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Regulatory and Technical Reports.Compilation for First Quarter 1985,January-March
	ML20128A383
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Issue date:	04/30/1985
From:	; NRC OFFICE OF ADMINISTRATION (ADM)
To:	;
References
	NUREG-0304, NUREG-0304-V10-N01, NUREG-304, NUREG-304-V10-N1, NUDOCS 8505240207
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NUREG-0304 Vol.10, No.1 Regulatory and Technical Reports (Abstract Index Journal) l l I Compilation for First Quarter 1985 J:nu ry - March D;ts Published: April 1985 Pclicy and Publications Management Branch Divi:i:n of Technical Information and Document Control Offica of Administration U.S. Nuclear Regulatory Commission l W=hington, D.C. 20555 f* *%,, 1

f,. ' 4 l. CONTENTS Preface...............................................................................v Index Tab Main Citation and Abstracts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

             . S taff Reports . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Conference Proceedings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Contractor Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Contractor Report Number index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Personal Author Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 S u bject index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 NRC Originating Organization Index (Staff Reports) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 NRC Contract Sponsor index (Contractor Reports) . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............ 6 Contractor i ndex . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Licensed Facility Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 n l

PREFACE 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: Dnneson of Technical Information and Document Control - Policy and Publications Management Branch Publishing and Translations Section Woodmont 501

               . U.S. Nuclear Regulatory Commission
              . Washington, D.C. 20665 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, and NUREG/CR-XXXX. These precede the following indexes:

Contractor Report Number index Personal Author index Subject Index NRC Originating Organi'ation Index (Staff Reports)

               . NRC Contract Sponsor Index (Contractor Reports)

Contractor Index Ucensed Facility Index A detailed explanation of the entries precedes each index. The bibliographic elements of the main citations are the following: Staff Report - NUREG-0608: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048 09570:200. Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for intemal NRC use). Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp. 8105280299 ANL-81-3. 08632:070.

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

Contractor Report NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.: BENNETT, P.R. Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242. Where the entries are (1) report number, (2) report title, (3) report authors. (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the microfiche address (for NRC intemal use). v ,

o t The following abbreviations are used to identify the document status of a report: ADD - ' addendum APP .- appendix DRFT -- draft ERR - errata N - number R - revmon S. - supplement V - volume

    - Availability of NRC Publications Copies of NRC staff and contractor reports n'sy be purchased either from the NRC-G'PO Sales Office or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the NRC-GPO Sales Office send a check or money order, payable to the Superintendent of
    - Documents, to the following address:

U.S. Nuclear Regulatory Commission ATTN: Sales Manager Washington, D.C. 20S55 You may charge any purchase to your GPO Deposit Account, Master Charge card, or VISA charge card by calling the NRC-GPO Sales Office on (301) 492-9530. Non-U.S. customers must make payment in advance either by International Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents.

    - NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor established codes such as ORNL/NUREG/TM-XXX and TREE--

NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of

   . the work being reported.

In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings. All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Technical Information and Document Control. vi

Main Citations and Abstracts The report listings in this compliation are arranged by report number, where NUREG-XXXX is an NRC staff originated report, NUREG/CP-XXXX is an NRC sponsored conference report, and NUREG/CR-XXXX is an NRC contractor-prepared report. The bibliographic information (see Preface for details)is followed by a brief abstract of the report. NUREG-OO2O VO8 N12: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of November 30,1984.(Gray Book I)

Division of Budget & Analysis. February 1985. 427pp. 8502210264. 29047:289.

The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and cccurately as possible. This information is collected by the Office of Resource Management f rom the Headquarters staf f of NRC 's Of fice of Inspection and Enforcement, from NRC's Regional Of fices, and from utilities. The three sections of the report are: monthly highlights cnd statistics for commercial operating units, and errata from previously reported datas a compilation of detailed information on cach unit, provided by NRC's Regional Offices, IE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the U.S. It is hoped the report is helpful to all egencies and individuals interested in maintaining an awareness of the U. S. energy situation as a whole. NUREG-OO2O VO9 NO1: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of December 31,1984.(Oray Book I).

Division of Budget & Analysis. February 1985. 396pp. 8503220010. 29487:265.

See NUREG-OO20,VO8,N12 abstract. NUREG-OO2O VO9 NO2: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of January 31,1985.(Gray Book I)

Division of Dudget

 & Analysis.      March 1985.       421pp. 8504090008. 29740:311.

See NUREG-OO20,VOB,N12 abstract. NUREG-OO40 VOB NO4: LICENSEM CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report,0ctober-December 1984. (White Book)

Division of GA, Safeguards & Insp Programs (830103-850212). January 1985. 230pp. 8502210401. 29056:014.

This periodical covers the results of inspections performed by the NRC's Vendor Program Branch that have been distributed to the inspected organizations tiring the period from October 1984 through December 1984. Also included in this issue are the results of certain 1

inopoctions parforced prior to Octobor 1984 that uoro not includod in 1 previous incuan of NUREG-OO40. NUREG-0304 VO9 NO4: REQULATORY AND TECHNICAL REPORTS. Annual Compilation For 1984.

Division of Technical Information & Document Control. [

   '~~'     January 1985. 501pp. 8502210096,        29049:001.                                                            ;

This Journal includes all formal reports in the NUREG series ; j prepared by the NRC staf,f and contractors, as well as proceedings of conferences and workshops. The entries in the compilation are indexed for access by title and abstract, contractor report number, personal i author, subject, NRC organization, contractor, and licensed facility. NUREG-0325 RO7: U. S. NUCLEAR REQULATORY COMMISSION FUNCTIONAL ORGANIZATION CHARTS.

Office of Resource Management, Director.

January 1985. 56pp. 8501180528. 28471:117. Funtional organization charts for the NRC Commission Offices, Divisions, and Branches are presented. i

NUREG-0540 VO6 N11: TITLE LIST OF DOCUENTS MADE PUBLICLY AVAILABLE.

November 1-30,1984.

Division of Technical Information & Document Control. January 1985. 569pp. 8502060305. 28754:136.

This document is a monthly publication containing descriptions of information received anJ generated by the U. S. NRC. This information includes (1) docketed material associated with. civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal
Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.

NUREG-0540 VO6 N12: TITLE LIST. 0F DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1984.

Division of Technical Information &

] Document Control. February-1985. 614PP. 8503200131. 29469:001. See NUREG-0540,VO6,N11 abstract. NUREG-0540 VO7 NO1: TITLE LIST OF DOCUENTS MADE PUBLICLY AVAILABLE. January 1-31,1985.

Division of Technical Information &

Document Control. ' March 1985. 665pp. 8504090020. 29742:015. See NUREC-0540,VO&,N11 abstract. i'

NUREG-0544 RO2: A HANDBOOK OF ACRONYMS AND INITIALISMS.

Division of Technical Information & Document Control. January 1985. 131pp. ,

8502150692. 28960:175. This Handbook records in alphabetical order abbreviations (acronyms, in i ti e.li sms, and other condensed forms) that have been used in the nuclear industry, both foreign and domestic. The present I volume is an attempt by the editorial staff of the Division of Technical Information and. Document Control to compile in one updated publication the abbreviations used in NRC staff and consultant reports. This issue, although not all inclusive, is to be treated as a compilation of available information. i 2 1 i-

NURE3-0606 VO7 NO1: UNRESOLVED SAFETY ISSUE 3

SUMMARY

.D3to An Of February 15, 1985.(Agua Book)

Division of Safety Technology.

March 1985.' 55pp. 8504040429. 29627: 321.

Provides an overview of the status of the progress and plans for resolution of the generic tasks addressing " Unresolved Safety Issues" as reported to Congress. NUREG-0675 S29: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF , DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company)

Division of Licensing. March 1985. 64pp. 8503280011. 29548:075.

Supplement No.-29 to the Safety Evaluation Report for Pacific Gas ! cnd Electric Company's application for licenses to operate Diablo ' Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared by the Office of Nuclear Reactor Regulation of-the U.S. Nuclear Regulatory Commission. This supplement presents the staff evaluation of the licensee's Internal Review Program for Diablo Unit 2 applicability and resolution of concerns that had been raised during the Diablo Unit i design verification by the Independent Design Verification Program, the licensee's Internal Technical Program ~ cnd the NRC staff. ' NUREG-0713 VOS: DCCUPATIONAL RADIATION EXPOSURE AT COMMERCIAL NUCLEAR POWER REACTORS - 1983 ANNUAL REPORT. BROOKS,B.G. Division of Radiation Programs & Earth Sciences (post 840429). March 1985. , 128pp. 8504050287. 29674:001. i- This report summarizes the occupational radiation exposure inf ormation that has b een reported to th e U. S. N. R. C. by commercial nuclear power reactors during the years 1969 through 1983. The bulk , of the data presented in the report was obtained from annual radiation oxposure-reports submitted in accordance with the requirements of 10 CFR 20.407 and license technical specifications. Data on workers terminating their employment at nuclear power facilities was obtained from. reports submitted pursuant to 10 CFR 20.408. The annual reports cubmitted by the 76 nuclear power plants that had completed at least one full year of operation as of December 31, 1983, indicated that the number of-personnel monitored during 1983 was 1367700 persons and the cnnual collective dose incurred by these individuals was 56,500 , can-rems (man-cSv). The average annual dose for each worker that received a measurable dose was 0.66 rems (cSv), and the average i collective dose per reactor was 753 man-rems (man-cSv). The i termination reports revealed that some 56,500 individuals completed their employment with one or-more reactor facilities during 1982.x Approximately 4,500 of these workers could be considered transients cnd they received an average dose of 1.11 rems (cSv). *The most recent year for which most of the termination data are available for enalysis. NUREG-0748 VO4 N12: OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data As'Of December 31,1984.(Drange Book)

Division of Budget &

Analysis. January 1985. 320pp. 8502210263. 29056:355. The Operating Reactors Licensing Actions Summary is designed to provide the Management of the Nuclear Regulatory Commission (NRC) with on overview of licensing actions dealing with the operating power and nonpower reactors. 3

NUREG-0743 VO5 NO1: OPERATING REACTbRS LICENSING ACTIONS

SUMMARY

. Data As Of January 31,1985.(Orange Book)

Division of Budget & Analysis.

March 1985. 3OOp p. 8504030070. 29598:326. See NUREG-0748,VO4 N12 abstract. I NUREG-0750 V2O 101: INDEXES TO NUCLEAR REQULATORY COMMISSION ISSUANCES I FOR JULY-SEPTEMBER 1984.

Division of Technical Information &

Document Control. January 1985. 78pp. 8503270298. 29540:261. Digests.and indexes for issuances of the Commission, the Atomic i Safety and Licensing Appeal Panel, the Atomic Safety and Licensing [ Board Panel, the Administrative Law Judge, the Director's Decisions, and-the Denials of Petitions for Rulemaking ere presented. NUREG-0750 V2O 102: INDEXES TO NUCLAER REQULATDRY COMMISSIDN ISSUANCES FOR JULY-DECEMBER 1984;

Division of Technical Information &

Document Control. March 1985, 105pp. 8504040416. 29618:193. See NUREG-0750,V20,101 abstvact. (~ NUREG-0750 V2O NO4: NUCLEAR REQULATORY COMMISSION ISSUANCES FOR OCTOBER 1984. Pages 1,055-1,435.

Division of Technical Information &

, Document Control. January 1985. 388p p. 8502060353. 28748:001. Legal issuances of the Commission, the Atomic Gafety and Licensing i Appeal Panel, the Atomic Safety and' Licensing Board Panel, the Administrative Law Judge,'and NRC Program Offices. I ! NUREG-0750 V2O N05: NUCLEAR REQULATORY COMMISSION ISSUANCES FOR NOVEMBER 1984. Pages 1,437-1,572.

Division of Technical.

Information & Document Control. January 1985. 143pp. 8502190320. 29012:226. See NUREG-0750,V20,NO4 abstract.

    ~    ,

NUREG-{)750 V2O N06: NUCLEAR REQULATORY COMMISSION ISSUANCES FOR DECEMBER 1984. Pa g e s 1, 573-1, 706. .

Division of Technical Information & Document Control. February 1985. 134pp. 8503110591.

29307:239. See NUREG-0750,V20,NO4 abstract. i NUREG-0750 V21 NO1: NUCLEAR REQULATORY COMMISSION ISSUANCES FOR JANUARY 1985. Pages.1-273.

Division of Technical Information & Document Control. March 1985. 281pp. 8504030299. 29624:081.

See NUREG-0750,V20,NO4 abstract. i NUREG-0787 S10: SAFETY EVALUATION REPORT RELATED TO THE OPERATIDN OF WATERFORD STEAM ELECTRIC STATION, UNIT 3. Docket No. 50-382. (Louisiana

                                                        ~
       . Power And Light Company)

Division of Licensing. March 1985.

35pp. 8503270297. 29541:279. Supplement 10 to the Safety Evaluation Report for the application filed by Louisiana Power & Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located ~ in St. Charles Parish, Louisiana, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of information submitted by 4

tho licensoo oinco tho Safoty Evoluotion Raport and its nino cupplements were issued. I NUREG-0797 907: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-446.(Texas Utilities Generating Company, et al)

Division of Licensing. January 1985. 100pp. 8502150150. 28958:071.

Supplement No. 7 to the Safety Evaluation Report for the Texas '- Utilities Generating Company application for a license to operate the Comanche Peak Steam Electric Station located in Somervell County, Texas has been jointig prepared by the Office of Nuclear Reactor Regulation and the Technical Review Team of the U.S. Nuclear Regulatory Commission. This Supplement provides the results of the otaff's evaluation and resolution of approximately 90 technical ; concerns and allegations in the areas of Electrical / Instrumentation 7 cnd Test Program regarding construction practices at the Comanche Peak facility. NUREG-0797 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK BTEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50-446.(Texas Utilities Generating Company, et al)

Division of Licensing. February 1985. 196pp. 8503110247. 29327:151.

Supplement No. 8 to the Safety Evaluation Report for the Texas Utilities Generating Company application for a license to operate the . Comanche Peak Steam Electric Station located in Somervell County, Texas has been Jointly prepared by the Office of Nuclear Reactor Regulation and the Technical Review Team of the U. S. Nuclear Regulatory Commission. This Supplement provides the results of the 4 otaff's' evaluation and resolution of approximately 80 technical concerns and allegations relating to civil / structural and Discellaneous issues regarding construction and plant readiness testing practices at the Comanche Peak facility. NUREG-0797 909: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAM STEAM ELECTRIC STATION, UNITS 1 AND 2. Doc ket Nos. 50-445 And 50-446.(Texas Utilities Generating Company)

Division of Licensing. . March 1985. 170pp.' 8504090015. 29730:116.

3 Supplement 9 to the Safety Evaluation Report for the Texas l Utilities Electric Company's application for a license to operate , Comanche Peak Steam Electric Station, Units 1 and 2 located in

Comervell County, Texas, has been prepared Jointig by the Office of

! ' Reactor Regulation and the Cemanche Peak Technical Review Team of the U. S. Nuclear Regulatory Commission. This supplement addresses Texas , Utilities analyses in support of its request to amend the Comanche Peak Final Safety Analysis Report to eliminate the commitment that coatings inside the reactor Containment Building be qualified for Units.1 and 2. In addition, this supplement provides the results of the staff's evaluation and resolution of 62 technical concerns and allegations in the coating area for Unit 1. Because of the favorable ' resolution of the items discussed in this report, the staff concludes l for the issues considered herein, that there is reasonable assurance l that the facility can be operated without endangering the health and i cafety of the public. NUREG-0798 905: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF ENRICO FERMI ATOMIC POWER PLANT, UNIT NO. 2. Docket No. 50-341. 5

i (Datroit Edican Cocpeng) o Division of Liconoing. Mnrch 1985. 2OOpp. . 8504040427. 29628:243. Supplement No. 5 to the Safety Evaluation Report (SER) related to l the operation of the Fermi-2 facility, provides the NRC staff's evaluation of additional information submitted by the applicant regarding the outstanding review issues identified in Supplement No. 4 i to the SER dated September 1984. This supplement contains the staff's conclusion that there are no outstanding issues which must be resolved prior.to issuance of a low power operating license (i.e.,-less than

;      five percent of full rated power) for the Fermi-2 facility.

Supplement No. 5 to the SER also summarizes the conditions which are placed in the Fermi-2 operating license. The Fermi-2 facility is located on Lake Erie in Monroe County,-almost 8 miles east-northeast of Monroe, Michigan. NUREG-0800 18.2 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To SRP Section 18.2, " Safety Parameter Display System (SPDS)."

Office of Nuclear Reactor Regulation, Director. January 1985. 11pp.

l 8502150068. 28962:310. This revision incorporates the guideline of' Task Action Plan Item 1.D.2 of NUREG-0660 as clarified in Supplement 1.to NUREG-0737,

      " Safety Parameter Display System."

i NUREG-0800 18.2A1 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To Appendix A To SRP Section 18.2, " Human Factors Review Guidelines For The Safety Parameter Display System (SPDS)."

Office of Nuclear Reactor Regulation, Director. January 1985. 46pp. 8502140040.

28943:317. i This revision incorporates.the guideline of Task Action Plan Item 1.D.2 of NUREG-0660 as clarified in Supplement 1 of NUREG-0737. Appendix A to SRP Section 18.2 was formerly draft NUREG-0835, " Human . Factors Acceptance Criteria for the Safety Parameter Display System," Draft Report issued for Comment. NUREG-0800 RO5: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 5 To SRP Table Of Contents. e _ Office of Nuclear Reactor Regulation, Director. ! January 1985. 24pp. 8502140198. 28955:140. Revision 5 to SRP Table of Contents. NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1984. JANO,J.; KRAMARIC,M.; COHEN,L. Region 1, Office of Director. January 1985. 146pp. 8502040043. 28727:001. This report.provides the status and results of the NRC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network. It presents the radiation levels measured in the vicinity of NRC licensed facility sites throughout the country for the third quarter of 1984. NUREG-0847 SO3: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Tennessee Valley Authority)

Division of Licensing.

6 o

Jcnucry 1985. 77pp. C502060553, 28745:199. This report supploacnto the Safoty Evaluation Ropert, NURE3-0847 (June 1982), Supplement No. 1 (September 1982), and Supplement No. 2 (January 1984) issued by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant. Units 1 and 2 (Docket i Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This cupplement provides recent information regarding resolution of some of the open confirmatory _ items and license conditions identified in the Safety Evaluation Report. NUREG-OS47 804: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391. (Tennessee Valley Authority)

Division of Licensing. March 1985.

45pp. 8504050283. 29673:180. This report supplements the Safety Evaluation Report, NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), and Supplement No. 3 (January 1985) issued by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the open and confirmatory l j items and license conditions identified in the Safety Evaluation Report. NUREG-0853 SO4: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF CLINTON POWER STATION UNIT 1. Docket No. 50-461.(I111nois Power

Company,et al)

Division of Licensing. February 1985. 70pp.

I G503210295. 29480:158. Supplement No. 4 to the Safety Evaluation Report on the opplication filed by Illinois Power Company, Sogland power Cooperative, Inc., and Western Illinois Power Cooperative., Inc., as l cpplicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The ' facility is located in Harp Township, DeWitt County, Illinois. This l Supplement reports the status of items that have been resolved by the j staff since Supplement No. 3 was issued. NUREG-0876 SO6: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF BYRON STATION, UNITS 1 AND 2. Docket Nos. 50-454 And 50-455. i (Commonwealth Edison Company)

Division of Licensing. February 1985. 52pp. 8503050075. 29243:290.

Supplement No. 6 to the Safety Evaluation related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County,. Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement provides recent information regarding resolution of the license conditions identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that the Byron Station, Unit 1 can be operated by the licensee at power levels 7

grooter then 5% uithout endcngering the hoolth end cofoty of the public. NUREG-0881 SO5: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF i WOLF CREEK GENERATING STATION, UNIT 1. Doc ket No. 50-482.(Kansas Gas And Electric Compangeet al)

Division of Licensing. March 1985.

2OOp p. 8504030055. 29599:298. Supplement No. 5 to the Safety Evaluation Report related to the operation of the Wolf Creek Generating Station, Unit No. 1 updates the information. contained in the Safety Evaluation Report, dated April 1982 and Supplements 1, 2, 3 and 4. dated August 1982, June 1983, August 1983, and December 1983, respectively. Supplement No. 5 also addresses open issues and items concerning the issuance of a five j' percent low power license. The Safety Evaluation and its supplements pertains.to the application for a license to operate the Wolf Creek Generating Station, Unit No. 1 filed by Kansas Gas and Electric Company on February 18, 1980. The Construction Permit No. CPPR-147 was issued on May 17.-1977. The facility is located in Coffey County, Kansas. NUREG-0885 104: U. S. NUCLEAR REQULATORY COMMISSION POLICY AND PLANNING QUIDANCE 1985.

Commissioners. February 1985. 25pp. 8503220005.

29495:096. The purposes of the Policy and Planning Guidance document are' to set forth the regulatory approach of the Nuclear Regulatory Commission and to provide the supporting principles to that approachs to state the major policies and planning objectives of the Commissions and to provide a common basis for the development of programs, for the establishment of priorities, and for the allocation of resources. NUREG-0887 S05: SAFETY EVALUATION REPORT RELATED T(' THE OPERATION OF THE PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2. Doc ks t Nos. 50-440 And 4 50-441.(Cleveland Electric Illuminating Company) e Division of Licensing. February 1905. 212pp. 8503050551. P9220: 135. Supplement No. 5 to the Safety Evaluation Report (NUREG-0887) pertains to the application filed by the Cleveland Electric Illuminating Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power l Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power P)ent, Units 1 and 2 (Docket Nos. 50-440 and 50-441). The report has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles i northeast of Cleveland, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 through 4. i l NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS. DEBEVEC,C.J.s HOLLAND,R.A. Emergency Preparedness Branch. March 1985. 21pp. 8504050280. IEB-79-12. 29673:144. i IE Circular 77-07 was issued on April 14, 1977 because of the occurrence of short period scram events at Dresden Unit 2 on December 28, 1976 and at Monticello on February 23, 1977. The circular advised BWR plants to revise their control rod withdrawal seguences and operating procedures to reduce the likelihood of future short period 1 ( 8

o crams. Horavare . ciciler ovanto cantinuod to occur. Th000 included cvents at Dyster Creek on December 14, 197Bs at Browns Ferry Unit 1 on January 18, 1979s and at Hatch Unit 1 on January 31, 1979. As a result of these events. IE Bulletin 79-12 was issued on May 31, 1979. This bulletin required'a written response from licensees of l CE-designed BWRs regarding specific actions listed in the bulletin. i ,All of the licensees responded in a satisfactory manner. No similar ; ovents have been reported since IE Bulletin 79-12 was issued. NUREG-0910 RO1 SO1: NRC COPPREHENSIVE RECORDS DISPOSITION GCHEDULE.

Division of Technical Information & Document Control. January 1985.

1Spp. 8502060493. 28749:329. In compliance with statutory requirements set forth in Title 44 U. S. Code, "Public Printing and Documents " and in the applicable i regulations cited in Title 41 Code of Federal Regulations, "Public l Contracts and Property Management," Chapter 101 Subchapter B,

    " Archives and Records," the U.S.      Nuclear Regulatory Commission oubmitted to the General Services Administration National Archives and Records Services, and to the Comptroller General a schedule (commonig referred to as a disposition or retention schedule) proposing the appropriate duration of retention and the final disposition for records created or maintained by the NRC.

4 NUREG-0910 RO1 SO2: NRC CDPPREHENSIVE RECORDS DISPOSITION SCHEDULE. * . Division of Technical Information & Document C~ontrol. February 1985. 24pp. 8503010298. 29188:287. See NUREG-0910,RO1,901 abstract. NUREG-0933 802: A PRIORITIZATION DF QENERIC SAFETY ISSUES. EMRIT,R.s MINNERS,W.s VANDER MOLEN.H.s et al. Division of Safstu Technology. January'1985. 288pp. 8502210083. 29053:003. The report presents the priority rankings for generic safety issues related to nuclear power plants. The purpose of these rankings :

   .is to assist in the timely and efficient allocation of-NRC resources

, for the resolution of those safety issues that have a significant ! potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts ostimated to result if resolutions of the safety issues were ; implemented, and the consideration of uncertainties and other I quantitative or qualitative factors. To the extent practical, ostimates are quantitative. I ( j NUREG-0936 VO3 NO4: NRC REGULATORY AGENDA.Guarterly i Report,0ctober-December 1984.

Division of Rules and Records.

February 1985. 207pp. 8502250788, 29094:043. The NRC Regulatory Agenda is a compilation of all rules on which the NRC has proposed or is considering action and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The. Regulatory Agenda is updated and issued each quarter. The Agendas for April and October are published in their entirety in the Federal Register while a notice of availability is published in the Federal Register for the January and i July Agendas. 9

NUREG-0940 VO3 NO4: ENFORCEMENT ACTIONS: SICNIFICANT ACTIONS RESOLVED. Guarterly Progress Report,0ctob er-December 1984.

Enforcement Staff. February 1985. 402pp. 8502250240. 29093:001. l This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (October - December 1984) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to licensees with respect to these ,

enforcement actions and the licensees' responses. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, in the interest of promoting pubic health and safety as well as common defense and security. NUREG-0979 903: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN APPROVAL OF THE GESSAR II BWR/6 NUCLEAR ISLAND DESIGN. Docket No. 50-447. (General Electric Company)

Division of Licensing. January 1985. 33pp. 8502110615. 28860:291.

Supplement 3 to the Safety Evaluation Report (SER) for the application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. Subject to

     . favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070.

NUREG-0981 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR ACCIDENT.

Director's Office, Office of Inspection and Enforcement.
Federal Emergency Management Agency. February 1985. 34pp. 8503060141. FEMA-51.

. 29263:287. Procedures have been developed by the U.S. Nuclear Regulatory Commission (NRC) and the Federal Emergency Management Agency (FEMA) which provide the response teams of both agencies with the steps to be taken in responding to an emergency at a commercial nuclear power plant. The emphasis of these procedures is mainly on the interface between NRC and FEMA at their respective Headquarters and Regional Offices and at the various sites at which such an emergency could occur. Detailed procedures are presented that cover for both agencies, notification schemes and manner of activation, organizations at Headquarters and the site, interface procedures, coordination of onsite and offsite operations, the role of the Senior FEMA official, and the coopc etive efforts of each agency's public information staff. NUREG-1021 RO1:-OPERATOR LICENSING EXAMINER STANDARDS.

Division of Human Factors Safety. February 1985. 195pp. 8503200229.

29473:058. The Operator Licensing Examiner Standards provide policy and ; guidance to NRC examiners and establish the procedures and practices ' for examining and licensing of Title 10 of the CODE OF FEDERAL REGULATIONS (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are 10

oubject .to rovician or other intorno1 sporotor oxccinntion licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience. , NUREG-1031 SO1: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MILLSTONE NUCLEAR POWER STATION, UNIT 3. Docket No. 50-423. (Northeast Nuclear Energy Company)

Division of Licensing. March 1985. 75pp.

8504090018. 29753:001. The Safety Evaluati,on Repor,t provides the results of the NRC staff review of Northeast Nuclear Energy Company's application for a. license to operate the Millstone Nuclear Power Plant, Unit No. 3. The ! 4 facility is located in Waterford Township, New London, Connecticut. ' , This Supplement No. 1 updates the information contained in the Safety Evaluation Report, dated July 1984. This supplement also addresses the ACRS Report issued September 10, 1984. NUREG-1047: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF NINE MILE POINT NUCLEAR STATION UNIT NO. 2. Docket No. 50-410. (Niagara Mohawk Power Corporation,et al)

Division of Licensing. February 1985. 652pp. 8502210335. 29054:001.

The Safety Evaluation Report for the application filed by the Niagara Mohawk Power Corporation, as app licant and co-owner, for a license to operate the Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410), has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located near Oswego, New York. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public. NUREG-1048 SO1: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF HOPE CREEK GENERATING STATION. Docket No. 50-354.(Public Service Electric and Gas Company)

Division of Licensing. March 1985.

77pp. 8503270534. 29542: 275.; Supplement No. 1 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company as applicant for itself and Atlantic City Electric Company, as owners, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of'the U.S.

Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement r reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report. NUREG-1073: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF RIVER BEND STATION. Docket No. 50-458.(Gulf States Utilities And Cajun Electric Power Cooperative)

Division of Licensing. January 1985.

140pp. 8501240053. 28559:174. This Final Environmental Statement contains the second assessment - of the environmental impact associated with the operation of River Bend Station, pursuant to the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51, as amended, of the Nuclear Regulatory Commission regulations. This statement examines the environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs. 11 .

4 4 NURED-1087: FINAL ENVIRONMENTAL GTATEMENT RELATED TO THE OPERATION OF VOGTLE ELECTRIC OENERATING PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Georgia Power Company)

Division of Licensing. March 1985.

400pp. 8504090235. 29748:035. l This Final Environmental Statement contains an assessment of the environmental impact associated with the operation of the Vogtle j Electric Generating Plant, Units 1 and 2, pursuant to the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51 (10 CFR 51), as amended, of the Nuclear L Regulatory Commission regulations. This statement examines the environmental impacts, environmental consequences and mitigating actions, and environmental and economic benefits and costs associated with station operation. NUREG-1089: TECHNICAL SPECIFICATIONS FOR FERMI-2. Docket No. 50-341. , (Detroit Edison Company)

Division of Licensing. March 1985.

501pp. 8504050281. 29672:001. -

;          The Fermi-2 Facility Technical Specifications were prepared by i

the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50. 36 of 10 CFR Part 50 f or the protection of the health and safety of the public, t i NUREG-1096: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FDR THE TRIGA TRAINING AND RESEARCH REACTOR AT THE UNIVERSITY DF UTAH. Docket No. 50-407. (University of Utah)

i Division of Licensing. -March 1985. 70pp. 8504090013. 29754:045.

This Safety Evaluation Report for the application filed by the ! University of Utah (UU) for a renewal of Operating License R-126 to l continue to operate a training and research reactor facility has been i prepared by the Office of Nuclear Reactor Regulation of the U.S. 1

Nuclear Regulatory Commission. The facility is owned and operated by the University of Utah and is located on its campus in Salt Lake City.

Salt Lake County, Utah. The staff concludes that this training reactor facility can continue to be operated by UU without endangering the health and safety of the public. t NUREG-1098: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL DF ! OPERATING LICENSE FOR THE RESEARCH REACTOR AT MANHATTAN COLLEGE. I Docket No. 50-199. (Manhattan College)

Division of Licensing.

February 1985. 58pp. 8503130281. 29360:064. This Safety Evaluation Report for the application filed by i Manhattan College (MC) for a renewal of Operating License R-94 to l [ continue to operate the MC O.1 W open-pool training reactor has been ! prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. ] i The facility is owned and operated by , MC and is located two blocks away from the MC main campus in the l Riverdale area of New York City, New York. The staff concludes that the reactor facility can continue to be operated by MC without L endangering the health and safety of the public. ! NUREG-1100 VO1: FY 1986 BUDGET ESTIMATES.

Division of Budget &

l Analysis. January 1985. 87pp. 8502190023. 29013:211. This report contains the fiscal year budget Justifications to 1 Congress. The budget estimates for salaries and expenses for fiscal I { 12 L __ . . . _ - - _ _ _

         -             .             .                            -            =                                                  - -

gocr 1986-87 provido for obligations of C429,000,000 to bo funded in l total bv a new appropriation. 1__ l NUREG-1103: CONTAMINATED MEXICAN STEEL. Importation Of Steel-Into The United States That Had Been Inadvertently Contaminated With Cobalt-60 As A Result Df Scrapping Of'A Teletherapy Unit. *- Safeguards & i Materials Program Branch. January 1985. 84pp. 8502110623. 28905:023. This report documents the circumstances contributing to the inadvertent melting of Co-60 contaminated scrap metal'in two Mexican e steel foundries and the subsequent distribution of contaminated steel , products into the United States. The report covers the tracing of the source to its origin, response actions to recover radioactive steel in the United States, and return of the contaminated materials to Mexico. Information outside of this scope is recounted as necessary, e. g . , details of the incident on the Mexican side of the border. The i incident resulted in.very significant exposure to citizens of the United States. i NUREG-1104: -TECHNICAL SPECIFICATIONS FOR WOLF CREEK OENERATING STATION, UNIT 1. Docket No. 50-482. (Kansas Gas And Electric Company)

Division of Licensing. March 1985. 500pp. 8504030425. 29601:354.

The Wolf Creek Generating Station, Unit 1 Technical . I Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. 4 NUREG-1105: REVIEW AND EVALUATION OF TE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM FOR FISCAL YEARS 1986 AND 1987.

ACRS -

Advisory Committee on Reactor Safeguards. February 1985. 59pp. 8503010055. 29186:001. Public Law 95-209 includes a requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on the safety research~ program of the Nuclear Regulatory Commission. This report presents the results of the ACRS review and evaluation of

       =the NRC safety research program for Fiscal Years 1986 and 1987.                                                               The

. report contains a number of comments and recommendations.

1 i NUREG-1106: TECHNICAL SPECIFICATIONS FOR CATAWBA NUCLEAR STATION, UNIT , 1. Docket No. 50-413.(Duke Power Company)

Division of Licensing.

January 1985. 525pp. 8502060481. 28746:001. The Catawba Nuclear Station, Unit 1, Technical Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public.

NUREG-1108: RADIOACTIVITY TRANSPORT FOLLOWING STEAM OENERATOR TUBE R UPTURE. HOPENFELD.J. March 1985.

Division of Accident Evaluation. 4 5p p. . 8504040001. 29630:269. A review of the capabilities of the CITADEL computer code as well as plant experience to' project radioactivity releases following steam generator tube rupture in PWR's shows that certain experimental data 13

  .is noodod for rolichlo offnito dono predictiano.           This crticlo dofinos fivo perccatoro ehich are the key.for such prodictions cnd discuscos the functional dependence of these parameters on various operational variables. A joint Westinghouse, Electric Power Research Institate, and the Nuclear Regulatory Commission program aimed at obtaining the five parameters empirically is described.           Present status of the CITADEL code is also reviewed.                                                             i NUR EG-1110: COMPARISON OF LICENSING ACTIVITIES FOR OPERATING PLANTS                        '

DESIGNED BY BABCOCK & WILCOX. THOMA J.O. Division of Licensing. January 1985. 29pp. 8502070590. 28807:175. This report provides a comparison of a number of licensing activities for the operating Babcock & Wilcox (B&W) plants with emphasis on Rancho Seco. The factors selected were a comparison of staff resources expended in FYB4, active licensing action reviews, implementation of NUREG-0737 modifications, exemptions to regulations, SALP reports, enforcement actions, and Licensee Event Reports (LERs). The eight licensed operating plants examined are as follows: Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3, Davis Besse, Oconee Units 1, 2, and 3. Rancho Seco, and Three Mile Island Unit 1 (TMI-1). NUREG-1112: ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF SPECI AL NUCLEAR MATERIAL LICENSE NO. SNM-368.(UNC Naval Products Division Of UNC Resources Inc)

Division of Fuel Cycle & Material Safety. January 1985. 74pp. 8502120056. 28871:266.

This Environmental Assessment is issued by the U.S. Nuclear Regulatory Commission (NRC) in response to an application by UNC Naval Products, Division of UNC Resources, Inc., for the renewal of Special Nuclear Material (SNM) License No. SNM-368 for the operation of the existing fuel fabrication facility. NUREG-1113: TECHNICAL SPECIFICATIONS FOR BYRON STATION UNITS 1 AND 2. Docket Nos.50-454 And 50-455.(Commonwealth Edison Company)

Division of Licensing. February 1985. 510pp. 8503110132.

29326:001. The Byron Station, Unit 1 and Unit 2 Technical Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. NUREG-1115: CATEGORIZATION OF REACTOR SAFETY ISSUES FROM A RISK P ERSPECTIVE.

Division of Risk Analysis & Operations (post 840429).

March 1985. 167pp. 8504030427. 29598:072. This report presents the results of an effort to identify and rank reactor safety and risk issues identified from past Probabilistic Risk Assessments (PRAs) and other safety analyses. Because of the varied scope of these analyses, the list of issues may be incomplete. Nevertheless, those studies comprised ordered analyses to whatever their respective depthss hence, they warranted scrutiny for whatever insights they could reveal with respect to issue importance. The top ranked issues in terms of their contribution to the uncertainty in risk are described in some detail. All of these risk issues are compared to the " generic safety issues" for completeness and omission. 14

NURE3-1117: TECHNICAL SPECIFICATIONS FOR WATERFORD STEAM ELECTRIC CTATION UNIT 3.Dsckot ND.' 50-382.(Leuioicna Pcwor And Light Corpcng) a - Office of Nuclear Reactor Regulation, Director. March 1985. 400pp. 8504030439. 29600:178. The Waterford, Unit 3 Technical Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear 4 reactor facility as set forth in Section 50. 36 of 10 CFR Part 50 f or the protection of the health and safety of the public. NUREG-1130: ENVIRONMENTAL ASSESSMENT FOR RENEWAL AND CONSOLIDATION OF MATERIALS LICENSE NOS. Sffi-362,SMB-405,08-00566-05, 08-00566-10.AND 08-00566-12.

Division of Fuel Cycle & Material Safety. March 1985. 45pp. 8504080548. 29713:038.

This Environmental Assessment is issued by the U.S. Nuclear Regulatory Commission (NRC) in_ response to an application by the U.S.

Department of Commerce, National Bureau of Standards, for the renewal and consolidation of five Materials Licenses for radiological activities at the National Bureau of Standards site.

NUREG/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFDRMATION MEETING. SZAWLEWICZ,S.A. Office of Nuclear Regulatory Research, Director. January 1985. 434pp. 8502040194.

28714: 188.

The papers published in this six volume report were presented at 1 the Twelfth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland during the week of October 22-26, 1984. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty-six different papers presented by researchers from seven European countries, Japan, and Canada. Volume 1 presents information on Plenary Session - I, Integral System Tests, Separate Effects. International Programs in Thermal Hydraulics, and Calculation of Appendix K Conservatisms.

 - NUREG/CP-OO58 VO2: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A.                                                               Office of Nuclear l     Regulatory Research, Director.                                                       January 1985.          459pp. 8502060596.

l 28753:039. l The papers published in this six volume report were presented at l the Twelfth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland during the week of October 22-26, 1984. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty-six different papers presented by researchers from seven European countries, Japan, and Canada. Volume 2 presents information on i Pressurized Thermal Shock, Code Assessment and Improvement, 2D/3D Research Program, and the Nuclear Plant Analyzer Program. NUREG/CP-OO50 VO3: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

Office of Nuclear Regulatory Research, Director. January 1985. 731pp. 8502060432. 28750:001. i The papers published in this six volume report were presented at the Twelfth Water Reactor Safety Research Information Meeting held at

15

tho Nsticnol Burocu of Standards, Gaithornburg, Mary 1cnd during the week of October 22-26, 1984. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty-six dif ferent papers presented by researchers from seven European countries, Japan, and Canada. Volume 3 presents information on Containment Systems Research, Fuel Systems Research, Accident Source , Term Assessment, and Japanese Industry Safety Research. NUREG/CP-OO58 VO4: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A. Office of Nuclear Regulatory Research, Director. January 1985. 388p p. 8502060428. 28752:012. The papers published in this six volume report were presented at the Twelfth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Caithersburg, Maryland during the week of October 22-26, 1984. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty-six dif ferent papers presented by researchers from seven European countries, Japan, and Canada. Volume 4 presents information on Materials Engineering Research. NUREC/CP-OO58 VO5: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A. Office of Nuclear Regulatory Research, Director. January 1985. 470pp. 8502060359. 28755:345. The papers published in this six volume report were presented at the Twelfth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland during the week of October 22-26, 1984. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty-six different papers presented by researchers from seven European countries, Japan, and Canada. Volume 5 presents information on

. Mechanical Engineering, Structural Engineering, Seismic Research, Process Control, Instrumentation and Control Program, and Equipment
Gualification and Nuclear Plant Aging.
NUREG/CP-OO58 VO6: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. SZAWLEWICZ,S.A. Office of Nuclear i

         - Regulatory Research, Director.                                                      January 1985.        515pp. 8502060357.

4 28757:094. , The papers published in this six volume report were presented at the Twelfth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland during the week of October 22-26, 1984. The papers describe progress and results ) of programs in nuclear safety research conducted in this country and l abroad. Foreign participation in the meeting included twenty-six different papers presented by researchers from seven European i countries, Japan, and Canada. Volume 6 presents information on ) i Plenary Session - II, Human Factors and Safeguards Research, Health l Effects and Radiation Protection, Risk Analysis, and EPRI Safety 1 Research. i i i 16

 , .ew-  - . - - - - - , - - -,-,,.---,--.we-,---,-----,-         - - - -  ---------,---w---w       - - - - - - - ,

NUREG/CR-2OOO VO3N12: . LICENSEE EVENT REPORT (LER) COMPILATION: For Manth Of December 1984.

Dak Ridge National Laboratory. January 1985.

49pp. 8501280397. . ORNL/NSIC-2OO. 28572:297.

                     ,This month 1g report contains Licensee Event Report-(LER) sperational information that was processed into the LER data file of the Nuclear Safety.Information Center (NSIC) during the one month period. identified on the cover of the document.         The LERs, from which this.information is derived, are submitted to the Nuclear Regulatory Commission (NRC) by nuclear power plant licensees in accordance with federal regulations.      Procedures for LER reporting for those. events.

(and revisions to those events) occurring prior to 1984 are described in NRC Regulatory. Guide 1.16 and NUREG-0161, _ Instructions for Preparation of Data Entry Sheets f or Licensee Event Reports. For those events occurring on and after. January 1. 1984, LERs are being cubmitted in accordance with the revised rule contained in Title 10 Part 50.73 of the Code of Federal Regulations (10 CFR 50.73-144) on

              ' July 26, 1983. NUREG-1022, Licensee Event Report System - Description of Systems and Guidelines for Reporting, provides supporting guidance and information on the revised LER-rule. -The LER summaries in this report are arranged alphabetically by facility name and then chronologically by event date for each facility.          Component, system, keyword, .sgstem and general keyword indexes are assigned by the computer using correlation tables-from the Sequence Coding and. Search System.

NURES/CR-2OOO YO4 N1: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of January 1985.

Dak Ridge National Laboratory. February 1985.

59pp. 8503150302. OR NL/NSIC-2OO. 29390:226. See NUREQ/CR-2OOO,VO3 N12 abstract. NUREO/CR-2OOO VO4 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month

              'Of February-1985.

Dak Ridge National Laboratory. February 1985.

70pp. 8504030412. ORNL/NSIC-2OO. 29604:359. See NUREO/CR-2OOO,VO3,N12 abstract. NURE9/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. WEISS.A.J. Brookhaven National Laboratory. -February 1985. 146pp. 8503090486. BNL-NUREG-51454. 29295:096.

This progress report will describe current activities and

! technical. progress in the programs at Brookhaven National Laboratory , ' sponsored by the Division of Accident Evaluation, Division of Engineering Technology, and Division of Risk Analysis & Operations of the U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. The projects reported are the following: High Temperature Reactor Research, SSC Development, Validation and Application, CRBR l Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, Development of Plant Analyz er, Code Assessment and l 7 Application (Transient and LOCA Analyses), Thermal Reactor Code Development (RAMONA-3B), Calculational Guality Assurance in Support of PTSs Stress Corrosion Cracking of PWR Steam Generator Tubing, Probability Based Load Combinations for Design of Category I Structures, Mechanical Piping Benchmark Problems, Identification of , Age-Related Failure Modess Analysis of Human Error Data for Nuclear Power Plant Safety Related Events, Human Factors Aspects of Cafety/ Safeguards Interactions, Emergency Action Levels, and j Protective Action Decision Making. 4 17

ww,.*wm -w-y

NUREG/CR-2482 VO6: REVIEW OF DOE WASTE PACKAEE PROGRAM. Subtesk 1.1 -

    -National Waste Package Program. October 1983 - March 1984. SOO, P.

Brookhaven National Laboratory. March 1985. 59pp. 8504030414. B NL-NUREG-51494. 29605:065. This report is part of an ongoing effort to review the national high level waste package program. The contributions of individual waste package components to containment and controlled release of radionuclides after emplacement in salt, basalt, tuff and granite repositories are evaluated. The U.S. crystalline (granite) repository program is reviewed and relevant foreign data are outlined. The use of crushed salt, bentonite and zeolite-containing packing materials is _ discussed. Temperatures and gamma irradiation are shown to be important environmental parameters in assessing waste package performance. NUREG/CR-2482 VO7: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 - National Weste Package Program April 1984 - September 1984. S00, P. Brookhaven National Laboratory. March 1985. 88pp. 8504030408. BNL-NUREG-51494. 29604:266. ' The present e'ffort is part of an ongoing task to review the national high level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, tuff, and granite repositories. In the current Biannual Report a review was carried out.to determine the ability of spent fuel cladding to provide additional radionuclide containment capability should the container /overpack system fail prematurely. NUREG/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIDACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981. BAKER,D.A.; PELOGUIN R.A. Battelle Memorial Institute, Pacific Northwest Laboratories. January 1985. 128pp. 8502060603. PNL-4221, 28745:070. Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1981. Fifty year dose commitments from a one year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each site. This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age group and organ. Also included for each site is a histogram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne patt. ways from 48 sites ranged from a high of 20 person-rem to a low of 0.008 person-rem with an arithmetic mean of 3 person-rem. The total population dose for all sites was estimated at 160 person-rem for the 98 million people considered at risk. The average individual dose commitment from all pathways on a site basis ranged from a low of 1 x 10(-5) mrem to a high of 0.05 mr om. No attempt was made in this study to determine the maximum dose commitment received by any one individual from the radionuclides released at any of the sites.

 .                                         18

_ _ - .. .- _ - - = -__ - . . . . . - _ - - - - _ . NURE3/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIDACTIVE HATERIALS. MONRDE,R.E.s i WOO,H.H. SEARS,R.G. Lawrence Livermore National Laboratory. March ! 1985. 16pp. 8504040007. UCRL-53044. , 29619:236. Welding and related operations are evaluated to assess the centrols required to prevent weld-related failure of shipping centainers used for transportation of radioactive materials. The rcport includes (1) recommended criteria for controlling welding as cpplied to shipping containers, and (2) a discussion of modifications of the recommended industry Codes as applied to shipping containers. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACGUISTION OF LICENSEE EVENT DATA. KATO,W.Y.s HALL,R.E.; TEICHMANN,T.s et al. Brookhaven National Laboratory. February 1985. 267pp. -8503080498. BNL-NUREG-51609, 29285:134. Brookhaven National Laboratory's Department of Nuclear Energy (DNE) has performed a study of the Licensee Event Report (LER) system. The objective of the study was to assess the feasibility of codifying the LER reporting system as proposed by NRC-AEOD, and/or developing an alternative plan that would in addition collect information about significant events amenable to statistical analysis, cuch as multi-case, multi-variate analysis. The study indicated that the LERs constitute reports from a large variety of events which have in most cases mang different plant parameters, both measured and currentig not measured, to characterire the event. In order to dotermine event-specific plant parameters required for statistical and doterministic analysis, a data matrix approach could be measured and rccorded, and those which are required for certain types of events involving thermal-hydraulics and neutronics as illustrative of events roquiring in-depth analysis. Also included in the study was a review of INPO's Nuclear Plant Reliability Data Systems NASA's Problem Roporting and Corrective Action (PRACA) programs Electricite de France's KIT system, an automatic computer-based reactor parameter osnitoring and recording systems and the regulatory relationship botween the FAA and the commercial airline industry.

NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS.

l R ANDOLPH. P. D. s ISAACSON, L. s AYERS, A. L. s et al. EG&G, Inc. January 30, 1985. 122pp. 8502130454. EQQ-2251. 28914:224. A study has been performed to evaluate problems associated with ! the existence of flammable or explosive gas mixtures in Pressurized l Water Reactor waste gas systems. Information on existing waste gas systems, waste gas concentrations, and gas monitoring instrumentation l ~ obtained from six operating nuclear power plants is summarized. A ccmparative risk evaluation has been performed for several generic types and configurations of PWR waste gas systems. Waste gas systems , in the plants visited are included and categorized as part of the risk l ovaluation. Existing data on the effect of initial pressure on flammability limits, as well as recently reported data on flammability

and detonability of hydrogen / air mixture has been collected and

! cummarized. A survey of commercially available instruments for ocnitoring hydrogen and oxygen concentrations has been performed and the results tabulated. A series of observations, conclusions and i recommendations are given. NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY DN THE RATES DF HYDROGEN GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS.

19

LOYOLA,V.M.s WOMELCDUFF,J.E. Brndio Natienol Lcbercterion. Januorg 1985. 40pp. 8502130366. SAND 83-1326. 29920:328. The rates of hydrogen generation are measured for the corrosion of. galvanized steel in three different light water cooled reactor (LWR) water chemistries. The results were obtained oven a temperature range of 100 degrees to 175 degrees centigrade and indicate that in a boiling water reactor (BWR) water chemistry, the reaction is. faster then.in those of two pressurized water reactors (PWR 's ) . A mechanism is proposed which would explain the observed results without requiring that the chemical additives come in direct contact with the corrodible unoxidized metal.. Such a mechanism is required because electron microprobe analysis suggests that_no chemical additives have diffused into the protective ZnO layer which forms on the unoxidized metal. Arrhenius parameters are calculated for the three chemistries, but some questions are raised about whether those parameters are associated with a diffusion process or with the actual hydrogen producing reaction. NURE9/CR-3430 VO2: NUCLEAR POWER PLANT OPERATING EXPERIENCE - 1982. Annual Report. SILVER,E.G. Oak Ridge National Laboratory. January 1995. 393pp. 8502150078. 29004:010. This report is the ninth in a series of reports issued annually that summarizes the operating experience of nuclear power plants in commercial operation in the United States. Power generation statistics, plant outages, reportable occurrencese fuel element performance, and occupational radiation exposure for each plant are presented and discussed, and summary highlights are given. The report includes 1982 data from 72 plants; 24 boiling-water-reactor plants, 47 pressurized-water-r,eactor plants, and 1 high-temperature gas-cooled reactor plant. NUREG/CR-34BS VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of

Trajectories, Concentration Patterns And MEBODIF Model Calculations.

l START G. E. s CATE,J.H.s SAGENDORF. J. F. s et al. Commerce, Dept. o f, j Natl. Oceanographic & Atmospheric Administration. February 1985. ? 75pp. 8503120094. 29340:116. The 1981 Idahe Field Experiment was conducted in southeastern Idaho over the Upper' Snake River Plain. Nine test-day case studies were measured between July 15 and 30, 1981. Eight-hour releases of SF(6) gaseous tracer were made from 46 m above the ground. Tracer was sampled hourly, for 12 seguential hours at about 100 locations within l an aree 24 km square. Also, a single total integrated sample of about

30 hours duration was collected at approximately 100 sites within an area 48 by 72 km (using 6 km spacings). Extensive tower profiles of meteorology at the release point were collected. RAWINSONDES, RABALS and PIBALS,were collected at 3 to 5 sites. Horizontal, low-altitude i

Winds were monitored using the INEL MESONET. SF(6) tracer plumes were t- marked with co-located oil fog releases and bi-hourig seguential

' levnches of tetroon pairs. Aerial LIDAR observations of the oil fog plume and airborne samples of SF(6) were collected. High altitude serial photographs of daytime plumes were also collected. The Idaho c

! Field Experiment is reported in three volumes. Volume 3 contains ! ' descriptions of the nine intensive measurement days. General ! meteorological conditions are described, trajectories and their relationships to analyses of gaseous tracer data are discussed, and

overviews of test day cases are presented. Calculations using the l ARLFRD MEBODIF model are included and related to the gaseous tracer l data. Finally, a summary and list of recommendations are presented, l

20

a 4 NURE0/CR-3498: TWO-DIMEN3IONAL MODELING OF INTRA-SUBA!SEMBLY HEAT. TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. l MHATIB-RAH 8ARs CAZZDLI,E.G. Brookhaven National Laboratory. January 1985, 179pp. 8501210102. BNL-NUREG-51713. 28496:154. Phenomenologica1'models and numerical techniques for prediction of coolant flow and' temperature fields during forced, mixed, and free convection-regimes of operation in LMFBR subassemblies are addressed. ,

;-                It.is shown that, simplified integral solutions provide an excellent                                                                                     !
                 .cpproach to assessing the importance of the intra-subassembly euoyancy                                                                                   i 1                  induced flow redistribution, and the transverse thermal conduction and

c1xing effects on the assembly wide peak coolant temperatures. ,

Furthermore, a more detailed steady-state and transient parabolic two-dimensional porous-body model, resulting in the TWIST computer code.is developed. Comparison of calculated results and out-of pile codium and water test data indicate generally good agreement in cross-assembly temperature profiles. However, the impact of fuel pin distortion and bowing, caused by large transverse power gradients on transverse distributions are found to be significant. j NURE9/CR-3516: A-SURVEY OF THE USES OF RADIDACTIVE MATERIALS IN LOUISIANA'S OFFSHORE WATERS. BENNETT,J.J.s HOOK,S.E.s PALAZZO,R.J.4 ot al. Louisiana, State of. February 1985. 37pp. 8503130140.' i 29360:122. I As a result of a contract agreement with the U.S. Nuclear Regulatory Commission, the State of Louisiana, and in particular, the Louisiana Nuclear Energy Division (LNED), conducted a survey of the use of radioactive materials in Louisiana's " offshore waters." Offshore waters are here defined as "that area of land and water on t cnd above the United States' Outer Continental Shelf." The objectives i of the survey were fourfold: 1) identification of those licensees using radioactive materials offshare Louisianas 2) identification of tork locations where radioactive materials are being useds 3) a , description of the types of work performeds and 4) performance of at ; , least three site visits to offshore locations where radioactive I

catorials are being used. By telephone survey, LNED attempted to l contact those licensees thought to be using radioactive materials I offshore. Of the 69 licensees reached by telephone, 43, or 61%,

indicated they have current offshore activities. The results of the j telephone survey, conducted in May-June 1983 are presented in detail in this report. To meet objective four of the survey, three visits core made to offshore rigs and platforms, two involving industrial radiography and one involving well-logging. Also included in this } report are summaries of these visits and a description of previous l cork done by LNED concerning radiation safety on " lay barges." l i l- NUREG/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT j REPORTS. VOSKA,K.J.s O 'BRIEN. J. N. Brookhaven National Laboratory. February 1985, 116pp. 8502220414. BNL-NUREG-51717, 29062:163. .; The objective of this report is to present a method for using " l field data from nuclear power plants to estimate human error , probabilities (HEPs). These HEPs are then used in probabilistic risk , cctivities. This method of estimating HEPs is one of four being ' pursued in NRC-sponsored research. The other three are (1) structured ,

orpert Judgment, (2) analysis of training simulator data, and (3) 1 performance modeling. The type.of field data analyzed in this report is from Licensee Event Reports (LERs) which are analyzed using a

                -cethod specifically developed for that purpose.                                              However, any type of field data.or human errors could be analyzed using this method with 21

cinse adjustsonto. Thio report acconcos the practicality, acceptability,1 and usefulness of estimating HEPs from LERs and comprehensively presents the method of use. NUREG/CR-3659: A MATHEMATIGAL MODEL FOR ASSESSING THE UNCERTAINTIES OF INSTdVMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. HES80N,0. M. 3 CLIFF,W.C.s STEVENS.D.L. Battelle Memorial Institute, Pacific Northwest Laboratories. February 1985. 48pp. 8503220007. PNL-4973. 29487:217. A method of assessing the quantitative uncertainties in the determination of PWR powers and coolant flows caused by measurement

                  . uncertainties'is provided.                                            The method defines the parameters entering into the calculation, the types and sources of measurement errors which must be considered, together with sources of quantitative data for the uncertainties.                              A mathematical model is developed which combines the measurement uncertainties in a rigorous statistical manner to give the overall uncertainty in the desired parameter together with a sample calculation.

NUREG/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly Induced By Earthquakes. RAVINDRA,M.K.s C AMPBELL, R. D. s KENNEDY,R.P.s et al. Lawrence Livermore National Laboratory. February 1985. 199pp. 8503130008. UCID-19988. 29360:183. The requirements to design the nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant loop (RCL) piping have led to excessive design costs, interference of normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/ Lawrence Livermore National Laboratory sponsored research program aimed at demonstrating that the probability of DEGB in RCL piping of nuclear power plants is acceptabig small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimated the probability of indirect DEGB in RCL piping as conseguence of seismic-induced structural failures within the containment of Westinghouse supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEOS was estimated to be about 3x10(-6) per year with a 10% to 90% subjective probability range approximately for 1x10(-7) per year to 4x10(-5) per year. NUREG/CR-3663 VO1: PROBABILITY DF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summery Report. HOLMAN,0.5.s LO T. s CHOU.C.K. Lawrence Livermore National i Laboratory. January 1985. 81pp. 8502120055. UCRL-53500 VO1. '

h. 20072:057.

[. As part of its reevaluation of the double-ended guillotine break j' (DEGB) as a design requirement for reactor coolant piping, the U.S. Nuclear Regulatory Commission (NRC) contracted with the Lawrence ,' Livermore National Laboratory (LLNL) to estimate the probability of ] 4 occurrence of.a DEGB, and to assess the effect that earthquakes have j en DEGB probability. This report describes a probabilistic evaluation i 1 of reactor coolant loop piping in PWR plants having nuclear steam

suppig systems designed by Combustion Engineering. Two causes of pipe {

break were considered: pipe fracture due to the growth of cracks ist welded Joints (" direct" DEGB), and pipe rupture indirect 1g caused by ; ! 22 i

  - - . a . - - -     .-c . - . .. - . - . , . - ~ ~ ,            - - - . . , - - - , - -         . . . . . . ~ . -

1 4 failuro of cecpansnt oupporto duo to en ocrthgucho (" indirect" DECB). d The probability of direct DEGB was estimated using a probabilistic

!          ' fracture mechanics model. The probability of indirect DEGB was.

ostimated by estimating support fragility and then convolving fragility with seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for j' reactor coolant loop piping in these plants, and that NRC should t therefore consider eliminating DEGB as a design basis in favor of more realistic criteria. ] 4 f

NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR CDOLANT; LOOPS OF COMBUSTION ENGIPEEERING PWR PLANTS, Volume 3: Double Ended Ou111otine Break Indirectiv Induced By Earthquakes. RAVINDRA M.K.-
C AMPBELL, R. D. s KENNEDY,R.P.s et al. Lawrence Livermore National ,
Laboratory. January 1985. 118pp. 8501280412. 28573: 001. '

i The.reguirements to design the nuclear power plants for the , 'offects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant loop (RCL) piping have led to excessive design i costs, interference of normal plant operation and maintenance )~ personnel. This report describes an aspect of the NRC/ Lawrence Livermore National Laboratory sponsored research program aimed at i demonstrating that the probability of DEGB in RCL piping of nuclear power plants is acceptably small and the requirements to design for , the DEGB effects (e.g., provision of pipe whip restraints) may be i removed. This study estimated the probability of indirect DEGB in RCL j piping as a conseguence of seismic-induced structural failures within

!            the containment of Combustion Engineering supplied pressurized water i            reactor nuclear power plants in the United States.                                                        The median 1             probability of indirect DEGB was estimated to be in the range of j             10(-6) per year of older plants, and less than 10(-8) per year-for i            codern plantss using very conservative assumptions, the 90% subjective

j. probability value (confidence) of P DEGB was found to be less than t i

5x10(-5) per year for older plants and less than 3 10(-7) per year for

.codern plants.

I NUREG/CR-3688 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT

. JUDGMENT. Volume 1: Main Report. COMER M.K.s BEAVER,D.A.s j BTILLWELL, W. G. s et al. General Physics Corp. January 1985. 61pp.

! G502210260. SAND 84-7115, 29028:272. ) 4 The U.S. Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness

cf several different methods for obtaining human reliability data and

, ootimates that can be used in nuclear power plant probabilistic risk j cssessments (PRA). One method, investigated as part of this overall rosearch program, uses expert Judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using

:omport Judgment: paired comparisons and direct numerical estimation.

Volume 1 of this report provides a brief overview of the background of 4 the project, the procedure for using psychological scaling techniques 1 to generate HEP estimates and conclusions from evaluation of the

tochniques. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to I ovaluate the techniques,,and HEP estimates generated as part of this e

project. The results of the evaluation indicate that techniques using i omport Judgment should be given strong consideration for use in developing HEP estimates. In addition, HEP estimates for 35 tasks t r

N rolct3d to boiling cotor roccters (BWRs) waro cbtainod no part of tho 4 e va,1ua t i on. These HEP estimates are also included in the report. 4 N REG /CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT i JUDGMENT. Volume 2: Appendices. COMER M.K.s SEAVER,D.A.; l STILLWELL,W.G.s et al. General Physics Corp. January 1985. 176pp. ' 8502210262. ' SAND 84-7115. 29028:095. ! See NUREG/CR-3688,VO1 abstract. NUREG/CR-3709: METHODS OF MINIMIZING GROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. DEUTSCH W.J.s MARTIN,W.J.s

  -        EARY,L.E.: et al.                  Battelle Memorial Institute, Pacific Northwest Laboratories.             March 1985.        145pp. 8504040003.              PNL-5319.

29629: 075.^

                       -This is the final report of a research project dealing with methods of minimiring ground-water contamination from in situ leach uranium mining.              Field work and laboratory experiments were conducted to identify excursion indicators for monitoring purposes during mining, and to evaluate effective aquifer restoration techniques following mining.                  Many of the solution constituents were found to be too reactive with the aguifer sediments to reliably indicate excursion 4

of leaching solution from the ore zones however, in many cases, the concentrations of chloride and sulfate and the total dissolved solids level of the solution were found to be good excursion indicators. l Aquifer restoration by ground-water sweeping consumed ground water and was not effective for the redox-sensitive contaminants often present

in the ore zone. Surface treatment methods were effective in lowering
the amount of water used, but also had the potential for creating i conditions in the aguifer under which the redox-sensitive contaminants would be mobile. In situ restoration by chemical reduction, in which a reducing agent is added to the solution recirculated through the ore zone during restoration, has the capability of restoring the ore zone sediment as well as the ground water. This method could lead to a stable chemical condition in the aquifer similar to conditions before mining, i

NUREG/CR-3723: STRESS-INTENSITY-FACTDR INFLUENCE CDEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. BALL.D.C.s BASS B.R.s BRYSON J.W.: et al. Oak Ridge National Laboratory. February 1985. 53pp.

8503290280. ORNL/CSD/TM-216. 29563: 295.

In the fracture-mechanics analysis of reactor pressure vessels, stress-intensity-factor influence coefficients are used in conjunction ' with superposition techniques to reduce the cost of calculating stress-intensity factors. The present study uses a finite-element ! code, together with a virtual crack extension technique, to obtain influence coefficients for semielliptical surface flaws in a cylinder, and particular emphasis was placed on mesh convergence (less than 1% error was sought in the results from any one mesh construction parameter), Comparison of the coefficients with those obtained by other investigators shows good agreement. Furthermore,

stress-intensity factors obtained by superposition for a severe thermal-transient loading condition agree within 1% of the values calculated by a direct finite-element method. Influence coefficients were calculated for three specific axially oriented semielliptical surface flaws. The first was a 2-m-long inner-surface flaw in a

nuclear reactor pressure vessel with depth-to-wall-thickness ratios between 0.2 and 0.9. The second was an inner-surface flaw in the 24 l

rcccter veccol with a curfcco-length-ta-depth vctie of 6 cnd with depth-to-wall-thicknoco retics between 0.05 and 0.2. The third was a 1-m-long flaw on the outer surface of a test vessel with depth-to-wall-thickness ration between 0.1 and 0.9. For the reactor vessel, separate coefficients were calculated for the cladding on the inner surface and for the base-material region. This allows for an - occurate accounting of the effect of thermal stresses in the cladding on the stress-intensity factor for surface flaws that extend through the cladding into the base material. NUR EQ/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review Of Data For Technetium. TILL,J.E. Radiological Assessments Corp. SHOR,R.W.s HOFFMAN,F.O. Oak Ridge National Laboratory. February 1985. 135pp. 8503120452. ORNL/TM-9150. 29340:293. Sources of (99)Tc releases to the environment are reviewed for the uranium fuel cycle considering the recycle of spent uranium fuel cnd no fuel recycling. Without recycling the only source of (99)Tc release is an extremely small amount associated with airborne omissions from the processing of high-level wastes. With recycling, (99)Tc releases are associated with the operation of reprocessing facilities, UF(6) conversion plants, uranium enrichment plants, fuel fabrication facilities, and low- and high-level waste processing and otorage facilities. Among these, the most prominent (99)Tc releases cre from the liquid effluents of uranium enrichment facilities. An oxtensive review of data estimate parameters for predicting the l environmental behavior and fate of (99)Tc indicates a reduced radiological significance for the ingestion of milk and meat. More important pathways of exposure to (99)Tc will probably be associated with drinking water and the consumption of aquatic organisms, garden vegetables, and eggs. For each parameter reviewed in this study, a range of values is recommended for radiological assessment calculations. Where obvious discrepancies exist between these range and the default values listed in USNRC Regulatory Guide 1.109, consideration for revision of the USNRC default values is recommended. NUREQ/CR-3744 V02: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. PUCH,C.E. Oak Ridge National Laboratory. January 1985. 244pp. 8502210332. ORNL/TM-9154/V2. 29052:001. The Heavy-Section Steel Technology (HSST) Program is an ongineering research activity conducted by the Oak Ridge National Laboratory for the Nuclear Regulatory Commission. The program i ' comprises studies related to all areas of the technology of materials fabricated into thick-section primary-coolant containment systems of light-water-cooled nuclear power reactors. The investigation focuses on the behavior and structural integrity of steel pressure vessels containing cracklike flaws. Current work is organized into ten tasks: (1) program management, (2) fracture-methodology and analysis. (3) material characterization and properties, (4) environmentally cssisted crack growth studies, (5) crack arrest technology. (6) irradiation effects studies, (7) cladding evaluations, (8) intermediate vessel tests and analysis, (9) thermal-shock technology, and (10) pressurized thermal-shock technology. NUR EQ/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. WALTERS,W.H.s SKAGGS R.L. Battelle Memorial Institute, Pacific Northwest Laboratories. January 25

l 1985. 55pp. C501280379. PNL-5069. 28572:072. Pacific Northwest Laboratory is studying the citigation of erosion of earthen radon suppression covers for uranium tailings impoundments. Because the covers will require erosion protection for upwards of 1000 years, rock riprap (armoring) has been proposed as the primary protection method. This study investigates the sensitivity of riprap de.-- procedures to extreme flood events that can generate high flow velocities and shear stresses. The study uses two decommissioned tailings sites (Grand Junction and Slick Rock, Colorado) as case studies to evaluate the sensitivity of design rock size with respect to variables such as flood discharge, side slope, specific gravity, safety factor, and channel roughness. The results indicate that design rock size can vary significantly for different design procedures. Other significant results indicate that embankment side slopes of about 4H:1V are optimum f or rock riprap and that the use of rock material with specific gravities less than 2.50 may prove too costly. NUR EQ/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. HARRINGTON,R.M.a FULLER,L.C. Oak Ridge National Laboratory. February 1985. 167pp. 8503120456. DRNL/TM-9163. 29338:179. The BWR-LTAS code was developed by the SASA program at Oak Ridge National Laboratory for the detailed study of specific accident sequences at Browns Ferry Unit One: station blackout, small break LOCA outside primary containment, loss of decay heat removal, loss of vessel water injection, and anticipated transient without scram. The primary use of the code has been to estimate the effects of operator actions on the timing and course of events during the part of the sequence leading up to but not including severe fuel damage. This report documents the basis of the methods used to simulate the response of reactor vessel, primary coolant system, primary containment, and other reactor systems; the output from a sample use of the code is presented. NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED DISPLAYS. BLACKMAN H.S.s GILMORE,W.E. EG&G, Inc. January 1985. 43pp. 8502210188. EGG-2308. 29035:284. The United States Nuclear Regulatory Commission (USNRC) is sponsoring an on going research program to develop methods of assessing various types of computer-generated displays currently being implemented in nuclear power plant control rooms. The purpose of this report is to present an interactive simulation technique for tra evaluation of computer-generated displays. The independent variables for this experiment were transient type (six levels), and display type including the levels of star + control panel, bar + control panel, meter + contrpl panel, pressure-temperature map + control panel, and control panel only. The dependent measures were deviations of parameter values comprising the safety functions at risk, percent of time these parameters were out of tolerence from onset of the transient, and accuracy of the operator path in transient mitigation. The results indicate that an interactive simulation method can be used to evaluate various display types, and that the workstation and computer / simulator is an effective configuration. The implications of these results for display evaluation and design are discussed. 26

' MURE3/CR-3772: RELAPS A7SESSMENT:SEMISCALE SMALL BREAK TESTD C-UT-1,S-UT-2, S-UT-6 S-UT-7 AND S-UT-8. PETERSON,A.C. Sandia National Laboratories. February 1995. 230pp. 8503040547. CAND84-OG84. 29198:001. " The RELAPS independent assessment project is part of an overall offort to evaluate the capability of various system codes to calculate

        .the detailed thermal / hydraulic response of LWRs during accident and eff-normal conditions.                       The RELAP5 - computer code is being assessed cgainst' test data from various integral and separate effects test fccilities.            As part of the assessment effort, several small break tests with and without upper hegd injection ( UHI) of emergency core coolant (ECC), performed in the Semiscale Mod-2A facility, have been analyzed.             The results show that RELAP5/ MOD 1 is capable of calculating ocme aspects of the important phenomena during small breaks both with and without UHI.                The times for the system to depressurize to the UHI cnd/or loop accumulator flow initiation were calculated cottsfactorily.              The correct. trends of the effects of break size and of UHI on.the system pressure response were also calculated. The-injection rate from the~UHI and loop accumulators was not always calculated correct 1gs the flows cycled on and off because large flow ourg0s caused the accumulator pressures to temporarily decrease below the sgstem pressure.                       This cycling of the flow had a significant i

offect on the system response during UHI accumulator flow. When the upper head was liquid-filled from UHI flow, a core liquid level j depression was calculated, but not measured, that resulted in a drgout During UHI flow the calculated densities in the upper of the core. plenum and near the top of the core were too high, which also affected the vessel mass distribution. The calculated break flow rates were too large, when the break uncovered later in the transients, centributing to a low liquid level in the vessel and late-time core heatup. Higher late-time core temperatures were calculated than f coasured both with and without UHI. NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF OE TYPE AK-2

      ' CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. DEAN,R.S.; FOLEY,W.J.s

. MILLS,W.R.s et al. Parameter. Inc. January 1985. 47pp. 6501280735. IEB-79-09. 28629:130. Twelve failures of General Electric Type AK-2 safety-related i circuit breakers reported in 1975, 1978 and 1979 are described in IE Bulletin 79-09. Because of these failures, the bulletin was issued April 17, 1979 to require responses and specific actions by all

licensees and holders of' construction permits. The failures were i- attributed to either binding within the linkage mechanism of the

' undervoltage trip device and trip shaft assembly or faulty adjustment lof that linkage mechanism. It was concluded that the twelve failures resulted from inadequate preventive maintenance. Because many occurrences of the same kind happened after 1979, a significant number of later NRC documents which are included in Appendix A were - issued. The bulletin has been closed out for 101 of the 129 current facilities uhich reported either that they had no Type AK-2 breakers in safety-related systems or none with undervoltage trip devices. Proposed followup items for the remaining 28 current facilities are presented in' Appendix C. Because followup is based on the roquirements of later Bulletins 83-04 and 83-08, Bulletin 79-09 is considered closed. 1 i NUREG/CR-3794: CLOSEQUT OF IE BULLETIN BO-25:DPERATING PROBLEMS WITH TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. FOLEY,W.J.s HENNICK A. l 27

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

k Percmotor, Inc. Jnnuary 1925. 44pp. 0501280090. PARAMETER IE-13. 285;74:157. During the three-month period beginning July 25, 1980, five , j events occurred involving.two types of malfunctions of Target Rock ' safety-relief valves .at Boston Edison Company 's Pilgrim Nuclear Power i Station Unit'1. The first three events were caused by direct failures , l of the valvens the last two events were caused by nitrogen supply ! system overpressure which led to valve-failure. IE Information Notice 80-40 was issued November 7, 1980 to call attention to the two nitrogen overpressure events. As a result of all five events. IE Bulletin 80-25 was issued December 19, 1980 for action to all 30 BWR facilities with operating licenses or near-term operating licenses, and for information onig to 24 facilities then under construction.

Actions were to be taken with' respect to (1) all Target Rock i two-stage, pilot-operated safety-relief valves (SRVs), (2) any make or i ' model of SRV which fails to function as designed, excepting for pressure setpoint requirements and (3) SRV nitrogen / air supply systems. Upon evaluation of utility responses and NRC inspection

_ reports, the bulletin has been closed out for eight of the 30 facilities to which the bulletin was issued for action. For use by NRC/IE, followup items for 22 current facilities with open bulletin status are proposed in Appendix C. Remaining areas of concern and continuing actions dealing with them are described. The bulletin has

served its purpose by resulting in identification of the need for correctiv6 actions at all of the 27 current operating facilities to which the bulletin was issued for action.

l f NUREG/CR-3802: RELAPS ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS. KMETYK,L.N.s BUXTON,L.D.s THOMPSON,S.L. Sandia j National Laboratories. February 1985. 30pp. 8503210472.

SAND 84-1013. 29479:072.

The advanced best-estimate systems codes currently being developed for the NRC are designed to provide realistic, rather than conservative, predictions of LWR plant behavior during a variety of 7 j accidents and transients. The RELAPS independent code assessment i project at Sandia National Laboratories is part of an overall code assessment effort funded by-the NRC to arrive at a qualified Judgment

of the accuracy with which these codes can predict the detailed

- thermal / hydraulic response of LWRs during accident and off-normal i conditions. The heart of the assessment project involves extensive

comparison of code results with test data. RELAPS/ MOD 1 has been assessed at Sandia against a variety of test data from both integral

                               -and separate effects test facilities. A11'these analyses have been documented in detail in individual topical reports and in an overall                                                                                 ,

summary and conclusions report. In this paper, we tabulate the

quantitative key parameters for those scenarios which involve leakage of coolant from the primary system to coatainment, as well as the run time statistics for all the analyses we have performed.

i i NUREG/CR-3804 VO3: PHYSICS OF REACTOR SAFETY.Guarterly , , Report. July-Leptember 1984.

Argonne National Laboratory. January 6 1985. 23pp. 8502210261. ANL-84-35 VO3. 29057:325.

                                         .This guarterly progress report summarizes work done during the months of July-September 1984 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of

' Reactor Safety Research in the U.S. Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety i 28

8 Approicolo Caction. Wark en roccter cero thoraal-hydtculico in i 4 performed at ANL's Components Technology Division, emphasizing dimensional-code development for LMFBR accidents under natural f s convection conditions. An executive summary is provided including a j otatement of the findings and recommendations of the report. i .NUREG/CR-3 SOS'VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task 2 II: Effects Of Ground Motion Characteristics On Structural Response + Cznsidering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. KENNEDY,R.P.s KINCAID R.H.s SHORT,S.A.

Otructural Mechanics Associates. March 1985. 150pp. 8504030436.

, 29604:105.

              'This report presents the results of part of a two-task study on

the engineering characterization of earthquake ground motion for nuclear power plant design. Task I of the study, which is presented

[ :in NUREG/CR-3905, Vol. 1.! developed a basis for selecting design rosponse spectra taking into account the characteristics of free-field ground motion found to be significant in causing structural damage. [- 1Tosk.II incorporates additional considerations of effects of spatial vcriations of ground motions and soil-structure interaction on foundation-motions and' structural response. The results of Task II cre presented in four parts: (1) effects of ground motion characteristics on structural response of a typical PWR reactor building with localized nonlinearities and soil-structure interaction offects: (2) empirical data on spatial variations of earthquake ground actions (3) soil-struc ture . interaction effects on structural responses and (4) summary of conclusions and recommendations based on Tasks I 3 and II studies. This report presents the results of the -first part of Task II. The results of the other parts will be presented in NUREG/CR-3805, Vols. 3-5. NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly _Roport July-September 1984. EDLER,S.K. Battelle Memorial Institute, Pccific Northwest Laboratories. February 1985. 35pp. 8503130105. PNL-5106-3. 29360:151. This document summarizes work performed by Pacific Northwest 7 Lcboratory from July 1 through September 30, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, U. S. Nuclear Regulatory Commission. Results from an instrumented fuel cosembig irradiation program being performed at Halden, Norway, are toported. Accelerated pellet-cladding interaction modeling is being < ccnducted to predict the probability of fuel rod failure under normal

' operating conditions. Experimental data and analytical models are 4

boing provided to aid in decision making regarding pipe-to-pipe icpacts following postulated breaks in high-energy fluid system l piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature

_catorials property tests are being conducted to provide data on severe

, core damage fuel behavior. Thermal-hydraulic models are being dcveloped to provide better digital codes to compute the behavior of i-full-scale reactor systems under postulated accident conditions. Covere fuel damage accident tests are being conducted in the NRU Reactor, Chalk River, Canada. NUREG/CR-3816 VO1: REACTOR SAFETY RESEARCH.Guarterly Roport. January-March 1984.

Sandia National Laboratories. January 29

l ) ! l~ 1985. 160pp. C5013C50074. SAND 84-1072. 28680:095. The overall objective of this report is to provida NRC with a 1 comprehensive data base essential to-(1) defining key safety issues, 3 (2) understanding risk-significant accident-sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number'of activities dealing with l current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to 4 understand important accident seguences and to serve as a basis for development and verification of the complex computer simulation models l and codes used in accident analysis and licensing reviews. Such a ' program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system j performance under a broad variety of abnormal conditions. i NUREG/CR-3817: DEVELOPMENT, USE AND CONTROL OF MAINTENANCE PROCEDURES ~ IN ! NUCLEAR POWER PLANTS. Problems And Recommendations. MORGENSTERN M.s B ARNES, V. E.1 RADFORD, L. R. s et al. Battelle Human Affairs Research Centers. January 1985. 120pp. 8501280403. PNL-5121. 28572:136. I This report describes the results of activities conducted to assess and document the need for guidance or regulatory involvement by the Nuclear: Regulatory Commission (NRC) in the development, upgrading, use and control of maintenance procedures in U.S. nuclear power

plants. Reported are the findings of the following four activities

j (1) a survey of current maintenance procedure practices in seven U. S. nuclear power plants, (2) a review and-analysis of plant administrative and maintenance procedures, (3) a survey of maintenance j procedure practices in industries that share some characteristics with the nuclear industry, and (4) a review of the research pertaining to Job performance mids and a brief analysisHof their applicability to maintenance in nuclear power plants. Based on these findings,.several recommendations for NRC action to upgrade maintenance procedure , j programs are offered. NUREG/CR-3825 VO3-4: ACDUSTIC EMISSION / FLAW RELATIONSHIP FDR IN-SERVICE MONITORING DF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 - September 1984. Volumes 3 and 4. HUTTON,P.H.s KURTZ,R.J. Battelle , Memorial Institute, Pacific Northwest Laboratories. March 1985. 22pp. 8504090010. PNL-5125. 29754:016. Technical progress toward continuous acoustic emission monitoring of nuclear reactor pressure boundaries for flaw detection is described for the period April-September 1984. A draft report of ZB-1 vessel test results was completed. Growth of machined flaws was detected by , AE during both 65 degree C and 285 degree C testing. AE data was generally proportional to crack growth. A key result was clear detection of a natural crack in a fabrication weld by AE. Crack growth rates estimated from AE data compared well with measured crack growth rates. In service hydro test monitoring gave mixed results. Impending failure conditions are readily detectable. However, with low overpressure (1.15 x operating pressure), flaws as deep as 70% through-wall did not produce significant AE. With higher overpressure (1.4 x operating pressure), flaws produced identifying AE. An engineering prototype AE monitor system has been completed for use in operational monitoring at Watts Bar Unit 1 reactor. A modified 30

l l , l l 1

      -cpprocch ta crcck greuth.AE signal identification in producing about      i 95% correct determinations on recorded waveforms from the 28-1 vessel    l 1

test. A report on results from AE monitoring hot functional testing l -at Watts Bar Unit.1 has been published. NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. BOWERMAN,B.S.s SWYLER,K.J.s DOUGHERTY,D.R.s et al. Brookhaven National Laboratory. March 1985. 167pp. 8503280017. BNL-NUREG-51784. 29548:142. The Technical Position on Waste Form and Container Materials (TP)

     -provides guidance to generators of low-level radioactive waste for coeting the regulations under 10.CFR Part 61 governing the disposal of

these wastes. Testing methods are recommended in the TP for assessing caterial properties relevant to long-term performance in shallow land Eburial. These tests were reviewed with respect to their application to specific materials: cement, bitumen, vinyl ester-styrene, and

i. polyethylene. In some cases, the applicability of the tests was found to be inadequate, and modifications to the existing tests or alternative methods were recommended. An experimental evaluation of one of the recommended _ biodegradation tests (the Bartha-Prnmer method) cos also-carried out. Conditions under which this test should be conducted are recommended.

  ' NUREG/CR-3830 VO2: AEROSDL RELEASE AND TRANSPORT PROGRAM. Semiannual         l l       Progress Report For April 1984-September 1984. ADAMS.R.E.3                l TOBIAS,M.L. Oak Ridge National Laboratory. January 1985. 50pp.

8502210193. 29035:233. l This report summarizes progress for the Aerosol Release and ' Transport Program sponsored by the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Division of Accident Evaluation, for the period April 1984-September 1984. Topics

     ' discussed include (1) the experimental program in the Fuel Aerosol Simulant Test Facility. (2) NSPP experiments involving an aerosol of limestone-aggregate concrete in a steam-air atmosphere, (3) revisions in the NSPP experimental program, (4) experiments relating to NSPP thermohydraulic conditions, (5) aerosol-moisture interaction test plans, (6) aerosol code implementation activities, (7) improvements in dcta processing procedures for NSPP experiments, and (8) a study ccmparing in-vessel and ex-vessel cascade impactor aerosol size coasurements in the NSPP.

l NURE9/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FDR NUCLEAR PLANT CDMPONENTS. Interim Report - Diesel Generators Batteries, Chargers And' Inverters. KAHL, W. K. s BORKDWSKI R.J. Oak Ridge National Laboratory. Fcbruary 1985. 109pp. 8502250363. ORNL/TM-9216. 29094:248. The objective of the In-Plant Reliability Data (IPRD) program is to develop a comprehensive, component-specific reliability data base far probabilistic risk assessment and for other statistical analyses rolevant to component reliability evaluations. This objective is i being attained through a cooperative effort with several utilities thich have provided' access to maintenance files and pertinent Population information. This pilot data base includes (1) a component population list.(for each plant) of selected electromechanical and cochanical equipment ( e. g. , pump, valves, etc.), and (2) comprehensive ccaponent failure and repair histories based on corrective maintenance actions on these components. This document is the product of a pilot

i. 31

1 study that was undottaken to demonstrato tho sothodology and i feasibility of app 1ging IPRDS techniques to develop and analyze the reliability characteristics of key electrical components in five nuclear power plants. These electrical components include diesel generators, batteries, battery charges and inverters. The sources used to develop the data base and produce the component failure rates i and mean repair times were the plant equipment lists, plant drawings, maintenance work requests, Final Safety Analysis Reports (FSARs), and interviews with plant personnel. The data spanned approximately 33

reactor-years of commercial operation.

NUREQ/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. KELMERS A.D.; ARNOLD,W.D.s MEYER,R.E.s et al. Oak Ridge National Laboratory. January 1985. 43pp. 8501280387. ORNL/TM-9191/V3. 28572:255. Geochemical information relevant to the retention of radionuclides by candidate high-level waste repositories being developed by Department of Energy (DOE) projects is being evaluated.by Oak Ridge National Laboratory (ORNL) for the Nuclear Regulatory Commission (NRC). During this report period, the project has i evaluated radionuclide sorption and solubility values applicable to the candidate repository site in the Columbia River basalts at the Hanford Reservation. The removal of technetium from pertechnetate-traced groundwater by McCog Canyon basalt under anoxic redox conditions (air excluded) at 27 degrees centigrade was found to be sensitive to the groundwater composition. Sorption of uranium from l groundwater by McCoy, Canyon basalt under oxic redox conditions at 60

degrees centigrade showed low sorption ratios (1.8 to 2.4 L/kg) similar to those previously obtained at 27 degrees centigrade. The average sorption ratio for strontium in groundwater onto McCoy Canyon basalt under oxic redox conditions at 27 degrees centigrade was 225 i

L/kg. Column chromatographic experiments with neptunium in groundwater to measure retardation factors at temperatures from 25 to 80 degrees centigrade gave calculated sorption ratio values that were in good agreement with the values previous 1g obtained in batch contact tests. NUREQ/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS. FISCHER,L.E.s LAI,W. Lawrence Livermore National Laboratory. March 1985. 17pp. 8504040417. UCRL-53544, 29619:007. Criteria are identified for controlling the fabrication of metal

components of shipping containers used for transporting radioactive materials. The criteria have been selected from the ASME Code and are based on the level of radioactive materials being transported and the nuclear safety function of the container's components. Criteria are identified for fabrication processes which are related to materials controls, forming, heat treatment, examination and acceptance testing.

Implementation of the criteria will ensure the structural integrity of shipping containers at levels consistent with the radioactive materials being transported. (

NOTE: Applies to all metals used in shipping containers construction except cast irons.)

NUREC/CR-3865: EVALUATION OF THE RADIOACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. MACKENZIE.D.R.s SMALLEY,J.F.s KEMPF C.R.s et al. Brookhaven National Laboratory. 32 _ _ - . - _ _ . , . ~ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ . - _

h J'nunty 1985. 196pp. C503040014. _BNL-NURED-51792. 29197:113. An: inventory has been compiled of the isotopes of half-life >5 . gears' buried in eight'of the trenches at the Sheffield radioactive ohipment records (RSRs). Pertinent information from some 32OO' fuel

cycle RSRs and 1700 non-fuel' cycle RSRs has been stored in a: ' computerized data base and used to develop the inventory.

Results of the compilation are in' disagreement with the two previous estimates for'H-3 and C-14. In particular, non-fuel cycle H-3 inventory for the oight trenches of the present study is approximately a factor of 2 higher than either previous estimates of total site inventory. a

     'Modeling of release processes has been carried out in order to obtain
      'ostimates of isotopic-release rates from waste packages to the trenches.                             This modeling is highly speculative, but believed to be otate-of-the-art.                               It required information not only on amounts of the different isotopes, but also on the waste forms and' containers holding them.                   Such information was generally not given on the RSRs and had to be obtained by-contact'with the generators.                                     Estimated numerical release-rate data are given for each trench for H-3, C-14, Cs-137, 4      Cr-90, and Co-60.                               I-129 is expected rto have been totally released eithin a year of container breaching by corrosion. Most of the Pu, in the form of oxide, will probably not be released at'a significant rate.

NUREG/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT. KNIGHT,T.D. Los Alamos ! Ccientific Laboratory. March 1985. 440pp. 8503220009. LA-10166-MS. 29488:307. 1This report documents the Los Alamos result's of the second

4 assessment phase, independent assessment, for TRAC-PD2. We documented the results of the developmental assessment for TRAC-PD2 in an earlier l report. This report describes calculations run with the released

     ; versions of TRAC-PD2.                               We' analyzed separate-effects tests to investigate the critical-flow calculation, the emergency-core-cooling
     .(ECC) bypass behavior, 'and the reflood-tracking capability.                                                  We i

cnalyzed integral tests to explore the gravity-driven reflood behavior, and the small-break LOCA behavior. The results show good agreement between the calculated parameters and the data for those tests related to large-break LOCA. In general, the comparisons to cmall-br.eak LOCA tests indicated that the code can be useful but that come model improvements are required. I i ( NUREG/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR

     -HIGH-LEVEL WASTE PACKAGING.Guarterly Report, July-September 1984.

. CTAHL, D, s MILLER,N.E. Battelle Memorial Institute, Columbus Laboratories. January'1985. 117pp. 8502130361. 28914:098. During this reporting period, it was found that glass-water

j. contact during the nonisothermal periods of leach testing may influence the test results. Modeling of wasteform degradation focused

     .cn dissolution /reprecipitation kinetics.                                   An experiment is planned to verify this model. 'A procedure was developed to disperse rug (2) in a      MCC 76-68 glass.                               Potentfodynamic polarization tests were performed to

[ determine the effects'of single chemical ~ species in groundwater on the

cracking and pitting susceptibility of carbon steel. Slow strain rate tests show that carbon steel is especially susceptible to L otress-corrosion cracking'in aqueous FeCl(3) at low strain rates. The otrength of commercial high purity iron was found not to be affected

!- by hydrogens however, ductility was somewhat reduced. The description of groundwater radiolysis was further refined during this quarter. !. Integral experiments are being prepared to provide information on

33 .

t-

          .ccchinod-offects'ar9conoco that may influenco the 1cng-tcra performancelof the maste package.

J NUREG/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS i STEEL'FOR LWR SERVICE.Guarterly Report April-June 1984. ATTERIDGE,D.G.: BRUEMMER,S.M.s PAGE,R.E. Battelle Memorial Institute, Pacific Northwest Laboratories. February 1985. 42pp. B503220003. PNL-5181. 29487:176. The Division of Engineering Technology, U. S. Nuclear Regulatorg

. Commission, is sponsoring a program at Pacific Northwest Laboratory to evaluate welded and repair-welded stainless steel piping for light-water reactor service. . Stainless steels often become ,

sensitized, or less resistant to stress corrosion cracking (SCC), after undergoing heating and cooling cycles such as those encountered in welding. The weld heat-effected zone is often the site of crack initiation. This program will therefore measure and model the

development of a sensitized microstructure and its. resultant

          -resistance to SCC in welded and repair-welded stainless steel pipe.
,          The result will be a metho'd to assess the effects of welding variables

on the SCC susceptibility of component-specific nuclear

' reactor / repairs. NUREO/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN !. ICE-CONDENSER CONTAINMENT. CAMP,A.L.s'DEHR,V.L.s HASKIN,F.E. Sandia National Laboratories. January 1985. 214pp. 8503050511. SAND 83-0501. ~ 29264: 001. The MARCH and HECTR computer codes are used in this study to examine hydrogen production, transport, and combustion in an

          = ice-condenser containment for a camber of hypothesized severe accidents.       Both degraded-core and core-meltdown accidents are treated. The sensitivity of the containment pressure-temperature response is assessed f or a number of fac tors, including the hydrogen s           and steam source-term assumptions, ignition and propagation limits, combustion completeness, flame speed, spray operation, and recirculation fan operation.        The highest containment pressures occur I          for those cases where the ignitors are assumed to fail, the                                       ,

L recirculation fans or containment sprays are assumed to fail, or very l large steam and hydrogen releases accompanying vessel breach are predicted. NURE9/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS PRESSURIZER TEST YO5. PETERSON,A.C. Sandia National Laboratories. ! February 1985. 60pp. 8503040543. SAND 84-1534, 29218:281. !- The. TRAC independent assessment project is part of an overall ef fort to determine the capability of various system codes to predict the detailed thermal / hydraulic response of light water reactors during accident and off-normal conditions. The TRAC computer code is being assessed against test data from various integral and separate effects test facilities. As part of this assessment effort, a separate

l. effects component test performed in the-NEPTUNUS pressurizer test facility for Thermal Power Engineering at Delft University of j' Technology was analyzed with TRAC-PF1/ MOD 1. The test simulated insurges, combined with spray flow, and outsurges from a pressurizer, i and was selected for code assessment because the capability of the computer codes used in safety analyses to calculate the correct pressurizer response is an important concern of the NRC. The report

{ summarizes results showing somewhat higher pressures and fluid 34 :

tcaporoturco ecro colculcted during incurgon eith sprog fles then coro coooured in tho toot. A centributing factor to tho calculatien of high pressures and fluid temperatures appears to be that the interfacial heat transfer from superheated vapor to subcooled liquid tas too low. The calculational results for the base analysis and some

  >                           codeling studies are discussed.                                                    A TRAC-PF1/ MOD 1 input listing of the base case model is also provided.

NUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. MURPHY,0.A. Dal

                             ' Ridge National Laboratory.                                              CASADA.M.L.s JOHNSON,M.P.s et al.                                                   JBF Associates. January 1985.                                            114pp.        8502120062. ORNL/NOAC-224.

28872:137. This report describes the first phase of an NRC-sponsored project that identified and svaluated system interaction (SI) events that have occurred at commercial nuclear power plants in the United Ctates. The project included: an assessment of nuclear power plant operating experience data sourcess the development of search' methods cnd event selection criteria for identifying SI events review of possible SI events and final evaluation and categorization of ovents. The report outlines each of these steps and presents the-results of the project. The results include 235 events identified as cdverse system interactions and 23 categories into which those events core assigned. The categories represent groups of similar events and i include areas such as: adverse interactions between normal or offsite s power and emergency power systems; degradation of safety systems by , vapor or gas intrusions degradation of safety-related equipment by fire protection systemss and flooding of safety-related equipment through plant drain systems. After evaluating each category (and the ovents contained in them), the emphasis on the potential for continued problems in these areas should be examineds and current system interaction analyses methods should be studied to determine their offectiveness for identifying system interaction events. (Phase II of this project, " Evaluation of System Interaction Methods," will assess the effectiveness of current methods using the events identified in this report). NUREG/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. MURPHY,0.A. Oak Ridge National Laboratory. CASADA.M.L.s JOHNSON,M.P.s et al. JBF Associates. 4 January 1985. 265pp. 8502120069. ORNL/NOAC-224. 28871:002. ' See NUREQ/CR-3922,VO1 abstract. NUREG/CR-3936: RELAP5 ASSESSMENT: CONCLUSIONS AND USER GUIDELINES. l' KMETYK, L. N. Sandia National Laboratories. January 1985. 182pp. C502010096. SAND 84-1122. 28700:001. The RELAPS independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal / hydraulic response of LWRs during accident and off-normal i conditions. The RELAP5/ MOD 1 code has been assessed at Sandia against o variety of test data from both integral and separate effects test facilities. All these analyses have been documented in detail in individual topical reports: in this paper we attempt to evaluate the overall code performance by comparing results from many different calculations, and to offer other users some guidelines based on our oxperience to date. All results show that good primary side steady 35 i

i ototo initici ,cnd/or operating conditieno cro rocdily cbtainod, given cdaquoto facility doccriptiano end ocm2 unor osporienco or guidolinon,

 '_although problems are usually encountered in the steam 3enerator secondary sides.

i NUREG/CR-3943: THE DWR PLAN ANALYZER. WULFF,W.s CHENG H.S.s LEMACH, S. V. ; et al. Brookhaven National Laboratory. February 1985. 345pp. 8503120461. B NL-NUREG-51812. 29339:131. This final report describes the modeling, the software and the hardware of=the plar.t analgrer. The reporg,also presents,the first developmental assessment and contains a uswr guide for the plant analyzer. A large number of transients have been simulated. The simulation encompasses the neutron kinetics, the thermal conduction in fuel structures and the hydraulics of nonequilibrium, nonhomogeneous two phase flow in the nuclear steam suppig system, steam Line dynamics, turbines, condensers, feedwater trains and the suppression pool, as well as the control and plant protection systems. All simulations can be carried out at speeds up to 10 times faster than real-time process speeds. The technology presented-here has been developed primarily for cost-effective' safety analyses, but is also

 -invaluable for plant monitoring, failure diagnosis and computer-aided mitigation of accidents.

NUREG/CR-3945: FATIGUE CRACK OROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT. CULLEN,W.H. Materials Engineering Associates, Inc. February 1985. 65pp. 8502150048. MEA-2055. 28958:202. ~ Fatigue crack growth rates of A 106 Or. C and A 516 Gr. 70 carbon steels, and A 351-CF8A stainless steel in PWR environments have been determined over a load ratio range (R) of 0.2 to 0.85, a temperature range of 93 degrees centigrade to 338 degrees centigrade, and a test frequency range of 17 mHz to 1 Ha using sinusoidal waveforms. In addition, growth rates have been determined for various orientations of the crack plane with respect to the product form. Crack. growth rates in 288 degrees centigrade air environments have been measured in order to provide a reference baseline. These results define the magnitude of and major influences on the environmentally-assisted fatigue crack growth rates for these piping steels, and are supported by fractographic observations of the fatigue fracture surface. NURE9/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM OENERATDR TUBING PRDORAM. Semiannual Progress Report For Period Ending June 30,1984. DODD C.V.s DEEDS.W.E.s SMITH, J.' H. s et al. Dak Ridge National Laboratory. January 1985. 14pp. 8502210202. DRNL/TM-9339/V1. 29028:333. Eddy-current inspection is the most suitable method for rapid boreside evaluation of steam generator tubing. However, small flaws can be masked by the effects of harmless variables, such as tube supports. To, identify the critical properties accurately and reliably in the presence of extraneous signals caused by variations of unimportant properties, sufficient information is needed to identify harmful variations and to reject harmless ones. For this reason we

 ~are developing instrumentation capable of measuring both the amplitude and phase of the eddy-current signal at several different frequencies and computer equipment capable of processing that data quickly and reliably. Our probes and test conditions are miso computer-optimited.      The most recent probe design embodies an array of 36

         'Caa11 f1ct "penccko" collo cnd i provas the dotecticn of cmall f1cun cnd the roJoctian of tubo cupport cignolo.                                       Wi edepted our nou IBM Cystem 9000 computer to take and process the larger amcunts of data required by additional variables, such as copper coating and
         'intergranular attack.                             We also completed construction of the hand-wired versions of the 8- and 16-coil arrays and the multiplexing circuitry and computer codes to handle the data.

NURES/CR-3950 VO1: FUEL PERFORMANCE ANNUAL REPORT FOR 1993. BAILEY, W. J. s DUNENFELD M. S. Battelle Memorial Institute, Pacific , Northwest Laboratories. March 1985. 123pp. 8503290020. PNL-5210.

;        .29547:139.

i This annual report, the sixth in a series, provides a brief description of fuel performance during 1983 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel curve 111ance programs, fuel operating experience, fuel prublems, high-burning fuel experience, and items of general significance are provided. References'to additional, more detailed information and related NRC evavations are included. l NUREG/CR-3954: HECTR ANALYSIS OF EQUIPPENT TEMPERATURE RESPONSES TO EELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. DANDINI,V.J.s MCCULLDCH,W.H. Sandia National Laboratories. February i 1985. 135pp. 8503280013. SAND 84-1704. 29547:001. i .The temperature response of three generic surface models in each of three-locations in an ice condenser containment building were calculated assuming a hydrogen deflagration event and using the HECTR code. The intent'of using the three-generic surfaces was to

.         conservatively represent surfaces of various types of safety

) oguipment. Analyses were performed for four accident seguence types eith variations. The general observations drawn from these analyses l cre that (1) higher equipment surface temperatures than calculated for 4 Q(2)D type arrested seguences were calculated for other sequence types of comparable core melt frequency, and (2) surface temperatures greater than qualification temperatures were calculated to occur for come seguence types. 1 i NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON i DF ANALYSIS TECHNIQUES. FAYER M.J.s MCKEDN T.J. Battelle Memorial 4 Institute, Pacific Northwest Laboratories. December 1984. 91pp. 1 E503040052. PNL-5222. 29198:274. Two empirical methods of settlement analysis (Terraghi's theory

!         cnd a simplified version of the Fredlund-Morgenstern two-stress-state

cpproach) were compared,to the computer code TRUNC, a modified version of the TRUST code for variably saturated flow in deformable porous cedia. The three methods were used to predict settlement of a 12.2-m-deep pile of tailings slimes with a drain at the bottom. The i.. simpler, empirical methods of settlement analysis were just as offective as TRUNC in predicting total settlement. For saturated tailings, predictions of total settlement by Terzaghi's theory and TRUNC were in close agreement (1.69 and 1,73 m, respectively). For partially saturated tailings, the simplified stress-state approach and i TRUNC predicted similar total settlements (0.52 and 0.51 m, l respectively).~ Terraghi's theory, as applied, overestimated the time l

Of settlement under saturated conditions (170 days versus 140 days i predicted by TRUNC) because it did not account for gravitational gradients. No empirical or analytical means were available to predict , 37 i.

    -tho tima of cottlocent under pertially octurctod ccnditieno.        If the ccgnitudo of portfolly octurated cottlocent to cenoidored significent,
   .then the time over which it occurs will most likely be.the deciding factor in determining when to place the cover on the tailings pile.

NUREG/CR-3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES. MCGOWAN,J.J. Oak Ridge National Laboratory. January 1985. 25pp. 8503110098 ORNL/TM-9516. 29328:206. The Heavy Section Steel Technology (HSST) program office conducted a series of experiments to determine the effect of neutron irradiation on the fracture toughness of nuclear pressure vessel materials. One plate (H88T plate 02) and four welds of A533 grade B class 1 steel were examined. The welds were made by current (about 1979) . practice. As part of this study, tensile properties were measured after irradiation to 2 x 10(23) neutrons /m(2) (>1 MeV) at 288 degrees C. The strength of all four welds increased with irradiation. Yield strength was about 10% more sensitive to irradiation than was ultimate strength. Tensile ductility was not affected significantly by irradiation. NUREG/CR-3980 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH. PROGRAMS. Guarterly Progress Report, April-June 1984. REST,J. Argonne National Laboratory. February 1985. 36pp. 8503290274. ANL-84-61 VO2. 29563:354. This progress report summarizes the Argonne National Laboratory work performed during April, May, and June 1984 on water reactor safety problems related to fuel and fuel cladding materials. The research and development areas covered are Transient Fuel Response and Fission Product Release and Clad Properties for Code Verification. NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. MAHN,B.s TURGEDN,K.S.s MARTINI,D.K.s et al. Georgia Institute of Technology, Atlanta, GA. February 1985. 170pp. 8502210154. 29051:001. Bluegill and catfish maintained in flow-through tanks were fed P-32 at two feeding levels. Fish were analgred in triplicate for P-32 and phosphorus at intervals of 1 - 8 day s. Additional aguaria experiments were performed to determine the effects of other factors i and to observe P-32 uptake from water by unfed fish (including fish ! with blocked esophagus). The bluegill showed a weight gain of 0.2 i X/d, a phosphorus turnover constant in muscle of 0.43 X/d, and a l BF(r)/BF ratio of 0.081 at the higher feeding rate, and 0.05 X/d, 0.34 l X/d, and 0.064 at the lower feeding rate. Hence, respective P-32 l BF(r) values are 6,000 and 4,000 at a ph osphorus BF of 70,000. The j BF(r) values for catfish were approximately twice as high. The i aguarium experiments suggest that the higher factors are due to a much i higher phosphorus intake, higher water temperature, higher retention l from pellets than from worms, and possible higher retention by catfish

j. than bluegill under the same conditions.

I i NUREG/CR-3984: BIOLOGICAL-CHARACTERIZATION DF RADIATION EXPOSURE AND i DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress

Report, April 1983 - March 1984. EIDSON,A.F. Inhalation Toxicology i Research Institute. January 1985, 21pp. 8501230614. LMF-11, 28536:228.

38

L: Tho prgbloco cddroecod cro tho protacticn of urcnius cill usrkoro , from accupatianol onpocuro to urcniva thrsugh rcutino biccocay programs and the assessment of accidental worker exposures. Chemical properties of refined uranium ore (gellowcake) and uranium , distribution patterns among organs are compared. These studies will 4 facilitate calculations of organ doses for potential exposures and cill identify important' bioassay procedures. Results of studies in rats to. investigate retention of yellowcake in a wound showed that retention of less soluble yellowcake from the body was significantly core prolonged than of more soluble gellowcake. However, retention , could not be guantitatively related to the chemical composition or in l- vitro dissolution behavior of the implanted powder. Studies of Beagle dogs following nose-only inhalation of aerosols of commercial yellowcake were continued. Histological observations showed kidney damage that appeared 4 to 8 days after exposure to.the more soluble yellowcake with repair occurring by 64 days after exposure. The ; concentration of uranium in kidney was 8-17 mg U/g kidney at 4-8 days , ofter exposure. No evidence of kidney damage was observed in dogs oxposed to the less soluble yellowcake form. i NUREG/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EQUATIONS FOR 4 MULTIPHASE FLOW. Part One: System Without Internal Solid Structures. , CHA W.T.s CHAO,B.T.s SDO, S. L. Argonne National Laboratory. Februarg

1985. 127pp. 8503110328. ANL-84-66. 29328: 001.

A set of rigorously derived conservation eguations of masse

;         comentum, and energy for multiphase systems without internal solid otructures via time-volume averaging of poins, instantaneous

conservation equations is presented. These equations are differential-integral equations in which the area integrals account i for the interfacial transport of mass, momentum, and energy. The

. oguations from volume averaging contain averages of the product of the

 ,        dependent variables which must be expressed in terms of the products

' of their averages. In nonturbulent flows, this is achieved by oxpressing the " point" variables as the sum of its intrinsic volume , overage and a spatial deviation. In turbulent flows for which further time-averaging is required, the " point" variable is then decomposed

into a low-frequency component and a high-frequency component. Time i

overaging following volume averaging preserves the identity of the dynamic phases. Under certain simplifying conditions, the proposed cet of rigorous 1g derived conservation equations reduces closely to

various forms that are current 1g " accepted" for two-phase flow i cnalysis. This set of conservation equations serves as a reference
point for modeling multiphase flow and provides theoretical guidance l cnd physical insight that may be useful to develop correlations for i

quantifying interfacial transport of mass. momentum, and energy. l l NURE9/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN { LIGHT-WATER REACTORS. DEITZ,V.R. Navy, Dept. o f, Naval Research Lab. March 1985. 190pp. 8504040424. 5528. 29629:058. j Nuclear grade carbons were systematically degraded by exposure to

unfiltered outdoor mir with decrease in radioactive methyl iodide i trapping. Local meteorological conditions of high humidity combined eith atmospheric pollutants in the test vicinity contributed jointly

] to the degradation. When service carbons were exposed to radiation l 1evels of 10(7) to 10(9) reds, the iodine isotope exchange capacity can regenerated. The adsorptive properties were onig slightig improved. It was possible to regenerate the iodine isotope-exchange officiencies by reaction with airborn chemical reducing agents such as

i. 39

   . .-.                - . - _     -        --          __        -  - - - -       .-. _ - -                ___ -  -    - - -      _ _ =

l hydronina for ccrbcno rensvod feco nucloor pecor sporaticnn. The i depth profilo in cathyl iodido-131 psnotration chcnged frco si plo i exponentional through new carbons to a non-linear profile for i weathered and service aged carbons. The behavior is attributed to the l chromatographic distribution of the contaminants that accumulate in l i the bed. The removal of radioactive iodine depends on a minimum of 3 e distinguishable processes: adsorption on the activated carbon, iodine , isotope exchange with impregnated iodine-127, and chemical combination ! with impregnated tertiary amines when present. When a carbon is new, j all 3 mechanisms are at peak performance. After the carbon is placed j in service, the 3 mechanisms degrade at different ratess the ! i adsorption process degrades faster than the others. ( NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND PARTIAL LOSSES i OF OFF-SITE POWER AT NUCLEAR POWER PLANTS. BATTLE,R.E. Oak Ridge l National Laboratory. February 1985. 63pp. 8502290542. j ORNL/TM-9384. 29169:202. ! Events involving loss of off-site power hat have occurred at j nuclear power plants through 1983 are describe 1 and categorized as

complete or partial losses. The events were frentfried as i plant-centered or grid-related failures. In asdid, ion, the causes of

! the failures were classified as weather, human reror, design error, or i i hardware failure. The plant-centered failures were usually of shorter ! daration than the weather-related grid failurvs. For this reason, the { weather-related events were reviewed in dete41. Design features that may be important factors affecting off-site power system reliability j were tabulated for most of the operating nuclear power plants. The i tabulated information was provided to NRC for a statistical analysis to determine the importance of these design features for losses of off-site power. The frequency of losses of off-site power versus l duration were estimated for three time periods. The frequency of loss of off-site power was estimated to be 0.09/ reactor-year based on industry-wide data for the years 1959 through 1983. NUREQ/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. ! B ARNER, J. D. s FITZSIMMONS,D. Battelle Memorial Institute, Pacific l Northwest Laboratories. February 1985. 104pp. 8503120445. P NL-524 5, 29340:190. i In a program sponsored by the Fuel Systems Research Branch of the U. S. Nuclear Regulatory Commission, a series of sin electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory primarily to determine the susceptibility of irradiated pressurized-water reactor Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures ; ! attributable to PCMI occurred during the six tests. This report l describes the testing methods, testing apparatus, fuel rod diametral l strain measuring device, and test matrix. Test results are presented and discussed. NURE0/CR-4000: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. GILLEN,K.T.s CLOUGH, R. L. Sandia National I Laboratories. January 1985. 51pp. 8503010341. SAND 84-1948. i 29186:186. ' i # 1 1

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

1 In Erdct to outropolcto motorial cccalcrotod cging dato. . .COthed81agico cunt bo_dovolapod booed en cufficient undorotending of the processes leading to material degradation. One of the most important. mechanisms leading to chemical dose-rate effects in polymers involves the breakdown of intermediate hydroperoxide species. A general model for this mechanism is derived based on the underlying a chemical steps. The results lead to a general formalism for

     ' understanding dose rate and sequential aging effects when l~

hydroperoxide breakdown is important. We apply the model to combined radiation / temperature aging data for a PVC material and show that this data is consistent with the model and that model extrapolations are in Oncellent agreement with 12-year real-time aging results from an octual nuclear plant. This model'and other techniques discussed in this report can aid in the selection of appropriate accelerated aging i cathods and can also be used to compare and select materials for use in safety-related components. This will result in increased assurance that equipment qualification procedures are adeguate. 4 NUREG/CR-4020: LHMS: A COMPUTER PROGRAM FOR TRANSIENT, THREE-DIMENSIONAL MIXING GASES. TRAVIS J.R. Los Alamos Scientific Laboratory, j February 1985. 52pp. 8503210474. LA-10267-MS. 29479:114. j A numerical technique has been developed for calculating the full

three-dimensional time-dependent equations of motion with multiple
opecies transport. The method is a modified form of the Implicit Continuous-fluid Eulerian (ICE) technique to solve the governing equations for low Mach number flows where pressure waves and local

! variations in compression and expansion are not significant. Large j j density variations, due to thermal and species concentration f

;     gradients, are accounted for without the restrictions of the classical 2'

Boussinesg approximation. Example calculations of the EPRI/HEDL

standard problems verify the feasibility of using this -

i finite-difference technique for analyzing hydrogen transport and ! , cizing within LWR containments, i NUREG/CR-4023: FIELD PERFORMANCE ASSESSMENT OF SYNTHETIC LINERS FOR URANIUM TAILINGS POND.A Status Report. MITCHELL.D.H.s SPANNER,0.E. Battelle Memorial Institute. Pacific Northwest Laboratories. January 1985. 78pp. 8502040788. PNL-5005. 28716:096. This report presents the status of Pacific Northwest Laboratory's i program through the end of FY-83 assessing the performance of

,     ognthetic liners used in uranium tailings ponds.                           Synthetic liner failure mechanisms, impoundment design, installation, and inspection techniques are presented from information collected from consultants, i      0111 operators, and the synthetic liner industry.                             Progress is

) rcported on laboratory tests on acceler.sted aging of liners, physical j properties of aged materials, and nondestructive examination of seams. I NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) j FRUCHTER,'J. S. s COWAN C.E.s ROBERTSON.D.E.s et al. Battelle Memorial i Institute,-Pacific Northwest Laboratories. March 1985. 56pp. 8504030419. PNL-5299. 29605:123. For the past several years, data on radionuclide migration in ground water at a low-level disposal site were collected. Most of the radionuclides were removed in the disposal basin and trench by either ! precipitation or adsorption mechanisms. However, three radionuclides j (&O)Co, (106) Rue and (125)Sb showed somewhat greater than expected l cobility. The elements of these three isotopes were found to be in t 41 i 'i.,, - -- . - . . _ _ .

oithor cnisnic or nsnionic chcrgo-forso. Cc=plomon eith both notural and man-made organics were implicated in the increased mobility, particularly in the case of (60)Co. Characterization studies of the organic fraction were performed. Ruthenium-103, (60)Co, and (125)Sb were found to be associated with the higher molecular weights greater

                                            .than 1000.                                                Studies were also performed that proved the hypothesis that the adsorption behavior of (235)Np on soils of the site is dominated by adsorption on iron hydroxide.                                                    Finally, geochemical modeling of the chemical and charge form data showed the ground water to be in equilibrium with several solids that could be important in controlling the concentrations of trace elements and radionuclides.

NUREQ/CR-4036: STRUCTURAL GEOLOGY OF SOUTHEASTERN ILLINOIS AND VICINITY. NELSDN W.J.s LUMM,D.K. Illinois, State of. November 1984. 138p p. 8501190501. 28485:078. Southeastern Illinois is the locus of several converging fault systems of differing orientations. The specific goals of this project were to determine the relationship of these fault systems (i.e. the Fluorospar District, the Wabash Valley, Shawneetown and the Cottage Grove) to each'other and whether or not any of them have been active in the Gunternary. Detailed field mapping indicated that southern Illinois and adjacent areas are currently subject to a compressional stress field in which tne major axis is east to east-northeast. North-tending thrust faults of small magnitude may be forming in response to these stresses, but the major fault systems are neither active or likely to be reactivated in this modern stress field. i j NUREQ/CR-4039: QAMMA-RAY CHARACTERIZATION DF THE TWO-YEAR IRRADIATION j EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. MAERKER,R.E. Oak j Ridge National Laboratory. January 1985, 21pp. 8502220412. ORNL/TM-9440. 29073:230. { Average gamma-ray group fluence rates are calculated for each of 1 the three exposures in the two year metallurgical blind test , experiment at the DRR-Poolside Facility in Oak Ridge, thus completing the character 1 ration of the radiation field for this experiment, which j is intended to serve as an international metalurgical benchmark. Heating rates in the steel derived from these calculations varied from about 0.23 watts / gram in the simulated surveillance capsule to 1. 4 milliwatts/ gram at the three quarters depth location in the simulated pressure vessel capsule, with secondaries arising from non-fission i reactions in the core and ex-core steel contributing between

seventy-seven and ninety-three percent of the tttal. Contributions from photofission to fission foil activities are estimated to be less: than five percent of those previously calculated arising from ,

neutron-induced fission. , 3 i NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. LARSON,J R. EG&Q. Inc. January 1985. 75pp. 8502210404. EGG-2354. T.9057: 349. , This handbook provides simple procedures for calculating the behavior of light water reactors during a variety of incidents. It provides an additional tool for assessment of ongoing and postincident behavior. The handbook consists of a main body describing generic ! procedures, an appendix providing specific design data for a limited number of plants for application with the procedures, and an appendix

listing existing and planned BWR and PWR plants by containment types

,                                            and thermal-hydraulic parameters.                                                    The procedures are currently limited to break flow rate, decay heat power and integrated power, l

42

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

l' otcom scncrctisn frea doccv hoot, maos balcnco, shutdswn margin,

!                         noturol circulation, noncandonochlo 300 gensroticn, dono coticaton, i                          and DN8 evaluation void formation in the upper head, and torus heatup.

i j, NURE9/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW i AND CONTAIPMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. HUANO,C.s EVANS,D.D. Arizona, Univ. o f, Tucson, AZ. January 1985. 116pp. 4 8502210181. 29051:171. i A 3-dimensional fracture generating scheme is presented which can be used to simulate water flow a'nd containment (solute) transport j through fracture system of a rock. It is presently limited to water ,

saturated conditions, zero permeability for the rock matrix, and

{ steady state water flow, but allows for transient solute transport.

The scheme creates finite planar plates of uniform thickness which

{ represent fractures in 3-dimensional space. A given fracture (plate)

has the following descriptors: center location, orientation, shape, areal extent and aperture. Each parameter can be described by an appropriate probability distribution. Individual fractures are 4

generated to form an assemblage of a certain fracture density. All j fracture intersections and boundary / frac ture intersections are determined and desdend fractures are eliminated. Flow through the ~ fracture assemblage is considered laminar and described by Poiseuille's law. The principle of mass conservation at each ! intersection is used to develop.the global matrix equation, which is ! solved subject to specified boundary conditions to vield the head and j flow distribution at each intersection. Solute transport is j considered to be advoctive between intersect *ons with complete mixing i at each intersection. Solutes added to the flow system can be explicitiv followed and concentration vs. time relationships can be [ Some examples are included. 4 determined anywhere in the system. 1 i i NURE0/CR-4045: LITERATURE REVIEW ON AEROSOL-SAMPLING DEVICES FOR l RESPIRATDRY FIELD STUDIES. SUTCLIFFE C.R. Los Alamos Scientific Laboratory. February 1985. 68pp. 8502250844. LA-9977-MS. ! 29095:216. l As part of the first phase of a Respirator Field Performance , project for the Occupational Safety and Health Administration / Nuclear } Regulatory Commissione,a critical review of the literature available l on respirator protection studies was completed. Little information l was available on experimental conditions, and when the information was ). available, each study was different in how the aerosol measurements { were made and in which parameters were controlled. Under these l conditions, it is difficult to compare results obtained from different i investigators. The literature was also surveyed for characteristics } desirable in an aerosol-sampling inlet in order to representative 1g l sample respirable particles. Available ambient aerosol samplers were j critically reviewed for their performance characteristics. Recommendations are made to avoid the pitfalls present in many ! respirator field studies and to help standardize these studies. 1 1 ) NURE0/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING l ! EXTRAPOLATION TECHNIGUES. JARRELL.D.B. EG&G, Inc. March 1985. j 28p p. 8504030456. E00-2357, 29605:220. 4 This report presents the methodology and documentation of the { calibration of the Loss-of-Fluid Test (LDFT) power-operated relief and 4 safety relief valve (PORV + SRV) for the L9-3 anticipated transient I without scram (ATWS) experiment. A multiposition globe valve was 43 i l

                                                                            . - ~ .

l l colibretod to prcduco ocolod high-proocuro flew ratos uoing o Ice-pronouro calibraticn fccility cnd a simplo RELAP5 critical flow ! model to extrapolate the calibration data to expected operating l pressures. It was demonstrated that an accurate high pressure,

'           multiphase flow calibration can be performed without the necessity of actual high-pressure testing.           This technique, when applied to large pressurized water reactor (LPWR) safety and relief valves, represents              '

j a potentially large savings in the capacity qualification procedure of l full-scale pressure reduction valves. i NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF U02 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. MITCHELL,G.W.3 OTTINGER,C.A.s MEISTER,H. Sandia National Laboratories. January 1985. B4pp. 8503050012.

SAND 84-1144. 29246:001.

The LMFBR Debris Coolability Program at Bandia National Laboratories investigates the coolability of particle beds which may form following a severe accident involving core disassembly in a nuclear reactor. The D series experiments utilize fission heating of j fully enriched UO(2) particles submerged in sodium to realistically l simulate decay heating. The D10 experiment is the first in the series to study the effects of bottom cooling of the debris which could be

provided in an actual accident condition by structural materials onto

, which the debris might settle. Additionally, the D10 experiment was designed to achieve maximum temperatures in the debris approaching the ! melting point of U0(2). The experiment was successfully operated for i over 50 hours and investigated downward heat removal in a packed bed at specific powers of.O.16 to 0.58 W/g. Dryout in the debris was ! achieved at powers from 0.42 to 0.58 W/g. Channels were induced in i the bed and channeled bed dryout was achieved at t owers of 1. 06 to 1.77 W/g. Maximum temperatures in excess of 2500 degrees centigrade were attained. i l NUREG/CR-4056: PARTICULATE AND GAS RELEASE FROM LIGHT-WATER-REACTOR (LWR) FUEL RODS STORED IN INERT AND DRY AIR ATMOBPHERES. OLSEN.C.S. l. EG&G, Inc. January 1985. 24pp. 8501210003. EGO-2359, 28497:013.

;                 A testing program using eight commercial pressurized water j           reactor (PWR) and boiling water reactor (BWR) spent fuel rods was t

conducted to investigate their long-term stability under a variety of I possible dry storage conditions. The ob Jective of this project is to I I provide the Nuclear Regulatory Commission (NRC) with information to confirm or establish licensing positions for dry spent fuel rod storage with regard to long-term, low-temperature (<250 degrees

centigrade), spent fuel rod behavior during dry storage and i

radioactive contamination arising from spallation of cladding crud. [ The results ~of the analyses of the crud, fuel particulate, and gas i release from these eight fuel rods is presented, which includes weight ! change measurements, delayed neutron measurements, and isotopic ! analysis of smears used to assess the particulate release. Gas i analyses of the fuel rod capsule environments were made to determine the fission gas release, and flow tests were performed to determine i the extent of filter blockage from particle entrapment. l NUREG/CR-4057: RADIOLOGICAL ASSESSMENT OF THE TOWN OF EDGEMONT.

JACKSON,P.O.s THOMAS.V.W.s YOUNG,J.A. Battelle Memorial Institute, j Pacific Northwest Laboratories. January 1985. 183pp. 8502040625.

j PNL-5320. 28727:151. This document is the final report for radiological surveys 44 )

ccnductcd in the cromunity of Edgocsnt, Csuth D2 keto fcr the purpoco Sf 1secting rooidual.redienctivo catcrials frca the urcniun precocaing industry. It contains a discussion of the historical justification for the surveys, and a summary of activities during the survey, from September 1980 through November 1984. The survey protocols are presented and discussed. The results of several studies of relevance to the surveys are also included. The results of the survey are presented in tabular form. NUREO/CR-4061: LEACHATE PLUME MIGRATION DOWNORADIENT FROM URANIUM TAILING 8 DISPOSAL IN MINE STOPES. NELSON,R.W.s MCKEON.T.J.1 CONSERE,W. Battelle Memorial Institute, Pacific Northwest Laboratories. February 1985. 82pp. 8504050284. PNL-5318.

     '29673:228.

A method previous 1g developed at Pacific Northwest Laboratory has been simplified and extended to better evaluate the environmental conseguences of below-water-table disposal of uranium-mill tailings in mine stopes. The method described uses analytical expressions for the velocity potential and examines numerically the convective transport of tailings liquor and leachate through the aguifer and into a water supply well located downgradient from the mine stope. The overall dependence of the leachate plume size and shape on the hydrologic parameters and the tailings disposal geometry are presented in graphical form for use in preliminary assessments. The graphical results are also used to set up worst-case scenarios for' return of the teachate constituents to the biosphere via the pumped water supply well. The interactive computer models developed to evaluate such worst-case conditions are presented, discussed, and used to evaluate four typical situations. NURE0/CR-4067: BUMMARY OF BARRIER DEGRADATIDN EVENTS AND BMALL ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. SAILOR,V.L.s CDLBERT,J.J. Brookhaven National Laboratory. March 1985. 66pp. 8504020086. BNL-NUREG-51842. 29585:135. The esperience of U.S. commercial nuclear power plants with respect to small accidents and events involving the breach of any of the various barriers to radioactive material release is reviewed and brief summaries are given of selected events. This report is intended to provide background information for the NRC staff evaluation of the proposed NRC saf ety goals. Included are events that resulted in the breach of one or more barriers (fuel cladding failures, primary coolant leakage, compromise of containment integrity), or in unintentional release of radioactive materials. Also included are miscellaneous small accidents or failures not resulting in radioactive releases, but which had special safety implications. The 1979 TMI-2 accident is not included. The report does not attempt to evaluate the significance of the events as potential precursors of more severe accidents (such evaluations are the subject of other studies). Rough statistics are presented on the frequency of events defined above for the periode 1974-1982. It is noted that none of the events resulted in fatalities or injuries attributable to radiological causes. NURE0/CR-4068:

SUMMARY

OF HISTORICAL EXPERIENCE WITH RELEASES OF RADIOACTIVE MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE UNITED STATES. SAILDR,V.L.: COLBERT.J.J. Brookhaven National Laboratory. March 1985. 73pp. 8503280025. BNL-NUREG-51843. 29548:007. 46

Thio rcpert proconto o cummary of tho hiotcrical oxperienco csncerning r0100000 of rcdicactiva catoriolo frca U.S. camm2rcial nuclear power plants. The material was compiled specifically to i provide background inf ormation for the Nuclear Regulatory Commission (NRC) Staff Evaluation of the proposed NRC Safety Goals. The types of available data on radioactive emissions are identified, reviewed and summarized. The annual 50 year population radiation dose commitments for the annular regions between 2 and 80 km surrounding each plant resulting from the radioactive emissions are summarized for the period, 1975-1981. These doses are compared with the annual population dose commitments from natural background radiation for the same areas, and with the proposed NRC Societal Safety Goal. The question of independent verification of licensee data on emissions is i examined. l NURE9/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT SHEFFIELD, ILLINOIS. PICIULO,P.L.s i- SHEA,C.E.: BARLETTA,R.E. Brookhaven National Laboratory. March 1985. 54pp. 8503200236. BNL-NUREG-51844, 29470:255. Soil samples and field resistivity data were collected from an area adjacent to the Sheffield site. Specimens of Peoria Loess, Roxana Silt, Radnor Till, sand from the Toulon member, Hulick Till, j and shale from the Pennsylvania system were collected and analyred. Resisitivities of the soils are all greater than 2500 ohm-cm, indicating an environment which can be moderately corrosive to steel. Measurements of soil pH range from 6.2 to 8.6. Determination of the total acidity of the soils indicates an alkaline environment. The moisture content of the soils are representative of a wet site. The

ion content of the soils show high levels of calcium consistent with l the calcareous nature of the soils. Both the extractable and j exchangeable concentrations of calcium, magnesium, potassium, and sodium in the soils are reported. The content of the following soluble anions is also given: carbonate, bicarbonate, sulfate,

[ sulfide, and chloride. l l i NUREQ/CR-4070 V02: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS. Volume 2: Current Status of Biofouling Surveillance And Control Techniques. DALING,P.M.s JOHNSON,K.I. Battelle Memorial i Institute, Pacific Northwest Laboratories. March 1985. 68pp. i 8503270015. PNL-53OO. 29541:211. l This report describes the current status of techniques for detection and control of cooling-water system fouling by bivalve mollusks at nuclear power plants. The effectiveness of these 4 techniques is evaluated on the basis of information gathered from a , literature review and in interviews with nuclear power plant

personnel. Biofouling detection techniques examined in this report j include regular maintenance, in-service inspection, and testing, Generally, these methods have been inadequate for detecting biofouling. Recommendations for improving biofouling detection capabilities are presented. Biofouling prevention (or control)

{ methods that are examined in this report include intake screen systems, thermal treatmente preventive maintenance, chemical treatment 1 alternatives, and antifoulant coatings. Recommendations for improving biofouling control methods at operating nuclear power plants are ' I presented. Additional techniques that could be implemented at future

power plants or that require further research are also described.

46 1

  -,,-,y-       ,,,,--ne-w,x. , , , , -           ,.__c   _,     - , . , - . ,  ,,.,,-,-.,,_,.--,n-,.e,_,.m.,
                                                                                             ,                . - - . - - - . ,                ,c,e,   . - , , . - - - .   ,nn- - . , .     -c,

            . . _ .        . . ~ _       _.     .. _ _ _ _ . _ _ _ _         _ ____ ___ _ _ _ _ _ __

^ NURE3/CR-4072: TM E]TIMATION OF ATMOSPHERIC DISPERJION AT NUCLEAR

       'POMR PLANTS UTILIZING REAL; TIME ANEMOMETER OTATISTICQ, LI,W.W.s l       MERONEY,R.N.          Colorado State Univ., Ft. Collins, CO.                                    January 1985.

236pp. 3502010665. 23703:001. Dispersion and turbulence measurements were conducted in a 3 '

      ^oimulated atmospheric boundary lay er..                         Field experiments and wind tunnel results for the behavior of lateral plume dispersion are compared to three semi-emperical expressions based on Taylor's diffusion theory.            Agreement between the field data and laboratory p        ceasurements supports using wind tunnel results to simulate ctmespheric transport phenomena.                        Eulerian space-time correlations l'       cith streamwise separations were measured for all three velocity

components in the simulated boundary _ layer. Results were compared to i previous measurements which were performed under different flow
configurations. A universal shape of the Eulerian space-time i correlation seems to exist when presented in a normalised time coordinate. Turbulence measurements of fixed-point Eulerian velocity

, otatistics were employed to estimate the Lagrangien velocity 1 ctatistics through the Baldwin and Johnson approach. The approach was

.       codified to account for the uniform shear stress effect in a

! homogenous turbulent flow field. The estimated Lagrangian integral . time scale agrees with estimates inferred from dispersion measurements j uithin only a 20% error. Such agreement supports the methodology of

using real time anemometer statistics to predict the atmospheric
. turbulent dispersion near a nuclear reactor site.

, NURE0/CR-4073: RESULTS OF THE SEMISCALE MOD-25 BTEAM GENERATOR TURE RUPTURE TEST SERIES. LDOMIS,0.0. E060. Inc. January 1995. 75pp. i 1 8502060492. E00-2363. 29745:278. ,

i. A series of esperiments was conducted in a scaled model of a 4

pressurized water reactor (Semiscale Mod-25) to investigate steam ,

generator tube rupture system signature response and recovery

, techniques. .The tube rupture was assumed to occur during normal full

. power operatica C15.6 MPa (2262 psia) system pressures 37 M (67: i. degrees fahrenheit) core differential temperature 3. From the

! omperimental results, the characteristic system signature responses for a wide range of number of tubes ruptured and rupture locations [, have been examined. In addition, recovery techniques requiring

operator actions.were esamined. These recovery technigues included l l the use of pressuriser muniliary spray and internal heaters, steam l l generator feed and steam, primary feed and bleed, and safety

! injection. The effectiveness of using these technigues for primary l l_ oystem pressure and subcooling control is discussed. l I NURE0/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT CAS AND DRY AIR STORAGE ATMOSPHERES. OL8EN,C.B. E060. Inc. January - 1995. 35pp. 8502220293. E00-2364. 29073:105. . i A testing program using eight commercial pressurized water [ reactor and boiling water reactor spent fuel rods was conducted to l investigate their long-term stability under a variety of possible dry i storage conditions. The objective of this project was to provide the Nuclear Regulatory Commission with information to confirm or establish , { dry spent fuel storage licensing positions for long-term, ' low-temperature (<250 degrees centigrade) spent fuel rod behavior ! during dry storage and radioactive contamination arising from spallation of cladding crud. The results of a nondestructive osamination of eight fuel rodge which included color closed-circuit 47 ~ -- ._ - __ - _-

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

I tolevioicn vicual oxaminatigno, celse phetsgrcphy, dic:noianol j measurements, and neutron radiography, are presented. , c NURE9/CR-4082 V01: DEGRADED PIPING PROGRAM - PHASE II. Semiannual , ,! Report. March 1984 - September 1984. WILKDWSKI,0.M.s AHMAD. J. s } L BARNE C.R.s et al. Battelle Memorial Institute, Columbus ! Laboratories. January 1985, 118pp. 8501290617. BMI-2120. 28574:202. . The objective of the Degraded Piping Program - Phase II is to ! j develop simple engineering analyses to assess the fracture behavior of nuclear piping. Such analyses must give realistic estimates of actual

,              fracture events.           Hence this is an intensely integrated program

! involving laboratory material property evaluation, analytical 1 developments, and full-scale pipe fracture experiments to verify the j simple engineering analyses. Both advance fracture mechanics analyses l l ( i. e. , J/T), and limit-load analyses will be assessed. This is a ! j 3-year program which began in March, 1984. Conseguently, this first i semiannual report describes work in progress rather than completed ef9 orts. i 1 NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIOACTIVE WA8TE : DISPOSAL SITES AT BARNWELL,8C AND RICHLAND,WA. PICIULO P.L.s ! SHEA C.E.; BARLETTA R.E. Brookhaven National Laboratory. March j 1985. 62pp. 8503290285. BNL-NUREG-51846. 29564:029. ; i To evaluate the performance of a buried waste form or weste ! container, considerati,on must include the interaction of the package ! with the burial environment. This report presents the results of l i physical and chemical measurements of soils from two currentiv l

;              aperating commercial radioactive waste disposal sitess one at i              Barnwell, BC, and the other near Richland, WA.                                                           Soil samples believed to be representative of the soil that will contact the buried waste forms were collected and analyzed.                                               Resistivity data given for soils
,              from both sites indicate mildly corrosive environments.                                                           The soil i               acidity measurements show the Barnwell site to have acidic soil,                                                                                            ,

l whereas, the Richland site has soils ranging from acidic to near *

!              neutral in pH.           The moisture content and the son content of the soils                                                                             !

! from each site are presented. The entractable ion content of the ! soils is given for the following ions: calcium, magnesium, potassium, ' { sodium, carbonate, bicarbonate, sulfate, sulfide, and chloride. Additionally, the exchangeable cations were measured for the soils from the two sites. NUMEG/CR-4087: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION i 1 CHARACTERISTICS. FAYER M.J. Battelle Memorial I n s t i t.u t e, Pacific Northwest Laboratories. February 1985. 44pp. 8503010322. ;

<              PNL-5339,       29196:059.                                                                                                                                  l-I                     Esperiments were conducted on uranium mill tailings from the                                                                                          ,

tailings pile in Grand Junction, Colorado, to determine their i consolidation characteristics. Three materials (sand, send / slimes min,' slimes) were loaded under saturated conditions to determine their 1 saturated consolidation behavior. During a separate esperiment, samples of the slimes material were kept under a constant load while the pore pressure was increased to determine the partially saturated i ! consolidation behavior. Results of the saturated tests compared well l with published data. Sand consolidated the least, while slimes consolidated the most. As each material consolidated, the measured i hydraulic conductivity decreased in a linear fashion with respect to i 4s l

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

tho void rotte. Pcrtictly caturcted osporiocnto eith the 011000 indiccted that thero con litt10 cens311dotica co tho psro proosuro was increased progressively above 7 kPa. The small amount of censo11dation that did occur was only a fraction of the amount of caturated consolidation. Preliminary measurements between pore pressures of 0 and 7 kPa indicated that measurable consolidation could occur in this range of pore pressure, but only if there was no load. NURE0/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CONTROL TEC69fIGUES FOR URANIUM MILL TAILINGS PILES. ELMORE M.R.s HARTLEY,J.N. Battelle Memorial Institute, Pacific Northwest Laboratories. January 1935. eSpp. 3502150694. Pft.-5340. 20960:305. Seventeen chemical stab 111:ers, rated as the most promising of these tested in earlier laboratory studies, were applied to test plots en a uranium mill tailings pile at the American Nuclear Corporation-Gas Hills Project mill site in central Wyoming. The durability of these materials when exposed to actual site conditions tas evaluated over time. In addition, eight commercially available cindscreens were field tested. Test panels of the eight materials core constructed at the Wyoming site to compare their relative resistance to weathering. A second test was conducted near Pacific Northwest Laboratory to evaluate the effectiveness of the windscreens at reducing wind velocity. Results of the field tests on the chemical otabilisers and windscreens are presented in this report, along with offectiveness-versus-cost information. Direct comparison of these two dust control methods is difficult due to the dependence of each on site-specific factors. However, simplified model case studies were developed to assess the cost of chemical stabilisation versus eindscreen systems for a hypothetical, inactive tailings pile. NURE0/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1994. MILLER,R.L s SAUMANN,5.L.s DOERGE.D.H. United Nuclear Corp. (subs, of UNC Resources. Inc.). January 1985. 124pp. 3502010524, 28702:207. This document summarises work performed during the 1984 fiscal year for the Nuclear Regulatory Commission's Evaluation of Nuclear Facility Decommissioning Projects program. This report describes actual work performed during the reporting period and work planned for the future. Included as appendices to this report are drafts of the current data from the TMI-2 recovery efforts and Shippingport Atomic Power Station decommissioning. NURE0/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAME ENVIRONMENTS. SIMPSON, H. J. s TRIER, R. M. s HERCZEO, A. L. s et al. Columbia Univ., New Yorke NY. January 1985. 72pp. 8501290372. 20572:001. The concentrations of a number of radioisotopes of some elements (Pu, U, The Pa, Ace Ra, 81, Po, Pb, Cs, Sr. and K) were measured in a group of lakes that are dominated by 90(4)(2-) ion in their anionic composition. Only Pu and the Th show possible enhancement of solubility at high sulfate concentrations in some Saskatchewan lakes. 01though never to the entent as observed for alkaline lakes. Surface caters of Green Lake (meromictic) which are at saturation with respect to calcite have approximately the same radionuclide content as seawater. The deepwaters (anonic) show a marked increase in Th and Re. This indicates that these elements may be coupled with the redos cycle of Fe and Mn which under oxygenated conditions effectively 49

octuootcr Pu, The cnd Ro co Fo(-Mn) oxyhydroxidon. Another possibility for tho enhanced (22a)Ra in the deep water is by coprecipitation with CACO (3) in the surface water and subsequent ) dissolution in the deep water thereby releasing radium into the water. NURES/CR-4100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT f OF YELLOWCAKE EMISSIONS. LEPEL E.A,s THOMAS,V.W. Battelle Memorial ,

              ' Institute, Pacific Northwest Laboratories.                 February 1985.       35pp.

8503200121.' PNL-5350. 29470:306. An evaluation of current sampling and analysis methods used for monitoring yellowcake emissions from uranium mill exhausts was performed by Pacific Northwest Laboratory. The representatives of , once per guarter sampling was felt to be guestionable. A more representative sample would be obtained by a continuous sampling system. The analysis could be performed by relatively newer instrumental methods. Direct-spectrometric and isotopically excited x-ray fluorescence instrumental analysis methods were evaluated. Because of a redirection in funding, the evaluation was not completed in terms of identifying instrumental interferences and field testing of the chosen methods. However, in light of readily available technology, a preferred rsethod for sampling and analysis of yellowcake from uranium mitt exhausks is proposed. This method would sample the ; enhaust stacks continuously using.a continuous, automatic, isokinetic ' I stack sampler with deposition o8 the exhaust gas particulates onto t filter paper. The deposited particulates would then be analyred by a-rev fluorescence using'(57)Co as an excitation source. It is also ! recommended that a papet;-tape sampler that houses an isotopic excitation source and detector be interfaced to a continuous stack sampler. This system would require evaluation and field testing after development. NURE0/CR-4112 V01: INVESTIGATION DF CABLE AND CABLE SYSTEM FIRE TEST ! PARAMETERS. Task A: IEEE Flame Test.

Underwriters Laboratory, Inc. L January 30, 1985. 10Spp. 8502130459. US 75-1, 28874:250.

The flame test in the Institute of Electrical and Electronics Engineers (IEEE) Standard 383 was investigated. The investigation was to develop possible modifications in test equipment and' test procedure : that would increase the repeatability of results and provide l additional information useful in assessing cable system performance in , response to a real fire. Several fire esperiments were conducted [ varying different test parameters. Tne experimental data were analyzed and modifications of both test equipment and test procedure i ! were developed to increase repeatability. These modifications were: r 1 An enclosure for the sample, defining cable damages cable fastening i ! and the cable tray to be useds establishing tolerances for exhaust of the enclosures starting temperature of the ambient air cable saeples } tocation of the burner and the flow rates of fuel and air into the burner. Sugge'sted also, was to report the maximum flame height versus , I time and the rate of heat released versus time as additional ! l Anformation that would be useful in assessing cable system

performance.

HURE0/CR-4112 V02: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST

PARAMETERS. Task B: Firestop Test Method. e Underwriters Laboratory, f i Jnc. January 1985. 71pp. 8502130568. US 75-2. 28922:351. L

i An experimental investigation was conducted to provide data j concerning the effects that changes in pressure differential, fire i f l 50 L 1 _._ _ _ _ _ ..__ _ _ , . _ _ .._.u___.__ _ _ . _

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

_ . . - _ - . - - = . . - . . - i- y Ospecuro cnd oceplo canotruction hevo an firootzp porfstmanco whon j ouposed to a standard fire test. Fifty-one fire test experiments were conducted using pressure dif ferentials between -12 to +120 Pa, different sample constructions and two fire exposure conditions.

Findings were that small changes in pressure differential did'not have o significant effect on firestop materials that did not have cracks-or through openings to allow passage of gases during fire esposure. If

,              the materials allowed passage for gases through cracks or. holes, such co those left open after pulling a cable, changing the pressure differential affected the firestop performance. Also, it was demonstrated that changing the size of the openings size, location and-i               type of the penetrating items installed through the openings and OcVerity of fire exposure affected the performance of the firestop.

] i i NURE0/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPENIMENT L2-5). Final Comparison Report. SURTT,J.D. EO&G, Inc. January 1985. ! 229pp. 8502210258. E00-2369. 29058:109. , j LDFT Emperiment L2-5 was designated International Standard 3 Problem 13 by the Organization for Economic Cooperation and e D velopment. Comparisons between measurements from Esperiment L2-5 , } ecre made with calculations frot. 11 international participants using j five different computer codes. LOFT Experiment L2-5 simulated a double ended guillotine cold lep rupture of a primary coolant loop of ( o large pressurized water reactor, coupled with a less of offsite l p cwor. , i 3 l NURE0/CR-4117: FAULTING AND JOINTING IN AND NEAR BURFACE MINE8 OF :

;            COUTHWESTERN INDIANA. AULT,C.H.s HARPER D.s BMITH C.R.s et al.

l Indiana Geological Survey. January 1985. 39pp. 8502070553. I 23795:316. 1 This project was directed towards the characterization of:, (1) the known large faults in southern Indiana, i. e. , the Georgetown Fault

l in Floyd. County and the newly named Crandall Fault in Harrison Countys and (2) the small scale fractures endemic to southwestern Indiana.

l The Georgetown and Crandall Faults are normal faults that have a coximum vertical displacement of about 65 feet. They are i post-Valmeyeran and pre-Pleistocene in age and are probably the result ^ j of hinge-line deformation between the subsiding. Illinois Basin and the l Cincinnati Arch. In contrast, abundant small-scale faults and joints i are related to regional compressive 11thospheric stress or to l 00dimentologic processes that operated penecontemporaneously with l deposition of the rocks in which they are found. Structures related 4 to regiunal stress include small-scale thrust faults with

displacements of a few inches to a few feet and joints that are !: cidespread in mines and outcrops in rocks of Mississippian and '

L Pcnnsylvanian age. The Jointing and most of the small-scale thrust i fculting indicate that southern Indiana is affected by the ; i Midcontinent Stress Province in the northern part of the study area ! cnd'by another stress field in the southern part. An east-west ! beundary can be defined between the two stress fields, i l NURE0/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK CDOLING ' PONDS. POLICASTRO, A. J s WASTAG,M.s DUNN,W.E.s et al. Argonne N3tional Laboratory. March 1985. 410pp. 8504050372. ANL/ES-143. ' i 29671:054. A general treatment of ultimate heat sink (UHS) cooling pond thermal performance is proposed through the application of a s 1 51 l t

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

w y .,three-dimensional grid model. Validation of the model has been shown

(through comparisons of predictions with data from a field and 1~aboratory pond. The advantage of the model lies in its ability to '

determine the detailed character of the flow field whether it be one. two, or three dimensional. Existing models require a priori knowledge l of the character and dimensionality of the flow field in such ponds. l Application of the model to a prototype UHS pond revealed that the balance of physical mechanisms involved in the thermal hydraulics of these ponds is gu,ite different than for ponds used in normal cooling. The small, heavily-loaded, irregularly-shaped nature of the UHS pond

should, in many cases, lead to a vertically mixed pond with only a one-dimensional (longitudinal) variation in pond temperature.

1 1 NURE9/CR-4121: EFFdCTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE { CRACM GROWTH RATES IN LWR ENVIRONMENTS. CULLEN,W.H. Materials . Engineering Associates, Inc. MEMPPAINEN M. s HANNINEN,H.s et al. j Finland, Govt. o f. February 1995. 49pp. 8503040021. MEA-2053. . 29198;223. i Fatigue crack growth rate tests, at a load ratio of 0.2, have i been conducted on steels of low, medium and high sulfur contents (0.004X, 0.013% and 0.025%) in PWR water at both low and high flow l j rates. Crack growth rates show no dependence on flow rate, but are - stron31v. dependent on sulfur content, with a large proportion of -

environmental assistance for the highest sulfur contents. Tests of low and high sulfur content steels at a load ratio of 0.7 show i relatively little environmental assistance in either case. The

]' fractography of these specimens shows the usual brittle appearance for environmentally-assisted fatigue crack growth. In addition, the , l opposing fracture surfaces match perfectly, indicating that little or i no dissolution of the metal matrix has occurred, and there is very [ little plastic flow associated with the fatigue cracking process. The l

X-ray photoelectron emission examination of the fracture surface oxides shows that Fe8 and Fes(2) coesist in the oxide layer, t j' suggesting that the conditions within the crack enclave involved near-neutral pH and cathodic potentials. 1 NURE9/CR-4123: SEIBMIC FRAGILITY OF REINFORCED CONCRETE STRUCTURES AND . j , COMPONENTS FOR APPLICATION TO NUCLEAR FACILITIES. GERGELY,P. j Lawrence Livermore National Laboratory. March 1985. 107pp. r 1 0503290022; UCID-20164. 29547:260.

                           '\the railure and fragility analyses of reinforced concrete                                                                    ,

j structures and elements in nuclear reactor facilities within the L Seismic-Safety Margins Research Program (SGMRP) at the Lawrence Livermore National Laboratory are evaluated. Uncertainties in l

i material modeling, behavior of low sheer walls, and seismic risk l assessment for nedlinear response receive special attention. Problems

!                 with ductility-based spectral deemplification and prediction of the
!                  stiffness of reinforced concrete walls at low stress levels are i                  esamined.                It is recommended to use relatively low damping values in

! connection with ductility-based response reductions. The study of l[ static nonlinear force-deflection curves is advocated for better 4 < nonlinear dynamic response predictions. Several details of the seismic risk analysis of the Zion plant are also evaluated. I I NURE9/CR-4145: EARTHOUAME RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ! ANALYSIC AND RANMING. H! LEMAN,J.A.1 MNOPDFF,L.s MANN,N.R.s et al. t 29471:001.

Earth Technology Co. March 1995, 142pp. 8503200110, N

1

Five methods for estimating earthquake recurrence were ranked. The methods represent those used, or proposed, in nuclear power plant otudies through 1982 and include Log Linear Poisson, Extreme Value, C;mi-Markov, Bayesian, and Uniform Hazard Method. Ranking focused on rocurrence estimates for earthquake sources, excluding attenuation and cite response. Scores were assigned to each method for 12 criteria cuch as accuracy, use of geologic data, and subjective Judgment. Criteria scores were weighted by their importance and summed. Different scoring and weighting schemes were used to identify any ocnsitivities. To aid in scoring statistical criteria, methods were tasted on synthetic earthquake catalogs with known statistics, and natural catalogs were tested against theoretical magnitude distributions. The uniform Hazard Method scored high because, in principal, expert Judgement draws upon all seismologic knowledge. The Boyesian Method scored low because data requirements are severe for practical cases. The other methods were intermediate. These observations seem insensitive to scorer, scoring approach, or coighting scheme. The semi-Markov Method scores were sensitive to the coighting scheme. NUR EG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. KATO,W.Y.s WEINSTOCK E. V. s CAREW,J.F.s et al. Brookhaven National Laboratory. Fobruary 1985. 327pp. 8502260121. BNL-NUREG-51858. 29109:001. Brookhaven National Laboratory has conducted a study on the need cnd feasibility of an independent organization to investigate significant safety events for the Office for Analysis and Evaluation of Operational Data. USNRC. This is being carried out in response to o Congressional request to the NRC for such a study. The study consists of three parts: the need for an independent or2anization to investigate significant safety events, alternative organizations to ccnduct investigations, and legislative requirements. The dotermination of need was investigated by reviewing current NRC investigation practices, comparing aviation and nuclear industry practices, and interviewing a spectrum of representatives from the nuclear industry, the regulatory agency, and the public sector. The advantages and disadvantages of alternative independent organizations core studied, namely, an Office of Nuclear Safety headed by a director roporting to the Executive Director for Operations (EDO) of NRCs an Office of Nuclear Safety headed by a director reporting to the NRC Ccamissioners; a multi-member NTSB-type Nuclear Safety Board independent of the NRC. The costs associated with operating a Nuclear Safety Board were also included in the study. The legislative roquirements, both new authority and changes to the existing NRC logislative authority, were studied. These legislative requirements I core based upon the Edwards-Udall Bill H.R. 6390 introduced in the 96th Congress and study of the NRC Organization Act. NUREG/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT EXPEP.IENCE TO THE U.S. NUCLEAR REGULATORY PROCESS. ANDREWS,W.B. Botte11e Memorial Institute, Pacific Northwest Laboratories. Fobruary 1985. 163pp. 8503130128. PNL-5388. 29359:207. This report is a summary of applications of probabilistic safety casessment (PRA) in the United States and foreign countries. It is intended to stimulate discussion on the applicability of foreign oxperience to the United States, provide information on foreign safety tochnology development and focus the United States goals for future pcrticipation in the activities of the Committee for the Safety of Nuclear Installations (CSNI), Principle Working Group 5 (PWG5). 53

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

Results indicate that the United States leads the surveyed countries l in the completion and application of comprehensive PSAs of public safety impacts from nuclear power plants. European experience has i focused on the use of reliability analyses in support of design and i operational decisions. It is recommended that use of probabilistic analyses be expanded in the United States for engineering applications based on the success in European countries. NUREG/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME

SIMULATIONS OF DISPERSION. LEWELLEN,W.S.s SYKES,R.I. Acronautical Research Associates of Princeton. March 1985, 180pp. 0503200126.

[ ARAP 472. 29468:191. g This report provides an evaluation of several availeble dispersion models to determine their suitability for providing the " capability for estimating the effects of accidental discharges of radioactive material at nuclear power plants. A critique of the assumptions involved and a review of existing verification studies are _ made for models ranging from the Gaussian plume with straight line winds to models which attempt a complete solution of the primitive equations of motion. It is demonstrated that although even the simple models are capable of providing reasonably accurate predictions under ideal conditions, there are reasons to expect relatively severe limits i on plume predictability when certain emission conditions are combined with certain meteorological conditions. The usefulness of a real-time y dispersion model is thus likely to be dependent on a complementary i estimate of the variability expected about the mean dispersion for the conditions existing at that time. This report is one of a set of three dealing with real-time dispersion models. The other two deal with the uncertainties involved in the deposition module of dispersion [ models and the results of testing some of the dispersion models reviewed in this report by comparing them with the data collected at the Idaho National Engineering Laboratory in July, 1981 during an NRC

sponsored field test.

E p NUREQ/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. POLKY,J.N.s MILLER D.D. Sigma 4 Research, Inc. March 1985. 42pp. 8504030453. 29605:178. [ The Phase-I feasibility study of residue number system (RNS) g image processing for SAFT inspection has successfully determined that an advanced' inspection system may be built using a correlation-reconstruction SAFT algorithm, implemented with RNS techniques and off-the-shelf electronic components. Images are s reconstructed in a number theoretic transform domain with simple y pointwise multiplication of the A-scan data volume by a custom point i spread function (PSF), all in a highly parallel computational architecture. These methods also allow image enhancement to be easily j performed for improved flaw visualization, and with negligible speed H reduction. It has been determined that high resolution three k dimensional flaw images may be generated and that a commercially ? viable product could result through development of a prototype f real-time RNS processor. The hardware is expected to be made up of 100 nsec bit slice microprocessor components and large RAM storage units. Based on the performance estimates of the Phase-I effort, this new image processing system has the potential to acquire and focus the equivalent of the 145 A-scans per second, which translates into more than 1000 cubic inches per min. inspection rate for typical pressure vessel specimens. 54

NURE3/CR-4172: A USER'O QUIDE FOR MERGE. FREEMAN-KELLYs JUNG,R.O. Battelle Memorial Institute, Columbus Laboratories. March 1985. 41pp. 8504040006 BMI-2121. 29629:233. The MERGE code acts as the interface between the MARCH-2 code, thich is used to determine overall accident progression, and the TRAP-MELT code, which is used to evaluate reactor coolant system fission product transport and deposition. MERGE uses MARCH-calculated . core exit flows and temperatures to perform a detailed gos-to-structures heat transfer analysis for the control volumes in the flow path through the reactor coolant system and converts these i results into a form required as input to TRAP-MELT. MERGE can treat i up to nine control volumes, containing up to five structures each. ' Roguired inputs include descriptions of the control volumes and their : j flow connections, as well as initial conditions. NURE0/CR-4173: CORSOR USER 'S MANUAL. KUHLMAN,M.R.3 LEHMICKE,D.J. Battelle Memorial Institute, Columbus Laboratories. MEYER,R.O. NRC

       - No Detailed Affiliation Given.          March 1985.          58pp. 8504040423.

BMI-2122. 29618:312. The CORSOR code simulates the release of fission products and ctructural materials from a reactor core during the in-vessel period of a severe accident in a light water reactor. The code is a simple, ocpirically based treatment of release and does not treat detailed nochanisms for release from high temperature fuel. The first-order release rate coefficients for the species considered are presented, the input requirements of the code are described, and an example input 4 ond output stream is supplied in an appendix. NUREO/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. DAEMEN,J.J.K.s CREER,W.B.s ADISOMA,0. S. s et al. Arizona, Univ. o f, Tucson, AZ. March 1985. 384pp. 8504090456. 29739:193. This report describes experimental borehole plugging performance casessments performed, started, or planned during June 1983 - May 1984. Results are given from field flow tests on three cement plugs installed in vertical boreholes in basalt and on one nearly horizontal coment plug. The horizontal plug and one vertical plug seal very wall. Hydraulic conductivity of two vertical field plugs has been 3 volatively high. Remedial action is described. Laboratory simulations of dynamic loading of cement plugs show no detrimental i offects. Drying of cement plug, especially for months, at elevated tcaperatures, increases the hydraulic conductivity of the plugs coverely, and reduces their bond strength along the interface. Microscopic inspection, strength and flow tests identify the drilling-induced damaged zone in basalt. While such a damaged zone omists, and has typical features (e.g. fracture density, size, 1ccation, orientation), it is thin and not likely to be a preferential flowpath. Engineering characteristics tests on bentonite plugs, chemical analgsfs and swelling tests. Experimental designs are given for the study of s24= and of thermal effects on plug performance. Preliminary results are presented. Results are included from ongoing cceent push-out tests and swelling measurements. 55

Contractor Report Number index This index lists, in alphabetical order, the contractor-issued report codes for the NRC contractor reports in this compilation. Each contractor code is cross-referenced to the NUREGICR for the report and to the 10-digit NRC Document Control System accession number. SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER 5523 NUREG/CR-3990 EGG-2369 NUREC/CR-4115 ANL-04-35 VO3 NUREG/CR-3804 VO3 FEMA-51 NUREG-0981 RO1 ANL-24-61 VO2 NUREG/CR-3980 VO2 IEB-79-09 NUREG/CR-3791 ANL-24-66 NUREG/CR-3989 IEB-79-12 NUREG-0905 ANL/ES-143 NUREC/CR-4120 IEB-80-25 NUREC/CR-3794 ARAP 472 NUREG/CR-4157 LA-10166-MS NUREG/CR-3866 BMI-2120 NUREG/CR-4082 VO1 LA-10267-MS NUREG/CR-4020 BMI-2121 NUREG/CR-4172 LA-9977-MS NUREQ/CR-4045 BMI-2122 NUREG/CR-4173 LMF-11 NUREQ/CR-3984 BNL-NUREG-51454 NUREG/CR-2331 VO4 N2 MEA-2053 NUREG/CR-4121 BNL-NUREG-51494 NUREQ/CR-2482 VO7 MEA-2055 NUREQ/CR-3945 BNL-NUREG-31494 NUREG/CR-2482 VO6 ORNL/CSD/TM-216 NUREG/CR-3723 BNL-NUREG-51609 NUREG/CR-3026 ORNL/NOAC-224 NUREQ/CR-3922 VO1 BNL-NUREG-51713 NUREG/CR-3498 ORNL/NDAC-224 NUREQ/CR-3922 VO2 BNL-NUREG-51717 NUREG/CR-3519 ORNL/NSIC-2OO NUREQ/CR-2OOO VO3N12 l BNL-NUREG-51784 NUREG/CR-3829 ORNL/NSIC-2OO NUREQ/CR-2OOO VO4 N1 BNL-NUREG-51792 NUREG/CR-3865 ORNL/NSIC-2OO NUREG/CR-2OOO VO4 N2 BNL-NUREG-51812 NUREG/CR-3943 ORNL/TM-9150 NUREG/CR-3738 BNL-NUREG-31842 NUREG/CR-4067 ORNL/TM-9154/V2 NUREG/CR-3744 VO2 BNL-NUREG-51843 NUREG/CR-4068 ORNL/TM-9163 NUREQ/CR-3764 BNL-NUREG-51844 NUREG/CR-4069 ORNL/TM-9191/V3 NUREG/CR-3851 VO3 BNL-NUREG-51846 NUREG/CR-4083 ORNL/TM-9216 NUREC/CR-3831 BNL-NUREG-51858 NUREQ/CR-4152 ORNL/TM-9339/V1 NUREG/CR-3949 VO! EGO-2251 NUREG/CR-3237 ORNL/TM-9384 NUREQ/CR-3992 EGO-2308 NUREG/CR-3767 ORNL/TM-9440 NUREQ/CR-4039 EQQ-2354 NUREQ/CR-4041 ORNL/TM-9516 NUREQ/CR-3978 EGG-2357 NUREG/CR-4046 PARAMETER IE-13 NUREQ/CR-3794 ' EGO-2359 NUREG/CR-4056 PNL-4221 NUREQ/CR-2850 VO3 EGO-2363 NUREQ/CR-4073 PNL-4973 NUREG/CR-3659 ,EQQ-2364 NUREG/CR-4074 PNL-5005 NUREQ/CR-4023 57 l

SECONDARY SECONDARY REPORT REPORT REPORT REPORT NUMBER NUMBER NUMBER NUMBER ,

                                                               ==     --__

J PNL-5069 NUREQ/CR-3752 SAND 84-1013 NUREQ/CR-3802 PNL-5106-3 NUREQ/CR-3810 VO3 SAND 84-1072 NUREQ/CR-3816 VO1 PNL-5121 NUREQ/CR-3817 SAND 84-1122 NUREQ/CR-3936 PNL-5125 NUREQ/CR-3825 VO3-4 SAND 84-1144 NUREQ/CR-4055 PNL-5181 NUREQ/CR-3911 VO2 SAND 84-1534 NUREQ/CR-3919 PNL-5210 NUREQ/CR-3950 VO1 SAND 84-1704 NUREQ/CR-3954 PNL-5222 NUREQ/CR-3972 SAND 84-1948 NUREQ/CR-4008 PNL-5245 NUREQ/CR-3999 SAND 84-7115 NUREQ/CR-3688 VO1 PNL-5299 NUREQ/CR-4030 SAND 84-7115 NUREQ/CR-3688 VO2 2 PNL-53OO NUREQ/CR-4070 VO2 UCID-19988 NUREQ/CR-3660 VO3 PNL-5318 NUREQ/CR-4061 UCID-20164 NUREQ/CR-4123 PNL-5319 NUREQ/CR-3709 UCRL-53044 NUREQ/CR-3019 PNL-5320 NUREQ/CR-4057 UCRL-53500 VO1 NUREQ/CR-3663 VO1 PNL-5339 NUREQ/CR-4087 UCRL-53544 NUREQ/CR-3854 PNL-5340 NUREQ/CR-4089 US 75-1 NUREQ/CR-4112 VO1 PNL-5350 NUREQ/CR-4100 US 75-2 NUREQ/CR-4112,VO2 PNL-5388 NUREQ/CR-4153 SAND 83-0501 NUREQ/CR-3912 SAND 83-1326 NUREQ/CR-3361 SAND 84-0884 NUREQ/CR-3772 1 i 1 N l i

Personal Author index l This index lists the personal authors of NRC staff and contractor reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by that author. If further information is needed, refer to the main citation by the NUREG number. ABEL,K.H. 3 NUREQ/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) ACKERMANN,G.R. NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MESODIF Model Calculations. ADAMS,R.E. NUREQ/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual Progress Report For April 1984-September 1984. ADISONA,G.S. NUREC/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual ReporteJune 1983 - May 1984. AHMAD,J. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984. AKGUN,H. NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. ANDERSON, R. F. NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. ANDREWS,W.B. NUREG/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT EXPERIENCE TO THE U.S. NUCLEAR REGULATORY PROCESS. ARNOLD,W.D. NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY l SITE PROJECTS. Report For April-June 1984. [ ATTERIDGE,D.G. NUREQ/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report April-June 1984. l AULT,C.H. NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF COUTHWESTERN INDIANA. l AYERS,A.L. NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. BADALAMENTE,R. NUREC/CR-3817: DEVELOPMENT, USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations. l BAILEY,W.J. NUREG/CR-3950 VO1: FUEL PERFORMANCE ANNUAL REPORT FOR 1983. ! 59

BAKER,D.A.: NURE9/CR-2850 VO3: POPULATION DOSE COMMITMENTO DUE TO RADIOACTIVE
RELEASES FROM NUCLEAR POER PLANT SITES IN 1981.

BALL,D.G. l NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS.

        -BANON,H.

NUREG/CR-3660 VO3: PROBAB.ILITY OF PIPE FAILURE IN THE REACTOR COOLANT ; LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirect 1g Induced By Earthquakes. NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS.DF COMBUSTION ENGIEERING PWR PLANTE. Volume 3: Double Ended ? Guillotine Break Indirect 1g Induced By Earthquakes. BARLETTA,R.E. NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH ' TECHNICAL POSITION ON WASTE FORMS AND CONTAIER MATERI ALS. NUREG/CR 3865: EVALUATION OF THE RADIDACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHJFFIELD LOW-LEVEL WASTE BURIAL SITE. NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL

  ~7            RADIDACTIVE WASTE DISPOSAL SITE AT SEFFIELD, ILLINOIS.

I NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIDACTIVE WASTE

     ]          DISPOSAL SITES AT BARNWELL,SC AND RICHLAND,WA.

BARNER,J.O. NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING l INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. BARNES,V.E. l NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN !. NUCLEAR POWER PLANTS. Problems And Recommendations. BARNS,C.R.

           . NUREG/CR-40tsJ VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984.

BASS,B.R. NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. BATTLE,R.E. NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND' PARTIAL LOSSES OF OFF-SITE POWER AT NUCLEAR POWER PLANTS. BAUMANN,B.L. NUREQ/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1984. BEHR,V.L. NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENBER CONTAINMENT. BENNETT,J.J. ! NUREG/CR-3516: A SURVEY OF THE USES OF RADIDACTIVE MATERIALS IN ! LOUISIANA'S OFFSHORE WATERS. , BLACKMAN,H.S. NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED DISPLAYS. BORK0WSKI,R.J. NUREG/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Report - Diesel Generators, Batteries, Chargers And. Inverters. I BOWERMAN, B. S. I NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE I BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. BROEK,D. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual l Report, March 1984 - September 1984. l 60

i r BROOK 2,B.O. l NUREG-0713 VO5: OCCUPATIONAL RADIATION EXPOSURE AT COMMERCIAL NUCLEAR POWER REACTORS - 1983 ANNUAL REPORT. BRUEMMER,S.M. NUREG/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report, April-June 1984. I BRYSON,J.W. I NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS.

    -BURTT,J.D.

NUREC/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT L2-5). Final Comparison Report. BUXTON,L.D.

.        NUREG/CR-3802: RELAP5 ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS.

CAMP,A.L. NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN

           . ICE-CONDENSER CONTAINMENT.

CAMPBELL,R.D. NUREC/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT : l LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly Induced By Earthquakes. : NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGIPEERING PWR PLANTS, Volume 3: Double Ended Guillotine Break Indirect 1g Induced By Earthquakes. , CAREW,J.F. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION.

CARHART,R.A.

NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PONDS. CASADA,M.L. NUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B.

        - NUREG/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND, SOURCES. Appendices C And D.

CATE,J.H. NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of l . Trajectories, Concentration Patterns And MESODIF Model Calculations. CAZZOLI,E.G.

NUREG/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUSASSEMBLY HEAT l ' TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS.

l CERBONE,R.J. l .NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. i CHAO,B.T. NUREG/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EQUATIONS FOR MULTIPHASE FLOW. Part One: System Without Internal Solid Structures. CHENG,H.S. l- NUR EG/CR-:.'943: THE BWR PLAN ANALYZER. CHEVERTON, R.'.'). l NUREG/CR-3',*23: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE F'.AWS IN PRESSURE VESSELS. CHOU,C.M. NUREG/CR-366J VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summary Report. < CLIFF,W.C. NUREQ/CR-3659: 4 MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES OF INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. CLOUGH,R.L. NUREQ/CR-4006: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. 61 1-

COMEN,L. NUREG-0837. VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1984. COLBERT,J.J. NUREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL

              . ACCIDENTS IN U.S.- COMMERCIAL NUCLEAR POWER PLANTS.

NUREG/CR-4068:

SUMMARY

-OF HISTORICAL EXPERIENCE WITH RELEASES OF RADIOACTIVE ' MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE >

              ; UNITED STATES.

COLMAR,R. NUREG-0933 SO2: A PRIORITIZATION OF GENERIC SAFETY ISSUES. COMER,M.K. NUREG/CR-36BB VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT

               -JUDOMENT. Volume 1: Main Report.

NUREQ/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDOMENT. Volume 2: Appendices. CONBERE,W. NUREG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. COWAN,C.E. NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final. Report)

       .CULLEN,W.H.

NUREG/CR-3945: FATIGUE CRACK GROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT. NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE

              . CRACK' GROWTH RATES IN LWR ENVIRONMENTS.

DAEMEN,J.J.K. NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE. ' PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. DALING,P.M. NUREC/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS. Volume 2: Current Status of Biofouling -Surveillance And Control Techniques. DANDINI,V.J. NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. DAVIS,R.E. NUREQ/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. DEAN,R.S. NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF QE TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. DEBEVEC,C.J. NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS. DEEDS,W.E. NUREQ/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. DEITZ,V.R. NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN LIGHT-WATER REACTORS. DEUTSCH,W.J. NUREG/CR-3709: ~ METHODS OF MINIMIZING GROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. DICKSON,C.R. NUREG/CR-3408 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MESODIF Model Calculations. DODD,C.V.

           ' NUREG/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING

_ PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. 62

DOERCE, D. H. NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. I Annbal Summary Report - Fiscal Year 1984. DOUGHERTY,D.R. NUREQ/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. DRAKE.J.D. NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE CDEFFICIENTS FOR

            ' SURFACE FLAWS IN PRESSURE VESSELS.                                        l DUNENFELD,M.S.                                                                       i NUREG/CR-3950 VO1: FUEL PERFORMANCE ANNUAL REPDRT FOR 1983.

DUNKERLY,S.J. NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. DUNN W.E. NUREC/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING P ONDS. EARY,L.E. NUREG/CR-3709: METHODS OF MINIMIZING QROUND-WATER CONTAMINATION FROM IN 4 SITU LEACH URANIUM MINING. Final Report. EDLER, S. K. NUREC/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS. Guarterly Report, July-September 1984. EIDSON,A.F. NUREC/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984. EL-SHINAWY,R.M.

;         NUREG/CR-3981: RIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH.

ELMORE M. R. NUR EG/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CONTROL TECHNIQUES FOR URANIUM MILL TAILINGS PILES. EMRIT,R.

    - NUREG-0933 SO2: A PRIORITIZATION OF GENERIC SAFETY ISSUES.

EVANS,D.D. NUREC/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. FAYER M.J. NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON , OF ANALYSIS TECHNIGUES. NUREG/CR-4087: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION CHARACTERISTICS. FISCHER,L.E. NUR EC/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS. FITZSIMMONS.D. NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING l INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. FDLEY,W.J. NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. NUREG/CR-3794: CLOSEOUT OF IE BULLETIN 80-25: OPERATING PROBLEMS WITH TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. FRAGOLA.J. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. FREEMAN-KELLY NUREG/CR-4172: A USER'S QUIDE FOR MERGE. FRUCHTER,J.S. NUREC/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) FUENK AJORN, K. NUR EC/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE 63

2. I PLUG PERFORMANCE. Annual Rsport, June 1933 - M::q 1984. FULLER,L.C. NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT l SIMULATION CODE. GADDY,C.D. , NUREC/CR-3688 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREG/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT, Volume 2: Appendiees. GAVIN,P. NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PONDS. GERGELY,P. NUREG/CR-4123: SEISMIC FRAGILITY OF REINFORCED CONCRETE STRUCTURES AND COMPONENTS FOR APPLICATION TO NUCLEAR FACILITIES. GILLEN K.T. NUREG/CR-4008: GEPERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. GILMORE,W.E. NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED DISPLAYS. GIRVIN,D.C. [ NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) GREER,W.B.

;            NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984.

GUPPY,J.G. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. HALL,R.E. ! NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. ! HAMMOND,R.A. NUREC/CR-3981: BI0 ACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. HANNINEN,H. NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. HARPER,D. NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. HARRINGTON,R.M. NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. HARTLEY,J.N. I NUREG/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CONTROL TEC W IGUES FOR URANIUM MILL TAILINGS PILES. HASKIN,F.E. NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. HENNICK.A. I NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2

              ' CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS.

, NUREG/CR-3794: CLOSEOUT OF IE BULLETIN 80-25: OPERATING PROBLEMS WITH TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. I HERCZEG,A.L. 4 NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. HESSON,G.M. NUREG/CR-3659: A MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES OF INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. 64 I

;  HILEMAN,J.A.

NUREG/CR-0105: EARTHGUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. HOFFMAN,F.O. I NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review l Of Data For Technetium. HOLLAND,R.A. NUREG-0905: CLOSEDUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS. HOLMAN,0.S. NUREG/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summary Report. HOOK.S.E. NUREG/CR-3516: A SURVEY OF THE USES OF RADIOACTIVE MATERIALS IN LOUISIANA'S OFFSHORE WATERS. HOPENFELD,J. NUREG-1108: RADIOACTIVITY TRANSPORT FOLLOWING STEAM QENERATOR TUBE RUPTURE. HUANG,C. NUREC/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. HUTTON,P.H. NUREQ/CR-3825 VO3-4: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 - September 1984. Volumes 3 and 4. ISAACSON,L.

NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS.

JACKSON,P.O. NUREC/CR-4057: RADIOLOGICAL ASSESSMENT OF THE TOWN OF EDGEMONT. JACOBS C. K. NUREG/CR-3851 VO3: PROGRESS IN EVALUATIDN DF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. JANO,J. NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITDRING NETWORK. Progress Report, July-September 1984. JARRELL D.B. NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING EXTRAPOLATIDN TECHNIGUES. JENNE.E.A. NUREC/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) JOHNSON,K.I. NUREG/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER ( SYSTEMS. Volume 2: Current Status of Biofouling Surveillance And Control Techniques. JOHNSON,M.P. NUREC/CR-3722 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. NUREC/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS l AND SOURCES! Appendices C And D. I JUNG,R.G. NUREQ/CR-4172: A USER'S GUIDE FOR MERGE. KAHL,W.K. 3 NUREG/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Report - Diesel Generators. Batteries, Chargers And Inverters. KAHN.B. NUR EG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. KATO,W.Y. l NUREC/CR-3026: FEASIBILITY STUDY ON THE ACGUISTION OF LICENSEE EVENT 66

1 I I DATA. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. KELMERS,A.D. NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL . l INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. l KEMPF,C.R. j NUREG/CR-3865: EVALUATION OF THE RADIOACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE.

,     KEMPP AI'EN, M.

NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATICUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. KENNEDY, R. P. NUREG/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirect 1g Induced By Earthquakes. NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGIPEERING PWR PLANTS. Volume 3: Double Ended Guillotine Break Indirect 1g Induced By Earthquakes. NUREG/CR-3805 VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task j II: Effects Of Ground Motion Characteristics O'n Structural Response Considering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. KHATIB-RAHBAR NUREG/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUBASSEMBLY HEAT l TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. KINCAID,R.H. NUREC/CR-3805 VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task II: Effects Of Ground Motion Characteristics On Structural Response Considering Localized Structural Nonlinearities And Soil-Structure ! Interaction Effects. MPETYK L. N. NUREG/CR-3802: RELAP5 ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS.

NUREC/CR-3936: RELAPS ASSESSMENT: CONCLUSIONS AND USER GUIDELINES.

KNIGHT,T.D. NUREG/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT. KNOPDFF,L. NUREQ/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. KOUSARI,B. NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOHEHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. KRAMARIC,M. NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-Septomber 1984. KRAMER,G. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984. l KUHLMAN,M.R. NUREC/CR-4173: CORSOR USER 'S MANUAL. KURTZ,R.J. NUREC/CR-3825 VO3-4: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE . MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 - ! September 1984. Volumes 3 and 4. LAI,W. NUREG/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS. LANDOW,M. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual 66

 ;           -Ropert,M rch 1984 - S3ptoribor 1984.

LARSON,J.R.

        . NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK.

. LEE,S.Y. NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL 4 INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE-PROJECTS. Report For April-June 1984. LEHMICKE,D.J. NUREG/CR-4173: .CDRSOR USER'S MANUAL. ! LEKACH,S.V. 4 NUREG/CR-3943: THE BWR PLAN ANALYZER.

;     LEPEL,E.A.

NUREG/CR-4100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT 4 0F YELLOWCAKE EMISSIONS. LEWELLEN,W.S. NUREG/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION.

   'LEYLAK,J.

i NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PDNDS. LI,W.W. NUREG/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR POWER PLANTS UTILIZING REAL TIME ANEMDMETER STATISTICS. LO, T.- NUREG/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summery Report. LOOMIS,G.G. NUREG/CR-4073: RESULTS OF THE SEMISCALE MOD-2B STEAM GENERATOR TUBE RUPTURE TEST SERIES. LOYDLA,V.M. NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN QENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. l LUCKAS,W.J. NUREG/CR-3026: FEASIBILITY STUDY DN THE ACQUISTION OF LICENSEE EVENT DATA. i LUMM,D.K. NUREG/CR-4036: STRUCTURAL GEOLOGY OF SOUTHEASTERN ILLINOIS AND , VICINITY. MACKENZIE,D.R.

       . NUREG/CR-3865: EVALUATION OF THE RADIOACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM.THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE.

MAERKER,R.E. ' -NUREG/CR-4039: QAMMA-RAY CHARACTERIZATION OF THE TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. l' MALLEN,A.N. NUREG/CR-3943: THE BWR PLAN ANALYZER. f MANN,N.R. i. NUREG/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. i MARSCHALL,C.W. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report March 1984 - September 1984. MARTIN,W.J. NUREQ/CR-3709: METHODS OF MINIMIZING GROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. MARTINI,D.K. l NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. MATHEWS,P. NUREG-0933 SO2: A PRIORITIZATION OF GEPERIC SAFETY ISSUES. 67

MAXEY,W. NURES/CR-d:082 VO1: DECRADED PIPING PROGRAM - PHASE II. S:naiennual Report, March 1984 - September 1984. l l MCCLUNO,R.W. NUREG/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. MCCULLAGH,C.M. , NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. MCCULLOCH,W.H. NURE0/CR-3954: HECTR ANALYSIS OF EQUIPPEENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. MCGOWAN,J.J. NUREG/CR-3'778: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES. MCGUIRE,R.K. NUREG/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS:

ANALYSIS AND RANKING.

MCKEON,T.J. NURE0/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON OF ANALYSIS TECHNIGUES. I NURE0/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. MEISTER,H. NURE0/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. MERONEY,R.N. NURE9/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR POWER PLANTS UTILIZING REAL TIME ANEMOMETER STATISTICS. MEYER,R.E. NUREO/CR-3851 VO3: PROGRESS IN EVALUATION DF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY j SITE PROJECTS. Report For April-June 1984. MEYER,R.O. NUREG/CR-4173: CORSOR USER 'S MANUAL. MILLER,D.D. NUREG/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. MILLER N.E. i NURE9/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.Guarterly Report, July-September 1984.

i. MILLER,R.L.

NURE0/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1984. j MILLS W.R. NUREG/CR-3791: CLOSEDUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. MILSTEAD,W. NUREG-0933 S02: A PRIORITIZATION OF GENERIC SAFETY ISSUES. MINNERS,W. . NUREG-0933 S02: A PRIORITIZATION OF QEFERIC SAFETY ISSUES. :

MITCHELL,D.H.

NURE9/CR-4023: FIELD PERFORMANCE ASSESSMENT OF SYNTHETIC LINERS FOR . URANIUM TAILINGS POND.A Status Report. MITCHELL,G.W. NURE0/CR-4055: THE D10 EXPERIMENT:COOLABILITY OF UD2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. MIZNER,A.A. NURE0/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. I MONROE,R.E. { NURE9/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION 68

OF SHIPPING CONTAINERS FOR RAD 7OACTIVE MATERIALS.

   -MORGENSTERN,M.

NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations. MU>LHEIM, M. D. NUREG/CR-3922 'VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. NUREG/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And'D. MURPHY,G.A. ' i NUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS

AND SOURCES. Main Report And Appendices A And B.

NUREG/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. NAKAGAKI,M. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984. NELSON,R.W. NUREG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. I NELSON,W.J. NUREG/CR-4036: STRUCTURAL GEOLOGY OF SOUTHEASTERN ILLINOIS AND VICINITY. NUMARI,N.H. NUREG/CR-3408 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories. Concentration Patterns And MESODIF Model Calculations. O 'BRIEN, J. N. NUREG/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT REPORTS.

  -OLSEN,C.S.

NUREG/CR-4056: PARTICULATE AND GAS RELEASE FROM LIGHT-WATER-REACTOR (LWR) FUEL RODS STORED IN INERT AND DRY AIR ATMOSPHERES. NUREG/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT I GAS AND DRY AIR STORAGE ATMOSPHERES. OTTINGER,C.A. NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF U02 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL.

  -PAGE,R.E.

NUREG/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report, April-June 1984.

PALAZZO,R.J.

NUREG/CR-3516: A SURVEY OF THE USES OF RADIOACTIVE MATERIALS IN

LOUISIANA'S OFFSHORE WATERS.

PAPASPYROPOULOS NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual r Report. March 1984 - September 1984. PASUPATHI,V. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984. , PELOGUIN,R.A. NUREG/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981. PETERSON,A.C. NUREG/CR-3772: RELAP5 ASSESSMENT: SEMISCALE SMALL BREAK TESTS S-UT-1,S-UT-2, S-UT-6,S-UT-7 AND S-UT-8. NUREG/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT:NEPTUNUS PRESSURIZER TEST YOS. PICIULO,P.L. NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT SHEFFIELD, ILLINOIS. l 69

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

NURED/CR-4083. ANALYSED 0F COILS FROM THE LOW-LEVEL RADIDACTIVE WASTE DISPOSAL SITES AT BARNWELL,SC AND RICHLAND,WA. PITTMAN,J. I NUREG-0933 SO2: A PRIORITIZATION OF QEERIC SAFETY ISSUES. Ic POLICASTRO,A.J. NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING i PONDS. t POLKY,J.N. NUREG/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. POPELAR,C. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual ReporteMarch 1984 - September 1984. PRICE,D.S. NUREG/CR-3516: A SURVEY OF THE USES OF RADIOACTIVE MATERIALS IN i- LOUISIANA'S OFFSHORE WATERS. 1 PUGH,C.E. NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. j RADFORD,L.R.

                       ' NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations.

RANDOLPH.P.D. NUREG/CR-3237: CONTROL OF EXPLDSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. , j, RAVINDRA,M.K.

NUREC/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectiv Induced By Earthquakes.

.                        NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGIEERING PWR PLANTS, Volume 3: Double Ended Guillotine Break Indirectly Induced By Earthquakes.
!                  REST,J.

NUREC/CR-3980 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH I PROGRAMS. Guarterly Progress Report, April-June 1984. 1 RIANI,L. NUREG-0933 SO2: A PRIORITIZATION OF GEERIC SAFETY ISSUES. RIGOS,R. NUREG-0933 SO2: A PRIORITIZATION. OF GEERIC SAFETY ISSUES. 4 ROBERTSON,D.E. NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) ROONEY,J.J.

NUREC/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B.

NUREQ/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. , SAGENDORF,J.F. NUREQ/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of

. Trajectories Concentration Patterns And MESODIF Model Calculations.
SAHA,P.

. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. SA! LOR,V.L. MJREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. r NUREQ/CR-4068:

SUMMARY

OF HISTORICAL EXPERIENCE WITH RELEASES OF RADIOACTIVE MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE-UNITED STATES. SAMANTA,P. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. 4 70

_~ __. - _ - .- - - -- . - - - . - - - - - . SAWYER,W.D. NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. SCOTT,P. NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March'1984.- September 1984. SEARS,R.G. NUREG/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIDACTIVE MATERIALS. -SEAVER,D.A. NUREG/CR-3688 VO1: 9ENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREG/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDOMENT. Volume 2: Appendices. SEGE, G. NUREG-0933 SO2: A PRIORITIZATION OF GEPERIC SAFETY ISSUES. BERNE,R.J. NUREG/CR-3709: . METHODS OF MINIMIZING GROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. SHA,W.T.

  - NUREG/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EGUATIONS FOR MULTIPHASE FLOW. Part One: System Without Internal Solid Structures.

SHEA,C.E. NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIDACTIVE WASTE DISPOSAL SITE AT SHEFFIELD, ILLINOIS. NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITES AT BARNWELL,SC AND RICHLAND,WA. t SHOR,R.W. NUREG/CR-3738: EfNVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE. A Review Of Data For Technetium. SHORT,S.A. NUREG/CR-3805 VO2: ENGINEERING CHARACTERIZATION OF QROUND MOTION. Task II: Effects Of Ground Motion Characteristics On Structural Response Considering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. SILVER,E.G. NUREG/CR-3430 VO2: NUCLEAR POWER PLANT OPERATING EXPERIENCE 1982. Annual Report. SIMPSON,H.J. NUREC/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION CDEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. DISKIND,B. NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE , BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. l SKAGGS,R.L. NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUP1 TAILINGS IMPOUNDMENTS. SMALLEY,J.F. NUREG/CR-3865: EVALUATION OF THE RADIDACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. SMITH,C.R. NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. SMITH,F.J. NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEDCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. CMITH,J.H. NUREC/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING 71

PROGRAM. S:micnnuel Prcgroon Report For Poried Ending Juno 30,1984. 900,P. NUREG/CR-2482 VO6: REVIEW - OF,. DOE WASTE PACKAGE PROGRAM. Subtas k 1.1 - , National Weste Package Program. October 1983 - March 1984. NUREG/CR-2482 VO7: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtes k 1.1 - National Weste Package Program April !?S4 - September 1984. 800 S. L. i NUREG/CR-3989: TIME AND VOLUME-AVERAGED CONSERVATION EGUATIONS FOR MULTIPHASE FLOW. Part One: System Without Internal Solid Structures. SPANNER,G.E. NUREG/CR-4023: FIELD PERFORMANCE ASSESSMENT-OF SYNTHETIC LINERS FOR URANIUM TAILINGS POND.A Status Report.

STAHL,D.

NUREG/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING. Guarterly Rep ort, July-September 1984. START,G.E. NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MESODIF Model Calculations. STEVENS,D.L. NUREG/CR-3659: A MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES OF

INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS.

STILLWELL,W.G. NUREG/CR-3688 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Rep ort. NUREG/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Vol ume 2: Ap p e nd ic e s. SUTCLIFFE,C.R.

;   NUREG/CR-4045: LITERATURE REVIEW ON AEROSOL-SAMPLING DEVICES FOR RESPIRATORY FIELD STUDIES.

SWYLER,K.J. NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. SYKES,R.I.

NUREG/CR-41.57: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION.

SZAWLEWICZ,S.A.* NUREG/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO58 VO2: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO58 VO4: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO58 VOS: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY

;      RESEARCH INFORMATION MEETING.

l NUREG/CP-OO58 VO6: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING.

TAYLOR,J.H.

NUREG/CR-3026: FEASIBILITY STUDY ON THE ACGUISTION OF LICENSEE EVENT DATA. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. TEICHMApei, T. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. THATCHER,D. NUREG-0933 802: A PRIORITIZATION OF QEPERIC SAFETY ISSUES. THOMA,J.O. NUREG-1110: COMPARISON OF LICENSING ACTIVITIES FOR OPERATING PLANTS DESIGNED BY BABCOCK & WILCOX. THOMAS,V.W. NUREG/CR-4057: RADIOLOGICAL ASSESSMENT-OF THE TOWN OF EDGEMONT. l 72 l

NUREO/CR-0100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT 0F YELLOWCAKE EMISSIONS. a ' THOMPSON,S.L. i NUREG/CR-3802: RELAPS ASSESSMENT: 0UANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS. THORNGREN,L.G. NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MESODIF Model Calculations. TILL,J.E. NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review 4 Of Data For Technetium. TOBIAS,M.L. NUREG/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual Progress Report For April 1984-September 1984. TORRONEN,K. NUREG/CR-4121: EFFECTS 0F SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE CRACK ORDWTH RATES IN LWR ENVIRONMENTS. TOSTE,A.P. NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) TRAVIS,J.R. l NUREG/CR-4020: HMS: A COMPUTER PROGRAM FOR TRANSIENT, THREE-DIMENSIONAL MIXING GASES. TRIER,R.M. ; J NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. . TSANG,F.Y. NUREC/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. TURGEON,K.S. NUREQ/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. TURNER,J.H. . NUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. NUREQ/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D.

URIBE,R.

NUREG/CR-3981: BI0 ACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. VANDER MOLEN,H. NUREG-0933 SO2: A PRIORITIZATION OF GElERIC SAFETY ISSUES. VOSKA,K.J.

NUREG/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT REPORTS.

WALTERS,W.H. NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. WASCOM,R.L. l NUREG/CR-3516: A SURVEY OF THE USES OF RADIDACTIVE MATERIALS IN I LOUISIANA'S OFFSHORE WATERS. WASTAG,M. l NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PONDS. WEINSTOCK,E.V. i NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. WEISS,A.J. ! NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. WHEELER,W.A. NUREG/CR-3817: DEVELOPMENT, USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations. WILK0WSKI,G.M. 73

i NURE3/CR-4082 VD1: DEORADED PIPING PROGRAM - PHASE II.Comicnnual RoporteMarch 1984 - September 1984. WILSON, M.' D. NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. WOMELSDUFF,J.E. NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN i GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. WOO, H. H. NUREG/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIOACTIVE MATERIALS. WRIGHT,M.A. NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. WULFF,W. NUREG/CR-3943: THE BWR PLAN ANALYZER. YAZDANDOOST,A. NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG FERFORMANCE. Annual Report, June 1983 - May 1984. YOUNG,J.A. NUREG/CR -4057: RADIOLOGICAL ASSESSMENT OF THE TOWN OF EDGEMONT. s

1 i

t 74

l l l I I Subject index This index was developed from keywords and word strings in titles and ab-stracts. During this development period, there will be some redundancy, which will be removed later when a reesonable thesaurus has been developed through experience. Suggestions for improvements are welcome. ALARA NUREO/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1984. ASPE Code NUREG/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIDACTIVE MATERIALS. ) ATWS NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT CIMULATION CODE. 4 NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING EXTRAPOLATION TECHNIGUES. A6ctrcet 4 NUREG-0304 VO9 NO4: REQULATORY AND TECHNICAL REPORTS. Annual Compilation Fcr 1984. Accidcnt l NUREG-09S1 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FOR RESPONSE l TO A COMMERCIAL NUCLEAR REACTOR ACCIDENT. NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. NUREO/CR-3943: THE BWR PLAN ANALYZER. ! NURE9/CR-3954: HECTR ANALYSIS OF EQUIPPENT TEMPERATURE RESPONSES TO CELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN l LIGHT-WATER REACTORS. I l NUREO/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF U02 DEBRIS IN SODIUM i WITH DOWNWARD HEAT REMOVAL. l NUREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. I NUREO/CR-4172: A USER 'S GUIDE FOR MERGE. AccCDtic Emission NURE0/CR-3825 VO3-4: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE l MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 - l Ocptember 1984. Volumes 3 and 4. AcrInvas And Initialisms NUREG-0544 RO2: A HANDBOOK OF ACRONYMS AND INITIALISMS. l Actinide Element l NURE9/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. Adacrption NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION 75

CDEFFICIENT3 OF ACTINIDE ELEMENTO IN SULFATE LAKE ENVIRONMENTC. Adv0rco CyotOco IntGrcctien NUREQ/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. NUREQ/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. Aerosol NUREQ/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual Progress Report For April 1984-September 1984. NUREQ/CR-3904: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984. NUREQ/CR-4045: LITERATURE REVIEW ON AEROSOL-SAMPLING DEVICES FOR RESPIRATORY FIELD STUDIES. Agenda NUREG-0936 VO3 NO4: NRC REQULATORY AGENDA.Guarterly Report,0ctober-December 1984. Aging NUREQ/CR-2331 VO4 N2: SAFETY RESEARCH PROGRANS SPONSORED BY OFFICE OF NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. NUREQ/CR-4008: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. Ambient Radiation Level NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1984. Anticipated Transient Without Scram NUREQ/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING EXTRAPOLATION TECHNIGUES. 4 Aquifer Restoration NUREQ/CR-3709: METHODS OF MINIMIZING QROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. Atmosphere NUR EQ/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT QAS AND DRY AIR STORAGE ATMOSPHERES. Atmospheric Dispersion NUREQ/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. B-Value NUREQ/CR-4145: EARTHGUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. BWR-LTAS NUREQ/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. Barrier Degradation Event NUREQ/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. Bayesian Method NUREQ/CR-4145: EARTHGUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. Below-Water-Table NUREQ/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. Bioaccumulation NUREQ/CR-3981: BI0 ACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. Bioassay NUREC/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984. Biofouling 76

i 1

NURED/CR-fi,70 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER; l. SYSTEMS Volume 2: Current Status of Biofouling Surveillance And Cont'rol Techniques.

1 Bcrchole NUREG/CR-4174: ROCK _ MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE l PLUG PERFORMANCE. Annual Report, June 1983 - May 1994. Brock Flow Rate NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. . .Brocker

NUREG/CR-3791: CLOSEQUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2

, CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. i Bubble Behavior ; NUREG/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual

Progress Report For April 1984-September 1984. ,

j Budget NUREG-1100 VO1: FY 1986 BUDGET ESTIMATES. Burial Environment i NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIDACTIVE WASTE

DISPOSAL SITES AT BARNWELL,SC AND RICHLAND,WA.

Burial Site NUREG/CR-3865: EVALUATION OF THE RADIDACTIVE INVENTORY IN,AND l ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. ! Buried Weste NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIDACTIVE WASTE } DISPOSAL SITES AT BARNWELL,SC AND RICHLAND,WA. i Burner Location-

,                   NUREG/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST i                            PARAMETERS. Task A: IEEE Flame Test.

CITADEL f NUREG-1108: RADI0 ACTIVITY TRANSPORT FOLLOWING STEAM GENERATOR TUBE ! RUPTURE. . COBRA; NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly l Report, July-September 1984. 1 CORSOR

NUREG/CR-41.73: CORSOR USER'S MANUAL.

CRT-Displays : ! NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED DISPLAYS. CRW = NUREG/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER , l SYSTEMS. Volume 2: Current Status of Biofouling Surveillance And i Control Techniques. Ccble i NUREG/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST ! PARAMETERS. Task A: IEEE Flame Test. l NUREG/CR-4112 VO2: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST l l PARAMETERS. Task B: Firestop Test Method. l < Chcoical Dose-Rate. , NUREG/CR-4008: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL , I DOSE-RATE EFFECT. L 4 Checical Stabilization 2 NUREG/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CONTROL TECFNIGUES FOR URANIUM MILL TAILINGS PILES. , Circulating Raw-Water NUREG/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS. Volume 2: Current Status of Biof ouling Surveillance And Control Techniques. .!' C1cdding N

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

_. ~ - _ - _ . _ _ _ - .- _ - . _ _ . - - _ - - l i

<        NURE2/CR-3744 V!2: HEAVY-SECTION GTEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1994.                              1 NUREG/CR-3980 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH      l PROGRAMS. Guarterly Progress Report, April-June 1984.                  l Closeout                                                                    l NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS.                                               ,

NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2 CIRCUIT SREAKERS IN SAFETY-RELATED SYSTEMS. NUREG/CR-3794: CLOSEOUT .OF IE BULLETIN B0-25: OPERATING PROBLEMS WITH l TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. j Cobalt-60 Teletherapy Incident i NUREG-1103: CONTAMINATED MEXICAN STEEL. Importation Of Steel-.Into The United States That Had Been Inadvertently Contaminated With Cobalt-60 , As A Result Of Scrapping Of A Teletherapy Unit. Code NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF j NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June i , 30,1984. NUREG/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUBASSEMBLY HEAT TRANSFER'AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. ! NUREG/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT, NUREG/CR-3972: SETTLEMENT (F URANIUM MILL TAILINGS PILES: A COMPARISON OF ANALYSIS TECFMIGUES. NUREG/CR-4172: A USER'S GUIDE FOR MERGE. i Cold Leg NUREG/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT L2-5). Final Comparison Report. l Commiteent NUREG/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TD RADIDACTIVE RELEASES FROM NUCLEAR POER PLANT SITES IN'19t51. Comparison Studies { NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. ! Annual Summary Report - Fiscal Year 1994. Comparison of Licensing Activities , NUREG-1110: COMPARISON OF LICENSING ACTIVITIES FOR OPERATING PLANTS , DESIGNED BY BABCOCK 8: WILCOX. Compilation Of Rules ; I NUREG-0936 VO3 NO4: NRC REGULATORY AGENDA.Guarterly Report,0ctober-December 1984. i Computer Code NUREG-1108: RADIOACTIVITY TRANSPORT FOLLOWING STEAM QENERATOR TUBE RUPTURE. NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. i ! NUREG/CR-3772: - RELAPS ASSESSMENT: SEMISCALE SMALL BREAK TESTS l

8-UT-1 S-UT-2, 8-UT-6.S-UT-7 AND S-UT-S.

NUREG/CR-3SO2: RELAP5 ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN

TIME STATISTICS. ,

i NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report, July-September 1984. ~ NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENBER CONTAINMENT. ', NUREG/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS PRESSURIZER TEST YOS. 4 NUREG/CR-3936: RELAPS ASSESSMENT: CONCLUSIONS AND USER GUIDELINES. . Computer Display , l NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED I i DISPLAYS. 1 j Computer Model ! l 78 4

                                                                             - . . - ._ -~ . - - _ - _ - - - ._.
 ,    NURED/CR-4042: A 3-DIMEN3IONAL COMPUTER MODEL TO DIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM.

! Cgeputer Program NUREfe/CR-4020: HMS: A COMPUTER PROGRAM FOR TRANSIENT, THREE-DIMENSIONAL

 !         MIXING GASES.

Censolidation Characteristics i NUREG/CN-4087: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION CHARACTERISTICS. Ccntainer NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. NUREG/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.Guarterly Report, July-Septemher 1984. Ccntainment NURE9/CR-3954: HECTR ANALYSIS OF EGUIPPENT TEMPERATURE RESPONSES TO , ^ SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. I NURE9/CR-4020: HMS: A COMPUTER PROGRAM FOR TRANSIENT, THREE-DIMENSIONAL MIXING GASES. NUREG/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW l AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. ' i Ccntaminated Steel Products NUREG-1103: CONTAMINATED MEXICAN STEEL. Importation Of Steel Into The United States That Had Been Inadvertently Contaminated With Cobalt-60 .l .'As A Result Of Scrapping Of A Teletherapy Unit. Centamination ! NURE9/CR-3709: METHODS OF MINIMIZING OROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. Coolant Boilaway NURE9/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly l Report, July-Septomber 1984. ! Cooling Pond Modeling

NURE9/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING

        ' PONDS.

Care Degradation l NUREG/CR-4173: CORSOR USER'S MANUAL. Care Disruptive Accident Analysis NUREO/CR-3804 VO3: PHYSICS OF REACTOR SAFETY.Guarterly , Report, July-Septomber 1984. l Ccre Meltdown NUREG/CR-3912: MARCH-HECTR ANALYSIS OF BELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. Coot NURE9/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1984. ,- Crcck Growth NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. Crcck NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHABE II. Semiannual Report, March 1984 - September 1984. Critical Flow NUREG/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT. D10 Experiment NURE0/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL.

   -DE!B l      NURE9/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT

! LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly l Induced By Earthquakes. NUREG/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS ,

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

p OF COMBUSTI(M ENGINEERING PWR PLANTS. Volu=o 1: Summary Ropert. Damage Definition L NUREG/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST-l- . - PARAMETERS. Tas k - A: IEEE Flame Te st. I Debris Cooling NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM

WITH DOWNWARD HEAT REMOVAL. '

Decay Heat NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. Decommissioning l NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1984. Degradation NUREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. Degraded Core NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. Deposition Velocity NUREG/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. Deposition NUREG/CR-3994: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND l DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress l Report, April 1983 - March 1984.

Diesel Generator Nm"EG/CP-Od31: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT l W M NTS. Interim Report - Diesel Generators, Batteries, Chargers And

, Inverters. Differential Integral Conservation Equation NUREG/CR-3999: TIPE- AND VOLUME-AVERAGED CONSERVATION EQUATIONS FOR j MULTIPHASE FLOW. Part One: System Without Internal Solid Structures. j Diffusion NUREG/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR POWER PLANTS UTILIZING REAL TIME ANEMOMETER STATISTICS. Digests And Indexes NUREG-0750 V2O 101: INDEXES TO NUCLEAR REQULATORY COMMISSION ISSUANCES l FOR JULY-SEPTEMBER 1984. NUREG-0730 V2O 102: INDEXES TO NUCLAER REQULATORY COMMISSION ISSUANCES FOR JULY-DECEMBER 1984. Dispersion Model NUREG/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. Disposal Site NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT SPEFFIELD, ILLINDIS. Disposal

NUREG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES.

I Disposition Schedule NUREG-0910 RO1 SO1: NRC COPPREHENSIVE RECORDS DISPOSITION SCHEDULE. j NUREG-0910 RO1 802: NRC COPFREHENSIVE RECORDS DISPOSITION SCHEDULE. Distribution Coefficient NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. l Dose NUREG/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIDACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981. l f l 1

j' >.

NURED/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress l Report, April 1983 - March 1984. I NUREG/CR-4068:

SUMMARY

OF. HISTORICAL EXPERIENCE WITH RELEASES OF , i RADIDACTIVE MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE l UNITED STATES. j Dauble-Ended Guillotine Break NUREG/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly Induced By Earthquakes. NUREG/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summary Report. > NUREG/CR-3663 VO3: PRGBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT 3 LOOPS OF COMBUSTION ENGIPEERING PWR PLANTS, Volume 3: Double Ended

                 . Guillotine Break Indirectly Induced By Earthquakes.

Drenward Heat Removal. NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. Dru Air Storage

   -         NUREG/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT
;                  GAS AND DRY AIR STORAGE ATMOSPHERES.

Dry Deposition i NUREG/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. Ecrthen Radon Suppression Cover NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. 7 Ecrthquake

j NUREG/CR-3660 VO3:c PROBABILITY OF PIPE FAILURE IN THE REACTOR CDOLANT

,                 LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly Induced By Earthquakes.

NUREG/CR-3805 VO2: ENGINEERING CHARACTERIZATION UF GROUND MOTION. Task , II: Effects Of Orcund Motion Characteristics On Structural Response l Considering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. NUREG/CR-4036: STRUCTURAL GEDLOGY OF SOUTHEASTERN ILLINDIS AND VICINITY. NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. NUREG/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PL/NTS: ANALYSIS AND RANKING. Eddy Current NURE9/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. ! l Effective Peak Acceleration ! NUREG/CR-3 SOS VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task I II: Effects Of Ground Motion Characteristics On Structural Response Considering Localized Structural Nonlinearities And Soil-Structure , Interaction Effects. l Erorgency Core Cooling NUREG/CR-3866: ' TRAC-PD2 INDEPENDENT ASSESSMENT. Escreency Response NUREG-0981 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FOR RESPONSE TG A COMMERCIAL NUCLEAR REACTOR ACCIDENT. Eaissions NUREG/CR-4100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT OF YELLOWCAKE EMISSIONS. Enforcement Actions NUREG-0940 VO3 NO4: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED. Guarterly Progress Report, Octob er-December 1984. I 81 , 1

l Envircnment i NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review Of Data For Technetium. Environmental Assessment l NUREG-1112: ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF SPECIAL NUCLEAR f MATERIAL LICENSE NO. SNM-368.(UNC Naval Products Division Of UNC l . Resources,Inc) Environmental Assesssment NUREG-1130: ENVIRONMENTAL ASSESSMENT FOR RENEWAL AND CONSOLIDATION OF MATERIALS LICENSE NOS. SNM-362,SMB-405,OS-DO566-05, 08-OO566-10,AND 08-00566-12. Environmental Conseguences NUREG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS' DISPOSAL IN MINE STOPES. Environmental Effect NUREG/CR-3945: FATIQUE CRACK ORDWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT. E'xaminer Standard NUREG-1021 RO1: DPERATOR LICENSING EXAMINER STANDARDS. Exhaust Flow NUREG/CR-4112 VO1: INVESTICATION OF CABLE AND CABLE SYSTEM FIRE TEST P ARAMETERS. Task A: IEEE Flame Test. Explosive-Gas NUREQ/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. Exposure NUREG-0713 VO5: OCCUPATIONAL RADIATION EXPOSURE AT COMMERCIAL NUCLEAR POWER REACTORS - 1983 ANNUAL REPORT. NUREG/CR-3984: BIDLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984. NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN

      . LIGHT-WATER REACTORS.

FAST-NUREG/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual Progress Report For April 1984-September 1984. FEMA NUREG-0981 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FDR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR ACCIDENT. FRAPCON NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PRDORAMS.Guarterly Report July-September 1984. Fabrication Criteria NUREG/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS. Failure NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR CDOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS, Volume 3: Double Ended Guillotine Break Indirect 1g Induced By Earthquakes. NUREQ/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Report - Diesel Generators, Batteries, Chargers And Inverters. NUREG/CR-3943: THE BWR PLAN ANALYZER. NUREG/CR-4123: SEISMIC FRAGILITY OF REINFORCED CONCRETE STRUCTURES AaD COMPONENTS FOR APPLICATION TO NUCLEAR FACILITIES. Fastener NUREG/CR-4112 VO1: INVESTICATION OF CABLE AND CABLE SYSTEM FIRE TEST , P ARAMETERS. Task A: IEEE Flame Test. Fatigue Crack Growth NUREG/CR-3945: FATIQUE CRACK QROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT. NUREQ/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIQUE 82

CRACK. GROWTH RATES IN LWR ENVIRONMENTS. Fcult NUREG/CR-4036: STRUCTURAL GEOLOGY OF SOUTHEASTERN ILLINOIS AND VICINITY. - NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. Fctritic Steel NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMI ANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. ' Filter Flow Test NUREG/CR-4056: PARTICULATE AND GAS RELEASE FROM LIGHT-WATER-REACTOR (LWR) FUEL RODS STORED IN INERT AND DRY AIR ATMOSPHERES. Final Environmental Statement NUREG-1073: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF RIVER BEND STATION. Docket No. 50-458.(Qulf States Utilities And Cajun Electric Power Cooperative) NUREG-1087: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF VDOTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Georgia Power Company) Fire Exposure NUREG/CR-4112 VO2: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task B: Firestop Test Method. Firestop NUREG/CR-4112 VO2: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task B: Firestop Test Method. Fish-NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. Fission Gas Release NUREQ/CR-4056: PARTICULATE AND QAS RELEASE FROM LIGHT-WATER-REACTOR

     ~(LWR) FUEL RODS STORED IN INERT AND DRY AIR ATMOSPHERES.

. Fission Product Modeling NUREG/CR-39BO VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH PROGRAMS. Guarterly Progress Report, April-June 1984. . Fission Product NUREG/CR-4172: A USER'S GUIDE FOR MERGE. Flame Test NUREQ/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task A: IEEE Flame Test. Flaw NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. l NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. NUREG/CR-3825 VO3-4: ' ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING DF NUCLEAR PRESSURE VESSELS.Guarterly Report. April 1984 - September 1984. Volumes 3 and 4. Flow Rate NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. Flow Redistribution NUREQ/CR-3498: TWO-DIMENSIONAL MODELING DF INTRA-SUBASSEMBLY HEAT TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. Fractography NUREQ/CR-3945: FATIGUE CRACK CROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT. NUREC/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE

    . CRACK CROWTH RATES IN LWR ENVIRONMENTS.

Fracture Mechanics NUREC/CR-3723:-STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. 83

l NUREG/CR-3744 VC2: HEAVY-SECTION STEEL TECHNOLODY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. I NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual l Report. March 1984 - September 1984. i I Fracture Toughness NUREG/CR-3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE , VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES. ; Fracture NUREG/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. Fragility NUREG/CR-4123: SEISMIC FRAGILITY OF REINFORCED CONCRETE STRUCTURES AND I COMPONENTS FOR APPLICATION TO NUCLEAR FACILITIES. Free Convection ' . NUREG/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUBASSEMBLY HEAT TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. l Fue1~ Aerosol Simulant Test-NUREG/CR-3830 VO2: AEROSDL RELEASE AND TRANSPORT PRDGRAM. Semiannual Progress Report For April 1984-September 1984. ! Fuel Flow NUREG/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST P ARAMETERS. Tas k A: IEEE Flame Test. Fuel Performance NUREG/CR-3950 VO1: FUEL PERFORMANCE ANNUAL REPORT FOR 1983. l- Fuel Rod Cladding NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. Fuel Rod NUREQ/CR-4056: PARTICULATE AND GAS RELEASE FROM LIGHT-WATER-REACTOR i (LWR) FUEL RODS STORED IN INERT AND DRY AIR ATMOSPHERES. Fugitive Dust NUREG/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CDNTROL TECHNIGUES FOR URANIUM MILL TAILINGS PILES. OE Type AK-2 Circuit NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF QE TYPE - AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. Galvanized Steel Corrosion NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN ! GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. } Gamma-Reg i NUREG/CR-4039: GAMMA-RAY CHARACTERIZATION OF THE TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. Gas Analyzer I i NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE. GAS SYSTEMS. Gaseous Tracer ; NUREQ/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories. Concentration Patterns And MESODIF Model Calculations. Generic Safety Issues NUREG-0933 SO2: A PRIORITIZATION OF GENERIC SAFETY ISSUES. I Ground Motion NUREG/CR-3805 VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task II: Effects Of Ground Motion Characteristics On Structural Response Considering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. l Ground Water NUREQ/CR-3709: METHODS OF MINIMIZING GROUND-WATER CONTAMINATIDN FROM IN SITU LEACH URANIUM MINING. Final Report. NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) Guidance NUREG-1021 RO1: OFERATOR LICENSING EXAMINER STANDARDS. 84 1

2. . . ._ _ _ l

HECTR NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICEvCONDENSER CONTAINMENT. NUREG/CR-3954: HECTR ANALYSIS OF EGUIPPENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. HMS NUREG/CR-4020: HMS: A COMPUTER PRDORAM FOR TRANSIENT THREE-DIMENSIONAL MIXING GASES. HSST

    - NUREQ/CR-3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE   l VESSEL PLATE ~AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES.       '

H:ndbook NUREG-0544 RO2: A HANDBOOK OF ACRONYMS AND INITIALISMS. ] NUREC/CR-4041: SYSTEM ANALYSIS HANDBOOK. H alth NUREC/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND

       , DOSE ESTIMATES FDR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984.

H2at Transfer NUREG/CR-3498: TWO-DIMENSIONAL.' MODELING OF INTRA-SUBASSEMBLY HEAT TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. Hsat-Affected Zone NUREG/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report, April-June 1984. Hacvg Section Steel Technology

NUREG/CR-3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE

VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES.

High-Burnup Fuel NUREG/CR-3950 VO'1: FUEL PERFORMANCE ANNUAL REPORT FOR'1983. High-Level Waste NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. NUREG/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FDR HIGH-LEVEL WASTE PACKAGING. Guarterly Rep ort, July-September 1984. High-Speed Simulation NUREQ/CR-3943: THE BWR PLAN ANALYZER. Human Error NUREQ/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT REPORTS. NUREG/CR-36BS VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREQ/CR-36BS VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 2: Appendices. Hunan Factors i NUREG-0800 18. 2 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To

SRP Section 18.2, " Safety Parameter Display System (SPDS)." NUREQ/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. Huran Reliability Estimate NUREG/CR-36BS VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREQ/CR-36BS VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT

      - JUDGMENT. Vol ume 2: Ap p e n d ic e s.

Huran Reliability e NUREC/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT R EPDR TS. j Hydrogen Burns i l 85

NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. Hydrogen Generation NUREQ/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN QENERATION FROM QALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. Hydrogen Transport NUREQ/CR-4020: HMS: A COMPUTER PROGRAM FOR TRANSIENT, THREE-DIMENSIONAL MIXING GASES. Hydrogen-Oxygen Flammability NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. Hydrogen NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. f NUREC/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIQUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. IE Bulletin 79-09 NUREQ/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF QE TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. IE Bulletin 79-12 NUREG-0905: CLOSEDUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS. IE Bulletin 80-25 NUREQ/CR-3794: CLOSEOUT OF IE BULLETIN 80-25: DPERATING PRDBLEMS WITH TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. IEEE 383 Flame Test NUREQ/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task A: IEEE Flame Test. IPRD NUREG/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Report - Diesel Generators, Batteries, Chargers And Inverters. Ice Condenser NUREQ/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. In Situ Leaching NUREC/CR-3709: METHODS OF MINIMIZING QROUND-WATER CONTAMINATIGN FRDM IN SITU LEACH URANIUM MINING. Final Report. In-Plant Reliability Data NUREQ/CR-3031: THE IN-PLANT RELI ABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Report - Diesel Generators, Batteries, Chargers And Inverters. ' Independent Assessment l NUREG/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS j PRESSURIZER TEST YO5. ! Independent Safety Organization l NUREC/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. Index NUREG-0304 VO9 NO4: REGULATORY AND TECHNICAL REPORTS. Annual Compilation For 1984. NUREG-0750 V2O 101: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JULY-SEPTEMBER 1984.

         . NUREG-0750 V2O 102: INDEXES TO NUCLAER REGULATORY COMMISSION ISSUANCES FOR JULY-DECEMBER 1984.

Inert Qas

NUREQ/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT QAS AND DRY AIR STORAGE ATMOSPHERES.

Influence Coefficient NUREC/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE CDEFFICIENTS FOR 86

d EURFACE FLAWS IN PRESSURE VESSELS. Infrared Analysis NUREC/CR-3984: BIDLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND

                                   -DDSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report. April 1983 - March 1984.

Inhalation

                        . NUREG/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress

, Report, April 1983 - March 1984. Inspection NUREG-OO40 VOS NO4: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report,0ctober-December 1984. (White Book) NUREQ/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. , Integral Systems i NUREG/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY' RESEARCH INFORMATION MEETING. Interfacial Transfer Integral NUREQ/CR-3989: TIME- AND VOLUME-AVERAGED CDNSERVATION EQUATIONS FOR MULTIPHASE FLOW. Part One: System Without Internal Solid Structures. International Probabilistic Safety Assessment NUREG/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT . EXPERIENCE TO THE U. S. NUCLEAR REQULATORY PROCESS. International Standard Problem 13 NUREC/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT L2-5). Final Comparison Report. l

             - Inventory NUREQ/CR-3865: EVALUATION OF THE RADIOACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE.

Indine i NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN j LIGHT-WATER REACTORS. ]. Irradiation NUREG/CR-3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES. , NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING ! INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. NUREC/CR-4039: GAMMA-RAY CHARACTERIZATIDN OF THE TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. l [ Isotope Exchange j NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN I I LIGHT-WATER REACTORS. Icotopic Release Rates l NUREG/CR-3865: EVALUATION OF THE RADIOACTIVE INVENTORY IN,AND t ESTIMATION OF ISOTOPIC RELEASE FROM, THE WASTE IN EIGHT TRENCHES AT l THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. J-Integral / Tearing Modulus , NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984. Jsb Performance Aid NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations. LER NUREG/CR-2OOO VO3N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of December 1984. NUREC/CR-2OOO YO4 N1: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of January 1985. 1 NUREQ/CR-2OOO VO4 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of Februarv~1985. l 87 i f

NUREG/CR-3026: FEASIBILITY CTUDY ON THE ACQUIDTION OF LICENSEE EVENT DATA. , LIDAR NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Compavison Of Trajectories, Concentration Patterns And MESODIF Model Calculations. LOCA NUREG/CR-3802: RELAPS ASSESSMENT: GUANTITATIVE MEY PARAMETERS AND RUN TIME STATISTICS. LOFT L2-5 NUREG/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT L2-5). Final Comparison Report. Leachate Movement 4 NUREG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. Legal Issuances NUREG-0750 V2O NO4: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR DCTOBER 1984. Pages 1,055-1,435. 4 NUREG-0750 V2O NOS: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR MOVEMBER 1984. Pagas 1,437-1,572. NUREG-0750 V2O N06: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR DECEMBER 1984. Pagas 1,573-1,706. NUREG-0750 V21 NO1: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JANUARY 1985. Pages 1-273. Licensed Operating Reactors NUREG-OO2O VOS N12: LICENSED OPERATING REACTORS STATUS

SUMMARY

 !      REPORT. Data As Of November 30,1984.(Gray Book I) i NUREG-OO2O VO9 NO1: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of December 31,1984.(Gray Book I). NUREG-OO2O YO9 NO2: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of January 31,1985.(Gray Book I) Licensee Event Report NUREG/CR-2OOO VO3N12: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of December 1984.

NUREG/CR-2OOO VO4 N1: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of January 1985.

NUREG/CR-2OOO VO4 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of February 1995. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. NUREQ/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT REPORTS. Load Combination NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. Load Ratio l NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. Loss Of Offsite Power NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND PARTIAL LOSSES OF OFF-SITE ~ POWER AT NUCLEAR POWER PLANTS. Loss-Of-Coolant Accident NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. ' NUREG/CR-3802: RELAP5 ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS.

NUREG/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT.

l NUREG/CR-3936: RELAP5 ASSESSMENT: CONCLUSIONS AND USER QUIDELINES. Loss-Of-Fluid Test NUREC/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING l a E_________ - _ _ _ _ _ _ _ . __ . _ - -

EXTRAPOLATION TECHNIGUES.

    -Lgw-Level Radioactive Waste
       ' NURE9/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE j-             BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS.

NUREG/CR-3865: EVALUATION OF THE RADIDACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. ' NUREO/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT SMEFFIELD, ILLINDIS. , MARCH-2 NUREG/CR-4172: A USER'S GUIDE FOR MERGE.' 1 -MARCH NURES/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN

ICE-CONDENSER CONTAINMENT.

MERGE NUREO/CR-4172: A USER'S QUIDE FOR MERGE. MEBODIF NUREC/CR-34SS VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MEBODIF Model Calculations. MINET l NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF i NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June i 30,19d4. MINTED NUREO/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. MOD 1 NUREQ/CR-3SO2: RELAPS ASSESSMENT: GUANTITATIVE MEY PARAMETERS AND RUN TIME STATISTICS. Maintenance NUREQ/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations. M:ndrel Loading Test NUREG/CR-3980 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH PROGRAMS. Guerterly Progress Report, April-June 1984. ' Mathematical Model NUREG/CR-3659: A MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES OF INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. MJosurement Uncertainties NUREG/CR-3659: A MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES OF INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. l Machanical Stability NUREC/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE i

          . BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS.

Maltdown [ NUREQ/CR-4172: A USER'S QUIDE'FOR MERGE. l Matallurgical Blind Test NUREC/CR-4039: QAMMA-RAY CHARACTERIZATION OF THE TWO-YEAR IRRADIATION l EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY.

   .M2teorology NUREG/CR-3408 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MESODIF Model Calculations.

, M2xican Incident NUREG-1103: CONTAMINATED MEXICAN STEEL. Importation Of Steel Into The

         ; United States That Had Been Inadvertently Contaminated With Cobalt-60 As A Result Of Scrapping Of A Teletherapy Unit.

Migration NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) Minicomputer Simulation l l

NURED/CR-3943: THE BWR PLAN ANALYZER. Mitigation of Erosion , NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. Model Uncertainty NUREG/CR-4157: -A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. Modeling NUMEG/CR-4173: CORSOR USER 'S MANUAL. i Monitoring NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1984. , NUREG/CR-3943: THE BWR-PLAN ANALYZER. f NUREG/CR-4100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT OF YELLOWCAKE EMISSIONS.

. Multiphase Flow System NUREG/CR-3999: TIME- AND VOLUME-AVERAGED CONSERVATION EQUATIONS FOR MULTIPHASE FLOW. Part One: System Without Internal Solid Structures.

4 Multiple Frequency NUREG/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. NEPTUNUS NUREG/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS l PRESSURIZER TEST YO5. t NPRDS NUREG/CR-3026: FEASIBILITY STUDY ON THE ACGUISTION OF LICENSEE EVENT , DATA. Natural Convection Core Cooling NUREQ/CR-3804 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July-Septomber 1984. Nuclear Waste NUREG/CR-38t,1 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. Nuclear-Grade Carbon NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN LIGHT-WATER REACTORS. Occupational Radiation Exposure NUREG-0713 VO5: OCCUPATIONAL RADIATION EXPOSURE AT COMMERCI AL NUCLEAR POWER REACTORS - 1983 ANNUAL REPORT. Official Records NUREG-0910 RO1 SO2: NRC C0ffREHENSIVE RECORDS DISPOSITION SCHEDULE.

,    Operating Experience NUREG/CR-3430 VO2: NUCLEAR POWER PLANT OPERATING EXPERIENCE -

1982. Annual Report.

Operating Reactors Licensing Actions NUREG-0748 VO4 N12: OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data ' As Of December 31,1984.(Orange Book) i NUREG-0748 VO5 NO1: OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data As Of January 31,1985.(Orange Book) Operational Procedures NUREG-0981 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR ACCIDENT. Organization Chart NUREG-0325 RO7: U. S. NUCLEAR REGULATORY COMMISSION FUNCTIONAL ORGANIZATION CHARTS. [ PHREEDE NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DDE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984.

.90

i PRA

              . NUREG-1115: CATEGORIZATION OF REACTOR SAFETY ISSUES FROM A RISK j                  PERSPECTIVE.
           .PVC
 ~

NUREG/CR-4008: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT.

          .Pollet-Cladding
               ' NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report, July-September 1984.

NUREG/CR-3999* ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING INTERACTION ' (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. '- Parmeameter Testing NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report June 1983 - May 1984. Potitions For Rulemaking NUREG-0936 VO3 NO4: NRC REQULATORY AGENDA.Guarterly R ep or t, 0c tob er-De c emb er - 1984. Phosphorus-32 NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. Pipe Failure NUREG/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summary Report. Pipe Whip Restraints NUREQ/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly Induced By Earthquakes; Pipe NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS Volume 3: Double Ended Guillotine Break Indirectly Induced By Earthquakes. NUREG/CR-3945: FATIQUE CRACK GROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT.

                                       -                                                       I NUREG/CR-4082 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report. March 1984 - September 1984.

l Planning Guidance ! NUREG-0885 104: U. S. NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING GUIDANCE 1985.

          'Plcnt Analyzer NUREG/CR-3943: THE BWR PLAN ANALYZER.

j Plcnt Safety

             ' NUREQ/CR-3943: THE BWR PLAN ANALYZER.

Plug NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Rep ort, June 1983 - May 1984. Policy And Planning NUREG-0885 104: U. S. NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING GUIDANCE 1985. Polyvinylchloride NUREG/CR-4008: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. Pond Surface Heat Transfer NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING P ONDS. Pcpulation NUREG/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981. Pont-LOCA Hydrogen Generation NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. Pontirradiation 91

NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Cuart:rly . R ep ort, July-Sep temb er 1984. Precipitation NUREG/CR-4157: A SCIENTIFIC CRITIQUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. ! Pressure Differential l NUREG/CR-4112 VO2: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST l PARAMETERS. Task B: Firestop Test Method. ,

              -Pressure Suppression Pool NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE.

Pressure Vessel NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLt'ENCE COEFFICIENTS FOR ! SURFACE FLAWS IN PRESSURE VESSELS. NUREG/CR-3744 VO2: W AVY-SECTION STEEL TE HNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FDE APRIL-SEPTEMBER 19'.s .

NUREG/CR-3825 VO3-4: ACOUSTIC EMISSIDN/FL nW RELATIONSHIP FOR IN-SERVICE i MONITORING DF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 -

September 1984. Volumes 3 and 4. i NUREG/CR-3978: TENSILE PROPERTIES OF IRF ADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES. Pressurized Thermal Shock NUREG/CR-2331 VO4 N2: SAFETY RESEARCH FtOGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterl, Progress Report, April 1-June 30,1984. l NUREG/CR-3723: STRESS-INTENSITY-FACTO; INFLUENCE CDEFFICIENTS FOR i L SURFACE FLAWS IN PRESSURE VESSELS. Prioritization NUREG-1115: CATEGORIZATION OF REACTJR SAFETY ISSUES FROM A RISK PERSPECTIVE. Probabilistic Risk Analysis NUREG-1115: CATEGORIZATION OF REA, TOR SAFETY ISSUES FROM A RISK PERSPECTIVE. Probabilistic Risk Assessment NUREG/CR-3519: HUMAN ERROR PROB.BILITY ESTIMATION USING LICENSEE EVENT REPORTS. NUREQ/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT EXPERIENCE TO THE U.S. NUCLEAR REQULATORY PROCESS. Probability Assessment NUREQ/CR-3688 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREQ/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 2: Appendices. ' Procedure NUREQ/CR-3817: DEVELOPMENT, USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations. Public Safety l NUREG/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT EXPERIENCE TO THE U.S. NUCLEAR REQULATORY PROCESS. RAMONA-3B NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY DFFICE OF NUCLEAR REQULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. l RCL NUREQ/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LODP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly Induced By Earthquakes. RELAP 5 NUREQ/CR-3772: RELAPS ASSESSMENT: SEMISCALE SMALL BREAK TESTS S-UT-1,S-UT-2, S-UT-6,S-UT-7 AND S-UT-S. 92 i

NUREG/CR-3802: RELAPS ASSESSMENT: QUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS. NUREG/C,R-3936: RELAPS ASSESSMENT: CONCLUSIONS AND USER QUIDELINES.

              -NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING EXTRAPOLATION TECHNIQUES.

RELAPS/ MOD 1 NUREG/OR-3936: RELAPS ASSESSMENT: CONCLUSIONS AND USER QUIDELINES. Rcdiation Exposure , NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOPMISSIONING PROJECTS. Annual Summary Report - Fiscal Year 1984. Rcdiation Stability , NUREC/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. Radiation NUREG/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984. NUREG/CR-4039: GAMMA-RAY CHARACTERIZATION OF THE TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. Radioactive Emission NUREG/CR-4068:

SUMMARY

OF HISTORICAL EXPERIENCE WITH RELEASES OF l RADIDACTIVE MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE UNITED STATES. Radioactive Iodine NUREG/CR-3990: CHARCDAL PERFORMANCE UNDER ACCIDENT CONDITIDNS IN i LIGHT-WATER REACTORS. Rcdioactive Isotopes NUREG/CR-3865: EVALUATION OF THE RADIDACTIVE INVENTDRY IN,AND j ESTIMATION DF ISOTOPIC. RELEASE FROM, THE WASTE IN EIGHT TRENCHES AT i THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. Rcdioactive Material NUREG/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIOACTIVE MATERIALS. NUREG/CR-3516: A SURVEY OF THE USES OF RADIDACTIVE MATERIALS IN LOUISIANA'S OFFSHORE WATERS. NUREG/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS. I Rcdioactive Release

 !             NUREG/CR-2850 VO3: POPULATION DOSE COMMITMENTG DUE TO RADIDACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981.

Radioisotope NUR EG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAME ENVIRONMENTS. Radiological Asseesment NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review Of Data For Technetium. NUREC/CR-4057: RADIOLOGICAL ASSESSMENT OF THE TOWN OF EDGEMONT. I Rcdionuclide

'             NUREO/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981.

NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIDNUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY j SITE PROJECTS. Report For April-June 1984. NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) Rocctor Coolant Loop NUREG/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR CDOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Guillotine Break Indirectly , Induced By Earthquakes. 1 NUREC/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summary Report. NUREC/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT t i 93

LOOPS OF COMBUSTION EN31EERING PWR PLANTC. Valuno 3: Doub10 Endad Guillotine Break Indirectly Induced By Earthquakes. 1 - Reactor Pressure Boundary ! NUREG/CR-3825 VO3-4: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 - September 1984. Volumes 3 and 4. Reactor Safety NUREG/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREC/CR-2331 VO4 N2: BAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH. Guarterly Progress Report, April 1-June 30,1984. NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. NUREG/CR-3804 VO3: PHYSICS OF REACTOR SAFETY.Guarterly . Report. July-September 1984. NUREC/CR-3810 VO3: REACTDR SAFETY RESEARCH PROGRAMS.Guarterly Report. July-September 1984. NUREG/CR-3816 VO1: REACTOR SAFETY RESEARCH.Guarterly Report. January-March 1984. Reactor NUREG/CR-3943: THE BWR PLAN ANALYZER. NUREG/CR-3954: HECTR ANALYSIS OF EQUIP E NT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT.

. Regulation NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN i

NUCLEAR PDWER PLANTS. Problems And Recommendations. Regulatory And Technical Report NUREG-0304 VO9 NO4: REQULATORY AND TECHNIC AL REPORTS. Annual Compilation For 1984. Regulatory Approach NUREG-OB85 104: U. S. NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING QUIDANCE 1985. Reliability NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF LICENSEE EVENT DATA. NUREC/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Rep ort - Diesel Generators, Batteries, Chargers And Inverters. NUREG/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT EXPERIENCE TO THE U.S. NUCLEAR REGULATORY PROCESS. Repeatability

NUREC/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task A: IEEE Flame Test.

Residue Number System , NUREQ/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSDR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. Residue Processor NUREC/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT

INSPECTION SYSTEM IMAGE ENHANCEMENT.

I Respirator NUREC/CR-4045. LITERATURE REVIEW ON AEROSOL-SAMPLING DEVICES FOR RESPIRATORY FIELD STUDIES. Risk Analysis . NUREG-1115: CATEQDRIZATION OF REACTOR SAFETY ISSUES FROM A RISK PERSPECTIVE. Rock Fracture System NUREG/CR-4042 A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. Rock Mais Sealing 1 94

NURED/CR-4174: ROCK MAS 3 SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFCRMANCE. Annual Report, June 1983 - May 1984. Rick'Riprep-NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. y SAFT-UT l NUREG/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. 4 SASA i NURE9/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT l l SIMULATION CODE. SPD NURE9/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED

.             DISPLAYS.                                                                I 4

NUREG-0800 18. 2 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY

            -ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To SRP Section 18.2. " Safety Parameter Display System (SPDS)."

NUREG-0800 18. 2A1 RO: STANDARD REVIEW PLAN FOR-THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To 4 Appendix A To SRP Section 18.2, ." Human Factors Review Guidelines For j .The Safety Parameter Display System (SPDS)." , Safeguard

?      NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN NUCLEAR POWER PLANTS. Problems And Recommendations.

s Ccfety Evaluation Report !' NUREG-0675 S29: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2. Doc ket Nos. 50-275 And 50-323.(Pacific Gas And Electric Company) NUREG-0787 S10: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF j WATERFORD STEAM ELECTRIC STATION, UNIT 3. Docket No. 50-382.-(Louisiana Power And Light Company) NUREG-0797 807: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF. 1 COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2. Doc ket Nos. 50-445 [ And 50-446.(Texas Utilities Generating Company, et al) NUREG-0797 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION, UNITS.1 AND 2. Docket Nos. 50-445

And 50-446.(Texas Utilities Generating Company, et al)

NUREG-0797 SO9: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos. 50-445

.            And 50-446.(Texas Utilities Generating Company) l NUREG-0798 905: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF

ENRICO FERMI ATOMIC POWER PLANT, UNIT NO. 2.Dochet No. 50-341.

} (Detroit Edison Company) NUREG-0847 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Tennessee Valley Authority) NUREG-OS47 SO4: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF. WATTS'BAR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391. ! (Tennessee Valley Authority) NUREG-0853 SO4:s SAFETY' EVALUATION REPORT RELATED TO THE OPERATION OF CLINTON POWER STATION, UNIT 1. Docket No. 50-461.(Illinois Power

           . Company,et al)

NUREG-0876 ' SO6: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF BYRON STATION, UNITS 1 AND 2. Docket Nos. 50-454 And 50-455. } (Commonwealth Edison Company) l NUREG-0881 905: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF l WOLF CREEK OENERATING STATION, UNIT 1.Dockat No. 50-482.(Kansas Gas And Electric Company,et al) NUREG-0887 805: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF , THE PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-440 And 95

i '53-441.(Clovolend Electric Illuainating Ccspcng) ! NUREG-0979 903: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESICN ' f APPROVAL OF THE GESSAR II BWR/6 NUCLEAR ISLAND DESIGN. Docket No. 50-447. (General Electric Company) NUREG-1031 SO1: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MILLSTONE NUCLEAR POWER STATION, UNIT 3. Docket No. 50-423. (Northeast Nuclear Energy Company) NUREG-1047: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF NINE 3 MILE POINT NUCLEAR STATION UNIT NO. 2. Docket No. 50-410. (Niagara i Mohawk Power Corporation,et al) 4 NUREG-1048 801: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF i HOPE CREEK GENERATING STATION. Docket No. 50-354.(Public Service Electric and Gas Company) ! NUREG-1096: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE l DPERATING LICENSE FOR THE TRIGA TRAINING AND RESEARCH REACTOR AT THE ) l'NIVERSITY OF UTAH. Doc ket No. 50-407. (University of Utah) NUREG-1098: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF OPERATING LICENSE FOR THE RESEARCH REACTOR AT MANHATTAN COLLEGE. Docket No. 50-199. (Manhattan College) Safety Fuel Systems Research

NUREQ/CR-3980 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH PROGRAMS. Guarterly Progress Report. April-June 1984.

Safety Goals NUREQ/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL

ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS.

j NUREG/CR-4068:

SUMMARY

OF HISTORICAL EXPERIENCE WITH RELEASES OF l RADI0 ACTIVE MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE l UNITED STATES. J Safety Parameter Display System j NUREG-0800 18. 2 RO: STANDARD REVIEW PLAN FOR THE REVIEW DF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To SRP Section 18.2, " Safety Parameter Display System (SPDS)." { j NUREG-0800 18.2A1 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY i ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To Appendix A To SRP Section 18.2, " Human Factors Review Guidelines For l The Safety Parameter Display System (SPDS)." NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FDR CRT-GENERATED t DISPLAYS. j Safety Relief Valve l NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING = EXTRAPOLATION TECHNIGUES. ! NUREG/CR-3794: CLOSEQUT OF IE BULLETIN 80-25: 0PERATING PROBLEMS WITH TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. Safety Research 4 NUREG-1105: REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM FOR FISCAL YEARS 1986 AND 1987. i NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF ! NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report. April 1-June l 30,1984.

Safety-Related System NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS.

Seal i NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. Seismic Hazard NUREG/CR-3660 VO3: PRO 3 ABILITY OF PIPE FAILURE IN THE REACTOR COOLANT i LOOP OF WESTINGHOUSE PWR PLANTS. Volume 3: Gui11otine Break Indirect 1g i Induced By Earthquakes, j Seismic Risk Assessment 4 j 96 i

NURE3/CR-G123: SEIBMIC FRAGILITY OF REINFORCED CONCRETE OTRUCTURES AND l

                .COMPQNENTS FOR APPLICATION TO NUCLEAR FACILITIED.

C:miscale' Mod-25 NUREO/CR-4073: RESULTS OF THE SEMISCALE MOD-2B STEAM GENERATOR TUBE RUPTURE TEST SERIES.

          .S iscale Small Break Tests NUREG/CR-3772: RELAPS . ASSESSMENT: SEMISCALE SMALL BREAK TESTS S-UT-1,B-UT-2, S-UT-6,S-UT-7 AND S-UT-8.

Corvice-Water

             . NURE9/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER

SYSTEMS. Volume 2: Current Status of Biofouling Surveillance And 4

Control Techniques. 4 S2ttlement NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON OF ANALYSIS TECHNIQUES. S:: vere ' Accident ! NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENBER CONTAINMENT. NURE0/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. NUREG/CR-4173: CORSOR USER 'S MANUAL. Shipping Container

NURE0/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION 0F SHIPPING CONTAINERS FOR RADIDACTIVE MATERIALS.

I NURE0/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS.

,          Chert-Period Scram                            -

NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS. Coil j NURE0/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIDACTIVE WASTE DISPOSAL SITE AT SPEFFIELD, ILLINDIS. NURE0/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIDACTIVE WASTE DISPOSAL-SITES AT BARNWELL,SC AND RICHLAND,WA. Solute Transport NUREO/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW l AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. Cpcco-Time Correlation NUREG/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR ' POWER PLANTS UTILIZING REAL TIME ANEMOMETER STATISTICS. Cpociation l NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) (- Spont Fuel Rod NUREG/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT l GAS AND DRY AIR STORAGE ATMOSPHERES. , Stchility Test l NUREO/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE I BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. Stainless Steel NURE0/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report, April-June 1994. NURE0/CR-3945: FATIQUE CRACK OROWTH RATES OF LOW-CARBON AND STAINLESS i PIPING STEELS IN PWR ENVIRONMENT. l Stcndard Review Plan NUREG-0800 18.2 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To SRP Section 18,2, " Safety Parameter Display Bystem (SPDS)." l- NUREG-0800 18. 2A1 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY j ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To 97 l

Appcndix A To SRP Cocticn 1C.2, " Human Fcctero R:vicw Cuidolin30 For Tho Cofoty PcrcmetOr Dicplcy Cyntco (SPDD)." NUREG-OSOO RO5: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 5 To SRP Table Of Contents. l Ctarting Temperature NUREG/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task A: IEEE Flame Test. Ctation Blackout I' NUREG/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND PARTIAL LOSSES j' OF OFF-SITE POWER AT NUCLEAR POWER PLANTS. Cteam Generation NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. Cteam Generator NUREG-1108: RADI0 ACTIVITY TRANSPORT FOLLOWING STEAM GENERATOR TUBE RUPTURE. NUREG/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1994. NUREG/CR-4073: RESULTS OF THE SEMISCALE MOD-28 STEAM QENERATOR TUBE RUPTURE TEST SERIES. Stress Corrosion Cracking NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984. NUREG/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report, April-June 1984. Ctress Field l NUREG/CR-4036: STRUCTURAL GEDLOGY OF SOUTHEASTERN ILLINOIS AND VICINITY: NOREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. , Stress-State Variable NUREG/CR-4087: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION CHARACTERISTICS. ! Ctructural Geology NUREG/CR-4036: STRUCTURAL GEOLOGY OF SOUTHEASTERN ILLINOIS AND VICINITY, Culfate Lake < NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION , COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. Culfur NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE l CRACK GROWTH RATES IN LWR ENVIRONMENTS. Cuper System

NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June

> 30,1984.

Superposition Technique NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS.

Curve 111ance i NUREG/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS. Volume 2: Current Status of Biof ouling Surveillance And Control Techniques.

Survey NUREG/CR-3516: A SURVEY OF THE USES OF RADIDACTIVE MATERIALS IN LOUISIANA'S OFFSHORE WATERS.
Synthetic Liner i

i I N 1

    -                                           .m,. . . , __      _ _ _ . . . . . , _ , . . _ , , . . . . _ , _ , , , . _ . . , . _ _ _ _ _ . _ _ _ _ . . _ , _ _ , . . _ _ . . . , _ _ _ , _ . , . _ , . _ _

NURE3/CR- 4023: FIELD PERFORMANCE AS!E!SMENT OF CYNTHETIC LINERS FOR URANIUM TAILING 3 POND.A Status Rsport. Cgstems Analysis l ;MUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. 1 NUREG/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. TLD NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING l' METWORK. Progress Report, July-September 1984. TRAC-PD2 NUREG/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT. J' TRAC-PF1/ MOD 1-i NUREG/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS PRESSURIZER TEST YO5. TRAP-MELT

!                                NUREG/CR-4172: A USER'S GUIDE FOR MERGE.

TRUNC NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON OF ANALYSIS TECHNIGUES. l TRUST

NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON OF ANALYSIS TECHNIQUES.

i TWIST l NUREG/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUBASSEMBLY HEAT

TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS.

L Technetium i NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review l Of Data For Technetium. Technical Specification NUREG-1089: TECHNICAL SPECIFICATIONS FOR FERMI-2. Docket No. 50-341. j (Detroit Edison Company) NUREG-1104: TECHNICAL SPECIFICATIONS FOR WOLF CREEK GENERATING STATION, } UNIT 1. Docket No. 50-482.(Kansas Gas And Electric Company)

NUREG-1106: TECHNICAL SPECIFICATIONS FOR CATAWBA NUCLEAR STATION, UNIT

! 1. Docket No. 50-413.(Duke Power Company) NUREG-1113: TECHNICAL SPECIFICATIONS FOR BYRON STATION UNITS 1 AND 2. > , Docket Nos.50-454'And 50-455.(Commonwealth Edison Company) NUREG-1117: TECHNICAL BPECIFICATIONS FOR WATERFORD STEAM ELECTRIC i STATION UNIT 3. Docket No. 50-382.(Louisiana Power And Light Company) i Temperature i NUREG/CR-3945: FATIGUE CRACK GROWTH RATES OF LOW-CARBON AND STAINLESS i ~ PIPING STEELS IN PWR ENVIRONMENT. NUREG/CR-3954: HECTR ANALYSIS OF EGUIPPENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. ( NUREG/CR-4112 VO2: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Task B: Firestop Test Method. Test NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE i BRANCH TECHNfCAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. l NUREG/CR-4073: RESULTS OF THE SEMISCALE MOD-2B STEAM GENERATOR TUBE RUPTURE TEST SERIES. Tetroon NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of I Trajectories Concentration Patterns And MESODIF Model Calculations, i Thermal Cycling i NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. Thermal Shock NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL 99

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

PROGRE%3 REPORT FOR APRIL-SEPTEMBER 1984. Thorcc1/Hydraulico , NUREG-1108: RADIOACTIVITY TRANSPORT FOLLOWING STEAM QENERATOR TUBE j RUPTURE. : NUREO/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY l RESEARCH INFORMATION MEETING. NUREQ/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April-1-June 30,1984. NUREO/CR-3802: RELAP5 ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS. NUREG/CR-3804 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report. July-September 1984. NUREO/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS PRESSURIZER TEST YO5. NUREO/CR-3936: RELAP5 ASSESSMENT: CONCLUSIONS AND USER QUIDELINES. NUREO/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PONDS. NUREQ/CR-4172: A USER'S GUIDE FOR MERGE. Thermoluminescent Dosimeter NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1984. Time Averaging NUREO/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EGUATIONS FOR MULTIPHASE FLOW. Part One: Sy stem Without Internal Solid Structures. Title List NUREG-0540 VO6 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. November 1-30,1984. NUREG-0540 VO6 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1984. NUREG-0540 VO7 NO1: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. January 1-31,1985. Transport NUREG/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual P.rogress Report For April 1984-September 1984. NUREO/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) Tube Rupture NUREG-1108: RADIDACTIVITY TRANSPORT FOLLOWING STEAM GENERATOR TUBE RUPTURE. NUREQ/CR-4073: RESULTS OF THE SEMISCALE MOD-2B STEAM GENERATOR TUBE RUPTURE TEST SERIES. Tubing Inspection NUREQ/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM QENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. - Turbulence NUREQ/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR POWER PLANTS UTILIZING REAL TIME ANEMOMETER STATISTICS. Turbulent Mixing NUREQ/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUBASSEMBLY HEAT TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. Turbulent Transfer NUREG/CR-4157: A SCIENTIFIC CRITIQUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. Twelfth Water Reactor Safety Research NUREG/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO58 VO2: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREQ/CP-OO58 VO3: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. 100

l NURED/CP-OO58 VOS: PROCEEDINGS OF'THE TWELFTH WATER REACTOR SAFETY

       -RESEARCH INFORMATION MEETING.

NUREG/CP-OO58 VOS: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY l

        .RESEARCH INFORMATIDN MEETING.                                             I 4

NUREG/CP-OOSS VO6: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY I RESEARCH INFORMATION MEETING. 1 UHI NUREG/CR-3772: RELAPS ASSESSMENT: SEMISCALE SMALL BREAK TESTS S-UT-1,B-UT-2, S-UT-6,S-UT-7 AND S-UT-8. Ultimate Heat Sink NUREG/CR-4120: MATHEMATICAL HODELING'OF ULTIMATE HEAT SINK COOLING , PONDS.

Uncertainty Bound NUREG/CR-3688 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report.

NUREC/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT . JUDGMENT. Yo lume 2: Ap p e nd ic e s. !- Unintentional Releases , NUREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL l . ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. Unresolved Safety Issue A-17 ! NUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS i -AND SOURCES. Main Report And Appendices A And B. NUREG/CR-3922 VO2:' SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS l AND SOURCES. Appendices C And D Unresolved Safety Issue A-44 4 NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND PARTIAL LOSSES OF OFF-SITE POWER AT NUCLEAR POWER PLANTS. , Unresolved Safety Issue ! NUREG-0606 VO7 NO1: UNRESOLVED SAFETY ISSUES

SUMMARY

. Data As Of ! February 15, 1985.(Aqua Book) ' Upper Head Injection NUREG/CR-3772: RELAPS ASSESSMENT: SEMISCALE SMALL BREAK TESTS i S-UT-1,S-UT-2, S-UT-6 S-UT-7 AND S-UT-8. Uranium Fuel Cycle-( NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM-FUEL CYCLE.A Review ' Of Data For Technetium. Uranium Mill Tailings NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON j' 0F ANALYSIS TECHNIGUES. l NUREQ/CR-4023: FIELD PERFORMANCE ASSESSMENT OF SYNTHETIC LINERS FOR j URANIUM TAILINGS POND.A Status Report. i NUREC/CR-4087: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION CHARACTERISTICS. NUREG/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CONTROL TECHNIQUES FOR URANIUM MILL TAILINGS PILES. Uranium Mining NUREG/CR-3709: METHODS OF MINIMIZING GROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. Uranium Tailings Impoundment NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. Uranium NUREG/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report. April 1983 - March 1984. User Guidelines NUREG/CR-3936: RELAP5 ASSESSMENT: CONCLUSIONS AND USER GUIDELINES. User's Guide NUREG/CR-4172: A USER'S QUIDE FOR MERGE. 101

User'o Manual NURE9/CR-4173: CORSOR USER 'S MANUAL. i Volume Averaging' NUREG/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EGUATIONS FOR MULTIPHASE FLOW.Part One: System Without Internal Solid Structures. ; Waste Container l .NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADIDACTIVE WASTE . DISPOSAL SITES'AT BARNWELL,SC AND RICHLAND,WA. 1 Weste Disposal NURE9/CR-4061: LEACHATE PLUME MIGRATION DOWNORADIENT FROM URANIUM < TAILINGS DISPOSAL IN MINE STOPES." Waste Gas NURE0/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. Weste Package

              .NURE9/CR-2482 VO6: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1,1 -

National Weste Package Program. October 1983 - March 1984.

              -NUREG/CR-2482 VO7: REVIEW OF DDE WASTE PACKAGE PROGRAM. Subtask 1.1 -

National Weste Package Program April 1984 - September 1994. NUREG/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING.Guarterly Report, July-Septomber 1984. Waste

NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE l BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS.

l Water Chemistry NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OF HYDROGEN ! GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS. Water Flow , NUREG/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW ~ l AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. ! Welding j ,NUREO/CR-3019: RECOMMENDED WELDED CRITERIA FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIDACTIVE MATERIALS. Wet Deposition NUREO/CR-4157: A SCIENTIFIC CRITIGUE OF AVAILABLE MODELS FOR REAL-TIME l SIMULATIONS OF DISPERSION.

        - Wind Tunnel

!- NUREO/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR

                . POWER PLANTS UTILIZING REAL TIME ANEMOMETER STATISTICS.

Windscreen , 1 NUREQ/CR-4089: EVALUATION OF FIELD-TESTED FUCITIVE DUST CONTROL TECHNIGUES FOR URANIUM MILL TAILINGS PILES. r

Yellowcake NUREO/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND; DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report. April 1983 - March 1984. -

l- NUREO/CR-4100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT ( OF YELLOWCAKE EMISSIONS. l' Zircaloy NUREG/CR-3900 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH l PROGRAMS. Guarterly Progress Report, April-June 1984. NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. ? 102 4 _ _ _ . _ . . , . . _ _ _

1 l

     .NRC Originating Organization index (Staff Reports)

This index lists those NRC organizations that have published staff reports. The index is arranged alphabetically by major NRC organizations (e.g., program offices) and then by subsections of these (e.g., divisions, branches) where ap-propriate. 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. ADVISORY COMMITTEE (S) ACRS - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUREG-1105: REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM FOR FISCAL Y2ARS 1986 AND 1987. OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) REGION 1. OFFICE OF DIRECTOR NUREG-0837 VO4 NO3: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, July-September 1984. l EDO - OFFICE OF ADMINISTRATION SIVISION OF TECHNICAL INFORMATION & DOCUMENT CONTROL NUREG-0304 VO9 NO4: REGULATORY AND TECHNICAL REPORTS. Annual Compilation For 1984. NUREG-0540 VO6 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. November 1-20,1984. NUREG-0540 VO6 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1-31,1984. NUREG-0540 VO7 NO1: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. January 1-31,1985. NUREG-0544 RO2: A HANDBOOK OF ACRONYMS AND INITIALISMS. NUREG-0750 V2O 101: INDEXES TO NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JULY-SEPTEMBER 1984. NUREG-0750 V2O 102: INDEXES TO NUCLAER REGULATORY COMMISSION ISSUANCES FOR JULY-DECEMBER 1984. NUREG-0750 V2O NO4: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR OCTOBER 1984. Pages 1,055-1,435. NUREG-0750 V2O N05: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR l NOVEMBER 1984. Pages 1,437-1,572. NUREG-0750 V2O N06: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR DECEMBER 1984. Pages 1,573-1,706. NUREG-0750 V21 NO1: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JANUARY 1985. Pages 1-273. i NUREG-0910 RO1 S01: NRC COMPREHENSIVE RECORDS DISPOSITION SCHEDULE. NUREG-0910 RO1 SO2: NRC COMPREHENSIVE RECORDS DISPOSITION SCHEDULE. ws l

CIVICION OF RULE] AND RECORDS l NURE -0936 VO3 NO4: NRC REGULATORY AGENDA.CucrtOrly ' l Rcpcrt,0ct bar-D:cerber 1984. l DFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) DIRECTOR'S OFFICE, OFFICE OF INSPECTION AND ENFORCEMENT NUREG-0981 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR ACCIDENT. ENFORCEMENT STAFF NUREG-0940 VO3 NO4: ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED.Guarterly Progress Report,0ctober-December 1984. EMERGENCY PREPAREDNESS BRANCH NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12: SHORT-PERIOD SCRAMS AT BOILING-WATER REACTORS. DIVISION OF GA, SAFEQUARDS & INSP PROGRAMS (830103-850212) NUREG-OO40 VO8 NO4: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT. Guarterly Report,0ctober-December 1984. (White Book) SAFEGUARDS & MATERIALS PROGRAM DRANCH NUREG-1103: CONTAMINATED MEXICAN STEEL. Importation OF Steel Into The United States That Had Been Inadvertently Contaminated With Cobalt-60 As A Result OF Scrapping Of A Teletherapy Unit. OFFICE OF NUCLEAR MATERI AL SAFETY & SAFEGUARDS DIVISION OF FUEL CYCLE & MATERIAL SAFETY NUREG-1112: ENVIRONMENTAL ASSESSMENT FOR RENEWAL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM-368.(UNC Naval Products Division Of UNC Resources,Inc) NUREG-1130: ENVIRONMENTAL ASSESSMENT FOR RENEWAL AND CONSOLIDATION OF I* MATERIALS LICENSE NOS. SNM-362,SMB-405,08-OO566-05, 08-OO566-10,AND 08-00566-12. U. S. NUCLEAR REQULATORY COMMISSION COMMISSIONERS NUREG-0885 104: U. S. NUCLEAR REGULATORY COMMISSION POLICY AND PLANNING GUIDANCE 1985. NRC - NO DETAILED AFFILIATION GIVEN NUREG/CR-4173: CORSOR USER 'S MANUAL. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/01) 0FFICE OF NUCLEAR REQULATORY RESEARCH, DIRECTOR NUREG/CP-OO58 VO1: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO50 VO2: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO58 VO3: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREG/CP-OO58 VO4: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY 104

RESEARCH INFORMATION MEETING. . NUREG/CP-0058 VOS: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING. NUREW/CP-OOSS VO6: PROCEEDINGS OF THE TWELFTH WATER REACTOR SAFETY

RESEARCH INFORMATION MEETING.

DIVISION OF ACCIDENT EVALUATION l NUREG-1108: RADIOACTIVITY TRANSPORT FOLLOWING STEAM GENERATOR TURE , RUPTURE. ! DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429) NUREG-1115: CATEGORIZATION OF. REACTOR SAFETY ISSUES FROM A RISK PERSPECTIVE. DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST 840429) NUREC-0713 VO5: OCCUPATI0hAL RADI ATION EXPOSURE AT COMPERCI AL NUCLEAR POWER REACTORS - 1983 AhNUAL REPORT. , f EDD-RESOURCE MANAGEMENT

!    0FFICE OF RESOURCE MANAGEMENT, DIRECTOR                                                   !
!. NUREG-0325 RO7: U. S. NUCLEAR REGULATORY COMMISSION FUNCTIONAL l         ORGANIZATION CHARTS.

DIVISION OF BUDGET & ANALYSIS NUREG-OO20 VOS N12: LICENBED OPERATING REACTORS STATUS

SUMMARY

j REPORT. Data As Of November 30,1984.(Gray Book I) NUREG-OO20 VO9 N01: LICENBED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of December 31,1984.(Gray Book I). j NUREG-OO20 V09 NO2: LICENBED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of January 31,1985.(Gray Book I)

NUREG-0748 V04 N12: OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data As Of December 31,1984.(Orange Book)

NUREG-0748 VOS N01: OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data i As Of January 31,1985.(Orange Book) , NUREG-1100 V01: FY 1986 BUDGET ESTIMATES. OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/B0) l . OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR i NUREG-0800 18.2 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision O To ? SRP Section 18.2. " Safety Parameter Display System (SPDS)." i NUREG-OS00 18.2A1 RO: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY . ANALYSIS REPORTS FDR NUCLEAR POWER PLANTS. LWR Edition. Revision 0 To ! Appendix A To SRP Section 18.2, " Human Factors Review Guidelines For The Safety Parameter Display System (SPDS)." , NUREG-OSOO RO5: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 5 To SRP Table Of Contents. ! NUREG-1117: TECHNICAL SPECIFICATIONS FOR WATERFORD STEAM ELECTRIC ! STATION UNIT 3. Docket No. 50-382.(Louisiana Power And Light Company) DIVISION OF HUMAN FACTORS SAFETY NUREG-1021 RO1: OPERATOR LICENSING EXAMINER STANDARDS. . DIVISION OF LICENSING NUREG-0675 529: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF ! 106 l l - -

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

l 4 DIABLO CANYON NUCLEAR POWER PLANT,UNITO 1 AND 2.Dacket Nao. 50-275 ! And 50-323.(Pacific Gas And Electric Company) NUREG-0787 810: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF

WATERFORD STEAM ELECTRIC STATION, UNIT 3. Docket No. 50-382. ;

(Louisiana Power And Light Company)

!'                             NUREG-0797 807: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF    l COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2. Docket Nos.
;                                 50-445 And 50-446. (Texas Utilities Generating Company, et al)
 !                             NUREG-0797 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF 4                                  COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos.

50-445 And 50-446.(Texas Utilities Generating Company, et al) j NUREG-0797 809: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHF PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2. Docket Nos.

50-445 And 50-446.(Texas Utilities Generating Company)

NUREG-0798 805: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF EPStICO FERMI ATOMIC POER PLANT, UNIT NO. 2. Docket No. 50-341. I (Detroit Edison Company) i NUREG-0847 803: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And I 50-391.(Tennessee Valley Authority)

NUREG-0847 SO4: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF l WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391. (Tennessee Valley Authority) NUREG-0953 804: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF 4 CLINTON POWER STATION, UNIT 1. Docket No. 50-461.(Illinois Power ! Company,et al) i NUREG-0876 506: BAFETY EVALUATION REPORT RELATED TO THE OPERATION OF BYRON STATION, UNITS 1 AND 2. Docket Nos. 50-454 And 50-455. j (Commonwealth Edison Company) l NUREG-0881 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF j WOLF CREEK GEERATING STATION, UNIT 1. Doc ket No. 50-4e2.(Kansas Gas i And Electric Compang,et al) l NUREG-0887 SOS: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF i T E PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2. Docket Nos. 50-440 And ; 50-441.(Cleveland Electric Illuminating Company) j NUREG-0979 SO3: SAFETY EVALUATION REPORT RELATED TO THE FINAL DESIGN j APPROVAL OF THE GESSAR II BWR/6 NUCLEAR ISLAND DESIGN. Docket No. { 50-447. (General Electric Company) NUREG-1031 801: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF l MILLSTONE NUCLEAR POWER STATION, UNIT 3. Docket No. 50-423. j (Northeast Nuclear Energy Company) NUREG-1047: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF NINE MILE POINT NUCLEAR STATION, UNIT NO. 2. Docket No. 50-410. (Niagara i Mohawk Power Corporation,et al) ; NUREG-1048 801: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF l i HOPE CREEK GENERATING STATION. Docket No. 50-354.(Public Servic.e i i Electric and Gas Company) i NUREG-1073: FINAL ENVIROPMENTAL STATEMENT RELATED TO THE OPERATION OF

;                                 RIVER BEND STATION. Docket No. 50-458. (Gulf States Utilities And

Cajun Electric Power Cooperative)

NUREG-1087: FINAL ENVIROPMENTAL STATEMENT RELATED TO THE OPERATION OF }' VOOTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425.(Georgia Power Company) NUREG-1089: TECHNICAL SPECIFICATIONS FOR FERMI-2. Docket No. 50-341. (Detroit Edison Company) NUREG-1096: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE I GPERATING LICENSE FOR THE TRIGA TRAINING AND RESEARCH REACTOR AT l THE UNIVERSITY OF UTAH. Docket No. 50-407. (University of Utah) NUREG-1098: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF l; OPERATING LICENSE FOR THE RESEARCH REACTOR AT MANHATTAN COLLEGE. i = l l. wew--vy- p-m-m *m--wm.--+=v y-m--==-.e-e

Dock 3t No. 50-199. (Manhetten Calicso)

NUREG- 1104: TECHNICAL SPECIFICATIONS FOR WOLF CREEK GEERATING STATION, UNIT 1. Docket No. 50-482.(Kansas Gas And Electric Company) NUREG-1106: TECFMICAL SPECIFICATIONS FOR CATAWBA NUCLEAR STATION, UNIT

1. Docket No. 50-413.(Duke Power Company)

NUREG-1110: COMPARISON OF LICENSING ACTIVITIES FOR OPERATING PLANTS , DESIGE D BY BABCOCK 8: WILCOX. NUREG-1113: TECHNICAL SPECIFICATIONS FOR BYRON STATION UNITS 1 AND 2. Docket Nos.50-454 And 50-455.(Commonwealth Edison Company) DIVISION (F SAFETY TECHNOLOGY NUREG-0606 VO7 NO1: UNRESOLVED SAFETY ISSUES

SUMMARY

. Data As Of February 15, 1985.(Agua Book) NUREG-0933 802: A PRIORITIZATION OF GENERIC SAFETY ISSUES. i i i i l l 107

1 l l NRC Contract Sponsor Index (Contractor Reports) 1 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 organization is followed by the NUREG/CR number and title of the report (s) prepared by that orge.nization. If further information is needed, refer to the main citation by the NUREG/CR number. EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA ' DIRECTOR'S OFFICE NUREG/CR-2OOO VO3N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of December 1984. NUREG/CR-2OOO VO4 N1: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of January 1985. NUREC/CR-2OOO VO4 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of February 1985. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) DIVISION OF EMERGENCY PREPAREDNESS & ENGINEERING RESPONSE (POST 830103) NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF QE TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. NUREG/CR-3794: CLOSE0VT OF IE BULLETIN 80-25: OPERATING PROBLEMS WITH , TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. DIVISION OF GA, SAFEGUARDS & INSP PROGRAMS (830103-850212) NUREC/CR-3516: A SURVEY OF THE USES OF RADIDACTIVE MATERIALS IN LOUISIANA'S OFFSHORE WATERS. I 0FFICE OF NUCLEAR MATERIAL SAFETY & SAFECUARDS DIVISION OF FUEL CYCLE & MATERIAL SAFETY NURE0/CR-3019: R8ECOMMENDED WELDED CRITERI A FOR USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADI0 ACTIVE MATERIALS. NUREG /CR-3865: EVALUATION OF THE RADI0 ACTIVE INVENTORY IN,AND ESTIMATION OF ISOTOPIC RELEASE FROM, THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. DIVISION OF WASTE MANAGEMENT NUREG/CR-2482 VO6: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1. 1 - National Weste Package Program.0ctober 1983 - March 1984. NUREG/CR-2482 VO7: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1. 1 - National Waste Package Program April 1984 - September 1984. NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN 109

M.._a5e_A wE, - - A---+a-i--=4sh--a.4 Jan A J ii agda__m.___a.--J

                         -TM EdANCH TECHNICAL POSITION ON WA8TE FORMS Ale CONTAINER
                         'MATERIALO.

NUREG/CR-3851 V03: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL IfFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY BITE. PROJECTS. Report For April-June 1984. NUREG/CR-4023: FIELD PERFORMANCE ASSESSMENT OF SYNTHETIC LINERS FOR l URANIUM-TAILINGS POND.A Status Report. i NUREG/CR-4057: RADIOLOGICAL ASSESSMENT OF THE TOWN OF EDGEMONT. ' NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE ATf 8EFFIELD, ILLINOIS. NUREG/CR-4083: ANALYSES OF BOILS FROM THE LOW-LEVEL RADI0 ACTIVE WASTE DISPOSAL' SITES AT- BARNELL, SC AND RICHLAND, WA.

                 ~U.B.' NUCLEAR. REGULATORY COMMIS810N OFFICE OF PG. ICY EVALUATIONS NUREG/CR-3026: FEASIBILITY STUDY ON THE ACGUISTION OF LICENBEE EVENT DATA.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POGT 4/05/81) OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING ' EXTRAPOLATION TECHNIGUES. , DIVISION OF ACCIDENT EVALUATION NUREG/CR-2331 V04 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June 30,1984.

                      .NUREG/CR-3361: THE EFFECT OF WATER CE MISTRY ON THE RATES OF HYDROGEN GENCRATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS.

NUREG/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUSA88EMBLY HEAT TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMF3R8. NUREG/CR-3764: BWR-LTA8: A BOILING WATER REACTOR LONG-TERM ACCIDENT SIMULATION CODE. 1 NUREG/CR-3772: RELAPS ASSESSMENT: SEMISCALE SMALL BREAM TEST 8 8-UT-1,8-UT-2, 8-UT-6,8-UT-7 AND 8-UT-8. NUREG/CR-3802: RELAPS ASSESSMENT: GUANTITATIVE MEY PARAMETERS AND RUN TIME STATISTICS. NUREG/CR-3804 V03: PHYSIC 8 OF REACTOR BAFETY.Guarterly Repor4, July-Septomber 1984. , NUREG/CR-3810 V03: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report, July-September 1984. NUREG/CR-3816 V01: REACTOR BAF'ETY RESEARCH.Guarterly

 -.w                    . Report,JanuarU-March 1984.

NUREG/CR-3830 V02: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual

            '-      -      Progress' Report For April 1984-September 1984.                                                                    f NUREG/CR-3866: TRAC-PD2-INDEPENDENT ASSESSMENT,                                                                        '

9 ,_ g .NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN

ICE-CONDENSER CONTAINMENT.

NUREG/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT: NEPTUNUS PRE 88URIZER TEST YOS. NUREG/CR-3936: RELAPS ASSESSMENT: CONCLUSIONS AND USER GUIDELINES. NUREG/CR-3943: THE BWR PLAN ANALYZER. i NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE RESPONSES TO 8 ELECTED HYDROGEN BURNS IN AN ICE CONDENGER CONTAINMENT. '

NUREG/CR-3980 V02: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH; PROGRAMS.'Guarterly Progress Report April-June 1984.

NUREG/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EQUATIONS FOR MULTIPHABE FLOW. Part One: System Without Internal Solid Structures. ! 110 j -

a. ; .- -.

                                                    , , , . . . .    .       .  - , . . ~ .. . . + . .            . .,. ,,    m.,.   . + . . , . . . . , . , . , , ._
   ,                                                                                                                                                                 J.;                             ;

9- + .. NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN ',',p. LIGHT-WATER REACTrJRS. p ', / , ~ NUREG/CR-3999: ELECTRICALLY HEATED EX-RFACTOR PELLET-CLADDING 4 INTERACTION (PCI) SIMULATIONS UTILIZ1c0 IRRADIATED ZIRCALOY :,r' ' -

CLADDING. N'Y~ ~

 /-                    NUREQ/CR-4020: HMS: A COMPUTER PROGRAM FOR TRANSIENT,THREE-DIMENSIONAL                                                                        ;

1 y ' ..

L MIXING GASES. iyf , ~ NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. il

                                                                                                                                                                                           .*              c

.o NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM ,. , . . k" WITH DOWNWARD HEAT REMOVAL. . 'T . s... NUREG/CR-4073: RESULTS OF THE SEMISCALE MOD-2B STEAM GENERATOR TUBE 1? .M [ RUPTURE TEST SERIES. N. . ,'[

5. ' ;,.A 1 V NUREG/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT

     .                       L2-5). Final Comparison Report.                                                                                                                      -:.

? NUREG/CR-4172: A USER 'S CUIDE FOR MERGE. .,:. f NUREG/CR-4173: CORSOR USER 'S MANUAL.

l' ' s

 ?                 DIVISION OF FACILITY OPERATIONS                                                                                                                   i O "*

NUREG/CR-3519: HUMAN ERROR PRODABILITY ESTIMATION USING LICENSEE 4'.i; I EVENT REPORTS. 1,

                                                                                                                                                                                    &H..

( DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429) '

                                                                                                                                                                          %,.&.G

( d.? Y x NUREG/CR-3688 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING ., U " ~.}y.,. ([, EXPERT JUDGMENT. Volume 1: Main Report. NUREG/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING : . ., , Q' t. EXPERT JUDGMENT. Volume 2: Appendices. NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED (J%i &

   \                         DISPLAYS.

y NUREG/CR-3831: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT f{ - COMPONENTS. Interim Report - Diesel Generators, Batteries, Chargers 5,t

      .                      And Inverters.                                                                                                                           [*k ,. 1

e. ' NUREC/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY , . -; -g' ASSESSMENT EXPERIENCE TO THE U.S. NUCLEAR REGULATORY PROCESS. ,f<.. ~. . : . , n,.

4 x ' .; - - e S DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST 840429) i i ._ a;. y NUREG/CR-3488 VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of % .E.' .$. t Trajectories, Concentration Patterns And MESODIF Model Calculations. 9 ". '

      )                NUREG /CR-3709 : METHODS OF MINIMIZING GROUND-WATER CONTAMINATION FROM                                                                              .cc..I c                        IN SITU LEACH URANIUM MINING. Final Report.                                                                                                  5/. ' $         '

NUREG/CR-3752: EFFECTS OF HYDROLOCAC VARIABLES ON ROCK RIPRAP DESIGN M., Y, '

 '..                         FOR URANIUM TAILINGS IMPOUNDMENTS.                                                                                                      27 .i L                   NUREG/CR-3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR                                                                                 ,5 '.c                   q f-

f. HIGH-LEVEL WASTE PACKAGING. Quarterly Report, July-September 1984. ff?1-YJ, ? .f.

y NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A

. COMPARISON OF ANALYSIS TECHNIQUES. f.*>;:

NUREG/CR-3981: BIDACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. ? % W, k NUREG/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND f '[

E DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual fl V O

'M                           Progress Report, April 1983 - March 1984.                                                                                                i.fJW$

NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) = f NUREG/CR-4036: STRUCTURAL GEOLOGY OF SOUTHEASTERN ILLINOIS AND ; ;$../ .- VICINITY. "fW., g ' NUREG/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW (

.3 AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. (j % h

                                                                                                                                                                                          . .f 3

f NUREG /CR-4045: LITERATURE REVIEW ON AEROSOL-SAMPLING DEVICES FOR *.).l ~ y , N' RESPIRATORY FIELD STUDIES. [ NUREG / C R-4061 : LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM

       .                      TAILINGS DISPOSAL IN MINE STOPES.                                                                                                       [h.',i1                        W I.~                   NUREG/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT y                             SERVICE-WATER SYSTEMS. Vo l ume 2: Curren t Status of Biofouling                                                                        g, .g . - .,
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l l l NUREG/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR l I POWER PLANTS UTILIZING REAL TIME ANEMOMETER STATISTICS. NUREC /CR-4097: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION CHARACTERISTICS. NUREG/CR-4089: EVALUSTION OF FIELD-TESTED FUGITIVE DUST CONTROL TECHNIQUES FOR URANIUM MILL TAILINGS P ILES. l NUREG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION l COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. NUREQ/CR-4100: EVALUATION OF INSTRUMENTAL METHODS FOR THE MEASUREMENT l OF YELLOWCAKE EMISSIONS. l NUREQ /CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF i SOUTHWEETERN INDIANA. i NUREQ /C R-4120 : MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PONDS. NUREG/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. NUR EG /CR-4157 : A SCIENTIFIC CRITIQUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOf.E PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. DIVISION OF ENGINEERING TECHNOLOGY NUREG/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summar y Report. NUREG/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS, Volume 3: Doub le Ended Guillotine Break Indirectly Induced By Earthquakes. NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. NUREG/CR-3805 VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task II: Effects Of Ground Motion Characteristics On Structural Response Considering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. NUREG/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly Report, July-September 1984 NUREG/CR-3825 VO3-4: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS. Quarterly Report, April 1984 - September 1984. Volumes 3 and 4. NUREG/CR-3854: FABRICATION CRITERI A FOR SHIPPING CONTAINERS. NUREG/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report, April-June 1984. NUREG/CR-3945: FATIQUE CRACK GROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT NUREG/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Semiannual Progress Report For Period Ending June 30,1984. NUREG/CR-3978: TENSILE PROPERTIES OF IPPADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR ' FOURTH HSST IRRADIATION SERIES. NUREG/CR-4008: GENERAL EXTRAPOLATION MODU FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. NUREG/CR-4039: GAMMA-RAY CHARACTERIZATION OF THE TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. NUREG 'CR-4056 : PARTICULATE AND GAS RELEASE FROM LIGHT-WATER-REACTOR (LkA) FUEL RODS STORED IN INER T AND DRY AIR ATMOSPHERES. NUREG/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT CAS AND DRY AIR STORACE ATMOSPHERES. NUREG/CR-4082 VO1: DEGRADED PIPING PROGR AM - PHASE II Semiannual Report, March 1984 - September 1984. NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING 112

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J PROJECTS. Annual Summary Report - Fiscal Year 1984. i NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE _; CRACK GROWTH RATES IN LWR ENVIRONMENTS. - NUREG/CR-4123: SEISMIC FRACILITY OF REINFORCED CONCRETE STRUCTURES -- AND COMPONENTS FOR APPLICATION TO NUCLEAR FACILITIES. NUREC/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT

INSPECTION SYSTEM IMAGE ENHANCEMENT.

EDO-RESOURCE MANAGEMENT - 1 DIVISION OF BUDGET & ANALYSIS ~ NUREO/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE ' RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981. 9 NUREC/CR-3430 VO2: NUCLEAR POWER PLANT OPERATING EXPERIENCE - 1982. Annual Report. OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80) g ASSISTANT DIRECTOR FOR MATLS, CHEM. & ENVIRON TECHNOLOGY (POST 830201) j NUREQ/CR-4112 VO1: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST - PARAMETERS. Task A: IEEE Flame Tes t. d NUREG/CR-4112 VO2: INVESTICATION OF CABLE AND CABLE SYSTEM FIRE TEST 3 PARAMETERS. Task B: Firestop Test Method. y DIVISION OF HUMAN FACTORS SAFETY ; NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES ; IN NUCLEAR POWER PLANTS. Problems And Recommendations. DIVISION OF SYSTEMS INTEGRATION (POST B11005) NUREQ/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE CAS - SYSTEMS. 4 NUREG/CR-3659: A MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES 0F INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. [ NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review OF Data For Technetium. NUREG/CR-3950 VO1: FUEL PERFORMANCE ANNUAL REPORT FOR 1983. DIVISION OF SAFETY TECHNOLOGY NUREQ/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS " AND SOURCES. Main Report And Appendices A And B. = NUREC/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS _ AND SOURCES. Appendices C And D. 3 NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND PARTIAL LOSSES OF OFF-SITE POWER AT NUCLEAR POWER PLANTS. ; NUREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL , ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. NUREQ/CR-4068:

SUMMARY

OF HISTORICAL EXPERIENCE WITH RELEASES OF RADIDACTIVE MATERIALS FROM COMMERCIAL NUCLEAR POWER PLANTS IN THE UNITED STATES. y

4 i m 113 Ma Y

. _ _ _ . .. _ . . . . . . . . . ., .. ~ Contractor index This Index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports. If further information is needed, refer to the main citation by the NUREG/CR number. AERONAUTICAL RESEARCH ASSOCIATES OF PRINCETON NUREQ/CR-4157: A SCIENTIFIC CRITIQUE OF AVAILABLE MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. ARGONNE NATIONAL LABORATORY NUREQ/CR-3804 VO3: PHYSICS OF REACTOR SAFETY.Guarterly Report, July-Septomber 1984. _ NUREQ/CR-3980 VO2: LIGHT-WATER-REACTOR SAFETY FUEL SYSTEMS RESEARCH PROGRAMS. Guarterly Progress Report, April-June 1984. NUREQ/CR-3989: TIME- AND VOLUME-AVERAGED CONSERVATION EQUATIONS FOR MULTIPHASE FLOW. Part One: System Without Internal Solid Structures. NUREQ/CR-4120: MATHEMATICAL MODELING OF ULTIMATE HEAT SINK COOLING PONDS. ARIZONA, UNIV. OF, TUCSON, AZ NUREQ/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO SIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A ROCK FRACTURE SYSTEM. NUREQ/CR-4174: ROCK MASS SEALING - EXPERIMENTAL ASSESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 - May 1984. BATTELLE HUMAN AFFAIRS RESEARCH CENTERS NUREQ/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTENANCE PROCEDURES IN _ NUCLEAR POWER PLANTS. Problems And Recommendatiuns. BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES

NUREQ/CR-:3900 VO2: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEs/EL WASTE PACKAGING.Guarterly Report, July-September 1984. NUREG/CR-4 382 VO1: DEGRADED PIPING PROGRAM - PHASE II. Semiannual Report, March 1984 - September 1984. - NUR EQ/CR-4172: A USER'S GUIDE FOR MERGE. NUREQ/CR-4173: CORSOR USER 'S MANUAL. _ BATTELLE NEMORIAL INSTITUTE, PACIFIC NORTHWEST LABORATORIES NUREQ/CR-2850 VO3: POPULATION DOSE COMMITMENTS DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981. ' NUREQ/CR-3657: A MATHEMATICAL MODEL FOR ASSESSING THE UNCERTAINTIES OF _' INSTRUMENTATION MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. NUR EQ/CR-3709: METHODS OF MINIMIZING CROUND-WATER CONTAMINATION FROM IN SITU LEACH URANIUM MINING. Final Report. . NUR EQ/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK RIPRAP DESIGN _ FOR URANIUM TAILINGS IMPOUNDMENTS. NUREQ/CR-3810 VO3: REACTOR SAFETY RESEARCH PROGRAMS.Guarterly " Report, July-September 1984. - NUREO/CR-3825 VO3-4: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE _ MONITORING OF NUCLEAR PRESSURE VESSELS.Guarterly Report, April 1984 - w 115 5

                                                                                                            . .i

Saptember 1984.Volumos 3 cnd 4. NUREC/CR-3911 VO2: EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Guarterly Report. April-June 1984. NUREC/CR-3950 VO1: FUEL PERFORMANCE ANNUAL REPORT FOR 1983. NUR EG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES: A COMPARISON OF ANALYSIS TECHNIGUES. NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET-CLADDING INTERACTION (PCI) SIMULATIONS UTILIZING IRRADIATED ZIRCALOY CLADDING. NUREC/CR-4023: FIELD PERFORMANCE ASSESSMENT OF SYNTHETIC LINERS FOR URANIUM TAILINGS POND.A Status Report. NUREG/CR-4030: RADIONUCLIDE MIGRATION IN GROUND WATER.(Final Report) NUREC/CR-4057: RADIDLOGICAL ASSESSMENT OF THE TOWN OF EDGEMONT. NUR EG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. NUREG/CR-4070 VO2: BIVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS. Volume 2: Current Status of Biofouling Surveillance And Control Techniques. NUREG/CR-4087: MEASUREMENTS OF URANIUM MILL TAILINGS CONSOLIDATION r... ; CHARACTERISTICS. :

                                                                                               ,.s c . .;

NUREG/CR-4089: EVALUATION OF FIELD-TESTED FUGITIVE DUST CONTROL . . , .f ' TECHNIGUES FOR URANIUM NILL TAILINGS PILES. j. ' <J NUREG/CR-4100: EVALUATION DF INSTRUMENTAL METHODS FOR THE MEASUREMENT g. ,3 '. i 0F YELLOWCAKE EMISSIONS. ' 'i ."- NUREC/CR-4153: APPLICATIONS OF FOREIGN PROBABILISTIC SAFETY ASSESSMENT }jJ [ EXPERIENCE TO THE U.S. NUCLEAR REGULATORY PROCESS. / '3" BROOKHAVEN NATIONAL LABORATORY '...'..' NUREG/CR-2331 VO4 N2: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF f 2 f[j NUCLEAR REGULATORY RESEARCH.Guarterly Progress Report, April 1-June k*dil 30,1984. Y.; .'l.l.] : NUREQ/CR-2482 VO6: REVIEW DF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 - *74 National Waste Package Program. October 1983 - March 1984. NUREG/CR-2482 VO7: REVIEW OF DOE WASTE PACKAGE PROGRAM. Subtask 1.1 - iC)1 '3 . National Waste Package Program April 1984 - September 1984. $ .d ? '.- NUREC/CR-3026: FEASIBILITY STUDY ON THE ACGUISTION OF LICENSEE EVENT ((T g{ DATA. p; f., NUREQ/CR-3498: TWO-DIMENSIONAL MODELING OF INTRA-SUBASSEMBLY HEAT TRANSFER AND BUOYANCY-INDUCED FLOW REDISTRIBUTION IN LMFBRS. [[ NUREG/CR-3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT ..;' REPORTS. , C. ' NUREG/CR-3829: AN EVALUATION OF THE STABILITY TESTS RECOMMENDED IN THE p.

2 5 "

BRANCH TECHNICAL POSITION ON WASTE FORMS AND CONTAINER MATERIALS. NUREG/CR-3865: EVALUATION OF THE RADIOACTIVE INVENTORY IN,AND jf .,E ., . ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT il THE SHEFFIELD LOW-LEVEL WASTE BURIAL SITE. f (( j.'. NUREG/CR-3943: THE BWR PLAN ANALYZER. 1. ' '" ,, NUREG/CR-4067:

SUMMARY

OF BARRIER DEGRADATION EVENTS AND SMALL [.@Y [. k

ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS. ftg e
- NUREG/CR-4068

SUMMARY

OF HISTORICAL EXPERIENCE WITH RELEASES OF ,o +

, RADIOACTIVE MATERIALS FRDM COMMERCIAL NUCLEAR POWER PLANTS IN THE .'$ ~, ,13 UNITED STATES. .% s .v NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA ADJACENT TO THE LOW-LEVEL d l.'

R ADI0 ACTIVE WASTE DISPOSAL SITE AT SHEFFIELD, ILLINDIS. [ $:y '

NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RADI0 ACTIVE WASTE p([C ' ? DISPOSAL SITES AT BARNWELL,SC AND RICHLAND,WA. n' 7 } .q[ l NUREC/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION. COLORADO STATE UNIV., FT. COLLINS, CD -2*

                                                                                                            .." e NUREC/CR-4072: THE ESTIMATION OF ATMOSPHERIC DISPERSION AT NUCLEAR        ,A - r .t .

POWER PLANTS UTILIIING REAL TIME ANEMOMETER STATISTICS. $Q?.S r COLUMBIA UNIV., NEW YORK, NY 2,3 j g NUR EG/CR-4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRIBUTION !:N f , g:fg.-:;; ig,QA- 7 i

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CDEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE LAKE ENVIRONMENTS. COMMERCE, DEPT. OF, NATL. OCEANOGRAPHIC & ATMOSPHERIC ADMINISTRATION NUREQ/CR-34BO VO3: IDAHO FIELD EXPERIMENT 1981. Volume 3: Comparison Of Trajectories, Concentration Patterns And MESODIF Model Calculations. EARTH TECHNOLDQY CD. I NUREQ/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NUCLEAR POWER PLANTS: l ANALYSIS AND RANKING. EE&Q, INC. NUREQ/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE CAS SYSTEMS. NUREQ/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED DISPLAYS. NUREQ/CR-4041: SYSTEM ANALYSIS HANDBOOK. NUREQ/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARACTERISTICS USING EXTRAPOLATION TECHNIQUES. NUREC/CR-4056: PARTICULATE AND GAS RELEASE FROM LIGHT-WATER-REACTOR (LWR) FUEL RODS STORED IN INERT AND DRY AIR ATMOSPHERES. NUREQ/CR-4073: RESULTS OF THE SEMISCALE MOD-2B STEAM GENERATOR TUBE RUPTURE TEST SERIES. _ NUREQ/CR-4074: THE PERFORMANCE OF DEFECTED SPENT LWR FUEL RODS IN INERT " QAS AND DRY AIR STORAGE ATMOSPHERES. NUREC/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT L2-5). Final Comparison Report. FEDERAL EMERCENCY MANAGEMENT AGENCY

 ,   NUREG-0981 RO1: NRC/ FEMA OPERATIONAL RESPONSE PROCEDURES FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR ACCIDENT.

FINLAND, GOVT. OF NUREQ/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIQUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. GENERAL PHYSICS CORP. MUREG/CR-36B8 VO1: CENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREQ/CR-3688 VO2: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 2: Appendices. GEORGIA INSTITUTE OF TECHNOLDQY, ATLANTA, GA NUREQ/CR-3981: BI0 ACCUMULATION OF P-32 IN BLUEGILL AND CATFISH. ILLINOIS, STATE OF NUREQ/CR-4036: STRUCTURAL QEOLOGY OF SOUTHEASTERN ILLINOIS AND VICINITY. INDIANA GEOLOGICAL SURVEY NUREQ/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE MINES OF SOUTHWESTERN INDIANA. INHALATION TOXICOLOGY RESEARCH INSTITUTE NUR EQ/CR-3984: BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILLING EFFLUENTS. Annual Progress Report, April 1983 - March 1984. t JBF ASSOCIATES NUREQ/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. NUREQ/CR-3922 VO2: SURVEY AND EVALUATIDN OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. LAWRENCE LIVERMORE NATIONAL LABORATORY NUREQ/CR-3019: RECOMMENDED WELDED CRITERIA FOR UGE IN THE FABRICATION OF SHIPPING CONTAINERS FOR RADIOACTIVE MATERIALS. NUREC/CR-3660 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOP OF WESTINCHOUSE PWR PLANTS. Volume 3: Quillotine Break Indirectly Induced By Earthquakes. NUREQ/CR-3663 VO1: PROBABILITY OF PIPE FAILURE IN REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS. Volume 1: Summary Report. NUREC/CR-3663 VO3: PROBABILITY OF PIPE FAILURE IN THE REACTOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR PLANTS, Volume 3: Double Ended 117

Guillotino Brook Indiroctly Inducod By Earthquakon. NUREQ/CR-3854: FABRICATION CRITERIA FOR SHIPPING CONTAINERS. NUREO/CR-4123: SEISMIC FRAGILITY OF REINFORCED CONCRETE STRUCTURES AND COMPONENTS FOR APPLICATION TO NUCLEAR FACILITIES. LOS ALAMOS SCIENTIFIC LABORATORY NUR EQ/CR-3866: TRAC-PD2 INDEPENDENT ASSESSMENT. NUREC/CR-4020: HMS: A COMPUTER PROGRAM FOR TRANSIENT. THREE-DIMENSIONAL MIXING GASES. NUREQ/CR-4045: LITERATURE REVIEW ON AEROSOL-SAMPLING DEVICES FOR RESPIRATORY FIELD STUDIES. LOUISIANA, STATE OF NUREG/CR-3516: A SURVEY OF THE USES OF RADIOACTIVE MATERIALS IN LOUISIANA'S OFFSHORE WATERS. MATERIALS ENGINEERING ASSOCIATES, INC. NUREQ/CR-3945: FATIGUE CRACK GROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMENT. NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRONMENTS. NAVY, DEPT. OF, NAVAL RESEARCH LAB. NUREQ/CR-3990: CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN LIGHT-WATER REACTORS. CAK RIDGE NATIONAL LABORATORY NUREC/CR-2OOO VO3N12: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of December 1984. NUREQ/CR-2OOO VO4 N1: LICENSEE EVENT REPORT (LER) COMPILATION:For Month Of January 1985. NUREC/CR-2OOO VO4 N2: LICENSEE EVENT REPORT (LER) COMPILATION: For Month Of February 1985. NUREQ/CR-3430 VO2: NUCLEAR POWER PLANT OPERATING EXPERIENCE - 1982. Annual Report. NUREG/CR-3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. NUREQ/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review Of Data For Technetium. NUREG/CR-3744 VO2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM SEMIANNUAL PROGRESS REPORT FOR APRIL-SEPTEMBER 1984. NUREQ/CR-3764: BWR-LTAS: A BOILING WATER REACTOR LONG iERM ACCIDENT SIMULATION CODE. NUREC/CR-3830 VO2: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual Progress Report For April 1984-September 1984. NUREG/CR-3S31: THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT COMPONENTS. Interim Report - Diesel Generators, Batteries. Chargers And Inverters. NUREG/CR-3851 VO3: PROGRESS IN EVALUATION OF RADIONUCLIDE GEOCHEMICAL INCORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS. Report For April-June 1984. NUREG/0R-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Main Report And Appendices A And B. NUREQ/CR-3922 VO2: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES. Appendices C And D. NUREG/CR-3949 VO1: EDDY-CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Gemiannual Progress Report For Period Ending June 30,1984. NUR EQ/CR-3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR THE FOURTH HSST IRRADIATION SERIES. NUREQ/CR-3992: COLLECTION AND EVALUA1 ION OF COMPLETE AND PARTIAL LOSSES OF OFF-SITE POWER AT NUCLEAR POWER PLANTS. NUREG/CR-4039: CAMMA-RAY CHARACTERIZATION OF THE TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY. PARAMETER, INC. NUR EG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09: FAILURE OF GE TYPE AK-2 1 118

                                                                     ... _ _ ___ ^Z_'

CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. NURE9/CR-3794: CLOSEOUT OF IE BULLETIN 80-25: OPERATING PROBLEMS WITH TARGET ROCK SAFETY-RELIEF VALVES AT BWRS. RADIOLOGICAL ASSESSMENTS CORP. NUREG/CR-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE.A Review Of Data For Technetium.

*SANDIA NATIONAL LABORATORIES NUREQ/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE RATES OC HYDROGEN GENERATION FROM GALVANIZED STEEL CORROSION AT POST-LOCA CONDITIONS.

NUREQ/CR-3608 VO1: GENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 1: Main Report. NUREQ/CR-36B8 VO2: CENERATING HUMAN RELIABILITY ESTIMATES USING EXPERT JUDGMENT. Volume 2: Appendices. NUREQ/CR-3772: RELAP5 ASSESSMENT: SEMISCALE SMALL BREAK TESTS S-UT-1,S-UT-2, S-UT-6,S-UT-7 AND S-UT-8. NUREQ/CR-3802: RELAP5 ASSESSMENT: GUANTITATIVE KEY PARAMETERS AND RUN TIME STATISTICS. NUREQ/CR-3816 VO1: REACTOR SAFETY RESEARCH.Guarterly Report. January-March 1984. NUREQ/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE-CONDENSER CONTAINMENT. NUREQ/CR-3919: TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT:NEPTUNUS PRESSURIZER TEST YOS. NUREC/CR-3936: RELAPS ASSESSMENT: CONCLUSIONS AND USER GUIDELINES. NUREQ/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CONDENSER CONTAINMENT. NUREC/CR-4008: GENERAL EXTRAPOLATION MODEL FOR AN IMPORTANT CHEMICAL DOSE-RATE EFFECT. NUREG/CR-4055: THE D10 EXPERIMENT: COOLABILITY OF UO2 DEBRIS IN SODIUM WITH DOWNKARD HEAT REMOVAL. SIGMA RESEARCH, INC. NUREQ/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. STRUCTURAL MECHANICS ASSOCIATES NUREG/CR-3805 VO2: ENGINEERING CHARACTERIZATION OF GROUND MOTION. Task II: Effects Of Ground Motion Characteristics On Structural Response Considering Localized Structural Nonlinearities And Soil-Structure Interaction Effects. UNDERWRITERS LABORATORY, INC. NUREG/CR-4112 VO1: INVESTICATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Tas k A: IEEE Flame Test. NUREC/CR-4112 VO2: INVESTIGATION OF CABLE AND CABLE SYSTEM FIRE TEST PARAMETERS. Tas k B: Firestop Test Method. UNITED NUCLEAR CORP. (SUBS. OF UNC RESOURCES, INC.) NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DECOMMISSIONING PRF'ECTS. Annual Summary Report - Fiscal Year 1984. 119

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 fur-ther information is needed, refer to the main citation by the NUREG number. 27-047 Barnwell, SC, Chem-Nuclear Systems. Inc., NUREQ/CR-4083 STN-50-454 Byrcn Station, Unit 1. Commonwealth Edison Co. NUREC-0876 S06 STN-50-454 Byron Station, Unit 1, Commonwealth Edison Co. NUREG-1113 STN-50-455 By ron Station. Unit 2. Commonwealth Edison Co. NUREC-0876 S06 STN-50-455 By ron Station, Unit 2. Commonwealth Edison Co. NUREG-1113 50-413 Catawba Nuclear Station, Unit 1. Duke Power Co. NUREG-1106 50-461 Clinton Power Station, Unit 1, Illinois Power Co. NUREC-0853 SO4 50-445 Comanche Peak Steam Electric Station, Unit 1 NUREG-0797 S09 50-445 Comanche Peak Steam Electric Station, Unit 1 NUREC-0797 S07 50-445 Comanche Peak Steam Electric Station. Unit 1 NUREC-0797 S08 50-446 Comanche Peak Steam Electric Station, Unit 2 NUREC-0797 SO9 50-446 Comanche Peak Steam Electric Station. Unit 2 NUREC-0797 S07 50-446 Comanche Peak Steam Electric Station, Unit 2 NUREG-0797 S08 70-0398 Commerce, Dept. o f. Washington, DC, NUREC-1130 50-275 Diablo Canyon Nuclear Power Plant, Unit 1. Pacific Cas & Electric Co NUREC-0675 S29 50-323 Diablo Canyon Nuclear Power Plant. Unit 2, Pacific Cas & Electric Co NUREG-0675 S29 50-341 Enrico Fermi Atomic Power Plant, Unit 2, Detroit Edison Co. NUREO-0798 505 50-341 Enrico Fermi Atomic Power Plant. Unit 2, Detroit Edison Co. NUREC-1089 STN-50-447 CESSAR-238, Caneral Electric Co. NUREC-0979 SO3 50-354 Hope Creek Nuclear Station, Unit 1, Public Service Electric & Cas Co NUREG-1048 Sol 50-199 Hanhattan College Research Reactor NUREC-1098 50-423 Millstone Nuclear Power Station, Unit 3. Northeast Nuclear Energy Co NUREC-1031 801 50-410 Nine Mile Point Nuclear Station, Unit 2, Niagara Mohawk Power Corp. NUREC-1047 50-440 Perry Nuclear Power Plant. Unit 1. Cleveland Electric Illuminating C NUREC-0887 SOS 50-441 Perry Nuclear Power Plant. Unit 2, Cleveland Electric Illuninating C NUREC-0887 S05 27-048 Richland, WA, U. S. Ecology. Inc., NUREQ/CR-4083 50-459 River Bend Station, Unit 1. Culf States Utilities Co. NUREG-1073 27-039 Sheffield, IL, U. S. Ecology, Inc., NUREC/CR-3865 27-039 Sheffield. IL, U. S. Ecology. Inc., NUREC/CN-aO69 70-0371 UNC Naval Products. Inc., Uncanville, CT, NUREC-1112 50-407 Univ. of Utah Research Reactor NUREG-1096 50-424 Vogtle Electric Generating Plant, Coorgia Power Co. NUREG-10"~ 50-425 Vogtle Electric Generating Plant, Georgia Power Co. NUR"'M J87 50-302 Waterford Generating Station. Unit 3. Louisiana Power & Light Co. NOREC-0787 609 ERR 50-302 Waterford Generating Station. Unit 3. Louisiana Power & Light Co. NUREG-0787 810 50-302 Waterford Cenerating Station, Unit 3. Louisiana Power & Light Co. NUREG-1117 50-390 Watts Bar Nuclear Plant, Unit 1. Tennessee Valley Authority NUREG-0847 S03 50-390 Watts Bar Nuclear Plant, Unit 1. Tennessee Valley Authority NUREG-0847 SO4 50-391 Watts Bar Nuclear Plant, Unit 2. Tenns esee Valley Authority NUR20-0047 S03 50-391 Watts Bar Nuclear Plant. Unit 2. Tennessee Valley Authority NUREC-0847 SO4 121

STN-50-482 Walf Creek 0snerating Station. Kansas Gas & Electric NURE3-0981 805 STN-50-482 Wolf Creek Generating Station. Kansas Ces & Electric NUREG-1104 122

8uRC pores 333 U.S. NUCLEAR REGULATORY COMulSSsON t REPORT NUMsER #Am papW by FIOC. ### Vor A'o. af eayi 12 441 ~

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UNITED STATES rounrw ctass unit NUCLEAR REGULATORY COMMISSION e sr* WASHINGTON, D.C. 20555 u's*= " ' wasw o c. etRMir No, G 47 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 Main Citations I and Abstracts Contractor Report - 2 Number index _: 3 Personal Author index 4 Subject Index 5 NRC Originating Organization index NRC Contractor 6 sponsor index 7 Contractor index Licensed _ Facility 8 Index (}}

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