposes of air sampling and how the pur-variations in future control systems resulting from application of poses affect the types of air samphng provided are discussed-computer technology, additional design and quahfcaton test re-The report discusses how to locate ar samplers to accurately quirements for future plants are recommended. determine the concentratons of iarbome radcactive matenals that workers will be exposed to. The need for and the methods NUREG-1461: REGULATORY ANALYSIS FOR THE RESOLU-of performing airflow pattern studies to improve the accuracy of TION OF GENERIC ISSUE 153: LOSS OF ESSENTIAL SERV-air samphng results are included. The report presents and give ICE WATER IN LWRS. SU.T.-M. Division of Safety issue Reso-examples of several techniques that can be used to evaluate lution (Post B80717). August 1993. 32pp. 9309030241. whether the arbome concentratons of material are representa-76328:181. tive of the er inhaled by workers. Methods to adjust derived air in this report, the staff of the U.S. Nuclear Regulatory Com-concentratons for particle size are described. Methods to cali-mission provides a regulatory analysis for the proposed resolu-brate for volume of ar sampled and est mate the uncertainty in ton of Generic issue 153 (GI-153), " Loss of Essential Service l the volume of ar sampled are described. Statistical tests for de-Water in LWRs." GI-153 deals with the concems pertaining to termining minimum detectable concentrations are presented. the reliability of essential service water (ESW) system and relat-How to perform an annual evaluaton of the adequacy of the ar ed problems for all bght water reactors except the seven multi-samphng is also discussed. unit sites addressed by GI-130 "Essental Service Water Pump NUREG-1423 V04: A COMPILATION OF REPORTS OF THE AD. failures at Multi-Unit Sites." On the basis of the technical find-VISORY COMMITTEE ON NUCLEAR WASTE. July 1992 - June ings of a scoping study for GI-153, the staff recommends that 1993.

• Advisory Committee on Nuclear Waste. August 1993.

the instghts gained from the study serve as a complement to 81pp. 9309210031. 76482:337. the on-going ESW performance inspecton orogram. The staff This compilation contains 17 reports issued by the Advisory also concludes that ESW system rehabikty is being addressed Committee on Nuclear Waste (ACNW) dunng the fifth year of its by vanous on-going regulatory programs. Therefore, the staff operabon. The reports were submitted to the Chairman and recommends that GI 153 should be considered " RESOLVED." l Commissioners of the U.S. Nuclear Regulatory Commissinn, the The need for future acton(s) on ESW reliability is expected to Executive Director for Operations, the Drector Office of Nucle-be determined from these on-going programs. or Material Safety and Safeguards, or to the Director, Division of NUREG-1463: REGULATORY ANALYSIS FOR THE RESOLU-High Level Waste Management, Office of Nuclear Matenal TION OF GENERIC SAFETY lSSUE 105: INTERFACING Safety and Safeguards. All reports prepared by the Committee SYSTEM LOSS-OF-COOLANT ACCIDENT IN LIGHT-WATER have been made urallable to the public through the NRC Public Document Room and the U.S. Library of Congress. REACTORS.

• Division of Safety issue Resolution (Post 880717). July 1993. 77pp. 9308160153. 76114:001.

NUREG-1449: SHUTDOWN AND LOW-POWER OPERATION AT An interfacing systems loss of coolant accident (ISLOCA) in-NUCLEAR POWER PLANTS IN THE UNITED STATES. Final volves failure or improper operaton of pressure isolation vatves Report.

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

(PlVs) that compose the boundary between the reactor coolant September 1993. 200pp. 9310130052. 76743:001. system and low-pressure rated systems. Some ISLOCAs can The report contains the results of the NRC staffs evaluabon bypass containment and result in drect release of fission prod-of shutdown and low-power operatons at U.S. commercial nu-ucts to the environment. A cost / benefit evaluaton, taing three clear poww plants. The report describes studies conducted by PWR analyses, calculated the benefit of two potential modifica-the staff in the following areas: operating experience related to tions to the plants. Altematrve I is improved plant operations to shutdown and low-power operations, probabihshc risk assess-optimize the operator's performance and reduce human error ment of shutdown and low-power conditions and utihty pro-probabihties. Attemative U adds pressure sensing dev' es, ca-c grams for planning and conduchng actrvities during perods the bhng, and instrumentabon between two PlVs to provide opera-plant is shut down. The report also documents evaluatons of a tors with continuous monitoring of the first PlV. These two alter-number of techncal issues regarding shutdown and low-power nattves were evaluated for the base case plants (Case 1) and operatens performed by the staff, including the pnncipal find-for each plant, assuming the plants had a particular auxiliary ings and conclusions. Potential new regulatory requirements are building design in which severe flooding would be a problem if 1

4 Main Citations and Abstracts an ISLOCA occurred. The auxihary building design (Case 2) was when they receive money; and revsewed the method used by selected from a survey that revealed a number of designs with the NRC to determine the cost of its vanous programs. features tnat provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The NUREG/CP-0040: PROCEED!NGS OF WORKSHOP V: FLOW results were judged not to provide sufficient basis for genenc AND TRANSPORT THROUGH UNSATURATED FRACTURED requirements It was concluded that the most viable course of ROCK - RELATED 10 HIGH4EVEL RADIOACTIVE WASTE action to resolve Generic issue 105 is licensee participation in DISPOSALHeld At Radisson Suite

Hotel, Tucson, individual plant examinations (IPEs).

Anzona, January 7-10,1991. EVANS D.D.; NICHOLSON,T.J. An-zona, Univ. of. Tucson, AZ. June 1993. 250pp. 9307220313. NUREG-1470 V02: CHIEF FINANCIAL OFFICER'S ANNUAL 75744:11" REPORT - 1993.

• Office of the Controller (Post 890205). Sep-The Workshop on Flow and Transport Through Unsaturated Fractured Rock Related to High-Level Radioactive Waste Dis-h h e F nan a ce et of 1990 requires the NRC p sal" was cosponsored by the NRC, the Center for Nuclear Chief Financial Officer to prepare and submit an aMual report Waste Regulatory Analyses, and the University of Arizona (UAZ) to the agency head and the Director of the Office of Manage-and was held in Tucson, Arizona, on January 7-10,1991. The ment and Budget. This 1993 report is the second annual report focus of this workshop, similar to the earlier four (the first being for the NRC and includes a desenption and analysis of the status of financial management for Fiscal Year 1993, an audited in 1982), related to hydrogeologic technical issues associated financial statement and audit reports for Fiscal Year 1992, and with possible disposal of commercial high-level nuclear waste a summary of the reports on internal accounting and administra.

(HLW) in a geologic repository within an unsaturated fractured live control syctoms for 1992. rock system which coincides with the UAZ field studies on HLW disposal. The presentat60ns and discussions centered on flow NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT and transport processes and conditions, relevant parameters, TO CONSTRUCT AND OPERATE A FACILITY TO as well as state-of-the-art measurement techniques, and model-RECEIVE. STORE, AND DISPOSE OF 11E.(2) BYPRODUCT ing capabilities. The workshop consisted of: four half-day techni-MATERIAL NEAR CLIVE. UTAH. Docket No. 40-8989 Envirocare cat meetings; a one-day field visit to the Apache Leap test site Of Utah.Inc. BRUMMETT.E.; ABU-EID R.; MULLINS,A.; et al. Di-to review ongoing field studies that are examining site charac-vision of Low-Level Waste Management & Decommissioning terization techniques and developing data sets for model valida-(Post 670413). August 1993. 206pp. 9309210022. 76481:269. tion studies; ar.d a final half-day session devoted to examining A Final Environmental impact Statement (FEIS) related to the research needs related to modeling groundwater flow and radio-heensing of Envirocare of Utah, inc/s proposed disposal facility nuclide transport in unsaturated, tractured rock. These proceed-in Tooele County, Utah (Docket No. 40-8989) for byproduct ma-i ings provide extended abstracts of the technical presentations tenal as defined in Section 11e.(2) of the Atomic Energy Act, as and short summaries of the research group reports. amended, has been prepared by the Office of Nuclear Matenal Safety and Safeguards. This statement describes and evalu-NUREG/CP-0129: PROCEEDINGS OF THE WORKSHOP ON ates: (1) the purpose of and need for the proposed action; (2) PROGRAM FOR ELIMINATION OF REQUIREMENTS MARGIN-the attematives considered; and (3) the environmental conse-AL TO SAFETY. DEY,M. Advanced Reactors Branch (Post quences of the proposed action. The NRC has concluded that the proposed action evaluated under the National Environmen-910830). ARSENAULT.F.; PATTERSON,M.; et al. SCIENTECH, tal Policy Act of 1969 (CPA) and 10 CFR Part 51, is to permit Inc. September 1993.194pp. 9310120259. 76741:245. the apphcant to proceed with the project as desenbed in this These are the proceedings of the Public Workshop on the Statement. U.S. Nuclear Regulatory Commission's Program for Elimination of Requirements Marginal to Safety. The workshop was held at NUREG 1479: RESULTS FROM TWO WORKSHOPS: STATE RA-the Holiday Inn, Bethesda, on April 27 and 28,1993. The pur-DIATION CONTROL PROGRAMS DEVELOPING AND AMEND-pose of the workshop was to provide an opportunity for public ING PEGULATIONS AND FUNDING. PARKER.G. Office of and industry input to the program. The workshop addressed the State Programs (Post 911117). September 1993. 34pp. institutionalization of the program to review regulations with the 9310130049,76743.190. purpose of ehminating those that are marginal. The objectrve is The first section of this document presents the results of a to avoid the dilution of safety efforts. One session was devoted technical workshop on the process of regulations development to discussion of the framework for a performance-based regula-and amendment sponsored by the Nuclear Regulatory Commis-tory approach. In addrtion, panelists and attendees discussed sion (NRC). This workshop focused on methods for reducing scope, schedules, and status of specific regulatory items: con-the time it takes to prornulgate regulations to help those States tainment leakage testing requirements, fire protection require-that are having difficulty meeting the three-year deadline for ments, requirements for environmental qualification of electrical adopting new NRC regulations. Workshop participants respond-equipment, requests for information under 10CFR50.54(f), re-ed to six questions, reviewed the procedures used by the vari _ quirements for combustible gas control systems, and quality as-ous States for revising and adopting changes to their regula-tions, and reviewed the time-flow charts used by vanous States. surance requirements. This workshop was designed to provide guidance to States that NUREG/CP-0130 V01: PROCEEDINGS OF THE 22ND DOE /NRC are promulgating and revising regulations. The second section NUCLEAR AIR CLEANING CONFERENCE. Sessions 1-8. Held in of this document summarizes the proceedings of a technical workshop, also sponsored by the NRC, on funding radiation Denver Colorado, August 24-27,1992. FIRST,M.W. Harvard School of Public Health, Boston, MA. July 1993. 500pp. I control programs that emphasized fee schedules and effective 9307270008. CONF-9020823. 75801:001. strategies for the 1990s. This workshop focused on determining l the true costs of running a program, on setting realistic fees for This document contains the papers and the associated dis-l the various categories of hcenses, and on the most efficient Cussions of the 22nd DOE /NRC Nuclear Air Cleaning Confer-methods for sending invoices, recordia receipts, depositing ence. Major topics are: (1) advanced reactors; (2) reprocessing; money received, and issuing licenses. Workshop participants re-(3) filter testing; (4) waste management; (5) instruments and sponded to seven questions; reviewed the methods various sampling; (6) reactor accidents; (7) filters and filter performance; States use to determine true costs; reviewed the procedure that (8) adsorber testing and performance; (9) carbon testing; and the vanous States use to produce invoices and licenses; re-(10) ventilation systems. viewed the procedures that the States are required to abide by l

Main Citations and Abstracts 5 NUREG/CP-0130 V02: PROCEEDINGS OF THE 22ND DOE /NRC This progress report summarizes work performed by Argonne NUCLEA7 AIR CLEANING CONFERENCE. Sessions 9-16. Held Natonal Laboratory on long-term thermal embnttlement of cast in Denver. Colorado. August 24-27, 1992. FIRST,M.W. Harvard duplex stainless steels in LWR systems during the six months Set ool of Public Health, Boston, MA. July 1993. 500pp. from April-September 1992. A procedure and correlations are 9307270050. CONF-9020823. 75803.070. presented for predicting Charpy-impset energy tensile flow See NUREG/CP-0130,V01 abstract-stress, fracture toughness J-R curves, tearing modulus and J(Ic) l NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT of aged cast stainless steels from known material informaton. NUCLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY The " saturation" impact stiength and fracture toughness of 't l OF SELECTED READINGS IN RADIATION PROTECTION AND specific cast stainless steel, i.e., the minimum value that woula l ALARA. KAUR;N.D.G.; KHAN,T.A.; SULLIVAN S.G.; et at be achieved for the matenal after long-term service, is estimat-i Brookhaven National Laboratory. July 1993. 106pp. ed from the chemical composition of the steel. Mechanical 9308160137. BNL-Nt lREG-51708. 76120:001. properties as a function of time and temperature of reactor l The ALARA Cente at Brookhaven Natonal Laboratory pub-service are estimated frorn impact energy and flow stress of the lashes a senes of bibliographies of selected readings in radiation unaged matenal and the kinetics of embnttlement, which are protecton and ALARA in the continuing effort to collect and dis-also determined from chemical composition. The J(Ic) valuws seminate informat:on on radiation dose reduction at nuclear are determined from the estimated J-R curve and flow stress. i l power plants Thic is volume 7 of the senes. The abstracts in Examples of estimating mechanical propertees of cast stainless this bibliography were selected from proceedings of technical steel componen5 dunng reactor service are presented. A meetings and conferences, journals, research reports, and common " lower-bound" J-R curve for cast stainless steels of searches of the Energy Science and Technology database of unknown chemical composition is also defined for a given grade the U.S. Department of Energy. The subject material of these of steel, ferrite content, and temperature, abstracts relates to radiaton protection and dose reduction, and ranges from use of robotics to operational health physics, to NUREG/CR-5404 V02: AUXILIARY FEEDWATER SYSTEM water chemistry. Matenal on the design, planning, and manage-AGING STUDY. Phase i Follow-On Study. KUECK,J.D. Oak ment of nuclear power statens is included, as well as informa-i Ridge National Laboratory. July 1993. 39pp. 9308160268. l ton on decommissioning and safe storage efforts. Volume 7 ORNL-6566/V1. 76118:327. contains 293 abstracts, an author index, and a subject index. The author index is specific for tbv. volume. The subject inder is This report documents the results of a Phase I follow-on cumulative and lists all abstract numbers from volumes 1 to 7. study of the Auxilia'Y Feedwater (AFW) System that has been The numbers in boldface indicate the abstracts in this volumo, conducted for the U S. Regulatory Commission's Nuclear Plant the numbers not in boldface represent abstracts in previous vol. Aging Research Program. The Phase I study found a number of

umes, significant AFW System functions that are not betreg adequately tested by conventional test methods and some that are actually NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACK

• being degraded by conventional testing. Thus, it was decided I

ING IN LIGHT WATER REACTORS. Semiannual I that this follow-on study would focus on those testmg omissions Report, October 1992 - March 1993. CHUNG.H.M.; CHOPRA.O K.; RUTHER,W.E.; et al. Argonne National Labora-and equipment degradation. The deficiencies in current monitor-ing and operating practice are categonzed and evaluated. Areas i to September 1993. 67pp. 9310120274. ANL-93/27. of component degradation caused by current practice are dis-This report summanzes work performed by Argonne National c ssed. Recommendatons are made for improved diagnostic j j Laboratory on fatigue and environmentally assisted cracking methods and test procedures. i i (EAC) in light water reactors (LWRs) during the six months from NUREG/CR-5754: BOILING-WATER REACTOR INTERNALS October 1992 to March 1993. Fatigue and EAC of piping, pres-sure vessels, and core components in LWRs are important con-AGING DEGRADATION STUDY. Phase 1. LUK,K.H. Oak Ridge National Laboratory. September 1993. 56pp. 9310120363. cerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel ORNL/TM-11876. 76740 333-i used in piping, steam generators, and reactor pressure vessels, This report documents the results of an aging assessment l (2) EAC of cast stainless steels (SSs), (3) radiation-induced study for boiling water reactor (BWR) internals. Major stressors segregaton and irradtation-assisted stress corrosion cracking of for BWR internals are related to unsteady hydrodynamic forces Type 304 SS after accumulation of relatively high fluence, and generated by the primary coolant flow in the reactor vesset. (4) EAC of low-alloy steets. Fatigue tests were conducted on Welding and cold-working, dissolved oxygen and impurities in medium-sulfur-content A106-Gr B piping and A533-Gr B pres-the coolant, applied loads and exposures to fast neutron fluxes sure vessel steels in simulated PWR water and in air. Additional are other important stressors. Based on results of a component crack growth data were obtained on fracture-mechanics speci-failure informaton survey, stress corrosion cracking (SCC) and l mens of cast austenitic SSs in the as-received and thermally fatigue are identified as the two major aging-related degradation j aged conditons and chromium-nickel-plated A533-Gr B steelin mechanisms for BWR internals. Significant reported failures in-simulated boiling-water reactor (BWR) water at 289 degrees C. clude SCC in jet-pump holddown beams, in core neutron flux The data were compared with predictions based on crack monitor dry tubes and core spray spargers. Fatigue failures growth correlations for ferritic steels in oxygenated water and were detected in feedwater spargers. The implementation of a correlations for wrought austenitic SS in oxygenated water de-plant Hydrogen Water Chemistry (HWC) program is considered veloped at ANL and rates in air from Section XI of the ASME as a promising method for controlling SCC problems in BWR. Code. Microchemical and microstructural changes in high-and More operating data are needed to evaluate its effectiveness commercial-purity Type 304 SS specimens from control-blade for internal components. Long-term fast neutron irradiation ef-absorber tubes and control-blade sheath from operating BWRs fects and high-cycle fatigue in a corrosive environment are un-were studied by Auger electron spectroscopy and scanning electron microscopy. certainty factors in the aging assessment process. BWR inter-nals are examined by visual inspectons and the niethod is l NUREG/CR-4744 V07 N2: LONG TERM EMBRITTLEMENT OF access limited. The presence of a large water gap and an ab-l CAST DUPLEX STAINLESS STEELS IN LWR sence of ex-core neutron flux monitors may handicap the use of SYSTEMS. Semiannual Report, April-September 1992. advanced inspection methods, such as neutron noise vibration CHOPRA,0.K. Argonne National Laboratory. July 1993. 54pp-measurements, for BWR. 9308160133. ANL-93/11. 76120:301.

l 6 Main Citations and Abstracts NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR sis provide the NRC with (1) data that could be used to assess POWER INDUSTRY. Annual Summary Of Program Performance the relative influence of a number of key snput parameters in the Reports.CY 1992. FLEMING.T.; WESTRA,C.; FIELD,l.; et al. Yankee Rowe PTS analysis and (2) data that can be used for Battelle Human Affairs Research Centers. July 1993. 140pp. readily determining the probability of vessel failure once a more 9308180128 PNL-8688. 76142:200. accurate indcation of vessel embnttiement becomes available. This report nummarizes the data from the semaannual reports This report is designated as HSST report No.117, on fitness-for-duty programs submitted to the NRC by 54 utilities for two reporting periods: January 1,1992 to June 30, 1992 NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE and from July 1,1992 to December 31,1992. Dunng CY 1992, WASTE RESEARCH AT CNWRA. July-December 1992. Icensees reported that they conducted 266,551 tests for the ABABOU.R.; AHOLA M.P.; BACA.R.G.; et at. Center for Nuclear presence of illegal drugs and alcohol. Of these tests,1,818 Waste Regulatory Analyses. July 1993. 200pp. 9309090005. (.68%) were confirmed positive. Positive test results vaned by CNWRA 92-02S. 76393:001. category of test and category of worker. Th3 majonty of positve Progress from July 1 to December 31,1992 on the nine NRC-test results (1,110) were obtained through pre-access testirg. sponsored research projects conducted at the Center for Nucle-Of tests conducted on workers having access to the protected er Waste Regulatory Analyses is described. lon-exchange ex-area, there were 461 positive tests from random testing, and periments between clinoptilolite and aqueous solutions of 178 positive tests from for-cause testing. Followup testing of Na(+) and Sr(2+) and three applicatons of reactorkpath mod-workers who had previously tested positive resulted in 69 posi-eling are described in the Unsaturated Mass Transport (Geo-tive tests. Positrve test results also varied by category of worker. chemistry) project. Numerical simulation of a laboratory-scale Overall, short-term and long-term contractor personnel had the non-isothermal two-phase flow is discussed in the Thermohy-highest rates of positive tests. Licensee employees had lower drology chapter. Methods for estimating rock joint roughness rates of positive test results. coefficient are the focus of the Seismic Rock Mechanics project NUREG/CR-5778 V03: NEW YORK /NEW JERSEY REGIONAL for which the Tilt Test, Tse and Cruden's equations, and fractal-SEISMIC NETWORK. Final Report For April 1985 - September based equations were tested and found to be unsatisfactory. In 1992. SEEBER.L; JOHNSON.D.; ARMBRUSTER,J. Lamont-Do-the integrated Waste Package Expenments chapter, investiga-herty Geological Observatory July 1993. 74pp. 9307270035. tions of pit initiaton and repassivation potential for alloys 825 75802:317. and C-22 and stainless steel 304L and 316L are described. l For almost 20 years, Lamont-Doherty Earth Observatory has Testing of the BIGFLOW computer code and visualization of operated the primary network for monitonng earthquake activity fracture topology is the theme of the Stochastic Hydrology in the New York State, northem New Jersey, and northwestem project. Preliminary analysis of feeld data from the Akrotiri site in l Vermont area, wth support by both NRC and the USGS. The Greece is developed in the Geochemical Analogs project. pnmary purpose of this research is directed toward the determi-Mechanistic modeling of sorpton using the MINTEOA2 code is nation of local seismicity and the possible identifcaton of asso-investigated as part of the Sorption project. Adaptive gridding ciated geologic and tectonc features. From April 1985 to Sep-and " modified equatons" methods for solving the flow and tember 1992, the network recorded and located 346 regonal transport equatons are described in the Performance Assess-l earthquakes. Scientific activity, pnmanly in the form of after-ment chapter. Finally, the Volcanism chapter focuses on using j shock monitonng, was concentrated upon a number of signifi-nonhomogeneous Poisson processes for estimating probability cant earthquakes: The Ardsley, NY: the Chardon, OH, the Ash-of volcanic events at the potential repository site. tabula, OH; and the Saguenay, Canada earthquakes; in additon to the Summit, NY event. These studies involved the deploy NUREG/CR-5829: AUXILIARY FEEDWATER SYSTEM RISK-ment of portable seismographs in the epicentral areas. Many of BASED INSPECTION GUIDANCE FOR THE DAVIS-BESSE NU-these sequences were in northeastem North America, but out-CLEAR POWER PLANT. NICKOLAUS.JR: MOFFITT,N E.: side the L-DEO seismic network and were not covered by other GORE,B.F.! et al. Battelle Memorial Institute, Pacife Northwest permanent networks. Spatial correlatons between structures Laboratory. September 1993. 38pp. 9310t20332. PNL-7905. and earthquakes were found at a wide range of scales, and sys-76742:139. temate searches of archival matenal were used to improve con-in a study sponsored by the U.S. Nuclear Regulatory Com-straints on historc sources. mission (NRC), Pacific Northwest Laboratory has developed and j NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABI-applied a methodology for deriving plant-specife risk-based in-LISTIC FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR specton guidance for the auxiliary feedwater (AFW) system at ) l YANKEE ROWE REACTOR PRESSURE VESSEL pressurized water reactors that have not undergone probabilistc DICKSON,T.L.; CHEVERTON.R.D.; BRYSON,J.W.; et al. Oak risk assessment (PRA). This methodology uses existing PRA re. l Ridge National Laboratory. August 1993.114pp. 930921C205. suits and plant operating experience information. Existing PRA-l ORNL/TM-11945. 76484:336. based inspection guidance information recently developed for The Nuclear Regulatory Commission (NRC) requested Oak the NRC for vanous plants was used to identify generic compo-l Ridge Natonal Laboratory (ORNL) to perform a pressurced-nent failure mocles. This information was then combined with thermal shock (PTS) probabiliste fracture mechanics (PFM) plant-specific and industry-wide component information and fall-i sensitivity analysis for the Yankee Rowe reactor pressure ure data to identify failure modes and failure mechanisms for vessel, for the fluences corresponding to the end of operating the AFW system at the selected plants. Davis-Besse was se-cycle 22, using a specific small-break-loss.of-coolant transient lected as one of a series of plants for study. The product of this as the loading condition. Regions of the vessel with distinguish-effort is a proritized listing of ACW failures which have occurred ing features were to be treated individually-upper axial weld, at the plant and at other PWRs. This listing is intended for use lower axial weld, circumferential weld, upper plate spot welds, by NRC inspectors in the preparaton of inspection plans ad-I upper plate regions between the spot welds, lower plate spot dressing AFW risk-important components at the Davis-Besse I welds, and the lower plate regions between the spot welds. The plant. fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-computer code, which was developed dunng the integrated BASED INSPECTION GUIDE FOR THE H.B. ROBINSON NU. Pressurized Thermal Shock (IPTS) Program. The NRC request CLEAR POWER PLANT. MOFFITT,N.E.; LLOYD.R.C.; specified that the OCA-P code be enhanced for this study to GORE.B.F.; et al. Battelle Memonal Institute, Pacife Northwest also calculate the conditonal probabilities of failure for subclad Laboratory. August 1993. 37pp. 9308160303. PNL-7907. flaws and embedded flaws. The results of this sensitivity analy-76119.293. l 1 i

l Main Citations and Abstracts 7 i in a study sponsored by the U.S. Nuclear Regulatory Com-a dry cavity and with a flooded cavity) and a Loss-Of-Coolant mission (NRC), Pacific Northwest Laboratory has oeveloped and Accident (LOCA) concurrent with complete loss of the Emer-applied a methodology for denving plant-specific nsk based in-gency Core Cooling System (ECCS) were analyzed for the spection guidance for the auxiliary foodwater (AFW) system at Peach Bottom Atomic Power Station (a BWR-4 with a Mark I pressunzed water reactors that have not undergone probabilistic containment). The results indicate that for the sequences ana-nsk assessment (PRA). This methodology uses existing PRA re-lyzed. the two codes predict similar total in-containment release suits and plant operating expenence information. Existing PRA-fractions for each of the element groups. However, the based inspection guidance information recently developed for MELCOR/CORBH Package predicts significantly longer time for the NRC for vanous plants was used to identify generic compo-vessel failure and reduced energy of the released matenal fed nent failure modes. This information was then combined Wh the station blackout sequences (when compared to the STCP plant-speci'ic and industry-wide component information and fail-results). MELCOR also calculated smaller releases into the en-ure data to identify failure modes and failure mechanisms for vironment than STCP for the statton blackout sequences. the AFW system at the selected plants. H. B. Robinson was se-lected as one of a senes of plants for study. The product of this NUREG/CR-5943: SENSmVITY ANALYSIS AND BENCHMARK-effort is a pnont: zed listing of AFW failures which have occurred ING C" THE BLT LOW-LEVEL WASTE SOURCE TERM CODE. at the plant and at other PWRs. This listing is intended for use SUEN.C.J.; SULLIVAN,T.M. Brookhaven National Laboratory. by NRC inspectors in the preparation of inspection plans ad-July 1993. 81pp 9307270012. BNL-NUREG-52346. 75802.236. dressing AFW nsk-important components at the H. B. Robinson To evaluate the source term for low-level waste disposal, a plant comprehensive model had been developed and incorporated ento a computer code, called BLT (Breach-Leach-Transport). NUREG/CR-5927 V01: EVALUATION OF A PCDCORMANCE AS-Since the release of the ong'nal version, many new features SESSMENi mc e nvDOLOGY FOR LOW-LEVEL HADIOAC-and improvements had also been added to the Leach model of TIVE WASTE DISPOSAL FACILITIES Evaluation Of Modeling the code. This report consists of two different studies based on Approaches. KOZAK.M.W,; OLAGUE.N E.; RAO.R.R.; et al. the new version of the BLT code: 1) a series of venfication/sen-Sandia National Laboratones. August 1993. 87pp. 9309210242. sativity tests; and 2) benchmarking of the BLT code using field SAND 912802. 76486:118-data. Based on the results of the venfication/ sensitivity tests, This report represents an update to our earlier reports on low-we concluded that the new version represents a significant im-level waste performance assessment. This update addresses provement and it is capable of providing more realistic simula-needed improvements and recommended approaches to the tions of the leaching process. Benchmarking work was camed existing state of the art in modeling, treatment of uncertainty, out to provide a reasonable level of confidence in the model and use of data. Greater attention is paid to developing an inte* predictions. In this study, the experimentally measured release grated approach to performance assessment than was done in curves for nitrate, technetium-99 and tntium from the saltstone earlier developments of the methodology. Furthermore, insights fysimeters operated by Savannah River Laboratory were used. are being developed by participating in validation exercises, and The model results are observed to be in general agreement i by evaluating which valdation data are needed to improve con-with the experimental data, within the acceptable limits of urb idence in the methodology. It is emphasized that the perform-certainty. i ance assessment methodology update is a work in progress; the recommendations given here will form the general directions NUREG/CR-5944: A CHARACTERIZATION OF CHECK VALVE toward which the methodology is heading, but some of the spe-DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLE-cific approaches may continue to evolve as the research pro-AR POWER INDUSTRY. CASADA.D.A.; TODD.M.D. Oak Ridge I gresses. National Laboratory. September 1993. 187pp. 9310120234. NUREG/CR-5928: ISLOCA RESEARCH PROGRAM Final Report. ORNL-6734. 76739:342. . Check valve operating problems in recent years have resulted GALYEAN,W.J.; KELLY.D.L; SCHROEDER,J.A.; et al. EG&G Idaho, Inc. Ju?y 1993 107pp. 9308160131. EGG 2685. m significant operating transients, increased cost and decreased 76120.108. system availability. As a result, additional attention has been This report contains a compilation of information generated gen to check valves by utilities (resulting in the formation of during the ISLOCA research program. Presented is a screening the Nuclear Industry Check Valve Group), as well as the U.S. analysis and a procedures guide for performing an ISLOCA Nuclear Regulatory Commission and the Amencan Society of evaluation. This methodology has been distilled from past analy. Mechanical Engineers Operation and Maintenance Committee. ses performed for the U.S. Nuclear Regulatory Commission and All these organizations havs the fundamental goal of ensunng documented in a senes of NUREG/CR reports. The methodolo-reliable operation of check valves. A key ingredient to an engi. gy compnses five distinct steps: (a) containment penetration neenng-oriented reliability improvement effort is a thorough un-screening; (b) interfaces for ISLOCA analysis; (c) mechanisms derstanding of relevant historical experience. A detailed review for failing the pressure boundary; (d) construction of event trees of histoncal failure data, available through tha Institute of Nucle-and estimation of rupture probabilities; and (e) quant:fetion of ar Power Operation's Nuclear Plant Reliability Data System, has j the event tree included in the methodology are steps required been conducted. The focus of the review is on check vatve fail-for a detailed human reliability analysis. In addition, this report ures that have involved significant degradation of the valve in-presents a BWR ISLOCA evaluation, a survey of PWR auxiliary temal parts. A vanety of parameters are considered, including building designs and identification of one design deemed most size, age, system of service, method of failure discovery, the 81-disadvantageous with respect to ISLOCA nsk, and a PWR fected valve parts, attributed causes, and corrective actions. ISLOCA cost / benefit analysis. NUREG/CR-5969: J AND CTOD ESTIMATION EOUATIONS FOR l NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHAR-SHALLOW CRACKS IN SINGLE EDGE NOTCH BEND SPECI-ACTERISTICS FOR SELECTED PEACH BOTTOM SE-MENS. KIRK,M.T.; DODDS,R.H. tilinois, Univ. of, Urbana, IL

• OUENCES PREDICTED BY THE MELCOR CODE.

Navy, Dept. of. July 1993. 28pp. 9308160121. UILU ENG91-CARBAJO,J.J. Oak Ridge National Laboratory. September 2013. 76121:333. 1993.345pp.9310120226. ORNL/TM-12229. 76739:001. Fracture toughness values determined using shallow cracked The purpose of this report is to compare in-containment single edge notch bend. SE(B), specimens of structural thick-source terms developed for NUREG 1159, which used the ness are useful for structural integrity assessments. Results Source Term Code Package (STCP), with those generated by from two dimensional plane strain finite-element analyses are MELCOR to identify significant differences. For this companson, used to develop J and CTOD estimation strategies appropnate two short-term depressunzed station blackout sequences (with for application to both shallow and deep crack SE(B) speci-I

8 Main Citations and Abstracts mens. Crack depth to specimen width (a/W) ratios between Recent computational studies of the stress and strain fields at 0 05 and 0.70 are modelled using Ramberg-Osgood strain hard-the tip of very sharp notches have shown that the stress and ening exponents (n) between 4 and 50. The estimation formulas strain fields are very weakly dependent on the initial geometry divide J and CTOD into small scale yielding (SSY) and large of the notch once the notch has been blunted to a radius that is scale yielding (LSY) components. For each case, the SSY com-6 to 10 times the initial root radius. It follows that if the fraClure ponent is oetermined by the linear elastic stress intensity factor, toughness of a matenal is sufficiently high so that fracture initi-K(1). The formulas differ in evaluation of the LSY component. ation does not occur en a specimen until the crack-tip opening The techniques considered include: estimating J or CTOD from displacement (CTOD) reaches a value from 6 to 10 times the plastic work based on load line displacement (A(p1)/LLD), from size of the init:al notch tip diameter, then the fracture toughness plaste work bawo on crack mouth opening displacement will be independent of whether a fatigue crack or a machined I (A(p1)/CMOD). and from the plastic component of crack mouth notch served as the initial crack. In this expenmentaf program l opening displacement (CMOD(p1)). A(p1)/CMOD provides the the fracture toughness (J(Ic) and J resistance (J-R) curve, and most accurate J estimation possible. CTOD) for several structural alloys was measured using speci-mens with conventional fatigue cracks and with kDM machined NUREG/CR-5970: APPROXIMATE TECHNIQUES FOR PREDICT-n tches. The results of this program have shown, in fact, that ING SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS m st smral maWals do not ach inmaton N vabes l (JC). KIR A M.T.; DODDS.R.H. lilinois, Univ. of, Urbana. IL, on the order of 6 to 10 times the radius of even the smallest Navy, Dept. of. July 1993. 38pp. 9308160116. UILU-ENG92-EDM notch tip presently achievable. It is found furthermore that 2016. 76122:001 This investigat$on examines the ability of an elaste T-stress tougher matenals do not seem to be less dependent on the type of notch tip present. Some matenals are shown to be analysis coupled with a modified boundary layer (MBL) solution much mure dependent on the type of initial notch tip used, but to predict stresses ahead of a crack tip in a variety of planar no simple pattern is found that relates this observed depend-geometnes. The approximate stresses are used as input to esti-enm 6 me maMnal hgm, WgWss, m stain hahng mate the effective dnving force for cleavage fracture (J(o)) using ra h the micromechancally based approach introduced by Dodds and Anderson. Finite element analyses for a wide vanety of NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMITY planar cracked geometnes are conducted which have elastic DOS 1 METERS--PILOT TEST. FOX,R.A ; HARTY,R.; biaxiahty parameters (d) rangmg from -0.99 (very low con-MCDONALD.J.C. Battelle Memonal Institute, Pacife Northwest stramt) to + 2.96 (very high constraint). The magnitude and sign Laboratory. July 1993. 51pp. 9308160263. PNL-8467. of 3 mdicate the rate at which crack-tip constraint changes 76119.169. with increasing applied load. All results pertain to a moderately A working group of the Health Physics Society Standards strain hardening matenal (strain hardening exponent (n) of 10). Committee (HPSSC) has issued a draft standard for extremity Tnese analyses suggest that # is an effective indicator of dosimeters. To determine the appropnateness of the proposed both the accuracy of T-MBL estimates of J(o) and of applicabih standard, Pacific Northwest Laboratory (PNL) has conducted sty limits on evolving fracture analysis methodologies (i.e. T-three separate evaluations of the performance by processors of MBL, J-O. and J/J(o)). Specifically, when 1#1 >0.4 these extremsty dosimeters. The dosimeters were tested in each of analyses show that the T-MBL approximation of J(o)is accurate the irradiation categories specified in the draft standard. high-to within 20% of a detailed finite-element analysis. As "structur-energy photons (general and accident dosimetry), low-energy al type" configurations, i e. shallow cracks in tension, generally photons (general and accident dosimetry), beta particles, neu-have 131 >0.4, it appears that only an elastic analysis may trons (first and second evaluatons only), and a mixture catego-I be needed to determine reasonably accurate J(o) values for ry, in the first evaluation only about 60% of the processors met l structural conditions. the draft standard's performance entena for accuracy and preci-l NUREG/CR-597t CONTINUUM AND MICROMECHANICS sion. The second evaluation showed an overall improvement of l TREATMENT OF CONSTRAINT IN FHACTURE. DODDS,R H. 15% to 18%, but most processors were still unable to meet the lilinois, Univ. of, Urbana, IL. SHIH C.F. Brown Univ., Providence, performance critena consistently in all irradiation categones, Rt. ANDERSON T.L Texas A&M Univ., College Station. TX. July After these evaluations, PNL suggested several changes to the 1993.47pp.9308160294. UILU-ENG92-2014. 76119.122. draft standard, including redefining some of the test categones Two complementary methodologies are desenbed to quantify and making the tolerance levels of the entena more consistent I the effects of crack-tip stress inaxiality (constraint) on the maC-with those in the standard for whole-body dosimetry. This report roscopic measures of elastc-plaste fracture toughness, J and summarizes the third evaluation, which y6elded an overall pass CTOD. In the continuum mechanics methodology, two param. rate of 87%. Suggestions are given to render the draft standard eters, J and O, suffce to charactenze the full range of near-tip generally more consistent with the criteria for whole-body do-environments at the onset of fracture. A micromechanics meth-S' metry. odology is desenbed which predcts the toughness locus using NUREG/CR-5993 V0t METHODS FOR DEPENDENCY ESTIMA. crack-tip stress fields and entical J-values from a few fracture TION AND SYSTEM UNAVAILABILITY EVALUATION BASED toughness tests. A robust micromechanics model for cleavage ON FAILURE DATA STATISTICS Summary Report. fracture has evolved from the observations of a strong, spatial AZARM,M.A.; HSU,F.; MARTINEZ-GURIDl; et al. Brookhaven self-similanty of crack-tip pnncipal stresses under increased i National Laboratory. July 1993. 37pp. 9307270039. BNL-toading and across different fracture specimens. This report ex-NUREG-52362. 75803:032. cres the fundamental concepts of the J-O desenpton of This report introduces a new perspective on the basic con-crack-tip fields, the fracture toughness locus and micromechan-cept of dependent failures where the deftniton of dependency ics approaches to predict the variability of macroscopic tracture is based on clustenng in failure temas of similar components. toughness with constraint under elastc-plastic conditions. Com-This perspective has two signifcant impicatons: firstly, it re-putational results are presented for a surface cracked plate con-laxes the conventional assumpton that dependent failures must taining a 6.1 semi-elliptcal, a =t/4 flaw subjected to remote un-be simuttaneous and result from a severe shock; secondly, it iaxial and biaxial tension. allows the analyst to use all the failures in a time continuum to NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE estimate the potential for multiple failures in a window of time MACHINED NOTCHES ON THE FRACTURE TOUGHNESS OF (e g,, a test intervaf), therefore amving at a rnore accurate value SEVERAL STRUCTL9AL ALLOYS. JOYCE.J.A. U.S. Naval for system unavailability. In additon, the models developed here Academy, Annapolis, MD. LINK,R E. Navy, Dept. of. September provide a method for plant-specific analysis of dependercy, re-1993. 90pp. 9310130032. 76744.290. flecting the plant-specifc maintenance practces that reduce or l

I l I I Main Citations and Abstracts 9 increase the contribution of dependent failures to system un-research laboratories. This report provides an overview of prin-r availability. The proposed methodology can be used for screen-cipal developments in each of the four program tasks from Jan-l ing analysis of failure data to estimate the fraction of dependent uary 1,1992 to December 31,1992. Planned activities under failures among the failures. In addition, the proposed method each of these tasks are also presented, i can evaluate the impact of the observed dependency on system unavailabihty and plant nsk. The formulations dorived in this NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED report have undergone various levels of vahdations through THERMAL-HYDROLOGIC-MECHAN ICAL-CHEMICAL PROC-computer simulation studies and pilot applicahons. The pilot ap-ESSES PERTINENT TO THE PROPOSED HIGH-LEVEL plications of these methodologies showed that the contnbution WASTE REPOSITORY AT YUCCA MOUNTAIN. of dependent failures of diesel generators in one plant was neg-MANTEUFELR.D.; AHOLA,M.P.; TURNER.D.R.; et al. Center hgible, while in another plant, it was quite significant. It also for Nuclear Waste Regulatory Analyses. July 1993. 240pp. showed that in the plant with significant contribution of depend-9308160184. CNWRA92-011, 76117:040. ency to Emergency Power System (EPS) unavailability, the con-A hterature review has been conducted to determine the state t inbution changed with time. Similar findings were reported for of knowledge available in the modehng of coupled thermal (T), the Containment Fan Cooler breakers. Drawing such conclu-hydrologic (H), mechanical (M), and chemical (C) processes rel - seons about system performance would not have been possible evant to the design and/or performance of the proposed high-with any other reported dependency methodologies, level waste (HLW) repository at Yucca Mountain, Nevada. The f NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMA-review focuses on identifying couphng mechanisms between in-TION AND SYSTEM UNAVAILABILITY EVALUATION BASED dividual processes and assessing the importance (i.e., if the ir ON FAILURE DATA STATISTICS. Detailed Description And / coupling is either important, potentially important, or negligible). plications. AZARM,M.A.; HSU,F-; MARTINEZ-GURIDl; et The significance of considering THMC-coupled processes lies sn whether or not the processes impact the design and/or per-j Brookhaven National Laboratory. July 1993.78pp.9307270044. BNL-NUF',EG-52362. 75802:143. formance objectives of the repository. A review, such as report-This report introduces a new perspective on the basse con-ed here, is useful in identifying which coupled effects will be im-cept of dependent failures where the definition of dependency portant, hence which coupled effects will need to be investigat- [ is based on clustenng in failure times of similar components. ed by the U.S. Nuclear Regulatory Commission in order to 6 This perspectrve has two signifi snt implications: firstty. it re. assess the assumptions, data, analyses, and conclusions in the taxes the conventional assumption that dependent failures must design and performance assessment of a geologic reoository. p be simultaneous and result from a severe shock; secondly, it Although this work stems from regulatory interest in the design e allows the analyst to use all the failures in a time continuum to of the geologic repository, it should be emphasized that the re-l estimate the potential for multiple failures in a window of time Pository design implicitly considers all of the repository perform-t (e.g., a test interval), therefore arriving at a more accurate value ance objectives, including those associated with the time after l for system unavailability. In addition, the models developed here permanent closure. The scope of this review is considered provide a method for plant-specifsc analysis of dependency, re-beyond previous assessments in that it attempts (with the cur-i flecting the plant-specific maintenance practices that reduce or rent state.of-knowledge) to determine which couplings are im-l increase the contribution of dependent failures to system un-portant, and identify which computer codes are currently avail-availabihty. The proposed methodology can be used for screen. able to model coupled processes. ing analysis of failure data to estimate the fraction of dependent failures among the failures. In addrieon, the proposed method NUREG/CR 6023: GENERIC ANALYSIS FOR EVALUATION OF LOW CHARPY UPPER-SHELF ENERGY EFFECTS ON can evaluate the impact of the observed dependency on system unavailabihty and plant nsk. The formulations denved in this SAFETY MARGINS AGAINST FRACTURE OF REACTOR report have undergone various levels of vahdations through PRESSURE VESSEL MATERIALS. DICKSON,T.L Oak Ridge computer simulatton studies and pilot applications. The pilot ap-National Laboratory. July 1943. 82pp. 9308160110. ORNL/TM-plications of these methodologies showed that the contribution 12340. M12&215. of dependent failures of diesel generators in one plant was neg-Appendix G to 10 CFR Part 50 requires that reactor pressure ligible, while in another plant, it was quite significant. It also vessel belthne matenals maintain Charpy upper-shelf energies showed that in the plant with significant contribution of depend-of no less than 50 ft-Ib dunng the plant operating life, unless it ency to Emergency Power System (EPS) unavailability, the con-is demonstrated in a manner approved by the Nuclear Regula-l tribution changed with time. Similar findings were reported for . tory Commission (NRC), that lower values of Charpy upper-shelf the Containment Fan Cooler breakers. Drawing such conclu-energy provide margins of safety against fracture equivalent to sions about system performance would not have been possible those in Appendix G to Sec+ ion XI of the ASME Code. Analyses i with any other reported dependency methodologies. based on acceptance enteria and analysis methods adopted m the ASME Code Case N-512 are described herein. Additionalin-l NUREG/CR-6015: STRUCTURAL AGING PROGRAM TECHNi-formation on material properties was provided by the NRC, CAL PROGRESS FOR PERIOD JANUARY - DECEMBER 1992. Office of Nuclear Regulatory Research, Materials Engineering NAUS.D.J.; OLAND C.B. Oak Ridge National Laboratory. July Branch. These cases, specified by the NRC, represent genenc 1993.164pp. 9309030224. ORNL/TM-12342. 76328:021. apphcations to boiling water reactor and pressurized water reac-The Structural Aging (SAG) Program is conducted for the Nu-tor vessels. This report is designe'ed as HSST Report No.140. j clear Regulatory Commission (NRC) by the Oak Ridge National j Laboratory (ORNL). The program has the overall objective of NUREG/CR 6026: THEORETICAL AND EXPERIMENTAL INVES-i prepanng an expandable handbook or report which will provide TIGATION OF THERMOHYDROLOGIC PROCESSES IN A l potential structural safety issues and acceptance criteria for use PARTIALLY SATURATED, FRACTURED POROUS MEDIUM. j by the NRC in nuclear power pLnt evaluations of continued GREEN R.T.; MANTEUFELR.D. Center for Nuclear Waste Reg - j service. Initial focus of the program is on concrete and con-ulatory Analyses. DODGE,F.T.; et at Southwest Research Insti-crete-related materials which comprise safety related (Category tute. July 1993. 225pp. 9309090010 CNWRA 92-006.

1) structures in hght-water reactor facilities. The SAG Program is 76390:064.

organized into four tasks: Task S.1+rogram Management Task The performance of a geologic repository for high-level nucle-S.2-Materials Property Data Base, Task S.3-Structural Compo-er waste will be influenced to a large degree by thermohydroio-nent Assessment / Repair Technology, and Task S.4-Ouantitative gic phenomena created by the emplacement of heat-generating Methodology for Continued Service Determinations, in meeting radioactive waste. The importance of these phenomena is mani-the individual objectives of these tasks resources are drawn fest in that they can greatly affect the movement of moisture - from ORNL with subcontract support from universities and other and the resulting transport of radionuclides from the repository. i

10 Main Citations and Abstracts Thus, these phenomena must be well understood pnor to a de-The Pacific Northwest Laboratory conducted a Phase I aging finitive assessment of a potential repository site. An investiga-assessment of chillers used in the essential safety air-condition-i tion has been undertaken along three separate avenues of ing systems of nuclear power plants. Centnfugal chillers in the analysis: (1) laboratory experiments; (2) mathematical models; 75-to 750-ton refngeration capacity range are the predominant and (3) similitude analysis. A summary of accomphshments to type used. The chillers used, and air-conditioning systems date la as follows: (1) A renew of the titerature on the theory of served, vary in design from plant-to-plant. It is crucial to keep heat and mass transfer 'n partially saturated porous medium; (2) chiller internals very clean and to prevent the leakage of water. i A development of the goveming conservation and constitutwe air, and other cont? Hnants into the refngerant Containment equateons; (3) A development of a dimensionless form of the system. Periodic opwtion on a weekly or monthly basis is nec-governing equation; (4) A numencal study of the importance and essary to remove moistura and nu-condensable gases that sensitivity of flow to a set of dimensionless groups; (5) A survey gradually build up inside the chiller. This is especially desirable if and evaluation of expenmental measurement techniques; (6) a chiller is required to operate only as an emergency standby Execution of laboratory expenments of nonisothermal flow in a unit. The pnmary stressors and aging mechanisms that affect porous medium with a simulated frccture. chillers incicde vibration, excessive temperatures and pressures, NUREG/CR-6029 V01: AGING ASSESSMENT OF NUCLEAR thermal cycling, chemical attack, and poor quality cooling water. AIR-TRF ATME NT SYSTEM HEPA FILTERS AND Aging is accelerated by moisture, non-condensable gases (e.g., ADSORBERS. Phose L WINEGARDNER.W. Battelle Memorial air), dirt, and other contamination within the refrigerant contain-institute Pacific Northwest Laboratory. August 1993. 48pp. ment system, excesswe start /stop cycling, and operating below 9309210238. PNL-8594. 76486-075. the rated capacity. Aging is also accelerated by corrosion and A N656 I aging assessment of high-efficiency particulate air fouling of the condenser and evaporator tubes. The pnncipal (HEPA) filters and actwated carbon gas adsorption units (ad-cause of chiller failures is lack of adequate monitonng. Lack of sorbers) was performed by the Pacific Northwest Laboratory performing scheduled maintenance and human errors also con-(PNL) as part of the U.S. Nuclear Regulatory Commission's tribute to failures. (NRC) Nuclear Plant Aging Research (NPAR) Program. Informa-tion concerning design features, failure expenence; aging mech-NUREG/CR-6048: PRESSURIZED WATER REACTOR INTER-anisms, effects, and stresso s; and surveillance and monitonng NALS AGING DEGRADATION STUDY. Phase 1. LUK,K.H. Oak methods for these key air-treatment system Components was Ridge National Laboratory. September 1993. 66pp. compiled. Over 1100 failures, or 12 percent of the filter instal!a-9310120328. ORNL/TM-12371. 76740:266. tions, were reported as part of a Department of Energy (DOE) This report documents the results of a Phase 1 study on the survey. Investigators from other national laboratones have sug-effects of aging degradations on pressunred-water reactor gested that aging effects could have contnbuted to over 80 per-(PWP) intomals. Prima y stressors for intemals are generated cent of these failures. Tensile strength tests on aged filter by the primary coolant flow in the reactor vessel, and they in-media specimens indicated a decrease in strength. Filter aging ciude unsteady hydrodynamic forces and pump-generated mechanisms range from those associated with particle loading ressure pulsations. Other stressors are applied loads, manu-to reactions that alter properties of sealants and gaskets. Low , g gg g 3 g radiotodine decontamination factors associated with the Three 9 fast neutron flures. A survey of reported aging-related failure in-E Mile island (TMI) accident were attnbuted to the premature formation indicates that fatigue, stress corrcsion cracking (SCC) aging of the carbon in the adsorbers Mechanisms that can lead and mechanical wear are the three major aging-related degra-to impaired adsorber performance include oxidation as well as the loss of potentially available active srtes as a result of the dation mechanisms for PWR intemals. Significant reported fail-adsorption of pollutants Stressors include heat, moisture, radi-ures include thermal shield flow-induced vibration problems, ation, and airborne particles and contaminants. SCC in guide tube support pins and core support structure bolts, faticx-induced core baffle water-jet impingement prob-

cess wear in flux thimbles. Many of the reported NUREG/CR-6034; OKLAHOMA SEISMIC NETWORK Final lems and $

Report. LUZA K.V.; LAWSON.J E. Oklahoma, Unrv. of, Norman, problems have been resolved by accepted engineenng prac-OK. July 1993. 42pp. 9308160178. 76122.314-tices. Uncertainties remain in the assessment of long term neu-The Nemaha uplift is composed of a number of crustal blocks tron irradiation effects and environmental factors in high-cycle tyrncally 3 to 5 miles (5 to 8 km) wide and 5 to 20 miles (8 to 32 fatigue failures. Reactor intemals are examined by visual in-km) long. In Oklahoma, several discontinuous uplifts, such as spections and the technique is access limited. Improved inspec-the Oklahoma City, Lovell, Garber, and Crescent uplifts, occur tion methods, especially one with an early failure detection ca-along the main axis of the Nemaha uplift. A statewide network pability, can enhance the safety and efficiency of reactor oper-of 12 stations records seismological data in Oklahoma. Six se-abons. mipermanent seismograph stations, four radiotetemetry statsons, the Oklahoma Geophysical Observatory's seismograph statioit NUREG/CR-6049: PIPING BENCHMARK PROBLEMS FOR THE and a borehole seismograph station at the Observatory com-GENERAL ELECTRIC ADVANCED BOILING WATER REAC-i prise the Oklahoma Geological Survey's seismic network. From TOR. BEZLER,P.; DEGRASSI,G.; BRAVERMAN J.; et al. Brook-January 1,1987, through December 31,1992,373 earthquakes haven National Laboratory. August 1993.178pp. 9309030213. were located by the Oklahoma seismic network. The distribution BNL-NUREG-52377. 76327:127* of the earthquakes by state is as follows: 315 in Oklahoma,28 To satisfy the need for venfication of the computer programs in Texas,14 in Kansas, 8 in Arkansas, 7 in Missouri, and 1 in Nebraska. Of the 352 earthquakes. 23 were reported fett. The and modeling techniques that will be used to perform the final earthquake epicentral data produces at least three seismic piping analyses for an advanced boiling water reactor standard trends. These trends are located in north-central Oklahoma, at design, three benchmark problems were developed. The prob-the eastem margin of the Anadarko basin, and in the Arkoma lems are representative piping systems subjected to representa-basin-Ouachita Mountains area tive dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard NUREG/CR-6043 V01: AGING ASSESSMENT OF ESSENTIAL design. It will be required that the combined license holders HVAC CHILLERS USED IN NUCLEAR POWER PLANTS. Phase demonstrate that their solutions to these problems are in agree-

1. BLAHNIK D E ; KLEIN.R.F. Battelle Memonal Inst:tute Pacific ment with the benchmark problem set.

Northwest Laboratory. September 1993.101pp. 9310120252. PNL-8614 76741:146. l

Main Citations and Abstracts 11 l NUREG/CR-6050: RADIATION EXPOSURE MONITORING AND cavity flooding. BWR drywell flooding, PWR depressurization, j INFORMATION TRANSMITTAL (REMIT) SYSTEM. User's and PWR feed and bleed. l Manual. CALE,R.; CLARK,T.; DIXSON,R.; et al. Science Apphca-tons intemational Corp. (formerfy Scence Applicatons, Inc.). NUREG/CR-6058: VIRGINIA REGIONAL SEISMIC June 1993. 300pg 9307220201. SAIC-93/1310-01. 75742:001. NETWORK. Final Report (1986 -1992). BOLLINGER,G.A.; The %r6ston Exposure Monitoring and Information Transmit-SIBOL,M.S.; CHAPMAN M.C.; et al. Virginia Polytechnic Institute tal (REMIT) system is designed to assist U.S. Nuclear Regula- & State Uruv., Blacksburg, VA July 1993.115pp. 9308160175. tory Commisson (NRC) licensees in meeting the reporting re-76122:084. quirements of the revised 10CFR20 and in agreement with the in 1986, the Virginia Regional Seisme Network was one of guidance contained in R.G. 8.7, Rev.1, instructions for Re-the few fully calibrated digital seismic networks in the United cording and Reporting Occupational Exposure Data." REMIT is States. Continued operation has resulted in the archival of sig-a personal computer (PC) based menu driven system that facili-nals from 2000+ local, regional and teleseismic sources. Seis-tates the manipulaton of data base files to record and report motectonic studies of the central Virginia seisme zone showed radiation exposure information. REMlT is designed to be user-the actmty in the western part to be related to a large antifor-fnendly and contains the full text of R.G. 8.7, Rev.1, on-line as mal structure while seismicity in the eastem porton is associal-well as context-sonsittve help throughout the program. The user ed spatially with dike swarms. The eastem Tennessee seismic can enter data directty from NRC Forms 4 or 5, REMIT allows zone extends over a 3OOx50 km area and is the result of a the user to view the individual's exposure in relaton to ragula-compressive stress field acting at the intersecton between two tory or administrative limits and alerts the user to exposures in large crustal blocks. Hydroseismicity, whch proposes a signife-excess of these limits. The system also proodes for the calcula-cant role for meteoric water in intraplate seismogenesis, found ton and summaton of dose from intakes and the determinaton support in the observaton of common cycicities between of the dose to the maximally exposed extremity for the monitor-streamflow and earthquake strain data. Seisme hazard studies ing year. REMIT can produce NRC Forms 4 and 5 in paper and have provided the following results: 1) Damage areas in the electronic format and can import / export data from ASCil and eastern United States are three to fwe times larger than those data base files. observed in the west; 2) Judged solely on the basis of cata-NUREG/CR-6052: METHODOLOGY FOR RELIABILITY BASED loged earthquake recurrence rates, the next major shock in the CONDITION ASSESSMENT. Appication To Concrete Structures southeast regica will probably occur outside the Charleston, in Nuclear Plants. MORI,Y.; ELL;NGWOOD.B. Johns Hopkins South Carolina, area; and 3) Investigatons yielded necessary Univ., Baltimore MD.

• Oak Ridge Natonal Laboratory. August hazard parameters (fur example, maximum magrutudes) for sev-1993.164pp. 9308200285. ORNLSUB93-SD684. 76159:048.

wal shes m the southeast Base to these WestWons was N Structures in nuclear power plants may be exposed to ag-development and maintenance of several seismological data bases. grossive environmental effects that cause their strength to de-crease over an extended penod of service. A major concern in NUREG/CR-6060: HYDROGEN MIXING STUDIES (HMS) AS-evaluating the L ntinued service of such structures is to ensure that in their current conditon they are able to withstand future SESSMENT MANUAL LA4K.L.: WILSON,T.L Los Alamos Na-extreme load events during the intended service life with a level tonal Laboratory. TRAVIS.J.R. Science Applications Intemation-of reliabihty suffcient for public safety. This report desenbes a al Corp. (formerly Science Applicatons, Inc.). June 1993. 94pp. methodology to facilitate quantitatue assessments of current 9308160049. LA-12593-M. 76121:148. i l end future structural reliability and performance of structures in This report documents some calculations performed to nuclear power plants. This methodology takes into account the assess the Hydrogen Mixing Studies (HMS) code. Results are nature of past and future loads, and randomness in strength presented first for some analytical test problems, including lam-End in degradation resulting from environmental factors. An inar flow and mass diffusion. The von Karman vortex street adaptwe Monte Carlo simulaton procedure is used to evaluate problem and the Sanda FLAME Facihty and Heiss Dampf Reak-time-dependent system rehabihty. The time-dependent reliability tor (HDR) containment facility test problems are then discussed. is sensittve to the time-varying load characteristes and to the For the analytical problems, the code gave results that agree choice of initial strength and strength degradation models but exceptionally well with the analytcal solutons Calculations for not to correlatirsn in component strengths within a system. In_ the von Karman vortex street problem were performed at se-spection/ maintenance strategies are identified that minimize the lected Reynolds numbers for several obstacle types. The com-expected future costs of keeping the failure probability of a puted flow pattems agree well with experimental observations-structure at or below an established target failure probabihty specifcally the occurrence of a vortex street (double row of vor-during its antcipated service period. tees) above a crincal Reynolds number. Calculatons for the von Karman vortex street problem were performed at selected NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF Reynolds numbers for several obstacle types. The computed SEVERE ACCIDENT MANAGEMENT STRATEGIES. flow pattems agree well with experimental observations-specifi-KASTENBERG,W E.; APOSTOLAKIS,G.; DHIR.V.K.: et al. Cali-cally the occurrence of a vortex street (double row of vortices) fornia, Univ. of. Los Angeles, CA. September 1993. 350pp. above a entcal Reynolds number The last assessment problem 9310130038. 76743.231. involves modeling the experiment T31.5. The expenment was Severe accident management can be defined as the use of carried out in the HDR containment building which is a large, existing and/or alternatwe resources, systems and actors to multi-compartment facility (11300 m(3) free volume in 72 com-prevent or mitigate a core-mett accident. For each accident se-partments). In the expenment, a steam-water mixture was first quence and each combination of severe accident management injected into the containment to simulate a large-break blow-strategies, there may be several options available to the opera-down of a pressure vessel, and then superheated steam was tor, and each involves phenomenological and operational con-injected that was followed by a release of helium-hydrogen light sideratons regarding uncertainty. Operatonal uncertainties in-gas. The calculated results (pressure, temperature, and gas ciude operator, systaTi and instrumentaton behavior during an concentratons) agree reasonably well with the experimental tecident. A framework based on decision trees and influence

data, diagrams has been developed which incorporates such criteria as feasibihty, effectiveness, and adverse effects, for evaluating MUREG/CR-6065: SYSTEMS ANALYSIS OF THE CANDU 3 RE-potential severe accident management strategies. The frame-ACTOR. WOLFGONG,J.R.: LINN.M.A.; WRIGHT,A.L: et r21. Oak work is also capable of propagating both data and model uncer-Ridge National Laboratory. July 1993. 334pp. 9308160298.

1:inty, it is apphed to several potential strategies including PWR ORNL/TM-12396. 76118:001.

I l l l l l 12 Main Citations and Abstracts This report presents the results of a systems failure analysis chanics methodologies are applied to investigate the crack 1 study of the CANDU 3 reactor design; the study was performed growth behavior observed in the hot leg of the model. These for the U S. Nuclear Regulatory Commission. As part of the are: the aK methodology (Paris law), AJ concepts and a re-study a review of the CANDU 3 design documentation was per-cently developed limit load stress-range entenon. The report in-l formed, a plant assessment methodology was developed, repre-ciudes a discussion on the pros and cons of the analysis in-l sentative plant initiating events were identified for detailed anal-volved in each of the methods, the role played by the key pr. ysis, and a plant assessment was performod. The results of the rameters influencing the formulation and a comparison of the olant assessment included classification of the CANDU 3 event results with the actual crack growth behavior observed in the vi-sequences that were analyzed, determination of CANDU 3 sys-bration test program. Some conclusions and recommendatons tems that are "significant to safety,*' and identification of key for improvement of the methodologies are also provided. operator actions for the analyzed events. NUREG/CR-6079: SEISMOLOGICAL INVESTIGATION OF NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUS-EARTHOUAKES IN THE NEW L1ADRID SEISMIC ZONE. Final TION BEHAVIOR OF HYDROGEN. AIR MIXTURES WITH TUR-Report. September 1986 - December 1992, HERRMANN,R.B.; BULENT JET IGNITION AT LARGE SCALE. DOROFEEV.S B.; NGUYEN.B. St. Louis Univ, St. Louis, MO. August 1993.75pp. BEZMELNITSIN,A.; EFIMENKO,A.A.; et al. Russia. June 1993. 9309030148. 76325 084. B3pp 9308160272. RRCKi-80-05/3. 76117:272. Earthquake actrvity in the New Madnd Saismic Zone had This report desenbes research carried out in the KOPER facil-been monitored by regional seismic networks since 1975. ity on spontaneous detonaten ignition in hydrogen-air mixtures Dunng this time penod, over 3700 earthquakes have been lo-by turbulent let ignitiorL The KOPER facility is a semi-confined cated within the region bounded by latitudes 35 degrees-39 de-volume of 47.7 m(3) and consists of a steel canyon and a grees N and longitudes 87 degrees-92 degrees W, Most of robust frame placed above it. The frame sides are sealed with these earthquakes occur within a 1.5 degree x 2 degree zone thin polyethylene sheet. A " jet" chamber of 0.55 m(3), located centered on the Missoun Bootheel. Source parameters of larger on the bottom of the canyon was used to produce 8 jet of hot earthquakes in the zone and in eastern North Amenca are de-gases, which was vented into the hydrogen-air mixture. The ef-termined using surface-wave spectral amplitudes and broad-fects of three vanables were investigated. hydrogen concentra-band waveforms for the purpose of determining the focal mech-tion (18-30% vol.); jet onf ce diameter (100-400 mm); and the anistr', source depth and seismic moment. Waveform modeling composition of combustion products in the turbulent jet (by of broadband data is shown to be a powerful tool in defining varying the hydrogen mole fracton in the jet"-chamber from 25 these source parameters when used complementary with re-to 50% vol) The possibility of initiation of turbulent combustion gional seismic network data, and in addition, in ventying the cor-and local detonation was demonstrated. Local detonation devel-rectness of previously published focal mechanism solutions. ops after a delay of 10-25 ms from ignition. For spontaneous e detonation initiaton, the minimum hydrogen concentraton is NUREG/CR-6081: ENHANCED REMOVAL OF RADIOACTIVE within the range of 20 to 25% vol., and the minimum jet onfice PARTICLES BY FLUOROCARBON SURFACTANT SOLU-diamete 1,es in the range of 100 to 200 mm for the KOPER fa-TIONS. KAISER,R.; HARLING.O.K. Affiliation Not Assigned. cihty. A minimum ratio of turbulent jet size L and mixture deto-August 1993. 97pp. 9309210192. 764B4:156. nation cell A. L/A = 12-13 is required for detonation initiation The proposed research addressed the application of ESI's which is supported by other type of turbulent jet initiation experi-particle removal process to the non-destructive decontamination ments (closed volume and continuous venting) and by theoreti-of nuclear equipment. The cleaning medium used in this proc-cal analysis. ess s a solution of a hegh molecular weight fluorocarbon surfac-tant in an inert perfluonnated hquid which results in enhanced NUREG/CR-6073: LYSIMETER LITERATURE REVIEW. ROGERS,R D.; MCCONNELL,J W. EG&G idaho, Inc. August particle removal The perfluonnated liquids of interest, which are 1993. 75pp. 9309210201. EGG-2706. 76484.257. recycled in the process, are nontoxic, nonfiammable, and envi-Many reports have been pubhshed concerning the use of lysi_ tonmentally compatible, and do not present a hazard to the ozone layer The information obtained in the Phase i program meters to obtain data on the performance of buned radioactive indicated that the proposed ESI process is technically effectwe waste. This document presents a review of some of those re. and economically attractwe. The fluorocarbon surfactant solu-ports. This review includes lysimeter studies using radioactive tons used as working media in the ESI process survived expo-waste forms at Savannah River Site, Hanford Site, Argonne Na_ sure of up to 10 Mrad doses of gamma rays, and are consid-tional Laboratory, and Oak Ridge National Laboratory; radiono, clide tracer studies at Whiteshell Nuclear Research Estabhsh. ered sufficiently radiaton resistant for the proposed process. UI-ment and Los Alamos National Laboratory; and water move. traconic cleaning in perfluonnated surfactant solutions was found to be an effective method of removing radioactive iron ment studies at the Nuclear Regulatory Commission's Bettsville, i (Fe 59) oxide particles from contaminated test pieces. Radcac-Maryland site. at the Hanford Site, and at New Mexico State University. The tests, results, and conclusions of each report live particles suspended in the process liquids could be quanti-tatively removed by filtration through a 0.1 um membrane filter. l are summarized, and conclusions concerning fysimeter technol. Projected economics indicate a pre tax pay back time of 1 ogy are presented from an overall analysis of the Lerature. month for a commercial scale system. NUREG/CR-6078: ANALYSIS OF CRACK IN!TIATION AND GROWTH IN THE HIGH LEVEL VfBRATION TEST AT TA-NUREG/CR-6082: DATA COMMUNICATIONS. PRECKSHOT G G. DOTSU. KASSIR,M.K.; PARK,Y.J., HOFMAYER.C.H.; et al. Lawrence Livermore National Laboratory. August 1993. 96pp. Brookhaven National Laboratory. August 1993. 82pp. 9309030171. UCRL-fD-114567. 76326:219. 9309210185. BNL-NUREG 52383. 76484:075. The purpose of this paper is to recommend regulatory guid-The High Level Vibration Test data are used to assess the ance for reviewers examining computer communication systems accuracy and usefulness of current engineering methodologies used in nuclear power plants. The recommendations cover for predicting crack initiation and growth in a cast stainless steel three areas important to these communications systems: pipe elbow under complex, large amplitude loading. The data system design, communication protocols, and communication were obtained by testing at room temperature a large scale media. The first area, system design, considers three aspects of modified model of one loop of a PWR pomary coolant system at system design -questions about archrtecture, specific risky the Tadotsu Engineenng Laboratory in Japan. Fatigue crack ini-design elements or omissions to look for in designs being re-tiation time is reasonably predicted by applying a modifet,d local viewed, and recommendations for multiplexed data communica-strain approach (Coffin-Mason-Goodman equation) in conjunc-tion systems used in safety systems. The second area reviews tion with Miner's rule of cumulative damage. Three fracture me-pertinent aspects of communicaton protocol design and makes r

Main Citations and Abstracts 13 recommendations for newly designed protocols or the selecton about the PLC manufacturing organtraton and the protection of existing protocols for safety system, information d: splay. and system engineenng organization. Supplementing this document non-safety control system use. The third area covers communi-are two appendices. Appendix A summanzes PLC characteris-cation media selecton, which differs significantly from traditional tscs. Specifically addressed are those charactenstics that make wire and cable. The recommendations for communication media the PLC more suitable for emergency shutdown systems than extend or enhance tne concems of published IEEE standards other electncal/ electronic-based systems, as well as character-about three subjects: data rate, imported hazards and maintain-istics that improve rehability of a system. Also covered are PLC ability. charactenstics that may create an unsafe operating environ-NUREG/CR-6083: REVIEWING REAL-TIME PERFORMANCE OF ment. Appendix B provides an overview of the use of program-NUCLEAR REACTOR SAFETY SYSTEMS. PRFCKSHOT.G G. mable logic controllers in emergency shutdown systems. The intent is to familianze the reader with the design, development, Lawrence Livermore National Laboratory. August 1993. 88pp. test, and maintenance phases of applying a PLC to an ESD 9309030112. UCRL-ID-114565. 76325.001 system. Each phase is desenbed in detail and information perte The purpose of this paper is to recommend regulatory guid-nent to the apphcaton of a PLC is pointed out. ance for reviewers examining real-time performance of comput-er-based safety systems used in nuclear power plants. Three NUREG/GR-0005 V02 P1: RISK-BASED INSPECTION-DEVELOP-areas of guidance are covered in this report. The first area MENT OF GUIDELINES. Light Water Reactor (LWR) Nuclear covers how to determine if, when, and what prototypes should Power Plant Compr'nents.

• Amencan Society of Mechanical be required of developers to make a convincing demonstraton Engineers. Jufy 1993.173pp. 9308200281. CRTD-VOL20-2.

that specific problems have been sotved or that performance 76159:209. goals have been met. The second area has recommendations Effective inservice inspect on programs can play a significant for timann analyses that will prove that the real-time system will role in minimizing equipment and structural failures. Most of the meet its safety-imposed deadlines. The third area has descrip-current inservice inspection programs for light water reactor tions of means for assessing expected or actual real-time per-(LWR) nuclear power plant components are based on experi-formance before, during, and after development is completed. ence and engineers' qualitatrve judgment. These programs in-To ensure that the delivered real-time software product meets clude only an implicit consideration of nsk, which combines the l performance goals, the paper recommends certain types of probability of failure of a component under its operation and j code-execution and communications scheduling. Technical loading conditons and the consequences of such failure, if it background is provided in the appendix on methods of timing occurs. This document recommends appropriate methods for analysis, scheduling real-time computatons, prototyping, real' establishing a nsk-based inspection program for LWR nuclear time software development approaches, modeling and measure-power plant components. The process has been built from a ment, and real-time operating systems. general methodology (Volume 1) and has been expanded to in-NUREG/CR 6085: UNITED STATES SEISMOGRAPHIC NET. volve five major steps: defining the system; evaluating fr/alita-WORK. BULAND,R. Interior, Dept. of, Geological Survey. Sep. tive risk assessment results; using Ws and information from tember 1993. 83pp. 9310120339. 76741:028. plant probabihstic risk assessments to perform a quantitative The concept of a United States National Seismograph Net. nsk anatysis; selecting target failure probabilities; and develop-work (USNSN) dates back nearty 30 years. The idea was re. ing an inspection program for components using economic deci-vrved several times over the decades, but never funded. For ex. sion analysis and structural reliability assessment methods. In-1 ample, a national network was proposed and discussed at great ciuded: extensive bibhography. Companion Volume 2 - Part 2 ] length in the so called Bolt Report (U.S. Earthquake Observa-document will recommend nsk-based inspection program for j tones: Recommendations for a New National Network, Natonal consideration by Section XI of the ASME Boiler and Pressure Academy Press, Washington, D.C., 1980,122 pp). From the be. VesselCode. l l ginning, a national network was viewed as augmenting and I complementing the relatrvely dense, predominantly short-period NUREG/lA-0085: ASSESSMENT OF FULL POWER TURBINE I vertical coverage of selected areas provided by the Regional TRIP START-UP TEST FOR C. TRILLO I WITH RELAPS/ I Seismograph Networks (RSN's) with a sparse, well-distributed MOD 2. LOZANO M.F.; MORENO.P.: DE LA CAL C.: et at Spain. l network of three-component, observatory quality, permanent Govt. of. July 1993. 73pp. 9308160158. ICSP-TR-TTRIP-R. stations. The opportunity finally to begin developing a national 70110 2I0-network arose in 1986 wrth discussions between the U.S. Geo. C. Trillo I has developed a model of the plant with RELAPS/ l logical Survey (USGS) and the Nuclear Regulatory Commission MOD 2/36.04. This model will be validated against a selected (NRC). Under the agreement signed in 1987, the NRC has pro-set of start-up tests. One of the transients selected to that aim vided $5 M in new funding for capital equipment (over the is the turbine inp, which presents very specific characteristics penod 1987-1992) and the USGS has provided personnel and that make it significantly different from the same transient in facihties to develop, deploy, and operate the network. Because other PWRs of different design, the main difference being that the NRC funding was earmarked for the eastern United States, the reactor is not tripped: a reduction in pnmary power is camed new USNSN station deployments are mostly east of 105 de. out instead. Pre-test calculations were done of the Turbine Tnp grees West longitude while the network in the westem United Test and compared against the actual test. Minor problems in States is mostly made up of cooperating stations (stations the first model, specially in the Control and Limitation Systems, meeting USNSN design goals, but deployed and operated by were identified and post-test calculations had been camed out. other instrtutions which provide a logical extension to the The results show a good agreement with data for all the com-USNSN). pared vanables. NUREG/CR-6090: THE PROGRAMMABLE LOGIC CONTROLLER NUREG/lA-0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A AND ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS. NATURAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PALOMAR J.; WYMAN,R. Lawrence Livermore National Labora-PLANT BORSSELE. WINTERS.L. Netherlands, Govt. of. July tory September 1993. 96pp. 9310120250. UCRL-ID-112900. 1993. 66pp. 9308160307. ECN-89-91. 76122:247. 76740:170. As part of the ICAP (Internatonal Code Assessment and Ap-This document provides recommendations to guide reviewers phcatons Program) agreement between ECN (Netherlands in the application of Programmable Logic Controllers (PLCs) to Energy Research Foundation) and USNRC, ECN has performed the control, monitonng and protection of nuclear reactors. The a number of assessment calculations for the thermohydraulic first topics addressed are system-level design issues, specifical-system analysis code RELAPS/ MOD 2/36.05. This document ly meluding safety. The document then discusses concems describes the assessment of this computer program versus a

l l 14 Main Citations and Abstracts natural circulaton expenment as conducted at the Borssele Nu-NUREG/lA-0112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST clear Power Plant. The results of this companson show that the ECN REFLOOD EXPERIMENTS WOUDSTRA.A.; code RELAPS/ MOD 2 predicts well the natural circulation be-VANDEBOGAARD.J.; STOOP,P.M. Netherlands. Govt. of. July havior of Nuclear Power Plant Borssele. The work has been 1993. 87pp. 9308160054. ECN-C-92 008. 76121:243. sponsored by the Dutch Licensing Authonty and ECN. As part of the ICAP (Intemational Code Assessment and Ap-placations Program) agreement between ECN (Netherlands ( NUREG/lA-0096: NUMERICS AND IMPLEMENTATION OF THE Energy Research Foundation) and USNRC, ECN has performed UK HORIZONTAL STRATIFICATION ENTRAINMENT OFF-a number of assessment calculations with the computer pro-TAKE MODEL INTO RELAPS/ MOD 3. BRYCE,W.M. Winfnth gram RELAP5. This report describes the fesults as obtained by Technology Centre (United Kingdom). June 1993. 47pp. ECN from the assessment of the thermohydraulic computer pro-9307220307. AEA-TRS-1050. 75743:253. gram RELAP5/ MOD 2/CY 36.05 versus a senes of reflood ex-This report presents the numencs and implementation details penments in a bundle geometry. A total number of seven se-to add the same improved discharge qualrty correlations into lected expenments have been analyzed, from the reflood exper-RELAPS/ MOD 3. In the light of expenence with the modifed 6 mental program as previously conducted by ECN under con. tract of the Cornmission of the European Communitics (CLC). In RELAPS/ MOD 2 code, some of the numencs have been slightly this document, the results of the analyses are presented and a changed for RELAPS/ MOD 3. The description is quite detailed in companson with the expe smental data is provided. order to facihtale change by some future code debeloper. A simple test calculation was performed to confirm the coding of NUREG/lA-0113: PRELIMINARY ASSESSMENT OF PWR the correlations implemented in RELAP5/ MOD 3. STEAM GENERATOR MODELLtNG IN RELAPS/ MOD 3. PREECE R.J.; PUTNEY,J.M. National Power (United Kingdom). NUREG/lA-0100: ASSESSMENT OF CCFL MODEL OF RELAPS/ July 1993. 26pp. 9309090021. 76381:154. MOD 3 AGAINST SIMPLE VERTICAL TUBES AND ROD A preliminary assessment of Steam Generator (SG) modelling BUNDLE TESTS. CHO.S.; ARNE,N. Korea Electnc Power Corp. in the PWR thermal-hydrauhc code RELAPS/ MOD 3 is present-CHUNG.B-D.; et al. Korea institute of Nuclear Safety. June ed. The study is based on calculations against a series of 1993.122pp. 9307220299. 75744:001. steady-state commissioning tests camed out on the Wolf Creek The CCFL model used in RELAPS/ MOD 3 version Sm5 has PWR over a range of load conditions. Data from the tests are been assessed against simple vertical tubes end rod bundle used to assess the modelling of pnmary to secondary side heat tests performed at a facility of Korea Atomic Energy Research transfer and, in particular, to examine the eftect of reverting to institute. The effect of changes in tube diameter and nodaliza-the standard form of the Chen heat transfer correlation in place lion of tube section were investigated. The roles of interfacial of the modified form applied rn RELAPS/ MOD 2. Compansons drags on the flooding charactenstics are discussed. Differences between the two versions of the code are also used to show be' ween the calculation and the expenment are also discussed, how the new interphase drag model in RELAP5/ MOD 3 affects A comparison between model assessment results and the test the calculation of SG liquid inventory and the void fraction pro-data showed that the calculated value lay well on the expen-file in the riser. mental flooding curve specified by user, but the pressure jurnp before onset of flooding was not calculated. NUREG/lA-0126: 2D/3D PROGRAM WORK

SUMMARY

REPORT. DAMERELL P.S.: SIMONS,J.W. MPR Associates. Inc. NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 9.1.B USING

• et al. Japan Atomic Energy Research Institute. June 1993.

RELAPS/ MOD 3. LEE.S.; CHUNG.B-Da KIM.H-J. Korea institute 400pp. 9307220220 GRS.100. 75745.042. of Nuclear Safety. June 1993. 250pp. 9307220272. 75743:001. The 2D/3D Prwam was camed out by Germany, Japan and i 2" cold leg break test 9.1.b, conducted at the BETHSY facility the United Staw to investigate the thermal-hydrauhes of a t was analyzed amno the RELAPS/ MOD 3 Version Sm5 code. The PWR large-bree JCA. A contributory approach was utilized m ~ test 9.1.b was conducted with the main objective being the in. which each cc( - contributed significant effort to the program vestigation of the thermal-hydraulic mechanisms responsible for and all three s uuntnes shared the research results. Germany the large core uncovery and fuel heat-up, requinng the imple. constructed and operated the Upper Plenum Test Facility mentation of an ultimate procedure. The present analysis dem-(UPTF), and Japan constructed and operated the Cylindrical onstrates the code's capability to predict, with sufficient accura-Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced cy, the main pheno-nena occumng in the depressurization tran-instrumentation to each of the three test facilitics, and assess-sient, both from a qualitative and quantitative point of view. ment of the TRAC computer code against the test results. Eval-Nevertheless, several differences regarding the evolution of uations of the test results were camed out in all three countries. j phenomena and affecting the timing order have to be pointed Ms mpod summanzes me 2M Wam n Wms of N ce out in the base calculaton. Three calculations were carried out inbuting efforts of the participants. to study the sensitivity to change of the nodalization in the com-ponents of the loop seal cross-over legs, and of the auxiliary NUREG/lA-0127: REACTOR SAFETY ISSUES RESOLVED BY ~ feedwater control logics, and of the break discharge coefficient. THE 2D/3D PROGRAM. DAMERELL.P.S; SIMONS.J.W. MPR j [*hp9 NUREG/lA-0107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 8160173 GR 0 76116 1 LOAD HEJECTION FROM 100% TO 50% POWER IN THE The 2D/3D Program studied multidimensional thermal-hydrau-l VANDELLOS ll NUCLEAR POWER PLANT. LLOPIS,C.; lics in a PWR core and primary system during tne end-of-blow-l MENDlZABAL,R.; PEREZ,J. Spain Govt. of. June 1993. 53pp. down and post-blowdown phases of a large-break L.OCA l l 9307220264. ICSP-V2-R50-R. 75742:241. (LBLOCA), and dunng selected smali-break LOCA (SBLOCA) l An assessrnent of RELAPS/ MOD 2 cycle 36.04 against a load transients. The program included tests at the Cylindrical Core rejection from 100% to 50% power in Vandellos il NPP (Spain) Test Facility (CCTF), the Stab Core Test Facility (SCTF), and is presented. The work is inscribed in the framework of the the Upper Plenum Test Facility (UPTF). and computer analyses Spanish contribution to ICAP Project. The model used in the using TRAC. Tests at CCTF investigated core thermal-hydrau-simulation consists of a single loop, a steam generator and a lics and overall system behavior while tests at SCTF concen-i steam line up to the steam header, all of them enlarged on a trated on multidimensional core thermal-hydraubes. The UPTF j scale of 3:1; and full-scaled reactor vessel and pressun2er. The tests investigated two-phase flow behavior in the downcomer, i results of the calculations have been in reasonable agreement upper plenum, tie plate region, and pnmary loops. TRAC analy-with plant measurements. ses evaluated thermal-hydraulic behavior throughout the pnmary

Main Citations and Abstracts 15 system in tests as well as in PWRs. This report summanzes the discussion is organtzed in terms of the reactor safety issues in-test and analysis results in each of the main areas where im-vestigated. proved information was obtained in the 2D/3D Program. The i J l l a [ L i

I a Secondary Report Number index This index lists, in alphabetical order, the performing organization issued report codes for the l NRC contractor and international agreement reports in this compilation. Each code is cross-i referenced to the NUREG number for the report and to the 10-digit NRC Document Control l System accession number. e i SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER ~ CEA-TRS-1050 NUREG/lA 0096 LA 12593-M. NURE j/CR4060 i AEEW.R2501 NUREG/lA-0096 MPR 1345 NURE m/lA-0126 ANL-93/11 NUREG/CR4744 V07 N2 MPR 1346 NUD'A/lA 0127 ANL-93/27 NUREG/CR-4667 V18 ORNL4566/V1

• UREG/CR-5404 V02 BHARC700/93/029 NUREG/CR-5758 V03 OR L/ M 11876 BNL-NUREG-51708 NUREG/CR 3469 Vn7 NL/TM-11945 NUREG/CR-5782 BNL-NUREG-52346 NUREG/CR-5943 ORNL/TM-12229 NUREG/CR-5942 BNL-NUREG42362 NUREG/CR 5993 VD1 ORNL/TM 12340 NUREG/CR4023 BNL-NUREG-52362 NUREG/CR-5993 V02 ORNL/TM-12342 NUREG/CR4015 BNL-NUREG-52377 NUREG/CR4049 OANL/TM 12371 -

NUREG/CR-6048 BNL-NUREG-52383 NUREG/CR4078 ORNL/TM-12396 NUREG/CR4065 CONSWCSMECR1792 NUREG/CR-5969 ORNLSUB93 SD684 NUREG/CR4052 CDNSWCSMECR1892 NUREG/CR-5970 PNL 7905 NUREG/CR-5829 l CNWRA 92406 NUREG/CR4026 PNL 7907 NUREG/CR-5833 CNWRA 92 02S NUREG/CR-5817 V03 N2 PNL-8467 NUREG/CR-5989 CNWRA92411 NUREG/CR4021 PNL-8594 NUREG/CR4029 V01 PNL-8614 NUREG/CR4043 V01 CONF-9020823 NUREG/CP 0130 V01 Kf40-05/3 CONF-9020823 NUREG/CP4130 V02 2 SAIC-93/131041 NUREG/CR-6050 'l CRTD-VOL20-2 NUREG/GR4005 V02 P1 ECN49-91 NUREG/lA 0091 SAND 91-2802 NUREG/CR-5927 V01 ECN C-92-008 NUREG/lA-0112 UCRL-ID 112900 NUREG/CR4090 i EGG-2685 NUREG/CR-5928 UCRL ID-114565 NUREG/CR-6083 EGG 2706 NUREG/CR4073 UCRL-ID 114567 NUREG/CR4082 GAS-100 NUREG/iA4126 UILU-ENG91-2013 NUREG/CR-5969 GR5-101 NUREG/lA4127 UlLU-ENG92-2014 NUREG/CR-5971 BCSP-TR-TTRIP-R NUREG/lA4085 UILU-ENG92-2016 NUREG/CR-5970 ICSP-V2-R50-R NUREG/lA4107 VARGOS-93/1 NUREG/CR4072 I l 1 1 I i 17 l i

Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by the author. If further information is needed, refer to the main cita-tion by the NUREG number. ABABOU,R. BEZLER,P. NUREG/CR-5817 V03 N2: NRC HIGH LEVEL RADIOACTIVE WASTE NUREG/CR-6049. PIPING BENCHMARK PROBRMS FOR THE GEN-RESEARCH AT CNWRA. July December 1992. EF.AL ELECTRIC ADVANCED BOlWNG WATER REACTOR. ABU-EID.R. BEZMELNITSIN,A. NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-CONSTRUCT AND OPERATE A FACluTY TO RECE!VF STORE. AND HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR NITION AT LARGE SCALE' CUVE. UTAH. Docket No. 40-8989.Envrocare Of Utah,Inc. AHOLA.M.P. BLAHNIK,D.E. NUREG/CR-5817 V03 N2-NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/CR-6043 V01. AGING ASSESSMENT OF ESSENTIAL HVAC RESEARCH AT CNWRA July-December 1992. CHILLERS USED IN NUCLEAR POWER PLANTS. Phase 1. NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL-HYDROLOGIC. MECHAN ICAL-CHEMICAL PROCESSES PERTINENT BOLLINGER,G.A. TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA NUREG/CR-6058: VIRGINIA REGIONAL SEISMIC NETWORK. Foal MOUNTAIN Report (1986 -1992). ALLENSPACH.F. BRAVERMAN,J. ATA UMM OF THE SYSTEMAT-NUREG/CR-6049 PIPING BENCHMARK PROBLEMS FOR THE GEN-C A SE M T F PE ERAL ELECTRIC ADVANCED BOluNG WATER REACTOR-ANDERSON,T L NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT BWMMETT,E. OF CONSTRAINT IN FRACTURE. NUREG 1476: FINAL' ENVIRONMENTAL IMPACT SIATEMENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE AND APOSTOLAKIS.G. DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF CLIVE. UTAH. Docket No. 40-8989,Ermrocare Of Utah.Inc. SEVERE ACCIDENT MANAGEMENT STRATEGIES. ARMBRUSTER.J[78 V03: NEW YORK /NEW JERSEY REGIONAL SEISMIC ^ NUREG/CR-5 HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE MODEL NETWORK. Foal Report For April 1985 - September 1992. INTO RELAPS/ MOD 3. ARNE.N. NUREG/lA-0100 ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 BRYSON,J.W. AGAINST CMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILtSTIC FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE ARSENAULT.F. ROWE REACTOR PRESSURE VESSEL NUREG/CP-0129: PROCEEDINGS OF THE WORKSHOP ON PROGRAM FOR ELIMINATION OF REOUIREMENTS MARGINAL TO SAFETY. BULAND,R. NUREG/CR 6085: UNITED STATES SEISMOGRAPHIC NETWORK. NUREG/CR.5033 V01: METHODS FOR DEPENDENCY ESTIMATION CALE,R. AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE NUREG/CR4050: RADIATION EXPOSURE MONITORING AND INFOR-NURE / 5 VO : T OR DEPENDENCY ESTIMATION MA ON MANSMM (REW SYSM M Wnual AND SYSTEM UNAVAILABluTY EVALUATION BASED ON FAILURE CARBAJO,JJ. DATA STATISTICS. Detailed Desenption And Applicatons. NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARACTER-BACA,R.G. ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED NUREG/CR-5817 V03 N2: NRC HiGH-LEVEL RADIOAf3vE WASTE BY THE MELCOR CODE. ' RESEARCH AT CNWRA. July-December N. BAGTZOGLOU.A.C. NUREG/CR-5944: A CHARACTERIZATION OF CHECK VALVE DEGRA. NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-RESEARCH AT CNWRA. July-December 1992. DUSTRY. BANDYOPADHYAY,K CHAPMAN,M.C. NUREG/CR 6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN NUREG/CR-6058: VIRGINIA REGIONAL SEISMIC NETWORKFmal THE HIGH LEVEL VfBRATION TEST AT TADOTSU. Report 0986 -1994 BASS.B.R NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC CHEYERTON,R.D. FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC ROWE REACTOR PRESSURE VESSEL FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE - ROWE REACTOR PRESSURE VESSEL BAUM.J.W. NUREG/CR 3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU. CHO,S. CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT. NUREG/lA-0100: ASSESSMENT OF CCFL MODEL OF RELAP5/ MOD 3 ED READINGS IN RADIATION PROTECTION AND ALARA. AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. 19

20 Personal Author index CHOPRA.O L DODGE F.T. NUREG/CH-4667 V16. ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE LIGHT WATER REACTORS. Semiannual Report. October 1992 - March RESEARCH AT CNWRAJulyCecember 1992. 1993 NUREG/CR4026. THEORETICAL AND EXPERIMENTAL INVESTIGA-NUREG/CR-4744 V07 N2: LONG TERM EMBRITTLEMENT OF CAST TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual SATURATED, FRACTURED POROUS MEDIUM. Report. April-September 1992. DOROFEEY,S.B. CHOWDHUR Y,A'H NUREG/CR-6072: EXPERIMEMAL STUDY ON THE COMBUSTION BE-NUREG/CR-581h V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-RESEARCH AT CNWRAJupDecember 1992 NUREG/CR4021: A LITERA.URE REVIEW OF COUPLED THERMAL. NITION AT LARGE SCALE. HYDROLOGIC MECHAN ICAL CHEMICAL PROCESSES PERTINENT TO T PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCP g $CR 5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTRY. Annual Summary Of Prog *am Performance Reports.CY CHUNG.5-D. 1992. NUREG/lA 0100: ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS EFIMENKO.A.A. NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 9.1.B USING NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-RELAPS/ MOD 3. HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-NITION AT LARGE SCALE. CHUNG,H.M. NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN EMGWOODA UGHT WATER REACTORS. Semiannual Report.Oc*,cbor 1992 March NUREG/CR4052: METHODOLOGY FOR RELIABILITY BASED CONDI-TION ASSESSMENT. Applicatum To Concrete Structures in Nuclear CICOTTE.G.R. Plants. NUREG-1400: AIR SAMPLING IN THE WORKPLACEfinal Report. ELLISON,P.G. CLARK,T. NUREG/CR-5928. ISLOCA RESEARCH PROGRAMfanal Report. NUREG/CR 6050: RADIATION EXPOSURE MONITORING AND INFOR. MATION TRANSMITTAL (REMIT) SYSTEM. User's Manual. EVANS.D.D. NUREG/CP-0040: PROCEEDINGS OF WORKSHOP V: FLOW AND TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE- -5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSAL. Held At RESEARCH AT CNWRAJuly-December 1992. Radisson Suite Hotel, Tucson, AnzonaJanuary 710,1991. COTTER.S.P. NUREG-1363 V05: ATOMIC SAFETY AND LICENSING BOARD PANEL FADDEN.M. ANNUAL REPORTJescal Year 1992. NUREG-0525 V02 R01: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL) January 1,1990 Through December 31,1992. NUREGICR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE FIELD,0. RESEARCH AT CNWRAJuly-December 1992. NUREG/CR-5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTRY. Annual Summary Of Program Performance Reports.CY DAMERELL,P.S. W2. NUREG/iA-0126 2D/3D PROGRAM WORK

SUMMARY

REPORT. NUREG/lA-0127: REACTOR SAFETY ISSUES RESOLVED BY THE 2D/ FIRSTEW. 3D PROGRAM. NUREG/CP 0130 V01: PROCEEDINGS OF THE 22ND DOE /NRC NU-DE LA CAL.C. CLEAR AIR CLEANING CONFERENCE.Gessions 1-8, Held in NUREG/lA 0085: ASSESSMENT OF FULL POWER TURBINE TRIP Derwer. Colorado, August 24-27,1992. START-UP TEST FOR C. TR!LLO f WITH RELAPS/ MOD 2. NUREGICP-0130 V02: PROCEEDINGS OF THE 22ND DOE /NRC NU-CLEAR AIR CLEANING CONFERENCE.Sessaons 916. Held in N RE -6049: PIPING BENCHMARK PROBLEMS FOR THE GEN-ERAL ELECTRIC ADVANCED BOILING WATER REACTOR. FLEMlNG T. NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTRY. Annual Summary Of Program Periremance Reports.CY N EG/CP-0129. PROCEEDINGS OF THE WORKSHCP ON PROGRAM . FOR EUMINATION OF REQUIREMENTS MARGINAL TO SAFETY. FONTANA,MR DHIR.V.K. NUREG/CR4056. A FRAMEWORK FOR THE ASSESSMENT OF NUREG/CR4065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR. SEVERE ACCIDENT MANAGEMENT STRATEGIES. FORSLUND,C. DICKSON T.L NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/CR-5782 PRESSURIZED THERMAL SHOCK PROBABILISTIC INDUSTRY. Annual Summary Of Program Periormance Reports.CY FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE 1992. ROWE REACTOR PRESSURE VESSEL NUREG/CR4023: GENERIC ANALYSIS FOR EVALUATION OF LOW FOK.R.A. CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS NUREG/CR-5DB9: PERFORMANCE TESTING OF EXTREMITY DOSI-AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERl-METERS--PILOT TEST. ALS. GAAL,M. MSON,R. NUREG/CP-0129: PROCEEDINGS OF THE WORKSHOP ON PROGRAM - NUREG/CR4050: RADtATION EXPOSURE MONITORING AND INFOR-FOR EUMINATION OF REQUIREMEICS MARGINAL TO SAFETY. MATON TRANSMITTAL (REMIT) SYSTEM. user's Manual GALYEAN,W.J. DODDS.RR NUREG/CR-5969: J AND CTOD ESTIMATON EQUATIONS FOR SHAL, NUREG/CR-5928: ISLOCA RESEARCH PROGRAMfinal Report LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS NUREG/CR-5970: APPROXIMATE TECHNIQUES FOR PREDICTING GARNER,LW, Si2E EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS UC) NUREG/CR 5833: AUXIUARY FEEDWATER SYSTEM RISK BASED IN. NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT SPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER - OF CONSTRAINT IN FRACTURE. PLANT. e

Personal Author index 21 GHOSH.A. JOHNSON,T.L NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADCACTfvE WASTE NUREG 1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO RESEARCH AT CNWRA. July-December 1992. CONSTRUCT AND OPERATE A FACILITY TO RECEIVE STORE, AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR R G CR 5829: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER JOYCE.J.A. PLANT. NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE MA-NUREG/CR-5833: AUXILtARY FEEDWATER SYSTEM RISK-BASED IN-CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL SPECTION GUOE FOR THE H.B. ROBINSON NUCLEAR POWER STRUCTURAL ALLOYS. PLANT. KAISER,R. NUREG/CR 6081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-NUREG/CR-5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER CLES BY FLUOROCARBON SURFACTANT SOLUTIONS. INDUSTRY. Annual Summary Of Program Performance Reports.CY 1992-KASSIR M.K. GREEN.R.T NUREG/CR 6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN NUREG/CR 5817 V03 N2: NRC HIGH-LEVEL RADICACTIVE WASTE THE HIGH LEVEL VIBRATION TEST AT TADOTSU. RESEARCH AT CNWRAJuly-December 1992. NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-KASSNER,T F TON OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN SATURATED. FRACTURED POROUS MEDIUM. LIGHT WATER REACTORS. Semannual Report. October 1992 - March 1993. HAGEMEYER.D. NUREG4713 V13: OCCUPATIONAL RADIATION EXPOSURE AT COM-KASTENBERG,W.E. MERCIAL NUCLEAR POWER REACTORS AND OTHER NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF FACILITIES.1991.Two ourth Annual Report SEVERE ACCIDENT MANAGEMENT STRATEGIES. NUREG/CR.6050: RADI ION EXPOSURE MONITORING AND INFOR-MATION TRANSMITTAL (REMIT) SYSTEM. User's Manual. KAURM.D HAMDAN,L CLEAR POWER PLANTS: ANNOTATED BIBLOGRAPHY ".# SELEt,5 NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO ED READINGS IN RADIATION PROTECTION AND ALARA. CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE. AND DISPO9E OF 11E.(2) BYPRODUCT MATERIAL NEAR KEENEY,J.A. CLIVE, UTAH. Docket No.40-8989 Ermrocare Of Utah,Inc. NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR

• ANMT HARLING,0.K.

ROWE REACTOR PRESSURE VESSEL NUREG/CR4081: ENHANCED REMOVAL OF RADtOACTIVE PARTI-CLES BY FLUOROCARBON SURFACTANT SOLUTIONS, KELLY,D.L NUREG/CR-5928: ISLOCA RESEARCH PROGRAM Foal Repop, NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMfTY DOSI. KHAN.T.A. METERS--PILOT TEST. NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT. NO-CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-HERRMANN.R.B. ED READINGS IN RADIATION PROTECTION AND ALARA. NUREG/CR-6079: SEfSMOLOGICAL INVESTIGATION OF EARTH-OUAKES IN THE NEW MADRfD SEISMIC ZONEFmal KIM,H-J. Report. September 1986 - December 1992. NUREG/lA-0100- ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 HICKEY.E.E. AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. NUREG/LA-0103: ASSESSMENT OF BETHSY TEST 9.1.B USING NUREG 1400: AIR SAMPLING IN THE WORKPLACEFnal Report. RELAPS/ MOD 3. HILL.B.H. NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE KIRK,M.T' CR-5969: J AND CTOD ESTIMATON EQUATONS FOR SHAL-NUREG/ RESEARCH AT CNWRA. July-December 1992. LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. ' HOFMAYER.C.H. NUREG/CR-5970 APPROXIMATE TECHNOUES FOR PREDICTING NUREG/CR-6078: ANALYSIS OF CRACK INITIATON AND GROWTH IN SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC). THE HIGH LEVEL VfBRATION TEST AT TADOTSU-KLEIN RI. HOPKINS,J.B. NUREG/CR-6043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC NUREG/CR-5829: AUXlLIARt FEEDWATER SYSTEM RISK-BASED IN-CHILLERS USED IN NUCLEAR POWER PLANTS. Phase L CTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER KOCHURKO A.S. NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-HSIUNG,S.M. HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE NITION AT LARGE SCALE. RESEARCH AT CNWRA. July December 1992. g HSU.F. NUREG/CR-5927 V01: EVALUATON OF A PERFORMANCE ASSESS-NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION MENT METHODOLOGY FOR LOW-LEVEL RADOACTIVE WASTE AND SYSTEM UNAVAILABILITY EVALUATON BASED ON FAILURE DISPOSAL FACILITIES.Evaluat on Of Modehn9 Approaches. DATA STATISTICS Summary Report NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATON KUECK,J.D. AND SYSTEM UNAVAILABIUTY EVALUATON BASED ON FAILURE NUREG/CR-5404 V02: AUXILIARY FEEDWATER SYSTEM AGING DATA STATISTICS. Detailed Desenpton And Apphcatons. STUDY. Phase 1 Follow-On Study. JAE,M. LAM,K.L NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF NUREG/CR-6060: HYDROGEN MIXING STUDIES (HMS) ASSESSMENT SEVERE ACCIDENT MANAGEMENT STRATEGIES. MANUAL-JOHNSON,D. LARREA.E. NUREG/CR-5778 V03: NEW YORK /NEW JERSEY REGONAL SEISMIC NUREG/lA-0085: ASSESSMENT OF FULL POWER TURBINE TRIP NETWORK. Foal Report For April 1985 - September 1992. START-UP TEST FOR C. TRILLO 1 WITH RELAPS/ MOD 2.

22 Personal Author index LAWSON.J.E. MCDONALD,J.C. NUREG/CR-6034: OKLAHOMA SEISMIC NETWORKTinal Report NUREG/CR 5989 FERFORMANCE TESTING OF EXTREMITY DOSI-METERS--PILOT TEST. LE E.S. NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 91.8 USING MCGUIRE,S.A. RELAPS/ MOD 3. NUREG-1400: AIR SAMPLING IN THE WORKPLACE Final Report LESLIE,B W. MCNAMARA,N. NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG OR37 V13 NO2. NRC TLD DIRECT RADIATION MONITORING RESEARCH AT CNWRA. July-December 1992. NETWORK. Progress Report Aprd-June 1993 LIM.H. MENDlZABAL.R. NUREG/CR4056: A FRAMEWORK FOR THE ASSESSMENT OF NUREG/lA-0107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A LOAD e SWERE CCIDENT MANAGEMENT STRATEGIES. REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS 11 NUCLEAR POWER PLANT. NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE MA. MICHAUD,W.F. CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN STRUCTURAL ALLOYS. LIGHT WATER REACTORS Semiannual Report. October 1992 - March LINN.M.A. NUREG/CR 6065. SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR. MillCt.T. NUREG/CR 6056. A FRAMEWORK FOR THE ASSESSMENT OF UE /lA-0107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A LOAD REJECTION FROM 100% TO $0% POWER IN THE VANDELLOS 11 MOFFITT,N.E. NUCLEAR POWER PLANT. NUREG/CR-5829. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-LLOYD,R.C. p q NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR-5833. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-SPECTION GUIDE FOR THE H B. ROBINSON NUCLEAR POWER SPECTION GUIDE FOR THE H B, ROBINSON NUCLEAR POWER PLANT. PLANT. W Z,A. MOFFITT'R NUREG/lA-0085: ASSESSMENT OF FULL POWER TURBlNE TRIP NUREG/CR-5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER START-UP TEST FOR C. TRILLO I WITH RELAPS/ MOD 2. INDUSTRY. Annual Summary Of Program Performance Reports.CY LOPEZ.E. 1992. NUREG/lA 0085 ASSESSMENT OF FULL POWER TURBINE TRIP MOR M P START-UP TEST FOR C. TRILLO I WITH RELAP5/ MOD 2. NUREG/lA 0085-ASSESSMENT OF FULL POWER TURBINE TRIP LOZANO.M.F. START-UP TEST FOR C. TRILLO I WITH RELAP5/ MOD 2. NUREG/IA 0085. ASSESSMENT OF FULL POWER TURBINE TRIP MO START-UP TEST FOR C. TRILLO I WITH RELAPS/ MOD 2. N RE'G/CR-6052: METHODOLOGY FOR RELIABILITY BASED CONDI-LUK,K.H. TION ASSESSMENT. Application To Concrete Svuctures in Naclear NUREG/CR-5754. BOILING. WATER REACTOR INTERNALS AGING Plants DEGRADATION STUDY. Phase 1. NUREG/CR-6048 PRESSURIZED-WATER REACTOR INTERNALS MULLINS.A. AGING DEGRADATION STUDY. Phase 1, NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE. AND LUZA.K.V. DISPOSE OF 11E (2) BYPRODUCT MATERIAL NEAR NUREG/CR-6034 OKLAHOMA SEISMIC NETWORKJinal Report CLIVE. UTAH. Docket No 40-8989 Envirocare Of Utah.Inc. MANTEUFELR.D. MURPHY,W.M. NUREG/CR-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/CR 5817 V03 N2: NRC HIGH LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. July-December 1992. RESEARCH AT CNWRAJuly December 1992. NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL-HYDROLOGIC-MECHAN l CAL-CHEMICAL PROCESSES PERTINENT N AUS.D.J. TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA NUREG/CR4015: STRUCTURAL AGING PROGRAM TECHNICAL MOUNTAIN PROGRESS FOR PERIOD JANUARY - DECEMBER 1992. NUREG/CR-6026: THEORETICAL AND EXPERIMENT AL INVESTIGA-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY NGUYEN.B. SSTURATED. FRACTURED POROUS MEDIUM. NUREG/CR-6079: SEISMOLOGICAL INVESTIGATION OF EARTH-OUAKES IN THE NEW MADRID SEISMIC ZONE.Fmal MARTINEZ-GURIDI ReportSeptember 1986 Decembet 1992. NUREG/CR-5993 V01: METHODS FOR DEPENDE'N ESTIMATION AND SYSTEM UNAVAILABILITY EVALUATION BA. ')N F AILURE NICHOLSON,T J. DATA STATISTICS Summary Report NUREG/CP-0040: PROCEEDINGS OF WORKSHOP V: FLOW AND NUREG/CR-5993 V02: METHODS FOR DEPENDENC. iSTIMATION TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE-AND SYSTEM UNAVAILABILITY EVALUATION BASEG ON FAILURE LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSALHold At DATA STATISTICS Detailed Desenption And Applications. Radisson Suite Hotel, Tucson. Anzona. January 7 10.1991. MATSUKOV,l.D. NICKOLAUS,.L8L NUREG/CR4072 EXPERIMENTAL STUDY ON THE COMBUSTION BE. NUREG/CR-5829: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-h HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG. SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER NITION AT LARGE SCALE. PLANT. MCCONNELL,J.W. NOVO M. NUREG/CR-6073: LYSIMETER LITERATURE REVIEW. NUREG/IA-0085 ASSESSMENT OF FULL POWER TURBINE TRIP START-UP TEST FOR C. TRILLO I WITH RELAP5/ MOD 2. MCCORD,J.T. NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS-OKRENT,D. MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE NURE G/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF DISPOSAL FACILITIES.Eva'uation Of Modeling Approaches. SEVERE ACCIDENT MANAGEMENT STRATEGIES.

Personal Author index 23 OLAGUE N.E. SANECKl J.E. NUREG/CR.5927 V01. EVALUATION OF A PERFORMANCE ASSESS-NUREG/CR4667 V16 ENVIRONMENTALLY ASSISTED CRACKING IN MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE LIGHT WATER REACTORS. Semaannual Repo tOctober 1992 - March DISPOSAL FACILITIES. Evaluation Of Modoing Approaches. 1993. OLAND.C.B. SANTAMARIAJ.G. NUREG/CR-6015. STRUCTURAL AGING PROGRAM TECHNICAL NUREG/lA.0085. ASSESSMENT OF FULL POWER TURBINE TRIP PROGRESS FOR PEROD JAWARY - DECEMBER 1992-RTART-UP TEST FOR C. TRILLO I WITH RELAP5/ MOD 2. OLSZEWSKI.M. NUREG/CR.6065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR' NUREG/ 28: ISLOCA REbuRCH PROGRAMhnal Report PABALAN,R.T. NUREG/CR-5817 V03 N2: NRC HIJH-LEVEL RADIOACTIVE WASTE SEEBER L RESEARCH AT CNWRA. July. December 1992. NUREG/CR.5778 V03. NEW YORK /NEW JERSEY REGIONAL SEISMIC NETWORK. Final Report For Apnl 1985 - September 1992. PALOMARJ. NUREG/CR 6090. THE PROGRAMMABLE LOGIC CONTROLLER AND SHACK,WJ. ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS. NUREG/CR-4667 V16. ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Serruannual ReportOctober 1992. March p NUREG/CR-6056: A FRAMEWORK FOR THE ASEESSMENT OF SEVERE ACCIDENT MANAGEMENT STRATEGIES-SHIH,C.F. NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT PARK,J.Y' CR-4667 V16: NUREG/ ENVIRONMENTALLY ASSISTED CRACKING IN L*GHT WATER REACTORS Semiannual Report, October 1992 - March SHTEYNGART.S. 0 NUREG/CR4078: ANALYSIS OF CRACK INITIATION AND GROWTH IN PARK,Y.J. THE HIGH LEVEL VfBRATION TEST AT TADOTSU. NUREG/CR-6078. ANALYSIS OF CRACK INITIATION AND GROWTH IN THE HIGH LEVEL VIBRATION TEST AT TADOTSU. SHUM D.K. NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC PARKER.G. FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG-1479. RESULTS FROM TWO WORKSHOPS. STATE RADI. ROWE REACTOR PRESSURE VESSEL. ATON CONTROL PROGRAMS DEVELOPING AND AMENDING REG-ULATIONS AND FUNDING SIBOL,M.S. NUREG/CR4058. VIRGINIA REGIONAL SEISMIC NETWORK. Final PATTERSON,M. Report (1986 1992). NUREG/CP-0129 PROCEEDINGS OF TEE WORKSHOP ON PROGRAM FOR ELIMINATION OF REQUIREMENTS MARGINAL TO SAFETY. SIDOROV,V.P. NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-UREG/C5I-5817 V03 N2. NRC HIGH-LEVEL RADIOACTIVE WASTENfT A LA G SCA E' RESEARCH AT CNWRA. July-December 1992. PEREZJ. SIMONS,J.W. NUREG/lA-0107 ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD NUREG/lA-0126: 2D/3D PROGRAbi WORK

SUMMARY

REPORT. REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS 11 NUREG/lA-0127: REACTOR SAFETY ISSUES RESOLVED BY THE 2D/ NUCLEAR POWER PLANT. 3D PROGRAM. PRECKSHOT.G.G. SNOKEJ.A. NUREG/CR4082: DATA COMMUNICATIONS NUREG/CR4058: VIRGINIA REGIONAL SEISMIC NETWORK Final NUREG/CR-6083. REVIEWING REAL. TIME PERFORMANCE OF NU. Report (1986 -1992). CLEAR REACTOR SAFETY SYSTEMS. sRIDHAR N. PREECE.RJ. NUREG/CR 5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/lA 0113. PRELIMINARY ASSESSMENT OF PWR STEAM GEN-RESEARCH AT CNWRA. July-December 1992. ERATOR MODELLING IN RELAPS/ MOD 3. STlREWALT,G.L PUTNEYJ.M. NUREG/CR.5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/lA-0113. PRELIMINARY ASSESSMENT OF PWR STEAM GEN. ERATOR MODELLING IN RELAPS/ MOD 3 RESEARCH AT CNWRA. July. December 1992. RADDATZ,C.T. STOETZEL,G.A. NUREG-0713 V13' OCCUPATIONAL RADIATION EXPOSURE AT COM. NUREG 1400: AIR SAMPLING IN THE WORKPLACE.F nal Report MERCIAL NUCLEAR POWER REACTORS AND OTHER STOOP.P.M. FACILITIES.1991. Twenty-Fourth Annual Report NUREG/lA 0112: ASSESSMENT OF RELAP5/ MOD 2 AGAINST ECN-RE-g FLOOD EXPERIMENTS NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS. STROM,DJ. MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE W ASTE DISPOSAL FACILITIES Evaluation Of Modeling Approaches NUREG.1400: AIR SAMPLING IN THE WORKPLACEBnal Report ROGERS.R.D. STRUCKMEYER,R. NUREG/CR.6073: LYS! METER LITERATURE REVIEW. NUREG-0837 V13 NO2: NRC TLD DIRECT RADIATION MONtTORING NETWORK. Progress Report ApHiJune 1993. RouRK.CJ. NUREG.1453-REGULATORY ANALYSIS FOR THE RESOLUTION OF SU,T.-M. GENERIC ISSUE 142: LEAKAGE THROUGH ELECTRICAL ISOLA. NUREG.1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF TORS IN INSTRUMENTATION CIRCUITS. GENERIC ISSUE 153. LOSS OF ESSENTIAL SERVICE WATER IN LWRS. RUTHER,W.E. NUREG/CR-4667 V16. ENVIRONMENTALLY AS$1STED CRACKING IN SUEN.CJ. LIGHT WATER REACTORS. Serruannual ReportOctober 1992 - March NUREG/CR.5943: SENSITIVITY ANALYSIS AND BENCHMARKING OF 1993. THE BLT LOW-LEVEL WASTE SOURCE TERM CODE.

24 Personal Author index SULLIVAN,5.G DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-CUVE, UTAH. Docket No 404989 Enwocare 01 Utah.Inc. CLEAR POWER PLANTS: ANNOT ATED BIBUDGRAPHY OF SELECT-ED READINGS IN RADnATION PROTECTON AND ALARA. WEBER.M. NUREG 1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO SULLIVAN.T.M. CONSTRUCT AND OPERATE A FACILITY TO RECEIVE. STORE. AND NUREGICR-5943, SENSITIVITY ANALYSIS AND BENCHMARKING OF DISPOSE OF 11E (2) BYPRODUCT MATERIAL NEAR THE BLT LOW LEVEL WASTE SOURCE TERM CODE. CLIVE. UTAH. Docket No. 40-8989 Enwocare O! Utah,Inc. SURMElERA WESTRA.C. NUREG-1476-FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR-5758 V03. FITNESS FOR DUTY IN THE NUCLEAR POWER CONSTRUCT AND OPERATE A FACluTY TO RECEIVE. STORE, AND INDUSTRY. Annual Summary Of Program Perfonnance Reports.CY DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR 1992. CUVE. UTAH Docket No. 40 8969 Enwocare Of Utah.Inc. WlBLIN.C.M. SVEDEMAN.S.J. NUREG.1400: AIR SAMPLING IN THE WORKPLACE.Fenal Report. NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY WILSON.T.L. SATURATED, FRACTURED POROUS MEDIUM. NUREG/CR 6060; HYDROGEN MIXING STUDIES (HMS) ASSESSMENT MANUAL SWIDER.J. NUHt.ti/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF WINEGARDNER,W. SEVERE ACCIDENT MANAGEMENT STRATEGIES. NUREG/CR 6029 V01: AGING ASSESSMENT OF NUCLEAR AIR-TREATMENT SYSTEM HEPA FILTERS AND ADSORDERS. Phase 8. TODD.M.D. NUREG/CR-5944. A CHARACTERIZATION OF CHECK VALVE DEGRA-WINTERS.L DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-NUREG/iA-0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NATU-DUSTRY. RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT TRAVis.J.R. NUREG/CR-6060 HYDROGEN MIX!NG STUDIES (HMS) ASSESSMENT WITTMEYER.G.W. MANUAL NUREG/CR4817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE a RESEARCH AT CNWRA.Juty-December 1992. NUREG/CR-5817 V03 N2: NRC H6GH-LEVEL RADCACTIVE WASTE WOLFOONG.J.R. RESEARCH AT CNWRA. July-December 1992. NUREG/CR-6065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR. NUREG/CR6021: A UTERATURE REVIEW OF COUPLED THERMAL-HYDROLOGIC-MECHAN ICAL CHEMICAL PROCESSES PERTINENT WOUDSTRA.A. TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA NUREG/lA4112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECNTIE. MOUNTAIN. FLOOD EXPERIMENTS. VANDEBOGAARDA WRIGHT.A.L NUREG/lA 0112. ASSESSMENT OF RELAP5/ MOD 2 AGAINST ECN.RE. NUREG/CR4065: SYSTEMS ANALYSIS OF THE CANDU S REACTOR. FLOOD EXPER!MENTS. WYMAN.R. VESELY.W.E. NUREG/CR4090: THE PROGRAMMABLE LOGIC CONTROLLER AND NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTiMATON ITS APPUCATION IN NUCLEAR REACTOR SYSTEMS. AND SYSTEM UNAVAILABILITY EVALUATON BASED ON FAILURE DATA STATISTICS Summary Report XING.L NUREG/CR4993 V02: METHODS FOR DEPENDENCY ESTIMATION NUREG/CR4056: A FRAMEWORK FOR THE ASSESSMENT OF AND SYSTEM UNAVAILABluTY EVALUATsON BASED ON FAILURE SEVERE ACCOENT MANAGEMEN1 STRATEGIES. DATA STATISTICS Detaned Desenption And Applications. YANKINAG. VO.T.V. NUREG/CR4072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-NUREG/CR-5P.29. AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-SPECTON GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER NITION AT LARGE SCALE. PLANT. NUREG/CR-5833 AUXIUARY FEEDWATER SYSTEM RISK-BASED IN-YARDUM1AN.J. SPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER NUREG 0525 V02 RO1: SAFEGUARDS

SUMMARY

EVENT LIST PLANT. (SSEL). January 1,1990 Through December 31,1992. CANG.Y.K. YOUNG,$.R. NUREG/CR4049 PIPING BENCHMARK PROBLEMS FOR THE GEN. NUREG/CR4817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE ERAL ELECTRIC ADVANCED BOluNG WATER REACTOR. RESEARCH AT CNWRAJuly-December 1992. CASTLER.S. YU.D. NUREG-1476 FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR4056: A FRAMEWORK FOR THE ASSESSMEN ' OF CONSTRUCT AND OPERATE A FACIUTY TO RECElVE. STORE. AND SEVERE ACCIDENT MANAGEMENT STRATEGIES. l__________________-__________._______-_______-______-____-________-__

Subject index This index was developed from keywords and word strings in titles and abstracts. During this development period, there will be some redundancy, which will be removed later when a rea-sonable thesaurus has been developed through experience. Suggestions for improvements are welcome. 2D/3D Program NUREG/CR-5829 AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/lA 0126; 2D/3D PROGRAM WORK

SUMMARY

REPORT. SPECTION GUIDANCE FOR THE DAVISBESSE NUCLEAR PCMIER NUREG/lA-0127: REACTOR SAFETY ISSUES RESOLVED BY THE 2D/ PLANT. 3D PROGRAM. NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN. AEOD SPECTION GUIDE FOR THE H.B ROBINSON NUCLEAR POWER P MNT* NUREG-1272 V07 Noi: OFFCE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA.1992 Annual Report - Power Reactors. BETHSY Test 9 NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 9.1.B USING EG/CR 3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU. CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT. BWR ED READINGS IN RADIATION PROTECTION AND ALARA NUREG/CR-5754: BOILING-WATER REACTOR INTERNALS AGING I Occurrence DEGRADATION STUDY. Phase 1. NUREG-0090 V16 N01: REPORT TO CONGRESS ON ABNORMAL NUREG/CR-5942-SEVERE ACCIDENT SOURCE TERM CHARACTER. OCCURRENCES. January-March 1993. ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED BY THE MELCOR CODE. Adm NUREG/CR-6049: PIPING BENCHMARK PROBLEMS FOR THE GEN-NOREG/CR4029 V01: AGING ASSESSMENT OF NUCLEAR AIR-ERAL ELECTRC ADVANCED BOILING WATER REACTOR. TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS. Phase L BWal TN Advloory Committee On Nuclear Weste NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT NUREG-1423 V04: A COMPfLATION OF REPORTS OF THE ADVISORY OF CONSTRAINT IN FRACTURE. COMMITTEE ON NUCEAR WASTEJuly 1992 June 1993. Boiling Water Reactor Aging NUREG/CR-5754: BOIL!NG. WATER REACTOR INTERNALS AGING NUREG/CR-5404 V02: AUXILIARY FEEDWATER SYSTEM AGING DEGRADATION STUDY. Phase 1. STUDY. Phase i Fotbw On Study NUREG/CR-5942: SEVERE ACCfDENT SOURCE TERM CHARACTER-NUREG/CR-5754: BOILING WATER REACTOR INTERNALS AGING ISTCS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED DEGRADATION STUDY. Phase 1. BY THE MELCOR CODE. NUREG/CR4015: STRUCTURAL AGING PROGRAM TECHNICAL NUREG/CR-6049: PIPING BENCHMARK PROBLEMS FOR THE GEN-PROGRESS FOR PERIOD JANUARY - DECEMBER 1992. ERAL ELECTRIC ADVANCED BOILING WATER REACTOR. NUREG/CR4029 V01: AGING ASSESSMENT OF NUCLEAR AIR-TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS Phase t Broadband Setsmology NUREG/CR4043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC NUREG/CR4079: SEISMOLOGICAL INVESTIGATION OF EARTH-CHILLERS USED IN NUCLEAR POWER PMNTS. Phase L QUAKES IN THE NEW MADRID SEISMIC ZONEFanal NUREG/CR-6048: PRESSURIZED-WATER REACTOR INTERNALS Report. September 1986 - December 1992. AGING DEGRADATION STUDY. Phase 1. NUREG/CR-6052: METHODOLOGY FOR RELIABILITY BASED CONDb Byproduct Material TION ASSESSMENT. Apphcation To Concrete Structures in Nuclear NUREG 1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO Plants. CONSTRUCT AND OPERATE A FAnlLITY TO RECEIVE. STORE. AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR ^" ~ N R G 147 ESULTS FROM TWO WORKSHOPS: STATE RADb ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG-CANCU 3 Reactor ULATtONS AND FUNDING. NUREG/CR-6065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR. Air Sampling CFO's Act NUREG-1400: AIR SAMPLING IN THE WORKPLACE.Fanal rem NUREG-1470 V02: CHIEF FINANCIAL OFF1CER'S ANNUAL REPORT - Alr-Treatment System 1993. NUREG/CR4029 V01: AGING ASSESSMENT OF NUCLEAR AIR-TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS. Phase L N E C 56: A FRAMEWORK FOR THE ASSESSMENT OF Airborne Radioactive Maternal SEVERE ACCIDENT MANAGEMENT STRATEGIES. NUREG-1400. AIR SAMPLING IN THE WORKPLACEfenal Report. C W py Annual Reporg NUREG/CR4023. GENERIC ANALYSIS FOR EVALUATION OF LOW NUREG-1145 V09: U.S NUCLEAR REGULATORY COMMISSION 1992 CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS ANNUAL REPORT. AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERI-NUREG-1470 V02: CHIEF FINANCIAL OFFICER'S ANNUAL REPORT - ALS. 1993. Check Valve Atonde Safety And Licens6ng Board Panel NUREG/CR-5944. A CHARACTERIZATION OF CHECK VALVE DEGRA-NUREG-1363 V05: ATOMIC SAFETY AND LICENSING BOARD PANEL DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-ANNUAL REPORTJescal Year 1992.

USTRY.

Auxillary Feedwater System e NUREG/CR-5404 V02: AUXILIARY FEEDWATER SYSTEM AGING aVREG/CR4043 V01: AGING ASSESSMENT OF EPSENTIAL HVAC STUDY. Phase 1 Follow.On Study. CHILERS USED IN NUCLEAR POWER PLANTS. Phase 1. 25

26 Subject Index Combustion Behav6or Electric Discharge NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTON BE-NUREG/CR-5981: THE E.FFECT OF ELECTRIC DISCHARGE MA-HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL NITION AT LARGE SCALE. STRUCTURAL ALLOYS. Communication Media Electr6 cal isolator NUREG/CR4D62. DATA COMMUNICATONS-NUREG-1453: REGULATORY ANALYSIS FOR THE RESOLUTION OF Cwek GENERIC ISSUE 142-LEAKAGE THROUGH ELECTRICAL ISOLA-TORS IN INSTRUMENTATION CIRCUITS. NUREG/CR-6015: STRUCTURAL AGING PROGRAM TECHNICAL PROGRESS FOR PERIOD JANUARY - DECEMBER 1992. Embrtttlement Concrete Structure NUREG/CR-4744 V07 N2: LONG-TERM EMDRITTLEMENT OF CAST NUREG/CR-6052. METHODOLOGY FOR RELIABtLITY BASED CONDI. DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual TION ASSESSMENT. Appleatton To Concrete Structures in Nuclear Report,Apni-September 1992. ants. Emergency Shutdown System Conngurable System NUREG/CR-6090: THE PROGRAMMABLE LOGIC CONTROLLER AND NUREG/CR4093: THE PROGRAMMABLE LOGIC CONTROLLER AND ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS. ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS. g g, Constra6nt NUREG-0940 V12 NO2: ENrORCEMENT ACTONS: SIGNIFICANT AC-NUREG/CR-5971: CONTINUUM AND MiCROMECHANICS TREATMENT TIONS RESOLVED.Ouarterly Progress Report. April-June 1993. OF CONSTRA!NT IN FRACTURE. Envirocare Containment Penetration NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR-5928 ISLOCA RESEARCH PROGRAM. Final Report-CONSTRUCT AND OPERATE A FACILITY TO RECE!VE. STORE, AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR HYDROGEN MIXING STUDIES (HMS) ASSESSMENT MANUAL-Essential Service Water NUREG 1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF Corros6on Fatigue NUREG/CR-4067 V16 ENVIRONMENTALLY ASSISTED CRACKING IN GENERIC ISSUE 153: LOSS OF ESSENTIAL SERVICE WATER IN LWRS. LIGHT WATER REACTORS. Semiannual Report. October 1992 March 9 93. Extremity Dosimetry Crack NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMITY DOSI-NUREG/CR4078: Af4ALYSIS OF CRACK INITIATON AND GROWTH IN METERS-PtLOT TEST. THE HIGH LEVEL VIBRATON TEST AT TADOTSU. FMwe Dau Cracking NUREG/CR-5944: A CHARACTERIZATION OF CHECK VALVE DEGRA. NUREG/CR-4667 V16. ENVIRONMENT ALLY ASSISTED CRACKING IN DATION AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN-LIGHT WATER REACTORS. Sem6 annual Report. October 1992 + March DUSTRY. 1993. F6nal Environmental impact Statement Data Communication NUREG-1476. FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR-6082: DATA COMMUNICATONS. CONS"t0CT AND OPERATE A FACILITY TO RECEIVE. STORE, AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR tU EG/CR 1: ENHANCED REMOVAL OF RADIOACTIVE PARTI. CLES BY FLUOROCARBON SURF ACTANT SOLUTONS Financial Management Dependency Estimation 3 NUREG/CR-5993 V01: METHODS FOR DEPENDENCY ESTlMATON AND SYSTEM UNAVAILABILITY EVALVATON BASED ON FAILURE Finger Dosimeter ^" NURE /R5 THO OR DEPENDENCY ESTIMATION E R OT T S ' AND SYSTEM UNAVAILAB!LITY EVALUATION BASED ON FAILURE DATA STATISTICS. Detailed Desenption And Applications-Fitness For Duty NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER Detonation S Annual Summay O gram Mormance Repods,W NUREG/CR4072: EXPERIMENTAL STUDY ON THE COMBUSTON BE-92' HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-NITION AT LARGE SCALE-Fluorocart>on NUREG/CR-6081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-Disposition Schedule NUREG-0910 R02 S01. NRC COMPREHENSfVE RECORDS DISPOSI-CLES BY FLOOROCARBON SURFACTANT SOLUTONS. TON SCHEDULE. Fracture Mechan 6cs Ductile Fracture NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC NUREG/CR-6023. GENERIC ANALYSIS FOR EVALUATION OF LOW FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY 'AARGtNS ROWE REACTOR PRESSURE VESSEL AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERi-NUREG/CR-5970: APPROXIMATE TECHNIOUES FOR PREDICTING ALS' SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC). NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT Dynam6e Load OF CONSTRAINT IN FRACTURE-NUREG/CR-6049 PIPING BENCHMARK PROBLEMS FOR THE GEN-Frae To g sa ERAL ELECTRIC ADVANCED BOILING WATER REACTOR. g Earthquake DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semannnual NUREG/CR4034. OKLAHOMA SEISMIC NETWORK Final Report. Report. April-September 1992. NUREG/CR4079-SEISMOLOGICAL INVESTIGATION OF EARTH-NUREG/CR-5969: J AND CTOD ESTIMATION EQUATONS FOR SHAL-QUAKES IN THE NEW MADRID SEISMIC ZONE. Final LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. Report September 1986 - December 1992. NUREG/CR-5970; APPROXIMATE TECHNIOUES FOR PREDICTING NUREG/CR4085: UNITED STATES SEISMOGRAPHIC NETWORK. SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC).

Subject index 27 NUREG/CR-59B1: THE EFFECT OF ELECTRIC DISCHARGE MA-NUREG/lA-0100: ASSESSMENT OF CCFL MODEL OF RELAP5/ MOD 3 i CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL AGAfNST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. 1 STRUCTURAL ALLOYS, NUREG/lA-0103. ASSESSMENT OF BETHSY TEST 9.1.B USING 1 RELAP5/ MOD 3. Frscrured Media NUREG/lA 0107: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS 11 RESEARCH AT CNWRA. July-December 1992. NUCLEAR POWER PLANT. NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-NUREG/lA-0112: ASSESSMENT OF REMPS/ MOD 2 AGAINST ECN.RE. TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY FLOOD EXPERIMENTS. SATURATED. FRACTURED POROUS MEDIUM. NUREG/lA-0113: PRELIMINARY ASSESSMENT OF PWR STEAM GEN-ERATOR MODELLING IN RELAP5/ MOD 3. NUREG/CP,0040: PROCEEDINGS OF WORKSHOP V: FLOW AND ISLOCA TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE4 NUREG-1463: REGULATORY. ANALYSIS FOR THE RESOLUTION OF LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSALHeld At GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF. Radisson Surte Hotel. Tucson. Anzona. January 710.1991. COOLANT ACCIDENT IN LIGHT-WATER REACTORS. Generic issue 142 ISLOCA Research Program NUREG-1453. REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG/CR 5928: ISLOCA RESEARCH PROGRAM.F nal Report. GENERIC ISSUE 142: LEAKAGE THROUGH ELECTRICAL ISOM-TORS IN INSTRUMENTATION CIRCUlTS. Inspect 6on Guidance 'l NUREG/CR-5829: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-Generic losue 153 SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER NUREG-1461: REGULATORY ANALYSIS FOR THE FIESOLUTION OF Pu NT. GENERIC ISSUE 153: LOSS OF ESSENTIAL SERVICE WATER IN LWRS. Inspection Guide NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-Generic Safety losue 105 SPECTION GUIDE FOR THE H B. ROBINSON NUCLEAR POWER i NUREG-1463: REGULATORY ANALYSIS FOR THE RESOLUTION OF. PUNT. ~' GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF-i l COOLANT ACCIDENT IN LIGHT-WATER RUF MS. Instrumentation Circuit HEPA Mer NUREG 1453 REGULATORY ANALYSlS FOR THE RESOLUTION OF 'i GENERIC ISSUE 142: LEAKAGE THROUGH ELECTRICAL ISOLA. I NUREG/CR-6029 V01: AGING ASSESSt 1 CLEAR AIR. TORS IN INSTRUMENTATION CIRCulTS' / TREATMENT SYSTEM HEPA FILTERS AND ADSORtmHS. Phase L

• • 0"U N' HVAC R /

8: 1 A 8 Port -j NUREG/CR-6043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC CHILLERS USED IN NUCLEAR POWER PLANTS, Phase L LOCA f H6gh-Level Radioactive Weste NUREG/LA-0126: 2D/3D PROGRAM WORK

SUMMARY

REPORT. NUREG 1423 V04: A COMPILATIOh OF REPORTS OF THE ADVISORY. COMMITTEE ON NUCLEAR WAS't E.Jutv 1992 - June 19D3. PR NUREG/CR-5817 V03 N2; NRC HIGH-LEVEL RADIOACTIVE WASTE j RESEARCH AT CNWRA. July-December 1992. l LWR NUREG-1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF. l' Highf.evel Rad 6oactive Weste D6sposal GENERIC ISSUE 153. LOSS OF ESSENTIAL SERVICE WATER IN NUREG/CP 0040: PROCEEDINGS OF WORKSHOP V; FLOW AND LWRS. TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE. NUREG-1463: REGULATORY ANALYSIS FOR THE RESOLUTION OF 3 LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSAL. Held At GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF-j COOLANT ACCIDENT IN LIGHT WATER REACTORS NUI CR 6026 HE ICA AD R L' INVESTIGA-NUREG/CR-4867 V16: ENVIRONMENTALLY ASSISTED CRACKING IN j TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY LIGHT WATER REACTORS. Semiannual Report. October 1992 - March r SATURATED. FRACTURED POROUS MEDIUM. N /CR-4744 V07 N2: LONG-TERM EMBRITTLEMENT OF CAST: k High-Level Waste Repoeltory DUPLEX STAINLESS STEELS'IN LWR - SYSTEMS.Sermannual i NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL. Report. April-September 1992. I HYDROLOGIC-MECHAN ICAL-CHEMICAL PROCESSES PERTINENT NUREG/GR 0005 V02 P1 RISK-BASED INSPECTION-DEVELOPMENT l TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA OF GUIDELINES. Light Water Reactor (LWR) Nuclear Power Plant l MOUNTAIN. Components Hydraulic Transoort Leakage THE B W E WA TE R TERM E GENE SSU 12 AVA R El.E ICAL TORS IN INSTRUMENT ATION CIRCulTS. l Hydrogen 6 NUREG/CR-6000: HYDROGEN MIXING STUDIES (HMS) ASSESSMENT Legal T ) MANUAL NUREG 0750 V37101: INDEXES TO NUCLEAR REGULATORY COM-' t MISSION ISSUANCES. Janua. -March 1993. i Hydrogen-Air NUREG 0750 V37 ND4: N R REGUMTORY COMMISSION IS-l NUREG/CR 6072: EXPERIMENTAL STUDY ON THE COMBUSTION BE-SUANCES FOR APRIL 1993.Pages 251354. HAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG. NUREGC750 V37 N05: NUCLEAR REGULATORY COMMISS4ON IS-l NITION AT LARGE SCALE. SUANCES FOR MAY 1993 Pa 355 418. I NUREG 0750 V37 N06: N AR REGULATORY COMMISSION IS-( Hydrogeology SUANCES FOR JUNE 1993.Pages 419-515. .l NUREG/CH-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE NU5EG 0750 V38 N01: NUCLEAR REGULATORY COMMISSION IS-RESEARCH AT CNWRA.Ju!y-December 1992. SUANCES FOR JULY 1993.Pages 124. ICAP Program Ught Weter Reactor NUREG/lA-0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NATU. NUREG-1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT GENERIC ISSUE 153: LOSS OF ESSENTIAL SERVICE WATER IN BORSSELE. LWRS. NUREG/lA-0096: NUMERICS AND IMPLEMENTATION OF THE UK NUREG-1463: FIEGULATORY ANALYSIS FOR THE RESOLUTION OF HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE MODEL GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSSOF. INTO RELAP5/ MOD 3. COOLANT ACCIDENT tN LIGHT-WATER REACTORS. I \\ 1 i I w

28 Subject Index idUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN Nuclear Air Cleaning LIGHT WATER REACTORS. Semiannual Report. October 1992 - March NUREG/CP-0130 V01: PROCECDINGS OF THE 22ND DOE /NRC NU-1993. CLEAR AIR CLEANING CONFERENCE.Semons 1-6 Held in MUREG/CR-4744 V07 N2: LONG-TERM EMBRITTLEMENT OF CAST Denver, Colorado. August 24-27,1992 DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Sermannual NUREG/CP 0130 V02; PROCEEDINGS OF THE 22ND DOE /NRC NU-Report. April-September 1992. CLEAR AIR CLEANING CONFERENCE Sessions 9-16 Held in WL* REG /GR 0005 V02 P1: RISK-BASED INSPECTION-DEVELOPIAT Denver, Colorado, August 24-27,1992. j OF GUIDEUNES Lsght Water Reactor (LWR) Nuclear Power Hant Components. Occupational Dose Reduction NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-Load Reject 6on CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-WUREG/lA-0107: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A LOAD ED READINGS IN RADIATION PROTECTION AND ALARA. REJECTION FROM 100% TO 50% POWER IN THE VANDELLOS ll i NUCLEAR POWER PLANT. Occupational Radletion Exposure 4 NUREG-0713 V13: OCCUPATIONAL RADIATION EXPOSURE AT COM-Log 6c Controller MERCIAL NUCLEAR POWER REACTORS AND OTHER NUREG/CR4090' THE PROGRAMMABLE LOGIC CONTROLLER AND FACILITIES.1991. Twenty-Fourth Annual Report NUREG/CR-6050: RADIATION EXPOSURE MONITORING AND INFOR. ITS APPUCATION IN NUCLEAR REACTOR SYSTEMS. MATION TRANSMITTAL (REVIT) SYSTEM User's Manual. Lose-Of-Coolant Accident NUREG-1463: REGULS. TORY ANALYSIS FOR THE RESOLUTION OF NUF EG491 R02 S01: NRC COMPREHENSIVE RECORDS DISPOSL I GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSSCF-TION SCHEDULE. COOLA4T ACCIDENT IN LIGHT-WATER REACTORS. NUREG/CR-5942 SEVERE ACCIDENT SOURCE TERM CHARACTER. Operating Experience ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED NUREG 1272 V07 N01: OFFICE FOR ANALYSIS AND EVALUATION OF BY THE MELCOR CODE. OPERATIONAL DATA.1992 Annual R - Power Reactors. NUREG/CR-5404 V02: AUXILIARY F EDWATER SYSTEM AGING pPebM STUDY. Phase t Follow On Study. i NUREG/CR-6023: GENERIC ANALYSIS FOR EVALUATION OF LOW NUREG/CR4043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS CHILLERS USED IN NUCLEAR POWER PLANTS. Phase I. AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERI-

ALS, Operator Act6on Low-Level Radioactive Waste NUREG-1423 V04: A COMPtLATON OF REPORTS OF THE ADVISORY PRA r

COMMITTEE ON NUCLEAR WASTE. July 1992 - June 1993. NL* LEG /CR-5829. AUXIUARY FEEDWATER SYSTEM RISK BASED IN-NUREG/CR-5943. SENSITIVITY ANALYSIS AND BENCHMARKING OF SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER THE BLT LOW LEVEL WASTE SOURCE TERM CODE. PLANT. NUREG/CR4833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-Low-Level Radioactive Waste Disposal SPECTION GUIDE FOR THE HB. ROBINSON NUCLEAR POWER NUREGICR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS-PLANT. MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE DISPOSAL FACluTIES Evaluation Of Modeling Approaches. PWR NUREG/CR 6048: PRESSURIZED-WATER REACTOR INTERNALS Low-Power AGING DEGRADATION STUDY. Phase 1. NUREG 1449. SHUTDOWN AND LOW-POWER OPERATION AT NU. CLEAR POWER PLANTS IN THE UNITED STATES Final R port. Performance Assessment l NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE ASSESS-l Lysimeter MENT METHODOLOGY FOR 1DW-LEVEL RADIOACTIVE WASTE NUREG/CR4073. LYSIMETER LTTERATURE REVIEW DISPOSAL FACluTIES Evaluation 01 Modeling Approaches. MELCOR Code Performance History NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARACTER. NUREG-1214 R12: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT. ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED IC ASSESSMENT OF UCENSEE PERFORMANCE. BY THE MELCOR CODE-Petitions For Rulemmung NRC Regulation NUREG-0936 V12 NO2: NRC REGULATORY AGENDA.Ouarterly NUREG-1479: RESULTS FROM TWO WORKSHOPS. STATE RADI-Report, April June 1993. t ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG-Pi n ULATIONS AND FUNDING. ERAL ELECTRIC ADVANCED BOlWNG WATER REACTOR. Natural Circulat6on NUREG/lA-0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NATU* Plant Transient RAL CIRCULATION EXPERIMENT IN NUCLEAR POWER PLANT NUREG/lA-0005: ASSESSMENT OF FULL POWER TURBINE TRIP BORSSELE. START-UP TEST FOR C. TRILLO I WITH RELAPS/ MOD 2. Nemaha Uplift Practice And Procedure Digest NUREG/CR4034 OKLAHOMA SEISMIC NETWORK. Final Report NUREG-0386 D06 R07: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE Nondestructive Evaluation DIGEST. Commission. Appeal Board And Ucensing Board NUREG/CR-6052: METHODOLOGY FOR REUABILITY BASED CONDi* Deciseons. July 1972 - September 1992. TION ASSESSMENT. Apphcotion To Concrete Structuros in Nut. lear Plants. Pressure Boundary NUREG/CR-5928: ISLOCA RESEARCH PROGRAM.Finni Report. NUREG/CP-0040: PROCEEDINGS OF WORKSHOP V: FLOW AND Pressurtred Thermal Shock TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE-NUREG/CR-57B2: PRESSURt2ED THERMAL SHOCK PROBABILtSTIC LATED TO HIGH-LEVEL RADIOACTIVE WASTE DISPOSALHeld At FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE Radesson Suite Hotel. Tucson, Anzona, January 710,1991. ROWE REACTOR PRESSURE VESSEL. i Notch Band Pressurtrod Water Reactor t NUREG/CR-5969: J AND CTOD ESTIMATION EQUATIONS FOR SHAL-NUREG/CR-6048 PRESSURIZED. WATER REACTOR INTERNALS l LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. AGING DEGRADATION STUDY. Phase 1. i l ___.,_______m_.

l Subject index 29 Primary Coolant Circuit Regulatory Agenda NUREG/CR4078: ANALYSIS OF CRACK INITIATON AND GROWTH IN NUREG4936 V12 NO2. NRC FIEGULATORY AGENDA. Quarterly THE HIGH LEVEL VIBRATION TEST AT TADOTSU. Peport April-June 1993. Program Performance Report Regulatory And Technical Report l NUREG/C45758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG4304 V18 NO2: REGULATORY AND TECHNICAL FIEPORTS j INDUSTRY. Annual Summary Of P'~ yam Performance ReportsCY (ABSTRACT INDEX JOURNAL). Compilaton For Second Quarter 1992. 1993. April-Junaf RELAP5/ MOD 2 Regulatory Rev6ew And improvement NUREG/lA 0085: ASSESSMENT /7 FULL POWER TURBINE TRIP NUREG/CP-0129; PROCEEDINGS OF THE WORKSHOP ON PROGRAM START-UP TEST FOR C. TRILLO e wtTH RELAPS/ MOD 2. FOR ELIMINATION OF REQUIREMENTS MARGINAL TO SAFETY. NUREG/lA-0091: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A NATU- ] RAL CIRCULATON EXPERIMENT IN NUCLEAR POWER PLANT - Rollability BORSSELE. NUREGICR-5944: A CHARACTERIZATION OF CHECK VALVE DEGRA. NUREG/lA4107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A LOAD DATON AND FAILURE EXPERIENCE IN THE NUCLEAR POWER IN- ' REJECTION FROM 100% TO $0% POWER IN THE VANDELLOS 11 DUSTRY* NUCLEAR POWER PLANT. NUREG/tA-0112 ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN-RE* Report To Congroms FLOOD EXPERIMENTS-NUREG-0090 V16 N01: REPORT TO CONGRESS ON ABNORMAL-t RELAPS/ MOD 3 OCCURRENCES. January-March 1993. l NUREG/lA-0096 NUMERICS AND IMPLEMENTATION OF THE UK Ring Doe 4 meter HORIZONTAL STRATIFICATION ENTRAINML.T OFF-TAKE MODEL NUREG/CR-5989: PERFORMANCE TESTING OF EXTREMITY DOSI- [ N E/ 0 00 SSMENT OF CCFL MODEL OF RELAP5/ MOD 3 - METERS--PILOT TEST, j AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. gi,g NUR /lA-01 ASSESSMENT OF BETHSY TEST 9.1.B USING NUREG-1449: SHUTDOWN AND LOW. POWER OPERATION AT NU-NUREG/lA4113: PRELIMINARY ASSESSMENT OF PWR STEAM GEN. CLEAR POWER PLANTS IN THE UNITED STATES.Foat Report - ERATOR MODELLING IN RELAP5/ MOD 3. gg,g gn,g,g, y REMIT NUREG/CR4056: A FRAMEWORK FOR THE ASSESSMENT OF - NUREG/CR-6050: RADIATION EXPOSURE MONITORING AND INFOR. SEVERE ACCIDENT MANAGEMENT STRATEGIES. - l MATION TRANSMITTAL (REMIT) SYSTEM User's Manual. { pg,g Radiation Control Program NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF NUREG-1479-RESULTS FROM TWO WORKSHOPS: STATE RADL SEVERE ACCIDENT MANAGEMENT STRATEGIES. { ATION CONTFIOL PROGRAMS DEVELDPING AND AMENDING REG- - [ ULATONS AND FUNDING ~ Reak-Saeed inapoction NUREG/GR-0005 V02 P1: R!SK-BASED INSPECTION-DEVELOPMENT i Radiat6on Protection OF GUIDELINES.Lght Water Reactor (LWR), Nuclear Power Plant l NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-Components. CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT. -+ ED READINGS IN RADIATION PROTECTION AND ALARA. Mah-Saeed Regulation ' -{ i NUREG/CP 0129 PROCEEDINGS OF THE WORKSHOP ON PROGRAM t l Radeonctive Particle FOR ELIMINATION OF REQUIREMENTS MARGINAL TO SAFETY. l NUREG/CR 6081: ENHANCED stEMOVAL OF RADIOACTIVE PARTI-CLES BY FLUOROCARBON SURFACTANT SOLUTONS. Rod Bundle NUREG/lA-0100: ASSESSMENT OF CCFL MODEL OF FIELAPS/ MOD 3 Radioactive Weste AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. t NUREG/CR4073: LYSIMETER LITERATURE REVIEW. Rules 6 Radlonucl6de M6eration NUREG4936 V12 NO2: NRC REGULATORY AGENDA.Ouarterly i NUREG/CH-5943: SENSITtVITY ANALYSIS AND BENCHMARKING OF Report. April-June 1963. ( THE BLT LOW LEVEL WASTE SOURCE TERM CODE. Rules Of Practice Reactor Component NUREG4386 006 R07: UNITED STATES NUCLEAR REGULATORY NUREGICR-5993 V01: METHODS FOR DEPENDENCY ESTIMATION COMMISSON STAFF PRACTICE ANG PROCEDURE I AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE DIGEST.Commisson, Appeal Board.And L6censmg Board I DATA STATISTICS. Summary Report. DecisionsJuly 1972 - September 1992. NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATION AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAfLURE Safeguardo Summary Event List t DATA STATISTICS. Detailed Deacnphon And Apphcatons. NUREG-0525 V02 RO1: SAFEGUARDS

SUMMARY

EVENT LIST I (SSEL) January 1,1990 Through December 31,1992. f i Reactor Pressure Vessel NUREG/CR-5782: PRESSURIZED THERMAL SHOCK PROBABILISTIC Satellite Telemetry } l FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR YANKEE NUREG/CR-6085: UNITED 61 ATES SEISMOGRAPHIC NETWORK. I ROWE REACTOR PRESSURE VESSEL NUREG/CR4023: GENERIC ANALYSIS FOR EVALUATION OF LOW Se6amic Effect CHARPY UPPER-?HELF ENERGY EFFECTS ON SAFETY MARGINS NUREG/CR-6078: ANALYSIS OF CRACK INITIATION AND GROWTH IN { AGAINST FRACTURE OF BEACTOR PRESSURE VESSEL MATERI-THE HIGH LEVEL VIBRATON TEST AT TADOTSU. ALS. I Setemic Network Reactor Safety NUREG/CR-5778 V03: NEW YORK /NEW JERSEY REGIONAL SEISMIC NUREG/CR 6049: PIPING BENCHMARK PROBLEMS FOR THE GEN-NETWORK.Fmal Report For April 1985 - September 1992. ERAL ELECTRIC ADVANCED BOILING WATER REACTOR. NUREG/CR 6034: OKLAHOMA SEISMIC NETWORK. Final Report. NUREG/CR4058: VIRGINIA REGIONAL SEISMIC NETWORK.Fmat Reactor Safety System Neport (1986 -1992). NUREG/CR4083: REVIEWING REAL TIME PEFIFORMANCE OF NU-NUREG/CR-6079: SEISMOLOGICAL INVESTIGATION OF EARTH-CLEAR REACTOR SAFETY SYSTEMS OUAKES IN THF NEW MADRID SEISMIC ZONE.Fmal ReporLSeptember 1988 - December 1992.. ) Real-Time Performance NUREG/CR4083: REVIEWING REAL TIME PERFORMANCE OF NU-Seismographic Network CLEAR FIEACTOR SAFETY SYSTEMS. NUREG/CR4085: UNITED STATES SEISMOGRAPHIC NETWORK. 1

30 Subject index Severe Acc6 dent NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATION NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARACTER-AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED DATA STATISTICS. Detailed Descrtpton And Applicatons. BY THE MELCOR CODE. NUREG/CR4056: A FRAMEWORK FOR THE ASSESSMENT OF Systematic Assessment Of Licensee Performance SEVERE ACCIDENT MANAGEMENT STRATEGIES. NUREG-1214 R12: HISTORICAL DATA

SUMMARY

OF THE SYSTEMAT-IC ASSESSMENT OF UCENSEE PERFORMANCE. Shanow Creek NUREG/CR4969:J AND CTOD ESTIMATION EQUATIONS FOR SHAL-LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. UREG-0837 V13 NO2: NRC TLD DIRECT RADIATtON MONITORING NETWORK. Progress Report Aprildune 1993. Shutdown NUREG-1449: SHUTDOWN AND LOW-POWER OPERATION AT NU-A 126 2D/3D PROGRAM WORK

SUMMARY

REPORT. CLEAR POWER PLANTS IN THE UNITED STATES.Fmal Report NUREGAA-0127: REACTOR SAFETY tsSUES RESOLVED BY THE 2D/ Source Term 3D PROGRAM. NUREG/CR4943: SENSITIVITY ANALYSIS AND BENCHMARKING OF ThermaMiydrologic THE BLT LOW-LEVEL WASTE SOURCE TERM CODE. NUREG/CR4021: A UTERATURE REVIEW f7 COUPLED THERMAL. HYDROLOGIC-MELHAN ICAL-CHEMCAL PROCESSES PERTINENT 0 N / 4744 V07 N2: t^NSTERM EMBRITTLEMENT OF CAST y0U A N' DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Serrmantmaat Report. April-September 1992. Thermohydraul6c NUREGAA-0112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN.RE-State Regulat6on FLOOD EXPERIMENTS. NUPEG-1479 RESULTS FROM TWO WORKSHOPS: STATE RADI-ATION CONTROL PROGRAMS DEVELOPING AND AMENDING REG-Thermohydrologic ULATIONS AND FUNDING. NUREG/CR4026: THEORETICAL AND EXPERIMEN%L INVESTIGA-TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY Steam Generator SATURATED. FRACTURED POROUS MEDIUM. NUREG/LA-0113: PRELIMINARY ASSESSMENT OF PWR STEAM GEN-ERATOR MODELLING IN RELAPS/ MOD 3. Mrmoluminescent Dos 6 meter 9 REG 0837 V13 NO2: NRC TLD DIRECT RADIATION MONITORING Stress Corrosion Cracking NETWORK. Progress Report April June 1993 NUREG/C"4754: BOILING-WATER REACTOR INTERNALS AGING DEGRADATION STUDY. Phase 1. Title Ust NUREG/CR4048: PRESSURIZED-WATER REACTOR INTERNALS NUREG-0540 V15 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY AGING DEGRADATION STUDY. Phase 1. AVAILABLE.May1-31,1993. trJREG-0540 V15 N06: TITLE UST OF DOCUMENTS MADE PUBLICLY Structural Ahoy AVAILABE. June 1-30.1993. E NUREG/C46981: THE EFFECT OF ELECTRIC DISCHARGE MA-NUREG-0540 V15 N07: TITLE LIST OF DOCUMENTS MADE PUBLICLY CHINED NOTCHES ON THu FRACTURE TOUGHNESS OF SEVERAL AVAILABLE July 1-31, 1993. STRUCTURAL ALLOYS. Turbine Trip Structural Assessment NUREG/lA-0085: ASSESSMENT OF FULL POWER TURBINE TRIP NUREG/CR-5970- APPROXIMATE TECHNIQUES FOR PREDICTING START-UP TEST FOR C. TRfLLO I WITH RELAP5/ MOD 2, SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS UC). Turbulent Jet Ignit6on Structurni Reitability NUREG/CR4072: EXPERIMENTAL STUDY ON THE COMBUSTION BE. NUREG/C46015: STRUCTURAL AGING PROGRAM TECHNICAL HAV40R OF HYDROGEN-AIR MIXTURES WITH TURBULENT JET IG-PROGRESS FOR PERIOD JANUARY - DECEMBER 1992. NITION AT LARGE SCAD. Substance Abuse UK Numerics And implementation NUREG/CR4758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG/lA-0096 NUMERICS AND IMPLEMENTATION OF THE UK INDUSTRY. Annual Summary Of Program Performance Reports.CY HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE MODEL 1992. INTO RELAP5/ MOD 3. Surface Crack Vendor inspect 6on t'UREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT NUREG-0040 V17 NO2: UCENSEE CONTRACTOR AND VENDOR IN. OF CONSTRAINT IN FRACTURE. SPECTION STATUS REPORT. QuarteMy Report. April 4une 1993.(White Surfactant Waste Reduction NUREG/CR4081: ENHANCED REMOVAL OF RADIOACTIVE PARTI-Vertical Tube CLES BY FLUOROCARBON SURFACTANT SOLUTIONS. NUREG/lA-0100 ASSESSMENT OF CCFL MODEL OF RELAPS/ MOD 3 AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE TESTS. System Analysis NUREG/CR4065: SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR. Vibration NUREG/CR4078 ANALYSIS OF CRACK INITIATION AND GROWTH IN System Des 6gn THE HIGH LEVEL VIBRATION TEST AT TADOTSU. NUREGiCR-6082. DATA COMMUNICATIONS. Yucca Mountain System Failure NUREG/CR4021: A LITERATURE REVIEW OF COUPLED THERMAL. NUREG/CR 5993 VD1: METHODS FOR DEPENDENCY ESTIMATION HYDROLOGIC-MECHAN ICAL-CHEMIOAL PROCESSES PERTINENT AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA DATA STATISTICS. Summary Report. MOUNTAIN.

t i NRC Originating Organization Index (Staff Reports) j 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-i 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. ADVtSORY COMMITTEE (S) NUREG-1272 V07 N01: OFFICE FOR ANALYSIS AND EVALUATION ADVISORY COMMITTEE ON NUCLEAR WASTE OF OPERATIONAL DATA.1992 Anrmal Report. Power Reactors. NUREG-1423 V04: A COMPILATION OF REPORTS OF THE ADVISO-RY COMMITTEE ON NUCLEAR WASTE. July 1992 - June 1993. EDO OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM (POST 861109) ATOMIC SAFETY BOARD (S) & PANEL (S) DivislON OF INFORMATION SUPPORT SERVICES (POST 890205) NUREG-0910 R02 S01: NRC COMPREHENSIVE RECORDS DISPOSI-NRG VOS i SAF UCENSING BOARD PANEL ANNUAL REPORT. Fiscal Year 1992. EDO - OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) REGION 1 (POST 820201) DIVISION OF LOW-LEVEL WASTE MANAGEMENT & DECOMMISSION-l NUREG 0837 V13 NO2: NRC TLD DIRECT RADIATION MONITORING ING (POST 870413) NETWORK. Progress Report April-June 1993. NUREG-1476: FINAL ENVIRONMENTAL IMPACT STATEMENT TO REGION 2 (POST 820201) CONSTRUCT AND OPERATE A FACluTY TO RECEIVE, STORE, NUREG/CR 5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED AND DISPOSE OF 11E.(2) BYPRODUCT MATERIAL NEAR INSPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER CUVE, UTAH. Docket No. 40-8989. Envrocare Of Utah,Inc. PLANT. OPERATIONS BRANCH OFC OF ENFORCEMENT (POST 870413) NUREG-0525 V02 R01: SAFEGUARDS

SUMMARY

EVENT UST NUREG 0940 V12 NO2: ENFORCEMENT ACTIONS: SIGNIFICANT AC-(SSEL). January 1,1990 Through Decernber St.1992. TIONS RESOLVED.Ouarterty Progress Report.Apni-June 1993. EDO F E I T ( E D 13 & POST 990205) OF EOT GE ERAL COU E ST 860701) p NUREG 1145 V09: U.S. NUCLEAR REGULATORY COMMISSION NUREG-0386 006 R07: UNITED STATES NUCLEAR REGULATORY 1992 ANNUAL REPORT. COMMISSION STAFF PRACTICE AND PROCEDURE DIVISION OF FREEDOM OF INFORMATION & PUBLtCATIONS SERV. DIGEST. Commission, - Appeal Board And Ucensing Board j ICES (POST 890205 Decisions. July 1972 - Septernber 1992. NUREG 0304 V18 NO2: REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL). Cornpilauon For Second Quarter EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405) 1993.Apnl-June. DIVISION OF REGULATORY APPLICATIONS (POST 870413) NUREG-0540 V15 N05: TITLE LIST OF DOCUMENTS MADE PUBUC. NUREG-0713 V13: OCCUPATIONAL RADIATION EXPOSURE AT LY AVAILABLE.May1-31,1993. COMMERCIAL NUCLEAR POWER REACTORS AND OTHER NUREG-0540 V15 N06: TITLE UST OF DOCUMENTS MADE PUBLIC. FACluTIES.1991. Twenty-Fourth Annual Report. LY AVAILABLE. June 1-30,1993. NUREG-1400- AIR SAMPUNG IN THE WORKPLACE. Final Report 4 NUREG-0540 V15 N07: TITLE UST OF DOCUMENTS MADE PUBLIC-DIVISION OF SAFETY ISSUE RESOLUTION (POST 880717) LY AVAILABLE. July 1-31,1993. NUREG-1453: REGULATORY ANALYSIS FOR THE RESOLUTION OF 1 NUREG-0750 V37101: INDEXES TO NUCLEAR REGULATORY COM-GENERIC ISSUE 142: LEAKAGE THROUGH ELECTRICAL ISOLA-MISSION ISSUANCES. January-March 1993. TORS IN INSTRI.tMENTATION CIRCUlTS. NUREG-0750 V37 N04: NUCLEAR REGULATORY COMMISSION IS-NUREG 1461: REGULATORY ANALYSIS FOR THE RESOLUTION OF NURE V7N RE U TORY COMMISSION IS-i NUREG 1463: REGULATORY ANALYSIS FOR THE RESOLUTION OF NURE 7 NO N CL RE TORY COMMISSION IS-GENERIC SAFETY ISSUE 105: INTERFACING SYSTEM LOSS-OF. SUANCES FOR JUNE 1993.P es 419-515 NUREG-0750 V38 N01: NUCLEIR REGULATORY COMMISSION IS-COOLANT ACCIDENT IN UGHT-WATER REACTORS. SUANCES FOR JULY 1993.Pages 1 ADVANCED REACTORS BRANCH (POST 910830) NUREG 0936 V12 NO2: NRC REGULATORY AGENDA.Ouarterty NUREG/CP-0129 PROCEEDINGS OF THE WORKSHOP DN PRO-Report, April-June 1993. GRAM FOR EUMINATION OF REQUIREVENTS MARGINAL TO SAFETY. ) EDO - OFFICE OF THE CONTROLLER (PRE 820418 & POST 890205) OFFICE OF THE CONTROLLER (POST 890205) EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 900428) NUREG 1470 V02: CHIEF FINANCIAL OFFICER'S ANNUAL REPORT PROJECT DIRMTORATE 111-3 - 1993. NUREG/CR.5829 AUXILIARY FEEDWATER SYSTEM RISK-BASED INSPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR OFFICE OF STATE PROGRAMS PRE B70413 & POST 911117) POWER PLANT DfVISION OF SYSTEMS SAFETY & ANALYSIS (POST 921004) I U EG 1479 E S 01 TWO SHOPS: STATE RADI-NUREG-1449: SHUTDOWN AND LOW-POWER OPERATION AT NU-ATION CONTROL PROGRAMS DEVELOPING AND AMENDING CLEAR POWER PLANTS IN THE UNITED STATES. Final Report. REGULATIONS AND FUNDING. DIVISION OF REACTOR INSPECTION & UCENSEE PERFORMANCE EDO-OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL (POST 921004) DATA NUREG-0040 V17 NO2: UCENSEE CONTRACTOR AND VENDOR IN. OFFICE FOR ANALYSIS & EVALUATION OF E,PERATIONAL DATA, DI-SPECTION STATUS REPORT, Quarterty Report.Apribiune RECTOR 1993.(White Book) NUREG-0090 V16 N01: REPORT TO CONGRESS ON ABNORMAL NUREG-1214 R12: HISTORICAL DATA

SUMMARY

OF THE SYSTEM-OCCURRENCESJanuary-March 1993. ATIC ASSESSMENT OF LICENSEE PERFORMANCE. I 31

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NRC Originating Organization Index (International Agreements) This index lists those NRC organizations that have published international agreement re-l 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 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 NUCLEAR REGULATORY RESEARCH (POST 820405) NUREG/lA-0107: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A OFFICE OF NUCLEAR REGULATORY RESEARCH (POST B60720) LOAD REJECTION FROM 100% TO 50% POWER IN THE VAN-NUREG/1A-0091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A DELLOS ll NUCLEAR POWER PLANT, NATURAL CIRCULATION EXPERIMENT IN NUCLEAR POWER NUREG/lA 0112. ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN-PLANT BORSSELE. REFLOOD EXPERIMENTS. NUREG/lA-0096: NUMERICS AND IMPLEMENTATION OF THE UK NUREG/lA-0113: PRELIMINARY ASSESSMENT OF PWR STEAM HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE GENERATOR MODELLING IN RELAPS/ MOD 3. MODEL INTO RELAPS/ MOD 3. NUREG/lA-0126 2D/3D PROGRAM WORK

SUMMARY

REPORT. NUREG/lA 0100: ASSESSMENT OF CCFL MODEL OF RELAPS/ NUREG/lA4127: REACTOR SAFETY ISSUES RESOLVED BY THE MOD 3 AGAINST SIMPLE VERTICAL TUDES AND ROD BUNDLE 2D/3D PROGRAM. TESTS. OlVISION OF ENGINEERING (POST 870413) NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 91.B USING NUREG/lA 0085: ASSESSMENT OF FULL POWER TURBINE TRIP RELAPS/ MOD 3. START-UP TEST FOR C. TRILLO f WITH RELAP5/ MOD 2. 1 i t i f 4 33

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m,,-----w; -e.,4 ww- .-ew_sa c a s --m-33,m.4 3,_ w.4 da Amm &&,& hums-M.mg e.m a h s -J ;.6,ag,, e m,,m, A A 3h.-4.. pip-- -- - - -*e pw wmae Em p-l d 1 l. 1-l i 1 4 l 4 4 i l .i t I 1 l l i i i l e 8 P W i i i

. =. 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. If further information is needed, refer to the main citation by the NUREG/CR number. EDO-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CR 5817 V03 N2 NRC HIGH-LEVEL RADIOACTIVE WASTE DIVISION OF HIGH-LEVEL WASTE MANAGEMENT (POST 870413) RESEARCH AT CNWRA. July-December 1992. NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL-NUREG/CR-5927 V01: EVALUATION OF A PERFORMANCE AS-HYDROLOGIC-MECH ANICAL-CHEMICAL PROCESSES PERTI-SESSMENT METHODOLOGY FOR LOW-LEVEL RACCACTIVE NENT TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT WASTE DISPOSAL FACluTIES. Evaluation Of Modeling Approaches. 1 YUCCA MOUNTAIN-NUREG/CR-5943: SENSITIVITY ANALYSIS AND BENCHMARKING yg(IIC UC RE OR ESEARCH (POST 820405) NUREG/CR 598 MA ET TN EXT ITY DOSI-g NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING METERS-PILOT TEST. i IN LIGHT WATER REACTORS. Sermannual Report, October 1992 NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA-March 1993. TlON OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY NUREG/CR-4744 V07 N2 LONG-TERM EMBRITTLEMENT OF CAST SATURATED, FRACTURED POROUS MEDIUM. DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUREG/CR4050: RADIATION EXPOSURE MONITORING AND IN-L Report.ApnLSeptember 1992. FORMATION TRANSMITTAL (REMIT) SYSTEM. User's Manual NUREG/CR-5404 V02: AUXILIARY FEEDWATER SYSTEM AGING NUREG/CR-6073: LYSIMETER LITERATURE REVIEW. STUDY. Phase i Followon Study. NUREG/CR-6081: ENHANCED FIEMOVAL OF RADIOACTIVE PARTI-NUREG/CR-5754: DOILING-WATER REACTOR INTERNALS AGING CLES BY FLOOROCARBON SURFACTANT SOLUTIONS. DEGRADATION STUDY. Phase 1-OlVISION OF SAFETY ISSUE RESOLUTION (POST 880717) NUREG/CR-5778 V03: NEW YORK /NEW JERSEY REGIONAL SEIS-NUREG/CR-5928: ISLOCA RESEARCH PROGRAM. Foal Report. MIC NETWORsLFmal Repor1 For Apn~l 1985 - September 1992. NUREG/CR-5942: SEVERE ACCIDENT SOURCE TERM CHARAC-NUREG/CR 5782: PRESSURIZED THERMAL SHOCK PROBABluS-TERISTICS FOR SELECTED PEACH BOTTOM SEOVENCES PRE-i TIC FRACTURE MECHANICS SENSITIVITY ANALYSIS FOR DICTED BY THE MELCOR CODE-Y ANKEE ROWE REACTOR PRESSURE VESSEL DIVISON OF SYSTEMS RESEARCH POST 880717) NUREG/CR-5944: A CHARACTERIZATION OF CHECK VALVE DEG-NUREG/CR.5993 V01: METHODS FOR DEPENDENCY ESTIMATION RDT AND FAILURE EXPERIENCE IN THE NUCLEAR AND SYSTEM UNAVAILABluTY EVALUATION BASED ON FAIL-NUREG/CR-5969-J 'AND CTOD ESTIMATION EOVATIONS FOR URE DATA STATISTICS. Summary Report. NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATON FHALLOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. NUREG/CR 5970: APPRCXIMATE TECHNIQUES FOR PREDICTING AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAIL-t SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JCy URE DATA STATISTICS. Detailed Desenption And Applicatena. NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREAT. NUREG/CR4056: A FRAMEWORK FOR THE ASSESSMENT OF MENT OF CONSTRAINT IN FRACTURE. SEVERE ACCIDENT MANAGEMENT STRATEGIES. NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE MA-NUREG/CR4060: HYDROGEN MIXING STUDIES (HMS) ASSESS-CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVER-MENT MANUAL l AL STRUCTURAL ALLOYS. NUREG/CR4065: SYSTEMS ANALYSIS OF THE CANDU 3 REAC-1 NUREG/CR4015: STRUCTURAL AGING PROGRAM TECHNICAL TOR. PROC 5lESS FOR PERIOD JANUARY - DECEMBER 1992. NUREG/CR-6072: EXPERIMENTAL STUDY ON THE COMBUSTION NUREGICR4023. GENERIC ANALYSIS FOR EVALUATION OF LOW BEHAVIOR OF HYDROGEN-AIR MIXTURES WITH TURBULENT CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MAR-JET IGNITION AT LARGE SCALE. GINS AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERIALS. EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 800420) NUREG/CR4029 V01: AGING ASSESSMENT OF NUCLEAR AIR-DIVISION OF REACTOR CONTROLS & HUMAN FACTORS (POST TREATMENT SYSTEM HEPA FILTERS AND ADSORBERS Phase 1. 921004) j NUREG/CR4034: OKtAHOMA SEISMIC NETWORK. Final Report. NUREG/CR-6082: DAT A COMMUNICATIONS. NUREG/CR-6043 V01: AGING ASSESSMENT OF ESSENTIAL HVAC NUREGJCR4083: REVIEWING REAL-TIME PERFORMANCE OF NU- [ CHILLERS USED IN NUCLEAR POWER PLANTS Phase 1. CLEAR REACTOR SAFETY SYSTEMS. i NUREGICR-6048: PRESSURIZED. WATER REACTOR INTERNALS NUREG/CR4090: THE PROGRAMMABLE LOGIC CONTROLLER AGING DEGRADATON STUDY. Phase 1 AND ITS APPUCATON IN NUCLEAR REACTOR SYSTEMS ~ a NUREG/CR-6052; METHODOLOGY FOR FIELIABluTY BASED CON-DIVISION OF SYSTEMS SAFETY & ANALYSIS (POST 921004) DITION ASSESSMENT. Applicat on To Concrete Structures in Nucle

• NUREG/CR-5829 AUXIUARY FEEDWATER SYSTEM RISK-BASED i

j INSPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR NI :I 4058: VIRGINIA REGIONAL SEISMIC NETWORK.Fmal POWER PLANT, Re (1986 -1992) NUR CR-6078: ANALYSIS OF CRACK INITIATION AND GROWTH NUREG/CR4833 AUllLIARY FEEDWATER SYSTEM RISK-BASED IN THE HIGH LEVEL VIBRATION TE3T AT TADOTSU. INSPECTION GUIDE FOR THE H.B. ROBINSON NUCLEAR POWER I NUREG/CR4079: SEISMOLOGICAL INVESTIGATION OF EARTH. PLANT. ] OUAKES IN THE NEW MADRID SEISMIC ZONEJoel DIVISION OF RADIATION SAFETY & SAFEGUARDS (POST 921004) NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR Report. September 1986 - December 1992 NUREG/CR 6085: UNITED STATES SEISMOGRAPHIC NETWORK. POWER INDUSTRY. Annual Summary Of Program Performance DIVISION OF REGULATORY APPUCATIONS (POST 870413) Reports.CY 1992. NUREG/CR 3469 V07: OCCUPATONAL DOSE REDUCTION AT NU-DIVISION OF ENGINEERING (POST 921004) CLEAR POWER PLANTS: ANNOTATED BIBUOGRAPHY OF SE-NUREG/CR4049: PIPING BENCHMARK PROBLEMS FOR THE GEN-LECTED READINGS IN RADIATION PROTECTION AND ALARA. ERAL ELECTRIC ADVANCED BOILING WATER REACTOR. 35 4

..e n... Ai 4 4 4 l t i e 1 1 i s 4 4 1 l s i i l t I { t i r i i e h 1 I h o I k

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. I ADVANCED SYSTEMS TECHNOLOGY, INC. CENTER FOR NUCLEAR WASTE REGULATORY ANALYSES NUREG-1400: AIR SAMPLING IN THE WORKPLACE. Final Report. NUREG/CR-5817 V03 N2: NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. July-December 1992. 1 AMERICAN SOCIETY OF MECHANICAL ENGINEERS NUREG/CR-6021: A LITERATURE REVIEW OF COUPLED THERMAL. I NUREG/GR-0005 V02 P1: RISK-BASED INSPECTION-DEVELOPMENT HYDROLOGIC. MECHAN ICAL-CHEMICAL PROCESSES PERTINENT OF GUIDELINES.Ught Water Reactor (LWR) Nuclear Power Plant TO THE PROPOSED HIGH-LEVEL WASTE REPOSITORY AT YUCCA Components MOUNTAIN. NUREG/CR-6026: THEORETICAL AND EXPERIMENTAL INVESTIGA- { ARGONNE NATIONAL LABORATORY TION OF THERMOHYDROLOGIC PROCESSES IN A PARTIALLY NUREG/CR-4667 V16: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semaannual Report. October 1992 - March SATURATED, FRACTURED POROUS MEDIUM. N /CR-4744 V07 N2-LONG-TERM EMBRITTLEMENT OF CAST EG&G mAHO, WC. i DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUREG/CR 5928: ISLOCA RESEARCH PROGRAM Foal Report. j NUREG/CR 6073: LYSIMETER LITERATURE REVIEW. ReporMSeptember 992. ARtION A, UNIV. OF, TUCSON. AZ HARVARD SCHOOL OF PUBLIC HEALTH, BOSTON, MA NUREG/CP-0040: PROCEEDINGS OF WORKSHOP V: FLOW AND NUREG/CP-0130 V01: PROCEEDINGS OF THE 22ND DOE /NRC NU. TRANSPORT THROUGH UNSATURATED FRACTURED ROCK - RE-CLEAR AIR CLEANING CONFERENCE. Sessions

18. Held in

" ^' NU E 1 0 R ED GS OF THE 22ND DOE /NRC NU. j a sson u Hote Tuc n .Janua 7 0. 99 CLEAR AIR CLEANING CONFERENCE. Sessions 9-16 Held in i CATTELLE HUMAN AFFAIRS RESEARCH CENTERS Denver. Colorado. August 24-27,1992. NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTRY. Annual Summary Of Program Performance Reports.CY ILLINOIS. UNIV. OF, URBANA, IL l 1992. NUREG/CR-5969. J AND CTOD ESTIMATION EQUATIONS FOR SHAL. LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. CATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NUREG/CR-5970: APPROXIMATE TECHNIQUES FOR PREDICTING LABORATORY SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC). NUREG 1400: AIR SAMPLING IN THE WORKPLACE. Final Report. NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATMENT NUREG/CR-5758 V03: FITNESS FOR DUTY IN THE NUCLEAR POWER OF CONSTRAINT IN FRACTURE. INDUSTRY. Annual Summary Of Program Performance Reports.CY 1992. INTERIOR, DEPT. OF, GEOLOGICAL SURVEY i NUREG/CR-5829. AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR-6085: UNITED STATES SEISMOGRAPHIC NETWORK. SPECTION GUIDANCE FOR THE DAVIS-BESSE NUCLEAR POWER PLANT. JOHNS HOPKINS UNIV., B ALTIMORE, MD d NUREG/CR-5833: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN-NUREG/CR-6052: METHODOLOGY FOR RELIABILITY BASED CONDI-SPECTION GUIDE FOR THE H B ROBINSON NUCLEAR POWER TION ASSESSMENT. Application To Concrete Structures in Nuclear NUR CR-5989: PERFORMANCE TESTING OF EXTREMITV DOSI- [ METERS -PILOT TEST. LAMONT-DOHERTY GEOLOGICAL OBSERVATORY NUREG/CR-6029 V01: AGING ASSESSMENT OF NUCLEAR AIR-NUREG/CR-5778 V03 NEW YORK /NEW JERSEY REGIONAL SEISMIC ( N at Repod 6 @ GB5 - Semenh M92. NU E / 0 A N SSES T H AC CHILLERS USED IN NUCLEAR POWER PLANTS. Phase t-LAWRENCE LIVERMORE NATIONAL LABORATORY NUREG/CR-6082: DATA COMMUNICATIONS. EROOKHAVEN NATIONAL LABORATORY TIME PERFORMANCE OF NU-NUREG/CR-3469 V07: OCCUPATIONAL DOSE REDUCTION AT NU-C RSF SY CLEAR POWER PLANTS-ANNOTATED BIBLIOGRAPHY OF SELECT-NUREG/CR-6090: THE PROGRAMMABLE LOGIC CONTROLLER AND 1 NU E 943 S IVI AN N EN HMdRKING OF ITS APPLICATION IN NUCLEAR REACTOR SYSTEMS. j S OS MMAL WORAM NUR G R 59 3

M H D EN CY ESTIMATION AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE NUREG/CR-6060: HYDROGEN MIXING STUDIES (HMS) ASSESSMENT j

MANUAL DATA STATISTICS Summary Report NUREG/CR-5993 V02: METHODS FOR DEPENDENCY ESTIMATION f NAVY, DEPT OF AND SYSTEM UNAVAILABILITY EVALUATION BASED ON FAILURE NUREG/CFI 5969: J AND CTOD ESTIMATION EQUATIONS FOR SHAL. NURE /R l'P EC R POLM bR THE GEN-LOW CRACKS IN SINGLE EDGE NOTCH BEND SPECIMENS. NUREG/CR-5970: APPROXIMATE TECHNIQUES FOR PREDICTING ERAL ELECTRIC ADVANCED BOILING WATER REACTOR NUREG/CR-6078: ANALYSIS OF CRACK INITIATION AND G'ROWTH IN SIZE EFFECTS ON CLEAVAGE FRACTURE TOUGHNESS (JC). THE HIGH LEVEL VIBRATION TEST AT TADOTSU. NUREG/CR-5981: THE EFFECT OF ELECTRIC DISCHARGE MA-CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL EROWN UNIV., PROVIDENCE. RI STRUCTURAL ALLOYS. NUREG/CR-5971 CONTINUUM AND MICROMECHANICS TREATMENT OF CONSTRAINT IN FRACTURE. OAK R!DGE NATIONAL LABORATORY j NUREGiCR 5404 V02: AUXILIARY FEEDWATER SYSTEM AGING i C ALIFORNIA, UNIV. OF, LOS ANGELES, CA STUDY. Phase 8 Follow-On Study. i NUREG/CR-6056: A FRAMEWORK FOR THE ASSESSMENT OF NUREG/CR-5754: BOILING WATER REACTOR INTERNALS AGING l SEVERE ACCIDENT MANAGEMENT STRATEGIES. DEGRADATION STUDY. Phase 1. 37

38 Contractor Index NUREG/CR 5762 PRESSURIZED THERMAL SHOCK PROBABILISTIC NUREG/CR 5993 V01 METHODS FOR DEPENDENCY ESilMATION FRACTURE MECHANICS SENSITIVITY ANALYSIS F OA YANVEE AND SYSTEM UNAV AILAB!LITY EVALUATION BASED ON F AILURE ROWE REACTOR PRE SSURE VE SSEL DAT A ST ATISTICS Summary Report NURE G/CR 5942 SEVERE ACCIDE NT SOURCE TE RM CHARACTER-NUREG/CR 5993 V02 METHODS FOR DEPENDEt4CY ESilMATION ISTICS FOR SELECTED PEACH BOTTOM SEQUENCES PREDICTED AND SYSTEM UNAVAILABILITY EVALUATION BASED ON F AILURE BY THE MELCOR CODE DAT A ST ATISTICS DetaM Desenption And Applications NUREG/CR 5944 A CHARACTERIZATION OF CHECK VALVE DEGRA-NUREG/CR-6050 RADIATION EXPOSURE MONITORING AND INFOR. DATION AND F AILURE EXPERIENCE IN THE NUCLEAR POWER IN' MATION TRANSMITT AL (REMIT) SYSTEM User's Manual DUSTRY NUREG/CR-6060 HYDROGEN MIXING STUDIE S (HMS) ASSE SSMENT NURE G/CR40 t t STRUCTURAL AGING PROGRAM TECHNICAL MANUAL PROGRE SS FOR PERIOD JANUARY - DF CE MBER 1992 NUREG/CR-6023 GENERIC ANALYSIS l OR EVALUATION OF LOW SCIE NTECH,1NC. CHARPY UPPER-SHELF ENERGY EFFECTS ON SAFETY MARGINS NUREG/CP-0129 PROCE EDINGS OF THE WORKSHOP ON PROGRAM AGAINST FRACTURE OF REACTOR PRESSURE VESSEL MATERI-FOR EllMINATION OF REQUIREMENTS MARGINAL TO SAFETY. ALS NUREG/CR4048 PRE SSURIZED4 ATER REACTOR INTE RN ALS SOUTHWEST RESEARCH INSTITUTE NUR CR 5 ET O L Y O 4 RELIABILITY BASED CONDl. NUREG/CR4026 THEORETICAL AND EXPER: MENTAL INVESTIGA-TION ASSESSMENT, Applicateon To Concrete Structures in Nuclear TION OF THERMOHYDROLOGIC PROCESSES IN A PAR 11 ALLY plants SATURATED. F RACTURED POROUS MEDIUM NUREG/CR4065 SYSTEMS ANALYSIS OF THE CANDU 3 REACTOR ST. LOUIS UNIV, ST. louts, MO OKLAHOM A, UNIV. OF, NORM AN. OK NUREG /CR4079 SEISMOLOGICAL INVE STIG ATION OF EARTH-NUREG/CR4034 OKLAHOMA SEISMIC NETWORK Final Report OUAKES IN THE NEW MADRID SEISMIC ZONE Fenal W ^ RUSSIA NUREG/CR-6072 EXPERIMENT AL STUDY ON THE COMBUSTION BE' TEX AS A&M UNIV., COLLEGE ST ATION, TX HAVIOR OF HYDROGEN AIR MIXTURES WITH TURBULE NT XT IG-NUREG/CR-5971: CONTINUUM AND MICROMECHANICS TREATME NT NITION AT LARGE SCALE' OF CONSTRAINT IN FRACTURE SANDIA NATIONAL LABORATORIES NUREG/CR 5927 V01 EVALUATION OF A PERFORMAt:CE ASSESS. U.S. NAVAL ACADEMY, ANNAPOLIS, MD MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE NUREG/CR4961: THE EFFECT OF E LECTRIC DISCHARGE MA. DISPOSAL F ACILITiES Evaluation Of Modeling Approaches CHINED NOTCHES ON THE FRACTURE TOUGHNESS OF SEVERAL STRUCTURAL ALLOYS SCIENCE APPLICATIONS INTERN ATIONAL CORP,(FORMERLY SCIENCE APPLICATIONS, VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV, BLACKSBURG, NUREG4713 V13 OCCUPATIONAL RADIATION EXPOSURE AT COM. VA MERCIAL NUCL E AR POWER REACTORS AND OTHER NUREG/CR 4058 VIRGINIA RE GIONAL SEISMIC NETWORK Final F ACILITIE S 1991 Twenty Fourth Annual Report Report (1986 1992)

l l t International Organization index This index lists, in alphabetical order, the countries and performing organizations that pre-t p red the NUREG/lA reports listed in this compilation. Listed below each country and per-forming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number. P i JAPAN CONSEJO DE SEGURIDAD NUCLEAR JAPAN ATOMIC ENERGY RESEARCH INSTITUTE NUREG/lA4085: ASSESSMENT OF FULL POWER TURBINE TRIP NUREG/lA4126: 2D/3D PROGRAM WORK

SUMMARY

REPORT. ST ART. UP TEST FOR C. TRILLO I WITH RELAP5/ MOD 2. e NUREG/lA4127: REACTOR SAFETY ISSUES RESOLVED BY THE 20/3D PROGRAM. - THE NETHERLANDS NETHERLANDS ENERGY RESEARCH FOUNDATION ECN REPUBLIC OF KOREA NUREG/lA4091: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A KOHEA ELECTRIC POWER CORPORATION NATURAL CIRCULATION EXPERIMENT IN NUCLJAR POWER NUREG/LA-0100- ASSESSMENT OF CCFL MODEL OF RELAPS/ PLANT BORSSELE. - MOD 3 AGAINST SIMPLE VERTICAL TUBES AND ROD BUNDLE ' NUREG/lA-0112: ASSESSMENT OF RELAPS/ MOD 2 AGAINST ECN-TESTS. REFLOOD EXPERIMENTS. KOREA INSTITUTE OF NUCLEAR SAFETY r NUREG/lA-0103: ASSESSMENT OF BETHSY TEST 9.1.B USING UNITED KINGDOM i RELAPS/ MOD 3. NATIONAL POWER NUREG/lA4113: PRELIMINARY ASSESSMENT OF PWR STEAM SPAIN GENERATOR MODELLING IN RELAP5/ MOO 3. ASOCIACION NUCLEAR DE VANDELLOS WINFRITH TECHNOLOGY CENTRE NUREG/lA-0107: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NUREG/lA4096: NUMERICS AND IMPLEMENTATION OF THE UK LOAD REJECTION FROM 100% TO 50% POWER IN THE VAN-HORIZONTAL STRATIFICATION ENTRAINMENT OFF-TAKE l DELLOS 11 NUCLEAR POWER PLANT. MODEL INTO RELAPS/ MOD 3. .j t { I i r ? I l l i 39 l l

a m.,- w u,, _a a_, a _pa E p l l l P p p l I I, I a p l 0 I i a 't a F J s 3 m e I t I i I t 1 ) i l B I i i l s l l l s l l a 4 4 .i l l I

' Licensed Facility'index This index lists the facilities that were the subject of NRC staff'or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is needed, refer to the main citation - by the NUREG number. y .-i S 346 Deve-Besse Nucient Power Staton, Unit 1. NUREG/C&5829 .54277 Peach Bottom Atome Power Stakm, Urdt 2, . NUREG/%5942 ' - Toledo Edson Co. Phdadedphe Electne Ca NUREG/CR-5942 i 4J 8989 Enwacare of Utah, Inc., Salt Lake City. UT, NUREG 1476 50 278 Peach Bottom Atome Power Staton, Unit 3, ' d l 54261 H.B. Rotmson Plani, Umt 2, Caroina Power & - NUREG/CR-5833 Phdodolphe Elecinc Co. NUREG/CR 5782

l l

bght Ca 5429 Yankee Rows Nucteer Power Staten, Yankee, l 54171 Peach Bonom Atome Power Staten, Und 1, NUREGICR-5942 Atome Electne Co. ] PMadowse Electne Co. l-l -l l f -l t t ? l 5 h i [ a P I 41 l ?

I NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION

1. REPORT NUMBER (2 89)

(Asz6gned ty NRC, Add Vol., NRCM 1102. Supp., Rev., and Addendum Num-32oi, 3202 BIBLIOGRAPHIC DATA SHEET t='* - " aav-) (see instructions on uw, reve se) NUREG-0304 ' Vol.18 No.3 2.1nLE AND svamLE

3. DATE REPORT PUBUSHED Regulatory and Technical Reports (Abstract Index journal)

Compilation for - November 1993

• Diird Quarter 1993

4. rn OR GRANT NUuBER July-September

6. AUTHOR (6)

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1 7, PERIOD COVERE.D (inclusive Dates) i I July-September 1992

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (ff NRC, provide DMs6on, Off60s or Reg 6cn, U.S. Nuclear Regulatory Commissen, and t

maihng address; if contractor, provide riamo and malling address.) i Division of Freedom of Information and Publications Senices Office of Administration U.S. Nuclear Regulatory Commission .j Washington, DC 20555 r

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Same as 8, above,

10. SUPPLEMENTARY NOTES 11 ABSTRACT (200 words or less)

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,

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13. AVAILAfsuTY ST ATEMENT Unlimited 14, SECURITY CLASSIFICATION compilation g7, abstract index Unclassified (This ReporO Unclassified

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