ML20206E573

Abstracts for Publications in the NUREG-SERIES.Annual Compilation for 1998

ML20206E573
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Issue date:	04/30/1999
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References
NUREG-0304, NUREG-0304-V23-N02, NUREG-304, NUREG-304-V23-N2, NUDOCS 9905050167
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NUREG-0304 Vol. 23, No. 2 a

} Abstracts for Publications in the NUREG-Series m

Annual Compilation for 1998

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NUREG-0304 Vol. 23, No. 2 Abstracts for Publications in the , NUREG-Series Annual Compilation for 1998 [ O = / [l hDV U.S. Nuclear Regulatory Commission Office of the ChiefInformation Officer f,, ,,, % Washington, DC 20555-0001 ig ) 990430 PDR

AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications ) NRC publications in the NUREG series, NRC regu- NRC Public Document Room lations, and Title 10, Energy, of the Code of Federa/ 2120 L Street, N.W., Lower Level l Regulations, maybepurchasedfromoneof thefol- Washington, DC 20555-0001 l lowing sources: < htt p://www. nrc. gov /N R C/PDR/pdr1.htm > l 1 -800-397-4209 or locally 202-634-3273

1. The Supen.n tendent of Documents U.S. Government Printing Office Microfiche of most NRC documents made publicly RO. Box 37082 available since January 1981 may be found in the Washington, DC 20402-9328 Local Public Document Rooms (LPDRs) located in I i

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previous paragraph) or through:

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703 -487-4650 Publicly released documents include, to name a  : The NUREG series comprises (1) brochures few. NUREG-series reports; Federal Register no-(NUREG/BR-XXXX), (2) proceedings of confer- tices; applicant, licensee, and vendor documents ences (NUREG/CP-XXXX), (3) reports resulting and correspondence; NRC correspondence and from international agreements (NUREG/l A-XXXX), internal memoranda; bulletins and information no-(4) technical and administrative reports and books tices, inspection and investigation reports; licens- [(NUREG XXXX) or (NUREG/CR-XXXX)), and (5) ee event reports; and Commission papers and compilations of legal decisions and orders of the their attachments. Commission and Atomic and Safety Licensing Boards and of Office Directors' decisions under Documents available from public and specialtech-Section 2.206 of NRC's regulations (NUREG. nical libraries include all open literature items, such XXXX). as books, journal articles, and transactions, Feder-al Register notices, Federal and State legislation, A single copy of each NRC draft report is available and congressional reports. Such documents as free, to the extent of supply, upon written request theses, dissertations, foreign reports and transla-as follows: tions, and non-NRC conference proceedings may be purchased from their sponsoring organization. Address: Office of the Chief information Officer Reproduction and Distribution Copies of industry codes and standards used in a Services Section substantive manner in the NRC regulatory process U.S. Nuclear Regulatory Commission are maintained at the NRC Library, Two White Flint Washington, DC 20555-0001 North, 11545 Rockville Pike, Rockville, MD E-mail: < DISTRIBUTION @nrc. gov > 20852-2738. These standards are available in the Facsimile: 301 - 415- 2289 library for reference use by the public. Codes and A portion of NRC regulatory and technicalinforma- standards are usually copyrighted and may be purchased from the originating organization or, if tion is available at NRC's World Wide Web site: they are American National Standards, from-

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American National Standards institute All NRC documents released to the public are avail. 11 West 42nd Street able for inspection or copying for a fee, in paper, New York, NY 10036-8002 microfiche, or, in some cases, diskette, from the < http://www ansi.org > Public Document Room (PDR): 212- 642 -4900 t A year's subscription of this report consists of two semiannualissues. I

NUREG-0304 l Vol. 23, No. 2 Abstracts for Publications in the NUREG-Series l Annual Compilation for 1998 Date Published: April 1999 L. L. Stevenson, Project Manager Publishing Services Branch Omce of the ChiefInformation Omcer U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

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c [s V i l l CONTENTS Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... ... ... ... v index Tab Main Citations and Abstracts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 o Staff Reports e Conference Proceedings e Contractor Reports e G: ant Reports e - International Agreement Reports Secondary Report Number index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Personal Author index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , .. .. ............... 3 Subject Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........................ ... 4 NRC Originating Organization Index (Staff Reports) . . . . . . . . . . . . . . . . . . . . . . . .. . ....... 5 NRC Originating Organization index (International Agreements) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 NRC Contract Sponsor index (Contractor Reports) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 7 Contractor index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 8 Intemational Organization Index . . . . . . ... .. ...................................... .... 9 l Licensed Facility index . . . . . . . . . . . . . . . .... .. . .. .. . ...... ........... ...... . 10  ; l I til

PREFACE The U.S. Nuclear Regulatory Commission (NRC) compiles bibliographic data and abstracts for publications in the NUREG-series available to the public. The compilation is published semiannually. In the first listing, the bibliographic data and abstracts for these publications are sequenced according to their NUREG-series number: publicaticns including reports or brochures prepared by the staff designated (NUREG-XXXX) or (NUREG/BR-XXXX); conference proceedings designated (NURiEG/CP-XXXX); reports prepared by an NRC contractor designated (NUREG/CR-XXXX); and publications resulting from intomational agreements designated (NUREG/lA-XXXX). After the principal listing, nine other indexes list the reports by-Secondary Report Number Personal Author Subject NRC Originating Organization index for Staff Reports NRC Originating Organization Index for Intemational Agreement Reports NRC Contractor index for Contractor Reports Contractor intomational Organization Licensed Facility Staff-Prepared Publication (1) NUREG-1552 (report number); (2) Fire Barrier Penetration Seals in Nuclear Power Plants (report title); (3) Bajwa, C. S., West, K. S. (report authors); (4) Office of Nuclear Reactor Regulation (organizational unit of authors); (5) July 1996 (publication date); (6) 55 pp. (number of pages); (7) 9608230207 (NRC Document Control System accession number-for NRC use); (8) 89455:045 (the microfiche address-for NRC use). Staff-Preoared Brochure (1) NUREG/BR-0164, Rev. 2 (report number); (2) NRC: Regulator of Nuclear Safety (report title); (3) None (report author); (4) Office of Public Affairs (organizational unit of author); (5) April 1997 (publication date); (6) 24 pp (number of pages); (7) 9705020298 (NRC Document Control System accession number-for NRC use); (8) 92700:001-031 (the microfiche address-for NRC use). Contractor-Prepared Publicatip.0 (1) NUREG/CR-6279 (report number); (2) Application of Fracture Toughness Scaling Models to the Ductile-to-Brittle Transition (report title); (3) Joyce, J. A. (report author); (4) U.S. Naval v i

Academy (organizational unit of author); (5) January 1996 (publication date); (6) 42 pp (number of pages); (7) 9602220350 (NRC Document Control System accession number-for NRC use); (8) 87234:102 (the microfiche address-for NRC use). Conference Proceedinos (1) NUREG/CP-0149, V01 (report number); (2) Proceedings of the Twenty-Third Water Reactor Safety Information Meeting: Plenary Session, High Bumup Fuel Behavior, Thermal Hydraulic Research (report title); (3) Monteleone, S. (report author); (4) Brookhaven National Laboratory (organization that compiled the proceedings); (5) March 1996 (publication date); (6) 278 pp. (number of pages); (7) 9604150352 (NRC Document Control System accession number-for NRC use); (8) 87868:001 (the microfiche address-for NRC use). l Intemational Aoreement Publication I (1) NUREG/lA-0133 (report number); (2) Development, Implementation and Assessment of I Specific Closure Laws for inverted-Annular Film-Boiling in a Two-Fluid Model (report title); (3) l DeCachard, F. (report author); (4) Paul Scherrer Institute (organizational unit of author); (5) l October 1906 (publication date); (6) 103 pp (number of pages); (7) 9611190277 (NRC Document Control System accession r. umber-for NRC use); (8) 90823:249 (the microfiche address-for NRC use). 1 Some NRC reports in the NUREG-series are posted on NRC's World Wide Web site under the Reference Library icon on the home page: <http://www.nrc. gov >. Availability of NRC Publications Copies of these publications may be purchased from one of the following sources: l l

1. The Superintendent of Documents U.S. Govemment Printing Office P.O. Box 37082 ]

Washington, DC 20402-9328

   <http://www. access.gpo. gov /su_ docs >

202-512-1800

2. The National Technical Information Service ,

Springfield, VA 22161-0002 i

  <http://www.ntis. gov /ordemow>

703-487-4650 1 1 vi

Main Citations and Abstracts The list of publications in this Compilation are arranged by number, where NUREG-XXXX is an i NRC staff-originated publication, NUREG/CP-XXXX is an NRC-sponsored Conference proceedings, NUREG/CR XXXX is an NRC Contractor-prepared publication, NUREG/lA XXXX is cn intomational agreement publication, and NUREG/BR-XXXX is a staff-originated j' publication. The bibliographic information (see Preface for details) is followed by a brief _ Ebstract of this publication. NUROG4040 V21 N04: LICENSEE CONTRACTOR AND VEN- NUREG0304 V2t N04: REGULATORY AND TECHNICAL RE-DOR INSPECTION STATUS REPORT. Quarterly Re- PORTS (ABSTRACT INDEX JOURNAL). Annual Compilation port, October-December 1997.(While Book)

• Olhos of Nuclear For 1997.
• NRC - No Detailed AflWeNon Given. Aprg 1998.

Reactor Regulation (Poet 941001). Apr# 1998. 81pp. 92pp.9805180333. A3436:187. 9006180328. A3431:300. Ses NUREG-0304,V22,NO3 abstract. 8 mwom #w resuNs W perknned W NURSO0304 V23 N01: ABSTRACTS FOR PUBLICATIONS IN nonce Branch have ,". E E "' j N abon 998' during #w period h Ochbor through December 0' 3pp. 9810220299. A5499:104. Tru's joumel includes au formel reports in the NUREG-eeries  : NUREG0000 V3R N01: LICENSEE CONTRACTOR AND VEN- prepared by the NRC staff and contractors; proceednge of con-DOR INSPECTION STATUS REPORT. Quarterly Re- forences and workshape; as wen as intemeNoned agreement re-port, January March 1998.(White Book)

• OfAce of Nuclear Reno- ports. The entries in this complistion are trutavars for access by

. tor Regulation (Post 941001). September 1998. 133pp.                 tulo and abstract, secondary report number, personal author,       !

9810000194. A6202:198. stelect NRC organization for staff and intemenonal a0'es-This periodlost covers the resuNs of ir=rarmans pusormed by I monts, contractor, intemenonal orgs *mnan, and hooneed tecik i the NRC's QueNty Assurance, Vendor Irmpanhan and Mainte- ity. In Vol. 23 No.1, of NUREG-0304, the the was changed nonce Branch, that have been detributed to the inopooled orge- from " Regulatory and Technicei Reports (Abstreet index Jour-  ; niaallons durin0 the period from January through March 1998. nel)." l NUREG0000 Vat NOR: LICENSEE CONTRACTOR AND VEN- NUREG0303 V01 Rit: DIRECTORY OF CERTIFICATES OF DOR INSPECTION STATUS REPORT. Quarterly Report,Apr#- COMPLIANCE FOR RADIOACTIVE MATERIALS PACK- 1 June 1998.(White Book)

• Ollios of Nuoleer Remotor Regulamon AGES. Report Of NRC. Approved Packages.
• Omos of Nuclear  ;

(Post 941001). October 1998. 94pp. 9810300023. A6826:191. Meterial Soloty & ' Sofoguards October 1998. 588pp. ) This periodosi covers the resuNs of inapareans performed by 9811100306. A5778:001. l r.w NRC's Quatty Assurance, Vendor Irmparean and Mainte- The purpoos of tNs directory is to make available a conven-  ! nonos Branch, that have been detributed to the inspected orge. isnt source of informaton on packe0 ings approved by the U.S. nizadono durin0 the period from Apr# through June 1998. I'"daar Regulatory Commission. To assist in Idonglying peck-NLlREGt,000 V30: REPORT TO CONGRESS ON ABNORMAL eGin0, en W by M Number and 6 Cornoste of Compuence Number is included et the front of Volumes 1 and O :CURRENCES. Fiscal Year 1997,

• Omos for Analysis & Evel-o . ol-ar afa =8. 2= "d" *,ne"1"rs',fSedCU""',eL"e*'j"s%""l" ,e ,

res,g,ogy,'org=y Seaton 208 of the En Reorganization Act of 1974 identi- 88ng usom e pg p, g, fles an abnormal cocunerne (AO) as an unscheduled inaldent de~ or event that the Nucieer Regulatory Commiselon (NRC) dolor-mines to be signAcent from the eter$oint of pubbo heeNh or NUREG0303 VOR Rti: DIRECTORY OF CERTIFICATES OF esiety. The Federal Reports Elimination and Sunset Act of 1996 COMPLIANCE FOR RADIOACTIVE MATERIALS PACK-requires tiet AOs be reported to Congrees on an annual beels. AGES.Certtostes Of Compliance.

• Omos of Nuclear Material TNs report includes pues evade that NRC has determined to Soloty & Salo0uards. October 1998. 540pp. 9811100309.

 . be AOs during Hecal year 1997. This report addresses two AOs        A5779:214.

c3 NRC-Nooneed tecBWes. One involved an event at a nuclew See NUREG-0383,V01,R21 abstract power plant, and one involved materials overeuposure. The re- NUREG0003 V0S R1e: DIRECTORY OF CERTIFICATES OF part also addresses tour Agreement State AOs Two of those COMPLIANCE FOR RADIOACTIVE MATERIALS PACK-AOs irwolved overeuposures and two involved radiopharme- AGES. Report Of NRC-Approved Queuty Assurance Programs oeudoel rniesdminletrabons. In addhon, Appendia C of the report For Redonouve Matenals Packages.

• Offloe of Nuclear Material includes Sve events of loss of control of llooneed meterials. Safety & Safeguards. October 1998. 73pp. 9811100312.

NUREG0004 VER NOS: REGULATORY AND TECHNICAL RE- A5781:051. PORTS (ABSTRACT INDEX JOURNAL). CompNeuon For Third See NUREG-0383,V01,R21 abstract Quarter 1997 Ju!y *C .7 i.

• NRC - No Detened AfRilation NUREG 0308 DOS: UNITED STATES NUCLEAR REGULATORY

 ' Given. January 1998. 41pp. 9802100108. A2079:175.                   COMMISSION STAFF PRACTICE AND PROCEDURE Di-This joumel includes au formal reports in the NUREG series       GEST. Commission, Appeal Board And Ucensing Board Doch prepared by tie NRC eleN and contractors; proceedn0s of con-        sions. July 1972 - June 1997.

• Offloe of the General Counsel levences and workshops; as well es intemedonal agreement re- (Poet 880701). October 1998. 800pp. 9811230287. A5918:001.

ports. The entries in this compuseon are indexed for acosas by This 9th edmon of the NRC Practice and Procedure digest Nuo and abstract, eeoondary report number, personal author, contains a digest of a number of Commission, Atomic Seisty subject, NRC orgenlaston for eleft and intemenonal agree- and Licensing Appeal Board, and the Atomic Safety and Ucens-meres, contractor, intemellonel organizauon, and licensed tecil- ing Board decisions leeued during the period of July 1,1972 to ily. June 30,1997, interpreting the NRC's Rules. 1

2 Main Citation 3 and Abstracts NUREG-0430 V16: LICENSED FUEL FACILITY STATUS RE- See NUREG-0540,V19,N11 abstract. PORT. Inventory Difference Data. July 1, 1995 - June 30, 1996.(Gray Book 11) PHAM,T.N. Office of Nuclear Material Safe. NUREG-0540 V20 N05: TITLE LIST OF DOCUMENTS MADE ty & Safeguards. February 1998. 19pp. 9802250133. PUBLICLY AVAILABLE.May 1-31, 1998.

• NRC - No Detailed A2282:176. Affiliation Given. July 1998. 300pp. 9902090137. A6756:001.

NRC is committed to the periodic publication of licensed fuel See NUREG-0540,V19,N11 abstract. o t and tion any ed nyes NUREG-0540 V20 N06: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. June 1-30, 1998.

• NRC No Detailed information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effec-AMiation @en. August 1998. 377pp. 981006M A5290:173.

See NUREG-0540,V19,N11 abstract. tive kilogram of special nuclear material. NUREG-0430 V17: LICENSED FUEL FACluTY STATUS RE- NUREG-0640 V20 N07: TITLE UST OF DOCUMENTS MADE i PORT. inventory Difference Data. July 1, 1996 June 30, PUBLICLY AVAILABLE. July 1-31, 1998.

• NRC - No Detailed 1997.(Gray Book 11) PHAM,T.N. Office of Nuclear Material Safe- Affiliation Given. September 1998. 350pp. 9810060309, ty & Safeguards. November 1998. 18pp. 9811230334. A5291:192.

A5940 296. See NUREG-0540,V19,N11 abstract. See NUREG-0430,V16 abstract NUREG-0540 V20 N06: TITLE LIST OF DOCUMENTS MADE i NUREG-0625 V02 R06: SAFEGUARDS

SUMMARY

EVENT UST PUBLICLY AVAILABLE. August 1-31,1998.* NRC - No Detailed l (SSEL). January 1,1990 Through December 31,1997. Affiliation Given. October 1998. 275pp. 9810300012. A5627:001, j DANIS,A.A. Operations Branch. July 1998. 39pp. 9809210231. See NUREG-0540,V19,N11 abstract. A5136 242. The Safeguards Summary Event List provides brief sum- NUREG-0640 V20 N09: TITLE LIST OF DOCUMENTS MADE maries of hundreds of safeguards-related events involving nu- PUBLICLY AVAILABLE. September 1-30, 1998.* NRC - No De-clear material or facilities segulated by the U.S. Nuclear Regu- tailed Affiliation Given. Noverribe' 1998. 375pp. 9812240081. latory Commission. Events are described under the categories: A6312:001. Bomt>related, intrusion, Missing /Allegedty Stolen, Transpor. See NUREG-0540,V19,N11 abstract. tatiorerelated, Tampering /Vandahsm, Arson, Firearms-related, Rau,v,ve.a; Sabotage, NorFradiological Sabotage, and Mis- NUREG-0640 V20 N10: TITLE LIST OF DOCUMENTS MADE cellaneous. Because of the public interest, the Miscellaneous PUBLICLY AVAILABLE. October 1-31, 1998.

• NRC - No De-category also includes events reported involving source mate- tailed Affiliation Given. December 1998. 300pp. 9901270231.

rial, byproduct material, and natural uranium, which are exempt A6621:031. from safeguards requirements. Information in the event descrip- See NUREG-0540,u19,N11 abstract. tions was obtained from official NRC sources. NUREG-0M R04: NRC COLLECTION OF ABBREVIATIONS. * . NUREG-0640 V19 N11: TITLE LIST OF DOCUMENTS MADE July 19%.147pp. 9808120162. A4603:187. PUBUCLY AVAILABLE. November 1-30, 1997.

• NRC - No De- This collection of abbreviations in common use in the nuclear talled Affiliation Given. January 1998. 284pp. 9002100166. Industry and regulatory community was compiled from Nuclear ument is a monthly publication containing descrip- she b t uthors r and a tions of information received and generated by the U.S. Nuclear holders in identifying abbreviations for the numerous organiza-Regulatory Commission (NHC). This information includes (1) tional, scientific, and engineering terms that appear in NRC docketed material associated with civilian nuclear power plants printed and electronic information. The co,npilation is descriptive and other uses of radioactive materials and, (2) nondocketed rather than prescriptive. No one abbreviation is recommended to rnatorial received and generated by NRC pertinent to its role as the exclusion of another because the same abbreviation may a regulatory agency. The following indexes are included: Per- with equal validity apply to two or more terms.

sonal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents. NUREG-0713 V18: OCCUPATIONAL RADIATION EXPOSURE NUREG-0640 V19 N12: TITLE LIST OF DOCUMENTS MADE AT COMMERCIAL NUCLEAR POWER REACTORS AND PUBLICLY AVAILABLE. December 1-31, 1997.

• NRC - No De. OTHER FACILITIES,1996. Twenty-Ninth Annual Report. THOM-tailed Affiliation Given. February 1998. 315pp. 9803030348. AS M.L Division of Regulatory Applications (Post 941217).

HAGEMEYER,D.A. Science Applications intemational Corp. (for-A2413 001. See NUREG-0540,V19,N11 abstract. merly Science Applications, Inc.). February 1998. 300pp. 9803180118. A2609:001. NUREG-0640 V20 N01: TITLE LIST OF DOCUMENTS MADE This report summarizes the occupational radiation exposure PUBLICLY AVAILABLE. January 1-31, 1998. MORRIS E.B. NRC information that has been reported to the NRC's Radiation Ex-

  -   No Detailed Affiliation Given. March 1998. 327pp.              posure infortnation Reporting System (REIRS). The bulk of the 9803240332. A2683:001.                                             data presented in the report was obtained from the 1996 annual See NUREG-0540,V19,N11 abstract.                                 radiation exposure reports submitted in accordance with the re-NUREG-0640 V20 N02: TITLE LIST OF DOCUMENTS MADE                      quirements of 10 CFR 20.2206. The 1996 annual reports sub-PUBLICLY AVAILABLE. February 1 28, 1998.

• NRC - No De- mitted by about 284 licensees indicated that approximately tailed Affiliation Given. April 1998. 300pp. 9805050440. 138,310 individuals were monitored,75,139 of whom were mon-A3318:039. itored by nuclear power facilities. They incurred an average indi-See NUREG-0540,V10,N11 abstract. vidual dose of 0.1 rem (cSv) and an average measurable dose f about 029 rem kS4. Analyses of transent wMer data M NUREG-0640 V20 NO3: TITLE LIST OF DOCUMENTS MADE cam M 22,M8 Muels compleM we ass %nments at two PUBUCLY AVAILABLE. March 1-31,1998.* NRC - No Detailed or more licensees during the monitoring year. The dose distribu-Affiliation Given. May 1998. 390pp. 9806010321. A3571:044. tions are adjusted each year to account for the duplicate report-See NUREG-0540,V19,N11 abstract. ing of transient workers by multiple licensees. In 1996, the aver-NUREG-0540 V20 N04: TITLE LIST OF DOCUMENTS MADE age measurable dose calculated from reported data was 0.24 PUBLICLY AVAILABLE. April 1-30, 1998.
• NRC - No Detalled cSv (rem). The corrected dose distribution resulted in an ever.

Affiliation Given. June 1998. 352pp. 9807060350. A4010:001. age measurable dose of 029 cSv (rem).

Mein Citations End Abstracts 3 NUREG-0713 V19: OCCUPATIONAL RADIATION EXPOSURE Legal issuances of the Commission, the Atomic Safety and Li-AT COMMERCIAL NUCLEAR POWER REACTORS AND censing Board Panel, the Administrative Law Judges, and NRC OTHER FACILITIES 1997. Thirtieth Annual Report. THOM- Program Offices are presented. AS,M.L Division of Regulatory Applications (Pos.t 941217); HAGEMEYER.D.A. Science Applications intemational Corp. (for NUREG-0750 V46101: INDEXES TO NUCLEAR REGULATORY morty Science Appications, Inc.). November 1998. 300pp. COMMISSION ISSUANCES. July-Septernber 1997.

• NRC - No 9812090010. A6091:013. Detailed Affiliation Given. March 1998. 31pp. 9803270317.

This report summarizes the occupational exposure data that A2767:305. Digests and indexes for issuances of the Commission, the are maintained in the U.S. Nuclear Regulatory Commission's (NRC) Radiation Exposure Information and Report System Atomic Safety and Licensing Board Panel, the Administrative (REIRS). The bulk of the information contained in the report was Law Judges, the Directors' Decisions, and the Decisions on Pe-compiled from the 1997 annual reports submitted by six of the titions for Rulemaking are presented, seven categories of NRC licensees subject to the reporting re. NUREG-0750 V46102: INDEXES TO NUCLEAR REGULATORY quirements of 10 CFR 20.2206. Since there are no geologic re- COMMISSION ISSUANCES. July-December 1997.

• NRC - No positories for high level waste currently licensed, only six cat- Detailed Affiliatinn Given. April 1998. 49pp. 9805050382.

egories will be considered in this report. Annual reports for 1997 A3320:196. were received from a total of 296 NRC licensees, of which 109 See NUREG-0750,V46,101 abstract. were operators of nuclear power reactorr in commercial oper-ation. Compilations of the reports submitted by the 296 licens. NUREG-0750 V46 NO3: NUCLEAR REGULATORY COMMISSION eos indicated that 142,730 individuals were monitored,75,291 of ISSUANCES FOR SEPTEMBER 1997. Pages 49-193.

• NRC - l whom received a rnessurable dose (Table 3.1). The collective No Detailed Affiliation Given. January 1998. 151pp. '

dose incurred by these individuals was 19, 841 person-rem 9802180098. A2195:001. which represents a 9% decrease from the 1996 value. The num- See NUREG-0750,V45 abstract. ber of workers receiving a measurable dose also decreased, re-sulting in the average measurable dose of 026 rem for 1997. NUREG-0750 V46 N04: NUCLEAR REGULATORY COMMISSION The average measurable dose es defined to be the total colleo- ISSUANCES FOR OCTOBER 1997. Pages 195-256.

• NRC -

tive dose (TEDE) divided by the number of workers receiving a No Detailed Affiliation Given. Fetruary 1998. 69pp. measurable dose. These figures have been adjusted to account 9802180103. A2195:155 for transient reactor workers, in 1997, the annual collective dose See NUREG-0750,V45 abstract. per reactor for, light water reactor licensees (LWRs) was 157 NUREG-0750 V46 N05: NUCLEAR REGULATORY COMMISSION persordrem. This represents a 9% decrease from the value re- ISSUANCES FOR NOVEMBER 1997. Pages 257-285.

• NRC -

ported for 1996. The annual collective dose per reactor for boil- No Detailed Affiliation Given. February 1998. 35pp. ing water reactors (BWRs) was 205 persordrem and for pressur- 9802180107. A2195:224. Ized water reactors (PWRs), it was 132 persorwem. Analyses See NUREG-0750,V45 abstract. of transient worker data indicate that 31,065 individuals com-pleted work assignments at two or more licensees during the NUREG-0750 V46 N06: NUCLEAR REGULATORY COMMISSION monitoring year. The dose distributions are adjusted each year ISSUANCES FOR DECEMBER 1997. Pages 287-319.

• NRC -

to account for the duplicate reporting of transient workers by No Detailed Affiliation Given. March 1998. 40pp. 9803270320. multiple licensees. In 1997, the average measurable dose cal- A2755:195. culated from reported data was 0.22 rem. The corrected dose See NUREG-0750,V45 abstract. distribution resulted in an average dose of 0.26 rem. NUREG-0750 V47101: INDEXES TO NUCLEAR REGULATORY NUREG-0725 R13: PUBLIC INFORMATION CIRCULAF FOR COMMISSION ISSUANCES. January-March 1998.

• NRC - No SHIPMENTS OF IRRADIATED REACTOR FUEL
• Office of Detailed Affiliation Given. June 1998. 17pp. 9806190291.

Nuclear Material Safety & Safeguards. October 1998. 34pp. A3903:323. 9811240189. A5955:249. See NUREG-0750,V46,101 abstract. This circular has been prepared to provide information on the NUREG-0750 V47102: INDEXES TO WUCLEAR REGUt.ATORY l shipment of irradiated reactor fuel (spent fuel) subject to reguta- ' COMMISSION ISSUANCES.Jarwary-June 1998.

• NRC No tion by the U.S. Nuclear Regulatory Commission (NRC). It pro-Detailed Affiliation Given. Ser'. ember 1998. 68pp. 9809210238. i vides a brief description of spent fuel shipment safety and safe-A5143:079 guards requirements of general interest, a summary of data for See NUFkEG-0750,V46.8J1 abstract.

1979-1997 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The en- NUREG-0750 V47 N01: NUCLEAR REGULATORY COMMISSION closed route information reflects specific NRC approvals that ISSUANCES FOR JAfeUARY 1998. Pages 1-12.

• NRC No have been granted in response to requests for shipments of Detailed Affikation Ghen. March 1998. 18pp. 9803270335.

spent fuel. This publ6 cation does not constitute authority for car- A2755:146. riers or other persons to use the routes described to ship spent See NUREG-0750N45 abstract. fuel, other categories of nuclear waste, or other materials. NUREG-0750 V47 N02: NUCLEAR REGULATORY COMMISSION NUREG-0750 Cl04: INDEXES TO NUCLEAR REGULATORY ISSUANCES FN FEBRUARY 1998. Pages 13-56.

• NRC - No COMMISSION ISSUANCES. January 1,1991 through Decembet Detailed AffilQtion Given. April 1998. 51pp. 9805060094.

31, 1995.

• NRC - No Detailed Affiliation Given. November A3321:277.

1997,452pp. 9803260270. A2725:001. See NUREG-0750,V45 abstract. Digests and indexes for issuances of the Commission, the Atomic Safety and Licensing Appeal Panel, the Atomic Safety NUREG-0750 V47 NO3: NUCLEAR REGULATORY COMMISSION cnd Licensing Board Panet, the Administrative Law Judge, the ISSUANCES FOR MARCH 1998.Pages 57-75.

• NRC - No De-Directors' Decisions, and the Decisions on Petitions for Rule. tailed Affiliation Given. April 1998. 25pp. 9805180293.

maidng are presented. A3428:328. See N REG-0750,W5 abstract. NUREG-0750 V45: NUCLEAR REGULATORY COMMISSION ISSUANCES. Opinions And Decisions Of The Nuclear Regu- NUREG-0750 V47 N04: NUCLEAR REGULATORY COMMISSION latory Commission With Selected Orders. January-June 1997.

• ISSUANCES FOR APRIL 1998.Pages 77-260.
• NRC - No De-NRC - No Detailed Affiliation Given. December 1997. 526pp. tailed Affiliation Given. June 1998. 191pp. 9807060212.

9802200041. A2245:001. A4009:044.

                                                                                                                                       )
                                                                                                                                        )

l i

4 4 Main Citation] and Abstracts See NUREG-0750,V45 abstract. The report presents the safety priority rankt ; for generic safety issues related to nuclear power plants.1he purpose of NUREG-0750 V47 N06: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR MAY 1998.Pages 261-306.

• NRC - No De_ mese rankigs is to assist in me Umely and efficent anocatim talled Affiliation Given. July 1998. 53pp. 9807270363. of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety prionty A4378:307 See NUREG-0750,V45 abstract. rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of NUREG-0750 V47 N06: NUCLEAR REGULATORY COMMISSION risk to costs and otner impacts estimated to result if resolution ISSUANCES FOR JUNE 1998.Pages 307-408.
• NRC - No De- of the safety issues were implemented, and the consideration of tailed AfEation Given. July 1998. 10Bpp. 9808310098. uncertainties and other quantitative or qualitative factors. To the A4864:120. extent practical, estimates are quantitative.

See NUREG-0750,V45 abstract. NURN0750 V48101: INDEXES TO NUCLEAR REGULATORY NUREG-0936 V16 NO2: NRC REGULATORY AGEN-DA. Semiannual Report. July-December 1997. Office of Admin-COMMISSION ISSUANCES. July -September 1998.

• NRC No Detailed Affiliation Given. December 1998. 46pp. 9902090263.

istration, Director (Post 940714). February 1998. 68pp. , 9803050116. A2424:115. l A6783 241 See NUREG-0750,V46'101 abstract. The NRC Regulatory Agenda is a cornpilation of all rules on l which the NRC has recently completed action, or has proposed l NUREG-0750 V48 N01: NUCLEAR REGULATORY COMMISSION action, or is considering action, and all petitions for rulemaking I ISSUANCES FOR JULY 1998.Pages 1-38.

• NRC - No Detailed which have been received by the Commission and are pending Affiliation Given. September 1998. 44pp. 9810060167, disposition by the Comrnission. The Regulatory Agenda is up.

A5294:248. dated and issued semiannually. See NUREG-0750,V45 abstract. NUREG-0936 V17 N01: NRC REGU AGEN-NUREG-0750 V48 NO3: NUCLEAR REGULATORY COMMISSION DA.Semanmal RepodamaWune MB. Mce,LATORY of Admne ISSUANCES FOR SEPTEMBER 1998. Pages 119-182.

• NRC , Drector (Post 940714). July 1998. 62pp. 9808310089.

     . No Detailed Affiliation Given. November 1998. 70pp.               {a     g;p See       EM36,W6,N02 had N RE O        V4 abstract NUREG-0750 V48 N04: NUCLEAR REGULATORY COMMISSION                        NUREG-0940 V16 N2 P1: ENFORCEMENT ACTIONS: StGNIFI-ISSUANCES FOR OCTOBER 1998. Pages 183-258.

• NRC . CANT ACTIONS RESOLVED INDIVIDUAL AC-No Detailed Affiliaton Given. December 1998. 82pp. TIONS. Semiannual Progress Report, July-December 1997.
• Ofc 9902030080. A6713:157. of Enforcement (Post 870413). April 1998. 416pp. 9805210430.

See NUREG-0750.V45 abstract A3499:001. This compilation summarizes significant enforcement actions NUREG4837 V17 NO3: NRC TLD DIRECT RADIATION MONI- at have Mn res@ed durng me pefod puty - Damber TORING NETWORK. Progress Report. July-September 1997, 1997) and includes copios of Orders and Notees of Violations STRUCKMEYER,R. Regon 1 (Post 820201). January 1998. sent h the Nuclear Regdatory Commisse to Mvduals wim 229pp' 9801260119. A1892:001 respect to hse enfounwnt achons. H is aMpated mat me This report provides the status and results of the NRC information in th,si publication will be widely disseminated to Thermoluminescent Dosimeter (TLD) Direct Radiation Moni- managers and ernplopes engagM M aMes brud by me toring Network. It presents the radiation levels rneasured in the NRC. The Commission believes this information may be useful vicinity of NRC licensed facilities throughout the country for the to licensees in making employment decisions. third quarter of 1997, NUREG-0910 R03: NRC COMPREHENSIVE RECORDS DIS- NUREG-0940 V16 N2 P2: ENFORCEMENT ACTIONS: SIGNIF1-CANT ACTIONS RESOLVED REACTOR LICENS-POSITION SCHEDULE.

• NRC No Detailed Affiliation Given.

February 1998. 380pp. 9803190155. A2629:090. EES. Semiannual Progress Report. July-December 1997.

• Ofc of Title 44 United States Code, "Public Printing and Docu. Enforcement (Post 870413). April 1998. 320pp. 9805180409.

l ments," regulations issued by the General Service Administra. A3433:027. tion (GSA) in 41 CFR Chapter 101, Subchapter B, " Manage. This compilation summarizes signifcant enforcement actions I rnent and Use of Information and Records." and regulations that have been resolved during the period (July December issued by the National Archives and Records Administration 1997) and includes copies of letters, Notices, and Orders sent (NARA) in 36 CFR Chapter Xil, Subchapter B, " Records Man. by the Nuclear Regulatory Commission to reactor licensees with agement," require each egency to prepare and issue a com. respect to these enforcement actions. It is anticipated that the l' prehensive records disposition schedule that contains the NARA information in this publication will be widely disseminated to approved records disposition schedules for records unique to managers and employees engaged in activities licensed by the l NRC, so that actions can be taken to improve safety by avoiding l the agency and contains the NARA's General Records Sched-ules for recortis common to several or all agencies. The ap- future violations similar to those described in this publication. l proved records disposition schedules specify the appropriate du- NUREG-0940 V16 N2 P3: ENFORCEMENT ACTIONS: SIGNIFi-ration of retention and the final disposition for records created CANT ACTIONS RESOLVED MATERIAL LICENS-or maintained by the NRC. NUREG-0910, Rev. 3, contains EES. Semiannual Progress Report, July-December 1997.

• Ofc of NRC's Comprehensive Records Disposition Schedule," and the Enforcement (Post 870413). April 1998. 407pp. 9806010324.

original authortzed approved citation numbers issued by NARA. A3570:001 Rev. 3 incorporates NARA approved changes and additions to This compilation summarizes significant enforcement actions l the NRC schedules that have been impamented since the last that have been resolved during the period (July - December revision dated March, 1992, reflects recent organizational 1997) and includes copies of letters, Notices, and Orders sent changes implemented at the NRC, and includes the latest by the Nuclear Regulatory Commission to material licensees version of NARA's General Records Schedule (dated August with respect to these enforcement actions. It is a7ticipated that 1995). the information in this pubication will be wdely disseminated to NUREG4933 $22: A PRIORITIZATION OF GENERIC SAFETY managers and employees engaged in activities licensed by the ISSUES. EMRIT,R. Division of Engineering Technology (Post NRC, so that actions can be taken to improve safety by avoiding 941217). March 1998. 398pp. 9804080010. A2923:001. future violations similar to those described in this publication.

Main Citations and Abstracts 5 NUREG-0940 V17 N1 P1: ENFORCEMENT ACTIONS: SIGNIFI- nuclear power plants. This revision was initiated to improve the CANT ACTIONS RESOLVED INDIVIDUAL AC- reporting guidelines related to 10 CFR 50.72 and 50.73 and to TIONS. Semiannual Progress Report, January-June 1998.

• Ofc consolidate these guidelines into a single reference document.

of Enforcement (Post 870413). November 1998. 375pp. A first draft of this document was noticed for pubic comment in 9812220093. A6310:001. the Federal Register on October 7,1991 (56 FR 50598). A seo-This cornpilation summarizes signifcant enforcement actions ond draft was noticed for comment in the Federal Register on that have been resolved during the period (January - June February 7,1994 (59 FR 5614). This document updates and su-1998) and includes copies of Orders and Notees of Violation persedes NUREG-1022 and its Supplements 1 and 2 (published sent by the Nuclear Regulatory Commission to individuals with in September 1983, February 1984, and September 1985, re-respect to these enforcement actions. It is anticipated that the spectively). It does not change the reporting requirements of 10 l information in this publication will be widely dissemiruted to CFR 50.72 and 50.73. managers and employees engaged in activities hcensed by the NRC. The Commission believes this information may be useful to licensees in making employment decisions. NUREG-1100 V14: BUDGET ESTIMATES. Fiscal Year 1999.

• Di- '

vision of Budget & Analysis (Post 890205). February 1998. l NUMEG 0940 V17 N1 P2: ENFORCEMENT ACTIONS: SIGNIFl. 176pp. 9802250137. A2282:001. l CANT ACTIONS RESOLVED REACTOR LICENS- This report contains the fiscal year budget justifcation to Con. EES. Semiannual Progress Report January-June 1998.

• Ofc of gress. The budget provides estimates for salaries and expenses Enforcement (Post 870413). November 1998. 275pp, and for the Offee of the inspector General for fiscal year 1999. ,

9812090015. A0000:001. l This compilation summarizes significant enforcement actions NUREG-1122 R02: KNOWLEDGE AND ABILITIES CATALOG that have been resolved during the period (January - June FOR NUCLEAR POWER PLANT OPERATORS. Pressurized 1998) and includes copies of letters, Notices, and Orders sent Water Reactors.

• Offee of Nuclear Reactor Regulation (Post by the Nuclear Regulatory Commission to reactor licensees with 941001). June 1998. 500pp. 9807270351. A4379:001.

respect to these enforcement actions. It is anticipated that the This document provides the basis for development of content-information in this publication will be widely disseminated to valid licensing examinations for reactor operators and serior re-managers and employees engaged in activities licensed by the actor operators. The examinations developed using this docu-NRC, so that actions can be taken to improve safety by avoiding ment will cover those topes hsted under Title 10, Code of Fed-future violdions similar to those desenbod in this publication. eral Regulations, Part 55, " OPERATORS' LICENSES." The NUREG 0040 V17 N1 P3: ENFORCEMENT ACTIONS; SIGNIFl. PWR catalog contains approximately 5,100 imowledge and abil-CANT ACTIONS RESOLVED MATERIAL LICENS. ity (K/A) statements for reactor operators and senior reactor op-EES. Semiannual Progress Report. January-June 1998.

• 01c of erators. Each K/A statement has been rated for its imoortance Enforcement (Post 870413). November 1998. 325pp. to safe operation of the plant in a manner ensuring personnel 9812240085. AB312:315. and public health and safety. THE PWR K/A catalog is orga-This compilation summarizes significant enforcement actions nized into six major sections: 1) Organization of the Catalog; 2) that have been resolved during the period (January - June Plant Wide Generic Knowledge and Abilities; 3) Plant Systems 1998) ard includes copies of letters, Notices, and Orders sent Grouped by Safety Function; 4) Emergency and Abnormal Plant by the Nuclear Regulatory Commission to material licensees Evolutions; 5) Components; and 6) Theory, with respect to these enforcement actions. It is anticipated that the information in this pubhcation will be widely disseminated to NUREG-1123 R02: KNOWLEDGE AND ABILITIES CATALOG managers and employees engaged in activities licensed by the FOR NUCLEAR POWER PLANT OPERATORS Boiling Water NRC, so that actioits can be taken to improve safety by avoiding Reactors.
• Office of Nuclear Reactor Regulation (Post 941001).

future violations similar to those described in this publication. June 1998. 500pp. 9807270345. A4377:001. { NUREG 0080 V01 N04: NUCLEAR REGULATORY LEGISLA- Ms docunwnt prwides the basis for development of content-TlON.104th Congress.

• Offee of the General Counsel (Post valid hcensing examinations for reactor operators and senior re-860701). December 1997. 594pp. 9804160178. A2992:160. actor operators. The examinations developed using this docu-This document is a compilation of nuclear regulatory legista, ment win cover those topes listed under Title 10, Code of Fed-tion and other relevant material through the 104th Congress,2d eral Regulations, Part SS, " OPERATORS' LICENSES." The Session. This compilation has been prepared for use as a re. BWR catalog contains approximately 7,000 knowledge and abil-source document, which the NRC intends to update at the end ity (K/A) statements for reactor operators and senior reactor op-of every Congress. The contents of NUREG-0980 include The erators. Each K/A statement has been rated for its importance Atomic Energy Act of 1954, as amended; Energy Reorganiza- to safe operahon of #w plant in a mamer ensunng persomel tion Act of 1974, as amended, Uranium Mill Tailings Radiation and public health and safety. The BWR K/A catalog is organized Control Act of 1978; Low Level Radioactive Waste Policy Act; into slx ma}or sections: 1) Organization of the Catalog; 2) Plant Nuclear Waste Policy Act of 1982; and NRC Authorization and Wide Generic Knowledge and Abilities; 3) Plant Systems Appropriations Acts. Other materials included are statutes and Grouped by Safety Function; 4) Emergency and Abnormal Plant treatees on export licensing, nuclear non-proli'eration, and envh Evolutions; 5) Compnnents; and 6) Theory.

ronmental protection. NUREG-1125 V19: A COMPILATION OF REPORTS OF THE AD-NUREG-0980 V02 N04: NUCLEAR REGULATORY LEGISLA- VISORY COMMITTEE ON REACTOR SAFEGUARDS.1997 An-TION.104th Congress.

• Offee of the General Counsel (Post 860701). December 1997. 523pp. 9804160185. A2991001.

nual.

• ACRS - Advisory Committee on Reactor Safeguards.

See NUREG-0980,V01,N04 abstract. April 1998. 222pp. 9806010317. A3572:070. This compilation contains 67 ACRS reports submitted to the NUREG-1022 R01: EVENT REPORTING GUIDELINES 10 CFR Commission, or to the Executive Director for Operations, during 50.72 AND 50.73. ALLISON,D.P.; HARPER,M.R.; JONES W.R.; calendar year 1997. It also includes a report to the Congress on d al. Office for Anatysis & Evaluation of Operational Data, Di- the NRC Safety Research Program. All reports have been made rector. January 1998.175pp. 9802100113. A2079:001. available to the public through the NRC Public Document Room, Revision 1 to NUREG-1022 clarifies the immediate notifcation the U. S. Library of Congress, and the latemet at http// requirements of Title 10 of the Code of Federal Regulations, www.nrc. gov /ACRSACNW The reports are categorized by the Part 50, Section 50.72 (10 CFR 50.72), and the 30. day written most appropriate Dener6c subject area and by chronological licensee event report (LER) requirements of 10 CFR 50.73 for order within the subject area.

l 6 Main Citation 3 and Abstracts l l l NUREG-1187 V01: PERFORMANCE INDICATORS FOR OPER- power reactors and presents an overview of the operating expe-ATING COMMERCIAL NUCLEAR POWER REACTORS. Data rience of the nuclear power industry from the NRC perspective, Through September 1997.

• Offce for Analysis & Evaluation of including comments about trends of some key performance Operational Data, Director. January 1998. 485pp. 9802110145, measures. The report also includes the principal findings and A2004:001.

issues identrfied in AEOD studies over the past year and sum-This Nuclear Regulatory Commission (NRC) report provides marizes information from such sources as licensee event reports performance indicator data, accounting for the different oper- and reports to the NRC's Operations Center. NUREG-1272, Vol. ational conditions, through September 1997 for 109 reactors. II, No. 2, covers nuclear materials and presents a review of the There are eight NRC Performance Indicators for Operating events and concems during 1997 associated with the use of li-Commercial Nuclear Power Plants: (1) automatic scrams while censed material in nonreactor applications, such as personnel critcal. (2) safety system actuations, (3) significant events, (4) overexposures and medical misadministrations. Both reports safety system failures, (5) forced outage rate, (6) equipment also contain a discussion of the incident Investigation Team pro-forced outages per 1000 commercial critcal hours, (7) collective gram and summarizes both the incident investigation Team and radiation exposure, and (8) cause codes. This report is based Augmented inspection Team reports. Each volume contains a on data extracted from Licensee Event Reports (LERs) sub- list of the AEOD reports issued from CY 1980 through 1997, mitted in accordance witn 10 CFR 5073, immediate notifications NUREG-1272, Vol.11, No. 3, covers technical training and pre-to the NRC Operations Center in accordance with 10 CFR sents the activities of the Techncal Training Center in support 50.72, monthly operating reports in accordance with plant tech- of the NRC's mission in 1997. nical specifcations, and screening of operating experience by NUREG 1272 V11 NO2: OFFICE FOR ANALYSIS AND EVALUA-NRC staff. Radiation exposure data are obtained from the Insti- TION OF OPERATCNAL DATA.1997 Annual Report (Nuclear tute of Nuclear Power Operations (INPO). Graphical presere Materia 4

• Office for Ardysis & Evaluation of Operational tations of each plant a data, including trends and deviatons Data, Director. November 1998. 108pp. 9901270174.

analyses are provided, as well as tabulated summaries of the A6623-220 i data. The trends and deviations analyses and tabulated sum- See NUlkEG-1272,V11,N01 abstract. ) maries have been presented and calculated accounting for the plants operabonal conditions. NUREG-1272 Vit N03: OFFICE FOR ANALYSIS AND EVALUA-TION OF OPERATIONAL DATA.1997 Annual Report (Technical NUREG 1272 V10 N01: OFFICE FOR ANALYSIS AND EVALUA- . Training).

• Office for Analysis & Evaluation of Operational Data, TlON OF OPERATIONAL DATA.1996 Annual Report.
• Office Director. November 1998. 40pp. 9901270192. A6623:325.

for Analysis & Evaluation of Operational Data Director. Decem- See NUREG-1272,V11,N01 abstract. ber 1997. 265pp. 9804080062. A2920:001. This annual report of the U.S. Nuclear Regulatory Commis- NUREG 1307 R08: REPORT ON WASTE BURIAL sion's Office for Analysis and Evaluation of Operational Data CHARGES. Changes in Decommissioning Waste Disposal Costs (AEOD) describes activities conducted during 1996. The report At Low-Level Waste Burial Facihties.

• Division of Regulatory is published in three parts. NUREG-1272, Vol.10, No.1, covers Applications (Post 941217). December 1998. 83pp.

power reactors and presents an overview of the operating expe. 9901270213. A6622:271. rience of the nuclear power industry from the NRC perspective. A requirement placed upon nuclear power reactor licensees including comments about trends of some performance meas- by the U.S. Nuclear Regulatory Commission (NRC) is that li-ures. The report also includes the principal findings and issues consees must annually adjust the estimate of the cost of decom-identified in AEOD studies over the past year and summarizes missioning their plants. in dollars of the current year, as part of information from such sources as licensee event reports and re- the procoss to provide reasonable assurance that adequate ports to the NRC's Operations Center. NUREG-1272, Vol.10, funds for decommissioning will be available when needed. This No. 2, covers nuclear materials and presents a review of the report, which is revised periodically, explains the formula that is events and concems during 1996 associated with the use of li- acceptab6e to the NRC for determining the minimum decommis-censed rnaterial in nonreactor applications, such as personnel sioning fund requirements for nuclear power riants. The sources overexposures and medical misadministrations. Both reports of infortnation used in the formula are identified, and the values also contain a discussion of the incident investigation Team pro- developed for the estimation of radioactive waste burial /disposi-gram and summarize both the incident Investigation Team and tion costs, by site and by year, are given. Licensees may use Augmented inspection Team reports. Each volume contains a the formula, coefficients, and burial / disposition adjustment fac-list of the AEOD reports issued from CY 1980 through 1996. tors from this report in their cost analyses, or they rnay use ad-NUREG.1272 Vol 10, No. 3, covers technical training and pre- justment factors at least equal to the approach presented here-sents the activities of the Technical Training Center in support in. of the NRC's mission in 1996. NUREG-1363 V07: ATOMIC SAFETY AND LICENSING BOARD NUREG 1272 V10 N02: OFFICE FOR ANALYSIS AND EVALUA- BIENNIAL REPORT. Fiscal Years 1995 - 1996.

• Atomic Safety TION OF OPERATIONAL DATA.1P96 Annual Report.
• Office & Licensing Board Panel. June 1998. 51pp. 9806290363.

for Analysis & Evaluation of Operational Data, Director. Decem- A3958 240. ber 1997.136pp. 9805050445. A3319:255. The Panet handled 33 cases in fiscal Year 1995 and 29 cases See NUREG-1272,V10,N01 abstract. in fiscal Year 1996. This report summarizes, highlights, and ana-lyzes how the wide-ranging issues raised in these cases were NUREG 1272 V10 NO3: OFFICE FOR ANALYSIS AND EVALUA- addressed by the Paners kcensing boards and presiding officers TION OF OPERATIONAL DATA.1996 Annual Report.

• Office during this period. This report also describes the Paners other for Analysis & Evaluation of Operational Data, Directo . Decem.

ber 1997. 42pp. 9805060124. A3321:069. responsibilities, addresses the status of Panet activities, and re-See NUREG-1272,V10.N01 abstract. ports on present and projected future caseloads. NUREG-1272 V11 N01: OFFICE FOR ANALYSIS AND EVALUA- NUREG-1415 V10 NO2: OFFICE OF THE INSPECTOR GEN. TlON OF OPERATIONAL DATA.1997 Annual Report (Reac- ERAL. Semiannual Report To Congress, October 1,1997 March tors).

• Office for Analysis & Evaluation of Operational Data, Di- 31,1998.
• Office of the inspector General (Post 890417). June rector. November 1998. 325pp. 9901270152. A6620:128. 1998. 36pp. 9807060272. A4009:318.

This annual report of the U.S. Nuclear Regulatory Commis- The inspector General Act of 1978, as arnended, requires that sion's Othee for Analysis and Evaluation of Operational Data inspectors General submit a " Semiannual Report to Congress" (AEOD) describes activities conducted during 1997. The report summarizing program actisit;es. The Inspector General's report is published in three parts. NUREG-1272, Vol. II, No.1, covers is submitted to the Chairman of the NRC not later than April 30

F I l Main Citations and Abstracts 7 l and October 31 for each reporting period. The Chairman com- test. These tests are performed in conjunction with an Elevated ments on the report and prepares the NRC's Semiannual Report Measurement Comparison to provide confidence that the radio-t to Congress as required by the Act. The Chairman then submits logical criteria specified for heense termination are met. The i the agency's report and the OlG's report to Congress no later Data Quality Objectives process is used for the planning of final l than November 30 and May 31, respectively, site surveys. This includes methods for determining the number NUREG 1415 Vit N01: OFFICE OF THE INSPECTOR GEN of samples needed to obtain statistically valid comparisons with ERALSemiannual Report To Congress April-September 1998. .- decommissioning criteria and the methods for conducting the Office of the inspector General (Post 890417). November 1998. statistical tests with the resulting sample data. 38pp. 9812110109. A6147:080. NUREG 1507: MINIMUM DETECTABLE CONCENTRATIONS See NUREG-1415,V10,N02 abstract. WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR NUREG-1423 V08: A COMPILATION OF REPORTG OF THE AD- VARIOUS CONTAMINANTS AND FIELD CONDITIONS. VISORY COMMITTEE ON NUCLEAR WASTE. July 1997 June ABELOUIST.E.W. Oak Ridge Associated Universities. 1998.

• Advisory Committee on Nuclear Waste. August 1998. BROWN.W.S. Brookhaven National Laboratory. POWERS.G.E.;

68pp. 9809230322. A5171294. et al. Division of Regulatory Applications (Post 941217). June This compilation contains 11 reporte, issued by the Advisory 1998.194pp. 9806190288. A3903:1$7. Committee on Nuclear Waste (ACNW) during the tenth year of This document describes and quantitatisely evaluates the ef-its operation. The reports were submitted to the Chairman and fects of various factors on the detection sensitivity of commer-Commissioners of the U.S. Nuclear Regulatory Commissiort. All cially available portable field instruments being used to conduct reports prepared by the Committee have been made available raddogical surveys in support of decommissioning. The U.S. to the public through the NRC Public Document Room, the U.S. Nuclear RegulEtory Commission (NRC) has amended its regula-Littary of Congress, and the intemet at http /www.nrc. l gov / tions to establish residual radioactivity enteria for decommis-ACRSACNW. sioning of hcensed nuclear facilities. to support of that rulo-ma no. m ssbn has pmpamd a Gmet EMrmmmtaf NUREG-1426 V03: COMPILATION OF REPORTS FROM RE- Impact Statement (GEIS), consistent with the National Envirore. SEARCH SUPPORTED BY THE ELECTRICAL. MATERIALS mental Policy Act (NEPA). The effects of this new rulemaking AND MECHANICAL ENGINEERING BRANCH, DIVISION OF on the overall cost of decommissloning are among the many ENGINEERING. SANTOS,C.G. Division of Engineering Tech- factors considered in the GEIS. The overall cost includes the hology (Post 941217). October 1998.115pp. 9811020047. c sts of decontamination, waste disposal, and radiological sur. A5648:23f'65, the Materials Engineering Branch, Division veys of En-to demonstrate compliance with the applicable guidelines. Since 19 An important factor affectingthe costs of such radiological gineering, of the Nuclear Regulatery Commission's Office of Nu-clear Hegulatory Research, and its predecessors dating back to sens e um e mnce@ah W) of hed survey instruments in relation to the residual rad.oactivity cri-the Atomic Energy Commission (AEC), have sponsored re- teria. The purpose of this study was two-fold. First, the cata search programs conceming the integrity of the primary system were used to determine the vahdity of the theoretcal minimum pressure boundary of light-water reactors. The components of a mnewawns Es) used h me GEIS. Semnd, concem in these research programs have included the reactor ms of e sW, @sW hemin, pmh gudana to pressure vessel (RPV), steam generators, and the piping. These Icensees for (a) selection and proper use of portable survey in-research programs have covered a broad range of topics, in- struments and (b) understanding the fsid conditions and the ex. ciuding fracture mechanics analysis and experimental work for tent to wM N WMes of mose hstrants can be hn> RPV and piping applications, inspection method development ited. The types of instruments commonly used in field radio-and qualification, and evaluation of irradiation effects on RPV 4 cal says wwe evaluaW inW. h part. gas p steels. This report provides as complete a listing as practical of po a, eiger- er , su e n ), and Mum formal technical reports submitted to the NRC by the investiga- * * ** tors working on these research programs. This listing includes topical, final, and progress reports and is divided by topic area. NUREG-1521 DRFT FC: TECHNICAL REVIEW OF RISK-IN-In many cases, a report will cover several topics (such as in the FORMED, PERFORMANCE-BASED METHODS FOR NU-case of progress reports of multi-faceted programs) but is listed CLEAR POWER PLANT FIRE PROTECTION ANALYSES. Draft under only one topic. Therefore, in searching for reports on a Report For Comment. DEY,M.K. Division of Systems Tech-specific topic, other related topic areas should be checked also, nology (Post 941217). AZARM.M.A.; TRAVIS,R.; et al. The separate volumes of this report cover the following periods: Brookhaven National Laboratory. July 1998. 200pp. Volume 1: 1965 1990, Volume 2: 1991 - 1993, and Volume 9809250050. A5200:181. 3: 1994 - 1998. The Nuclear Regulatory Commission (NRC) has instituted an initiative for regulatory improvement to focus licensee and NRC NUREG-1505 R01: A NONPARAMETRIC STATISTICAL METH- resources on WAsign:ficant activities, and decrease the ODOLOGY FOR THE DESIGN AND ANALYSIS OF FINAL prescriptiveness of its regulations. The NRC has identified risk-STATUS DECOMMISSIONING SURVEYS. Interim Report For int rmed methods utilizing insights from probabiliste nsk anal-Use And Comment. GOGOLAK,C.V. Energy, Dept. of. POW- ysis (PRA) as a major tool for achieving its goal for regulatory ERS.G.E.; HUFFERT,A.M. Division of Regulatory Applications focus. Fire protection requirements has been identified as a reg-(Post 941217). June 1998. 200pp. 9807270442. A4384:001. ulatory area in which NRC will pursue regulatory improvement. This report describes a nonparametric statisteal methodology This report presents a technical review and analysis of alter-fur the design and analysis of final status decommissioning sur- native risk-informed, performance-based methods to those in veys in support of the final rulemaking on Radiological Criteria current prescriptive fire protection requirements or guidance that for License Termination published by the Nuclear Regulatory could allow cost-effective methods for implementation of safety Commission in the Federal Register on July 21,1997. The tech. objectives, focus licensee efforts, and achieve greater efficiency l niques described are expected to be applicable to a broad in the use of rescurces for plant safety. A technical analysis of i range of circumstances, but do not preclude the use of alter- the usefulness of the results and insights derived from these native methods as particular situations may warrant. Nonpara- methods (including accounting for the uncertainties in the re-metric statistical methods for testing compliance with decommis- suits) in improving regulatory decision making is presented. sioning criteria are provided both for the case in which the radio-nuclides of concem occur in background and also for the case NUREG 1542 V03: ACCOUNTABILITY REPORT FISCAL YEAR in which they do not occur in background. The tests described 1997. CONNELLY,S.R. Office of the Controller (Post 890205). cre the Sign test, the Wilcoxon Rank Sum test, and a Quantile March 1998. 02pp. 9804200258. A3031:011. l l 1 l

8 Main Citation] and Abstracts The U.S. Nuclear Reguiivy Cornmission (NRC) is one of NUREG 1552 $01 DR FC: FIRE BARRIER PENETRATION several Federal agencies pacipating in a pilot project to SEALS IN NUCLEAR POWER PLANTS. Draft Report For Com-streamline financial management eporting. The goal of this pilot ment. BAJWA,C.S.; WEST,K.S. June 1998. 67pp. is to consolidate performance-rellled reporting into a single so- 9809250036. A5109:291, countabihty report in andance with the Govemment Manage. Nuclear power plants use the " defense in depth" concept of ment Reform Act (GMRA) e 1994. The NRC's third account- echelons of fire protection to achieve a high degree of fire safe-ability report consolidates t!1 information previously reported in ty. The objective of this concept is to (1) prevent fires from start-the NRC's annual financial st% ment required by the Chief FF ing; (2) rapidy detect, control, and extinguish those fires that do nancial Offcers Act of 1990, as amended; the chairman's an- occur; and (3) protect structures, systems, and components im-r ual report to the President and the Congress, required by the portant to safety so that a fire that is not promptly extinguished Federal Managers Financial integrity Act of 19C2; and the Chair. will not prevent the safe shutdown of the plant. Fire barrier pen-man's semiannual report to the Congress on management deci- etration seats, which are but one element of the fire protection ( sions and final actions on Office of Inspector General (OlG) defense-irMepth concept, are designed to confine a fire to the ' audit recommendations, required by the inspector General Act area in which it started or to protect plant systems and compo-of 1978, as amended. This report also includes performance nents within an area from a fire outside the area. On the basis measures, as required by the Chief Financial Offcers Act, the of everything it found ard considered, it is the staffs judgement ] Govemment Performance Results Act of 1993, and the Chair. that, overall, the issue of the potential fire barrier penetration man's statement on the compliance of the agency's financial seal defciencies does not affect safety. For the reasons given - management systems with the Federal Financial Management in this paper, typical penetration seal deficiencies do not nec-Improvement Act of 1996. essarily equate to a lack of adequate protection or result in  : NUREG-1549 DRFT FC: DECISION METHODS FOR DOSE AS- tha e nsee o exis'ti penetra s al l SESSMENT TO COMPLY WITH RADIOLOGICAL CRITERIA spections are adequate (1) to ensure that penetration seal prob-FOR UCENSE TERMINATION. Draft Report For Comment.

• Di- tems are discovered and resolved and (2) to maintain public vision of Regulatory Applications (Post 941217). July 1998. health and safety. ]'

96pp. 9902090154. A6757:305. This draft NUREG descrbes a methodology for calculating NUREG-1556 V02: CONSOLIDATED GUIDANCE ABOUT MATE-doses to demonstrate compliance with the radiological criteria RIALS LICENSES. Program-Specife Guidance About industrial for decommissioning and Icense termination. The methodology Radsography Licenses. Final Report. CARRICO J.B.; COL-is designed to allow each heensee the fledbility to optimize their LINS.D.; WHITE,D.; et al. Division of Industrial & Medical Nu-decommissioning activities within the context of the License Ter. clear Safety (Post 870729). August 1998. 200pp. 9809100204. mination rule. It is important to note that, although this document A5008:129. is dvided into multiple sections to simplify the dscussion for dif- This document is intended for use by applicants, licensees, forent situations, the undertying modeling process is a smooth and NRC staff and will also be available to Agreement States. continuum of options. The simplest method for calculating dose, This guidance corresponds with the revision to 10 CFR Part 34 generic screening (see Chapter 3), uses models and default pa- published in May 1997. This document combines and super-rameters developed by the NRC for the purpose of screening sedes the guidance previously found in draft Regulatory Guide [ Kennedy and Strenge,1992). The generic models and default FC 401-4," Guide for the Preparation of Applications for the Use parameters are intended to estimate the upper range of the of Sealed Sources and Devices for Performing Industrial Radi-dose that an individual could receive, and are expected to over, ography," and in NMSS Pohey and Guidance Directive FC 64-estimate the dose for rnost sites. The purpose of generic 15. " Standard Review Plan for Applications for the Use of screerung is to allow the licensee o aimple and cost-effective Sealed Sources and Devices for Performing Industrial Radiog-method to demonstrate compliance using a minimum amount of raphy." This report, where applicable, provides a more risk-in-site-specific information. Such a screening approach is based on formed, performarce-based approach to industrial radiography reasonably conservative assumptions, since it must provide a hcensing consistet wnh the current regulations. On September reasonable level of assurance and must be appleable to a wide 17,1997, (62 FR 48904), NRC announced the aveitability of range of licensees, radionuclides, and processes. As such, it is draft NUREG 1556, VOL 2, and requested comments on it. The expected to be appropriate for NRC licensees who have rel- final document contains a compilation of the comments and the atively simple decommissioning situations. The calculated value staffs responses. The comments were considered in preparing under genere screening conditions is simply a marker used to the final NUREG Report. The draft and final NUREG Reports demonstrate compliance, and is not intended to be a realistic may differ, if your Icense was issued or amended based on rec-dose estimate. Generic screening may not be appropriate for h- ommendations in the draft NUREG Report and you feel that the consees who have complex mixtures of radionuclides, unusual final guidance is more advantageous to you, you may choose or unique decommissioning situations, or where the use of very to request an aedet. conservative assumptions would result in unwarranted costs or NUREG-1556 V03: CONSOLIDATED GUIDANCE ABOUT MATE-inefficient and !! logical remediation requirements. Licensees who RIALS 1.lCENSES. Applications For Sealed Source And Device prefer to use site-specife information can use the same models Evaluation Ard Registration. Final Report. LUBINSKI,J.; as are used for generic screening, but substitutes site-specific BAGGETT,S.; BROADDUS,0.; et al. Divis!on of trdustrial & values in place of some or all of the genede default parameters Medical Nuclear Safety (Post 670729). July 1998.158pp. (see Chapters 4 and 5). The resulting dose estimates are ex- 9808210282. A4711:185. pected to be more realistic and provide less of an over-estimate NUREG 1556, Vol. 3. "Consoldated Guidance about Mate-of dose than that provided by the generic screening approach. rials Licenses: Applications for Sealed Source and Device Eval-Site-specific screening utilizes additional site specific data to uation and Registration," dated July 1998, is designed to pro-support the mod:fication or elimination of a particular scenario or vide applicants for requests for a sealed source or device safety pathway, or to demonstrate that a parameter or group of pararn evaluations, and reviewers of such requests, with the informa-eters can be better represented by site specific values. Alter- tion and materials necessary to make determinations that the native exposure scenarios may be appropriate based on site- products are acceptable for hcensing purposes. It provides the specife factors that affect the likelihood and extent of potential applicants and reviewtsrs with information conceming how to file future exposure to residual radioactivity. Guidarre has been in- a request, a hsting of the applicable regulations and industry cluded in this document regarding sources of information avall- standards, policies affecting evaivation and registration, certain able to licensees that can be used to support modifcation of pa- administrative procedures to be followed, information on how to rameter values. perform the evaluation and write a registration certifcate, and

F l l Maln Citations and Abstracts 9

the responsibilities of the registration certifcate holder. This doo- Revision (Rev.) 1, " Guide for the Preparation of Applications for ument combines the guidance previously found in NUREG- Licenses for the Use of Self-Contained Dry Source-Storage 1550,

• Standard Review Plan for Appucations for Sealed Source Gamma irradiators," dated December 1988, and in NMSS Pol-and Device Evaluations and Registrations," Regulatory Guide icy and Guidance Directive (P&GD) FC 84-16, Rev.1, " Stand-10.10, " Guide for the Preparation of Applications for Radiation ard Review Plan for Appleations for Use of Self-Contained Dry Safety Evaluation and Registration of Devices Containing By- Source-Storage Garnma irradiators," dated January 26, 1989.

product Material," Regulatory Giuide 10.11. " Guide for the Prep- This final report takes a more risk-informed, performance-based aration of Applications for Radiation Safety Evaluation and Reg- approach to licensing self-shielded irradiators, and reduces the istration of Sealed Sources Containing Dyproduct Material," and information (amount and level of detail) needed to support an the Offee of Nuclear Material Safety and Safeguards Policy and application to use these devices, it incorporates suggestions re-Guidance Directives 84-22,"What Source and Device Designs ceived during the comment period on draft NUREG-1556, Vol. Require an Evaluation," and 84-5, " Source and Device Evalua- 5. When published, this final report should be used in preparing tion Technical Assistance Request" This report incorporates self-shielded irradiator license appleations. NRC staff will use  ! suggestions submitted during the comment period on draft this final report in reviewing these applications. I NUREG-1556, Vol. 3. When published, this final report should be used in preparing sealed source and device applications. NUREG-1556 V07 DR FC: CONSOUDATED GUIDANCE ABOUT NRC staff will use this final report in reviewing these applica- MATERIALS UCENSES. PROGRAM-SPECIFIC GUIDANCE hons. ABOUT ACADEMIC,RESEARCH AND DEVELOPMENT AND OTHER LICENSES OF LIMITED SCOPE. Draft Report For Com-NUREG-1556 V04: CONSOUDATED GUIDANCE ABOUT MATE- ment. FULLER M.; HAYS,R.; LODHl,A.S.; et al. Division of in- l RIALS UCENSES. Program-Specife Guidance About Fixed dustrial & Medical Nuclear Safety (Post 870729). May 1998. l Gauge Licenses. Final Report. HENDERSON,P.J.; KIRK- 166pp. 9807270190. A6805:001. WOOD,A.S.; LEWIS.S.H. Division of Industrial & Medical Nu- As part of its redesign of the materials licensing process, NRC l clear Safety (Post 870729). October 1998. 300pp. 9811120288. is consolidating and updating numerous guidance documents j A5825:001. Into a single comprehensive repository as described in NUREG-As part of its redesign of the materials licensing process, NRC 1539, " Methodology and Findings of the NRC's Materials LL b consolidating and updating numerous guidance documents censing Process Redesign," dated April 1996, and draft into a single comprehensive repository as described in NUREG- NUREG-1541," Process and Design for Consolidating and Up-1539 and draft NUREG-1541, NUREG-1556, Vol. 4, "Consoli- dating Materials Ucensing Guidance," dated April 1996. Draft dated Guldance about Materials Licenses: Program-Specife NUREG-1556, Vol. 7. " Consolidated Guidance about Materials Guidance about Fixed Gauge Ucenses," dated October 1998, is Ucenses: Program-Specife Guidance about Academic, Re-the fourth prograrmspecific guidance developed for the new search & Development, and Other Licenses of Umited Scope," process and is intended for use by applicants, licensees, and dated May 1998, is the seventh program-specific guidance de-NRC staff, and will also be available to Agreement States. This veloped for the new process and is intended for use by appli-document combines and updates the guidance found in Draft cants, licensees, and NRC staff and will also be available to Regulatory Guide and Value/ impact Statement, FC 404-4, Agreement States. This document combines and updates the "Gude for the Preparation of Applications for Licenses for the Guidance for applicants and licensees previously found in; (1) Use of Sea led Sources and Nonportable Gauging Devices," Regulatory Guide 10.2, Revision 1. " Guidance To Academic in-dated January 1985, and in NMSS Policy and Guidance Direc- stitutions Applying For Spectre Byproduct Material Ucenses of tive, FC 85-4, " Standard Review Plan for Applications for Use Umited Scope," dated Decerrt>er 1976, (2) Regulatory Guide of Sealed Sources and Nonportable Gauging Devices," dated 10.7. " Guide For The Preparation Of Applications For Ucenses February 6,1985. This report takes a more risk-informed per- For Laboratory and Industrial Use of Small Quantities of Byprod-fomiance-based approach to Icensing fixed gauges, and re- uct Material," dated August 1979, and (3) Draft Regulatory duces the information (amount and level of detail) needed to Guide FC 405-4, " Guide for the Preparation of Applications for support an application to use these devices. On December 23, Ucenses for the Use of Sealed Sources in Gas Chromatography 1997 (62 FR 67100), NRC announced the availability of draft Devices and X-Ray Fluorescence Analyzers," dated February j NUREG-1556 Vol. 4, and requested comments on it. The final 1985. This draft report takes a more risk-informed, performance-documerd contains a compilation of the comrnents and the ( based approach to the information needed to support an appli-staffs responses. The comments were considered in preparing cation for the use of byproduct material. Note that this document the final NUREG Report. This document is available for use by is strictly for public comment and is not for use in preparing or NRC licensees, applicants, and reviewers, and is also available reviewing academic, research & development, and other li-to Agreement States. censes of limited scope (ARDL) until it is published in final form. NUREG-1556 V05: CONSOUDATED GUIDANCE ABOUT MATE- NUREG-1556 V08: CONSOLIDATED GUIDANCE ABOUT MATE-RIALS LICENSES. Program-Specife Guidance About Self- RIALS UCENSEES. Program-Specific Guidance Exempt Dis-Shielded irradiator Licenses. Final Report VACCA,P.C.; COL- tribution Ucenses. Final Report. GREENE,S.: CAMPER,L; UNS.D.J.; MITCHELL.M.W.; et al. Division of industrial & Med- RICH,T, Division of Industrial & Medical Nuclear Safety (Post leal Nuclear Safety (Post 870729). October 1998. 214pp. 870729). September 1998. 200pp. 9810080056. A5352:001. 9902090210. A6756:313. NUREG-1556, Vol 8, "Consohdated Guidance about Mate-As part of its redesign of the materials licensing process, the rials Licenses: Program-Specife Guidance about Exempt Dis-Nuclear Regulatory Commission (NRC) is consolidating and uo- tribution Licenses," dated July,1998, is the eighth program-spe-cating numerous guidance documents into a single comprehen- cific guidance developed under a new process and is intended siva repository as described in NUREG-1539, " Methodology and for use by applicants, licensees, and NRC staff, and will also be Findings of the NRC's Materials Ucensing Process Redesign," available to Agreement States. On April 7,1997 (62 FR 16630), dated April 1996, and draft NUREG-1541, " Process and Design NRC announced the availabihty of draft NUREG-1562, " Stand-for Consolidating and Updating Materials Licensing Guidance," ard Review Plan for Applications for Licenses to Distribute By-dated April 1996. NUREG-1556, Vol. 5,

• Consolidated Guidance product Material to Persons Exempt from the Requirements for about Materials Licenses: Program-Specife Guidance about an NRC License," dated January 1997, and requested corre

{ Self-Shielded irradiator Licenses," dated October 1998, is the ments on it. The final version of NUREG-1562 will be published l fifth program-epocific guidance developed for the new process as NUREG-1556, Vol. 8, " Consolidated Guidance about Mate- } and is intended for use by applicants, licensees, and NRC staff rials Ucenses: Program-Specirc Guidance about Exempt Dis- ) and will also be available to Agreement States. This document tribution Licenses," dated August 1998, in finalizing the NUREG [ supersedes the guidance found in Regulatory Guide (RG) 10.9, report, the NRC staff considered all of the comments, including

10 Main Citations and Abstracts constructive suggestions to improve the document. This docu- censes," dated October 1998, is the tenth program-specifc ment combines, updates and supersedes the guidance found in guidance developed for the new process and is intended for use Draft NUREG-1562, " Standard Review Plan for Appications for by Federal applicants and licensees, and NRC staff. This docu-Licenses to Distribute Byproduct Material to Persons Exempt ment combines and updates the guidance for appleants and h-from the Requirements for an NRC License," and incorporates censees previously found in Policy and Guidance Directive PG suggestions and comments received during the comment pe- 6-02, Revision 1:" Standard Review Plan (SRP) for License Ap-riod. When published, this final report should be used in applica- plcation for Master Material License," dated September 25, tions for exempt distribution. NRC staff will use this final report 1997. Note that this document is stnctly for public comment and in reviewing these appleations. The guidance contained within is not for use in preparing or reviewing applcations for Master the document does not represent new or proposed regulatory Materials licenses until it is published in final form. requirements, and licensees wil! not be inspected against any portion of it. Additionally, regulatory compliance with all applica- NUREG 1556 V11 DR FC: CONSOLIDATED GUIDANCE ABOUT ble regulations is not assured by Icensees who adopt any por. MATERIALS LICENSES.Progran> Specific Guidance About Spe-tion of, or apply the principles described in, this guidance, cific Licenses Of Broadscope. Draft Report For Comment. DWYER.J.P.; BAILEY,0.M.; MULLAUER,J.R.; et al. Division of NUREG-1556 V09 DR FC: CONSOLIDATED GUIDANCE ABOUT Industrial & Medical Nuclear Safety (Post 870729). August 1998. MATERIALS LICENSES.Progran> Specific Guidance About Medical Use Licenses. Draft Report For Comment. 00pp. M020902% A6765M,' Draft NUREG-1556, Vol.11, Consolidated Guidance about LANZISERA,P.A.; JONES,A.R.; GATTONE,R.G.: et af. Division Matenals Licenses: Program-Specific Guidance about Licenses of industrial & Medical Nuclear Safety (Post 870729). August of Broad Scope," dated August 1998, is the eleventh program-1998. 200pp. 9808250203. A4755:001. specific guidance document developed for the new process and This draft guide has been developed in parallel with the pro- is intended for use by applicants, licensees, and NRC staff and posed revision of 10 CFR Part 35, " Medical Use of Byproduct will also be available to Agreement States. This document com-Material." Comments received m response to pubication of this bines, updates and supersedes the guidance for appicants and dratt will be considered in developing the final guide. Finalization licensees previously found in Draft Regulatory Guide DG-0005 of the guidance will continue to parallel the rulemaking; resulting dated October 1994. Included in this guidance is a new option in a guidance document that is consistent with the final rule. Draft NUREG-1556, Vol. 9. " Consolidated Guidance about Ma_ for Type A licensees of broad scope to have increased flexibility to make changes in some program areas and revise some pro-tenals Licenses: Program Specific Guidance About Medical Use cedures previously approved by the NRC without amendment of Ucenses," dated July 1998, is the ninth program-specific guid_ the hcense. This option is discussed in detail in Chapter 1 of this ance document developed for this guidance senes, and is in- document. Draft NUREG-1556, Volume 11, is not intended to be tended for use by apphcants, licensees, and NRC staff and will used alone. Because broad scope Icensees may be involved in also be available for use by Agreement States. This document combines and supersedes the guidance previously found in many different program areas (e.g., medicine, research and de-velopment, manufacturing and distribution, etc.), this document Regulatory Guide (RG) 10.8, Revision 2, " Guide for the Prepa-frequently refers the user to other more program-specife guid-ration of Appications for Medical Use Programs"; Appendix X to ance documents in the NUREG-1556 series. A single document RG 10.8, Revision 2, " Guidance on Complying With New Part containing all of the guidance that might be required by a broad 20 Requirements"; Draft RG DG-0009, " Supplement to Regu_ scope licensee or an applicant for a broad scope license would latory Guide 10.8, Revision 2. " Guide for the Preparation of Ap-be unw;eldy and would quickly become obsolete as guidance in pications for Medical Use Programs"; Draft RG FC 414-4, the individual proDram areas is revised. This document takes a

     " Guide for the Preparation of Apphcations for Licenses for Med.

m re risk-informed, perfotmanc& based approach to the infor-ical Teletherapy Programs"; Policy and Guidance Directive mation neeoed to support an appleation for a license of broad (P&GD) FC 87-2, " Standard Review Plan for License Applica-tions for the Medcal Use of Byproduct Matenal"; P&GD FC 86- scope. Note that this document is strictly for puble comment and is not for use in preparing or reviewing Icenses of broad 4, Revision 1, "information Required for Licensing Remote scope until it is published in final form. Afterloading Devices"; Addendum to Revision 1 to P&GD FC 86-4, "Information Required for Licensing Remote Afterloading NUREG-1560 V01 P1: INDIVIDUAL PLANT EXAMINATION PRO-Devices- Increased Source Possession Umits"; P&GD 3-15, GRAM: PERSPECTIVES ON REACTOR SAFETY AND PLANT

• Standard Review Plan for Review of Quahty Management PERFORMANCE. Summary Report.
• Division of Systems Tech-Programs"; RG 8.39, " Release of Patients Administered Radio- nology (Post 941217). Decernber 1997, 257pp. 9802200064.

active Materials"; RG 8.33, " Quality Management Program"; A2254 001. P&GD 317," Review of Training and Experience Documentation This report provides perspectives gained by reviewing 75 Indi-Submitted by Proposed Physcian User Appicants"; and RG vidual Plant Examination (tPE) submittals periaining to 108 nu-8.23. " Radiation Safety Surveys at Medical Institutions, Revision clear power plant units. IPEs are probabihste analyses that esti-1 ". Note that this document is strictly ior public comment and mate the core damage frequency (CDF) and containment per. is NOT for use in preparation or review of medical use licenses formance for accidents initiated by intemal events (incheng in-until it is published in final form. temal floods, but excluding intemal fire). The U.S. Nuclear Reg-NUREG 1556 V10 DR FC: CONSOLIDATED GUIDANCE ABOUT ulatory Commission (NRC), Offee of Nuclear Regulatory ne-MATERIALS LICENSES. Program Spoofic Guidance About Mas- search, reviewed the IPE submittals with the objective of gaining ter Matenals Licenses. Draft Report For Comment. JONES,J.D.; perspectives in three major areas: (1) improvements made to in-HENSON,J.; HOWE D.B.; et al. Division of Industrial & Medical dividual plants as a resutt of their IPEs and the collective results Nuclear Safety (Post B70729). October 1998. 50pp. of the IPE program, (2) plant-specific design and operational 9902090277. A6765:263. features and modeling assumptions that signifcantty affect the As part of its redesign of the materials Icensing process, NRC estimates of CDF and containment performance, and (3) is consohdating and updating numerous guidance documents strengths and weaknesses of the models and methods used in into a single comprehensive repository as described in NUREG- the IPEs. These perspectives are gained by assessing the core 1539, " Methodology and Findings of the NRC's Materials L6- damage and containment performance results, including overall censing Process Redesign," dated April 1996, and draft CDF, accident sequences, dominant contributions to the design NUREG-1541, " Process and Design for Consolidating and Up- and operational characteristes of the various reactor and con-dating Materials Licensing Guidance," dated Aprl! 1996. Draft tainment types, and by comparing the IPEs to probabilistic risk NUREG-1556, Vol.10, " Consolidated Guidance about Materials assessment characteristics. Methods, data, boundary conditions, Licenses: Program-Specific Guidance about Master Matenals Li- and assumptions used in the IPEs are considered in under-i

F Main Citations cnd Abstracts 11 standing the difference and similarities observed among the var- the Commission's Enforcement Policy. However, this is a policy lous types of plants, statement and not a regulaton. The Commisson may deviate { NUREG-1560 V02 P2-5: INDIVIDUAL PLANT EXAMINATION from this statement of policy and procedure as appropriate PROGRAM: PERSPECTIVES ON REACTOR SAFETY AND under the circumstances of a particular case. PLANT PERFORMANCE.

• Division of Systems Technology NUREG 1608: CATEGORIZING AND TRANSPORTING LOW (Post 941217). December 1997. 546pp. 9802200072. SPECIFIC ACTIVITY MATERIALS AND SURFACE CONTAMI-A2246:163. NATED OBJECTS. COOK,J.; LEWIS.R.; EASTON.E.; et al. Of-See NUREG-1560,V01,P1 abstract. fice of Nuclear Material Safety & Safeguards. July 1998. 62pp.

NUREG 1560 V03 P6: INDIVIDUAL PLANT EXAMINATION PRO-80806006E A4516 . , The primary purpose of this guidance is to assist shippers in GRAM: PERSPECTIVES ON REACTOR SAFETY AND PLANT anng I w sp a mateftals 4SA) and sMace com PERFORMANCE. Appendices.

• Division of Systems Technology (Post 941217). December 1997. 46pp. 9802200077. A2249:302. taminated objects (SCOs) for shipment in compliance with Fed-See NUREG-1560,V01,P1 abstract. eral reg abes. Mance W prom in qidon and answer i format on the categorization, packaging a1d Tansportation of j NUREG 1570: RISK ASSESSMENT OF SEVERE ACCIDENT-IN. LSA and SCOs, including the definition of t.SA and SCOs the i DUCED STEAM GENERATOR TUBE RUPTURE.
• Office of determination of distributon of activity i9 LSA matenal or on Nuclear Reactor Regulation (Post 941001). March 1998. 218pp. SCO surfaces, mixing LSA and SCCM in a package, radiation 9803310390. A2839:001. level measurement, and various sther aspects of transporting This report describes the basis, results, and related nsk impli- LSA and SCOs. There are mray requirerrents, other than those cations of an analysis performed by an ad hoc working group addressed herein, impnN in the shipment of LSA and SCOs.

to assess the containment bypass potential attributable to steam The guidance ra%ents ore or more metheds of demonstrating i generator tube rupture (SGTR) induced by severe accident con. compliance with the regulatory requirerrents for LSA material I ditions. The SGTR Severe Accident Working Group, comprised and SCOs that have been found acceptable to NRC and DOT; of staff members from the NRC's Offices of Nuclear Reactor however, additional methods may also be found to be accept-Regulat on (NRR) and Nuclear Regulatory Research (RES), un. able with adequate justification. dertook the analysis beginning in December 1995 to support a NUREG-1619: STANDARD REVIEW PLAN FOR PHYSICAL proposed steam generator integrity rule. The work drew upon PROTECTION PLANS FOR THE INDEPENDENT STORAGE previous risk and thermal-hydraulic analyses of core damage OF SPENT FUEL AND HIGH-LEVEL RADIOACTIVE WASTE. sequences, with a focus on the Surry plant as a representative DWYER,P.A. Division of Fuel Cycle Safety & Safeguards (Post example. This analysis yielded new results, however, derived by 930207). July 1998. 33pp. 9808120167. A4604:070. predicting thermal-hydraulc conditions of selected severe accb This document is a standard review Plan (SRP) for evaluating dent scenarios using the SCDAP/RELAPS computer code, plans for the physical protection of spent fuel and high-level ra-flawed tube failure modeling, and tube failure probability esth dioactive waste stored at (1) independent spent fuel storage in-mates. These results, in terms of containment bypass prob- stallations, (2) monitored retrievable storage installations, and ability, form the basis for the findings presented in this report. (3) the geologic repository operations area. Conducting a review NUREG-1575: MULTI-AGENCY RADIATION SURVEY AND SITE according to an SRP ensures that license apphcants address INVESTIGATION MANUAL (MARSSIM). Final Report.

• NRC . every pertinent Nuclear Regulatory Commission (NRC) require-No Detailed Affinaten Given.
• Defense, Dept. of. *: et at En- ment in their NRC-approved physical protection plan and com-ergy, Dept. of. December 1997. 665pp. 980220C046. EPA. prehensiveness in the NRC review of the plans. The information 402R-97-016. A2248:001, here takes a new matrix or " modular
• format to streamline the The MARSSIM provides information on planning, conducting, information and facihtates its use.

cvaluating, and documenting building and surface soll final sta- NUREG-1622: NRC ENFORCEMENT POLICY REVIEW. July 1995 tus radiological surveys for demonstrating compliance with dose . July 1997, LIEBERMAN.J.; PEDERSEN.R.M. Ofc of Enforce-or risk-based regulations or standards. The MARSSIM is a ment (Post 870413). April 1998. 66pp. 9805060112. A3321:001. multi-agency consensus document that was developed collabo- On June 30,199b, the Nuclear Regulatory Commission  ! ratively by four Federal agencies having authonty and control (NRC) issued a complete revision of its General Statement of I over radioactive materials: Department of Defense (DOD), De- Policy and Procedure for Enforcement Actions (Enforcement partment of Energy (DOE), Environmental Protection Agency Policy) (60 FR 3481). In approving the 1995 revision to the En-(EPA), and Nuclear Regulatory Commission (NRC). The forcement Pohey, the Commission directed the staff to perform MARSSIM's objective is to describe a consistent approach for a review of its implementation of )he Policy after approximately building and surface soil final status surveys to meet established 2 years of experience and to consider public comments. This re-dase or risk. based release critoria, while at the same time en- port represents the results of that review. couraging an effective use of resources. NUREG-1624 DRFT FC: TECHNICAL BASIS AND IMPLEMEN-NUREG-1600 R01: GENERAL STATEMENT OF POLICY AND TATION GUIDELINES FOR A 1ECHNIQUE FOR HUMAN PROCEDURE FOR NRC ENFORCEMENT AC- EVENT ANALYSIS (ATHEANA). Draft Report For Comment.

• TiONS. Enforcement Policy.
• Ofc of Enforcement (Post 870413). Probabilistic Risk Analysis Branch (Post 941217). May 1998.

May 1998. 32pp. 9806030386. FACA. A3636:272. 404pp. 9806080242. A3687:001. This document includes the U.S. Nuclear Regulatory Commis- This report introduces a next-generation HRA method called sion's (NRC's or Commission's) revised General Statement of "A Technique for Human Event Analysis," (ATHEANA). i Policy and Procedure for Enforcement Actions (Enforcement ATHEANA was developed to address limitations identified in Policy) as it was published in the Federal Register on May 13, current HRA approaches by: (1) addressing errors of commis-1998 (63 FR 26630). The Enforcement Pokey is a general state- sion and dependencies; (2) more realistically representing the ment of policy explaining the NRC's policies and procedures in human-system interactions that have played important roles in initiating enforcement actions, and of the presiding officers and accident response; and (3) integrating advances in psychology the Commission in reviewing these actions. This policy state- with engineenng, human factors, and PRA disciplines. This re-ment is applicable to enforcement matters involving radeological port is the ste@y-step guidebook for applying the method. It health and safety of the pubic, including employees' health and describes how to: (1) select and organize the ATHEANA team, safety, the common defense and security, and the environment. (2) perform and control the structured search processes for This statement of 9eneral policy and procedure is published as human failure events and unsafe acts, including a discussion of j NUREG-1600, Rev.1 to provide wide-spread dissemination of the reasons that such events occur (i.e., the elements of error- ) i

12 Main Citations and Abstracts forcing context), (3) use the knowledge encoded in the PRA the NRC staff. In respording to the questions, the NRC staff at-along with the specialized knowledge and experience of the tempted to answer in a clear and norFlechnical form, one that ATHEANA team to focus the searches on those events and rea- an individual with no or little technical training could understand. sons that are most likeh to aifect the risk, and (4) quantify the Questions are posed on the following categories: the decommis-error-forcing contexts and probability of each unsafe act, given sioning process and decommissioned sites; licensing; regula-its context. tions; the inspection program; spent fuel, spent fuel pools, and spent fuel storage; radioactive low-level waste; transportation, li-NUREG-1625 DRF FC: PROPOSED STANDARD TECHNICAL cense termination and the ultimate disposition of the facility; SPECIFICATIONS FOR PERMANENTLY DEFUELED WES- hazards; finances; and public involvement. This document is TINGHOUSE PLANTS. Draft Report For Comment.

• March being issued for public comment. As a result of this comment, 1998.122pp. 9807270162. A4388:131.

peer review, and discussions, the final document may be mode-This NUREG report describes the staff's proposed Standai . fied from this draft. Technical Specifcations for Permanently Defueled Westing-house Plants (STS PDW). The report includes a detailed discus- NUREG-1629: THE CHARACTERIZATION OF VICKER'S MICRO-sion of the strategy followed for determining the contents of the HARDNESS INDENTATIONS AND PILE-UP PROFILES AS A STS PDW. The proposed STS PDW is being published to pro- STRAIN-HARDENING MICROPROBE. SANTOS,C. Division of vide the general public and the nuclear cornmunity with an op. Engineering Technology (Post 941217). ODETTE G.R.; portunity for comment. The contents of the proposed STS POW LUCAS.G.E.; et al. California, Univ. of, Santa Barbara, CA. April are based primarily on the Standard Technical specifcations, 1998.153pp. 9805180231. A3428:008. Westinghouse Plants (NUREG-1431, Revision 1, April 1995), Mcrohardness measurements have long been used to exam-which in tum were based on the criteria in the Nuclear Regu. ine strength properties and changes in strength propertier, in latory commission (NRC) Final Policy Statement on Techncal metals, for example, as induced by irradiation. Merohardness Specifications improvements for Nuclear Power Reactors affords a relatively simple test that can be applied to very small (SECY-93-067, 58 FR 39132, July 22,1993). The proposed volumes of material. Microhardness is nominally related to the STS PDW reflect the experience gained in the development of flow stress of the material at a fixed level of plastic strain. Fur-the Permanently defueled Techncal specifications (PDTS) for ther, the geometry of the pile-up of material around the indenta-the Trojan Nuclear Plant, the first PDTS approved by the NRC tion is related to the strairFhardening behavior of the material; that were based on the improved STS for Westinghouse Plants. steeper pile-ups correspond to smaller strain hardening rates. In As licensees begin to plan permanent shutdown of their nuclear this study the relationship between pile-up profiles and strain power plants, they are encouraged to adopt the STS PDW to hardening is examined using both experimental and analytical an extent that is practical and consistent with their licensing methods. Vicker's microhardness tests have been performed on besis. a variety of metal alloys including low alloy, high Cr and aus-en sa ss steels. W @@ Wogy amund the in@nta-NUREG-1626: FINAL ENVIRONMENTAL IMPACT STATEMENT ms has Men quank usN confocal mcmscopy t@ues. FOR THE CONSTRUCTION AND OPERATION OF AN INDE- in amn, the in@ntaMn and pb@ geonWry has bem sinn PENDENT SPENT FUEL STORAGE INSTALLATION TO lated using finite element method techniques. These results STORE THE THREE MILE ISLAND UNIT 2 SPENT FUEL AT have been used to develop improved quant feation of the re:a-THE IDAHO NATIONAL ENGINEERING AND ENVIRON- ~

• hadeNng cm MENTAL.
• Office of Nuclear Material Safety & Safeguards. , , ,'yh 'the test mat a March 1998. 219pp. 9803180129. A2611:303.

This Final Environmental impact Statement (FEIS) was pre. NUREG-1631: SOURCE DISCONNECTS RESULTING FROM pared by the U.S. Nuclear Regulatory Commission in accord. RADIOGRAPHY DRIVE CABLE FAILURES. Final Report. ance with the requirements of 10 CFR Part 51. The FEIS con. CAMPER,LW.; BROADDUS,0.A.; PELCHAT,J.M.: et al. Divh tains an assessrnent of the potential environmental impacts of sion of Industrial & Medical Nuclear Safety (Post 870729). June the construction and operation of an Indopendent Spent Fuel 1998. 250pp. 9807270358. A4378:060. l Storage installation (ISFSI) for the Three Mile island Unit 2 From November through Decernber 1997, the NRC received (tmh 2) fuel debris at the Idaho National Engineering and Envh three reports of drive cable failures associated with the ronmental Laboratory (INEEL). The NRC proposes to issue a 19 Amersham Model 6608 radiography system. All three failures I conse to the U.S. Department of Energy-Idaho Operations Of. occurred immediately behind the male connector and appeared fice (DOE-ID) which will authorize DOE-ID to store the TMi-2 to be generic in nature. Although drive cable failures have oc-fuel debris in an ISFSI DOE-ID is proposing to design, con. curred periodically within the industrial radiography industry, it struct, and operate at the Idaho Chemmal Processing Plant was uncommon to experience so many apparently identcal fait. (ICPP). The TMI-2 fuel debris would be removed from wet stor. ures within such a brief period of time. The events were re-age at the Test Area North pool, transported to the ISFSI at the viewed by the NRC to determine K the failures met the criteria ICPP, and placed in storage modules on a concrete basemat. In NRC Management Directive 8.3,"NRC incident Investigation Program," for initiating an inspection as either an Augmented in-NUREG-1627 V01: PERFORMANCE PLAN FY 1999* s n am ma ent inestgahn Team (M N FUNCHES,J.L NRC - No Detailed Affiliation Given. Februa'Y was decided that the reported failures did not satisfy all of the Mea aW hse categoM d Mspehs, M the ap-Rds perto plan rrplements the agency's stra- " # ** *"" E' '" 0 " E tegic plan by setting annual goals with measurable target levels p sure t radi graphers, and the p'ossibility that the issue went of performance for FY 1999' as required by the Govemment beyond NRC junsdiction thus affecting Agreement States war. Performance and Resutts Act, ranted NRC's attention. As a result, a Special Team inspection NUREG 1628 DRF FC: STAFF RESPONSES TO FREQUENTLY was initiated on December 22,1997. The Team, led by a Senior ASKED QUESTIONS CONCERNING DECOMMISSIONING OF Executive Service (SES) executive, included members with a NUCLEAR POWER PLANTS. Draft Report For Comment. broad knowledge in health physics, mechanical engineering, and MINNS,J.L; MASNIK,M.T. , April 1998. 74pp. 9807270179. industrial radiography operations. The inspection involved inter. A4386:280. action with three Agreement States including close coordination Through a question-and-answer format, this document pre- of inspection activities conducted within their jurisdction. This sents information to the public on decommissioning. The ques- report describes the investigation of the initially reported drive tions were taken from a variety of sources over the past several cabie failures, other failures identified during the inspection, the years, including written inquiries to the NRC ard questbns rnethodology used in the inspection, and presents the Team's asked at public meetings and during informal discussions with findings, conclusions, and recommendations.

1 Main Citations and Abstracts 13 NUREG-1632: EVALUATION OF AP600 CONTA!NMENT THER- the Reactor Safety Research Program, among other things, in MAL-HYDRAULIC PERFORMANCE. CAMPE.K.M.; the Staff Requirements Memorandum of September 9,1997. KUDRICK,J.D. Offce of Nuclear Reactor Regulation (Post This report provides observations and recommendations on en. 941001). June 1998.139pp. 9807270348. A4387:001. gineering the Safety Research Program, comments on specife The thermal-hydraulic performance of the AP600 containment research actMties, waste management research at the NRC, with respect to selected design-basis accidents (DBAs) was and the continued need for the Nuclear Safety Research Review evaluated using the CONTEMPT and CONTAIN codes. The Committee (NSRRC) function. AP600 containment design includes a passive cooling system (PCS) in the form of gravity driven water flowing on the extemal NUREG/CP-0152 V02: PROCEEDINGS OF THE FIFTH NRC/ surface of a steel conta#nment shell. This design feature cannot ASME SYMPOSIUM ON VALVE AND PUMP TEST'NG. be modeled directly within the CONTEMPT code. The CONTAIN COLACCINO,J.; SCARBROUGH,T.G. Office of Nuclear Reactor Regulation (Post 941001).

• American Society of Mechanical code was used to estimate the PCS heat transfer coeffeients Engineers. July 1998. 550pp. 9808120158. A4602:001, that were provided as input into the CONTEMPT code. The re.

suits show fair agreement in terms of containment pressure and The 1998 Symposium on Valve and Pump Testing, jointly sponsored by the Board on Nuclear Codes and Standards of the terrperature response to selected DBAs. Confirmatory analyses were made using the CONTAIN code with the intent of venfying American Society of Mechanical Engineers and by the U.S. Nu- < the Westinghouse analyses that were performed using the clear Regulatory Commission, provides a forum for exchanging WGOTHIC code. The results indicate that the Westinghouse information on technical and regulatory issues associated with pressure and temperature estimates for the AP600 appear to be the testing of valves and pumps used in nuclear power plants. reasonable and are within the applicable acceptance criteria of The symposium provides an opportunity to discuss the need to improve that testing in o~1er to help ensure the reliable perform-the Standard Review Plan. The CONTAIN code also was used to conduct a sedes of sensitivity analyses. The purpose of these ance of valves and pumps. The partcipation of industry rep-resentatives, regulators, and consultants ensures the discussion analyses was to assess the relative importance of the key < AP600 containment design features and operatng conditions. of a broad spectrum of ideas and perspectves regarding the im-One aspect of the sensitivity analyses involved the consideration provement of testing programs and methods at nuclear power plants, of selected limiting assumptions regarding the principal modes of heat transfer with respect to the containment shell and the in. NUREG/CP-0160: PROCEEDINGS OF THE OECD/CSNI SPE-temal heat sinks. CIALIST MEETING ON ADVANCED INSTRUMENTATION AND MEASUREMENT TECHNIOUES. Held in Santa Bar. NUREG-1633 DRFT FC: ASSESSMENT OF THE USE OF PO- bara.CA, March 17-20, 1997.

• Organization for Economic Co.

TASSIUM IODIDE (KI) AS A PUBLIC PROTECTIVE ACTION operation & Development. September 1998. 743PP. DURING SEVERE REACTOR ACCIDENTS. Draft Report For 9809290050. A5218:001. Comment. CONGEL,F.J.; MOHSENI A.S. Incident Response This report contains papers presented at the OECD/CSNI Branch. WILLIS,C.A.; et al. Office of Nuclear Reactor Regulation Specialist Meeting on Advanced instrumentation and Measure-(Post 941001) July 1998. 52pp. 9902090296. A6767:263. ment Techniques held at Fess Parker's Red Lion Resort in The use of potassium lodido (KI) as a supplemental protective Santa Barbara, CA on March 17-20,1997. The papers are print-action for the general public during severe reactor accidents is ed in the order of their presentat on in each session. The Spe-evaluated. A brief history of reactor accidents leading to an cialist Meeting followed soon after the CSNI Workshop on Ther-overview of severe reactor accident source terms is presented. mahHydraulics and Neutronics Codes (November 1996). At Next, thyroid and whole-body dosimetry, their associated risk these meetings, the current and future modeling needs and cur-assessments, and their relationship to severe reactor accident rent instrumentation capabilities to support these modeling ef-source terms are discussed. The medical aspects of K! use are forts were reviewed in detall. This Specialist Meetng was orga-discussed. An estimation of offsite doses for several accident nized to bring together the intemational experts on instrumenta-scenarios, with and without KI, is made. These findings are then tion, experiment, and modeling. The recent developments on related to the Chemobyl accident. A summary of intemational advanced two-phased flow instrumentations are reported in de-Duldance and practices is provided. tall. The future direcbons of instrumentation developments, ex-NUREG-1634: 1997 LOST SOURCE EXERCISE.An Exercise Of periments, and rnodeling were discussed. All the goals set for Radiological Response ThrouDb Cooperation And Coordination this meeting have been reached. Of Local, State And Federal Resources Under The National Con- NUREG/CP-0162 V01: PROCEEDINGS OF THE TWENTY-FIFTH tinDency Plan. BELANGER,B.: STEUTEVILLE.B. Environmental WATER REACTOR SAFETY INFORMATION MEETING. Plenary Protection Agency. CHAWAGA,0.; et al. Office for Analysis & Sessions, Pressure Vessel Research,BWR Strainer Blockage Evaluation of Operabonal Data, Director. August 1998. 50pp. And Other Generic Safety lasues. Environmentally Assisted Deg-9808280267. EPA 903-K-98-002. A4815295. radation Of LWR.._ MOtiTELEONE,S. Brookhaven National This was the first major exercise of its type with the Environ- Laboratory. March 1998. 370pp. 9805180401. A3427:001. mental Protection Agency as the lead Federal Agency under the This three-volume report contains papers presented at the Federal Radiological Emergency Response Plan utilizing the re- Twenty-Fifth Water Reactor Safety information Meeting held at sources of the National ContinDency Plan. The Nuclear Regu- the Bethesda Marriott Hotel, Bethesda Maryland, October 20-latory Commission acted in a support role as the EPA utilized 22,1997. The papers are printed in the order of their presen-the unified command structure to support the needs of owner / tation in each session and describe progress and results of pro-operator, State and local organizationc. The lessons teamed grams in nuclear safety research conducted in this country and provide valuable Guidance for organizations encountering similar abroad. Foreign participation in the meeting included papers circumstances in the future. presented by researchers from France, Japan, Norway, and NUREG-1635 V01: REVIEW AND EVALUATION OF THE NU- Russia. The titles of the papers and the names of the authors CLEAR REGULATORY COMMISSION SAFETY RESEARCH have been updated and may differ from those that appeared in PROGRAM.A Report To The US Nuclear Regulatory Commis- the final program of the meeting. stort

• ACRS - Advisory Committee on Reactor Safeguards. NUREG/CP-0162 V02: PROCEEDINGS OF THE TWENTY-FIFTH June 1998. 91pp. 0808250189, A4756:203. WATER REACTOR SAFETY INFORMATION MEETING. Human This report provides a response to the Commission's request Reliability Analysis And Human Performance Evaluation Tech-for the Advisory Committee on Reactor Safeguards (ACRS) to nical lasues Related To Rulemakings Risk-informed, Perform-review the NRC Safety Research Program. The Commission ance-Based initiatives... MONTELEONE S. Brookhaven National asked the ACRS to examine the need, scope, and balance of Laboratory. March 1998. 235pp. 9805180394. A3425:048.

14 Main Citation 3 and Ab tract 3 See NUREG/CP-0162,V01 abstract. ceedings. Irnprovements in ED capability, regulatory guidance, and more confidence in the use and performance of EDs NUREG/CP-0162 V03: PROCEEDINGS OF THE TWENTY-FIFTH emerged as concerns with regard to the use of the ED. WATER REACTOR SAFETY INFORMATION MEET-ING. Thermal-Hydraulic Research And Codes, Digital Instrumen- NUREG/CP-0165: TRANSACTIONS OF THE TWENTY-SIXTH tation And Control, Structural Penurmance. MONTELEONE,S. WATER REACTOR SAFETY INFORMATION MEETING. Brookhaven National Laboratory. April 1998. 358pp. MONTELEONE,S. Brookhaven National Laboratory. October 9805180351. A3423:001. 1998.148pp. 9810270361. A5560:091. See NUREG/CP-0162,V01 abstract. This contains summaries of papers to be presented at the Twenty-Sixth Water Reactor Safety Information Meeting held at NUREGICP-0163: PROCEEDINGS OF THE WORKSHOP ON the Bethesda Marriott Hotel, Bethesda, Maryland, October 26-REVIEW OF DOSE MODELING METHODS FOR DEM, 28,1998. The summaries briefly describe the programs and re-ONSTRATION OF COMPLIANCE WITH THE RADIOLOGICAL suits of nuclear safety research sponsored by the Office of Nu-CRITERIA FOR LICENSE TERMINATION. NICHOLSON.T.J. clear Regulatory Research, U.S. NRC. Summaries of invited pa-Division of Regulatory Applications (Post 941217). pas conceming nuclear safety issues kom RS. govemment PARROTT.J.D. Division of Waste Management (NMSS 940403). , laboratories, the electric utilities, the nuclear industry, and from May 1998.123pp. 9806080229. A3688:041. fmeign govemments and industry are also ,ncluded.i The sum-The public " Workshop on Review of Dose Modeling Methods manes have been compiled in one report to provide a basis for for Demonstration of Compliance with the Radiological Criteria rneaningful discussion and information exchange during the for License Termination" was held at the NRC Headquarters Au-course of the meeting, and are given in the order of their pres-ditorium, Rockville, Maryland, on November 1314,1997. The entation in each session. workshop was one in a series to support NRC staff development of guidance for implementing the final rule on " Radiological Cri- NUREG/CR-4219 V13 N2: HEAVY-SECTION STEEL TECH-tena for License Terminaton." The workshop topics included NOLOGY PROGRAM. Semiannual Progress Report For April - discussion of: dose models used for decommissioning reviews; September 1996. PENNELL,W.E. Oak Ridge National Labora-identification of cnteria for evaluating the acceptability of dose tory. August 1998. 103pp. 9808310131. ORNL/TM-9593. models; and selecton of parameter values for demonstrating A4851:029. compliance with the final rule. The 2-day public workshop was The Heavy-Section Steel Technology (HSST) Program is con-jointly organized by RES and NMSS staff responsible for review- ducted for the U.S. Nuclear Regulatory Commission (NRC) by ing done modeling methods used in decommissioning reviews, the Oak Ridge National Laboratory (ORNL). The program focus The wJ dshop was noticed in the Federal Register (62 FR is on the development and vahdation of technology for the as-51706). The workshop presenters included: NMSS and RES sessment of fracture-prevention margins in commercial nuclear staff, who discussed both dose modeling needs for licensing re- reactor pressure vessels. The HSST Program is organized in views, and development of guidance related to dose modeling seven tasks: (1) program management (2) constraint effects an-and parameter selection needs; DOE national laboratory sci- alytcal development and validaton, (3) evaluaton of cladding entists, who provided responses to earlier NRC staff-developed effects, (4) ductile to cleavage fracture mode conversion, (5) questions and discussed their various Federally-sponsored dose fracture analysis methods development and applications, (6) modem (i.e., DandD, RESRAD, and MEPAS codes); and an material property data and test methods, and (7) integration of EPA scientist, who presented details on the EPA dose assess- results into a state-of-the-art methodology. The program tasks ment model (i.e., PRESTO coce). The workshop was formatted have been structured to place emphasis on the resolution frac-to provide opportunities for the attendees to observe computer ture issues with near-term licensing signifcance. Resources to demonstrations of the dose codes presented. More than 120 execute the research tasks are drawn from ORNL with sub-workshop attendees from NRC Headquarters and the Re0 ions, contract support from universities and other research labora-Agreement States; as well as industry representatives and con- tories. Close contact is maintained with the sister Heavy-Section sultants; scientists from EPA, DOD, DNFSB, DOE, and the na- Steel Irradiation Program at ORNL and with related research tional laboratories; and interested members of the public partici- programs both in the United States and abroad. This re, port pro-pated. A complete transcript of the workshop, including vides an overview of principal developments in each of the viewgraphs and attendance lists, is available in the NRC Pubic seven program tasks from April 1996 - September 1996. Document Room. This NUREG/CP documents the formal pres-NUREG/CR-4219 V14 N1: HEAVY-SECTION STEEL TECH-entations made dunng the workshop, and provides a preface NOLOGY PROGRAM. Semiannual Progress Report For October outlining the workshop's focus, objectives, background, topics 1996 - March 1997. PENNELL W.E. Oak Ridge National Lab-and questions provided to the invited speakers, and those oratory. November 1998.100pp. 9812160006. ORNL/TM-9593. raised during the panel discussion. NUREG/CP-0163 also pro- A6218:120 vides techncal bases supporting the development of decommis- The Heavy-Section Steel Technology (HSST) Program is con-sionsg guidance. ducted for the U.S. Nuclear Regulatory Commission (NRC) by NUREG/CP-0164: PROCEEDINGS OF THE WORKSHOP ON the Oak Ridge National Laboratory (ORNL). The program focus ELECTRIC DOSIMETRY. Held in Gaithersburg, Maryland On is on the development and validation of technology for the as-October 14-16, 1997. SWINTH,K.L August 1998. 203pp. sessment of fractureprevention margins in commercial mclear 9808250171. A4756:001. reactor pressure vessels. The HSST Program is organized in From October 14-16, 1997, a workshop on electronic dosim- seven tasks: (1) program management, (2) constraint effects an-etry was held in Gaithersburg, MD with representatives attend- a!ytcal development and validation, (3) evaluation of cladding ing from Canada, France, Great Bntain and South Africa as well effects, (4) ductile to cleavage fracture mode conversion, (5) as the U.S. The purpose of the workshop was to discuss the fracture analysis methods development and applications, (6) present status of electronic dosimetry and, in particular, the ap- material property data and test methods, and (7) integration of pication of electronic dosimeters (EDs) for primary dosimetry, results into a state-of-the-art methodology. The program tasks The workshop proceedings contain presentations and sum- have been structured to place emphasis on the resolution frac-maries of breakout discussions from the meeting. Electronic ture issues with near-term licensing significance. Resources to dosimeters are being used for primary dosimetry in hmited appli- execute the rer,earch tasks are drawn from ORNL with sub-cations. Information on the mechanism for approval as primary contract support from universities and other research labora- l tories. Close contact is maintained with the sister Heavy-Section dosimetry in Great Britain was presented. Information on tech-nological advances, the status of performance standaros, and Steel Irradiation Program at ORNL and with related research

                                                                                                                                           )

current experience with the ED are also provided in the pro- programs both in the United States and abroad. This report pro. I

Main Citations cnd Abstracts 15 vides an overview of principal developments in each of the (CGR) tests were completed on compact-tension specimens I seven program tasks from October 1996 - March 1997, from several heats of Alloys 600 and 690 in air, high-punty I NUREGICR-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS water, and simulated pressurized water reactor environments. ' SYSTEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM CGR correlations were developed as a functon of load:ng and FOR SHIPPING CASK DESIGN REVIEW. User's Manual to environmental parameters. Version 3a. MOK G.C.; THOMAS G.R.; GERHARD,M.A.; et al. NUREGICR-4674 V25: PRECURSORS TO POTENTIAL SEVERE l Lawrence Livermore National Laboratory. March 1998. 219pp. CORE DAMAGE ACCIDENTS: 1996. A Status Report. 9803260397. UCID-20674. A2727:001. BELLES,R.J.; CLETCHER,J.W.; COPINGER,0.A.; et al. Oak l SCANS (Shipping Cask Analysis System)is a microcomputer Ridge National Laboratory. December 1997. 271pp. l based system of cornputer programs and databases developed 9802200043. ORNL/NOAC-232. A2250:001. l ct the Lawrence Livermore National Laboratory (LLNL) for eval- This report describes the 14 operational events in 1996 that  ! uating safety analysis reports on spent fuel shipping casks. affected 13 commercial light-water reactors and that are consid- j SCANS is an easy-to-use system that calculates the global re- ered to be precursors to potential severe core damage acci- j sponse to impact loads, pressure loads and thermal conditions, dents. All these events had conditional probabilities of subse-providing reviewers with an independent check on analyses sub- quent severe core damage greater than or equal to 1.0 x 10(- l i mitted by licensees. SCANS is based on microcomputers com- 6). These events were identified by first computer-screening tne l patible with the IBM-PC famny of computers. The system is 1996 licensee event reports from commercial light-water reac-composed of a series of menus, input programs, crask analysis tors to identify those events that could potentially be precursors. programs, and output display programs. All data is entered Candidate precursors were selected and evaluated in a process through fill-in-the blank input screens that contain descriptive similar to that used in previous assessments. Selected events data requests. Analysis options are based on regulatory cases underwent engineering evaluation that identified, analyzed, and described in the Code of Federal Regulations 10 CFR 71 and documented the precursors. Other events designated by the Nu-Regulatory Guides published by the U.S. Nuclear Regulatory clear Regulatory Commission (NRC) also underwent a similar Commission in 1977 and 1978, evaluation. Finally, documented precursors were submitted for NUREG/CR-4667 V24: ENVIRONMENTALLY ASSISTED CRACK- review by licensees and NRC headquarters and regional offices ING IN LIGHT-WATER REACTORS. Semiannual Re- to ensure the plant design and its response to the precursor port, January-June 1997. CHOPRA,0.K.; CHUNG,H.M.; were correctly characterized. This study is a continuation of ear. GRUBER E.E.; et al Argonne National Laboratory. April 1998. lier work, which evaluated 1969-1995 events. The report dis-115pp. 9805180239. ANL-98/6. A3428:161, cusses the general rationale for this study, the selection and . This report summarizes work performed by Argonne National documentation of events as precursors, and the estimation of l Laboratory on fatigue and environmentally assisted cracking conditional probabilities of subsequent severe core damage for 1 (EAC) in light water reactors from January 1997 to June 1997. the events. Topics that have been investigatodinclude (a) fatigue of carbWtJREG/CR-4674 V26: PRECURSORS TO POTENTIAL SEVERE low-alloy, and austenitic stainless steels (SSs) used in reactor CORE DAMAGE ACCIDENTS:1997. A Status Report. i piping and pressure vessels, b) Irradiation-assisted stress corro- BELLES.R.J.; CLETCHER.J.W.; COPINGER,0.A.; et al. Oak l sion cracking of Typs 304 and 304L SS, and (c) EAC of Alloys Ridge National Laboratory. November 1998. 200pp. 1 600 and 690. Fatigue tests were conducted on ferritic and aus- 9812160008. ORNL/NOAC-232. A6219.001. tenitic SSs in water that contained various concentrations of dis- This report describes the five operational events in 1997 that sofved oxygen (DO) to deterrrune whether a slow strain rate ap- affected five commercial light-water reactors (LWRs) and that plied during various portions of a tensile-loading cycle is equally are considered to be precursors to potential severe core dam-effective in decreasing fatigue * +. Slow-strain-rtte-tensile tests age accidents. All these events had conditional probabilities of were conducted in simulated bumng water reactor (BWR) water subsequent severe core damage greater than or equal to 1.0 x at 288 degrees C on SS specimens irradiated to a low fluence 10-6. These events were identified by first computer-screening in the Halden reactor and the results were compared with simi- the 1997 licensee event reports from commercial LWRs to iderb tr data from a controhblade sheath and neutron-absorber tubes lify those events that could be precursors. Candidate precursors irradiated in BWRs to the same fluence level Crack-growth-rate were selected and evaluated in a process similar to that used t:sts were conducted on compact-tension specimens from sev- in previous assessments. Selected events underwent engineer-er;l heats of Alloys 600 and 690 in low-DO, simulated pressur- ing evaluation that identified, analyzed, and documented the ized water reactor environments. precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, NUREG/CR-4667 V25: ENVIRONMENTALLY ASSISTED CRACK-ING IN LIGHT. WATER REACTORS. Semiannual Report, July- documented precursors were submitted for review by licensees Decernber 1997. CHOPRA,G.K.; CHUNG,H.M.; GRUBER,E.E.; and NRC headquarters to ensure that the plant design and its response to the precursor were correctly charactenzed. This et al. Argonne National Laboratory. Septener 1998. 106pp. study is a continuation of earlier work, which evaluated 1969-9810080067. AML-98/18. A5352:154. 1996 events. The report discusses the general rationale for this This report summarizes work petformed by Argonne National study, the selection and documentation of events as precursors, Laboratory on fatigue and environmentally assisted cracking and the estimation of conditional probabilities of subsequent se-(EAC) in light water reactors from July 1997 to December 1997, vere core damage for the events. Topics that have been investigated include (a) fatigue of aus-tenitic stainless steels (SSs), (b) Irradiation-assisted stress cor- NUREG/CR-5342: ASSESSMENT AND RECOMMENDATIONS rosion cracking of austenitic SSs, and (c) EAC of Alloys 600 and FOR FISSILE-MATERIAL PACKAGING EXEMPTIONS AND 690. Fatigue tests were conducted on austenitic SSs in water GENERAL LICENSES WITHIN 10CFR PART 71. PARKS,C.V.; that contained various concentrations of dissolved oxygen to de- HOPPER,C.M.; LICHTENWALTER,J Oak Ridge National Lab-termine whether a slow strain rate applied during various por- oratory. July 1998. 110pp. 9808310084. ORNL/TM-13607, tions of a tensile-loading cycle la equally effective in decreasing A4849:179. fatigue life. Slow-strain-rate-tenslie tests were conducted in sim- This report provides a technical and regulatory assessment of ulated boiling water reactor (EWR) water at 288 degrees C on the fissile material general licenses and fisslie material exemp-SS specimens irradiated to a low and medium fluence in the tions within Title 10 of the Code of Federal Regulations Part 71. Halden reactor, and the results were compared with similar data The assessment included literature studies and calculational from a controhblade sheath and neutron-absorber tubes irradi- analyses to evaluate the technical criteria; review of current in-ated in BWRs to the same fluence levels. Crack-growth-rate dustry practice and concems; and a detailed evaluation of the

16 Main Citation 3 and Abstracts regulatory text for clanty, consistency and relevance. Rec- pansion anchors. These tend to pull out and pull through under ommendations for potential consideration by the Nuclear Regu- dynamic loading. Wadg&-iype expansion anchors should be latory Commission staff are provided. The recommendations call evaluatwo ,'iveaually to determine their seismic adequacy, for a simplification and consolidation of the general licenses and huREG/CR-5485: GUIDELINES ON MODELING COMMON-a change in the technical criteria for the first fissle material e" CAUSE FAILURES IN PROBABILISTIC RISK ASSESSMENT. emptions. MOSLEH,A. Maryland. Univ. of, College Park, MD. NUREG/CR-5361: SEISMIC ANALYSIS OF PIP %G. Final Program RASMUSON,D.M. Division of Safety Programs (Post 870413). Report. JAQUAY,K. June 1998 400pp. 9807060324. MARSHALL,F.M. Idaho National Engineering & Environmental A4008:001. Laboratory. November 1998. 200pp. 9812240092. This report provides a sumanary of the work conducted by the INEELEXT9701327. A6313:328. Energy Technology Engineering Center (ETEC) under the U.S. This report provides a set of guidelines to help probabilistic Nuclear Regulatory Commission Seismic Analysis of Piping Pro. risk assessment (PRA) analysts in modeling common cause fail-gram. ETEC was contracted by the NRC to review the technical ure (CCF) events in commercial nuclear power plants. The aim bases for new rules in the ASME Boiler and Pressure Vessel is to enable the analyst to identify important comrnon cause Code, Section lli related to seismic analysis of piping systems vulnerabilities, incorporate their impact into system reliabihty in nuclear power plants, and evaluate the cumulatrve impact of rnodels, perform data analysis, and quantify system unavaih these changes in design criteria on overall safety margins of ability in the presence of CCFa. Much of the material in this re-these piping systems. The ETEC effort is documented in this re. port has been presented in previous reports issued by United port. States Nuclear Regulatory Commission (NRC). The present document brings together the key aspects of these procedural NUREG/CR-5372: EXPERIMENTS ON INTERACTIONS BE- guidel nes supplemented by additional insights gained from their TWEEN ZlRCONIUM-CONTAINING MELT AND WATER. application, and enhanced by the capabilities of the CCF soft-CHO,D.H.; ARMSTRONG,0.R.; GUNTHER,W.H. Argonne Na- ware and its data analysis capabilities, recently developed by tional Laboratory. October 1998.176pp. 9810230178. FIN A- the NRC. 2284. A5512d3L The resutts of two series of experiments on explosrve inter. NUREG/CR-5496: EVALUATION OF LOSS OF OFFSITE POWER actions between zirconiurmcontaining melt and water are de- EVENTS AT NUCLEAR POWER PLANTS: 1980 - 1996. AT-WOOD,C.L; KELLY,D.L.; MARSHALL,F.M.; et al. Idaho Na-scribed. The first senes of experiments involved dropping 1 kg tional Engineering & Environmental Laboratory. November 1998. batches of zirconium / zirconium dioxide (Zr/ZrO2) mxture melt 124pp.9811270096. INEELEXT9700887. A5962:001. into a column of water while the second series employed 1.2 kg it is recognized that the availability of AC power to commer-batches of zirconium / stainless steel (Zr/SS) mixture melt. A total cial nuclear power plants is essential for safe operations and ao-of 14 successful tests (nine triggered and five untriggered) were cident recovery. A loss of offsite power (LOSP) event, therefore conducted with Zr/Z 02 mixtures and a total of 8 tests (five trig- is considered an important contritutor to total risk at nuclear gered and three untriggered) were conducted with Zr/SS mix- power plants. In 1988, the NRC published NUREG-1032 to re-tures. The test results indicate that the explosion energetics port on an evaluation of the risk from actual LOSP events that were augmented by the zirconium-water reaction, the extent of had occurred at nuclear power plants within the United States augmentation being dependent on the zirconium content in the up through 1985. This report documents a similar study whose melt. The explosion energies for tests involving high zirconium primary objective was to update the LOSP model parameters, contents were mostly in the range of 2-3% of the combined ther- frequercy and recovery time, using power plant event data from mal and chemical energy available, while the energy conversion 1980-1996. An additional objective is to re-examine the engi-ratios for Zr/SS mxtures of low zirconium content were consid- neering insights conceming LOSP events. NUREG/CR-5497: COMMON CAUSE FAILURE PARAMETER NUREG/CR-5434: ANCHOR BOLT BEHAVIOR AND STRENGTH ESTIMATIONS. MARSHALL,F.M. Idaho National Engineering & DURING EARTHOUAKES. KLINGNER,R.E.; HALLOWELL,J.M.; Environmental Laboratory. MOSLEH,A. Maryland, Univ. of, Coi-LOTZE,D.; et at Texas, Univ. of, Austin, TX. August 1998- lege Park, MD. RASMUSON.D.M. Division of Safety Programs 389pp. 9808310094. A4850:001. (Post 870413). October 1998. 381pp. 9811170329. This is the final report for NRC Contract No. 03-92-05 (" Ark INEELEXT9701328. A5847:001. chor Bolt Behavior and Strength During Earthquakes"). The ot> This report documents the quantitative results of the common jective of this project was to obtain technical information to cause failure (CCF) data collection effort described in Volumes verify, by testing, the adequacy of the assumption used in the 1 - 4 of the Common Cause Failure System Database and Anab US nuclear power plant designs that the behavior and strength ysis System (References 2-5), as well as some qualitative in-of anchor bolts (cast-irkplace, expansion, and bearing-type (un- sights about the data. These results are for use in Probabilistic dercut) and their supporting concrete under seismic loads do not Risk Assessment (PRA) studies of commercial nuclear power differ significantly from those for static conditions (Klingner plants in the U.S. It summarizes the results of the parameter es-1991). To that end, a research program was carried out on the timation quantification process, performed on the CCF data, as dynamic behavior of anchors in concrete, in this report, that re- described in Volume 2 of that series of reports. Equipment faih search program is described; the principal results are summa- ures that contribute to CCF events are identified during rized; and the principal conclusions are given. Four documents searches of Licensee Event Reports and Nuclear Plant Reh-have already been submitted to the Nuclear Regulatory Com- ability Data System failure reports. Once CCF events are identi-mission, giving detailed resutts of this research: Rodriguez fied by screening reports of equipment failures, they are coded (1995); Hallowell (1996); Lotze (1996); and Zhang(1997). The for entry into a personal computer storage system. Once all data intent of this report is to summarize and synthesize those pre- for a specific system and component data set have been erd viously submitted documents, and to guide the reader in obtain- tered, parameter estimations are performed, producing the re-ing more detailed information from them. Most of this report is suits. The results of the database analysis are presented here adapted from those documents. The basic conclusions of the re- as a summary of the entire database, and as individual reports search are: 1) Under seismic-type loading, the capabilities of for individual system / component combinations describe the sys-most anchors tested in this study were at least as high as under tem and component boundaries, along with the guidelines for quashstatic loading. 2) As a result, most anchors tested in this identifying CCF events that may be unique to the data set. The study, if designed for ductile behavior under quashstatic loading, quantitative results are presented as both alpha factors and muh I would behave in a ductile manner under seismic-type loading as tiple Greek letter parameter estimations. The a'pha factor uncer- l well. 3) The above conclusions are not true for wedge-type ex- tainty distributions are also presented.

Main Citations Cnd Abstracts 17 NUREG/CR-5498: SINGLE-PHASE AND TWO-PHASE NATURAL The U.S. Nuclear Regulatory Commission (NRC) has sup-CIRCULATION TESTS IN THE PUMA FACILITY. ported research at Pacific Northwest National Laboratory REVANKAR.S.T.; ISHil.M.; XU,Y.; et al. Purdue Univ., West La- (PNNL) to establish bases for estimating the number and sizes tyette, IN. August 1998. 200pp. 9808310127. PU/NE-98-7. of flaws in reactor pressure vessel welds. This report describes A4852:001. a collaborative effort between PNNL and Rolls Royce and Asso-Single-phase and tw& phase natural circu!ation tests were ciates (RRA), which developed an approach to predct flaw dis-conducted at the Purdue University MulthDimensional Integral tributions based on knowledge of the vessel dimensions, weld-Test Assembly (PUMA) using the reactor pressure vessel and ing practices, and inspect;on procedures. The approach uses isolation condensers. The experimental data for steady state knowledge of welding experts and mathematical rnodelling to single-phase natural circulation are presented. Transient single- simulate the weld manufacture and the errors that lead to dif-phase natural circulation data are also dscussed. Theoretcal ferent types of defects. The model addresses main vessel welds predictions of the steady state natural circulation rate and the and cladding welds. Two meetings were held to enable RRA temperature difference between the hot and cold legs compared and PNNL to engage in discussions with experts in the areas well with the measured data. In two-phase natural circulation of vessel welding and inspection practices. The discussions con-flow, depending on the RPV water level, the entrance flow to the firmed the basic soundness of the original RRA methodology, hot leg was either single-phase water, two-phase mixture, or and provided insights which improved the quality of the assump-steam. The flow was unsteady when the entrance flow was a tions and inputs to the model. The final product of the project tw& phase mixture. Analysis of the unateady data using time se- was the RR-PRODIGAL computer code, which has boen applied ries was performed and the flow was identified i.s cyclic, near to predict the flaws in the Pressure Vessel User Research Facil-cyclic or chaotic. ity (PVRUF) vessel, with good agreement being observed. This report describes the flaw simulation rnethodology, provides guid-NUREG/CR-5500 V01: RELIABILITY STUDY: AUXILIARY /EMER- ance in use of the computer code, and describes exampie cal-

 ,        GENCY FEEDWATER SYSTEM, 1987-1995. POLOSKI,J.P.;                         culations.

n t A ust 8. " - NUREG/CR-5534: CRITICAL HEAT FLUX (CHF) PHENOMENON 9808210000. INEELEXT9700740. A4712:001" ON A DOWNWARD FACING CURVED SURFACE: EFFECTS This report documents an analysis of the safety-related per- OF THERMAL INSULATION. CHEUNG,F.B.; Llu,Y.C. Pennsyh formance of the auxiliary /ernergency feedwater (AFW) system at vania State Univ., University Park, PA. September 1998. 92pp. United States commercial pressurized water reactor plants dur- 98 14 SU

                                                                                                      ,   na f u al        ectio  ling and critical heat ing the period 1987-1995. Both a risk-based analysis and an en-gineering analysis of trends and pattems were performed on               flux on the outer surface of a heated hemispherical vessel sur-data from AFW system operational events to provide insights              rounded by a thermal insuh she were MvestWed &

into the performance of the AFW system throughout the industry perimentally in the subscale boundary layer boiling (SBLB) test and at a plant-specife level. Comparisons were made to prob. facility. The objectives were to measure the rate of boiling heat abilistic risk assessments and individual plant evaluations for 72 We to oMerve N Wor of N MilWW h phase motion in the annular channel between the hemispherical plants to indicate where operational data either support or fail to support the assumphons, models, and data used to develop vessel and the insulation structure, and to determine the flow ef-the AFW system unreliability estimates. fect on the critical heat flux on the downward-facing hemi-spherical surface. High-speed photographic records revealed the NUREG/CR-5501: ADVANCED INSTRUMENTATION AND MAIN. presence of violent cyclic ejection of the vapor masses gen-TENANCE TECHNOLOGIES FOR NUCLEAR POWER erated by boiling on the vessel outer surface whch resutted in PLANTS. HASHEMIAN,H.M.; RIGGSBEE,E.T.; MITCH- a buoyancy-driven, upward, co-current two-phase flow through ELL.D.W.; et al. Analysis & Measurement Services Corp. August the channel. A one-dimensional analysis of the co-current flow 1998. 500pp. 9810060181. A5289:001. in the channel indicated that the induced mass flow rate was a Advanced sensors and new testing and maintenance tech- strong function of the wall heat flux and the size of the channel. nologies have become available over the last ten years that Measurements of the local boiling heat fluxes and the local wall have the potential for use in nuclear power plants to replace superheats showed a signifcant spatial variation of the nucleate outdated, obsolete, and troublesome instruments, provide for bolling heat transfer, evidently due to the effect of the upward management of aging of critcal plant equipment. optimize main- cocurrent flow. At high heat flux levels, the steam venting proc-tenance activities, reduce rnaintenance costs and personnel ra- ess through the minimum gap of the channel was found to be diation exposure, and at the same time, improve plant safety highly unsteady and chaotic. Owing to the chaotic steam venting and availability. Some of these new developments are described process, the local critical heat flux had the lowest value near the in this report. In addition, the results of an assessment of their minimum gap. However, this lowest value is still higher than the qualification and potential for nuclear power plants is presented. corresponding local CHF value for the case without thermal in-NUREG/CR-5502: ENGINEERING DRAWINGS FOR 10 CFR PART 71 PACKAGE APPROVALS. SHEAFFER,M.K.; THOM- NUREG/CR-5549: ENVIRONMENTAL ASSESSMENT RENEWAL AS,G.R.; DANN R.K.; et al. Lawrence Uvermore National Lat>- OF MATERIAL LICENSES FOR ALARON CORP. NORTHEAST oratory. May 1998. 20pp. 9806100427. URCL-lD-130438. REGIONAL SERVICE FACILITY, WAMPUM, PENNSYLVANIA. A3728:158. BLASING,T.J.; CADA,G.F.; EASTERLY C.E.; et al. Oak Ridge This report provides information for preparing drawings of National Laboratory. November 1998. 59pp. 9812220072. transportation packages submitted in an application for approval A6303 274. under 10 CFR Part 71. It dscusses the purpose of these draw. Under material licenses from the U.S. Nuclear Regulatory ings and describes the recommended format and technical con- Commission (NRC), ALARON Corporation's Northeast Regional tent appr@ie for package applications. Examples of fre- Service Facility at Wampum, Pennsylvania, provides treatment, quently used drawing symbols are also provided. decontamination, compaction, and repackaging services for gen-erators of radioactively contaminated materials. The NRC is NUREG/CR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING considering renewal of the matenal licenses and two modifica-THE PROBABluTIES OF DEFECTS IN REACTOR PRESSURE tions proposed by the licensee (increased storage time for some VESSEL WELDS. CHAPMAN,0.J. United Kingdom. materials and changes to a building exhaust system). This Envi-SIMONEN,F.A. Battelle Memorial Institute. Pacific Northwest ronmental Assessment evaluates whether the proposed action National Laboratory. October 1998.152pp. 9811060235. PNNL- could have or contribute to sigraficant impacts on the human en-11898. A5755:133. vironment, and it assesses the potentialimpacts of two forms of , I l

18 Main Citation 3 and Abstracts a no action alternative: renewing the Icenses without change methods are being improved, resulting in a large selection of and terminating the current licenses. The results of the impact methods and greater accuracy and applicability for most meth-analysis indcate that (1) the proposed action would not have ods. In addition, most dating methods are undergoing continued signifcant impacts on environmental resource areas, (2) the po- testing to better understand their limitations and applicability. tential impacts of renewing the Icenses without change would This has led to more effective application, and occasionally, de-be the same as those of the proposed action, and (3) termi- creased use of specific methods. Despite the many dating meth-nating the licenses would eliminate the small impacts of facility ods available and these new advances, obtaining accurate and operation but eliminating or replacing facility capabilities could precise age estimates of Quaternary deposits and landforms re-lead to environmental impacts elsewhere. mains a challange. Best results are obtained when the NUREG/CR-5559: SINGLE- AND CROSS-HOLE PNEUMATIC paleosessmologist and geochronologist closely collaborate, when TESTS IN UNSATURATED FRACTURED TUFFS AT THE age estimates are venfied by the application of multiple dating l APACHE LEAP RESEARCH SITE: PHENOME- methods, when error analysis accounts for all sources of uncer- 1 NOLOGY, SPATIAL VARIABILITY,CONNECTIVITY AND tainty, and when studies undergo technical peer review. SCALE. ILLMAN,W.A.; THOMPSON.D.L.; VESSELINOV V.V. NUREG/CR-5570: APPLICATION OF THE NCSA HABANERO Arizona, Univ. of, Tucson, AZ, November 1998. 212pp. TOOL FOR COLLABORATION ON STRUCTURAL INTEGRITY 9811100329. FIN W 6388. A5781:129. ASSESSMENTS. BASS,B.R.; KRUSE,K. Oak Ridge National This report documents research results from a series of field Laboratory. DODDS R.H.; et al. Illinois, Univ. of, Urbana. IL. No-experiments and analyses used to test interpretive models for vember 1998.52pp.9811270094. ORNUTM-13674. A5962:191. , investigating the role of fractures in fluid flow through unsatu- The Habanero software was developed by the National Cen-rated, fractured tuffs. The report summarizes the experimental ter for Supercomputing Applications at the University of Illinois, l design of single-holt, : J cross-hole pneumate injection tests, Urbana-Champaign, as a framework for the collaborative shar-including borehole configuration and testing schedules, data col- ing of Java applications. The Habanero tool performs distributed lection system, interpretive models developed and tested, data, communication of single-user, computer software interactions to and conclusions. Single-hole tests initially used steady-state a multiuser collaborat:ye environment. An investigation was con-analysis to obtain permeabihty values based solely on late pres- ducted to evaluate the capabilities of the Habanero tool in pro-sure data, and subsequently used pressure and pressure-deriv- viding an internet-based collaborative framework for researchers ] ative typewve analyses on the transient data. Air located at different sites and operating on different workstations. permeabihties determined from transient analysis agree well with These collaborative sessions focused on the sharing of test data those derived from steady state analysis. Single-and cross hole and analysis results from materials engineering areas (i.e., frac-pneumatic tests were analyzed using a graphical matching pro- ture mechanics and structural integrity evaluations) related to re-cedure based on type-curves of pressure and pressure-deriva- actor pressure vessel safety research sponsored by the U.S. , tives, and an automatic parameter estimation rnethod based on Nuclear Regulatory Commission. This report defines collabo- l a three-dimensional, finite-volume code (FEHM) coupled with an rative-system requirements for engineering applications and pro- ) inverse code (PEST). Analyses of pressure data from individual vides an overview of collaborative systems within the project. monitored intervals using these two methods, under the as- The installation, application, and detailed evaluation of the per-sumption that the rock acts as a uniform and isotropic fractured formance of the Habanero collaborative tool are compared to porous continuum, yield comparable results. These results ird those of another commercially available collaborative product. ciude information about pneumatic connections between the in. Recommendations are given for future work in collaborative t jection and monitored intervals, corresponding directional air communications. l permeabilities, and air-filled porosities. Together with the results NUREG/CR-5591 V04 Nt: HEAVY-SECTION STEEL IRRADIA-of earlier site investigations, s!ngle- and cross-hole test analyses TION PROGRAM. Semiannual Progress Report For October reveal that, at the Apache Leap Research Site: (1) the pneu- 1992 ihrough March 1993. CORWIN.W.R. Oak Ridge National matic pressure behavior of fractured tuff is amenable to analysis Laboratory. April 1998. 50pp. 9805180222. ORNUTM-11558. by rnethods that treat the rock as a continuum on scales ranging A3428 277. from rneters to tens of meters; (2) this continuum is representa- The primary goal of the Heavy-Section Steel Irradiation Pro-tive primarily of interconnected fractures; (3) its pneumatic prop- gram is to provide a thorough, quantitative assessment of ef-erties vary strongly with location, direction and scale, in par- fects of neutron irradiation on material behavior, and in par-ticular, the mean of pneumatic permeabilities increases, and ticular the fracture toughness properties, of typical pressure ves-their variance decrease, with scale; (4) this scale effect is rnost sel steels as they relate to light-water reactor pressure-vessel I probably due to the presence in the rock of various size frac- integrity. Effects of specimen size, material chemistry, product tures that are interconnected on a variety of scales; and (5) l form and microstructure, irradiation fluence, flux, temperature given a sufficiently large sample of spatially-varying pneumatic and spectrum, and post-irradiation annealing are being exam-rock properties on a given scale of measurement, these prop- ined on a wide range of fracture properties. During this reporting erties are amenable to analysis by geostatistical methods, which period, irradiated crack-arrest specimens were tested; charpy V- i treat them as correlated random fields defined over a con- notch specimens of high-cooper weld metal were annealed and I tinuum. tested; a fracture mechanics evaluation of the unirradiated Mid-NUREGICR-5562: DATING AND EARTHOUAKES: REVIEW OF land low upper-shelf weld was nearty completed; irradiation of OUATERNARY GEOCHRONOLOGY AND ITS APPLICATION the first large Midland capsule was completed; refined calcula-TO PALEOSEISMOLOGY. SOWERS.J.M.; LETTIS,W.R. Affili- tions and detailed experimental measurements of the exposure

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ation Not Assigned. NOLLER,J.S. Vanderbilt Univ., Nashville, parameters in the High Flux Isotope Reactor were evaluated; in-TN March 1998. 850pp. 9807100109. A4103:001. cavity irradiation of vessel support materials were completed; Quatemary geochronology, or the dating of Quatemary de- unirradiated microstructural characterization of a Russian reactor posits and landforms, is critical to paleoseismology; it provides vessel steel was completed; collaborative investigations of in-the means of assessing rates of deformation and the timing of cascade point-defect generation experiments and investigations past displacements. This report provides: (1) reviews of twenty- of a very wide range of flux levels on low-temperature embntile-two Quatemary geochronologic methods, each authored by an ment were initiated; baseline testing for an ASTM round robin active researcher, (P.) a discussion of the application of on reconstituted Charpy V-notch specimens was completed; in-geochronology to paleoseismology, including twelve separately formal agreement was reached on collaboration on pressure authored case studies, and (3) the results of four original field vessel material from the Japan Power Demonstration Reactor; and laboratory studies. Quatemary Doochronology is a growing impact and tensile specimens of two U.S. reactor vessels mate- l f field in which new methods are being developed and existing rials were encapsulated and irradiation begun in a Russian reac-l l l 1

Main Citations and Abstracts 19 tor; and tensile and impact specimens were tested for three ANATECH Research Corporation under contract with Sandia stalniess steel welds aged for up to 20,000 h. National Laboratories. Model construction and hydraulic testing NUREGICR-5601 V08 N1: HEAW SECTION STEEL IRRADIA- were commissioned by the Central Electricity Generating Board TION PROGRAM. Semiannual Progress Report For October (CEGB) in the United Kingdom. To further its understanding of 1996 Through March 1997 ROSSEEL,T.M. Oak Ridge National containment behavior, the United States Nuclear Regulatory Laboratory. February 1998. 66pp. 9803180030. ORNUTM- Commission (USNRC) also partcipated in the test program 11568. A2612:158 through an agreement with the United Kingdom Atome Energy Maintaining the integrity of the reactor pressure vessel (RPV) Authority (UKAEA). Sandia National Laboratories served as in o light-water-cooled nuclear power plant is crucial in pre-tecMcal agent for N MRC. venting and controlling severe accidents that have the potential NUREG/CR-6119 V01 R1: MELCOR COMPUTER CODE MANU- l for major contamination release. Because the RPV is the only ALS. Primer And Ussrs' Guides. Version 1.8.4,Ju'y 1997. l key safety-related component of the plant for which a redundant GAUNTT,R.O.; COLE.R.K. Sandia National Laboratories.  ! backup system does not exist, it is imperative to fully under. HODGE,S.A.; et al. Oak Ridge National Laboratory. May 1998. l stand the degree of irradia3orkinduced degradation of the RPV's 625pp. 9806100430. SAND 97-2398 A3726:001, ) fracture resistance that occurs during service. For this reason, MELCOR is a fully integrated, engineering-level computer j the Heavy-Section Steel irradiation (HSSI) Program has been code that models the progression of severe accidents in light ' established its primary Doal is to provide a thorough, quarp water reactor nuclear power plants. MELCOR is being devel-titative assessment of the effects of neutron irradiation on the oped at Sandia National Lauoratories for the U.S. Nuclear Reg-matorial behavior and, in particular, the fracture toughness prop. ulatory Commission as a second-generation plant risk assess-erties of typical pressure-vessel steels os they -late to light- ment tool and the successor to the Source Term Code Package, water RPV integrity. Effects of specimen size, material chem. A broad spectrum of severe accident phenomena in both boiling istry; product form and microstructure; irradiation fluence, flux, and presssrized water reactors is treated in MELCOR in a uni-temperature, and spectrum, and postirradiation annealing are fied framework. These include thermal-hydraulic response in the being examined on a wide range of fracture properties. The reactor coolant system, reactor cavity, containment, and confine. HSSi Program is arranged into seven tasks: (1) program man- ment buildings; core heatup, degradation, and relocation; core- l agement (2) irradiation effects in engineering materials (3) an- concrete attack; hydrogen production, transport, and combus-nealing, (4) microstructural analysis of radiation effects, (5) in- tion; fission product release and transport behavior. Current service irradiated and aged material evaluations, (6) fracture uses of MELCOR include estimation of severe accident source toughness curve shift method, (7) special technical assistance, terms and their sensitivities and uncertainties in a variety of ap-and (8) foreign research interactions. The work is performed by plications. This publication of the MELCOR computer code the Oak Ridge National Laboratory. manuals corresponds to MELCOR 1.8.4, released to users in July 1997. Volume 1 contains a primer that describes NUREGICR-6621: GROUND-WATER MODELS IN SUPPORT OF MELCOR's phenomenological scope, organization (by package), l NUREG/CR-5512. COLE,C.R.; WILLIAMS,M.D.; PERKINS.W.A.; and documentation. The remainder of Volume 1 contains the et d. Battelle Mernorial institute, Pacific Northwest National Lab- MELCOR User's Guides, which provide the input instructions oratory. Decernber 1998. 218pp. 9812220079. A8303:059. and guidelines for each package. Volume 2 contains the This report describes an investigation of the modeling rneth- MELCOR Reference Manuals, which describe the phenomeno-odology laid out in NUREG/CR-5512, Volume 1, for determining logical models that have teen implemented in each package. site-specific radiological criteria for license termination. Four NUREG/CR-6119 V02 R1: MELCOR COMPUTER CODE MANU-computer codes wer6 developed; the standard model, an imple. ALS. Reference Manuals, Version 1.8.4, July 1997. mentation of the three compartment model of NUREG/CR-5512, GAUNTT.R.O.; COLE,R.K. Sandia National Laboratories. Volume 1; an extended rnodel that replaces the single-compart. HODGE,S.A.; et al. Oak Ridge National Laboratory. May 1998. rnent vadose zone of the standard model with multiple compart- 800pp. 9806100437. SAND 97-2398. A3723:001. ments; and two hybrid codes constructed by replacing either the See NUREG/CR-6119,V01,R01 abstract. vadose zone or the aquifer compartment in the standart.; or ex-tended NUREG/CR-5512 code with a more realistic numerical NUREG/CR-6131: VICTORIA 2.0; A MECHANISTIC MODEL FOR code, STOMP for the vadose zone and CFEST for the aquifer. RADIONUCLIDE BEHAVIOR IN A NUCLEAR REACTOR Sensitivity analyses were used to determine which parameters COOLANT SYSTEM UNDER SEVERE ACCIDENT CONDI-must be "hard wired" to assure prudent conservatism and which TIONS. BIXLER,N.E. Sandia National Laboratories. December can be varied with site data to extend the screening range. 1998. 300pp. 9901270217. SAND 93-2301. A6621:335. Comparisons between the models demonstrated the rea!!stic VICTORIA is a computer code that is intended to rnechanis-conditions represented by the standard and extended models toally treat fission product behavior and related phenomena in when using generic parameter values. The results also indicated the reactor coolant system of commercial light water reactors, that the conditions under which the NUREG/CR-5512 screening Processes that are treated by the VICTORIA rnodels include the approach is not appropriate are quite simple. The NUREG/CR-release d Ossim products and nonradioa@e materials from 5512 methodology is not appropriate if there already exists deep fuel rods and degra@d forms d fwl, the release d control rod elements and other structural materials, chemical interactions contamination at the site, if there are signifcant risks associated with contamination at the site (for example it overlies a sole and change of phase, aerosol formation and interactions, trans-source aquifer), or if the basic conceptual model and scenarios port W deposhn of aerosols and vapors, resuspenson d de-posited aerosols, and decay heating effects due to transport and that are part of the NUREG/CR-5512 rnethodology are inappro-

                                                                       &posen of hssim pro & cts. Emed kom Ws hst are mermal priate (e.g., karst terrain, located in a flood plain, highly frac-  hydraules and rnost aspects of fuel degradation and relocation.

tured bed rock aYstem)~ lt is possible to decouple these aspects of accident analysis NUREG/CR-5671: PRETEST PREDICTION ANALYSIS AND from fission product behavior and transport because the cou-POSTTEST CORRELATION OF THE SIZEWELL-B 1:10 phng between these phenomena is primarily unidirectional, i.e., SCALE PRESTRESSED CONCRETE CONTAINMENT MODEL thermal hydraulics influence fission product behavior, but not TEST. DAMERON,R.A.; RASHID,Y.R.: SULLAWAY,M.F. vice versa. VICTORIA provides the user with a detailed mass ANATECH Research Corp. July 1998.180pp. 9808100114. balance for each of the 26 elements that it treats it also pro. SAND 90-7117. A4547:142. vides information on chemical speciation of vapors and con-1 This report describes pretest predictions and posttest correla- densed-phase materials. Finally, it provides a detailed descrip-tions of a one-tenth-scale model of the "Sizewell-B" prestressed tion of the aerosols that remain suspended, of the cuantities that concrete contairvnent building. The analyses were performed by deposit and of the mechanisms that lead to deposition.

20 Main Citation 3 and Abstracts NUREG/CR-6150 V01 R1: SCDAP/RELAPS/ MOD 32 CODE NUREGICR-6150 V04 R1: SCDAP/RELAPS/ MOD 3.2 CODE MANUALinterface Theory.

• Idaho National Engineering & Envi- MANUAL.MATPRO - A Library Of Materials Properties For ronmental Laboratory. July 1998. 70pp. 9807280099. INEL-96/ Light-Water Reactor Accident Analysis.
• Idaho National Engi-0422. A4397:208. neering & Environmental Laboratory. July 1998. 600pp.

The SCDAP/RELAPS code has been developed for best+sti. 9807280111. INEL-96/0422. A4394:001. mate transient simulation of light water reactor coolant systems The SCDAP/RELAP5 code has been developed for best-esti-during a severe accident. The code models the coupled behav. mate transient simulation of light water reactor coolant systems lor of the reactor coolant system core, fission product released during a severe accident. The code models the coupled behav-during a severe accident transient as well as large and smalg ior of the reactor coolant system core, fission product released break loss-of-coolant accidents, operational transients such as during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled used that pemits as much of a partiwlar system to be modMed as necessary. Control system and secondary system compo-as necessary. Control system and secondary system compo-nents are included to permit modeling of plant controls, turbines, nents are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the inter. Volume 4 describes the material properties correlations and face between severe accident rnodels which are resident in the computer subroutines (MATPRO) used by SCDAP/RELAPS. SCDAP portion of the code and hydrodynamic models which are Formulation of the materials properties are generally resident in the RELAP5 portion of the code. A description of the organization and structure of SCDAP/RELAPS is presented. Ad-spe5M M nahe. De mateMs property sNnes con-tained in this document are for uranium, uranium dioxide, mixed ditional information is provided regarding the manner in which uraniun> plutonium, dioxide fuel, Zircaloy cladding, zirconium di-models in one portion of the code impact other parts of the oxide, stainless steel, stainless steel oxide, silver-indiurrH:ad-code, and models which are dependent on and derive informa- mium alloy, cadmium, boron carbide, inconel 718, Zirconium-tion kom othw subcodes. uranium-oxyDen melts, fill gas mixtures, carbon steel and tung-ston. This document also contains descriptions of the reaction NUREGICR-6150 V02 R1: SCDAP/RELAPS! MOD 3.2 CODE MAN- and solu

• n rate models needed to analyze a reactor accident.

UAL. Damage Progression Model Theory.

• Idaho National Engi- Revision 1 includes changes incorporated since the manuals neering & Environmental Laboratory. July 1998. 216pp. were released in June 1995.

9807280103. INEL-96/0422. A4397:001. NUREG/CR-6150 V05 R1: SCDAP/RELAP5/ MOD 3.2 CODE The SCDAP/RELAP5 code has been developed for best-esti- MANUAL. Developmental Assessment.

• Idaho National Engi-mate trans6ent simulation of light water reactor coolant systems neering & Environmental Laboratory. July 1998. 250pp.

during a severe accident. The code models the coupled behav- 9807280113. INEL-96/0422. A4393:040. ior of the reactor coolant system cue, fission product released The SCDAP/RELAPS code has been developed for best-estL during a severe accident transient as well as large and small mate transient simulation of light water reactor coolant systems break loss-of-coolant accidents, operational transients such as during a severe accident. The code models the coupled behav-anticipated transient without SCRAM, loss of offsite power, loss for of the reactor coolant system core, fission product released of feedwater, and loss of fiow. A generic modeling approach is during a severe acc. dent transient as well as large and small used that permits as much of a particular system to be modeled break loss-of-coolant accidents, operational transients such as as necessary. Control system and secondary system compo- anticipated transient without SCRAM, loss of offsite power, loss nents are included to permit rnodeling of plant controls, turbines. of feedwater, and loss of flow. A generic modeling approach is condensers, and secondary feedwater conditioning systems, used that permits as much of a particular systern to be modeled This volume contains detailed descriptions of the severe acci- as necessary. Control system and secondary system compo-dent models and correlations. It provides the user with the un- nents are included to permit modeling of plant controls, turbines, derlying assumptions and simplifications used to generate and condensers, and secondary feedwater conditioning systerns. irnplement the basic equations into the code, so an intelligent This volume contains detailed code-to#*a calculations per-assessment of the applicability and accuracy of the resulting cal- formed with SCDAP/RELAPS/ MOD 3.2 as ell as comparison culation can be made. calculations performed with the earlier code version, SCDAP/ RELAP5/ MOD 3.1. In addition to the comparison of code calcula-NUREG/CR-6150 V03 R1: SCDAP/RELAP5/ MOD 32 CODE tions to experimental data, comparisons of specific SCDAP MANUAL. User's Guide And input Manual.

• Idaho National En- models to stand-alone calculations of rnotten pool behavior and gineering & Environmental Laboratory. July 1998. 300pp. crust failures are also described. The results of nodalization and 9807280108. INEL-96/0422. A4396:001. time step studies are also reported.

The SCDAP/RELAPS code has been developed for best-esth NUREG/CR-6210 S01: COMPUTER CODES FOR EVALUATION mate transient simulation of light water reactor coolant systems OF CONTROL ROOM HABITABILITY (HABIT V1.1). during a severe accident. The code models the coupled behav- RAMSDELL,J.V. Battelle Memorial Institute, Pacific Northwest ior of the reactor coolant system core, fission product released National Laboratory. STAGE,S.A. Affiliation Not Assigned. Octo-during a severe accident transient as well as large and small ber 1998.133pp. 9811130266. PNNL-10496. A5867:006. break loss-of-coolant accidents, operational transients such as Over the years the Nuclear Regulatory Commission has de-anticipated transient without SCRAM, loss of offsite power, loss veloped computational tools for use in evaluation of control of feedwater, and loss of flow. A Deneric rnodeling approach is room habitability in the event of postulated accidents at nuclear used that permits as much of a particular system to be modeled reactors. Five of these tools, which were developed individua!!y, as necessary. Control system and secondary system compc- were combined in a single Control Room Habitability software nents are included to permit modeling of plant controls, turbines, packa;e called HABIT. Advances in the understanding of the re-condensers, and secondary feedwater conditioning systems. actor source terms and changes in regulations have expanded This volume provides guidelines to code users based upon les- the arena in which two of the codes, TACTS and CONHAB, may sons teamed during the developmental assessment process. A be applied. Consequernly, these codes and the HABIT user description of problem control and the installation process is in. interface were modified to simplify control room habitability as-cluded. Appendix A contains the description of the input require- sessments in the current regulatory environment. This report de- l ments, scribes the changes made to HABIT, TACTS, and CONHAB. It l I i t l l

Main Citations and Abstracts 21 supplements, but does not replace, the original desenption of to common cause failure events at commercial nuclear power HABIT. The changes include modifcations (1) to permit calcula- plants in the U.S. are identified during Licensee Event Report ik n of doses using updated source term insights from NUREG- (LER) and Nuclear Plant Reliability Data System (NPRDS) fail-1465, " Accident Source Terms for Ught-Water Nuclear Power ure report reviews. Once equipment failures that contrbute to a Plants," (2) to permit calculation of total effective dose equiva- common cause failure event are identified, the common cause lent (TEDE) and maximum two-hour dose discussed in recent failure events are coded for entry into a personal computer stor-revisions to 10 CFR Parts 50 and 100, and (3) to limit routine age system using the method presented in this volume. The output to doses of interest. In addition, new radionuclide data database resulting from coding common cause failure events is files were prepared to permit dose calculations using dose fac- used to estimate common cause failure parameters for use in tors from Federal Guidance Reports 11 and 12. various probabilistic risk assessment (PRA) CCF rnodels. NUREG/CR4268 Voit COMMON CAUSE FAILURE DATABASE NUREG/CR4268 V04: COMMON CAUSE FAILURE DATABASE AND ANALYSIS SYSTEM. Volume 1: Overview. MAR- AND ANALYSIS SYSTEM. Volume 4: Software Reference Man. SHALL,F.M. Idaho National Engineering & Errvironmental Lab. ual. MARSHALL,F.M.; KVARFORDT,K.J.; CEBULL,M.J.; et at oratory. MOSLEH,A. Maryland, Univ. of College Park, MD. Idaho National Engineering & Environmental Laboratory. June RASMUSON D.M. Division of Safety Programs (Post 870413). 1998.96pp.9807270430. INEELEXT9700696. A4387:244. June 1998. 37pp. 9807270385. INEELEXT9700696. A4381:314. This volume of the Common-Cause Failure Database and This volume of the Common Cause Failure Database and Analysis System report provdes an overview of common cause Analysis System report presents an overview of common cause failure methods for use in the U.S. commercial nuclear power failure methods for use in the U.S. commercial nuclear power industry. It summarizes how data (on common cause failure industry, it summarizes how data (on common cause failure events) are gathered, evaluated, and coded. It then describes events) are 9athered, evaluated, and coded. It then describes the process for estimating probabilistic risk assessment (PRA) the process for estimating probabilistic risk assessment (PRA) common cause failure pararneters. It also references other vol-common cause failure parameters. It also references other vol- umes of this report for specific details. Equipment failures that umes of this report for specif'c details. Equipment failures that contribute to common cause failure events are identsfied through contribute to common cause failure events are identified through searches of Ucensee Event Reports (LERs) and Nuclear Plant searches of Ucensee Event Reports (LERs) and Nuclear Plant Reliability Data System (NPRDS) failure reports. Once common Reliability Data System (NPRDS) failure reports. Once common cause failure events are identified by reviewing reports of equip. cause failure events are identified by reviewing reports of equip- ment failures, INEEL staff enters the event information into a ment failures, INEEL staff enter the event information into a per- personal cornputer data analysis system (CCF system) usitig sonal computer data analysis system (CCF system) using the the rnethod presented in this and companion volumes. The method presented in this and companion volumes. The events events stored in the CCF system are utilized for common cause stored in the CCF system are used for common cause failure faMrs PRA parameter estimations using common cause failure PRA parameter estimations using common cause failure quan- quantification methods. Scation methods. NUREG/CR4274: PALEOSEISMIC STUDIES IN THE SOUTH-NUREGICR4268 V02: COMMON CAUSE FAILURE DATABASE EASTERN UNITED STATES AND NEW ENGLAND. AND ANALYSIS SYSTEM. Volume 2: Event Definition And Clas- GELINAS.R.; CATO,K.; AMICK.D.; et al. , August 1998.475pp. sification, MARSHALL,F.M. Idaho National Engineering & Envh 9809100199. A5009:001. ronmental Laboratory. MOSLEH.A. Maryland, Univ. of, College A mul0 faceted approach toward palooseismic studies was Park, MD. RASMUSON,D.M. Division of Safety Programs (Post conducted in selected inland areas of the southeastem United 870413). June 1998. 57pp. 9807270414. INEELEXT9700696. States and New England. The purpose of these investigations A4380 293. was to provide information on the spatial and temporal distribu. This volume of the Common Cause Failure Database and tion of large prehistoric earthquakes in areas that have had low Analysis System report provides the definition and classification to moderate levels of seismic actMty during historical times. method used for identifying, coding, and quantifying common Some of the varied neotectonc techniques employed during the cause failure (CCF) probabiliste risk assessment (PRA) param- studies included extensive field mapping and trenching of poten-eter estimates on the data that are stored in the CCF database. tially liquefiable sediments and soft sediment deformation fea-Equipment failures that contribute to CCF events at commercial tures, evaluation of geophysical data for earthquake-induced nuclear power plants in the U.S. will be identified during a ground failure features both on-land and in lake-bottom sedi-search and review of Licensee Event Report (LER) and Nuclear ments, evaluation of giant block landslides for a seismic trig. Plant Reliability Data System (NPRDS) failure reports in accord- gering mechanism, and investigation of whether or not struc-ance with the criteria specified in this document. The equipment tures in caves could be used as indicators of past seismic activ-failures that contribute to a CCF event are identified and coded ity. for entry into a personal computer storage system using the method presented in this volume. The events stored in the sys- NUREG/CR4359 V01: RAMONA-4B: A COMPUTER CODE ter' are used to perform CCF PRA parameter estirnations using WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR CCF quantification methods (also described in this volume). The AND SBWR SYSTEM TRANSIENTS.Models And Correlations. ROHATGI,U.S.; CHENG,H.S.; KHAN,H.J.; et al. Brookhaven database resulting from coding CCF events is used to estimate C failure parameters for use in various CCF rnodels used in National Laboratory. March 1998. 453pp. 9803190143. BNL-NUREG-52471. A2628:001. Ramona-4B is a systems transient code for application to dif-NUREGICR4268 V03: COMMON CAUSE FAILURE DATABASE ferent versions of Bolling Water (BWR) such as the current AND ANALYSIS SYSTEM. Volume 3: Data Collection And Event BWR, the Advanced Boiling Water Reactor (ABWR), and the Coding. MARSHALL,F.M. Idaho National Engineering & Environ- Simplified Boiling Water Reactor (SBWR). This code uses a mental Laboratory. MOSLEH,A. Maryland, Univ. of, College three-dimensional neutron kinetics model coupled with a multh Park, MD. RASMUSON,D.M. Division of Safety Programs (Post channel nonequilibnum, drift-flux, two-phase flow formulation of 870413). June 1998. 66pp. 9807270421. INEELEXT9700696. the thermal hydraulics of the reactor vessel. The code is de-A4385 293. signed to analyze a wide spectrum of BWR core and system This volume of the Common Cause Failure Database and transients and instability issues. Chapter 1 is an overview of the Analysis System report documents the method used for coding code's capability and limitations; Chapter 2 discusses the neu-common cause failure (CCF) events that are stored in the com- tron kinetics modeling and the implementation of reactivity edits. mon cause failure database. Equipment failures bat contribute Chapter 3 is an overview of the heat conduction calculations.

22 Main Citations and Abstracts Chapter 4 presents modifications to the thermal hydraulics chemistry of the vauft disposal system may have on radionuchde model of the vessel, recirculation loop, steam separators, boron release. The geochemistry of pore waters buffered by cementi-transport and SBWR specific components. Chapter 5 describes tious materials in the disposal system will be different from the rnodehng of the plant control and safety systems. Chapter 6 pre- local ground water. Therefore, the cement-buffered environment sents the modeling of Balance of Plant (BOP). Chapter 7 de- needs to be considered within the source term calculations if scribes the mechanistic containment model in the code. The credit is taken for solubility limits and/or sorption of dissolved content of this report is complamentary to the RAMONA-3B radionuchdes within disposal units. A literature review was cord code description and assessment document. ducted on methods to model pore-water compositions resulting fr m reactions with cement, experimental studies of cement / NUREG/CR-6359 V02: RAMONA-4B: A COMPUTER CODE wats systems, natural analogue studes of cement and con-WITH THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR crete, and radionuclide solubihties experimentally determined in AND SBWR SYSTEM TRANSIENTS. User's Manual ceneM pore waters. Based on Ws review, geochental mod-ROHATGl,U.S.; CHENG,H.S.; KHAN,H.J.; et al. Brookhaven National Laboratory. March 1998. 350pp. 9803190151. BNL- 8. ling was used to calculate maximum concentrations for ameri-cium, neptunium, nickel, plutonium, radium, strontium, thorium, NUREG-52471. A2630:106. and uramm fw pwe-water composWions budemd by cement This document is the User's Manual for the Boiling Water Re- and local grou# water. Anothw Mwatur actor (BWR), and Simplified Boiling Water Reactor (SBWR) sys- on radionuclide sorption behavior onto

                                                                                                                  , fresh'    ,o review cement / concretewas compj tems transient code RAMONA-4B. The code uses a three-di-             where N pore water pH wm be 210. Based on Ws reWew, mensional neutron-kinetics model coupled with a rnultichannel        a database was developed of preferred minimum distribution co-nonequilibrium, drift-flux, two-phase flow rnodel of the thermal                                                                                   l efficient (K(d)) values for these radionuclides in cement / concrete hydraulics of the reactor pressure vessel. The cods is designed     enuments.                                                                      I to analyze a wide spectrum of BWR and SBWR core and syrr tem transients. Chapter 1 gives an overview of the code's capa-     NUREG/CR-6408: TECHNICAL ASSISTANCE IN REVIEW OF bilities and limitations. Chapter 2 describes the code's structure,  SOURCE TERM-RELATED ISSUES OF ADVANCED REAC-lists major subroutines, and discusses the computer require-        TORS. BEAHM.E.C.; WEBER C.F.; DlLLOW T.A. Oak Ridge ments. Chapter 3 provides the instructions for installing and run-   National Laboratory. October 1998. 28pp. 9811050323. ORNU ning the RAMONA-48 code on sun SPARC and IBM                         TM-13144. A5695:073.

workstations. Chapter 4 contains component descriptions and The distribution of iodine in containment during an AP-600 de-detailed car &by-card input instructions. Chapter 5 gives sam- sign-basis accident was evaluated using rnodels in the pies of the tabulated output for the steady-state and transient " TRENDS" code. The AP-600 3BE accident saquence calcula-calculations and discusses the plotting procedures for the tions showed that a pH greater than 7 was rnaintained for at steady-state and transient results. Three appendices contain im. least 30 days. Because the pH was rnaintained at this level, portant user and programmer information: lists of plot variables most of the iodine was in the form of iodide; only 3 x 10(-3) % (Appendix A), listings of input deck for sample problem (Appen- was present as aqueous f(2), and only 1 x 10(-6) % was present dix B), and a desenption of the plotting program PAD (Appendix as 1(2) in the vapor phase. C). NUREG/CR-6410: NUCLEAR FUEL CYCLE FACILITY ACCl-NUREG/CR-6364: HUMAN PERFORMANCE IN RADIOLOGICAL DENT ANALYSIS HANDBOOK.

• Science Applica-SURVEY SCANNING. BROWN,W.S. Brookhaven National Lab- tionsinternational Corp. (forrnerly Science Applications, Inc.).

oratory. ABELOUIST.E.W. Oak Ridge Associated Universities. March 1998. 637pp. 9804060094. A2880:001. March 1998. 54pp. 9803180087. BNL-NUREG-52474. The purpose of this Handbook is to provide guidance on how A2609:303, to calculate the characteristics of releases of radioactive mate-The probability of detecting residual contamhation in the field rials and/or hazardous chemicals from nonreactor nuclear facili-using portable radiological survey instruments depends not only ties, in addition, the Handbook provides guidance on how to cah on the sensitivity of the instrumentatson used in scanning, but culate the consequences of those releases. There are four also on the surveyor's performance. This report provides a basis major chapters: Hazard Evaluation and Scenario Development; for taking human perforrnance into account in determining of the Source Term Determination; Transport Within Containment /CorF minimum level of activity detectable by scanning. A theoretical finement; and Atmospheric Dispersion and Consequence Mod-framework was developed (based on signal detection theory) eling. These chapters are supported by Appendices, including: which allows influences on surveyors to be antscipated and un- a summary of chemical and nuclear information that contains derstood, and supports a quantitative assessment of perform- descriptions of various fuel cycle facilities; details on how to cah ance. The performance of surveyors under controlled yet reah culate the characteristics of source terms for releases of haz-istic field conditions was examined to gain insight into the task ardous chemicals; a comparison on NRC, EPA, and OSHA pro-and to develop rneans of quantifying performance. Then, their grams that address chemical safety; a summary of the perform-performance was assessed under laboratory conditions to quan- ance of HEPA and other filters; and a discussion of uncertain-tify more precisely their ability to make the required discrimina- ties. Several sample problems are presented: a free-fall spill of tions. The information was used to characterize sunreyors' per- powder; an explosion with radioactive releases; a fire with radio-formance in the scanning task and to provide a basis for pre- active releases; filter failure; hydrogen fluoride release from a dicting levels of radioactivity that are likely to be detectable tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in under various conditions by surveyors using portable survey in- a vitrification plant; and a criticahty incident. Finally, this Hand-struments. book includes a computer model, LPF#1b, that is intended for use in calculating leakpath factors. NUREG/CR-6377: EFFECTS ON RADIONUCLIDE CONCENTRA-TlONS BY CEMENT / GROUND-WATER INTERACTIONS IN NUREG/CR-6412: AGING AND LOSS-OF-COOLANT ACCIDENT SUPPORT OF PERFORMANCE ASSESSMENT OF LOW- (LOCA) TESTING OF ELECTRICAL CONNECTIONS. NEL-LEVEL RADIOACTIVE WASTE DISPOSAL FACILITIES. SON,C.F. Sandia National Laboratories. January 1998.109pp. KRUPKA,K.M.; SERNE,R.J. Battelle Memorial Institute, Pacific 9803180097. SAND 97-3170. A2611:196. Northwest National Laboratory. May 1998.154pp. 9806010310. This report presents the results of an experimental program PNNL-11408. A3572:292. to determine the aging and loss-of-coolant accident (LOCA) be-The U.S. Nuclear Regulatory Commission is developing a havior of electrical connections in order to obtain an initial technical position document that provides guidance regarding scoping of their performance. Ten types of connections com-the performance assessment of low-level radioactive waste dis- monly used in nuclear power plants were tested. These included I posal facilities. This guidance considers the effects that the 3 types of conduit seals,2 types of cable-tcWevice connectors,

Main Citation 3 and Abstracts 23 3 types of cable-tx.able connectors, and 2 types of in-line NUREG'CR-6471 V01; CHARACTERIZATION OF FLAWS IN U.S. splices. The connections were aged for 6 months under simulta- REACTOR PRESSURE VESSELS. Density And Distribution Of neous thermal (99 degrees C) and radiation (46 Gy/hr) condi- Flaw Indications in PVRUF, SCHUSTER,G.J.; DOCTOR,S.R.; tions. A simulated LOCA consisting of sequential high dose-rate HEASLER,P.G. Battelle Memorial institute Pacife Northwest irradiation (3 kGy/hr) and high-temperature steam exposures fol- National Laboratory. November 1998. 450pp. 9901270225. Iowed the aging. Connection functionality was monitored using PNNL-11143. A6619:001. Insulation resistance measurements during the aging and LOCA Characterization of Flaws in U.S. Reactor Pressure Vessels is cxposures. Because only 5 of the 10 connection types passed a multi-volume report. Volume 1, this document, provides the re-O post-LOCA, submerged dielectric withstand test, further de. sults of a nondestructive examination conducted at the Oak tailed investigation of electrical connections and the effects of Ridge National Laboratory's Pressure Vessel Research User Fa-cable jacket integrity on the cable connection system is war- cility (PVRUF) on a vessel fabricated for a canceled nuclear j ranted. power plant. Volume 2, in preparation, wd,l document the results I of Pacific Northwest National Laboratory's (PNNL) destructive validation of the flaw rates in the PVRUF evaluation, Twenty lirw NUREG/CR-6418: RISK IMPGCTANCE OF CONTAINMENT AND ear meters of weldment were inspected by SAFT-UT including RELATED ESF SYSTEM PERFORMANCE REQUIREMENTS. the entire circumferential beltline weld of the vessel. There were NOURBAKHSH,H.P.; HANSON,A.L.; PRATT,W.T. Brookhaven 2500 detectable indications in the SAFT-UT inspections of the National Laboratory. November 1998.105pp. 9811230224. BNL- PVRUF vessel, The largest number of these,982, were found NUREG-52489. A5931243. at the clad-to-base metal interface, but 978 of these were less A study has been performed determining the potential than 2mm (0.08 in.) in size, in the near surface zone, the weld ) changes in plant risk as a result of implementing revised source metal contained 98 detectable planar indcations. The density of I terms in nuclear power plants that could result in changes in indications was four times higher in the weldment than in the i plant design. The effect of revising the containment and related base metal The distribution of the empirical data provided ' ESF system's performance requirements on overall nuclear re- enough information to apply a parametric model of the cumu-actor accident risk was investigated. This study takes into ac- lative flaw rate to six different subsets of the data. and to obtain 1 count recent insights into severe accident phenomena and ex- reasonable confidence bounds on the resutts. Recommenda-amines consequerces in a risk based format consistent with the tions are given for validating the indication rates by selective de-quantitative health objectives (OHOs) of the NRC Safety Goat structive analysis to provide the necessary high quality flaw sta-PoHey. tistics for use in fracture mechanics calculations such as those used in pressurized thermal shock (PTS) analysis. NUREG/CR-6447: RESULTS OF CRACK-ARREST TESTS ON NUREG/CR-6472: PRELIMINARY PHENOMENA IDENTIFICA-IRRADIATED A 508 CLASS 3 STEEL ISKANDER,S.K.; TlON AND RANKING TABLES FOR SIMPLIFIED BOILING MILELLA,P.P.; PlNI.A. Oak Ridge National Laboratory. February WATER REACTOR LOSS-OF-COOLANT ACCIDENT SCE-1998. 97pp. 9802270188. ORNL-6894. A2342:122. NARIOS. KROGER,P.G.; ROHATGI,U.S.; JO J.H.; et al. Crack-arrest specimens of irradiated A 508 class 3 forging Brookhaven National Laboratory. April 1998 162pp. steel were tested and evaluated according to the Arnerican So. 9805050390. BNL-NUREG-52501. A3317:239. ciety for Testing and Materials Standard Test Method for Deter. A set of Phenomena identifcation and Ranking Tables (PIRT) mining Plain-Strain Crack-Arrest Fracture Toughness, K(la), of for three potential Loss-of-Coolant Accident (LOCA) scenanos in Ferritic Steels, E 1221-88. The irradiation-induced shifts while the General Electric Simplified Boiling Water Reactor is pre-small, averaging only about 10 K, are approximately the same sented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Anal-as the Charpy 41-J temperature shifts. The specimens were ir- , radiated at temperatures ranging from 243 to 280 degrees C to ysis Reports. The method used to develop the PIRTs is de-scribed. Following a discussion of the transient scenarios, the fluences varying from 1.7 to 2.7 x 10(19) neutrons /cm(2)(>l PIRTs are presented and discussed in detailed and summarized y,y)' form. A procedure for future validation of the PIRTs. to enhance NUREG/CR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCHMARK. REMEC,1.; KAM.F.B. Oak Rdge National Lab. NUREG/CR4475: RESOLUTION OF THE DIRECT CONTAIN-MENT HEATING ISSUE FOR COMBUSTION ENGINEERING oratory. February 1998. 58pp. 9803050078. ORN11TM-13204. PLANTS & BABCOCK & WILCOX PLANTS. PILCH,M.M.; A2424:056 ALLEN,M.D.; POWELL J.L; et al. Sandia National Laboratones. The HBFI-2 benchmark is specified and analyzed in this re-port. Analysis of the HBR-2 benchmark can be used as partial W ^" 00 fulfillment of the requirements for the qualifcation of the method- This report applies the DCH assessment methodology devel-ology for calculating neut on fluence in pressure vessels, as re- oped in NUREG/CR-6075 and NUREG/CR-6075, Supplement 1, quired by the U.S. Nuclear Regulatory Commission Regulatory to all Combustion Engineering (CE) and Babcock & Wilcox Guide DG-1053, " Calculational and Dosametry Methods for De- (B&W) PWRs. Using screening-level load / strength analyses, the termining Pressure Vessel Neutron Fluence." Secten 1 of this CCFP is less than 0.01 for the following plants: Fort Calhoun report provides all the dimensions, material compositions, and (CE), Palo Verde (CE), St. Lucie (CE), Waterford (CE), ANO-1 neutron source data necessary for the anatysis. The measured (B&W), Crystal River (B&W), Oconee (B&W), and Three Mile Is-quantities, to be cornpared with the calculated values, are the land (B&W). DCH is considered resolved for these plants based specific activities of the neutron dosimeters, on both sides of the on loadstrength comparisons alone and no additional anatyses pressure vessel: in the surveillance capsule attached to the ther. are required. The following plants have CCFPs between 0.01 rnal shield and in the reactor cavity. Section 2 describes the and 0.1 based on the screening-level load / strength analyses:

analysis of the HBR-2 benchmark with the computer code ANO-2 (CE), Millstone 2 (CE), Palisades (CE), San Onofre DORT and three ENDF/B-VI based multigroup libraries. The av- (CE), and Davis Besse (B&W). These plants fall the screening l erage ratio of the calculated-to-measured specifc activities (C/ criterion established by the rnethodology. These plants were re-M) for the six dosimeters in the surveillance capsule was 0.90 solved by factoring in the low likelihood of high system pres-10.04 for all three libraries. The average C/Ms for the cavity sures at vessel breach. Two plants, Calvert Cliffs (CE) and dosimeters (without neptunium dosimeter) were 0.8910.10, Maine Yankee (CE), have CCFPs (based only on load' strength 0.91 1 0.10, and 0.90 1 0.09 for the BUGLE-93, SAILOR-95, comparisons) greater than the 0.1 success criterlon of the meth-and BUGLE-96 libraries, respectively, odology (Catvert Cliffs CCFP.O.149; Main Yankee

24 Main Citation 3 and Abstracts CCFP-0.211). DCH load distributions used by Calvert Cliffs and the spacing of the orifice plates is one tube diameter. A stand-Maine Yankee in their IPEs are higher than the loads distribu- ard automobile diesel engine glow plug was used to ignite the tions calculated for this study. Hence, it is also reasonable to test mixture at one end of the tube. Hydrogen-air-steam mix-conclude that DCH is resolved from this integrated perspective tures were tested at a range of temperatures up to 650K and assuming that IPE treatment of DCH in its event tree is com- at an initial pressure of 0.1 MPa. It was also observed that the prehensive and valid. distance required for the flame to accelerate to the point of deto-nation initiation, referred to as the run-up distance, was found NUREG/CR-6479: TECHNICAL BASIS FOR ENVIRONMENTAL a bndon of M N Wyn nWe #achon and tM h QUALIFICATION OF MICROPROCESSOR-BASED SAFETY- * '***

• RELATED EQUIPMENT IN NUCLEAR POWER PLANTS- "' " # * " #"* * "

or increasing the in4ial rntxture temperature resulted in longer KORSAH,K. Oak Ridge National Laboratory. HASSAN,M. run-up distances. The density ratio across the flame and the Brookhaven Nahonal Laboratory. TANAKA.T.J.; et al Sandia speed of sound in the unburned mixture were found to be two National Laboratories. January 1998. 128pp. 9803180022. parameters which influence the run-up distance. ORNUTM-13264. A2610:197. This document presents the results of studies sponsored by NUREG/CR-6511 V02: STEAM GENERATOR TUBE INTEGRITY the Nuclear Regulatory Commission (NRC) to provide the tech- PROGRAM. Annual Report, August 1995 September 1996. nical basis for environmental qualification of computer 4ased DIERCKS.D.R.; BAKHTIARl,S.; KASZA,K.E.; et al. Argonne Na-safety equipment in nuclear power plants. The studies were tional Labomory. February 1998.193pp. 9803180026. ANL-97/ conducted by Oak Ridge National Laboratory (ORNL), Sandia 3. A2611:001. National Laboratories (SNL), and Brookhaven National Labora. This report summarizes work performed by Argonne Nato.nal tory (BNL). The studies address the following: (1) adequacy of Laboratory on the Steam Generator Tube integrity Program from the present test methods for qualification of digital l&C systems; the inception of the program in August 1995 through September (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) 1996. The program is divided into five tasks: (1) Assessment of rn.winis,Cmd stressors to be included in the qualification proc- Mspechon Reliability, (2) Research on ISI (inservice-inspection) ess during type testing: (4) resolution of need for accelerated Technology, (3) Research on Degradation Modes and integrity, aging for equipment to be located in a benign environment; and (4) Tube Removals from Steam Generators, and (5) Program (5) determination of an appropriate approach for addressing the Management. Mr Task 1, mess is mM m h prepa-impact of smoke in digital equipment qualification programs. Sig- ration of facilites and evaluation of nondestruchve evaluation nificant findings from the studies form the technical basis for a tecMques fw Mspechng a mock-up steam generatw for rod recommended approach to the environmental qualification of robin testing, the development of better ways to correlate failure microprocessor-based safety-related equipment in nuclear Pressure and leak rate with eddy current (EC) signals, the n-power plants' spection of sleeved tubes, workshop and training scuvities, and the evaluation of emerging NDE technology. Results are re. NUREG/CR4502: ACTION REQUIREMENTS FOR AFW SYS- ported in Task 2 on closed-form solutions and finite-element TEM FAILURES.An Analysis For Four Nuclear Power Plants. electromagnetic modeling of EC probe responses for various MANKAMO,T, Finland, Govt. of. KIM,LS.; SAMANTA,P.K.; et al. probe designs and flaw characteristics. In Task 3, facilities are Brookhaven Nabonal Laboratory. July 1998. 180pp. being designed and built for the production of cracked tubes 9807280073. BNL-NUREG-52513. A4398:001. under aggressive and near-prototypical conditions and for the The current Technical Specifications for the auxiliary testing of flawed and unflawed tubes under normal operating, feedwater (AFW) systems of pressurized water reactors (PWRs) accident, and severe-accident conditions. Crack behavior and typically require that the plant is shut down irnmediately when stability are also being modeled to provide guidance for test fa-there are multiple failures in the AFW pump-trains, l.e.,2- or 3- cility desip1, develop an improved understanding of the ex-train failures in the typically 3-train system (in some cases, im- pected rupture behavior of tubes with circumferential cracks, mediate repair of one train is requireo when all AFW trains fail). and predict the behavior of flawed and unflawed tubes under se-Because the AFW system is used to remove decay heat during vere accident conditions. Task 4 is concerned with the acquisi-shutdown in PWRs, shutting down when the availability of this tion of tubes and tube sections from retired steam generators for system is degraded may incur a considerable risk, which may use in the other research tasks. Progress on the acqubition of even surpass the risk of continuing power operation giving pri- tubes from the Salem and McGuire I nuclear plants is reported. ority to the repairs. In this study, we developed a simplified method to analyze the nsks of continued power operation and NUREG/CR4511 V03: STEAM GENERATOR TUBE INTEGRITY of plant shutdown associated with the technical specifications PROGRAM. Semiannual Report, October 1996 - March 1997, action requirements for failures of stan@y safety systems during DIERCKS,D.R.; BAKHTIARI.S.; KASZA K.E.; et al. Argonne Na-power operation. This method was applied to various failures in tional Laboratory. August 1998.124pp. 9808310079. ANL-9F/7. A4851:130 the AFW systems of four PWRs, San Onofre 3. Sequoyah 1 system 80+, and Surry 1. The results suggest that the nsks of This report summarizes work performed by Argonne mtional continued operation and plant shutdown both are substantial' Laboratory on the Steam Generator Tube ir tegrity Program dur-but the nsk of shutting down the plant with the equipment ine ing the period from October 1996 through March 1997, Under erable is much higher than the corresponding risk of continuing Task 1, progress is reported on the assembly of a steam gener-stor tube mock-up for round robin studies on currently practiced power operation in most multiple failures of the AFW systems. NDE procedures and on the evaluation of NDE techniques for NUREG/CR4500: THE EFFECT OF INITIAL TEMPERATURE ON characterizing the tubes going into the mock-up. Eddy, current FLAME ACCELERATION AND DEFLAGRATION-TO-DETON%- inspection data from the McGuire station have been evaluated TION TRANSITION PHENOMENON. CICCARELLI,G.; in an effort to optimize the selection of tubes for a planned tube BOCCIO.J.L; GINSBERG,T.; et al. Brookhaven National Lab- removal effort. Under Task 2, results are reported on the appli-oratory. May 1998. 75pp. 9806080235. BNL-NUREG-52515. cation of signal processing, visualization, and data analyses A3688:162, schemes to improve the NDE of service-degraded tubing. Se-The High Temperature Combustion Facility at BNL was used lected outcomes are presented on the implementation of a multi- 4 to conduct deflagration-to-detonation trcnsitica (DDT) experi- variate regression rnodel to correlate bobbin probe indications of ments. Periodic onfice plates were installed inside the entire cracking at support plate intersections with tube failure pressure, length of the detonation tube in order to promote flarne accel- in Task 3, a model boiler multitube corrosion cracking facility is eration. The orifice plates are 27.3-cm outer diameter, which is being designed, and a pressure and leak-rate test facility is , equivalent to the inner diameter of the tube, and 20.6-cm-inner being built. An autoclave system has been constructed to I diarneter. The detonation tube length is 21.3-rneters long, and produce cracked specimens for use in subsequent pressure and

Main Citation 3 and Abstracts 25 leak-rate tests and NDE studies, and room-temperature tech- NUREG/CR-6524: THE EFFECT OF LATERAL VENTING ON niques for producing cracked tubes are also being developed. DEFLAGRATION-TO-DETONAT ION TRANSITION IN HYDRO-High-temperature, severe-accident tests have been conducted GEN-AIR-STEAM MIXTURES AT VARIOUS INITIAL TEM-on tubes with a single or two symmetrical circumferential cracks PERATURES. CICCARELLl,G.; BOCCIO,J.L; GINSBERG,T.; et under a constant internal pressure and a linear temperature al. Brookhaven National Laboratory. November 1998. 69pp. rarnp to simulate the behavior of tubes with and without cor>. 9811270092. BNL-NUREG-52518. A5962:122. straint against bending. Agreement has been reached with Duke The influence of gas venting on flame acceleration in an oty Power Company for the removal of service-degraded tubes and stacle-laden tube has been investigated in the High Tempera-tube sections from the retired steam Generators from the ture Combustion Facility (HTCF) at BNL in these venting ex-McGuire 1 Nuclear Power Station. periments, the flame was observed to accelerate very quickly in the first tubo section before the first verd section. For lean hy-NUREGICR4617: ROUND ROBIN PRETEST ANALYSES OF A drogen mixtures, after the first vent section, the flame velocity STEEL CONTAINMENT VESSEL MODEL AND CONTACT decayed to a velocity on the order of the laminar buming veloc-STRUCTURE ASSEMBLY SUBJECT TO STATIC INTERNAL ity. For rnore sensitive mixtures, the flame reached a quash PRESSURIZATION. LUK,V.K.; KLAMERUS.E.W. Sandia Na- steady flame velocity similar to llame propagation in the choking tional Laboratories. August 1998. 550pp. 9810160100. SAND 96- regime observed in tests without venting. For all initial tempera-tures, the lean limit for signifcant flame acceleration (i.e., chok. 2899. A5415:001. ing regime limit) with venting increased over the nonventing The Nuclear power Engineering Corporation (NUPEC) and case by an average of 2 percent hydrogen. In the choking re-the U.S. Nuclear Regulatory Commission (NRC) are co-spork gime, the flame was observed to accelerate in the tube section soring and jointly funding a research project at Sandia National to a maxtrnum velocity close to the speed of sound in the prod-Laboratories (SNL) to conduct a failure test of a steel contairk ucts and then decelerate across the vent section. At the limited ment vessel (SCV) model and contact structure assembly. The temperatures tested where DDT was observed, the minimum SCV model, representative of an improved Mark-Il Boiling Water hydrogen concentration required for transition to detonation irk Reactor (BWR) containment vessel, is scaled 1:10 in Geornetry creased with venting present as compared to without venting. In and 1:4 in shell thicknesses. The contact structure, a thick bell- all cases, after a certain propagation distance, the detonation shaped steel shell, provides a simplified representation of a con- wave failed due to local venting effects and continued to propa-crete reactor shield building in the actual plant. The failure test gate at a velocity characteristic of the choking regime. wi3 be conducted in December 1996 to provide data on the re-sponse of the composite structure up to its failure in order to NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE CALCULATION OF STEADY-STATE, THERMAL-ME-validate analytical modeling, to find the pressure capacity of the model, and to observe the failure mechanistns. Eight inter- CHANICAL BEHAVIOR OF OXIDE FUEL RODS FOR HIGH BURNUP. BEYER,C.E.; LANNING.D.D. Battelle Memorial insti-r ational organizations performed model analyses and provided predictions of the model behavior at 43 specified locations and tute, Pacific Northwest National Laboratory. DAVIS,K.L.; et al. Idaho National Engineering & Environmental Laboratory. De-the failure pressure and mechanisms. This report describes the cember 1997.111pp. 9803050101. PNNL-11513. A2423:211, analysis models and tabulates the pretest predictions submitted FRAPCON-3 is a FORTRAN IV computer code that calculates by each organization. A pretest meeting of all participants will the steady-state response of light water reactor fuel rods during be held on October 1-2,1996, to discuss modeling approaches long-term bumus. The code calculates the temperature, pres-and analysis results. sure, and defor' nation of a fuel rod as functions of time-depend-ent fuel rod power and coolant boundary conditions. The phe-NUREGICR4521: ESTIMATING PROBABLE FLAW DISTRIBU- nomena modeled by the code include 1) heat conduction TIONS IN PWR STEAM GENERATORS. GORMAN J.A.t TURN- through the fuel and cladding,2) cladding elastic and plastic de-ER,A.P.; KREIDER,M.A.; et al. Argonne National Laboratory, formation,3) fuel-cladding rnechanical interaction, 4) fission gas August 1998.182pp. 9808310136. ANL-96/20. A4853:001. release, 5) fuel rod internal gas pressure, 6) heat transfer be-This report describes methods for estimating the number and tween fuel and cladding, 7) cladding oxidation, and 8) heat siz] distributions of flaws of various types in PWR steam gener- transfer from cladding to coolant. The code contains necessary ator tubes and to provide some typical results for typical hypo- material properties, water properties, and heat- transfer correla-thetical units. Such estimates are needed, for example, when tions. The codes

• Integral predictions of mechanical behavior calculating the probable primary to secondary leakage through have not been assessed against a data base, e.g., cladding steam generator tubes under postulated accidents. The basic strain or failure data. Therefore, it is recornmended that the assumption of the approach used is that estimating the ex- code not be used for analyses of cladding stress or strain.

pected distribution of flaws can be broken into two steps: (1) es- FRAPCON-3 is programmed for use on both mainframe com-timating the number of tubes with detectable flaw , as a function puters and UNIX-based workstatior- sch as DEC 5000 or SUN of time using a " defect occurrence distribution" in which the Sparcstation 10. It is also prograns.,ed for personal computers number of detectable flaws varies with time, and (2) distributing with FORTRAN compiler software and at least 8 to 10 rnega-the size of these flaws using a second " flaw size distribution" bytes of random access rnemory (RAM). that is either independent of time or varies with time in some NUREG/CR-6534 V03: FRAPCON-3: INTEGRAL ASSESSMENT. known way. The report discusses (1) how to estimate the num- LANNING,D.D.; BEYER,C.E. Battelle Memorial institute, Pacific ber of tubes with detectable flaws for several flaw types and Northwest National Laboratory. BERNA.G.A. Affiliation Not As-classes of steam generators, (2) the methods typically used to signed. December 1997. 210pp. 9803050091, PNNL-11513. characterize the size distributions of these flaws, and (3) how A2423:001. the above two steps are combined to provide an integrated pre- Fuel rod material properties and performance models have diction. The types of flaws of most current interest to the indus- been updated for the FRAPCON steady-state fuel rod perform-try are covered, including circumferential cracks at the top of the ance code to account for changes in behavior due to extended tube sheet, axial primary water stress corrosion cracks at roll fuel bumup. The updated code is named FRAPCON-3 and is in-transitions, freespan cracks, axial outer-diameter cracks at tube tended to replace the earlier codes FRAPCON-2 and GAPCON-support plate dents, axial cracks associated with intergranular THERMAL-2. The property and rnodel updates are described in attack / stress corrosion cracking in sludge piles, and flaws due Volume 1 of this report. Volume 2 of this report constitutes the to damage from loose parts. Examples were developed for hy- code description document and includes the input instructions. pothetical moderately affected, severely affected, and lightly af- This document (Volume 5) provides the results of the assess-focted plants with Westinghouse Model 51 steam generators. ment of the integral code predictions to measured data for var-

F 26 Main Citation 3 and Abstracts ious performance parameters. In the case of fuel temperature ther information, and more importantly, describes the accuracy and fission Das release (FGR) predictions, comparison is made of the models. Nevertheless, the report still contains over 150 to both benchmark data sets and independent benchmark data equatiort, and over 400 references. The main sections of this sets. The benchmark data sets are desenbed in Section 2.0. Ap- report describe: (1) the evolution of piping fracture mechanics pendix A describes each individual set of benchmark data and history relative to the developments of the nuclear industry, (2) gives the code input for each data comparison. The data are technical developments in stress analyses, material property as-Cawn from a wide range of burnup levels and operating condi- pects, and fracture mechanics analyses, (3) unresolved issues j tions for both PWR and BWR type rods. and technically evolving areas, and (4) a summary of conclu-sions of major developments to date. NUREG/CR4536: VERIFICATION OF THE LWRARC CODE FOR LIGHT-WATER-REACTOR AFTERHEAT RATE CALCULA. NUREG/CR4544: METHODOLOGY FOR ANALYZING PRECUR- J TIONS. MJRPHY,B.D. Oak Rioge National Laboratory. Feb. SORS TO EARTHQUAKE-INITIATED AND FIRE-INITIATED 4 ACCIDENT SEQUENCES. BUDNITZ,R.J.; LAMBERT,H.E. Fu-ruary 1998.15pp. 9802270183. ORN1/TM-13390. A2342:107. ture Resources Associates, Inc. APOSTOLAKIS,G.; et al. Mas- f This report describes verification studies carried out on the LWRARC (Light-Water-Reactor Afterheat Rate Calculations) sachusetts Institute of Technology, Cambridge, MA. April 1998. computer code. The LWRARC code is proposed for automating 149pp. 9805050400. A3320:027. the implementation of procedures specified in Draft Revision 1 This report covers work to develop a methodology for ana-of the U.S. Nuclear Regulatory Commission (NRC) Regulatory lyzing precursors to both earthquake-initiated and internal fire-Guide 3.54, " Spent-Fuel Heat Generation in an independent initiated accidents at commercial nuclear power plants. Cur- l Spent-Fuel Storage Installation," which gives guidelines on the rently, the U.S. Nuclear Regulatory Commission sponsors a l calculation of decay heat for spent nuclear fuel. Draft Regulatory large ongoing project, the Accident Sequence Precursor project, l Guide 3.64 allows one to estimate decay-heat values by means to analyze the safety significance of other types of accident pre- ( of a table lookup procedure with interpolation performed be- cursors, such as those arising from intemally4nitiated transients tween table-entry values. The tabulated values of the relevant and pipe breaks, but earthquakes and fires are not within the parameters span ranges that are appropriate for spent fuel from current scope. The results of this project are that (i) an overall e boiling-water reactor (BWR) or a pressurized-water reactor step-by-step methodology has been developed for precursors to (PWR), as the case may be, and decay-heat rates are obtained both fire-initiated and seismic-initiated potential accidents; (ii) for spent fuel whose properties are within those parameter lim- some stylized case-study examples are provided to demonstrate its. In some instances, where these limits are either exceeded how the fully-developed methodology works in practice, and (iii) or where they approach critical regions, adjustments are invoked a generic seismic-fragility data base for equipment is provided following table lookup. The LWRARC cornputer code is intended for use in seismic-precursor analyses. to replicate this rnanual process. NUREG/CR4545 V01: PROBABILISTIC ACCIDENT CON-NUREG/CR4537: INFLUENCE OF LONG-TERM THERMAL SEQUENCE UNCERTAINTY ANALYSIS. Early Health Effects AGING ON THE MICROSTRUCTURAL EVOLUTION OF NU- Uncertainty Assessment. Main Report. HASKIN,F.E. New Mex-CLEAR REACTOR PRESSURE VESSEL MATERIALS.An Atom ico, Univ, of, Albuquerque, NM. HARPER,F.T. Sandia National Probe Study. PAREIGE P. France. RUSSELL,K.F.; Laboratories. GOOSSENS,L.H.J.; et al. Delft University of Tech-STOLLER,R.E.; et al. Oak Ridge National Laboratory. March nology. December 1997. 64pp. 9804240196. EUR 16775. 1998,28pp. 9803180083. ORN!JTM-13406. A2610:324 A3158:001. Atom probe field ion microscopy (APFIM) investigations of the The development of two new probabilistic accident con-microstructure of unaged (as-fabricated) and long-term thermally sequence codes, MACCS and COSYMA, was completed in aged ( 100,000 h at 280 degrees C) surveillance mater',als 1990. These codes estimate the consequences from the acci-from commercial reactor pressure vessel steels were performed. dental releases of radiological material from hypothesized acci-This combination of materials and conditions permitted the in- dents at nuclear installations. In 1991, the U.S. Nuclear Regu-vestigation of potential thermal-aging effects. This microstruc- latory Cornmission and the Commission of the European Com-tural study focused on the quantifcation of the compositions of munities began cosponsoring a joint uncertainty analysis of the the matrix and carbides. The APFIM results indicate that there two codes. The ultimate objective of this joint effort was to sys-was rio significant microstructural evolution after a long-term tematically develop credible and traceable uncertainty distribu-thermal exposure in weld, plate, or forging materials. The matrix tions for the respective code input variables. A formal expert depletion of copper that was observed in weld materials was judgment elicitation and evaluation process was identified as the consistent with the copper concentration in the matrix after the best technology available for developing a library of uncertainty stress-relief heat treatment. The compositions of cementite car. distributions for these consequence parameters. This report fo-bides aged for 100,000 h were compared with the cuses on the results of the study to develop distribution for vari-Thermocalc(TM) prediction. The APFIM comparisons of mate. ables related to the MACCS and COSYMA early health effects rials under these conditions are consistent with the measured models. change in mechanical properties such as the Charpy transition NUREGICR4545 V02: PROBABILISTIC ACCIDENT CON-temperature. SEQUENCE UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assessment. Appendices. HASKIN,F.E. New Mexico, NUREGICR4540: STATE-OF-THE-ART REPORT ON PIPING Univ. of, Albuquerque, NM. HARPER,F.T. Sandia National Lab-FRACTURE MECHANICS. WILKOWSKI,G.M.; OLSON,R.J.; SCOTT,P.M. Battelle Memorial Institute, Columbus Laboratorhs. oratories. GOOSSENS,LH.J.; et al. Delft University of Tech-January 1998. 385pp. 9802100139. BMI-2196. A2077:001, nology. December 1997. 350pp. 9804240244. EUR 16775. This report is an in-depth summary of the state-of-the-art in A3158:068. See NUREG/CR-6545,V01 abstract. nuclear piping fracture mechanics. It represents the culmination of 20 years of work done primarily In the U.S., but also attempts NUREG/CR4646: A DAMAGE MECHANICS BASED APPROACH to include important aspects from other intemational efforts. Al- TO STRUCTURAL DETERIORATION AND RELIABILITY. though the focus of this work was for the nuclear industry, the BHATTACHARYA B.; ELLINGWOOD,B. Johns Hopkins Univ., technology is also applicable in many cases to fossil plants, pe- Baltimore, MD.

• Oak Ridge National Laboratory. February trochemical/ refinery plants, and the oil and gas industry, in com- 1998. 200pp. 9803180018. ORNLSUB96 SP638. A2610:001.

piling this detailed summary report, all of the equations and de- Structural deterioration often occurs without perceptible mani-tat :'iN analysis procedure or experimental results are not festation. Continuum damage mechanics defines structural dam-necessarily included. Rather, the report describes the important age in terms of the material microstructure, and relatos the dam-aspects and limitations, tells the reader where he can go for fur- age variable to the macroscopic strength or stiffness of the l

Main Citations cnd Abstracts 27 structure. This enables one to predict the state of damage prior achievable principle, and lessons learned from the MartMe Hill to the initiation of a macroscopic flaw, and allows one to esti- annealing demonstration. mate residual strength / service life of an existing structure. The accumulation of damage is a dissipative process that is gov-emed by the laws of thermodynamics. Partial differential equa- NUREG/CR-6554: FINITE ELEMENT ANALYSES FOR SEISMIC tions for damage growth in terms of the Helmholtz free energy SHEAR WAL'. INTERNATIONAL STANDARD PROBLEM- . are derived from fundamental thermodynamical conditions. PARK,Y.J.; HOFIMYER,C.H. Brookhaven National Laboratory. Closed-form solutions to the equations are obtained under pril 1998. 27tto. 9805050413. BNL-NUREG-52530. uniaxial loading for ductile deformation damage as a function of A3318:343. plastic strain, for creep damage as a function of time, and for Two identical reinfo ced concrete (RC) shear walls, whbh fatigue damage as function of number of cycles. The proposed consist of web, flanges and massive top and bottom slabs, were damage growth model is extended into the stochastic domain by tested up to ultimate failure under earthquake motions at the considering fluctuations in the free energy, and closed-form so- Nuclear Power Engineering Corporation's (NUPEC) Tadotsu En-lutions of the resulting stochastic differential equation are ob. gineering Laboratory, Japan. NUPEC provided the dynamic test tained in each of the three cases mentioned above. A reliability results to the OECD (Organization for Econome Cooperation analysis of a ring-stiffened cylindrical steel shell subjected to and Development), Nuclear Energy Agency (NEA) for use as an corrosion, accidental pressure, and temperature is performed. International Standard Problem (ISP). The shear walls were in. NUREG/CR-6551: IMPROVED EMBRITTLEMENT CORRELA- tended to be part of a typical reactor building. One of the rnajor TlONS FOR REACTOR PRESSURE VESSEL STEELS. objectives of the Seismic Shear Wall ISP (SSWISP) was to 1 EASON.E.D.; WRIGHT.J.E. Modeling & Computer Services. evaluate various seismic analysis methods for concrete struc- j ODETTE,G.R. California, Univ of. Santa Barbara, CA. Novem- tures used for design and seismc margin assessment. It also ber 1998.127pp. 9811200054. MCS 970501. A5901:204. offered a unique opportunity to assess the state.cf-the-art in The reactor pressure vessels of commercial nuclear power nonlinear dynamic analysis of reinforced concrete shear wall plants are subject to embrittlement due to exposure to high en. structures under severe earthquake loadings. As a participant of ergy neutrons from the core. The purpose of the reported work the SSWISP workshops, Brookhaven National Laboratory (BNL) was to improve on the correlation models in the current Regu. performed finite element analyses under the sponsorship of the latory Guide 1.99 Revision 2 (RG1.99/2) using the broader data U.S. Nuclear Regulatory Commission (USNRC). Three types of base now evailable. The embnttlement data base used for this analysis were performed, i.e., rnonotonic static (push-over), cy-analysis was derived primarily from the Power Reactor Embrit- clic static and dynamic analyses. Additional rnonotonic state tiement Data Base developed at Oak Ridge National Laboratory. analyses were performed by two consultants. F. Vecchio of the Shifts in transition temperature and drops in upper shelf energy University of Toronto (UT) and F. Filippou of the University of were calculated on a consistent basis from individual fits to Califomia at Berkeley (UCB). The analysis results by BNL and unirradiated and irradiated raw Charpy data. The final transition the consultants were presented during the second workshop in temperature shift (TTS) and upper shelf energy (USE) models Yokohama, Japan in 1996. A total of 55 analyses were pre-include fluence, copper, nickel, phosphorous content, and prod- sented during the workshop by 30 parteipants from 11 different uct form; the TTS model also includes coolant temperature and countries. The major findings on the presented analysis rneth-neutron flux. The models were developed using multivariable ods, as well as engineering insights regarding the applicability surface-fitting techniques, based on pattem recognition, under- and reliability of the FEM codes are described in detail in this standing of the TTS mechanisms, and engineering jud0ement. report. The key variable trends, such as the copper nickel dependence in the new TTS model, are much improved over RG1.99/2 and are well supported by independent elata and current under. NUREG/CR-6555 V01: PROBABILISTIC ACCIDENT CON-standing of embrittlement mechanisms. The improved TTS SEQUENCE UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assessment. Main Report. LITTLE,M.P.; model reduces scatter significantly relative to RG1.99/2 on the MUIRHEAD,C.R. United Kingdom GOOSSENS,LH.J.; et al. currently-available data base for plates, forgings, and welds

• Delft University of Technology. December 1997. 64pp.

NUREG/CR-6552: MARBLE HILL ANNEALING DEMONSTRA- 9802230110. EUR 16774. A2257:293. TION EVALUATION. OLAND,C.B.; BASS B.R.; BRYSON.J.W.; The development of two new probabilistic accident con-et al. Oak Ridge National Laboratory. February 1998.155pp. sequence codes, MACCS and COSYMA, was developed in 9807270436. ORNUTM-13446. A4385:139. 1990. These codes estimate the consequence from the acci-During the summer of 1996, an unirradiated reactor pressure dental releases of radiological material from hypothesized acci-vessel at the abandoned Marble Hill nuclear power plant was dents at nuclear installations. In 1991, the U.S. Nuclear Regu-successfully annealed thereby demonstrating that existing tech- latory Commission and the Commission of the European Com-nology can be used to thermally anneal reactor pressure ves. munities began cosponsoring a joint uncertainty analysis of the sels at commercial pressurized water reactor nuclear power two codes. The ultimate objectrve of this joint effort was to sys-plants an the United States. An independent evaluation of engi- tematically develop credible and traceable uncertainty distribu-neenng issues associated with the annealing demonstration was tions for the respective code input variables. A formal expert conducted at the Oak Ridge National Laboratory for the U.S. Judgment elicitation and evaluation process was identified as the Nuclear Regulatory Commission. Temperature, strain, and dis-best technology available for developing a library of uncertainty placement data acquired during the annealing demonstration distributions for these consequence parameters. This report fo. were used to verify thermal-hydraulic and structural analysis re-cuses on the results of the study to develop distribution for vari-sults. Based on findings and observations from the annealing demonstration and analytical results, an instrumentation system ables related to the MACCS and COSYMA late health effects models* was developed for use in assessing thermal annealing at other nuclear power plants similar to Marble Hill. Objectives of the in-strumentation system are to provide sufficient data for deter. NUREG/CR-6555 V02: PROBABILISTIC ACCIDENT CON-mining if the observed time and temperature profile satisfies or SEQUENCE UNCERTAINTY ANALYSIS. Late Health Effects exceeds the required thermal annealing conditions, and for Uncertainty Assessment.Apperdces. LITTLE.M.P.; verifying thermal and structural analysis results. Development of MUIRHEAD,C.R. United Kingdom GOOSSENS,L.H.J.; et al. the instrumentation system involved consideration of technical Delft University of Technology. December 1997. 223pp. requirements, issues related to minimizing occupational expo- 9802230116. EUR 16774. A2257:070 sure to radiation in accordance with the as low as is reasonably See NUREG/CR-6555,V01 abstract.

28 Main Citations and Abstract] NUREG/CR-6556: A COMPREHENSIVE STUDY OF THE EAST- drops range from abuut 1 to 100 bars. The West Texas earth. ERN TENNESSEE SEISMIC ZONE. POWELL C.A.; quake has lower values of a few bars to a few tens of bars. VLAHOVIC,G. North Carolina, Univ. of, Chapel Hill, NC. CHAP-NUREG/CR-6569: LOW-LEVEL WASTE DATA BASE DEVELOP. MAN,M.C.; et al. Virginia Polytechnic Institute & State Univ., MENT PROGRAM. MCCONNELL,J.W.; ROGERS R.D. Idaho Blacksburg, VA. December 1998. 77pp. 9902220175. FIN G- National Engineering & Environmental Laboratory. 6107. A6874280. JASTROW,J.D.; et al. Argonne National Laboratory. September A joint hypocenter-velocity inversion for the eastem Ten- 1998.400pp.9810130250. INEELEXT9700925. A5369:001. nessee seismic zone has resolved velocity features in basement The Field Lysimeter investigations: Low-Level Waste Data rock below detached Appalachian thrust sheets. There is a sys- Base Development Program, funded by the U.S. Nuclear Regu-tematic velocity increase with depth and a stable VPNS ratio latory Commission, (a) studied the degradation effects in between 1.73 and 1.74. Three-D P-wave and S-wave solutions EPICOR-il organic lon-exchange resins caused by radiation, (b) are very similar, Relative to the 1-D model, velocity anomalies examined the adequacy of test procedures recommended in the range from -8% to +16% in the upper 5 km and between e 7,% " Technical Position on Waste Form" to meet the requirements in deeper layers. Prorrunent velocity anomalies parallel the seis- of 10 CFR 61 using solidified EPICOR-il resins, (c) obtained mic zone and are consistent from layer to layer. The most per- performance information on solidified EPICOR-Il ion-exchange sistent anomaly is a low velocity region that borders the seismic resins in a disposal environrnent, and (d) detemnined the condi- I zone to the northwest and is flanked on either size by regions tion of EPICOR-Il liners. Results of 10 years of data acquisition of anomalously high velocity. The NY-AL magnetic lineament from operation of the field testing in a disposal environment are coincides with or lies close to the southeast boundaty of the presented and discussed. During the field testing, both portland prominent velocity low. The spatial coincidence between veloc- type hil cement and Dow vinyl ester-styrene waste forms were ity, gravity and magnetic gradients suggests that major dis- tested in fysimeter arrays located at Argonne National Labora-continuities are present in the basement. Earthquakes are cor- tory-East in Illinois and at Oak Ridge National Laboratory. The related with the velocity anomalies and probably occur on an experimental equipment is described. The study was oesigned cient faults and fractures. Focal mechanistm suggest that strike- to provide continuous data on nuclide release and movement, i slip motion on steeply dipping planes is the dominant mode of as well as environmental conditions, over a 20-year period. At faulting throughout the 300 km long seismic zone, the end of the tenth year, the experiment was closed down. Ex-NUREG/CR4559: LARGE-SCALE VIBRATION TESTS OF MAIN amination of soil and waste forms was conducted, and the re-STEAM AND FEEDWATER PIPING SYSTEMS WITH CON- sutts are reported here. Results of waste form characterization VENTIONAL AND ENERGY-ABSORBING SUPPORTS. using tests recommended by the " Technical Position on Waste PARK,Y.J.; DEGRASSI,G.; BEZLER P.; et al. Brookhaven Na- Form" are presented and used as a basis of compan, son to the tonal Laboratory. August 1998. 400pp. 9809100252. BNL- field testing waste form final characterizations. NUREG-52532. A5011:001. NUREG/CR-6571 V01: PROBABILISTIC ACCIDENT CON-As part of collaborative efforts between the United States and SEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assess-Japan, the U.S. Nuclear Regulatory Commission (USNRC) and ment For Intemal Dosimetry. Main Report. GOOSSENS,L.H.J.; Brookhaven National Laboratory (BNL) participated in the Seis- KRAAN,B.C.P.; et al. Delff uwersity of Technology. HAR-mic Proving Test program of rnain steam and feedwater sys- RISON,J.D. National Radiok .#.;el Pr6ection Board. April 1998. l tems (MS) conducted by the Nuclear Power Engineering Cor- 70pp. 9805180257. EUR 16773. W252/ !. poration (NUPEC) foi the Ministry of Intemational Trade and in- The development of two new probabilistic accident con-dustry (MITI) of Japan. Scaled models of main steam piping for sequence codes, MACCS and COSYMA, was completed in a typical PWR plant and feedwater piping for a BWR plant were 1990. These codes estimate the consequence from the acci-fabricated by NUPEC and subjected to a large number of earth- dental releases of radiological material from hypothesized acck quake motions at NUPEC's Tadotsu Engineering Laboratory. Ini- dents at nuclear installations. In 1991, the U.S. Nuclear Regu-tialty, the piping systems were supported by conventional snub. latory Commission and the Commission of the European Com-bers and subjected to design level earthquakes as well as exci- munitios began cosponsoring a joint uncertainty analysis of the tations beyond the design level. Then the snubbers were re- two codes. The ultirnate objective of this joint effort was to sys-placed by energy absortxng devices, and tests at various exci- tematically develop credible and traceable uncertainty distribu-tation levels up to the deformation limits of the energy absorbers tions for the respective code input variables. A formal expert were perforrned. This report describes the evaluation of the test judgment elicitation and evaluation process was identified as the results and BNL's post-test analyses. best technology available for developing a library of uncertainty distributions for these consequence parameters. This report to. NUREG/CR-6564: ANALYSES OF SOURCE SPECTRA, AT- cuses on the results of the study to develop distribution for varl-TENUATION, AND SITE EFFECTS FROM CENTRAL AND ables related to the MACCS and COSYMA intemal dosimetry EASTERN UNITED STATES EARTHOUAKES. LINDLEY,G. models. a a, n of, Barbara, CA. February 1998.100pp. NUREG/CR-6571 V02: PROBABILISTIC ACCIDENT CON-Results from 27 previous studies were used to analyze stress SEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assess-  ; drop vs. magnitude in eastem North America. Stress drop was ment For Intemal Dosimetry. Appendices. GOOSSENS,LH.J.; ' not constant, but increased approximately with the square root KRAAN.B.C.P.; et al. Delft University of Technology. HAR-of the seismic moment from 3 bars at 10(20) dyne cm to 690 RISON,J.D. National Radiological Protection Board. April 1998. bars at 10(25) dyne-cm. O(ig) as a function of frequency was 317pp. 9805180287. EUR 16773. A3426:001. See NUREG/CR-6571,V01 abstract. analyzed in five regions of the contiguous United States. Simul. taneous inversions using Fourier amplitude spectra were com- NUREG/CR-6573: " INVESTIGATING SEISMOTECTONICS IN l puted to determine attenuation, site responses, and source THE EASTERN UNITED STATES USING A GEOGRAPHIC IN- I spectra. Unlike some previous studies, O(Ig) in the central and FORMATION SYSTEM." EBEL,J.E.; LAZAREWICZ,A.R.; ) northeastern U.S. was found to be nearty identical from 2 to 10 KAFKA,A.L Boston College, Weston, MA. February 1998. Hz. Q(ig) in the southeastem U.S. is about 20% lower. Anelastic 109pp. 9803030332. A2379-209. attenuaton of four regional phases, and source parameters of A Geographic information System (GIS) database has been 27 earthquakes, including the 1995 West Texas earthquake assembled to use in regional analyses looking for (M(b) 5.6) were also estimated. L(g) attenuation is in good seismotectonically active features in the central and eastam agreement with previous estimates for the central and eastem U.S. (CEUS). Included in the database for the region are topog-U.S. Assuming a single comer frequency source rnodel, stress raphy, earthquakes, stress measurements, gravity residual field,

r ) I

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Main Citations old Abstracts 29 1 magnetic residual field, major rivers and regional geology, espe- risk (core damage frequency impacts of the stressors), indicate l cially faults. Observables from this database were extracted for that humidity, EMI from lightning, and smoke can be potentially the seismically active areas of the northeastern, southeastern risk-significant. Risk from other sources of EMI could not be and central U.S. for use in multivariate statistical analyses. evaluated for a lack of data. Risk from temperature appears to These analyses indicate that the earthquakes of the CEUS do be insignificant as that from the assumed levels of vibrations. A tend to associate with faults and other deformation structures, comparison of the hardware unavailability of the existing analog but that the geologic characteristics are not very similar between Safety injection Actuation System (SIAS) in the example plant earthquakes in different regions. The discriminant function anal- with that of an assumed digital upgrade of the system indicates ysis shows some ability to differentiate between seismic and that system unavailability may be rnore sensitive to the level of nonseismic areas. redundancy in elements of the digital system than to the envh l NUREGICR4575: FAILURE BEHAVIOR OF INTERNALLY PRES- ronmental and operational variations involved. The findings of SURIZED FLAWED AND UNFLAWED STEAM GENERATOR this study can be used to focus activities relating to the regu-TUBING AT HIGH TEMPERATURE -EXPERIMENTS AND latory basis for digital I&C upgrades in NPPs, including identi-COMPARISON WITH MODEL PREDICTIONS. MAJUMDAR,S.; ncation of domina,nt stressors, data $athering, equipment quali-SHACK,W.J.; DIERCKS,0.R.; et al Argonne National Labora- fication, and requirements to limit the effects of environmental tory. March 1998.102pp. 9803260391. ANL-97/17. A2738221. stressors. This report summartzes experimental work performed at Ar-gonne National Laboratory on the failure of intemally pressur- NUREGICR-6580: PERFORMANCE TESTING OF PASSIVE AUTOCATALYTIC RECOMBINERS. BLANCHAT,T.K. Sandia ized steam generator tubing at high temperatures (2 700 de-National Laboratories. MALLIAKOS,A. Division of Systems grees C). A model was developed for predicting failure of flawed Technology (Post 941217). June 1998. 212pp. 9809210241* and unflawed steam generator tubes under Iraernal pressure SAND 97-2632. A5130:001.  ! and temperature histories postulated to occur during severe ac- Performance tests of a scaled passive autocatalytic recom-cidents. The model was validated by failure tests on specimens ) biner (PAR) were performe0 in the Surtsey test vessel at Sandia with part-throughall axial and circumferential flaws of various National Laboratories. The test program included experiments i lengths and depths, conducted under various constant and ~ to: 1) define the startup characteristics of PARS,2) confirm a hy-rarrped intemal pressure and temperature conditions. The fall drogen depletion rate curve of PARS,3) define the PAR per-ure temperatures predicted by the model for two temperature formance in the presence of steam, 4) evaluate the effect of and pressure histories, calculated for severe accidents initiated scale (number of cartridges) on the PAR performance at both by a station blackout, agree very well with testa performed on low and high hydrogen concentrations,5) define the PAR per-both flawed and unflawed specimens. formance with and w:thout the hydrophobic cod,6) determine NUREG/CR-6677: U.S. NUCLEAR POWER PLANT OPERATING if the PAR could Ignite hydrogen mixtures, 7) define the PAR COST AND EXPERIENCE SUMMAR3ES. KOHN,W.E.; performance in well-mixed conditions, and 8) define the PAR REID,R.L: WHITE,V.S. Oak Ridge National Laboratory. Feb. performance in a low oxygen environment. The tests determined ruary 1998. 433pp. 9802230118. ORNIJTM-13494. A2256 001. that the PAR startup delay times decrease with increasing hy-This report has been prepared to provide historical operating drogen concentrations in steamy environments. Measured de-cost and experience information on U.S. commercial power pletion rate data were obtained and compared with previous plants. Costs incurred after initial construction are characterized work. Depletion rate appears to be proportional to scale. PAR as annual production costs, representing fuel and plant oper. performance in steamy environments and the lack of hydro-ating and maintenance expenses, and capital expenditures re. phobic coating was investigated. Placement of the PAR near a lated to facility additions /rnodifications which are included in the wall (as opposed to a center location) appeared to have an ef-plant capital asset base. As discussed in the report, annual data fact on depletion rates. The PAR ignited hydrogen at relatively for these two cost categories were obtained from publicly avaip high concentrations (5-10 mole %). Low oxygen concentrations able reports and must be accepted as having different degrees appeared to have an effect on the hydrogen / oxygen recombina-of accuracy and completeness. Treatment of inconclusive and tion rate. The effect of well-mixed condrtions during depletion incomplete riata is discussed. As an aid to understanding the rate measurements were inconclusive, fluctuations in the cost histnries, operations summaries for each nuclear unit are provided. The intent of these summaries is to NUREG/CR-6583: EFFECTS OF LWR COOLANT ENVIRON-identify important operating events; refueling, major mainte- MENTS ON FATIGUE DESIGN CURVES OF CARBON AND nance, and other significant outages; operating milestones; and LOW-ALLOY STEELS. CHOPRA,0.K.; SHACK,W.J. Argonne significant licensing or enforcement actions. Information used in National Laboratory. March 1998.128pp. 9803260384. ANL-97/ the surnmaries is condensed from annual operating reports sub- 18 A2727:220 khe ASME Boiler and Pressure Vessel Code provides rules mitted by the licensees, plant histories contained in Nuclear for the construction of nuclear power plant components. Figures Power Experience, trade press articles, and the Nuclear Regu-latory Commission (NRC) web site (www.nrc. gov). l-9.1 through I-9.6 of Appendix 1 to Section lli of the Code speci-fy fatigue design curves for structural materials. While effects of NUREG/CR-6579: DIGITAL l&C SYSTEMS IN NUCLEAR reactor coolant environments are not explicitly addressed by the POWER PLANTS. Risk-Screening Of Environmental Stressors design curves, test data indicate that the Code fatigue curves And A Comparison Of Hardware Unavailability With An Existing may not always be adequate in coolant environments. This re-Analog System. HASSAN,M. Brookhaven National Laboratory. port summarizes work performed by Argonne National Labora-VESELY,W.E. Science Applications intemational Corp. (formeity tory on fatigue of carbon and low-alloy steels in light water reac-Science Applications, Inc.). January 1998,128pp. 9802130002. tor (LWR) environments. The existing fatigue S-N data have BNL-NUREG-52536. A2144:068. been evaluated to estaolish the effects of various material and in this report, we present a screening study to identify envi- loading variables such as steel type, dissolved oxygen level, ronmental stressors for digital instrumentation and control (l&C) strain range, strain rate, temperature, orientation, and sulfur systems in a nuclear power plant (NPP) which can be potentialty content on the fatigue life of these steels. Statistical models risk-significant, and compare the hardware unavailability of such have bean developed for estimating the fatigue S-N curves as a system with that of its existing analog counterpart. The a function of material, loading, and erwironmental variables. The stressors evaluated are temperature, humidity, vibration, radi- results have been used to estimate the probability of fatigue ation, electro- magnetic interference (EMI), and smoke. The re- cracking of reactor components. The different methods for ircor-suits of risk-screening for an example plant, subject to some porating the effects of LWR coolant environments on the ASME bounding assumptions and based on relative changes in plant Code fatigue design curves are presented.

30 Main Citations and AbDtracts NUREG!CR-6589: THE EFFECTS OF SURFACE CONDITION method, while the other technique uses low frequencies (350 ON AN ULTRASONIC INSPECTION: ENGINEERING STUDIES kHz) and a synthetic aperture focusing technique (SAFT). The USING VALIDATED COMPUTER MODEL. GREENWOOD,M.S. primary focus of this work is to provide information to the NRC Battelle Memorial institute, Pacific Northwest National Labora- on the effectiveness and reliability of these two ultrasonic testing tory. April 1998.154pp. 9805200010. PNNL-11751. A3467:217. (UT) techniques for the inservice inspection of components in This report documents work perforrned at Pacific Northwest pressurized water reactors (PWRs). For low frequency ultrasonic National Laboratory (PNNL) on the effects of surface roughness inspections of centrifugally cast stainless steel (CCSS), the data on the reliability of an ultrasonic inservice inspection. The pri- reported from laboratory work indicates that an ultrasonic tech-mary objective of this research is to develop ASME Code rec- nique utilizing 0 degrees,30 degrees,45 degrees, and 60 de-ommendations in order to limit the adverse effects of a rough grees incident longitudinal waves in a pitch-catch configuration surface and thereby increase the reliability of ultrasonic inserv- at 350 kHz has demonstrated promising results for coarse-ice inspections. In order to achieve this objective, engineering grained material inspections. Utilization of SAFT coupled with studies were conducted that included experimental validation of the low frequency UT technique in a blind test provided positive computer codes, developed at the Center for Nondestructive performance results, yielding a relatively low false call prob-Evaluation (CNDE) at lowa State University as a result of a co- ability (30%) and relatively high probability of detection (70%). operative effort between the Electric Power Research Institute Fracture mechanics evaluations estimate that the relatively low (EPRI) and the Nuclear Regulatory Commission. The basic service loads (including seismic events) on PWR primary cool-problem associated with a rough surface in an inservice inspec- ant piping result in high flaw tolerances even for severely aged tion is that as the transducer rotates slightly to accommodate CCSS and circumferential cracks approximately 50% in through-the rough surface, the beam direction in the metal changes and wall size are important to structural integrity, the timedflight of the echo changes as well One problem is NUREG/CR-6598: AN INVESTIGATION OF TENDON SHEATH-the excessive weld crown, where weld material protrudes above ING FILLER MIGRATION INTO CONCRETE. NAUS,D.J.; the adjoining surfaces. In this research this condition is modeled OLAND,C.B. Oak Ridge National Laboratory. March 1998.81pp. by considering a step discontinuity on the top surface. CNDE 9807060348. ORNtJTM-13554. A4009:236. developed several rnodels of increasing complexity in order to During some of the inspections at nuclear power plants with model an inservice inspection. This report describes the valida- prestretaed concrete containments, it was observed that the tion of four computer codes. These codes were used to mimic containments had experienced leakage of the tendon sheathing an inservice inspection in order to understand effects associated filler (i.e., streaks). The objective of this activity was to provide with rotation of the transducer as it traverses a step disconte- an indication of the extent of tendon sheathing filler leakage into nuity. Studies resulted in ASME Section XI Code recommenda" the concrete and its affects on concrete properties. Literature tions. was reviewed and concate core samples were obtained from NUREG/CR4593: CRUSTAL STRUCTURE AND GROUND MO- the Trojan Nuclear PlaM snd tested. The literature primarily ad-TION MODELS IN THE EASTERN AND CENTRAL UNITED dressed effects of crude or lubricating oils that are known to STATES FROM NATIONAL SEISMOGRAPHIC NETWORK cause concrete damage. However, these materials have signifi-DATA. SAIKIA,C.K.; SOMERVILLE.P.G.; THIO,H.K.; et al. . De cantly different characteristics relative to the materials used as cember 1998.170pp. 9902090217. A6757:133. terdcr4 sheathing fillers. Examination and testing of the concrete The aim of this study was to analyze crustal structure models, cores indicated that the appearance of tendon sheathing filler on regional seismicity and ground motion in the eastern and central the concrete surface was due to leakage from the conduits and United States. We inverted long-period regional waveforms r3- its subsequent migration through cracks that were present. Mi-corded by National Seismograph Network (NSN) stations to de- gration of the tendon sheathing filler was confined to the cracks termine source parameters. NSN seismograms were used to and there was no perceptible movement into the concrete. Re-calibrate crustal rnodels by fitting travel times of regional suits of compressive strength testing indicated that the concrete phases. The final crustal model was then used to synthesize quality was consistent in the containment and that the strength ground motion attenuation. Broadband acceleration time his. had increased over 40% in 25.4 years relative to the average toriesof New Madrid earthquakes were simulated using a hybrid compressive strength at 26-days age. method. High-frequency (f > 1 Hz) and low-frequency (f < 1 Hz) NUREG/CR4599: LODINE VOLATILITY AND PH CONTROL IN time histones, simulated separately, using a semi-ewirical THE AP-600 REACTOR. WEBER.C.F.; BEAHM,E.C. Oak Ridge method for high frequencies and a deterministic method for low National Laboratory. October 1998. 25pp. 9811050329. ORNll frequencies were added using a matched filter. The method has TM-13555. A5695:191. been validated against many recorded accelerograms of large Two design-basis accidents for the AP400 reactor are formu-earthquakes. Path effect was included using a velocity model lated and evaluated, in which significant bypass of the principal that was developed for the entire region by fitting-empirical at- pH control system occurs. Some iodine released from the reac-tenuation curves for Lg waves recorded by the NSN stations at tor primary system is retained in the incontainment Refueling distances beyond 100 km. The crustal structure rnodel contains Water Storage Tank (IRWST) water, never entering the contain-attemating high and low velocity layers to trap high frequency ment, where trisodium phosphate produces a high pH. Some of Lg waves with extended duration compared to those models this iodine is volatilized and is transported into the reactor con-that do not have such layering structure. Regional seismograms tainment airspace. In the worst case, a small fraction is released generated using our model have also produced a reasonable to the environment at design basis leak rate, yielding a total cu-agreement to response spectral values computed for small mutative iodine release at 30 days of 0.0352 mol (0.023% of events. Ground motion estimates for targe earthquakes gen- core iodine inventory) due to the iodine volatilization bypassing erally lie between the median and 84th percentile estimates of the pH control system. No fission product removal in the con-the stochastic model of Toro et al. (1997). tainment atmosphere (i.e., natural deposition sprays) is consid-ered. NUREG/CR4594: EVALUATION OF ULTRASONIC INSPECTION I TECHNIQUES FOR COARSE-GRAINED MATERIALS. NUREG/CR4600: NEUTRON EXPOSURE PARAMETERS FOR I DIAZ,A.A.; DOCTOR,S.R.; HILDEBRAND.B.P.; et al. Battelle CAPSULE 10.05 IN THE HEAVY-SECTION STEEL IRRADIA-Memonal Institute, Pacific Northwest National Laboratory. Octo- TlON PROGRAM TENTH IRRADIATION SERIES. REMEC,l.; l ber 1998. 200pp. 9812220125. PNNL-11171. A6311:023. BALDWIN,C.A.; KAM,F.B.K. Oak Ridge National Laboratory. l This report documents work performed at PNNL on evaluating October 1998. 48pp. 9810230175 ORNt/TM-13548.  ; two inherently different uttrasonic techniques used for inspection A5509:266. f of coarse-grained materials. One technique uses focused higher This report describes the computational methodology for the frequencies (2.25 MHz) and an adaptsve signal processing least-squares adjustment of dosimetry data from the HSSI 10.05

I Main Citation] and Abstracts 31 l > 1 capsule with neutronica calculations, it presents exposure pa- code uses a combination of tables and/or numerical models of l rameters for the metallurgical specimens irradiated in the cap- source term reduction phenomena to determine the time de-sule. The exposure pararneters reported are the neutron fluence pendent dose at user specified locations for a given accident l greater than 1.0 MeV, fluence greater than 0.1 MeV, and dis- scenario. The code also provides the inventory, decay chain, l placements per atom. Exposure parameter distiibutions are also and dose conversion factors needed for the dose calculation. i described in terms of three-dimensional fitting functions. When The RADTRAD code can be used for occupational radiation ex-i fitting functions are used, it is recommended that an uncertainty posure assessments, typically in the control room, for site l of 6% (1 O ) be associated with the exposure parameters. boundary dose estimates, and for dose attenuation estimates NUREG/CR-6601: NEUTRON EXPOSURE PARAMETERS FOR due to facility or accident sequence modifications. THE DOSIMETRY CAPSULE IN THE HEAVY SECTION STEEL IRRADIATION PROGRAM TENTH IRRADIATION SERIES. NUREGICR-6605: AN EVALUATION OF HUMAN FACTORS RE-l REMEC,1.; BALDWIN,C.A.; KAMf.B.K. Oak Ridge National SEARCH FOR ULTRASONIC INSERVICE INSPECTION.

Laboratory. October 1998. 46pp. 9810230178. ORNtJTM-13549. POND,D.J.; DONOHOO,D.T.; HARRIS,R.V. Battelle Memorial l l A5509220. Institute, Pacific Northwest National Laboratory. March 1998. I j This report describes the computational methodology for the 41pp. 9803260380. PNNL-11797. A2726291.

least-squares adjustment of the dosimetry data from the HSSI This work was undertaken to determine if human factors re-10.00 dosimetry capsute with neutronics calculations. It pre- search has yleIded information applicable to upgrading require-l sents exposure rates at each dosimetry location for the neutron ments in ASME Boller and Pressure Vessel Code Section XI l l fluence greater than 1.0 MeV, fluence greater than 0.1 MeV* improving methods and techniques in Section V, and/or sug-and displacements per atom. Exposure parameter distributions gesting relevant research. A preference was established for in-are also described in terms of three-dimensional fitting functions, formation and recommendations which have become accepted When fitting functions are used it is suggested that an uncer-and standard practice. Manual Ultrasonic Testing / Inservice In-s The fca ity of a met at nd of irr - spection WSI) is a complex task subject to inhnu by doz-diation is listed in the Appendix. ens of variables. This review frequently revealed equivocal find-ings regarding effects of environmental variables as well as re-NUREGICR-6603: CHARACTERIZATION OF RETARDATION peated indications that inspection performance may be more, MECHANISMS IN SOIL. WESTRICH,H.R.; BRADY,P.V.; and more reliability, influerred by the workers' social environ. CYGAN,R.T.; et al Sandia National Laboratories. April 1998. ment, including rnanagerial practices, than by other situational 32pp. 9807280076. SAND 9%419. A4397:273. variables. Also of significance are each inspectors relevant Performance assessment of radioactive waste sites is de- knowledge, skills, and abilities, and determination of these is pendent upon our understanding of the retardation of radio- seen as a necessary first step in upgrading requirements, rneth-l nuclides in soils under various geochemical conditions. Im- ods, and techniques as well as in focusing research in support proved fate and transport codes should include a mechanistic of such programs. While understanding the effects and medi-model of radeonuclide retardation so as to minimize risk uncer- ating mechanisms of the variables impacting inspection perform-tainties. We have investigated metal sorption (Ca(+), Sr(2+), and ance is a worthwhile pursuit for researchers, initial improve-Ba(2+)) on kaolinite to understand better interactions of clays ments in industrial UT/ISI performance may be achieved by irn-with contaminated groundwaters. Spectroscopic analyses show lementing practices already known to mitigate the effects of po-that Cs is sorbed on edge and basal sites of kaolinite, while AFM results suggest that edges comprise from 10 to 50% of the tenbah adverse conMons. surface area. Proton adsorption isotherms for kaolinite and mo-tecular rnodeling results show that aluminol edge binding sites NUREG/CR-6606: INVESTIGATION OF TECHNIQUES FOR THE are more reactive at typical soll pH than silanol edge sites, and DEVELOPMENT OF SEISMIC DESIGN BASIS USING THE are the most likely sites for metal sorption on kaolinite in natural PROBABILISTIC SEISMIC HAZARD ANALYSIS. , solutions. Recent rneasurements and characterization of Sr(2+) BERNREUTER.D.L.; BOISSONNADE.A.; SHORT,C.M. Law- l and Ba(2+) sorption onto montmorillonite clay indicate that the rence Livermore National Laboratory. April 1998. 168pp. ) basal plane residual charge greatty influences metal sorption. 9805060107. UCRL-ID-128920. A3321:113. On the other hand, phase transformation kinetics (ferrthydrite to The Nuclear Regulatory Commission asked Lawrence Liver-goethite) may be more important control over metal sorption and more Laboratory to form a group of experts to assist them in re-desorption from Fe-oxyhydroxides. These results provide the vising the seismic and geologic siting criteria for nuciaar power basis to understand and predict metal sorption onto clays, and plants, Appendix A to 10 CFR Part 100. This document de-O framework to charactertze sorption processes on more com- scribes a deterministic approach for determining a safe Shut-plex soll minerals. down Earthquake (SSE) Ground Motion for a nuclear power NUREQlCR 6604: RADTRAD: A SIMPLIFIED MODEL FOR plant site. One disadvantage of this approach is the difficulty of RADIONUCUDE TRANSPORT AND REMOVAL AND DOSE integrating differences of opinions and differing interpretations ESTIMATION. HUMPHREYS,S.L.; MILLER,LA.; et al. Sandia into seismic hazard characterization. In answer to this, prob-National Laboratories. HEAMES,T.J. . April 1998. 408pp. abilistic seismic hazard assessment methodologies incorporate 9805180342. SAND 98-0272. A3424:001. differences of opinions and interpretations among earth science This report documents the RADTRAD computer code deveL experts. For this reason, probabilistic hazard methods were se-oped for the U.S. Nuclear Regulatory Commission, Office of Nu. lected for determining SSEs for the revised regulation,10 CFR clear Reactor Regulation to estimate transport and removal of Part 100.23. However, because these methodologies provide a radionuclides and dose at selected receptors. The document in, composite analysis of all possible earthquakes that may occur, cludes a users' guide to the code, a description of the technical they do not provide the familiar link between seismic design basis for the code, the quality assurance and code acceptance loading requirements and engineering design practice. There-testin0 documentation, and a programmers' guide. The fore, approaches used to characterize seismic events (mad-l RADTRAD code can be used to estimate the containment re- nitude and distance) which best represent the ground motion lease using either the TID-14844 or NUREG-1465 source terms, level determined with the probabilistic hazard analysis were in-and assumptions, or a user-specified table, in addition, the code vestigated. This report summarizes investigations conducted at can account for the reduction in the quantity of radioactive mate- 69 nuclear reactor sites in the central and eastam U.S. for de-rial due to containment sprays, natural deposition, filters, and termining SSEs using probabilistic analyses. Attemative tech-other natural and engineered safety features. The RADTRAD niques are presented along with justification for key choices.

32 Main Citations and Abstracts NUREG/CR-6608:

SUMMARY

AND EVALUATION OF LOW-VE- nological modeling and new output options, initial installation of LOCITY IMPACT TEST OF SOLID STEEL BILLET ONTO CON- the code, written in FORTRAN 77, requires a 486 or higher IBM-CRETE PADS. WITTE,M.C.; HOVINGH J.; MOK,G.C.; et al. enmpatible PC with 8 MB of RAM. Lawrence Livermore National Laboratory. February 1998. NUREG/CR-6613 V02: CODE MANUAL FOR 184pp. 9802250129. UCRL 129211. A2281:001. MACCS2. Preprocessor Codes COMIDA2, FGRDCF, IDCF2. Spent fuel storage casks intended for use at independent spent fuel storage installations are evaluated during the applica-CHANIN,0. Technadyne Engineering Consultants Inc YOUNG,M.L Sandia National Laboratones. May 1998. '102pp tion and review process for low-velocity impacts representative of possible handling accidents. In the past, the analyses in- 9806150062. SAND 97-0594. A3799:129-volved in these evaluations have assumed that the casks This report is a user's guide for the preprocessors developed dropped or tipped onto an unyielding surfaces conservative and fM the MACCS2 code. MACCS2 mpresents a major enhanc& simplifying assumption. Applicants are currently seeking a more ment of its predecessor MACCS, the MELCOR Accident Con-realistic model for the anaryses to predict the effect of a cask sequence Code System. MACCS, distributed by govemment dropping onto a reinforced concrete pad, including energy ab- code centers since 1990, was developed to evaluate the im-sorbing aspects such as cracking and flexure. To develop data pacts of severe accidents at nuclear power plants on the sur-suitable for benchmarking these analyses, the NRC has con- roundi g public. The principel phenomena considered are at-ducted several series of drop-test studies of a solid steel billet mospheric transport and deposition under tirne-variant meteor-and of a near-fulkscale ernpty cask. This report contains a sum- ology, short and long-term mitigative actions and exposure path-rnary and evaluation of all steel billet testing conducted by ways, deterministic and stochastic health effects, and economic Sandia National Laboratories and Lawrence Livermore National costs. MACCS2 was developed as a generahpurpose tool appli-Laboratory. A series of finite element analyses of the billet test. cable to diverse reactor and nonreactor facilities licensed by the ing is described and benchmarked against the test data. A Nuclear Regulatory Commission or operated by the Department method to apply the benchmarked firute element model of the of Energy or the Department of Defense. The preprocessors soil and concrete pad to an analysis of a full-size storage cask available for use with the MACCS2 code are COMIDA2, is provided. In addition, an application to a " generic" full-size DOSFAC2, FGRDCF, and IDCF2. The COMIDA2 code contains cask is presented for side and end drops, and tipover events. a semidynamic food chain rnodel and generates a file of dose-NUREG/CR-6611: RESULTS OF PRESSURE LOCKING AND source mme factas M are usM W MS2 in @ a f inpskn doses. DOSFAC2, MRW, and N2 THERMAL BINDING TESTS OF GATE VALVES gewrate a Se of dose comwse factus mat am Wmd b DEWALL,K.G.; WATKINS J.C.; MCKELLAR,M.G.; et al. Idaho MACCS2 dose calculations. The preprocessors, written in National Engineering & Environmental Laboratory. May 1998. FORTRAN 77, requke a 486 or higher IBM <ompatible PC, 69pp. 9806080225. INEELEXT9800161. A3688:237. The U.S. Nuclear Regulatory Commission (NRC), Office of NUREG/CR-6614: FEASIBILITY OF HIGH FREQUENCY ACOUS-Nuclear Regulatory Research, is funding the Idaho National En- TIC IMAGING FOR INSPECTION OF CONTAINMENTS. gineering and Environmental Laboratory (INEEL) in performing BONDARYK,J.E.; CORRADO,C.N. Oak Ridge National Labora-research investigating the performance of gate valves subjected tory. GODINO.V. . August 1998. 56pp. 9810300021. to pressure locking and thermal birdng conditions. Pressure ORNLSUB97SX754V. A5626287. locking and therrnal binding are phenomena that make a closed This numerical feasibility study investigated the use of high-gate valve difficult to open. If the loads associated with pressure frequency bistatic acoustic imaging techniques for inspection of locking or thermal birdng are very high, the actuator might not inaccessible portions of the metallic pressure boundary of nu-have the capacity to open the valve. We tested a flexible-wedge clear power plant containments. High frequency vibrational gate valve and a double 4isc gate valve under pressure locking sources were used to excite elastic waves in the steel Waves and thermal binding conditions. The results show that these that reflect and scatter from surface roughness caused by deg-valves are susceptible to pressure locking; however, they are radations (e.g., corrosion) are detected and used to identify and not significantly affected by thermal binding. The results also map the steel degradation. Variables in the study included fre- i show that seat leakage affects the bonnet pressurization rate quency, flaw size, interrogation distance, and sensor incident l when the valve is subjected to thermally induced pressure lock- angle. Results for portions of steel containments embedded in ing conditions. concrete indicate that the intrinsic backscatter from degradations repmseming mickmss re&ctions from 10 to 8% me shen NUREGICR-6613 V01: CODE MANUAL FOR MACCS2. User's , thickness are sufficient to permit detection. For the embedded  ! Guide. CHANIN,D. Technadyne Engineering Consultants, Inc. steel liner of reinforced concrete containments, the thin steel YOUNG,M.L Sandia National Laboratories. May 1998. 300pp. layer and high signal losses to the concrete indicate that it is 9806150058. SAND 97-0594. A3823:001. This report describes the use of the MACCS2 code. The doc- unlikely that acoustic imaging technology can be applied to this ument is primarily a user's guide, though some model descrip- scenano. H is mcommended mat a controM experimeMal p tion information is included. MACCS2 represents a major en- gram be conducted in which sensor levels are calibrateq against hancement of its predecessor MACCs, the MELCOR Accident degradations to determine if current sensor technology can input Consequence Code System. MACCS2, distributed by govem- sufficient power into the system to provide return levels within the dynamic range of the receivers. rnent code centers since 1990, was developed to evaluate the impacts of severe accidents at nuclear power plants on the sur- NUREG/CR-6615: A SURVEY OF REPAIR PRACTICES FOR 4 rounding public. The principal phenomena considered are at- NUCLEAR POWER PLANT CONTAINMENT METALLIC PRES- I mospheric transport and deposition under time-variant meteor- SURE BOUNDARIES. OLAND,C.B.; NAUS,0.J. Oak Ridge Na-ology, short and long-term mitigative actions and exposure path- tional Laboratory. May 1998.128pp. 9806080222. ORNtJTM-ways, deterministic and stochastic health effects, and economic 13601. A3686:178. costs. No other U.S code that is publicly available at present The Nuclear Regulatory Commission has initiated a program offers all these capabilities. MACCS2 was developed as a gen- at the Oak Ridge National Laboratory to provide assistance in eral-purpose tool applicable to diverse reqctor and nonreactor their assessment of the effects of potential degradation on the facilities heensed by the Nuclear Regulatory Commission or op- structural integrity and leaktightness of rnetal containment ves-ersted by the Department of Energy or the Department of De- seis and steel liners of concrete containments in nuclear power fense. The MACCS2 package includes three primary enhance- plants. One of the program objectives is to identify repair prac-ments: (1) a more flexible emergency-response model, (2) an tices for restoring metallic containment pressure boundary com-expanded library of radionuclides, and (3) a semidynamic food ponents that have been damaged or degraded in service. This chain model. Other improvements are in the areas of phenome- report presents issues associated with inservice condition as-

I Main Cltations cnd Abstracts 33 sesstnents and continued service evaluations and ident!fies the duced two types of microstructures in both commercial and core rules and requirements for the repair and replacement of non- trolled-purity alloys: one dominated by grain boundary carbidos conforming containment pressure boundary components by and one dominated by intragranular carbides. CERT tests con-welding or rnetal removal. Discussion topics include base and ducted over a range of strain rates and at two temperatures welding materials, welding procedure and performance qualifica- showed that in all samples, IGSCC was the dominant failure tions, inspection techniques, testing rnethods, acceptance cri- mode. For both the commercial alloy and the controlled 9urity teria, and documentation requirements necessary for making re- alloys, the microstructure with grain boundary carbides showed pairs and replacements so that the plant can be retumed to a delayed crack initiation and shallower crack depths than did the safe operating condition. intragranular carbide microstructure under all experimental con-NUREG/CR-6616: RISK COMPARISON OF SCHEDULING PRE- ditions, indcating that a grain boundary carbide microstructure VENTIVE MAINTENANCE DURING SHUTDOWN VS. DURING is more resistant to IGSCC than an intragranular carbide micro-POWER OPERATION FOR PWRS. YANG,J.W.; CHU,T-L; structure. Observations support both the film rupture / slip dissolu-MARTINEZ-GURIDI Brookhaven National Laboratory. December tion mechanism and enhanced localized plasticity. Crack growth 1998.200pp.9901270229. BNL-NUREG-52549. A6623:001, rates increased with increasing strain rate according to a power Preventive maintenance is performed to assure the reliability law relation with a strain rate exponent between 0.40 and 0.64. of safety-system components. This preventive maintenance typi- However, crack growth rate measured in m/ unit strain de-cally is planned and scheduled ahead of time which is suitable creased with increasing strain rate indicating an effect of either from multiple considerations including risk impact, cost, avail- the environment or creep. The temperature dependence of the ability of personnel, and operating reliability / life of the compo- crack growth rate was consistent with the literature, , nent. Since such maintenance can take place during power op-eration or different periods in a plant shutdown, including cold NUREG/GR-0017: DATING OF LIQUEFACT:ON IN THE NEW i shutdown and refueling periods, the relative risk-impact of a par. MADRID SEISMIC ZONE AND IMPLICATIONS FOR EARTH- l ticular maintenance during these times is an important input in OUAKE HAZARD. TUTTL E.M.P.; LAFFERTY,R.H.; I the overall decision on scheduling. In this report, we use prob- SCHWEIG E.S. Maryland, Univ of, College Park, MD. Sep-abilistic risk assessment (PRA)-based methods and apply them tember 1998,122pp. 9812240089. A6314:163. to a specific pressurized water reactor (PWR) plant to address Resutts of paleoseismological investigations of earthquake-in-the needs for evaluation and the insights obtained in scheduling duced liquefaction features at nine sites indicate that significant preventive maintenance to control risk impacts. To conduct such earthquakes struck the New Madrid region in A.D. 90r A 100 a comparison, we developed PRA models comparable for the yr and A.D. 15301 130 yr. This finding is consistent with other cold shutdown and refueling periods to that for the power oper- paleoseismological studies in the region. The A.D. 900 event ation period. Both the core damage frequency and consequence was probably similar to the 1811 1812 earthquake sequence. irripacts are analyzed, i.e., risk measures calculated using PRA Less is known about the A.D.1530 event, but it is likely to have Level 1,2, and 3 models. The results provide insights on sched- been very large as well. These data suggest that very large uling preventive maintenance that may be generally applicable earthquakes occurred in the New Madrid seismic zone every across PWR plants, and also identify plant-specific evaluations 200 to 900 yr during the past 1200 yr. In addition, liquefaction useful for specific considerations. In general, risk-comparison in- features along the Black and Current Rivers suggest an earth-sights can play an integral part in distributing maintenance quake source, possibly associated with the Commerce geo-across different periods to control or reduce any of its adverse physical lineament, in the western Lowlands. Much work re-effects. Based on risk characteristics, plants can be grouped to rnains to refine estimates of the timing, magnitude, and source defi:w guidance for maintenance scheduling, and to review the areas of prehistoric earthquakes in the New Madrid region and schedules being used. to determine if the recent high rate of seismcity reflects the NUREG/CR-6617: THE PRICE-ANDERSON ACT - CROSSING 30Vterm hazard. THE BRIDGE TO THE NEXT CENTURY: A REPORT TO CON-GRESS. BAILEY,P.; BLAKE.K.; BROWN,M.; et al. ICF, Inc. Oc- NUREG/lA-0024: APPLICATION OF RELAPS/ MOD 3.1 CODE TO tober 1998.179pp. 9810260052. A5550:248. THE LOFT TEST L3-6. PYLEV,S.S.; ROGINSKAGA,V.L Rus-This report fulfills the mandate of Subsection 170p. of the sia. February 1998. 6Gpp. 9802100124. A2059:287 Atomic Energy Act of 1954, as amended, which requires that A calculation of LOFT Experiment L3-6, a small-break equiva-the Commission submit to the Congress by August 1,1998, a lent to a 4-inch diameter rupture in the cold leg of a four-loop detailed report on the need for continuation or modification of commercial pressurized water reactor, has been performed to Section 170 of the Act, the Price-Anderson provisions. Part 1 help validate RELAP5/ MOD 3.1 for this application. The version presents an overview of the Price-Anderson system. Part 2 ex- of the code to be used is SCDAP/RELAP5/ MOD 3.1.8do. Three amines the issues that the Commission is required by statute to calculations were carried out in order to study the sensitivity to study (i.e., condition of the nuclear industry, state of knowiedge change of the break nozzle superheated discharge coeffcient. i of nuclear safety, and availability of private insurance). Part 3 Conducted comparative analysis of the LOFT L3-6 experiment I covers other issues of interest and importance to the Congress shows on the whole a reasonable agreement between cal- ) and to the public, such as proof of causation and international culated and measured data. Some discreoancies in the system ' agreements relevant to Price-Anderson. Part 4 of the report con- pressure do not distort a picture of the transient. tains conclusions and recommendations. Part 5 is the list of ret-erences. Appendix A is an evaluation of the affordability of ccr- NUREG/lA-0025: RELAP5/ MOD 3 SUBCOOLED BOILING MODEL tain Price-Arderson assessments. ASSESSMENT. DEVKIN,A.S.; PODOSENOV,A.S. Russian Re-NUREG/GR-0016: THE ROLE OF TIME-DEPENDENT DEFOR- search Centor (Kurchatov Institute). May 1998. 83pp. MATION IN INTERGRANULAR CRACK INITIATION OF ALLOY 9805200009. A3468:097. 600 STEAM GENERATOR TUBING MATERIAL. WAS.G.S.; This report presents the assessment of the RELAPS/ Mod 3 LIAN.K. Michigan, Univ. of, Ann Arbor, MI. March 1998. 41pp. (5m5 version) code subcooled boiling process model, which is 9804200284. A3031:101, based on a variety of experiments. The accuracy of the rnodel intergranular stress corrosion cracking (IGSCC) of two com. is confirmed for a wide range of regime parameters for the case ( mercial alloy 600 conditions and controlled-purity Ni-18Cr-9Fe of uniform heating along the channel. The condensation rate is l alloys were investigated using constant extension rate tensile rather underpredcted, which may lead to considerable errors ,n i (CERT) tests in primary water with 1 bar hydrogen overpressure void fraction behavior prediction in subcooled boiling regimes for at 360 degrees C and 320 degrees C. Heat treatments pro- nonuniformly or unheated channels.

34 Main Citation 3 and Abstracts  ! NUREG/lA-0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING dows 95 environment. The "dinstis* script with proper modifca-LOFT LARGE BREAK LOCA TEST,LP-02-6. CHOi,T.S.; tions was used to extract the source for conversion and then LEE,J.H.; PARK,B.S.; et al. Korea, Democrate Peoples Reput> several modifcations were done for installation on PC. Compila-lic (North Korea). August 1998.118pp. 9809210277, A5126 230. tion and linking has been done using a MAKE utility and genera-The LOFT expenment LP-02-6 was simulated using the tion of TPFH2O has been done also. Four test cases were used RELAP5/ MOD 32 code to assess its capability to predet the to verify the conversion and installaton of RELAPS/ MOD 32 on thermal-hydraulic phenomena in LBLOCA of the PWR. The re- PC. CPU time benchmark calculation was also done. The re-actor vessel was modeled with two core channels and split suits show that the use of PC version could be an option for tho downcomer for a base calculation. The resutts of the base cal- users based on the availability of hardware and the speed of the culation show that the code can not predict the early bottom-up CPU. quenching whch is a distinguished phenomenon of the experi- NUREG/lA-0143: ASSESSMENT OF RELAPS/ MOD 3.2 WITH ment LP-02-6, mainly due to the deficiency of break flow model* THE LSTF EXPERIMENT SIMULATING A LOSS OF RESID-The discharge coefficient sensitivity study was performed to UAL HEAT REMOVAL EVENT DURING MID-LOOP OPER-show that the calculated subcooled break flow which might sig- ATION. SEUL,K.W.; BANG,Y.S.; LEE,S.; et al. Korea institute ot nificantly Nuclear Safety. August 1998.134pp. 98092103n4. A5128:169. the coeffe, affect ient. Morethe earlymodeling detailed bottom-up quenching of the cross flow is in dependent the o The potential for the RELAP5/ MOD 32 code was assessed for split downcomer was performed, but, resulted in little improve- the loss of residual heat reinoval (RHR) event during the mid-rnent on the predictability of bottom-up quenching. Additional loop operation. The predictability of major thermal hydraulic phe-calculation using the RELAPS/ MOD 3.1 instead of RELAPS/ nomena was evaluated for the long term transient. The results MOD 3.2 showed that there is no large difference between the of two typical cases, cold leg opening (CLO) case with water-versions in the simulation of LBLOCA. filled steam generators (SGs) and pressurizer opening (PRO) NUREG/lA-0140: DEVELOPMENT ASSESSMENT OF RELAP5/ case with emptied SGs were compared with experimental data MOD 3.1 WITH SEPARATE-EFFECT AND INTEGRAL TEST conducted at ROSA-IV/LSTF in Japan. It was found that the EXPERIMENTS: MODEL CHANGES AND OPTIONS. code was capable of simulating the system responses to the ANALYTIS G.T, Paul Scherrer Institute. October 1998. 46pp. loss-of-RHR event during the reduced inventory operation. The 9811050306. A5695:026. thermal hydraulic transport process including noncondensable A summary of modifcations and options introduced in gas behavior was reasonably predcted with an appropriate time RELAPS/ MOD 3.1 (R5M3.1) is presented and is shown that the setup and CPU time. Overall, the code well predcted the major predcting capabilities of the modified version of the code are thermal hydraulic phenomena during the transient. greatly improved, while the general philosophy we followed in NUREG/lA-0144: ASSES 7 MENT OF RELAPS/ MOD 32 WITH urrving at these modifications as also outlined. These changes THE SEMISCALE NATUHAL CIRCULATION EXPERIMENT, S-which are the same ones we implemented in the past version NC-8B. BANG,Y.S.; SEUL,K.W.; LEE,S.; et al. Korea Institute of

7) of the code, include 2 different heat transfer packages (one Nuclear Safety. August 1998. 200pp. 9809210310. A5136:024.

of them activated dunng reflooding), modification of the low The predictability of RELAPS/ MOD 32 code is assessed for mass-flux Groeneveld CHF look-up table and of the dispersed the natural circulation induced by small break loss of coolant ac-flow interfacial area (and shear) as well as of the criterion for , cident in the pressurized water reactor by using Semiscale ex-transition into and out from this regime, almost complete elimi- periment S-NC-88. The Semiscale Mod-2A facility is modeled nation of the under-relaxation schemes of the interfacial closure as a base case, by using single core channel model. The base coefficients etc. The modified R5M3.1 code is assessed against case calculation is executed, the result is compared with the ex-a number of separate-effect and integral test experiments and periment data and code predictability on the important thermal-in contrast to the frozen version, is shown to result in physically hydraulic phenomena is discussed. Sensitivity calculations are sound predictions which are much closer to the measurements, attempted to figure out the problems in base case predction, while almost all the predcted variables are free of unphysical and to find out the effects of two core channel model and spunous oscillations. The modifications introduced solve a num- ECCMIX component model on the improvement code predct-ber of problems associated with the frozen version of the code ability. The important thermal-hydraulic phenomena include sys-and result in a version which can be confidently used for both tem depressurization, break flow in saturated and stratified con-SB-LOCA and LB-LOCA analyses, ditions, natural circulation in twophase mode and reflux mode, NUREG/tA-0141: RESULT OF BETHSY TEST 9.1.B USING loop seal behavior at crossover legs, and accumulator injection RELAPS/ MOD 3. PETELIN,S.; MAVKO.B.; GORTNAR,0.; et al. behavior. The base calculation shows the RELAP5/ MOD 3.2 can Josef Sefan institute. August 1998. 73pp. 9809210298. predict the overall thermal-hydraulic behavior such as system A5143:147. depressurization, with the exception of underprediction of satu-RELAPS computer code was used to simulate an experiment rated break flow, deviation of loop seal behavior, and resultant designated 9.1.b. (2" Cold Leg Break without HPSI and with De- discrepancy in core thermal response. Two core channel model layed Ultimate Procedure) performed on BETHSY integral test can improve the predictability on loop seal behavior. ECCMIX facility. This test is characterized as beyond design transients component can improve an early accumulator injection behavior scenarios with unavailability of some safety and protection sys- and core thermal response. However, discontinuous accumu-tems. The calculations which have been completed using the lator injection is one of the problems in two core channel model i computer Sun Sparcstation 20 aim to evidence the difference calculation. To resolve the accumulator injection problem, the l between experimental and computed data. Generally, an agree- extensive modeling study and/or code model irnprovement are , ment of major transient trends is shown to be obtained in the i, ceded.  !

   "I*"I*U "'                                                                NUREG/lA-0146: REMP5 ASSESSMENT AGAINST PACTEL NUREQ/lA-0142: INSTALLATION OF RELAP5/ MOD 3.2 ON                            EXPERIMENTAL DATA (REVISION 1). PARZER,1.; MAVKO,B.;

80486 AND PENTIUM BASED PERSONAL COMPUTERS. PETELIN S. Josef Sefan institute. November 1998. 104pp. CHO.C.S.; LEE,G.W.; PARK,J.Y.; et al. . October 1998. 88pp. 9812040132. A6022:158. 9811050313. A5695:100. The results of the pre-test and post-test calculations of OECD REMPS/ MOD 3.2 has been installed on Intel Chip based per- International Standard Problem no.33 (ISP-33) are presented. sonal computers at KNFC. This report is to present the installa- The frozen version of RELAP5 code version MOD 2 and the lat-tion procedures and test resutts for CPU time comparison and est released version MOD 3.2 have been assessed against ex-installation verification, installation of RELAPS/ MOD 32 on PC perimental data from PACTEL facility. Generally, prodctions has been done using Lahey Fortran F77L3 compiler under Wirr were in good agreement with experimental data, except in the l.

1 Main Citations and Abstracts 35 l rnost interesting part of the transient only MOD 32 was able to dicts the heat transfer coefficients, but that the alternative model l follow periodic hot leg loop seal clearance phenomena and prb of RELAP5/ MOD 3.2 over-predicts them throughout the con. I mary pressure oscillatons closely. densing tube. NUREG/lA-0146: IMPLEMENTATION AND ASSESSMENT OF NUREG/lA-0148: ASSESSMENT OF RELAPS/ MOD 3.1 USING IMPROVED MODELS AND OPTIONS IN TRAC-BF1. LSTF TEN-PERCENT MAIN STEAM-LINE-BREAK TEST RUN ANALYTIS,G.T. Paul Scherrer institute. October 1998. 48pp. SB-SL-01. OH,J.G.; LEE,H.D.; JEE,K.K.; et al. Korea Power En-9811020050. A5650:001, gineering Co., Inc. September 1998. 125pp. 9810160058. A summary of modifications and options introduced in TRAC- AS417f.09. BF1 is presented and is shown that the predicting capabilities Results produced by the RELAP5/ MOD 3.1 computer code of the modified version of the code are greatly improved. These were compared with the experimental data from JAERI's LSTF changes include the introduction of a different heat transfer Test Run SB-SL-01 for a 10% main steam line break transient package during reflooding, the implementation of a simple sin- in a pressurized water reactor. The code simulation for the base gle-phase limit procedure for forcing the two phases to acquire case included a total of 189 fluid control volumes and 199 flow the same velocity if one phase disappears, a close assessment junctions to model the transient two-phase phenomena. Also, a of the annular flow interfacial shear correlation, implementation total of 180 heat slabs were used to model the system heat of a simple radiation model which seems to alleviate some nu- transfer. The code predictions of the experimental results are merical-oscillation problems induced by the existing highly com generally satisfactory for the trends of key parameters. Sensi-  ; plex model Furthermore, different options were introduced and tivity studies performed for the break discharge coefficient, the  ! tested like upwinding some terms of the momentum equations, separator drain line loss coefficient, and the number of steam I the second upwind scheme for the convective terms of the pha- generator nodes did not reveal any strong dependencies. Never-  ! sic momentum equations and the implementation and assess- theless, optimal values of these parameters that led to the low-ment of a cornpletely different annular flow interiacial shear cor- est overall statistical error were obtained, and these values were relation. Modified TRAC-BF1 is assessed against some bottom- subsequently used in the " Base Case' analysis, flooding separate-effect experiments, a " benchmark" top flood-ing simulation as well as against the TLTA test No. 6423. In the NUREG/lA-0149: ASSESSMENT OF RELAPS/ MOD 32-NPA3.4 process of this task, the different options are assessed and dis. AGAINST A TRANSIENT OF HIGH NUCLEAR FLUX VARI-cussed and is shown that the predictions of the modified code ATION REACTOR TRIP, NATURAL CIRCULATION AND THE are physically sound and close to the measurements, while al. START OF A MAIN PUMP IN THE VANDELLOS ll NUCLEAR most all the predicted variable are free of unphysical spurious POWER PLANT. LLOPIS.C.; MARTIN,M. Spain, Govt. of. No. oscillations. The modifications introduced solve a number of vember 1998. 61pp. 9812160015. A6218270. problems associated with the frozen version of the code and re- This report is an assessment calculation with RELAPS/ sult in a version which can be confidently used for LBLOCA MOD 32 - NPA3.4 against a transient that took place in analyses. Vandellos 11 Nuclear Power Plant on December 2,1991. This w a pa e pan % coNon 4 h M Assessment NUREG/lA 0147: ASSESSMENT OF RELAPS/ MOD 32 FOR and Maintenance Program (CAMP). Vandellos 11 is a member of STEAM CONDENSATION EXPERIMENTS IN THE PRESENCE UNIDAD ELECTRICA, S.A. (UNESA). Vandellos 11 is a 2775 OF NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. Mwt three loop Westinghouse PWR owned by ENDESA, SA PARK,H.S.; NO,H.C. Korea Advanced institute of Science and (72%) and IBERDROLA (28%); it is on the Mediterranean coast. Technology. BANG,Y.S.; et al. Korea Institute of Nuclear Safety, near Tarragona Spain). The commercial operation began in September 1998.108pp. 9810160173. AS417:001. 1988. The simulation has been running on a Pentium 11266MHz This report deals With the application of RELAPS/ MOD 32 to PC under Windows NT. This version has been tested and re-condensation experiments in the presence, of noncondensable sults are exactly the same as those obtained with a DECStation gases in a vertical tubo of Passive Containment Cooling Sys- 500/200* tem. When steam-noncondensable gas mixture was injected into the vertical tube, steam was condensed on the inner surface of NUREG/tA-0150: STUDY OF TRANSIENTS RELATED TO the condensing tube but the noncondensable gas greatly inhib- AMSAC ACTUATION. SENSITIVITY ANALYSIS. POSADA,J.M.; ited the condensation of the f, team. As the scattering of previous REVETOS,F.; SANCHEZ-BAPTIST: et at Spain, Govt. of. No-experimental data was large, the present experimental appa- vember 1998. 60pp. 9812160018. A6218217. ratus was set up to get reliable data on the condensation heat The Asociacion Nuclear Asco (ANA) has prepared a model of transfer coefficient of the steanw-(mde sable gas mixture in Asco NPP using RELAPS/ MOD 3.2. This model, which includes a vertical tube. The experimental results show that the con- thermahhydraulics, kinetics and protection and control systems, densation heat transfer coefficient increases as the inlet steam- has been qualified in previous calculations of several actual air rnixture flow rate increases, the inlet air mass fraction de- plant transients. Asco NPP is a two unit station of three loop creases, and the inlet saturated steam temperature decreases. Pressurized Water Reactor (PWR) of Westinghouse design op. There are two wall film condensation models, the default model ersted by ANA. ANA is a Spanish utility that contributes to the and the alternative model, in RELAPS/ MOD 3.2. After a con- Code Application and Maintenance Project (CAMP) as a mem-densation database was constructed, two models were as- ber of UNIDAD ELECTRIC A S.A. (UNESA). This report sum-sessed directly with the data of the database. The experimental marizes the results obtained with Asco NPP model for a loss of apparatus was also modeled with RELAP5/ MOD 32, and simula- normal feedwater ATWS event and presents a sensivitity anah tions were performed for several sub-tests to be cornpared with ysis to kinetic parameters for the same transient. The phenome-the experimental results. The simulation results show that in an nology prediction has been useful from the operation and safety overall sense the default model of RELAPS/ MOD 3.2 under-pre- point of view.

1 f

   +

( 1 1 l 2

E1 Main Citations and Abstracts 35A NUREGISR 0060: ADVISORY COMMITTEE ON NU. CLEAR WASTE - 1998 STRATEGIC PtAN AND PRI- , ORITY ISSUES AND ACTIVITIES.

• ACNW - Advisory I Committee on Nuclear Weste. March 1998.12pp. .

9804200219. A3031:001. l The Advisory Committee on Nuclear Waste (ACNW) has developed a strategic plan that establishes a frame-work to guide it in providing independent and timely tech-nical advice to the Nuclear Regulatory Commission on nuclear waste disposal and management issues. The plan includes near-term priority issues the Committee will con-sider in 1998, as well as longer term issues the Committee plans to consider in 1999 and beyond. The ACNW's strate-  ! gic plan is anchored to the NRC's Strategic Plan for Fiscal years 1997 - 2002 and supports the mission, vision, and relevant goals, strategies, and substrategies identified by the agency. NUREGIBR4117N97-4: NMSS LICENSEE NEWSLETTTER.

• Office of Nuclear Material Safety and Safe-guards. February 1998.12pp. 9804200291. A3031:143.

This newsletter contains articles that discuss recent re-gulatory issues and provide administrative information. It I includes descriptions of recent Federaf Register notices, l generic communications, significant enforcement actions, and significant operational events. NUREGIBR 0164, Rev. 3: NRC REGULATOR OF NUCLEAR SAFETY.

• Office of Public Affairs.

December 1998. 28pp. 9902090259. A6765:315. This brochure describes the activities regulated by the Nuclear Regulatory Commission, using color photo-graphs and diagrams. It includes an explanation of the inr>er workings of nuclear power plants and describes NRC's inspection program. The brochure also explains how radioactive materials are used in industry and medicine, and offers a brief overview of NRC's responsibilities in emergency planning, transportation, and radioactive waste. NUREGIBR-0240: REPORTING SAFETY CONCERNS TO THE NRC.

• Office of Public Affairs. October 1998.16pp.

9810230170. A5511:274. This brochure provides information on how nuclear workers such as yourself- can report safety concerns to the NRC, what degree of protection can be afforded to a worker's identity, and the NRC process for handling a worker's allegation of discrimination that may result from reprisals by licensees, their contractors, or subcontractors.

l Main Citations and Abstracts 35B NUREG/BR 0249: THE ATOMIC SAFETY AND LICEN. j SING BOARD PANEL

• Office of Public Affairs.

February 1998. 6pp. 9803190405. A2670:015. Through the Atomic Energy Act, Congress made it possible for the public to get a full and fair hearing on civilian nuclear matters. IndMduals who are directly affected by a licensing action invoMng a facility producing or utilizing nuclear materials may participate in a formal hear-ing, on the record, before independent judges on the Atomic Safety and Licensing Board Panel. Hearings, routinely involve difficult interrelated questions, often at the cutting edge of science and technology, confronting highly technical and scientific theories, opinions, and research findings. In addition, NRC hearings air local concerns about the consequences of severe accidents and continue the national debate over the role nuclear power should play in meeting the nation's energy needs. NUREGIBR 0252: USER'S GUIDE TO NRC-PUBLISHED PHYSICAL PROTECTION DOCUMENTS.

• Office of Nuclear Material Safety and Safe-guards. November 1998. 58pp. 9811230217. A5946:079.

This report is a compilation of physical protection documents published by the U.S. Nuclear Regulatory Commission. It is intended to serve as a user's gu'de to assist in conducting information searches about physical prote:: tion subjects. Given for each document is a reference number, title, publication date, and abstract to further aid in identifying available physical protection information of interest. I i

g , r i l t i t Secondary Report Number index L This index lists, in alphabetical order, the performing organization-issued report codes for the l NRC contractor and international agreement reports in this compilation. Each code is cross-referenced to the NUREG number for the report and to the 10-digit NRC Document Control System accession number. i-ECONDARY REPORT NUGAGER REPORT NUGASER SECONDARY REPORT NUGASER REPORT NUMBER ANL-00/20 NUR6@CR4621 ORNWM-11668 ANL47/17 NUREG/CR-6601 V04 N1 NURE@CR4676 ORNL/TM-13144 NUREG/CR4408 ANL47/18 . NUREGCR4683 ORNWM-13204 NUREG/CR4463 ANL47/3 - NUME@CR4611 V02 ORNL/TE13264 NUREGCR4479 ANL40/18 NURE@CR-4067 V26 ORNL/TM-13396 NUREG/CR 0636 ANL40/6 NUREG/CR-4067 V24 ORNL/TM-13408 NUREGCR4637

  ' ANL-90/7                  NUREG/CR4611 V03           ORNL/TM-13446 SMI-2100                                                                       NUREG/CR4662 NUREGICR4640               ORNL/TM-13494              NUREG/CR4677 l     BNL NUREG42471           NUREG/CR4369 V01           ORNUTE13648 8NL NUREG42471                                                                 NUREG/CR4000 NUREGICR4369 V02           ORNL/TM-13649              NUREG/CR4001 BNLWOREG42474            NUREGICR4384               ORNL/TM-18664 ONL-NUREG42400            NUREG/CR4418                                          NUREG/CR4698 BNLWUREG42001                                        ORNL/TE13666 ~             NUREG/CR4699 NUREG/CR4472               ORNUTM-13001 ONL-NUREG42613            NURE@CR4602                                           NUREG/CR4616 ONLWUREG 62616            NURE@CR-6609 ORNL/TM-13807              NUAEG/CR4342 BNL NUREG42618            NUREGICR4624               ORNL/TM 13674              NUREGICR4670 BNL NUREG42630            NUREGICR4664               ORNL/TM4603                NURE@CR-4219 V13 N2 ONL-NUREG42632            NURE@CR4669                ORNL/TM-9693               NUREGICR-4219 V14 N1 DNL NUREG42638            NUREGICR-0679              ORNLSUS96-SP638            NUREG/CR-0646 BNLWUREG42649             NUREGICR4616               ORNLSU897SX764V            NUREGICR4614 EPA-402R47 016            NUREG-1676                 PNNL-10496                 NUREGICR4210 801 EPA 008 K-96402           NUREG-1634                 PNNL 11143                 NUREG/CR4471 V01 EUR 18773                 NUREGCR-0671 V01           PNNL-11171                 NUREG/CR4694

! EUR 167T3 NUREGICR4671 V02 PNNL 11408 NUREG/CR4377 l EUR 18774 NUREGICR4666 V01 PNNL-11613 NUREGICR4634 V03 EUR 16774 NUREGICR4666 V02 PNNL-11613 NUREGICR4634 V02 EUR 18776 NUREGCR4646 V01 PNNL-11761 NUREGICR4600 EUR 16776 NUREG/CR4646 V02 PNNL-11797 NUREGCR4006 FACA NUREG 1000 R01 PNNL-11898 NUREGCR-6606 FIN A-2244 NUMEGICR4372 PSU/ME 08-7321 NUREG/CR-6634 FIN F-0847 NUREGICR-6372 PU/NE-08-7 NUREG/CR4de8 FIN G4107 - NURE@CR4668 SAND 00-7117 NUREGICR4671 FIN W4308 NUREG/CR4669 SAND 93-2301 NUREG/CR4131 INEELEX79700006 NUREGICR4268 V01 SAND 964099 NURE@CR4617 INEELEXT9700096 NUREG/CR4268 V02 SAND 07-0604 NUREG/CR4613 V01 INEELEXT9700006 NUREGICR4268 V03 SAND 974604 NUREG/CR4613 V02 l INEELEXT9700006 NURE@CR4208 V04 SAND 07-0887 NUREG/CR4476 ! INEELEXT9700740 NUREG/CR4600 VOI SAND 97-2322 NUREGCR4665 V01 INEELEXT9700687 NUREGICR4496 SAND 97-2322 NUREG/CR4666 V02 INEELEX79700026 ' NUREGICR4689 SAND 97-2398 NUREGCR4119 V01 R1 INEELEXT9701327 NUREG/CR4406 SAND 07-2308 NUREG/CR4119 V02 R1 INEELEXT9701328 NUREGICR4497 SAND 07-2632 NUREGICR4600 INEELEXT9000161 NUREGICR4611 SAND 97-2600 NUREGICR4646 V01 INEL 06/0422 NUREG/CR4160 V01 R1 SAND 07-2600 NUREG/CR4646 V02 INEL-08/0422 NUREGICR4160 V02 R1 SAND 07-3170 NUREG/CR-6412 INEL-08/0422 NUREGICR4160 V03 R1 SAND 98-0119 NUREG/CR-0671 V01 INEL 06/0422 NURE@CR4160 V04 R1 SAND 98 0119 NURE@CR4671 V02 l INEL-06/0422 NUREG/CR4160 V06 R1 SAND 98 0272 NUREGICR-8004 t MCS 970001 NUREGICR-6661 SAND 964419 NUREG/CR4003 i ORNL4004 NUREG/CR4447 UCID40674 NUREGICR 4664 V01 R2 i ORNL/NOAC-232 NURE@CR-4674 V26 UCRL-129211 NUREG/CR-8008 i ORNL/NOAC 232 NUREGICR-4674 V26 UCRL-ID-120020 NUREG/CR-8006 l ORNL/TM-11608 NUREGCR4601 V08 N1 URCL-ID-130438 NUREGICR-6602 l 37

l l

 )

B I

Personal Author index.

                  .                                                                                                                                           i This index lists the personal authors of NRC staff, contractor, and international agreement reports in al the report (s)phabetical       prepared by theorder. author.Each       name If further         is followed information                  by the is needed,          NUREG refer   to the main    numbercita- and the title o 4

tion by the NUREG number.

   ==== auiaT,E.W.                                                                mAeWA.C.S.

NURE41507: MINIMUM DETECTABLE CONCENTRATIONS WITH NURE41662 801 DR FC: FIRE BARRIER PENETRATION SEALS N TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- NUCLEAR POWER PLANTS. Drelt Report For Comment. IN RADIOLOGICAL SUR- BAIDfTIARIA VEY 8 CANNING NUREGOR4611 V02: STEAM GENERATOR TUBE INTEGRITY PRO-Asa as n NUR GCR4476: RESOLUTION OF THE DIRECT CONTAINMENT 1 kl[E l'NTEGRITY PRO-GRAM.8emiennual Report. October 1908 March 1997. HEATNG ISSUE FOR COMBUSTON ENGINEERING PLANTS & as a nanna n *.

NEUTRON EXPOSURE PARAMETERS FOR CAP-SULE 10.06 IN THE HEAVY-8ECTON STEEL IRRADIATION PRO.

NURE41022 R01: EVENT REPORTING GUIDEUNES 10 CFR 60.72 GRAM TENTH IRRADIATION SERIES. AND 80.73. NUREGCR4001: NEUTRON EXPOSURE PARAMETERS FOR THE DO61 METRY CAPSULE N THE HEAVY 8ECTION STEEL 1RF.ADIA-74: PALEOSEl8MIC STUDIES N THE SOUTHEASTERN

• UNITED STATE 8 AND NEW ENGLAND BAfsG,Y.8.

NUREGilA4130: ASSESSMENT OF RELAP6AsODS.2 USING LOFT NUMEGim41'40: DEVELOPMENT ASSES 8 MENT OF RELAP6 MODS.1NU WITH SEPARATE-EFFECT AND NTEGRAL TEST EXPERIMENTS: IELAP6 MODS.2 ON 00400 AND PENTIUM BASED PE L COMPUTERS MODEL NUREGilA4143: ASSE NUREGilA 148: NTA AND A88E88 MENT OF M- NT OF RELAPSMODS.2 WITH THE LSTF PROVED MODEL8 AND OPTIONS IN TRAC 8F1. EXPERIMENT SIMULATNG A LOSS OF RESIDUAL HEAT REMOVAL EVENT DURING OOP OPERATION. NUREGilA4144: 88 MENT OF RELAP6 MOD 3.2 WITH THE

EVALUATION OF ULTRA 8ONIC INSPECTION TECH- "^

NIQUE8 FOR COARSE-GRAINED MATERIALS. NU 7 FR TEAM CONDENSATION EXPERIMENTS N THE PRESENCE OF PCCS. NURE CHARACTENZATION OF RETARDATON MECHA- NUNONCONDEN84g G 4e IN A V RT N0 LSTF M MA N E-8R TEM @ A80004,Y* EA8E A A-NUREGllA4148: ASSES 8 MENT OF RELAP6 MOD 3.1 USING LSTF NUREGCR4670: APPUCATION OF THE NC8A HA8ANERO TOOL TEN PERCENT MAN STEAM 4lNE-BREAK TEST RUN SS4L41. FOR COLLABORATION ON STRUCTURAL WTEGRITY ASSESS-ApD870LAs08.8. NUREb4662: MAROLE HILL ANNEAUNG DEMONSTRATION NUREGCR4644: METHODOLOGY FOR ANALYZWG PRECUR8 ORS EVALUATION. EAR E-INITIATED AND FIRE-INITIATED ACCIDENT SE-

                    "                                                            SEMMAC.

NUREGCR4400: TECHNCAL ASSISTANCE IN REVIEW OF 800F'E GCR4372: EXPERIMENT 8 ON INTERACTIONS BETWEEN ZlR- NLflE :1 N TYAN PH IN THE AP. CONIUH CONTAINING MELT AND WATER. 800 REACTOR. ATWOOO.C.L SELA880 ERA NUREGCR4408. EVALUATON OF LOS8 OF OFFSITE POWER NUREG 1834: 1997 LOST SOURCE EXERCISE.An Emeroise Of Radlo-EVENTS AT NUCLEAR POWER PLANTS: 1000 1998. logical Response Through e .. ~. And Coonsnanon Of AIARIA,00.A. Looni,80sde And Federal Resources Under The Nabonel Conengency Plan. NURE41821 DRFT FC: TECHNICAL REVIEW OF RISK-INFORMED. PERFORMANCE 8ASED METHODS FOR NUCLEAR POWER PLANT SELL 88AJ. FIRE PROTECTION ANALYSES.Dren Report For Commert NUREGCR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE BAhas amat V. NU -487 V28 E O ENTIAL SEVERE CORE NUREGCR4400: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- DAMAGE ACCIDENTS:1997. A Status Report. CULATION TESTS IN THE PUMA FACluTY. SA08ETT,8. NUREGCR4634 V02: FRAPOON-3: A COMPUTER CODE FOR THE NUREG 1688 V03: CON 800 DATED GUIDANCE ABOUT MATERIALS CALCULATION OF STEADY 4 TATE. THERMAL-MECHANICAL BE-UCENSE8 a r m For Sealed Source And Devlos Evoluenon And HAVOR OF OXIDE FUEL FOR Hi BURNUP. Registremon. Final Report. NUREGCR4634 V03: F 3: NTE L ASSESSMENT. SAILEY 0.H. BERfdREUTER,D.L. NURI:#41908 V11 DR FC: CON 80UDATED GUIDANCE ABOUT MATE- NUREGOR4006: NVESTIGATION OF TECHNOUES FOR THE DE-MALS UCENBE8.7. _. _.1 - Guldence About apar* Uoenese VELOPMENT OF SEISMIC DESIGN BASIS USING THE PROS-Of Braa,emanpa Droit Report For Commert A88USTIC SEISMIC HAZARD ANALYSIS. BAE.EY,P. BERT00A100.ht.L. NUREOCR4017: THE PRCE-ANDER8ON ACT CRO881NG THE NUREGOR4498: SINGLE-PHA8E AND TWO-PHASE NATURAL CIR. BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. CULATION TESTS N THE PUMA FACluTY. 39

40 Personal Author index l BEVERLY,D.D. NUREG/CR4364: HUMAN PERFORMANCE IN RADIOLOGICAL SUR- ) ! NUREG/CR4501: ADVANCED INSTRUMENTATION AND MAINTE- VEY SCANNING. l I NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. I' ! BRYSON,J.W. BEYER.C.E. NUREGCR4552: MARBLE HILL ANNEALING DEMONSTRATION l NUREGCR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE EVALUATION. ! CALCULATION OF STEADY-STATE, THERMAL-MECHANICAL BE- ! HAVIOR OF OXIDE FUEL RODS FOR HIGH BURNUP. BUDNITZ,R.J. l NUREG/CR4534 V03: FRAPCON-3: INTEGRAL ASSESSMENT. NUREGCR4544: METHODOLOGY FOR ANALYZING PRECURSORS TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE-BEZLER.P. QUENCES. l NUREG/CR4559: LARGE SCALE VIBRATION TESTS OF MAIN STEAM AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND BURGESS,M. ENERGY-ABSORBING SUPPORTS. NUREG-1556 V03: CONSOLIDATED GUIDANCE ABOUT MATERIALS UCENSES.Applicatons For Sealed Source And Device Evaluston And BHATTACHARYA.B. Registration. Final Report. NUREG/CR4546: A DAMAGE MECHANICS BASED APPROACH TO STRUCTURAL DETERIORATION AND RELIABluTY. CADA,Q.F. NUREG/CR-5549: ENVIRONMENTAL ASSESSMENT RENEWAL OF BlXLER N.E. MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-NUREGCR4131: VICTORIA 2.0 A MECHANISTIC MODEL FOR GIONAL SERVICE FACluTY, WAMPUM, PENNSYLVANIA. 8voNt1CUDE BEHAVIOR IN A NUCLEAR REACTOR COOLANT SYSTEM UNDER SEVERE ACCIDENT CONDITONS. CAMPE,K.M. NUREG-1632: EVALUATION OF AP600 CONTAINMENT THERMAL-HY. UR G"/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. CAMPER,L NUREG-1556 V08: CONSOUDATED GUIDANCE ABOUT MATERIALS NSEES%arnWe Guence Exempt Dshn W U  : PERFORMANCE TESTING OF PASSfVE censesh Report. AUTOCATALYTIC RECOMBINERS. CAMPER,LW. RI 5549: ENVIRONMENTAL ASSESSMENT RENEWAL OF NUREG 1631: SOURCE DISCONNECTS RESULTING FROM RADIOG-U RAPHY DRIVE CABLE FAILURES. Final Report. MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-GIONAL SERVICE FACluTY, WAMPUM, PENNSYLVANIA. CARRICO.J.B. BOCCIO,J1, NUREG-1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS UCENSES.Prograrn-Specific Guidance About Industnal Radiograpny U-NUREG/CR4509: THE EFFECT OF INITIAL TEMPERATURE ON conses. Final Report. FLAME ACCELERATION AND DEFLAGRATION TO-DETONATION TRANSITION PHENOMENON. CATO,K. NUREG/CR-6524: THE EFFECT OF LATERAL VENTING ON DEFLA* NUREGCR4274: PALEOSEISMIC STUDIES IN THE SOUTHEASTERN GRATION-TO-DETONAT ION TRANSITION IN HYDROGEN-AIR- UNITED STATES AND NEW ENGLAND

• STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES.

CEBULL.M.J R4 : INVESTIGATION OF TECHNIQUES FOR THE DE- NUREGCli4268 V04: COMMON CAUSE FAILURE DATABASE AND U ANALYSIS SYSTEM. Volume 4: Software Reference Manual. VELOPMENT OF SEISMIC DESIGN BASIS USING THE PROB-ABILISTIC SEISMIC HAZARD ANALYSIS. CERNE,G. BONDARYK J.E. NUREG/lA 0141: RESULT OF BETHSY TEST 9.1.8 USING RELAP5/ MOD 3. NUREGCR4614: FEASIBILITY OF HIGH FREQUENCY ACOUSTIC IM-AGING FOR INSPECTION OF CONTAINMENTS. CHANIN,D. NUREG/CR4613 V01: CODE MANUAL FOR MACCS2. User's Guide.

                                                                                                           ^

• N R G-1608: CATEGORIZING AND TRANSPORTING LOW SPECIFIC DA2 R F F' ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS.

CHAPMAN,M.C. SUR R CHARACTERIZATION OF RETARDATION MECHA. TEN S EE SEISM BRAMWELL,D CHAPMAN.O.J. NUREGCR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE NUREG/CR4611: RESULTS OF PRESSURE LOCKING AND THERMAL { BINDING TESTS OF GATE VALVES. PROBABluTIES OF DEFECTS IN REACTOR PRESSURE VESSEL WELDS. BREY,R.R. NUREG/C 9: LOW-LEVEL WASTE DATA BASE DEVELOPMENT CH NU 1 : 1997 LOST SOURCE EXERCISE.An Exercise Of Radio-logical Response Through Cooperation And Coordination Of BROADDUS,0. Local, State And Federal Resources Under The National Contingency z NUREG-1556 V03: CONSOUDATED GUIDANCE ABOUT MATERIALS Plan. j UCENSES. Applications For Sealed Source And Device Evaluation And i NU GCR4559: SINGLE. AND CROSS-HOLE PNEUMATIC TESTS IN BROADDUS,0.A. UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE. NUREG-1631: SOURCE DISCONNECTS RESULTING FROM RADIOG. SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA. RAPHY DRIVE CABLE FAILURES. Final Report. BILITY,CONNECTIVITY AND SCALE. BROWN,M. CHEN,T.F. NUREGCR4617: THE PRICE-ANDERSON ACT CROS$ LNG THE NUREG/CR4608:

SUMMARY

AND EVALUATION OF LOW-VELOCITY BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. IMPACT TEST OF SOUD STEEL BILLET ONT0 CONCRETE PADS. BROWN.W.S. CHENG,H.S. NUREG-1507: MINIMUM DETECTABLE CONCENTRATIONS WITH NUREGCR4359 V01: RAMONA-4B: A COMPUTER CODE WITH TYPICA', RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR TAMINANTS AND FIELD CONDITIONS. SYSTEM TRANSIENTS.Models And Correlatens.

Personal Author index 41 NUREGCR4300 V02: RAMONA-48: A COMPUTER CODE WITH - CONGEL,F.J. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND S8WR NUREG-1633 DRFT FC: ASSESSMENT OF THE USE OF POTASSIUM SYSTEM TRANSIENTS. User's Manuel IODIDE (KI) AS A PUBLIC PROTECTIVE ACTION DURING SEVERE REACTOR ACCIDENTS. Draft Report For Comment. NUREGCR4634: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A - CONNELLY,SA DOWNWARD FACING CURVED SURFACE: EFFECTS OF THERMAL NUREG 1642 V03: ACCOUNTABILITY REPORT FISCAL YEAR 1997. INSULATION. CO M CHO.C.S. NUREG 1600: CATEGORIZING AND TRANSPORTING LOW SPECIFIC NUREGAA 0139: ASSESSMENT OF RELAP6/ MOD 3.2 USING LOFT ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS. NU IN AL I NELAP6/ MOD 3.2 ON 60406 AND 000KEAM. PENTIUM BASED PERSONAL COMPUTERS. NUREGCR4666 V01: PROSA81USTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS.Lele Health Effecte Uncertainty Asesse-CMO.D.M. ment. Main NUREGCR4372: EXPERIMENTS ON INTERACTIONS BETWEEN ZlR- NUREGC V02: PROSABILISTIC ACCIDENT CONSEQUENCE CONIUM OONTAINING MELT AND WATER. UNCERTANTY ANALYSIS.Lete Health Effecte Uncertainty Aseees. { CNol,T.S.' - - . NUY V01: PROBASILISTIC ACCIDENT CONSEQUENCE NUREG/lA 0139: ASSES 8 MENT OF RELAP6/ MOD 3.2 USING LOFT UNCER1 ANTY ANALYSIS. Uncertainty Assosoment For intemel Do-LARGE BREAK LOCA TEST.LP-024. . Main Report. NURE 4671 V02: PROSA81USTIC ACCIDENT CONSEQUENCE 7 V24: ENVIRONMENTALLY ASSISTED CRACKING IN NUR N ALL C i IN COMNGER,0.A. NU  : C F l TS ACCIDENT 1 A FATIGUE DESIGN CURVE 80F CARBON AND LOW-ALLOY STEELS. N EG 4674 V26: PRECURS O ENTIAL SEVERE CORE DAMAGE ACCIDENTS:1997. A Status Report. g NUREGCR4616: RISK COMPARISON OF SCHEDULING PREVENTIVE CORRADO,C.N. MANTENANCE DURNG SHUTDOWN VS. DURING POWER OPER- NUREG/CR4614: FEASIBiUTY OF HIGH FREQUENCY ACOUSTIC IM-ATION FOR PWRS. AGING FOR INSPECTION OF CONTAINMENTS. CHUNGEM. CORWIN,WA NUREGCR4667 V24: ENVIRONMENTALLY ASSISTED CRACKNG IN NUREGCR-6601 V04 N1: HEAVY-SECTION STEEL IRRADIATION .I

• 8*** *E 'I U NU G V26 NI ALL CRA INGlN '

UGHT-WATER REACTORS. Senuannual Report, July-Decenter 1997. CRASTREE,J.A. 4000: THE EFFECT OF INITIAL TEMPERATURE ON EVA AT

• l FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION NURE 4 H N OF LATERAL VENTING ON DEFLA- N RE' R-6003: CHARACTERIZATION OF RETARDATION MECHA-GRATION-TO-DETONAT ION TRANSITION N HYDROGEN-AIR- NISMS IN SOIL STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES.

DAMERONAA. CLETCHER,J.W. NUREGCR-6671: PRETEST PREDICTION ANALYSIS AND POSTTEST NUREGCR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE CORRELATION OF THE SIZEWELL-81:10 SCALE PRESTRESSED R Gl 26 E O ENTIAL SEVERE CORE DAMAGE ACCIDENTS:1997. A Status Report. DANIS,A.A. - NUREG-0625 V02 R06: SAFEGUARDS

SUMMARY

EVENT UST GbR4660: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM. DANN,RK NUREG/CR 6602: ENGINEERING DRAWINGS FOR 10 CFR PART 71 COLACCNO.J. PACKAGE APPROVALS. NUREGCP-0162 V02: PROCEEDINGS OF THE FIFTH NRC/ASME DAVIS,E.C. SYMPOSIUM ON VALVE AND PUMP TESTNG. NUREG/CR4689: LOW-LEVEL WASTE DATA BASE DEVELOPMENT COLE,CA PROGRAM. NUREGCR4621: GROUND-WATER MODELS IN SUPPORT OF NUREGCR4612. DAVISEL. NUREGCR4634 V02: FRAPCON-3: A COMPUTER CODE FOR THE COLE.R.K. CALCULATION OF STEADY-STATE. THERMAL-MECHANICAL BE-NUREGCR4119 V01 R1: MELCOR COMPUTER CODE MANU- , HAvlOR OF OXIDE FUEL RODS FOR HIGH BURNUP. NU R41W AE P CODE MANU- DEGRAssi,0. ALS.Relerence Manuele. Version 1.6.4, July 1997. NUREGCR4660: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND EM-AMM MMS. NU '666 V02: CONSOLIDATED GUIDANCE ABOUT MATERIALS UCENSES.N .. L. 7 Guldence About trestrial Radio 0rephy U- DEVKN,A.S. conees. Final Report. NUREG/lA-0026: RELAP6/ MOD 3 SUBCOOLED BOILING MODEL AS-SESSMENT. COLLINS,0.J. NUREG 1666 V06: CON 80UDATED GUIDANCE ABOUT MATERIALS DEWALLEG. UCENSES.N, ..?;_"- Guldence About Self-SNeided irredletor NUREGCR4611:RESULTS OF PRESSURE LOCKING AND THERMAL Ucensee. Final Report. SNDING TESTS OF GATE VALVES. DEY,ML COMPTON E. NUREG-1666 V03: CONSOLIDATED GUIDANCE ABOUT MATERIALS NUREG-1521 DRFT FC: TECHNICAL REVIEW OF RISK-NFORMED. PERFORMANCE-8ASED METHODS FOR NUCLEAR POWER PLANT UCENSES.Aararene For Sealed Source And Device Evaluation And FIRE PROTECTION ANALYSES.Drelt Report For Comment. Registration. Final Report.

42 Personal Author inden naar a a FINFROCK C. NUREGCR 0804: EVALUATION OF ULTRA 80NC INSPECTION TECH.- NUREGCR 0600: THE EFFECT OF INITIAL TEMPERATURE ON NIQUE8 FOR COAR8E GRAINED MATERIALS. FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION DIERCK8A.R. NUR 4 624 H F T OF LATERAL VENTNG ON DEFLA. NUREGCR4611 Vot: STEAM GENERATOR TUBE INTEGRITY PRO- GRATON-TO-DETONAT lON TRANSITON N HYDROGEN-AIR-I g g g896 gy STEAM MIXTURES AT VARIOUS INITML TEMPERATURES. 09AM.^ ~ October 1996 March 1997. Pl8C6ER L.E. i NUREGCR406:_7All BEHAVOR OF NTERNALLY PRE 88UR- NUREGCR4008:

SUMMARY

AND EVALUATION OF LOW-VELOCITY l IZED FLAWED AND UNFLAWED STEAM GENERATOR TU81NG AT IMPACT TEST OF SOUD STEEL 81LLET ONTO CONCRETE PADS. 1 HIGH TEMPERATURE . EXPERIMENTS AND COMPARISON WITH MODEL PREDICTIONS. FRA81KLINA NUREGCR4676: FAILURE BEHAVIOR OF INTERNALLY PRESSUR-DILLOW.T.A. IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT NUREGCR4400: TECHNICAL ASSISTANCE IN REVIEW OF SOURCE HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH TERM RELATED 188UE8 OF ADVANCED REACTORS. MODEL PREDICTIONS. nnans e a pupmasanans as NUREGCR4476: RESOLUTION OF THE DIRECT CONTAINMENT NUREGCR4600 LOW-LEVEL WASTE DATA BASE DEVELOPMENT l HEATING ISSUE FOR COM808 TION ENGNEERING PLANTS & PROGRAM i SA8 COCK & WILCOX PLANTS. j DOCTOR,8.R. . . NUREG-1666 V07 DR FC: CON 800 DATED GUIDANCE A80UT MATF-NUREGCR4471 V01: CHARACTERIZATION OF FLAWS M U.S. RE. RIALS UCENSES. PROGRAM-SPECIFC GUIDANCE ABOUT ACA. ACTOR PRES 8URE VES8ELS. Donelty And Delrtbution Of Flow laus . DEMC.RESEARCH AND DEVELOPMENT AND OTHER LEENSES estions in PVRUF. OF LIMITED SCOPE.Drelt Report For Comment. NUREGCR4004: EVALUATION OF ULTRA 80NC INSPECTION TECH- ) NIQUES FOR COARSE GRAMED MATERIALS. FU10CHE8,J.L. NUREG-1827 V01: PERFORMANCE PLAN FY 1900. DODOS,R.H. NUREGCR4670: APPUCATION OF THE NCSA HABANERO TOOL GALYEAN,W.J. FOR COLLABORATION ON STRUCTURAL WTEGRITY ASSESS. NUREGCR-6600 V01: RELIA 81UTY STUDY: AUXILMRY/ EMERGENCY MENTS. FEEDWATER SYSTEM. 1987-1906. 00 LANA.W. GATTONE,R.G. NUREGOR 4874 V26: PRECURSORS TO POTENTIAL SEVERE CORE NUREG 1666 V00 DR FC: CONSOUDATED GUIDANCE A80UT MATE- 1 ACCIDENTS: 1996. A Stelue Report. RIALS LICENSES.P e. - - Guidance About Meecal Use U- ( NURE 74 V20: PRECURSORS TO POTENTIAL SEVERE CORE coness. Droit Report For Comment. DAMAGE ACCIDENTS:1007. A Stelus Report. g DONOH00A.T. NUREGCR4110 V01 R1: MELCOR COMPUTER CODE MANU-NUREGCR4006: AN EVALUATION OF HUMAN FACTORS RESEARCH ALS. Primer And Users' V 1.8.4 1997. FOR ULTRASONIC INSERVICE NSPECTON. NUREGCR4119 V02 1: L ER CODE MANU-ALS. Reference Monumis, Version 1.8.4. July 1997. DugacK,P. - . NUREGCR4017: THE PRCE-ANDERSON ACT CROSSING THE "" " ^

• 4 8 RIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. NUREGCR4274: PALEOSEISMIC STUDIES M THE SOUTHEASTERN UNITED STATES AND NEW ENGLAND DWVER J.P.

NUREG-1668 V11 DR FC: CON 8000ATED GUIDANCE ABOUT MATE. GANTILLON,C.D. RIALS UCENSES. Program-Specinc Guldence About Spoolflc Ucenses NUREGCR4600 V01: REUA81UTY STUDY: AUXIUARY/ EMERGENCY og g, ,8=aaP= Dren Report For Comment FEEDWATER SYSTEM, 1987-1906. DWYER.P.A. GERHARD,M.A. NUREG 1619: STANDARD REVIEW PLAN FOR PHYSICAL PROTEC- NUREGCR-4664 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS-TION PLANS FOR THE INDEPENDENT STORAGE OF SPENT FUEL TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP-AND HIGH-LEVEL RADOACTIVE WASTE. PNG CASK DESIGN REVIEW. User's Manuel to Version 3a. saanas s n GERLACH.L. NUREGCR4661: IMPROVED EMBRITTLEMENT CORRELATIONS FOR NUREGCR4600: THE EFFECT OF INITIAL TEMPERATURE ON REACTOR PRES 8URE VES8EL STEELS. FLAME ACCELERATON AND DEFLAGRATION-TO-DETONATION TRANSITION PHENOMENON. EASTERLY,C.E. NUREGCR4624: THE EFFECT OF LATERAL VENTING ON DEFLA-NUREGCR.6640: ENVIRONMENTAL ASSESSMENT RENEWAL OF GRATION-TO-DETONAT ION TRANSITION IN HYDROGEN-AIR-MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE. STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. GONAL SERVICE FACIUTY. WAMPUM, PENNSYLVANIA. gg, EA87018,E. . . . NUREGCR4602: ACTION REOUIREMENTS FOR AFW SYSTEM FAIL-NUMEG-1808: CATEGORIZNG AND TRANSPORTING LOW SPECIFC URES.An Analysis For Four Nuclear Power Plants. f ACTIVITY MATERIAL 8 AND SURFACE CONTAMINATED OBJECTS. GM88E6 mes .s s NUREGCR4600: THE EFFECT OF WITIAL TEMPERATURE ON NUREGCR4673: "WVESTIGATING SEl8MOTECTONCS IN THE FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION

    . EASTERN UNITED STATES USNC A GEOGRAPHC INFORMATON                   TRANSITION PHENOMENON.
    - SY8 TEM.=                                                         NUREGCR 0624: THE EFFECT OF LATERAL VENTING ON DEFLA-GRATION-TO-DETONAT ION TRANSITION IN HYDROGEN-AIR-ELL 80GWOODA.                                                            STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES.

NUREGCR4648: A DAMAGE MECHANOS BASED APPROACH TO STRUCTURAL DETERIORATION AND REUA81UTY. GODINO,V. NUREGCR4614: FEASIBluTY OF HIGH FREQUENCY ACOUSTIC IM-i EINWT.R.- AGING FOR INSPECTON OF CONTAINMENTS. NUREG 0033 822: A PRIORITIZATION OF GENERIC SAFETY ISSUES. G000LAK,C.V. F80DLAYALW. - NUREG-1606 R01: A NONPARAMETRIC STATtBTICAL METHOD-NUREGCR4689: LOW-LEVEL WASTE DATA BASE DEVELOPMENT OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-PROGRAM COMMISSIONING SURVEYS. Interim Report For Use And Comment. I

Personal Author index 43 mannene s MJ NUREGCR4646 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREGCR4646 V01: PROSA81USTC AOCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Early Hoenh Eftects Uncerterey Assese-UNCERTANTY ANALYSIS. Early Heath Eflects Uncertalrny Aesses- ment. . mort. Main NURE 4666 V01: PROBABluSTIC ACCIDENT CONSEQUENCE NUREGCR V02: PROSABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS.Lete Heath ENects Uncerteney Assees-NU C V02: PROBABILISTIC ACCIDENT CONSEQUENCE NU V01: PROSA88USTIC ACCIDENT CONSEQUENCE UNCERTANTY ANALYSIS.Lele Health Effecte UncertowWy Assees-NU V01: PROBABILISTIC ACCIDENT CONSEQUENCE NURE V02: PROSA88USTC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assosoment For inlemal Do-UNCERTANTY ANALYSIS. Late Hoehh Ellecte Uncertelnty Asseen. NUNb 467  : PROBABluSTIC ACCIDENT CONSEQUENCE NUREGGl4671 V01: PROSA810STIC ACCIDENT CONSEQUENCE UNCERTA8NTY ANALYSIS. Uncertemty Aseeeemers For intemel Do-UNCERTAINTY ANALYSIS. Uncertainly Asseeement For intemel Do- esmetry. Appendices N

     . ~ p. _ ty                                                                  2 ROI: EVENT REPORTING GUIDEUNES 10 CFR 60.72 anneaans a a                                                       MARRISJL
                                                                     "                 1ET            PROBABLE FLAW DISTRIBUTIONS IN NUREGCR4621: ESTIMATING PROSA8LE FLAW DISTRIBUTIONS IN                       T PWR STEAM GENERATORS.

• HARRIS,R.V.

00fWHARA NUREG/lA-0141: RESULT OF BETHSY TEST 9.1.8 USING RELAP6/ NUREGOR4606: AN EVALUATION OF HUMAN FACTORS RESEARCH MOD 3. FOR ULTRASONIC INSERVICE INSPECTION. GRAtlT,0.M. HARRISONJ.D. NUREGCR4671 V01: PROBABluSTIC ACCIDENT CONSEQUENCE NUREGCR-6000 V0v: REUA88UTY STUDY: AUXILIARY / EMERGENCY UNCERTAINTY ANALYSIS. UncerterWy Aseeenment For Inlemal Do-FEEDWATER SYETEM,1987-1996.

                                                                               . Main Report.

NURE R4671 V02: PROBA81LISTIC ACCIDENT CONSEQUENCE

            $68 VOS: CONSOUDATED GUIDANCE ABOUT MATERIALS              eims UCENSEES.L _ .. "+ Guudence Emempt Deelettiulion U-coness. Fine! Report.                                         HASHEMBAN,H.M.

GREEglulOOD,M1 NUREGCR4601: ADVANCED INSTRUMENTATION AND MAINTE-NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS.

. NUMEGCR4689: THE EFFECTS OF SURFACE CONDITION ON AN ULTRA 80NIC INSPECTION: ENGINEERWG STUDIES USING VAU-         HASHEMLAN,M.

DATED COMPUTER MODEL NUREGCR-6601: ADVANCED INSTRUMENTATION AND MAINTE-NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. GRUSER,El NUREGCR4087 V24: ENVIRONMENTALLY ASSISTED CRACKING IN HASIGN# 1 UGHT-WATER REACTORS. Semiennual Report. January June 1997. NUREG/CR4646 V01: PROBABluSTC ACCIDENT CONSEQUENCE NUREGCR eas7 V26: ENVIRONMENTALLY ASSISTED CRACKING IN UNCERTAINTY ANALYSIS. Early Health Eftects Uncertairny Assese-UGHT-WATER REACTORS. Semiennual Report, July-December 1997. ggg pp GRUENHAGEN.S.E. UNCERTAINTY ANALYSIS. Early Hoekh Eflects Uncertainty Assees-NUREGCR4003: CHARACTERIZATION OF RETARDATION MECHA- menLAppendices NISMS N 80lL. HASSAN,M. GRUPAJ,3. NUREGCR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAU., NUREGCR4646 V01: PROBA84USTIC ACCIDENT CONSEQUENCE FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED

                                                           ^***~

mort :grt. NR CR46 TAL YST S IN NUCLEAR POWER PLANTS.Riek-Screening Of Environmental Streasore And A Compo GUNTHER,W.H. Son Of Hardware Unsweilatilley With An Extehng Analog Syelem. NUREGCR4372: EXPERIMENTS ON INTERACTIONS BETWEEN ZlR-CONIU14CONTAINING MELT AND WATER. HA R4604: EVALUATION OF ULTRASONIC INSPECTION TECH. HAGWAEYER,0.A. NIQUES FOR COARSE GRAINED MATERIALS. NUREG-0713 V18: OCCUPATIONAL RADIATION EXPOSURE AT COM-MERCIAL NUCLEAR POWER REACTORS AND OTHER FAClu-

                      " ~

HAYSA"G NURE 1666 V07 DR FC: CONSOUDATED GUIDANCE AB NL 0 i9: O6 CUP IONA DIATION EXPOSURE AT COM- RIALS UCENSES. PROGRAM-SPECIFIC GUIDANCE ABOUT ACA-MERCIAL NUCLEAR POWER REACTORS AND OTHER FACluTIES DEMIC.RESEARCH AND DEVELOPMENT AND OTHER UCENSES 1997.Thir9elh Annual Report. OF UMITED SCOPE. Draft Report For Comment HE.W NU GC  : ANCHOR SOLT BEHAVIOR AND STRENGTH DUR- RE P P e to NG EARTHOUAKES. HEAMES,T.J. HANJ.T. NUREGCR4604: RADTRAD: A SIMPUFIED MODEL FOR RADIO-NUREGCR-6det: SINGLE-PHASE AND TWO PHASE NATURAL CIR-NUCUDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION. CULATION TESTS IN THE PUMA FACILITY. HEASLER,P.G. HANSON,A.L. NUREGCR4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE-NUREGCR4418: RISK IMPORTANCE OF CONTANMENT AND RE- ACTOR PRESSURE VESSELS. Denemy And Distritsution Of Flow indh LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. catione in PVRUF, MARPERf.T. HENDERSON,PJ. NUREGCR4646 V01: PROBA81USTIC ACCIDENT CONSEQUENCE NUREG 1666 V04: CONSOUDATED GUIDANCE ABOUT MATERIALS UNCERTANTY ANALYSIS. Early Health Eflects Uncerle6nty Assese- UCENSES.Progrem-Specife Guidance About Fixed Gauce U-ment Main Report. cenees. Final Report.

44 Personal Author index HENSON,J. JEE,K K. NUREG-1556 V10 DR FC: CONSOUDATED GUIDANCE ABOUT MATE- NUREG/lA 0148: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF RIALS LICENSES. Program Specife Guidance About Master Matenats TEN-PERCENT MAIN STEAM-LINE-BREAK TEST RUN SB-SL-01. Uoonses. Draft Report For Comment. HILDEBRAND,B.P, NUREGOR4472: PREUMINARY PHENOMENA IDENTIFICATION AND NUREGCR4594: EVALUATION OF ULTRASONIC INSPECTION TECH- RANKING TABLES FOR SIMPLIFIED BOILING WATER REACTOR NIQUES FOR COARSE GRAINED MATERIALS. LOSS OF COOLANT ACCIDENT SCENARIOS. HILTON.LD. JOHNSON,0.L i NUREGOR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT NUREGCR4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS- 1 PROGRAM. TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP-PING CASK DESIGN REVIEW. Users Manual to Version 3a. NUREGCR4119 V01 R1: MELCOR COMPUTER CODE MANU- JONES,A.R. ALS Primer And Users' Guides. Version 1.8.4. July 1097. NUREG-1556 V09 DR FC: CONSOLIDATED GUIDANCE ABOUT MATE-NUREGCR4119 V02 RI: MELCOR COMPLAER CODE MANU- RIALS LICENSES. Program-Specife Guidance About Medcal Use U-ALS. Reference Manuals,Versson 1.8.4, July 1997. censes. Draft Report For Comment. HOFMAYER,C.H. JONES.J.D. l NUREGCR4554: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR ! NUREG-1556 V10 DR FC: CONSOUDATED GUIDANCE ABOUT MATE-l WALL INTERNATIONALjCALE NUREGCR4559: LARGE TANDAFID PROBLEM. VIBRATION TESTS OF MAIN STEAM RIALS UCENSES. Program Specife Guidance About Master Materials Ucenses. Draft Report For Comment

                                                                                                                                                    )J AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND ENERGY-ABSORBING SUPPORTS.                                           JONES,W.R.

NUREG-1022 ROI: EVENT REPORTING GUIDEUNES 10 CFR 50.72 HOPPER,C.M. AND 50.73. NUREGCR-5342: ASSESSMENT AND RECOMMENDATIONS FOR FISSILE-MATERtAL PACKAGING EXEMPTIONS AND GENERAL U. KAFKA.A.L CENSES WITHIN 10CFR PART 71. NUREGCR4573: " INVESTIGATING SEISMOTECTONICS IN THE EASTERN UNITED STATES USING A GEOGRAPHIC INFORMATION HORA,8.C. SYSTEM." NUREGCR4555 VO1: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- KAM,F.B. ment. Main Re . NUREG/CR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH. NUREGCR455pV02: a PROBABluSTC ACCIDENT CONSEQUENCE MARK. UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess-NU 1$C1: PROBABILISTIC ACCIDENT CONSEQUENCE NU E R4600: NEUTRON EXPOSURE PARAMETERS FOR CAP. UNCERTAINTY ANALYSIS. Uncertanty Assessment For intamal Do. SULE 10.05 IN THE HEAVY-SECTION STEEL 1RRADIATION PRO-simetry. Main Report GRAM TENTH IRRADIATION SERIES. NUREGCR4571 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREGCR4601: NEUTRON EXPOSURE PARAMETERS FOR THE UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do. DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRADIA-simetry.Appendoes, TION PROGRAM TENTH IRRADIATION SERIES. HOVINGH,J. KANG.S.K. NUREGCR4608:

SUMMARY

AND EVALUATION OF LOW-VELOCITY NUREG/lA-0148: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. TEN-PERCENT MAIN STEAM-UNE-BREAK TEST RUN SB-SL-01. HOWE,0.B. KARLSEN.T.M. NUREG 1556 V10 DR FC: CONSOUDATED GUIDANCE ABOUT MATE. NUREGCR4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN HlALS UCENSES. Program Speelfe Guidance About Master Materials UGHT-WATER REACTORS. Semiannual Report. January-June 1997. Ucenses. Draft Report For Comment. KASSNER,T.F. HUFFERT,A.M. NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG 1505 Rot: A NONPARAMETRIC STATISTICAL METHOD- UGHT-WATER REACTORS. Semiannual Repart. January 4une 1997. NUREGCR4667 V25: ENVIRONMENTALLY ASSISTED CRACKING IN OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE- UGHT-WATER REACTORS. Semiannual Report. July-December 1997. COMMISSIONING SURVEYS.Intenm Repoit For Use And Comment. NUREG-1507: MINIMUM DETECTABLE CONCENTRATIONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- KASZA NUREGCK.E'R4511 V02: STEAM GENERATOR TUBE INTEGR'TY PRO TAMINANTS AND FIELD CONDITIONS. GRAM. Annual Report,Auaust IWS - September 1996. I NUREG/CR4511 v03: STEAM GENERATOR TUBE INTEGRITY PRO-Senannual , Och M - Mae W. NUR 4 : RADTRAD: A SIMPUFIED MODEL FOR RADIO-NUCLlDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION. KELLY,D.L NUREGCR4496: EVALUATION OF LOSS OF OFFSITE POWER

                                                                                                                               ~

N -5559: SINGLE AND CROSS-HOLE PNEUMATIC TESTS IN UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE- KEMPPINEN,H. SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA- NUREGCR4274: PALEOSEISMIC STUDIES IN THE SOUTHEASTERN BluTY,CONNECTIVITY AND SCALE. UNITED STATES AND NEW ENGLAND. l ISHil,M. KHAN,H.J. 1 l NUREQCR4498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- NUREGCR4359 V01: RAMONA4B: A COMPUTER CODE WITH CULATION TESTS IN THE PUMA FACluTY. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR l ISKANDER,S.K. NI o B C UTER CODE WITH NUREGCR4447: RESULTS OF CRACK-ARREST TESTS ON IRRADI- THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR ATED A 508 CLASS 3 STEEL. SYSTEM TRANSIENTS. Users Manual. JAQU AY,K. KIM,E.K. I NUREGCR4361: SEISMIC ANALYSIS OF PIPING. Final Program Re- NUREG/lA4142: INSTALLATION OF RELAP5/ MOD 3.2 ON 80486 AND port. PENTIUM BASED PERSONAL COMPUTERS. JASTROW.J.D. KIM,H.J. NUREGCR4560: LOW-LEVEL WASTE DATA BASE DEVELOPMENT NUREG/lA-0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING LOFT PROGRAM. LARGE BREAK LOCA TEST LP 024.

l Personal Author index 45 l l NUREGAA 0142: INSTALLATION OF RELAP5 MOD 3.2 ON 80486 AND KROGER.P.G. ) PENTIUM BASED PERSONAL COMPUTERS. l NUREG/CR4472: PRELIMINARY PHENOMENA IDENTIFICATION AND NUREG/lA 0143: ASSESSMENT OF RELAPfdWOD32 WITH THE LSTF ! RANKING TABLES FOR SIMPLIFIED BOluNG WATER REACTOR ' EXPERIMENT SIMULATING A LOSS OF RESIDUAL HE LT REMOVAL LOSS OF COOLANT ACCIDENT SCENARIOS. , EVENT DURING MID-LOOP OPERATION. 1 NUREGAA 0144: ASSESSMENT OF RELAP5 MOD 3.2 WITH THE KRUPKA.K.M.  ; SEMISCALE NATURAL CIRCULATION EXPERIMENT, S-NC48. NUREGCR4377: EFFECTS ON RADIONUCUDE CONCENTRATIONS i NUREGAA4147: ASSESSMENT OF RELAP5/ MOD 3.2 FOR STEAM BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF I CONDENSATION EXPERIMENTS IN THE PRESENCE OF PERFORMANCE ASSESSMENT OF LOW-LEVEL RADIOACTIVE NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. WASTE DISPOSAL FACluTIES. KIM LS. KRUSE,K. NUREG/CR4502: ACTION REQUIREMENTS FOR AFW SYSTEM FAIL-  ! URES.An Analysis For Four Nuclear Power Plants. NUREG/CR-5570: APPUCATION OF THE NCSA HABANERO TOOL FOR COLLABORATION ON STRUCTURAL INTEGFOTY ASSESS-KIRKWOOD,A.S.

 ' NUREG.1566 V04: CONSOUDATED GUIDANCE ABOUT MATERIALS              gyongCK,J.D.

l LICENSES. Program-Spectre Guidance About Fixed Gauge Lk connes. Final Report. NUREG-1632: EVALUATION OF AP600 CONTAINMENT THERMAL-HY. DRAUUC PERFORMANCE. ' l KLAMERUS.E.W. KUMAMARU,H. ' NUREG/CR4517: ROUND ROBIN PRETEST ANALYSES OF A STEEL NUREG/lA-0148: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF CONTAINMENT VESSEL MODEL AND CONTACT STRUCTURE AS-TEN-PERCENT MAIN STEAM-UNE-BREAK TEST RUN SS SL 01 SEMBLY SUBJECT TO STATIC INTERNAL PRESSURt2ATION. KUPPERMAN,0.5. KLINGINOMITH,0. NUREG/CR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO-NUREG-1829: THE CHARACTERIZATION OF VICKER'S MICROHARD- GRAM. Annual Report.AuDust 1995 - September 1996. NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN- NUREG/CR4511 V03: STEAM GENERATOR TUBE INTEGRITY PRO-HARDENING MICROPROBE. GRAM. Semiannual Report, October 1996 - March 1997. KLINGNER,R.E. KVARFORDT.K.J. NUREG/CR4434: ANCHOR BOLT BEHAVIOR AND STRENGTH DUR- NUREG/CR4268 V04: COMMON CAUSE FAILURE DATABASE AND ING EARTHOUAKES. ANALYSIS SYSTEM. Volume 4: Software Reference Manual. KNOSLICH.L LAFFERTY,R.H. NUREG/CR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR- NUREG/GR-0017: DATING OF LIQUEFACTION IN THE NEW MADRID IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT SEISMIC ZONE AND IMPUCATIONS FOR EARTHOUAKE HAZARD. HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH MODEL PREDICTIONS. LAWSERT,H.E. NUREG/CR.6544: METHODOLOGY FOR ANALYZING PRECURSORS KNUDSEN.J.K. TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE-NUREG/CR-5600 V01: RELIABluTY STUDY: AUXILIARY / EMERGENCY OVENCES. FEEDWATER SYSTEM,1987-1995. LANNING,0.D. KNUDSON.D.L. NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE NUREGlCR4475: RESOLUTION OF THE DIRECT CONTAINMENT CALCULATION OF STEADY-STATE, THERMAL-MECHANICAL BE-HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & HAVIOR OF OXIDE FUEL RODS FOR HIGH BURNUP. BABCOCK & WILCOX PLANTS. NUREG/CR4534 V03: FRAPCON-3: INTEGRAL ASSESSMENT. KOHN,W.E. LANZISERA.P.A. NUREG/CR4577: U.S. NUCLEAR POWER PLANT OPERATING COST NUREG-1556 V09 DR FC: CONSOUDATED GUIDANCE ABOUT MATE-AND EXPERIENCE SUMMARIES. RIALS LICENSES. Program Specife Guidance About Medcol Use U-connes. Draft Report For Comment NUREGCR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAU- LARSON,LL. FICATION OF MICROPROCESUOR-BASED SAFETY RELATED NUREG/CR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT EQUIPMENT IN NUCLEAR POWER PLANTS. PROGRAM. KRAAN.S.C.P. LAURENSON,J. NUREG/CR4545 V01: PROBABluSTIC ACCIDENT CONSEOUENCE NUREG/CR4617: THE PRICE-ANCERSON ACT - CROSSING THE UNCERTAINTY ANALYSIS. Earty Health Effects Uncertainty Assess- BRIDGE TO THE IdXT CENTURY: A REPORT TO CONGRESS. mont. Main Report. NUREG/CR4545 V02: PROBABluSTIC ACCIDENT CONSEOUENCE LAZAREWICZ,A.R. UNCERTAINTY ANALYSIS. Earty Health Elfacts Uncertainty Assess- NUREG/CR4573: INVESTIGATING SEISMOTECTONICS IN THE ment A . EASTERN UNITED STATES USING A GEOGRAPHIC INFORMATION NURE 4555 VOI: PROBABluSTIC ACCIDENT CONSEQUENCE SYSTEM? UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assens-ment. Main Report. LEE,0.W. NUREG/CR4555 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/LA 0142: INSTALLATION OF RELAP5/ MOD 3.2 ON 80486 AND UNCERTAINTY ANALYSIS. Late Health Eflects Uncertainty Assess- PENTIUM BASED PERSONAL COMPUTERS. NU G V01: PROBABluSTIC ACCIDENT CONSEQUENCE LEE,H.D. UNCERTAINTY ANALYSIS. Uncertainty Assessment For intemal Do- NUREG/LA 0148: ASSESSMENT OF RELAPS/ MOD 3.1 USING LSTF s6 metry. Main Report. TEN-PERCENT MAIN STEAM-LINE-BREAK TEST RUN SB-SL.01. NUREG/CR4571 V02: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assessment For intemal Do- LEE,J.H. simetry. Appendices. NUREG/lA-0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING LOFT LARGE BREAK LOCA TEST,LP-024. KREIDER,M.A. NUREG/CR4521: ESTIMATING PROBABLE FLAW tilSTRIBUTIONS IN LEE.S. PWR STEAM GENERATORS. NUREG/lA4143: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE LSTF EXPERIMENT SIMULATING A LOSS OF RESIDUAL riEAT REMOVAL l KRILL,S. EVENT DURING MID-LOOP OPERATION. NUREG/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE NUREG/lA-0144: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. SEMISCALE NATURAL CIRCULATION EXPERIMENT, S-NG.88.

48 Personal Author index LacNARol.T. MACKNNON,J.S. NUREGCR-6400: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- NUREG 1022 ROI: EVENT REPORTING GUIDEUNES 10 CFR 50.72 CULATION TESTS IN THE PUMA FACILITY. AND 50.73. LETT18,W.R. MAJUISAR,S. NUREG/CR-6662: DATING AND EARTHOUAKES: REVIEW OF QUA- NUREGCR4611 V02: STEAM GENERATOR TUBE INTEGRITY PRO- ] TERNARY GEOCHRONOLOGY AND iTS APPLICATION TO GRAM. Annual .A 1996 - 1996. PALEOSEISMOLOGY. NUREGCR4611 :S GENE OR TUBE INTEGRITY PRO-GRAM.Somiennual Report, October 1996 - March 1997 LEWWE.R. NUREGCR4575: FAILURE BEHAVIOR OF NTERNALLY PRESSUR-NUREG 1521 DRFT FC: TECHN6 CAL REVIEW OF RISK-lNFORMED, lZED FLAWED AND UNFLAWED STEAM GENERATOR TUBNG AT PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH FIRE PROTECTION ANALYSES. Draft Report For Comment- MODEL PREDICTIONS. LEWISR. MALIK S.N.M. NUREG-1606: CATEGORIZING AND TRANSPORTING LOW SPECIFIC NUREGCR-5670: APPUCATION OF THE NCSA HABANERO TOOL ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS. FOR COLLABORATION ON STRUCTURAL WTEGRITY ASSESS-LEWIS,8.H. NUREG.1566 V04: CONSOLIDATED GUIDANCE ABOUT MATERIALS MALLEN,A.N. LICENSES. Pro 9 tem. Spec 6fic Gu6dence About Fised Gau9e. U- NUREGCR4359 V02: RAMONA 48. A COMPUTER CODE WITH oeness. Final Report. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR SYSTEM TRANSIENTS. User's Manuel LIAN,K. NUREG/GR-0016: THE ROLE OF TIME-DEPENDENT DEFORMATION MALLIAKOS,A. IN WTERGRANULAR CRACK INITIATION OF ALLOY 600 STEAM NUREGCR4500: THE EFFECT OF INITIAL TEMPERATURE ON GENERATOR TUBING MATERIAL FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION TRANSITION PHENOMENON LICHTENWALTER J NUREGCR4524: THE EFFECT OF LATERAL VENTING ON DEFLA-NUREGCR4342: ASSESSMENT AND RECOMMENDATIONS FOR GRATION-TO-DETONAT lON TRANSITION IN HYDROGEN-AIR-FISSILE-MATERIAL PACKAGING EXEMPTIONS AND GENERAL U- STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. CENSES WITHIN 10CFR PART 71. NUREGCR4600: PERFORMANCE TESTNG OF PASSIVE AUTOCATALYTIC RECOMBINERS. NUREG-1622: NRC ENFORCEMENT POLICY REVIEW. July 1995 July MANKAMO,T. 1997. NUREGCR4602: ACTION REQUIREMENTS FOR AFW SYSTEM FAIL-URES.An Analysie For Four Nuclear Power Plants. NUREGCR4664: ANALYSES OF SOURCE SPECTRA ATTENUATION. MAngNALL F.M. AND SITE EFFECTS FROM CENTRAL AND EASTERN UNITED NUREGCR4485: GUIDELINES ON MODELNG COMMON-CAUSE STATES EARTHOUAKES. FAILURES IN PROBASILISTIC R?SK ASSESSMENT. NUREG/CR-6496: EVALUATION OF LOSS OF OFFSITE POWER LITTLE.M.P. EVENTS AT NUCLEAR POWER PLANTS: 1980 1996. NUREGCR4656 V01: PROBABILISTIC ACCIDENT CONSEOUENCE NUREGCR-6497: COMMON CAUSE FAILURE PARAMETER ESTi-UNCERTAINTY ANALYSIS. Late Health Effects Uncerterwy Asesse- MATIONS. mont. Main Report. N NUREGCR-6208 V01: COMMON CAUSE FAILURE DATABASE AND NUREGCR4666 V02: PROBABluSTIC ACCIDENT CONSEQUENCE ANALYSIS SYSTEM. Volume 1:Overviour. UNCERTAINTY ANALYSIS.Lete Health Eflects Uncertainty Assees. NUREGCR4208 V02: COMMON CAUSE FAILURE DATABASE AND ment.Appendmea, ANALYSIS SYSTEM. Volume 2: Event Definition And C6eselfication. NUREGCR4268 V03: COMMON CAUSE FAILURE DATABASE AND LIU,Y.C. ANALYSIS SYSTEM. Volume 3: Daleramarman And Event Coeno. NUREGCR-6634: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A NUREGCR4208 V04: COMMON CAUSE FAILURE DATABASE AND DOWNWARD FACNG CURVED SURFACE: ErFECTS OF THERMAL ANALYSIS SYSTEM. Volume 4: Soliware Reference Manual. WSULATION. LLOPIS.C. NUREG/lA 0149: ASSESSMENT OF RELAP6 MOD 3.2-NPA3.4 AGAWST NUREG/lA 0149: ASSESSMENT OF RELAP5/ MOD 3.2 NPA3.4 AGANST A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR TRIP, NATURAL CIRCULATION AND THE START OF A MAIN PUMP TRIP, NATURAL CIRCULATION AND THE START OF A MAIN PUMP IN THE VANDELLOS 11 NUCLEAR POWER PLANT. IN THE VANDELLOS 11 NUCLEAR POWER PLANT, LOOHl A.S. . NUREG-1521 DRFT FC: TECHNICAL REVIEW OF RISK-INFORMED. NUREG-1566 V07 DR FC: CONSOUDATED GUIDANCE ABOUT MATE- PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT RIALS UCENSES. PROGRAM-SPECIFIC GUIDANCE ABOUT ACA- F!RE PROTECTION ANALYSES. Draft Report For Commer4 DEMIC,RESEARCH AND DEVELOPMENT AND OTHER LICENSES OF LIMITED SCOPE.Drott Report For Comment. MARTINEZ GURIDI ( NUREGCR4616 RISK COMPARISON OF SCHEDUUNG PREVENTIVE LOTZE,0. MANTENANCE DURING SHUTDOWN VS. DURING POWER OPER- ( NUREGCR-5434: ANCHOH BOLT BEHAVIOR AND STRENGTH DUR- ATION FOR PWRS. NG EARTHOUAKES. MAGNIK,M.T. LUSINSKI,J. NUREG-1628 DRF FC: STAFF RESPONSES TO FREQUENTLY ASKED NUREG-1566 V03: CONSOLIDATED GUIDANCE ABOUT MATERIALS QUESTIONS CONCERNING DECOMMISSIONING OF NUCLEAR UCENSES.AM*amano For had Source And Dev6ce Evolushon And POWER PLANTS. Draft Report For Comment Registration. F6nel Report. MAVKO,s. LUCAS,G.E. NUREG/lA4141: RESULT OF BETHSY TEST 9.1.B USING RELAP5/ . NUREG-1629: THE CHARACTERIZATION OF VICKER'S MICROHARD- MOD 3.  ! NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN- NUREG/lA-0145: RELAPS AESESSMENT AGAINST PACTEL EXPERI-HARDENING MICROPROBE. MENTAL DATA (REVISION 1). LUK,V.K. MCCOLD.LN. NUREG/CR4617: ROUND ROBIN PRE 1EST ANALYSES OF A STEEL NUREGCR-6549: ENVIRONMENTAL ASSESSMENT RENEWAL OF CONTANMENT VESSEL MODEL AND CONTACT STRUCTURE AS- MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-SEMBLY SUIDECT TO STATIC INTERNAL PRESSURIZATION. GIONAL SERVICE FACluTY, WAMPUM, PENNSYLVANIA.

E Personal Author index 47 l MocONNELL.J.w. MORRIS.E.S. l' I NUREGICR4669: LOW-LEVEL WASTE DATA BASE DEVELOPMENT NUREG-0540 V20 N01: TITLE LIST OF DOCUMENTS MADE PUBLICLY PROGRAM. AVAILABLE. January 1-31,1996. MCDONALD,G.P- MORTON,G.W. t NUREGICR4604: EVALUATON OF ULTRASONIC INSPECTION TECH- NUREG/CR4501: ADVANCED INSTRUMENTATION AND MAINTE-NLUES FOR COARSE GRAINED MATERIALS. NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. BACKELLAR M.G. RADSLEH,A. NUREG/CR 9611: RESULTS OF PRESSURE LOCKING AND THERMAL NUREG/CR-5485: GUIDEUNES ON MODELING COMMON-CAUSE BINDING TESTS OF GATE VALVES. guYM34$77gDggiLggRAMETER ESTi- E R R-6621: GROUND-WATER MODELS IN SUPPORT OF S T NUREGICR4512. NUR 4 UN FAILURE DATABASE AND l teLELLA.P.P. NU E O U A LU SIAND NUREGICR-6447: RESULTS OF CRACK-ARREST TESTS ON IRRADI- ANALYSIS SYSTEM.V  : Datta Collection And Event Coeng. ' ATED A 606 CLASS 3 STEEL NUREGCR4268 V04: CAUSE FAILURE DATABASE AND ANALYSIS SYSTEM. Volume 4: Software Reference Manual MILLER,LA. NUREGCR4004: RADTRAO: A SIMPUFIED MODEL FOR RADIO. MRUK,K. NUCLIDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION. NUREGCR4575: FAILURE BEHAVOR OF INTERNALLY PRESSUR. tZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT MILLER.M.K. HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH NUREG/CR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON MODEL PREDICTIONS. THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR PRESSURE VESSEL MATERIALS.An Atom Probe Study. 00U E i RANARICKJ.W DAMAGE ACCIDENTS:1996. A . NUREGCR5674 V25: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-4674 V26: PRECUR S O TENTIAL SEVERE CORE  ! DAMAGE ACCIDENTS: 1996. A Status Report. DAMAGE ACC; DENTS:1997. A Status Report.  ; NUREGCR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE MUIRHEAD,C.R. DAMAGE ACCIDENTS:1997. A Status Aerort- NUREGCR4555 V01: PROBABILISTIC ACCIDENT CONSEQUENCE gggggj,g UNCERTAINTY ANALYSIS. Late Health Eflects Uncertanly Assees-NUREG 1628 DRF FC: STAFF RESPONSES TO FREQUENTLY ASKED NUNa/C bO2: PROBABluSTIC ACCIDENT CONSEQUENCE i OUESTIONS CONCERNING DECOMMISSONING OF NUCLEAR UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Aseees. POWER PLANTS. Draft Report For Comment rnent W . IM CHELL.D.W. gyLLaugn,J,g, j NUREGCR-6601: ADVANCED INSTRUMENTATON AND MAINTE- NUREG-1556 VII DR FC: CONSOUDATED GUIDANCE ABOUT MATE-NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. RIALS UCENSES.Prograrr> Specific Guidance About Specific Uoonees Of Broadacope. Draft Report For Comment INTCHELL.M.W. NUREG-1566 V05: CONSOUDATED GUIDANCE ABOUT MATERIALS MURPHY,8.D. UCENSES. Program-Specife Guidance About Self-Shielded irradetor NUREGCR4536: VERIFICATION OF THE LWRARC CODE FOR Licennes. Final Report. UGHT-WATER-REACTOR AFTERHEAT RATE CALCULATIONS. 800HOENt A.S. MURTY,8.8. l NUREG-1633 DRFT FC: ASSESSMENT OF THE USE OF POTASSIUM NUREGCR-6606:

SUMMARY

AND EVALUATION OF LOW-VELOCITY j ODIDE (KI) AS A PUBUC PROTECTIVE ACTION DURING SEVERE IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. REACTOR ACCIDENTS. Draft Report For Comment. NAQY,K.L teDK,G.C. NUREG/CR4603: CHARACTER!ZATION OF RETARDATON MECHA. NUREGICR-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS- NISMS IN SOIL TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP- NAUS,D.J. ~ PING CASK DESIGN REVIEW. User's Manual to Verson Sa. NUREG/CR4596: AN INVESTIGATION OF TENDON SHEATHING NUREGCR-6608:

SUMMARY

AND FVALUATION OF LOW-VELOCITY FILLER MiGRATON INTO ' RETE. IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. NUREGCR4616: A SURVEY EPAIR PRACTICES FOR NUCLEAR POWER PLANT CONTAINMENT METALLIC PRESSURE BOUND- , 000NNOE.D.K. ARIES. I NUREGCR4604: RADTRAD: A SIMPLIFIED MODEL FOR RADIO-i NUCUOE TRANSPORT AND REMOVAL AND DOGE ESTIMATION. NEILSON,R.M. l NUREGCR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT MDNTELEONE,S. PROGRAM. NUREGCP 0162 V01: PROCEEDINGS OF THE TWENTY-FIFTH WATER REACTOR SAFETY INFORMATION MEETING. Plenary Soo- NELLIS,0. I sons. pressure Vessel Roeweh,BWR Strainer Beckage And Omer NUREG-1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS Generic salsty issuns ErMronmentally Assisted Degradaten Of LWR.... LICENSES. Program-Specife Guidance About industrial Radography U-NUREGCP 0162 V02: PROCEEDINGS OF THE TWENTY-FIFTH conses. Final Report. WATER REACTOR SAFETY INFORMATION MEETING. Human Reli- NELSON,C.F ability Analysis And Human Performance Evmiunton, Technmal lasues NUREGCd6412: AGING AND LOSS OF COOLANT ACCIDENT (LOCA) Related To Rulemakings, Risk-informed, Performance-Based initia- TESTING OF ELECTRICAL CONNECTIONS. 16ves-NUREG/CP 0162 V03: PROCEEDINGS OF THE TWENTY-FIFTH NEUMAN,8.P. WATER REACTOR SAFETY INFORMATON MEETING. Thermal-Hy- NUREGCR-5559: SINGLE- AND CROSS-HOLE PNEUMATIC TESTS IN l l draulic Research And Codes, Digital Instrumentation And Control, UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-Structural Performance. SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA. NUREGCP-0165: TRANSACTIONS OF THE TWENTY SIXTH WATER BluTY,CONNECTIVITY AND SCALE. REACTOR SAFETY INFORMATION MEETING. NEYMOTIN,LY. teDRENO P. NUREGCR4359 V01: RAMONA-4B: A COMPUTER CODE WITH NUREG/lA-0150: STUDY OF TRANSIENTS RELATED TO AMSAC AC- THRES-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR TUATION. SENSITIVITY ANALYSIS. SYSTEM TRANSIENTS.Models And Correlations. l l.

48 Personal Author index NUREGOR4369 V02: RAMONA 48: A COMPUTER CODE WITH NUREGCR4559: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND ENERGY-ABSORBING SUPPORTS.

                                                                                                                                            ]

SYSTEM TRANSIENTS. User's Manuel NICHOLSON.T.J. PARKS.C.V. NUREGCP4163: PROCEEDINGS OF THE WORKSHOP ON REVIEW NUREGCR-5342: ASSESSMENT AND RECOMMENDATIONS FOR OF DOSE MODEUNG METHODS FOR DEMONSTRATION OF COM- FISSILE-MATERIAL PACKAGING EXEMPTIONS AND GENERAL U-PUANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TER- CENSES WITHIN 10CFR PART 71, PARROTT J.D. NO,H.C. NUREGCP 0163: PROCEEDINGS OF THE WORKSHOP ON REVIEW NUREG/lA-0147: ASSESSMENT OF RELAP5/ MOD 3.2 FOR STEAM OF DOSE MODEUNG METHODS FOR DEMONSTRATION OF COM-CONDENSATION EXPERIMENTS IN THE PRESENCE OF PUANCE WITH THE RADIOLOGICAL CRITERIA FOR UCENSE TER-NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. MINATION. NOLLER.J.S. PARZER,L NUREGCR4662: DATING AND EARTHQUAKES: REVIEW OF QUA- NUREG/LA4145: RELAPS ASSESSMENT AGAINST PACTEL EXPERI-TERNARY GEOCHRONOLOGY AND ITS APPUCATION TO MENTAL DATA (REVISION 1). PALEOSElSMOLOGY. NOURSAKHSH,H.P. NUREG-1622: NRC ENFORCEMENT POUCY REVIEW. July 1995 - July r NUREGCR4416: RISK IMPORTANCE OF CONTAINMENT AND RE- 1997. LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. PELCHAT,J.M. ODETTE,G.R. NUREG-1631: SOURCE DISCONNECTS RESULTING FROM RADIOG-NUREG 1629: THE CHARACTERIZATION OF VICKER'S MICROHARD- RAPHY DRIVE CABLE FAILURES. Final Report. NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-HARDENING MICROPROBE. PENNELL,W.E. NUREGCR4551: IMPROVED EMBRITTLEMENT CORRELATIONS FOR NUREGCR-4219 V13 N2: HEAVY-SECTION STEEL TECHNOLOGY REACTOR PRESSURE VESSEL STEELS. PROGRAM. Semiannual Proereas Report For April - September 1996. NUREGCR-4219 V14 N1: HEAVY-SECTION STEEL TECHNOLOGY OH,J.G. PROGRAM.Sermannual Progrees Report For Ociober 1996 March NUREG/lA-0146: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF Iggy, TEN PERCENT MAIN STEAM-LINE-BREAK TEST RUN SB-SL-01. PEREZ-NAVAS,A. OLANO,C.S. NUREG/lA-0150: STUDY OF TRANSIENTS RELATED TO AMSAC AC-NUREGCR4552: MARBLE HILL ANNEAUNG DEMONSTRATION TUATION SENSITIVITY ANALYSIS. EVALUATION. NUREGCR4596: AN INVESTIGATION OF TENDON SHEATHING PERKINS.W.A. NU  : UVY A IS PRACTICES FOR NUCLEAR UR 2 POWER PLANT CONTAINMENT METALUC PRESSURE BOUND-ARIES. PETELIN,S. OLSON,R.J. o NUREGCR4540 STATE-OF-THE-ART REPORT ON PIPING FRAC- NUREG/iA 0145: RELAPS ASSESSMENT AGAINST PACTEL EXPERI-TURE MECHANICS. MENTAL DATA (REVISION 1). OTT,L.J- PHAM.T.N. NUREGCR4552: MARBLE HILL ANNEAUNG DEMONSTRAT ON NUREG-0430 V16: UCENSED FUEL FACluTY STATUS RE-EVALUATION. PORT. inventory Difference Data. July 1,1995 - June 30,1996.(Gray NUR R4537: INFLUENCE OF LONG TERM THERMAL AGING ON NU G 30 V17: UCENSED FUEL FACluTY STATUS RE-PORTInventory Difference Data. July 1,1996 June 30,1997.(Gray THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR PRESSURE VESSEL MATERIALS.An Atom Probe Study. PILCH,M.M. NUREGCR4475: RESOLUTION OF THE DIRECT CONTAINMENT NUR A 0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING LOFT LARGE BREAK LOCA TEST,LP-024. HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & BABCOCK & WILCOX PLANTS. PARK,H.G. NR R-54 : HOR BOLT BEHAVIOR AND STRENGTH DUR-R"EGCR4447: RESULTS OF CRACK-ARREST TESTS ON IRRADI-ATED A 506 CLASS 3 STEEL. PARK,H.S. PBSKURA,0. NUREG/lA-0147: ASSESSMENT OF RELAP5%IOD3.2 FOR STEAM CONDENSATION EXPERIMENTS IN THE PRESENCE OF NUREG-1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS ( NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. UCENSES. Pro 9 ram Specific Guidance About industrial Radiography U- j conses. Final Report. 1 PARK J.H. NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN PISKURA,D.A. UGHT-WATER REACTORS. Semiannual Report, January-June 1997. NUREG-1631: SOURCE DISOONNECTS RESULTING FROM RADIOG-  ; RAPHY DRIVE CABLE FAILURES. Final Report. PARK.J Y. NUREG/CR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO. PITARKA.A. GRAM. Annual Report.Auoust 1995 - September 1996. NUREGCR4593: CRUSTAL STRUCTURE AND GROUND MOTION NUREGOR4511 VD3: STEAM GENERATOR TUBE INTEGRITY PRO. MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM

  - GRAM.Somlannual Report. October 1996 March 1997.                 NATIONAL SEISMOGRAPH 6C NETWORK DATA.

NUREG/lA 0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING LOFT LARGE BREAK LOCA TEST LP-024. PODOSENOV.A.S. NUREG/lA 0142: INSTALLATl6N OF RELAP5/ MOD 3.2 ON 80486 AND NUREG/lA-0025: RELAPS/ MOD 3 SUBCOOLED BOlWNG MODEL AS-PENTIUM BASED PERSONAL COMPUTERS, SESSMENT, PARK,YJ. POLOSKI J.P. NUREGCR4564: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR NUREGCR4500 V01: RELIABluTY STUDY: AUXILIARY / EMERGENCY WALL INTERNATIONAL STANDARD PROBLEM. FEEDWATER SYSTEM,19671995.

p 1 Personal Author Index 49 POND,D.J. RASMUSON,0.M. l NUREGCR4605: AN EVALUATION OF HUMAN FACTORS RESEARCH NUREGCR 5485: GUIDEUNES ON MODEUNG COMMON-CAUSE l l FOR ULTRASONIC INSERVICE INSPECTION. FAILURES IN PROBABiUSTIC RISK ASSESSMENT. NUREGCR4497: COMMON CAUSE FAILURE PARAMETER ESTi-POPE,R. MATIONS. l NUREG 1608: CATEGOR! ZING AND TRANSPORTING LOW SPECIFIC NUREGCR4268 V01: COMMON CAUSE FAILURE DATABASE AND ! ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS. ANALYSIS SYSTEM. Volume 1:Overv6ew. NUREGOR4268 V02: COMMON CAUSE FAILURE DATABASE AND POSADA,J.M. ANALYSIS SYSTEM. Volume 2: Event Definition And Classifcation. NUREG/tA 0150: STUDY OF TRANSIENTS RELATED TO AMSAC AC- NUREGCR4268 V03: COMMON CAUSE FAILURE DATABASE AND TUATION SENSITMTY ANALYSIS. ANALYSIS SYSTEM. Volume 3: Data Collection And Event Codng. POWELL,C.A. RAVINDRA,M.K. NUREGCR4556: A COMPREHENSIVE STUDY OF THE EASTERN NUREGCR4544: METHODOLOGY FOR ANALYZING PRECURSORS TENNESSEE SDSMIC ZONE. TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE-QUENCES. l NUREGCR4475: RESOLUTION OF THE DIRECT CONTAINMENT RAVNIKAR.L I HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & NUREG/LA 0141: RESULT OF BETHSY TEST 9.1.B USING RELAPS/ BABCOCK & WILCOX PLANTS. MOD 3. POWERS,G.E. REID R.D. NUREG-1505 R01: A NONPARAMETRIC STATISTICAL METHOD- NUREG-1556 V09 DR FC: CONSOUDATED GUIDANCE ABOUT MATE-OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE- RIALS UCENSES. Program-Specific Guidance About Medcal Use LL j COMMISSIONING SURVEYS. Interim Report For Use And Comment. conses. Draft Report For Comment i NUREG-1507: MINIMUM DETECTABLE CONCENTRATIONS WITH I TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- REID.R.L TAMINANTS AND FIELD CONDITIONS. NUREGCR4577: U.S. NUCLEAR POWER PLANT OPERATING COST AND EXPERIENCE SUMMARIES. p NUREGCR4418: RISK IMPORTANCE OF CONTAINMENT AND RE- REMEC,L LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. NUREGCR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH-MARK. PRAWDZlK,0.A. NUREG/CR4600: NEUTRON EXPOSURE PARAMETERS FOR CAP-NUREG/CR4496: EVALUATION OF LOSS OF OFFSITE POWER SULE 10.05 IN THE HEAVY-SECTION STEEL IRRADIATION PRO-EVENTS AT NUCLEAR POWER PLANTS: 1980 - 1996. GRAM TENTH 1RRADIATION SERIES. NUREG/CR4601: NEUTRON EXPOSURE PARAMETERS FOR THE PRENDERGAST,K. DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRADIA-NUREG-1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS TION PROGRAM TENTH IRRADIATION SERIES. UCENSES. Program-Specific Guidance About industrial Radiography U-conses. Final Report. REVANKAR,5.T. NUREGOR4498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR-PURDY,0. CULADON TESTS IN THE PUMA FACluTY. NUREG-1556 V07 DR FC: CONSOUDATED GUIDANCE ABOUT MATE-RIALS UCENSES. PROGRAM-SPECIFIC GUIDANCE ABOUT ACA- REVETOS,F. DEMIC,RESEARCH AND DEVELOPMENT AND OTHER UCENSES NUREG/LA-0150 STUDY OF TRANSIENTS RELATED TO AMSAC AC-OF UMITED SCOPE. Draft Report For Comment. TUATION. SENSITIVITY ANALYSIS. PYLEY,$ S. RICH,T. NUREG/lA-0024: APPUCATON OF RELAP5/ MOD 3.1 CODE TO THE NUREG-1556 V03: CONSOLIDATED GUIDANCE ABOUT MATERIALS LOFT TEST L34. LICENSES.Appleatons For Sealed Source And Device Evaluation And l QUICK K.S. NUR 15 V0 SOUDATED GUIDANCE ABOUT MATERIALS l NUREG/CR4475: RESOLUTION OF THE DIRECT CONTAINMENT LICENSEES. Program-Specife Guidance Exempt Distribution U-HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & consesRnal Report. BABCOCK & WILCOX PLANTS. RIGGSBEE,E.T. RADCLIFTE,W.H. NUREGCR4501: ADVANCED INSTRUMENTATION AND MAINTE-l NUREG-1556 V04: CONSOUDATED GUIDANCE ABOUT MATERIALS NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. UCENSES. Program-Specife Guidance About Fixed Gauge U-conses. Final Report. RODRIGUEZ,M. NUREG.1556 V05: CONSOUDATED GUIDANCE ABOUT MATERIALS NUREG/CR4434: ANCHOR BOLT BEHAVIOR AND STRENGTH DUR-UCENSES. Program Specife Guidance About Self-Shielded irradiator ING EARTHOUAKES. Licenses. Final Report. RODRIGUEZ,S.B. RAMSDELL,J.V. NUREGCR4119 V01 R1: MELCOR COMPUTER CODE MANU-NUREG/CR4210 S01: COMPUTER CODES FOR EVALUATION OF ALS.Pnmer And Users' Guides. Version 1.8.4. July 1997. CONTROL ROOM HABITADUTY (HABIT V1.1). NUREGCR4119 V02 R1: MELCOR COMPUTER CODE MANU-ALS. Reference Manuals, Version 1.8.4. July 1997. RANDALL.K. NUREG-1556 V03: CONSOUDATED GUIDANCE ABOUT MATERIALS ROGE RS,J.W. UCENSES.Applicrtions For Scaled Source And Device Evaluation And NUREGCR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT Regletration. Final Report. PROGRAM. l RANSOM,V.H. ROGERS,R.D. f NUREG/CR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT I NUREG/CR4498: SINGLE-PHASE AND TWO PHASE NATURAL CIR-CULATION TESTS IN THE PUMA FACILITY. PROGRAM. RAPP M. ROQlNSKAGA,V.L. NUREG/lA-0024: APPUCATION OF RELAP5/ MOD 3.1 CODE TO THE j NUREG/CR4498: SINGLE-PHASE AND TWO-PHASE NATURAL C!R-CULATION TESTS IN THE PUMA FACluTY, LOFT TEST L34. RASHID,Y.R. ROHATGI,U.S. NUREG/CR-66711 PRETEST PREDICTION ANALYSIS AND POSTTEST NUREG/CR4359 V01: RAMONA-4B: A COMPUTER CODE WITH CORRELATION OF THE SIZEWELL-B 1:10 SCALE PRESTRESSED THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR CONCRETE CONTAINMENT MODEL TEST. SYSTEM TRANSIENTS.Models And Correlations, f 1

50 Personal Author index NUREGCR4359 V02: RAMONA-4B: A COMPUTER CODE WITH SCHUSTER,G.J. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR NUREGCR4471 V01: CHARACTERIZAT ON OF FLAWS IN U.S. RE-SYSTEM TRANSIENTS User's Manual. ACTOR PRESSURE VESSELS. Density And DistrbJtton Of Flaw Indi-NUREGCR4472: PREUMINARY PHENOMENA IDENTIFICATION AND cetions in PVRUF. RANKING TABLES FOR SIMPLIFIED BOluNG WATER REACTOR NUREGCR4594: EVALUSTION OF ULTRASONIC INSPECTION TECH-LOSS OF-COOLANT ACCIDENT SCENARIOS. NIQUES FOR COARSE-GRAINED MATERIALS. ROSSEEL T.M. SCHWARTZ,M. NUREGCR-5591 V0B N1: HEAVY-SECTION STEEL IRRADIATION NUREG 1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS PROGRAM Serniannual Progress Report For October 1996 Through UCENSES.Prograrn-Specife Guidance About industrial Radiography U-March 1997. censes. Final Report. RUSSELL.E.W. SCHWARTZ,M.E. NUREGCR-5502: ENGINEER!NG DRAWING 3 FOR 10 CFR PART 71 NUREG-1556 V05: CONSOLIDATED GUIDANCE ABOUT MATERIALS PACKAGE APPROVALS. UCENSES.Prograrn-Specife Guidance ADout Self-Shielded irradator RUSSELL,K.F. NUREGCR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON SCHWElG.E.S. THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR NUREG/GR-0017: DATING OF UQUEFACTION IN THE NEW MADRID PRESSURE VESSEL MATERIALS.An Atorn Probe Study. SEISMIC ZONE AND IMPUCATIONS FOR EARTHQUAKE HAZARD. RUTHER,W.E. SCOTT,P.M. NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4540: STATE-OF-THE-ART REPORT ON PIPING FRAC-UGHT-WATER REACTORS. Semiannual Report. January June 1997. TURE MECHANICS. RUTHER,W.K. SERNE,RA NUREGCR-4667 V25: ENVIRONMENTALLY ASSISTED CRACKING IN NUREGCR4377: EFFECTS ON RADIONUCLIDE CONCENTRATIONS UGHT-WATER REACTORS. Semiannual Report. July-Decernber 1997. BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF SAIKIA.C.K. PERFORMANCE ASSESSMENT OF LOW-LEVEL RADIOACTIVE NUREGCR4593: CRUSTAL STRUCTURE AND GROUND MOTION WASTE DISPOSAL FACILITIES. MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM SEUL,K.W NATIONAL SEISMOGRAPHIC NETWORK DATA. NUREG/IA 0143: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE LSTF SALASA,J.K. EXPERIMENT S!MULATING A LOSS OF RESIDUAL HEAT REMOVAL NUREGCR4544: METHODOLOGY FOR ANALYZING PRECURSORS EVENT DURING MID-LOOP OPERATION. TO EARTHQUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE- NUREG/tA 0144: ASSESSMENT OF RELAP5/ MOD 3 2 WITH THE SEMISCALE NATURAL CIRCULATION EXPERIMENT, S-NC48. QUENCES' NUREG/lA-0147: ASSESSMENT OF RELAP5/ MOD 3.2 FOR STEAM SALTZMAN.J. CONDENSATION EXPERIMENTS IN THE PRESENCE OF NUREGCR4617: THE PRICE-ANDERSON ACT - CROSSING THE NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. NUREG/lA 0148. ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF TEN-PERCENT MAIN STEAM-UNE-BREAK TEST RUN SB-SL-01. SAMANTA,P.K. URE An nelys s F N P s. NU E 0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING LOFT LARGE BREAK LOCA TEST,LP-02-6. SANCHEZ. BAPTIST NUREG/lA 0150: STUDY OF TRANSIENTS RELATED TO AMSAC AC- SEXTON,C.D. TUATION SENSITIVITY ANALYSIS. NUREGCR-5501: ADVANCED INSTRUMENTATION AND MAINTE-NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. SANDERS,R.L. NUREG/CR4119 V01 R1: MELCOR COMPUTER CODE MANU- SHACK,W.J. ALS. Primer And Users' Guedes. Version 1.8.4.Jufv 1997. NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN NUREGCR4119 V02 R1: MELCOR COMPUTER CODE MANU- LIGHT-WATER REACTORS. Semiannual Report. January-June 1997. ALS. Reference Manuals, Version 1.8.4.Jufy 1997. NUREG/CR-4667 V25: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT-WATER REACTORS. Semiannual Report. July-December 1997. SANDIN.S.S. NUREGCR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO-NUREG-1022 RO1: EVENT REPORTING GUIDELINES 10 CFR 50.72 GRAM. Annual Report, August 1995 - September 1996. AND 50.73. NUREG/CR-6511 V03: STEAM GENERATOR TUBE INTEGRITY PRO-GRAM. Semiannual Report, October 1996 - March 1997. SANECKI,J.E. NUREGCR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR-NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT UGHT-WATER REACTORS. Serniannual Report. January June 1997. HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH SANFORD,W'E. MODEL PREDICTIONS. NUREGCR-6569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT NUREG/CR4583 EFFECTS OF LWR COOLANT ENVIRONMENTS ON PROGRAM. FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY STEELS. SANTOS,C. NUREG-1629: THE CHARACTER!ZATION OF VICKER'S MICROHARD- SHEAFFER,M.K. NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN. NUREGCR-5502: ENGINEERING DRAWINGS FOR 10 CFR PART 71 HARDENING MICROPROBE. PACKAGE APPROVALS. SANTOS,C.G. SHORT,C.M. NUREG-1426 V03: COMPILATION OF REPORTS FROM RESEARCH NUREGCR-6600: INVESTIGATION OF TECHNIQUES FOR THE DE-SUPPORTED BY THE ELECTRICAL, MATERIALS AND MECHANICAL VELOPMENT OF SEISMIC DESIGN BASIS USING THE PROB-ENGINEERING BRANCH, DIVISION OF ENGINEERING. ABluSTIC SEISMIC HAZARD ANALYSIS. SCARBROUGH,T.G. SIBOL.M.S. NUREGCP-0152 V02: PROCEEDINGS OF THE FIFTH NRC/ASME NUREG/CR4556: A COMPREHENSIVE STUDY OF THE EASTERN SYMPOSIUM ON VALVE AND PUMP TESTING. TENNESSEE SEISMIC ZONE. SCHROETER.B. SIMONEN,F.A. NUREG-1629: THE CHARACTERIZATION OF VICKER'S MICROHARD- NUREG/CR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN- PROBABluTIES OF DEFECTS IN REACTOR PRESSURE VESSEL HARDENING MICROPROBE. WELDS.

I l i Personal Author Index 51 i ! NUREG/CR4594: EVALUATION OF ULTRASONIC INSPECTION TECH- SULLIVAN.T.M. j NIQUES FOR COARSE 4 RAINED MATERIALS. NUREG/CR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT SLOVIK,0.C. l NUREG/CR4472: PREUMINARY PHENOMENA IDENTIFICATION AND SUMMERS R.M. I l RANKING TABLES FOR SIMPUFIED BOILING WATER REACTOR NUREG/CR4119 V01 R1: MELCOR COMPUTER CODE MANU-l LOSS 4r-COOLANT ACCIDENT SCENARIOS. ALS. Primer And Users' Guides, Version 1.8.4 1997. S M TH,8, NUREG/CR4119 V02 RI: MELCOR COMP ER CODE MANU-i ALS. Reference Manuals, Version 1.8.4. July 1997. NUREG-1556 V03: CONSOUDATED GUIDANCE ABOUT MATERIALS l' UCENSES.Apphcotions For Sealed Source And Device Evaluation And SWINTH,K.L. , Re96etre 6on. Final Report. NUREG/CP-0164: PROCEEDINGS OF THE WORKSHOP ON ELECTRIC l DOSIMETRY. Held in Gaithersburg. Maryland On October 14-16, 1997. , SMTH,J.L l NUREG/CR4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN TAGAWA H. LIGHT WATER REACTORS. Semiannual Report. January 4une 1997. NUREG/CR4509: THE EFFECT OF INITIAL TEMPERATURE ON NUREG/CR4667 V25: ENVIRONMENTALLY ASSISTED CRACKING IN FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION UGHT-WATER REACTORS. Semiannual Floport, July-December 1997. TRANSITION PHENOMENON. NUREG/CR4524: THE EFFECT OF LATERAL VEN'ING ON DEFLA-SMITH,N.F. GRATION-TO DETONAT ION TRANSITION IN HYDROGEN-AIR-

NUREG/CR4593: CRUSTAL STRUCTURE AND GROUND MOTION STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. I l MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM NATIONAL SEISMOGRAPHIC NETWORK DATA. TANAKA,T.J.

NUREG/CR4479. TECHNICAL BASIS FOR ENVIRONMENTAL QUAU-SMITH,R.C. FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED l NUREGiCR4119 V01 R1: MELCOR COMPUTER CODE MANU- EQUIPMENT IN NUCLEAR POWER PLANTS. 1997.- ALS. Primer AndV02 NUREG/CR4119 Users' R1: Guides, MELCOR version COMPui1.8.4. July _ERMANU-CODE TAYLOR,T.M. ALS. Reference Manuals, Version 1.8.4, July 1997. NUREG-1556 VII DR FC: CONSOLIDATED GUIDANCE ABOUT MATE-RIALS LICENSES. Program-Specific Guidance About Specific Ucenses ! SOMERVILLE,P.G. Of Broadacope. Draft Report For Comment. I NUREG/CR4593: CRUSTAL STRUCTURE AND GROUND MOTION MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM THlo,H.K. NATIONAL SEISMOGRAPHIC NETWORK DATA. NUREG/CR4593: CRUSTAL STRUCTURE AND GROUND MOTION MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM SOPPET,W.K. NATIONAL SEISMOGRAPHIC NETWORK DATA. NUREG/CM667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN i UGHT-WATER REACTORS. Semiannual Report, January 4une 1997. THOMAS,G.R. l NUREG/C!&4667 V25: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS- , UGHT-WATER REACTORS. Semiannual Report. July-December 1997. TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP-l PING CASK DESIGN REVIEW. User's Manual to Versen 3a. i SOWERS,J.M. NUREG/CR-5502: ENGINEERING D?tAWINGS FOR 10 CFR PART 71 ( NUREG/CR-5562: DATING AND EARTHOUAKES: REVIEW OF QUA- PACKAGE APPROVALS.

TERNARY GEOCHRONOLOGY AND ITS APPUCATION TO I PALEOSEISMOLOGY. THOMAS,M.L.

l NUREG-0713 V18: OCCUPATIONAL RADIATION EXPOSURE AT COM-l STAGE,8.A. MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILI- ! NUREG/CR4210 S01: COMPUTER CODES FOR EVALUATION OF TIES 1996. Twenty Ninth Annual Report. CONTROL ROOM HABITABluTY (HABIT V1.1). NUREd 0713 V19: bCCUPATIONAL RADIATION EXPOSURE AT COM- ! MERCIAL NUCLEAR POWER REACTORS AND OTHER FACluTIES STETKAR,J.W. 1997. Thirtieth Annual Report. NUREG/CR-5496: EVALUATION OF LOSS OF OFFSITE POWER EVENTS AT NUCLEAR POWER PLANTS: 1980 1996. THOMPSON D.L NUREG/CR-5559: SINGLE- AND CROSS-HOLE PNEUMATIC TESTS IN ! STEUTEVILLE,8. UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-

NUREG-1634

1997 LOST SOURCE EXERCISE.An Exercise Of Radio- SEARCH SITE: PHENOMENOLOGY,SPATLAL VARIA-lo9ical Response Through Cooperation And Coordinahan Of BluTY,CONNECTIVITY AND SCALE.

Local. State And Federal Resourceo under The National Conhngency I pian, THOMPSON,T. NUREG-1556 V10 DR FC: CONSOUDATED GUIDANCE ABOUT MATE- [ STOLLER,R.E. RIALS UCENSES. Program Specific Guidance About Master Matenals l NUIEG/CR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON Ucenses. Draft Report for Comment ' THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR PRESSURE VESSEL MATERIALS.An Atom Probe Study. TRAY S E 1521 DAFT FC: TECHNICAL REVIEW OF RISK-INFORMED, ST RA~",R.V. PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT ! FIRE PROTECTION ANALYSES. Draft Report For Comment NUREG/CR4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN l LIGHT-WATER REACTORS. Semiannual Report.Janua une 1997. l NUREG/CR4667 V25: ENVIRONMENTALLY ASSISTED RACKING IN TRUMMER,D.J. I UGHT-WATER REACTORS. Semiannual Report, July-December 1997, NUREG/CR4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS-l TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP-l STRUCKMEYER R. PING CASK DESIGN REVIEW. User's Manual to Version 3a. ! NUREG-0837 V17 NO3: NRC TLD DIRECT RADIATION MONITORING NURE R4521: ESTIMATING PROBABLE FLAW DISTRIDUTIONS IN STUART,D.S. PWR STEAM GENERATORS. NUREG/CR4119 VD1 R1: MELCOR COMPLTTER CODE MANU-1997. TUTTLE M.P. ALS. Primer And NUREG/CR4119 V02Users' R1: Guides. MELCOR Version COMPUT 1.8.4. July _ER MANU-CODE NUREG/GR 0017: DATING OF UQUEFACTION IN THE NEW MADRID ALS. Reference Manuals, Version 1.8.4, July 1997. SEISMIC ZONE AND IMPUCATIONS FOR EARTHOUAKE HAZARD. SULLAWAY,M.F. VACCA,P.C. NUREG/CR4671: PRETEST PREDICTION ANALYSIS AND POSTTEST NUREG 1556 V05: CONSOUDATED GUIDANCE ABOUT MATERIALS CORRELATION OF THE SIZEWELL B 1:10 SCALE PRESTRESSED UCENSES. Program-Spect'ic Guidance About Self-Shielded irradator CONCRETE CONTAINMENT MODEL TEST. Ucenses. Final Report. L __ _ __ _ _

52 Personal Author index VESELY,W.E. WILLIS,C.A. NUREGOR4579: DIGITAL l&C SYSTEMS IN NUCLEAR POWER NUREG-1633 DRFT FC: ASSESSMENT OF THE USE OF POTASSIUM PLANTS. Risk-Screerung Of Environmental Stressors And A Compark IODIDE (K!) AS A PUBUC PROTECTIVE ACTION DURING SEVERE son Of Hardware Unavallatility With An Existing Analog Systern. REACTOR ACCIDENTS. Draft Report For Comment VESSELINOV,V.V. WITTE*M'C' NUREGCR4559: SINGLE- AND CROSS-HOLE PNEUMATIC TESTS IN NUREGCR4608:

SUMMARY

AND EVALUATION OF LOW-VELOCITY l UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-PHENOMENOLOGY, SPATIAL VARIA- IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. SEARCH SITE: BluTY,CONNECTIVITY AND SCALE. WOOD,R.T. VISKANTA,R. NUREGCR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAll-NUREGCR4498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- FiCATION OF MICROPROCESSOR-BASED SAFETY 4tELATED l EQUIPMENT IN NUCLEAR POWER PLANTS. I CUI.ATION TESTS IN THE PUMA FACluTY. VLAHOVIC,G. WOOD,8.T. NU,REGCR4556: A COMPREHENSIVE STUDY OF THE EASTERN NUREGCR4268 V04: COMMON CAUSE FAILURE DATABASE AND

    ,ENNESSEE SEISMIC ZONE.                                                   ANALYSIS SYSTEM. Volume 4: Software Reference Manual.

GS WOODS.S.S. CAS'R NU G/GR-0016: THE ROLE OF TIME-DEPENDENT DEFORMATION NUREGCR4593: CRUSTAL STRUCTURE AND GROUND MOTION / I IN INTERGRANULAR CRACK INITIATION OF ALLOY 600 STEAM ' GENERATOR TUBING MATERIAL MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM NATIONAL SEISMOGRAPHIC NETWORK DATA. CATKINS,J.C. NUREOCR4611: RESULTS OF PRESSURE LOCKING AND THERMAL WRIGHT J.E. BINDING TESTS OF GATE VALVES. NUREGCR4551: IMPROVED EMBRITTLEMENT CORRELATIONS FOR CATSON,G.M. NUREG-1556 V04:CONSOUDATED GUIDANCE ABOUT MATERIALS WU,J.S. UCENSES. Program-Spec 6fc Guidance About Fixed Gauge Lk NUREGCR4544: METHODOLOGY FOR ANALYZING PRECURSORS conseafinal Report TO EARTHQUAKE-INITIATED AND FJRE-INITIATED ACCIDENT SE-WESER,C.F. QUENCES. NUREGCR4408: TECHNICAL ASSISTANCE IN REVIEW OF SOURCE XU,Y. TERM 4 ELATED ISSUES OF ADVANCED REACTORS. NUREGCR4599: LODINE V6LATluTY AND PH CONTROL IN THE AP- NUREGCR4498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR-600 REACTOR. CULATION TESTS IN THE PUMA FACluTY. WEINSTEIN,E. YAMAMOTO,T, NUREG-1634: 1997 LOST SOURCE EXERCISE.An Exercise Of Rad

• NUREG 1629: THE CHARACTERIZATION OF VICKER'S MICROHARD-logical Response Through Cooperation And Coordination Of NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-Local. State And Federal Resources Under The National Contingency HARDENING MICROPROBE.

Plan. YANG,J.W. NU 1552 S01 DR FC: FIRE BARRIER PENETRATION SEALS IN NUREGCR4502: ACTION REQUIREMENTS FOR AFW SYSTEM FAIL-NUCLEAR POWER PLANTS. Draft Report For Comment NURE 46 : ISK PA N S UUNG PREVENTIVE WESTRICH,H.R. MAINTENANCE DURING SHUTDOWN VS. DURING POWER OPER-NUREGCR4603: CHARACTERIZATION OF RETARDATION MECHA- ATION FOR PWRS. NISMS IN SOIL YOUNG,M.F. WHITE,D. NUREGCR4119 V01 R1: MELCOR COMPUTER CODE MANU-NUREG 1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS ALS. Primer And Users' Guides, Version 1.8.4. July 1997. UCENSES. Program-Specific Guidance Aboul Industrial Radiography LL NUREGCR-6119 V02 R1: MELCOR COMPUTER CODE MANU-connes. Final Repat. ALS. Reference Manuals, Version 1.8.4,Juty 1997. WHffE.MS. Y N NURE GROUND-WATER MODELS IN SUPPORT OF R R4613 V01: CODE MANUAL FOR MACCS2. User's Guide. NUREGOR4613 V02: CODE MANUAL FOR MACCS2. Preprocessor WHITE V.S. Codes COMIDA2, FGRDCF IDCF2. NUREGCR4577: U.S. NUCLEAR POWER PLANT OPERATING COST AND EXPERIENCE SUMMARIES. ZALUZEC,N.J. NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKlfeG IN

                                                                                                ^         ' ""                   '"    "*

• NURE 4 9: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM. ZHANG,Y-G. (

WILK0WSKI,0.M. NUREGCR4434: ANCHOR BOLT BEHAVIOR AND STRENGTH DUR- { NUREGCR4540: STATE OF-THE-ART REPORT ON PIPING FRAC, ING EARTHOUAKES. S' ZIMMERMAN,G.P. WILLIAMS.M.D. NUREGCR4549: ENVIRONMENTAL ASSESSMENT RENEWAL OF NUREGCR-6621: GROUND-WATER MODELS IN SUPPORT OF MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-NUREGCR4512. GlONAL SERVICE FACILITY, WAMPUM, PENNSYLVANIA.

1 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 aro welcome. 10 CFR Port 71 Accountabimy Report NUREG/CR4502: ENGINEERING DRAWINGS FOR 10 CFR PART 71 NUREG-1542 V03: ACCOUNTABILITY REPORT FISCAL YEAR 1997. PACKAGE APPROVALS. A SOS Class 3 Steel NUREG/CR4614: FEASIBILITY OF HIGH FREQUENCY ACOUSTIC IM-NUREG/CR4447: RESULTS OF CRACK ARREST TESTS ON IRRADI- AGING FOR INSPECTION OF CONTAINMENTS. ATED A 508 CLASS 3 STEEL ACMS Report NUREG-0544 RO4: NRC COLLECTION OF ABBREVIATIONS. NUREG-1125 V19: A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.1997 Annual. Adv tor AEOO TEAM-RELATED ISSUES OF ADVANCED REACTORS. NUREG-1272 V10 N01: OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA.1996 Annual Report. Advloory Committee On Nuclear Weste NUREG-1272 V10 NO2: OFFICE FOR ANALYSIS AND EVALUATION OF NUREG-1423 V08: A COMPILATION OF REPORTS OF THE ADVISORY OPERATONAL DATA.1996 Annual Report. COMMITTEE ON NUCLEAR WASTE. July 1997 - June 1998. NUREG-1272 V10 NO3: OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA.1996 Annual Report. Afterheet Rete Calculation NUREG-1272 VII N01: OFFICE FOR ANALYSIS AND EVALUATION OF NUREG/CR-6536: VERIFICATION OF THE LWRARC CODE FOR OPERATIONAL DATA.1997 Annual Report (Reactors . UGHT-WATER-REACTOR AFTERHEAT RATE CALCULATIONS. NUREG-1272 VII NO2: OFFICE FOR ANALYSIS AND) EVALUATION OF N RE 2 V CE Y AND L TION OF EG/CR4412: AGING AND LOSS-OF COOLANT ACCIDENT (LOCA) OPERATONAL DATA.1997 Annual Report (Technical Training). gglglbE E BASED APPROACH TO AFW System STRUCTURAL DETERIORATION AND RELIABILITY, NUREG/CR-5500 V01: REUABluTY STUDY; AUXIUARY/ EMERGENCY FEEDWATER SYSTEM, 1987-1995. Al g ALARA IN INTERGRANULAR CRACK INITIATION OF ALLOY 600 STEAM NUREG-1556 V03: CONSOUDATED GUIDANCE ABOUT MATERIALS GENERATOR TUSING MATERIAL UCENSES. Applications For See6ed Souece And Device Evaluation And UE 4 579: DIGITAL I&C SYSTEMS IN NUCLEAR POWER ALARON Corporat6on PLANTS. Risk-Screening Of Environmental Stressors And A Cornpart-NUREG/CR4549: ENVIRONMENTAL ASSESSMENT RENEWAL OF son Of Hardware Unavailability With An Existing Analog System. MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-GeONAL SERVICE FACluTY, WAMPUM, PENNSYLVANIA. A g g g AP900 ING EARTHOUAKES. NUREG-1632; EVALUATION OF AP600 CONTAINMENT THERMAL-HY-DRAUUC PERFORMANCE. Annual Report NUREG/CR-6599: IODINE VOLATILITY AND PH CONTROL IN THE AP. NUREG-1272 V10 NO2: OFFICE FOR ANALYSIS AND EVALUATION OF 600 NOR. NURE V1 NF C OR SIS AND EVALUATION OF ATWS OPERATIONAL DATA.1996 Annual Report. NUREG/lA 0150: STUDY OF TRANSIENTS RELATED TO AMSAC AC. TUATION. SENSITIVITY ANALYSIS. Apu E -5559: SINGLE- AND CROSS-HOLE PNEUMATIC TESTS IN Abbrewletion UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-NUREG-0544 A04: NRC COLLECTION OF ABBREVIAT!ONS- SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA-BluTY,CONNECTIVITY AND SCALE. NUREG 0090 V20: REPORT TO CONGRESS ON ABNORMAL OCCUR- Atom Probe RENCES. Fiscal Year 1997. NUREG/CR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR NU G43N V23 N01:ABSTRACTS FOR PUBUCATIONS IN THE "#* NUREG-SERIES. Semiannual Compilation for January, June 1998. Atomic Safety And Licensino Board NUREG-1363 V07: ATOMIC SAFETY AND UCENSING BOARD BIEN-NU 4 V04 R1: SCDAP/RELAPS/ MOD 3.2 CODE MAN- *

• UALMATPRO - A Ubrary Of Matenals Properties For Ught-Water Re- Auxillery Feedwater actor Accident Analysis. NUREG/CR-5500 V01: REUABILITY STUDY: AUXIUARY/ EMERGENCY

                     ,                                                        FEEDWATER SYSTEM, 1987 1995.

NUREG/CR-4674 V25: PRECURSORS TO POTENTIAL SEVERE CORE Auxliiery Foodwater System DAMAGE ACCIDENTS: 1996. A Status Report. NUREG/CR4502: AulON REQUIREMENTS FOR AFW SYSTEM FAIL-NUREG/CR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE URES.An Analysis For Four Nuclear Power Plants. DAMAGE ACCIDENTS:1997. A Status Report. NUREG/CR4544: METHODOLOGY FOR ANALYZING PRECURSORS BETHSY TO EARTHOUAKE-lNITIATED AND FIRE-INITIATED ACCIDENT SE- NUREG/lA4141: RESULT OF BETHSY TEST 9.1.8 USING RELAP5/ OUENCES. MOD 3. t 53

54 SubjoCt index BWR NUREGOR4150 V02 R1: SCDAP/RELAP5/ MOD 32 CODE MAN-NUREG-1123 R02: KNOWLEDGE AND ABluTIES CATALOG FOR NU- UAL.Demeos Progression Model Theory. CLF.AR POWER PLANT OPERATORS Boiling Water Reactors. NUREG/CR4150 v03 R1: SCDAP/RELAPS/ MOD 3.2 CODE MAN-NUREG/CR4472: PREUMINARY PHENOMENA IDENTIF CATION AND UALUser's Guide And input Manuel. RANKING TABLES FOR SIMPUFIED BOILING WATER REACTOR NUREG/CR4150 VOS R1: SCDAP/RELAPS/ MOD 32 CODE MAN-LOSSCF-COOLANT ACCIDENT SCENARIOS. UALDevelopmental Assessment

                                                                                                                                                           ]

Bebook & Wilcon Plant Cold Lee NUREG/CR4475: RESOLUTON OF THE DIRECT CONTAINMENT NUREG/lA4024: APPUCATION OF RELAP5/ MOD 3.1 CODE TO THE HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & LOFT TEST L34. BABCOCK & WILCOX PLANTS. Besentent Faulte NUREG/CR4558: A COMPREHENSIVE STUDY OF THE EASTERN NUREG/CR4570: APPUCATION OF THE NCSA HABANERO TOOL ) FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS-  ! TENNESSEE SEISMIC ZONE. MENTS. Behavior Prediction Combustion Engineert Plant NUREG/iA4025: RELAP5/ MOD 3 SUSCOOLED BOILING MODEL AS- NUREG/CR4475: RE UTON OF THE DIRECT CONTAINMENT SESSMENT, HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & BABCOCK & WILCOX PLANTS. Bachewk NUREG/CR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH- Common Cause Fellure MARK. NUREG/CR4485: GUIDELINES ON MODEUNG COMMON-CAUSE g,gg,g wm, p., ego, FAILURES IN PR06ABILISTIC RISK ASSESSMENT. NUREG-1123 R02: KNOWLEDGE AND ABluTIES CATALOG FOR NU- NUREG/CR-5497: COMMON CAUSE FAILURE PARAMETER ESTb N R4472 R M NARY NOME A E I ASON AND NU E R 6268 V01: COMMON CAUSE FAILURE DATABASE AND RANKING TABLES FOR SIMPUFIED BOluNG WATER REACTOR NU E 42 V : CAU$ FAILURE DATABASE AND LOSS-OF COOLANT ACCIDENT SCENARIOS. ANALYSIS SYSTEM.Vo6ume 2: Event Definition And Closemcation. Bolt NUREG/CR4268 V03: COMMON CAUSE FAILURE DATABASE AND NR R44 : HOR BOLT BENVIOR AND STRENGTH DUR. A{SjS STEM. Volume Date gEgg ANALYSIS SYSTEM. Volume 4: Software Reference Manuel. G 6 V11 DR FC: CONSOUDATED GUIDANCC ABOUT MATE. CompHetion RIALS UCCNSES. Program-Specific Guidance About Specife Ucenses NUREG-0304 V23 N01: ABSTRACTS FOR PUBUCATIONS IN THE Of . Drn8 Report For Comment. NUREG-SERIES.Semiennual Complistion for January 4une 1998. BudO8, M .elve Record NUREG-1100 V14: BUDGET ESTIMATES. Fiscal Year 1999. NUREG-0910 R03: NRC COMPREHENSIVE RECORDS DISPOSITON SCHEDULE. Bypees Event COGYMA NUREG/CR4359 V01: RAMONA-4B: A COMPUTER CODE WITH NUREG/CR4555 V01: PROBABILISTIC ACCIDENT CONSEQUENCE THREE-DIMENSONAL NEUTRON KINETICM FOR BWR AND SBWR UNCERTAINTY ANALYSIS.Lete Health Effects Uncertelnty Assess. SYSTEM TRANSIENTS.Models And Correlatons, ment. Mein Reoort. NUREG/CR4350 '/02: RAMONA-4B: A COMPUTER CODE WITH NUREG/CR4555 V02: PROBABILISTIC ACCIDENT CONSEQUENCE THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR UNCERTAINTV ANALYSIS.Lele Health Effects Uncertainty Assess- SYSTEM TRANSIENTS. Users Manuel. ment.Appendcas NUREG/CR4534 V02: FRAPCON 3: A COMPUTER CODE FOR THE CALCULATION OF STEADY-STATE. THERMAL-MECHANICAL BE-UR 4608:

SUMMARY

AND EVALUATION OF LOW VELOCITY NU G/CR :F 3 E L ESSMENT. IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. Concrete R 94: EVALUATION OF ULTRASONIC INSPECTION TECH- N R A S NIQUES FOR COARSE GRAINED MATERIALS. NUREG/CR45W8: AN INVESTIGATION OF TENDON SHEATHING CwWcotes Of CompHmco FILLER MIGRATON INTO CONCRETE. NUREG 0383 V01 R21: DIRECTORY OF CERTIFICATES OF COMPLk Concrete Pod ANCE FOR RADCACTIVE MATERIALS PACKAGES. Report Of NRC-NUREG/CR4608:

SUMMARY

AND EVALUATON OF LOW-VELOCITY NU G 02 R2'1: DIRECTORY OF CERTIFICATES OF COMPLL IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certifcetes Of Confocal M6croscopy nce. NUREG-1629: THE CHARACTERIZATION OF VICKER'S MICROHARD-NURE 3 V03 R18: DIRECTORY OF CERTIFICATES OF COMPU-ANCE FOR RADOACTIVE MATERIALS PACKAGES. Report Of NRC-NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-HARDENING MICROPROBE. Approved Quality Assurance Programs For Redonctive Materials Peck-ages. Co. - _ m Analyele g NUREG/CR4613 V01: CODE MANUAL FOR MACCS2. Users Guide. NUREG/CR4551: IMPROVED EMBRITTLEMENT CORRELATIONS FOR Consolidated Guldence REACTOR PRESSURE VESSEL STEELS. NUREG-1556 V02: CONSOLOATED GUIDANCE ABOUT MATERIALS Circulsele UCENSES.Progrem Spectic Guldence About industrial Radiography LL NUREG/lA4144: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE SEMISCALE NATURAL CIRCULATION EXPERIMENT, S-NC-88. NkE 1 V 3 CONSOUDATED GUIDANCE ABOUT MATERIALS l UCENSES. Applications For Seeled Source And Device Evolustion And 1 CoerWned Meterial Rooistration. Final Report. l NUREG/CR4594: EVALUATION OF ULTRASONIC INSPECTION TECH- NUREG-1556 V04: CONSOUDATED GUIDANCE ABOUT MATERIALS j NtOUES FOR COARSE-GRAINED MATERIALS. UCENSES. Program Specife Guldence About Fixed Gauge Lk - connes. Final Repon. Code Manuel NUREG-1556 VOS; CONSOUDATED GUIDANCE ABOUT MATERIALS NUREG/CR4150 V01 R1: SCDAP/RELAPS/ MOD 3.2 CODE MAN- UCENSES. Program-Specl1ic Guldence About Self-Shielded irredator UALintertece Theory. Ucenses. Final Report.

1 I Subject lfHlex 55 NURE41886 V07 DR FC: CON 9000ATED GUIDANCE ABOUT MATE- Data Collessen RIALS LICENBES. PROGRAM 4PECIFC GUIDANCE ABOUT ACA. NUREGCR4268 V03: COMMON CAUSE FAILURE DATABASE AND

            - DEMC.RESEARCH AND DEVELOPMENT AND OTHER LCENSES                                                 ANALYSIS SYSTEM. Volume 3: Data Collection And Event Codmg.
          - OF UMITED 8 COPE. Droit Report For Comment NUREG-1858 V08: CON 800 DATED GUIDANCE ABOUT MATERIALS -                                       Date Quality CBjoseve                                                         a UCENBEES. Program Spec 8c Guldence Emempt Distreunion U.                                       NUREG-1676: MULTI-AGENCY RADIATION SURVEY AND SITE NVES-                    l
                     . Final Report.                                                                          TIGATION MANUAL (MARSSIM). Final Report.                                   !

i 1888 Vos DR FC: CONSOUDATED GU6 DANCE ABOUT MATE-h

        ' NUN 1888 V DR FC: CONSOUDATED GUIDANCE ABOUT MATE.

NURE41307 600: REPORT ON WASTE SURIAL CHARGES.Changee in C----1 4 Weste Disposal Coeto At Low-Level Weste Bur 6el I l NUR G- 606 R01: A NONPARAMETRIC STATISTICAL METHOD-888 V1 FC: CONSOUDATED GUIDANCE ABOUT MATE. OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-  ! Report NU 7 iMUM ABL NTRA IONS l'TH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON-li Centset SIReeture AenemBey TAMtNANTS AND FIELD CONDITIONS. NUREGCM617: ROUND ROSN PRETEST ANALYSES OF A STEEL NUREG-1548 DRFT FC: DECISION METHODS FOR DOSE ASSESS-CONTANMENT VESSEL MODEL AND CONTACT STRUCTURE AS- MENT TO COMPLY WITH RADIOLOGICAL CRITERIA FOR LICENSE SEMBLY SUBJECT TO STATIC NTERNAL PRESSUR4ZATION. yE W For . Dentainmeng TIGATION MANUAL . Final Report. ' i NUREG 1832: EVALUATION OF AP900 CONTAINMENT THERMAL-HY, NUREG-1626 DRF FC: D STANDARD TECHNICAL SPEC 4- ! DRAUUC PERFORMANCE. FCATIONS FOR PERMANENTLY DEFUELED WESTINGHOUSE NUREGCR4871: PRETEST PREDICTION ANALYSIS AND POSTTEST PLANTS. Draft Report For Comment. CORRELATION OF THE SIZEWELL 81:10 SCALE PRESTRESSED NUREG-1628 DRF FC: STAFF RESPONSES TO FREQUENTLY ASKED i CONCRETE CONTAINGENT MODEL TEST. QUESTIONS CONCERNWG DECOMMISSIONING OF NUCLEAR I NUREGCR4646: A DAMAGE MECHANICS BASED APPROACH TO POWER PLANTS. Droit For Comment. I RAL DETERIORATION RE UTY. NUREGCP 0181: PROCEE OF THE WORKSHOP ON REVIEW i OF DOSE MODEUNG METHODS FOR DEMONSTRATION OF COM-NURE 14: FEAS181UTY OF FRE NCY ACOUSTIC IM-AGING FOR NSPECTION OF CONTAINMENTS. PUANCE WITH THE RADCLOGICAL CRITERIA FOR LICENSE TER- I l NUREGOR4815: A SURVEY OF REPAIR PRACTICES FOR NUCLEAR MtNATION. I POWER PLANT CONTANMENT METALUC PRESSURE SOUND- NUREGCR4621: GROUND-WATER MODELS IN SUPPORT OF ARIES. - NUREGrCR-6612. NUREGCR4384: HUMAN PERFORMANCE IN RADCLOGICAL SUR-Centelament PeNure ' VEY SCANNNG. Centred Reem NUREGCR4210 801: 03MPUTER CODES FOR EVALUATION OF U_FIEGCR4600: NU -__ To Deteneden THE EFFECT Treneltlen OF INITIAL TEMPERATURE ON FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION  ; i CONTROL ROOM NUREGOR4004: RADTRAD: HA8tTA81UTY A SIMPUFIED .(HABIT V1.1)MODEL FORTRANSITION RADIO- PHENOMENON. <

           - NUCLIDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION.

Desien Bede Aeoident Coolant Sartrennent NUREGC04210 801: COMPUTER CODES FOR EVALUATION OF NUMEGCR4603: EFFECTS OF LWR COOLANT ENVIRONMENTS ON CONTROt. ROOM HABITA84WTY (HABIT V1.1). FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY NUREGCR 6408: TECHNICAL ASSISTANCE IN REVIEW OF SOURCE N EG  : lODIN V L TY A IN THE AP- , Corseelen - 800 REACTOR. l l NURE 7 V24: ENVIRONMENTALLY ASSISTED CRACKING N NUREGAA-0141: RESULT OF BETHSY TEST 9.1.8 USING RELAP5/ i UGHT-WATER REACTORS. Semlennua! ReportM 1997. MOD 3. NUMEGCRm7 V26: ENVIRONMENTALLY ASSISTED CRACKNG H " UGHT-WATER REACTORS. Semiennual Report July-Decomtier 1997. P--- NUf (EGCR4003: CHARACTERIZATION OF RETARDATION M Cost Sellmate - NISMS W SOIL NUMEG-1307 R08: REPORT ON WASTE BURIAL CHARGES.ChanDes in Decommiseloning Weste Disposal Costs At Low-Level Weste Bur 6el Detenstlen Fac8 lies. NUREGCR4600: THE EFFECT OF INITIAL TEMPERATURE ON FLAME ACCELERATION AND DEFLAGRATION TO-DETONATION t Creek Gros e . TRANSITION PHENOMENON. l NUREGCR-4087 V26: ENVIRONMENTALLY ASSISTED CRACKNG IN UGHT-WATER REACTORS. Semennual Report. July Doosneer 1997. L- _ _ Aseseement I- NURIGCR4150 V06 R1: SCDAP/RELAP6 MOD 3.2 CODE MAN-Crea4>Arvoet Test UAL.Developmented Acessement. NUREGCR4447: RESULTS OF CRACK-ARREST TESTS ON IRRADI-EG-1566 V03: CONSOUDATED GUIDANCE ABOUT MATERIALS Creleel Nest Plus - . UCENSES. Applications For Sealed Source And Device Evalusten And NUREGCR4634: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A Regletreten. Final Report. DOWNWARD FACING CURVED SURFACE: EFFECTS OF THERMAL ! EGC  : DIGITAL IAC SYSTEMS IN NUCLEAR POWER l PLANTS. Risk-Screening Of Environmental Stressors And A Comparl-NURE . ASSESSMENT AND RECOMMENDATIONS FOR son Of Hardware Uneveliabliity With An Existing Analog System. FIBSILE MATERIAL PACKAGNG EXEMPTIONS AND GENERAL U-CENSES WITHIN 10CFR PART 71. Digined inetnamentation And Control NUREGCP-0162 V03: PROCEEDINGS OF THE TWENTY FIFTH Crustel Structure WATER REACTOR SAFETY NFORMATION MEETING.Thermel-Hy-NUREGCR4683. CRUSTAL STRUCTURE AND GROUND MOTION draulic Reneerch And Codes, Digital Instrumentation And Control, MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM ' Structural Performance NATIONAL SEISMOGRAPHIC NETWORK DATAc Damage tischenic NUREGCR4475: RE UTION OF THE DIRECT CONTAINMENT NUREGCR4646: A DAMAGE MECHANICS BASED APPROACH TO HEATING ISSUE FOR COMSUSTION ENGINEERING PLANTS & STRUCTURAL DETERIORATION AND RELIA 8luTY. BABCOCK & WILCOX PLANTS. I l i i

l. . . . . . .

56 Subject index Displacement Transducer Emergency Foodwater NUREGOR45ti2: MARBLE HILL ANNEAUNG DEMONSTRATION NUREGCR-5500 VOI: REUABluTY STUDY: AUXIUARY/ EMERGENCY EVALUATION. FEEDWATER SYSTEM. 1987-1995. Disposition Schedule Emergency Notification System NUREG0910 R03: NRC COMPREHENSIVE RECORDS DISPOSITION NUREG-1022 ROI: EVENT REPORTING GUIDELINES 10 CFR 50.72 SCHEDULE. AND 50.73. Does Aseesament Emergency Response NUREG-1549 DRFT FC: DECISION METHODS FOR DOSE ASSESS. NUREG-1634: 1997 LOST SOURCE EXERCISE.An Exercise Of Radio-MENT TO COMPLY WITH RADIOLOGICAL CRITERIA FOR UCENSE logica: Response Through Cooperation And Coordination Of . TERMINATION. Draft Report For Comment. Local, State And Federal Resources Under The National Contingercy l Plan. I Does Conversion i NUREGCR4613 V02: CODE MANUAL FOR MACCS2. Preprocessor Energy-Absort> lng Support l Codes COMIDA2, FGADCF, IDCF2. NUREGCR4559: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND Does Estimate ENERGY-ABSORBING SUPPORTS. NUREGCR4210 S01: COMPUTER CODES FOR EVALUATION OF j CONTROL ROOM HABITABluTY (HABIT V1.1). Enforcement Action NUREG-0940 V16 N2 P1: ENFORCEMENT ACTIONS: SIGNIFICANT Does ModoHng ACTIONS RESOLVED INDIVIDUAL ACTIONS.Somiannual Progress NUREGCP-0163: PROCEEDINGS OF THE WORKSHOP ON REVIEW ReportJuly-December 1997. OF DOSE MODEUNG METHODS FOR DEMONSTRATION OF COM- NUREG-0940 V16 N2 P2: ENFORCEMENT ACTIONS: SIGNIFICANT PUANCE WITH THE RADIOLOGICAL CRITERIA FOR UCENSE TER- ACTIONS RESOLVED REACTOR LICENSEES.Sem! annual Progress MINATION. ReportJuh-December 1997. NUREG-0940 V16 N2 P3: ENFORCEMENT ACTIONS: SIGNIFICANT Dos 6 metry ACTIONS RESOLVED MATERIAL UCENSEES. Semiannual Progress NUREGCR4571 V01: PROBABluSTIC ACCIDENT CONSEQUENCE ReportJuly-December 1997. UNCERTAINTY ANALYSIS. Uncertainty Assessment For Internal Do- NUREG-0940 V17 N1 P1: ENFORCEMENT ACTIONS: SIGNIFICANT simetry Maen Report. ACTIONS RESOLVED INDIVIDUAL ACTIONS. Semiannual Progress NUREGCR4600: NEU1RON EXPOSURE PARAMETERS FOR CAP- Repo 1 January 4une 1998. SULE 10.05 (N THE HEAVY-SECTION STEEL IRRADIATION PRO- NUREG-0940 V17 N1 P2: ENFORCEMENT ACTIONS: SIGNIFICANT GRAM TENTH IRRADIATION SERIES. ACTIONS RESOLVED REACTOR LICENSEES. Semiannual Progress NUREGCR4601: NEUTRON EXPOSURE PARAMETERS FOR THE Report, January 4une 1998. DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRADIA- NUREG-0940 v17 N1 P3: ENFORCEMENT ACTIONS: SIGNIFICANT TlON PROGRAM TENTH IRRADIATION SERIES. ACTIONS RESOLVED MATERIAL LICENSEES.Serniannual Progress ReportJanuary4une 1998. NUREG-1631: SOURCE DISCONNECTS RESULTING FROM RADIOG- Engineered Safety System i RAPHY DRIVE CABLE FAILURES. Final Report. NUREG/CR4418: RISK IMPORTANCE OF CONTAINMENT AND RE-( LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. NUREGCR4559: URGE-SCALE VIBRATION TESTS OF MAIN STEAM Engineering Drawing AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND NUREGCR-5502: ENGINEERING DRAWINGS FOR 10 CFR PART 71 ! ENERGY-ABSORBING SUPPORTS. PACKAGE APPROVALS. ENDF/8-Vi Environmental Assessment NUREGCR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH- NUREGCR-5549: ENVIRONMENTAL ASSESSMENT RENEWAL OF MARK. MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-GIONAL SERVICE FACluTY, WAMPUM, PENNSYLVANIA. NUREGCR-6434. ANCHOR BOLT DEHAVIOR AND STRENGTH DUR- Environmental Qualificeuon ING EARTHOUAKES. NUREGCR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAll-NUREGCR-5562: DATING AND EARTHOUAKES: REVIEW OF QUA- FICATION OF MICROPROCESSOR-BASED SAFETY-RELAT ED TERNARY GEOCHRONOLOGY AND ITS APPUCATION TO EQUIPMENT IN NUCLEAR POWER PLANTS. PALEOSEISMOLOGY, NUREGCR4544: METHODOLOGY FOR ANALYZING PRECURSORS Environmental Stressor TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE- NUREGCR4579: DIGITAL I&C SYSTEMS IN NUCLEAR POWER QUENCES. PUNTS. Risk-Screening Of Environmental Stressors And A ComparL NUREG/CR4564: ANALYSES OF SOURCE SPECTRA, ATTENUATION. son Of Hardware Unavailability With An Existing Analog System. AND SITE EFFECTS FROM CENTRAL AND EASTERN UNITED STATES EARTHOUAKES. Environmentally Assisted Cracking NUREG/GR-0017: DATING OF LIQUEFACTION IN THE NEW MADRID NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACK!NG IN SE1SMIC ZONE AND IMPUCATIONS FOR EARTHOUAKE HAZARD. UGHT-WATER REACTORS. Semiannual Report. January 4une 1997. Eastern Tennessee Event Coding NUREGCR4556: A COMPREHENSIVE STUDY OF THE EASTERN NUREGCR4268 V03: COMMON CAUSE FAILURE DATABASE AND TENNESSEE SEISMIC ZONE. ANALYSIS SYSTEM. Volume 3: Data Collection And Event Coding. Electric Dosametry Event Definition NUREGCP-0164: PROCEEDINGS OF THE WORKSHOP ON ELECTRIC NUREGOR4268 V02: COMMON CAUSE FAILURE DATABASE AND DOS! METRY. Held in Gaithersburg. Maryland On October 14-16. 1997. ANALYSIS SYSTEM. Volume 2: Event Definition And Classificat6on. Electrical Connection Event Reporting Guidelines NUREGCR4412: AGING AND LOSS OF-COOLANT ACCIDENT (LOCA) NUREG-1022 R01: EVENT REPORTING GUIDELINES 10 CFR 50.72 TESTING OF ELECTRICAL CONNECTIONS. AND 50.73. Embetttlement Exempt Distribution NUREGCR4447: RESULTS OF CRACK-ARREST TESTS ON IRRADI- NUREG-1556 V08: CONSOUDATED GUIDANCE ABOUT MATERIALS ATED A 508 CLASS 3 STEEL LICENSEES Program-Specific Guidance Exempt Distribution U-NUREGCR4537: INFLUENCE OF LONG-TERNI THERMAL AGING ON censes. Final Report. THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR PRESSURE VESSEL MATERIALS.An Atom Probe Stuoy. Externet Exposure NUREG/CR4551: IMPROVED EMBRITTLEMENT CORRELATIONS FOR NUREGCR4613 V02: CODE MANUAL FOR MACCS2. Preprocessor REACTOR PRESSURE VESSEL STEELS. Codes COMIDA2, FGRDCF, IDCF2.

I 1 Subject index 57 i FRAPCON3 NUREG/CR4521: ESTIMATING PROBABLE FLAW DISTRIBUTIONS IN ! NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE PWR STEAM GENERATORS. i CALCULATION OF STEADY-STATE. THERMAL-MECHANICAL BE- ! HAVIOR OF OXtDE FUEL RODS FOR HIGH BURNUP, Flow Menourement I NUREG/CR4534 V03: FRAPCON-3: INTEGRAL ASSESSMENT. NUREG/CR-5501: ADVANCED INSTRUMENTATION AND MAINTE-p p NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. NUREG/CR4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE- Fracture Mechen6ce l ACTOR PRESSURE VESSELS. Donalty And Distribuhon Of Flow Indi- NUREG 1426 V03: COMPILATION OF REPORTS FROM RESEARCH t cations in PVRUF. SUPPORTED BY THE ELECTRICAL. MATERIALS AND MECHANICAL ENGINEERING BRANCH, DIVISION OF ENGINEERING. NU EG. 10: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANAL- TUR E HA IC l YSIS HANDBOOK. I Fracture Toughnese

                                                                                                         ^

EGC 4 FFECTS OF LWR COOLANT ENVIRONMENTS ON ATE $08 SS 3 STEEL. j FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY l STEELS. Fragility NUREG/CR4361: SEISMIC ANALYSIS OF PlPING. Final Program Re-Feedwater Piping System E NUREG/CR4559: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND Fuel Rod ENERGY-ABSORBING SUPPORTS. NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE l CALCULATION OF STEADY-STATE. THERMAL-MECHANICAL BE. I U G/CR4603: CHARACTERIZATION OF RETARDATION MECHA-

                                                                         "^                      L c"

NURE R FRAP 3 E 5 MENT. I NISMS IN SOIL Fuel-Cooient interaction URE F A V NTAL IMPACT STATEMENT FOR AN EL N W T R' l THE CONSTRUCTION AND OPERATION OF AN INDEPENDENT l SPENT FUEL STORAGE INSTALLATION TO STORE THE THREE Gate Velve i MILE ISLAND UNIT 2 SPENT FUEL AT THE IDAHO NATIONAL ENGi- l NUREG/CR4611: RESULTS OF PRESSURE LOCKING AND THERMAL NEERING AND ENVIRONMENTAL . BINDING TESTS OF GATE VALVES. l i Financial Statement Generic Safety leeuse ! NUREG-1542 V03: ACCOUNTABILITY REPORT FISCAL YEAR 1997. NUREG-0933 S22: A PRIORITIZATION OF GENERIC SAFETY ISSUES. i Finite Element Goochronology l NUREG/CR4554: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR NUREG/CR-5562: DATING AND EARTilOUAKES: REVIEW OF OUA-l WALL INTERNATIONAL STANDARD PROBLEM. TERNARY GEOCHRONOLOGY AND ITS APPLICATION TO PALEOSEISMOLOGY. l Fim Serhr l NUREG 1552 801 DR FC: FIRE BARRIER PENETRATION SEALS IN Geographic information System NUCLEAR POWER PLANTS. Draft Report For Comment- NUREG/CR4573: "lNVESTIGATING SEISMOTECTONICS IN THE pg, m EASTERN SYSTEM., UNITED STATES USING A GEOGRAPHIC INFORMATION ! NUREG-1521 DRFT FC: TECHNICAL REVIEW OF RISK-INFORMED. PERFORMANCE-BASED METHODS FOR NUC' EAR POWER PLANT Geology FIRE PROTECTION ANALYSES. Draft Report For Comment- NUREG/CR4573: "lNVESTIGATING SEISMOTECTON!CS IN THE pg,,gg, g,,,gg EASTERN UNITED STATES USING A GEOGRAPHIC INFORMATION NUREGCR4342: ASSESSMENT AND RECOMMENDATIONS FOR SYSTEM." l FISSILE-MATERIAL PACKAGING EXEMPTONS AND GENERAL LI- Ground Motion CENSES WITHIN 10CFR PART 71. NUREGCR4593: CRUSTAL STRUCTURE AND GROUND MOTION Floelon Gas MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM , NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE NATIONAL SEISMOGRAPHIC NETWORK DATA. i ! CALCULATION OF STEADY-STATE. THERMAL-MECHANICAL BE- Ground-Water interaction NUREG/CR4377: EFFECTS ON RADIONUCLIDE CONCENTRATIONS NURE R45 0 : FRh 3 E L ES5 MENT

• BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF Flesion Product PERFORMANCE ASSESSMENT OF LOW LEVEL RADIOACTIVE NUREGCR4408: TECHNICAL ASSISTANCE IN REVIEW OF SOURCE WASTE DISPOSAL FACILITIES.

TERM-RELATED ISSUES OF ADVANCED REACTORS. NUREGCR4599: LODINE VOLATILITY AND PH CONTROL IN THE AP. Ground-Weter Modeling 600 REACTOR. NUREG/CR-5621: GROUND-WATER MODELS IN SUPPORT OF NUREG/CR 5512. Flued Gauge NUREG-1556 V04: CONSOLIDATED GUIDANCE ABOUT MATERIALS Hebenero Tool l LICENSES. Program-Specific Guidance About Fixed Gauge u- NUREG/CR-5570: APPUCATION OF THE NCSA HABANERO TOOL ! conoes. Final Report. FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS-l MENTS. Flame Accelerotion NUREG/CR4509: THE EFFECT OF INITIAL TEMPERATURE ON Habitability '! FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION NUREG/CR4210 S01: COMPUTER CODES FOR FVALUATION OF

!      TRANSITION PHENOMENON.                                             CONTROL ROOM HABITABluTY (HABIT V1,1).

NUREG/CR4524: THE EFFECT OF LATERAL VENTING ON DEFLA-I GRATION-TO DETONAT ION TRANSITION IN HYDROGEN-AIR. Health Effect STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. NUREG/CR4545 V01: PROBABILISTIC ACCIDENT CONSEOUENCE UNCERTAINTY ANALYSIS. Earty Health Effects Uncertainty Assees-Flow Distribution rnent. Main Report. NUREGCR-5505: RR-PRODIGAL A MODEL FOR ESTIMATING THE NUREGCR4545 V02: PROBABILISTIC ACCIDENT CONSEOVENCE PROBABILITIES OF DEFECTS IN REACTOR PRESSURE VESSEL UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assese-WELDS. menLAppendices. I l l l

58 Subject index NUREG/CR4555 VOI: PROBABILISTIC ACCIDENT CONSEQUENCE abiltty Analysis And Human Performance Evaluation Technscal lasues UNCERTAINTY ANALYSIS.Lete Health Effects Uncertainty Assess- Related To Rulemakings, Risk-informed, Performance-Based initia-ment. Main Report. tives. NUREGICR4555 V02: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- Hydraulic menLAppendices NUREG/CR4472: PRELIMINARY PHENOMENA IDENTIFICATION AND RANKING TABLES FOR SIMPUFIED DOILING WATER . REACTOR Hest Eschenger LOSS-OF-COOLANT ACCIDENT SCENARIOS. NUREG/CR4552: MARBLE HILL ANNEALING DEMONSTRATION EVALUATION, Hydrogen Combustion NUREGCR4524: THE EFFECT OF LATERAL VENTING ON DEFLA-Heat Trenefer GRATION-TO-DETONAT ION TRANSITION IN HYDROGEN-AIR-NUREGCP-0160: FROCEEDINGS OF THE OECD/CSNI SPECIAUST STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. MEETING ON ADVANCED INSTRUMENTATION AND MEASURE. NUREG/CR4580: PERFORMANCE TESTING OF PASSIVE l MENT TECHNIQUES. Held in Santa Barbara,CA March 17-20,1997. AUTOCATALYTIC RECOMBINERS. [ NUREGCR4472: PRELIMINARY PHENOMENA IDENTIFCATION AND RANKING TABLES FOR SIMPUFIED BOluNG WATER REACTOR Hydrogen Depletion LOSS-OF-COOUNT ACCIDENT SCENARIOS. NUREG/CR4580: PERFORMANCE TESTING OF PASSIVE AUTOCATALYTIC RECOMBINERS.

=;j t m= Steel irradiation Progro NUREG/CR-5591 V08 N1: HEAVY-SECTION STEEL IRRADIATION                 ISFSI PROGRAM.Semlannual Progress Nport For October 1996 Through             NUREG-1628 DRF FC: STAFF RESPONSES TO FREQUENTLY ASKED March 1997,                                                              QUESTIONS CONCERNING DECOMMISSIONING OF NUCLEAR POWER PLANTS. Draft Report For Comment.

=c

• Mon Steelirradiation Program NUREG/CR-5591 V04 N1: HEAVY-SECTION STEEL IRRADIATION b NU R CRITICAL HEAT FLUX (CHF) PHENOMENON ON A R M Semiannual Progress Report For October 1992 Through DOWNWARD FACING CURVED SURFACE: EFFECTS OF THERMAL INSULATION.

Heavy-Section Steel Technology Program NUREG/C 219 V N2: HEAVY-SECTION STEL TE L R G/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE NUREG/CR-4219 V14 N1 AVY-S CTION WEEL ECHNOLOGY BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. R RAM.SerNannual Progress Report For October 1996 - March individual Plant Examination Program NUREG-1560 V01 PI: INDIVIDUAL PLANT EXAMINATION PROGRAM: High Burn-Up Fuel Reneerch PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM-NUREG/CP-0162 V02: PROCEEDINGS OF THE TWENTY FIFTH ANCE. Summary _ Report. NUREG-1560 V02 P2-5: INDIVIDUAL PLANT EXAMINATION PRO-WATER REACTOR SAFETY INFORMATION MEETING. Human Reli- M RSPECTIVES ON REACTOR SAFETY AND PLANT PER-ebility Analysis And Human Performance Evaluation, Technical lasues Related To Rulemakings Esk-informed, Performanc6-Based initia- NUREG-1560 V03 P6: INDIVIDUAL PLANT EXAMINATION PROGRAM: U'** - PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM-High Burnup ANCE. Appendices. NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE Induetel Redlography CALCULATION OF STEADY-STATE, THERMAL-MECHANICAL BE- NUREG-1556 V02: CONSOLIDATED GUIDANCE ABOUT MATERIALS NURE R 5 FRA 3 AL ES MENT. EEESAogram-Spee Gh h indumrial Radiography S conses. Final Report. U 14: FE I (TY OF G FRE ENCY ACOUSTIC IM- R G/C 4613 V02: CODE MANUAL FOR MACCS2. Preprocessor Codes COMIDA2, FGRDCF, IDCF2. H6pn Temperature I"" NUREG/CR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR- N REG /C 509: THE EFFECT OF INITIAL TEMPERATURE ON IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT FLAME ACCELERATION AND DEFLAGRATION-TO-DETONAllON HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH TRANSITION PHENOMENON. MODEL PREDICTIONS. Inittellem G1 : D RD REVIEW PLAN FOR PHYSICAL PROTEC-NUREMW M MC MMM & ABMWN TION PLANS FOR THE INDEPENDENT STORAGE OF SPENT FUEL Inservice inspection AND HIG+ LEVEL RADIOACTIVE WASTE. NUREGOR4511 V03: STEAM GENERATOR TUBE INTEGRITY PRO-gg GRAM. Semiannual Report October 1996 - March 1997. NUREG/CR4524: THE EFFECT OF LATERAL VENTING ON DEFLA- Innervice Testing GRATION-TCFDETONAT ION TRANSITION IN HYDROGEN-AIR- NUREG/CP-0152 V02: PROCEEDINGS OF THE FIFTH NRC/ASME STEAM MlXTURES AT VARIOUS INITIAL TEMPERATURES. SYMPOSIUM ON VALVE AND PUMP TESTING. Human Event Analyale instrumentation NUREG-1624 DRFT FC: TECHNICAL BASIS AND IMPLEMENTATION NUREGCR-5501: ADVANCED INSTRUMENTATION AND MAINTE-GUIDEUNES FOR A TECHNIQUE FOR HUMAN EVENT ANALYSIS NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. (ATHEANA). Draft Report For Comment instrumentation And Control Human Factor NUREG/CR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUALl-NUREGCR4605: AN EVALUATION OF HUMAN FACTORS RESEARCH FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED FOR ULTRASONIC INSERVICE INSPECTION. EQUIPMENT IN NUCLEAR POWER PLANTS. Human Performance Interface Theory I NUREGCR4364: HUMAN PERFORMANCE IN RADIOLOGICAL SUR- NUREGCR4150 V01 R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN- l VEY SCANNING. VAL. interface Theory. Human Rollability Analyene intertecial Sheer NUREGCP-0162 V02: PROCEEDINGS OF THE TWENWFIFTH NUREG/lA-0146: IMPLEMENTATION AND ASSESSMENT OF IM-WATER REACTOR SAFETY INFORMATION MEETING. Human Reli- PROVED MODELS AND OPTIONS IN TRAC-BF1.

1 Subject Index 59 i I intoryonular Crack Look Rate NUREGOR-0016: THE ROLE OF TIME-DEPENDENT DEFORMATION NUREG/CR4540 STATE OF-THE-ART REPORT ON PlPING FRAC-IN INTERGRANULAR CRACK INITIATION OF ALLCY 600 STEAM TURE MECHANICS. GENERATOR TUBING MATERIAL Legal leeuences intemel Doelmetry NUREG-0750 C104: INDEXES TO NUCLEAR REGULATORY COMMIS-NUREG/CR4571 V02: PROBABILISTIC ACCIDENT CONSEQUENCE SiON ISSUANCES.Jama UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do- NUREG-0750 V45: 1,1991 R N CLEAR thrNh December ULATORY 31 1995. CIOMMISSION anmetry.Appndices. ISSUANCES. Opinions And Decissorm Of The Nuclear Regulatory Corn-rnession With Selected Orders. January-June 1997. Intemel Fire NUREG-0750 V46101: INDEXES TO NUCLEAR REGULATORY COM-NUREG/CR4544: METHODOLOGY FOR ANALYZING PRECURSORS MISSION ISSUANCES. July-September 1997. TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE- NUREG0750 V46102: INDEXES TO NUCLEAR REGULATORY COM-QUENCES. MISSION ISSUANCES.Julv-December 1997. Intemel Pressurization NUREG-0750 V46 NO3: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR SEPTEMBER 1997. Pages 49-193. NUREG/CR-6517: ROUND ROBIN PRETEST ANALYSES OF A STEEL NUREG-0750 V46 N04: NUCLEAR REGULATORY COMMISSION CONTAINMENT VESSEL MODEL AND CONTACT STRUCTURE AS- ISSUANCES FOR OCTOUER 1997. Pages 195-256. SEMBLY SUBJECT TO STATIC INTERNAL PRESSURIZATION. NUREG-030 V46 N05. NUCLEAR HEGULATORY COMMISSION lodine ISSUANCES FOR NOVEMBER 1997. P 257-285. . NUREG-0750 V46 N06: NUCLEAR R GULATORY COMMISSION i NUREGCR4408: TECHNICAL ASSISTANCE IN REVIEW OF SOURCE ISSUANCES FOR DECEMBER 1997. Pages 287-319. TERM-RELATED ISSUES OF ADVANCED REACTORS. NUREG-0750 V47101: INDEXES TO NUCLEAR REGULATORY COM-NUREG/CR-6599: LODINE VOLATluTY AND PH CONTROL IN THE AP- MISSION ISSUANCES. January-March 1998. i 600 REACTOR. i NUREG4750 V47102: INDEXES TO NUCLEAR REGULATORY COM- l Irradleted Reactor Fuel MISSION ISSUANCES.Jan 1998. NUREG-0750 V47 N01: NU LEAR REGULATORY COMMISSION NUREG-0725 R13: PUBLIC INFORMATION CIRCULAR FOR SHIP- ISSUANCES FOR JANUARY 1998. Pages 1 12. MENTS OF IRRADIATED REACTOR FUEL NUREG-0750 V47 NO2: NUCLEAR REGULATORY COMMISSION Irradiation ISSUANCES FOR FEBRUARY 1998. P 1M6. NUREG-0750 V47 NO3: NUCLEAR GULATORY COMMISSION NUREGCR4600: NEUTRON EXPOSURE PARAMETERS FOR CAP. ISSUANCES FOR MARCH 1998.Pages $7 75. SULE 10.05 IN THE HEAVY-SECTION STEEL 1RRADIATION PRO- NUREG 0750 V47 N04: NUCLEAR REGULATORY COMMISSION GRAM TENTH IRRADIATION SERIES. ISSUANCES FOR APRIL 1998.Pages 77-260. NUREG/CR4601: NEUTRON EXPOSURE PARAMETERS FOR THE NUREG-0750 V47 N05: NUCLEAR REGULATORY COMMISSION DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRADIA- ISSUANCES FOR MAY 1998.Paoes 261-306. TION PROGRAM TENTH IRRADIATION SERIES. NUREG-0750 V47 N06: NUCLEAR REGULATORY COMMISSION Jave ication ISSUANCES FOR JUNE 1998 P s 307-408. NUREG 07bo V48101: INDEXES NUCLEAR REGULATORY CCR NU G/CR-5570: APPLICATION OF THE NCSA HABANERO TOOL MISSION ISSUANCES. July -Septernber 1998. FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS- NUREG-0750 V48 N01: NUCLEAR REGULATORY COMMISSION MENTS. ISSUANCES FOR JULY 1998.Pages 1-38. LOCA NUREG-0750 V48 NO3: NUCLt.AR REGULATORY COMMISSION ISSUANCES FOR SEPTEMBER 1998. Pages 119-182. NUREG/lA-0139: ASSESSMENT OF RELAPS/ MOD 3.2 USING LOFT NUREG-0750 V48 N04: NUCLEAR REGULATORY COMMISSION 4 j N 1 E L T SSESSMENT OF RELAP5/ MOD 3.1 WITH SEPARATE-EFFECT AND INTEGRAL TEST EXPERIMENTS: License Termination MODEL CHANGES AND OPTIONS. NUREG1549 DRFT FC: DECISION METHODS FOR DOSE ASSESS-NUREG/LA 0146: IMPLEMENTATION AND ASSESSMENT OF IM- MENT TO COMPLY WITH RADIOLOGICAL CRITERIA FOR UCENSE PROVED MODELS AND OPTIONS IN TRAC-BF1. TERMINATION.Drafi R NUREG/LA-0148: ASSESSMENT OF RELAP5fMOD3.1 USING LSTF For Comment. NUREG-1628 DRF FC: S FF RESPONSES TO FREQUENTLY ASKED TEN-PERCENT MAIN STEAM-LINE-BREAK TEST RUN SB-SL 01. QUESTIONS CONCERNING DECOMMISSIC41NG OF NUCLEAR LOR POWER PLANTS Draft For Comment. NUREG/LA-0130: ASSESSMENT OF RELAP5/ MOD 32 USING LOFT NUREG/CP-0163: PROCEE INGS OF THE WORKSHOP ON REVIEW LARGE BREAK LOCA TEST.LP-024. OF DOSE MODELING METHODS FOR DEMONSTRATION OF COM-PLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TER-Lory y g L3.g MINATION. NUc!EG/lA-0024: APPUCATION OF RELAP5/ MOD 3.1 CODE TO THE LOFT TEST LM. Licensed Fuel Facillip Status Report NUREG-0430 V16: UCENSED FUEL FACILITY STATUS RE-LWR PORT. inventory Difference Data. July 1,1995 - June 30,1996.(Gray NUREG/CR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN Book 11) June 1997. NUREG-0430 V17: LICENSED FUEL FACluTY STATUS RE-UGHT. WATERV04 NUREG/CR4150 REACTORS. Semiannual R1: SCDAP/RELAP5/ MODReport,JanuaryCODE 32 MAN- PORT inventory Difference Data. July 1,1996 - June 30,1997.(Grey , UALMATPRO A Ubrary Of Matenals Properties For Ught-Water Re- Book 11) actor Accident Ana is. NUREG/CR4536: V RIFICATION OF THE LWRARC CODE FQR Licensee Event Report UGHT-WATER-REACTOR AFTERHEAT RATE CALCULATIONS. NUREG-1022 RO1: EVENT REPORTING GUIDEUNES to CFR 50.72 NUREG/CR4583: EFFECTS OF LWR COOLANT ENVIRONMENTS ON AND $0.73. FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY NUREG-1187 V01: PERFORMANCE INDICATORS FOR OPERATING STEELS. COMMERCIAL NUCLEAR POWER REACTORS. Data Through Sep-tomber 1997. LWRARC Codo NUREG/CR-4674 V25: PRECURSORS TO POTENTIAL SEVERE CORE NUREGCR4536: VERIFICATION OF THE LWAARC CODE FOR DAMAGE ACCIDENTS: 1996. A Status Report. UGHT-WATER-REACTOR AFTERHEAT RATE CALCULATIONS, NUREG/CR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1997. A Status Report. NUREG/CR4559: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM Light Water Reactor AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND NUREG/CR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN ENERGY ABSORBING SUPPORTS. June 1997. UGHT-WATERV04 NUREG/CR4150 REACTORS. Semiannual R1: SCDAP!RELAPS/ Repod.JanuaryCODE MOD 3.2 MAN-Lateral Venting UAL.MATPRO - A Ubrary Of Materials Properties For Ught-Water Re-NUREG/CR4524: THE EFFECT OF LATERAL VENTING ON DEFLA- actor Accident AnaYsis. GRATION-TO DETONAT ION TRANSITION IN HYDROGEN-AIR- NUREG/CR4536: VERIFICATION OF THE LWRARC CODE FOR , STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. UGHT-WATER-REACTOR AFTERHEAT RATE CALCULATIONS. l l

60 Subject index NUREG/CR4583: EFFECTS OF LWR COOLANT ENVIRONMENTS ON Main Steam Line FATIGUE DESIGN CURVES OF CARBON AND LOW. ALLOY NUREG/lA-0148: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF STEELS. TEN-PERCENT MAIN STEAM-LINE-BREAK TEST RUN SB-SL-01. Llquelection Maintenance i NUREG/CR4274: PALEOSEISMIC STUDIES IN THE SOUTHEASTERN NUREG/CR-5501: ADVANCED INSTRUMENTATION AND MAINTE-UNITED STATES AND NEW ENGLAND. NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. NUREG/GR4017: DATING OF LIQUEFACTION IN THE NEW MADRID SEISMIC ZONE AND IMPUCATIONS FOR EARTHOUAKE HAZARD. Maintenance Scheduling NUREG/CR4616: RISK COMPARISON OF SCHEDULING PREVENTIVE Loos Of Offene Power MAINTENANCE DURING SHUTDOWN VS. DURING POWER OPER-NUREGlCR4496: EVALUATION OF LOSS OF OFFSITE POWER ATION FOR PWRS. EVENTS AT NUCLEAR POWER PLANTS: 1980 -1996. Lose-Of-Coolant Accident NUREG/CR4552: MARBLE HILL ANNEAUNG DEMONSTRATION NUREG/CR4412: AGING AND LOSSCF-COOLANT ACCIDENT (LOCA) EVALUATION. TESTING OF ELECTRICAL CONNECTIONS. NUREG/CR4472: PRELIMINARY PHENOMENA IDENTIFICATION AND Material Licensee RANKING TABLES FOR SIMPLIFIED BOluNG WATER REACTOR NUREG/CR-5549: ENVIRONMENTAL ASSESSMENT RENEWAL OF LOSS-OF-COOLANT ACCIDENT SCENARIOS. MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE-GIONAL SERVICE FACluTY, WAMPUM, PENNSYLVANIA. Lost Source Materiale Event NUREG-1834: 1997 LOST SOURCE EXERCISE.An Exercise Of Radio-NUREG-1634: 1997 LOST SOURCE F.XERCISE.An Exercise Of Radio-logical Response Through Cooperation And Coordination Of logcal Response Through Cooperation And Coordination Of Local, State And Federal Resources Under The National Contingency Local. State And Federal Resources Under The National Contingency Plan. Plan. Low Level Radloactive Weste NUREG/CR4377: EFFECTS ON RADIONUCLIDE CONCENTRATIONS U 1 6 02: CONSOUDATED GUIDANCE ABOUT MATERIALS BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF UCENSES.Prograrn-Specife Guidance About industrial Radiography Lk PERFORMANCE ASSESSMENT OF LOW-LEVEL RADIOACTIVE conses. Final Report. WASTE DISPOSAL FACIUTIES. NUREG-1556 V03: CONSOUDATED GUIDANCE ABOUT MATERIALS Low Specific Activity Matertal "*^" '" ra NUREG-1608: CATEGORIZING AND TRANSPORTING LOW SPECIFIC NURIG-1556 V04: SOUDATED GUIDANCE ABOUT MATERIALS ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS- UCENSES. Program-Specife Guidance About Fized Gauge Lk Low Alloy Steel N$E 556 0 CONSOUDATED GUIDANCE ABOUT MATERIALS NUREG/CR4583: EFFECTS OF LWR COOLANT ENVIRONMENTS ON LICENSES. Program-Specife Guidance About Self-Chielded irradiator FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY Licenses. Final Report. STEELS. NUREG 1556 V07 DR FC: CONSOUDATED GUIDANCE ABOUT MATE. RIALS LICENSES. PROGRAM-SPECIFIC GUIDANCE ABOUT ACA-E " " " N E R4 : W-L VEL WASTE DATA BASE DEVELOPMENT h,,R E D an PFIOGRAM- NUREG-1556 V08: CONSOLIDWED GUIDANCE ABOUT MATERIALS g ,,y g,p,,, UCENSEES. Program-Specife Guidance Exempt Distribution Lk NUREG/CR4608:

SUMMARY

AND EVALUATION OF LOW-VELOCITY NUYE$1 V09 FC: CONSOUDATED GUIDANCE ABOUT MATF-IMPACT TEST OF SOLID STEEL BILLET ONTO CONCRETE PADS. RIALS UCENSES. Program-Specific Guidance About Medical Use Lk LY'I** NU E 5 V FC: N UDATED GUIDANCE ABOUT MATE-NUREG/CR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT RIALS UCENSES. Program Specife Guidance About Master Materials PROGRAM. Licenses. Draft Report For Comment. MACCS NUREG-1556 V11 DR FC: CONSOUDATED GUIDANCE ABOUT MATE-RIALS LICENSES Program-Specife Guidance About Specife Ucenses NUREG/CR4555 V01: Ph0BABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Late Health Eficcts Uncertainty Assess- Of &oadscope. Dran Report For Comment. ment. Main Report. Measurement Planning NUREG/CR4555 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG-1575: MULTI-AGENCY RADIATION SURVEY AND SITE INVES-UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess ~ TIGATION MANUAL (MARSSIM). Final Report. ment.Appendees. Measuring instrument MACCS2 NUREG/CP-0160: PROCEEDINGS OF THE OECD/CSNI SPECIAUST NUREG/CR4613 V01: CODE MANUAL FOR MACCS2. User's Guide. MEETING ON ADVANCED INSTRUMENTATION AND MEASURE-NUREG/CR4613 V02: CODE MANUAL FOR MACCS2. Preprocessor MENT TECHNIQUES. Held in Santa Barbara,CA, March 17-20,1997. Codes COMIDA2, FORDCF, IDCF2. ( Metal Retardation NU 1507: MINIMUM DETECTABLE CONCENTRATIONS WITH N SM N TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON-TAMINANTS AND FIELD CONDITIONS. Microhardness NUREG-1575: MULTkAGENCY RADIATION SURVEY AND SITE INVES- NUREG-1629: THE CHARACTERIZATION OF VICKER'S MICROHARD-TIGATION MANUAL (MARSSIM). Final Report. NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-MELCOR Computer Code NUREG/CR4119 V01 R1: MELCOR COMPUTER CODE MAIO- Microprocessor ALS. Primer And Users' Guides. Version 1.8.4 July 1997. NUREG/CR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAll-NUREGlCR4119 V02 R1: MELCOR COMPUTER CODE MANU- FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED ALS. Reference Manuals, version 1.8.4, July 1997, EOulPMENT IN NUCLEAR POWER PLANTS. Mein Pump Motor-Operated Valve NUREG/lA 0149: ASSESSMENT OF RELAP5/ MOD 3.2-NPA3.4 AGAINST NUREG/CP-0152 V02: PROCEEDINGS OF THE FIFTH NRC/ASME A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR SYMPOSIUM ON VALVE AND PUMP TESTING. TRIP. NATURAL CIRCULATION AND THE START OF A MAIN PUMP NUREG/CR4611: RESULTS OF PRESSURE LOCKING AND THERMAL IN THE VANDELLOS 11 NUCLEAR POWER PLANT. BINDING TESTR OF GATE VALVES.

Subject Index 61 NRC Enforcement Guldence Nuclear Criticality NUREG 1600 R01. GENERAL STATEMENT OF POUCY AND PROCE- NUREG/CR4410: NUCLEAR FUEL CYCLE FACluTY ACCIDENT ANAL. DURE FOR NRC ENFORCEMENT ACTIONS. Enforcement Pclicy. YSIS HANDBOOK. NRC Enforcement Polley Nuclear Fuel NUREG-1600 RO1: GENERAL STATEMENT OF POUCY AND PROCE- NUREG/CR4410: NUCLEAR FUEL CYCLE FACluTY ACCIDENT ANAL-DURE FOR NRC ENFORCEMENT ACTIONS Enforcement Policy. YSIS HANDBOOK. NUREG-1622: NRC ENFORCEMENT POUCY REVIEW. July 1995 - July 1997. Nuclear Liability insurance NUREG/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE 0 52 V : PROCEEDINGS OF THE FIFTH NRC/ASME

   ' SYMPOSIUM ON VALVE AND PUMP TESTING.                            Hugleer Power Plant NUREG/CR-5496: EVALUATION OF LOSS OF OFFSITE POWER NUREG-Series PubliceHon                                                 EVENTS AT NUCLEAR POWER PLANTS: 1980 - 1996.

NUREG 0334 V23 N01: ABSTRACTS FOR PUBUCATIONS IN THE NUREG/CR4501: ADVANCED INSTRUMENTATION AND MAINTE-NUREG-SERIES. Semiannual Compilation for January-June 1998. NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS.

           .                                                          NUREG/CR4577: U.S. NUCLEAR POWER PLANT OPERATING COST R G/CR45h3       CR.      STRUCTURE AND GROUND MOTION MODELS IN THE EASTERN AND CENTRAL UNITED STATES FROM            Nuclear Reactor NATIONAL SEISMOGRAPHIC NETWORK DATA.                             NUREG-1635 V01: REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM.A Fle-U EG/b          : SINGLE-PHASE AND TWO-PHASE NATURAL CIR-                                        ulaW hsWon.

CULATION TESTS IN THE PUMA FACluTY. Nuclear Reactor Safety NUREG/lA 0149: ASSESSMENT OF RELAP5/ MOD 3 2-NPA3.4 AGAINST NUREG/CR4372: EXPERIMENTS ON INTERACTIONS BETWEEN ZlR-A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR CONIUM-CONTAINING MELT AND WATER. TRIP NATURAL CIRCULATION AND THE START CF A MAIN PUMP NUREG/CR4119 V01 R1: MELCOR COMPUTER CODE MANU-IN THE VANDELLOS 11 NUCLEAR POWER PLANT. ALS. Primer And Users' Guides, Version 1.8.4. Jut 1997. NUREG/CR4119 V02 R1: MELCOR COMPuiER CODE MANU-UE R 5534 C AL HEAT FLUX (CHF) PHENOMENON ON A DOWNWARD FACING CURVED SURFACE: EFFECTS OF THERMAL Nuclear Regulatory Leglotation INSULATION. NUREG-0980 V01 N04: NUCLEAR REGULATORY LEGISLATION.104th Neutron Dosimetry NU 80 V02 N04: NUCLEAR REGULATORY LEGISLATION.104th NUREG/CR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH- Congress. MARK. Nuclear Safety Research Neutron Exposure NUREG/CP-0165: TRANSACTIONS OF THE TWENTY-SIXTH WATER NUREG/CR4600: NEUTRON EXPOSURE PARAMETERS FOR CAP. REACTOR SAFETY INFORMATION MEETING. SULE 10.05 IN THE HEAVY <SECTION STEEL IRRADIATION PRO-GRAM TENTH IRRADIATION SERIES. Nuclear Term NUREG/CR4601: NEUTRON EXPOSURE PARAMETERS FOR THE NUREG-0544 R04: NRC COLLECTION OF ABBREVIATIONS. DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRADIA-YlON PROGRAM TENTH IRRADIATION SERIES. Occupationel Redlauon Exposure NUREG 0713 V18: OCCUPATIONAL RADIATION EXPOSURE AT COM-Neutron Kinetic MERCIAL NUCLEAR POWER REACTORS AND OTHER FAClu-NUREG/CR4359 VO1: RAMONA-4B: A COMPUTER CODE WITH TIES 1996.Twent THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR NUREd-0713 V19:y-Ninth Annual Report. OCCUPATIONAL RADIATION EXPO SYSTEM TRANSIENTS.Models And Currelations. MERCIAL NUCLEAR POWER REACTORS AND OTHER FACIUTIES NUREG/CR4359 V02: RAMONA4B: A COMPUTER CODE WITH 1997,Thart eth Annual Report. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR SYSTEM TRANSIENTS. User's Manual Office of The impector Gemrel NUREG-1415 V10 NO2: OFFICE OF THE INSPECTOR GEN-New Madrid Seism 6c Zone ERAL. Semiannual Report To Congress,0ctober 1,1997 - March 31, i NUREG/GR-0017: DATING OF LIQUEFACTION IN THE NEW MADRID 1998. SEISMIC ZONE AND IMPUCATIONS FOR EARTHOUAKE HAZARD. Operating Coat Nondestructive Evolustion NUREG/CR4577: U.S. NUCLEAR POWER PLANT OPERATING COST NUREG/CR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO- AND EXPERIENCE SUMMARIES. GRAM. Annual Report.Avaust 1995 - Sat 4 ember 1996. NUREG/CR4511 V03: STEAM GENERATOR TUBE INTEGRITY PRO- Operating Expertence GRAM. Semiannual Report, October 1996 - March 1997. NUREG/CR4577: U.S. NUCLEAR POWER PLANT OPERATING COST AND EXPERIENCE SUMMARIES. f NUREG/CR4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE- Operational Event ACTOR PRESSURE VESSELS. Density And Distribution Of Flaw Indi- NUREG/CR-4674 V25: PRECURSORS TO POTENTIAL SEVERE CORE cetions in PVRUF. DAMAGE ACCIDENTS: 1996. A Status Report. NUREG/CR4614: FEASIBluTY OF HIGH FREQUENCY ACOUSTIC IM- NUREG/CR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE AGING FOR INSPECTION OF CONTAINMENTS. DAMAGE ACCIDENTS:1997. A Status Report. Nondestructive Testing Operator Licensing NUREG/CR4589: THE EFFECTS OF SURFACE CONDITION ON AN NUREG-1122 R02: KNOWLEDGE AND ABILITIES CATALOG FOR NU-ULTRASONIC INSPECTION: ENGINEERING STUDIES USING VAU- CLEAR POWER PLANT OPERATORS.Pressurtred Water Reactors. DATED COMPUTER MODEL NUREG-1123 R02: KNOWLEDGE AND ABILITIES CATALOG FOR NU-NUREG/CR4594: EVALUATION OF ULTRASONIC INSPECTION TECH- CLEAR POWER PLANT OPERATORS.Boihng Water Reactors. NIQUES FOR COARSE GRAINED MATERIALS. Optical System Nuclear Accident Analyste NUREG/CP-0160 PROCEEDINGS OF THE OECD/CSNI SPECIAUST NUREG/CR4545 V01: PROBABluSTIC ACCIDENT CONSEQUENCE MEETING ON ADVANCED INSTRUMENTATION AND MEASURE-UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assess- MENT TECHNIQUES. Held in Santa Barbara,CA March 17-20,1997. ment. Main Report. NUREG/CR4545 V02: PROBABluSTIC ACCIDENT CONSEQUENCE Overview UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assesa- NUREG/OR4268 V01: COMMON CAUSE FAILURE DATABASE AND ment. Appendices ANAlii:IS SYSTEM. Volume 1: Overview. l f f , l )

62 Subject index PACTEL Egertmental Date Petit 6one For Rulemaking NUREG4A-0146: RELAP6 ASSESSMENT AGAINST PACTEL EXPERI. NUREG-0936 V16 NO2: NRC REGULATORY AGENDA 4emiennual Re-MENTAL DATA (REVISION 1). gful 1 7 p PCCS portJenuary-June 1996. NUREG4A4147: ASSESSMENT OF RELAP5/ MOD 3.2 FOR STEAM CONDENSATION EXPERIMENTS IN THE ,RESENCE OF ~_- - NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. NUREGCR-5669: SINGLE AND CROSS-HOLE PNEUMATIC TESTS IN UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-PUMA Feelalty SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA-NUREGCR-6498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- BluTY,CONNECTIVITY AND SCALE. CULATON TESTS IN THE PUMA FACluTY. Physical Protection Plan PWR NUREG-1619: STANDARD REVIEW PLAN FOR PHYSICAL PROTEC-NUREG-1122 R02: KNOWLEDGE AND A81UTIES CATALOG FOR NU- TION PLANS FOR THE INDEPENDENT STORAGE OF SPENT FUEL CLEAR POWER PLANT OPERATORS. Pressurized Weier Reactors. AND HOH LEVEL RADCACTIVE WASTE. NUREG/lA-0024: APPUCATION OF RELAPbMOD3.1 CODE TO THE LOFT TEST L34. Piping NUREGCR-5361: SEISMIC ANALYSIS OF PIPING. Final Program Re-6602: ENGINEERING DRAWINGS FOR 10 CFR PART 71 NUYE'GCR4540: STATE OF-THE-ART REPORT ON PIPING FRAC-PACKAGE APPROVALS. TURE MECHANICS. Pecheging Plant Performance NUREG-0383 V01 R21: DIRECTORY OF CERTIFICATES OF COMPU- NUREG-1660 V01 P1: INDIVIDUAL PLANT EXAMINATION PROGRAM. ANCE FOR RADOACTIVE MATERIALS PACKAGES. Report Of NRC- PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM-

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NU R2'1: DIRECTORY OF CERTIFICATES OF COMPU- NU II  : INDIVIDUAL PLANT EXAMINATION PRO-ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certificales Of GRAM: PERSPECTIVES ON REACTOR SAFETY AND PLANT PER-NU 'V03 R18: DIRECTORY OF CERTIFICATES OF COMPL1- NLR G- 03 P6: INDIVOUAL PLANT EXAMINATION PROGRAM: ANCE FOR RADCACTIVE MATERIALS PACKAGES. Report Of NRC- PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM- , Approved Quality Assurance Programs For Redonctive Meterials Peck- ANCE."R 1 . j Pneumatic Test WW NUREG/GR 0017: DATING OF UQUEFACTION IN THE NEW MADRID NUREGCH-5559: SINGLE AND CROSS-HOLE PNEUMATIC TESTS IN UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-SElSMIC ZONE AND IMPUCATIONS FOR EARTHOUAKE HAZARD. SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA-BILITY,CONNECTIVITY AND SCALE. ' NUREGCR4T62: DATING AND EARTHOUAKES: REVIEW OF QUA- Post-Tensioning W TERNARY GEOCHRONOLOGY AND ITS APPUCATION TO NUREGCR4691L AN INVESTIGATION OF TENDON SHEATHING N pal OSEISMIC STUDIES IN THE SOUTHEASTERN UNITED STATES AND NEW ENGLAND. Potasolum lodide Parameter Betimshon NUREG-1633 DAFT FC: ASSESSMENT OF THE USE OF POTASSIUM NUREGCR-6497: COMMON CAUSE FAILURE PARAMETER EST1- IODIDE (Kl) AS A PUBUC PROTECTIVE ACTION DURING SEVERE REACTOR ACCIDENTS. Draft Report For Comment. MATIONS. Proottee And Procedure Digest % u. _ NUREG-0386 DOS: UNITED STATES NUCLEAR REGULATORY COM-NUREGCR46ef PERFORMANCE TESTING OF PASSIVE AUTOCATALYTIC RECOMBINERS' MISSION STAFF PRACTICE AND PROCEDURE Dh GEST. Commission. Appeal Board And ucono6ng Bcard DecisionsJuly Penelve Catalync FHeer 1972 - June 1997. NUREGCR-0600: PERFORMANCE . TESTING OF PASSIVE AUTOCATALYTIC RECOMBINERS. GC ALEOSEISMIC STUDIES IN THE SOUTHEASTERN Penetreden Seel UNITED STATES AND NEW ENGLAND. NUREG-1662 S01 DR FC: FIRE BARRIER PENETRATION SEALS IN NUCLEAR POWER PLANTS. Draft Report For Comment. P ) Performenee Aeoseement POWER PLANT CONTAINMENT METALLIC PRESSURE BOUND-NUREGCR4377: EFFECTS ON RADIONUCUDE CONCENTRATONS ARIES. BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF ERF ES ENT OF LOW-LEVEL RADIOACTIVE G 46 1: RESULTS OF PRESSURE LOCKING AND THERMAL BINDING TESTS OF GATE VALVES. $ Performones indiostor j NUREG-1187 V01: PERFORMANCE INDICATORS FOR OPERATING Pressure Vessel COMMERCIAL NUCLEAR POWEH REACTORS.Dete Through Sep. NUREGCR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH-tomber 1997. NU E R4537: INFLUENCE OF LONG-TERM THERMAL AGING ON Performance Measure THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR NUREG-1542 V03: ACCOUNTABluTY REPORT FISCAL YEAR 1997. PRESSURE VESSEL MATERIALS.An Atom ProbeStudy. l Performones Plan Pressure Vessel Roemerch NUREG-1827 V01: PERFORMANCE PLAN FY 1999 NUREGCP-0162 V01: PROCEEDINGS OF THE TWENTY-FIFTH WATER REACTOR SAFETY INFORMATION MEETING. Plenary See-Permanently Defueled eions, Pressure Vessel Research.BWR Strainer Blocks 0e And Other NUREG 1825 DRF FC: PROPOSED STANDARD TECHNICAL SPECl- Generic Safely lesues.Erwironmentepy Assisted Degradation Of LWR. FICATIONS FOR PERMANENTLY DEFUELED WESTINGHOUSE PLANTS. Draft Report For Comment Presourtaed Wotor Reactor NUREG-1122 R02: KNOWLEDGE AND ABluTIES CATALOG FOR NU-Personal Computer CLEAR POWER PLANT OPERATORS. Pressurized Water Reactors. NUREG4A 0142: INSTALLATION OF RELAP5 MOD 3.2 ON 80486 AND NUREG/lA-0024: APPUCATON OF RELAPSMOD3.1 CODE TO THE PENTIUM BASED PERSONAL COMPUTERS. LOFT TEST L&4.

y Subject index 63 Prostreseed Concrete RADTRAD NUREG/CR-5671: PRETEST PREDICTION ANALYSIS AND POSTTEST NUREG/CR4604: RADTRAD: A SIMPLIFIED MODEL FOR RADIO-CORRELATION OF THE SIZEWELL-B 1:10 SCALE PRESTRESSED NUCUDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION. CONCRETE CONTAINMENT MODEL TEST. RAMONA-48 Protest Prediction NUREG/CR4359 V01: RAMONA-4B: A COMPUTER CODE WITH NUREG/CR4671: PRETEST PREDICTION ANALYSIS AND POSTTEST THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR CORRELATION OF THE SIZEWELL-B 1:10 SCALE PPESTRESSED SYSTEM TRANSIENTS Models And Correlations. CONCRETE CONTAINMENT MODEL TEST. NUREG/CR4359 V02: RAMONA-48: A COMPUTER CODE WITH Provenew Mainknence THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR NUREG/CR4616: RISK COMPARISON OF SCHEDULING PREVENTIVE SYSTEM TRANSIENTS. User's Manual MAINTENANCE DURING SHUTDOWN VS. DURING POWER OPER- RELAPS ATION FOR PWRS. NUREG/CR4150 V01 R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN-ProbabiHatic Accident UALinierface Thearv. NUREGCR4555 VO1: PR IUSTIC ACCIDENT CONSEQUENCE NUREG/CR 6150 V02 RI: SCDAP/RELAP5/ MOD 32 CODE MAN-UNCERTAINTY ANALYSIS.Lete Health Effects Urcertainty Assess-NURE 5 03  : AP L'APS/ MOD 3.2 CODE MAN-NU N45bO2; PROBABILISTIC ACCIDENT CONSEQUENCE UALUser's Guide And Irput Manual. UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assus- NUREGCR-6150 V04 R1: SCDAP/RELAP5/ MOD 32 CODE MAN-UALMATPRO - A Ubrary Of Materials Proportes For Ught-Water Re-NURE' 4571 V01: PROBABluSTIC ACCIDENT CONSEQUENCE actor Amident Analysis. UNCERTAINTY ANALYSIS. Uncertainty Assessment For intomal Do, NUREG/CR4150 V05 R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN-simetry. Main Report. UALDevelopmentai Assessment. NUREG/CR4571 V02: PROBABILISTIC ACCIDENT CONSEQUENCE UN NTY NALYSIS. Uncedainty Assessment For Intemal Do- M NURE  : RELAPS/ MOD 3 SUBCOOLED BOlWNG MODEL AS-SESSMENT. ProbabiNatic Risk Analyele NUREG/lA C141: RESULT OF BETHSY TEST 9.1.B USING RELAP5/ NUREG-1521 DRFT FC. TECHNICAL REVIEW OF RISK-INFORMED, MOD 3. PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT NUREGAA 0024: APPUCATION OF RELAPWOD3.1 CODE TO THE FIRE PROTECTION ANALYSES. Draft Report For Comment. FT TEST EVELOPMENT ASSESSMENT OF RELAP5/ MOD 3.1 Probabillotic Risk Assosoment WITH SEPARATE-EFFECT AND INTEGRAL TEST EXPERIMENTS: NUREG-1624 DRFT FC: TECHNICAL BASIS AND IMPLEMENTATION MODEL CHANGES AND OPTIONS. GUIDEUNES FOR A TECHNIQUE FOR HUMAN EVENT ANALYSIS NUREG/lA-0148: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF (ATHEANA). Draft Report For Comment. TEN-PERCENT MAIN STEAM-UNE-BREAK TEST RUN SB-SL-01. NUREG/CR-5485: GUIDEUNES ON MODEUNG COMMON-CAUSE NUREG/tA-0139: ASSESSMENT OF RELAP5/ MOD 3.2 USING LOFT FAILURES IN PROBABluSTIC Al*K ASSESSMENT. LARGE BREAK LOCA TEST,LP-024. i NUREG/CR4119 Vol R1: MELdOR COMPUTER CODE MANU- NUREG/lA4142: INSTALLATION OF RELAP5/ MOD 3.2 ON 80486 AND I 1997. PENTIUM BASED PERSONAL COMPUTERS. ALS. Primer AndV02 NUREG/CR4119 Users' R1:Guides IJELwRVeAion 1.8 4 July _ER COMPu1 CODE MANU- NUREGAA-0143: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE LSTF ALS. Reference ManusJa,Versson 1.8.4, July 1997. EXPERIMENT SlMULATING A LOSS OF RESIDUAL HEAT REMOVAL _ - . . - . EVENT DURING MID-LOOP OPERATION. NUR N3Ei31556 V05: CONSOLIDATED GUIDANCE ABOUT MATERIALS UCENSES. Program-Specife Guidance About Sell-Shielded irradiator $EM ALE ATLR RCULA ON EX ER ENT S-N . NUREG/lA-0145: RELAP5 ASSESSMENT AGAINST PACTEL EXPERI-N N5 07 C: CONSOUDATED GUIDANCE ABOUT MATE-NUREGAA 01 ASS E OF RELAP5/ MOD 3.2 FOR STEAM RIALS LICENSES. PROGRAM-SPECIFIC GUIDANCE ABOUT ACA-DEMIC.RESEARCH AND DEVELOPMENT AND OTHER UCENSES CONDENSATION EXPERIMENTS IN THE PRESENCE OF NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. NU E 5 V R UDA DANCE ABOUT MATE- NUREGAA-0149: ASSESSMENT OF RELAP5/ MOD 3.2-NPA3.4 AGAINST A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR ] RIALS UCENSES. Program-SpecWe Guidance About Specific Ucenses O """- " Dran Report For Comment' TRIP NATURAL CIRCULATION AND THE START OF A MAIN PUMP  ; IN THE VANDELLOS 11 NUCLEAR POWER PLANT. { Protective Action NUREG/lA-0150 STUDY OF TRANSIENTS RELATED TO AMSAC AC-NUREG-1633 DRFT FC: ASSESSMENT OF THE USE OF POTASSIUM TUATION.SENSITMTY ANALYSIS. IODIDE (KI) AS A PUBUC PROTECTIVE ACTION DURING SEVERE l REACTOR ACCIDENTS. Draft Report For Comment. M N RE A 0143: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE LSTF Public Information Circular EXPERIMENT SIMULATING A LOSS OF RESIDUAL HEAT REMOVAL NUREG-0725 R13: PUBUC INFORMATION CIRCULAR FOR SHIP. EVENT DURING MID-LOOP OPERATION. MENTS OF IRRADIATED REACTOR FUEL Radletion Control Public Liability NUREG-1556 V10 DR FC: CONSOLIDATED GUIDANCE ABOUT MATE-NUREG/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE RIALS LICENSES. Program Specific Guidance About Master Materials $ BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. Lcenses. Draft Report For Comment, l Radiation Embrittlement PumfREG/CP-0152 Nt V02: PROCEEDINGS OF THE FIFTH NRC/ASME NUREG-1426 V03: COMPILATION OF REPORTS FROM RESEARCH SYMPOSIUM ON VALVE AND PUMP TESTING. SUPPORTED BY THE ELECTRICAL, MATERIALS AND MECHANICAL Quality Amourence ENGINEERING BRANCH, DIVISION OF ENGINEERING. NUREG 1505 RO1: A NONPARAMETRIC STATISTICAL METHOD-OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE. R on x sure COMMISSIONING SURVEYS. Interim Report For Use And Comment. COMMERCIAL NUCLEAR POWER REACTORS. Data Through Sep-QuentHe Test tomber 1997. NUREG-1505 RO1: A NONPARAMETRIC STATISTICAL METHOD-OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE. Radiation Model COMMISSIONING SURVEYS. Interim Report For Use And Comment. NUREGAA 0146: IMPLEMENTATION AND ASSESSMENT OF IM-PROVED MODELS AND OPTIONS IN TRAC-BF1. NU  : DATING AND EARTHOUAKES: REVIEW OF QUA- Radiation Protection TERNARY GEOCHRONOLOGY AND ITS APPUCATION TO NUREG/CR4418: RISK IMPORTANCE OF CONTAINMENT AND RE-PALEOSEISMOLOGY, LATED ESF SYSTEM PERFORMANCE REQUIREMENTS.

l 64- Sub)SCt ifHl0X Redenen assoty commities Resetor Pressure Vessel , NUREG 1666 V10 DR FC: CONSOUDATED GUIDANCE ABOUT MATE- NUREG-1426 V03: COMPILATION OF REPORTS FROM RESEARCH  ; RIALS LICENSES. Program Specific Guidance About Meeler Materiais SUPPORTED BY THE ELECTRICAL, MATERIALS AND MECHANICAL i Licenses. Draft Report For Comment ENGINEERING BRANCH, DIVISION OF ENGINEERING. 1 NUREG/CR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE 7 5 RONMENTALLY ASSISTED CRACKING IN E LIGHT-WATER REACTORS. Semannual Report. July-December 1997. NUREG/CR4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE-g ACTOR PRESSURE VESSELS. Density And Distribution Of Flaw indi-NUREG-0363 V01 R21: DIRECTORY OF CERTIFICATES OF COMPLl- NU EG/C 51 MPROVED EMBRITTLEMENT CORRELATIONS FOR ANCE F RADOACTIVE MATERIALS PACKAGES. Report Of NRC-REACTOR PRESSURE VESSEL STEELS. NU G-0363 V02 R21: DIRECTORY OF CERTIFICATES OF COMPLI-ANCE FOR RADOACTIVE MATERIALS PACKAGES. Certificates Of - G/C :T HNICAL BASIS FOR ENVIRONMENTAL QUAll-NU 'V03 R16: DIRECTORY OF CERTIFICATES OF COMPLl. FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED ANCE FOR RADOACTIVE MATERIALS PACKAGES. Report Of NRC- EQUIPMENT IN NUCLEAR POWER PLANTS. Approved Quality Assurance Programs For Radoective Malerials Pack-80es. NUREG-1560 V01 P1: INDIVIDUAL PLANT EXAMINATION PROGRAM: Racesotive metadal PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM-NUREG/CR-4554 V01 R2: SCANS $ HIPPING CASK ANALYSIS SYS- ANCE. Summary Report. TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP. NUREG-1560 V02 P24: INDIVIDUAL PLANT EXAMINATON PRO-PING CASK DESIGN REVIEW. User's Manual to Version 3a. GRAIA PERSPECTIVES ON REACTOR SAFETY AND PLANT PER-FORMANCE. 7- NUREG-1560 V03 P6: INDIVIDUAL PLANT EXAMINATON PROGRAM: NUlIEd-1631: SOURCE DISCONNECTS RESULTING FROM RADIOG- PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM. RAPHY DRIVE CABLE FAILURES. Final Report. ANCE. Appendices. NUREG/CP-0162 V01: PROCEEDINGS OF THE TWENTY-FIFTH RoselegleelCettede WATER REACTOR SAFETY INFORMATION MEETING. Plenary See-NUREG-1549 DRFT FC DECISION METHODS FOR DOSE ASSESS- sene, Pressure Veneel Research,BWR Strainer 86ockage And Other MENT TO COMPLY WITH RADIOLOGICAL CRITERIA FOR LICENSE Generic Safety leaues. Environmentally Assisted Degradotion Of LWR.. TERMINATION. Draft Report For Comment NUREG/CP 0162 V02: PROCEEDINGS OF THE TWENTY-FIFTH NUREG/CP-0163: PROCEEDINGS OF THE WORKSHOP ON REVIEW WATER REACTOR SAFETY INFORMATION MEETING. Human Reli-OF DOSE MODELING METHODS FOR DEMONSTRATION OF COM- ability AnrJysis And Human Performance Evaluelen Technscal leeues PLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TER- Related To Rulemakings, Risk-informed Performance-Based inilm-MINATlON* tivos. __ . NUREG/CP-0162 V03: PROCEEDINGS OF THE TWENTY FIFTH WATER REACTOR SAFETY INFORMATION MEETING.Thermel-Hy-NURED/CR 3 V01: CODE MANUAL FOR MACCS2. User's Guide. draulic Reneerch And Codes, Digital instrumonlation And Control, Resoinglesi Response Structural Performance. NUREG-1634: 1997 LOST SOURCE EXERCISE.An Exercise Of Radio. NUREG/CR4418: RISK IMPORTANCE OF CONTAINMENT AND RE-

     . logical Response Through Cooperation And ' Coordination Of               LATED ESF SYSTEM PERFORMANCE REQUIREMENTS.
             . State And Federal Resources Under The National Contingency NUREG/CP-0165: TRANSACTIONS OF THE TWENTY-SIXTH WATER
                " Survey                                                        REACTOR SAFETY INFORMATION MEETING.

NURED-1507: MINIMUM DETECTABLE CONCENTRATONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON. Reactor Shutdown TAMINANTS AND FIELD CONDITIONS. NUREG/CR-6502: ACTION REQUIREMENTS FOR AFW SYSTEM FAIL-

  ' NUREG/CR4364: HUMAN PERFORMANCE IN RADIOLOGICAL SUR-                        URES.An Analysis For Four Nucieer Power Plants.

_ VEY SCANNING. Redenuellde NUREG/lA-0149: ASSESSMENT OF RELAP5/ MOD 3.2-NPA3.4 AGAINST NUREG/CR4377: EFFECTS ON RADONUCLIDE CONCENTRATONS A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF TRIP,NATilR* CIRCULATION AND THE START OF A MAIN PUMP PERFORMANCE ASSESSMENT OF LOW-LEVEL RADIOACTIVE IN THE V.4NDaLOS 11 NUCLEAR POWER PLANT, WASTE DISPOSAL FACILITIES. Regulation Radionuellde Behavior NUREG-1608: CATEGORIZING AND TRANSPORTING LOW SPECIFIC NUREG/CR4131: VICTORIA 2.0: A MECHANISTIC MODEL FOR ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS. RADONUCLIDE BEHAVIOR IN A NUCLEAR REACTOR COOLANT NUREG/CR-5342: ASSESSMENT AND RECOMMENDATIONS FOR SYSTEM UNDER SEVERE ACCIDENT CONDITIONS. FISSILE-MATERIAL PACKAGING EXEMPTIONS AND GENERAL Li-CENSES WITHIN 10CFR PART 71 NUREG/CR4604: RADTRAD: A SIMPLIFIED MODEL FOR RADIO- Regulatory Agende M N. NUREG-0936 V16 NO2: NRC REGULATORY AGENDA. Semiannual Re-  ; NUR G 1 0 C L MAC . e port. July-Docember 1997. Reacter Coolant System NUREG-0930 V17 N01: NRC REGULATORY AGENDA.Somennual Re-NUREG/CR4131: VICTORIA 2.0: A MECHANISTIC MODEL FOR port JanuarNune 1998. RADIONUCLIDE BEHAVIOR IN A NUCLEAR REACTOR COOLANT " I SYSTEM UNDER SEVERE ACCIDENT CONDITIONS. g y22 I LATORY AND TECHNICAL REPORTS Reester Opeestion (ABST'4ACT INDEX JOURNAL). Compilation For Third Quarter ) NUREG-1635 V01: REVIEW AND EVALUATION OF THE NUCLEAR 1997. July-September, j REGULATORY COMMISSION SAFETY RESEARCH PROGRAM.A Re- I port To The US Nuclear Regulatory Commission. Reguhtory And Technical Report NUREG-0304 V22 N04: REGULATORY AND TECHNICAL REPORTS Remoter Oposetor (ABSTRACT INDEX JOURNAL). Annual Compilaten For 1997, l NUMEG-1122 R02: KNOWLEDGE AND ABILITIES CATALOG FOR NU-CLEAR POWER PLANT OPERATORSPressur! red Water Reactors. Reinforced Concrete

   . NOREG.1123 R02: KNOWLEDGE AND ABILITIES CATALOG FOR NU-                   NUREG/CR4554: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR CLEAR POWER PLANT OPERATORS.Bo6 ling Water Reactore.                     WALL INTERNATIONAL STANDARD PROBLEM.

Subject index 65 Renesee Frequency Safety Research Program NUREG-1635 V01: REVIEW AND EVALUATION OF THE NUCLEAR E R4615: A SURVEY OF REPAIR PRACTICES FOR NUCLEAR TT S R at n POWER PLANT CONTAINMENT METALUC PRESSURE BOUND-ARIES. Sealed Source NUREG-1556 V03: CONSOLIDATED GUIDANCE ABOUT MATERIALS E NSES.Appkatons 6 SeaW Soua And Due Evaluaton W V REPORT TO CONGRESS ON ABNORMAL OCCUR-RENCES. Fiscal Year 1997. Registraton. Final Report. NUREG-1415 V10 NO2: OFFICE OF THE INSPECTOR GEN- Seismic ERAL. Semiannual Report To Congress October 1,1997 - March 31, 1990- NUREG/CR-5434: ANCHOR BOLT BEHAVIOR AND STRENGTH DUR-ING EARTHOUAKES. Residual Heat Removal NU ^

                         ^                                   " "           NUR EXP M           S MU       NG       SO R     UAL    T E     A                 5 1: SEISMIC ANALYSIS OF PIPING. Final Program Re-EVENT DURING MID-LOOP OPERATION.                                  NU%GCR4554:FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR Risk Assessment                                                          WALL INTERNATIONAL STANDARD PROBLEM.

NUREG-1570: RISd ASSESSMENT OF SEVERE ACCIDENT-INDUCED l NF i EhTS FOR AFW SYSTEM FAIL- NUR GCR4 : INVESTIGATION OF TECHNIQUES FOR THE DE-URES.An Analysis For Four Nuclear Powes Plants. /ELOFMENT OF SEISMIC DESIGN BASIS USING THE PROB-NUREGCR4579: DIGITAL l&C SYSTEMS IN NUCLEAR POWER ABluSTIC SEISMIC HAZARD ANALYSIS. PLANTS. Risk-Screening Of Environmental Stressors And A Compan-an Haware Unavam mh An Emsting Analog Men W N 4 : 1 ESTIGATION OF TECHNIQUES FOR THE DE- l Riek Comparison VELOPMENT OF SEISMIC DESIGN BASIS USING THE PROB- 1 NUREG/CR4616: RISK COMPARISON OF SCHEDULING PREVENTIVE ABlUSTIC SEISMIC HAZARD ANALYSIS. I MAINTENANCE DURING SHUTDOWN VS. DURING POWER OPER-ATION FOR PWRS. Seismic Mount l NUREGCR4564: ANALYSES OF SOURCE SPECTRA. ATTENUATION, Risk 4nformed AND SITE EFFECTS FROM CENTRAL AND EASTERN UNITED NUREG-1556 V09 DR FC: CONSOUDATED GUIDANCE ABOUT MATE- STATES EARTHQUAKES. RIALS LtCENSES. Program-Specific Guidance About Medical Use U-NUR 59: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM Rink-informed Regulation AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND NUREG 1521 DRFT FC: TECHNICAL REVIEW OF RISK-INFORMED, ENERGY-ABSORBING SUPPORTS. PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT FIRE PROTECTION ANALYSES. Draft Report For Comment. Se s u 4 556: A COMPREHENSIVE STUDY OF THE EASTERN Rules TENNESSEE SEISMIC ZONE. NUREG-0936 V16 NO2: NRC REGULATORY AGENDA. Semiannual Re-portJulv-December 1997. Seismotecton6c NUREG-6936 V17 NOI: NRC REGULATORY AGENDA. Semiannual Re- NUREGCR4573: "lNVESTIGATING SEISMOTECTONICS IN THE portJanuary-June 1998. EASTERN UNITED STATES USING A GEOGRAPHIC INFORMATION Rules Of Practice NUREG 0386 D09: UNITED STATES NUCLEAR REGULATORY COM- Self-Shielded trredletor MISSION STAFF PRACTICE AND PROCEDURE DI- NUREG-1556 V05: CONSOLIDATED GUIDANCE ABOUT MATERIALS GEST.Comrmssion, Appeal Board And Ucensing Board DecisonsJuly UCENSES. Program-Specific Guidanw About Self-Shielded irradiator 1972 - June 1997. Ucenses. Final Report.

     $5LOCA                                                              Semlannual Report To Congress NUREG/lA 0144: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE               NUREG 1415 VII N01: OFFICE OF THE INSPECTOR GEN-Senwannual ep n          ngress.AprWSeper W98.

NU E 1 A P5 S S T GA AE E PERI- , MENTAL DATA (REVISION 1). Semiscale NUREG/lA-0144. ASSESSMENT OF RELAP5' MOD 32 WITH THE SCANS SEMISCALE NATURAL CIRCULATION EXPERIMENT, S-NC-8B. NUREG/CR-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS-TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP- Sensitivtty Analysis PING CASK DESIGN REVIEW. User's Manual to Verson 3a. NUREGCR 5621: GROUND-WATER MODELS IN SUPPORT OF NUREGCR-5512. SCDAP NUREG/lA-0150: STUDY OF TRANSIENTS RELATED TO AMSAC AC-NUREGCR4150 V01 R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN- DMUEE AW66 UAL1nterface Theory. NUREGOR4150 V02 R1: SCDAP/PELAP5/ MOD 3.2 CODE MAN- Severe Accident NUREG 1560 V01 P1: INDIVIDUAL PLANT EXAMINATION PROGRAM: NR / 15 b: S P L'APS/ MOD 32 CODE MAN-PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM-UALUser's Guide And input Manual. ANCE. Summary Report. - NULEGCR4150 V04 RI: SCDAP/RELAP5/ MOD 3.2 CODE MAN. NUREG-1560 V02 P2-5. INDIVIDUAL PLANT EXAMINATION PRO-UAL.MATPRO - A Ubrary Of Matenals Properties For Ught-Water Re-actor Accident Anafysis. GRAM: PERSPECTIVES ON REACTOR SAFETY AND PLANT PER-NURE R4 V05 DAP/RELAP5/ MOD 3.2 CODE MAN. FR E p PERSPECTIVES ON REACTOR SAFETY AND PLANT PERFORM-Safeguardo Summary Event List ANCE.Annendices. NUREG 0525 V02 ROS: SAFEGUARDS

SUMMARY

EVENT UST NUREG-1670: RISK ASSESSMENT OF SEVERE ACCIDENT-INDUCED i STEAM GENERATOR TUBE RUPTURE.  ; (SSEL) January 1,1990 Through December 31,1997. NUREG-1624 DRFT FC: TECHNICAL BASIS AND IMPLEMENTATION i Safety Analysis Report GUIDEUNES FOR A TECHNIQUE FOR HUMAN EVENT ANALYSIS l NUREGOR-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS- (ATHEANA). Draft Report For Comment. TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP- NUREGCP4162 vot: PROCEEDINGS OF THE TWENTY FIFTH PING CASK DESIGN REVIEW. User's Manual to Verson 3a. WATER REACTOR SAFETY INFORMATION MEETING. Plenary See-l \ l_ __ ._ ____ - - - -

66 Subject index sens.Preneure Vesest Research.BWR Stromer Blocke9e And Other Generic Soloty leeues Environmenteily Assisted Dooredeten Of LWR.. NUREG-1628 DRF FC: STAFF RESPONSES TO FREQUENTLY ASKED ) QUESTONS CONCERNING DECOMMISSIONING OF NUCLEAR ' NUREGOR4312: EXPdRIMENTS ON INTERACTIONS BETWEEN ZlR- POWER PLAN 1S. Draft Report For Comment CONIUM CONTAINING MELT AND WATER. NUREGCR4119 V01 R1: MELCOR COMPUTER CODE MANU- Spent Nucteer Fuel ALS.Pnmer And Users' Guidea. Vernon 1.8 4. July 1997. NUREG-1626: FINAL ENVIRONMENTAL IMPACT STATEMENT FOR NUREGCR4119 V02 R1: MELCOR COMPUTER CODE MANU- THE CONSTRUCTON AND OPERATION OF AN INDEPENDENT ALS. Reference Manuals.Vereen t.8.4, July 1997. SPENT FUEL STORAGE INSTALLATION TO STORE THE THREE NUREGCR4131: VICTORIA 2.0- A MECHANISTIC MODEL FOR MILE ISLAND UNIT 2 SPENT FUEL AT THE IDAHO NATIONAL ENGi-RADONUCUDE BEHAVOR IN A NUCLEAR REACTOR COOLANT NEERING AND ENVIRONMENTAL SYSTEM UNDER SEVERE ACCIDENT CONDITIONS. NUREGCR4475: RESOLUTION OF THE DIRECT CONTAINMENT Splitting Tenelle Strength HEATING ISSUE FOR COMBUSTON ENGINEERING PLANTS & NUREGCR4598: AN INVESTIGATION OF TENDON SHEATHING BABCOCK & WILCOX PLANTS. FILLER MiGRATON INTO CONCRETE. NUREGCR4575: FAILURE BEHAVIOR OF INTERNALLY PRFSSUR-12ED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT Stelnises Steel HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH NUREGCR4594: EVALUATION OF ULTRASONIC INSPECTION TECH. MODEL PREDCTIONS. . NIQUES FGA COARSE-GRAINED MATERIALS. NUREGCR4604: RADTRAD: A SIMPUFIED MODEL FOR RADIO-NUCLIDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION. 8 G- S DARD REVIEW PLAN FOR PHYSICAL PROTEC-Severe Core Damage TION PLANS FOR THE INDEPENDENT STORAGE OF SPENT FUEL NUREGCR-4674 V25: PRECURSORS TO POTENTIAL SEVERE CORE AND HIGH-LEVEL RADOACTIVE WASTE. DAMAGE ACCIDENTS: 1996. A Status Report. NUREGCR-4874 V26: PRECURSORS TO POTENTIAL SEVERE CORE Standard Technical Specificatione DAMAGE ACCIDENTS:1997. A Status Report. NUREG-1625 DRF FC: PROPOSED STANDARD TECHNICAL SPECl-FICATIONS FOR PERMANENTLY DEFUELED WESTINGHOUSE Severo Remotor Accident PLANTS. Draft Report For Cornment. NUREG-1633 DRFT FC: ASSESSMENT OF THE USE OF POTASSIUM IODIDE (KI) AS A PUBUC PROTECTIVE ACTION DURING SEVERE Static REACTOR ACCIDENTS. Draft Report For Comment, NUREGCR4517: ROUND ROBIN PRETEST ANALYSES OF A STEEL CONTAINMENT VESSEL MODEL AND CONT ACT STRUCTURE AS-Sheer Well SEMBLY SUBJECT TO STATIC INTERNAL PRESSURIZATION. NUREGCR4554: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR WALL INTERNATIONAL STANDARD PROBLEM. Steam Condoneet6on NUREG/lA-0147: ASSESSMENT OF RELAP5/ MOD 3.2 FOR STEAM Shipment CONDENSATION EXPERIMENTS IN THE PRESENCE OF NUREG4725 R13: PUBLIC INFORMATION CIRCULAR FOR SHIP- NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. MENTS OF IRRADIATED REACTOR FUEL Steam Empicolon Cook Design NUREGCR-5372: EXPERIMENTS ON INTERACTIONS BETWEEN ZlR-NURE R-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS- CONIUMCONTAINING MELT AND WATER. TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP-PING CASK DESIGN REVIEW. User's Manuel to Vernon 3e, or Shutdoum STEAM GENERATOR TUBE RUPTURE. NUREGCR4618: RISK COMPARISON OF SCHEDUUNG PREVENTIVE NUREGCR4511 V02. STEAM GENERATOR TUBE INTEGRITY PRO-NTENANC D RING SHUTDOWN VS. DURING POWER OPER- RAgAgl Agggig-g Ig. GRAM Somaannual Report October 1996 - March 1997. Sitoone Foam NUREGCR4521: ESTIMATING PROBABLE FLAW DISTRIBUTIONS IN NUREG-1552 S01 DR FC: FIRE BARRIER PENETRATION SEALS IN PWR STEAM GENERATORS. NUCLEAR POWER PLANTS. Drelt Repor1 For Comment NUREGCR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR-IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT Sofhuere Reference Manuel HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH NUREGCR4268 V04: COMMON CAUSE FAILURE DATABASE AND MODEL PREDICTIONS. ANALYSIS SYSTEM. Volume 4: Software Reference Manual. NUREG/GR-0016: THE ROLE OF TIME-DEPENDENT DEFORMATION IN INTERGRANULAR CRACK INITIATION OF ALLOY 600 STEAM Soll GENERATOR TUBING MATERIAL NUREGCR4603: CHARACTERIZATION OF RETARDATION MECHA-NISMS IN SOIL Steel NUREGCR4551: IMPROVED EM83RITTLEMENT CORRELATIONS FOR

• • "h"~t REACTOR PRESSURE VESSEL STEELS.

NUREGCR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM. Steel Siliet NUREGCR4608:

SUMMARY

AND EVALUATON OF LOW-VELOCITY Source Disconnect IMPACT TEST OF SOUD STEEL BILLET ONTO CONCRETE PADS. NUREG-1631: SOURCE DISCONNECTS RESULTING FROM RADIOG-RAPHY DRIVE CABLE FAILURES. Fool Report. Steel Containment Vessel NUREGCR4517: ROUND ROBIN PRETEST ANALYSES OF A STEEL Source Spectre CONTAINMENT VESSEL MODEL AND CONTACT STRUCTURE AS-NUREGCR4564: ANALYSES OF SOURCE SPECTRA. ATTENUATION. SEMBLY SUBJECT TO STATIC INTERNAL PRESSURIZATION. AND SITE EFFECTS FROM CENTRAL AND EASTERN UNITED STATES EARTHOUAKES. St 629 THE CHARACTERIZATION OF VICKER'S MICROHARD-Source term NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-NUREGCR4410: NUCLEAR FUEL CYCLE FACluTY ACCIDENT ANAL- HARDENING MICROPROBE. YSIS HANDBOOK. NUREGCR4418: RISK IMPORTANCE OF CONTAINMENT AND RE- Strainer Stockage LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. NUREGCP-0162 VO1: PROCEEDINGS OF THE TWENTY-FIFTH WATER REACTOR SAFETY INFORMATION MEETING. Plenary See-  ! Spent Fuel sons. Pressure Vessel Research.BWR Strainer B6ockage And Other l NUREG0725 R13: PUBUC INFORMATION CIRCULAR FOR SHIP- Genene Safety issues. Environmentally Assisted Degradeten Of LWR. MENTS OF IRRADIATED REACTOR FUEL NUREG-1619: STANDARD REVIEW PLAN FOR PHYSICAL PROTEC- Stresa Corros6on Crock 6ng TION PLANS FOR THE INDEPENDENT STORAGE OF SPENT FUEL NUREGCR-4667 V25: ENVIRONMENTALLY ASSISTED CRACKING IN AND HIGH-LEVEL RADIOACTIVE WASTE. UGHT. WATER REACTORS. Semiennual Report. July-December 1997.

Subject Index 67 NUREG/CR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO- The Price Anderson Act GRAM. Annual Report.Aunuat 1995 Seplerrter 1996. NUREG/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE NUREG/CR4511 V03: STEAM GENERATOR TUBE INTEGRITY PRO- BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS. GRAM.Sem ennual Report October 1996 - March 1997. NUREG/CR4521: ESTIMATING PROBABLE FLAW DISTRIBUTIONS IN Thermal Aging PWR STEAM GENERATORS. NUREG/CR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON NUREG/GR-0016: THE ROLE OF TIME-DEPENDENT DEFORMATION THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR l IN INTERGRANULAR CRACK INITIATION OF ALLOY 600 STEAM PRESSURE VESSEL MATERIALS.An Atom Prote Study. I GENERATOR TUBING MATERIAL l Thermal Anneeling Strees NURE R4564: ANALYSES OF SOURCE SPECTRA, ATTENUATION- NUREG/CR4552: MARBLE HILL ANNEAUNG DEMONSTRATION EVALUATION. AND SITE EFFECTS FROM CENTRAL AND EASTERN UNITED STATES EARTHQUAKES. Therrnal Sindl NUREG/CR 11: RESULTS OF PRESSURE LOCKING AND THERMAL G R MARBLE HILL ANNEAUNG DEMONSTRATION EVALUATION. Thermal Hydraulle Reeeerch Structural Eng6neerin9 NUREG/CP-0162 V03: PROCEEDINGS OF THE TWENTY-FIFTH WATER REACTOR SAFETY INFORMATION MEETING. Thermal-Hy-NUREG/CR4546: A DAMAGE MECHANICS BASED APPROACH TO draulic Research And Codes, D+gita: Instrumentation And Control, STRUCTURAL DETERIORATION AND REUABluTY. Structural Performance. Stucturali Thermalineutetion NUREG/CR- 7: APPUCATION OF THE NCSA HABANERO TOOL FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS- NUREG/CR-5534: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A MENTS. DOWNWARD FACING CURVED SURFACE: EFFECTS OF THERMAL INSULATION. Structural Performance T NUREG/CP-0162 V03: PROCEEDINGS OF THE TWENTY-FIFTH WATER REACTOR SAFETY INFORMATION MEETING. Thermal-Hy- 63 EVALUATION OF AP600 CONTAINMENT THERMAL-HY. dreule Research And Codes. Dgtal Instrumentation And Control. DRAUUC PERFORMANCE NUREG/CR-5498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR-Structural Performance CULATION TESTS IN THE PUMA FACILITY. m g, NUREG/CR4359 V02: RAMONA-4B: A COMPUTER CODE WITH NUREG/IA 002 5/ MOD 3 SUBCOOLED BOILING MODEL AS- THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR SESSMENT. SYSTEM TRANSIENTS. User's Manual. Surface Condition Thermolum6neeoent Dosimeter NUREG/CR4589: THE EFFECTS OF SURFACE COrJDITION ON AN NUREG0837 V17 NO3: NRC TLD D RECT RADIATION MONITORING ULTRASONIC INSPECTION: ENGINEERING STUDIES USING VAU- NETWORK. Progress Report. July-September 1997 DATED COMPUTER MODEL. Tule um Surface Contaminated Object NUREG-0540 V19 N11: TITLE LIST OF DOCUMENTS MADE PUBUCLY NUREG-1606: CATEGORIZING AND TRANSPORTING LOW SPECIFIC AVAILABLE. November 1-30 1997. ACTIVITY MATERIALS AND SURFACE CONTAMINATED Oa;ECTS. NUREG-0540 V19 N12: TITLd UST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1 31 1997. Surteos Creek NUREG 0540 V20 Nol: TITLd LIST OF DOCUMENTS MADE PUBUCLY NURE R4540 ATE-OF-THE-ART REPORT ON PIPING FRAC' NU N N: LNIST OF DOCUMENTS MADE PUBLICLY AVAILABLE.Februarv 148,1998. Survey instrument NUREG-0540 V20 N01 TITLE UST OF DOCUMENTS MADE PUBUCLY NUREG-1507: MINIMUM DETECTABLE CONCENTRATIONS WITH AVAILABLE. March 1-31 1998. TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- NUREG-0540 V20 N04: TITLE UST OF DOCUMENTS MADE PUBUCLY TAMINANTS AND FIELD CONDITIONS AVAILABLE. April 1-30 1998. NUREG'CR4364: HUMAN PERFORMANCE IN RADIOLOGICAL SUR- NUREG-0540 V20 N05: TITLE UST OF DOCUMENTS MADE PUBUCLY 0

• NUI
• V2 10 TITL UST OF DOCUMENTS MADE PUBUCLY System Transient AVAILABLE. June 1-30 1998.

NUREG/CR4359 V01: RAMONA 4B: A COMPUTER CODE WITH NUREG 0540 V20 N07:IITLE UST OF DOCUMENTS MADE PUBUCLY EM RANS NTS nd ela NU k  :'Tib LIST OF DOCUMENTS MADE PUBUCLY AVAILABLE. August 1-31 1998. NUREGCR4359 V02: RAMONA-48: A COMPUTER CODE WITH THREE DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR NUREG-0540 V20 N09: title UST OF DOCUMENTS MADE PUBUCLY SYSTEM TRANSIENTS. User's Manual. NU V2YN IT'L Li T OF DOCUMENTS MADE PUBLICLY System UnreHebHNy AVAILABLE. October 1-31,1998. NUREG/CR-5500 V01: RELIABluTY STUDY: AUXtUARY/ EMERGENCY " FEEDWATER SYSTEM, 1987 1995. UR A 0150 STUDY OF TRANSIENTS RELATED TO AMSAC AC-TLD TUATION, SENSITIVITY ANALYSIS, NUREG-0837 V17 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report. July-September 1997. U G V01 R21: DIRECTORY OF CERTIFICATES OF COMPU-TMk2 ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC-NUREG 1626: FINAL ENVIRONMENTAL N8ACT STATEMENT FOR Atoroved Packages. THE CONSTRUCTION AND OPERATION OF AN INDEPENDENT NUREG-0383 V02 R21: DIRECTORY OF CERTIFICATES OF COMPLi-SPENT FUEL STORAGE INSTALLATION TO STORE THE THREE ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certifu:ates Of

                                                 "               ^

N $V03 R18. DIRECTORY OF CERTIFICATES OF COMPU-IN AND V RO TA ~ ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC-TRAC SF1 Approved Quality Assurance Programs For Radioactive Materials Pack-NUREG/lA-0146: IMPLEMENTATION AND ASSESSMENT OF IM- a0es. PROVED MODELS AND OPTIONS IN TRAC-BF1. NUREG-1608: CATEGORIZING AND TRANSPORTING LOW SPECIFIC ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS. Tendon Sheath 6ng NUREG/CR-5342: ASSESSMENT AND RECOMMENDATIONS FOR NUREG/CR4595: AN INVESTIGATION OF TENDON SHEATHING FISSILE-MATERIAL PACKAGING EXEMPTIONS AND GENERAL U. FILLER MIGRATION INTO CONCRETE. CENSES WITHIN 10CFR PART 71.

68 Subject index Tube Vadose Zone NUREG/CR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO- NUREGICR4621: GROUND-WATER MODELS IN SUPPORT OF NUREG/CR-5512. N R G/CR 1 TWM NE TUB INTEGRITY PRO-N REG /C NS INNG LEN TRIBUT60NS IN Valvo PWR STEAM GENERATORS. NUREG/CP-0152 V02: PROCEEDINGS OF THE FIFTH NRC/ASME SYMPOSIUM ON VALVE AND PUMP TESTING. Tube Rupture NUREG-1570: RISK ASSESSMENT OF SEVERE ACCIDENT-INDUCED Vandellos 11 STEAM GENERATOR TUBE RUPTURE. NUREG/lA 0149: ASSESSMENT OF RELAP5 MOD 32-NPA3.4 AGAINST Tubing A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION REACTOR NUREG/CR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR- TRIP, NATURAL CIRCULATION AND THE START OF A MAIN PUMP IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT IN THE VANDELLOS 11 NUCLEAR POWER PLANT. H'GH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH MODEL PREDICTIONS. Vendor inspection NUREG-0040 V21 N04: UCENSEE CONTRACTOR AND VENDOR IN-SPECTION STATUS REPORT. Quarterly Report, October-December NUR 60: PROCEEDINGS OF THE OECD/CSNI SPECIAUST MEETING ON ADVANCED INSTRUMENTATION AND MEASURE-1887 "I MENT TECHNIQUES. Held in Santa Barbara CA. March 17-20,1997. NUREG'I#hi o040 V22 N01:0 LICENSEE CCNTRACTOR AND VENDOR IN-SPECTION STATUS REPORT. Quarterty Report. January-March Ultrasonic innervice inspect 6on 1998.(White Book) NUREG/CR4605: AN EVALUATION OF HUMAN FACTORS RESEARCH NUREG-0040 V22 NO2: UCENSEE CONTRACTOR AND VENDOR IN-FOR ULTRASONIC INSERVICE INSPECTION. SPECTION STATUS REPORT. Quaderly Report.Apnl-June Ultrasonic inspect 6on 98 M e Boom NUREG/CR4589: THE EFFECTS OF SURFACE CONDITION ON AN Vertical Tube LT ASO C PEC I ENGINEERING STUDIES USING VAU-NUREG/lA-0147: ASSESSMENT OF RELAP5 MOD 3.2 FOR STEAM CONDENSATION EXPERIMENTS IN THE PRESENCE OF Ultrasonic Sensor NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. NUREG/CR4501: ADVANCED INSTRUMENTATION AND MAINTE-NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. Vicker NUREG-1629: THE CHARACTERIZATION OF VICKER'S MICROHARD-N R4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE- NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-ACTOR PRESSURE VESSELS. Density And Distnbution Of Flaw Ird- HARDENING MICROPROBE. cations in PVRUF* 5 U ncertelnty Analysis NUREG/CR4545 V01: PROBABILISTIC ACCIDENT CONSEQUENCE Weste Burial NUREG 1307 R08: REPORT ON WASTE BURIAL CHARGES Changes in Decommissioning Waste Disposal Costs At Low-Level Waste Bunal UNCERTAINTY ANALYSIS. Early Health Ef1ects Uncertainty Assess- Facihties. NUYE 4 V02: PROBABluSTIC ACCIDENT CONSEQUENCE Wold UNCERTAINTY ANALYSIS. Earty Health Effects Uncertainty Assess-NUREG/CR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE NUYE V01: PROBABluSTIC ACCIDENT CONSEQUENCE PROBABILITIES OF DEFECTS IN REACTOR PRESSURE VESSEL UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess. WELDS. NU 4 V02: PROBADILISTIC ACCIDENT CONSEQUENCE Welding UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- NUREG/CR4615: A SURVEY OF REPAIR PRACTICES FOR NUCLEAR NU V01: PROBABlUSTIC ACCIDENT CONSEQUENCE

                                                                                       "               ^           ^        "

A ' UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do-UC A TY LYSI U i A s nt er UREG-1625 DR FC: PROPOSED STANDARD TECHNICAL SPECl-simetry.Appendees. FICATIONS FOR PERMANENTLY DEFUELED WESTINGHOUSE PLANTS. Draft Report For Comment. Unnatursted Fractured Tuff NUREG/CR4559: SINGLE- AND CROSS-HOLE PNEUMATIC TESTS IN Wilcoxon Rank Sum Test UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-SEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA. NUREG 1505 R01: A NONPARAMETRIC STATISTICAL METHOD-BILITY,CONNECTIVITY AND SCALE. OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-COMMISSIONING SURVEYS.Intenm Report For Use And Comment. VICTOR 6A 2 NUREG/CR4131: VICTORIA 2.0: A MECHANISTIC MODEL FOR Workshop RADIONUCUDE BEHAVIOR IN A NUCLEAR REACTOR COOLANT NUREG/CP-0164: PROCEEDINGS OF THE WORKSHOP ON ELECTRIC SYSTEM UNDER SEVERE ACCIDENT CONDITIONS. DOSIMETRY. Held in Garthersburg. Maryland On October 14-16, 1997. I l f I _. .. .. . . . _ _ _ _ _ _ _ _ _ _ . _

n l NRC Originating Organization index (Staff Reports) This index lists those NRC organizati6ns that have published staff reports. The index is ar-r nged alphabetically by major NRC organizations (e.g., program offices) and then by sub-sections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number. NUREG-1272 V10 NO2: OFFICE FOR ANALYSIS AND EVALUATION ADVISORY COMMITTEE (S)ON NUCLEAR WASTE ADVISORY COMMITTEE OF OPERATIONAL DATA.1996 Annual Report NUREG-1423 V08: A COMPILATION OF REPORTS OF THE ADVl- NUREG-1272 V10 NO3: OFFICE FOR ANALYSIS AND EVALUATION SORY COMMITTEE ON NUCLEAR WASTE.Jutv 1997 - June 1998. OF OPERATIONAL DATA.1996 Annual Report ACRS - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUREG-1272 VII N01: OFFICE FOR ANALYSIS AND EVALUATION NOREG-1125 V19: A COMPILATION OF REPORTS OF THE ADV1- OF OPERATIONAL DATA.1997 Annual Report (Reactors) SORY COMMITTEE ON REACTOR SAFEGUARDS.1997 Annual. NUREG-1272 VII NO2: OFFICE FOR ANALYSIS AND EVALUATION NUREG-1635 V01: REVIEW AND EVALUATION OF THE NUCLEAR OF OPERATIONAL DATA.1997 Annual Repor1(Nuclear Matenals). REGULATORY COMMISSION SAFETY RESEARCH PROGRAM.A NUREG-1272 VII NO3: OFFICE FOR ANALYSIS AND EVALUATION Report To The US Nuclear Regulsdory Commission. OF OPERATIONAL DATA.1997 Annual Report (Techrncal Training). NUREG-1634: 1997 LOST SOURCE EXERCISE.An Exercise Of Radio. ATOMIC SAFFTY SOAR logical Response Through Cooperation And Coordination Of S & PANE % Local. State And Federal Resources Under The National Contingency N VD : OM A TY A L ENSING BOARD BIEN-NIAL REPORT. Fiscal Years 1995 - 1996. INCL T RESPONSE BRANCH OFF OF G CUT E RECTOR FOR OPERATIONS (EDO) S UM O D (K) A A PUBL C PROT CTIVE I D RING NUREG 0837 V17 NO3: RC TLD DIRECT RADIATION MONITORING SEVERE REACTOR ACCIDENTS. Draft Rg For Comrnent N J 8 - OFC OF C I S'T8po' U EG R  : UI U ODEU COMMON.CAUSE NUREG 0940 V16 N2 P : ENFORCE 4 ENT ACTIONS: SIGNIFICANT FAILURES IN PROBABILISTIC RISK ASSESSMENT. ACTIONS RESOLVED INDIVIDUAL ACTIONS. Semiannual Progress NUREG/CR 5497: COMMON CAUSE FAILURE PAMMETER ESTI-i NUR G NUREG/CRI6268 V01: COMMON CAUSE FAILURE DATABASE AND P2: NFORCEMENT ACTIONS: SIGNIFICANT ^ l ACTIONS RESOLVED REACTOR UCENSEES. Semiannual Progress NUR R 0. "M A CE FAILURE DATABASE AND NUR 4 P ENFORCEMENT ACTIONS: SIGNIFICANT NU E I  : SE I E D TY NND ACTIONS RESOLVED MATERIAL UCENSEES. Semiannual ANALYSIS SYSTEM. Volume 3: Data Collection And Event Coding. Progress Report. July-December 1997. NUREG 0940 v17 N1 P1: ENFORCEMENT ACTIONS: SIGNIFICANT EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS ACTIONS RESOLVED INDIVIDUAL ACTIONS. Semiannual Progress OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS Report. January June 1998. NUREG 0383 V01 R21: DIRECTORY OF CERTIFICATES OF COM-l NUREG-0940 V17 N1 P2: ENFORCEMENT ACTIONS: SIGNIFICANT PLIANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of ACTIONS RESOLVED REACTOR LICENSEES. Semiannual Progress NRC- ved Padages. l Report. January June 1998. NURE 83 V02 R21: DIRECTORY OF CERTIFICATES OF COM-NUREG 0940 V17 N1 P3: ENFORCEMENT ACTIONS: SIGNIFICANT PUANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates ACTIONS RESOLVED MATERIAL UCENSEES. Semiannual of Compliance. ! Progress Report. January-June 1998. NUREG-0383 V03 R18: DIRECTORY OF CERTIFICATES OF COM- ! NUREG-1600 M01: GENERAL STATEMENT OF POUCY AND PRO. PUANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of CEDURE FOR NRC ENFORCEMENT ACTIONS. Enforcement Policy. NRC-Approved Quality Assurance Programs For Radioactive Mate. NUREG1622: NRC ENFORCEMENT POUCY REVIEW. July 1995 - rials Packages. Juty 1997. NUREG-0430 V16: UCENSED FUEL FACILITY STATUS RE-PORT. inventory Difference Data. July 1,1995 - June 30.1995.(Gray EDO . OFFICE OF ADMINISTRATION (PRE 870413 & POST 800206) Book 11) OFFICE OF ADMINISTRATION, DIRECTOR (POST 940714) NUREG-0430 V17: UCENSED FUEL FACILITY STATUS FiE-NUREG0936 V16 NO2: NRC REGULATORY AGENDA. Semiannual PORT. inventory Diflerence Data. July 1,1998 - June 30,1997.(Gray ReportJuh-December 1997. Book 11) NUREG-0936 V17 N01: NRC REGULATORY AGENDA. Semiannual NUREG-0725 R13: PUBLIC INFORMATION CIRCULAR FOR SHIP-Report. January June 1998. MENTS OF IRRADI ATED REACTOR FUEL. NUREG-1608: CATEGORIZING AND TRANSPORTING LOW SPE-EDO . OFFICE OF THE CONTROLLER (PRE $20418 & POST 890206) CIFIC ACTIVITY MATERIALS AND SURFACE CONTAMINATED OFFICE OF THE CONTHOLLER (POST OEVECTS ! NUREG-1542 V03: ACCOUNTABluTY REPOR890205)T FISCAL YEAR NUREG-1f.iti iINAL 1997. ENVIRONMENTAL lMPACT STATEMENT FOR , 1 EG 0 :BU G ES I E F sea r 1999. THE CONSTRUCTION AND OPERATION OF AN INDEPENDENT l SPENT FUEL STORAGE INSTALLATION TO STORE THE THREE l EDO. OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL MILE ISLAND UNIT 2 SPENT FUEL AT THE IDAHO NATIONAL EN-DATA GINEERING AND ENVIRONMENTAL.... OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST DIRECTOR 870729) NUREG-0090 V20: REPORT TO CONGRESS ON ABNORMAL OC- NUREG-1556 V02: CONSOUDATED GUIDANCE ABOUT MATERIALS j ! CURRENCES. Fiscal Year 1997. UCENSES. Pro 0 ram-Specific Guidance About industrial Radiography . NUREG. RO1: EVENT REPORTING GUIDELINES 10 CFR 50.72 n

                                                                                      .56             SOUDATED GUIDANCE ABOUT MATERIALS NUREG-1187 V01: PERFORMANCE INDICATORS FOR OPERATING                  UCENSES. Applications For Sealed Source And Device Evaluation ERCIAL fdVCLEAR POWER REACTORS. Data Through Sep-            A ga        :           ATED GUIDANCE ABOUT MATERIALS NUREG-1272 V10 N01: OFFICE FOR ANALYSIS ANC EVALUATION                LICENSES. Program-Specific Guidance About Fixed Gauge U-OF OPERATIONAL DATA.1996 Annual Report.                             censes. Final Report.

69 i

70 NRC Originating Organization index (Staff Reports) NUREG-1556 VOS: CONSOLIDATED GUIDANCE ABOUT MATERIALS NUREG-0750 V45; NUCLEAR REGULATORY COMMISSION LICENSES. Program-Specife Guidance About Self-Shielded irradiator ISSUANCES.Opirnons And Decisions Of The Nuclear Regulatory Licenses. Final Report. Commission With Selected Orders. January-June 1997. NUREG-1556 V07 DR FC: CONSOLIDATED GUIDANCE ABOUT MA- NUREG-0750 V46101: INDEXES TO NUCLEAR REGULATORY COM-TERIALS LICENSES. PROGRAM-SPECIFIC GUIDANCE ABOUT MISSION ISSUANCES. July-September 1997. ACADEMIC.RESEARCH AND DEVELOPMENT AND OTHER L1- NUREG-0750 V46102; INDEXES TO NUCLEAR REGULATORY COM-CENSES OF LIMITED SCOPE. Draft Report For Comment MISSION ISSU ANCESJuly-December 1997, i NUREG-1556 V08: CONSOLIDATED GUIDANCE ABOUT MATERIALS NUREG 0750 V46 NO3: NUCLEAR REGULATORY COMMISSION I LICENSEES. Program-Specific Guidance Exempt Distributton Lk ISSUANCES FOR SEPTEMBER 1997. Pages 49-193. censes. Final Report. NUREG-0750 V46 N04: NUCLEAR REGULATORY COMMISSION , NUREG-1556 V09 DR FC: CONSOLIDATED GUIDANCE ABOUT MA- ISSUANCES FOR OCTOBER 1997. Pages 195-256.  ! TERIALS UCENSES. Program-Specific Guidance About Medcal Use NUREG-0750 V46 NOJ: NUCLEAR REGULATORY COMMISSION I Licer ses. Draft Report For Comment. ISSUANCES FOR NOVEMBER 1997. Pages 257-285. l NUREG-1556 V10 DR FC: CONSOLIDATED GUIDANCE ABOUT MA- NUREG4750 V46 N06: NUCLEAR REGULATORY COMMISSION j TERIALS LICENSES. Program Specife Guidance About Master Mate- ISSUANCES FOR DECEMBER 1997. Pages 287419. I nals Licenses. Draft Report For Comment. NUREG-0750 V47101: INDEXES TO NUCLEAR REGULATORY COM-NUREG-1556 V11 DR FC: CONSOLIDATED GUIDANCE ABOUT MA- MISSION ISSUANCESJanuary-March 1998. TERIALS LICENSES. Program-Specif'c Guidance About Specific Lk NUREG 0750 V47102: INDEXES TO NUCLEAR REGULATORY COM-eenses Of Broadscape. Draft Report For Comment. June 1998. NUREG-1631: SOURCE DISCONNECTS RESULTING FROM RADI- MISSION NUREG 0750ISSUANCES.JanuaryEAR V47 N01: NUGL REGULATORY COMMISSION { OGRAPHY DRIVE CABLE FAILURES. Final Report. ISSUANCES FOR JANUARY 1998. Pages 1 12. DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207) NUREG-0750 V47 NO2: NUCLEAR REGULATORY COMMISSION NUREG-1619: STANDARD REVIEW PLAN FOR PHYSICAL PROTEC- ISSUANCES FOR FEBRUARY 1998. Pages 13-56. TION PLANS FOR THE INDEPENDENT STORAGE OF SPENT NUREG 0750 V47 NO3: NUCLEAR REGULATORY COMMISSION i FUEL AND HIGH-LEVEL RADIOACTIVE WASTE. ISSUANCES FOR MARCH 1998.Pages 57-75. l OPERATIONS BRANCH NUREG-0750 V47 N04: NUCLEAR REGULATORY COMMISSION l NUREG-0525 V02 ROS: SAFEGUARDS

SUMMARY

EVENT LIST ISSUANCES FOR APRIL 1998.Pages 77-260. 1 (SSEL) January 1,i990 Through December 31,1997. NUREG-0750 V47 N05: NUCLEAR REGULATORY COMMISSION DIVl$10N OF WASTE MANAGEMENT (NMSS 940403) ISSUANCES FOR MAY 1998.Pages 261-306. NUREGfCP-0163: PROCEEDINGS OF THE WORKSHOP ON REVIEW NUREG-0750 V47 N06: NUCLEAR REGULATORY COMMIGSION OF DOSE MODELING METHODS FOR DEMONSTRATION OF ISSUANCES FOR JUNE 1998.Pages 307-408. COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR Li- NUREG-0750 V48101: INDEXES TO NUCLEAR REGULATORY COM-CENSE TERMINATION. MISSION ISSUANCESJuly -September 1998. NUREG-0750 V48 N01: NUCLEAR REGULATORY COMMISSION U.S. NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JULY 1998.Pages 1,38. OFFICE OF THE GENERAL COUNSEL (POST 860701) NUREG-0750 V48 NO3: NUCLEAR REGULATORY COMMISSION NUREG-0386 009: UNITED STATES NUCLEAR REGULATORY COM- ISSUANCES FOR SEPTEMBER 1998. Pages 119-182. MISSION STAFF PRACTICE AND PROCEDURE Di* NUREGC750 V48 N04: NUCLEAR REGULATORY COMMISSION GEST.Comrnisson, Appeal Board And Licensin9 Board Dech ISSUANCES FOR OCTOBER 1998. Pages 183-258. sons. July 1972 - June 1997. NUREG-0910 R03: NRC COMPREHENSIVE RECORDS DISPOSI-NUREG-0980 V01 N04: NUCLEAR REGULATORY LEGISLA- TION SCHEDULE. TION.104th Congress. NUREG-1575: MULTI-AGENCY RADIATION SURVEY AND SITE IN-NUREG-0980 V02 N04: NUCLEAR REGULATORY LEGISLA- VESTIGATION MANUAL (MARSSIM). Final Report. TION.104th Con 0ress. NUREG-1627 vot: PERFORMANCE PLAN FY 1999. OFFICE OF THE INSPECTOR GENERAL (POST 890417) NUREG-1415 V10 NO2: OFFICE OF THE INSPECTOR GEN- EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST ERALSemiannual Report To Congress,0ctober 1,1997 - March 31, 820406) 1998. DIVISION OF ENGINEERING TECHNOLOGY (POST 941217) NUREG-1415 V11 N01: OFFICE OF THE INSPECTOR GEN- NUREG 0933 S22: A PRIORITIZATION OF GENERIC SAFETY ERALSemiannual Report To ConDress, April-September 1998. ISSUES. NRC - NO DETAILED AFFILIATION GIVEN NUREG-1426 V03: COMPILATION OF REPORTS FROM RESEARCH NUREG-0304 V22 NO3: REGULATORY AND TECHNICAL REPORTS SUPPORTED BY THE ELECTPICAL MATERIALS AND MECHAN-(ABSTRACT INDEX JOURNAL) Compilation For Third Quarter ICAL ENGINEERING BRANCH, OlVISION OF ENGINEERING. 1997. July-September. NUREG-1629: THE CHARACTERl2ATION OF VICKER'S MICRO-NUREG-0304 V22 N04: REGULATORY AND TECHNICAL REPORTS HARDNESS INDENTATIONS AND PILE-UP PROFILES AS A (ABSTRACT INDEX JOURNAL). Annual Compilation For 1997. l STRAIN-HARDENING MICROPROBE.  ! NUREG-0304 V23 NOI: ABSTRACTS FOR PUBLICATIONS IN THE NUREGrCR-5570: APPLICATION OF THE NCSA HABANERO TOOL  ! NUREG-SERIES. Semiannual Compilation for January-June 1998. FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS- { NUREG 0540 V19 N11: TITLE LIST OF DOCUMENTS MADE PUB- MENTS. LICLY AVAILABLE.Nosember 1-30,1997. I NUREG4540 V19 N12: TITLE UST OF DOCUMENTS MADE PUB- DIVISION OF REGULATORY APPLICATIONS (POST 941217) LICLY AVAILABLE. December 1 31,1997. NUREG0713 V18: OCCUPATIONAL RADIATION EXPOSURE AT j COMMERCIAL NUCLEAR POWER REACTORS AND OTHER FA- ' NUREG-0540 V20 N01: TITLE LIST OF DOCUMENTS MADE PUS- CILITIES,1996. Twenty-Ninth Annual Report. j LICLY AVAILABLEJanuary 1-31,1998. NUREG 0713 V19; OCCUPATIONAL RADIATION EXPOSURE AT I NUREG-0540 V20 NO2: TITLE LIST OF DOCUMENTS MADE PUB-COMMERCIAL NUCLEAR POWER REACTORS AND OTHER FA-N E O N ITYE IST DOCUMENTS MADE PUB" CLT S 1997.T h R , NU N E N TL LIST OF DOCUMENTS MADE PUB- CHARGES.ChanDes in Decommissioning Waste Disposal Costs At E L w-Level Waste Burial Facilities. N E 20 N TT E LS OF DOCUMENTS MADE PUB- NUREG-1505 RO1: A NONPARAMETRIC STATISTICAL METHOD-LICLY AVAILABLE.May 1 31,1998 OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS NUREG0540 V20 N06: TITLE LIST' OF DOCUMENTS MADE PUB- DECOMMISSIONING SURVEYS. Inter 6m Report For Use And Corn-LICLY AVAILABLE. June 1-30, 1998. ment. NUREG-0540 V20 N07: TITLE LIST OF DOCUMENTS MADE PUB- NUREG-1507: MINIMUM DETECTABLE CONCENTRATIONS WITH LICLY AVAILABLEJuly 1-31,1998. TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS NUREG4540 V20 N08: TITLE LIST OF DOCUMENTS MADE PUB. CONTAMINANTS AND FIELD CONDITIONS. LICLY AVAILABLE. August 1-31, 1998. NUREG-1549 DRFT FC: DECISION METHODS FOR DOSE ASSESS-NUREG4540 V20 N09: TITLE UST OF DOCUMENTS MADE PUB- MENT TO COMPLY WITH RADIOLOGICAL CRITERIA FOR LL LICLY AVAILABLE. September 1-30,1998. CENSE TERMINATION. Draft Report For Comment. NUREG4540 V20 N10. TITLE LIST OF DOCUMENTS MADE PUB- NUREG/CP-0163: PROCEEDINGS OF THE WORKSHOP ON REVIEW LICLY AVAILABLE. October 1-31,1998. OF DOSE MODELING METHODS FOR DEMONSTRATION OF NUREG-0750 C104: INDEXES TO NUCLEAR REGULATORY COM- COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LL MISSION ISSUANCESJanuary 1,1991 through December 31,1995. CENSE TERMINATION.

NRC Originating Organization index (Staff Reports) 71 DIVISION OF SYSTEMS TECHNOLOGY (POST 941217) EDO . OFFICE OF NUCLEAR REACTOR REGULATION (POST 000428) NUREG 1521 DRFT FC: TECHNICAL REVIEW OF RISK-INFORMED. OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001) PERFORMANCE-BASED METHODS FOR NUCLEAR POWER NUREG0040 V21 N04:UCENSEE CONTRACTOR AND VENDOR IN-NUR G 1 V1 U ION 0- , 7 GRAM: PERSPECTIVES ON REACTOR SAFETY AND PLANT PER-NUREG4040 V22 N01: UCENSEE CONTRACTOR AND VENDOR IN-NUR 560  : VIDUAL PLANT EXAMINATION PRO. SPECTION STATUS REPORT, Quarterty Report. January-March PECTIVES ON REACTOR SAFETY AND PLANT PER-NUR V22 02: UCENSEE CONTRACTOR AND VENDOR IN. NUREG-1560 V03 P6: INDIVIDUAL PLANT EXAMINATION PRO, SPECTION STATUS REPORT. Quarterty Report.Aprlhlune GRAM: PERSPECTIVES ON REACTOR SAFETY AND PLANT PER- 1996.(White Book) FORMANCE.Appendloes. NUREG-1122 R02: KNOWLEDGE AND ABluT!ES CATALOG FOR 1 NUREGCR-5498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- NUCLEAR POWER PLANT OPERATORS. Pressurized Water Reac-CULATION TESTS IN 'iHE PUMA FACILITY, lt tors. NUREGOR-6500: THE EFFECT OF INITIAL TEMPERATURE ON NUREG-1123 R02: KNOWLEDGE AND ABluTIES CATALOG FOR FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATIC" l NUCLEAR POWER PLANT OPERATORS.Boill Water Reactors. 1 N E -6524: H F CT OF LATERAL VENTING ON DEFLA-GRATION-TO-DETON ATION TRANSITION IN HYDROGEN-AIR-DUC S GENE T TU E U RE STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. NUREG-1632: EVALUATION OF AP600 CONTAINMENT THERMAL-NUR GCR MA E TESTING OF PASSIVE C NU EG- 6 DR SSMENT OF THE USE OF POTAS-PROBABluSTIC RISK ANALYSIS BRANCH (POST 94121n SIUM IODIDE (KI) AS A PUBLIC PROTECTIVE ACTION DURING NUREG-1624 DRFT FC: TECHNICAL BASIS AND IMPLEMENTATION SEVERE REACTOR ACCIDENTS. Draft Report For Comment. GUIDELINES FOR A TECHNIQUE FOR HUMAN EVENT ANALYSIS NUREGCP-0152 VQ2: PROCEEDINGS OF THE FIFTH NRC/ASME (ATHEANA). Draft Report For Comment SYMPOSlVM ON VALVE AND PUMP TESTING. l

NRC Originating Organization index (Staff Reports) 71A ADVISORY COMMITTEE (s) ACNW- ADVISORY COMMITTEE ON NUCLEAR WASTE NUREG/BR4050: ADVISORY COMMITTEE ON NUCLEAR WASTE .1998 STRATEGIC PLAN AND PRIORlTY ISSUES AND ACTIVITIES. EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/BR-0117N97-4: NMSS LICENSEE NEWSLETTER. NUREG/BR4252; USER'S GUIDE TO NRC PUBLISHED PHYSICAL PROTECTION DOCUMENTS. U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF PUBLIC AFFAIRS NUREG/BR4164, REV.3: NRC REGULATOR OF NUCLEAR SAFETY. NUREG/BR4240; REPORTING SAFETY CONCERNS TO THE NRC. NUREG/BR4249: THE ATOMIC SAFETY AND LICENSING BOARD PANEL.

NRC Originating Organization Index (International Agreements) This index lists those NRC organizations that have published international agreement reports. Th3 index is arranged alphabetically by major NRC organizations (e.g., program offices) and th:n by subsections of these (e.g., d, ivisions, branches) where appropnate. Each entry is fol-lowed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number. EDO OFFICE OF NUCLEAR REOULATORY RESEARCH (POST DIVISION OF SYSTEMS TECHNOLOGY (POST 941217) OF LOF TE NUCLEAR REGULATORY e RESEA CH ' 'POST uS'" 941217g 5: ae'^rs'uoo'suecoo'co Bol'l"G Mooe' ^S-

     "Y"AffR?htSs*TTSr"'P J ae'^"                                   auaeg^=d6:

sS NUREG/lA-0140: DEVELOPMENT ASSEESMENT OF RELAPS/ NUREG/LA 01 IMPLEMENTATION AND ASSESSMENT OF IM-

                                      ^    NTEGRAL TEST EX-           PROVED MODELS AND OPTIONS IN TRAC-BF1.

i N MODEL A SAD T NU A0 7: SS ENT R LAPS / MOD F R STEAM NUR A 0141: RESULT OF BETHSY TEST 9.1.8 USING RELAPS/ NUREGliA4142: INSTALLATION OF RELAP5/ MOD 3.2 ON 80486 NONCONDENSIBLES IN A VERTICAL TUBE OF PCCS. AND PENTIUM BASED PERSONAL COMPUTERS. NUREG/lA4148: ASSESSMENT OF RELAP5/ MOD 3.1 USING LSTF NUREG/IA-0143: ASSESSMENT OF RELAP5 MOD 3.2 WITH THE TEN-PERCENT MAIN STEAM-UNE-BREAK TEST RUN SB SL41. LSTF EXPERIMENT SIMULATING A LOSS OF RESIDUAL HEAT NUREG/lA4149: ASSESSMENT OF RELAP5/ MOO 3.2.NPA3.4 REMOVAL EVENT DURING MID-LOOP OPERATION. AGAINST A TRANSIENT OF HIGH NUCLEAR FLUX VARLATION NUREG/lA-0144: ASSESSMENT OF RELAP5/ MOD 3.2 WITH THE REACTOR TRIP. NATURAL CIRCULATION AND THE START OF A SEMISCALE NATURAL CIRCUt.ATION EXPERIMENT S-NC-88 MAIN PUMP IN THE VANDELLOS 11 NUCLEAR POWER PLANT. NUREG/lA-0145: RELAPS ASSESSMENT AGAINST PAdTEL EXPERI- NUREG/lA-0150: STUDY OF TRANSIENTS RELATED TO AMSAC MENTAL DATA (REVISION 1). ACTUATION. SENSITIVITY ANALYSIS. l I l l 73 l t

I l i l l 1 l l

r NRC Contract Sponsor index (Contractor Reports) This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabeticall cnd then by subsections of these (e.g., y bywhere divisions) majorappropriate. NRC organizationThe sponsor (e.g., program offi 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 FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR-5562: DATING AND EARTHOUAKES: REVIEW OF QUA-DATA TERNARY GEOCHRONOLOGY AND ITS APPUCATON TO DIVISION OF SAFETY PROGRAMS (POST 870413) PALEOSEISMOLOGY. NUREGCR-4674 V25: PRECURSORS TO POTENTIAL SEVERE NUREG/CR 5570: APPLICATION OF THE NCSA HABANERO TOOL CORE DAMAGE ACCIDENTS: 1996. A Status Report. l FOR COLLABORATON ON STRUCTURAL INTEGRITY ASSESS- l NUREG/CR-4674 V26: PRECURSORS TO POTENTIAL SEVERE MENTS. CORE DAMAGE ACCIDENTS:1997. A Status Report. NUREG/CR-5591 V04 N1: HEAVY-SECTON STEEL IRRADIATON NUREGCR-5485: GUIDEUNES ON MODELING COMMON-CAUSE PROGRAM. Semiannual Progress Report For October 1992 Through FAILURES IN PROBABILISTIC RISK ASSESSMENT. March 1993. NUREG/CR 5496: EVALUATION OF LOSS OF OFFSITE POWER NUREGCR-5591 V08 NI: HEAVY-SECTION STEEL IHRADIATON EVENTS AT NUCLEAR POWER PLANTS: 1960 -1996. PROGRAM.Sermannual Progress Report For October 1996 Through NUREGCH-5497: COMMON CAUSE FAILURE PARAMETER ESTl- March 1997. MATONS. NUREG/CR-5671: PRETEST PREDICTON ANALYSIS AND NUREG/CR-5500 Voi: REUABILITY STUDY: AUXIUARY/EMER- POSTTEST CORRELATION OF THE SIZEWELL-B 1:10 SCALE GENCY FEEDWATER SYSTEM, 1987-1995. i PRESTRESSED CONCRETE CONTAINMENT MODEL TEST. l NUREG/CR4268 V01: COMMON CAUSE FAILURE DATABASE AND NUREG/CR-6274: PALEOSEISMIC STUDIES IN THE SOUTH- ' ANALYSIS SYSTEM.Volurne 1: Overview. EASTERN UNITED STATES AND NEW ENGLAND. NUREG/CR4268 V02: COMMON CAUSE FAILURE DATABASE AND NUREG/CR4412: AGING AND LOSS 4F-COOLANT ACCIDENT  ; ANALYSIS SYSTEM.Volurne 2: Event Dehniton And Classification. (LOCA) TESTING OF ELECTRICAL CONNECTIONS. ' NUREG/CR4268 V03: COMMON CAUSE FAILURE DATABASE AND NUREG/CR-6447: RESULTS OF CRACK-ARREST TESTS ON IRRA-ANALYSIS SYSTEM.Volorne 3: Data Collection And Event Coding. DIATED A 508 CLASS 3 STEEL NUREG/CR4268 V04: COMMON CAUSE FAILURE DATABASE AND NUREG/CR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH-ANALYSIS SYSTEM. Volume 4: Software Reference Manual. MARK. NUREG/CR4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE-EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEQUARDS ACTOR PRESSURE VESSELS. Density And Distnbution Of Flaw In-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS ecations in PVRUF. NUREGCR-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS- NUREGCR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO. TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR GRAM. Annual Report, August 1995 - September 1996. SHlPPING CASK DESIGN REVIEW User's Manual to Version 3a. NUREG/CR4511 V03: STEAM GENERATOR TUBE INTEGRITY PRO-NUREGCR4342: ASSESSMENT AND RECOMMENDATONS FOR GRAM. Semiannual Report, October 1996 - March 1997. FISSILE-MATERIAL PACKAGING EXEMPTONS AND GENERAL U- NUREGCR4517: ROUND ROBIN PRETEST ANALYSES OF A CENSES WITHIN 10CFR PART 71. STEEL CONTAINMENT VESSEL MODEL AND CONTACT STRUC-NUREGCR 5502: ENGINEERING DRAWINGS FOR 10 CFR PART 71 TURE ASSEMBLY SUBJECT TO STATIC INTERNAL PRESSURIZA-PACKAGE APPROVALS. TON. NUREGCR 5549: ENVIRONMENTAL ASSESSMENT RENEWAL OF NUREGCR-6521: ESTIMATING PROBABLE FLAW DISTRIBUTIONS MATERIAL UCENSES FOR ALARON CORP. NORTHEAST RE- IN PWR STEAM GENERATORS. GONAL SERVICE FACILITY, WAMPUM. PENNSYLVANIA. NUREG/CR4537: INFLUENCE OF LONG-TERM THERMAL AGING NUREGCR4608:

SUMMARY

AND EVALUATION OF LOW-VELOC- ON THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REAC-ITY IMPACT TEST OF SOUD STEEL BILLET OUTO CONCRETE TOR PRESSURE VESSEL MATERIALS.An Atom Probe Study. PADS. NUREGCR4540: STATE.OF-THE-ART REPORT ON PIPING FRAC-DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207) TURE MECHANICS. NUREGCR4410: NUCLEAR FUEL CYCLE FACILITY ACCIDENT NUREGCR-6546: A DAMAGE MECHANICS BASED APPROACH TO ANALYSIS HANDBOOK. STRUCTURAL DETERORATION AND REUABILITY. DIVISION OF WASTE MANAGEMENT (NMSS 940403) NUREGCR-6551: IMPROVED EMBRITTLEMENT CORRELATIONS NUREGCR4377: EFFECTS ON RADONUCLIDE CONCENTRA- FOR REACTOR PRESSURE VESSEL STEELS. TONS BY CEMENT / GROUND-WATER INTERACTIONS IN SUP. NUREGCR-6552: MARBLE HILL ANNEAUNG DEMONSTRATION PORT OF PERFORMANCE ASSESSMENT OF LOW-LEVEL RA. EVALUATON. DIOACTIVE WASTE DISPOSAL FACILITIES. NUREGCR4554: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR WALL INTERNATIONAL STANDARD PROBLEM. EDO . OFFICE OF NUCLEAR REGULATORY RESEARCH (POST NUREGCR-6556: A COMPREHENSIVE STUDY OF THE EASTERN 820405) TENNESSEE SEISMIC ZONE. DIVISION OF ENGINEERING TECHNOLOGY (POST 941217) NUREGCR4559: LARGE-SCALE VIBRATION TESTS OF MAIN NUREGCR-4219 V13 N2: HEAVY-SECTION STEEL TECHNOLOGY STEAM AND FEEDWATER PIPING SYSTEMS WITH CONVEN-PROGRAM.Semaannual Progress Report For April- Seplember 1996. TIONAL AND ENERGY-ABSORBING SUPPORTS. NUREGOR-4219 V14 N1: HEAVY SECTON STEEL TECHNOLOGY NUREGCR4564: ANALYSES OF SOURCE SPECTRA. ATTENU-PROGRAM. Semiannual Progress Report For October 1996 - March ATION, AND SITE EFFECTS FROM CENTRAL AND EASTERN 1997. UNITED STATES EARTHQUAKES. NUREGCR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING NUREGCR4573: " INVESTIGATING SEISMOTECTONICS IN THE IN UGHT-WATER REACTORS. Semiannual Report. January-June EASTERN UNITED STATES USING A GEOGRAPHIC INFORMA-1997. TON SYSTEM." NUREGCR 5361: SEISMIC ANALYSIS OF PIPING. Final Program Re. NUREGCR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR-port. IZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING NUREG/CR 5434: ANCHOR BOLT BEHAVIOR AND STRENGTH AT HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON DURING EARTHOUAKES. WITH MODEL PREDICTONS. NUREGCR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE NUREGCR-6583: EFFECTS OF LWR COOLANT ENVIRONMENTS PROBADiLITIES OF DEFECTS IN REACTOR PRESSURE VESSEL ON FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY WELDS. STEELS. 75

76 NRC Contract Sponsor index NUREG/CR4589: THE EFFECTS OF SURFACE CONDITION ON AN NUREG/CR4359 V02: RAMONA-4B: A COMPUTER CODE WITH ULTRASONIC INSPECTION: ENGINEERING STUDIES USING THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND VAllDATED COMPUTER MODEL. SBWR SYSTEM TRANSIENTS. Users Manual. NUREG/CR4593; CRUSTAL STRUCTURE AND GROUND MOTION NUREG/CR4418: RISK IMPORTANCE OF CONTAINMENT AND RE. MODELS IN THE EASTERN AND CENTRAL UNITEDSTATES LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. FROM NATIONAL SEISMOGRAPHIC NETWORK DATA. NUREG/CR4472: PRELIM NARY PHENOMENA IDENTIFICATION NUREG/CR4594: EVALUATICN OF ULTRASONIC INSPECTION AND RANKING TABLES FOR SIMPLIFIED BOILING WATER REAC-TECHNIQUES FOR COARSE GRAINED MATERIALS. TOR LOSS-OF-COOLANT ACCIDENT SCENARIOS. NUREG/CR4598: AN INVESTIGATION OF TENDON SHEATHING NUREG/CR4475: RESOLUTION OF THE DIRECT CONTAINMENT NU 00: N UT N E POS E PARAMETERS FOR CAP. B K WI C X PLANTS SULE 10.05 IN THE HEAVY-SECTION STEEL IRRADIATION PRO- NUREG/CR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAL-IFICATION OF MICROPROCESSOR-BASED SAFETY-RELATED NURE CR4601: NEU E l RE PARAMETERS FOR THE AR POW P NT . EQU MEN NU DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRA- g DIATION PROGRAM TENTH IRRADIATION SERIES. FAILURES.An Analysis For Four Nuclear Power Plants. NUREG/CR4605: AN EVALUATION OF HUMAN FACTORS RE- NUREG/CR-6509: THE EFFECT OF INITIAL TEMPERATURE ON SEARCH FOR ULTRASONIC INSERVICE INSPECTION. NUREG/CR4606: INVESTIGATION OF TECHNIQUES FOR THE DE. FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION VELOPMENT OF SEISMIC DESIGN BASIS USING THE PROB- TRANSITION PHENOMENON. ABILISTIC SEISMIC HAZARD ANALYS!S. NUREG/CR4524: THE EFFECT OF LATERAL VENTING ON DEFLA-NUREG/CR-6611: RESULTS OF PRESSURE LOCKING AND THER. GRATION-TO-DETON ATION TRANSITION IN HYDROGEN-AIR. MAL BINDING TESTS OF GATE VALVES. STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. NUREG/CR-6614: FEASIBILITY OF HIGH FREQUENCY ACOUSTIC NUREG/CR-6534 V02: FRAPCON-3: A COMPUTER CODE FOR THE IMAGING FOR INSPECTION OF CONTAINMENTS. CALCULATION OF STEADY-STATE, THERMAL-MECHANICAL BE-NUREG/CR-6615: A SURVEY OF REPAIR PRACTICES FOR NU- HAVIOR OF OXIDE FUEL RODS FOR HIGH BURNUP. l CLEAR POWER PLANT CONTAINMENT METALLIC PRESSURE NUREG/CR4534 V03: FRAPCON-3: INTEGRAL ASSESSMENT. 1 BOUNDARIES. NUREG/CR4544: METHODOLOGY FOR ANALYZING PRECURSORS TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT DIVISION NUREG/CR-5559:OF REGULATORY APPLICATIONS SINGLE- AND CROSS-HOLE PNEUMATI (POST 941217) SEQUENCES. C TESTS IN UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP NUREG/CR-6545 V01: PROBABILISTIC ACCIDENT CONSEQUENCE RESEARCH SITE: PHENOMENOLOGY, SPATIAL VARIA- UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assess-BILI V CONNECTIVITY AND SCAL E. rnent. Main Report NUREG/6R-5621: GROUND-WATER MODELS IN SUPPORT OF NUREG/CR4545 V'02: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assess-NURE HUMAN PERFORMANCE IN RADIOLOGICAL SUR- mentAppedes. VEY SCANNING" VERIFICATION OF THE LWRARC CODENUREG/CR4555 NUREG/CR4536: FOR V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Late Hearth Effects Uncertainty Assess-LIGHT WATER-REACTOR AFTERHEAT RATE CALCULATIONS. ment. Main Report. NUREG/CR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM. NUREG/CR4555 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4603: CHARACTERIZATION OF RETARDATION MECHA- UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess-NISMS IN SOIL. ment. ndaces. OlVISION OF SYSTEMS TECHNOLOGY (POST 941217) NURE R4571 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-5372: EXPERIMENTS ON INTERACTIONS BETWEEN UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do-ZlRCONIUM CONTAINING MELT AND WATER. simetry. Main Report. NUREG/CR-5498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR- NUREG/CR4571 V02: PROBABILISTIC ACCIDENT CONSEQUENCE

                                                                                               ^          " "       ^       '""'     '"'"'

NURE 1 ANCED NS E TATION AND MAINTE- i ry c 657E DIGITAL l&C SYSTEMS IN NUCLEAR POWER NURE -5 CR 1 L A FLUX CF EN ON ON S sbScreeg O Monmental hessors And A Conpan. A DOWNWARD FACING CURVED SURFACE: EFFECTS OF THER- son Of Hardware Unavailabihty With An Existing Analog Systers MAL INSUt.ATION NUREG/CR4580: PERFORMANCE TESTING OF PASSIVE NUREG/CR4119 V01 R1: MELCOR COMPUTER CODE MANU-NU 11 R L k PUT. ODE MANU- NUR G/ R 1 0:C E MANU'AL FOR MACCS2. User's Guide. NUREG/CR-6613 V02: CODE MANUAL FOR MACCS2. Preprocessor ALS. Reference Manuals. Version 1.8.4,Ju!y 1997. Codes COMIDA2, FGROCF, IDCF2. NUREG/CR-6131: VICTORIA 2.0: A MECHANISTIC MODEL FOR RADIONUCLIDE BEHAVIOR IN A NUCLEAR REACTOR COOLANT NUREG/CR4616: RISK COMPARISON OF SCHEDULING PREVEN-SYSTEM UNDER SEVERE ACCIDENT CONDITIONS. TlVE MAINTENANCE DURING SHUTDOWN VS. DURING POWER NUREG/CR4150 V01 R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN- OPERATION FOR PWRS. UAL. interface Theory. NUREG/CR4150 V02 R1: SCDAP/RELAPS/ MOD 3.2 CODE MAN- EDO. OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428) UAL.Damaos Progression Model Theory. OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001) NUREG'CR4150 v03 R1: SCDAP/RELAPS/ MOD 3.2 CODE MAN. NUREG/CR4408: TECHNICAL ASSISTANCE IN REVIEW OF UAL . Users Guide And input Manual. SOURCE TERM-RELATED ISSUES OF ADVANCED REACTORS. @ NUREG/CR-6150 V04 R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN- NUREG/CR4577: U.S. NUCLEAR POWER PLANT OPERATING UAL.MATPRO - A Library Of Matenals Properties For Light-Water COST AND EXPERIENCE SUMMARIES. Reactor Accident Analysis. NUREG/CR4599: LODINE VOLATILITY AND PH CONTROL IN THE NUREG/CR4150 VOS R1: SCDAP!RELAPS/ MOD 3.2 CODE MAN- AP-600 REACTOR. UAL.Devekipmental Assessment. NUREG/CR-6604: RADTRAD: A SIMPLIFIED MODEL FOR RADIO-NUREG/CR4359 V01: RAMONA-4B: A COMPUTER CODE WITH NUCLIDE TRANSPORT AND REMOVAL AND DOSE ESTIMATION. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND NUREG/CR-6617: THE PRICE-ANDERSON ACT . CROSSING THE SBWR SYSTEM TRANSIENTS.Models And Correlations. BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS.

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. ADVANCED SYSTEMS CONCEPTS ASSOCIATES NUREGCR4589: THE EFFECTS OF SURFACE CONDITION ON AN NUREG/CR4544: METHODOLOGY FOR ANALYZING PRECURSORS ULTRASONIC INSPECTION: ENGINEERING STUDIES USING VAU-TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE- DATED COMPUTER MODEL. QUENCES. NUREG/CR-6594: EVALUATION OF ULTRASONIC INSPECTION TECH-NIOUES FOR COARSE-GRAINED MATERIALS. AMERICAN SOCIETY OF MECHANICAL ENGINEERS NUREG/CR4005: AN EVALUATION OF HUMAN FACTORS RESEARCH NUREG/CP-0152 V02: PROCEEDINGS OF THE F!FTH NRC/ASME FOR ULTRASONIC INSERVICE INSPECTION. SYMPOSIUM ON VALVE AND PUMP TESTING. BOSTON COLLEGE, WESTON, M A ANALYSIS & MEASUREMENT SERVICES CORP. NUREG/CR4573: "lNVESTIGATING SElSMOTECTONICS IN THE NUREG/CR 5501: ADVANCED INSTRUMENTATION AND MAINTE. EASTERN UNITED STATES USING A GEOGRAPHIC INFORMATION NANCE TECHNOLOGIES FOR NUCLEAR POWER PLANTS. SYSTEM." ATATECH RESEARCH CORP. BROOKHAVEN NATIONAL LABORATORY NUREG/CR 5671: PRETEST PREDICTION ANALYSIS AND POSTTEST NUREG-1507: MINIMUM DETECTABLE CONCENTRATIONS WITH CORRELATION OF THE SIZEWELL-B 1:10 SCALE PRESTRESSED TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- l CONCRETE CONTAINMENT MODEL TEST. j[A S AN FgDg Ig ARGONNE NATIONAL LABORATORY PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT ) NUREG/CR-4667 V24: ENVIRONMENTALLY ASSISTED CRACKING IN FIRE PROTECTION ANALYSES. Draft Report For Cor9 ment. J LIGHT-WATER REACTORS. Semiannual Report. January-June 1997* NUREG/CP-0162 V01: PROCEEDINGS OF THE TWENTY-FIFTH I NUREG/CR-4667 V25: ENVIRONMENTALLY ASSISTED CRACKING lN WATER REACTOR SAFETY INFORMATION MEETING. Plenary See LIGHT-WATER REACTORS. Semiannual Report. July-Decernber 1997, sions. Pressure Vessel Research,BWR Strainer Blockage And Other NUREG/CR-5372: EXPERIMENTS ON INTERACTIONS BETWEEN ZlR. Genene Safety issues.Environmenta'ly Assisted Dearadate Of LWR.... CONIUM-CONTAINING MELT AND WATER. NUREG/CP 0162 V02: PROCEEDINGS OF THE TWENTY-FIFTH NUREG/CR4511 V02: STEAM GENERATOR TUBE INTEGRITY PRO- WATER REACTOR SAFETY INFORMATION MEETING. Human Rele GRAM. Annual Report. August 1995 - September 1996. ability Analysis And Human Performance Evaluation, Technscal lasues NUREGCR4511 V03: STEAM GENERATOR TUBE INTEGRITY PRO- Related To Rulemakings, Risk-informed, Performance-Based initia-GRAM Semiannual Report. October 1996 - March 1997. tives... i NUREC/CR4521: ESTIMATING PROBABLE FLAW DISTRIBUTIONS IN NUREG/CP 0162 V03: PROCEEDINGS OF THE TWENTY-FIFTH i PWR STEAM GENERATORS. WATER REACTOR SAFETY INFORMATION MEETING. Thermal-Hy-NUREGCR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT drauhc Research And Codes, Digital Instrumentation And Control, PROGRAM. Structural Performance. NUREG/CR4575: FAILURE BEHAVIOR OF INTERNALLY PRESSUR- NUREG/CP 0165: 1RANSACTIONS OF THE TWENTY-SIXTH WATER lZED FLAWED AND UNFLAWED STEAM GENERATOR TUBING AT REACTOR SAFETY INFORMATION MEETING. HIGH TEMPERATURE -EXPERIMENTS AND COMPARISON WITH NUREG/CR4359 V01: RAMONA-4B: A COMPUTER CODE WITH MODEL PREDICTIONS. THREE-DIMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR NUREG/CR4583 EFFECTS OF LWR COOLANT ENVIRONMENTS ON SYSTEM TRANSIENTS.Models And Correlations. FATIGUE DESIGN CURVES OF CARBON AND LOW-ALLOY NUREG/CR4359 V02: RAMONA-48: A COMPUTER CODE WITH STEELS. THREE-OlMENSIONAL NEUTRON KINETICS FOR BWR AND SBWR SYSTEM TRANSIENTS. User's Manual. i ARIZON A, UNIV. OF, TUCSON, AZ NUREG/CR4364: HUMAN PERFORMANCE IN RADIOLOGICAL SUR-NUREGCR-5559: SINGLE- AND CROSS-HOLE PNEUMATIC TESTS IN UNSATURATED FRACTURED TUFFS AT THE APACHE LEAP RE-N EC 1 RISK IMPORTANCE OF CONTAINMENT AND RE-PHENOMENOLOGY, SPATIAL LATED ESF SYSTEM PERFORMANCE REQUIREMENTS. SEARCH SITE: VARIA' NUREG/CR4472: PRELIMINARY PHENOMENA IDENTIFICATION AND BILITY,CONNECTIVITY AND SCALE. RANKING TABLES FOR SIMPLIFIED BOILING WATER REACTOR BATTELLE MEMORIAL INSTITUTE, COLUM8US LABORATORIES N G/C 44 i L AS IRONMENTAL QUAll-NUREGCR4540 STATE OF-THE-ART REPORT ON PIPING FRAC- FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED TURE MECHANICS. EQUIPMENT IN NUCLEAR POWER PLANTS. B ATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NATK)NAL NUREG/CR4502: ACTION REQUIREMENTS FOR AFW SYSTEM FAIL-URES.An Ana sis For Four Nuclear Power Plants. LABORATORY NUREG/CR4 : THE EFFECT OF INITIAL TEMPERATURE ON NUREG/CR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE FLAME ACCELERATION AND DEFLAGRATION-TO DETONATION PROBABILITIES OF DEFECTS IN REACTOR PRESSURE VESSEL TRANSITION PHENOMENON. WELDS. NUREG/CR4524: THE EFFECT OF LATERAL VENTING ON DEFLA-NUREG/CR-5621: GROUND-WATER MODELS IN SUPPORT OF GRATION-TO-DETONAT ION TRANSITION IN HYDROGEN-AIR-NUREG/CR-5512. STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. j NUREG/CR4210 S01: COMPUTER CODES FOR EVALUATION OF NUREGCR4554: FINITE ELEMENT ANALYSES FOR SEISMIC SHEAR j CONTROL ROOM HABITABILITY (HABIT V1.1). WALL INTERNATIONAL STANDARD PROBLEM. i NUREGCR4377: EFFECTS ON RADIONUCLIDE CONCENTRATIONS NUREG/CR4559: LARGE-SCALE VIBRATION TESTS OF MAIN STEAM l BY CEMENT / GROUND-WATER INTERACTIONS IN SUPPORT OF AND FEEDWATER PIPING SYSTEMS WITH CONVENTIONAL AND i PERFORMANCE ASSESSMENT OF LOW-LEVEL RADIOACTIVE ENERGY-ABSORBING SUPPORTS. I WASTE DISPOSAL FACILITIES. NUREG/CR-6569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT NUREGCR4471 V01: CHARACTERIZATION OF FLAWS IN U.S. RE- PROGRAM. ACTOR PRESSURE VESSELS. Density And Distributon Of Flaw Indi- NUREG/CR4579: DIGITAL I&C SYSTEMS IN NUCLEAR POWER cations in PVRUF. PLANTS Risk-Screening Of Environmental Stressors And A Compari-NUREGCR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE son Of Hardware Unavailability With An Existing Analog System. CALCULATION OF STEADY-STATE, THERHAL-MECHANICAL BE- NUREGCR4616: RISK COMPARISON OF SCHEDULING PREVENTIVE HAVIOR OF OXIDE FUEL RODS FOR HIGH BURNUP. MAINTENANCE DURING SHUTDOWN VS. DURING POWER OPER-NUREG/CR4534 V03: FRAPCON 3: INTEGRAL ASSESSMENT. ATION FOR PWRS. 77

r 78 Contractor index CALIFORNIA, UNIV. OF, SANTA B AR8 ARA, CA NUREG/CR4555 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG 1629: THE CHARACTERIZATION OF VICKER'S MICROHARD- UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess-NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN- ment.A s. HARDENING MICROPROBE. NURE $71 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4551: IMPROVED EMBRITTLEMENT CORRELATIONS FOR UNCERTAINTY ANALYSIS. Uncertainty Assessment For intomal Do-R simetr' . Main Report. NURE PRESSURE 4564: ANALYSES OF VESSEL SOUR STFE . SPECTRA. ATTENUATION. NURE R4571 V02: PROBABluSTIC ACCIDENT CONSEQUENCE , AND SITE EFFECTS FROM CENTRAL AND EASTERN UNITED UNCERTAINTY ANALYSIS. Uncertainty Assessment For Intemal Do-l STATES EARTHOUAKES. simetry.Apperdcas. COLORADO, UNIV. OF ICF MC NUREG/CR4603: CHARACTERIZATION OF RETARDATION MECHA- NUREG/CR4617: THE PRICE-ANDERSON ACT - CROSSING THE

  ' NISMS IN SOIL                                                                BRIDGE TO THE NEXT CENTURY: A REPORT TO CONGRESS.

DE IDAHO NATIONAL ENGMEERING & ENVIRONMENTAL LABORATORY NU E 1 5 TI-AGENCY RADIATION SURVEY AND SITE INVES-NUREG/CR-5485: GUIDEUNES ON MODEUNG COMMON CAUSE TIGATION MANUAL (MARSSIM). Final Report ~ FAILURES IN PROBABILISTIC RISK ASSESSMENT. DELFT UNIVERSITY OF TECHNOLOGY NUREG/CR-5496: EVALUATION OF LOSS OF OFFSITE POWER NUREG/CR4545 V01: PROBABluSTIC ACCIDENT CONSEQUENCE EVENTS AT NUCLEAR POWER PLANTS: 1980 - 1996. UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assess. NUREG/CR-5497: COMMON CAUSE FAILURE PARAMETER ESTI-NU 4 02: PROBABILISTIC ACCIDENT CONSEQUENCE NURE 5500 VO1: REUABluTY STUDY: AUXIUARY/ EMERGENCY UNCERTAINTY ANALYSIS. Earty Health Ef1ects Uncertainty Assess. F TER SY TEM 198 1 g NUYE 901: PROBABluSTIC ACCIDENT CONSEQUENCE UALinterface T  ; UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- NUREG/CR4150 V R1: SCDAP/RELAP5/ MOD 3.2 CODE MAN-NU R45 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NLRE 50 [R AP L'AP5/ MOD 32 CODE MAN-UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- ' NtRE 6150 V04 0 /RELAPS/ MOD 3.2 CODE MAN-NU 901: PROBABluSTIC ACCIDENT CONSEQUENCE UAL.MATPRO - A Ubrary Of Matenals Properties For Ught-Water Re-UNCER NTY ANALYSIS. Uncertainty Assessment For Intemal Do. gar A AgsisR1: SCDAP/RELAPS/ MOD 3.2 CODE MAN-NU R4571 V02: PROBABluSTIC ACCIDENT CONSEQUENCE UAL Assessment. UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do. NUREG/CR 68 VO1: COMMON CAUSE FAILURE DATABASE AND simetry. Appendices ANALYSIS SYSTEM. Volume 1: Overview. NUREGCR4268 V02: COMMON CAUSE FAILURE DATABASE AND DOMW60N ENOMEERWG, MC. ANALYSIS SYSTEM. Volume 2: Event Definition And Classification. NUREG/CR4521: ESTIMATING PROBABLE FLAW DISTRIBUTIONS IN NUREG/CR4268 V03: COMMON CAUSE FAILURE DATABASE AND PWR STEAM GENERATORS. ANALYSIS SYSTEM. Volume 3: Data Collection And Event Codire NUREG/CR4260 V04: COMMON CAUSE FAILURE DATABASE AND ENERGY DEPT.OF ANALYSIS SYSTEM. Volume 4: Software Reference Manual. NUREd-1505 RO1: A NONPARAMETRIC STATISTICAL METHOD- NUREG/CR4475: RESOLUTION OF THE DIRECT CONTAINMENT OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE- HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & COMMISSIONING SURVEYS. interim Report For Use And ment. BABCOCK & WILCOX PLANTS. NUREG-1575: MULTI-AGENCY RADIATION SURVEY AND 1 E INVES- NUREG/CR4534 V02: FRAPCON-3: A COMPUTER CODE FOR THE TIGATION MANUAL (MARSSIM). Final Report. CALCULATION OF STEADY-STATE, THERMAL-MECHANICAL BE-N  : UL D A ION SURVEY AND SITE INVES- A NUREGCR4' 611: RESULTS OF PRESSURE LOCKING AND THERMAL NL 163 S E X$bE.An Exercise Of Radio- BINDING TESTS OF GATE VALVES. logical Response Through Cooperation And Coordination Of j Local. State And rederal Resources Under The National Contingency IDAHO STATE UNfV., POCATELLO,lD I Plan. NUREGCR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM. EQE MTERNATIONAL NUREGCR4544: METHODOLOGY FOR ANALYZING PRECURSORS ILLMOIS UNIV.OF URBANA IL TO EARTHOUAKE-INITIATED AND FIRE-INlTIATED ACCIDENT SE* NUREG/CR-5570  ! APPUCATION OF THE NCSA HABANERO TOOL QUENCES. FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS-EUROPEAN COMMUNITIES, COMMISS60N OF MENTS. l I NUREGCR4571 V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assessment For Intemal Do-U 1 THE C CTER Tl OF VICKER'S MICROHARD-NUR G 71 2: PROBABluSTIC ACCIDENT CONSEQUENCE NESS INDENTATIONS AND PILE-UP PROFILES AS A STRAIN-UNCERTAINTY ANALYSIS. Uncertainty Assessment For Internat Do. HARDENING MICROPROBE. simetry. Appendices

• JOHNS HOPKWS UNIV., BALTIMORE, MD FINLAND, OOVT. OF NUREGCR4546: A DAMAGE MECHANICS BASED APPROACH TO NUREGCR4502: ACTION REQUIREMENTS FOR AFW SYSTEM FAIL. STRUCTURAL DETERIORATION AND REUABluTY.

URES.An Analysis For Four Nuclear Power Plants. LAWRENCE LfVERMORE NATIONAL LABORATORY FRANCE NUREGCR-4554 V01 R2: SCANS (SHIPPING CASK ANALYSIS SYS-NUREGCR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON TEM) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR SHIP. THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR PING CASK DESIGN REVIEW. User's Manual to Version Sa. PRESSURE VESSEL MATERIALS.An Atom Probe Study. NUREGCR4502: ENGINEERING DRAWINGS FOR 10 CFR PART 71 PACKAGE APPROVALS. FUTURE RESOURCES ASSOCIATES,INC. NUREGCR4606: INVESTIGATION OF TECHNIQUES FOR THE DE-NUREGOR4544: METHODOLOGY FOR ANALYZING PRECURSORS VELOPMENT OF SEISMIC DESIGN BASIS USING THE PROB-TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE- ABluSTIC SEISMIC HAZARD ANALYSIS. QUENCES. NUREGCR4608:

SUMMARY

AND EVALUATION OF LOW-VELOCITY HAWAll, UNIV. OF, HILO, HI NUREGCR4555 V01: PROBABluSTIC ACCIDENT CONSEQUENCE MARYLAND, UNIV. OF, COLLEGE PARK, MD UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- NUREGOR 5485: GUIDELINES ON MODEUNG COMMON-CAUSE ment. Main Report. FAILURES IN PROBABILISTIC RISK ASSESSMENT.

Contractor index 79 NUREG/CR-5497: COMMON CAUSE FAILURE PARAMETER EST6 NUREG/CR 5342: ASSESSMENT AND RECOMMENDATONS FOR MATICNS. FISSILE-MATERIAL PACKAGING EXEMPTIONS AND GENERAL Ll-NUREG/CR4268 V01: COMMON CAUSE FAILURE DATABASE AND CENSES WITHIN 10CFR PART 71. ANALYSIS SYSTEM. Volume 1: Overview. NUREG/CR4549: ENVIRONMENTAL ASSESSMENT RENEWAL OF NUREG/CR4268 V02: COMMON CAUSE FAILURE DATABASE AND MATERIAL LICENSES FOR ALARON CORP. NORTHEAST RE-ANALYSIS SYSTEM. Volume 2: Evert Definition And Classificate. GONAL SERVICE FACILITY WAMPUM, PENNSYLVANIA NUREG/CR4268 V03: COMMON CAUSE FAILURE DATABASE AND NUREG/CR-5570: APPLICATI N OF THE NCSA HABANERO TOOL ANALYSIS SYSTEM. Volume 3: Data Collectm And Event Codna. FOR COLLABORATION ON STRUCTURAL INTEGRITY ASSESS-NUREG/CR4268 V04: COMMON CAUSE FAILURE DATABASE AND MENTS. ANALYSIS SYSTEM. Volume 4: Software Reference Manual. NUREG/CR-5591 V04 N1: HEAVY-SECTION STEEL IRRADIATION NUREG/GR-0017: DATING OF LIQUEFACTION IN THE NEW MADRID PROGRAM. Semiannual Progress Report For October 1992 Through SEISMIC ZONE AND IMPLICATIONS FOR EARTHQUAKE HAZARD. March 1993. MA&&ACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE. MA NUREG/CR4591 V08 N1: HEAVY-SECTION STEEL IRRADIATION PROGRAM. Semiannual Pro 9ress Report For October 1996 Through NUREG/CR4544: METHODOLOGY FOR ANALYZING PRECURSORS March 1997. TO EARTHOUAKE-INITIATED AND FIRE-INITIATED ACCIDENT SE- NUREG/CR4119 V01 R1: MELCOR COMPUTER CODE MANU-QUENCES. 1997. ALS. Primer And NUREG/CR4119 Users' V02 Guidos. Version R1: MELCOR COMPUT 1.8.4, July CODEERMANU-MICHIGAN, UNIV. OF, ANN ARSOR, M1 ALS. Reference Manuals. Version 1.8.4 July 1997. NUREG/GR-0016: THE ROLE OF TIME-DEPENDENT DEFORMATION NUREG/CR4408: TECHNICAL ASSISTANCE IN REVIEW OF SOURCE IN INTERGRANULAR CRACK INITIATION OF ALLOY 600 STEAM TERM-RELATED ISSUES OF ADVANCED REACTORS. GENERATOR TUBING MATERIAL NUREG/CR4447: RESULTS OF CRACK-ARREST TESTS ON IRRADI-ATED A 508 CLASS 3 STEEL MODELING & COMPUTER SERVICES NUREG/CR4453: H. B. ROBINSON-2 PRESSURE VESSEL BENCH-NUREGOR4551: IMPROVED EMBRITTLEMENT CORRELATIONS FOR MARK. REACTOR PRESSURE VESSEL STEELS. NUREG/CR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAll. FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERLY EQUIPMENT IN NUCLEAR POWER PLANTS. NATIONAL SUREAU OF NUREG/CR4536: VERIFICATION OF THE LWRARC CODE FOR NUREG-1521 DRFT FC: TECHNICAL REVIEW OF RISK-INFORMED, LIGHT-WATER-REACTOR AFTERHEAT RATE CALCULATIONS. PERFORMANCE-BASED METHODS FOR NUCLEAR POWER PLANT NUREG/CR4537: INFLUENCE OF LONG-TERM THERMAL AGING ON FIRE PROTECTION ANALYSES. Draft Report For Comment THE MICROSTRUCTURAL EVOLUTION OF NUCLEAR REACTOR PRESSURE VESSEL MATERIALS.An Atom Probe Study. ( NATIONAL RADIOLOGICAL PROTECTION SOARD NUREG/CR4546: A DAMAGE MECHANICS BASED APPROACH TO NUREG/CR4571 V01: PROBABILISTIC ACCIDENT CONSEQUENCE STRUCTURAL DETERORATION AND REUABILITY. UNCERTAINTY ANALYSIS. Oncertanty Assessment For intomal Do. NUREG/CR4552: MARBLE HILL ANNEALING DEMONSTRATION simet . Main Report EVALUATION. NURE R4571 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4569: LOW-LEVEL WASTE DATA BASE DEVELOPMENT UNCERTAINTY ANALYSIS. Uncertainty Assessment For Intemal Do-e% Appendices. NU EG/C 4577: U.S. NUCLEAR POWER PLANT OPERATING COST AND EXPERIENCE SUMMARIES. NETHERLANDS ENERGY RESEARCH FOUNDATION ECN NUREG/CR4598. AN INVESTIGATON OF TENDON SHEATHING NUREGCR4545 V01: PROBASILISTIC ACCIDENT CONSEQUENCE F T UNCERTAINTY ANALYSIS. Early Health Effects Uncertairty Assess- NU E l OLA L ND PH CONTROL IN THE AP-menLMain Repor1. 600 REACTOR NUREG/CR46005 NEUTRON EXPOSURE PARAMETERS FOR CAP-NEW MEXICO, UNIV. OF, ALSUQUEROUE. NM SULE 10.05 IN THE HEAVY-SECTION STEEL IRRADIATON PRO-NUREGOR4545 V01: PROBABILISTIC ACCIDENT CONSEQUENCE ^ E IR AT ON E UNCERTAINTY ANALYSIS. Earty Heath Effects Uncertaarty Assess-N RE R DOSIMETRY CAPSULE IN THE HEAVY-SECTION STEEL IRRADIA-NU G/C 454 02: PROBABILISTIC ACCIDENT CONSEQUENCE NU G 46 :FE Li Y li UENCY ACOUSTIC IM-UNCEPTAINTY ANALYSIS. Earty Heath Effects Uncertainty Assens- AGING FOR INSPECTION OF CONTAINMENTS. menLAppendices NUREG/CR4615: A SURVEY OF REPAIR PRACTICES FOR NUCLEAR NORTH CAROLIN A, UNIV. OF, CHAPEL HILL, NC g

• NUREG/CR4556: A COMPREHENSIVE STUDY OF THE EASTERN TENNESSEE SEISMIC ZONE. ORGANIZATION FOR ECONOMlC COOPERATION & DEVELOPMENT NUREG/CP-0160: PROCEEDINGS OF THE OECD/CSNI SPECIALIST NUCLEAR POWER ENGINEERING CORP.

MEETING ON ADVANCED INSTRUMENTATON AND MEASURE-NUREG/CR4509: THE EFFECT OF INITIAL TEMPERATURE ON MENT TECHNIOUES. Held in Santa Barbara CA March 17-20,1997. FLAME ACCELERATION AND DEFLAGRATION-TO-DETONATION NUREG/CR-4667 W ENVIRONMENTALLY 'AS8tSTED CRACKING IN NUR 4524: H E CT OF LATERAL VENTING ON DEFLA-GRATION-TO DETONAT ION TRANSITON IN HYDROGEN-AIR- PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA STEAM MIXTURES AT VARIOUS INITIAL TEMPERATURES. NUREG/CR-5534: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A DOWNWARD FACING CURVED SURFACE: EFFECTS OF THERMAL OAK PilDGE ASSOCIATED UNIVERSITIES INSULATION. NUREG-1507: MINIMUM DETECTABLE CONCENTRATONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS CON- PURDUE UNIV., WEST LAFAYETTE, IN TAMINANTS AND FIELD CONDITIONS. NUREG/CR-5498: SINGLE-PHASE AND TWO-PHASE NATURAL CIR-NUREGCR4364: HUMAN PERFORMANCE IN RADOLOGICAL SUR- CULATION TESTS IN THE PUMA FACILITY. SANDIA NATONAL LABORATORIES OAK RIDGE NATONAL LABORATORY NUREGCR4119 V01 R1: MELCOR COMPUTER CODE MANU-NUREG 1608: CATEGORIZING AND TRANSPORTING LOW SPECIFIC ALS. Primer And Users' Guides, Version 1.8.4, July 1997. ACTIVITY MATERIALS AND SURFACE CONTAMINATED OBJECTS. NUREG/CR4119 V02 R1: MELCOR COMPUTER CODE MANU-NUREG/CR-4219 V13 N2: HEAVY-SECTION STEEL TECHNOLOGY ALS. Reference Manuals. Version 1.8.4. July 1997. PROGRAM. Semiannual Progress Report For April - September 1996. NUREG/CR4131: VICTORIA 2.0: A MECHANISTIC MODEL FOR NUREG/CR-4219 V14 N1: HEAVY-SECTION STEEL TECHNOLOGY RADIONUCLIDE BEHAVOR IN A NUCLEAR REACTOR COOLANT PROGRAM.Somicanual Progress Report For October 1996 - March SYSTEM UNDER SEVERE ACCIDENT CONDITIONS. S997. NUREG/CR4412: AGING AND LOSS-OF-COOLANT ACCIDENT (LOCA) NUREGCR-4674 V25: PRECURSORS TO POTENTIAL SEVERE CORE TESTING OF ELECTRICAL CONNECTIONS. DAMAGE ACCIDENTS: 1996. A Status R . NUREG/CR4475: RESOLUTION OF THE DIRECT CONTAINMENT NUREGOR-4674 V26: PRECURSORS TO TENTIAL SEVERE CORE HEATING ISSUE FOR COMBUSTION ENGINEERING PLANTS & DAMAGE ACCIDENTS:1997. A Status Report. BABCOCK & WILCOX PLANTS.

80 Contractor index NUREG/CR4479: TECHNICAL BASIS FOR ENVIRONMENTAL QUAll- NUREG/CR-4674 V26: PRECURSORS TO POTENTIAL SEVERE CORE FICATION OF MICROPROCESSOR-BASED SAFETY-RELATED DAMAGE ACCIDENTS:1997. A Status Report. EQUIPMENT IN NL NUREG/CR4410: NUCLEAR FUEL CYCLE FACILITY ACCIDENT ANAL-NUREG/CR4517: Ru) CLEAR POWER PLANTS.uND ROBIN PRETEST YSIS HANDBOOK. ANALYSES OF A STEEL CONTAINMENT VESSEL MODEL AND CONTACT STRUCTURE AS- NUREGCR4579: DIGITAL l&C SYSTEMS IN NUCLEAR POWER ' SEMBLY SUBJECT TO STATIC INTERNAL PRESSURIZATION. PLANTS. Risk-Screerung Of Environmental Stressors And A Compart-NUREGCR4$45 V01: PROBABILISTIC ACCIDENT CONSEQUENCE son Of Hardware Unavailability With An Existing Analog System. UNCERTAINTY ANALYSIS. Earty Health Effects Uncertainty Assess-CHNADM ENGNEEM CONSMANTS,INC NU Y454 V02: PROBABILtSTIC ACCIDENT CONSEQUENCE NUREGlCR4613 V01: CODE MANUAL FOR MACCS2. User's Guide, UNCERTAINTY ANALYSIS. Early Health Effects Uncertainty Assess- NUREG/CR4613 V02: CODE MANUAL FOR MACCS2. Preprocessor ment#A es. Codes COMIDA2, FGRDCF, IDCF2. NURE 4555 V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Late Health Effects Untertainty Assess-TEXAS, UNIV. OF, AUSTIN, TX NU R45 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-5434: ANCHOR BOLT BEHAVIOR AND STRENGTH DUR-UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess- ING EARTHOUAKES. NU V01: PROBABILISTIC ACCIDENT CONSEQUENCE TRANSPORTATION. DEPT. OF UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do- NUREG-1608: CATEGORIZING AND TRANSPORTING LOW SPECIFIC

                                                                                                                                                        ^

NL$ 71  : PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assessment For internal Do- UNITED KINGDOM NUREG/CR-5505: RR-PRODIGAL - A MODEL FOR ESTIMATING THE simetrv.AW.. PERFORMANCE TESTING OF PASSIVE NUREGCFM580: PROBABILITIES OF DEFECTS IN REACTOR PRESSURE VESSEL NU 3 T ION OF RETARDATION MECHA-NUREG R4555 VOI: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess-NUREG/CR4604: RADTRAD: A SIMPUFIED MODEL FOR RADIO. TRANSPORT AND REMOVAL AND E ESTIMATION. ment. Main Report. NU NU 4613 V01:CQDE MANUAL FOR MA . er's Guede. NUREG/CR4555 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUR R4613 V02: wDE MANUAL FOR MA . Preprocessor UNCERTAINTY ANALYSIS. Late Health Effects Uncertainty Assess-Codes COMIDA2, FGRDCF, IDCF2- rnent.Appendcas. SCIENCE APPLICATIONS INTERNATIONAL CORP (FORMERLY VANDERBILT UNIV, NASHVILLE, TN NUREG/CR-5562: DATING AND EARTHOUAKES: REVIEW OF QUA-NU 713 8 TiONAL RADIATION EXPOSURE AT COM-

 #                                                                                                 TERNARY GEOCHRONOLOGY AND ITS APPLICATION TO MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILi-                                           PALEOSElSMOLOGY.

TIES 1996.Twent NUREM/13 V19:y-Ninth Annual Report. OCCUPATIONAL RADIATION EXPOSURE AT COM-VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV BLACKSBURG, MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES VA 1997. Thirtieth Annual Report. NUREG/CR 4674 V25: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4556: A COMPREHENSIVE STUDY OF THE EASTERN DAMAGE ACCIDENTS: 1996. A Status Report. TENNESSEE SEISMIC ZONE. l 9 1 l (

m  ; International Organization index This index lists, in alphabetical order, the countries and performing organizations that pre-lured the NUREG/lA reports listed in this compilation. Listed below each country and per-f lorming organization are the NUREG/lA numbers and titles of their reports. If further informa-tion is needed, refer to the main citation by the NUREG/lA number. i l l RSPU8LIC OF KOR8A SLOVENIA 1 j KOREA ADVANCED INSTITUTE OF SCIENCE AND TECHNOLOGY UNIVERSITY OF LJUSLJANA ) l NUREGAA4147: ASSESSMENT OF REuP6 MOD 32 FOR STEAM NUREGAA-0141: RESULT OF BETHSY TEST 9.1.8 USING RELAP5/ i j CONDENSATION EXPERMENTS IN THE PRESENCE OF MODS. I go, g,7yo,EggRTI

                 ' 8L                 TUSE OF PCCS.                  NUREG/lA-0145. RELAP ASSESSMENT AGAINST PACTEL EXPERI-   l NUREGAA4143: ASSESSMENT OF RELAP5 NOD 3.2 WITH THE                 MENTAL DATA (REVISK)N 1).                              i

, LSTF EXPERWENT SMULATING A LOB 8 OF RESIDUAL HEAT gpgg l CENTRAL NUCLEAR VANDELLOS 11 NURE A 1 : ASSE OF E '2 WITH THE SEM18CALE NATURAL CIRCULATION EXPERMENT,S-NC48- NUREGAA-0149: ASSESSMENT OF RELAP5 MOD 3.2-NPA3.4 ) AGAINST A TRANSIENT OF HIGH NUCLEAR FLUX VARIATION i KOREA NUCLEAR FUEL COMPANY NUREGAA4130: ASSESSMENT OF RELAP5 NOD 32 USNG LOFT REACTOR TRIP, NATURAL CIRCULATION AND THE START OF A BREAK LOCA TEST.LP 024. MAIN PUMP IN THE VANDELLOS 11 NUCLEAR POWER PLANT. NURE 142: MSTALLATION OF RELAP6 MOD 3.2 ON 80486 P.M.S.A. AND PENTIUM 8ASED PERSONAL COMPUTERS. NUREG/lA-0150: STUDY OF TRANSIENTS RELATED TO AMSAC NUREGA 148 ASSE NT OF R LAP 6 MOD 3.1 USING LSTF

                                                                            ^    '   ""              '*

i TEN-PERCENT MAIN STEAM-LINE-8REAK TEST RUN SS-SL-01. SWITIERLAND I PAUL SCHERRER INSTITUTE AU NUREGAA-0140: DEVELOPMENT ASSESSMENT OF RELAP5/ SAFETY INSTITUTE NUREGAA-0024: APPLICATION OF RELAP5 MOD 3.1 CODE TO THE MOD 3.1 WITH SEPARATE-EFFECT AND INTEGRAL TEST EX-LOFT TEST L34. PERWENTS:MODEL CHANGES AND OPTIONS. NUREG4A-0026: RELAPSNOD3 SUSCOOLED BOluNG MODEL AS. NUREG/lA4146: IMPLEMENTATION AND ASSESSMENT OF N-SE88 MENT. PROVED MODELS AND OPTIONS IN THAC-BF1. I 81

I 1 I i j I B

s Licensed Facility Index Thb 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 cnd followed by the report number. If further information is needed, refer to the main citation by the NUREG nurnber. ca: u m0s ,opw.o n.weennoo m ainz mai na m.on w um 2.c % a acaw E403 Sienduti Pts D.sqpi, Wesanghouse NUREG O Wl600 72420 INE TML2 tSFS, m41826 f

NRC FORM 336 U.s. NUCLEAR REGULATORY COMMiss4ON 1. REPORT NUMBE3 94e) (Assioned by NRc, Add vol., supp new., E"3E Bl!UOGRAPHIC DATA SHEET **'*"'""""""*"*"'l (s= *=tucnons en e.,ev => L TrrLE AND sV8 TITLE NUREG-0304 Vol. 23, No. 2 Abstracts for Publications in the NUREG-Series

3. DATE REPORT PUBUSHED AnnualCompilation for 1998 MONN YEAR l

April 1999 in Vol. 23, No.1 of NUREG4304, the title was changed from " Regulatory and Technical Reports

4. FIN OR GRANT NUMBER (Abstract index Joumal)"

5. AUTHOR (S) s. TYPE OF REPORT

7. PERIOO COVERED (anctusse osses;

8. PERFORMING ORGANIZATION - NAAE AND ADDRESS (rNRc; provade Dvam omos a Aepm u3. Nuedear Aepusetry c_  ; and memna addoss; acontecer prov,no nome and mes, oneus>

Publishing Services Branch Offica of the Chief information Officer U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

9. SPONSORING ORGANIZATION . NAME AND ADDRESS (rNRc type 'Same as a6ove' #contecer, provafe NRC Dvem oree a Repson, u 3. Nucher Repudskry commusion, and memne eatess}

Sims as 8, above.

10. SUPPLEMENTARY NOTES L. L. Stevenson, Project Manager

11. ABSTRACT 900 words aJess)

This journalincludes all formal reports in the NUREG-series prepared by the NRC staff and contractors; proceediPgs of conferences and workshops; as well as international agreement reports. The entries in this compilation are ind: sed for access by titlJ and abstract, secondary report number, personal author, subject, NRC organization for staff and intemt.donal agreements, i contractor, intemational organization, and licensed facility. In Vol. 23, No.1, of NUREG-0304, the title was changed from

• Regulatory and Technical Reports (Abstract index Journal)." i l

12. KEY WORDSOESCRIPTORS (ust woes a pirems yet wd uset reseechars a bee 6ap pe rspar) 13. AVAILABUTY STATEMENT i

Unlimited Ccmpilation 14 SECURTrYCLASSIFCATON cbstractindex (7ms pays; Ebstract NRC publications Unclassified NUREG-series publications g% Unclassified

15. NUMBER OF PAGES is. PRCE NRC FoR' J 336 Q40) The form was electorucelly r,toduced by Eine Federal Forms. Inc.

g..., i i Printed on recycled paper i Federal Recycling Program

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