ML20149D437

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Second Quarter 1993,April-June

ML20149D437
Person / Time
Issue date:	08/31/1993
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V18-N02, NUREG-304, NUREG-304-V18-N2, NUDOCS 9309210040
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NUR EG-0304 Vol.18, No. 2 4

Regu:atory anc Tecanica Repor:s Tabs:ract Inc ex JournaD)

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l Compilation for Second Quarter 1993 April - June  !

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NUREG-0304 Vol. ]8. No. 2 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for Second Quarter 1993 April- June Date Published: August 1993 Regulatory Publications Ilranch Division of Freedom ofInformation and Publications Senices Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 p ~.vu t

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CONTENTS  !

J Preface. . . . v t

index Tab  !

Main Citations and Abstracts . - , . 1

• Staff Reports

• Conference Proceedings

• Contractor Reports -

• International Agreement Reports Secondary Report Number Index . 2 Personal Author Index 3 -i 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 international Organization Index . . . . . . 9 Licensed Facility Index . . . . . .10 i

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l PREFACE  !

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

Regulatory Publications Branch Division of Freedom of Information and Publications Services [

P-223 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 [

The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, '!

NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA XXXX. These precede the following indexes:

Secondary Report Number Index '

Personal Author index Subject index NRC Originating Organization index (Staff Reports) 'i NRC Originating Organization index (International Agreements) i NRC Contract Sponsor Index (Contractor Reports)

Contractor index International Organization Index Licensed Facility index -

A detailed explanation of the entries precedes each index.

The bibliographic elements of the main citations are the following:

• Staff Report I

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

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

Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author,15) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for internal NRC use).

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

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

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

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

i Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of . '

authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC i Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC intemal use).

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

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

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

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

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

NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replacef contractor-estab!ished codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported, in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/lA is used for international agreement reports.

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

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1 Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG-XXXX is j an NRC staff-originated report, NUREG/CP-XXXX is an NRC-sponsored conference report, i NUREG/CR-XXXX is an NRC contractor-prepared report, and NUREG/lA-XXXX is an inter-  ;

national agreement report. The bibliographic information (see Preface for details) is followed i by a brief abstract of this report.

NUREG-0040 V17 N01: LICENSEE CONTRACTOR AND NRC is committed to the penodic publication of licensed fuel VENDOR INSPECTION STATUS REPORT. Quarterly facilities inventory difference data, following agency review of .;

Report. January-March 1993 (White Book)

• Division of Reactor the information and completion of any related NRC investga-  !

Inspection & Licensee Performance (Post 921004). May 1993- tions. Information in this report includes inventory difference 226pp 9306180278. 75388 204.

data for active fuel fabncation facilities possessing more than This penodical covers the results of snspections performed by one effective kilogram of high ennched uranium, low ennched the NRC's Vendor inspecten Branch that have been distnbuted uranium, plutonium, or uranium-233.  ;

to the inspected organciations dunng the penod from January through March 1993. ,

NUREG-0540 V15 NO2: llTLE LIST OF DOCUMENrS MADE i "7 * **"

NUREG-0090 V15 N04: REPORT TO CONGRESS ON ABNOR- " '" * * ' * " "" #""

MAL OCCURRENCES October-December 1992.

• Office for 99 2pp.930 25 1 9 750 3 s o n samn caton containing desenp- ,

Section 208 of the Energy Reorganization Act of 197/ adenti- ons f maton mceived and generated by the U.S. Nuclear J fies an abnormal occurrence as an unscheduled incident or Regulatory Commission (NRC). This information !ncludes (1) event that the Nuclear Regulatory Commission determ'nes to be docketed material associated with civihan nuclear power plants significant from the standpoint of public health and safety and and othar uses of radioactive materials, and (2) nondocketed requires a quarterly report of such events to be made to Con- matenal received and generated by NRC pertinent to its role as gress. This report covers the period October through December a regulatory agency. The following indexes are included: Per-l 1992. There were two abnormal occunences at nuclear power sonal Author, Corporate Source, Report Number, and Cross  !

plants. Six abnormal occurrences involving medicat midadminis- Reference of Enclosures to Pnncipal Documents.  ;

trations (all therapeutic) at NRC-hcensed facihties are discussed in this report No abnormal occurrences were reported by NUREG-0540 V15 NO3: TITLE UST OF DOCUMENTS MADE NRC s Agreement States. The report also contains information PUBUCLY AVAILABLE' March 1-31, 1993.

• Division of Free- i updating' previously reported abnormal occurrences _ d m of information & Pubhcatons Services (Post 890205). May 1993. 400pp. 9306010264. 75057:001.

NUREG-0304 Vi8 N01: REGULATORY AND TECHNICAL RE- See NUREG-0540,V15 N02 abstract PORTS (ABSTRACT INDEX JOURNAL). Compilation For First Quarter 1993.JanuaryMarch.

• Division of Freedom of Informa- NUREG-0540 V15 N04: TITLE LIST OF DOCUMENTS MADE tion & Publications Services (Post 890205) May 1993. 48pp. PUBLICLY AVAILABLE.Apnl 1-30, 1993.

• Division of Freedom 9306110059. 75338 252. of informaton & Pubhcations Services (Post 890205) June This joumal includes all formal reports in the NUREG senes 1993 350pp. 9306290175. 75499:053. i prepared by the NRC staff and contractors; proceedings of con- See NUREG-0540,Vt 5,N02 abstract I forences and workshops; as well as international agreement re- '

ports. The entnes in this compilation are indexed for access by NUREG-0725 R09: PUBLIC INFORMATION CIRCULAR FOR

, title and abstract, secondary report number, personal author, SHIPMENTS OF IRRADIATED REACTOR FUEL.

• Division of subject, NRC organizaton for staff and intemational agree- Safeguards & Transportation (870413-930206). March 1993.

monts, contractor, international organization, and hcensed facib- 37pp. 9304190142. 74624:305.

J ty.

This circular has been prepared to provide information on the NUREG-0386 D06 R06: UNITED STATES NUCLEAR REGULA- shipment of irradiated reactor fuel (spent fuel) subject to regula- '

TORY COMMISSION STAFF PRACTICE AND PROCEDURE ton he Nucbar Ngdaton Comrnssen M, aM to med ,

DIGEST. Commission, Appeal Board And Licensing Board the requirements of Pubhc Law 96-295. The report provides a Decisions. July 1972 - June 1992.

• Office of the General Coun- bnef descnrnion of NRC authority for certain aspects of trans-sel (Post 860701L May 1993. 600pp. 9306010258. 75055:001. porting spent fuel it provides desenptive statistics on spent fuel This 6th revision of the sixth edition of the NRC Practice and shipments regulated by the NRC from 1979 to 1992. It also hsts detaned highway and railway segments used within each state Procedure Digest contains a digest of a number of Commission,

! Atomac Safety and Licensing Appeal Board, and Atomic Safety from October 1,1987 through December 31,1992.

1 and Licensing Board decisions issued during the period of July 1 NUREG-0750 V36101: INDEXES TO NUCLEAR REGULATORY i CRP I'992, in orpreting the NRC s Rules of Prac- COMM!SSION ISSUANCES. July-September 1992.

• Divison of Freedom of information & Publications Services (Post 890205).

NOREG-0430 V12: LICENSED FUEL FACILITY STATUS April 1993. 52pp. 9305250147. 75005:052-

REPORT. inventory Dstierence Data. July 1,1991 June 30, Digests and indexes for issuances of the Commission, the 1992,(Gray Book II) JOY,Da BROWN.C Office of Nuclear Mate- Atomic Safety and Licensing Board Panel, the Administrative I nal Safety & Saloguards. April 1993. 20pp. 9305100071. Law Judges, the Directors' Decisions, and the Denials of Peti-74857

295 tions for Rulemaking are presented.

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2 Main Citations and Abstracts i NUREG-0750 V36102: INDEXES TO NUCLEAR REGULATORY certainties and other quantitative or qualitative factors. To the COMMISSION ISSUANCES. July-December 1992.

• Division of extent practical, estimates are quantitative.

NUREG-0936 V12 N01: NRC REGULATORY AGENDA.Ouarterly ne 993 73 p 93 060 47 00 See NUREG-0750,V36.101 abstract. Report. January-March 1993.

• Division of Freedom of Informa-tion & Publications Sennces (Post 890205). April 1993.135pp.

NUREG-0750 V37 NO2: NUCLEAR REGULATORY COMiWSSION 9305250050. 75004:072.

ISSUANCES FOR FEBRUARY 1993. Pages 55-134.

• Divison The NRC Regulatory Agenda is a compilanon of all rules on of Freedom of Information & Pubhcations Services (Post which the NRC has recently completed acton, or has proposed l 890205) April 1993. 85pn 9305250140. 75004:213. action, or is considering action, and all petitions for rulemaking '

legal issuances of the Commissson, the Atomic Safety and Li- which have been received by the Commission and are pending censing Board Panel, the Administrative Law Judges, and NRC disposition by the Commission. The Regulatory Agenda is up- ,

Program Offices are presented. dated and issued each quarter.

NUREG-0750 V37 NO3: NUCLEAR REGULATORY COMMISSION NUREG-0940 V12 N01: ENFORCEMENT ACTIONS.SIGNIFICANT ISSUANCES FOR MARCH 1993. Pages 135-249.

• Division of ACTIONS RESOLVED.Ouarterly Progress Report January-March Freedom of information & Publications Services (Post 890205). 1993.

• Ofc of Enforcement (Post 870413). June 1993. 250pp.

May 1993.150pp. 9306210229. 75403:016. 9306210211. 75406:001.

See NUREG-0750,V37,N02 abstract- This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (January -

NUREG-0797 S27: SAFETY EVALUATION REPORT RELATED March 1993) and includes copies of letters, Notices, and Orders TO THE OPERATION OF COMANCHE PEAK STEAM ELEC- sent by the Nuclear Regulatory Commission to licensees with TRIC STATION, UNIT 2. Docket No. 50-446 (Texas Utilities Elec_

respect to these enforcement actions. It is anticipated tnat the inc Company et al.)

• D' vision of Reactor Projects - til,lV,V (Post informaton in this pubhcaton will be widely disseminated to 901216). April 1993. 49pp. 9305100006. 74859:074. i managers and employees engaged in activities licensed by the Supplement No. 27 to the Safety Evaluation Report related to NRC, so that actions can be taken to improve safety by avoid-the operation of the Comanche Peak Steam Electric Station, Unit 2, has been prepared by the Office of Nuclear Reactor ing future violations similar to those desenbed in this publica-

' tion.

Regulation of the U.S. Nuclear Regulatory Commission. The fa-cility is located in Somervell County, Texas, approximately 40 NUREG-1100 V09: BUDGET ESTIMATES. Fiscal Years 1994-miles southwest of Fort Worth, Texas. This supplement reports 1995.

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

the status of certain issues that had not been resolved when 213pp. 9304160046. 74643:208.

the Safety Evaluation Report and Supplements 1,2,3.4,6,12. This report contains the fiscal year budget justification to Con-21, 22, 23, 24, 25, and 26 to that report were pubisshed. This gress The budget provides estimates for salaries and expenses supplement deals pnmarily with Unit 2 issues and for the Office of the inspector General for fiscal years 1994 and 1995.

NUREG0837 V13 N01: NRC TLD DIRECT RADIATION MONI.

TORING NETWORK. Progress Report. January-March 1993. NUREG-1125 V14: A COMPILATION OF REPORTS OF THE AD.

STRUCKMEYER R.; MCNAMARA,N, Region 1 (Post 820201). VISORY COMMITTEE ON REACTOR SAFEGUARDS.1992

^ '

This report p des the at and results of the NRC Ther- 93 2 930 1 762 1 moluminiscent Dosimeter (TLD) Direct Radiation Monitonng Net. This compilation contains 50 ACRS reports submitted to the work It presents the radiation levels measured in the vicinity of Commission, Executive Director for Operations, or to the Offic e NRC licensed facilities throughout the country for the first quar. of Nuclear Regulatory Research, dunng calendar year 1992. .t ter of 1993' also includes a report to the Congress on the NRC Safety Re-NUREG-0847 S11: SAFETY EVALUATION REPORT RELATED search Protram. All reports have been made available to the TO THE OPERATION OF WATTS BAR NUCLEAR public through the NRC Public Document Room and the U.S. Li-PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Ten- brary of Congress. The reports are divided into two groups: Part nesee Valley Authonty)

• Division of Reactor Projects - 1/11 (Post 1: ACRS Reports on Project Reviews, and Part 2: ACRS Re- '

B70411). Apnl 1993 50pp. 9305250123. 75005:001. ports on Genenc Subjects. Part 1 contains ACRS reports alpha-Supplement No.11 to tne Sefety Evaluation Report for the bettzed by project name and by chronological order within apphcation filed by the Tennessa Valley Authority for license to project name. Part 2 categonzes the reports by the most appro-operate Watts Bar Nuclear Plant, Units t and 2, Docket Nos. priate generic subject area and by chronological order within 50-390 and 50-391, located in Rhea County, Tennessee, has subject area. ,

been prepared by the Office of Nuclear Reactor Regulation of I NUREG-1266 V07: NRC SAFETY RESEARCH IN SUPPORT OF the Nuclear Regulatory Commission. The purpose of this sup- REGULATION - FY 1992.

• Office of Nuclear Regulatory Re-piement is to update the Safety Evaluation of: (1) additional in- search (Post 860720). May 1993. 76pp. 9306210373.

formation submitted by the applicant since Supp!ement No.10 75401:319

• was issued, and (2) matters that the staff had under review This report, the eighth in a series of annual reports, was pre-when Supplement No.10 was issued pared in response to congressional inquines concoming how NUREG-0933 S15: A PP ORITIZATION OF GENERIC SAFETY nuclear regulatory research is used. It summarizes the accom-i ISSUES. EMRIT.R. Division of Safety issue Resolution (Post plishments of the Office of Nuclear Regulatory Research during 880717). May 1993.186pp. 9306110046. 75338:007. FY 1992. A special emphasis on accomplishments in nuclear ,

The report presents the pnonty rankings for genene safety power plant aging research reflects recognition that a number of issues related to nuclear power plants. The purpose of these plants are entering the final portion of their original 40-year op-rankings is to assist in the timely and efficient allocation of NRC erating licenses and that, in addition to current aging effects a resources for the resolut;on of those safety issues that have a focus on safety considerations for license renewal becomes i significant potential for reducing nsk. The safety priority rankings timely. The primary purpose of performing regu!atory research is are HIGH, MEDIUM, LOW, and DROP and have been assigned to develop and provide the Commission and its staff with the on the basis of nsk significance estimates, the ratio of risk to technical bases for regulatory decisions on the safe operation costs and other impacts estimated to resu!t if resolutions of the of licent.ed nuclear reactors and facilities, to find unknown or '

safety issues were implemented, and the consideration of un- unexpected safety problems, and to develop data and related 5 1

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Main Citations and Abstracts 3 information for the puroose of revising the Commission's rules, mended that this genenc issue be resolved simply by making regulatory guides, or other guidance. these results available en a genenc letter. This information may MUREG-1307 R03: REPORT ON WASTE BURIAL help licensees in their plant evaluations recommended by Ge-CHARGES Escalation Of Decommissioning Waste Disposal nenc Letter 88-20, Supplement 4, " Individual Plant Exarr.

Costs At Low-Level Waste Bunal Facahties.

• D vision of Regula.

of Extemal Events for Severe Accident Vulnerabihtees,; nation June tory Applications (Post 870413). May 1993. 59pp 9306110042. 28 N t 75338.203-NUREG-1415 V05 NO2: OFFICE OF THE INSPECTOR ,

One of the reouirements placed upon nuclear power reactor GENERAL. Semiannual Report,0ctober 1, 1992 . March 31, '

licensees by the U.S. Nuclear Regulatory Commission (NRC) is 1993.

• Office of the inspector General (Post 890417). Aprit for the licensees to penodically adlust the estimate of the cost 1993 38pp. 9306210379. 75401:278.

of decsmmissioning their plants, in dollars of the current year, The inspector General is required by statute to prepare a se-

• as part of the process to provide reasonable assurance that miannuaireport to Congress which summanzes the significant in-adequate funds for decommissioning will be available when vestigative and audit activities of the office. The 6-month report-needed This report, which is scheduled to be revised annually, ing penod ends March 31 and September 30. The report is sub-contains the development of a formula for escalating decom- mitted to the Chairman not later than Apnl 30 and October 31, missioning cost estimates that is acceptable to the NRC, and respectively, of each year. The Chairman prepares comments contains values far the escalation of radioactrve waste bunal and his own report and submits both reports to Congress.

costs, by tite and by year. The licensees may use the formula, the coefficients, and the burial escalation from this report in NUREG-1474: EFFECT OF HURRICANE ANDREW ON THE their escalation analyses, or they may use an escalation rate at TURKEY POINT NUCLEAR GENERATING STATION FROM least equal to the escalation approach presented herein. Rev,_ AUGUST 20-30, 1992. HEBDON.F.J. Office for Analysis & Eval-sion 3 of this report corrects several errors in the calculations uation of Operational Data, Director.

• Institute of Nuclear and disposal costs for the reference PWR and the reference Power Operations March 1993 80pp. 9307060041. 75585.001.

BWR. On August 24,1992, Humcane Andrew, a Category 4 hurn-cane, struck the Turkey Point Electncal Generating Station with MUREG-1350 V05: NUCLEAR REGULATORY COMMISSION IN' sustained winds of 145 mph (233 km/h). This is the report of FORMATLON DtGEST.1993 Edition. OltVE.KL Dmsion of the team that the U S. Nuclear Regulatory Commission and the Budget & Analysis (Post 890205) March 1993. 127pp- Institute of Nuclear Power Operations jointly sponsored: (1) to 9305250029. 75005-098.

review the damage that the hurricane caused the nuclear units The Nuclear Regulatory Commission information Digest and the utikty's actions to prepare for the storm and recover (dagest) provides a summary of information about the U S. Nu-from it, and (2) to compile lessons that might benefit other nu-clear Regulatory Commission (NRC), NRC's regulatory responsi- clear reactor facilities.

bikties, the activities NRC hcenses, and general information on domestic and worldwide nuclear energy. The digest, pubbshed NUREG-1477 DRFT FC: VOLTAGE-BASED INTERIM PLUGGING annually, is a compilation of nuclear and NRC-related data and CRITERIA FOR STEAM GENERATOR TUBES Draft Report For is designed to provide a quick reference to major facts about Comment.

• IPC Task Group. June 1993,120pp. 9307060061.

the agency and the industry it regulates. In general, the data 75563.068.

cover 1975 through 1992, with exceptions noted information on This report presents the prehrmnary results of a special U.S.

generating capacity and average capacity factor for operating Nuclear Regulatory Commission (NRC) task group established:

U S. commercial nuclear power reactors is obtarned from (1) to review the technical bases for and outstanding issues re-monthly operating reports that are submitted directly to the NRC lated to interim approval of voltage-based intenm plugging ente-by the hcensee. This information is reviewed by the NRC for na for outside diameter stress corrosion cracking (ODSCC) of consistency only and no independent vahdation and/or venfica- steam generator tubes; and (2) to prepare conclusions and rec-tron is performed. ommendations concerning implementation of these enteria. The NUREG-1364; REGULATORY ANALYSIS FOR THE RESOLU-

• 9' "P * * " "' "* **"**"

TlON OF GENERIC SAFETY ISSUE 106: PlPING AND THE " "" 9 "'#"***** "* * * #

USE OF HIGHLY COMBUSTIBLE GASES IN VITAL AREAS'

" 9 "9 9 * "9 "9 "

GRAVES.C C. Division of Safety Issue Resolution (Post ODSCC. Most of these issues are relevant to the long-term ap-880717). June 1993. 48pp. 9307130111. 75653:153. proval of voftage-based plugging cntena. This report descnbes Highty combustible gases such as hydrogen, propane, and the results of the task group's review and evaluation of: (1) the acetylene are used at all nuclear power plants. Hydrogen is of issues related to tube integnty, including the potential for tube rupture or feakage under postulated sccident conditions and the particular importance because it is stored in large quantities and is distributed and used continuousty in buildings containing safety impbcations of these issues; (2) the radsological doses safety-related equipment. Large hydrogen releases at the hydro- and the potential for core damage associated with a range of gen storage facilsties or in these buildings could lead to fires or assumed pnmary-to-secondary leak rates; and (3) the safety explosions that might result in loss of safety-related equipment. significance of ODSCC of steam generator tubes.

This report grves the regulatory analysis for the resolution of NUREG-1485: UNAUTHORIZED FORCED ENTRY INTO THE Genenc Safety issue 106, " Piping and the Use of Highly Com- PROTECTED AREA AT THREE MILE ISLAND UNIT 1 ON FEB-bustible Gases in Vital Areas." Scoping analyses showed that RUARY 7,1993.

• Ofc of the Executive Director for Operations.

tne nsk associated with the storage and distnbution of hydrogen April 1993.142pp. 9304210263. 74677:061, for cooling electne generators at boihng water reactors (BWRs). On February 7,1993, at 6:53 a.m. Eastern Standard Time the off-gas system at BWRs, the waste gas system at pressur. (EST) an intruder drove into the site owner-controlled area, tred-water reactors (PWRs), and station battery rooms and port- through a gate into the protected area of Three Mile Island Nu-able bottles of combustible gas used for maintenance at PWRs clear Generating Station, Unit 1 (TMi-1) and crashed trerough a and BWRs is small. On the basis of genene evaluations, the rotl-up door on the Turbine Building. TMI Secunty reported this NRC statt has concluded that several possible methods to event to the U S. Nuclear Regulatory Commission's (NRC's) reduce nsk could provide cost-effective safety benefits at some Headquarters operations officer and declared a Secunty Emer-plants. However, in view of the observed large differences in gency upon determining that the protected area of the plant had plant-specific charactenstics affecting the nsk associated with been compromised. At 7.23 a.m, the TMI-1 shift supervisor offt-the use of hydrogen, and the marginal genenc safety benefits cially notified the NRC Headquarters operations officer that he that can be achieved in a cost-effective maner, tt is recom- had declared a Site Area Emergency effective at 7:05 a.m.

4 Main Citations and Abstracts Upon considering the possible significance to physical secunty This is the fourth semiannual report of the U.S. Nuclear Regu-and the regulatory questions that could result from the event, latory Commisson's Short Cracks in Piping and Piping Welds the NRC Executive Director for Operat ons established an inci- research program. This 4. year program began in March 1990.

dent investigation team to determine what happened and make The overall objective of this program is to venty and improve appropnate findings and conclusions. In this report the team de- fracture analyses for circumferentially cracked large-diameter l I

Scnbed the event and the response to the event, evaluated the nuclear piping with crack sizes typically used in leak-before-regulatory requirements, and presented the team's findings and break analyses or in-service flaw evaluations. Progress during ,

conclusions. this reporting period involved: (1) completing two through-wall-cracked pipe expenments and supplementary material property NUREG/CP-0126 V0t: PROCEEDINGS OF THE TWENTIETH data (2) an internal circumferential surface-cracked pipe experi- t WATER REACTOR SAFETY INFORMATION MEETING, ment was completed which showed that the R/t effects on the WEISS.A.J. Brookhaven National Laboratory. March 1993 Net-Section-Collapse predcted loads for surface-cracked pipe i- 535pp. 9304260111. 74713 001. to be independent of crack size, (3) the anisotropy investigation This three-volume report contains 93 papers out of the 108 showed that pipe dimensions may be as important in determin-that were presented at the Twentieth Water Reactor Safety in, ing the out-of-plane crack growth angle as the anisotropy of the formation Meeting held at the Bethesda Marriott Hotel, Bethes.

toughness, (4) we initiated a probabilistic analysis of LBB to da, Maryland, during the week of October 21-23,1992. The assess the potential changes in the leakage detection criteria in papers are pnnted in the order of their presentation in each ses.

sion and desenbe progress and results of programs in nuclear NRC Reg Guide 1.45, and (5) other efforts involved a sensitivity safety research conducted in this country and abroad. Foreign study on the effect of thermal aging of cast stainless steel on the moment-carrying capacity of the pipe as a function of time. l participation in the meeting included 10 different papers pre.

sented by researchers from CEC, China, Finland, France, Ger-NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACK- }

many, Japan. Spain and Taiwan. The titles of the papers and ING IN LIGHT WATER REACTORS. Semiannual Report, April-the names of the authors have been updated and may differ September 1992. RUTHER.W E.; CHUNG.H.M4 CHOPRA.O.K.;

from those that appeared in the final program of the meeting. et at Argonne National Laboratory June 1993. 76pp.

NUREG/CP-0126 V02: PROCEEDINGS OF THE TWENTIETH 9306290083. ANL-93/2. 75496.107.

WATER REACTOR SAFETY INFORMATION MEETING. This report summarrzes work performed by Argonne National WEISS,A.J. Brookhaven National Laboratory. March 1993- Laboratory on fatigue and environmentally assisted cracking l 555pp. 9304190226. 74621:001. (EAC) in light water reactors (LWRs) during the six months from  !

i Soo NUREG/CP-0126,V01 abstract. April 1992 to September 1992. Topics that have been investi-WUREG/CP-0126 V03: PROCEEDINGS OF THE TWENTIETH gated include: (1) f atigue and stress corrosion cracking (SCC) of Iow-alloy steel used in piping steam generators, and reactor j WATER REACTOR SAFETY INFORMATION MEE TING.

WEISS,A.J. Brookhaven National Laboratory March 1993. pressure vessels: (2) EAC of cast stainless steels (SSs), and (3)  !

585pp 9304190227,74623t01. radiation-induced segregation and irradiation-assisted SCC of See NUREG/CP-0126,V01 abstract. Type 304 SS after accumulation of relatrvely high fluence. Data  ;

n fa gw aH y s en n n s am Wn e  :

NUREG/CR-4214 R1P2A2: HEALTH EFFECTS MODELS FOR NUCLEAR POWER PLANT ACCIDENT CONSEQUENCE y asd on kactuemdaNcs Mels and eginwng i

ANALYSIS Modification Of Models Resulting From Addition Of I" 9*"*" 9"'""**

• are mns s n" w

• a'"va'9"* dane faMM dah Dad WM daj Effects Of Exposure To Alpha-Emitting Radionuchdes Part it w re btained on fracture-mechanics specimens of A533-Gr B  ;

Scientific Bases For Health... ABRAHAMSON S. Wisconsin aM NG B fece sWs and on cast ausesc SSs m N ,

Unw. of, Madison, Wl. BENDER M.A. Brookhaven National Lab' asmeW ad Nmah agM mWons in nan M oratory. BOECKER B B.; et al inhalation Toxicology Research water at 289 degrees C. The data were compared with predic-Institute. May 1993. 87pp. 9306020013. LMF 136. 75082:159. tions based on crack growth correlations for femtic steels in ox-Several studies designed to identrfy and quantify the potential ygenated water and correlations for wrought austenitic SS in ox-health effects of accidental releases of radionuclides from nu- ygenaW water daveloped at ANL and rates in air from Section clear power plants have been sponsored by the Nuclear Regu- f the ASt. E Code, Microchemical and microstructural latory Commission. Report NUREG/CR-4214, Rev.1. Part li changes in high and commercial-purity Type 304 SS specimens (NRC,1989a) desenbes in detail the most recent health effects from control-blade absorber tubes and a control-blade sheath models that have evolved from these efforts. Since the Part il from operating BWRs were studied by Auger ele: tron spectros- ,

report was published in 1989, two addenda to that report have copy and scanning electron microscopy. Slow-strain-rate-tensile r been prepared to 1) incorporate other scientifc information re- tests were conducted on irradiated specimens in air and simu- ,

lated to low-LET health effects models and 2) extend the '# "" I models to consider the possible health consequences of includ-i ing alpha-emitting actinide radionuclides in the exposure source NUREG/CR-4735 V08: EVALUATION AND COMPILATION OF term. The first addendum was published as NUREG/CR-4214, DOE WASTE PACKAGE TEST DATA. Biannual Report, August Rev.1, Part 11, Addendum 1 (NRC,1991). This report, the 1989 - January 1990. INTERRANTE C.G. Geology & Engineer. I second addendum to the Part il report, extends the health ef- ing Branch (Post 910506). FRAKER,A C.; ESCALANTE,E. Na-fects models to consider chronic stradiation from elpha emrtting tional institute of Standards & Technology (formerly National k radionuclides as well as low-LET sources. Consistent with the Bureau of Standa. June 1993.114pp. 9306290106. 75496:183.

organization of past reports. this report has three main sections This report summanzes evaluations by the National Institute .

that address early-occurring and continuing effects, late somatic of Standards and Technology (NIST) of some of the Depart-effects, and genetc effects. These results should be used with ment of Energy (DOE) activities on waste packages designed the basic NUREG/CR-4214 report and Addendum 1 to obtain for containment of radioactive high-level nuclear waste (HLW) current views on potential health effects models for radionu- for the six-rnonth penod, August 1989 - January 1990. This in- .

chdes released accidentally from nuclear power plants. cludes reviews of related matenals research and plans, informa-NUREG/CR-4599 V02 N2: SHORT CRACKS IN PIPING AND tion on the Yucca Mountain, Nevada disposal site activities, and PIPING WELDS Semiannual Report, October 1991 - March other information regarding supporting research and special as- ,

1992. WILKOWSKI,G M.; BRUST,F.; FRANCINI.R.; et al Bat- sistance. Short discussions are given relating to the publications telle Memonal Institute, Columbus Laboratories. May 1992 reviewed and complete reviews and evaluations are included.

53pp. 9306180295. BML2173. 75389.304. Reports of other work are included in the Append'ces. F

' = * ' ' -- u -a., - - _,, _ _ ,

Main Citations and Abstracts 5 NUREG/CR-4744 V07 N1: LONG-TERM EMBRITTLEMENT OF damage frequency. A detailed analysis of the fire nsk resulted in CAST DUPL EX STAINLESS STEELS IN LWR a total (mean) core damage frequency of 3 21E-5 per year.

SYSlEMS. Semiannual Report. October 1991 March 1992.

CHOPRA,0 K. Argonne National Laboratory. May 1993.152pp. NUREG/CR-5247 V02: RASCAL VERSION 2.0 WORKBOOK.

9306180315. ANL-92/42. 75387.001. ATHEY,G F. Athey Consulting. MCKENNA.T.J. Incident Re-This progress report summanzes work performed by Argonne sponse Branch. May 1993.105pp 9306110031. 75321:069.

National Laborator, on long-term thermal embnttlement of cast The Radiological Assessment System for Consequence Anal-duplex stainless stet % in LWR systems dunng the six months YS's, Version 2.0 (RASCAL 2.0) has been developed for use by from October 1991 to March 1992. Charpy-impact, tensile and the NRC personnel who respond to radiological emergencies.

fracture toughness J4 mrve data are presented for several This workbook as intended to complement the RASCAL 2.0 heats of cast stainless stee; that were aged 10.000-58,000 h at User's Guide (NUREG/CR 5247, Vol.1). The workbook con-290, 320, and 350 degrees C. The results indicate that thermal tains exercises designed to famihanze the user with the comput-  ;

aging decreases the fracture toughness of cast stainless steels. er based tools of RASCAL through hands-on problem solving.

In general CF 3 steels are the least sensitive to thermal aging The workbook is composed of four major sections. The first part and CF-BM steels are the most sensitwe. The values of fracture is a RASCAL famihanzation exercise to acquaint the user with toughness J(IC) and teanng moduius for CF-BM steels can be the operation of the forms, menus, on-line help, and documen-as low as = 90 kJ/m(2) and = 60. respectively. The fracture tation. The latter three parts contain exercises in using the three toughness data are consistent with the Charpy impact results, i e.. tools of RASCAL Version 2.0; DECAY, FM-DOSE, and ST-unaged and aged steels that show low impact energy also DOSE. Each section of exercsses is followed by discussion on a exhibit lower fracture toughness. All steels reach a minimum how the tools could be used to solve the problem.

• saturation fracture loughness after thermal aging; the time to reach saturation depends on the aging temperature The results NUREG/CR-5305 V02 P1: INTEGRATED RISK ASSESSMENT also indicate that low-strength cast stainless steels are general-FOR THE LASALLE UNIT 2 NUCLEAR POWER ly insensitive to thermal aging PLANT.Phenomenology And Risk Uncertainty Evaluation Pro-gram (PRUEP) Appendices A-C. BROWN,T.D.; PAYNE,A.C.;

NUREG/CR-4832 V05: ANALYSIS OF THE LASALLE UNIT 2 NU. MILLER.L.A.; et al Sandia National Laboratones. May 1993.

CLEAR POWER PLANT- RISK METHODS INTEGRATION AND 200pp. 9306210243. SAND 92-2765. 75408.056.

EVALUATION PROGRAM Parameter Estimation Analysis And This volume contains a desenption of the codes and input /

Screening Human Reliability Analysis. WHEELER,T.A.; output files used to perform the LaSalle Level 11/111 Probabilistic SWAIN.A Da LAMBRIGHT.J A.; et al. Sandia National Laborato- Risk Assessment. A chart showing the process flow is present-nes. March 1993. 208pp. 9304190161. SAND 92 0537. ed and the relationship between the codes and the needed 74643 007_ input and output data is discussed Code listings for codes not This volume desenbos the methodologies used in the data documented otsewhere and complete or sample listings of the analysis, the screening human error analysis, and the common input and output files are also presented.

mode human error analysis performed rn support of the LaSalle PRA. Selected results are presented in this volume. The remain- NUREG/CR-5305 V02 P2: INTEGRATED RISK ASSESSMENT

• der of the results are presented in other volumes of this report FOR THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Risk Uncertainty Evaluation Pro- i where they are actualty used The data review process used in gram (PRUEP). Appendices D-G. BROWN.T.D.; PAYNE A.C.; I the determination of the data used for the initial screenir.g anal- MILLER.LA.; et al. Sandia National Laboratories May 1993.  ;

ysis is descnbed and the final screening data base is given. The 393pp. 9306210303. SAND 92-2765. 75404:246. '

final data selection process is descnbed and the final data dis' See NUREG/CR-5305,V02.P1 abstract. f inbutions are presented. The actual implementation of the data +

n base for the integrated accident sequence quantification is de- NUREG/CR-5410: STATISTICALLY BASED REEVALUATION OF scnbed in Volume 2 of this report on integrated Ouantification PISC-Il ROUND . ROBIN TEST DATA. HEASLER.P.G ;  ;

and Uncertainty Analysis. Several new methods developed for TAYLOR,T.T.; DOCTOR,S R. Battelle Memorial Institute, Pacific  !

, use sn analyzing both pre- and post- accident human errors for Northwest Laboratory. May 1993.1 ~ :pp. 9306020025. PNL-  !

the initial screening analysis are described. Most of the actual 8577. 75082:001. I results are given in other volumes of this report under the ap- This report presents a re-analysis of an international PISC-Il  !

propriate sub analysis desenptions. A method for determining round-robin inspection results using formal statistical techniques I procedural common mode analysis is descobod and the results to account for experimontal error. The analysis examines: U.S. -

presented team performance vs. Other participants performance; flaw l sizing performance and errors associated with flaw sizing; fac-  ;

NUREG/CR-4832 V09: ANALYSIS OF THE LASALLE UNIT 2 NU' tors influencing flaw detection probability; and performance of I CLEAR POWER PLANT: RISK METHODS INTEGRATION AND all participants with respect to recently developed ASME Sec- '

EVALUATION PROGRAM (RMIEP) Intemal Fire Analysis. tion XI flaw detection performance demonstration requirements, I LAM 8RIGHT,J.A.; BROSSEAU,D.A.; PAYNE,A.C.; et al. Sandia and develops conclusions concerning ultrasonic inspection ca-Nationa' Laboratones March 1993. 600pp. 9304260106. pability.

SAND 92-0537. 74711:001 This report is a desenption of the internal fire analysis per- NUREG/CR-5471: ENHANCEMENTS TO DATA COLLECTION formod on the LaSalle County Nuclear Generating Station Unit AND REPORTING OF SINGLE AND MULTIPLE FAILURE 2 As part of this effort, a new data base for fires was construct- EVENTS. WHITEHEAD.D.W. Sandia National Laboratories.

ed (NUREG/CR-4586). This data base aided in quantification of PAULA,H M. JBF Associates, Inc. PARRY,G W.; et al. NUS fire initiating event frequencies. The most detailed integration Corp. March 1993. 149pp. 9304190151. SAND 89 2562.

between fire nsk asse'ssment and intemal events analysis, to 74622:204.

date, was also accomplished The same system fault trees used This document presents recommendations on how the collec-for internal events were utilized for the fire analysis, which in- tion and documentation of failure events at nuclear power cluded modeling of components down to the contact pair level. plants can be improved. These recommendations, if adopted, Subsidiary equations were created to map the effects of cable should enhance the reliability improvement and nsk assessment j failures and sp inous actuations. All component and associated programs that are dependent on such information. The report cable locations were traced and mapped into the fault trees. A concentrates on how the recommendations should provide the detailed screening analysis was performed which showed most information necessary to improve the parameter estimations for plant areas had a negligible contnhution to fire-induced core both independent and dependent events in a probabilistic risk i

2 1

i i

! 6 Main Citations and Abstracts a" 'ssment and alludes to the fact that this same information Detailed guidehnes are provided for Phase til to aid the analyst {

ct . De used to enhance other nuclear power plant activities. In using this quahtative information and generic data in develop-Several ensting data bases are reviewed and areas where infor- ing a plant-specific CCF base. Depending on the overall objec- ,

mation is lacbng. either because certain information is not re- tive of the study, CCF analysis can stop at the end of any of the j cured to be reported or because required information was three phases. l wmply not reported, are identified. Finally, data needs identified from recent PRAs are discussed NUREG/CR-5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE f RESEARCH AT CNWRA. Calendar Year 1991. ABABOU,R; j NUREG/CR-5759: RISK ANALYSIS OF HIGHLY COMBUSTIBLE BAGTZOGLOU A.C.; CHOWDHURY,A H.; et al. Center for Nu- ,

GAS STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN clear Waste Regulatory Analyses. May 1993 500pp. j PRESSURIZED WATER REACTOR PLANTS. SIM1ON.G P. Sci- 9306290022. CNWRA 91-01 A. 75494:144-  ;

ence Applications International Corp. (formerly Science Applica- This is an annual status report on the results of research con- p tions. Inc ). VANHORN.R.L; SMITH.C.L.; et al. EGSG idaho. ducted on behalf of the U.S. NRC by the Center for Nuclear  ;

inc June 1993. 250pp. 9306290067. EGG-2640. 75495:243. Wasto Regulatory Analyses in support of activities under the l

. This report presents the evaluation of the potential safety Nuclear Waste Policy Act, as amended. Nine specific projects l concems for pressunzed water reactors (PWRs) identified in are underway; eight of which are reported here. The Geochem-Genenc Safety issue 106. Piping and the Use of Highly Com- istry project is using laboratory methods and computer calcula- l bustibie Gases in Vital Areas. A Westinghouse four-loop PWR tions to assess key geochemical constraints and to evaluate  !

, plant was analyzed for the nsk due to the use of Combustible $0rpt've properties of Zeolites present at the proposed reposi-gases (predominantly hydrogen) within the plant. The analysis tory site. The Thermohydrology project has as its focus im- }

evaluated an actual hydrogen distribution configuration and con- proved understanding of heat and fluid flow in unsaturated ducted several sensitivity studies to determine the potential vari- media Laboratory, field, and calculational studies are combined  ;

abikty among PWRs. The sensitwity studies were based on hy. n the Seismic Rock Mechanics project to examine the effects l drogen and safety-related equipment configurations observed at of repeated seismic loadings on the rock-mechanical and hydro-other PWRs within the United States. Several options for im- ;ogical responses of rock masses. The integrated Waste Pack- l I

proving the hydrogen distnbution system design were identified age Expenments have been initiated to evaluate degradation and evaluated for their effect on nsk and core damage frequen- modes of candidate waste container alloys. Three-dimensional  !

cy. A cost / benefit analysis was performed to determine whether computer analysis techniques are being used to investigate spa-atternatives consioered were justifiable based on the safety im- t al vanabihty of flow and transport in variably saturated frac-provement and economics of each poss'ble improvement. tured porous media in the Stochastic Flow and Transport j NUREG/CR-5776: DAMPING IN LOW-ASPECT. project. The recently initiated Geochemical Analogs project RATIO, REINFORCED CONCRETE SHEAR WALLS. seeks to inveshgate the role of such analogs in the licensing ,

F ARR AR.C R Los Alamos National Laboratory BAKER.W E. process, and is currently focused on charactenzing and evaluat-  !

ing a potential site for investigation. The Sorption Modehng  !

l New Mexico. Univ. of. Aibuquerque, NM. May 1993. 83pp.

93062103:s7. LA-12201-MS. 75403:242.

Project has as its objectwe the evaluation and eventual selec- i This report summanzes the information obtained from static tion of model(s) of sorption processes which are deemed tech-nically acceptable in the context of repository licensing. Finally,  ;

and dynamic tests of scale-model Seismic Category 1 structures

$ (exclusive of containment) on the damping of low-aspect-ratio. the Performance Assessment project is directed toward devel- j reinforced concrete shear walls. The report reviews expenmen. oping and evaluating methodologies for evaluation of the long- l 1 tal assessments of damping in low-aspect-ratio shear walls that term performance of the proposed repository. p have been reported in the hterature and presents a summary of NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE the types of structures and structural elements tested It dis- WASTE RESEARCH AT C4WRA. January-Jurie 1992.

cusses the testing methods and the methods used to determine ABABOU,R; AHOLA.M.; BACA.R.; et al. Center for Nuclear l equrvalent viscous damping ratios (both directly and indirectly). Waste Regulatory Analyses. May 1993.105pp. 9306210312.

a numental study that examines the accuracy of vanous meth- CNWRA 92-01S. 75404:022.

ods for estimating damping from measured acceleraton input This is a semi-annual status report on the results of research  !

and response data, and tabulates the damping results. The conducted on behalf of the U.S. Nuclear Regulatory Commis- ,

report concludes by graphically showing the changes in the soon by the Center for Nuclear Waste Regulatory Analyses in damping of the shear walls as a function of the peak nominal support of activities under the Nuclear Waste Pohey Act, as l 1

base shear stress expenenced by the structure dunng simulated amended. Nine specific projects are under way as reported I seismic events Also included are comparisons of the damping '

he Geohm P@ sM s q Wram 6 results obtained in this program with those obtained by other in- ods and computer calculations to assess key geochemical con- i

, vestigators. straints and to evaluate sorptive properties of zeolites present  !

NUREG/CR-5801: PROCEDURE FOR ANALYSIS OF COMMON- at the proposed repository site. The Thermo'iydrology Project  !

j CAUSE FAILURES IN PROBABILISTIC SAFETY ANALYSIS has as its focus improved understanding of heat and fluid flow  !

MOSLEH.A. Maryland. Univ. of. College Park, MD

• Sandia Na- en unsaturated media. Laboratory, field, and calculational studies  !

are combined in the Seismic Rock Mechanics Project to exam-tional Laboratones Apnl 1993 51pp 9306010342. SAND 91-  !

l 7087. 75062.001. ine the efIects of repeated seismic loadings on the rock-me-This report provides practical guidelines for treatment of chanical and hydrological responses of rock masses. The Inte-  !

common-cause failures (CCF) in nsk and reliability studies The grated Waste Package Expenments have been initiated to  ;

procedures outhned in this report are organized according to evaluate degradation modes of candidate waste container  ;

three phases of analysts, screening analysis, detatied quaktatwe alloys. Three-dimensional computer analyses techniques are analysts. and detatied quantitatwe analysis. The results of the being used to investigate spatial variabihty of flow and transport screening analysis phase include conservatwe identificaton of in vanably saturated fractured porous media in the Stochastic i potential common cause vulnerabihties and determination of the Flow and Transport Project. The Geochemical Analogs Project {

scope and 'ocus for more detailed analysis in Phases il and til. staff seeks to investigate the role of such analogs in the licens- l ing process, and is currently focused on charactenzing and j 4 Phase II, the detailed qualitatwe analysis, provides a better un-derstanding of the plant specific susceptibikties of the systems evaluating two potential sites for investigatiott The Sorpton j and components to causes and couphng mechanisms of CCF. Modehng Project has as its objectwo the evaluation and eventu- i The infor~ation from this phase can then be used as a basis al selection of model(s) of sorptori processes which are j

' for a pla apeClfiC quantitative assessment of CCF frequencies. deemed technically acceptable in the context of repository h- l l

i I

Main Citations and Abstracts 7 censing. The Pertormance Assessment Project is directed egonzation of four PWR groups based upon their porreivod late toward developing and evaluating methodologies for evaluation depressunzation capabihty. In this report, a PWR representative of the long term performance of the proposed repository Final- of each of the four PWR groups was chosen for detaded analy-

)y, in the recently initiated Volcanism Protect, a compretensive us of its capabihty to intentionally depressunze employing the evaluation of tre state of knowledge regarding volcarnm in the late depressunzation strategy. The phenomenological behavior, basin and range province has been completed. hardware performance, and operational performance of these NUREG/CR-5882: TRAC-B THERMAL-HYDRAULIC ANALYSIS PWRs dunng the intentional depressunzation strategy were con-OF THE BLACK FOX Dolt LNG WATER RfACTOR sid r d. The phenomenological behavior was analy7ed using ,

MARTIN.R P. EG&G Idaho, Inc May 1993. 68pp. 9306180322l the SCDAP/RELAPSIMOD3 severe accident analysis code. The

{

EGG-2677. 75387:169 results of these evaluations were then extended to the remain-  :

Thermal-hydraukc analyses of 9x hypothetical accident sco, ing s compnwng es@ M group naros for the General Electoc Black Fox Nuclear Project boikng water reactor were performed using the TRAC-BF1 computer NUREG/CR-5955: MATERIALS AND DESIGN BASES ISSUES IN l ASME CODE CASE N-4 7. HUDDLESTON.R L; 1 code This work is sponsored by the U S Nuclear RegulatoW f Commission and is t>eing done in conjunction with future analy' SWINDEMANRW Oak Ridge National Laboratory Apnl 1993.

sis work at the U.S Nuclear Regulatory Commission Technical 42pp 9306010335. ORNL/TM 122t>6 75062:284. I Training Center in Chattanooga. Tennessee. These accident A prekminary evaluation of the design bases (pnncipalty scenanos were chosen to assess and benchmark the thermal- ASME Code Case N-47) was conducted for design and oper-hydraulic capabilities of the Black Fox Nuclear Project simulator aton of reactors at elevated temperatures where the time-de-at the Techrucal Training Center to model abnormal transient pendent effects of creep. creep-fatigue, and creep ratchetinq  !

conditions are significant A'eas where Code rules or regulatory guides  !

rnay be lacking or inadequate to ensure the operation over the [

NURE G/CR-5894: RADIONUCLIF'E CHARACTERIZATION OF expected life cycles for the next generation advanced high tem-  !

REACTOR DECOMMISSIONING WASTE AND NEUIRON AC. perature reactor systems, with designs to bo certiied f by tho '

T!VATED METALS HOBERTSON.D Ea THOMAS,C W.; U S. Nuclear Regulatory Commission, have been identtfied as i WYNHOF F,N L.; et i.i Battelle Memonal Instituto. Pacific North. unresotved nsues Twonty-two unresolved issues were idente

• wust t.aDoratory M e 1993 80pp. 9307060129 PNL 8106 fied and bnef scoping plans developed for resolving these  ;

7 % 72213 issues Thm study is preading ste NRC and hcensees with a mote comprehensivo Md defenuble data t>ase and regulatory as- NUREG/CR-5957: SYSTEM 80 4 (TM) CONT AtNMENT - STRUC- i sessment of the radiological factors associated with reactor do. TURAL DESIGN REVIEW GREIMANN.L, FANOUS.F4 3 commissioning and disposal of wastes generated dunng these CHALLA.R ; et at. Iowa State Univ , Ames. IA May 1993 112pp l activities. The objectives of this study are bmng accomphshed 9306010330 15-5083 75063 001. i dunng a two-phase sampling. measurement, and assessment A review of the structural descgn of the Combustion Engineer.  !

program involving the actual decommissioning of Shippingport eng (CE) System 80 4 (TM) stoel contsnment was completed  :

Staten and the detailed analysis of neutron-activated matonals The stress analysis and the evaluation of the structure against buckling were performed by using BOSOR4 and BOSOR5 finite {

trom commercial reactors. The radiological charactentation ,

studies of Shippingport decommissioning matenals have now difference software, respectuely. The CE System 80 + (TM) con- j been completed, and analyses of dismantled ptoang and scab- twnment was modelled as an axsymmetnc shell consisting of  ;

blod concrete have shown that neutron actw ion products, different segments and mesh points with the additional mass of j dominated by (60)Co, compnsed the residuat radionuclido inverg the penetratsons and appurtenance being smeared around the i tory. Fission products and transuranic radonuchdes were essun. cam %e The transition region was modelled using elastic l tsally abbent. Waste classification assessments have shown that springs with a foundation modulus of 180 lbshn(3) The (

all decommissoning matenals (except reactor pressure vessel stresses due to the individual loads (dead loads, intemal and '

internak) could be disposed of as Class A waste. Spent fuet eAnal pressures and temperatures) were computed using the disassembly hardware from the Shippingport Gore-3 was ana. stress analysts option in the BOSOR4 program. The stresses l

lyzed for long-hved activation products specified in 10 CFR 61, from mdividual loads were combined according to ASME Code  !

and the hardware was etassified with respect to 10 CFR 61 'nto stress intensities Service Levet B loadings produced a 20  ?

waste disposal rules Niobium-94 and (63)Ni concentrations in percent over-stress in a small zone just above the transition j inconel-X750 and stainless steel components cuteeded their reg:on. All other stress intensities were within allowable hmits.  ;

Class C hmits. For the System B04 (TM), the perfect shell with an elastic mate- ,

nal was initialty analyzed The calculated factor of safety values HUREG/CR-5937: INT ENTIONAL DEPRESSURIZATION ACCl- were 2.3 (Level B) and 1.59 (Levels C and D) Finally, senutivtry i DENT MANAGEMENT STRATEGY FOR PRESSURflED studies were conducted to investrgate the efiocts of mesh size WATER REAClORS BROWNSON.D AJ. HANEY,L N ;  !

and transition zone stiffness on the controthng buckhng load.

CHIEN,N D. EG&G Idaho, Inc. Apnl 1993.165pp 9305100005. f EGG-2688. 74858:270. NUREG/CR-5966: A SIMPLIFIED MODEL OF AEROSOL RE-in a previous investigation of the Surry nuclear power station, h MOVAL BY CONTAINMENT SPRAYS POWERS.D A. Sandia 11 was concluded that intentional depressunzation of the reactor Nationat LaboratorKrs. BURSON S B. Severe Accidnnt issuns  ;

coolant system (RCS) could prevent of mitigate the effects of Brancti June 1993. 180pp. 9306210236. SAND 32 2689 5 direct contmnment heating (DCH) during a station blackout tran- 75407:095.

sient. Two strategies, early and lato depressurizat on, were in- l Spray system in nuclear reactor containment 9 are descnbed.

vestigated as methods to mitigate DCH The investigaton con. The scrubbing of aerosols from containment atmospheres by  ;

cluded that since there are greater opportunities to recover spray droplets is discussed. Uncertainties are ident!fted in the f plant functons before core damage occurs and operator re- prediction of spray performance when the sprays are used as a i sponse uncertainties are lest.ened, the strategy of late depres- means for decontaminating containment atmospheres. A sunzation is preferred over early depressunration. The resutts of f mechanistic model based on current knowledge of the physical }

the Sorry analysis were extended to other U.S pressurned phenomena involved in spray r,arformance is developed. With water reactors (PWRs) in order to evaluate their capabihty to this model, a quantitative uncertenty analysis of spray perform-successtutty employ the late depressunraton strategy to pre- {

ance is conducted using a Monte Carlo method to sample 20 i Vent or mitigate DCH By applying appropriate scahng 18Ctois to uncertain Quantities related to phenomena of spray droplet be-the sowieted key parameters, this evaluation resutted in tne cai l havior as well as the initial and boundary conditons expected to l i

f i

I

l 1

i 8 Main Citations and Abstracts be associated with severe reactor accidents. Results of the un- selected to demonstrate the feasibility of developing train-level certainty analysis are used to construct simplified expressions databases. Five different methods for developing train-level da-for spray decontamination coefficients. Two vanables that affect tabases were hypothesized and are examined. Ultimately, two aerosol capture by water droplets are not treated as uncertain; train-level databases were developed using the Peach Bottom they are (1) 'O', spray water flux into the containment, and (2) Unit 2 PRA and one train-level database was developed using

'H; the total fall distance of spray droplets. The choice of the Beaver Valley Unit 2 iPE. The development, use, hmitations, values of these vanables is left to the user since they are plant and results of these train-level databases are discussed.

and accident specific. Also, they can usually be ascertained with .

Some degree of certainty. The spray decontamination coeffi- MANAGEMENT

• cients are found to be sufficiently dependent on the extent of SPRAYS kN CONTAINMENT NOURBAKHSH,H.P.; PEREZ,S E.; LEHNER.J.R. Brookhaven decontamination that the fraction of the initial aerosol remaining in the atmosphere, m(f), is explicitly treated in the simphfied ex.

National Laboratory. May 1991 54pp. 9306180261. BNL- l pressions. NUREG-52354. 75427:109.  !

A limited study has been performed assessing the effective-NUREG/CR-5972; EFFECTS OF NONSTANDARD HEAT TREAT- ness of containment sprays to mitigate particular challenges i

MENT TEMPERATURES ON TENSILE AND CHARPY IMPACT which may occur during a severe accident. Certain aspects of ,

PROPERTIES OF CARBON-STEEL CASTING REPAIR WELDS. three specific topics related to using sprays under severe acci-NANSTAD R.K.; GOODWIN,G M.; SWINDEMAN,M.J. Oak Ridge dent conditions were investigated. The first was the effective-  !

National Laboratory. April 1993.118pp. 9304210258. ORNL/ ness of sprays connected to an alternate water supply and

, TM-12280. 74677.192. pumping source because the actual containment spray pumps i

! Carbon steel castings are used for a number of different com- are inoperable. This situation could occur during a station black.  !

ponents in nuclear power plants. including valve bodies and out. Tne second topic concerned the adverse as well as benefi- F bonnets. Components are often repaired by welding processes, cial effects of using containment sprays dunng severe accident j j and both welded components and the repair welds are subject" scenanos where the containment atmosphere contains substan- i' ed to a vanety of postweld heat treatments (PWHT) wrth tem- tial quantities of hydrogen along with stearn. The third topic was i peratures as high as 899 degrees C (1650 degrees F), well the feasibility of using containment sprays to moderate the con- s above the normal 593 to 677 degrees C (1100 to 1250 degrees sequences of DCH. .

F) temperature range. The temperatures noted are above the i A1 transformation temperature for the matenals used for these NUREG/CR 5983: SAFETY ASPECTS OF FORCFC ROW components. A test program was conducted to investigate the COOLDOWN TRANSIENTS IN MODULAR HIGH TEMPERA-  !

potential effects of such " nonstandard" PWHTs on mechanical TURE GAS-COOLED REACTORS. KROEGER P.G. Brookhaven l National Laboratory. May 1993. 24pp. 9306010323. BNL-properties of carbon steet casting welds. Four weldments were fabncated, two each with the shielded-metal-arc (SMA) and flux- NUREG-52355. 75058.297.

cored-arc (FCA) pr: cesses, with a high-carbon and low-carbon Dunng some of the design basis accidents in Modular High

' j filler metal in each case. All four welds were sectioned and Temperature Gas Cooled Reactors (MHTGRs), the main Heat grven simulated PWHTs at temperatures from 621 to 899 de- Transport System (HTS) and the Shutdown Coohng System ,

I 4

grees C (1150 to 1650 degrees F)in increments of 56 degrees ;SCS) are assumed to have failed. Decay heat is then removed C (100 degrees F) and for times of 5,10,20, and 40 h at each by the passive Reactor Cavity Cooling System (RCCS) only. If i temperature. Hardness, tensile, and Charpy V-notch (CVN) either forced flow coohng system becomes available dunng  ;

impact tests were conducted for the as-welded and heat-treated such a transient, its restart could significantly reduce the down- j conditions. Results were plotted versus a time-temperature rela- time. This report used the THATCH code to examine whether -

4 tionship (tempenng parameter) to enable a more direct compan- such restart, during a penod of elevated core temperatures, can 3

son of the effects of the vanous PWHT conditions Heat treat- be accomphshed within safe hmrts for fuel and metal compo- ,

ments at 621 and 677 degrees C (1150 and 1250 degrees F) nent temperatures. If the reactor is scrammed, either system  !

gave resutts amenable to prediction, and regression analyses can apparently be restarted at any time, without exceeding any l are presented for those conditions. Heat treatments at 732 to safe hmits. However, under unscrammed conditions a restart of B99 degrees C (1350 to 1650 degrees F), however, resulted in forced coohng can lead to reenticality, with fuel and metal tem- i substantial changes in mechanical properties of these SMA and peratures significantly exceeding the safety hmits. i FCA welds, with the changes not amenable to prediction and [

4 highly dependent on the weld metal. Heat treatments in that NUREG/CR-5984: CODE AND MODEL EXTENSIONS OF THE i temperature range should not be applied to these matenals THATCH CODE FOR MODULAR HIGH TEMPERATURE GAS-  !

COOLED REACTORS. KROEGER.P.G.; KENNETT,R.J. Brook- t without pnor qualification for the mtended use.

haven National Laboretory. May I993. 46pp. 9306010318. BNL-

• NUREG/CR-5976: DEVELOPMENT AND USE OF A TRAIN- NUREG-52356. 75062.166.

LEVEL PROBABILISTIC RtSK ASSESSMENT. SMITH C.L.; This report documents several model extensions and im- ,

FOWLER.R.D.; WOLFRAM.L M. EG&G Idaho, Inc. April 1993. provements of the THATCH code, a code to model thermal and ,

74pp. 9306010327. EGG 2694. 75062:209. fluid flow transients in High Temperature Gas-Cooled Reactors.

The Idaho National Engineenng Laboratory examined the po- A heat exchanger model was added, which can be used to rep-tential for the development of train-level probabilistic risk as- resent the steam generator of the main Heat Transport System .

sessment (PRA) databases These train-level databases will or the auxiliary Shutdown Coohng System. This addition permits  !

allow the Nuclear Regulatory Commission to investigate effects the modehng of forced flow cooldown transients with the f on plant core damage frequency (CDF) given a train is failed or THATCH code. An enhanced upper head model, considering j taken out of service. The intent of this task was to develop the actual conical and spherical shape of the upper plenum and i user-fnendly databases that required a minimal amount of per. reactor upper head was added, permitting more accurate mod-  !

sonnel involvement to be usable. It was onginally intended that elling of the heat transfer in this region. The revised models are :

the train-level models would not be expanded to include basic described, and the changes and addition to the input records  !

events below the top gate of a train, with the possible exception are documented. I of including some of the trajor train-related components (e g.,  !

. important pumps and motor-operated valves). ft was found that NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES 1

! a database similar to the unginal plant PRA provided the accu. OF LARGE-SCALE INTERNATIONAL REFERENCE EXPERI-

' racy needed to measure the changes in plant CDF. The Peach MENTS (PROJECT FALSIRE). BASS,B.R.; PUGH.C.E.;

a Bottom Unit 2 NUREG-1150 PRA (a large fautt tree model) and KEENEY WALKER,J, et al. Oak Rsdge National Laboratory.  ;

the Beaver Valley Unit 2 iPE (a large event tree model) were June 1993.150pp. 9307020010. ORNL/TM-12307. 75585:204.  !

i

Main Citations and Abstracts 9 4

This report summanzes the recently completed Phase I of the NUREG/CR-6014: HIGH PRESSURE COOLANT INJECTION Project for Fracture Anarysis of Large-Scale international Refer- SYSTEM RISK-BASED INSPECTION GUIDE FOR HATCH NU-ence Expenments (Project FALSIRE). Project FALSIRE was cre- CLEAR POWER STATION. DIBIASIO.A M. Brookhaven National ated by the Fracture Assessment Group (FAG) of Pnncipal Laboratory May 1993. 57pp. 9306110025, BNL-NUREG-52367.

Working Group No. 3 (PWG/3) of the Organization for Econom- 75336:303.

sc Cooperation and Development (OECD)/ Nuclear Energy Agen.

ey s (NEA's) Committee on the Safety of Nuclear installations A review of the operating expenence for the High Pressure Coolant injection (HPCI) system at the Hatch Nuclear Power (CSNI) Motivation for the protect was denved from recognition by the CSNi-PWG/3 that inconsistencies were being revealed in Station, Unsts 1 and 2, is desenbed in this report. The informa-predictive capabihties of a vanety of fracture assessment meth- tion for this review was obtained trorn Hatch Licensee Event I ods, especially in ductile fracture applications. As a conse- Reports (LERs) that were generated between 1980 and 1992.  !

Wse LERs han been categow.in. to 23 fahre mes M quence, the CSNI/ FAG was formed to evaluate fracture predic-i tion capabihties currently used in safety assessments of nuclear have been pnoritized based on r .babikstic nsk assessment  ;

j considerations. In addition, the results of the Hatch operating components. Members are from laboratones and research orga.  !

nizations in Western Europe, Japan, and the United States of exponence review have been compared with the results of a l

Amenca (USA) On behalf of the CSN1/ FAG. the U.S. Nuclear similar, industry wide operating expenonce review. This compan- p Regulatory Commission's (NRC's) Heavy-Section Steet Technob son provides an indication of areas in the Hatch HPCI system  ?

ogy (HSST) Program at the Oak Ridge National Laboratory that should be given increased attenteon in the pnontization of (ORNL) and the Gesellschaft for Anlagen-und Reaktorsicher. inspection resources.

heit (GRS). Koln, Federal Republic of Germany (FRG) had re- t sponsibility for organization arrangements related to Project NUREG/CR-6018: SURVEY AND ASSESSMENT OF CONVEN.

FALSIRE. The group is chaired by H Schutz from GRS, Koln, l llONAL SOFTWARE VERIFICATION AND VALIDATION METH-  ;

FRG. ODS, MILLERLA.; GROUNDWATER.E.; MIRSKY.S.M; et al.  !

IdUREG/CR-5999: INTERIM FATICUE DESIGN CURVES FOR Science Apphcations international Corp. (formerly Science Ap- l CARBON, LOW-AL LOY, AND AUSTENITIC STAINLESS phcations, Inc.). Apnl 1993 186pp. 9305100042. EPRI TR.  ;

STEELS IN LWR ENVIRONMENTS. MAJUMDAR.Sa 102106. 74858 001.  ;

CHOPRA.O K4 SHACK.W J Argonne National Laboratory Apnl This report documents the results of the first (of ten) tasks l 1993. 34pp 9305100010. ANL-93/3. 74858.189_ being performed under a contract jointly funded by the USNRC ,!

Existing data in the literature on fatigue of carbon, k walloy, and EPR to develop and document guidehnes for the verifica- [

d and austenitic stainless steels in LWR environments ce re- tion and vahdation of expert systems in the nuclear industry.  ;

viewed it is found that both temperature and dissolved-ox ' gen This task conducted an extensive survey of conventional soft-concentiation in water significantly affect fatigue hie. At the 'ery ware venfication and validation (V8V) methods. A total of 134 low d$ solved-orygon levels charactenstic of pressunred w ter different methods were identified which can be apphed to either a teactors and boshng water reactors with hydrogen-water cher ns- the requirements-design of implementation phases of software.

try, environmental effects un fatigue hfe are modest. Howe ver.

at higher dissolved. oxygen levels (2100 ppb) significant r( duc-These methods were classified by a sequential hfecycle model, .

tions in fatigue life can occur. The susceptibehty of carbon anc low' charactenred by factors of power and ease-of-use, and as-

, alloy steels to reduced fatigue hie is strongly related to sessed according to their appkcability to expert systems. Expert  !

sulfur concentration Although the fatigue hves of austenstic systems were decomposed into four components: knowledge i stainless steels may be reduced, the reductions are much base, inference engine, interfaces, and tools / utilities. The con-smaller than those observed in high-sulfur carbon and low alloy ventional software V&V methods were found to be directly or, steels in ovygenated water, fatigue hfe depends strongly on by extension, apphcable to all of the expert system techniques strain rate. Intenm tatigue design curves are proposed that take except the knowledge base. ,

into account temperature, dissolved oxygen level in the water the sulfur level in the steel and strain rate Design curves for' NUREG/CR-602h PRELIMINARY EVALUATION OF SNUBBER .

on and low ahoy steels for hves up to 10(8) cycles are als SINGLE FAILURES. WARE,A.GJ BLANDFORD,R.K/

KELLY.D L.; et al. EG&G Idaho. Ir$c. April 1993. 49pp 9305100059. EGG-2697. 74857.316.

NUREG/CR-6013: METHODS USED FOR THE TREATMENT OF The United States Nuclear Regulatory Commission developed -

NON-PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS CONOSCENTE J P.; MASLENIKOV.O.R. JOHNSON.JJ ECE FIN L2430. Prehminary Evaluation of Snubber Single Failures, Engineenng Consultants (formerly EOE Engineenng. Inc.) May with the objective of performing a prehminary evaluation of the 1993. 62pp. 9306210371 75402:286 safety imphcation of a potential single failure of a snubber used  !

Non-proportional or norectassical damping is defined as a to support safety-related piping or equipment. The Idaho Nation-torm of viscous darnping that introduces couphng between the al Engineering Laboratory staff conducted a quahtative review of  ;

undamped modal coordinates of motion. Such problems have a large number of hght water reactor systems, ard a pantita-practical apphcations in the dynamic analysis of soil-structure trve stress analysis of four systems. A candidate hst was deveh  !

systems, structure-equipment systems, and structural systems opod that ranked the systems as having a high, medium, or low  !

made of matenals with dif'erent energy dissipation capacities. probabihty of causing a significant increase in the core damage  ;

Presented in this report is a review of the methods most com. frequency should a single snubber fail to function. Two systems  !

monly used in structural analysis for the solution of the dynamec were ranked high, a PWR ice condenser main steam contain-response of systems with non-proportional damping. Both ngor. ment penetration and a BWR Mark 1 torus. Stx systems were ous and approximate methods are described. Since ngorous ranked medium and the remaining 30 were ranked as low or  ;

methods usually require large computational efforts, approxL low-to-medium. The two systems ranked high and two systems mate methods using undamped mode shapes are often pro. ranked medium, a PWR ice condenser auxshary feedwater hne ferred. In the study described here, the accuracy of three ap- and a PWR reactor coolant system loop drain kne, were chosen proximate methods was evaluated for three benchmark prob- for a quantitative stress analysis. Of the four systems analyzed, i tems, with vanous parametne vanations. Results were compared only the PWR ice condenser main steam containment penetra-with the exact solution for difierent combinations of structura! tion is judged to be srgnificantly susceptible to the failure of a properties Based on these results conclusions and recommon-single snubber. .

. dations are presented for the use of the selected approximate l l methods. '

4 5

f B

I i I 10 Main Citations and Abstracts. l l

NUREG/CR-6028: B.GHOW. A NUMERICAL CODE FOR S MU- NUREG/CR4035: FEASIBILITY STUDY FOR IMPROVED LATING FLOW IN VARIABLY SATURATED, HETEROGENE- STEADY-STATE INITIALIZATION ALGORITHMS FOR THE OUS CEOLOGIC MEDIA. Theory And Users Manual - Version RELAp5 COMPUTER CODE. PAULSEN M.P.; PETERSON3E.;

11. ABABOU.R. France. BAGTZOGLOUAC. Center for Nucle- KATSMAKR Computer Simulaton & Analys4s. inc. April 1993.

i at Waste Regulatory Analy ses. June 1993.160pp 9307060141- 75pp. 9306010277. 75058 225.

CNWRA 92-026. 75S72 290 A design for a new steady-state initialization method is pre-This report documents BIGFLOW 1.1. a numencal code for sented that represents an improvement over the current method simulating flow in variably saturated heterogeneous geologic used in RELAPS. Current initiahzation methods for RELAP5 med6a. It contains the underlying mathematical and numencal solve the transient fluid flow balance equatons simulatrng a models, test proolems, benchmass, and apphcations of the transient to achieve steady state conditions. Because the tran-BIGFLOW code. The BIGFLOW software package is composed s>ent solution is used, the initial cond:tsons may change from the j of a simulation and an :nteractive data processing code (DATA. desired values requinng the use of controllers and long tran- l FLOW) The simulat or, code solves knear and nonhnear porous sient running imes to obtain steady-state conditions for system rned;a flow equatons bised on Darcy's law, appropnatefy gen- problems. The new initialization method allows the user to fix l erahzed to account for 3D. deterministic, of random heterogene- thermal-hydrauhc values in volumes and lunctions where the ity A mod:f ed Picard Scheme is used for bnearmng unsaturated conditons are best known and have the code compute tne ini-flow equations, and preconditioned iterative methods are used taal conditions in other areas of the system The steady-state .

J for solving the resulteng maton systems. The data processor balance equations and solution methods are presented The

! DATAFLOW) allows interactive data entry, manipulation, and const:tutrve, component, and special purpose models are re- .

analysis of 3D datasets The report contaos analyses of com- y,ewed with respect to mod 6catons required for the new {

putatonal performance camed out using Csay-2 and Cray-Y/ steady-state atiahzation method. The requirements for user '

MPB supercomputers. Benchmark tests include compansons input are deftned and the feasibility of the method is demon-with other independently developed codes, such as PORFLOW strated with a testbed code by initializing some simple channel l l and CMVSFS. and with anaYticat or sembanalytical solutions problems. The ntialization of the sample problems using the I

' old and the new methods are compared. {

NUREG/CR-6031: CAvlT ATION GU:DE FOR CONTROL VALVES TULU5.J p Tums Engineenng Consultants. Apnl NUREG/CR-6036: INITIAL RESULTS OF THE INFLUENCE OF P BIAX1AL LOADING ON FRACTURE TOUGHNESS. THEtSS.T.J ; ,

de e f ndamentais of cavitaton to G d at Oak Ridge National Laborato-l provde tne reader with an understanding of what causes cavita-MsA ry June 1993 94pp 9306290012. ORNL/TM-12349.

ton when it occurs and the potential problems cavrtaten can 75494 056 ,

cause to a valve and piping system, The document provides

~

A msung Nam to name m hn d bad bad m gudehnes for understanding how to reduce the cavitaten and/ the fracture toughness of shallow-flaw specirnens under conds or select control va!ves for a cavitating system The guide pro- lions prototypic of a reactor pressure vessel was begun. Exist-vides a method for predicting the intens,ty of control valves and sng data suggest that shallow-flaw spec: mens under biaxial load-how the effect of cavitation on a system will vaqr with valve ,

ing will exhibit a toughness reduction compared to comparable j type, wa!ve size, vaive function operating pressure, duration of unianial specimens. Quantification of this toughness reducton is  ;

operaton and details of the piping mstallation. The gu de de. the main goal of the biaxiat fiacture toughness program. A cru- j fines six cavitation hmits dentifying cavitation intensitees rangmg ciform specimen with a two-dimensional shahow through-thick-from inception to the maximum intensity possible The intensity 7 of the cavitaton at each hmit is desenbed including a bnef dis. ness flaw under a biaxial load fatto of 0.6:1 was used for bearial  ;

tracture toughness testing. The entical fracture load for each j cusson of how each level of cavitation influences the vatve and specimen was approximately the same, but the unsaxial spec 9 j system Enamples are included to demonstrate how to apply the men withstood substantially more determaton at failure than did p method, including making both size and pressu'e scale eMects tne biamiat specimens. Three-dimensional, elastic plastic, finite-

' correct.ons Methods of contro!bng cavitat on are discussed pro, element posttest analyses were necessary to estimate fracture ,

vding information on vanous techniques which can be used to toughness in all cases, agreement between the measured and deseqn a new system or moddy an existing one so it can oper. computed load vs deformation responses was excellent. Tough-ate at a desired level of cavitat<on ness values for the cruciform specimens were compared with  !

2 NilREG/CR4032: SOLIDUS AND LIQUIDUS TEMPERATURES data from prevously tested. deep- and shallow-crack specs-vF CORE CONCRETE MIXTURES ROCHE.M F ; mens Resu".s from these tests indicate that the shallow-crack

• LEIBOWITZ La FINK.J K : et al Argonne Nat ona! Laboratory toughnesit increase is part: ally. but not totally, removed by the June 1993 E5pp 9306290008 ANL-93/9 75494 001. application of biarial loading. However, additonal data are re- r Sohdus and hauidus temperatures were measured by a com- quired to sohdify those conclusions. A proposed test matrix for  ;

bination of d:ffment6ai thermal analysts and rotatonat viscometry additonal uniaxial and biaxial testing is described This report i for four types of concrete (kmestone, hmestone sand, basatt. has been designated HSST Report No.13B  !

and mbceous) and for their mixtures with urania and 2irconia.

The rneasured sohdus temperatures for the urania-zirconia-con- NUREG/CR404t DISPOSAL UN!T SOURCE TERM (DUST) crete mixtur es were segnificantly lower (hundreds of degrees) DATA INPUT GUIDE. SULLIVAN.T.M Brookhaven Natonal than those employed in the CORCON Mod 2 thermal hydrauhc Laboratory May 1993.92pp 9306180317. BNL-NUREG-52375. ,

I code, and the measured haucus temperatures were sign 6cantly 75388 001.

higher (also hundreds of degrees) The houndus temperatures Performance assessment of a low-level waste (LLW) disposal ,

for urania-zirconia-concrete mixtures containing kmestone or tacihty begins with an estimation of the rate at which radionu- [

umestone4and concrete were general!y above 2850 K, which ckdes migrate out of the facihty (i.e., the source term). The -

was the upper temperature hmit of our expenments The revised focus of trns work is to develop a methodology for calculating T soldus and hquidas temperatures are to be incorporated in the the source term. In general, the source term is influenced by ,

CORCON-Mod 3 thermal hydraule code which is an integral the radionuchde inventory, the wasteforms and containers used {

pa'l of the U.S. Nuclear Regulatory Commisson's MELCOR to dispose of the inventory, and the physical processes that  ;

Code. DTA was also employed to redetermine the calcia-urania lead to release from the facility (fluid flow, container degrada- [

(CaO-UO(2)) phase diagram which is required sn comput(+r pro- ton, wasteform leaching. and rad.onucide transport). The com-  ;

grams that calculate the phase diagrams (and sohdus and haun puter code DUST (Disposa! Unit Source Term) has been devel- j dus temperatures) of urania-zerconie concrete systerns from the oped to model these processes This document presents the  ;

phase diagrams of sempier systems. models used to calculate release from a disposal facshty, vente

[

t e

. . _ - . _ . . -- . - - , - _.?

i J

Main Citations and Abstracts 11 cation of the model, and instructions on the use of the DUST control systems for the control rod, feedwater, steam generator code. In addition to DUST, a preprocessor, DUSTIN, which level, steam dump, pressunter level and pressure are modeled helps the code user create input decks for DUST and a post- to be functoned automatically until the power level decreases processor, GRAFXT, which takes selected output files and plots below 30% nuclear power. A sensitivity study on control rod them on the computer terminal have been wntten. Use of these codes is also desenbed. wortn was carned out and it was found that vanable rod worth should be used to achieve good prediction of neutron power.

NUREG/CR-6061: DETERMINATION OF THE BIAS IN LOFT The results obtained from RELAPS/ MOD 2 simulation agree we!!

FUEL PEAK CLADDING TEMPERATURE DATA FROM THE with the plant operating data and it can be concluded that this BLOWDOWN PHASE OF LARGE-BREAK LOCA EXPERI- code has the capabikty in analyzing the transient of this type in MENTS, BERTA,V.T.; HANSONRG; JOHNSEN.G Wa et al. a best estimate means. l Idaho National Engineenng Laboratory. May 1993 82pp- '

9306210362. EGG-26t0. 75403.140 NUREG/lA-0094: ASSESSMENT OF RELAPS/ MOD 3 AGAINST Data from the Loss-of-Fluid Test (LOFT) Program help quanti-TWENTY-FlVE POST-DRYOUT EXPERIMENTS PERFORMED ty the margin of safety inherent in pressurced water reactors AT THE ROYAL INSTITUTE OF TECHNOLOGY. NILSSON.L dunng postulated loss-of-coolant accidents (LOCAs). This report Swedish Nuclear Power inspectorate (Statens Kamkraftinspek-analyzes how well erlernally-mounted fuel rod cladding surface tion) May 1993. 92pp. 9306110035. STUDSVIK.NS90/93.

thermocouples in LOFT accurately reflected actual cladding 75377:080 temperature dunng large-break LOCA expenments. The analysis Assessment of RELAPS/ MOD 2 has been made against vari-shows that there can be a significant difterence (referred to as ous experimental data, among other data from twenty-five post-bias) between the surface-mounted thermocouple reading and dryout expenments conducted at the Royal institute of Technol-  ;

the actual cladding temperature, and that the magnitude of this ogy (RIT) in Stockholm. As the MOD 3 version of RELAPS has ,

b4as depends on the rate of heat transfer between the fuel rod now been released, incorporating a different method of calculat-cladding and coolant. Further, it is shown that, in terms of peak ing enhcal I, Sat flux compared to RELAPS/ MOD 2, it seemed i cladding temperature recorded dunng LOFT large-break LOCA I 9 R$

l i

expenments. the mean bias is 1142 16.2 K (20.5 1 29 2 de- data The results shosv that the axial dryout position is generally grees F) The best.esumate value of peak cladding temperature better predicted by the MOD 3 than by the MOD 2 version. The for LOFT LP-02-6 is 1104 8 K. The best estimate peak cladding prediction is, however, still nonconservative, te. the calculated I temperature for LOFT LP-LB-1 is 12B4 0 K. dryout position falls in most cases downstream the actual meas- ,

ured point. While the pre-dryout heat transfer seems to be equal NUREG/GR-0006: DEPOSITION: SOFTWARE TO CALCULATE for MOD 2 and MOD 3, both versions giving slightly higher wall PARTICLE PENETRATION THROUGH AEROSOL TRANS- temperatures than the experiments, there is a considerable dif-PORT SYSTEMS Finat Report. ANAND.N K.; terence in the post-dryout heat transfer. The results of the RIT '

MCFARLAND.A R.; WONG.F.Sa et al. Texas A&M Univ., Col-lege Station. TX. Apnl 1993. 45pp 9305100008. 74858.223- data companson indicate that MOD 3 underpredicts the post- [

User-fnendly software (DEPOSITION 2.0) has been devel- dryout wall temperatures remarkably while MOD 2 gave reasona-oped which permits charactenzation of aerosol particle losses in ble agreement. In this respect RELAP/ MOD 3 shows no im- j provement over RELAPS/ MOD 2.

transport systems. The sub-models which compnse the DEPO-  ;-

SITION code are presented and the hmitations of these sub-NUREG/lA-0095: RELAPS ASSESSMENT USING LSTF TEST models are noted. These sub-mode!s have all been previously DATA SB-CL-18 LEE,S; CHUNG,B-D; KIM,H-J. Korea Institute l pubhshed in the peer rev ewed hierature. The software can be of Nuclear Safety, May 1993.100pp. 9306210225. 75402:065.

used to determine the penetration of aerosol through existing 5% cold leg break test, run SB-CL-18, conducted at the

, transport systems; it will provide the optimal tube diameter for a Large Scale Test Facihty (LSTF) was analyzed using RELAPS/

transport system operated at a given flow rate and at a given MOD 2 Cycle 36.04 and RELAPS/ MOD 3 Version Sm5 codes.

l particle size; it will provide a value for the maximum penetration for a transport system that would connect two points in three- The test was conducted with the main objective being the in-vestigation of thermal-bydraulic mechanisms responsible for dimensional space; and, it will provide tables of data and create early core uncovery, including manometric effect due to an output files for parametric studies on the effects of varying parti- asymmetfic coolant holdup in the steam generator upflow and cte size, flow rate and tube diameter. Use of this software for downflow side. The present analysis, camed out with RELAP5/

specific examples is given herewith in an Appendix. Reference MOD 2 and MOD 3 codes, demonstrates the code's capabihty to to this software is included in NRC Regulatory Guide 8.25 predict, with sufficient accuracy, the main phenomena occumng (1992) where at is considered to be an acceptable method for l in the depressurization transient, both from a quahtative and calculahng the penetration of particles through samphng sys- quantitative point of view. Nevertheless, several differences re-tems. >

garding the evolution of phenomena and affecting the timing v NUREG/lA-0090: ASSESSMENT OF RELAPS/ MOD 2 USING THE order have to be pointed out in the base calculations. The sen-TEST DATA OF REWET-It REFLOODING EXPERIMENT SGI/ sitivity study on the break flow and the nodalization study in the R, HAMALAINEN.A. Technical Research Centre of Finland components of the steam generator U-tubes and the cross-over

' (VTT). May 1993. 44pp. 9306230098. 75462:278. legs were also camed out. The RELAP5/ MOD 3 calcutation with An analyses of a reflooding expenment with RELAPS/ MOD 2 the nodahzation change resulted in good predictions of the 4

cycle 36 04 is presented. The expenmcint had been carned out major thermal-hydraubc phenomena and their timing order.

in the REWET il facility simulating the reactor core with a bundle of 19 electncally heated rods. On the basis of the results NUREG/lA-0099: RELAPS ASSESSMENT USING SEMISCALE -

of two calculations recommendations for the core nodalization SBLOCA TEST S-NH-1. LEE,E Ja CHUNG,B-D.; KIM.H-J. Korea .

are presented, and a modification to the code is proposed' institute of Nuclear Safety. June 1993. 250pp. 9306290162. l 75498:090.

• NUREG/lA-0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUT. 2-inch cold leg break test S.NH-1, conducted at the 1/1705 '

ER CODE AGAINST THE NET LOAD TRIP TEST DATA FROM volume scaled facility Semiscale, was analyzed using RELAP5/

YONG-GWANG. UNIT 2. ARNE N; CHO.S. Korea Electnc Power MOD 2 Cycle 36.04 and MOD 3 Version Sm5. Loss of HPIS was '

Corp. LEE,5.H. Korea Institute of Nuclear Safety. June 1993. assumed, and reactor tnp occurred on a low PZR pressure  !

75pp. 9306290172. 75498:339.

signal (13.1 MPa), and pumps began an unpowered coastdown The results of the RELAPS/ MOD 2 computer code simulation ,

on SI signal (12.5 MPa). The system was recovered by opening j for the 100% Net Load Trip Test in Yong-Gwang Unit 2 are ADV's when the PCT became higher than 811 K. Accumulator analyzed here and compared with the plant operation data. The was finally injected into the system when the pnmary system i h

- . . . .- = .. . ._-

12 Main Citations and Abstracts I pressure was less than 4.0 MPa. The expenment was terminat- behavior of pnmary pressure dunng pressurizer PORV actuation ed when the pressure reached the LPIS actuation set point. is poorly predicted because the actual behavior of pressunter RELAPS/ MOD 2 analysis demonstrated its capability to predict, PORV could not be modelled in the present simulation. .

with a sufficient accuracy, the main phenomena occumng in the NUREG/lA-0106: ASSESSMENT OF PWR STEAM GENERATOR depressurization transient, both from a qualitative and quantita- MODELLING IN RELAPS/ MOD 2. PUTNEY,J M.; PREECE,R.J.

tive points of view. Nevertheless, several d:fferences were National Power (United Kingdom). June 1993. 123pp noted regarding the break flow rate and inventory distribution 9307120157. TEC/L/0471/R91. 75623.025.

due to deficiencies in two-phase choked flow model, horizontal An assessment of Steam Generator (SG) modelling in the p stratification interfacial drag, and a CCFL model The main PWR thermal-hydrauhc code RELAPS/ MOD 2 is presented. The reason for the core to remain nearfy fully covered with the hquid .

was the under-prediction of the break flow by the code. Several assessment is based on a review of code assessment calcula-tions performed in the UK and elsewhere, detailed calculations  :

sensstrvity calculations were tried using the MOD 2 to improve against a series of commissioning tests carned out on the Wolf the results by using the different options of break flow modeling Creek PWR and analytical investigations of the phenomena in- ,

(downward, homogeneous, and area increase). The break area volved in normal and abnormal SG operation. A number of mod-compensating concept based on "the integrated break flow elling deficiencies are identified and their implications for PWR t

matching" gave the best results than downward junction and safety analysis are discussed - including methods for compen-homogeneous options. And the MOD 3 showed improvement in sating for the deficiencies through changes to the input deck.

predicting a CCFL in SG and a heatup in the core. Consideration is also given as to whether the deficiencies will NUREG/tA-0104: REuP5/ MOD 3 ASSESSMENT USING THE stdl be present in the successor code RELAP/ MOD 3.

SEMISCALE 50% FEED LINE BREAK TEST S-FS-11. LEE E-J.; ,

CHUNG.B-D.; KIM.H-J. Korea Institute of Nuclear Safety. June NUREG/lA-0108: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A

'993,200pp. 9306290153 75497:272. TURBINE TRIP FROM 100% POWER IN THE VANDELLOS 11 POWER PLANT. LLOPIS.C.; PEREZ,J.;

The RELAPS/ MOD 3 Sm5 code was assessed using the 1/ NUCLEAR 1705 volume scaled Semiscale 50% Feed Line Break (FLB) MENDIZABAL,R. Spain, Govt. of. June 1993. 58pp.

test S-FS 11. Test S-FS-11 was designed in three phases: (a) 9306210328. ICSP-V2-R100-R. 75403:325.

blowdown phase, (b) stabilization phase, and (c) refill phase. An assessment of RELAPS/ MOD 2 cycle 36.04 against a tur- ,

The fist ob}ective was to assess the code applicability to 50% bine trip from 100% power in Vandeflos il NPP (Spain) is pre- i FLB situation, the seced was to evaluate the FSAR conserv- sented. The work is inscribed in the framework of the Spanish atisms regarding SG hset transfer degradation, steam kne contnbution to ICAP Project. The model used in the simulation check valve failure, break %w state, and peak pnmary system consists of a single loop, a steam generator 6 ! a steam line pressure, and the third was 3 validate the EOP effectiveness. up to the stearn header all of them enlarged on a scale of 3:1; The code was able to simo - te the major T/H parameters and full scaled reactor vessel and pressunzer. The results of the except for the two-phase break low and the secondary convec- Calculations have been in reasonable agreement with plant tive heat transfer rate. The twowhase break flow had still defi- measurements. An additional study has been performed, to ciencies. The current boihng heat transfer rate was developed check the ability of a model m which all the plant components from the data for flow inside of a heated tube, not for flow are full-scaled to reproduce the transient. A second study has around heated tubes in a tube bundle. Results indicated that been performed using the Homogeneous Equilibrium Model in the assumption of 100% heat transfer until the liquid inventory the pressurizer trying to elucidate the influence of the velocity depletion was not conservative, the failed affected steam gener- slip in the primary depressunzation rate.

ator main steam line check valve assumption was not either conservative, the measured break flow expenenced all types of NUREG/lA-0109: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A flow conditions, the relative proximity to the 110% design pres. 10% LOAD REJECTION TRANSIENT FROM 75% STEADY sure limit was conservative. The automatic actions dunng the STATE IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

LLOPIS,C.; PEREZ,J.; CASALS.A.; et al. Spain, Govt. of. May  !

blowdown phase were eficctive in mitigating the consequences. 1993. 59pp. 9306210175. UNID-91-08. 75401:065.

The stabilization operation performed by operator actions were The Consejo de Segundad Nuclear (CSN) and the Asociacion effective to permit natural circulation cooldown and depressuri. Nuclear Vandellos (ANV) have developed a model of Vandellos  ;

zation, The voided secondary refill operations also venfied the effectiveness of the operations while recovering the inventory in il Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simula-a voided steam generator. e onkot syskms and & W compoMs, as

! on NUREG/lA-0105: ASSESSMENT OF RELAPS/ MOD 3 VERSION well as in the results analysis. The obtained model has been as-SM5 USING INADVERTENT SAFETY INJECTION INCIDENT sessed against the following trans ents occurring in the plant: a DATA OF KORI UNIT 3 PLANT. K!M,K.T.; CHUNG.B-D.; inp from the 100% power level (CSN); a load rejection from )

KIM,l.G ; et al. Korea Institute of Nuclear Safety. May 1993. 100% to 50% (CSN); a load rejection from 75% to 65% (ANV);

67pp. 9306180330. 75387:232. and a feedwater turbopump inp (ANV). This copy is a report of An inadvertent safety injection incident occurred at Kon Unit 3 the load rejection from 75% to 65% transient simulation. This in September 6,1990. It was analyzed using the RELAPS/ transient was one of the tests camed out in Vandellos 11 NPP 1 MOD 3 code. The event was initiated by a closure of main feed ~ dunng the startup tests. i water control valve of one of three steam generators. High pressure safety injection system was actuated by the low pres- NUREG/lA-0116: ASSESSMENT OF RELAPS/ MOD 3/V5M5 sure signal of main steam kne. The actual sequence of plant / GAINST THE UPTF TEST NUMBER 11 (COUNTERCURRENT transient with the proper estimations of operator actons was in- FLOW (N PWR HOT LEG). CURCA-TMG.F. Siemens AG - ,

vestigated in the present calculaton. The asymmetric loop be- KWU Group (formerly Siemens AG . Bereich Energieorzeugung l' haviors of the plant were also considered by nodalizing the (KWU)). May 1993.100pp. 9306210180. KWU E412/91/E10.

loops of the plant into three. The calculatonal results are com- 75401:122. I pared with the plant transient data. It is shown that the overall Analysis of the UPTF Test No.11 using the "best-estimate" plant transient depends strongly on the auxthary feedwater flow- computer code RELAPS/ MOD 3/ Version Sm5 is presented. Test rate controlled by the operator and that the code gives an ac- No.11 was a quasi-steady state, separate effect test designed ceptable prediction of the plant behavior with the proper as- to investigate the conditions for countercurrent flow of steam sumptons of the operator actons. The resutts also show that and saturated water in the hot leg of a PWR. An unphysical the solidification of pressunzer does not occur and the liquid- result was received using a CCFL correlation of the Walks type vapor mixture does not flow out through pressurizer PORV. The with the intercept C = 0.644 and the slope m = 0.8. The un- l l

l l

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Main Citations and Abstracts 13 physical prediction is an endication of possible programming wider use and to test the code capabilities As a result of the i errors in the CCFL model of the RELAP5/ MOD 3/V5m5 comput- analysis, it is felt that TRAC-BF1 is capable of reproducing the l ercode. plant behavior with an acceptable degree of accuracy although '

NUREG/lA-0118: ANALYSIS OF LOFT TEST LS-1 USING better models are clearly needed, in addition to further noding RELAP5/ MOD 2. COOPER,S. United Kingdom. May 1993.43pp. work and code improvements. The code tools almost 14000 9306210192. TD/SPB/ REP /0130. 75401:234. sec. which makes a 1/230 calculation time to real time ratio.

The RELAPS/ MOD 2 code. Reference 1, is being used by Nu- For this transient a mechanistic separator model is needed. It .

clear Electric for the calculation of Small Break Loss of Coolant will also help to cut down running costs if the vessel noding  !

Accidents (SBLOCA) and pressunzed transient sequences in could have a different number of cells at different heights. '

the Sizewell "B" PWR To validate the code for this purpose, il Though not very important for this transient, the entical flow has been used to model expenments of this type of transient model will allow for reafistic RV flow assumptions. There are not carned out in vanous integral test facilities. A number of these guidelines available for separator modelling in transients. It has studies have been for expenments carried out in the LOFT ex, twen found that a detailed noding in the separator region may ,

penmental reactor, Reference 2, and are desenbod in Refer. be needed to represent steam-water interaction. '

ences 3,4, 5, 6, and 7. To assist in assessing the capability of HELAP5/ MOD 2, the LOFT test LS-1 has been selected for NUREG/lA-0123: APPLICATION OF FULL POWER BLACKOUT analysis This test was designed to simulate the rupture of a FOR C.N. ALMARAZ WITH RELAP5/ MOD 2. LECHAS.A.L.

single 14 inch diameter accumulator injection line in a commer- S a n, Govt of. June 1993.100pp. 9306290128 ICSP-AL- l cial PWR. equivalent to a 25% area break in the broken loop BO 496 3 .

cold leg Early in the transient the pumps wer6 inpped and the g, oup of Almaraz Nuclear Power Plant has de-HPlS injuction initiated, towards the end of the transient, accu- velo ed a model of the plant with RELAP5/ MOD 2/36.04. This mulator and LPIS injection began, it should be noted that for model is the result of the work-exponence on the code Sizewell "B" analyses a 25% break is classified as large RELAP5/ MODI that was the standard code dunng the period whereas in this report, as in the cirternal literature, this break' 1984/1989. Different solutions were adopted in the network to size is referred to as intermediate. adequate the model to RELAPS/ MOD 2 Computer Code. This transient was selected for ICAP because it presents an experi.

MUREG/lA-0119: ASSESSMENT AND APPLICATION OF BLACK- ence with the same transient calculated with RELAP5/ MOD 1/

OUT TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH CY 29 Computer Code. The companson between both analysis RELAPS/ MOD 2. REVENTOS,F.; BAPTISTA.J.S ; NAVAS,A P.; will be interesting.

et al. Spain, Govt of. June 1993.63pp 9306290137. ICSP-AS-BOUT-R 75497:140. NUREG/lA-0124: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A The Asociacion Nuclear Asco has prepared a model of Asco PRESSURIZER SPRAY VALVE INADVERTED FULLY OPEN-NPP using RELAP5/ MOD 2. This model, which include thermal- ING TRANSIENT AND RECOVERY BY NATURAL CIRCULA.

hydraulics, kinetics and protection and controls, has been quali. TION IN JOSE CABRERA NUCLEAR STATION. ARROYO,R.;

fied in previous calculations of several actual plant transients REBOLLO.L. Union Electnca Fenosa S.A June 1993.125pp.

The first part of the transient presented in this report is an 9307060121. ICSP-JC.SPR-R. 75572:049.

actual black-out and one of the transients of the qualdication This document presents the comparison between the simula-process. The results are in agreement with plant data. The tion results and the plant measurements of a real event that i second part of the transient is a hypothetical case. It consists in took place in Jose Cabrera nuclear power plant in August 30, re-starting a pnmary pump and assume a new black-out. The 1984. The event was onginated by the local, continuous and in-phenomenology prediction of this second part has been useful adverted opening of the pressunzer spray valve PCV-400A. '

from the operation and safety point of view, Jose Cabrera power plant is a single loop Westinghouse PWR NUREG/tA-0120: ASSESSMENT OF THE TURBlNE TRIP TRAN. belonging to UNION ELECTRICA FENOSA, S.A. (UNION SIENT IN COFRENTES NPP WITH TRAC-BF1. CASTRfLLO.F. FENOSA), a Spanish utility which participates in the Internation-Hidroelectnca Espanola GOMEZ,A: GALLEGO.L; et at Union al Code Assessment and Applications Program (ICAP) as a lberoamencana De Tecnologia. June 1993 74pp.9306290147. member of UNIDAD ELECTRICA, S.A. (UNESA). This is the EST-SIAN-22. 75497:203 second of its two contributions to the Program; The first one .

This report presents the results of the assessment of TRAC. was an application case and this is an assessment one. The BF1 (G1 J1) code with the model of C. N. Cofrentes for simula- simulation has been performed using the RELAPS/ MOD 2 cycle i tion of the transiont originated by the manual trip of the main 36.04 code, running on a CDC CYBER 180/830 computer turbine. C. N. Cofrontec : ; General Electnc designed BWR/6 under NOS 2.5. operating system. The main phenomena have plant, with a nominal core thermal power of 2894 Mwt, in com_ been calculated correctly and some conclusions about the 3D mercial operation since 1985, owned and operated by Hidroe- characteristics of the condensation due to the spray and its sim-tectnca Espanosa, S A. The plant incorporates all the character- ulation with a 1D tool have been reached. ,

istics of BWR/6 reactors, with two turbine dnven FW pumps. As '

a result of this assessment a model of C. N, Cofrentes has NUREG/lA-0125: ASSESSMENT OF RELAP5/ MOD 2 COMPUT-been developed for TRAC-BF1 that fairly reproduces operation- ER CODE AGAINST THE NATURAL CIRCULATION TEST i DATA FROM YONG GWANG UNIT 2. ARNE.N.; CHO S. Korea al transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC- Electric Power Corp KIM H-J. Korea Instttute of Nuclear Safety. l BF1, from the 3D simulator. June 1993.110pp. 9306290132. 75497:032.

NUREG/lA-0122; ASSESSMENT OF MSIV FULL CLOSURE FOR The results of the RELAPS/ MOD 2 computer code simulation l for the Natural Circulation Test in Yong-Gwang Unit 2 are ana- l SANTA MARIA DE GARONA NUCLEAR POWER PLANT lyzed here and compared with the plant operation data. The USING TRAC-BF1 (G1J1) CRESPO.J L.; FERNANDEZ.R.A. result of companson reveals that the code calculation does Cantabna, Univ. of, Spain. June 1993 46pp. 9307060112. ICSP- present well the overall macroscopic behaviors of thermalhy-

, G A-M SIV-T. 75572.167. draulic parameters in primary and secondary system compared An assessment of the first 60 seconds of a spurious Main with the plant operating data. The sensitivity study is performed Steam insolation Vatve (MSIV's) closure for Santa Mana to find out the effect of steam dump flow rate on the primary Garona Nuclear Power Plant using TRAC-BF1 code is present- temperatures and it is found that the pnmary temperatures are ed Reasonable and realistic adjustments have been made in very sensitive to the steam dump flow rate dunng the Natural the model to improve its performance. This work is part of the Circulation. Because of the inherent uncertainties in the plant vahdation set for the TRAC model that is being developed for data. the assessment work is focussed on phenomena whereby

14 Main Citations and Abstracts the comparison between plant data and calculated data is based more on trends than on absolute values.

I I

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

SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER ANL-92/42 NUREG/CH-4744 V07 N1 1 CSP-GA MSty T NUREG/lA 0122 v

i ANL-93/2 NUREG/CR-4667 V15 IC5P-JC-SPR R NURE G/lA-0124 ANL 93/3 NUREG/CR 5999 ICSP V2-R100-R NUREG/lA-0108 j ANL-93/9 NUREG/CR-6032 DMI-2173 NUREG/CR 4599 V02 N2 fE412/91/E10 UR A-0 BNL NUREG 52354 NUREG/CR 5982 L A 12201-MS NUREG/CR-5776 BNL NUREG 52355 NUREG/CR-5983 (MF 136 NUREG/CR.4214 R1P2A2 DNL,NUREG 52356 ORNL/T M-12266 NURE G/CR 5965 NUREG/CR 5984 ORNL/T M- 12280 BNL,NUREG 52367 NUREG/CR-6014 NUREG/CR-5972 ORNL/TM 12307 NUREG/CR-5997 DNL NURE G-52375 NUREG/CR-6041 ORNL /TM-12349 CNWRA 91-01 A NUREG/CR 6036 NUREG/CR-5817 VO2 PNL-8106 NUREG/CR-5894 CNWRA 92 01S NUREG/CR-5817 V03 N1 PNL -8577 NUREG/CR-5410 CNWRA 92 026 NURE G/CR-6028 SAIC-91/6660 NUREG/CR-6018 EGG 2610 NURE G/CR-6061 SAND 89 2562 NUREG/CR-5471 E GG 2640 NUREG/CR 5759 SAND 9170H7 NURE G/CR 5801 EGG 2077 NUREG/CR 5882 SAND 92-0537 NUREG/CR-4832 V05 EGG-2688 NURE G/CR-5937 SAND 924537 NUREG/CR-4832 V09 E GG-2694 NUREG/CR-5976 SAND 92-2689 NURE G/CR 5W4 E GG 2697 SAND 92-2765 NUREG/CR-5305 V02 P1 1 NUREG/CR-6027 SAND 92-2765 EPRf TR-102106 NURE G/CR-6018 NUREG/CR 5305 V02 P2 l E ST SIAN 22 s1UDSVIKNS90/93 NUREG/lA-0094 NUREG/lA 0120 TD/SP8/ REP /0130 NUREG/lA 0118 ICSP AL-BOUT-R NUREG/lA-0123 iCSP-AS BOUT-R TEC/L/0471/R91 NUREG/iA-0106 NURE G/lA-0119 UNID 91 -0B NURE G/lA-0109 P

I 15

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l Personal Author index l

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

ABABOU.R. NUREGICR 6036: INITIAL RESULTS OF THE INFLUENCE OF BIAxlAL NUREG/CR 5817 V02. NRC HIGH-LEVEL RADIOACTIVE WASTE RE. LOADING ON FRACTURE TOUGHNESS.

SEARCH AT CNWRA Ca6endar Year 1991.

NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE BENDER,M.A.

RESE ARCH AT CNWRA January-June 1992 NURE G/CR4026- BIGr LOW. A NUMERICAL CODE F OR SIMULATING NUREG/CR 4214 R1P2A2. HEALTH EFFECTS MODELS FOR NUCLE- -

FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC AR POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Modificaten Of Models Resulting From Additon Of Effects '

MEDIA. Theory And User's Manual- Verson 1.1 Of Exposure To Alpha-Emit 1ing Radionuchdes Part 11: Scientdsc Bases ABRAHAMSON,S. For Health-.

NUREG/CH-4214 R1P2A2 HEALTH EF FECTS MODELS FOR NUCLE-  ;

AR POWER PLANT ACCIDENT CONSEQUENCE BERTA V.T.

ANALYSIS Modification Of Models Resutting From Addition Of Ettects NUREG/CR4061: DETERMINATION OF THE BIAS IN LOFT FUEL Of E.xposure To Alpha Emitting Radionuclades Part il Scientifc Bases PEAK CLADDING TEMPERATURE DAT A FROM THE BLOWDOWN For Health .. PHASE OF LARGE-BREAK LOCA EXPERIMENTS.

AHOLA,M. BICKEL,J.H.

NUREG/GR 5817 V03 N1. NRC HIGH-LEVEL RADIOACTIVE WASTE NUREG/CR-5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS RL SEARCH AT CNWRA. January-June 1992. STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-ANAND/l.K. RED WATER REACTOR PLANTS.

NUREG/GR-0006: DEPOSITION: SOFTWARE TO CALCULATE PARTI.

CLE PENETRATION THROUGH AEROSOL TRANSPORT SYSTEMS Final Report. NUREG/CR-6027. PRELIMINARY EVALUATION OF SNUBBER SINGLE FAILURES.

ARNE.N.

NUREG/lA-OO92 ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE BLUHM D.

AGAINST THE NET LOAD TRIP TEST DATA FROM YONG. NUREG/CR-5957: SYSTEM 80 + (TM) CONTAINMENT - STRUCTURAL GWANG. UNIT 2. DESIGN REVIEW.

NUREGilA-0125: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG- BOECKER B.B.

GWANG UNIT 2. NUREG/CR-4214 R1P2A2. HEALTH EFFECTS MODELS FOR NUCLE-AR POWER PLANT ACCIDENT CONSEQUENCE

^ **'*" " "#"E "^ " "

NU EG IA-0124. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES-Ex sure To Alpha Emit 1ng Radionuchdes.Part it: Scientife Bases SURl2ER SPRAY VALVE INADVERTED FULLY OPENING TRAN-

~

S!ENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA-BRE RA NUCLEAR STATION. BROSSEAU,D.A.

ATHE Y,G.F. NUREG/CR.4832 V09 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR 5247 V02 RASCAL VERSION 2.0 WORKBOOK. POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP). Internal Fire Analysis.

BaCA.R.

NUREG/CR 5817 V03 N1. NRC HIGH-LEVEL RADIOACTIVE WASTE BROWN,C.

RESEARCH AT CNWRA. January-June 1992 NUREG-0430 V12' LICENSED FUEL FACILITY STATUS BAGTZOGLOU.A.C.

NUREG/CR-5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE- 00 N SE' ARCH AT CNWRA Calendar Year 1991 NUREG/CR 6028 BIGFLOW A NUMERICAL CODE FOR SlMULATING BROWN'T.D' FLOW IN VARIABLY SATURATED. HETEROGENEOUS GEOLOGIC NUREG/CR 5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE MEDIA. Theory And User's Manual- Version 11. LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenornonology And Rmk Uncertainty Evaluaton Program (PRUEP) Appendees A C.

BACCER.L NUREG/CR-5305 V02 P2: INTEGRATED RISK /sSSESSMENT FOR THE NUREG/CR4032: SOLIDUS AND LIQUIDUS TEMPERATURES OF LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And CORECONCRETE MIXTURES. Risk Uncertainty Evaluation Program (PRUEP). Appendices D-G.

B AK E R,W.E.

BROWNSON D.A.

NUREG/CR-5776. DAMPlNG IN LOW- ASPECT RATIO. REINFORCED .

NUREG/CR.5937 INTENTIONAL DEPRESSURIZATION ACCIDENT CONCRETE SHEAR W ALLS MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC-BAPTIST A.J.S. TORS.

NUREG/LA-0119 ASSESSMENT AND APPLICATION OF BLACKOUT BR TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

NUR /CR-4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING WELDS.Somiannual Report. Octorier 1991 - March 1992.

BOSS.B.R.

NUREG/CR 5997. CSNI PROJECT FOR FRACTURE ANALYSES OF BR Y SON.J.W.

LARGE-SCALE INTERNATIONAL REFERENCE EXPERLMENTS NUREG/CR4036 INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL (PROJECT FALSIRE) LOADING ON FRACTURE TOUGHNESS i

I 17

18 Personal Author index 4

BULMAHN,K.D. NUREG/CR4817 V03 N1. NRC HIGH-LEVEL RADIOACTIVE WASTE J

NUREG/CR-5759. RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS RESEARCH AT CNWRA. January-June 1992.

STORAGE. SUPPLY, AND DtSTRIDUTION SYSTEMS IN PRESSUR-(ZED WATER REACTOR PLANTS CRESPO J.L NUREG/lA-0122: ASSESSMENT OF MSIV FULL CLOSURE FOR BURSON,$8. SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING NUREG/CR 5966. A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY TRAC-BF1 (G1J1).

CONTAINMENT SPRAYS.

CURCA TIVIG.F.

CASALS.A. NUREG/lA-0116: ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST NUREG/LA.0109 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10*4 THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWR LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE HOT LEG)

VANDELLOS 11 NUCLEAR POWER PLANT.

DANIEL,S.L.

CASTRILLO,F. NUREG/CR4832 V09: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR i

NUREG/lA4120 ASSESSMENT OF THE TURBINE TRIP TRANSIENT POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION l lN COFRENTES NPP WITH TR AC-BF1 PROGRAM (RMIEP). Internal Fire Analyt,is.  !

CHALLA R O 0 A$

i NURL R 9 7 SYSTEM 80+ (TM) CONTAINMENT -- STRUCTURAL R 'C/ 14: HIGH PRESSURE COOLANT INJECTION SYSTEM RISK-BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER j CHANIN,0.L STATION.

NUREG/CR4305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE DOCTOR,S.R.

LASALLE UNIT 2 NUCLEAR POWER PLANT.Phonomerciogy And Risk Uncertainty Evaluation Program (PRUEP) Appendices A C NUREG/CR4410: STATISTICALLY BASED REEVALUATION OF PISC-l:

NUREG/CR 5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR THE ROUND ROBIN TEST DATA LASALLE UNIT 2 NUCLEAR POWER PLANT:Phenomenology And Risk Uncertainty Evaluation Program (PRUEP) Appendices D-G M '

RG CHIEN,N.D. SEARCH AT CNWRA. Calendar Year 1991 NUREG/CR-5937: INTENTIONAL DEPRESSURIZATION ACCIDENT l EMRIT,R.  ;

MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC-Tops. NUREG-0933 S15. A PRIORITIZATION OF GENERIC SAFETY ISSUES.

ESCALANTE E.

CHO.S.  !

NUREG/lA-0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE NUREG/CR4735 V08: EVALUATION AND COMP!LAT60N OF DOE '

WASTE PACKAGE TEST DATA. Biannual Report. August 1989 - Janu-AGAINST THE NET LOAD TRIP TEST DATA FROM YONG.

GWANG. UNIT 2. ary 1990 NUREG/lA-0125. ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE I F ANOUS,F.

AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG.

GWANG UNIT 2. NUREG/CR4957: SYSTEM 80+ (TM) CONTAINMENT STRUCTURAL DESIGN REVtEW.

CHOPRA,0.K. I FARRAR,C.R.

NUREG/CR4667 Vts. ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual Report. April-September 1992. NUREG/CR-5776: DAMPING IN LOW-ASPECT-RATIO. REINFORCED i NUREG/CR-4744 V07 N1: LONG-TERM EMBRITTLEMENT OF CAST CONCRETE SHEAR WALLS.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual Report. October 1991 March 1992 FERNANDEZ R. A. ,

NUREG/CR-5999: INTERIM F ATIQUE DESIGN CURVES FOR CARBON, NUREG/lA 0122: ASSESSMENT OF MSIV FULL CLOSURE FOR LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVh SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING RONMENTS TR AC-BF1 (G1J1) >

CHOWDHURY,A.H. FINK,J.K. i NUREG/CR4817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE- NUREG/CR-6032: SOLIDUS AND LIQUIDUS TEMPERATURES OF SEARCH AT CNWRA Calendar Year 1991 CORE-CONCRETE MIXTURES.

NUREG/CR-5817 V03 N1: NRC HIGH. LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA January-June 1992. FOWLER.R.D.

NUREG/CR 5976. DEVELOPMENT AND USE OF A TRAIN-LEVEL {

NURE"G/ A-0095: RELAPS ASSESSMENT USING LSTF TEST DATA SB-CL-18 FRAKER,A.C.

NUREG/lA 0099- RELAPS ASSESSMENT USING SEMISCALE SBLOCA NUREG/CR-4735 V08: EVALUATf0N AND COMPILATION OF DOE

1. ^ U '8""" "9"" ' *

• NU / k 01 4' RELAPS/ MOD 3 ASSESSMENT USING THE SEMIS- 87 CALE 50% FEED LINE BRE AK TEST S-FS-11.

NUREG/lA 0105. ASSESSMENT OF RELAPS/ MOD 3 VERSlON SM5 FRANCINI,R.

USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF NUREG/CR4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING KORI UNIT 3 PLANT. WELDS Semiannual Report. October 1991 March 1992.

CHUNG.H M. Ol NUREG/CR4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN ji .0120. ASSESSMENT OF THE TURBINE TRIP TRANSIENT LIGHT WAT ER REACTORS. Semiannual Report.ApribSeptember 1992 IN COFRENTES NPP WITH TRAC-BF1.

CONOSCENTE.J.P. CHADIAll,N.

NUREG/CR-6013 METHODS USED FOR THE TREATMENT OF NON- NUREG/CR-4599 V02 N2: SHOR1 CRACKS IN PIPING AND PtPING PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS. WELDS Semiannual Report. October 1991 - March 1992. ,

COOPER S. GILBERT,E S.

NUREG/lA-0118. ANALYSIS OF LOFT TEST L5-1 USING RELAPS/ NUREG/CR4214 R1P2A2: HEALTH EirFECTS MODELS FOR NUCLE.

MOD 2. POWER PLANT ACCIDENT CONSEQUENCE AR ,

CHAGNOLINO,G. ANALYSIS. Modification Of Models Resulting From Addition Of Effects >

NUREG/CR4817 V02: NRC HIGH LEVEL RADIOACTIVE WASTE RE- Of Erposure To Alpha-Emrmng Radionuclides.Part I!: Scientific Bases SEARCH AT CNWRA. Calendar Year 1991. For Health...

- . . . - - - - . - .-. - , . ~ . - .

9 Personal Author index 19 GOME2.A. JOHNSON,J D.

NUREG/lA-0120. ASSESSMENT OF THE TURBINE TRIP 1RANSIEN1 NUREG/CR 5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE IN COFRENTES NPP WITH TRAC BF1 LASALLE UNIT 2 NUCLE AR POWER PLANT Phonomenology And GOODWIN,G M. Rmk Uncertainty Evaluation Program (PRUEP) Appendices A-C NUREG/CR 5305 V02 P2. INTEGRATED RISK ASSESSMENT F OR THE NUREG/CR 5972 EFFECTS OF NONSTANDARD HEAT TRE ATMENT LASALLE UNIT 2 NUC! EAR POWER PLANT Phenornenology And TEMPERATURES ON TENSILE AND CHARPV IMPACT PROPERTIES Rcsk Uncartainty Evaluation Program (PRUEP). Appendices D-G OF CARBON-STEEL CASTING RE PAIR WELDS. {

GH AVE S.C.C. JOHNSON.J.J. l NURE G 136dl. REGULATORY ANALYS!S FOR THE RESOLUTlON OF NURE G/CR-6013 METHODS USED FOR THE TREATMENT OF NON.  ;

GENERIC SAFETY ISSUE 106. PIPING AND THE USE OF HIGHLY PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS. i COMBUSTIBLE GASES IN VIT AL ARE AS JOYA l GREE N.R.T. NUREG 0430 V12 LICE NSED FUEL FACILITY STATUS NUREG/CR 5817 V02 NRC HIGH LEVEL RADIOACTIVE WASTE RE. REPORT. inventory Ditterence Data July 1,1991 - June 30,1992 (Gray SE ARCH AT CNWRA Calendar Year 1991. BOOK ll)  !

NUREG/CR 5817 V03 Nt: NRC HIGH-LEVEL RADIOACTIVE WASTE RESE ARCH AT CNWRA, January-June 1992 K ASSNE R.T.F.

NUREG/CR-4667 V15. ENVIRONMENTALLY AS$1STED CRACKING IN GREIMANNL LIGHT WATER RE ACTORS. Semiannual Report.Aptd-Septemter 1992.

NUREG/CR 5957: SYS1EM 80 + (TM) CONTAINMENI - STRUCTURAL DESIGN REVIEW K ATSM A.K.R.

NUREG/CR 6035 F EAStBILITY STUDY FOR IMPROVED STEADY-GROUNDWATER.E. STATE INiilALIZATION ALGORITHMS FOR THE RELAPS COMPUT.

NURE G/CR 6018 $URVEY AND ASSESSMENT OF CONVENTIONAL E R CODE.

SOFTWARE VERIFICATION AND VALIDATION METHODS l KE ENE Y-W ALKER.J .

NU EG .2894 RADIONUCLIDE CHARACTER 12AilON OF RE AC.

TOR DE COMMISS:CNfNG W AST E AND NEUTRON- ACTIV AT ED fP E[C A Si i METALS HAMALAINEN.A. OEUA i J

NUREG/lA-0990 ASSESSMC NT OF RE LAPS / MOD 2 USING THE TEST NUREG/CR 6027: PRELIMINARY EVALUATION OF SNUBBE R SINGLE ,

FAILURES j DAT A OF REWET-il HEFLOODING EXPLRIMENT SGl!R H ANE Y,LN. KE NNETT.R.J. I NURE G/CR 5937. INTE NT IONAL DEPRESSUR12ATION ACCIDENT NUREG/CR-5984 CODE AND MODEL EXTENSIONS OF THE THATCH l MANAGEMENT STR AT EGY FOR PRESSUR! ZED WATER RE AC- CODE FOR MODULAR HIGH TEMPERATURE GAS COOLED REAC. -

TORS TORS <

f HANSON,RG. KILINSKI.T.

s NUREG/CR 6061 DETERMINATION OF THE BIAS IN LOFT FUEL NUREG/CR-4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING  ;

PEAK CLADDING TE MPERATURE DATA FROM THE BLOWDOWN WELDS Sermannual Report, Octot or 1991 March 1992. [

PHASE OF LARGE-BREAK LOCA EXPERIMENTS KtM.H-J.

HEASLER.P.G. NUREG/lA 0095. RELAPS ASSESSMENT USING LSTF TEST DATA SO NUREG/CR 5410. STATISTICAL LY BASED REEVALUATION OF DISC-Il CL-18 ROUND ROBIN TEST DATA NUREG/lA-0099. RELAPS ASSESSMENT USING SEMISCALE SBLOCA TE ST S-NH 1.

HEBDONLJ. NUREG/lA-0104. RELAP5/ MOD 3 ASSESSMENT USING THE SEMIS-NURE G-14 74 EFFECT OF HURRICANE ANDREW ON THE TURKEY CALE 50% FEED LINE BREAK TEST S-FS-11.

PO:NT NUCL EAR GENERATING STATION FROM AUGUST 20 30, NUREG/lA 0105: ASSESSMENT OF RELAPS/ MOD 3 VERSION SMS 1992 '

USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF KORI UNIT 3 PLANT.

HIGGINS.S.J. NUREG/lA-0125' ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE ,

NUREG/CR 5305 V02 Pt. INTEGRATED OISK ASSESSMENT FOR THE AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG-  !

LASALLE UNIT 2 NUCL E AR POWE R PLANT Phenomrmology And GWANG UNIT 2.  !

Rsk Uncertainty Evaluation Program (PRUEP) A  !

NUREG/CR 5305 V02 P2 INTEGRATED RISK Abnondices FOR THE' A-C. KIM.LG.

ESSMENT i LASALLE UN!T 2 NUCLE AR POWER PLANT Phenomenoicqy And NURECJIA-0105- ASSESSMENT OF RELAPS/ MOD 3 VERSION SMS R*k Uncertainty Evaluahon Pcgram (PRUEP) Appendeces D-G USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF i HINS,A.G. KORI Uf IIT 3 PLANT. ,

NUREG/CR-4667 V15- ENVIRONMENT ALLY ASSISTED CRACKING IN i LIGHT W ATER RE ACTORS Semiannual Hoporq.Aptd-September 1992- KIM K T'G/lo-0105 NUFkE ASSESSMENT OF RELAPS/ MOD 3 VERSION SM5 HSIUNG.S.H USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF NUREG/CR 5817 V02 NRC HIGH-LEVEi RADIOACTIVE WASTE RE- KORl UNIT 3 PL ANT.

SE ARCH AT CNWRA Calorutar Year 1991 KOCMOUD,C.J '

"^ ^

RES AR T Cf RAJr ry ute 19 NUREG/GR-0006. DEPOSITION. SOFTWARE TO CALCULATE PARTI- )

CLE PENETRATION THROUGH AEROSOL TRANSPORT HUDDLESTON RL SYSTE MS. Final Report.

NUREG!CR-5955 MATERIALS AND DESIGN BASES ISSUES IN ASME CODE CASE N47 KRISHNASWAMY,P.

NUREG/CR4599 V02 N2. SHORT CRACKS !N PlPING AND PIPING INTERRANTE.C.G. WE LDS,Semsannual Report. October 1991 - March 1992 NUREG/CR-4725 Voll FVALUATION AND COMPILATION OF DOE WASTE PACKAGE TEST DATA Biannual Report. August 1989 Janu. KROEGER.P.G.

ary 1990 NUREG/CR4983. SAFETY ASPECTS OF FORCED FLOW COOLDOWN TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED JOHNSEN.G.W. REACTORS.

NURE G/CR-6061: DETERMINATION OF THE BIAS IN LOFT FUEL NUREG/CR 5984: CODE AND MODEL EXTENSIONS OF THE THATCH PEAK CLADD'NG TE MPLRATURE DATA FROM THE BLOWDOWN CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLE.D REAC.

PHASE OF LARGE.BRE AK LOCA EXPERIMENTS TORS.

I i

l l

I 20 Personal Author index LAMBRIGHT.J.A. MCK ENN A,T.J.

NUREG/CR4832 V05: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR-5247 V02: RASCAL VERSION 2.0 WORKBOOK. [

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM Parameter Estimaton Analysis And Screening Human Reli. MCNAMARA N. i I

ability Analysis. NUREG.0837 V13 N01: NRC TLD DIRECT RADIATION MONITORING NUREG/CR4832 V09: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NETWORK. Progress Report. January-March 1993.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM (RMIEP). Internal Fra Analysis. MENDIZABAL,R.

NUREG/lA-0108. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A TUR-NU EG'/C'R4599 VO2 N2- SHORT CRACKS IN PIPING AND PIPING WELDS. Semiannual Report. October 1991 - March 1992.

NU EG/ A 010 ASSESSMENT OF RELAP5/ MOD 2 AGAINST A 10% (

LECHAS.A.L LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE NUREG/lA-0123: APPLICATION OF FULL POWER BLACKOUT FOR VANDELLOS 11 NUCLEAR POWER PLANT. ,

C.N ALMARAZ WITH RELAPS/ MOD 2.

LEE.E-J. NUREG/CR-5305 V02 P1; INTEGRATED RISK ASSESSMENT FOR THE NUREG/lA 0009. RELAPS ASSESSMENT USING SEMISCALE SBLOCA LASALLE UNIT 2 NUCLEAR POWE.R PLANT.Phenomenology And TEST S-NH-1. Risk Uncertainty Evaluaton Program (PRUEP) Appendices A.C.

NUREG/lA 0104. RELAPS/ MOD 3 ASSESSMENT USING THE SEMIS- NUREG/CR-5305 V02 P2; INTEGRATED RISK ASSESSMENT FOR THE CALE 50% FEED LINE BREAK TEST S-FS-11. LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Risk Uncertainty Evaluaton Program (PRUEP). Appendices D G.

LEE.S ^

NUREG/lA.0095 RELAPS ASSESSMENT USING LSTF TEST DATA SB-SOFTWARE VERIFICATION AND VALIDATION METHODS.

Cb18.

MIRSKY,S.M.

LE E.S.H.

NUREG/IA-0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE NUREG/CR-6018 SURVEY AND ASSESSMENT OF CONVENTIONAL AGAINST THE NET LOAD TRIP TEST DATA FROM YONG. SOFTWARE VERIFICATION ANO VALIDATION METHODS.

GWANG. UNIT 2 LEHNER,J.R. NUREG/lA 0119: ASSESSMENT AND APPUCATION OF BLACKOUT NUREG/CR-5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN TRANSlENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

CONTAINMENT MANAGEMENT. MOD 2. ,

LilBOWITZ,L MORTON.D.K.

NUREG/CR-6032: SOLIDUS AND LlOUIDUS TEMPERATURES OF NUREG/CR 6027: PREUMINARY EVALUATION OF SNUBBER SINGLE CORE CONCRETE MIXTURES- FAILURES.

LESLIE.8.W.

^ NU EG CH-5801: PROCEDURE FOR ANALYSIS OF COMMON-CAUSE  !

SE RCH T A andar Y ar 1.

FAILURES IN PROBABiUSTIC SAFETY ANALYSIS.

NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE 3 RESEARCH AT CNWRA January-June 1992. MURPHY,W.M. {

~

LLOPIS C. NUREG/CR 5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREG/ lao 100: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A TUR- SEARCH AT CNWRA. Calendar Year 1991.

BINE TRP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE l POWER PLANT. RESEARCH AT CNWRA. January June 1992. i F

NUREG/lA-0109. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10%

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE NANSTAD,R.K.

VANDELLOS Il NUCLEAR POWER PLANT NUREG/CR-5972. EFFECTS OF NONSTANDARD HEAT TREATMENT '

TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES MAJUMDAR,S. OF CARBON-STEEL CASTING REPAIR WELDS. l i

NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN Nf) REG /CR-6036: INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL '

LOADING ON FRACTURE TOUGHNESS.

N E /C 99 i M A OE EI N d S R C RB .

LOW ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI- NAVAS A.P.

RONMENTS. NUREG/lA-0119. ASSESSMENT AND APPLICATION OF BLACKOUT TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/

M ANTEUFEL R.D. MOD 2.

NUREGICR-5817 V02 NRC HIGH-LEVEL RADIOACTIVE WASTE RE. .

SEARCH AT CNWRA Calendar Year 1991- HILSSON,L j NUREG/CR-5817 V03 N1- NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA January-June 1992 NUREG/lA-0094. ASSESSMENT OF RELAPS/ MOD 3 AGAINST i l

TWENTY-FIVE POST DRYOUT EXPERIMENTS PERFORMED AT THE MARSCHALLC.W. ROYAL INSTITUTE OF TECHNOLOGY. f l

NUREG/CR4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING '

WELDS. Semiannual Report. October 1991 March 1992 NITZELM.E.

NUREG/CR-6027; PRELIMINARY EVALUATION OF SNUBBER SINGLE MARTIN.R.P. FAILURES ~

NUREG/CR-5882 TRAC-B THERMAL-HYDRAUllC ANALYSIS OF THE BLACK FOX BOlUNG WATER REACTOR NOURBAKHSH,H.P.

NUREG/CR-5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN {

MASLENIKOV,0.R. CONTAINMENT MANAGEMENT.

NUREG/CR-6013. METHODS USED FOR THE TREATMENT OF NON-PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS OLIVE.K.L.

NUREG-1350 V05. NUCLEAR REGULATORY COMMISSION INFORMA-MCAFEE.W.J.

TlON DIGEST.1993 Editon.

NUREG/CR 6036: INITIAL RESULTS OF THE INFLUENCE OF BIAXtAL LOADING ON FRACTURE TOUGHNESS PABALAN R.T.

MCF ARLAND,A.R. NUREG/CR-5817 V02. NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREG/GR 0006' DEPOSITION SOFTWARE TO CALCULATE PARTI. SEARCH AT CNWRA Calendar Year 1991.

CLE PENETRATION THROUGH AEROSOL TRANSPORT NUREG/CR 5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE SYSTEMS Fina! Report. RESEARCH AT CNWRA. January-June 1992.

Personal Author index 21 PARK,J.Y. R AO.M.C-NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6036: INITIAL RESULTS OF THE INFLUENCE OF BIAXlAL LIGHT WATER REACTORS Semiannual Report,Apnt-September 1992. LOADING ON FRACTURE TOUGHNESS.

PARRY,G.W.

R ASMUSSON.D.M. l f4UREG/CR-5471 ENHANCEMENTS TO DATA COLLECTION AND RE- NUREG/CR-5471: ENHANCEMENTS TO DATA COLLECTION AND RE-PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS- l PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

PAULA.H.M.

NUREG/CR 5471, ENHANCEMENTS TO DATA COLLECTION AND RE^ REBOLLO.L NUREG/lA-0124- ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES-PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN-PAULSE N.M.P. SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA-NUREG/CR-6035- F E ASIBILITY STUDY FOR IMPROVED STEADY. BRERA NUCLEAR STATION.

STA INITIAll2ATION ALGORITHMS FOR THE RELAPS COMPUT.

~

REVENTOS F. ,

NUREG/lA 0119- ASSESSMENT AND APPLICATION OF BLACKOUT '

PAYNE,A.C. TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/ l' NUREG/CR-4832 VOS- ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR MOD 2.

POWER PLANT; RISK METHODS INTEGRATION AND EVALUATION PROGRAM Parameter Estimaton Analysis And Screening Human Reto ROBERTSON.D.E.

abilQ Analysis. NUREG/CR-5894; RADIONUCLIDE CHARACTERIZATION OF REAC-NUREu/CR-4832 V09 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR TOR DECOMMISSIONING WASTE AND NEUTRON-ACTIVATED POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION METALS.

PROGRAM (RMtEP) Internal Fire Analysis NUREG/CR-5305 V02 P1. INTEGRATED RISK ASSESSMENT FOR THE ROCHE M.F. >

LASALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And NUREG/CR 6032: SOLIDUS AND LIQUIDUS TEMPERATURES OF l Risk Uncertainty Evaluaton Program (PRUEP) Appendices A-C CORE CONCRETE MIXTURES.

NUREG/CR-5305 VO2 P2 INTEGRATED RISK ASSESSMENT FOR THE LASALLE UNIT 2 NUCLEAR POWER PLANT:Phenomenology And RUTHER,W.E.

Risk Uncertainty Evaluaton Program (PRUEP). Appendices D-G NUREG/CR-4067 V15 ENVIRONMENTALLY ASSISTED CRACKING IN  !

LIGHT WATER HEACTORS. Semiannual Report Apnt-September 1992.

NUREG/CR.5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE- SAGAR.B.

SEARCH AT CNWRA Calendar Year 1991. NUREG/CR-5817 VO2: NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREG/CR4817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE SEARCH AT CNWRA. Calendar Year 1991.

RESEARCH AT CNWRA January-June 1992 NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE ,

RESEARCH AT CNWRA. January-June 1992.

NUREG/CR4036: INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL SANECKI,J E.

LOADING ON FRACTURE TOUGHNESS NUREG/CR-4667 V15: ENVIRONMENTALLY ASSISTED CRACKING IN PER W LIGHT WATER REACTORS Semiannual Report,Apni-September 1992.

NUREG/lA-0198 ASSESSMENT OF RELAP5/ MOD 2 AGAINST A TUR-B TRIP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR N EG C 5759: R!SK ANALYSIS OF HIGHLY COMBUST!BLE GAS STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-NUREG/LA-0109. ' ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10%

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE IZED WATER REACTOR PLANTS.

VANDELLOS 11 NUCLEAR POWER PLANT, SCHULTZ,R.R.

PE R EZ.S.E. NUREG/CR.6061: DETERMINATION OF THE BIAS IN LOFT FUEL NUREG/CR-5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN CONTAINMENT MANAGEMENT PHASE OF LARGE-BREAK LOCA EXPERIMENTS. ,

PETERSON,C.E. SCHULZ,H.

NUREG/CR4035: FEASIBILITY STUDY FOR IMPROVED STEADY. NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF ,

STATE INITIALIZATION ALGORITHMS FOR THE RELAP5 COMPUT. LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS ER CODE. (PROJECT FALSIRE)

POWERS.D.A. SCOTT.B.R. (

NUREG/C45966. A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY NUREG/CR-4214 R1P2A2: HEALTH EFFECTS MODE'LS FOR NUCLE- t CONTAINMENT SPRAYS. AR POWER PLANT ACCIDENT CONSEQUENCE i ANALYSIS Mod 6 cation Of Models Resulting From Additon Of Effects 4 PREECE.R.J. Of Exposure To Alpha-Emitting Radonuclides Part 11: Scientific Bases NUREG/lA 0106; ASSESSMENT OF PWR STEAM GENERATOR MOD- For Health...,

ELLING IN RELAPS/ MOD 2.  ;

SCOTT,P.

CF5-5817 V02; NRC HIGH-LEVEL RADIOACTIVE WASTE RE- # "

JR SEARCH AT CNWRA. Calendar Year 1991, WELDS Sermannual Report October 1991 - March 1992.

NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE SHACK,W.J. '

RESEARCH AT CNWRA, January June 1992. NUREG/CR-4667 V15; ENVIRONMENTALLY ASSISTED CRACKING IN ,

PUGH.C E LIGHT WATER REACTORS. Semiannual Report,Apr$ September 1992.  !

NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF NUREG/CR-5999: INTERIM FATIQUE DESIGN CURVES FOR CARBON, LARGE SCALE INTERNATIONAL REFERENCE EXPERIMENTS LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI-

• (PROJECT F ALSIRE) RONMENTS. '

PUTNEY,J.M. SHIVERAW.

NUREG/lA 0106 ASSESSMENT OF PWR STEAM GENERATOR MOD, NUREG/CR-5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE ELLING IN RELAP5/ MOD 2. LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Risk Uncertainty Evaluaton Program (PRUEP) Appendices A C.

l RAHMAN.S. NUREG/CR-5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR THE NUREG/CR-4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING LASALLE UNIT 2 NUCLEAR POWER PLANTPhenomenology And WELDS Sennannual Report. October 1991 March 1992. Risk Uncertainty Evaluation Program (PRUEP) Appendices D-G i

1

22 Personal Author index SIEVERS.J. THOM AS,C.W.

NUREG/CR SW7 CSNI PRt :JE r* FOR F RACTURE ANALYSE S OF NUREG/CR SA94 R ADiONUCLlDE CHARACTE Rl2ATION OF RE AC-LARGE SCALE INTERNA > vNAL REF E RENCE E XPE R: MENT S TOR DECOMMISSIONING W AST E AND NEUTRON ACTIVATED iPROJE CT f At SIRE ) METALS SIMION.G P TULLIS J.P.

NUREG/CR 57b9 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS NURE G/CR 6031 CAVITATION GUIDE FOR CONTROL VALVE S STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR TURNER.D.R.

12ED W ATER RE ACTOR PLANTS NUREG/CR4817 V02: NRC HIGH, LEVEL RADIOACTIVE W ASTE RE-SE ARCH AT CNWRA Calendar Year 1991 SMITH.C L NUREG/CR 5817 V03 N1. NRC HIGH. LEVEL RADIOACTIVE WASTE NUREG/CR 5759 RISK ANALYS!S OF HIGHL Y COMBUSTIBL E GAS RESEARCH AT CNWRA January-June 1992.

STOR AGE. SUPPL Y, AND DISTH BUTION SYSTEMS IN PRESSUR-i2E D WATER RE ACTOR PLANTS VANHORN R.L NUREG/CR 5976 DE VE L OPME NT AND USE OF A TRA!N LEVE L NUREG/CR 5759 RISK ANALYSFS OF HIGHLY COMBUSTIBLE GAS .

PROBABILISilC RISK ASSESSMENT STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR. l IZED WATE R RE ACTOR PLANTS SRIDHAR.N.

NURE G/CR 5817 V02 NRC HIGH LEVE L RADIOACTIVE WASTE RE. W ARE.A.G. I SE ARCH AT CNWRA Calerviaf Year 1991 NUREG/CH.6027 PRELIMINARY EV AL UATION OF SNUBBER SINGLE NURE G/GR $617 V03 NI NRC HIGH LEVEL RADIOACTIVE WASTE F AILURES RESE ARCH AT CNWRA January. June 1992 WE ISS. A.J.

STlRE WA LT,G. NUREG/CP-0126 V01: PROCEE DINGS OF THE TWENTIETH WATE R NURE G/CR 5017 V03 N1 NRC HIGH LEVE L RADIOACTIVE WASTE RE ACTOR SAFETY INrORMATION MEETING NURE G/CP-0120 V02. PROCE ED!NGS OF THE TWENTIETH WATER RE SE ARCH AT CNWRA January June IW2 RE ACTOR SAFETY INF ORMATION MEE TING STRUCK M E YER,R.

NURE GICP U126 V03 PROCEEDINGS OF THE TWENTIETH WATER FiE ACTOR SAFETY INFORMATION MEETING NURE G OH37 V13 NOI NRC TL D DIRECT RADIATION MONITORING NLTWORK P oyess Report Janusy March 1993 WHEELER,T.A.

NUREG/CR-4832 V05 ANALYSIS Or THE LASALLE UNIT 2 NUCLEAR SULLIVAN,T.M ^

NURE G/CR 6041 D6POSAL UNIT SOURCE TE RM (DUST) DATA awM a n Anahm AM krmg Human %

INPUT GUIDF stulity Analyms SW AIN,A D. WHITEHEAD.D W.

NURE G/CR.4832 VU5. ANAL YSIS OF THE L ASALLE UNIT 2 NfLE AR NLlRE G/CR-54 71 ENHANCEMENTS TO DATA COLLECTION AND RE-POWER PLANT RISK METHODS IN1EGRATION AND EV AtUATION PORTING OF SINGLE AND MULTIPLE F AILURE EVENTS PROGRAM Parameter E stimation Analysis And Screenmg Hurnan Reh-atxhty Anatym WILKOWSKI,G M.

NUREG/CR 4599 V02 N2 SHORT CRACKS IN PIPING AND PIPING SWINDE M AN.M.J. WE LDS Semiannuai Report. Octotar 1991 - March 1992.

NURE G/CR 5972 E F F E CT S OF NONST ANDARD HE AT TRLATME N1 TEMPERATURES ON TENS!LE AND CHARPY IMPACT PROPERTIE S WfTTME YE R.G.W.

OF CARBON STE E L CASTING RE PAIR WELDS NUREG/CR 5817 V02 NRC HIGH. LEVEL RADIOACTIVE WASTE RE-SE ARCH AT CNWR A Calendar Year 1991 SWINDE M AN,R W. NUREG/CR-5817 V03 N1 NRC HIGH LEVEL RADIOACTIVE WASTE RE.SE ARCH AT CNWRA January-June 1992.

NURE GeCR 5955 MATE RIALS AND DE SIGN DASE S ISSUES IN A5ME CODE CASE N-47 WOLFRAM.LM.

NUREG/CR 5976 DEVELOPMENT AND USE OF A TRAIN LEVEL SY PE.,T .T.

PROBABILISTIC RISK ASSESSMENT.

NUREG,CR 5305 V02 P1 INTEGRATED RISK ASSESSMENT FOR THE LASALL E UNIT 2 NUCLE AR POWE R PLANT Phenomenology And WONGJ.S.

Risk Uncertamry E valuation Program (PRUE P) Appendices A.C NUREG/GR.0006 DEPOSITION SOFTW ARE TO CALCULATE PARTI-NURL G/CR 5305 V02 P2 INTE GRATED RISK ASSE SSML NT FOR THE CLE PE NE TR ATION THROUGH AEROSOL TRANSPORT L ASAl tg UNIT 2 NUCLE AR POWER PLANT Phenomonoiogy And SYSTEMS hnal Report Risk Urx ertainty E valuation Pro;pam (PRUE P) Append *ces D G WYNHOF F,N.L T AYLOR.T.T. NUr,EG/CR.5894 RADIONUCLtDE CHARACTERi2AilON OF RE AC-NURE G/CR 5410 ST ATishCAL L Y BASE D REE VALUATION OF PISC.li TOR DECOUM!SSIONING W ASTE AND NEUTRON ACTIVATE D ROUND ROBIN TEST DAT A ME T AL S.

T H E ISS,T.J- YOUNG.S.

NURE G/CR W.16 IN!TIAL RE SULTS Of THE INF LUENCE Of BiAX!AL NUREG'CR 5817 V03 N1 NRC HIGH-L EVE L RADtOACTIVE WASTE LOAD!NG ON F RACTURE TOUGHNE SS RE SE ARCH AT CNWRA JanuarpJune 1992

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

I A533 8 Steel Blackout '

NUREG/CR4030 INITIAL RESULTS OF THE INFLUENCE OF BlAXiAL NUREG/lA 0119. ASSESSMENT AND APPLICATION OF BLACKOUT LOADING ON FRACTURE TOUGHNESS TRANS!ENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/

ACRS Report MOD 2 NUREG/lA 0123 APPLICATION OF FULL POWER BLACKOUT FOR NUREG-1125 V14. A COMPil ATION OF REPORTS OF THE ADVISORY C N ALMARAZ WITH RELAPS/ MOD 2 COMMITTEE ON REACTOR SAFEGUARDS 1992 Annual Bolling Water Reactor NUR R5 ATERIALS AND DESIGN BASES ISSUES IN ASME

" ^ '

CODE CASE N-47. B OW WW MN Abnormal Occurrence Budget NUREGM90 V15 N04 REPORT TO CONGRESS ON ABNORMAL NUREG-1100 V09 BUDGET ESTIMATES Fscal Years 1994-1995 OCCURRENCES. October-Decemte 1992. g,gg Abnormal Transient Condition NUREG/CR-5972. EFFECTS OF NONSTANDARD HEAT TREATMENT I NUREG/CR 5882. TRAC-B THERMAL HYDRAULIC ANALYslS OF THE TEMPERATURES ON TE.NSILE. AND CHARPY IMPACT PROPERTIES h BLACK FOX BOILING WATER REACTOR OF CARBON-STEEL CASTING REPAIR WELDS.  ;

Accident Consequence Analysis Cavitation i NUREG/CR-4214 R1P2A2; HEALTH EFFE. CTS MODELS FOR NUCLE. '

NUREG/CR 6031: CAVITATION GUIDE FOR CONTROL VALVES AR POWER PLANT ACCIDENT CONSEQUENCE ANALYSIS Modificabon Of Models Rosvihng From Ashon Of Effects Charpy impact Toughness Of Esposure To Alpha-Emdhng Radionuchaos Part 9 Scienhhc Bases NUREG/CR-5972: EFFECTS OF NONSTANDARD HEAT TREATMENT I For Health _. TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPERTIES OF CARBON-STELL CASTING REPAIR WELDS.

• Accident Management NUREG/CR-5937, INTENT ONAL DEPRESSURl2ATION ACCIDE.NT Combustible Gas MANAGEMENT STRATEGY FOR PRESSURf2ED WATER REAC. NUREG-1364 REGULATORY ANALYSIS FOR THE RESOLUTION OF TORS GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY COMBUSTIBLE GASES IN VITAL AREAS.

Advanced Reactor NUREG/CR 5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS NUREG/CR-5955 MATERIALS AND DESIGN BASES ISSUES IN ASME STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR CODE CASE N-47 IZED WATER REACTOR PLANTS Aerosoi Common Cause failure NUREG/CR-5%b A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY NUREG/CR-5801: PROCEDURE FOR ANA! YSIS OF COMMON-CAUSE CONTAINMENT SPRAYS FAILURES IN PRODABILISTIC SAFETY ANALYSIS Aerosol Transport System Concrete NUREG/GR4006 DEPOSITION SOFTWARE TO CALCULATE PARTI' CLE PENET RATION THFIOUGH AEROSOL TRANSPORT NUREG/CR-6032- SOLIDUS AND LlOulDUS TEMPERATURES OF SYSTE MS Final Report CORE-CONCRETE MIXTURES Air Sampling Containment Penetration ,

NUREG/GR4006 DEPOSITION SOFTWARE TO CALCULATE PARTI- NUREG/CR-6027: PRELIMINARY EVALUATION OF SNUBBER SINGLE CLE PE NETRATION THROUGH AEROSOL F AILURE S.

TRANSPORT SYSTEMS Final Report.

Containment Spray Atmospheric Dispersion NUREG/CR-5966 A SIMPLIFIED MODEL OF ALFIOSOL REMOVAL BY NUREG/CR5247 V02 RASCAL VERSION 2 0 WORKBOOK- CONTAINMENT SPRAYS.

NUREG/CR-5982- EFFECTIVENESS OF CONTAINME.NT SPRAYS IN Austenttic Stainless Steel CONTAINMENT MANAGEMENT.

NUREG/CR-5999 tNTERlM F ATIQUE DESIGN CURVES FOR CARBON. ,

LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENV). Control Valve j RONMENTS. NUREG/CR-6031: CAVITATION GUtDE FOR CONTROL VALVE S.

BIGFLOW Coo 6down Trans6ent NUREG/CR 6028 DIGFLOW. A NUMERICAL CODE FOR SIMULATING NUREG/CR-5983. SAFETY ASPECTS OF FORCED FLOW COOLDOWN FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED MEDIA Theory And User's Manual - Verseon 11. REACTORS BWR Corroston Fatigue NUREGICR-5882. TRAC-B THERMAL. HYDRAULIC ANALYSIS OF THE NUREG/CR-4667 V15 ENVIRONMENT ALLY ASSISTED CRACKING IN DLACK FOX BOILING WATER REACTOR. LIGHT W ATER REACTORS. Semiannual Report. April-Septembnr 1992.

Blautal Lond6ng Crack NUREG/CR-6036 INITIAL RESULTS OF THE INFLUENCE OF B1AX1AL NUREG/CR-4599 V02 N2 SHORT CRACKS IN PIPING AND PIPING LOADING ON FRACTURE TOUGHNESS. WELDS Semsannual Report, Octoter 1991 March 1992.

23

- - - - -. .. =- . - . .

24 Subject index Cracking Fatique Design Curve NUREG/CR 4667 V15. ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5999 INTERIM FATIOUE DESIGN CURVES FOR CARBON, LIGHT WATER REACTORS Semiannual Report,Apnt September 1992. LOW. ALLOY. AND AUSTENITIC STAINLESS STEELS IN LWR ENV8-RONMENTS DEPOSITION Computer Code NUREG/GR-0006. DEPOSITION SOFTWARE TO CALCULATE PARTI- Feed une Break CLE PENETRATION THROUGH AEROSOL TRANSPORT NUREG/lA-0104: RELAP5/ MOD 3 ASSESSMENT USING THE SEMIS-SYSTEMS Final Report CALE 50% FEED LINE BRE AK TEST S-FS-11.

DOE Waste Package Fiscal Year NUREG/CR-4735 V08. EVALUATION AND COMPILATION OF DOE NUREG 1100 V09 BUDGET ESTIMATES Fiscal Years 1994-1995.

WASTE PACKAGE TEST DATA B6 annual Report, August 1989 - Janu-ary 1990. Forced Entry i NUREG 1485: UNAUTHORIZED FORCED ENTRY INTO THE PROTECT-Damping ED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7,1993 j NUREG1CR-5770 DAMPING IN LOW-ASPECT-R ATIO. REINFORCE D .

CONCRETE SHEAR WALLS Fracture Analyses l NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF Data Collection LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS NUREG/CR 5471: ENHANCEMENTS TO DATA COLLECTION AND RE- (PROJECT FALSIRE).

PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

Fracture Mechanics Decommissioning NUREG/CR-4599 V02 N2. SHORT CRACKS IN PlPING AND PIPING NUREG-1307 R03 REPORT ON WASTE BURIAL CHARGES Escalation WELDS Semiannual Report, Octots 1991 - March 1992 Of Decommmsioning Waste Disposal Costs At Low-level Waste Bunal NUREG/CR4410: STATISTICALLY BASED REEVALUATION OF PISC41 l Facanes. ROUND ROBIN TEST DATA. I NUREG/CR-5999 INTERIM FATIQUE DESIGN CURVES FOR CARBON, ,

NUREG/CR 5894 RADIONUCLIDE CHARACTERIZATION OF REAC.

TOR DECOMMISSIONtNG WASTE AND NEUTRON-ACTIVATED LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI-METALS RONMENTS Decontamination Fracture Toughness NUREG/CR-5966. A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF ,

CONTAINMENT SPRAYS LARGE SCALE INTERNATIONAL REFERENCE EXPERIMENTS (PROJECT FALSIRE1 Depressurtration NUREG/CR 6036: INITIAL RESULTS OF THE INFLUENCE OF BIAXIAL 4 v

NUREG/CR 5937. INTENTIONAL DEPRESSURIZATION ACCIDENT LOADING ON FRACTURE TOUGHNESS.

MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC- i TORS.

Fractured Media NUREG/CR-5817 V03 N1: NRC H!GH-LEVEL RADIOACTIVE WASTE Design Bases RESEARCH AT CNWRA. January-June 1992 NUREG/CR 5955 MATERlALS AND DESIGN BASES ISSUES IN ASME Fuel Peak Cladding CODE CASE N47.

NUREG/CR-6061: DETERMINATION OF THE BLAS IN LOFT FUEL Design Basis Event PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN i NUREG/CR 6027: PRELIMINARY EVALUATION OF SNUBDER SINGLE PHASE OF LARGE-BREAK LOCA EXPERIMENTS r FAILURES '!

Genochemistry Disposal unit Source Term NUREG/CR-5817 V02 NRC HIGH-LEVEL RADIOACTIVE WASTE RE-NUREGICR4041 - DISPOSAL UNIT SOURCE TERM (DUST) DATA SEARCH AT CNWRA. Calendar Year 1991. [

INPUT GUIDE Generic Safety issue i Dose Assessment NUREG-0933 SIS: A PRIORiTIZATION OF GENERIC SAFETY ISSUES.  :

NUREG/CR-5247 V02: RASCAL VERSION 2 0 WORKBOOK i Embrittlement NUREG-1364: REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREGICR-4744 V07 N1. LONG-TERM EMBRITTLEMENT OF CAST GENERIC SAFETY ISSUE 106. PIPING AND THE USE OF HIGHLY  !

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Som> annual COMBUSTIBLE GASES IN VITAL AREAS Report. October 1991 - March 1992.

Geologic Media l Emergency Preparedness NUREG/CR 6028' BIGFLOW: A NUMERICAL CODE FOR SIMULATING i NUREG-1474: EFFECT OF HURRICANE ANDREW ON THE TURKEY FLOW IN VARIABLY SATURATED, HETEROGENEOUS GEOLOGIC .

+

POINT NUCLE AR GENERATING STATION FROM AUGUST 20-30, MEDIA Theory And User's Manual - Version 1.1.

1992.

NUREG-1485. UNAUTHORIZED FORCED ENTRY INTO THE PROTECT.

HTGR Type Reactor +

ED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7,1993 NUREG/CR-5983: SAFETY ASPECTS OF FORCED FLOW COOLDOWN TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED j Energy Dissipation REACTORS.

' NUREG/CR 5984. CODE AND MODEL EXTENSIONS OF THE THATCH NUREG/CR-6031:CAVIT ATION GUIDE FOR CONTROL VALVES ,

CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLED REAC- i Enforcement Action TORS NUREG 0940 V12 N01: ENFORCEMENT ACTIONS.SIGNIFICANT AC.

TIONS RESOLVED Ouarterly Progrea Report,,lanuary-March 1993. Health Effect p

~

NUREG/CR4214 R1P2A2 HEALTH EFFECTS MODELS FOR NUCLE-Event Analysis AR POWER PLANT ACCIDENT CONSEQUENCE ,

NUREG/CR-5976: DEVELOPMENT AND USE OF A TRAIN-LEVEL ANALYSIS Modificabon Of Models Resulting From Additon Of Effects PROBABILISTIC RISK ASSESSMENT, Of Exposure To Alpha-Emrtting Radionuclides.Part il Scentific Bases i For Health..

l Expert System

, NUREG/CR401B. SURVEY AND ASSESSMENT OF CONVENTIONAL Heat Transfer SOFTWARE VERIFICATION AND VALIDATION METHODS. NUREG/CR-5982. EFFECTIVENESS OF CONTAINMENT SPRAYS IN  !

CONTAINMENT MANAGEMENT. 4 Failure Event NUREG/CR 5983 SAFETY ASPECTS OF FORCED FLOW COOLDOWN i NUREG/CR-5471 ENHANCEMENTS TO DATA COLLECTION AND RE- TRANS!ENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED [

PORilNG OF SINGLE AND MULTIPLE FAILURE EVENTS. REACTORS ,

1

l Subject index 25 NUREG/CR 5984 CODE AND MODEL EXTENSIONS OF THE THATCH Information Circular CODE FOR MODULAR HISH TEMPERATURE GAS-COOLED REAC- NUREG-0725 A09: PUBLIC INFORMATION CIRCULAR FOR SHIP-TORS. MENTS OF IRRADIATED REACTOR FUEL Heat Treatment information Digest NUREG/CR4972. EFFECTS OF NONSTANDARD HEAT TREATMENT NUREG-1350 V05. NUCLEAR FtEGULATORY COMMISSION INFORMA-TEMPERATURES ON TENS:LE AND CHARPY IMPACT PROPERTIES TION DIGEST.1993 Edition.

OF CARBON-STEEL CASTING REPAIR WELDS inspection Guide High Pressure Coolant injection System NUREG/CR4014. HIGH PRESSURE COOLANT INJECTION SYSTEM NURE G/CR4014 HIGH PRESSURE COOLANT INJE.CTION SYSTEM RISK-BASED INSPECTION GUIDE FOR HATCH NUCLE AR POWER RIS0 BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER STATION.

STATION Integrated Risk Assessment G CR / 02: N C HIGH-LEVEL RADIOACTIVE WASTE RE- LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NURE /CR 581 Risk Uncenainy baluahon Wogram NER Apnbs M 03 L EL RADIOACTIVE WASTE RESEARCH Al CNWRA January-June 1992' NUREG/CR-5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Human Rehability Analysis Risk Uncertainty Evaluaton Program (PRUEP) Appendices D-G.

NUREG/CR-4832 V05 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR Internal Fire Analysis POWER PLANT. RISK METIIODS INTEGRATION AND EVALUATION PROGRAM Parameter Eshmahon Analysis And Screening Human Reh- NUREG/CR-4P32 V09: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR atukty Analysis POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION PROGRAM (RMiEP)Intemal Fire Analysis. l Hurricane Andrew NUREG-1474 EFFECT OF HURRICANE ANDREW ON THE TURKEY #

POINT NUCLEAR GENERATING STATION FROM AUGUST 20-30, NUREG/CR-6061: DETERMINATION OF THE BIAS IN LOFT FUEL 1992. PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN PH.ASE OF LARGE-BREAK LOCA EXPERIMENTS.

Hydrogeology NUREG/CR-5817 V03 N1: NRC HIGH-LEVEL RADIOACTIVE WASTE LOFT Test L5-1 RESEARCH AT CNWRA. January June 1992 NUREG/lA-0118. ANALYSIS OF LOFT TEST LS-1 USING RELAP5/ .

MOD 2 ICAP Program NUREG/lA 0090: ASSESSMENT OF RELAP5/ MOD 2 USING THE TEST LSTF Data SB-CL-18 DATA OF REWET-il REFLOODING EXPERIMENT SGl/R. NUREG/lA-0095. RELAPS ASSESSMENT USING LSTF TEST DATA SB-NUREG/lA-0092: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE CL-16 AGAINST THE NET LOAD T RIP TEST DATA FROM YONG-GWANG, UNIT 2 LWR NUREG/lA-0094. ASSESSMENT OF RELAP5/ MOD 3 AGAINST NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN TWENTYSIVE POST-DRYOUT EXPERIMENTS PERFORMED AT THE LIGHT WATER REACTORS. Serruannual Report.ApnLSeptember 1992.

ROYAL 'NSTITUTE OF TECHNOLOGY, NUREG/CR-5999. INTERIM FATIQUE DESIGN CURVES FOR CARBON, i NUREG/lA-0095. RELAPS ASSESSMENT USING LSTF TEST DATA SB- LOW-ALLOY, AND AUSTENTIC STAINLESS STEELS IN LWR ENVI-CL 18- RONMENTS.

NUREG/lA-0099. RELAP5 ASSESSMENT USING SEMISCALE SBLOCA NUREG/CR-4744 V07 N1: LONG-TERM EMBRITTLEMENT OF CAST TEST S NH-t- DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUREG/1A-0104- RELAPS/ MOD 3 ASSESSMENT USING THE SEMtS Report,0ctober 1991 - March 1992.

CALE 50% FEED LINE DREAK TEST SFS 11.

NUREG/lA-0105. ASSESSMENT OF RELAP5/ MOD 3 VERSION SMS Large-Break LOCA USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF NUREG/CR-6061: DETERMINATION OF THE BIAS IN LOFT FUEL NURE / A 01 6 SSESSMENT OF PWR STEAM GENERATOR MOD- HASE O RGE BREAK L A EX R MENTS ELLING IN RELAPS/ MOD 2.

NUREG/lA-0118. ANALYSIS OF LDFT TEST LS 1 USING RELAPS/ Legal issuances NU EG/lA-0119. ASSESSMENT AND APPLICATION OF BLACKOUT NUREG-0750 V36101: INDEXES TO NUCLEAR REGULATORY COM-TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP51 NU GO V3 I IN E S TO CLEAR REGULATORY COM-

^

NC S NPP W H HA BF NU V37 NO2 C EG LATORY COMMISSION IS NUREG/lA 0122: ASSESSMENT OF MSiv FULL CLOSURE FOR SUANCES FOR FEBRUARY 1993. Pages 55134 SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING NUREG-0750 V37 NO3- NUCLEAR REGULATORY COMMISSION IS-TR AC-BF 1 (G1J1). SUANCES FOR MARCH 1993, Pages 135-249.

NUREG/lA-0123. APPLICATION OF FULL POWER BLACKOUT FOR C N ALMARAZ WITH RELAPS/ MOD 2 Licensed Fuel facility Status Report NUREG/lA-0124: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES- NUREG-0430 V12- LICENSED FUEL FACILITY STATUS SUR12ER SPRAY VALVE INADVERTED FULLY OPENING TRAN. REPORT. inventory Difference Data. July 1,1991 - June 30.1992.(Gray SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA- OCCk IU BRERA NUCLEAR STATION NUREG/tA-0125: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE Light Water Reactor AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG. NUREG/CR-4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN GWANG UNIT 2 LIGHT WATER REACTORS. Semiannuat Report Aprd-September 1992.

NUREG/CR 4744 V07 N1: LONG-TERM EMBRITTLEMENT OF CAST ICAP Report DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual NUREG/iA-0116 ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST Report. October 1991 - March 1992.

THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWR NUREG/CR-5999 INTERIM FATIOUE DESIGN CURVES FOR CARBON, HOT LEG). LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVI.

RONMENTS Inadvertent Safety injection NUREG/lA-0105: ASSESSMENT OF RELAPS/ MOD 3 VERSION SM5 Liquidus Temperature USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF NUREG/CR 6032: SOLIOUS AND LlOUIDUS TEMPERATURES OF KORI UNIT 3 PLANT CORE-CONCRETE MIXTURES.

_- - _ ~ . . - -. .. - . . -

26 Subject index I

Load Rejection Transient Piping NUREG/lA-0109. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10% NUREG-1364 REGULATORY ANALYSIS FOR THE RESOLUTION OF -l LOAD REJECTION TRANSIENT FROM 75's STEADY STATE IN THE GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY VANDELL OS 11 NUCLEAR POWER PLANT. COMBUSTIBLE GASES IN VITAL AREAS.

NUREG/CR-A599 V02 N2: SHORT CRACKS IN PIPING AND PIPING ,

N Al /lA 0092: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE AGAINST THE NET LOAD TRIP TEST DATA FROM YONG' Plugging Criterla GWANG, UNIT 2 NUREG-1477 DRFT FC: VOLTAGE BASED INTERIM PLUGGlNG CRITE-RIA FOR STEAM GENERATOR TUBES Draf1 Report For Comment.

Loss-Of Coolant NUREG/CP4)126 V01: PROCEEDINGS OF THE TWENTIETH WATER Pract6ce And Procedure Digest REACTOR SAFETY INFORMATION MEETING NUREG-0386 D06 R06: UNITED STATES NUCLEAR REGULATORY NUREG/CP 0126 V02; PROCEEDINGS OF THE TWENTIETH WATER COMMISSION STAFF PRACTICE AND PROCEDURE {

REACTOR SAFETY fNFORMATION MEETING DIGEST. Commission.Appea! Board And L censing Board Decisions July NUREG/CP-0126 V03. PROCEEDINGS OF THE TWENTIETH WATER 1972 - June 1992. I REACTOR SAFETY INFORMATION MEETING. 1 Pressurtred Water Reactor i Low-Level Radioactive Waste NUREG/CR-6041: DISPOSAL UNIT SOURCE TERM (DUST) DATA NUREG/CR-5759 RISK ANALYS!S OF HIGHLY COMBUSTIBLE GAS INPUT GUIDE- STORAGE. SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR- ,

lZED WATER REACTOR PLANTS.

MSIV NUREG/CR-5937: INTENTIONAL DEPRESSURIZATION ACCIDENT NUREG/lA-0122. ASSESSMENT OF MSIV FULL CLOSURE FOR MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC-  ;

SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING TORS' TRAC-DF 1 (G1J1).

Pressurizer Spray Vafve f 4 Diatural Circulation NUREG/lA-0124: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES- l NUREG/lA 0124: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A PRES. 1 SURIZER SPRAY VALVE INADVERTED FULLY OPEfMG TRAN-SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN. S1ENT AND RECOVERY BY NATURAL CtRCULATION ;N JOSE CA-SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA.

BRERA NUCLEAR STATION.

BRERA NUCLEAR STATION NUREG/lA-0125: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE Probabmetic Risk Assessment .

AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG-NUREG/CR-4832 V05 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR GWANG UNIT 2.

POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION #

Nondestructive Evaluation PROGRAM. Parameter Estimation Analysis And Screening Human Reh- .

3 4

NUREGICR-5410. STATISTICALLY BASED REEVALUATION OF PISC-Il abthty Analysis.

, ROUND ROBIN TEST DATA. NUREG/CR 4B32 V09: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR 1 POWER PLANT: AtSK METHODS INTEGRATION AND EVALUATION Nuclear Regulatory Research PROGRAM (RMIEP)Intemal Fire Analysis.

NUREG 1266 V07: NRC SAFETY RESEARCH IN SUPPORT OF REGU- NUREG/CR-5471: ENHANCEMENTS TO DATA COLLECTION AND RE- ,

LATION - FY 1992 PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS.

NUREG/CR-5759. RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS

" " "~

N R G/C 4735 V08. EVALUATION AND COMPILATION OF DOE ED kR R C P TS WASTE PACKAGE TEST DATA. Biannual Report. August 1989 Jana NUREG/CR-5976. DEVELOPMENT AND USE OF A TRAIN-LEVEL r ary im PROBABILISTIC RISK ASSESSMENT Office Of The inspector General Probabilistic Safety Analysis NUREG 1415 V05 NO2 OFFICE OF THE INSPECTOR NUREG/CR-5801: PROCEDURE FOR ANALYSIS OF COMMON-CAUSE GENERAL. Semiannual Report. October 1.1992 - March 31,1993.

FAILURES IN PROBABILISTIC SAFETY ANALYSIS.

+

PISC-Il NUREG/CR-5410 STATISTICALLY BASED REEVALUATJON OF P)SC.}l Protected Area ROUND ROBIN TEST DATA. NUREG-1485: UNAUTHORIZED FORCED ENTRY INTO THE PROTECT-ED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7,1993. ,

PRUEP 1 NUREG/CR-5305 V02 P1: INTEGRATED RISK ASSESSMENT FOR THE RASCAL l i

LASAtlE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And NUREG/CR 5247 V02: RASCAL VERSION 2.0 WORKBOOK, Risk Uncertamty Evaluation Program (PRUEP) i ndees A-C.

NUREG/CR 5305 V02 P2: INTEGRATED RISK A ESSMENT FOR THE RELAP5  :

LASALLE UNIT 2 NUCLEAR POWER PLANT:Phonemenology And NUREG/lA-0095 RELAPS ASSESSMENT USING LSTF TEST DATA SB- ,

Rok Uncertainty Evaluation Program (PRUEP). Appendices D-G. CL 18.

l NUREG/lA-0099: RELAPS ASSESSMENT USING SEMISCALE SBLOCA i

UREG/CR 5759: RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS STORAGE. SUPPLY. AND DtSTRtBUTION SYSTEMS IN PRESSUR- RELAPS Computer Code (ZED W ATER REACTOR PLANTS. NUREG/CR-6035: FEAStBILITY STUDY FOR IMPROVED STEADY-NUREG/CR 5937: INTENTIONAL DEPRESSURIZATION ACCIDENT STATE INITIALIZATION ALGORfTHMS FOR THE RELAP5 COMPUT-d MANAGEMENT STRATEGY FOR PRESSURIZED WATER REAC- ER CODE.

TORS 002 i Parameter Estimation Analysis NUREG/lA-0090: ASSESSMENT OF RELAP5/ MOD 2 USING THE TEST j NUREGICR-4832 V05 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR DATA OF REWET il REFLOODING EXPERIMENT SGl/R.

POWER PLANT: R!SK METHODS INTEGRATION AND EVALUATION NUREG/lA-0092: ASSESSMENT OF RELAP5/ MOD 2 COMPUTER CODE PROGRAM Parameter Estimation Analysis And Screenmg Human Reli- AGAINST THE NET LOAD TRIP TEST DATA FROM YONG- i aWy AnaWs GWANG. UNIT 2. I NUREG/lA-0106: ASSESSMENT OF PWR STEAM GENERATOR MOD-Particle Penetration ELLING IN RELAPS/ MOD 2.

NUREG/GR-0006: DEPOSITION: SOFTWARE TO CALCULATE PARTI. NUREG/lA-0108: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A TUR-CLE PENETRATION THROUGH AEROSOL TRANSPORT SYSTEMS Final Report. BINE TRIP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR POWER PLANT.

Petitions For RulemaMng NUREG/lA-0109 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10%

NUREG-0936 V12 N01: NRC REGULATORY AGENDA.Ouarterty LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN THE Report. January-March 1993 VANDELLOS ll NUCLEAR POWER PLANT.

l

. --. ._ - - - - , - . ~ . . . . . .

Subject Index 27 NUREG/lA.0118. ANALYSIS Of LOFT TEST LS-1 USING RELAP5/ NUREG/CP-0126 V03' PROCEEDINGS OF THE TWENTIETH WATER MOD 2 REACTOR SAFETY INFORMATION MEETING.

NUREG/lA-0119 ASSESSMENT AND APPLICATION OF BLACKOUT TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAP5/ Referen:e Espertment MOD 2. NUREG/CR-5997. CSNI PROJECT FOR FRACTURE ANALYSES OF ,

NUREG/lA4123 APPLICATION OF FULL POWER BLACKOUT FOR LARGE-SCAL E INTERNATIONAL REFERENCE EXPER:MENTS I C N ALMARAZ WITH RELAPS/ MOD 2 (PROJECT FALSIRE).

NUREG/lA-0124 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A PRES- ]

SURIZER SPRAY VALVE INADVERTED FULLY OPENING TRAN- Regulatory Agenda  ;

j SIENT AND RECOVERY BY NATURAL CIRCULATION IN JOSE CA- NUREG 0936 V12 N01. NRC REGULATORY AGENDA Ouarterly i ORERA NUCLEAR STATION Report. January-March 1993 NUREG/lA-0125: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER CODE AGAINST THE NATURAL CIRCULATION TEST DATA FROM YONG- Regulatory And Technical Report $

GWANG UNIT 2. NUREG 0304 V18 N01: REGULATORY AND TECHNICAL REPORTS ]

(ABST RACT INDEX JOURNAL) Compilatson For Fwst Ouarter i FIELAP5/ MOD 3 1993, January March. I NUREG/CR-6061 DETERMINATION OF THE BIAS IN LOFT FUEL s PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN Reinforced Concrete 'i PHASE OF LARGE-BRE AK LOCA EXPERIMENTS NUREG/CR-5776: DAMPING IN LOW-ASPECT-RATIO. REINFORCED

, NUREG/lA-0094- ASSESSMENT OF RELAPS/ MOD 3 AGAINST CONCRETE SHE AR WALLS. i TWENTY-FIVE POST-DRYOUT EXPERIMENTS PERFORMED AT THE ROYAL INSTITUTE OF TECHNOLOGY Fleport To Congress NUREG/1A-0104 RELAPS/ MOD 3 ASSESSMENT USING ThE SEMIS- NUREG-0090 V15 N04 REPORT TO CONGRESS ON ABNORMAL i CALE 50% FE E D LINE BRE AK TEST S-FS 11 OCCURRENCES.0ctoter Decemtser 1992.

NUREG/iA-0105 ASSESSMENT OF RELAP5/ MOD 3 VERSION SMS i USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF Risk Methods integration KORI UNIT 3 PLANT. NUREG/CR-4832 V09. ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION URE I 0116 ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST

" * ^""

• THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWR Rules HOT LEG) NURE G-0936 V12 N01 - NRC RE GULATORY AGENDA.Ouarterly j REWET-il Reflood4ng Experiment ReWanuan Mad W NUREG/lA-0090 ASSESSMENT OF RELAP5/ MOD 2 USING THE TEST Rules Of Practice DATA OF REWET il REFLOODING EXPERIMENT SG1/R NUREG-0386 DOB R06: UNITED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE j Radiat6on '

DIGEST.Comrnission, Appeal Board And Licensing Board Decisions July NUREG/CR-4214 RIP 2A2: HEALTH EFFECTS MODELS FOR NUCLE-AR POWER ACCIDENT -

1972 + June 1992.

PLANT CONSEQUENCE ANALYSIS Modificahon Of Models Resulting From Addition Of Effects Safety Evaluation Report '

Of Exoosure To Alpha Emethng Radionuchdes Part 11- Scientific Bases NUREG-0797 S27: SAFETY EVALUATION REPORT RELATED TO THE I 6 Hea h OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION, {

Radiation Transport UNIT 2 Docket No 50-446 (Texas Uhhteen Electnc Compa et at ) '

WR 0847 S10 SAHW Mpp WW W W M M  ;

NUREG/CR4247 V02 RASCAL VERSION 2 0 WORKDOOK OPERATION OF WATTS BAR NUCLEAR PLANT,0 NITS 1 AND j Fladlonuclide 2. Docket Nos. 50-390 And 50-391.(Tennesee Valley Authonty)  ;

NURE0 /CR 5694 RADIONUCLIDE CHARACTEntZATION OF REAO DECOMM;SSIONING WASTE AND NEUTRON-ACTiVAT ED 88 ][G/ 01 V01 PROCE'EDINGS OF THE TWENTIETH WATER REACTOR SAFETY INFORMATION MEETING.

Radionuci'le Migration NUREG/CP-0126 V02: PROCEEDINGS OF THE TWENTIETH WATER NUREG/CA6041 DtSPOSAL UNIT SOURCE TERM (DUST) DATA REACTOR SAFETY INFORMATION MEETING. r Safety Research Reactor Acetdent NUREG 1266 V07: NRC SAFETY RESEARCH IN SUPPORT OF REGU- ,

, NUREG/CP-0126 V01: PROCEEDINGS OF THE TWENTIETH WATER LATION - FY 1992.

REACTOR SAFE TY INFORMATION MEETING NUREG/CP-0126 V02 PROCEE DINGS OF THE TWENTIETH WATER Safety informat6on  ;

REACTOR SAFETY INFORMATION MEETING. NUREG/CP-0126 V03 PROCEEDINGS OF THE TWENTIETH WATER NUREG/CP 01?6 V03. PROCEEDtNGS OF THE TWE NTIETH WATER REACTOR SAF ETY INFORMATION MEETING REACTOR SAFETY INFORMATtON MEETING.

NUREG/CR 59a2, EFFECTIVE NESS OF CONTAINMENT SPRAYS IN Security Event CONTAINMENT MANAGEMENT. NUREG 1485. UNAUTHORIZED FORCED ENTRY INTO THE PROTECT. [

ED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7.1993.

Reactor Coollng System

• NUREG/CR 5983. SAF ETY ASFECTS OF FORCED FLOW COOLDOWN Seismic Analysis TRANS:ENTS IN MODULAR HIGH TEMPERATURE GAS-COOLED NUREG/CR-6013. METHODS USED FOR THE TREATMENT OF NON-- ,

REACTORS PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS  !

NUREG/CR-5984 CODE AND MODEL EXTENSIONS OF THE THATCH CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLED REAC- Se6sm6c Event TORS NUREG/CR-6027; PRCtJMl NARY EVALUATION OF SNUBBER SINGLE i F AILURES. .

Reactor Pressure Vessel i NUREG/CR 5410 STATISTICALLY BASED RFE VALUATION OF PISO.tl Semiscale S-FS-11 ROUND ROBIN TEST DATA. NUREG/lA-0104: RELAPS/ MOD 3 ASSESSMENT USING THE SEMIS- )

NUREG/CR 5997: CSNI PROJECT FOR FRACTURE ANALYSES OF CALE 50% F EED LINE BREAK TEST S-FS 11.  !

LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS

{ PROJECT FALSIRE). Semiscale SBLOCA NUREG/lA-0099: RELAP5 ASSESSMENT USING SEMISCALE SBLOCA Flesctor Sately TEST S-NH-1.

NUREG/CP 0126 V01: PROCEEDINGS OF THE TWENTIETH WATER  !

RE ACTOR SAFETY INFORMATION MEETING Severo Reactor Accident ,

NUREG/CP-0126 V02 PROCEEDINGS OF THE TWENTIETH WATER NUREG/CR-5906: A SIMPLIFIED MODEL OF AEROSOL REMOVAL BY ,

RE ACTOR SAF ETY INFORMATION MEETING. CONTAINMENT SPRAYS

28 Subject Index Shear Wall TR AC-BF1 NUREG/CR-5776, DAMFiNG IN LOW-ASPECT.R ATIO. REINFORCED NUREG/CR 5882: TRAC-B THERMAL-HYDRAULIC ANALYSIS OF THE CONCRETE SHEAR WALLS BLACK FOX BOIUNG WATER REACTOR NUREG/lA 0120: ASSESSMENT OF THE TURBINE TRIP TRANSIENT Shipment IN COFRENTES NPP WITH TRAC-BF1.

NUREG-0725 A09 PUBLIC INFORMATION CIRCULAR FOR SHIP- NUREG/lA-0122: ASSESSMENT OF MSiv FULL CLOSURE FOR MENTS OF 1RRADIATED REACTOR FUEL. SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING TRAC-BF1 (G1J1).

NUREG/CR-6027: PREUMl NARY EVALUATION OF SNUBBER SINGLE ThermaFHydraulic F AILURE S. NUREG/CR-5882 TRAC-B THERMAL-HYDRAULIC ANALYSIS OF THE Software Verification BLACK FOX BOILING WATER REACTOR.

NUREG/CR-6018 SURVEY AND ASSESSMENT OF CONVENTIONAL Thermohydraulic SOFTWARE VERIFICATION AND VALIDATION METHODS NUREG/CR-5817 VC2 NRC HIGH-LEVEL RADIOACTIVE WASTE FIE-Solidus Temperature SEARCH AT CNWRA Calendar Year 1991. l NUREG/CR-6032- SOLIDUS AND LIOUiDUS TEMPERATURES OF Dennoluminescent Dosimeter I

CORE-CONCRETE MIXTURES. l NUREG-0837 V13 N01: NRC TLD DIRECT RADIATION MONITORING Spent f uel NETWORK. Progress Report January-March 1993 f NUREG-0725 R09 PUBUC INFORMATION CIRCULAR FOR SHIP-  !

D3' U i MENTS OF IRRADIATED REACTOR FUEL NUREG/CR-5894 RADIONUCLIDE CHARACTER 12ATION OF REAC. NUREG-0540 V15 N02: TITLE LIST OF DOCUMENTS MADE PUBLICLY TOR DECOMMISSIONING WASTE AND NEUTRON ACTIVATED AVAILABLE. February 1 46,1993.

METALS. NUREG-0540 V15 NO3. TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE Marcn 1 31,1993.

Stainless Steel NUREG-0540 V15 N04. TITLE UST OF DOCUMENTS MADE PUBLICLY ,

NUREG/CR-4744 V07 N1 LONG-TERM EMBRITTLEMENT OF CAST AVAILABLE.Apnl 1-30,1991 r DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual Report, October 1991 - March 1992. Tube Degradat6on NUREG-1477 DRFT FC: VOLTAGE-BASED INTERlM PLUGGING CRITE-Steady-State Algortthms RlA FOR STEAM GENERATOR TUBES Drait Report For Comment.

NUREG/CR-6035. FEAStBILITY STUDY FOR IMPROVED STEADY-STATE INITIAUZATION ALGORITHMS FOR THE RELAPS COMPUT. Turbine Trip ER CODE. NUREG/lA-0108. ASSESSMENT OF RELAP5/ MOD 2 AGAINST A TUR- '

BINE TRIP FROM 100% POWER IN THE VANDELLOS 11 NUCLEAR Steam Generstor POWER PLANT.

l NUREG-1477 DRFT FC- VOLTAGE-BASED INTERIM PLUGGING CRITE-E RiA FOR STEAM GENERATOR TUBES Draft Report For Cornment Turbine Trip Transient NUREG/iA 0106. ASSESSMENT OF PWR STEAM GENERATOR MOD. NUREG/lA-0120 ASSESSMENT OF THE TURBINE TRIP TRANSIENT ,

ELLING IN RELAPS/ MOD 2 IN COFRENTES NPP WITH TRAC-BF1 SteelContainment UPTF Test 11 i

NUREG/CR-5957: SYSTEM 80 4 (TM) CONT AINMENT - STRUCTURAL NUREG/lA-0116- ASSESSMENT OF RELAPS/ MOD 3/V5M5 AGAINST DESIGN REVIEW. THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN PWR ,

Stress Corrosion NUREGICR-4735 V08. EVALUATION AND COMPILATION OF DOE Unsaturated Flow WASTE PACKAGE TEST DATA. Biannual Report. August 1989 - Janu- NUREG/CR4028: BIGFLOW: A NUMERICAL CODE FOR SIMULATING ary 1990 FLOW IN VARIABLY SATURATED. HETEROGENEOUS GEOLOGIC

• "'~

Stress Corrosion Cracking NUREG-1477 DRFT FC. VOLT AGE-BASED INTERIM PLUGGING CRITE- Vendor inspection RlA FOR STEAM GENERATOR TUBES Draft Report For Comment NUREG-0040 V17 N01: UCENSEE CONTRACTOR AND VENDOR IN-Structural Design SPECTION STATUS REPORT. Quartedy Report, January-March

NUREG/CR-5957. SYSTEM 80 4 [1M) CONTAINMENT -- STRUCTURAL I9 8b DESIGN REVIEW Viscometry Structural System NUREG/CR-6032. SOLIDUS AND LIOUlDUS TEMPERATURES OF NUREG/CR-6013 METHODS USED FOR THE TREATMENT OF NON. CORE-CONCRETE MIXTURES PROPORTIONALLY DAMPED STRUCTU9AL SYSTEMS Vital Gas Structure Damping NUREG-1364 REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG/CR-6013 METHODS USED FOR THE TREATMENT OF NON. GENERIC SAFETY ISSUE 106 PIPING AND THE USE OF HIGHLY PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS. COMBUSTIBLE GASES IN VITAL AREAS.

System 80, Waste Burial NUREG/CR 5957: SYSTEM 80 +(TM) CONT AINMENT - STRUCTURAL NUREG-1307 R03 REPORT ON WASTE BURIAL CHARGES Escalation DESIGN REVIEW Of Decommissioning Waste Disposal Costs At Low-Level Waste Bunal Facilitses System Train i

NUREG/CR-5976: DEVELOPMENT AND USE OF A TR AIN-LEVEL Weld l PROBABILISTIC RISK ASSESSMENT. NUREG/CR-4599 V02 N2: SHORT CRACKS IN PIPING AND PIPING WELDS Sermannual Report. October 1991 - March 1992.

l l THATCH Code NUREG/CR-5972s EFFECTS OF NONSTANDARD HEAT TREATMENT NUREGICR-59B4. CODE AND MODEL EXTENSIONS OF THE THATCH TEMPERATURES ON TENSILE AND CHARPV IMPACT PROPERT!ES CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLED REAC- OF CARBON-STEEL CASTING REPAIR WELDS TORS Yucca Mountain TLD NUREG/CR-4735 V0B. EVALUATION AND COMPILATION OF DOE NUREG 0337 V13 N01. NRC TLD DIRECT RADIATION MONITORING WASTE PACKAGE TEST DATA. Biannual Report August 1989 - Janu-NETWORK. Progress Report, January-March 1993. ary 1990.

i

i l

NRC Originating Organization index (Staff Reports)

This index lists those NRC organizations that have published staff reports. The index is ar- ,

ranged alphabetically by major NRC organizations (e.g., program offices) and then by sub-sections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number. l i

QDVISORY COMMITTEE (S) EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS ACRS - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG-1125 V14. A COMPILATION OF REPORTS OF THE ADVISO. NUREG-0430 V12; LICENSED FUEL FACILITY STATUS RY COMMITTEE ON REACTOR SAFEGUARDS 1992 Annual REPORT. inventory Difference Data. July 1, 1991 - June 30, 1992 (Gray Book 11)

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) DIVISION OF SAFEGUARDS & TRANSPORTATION (870413-930206)

OFC OF THE EXECUTIVE DIRECTOR FOR OPERATIONS NUREG-0725 R09: PUBLIC INFORMATION CIRCULAR FOR SHIP-NUREG-1485 UNAUTHORIZED FORCED ENTRY INTO THE PRO- MENTS OF IRRADIATED REACTOR FUEL TECTED AREA AT THREE MILE ISLAND UNIT 1 ON FEBRUARY 7' GEOLOGY & ENGINEERING BRANCH (POST B10506) ggg3. NUREG/CR-4735 V08 EVALUATION AND COMPILATION OF DOE REGION 1 (POST 820201) WASTE PACKAGE TEST DATA. Baannual Report, August 1989 - Jan-NUREG-0837 V13 N01: NRC TLD DIRECT RADIATION MONITORING uary 1990.

OFC F f RC E T (POST 8704 3 U.S. NUCLEAR REGULATORY COMMISSION NUREG4940 V12 NC1. ENFORCEMENT ACTtONS SIGNIFICANT AC- N EG 0386 R U TED N L R REGULATORY TIONS RESOLVED.Ouarterty Progress Report. January-March 1993- COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST. Commission. Appeal Board And Licensing Board EDO - OFFICE OF ADMINISTRATION (PRE 870413 & POST 890205)

DIVISION Of FREEDOM OF INFORMATION & PUBLICATIONS SERV- OFF C O ik i C NERAL (POST 890417)

N EG-O V1 01: REGULATORY AND TECHNICAL REPORTS ' *

(ABSTRACT INDEX JOURNAL) Compilation For First Quarter **"""" # "'

1993. January-March. goo . OFFICE OF NUCLEAR REGULATORY RESEARCH (POST B20405)

NUREG-0540 V15 NO2: TITLE LIST OF DOCUMENTS MADE PUBLIC- OFFICE OF NUCLEAR REGULAlORY RESEARCH (POST 860720)

LY AVAILABLE February 1-28.1993. NUREG-1266 V07: NRC SAFETY RESEARCH IN SUPPORT OF REG-NUREG-0540 V15 NO3: TITLE LIST OF DOCUMENTS MADE PUBLIC- ULATION - FY 1992.

LY AVAILABLE. March 1-31,1993 OfVISION OF REGULATORY APPLICATIONS (POST S70413)

NUREG-0540 V15 N04 TITLE LIST OF DOCUMENTS MADE PUBLIC- NUREG 1307 R03: REPORT ON WASTE BURIAL [

LY AVAILABLE.Apnl 1 30.1993. CHARGES. Escalation Of Decommissioning Waste Disposal Costs At t NUREG 0750 V36101; INDEXES TO NUCLEAR REGULATORY COM. Lo* Level Waste Burial Facilitses. i MISSION ISSUANCESJuty September 1992 OfVISION OF SAFETY ISSUE RESOLUTION (POST 880717)

NUREG-0750 V36102. INDEXES TO NUCLEAR REGULATORY COM- NUREG-0933 S15: A PRIORITIZATION OF GENERIC SAFETY MISSION ISSUANCES. July-December 1992.

N EG 64: REGULATORY ANALYSIS FOR THE RESOLUTION OF NUREG-0750 V37 NO2: NUCLE AR REGULATORY COMMIS$10N IS.

GENERIC SAFETY ISSUE 106: PIPING AND THE USE OF HIGHLY SUANCES FOR F EBRUARY 1993. Pages 55-134 NUREG-0750 V37 NO3: NUCLEAR REGULATORY COMMISSION IS- SEVE E T SS ES B AN H SUANCES FOR MARCH 1993. Pages 135-249. NUREG/CR-5966: A SIMPLIFIED MODEL OF AEROSOL REMOVAL NUREG-0936 V12 N01: NRC REGULATORY AGENDA Quarterty BY CONTAINMENT SPRAYS.

Report. January-March 1993.

INTRA-AGENCY COMMITTEES, REVIEW GROUPS ETC.

EDO - OFFICE OF THE CONTROLLER (PRE B20418 & POST 890205) IPC TASK GROUP DIVISION OF BUDGET & ANALYSIS (POST 890205) NUREG-1477 DRFT FC: VOLT AGE-BASED INTERIM PLUGGING CRI-NUREG-1100 V09 BUDGET ESTIMATES Fiscal Years 1994-1995 TERIA FOR STEAM GENERATOR TUBES Draft Report For Com-NUREG-1350 VOS: NUCLEAR REGULATORY COMMISSION INFOR- ment.

MATION DIGEST.1993 Edition.

EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428)

EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DIVISION OF REACTOR PROJECTS - t/II (POST 870411)

DATA NUREG-0847 S11: SAFETY EVALUATION REPORT RELATED TO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA, DI- THE OPERATION OF WAITS BAR NUCLEAR PLANT, UNITS 1 AND RECTOR 2. Docket Nos. 50 390 And 50-391.(Tennesee Valley Authonty)

NUREG-D090 V i5 N04: REPORT TO CONGRESS ON ABNORMAL DIVISION OF REACTOR PROJECTS - Ill.IV.V (POST 901216)

OCCURRENCES October-December 1992. NUREG-0797 S?7: SAFETY EVALUATION REPORT RELATED TO NUREG 1474: EFFECT OF HURRICANE ANDREW ON THE TURKEY THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC STA-POINT NUCLEAR GENERATING STATION FROM AUGUST 20-30. TlON, UNIT 2 Docket No. 50 446.(Teras Utahtes Electne Company.et 1992. al.)

INCIDENT RESPONSE BRANCH DIVISION OF REACTOR INSPECTION & LICENSEE PERFORMANCE NUREG/CR-5247 V02. RASCAL VERSION 2 0 WORKBOOK. (POST 921004)

TRENDS & PATTERNS ANALYSIS BRANCH NUREG-0040 V17 N01: LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR-5471: ENHANCEMENTS TO DATA COLLECTION AND SPECTION STATUS REPORT. Quarterly Report. January-March REPORTING Or SINGLE AND MULTIPLE F AILURE EVENTS. 1993 (White Book) 29

.Ai.

I l

t B

l l

I I

- - ___ _ ~ " -- - __ __

NRC Originating Organization Iridex (International Agreements)

This index lists those NRC organizations that have published international agreement re- .

ports. The index is arranged alphabetically by rnajor NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

EDO - OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405) NUREG/lA4109: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 10% i OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 860720)

LOAD REJECTION TRANSIENT FROM 75% STEADY STATE IN NURE G/lA-0090: ASSESSMENT OF RELAP5/ MOD 2 USING THE THE VANDELLOS 11 NUCLEAR POWER PLANT.

TEST DATA OF REWET.ll REFLOODING EXPERIMENT SGt/R. NUREG/lA-0116: ASSESSMENT OF RELAP5/ MOD 3/V5MS AGAINST NUREG/lA-0092- ASSESSMENT OF RELAP5/ MOD 2 COMPUTER THE UPTF TEST NUMBER 11 (COUNTERCURRENT FLOW IN CODE AGAINST THE NET LOAD TRIP TEST DATA FROM YONG_ PWR HOT LEG).

GWANG. UNIT 2. NUREG/1A-0118: ANALYSIS OF LOFT TEST L5-1 USING RELAP5/

NUREG/lA-0094: ASSESSMENT OF RELAPS/ MOD 3 AGAINST MOD 2.

TWENTY FIVE POST-DRYOUT EXPERIMENTS PERFORMED AT NUREG/lA-0119: ASSESSMENT AND APPLICATION OF BLACKOUT THE ROYAL INSTITUTE OF TECHNOLOGY. TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

NUREG/lA-0095- RELAPS ASSESSMENT USING LSTF TEST DATA NUREGilA-0120: ASSESSMENT OF THE TURBINE TRIP TRANSIENT NUREG/lA 0099. RELAPS ASSESSMENT USING SEM: SCALE N E / A-0 AS E MEN MSIV FULL CLOSURE FOR N REG 04 F[E 5/ MOD 3 ASSESSMENT USING THE SEMIS- T OF (G1J1 CALE 50% FEED LINE BREAK TEST S-FS-11. NUREG/lA-0123: APPLICATION OF FULL POWER BLACKOUT FOR NUREG/lA-0105, ASSESSMENT OF RELAP5/ MOD 3 VER$lON SMS C N ALMARAZ WITH RELAP5/ MOD 2.

USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF NUREG/lA4124. ASSESSMENT OF RELAPS/ MOD 2 AGAINST A KORl VNIT 3 PLANT. PRESSURIZER SPRAY VALVE INADVERTED FULLY OPENING NUREG/lA-0106 ASSESSMENT OF PWR STEAM GENERATOR TRANSIENT AND RECOVERY BY NATURAL CIRCULATION IN MODELLING IN RELAPS/ MOD 2- JOSE CABRERA NUCLEAR STATION.

NUREG/lA 0100: ASSESSMENT OF RELAP5/ MOD 2 AGAINST A NUREG/lA-0125: ASSESSMENT OF RELAPS/ MOD 2 COMPUTER TURBtNE TRIP FROM 100% POWER IN THE VANDELLOS 11 NU. CODE AGAINST THE NATURAL CIRCULATION TEST DATA FROM CLEAR POWER PLANT, YONG.GWANG UNIT 2. 5 i

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I NRC Contract Sponsor Index (Contractor Reports) l This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program office) and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) prepared by that organi- l zation. if further information is needed, refer to the main citation by the NUREG/CR number. i l

l EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DIVISION OF SAFETY ISSUE RESOLUTION (POST 890717)

DATA NUREG/CR-4832 VO5: ANALYSIS OF THE LASALLE UNIT 2 NUCLE-DIVISION OF OPERATIONAL ASSESSMENT (POST B70413) AR POWER PL ANT: RISK METHODS INTEGRATION AND EVALUA.

NURF G/CR 5247 V02' R ASCAL VERSION 2 0 WORKBOOK-TION PROGRAM.Paramete* Estrmation Analysis And Screening EDO- OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NLRE CR 83 09 DIVISION OF HIGH-LEVEL WASTE MANAGEMENT (POST 870413) LYSIS OF THE LASALLE UNIT 2 NUCLE-NUREG/CR-4735 V08 EVALUATION AND COMPILATION OF DOE AR POWER PLANT: RISK METHODS INTEGRATION AND EVALUA-W ASTE PACKAGE TEST DATA. B4 annual Report. August 1989 - Jan~

NL EG/CR 305 V02 1 1 TE D 1 ASSESSMENT FOR DIVNOf LOW-LEVEL WASTE MANAGEMENT & DECOMMISSION. THE LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology ING (POST 870413) And Rmk Uncertainty Evaluation Program (PRUEP) Appendices A-C.

NUREG/CR4041 DISPOSAL UNIT SOURCE TEAM (DUST) DATA NUREG/CR-5305 V02 P2: INTEGRATED RISK ASSESSMENT FOR INPUT GUIDE. THE LASALLE UNIT 2 NUCLEAR POWER PLANT;Phenomenology And Risk Uncertainty Evaluation Program (PRUEP) Appendices D-G.

EDO OFFICE OF NUCLEAR REGULATORY RESEARCH(POST 820405) NUREG/CR 5471: ENHANCEMENTS TO DATA COLLECTION AND DIVISION OF ENGINEERING (POST B70413) REPORTING OF SINGLE AND MULTIPLE FAILURE EVENTS .

NUREG/CR-4599 V02 N2. SHORT CRACKS IN PIPING AND PIPING NUREG/CR-5759 RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS '

WELDS Semtannual Report. October 1991 - March 1992 STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-NUREG/CR-4667 V15: ENVIRONMENTALLY ASS:STED CRACKING IZED WATER RE ACTOR PLANTS.

IN LIGHT WATER REACTORS Semiannual Report. April September NUREG/CR-5801: PROCEDURE FOR ANALYSIS OF COMMON- l 1992 CAUSE FAILURES IN PROBABILISTIC SAFETY ANALYSIS.

NUREG/CR-4744 V07 N1 LONG TERM EMBRITTLEMENT OF CAST NUREG/CR-5966: A SIMPLIFIED MODEL OF AEROSOL REMOVAL DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual BY CONTAINMENT SPRAYS.

Report October 1991 - March 1992. NUREG/CR-5976: DEVELOPMENT AND USE OF A TRAIN 4EVEL NUREG/CR 5410- STATISTICALLY BASED REEVALUATION OF PROBABILISTIC RISK ASSESSMENT. I P!SC-il ROUND ROBIN TEST DATA NUREG/CR 6027; PRELIM! NARY EVALUATION OF SNUBBER i NUREG/CR 5776 DAMPING IN LOW-ASPECT RATIO REINFORCED SINGLE FAILURES. '

CONCRETE SHEAR WALLS DIVISION OF SYSTEMS RESEARCH (POST 880717)

NURE G/CR-5955. MATERIALS AND DESIGN BASES ISSUES IN NUREG/CR-5882: TRAC-B THERMAL-HYDRAULIC ANALYSIS OF ASME CODE CASE N-47 THE BLACK FOX BOILING WATER REACTOR.

NUREG/CR-5972 EFFECTS OF NONS"ANDARD HEAT TREATMENT NUREG/CR 5937: INTENTIONAL DEPRESSURIZATION ACCIDENT TEMPERATURES ON TENSILE AND CHARPY IMPACT PROPER- MANAGEMENT STRATECY FOR PRESSURIZED WATER REAC-TIES OF CARBON STEFL CASTING REPAIR WELDS. TORS.

NUREG/CR 5997: CSNI PROJECT FOR FRACTURE ANALYSES OF N'JREG/CR-5982: EFFECTIVENESS OF CONTAINMENT SPRAYS IN LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS CONTAINMENT MANAGEMENT (PROJECT F ALSIRE) NUREG/CR-5983: SAFETY ASPECTS OF FORCED FLOW COOL.

NUREG/CR-5999 INTERIM FATIQUE DESIGN CURVES FOR DOWN TRANSIENTS IN MODULAR HIGH TEMPERATURE GAS-CARBON, LOW ALLOY, AND AUSTENITIC STAINLESS STEELS IN COOLED REACTORS.

LWR ENVIRONMENTS NUREG/CR-5984: CODE AND MODEL EXTENSIONS OF THE NUREG/CR-6013- METHODS USED FOR THE TREATMENT OF THATCH CODE FOR MODULAR HIGH TEMPERATURE GAS-NON-PROPORTIONAL LY DAMPED STRUCTURAL SYSTEMS COOLED REACTORS.

NUREG/CR-6031 CAV1TATION GUIDE FOR CONTROL VALVES NUREG/CR-6018: SURVEY AND ASSESSMENT OF CONVENTIONAL NUREG/CR-6036 INITIAL RESULTS OF THE INFLUENCE OF BIAX- SOFTWARE VERiflCATION AND VALIDATION METHODS.

IAL LOADING ON FRACTURE TOUGHNESS NUREG/CR-6032: SOUDUS AND LIQUIDUS TEMPERnTURES OF DIVISION OF REGULATORY APPLICATIONS (POST 870413) CORE-CONCRETE MIXTURES NUREG/CR-4214 R1P2A2. HEALTH EFFECTS MODELS FOR NU-CLEAR POWER PLANT ACCIDENT NUREG/CR 6035: FEASIBILITY STUDY FOR IMPROVED STEADY- -

CONSEQUENCE ANALYSIS Modification Of Modeis Resultog From Addition Of Ef- STATE INITIALtZATION ALGORITHMS FOR THE RELAP5 COM- [

PUTER CODE.

tects Of Exposure To Alpha Emittmg Radeonuchdes Pad 11: Scientific Bawn For Heauh- NUREG/CR 6061: DETERMINATION OF THE B AS IN LOFT FUEL PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN NUREG/CH-5817 V02. NRC HiGH-LEVEL RADIOACTIVE WASTE RE-SE ARCH AT CNWRA Calendar Year 1991 PHASE OF LARGE-BREAK LOCA EXPERIMENTS NUREG/CR-5817 V03 NY: NRC HIGH-LEVEL RADIOACTIVE WASTE RESEARCH AT CNWRA. January-June 1992 EDO - OFFICE OF NUCLEAR REACTOR HEGULATION (POST 800428)

DIVISION OF SYSTEMS SAFETY & ANALYSIS (POST 921004)

NUREG/CR 5894. RADIONUCLIDE CHARACTERIZATION OF REAC- NUREGICR-6014. HIGH PRESSURE COOLANT INJECTION SYSTEM TOR DECOMMISSIONING WASTE AND NEUTRON-ACTIVATED RISK-BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER METALS- ST ATION NUREG/CR-6028- BIGFLOW. A NUMERICAL CODE FOR SIMULAT- DIVISION OF ENGINEERING (POST 921004)

ING F LOW IN VARIABLY SATURATED. HETEROGENEOUS GEO- NUREG/CR-5957. SYSTEM Bo + (TM) CONT AINMENT - STRUCTUR-LOGIC MEDIA. Theory And User's Manual - Version 11. AL DESIGN REVIEW.

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Contractor Index 1

This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports. If further information is needed, refer to the main citation by the NUREG/CR number.

I i

ARGONNE NATIONAL LABORATORY EG&G IDAHO, INC.

NUREG/CR4667 V15 ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-5759: RISK ANALYSIS OF HIGHLY COMBUSTIBLE GAS  ;

LIGHT WATER REACTORS Semiannual Report.ApnLSeptember 1992. STORAGE, SUPPLY. AND DISTRIBUTION SYSTEMS IN PRESSUR i NUREG/CR4744 V07 N1: LONG TERM EMBRITTLEMENT OF CAST LZED WATER REACTOR PLANTS.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semiannual NUREG/CR-5882: TRAC-B THERMAL-HYDRAULIC ANALYSIS OF THE Report. October 991 - March 1992 BLACK FOX BOILING WATER REACTOR.

• NUREG/CR-5999. INTERIM F ATIOUE DESIGN CURVES FOR CARBON- NUREG/CR-5937: INTENTIONAL DEPRESSUR12ATION ACCIDENT t LOW-ALLOY, AND AUSTENITIC STAINLESS STEELS IN LWR ENVL RONMENTS MANAGEMENT STRATEGY FOR PRESSURIZEO WATER REAC-NUREG/CR-6032: SOLIDUS AND LtOUIDUS TEMPERATURES OF TORS CORE CONCRETE MIXTLAES NUREG/CR-5976: DEVELOPMENT AND USE OF A TRAIN-LEVEL PROBABILISTIC RtSK ASSESSMENT.

QTHEY CONSULTING NUREG/CR-6027: PRELIMINARY EVALUATION OF SNUBBER SINGLE l NUREG/CR 5247 V02. RASCAL VERSION 2.0 WORKBOOK.

NUR G/CR4061 DETERMINATION OF THE BIAS IN LOFT FUEL BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN NUREG/CR4599 V02 N2 SHORT CRACKS IN PIPING AND PIPING PHASE OF LARGE-BREAK LOCA EXPER:MENTS.

WELDS Semiannual Report, October 1991 - March 1992.

ELECTRIC POWER RESEARCH INSTITUTE i BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NUREG/CR-6018: SURVEY AND ASSESSMENT OF CONVENTIONAL  ;

LABORATORY SOFTWARE VERIFICATION AND VAllDATION METHODS.

NUREG/CR4214 R1P2A2: HEALTF EFFECTS MODELS FOR NUCLE-AR POWER PLANT ACCIDENT CONSEQUENCE EOE ENGINEERING CONSULTANTS (FORMERLY EDE ENGINEERING,  !

ANALYSIS Modifcation Of Moels Resulting From Addition Of Effects INC.)

Of Exposure To Alpha Emrng Radeonuclties.Part II: Scient fc Bases NUREG/CR-6013 METHODS USED FOR THE TREATMENT OF NON-For Health- PROPORTIONALLY DAMPED STRUCTURAL SYSTEMS.

NUREG/CR-5410 STATIS' eCALLY BASED REEVALUATION OF PISC-li ROUND ROBIN TEST T #A FRANCE NUREG/CR-5894 RAD' >NUCLlDE CHARACTER 12ATION OF REAC-NUREG/CR4028. BIGFLOW: A NUMERICAL CODE FOR SIMULATING i TOR DECOMMISS( NING WASTE AND NEUTRON. ACTIVATED METALS FLOW IN VARIABLY SATURATED. HETEROGENEOUS GEOLOGIC i MEDIA Theory And User's Manual- Versson 11. f BROOKHAVEN NATIOS AL LABORATORY GESELLSCHAFT FUR REAKTORSICHERHEIT i' REAC R FET INFO AT N E Tl NUREG/CR-5997: CSNI PROJECT FOR FRACTURE ANALYSES OF NUREG/CP-0126 V02 PROCEEDINGS OF THE TWENTIETH WATER LARGE-SCALE INTERNATIONAL REFERENCE EXPERIMENTS l REACTOR SAFETY INFORMATION MEETING. (PROJECT FALSIRE). l NUREG/CP 0126 V03. PROCEEDINGS OF THE TWENTIETH WATER '

RE ACTOR SAFETY INFORMATION MEETING. IDAHO NATIONAL ENGINEERING LABORATORY i NUREG/CR4214 R1P2A2: HEALTH EFFECTS MODELS FOR NUCLE- NUREG/CR-6061: DETERMINATION OF THE BIAS IN LOFT FLFL <

AR POWER PLANT ACCIDENT CONSEOUENCE PEAK CLADDING TEMPERATURE DATA FROM THE BLOWDOWN  !

ANALYSIS Modifcation Of Models Resulting From Additon Of Effects PHASE OF LARGE-BREAK LOCA EXPERIMENTS.

Of Exposure To Alpha Emitting Radeonuchdes Part II. Scentific Bases For Heattn .. INHALATION T0XICOLOGY RESEARCH INSTITUTE NUREG/CR-5982. EFF'!CTIVENESS OF CONTAINMENT SPRAYS IN NUREG/CR-4214 R1P2A2: HEALTH EFFECTS MODELS FOR NUCLE- .

CONTAINMENT MANAGEMENT. AR POWER PLANT ACCIDENT CONSEQUENCE f NUREG/CR-5983: SAFETY ASPECTS OF FORCED FLOW COOLDOWN ANALYSIS Modifcation Of Models Resultin0 From Additon Of Effects TRANSIENTS IN MODULAR HIGH TEMPERATURE GASCOOLED Of Exposure To Alpha-Emittsng Radonuctedes Part it. Scentifc Bases RE ACT ORS.

For Health ~ '

NUREG/CR-5984 CODE AND MODEL EXTENSIONS OF THE THATCH CODE FOR MODULAR HIGH TEMPERATURE GAS-COOLED REAC- INSTITUTE OF NUCLEAR POWER OPERATIONS TORS. l NUREG/CR.6014. HIGH PRESSURE COOLANT INJECTION SYSTEM NUREG 1474: EFFECT OF HURRICANE ANDREW ON THE TURKEY RISK-BASED INSPECTION GUIDE FOR HATCH NUCLEAR POWER POINT NUCLEAR GENERATING STATION FROM AUGUST 20-30'

• 1992' '

STATION NUREGICR4041: DISPOSAL UNIT SOURCE TEAM (DUST) DATA BOWA STATE UNIV., AMES,lA INPUT GUIDE.

NUREG/CR-5957: SYSTEM 804 (TM) CONTAINMENT - STRUCTURAL CENTER FOR NUCLEAR WASTE REGULATORY ANALYSES DESIGN REVIEW.

NUREG/CR 5817 V02: NRC HIGH-LEVEL RADIOACTIVE WASTE RE-SEARCH AT CNWRA. Calendar Year 1991. JBF ASSOCIATES,INC.

NUREG/CR-5817 V03 N1: NRC HIGH LEVEL RADIOACTIVE WASTE NUREG/CR4471: ENHANCEMENTS TO DATA COLLECTION AND RE. +

RESEARCH AT CNWRA. Januarydune 1992. PORTING OF SINGLE AND MULTIPLE FAILURE EVENTS. i NUREG/CR4026: BIGFLOW. A NUMERICAL CODE FOR SIMULATING  !

FLOW IN VARIABLY SATURATED HETEROGENEOUS GEOLOGIC LOS ALAMOS NATIONAL LABORATORY I

MEDIA. Theory And User s Manuat - Verson 1.1. NUREG/CR-5776: DAMPING IN LOW-ASPECT-RATIO. REINFORCED CONCRETE SHEAR WALLS.

I COMPUTER SIMULATION & ANALYSIS,INC. '

I NUREG/CR-6035: FEASIBILITY STUDY FOR IMPROVED STEADY. M ARYLAND, UNIV. Or, COLLEGE PARK, MD STATE INITIAtl2ATION ALGORITHMS FOR THE RELAP5 COMPUT- NUREG/CR 5801: PROCEDURE FOR ANALYSIS OF COMMON CAUSE ER CODE. FAILURES IN PROBABILISTIC SAFETY ANALYSIS. i 35 r

36 Contractor index NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERLY NUREG/CR 5335 V02 P2: INTEGRATED RISK ASSESSMENT FOR THE NATIONAL BUREAU OF LASALLE UNIT 2 NUCLE AR POWER PLANT.Phenomenokgy And NUREG/CR 4735 V96 EV ALU ATION AND COMPILATION OF DOE Risk Uncedainty Evaluation Pmgram (PRUEP) Appendices D-G W ASTE PACKAGE TEST DAT A. Ebannual Repo*t. August 1989 Janu. NUREG/CR-54'71 E NHANCEMENTS TO DAT A COLLECTION AND RE-a y 1990. PORTING OF S:NGLE AND MULTIPLE FAILURE EVENTS NUREG/CR-5801: PROCEDURE FOR ANALYSIS OF COMMON-CAUSE NE W ME XICO, UNIV. OF, ALBUQUERQUE, NM FAILURES IN PROBABILISTIC SAFETY ANALYSIS NUREG/CR 5776 DAMPING IN LOW ASPECT RATIO REINFORCED NUREG/CR-5966 A S'MPLIFiED MODEL OF AEROSOL REMOVAL BY CONCRETE SHEAR W ALLS CONTAINMENT SPRAYS NUS CORP. SCIENCE APPLICATIONS INTERNATIONAL CORP-(FORMERLY NURE G/CH 5471 ENHANCEMENTS TO DATA COLLECTION AND RE. SCIENCE APPLICATIONS, PORTING OF SINGLE AND MULT'PLE FA! LURE F VENTS NUREG/CR 5759. RISK ANALYSIS OF HIGHLY COMBUSTlBLE GAS STORAGE, SUPPLY, AND DISTRIBUTION SYSTEMS IN PRESSUR-OAK RIDGE NATIONAL LABORATORY IZED WATER REACTOR PLANTS NUREG/CR 5955 MATE RIALS AND DESIGN BASES ISSUES IN ASME NUREG/CR-6010 SURVEY AND ASSESSMENT OF CONVENTIONAL CODE CASE N-47 SOFTWARE VERiflCATION AND VALIDAflON METHODS NUREG/CR.5972 EFrECTS OF NONST ANDARD HEAT TRE ATMENT TEMPERATURES ON TENS!LE AND CHARPY IMPACT PROPERTIES SCIE NTECH, INC.

OF CARDON STEEL CAST lNG REPAIR WELDS NUREG/CR-5759. RISK ANALYSIS OF HIGHLY COMBUSTlBLE GAS NUREGICR 5997. CSNI PROJECT FOR F RACTURE ANALYSES OF STORAGE. SUPPLY, AND D!STR:BUTION SYSTEMS IN PRESSUR-LARGE SCALE INT E RN AllON AL REFE RENCE E)PER ME NTS 17ED W ATER RE ACTOR PLANTS iPROJECT F ALSIREl Nt REG /CRfC36- INITIAL RESULTS OF THE INFLUENCE OF B:AKiAL TEX AS A&M UNIV., COLLEGE STATION, TX LOADING ON FRACTURE TOUGHNESS NUREG/GR-0006. DEPOSITION SOFTWARE TO CALCULATE PARTI-CLE PENETRATION THROUGH AE ROSOL TRANSPORT SANDIA NATIONAL LABORATORfES SYSTEMS Finat Report NUREG/CR-4832 VOS ANALYSIS OF THE LASALLE UN:T 2 NUCLEAR POWER PLANT RISK METHODS INTEGRATION AND EVALUATION TULLIS ENGINEERING CONSULTANTS PROGRAM Parameter E stimation Analysis And Screening Human Reit- NUREG/CR-E031 CAVITATION GUtDE FOR CONTROL V ALVES.

afnhty Anaiysis wtSCONSIN, UNIV. OF, MADISON WI NUREo/CR-4832 V03 ANALYST $ OF THE LASALLE UN;T 2 NUCLEAR PCWER PLANT HISK METHODS INTEGRATION AND E VALUATION NUREG/CR-4214 R1P2A2 HEALTH EFFE CTS MODELS F OR NUCLE-PROGR AM (RM EP) Internal Fue Analysis AR POWER PLANT ACCIDENT CONSEQUENCE NUREG/CR-5305 V02 P1. INTEGRATED R!SK ASSES 5 MENT FOR THE ANALYSIS Mod;fecation Of Models Resulting From Ati: tion Of Effects LASAtl.E UNIT 2 NUCLEAR POWER PLANT Phenomencegy And Of Exposure To Alpha Emitting Radionuchdes Pa t it Sc ent< tic Bases Amk Uncenamty Evaluation Program (PRUEP) Appen$ces A-C. F or Health ..

9 ..

• • .4

International Organization index This index lists, in alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed below each country and per-forming organization are the NUREG/lA numbers and titles of their reports, If further infor-mation is needed, refer to the main citation by the NUREG/lA number.

FEDERAL REPUBLIC OF GERMANY NUREG/lA-0109; ASSESSMENT OF RELAPS/ MOD 2 AGAINST A 1%

StEMENS AG-KWU GROUP LOAD REJECTION TRANSIENT FROM 75% STEADY STATE N NUREG/lA 0116. ASSESSMENT OF RELAP5/ MOD 31V5M5 AGAINST THE VANDELLOS tl NUCLEAR POWER PLANT.

i THE UPTF TEST NO.11 (COUNTERCURRENT FLOW IN PWR HOT ASOCIACION NUCLEAR ASCO l LEF) NUREG/lA-0119. ASSESSMENT AND APPLICATION OF BLACXOtfi TRANSIENTS AT ASCO NUCLEAR POWER PLANT WITH RELAPS/

l FINLAND MOD 2.

[ TECHNICAL RESEARCH CENTRE OF FINLAND . C N ALMARAZ l Y ll NUREG/lA-0090: ASSESSMENT OF RELAPS/ MOD 2 USING THE NUREG/iA-0123: APPLICATION OF FULL POWER BLACKOUT FOR TEST DATA OF REWET il REFLOODING EXPERIMENT SGI/R. C.N ALMARAZ WITH RELAP5/ MOD 2.

UN! DAD ELECTRICA.S A.

REPUBLIC OF KOREA NUREG/lA 0120: ASSESSMENT OF THE TURBINE TRIP TRANSIENT KOREA ELECTRIC POWER CORPORATION IN COFRENTES NPP WITH TRAC-BF1.

NUREG/lA-0092- ASSESSMENT OF RELAP5/ MOD 2 COMPUTER UNION ELECTRICA FENOSA CODE AGAINST THE NET LOAD TRIP TEST DATA FROM YONG. NUREG/lA 0124: ASSESSMENT OF RELAPS/ MOD 2 AGAINST A GWANG. UNIT 2. PRESSURIZER SPRAY VALVE INADVERTED FULLY OPENING NUREG/lA-0125- ASSESSMENT OF RELAP5/ MOD 2 COMPUTER TRANSIENT AND RECOVERY BY NATURAL CIRCULATION IN CODE AGAINST THE NATURAL CIRCULATION TEST DATA FROM JOSE CABRERA NUCLEAR STATION. ,

YONG-GWANG UNIT 2 UNIVERSITY OF CANTABRIA  ;

NUREG/lA 0122. ASSESSMENT OF MSIV FULL CLOSURE FOR KOREA INSTITUTE OF NUCLEAR SAFETY NUREG/lA-0095: RELAPS ASSESSMENT USING LSTF TEST DATA SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING ,

sB.CL-18 TRAC-BF1 (G1J1)

• NUREG/lA-0099- RELAPS ASSESSMENT USING SEMISCALE NURLG A- LA / MOD 3 ASSESSMF9 USING THE SEMIS-

^ "^

NUR /A SS E F ELAPS / MOD 3 AGAINST ,

NURE / 10 A ES M RELA / MOD 3 VERSION SMS EROA ST T O CH O Y' USING INADVERTENT SAFETY INJECTION INCIDENT DATA OF KORI UNIT 3 PLANT- UNITED KINGDOM NATIONAL POWER SPAIN NUREG/lA 0106- ASSESSMENT OF PWR STEAM GENERATOR ASOCiACION NUCLEAR VANDELLOS MODELLING IN RELAP5/ MOD 2 NUREG/lA-0108 ASSESSMENT OF RELAPS/ MOD 2 AGAINST A NUCLE AR ELECTRIC TURBINE TRIP FROM 100% POWER IN THE VANDELLOS 11 NU- NUREG/lA4118: ANALYSIS OF LOFT TEST LS-1 USING RELAPS/

CLE AR POWE.R PLANT. MOD 2 T

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Licensed Facility Index This index lists the facilities that were the subject of NRC staff or contractor reports. The  ;

facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

l STN 50 556 Black For Statm Unit 1, Pvt,6c Serwce of NUREGICR 58E2 W374 LaSalie Coety Staton, Umt 2. Commonwea!m NUREG/CR 5305 V32 P2 Oklahoma Esson Co STN 50 557 B:ack For Statm Unn 2. Putec Sevre of NURE G'CR-5662 52-002 System 80+ Standar$ zed Nockar Power Plant NUREG/CR 5957 hhoma i 50-446 Des. Combuston Erynee Comanche Pean Steam Electnc State Und 2. NJAEG-0797 S27 50 289 Ttvee We haard Nuclear Stamn. Umt 1 NUREG 1485 Texas Utdees EM 50 321 gen,,a: PutAc Utdmes Edom i Hatch Nucear Plant. Um! 1. Geortpa NUREG/CR 6014 50 390 Warts Bae Nuciear Plant. Und 1. Tennessee NUREG-0847 511 W366 E$wn Hatch Naciear Plant. Umt 2. Georpa NUREG 'CR&t4 50 391 Wat Bar xte Plant. Und 2, Tennessee NUREG4847 S11

% 374 {

LaSa!ie Corty Staten, Unit 2 Commonwealth NUREG!CR 53D5 V02 P1 Vailey Authonty E$ son Ce s

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t#4C FORM 335 1, RCPORT f JUMBER p-09) U. S NUCLE AR ALGULATORY COMMISSION r#4CM 1102, (Assigned by FJtC. Add Vol...

Supp. , Anv. , and Addendum Num-3* 32G2  ;

BIBLIOGRAPHIC DATA SHEET **"5- d

• avl  ;

(s+.e enwucienm on ne re msei NU111!G-0304

e nia AND suuiint Vol.18, No. 2 5

3. DA T E RE PORT PUUUSHE D llegulatory and Technical iteports ( Abstract Index Journal)

MONT H I YEAR  !

I Compilation for August t

1993 Second Quarter 1993 Apni-June 4. rlN OR GRANT NUMBER .

b.Avi'*"!N

b. TYPE Of HLPORT Reference <

7. PERIOD COVERED (inclusive Dates)

April-June 1993 8 Pt HFOHMING OHGAN;2AllON

- NAML AND ADDHE SS (if NRC, provios Dmsion. Off 4ce or Hegson, U, S. Nuclear Regulatory Commiss60n. and rnathng addr0sS; it coritractur, prevede name and malbng ad@ess )

Division of Freedom of Information and Publications Services Office of Administration U.S. Nuclear llegulatory Commission Washington, DC 20555  :

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SPONSORING ORGANIZATION - NAME AND ADDRESS (if tJRC, type *Same as above , it contractor, provede NRC Division. Office or Region.

O.S F Ascicar kegulatory Comrmssion, and enaihng address. )

Same as 8. above.  !

10. SUI'PLLME NT AhY NOTt s

11. AesinAcT gca worcs or iess)

'Ihis journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceed-ings of conferences and workshops; as well as international agreement reports. The entries in this compilation are  !

indexed for access by title and abstract, secondary report number, personal author, subject, NRC organization for ,

staff and international agreements, contractor, international organization, and licensed facility.  ;

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12. VIY WORDS/DESCRIPTORS (Lsst words or phrases that will assist researchins in locating the report.) 13. AVAR.AD:LITY STATEMENT Unlimited  !

com ilation 14. SECURITY CLASSIFICATION (This Page) i abstract mdex +

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Unclassified (This Report)

Unclassified

15. NUMBLR OF PAGES L

16. fHICE -I NRC FORM 335 (2-89)

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