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Regulatory and Technical Reports.Compilation for Second Quarter 1986,April-June
	ML20205G571
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Issue date:	07/31/1986
From:	; NRC OFFICE OF ADMINISTRATION (ADM)
To:	;
References
	NUREG-0304, NUREG-0304-V11-N02, NUREG-304, NUREG-304-V11-N2, NUDOCS 8608190550
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,y NUREG-0304 l

Vol.11, No. 2 l

l Regulatory and Technical Reports l (Abstract Index Journal)

Compilation for Second Quarter 1986 April - June l

U.S. Nuclear Regulatory Commission Office of Administration

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Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.

I Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 i'

l NUREG-0304 Vol.11, No. 2 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for Second Quarter 1986 April - June D:ta Published: July 1986 Policy and Publications Management Branch Divi: ion of Technical information and Document Control Office of Administration U.S. Nuclear Regulatory Commission W:shington, D.C. 20566 g,/' A.i

CONTENTS Preface...........................................................................v i

Index i

Tab Main Citation and Abstracts........................................................... 1 S ta f' R epo rts....................................................................

Conferencs Proceedings...........................................................

Contract or Reports..................................................................

Contractor Report Number Index......................................................... 2 Pww al Au'.l ior i ndex................................................................. 3 S ubj*ect I ndex....................................................................... 4 NRC Originating Organization Index (Staff Reports)....................................... 5 NRC Con yact Sponsor Index (Contractor Reports)................

....... 6 Contractor in dex..................................................................... 7 Licensed Facilit y Index............................................................... 8 l

1 l

l HI 1

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:

Division 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. 20565 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 Organization index (Staff Reports)

NRC Contract Sponsor Index (Contractor Reports)

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

The bibliographic elements of the main citations are the following:

Staff Report NUREG-0508: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

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

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

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

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

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

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

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

v

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

ADD - addendum APP - appendix DRFT - draft ERR errata N

number R

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

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

NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-gene ated 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 comg)ilation are ar-is an NRC contractor-prepared report. The ranged by report number, where NUREG-bibliographic information (see Preface for XXXX is an NRC staff-originated report, detalis) is followed by a brief abstract of this NUREG/CP-XXXX is an NRC-sponsored report.

conference report, and NUREG/CR-XXXX NUREG-0020 V10 N03: LICENSED OPERATING REACTORS NUREG-0304 Vit Not: REGULATORY AND TECHNICAL STATUS

SUMMARY

REPORT. Data As Of February REPORTS. Compilation For First Quarter 1986, January-March.

28,1986.(Gray Book 1) ROSS P.A.; BEEBE,M.R. Division of Drvision of Technical information & Document Control April Budget & Analysis. Apnl 1986. 437pp. 8605210414. 36053:112.

1986. 73pp. 8605020454. 35808:348.

The OPERATING UNITS STATUS REPORT - LICENSED OP.

TNs joumal includes all formal reports in the NUREG series prepared by the NRC staff and contractors, as well as proceed-ERATING REACTORS provides data on the operation of nucle.

ings of conferences and workshops. The entries in the compila-at units as bmely and accurately as possible. This information is bon are indexed fa access by etle and abskact, conuteta collected by the Office of Resource Management from the report number, personal author, subject, NRC organization, con-Headquarters staff of NRC's Office of Inspection and Enforce-tractw, and licensed facW ment, from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for NUREG-0386 D04 R01: UNITED STATES NUCLEAR REGULA-commercial operating units, and errata from previously reported TORY COMMISSION STAFF PRACTICE AND PROCEDURE data; a compilation of detailed information on each unit, provid-DIGEST JULY 1,1972 SEPTEMBER 30,1985.

Office of the ed by NRC's Regional Offices, IE Headquarters and the utilities; Executive Legal Director.

Aspen Systems, Inc. April 1986.

and an appendix for miscellaneous information such as spent 762pp. 8605220018. 36063.033.

fuel storage capability, reactor-years of experience and non-TNs edition of the NRC Staff Practice and Procedure Digest power reactors in the U.S. It is hoped the report is helpful to all contains a digest of a number of Commission, Atomic Safety agencies and individuals interested in maintaining an awareness and Ucensing Appeal Board, and Atomic Safety and Ucensing of the U.S. energy situation as a whole.

Board decisions issued during the period July 1,1972 to Sep-tember 30,1985, interpreting the NRC's Rules of Practice in 10 NUREG-0020 V10 N04: LICENSED OPERATING REACTORS CFR Part 2. TNs edition replaces earlier editions and supple-STATUS

SUMMARY

REPORT. Data As Of March 31,1986.(Gray ments and includes appropriate changes reflecting the amend-Book I) ROSS,P.A.; BEEBE M.R. Division of Budget & Analysis.

ment to the Rules of Practice effective September 30,1985.

NUREG-0540 V00 N02: TITLE UST OF DOCUMENTS MADE N RE abstr c PUBLICLY AVAILABLE. February 1 28,1986.

DNision of Tech-nical Information & Document Control April 1986. 575pp.

NUREG4040 V10 N01: LICENSEE CONTRACTOR AND 8604170623. 3S619:329.

VENDOR INSPECTION STATUS REPORT. Quarterty This document is a monthly publication containing descrip-Repart. January 1986 - March 1986.(White Book)

Division of tions of information received and generated by the U.S. NRC.

CA, Vendor & Technical Training Center Programs (Post TNs informaHon includes (1) docketed matenal associated we 850212). May 1986.158pp. 8605300516. 36163:127.

civilian nuclear power plants and other uses of radioactive ma-TNs periodical covers the results of inspections _ _ m,,,,,

by W W nondocW maid recM and OmMM by m_m r

the NRC's Vendor Program Branch that have been distnbuted NRC pertinent to its role as a regulatory agency. The following t) the inspected organizations during the period from January indexes are included: Personal Author Index, Corporate Source 1986 through March 1986. Also included in this issue are the Index, Report Number index, and Cross Reference to Principal results of certain inspections performed pn,or to the January Documents index*

1986 that were not included in prev 60us issues of NUREG-0040.

NUREG-0540 V00 NO3: TnLE LIST OF DOCUMENTS MADE NUREG0090 V06 N04: REPORT TO CONGRESS ON ABNOR-PUBLICLY AVAILABLE. March 1 31,1986.

Division of Techni-MAL OCCURRENCES. October December 1985.

AEOD, Di-cal Information & Document Control. May 1986. 467pp.

rector's Office. May 1986. 47pp. 8606120901. 36435.001.

8605210558. 36049.001.

Section 208 of the Energy Reorganization Act of 1974 identi-See NUREG-0540,V08,N02 abstract.

fies an abnormal occurrence as an unschedufed incident or NUMEG-0540 V04 N04: TITLE UST OF DOCUMENTS MADE event wNch the Nuclear Regulatory Commission determines t PUBUCLY AVAILABLE. April 1 30, 1986.

Dhrsion of Technical be significant from the standpoint of public he,alth and safety information & Document Control. June 1986. 400pp, and requires a quarterfy report of such events to be made to 8606 Congress. TNs report covers the period October 1 to December S

RE

'V

.N02 abstract.

31,1985. During the report period, there were two abnormal oc-currences at the nuclear power plants licensed to operate. The NUREG-0675 S33: SAFETY EVALUATION REPORT RELATED first invoNed inoperable main steam isolation vanes and the TO THE OPERATION OF DIABLO CANYON NUCLEAR second invoNed management deficiencies at Fermi Nuclear POWER. UNITS 1 AND 2. Docket Nos.50-275 and 50-323.(Pacif-Power Station. There were three abnormal occurrences at the ic Gas And Electric Company) SHIERLING,H. Division of Pres-other NRC licensees. Two inveNed diagnostic medical misad-surtzed Water Reactor Ucensing. A (post 851125). May 1986.

mirustrations and the other invoNed a therapeutic medical mis-450pp. 8605210480. 36051:131.

admirwstration. There were no abnormal occurrences reported Supplement 33 to the Safety Evaluation Report for the Pacific by the Agreement States. The report also contains information Gas and Electric Company's Diablo Canyon Nuclear Power updating some previously reported abnormal occurrences.

Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been 1

2 Mein Citations and Abstracts prepared jointly by the Office of Nuclear Reactor Regulation its independent design assessment. The NRC staff concludes and the Region V Office of the U.S. Nuclear Regulatory Com-that the CPRT Program Plan provides an overall structure for miseson. The supplement reports on the status of the staff's in-addressing all existing issues and any future issues which may vestigation, inspection, and evaluation of allegations and con-be identified from further evaluations, and if property implement.

cems that have been identified to the NRC through March ed will provide important evidence of the design and construc-1986. The report includes a complete listing of all allegations tion quality of CPSES, and will identify any needed corrective and concoms, indicating the status of their resolution. The NRC action. The report identifies items to be addressed by the NRC staff concludes that the technical issues raised in the allege-staff during the implementation phase.

Uons with regard to the design, construction, and safe operation.

of Diablo Canyon Units 1 and 2 have been satisfactorily re-NUREG-0000 09.2.1 R4: STANDARD REVIEW PLAN FOR THE solved and no further action is required.

REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Edition. Revision 4 to Section 9.2.1, NUREG-0750 V23101: INDEXES TO NUCLEAR REGULATORY "Stanon Semce Water System?

Ne of Nuch Reacer COMMISSION ISSUANCES. January-March 1986.

Dmeson of Regulabon, Dkecer (post 85H 25). AprH 1986. U pp.

Technical Information & Document Control. June 1986. 51pp.

8607080183. 36918:143.

8607070473. 36899:156.

Digests and indexes for issuances of the Commission, the Revicion 4 to SRP Section 9.2.1 incorporates the rMWn Atomic Safety and Licensing Appeal Panel, the Atomic Safety CI Generic issue 36 " Loss of Service Water System," and and Licensing Board Panel, the Administrative Law Judge, the mis clanficabon 2 acceptance enteria 6 emphasize mat au Directors' Decessons, and the Denials of Petitions for Rulemak-requirenes of General Design CrHena 4 nwst sbu be met. Ms ing are presented.

revision does not incorporate any new guidelines or require-ments.

NUREG-0750 V23 N02: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR FEBRUARY 1986. Pages49-111.

Division NUREG-0000 09.2 2 R3: STANDARD REVIEW PLAN FOR THE of TechrucalInformation & Document Control. April 1986.78pp.

REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR 8605010547. 35803:047.

POWER PLANTS. LWR EdiuurtRevision 3 to Section 9.2.2,"Re-Legalissuances of the Commission, the Atomic Safety and U-actor Auxiliary Cooling Water Systems."

Office of Nuclear Re-censing Appeal Panel, the Atomic Safety and Licensing Board actor Regulation, Director (post 851125). June 1986.13pp.

Panel, the Administrative Law Judge, and NRC Program Offices.

8607080200.36918:154.

NUREG-0750 V23 N03: NUCLEAR REGULATORY COMMISSION Revision 3 to SRP Section 9.2.2 incorporates the resolution ISSUANCES FOR MARCH 1986. Pages 113-232.

Dnnaion of of Generic Issue 36 " Reactor Auxiliary Cooling Water Systems,"

Technical Information & Document Control. May 1986.129pp.

and minor clanficadon 6 acceptance crHeria 2 emphasize mat 8606120550. 36517:001.

all requirements of General Design Criteria 4 must still be met.

See NUREG-0750,V23,N02 abstract.

This revision does not incorporate any new guidelines or re.

NUREG 0750 V23 N04: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR APRIL 1986. Pages 233-464.

Division of NUREG-0637 V06 N04: NRC TLD DIRECT RADIATION MONI-Technical Information & Document Control. June 1986. 235pp.

TORING NETWORK. Progress Report, October-December 1965.

8607090195. 36931:014.

JANG,J.; RABATIN,K.; COHEN,L Region 1 Office of Director.

See NUREG-0750,V23.N02 abstract.

May 1986. 320pp. 8605270119. 36625:046.

NUREG 0781: SAFETY EVALUATION REPORT RELATED TO This report provides the status and results of the NRC Ther-THE OPERATION OF THE SOUTH TEXAS PROJECT, UNITS 1 moluminescent Dosimeter (TLD) Direct Radiation Monitoring AND 2. Docket Nos. 50-498 And 50-499.(Houston Lighting And Network, it presents the radiation levels measured in the vicinity Power Company)

Division of Pressurtzed Water Reactor Li-of NRC licensed facility sites throughout the country for the censing A (post 851125). April 1986. 746pp. 8605060475.

fourm quarter of 1985.

358713m NUREG-0837 V06 Not: NRC TLD DIRECT RADIATION MONI-This report provides the results of the NRC staff review of TORING NETWORK. Progress Report, January-March 1986.

Houston Lighting and Power Company's application for licenses JANG.J.; RABATIN,K.; COHEN,L Region 1, Office of Director.

to operate the South Texas Project. The facility consists of two J

986 50pp 70804 36936 pressurized water nuclear reactors located in Matagorda County, Texas. Subject to favorable resolution of the items dis-moluminescent Dosimeter (TLD) Direct Radiation Monitoring cussed in the Gafety Evaluation Report, the staff concludes that Network. It presents the radiation levels measured in the vicinity the facility can be operated by the applicant without endanger-of NRC licensed facility sites throughout the country for the first ing the health and safety of the public.

q d 1986.

NUREG-0797 S13: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-NUREG 0057 S10: SAFETY EVALUATION REPORT RELATED TRIC STATION UNITG l AND 2. Docket Nos. 50-445 And 50-TO THE OPERATION OF PALO VERDE NUCLEAR GENERAT.

446.(Texas Utilities Generating Company)

Division of Pressur-ING STATION, UNITS 1,2 And 3. Docket Nos. 50-528,50-529 ized Water Reactor Licensing. A (post 851125). May 1986.

And 50-530.(Artzona Public Service Company)

Division of 64pp. 8606160009. 36548.001.

Pressurtzed Water Reactor Licensing A (post 851125). Apnl Supplement No.13 to the Safety Evaluation Report related to 1986. 50pp. 8605130554. 35962:181.

operation of the Comanche Peak Steam Electric Station, Units Supplement No.10 to the Safety Evatuation Report for the I and 2 (NUREG-0797), has been prepared by the Office of Nu-application filed by Arizona Public Service Company, et af, for clear Reactor Regulation of the U.S. Nuclear Regulatory Com-licenses to operate the Palo Verde Nuclear Generating Station, mission. The facitety is located in Somervell County, Texas, ap-Units 1,2, and 3 (Docket Nos. STN 50-528/529/530) located in proximately 40 miles southwest of Fort Worth, Texas. This sup.

Maricopa County, Artzona, has been prepared by the Office of i

I piement presents the staff evaluation of the Comanche Peak Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Response Team Program Plan which was formulated by the Ap-Commission. The purpose of this supplement is to update the plicant to resolve various construction and desgn issues raised Safety Evaluation Report by providing an evaluation of (1) addi-t by the Atomic Safety and Licensing Board, allegers, intervenor tional information submitted by the licensees since Supplement I

Citizens Association for Sound Energy (CASE), NRC inspections No. 9 was issued and (2) other matters requiring staff review of various types, and Cygna Energy Services while conductog since Supplement No. 9 was issued.

Main Citations and Abstracts 3

NUREG-0008 904: SAFETY EVALUATION REPORT RELATED Lake Wylie. This supplement provides additional information TO THE OPERATION OF SEABROOK STATION UNITS 1 AND supporting the license for operation above 5% power and

2. Docket Nos.50-443 And 50-444. (Public Sennce Corr;riny of power ascensson to full-power operation for Unit 2.

New HampsNre,et al)

Dwieson of Pressurized Water Reactor NUREG 0079 S05: SAFETY EVALUATION REPORT RELATED -

A (post 851125). May 1986.100pp. 8606200186.

TO THE DESIGN APPROVAL OF THE GESSAR 11 BWR/6 NU-CLEAR ISLAND DESIGN. Docket No. 50-447. (General Electric Supplement No. 4 to the Safety Evaluation Report documents Company)

Dwisson of Boiling Water Reactor (BWR) Licensing.

the rmium of outstandng issues and confirmatory items in the Safety Evaluation Report and Supplements 1,2, and 3. The May 1986. 35pp. 860617m 36563268.

report relates to the application filed by the Public Service Com-Supplement 5 to the Safety Evaluation Report (SER) for the pony of New HampsNro for licennes to operate the Seebrook application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design Station, Units 1 and 2 in Rockingham County, New HampsNre.

(GESSAR fl) has been prepared by the Office of Nuclear Reac-NUREG-0000 R01: NUCLEAR POWER PLANT SEVERE ACCI-tor Regulation of the Nuclear Regulatory Cv...;.Ga This DENT RESEARCH PLAN. MARINO,G.P. Office of Nuclear Reg-report supplements the GESSAR 11 SER (NUREG-0979) issued ulatory Research, Director. Apnl 1986.129pp. 8604180344.

in April 1983 soir,,T,knq the results of the staff's safety review 35624:324.

of the GESSAR ll BWR/6 nuclear island design; Supplement 1 Over the past six years a maior research effort has been un-issued in July 1983; Supplement 2, issued in November 1984; derway by the NRC to develop an impruved understanding of Supplement 3, issued in January 1985; and Supplement 4, severe accidents and to provide a technical basis to support issued in July 1985. Subject to favorable resolution of the items regulatory decinons. A severe Accident Research Plan (SARP) discussed in the Final Design Approval FDA-1, Amendment No.

was issued in January 1963 as NUREG-0900. The purpose of

2) the staff concludes that the GESSAR 11 design satisfactonly this revision of the SARP is to desenbe current plans for the addresses the severe-accident concems described in the Com-wii.p;eivn and extension of tNs research. The plan is focused mission's Policy Statement on Severe Reactor Accidents Re-on technical issues and does not have as a purpose the justifi-garding Future Designs and Existing Plants, and that subject to cation of the need for the researc*i programs. TNs justification the approval of the balance-of-plant design, applications refer-is provided in the Long Range Research Plan (NUREG-1080).

encing GESSAR 11 can conform with provisions of the Atomic The discussion of the research to be accomplished during 1986 Energy Act of 1954, as amended and the regulations of the Nu-and 1987 is presented in tNs report in three general areas, clear Regulatory Commission.

each of which is duided into two subareas. The subareas of re-search are (1) risk reducten and evaluation, (2) severe accident NUREG-0000 R02: NUCLEAR REGULATORY LEGISLATION.

sequence analysis, (3) in-vessel melt progression and fission HOSPOOOR,S. Office of the Executwe Legal Director. Apnl product behavior, (4) conteenment loads and ex-vessel fission 1986. 400pp. 8605010553. 35804:112.

product behavior, (5) containment performance, and (6) equip.

NUREG-0980 is a compilation of nuclear regulatory legislation mont survivabihty. Chapter 3 of this report discusses for each and other relevant material through the 98th Congress, 2nd subarea the scope of research, the research accomplishments, Session. This compilation has been prepared for use as a re-to date, the outstanding issues, and the planned actrvities. The source document, which the NRC intends to update at the end plan reviews the 18 major NRC/IDCOR technical issues. The of every Congress. Contents of NUREG-0980 include: The plan also covers the eight mejor areas of uncertainty in source Atomic Energy Act of 1954, as amended; Energy Reorganiza-term analysis identified in NUREG-0956.

tion Act of 1974, as amended; Uranium Mill Tashngs Radiation Control Act of 1978; Low-Level Radioactive Waste Policy Act; NUREG-0040 V06 N01: ENFORCEMENT ACTIONS.SIGNIFICANT Nuclear Waste Policy Act of 1982; and NRC Authorization and ACTIONS RESOLVED.Ouarterfy Progress Report January-March Appropriations Acts. Other materials included are statutes and 1986.

Director's Office, Office of Inspection and Enforcement treaties on export hcensing, nuclear non-proliferation, and envi-May 1986. 368pp. 8606110698. 36418.001.

ronmental protection. Sections of Title 5, United States Code, This compilation summartzes segnrficant enforcement actons on Administrative Procedure are also included.

that have been resolved during one quarterfy period (January -

March 1986) and includes copies of letters, notices, and orders NUREG-0906 R02: U.S. NUCLEAR REGULATORY COMMISSION sent by the Nuclear Regulatory Comrmsson to licensees with HUMAN FACTORS PROGRAM PLAN.

Duision of Human respect to these enforcement actions and the ftensees' re.

Factors Technology (post 851125). Apnl 1986. 72pp.

sponses. It is anticipated that the information in this publication 8605020448. 35808:279.

will be widely disseminated to managers and employees en-This document is the Second Annual Revision to the NRC geged in actrwities hcensed by the NRC, in the interest of pro-Human Factors Program Plan. The first edition was published in moting public health and safety as well as common defense August 1983. Revision 1 was published in July of 1984. The and secunty.

purpose of the NRC Human Factors Program is to ensure that P*'

NUREG-0964 See: SAFETY EVALUATION REPORT RELATED nd ope ion f er er pla This men des TO THE OPERATION OF CATAWBA NUCLEAR the plans of the Office of Nuclear Reactor Regulation to ad-STATION, UNITS 1 AND 2. Docket Nos. 50-413 And 50-414.

dress high priority human factors concerns of importance to re-(Duke Power Company,et af)

Diviseon of Pressurtzed Water actor safety in FY 1986 and FY 1987. Revision 2 of the plan Reactor Ucensing - A (post 851125). May 1986. 25pp.

incorporates recent Commission decisions and policies bearing

- E**

s nts the Safety Evaluation Report a few exceptons, the principal changes from prior edition's re-(NUREG-0954) issuerf in Februa'Y 1983 by the Office of Nucle-flect a shift from developing new requirements to staff evalua-at Reactor Regulaton of the U S. Nuclear Regulatory Commis-tion of industry progress in resolving human factors issues. The sion with respect to the application filed by Duke Power Compa-plan addresses seven major program elements: (1) Training, (2) ny, North Carohna Murwcipal Power Agency Number 1, North Licensing Examinatens, (3) Procedures, (4) Man-Machine Inter.

Carohna Membership Corporation Saluda Rwer Electric Cooper-face (5) Staffing and Quahfications, (6) Management and Orga-atwo, Inc., and Piedmont Municipal Power Agency, as applicants nizaton, and (7) Human Performance.

and owners, for hcenses to operate the Catawba Nuclear Sta-tion, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectwo-NUREG-1021 R02: OPERATOR LICENSING EXAMINER STAND-ly). The facehty is located in York County, South Carohns, ap-ARDS. SZYMANSKI,T. Divison of Human Factors Technology proximately 96 km (6 rn) north of Rock Hill and adjacent to (post 851125). April 1986. 200pp. 8605210463. 36058.299.

l

4 Main Citations and Abstracts The Operator Licorning Examiner Standards provide policy brary of Congress. No classefied or other controlled information and guidance to NRC examiners and establish the procedures was prepared in 1985. The reports are drvided into two groups, and practces for exar.mung and licensing of applicants for NRC Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS operator hcenses pursuant to Part 55 of Title 10 of the CODE Reports on Generic Subjects. Part 1 contains ACRS Reports al-OF FEDERAL REGdLATIONS (10 CFR 55). They are intended phabetized by project name and within project name by chrono-to assist NRC examiners and facihty hcensees to understand logical order. Part 2 categorizes the reports by the most appro-the examinabon process better and tu provide for equitable and pnate genenc subject area and within subject area by chrono-consistent administratton of examinations to all applicants by logical order.

NRC examinen. These standards are not a subsbtute for the operator heeru,ing regulations and are subject to revision or NUREG-1137 S02: SAFETY EVALUATION REPORT RELATED other internal operator examination licensing policy changes. As TO THE OPERATION OF VOGTLE ELECTRIC GENERATING appropnate, these standards will be revised periodically to ac-PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425. (Geor.

U,JJ,04.te comments and reflect new information or experi-gia Power Company,et al)

Division of Pressurtzed Water Reac-ence.

ter Licensing A (post 851125). May 1986.135pp. 8606190590.

3662 M 71.

NUREG-1038 S03: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF SHEARON HARRIS NUCLEAR in June 1985, the staff of the Nuclear Regulatory Commission POWER PLANT, UNIT 1. Docket No.50-400. (Carolina Power And issued its Safety Evaluation Report (NUREG-1137) regarding Light Company And North Carolina Eastern Muncipal Power the apphcation of Georgia Power Company, Municipal Electric Agency)

Division of Pressurized Water Reactor Licensing A Authonty of Georgia, Oglethorpe Power Corporation, and City of (post 851125). May 1986.123pp. 8605210455. 36051:001.

Dalton, Georgia, for a license to operate the Vogtle Electric Supplement No. 3 to the Safety Evaluation Report for the ap-Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 425).

plicaten filed by Carolina Power and Light Company, et al., for Supplement 1 to NUREG-1137 was issued by the staff in Octo-hcense to operate the Shearon Harris Nuclear Power Plant, Unit ber N. The facihty is located in Burke County, Gewgia, ap No.1, located in Wake County, North Carolina, has been pre-proximately 26 miles south-southeast of Augusta, Georgia, and pared by the Office of Nuclear Reactor Regulation of the Nucle-on the Savannah River. This second supplement to NUREG.

ar Regulatory Comrrussion. The purpose of this supplement is to 1137 provides recent infamauon regardng resolubon of some update the Safety Evaluation Report of (1) addibonal informa-of me open and conhrmatay Hems met remained umosoked at tion subtr9ted by the applicants since Supplement No. 2 was the time the Safety Evaluation Report was issued. This supple-issued, and (2) matters that the staff had under *eview when ment also discusses some new open and confirmatory items.

Supplement No. 2 was issued.

NUREG 1152: MILLSTONE 3 RISK EVALUATION REPT:AN NUREG-1044 806: SAFETY EVALUATION REPORT RELATED OVERALL REVIEW AND EVALUATION OF THE MILLSTONE TO THE OPERATION OF HOPE CREEK GENERATING UNIT 3 PROBABILISTIC SAFETY STUDY, KELLY,G.;

STATION. Docket No.50-354. (Pubhc Service Electnc And Gas BARRETT,R.: BUSLIK,A. Division of Safety Review & Oversight Company.Atlante City Electric Company).

Division of Boiling (post 851125). June 1986.140pp. 8607020205. 36854:128.

Water Reactor (BWR) Licensing. Apnl 1986. 216pp.

In 1981, the U.S. Nuclear Regulatory Commisson (NRC) re-8604250242. 35733 270.

quested Northeast Ublities to perform a design-specific probabi-Supplement No. 5 to the Safety Evaluation Report on the ap-hstic safety study (PSS) for Millstone Nuciear Power Station, plication filed by Public Service Electnc and Gas Company on Unit No. 3 (Millstone 3) In 1983, Northeast Utdibos submitted its own behalf as co-owner and as agent for the other co-the Millstone 3 Probabihetic Safety Study (PSS) for rev6ew by owner, the Atlante City Electric Company, for a license to oper-the NRC staff. The NRC staff prepared the Mdistone 3 Risk ate Hope Creek Generating Station has been prepared by the Evaluaton Report, which discusses the findings regarding the Office of Nuclear Reactor Regutabon of the U.S. Nuclear Regu-PSS. The PSS eshmates that the mean annual core damage latory Comtrussion. The facihty is located in Lower Alloways frequency due to intomal and extemal events is 5x10( 5) and Creek Township in Salem County, New Jersey. This supplement 2x10(-5), respectively. The NRC staff's Risk 'Ivaluation Report reports the status of certain items that had not been resolved at estimates that the mean annual core damage frequency is the bme of the publication of the Safety Evaluation Report.

about 2x10(4) for intomal events and hos between 1x10(-5)

NUREG-1057 S01: SAFETY EVALUATION REPORT RELATED and 2x10(-4) for extemal events. The NRC staff estimates that station blackout dominates intemal and extemal event core TO THE OPERATION OF BEAVER VALLEY POWER STATION UNIT

2. Docket No.

50-412.(Duquesne Light damage frequencies. The staff recommends that Northeast Utih-Company.et al)

Divison of Pressunzed Water Reactor Licens.

Ms pehm an egineedng analysis on upgradng the desel A (post 851125). May 1986. 94pp. 8606110036.

generator tube od cooler anchorage syMem and on aMng a ing 36421:252 manuaHy operated, AQndependent containment spray system.

Supplement No.1 to the Safety Evaluation Report for the ap-Tre staff also recommends that Northeast Utilities prepare two picaton filed by Duquesne Light Company, et al., for heense to emergency procedures (loss of room cooling and relay chatter operate the Beaver Vallav Power Staten, Unit 2 i ocket No.

d's to an earthquake) to help reduce uncMainks.

D 50-412), located in Beaver County, Pennsytvania, tas been pre-NUREG 1154: ACCURACY AND DETECTION LIMITS FOR BIO-pared by the Office of Nuclear Reactor Regutabon of the Nucle-ASSAY MEASUREMENTS IN RADIATION PROTECTION -

ar Regulatory Commission. The purpose of this supplement is to STATISTICAL CONSIDERATIONS. BRODSKY,A. Division of update the Safety Evaluation Report of (1) additional informa-Radiation Programs & Earth Sciences (post 840429). April 198C.

ton submitted by the apphcants since the Safety Evaluation 99pp. 8605130560. 35960.291.

Report was issued, and (2) matters that the staff had under This report presents statisteal concepts and formulas for de-review when the Safety Evaluaton Report was issued fining minimum detectable amount (MDA), b6as and precision of NUREG 1125 V07: A COMPILATION OF REPORTS OF THE AD-sample analytcal measurements of radioactivity for radiobcas.

VISORY COMMITTEE ON REACTOR SAFEGUARDS,1985.

say purposes. The defined statisteal quantities and accuracy ACRS Advisory Committee on Reactor Safeguards. May 1986.

cntena were developed for use in standard performance criteria 238pp. 8606200289. 36651:001, for radiobicassay, but are also useful in intrataboratory quahty This compilation contains 63 ACRS reports submitted to the assurance programs. This report also includes a hterature Cynmisson or to the Executive Director for Operatens dunng review and anatysis of accuracy recommendatons of national calendar year 1985. All reports have been made available to the and intemabonal scientific orgaruzations for radiabon or radioac-pubhc through the NRC Pubhc Document Room and the U S. Li-tivity measurements used for radiation protecton purposes.

Main Citations and Abstracts 5

Computer programs are also included for calculatog the prob-facihty. On March 13,1966, a Model 48X cyhnder was overfilled abshtes of pesoing or fasting multpie analybcal tests for d.fteront durmg a special one-time draining procedure; however, no re-acceptable ranges of bias and precisen, loose of UF6 occurrei An Augmented inveengetion Team in.

vestigated this second incident. TNs report, NUREG-1179, NUREG 1100: INTERNATIONAL COOPERATION DURING RADI-Volume 2, presents the fmdings made by the Augmented inves-OLOGICAL EMERGENCIES. NRC Program Guedence For The tigation Team of the March 13 incident and the report of the de-Provision Of Technical Advice To Foreign Counterpart Organiza-tions. SENSENEY,RS. Office of Intemational Programs, Direc-tailed metaNurgical exammaten conducted by BaMe Columbus tot. Apnl 1986. 34pp. 8605150364. 35994:102.

Div6een of the cylinder damaged on January 4,1986.

TNs report dennes the scope, application and hmits of the NUREG 1101: TECHNICAL SPECIFICATIONS FOR PALO VERDE tec c operaten Nuclear Regulatory Comrmesen NUCLEAR GENERATING STATION, UNIT 2. Docket No. 50-(NRC) would prov6de, upon request, to a foreign regulatory 52g.(Arizona Public Sennce Company)

Divieson of Pressurized agency in a nuclear emergency. It outhnes the basis for such Water Reactor Licensing. B (post 85H25b AprW M 517pp.

cooperation, offers a model wntten agreement, and descnbos 8605MM 35979:062.

recent cases of NRC assistance, it also identifies non-NRC The Palo Verde Urwt 2 Technical Specifications were pro-sources of emergency advisory assistance available to foreign pared by the U.S. Nuclear Regulatory Commission to set forth organizat6ons.

the hmits, operating conditons, and other requirements apphca-NUREG-1146: DRAFT ENVIRONMENTAL STATEMENT FOR DE-ble to a nuclear reactor facihty as set forth in Secton 50.36 of COMMISSIONING HUMBOLDT BAY POWER PLANT, UNIT 10 CFR Part 50 for the protection of the health and safety of

3. Docket No. 50133.(Pacific Gas And Electric Company)

As-the pubHc.

sistant Drector for Techrwcal Support (PWR-B). Apnl 1986.

54pp. 8604290114. 35756:256.

NUREG 1143: NONRADIOLOGICAL GROUNDWATER OUALITY TNs Draft Environmental Statement contains the assessment AT LOW-LEVEL RADIOACTIVE WACTE DISPOSAL SITES.

of the environmental impact associated with decommissionmg GOODE D.J. Division of Weste Management. April 1986.242pp.

the Humboldt Bay Power Plant Unit 3 pursuant to the National 8605270414. 36091:000.

Environmental Policy Act of 1969 (NEPA) and Trtle 10 of the The NRC is investigating appropriate regulatory opuons for Code of Federal Regulabons, Part 51, as amended, of the Nu-disposal of low-level radoective weste containing nonrediologi-clear Regulatory Commiseson regulations. The proposed decom-cat hazardous constituents, as defined by EPA regulations.

rmoeioning would invo ve safe storage of the facihty for about 30 Standard EPA / RCRA procedures to determine hazardous or-years, after which the residual radcactivity would be removed genics, metals, indicator parameters, and general water quehty so that the facihty would be at levels of radioactivity accep'able are apphed to samples from youndwater monitoring wells at for rolesse of the facihty to unrestricted access, two commercial low-level radioactive waste disposal sites. At NUREG 1175: NRC SAFETY RESEARCH IN SUPPORT OF REG-the Sheffield, il site (non-operating) several typical organic sol.

ULATION Selected Highhghts.

Office of Nuclear Regulatory vents are identified in elevated concentrations in oneite wells Research. Director. May 1966. 51pp. 6605220009. 36073.018.

and in an offsite area exNbieng elevated tritium concentrations.

The report presents selected Nghlights of how research has At the Bamwell, SC site (operating), only very low concentra-contnbuted to the regulatory effort. It explains the research role tions of three organics are found in wells adjacent to disposal of the NRC and nuclear safety research contnbutons in the urwts. Hydrocarbons associated with petroleum products are de-areas of; pressure vessel integrity, piping, small-and large-tected at both sites. Hazardous constituents associated with break loss-of-coolant accidents, hydrogen and containment, previously identified maior LLW mixed weste streams, toluene, source term analysis, seismic hazards and high-level waste xylene, chromium, and lead are at or below detecton lirnits or at management. The report also provides a summary of current background levels in all samples. Rev6ew of previously collected and future research directions in support of regulation.

data also supports the conclusion that organic solvents are the

'*"'*d*' 9' **"'* *I"*"'" *

~

NUREG-1177: SAFETY EVALUATION REPORT RELATED TO THE RESTART OF DAVIS-BESSE NUCLEAR POWER E0**

STATION UNIT 1 FOLLOWING THE EVENT OF JUNE NUREG-1166: TECHNICAL SPECIFICATIONS FOR HOPE 9,1965.Dochet No. 50-346-(Toledo Edison Company) DE CREEK GENERATING STATION. Docket No. 50-354.(Public AGAZIO,A.W. Office of Nuclear Reactor Regulation, Drector Service Electric And Gas Company)

Division of Boihng Water (post 651125). June 1966. 200pp. 6607070490. 36660.062.

Reactor (BWR) Licensing. April 1966. 530pp. 6604260294 On June 9,1965, the Davis-Besse Nuclear Power Station ex-35752.016 perienced a partialloss of main feedwater wNie at 90% power.

The Hope Creek Generating Station Technical Specificatens Following a reactor trip, other malfunctions and operator errors were prepared by the U.S. Nuclear Regulatory Commission to led to a total loss of feedwater for a short period. Before opera-set forth the hmets, operating conditions, and other requirements tors were able to restore feedwater, both steam generators apph @ to a numer reecW W as ed fath in Sen boiled dry. This report preson'.2 *e staff a evaluation of the cor-50.36 of 10 CFR Part 50 for the protection of the health and rective act one taken by the bcensee to prevent recurrence and

- Y O' to improve overall performance of Davis-Bosse wtth respect to safety. The Safety Evaluabon supports the restart of the facehty.

NUREG 1166: THE AUBURN STEEL COMPANY RADIOACTIVE NUREG 1179 V02: RUPTURE OF MODEL 46Y UF6 CYLINDER CONTAMINATION INCIDENT.

Office of State Programs, Di-AND RELEASE OF URANIUM HEXAFLUORIDE.Cyfinder rector. CABASINO L; KELLY,R; et al. New York, State of. April Overfill. March 12 13,1986. Investigation Of A Feeled UF6 Ship-1966. 32pp. 6605160200. 36007:264.

ping Cyhnder, SMITH,R.D.; CAIN.C.L; LONG.J.T.: et af. NRC.

On February 21,1963, workers at the Aubum Steel Company, No Detailed Affiliation Given. June 1986. 250pp. 6606260195.

Aubum, New York discovered that about 120 tons of steel 36764:200.

poured that day had become contaminated wtth (60)Co. In addi-NUREG 1179, Volume 1, reported on the rupture of a Model hon to the steel, the air clean 6ng system and portions of the mell 48Y uranium hexafluoride (UF6) cylinder and the subsequent re-used in caebng the steel were contaminated. Approximately 25 lease of UF6. At the time of publ6 cation, a detailed metallurg6 cal cunes of (60)Co were involved. Decontamination and disposal enarrunabon of the damaged cyhnder was under way and results of the contaminat6on cost in arcoes of $2.200,000. This report were not available. Subsequent to the pubhcaton of Volume 1, details the discovery of the contemnation, decontaminaten of a second incident occurred at the Sequoyah Fuels Corporaten the plant and disposal of the contaminat>on.

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

i i

t i

G ERbt CllSUORS GRd AheWSS$8 NMRSS 1191: TECHNICAL SPECIFICATIONS FOR CATAWBA pebaues of members, operemon, and results of the NRC Dev6s-NUCLEAR STATION UNITS 1 AND 2.Dochet Nos. 50-413 And Besse HT, and the use to wNch as report was put by the reguis-50 414.(Duke Power Company)

DMelon of Pressurised Water tory staff.

i j-Reactor uneneing A (post 061126). May ISOS. Stepp.

9006300475. 38100 067.

NURES 1800: PROGRAM PLAN FOR ENVIRONMENTAL QUAU-The Catawba Nuclear Stenon. Units 1 and 2. Technical Speck FICATION OF MECHANICAL AND DYNAMIC (INCLUDING Incetone were prepared by the U.S. Nuclear Reguistory Com-SEISMIC) OUAUFICATION OF MECHANICAL AND ELECTRI-i minelon to est forth the Ilmits, opereung comBuona, and other CAL EQUIPMENT PROGRAM (EDQP). WEiOENHAMER G. DM.

l l

reclurements appuombio to a nuclear reactor facely as est forth sion of Engineering Technology. June 1988.30pp.9007000100.

I j

in Secuan 50.38 of 10 CFR 80 for the protocelon of me heelsh 30014:332.

and esfoty of the puhuc.

This report doecethos the eNort for FY 1000 and FY 1987 to nUREs 1100: AN pdVESTIGATION OF THE CONTROUTORS be performed for each of to tashe mehing up this equipment TO WRONG UNIT OR WRONG TRAIN EVENTS. RAMEY.

Iluogheellon program. The resulle of ins resserch are incomied SMITH,A.; PERSINNO.D. DMeion of Human Factors Technology to provide the technical basis for resoMng uncertaintles in exist.

(post 061125). Apre 1908. 37pp. 0006100013. 30034:20s, ing equipment gumallommon siendants. In ademon, the results Wrong unit / wrong train events result when acuano are per.

are conMbueng to the resolullon of solely issues G8-23. GI-47 formed on the wrong train of systems with redundant keins or and USI-A44, used, " Reactor Coolant Pump Seal Failure,"

on the wrong unit of a muluunit facely. This type of human

" Failure of HPCI Steam Une Wilhout teolomon," and "Stemon error was invesegated at 10 plant enes through irHispeh inter-mad =4" respecevely. Also, resserch eNort is bein0 directed at views with krP plant personnel and through plant provkang informegen on the behestor of containment lealeton wouothroughs. The purpose of the interviews was to determine valves under severe aceklont environments ANhough the re-the conertbutors to wrong unit / wrong train human errors in order suns of the laser resserch we not contreute to resoMng uncer.

to develop a strategy to reduce the incklence of such events.

tainues in spangAceben elendants, it has proven cost eNeceve to Thoes factors that conethuted to wrong unit / wrong train errors obtain the informenon under INe propam, are idenuAed and are renhed acconang to their releuve inch dance of involvement. Observatone perunent to wrong unit /

81URES/CP 4007 VS1: 1 ROCEEDINGS OF THE SECOND IAEA wrong train events and recommendeuona for NRC acuan to SPECIAUSTS' MEETING ON SUSCRITICAL CRACK reduce the incidence or wrong unit / wrong train events are GROWTH. Sessions i And fl, Held At Sendel, Japan,Mer 15-given.

17,1006. CULLEN,W.H. Melonels Engineering A-Inc.

IIUREG 1190: RELEASE OF UFO FROM A RUPTURED MODEL A 'N 1908,400pp. - MEA 2000. 38000m P

40Y CYUNDER AT SEQUOYAH FUELS CORPORATION TNs report la a compdenen of the papers wNch wem pmeent.

FACluTY.Leesons-Leemed Report.

Leesons Leemed Group.

ed at the Second IAEA Spooletets' Meeung on Suborecel June 1906. 72pp. 8007000100. 30030:192.

Crack Growth, held at the Ministry of Trade and Commerce, in The uranium hexanuortdo (UF6) rolesse of January 4,1906 at Sendel, Japan, on May 15 17, 1986. TNs mesung took place the Sequoyeh Fuels Corporeuen factly hee been rev6 owed by a four years aRet the Aret moeung, held on May 1315,1901, in NRC Leesons-Leemed Group. A Model 40Y cyander conleining Freiburg, Germany. All contribuunne address the leeue of subart-UF6 ruptured upon being heated eher it was grossly overnsed scal crack growth retos and prossenes in etnole and sNoys for The UF6 rolessed upon rupture of the cyunder rgacted wem ar-nuclear reactor preneure veneele, piping and reactor intemels home moteture to produce hydronuorte acid (HF) and uranyl flu.

TNe sympoelum was dMded into four eseeinns, deseng with: (a) oride (UO2F2). One trutvklusi caed from signoeure to airtome test methods and interleboratory comparison test programs, (b)

HF and esveral othere were injured. There were no signNicent recent results from test proyams, (c) machenlems for environ-immedlate ellects from suposure to uranyt fluorlds. TNe report monteNy-assisted crecieng, and (d) utReemon of dele and of :he Leseowlesmed Group presents discussions and recom-models Several crtueel verteblos have been idanunod, and the mondemons on the process, opershon and of the fechty, guidennes for processing and applying fougue crack growth date as won as on the responess of the beenese, and other have been developed.

local, state and huletal agencies to the incident it also provides recommendemons in the areas of NRC Rooneing and inspecuon IIURES/CP4007 Vee: PROCEEDINGS OF THE SECONO IAEA of fuel facely and certain other NMSS liconees. The implemen-SPECIAUSTS MEETING ON SUSCRITICAL CRACK tenon of some recommendemons well depend on decisens to be GROWTH. Sessions ill & IV Held At Sendel, Japan,May 15 modo regarding the scope of NRC responstimeles with respect 17,1006. CULLEN,W.H. Metertels Engineering Am inc.

to thoes sepsets of the doelen and operemon of such facenes Apr01906. 623pp. annananm75. MEA-2000. 36000:115.

that are not directly related to radiological safety.

See NUREG/CP 0067,V01 abstract.

10UREG 1991: REPORT OF THE INDEPENDENT AO HOC GROUP FOR THE DAVIS-8 ESSE INCIDENT,

Team on Deves-IIURES/CP 0077: PROCEELINGS OF THE SEMINAR ON LEAK.

Seeme Event. June 100s. espy edO6190671. 30024.040.

BEFORE. BREAK: INTERNATIONAL POUCIES AND SUPPORT.

The Nucteer Regulatory Commiseen estabgehed an inde-ING RESEARCH

DMelon of Engineering Technology.

Set.

pendent Ad Hoc Group in January 100s to review issuse subee-

' Mamo'tet ineutute, Columbus Laboratories Jurm 1986.

guent to a complete lose of feedweler event at Dev6s-Besee Nu.

336pp. 9007000174. 30020-031.

cieer Power Stemon on June e,1986, including the NRC inci-On October 20,20 and 30,1906 the U.S. Nucieer Reguietory dont inveaugemon Team (IIT) inveaugemon of that event. The Commiseen and Bettese's Columbus DMeion cosponsored a Commuoion asked the Group to identNy addluonal lessons that seminer in Columbus, ONo on incomenonal ponctos for look.

might be loomed and from these to mehe recommendemons to before break and the resserch in support of those pohcess. The improve NRC oversight of reactor heeneses. To fulful its charter, purposes of the seminer were to foster communicouen in the the Ad Hoc Group examined the following: (1) pre-event interac.

intemenonal communNy concoming the evoMng regulatory poll-none between the beenese and NRC concoming rehebety of the cies and the receerch being conducted in support of those poil-aumihery feedweler system and associated eyetems, (2) pre-cies, and to examine changes in the seek before break pohcies event pramm acessements of the remetmhty of plant eefety and associated technology bene in the two years since the systems; (3) heenese management, operemon and memeenance CSNI e wma moeung held at Monterey, CaWomie in 1903.

r programs as they may have conenbuted to equipment felures TNs report documente the presentenone made dunng the meet.

and NRC overught of such programs; and (4) the mendete, ca.

ing.

Main Citations and Abstracts 7

NUREG/CR-2000 V06 N3: LICENSEE EVENT REPORT (LER)

NUREG/CR-2676 V06: RELEVANCE OF BIOTIC PATHWAYS TO COMPILATION.For Month Of March 1986.

Oak Ridge Nation.

THE LONG-TERM REGULATION OF NUCLEAR WASTE DIS-al Laboratory. April 1986.125pp. 8604290106. ORNL/NSIC-POSAL (Esnmation Of Radiation Does To Man Resulting From 200. 35756:131.

Biote Transport:The BIOPORT/ MAX 11 Software Package).

TNs montNy report contains Ucensee Event Report (LER)

MCKENZIE,D.H.; CADWELL,LL: GANO,K.A.; et al. Battelle Me-operational informabon that was processed into the LER data morial Insbtute, Pacific Northwest Laboratories. October 1985.

file of the Nuclear Safety Information Center (NSIC) during the 334pp. 8605130566. PNL 4241,35961:037.

one month period identified on the cover of the document. The BIOPORT/ MAX 11 is a collecten of five computer codes de-LERs from which tNo information is derived, are submitted to signed to estimate the potential magrwtude of the radiation dose the Nuclear Regulatory Commission (NRC) by nuclear power to man resultmg from biotic transport processes. Dose to man plant hcensees in accordance with federal regulations. Proce.

is calculated for ingesbon of agdculual crops yown in con-dures for LER reporbng for revisions to those events occurring taminated soil, inhalaton of resuspeded redenuchdes, and prior to 1984 are desenbod in NRC Regulatory Guide 1.16 and direct aposure to penetrating radabon rewtung from me rado-NUREG 1061, INSTRUCTIONS FOR PREPARATION OF DATA nuchde concentraums estabhshed in the available soil surface ENTRY SHEETS FOR LICENSEE EVENT REPORTS. For those by the biobc transport model. TNs document is designed as events occurdng on and after January 1,1984, LERs are bemg both an instructional and reference document for the BIO-autmtted in accordance with the revised rule contained in Title PORT / MAX 11 computer software package and has been wntten 10 Part 50.73 of the Code of Federal Rowlations (10 CFR fw two malw ausences. h first ausence incbdes persons 50.73 Licensee Event Report System) which was pubhshed in concemed eth me mathemencal models of bobgical transport the Federal Register (Vol. 48, No.144) on July 26, 1983.

of commercial bwM redoachve wastes and N computer al-NUREG-1022. LICENSEE EVENT REPORT SYSTEM DE.

gonthms used to implement those models The second audi-SCRIPTION OF SYSTEMS AND GUIDELINES FOR REPORT-ece includes pwsons concemed wim cercising me canputer ING, provides supporting guidance and informahon on the re, program and expoon scearios to obtain rewRs fw W vised LER rule. The LER summades in tNs report are arranged apphcaums. W repet contains seches escdbmg me meh-alphabetcally by fac6hty name and then chronologically by event matical models, user operation of the computer programs, and date for each facshty. Component, system, keyword, and compo-progam smo. Input and Mpd tw Ne sample proNoms nont vendor indexes follow the summaries. Vendors are those are 6ncluded. In amn, Hahngs of h canputer progame, idenbfied by the utsty when the LER form is irwtiated; the key-data hbraries, and dose converson factors are provided 6n ap-words for the Wu,w.t, system, and general keyword indexes pesces.

are assigned by the computer using correlation tables from the NUREO/CR-2860 V04: POPULATION DOSE COMMITMENT DUE Sequence Codmg and Search System.

TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1982. BAKER,DA: PELOQUIN,RA Batteile NUREO/CR-2000 V06 N4: LICENSEE EVENT REPORT (LER)

Memorial Insttute, Pacific Northwest Laboratories. June 1986.

COMPILATION For Month Of Apnl 1986.

Oak Ridge Natonal 150pp. 8907080241. PNL 422L 36919:011.

Laboratory. May 1986.140pp. 8606110035. ORNL/NSIC 200.

Population redaten dose commitments have been estimated 36420104.

from reported redonuchde releases from commercial power re.

See NUREG/CR-2000,V05.NO3 abstract.

actors operating during 1982. In addition doses dortved from the NUREG/CR 2000 V06 NS: LICENSEE EVENT REPORT (LER)

COMPILATIONFor Month Of May 1986.

Oak Ridge Natonal c

t om a

t r a for Laboratory. June 1986.136pp. 8606300479. ORNL/NSIC-200.

populaton groups (infant, cnid, teenager and adult) residing be-36909 001, tween 2 and 80 km from each site. TNs report tabdates the re-See NUREG/CR 2000,V05,NO3 abstract.

suits of these calculanons, show6ng the dose commitments for both liquid and airborne pathways for each age group and NUREG/CR-2331 V06 N3: SAFETY RESEARCH PROGRAMS organ. Also included for each site is a Netogram showing the SPONSORED BY OFFICE Of' NUCLEAR REGULATORY fraction of the total population within 2 to 80 km around each RESEARCH.Ouarterfy Progress Neport.Jufy September 1985.

site receMng various average dose commitments from the air.

WEISS.A.J. Brookhaven Natiorwil Laboratory. March 1986.

bome pathways. The total dose commitment from both hquid 131PP,8605300346. BNL.NUREG-51454. 36171:060.

and airbome pathways ranged from a Ngh of 30 person-rem to TNs progress report will desenbe current activities and techni-a low 0.007 person-rom with an arithmetic mean of 3 persork cal proyess in the programs at Brookhaven Nanonal Laborato y rom. The total population dose from all sites was estimated at sponscred by the DMaion of Accident Evaluaton, Orvision of 130 person-rem for the 100 milhon people considered at risk.

Engineering Technology, and DMsion of Risk Analyss & Oper.

The average indMdual dose commitment from all pathways on atons of the U S. Nuclear Regulatory Commission, Office of Nu-a sto bass ranged from a low of 6 x 10( 7) frrem to a Ngh of l

clear Regulatory Researeft Trm r%octs reported are the fol.

0.06 mrom. No attempt was made in this study to determine the lowing' High Temperature Reactor Research, SSC/MINET De-maximum dose commitment received by any one indMdual from l

velopment, Vahdahon and Apphcat on. Thermal-Hydrauhc Reac-the radonuclides released at any of the sites j

tor Safety Expenments, Plant Analyrer, Code Assessment and l

Apphcation, Code Maintenance (RAMONA 30), Benchmarking NUMEG/CR 3064 V01: COMPUTATIONAL METHODOLOGY FOR i

and Venficahon of LWR Severe Accident Codes, Pool Vernon OAK RIDGE RESEARCH REACTOR (ORR) AND BULK l

of the SSC Code, Uncertainty Analysis of the Source Term; SHIELDING REACTOR (BSR) Cross Section Generston And i

Stress Corrosion Cracking of PWR Steam Generator Tubing, Vahdation, Volume 1. MILLER,LF.; WILUAMS.M L Oak Ridge l

Probabihty Based Load Combinabons for Desgn of Category i National Laboratory. March 1986. 83pp. 8605120159. ORNL/

Structures, Soil-Structure Interaction Evaluabons, Setsmec Re.

TM.9968/V1, 35942.082.

search Coordination and Teclinology Transfer Transfer and Use A neutronics hbrary suitable for low. enriched-uranium (LEU) of the SMACS Code at BNL, Identification of Age Related Fail.

and high ennched. uranium (HEU) fueled cores for both the Oak ure Modes, Combinatonal Procedures for Piping Response Ridge Research Reactor (ORR) and the Bulk Shielding Reactor Spectra Analyses; Apphcation of HRA/PRA Results to Resolve (BSR) is documented herein. The library is obtained from vor.

Human Reliabihty and Human Factors Safety issues, Protective soon V of the Evaluated Nuclear Data File (ENDF/B V) and cork Action Decisionmaking. Operatonal Safety Rehabihty Research, tains 223 nuchdes weighted over a vanety of region-dependent and Venfication of Source Term Code Package Calculatinns neutron spectra. SOlf. shielding and tone. weighting effects are l

l

8 Main Citatione and Abetracts incorporated with 227. group calculations for several reactor-NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER core configurations. Ubraries are arcNved for both transport REACTOR PRESSURE BOUNDARY COMPONENTS. Annual and diffuson theory seven-group calculations. Complete listings Report for 1965.

Materials Engineering Associates, Inc. June of processing deteels are included so that libraries with different 1986. 205pp. 8607020354. MEA-2146. 36666:136.

specefications can be seeily obtaened. Results from validation TNs program is being conducted for the NRC for the purpose calculations indicate that the neutronics hbraries obtained from of providing analytical and experimental methods and data that this effort are suitable for neutrorucs comnutatens for the ORR are necessary to ensure the structural safey and reRebiby of and BSR.

prusure boundary components in Eght water reactor compo-nonts of U.S. commerc6el power systems. Emphasis le placed NUREG/CR-3084 V02: COMPUTAT10flAL METHODOLOGY FOR on characterization of material properties performance in a nu-OAK RIDGE RESEARCH REACTOR (ORR) AND BULK clear environment for the applicamon to plant-afe extension and SHIELDING REACTOR (BSR):The VICTORR Input Processing nhganon of the consequences of postulated accident scener-Code For The Bold Venture System, Volume IL WILLIAMS.M.L; los. Current work is organized into three major tasks: (1) frec-MILLER.LF. Oak Ridge National Laboratory. April 1966. 68pp.

ture mechanics inveengstone, (2) 6,mC; assisted 8605300170. ORNL/TM-9966/V2. 36157.306, crack gmwm in Ngh temperature, pnmary reecer water, and (3)

TNs report is the second volume in a two-volume set which redanon sene% and posenedemon proporses recmory. Re-desenbos the development of a standardized computational search progress in these tasks for 1985 is summarized in tNe methodology for anatyring the neutron environment in the Oak

PC'I-Rwige Research Reactor (ORR) and the Bulk Shielding Reactor NUREG/CR-3282 V01: COBRA-NC:A THERMAL-HYDRAULIC (BSR). The first volume desenbos the development of a stand-CODE FOR TRANSlENT ANALYSIS OF NUCLEAR REACTOR ard seven-group cross.section hbrary based on ENDF/B-V data COMPONENTS. Volume 1: Equations And Conettutve Modois, uteh has been tailored specifically for analyses of these reac-WHEELER C.L; THURGOOD,M.J.; GUIDOTTI.T.E.; et al. Bet-tors. The VICTORR (VENTURE Input Code for Treating Oak telle Memorial Instrtute, Pacsfic Northwoot Laboratories. May Hidge Reactors) program prepares input for neutronics calcula-1986.156pp. 8605210444. PNL-4710. 36060:202.

fons based on the BOLD VENTURE computatonal system. The COBRA-NC is a digital computer program written in FOR-VENTURE module in this system is a multsgroup, three-dmen-TRAN IV that simulates the response of nuclear reactor compo-sional (3.D) diffuseon theory code, which performs core physics nents and systems to thermel-hydraulic transients. The code analyses of complete reactor configuratons. However, the VEN-solves the multicomponent, compressible, three-dimensional, TURE 6nput for defin6ng complex 3.D problems is very involved two fluid, three. field equeuons for two-phase flow. The three vo-end time consumng to prepare, even for the exponenced user; locity fields are the vapor /ges fleid, the continuous liquid field, for the nov6ce, the burden of input preparaton to VENTURE can and the liquid drop field. The code has been used to model flow be overwhelmeng. Thus, VICTORR was wntten to serve as an and heat transfer witNn the reactor core, the reactor vessel, the input processing code whicts requires very httle input from the steam generators, and in the nuclear conteenment The conser-user in order to prepare a complicated 3-D VENTURE model for vaten equatens, equations of state, and physical models that the two Oak Ridge reactors, ORR and BSR, as well as to venfy are common to all applications are presented in tNe volume of the accuracy of the model. From the user's point of view, the the code documentation.

VICTORR. VENTURE calculation is automated so that only the input to VICTORR is required to obtain the VENTURE results.

NUREG/CR 3242 V02: COBRA-NC:A THERMAL HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUMEG/CM 3180: SEISMICITY AND TECTONIC RELATION.

COMPONENTS. Volume 2: COBRA NC Numerical Solution Meth-SHIPS FOR UPPER GREAT LAKES PRECAMBRIAN SHIELD ods. THURGOOD,M.J.; GEORGE.T.L; WHEELER,C.L Batteile PROVINCE FINAL REPORT July 1981 December 1982.

Memorial institute, Pacific Northwest Laboratories. April 1966.

MOONEY,H.; WALTON.M. Minnesota, Univ. of. Minneapolis, 61pp. 8605220014. PNL 5515. 36065:126.

MN. June 1986. 58pp. 8607010528. 36837:317.

The COBRA NC computer program has been developed to The Central Minnesota Seismic Array operated from 1 Janu, predict the thermal-hydraulic response of nuclear reactor com-ary 1977 to 14 July 1982. The intent of the investigation was to ponents m monnabhydaunc transients. The code soNu h provide a seismicity data base for the Upper Great Lakes Pre-mumcomponent, canpre thrmi, WM, cambrian SNeld Province. The results of the investgation are three-f6 eld equebons for two-phase flow. The three fleide are the presented as a table kehng parameters for 16 earthquakes. The vapor neW, me contmous hquid neW, and N hqW Wop W The code has been used to model flow and heet transfer wttNn epicenter locatons are shown on a tectonic map.

the reactor core, the reactor vessel, the steem generators, and NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT li.The in the nuclear containment lhes volm decnbu N Anne-Sandia Waste-isolation Flow And Transport Model For Frac-volm equenons and N mmedcel soluton mmode used 2 tured Media Release 4.84.

REEVES,M.; WARD,0 S.;

sono mee equebons. It is directed toward the user who is in.

JOHNS,N D. et al. Sandia National Laboratories. April 1986.

Wated in gaieg a mm canpWe undemandng of me w 150pp. 8606200217. SAND 83-0242. 36649, t 13.

al memods used to oMain a soMon to N @@M This report is one of three wNch desenbos the SWIFT 11 com-

'9"*'*"

puter code. The code simulates flow and transport processes in NUREG/CR 3262 V07: COBRA.NC:A THERMAL HYDRAULIC geologic moda which may be fractured. SWIFT 11 was devel-CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR oped for use in the analysis of daep geologte facilities for nucle-COMPONENTVolume 7. Assessment Manual for Containment ar weste disposal. This user's manual should permit the analyst Applications.

WHEELER,C.L; THURGOOD,M.J.;

to use the code effectively by facilitating the preparation of input GUIDOTTI T.E.; et al. Battelle Memorial institute, Pacific North-data. A second companion document, Reeves et al. (1985a),

west Laboratories. April 1986. 251pp. 8605270124. PNL 6515.

docusses the theory and implementaten of the models em-36095 207.

pioyed by the SWIFT ll code. A therd document, Reeves et al.

CCORA.NC is a digital computer program wntten in FOR.

(1985b), provides illustratrve problems for instructonal pur.

TRAN IV that simulates the response of nuclear reactor compo-poses. This report contains detailed descriptons of the input nents and systems to thermal-hydraulic transeents. The code data along with an appendix of the input diagnostics. The use of solves multicomponent compresseblo, three-dimensional, two-auniliary files, unit conversions, and program variable descrip-fluid, three-field equations for two-phase flow. The three fields lors also are 6ncluded in this document.

are the vapor / gas field, the continuous hquid field, and the

Main Citations and Abstracts 9

hquid drop feld. TNs volume of the manual provides the user ally drected toward the metabolism and dosemetry of vanous wth the results of compensons between COBRA-NC predc-chemical forms of a redonuclide rather than toward health of-bons and dem otatained from containment systems expenments.

facts per so. TNs study is needed because the chemical spe-These data compensons provide an indcahon of the code's cies of redoelements released to the environment from a high-abihty to predct the response of muttcompartment nuclear con-level waste repoestory may not be adequately descnbod by the teenment systems to postulated loss-of-coolant accidents that metabolic and dosemetnc models of Publication 30 of the Inter.

result in the release of steam, water, and/or i,0c,w,esable natonal Commission on Radologscal Protecton. Our prev 6ous gases into nuclear contaenments.

report dealt wth two main topics: (1) identfying those chemical forms of redenuchdes whch are hkely to reach humans after NUMEG/CR-3441: RADONE:A COMPUTER CODE FOR SIMU.

LATING FAST TRANSIENT ONE-OlMENSIONAL HYDRODY-migrabon kom a waste repoestory and (2) ideWng moes a>

NAMIC CONDITIONS AND TWO-LAYER RADIONUCLIDE pects d me bod (s mtabohem met deped upon me chenical Wm. H was poented out in me prwbus report met one d me CONCENTRATIONS INCLUDING THE EFFECT OF BED-DEPO-SITION IN CONTROLLED AlVERS AND TIDAL ESTUARIES.

grutest uncatainhos now preent in whmanng organ doses ERASLAN,A.H.; ABDEL-RAZEK.M. Oak Ridge Nabonal Labora, from given m%nmntal expown to vedous cheemuel Wms d radonuchdu un in me nemate W me frac #on abebed tory. Apnl 1986. 500pp. 8606260207. ORNL/TM-8870.

through the gastrointestinal tract to blood. TNs report desis with 36785:313.

two topes on absorpton through the gastrointestinal tract (1)

RADONE is a computer code for predcting the transient, ono<hmensional transport of redonuchcles in reconnng water an upper bound fw me absorpbon fracton W plutonium tw adults, wNch is based on data from adult humans, is derived, bodes. The model formulaton considers the one-dimensional (cross-sectonally averaged) conservehon of mass and momen.

and (2) the important topic of absorption of redenuchdes (stron-tum equatons and the two coupled, depth-averaged redonu.

tium and acterudes) by neonates and juvenilee is reviewed. and chde transport equations for the water layer and the bottom a table of recommended absorption fractons for neonates, in-sediment layer. The couphng condtons incorporate bottom fants, cNidren, and adults is presented.

deposaton and resuspenson effects. The computer code uses a NUREG/CR-3447: IDENTIFICATION AND EVALUATION OF FA-decrete-element method that offers vanable nver cross-section CILITATION TECHNIQUES FOR CECOMMISS:ONING UGHT spacing, accurate representahon of cross-sectonal geometry, WATER POWER REACTORS. LAGUARDIA,T.S.; RISLEY,J.F.

and numencal accuracy. A sample apphcahon is provided for TLG Engineenng, Inc. June 1966. 225pp. 8607000464 the problem of hypothetical accidental releases and actual rou-369 hne releases of redonuchdes to me Hudson River.

Q,g NUMEG/CM-3463: ELECTRONIC ISOLATERS USED IN SAFETY Regulatory Commessen to idenbfy practical techniques to faceli-SYSTEMS OF U.S. NUCLEAR POWER PLANTS. NIELSEN,J R.

tate the decomrmsesoning of nuclear power generating fac6hties.

EGAG Idaho, Inc. (subs. of EG40, Inc.). March 1986.104pp.

The othectives of these "fac6ktation techniques" are to reduce 8605090468. EGO 2444. 35922:240.

pubhc/occupahonal exposure and/or reduce volumes of redo-TNs report presents the results from an evaluaten program active waste generated during the decomm6eesoning process.

for electrornc isolators used in safety systems nf U.S. nuclear The report presents the posesbie facihtanon techniques idenu-power plants. Included are recommendatons for test methods fied dunng the study and docusses the corresponding facif,ta-that can be used to ensure that isolabon devices are being tion of the decomrmesioning process. Techruques are catego-quakt ed adequately to satisfy IEEE 279(1) requirements. These rtzed by their apphcabshty of being implemented durtn0 the three recommendatons are based on studes made of National stages of power reactor hfe: desegn/ construction, operation, or Standards, conversabons held with utihty personnel, Nuclear decomrmesiomn0. Detasted cost.benef t analyses were per-Steam System Supphers Architect Engineers, and the isolator formed for each technique to determine the anticipated expo-vendor staff, and analysis of actual tests performed on sample sure and/or radoactive waste reduction; the estimated cost for isolators.

Implementing each techn6que was then calculaied. Finally, NUMEG/CM-3472 V02: SURFACE PROPERTIES AND PER.

these techniques were ranked by their effectiveneet to facihtete FORMANCE PREDICTION OF ALTERNATIVE WASTE the decomrmseiomng process. TNs study is a part of the NRC's FORMS Final Report. HENCH,LL; CLARKE,D E. Flortda, Univ.

evaluaton of decomm%oning policy and modWication of regu-of, Gainesville, FL June 1966. 249pp. 8607080233. 36919:147.

labons portaimng to the decommissiomng process. The findings can be used by the utilrbes in the planning and estabhehment of The main obtective of this research was to study the basic i

mechanisms of waste glass / water interactions and effects of the actrvities to ensure all objectives of decommissiorwng will be acNeved.

experimental vanablos such as glass composeton, pH and solu.

ton composition on 'he chormcal durabshty of glass in aqueous NUMEG/CR-3620 901: INTRUDER DOSE PATHWAY ANALYSIS solutions. The teachate and glass surface were analyred by var-FOR THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The lous soluton and surface analytical techmques such as ICP AE, ONSITE/ MAX 11 Computer Program.

KENNEDY,W.E.;

FTIR, SEM.EDS, AES-lM and XPS to identFy leaching mecha.

PELOQUINAA.; NAPIER.B A; et al. Battelle Momorial Institute, rusms it was found that in the lencNng process, there is a short Pacshc Northwest Laboratories. May 1986,188pp.8C06120891.

period of allall proton or hydromum 6on-exchange followed by a PNL 4054. 36435.046 l

combnation of matrtu desolution and soluten/ precipitation re-The document entitled INTRUDER DOSE PATHWAY ANALY.

actions. Solubhty of the Beached glass species, especially that SIS OF THE ONSITE DISPOSAL OF RADIOACTIVE WASTES:

of sehca, is the dormnant factor Controlhng the chemical durabih-THE ONSITE/MAXll COMPUTER PROGRAM (1964) by Napier ty of weste glasses in aqueous solutions' et al. summarlies our initia efforts to develop human 4ntrusion l

NUMEG/CR 3672 V02: DETERMINATION OF META 00LIC DATA scenarios and a modhed vesson of the MAXl computer pro-APPROPRIATE FOR HLW DOSIMETRY.fl.Gastrointoshnal Ab-gram for potenhal use by the NRC in reviewing appbcatons for l

sorption. CRISTY M.; LEGGETT.R W. Oak Ridge Natonal Labo.

onsste radioactive waste disnosal. TNs document is a supple-retory. February 1986. 75pp. 8605090475. ORNL/TM-8939/V2.

ment to that document and summarises our efforts to further 35922.169.

modify and improve the O'mlTE/MAXIt software package. To TNs report and a previous one (NUREG/CR 3572) evaluate facshtste cross-referenctng, it follows the same format. Notable the dependence on chemical forms of estmates of tealth of-improvements to the software package include the capabehty to fects in Ngh-level waste (HLW). As in the previous report, et is account for sNelding condibons that represent noncompacted assumed that the organ dose to a suitable inden for health of.

trash wastes and the ophon to indcate altematrve land-use con-fects from enposure to radionuchdet, and our discussion is usu-ditone. This supplement contains a descriphon of the imple-

O Main Cltallene enti Abetracts mentauon of these modificatons. In addison, a series of discus-NUREG/CR-3770: PRELIMINARY DEVELOPMENT OF AN INTE.

sions are included in an attempt to increase the user's under-GRATED APPROACH TO THE EVALUATION OF PRESSUR-standing of the scenanos and does calculabon methods. These IZED THERMAL SHOCK AS APPLIED TO THE OCONEE UNIT

- =m roepond to frequently asked questions about the 1 NUCLEAR POWER PLANT.W/TWO OVERSIZE DRAWINGS.

mothematcal models and use of the software. Computer heengs BURNS T.J.; CHEVERTON R.D.; FLANAGAN,G.F.; et al. Oak of the ONSITE/MAXII computer program are included as Ap-Ridge National Laboratory. December 30, 1965. 800pp.

mndices A and 8 Of tNo document. Appendix C bets extemal 8606200340. ORNL/TM-9176. 36646:110.

esposure dose-rate factor libraries.

An evaluation of the risk of pressurized thermal shock (PTS) resulting in a through-the-wall crack in a reactor pressure vessel NUREG/CR 3701: REMtX:A COMPUTER PROGRAM FOR 1EM.

was performed for the Oconee Urut I nuclear power plant. The PERATURE TRANSIENTS DUE TO HIGH PRESSURE INJEC-information presented in tNs report covers one of three plant.

T)ON AFTER INTERRUPTION OF NATURAL CIRCULATION.

specific studies performed for NRC. The other two studies, for lYER,K.; NOURBAKHSH,H.; THEOFANOUS.T.G. Purdue Univ.,

Calvert Chffs Urut 1 and H.B. Rotunnon Unit 2, are documented Wet Lafayette, IN. May 1966.100pp. 8606110702. 36416:05g.

in NUREG/CR-4022 and NUREG/CR-4183, respecevely. The TNs report desenbos the features and use of several comput.

specific objectives of the Oconee study were (1) to develop a er programs developed on the beeis of the Regional Mixing basic methodology for evaluathg the risk of PTS that could be Model (RMM). TNs model provdes a phenomenologicaNy-based refined later in the other two studies, (2) to provide a best esti-and analytical descrtphon of the stratified flow and temperature mate of the frequency of a through the-wall crack for the fleide roeuthng from High Prosauro Safety injechon (HPI) in the Oconee Unit i vessel, (3) to deterone the donnant PTS so-stagnated loops of a Preseurtzed Water Reactor (PWR). The quences for the unit, and (4) to evaluate the effecsveness of

)

beelc program is called REMIX and is intended for thermaity-in.

potent 6al correenve measures. The exannation of tens of thou-duced stratftcation at low Froude number injections. The sands of transients indicated that PTS was not an important REMIX S version is intended for soluto-induced strabfication com melt Irwuabr for Oconee Urut 1. The dominant nok so-with or without thermal effects as found in several expenmental quences were determined to be large steam-line break events.

nimulations. The NEWMtX program is a derivauve of REMIX representing the hmet of max 6 mum poseble mixing w6tNn the NUREG/CR-3882: A METHOO TO CHARACTERIZE LOCAL ME.

cold leg and is intended for Ngh Froude number injections. The TEOROLOGY AT NUCLEAR FACILITIES FOR APPLICATION TO EMERGENCY RESPONSE NEEDS. LINDSEY,C.G.;

NEWMIX-S version accounts for solute effects. Lisbngs of all programs and sample problem input and output files are includ.

GLANTZ,C.S. Battelle Memorial Institute, Pacif6c Northwest Lab-orator 6es. April 1966. 66pp. 8605160024. PNL-5155. 36034:195.

TNs report summertzes the design, application and use of a NUREG/CR 3702: BUOYANCY EFFECTS IN OVERCOOLING weehW mdel to anahre local mtemology at nuclear TRANSIENTS CALCULATED FOR THE NRC PRESSURIZED facihtm in a chmatological kamwk. Wind and s% data THE9 MAL SHOCK STUDY, THEOFANOUS,T.G.; lYER,K,

wn inskumnkd tom am skaN h me of M Mace NOURBAKHSH.H.P.; et al. Purdue Univ., West Lafayette, lg wee feaues and me synopbc scam sudam pmseum M ent pmva in emgen ata cW M May 1966, 200pp. 6606110733. 36417.041. The thermal-hy-draunc responsee of three PWRs (Oconee, Calvert Cl;ffs, and TNs approach isolates and descnbes local meteorology affect.

H.B. Rouineon), to postulated Pressurtred Thermal Shock (PTS) ed by largewak weamer cmdsbons korn utowolopcal 6 hsr ng pnncipah to mecem cWons fand in scenarios, which were originally determined by RELAP5 and TRAC calculations, are being further developed here with demnskated Wng W and knperam" data fmm mm N

EE regard to buoyancy /strabf6cabon effects. These three PWRs "I

were the subject of the NRC PTS study, and the present resulte helped def6ne the thermal-hydraulic conditions utilized in the

""I fracture mechanics calcula' ions carried out at ORNL The com-near Tms gun coast. W res of Neo anaW M ca at characted wmd and staW cmeda can h &

puter program REMIX, wNch is based on the Regional Mixing a

as a Won of N he synW WM W Model (RMM), was the analyncal tool employed, while Purduo's

- '*D"*

1/2 Scale HPl Thermet Mixing facihty provided the basis for ex-Ma a and Wre @e aM manWnhe% M We portmental support. Important mixing and well heat transfer re.

gimes are deelneated on the bases of these results. We con-have also been idenbfied. The results of these analyses are clude that stratification is important only in cases of complete pm ed uparak Me w k W s. h W e

E*CU"O**E*C loop stagnat60n and that muod-convecten effects are important for downcomer flow velocitses below. 0.25 m/s.

dispmion mdels dunng an wnwgency mponse Wabon.

NUREG/CR 396h RELIABILITY ASSESSMENT AND PROBABILl-NUMEG/CR 3706: IMPROVED MODELING AND NUMERICS TO TY BASED DESIGN OF REINFORCED CONCRETE CONTAIN.

SOLVE TWO' DIMENSIONAL ELLIPTIC FLUID FLOW AND MENTS AND SHEAR WALLS. Summary Report. HUANG,H.;

HEAT TRANSFER PROBLEMS. CHAN.B C. Brookhaven N.abon-REICH,M.; ELLINGWOOD,0; et al. Brookhaven National Labo-d Laboratory. May 1966. 65pp. 6607070463. BNL NU9EG-retory. March 1966.110pp. 6605130261, BNL NUREG-51956.

51751, 36000 299.

35953.335.

A basic, limited scope, fast-running computer model is pre-TNs report summartzes work completed under the program sented for the solution of two-dimensional, transient, thermalty-entitled, "Probatzhty. Based Load Combinations for Desegn of coupled fluid flow problems. TNs model is to be the module in Category i Structures, Under this program, the probatxlestic the SSC (an LMFBR thermal.hydrauhe systems code) for pre-models for various stehc and dynamic loads were formulated.

dicting compion flow behaylor, as occurs in the upper plenum of The randomness and uncertainties in material strengths and the loop type design or in the sod 6um pool of the pool-type structural resistance were estabhshed. Several limit states of dosion. The nonhnear Nav6er. Stokes equations and the two-concrete containments and shear walls were identified and ana-equation (two-verleble) transport model of turbulence are re-fytically formulated. Furthermore, the reliability analysis methods duced to a set of unear algebraic equahone in an imphcit finite can be used to evaluate the safety levels of nuclear structures difference scheme, based on the control volume approach.

under various comtunahone of static and dynamic loads. They These equations are solved storattvely in a hne-by hne proco-can also be used to generate anatytically the frag $ty data for chare using the trksagonal matrix algonthm. The results of cal.

PRA studies. In addition to the development of reliability analy-culosone ouemples are shown in the computer generated plots.

sit methods, probatxhty based design entena for concrete con.

Main Citations and Abstracts 11 tainments and shear wall structures have also been developed.

bined thermal and mechancal loadings. From the results of The proposed design cnteria are in the load and resistance theory and the expenments, a design rule for combining me-factor design (LRFD) format. The load and resistance factors chanical and thermal stress in graphite structural components is are determned for several lint states and target limit sta's proposed.

probabilibes. Thus, the proposed design cnteria are risk-consist-ent and have a well-estabhshed rationale.

NUREG/CR-3970: TRAC-PF1/ MOD 1 INDEPENDENT ASSESS-MENT: LOBI INTERMEDIATE BREAK TEST B-R 1 M.

NUREG/CR-3M0: CLOSEOUT OF IE BULLETIN 80-01. Ope.abil-KMETYK,L.N. Sandia Nabonal Laboratones. February 1986.

6ty Of Automate Depressurtzation System (ADS) Valve P eu-92pp. 8605300396. SAND 85 2264. 36168.018.

mate Supply. FOLEY,W.J.; DEAN.R.S.; HENNICK A. Parameter.

The TRAC-PF1/ MOD 1 independent assessment project at Inc. June 1986. 39pp. 8607010443. IEB-80-01. 36838.055.

Sandia National Laboratories is part of an overall effort funded On January 10, 1980, the Phdadelphia Electric Company by the NRC to determine the ability of various system codes to (PECO) Informed the NRC that the pneumats supply for the predet the detailed thermal / hydraulic response of LWRs dunng Automate Depressurization System (ADS) at Peach Bottom accident and off. normal conditions. The TRAC code is being as-Units 2 and 3 might not be operable for all possible events. On sessed at SNLA against test data from various integral and sep-January 11,1980, IE Dullebn 80-01 was issued by the NRC to arate effects test facilities. As part of this assessment matnx, hcensees of operabng boshng water reactors (BWRs), because two 25% intermediate break counterpart tests, B-RIM per-of concern about the possible safety-related defcsency reported formed at the LOGI facility and S-IB-3 performed at the Semis-by PECO. All DWRs were 6ncluded in the issuance, including cafe Mod 2A facility, have been analyzed. The LOBI B-RIM re-several whch reported that they did not have pneumatically op-suits show that no core heatup was measured, and none was eratod ADSs. The hcensees were required to take six specific calculated, there are some 6fferences between calculated re-actions. On May 7,1980, per TMI Action Plan item II.K.3.28, suits and data, but generally the agreement is good. Compart-NRC/NRR issued a letter to all operating DWR Icensees requir*

son of our LOBI results with corresponding results from our ing venfication of quahfication of ADS accumulators and analyses of the counterpart test S-lB-3 show that simslanties chances to technical specificahons. Earty in 1983,20 hcensees and differences in most of the maior phenomena were correctly of current DWRs were requested by NRC/NRR to supply addb predicted.

bonalinformation per Multi-Plant Action F 55. Evaluabon of utili-ty responses, NRC/IE inspection reports and NRC/NRR safety NUREG/CR-402h TRAC-PF1/ MODI INDEPENDENT evaluation reports shows that the bulletin can be closed out per ASSESSMENT. Condensation in Stratified Cocurrent Flow.

specific entena for all of the 26 current OWR facihties to which it DYERS R.K.

Sandia Nabonal Laboratories. February 1986.

was issued. Documentabon applying to related item II.K.3 28 112pp. 8605290040. SAND 84 2161. 36148.035.

and Action F 55 is included in doterminat on of bulletin closcout The USNRC is funding efforts at several laboratones to if necessary. TMl ll K.3 28 and Mulu-Plant F 55 action items for assess the adequacy of vanous advanced, best-eshmate sys-nine facihties that have not been closed out by regionalinspec-tems codes for predicting the behavior of LWRs in accident and ton are included for reference abnormal conditons. Sandia's participation in this protect in.

NUREG/CR-3962: CLOSEOUT OF IE DULLETIN 80-20 Failures cludes the use of TRAC-PFt/ MODI to model stratfied, horizon-Of Washnghome Type W.2 Spnng Return To Neutral Control tal cocurrent flow, for comparison with expenmental data pro.

Switches. DE).N,R.S.: FOLEY,W.J.; HENNICK.A. Parameter, Inc.

duced at Northwestern University. The expenments are very June 1986. 44pp. 8607010448. IED 80 20. 36838 012.

simple, and the results should duplay the effects of mass, mo-On June 18, 1980, Commonwealth Edison Company subret, mentum, and energy transfer at the interface, as well as those ted Lacervee Event Report (LER) 50-295/80-24 to the NRC, de.

of wall inction. Anafyses were performed for four of the North-scribing a malfunction of a Westinghouse Type W 2 control western experiments, which involved condensing steam / water switch,mportant to safety at Zion Unit 1, On the same date, flow in a rectangular channel. The study showed that the code's Weshr#,ouse submitted a prehminary issue of Technical Bulle-timestep control algonthm and enteria for steady state conver-tin N3D-TD 80-9 to the NRC on the subloct switches On July gence need attention, and that the interfac6al heat transfer 31,1980, IE r)ulletin 80-20 was issued to all power plant hcens, model geerapy owpredicts the rate of phase change for con-ees and permit holdors, requinng them to take specific actions ditons of the experiments. In TRAC, horizontal strabfied flow is and report results. Evaluation of utshty responses and NRC/IE assumed to occur in a channel of circular cross section; this inspection reports shows that the bulletin can be closed out per precludes a simple and detailed quantitative compenson be-specific cntena for 122 (98%) of the 124 current facilities to tween calculated results and the reported expenmental data.

which it was issued A followup item for the remaining two facill-However, the quaktative effects of vanous changes 6n experb ties is proposed for use by NRC/IE, to ensure satisfactory com-mental conditions are well predicted in most cases. A very pletion of corrective action.

simple ad hoc modification to the 6nterface treatment, based on boundary layer theory, was able to remove some of the larger NUREG/CR 396S: AN INVESTIGATION OF THE STRENGTH OF discrepancies between the expenmental and calculated results.

H440 GRAPHITE WHEN SUBJECTED TO COMBINED PRI.

Further improvements could probably result from analysis of the MARY AND SECONDARY STRESS. ANDERSON,C. A.;

data in a different way from that presented in the experiment FLY,0 W; LUNDBERG,L 0 ; et al. Los Alamos Sc6entific Labo.

report, but this possibikty was only br6efty examined.

retory Apnl 1986. 81pp. 8605290072. LA 10652 MS 36148 263.

NUREG/CR 404h AN ASSESSMENT OF THE SAFETY IMPLICA.

An expenmental and anatyticalinvestigation of the strength of TIONS OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT a fine grained graphite, H440, under combined mechanical and FINAL REPORT. STONE,R S; CLARK,F.H ; SMITH,0.L.; et af.

thermal stress is desenbod Small sample laboratory tests were Oak Ridge Nabonal Laboratory. March 1980. 325pp.

camed out to estabhsh a mechanical property data base from 8605290031. ORNL/TM 9444. 36149 180.

which statistical parameters could be determined and then used As part of the NRC's Unresolved Safety issue on Safety im.

in finite element codes for predicting failure probabehties of large phcatone of Control Systems (USI A 47) Program, an investiga.

graphite structura' Components under load The theory was ap-tion of nonsafety-grade control system failures at Oconee 1 that phed to graphite nngs under en 6mposed thermal stress from a could lead to rapid steam generator overfall, reactor overcochng, heat flun apphed to the inner surface of the nnqs and under me-or reactor overheating (inadequate core coohng) has been com-chanical stress caused by diametncatty opposed concentrated pleted Transients that result from loss of control system power loads apphed to the outer surface of the nngs Rings of H440 and that threaten plant safety have also been investigated De-graphite were fabncated in two sues and tasted to the com.

taded enarmnations of all Oconee 1 major plant systems were

i 12 Main Citations and Abstracts performed, followed by a logical evaluation of their influence in Lawrence Livermore Natonal Laboratory (LLNL) has conduct-the transients of interest. Broad failure rrode and effects analy-ed a review of the Meilstone Unit 3 (MP 3) Prohahdiate Safety ses (FEMA) were conducted on each of the candidate plant Study (PSS) for the Office of Nuclear Reactor Regulation, U.S.

J systems to determine their safety consequences. In these anal.

Nuclear Regulatory Comrmsson (NRC). TNs probetxlistic safety ytes, the effects of common-cause failures upon Oconee 1 con-study was performed by Northwest Utikties (NU) in response to trol systems were exermned. Sequence analyses using the re-a 1981 request from the NRC, The objective of LLNL's review suits of the FMEAs were then performed to assess comtuna-was to review those aspects of the MP 3 PSS leading to esti-tions of failures and the estimated frequencies of these accident mates of the plant core damage frequency LLNL estimated sequences. Where simple cause-and-effect relationeNps could core damage frequency from intemel events at MP 3 to be not be demonstrated (i e., in cases where feedback was found about 1 X 10(-4) per year. LLNL rev6ewed major areas of the b exist between the failure outcome and the initiating event), a PSS, includmg trutseting events, event trees, succese cnterte, hybrid computer model was used to augment the FMEAs. TNs fault trees, human factors, component and operating exper6ence model, wNch was developed with techniques similar to those data, and treatment of uncertainty. The review of extemel currently used in other maior systems class codes, combines events included certhquakes, fires, external and intomal flood-thermal hydrauhcs, neutrorucs, and control system packages to ing, extreme winds, aircraft accidents, hazardous meterials, and produce a total-plant simulator with emphasis on control system turtune missiles. The MP 3 PSS treeted extemel events, other dynarrucs.

than seismic and fire, in a cursory manner. LLNL's seismic NUREG/CR-4046: A METHODOLOGY FOR ALLOCATING RELI-rmw obt was curtasted h me stan h-of ongoing seis.

ABILITY AND RISK. CHO.N.Z.; PAPAZOGLOU,lA: BARl,RA nuc analyms rmons W M Brookhsven Nat,onal Laboratory. May 1986.

182pp-NUMEG/CR 4807: FAULT TREE APPUCATION TO THE STUDY 6606t90000. BNL NUREG-51834. 36621:292.

OF SYSTEMS INTERACTIONS AT INDIAN POINT 3.

TNs report desenbos a methodology for rehabelity and risk al.

YOUNGBLOOD.R.W.; HANAN,N.; FITZPATRICK.R.; et al.

locaton in nuclear power plants. The work investigates the Brookhaven National Laboratory. March 1985. 332pp.

techrncal fossetxhty of allocating rehatxhty and nok expressed in 8604180302. BNL-NUREG-51872. 35623.305.

O set of global safety critena to vanous reactor systems, sub-TNs report desenbes an applicaton of fault t ce methods to systems, components, operahons, and structures in a consistent search for systems interactions at Indian Point 3. This project manner. The problem is formulated as a multiettnbute decision was carried out in support of the resolution of UnresoNed analysis paradsm. The work mainly addresses the first two Safety lasue A 17 on Systems interaction. Here, tre methods steps of a typical decision analysis, i.e., (1) identifying altema.

are introduced, the findings are presented, and comments on tives and (2) generating information on outcomes of the altema-the methods are offered. Findings are presented in the follow 6ng tives, by performing a multiobjective optimization on a PRA manner. Systems interactons wNch may auelitari violate model and rehabehty cost functions. The muhtotdectrve optimiza-regulatory requirements (regardless of theer probabihty) are dis-tson serves as the guiding pnnciple to rehabihty and risk alloca-cussed, addihonally, a probabil6stically ranked list of system hon. The concept of "noninfenonty" is used in the rnultiobrec.

interactons is provided. This study resulted in the discovery of a ttve opundrahon problem. Findng the nonenferer soluton set is previously undetected active single failure causing loss of Icw the rwn theme of the current approach. The final step of dec6*

pressure injection. After vonfy6ng tfwe findng, the hcensee took son analysis, lo., assessment of the decision maker's prefer

immediate corrective actons, inclueng a design modification to ences could then be performed more easily on the noninfenor the switching logic for one of the safety buses, as well as pro-solution set. Results of the methodology applications to a non-cedural changes.

Invlat risk model are provided, and several outstandire issues such as genenc allocation, preference assessment, and uncer.

NUMEG/CM-4236 V03: PROGRESS IN EVALUATION OF RADIO-tainty are dLcussed.

NUCLIDE GEOCHEMISTRY INFORP.lATION DEVELOPED BY NUMEG/CM 4124 V02: NDE OF STAINLESS STEEL AND ON.

DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE LINE LEAK MONITORING OF LWRS. Annual Report Octobe, PROJECTS. Report For Apni June 1965. KELMERS.A D.;

1984 September 1985. KUPPERMAN,0.S.; CLAYTOR,T.N.;

ARNOLD W.D.; BLENCOE,J.G.; et at Oak Ridge National Labo-MATHIESON,T.; et al. Argonne National Laboratory. February retory. April 1966. 26pp. 8604250237. ORNL/TM 9614/V3.

t986. 58pp. 8605290104. ANL-86 7. 36147.231.

35734 215.

An informal workshop on NDE of pipes with weld overtsys Geochemical informahon relevant to the potential mobihty of was held at ANL in January 1985. Personnel from four research radonuchdes at the Hanford Site and the Yucca Mounta6n sete, 6nstitutions and one utihty attended the workshop. Two pipe-to.

candidate sites for high-level nuclear waste geologic repositor.

endcap weldments with overisys were examined. Because the les being developed by Department of Energy projects,is being amount of cracking in the specimens was hmited, the emphasis evaluated by Oak Ridge National Laboratory (ORNL) for the was on trying to understand the nature of crack overcathng.

U S. Nuclegar Regulatory Commission. Neptun6um(V) sorption One mator source of spunous echoes that may be erroneously isotherms in three efferent basalt / synthetic groundwater sys-interpreted as originating from a crack has been ldentified. De, tems were initiated this guarter. Uranium (VI) sorption isotherms cause of the comples microstructure of the overlay, a situation were completed with McCoy Canyon basalt and synthetic may exist where in addition to the conventional angled shear groundwater GR.2. The control of U(VI) solutions at a level of twam, a second shear beam is created which propagate, 1004) mol/L is apparentfy dominated by the precipitation of normal to the surface of the specimen. This beam has a polar.

sodum boltwoodite. Different apparent concentration hmits for liahon that makes it impossible to detect by standard finger-uranium were obtained with GR 2 and GR 4 solutions. The re-damping techniques, for this reason, it was not property charac.

sufts suggest that uranium sorption /solubehty behavior could be tented oather. We concluded that it is difficult to 6nspect pipes substantially different in the vanous basalt unite likely to be en-with oveitays because of unpredictable beam distortion due to countered by groundwater during migration from the waste the overlay and the absence of effective reference pipes. The package to the accessible environment. The E03/6 softwater use of 1. Mitt longitudinal angle beam probes rather than shear.

package was implemented and tested on the ORNL computer wave probes may facihtate inspecton of such pipes, system.

NUREQ/CR 4142: A REVIEW OF THE MILLSTONE 3 PROBABl.

NUREG/CR 4241: CHEMICAL ASPECTS OF CESIUM r>DIDE LISilC SAFETY STUDY. GARCI A.A ; DERNEUTER.D ;

INTERACTION IN STEAP.1 WITH 304 STAINLESS STFil AND MCKONE.T; et al Lawrence Livermore National Laboratory.

INCONEL 600. SALLACH.R A. Sandia Nabonal La'etatonet April 1986 349pp. 6605160305 36006 295.

Apnl 1986 40pp. 8607020201 SAND 84 0749. 3685e 256.

Main Citations and Abstracts 13 WNie being oxidized by steam at 1273K, the alloys 304 stairw loop piping, in which the probabihty of direct DEGB had been less steel and inconel-600 were exposed to Cal vapor. The explicotty estimated using a probabilistic fracture mechanics chem 6 cal data from subsequent analyses are presented and dis-model, no detailed fracture mechanics calculations were per-cussed. Some cesium, but not iodine, was retamed on the oxide formed. Instead, a comparison of relevant plant data, mainly re-formed on inconel-600. There is a moderate correlation of actor coolant loop stresses, for one representative B&W plant cesium with the sehcon content of the onde (correlaten coeffi-with equivalent information for Westnghouse and C-E systems cients near 0.7) but no correlation with any other element For-Inferred that the probabikty of direct DEG8 should be sirmierty mation of a cesium sehcate is indcated. No cesium or iodine low (less than to 10 per reactor year). The probabihty of indirect was detected in the oxide layers formed on 304 stamless steel, DEGB, on the other hand, was explicitly estimated for two rep-in sharp contrast to pievious expenments with CsOH vapor.

recentative plants. The results of tNo study indicate that the NUREG/CR-4381: ASSESSMENT OF SYSTEM INTERACTION probabihy of a DEGB W esmer cause is very W fw ream EXPERIENCE IN NUCLEAR POWER PLANTS. MURPHY,0.A.;

calent W piping M h specife plants and, becem of se KIMMINS,A.D.; THORNTON,RH.; et al. Oak Ridge Natonal Nany in demon, infer hat me probabikh of DEGB is generah Laboratory. June 1966.110pp. 8606260204. ORNL/NOAC-227.

very low B&W re calent imp W De E M 36785 053 wec nedw ung DEM as a W m mnt in TNs report desenbos the second phase of an NRC spon-

"OI""*"'N # N sored project that idenhfied and assessed the system interac.

NUREG/CR-4300: VALENCE EFFECTS ON SOLUBluTY AND ton events that have occurred at commercial nuclear poww SORPTION.The Solutulity Of Tc(IV) Ouides. MEYER,R.E.;

plants in the United States. The second phase of the project l'k ARNOLD,W.D.; CASE F.L Oak Ridge National Laboratory, cluded a summary and assessment of methods to find adverse March 1986. 40pp. 8605130355. ORNL 6199. 35953.095.

system interactons; plus, for each of the 23 categones identi-The solutzhties of technebum (IV) oxides have been deter-fied in Phase 1 (reported in NUREG/CR-3922), identified the safety concerns, assess previous regulatory or industry action to rmned in solutens of Nacl, hcl, and synthetic groundwaters in address the concern, and to make recommendetens to resolve the pH range 0 to 10. Oxides were prepared either by electrore-safety concerns needmg further attenton. Two general areas of duction from Tc(Vit) solubons onto platinum electrodes or by concern were identsfied as needing further attenton: (1) adverse precepitehon of Tc from Tc(Vil) solutens using hydrarene. In interactions between normal or offste power systems and some of the determinations, the oxide was precipitated onto pu-nfled sand by use of hydrazine. The oxide-covered send was emergency power systems, and (2) potenbal seismic, induced spatsal interactons. The results of Phase I and 11 will be integrat-then placed into a column and the test solution contmuously re-ed into the technical resoluton of Unresolved Safety lasue A 17 circulated through the column. The onlde plated on plehnum System Interactions in Nuclear Power Plants.

was placed into a small cell and the test solution stirred. Solubi-hties were determmed by rneasuring the beta radiation of (99)Tc NUMEG/CR 4265 V01: AN ASSESSMENT OF THE SAFETY IM-in the stirred or circulated soluton in contact with the oride. In PLICATIONS OF CONTROL AT THE CALVERT CLIFFS 1 NU-the pH range 4 to 10, solut>hties were in the range 7.5 x 10( 9)

CLEAR PLANT. BALL S.J. Oak Ridge Natonal Laboratory. Apnl to 5 x 10(-8) mol/L, and most of the determinations were in the 1986.239pp.8607070481. ORNL/TM-9640/V1, 36900.065.

range 1 x 10(.8) to 2 x 10(.8) rnol/L in acid solutions the solu-TNs study examined the consequences of possible control bihtses were generally Ngher for hydrarme precipitated oxide system malfunctions at the Calvert Chffs 1 nuclear power plant than for the electrodeposted oxide, and they increased as the as technical support for an NRC program to assess the safety pH was lowered. Solutphbes of Tc(IV) oxides will be useful for emphcations of ruclear power plar,t control systems. Plant sys.

calculaton of transport rates of technetium for the case of solu-tems capable of initiabng piant overcoorng and undercoohng tzhty-hmited transport in nuclear weste repostory environments.

were identified, as well as those with potential for overfill events in the steam generaten. Failure mode and effects analyses NUMEG/CR 4315 V01: EVALUATION OF NUCLEAR FACILITY were conducted on these candidate plant systems, with com.

DECOMMISSIONING PROJECTS. Summary Status Report puter analysis apphed whwe appropnate. TNs latter process utl-Three Mile letand Unit 2 Reactor Coolant System & Systems bred a detailed RETRAN model of the Calvert Chffs plant using Decetaminaton. DOERGE.D H.; MILLER,R L; SCOTTI.K S.

adaptations made as put of tNs program. Where failures with UNC Nuclear Industries. May 1966. 43pp. 8606240547, safety consequences were found, probatHhties of the pertinent W R251 scenarios were develoPid. Several control system failures were TNs document summantes informat6on concerning the de-identified as being of gossible safety concern. Of these, two contam6 nation and restorat6on of the Three Mile Island Unit 2 were selected as being of sufficient interest to warrant further reactor coolant system and other plant systems wNch incurred study and followup us ry plant simulations.

damage dunng the loss &oolant accident esportenced on March 28,1979. The data collected from actMty reports, reac.

NURFG/CR 4290 VOI: PROBA0lLITY OF PIPE FAILURE IN THE tor containment entry records, and other sources were placed in REACTOR COOLANT LOOPS OF DADCOCK AND WtLCOX a computenrod informaton retneval/manipulebon system wNch PWR PLANTS Volume i Summary Report. HOLMAN.G S ;

permits outraction/ manipulation of specific data which can be CHOU.C K. Lawrence Livermore National Laboratory. May 1986.

utslized in planning for recovery activities should a similar acci.

69pp 6606200205. UCRL 63644 Vol. 36648 286.

dont occur in a nuclear generating plant The informahon is pre.

As part of its reevaluahon of the double-ended guillotine break (DEGO) of reactor coolant piping as a dssign base event sented in both computer output form and in a manually assem.

bled tummary. TNs report contains only the man. hours spent for nuclear power plants, the U S. Nuclear Regulatory Commis-and radiation esposure actually incurred during performance of sion (NBC) contracted the Lawrence Livermore Natonal Labo-tasks in radiation areas and does not include support activities retory (LLNL) to estimate the probabihty of necurrence of a outade of radiahon areas.

DEGH, and to assess the effect that earthquakes have on DEGO probatulity. This report describes an evaluehon of reactor NUMEG/CR 4315 V02: EVALUATION OF NUCLEAR FACILITY coolant loop piping n PWR plants hav ng nuclear steam supply DECOMMISSIONING PROJECTS. Summary Status Report systems deogned by Dabcock & Wilcot Two causes of pipe Three Mde Island Unit 2 Reactor Duilding Decontammaton.

break were considered. pipe fracture due to the g7wth of DOERGE,D H ; MILLER,R L: MCOTTI,K S. UNC Nuclear Indus-cracks at welded nts (" direct" DEGO), and pipe rupture inde-tries. May 1986 163pp. 8606240386 36710.108, rectty caused by failure of heavy component supports due to an TNs document summartres informahon concorrwng decon-earthquake ("6ndirect" DE00) Unlike n earlier evaluatens of taminaten of the Three Mde Island Unit 2 reactor buelding wNch Westinghouse and Combuston Engineering reactor coClant was grossly contaminated as a result of the loss of coolant ac-

14 Main Citations and Abstracto cident experienced on March 28,1979. Data collected from ac-NUREG/CF-4319: NUCRAC A CODE FOR THE ESTIMATION j

tmty reports, reactor contenment entry records, and other OF ADVERSARY ACTION CONSEQUENCES IN THE NUCLE-j sources were placed in a computertred informahon retrieval /

AR POWER FUEL CYCLE. KAUL D.C.; RITZMAN,R L; manipulahon system wNch pernts extraction / manipulation of ROBERTS,J.A.; et al. Brookhaven Nabonal Laboratory. February specife data wNch could be utshred in planning for recovery ac-1986.431pp.8605290025. BNL NUREG-51904. 36152:237.

tmhes should a smier accident occur in a nuclear generahng A program sponsored by the Nuclear Regulatory Commission plant. The ir.formahon is presented in both computer output (NRC) and designed to estimate the potenhal consequences of form and a manually aseembled summartration. TNs report con-adversary schons in the nuclear power fuel cycle has been tens orWy the man-hours spent and radahon exposure actuaNy completed. So that the results of this consequence analysis incurred dunng performance of tasks in radahon areas and would be comparable to that of the Reactor Safety Study does not include support actmties outside of radation areas.

(RSS), the methodology descnbod in the RSS and implemented Na covers the period of activrhes from June 1,1979 in the Calculebon of Reactor Accident Consequences (CRAC) code served as the baseline for consequences evaluation in NUMEG/CM-43tl V03: EVALUATION OF NUCLEAR FACIUTY tNs study the atmosphenc depersion model,the inhalation dose DECOMMISSIONING PROJECTS. Summary Status Report factors, the enterion for earty mortality from lung dose, and the Three Mde Island Unit 2 Reactor Defueleng & Disassembly.

model for chronic pathways to man. Implementahon of these DOERGE,0.H.; MILLER R.L; SCOTTI,K.S. UNC Nuclear Indus.

modrcations to the CRAC code resulted in the preparahon and tries. May 1966. 83pp. 8606240370. 36710 271.

appication of a revised code termed NUCRAC, These modifca This document summartres information concoming Three tions are descrAJ in detail. Detailed instructions for the oper-Mile Island Unit 2 reactor defuehng and esassembly actMhee ation of NUCRAC are presented in the form of a user manual.

being performed concurrently with decontamnahon of the facih.

Inputs and outputs for an example calculahon are also present-ty dunng the penod of Apnl 23,1979 to April 16,1985. Data ed.

collected from actmty reports, reactor contenment entry NUMEG/CM 4330 V01: REVIEW OF LIGHT WATER REACTOR records, and other sources were placed in a computerized infor.

REGULATORY REQUIREMENTS. Volume 1: Identification Of I

mahon retneval/manipulabon system wNch permits extracton/

l marnpulation of specife data wNch could be utilized in planning Regulatory Requirements That May Have importance To Risk.

for recovery actmtses should a similar accident occur in a nucle.

SCOTT,W B.; BICKFORD,W.E.; BEOGEL A.J.; et al. Battelle Me-I at generahng plant. The informahon is presented in both com.

monal Institute, Pacife Northwest Laboratones. April 1986.

puter output form and a manually assembled summartrabon.

84pp. 8605020027 PNL 5609. 35807:293.

TNs report contons only the memhours spent and radiation ex-In a study commissioned by the Nuclear Regulatory Commis-l posure actualty 6ncurred during performance of tasks in radiaton sion, Pacife Northwest Laboratory (PNL) idenhfied burdensome l

areas and does not include support actmties outside of redl.

light water reactor regulatory requirements that appeared to ation areas, TNs report covers the period of activit>es from April have marginal importance to nok. PNL obtained industry and i

23 through Apnl 16 1965.

NRC input to the idenhfication process through fermal inter-I views and questionnaeres. Over 40 regulatory requirements were NUREG/CR 43tl V09: EVALUATION OF NUCLEAR FAClurY identified and these are docussed in the report. Based on the DECOMMISSIONING PROJECTS. Summary Status Report information collected, the potential savings in terms of reduced Three Mde faland Unit 2 Radoactrve Weste And Laundry ShP regulatory burdens, both for NRC and the industry, appear to be ments. DOERGE.D.H.; MILLER.R L; SCOTTI,K.S. UNC Nuclea' substantial without compromising pubic health and safety.

Industnes. May 1906. 39pp. 8606240409. 36711:102.

l TNs document summanres informahon concerrung radoac.

NUREG/CR-4330 V02: REVIEW OF LIGHT WATER REACTOR

(

trve waste and laundry sNoments from the Three Mle leiend REGULATORY REQUIREMENTS ASSESSMENT OF SELECT.

l Nuclear Stahon Urut 2 to radoactive waste esposal artes and to ED REGULATORY REQUIREMENTS THAT MAY HAVE MAR-protecttve clotNng decentaanation facdihes (taundnes) since GINAL IMPORTANCE TO RISK. Reactor Containment Leakage the loss of coolant accident exponenced on March 29, 1979.

Rates Main Steam Isofahon Valve Leakage... MULLEN,M F.;

Data were collected kom radioactive shipment records, summa DAILEY,W.J.; DEYER.C E.; et al. Battelle Memonal institute, Pa-fired, and placed in a computenrod information retnevaf/ manip.

cife Northwest Laboratones. June ifr86.130pp. 8606240427, ulation system wNch permits entraction of specific informahon PNL 5809. 36110.353.

as required. Informat>on contained in tNs report includes: waste in a study commissioned by the Nuclear Regulatory Commis-deposal site locahons, dose rates, cune content, waste descrip-soon (NRC), Pacife Northwest Laboratory (PNL) evaluated the tion, container type and number, volumes and weights.

costs and benefits of streamhning regulatory requirements in the NUREG/CR 43to: EVALUAflON OF NUCLEAR FACILITY DE.

areas of reactor containment leakage rate, main steam isolation COMMISSIONING PROJECTS STATUS REPORT HUM.

valve leakage control systems in boshng water reactors (BWHs),

00LDT DAY POWER PLANT tJNIT 3 SAFSTOR DECOMMts.

and NRC fuel system safety rev6ews. The basse framework for SiONING. DAUMANN.0 L; HAFFNER.D R ; MILLER.R L; et al.

the analyses was that presented in the Regulatory Analysis UNC Nuclear Industnes. June 1986. 104pp. 8606250022.

Guidelines (NUREG/DR 0058) and in the Handbook for Value-36739 022.

Impact Assessment (NUREG/CR.3568). The effects of stream-TNs document summarties information concoming the SAF.

lined regulatons were evaluated 6n terms of such factors as STOR decommessorwng of the Humboldt Day Power Plant Unit populabon dose, 6ndiv6 dual dose, prompt fataktes and 6njunes, j

3 Preparahons putting this facility into a custodial safestorage and costs to industry and NRC. The results 6ndicate that

[

(SAFSTOR) mode are scheduled for complehon by January 1, streamlining the regulatory requirements ln all three areas wrwsid i

1966 TNs report grves the current status of those efforts. A have little impact on pubhc risk Substanhal savings in operating final report wdl be 6ssued after completon of these preparations costs may be reahted in the areas of containment leakage rates for custodial SAFSTOH The informaton collected from the fa-and leakage control systems for OWR main steam isolabon cdity decommworung plan, environmental report, and other vatves The cost anafysis indicates that only marginal benefits sources inade availat le by the licensee were placed in a Com*

may be g4'ned by streamlining NRC's safety review of fuel puter data base system wfW.h permits data manipulahon and gystem designs summanistinn. These computer generated reports and back-grouruf information are included in th's document l

l

Main Citations and Abstracts 15 NUREG/CR 4333: DESIGN AND FINAL SAFETY ANALYSIS and axial-history dependent models. These correlations have REPORT FOR VERTICAL FISSION PRODUCT RELEASE AP-been compared with more than 500 data points reported in PARATUS IN HOT CELL B. BUILDING 4501. OSBORNE.M.F.;

NUREG/CR-33t>2.

COLLINS.J.L; HAAS P.A.; et at Ook Ridge National Laboratory.

March 1986. 45pp. 8005000465. ORNL/TM-9720. 35922:127.

NUREG/CM-4374 V03: A REVIEW OF THE OCONEE 3 PROB-An enietng horizontal fumece and a fleeion product collection ABILISTIC RISK ASSESSMENT CONTAINMENT system, used for teetng light.weter reactor fuel under simulated PERFORMANCE RADIOLOGICAL SOURCE TERMS AND RISK accident condluons, were rebuilt for operaton in the vertical ork ESTIMATES. PARK C.K.; AGRAWAL,A.K.; KHATIB-RAHBAR; et entatort Mejor changes in the design of the fumace and irt the al. Brookhaven National Laboratory. June 1986. 90pp.

loading /unioeding procedure were required. Modrincetons to the 8007080211. BNL-NUAEG-51917,36918:167.

Aseim product coNecton components and the steel contain-A technical review of the Oconoe-3 Probabsliebc Riek Assess-mont box for the entre apparatus, however, were rela 0W/

ment (OPRA) hee boon performed with the objective of evaluet-rnina Because me fuel specimens, me fumace constuccon, ing containment response, radiological source terms, and off-and the general mode of test operaton for this new facility are site consequences. In the OPRA study, a detasled structural all aimeter to those of the previously operated facihty, no signifi.

andysse fw demsneton of uitmeM fedwo pm of me cent changes in the safety hazards are apparent. Adherence to containment has not been performed. A seneMevity study shows aN queMy assurance Weining, and operatonal safety that the off-site conesquences as well as the frequencies of re.

ieese caiego,ies - be changed by a is,ge maganua. pend.

NUREG/CR-4330: TELLURIUM BEHAVIOR IN CONTAINMENT ing on the assumphone used on the dlstributonal parameters UNDER LIGHT WATER REACTOR ACCIDENT CONDITIONS.

and the probability of contesnment isolation failure. The ansees-BEAHM.E.C. Ook Ridge National Laboratory. February 1986.

ment of the OPRA shows that very late overpresourtraton fail-29pp.8605130356. ORNL/TM 9726. 35953.066.

uro is the dominent containment failure mode. The OPRA redlo-Interactions of tellurtum in contesnment can result in changes logice rWesses we comparable in magnstude to thou used in of physical form and therefore in its transport properbes. This other recent studies. Cost-effectiveness analysis by a muill-ob-report <tamaa=a the most probable forms of teNurium in a con-joctive optimizaton approach shows that vertous containment lainment environment under LWR accident conditons. The are doesgned coneamenwy phyescal and chemical form of inorganic tellurium speces will be determined by condensation, oxideton, and dissoluton in water.

NUREG/CR-4379 V04: LONG TERM PERFORMANCE OF MATE-Of the three volatlie tellurtum chemical forms, Ta(2) (ges),

H(2)Te, and organic teNur' des, only organic telluridos hevo the RIALS USED.r OR HIGH-LEVEL WASTE PACKAGING. Annual g, g c4r. April 1985 March 1986. STAHL,D -

potenhal to remsen in the gas phase in a contamment atmos-MILLER.N E. BatteNo Memorial inettute, Columbus Laborat[

phere. There is a general lack of informehon on the formation rtes. June 1966.175pp. 0007070478. BMI-2128. 36890:199.

and removal of orgerwc tellurides under LWR accident condi-tions.

Wask fem studies how been Wocted toward inwetgegng spent fuel leeching /dissolut6on behav6or. The experimental vai6-NUREG/CM 4340: LSL M2:A COMPUTER PROGRAM FOR demon of a glau dissoluton/reprecipMason model hee been LEAST SOUARES LOGARITHMIC ADJUSTMENT OF NEU*

concluded. Expenmente are generanne date on UO(2) and TRON SPECTRA. STALLMANN F.W. Oak Ridge National Labo-opent-fuel leech rates in simulated anonic groundwaters, and ini-re ch 1986. 110pp. 8605090459. ORNL/TM 9933.

tal data indicaM met only brine leechante contain any mesewa-LSL M2'le a package of computer programs for the adjust-E 0"

I ment of neutron spectra in nuclear reactors beoed on the com-e suscephbilHy d cast Wed M PNung cWW and ewes &

benebon of neutron transport calculehone and radiometric or core cracWng is W Wudied by elechchwn6 cal e other integral doeimetry measuremente and their uncertainties in niques. Polential cracking agente are being inveengated by slow the form of variances and corrWooons. The undertying algorithm ween rak experwnets. The pneng-conmion modd wee twmer le a tenet equeres logenthmic statteccal setsmation procedure.

t 23+1 taking into account cation dissolubon at the pit base Spectra at several different locatione can be processed simulta.

and chenucaNy acew pH wens. GroundwateNedioheie modehng neously which increaeos the accuracy of the adjusted values hee congnued, wMh the descripton being exteded to include and allows the adjustment of spoetra at locat6ons without do.

additional spec 6es. Spent fuel specimens are bemg used in inte-esmetry. The primary output le adjusted damage parameter yet tests with flowing simulated groundwater to study the role values (e g, fluence > 1.0 MeV ano dpe) with uncertamties.

of cledd6ng in radionuclide reisese and to identfy posetle com-bineMocte procean NUMEG/CR 4363: ASSESSMENT OF POST-CRITICAL HEAT FLUX NODELS WITH LEHIGH NONEQUILIBRIUM OATA.

NUREG/CR-4364: BREAK SPECTRUM ANALYSIS FOR SMALL WEBB.S W4 CHEN.J C. Lehigh Uruv. Bethlehem, PA. April BREAK LOSSOF COOLANT ACCIDENTS IN A RESAR 3S CHER,@ MERG,3 NG Maho, M ne' of convective film boelmg fWe r e c r e.

model predictions to the poet <rttical-heet flux esper6 mental data (subs. of N, In4 March m 78pp. 860423m31. EGG.

2416. 35681:341.

obtamed at Lehegh UrWversely. At Lehigh Universety, meeeure.

monts of aulaNy varying vapor superheate 6n diepersed flow con-A sortes of thermal-hydraulic analyses were performed to in-vectfve film boilmg have been obtained for water in a vertical wohgek phewnea occwnng dwing eman break iceed-cool-tube et low preeeure and mese flux conditions. These data were ant-accident (LOCA) eequences in a RESAR-3S preeeurtred taken uomo a slow "reflood" procese allowmg for measurement water reactor The analyses included semulebone of plant behav-of won and vapor temperature as a function of the distance lor uomg the TRAC-PFt and RELAPS/ MOO 2 corr.puter codes.

from the quench front. The esperimental set up and the film Sones of calculations were performed using both codet for dif.

boilmg data are summettred in a separate report by Evans, forent break eiros. The anatyees presented here ateo served an Webb and Chen [NUREG/CH 3363). Two types of convective audit function in that the results shown here were used by the film boilmg modois have been evaluated (t) equehbnum modele, U.S. Nuclear Regulatory Commietion (NRC) as an independent which do not allow for vapor superheating, and (2) nonequili-confirmation of eimolar anatyeet performed by Weehnghouse bnum models which attempt to predict the actual Quality and Electnc Company uomg another computer code, vapor superheats The equelibnum correlettone are reetncted to local models wtvle the nonequehhnum correlatione include local NUMEG/CR 4402 V03: HIGH TEMPERATURE GAS COOLED RE-

16 IInin C10milone ansi Atletracts i

ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT venish at temperatures between 100 depees coneyede and EVALUATION. Quarterly Progrees Report. July 1 September 1100 degrees eeneyede for the best-eemmele Comoert-Powel 30.1006. sALL S.J.; CLEVELAND.J.C.; CONKLIN.J.C.; et al. Oak correlamon. Use of the Cathcart-Powel correlamon for LOCA ces-Ridge Neuonal Laboratory. Apru 1000, 22pp. 8000200273.

culetone con be juseRed at me 06% to 80% conAdence level N ORNL/TM-0700/V3. 30042:340.

cootng rate effects con be neglected.

Further accident sceneno analyses were done using simulo-tlone for the modular High-Temperature GeoCooled Reactor IAURES/CR 441k LOSS OF CONTROL AIR AT BROWNS (HTGR), including en uncertainty onelysis of the reactor veneel FERRY UNIT ONE. ACCIDENT SEQUENCE ANALYSIS.

enial temperatures for long. term heetup accidente. More sec.

HARRINGTON,R.M.; HODGE SA Ook Ridge Neuonal Labore-mons of me HTGR SAFETY HANDBOOK were completed. Sev.

tory. March 1000. 71pp. 0006300333 ORNL/TM 0028-eral L,.

' teoks were completed for the Fort St.

30157:010.

J Vrain HTOR.

This study describes the predcted response of to Browns IIURES/CIM404: ANALYSIS OF ALLOWED OUTAGE TIMES AT Tfw feNur plant contal cessados o BYRON GENERATING STATION. CHO N.Z.; CHU,T.L; XUE D.;

um ices of drywou omWW Wr m Urms 1 and 2. Nevermehen, et al. Brookhaven Neuonal Laboratory _ June 1000. 130pp.

me W a benign acident urdens compounded by shnuheneous 0000240602. BNL-NUREG-6103130711:141.

h in um m lup pressure % eymems, This report provides a crecal review of the methods used in Accident emquence ceiculetons are presented for Lees of Con-WCAP 10626 which proposed met showed outage smes (AOTs)

WW Air emquences wnh eseurned hdure upon demand W tw for a number of salmy systeme in the syron Genereung Stenen Reactor Core lealebon Coceng (RCIC) and tw High Pressure be incrossed kom 3 m 7 days, and preente en independent Cooient iniection (HPCI) systems at Unit 1. Seguences wnh and soumete of me change in risk invol.sd in me ACT entonoton. It womout opwWw acuan we conddmed. fhouns show mm me eleo presents reeuns W samed emesenty mudes. Amo inchesed operatore can prevent core uncovery W may tehe ecuan to us.

are a survey of rnemodo that can be used to evaluate nucces' nae me C,atW Rod Drive Hydreute SyWorn m a W Pup power plant techrucal Wamne and a description of pair-p,,,,,,,

wies importance measures NURES/CR-4400: DATA BASE ON NUCLEAR POWER PLANT 10URES/CIH440 A PWR HYORIO COMPUTER MODEL FOR

(

DOSE REDUCTION RESEARCH PROJECTS. KHAN,T.A.;

ASSESSING THE SAFETY IMPLICATIONS OF CONTROL SYS-DIONNE,0.J.; BAUM J.W. Brookhaven Nabonel Laboratory. Do.

TEMS. SMITH,0.L; REINER,J.P.; DIFILIPPO,F.C.; et al. Ook cember 1806. 300pp. 8000200202. BNL NUREG-51034.

Ridge Neuonal Laboratory. March 1000. 207pp. 8000300872.

30060.002.

ORNL/TM-0000. 30171:192.

This report containe praiset informenon on the resserch and The ORNL eludy of the eefoly-related espects of nuclear

-t

,._.; aconses of the nuclear power industy in uw eres power plant contos eyetems consists of two interrelated tasks:

of does reducson. It le bened on a data bene of informenon set (t) fedure made end enacts ennsyeis (FMEA) met idonelled up at me ALARA Center of Brochheven Neuonal Laboratory, single and muluple component fedures that mipt lead to signin-One purpose of mis report is to drew enenson to work in cent pient upsete and (2) computer modele met used mese tes-proyees and to enable resserchere end subscreers to obtain ures as innel condmone end traced the dynamic impoet on me turmer incormenon from ww invenuestore and protect managere.

coreal symem and remainder of the piant TNs report demorese ensormenon is provused on 100 prosecte, dMded acconsne to the simuseman of Ocones Unit 1, the erst plant ansysed. A arst-whether they are oriented to Enynooring Reneerch or to HeeNh prmcipios, best eenmate model wee developed and implement Physics Technology. Tlw report contane indcas on main cato.

ed on a hytetd computer consioung of AD4 enalog and PDP 10 gory, pro 6sct manager, principal inveselgator, sponsoring organi.

dguel mecNnes Contale wem pleoed prknergy on me snelog l

remon, contracung organisecon, and sub6sct. TNe is en IrWeel to use its interactve capahdley to almulate operator seson. The report. It is intended that periodc updelse be leeued whenever irunel step in model development was to deAne a sumable inter-sumceent metonal hee boon accumulated face betwoon the FMEA and computer simulomon teshe. The i

overall approach predominently used adsung advanced elete-r 10MRES/CfM410: AN ASSESSMENT OF SAFETY MARGINS IN og.Ww-st proomiures evadable in producson oodse w in Ww Nt-ZlRCALOY OXIDATION ANO EMORITTLEMENT CRITERIA eroture. At the apenee of generehty, enempts were made to

]

FOR ECCS ACCEPTANCE. WILLIFORD,R.E. Bettelle Memortal simpWy and stroemene proyamming, tenor it to e specac plant, Ineslute, Pacisc Northwest Laboratortes. April 1986. 71pp.

and improve computsuond speed and maneuverebdNy as com-000$t00000. PNL 500s. 30034.200.

pared wnh large producson codes. The use of conArmed tech-l Current Emergency Core Cochng System (ECCS) Acceptenos n6 ques helped with wereceton. Vandagen showed escement Cenerte for hght-weter reactore include certewi requirements po'*

eycoment veti plant data es wen es weri two widely used pro-tenung to calculemune of com performance durmg a Loseof ducton codes, RELAPS and TRAC, Cooient Accident (LOCA). The Beher Just correlamon must be used to ceiculate Zwceioy.eleem ontdelion, ceiculated peak 00URES/CH 4460 Vet: LIGHT. WATER-REACTOR FUEL SAFETY eledeng temperatures (PCT) must not exceed 1204 doyees SYSTEMS RESEARCH PROGRAMS. Quarterly Proyees j

eeneyede, and ceiculeled onwletion must not onceed 17 per.

Report,Apni June 1806. REST,J.; CHUNG.H.M. Argonne Neuen-t cent equsweient c6edeng reacted (17% ECR). The rninimum al Laboratory. March 1000, 31pp. 9006130344. ANL.06 71 Vll.

maryn of esfoty wee seemeted for each of these cenerte, bened 36063 036.

l on reeserch portormed in the last decade. Maryne were deaned Ttwo progrees report summertsee the Argonne Neuenal Labo-l es the amourWe of conservenom over and above the espected retory work performed during Apnt, May, and June 1996 on I

entreme values computed from the date beee et specded conn-water reactor esfoty protdome related to fuel and fuel ciedeng donce levole. The currongly regoed Seher Just omedemon corre-motortale. The reeeerch and development erees covered are lemon prov6 des mergne only over the 1100 degrees coneyede Trenoient Fuel Respones and Fiesson Product Release and Clad 4

to 1SOO degrees cer* grade temperature range et the 06% con.

Properuse for Code Vennceton.

l fidence level. The PCT meryne for thermal shock and handbng ladures are adequate et ontdelson temperatures above 1204 de-NURES/CR 4444: RELAPS/ MOD 2 CODE ASSESSMENT AT THE I

yees conterede for 210 and 100 seconde, ra -.;;4, et the IDAHO NATIONAL ENGINEERING LA80RATORY, 96% coredence level. ECR thermal shock and handhng mer.

WHEATLEY,P.D.; BOLANDER.M A.; DAVIS,C.B.; et al EG4G one et the 60% and $$% coredence levels, roepectrvely, range ideho, Inc. (sube, of EG4G, Inc.). March 1906. 34pp.

between 2% and 7% ECR for the Beher Joet correlamon, but 9004230120. EGO.2420. 35000.242.

t i

i

~

Main Citatiens and Abstracts 17 Independent assessment of the RELAPS code continued wth 25 e!ements was determined with the MELCOR Accident Con-the assessment of RELAPS/ MOD 2 during 1985. RELAPS was sequence Code System (MACCS). 60 nuclides of the 25 ele-assessed using a range of integral and separate-effects data.

monts were exarmned using the inventory of a 3412 MW PWR.

Semiscale tests S-UT-8, S-UT-6, and S-PL 4 simulahng small-The relative importance of each element was determined for break transients were used for assessme

Other tests, both early and long-term consequences. A number of elements GERDA 1605AA and Model Doiler-2, were a used. The could be important to early and long term doses if released in crossflow junction capability in RELAP5 was assessed using the sufficiently large quantities. However, to be important for early Electric Power Research Institute's single-phase liquid, sub-health effects, an element must be released to the environment channel-blockage test data. International Standard Problem 18 before there hae been sufficient time for an effective emergency was also reviewed as part of the assessment of RELAPS/

response.

MOD 2. Results of the independent assessments are document-ed in this report.

NUMEG/CR-4443: REACTOR PRESSURE VESSEL FAILURE NUREO/CR-44el: TORNADO CLIMATOLOGY OF THE CONTIG-PROlg&lTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSURIZED THERMAL SHOCK EVENTS.

UOUS UNITED STATES. RAMSDELL J.V.; ANDREWS,G.L Bat.

telle Memorial Institute, Pacific Northwest Laboratones. May SIMONEN,F.A.; GARNICH,M.R.; SIMONEN,E.P.; et at Battello 1986. 200pp. 8606190560. PNL-5697. 36624:113.

Mernonal Institute, Pacific Northwest Laboratories. April 1986.

The characterishes of tomadoes that were reported in the 251pp. 8605210422. PNL 5727. 36052:221.

~

contguous Uruted States for the penod from January 1,1954 A fracture mechanics model was developed at the Pacific through December 31,1983 have been computed from data in Northwest Laboratory (PNL) to predict the behavior of a reactor the Nabonal Severo Storms Forecast Center tomado data base.

pressure vessel following a through-wall crack that occurs The charactenstics summarized in this report include frequency dunng a pressurized thermal shock (PTS) event. This study con-and locations of tomadoes, and their lengths, udths, and areas.

tnbuted to a U.S. Nuclear Regulatory Comrmesion (NRC) pro-Tomado stnke and intensity probabilities have been estimated gram to study PTS risk. The model predicts the arrest, reinitie-on a regional basis, and these estimates have been used to tien, and direction of crack growth for a postulated through-wall compute wind speeds wth 10( 5),10(-6) and 10(.7) yr( 1) prob.

crack and thereby predicts the mode of vessel failure. A Monte-abilebes of occurrence. The 10(-7) yr( 1) wind speeds range Carlo type of computer code was etten to predict the pmbebil-from below 200 mph in the westem Urvted States to about 330 ities of the attemative failure modes wth the fracture mechanics mph in the vicinity of Kansas and Nebraska. The appendices properties of the various welds and plates of a vessel as contain extensive tabulations of tornado s'atistics. Variations of random variables. Plant-specific calculations for the Oconee-1, the charactenstics wthin the conbguous United States are pre.

Calvert Chffs-1, and H.B. Robinson-2 reactor pressure vessels sented in the summades. Separate tabulebons are provided for for the conditions of postulated transients predicted that 50% or the contiguous United States, for each state, for each 5 degree more of the through-wall axial cracks will tum to follow a cir.

and 1 degree latitude and longitude box, and for the eastem cumferential weld. This predicted failure mode results in a po-and westem United States.

tential vertically directed missile consisting of the upper head NUREG/CR-4443: HUMAN FACTORS IN ANNUNCIATOR /

missiles, as well as all potential horizontally directed fragmenta-ALARM SYSTEMS ANNUNCIATOR EXPERIMENT PLAN 1.

ROSCOE.B.J.; WESTON.L.M. Sandia Nabonal Laboratones.

tion type missiles, will be confined to the vessel enclosure May 1986. 64pp. 8607010419. SAND 85-2545. 36839.227.

cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure This report presents a plan for the experimental evaluahon of alarm reduction techniques on a full scale nuclear power plant modes that are most probable for postulated PTS events.

simulator. The alarm reduchon techniques include suppression NUREG/CR-4444: STATUS OF ACTIVITIES FOR INSPECTING of irrelevant alarms based on plant mode, time delay of alarm WELD OVERLAID PIPE JOINTS. GOOD,M.S.; VAN FLEET,LG.

groups, and separation of component status annunciators from Battelle Memortal Institute, Pacific Northwest Laboratories. Feb-alarms. Five hypotheses about the benefits of alarm reduction ruary 1986. 69pp. 8604170642. PNL-5729. 35606.343.

are stated. The independent vanables whose effects are to be evaluated in the experiment are identified and their possible Pacific Northwest Laboratory (PNL) evaluated the ultrasonic loteractions are desenbed. The performance measures are de-inspectabihty of weld overteid pipe joints. As part of this task, fined in terms of operator decision levels, operator task per-PNL la providing NRC staff with conclusions and recommende-formance, operator perceived work load, and operator control of tions concoming the effectiveness of ultrasonic inspections per.

key process variables. The data obtained from measurements formed on weld overiaid pipe joints. PNL evaluated data from of the performance measures in response to changes in the in-availr.ble technical literature, conducted expenments to deter.

dependent variables mil be analyzed by statistical analysis of mine the distortonal effects of weld overlay on ultrasound, and reviewed data from the weld overlay inspection development of-forts of the Electric Power Research Instituto NDE Center.

or conti dic ch o h

s Based on these reviews and expertments, PNL concluded that NUREG/CR 4447: RELATIVF IMPORTANCF OF IND M DtJAL ultrason6c inspection of weld overteld pipe joints has not been ELEMENTS TO REACTOR ACCIDENT CONSEQUENCES AS.

demonstrated to be reliable, for two reasons. First, insuff;cient SUMING EQUAL RELEASE FRACTIONS. ALPERT D J.;

data exists to demonstrate the reliable detection and sizing of CHANIN.D.I.; RITCHIE.LT. Sandia National Laboratones. March intergranular stress corrosion cracks. Second, the detection of 1986 36pp. 8605300237. SAND 85 2575. 36172:233.

unacceptable fabrication flaws contained within the weld overisy The retanve importance to offsite health and economic con-material has a low reliability due to poor signal to-noise ratios.

sequences of the radioactrve elements in a nuclear reactor core However, as current research and development programs lead is examined, assuming an equal fraction of the inventory of to a more comprehensive engineering database, these conclu-I each element is released. Studies of reactor accident source sions may change.

terms can use these results to identify the elements (nat are likely to be the more important contnbutors to o'fsite conse-NUREG/CR 4444: VENTING OF NONCONDENSIBLE GAS quences. Radionuchde inventones for a 3412 ard a 1518 MW FROM THE UPPER HEAD OF A dew REACTOR VESSEL thermal PWR and a 3578 MW DWR were calculated with the USING HOT LEG U BEND VALVES. WATERMAN.M E.;

SANDIA ORIGEN code. Of the 89 elemente considered t'y KULLDERG,C M.; WHEATLEY,P.O. EGAG Idaho, Inc. (subs. of ORIGEN. only about 25 could be important contnbutors to off.

EGaG, Inc) March 1986. 42pp. 8605290063. EGO 2436.

site consequences The relatrve radiological importance of the 36147:190.

l

18 Main Citations and Abstracts This report desenbos RELAPS/ MOD 2 thermal-hydraulic analy-wall of the test vessel. A model based on experiments is recom-ses of riswesible gas removal from Babcork and Wilcox mended for prodction of the steam generation rate, the particu-(B&W) reactor systems before and during natural circulation late temperatures and the frontal propagation speed during the conditions following a severe core damage accident. Hot leg U-quenching of superheated beds. The model includes a critenon bend vont valves were modeled as the pnncipal noncondensable for single-vs. two-stage quenching and assumes that the steam venting pathway. The analyses will assist the NRC in determing is superheated during its passage through the bed. The two-whether three B&W plants should receive permanent exemp.

phase hydrodynamics model proposed by Upinski for the tions from a reactor vessel upper head vent requirement. The steady-state dryout heat flux is used with modificetions for fea-raised-loop plant analysis determined the effect of a reactor tures of the quench process. It is concluded that the rate of vessel upper head vent line on plant refill and recovery of natu-containment pressurization resulting from quench of a super-ral circulation and showed that the vent line should be connect.

heated bed of solid debris by water from an overlying pool is ed to the loop with the pressurizer. The lowered-loop plant anal-limeted by two-phase countercurrent flow processes. On the ysis investigated the removal of noncondensible gas during nat-other hand, debris quenching imposes an additional constraint ural circulation and showed that 59% of the onginal inventory on conditions for bed coolability.

could be removed in 6900 s with a removal rate of 1% per 100

s. In both enalyses, significant amounts of noncondensible gas NUREG/CR-4497-NRCPAGE APPLICATIONS MANUAL were removed. Additionally, no fuel rod cladding temperature in-PICARD.R.R. Los Alamos Scientific Laboratory. Ap1 1986.

creesee were predicted during the periods of loop stagnation.

5 86042 LA M

60 NUREG/CR-4400: HISTORICAL

SUMMARY

OF THE HEAVY-viewed and related test procedures are desenbed. A FORTRAN SECTION STEEL TECHNOLOGY PROGRAM AND SOME RE-computer code (NRCPAGE) is provided that efficiently performs LATED ACTIVITIES IN LIGHT WATER REACTOR PRESSURE the calculations for one such procedure (Page's test) widely ad-VESSEL SAFETY RESEARCH. WHITMAN,G.D. Oak Ridge Na-vocated in the safeguards literature. Use of the code is illustrat-tional Laboratory. March 1986.180pp. 8605130122. ORNL-ed with a sample data set.

6259. 35953:151.

The accomplishments of the Heavy Section Steel Technology NUREG/CR-4498: FIELD TESTING OF WASTE FORMS CON.

Program and other programs having a close relatiorship to the TAINING EPICOR-il ION EXCHANGE RESINS USING LYSI.

development of information used in the assessment of light.

METERS. ROGERS,R D.; MCCONNELL J.W.; DAVIS,E.C.; et at water reactor pressure vessel integnty are reviewed. The earty EG&G Idaho, Inc. (subs. of EG&G, Inc.). June 1986. 56pp.

Pressure Vessel Research Committee planning, the principles 8607020359. EGG-2438, 36855:083.

contnbuting to program formulation, the role of the U.S. Atomic A field study was designed to monitor the release (if any) of Energy Commission, and the developments under the U.S. Nu.

beta-and gamma-producing radionuclides from solidified clear Regulatory Commission sponsorship are identified. The EPICOR.ll ion exchange resins. Both Portland Type I-11 cement need for major research and development accomplishments in and Dow vinyt ester-styrene waste forms are being tested in ly-fracture mechanics, heavy-section steel procurement, material simeter arrays located at Argonne National Laboratory in Illinois properties, irradiation effects, fatigue crack growth, and structur, (ANL-E) and Oak Ridge National Laboratory (ORNL). The study d testing are summartzed. The impact of program results on is designed so that continuous data on nuclide release and regulatory issues and the development of data used in the prep-movement, as well as environmental conditions, will be obtained aration of codes, standards, and guides are discussed. Continu.

over a 20-yr period. Details on waste form formulation, lysimeter ing activtties and recommendations for future research and de.

design, installation, instrumentation, and data acquisition and velopment in support of pressure vessel integrity assessments storage are provided.

are presented.

NUREG/CR-4303 V01: LONG TERM EMBRITTLEMENT OF NUREG/CR-4490 V01: LIGHT WATER-REACTOR SAFETY MA.

CAST-DUPLEX STAINLESS STEELS IN LWR SYS. Annual TERIALS ENGINEERING RESEARCH PROGRAMS:Ouarterty Report,0ctober 1984 - September 1985. CHOPRA,D.K.;

Progress Report. January. March 1985. SHACK W.J. Argonne CHUNG,H.M. Argonne National Laboratory. January 1986.

National Laboratory. March 1986. 84pp 8605290090. ANL 43pp. 8604180327, ANL-86-3. 35624:280.

75. 36149:001.

This progress report summartzes work performed by Argonne This progress report summertzes the Argonne Natonal Labo.

National Laboratory during the twelve months from October retory work performed during January February, and March 1984 to September 1985 on long-term embrtttlement of cast-1985 od water reactor safety problems related to out of-core duplex stainless steels used in 16ght water reactors.

materials. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors and NUREG/CR-4506: A SCOPING STUDY OF THE POTENTIAL EF-Lont orm Embnttlement of Cast Duplex Stainless Steels in FECTIVENESS OF AN OPERATIONAL SAFETY RELIABILITY T

PROGRAM IN ADDRESSING GENERIC SAFETY PROBLEMS.

MUELLER,C.J.; DEZELLA,W.A.; HCINEMAN,J.B.; et al. Argonne NUREG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL IN-National Laboratory. January 1986.137pp. 8605300232, ANL.

VESTIGATION OF QUENCHING OF SUPERHEATED DEDRIS 86-4. 36168:179.

BEDS UNDER TOP-REFLOOD CONDITIONS Final Report.

This report discusses the potential effectiveness of an Oper-GINSBERG,T.; KLEIN.J.; KLAGES.J.; et al. Drookhaven National ational Safety Reliability Program (RP) in addre. sing generic Laboratory. January 1986.154pp. 8605010580. DNL NUREG-safety problems. For the current study, the Program was broken 51951. 35795:341, down into 8 major elements: in-plant data morwtoring, industry.

A report is presented of a combined experimental and analyti-wide data monitoring, processing of performance data, determi-cal investigation of the quench characteristics of packed beds nation of performance goals / alerts, analysis of performance i

of superheeted partseles which are cooled by water supplied by data, management functions, priontization of crttcal items, and l

an overtying pool. Experiments were camed out using stainless corrective actions. Genenc safety problems investigated con-steel particles of diameter in the range 0 89 12.7 mm to simo-sisted of the TMI-related action lists, generic safety issues, late the partsculate core debris. Packed beds of up to 400 mm human factors issues, and severe accident precursors occumng j

in height were studied in a 108-mm diameter test vessel. The in 1980-8t; and all published post-TMI abnormal occurrences particles were heated to temperatures up to 1000K and were through June 1984. The study concludes that the Reliabihty Pro-1 quenched, with no internal heat source, using saturated water gram, as outlined, hs strong potential for offectively deshng as coolant at ambient pressure The steam generation rate was with the above safety issues, especially the new genenc issues.

measured along with temperatures within the bed and on the it also was shown to have potential for either preventing or fa-

T' Main Citations and Abstracts 19 cihtating problem closures to the great majonty of the aforemere-meneurate with the needs for radiation field measurements rou-tioned abnormal occurrences and severe accident precursors.

tinely encountered in radiation protection.

Thus, the Program outhned provides a basehne framework upon NUREG/CR-4516: INTERNATIONAL SAFEGUARDS AT FACIL1-which an effective hcensee Operational Safety Reliatukty Pro.

TIES EMPLOYING SPENT FUEL ROD CONSOUUTION.

yam can be further developed and implemented.

ROBERTS.F.P.; HARMS,N.L; SEWART,G.H. Battelle hemonal NUREG/CR-4808. AN OPERATIONAL SAFETY RELIABILITY Institute, Pacific Northwest Laboratories. April 1986. 43pp.

PROGRAM APPROACH WITH PECOMMENDATIONS FOR 8005050546. PNL-5480. 35843:172.

FURTHER DEVELOPMENT AND EVALUATION.

In a study sponsored by the Nuclear Regulatory Comrnession, MUELLER,C.J.; BEZELLA,W.A. Argcnne National Laboratcry.

the Padf;c Northwest Laboratory has analyzed the impacts that January 1986.146pp. 8605300260. ANL-86-5. 36175.170.

spent fuel rod consohdation may have on implementation of This report oocuments the development of a Reliability Pro.

Intemational safeguards. Potential processes for coneohdation gram structure, its functons, and associated activities for the at reactors and at away-from-reactor independent spent fuel operating phase of a nuclear power plant. This development storage fechties wwe examined and posssbie eMects on the ca-provides a baseline for tailoring utility reliabihty programs to spe.

pabihty to perform nuclear material flow and inventory verifica-cific equipment or operations. Previous recommendations of tion by International Atomic Energy Agency inspections were earlier industry and govemment groups and consultants are dis.

Identified. Facihty design features to enhance inspecten effec-cussed and factored into this structure. Synopses of current li.

tiveness and methods for venfication of flow and inventory are consee practices and regulatory requirements that address the identified.

same safety issues addressed by the Rehabihty Program are NUREG/CR 4519: TECHNOLOGY, SAFETY AND COSTS OF DE-provided. Reievent new industry initiatives or regulatory activl-COMMISSIONING NUCLEAR FUEL CYCLE FACILITIES CLAS-ties are cited. Studies demonstrating the Programs usefulness S4FICATION OF DECOMMISSIONING WASTE. ELDER,H.K.

In addressing operational safety issues are also cited. Finally' Battelle Memonal Institute, Pacific Northwest Laboratories. May cost / benefit considerations are identified. The Reliability Pro-1986. 32pp. 8606110722. PNL-5586. 36417:004.

gram es shown to provide a systematic framework for identHying The radioactive wastes expected to result from decommis-weaknesses or degradations in equipment and operations, de-sioning nucker fuel cycle facihties are rM and classified in termening the root-causes of these degradations, and integrating accordance with 10 CFR 61. Most of the wastes from the MOX problem solutions into day-to-day operations. The Program does plant (exclusive of the lagoon wastes) will require interim stor-not introduce new" concepts or procedures. Rather, it ties to-age (11% Class A 49 m3; 89% interim storage,383 m3). All of gether utshties, some much more than others. It is also shown the wastes from the U-Fab and UF6 plants are designated as that the Program factors proper operational safety reliability Class A waste (U-Fab 1090 m3, UF61259 m3).

considerations into the requerements in Technical Specifications, test and maintenance strategies, system configuration, and NUREG/CR-4520: PREDICTIVE GEOCHEMICAL MODELING OF actual performance of operating and maintenance functions.

CONTAINMENT CONCENTRATIONS IN LABORATORY COL-UMNS AND IN PLUMES MIGRATING FROM URAN!UM MILL NUREG/CR-4807: HECTR VERSION 1.5 USER'S MANUAL TAILINGS WASTE IMPOUNDMENTS.

Final Report.

DINGMAN S.E.; CAMP.A.L; WONG C.C.; et al. Sandia National PETERSON S.R.; MARTIN,W.J.; SERNE,R.J. Battelle Memorial Laboratories. April 1966. 369pp. 8607070488. SAND 86-0101.

Institute, Pacif;c Northwest Laboratories. April 1986. 157pp.

36901:004.

8605130393. PNL-5788. 35955.006.

TNs report desenbes the features and uses of HECTR Ver*

A computer-based conceptual chemical model was applied to soon 1.5. HECTR is a relatively fast-running, lumped-volume predict contaminant concentrations in plumes migrating from a 4

containment analysis computer program that is most useful for uranium mill tailings waste impoundment. The schos chosen for performing parametric studies. The main purpose of HECTR is inclusion in the conceptual model were selected based on re-to analyze nuclear reactor accidents involving the transport and views of the literature, on ion speciation/ solubility calculations combustion of hydrogen, but HECTR can also function os an performed on the column effluent solutions and on mineralogi-experiment analysis tool and can solve a limited set of other cal characterization of the contacted and uncontacted sedi-types of containment problems. New models added to HECTR monts. The mechanism of absorption included in the conceptual Version 1.5 include fan coolers, containment leakage, continu-chemical model was chosen based on results from semiselec-ous buming. and the capabihty to treat carbon monoxide and di-tive extraction experiments and from mineralogical characteriza-oxide. Models for the ice condenser, sumps, and Mark lli sup-tion procedures performed on the sediments. This aonceptual pression pool were upgraded.

chemical model was further developed and partially didated in laboratory experiments wf.ere assorted acidic uraniur,1 mill tail-NUREG/CR-4511: ASSESSMENT OF THE ADEOUACY OF THE ings solutions percolated through various sediments. This docu-CALIBRATIONS PERFORMED BV COMMEHCIAL CALIBRA.

ment contains the results of a partial field and laboratory valida.

TION SERVICES FOR IONIZING RADIATION SURVEY IN-tion (i.e., test of coherence) of this chemical model. Macro con.

STRUMENTS. COOKt-,R.H.; DOLECEK,E.H.; MOE H.J.; et al.

stituents (e.g., Ca, SO4, Al, Fe, and Mn) of the tailings solution Argonne National Laboratory. May 1986.125pp. 8606120902.

were predicted closely by considering their concentrations to bw 36434:122.

controlled by the precipitation / dissolution of solid phases. Trace TNs report discusses a blind study aimed at assessing the elements, however, were generalty predicted to be undersatur-adequacy of commercial services for the cahbration of portable ated with respect to plausible solid phase controls. The concen-radiation survey instruments. In addition to a quantitative analy*

tration of several of the trace elements were closely predicted sis of the achieved cahbration values, the results also address by considering their concentrations to be controlled by adsorp-i associated quehtative aspects. Data presented reviews the per*

tion onto the amorphous iron oxyhydroxides that precipitated.

formance of a cahbration serv 6ce for an Individual instrument cakbration as well as their overall performance for a variety of NUREG/CR-4526: FINITE ELEMENT ANALYSIS OF THE 2240 instruments. Collective cahbrator performance is addressed.

MW HTGR PCRV FUGELSO,LE. Los Alamos Scientific Labo-Certain recommendations on guidehnes for the development of ratory. April 1986. 43pp. 8606200248. LA 10663-MS.

appropriate standard cahbratnn procedures and ior documenta.

36649:266.

tion are provided. Although this study was somewhat hmited in Three-dimensional finite element calculations for the re-scope and could not cover all cabbration serv 6ces or all portable sponse of the prestressed concrete reactor vessel for the 2240 survey instruments, the results indicate that Current Cahbration MW HTGR which evaluated the stress distnbutions and concen-services generalty provide a level of acceptable accuracy com-trations were accomphshed. Constitutive equations utshzed in 1

l

._.m._

20 Main Citations and Abstracts this evaluation were linear elas6c, Von Mises elastic-plastic and tatively concluded that a correlation between partcle, particle the empencal Kotsovos-Newman concrete fit with and without energy, and material damage (as measured by changes in rne-steel reinforcing. Ultinwte values of the intomal pressures with-terial elongation and/or tensile strength) has been demonstrat-out trutial prestress were obtained. Also stresses in the a.inular ed.

concrete retaming cover over the steam generator were evalu-NUREG/CR-4548: CORRELATION OF ELECTRICAL REACTOR CABLE FAILURE WITH MATERIALS DEGRADATION.

NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WELDED STUETZER,0.M. Sandia National Laboratories. March 1986.

TYPE 304 STAINLESS STEEL PIPE: J-R CURVE CHARACTER-123pp. 8605290055. SAND 86-0494. 36148:140.

IZATION AND UMIT LOAD ANALYSIS. HAYS R.A.;

Complete circuit failure (shortout) of electrical cables typically VASSILAROS,M.G.; GUDAS,J.P. David W. Taylor Naval Re-used in nuclear power plant containments is invesugated. Fail-I search & Development Center. May 1986. 49pp. 8606180262.

ure modes are correlated with the mechanical deterturation of 36545:150.

the elastomeric cable matenals. It is found that for normal reac-An expenmental investigation was performed to determine the tor operation, electrical cables are reliable and safe over very fracture resistance of 4 in. diameter circumferentia&/ welded long penods During high temperature excurssons, however, type 304 staeniess pipe at 550 degrees fahrenheit (288 degrees cables pulled across comers under high stress may short out centigrade). Two crack geometnes were investigated. These due to conductor creep. Severe cracking will occur in short were a circumferential through wall crack (simple) and circum-times during high temperatures (>150 degrees centigrade) and ferential through wall crack supenmposed on a 360 degree in times of the order of years at elevated temperatures (100 de-radial crack on the inside diameter of the pipe (complex). Test grees centigrade 140 degrees centigrade). A theoretical treet-results were analyzed using J-integraf and limit load techniques.

ment of stress distnbution responsible for creep and for crack-Additionally, J-integral resistance curve tests were performed on ing by J. E. Reaugh of Science App;a 2,6., Inc. is contained in large plan-size compact tension specimens for comparison with the Appendix.

the pipe specimen results. Results of the J-integral analysis ind-cates that J-initiation for pipes containing simple cracks was ap.

NUREG/CR-4549: DETERMINATION OF APDENDIX K CON-proximately 1120 kJ/m2 (6400 in-lb/in2) and a factor of four do.

SERVATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE crease in J-irutiation between pipe specimens containing the PWR USING TRAC-PD2/ MOD 1 CODE. HOHATGI,U.S.;

simple crack geometry and shown at J-initiation between pipe YUELYS-MIKSIS, SAHA,P. Brookhaven National Laboratory.

specimens contairung the simple crack geometry and compact April 1986.

114pp.

8606200207.

BNL-NU"EG-51967.

tension specimens. The accuracy of the hmit load analysis was 36649:001.

variable for pipes containing the simple crack geometry with the A 200% cold leg break accident in a Westinghouse four-loop average predicted limit load calculated using the ASME Code RESAR-3S plant has been rqalyzed using the best-estimate flow stress being 8.7% higher than that actually attained in the code TRAC-PD2/ MOD 1/ Version 27 with updates. Three TRAC tests. The calculated limit loads based on the ASME Code now calculations have been performed. The first calculation used the stress were conservative for the complex crack cases.

best-estimate or realistic initial and boundary condrbons and scenarios, while the other two calculations, one with and one NUREG/CR-4539: INVESTIGATION OF TEARING INSTABILITY without locked rotor resistance, used the licensing conditions.

PHENOMENA IN ASTM A106 STEEL LINK,R.E.; HAYS,R A-These calculations produced peak cladding temperatures David W. Taylor Naval Research & Development Center. May (PCTs) of 800.5 degrees K.1072 degrees K, and 1153 degrees 1986. 28pp. 8606170007. 36563;102.

K, respectively. Comparison of these results with the Westing-An expenmental investigation was performed to evaluate tear-house licensing calculatior s Ivrformed in accordance with *he ing instability theory by varying the applied tearing modulus, guidelines in Appendix K of 10CFR50 shows an overall safety (T) applied, so that fracture instability would be initiated at vari-margin of 663 degrees K, of which 352.5 degrees K is due to ous levels of stable crack extension. This is an extension of conservatrve initial and boundary cordtions and scervito. The past investigations of tearing instability theory in that crack ex-remaining 310.5 degrees K is due to conservatrve physical tension was monitored continuously using the D.C. potential models. The locked rotor resistance contributed about 81 de-drop technique, enabling the apped and material tearing moduli grees K in PCT.

to be calculated at the point of Iristability. The resutts of this in-i l

vestigation indicate that, in trust cases, fracture instability oc-NUREG/CR-4557: A REVIEW OF ISSUES RELATED TO IM-curred when the difference between the applied and material PROVING NUCLEAR POWER F~. ANT DIESEL GENERATOR tearing moduli was on the order of 10%. Variations in the load RELIABILITY. HIGGINS,J.C.; C2AJKOWSKI.C.J.; TINGLE A.G.

versus displacement records of the specimens near maximum Brookhaven National Laboratory. April 1986. 221pp.

load due to local instabilities and friction in the load train pre.

8604280425. BNL NUREG-51969. 35737:286.

cluded measurement of a smooth applied tearing modulus The purpose of the report was to analyze the data and rec-curve.

ommendations made by the various groups associated with nu-

' E #

- - '0*""

NUREG/CR-4543: FIRST RESULTS FROM ELECTION-PHOTON ommendations of each group. Those groups making recommen.

DAMAGE EQUIVALENCE STUDIES ON A GENERIC ETHYL' dations includod nuclear utilities, industry organizations (such ENE. PROPYLENE RUBBER. DUCKALEW,W.H. Sandia National as INPO a-o ASME), DG manufacturers or vendors, foreign DG Laboratories. April 1986. 43pp. 8605290081. SAND 86-0462.

cwrs, ttw Ady'sory Committee on Reactor Safeguards (ACRS),

36150-1'l7.

and some miscellaneous groups. The report presents those As pert of a simulahr adequacy assess.oent program, the rel-areas having broad documented support.

atrw effectiveness of electrons and photons to produce damage in a genenc ethylene-propylene rubber (EPR) has been NUREG/CR-4561: FIPaC USER'S MANUALA COMPUTER investigated. The investigation was limited in extent in that a CODE TO SIMULATE FIRE ACCIDENTS IN NUCLEAR FACIL1-single EPR material, in three thicknesses, was exposed to TIES. NICHOLS,8.D.; GREGORY,W.S. Los Alamos Scientific Cobatt 60 photons and three electron beam energies. Basing Laboratory. April 1986. 308pp. 8605130352. LA 10678-M.

material damage on changes in the EPR mechanicel properties 35959.074.

elongation and tensile strength, we observed that EOR damage This user's manual supports the fire accident analysis com-was a smoothly varying function of absor"ed energy and inde-puter code FIRAC. FIRAC is designed to estimate radioactive pendent of irradiating particle type. EPR damage tracked equal-and nonradioactive source terms and to predict fire-induced fy well as a function of both incident particle energy and materi-flows and thermal and material transport within the ventilation al front surface dose Based on these prehminary data, we ten-systems of nuclear fuel cycle facilities. FIRAC has been expand-l l

l l

Main Citations and Abstracts 21 ed and modified to include the capabilities of the zone-type on J-tearing theory. This aproach is intended to provde a con-I u.s.p.itirent fire model computer code FIRIN developed by servative approxiaten of the appleed crack driving parameter, J, Battelle Pacific Northwest Laboratories. The two codes have for postulated through-wall leakage-size cracks in nuclear power been coupled to provide an improved simulation of a fire-in-plant pipes. Piping integrity evaluations can then be accom-I duced transient within a facility. The basic matenal transport ca-plished for various loading conditions and assumed flaw sizes.

pability of FIRAC has been retained and includes estimates of Because the method can be used to obtain a rather rapid com-entrainment, convecten, depositen, and filtrabon of material.

puter generated approximation of the applied crack driving pa-Also, the interrelated effects of filter plugging, heat transfer, gas rameters, NRC evaluation of application or licensee submittals dynamics, material transptat, and fire and radioactive source can be accomplished in an expeditious manner without resorting terms are simuisted. This report summarizes the physical to elaborate firute element techniques. The NRC program I

models that desenbe the gas dynamic, material transport, heat should not be considered as fixed in time. As piping fracture transfer, and source term processes and illustrates how a typi-mechanics technology matures, it may be refined in the future.

cal facility is modeled using the code. The modifications re-quired to couple the code to FIRIN also are presented. Finally, NUREG/CR-4579: APPLICATION OF THE KEY CURVE AND the input and code-calculated output for several sample prob.

MULTI-SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE lems that illustrate some of the capabihties of the code are de.

TESTING OF ALLOY STEEL JOYCE,J.A. U.S. Naval Academy, scnbed.

Annapolis, MD. HACKETT.E.M. David W. Taylor Naval Re-search & Development Center. April 1986. 37pp. 8604280316.

NUREG/CR-4568: A HANDBOOK FOR QUICK ESTIMATES:A 35751:037.

Method For Developing Ouick Approximate Estimates Of Costs For Generic Actons For Nuclear Power Plants. BALLJ.R. Ar-J-integral R-curve tests were performed on three point bend specimens of a 3-Ni steel at three loading rates - quasi-static, gonne National Laboratory. Apnl 1986.134pp. 8604290072.

ANL/EES-TM-297. 35756.005.

intermediate (25mm/sec) and a drop tower rate (2.54mm/sec).

The key curve and multi-specimen procedures were employed This document provides guidance for developing " quick" ap-for the higher rate tests and this investigation is focused primar-proximate estimates of the cost of implementing generic regula-ily on details of the test metnod development. The multi-speci-tory requisements for nuclear power plants. A method is pre-sented for relating the known constructen costs for new nucle-men and key curve techniques were found to yield upper shelf J-R curves which were in substantial agreement at the elevated ar power plants (as contained in the Energy Economic Data Base) to the cost of performing simstar work, on a back-fit basis, loading rates. Numerical smoothing techniques required to apply at existing plants. Cost factors are presented to account for a key curve method appear to separate the oscillatory high fre-quency component from the load-displacement record. For the variations in such important cost areas as construction labor 3-Ni steel tested for this investigation both J(Ic) and T were productivity, engineenng and quality assurance, replacement found to be elevated with increasing loading rate.

energy, reworking of existing features, and regional variations in the cost of materials and labor. Other cost categories ad-NUREG/CR-4540: STONY BROOK SEISMIC NETWORK ON dressed in this handbook include those for changes in plant op-LONG ISLAND.NEW YORK. Final Report (September 1979 -

erating personnel and plant documents, licensee costs, NRC March 1985). LIEBERMANN,R.C.; THURBER,C.H. New York, costs, and costs for other govemment agencies. Data sheets, State Univ. of, Stony Brook, NY. April 1986. 113pp.

worksheets, and appropriate cost algonthms are included to 8605160018. 36034:081.

guide the user through preparation of rough estimates. A The seismology group in the Department of Earth and Space satnple estimate is prepared using the method and the estimat-Science at the State University of New York at Stony Brook cur.

ing tools provided. This document is a stand-alone supplement rently operates two short period seismic stations on Long to NUREG/CR-3971,"A Handbook for Cost Estimating."

island, New York. These stations are part of the larger North-NUREG/CR-4569 A REVIEW OF THE SEVERE ACCIDENT RISK eastern United States Seismic Network (NEUSSN) operated by REDUCTION PROGRAM (SARRP) CONTAINMENT EVENT several university groups in New York, New Jersey, Pennsyfva-TREES.

Wisconsin, Univ. of, Madison, W1. May 1986.190pp.

nia and New England. The Stony Brook stations provide data 8606170012.36562:230.

for specific studies of earthquakes in the vicinity of Long Island, As part of the Severe Accident Risk Reduction Program, re-including southeastern New England and offshore areas. The searchers at Sandia National Laboratones have constructed a installation, operation and maintenance of the Stony Brook sta-proup of containment event trees to be used in the analysis of tions has been supported primarily by funds from the United key accident sequences for light water reactors (LWR) during States Geological Survey (USGS), the United States Nuclear postulated severe accidents. The ultimate goal of the program Regulatory Commission (NRC), and SUNY at Stony Brook. This is to provide to the NRC staff a current assessment of the risk report discusses network operation and results of research from severe reactor accidents for a group of five light water re, using NEUSSN and related data during Phase I of the contract actors. This review specifically focuses on the development and with the NRC (1979-1985).

construction of the containment event trees and the results for NUREG/CR-4541: DRYOUT FRONT MODELING FOR PWR containment failure probabihty, modes and timing. The mport first gives the background on the program, the review cnteria' THERMAL HYDRAULIC ANALYSIS. BROWN,0.; GRIFFITH.P.

and a summary of the observations, findings and recommenda.

Massachusetts Institute of Technology, Ca nbridge, MA. May tions. Secondly, the individual reviews of each committee 1986. 66pp. 8605300523. 36163:058.

member on the event trees is presented. Finally, a review is An expenmental and analytical study has been camed out on ovided e computer model used to construct and evaluate the effects of a steamline break on the heat transfer on the

~

secordary side of a PWR U-tube steam generator. The study is focused on prediction of the dryout front location during blow.

NUREG/CR-4572: NRC LEAK BEFORE BREAK (LBB NRC) down. Prediction of the dryout front is essential for determining ANALYSIS METHOD FOR CIRCUMFERENTIALLY THROUGH.

the wetted heat transfer area. The bulk of the heat transfer from WALL CRACKED PIPES UNDER AXIAL PLUS BENDING the pnmary to secondary side occurs in the wetted area. To un-l LOADS. Topical Report.

KLECKER.R.;

BRUST,F.;

derstand the behavior of the dryout front, an expenment using a WILKOWSKl.G. Battelle Memonal Institute. Columbus Laborato-simulated steam generator test facility was performed for a wide rios. May 1986. 97pp. 8605290019. BMI-2134. 33149.085.

range of simulated steamline break sizes. The thermoenuple The fracture mechanics analysis procedure used by the NRC data obtained from the expenment was used to determir.e the to evaluate utility leak-before-break submittals is desenbed in dryout front propagation. Examination of the data provided this report. This methodology is an estimation technique based godelines for modehng the multiple water levels, the flooding

22 Mein Citations and Abstracts charactonstics of tube support plates were determned and tween a model and a prototype. The present study demon-input into the code. Using the predicted womr levels and the strates a new approach to this classecal problem using the two-thermocouple data, the dryout zone could then be deternned.

phase flow scaling criteria. It indicates that a real time scaling is An empwical correlation for dryout zone height as a function of not a practical soluton, and a scaled-down model should have the superficial velocity of the vapor is presented. The zone turns an accelerated (shortened) time scale. One of the most impor-out to be O to 8 inches in thickness.

tant results is the new scaling methodology for simulating pres-NUREG/CR-4682: TEMPERATURE EFFECTS ON THE SOLUBIL-swe tra.m it is obtained by considenng me changes of me ITY AND SPECIATION OF SELECTED ACTINIDES.

M propedy groups which appear within the two-phase similari.

NITSCHE,H. Lawrence Berkeley Laboratory. June 1986. 52pp.

ty parameters and the single-phase to two-phase few transition 8607000452. LBL-20387. 36934:290.

parameters. Some sample calculaties are given fw modehng a The objective cf this study was to determene experimentally hig4 pressure water system by a low-pressure water system and the trends in the temperature dependence of the solubihties and Freon systems.

spuciation of neptunium, plutonium, and americium. In ground-NUREG/CR-4588 V01: SOIL-STRUCTURE INTERACTION.Vol waters hydroxide and carbonate anions are consedered to play a moet important role in the formation of insoluble precipitates 1: Influence Of Layerity. PHILIPPACOPOULO Brookhaven Na-and soluble complexes of actinides (ALLARD 1982, NRC 1984).

tional Laboratory. April 1986. 96pp. 8606230295. BNL-NUREG-51983. 3668t272.

Therefore, solubehty measurements were made in 0.01 M non-complexing NaC10(4) soluton at pH 6.0 and 8.5 at both 25 de-This study has been performed for the Nuclear Regulatory grees centigrade and 60 degrees centigrade. The total cart)on.

Commission (NRC) by the Structural Analyses Division of Brook-ate concentration was held constant at 120 ppm, a value close haven National Laboratory (BNL). The study was conducted to that found in some groundwater. This study showed no clear during the fiscal year 1985 on the program entitled " Bench-change in solutzitty as the temperature changed from 20 de.

marking of Structural Engineenng Problems" sponsored by grees centigrade to 60 degrees centigrade. Many of the solid NRC. The program considered three separate but complemen-phases that formed, although crystalline, remain unidentified be.

tary problems, each associated with the soil-structure interac-cause of the lack of reference data in the Eterature. A compart.

tion (SSI) phase of the seismic response analysis of nuclear son of the experimental results with solubikty predictions from plant facihties. In this volume, " Influence of Layering" approxi-modehng calculations (SILVA 1984) showed significant differ, mate analytic formulations were developed to obtain frequency onces. These differences are probably caused by the existence dependent interaction coefficients apphcable to the case of a in the experiment of unknown solubihty-controlling solid phases rigid circular disk at the surface of an elastic, layered halfspace.

that were not included in the modeling data base. The results of With these approximations, SSI calculations can be made in a this study demonstrate the need to study radionucHde solubili-direct manner, eliminating the need to perform complicated and ties experimentally in groundwaters from a prospective reposi.

expensive interaction calculations by means of finite element or tory site to accurately predict the solubility hmits needed for li-CLASSI-type computer programs. Such simphfications are par-consang a nuclear waste repository.

ticularfy important when performing prenminary designs of nucle.

ar a w

n ahnpW to cM me adequacy of me NUREG/CR-4583 V01: DEVELOPMENT AND VALIDATION OF A p

cunpuw h h rnemods developed we ap REAL-TIME SAFE-UT SYSTEM FOR THE INSPECTION OF phed to both rigid and flexible structural models of actual plant LIGHT WATER COMPONENTS. Semi-Annual Report For April fa Bom ham and tranM responus we cone 1984-September 1964.

DOCTOR,S.R -

BUSSE.L.J -

CRAWFORD,S.L; et al. Batteile Memorial Institute, Pacifi; ed Wng me approximate and exact inwacnon coems.

o em a n func s aM ha reponse spectra sM Northwest Laboratories. May 1986. 83pp. 8605210471. PNL-

$622. 36050:108.

good W ahon.

The Pacific Northwest Laboratory is working to design, fabri-NUREG/CR-4648 V02: SOIL-STRUCTURE INTERACTION.Vol cate, and evaluate a real-time flew detection and characteriza-

2. Influence Of Lift Off. MILLER,C.A. Brookhaven National Labo-tion system based on the synthetic aperture focusing technique ratory. April 1986. 62pp. 8606200277. BNL-NUREG-S t 983.

for ultrasonic teseng (SAFT-UT). The system is for inservice in-36650:001 spection of light water reactor swe,n.6ts. Included objectives This study has been performed for the Nuclear Regulatory of tfus program for the.ucicer Regulatory Commission are to Commission (NRC) by the Structural Analysis Drvision of Brook-develop procedures for % stem cahbration and field operation, haven National Laboratory (BNL). The study was conducted to vandate the system through laboratory and field inspections, and to generate an engineering data base to support ASME during the fiscal year 1985 on the program entitled " Bench-Code acceptance of the technology. This progress repat marking of Structural Ergneenng Problems" sponsored by NRC. The program considered three separate but complemen.

cov the programmauc work from April 1984 through Septem-tary problems, each associated with the soil-structure interac.

tion (SSI) phase of the seismic response analysis of nuclear NIHetrG/CR-49adt RF00CED PRFMilRE AND FLUID TO FLUID plant facilities. This report presents a summary of the work per.

SCALING LAWS FOR TWO-PHASE FLOW LOOP. KOCAMUS-formed defirung the influence httoff has on the seismic response TAFAOGUL: ISHil,M. Argonne National Laboratory. April 1986.

of nuclear power plant structures. The standard lumped param-48pp. 8606200237. ANL-86-19. 36649:310.

eter analysis method was modified by incorporating into the Scanng criteria for a natural circulation loop under single-lumped sod / structure interaction model nonhnear foundation phase and two. phase flow conditions are dertved. The criteria stiffness. Additional interaction damping is included to account include the effects of primary fluid properbes so one can use for for the energy dissipated as a porbon of the foundation which modehng a high-pressure water system by a low-pressure water has separated comes back into contact with the soil. Data is system or by a different fluid system. In addition to the standard presented which identifies the peak acceleration required to single-phase flow dimensionless groups, it is shown that the cause fiftoff. For parameters typical of nuclear power plant phase change, subcoohng, drift-flux and Froude numbers are structures hftoff was fourd to occur when the peak accelera-particular1y important for a two phase flow simulation. Based on tions are in the range of 0.3 to 0.6 G's. Studies were then per-the results obta6ned from the rigorous simdarity ana ysis, practi-formed to evaluate the consequences of neglecting hftoff when s

cal appucahons for desegrung a scaled-down model are consid-it occurs. Significant special differences were found for the rock-ered. Special emphasis is given on scahng at reduced pressure ing and vertical motxms between the analyses performed con-levels relative to a prototype system and fluid to fluid properties sidering and neglecting hftoff effects. It was found that peak ac-due to a large number of simdarity groups to be matched be-celershons 1.33 that were required to cause hftoff (0.4 to 0.8

Main Citations and Abstracts 23 G's) result in unconservative response spectra when the Eftoff ed consequences. Matigating actions designed to prevent con-effects are neglected in the SSI analyses.

tainment failure as well as actions desagned to prevent core NUREG/CR-4568 V03: SOIL-STRUCTURE INTERACTION.Vol men gun containment fahe are also discussed. De stdy 3: Influence Of Ground Water. COSTANTINO C.J. Brookhaven concludes mat, whHe sane of me sequences couW potenbah result in cwe ne and cause segnScant releases of 6ssion Nabonal Laboratory. April 1986. 84pp. 8607010437. BNL-products, me frequency of mese sequences are M M be NUREG-51963. 36838:093 suf6cieUy sman to remove mem as signi6 cant cMnbutws to This study has been performed for the Nuclear Regulatory pubHc risk.

Commession (NRC) by the Structural Analysis Drvisson of Brook-haven National Laboratory (BNL). The study was conducted NUREG/CR-4595: ENHANCEMENT TO THE LAFM COMPUTER during the fiscal year 1985 on the program entitled ' Bench-CODE. BAARS R.E. Los Alamos Scientific Laboratory. April marking of Structural Engineenng Problems" sponsored by 1986. 25pp. 8606200210. LA-10711-MS. 36646:081.

NRC. The program conssdeved three separate but complemen-Segnificant modifications to the Los Alcmos Failure Model tary problems, each associated with the soil-structure interac-(LAFM) computer code are desenbed. The material in this tion (SSI) phase of the seestruc response analysis of nuclear report is limited to changes made since a desenpbon of LAFM plant facihties. This report presents a summary of the first year's was pubhshed in 1979. The most significant improvement allows effort on the subsect of the influence of foundation ground water continuation of the calculation beyond exhaustion of void space on the SSI phenomenon. A finite element computer program within the cladding. Updated input instructions also are included.

was developed for the two-phased formulation of the combined soil-water problem. This formulation is based on the Biot dy.

NUREG/CR-4601: TECHNICAL CONSIDERATIONS AFFECTING namic equations of motion for both the solid and fluid phases of PREPARATION OF ION-EXCHANGE RESINS FOR DISPOSAL a typical soil. Frequency dependent interaction coefficients were BOWERMA,B.S.; PICIULO,P.L Brookhaven National Laboratory.

generated for the two-dimensional plane problem of a rigid sur.

May 1986.

103pp.

8605300437.

BNL-NUREG-51987.

face footing moving against a saturated Enear soil The results 36162:009.

indicate that interaction coefficients are significantly modified as Three incidents invoNing low-level waste (LLW) from separate compared to the comparable values for a dry soil, particularfy nuclear power plants, i.e., dewatered ion-exchange resins or for the rocking mode of response. Calculations were made to dewatered filter media, occurred during 1983 and 1984. This study the impact os the modified interaction coefficients on the report summarizes and reviews the investigations into the response of a typical nuclear reactor building. The amplification causes of each incident. Factors unique to each incident are factors for a stick model placed atop a dry and saturated soil discussed and remun,endetions are given on the basis of were computed. It was found that pore water caused the rock.

these factors which may help limit such occurrences in the ing response to decrease and translational response to in.

future.

crease over the frequency range of interest, as compared to the NUREG/CR-4602: UNIQUENESS OF BOILING WATER REAC-response on dry son.

TOR PRIMARY WATER CHEMISTRY. Final Report, October NUREG/CR-4549: REVIEW OF SELECTED AREAS OF YANKEE 1985. March 1986. FOX,M.J. Fox Consulting Ertneers & Ge-ROWE PROBABILISTIC SAFETY STUDY.

Brookhaven Na-ologists. May 1986.125pp. 86062500J9. FOX /NRC 8601.

tional Laboratory. June 1986.87pp.8607090187. BNL-NUREG-36738:172.

51984.36932:071.

This report is a detailed review of boiling water reactor (BWR)

The Yankee Nuclear Power Station Probabilistic Safety Study primary water chemistry. The main question addressed is has been reviewed in three specific areas. These areas are (1) whether the primary water of a BWR is unique from plant to treatment of initiating events, (2) treatment of human actions, plant with respect to its ability to facilitate intergranular stress and (3) treatment of the emergency ac and dc power systems, corrosion cracking (IGSCC) of sensitized Type 304 stainless The results reported here are based on three individual and steel. The results indicate that BWRs are unique with respect to highly-focused reviews. Therefore, the conclusions offered are the ability of their primary water to facilitate IGSCC. This unique-based within the context of each individual review, russ exists under high purity water widsOne, and is not a U' '*"

NUREG/Cf' 1694: ESTIMATED SAFETY SIGNIFICANCE OF GE-P *" "'

E"**'Y **

NERIC SAFETY ISSUE 61. LEHNER.J.R.; PERKINS,K.R -

O'

'O*

ECONOMOS.C. Brookhaven National Laboratory. June 1986-ma pwa ng in a wa CMW pwan 119pp.8607010423.

BNL-NUREG-51986. 36838:179' cident se-M *" **

0*"

drogen ratios are significa'ntly different from o' ne plant to an-The potential threat posed by transient initiated ac quences irwolving BWR systems capable of releasing steam other. These variations appear to be real and are difficult to ex-into the wetwell airspace and thereby pressunzing the contain-plain with data presently available. It is evident that a serious mer t has been examined. This study estimates the hkelihood of gap exists in the present day understanding of BWR primary a rupture in one of a number of high pressure steam lines which water chemistry. The report also points out that hydrogen water pass through the wetwell aarspace before entering the suppres-chemistry does not eliminate stress corrosion cracking, but sion pool. If the broken steam line is connected to an active changes the mode of cracking from intergranular to transgranu-steam source, such as a stuck open relief valve, which supplies lar, and that the practical significance of the transgranular steam at a high enougn rate and over a long enough interval, cracking should not be discounted without sufficient experimen-the suppression pool bypass may result in containment over-tal evidence

presstire failure. The three BWR plant systems identified as having components whose failure could lead to a steam dis-NUREG/CR-4603: APPRAISING ATMOSPHERIC TRANSPORT charge into the wetwell airspace and pressurization of the con-AND DIFFUSION MODELS FOR EMERGENCY RESPONSE tainment are: the Main Steam Relief Vanes and associated dis-FAC!LITIES. SAGENDORF.J.F. Commerce, Dept. of, National charge lines, the High Pressure Coolant injection turbine ex-Oceanic & Atmospheric Administration. FAIROBENT,J.E. NRO -

haust and the steam condensing relief lines in the Residual No Detailed Affiliation Given. May 1986. 77pp. 8606110023.

Heat Removal System. This study outlines the postulated acci-36421:013.

dent sequences, estimates their frequency, and calculates con-As described in Supplement 1 to NUREG-0737, the NRC will tainment response to the proposed steam discharges using the conduct post-implementation reviews of emergency response computer code CONTEMPT 4. Based on the predicted contain-facilities (ERFs) at operating nuclear power plants. These re-ment response, the investigation estimates the probability of views include examination of the dose assessment methodology core melt, the possible fission product release and the associat-for emergency response. This report describes a methodology

24 Main Citations and Abstracts for evaluating the atmosphenc transport and diffusion models tion can be discussed in a positive framework with an assess-incorporated in dose consequence assessment by O.niwk.,6 ment of current capabilities and future needs in this area of re-with reference models. Two Gaussian puff-advoction rnodels, search. This paper provides informatLn and sets of data that one segmented plume model and one straight line Gaussian will be useful to the modelers in meeting the objectives of the plume model have been identified as reference models. These modeling study. The information and data sets include (1) a de-models are exercised with a series of data sets, and serve as scription of the experimental design and rnethods used in ob-points of reference for model compensons. A series of data taining solute transport data, (2) supporting data that may be sets have been assembled to exercise model performance in a useful in modeling the data set of interest, and (3) the data set venety of meteorological situations, ranging from simple invar-to be modeled.

lant conditions to more complicated conditions of river valley or coastal environments. These data sets are intended to be of.

NUREG/CR-4619: STRESS CORROslON CRACKING TESTS ON fered to licensees prior to the ERF appraisal of their particular HIGH-LEVEL-WASTE CONTAINER MATERIALS IN SIMULAT.

facility.

ED TUFF REPOSITORY ENVIRONMENT 3.

ABRAHAM,T.;

JAIN,H.; SOO,P. Brookhaven National Laboratory. June 1986.

NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND 85pp.8607080462. BNL-NUREG-51996. 36936-015.

GAMMA IRRADIATION ON THE MECHANICAL PROPERTIES Types 304L,316L and 321 austenitic stasniosa steel and Inco-OF HIGH DENSITY POLYETHYLENE. SOO.P.; ARORA,H.;

loy 825 are being considered as candidate contasner materials SWYLER,K.J.; et al. Brookhaven National Laboratory. June for emplacing high level waste in a tuff repository. The stress 1986.106pp.8606230288. BNL-NUREG-51991. 36682:006.

corrosion cracking susceptibility of these materials under simu-An evaluation was made of the effects of environment and lated tuff repository conditions was evaluated by using the gamma irradiation on the short. term tensile and creep proper

ties of Marlex CL-100, a highly cross-linked high-density poly-notched C-ring method. The tests were conducted in boiling i

synthetic groundwater as won as in the steam / air phase above ethylene. This material is being considered as a constructional the boiling solutions. AH specimens were in contact with material for a low-level radioactive waste high-integrity contai*

crushed Topopah Spring tuff. The investigation showed that mi-er. It was found that the chemical environments studicd could crocracks are frequently observed after testing as a result of be beneficial or detnmental to strength and ductility depending stress corrosion cracking or intergranular attack. Results show-on the type of mecharucal property test and the nature of the ing changes in water chemistry during test are also presented.

Exposure of the polyethylene to the chemical environment.

Gamma irradiation to a sufficiently high dose prior to tensile or NUREG/CR-4620: METHODOLOGIES FOR EVALUATING LONG-creep testing increased the strength and decreased the ductility.

TERM STABILIZATION DESIGNS OF URANIUM MILL TAIL-In-test irradiation, however, could increase or decrease the INGS IMPOUNDMENTS. NELSON.J.D.; ABT,S.R.; VOLPE.R.L; creep rate depending on the dose rate and applied stress.

et al. Colorado State Univ., Ft. Collins, CO. June 1986.157pp.

NUREG/CR-4609: EFFECTS OF EARTHOUAKES ON UNDER-8606260211. ORNL/TM-10067. 36785:164.

GROUND FACILITIES.Literaturo Review And Discussion Uranium mill tailings impoundments require long-term (200-CARPENTER D.W.; CHUNG D.C. Lawrence Livermore National 1000 years) stabiHzation. This repwt reviews currently available Laboratory. June 1986. 61pp. 8607080458. UCID-20505.

methodologies for evaluating factors that can have a significant 36935:009.

influence on tailings stabilization and develops methodologies in A review of literature concerning effects of ground motion on technical areas where none presently exist. Min operators can underground facilities has been completed, and an annotated use these methh to assist with (1) the selection of sites bibliography has been prepared. This information provides fw mix tailings impwndments, (2) the design of stable impme useful background for the science and engineering of under-ments, and (3) the development of reclamation plans for exist-ground nuclear waste management facility development. While ing impendmets. These mew wwid also be useful some conflicts are evident in the literature reviewed, the follow, for regulatory agency evaluations of proposals in permit or li-ing tentatrve conclusions may be drawn from the available infor.

cense appHcations. Methodologies were reviewed w developed mation: (A) Darr2ge is expectable if fault displacement occurs in the following technical areas: (1) prediction of the Probable through a site, but damage from shaking alone is generally con-Maximum Precipitation (PMP) and an ewvw.nying Probable w

fined to facilities located within the epicentral region and may Maximum Flood (PMF); (2) prediction of the stability of local and be less than to surface facilities at the same site. (B) Seismic regional fluvial systems; (3) design of impoundment surfaces re-data is mixed, but favors reduction of amplitude with depth; ob-sistant to gully erosion; (4) evaluation of the potential for sur-servations appear quite dependent upon station characteristics.

face sheet erosion; (5) design of riprap for protecting embank-(C) The frequency content of earthquake motions is important ments from channel flood flow and overland flow; (6) selection to the stability of underground openings and the applicability of of riprap with appropriate durability for its intended use; and (7)

Ettenuation relationships developed in areas where geologic evaluation of oversizbg required fu marginal quaHy riprap, and tectonic characteristics favor high attenuation rates to mid-NUREG/CR-4621: FLOW VISUALIZATION EXPERIMENT ON continental sites is questionable. (D) Model studies indicate HOT-LEG U-BEND TWO-PHASE NATURAL CIRCULATION problems for shafts and the potential for problems with waste-handling equipment in shafts. The results of the review indicate PHENOMENA. KIM.S.B.; ISHil,M. Argonne National Laboratory.

the need to assure that site-specific response spectra and at-May 1986. 37pp. 8606200220. ANL-86-27. 36648:105.

tenuation relationships are developed for pioposed sites, and A hot-leg U-bend experiment was p Led. The experimen-that detailed assessments of seismic aspects of shaft designs, tal condition simulated the two-phase flow in a B&W primary hoists and in-shaft waste-handling equipment are required.

loop during a small break loss of coolant accident or during some other abnormal transients. The loop design was based on NUREG/CR-4615 V01: MODELING STUDY OF SOLUTE TRANS-the scaling criteria developed previously and the loop was oper.

PORT IN THE UNSATURATED ZONE. Information And Data ated either in a natural circulation mode or in a forced circula-Sets. POLZER.W.L; FUENTES,H.R.; SPRINGER,E.P.; et al. Los tion mode using nitrogen gas and water. The two-phase flow re.

Alamos Scientific Laboratory. May 1986.134pp. 8606250043.

gimes at the hot-leg were identified on the basis of visual obser-LA.10730-MS. 36738:243.

vation. The phase separation at the top of the inverted U-bend The Environmental Science Group (HSE-12) is conducting a was observed at low gas flow rate. The void fractions were study to compare various approaches of modeling water and measured using differential pressure transducers and compared solute transport in porous media. Various groups representing with the prediction from the drift-flux model. The natural circula-different approaches will model a common set of transport data tion flow interruption occurred in two different modes, namely, so that the state of the art in modeling and field experimenta-quasi-periodic and semi-permanent modes. This phenomenon is

Main Citations and Abstracts 25 mairdy dependent on the dfference in the hydrostatic head in Rough Creek Graben is practically aseismic, probably in large the neer and dvwiKiss, and the flow regime at hot-leg. Be-part due to its orientation in the current stress field. The north-sides this flow interruption ih ivi.4..vii, dynamic flow instabil-west-trending St. Louis Arm of the proposed rift complex in-ities of considerable amplitudes have been observed.

ciudes a pattern of seismicity that extends from southern Illinois along the Mississippi River. This arm must be considered to NUREG/CR-4627: GENERIC COST ESTIMATES. Abstracts From have seismic risk, but because of me lack of development of a Generic Studes For Use in Prepanng Regulatory impact Analy-graben associated we me ann and me mientabon of me ann ses.

Soonce & Engineering Associates, Inc.

S. Cohen & As-In me cunent stmas hem, me M appeam to be W than in me sociates, Inc.

Mathtech, Inc. June 1986.141pp. 8607010432.

ReeNoot porbon of me Mt complex.

36840:003.

The Nuclear Regulatory Cvnoaan.'s Cost Analysis Group NUREG/CR-4634: DEVELOPMENT OF A REAL-TIME RESIDUE (CAG) has sponsored a number of genenc cost estimating stud-NUMBER PROCESSOR FOR SAFT INSPECTION. Phase 11 Final les. These studies were prepared to aid NRC analysts in prepar-Report, September 1984 - April 1986. POLKY,J.N. Sigma Re-l ing Regulatory impact Analyses (RIA's). They were initiated by search, Inc. May 1986.117pp. 8606110715. 36416:240.

the CAG in order to provide cost estimates that would have A high speed SAFT imaging system has been designed using wide application to a large number of Regulatory Analyses residue number system (RNS) computational methods. The im-being performed throughout the NRC. These generic studes aging system is based on a new frequency domem correlation deal primarily with repaar and modfication actuties that may be process applied to conventional pulse-echo ultrasonic data, imposed on nuclear power plants as a result of regulatory ac-wherein the data is collected over a two dimensional aperture.

tions. Abstracts of each of the genenc cost estimating studies The resulting threeh data set is x, y, and ' time-of-have been prepared and assembled in this catalog. These ab-flight' may be processed by the frequency domain SAFT stracts present the results of the more detailed studies in a (FSAFT) system in either real-time or batch (post-test) modes.

compact, easily understood and readly useable format. Individ-In the real-time mode it is expected that true flew re,vyeni ual abstracts have been developed to treat the main-line topics would be of pnmary interest and the resultog images would be of the generic studies. In addition, abstracts have been pre-competitive with current time-domam SAFT (TSAFT) techniques.

pared covering important sub-topics or " stand-alones" which However, the greatest benefit of FSAFT is for detailed analysis are of broad interest in RIA preparation. This abstract catalog of cntical flaw types using the high speed batch or fast inspec-will be expanded and modified as additional generic cost stud-tion mode. The system's performance results from using custom les are completed and as abstracts are modified to reflect up-RNS hardware to speed the correlation process, which for typi-dated conditions.

cal sub-volumes of 64 x 128 x 400 samples points would exe-NUREG/CR-4631: ELECTROMAGNETIC ACOUSTIC TRANS.

cute in 11 seconds. This rapid execution time includes all DUCER (EMAT) DEFECT CHARACTERIZATION OF NUCLEAR memory exchange overhead as weH as forward and inverse REACTOR PIPING WELDS. Phase 1 Final Report, October 1985 -

number theoretic transforms (NTTs) and point spread function March 1986. DAVIS,T.J.; THOME D.K. Sigma Research, Inc.

(PSF) multiplication. The hardware design concentrated on a May 1986. 48pp. 8606110710. 36416:193, custom memory management processor and RNS computation-This work was directed at determining the most promising al modules. Although no hardware was fabricated, it is believed methods for application of EMATs to stainless steel piping ex.

that a commercial system could be realized with conventional amination. It consisted of a literature review, evaluation of shear electronic components operating at a base clock frequency of and longitudinal wave inspection modes, and evaluation of sev.

10MHz.

eral signal processing techniques to enhance signal / noise NUREG/CR-4641: EXPERIMENTAL ASSESSMENT OF BORE-ratios. The work involved both hardware and software develop-HOLE WALL DRILLING DAMAGE IN BASALTIC ROCKS.

ment. The work indicated that defects as small as 0.1 cm deep could be detected in wrought stainless piping, and that defects FUENKAJORN,K.; DAEMEN.J.J. Arizona, Univ. of, Tucson, AZ.

in thick centnfugally cast stainless samples could also be de.

June 1986. 287pp. 8607080221. 3693W1.

tected. In addition, the techniques showed promise for sizing Ring tension tests, permeabimy tests, and um,myn. frac-ture studies have been performed to investigate the borehole the flaws. These results were achieved through a combination of synthetic aperture processing, temporal averaging and low damage induced at low confining pmssure by three Mng h frequency illumination. Additional techniques were evaluated, in.

niques (damond, percussion and mtary). Specimens am drmed with three hole sizes (38, 76 and 102 mm diameter) in Pomona cluding frequency analysis, angle beam scanning and multimode basalt and Grande basaltic andesite. The damaged zone is inspection, but were shown to be of limited benefit for the sam.

pies available. However, these techniques may offer potential characterized in terms of fractures and fracture pattems around the hole, and in terms of tensile strength reduction of the rock for discriminating between cracks and geometric reflectors.

around the holes. Experimental results show that the thickness NUREG/CR-4632:

NEW MADRID SEISMOTECTONIC of the damaged zone around the hole ranges from 0.0 to 1.7 PROGRAM, Final Report. BUSCHBACH T.C. St. Louis Univ., St.

mm. A larger drill bit induces more wan damage than does a Louis, MO. June 1986. 75pp. 8607080250. 36918:257-smaller one. Different drilling techniques show different damage The New Madrid Seismotectonic Program was a large-scale characteristics (itensity and distnbution). Damage characteristics multidisciplinary effort that was designed to define the structural are govemed not only by asilling parameters (bit size, weight on setting and tectonic history of the New Madrid area in order to bit, rotational speed, diamond radius, and energy), but also by realistically evaluate earthquake risks in the siting of nuclear fa-properbes of the rock. The weaker rock tends to show more in-cilities. The tectonic model proposed to explain the New Madrid tense damage than does the stronger one. Cracks within grains seismicity is the " zone of weakness model, which suggests or cleavage fractures are predominent in slightly coarser that an ancient nft complex formed a zone of weakness in the grained rock (larger than 0.5 mm grain size), while intergranular earth's crust along which regional stresses are relieved. The cracks are predominant in very fined grained rock (smaller than Reelfoot Rift portion of the proposed rift complex is currently 0.01 mrn grain size). The damaged zones play no significant seismically active, and it must be considered capable and likely role in the flow path around a borehole plug.

to be exposed to large-magnitude earthquakes in the future.

i Earthquakes that occur in the Wabash Valley area are less NUREG/CR-4642: ROCK MASS SEALING - EXPERIMENTAL AS-abundant and generally have deeper hypocenters than earth-SESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual quakes in the New Madrid area. The area of the Southem Indi-Report June 1984 - May 1985. DAEMEN.J.J.; GREER,W.B.;

ana Arm must be considered to have seismic risk, although a FUENKAJORN,K.; et al. Arzona, Univ. of, Tucson, AZ. June lesser extent than the Reelfoot Rift. The east-west trending 1986. 485pp. 8607080472. 36937:001.

26 Main Citations and Abstracts Field flow tests on cement and bentonite plugs in granite and flow studies conducted on split blocks indicate a complex inter-basalt show results that range from very good, i.e. conductmbes action between plug (swelling, shrinkage) and rock fracture ap-smlar to intact very low permeabihty rock, to moderate. Installa-erture. Fracture grouting experiments confirm nonuruform grout tion procedures can dominate borehole plug sealing perform.

flow (fingering), provide estimates on groutability and grout vis-ance. Tests on cement plugs at 90-95 degrees centigrade show cosity influence, and point out the need for appropriate monitor-detenoration (increase in conductnnty). Push-out tests at 50 de-ing if sealing effectiveness of fracture grouting is to be deter-grees centigrade show little interface strength reduction com-mined correctly. Preliminary results are given of plug-rock inter-pared to room temperature experiments. Axial splitting of basalt face studies, cement and bentonite test installations at Cargo-cylinders during heated flow test suggests possible stress corro-dera Canyon, and sealing performance of bentonite / crushed sion, as does splitting of blocks in size effects studies. Fracture basalt plugs.

Contractor Report Number Index l

l This index lists, in alphabetical order, the NUREG/CR for the report and to the 10-contractor-issued report codes for the NRC digit NRC Document Control System acces-contractor reports in this compilation. Each sion number.

i contractor code is cross-referenced to the SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMSER ANL-85-71 Vil NUREG/CR-4453 V02 MEA-2148 NUREG/CR-3228 V04 ANL 85-75 NUREG/CR-4490 V01 ORNL-6199 NUREG/CR-4309 ANL-86-19 NUREG/CR-4584 ORNL-62b9 NUREG/CR-4489 ANL-8627 NUREG/CR-4621 ORNUNOAC-227 NUREG/^>I-4261 ANL-86-3 NUREG/CR 4503 V01 ORNUNSIC-200 NUREG/CR-2000 V05 N3 ORNUNSIC-200 NUREG/CR-2000 V05 N4 ANL-86-4 NUREG/CR-4505 ORNUNSIC-200 NUREG/CR-2000 V05 N5 ANL-86 5 NUREG/CR-4506

[MI hhh U

ANL 86-7 NUREG/CR-4124 V02 ANL/EES TM-297 NUREG/CR-4568 ORNUTM-8939/V2 NUREG/CR-3572 V02 BMI-2128 NUREG/CFI 4379 V04 ORNUTM-9176 NUREG/CR-3770 BMl-2134 NUREG/CR-4572 ORNUTM-9444 NUREG/CR-4047 BNL-NUREG-51454 NUREG/CR-2331 VOS N3 ORNUTM-9614/V3 NUREG/CR-4236 V03 BNL-NUREG-51751 NUREG/CR-3705 ORNUTM-9640/V1 NUREG/CR-4265 V01 BNL-NUREG-51834 NUREG/CR-4048 ORNUTM-9720 NUREG/CR-4332 BNL-NUREG-51872 NUREG/CR-4207 ORNUTM-9726 NUREG/CR-4338 BNL-NUREG-51904 NUREG/CR-4319 ORNUTM-9798/V3 NUREG/CR-4402 V03 BNL-NUREG-51917 NUREG/CR-4374 V03 ORNUTM-9826 NUREG/CR-4413 BNL-NUREG-51930 NUREG/CR-4404 ORNUTM-9068 NUREG/CR-4449 BNL-NUREG-51934 NUREG/CR-4409 ORNUTM-9933 NUREG/CR-4349 BNL-NUREG-51951 NUREG/CR-4493 ORNUTM-9968/V1 NUREG/CR-3064 V01 BNL-NUREG-51956 NUREG/CR-3957 ORNUTM-9968/V2 NUREG/CR-3064 V02 BNL NUREG-51967 NUREG/CR-4549 PNL-4054 NUREG/CR-3620 S01 BNL-NUREG-51969 NUREG/CR-4557 PNL-422L NUREG/CR-2850 V04 BNL-NUREG-51983 NUREG/CR-4588 V02 PNL-4241 NUREG/CR-2675 VOS BNL-NUREG-51983 NUREG/CR-4588 V01 PNL-4710 NUREG/CR-3262 V01 BNL-NUREG-51983 NUREG/CR-4588 V03 PNL-5155 NUREG/CR-3882 BNL-NUREG-51984 NUREG/CR-4589 PNL-5480 NUREG/CR-4516 PNL-5515 NUREG/CR-3262 V02 BNL-NUREG-51986 NUREG/CR-4594 PNL-5515 NUREG/CR-3262 V07 BNL-NUREG-51987 NUREG/CR-4601 5

BNL-NUREG-51991 NUREG/CR-4607 560 BNL-NUREG-51996 NUREG/CR-4619 PNL-5697 NUREG/CR-4461 EGG-2416 NUREG/CR 4384 PNL-5727 NUREG/CR-4483 EGG-2428 NUREG/CR-4454 PNL-5729 NUREG/CR-4484 EGG-2436 NUREG/CR-4488 PNL-5788 NUREG/CR-4520 EGG-2438 NUREG/CR-4498 PNL-5809 NUREG/CR-4330 V01 EGG-2444 NUREG/CR-3453 PNL-5809 NUREG/CR-4330 V02 FOX /NRC 8601 NUREG/CR-4602 PNL-5822 NUREG/CR-4583 V01 IE B-80 01 NUREG/CR-3960 SAND 83-0242 NUREG/CR-3162 IEB-80-20 NUREG/CR-3962 SAND 84-0749 NUREG/CR-4241 LA-10638-M NUREG/CR-4497 SAND 84-2161 NUREG/CR-4027 LA-10652-MS NUREG/CR-3965 SAND 85-2264 NUREG/CR-3970 LA-10663-MS NUREG/CR-4526 SAND 85-2545 NUREG/CR-4463 LA-10678-M NUREG/CR-4561 SAND 85-2575 NUREG/CR-4467 LA-10711-MS NUREG/CR-4595 SAND 86 0101 NUREG/CR-4507 LA-10730-MS NUREG/CR-4615 V01 SAND 86-0462 NUREG/CR-4543 LBL-20387 NUREG/CR-4582 SAND 86-0494 NUREG/CR-4548 MEA-2090 NUREG/CP-0067 V01 UCID-20505 NUREG/CR-4609 MEA-2090 NUREG/CP-0067 V02 UCRL-53644 V01 NUREG/CR-4290 V01 i

l I

27 I

I l

l l

I i

l l

4

Personal Author index This index lists the personal authors of NRC report (s) prepared by the author. If informa-staff and contractor reports in al)habetical tion is needed, refer to the main citation by order. Each name is followec by the the NUREG number.

NUREG number and the title of the A8 DEL-RAZEK,M.

AWAl,A.

NUREG/CR-3441: RADONEA COMPUTER CODE FOR SIMULATING NUREG-1188: THE AUBURN STEEL COMPANY RADIOACTIVE CON-FAST. TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDI-TAMINATON INCIDENT.

TiONS AND TWO4AYER RADIONUCUDE CONCENTRATIONS IN-CLUDING THE EFFECT OF BED-DEPOSITION IN CONTROLLED BAARS,R.E RIVERS AND TIDAL ESTUARIES.

NUREG/CR-4595: ENHANCEMENT TO THE LAFM COMPUTER CODE.

ASRAHAM,T.

BAILEY,WJ.

NUREG/CR-4619: STRESS CORROslON CRACKING TESTS ON HIGH-NUREG/CR-4330 V02: REVIEW OF UGHT WATER REACTOR REGU-LEVEL-WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-POSITORY ENVIRONMENTS.

TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE ABT,S.R.

TO RISK; Reactor Cor:tainment Leakage Rates. Main Steam isolation NUREG/CR-4620: METHODOLOGIES FOR EVALUATING LONG-TERM Valve Leakage ST UZATON DESIGNS OF URANIUM MILL TAluNGS IMPOUND-BAKER D.A.

NUREG/CR-2850 V04: POPULATON DOSE COMMITMENT DUE TO AGRAWAL A.K.

RADCACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR-4374 V03: A REVIEW OF THE OCONEE-3 PROBABluSTIC IN 1982.

RISK ASSESSMENT CONTAINMENT PERFORMANCE RADIOLOGICAL SOURCE TERMS AND RISK ESil-BALDWIN,AJ.

MATES.

NUREG/CR-4583 V01: DEVELOPMENT AND VAUDI. TON OF A REAL-TIME SAFE-UT SYSTEM FOR THE INSPECTION OF UGHT WATER NU E /CR-4642: ROCK MASS SEAUNG EXPERIMENTAL ASSESS-MENT OF BOREHOLE PLUG PERFORMANCE. Annual ReportJune BALL.D.G.

1984 - May 1985.

NUREG/CR-3770- PRELIMINARY DEVELOPMENT OF AN INTEGRAT-ALPERT,DJ.

ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL NUREG/CR-4467: RELATIVE IMPORTANCE OFylNDIVIDUAL ELE-SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER MENTS TO REACTOR ACCIDENT CONSEQUNCES ASSUMING PLANT.

EOUAL RELFASE FRACTONS.

BALL J.R.

AMICO,P.

NUREG/CR-4568: A HANDBOOK FOR QUICK ESTIMATES:A Method NbREG/CR-4142: A REVIEW OF THE MILLSTONI.3 PROBABluSTIC For Developing Quick Approximate Estimates Of Costs For Generic Ac-SAFETY STUDY.

tions For Nuclear Power Plants.

ANDERSON,C.A.

3ALL,3.J.

NUREG/CR-3965: AN INVESTIGATON OF THE S7ENGTH OF H440 NUREG/CR-4265 V01: AN ASSESSMENT OF THE SAFETY IMPUCA-GRAPHITE WHEN SUBJECTED TO COMBINED ylMARY AND SEC-TONS OF CONTROL AT THE CALVERT CUFFS-1 NUCLEAR PLANT.

ONDARY STRESS.

NUREG/CR-4402 V03: HIGH-TEMPERATURE GAS-COOLED REACTOR ANDERSON,W.E SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NUREG/CR-4483: REACTOR PRESSURE VESSEL FAILURE PROBA-EVALUATION.Ouarterfy Progress Report, July 1 - September 30,1985.

BluTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-l{ED THERMAL SHOCK EVENTS.

NU

/CR-4048: A METHODOLOGY FOR ALLOCATING REUABluTY ANDREWS.G.L AND RISK.

NUHEG/CR-4461: TORNADO CUMATOLOGY OF THE CONTIGUOUS UNITED STATES.

BARRETT,R.

NUREG-1152: MILLSTONE 3 RISK EVALUATON REPT:AN OVERALL APOSTOLAKIS G.

REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBABI-NUREG/CR-4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC USTIC SAFETY STUDY.

SAFETY STUDY.

BATTLE R E ARNOLD W.D.

NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS NUREG/CR-4236 Vf 3: PROGRESS IN EVALUATON OF RADIONU-OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL CUDE GEOCHEh/GTRY INFORMATION DEVELOPED BY DOE HIGH-REPORT' LEVEL NUCLEAR'NASTE REPOSITORY SITE PROJECTS. Report For NU G EN E E F CT ON SOLUBluTY AND N R G/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE REDUCTION RESEARCH PROJECTS.

ARORA.H

NUREGYR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR.

BAUMANN.B.L RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY NUREG/CR-4316: EVALUATION OF NUCLEAR FACIUTY DECOMMIS-POLYETHYLENE.

SiONING PROJECTS STATUS REPORT - HUMBOLDT BAY POWER PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

AVERY,T.S.

NUREG/CR-4642: ROCK MASS SEAUNG - EXPERIMENTAL ASSESS-BEAHM,E.C.

MENT OF BOREHOLE PLUG PERFORMANCE. Annual RepotJune NUREG/CR-4338: TELLURIUM BEHAVOR IN CONTAINMENT UNDER 1984 - May 1985.

UGHT WATER REACTOR ACCIDENT CONDITONS.

29

30 Personal Author Index eEcKER,W.

auCKALEw,W.H.

NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-NUREG/CR-4543: FIRST RESULTS FROM ELECTION-PHOTON RADIATION ON Til: MECHANICAL PROPERTIES OF HIGH DENSITY DAMAGE EQUIVALENCE STUDIES ON A GENERIC ETHYLENE-PRO-POLYETHYLENE.

PYLENE RUBBER.

SEEEE MJt.

SURNS,TJ.

i NUREG-0020 V10 NO3: UCENSED OPERATING REACTORS STATUS NUREG/CR-3770 PREUMINARY DEVELOPMENT OF AN INTEGRAT-

SUMMARY

REPORT. Data As Of Fetm28.1986 (Gray Book 0 ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL NUREG-0020 V10 N04: UCENSED OPEMATING REACTORS STATUS SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER

SUMMARY

REPORT. Data As Of March 31,1986.(Gray Book f)

PLANT.WITWO OVERSIZE DRAWINGS.

BEOGEL.AJ NUREG/Chl-4330 V01: REVIEW OF UGHT WATER REACTOR REGU-SUSCHSACH,TI.

NUREG/CR4632: NEW MADRID SEISMOTECTONIC PROGRAMFmal LATORY REQUIREMENTS. Volume 1:ldenbfication Of Regulatary Re-Report.

qurements That May Have importance To Rrak.

EU8LIK A BERNEUTER,D.

NUREG/CR-4142 A REVIEW OF THE MILLSTONE 3 PROBABlWSTIC NUREG-1152 MILLSTONE 3 RISK EVALUATION REPT AN OVERALL REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBA81-SAFETY STUDY.

USTIC SAFETY STUDY.

BEYER C.E.

NUREG/CR-4330 V02: REVIEW OF UGHT WATER REACTOR REGO-BUStfE,LJ.

LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-NUREG/CR-4583 V01: DEVELOPMENT AND VAUDATION OF A REAL-TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE TIME SAFE-UT SYSTEM FOR THE INSPECTON OF UGHT WATER TO RISK. Reactor Contamment Leakage Rates.Mem Stearn isolation COMPONENTS.Sem> Annual Report For April 1964-September 1964.

Valve Leekage BE2ELLA,W.A.

NUREG/CR-4027:

TRAC-PF1/ MOD 1 INDEPENDENT NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC-ASSESSMENT. Condensation in Stratified Cocurrent Flow.

TIVENESS OF AN OPERATIONAL SAFETY RELIABluTY PROGRAM IN ADDRESSING GENERIC SAFETY PROBLEMS.

CASASetO.L NUREG/CR-4506: AN OPERATIONAL SAFETY REUABluTY PROGRAM NUREG-1188: THE AUBURN STEEL COMPANY RADCACTIVE CON-APPROACH WITH RECOMMENDATIONS FOR FURTHER DEVELOP.

TAMINATON INCIDENT.

MENT AND EVALUATION.

CADWELL.LL BIANAH-NUREG/CR-2675 V05: RELEVANCE OF BOTIC PATHWAYS TO THE NUREG/CR-4483: REACTOR PRESSURE VESSEL FAILURE PROBA-LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esti-BILITY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-mation Of Radelion Dose To Man Resultm0 rom Biotic TransportThe F

IZED THERMAL SHOCK EVENTS.

BIOPORT/ MAX 11 Software Package).

BICKFORD,W.E.

CAIN.C.L NUREG/CR-4330 V01: REVIEW OF UGHT WATER REACTOR REGU-NUREG-1179 V02: RUPTURE OF MODEL 48Y UF6 CYUNDEP AND RE-LATORY REQUIREMENTS. Volume 1:Identrhcahon Of Regulatory Re-LEASE OF URANIUM HEXAFLUORIDE.Cyhnder Overfillrarch 12-qurements That May Have importance To Risk.

13,1986. Investigation Of A Failed UF6 Shipping Cyhnder SLENCOE J.G.

NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADIONU-CAMPAL NUREG/CR-4507: HECTR VERSION 1.5 USER'S MANUAL CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For CARMNM D.W-April-June 1985' NUREG/CR-4609: EFFECTS OF EARTHOUAKES ON UNDERGROUND BOLANDER M.A.

FACluTIES.Uterature Review And Discuseson.

NUREG/CR-4454: RELAP5/ MOD 2 CODE ASSESSMENT AT THE C

IDAHO NATONAL ENGINEERING LABORATORY.

pp DOOTH.R.Sw ENCE IN NUCLEAR POWER PLANTS.

NUREG/CR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS.

CASE,F.L NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND BOWERMAAS.

SORPTION.The Solubikty Of Tc(IV) Oxides.

NUREG/CR-4601: TECHNICAL CONSIDERATONS AFFECTING PREP.

ARATION OF ION-EXCHANGE RESINS FOR DISPOSAL CHAMSERS,R.

NUREG/CR-4454: RELAPS/ MOD 2 CODE ASSESSMENT AT THE N E/ -4207: FAULT TREE APPLICATON TO THE STUDY OF SYSTEMS INTERACTIONS AT INDIAN POINT 3.

CHAN.B.C.

NUREG/CR-4404. ANALYSIS OF ALLOWED OUTAGE TIMES AT NUREG/CR-3705: IMPROVED MODEUNG AND NUMERICS TO SOLVE BYRON GENERATING STATION.

TWO. DIMENSIONAL ELLIPTIC FLUID FLOW AND HEAT TRANSFER PROBLEMS.

SRADLEYf.J.

NUREG-1188: THE AUBURN STEEL COMPANY RADCACTIVE CON-TAMINATON INCIDENT.

NUR / -4467: RELATIVE IMPORTANCE OF INDIVIDUAL ELE.

MENTS TO REACTOR ACCIDENT CONSEQUENCES ASSUMING BRODSKY A.

NUREG-1156: ACCURACY AND DETECTON UMITS FOR BIOASSAY EQUAL RELEASE FRACTIONS.

MEASUREMENTS IN RADIATON PROTECTION - STATISTICAL CHEN J.C.

CONSIDERATIONS.

NUREG/CR-4353: ASSESSMENT OF POST 4RITICAL HEAT FLUX 3ROWN.D.

MODELS WITH LEHIGH NONEQUluBRIUM DATA.

NUREG/CR-4581: DRYOUT FRONT MODEUNG FOR PWR THERMAL NUREG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-HYDRAUUC ANALYSIS.

TION OF QUENCHING OF SUPERHEATED DEBRIS BED 3 UNDER TOP-REFLOOD CONDITIONSFmal Report.

BRUSTf.

NUREG/CR-4572 NRC LEAK.BEFORE-BREAK (LBB.NRC) ANALYSIS CHEVERTON,R.D.

METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED NUREG/CR-3770: PREUMINARY DEVELOPMENT OF AN INTEGRAT.

PIPES UNDER AXIAL PLUS BENDING LOADS. Topical Report.

ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL

Personal Author Index 31 SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER COZZUOL,J.M.

PLANT.W/TWO OVERSIZE DRAWINGS.

NUREG/CR-4454: RELAPS/ MOD 2 CODE ASSESSMENT AT THE IDAHO NATIONAL ENGINEERING LABORATORY.

NUREG/CR-4048: A METHODOLOGY FOR ALLOCATING REUABluTY CRADDICK,W.G.

AND RISK.

NUREG/CR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-NUREG/CR-4404: ANALYSIS OF ALLOWED OUTAGE TIMES AT ING THE SAFETY IMPLICATIONS OF CONTROL SYSTEMS.

BYRON GENERATING STATION.

CRANWELL R.M.

,g, NUREG/CR-3162-DATA INPUT GUIDE FOR SWIFT li.The Sanda NUREG/CR-4503 V01: LONG TEAM EMBR'TTLEMENT OF CAST.

Weste-leolaton Flow And Transgnt Model For Fractured Media Re-DUPLEX STAINLESS STEELS IN LWR SYS. Annual Report,0ctober lease 4.84.

l 1984 September 1985.

CHOU.C.K.

CRAWFORD S.L.

NUREG/CR-4290 V01: PROBABluTY OF PlPE FAILURE IN THE REAC.

NUREG/CR-4583 V01: DEVELOPMENT AND VAUDATON OF A REAL.

TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR TIME SAFE-UT SYSTEM FOR THE INSPECTION OF UGHT WATER PLANTS. Volume 1: Summary Report.

COMPONENTS. Semi-Annual Report For April 1984-September 1984.

CHU,T.L.

CRISTY,M.

NUREG/CR-4404: ANALYSIS OF ALLOWED OUTAGE TIMES AT NUREG/CR-3572 V02: DETERMINATION OF METABOUC DATA AP-BYRON GENERATING STATION.

PROPRIATE FOR HLW DOSIMETRY.tl. Gastrointestinal Absorphort CHUNG D.C.

CULLEN,W.H.

NUREG/CR-4609: EFFECTS OF EARTHOUAKES ON UNDERGROUND NUREG/CP-0087 V01: PROCEEDINGS OF THE SECOND IAEA SPE.

FACIUTIES.Uterature Review And Discussen.

CIAUSTS' MEETING ON SUBCRITICAL CRACK GROWTH. Sessions i And II. Held At Sendai, Japan.May15-17,1985.

CHUNG,H.M.

NUHEG/CP-0067 V02: PROCEEDINGS OF THE SECOND IAEA SPE-NUREG/CR-4453 V02; UGHT-WATER-REACTOR FUEL SAFETY SYS-CIAUSTS MEETING ON SUBCRITICAL CRACK GROWTH. Sessions ill TEMS RESEARCH PROGRAMS. Quarterly Progress Rep @

& IV Held At Sendai, Japan,May 15-17,1985.

1985.

NUREG/CR-4503 V01: LONG TERM EMBRITTLEMENT OF CAST-CZAJKOWSKI C.J.

DUPLEX Si AINLESS STEELS IN LWR SYSAnnual Report. October NUREG/CR-4557. A REVIEW OF ISSUES RELATED TO IMPROVING 1984 - September 1985.

NUCLEAR POWER PLANT DIESEL GENERATOR REUABluTY.

CLA M,N E DAEMEN.J.J NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPLICATONS NUREG/Chl-4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

NUREG/CR-4642: ROCK MASS SEAUNG EXPERIMENTAL ASSESS-NURE

-4449-A PWR HYBRID COMPUTER MODEL FOR ASSESS-MENT OF BOREHOLE PLUG PERFORMANCE. Annual ReportJune ING THE SAFETY IMPLICATIONS OF CONTROL SYSTEMS.

1984 - May 1985.

CLARK F.H.

NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS DAWS,3 OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL NUREG/CR-4454: RELAP5/ MOD 2 CODE ASSESSMENT AT THE REPORT.

IDAHO NATONAL ENGINEERING LABORATORY.

CLARKE,D1 DAVIS,E.C.

NUREG/CR-3472 V02: SURFACE PROPERTIES AND PERFORMANCE NUREG/CR 4498: FIELD TESTING OF WASTE FORMS CONTAINING PREDICTON OF ALTERNATIVE WASTE FORMS. Final Report.

EPICOR-il ON EXCHANGE RESINS USING LYSIMETERS.

CLAYTOR.T.N.

DAVIS,P.

NUREG/CR-4124 V02: NDE OF STAINLESS STEEL AND ON-LINE NUREG/CR-4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC LEAK MONITORING OF LWRS. Annual Report October 1984 - Sep-SAFETY STUDY.

tomber 1985.

DAVIS,T.J.

CLEVELAND,J.C.

NUREG/CR-4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER NUREG/CR-4402 V03: HIGH-TEMPERATURE GAS COOLED REACTOR (EMAT) DEFECT CHARACTERl2ATON OF NUCLEAR REACTOR SAFETY STUDIES FOR THE OfVISION OF ACCIDENT PIPING WELDS. Phase i Fr' ial Report,0ctober 1985 March 1986.

EVALUATON.Ouarterty Progress Report. July 1 September 30,1985.

DE AGAZIO,A.W.

COMEN,L.

NUREG-1177: SAFETY EVALUATON REPORT RELATED TO THE RE-NUREG-0837 V05 N04: NRC TLD DIRECT RADtATON MONITORING START OF DAVIS-BESSE NUCLEAR POWER STATION. UNIT 1.FOL.

NUR 4 83 1

R A

MONITORING NETWORK. Progress Report, January-March 1986.

DEAN,R.S.

g NUREG/CR-3960- CLOSEOUT OF IE BULLETIN 8001. Operability Of NUREG/CR-4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT FOR VERTICAL FISSON PRODUCT RELEASE APPARATUS IN HOT NU E /CR-3962 OSEO E

80-es Wes-CELL B. BUILDING 4501.

tinghouse Type W-2 Spring Return To Neutral Control Switches.

CONKuN,J.C.

DEBELUS,DL NUREG/CR-4402 V03: HIGHTEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE OfVISON OF ACCIDENT NUREG/CR-3262 V01: COBRA-NC:A THERMAL-HYDRAULIC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR EVALUATION Quarterty Progress Report. July 1 - September 30,1985.

COMPONENTS. Volume 1: Equations And Constitutive Models.

COOKE.R.H.

NUREG/CR-3262 V07: COBRA-NC A THERMAL-HYDRAUUC CODE NUREG/CR-4511: ASSESSMENT OF THE ADEQUACY OF THE CAU.

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR BRATIONS PERFORMED BY COMMERCIAL CALIBRATON SERV-CCMPONENT. Volume 7: Assessment Manual for Cor'tainment Applica-ICES FOR IONIZING RADIATON SURVEY INSTRUMENTS.

tions.

COSTANTINO,C.J.

DIFILIPPO,F.C.

NUREG/CR-4588 V03: SOIL-STRUCTURE INTERACTON Vol 3.Influ-NUREG/CR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-ence Of Ground Water.

ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS.

32 Personal Author index nananasaas a r FOK,MJ.

NUREG/CR-4507: HECTR VERSON 1.5 USER'S MANUAL NUREG/CR-4602: UNIQUENESS OF BOluNG WATER REACTOR PRI-MARY WATER CHEMISTRY. Final Report,0ctober 1985 March 1986.

NUREG/CR-4409-DATA BASE ON NUCLEAR POWER PLANT DOSE FRESCO,A.

REDUCTION RESEARCH PROJECTS.

NUREG/CR-4207: FAULT TREE APPUCATION TO THE STUDY OF DOCT M SYSTEMS INTERACTIONS AT INDIAN POINT 3.

NUREG/CR-4583 VOI: DEVELOPMENT AND VAUDATION OF A REAL*

FUENKAJORN,K.

TIME SAFE-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER NUREG/CR-4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE COMPONENTS. Semi Annual Report For Apnl 1984-September 1984.

WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

NUREG/CR-4642: ROCK MASS SEAUNG - EXPERIMENTAL ASSESS-U CR-4315 V01: EVALUATON OF NUCLEAR FACluTY DECOM-

{M MISSONING PROJECTS. Summary Status Report Three Mde Island 198 '

Unit 2 Reactor Coolant System & Systems Decontamination.

NUREG/CR-4315 V02: EVALUATON OF NUCLEAR FACluTY DECOM-FUENTES,H.R.

MISSONING PROJECTS. Summary Status Report Three Mile Island NUREG/CR-4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT Urut 2 Reactor Buildino Decontamanation.

IN THE UNSATURATED ZONE. Information And Data Sets.

~

NUREG/CR-4315 V03: EVALUATION OF NUCLEAR FACILITY DECOM-MISSIONING PROJECTS. Summary Status Report Three Mile Island FUGELSO,LE.

Unit 2 Reactor Defueling & Disassen#y.

NUREG/CR-4526: FINITE ELEMENT ANALYSIS OF THE 2240 MW NUREG/CR-4315 V09 EVALUATION OF NUCLEAR FACIUTY DECOM-HTGR PCRV-MISSIONING PROJECTS. Summary Status Report Three Mile Island Unit 2 Radioactive Waste And Laundry Shipments.

GANO.K.A.

NUREG/CR-2675 V05: RELEVANCE OF BIOTIC PATHWAYS TO THE DOLECEK,E.H.

LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esti-NUREG/CR4511: ASSESSMENT OF THE ADEOUACY OF THE CAU_

mation Of Radiation Dose To Man Result ng From Biotic TransportThe BRATIONS PERFORMED BY COMMERCIAL CAUBRATON SERV-BIOPORT/MAXII Software Package).

ICES FOR IONIZING RADIATION SURVEY INSTRUMENTS.

GARCIA A.

ECONot008,C.

NUREG/CR-4142-A REVIEW OF THE MILLSTONE 3 PROBABluSTIC NUREG/CR4594: ESTIMATED SAFETY SIGNIFICANCE OF GENERIC SAFETY STUDY.

A iS 6L GARNICH,M.R.

ELDER,H.K.

NUREG/CR-4483. REACTOR PRESSURE VESSEL FAILURE PROBA-NUREG/CR-4519: TECHNOLOGY. SAFETY AND COSTS OF DECOM-BluTY FOLLOWING THROUGH. WALL CRACMS DUE TO PRESSUR-MISSIONING NUCLEAR FUEL CYCLE FACILITIES CLASSIFICATON 12ED THERMAL SHOCK EVENTS.

OF DECOMMISSIONING WASTE.

ELLINGWOOD,8.

NUREG/CR-4507: HECTR VERSON 1.5 USER'S MANUAL i

l NUREG/CR-3957: REUABluTY ASSESSMENT AND PROBABluTY BASED DESIGN OF REINFORCED CONCRETE CONTA!NMENTS GEORGE.T.L AND SHEAR WALLS. Summary Report NUREG/CR-3262 V02 COBRA-NC-A THERMAL-HYDRAUUC CODE

[

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR ERASLAN A H.

COMPONENTS. Volume 2: COBRA-NC Numerical Solution Methods.

NUREG/CH-3441: RADONE A COMPUTER CODE FOR SIMULATING FAST-TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDl-GHERSON,P.

TONS AND TWO LAYER RADIONUCUDE CONCENTRATONS IN-NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-CLUDING THE EFFECT OF BED-DEPOSITON IN CONTROLLED SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL j

RIVERS AND TCAL ESTUARIES.

SHOCK STUDY.

FAIROBENT J.E GINS 8 ERG,T.

NUREG/CR-4603. APPRAISING ATMOSPHERIC TRANSPORT AND NUREG/CR4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-OlFFUSON MODELS FOR EMERGENCY RESPONSE FACluTIES.

TION OF OUENCHING OF SUPERHEATED DEBRIS BEDS UNDER y

TOP-REFLOOD CONDITIONS. Final Report NUREG/CR4498: FIELD TESTING OF WASTE FORMS CONTAINING GLANTZ,C.S.

EPICOR-il CN EXCHANGE RESINS USING LYSIMETERS.

NUREG/CR-3882 A METHOD TO CHARACTERIZE LOCAL METEOR-OLOGY AT NUCLEAR FACIUTIES FOR APPUCATON TO EMER-RE /CR 207: FAULT TREE APPUCATON TO THE STUDY OF SYSTEMS INTERACTONS AT INDIAN POINT 3.

GOOD,M.S.

I FLANAGAN,0.F'770 PREUMINARY DEVELOPMENT OF AN INTEGRAT-NUREG/CR 4484: STATUS OF ACTMTIES FOR INSPECTING WELD NUREG/CR-3 OVERLAID PIPE JOINTS.

l ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL SHOCK AS APPLIED TO THE OCONEE UNIT 1 NUCLEAR POWER g;

PLANT.W/TWO OVERSIZE DRAWINGS.

NUREG-1183: NONRADIOLOGICAL GROUNDWATER QUAUTY AT LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITES.

FLETCHER,C.D.

NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK GREER,W.8.

LOSS-OF COOLANT ACC DENTS IN A RESAR-3S PLANT.

NUREG/CR4642: ROCK MASS SEAUNG EXPERIMENTAL ASSESS-l' MENT OF BOREHOLE PLUG FERFORMANCE. Annual Report. June FLY,0.W.

1984 - May 1985.

NUREG/CR-3965: AN INVESTIG\\ TON OF THE STRENGTH OF H440 GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND SEC.

GREGORY,W.S.

ONDARY STRESS.

I NUREG/CR-4561: FIRAC USER'S MANUAL-A COMPiTER CODE TO SIMULATE FIRE ACCIDENTS IN NUCLEAR FACluw3.

FOLEY,WJ.

NUREG/CR-3960- CLOSEOUT OF IE BULLETIN 8001.Operatuhty Of GRl88LE,R.P.

Autorratic Depressurization System (ADS) Valve Pneumatic Supply.

NUREG/CR4583 V01: DEVELOPMENT AND VAUDATION OF A REAL-NUREG/CR.3962 CLOSEOUT OF IE BULLETIN 80-20.Fadures Of Wes-TIME SAFE-UT SYSTEM FC 1 THE INSPECTON OF UGHT WATER Dnghouse Type W.2 Spnng Return To Neutral Control Switches.

COMPONENTS. Semi-Annual Reprt For April 1984-September 1984.

Personal Author index 33 GMFFITH P.

HOLMAN,0.S.

NUREG/CR-4581: DRYOUT FRONT MODEUNG FOR PWR THERMAL NUREG/CR-4290 V01: PROBABluTY OF PIPE FAILURE IN THE REAC-HYDRAUUC ANALYSIS.

TOR COOLANT LOOFS OF BABCOCK AND WILCOX PWR GUDASAP.

PLANTS. Volume 1: Summary Report NUREG/CR4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 HOSPODOR,S.

STAINLESS STEEL PlPE: J-R CURVE CHARACTERIZATION AND NUREG-0980 R02-NUCLEAR REGULATORY LEGISLATON.

UMIT LOAD ANALYSIS.

GUIDOTTI.T.E.

HUANG,H.

NUREG/CR-3262 V01: COBRA-NC A THERMAL-HYDRAUUC CODE NUREG/CR-3957: REUABluTY ASSESSMENT AND PROBABluTY FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS COMPONENTS. Volume 1:Equatons And Constitutrye Models.

AND SHEAR WALLS. Summary Report NUREG/CR 3262 V07: COBRA.NC.A THERMAL-HYDRAUUC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR ISHIt,M.

COMPONENT. Volume 7 Assessment Manual for Contanment Apphca-NUREG/CR-4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL-i tions.

ING LAWS FOR TWO-PHASE FLOW LOOP.

I NURE3/CR-4621: FLOW VISUAUZATION EXPERIMENT ON HOT-LEG

/CR-4332 DESIGN AND FINAL SAFETY ANALYSIS REPORT FOR VERTICAL FISGON PRODUCT RELEASE APPARATUS IN HOT BYER,K.

CELL B. BUILDING 4501.

NUREG/CR-3701: REMIX A COMPUTER PROGRAM FOR TEMPERA-HACKETT,E.M.

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER NUREG/CR4579-APPLICATION OF THE KEY CURVE AND MULTI.

INTERRUPTON OF NATURAL CIRCULATION.

SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-ALLOY STEEL SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL SHOCK STUDY.

HAFFNER D.R.

NUREG/CR4316: EVALUATON OF NUCLEAR FACIUTY DECOMMIS-JACOSS,0.K.

SONING PROJECTS STATUS REPORT - HUMBOLDT BAY POWER NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADIONU-PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

CUDE GEOCHEMISTRY INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PRCLiECTS. Report For N E /CR4583 V01: DEVELOPMENT AND VAUDATON OF A REAL-TIME SAFE-UT SYSTEM FOR THE INSPECTON OF UGHT WATER JAIN,H.

COMPONENTS. Semi-Annual Report For April 1984-September 1984.

NUREG/CR4619: STRESS CORROSION CRACKING TESTS ON HIGH-HANAN,N.

LEVEL-WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-NUREG/CR-4207: FAULT TREE APPUCATON TO THE STUDY OF POSITORY ENVIRONMENTS.

SYSTEMS INTERACTONS AT INDIAN POINT 3.

HARMS,NL NUREG-0837 V05 N04: NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4516: INTERNATIONAL SAFEGUARDS AT FACIUTIES EM-NETWORK. Progress Report, October-December 1985.

PLOYING SPENT FUEL ROD CONSOUDATION.

NUREG-0837 V06 Not NRC TLD DIRECT RADIATON MONITORING HARRlNGTON,R.M.

NUREG/CR-4402 V03: HIGH-TEMPERATURE GAS-COOLED REACTOR JOHNS,N.D.

SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ll.The Sandia

^

NU E

-4413 O O ERO TB RR U IT Waste-Isolation Flow And Transport Model For Fractured Media Re-lease 484.

ONE - ACCIDENT SEQUENCE ANALYSIS.

HAYS R.A.

JOHNSON,M.P.

NUREG/CR4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTION EXPERI-STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATON AND ENCE IN NUCLEAR POWER PLANTS.

LIMIT LOAD ANALYSIS.

NUREG/CR-4539: INVESTIGATION OF TEARING INSTABluTY PHE-JOYCE,J A.

NGMENA IN ASTM A106 STEEL NUREG/CR-4579: APPUCATON OF THE KEY CURVE AND MULTI-SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF HEINEMAN.J.S.

ALLOY STEEL NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC.

TIVENESS OF AN OPERATIONAL SAFETY RELIABluTY PROGRAM KASyK,0.

IN ADDRESSING GENERIC SAFETY PROBLEMS-NUREG-1188: THE AUBURN STEEL COMPANY RADIOACTIVE CON-TAMINATION INCOENT.

HENCH,LL NUREG/CR4472 V02: SURFACE PROPERTIES AND PERFORMANCE PREDICTON OF ALTERNATIVE WASTE FORMS. Final Report NU G/CR-4319. NUCRAC - A CODE FOR THE ESTIMATION OF AD.

HENNICK,A.

VERSARY-ACTION CONSEQUENCES IN THE NUCLEAR POWER NUREG/CR-3960: CLOSEOUT OF IE BULLETIN 80-01.Operatxtity Of FUEL CYCLE.

Automate Depressurtraton System (ADS) Valve Pneumatic Supply.

NUREG/CR-3962 CLOSEOUT OF IE BULLETIN 80-20. Failures Of Wes-KELLY,G.

tmghouse Type W-2 Spnng Return To Neutral Control Switches.

NUREG-1152 MILLSTONE 3 RISK EVALUATON REPTAN OVERALL REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBABI-

-4557: A REVIEW OF ISSUES RELATED TO IMPROVING NUCLEAR POWER PLANT DIESEL GENERATOR REUABlWTY.

KELLY,R.

HINKLE,N.E.

NUREG-1188: THE AUBURN STEEL COMPANY RADOACTIVE CON.

NUREG/CR-4620: METHODOLOGIES FOR EVALUATING LONG-TERM TAMINATION INCIDENT.

STAB UZATON DESIGNS OF URANIUM MILL TAILINGS IMPOUND.

KEL.MERS,A.D.

NUREG/CR4236 V03: PROGRESS IN EVALUATON OF RADIONU-HODGE,S.A.

CUDE GEOCHEMISTRY INFORMATON DEVELOPED BY DOE HIGH-NUREG/CR4413: LOSS OF CONTROL AIR AT BROWNS FERRY UNIT LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For ONE. ACCIDENT SEQUENCE MLYSIS.

Apnt, June 19e5.

i

34 Personal Author index KENNEDY,W.E.

LAMONICA,L8.

NUREG/CR-2675 VOS: RELEVANCE OF BIOTIC PATHWAYS TO THE NUREG/CR-3770- PREUMINARY DEVELOPMENT OF AN INTEGRAT.

LONG. TERM REGULATON OF NUCLEAR WASTE DISPOSAL (EstL ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL motion Of Radiation Dose To Man Resulting From Biote TransportThe SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER BIOPORT/ MAX 11 Software Package).

PLANT.

NUREG/CR-3820 S01: INTRUDLR DOSE PATHWAY ANALYSIS FOR THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

LEGGETTAW.

MAX 11 Computer Program.

NUREG/CR-3572 V02: DETERMINATON OF METABOUC DATA AP.

KHAM T.A.

PROPRIATE FOR HLW DOSIMETHY.ll. Gastrointestinal Absorption.

NUREG/CR4409 DATA BASE ON NUCLEAR POWER PLANT DOSE LEHNER,J.R REDUCTION RESEARCH PROJECTS.

NUREG/Ck-4594: ESTIMATL D SAFETY SIGNIFICANCE OF GENERIC KHAT18-RAHSAR SAFETY ISSUE 61.

NUREG/CR-4374 V03: A REVIEW OF THE OCONEE-3 PROBABluSTIC RISK ASSESSMENT CONTAINMENT UE8ERMANNAC.

PEFiFORMANCE. RADIOLOGICAL SOURCE TERMS AND RISK ESTl-NUREG/CR-4580: STONY BROOK SEISMIC NETWORK ON LONG MATES.

ISLAND,NEW YORK. Final Report (September 1979 March 1985).

KIM,S.8.

UND8EY,C.G.

NUREG/CR4621: FLOW VISUALIZATON EXPERIMENT ON HOT-LEG NUREG/CR-3882: A METHOD TO CHARACTERIZE LOCAL METEOR-U-BEND TWO PHASE NATURAL CIRCULATON PHENOMENA.

OLOGY AT NUCLEAR FACluTIES FOR APPUCATION TO EMER-GENCY RESPCNSE NEEDS.

KIMSRELL,A.F.

NUREG/CR4642: ROCK MASS SEALING - EXPERIMENTAL ASSESS-UNKAE.

MENT OF DOREHOLE PLUG PERFORMANOE. Annual Report, June NUREG/CR-4539: INVESTIGATION OF TEARING INSTABlUTY PHE-1984 - May 1985.

NOMENA IN ASTM A106 STEEL KIMM6NSAD.

UTTLE,W.W.

NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTION EXPERl-NUREG/CR-4330 V01: REVIEW OF LIGHT WATER REACTOR REGU-ENCE tH NUCLEAR POWER PLANTS.

LATORY REQUIREMENTS. Volume 1: Identification Of Regulatory Re-KING.D.8.

quirements That May Have Irnportance To Risk.

NUREG/CR-4507: HECTR VERSION 1.5 USER'S MANUAL g

y KLAGES J.

NUREG-1179 V02. RUPTURE OF MODEL 48Y UF6 CYLINDER AND RE-NUREG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA.

LEASE OF URANIUM HEXAFLUORIDE. Cylinder Overfill. March 12-TION OF QUENCHING OF SUPERHEATED DEBRIS BEDS UNDER 13,1986. Investigation Of A Failed UF6 SNpping Cylinder TOP-REFLOOD CONDITONSTinal Report LORENZAA.

KLECKERA NUREG/CR-4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT NUREG/CR-4572: NRC LEAK-BEFORE-BREAK (LBB.P:RC) ANALYSIS FOR VERTICAL FISSION PRODUCT RELEASE APPARATUS IN HOT METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED CELL B. BUILDING 4501.

PIPES UNDER AXIAL PLUS BENDING LOADS. Topical Report LUND8 ERG,L8.

KLEIN.J.

NUREG/CR-3965: AN INVESTIGATON OF THE STRENGTH OF H440 NUREG/CR4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND SEC-TON OF QUENCHING OF SUPERhEATED DEBRIS BEDS UNDER ONDARY STRESS' TOP.REFLOOO CONDITONS. Final Report MACDONALDA NU E CR-3970- TRAC.PF1/ MOD 1 INDEPENDENT ASSESSMENT:

NUREG/CR-4207: FAULT TREE APPUCATION TO THE STUDY OF LOBt INTERMEDIATE BREAK TEST B-R1M.

SYSTEMS INTERAOTONS AT INDIAN POINT 3.

KOCAMUSTAFAOGUL MARINO,0.P.

NUREG/CR4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL.

NUREG-0900 RO1: NUCLEAR POWER PLANT SEVERE ACCIDENT RE-ING LAWS FOR TWO-PHASE FLOW LOOP.

SEARCH PLAN.

KON2EK,G.J.

MARTIN,W.J.

NUREG/CR-4330 V02: REVIEW OF LIGHT WATER REACTOR REGU-NUREG/CR-4520: PREDICTNE GEOCHEMICAL MODEUNG OF CON-LA10RY REOUIREMENTS ASSESSMENT OF SELECTED REGULA-TAINMENT CONCENTRATONS IN LABORATORY COLUMNS AND IN TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE PLUMES MIGRATING FROM URANIUM MILL TAluNGS WASTE IM-TO RISK. Reactor Containment Leakage Rates Main Steam teolation POUNOMENTS. Final Report Valve Leakage..

MATHIESON,T.

KOUSARIA NUREG/CR4124 V02: NDE OF STAINLESS STEEL AND ON-UNE NUREG/CR-4642 ROCK MASS SEALING - EXPERIMENTAL ASSESS-LEAK MONITORING OF LWRS. Annual Report October 1984 - Sep-MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June tomber 1985.

1984 May 1985.

MAZOUR,T.

NURE 4207: FA L REE AP ICA ON TO THE STUDY OF N

G/

4 84: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK LOSSOF COOLANT ACCIDENTS IN A RESAR-3S PLANT.

NUREG/CR-4488: VENTING OF NONCONDENSIBLE GAS FROM THE R HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-MC8 RIDE,A.F NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATONS OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL KUPPERMAN,D.S.

REPORT.

NUREG/CR-4124 V02: NDE OF STAINLESS STEEL AND ON4JNE LEAK MONITORING OF LWRS. Annual Report October 1984 Sep-MCCANN,M.

tomber 1985.

NUREG/CR4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC SAFETY STUDY.

LAGUARDIA.T.S.

NUREG/CR-3587: OENTIFICATON AND EVALUATON OF FACILITA-MCCONNELL.J.W.

TION TECHNfQUES FOR DECOMMISSIONING UGHT WATER NUREG/CR-4498: FIELD TESTING OF WASTE FORMS CONTAINING POWER REACTORS.

EPICOR-li ON EXCHANGE RESINS USING LYSiMETERS.

Personal Au' hor index 35 MCKENZIE,D.H.

NUREG/CR-4330 V02: REVIEW OF UGHT WATER REACTOR REGU-NUREG/CR-2675 V05: RELEVANCE OF BOTIC PATHWAYS TO THE LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-LONG-TERM REGULATON OF NUCLEAR WASTE DISPOSAL (Est6-TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE meton Of Radlebon Do&a To Man Resultmg From Betic Transport:The TO RISK. Reactor Containment Leakage Rates. Main Steam Isolaton BIOPORT/MAXII Software Packare).

Valve Leakage.

MCKONE.T.

MURPHY,0.A.

NUREG/CR-4142 A REVIEW OF THE MILLSTONE 3 PROBABluSTIC NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTON EXPERI-SAFETY STUDY

ENCE IN NUCLEAR POWER PLANTS.

MEYERAE.

NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADONU-NAPIER,5.A.

I CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-NUREG/CR-2675 V05: RELEVANCE OF BIOTIC PATHWAYS TO THE LEVEL NUCLEAR WASTE REPOSITORY SITE PRCUECTS. Report For LONG-TERM REGULA lON OF NUCLEAR MASTE DISPOSAL (Esti-Aprklune 1985.

maton Of Radiation Dose To Man Resulting From Betic Transport.The l

NUREG/CR-4309-VALENCE EFFECTS ON SOLUBluTY AND BIOPORT/MAXII Software Package).

SORPTON.The Solutphty Of Tc(IV) Oxides.

NUREG/CR-3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR THE ONSITE DISPOSAL OF RADCACTIVE WASTES.The ONSITE/

MILLER,C.A-MAXII Computer Program.

NUREG/CR-4588 V02: SOIL-STRUCTURE INTERACTON.Vol 2:Influ-ence Of Uftoff.

NELSON,J.D.

NUREG/CR-4620: METHODOLOGIES FOR EVALUATING LONG-TERM STAB UZATION DESIGNS OF URANIUM MILL TAluNGS IMPOUND-N CR-3064 V01: COMPUTATIONAL METHODOLOGY FOR OAK RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-NL R G CR COMPUTATON E

FdR OAK "O"I'EE RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-NUREG/CR-4561: FIRAC USER'S MANUAL A COMPUTER CODE TO TOR (BSR):The VICTORR input Processang Code For The Bold Ven.

SIMULATE FIRE ACCIDENTS IN NUCLEAR FACluTIES.

ture System, Volume IL NIELSEN,J R.

MILLER N.E.

NUREG/CR-3453: ELECTRONIC ISOLATERS USED IN SAFETY SYS-NUREG/CR 4379 V04: LONG TERM PERFORMANCE OF MATERIALS TEMS OF U.S. NUCLEAR POWER PLANTS.

USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, Year Four Apr# 1965 - March 1986.

NITSCHE,H.

NUREG/CR-4582 TEMPERATURE EFFECTS ON THE SOLUBluTY MILLERAL-AND SPECIATON OF SELECTED ACTINIDES.

NUREG/CR-4315 V01: EVALUATON OF NUCLEAR FACILTY DECOM-MISSONING PROJECTS. Summary Status Report Three Mde Island NOMURA,K.K.

Unit 2 Reactor Coolant System & System. s Decontaminehon.

NUREG/CR-4483: REACTOR PRESSURE VESSEL FAILURE PROBA-NUREG/CR-4315 V02 EVALUATION OF NUCLEAR FACluTY DECOM-BlWTY FOLLOWING TWOUGH. WALL CRACKS DUE TO PRESSUR.

MISSONING PROJECTS. Summary Status Report Three Mde Island IZED THERMAL SHOC OVENTS.

NR /CR 5

OF NUCLEAR FACluTY DECOM-NOURBAKHSH,H.

urnrn tus se e ist nd

& Di NUREG/CR-3701: REMIX A COMPUTER PROGRAM FOR TEMPERA-NUHEG/CR-4315 ALUATON ' NUCLEAR FACluTY DECOM.

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER MISSONING PRCMECTS. Summary Status Report Three Mde Island INTERRUPTION OF NATURAL CIRCULATON.

Urvt 2 Radoeceve Weste And ts.

NUREG/CR-431ft: EVALUATON OF FACIUTY DECOMMIS.

NOURSAKHSH,H.P.

SONING PROJECTS STATUS REPORT HUMBOLDT BAY POWER NUREG/C9-3702-BUOYANCY EFFECTS IN OVERCOOUNG TRAN-PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL SHOCK STUDY.

MITRA S.

NUREG/CR-4207: FAULT TREE APPUCATION TO THE STUDY OF NUEDERS.M.

SYSTEMS INTERACTIONS AT INDIAN POINT 3.

NUREG/CR-3620 S01: It!TRUDER DOSE PATHWAY ANALYSIS FOR THE ONSITE DISPOSAL OF RADCACTIVE WASTES.The ONSITE/

G/CR-4511: ASSESSMENT OF THE ADEQUACY OF THE CAU.

MA Computer Rogras BRATIONS PERFORMED BY COMMERCIAL CAUBRATON SERV-NYHAN,J.W.

ICES FOR IONIZING RADIATION SURVEY INSTRUMENTS.

NUREG/CR 4615 V01: LODEUNG STUDY OF SOLUTE TRANSPORT MOONEYg IN THE UNSATURATED ZONE. Information And Data Sets.

NUREG/CR-3150- SEISMICITY AND TECTONIC RELATONSHIPS FOR UPPER GREAT LAKES PRECAMBRIAN SHIELD PROVINCE FINAL MSONA REPORT -July 1981 -December 1982.

NUREG/CR-3770 PREUMINARY DEVELOPMENT OF AN INTEGRAT-ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL MORRIS,0.0-SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER NUREG/CR-4449 A PWR HYBRID COMPUTER MODEL FOR ASSESS-PLANT.

ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS.

OSSORNE,M.F.

MUELLER,CA NUREG/CR-4332 DESIGN AND FINAL SAFETY ANALYSIS REPORT NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC-FOR VERTICAL FISSION PRODUCT RELEAS. APPARATUS IN HOT TIVENESS OF AN OPERATIONAL SAFETY REUABluTY PROGRAM CELL B. BUILDING 4501.

IN ADDRESSING GENERO SAFETY PROBLEMS.

NUREG/CR-4506: AN OPERATIONAL SAFETY REUABIUTY PROGRAM PAPAZOGLOUJ.A.

APPROACH WITH RECOMMENDATONS FOR FURTHER DEVELOP-MENT AND EVALUATON' NUREG/CR-4048: A METHODOLOGY FOR ALLOCATING REUABILITY AND RISK.

MUMLHEIM,M.g NUREG/CR-4207: FAULT TREE APPLICATON TO THE STUDY OF NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTON EXPERI.

SYSTEMS INTERACTONS AT INDIAN POINT 3.

ENCE IN NUCLEAR POWER PLANTS.

PARK.C.K.

MULLEN.M.F.

NUREG/CR-4374 V03: A REVIEW OF THE OCONEE-3 PROBABluSTIC NUREG/CR-4330 VO1: REVIEW OF UGHT WATER REACTOR REGU-RISK ASSESSMENT CONTAINMENT LATORY REQUIREMENTS. Volume 1:ldentificatson Of Regulatory Re-PERFORMANCE,RADOLOGICAL SOURCE TERMS AND RISK ESTI-quw.unts That May Have importam iv Risk.

MATES.

l

36 Personal Author index PEDERSEN,LT.

NUREG-0637 V06 N01: NRC TLD DIRECT RADIATION MONITORING NUREG/CR4483: REACTOR PRESSURE VESSEL FAILURE PROBA-NETWORK. Progress Report, January-March 1986.

BluTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-12ED THERMAL SHOCK EVENTS.

RAOLAN3,W.A.

NUREG/CR-4505: A SCOPIN3 STUDY OF THE POTENTIAL EFFEC-PELOOUW8,RA TIVENESS OF AN OPERATIONAL SAFETY REUABluTY PROGRAM I

NUREG/CR-2675 V05: RELEVANCE OF BOTIC PATHWAYS TO THE IN ADDRESSING GENERIC SAFETY PROBLEMS.

LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esti-maton Of Radiaton Does To Man Resuming From Biotic Transport:The RAMEY-SMITH,A.

NUREG-1192: AN INVESTIGATION OF THE CONTRIBUTORS TO NUR C 28 V04 T

DOSE COMMITMENT DUE TO WRONG UNIT OR WRONG TRAIN EVENTS.

RADCACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES N

G CR-3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR RAMODELL,J.V.

THE ONSITE DSPOSAL OF RADIOACTIVE WASTES.The ONSITE/

NUREG/CR4461: TORNADO CUMATOLOGY OF THE CONTIGUOUS MAXit Computer Program.

UNITED STATES.

PELTO.PJ.

REEO,J.

NUREG/CR-4330 V01: REVIEW OF UGHT WATER REACTOR REGU-NUREG/CR4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC LATORY REQUIREMENTS. Volume 1;ldentslication Of Regulatory Re-SAFETY STUDY.

ga'rements That May Have Importance To Rek.

NUMEG/CR-4330 V02: REVIEW OF UGHT WATER REACTOR REGU-REEVES,M.

LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-NUREG/CR-3162 DATA INPUT GUIDE FOR SWIFT II.The Sanda TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE Weste-Isolation Flow And Transport Model For Fractured Moda Re-TO RISK. Reactor Contanment Leakage Ratos. Main Steam Isolaton lease 4.84.

Valve Leakage.-

REICH,M.

^

NUR / 4 594: ESTIMATED SAFETY SIGNIFICANCE OF GENERC B

D N

R OR ET CO M

SAFETY ISSUE 61.

AND SHEAR WALLS. Summary Report.

PERSasMO.D.

NUREG-1192: AN INVESTIGATION OF THE CONTRIBUTORS TO REINER,J.P.

WRONG UNIT OR WRONG TRAIN EVENTS.

NUREG/CR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-ING THE SAFETY IMPLICATONS OF CONTROL SYSTEMS.

PETERSON,ER.

NUREG/CR-4520t PREDICTIVE GEOCHEMICAL MODEUNG OF CON-REST J.

TAINMENT CONCENTRATONS IN LABORATORY COLUMNS AND IN NUREG/CR4453 V02: UGHT-WATER-REACTOR FUEL SAFETY SYS-PLUMES MIGRATING FROM URANIUM MILL TAluNGS WASTE IM-TEMS RESEARCH PROGRAMS. Quarterty Progress Report,Apnl. June POUNDMENTS. Final Report.

1965.

PHILIPPACOPOULO RISLEY,J.F.

NUREG/CR-4588 VOI: SOIL-STRUCTURE INTERACTION.Vol 1;lnflu-NUREG/CR-3587: IDENTIFICATION AND EVALUATION OF FACIUTA-ence Of Layenng.

TION TFCHNIQUES FOR DECOMMISSIONING UGHT WATER pgCARO,R.R.

POWER REACTORS.

NUREG/CR-4497: NRCPAGE APPUCATIONS MANUAL l

PICIULO,P.L NUREG/CR-4467: RELATIVE IMPORTANCE OF INDMDUAL ELE-NUREG/CR-4601: TECHNICAL CONSIDERATIONS AFFECTING PREP-MENTS TO REACTOR ACCIDENT CONSEQUENCES ASSUMING ARATiON OF ION-EXCHANGE RESINS FOR DISPOSAL EQUAL RELEASE FRACTONS.

POLKY,J.M.

RIT2MA88,R.L NUREG/CR-4634; DEVELOPMENT OF A REAL-TIME RESIDUE NUREG/CR-4319 NUCRAC - A CODE FOR THE ESTIMATION OF AD.

NUMBER PROCESSOR FOR SAFT INSPECTION. Phase il Final VERSARY ACTON CONSEQUENCES IN THE NUCLEAR POWER Report Sepiomber 1964. April 1986.

FUEL CYCLE.

POL 2ER,W.L NUREG/CR4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT ROGERTS,F#.

IN THE UNSATURATED ZONE. Information And Data Sets.

NUREG/CR-4516: INTERNATIONAL SAFEGUARDS AT FACIUTIES EM-PLOYING SPENT FUEL ROD CONSOUDATON.

POWERS,T.S.

NUREG/CR-4330 V01: REVIEW OF UGHT WATER REACTOR REGU.

ROGERTS,JA LATORY REQUIREMENTS. Volume 1:ldentrfication Of Regulatory Re.

NUREG/CR-4319 NUCRAC - A CODE FOR THE ESTIMATION OF AD-

)

quirements That May Have importance To Reek.

VERSARY-ACTON CONSEQUENCES IN THE NUCLEAR POWER FUEL CYCLE.

l PRATT.W.T.

NUREG/CR4374 V03: A REVIEW OF THE OCONEE-3 PROBABluSTIC ROGERS,R.D.

RISK ASSESSMENT CONTAINMENT NUREG/CR4498: FIELD TESTING OF W A3TE FORMS CONTAINING i

PERFORMANCE,RADOLOGICAL SOURCE TERMS AND RISK ESTl-EPICOR-il CN EXCHANGE RESINS USING LYSIMETERS.

MATES.

ROHATGI,U.S.

E""EMI'R-4124 V02:

NUREG/CR-4549: DETERMINATON OF APPENDtX K CONSERV.

NUREG/C NDE OF STAINLESS STEEL AND ON-UNE ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR LEAK MONITORING OF LWRS. Annual Report October 1964 Sep-USE TRAC-PD2/ MOO 1 CODE.

tomber 1965.

pnopananasR,LA.

ROKO.R.O.

NUREG/CR-2675 V05: RELEVANCE OF BOTIC PATHWAYS TO THE NUREG/CR-4642 ROCK HASS SEALING. EXPERIMENTAL ASSESS-l LONG TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esti.

MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June maton Of Radiation Duse To Man Resulting from Botic Transport;The 1964 May 1985.

BIOPORT/ MAX 11 Software Package).

MASATIN,K.

NUREG/CR-3965: AN INVESTIGATION OF THE STRENGTH OF H440 NUREG 0637 V05 N04: NRC TLD DIRECT RADIATON MONITORING GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND SEC-NETWORK. Progress Report October-December 1985.

ONDARY STRESS.

Personal Author index 37 ROSCOE,BJ.

SERNE,RJ.

NUREG/CR-4463: HUMAN FACTORS IN ANNUNCIATOR / ALARM NUREG/CR-4520: PREDICTIVE GEOCHEMICAL MODEUNG OF CON-SYSTEMS. ANNUNCIATOR EXPERIMENT PLAN L TAINMENT CONCENTRATIONS IN LABORATORY COLUMNS AND IN i

ROSS.P A.

PLUMES MIGRATING FROM URANIUM MILL TAIUNGS WASTE IM-NUREG-0020 V10 NO3: UCENSED OPERATING REACTORS STATUS POUNDMENTS. Fm' al Report.

EUMMARY REPORT.Deta As Of February 28.1986(Gray Book 11 SEWART,G.H.

NUREG-0020 V10 N04: UCENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of March 31,1986.(Gray Book 1)

NUREG/CR-4516: INTERNATIONAL SAFEGUARDS AT FACIUTIES EM-PLOYlNG SPENT FUEL ROO CONSOUDATION.

Ramaam a a NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC.

SHACK,WJ.

TIVENESS OF AN OPERATIONAL SAFETY REUABILITY PROGRAM NUREG/CR4490 VOI: UGHT. WATER-REACTOR SAFETY MATERIALS IN ADDRESSING GENERIC SAFETY PROBLEMS.

ENGINEERING RESEARCH PROGRAMS:Ouarterty Progrees Report, January-March 1965.

CR4319: NUCRAC - A CODE FOR THE ESTIMATION OF AD.

8" O

NUREG-0' 75 S33: SAFETY EVALUATION REPORT RELATED TO 6

YERSARY. ACTION CONSEOUENCES IN THE NUCLEAR POWER FUEL CYCLE.

OPERATION OF DIABLO CANYON NUCLEAR POWER, UNITS 1 AND

2. Docket Nos.50-275 and 50-323.(Pacific Gas And Electric Company)

SAGENDORF,J.F.

NUREG/CR4633: APPRAISING ATMOSPHERIC TRANSPORT AND SHWOZUKA,M.

DIFFUSION MOCELS FOR EMERGENCY RESPONSE FACluTIES.

NUREG/CR-3957: REUABluTY ASSESSMENT AND PROBABluTY BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS SAHA.P.

AND SHEAR WALLS. Summary Report.

NUREG/CR4545 DETERMIN4 TION OF APPENDIX K CONSERV.

ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR SeMMONS,M.A.

USING TRAC.PD2/ MODI CODE.

NUREG/CR-2675 V05: RELEVANCE OF BIOTIC PATHWAYS TO THE LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Est-SALLACH R.A.

mation Of Radiation Done To Man Resulting From Bione TransportThe NUREG/CR-4241: CHEMICAL ASPECTS OF CESIUM IODIDE INTERAC-BIOPORT/ MAX 11 Softwere Pecksge).

TION IN STEAM WITH 304 STAINLESS STEEL AND INCONEL 600.

SIMONEN,E.P.

SANeORN,0.F.

NUREG/CR-4483: REACTOR PRESSURE VESSEL FAILURE PROBA-NUREG-1179 V02: RUPTURE OF MODEL 48Y UF6 CYUNDER AND RE.

BlUTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-LEASE OF URANIUM HEXAFLUORIDE.CykrWar Overfilt, March 12-lZED THERMAL SHOCK EVENTS.

13,1986. Inv3stigation Of A Failed UF6 Shipping Cylinder.

SANDORN,Y.

SIMONEN,F.A.

NUREG/CR4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-NUREG/CR4483: REACTOR PRESSURE VESSEL FAILURE PROBA-TION OF OUENCHING OF SUPERHEATED DEBRIS BEDS UNDER BluTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-lZED THERMAL SHOCK EVENTS.

TOP-REFLOOD CONDITIONS. Final Report SCHAMR A.

SMITH,0.L.

NUREG/CR-4642-ROCK MASS SEAUNG EXPEH!?AENTAL ASSESS-NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL 1984. May 1985.

REPORT.

NUREG/CR4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-SCHWARZ,C.E.

ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS.

NUREG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA.

TION OF OUENCHING OF SUPERHEATED DEBRIS BEDS UNDER gggy,p TOP REFLOOD CONDITIONS. Final RepnrL NUREG/CR-4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC SAFETY STUDY.

SCOTT.W.S.

NUREG/CR-4330 V01: REVIEW OF UGHT WATER REACTOR REGU-SMINAD.

LATORY REQUIREMENTS. Volume 1:fdentification Of Regulatory Re-NUREG-1179 V02: RUPTURE OF MODEL 48Y UF6 CYUNDER AND RE.

qurements That May Have importance To Risk.

LEASE OF URANIUM HEXAFLUORIDE. Cylinder Overfill, March 12-NUREG/CR-4330 V02: REVIEW OF UGHT WATER REACTOR REGU.

13,1986. Investigation Of A Faded UF6 Shipping Cylinder LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE SOSEL,E.

TO RISK. Reactor Containment Leakage RatesMain Steam Isolation NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-Valve Leaka0s.

RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY POLYETHYLENE-SCOTTI.K S.

NUREG/CR4315 V01: EVALUATION OF NUCLEAR FACluTY DECOM.

SOO,P.

MISSIONING PRWECTS. Summary Status Report Three Mile Island NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-Urut 2 Reactor Coolant System & Systems Decontamination.

RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY NUREG/CR4315 V02-EVALUATION OF NUCLEAR FACluTY DECOM-POLYETHYLENE.

MISSIONING PROJECTS. Summary Status Report Three Mde Island NUREG/CR-4619: STRESS CORROSION CRACKING TESTS ON HIGH-Urut 2 Reactor Building Decontamination.

LEVEL. WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-NUREG/CR4315 V03: EVALUATION OF NUCLEAR FACluTY DECOM-POSITORY ENVIRONMENTS.

MISSIONING PRWECTS. Summary Status Report Three Mde Island Unit 2 Reactor Defuehng & Disassembly.

SOZER,A.

NUREG/CR-4315 V09: EVALUATION OF NUCLEAR FACluTY DECOM-NUREGICR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-MISSIONING PRCUECTS. Summary Status Report Three Mde Island ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS, Urut 2 Radioactive Weste And Laundry Shipments.

NUREG/CR4316: EVALUATION OF NUCLEAR FACluTY DECOMMIS.

SPRINGER,E.P.

SiONING PROJECTS STATUS REPORT. rtVMBOLDT BAY POWER NUREG/CR-4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT PLANT UP.lT 3 SAFSTOR DECOMMISSION NG.

IN THE UNSATURATED ZONE. Information Ard Data Ses.

NNSENEVAS.

STAHL,0.

NUREG-1160 INTERNATIONAL COOPERATsON DURING RADIOLOGi-NUREG/CR-4379 V04: LONG-TERM PERFORMANCE OF MATERIALS CAL EMERGENCIES. NRC Program Guidance For The Provision Of USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, Year Techrucal Advice To Fore'On Counterpart Orga wrations.

Four - April 1985 March 1986.

l

38 Personal Author index STARMANN,F.W.

VASSILAROS,M.G.

NUREG/CR-4349 LSL M2A COMPUTER PROGRAM FOR LEAST-NUREG/CR-4538 VO1: FRACTURE ANALYSIS OF WELDED TYPE 304 SOUARES LOGARITPMO ADJUSTMENT OF NEUTRON SPECTRA.

STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATON AND UMIT LOAD ANALYSIS.

NUREG/CR4620 METHOOOLOGIES FOR EVALUATING LONG-TERM YOLPE R L.

STABau2ATION DESIGNS OF URANIUM MILL TAluNGS IMPOUND-NUREG/CR-4620 METHODOLOGIES FOR EVALUATING LONG-TERM MENTS.

STABlu2ATION DESIGNS OF URANIUM MILL TAluNGS IMPOUND-STEINER J.L NUREG/CR4454: RELAP5/ MOD 2 CODE ASSESSMENT AT THE WALTON,M.

IDAHO NATIONAL ENGINEERING LABORATORY.

NUREG/CR-3150- SEISMICITY AND TECTONIC RELATONSHIPS FOR UPPER GREAT LAKES PRECAMBRIAN SHIELD PROVINCE FINAL N

G CR 4047; AN ASSESSMENT OF THE SAFETY IMPUCATIONS OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL WARD,D.S.

REPORT.

NUREG/CR.3162: DATA INPUT GUIDE FOR SWIFT ll.The Sandie l

ton Flow And Transport Model For Fractured Media Re-STUETZER,0.M.

W**'*j NUREG/CR4548. CORRELATION OF ELECTRICAL REACTOR CABLE FAILURE WITH MATERIALS DEGRADATON.

WATERMAN,M.E.

NUREG/CR4488: VENTING OF NONCONDENSIBLE GAS FROM THE SWYLER K.J UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-NUREd/CR4607: THE EFFECTS OF ENVIRONMENT AND GAMMA iR-BEND VALVES.

RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY POLYETHYLENE-WEps,S.W.

NUREG/CR-4353: ASSESSMENT OF POST-CRITICAL-NEAT FLUX SZY W SMLT MODELS WITH LEHIGH NONEOUILIBRIUM DATA.

NUR?.G-1021 R02: OPERATOR UCENSING EXAMINER STANDARDS.

WEBSTER,C.S.

THEOFANOUS,T.G.

NUREG/CR-4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT NUREG/CR-3701: REMIXA COMPUTER PROGRAM FOR TEMPERA.

FOR VERTICAL FISSION PRODUCT RELEASE APPARATUS IN HOT TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER CELL 8,8UILDNG 4501.

INTERRUPTON OF NATURAL CIRCULATON.

NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-S S CAL LATED FOR THE NRC PRESSURIZED THERMAL N EG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA.

TON OF QUENCHING OF SUPERHEATED DEBRIS BEDS UNDER THOME.D.K.

TOP-REFLOOD CONDITIONSFmal Report.

NUREG/CR4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER W

AMER,E (EMAT) DEFECT CHARACTERIZATON OF NUCLEAR REACTOR NOREG-1209: PROGRAM Pt.AN FOR ENVIRONMENTAL QUAUFICA.

PlPING WELDS Phase 1 Final Report,0ctober 1985. March 1986' TION OF MECHANICAL AND DYNAMIC (INCLUDING SEISMIC)

QUAUFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENT THORNTONR.H.

NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTION EXPERI.

PROGRAM (EDOP).

ENCE IN NUCLEAR POWER PLANTS.

THURSER.C.H.

NUREG/CR-2331 VOS N3: SAFETY RESEARCH PROGRAMS SPON-NUREG/CR-4580- STONY BROOK SEISMIC NETWORK ON LONG SORED BY OFFICE OF NUCLEAR REGULATORY ISLAND,NEW YORK. Final Report (Septernber 1979. March 1985).

RESEARCH.Ouarterty Progress Report. July-Septernber 1985.

THUR0000,M.J.

WESTON,LM.

NUREG/CR-3262 V01: COBRA-NCA THERMAL-HYDRAULIC CODE NUREG/CR-4463: HUMAN FACTORS IN ANNUNCIATOR / ALARM FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR SYSTEMS ANNUNCIATOR EXPERIMENT PLAN L COMPONENTS.Volurne 1. Equations And Constitutrve Models.

NUREb/CR-3262 V02: COBRA.NC-A THERMAL-HYDRAUUC CODE WHATLEY,S.K.

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADIONU-CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-COMPONENTS. Volume 2: COBRA-NC Numencal Solution Methods.

NUREG/CR-3262 V07: COBRA-NC.A THERMAL-HYDRAUUC CODE LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR April-June 1985.

COMPONENT. Volume 7 Assessment Manual for Contamment Applica-NUREG/CR-4454: RELAP5/ MOD 2 CODE ASSESSMENT AT THE TINGLE A.G.

IDAHO NATIONAL ENGINEERING LABORATORY.

NUREG/CR4488: VENTING OF NONCONDENSIBLE GAS FROM THE NUREG/CR-4557: A REVIEW OF ISSUES RELATED TO IMPROVING UPPER HEAD OF A B4W REACTOR VESSEL USING HOT LEG U-NUCLEAR POWER PLANT DIESEL GENERATOR REUABluTY.

BEND VALVES.

TRAVIS,J.R.

NUREG/CR4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT WHEELER,C L.

FOR VERTICAL FISSION PRODUCT RELEASE APPARATUS IN HOT NUREG/CR-3262 V01: COBRA-NCA THERMAL-HYDRAUUC CODE FOR TRAN* LENT ANALYSIS OF NUCLEAH REACTOR CELL 8.BUILDNG 4501.

COMPONENTS. Volume 1. Equations And ConstrtutNo Models.

NdREG/CR 3262 V02-COBRA-NC:A THERMAL-HYDRAULIC CODE VAN FLEET.LG.

NUREG/CR4484: STATUS OF ACTIVITIES FOR INSPECTING WELD FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR COMPONENTS. Volume 2: COBRA-NC Numerical Solution Methods.

OVERLAID PIPE JOINTS.

NUREG/CR-3262 V07; COBRA-NCA THERMAL-HYDRAULIC CODE VAN HOUTEM,L.P.

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUREG/CR-4583 VC1: DEVELOPMENT AND VAUDATION OF A REAL-COMPONENT. Volume 7 Assessment Manual for Containment Applica.

TIME SAFE-UT SYSTEM FOR THE INSPECTION OF UGHT WATER tions.

l COMPONENTS Semi-Annual Report For Apnl 1984-September 1984.

I WHITE.J.D.

VAN ZYL.D.

NUREG/CR-3770: PRELIMINARY DEVELOPMENT OF AN INTEGRAT.

NUREG/CR4620- METHODOLOGIES FOR EVALUATING LONG-TEHM ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL STABlu2ATION DESIGNS OF URANIUM MILL TAILINGS IMPOUND.

SHOCK AS APPLIED TO THE OCONEE UNIT 1 NUCLEAR POWER MENTS.

PLANT.

Personal Autitor Index 39 WNmaan.o o.

WONo,C.C.

NUREG/CR44% HISTORICAL

SUMMARY

OF THE HEAVY-SECTON NUREG/CR4507: HECTR VERSION 1.5 USER'S MANUAL STEEL TECHNOLOGY PROGRAM AND SOME RELATED ACTIVITIES IN UGHT-WATER REACTOR PRESSURE VESSEL SAFETY RE-WYNVEENAA.

NUREG/CR4511: ASSESSMENT OF THE ADEQUACY OF THE CALI-WIugoWSKI,o.

BRATIONS PERFORMED BY COMMERCIAL CAUBRATION SERV-NUREG/CR-4572: NRC LEAK-BEFORE-BREAK (LBB.NRC) ANALYSIS ICES FOR ONIZING RADIATON SURVEY INSTRUMENTS.

METHOD FOR CIRCUWFERENTIALLY THROUGH-WALL CRACKED PtPES UNDER AXIAL PLUS BENDING LOADS.Topcal Report XUE,D.

NUREG/CR-4207: FAULT TREE APPUCATON TO THE STUDY OF g

NUREG/CR-4642 ROCK MASS SEAUNG EXPERIMENTAL ASSESS-SYSTEMS INTERACTIONS AT INDIAN POINT 3.

MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June NUREG/CR4404: ANALYSIS OF ALLOWED OUTAGE TIMES AT 1964 - May 1965.

BYRON GENERATING STATON.

l WILLIAa8SAS.L.

YAZDAN0006T.A.

NUREG/CR-3064 V01: COMPUTATIONAL METHODOLOGY FOR OAK NUREG/CR-4642: ROCK MASS SEAUNG EXPERIMENTAL ASSESS-1 RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC.

MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June TOR (BSR)-Croes-Secton Generston And Vahdoton Volume 1.

NUREG/CR-3064 V02 COMPUTATONAL METHOD 6 LOGY FOR OAK 1964 May 1965.

RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-TOR (BSR).The VICTORR Input Proceemng Code For The Bold Ven.

YOUNo8LOOOAW.

ture System, volume 11.

NUREG/CR-4207: FAULT TREE APPUCATION TO THE STUDY OF SYSTEMS INTERACTONS AT INDIAN POINT 3.

^

/CR 12 AN ASSESSMENT OF SAFETY MARGINS IN ZlRCA-BYRON GENERATING STATON.

LOY OXfDATON AND EMBRITTLEMENT CRITERIA FOR ECCS AC-YUELYS-INKSl8 WILSON,J.M.

NUREG/CR4549: DETERMINATION OF APPENDIX K CONSERV-NUREG/CR4402 V03: HIGH TEMPERATUf1E GAS. COOLED REACTOR ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT USING TRAC-PD2/ MODI CODE.

EVALUATON.Ouarterty Progrose Report.My 1 September 30,1965.

m 1

i I

b I

l r

I I

I 4

1 i

l l

)

l

_,.,_________,y_.

Subject Index This index was developed from keywords moved later when a reasonable thesaurus and word strings in titles and abstracts.

has been developed through experience.

During this development period, there will Suggestions for improvements are wei-be some redundancy, which will be re-come.

S.O Finite Difference Code Aging NUREG/CR3262 V02: COBRA-NC:A WERMAL-HYDRAUUC CODE NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR SORED BY OFFICE OF NUCLEAR REGULATORY COMPONENTS. Volume 2 COBRA-NC Numencal Soluton Methods.

RESEARCH.Ouarterty Progress Report. July-September tir4 NUREG/CR3262 V07: COBRA-NC:A THERMAL-HYDRAULIC CODE NUREG/CR4490 V01: UGHT-WATER-REACTOR SAFETY 'AATERIALS FOR TPANSIENT ANALYSIS OF NUCLEAR REACTOR ENGINEERING RESEARCH PROGRAMS.Quartert Progress COMPONENT. Volume 7. Assessment Manual for Cont.unment Applica-Report. January-March 1985.

hone.

l NUREG/CR-4503 V01: LONG TERM EMBRITTLEMENT OF CAST-304 Steiniese Steel DUPLEX STAINLESS STEELS IN LWR SYS. Annual Report. October i

NUREG/CR4241: CHEMICAL ASPECTS OF CESIUM IODOE INTERAC-1964 September 1985.

TION IN STEAM WITH 304 STAINLESS STEEL AND INCONEL-600.

i AC Power System NUREG/CR4463: HUMAN FACTORS IN ANNUNCIA TOR / ALARM NUREG/CR-4589: REVIEW OF SELECTED AREAS OF YANKEE ROWE SYSTEMS. ANNUNCIATOR EXPERIMENT PLAN 1.

PROBA81USTIC SAFETY STUDY.

Alarm Reduction NUREG/CR-4463: HUMAN FACTORS IN ANNUNOATOR/ ALARM EG/CR4316: EVALUATION OF NUCLEAR FAQUTY DECOMMIS-SYSTEMS. ANNUNCIATOR EXPERIMENT PLAN L SiONING PROJECTS STATUS REPORT HUMBOLDT BAY POWER PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

g ASTM A108 Stool NUREG/CR4404: AtlALYSIS OF ALLOWED OUTAGE TIMES AT NUREG/CR-4539: INVESTIGATION OF TEARING INSTABluTY PHE-BYRON GENERATING STATION.

NOMENA IN ASTM A106 STEEL All0FSteel Almormal Occurrences NUREG/CR4579: APPUCATION OF THE KEY JJRVE AND MULTI-NUREG4000 V00 N04: REPORT TO CONGRESS ON ABNORMAL SPEOMEN TECHNIOUES TO DYNAMIC J-R CURVE TESTING OF OCCURRENCES October -December 1985.

ALLOY STEEL Abstract Altenotive Weste Form NUREG4304 V11 N01: REGULATORY AND TECHNICAL REPORTS.Compdation For First Quarter 1986. January-March.

NUREG/CR3472 V02 SURFACE PROPERTIES AND PERFORMANCE PREDICTION OF ALTERNATIVE WASTE FORMS. Final Report.

Acektent NUREG-0900 Rot: NUCLEAR POWER PLANT SEVERE ACCIDENT RE.

Ambient Redletion Level SEARCH PLAN NUREG4837 V05 N04: NRC TLD DIRECT RADIATION MONITORING NUREG 1160 INTERNATIONAL COOPERATION DURING RADIOLOGI-NETWORetProgress Report. October December 1985.

CAL EMERGENOES. NRC Program Guidance For The Provioson Of NUREG4837 V06 N01: NRC TLD DIRECT RADIATION MONITORING ess RW JanuaWa@ m NUR C

-3 RA

/

DE D

ASSESSMENT:

LOBIINTERMEDIATE BREAK TEST B-R1M.

NUREG/CR4027:

TRAC-PF1/ MOD 1 INDEPENDENT Annual Report ASSESSMENT. Condensation in Stratifled Cocurrent Flow.

NUREG/CR3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-NUREG/CR4315 V01: EVALUATION OF NUCLEAR FAOUTY DECOM.

TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

MISSIONING PRCUECTS. Summary Status Report Three Mde taland NUREG/CR4124 V02 NDE OF STAINLESS STEEL AND ON-UNE Urut 2 Reactor Coolant System & Systems Decontamination.

LEAK MONITORING OF LWRS. Annual Report October 1984 - Sep-NUREG/CR4315 V02: EVALUATION OF NUCLEAR FAQUTY DECOM-tomber 1985.

MISSIONING PRCUECTS. Summary Status Report Three Mde Island N'JREG/CR4379 V04: LONG-TERM PERFORMANCE OF MATERIALS Urvt 2 Reactor Bud' ding Decontaminat6an.

USED FOR HIGH-LEVEL WASTE PACKAGING Annual Report. Year NUREG/CR4315 V03: EVALUATON OF NUCLEAR FAOUTY DECOM-Four April 1985 March 1986.

MISFIONING PROJECTS. Summary Status Report Three Mde Island NR /

38 U BE A N CONTAINMENT UNDER CR4 M

FACTORS IN ANNUNCIATOR / ALARM NUREG/CR4467: RELATIVE IMPORTANCE OF INDIVIDUAL ELE.

SYSTEMS: ANNUNCIATOR EXPERIMENT PLAN L MLNTS TO REACTOR ACODENT CONSEQUENCES ASSUMING I

EQUAL RELEASE FRACTIONS.

Appecadone Manual NUREG/CR-4488: VENTING OF NONCONDENSIBLE GAS FROM THE NUREG/CR4497: NRCPAGE APPUCATIONS MANUAL UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-BEND VALVES.

Acessement Manuel NUREG/CR4569: A REVIEW OF THE SEVERE ACODENT RISK RE-NUREG/CR-3262 V07: COBRA-NC:A THERMAL-HYDRAULIC CODE DOCTION PROGRAM (SARRPI CONTAINMENT EVENT TREES.

FOR TRANS.ENT ANALYSIS OF NUCLEAR REACTOR NUREG/CR-4603-APPRAISING ATMOSPHERIC TRANSPORT AND COMPONENT. Volume 7: Assessment Manual for Containment Applica-DIFFUSION MODELS FOR EMERGENCY RESPONSE FACluTIES.

tions.

Actinkle Solubluty

.m NUREG/CR-4 PPRAISING ATMOSPHERIC TRANSPORT AND AND ATION OF ELE D T N DES DIFFUSION MODELS FOR EMERGENCY RESPONSE FACILITIES.

Ager,-.u s NUREG/CR3572 V02 DETERMINATION OF METABOUC DATA AP.

Atomic Energy Act PROPRIATE FOR HLW DOSIMETRY.ll. Gastrointestinal Absorption.

NUREG4980 R02 NUCLEAR REGULATORY LEGISLATION.

41

42 Subject Index l

Austenleic Steintees Steel NUREG/CR4642 ROCK MASS SEAUNG. EXPERIMENTAL ASSESS.

NUREG/CR4619 STRCSS CortROSON CRACK!NG TESTS ON HIGH-MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June LEVEL-WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-1964. May 1985.

POSITORY ENVIRONMENTS.

Borehole Well OrtHing Damage h Depresourtantion System NUREG/CR 4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE I

NUREG/CR-3980 CLOSEOUT OF IE BULLETIN 8041.Operebility Of WALL DRILLING DAMAGE IN BASALTO ROCKS.

Automatic Depressurtration Systern (ADS) Velve Pneumat6c Supply.

BoroeHiosto Glees Weste Form NUREG/CR-3472 V02: SURFACE PROPERTIES AND PERFORMANCE N

G 00 3

DARD REVIEW PLAN FOR THE REVIEW PPEDN OF AGERNAM WASTE MSBnal W OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Editon. Revision 3 to Section 9.2.2. " Reactor Auxshery Brealmen Test Coohng Water Systems."

NUREG/CR-4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE Ausniery Feodoieter System WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

NUREG-1201: REPORT OF THE INDEPENDENT AD HOC GROUP FOR THE DAVIS BESSE INCIDENT.

Away-FrosHleector Storage HUMAN FACTORS PROGRAM PLAN.

NUREG/CR-4516: INTERNATONAL SAFEGUARDS AT FACluTIES EM-PLOYING SPENT FUEL ROO CONSOUDATON.

Bulk Shleiding Reacter NUREG/CR-3064 V01: COMPUTATONAL METHODOLOGY FOR OAK BeOPOftT/ MAX 11 RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-NUREG/CR2675 V05: RELEVANCE OF BOTIC PATHWAYS TO THE TOR (BSR)-Crose-Section Generston And Vendston, Volume 1.

LONG TERM REGULATON OF NUCLEAR WASTE DISPOSAL (Esti-mation Of Radletion Dose To Men Resulting From Beotic TransportThe Buoyant Plume BIOPORT/ MAX 11 Software Package)-

NUREG/CR-3701: REMIX.A COMPUTER PROGRAM FOR TEMPERA.

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTION AFTER DMD VENWRE INTERRUPTON OF NATURAL CIRCULATION NUREG/CR-3064 V02: COMPUTATIONAL METHOOOLOGY FOR OAK NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-RfDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL TOR (BSR).The VICTORR Input Processing Code For The Bold Ven.

SHOCK STUDY ~

ture System, Volume 14.

COSRA-NC gas,sg NUREG/CR4642 ROCK MASS SEALING. EXPERIMENTAL ASSESS.

NUREG/CR 3262 V01: COBRA-NC:A THERMAL-HYDRAUUC CODE MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR 1964. May 1985.

COMPONENTS. Volume 1:Equatione And Conettusveunma NUREG/CR-3262 V02 COBRA-NC-A THERMAL-HYDRAUUC CODE Seesitic Roche FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUREG/CR4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE COMPONENTS. Volume 2: COBRA-NC Numerical Soluton Methods.

WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

NUREG/CR-3262 V07: COBRA NCA THERMAL-HYDRAUUC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR

~

G/CR-4588 V01: SOfL-STRUCTURE INTERACTON.Vol 1:Influ-ence Of Layenng.

CONTEMPT 4 w

NUREG/CR4594: ESTIMATED SAFETY SIGNIFICANCE OF GENERIC NUREG/CR4330 V01: REVIEW OF UGHT WATER REACTOR REGO-SAFETY ISSUE 61.

l LATORY REQUIREMENTS. Volume 1:ldentrhcation Of Regulatory Re-NU CR 4330 [RINI W TER REACTOR REGU-l Cast Steinises Steel NUREG/CR-4124 V02: NDE OF STAINLESS STEEL AND ON-UNE LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA.

LEAK MONITORING OF LWRS. Annual Report October 1964. Sep-i TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE Al eactor Containment Leakage Ratee. Main Steam leolation NU 90 V01: UGHT WATER-REACTOR SAFETY MATERIALS

~~

ENGINEERING RESEARCH PROGRAMS. Quarterly Progrees gese Report. January Marcti 1985.

NUREG-1156 ACCURACY AND DETECTON UMITS FOR BOASSAY NUREG/CR4503 V01: LONG TERM EMBRITTLEMENT OF CAST.

MEASUREMENTS IN RADIATON PROTECTION. STATISTICAL DUPLEX STAINLESS STEELS IN LWR SYS. Annual Report. October CONSOERATONS.

1964. September 1965.

Coment Swelling Pressure Oneteral t, :. INTERNATONAL COOPERATON DURING RADIOLOGI-NUREG/CR-4642 ROCK MASS SEAUNG. EXPERIMENTAL ASSESS-NUREG 1160:

CAL EMERGENCIES. NRC Program Gudence For The Provision Of MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June Technical Advice To Foreign Counterpart Orgaruzations.

1984. May 1985.

M Cesium lodido Chemistry NUREG-1156: ACCURACY AND DETECTION LIMITS FOR BIOASSAY NUREG/CR-4241: CHEMICAL ASPECTS OF CESIUM IODIDE INTERAC.

MEASUREMENTS IN RADIATON PROTECTON. STATISTICAL TON IN STEAM WITH 304 STAINLESS STEEL AND INCONEL-800.

CONSIDERATONS.

Chemical Hemord NUREG 1198: RELEASE OF UF6 FROM A RUPTURED MODEL 48Y 601: TECHNICAL CONSIDERATIONS AFFECTING PREP-CYUNDER AT SEQUOYAH FUELS CORPORATON ARATON OF ON-EXCHANGE RESlNS FOR DISPOSAL FACluTY. Lessons-Leamed Report 560 tic Transport Circuit Fauvre l

NUREG/CR2675 V05: RELEVANCE OF BOTO PATHWAYS TO THE LONG-TERM FiEGULATON OF NUCLEAR WASTE DISPOSAL (Esti.

NURFG/CR-4548: CORRELATION OF ELECTRICAL REACTOR CABLE mation Of Radiahon Dose To Man Resulting From Biobe TransportThe FAILURE WITH MATERIALS DEGRADATION.

BIOPORT/MAxit Software Package).

g Borehole NUREG/CR4572: NRC LEAK-BEFORE-BREAK (LBB.NRC) ANALYSIS NUREG/CR4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

PIPES UNDER AXIAL PLUS BENDING LOADS. Topical Report

Subject index 43 Clodding Comperleon Studlee NUREG/t.A-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC.

NUREG/CR4316: EVALUATON OF NUCLEAR FACluTY DECOMMIS-TOR PRESSURE BOUNDARY COMPONENTS Annual Report for 1985.

SiONING PROJECTS STATUS REPORT - HUMBOLDT BAY POWER NUREG/CR-4595: ENHANCEMENT TO THE LAFM COMPUTER CODE-PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

Closed Loop Cooling Water System Component Cco6 ng Water System NUREG4600 09 2 2 R3: STANDARD REVIEW PLAN FOR THE REVIEW NUREG-0800 C9.2.2 R3: STANDARD REVIEW PMN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS LWR EditertRevision 3 to Secton 9.2.2, " Reactor Auxiliary PLANTS. LWR Edition. Revision 3 to Secten 9.2.2, " Reactor Auxiliary Coolmg Water Systems."

Cooling Water Systems."

Closoout Computettenal Methodology NUREG/CR-3960: CLOSEOUT OF IE BULLETIN 80-01.Operatulity Of NUREG/CR-3064 V02: COMPUTATONAL METHODOLOGY FOR OAK Automate Depressuruaton System (ADS) Valve Pneumate Supply.

RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-NUREG/CR-3962: CLOSEOUT OF IE BULLETIN 80 20 Failures Of Wes-TOR (BSR).The VICTORR Input Processing Code For The Bold Ven-tinghouse Type W-2 Spnng Return To Neutral Control Switches-ture System, Volume IL Cohaue Computer program NUREG/CR-4543: FIRST RESULTS FROM ELECTION-PHOTON NUREG/CR-2675 V05: RELEVANCE OF BIOTIC PATHWAYS TO THE DAMAGE EQUIVALENCE STUDIES ON A GENERIC ETHYLENE-PRO.

LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esth PYLENE RUBBER.

mation Of Radeten Dose To Man Resulting From Beobe TransportThe BIOPORT/MAXII Software P REG /CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-R E

REACTOR ORR) AN U R

SORED BY OFFICE OF NUCLEAR REGULATORY l

RESEARCH.Ouarterty Progress aeport. July-September 1985-TOR (BSR):The VICTORR input Processmg Code For The Bold Ven.

ture System. Volume il' INTRUDER DOSE PATHWAY ANALYSIS FOR NUREG/CR-3162. DATA INPUT GUIDE FOR SWIFT ll.The Sanda NUREG/CR-3620 S01:

Waste-Isolation Flow And Transport Model For Fractured Media Re-THE ONSITE DISPOSAL OF RADOACTi%E WASTES.The ONSITE/

NURE /CR 3262 V01: COBRA-NC:A THERMAL-HYDRAUUC CODE NUR /CR 1 R X COMPUTER PROGRAM FOR TEMPERA-FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR TURE TRANSIENTS DUE TO HIGH PRESSURE NECTON AFTER COMPONENTS. Volume 1 Equations And Constitutive Models.

NUREG/CR-3262 V02: COBRA-NC:A THERMAL-HYDRAUUC CODE INTERRUPTION OF NATURAL CIRCULATION NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR COMPONENTS Volume 2. COBRA-NC Numencal Solution Methods.

SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL SHOCK STUDY NUREG/CR-3262 V07: COBRA-NC:A THERMAL HYDRAUUC CODE NUREG/CR-397(N TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT:

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR LOB 1 INTERMEDIATE BREAK TEST B-A1M COMPONENT. Volume 7. Assessment Manual for Containment Applica-NUREG/CR-4027:

TRAC-PF1/ MOD 1 INDEPENDENT A

SS N ton in Stem Coment h NU EG/CR-3441: RADONE:A COMPUTER CODE FOR SIMULATING FAST TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDI-Concrete Containment TIONS AND TWO4AYER RADIONUCUDE CONCENTRATIONS IN-NUREG/CR-3957: REUABluTY ASSESSMENT AND PROBABluTY CLUDING THE EFFECT OF BED-DEPOSITION IN CONTROLLED BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS NUR G/CR 7 EL M NA Y DEVELOPMENT OF AN INTEGRAT.

ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL Container SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER NUREG/CR-4315 V09 EVALUATION OF NUCLEAR FACIUTY DECOM-PLANT W/TWO OVERSIZE DRAW)NGS.

NUREG/CR4319: NUCRAC - A CODE FOR THE ESTIMATION OF AD" MISSIONING PROJECTS. Summary Status Report Three Mile Island VERSARY-ACTION CONSEQUENCES IN THE NUCLEAR POWER Urut 2 Radioactive Waste And Laundry Shipments.

NUREG/CR4601: TECHNICAL CONSIDERATIONS AFFECTING PREP.

FUEL CYCLE.

ARATION OF ON-EXCHANGE RESINS FOR DISPOSAL NUREG/CR4349: LSL-M2.A COMPUTER PROGRAM FOR LEAST*

NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-SOUARES LOGARITHMIC ACMUSTMENT OF NEUTRON SPECTRA.

RADIATON ON THE MECHANICAL PROPERTIES OF HIGH DENSITY NUREG/CR4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK POLYETHYLENE LOSS-OF COOLANT ACCIDENTS IN A RESAR-3S PLANT

NUREG/CR-4619: STRESS CORROSION CRACKING TESTS ON HIGH-NUREG/CR4384-BREAK SPECTRUM ANALYSIS FOR SMALL BREAK LOSS-OF4OOLANT ACCIDENTS IN A RESAR-3S PLANT.

LEVEL WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-POSITORY ENVIRONMENTS.

NUREG/CR-4467: RELATIVE IMPORTANCE OF INDIVIDUAL ELE.

MENTS TO REACTOR ACCIDENT CONSEQUENCES ASSUMING Containment NUR G/ 4497 A E UCATIONS MANUAL NUREG/CR4507: HECTR VERSION 1'5 USER'S MANUAL SEARCH PLAN.

NUREG/CR-4549. DETERMINATION OF APPENDIX K CONSERV-NUREG-1152-MILLSTONE 3 RISK EVALUATION REPT;AN OVERALL ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBABI-LISTIC SAFETY STUDY.

USING TRAC-PD2/ MOD 1 CODE' MANUALA COMPUTER CODE TO NUREG/CR-4561: FIRAC USER'S NUREG/CR4330 V02: REVIEW OF UGHT WATER REACTOR REGU-SIMULATE FIRE ACCIDENTS IN NUCLEAR FACIUTIES LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-NUREG/CH-4595: ENHANCEMENT TO THE LAFM COMPUTER CODE.

TOAY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE TO RISK. Reactor Containment Leakage Rates Main Steam isolation CodeC2-, _..t Valve Leakage..

NUREG4900 R01: NUCLEAR POWER PLANT SEVERE ACCIDENT RE.

NUREG/CR4338: TELLURIUM BEHAVIOR IN CONTAINMENT UNDER SEARCH PLAN LIGHT WATER REACTOR ACCIDENT CONDITONS.

NUREG/CR4493: AN EXPERlWENTAL AND ANALYTICAL INVESTIGA-Columnar Grain Structure TION OF QUENCHING OF SUPERHEATED DEBRIS BEDS UNDER NUREG/CR4484: STATUS OF ACTINITIES FOR INSPECTING WELD TOP-REFLOOD CONDITIONS Final Report OVERLAID PIPE JOINTS.

NUREG/CR4569: A REVIEW OF THE SEVERE ACCIDENT RISK RE-DUCTON PROGRAM (SARRP) CONTAINMENT EVENT TREES.

Combustion NUREG/CR-4594: ESilMATED SAFETY SIGNIFICANCE OF GENERIC NUREG/CR4507. HECTR VERSION 1.5 USER'S MANUAL SMETY ISSUE 61.

Commercial Calibretton Service Containment Analyste NUREG/CR4511: ASSESSMENT OF THE ADEQUACY OF THE CALi-NUREG/CR.3262 V02: COBRA-NC.A THERMAL-HYDRAUUC OODE BRATIONS PERFORMED BY COMMERCIAL CAUBRATON SERV-FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR ICES FOR lONIZING RADIATION SURVEY INSTRUMENTS.

COMPONENTS Volume 2 COBR A-NC Numencal Solution Methods.

i

44 Subject Index NUREG/CR-3262 V07: COBRA-NC A THERMAL-HYDRAUUC CODE NUREG/CR4568: A HANDBOOK FOR OUICK ESTIMATES:A Method FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR For Developmg Quid Approxunate Estimates Of Costs For Genenc Ac-COMPONENT.Voksne 7: Assessment Manual for Contamrnent Applica-twwis For Nuclear Power Plants.

tions.

NUREG/CR-4627; GENERIC COST ESTIMATES. Abstracts From Genenc Studies For Use in Propenng Regulatory impact Analyses.

NUREG/CR-4374 V03: A REVIEW OF THE OCONEE-3 PROBABluSTIC Cost-Effectivenees Analyele RISK ASSESSMENT CONTAINMENT NUREG/CR4374 V03: A REVIEW OF THE OCONEE 3 PROBABluSTIC PERFORMANCERADOLOGICAL SOURCE TERMS AND RISK ESil.

RISK ASSESSMENT CONTAINMENT MATES.

PERFORMANCE. RADIOLOGICAL SOURCE TERMS AND RISK ESTi-MATES.

NUREG/CR-4520 PREDICTIVE GEOCHEMICAL MODEUNG OF CON-Crecis Or m TAINMENT CONCENTRATIONS IN LABORATORY COLUMNS AND IN PLUMES MIGRATING FROM URANIUM MILL TAluNGS WASTE IM-NUREG/CP 0067 V01: PROCEE DINGS OF THE SECOND lAEA SPE-POUNDMENTS. Final Report.

CIAUSTS' MEETING ON SUBCRITICAL CRACK GROWTH. Sessions 1 And ll Held At Sendai, Japan.May 15 17,1985.

Contemenellen NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-NUREG 1188: THE AUBURN STEEL COMPANY RADIOACTIVE CON.

TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

TAMINATION INCIDENT, NUREG/CR-4490 VOI: UGHT-WATER-REACTOR SAFETY MATERIALS ENGINEERING RESEARCH PROGRAMS;Ouarterty Progress Control Air System Report. January-March 1985.

NUREG/CR4413: LOSS OF CONTROL AIR AT BRC4-S FERRY UNIT ONE - ACCIDENT SEQUENCE ANALYSIS.

Curie NUREG/CR4315 V09: EVALUATION OF NUCLEAR FACIUTY DECOM-Control Room MISSIONING PROJECTS. Summary Status Report Three Mele Island NUREG4985 R02:

U.S.

NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.

Unit 2 Radioactrve Waste And Laundry Shipments.

Cytender Rupture EG 26 AN ASSESSMENT OF THE SAFETY IMPLICA-NUREG t179 V02: RUPTURE OF MODEL 48Y UF6 CYLINDER AND RE-I TONS OF CONTROL AT THE CALVERT CUFFS 1 NUCLEAR PLANT.

LEASE OF URANIUM HEXAFLUORIDE. Cylinder Overfill. March 12-13.1986. Investigation Of A Failed UF6 SNppmg Cylmder.

Control Systems Program Aseseement NUREG-1198. RELEASE OF UF6 FROM A RUPTURED MODEL 48Y NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS CYUNDER AT SEQUOYAH FUELS CORPORATON OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL FACluTY. Lessons-Learned Reoort.

REPORT.

NUREG/CR-4449 A PWR HYBRID COMPUTER MODEL FOR ASSESS.

DC Power System ING THE SAFETY IMPUCATONS OF CONTROL SYSTEMS.

NUREG/CR4589: REVIEW Of SELECTED AREAS OF YANKEE ROWE PROBABluSTIC SAFETY STUDY.

Convection Film Selling Model NUREG/CR4353: ASSESSMENT OF POST-CRITICAL-HEAT FLUX Damaged Zone MODELS WITH LEHIGH NONEQUluBRIUM DATA.

NUREG/CR-4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE l

Core Aeoseement WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

NUREG/CR-4642 ROCK MASS SEALING. EXPERIMENTAL ASSESS-NUREG/C%4364: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK LOSS 6 COOLANT ACCIDENTS IN A RESAR-3S PLANT.

MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June 1984 May 1985.

Core Demog>

NUR 94 ESilMATED SAFETY SIGNIFICANCE OF GENERIC RE 15 V01: EVALUATION OF NUCLEAR FACluTY DECOM-l MISSIONING PROJECTS. Summary Status Report Three Mile Isisnd Core Melt Unit 2 Reacter Coolant System & Systems Decontammation.

r NUREG-184: MILLSTONE 3 RISK EVALUATON REPT:AN OVERALL NUREG/CR4315 V02 EVALUATON OF NUCLEAR FACIUTY DECOM-REVIE' d AND EVALUATON OF THE MILLSTONE UNIT 3 FAOBABI-MISSIONING PROJECTS. Summary Status Report Three Mile Island t

LISTIMlU'ETY STLDY.

Urvt 2 Reactor Building Decontammation.

i f

NURFUT4-4142 A REVIEW OF THE MILLSTONE 3 PROBABILISTIC NUREG/CR-4315 V03: EVALUATON OF NUCLEAR FACIUTY DECOM-l SAFETY STUDY.

MISSIONING PROJECTS. Summary Status Report Three Mile island NUREG/CR4569: A R2 VIEW OF THE SEVERE ACCIDENT RISK RE-Urut 2 Reactor Defuelmg & Disassembly.

(

DUCTION PROGRAM (SARRP) CONTAINMENT EVENT TREES.

Data Compliation R

CR-V02 COMPUTATIONAL METHODOLOGY FOR OAK RED ES R PRO CTS RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC.

TOR (BSR).The VICTORR Input Processing Code For The Bold Ven.

Data input Guide ture System. Volume ll-NUREG/CR-3182 DATA INPUT GUIDE FOR SWIFT ll.The Sandia p

Waste-Isolation Flow And Transport Model For Fractured Media Re-NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC.

I**** 444-TOR PRESSURE BOUNDARY COMPONENTS Annual Report for 1985.

Debrte Quenching Cool NUREG/CR-4493: AN EXPERIMENTAL AND AP.ALYTICAL INVESTIGA-NUREG/CR-4316. EVALUATION OF NUCLEAR FACluTY DECOMMIS.

TION OF OUENCHING OF SUPERHEATED DEBRIS BEDS UNDER SiONING PROJECTS STATUS REPORT HUMBOLDT BAY POWER TOP-REFLOOD CONDIT'ONS Final Report.

PLANT UNIT 3 SAFSTOR DECOMMISSIONING NUREG/CR-4330 V01: REVIEW OF LIGHT WATER REACTOR REGU.

Decay Heat Removal LATORY REQUIREMENTS. Volume 1.identrlication Of Regulatory Re.

NUW/CR-4621: FLOW VISUAU2ATON EXPERIMENT ON HOT-LEG queroments That May Have importance To Risk.

U. BEND TWO-PHASE NATURAL CIRCU' ATION PHENOMENA.

J NUHLG/CR4330 V02 REVIEW OF UGHT WATER REACTOR REGU-LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-Dec-a TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE NUREGICR-3587: IDENTIFICATION AND EVALUATON OF FACluTA.

TO RISK.Reactar Contamment Leakage RatesMac Steam Isolation TON TECHNIOUES FOR DECOMMISSIONING LIGHT WATER Vapve Leakage...

POWER REACTORS.

NUREG/CR-45tt TECHNOt OGY. SAFETY AND COSTS OF DECOM-NUREG/CR-4315 V01: EVALUADON OF NUCLEAR FACILITY DECOM-MISSIONING NUCLEAR FUEL CYCLE FACluTIES CLASSIFICATON MISSIONING PRGjECTS. Surrmary Status Report Three Mile Island OF DECOMMISSONING WASTE.

Urut 2 Reactor Coolant System & Systems Decontammation.

l l

L

Subject index 45 NUREG/CR-4315 V02 EVALUATON OF NUCLEAR FACluTY DECOM-TIONS AND TWO-LAYER RADIONUCUDE CONCENTRATIONS IN-M4SSONING PROJECTS. Summary Status Report Three Mile Island CLUDING THE EFFECT OF BED-DEPOSITION IN CONTROLLED Unit 2 Reactor Buildme Decontammation.

RIVERS AND TIDAL ESTUARIES.

NUREG/CR-4315 V03: EVALUATION OF NUCLEAR FAQUTY DECOM-MISSIONING PROJECTS. Summs y Status Report Three Mde island N=Paaa' Unit 2 Reactor DefusenD & r" NUREG/CR4315 V09: EVALUATION @"" NUCLEAR FACIMTY DECOM-NUREG-1188: THE AUBURN STEEL COMPANY RADIOACTIVE CON-TAMINATON INCIDENT.

MISSONING PRCUECT3. Summary Status Report Three Mde Island NUREG/CR-4801: TECHNICAL CONSIDERATIONS AFFECTING PREP.

NUR G/

EVA AION F' OUTY DECOMMIS.

SONING PROJECTS STATUS REPORT. HUMBOLDT BAY POWER Disturtmed Zone PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

NUREG/CR-4841: EXPERIMENTAL ASSESSMENT OF BOREHOLE NUREG/CR-4519 TECHNOLOGY, SAFETY AND COSTS OF DECOM-WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

MISSIONING NUCLEAR FUEL CYCLE FACluTIES CLASSIFICATION OF DECOMMISSIGNING WASTE.

Does NUREG/CR-2850 V04: POPULATON DOSE COMMITMENT DUE TO THE AUBURN STEEL COMPANY RADCACTIVE CON-ES mW NM NR MM MS N E C 15 VO EV'ALUATON OF NUCLEAN FACluTY DECOM-NUREG/C'R-3572 V02: DETERMINATION OF METABOUC DATA AP-MISSONING PRCMECTS. Summary Status Report Three Mile island NU E S01 N HWA ANAL FOR NU G/

5 E

T UCLEAR Cl Y DECOM-THE ONSITE DISPOSAL OF RADOACTIVE WASTES.The ONSITE/

' W j

U NUR /CR 5V EV LUATION OF NUCLEAR FACluTY DECOM-NUREG/CR-4315 VALUATION OF ' NUCLEAR FACluTC DECOM.

MSSONING PRWECTS. Summary Status Report Three Mile Island MISSONING PRCJECTS. Summary Statss Report Three Mile Island Urvt 2 Radioactive Waste And Laundry Shipments.

Unit 2 Reactor Defuehng & r'-

.;4 NUREG/CR4319: NUCRAC A CODE FOR THE ESTIMATON OF AD-VERSARY-ACTON CONSEQUENCES IN THE NUCLEAR POWER Defueling FUEL CYCLE.

NUREG/CR-4315 V03: EVALUATION OF NUCLEAR FACluTY DECOM-NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE MISSONING DROJECTS. Summary Status Report Three Mde Island REDUCTON RESEARCH PROJECTS.

Una 2 Reactor % & Deasm Degraded Electosner Motorial NUREG/CR-4803: APPRAISING ATMOSPHERIC TRANSPORT AND NUREG-0090 V08 N04: REPORT TO CONGRESS ON ABNORMAL DIFFUSION MODELS FOR EMERGENCY RESPONSE FACIUTIES.

OCCURRENCES. October -December 1985.

Dose Reduction Design Cr#erte NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE NUREG/CR-3957; REUABluTY ASSESSMENT AND PROBABlWTY REDUCTION RESEARCH PROJECTS.

BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS AND SHEAR WALLS. Summary Report.

Dosimetry NUREG/CR-4349: LSL-M2-A COMPUTER PROGRAM FOR LEAST-SO RES LM@E MSWENT & WM MN E CR TORNADO CUMATOLOGY OF THE CONTIGUOUS UNITED STATES.

Draft Environmental Statement Detec#on Urnets NUREG-1168: DRAFT ENVIRONMENTAL STATEMENT FOR DECOM-NUREG 1158: ACCURACY AND DETECTION UMITS FOR BIOASSAY MISSIONING HUMBOLDT BAY POWER PLANT, UNIT 3. Docket No.

MEASUREMENTS IN RADIATION PROTECTON - STATISTICAL 50-1334 Pacific Gas And Electric Company)

CONSIDERATONS.

Dynamic Loading L.

J.

Gee Generadon NUREG/CR-4579 APPUCATON OF THE KEY CURVE AND MULTI-NUREG/CR4801: TECHNICAL CONSIDERATIONS AFFECTING PREP.

SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF ARATION OF CN-EXCHANGE RESINS FOR DISPOSAL ALLOY STEEL Diesel Generator ECCS _t, _ Criterte NUREG/CR 4557: A REVIEW OF ISSUES RELATED TO IMPROVING NUREG/CR4412: AN ASSESSMENT OF SAFETY MARGINS IN ZIRCA.

NUCLEAR POWER PLANT DIESEL GENERATOR REUABluTY.

LOY OXIDATION AND EMBRITTLEMENT CRITERIA FOR ECCS AC-CEPTANCE.

NUREG/CR4803: APPRAISING ATMOSPHERIC TRANSPORT AND EDGP DIFFUSON MODELS FOR EMERGENCY RESPONSE FACILITIES.

NUREG-1209-PROGRAM PLAN FOR ENVIRONMENTAL QUAUFICA-TION OF MECHANICAL AND DYNAMIC (INCLUDtNG SEISMIC)

G 750 101: INDEXES TO NUCLEAR REGULATORY COM-fR DOP' MISSION ISSUANCES. January-March 1988.

E CR45'83 V01[ DEVELOPMENT AND VAUDATION OF A REAL-NUREG/CR-4831: ELECTROMAGNETIC ACOUSTIC TRANSDUCER TIME SAFE 4li SYSTEM FOR THE INSPECTION OF UGHT WATER (EMAT) DEFECT CHARACTERl2ATION OF NUCLEAR REACTOR COMPONENTS. Semi-Annual Report For April 1984-September 1984.

PIPING WELDS. Phase 1 Final Report,0ctober 1985. March 1988.

Derect Radletion Monitoring mM EPICOR-11 NUREG.0837 V05 N04. NRC TLD DIRECT RADIATON MONITORING NUREG/CR4498: FIELD TESTING OF WASTE FORMS CONTAINING NETWORK.

ese Report. October-December 1985.

EPICOR-il CN EXCHANGE RESINS USING LYSIMETERS.

NUREG-0837 N01: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, January-March 1988.

E e

Dieseeemhey BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS NUREG/CR-4315 V03: EVALUATION OF NUCLEAR FACIUTY DECOM-AND SHEAR WALLS. Summary Report.

MISSIONING PROJECTS. Summary Status Report Three Mile island NUREG/CR4580- STONY BROOK SEISMIC NETWORK ON LONG Urut 2 Reactor Defuelmg & Disassembly ISLAND.NEW YORK Final Report (September 1979 March 1985).

NUREG/CR4609 EFFECTS OF EARTHOUAKES ON UNDERGROUND D6ecrete Element FACI'ITIES Uterature Review And Discussion.

NUREG/CR-3441: RADONE.A COMPUTER CODE FOR SIMULATING NUREG/CR4832-NEW MADRID SElGMOTECTONIC PROGRAM Final FAST TRANSIENT ONE-DIMENSONAL HYDRODYNAMIC CONDI-Report.

46 SublGCt lfMiGX Electic-Mas 4c Fractwo Exposure NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-NUREG/CR-4315 V01: EVALUATON OF NUCLEAR FAQUTY DECOM-TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985 MISSONING PROJECTS. Summary Status Report Three Mile telend NUREC/CH-4572-NRC LEAK-BEFORE-BREAK (LBB.NRC) ANALYSIS Unit 2 Reactor Cooient System & Systems Demntaminamon.

METHOD FOR ORCUMFERENTIALLY THROUGH-WALL CRACKED NUREG/CR-4315 V02: EVALUATON OF NUCLEAR FAQUTY DECOM-PtPES UNDER AXIAL PLUS BENDING LOADS. Topical Report.

MISSONING PROJECTS. Summary Status Report Three Mile falend Elec'rical Cable NUR /CR 5

OF NUCLEAR FACIUTY DECOM-NUREG/CR-4548: CORRELATION OF ELECTRICAL REACTOR CABl MISSIONING PROJECTS. Summary Status Report Three Mile soland FAILURE WITH MATERIALS DEGRADATON.

UM 2 Reacia % & f7 4

Electromagnetic Acoustic Transducer Event NUREG/CR-4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER NUREG 1152: MILLSTONE 3 RISK EVALUATON REPT;AN OVERALL (EMAT) DEFECT CHARACTERIZATION OF NUCLEAR REACTOR REVIEW AND EVALUATION OF THE MILLSTONE UNIT 3 PROBASI-PlPING WELDS. Phase i Final Report. October 1985 March 1966.

USTIC SAFETY STUDY.

Electror> Photon Demage NUREG/CR-4543: FLRST RESULTS FROM ELECTION-PHOTON FIRAC NUREG/CR-4561: FIRAC USER'S MANUAL-A COMPUTER CODE TO DAMAGE EQUIVALENCE STUDIES ON A GENERIC ETHYLENE-PRO.

SIMULATE FIRE ACODENTS IN NUCLEAR FAQUTIES.

PYLENE RUB 8ER.

FIRIN Electronic leoletor NUREG/CR-4561: FIRAC USER'S MANUAL A COMPUTER CODE TO NUREG/CR-3453: ELECTRONIC ISOLATERS USED IN SAFETY SYS-SIMULATE FIRE ACODENTS IN NUCLEAR FAQUTIES.

TEMS OF U S. NUCLEAR POWER PLANTS.

FORTRAN Emhrtr.nement NUREG/CR 4490 V01: UGHT WATER-REACTOR SAFETY MATERIALS NUREG/CR-4497: NRCPAGE APPUCATIONS MANUAL ENGINEERING RESEARCH PROGRAMS: Quarterly Progress FORTRAN IV Report. January-March 1985.

NUREG/CR-4503 VOI: LONG TERM EMBF:ITTLEMENT OF CAST-NUREG/CR-3262 V07; COBRA-NC A THERMAL-HYDRAUUC CODE DUPLEX STAINLESS STEELS IN LWR SYS. Annual Report,0ctober FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR COMPONENT. Volume 7: Assessment Manuel for Containment Applica-1984 September 1985.

tions.

Emergency Planning NUREG/CR4603: APPRAISING ATMOSPHERIC TRANSPORT AND Facultation Technique DIFFUSION MODELS FOR EMERGENCY RESPONSE FAOUTIES.

NUREG/CR-3587: IDENTIFICATION AND EVALUATON OF FAQUTA-TION TECHNIQUES FOR DECOMMISSIONING UGHT WATER 1

LEASE OF UF6 FROM A RUPTURED MODEL 48Y CYUNDER AT SEQUOYAH FUELS CORPORATON Fanure FACtUTY. Lessons-Leamed Repat NUREG 0090 V08 N04: REPORT TO CONGRESS ON ABNORMAL OCCURRENCES. October -December 1965.

G CR METHOO TO CHARACTERIZE LOCAL METEOR-OLOGY AT NUCLEAR FACIUTIES FOR APPUCATION TO EMER-NU CR 7: AN OF ETY MPUCATONS CONM AT WE NEE 1 NM N M NUR C

Pf I NG ATMOSPHERIC TRANSPORT AND DIFFUSON MODELS FOR EMERGENCY RESPONSE FAQUTIES.

NURE

-4265 V01: AN ASSESSMENT OF THE SAFETY IMPUCA-TONS OF CONTROL AT THE CALVERT CUFFS-1 NUCLEAR PLANT.

Energy Reorganization Act NUREG/CR-4290 V01: PROBABluTY OF PlPE FAILURE IN THE REAC-NUREG-0980 R02: NUCLEAR REGULATORY LEGISLATON.

TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR PLANTS. Volume 1:WR..eport.

Enforcement Actione NUREG/CR-4330 V02: REvitw OF UGHT WATER REACTOR REGO-NUREG-0940 VOS N01: ENFORCEMENT ACTONS;SIGNIFICANT AC-LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-TIONS RESOLVED.Quarterty Progress Report. January-March 1986.

TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE Enytronmental Ouenfication TO RISK. Reactor Containment Leekage Rates. Main Steam teolation NtJREG 1209: PROGRAM PLAN FOR ENVIRONMENTAL QUAUFICA-Valve Leakage TON OF MECHANICAL AND DYNAMIC (INCLUDING SEISMIC)

NUREG/CR-4374 V03: A REVIEW OF THE OCONEE-3 PROBABluSTiu OUAUFICATON OF MECHANICAL AND ELECTRICAL EQUIPMENT RISK ASSESSMENT CCNTA!NMENT PERFORMANCE, RADIOLOGICAL SOURCE TERMS ANC RISK ESTi-PROGRAM (EDOP).

MATES

~

^'

N EG 2-OCE INGS OF THE SECOND IAEA SPE-E-

T O NCE N YSlS CALISTS MEETING ON SUBCRITICAL CRACK GROWTH. Sessions ill NUREG/CR-4449: A PWR HYBRID COMPUTER MODCL FOR ASSESS-

& IV Held At Sendai, Japan,May 15-17,1965.

ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS, NUREG/CR-4483: REACTOR P9 ESSURE VESSEL FAILURE PROBA.

Bl

/CR ECTRONIC ISOLATERS USED IN SAFETY SYS-gED E S

TEMS OF U.S. NUCLEAR POWER PLANTS-NUREG/CR-4548: CORRELATON OF ELECTRICAL REACTOR CABLE FAILURE WITH MATERIALS DEGRADATION.

' - -- Rutstper NUREG/CR-4548: CORRELATION OF ELECTRICAL REACTOR CABLE Eth' VIE'G'/UU'U FIRST N

RESULTS FROM ELECTION-PHOTON FAILURE WITH MATERIALS DEGRADATION.

DAMAGE EOUIVALENCE STUDIES ON A GENERIC ETHYLENE-PRO-PYLENE RUBBER.

FanSide inepection NUREG/CR-4484: STATUS OF ACTIVITIES FOR INSPECTING WELD Evolustion Model Calculation OVERLAID PIPE JOINTS.

NUREGICR-4549-DETERMINATON OF APPENDIX K CONSERV.

ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR F

USING TRAC-PD2/ MOD 1 CODE.

EG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

Eneminor Stander (e NUREG 1021 R02: OPERATOR UCENSING EXAMINER STANDARDS.

p NUREG-1156: ACCURACY AND DETECTON UMITS FOR BIOASSAY Esothermic Reaction NUREG/CR-4601: TECHNICAL CONSIDERATIONS AFFECTING PREP-MEASUREMENTS IN RADIATION PROTECTON - STATISTICAL ARATION OF ION-EXCHANGE RESINS FOR DISPOSAL CONSIDERATIONS.

Subject index 47 NUREG/CP4067 V01: PROCEEDINGS OF THE SECOND IAEA SPE-Fragility CIAUSTS' MEETING ON SUBCRITICAL CRACK GROWTH. Sessions 1 NUREG/CR-3957: REUABluTY ASSESSMENT AND PROBABIUTY And ll. Held At Sendai, Japan,May 15-17 1985.

BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS NUREG/CP-0067 V02: PROCEEDINGS DF THE SECOND lAEA SPE-AND SHEAR WALLS.Sumrnary Report.

QAUSTS MEETING ON SUBCRITICAL CRACK GROWTH. Sessions til

& IV Held At Sendai, Japan,May 15 17.1985.

Frequency Domain SAFT NUREG/CR 4634: DEVELOPMENT OF A REAL TIME RESIDUE URE /CR-4207: FAULT TREE APPUCATON TO THE STUDY OF R

1 986' SYSTEMS INTERACTIONS AT INDIAN POINT 3.

Froude Number Fleed Study NUREG/CR4498. FIELD TESTING OF WASTE FORMS CONTAINING NUREG/CR-3701: REMIX A COMPUTER PROGRAM FOR TEMPERA-EPICOR-Il lON EXCHANGE RESINS USING LYSIMETERS.

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER INTERRUPTION OF NATURAL CIRCULATION.

Filter Medle NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-NUREG/CR4601: TECHNICAL CONSIDERATONS AFFECTING PREP-SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL ARATION OF ON-EXCHANGE RESINS FOR DISPOSAL SHOCK STUDY.

Final Report Fuel Cycle NUREG/CR4632: NEW MADRID SEISMOTECTONIC PROGRAM Final NUREG/CR-4519: TECHNOLOGY, SAFETY AND COSTS OF DECOM.

Report MISSIONING NUCLEAR FUEL CYCLE FACluTIES CLASSIFICATON DmWSSNG WASTE.

Fire Accident NUREG/CR4581: FIRAC USER'S MANUALA COMPUTER CODE TO FuelRed SIMULATE FIRE ACCIDENTG IN NUCLEAR FACluTIES.

NUREG/CR-4516: INTERNATONAL SAFEGUARDS AT FACluTIES EM-Flesion Product PLOYING SPENT FUEL ROD CONSOUDATON.

NUREG/CR 4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT V RTICAL ISSION PRODUCT RELEASE APPARATUS IN HOT URE 4507: HECTR VERSION 1.5 USER'S MANUAL NUREG/CR-4338: TELLURIUM BEHAVOR IN CONTAihMENT UNDER UGHT WATER REACTOR ACCIDENT CONDITONS.

Gastrointestinal _ J.._

NUREG/CR 4453 V02: LIGHT-WATER-REACTCH FUEL S/ 'ETY SYS-NUREG/CR-3572 V02 DETERMINATION OF METABOUC DATA AP-TEMS RESEARCH PROGRAMS. Quarterty Progress Report.Aoni-June PROPRIATE FOR HLW DOSIMETRY.lLGastrointestinal Absorpton.

1985.

Fleelon Product Releene NUREG/CR-4568: A HANDBOOK FOR QUICK ESTIMATES.A Methud NUREG-0900 RO1: NUCLEAR POWER PLANT SEVERE ACCIDENT RE-For Developing Quick Approxirnate Estimates Of Costs For Genene Ac-SEARCH PLAN.

tons For Nuclear Power Plants.

Flow Terminetton Generic Safety leeue ti NUREG/CR4621: FLOW VISUALIZATON EXPERIMENT ON HOT-LEG NUREG/CR-4594: ESTIMATED SAFETY SIGNIFICANCE OF GENERIC U-BEND TWO-PHASE NATURAL CIRCULATON PHENOMENA.

SAFETY ISSUE 61.

Fluid Jet Mixin0 I*

I'*"**

NUREG/CR 3701: REMIX-A COMPUTER PROGRAM FOR TEMPERA TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER NUREG/CR-4905: A SCOPING STUDY OF THE POTENTIAL EFFEC-INTERRUPTION OF NATURAL CIRCULATON TIVENESS 03 A!4 OPERATONAL SAFETY REUABluTY PROGRAM NUREG/CR-3702 BUOVANCY EFFECTS IN OVERCOOUNG TRAN.

IN ADDRESS NG GENERIC SAFETY PROBLEMS.

SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL SHOCK STUDY' NUREG/CR-4520 PREDICTIVE GEOCHEMICAL MODEUNG OF CON.

Fluid to Fluid Modeling TAINMENT CONCENTRATONS IN LABORATORY COLUMNS AND IN NUREG/CR-4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL.

PLUMES MIGRATING FROM URANIUM MILL TAluNiS WASTE IM-ING LAWS FOR TWO' PHASE FLOW LOOP.

POUNDMENTS. Final Report.

Foreign Policy Geochemical Conditione NUREG-1160: INTERNATIONAL COOPERATION DURING RADIOLOGI-NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADIONU.

CAL EMERGENCIES. NRC Program Guidance For The Provison Of CLIDE GEOCHEMISTRY INFORM ATION DEVELOPED BY DOE HIGH-Technical Advice To Foreign Counterpart Organizat ons.

LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For Fractography NUREG/CP-0067 V01: PROCEEDINGS OF THE SECOND IAEA SPE-Goosogne weds.

CIAUSTS' MEETING ON SUBCRITICAL CRACK GROWTH.Sessons 1 NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ll.The Sandia as sotaten And hanW M N Rectured h b NUR CP 7

R I HE SECOND lAEA SPE-CIAUSTS MEETING ON SUBCRITICAL CRACK GROWTH.Sessons til

& IV Held At Sendal, Japan,May 15-17,1985.

Granite Fracture Mechanico NUREG/CR-4642: ROCK MASS SEALING EXPERIMENTAL ASSESS-NUREG/CP 0077: PROCEEDINGS OF THE SEMINAR ON LEAK-MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June BEFORE-BREAK: INTERNATONAL POUCIES AND SUPPORTING 1984 - May 1985.

RESEARCH NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC.

Graphite H440 TOR PRESSURE BOUNDARY COMPONENTS Annual Report for 1985.

NUREG/CR-3965: AN INVESTIGATON OF THE STRENGTH OF H440 NUREG/CR-3770: PREUMINARY DEVELOPMENT OF AN INTEGRAT.

GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND SEC-ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL ONDARY STRESS.

SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER PLANT W/TWO OVERSIZE DRAWINGS.

Graphne Ringe Stroes NUREG/CR-4539: INVESTIGATON OF TEARING INSTABluTY PHE.

NUREG/CR-3965: AN INVESTIGATION OF THE STRENGTH OF H440 NOMENA IN ASTM A106 STEEL GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND SEC-ONDARY STRESS.

Fracture Toughnese NUREG/CR-4538 VO1: FRACTURE ANALYSIS OF WELDED TYPE 304 Ground Motion STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION AND NUREG/CR-4609: EFFECTS OF EARTHOUAKES ON UNDERGROUND LIMIT LOAD ANALYSIS.

FACluTIES Uterature Review And Discusson.

48 Sut> ject hulex Groundwater Human Action NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND NUREG/CR-4589: REVIEW OF SELECTED AREAS OF YANKEE ROWE SORPTION.The Solutulity Of Tc0V) Oxides.

PROBABILLSTIC SAFETY STUDY.

NUREG/CR-4588 V03. SOIL-STRWTURE INTERACTON.Vol 3:Influ-ence Of Ground Water.

Human Error NUREG-1192: AN INVESTIGATION OF THE CONTalBUTORS TO G-1160: INTERNATIONAL COOPERATION DURING RADIOLOGI-CAL EMERGENCIES. NRC Program Guidance For The Provtsson Of Human Factore Tectncal Advice To Foreign Counterpart Orgarnations.

NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-HECTR SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-4507: HECTR VERSION 1.5 USER S MANUAL NURE M

A N

CIATOR/ ALARM Hapahn ar SYSTEMS;ANNUNCIATCA EXPERIMENT PLAN L NUREG/CR-4568: A HANDBOOK FOR QUICK ESTIMATES:A Method Quick Appro to Estimates Of Costs For Genenc Ac-EG V

NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN.

Hometh Effects NUREG/CR 3572 V02 DETERMINATON OF METABOUC DATA AP-Hybrid Computer Model PROPRIATE FOR HLW DOSIMETRY.ll. Gastrointestinal Absorpbon.

NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL Health Pnyeice REPORT.

NUREG 1156; ACCURACY AND DETECTON UMITS FOR BIOASSAY NUREG/CR-4449: A PWR HYBRID COMPUTER PAODEL FOR ASSESS-MEASUREMENTS IN RADIATION PROTECTON - STATISTICAL ING THE SAFETY IMPUCATONS OF CONTROL SYSTEMS.

CONSIDERATONS.

Hydrolyste Heat Trenefer NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND NUREG/CR-3705: IMPROVED MODEUNG AND NUMERICS TO SOLVE SORPTION.The Solubsty Of Tc(IV) Oxides.

TWO-DIMENSIONAL ELUPTIC FLUID FLOW AND HEAT TRANSFER PROBLEMS.

lAEA NUREG/CR-4027:

TRAC #F1/ MODI INDEPENDENT

. NUREG-1160: INTERNATONAL COOPERATON DURING RADIOLOGI-ASSESSMENT. Condensation in Stratified Cocurrent Flow.

CAL EMERGENCIES. NRC Program Guidance For The Provision Of NUREG/CR-4581: DRYOUT FRONT MODEUNG FOR PWR THERMAL Tectncal Advice To Foreign Counterpart Orgarnations.

HYDRAUUC ANALYSIS.

IE Bulletin 30-01 High Enttched Uranium NUREG/CR-3064 Vot: COMPUTATIONAL METHODOLOGY FOR OAK NUREG/CR-3960; CLOSEOUT OF IE BULLETIN 80-01. Operability Of RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC.

Automatic Depressunzation System (ADS) Valve Pneumatic Supply.

TOR (BSR) Cross-Section Generation And Vahdation Volume 1.

lE Bulletin 80-20 High hvel Wash NUREG/CR-3962: CLOSEOUT OF IE BULLETIN 80-20. Failures Of Wes-NUREG/CR-3572 V02: DETERMINATION OF METABOUC DATA AP-tingtmse Type W-2 Spnng Return To Neutral Control Sutches.

PROPRIATE FOR HLW DOSIMETRY.ll. Gastrointestinal Absorption-IOSCC High hvel Wash phpoeNory NUREG/CR-4602 UNIQUENESS OF BOluNG WATER REACTOR PRI-NUREG/CR-4236 V03: PROGRESS IN EVALUATON OF RADIONU.

MARY WATER CHEMISTRY. Final Report, October 1985 - March 1986.

CUDE GEOCHEMISTRY INFORMATON DEVELOPED BY DOE HIGH-lcond-400 LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For M85.

NUREG/CR-4241: CHEMICAL ASPECTS OF CESIUM ODIDE INTERAC-TON IN STEAM WITH 304 STAINLESS STEEL AND INCONEL-600.

High Pressure injection

, _ _ _- ------- A_

NUREG/CR 3701: REMIX:A COMPUTER PROGRAM FOR TEMPERA-TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER NUREG/CR-4620: METHODOLOGIES FOR EVALUATING LONG-TERM INTERRUPTICN OF NATURAL CIRCULATON.

STABluZATION DESIGNS OF URANIUM MILL TAluNGS IMPOUND-MENTS.

High Temperature Graphite Reactor NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON.

Ir@rogrees MonMoring SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR 4497: NRCPAC' APPLICATONS MANUAL RESEARCH.Ouarterty Progress Report July-September 1985.

Incident investigation Team High Density Polyethylene NUREG 1201: REPORT OF THE INDEPENDENT AD HOC GROUP FOR NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR.

THE DAVIS-BESSE INCIDENT.

RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY N E CR-4619-STRESS CORROSION CRACKING TESTS ON HIGH-High Level Warte LEVEL WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-NUREG/CR-3472 V02: SURFACE PROPERTIES AND PERFORMANCE POSITORY ENVIRONMENTS.

PREDICTION OF ALTERNATIVE WASTE FORMS Final Report.

NUREG/CR-4379 V04: LONG-TERM PERFORMANCE OF MATERIALS Index USED FOR HIGH-LEVEL WAf'E PACKAGING. Annual Report, Year NUREG-0304 VII

)RY AND TECHNICAL Four - April 1985 - March if a6 REPORTS.Compdatir 486. January-March.

NUREG/CR4619: STRESS CORROSION CRACKING TESTS ON HIGH-NUREG4750 V2310) duCLEAR REGULATORY COM-LEVELWASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-MISSION ISSUANC

.h 1986.

POSITORY ENVtRONMENTS.

Induction Furnace High Temperature Gae-Cooted Reactor NUREG/CR-4332: DESaGN

._..e FINAL SAFETY ANALYSIS REPORT NUREG/CR4402 V03: HIGH TEMPERATURE GAS-COOLED REACTOR FOR VERTICAL FISSION PRODUCT RELEASE APPARATUS IN HOT SAFETY STUDIES FOR THE DIVISION OF ACCIDENT CELL B. BUILDING 4501.

EVALUATION Ouarterty Progress Raport. July 1 - September 30.1985.

NUREGICR-4526; FINITE ELEMENT ANALYSIS OF THE 2240 MW initiating Event HTGR PCRV.

NUREG/CR-4589: REVIEW OF SELECTED AREAS OF YANKEE ROWE PROBABluSTIC SAFETY STUDY.

NUREG/CR-4621: FLOW VISUAUZATON EXPERIMENT ON HOT-LEG Innovation Sequence U-BEND TWO' PHASE NATURAL CIRCULATON PHENOMENA.

NUREG/CR-4497: NRCPAGE APPLICATIONS MANUAL l

1

Subject index 49 Inoperable Pilot Valve NUREG/CP-0067 V02: PROCEEDINGS OF THE SECOND IAEA SPE-NUREG-0090 VOS N04: REPORT TO CONGRESS ON ABNORMAL CIAUSTS MEETING ON SURCRITICAL CRACK GROWTH.Sessons til OCCURFsENCES. October -December 1965.

& IV Held At Sendai, Japan,May 15-17,1985.

Oneervice inspection Key Curve NUREG/CR-4484: STATUS OF ACTMTIES FOR INSPECTING. WELD NUREG/CR-4579-APPUCATION OF THE KEY CURVE AND MULTI-OVERLAID PIPE JOINTS.

SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF NUREG/CR-4583 V01: DEVELOPMENT AND VAUDATION OF A REAL*

ALLOY STEEL TIME SAFE-UT SYSTEM FOR THE INSPECTION OF UGHT WATER COMPONENTS. Semi-Annual Report For April 1984-September 1984.

Keywords NUREG/CR-3262 V02: COBRA-NCA THERMAL-HYDRAUUC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUREG-1198: RELEASE OF UF6 FROM A RUPTURED MODEL 48Y COMPONENTS. Volume 2: COBRA-NC Numencal Solubon Methods.

CYUNDER AT SEQUOYAH FUELS CORPORATON FACIUTY. Lessons-Learned Report.

LAFM

- r Conomion NUREG/CR-4595: ENHANCEMENT TO THE LAFM COMPUTER CODE.

NUREG/CR-4619 STRESS CORRCSION CRACKING TESTS ON HIGH-LER LEVEL-WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE' NUREG/CR-2000 V05 N3: LICENSEE EVENT REPORT (LER)

POSITORY ENVIRONMENTS.

COMPILAMFor M Of March 1986.

Intergranuier Stroes Conoelon Cracking NUREG/CR-2000 V05 N4: UCENSEE EVENT REPORT (LER)

NU G/CR4484 US OF ACTIVITIES FOR INSPECTING WELD NU V05 N5 LICEN E EVENT REPORT

(* ER)

NUREG/CR-4602 UNIQUENESS OF BOluNG WATER REACTOR PRI-COMPtLATION:For Month Of May 1986.

MARY WATER CHEMISTRY. Final Report, October 1985 - March 1996.

LSbM2 Internal Event NUREG/CR4349: LSL-M2-A COMPUTER PROGRAM FOR LEAST.

NUREG-1152-MILLSTONE 3 RISK EVALUATION REPTAN OVERALL SOUARES LOGARITHMIC ADJUSTMENT OF NEUTRON SPECTRA.

REVIEW AND EVALUATION OF THE MILLSTONE UNIT 3 PROBABI-USTIC SAFETY STUDY.

Labor NUREG/CR-4142: A REVIEW OF THE MILLSTONE 3 PROBABlUSTIC NUREG/CR-4627: GENERIC COST ESTIMATES Abstracts From Genenc SAFETY STUDY.

Studies For Use in Prepanng Regulatory trnpact Analyses.

Internettonal Cooperation Large Break LOCA NUREG 1160- INTERNATIONAL COOPERATION DURING RADCLOGl.

NUREG/CR-4549-DETERMINATION OF APPENDIX K CONSERV-CAL EMERGENCIES. NRC Program Guidance For The Frovision Of ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR Techncal Advice To Foreign Counterpart Orgarnations.

USING TRAC.PD2/ MOD 1 CODE.

International Setoguardo Lavoring NUREG/CR-4516-INTERNATIONAL SAFEGUARDS AT FACIUTIES EM-NUREG/CR-4588 V01: SOIL-STRUCTURE INTERACTION.hi 1:Influ-PLOYING SPENT FUEL ROD CONSOUDATON.

ence Of Layenno-lodine Procursor Leachability NUREG/CR4338: TELLURIUM BEHAVIOR IN CONTAINMENT UNDER NUREG/CR-3472 V02-SURFACE PROPERTIES AND PERFORMANCE UGHT WATER REACTOR ACCIDENT CONDITONS.

PREDICTION OF ALTERNATIVE WASTE FORMSFinal Report lor > Exchange Reelne Leek Monitoring NUREG/CR4601: TECHNICAL CONSIDERATIONS AFFECTING PREP-NUREG/CR-4124 V02: NDE OF STAINLESS STEEL AND ON-UNE ARATION OF ION-EXCHANGE RESINS FOR DISPOSAL LEAK MONITORING OF LWRS. Annual Report October 1984 - Sep-tomber 1985.

Inedlated Zircatoy Cladding NUREG/CR-4453 V02: UGHT-WATER-REACTOR FUEL SAFETY SYS-Leek Rate TEMS RESEARCH PROGRAMS. Quarterty Progress Reph NUREG/CR-4572: NRC LEAK-BEFORE BREAK (LBB.NRC) ANALYSIS 1985.

METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED trradief'Jn Effect PIPES UNDER AXIAL PLU3 BENDING LOADS. Topical Report.

NURaiG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-t,,g, RADIA ON THE MECHANICAL PROPERTIES OF HIGH DENSITY NUREG/CP-0077: PROCEEDINGS OF THE SEMINAR ON LEAK-BEFORE-BREAK: INTERNATIONAL POUCIES AND SUPPORTING J-Integr g RESEARCH.

NUPiG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-NUREG/CR4572: NRC LEAK-BEFORE-BREAK (LEB.NRC) ANALYSIS T )R PRESSURE BOUNDARY COMPONENTS. Annual R for 1985.

METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WE D TYPE 304 PIPES UNDER AXtAL PLUS BENDING LOADS. Topical Report.

STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION AND UMIT LOAD ANALYSIS.

Leakage NUREG/CR4539: INVESTIGATION OF TEARING INSTABluTY PHE-NUREG/CR 4330 V02: REVIEW OF LIGHT WATER REACTOR REGU-NOMENA IN ASTM A106 STEEL LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE J-R Curve TO RISK. Reactor Conta% ment Leakage Rates. Main Steam Isolation NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 Valve Leakage.

STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION AND UMIT LOAD ANALYSIS.

Legalleeuences NUREG/CR4572: NRC LEAK-BEFORE-BREAK (LBB NRC) ANALYSIS NUREG4304 V11 N01: REGULATORY AND TECHNICAL METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED REPORTS. Compilation For First Quarter 1986. January-March.

PIPES UNDER AXIAL PLUS BENDING LOADS. Topical Report.

NUREG-0750 V23 NO3: NUCLEAR REGULATORY COMMISSION IS-NUREG/CR-4579: APPLICATON OF THE KEY CURVE AND MULTI-SUANCES FOR MARCH 1986. Pages 113 232.

SPECIMEN TECHN1 QUES TO DYNAMIC J-R CURVE TESTING OF NUREG4750 V23 N04: NUCLEAR REGULATORY COMMISSION IS-ALLOY STEEL SUANCES FOR APRIL 1986. Pages 233-4G4.

Japan Lessone Learned NUREG/CP4067 V01: PROCEEDINGS OF THE SECOND IAEA SPE.

NUREG 1198: RELEASE OF UF6 FROM A RUPTURED MODEL 48Y CIALISTS' MEETING ON SUBCRITICAL CRACK GROWTH Sessions I CYUNDER AT SEQUOYAH FUELS CORPORATON And ll. Held At Sendat, Japan.May 15 17,1985.

FACILnY. Lessons-Learned Report

50 Subject Irulex Ucensed Opereung Reactor NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR.

NUREG-0020 V10 N04: UCENSED OPFRATING REACTORS STATUS RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY

SUMMARY

REPORT. Data As Of March 31,1986.(Gray Book I)

POLYETHYLENE.

Ucer.eed Opereung Reactore Lysimeter NUREG4020 V10 NO3: UCENSED OPERATING REACTORS STATUS NUREG/CR-4498: FtELD TESTING OF WASTE FORMS CONTAINING

SUMMARY

REPORT.Deta As Of February 28,1986.(Grey Book 1)

EPICOR-il ION EXCHANGE RESINS USING LYSIMETERS.

Ucensee Conerector And Vender inspe-tion MELCOR NUREG4040 V10 N01: UCENSEE CONTRACTOR AND VENDOR IN-NUREG/CR4467; RELATIVE IMPORTANCE OF INDIVIDUAL ELE-SPECTION STATUS REPORT. Quarterty Report, January 1986 - March MENTS TO REACTOR ACCIDENT CONSEQUENCES ASSUMING 19861 White Book)

EQUAL RELEASE FRACTIONS.

Ucensee Event Report MNGET NUREG/CR-2000 VOS N3: LICENSEE EVENT REPORT (LER)

NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY NU V05 N4 NSE EVENT REPORT (LER)

COMP 1LATION.For Month Of Apre 1986.

RESEARCH.Quarterty Progress Report July-September 1985.

NUREG/CR-2000 V05 N5: UCENSEE EVENT REPORT (LER)

LATimh MonW May 1906-NURE3-0090 V08 N04: REPORT TO CONGRESS ON ABNORMAL Ucensing OCCURRENCES. October -December 1985.

NUREG-1198: RtELEASE OF UF6 FROM A RUPTURED MODEL 48f fA b arned G4090 N04 REPORT TO CONGRESS ON ABNORMAL NUREG/CR4549-DETERMINA ION OF APPENDIX K CONSERV-OCCURRENCES. October December 1985.

ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWH G

V0.t UC.HT-WATER-REACTOR FUEL SAFETY SYS-Utt-Off TEMS RESEARCH PROGRAMS. Quarter 1y Frogress Report, April-June NUREG/CR-4588 V02: SOIL STRUCTURE INTERACTION.Vol 2:Influ-1985.

ence Of Urt-Off.

Meterial Bolence sequencer umit States NUREG/CR-4497: NRCPAGE APPUCATIONS MANUAL NUREG/CR-3957: REUAD!UTY ASSEOSMENT AND PROBABlUTY BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS Mothematicet Model AND SHEAR WALLS. Summary Report.

NUREG/CR-3705: IMPROVED MODEUNG AND NUMERICS TO SOLVE TVrO-DIMENSIONAL ELUPTIC FLUID FLOW AND HEAT TRANSFER Uguid4Betal Fast Breeder Reactor PROBLEMS.

NUREG/CR4595: ENHANCEMENT TO THE LAFM COMPUTER CODE.

Mechanical Propertlee Load Combinatione NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-NUREG/CR-2331 VOS N3: SAFETY RESEARCH PROGRAMS SPON-RADIATION ON THE MECHANICAL PROPERTIES OF HIGH DENSITY SORED BY OFFICE OF NUCLEAR REGULATORY POLYETHYLENE.

RESEARCH.Ouarterty Progress Report. July-September 1985.

Melt Proereseson RE CR-3957: REUABILITY ASSESSMENT AND PROBABluTY S R

^

BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS AND SHEAR WALLS. Summary Report MetaboHom NUREG/CR-3572 V02 DETERMINATION OF METABOUC DATA AP-W Reno PROPRIATE FOR HLW DOSIMETRY.ll. Gastrointestinal Abeorptort NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC.

TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

MeteNurgical Examinadon L,., m NUREG-1179 V02 RUPTURE OF MODEL 48Y UF6 CYUNDER AND RE-NUREG/CR-3970- TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT:

LEASE OF URANIUM HEXAFLUORIDE.Cyhnder Overfill, March 12-LOBI INTERMEDIATE BREAK TEST B-R1M.

13,1986. Investigation Of A Failed UF6 Shipping Cylinder Loos of Foodwater Event M#'880eV NUREG-1201: REPORT OF THE INDEPENDENT AD HOC GROUP FOR NUREG/CR-3882: A METHOD TO CHARACTERIZE LOCAL METEOR-THE DAVIS-BESSE INCIDENT.

OLOGY AT NUCLEAR FACIUTIES FOR APPUCATION TO EMER-GENCY RESPONSE NEEDS.

Lose-Of-Cooient A~** ant NUREG/CR-4315 V01: EVALUATION OF NUCLEAR FACIUTY DECOM.

Microecopic Fracture MISSIONING PROJECTS. Summary Status Report Three Mde Island NUREG/CR4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE Unit 2 Reactor Coolant System & Systems Decontamination.

WALL DRILUNG DAMAGE IN BASALTIC ROCKS.

NUREG/CR4315 V02-EVALUATIOff OF NUCLEAR FACIUTY DECOM-MISSIONING PROJECTS. Summary Statue Report Three Mde Island Microecepy Unit 2 Reactor Building Decontaminabort NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-NUREG/CR4315 VO9-EVALUATION OF NUCLEAR FAC3UTY DECOM-RADIATION ON THE MECHANICAL PROPERTIFS OF HIGH DENSITY MISSIONING PROJECTS. Summary Status Pepoit Three Mile Island POLYETHYLENE.

Unit 2 Radioec:lve Waste And Laundry Shipments.

Low Enriched Urenium NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND NUREG/CR-3064 VOI: COMPUTATIONAL METHODOLOGY FOR OAK SORPTION.The Solubdity Of Tc(IV) Oxides.

RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-TOR (BSR) Cross-Section Generation And ValidaDon, Volume 1.

Minoretogical Cherectertteson NUREG/CR4236 V03: PROGRESS IN EVALUATION OF RADIONU-Low-Level Weste CUDE GEOCHEMISTRY INFORMATMN DEVELOPED BY DOE HIGH-i NUREG/CR-2675 V05: RELEVANCE OF BIOTIC PATHWAYS TO THE LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For LONG-TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esti-Apni, June 1985.

mation Of Radiabon Dose To Man Resutbng From Betic TransportThe BIOPORT/ MAX 11 Software Pack Modeling NUREG/CR4001: TECHNICAL CCRAT!ONS AFFECTING PREP-NUREG/CR-4520: PREDICTIVE GEOCHEMICAL MODELING OF CON-ARATION OF ION-EXCHANGE RESINS FOR DISPOSAL TAINMENT CONCErlTRATIONS IN LA3 ORATORY COLUMNS AND IN I

l

Subject index 51 PLUMES MIGRATING FROM URANIUM MILL TAluNGS WASTE IM-NUREG/CR-4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER POUNDMENTS. Fir:al R (EMAT) DEFECT CHARACTERIZATION OF NUCLEAR REACTOR NUREG/CH-4615 V01:

UNG STUDY OF SOLUTE TRANSPORT PIPING WELDS. Phase i Final Report,0ctober 1965 - March 1986.

IN THE UNSATURATED ZONE Infonastion And Data Sets.

Nonalostrucuve Examinetton 6 Technique NUREG/CR4583 V01: DEVELOPMENT AND VAUDATION OF A REAL-NUREG/CR-4579: APPLICATION OF THE KEY CURVE AND MULTI-TIME SAFE-UT SYSTEM FOR THE INSPECTION OF UGHT WATER SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF COMPONENTS.Serru-Annual Report For April 1964-September 1984.

ALLOY STEEL Nonequlubrium Heat Transfer

-4048 HODOLOGY FOR ALLOCATING REUABluTY NUREG/CR-4353: ASSESSMENT OF POST-CRITICAL-HEAT FLUX AND RISK.

MODELS WITH LEHIGH NONEOUIUBRIUM DATA.

Muluple FaHure Nonsafety-Gresle Control System Failure NUREG/CR-4207: FAULT TREE APPUCATION TO THE STUDY OF NUREG/CR 4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS SYSTEMS INTERACTIONS AT INDIAN POINT 3.

OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL REPORT.

NDE NUREG/CR4124 V02 NDE OF STAINLESS STEEL AND ON-UNE Mucieer Meterial Accounting LEAK MONITORING OF LWRS. Annual Report October 1964 - Sep-NUREG/CR-4516: INTERNATIONAL SAFEGUARDS AT FACIUTIES EM-tomber 1965.

PLOYING SPENT FUEL ROD CONSOUDATION.

NEWMIX Nuclear Regulatory Legleistion NUREG/CR-3701: REMIX A COMPUTER PROGRAM FOR TEMPERA.

NUREG 0980 R02 NUCLEAR REGULATORY LEGISLATION.

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTION AFTER INTERRUPTION OF NATURAL CIRCULATION.

Numerical Smoothing NUREG/CR-4579: APPUCATION OF THE KEY CURVE AND MULTI-NRC AuthoriseWon Anti Approprisuone Act SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF NUREG-0980 R02 NUCLEAR REGULATORY LEGISLATION.

ALLOY STEEL NRCpAGE NUREG/CR-4497: NRCPAGE APPUCATIONS MANUAL

/CR COBRA-NC A THERMAL-HYDRAUUC CODE NUCRAC FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUREG/CR-4319: NUCRAC - A CODE FOR THE ESTIMATION OF AD-COMPONENTS. Volume 2: COBRA-NC Numencal Solution Methods.

VERSARY-ACTION CONSEQUENCES IN THE NUCLEAR POWER FUEL CYCLE.

OCA P NUREG/CR-3770: PREUMINARY DEVELOPMENT OF AN INTEGRAT-Natural Circulation ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL NUREG/CR-3701: REMIX.A COMPUTER PROGRAM FOR TEMPERA-SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER TUAE TRANSENTS DUE TO H!GH PRESSURE INJECTION AFTER PLANT.W/TWO OVERSIZE DRAWINGS.

INTERRUPTION OF NATURAL CIRCULATION.

NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-ONSITE/MAXl1 SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL -

NUREG/CR-3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOH SHOCK STUDY.

tnt: ONSliE DISPOSAL OF HADICACTIVE WASTES.The ONSITE/

NUREG/CR-4564: REDUCED PRESSURE AND FLUID TO FLUID SCAL

MAXII Computer Program.

ING LAWS FOR TWO. PHASE FLOW LOOP.

NUREG/CR-4621: FLOW VISUAUZATION EXPERIMENT ON HOT-LEG ORIGEN U-BEND TWO-PHASE NATURAL CIRCULATION PHENOMENA.

NUREG/CR-4467: RELATIVE IMPORTANCE OF INDIVIDUAL ELE-Natural Circulomon Loop MENTS TO REACTOR ACCIDENT CONSEQUENCES ASSUMING NUREG/CR-4564: REDUCED PRESSURE AND FLUID TO FLUID SCAL-EQUAL RELEASE FRACTIONS.

ING LAWS FOR TWO-PHASE FLOW LOOo.

Occupational Rassiamon Esposure Neutral Control Switch NUREG/CR-4627: GENERIC COST ESTIMATES. Abstracts From Generic NUREG/CR-3962: CLOSEOUT OF IE BULLETIN 80-20. Failures Of Wes-Studies For Use in Preparing Regulatory impact Analyses tmghouse Type W-2 Spring Retum To Neutral Control Switches.

Neutron Spectra NUREG/CR-3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR NUREG/CR-4349: LSL-M2:A COMPUTER PPOGRAM FOR LEAST-THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

SQUARES LOGARITHMIC ADJUSTMENT OF NEUTRON SPECTRA.

MAXII Computer Program.

Neutronico Ubrary Operemonal Safety NUREG/CR-3064 V01: COMPUTATIONAL METHODOLOGY FOR OAK NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC-RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-TIVENESS OF AN OPERATIONAL SAFETY RELIABluTY PROGRAM TOR (BSR): Cross 4ection Generation And Validation, Volume 1.

IN ADDRESSING GENERIC SAFETY PROBLEMS.

NUREG/CR-4506: AN OPERATIONAL SAFETY REUABluTY PROGRAM EG V03 OG EVALUATION OF RADONU-APPROACH WITH RECOMMENDATIONS FOR FURTHER DEVELOP-CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-PIENT AND EVALUATON.

LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For Operator Ucensing Examiner 1985.

NUREG-1021 R02: OPERATOR UCENSING EXAMINER STANDARDS.

New Madrid NUR CR-4632-NEW MADRID SEISMOTECTONIC PRCGRAM. Final G/ -4338: TELLUR!UM BEHAVIOR IN CONTAINMENT UNDER UGHT WATER REACTOR ACCIDENT CONDITIONS.

Nonconstenalbie Gas NUREG/CR-4486: VENTING OF NONCONDENSIBLE GAS FROM THE Overcoonng UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U.

NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-BEND VALVES.

SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL SHOCK STUDY.

Nonstestructive Evaluetton NUREG/CR-3770 PRELIMINARY DEVELOPMENT OF AN INTEGRAT-NUREG/CR-4124 V02: NDE OF STAINLESS STEEL AND ON-UNE ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL LEAK MONITORING OF LWRS. Annual Report October 1984 Sep-SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER tomber 1965.

PLANT.W/TWO OVERSIZE DRAWINGS.

1

52 Subject Index NUREG/CR-4265 V01: AN ASSESSMENT OF THE SAFETY IMPUCA-NUREG/CR 4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER TIONS OF CONTROL AT THE CALVERT CUFFS-1 NUCLEAR PLANT.

(EMAT) DEFECT CHARACTERIZATION OF NUCLEAR REACTOR PIPING WMLS Phase i Final Report,0ctober 1985 - March 1968.

NUREG-1179 V02 RUPTURE OF MODEL 48Y UF6 CYUNDER AND RE-Pipe Crack LEASE OF URANIUM HEXAFLUORIDE.Cylmder Overfill, March 12-NUREG-0090 V36 N04: REPORT TO CONGRESS ON ABNORMAL 13,1986. Investigation Of A Feifed UF6 Shippog CWoder OCCURRENCES. October -December 1985.

Overlyens Pool P"*""*** 8"PP'I NUREG/CR4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-NUREG/CR-3060 CLOSEOUT OF IE BULLETIN 80-01. Operability Of TION OF QUENCHING OF SUPERHEATED DEBRIS BEDS UNDER Automatic Depressurization System (ADS) Valve Pneumatic Supply.

TOP-REFLOOD CONDillONSFmal Report Overpeck Population Dose Commitment NUhEG/CR-4379 V04: LONG-TERM PERFORMAflCE OF MATERIALS NUREG/CR-2850 V04: POPULATION DOSE COMMITMENT DUE TO USED FOR H:GH-!EVEL WASTE PACKAGING. Annual Repwt, Year RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES Four - April 1985 - March 1986.

LN 1982.

PRA Post-Critical Heat Flux NUREG 1152: MILLSTONE 3 RISK EVALUATION REPT:AN OVERALL NUREG/CR-4353: ASSESSMENT OF POST-CRITICAL-HEAT FLUX REVIEW AND EVALUATION OF THE MILLSTONE UNIT 3 PROBAB1-MODELS WITH LEHIGH NONEOUluBRIUM DATA.

USTIC SAFETY STUDY.

i NUREG/CR-3770- PRELIMINARY DEVELOPMENT OF AN INTEGRAT-Prm And Procedu.ee Digest ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL NUREG-0386 D04 Rot: UNITED STATES NUCLEAR REGULATORY SHOCK AE APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST JULY PLANT.W/TWO OVERSIZE DRAWINGS.

1,1672 SEPTEMBER 30,1985.

NUREG/CR-4142-A REVIEW OF THE MILLSTONE 3 PROBA01USTIC SAFETY STUDY.

PTS NUREG/CR-3150- SEISMICITY AND TECTONIC RELATIONSHIPS FOR NUREG/CR-U70 PRELIMNARY DEVELUPMENT OF AN INTEGRAT-UPPER CREAT LAKES PRECAMBRIAN SHIELD PROVINCE FINAL ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL REPORT - July 1981 -December 1982.

SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER PLANT.W/TWO OVERSIZE DRAWINGS.

Pressure Boundary NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-i l

Pocked Bede TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

NUREG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-TION OF QUENCHING OF SUPERHEATED DEBRIS BEDS UNDER Proeeure Vessel TOP-REFLOOD CONDITIONSFmal Report-NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-TOR PRESSURE SOUNDARY COMPONENTS. Annual Report for 1985.

N NUREG/CR-3770; PREUMINARY DEVELOPMENT OF AN INTEGRAT-NUREG/CR-4543: FIRST RESULTS FROM ELECTV)N-PHOTON ED APPPOACH TO THE EVALUATION OF PRESSURIZED THERMAL DAMAGE E LENCE STUDIES ON A GENERIC ETHYLENE-PRO-SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER PLANT.W/TWO O'ERSIZE DRAWINGS.

Poocolo Arch NUREG/CR-4463: REACTOR PRESSURE VESSEL FAILURE PROBA-NUREG/CR-4632: NEW MADRID SEISMOTECTONIC PROGRAM Final BIUTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-IZED THERMAL SHOCK EVENTS.

Report.

NUREG/CR-4489-HISTORICAL

SUMMARY

OF THE HEAVY-SECTION l

Permesmew Testing STEEL TECHNOLOGY PROGRAM AND SOME RELATED ACTIVITIES i

l NUREG/CR-4642: ROCK MASS SEAUNG - EXPERIMENTAL ASEE,lS-IN LIGHT WATER REACTOR PRESSURE VESSEL SAFETY RE-MENT OF BOREHOLE PLUG PERFORMANCE. Annual Repoit, June SEARCH.

l 1984 - Hay 1985.

Procoure Vessel Integrity t

^

V06 N04: REPORT TO CONGRESS ON ABNORMAL T

f-ed HigN OCCURRENCES. October -Decemoor 1985.

Phase Precepetetion And Identification NUREG/CR-4594: ESTIMATED SAFETY SIGNIFICANCE OF GENERIC NUREG/CR-4503 V01: LONG TERM EMBRITTLEMENT OF CAST.

SAFETY ISSUE 61.

DUPLEX STAlic.ESS STEELS IN LWR SYS. Annual Report.O:.cber 1984 September 1985.

4 g

Physical Modificaton NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOLING TRAN-NUREG/CR-4627: GENERC COST ESTIMATES. Abstracts From Generic SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL Studies For Use in Preparing Regt4atory Irnpact Analyses.

SHOCK STUDY.

NUREG/CR 3770- PREUMINARY DEVELOPMENT OF AN INTEGRAT.

(

P'P' ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL NUREG-1175: NRC SAFETY RESEARCH IN SUPPCET OF REGULA-SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER TION Se4ected R '

s-PLANT.W/TWO OVERSIZE DRAWINGS.

NUREG/CP 0077:

EDINGS OF THE SEMINAR ON LEAK-NUREG/CR-4483: REACTOR PRESSURE VESSEL FAILURE PROBA-BEFORE-BREAK: INTERNATIONAL POUCIES AND SUoPORTING BluTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR-IZED THERMAL SHOCK EVENTS.

NUREG/CR 28 V04: STRUCTURAL INTEGRITY OF WATER REAC-TOR PRESSURE BOUNDARY COMPONENTS Annual Report for 1985.

Prestressed Concrete Reector Vessel NUREG/CR-4290 V01: PROBABluTY OF PIPE FAILURE IN THE REAC-NUREG/CR-4526: FINITE ELEMENT ANALYSIS OF THE 2240 MW l

TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR HTGR PCRV.

PLANTS. Volume 1 Summary Report.

I NUREG/CR-4484: STATUS OF ACTIVITIE? FOR INSPECTING WELD Prtrnary Wow OVERLAID PIPE JOINTS.

a NUREC/CR-4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 NUREG/CR-4602: UNiOUENESS OF BOILING WATER REACTOR PRI-STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION AND MARY WATER CHEMISTRY. Final Report, October 1985 - March 1986.

UMIT LOAD ANALYSIS.

NUREG/CR-472: NRC LEAK BEFORE-BREAK (LBB.NRC) ANALYSIS Probabilistic RenetWitty Aneiyele METHOD FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED NUREG-1201: REPOAT OF THE INDEPENDENT AD HOC GROUP FOR PIPES UNOF.3 AXIAL PLUS BENDING LOADS. Topical Report.

THE DAVIS-BESSE INCIDENT.

i

Subject index 53 Pr-Alek Aseeeement NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-NUREG-1152 MILLSTONE 3 RISK EVALUATION REPT.AN OVERALL S:ENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBABI-SHOCK STUDY.

LISTIC SAFETY STUDY.

NUREG/CR-3770 PREUMINARY DEVELOPMENT OF AN INTEGRAT-NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL SORED BY OFFICE OF NUCLEAR REGULATORY SHOCK AS APPLIED TO THE OCONEE UNIT 1 NUCLEAR POWER NURE

-4048 ET OCA NG R UABluTY AND RISK.

RESAR-38 NUREG/CR-4142 A REVIEW OF THE MILLSTONE 3 PROBABluSTO NUREG/CR4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK NU CR-4 74 03: A REVIEW OF THE OCONEE-3 PROBABluSTIC LOSS-OF-COOLANT ACCIDENTS IN A RESAR-3S PLANT.

RISK ASSESSMENT CONTAINMENT RHR LOCA PERFORMANCE,RADOLOGICAL SOURCE TEHMS AND RISK ESTI-NUREG/CR-4142: A REVIEW OF THC MILLSTONE 3 PROBABluSTIC N RE /CR-4404: ANALYSIS OF ALLOWED OUTAGE TIMES AT SAFETY STUDY.

NU E C 9R LECTED AREAS OF YANKEE ROWE

/

8 LATON DOSE COMMITMENT DUE TO RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES Frah=h"=* Safety Study IN 1982.

NUREG-1152 MILLSTONE 3 RISK EVALUATON REPTAN OVERALL REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBABl.

Radiation Effects USTIC SAFETY STUDY.

NUREG/CR4489: HISTORICAL

SUMMARY

OF THE HEAVY-SECTION STEEL TECHNOLOGY PROGRAM AND SOME RELATED ACTIVITIES Productivity Factor IN LIGHT-WATER REACTOR PRESSURE VESSEL SAFETY RE-NUREG/CR-4627: GENERIC COST ESTIMATES. Abstracts From Generic SEARCH.

l Studies For Use in Prepanng Regulatory impact Analyses.

Radiation Esposure Program Management NUREG/CR-4315 V01: EVALUATON OF NUCLEAR FACluTY DECOM-NUREG-0985 R02:

U.S. NUCLEAR REGULATORY COMMISSION HUMAN FACTORS PROGRAM PLAN-MISSIONING PROJECTS. Summary Status Report Three Mile Island Urvt 2 Reactor Coolant System & Systems Cm,v,o_...

tion.

Program Plan NUREG/CR4315 V02-EVALUATION OF NUCLEAR FACIUTY DECOM-NUPEG/CR-4316: EVALUATON OF NUCLEAR FACluTY DECOMMIS-MISSIONING PROJECTS. Summary Status Report Three Mile Island UN ST D M i ING-NUR /C -4 5

i N OF NUCLEAR FACluTY DECOM-MISSIONING PROJECTS. Summary Status Report Three Mde Island Quanty Assurance Urut 2 Peactor Defueling & Disassembly.

NUREG 1156. ACCURACY AND DETECTION UMITS FOR BIOASSAY NUREG/CR4316: EVALUATION OF NUCLEAR FACIUTY DECOMMIS-MEASUREMENTS IN RADIATION PROTECTION - STATISTICAL SiONING PROJECTS STATUS REPORT - HUMBOLDT BAY POWER CONSIDERATONS.

PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

Quarterly Progrees Report Radletion Measurement NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-NUREG-1156: ACCURACY AND DETECTON LIMITS FOR BIOASSAY SORED BY OFFICE OF NUCLEAR REGULATORY MEASUREMENTS IN RADIATION PROTECTION - STATISTICAL RESEARCH.Ouarterty Progress Report, July-September 1985.

CONSIDERATIONS.

NUREG/CR-4402 V03: HIGH-TEMPERATURE GAS COOLED REACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT Radletion Protection EVALUATION.Ouartetty Progress Report. July 1 - September 30,1985.

NUREG-1156: ACCURACY AND DETECTON UMITS FOR BIOASSAY RCme MEASUREMENTS IN RADIATON PROTECTION - STATISTICAL CONSIDERATIONS.

NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

REDUCTON RESEARCH PROJECTS.

RADONE NUREG/CR-3441: RADONE:A COMPUTER CODE FOR SIMULATING Radletion Safety Program NUREG-0940 V05 N01: ENFORCEMENT ACTONS.SIGNIFICANT AC-FAST TRANSIENT ONE-OlMENSIONAL HYDFtODYNAMIC CONDI-TONS AND TWO-LAYER RADIONUCUDE CONCENTRATONS IN-TIONS RESOLVED.Quarterty Progress Report, January-March 1986.

CLUDING THE EFFECT OF BEDDEPOSITON IN CONTROLLED Radioactive Release RIVERS AND TIDAL ESTUARIES.

NUREG/CR-2850 V04: POPULATION DOSE COMMITMENT DUE TO RAMONA-38 RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-IN 82.

SORED BY OFFICE OF NUCLEAR REGULATORY Radioactive Weste RESEARCH.Quarterty Progress Report, July-September 1985.

NUREG/CR-4315 V09: EVALUATION OF NUCLEAR FACluTY DECOM-RELAP5 MISSiONING PROJECTS. Summary Status Report Three Mile Island NUREG/CR-3770- PREUMINARY DEVELOPMENT OF AN INTEGRAT-Urut 2 Radioactive Weste And Laundry Shipments.

ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL p

y SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER PLANT.W/TWO OVERSIZE DRAWINGS.

NUREG-1156: ACCURACY AND DETECTON LIMITS FOR BIOASSAY MEASUREMENTS IN RADIATION PROTECTION - STATISTICAL RELAP6/ MOD 2 CONSIDERATIONS.

NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK LOSS-OF-COOLANT ACCIDENTS IN A RESAR-3S PLANT.

R6 SeMy NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK NUREG-1198: RELEASE OF UF6 FROM A RUPTURED MODEL 48Y LOSSOF400LANT ACCIDENTS IN A RESAR-3S PLANT.

CYUNDER AT SEQUOYAH FUELS CORPORATION NUREG/CH-4488: VENTING OF NONCONDENSIBLE GAS FROM THE FACILITY. Lessons-Learned Report UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-NUREG-1198. RELEASE OF UF6 FROM A RUPTURED MODEL 48Y BEND VALVES.

CYLINDER AT SEQUOYAH FUELS CORPORATION FACluTY. Lessons-Learned Report NUREG/CR-3701: REMIX:A COMPUTER PROGRAM FOR TEMPERA-Radiological Source Term TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTION AFTER NUREG/CR 4374 V03: A REVIEW OF THE OCONEE-3 PROBABILISTIC INTERRUPTON OF NATURAL CIRCULATION.

RISK ASSESSMENT CONTAINMENT

54 Subject IfM0SX PERFORMANCE RADOLOGICAL SOURCE TERMS AND RISK ESTI-NUREG/CR4557: A REVIEW OF ISSUES RELATED TO IMPROVING MATES.

NUCLEAR POWER PLANT DIESEL GENERATOR REUABluTY.

Radionuclide Sorpeen Rollability m-Man NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADONU-NUREG/CR-4048: A METHODOLOGY FOR ALLOCATING RELIABluTY CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-AND RISK.

LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For April, June 1985.

RollelWilty Asessement NUREG/CR-3957: REUABluTY ASSESSMENT AND PROBABlUTY CR ADONE-A COMPUTER CODE FOR SIMULATING N

WA FAST-TRANSIENT ONE-DIMENSIONAL HYDRODYNAMC CONDI-TIONS AND TWO-LAYER RADONUCUDE CONCENTRATIONS IN-Remote Handling Equipment CLUDING THE EFFECT OF BED-DEPOSITION IN CONTROLLED NUREG/CR-4409-DATA BASE ON NUCLEAR POWER PLANT DOSE RIVERS AND TIDAL ESTUARIES.

REDUCTION RESEARCH PROJECTS.

Remepo Fault M Sensing NUREG/CR-4580 STONY BROOK SE!SMIC NETWORK ON LONG NUREG/CR-4409-DATA BASE ON NUCLEAR POWER PLANT DOSE ISLAND,NEW YORK Final Report (September 1979 March 1985).

REDUCTION RESEARCH PRalECTS.

Reactor Cooient Loop NUREG/CR-4290 V01: PROBABluTY OF PIPE FAILURE IN THE REAC-Report to Congrees TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR NUREG-0000 V06 N04: REPORT TO CONGRESS ON A8 NORMAL OCCURRENCES. October -December 1985.

PLANTS. Volume 1:Sumrnary Report Reactor Cooient System Repoellory NUREG/CR-4315 V01: EVALUATON OF NUCLEAR FACIUTY DECOM-NUREG/CR-4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE MISSIONING PROJECTS. Summary Status Report Three Mile Island WALL DRILLING DAMAGE IN BASALTIC ROCKS.

Urut 2 Reactor Coolant System & Systems Decontammaton.

Research Contributions To Regulellon Reactor Operation NUREG 1175. NRC SAFETY RESEARCH IN SUPPORT OF REGULA-NUREG-1125 V07: A COMPILATION OF REPORTS OF THE ADVISORY TION - Selected Highhghts.

COMMITTEE ON REACTOR SAFEGUARDS,1985.

Reeeerch Project Reactor Safety NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE NUREG-1125 V07: A COMPILATION OF REPORTS OF THE ADVISORY REDUCTION RESEARCH PROJECTS.

COMMITTEE ON REACTOR SAFEGUARDS 1985.

NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-Residue Number System SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-4634: DEVELOPMENT OF A REAL-TIME RESIDUE NUMBER PROCESSOR FOR SAFT INSPECTON. Phase ll Final NUREG CR 40 V03 H M

TU E D REACTOR Report, September 1984 - April 1986.

SAFETY STUDIES FOR THE DMSION OF ACCIDENT EVALUATION Ouarterly Progress Report. July 1 - September 30,1985.

Reeln NUREG/CR-4498: FIELD TESTING OF WASTE FORMS CONTAINING Recurring Loes Detocuen EPICOR-il ON EXCHANGE RESINS USING LYSIMETERS.

NUREG/CR-4497: NRCPAGE APPUCATIONS MANUAL RNt Redos Process NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND NUREG/CR-4632 NEW MADRID SEISMOTECTONIC PROGRAM Fhel SORPTION.The Solutnisty Of Tc(IV) Oxides.

Report Reduced Preneure Scaling Ring Tension Test NUREG/CR-4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL.

NUREG/CR-4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE ING LAWS FOR TWO. PHASE FLOW LOOP.

WALL DRlLUNG DAMAGE IN BASALTIC ROCKS.

Reetfoot Rift Risk NUREG/CR-4632: NEW MADRID SEISMOTECTONIC PROGRAMFmal NUREG-0900 Rot: NUCLEAR POWER PLANT SEVERE ACCIDENT RE-Report SEARCH PLAN.

NUREG-1152-MILLSTONE 3 RISK EVALUATON REPT:AN OVERALL Reguietory Analyste REVIEW AND EVALUATION OF THE MILLSTONE UNIT 3 PROBASI-NUREG/CR.4568: A HANDBOOK FOR QUICK ESTIMATES:A Method LISTIC SAFETY STUDY.

For Developmg Quick Approxirnate Estimates Of Costs For Generic Ac-NUREG/CR-3770- PRELIMINARY DEVELOPMENT OF AN INTEGRAT.

tons For Nuclear Por*er Plants.

ED APPROACH TO THE EVALUATION OF PRESSURIZED THERMAL SHOCK AS APPLIED TO THE OCONEE UNIT 1 NUCLEAR POWER Remy And Tm W PLANT.W/TWO OVERSIZE DRAWINGS NUREG4304 V11 N01: REGULATORY AND TECHNICAL NUREG/CR-4315 V02: EVALUATION OF ' NUCLEAR FACluTY DECOM-REPORTS.Compdation For First Quarter 1986, January-March.

MISSIONING PROJECTS. Summary Status Report Three Mile leiend Urut 2 Reactor Building Decontamination.

Reguietory Requirement NUREG/CR-4330 V01: REVIEW OF UGHT WATER REACTOR REGU-NUREG/CR-4315 V02-EVALUATON OF NUCLEAR FACIUTY DECOM-LATORY REQUIREMENTS. Volume 1: Identification Of Regulatory Re-MISSIONING PROJECTS. Summary Status Report Three Mde Island quirements That MafHave importance To Risk-Urut 2 Reactor Buddog Decontammation.

NUREG/CR-4330 V01: REVIEW OF LIGHT WATER REACTOR REGU-UR 569: A REV;EW OF THE SEVERE ACCIDENT RISK RE-The May nc R

DUCTON PROGRAM (SARRP) CONTAINMENT EVENT TREES.

Reneeee NUREG-1198: RELEASE OF UF6 FROM A RUPTURED MODEL 48y Rock Drsling CYLINDER AT SEQUOYAH FUELS CORPORATION NUREG/CR-4641: EXPEITMENTAL ASSESSMENT OF BOREHOLE WALL DRILUNG DAM / GE IN BASALTIC ROCKS.

FACiUTY. Lessons-Leamed Report RenabHfty Rock Mees Seeling NUREG/CR-4506: AN CPERATIONAL SAFETY REUABILITY PROGRAM NUREG/CR-4642-ROCK MASS SEAUNG - EXPERIMENTAL ASSESS-APPROACH WITH RECOMMENDATIONS FOR FURTHER DEVELOP-MENT OF BOREHOLE PLUG PERFORMANCEAnnual Report. June MENT AND EVALUATON.

1964 May 1985.

Subject Index 55 Rock Permeability Nos.50-443 And 50-444. (PutAc Service Company of New NUREG/CR-4642-ROCK MASS SEAUNG EXPERIMENTAL ASSESS-Hampshre,et al)

MENT OF GOREHOLE PLUG PERFORMANCE. Annual Report. June NUREG-0954 S06: SAFETY EVALUATION REPORT RELATED TO THE 1984 - May 1985.

OPERATION OF CATAWBA NUCLEAR STATON, UNITS 1 AND Rules Of Procuce

2. Docket Nos. 50 413 And 50-414. (Duke Power Company.et el)

NUREG-0979 S05: SAFETY EVALUATION REPORT RELATED TO THE NUREG4386 D04 R01: UNITED STATES NUCLEAR F$EGULATORY DESIGN APPROVAL OF THE GESSAR 18 BWR/6 NUCLEAR ISLAND COMMISSON STAFF PRACTICE AND PROCEDURE OtGEST JULY DESIGN. Docket No. 50447. (General Electric Company) 1,1972 - SEPTEMBER 30,1985.

NUREG-1038 S03: SAFETY EVALUATON REPORT RELATED TO THE OPERATION OF SHEARON HARRIS NUCLEAR POWER PLANT. UNIT NUREG-1179 V02: RUPTURE OF MODEL 48Y OF6 CYUNDER AND RE-1.Docw No.50400. (Candne Power And Ught Company And North LEASE OF URANtUM HEXAFLUORIDE.Cytinder Overfill, March 12 CenAne Eastern Municipal Power A9ency) 13,1986. Inves5getion Of A Feded UF6 SNpping Cylinder.

NUREG-1048 S05: SAFETY EVALUATON REPORT RELATED TO THE OPERATION OF HOPE CREEK GENERATING STATION. Docket SAFSTOR No.50-354. (PutAc Service Electric And Gas Company,Atlanbc Oty NUREG/CR-4316: EVALUATION OF NUCLEAH FACluTY DECOMMIS-Electric Company).

SiONING PROJECTS STATUS REPORT - HUMBOLDT BAY POWER NUREG-1057 S01: SAFETY EVALUATION REPORT RELATED TO THE PLANT UNIT 3 SAFSTOR DECOMMISSONING.

OPERATION OF BEAVER VALLEY POWER STATON, UNIT 2. Docket No. 50-412.(Duquesne Ught Company.et al)

SAFT-UT NUREG-1137 S02-SAFETY EVALL,ATION REPORT RELATED TO THE NUREG/CR4634: DEVELOPMENT OF A REAL TIME RESIDUE OPERATION OF VOGTLE ELECTRO GENERATING PLANT, UNITS 1 NUMBER PROCESSOR FOR SAFT INSPECTON. Phase 11 Fine!

Report. September 1984 April 1996.

AND 2. Docket Nos. 50424 And 50-425. (George Power Company et ef)

gagg, NUREG-1177: SAFETY EVALUATON REPORT RELATED TO THE RE-NUREG/CR-4569: A REV!EW OF THE SEVERE ACCIDENT RISK RE-START OF DAVIS-BESSE NUCLEAR POWER STATON, UNIT 1.FOL.

DUCTION PROGRAM (SARRP) CONTAINMENT EVENT TREES.

LOWING THE EVENT OF JUNE 9,1985. Docket No. 50-346.(Toledo Edison Company)

SMACS NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON_

SofWy Goele SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4048: A METHODOLOGY FOR ALLOCATING REUA8iUTY RESEARCH.Ouenerty Progress Report, July-September 1985.

AND RISK SWIFT n Safety Margin NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ILThe Sandia NUREG/CR-4412: AN ASSESSMENT OF SAFETY MARGINS IN ZlRCA.

Weste-isolation Flow And Transport Model For Fractured Media Re-LOY OXIDATON AND EMBRITTLEMENT CRITERIA FOR ECCS AC-lease 4.84.

CEPTANCE.

Sabotage NUREG/CR-4319: NUCRAC - A CODE FOR THE ESTIMATION OF AD-Safety Roemerch VERSARY-ACTION CONSEQUENCES IN THE NUCLEAR POWER NUREG-1125 V07: A COMPILATION OF REPORTS OF THE ADylSORY E.

COMMITTEE ON REACTOR SAFEGUARDS.1985.

NUREG-1175: NRC SAFETY RESEARCH IN SUPPORT OF REGULA-Safeguards TlON - Selected Highlights.

NUREG/CR-4516: INTERNATONAL SAFEGUARDS AT FACluTIES EM.

NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-PLOYING SPENT FUEL ROD CONSOLIDATON.

SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Ouarterty Progress Report. July-September 1985.

Sefectorage NUREG/CR-4489-HISTORICAL

SUMMARY

OF THE HEAVY-SECTION NUREG/CR4316: EVALUATON OF NUCLEAR FACluTY DECOMMIS-STEEL TECHNOLOGY PROGRAM AND SOME RELATED ACTIVITIES SiONING PROJECTS STATUS REPORT - HUMBOLDT BAY POWER IN UGHT-WATER REACTOR PRESSURE VESSEL SAFETY RE-PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

SEARCH.

SafWy safety Sheer Wall Structure NUREG/CR4265 V01: AN ASSESSMENT OF THE SAFETY IMPLICA-NUREG/CR-3957: REUA8luTY ASSESSMENT AND PROBABluTY PLANT NUR CR-4 ECH G

ET AN CO O DECOM:

BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS MISSONING NUCLEAR FUEL CYCLE FACIUTIES CLASSIFICATION AND SHEAR WALLS. Summary Report.

OF DECOMMISSIONING WASTE.

SatWy Signinconce Safety Engineering NUREG/CR-4594: ESTIMATED SAFETY SIGNIFICANCE OF GENERIC NUREG-1125 V07: A COMPILATON OF REPORTS OF THE ADVISORY SAFETY ISSUE 61.

COMMITTEE ON REACTOR SAFEGUARDS,1985.

Safety System Safety Evolustion NUREG/CR-3453: ELECTRONIC ISOLATERS USED IN SAFETY SYS-NUREG-%40 V03 NOI: ENFORCEMENT ACTONS.SIGNIFICANT AC-TEMS OF U S. NUCLEAR POWER PLANTS.

TIONS RESOLVED.Ouarterty Progress Report. January-March 1986.

Scale Model Safety Evolueuen Report NUREG-0675 S33: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL-ING LAWS FOR TWO-PHASE FLOW LOOP.

OPERATION OF DIABLO CANYON NUCLEAR POWER, UNITS 1 AND

2. Docket Nos 50-275 and 50-323(Pacific Gas And Electric Company)

NUREG-0781: SAFETY EVALUATON REPORT F1 ELATED TO THE OP-Scaling ERATON OF THE SOUTH TEXAS PROJEC, UNITS 1 AND 2 Docket NUREG/CR-4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL.

Nos 50-498 A9d 50-499 (Houston '

Ar Power Company)

ING LAWS FOR TWO-PHASE FLOW LOOP.

NUREG4797 S13: SAFETY EVALUA ON EPCRT RELATED TO THE OPERATON OF COMANCHE PEAK STEAM ELECTRIC Scoping Study STATION UNITS 1 AND 2. Docket Nos. 50-445 And 50-446.(Texas Utili.

NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC.

ties Generating Company)

TIVENESS OF AN OPERATIONAL SAFETY REUABluTY PROGRAM NUREG4857 SfD-SAFETY EVALUATION REPORT RELATED TO THE IN ADDRESSING GENERIC SAFETY PROBLEMS OPERATON OF PALO VERDE NUCLEAR GENERATING STATION, UNITS 1,2 And 3. Docket Nos. 50-528,50-529 And 50-Seel N -

3 530 (Artrone Publec Sennce Company)

NUREG/CR-4642: ROCK MASS SEAUNG - EXPERIMENTAL ASSESS-NUREG4896 SO4-SAFETY EVALUATION REPORT RELATED TO THE MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June OPERATION OF SEABROOK STATION, UNITS 1 AND 2. Docket 1984. May 1985.

c

r 56 Subject Itulex Sealing Sos-Structure interaction NUREG/CR-4642 ROCK MASS SEALING EXPERIMENTAL ASSESS-NUREG/CR-*588 V01: SOIL-STRUCTURE INTERACTON.Vol 1;lnflu-MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report. June ence Of Layering.

1984 May 1985.

NUREG/CR-4588 V02: SOIL-STRUCTURE INTERACTON.Vol 2.Influ-ence Of Utt-Off.

Sediment Layer NUREG/CR-3441: RADONE:A COMPUTER CODE FOR SIMULATING Solubility FAST-TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDI-NUREG/CR-3472 V02: SURFACE PROPERTIES AND PERFORMANCE TIONS AND TWO-LAYER RADIONUCUDE CONCENTRATIONS IN-PREDICTON OF ALTERNATIVE WASTE FORMS. Final Report.

CLUDING THE EFFECT OF BED-DEPOSITON IN CONTRO' LED NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND RIVERS AND TIDAL ESTUARIES.

SORPTON.The SohMty Of Tc(IV) Oxides.

NUREG/CR-4582: TEMPERATLRE EFFECTS ON THE SOLUBluTY Seismic Reeponse AND SPECIATION OF SELECTED ACTINIDES.

NUREG/CR-4588 V02 SOfL-STRUCTURE INTERACTON.Vol 2.Influ-ence Of Utta solute Transport NUREG/CR-4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT Setemicity IN THE UNSATURATED ZONE. Information And Data Sets.

NUREG/CR-3150t SEISMICITY AND TECTONIC RELATIONSHIPS FOR UPPER GREAT LAKES PRECAMBRIAN SHIELD PROVINCE FINAL 80'Pl88" NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND REPORT - July 1981 -December 1982.

NUREG/CR-4580 STONY BROOK SEISMIC NETWORK ON LONG SORPTON.The SokMty Of Tc(IV) Oxides.

ISLAND,NEW YORK. Final Report (September 1979 - March 1985).

Solomographic Networtt NUREG4900 Rot: NUCLEAR POWER PLANT SEVERE ACCOENT RE-NUREGICR-4580: STONY BROOK SEISMIC NETWORK ON LONG SEARCH PLAN.

NUREG/CR 2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-ISLAND,NEW YORK. Final Report (September 1979 - March 1985).

SORED BY OFFICE OF NUCLEAR REGULATORY Semlocale RESEARCH.Ouarterty Progress Report. July-September 1985.

NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK NUREG/CR4338: TELLURIUM BEHAVIOR IN CONTAINMENT UNDER LOSS OF COOLANT ACCIDENTS IN A RESAR-3S PLANT.

UGHT WATER REACTOR ACCIDENT CONDITIONS.

NUREG/CR-4561: FIRAC USER'S MANUAL:A COMPUTER CODE TO SeN Mod-2A Facility SIMULATE FIRE ACCOENTS IN NUCLEAR FACluTIES.

NUREG/CR-3970- TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT:

LOBI INTERMEDIATE BREAK TEST B-R1M.

U CR-4582-TEMPERATURE EFFECTS ON THE SOLUBluTY Sonettlaed Type 304 Stainiees Steel AND SPECIATION OF SELECTED ACTINIDES.

NUREG/CR-4602 UNIQUENESS OF BOluNG WATER REACTOR PRi-MARY WATER CHEMISTRY. Final Report. October 1985 - March 1986.

RE CR-4516: INTERNATONAL SAFEGUARDS AT FACluTIES EM-Service Water System PLOYING SPENT FUEL ROD CONSOUDATON.

NUREG-0800 09.2.1 R4: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Steiniees Steel PLANTS. LWR Editon. Revision 4 to Section 9.2.1, " Station Service NUREG/CR-4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER (EMAT) DEFECT CHARACTERIZATION OF NUCLEAR REACTOR Water System."

PIPING WELDS. Phase i Final Report. October 1985 March 1986.

Swore Accident NUREG-0900 RO1: NUCLEAR POWER PLANT SEVERE ACCIDENT RE.

Stainless Steel Pipe NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 SEARCH PLAN.

NUREG/CR4413: LOSS OF CONTROL AIR AT BROWNS FERRY UNIT STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION AND ONE - ACCIDENT SEQUENCE ANALYSIS.

UMIT LOAD ANALYSIS.

NUREG/CR-4569: A REVIEW OF THE SEVERE ACCIDENT RISK RE-Standard Review Plan DUCTON PROGRAM (SARAP) CONTAINMENT EVENT TREES.

NUREG 0800 09.2.1 R4: STANDARD REVIEW PLAN FOR THE REVIEW Sleerpotectonic OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUHEG/CR-4632: NEW MADRID SEISMOTECTONIC PROGRAM. Final PLANTS. LWR EditiortRevision 4 to Section 9.2.1, " Station Service Report.

Water System."

NUREG 0800 09.2.2 R3: STANDARD REVIEW PLAN FOR THE REVIEW Signal WC - -

-'4 OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR4583 V01: DEVELOPMENT AND VAUDATION OF A REAL-PLANTS. LWR Edition. Revision 3 to Section 9.2.2, " Reactor Auxiliary TIME SAFE-UT SYSTEM FOR THE INSPECTON OF UGHT WATER Coohng Water Systems."

COMPONENTS. Semi-Annual Report For April 1984-September 1984.

NUREG/CR4634: DEVELOPMENT OF A REAL TIME RESIDUE Station Blackout NUMBER PROCESSOR FOR SAFT INSPECTION Phase 11 Final NUREG-1152: MILLSTONE 3 RISK EVALUATION REPTAN OVERALL Report. September 1964 Apnl 1986.

REVIEW AND EVALUATION OF THE MILLSTONE UNIT 3 PROBABl-LISTIC SAFETY STUDY.

Similartty Lawe NUREG/CR-4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL.

Station Service Water System ING LAWS FOR TWO PHASE FLOW LOOP.

NUREG-0800 09.2.1 R4; STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Single Failure Criterton PLANTS. LWR Editon. Revision 4 to Section 9.2.1, " Station Service NUREG/CR-4207: FAULT TREE APPLICATON TO THE STUDY OF Water System."

SYSTEMS INTERACTONS AT INDIAN POINT 3.

Steam Condoneing Mode Small Broek LOCA NUREG/CR-4594: ESTIMATED SAFETY SIGNIFICANCE OF GENERIC NUREG/CR4142 A REVIEW OF THE MILLSTONE 3 PROBABILISTIC SAFETY ISSUE 61.

SAFETY STUDY.

NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK Steam Gerierator LOSSOF-COOLANT ACCIDENTS IN A RESAR-3S PLANT.

NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATONS NUREG/CR-4621: FLOW VISUAUZATION EXPERIMENT ON HOT-LEG OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL U-BEND TWO-PHASE NATURAL CIRCULATON PHENOMENA.

REPORT.

NUREG/CR 4265 V01: AN ASSESSMENT OF THE SAFETY IMPUCA.

Soll Structure interaction TIONS OF CONTROL AT THE CALVERT CLIFFS-1 NUCLEAR PLANT.

NUREG/CR4588 V03. SOIL-STRUCTURE INTERACTION Vol 3.Influ-NUREG/CR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-ence Of Ground Water.

ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS.

Subject index 57 NUREG/CR4581: DRYOUT FRONT MODELING FOR PWR THERMAL System RollelWIlly HYDRAUUC ANALYSIS.

NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC-TlVENESS OF AN OPERATIONAL SAFETY REUA31LITY PROGRAM E CR40 :

TRAC.PF1/ MOD 1 INDEPENDENT ASSESSMENT.Condoneeton in Stratriled Cocurrent Flow.

Systems inteaction NUREG/CR-4207: FAULT TREE APPUCATON TO THE STUDY OF

/

701: REMIX-A COMPUTER PROGRAM FOR TEMPERA.

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTON AFTER TLD N

G/CR 7 UO CT OVERCOOUNG TRAN-N K

no SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL NUREG-0837 V06 N01: NRC TLD DIRECT RADIATION MONITORING SHOCK STUDY-NETWORK. Progress Report, January-March 1986.

TRAC NUREG/CR 4528: FINITE ELEMENT ANALYSIS OF THE 2240 MW NUREG/CR-3770- FREUMINARY DEVELOPMENT OF AN INTEGRAT-ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL Strees Corrosion SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER NUREG/CR-4409-DATA BASE ON NUCLEAR POWER PLANT DOSE PLANT.W/TWO OVERSIZE DRAWINGS.

REDUCTON RESEARCH PROJECTS.

g j

Strees Corrosion Crecidng NUREG/CR-4549-DETERMINATION OF APPENDIX K CONSERV.

NUREG/CP-0067 V01: PROCEEDINGS OF THE SECOND IAEA SPE-ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR CIAUSTS' MEETING ON SUBCRITICAL CRACK GROWTH.Sessons i USING TRAC-PD2/ MODI CODE.

And ll, Held At Sendai, Japan,May 15-17,1985.

NUREG/CP-0087 V02: PROCEEDINGS OF THE SECOND IAEA SPE.

TRAC-PET CIAUSTS MEETING ON SUBCRITICAL CRACK GROWTH.Sessons ill NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK

& IV Held At Sendel, Japan,May15-17,1985.

LOSS-OF COOLANT ACCIDENTS IN A RESAR-3S PLANT.

NUREG/CR-2331 V05 N3: SAFETY FsESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY TRAC-PF1/ MOO 1 RESEARCH.Ouarterly P ogrese Report. July-September 1965.

NUREG/CR-3970: TRAC-PF1/ MODI INDEPENDENT ASSESSMENT:

NUREG/CR-4490 V01: LIGHT WATER-REACTOR SAFETY MATERIALS LOBIINTERMEDIATE BREAK TEST B-R1M.

ENGINEERING RESEARCH PROGRAMS.Quarterty Progress NUREG/CR-4027:

TRAC-PF1/ MOD 1 INDEPENDENT Report January-March 1985.

ASSESSMENT.Condensaton in Stratified Coeurrent Flow.

NUREG/CR-4619: STRESS CORROSON CRACKING TESTS ON HIGH-LEVEL WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE.

Tallinge :, _ r _

POSITORY ENVIRONMENTS.

NUREG/CR-4620- METHODOLOGIES FOR EVALUATING LONG-TERM STABlu2ATON DESIGNS OF URANIUM MILL TAluNGS IMPOUND-Semdural Integr#y MENTS.

NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC.

TOR PRESSURE BOUNDARY COMPONENTS.An iual Report for 1985.

Toering instatWilty NUREG/CR-4539: INVESTIGATON OF TEARING INSTABluTY PHE-

/CR-4489: HISTORICAL

SUMMARY

OF THE HEAVY-SECTON STEEL TECHNOLOGY PROGRAM AND SOME RELATED ACTMTIES Toering anon.

IN LIGHT WATER REACTOR PRESSURE VESSEL SAFETY RE-NUREG/CR-4579: APPUCATON OF THE KEY CURVE AND MULTI-SEARCH.

SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF ALLOY STEEL h CM M NUREG/CP-0067 V01: PROCEEDINGS OF THE SECOND IAEA SPE-Technical Socification CIALISTS' MEETING ON SUBCRITICAL CRACK GROWTH.Sessons i NUREG-1181: TECHNICAL SPECIFICATONS FOR PALO VERDE NU-Ohi

^

NUR 7

HE SECOND lAEA SPE-I CIALISTS MEETING ON SUBCRITICAL CRACK GROWTH.Sessons ill

& IV Held Al Sendai Japan,May 15-17,1985.

Technical Specification Super System Code NUREG-1186: TECHNICAL SPECIFICATIONS FOR HOPE CREEK GEN-NUREG/CR-2331 VOS N3: SAFETY RESEARCH PROGRAMS SPON-ERATING STATION. Docket No. 50-354.(Public Service Electric And SORED BY OFFICE OF NUCLEAR REGULATORY NUR G ECHNICAL SPECIFICATONS FOR CATAWBA NUCLE-RESEARCH.Quarterfy Progress Report. July-September 1985.

AR STATON UNITS 1 AND 2. Docket tios. 50 413 And 50-414.(Duke Supertiested Paractee Power Company)

NUREG/CR-4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-OU NCH NG RHWED DEBRIS BEDS UNDER U

-4519: TECHNOLOGY, SAFETY AND COSTS OF DECOM-MISSIONING NUCLEAR FUEL CYCLE FACluTIES CLASSIFICATION Survey Instrument OF DECOMMISSIONING WASTE.

NUREG/CR-4511: ASSESSMENT OF THE ADEQUACY OF THE CAU-BRATIONS PERFORMED BY COMMERCIAL CAllBRATION SERV-Tectonles ICES FOR ONIZING RADIATION SURVEY INSTRUMENTS.

NUREG/CR-3150: SEISMICITY AND TECTONIC RELATIONSHIPS FOR UPPER GREAT LAKES PRECAMBRIAN SHIELD PROVINCE FINAL Synthetic Aperture Focusing REPORT - July 1981 -December 1982.

NUREG/CR-4583 V01: DEVELOPMENT AND VALIDATON OF A REAL.

TIME SAFE-UT SYSTEM FOR THE INSPECTION OF LIGHT WATER Temperature COMPONENTS. Semi-Annual Report For April 1984-September 1984.

NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

System Analysie NUREG/CR-4582: TEMPERATURE EFFECTS ON THE SOLUBluTY NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTION EXPERI.

AND SPECIATON OF SELECTED ACTINIDES.

ENCE IN NUCLEAR POWER PLANTS.

Temperature Transient System Interaction Event NUREG/CR-3701: REMIX:A COMPUTER PROGRAM FOR TEMPERA.

NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTON EXPERI.

TURE TRANSIEteTS DUE TO HIGH PRESSURE INJECTON AFTER ENCE IN NUCLEAR POWER PLANTS.

INTERRUPTION OF NATURAL CIRCULATION.

v-.

58 Sutdoct Index Test Two phase Flow Regimes NUREG/CR-4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-NUREG/CR 4621: FLOW VISUAUZATION EXPERIMENT ON HOT-LEG RADIATON ON THE MECHANICAL PROPERTIES OF HIGH DENSITY U-BEND TWofi(ACC NATUR AL CIRCULATION PHENOMENA.

POLYETHYLENE.

Two-phees Flow Simulellon Testing NUREG/CR4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL-NUREG/CR 4330 V02-REVIEW OF UGHT WATER REACTOR REGU-ING LAWS FOR 1WO. PHASE FLOW LOOP.

LATORY REOUIREMENTS ASSESSMENT OF SELECTED REGULA-TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE Ultrasonic inepection TO RISK. Reactor Contammens Leekage Rates. Man Steam isolabon NUREG/CR-4484: STATUS OF ACTIVITIES FOR INSPECTING WELD Valve Leakage-OVERulD PlPE JOINTS.

Thermal Hydreuste

/

583 V01: DEVELOPMENT AND VAUDATON OF A REAL-ED OFF OF R ULAT Y TIME SAFE-UT SYSTEM FOR THE INSPECTION OF UGHT WATER RESEARCH Ouarterty ess R ember 1985 NUREG/CR-3262 V01:

RA

.A T R AL-HYDRAUOC CODE COMPONENTS. Semi-Annual Report For April 1984-Soptember 1984.

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR Mtremonico COMPONENTS, Volume 1 Equenons And Constitutive Models.

NUREG/CR-3262 V02: COBRA-NC A THERMAL-HYDRAUUC CODE NUREG/CR-4634: DEVELOPMENT OF A REAL TIME RESIDUE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUMBER PROCESSOR FOR SAFT INSPECTION. Phase il Final COMPONENTS. Volume 2: COBRA-NC Numencal Solubon Methods.

Report September 1984 - April 1986.

NUREG/CR-3262 V07; COBRA-NCA THERMAL-HYDRAUUC CODE FOR TRANSIENT ANALYSIC OF NUCLEAR REACTOR Uncertainties COMPONENT. Volume 7: Assessment Manual for Contamment 4 plica-NUREG/CR-3770- PREUMINARY DEVELOPMENT OF AN INTEGRAT-bons.

ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL NUREG/CR-3970- TRAC PF1/ MOD 1 INDEPENDENT ASSESSMENT:

SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR POWER LOBI INTERMEDIATE BREAK TEST B-R1M.

PLANT.W/TWO OVERSZE DRAWINGS.

NUREG/CR-4488 VENTING OF NONCONDENSIBLE GAS FROM THE UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-Undercooling Nt E 8 DRY FRONT MODEUNG FOR PWR THERMAL T

OF T TH CA CF Underground Facility ThermalHydraulle Analysie NUREG/CR-4609: EFFECTS OF EARTHOUAKES ON UNDERGROUND NUREG/CR-4384: BREAK SPECTRUM ANALYSIS FOR SMALL BREAK FACluTIES.uterature Review And Discueeson.

LOSS-OF-COOLANT ACCIDENTS IN A RESAR-3S PLANT.

Thermolumineecent Doeimeter Unexpected Criticality NUREG4837 V05 NO4: NRC TLD DIRECT RADIATON MONITORING NUREG-0090 VOS N04: REPORT TO CONGRESS ON ABNORMAL NETWORK. Progress Report, October-December 1985.

OCCURRENCES. October -December 1985.

NUREG4837 V06 N01: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report, January-March 1986.

Unified Transport Approach NUREG/CR-3441: RADONE A COMPUTER CODE FOR SIMULATING Through-Well Crack FAST TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDl-NUREG/CR-4483: REACTOR PRESSURE VESSEL FAILURE PROBA-TIONS AND TWO-LAYER RADIONUCUDE CONCENTRATIONS IN-BluTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRESSUR*

CLUDING THE EFFECT OF BED-DEPOSITON IN CONTROLLED IZED THERMAL SHOCK EVENTS-RIVERS AND TIDAL ESTUARIES.

D' Unresolved Sofety feeue A-17 NURE O

NO2: Ti ST OF DOCUMENTS MADE PUBLICLY NUREG/CR-4207: FAULT TREE APPUCATION TO THE STUDY OF NUREG4540 V08

TITI'E UST OF DOCUMENTS MADE PUBUCLY SYSTEMS INTERACTONS AT INDIAN POINT 3.

AVAILABLE. March 1 31 1986 NUREG4540 V06 N04: Tl'fLE OST OF DOCUMENTS MADE PUBUCLY Unresolved Safety leeue A-47 NUREG/CR4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS AVAILABLE. April 1-30,1986.

OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL Tornado REPORT.

NUREG/CR-4461: TORNADO CLIMATOLOGY OF THE CONTIGUOUS NUREG/CR-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-UNITED STATES.

ING THE SAFETY IMPUCATONS OF CONTROL SYSTEMS.

Trenoient Uneaturated Zone NUREG/CR-3262 V01: COBRA-NC:A THERMAL HYDRAUUC CODE NUREG/CR-4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR IN THE UNSATURATED ZONE. Information And Data Sets.

COMPONENTS. Volume 1 Equations And Consttutrve Models.

NUREG/CR-3262 V02: COBRA NC:A THERMAL-HYDRAUUC CODE Uranium Hesefluoride FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NUREG 1179 V02: RUPTURE OF MODEL 48Y UF6 CYUNDER AND RE-COMPONENTS. Volume 2 COBRA-NC Numerical Solution Methods.

LEASE OF URANIUM HEXAFLUORIDE.Cylmder OverfillMarch 12-NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN-13.1986. Investigabon Of A Failed UF6 Shippmg Cytinder.

SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL NUREG-1198: RELEASE OF UF6 FAOM A RUPTURED MODEL 48Y

' ESTIMATED SAFETY SIGNIFICANCE OF GENERIC CYLINDER AT SEQUOYAH FUELS CORPORATION NU

/C 5 :

FACIUTY. Lessons-Leamed Report.

SAFETY ISSUE 61.

Tuff Urenlum Mill NUREG/CR-4236 V03: PROGRESS IN EVALUATON OF RADIONU-NUREG/CR 4620 METHODOLOGIES FOR EVALUATING LONG-TERM CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-STABluZATION DESIGNS OF URANIUM MILL TAILINGS IMPOUND-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For MENTS.

NUI G -

STRESS CORROSION CRACKING TESTS ON HIGH-User's Manuel LEVEL-WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE.

NUREG/CR-4507: HECTR VERSION 1.5 USER'S MANUAL POSITORY ENVIRONMENTS.

NUREG/CR-4561: FIRAC USER'S MANUAL A COMPUTER CODE TO SIMULATE FIRE ACCIDENTS IN NUCLEAR FACIUTIES.

Two D6mensional Elhptic Fluid Flow NUREG/CR 3705: IMPi'OVED MODEUNG AND NUMERICS TO SOLVE VICTORR TWO-DIMENSIONAL ELUPTIC FLUID FLOW AND HEAT TRANSFER NUREG/CR-3064 V02: COMPUTATIONAL METHODOLOGY FOR OAK PROBLEMS.

RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-

Subject index 59 TOR (BSR):The VICTORR Input Procesemg Code For The Bold Ven-Water Chemistry ture System, Volume IL NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE REDUCTON RESEARCH PROJECTS.

Valence State NUREG/CR-4490 V01: LIGHT-WATER-REACTOR SAFETY MATERIALS NUREG/CR-4309: VALENCE EFFECTS ON SOLUBluTY AND ENGINEERING RESEARCH PROGRAMS:Ouarterly Progrees SORPTON.The Solutwhty Of Te(IV) Oxides.

Report January-March 1985.

Velve impact Water Layer NUREG/CR-4627: GENERIC COST ESTIMATES. Abstracts From Genenc NUREG/CR-3441: RADONE:A COMPUTER CODE FOR S:MULATING Studies For Use in Propenng Regulatory impact Analyses.

FAST. TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDI-TONS AND TWO-LAYER RADIONUCUDE CONCENTRATIONS IN.

Vent Valve CLUDING THE EFFECT OF BED-DEPOSITION IN CONTROLLED NUREG/CR4488: VENTING OF NONCONDENSIBLE GAS FROM THE RIVERS AND TIDAL ESTUARIES.

UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-

[ REG BEND VALVES.

15 V01: MODEUNG STUDY OF SOLUTE TRANSPORT l

Vertical Flesion Product Release Apparatus IN THE UNSATURATED ZONE. Information And Data Sets.

l NUREG/CR4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT Wave DWortion l

FOR VERTICAL FISSION PRODUCT RELEASE APPARATUS IN HOT NUREG/CR-4484: STATUS OF ACTIVITIES FOR INSPECTING WELD CELL B. BUILDING 4501.

OVERLAID PIPE JOINTS.

Volumes and Weight Wold NUREG/CR-4315 V09 EVALUATON OF NUCLEAR FACluTY DECOM-NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 MISSIONING PROJECTS. Summary Status Report Three Mile Island STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATON AND Urut 2 Radioactive Waste And Laundy Stupments.

UMIT LOAD ANALYSIS.

NUREG/CR-4631: ELECTROMAGNETIC ACOUSTIC TRANSDUCER Weste Closelfication (EMAT) DEFECT CHARACTERIZATION OF NUCLEAR REACTOR NUREG/CR-4519-TECHNOLOGY. SAFETY AND COSTS OF DECOM-PIPING WELDS. Phase 1 F' al Report,0ctober 1985 - March 1986.

m MISSIONING NUCLEAR FUEL CYCLE FACIUTIES CLASSIFICATION OF DECOMMISSIONING WASTE.

W R

4124 V02: NDE OF STAINLESS STEEL AND ON-UNE Weste Disposal LEAK MONITORING OF LWRS. Annual Report October 1984 - Sep-NUREG/CR 2675 VOS: RELEVANCE OF BIOTIC PATHWAYS TO THE tomber 1985.

LONG-TERM REGULATON OF NUCLEAR WASTE DISPOSAL (Esti.

NUREG/CR-4484: STATUS OF ACTIVITIES FOR INSPECTING WELD mation Of Radiation Dose To Man Resulting From Biotic TransportThe OVERLAID PIPE JOINTS.

BIOPORT/MAXII Software Package).

Wht NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ll.The Sandia NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-Waste-isolaton Flow And Transport Model For Fractured Media Re-TOR PRESSURE BOUNDARY COMPONENTS. Annual Report for 1985.

lease 4.84.

NUREG/CR-3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR Wetwell Alrapece THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

NUREG/CR-4594: ESTIMATED SAFETY StGNIFICANCE OF GENERIC MAX 11 Computer Program.

SAFETY ISSUE 61.

NUREG/CR-4315 V09: EVALUATON OF NUCLEAR FACIUTY DECOM-MISSIONING PROJECTS. Summary Status Report Three M4e Island wrong c o p _..

Urvt 2 Radoactive Waste And Laundry Shipments.

NUREG-1192: AN INVESTIGATON OF THE CONTRIBUTORS TO WRONG UNIT OR WRONG TRAIN EVENTS.

Weste Flow NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ll.The Sarxha Wrong Train Event Waste-Isoletion Flow And Transport Model For Fractured Media Re-NUREG-1192 AN INVESTIGATON OF THE CONTRIBUTORS TO lease 4 84.

WRONG UNIT OR WRONG TRAIN EVENTS.

M AN WSNM & M NMMS M UREG CR-4379 V04: LONG-TERM PERFORMANCE OF MATERIALS WRONG M OR WRONG NN MM USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, Year Four Apnl 1985 March 1986-Zircaloy Fracture NUREG/CR-4453 V02 LIGHT WATER-REACTOR FUEL SAFETY SYS-UREG CR 379 V04: LONG-TERM PERFORMANCE OF MATERIALS 5'

USED FOR HIGH-LEVEL WASTE PACKAGING. Annual Report, Year Four. Apnl 1985. March 1986.

Zircoloy Oxidation And Embrtttlement NUREG/CR-4412: AN ASSESSMENT OF SAFETY MARGINS IN ZlRCA.

Weste Proceseine LOY OXIDATION AND EMBRITTLEMEhI CRITERIA FOR ECCS AC.

NUREG/CR-4601: TECHNICAL CONSIDERATIONS AFFECTING PREP-CEPTANCE.

ARATION OF ION-EXCHANGE RESINS FOR DISPOSAL NilREG-116'70 RAFT ENVIRONMENTAL STATEMENT FOR DECOM-Weste Repository 6

NUREG/CR-3572 V02: DETERMINATION OF METABOLIC DATA AP-MISSIONING HUMBOLDT BAY POWER PLANT, UNIT 3. Docket No.

PROPRIATE FOR HLW DOSIMETRY.ll.Gastromtestmal AbsorptKm.

50-133.(Pacific Gas And Ehttric Company)

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NRC Originating Organization Index (Staff Reports)

This index lists those NRC organizations branches) where appropriate. Each entry is that have published staff reports. The index followed by a NUREG number and title of b arranged alphabetically by major NRC or-the report (s). If further information is ganizations (e.g., program offices) and then needed, refer to the main citation by by subsections of these (e.g., divisions, NUREG number.

l ADVISORY COutNTTEE(S)

U.S. NUCLEAR REGULATOPY COMMISSION ACRS ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NRC - NO DETAILED AFFILIATION GIVEN NUREG 1125 V07: A COMPILATION OF REPORTS OF THE ADVISO.

NUREG-1179 V02: RUPTURE OF MODEL 48Y UF6 CYUNDER AND RY COMMITTEE ON REACTOR SAFEGUARDS 1985.

RELEASE OF URANIUM HEXAFLUORIDE.Cytinder Overfill. March 12 13.1986. Investigation Of A Failed UF6 Shippng Cyhnder.

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

NUREG/CR-3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR REGON 1. OFFCE OF DIRECTOR THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

NUREG-0837 VOS N04: NFO TLD DIRECT RADIATON MONITORING MAX 11 Computer Program.

j NETWORK.Progresa Report. October-December 1985.

NUREG/CR-4603: APPRAISING ATMOSPHERIC TRANSPORT AND NUREG-0837 V06 N01: NRC TLD DIRECT RADIATON MONITORING DIFFUSION MODELS FOR EMERGENCY RESPONSE FACluTIES.

i l

NETWORK. Progress Report, January-March 1986.

OFFICE OF NUCLEAR REGULATORY REEEARCH (POST 4/05/81)

EDO. OFFICE OF ADMINISTRATION CFFCE OF NUCLEAR REGULATORY RESEARCH. DIRECTOR DIVISON OF TECHNICAL INFORMATON & DOCUMENT CONTROL NUREG-0900 RO1: NUCLEAR POWER PLANT SEVERE ACCIDENT NUREG-0304 V11 N01: REGULATORY AND TECHNICAL RESEARCH PLAN.

REPORTS. Compilation For First Quarter 1986. January-March.

NUREG-1175: NRC SAFETY RESEARCH IN SUPPORT OF REGULA-NUREG-0540 V08 NO2: TITLE UST OF DOCUMENTS MADE PUBUC-TON Selected Highlights.

LY AVAILABLE. February 1-28.1986.

OlVISION OF RADIATON PROGRAMS & EARTH SCIENCES (POST NUREG4540 V08 NO3: TITLE UST OF DOCUMENTS MADE PUBUC-840429)

LY AVAILABLE. March 1-31.1986.

NUREG 1156: ACCURACY AND DETECTION UMITS FOR BIOASSAY NUREG-0540 V08 N04: TITLE UST OF DOCUMENTS MADE PUBUC-MEASUREMENTS IN RADIATION PROTECTON - STATISTICAL LY AVAILABLE. April 130,1986.

CONSIDERATONS.

NUREG4750 V23101: INDEXES TO NUCLEAR REGULATORY COM.

DIVISION OF ENGINEERING TECHNOLOGY MISSON ISSUANCES. January-March 1986.

NUREG-1209: PROGRAM PLAN FOR ENVIRONMENTAL QUAUFICA-NUREG-0750 V23 NO2-NUCLEAR REGULATORY COMMISSION IS.

TION OF MECHANICAL AND DYNAMIC (INCLUDING SEISMIC)

)

SUANCES FOR FEBRUARY 1986. Pages49-111.

QUALIFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENT j

NUREG4750 V23 NO3: NUCLEAR REGULATORY COMMISSION IS-PROGRAM (EDOP).

SUANCES FOR MARCH 1986. Pages 113-232.

NUREG/CP-0077: PROCEEDINGS OF THE SEMINAR ON LEAK-NUREG4750 V2 N04: NUCLEAR REGULATORY COMMISSION IS.

BEFORE-BREAK: INTERNATONAL POUC1ES AND SUPPORTING J

SUANCES FOR APRIL 1986. Pages 233-464.

RESEARCH.

EDO. OFFICE OF INTERNATIONAL PROGRAMS INTRA-AGENCY COMMITTEES, REVIEW GROUPS, ETC.

OFFCE OF INTERNATONAL PROGRAMS, DIRECTOR LESSONS LEARNED GROUP NUREG-1180: INTERNATONAL COOPERATION DURING RADIO.

NUREG-1198: RELEASE OF UF6 FROM A RUPTURED MODEL 48Y LOGICAL EMERGENCIES. NRC Program Guidance For The Prov6 CYUNDER AT SEQUOYAH FUELS CORPORATION sion Of Technical Advice To Foreign Counterpart Orgarw2ations.

FACIUTY. Lessons-Leamed Report.

TEAM ON DAVIS-BESSE EVENT EDO. 0FFICE OF EXECUTIVE LEGAL DIRECTOR NUREG 1201: REPORT OF THE INDEPENDENT AD HOC GROUP OFFICE OF THE EXECUTIVE LEGAL DIRECTOR FOR THE DAVIS-BESSE INCIDENT.

NUREG4386 004 R01: UfitTED STATES NUCLEAR REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE DIGEST JULY EDO-RESOURCE MANAGEMENT 1.1972 SEPTEMBER 30.1985 DIVISION OF BUDGET & ANALYSIS NUREG4980 R02-NUCLEAR REGULATORY LEGISLATON.

NUREG-0020 V10 NO3: UCENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of February 28,1986.(Gray Book 1)

EDO OFFICE OF STATE PROGRAMS NUREG-0020 V10 N04: UCENSED OPERATING REACTORS STATUS OFFICE OF STATE PROGRAMS. DIRECTOR

SUMMARY

REPORT. Data As Of March 31.1986.(Gray Book f)

NUREG-1188: THE AUBURN STEEL COMPANY RADIOACTIVE CON-TAMINATON INCIDENT' OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL 8 2 AEOD, DIRECTOR'S OFFICE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR NUREG4090 V08 N04: REPORT TO CONGRESS ON ABNORMAL POWER PLANTS. LWR EditiortRevision 4 to Section 9.2.1, " Station OCCURRENCES October -December 1985.

Service Water System."

NUREG-0800 09.2.2 R3: STANDARD REVIEW PLAN FOR THE OFFICE OF INSPECTON & ENFORCEMENT (POST 12/11/80)

REVIEW OF SAFETY ANALYSIS REPORTS FCP. NUCLEAR DIRECTOR'S OFFCE. OFFCE OF INSPECTON AND ENFORCEMENT POWER PLANTS. LWR Edition. Revision 3 to Section 9 2.2. " Reactor NUREG4940 V05 N01: ENFORCEMENT ACTIONS SIGNIFICANT AC-Auxiliary Coohng Water Systems."

TONS RESOLVED.Quartetty Progress Report. January-March 1986.

NUREG-1177: SAFETY EVALUATION REPORT RELATED TO THE DIVISION OF OA, VENDOR & TECHNICAL TRAINING CENTER PRO.

RESTART OF DAVIS-BESSE NUCLEAR POWER STATION, UNIT GRAMS (POST 85021 1.FOLLOWING THE EVENT OF JUNE 9.1985. Docket No. 50 NUREG-0040 V10 N01: UCENSEE CONTRACTOR AND VENDOR IN-346 (Toledo Edison Company)

SPECTION STATUS REPORT. Quartetty Report. January 1986 DIVISION OF HUMAN FACTORS TECHNOLOGY (POST 851125)

March 1986 (White Book)

NUREG 0935 R02: U.S NUCLEAR REGULATORY COMMISSON HUMAN FACTORS PROGRAM PLAN.

OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG 1021 R02. ODERATOR UCENSING EXAMINER STAND-DIVISION OF WASTE MANAGEMENT ARDS.

NUREG-1183: NONRADIOLOGICAL GROUNDWATER OUAUTY AT NUREG-1192: AN INVESTIGATION OF THE CONTRIBUTORS TO LOW-LEVEL RADOACTIVE WASTE DISPOSAL SITES.

WRONG UNIT OR WRONG TRAIN EVENTS.

61

62 NRC Originating Organization Index OlvlSON OF PRESSURIZED WATER REACTOR LICENSING A

NUREG-1137 S02: SAFETY EVALUATON REPORT RELATED TO (POST 851125)

THE OPERATON OF VOGTLE ELECTRIC GENERATING NUREG4675 S33: SAFETY EVALUATON REPORT RELATED TO PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50-425. (Georgia THE OPERATION OF DIABLO CANYON NUCLEAR POWER. UNITS Power Company.et al) 1 AND 2. Docket Nos.50 275 and 50-323.(Paonc Gas And Electnc NUREG 1191: TECHNICAL SPECIFICATONS FOR CATAWBA NU-CLEAR STATON UNITS 1 AND 2. Docket Nos. 50-413 And 50-NL 1: SAFETY EVALUATON REPORT RELATED TO THE 414 (Duke Power Company)

OPERATION OF THE SOUTH TEXAS PROJECT, UNITS 1 AND DIVISON OF PRESSURIZED WATER REACTOR LICENSING - B 2 Docket Nos. 50-498 And 50499(Houston Ughting And Powe, (POST 851125)

NURE 4 7 S13: SAFETY EVALUATION REPORT RELATED TO NUREG-1181: TECHNICAL SPECIFICATONS FOR PALO VERDE NU.

THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC CLEAR GENERATING STATON, UNIT 2. Docket No. 50-529.(Arizona STATION. UNITS 1 AND 2. Docket Nos. 5(M45 And 50-446.(Texas Pubhc Service Company)

Ubbtes Generating ASSISTANT DIRECTOR FOR TECHNICAL SUPPORT (PWR-8)

NUREG4857 S10 SAF Y ALUATON REPORT RELATED TO NUREG-1166: DRAFT ENVIRONMENTAL STATEMENT FOR DECOM-THE OPERATION OF PALO VERDE NUCLEAR GENERATING MISSIONING HUMBCLDT DAY POWER PLANT, UNIT 3. Docket No.

STATION, UNITS 1,2 And 3 Docket Nos. 50-528,50-529 And 50-50-133.(Pacshc Gas And Electric Company) 530 (Arizona P@lic Sennce Company)

OtVISION OF BOluNG WATER REACTOR (BWR) LICENSING NUREG4896 SO4: SAFETY EVALUATION REPORT RELATED TO NUREG4979 S05. SAFETY EVALUATON REPORT RELATED TO THE OPERATION OF SEABROOK STATON, UNITS 1 AND 2. Docket THE DESIGN APPROVAL OF THE GESSAR 11 BWR/8 NUCLEAR Nos.50-443 And 50-444. (Pubhc Service Company of New ISLAND DESIGN. Docket No. 50-447. (General Electnc Company)

NUREG-1048 S05: SAFETY EVALUATON REPORT RELATED TO NUR

. SAFETY EVALUATON REPORT RELATED TO THE OPERATON OF HOPE CREEK GENERATING THE OPERATION OF CATAWBA NUCLEAR STATION, UNITS 1 STATON Docket No.50-354. (Pubhc Service Electnc And Gas AND 2. Docket Nos. 50-413 And 50-414. (Duke Power Company,et Company,Atlanbc City Electric Company).

,9 NURE31038 S03: SAFETY EVALUATION REPORT RELATED TO NUREG.1186. TECHNICAL SPECIFICATIONS FOR HOPE CREEK THE OPERATION OF SHEARON HARRIS NUCLEAR POWER GENERATING STATION. Docket No. 50-354.(Pubhc Sennce Electnc PLANT, UNIT 1. Docket No.50-400. (Caronna Power And Ught Com.

And Gas Company) pany And North Carohna Eastern Munecipal Power A

)

OfVISION OF SAFETY REVIEW & OVERSIGHT (POST 851125)

NUREG-1057 S01: SAFETY EVALUATION REPOR R LATED TO NUREG 1152: MILLSTONE 3 RISK EVALUATION REPT:AN OVERALL THE OPERATION OF BEAVER VALLEY POWER STATON, UNIT REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROB-

2. Docket No. 50-412.(Duquesne Light Company.et al)

ABluSTIC SAFETY STUDY.

6 l

l l

NRC Contract Sponsor Index (Contractor Reports)

This index lists the NRC organizations that sponsor organization is followed by the sponsored the contractor reaorts listed in NUREG/CR number and title of the this compilation. It is arranget alphabetically report (s) prepared by that organization. If by major NRC organization (e.g., program further information is needed, refer to the 4

office) and then by subsections of these main citation by the NUREG/CR number.

(e.g.,

divisions) where appropriate. The EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR-3262 V07: COBRA-NCA THERMAL HYDRAUUC CODE DATA FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR AEOD, DRECTOR'S OFFICE COMPONENT. Volume 7Amesesment Manuel for Contamment App 6 NUREG/CR2000 V05 N3. LCENSEE EVENT REPORT (LER) cations.

COMPILATION For Month Of March 1966.

NUREG/CR-3701: REMIX.A COMPUTER PROGRAM FOR TEMPERA-NUREG/CR2000 VOS N4: UCENSEE EVENT REPORT (LER)

TURE TRANSIENTS DUE TO HIGH PRESSURE INJECTION AFTER COMPILATON.For Month Of April 1966.

INTERRUPTON OF NATURAL CIRCULATION.

NUREG/CR2000 VOS N5: UCENSEE EVENT REPORT (LER)

NUREG/CR-3702 BUOYANCY EFFECTS IN OVERCOOUNG TRAN.

COMPILATON For Month Of May 1966.

SIENTS CALCULATED FOR THE NRC PRESSURIZED THERMAL SHOCK STUDY' IMPROVED MODEUNG AND NUMERICS TO yyg" NUREG/CR-3705:

EN PR PAREDN

& NG N RING RE-SOLVE TWO-OlMENSIONAL ELUPTIC FLUID FLOW AND HEAT SPONSE (POST 630t03)

NUFIEG/CR3662 A METHOD TO CHARACTERIZE LOCAL METEOR-TRANSFER PROBLEMS.

OLOGY AT NUCLEAR FACluTIES FOR APPUCATION TO EMER.

NUREG/CR-3970 TRAC.PF1/ MODI INDEPENDENT ASSESSMENT:

GENCY RESPONSE NEEDS LOBI INTERMEDIATE BREAK TEST B-R1M.

NUREG/CR3960- CLOSEOUT OF IE BULLETIN 60-01.Operatality Of NUREG/CR4027:

TRAC-PF1/ MOD 1 INDEPENDENT ASSESSMENT.Condeneata in Stratmed Cocurrent Flow.

Automatic Depressunration System (ADS) Valve Pneumanc Suppty.

NUREG/CR3962: CLOSEOUT OF IE BULLETIN 60-20.Fadures Of NUREG/CR4241: CHEMICAL ASPECTS OF CESIUM IODOE INTER-Westmghouse Type W-2 Spring Retum To Neutral Control Switches.

ACTION IN STEAM WITH 304 STAINLESS STEEL AND INCONEL-600.

OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDg NUREG/CR-4332: DESIGN AND FINAL SAFETY ANALYSIS REPORT DVISON OF FUEL CYCLE & MATERIAL SAFETY FOR VERTICAL FISSION PRODUCT RELEASE APPARATUS IN NUREG/CR4561 FIRAC USER S MANUAL:A COMPUTER CODE TO HOT CELL B. BUILDING 4501.

SIMULATE FIRE ACCIDENTS IN NUCLEAR FACluTIES.

NUREG/CR4353: ASSESSMENT OF POST 4RITICAL-HEAT FLUX DVISION OF SAFEGUARDS MODELS WITH LEHIGH NONEQUtuBRIUM DATA.

NUREG/CR4497: NRCPAGE APPUCATONS MANUAL NUREG/CR4364: BREAK SPECTRUM ANALYSIS FOR SMALL DVISION OF WASTE MANAGEMENT BREAK LOSSOF-COOLANT ACCOENTS IN A RESAR-35 PLANT, NUREG/CR 3572 V02: DETERMINATON OF METABOUC DATA AP.

NUREG/CR4402 V03: HIGH TEMPERATURE GAS 400 LED REAC.

PROPRIATE FOR HLW OOSIMETRY ll Gastrointestinal Absorption.

TOR SAFETY STUDIES FOR THE DIVISON OF ACCIDENT NUREG/CR3620 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR Snptember EVALUATON Quarterty Progrees Report. July 1 THE ONSITE DISPOSAL OF RADOACTIVE WASTES.The ONSITE/

30 1965 MAW Computer Program.

NUR'EG/CR4412: AN ASSESSMENT OF SAFETY MARGINS IN ZlR.

NUREG/CR4236 V03: PROGRESS IN EVALUATON OF RADIONU-CALOY OXIDATION AND EMBRITTLEMENT CRITERIA FOR ECCS CUDE GEOCHEMISTRY INFORMATON DEVELOPED BY DOE ACCEPTANCE HOH-LEVEL NUCLEAR WASTE REPOSITORY SITE NUREG/CR441il: LOSS OF CONTROL AIR AT BROW'13 FERRY PROJECTS Report For April-June 1965.

UNIT ONE. ACCIDENT SEQUENCE ANALYSIS NUREG/CR4496: FIELD TESTING OF WASTE FORMS CONTAINING NUREG/CR4453 V02-UGHT WATERREACTOR FUEL SAFETY SWEM pm M N.

he %se NUR C

1 EC i IDERAT S A FECTING CA M2 COOE ASSESSM AT WE NU E /CR EF TS A S O UNDER-GROUND FACluTIES Uterature Review And Docussion OAHO NATONAL EN%EERING MAM NUREG/CR4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT NUREG/CR4493: AN EXPERIMENTAL AND ANALYTICAL INVESTl-IN THE UNSATURATED ZONE. Information And Data Sets.

GATON OF OUENCHING OF SUPERHEATED DEBRIS BEDS NUREG/CR4619: STRESS CORROSION CRACKING TESTS ON UNDER TOP-REFLOOD CONDITIONS. Final Report HIGH LEVEL WASTE CONTAINER MATERIALS IN SIMULATED NUREG/CR4507: HECTR VERSON 1.5 USER'S MANUAL NUREG/CR4549 DETERMINATION OF APPENDIX K CONSERV-TUFF REPOSITORY ENVIRONMENTS'FOR NUREG/CR4620: METHODOLOGIES EVALUATING LONG-ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR T

ABil ATON DESIGNS OF URANIUM MILL TAluNGS IM-NL R 4 56 O

ONT MODEUNG FOR PWR THER-MAL HYDRAUUC ANALYSIS.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/01)

NUREG/CR4564: REDUCED PRESSURE AND FLUID TO FLUID OFFICE OF NUCLEAR REGULATORY RESEARCH, DtRECTOR SCAUNG LAWS FOR TWO PHASE FLOW LOOP.

NUREG/CR3770: PRELIMINARY DEVELOPMENT OF AN INTEGRAT-NUREG/CR4595: ENHANCEMENT TO THE LAFM COMPUTER ED APPROACH TO THE EVALUATION OF PRESSURIZED THER CODE.

MAL SHOCK AS APPUED TO THE OCONEE UNIT 1 NUCLEAR NUREG/CR-4621: FLOW VISUAUZATION EXPERIMENT ON HOT.

POWER PLANT W/TWO OVERSIZE DRAWINGS.

LEG U-8END TWOLPHASE NATURAL CIRCULATON PHENOM-DIVISON OF ACCIDENT EVALUATION (POST 840101)

ENA.

NUREG/CR2331 V05 N3. SAFETY RESEARCH PROGRAMS SPON.

OfVISON OF HEALTH, SITING & WASTE MANAGEMENT (PRE SORED BY OFFICE OF NUCLEAR REGULATORY 640429)

RESEARCH Ouarterty Progress Report. July September 1965.

NUREG/CR3150 SEISMICITY AND TECTONIC RELATIONSHIPS NUREG/CR3262 V01: COBRA.NC A THERMAL-HYDRAUUC CODE FOR UPPER GREAT LAKES PRECAMBRIAN SHIELD PROVINCE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR FINAL REPORT. July 1961 December 1962.

COMPONENTS Volume 1 Equations And Constitutrve Models.

DIVISION OF RISK ANALYSIS & OPERATONS (POST 840429)

NUREG/CR3262 V02 COBRA-NC A THERMAL HYDRAULIC CODE NUREG/CR4319 NUCRAC A CODE FOR THE ESTIMATION OF FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR ADVERSARY ACTION CONSEQUENCES IN THE NUCLEAR COMPONENTS Volume 2 COBRA NC Numencal Solution Methods.

POWER FUEL CYCLE.

63

64 NRC Contract Sponsor index NUREG/CR4330 V01: REVIEW OF UGHT WATER REACTOR REGU-NUREG/CR-3957: REUA81UTY ASSESSMENT AND PROBA81UTY LATORY REQUIREMENTS. Volume 1:ldentrlicaton Of Regulatory Re.

qurements That May Have importance To Riek BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS NUREG/CR4330 V02: REVIEW OF UGHT WATER REACTOR REGU-AND SHEAR WALLS. Summary Report.

LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGU-NUREG/CR-3965: AN INVESTIGATION OF THE STRENGTH OF H440 LATORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPOR-GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND TANCE TO RISK. Reactor Contamment Leekage Rates.Mem Steam SECONDARY STRESS.

19aistun Velve Leekage. -

NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCA-NUREG/CR-4463: HUMAN FACTORS IN ANNUNCIATOR / ALARM TIONS OF CONTROL AT THE OCONEE t NUCLEAR PLANT FINAL SYSTEMS. ANNUNCIATOR EXPERIMENT PLAN L REPORT.

NUREG/CR-4467: RELATIVE IMPORTANCE OF INDIVIDUAL ELE-NUREG/CR-4t24 V02 NDE OF STAINLESS STEEL AND ON-UNE MENTS TO REACTOR ACCIDENT CONSEQUENCES ASSUMING LEAK MONITORING OF LWRS. Annual Report October 1984. Sep.

EQUAL RELEASE FRACTIONS ^

tomber 1985 NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC.

NUREG/CR-4265 V01: AN ASSESSMENT OF THE SAFETY IMPUCA-TtVENESS OF AN OPERATIONAL SAFETY REUA8luTY PRO-TIONS OF CONTROL AT THE CALVERT CUFr0 t NUCLEAR NUREG 4

AN OPE A A SA TY ELIA IUTY PRO-NU R4290 V01: PROSA81UTY OF PtPE FAILURE IN THE RE.

GRAM APPROACH WITH RECOMMENDATIONS FOR FURTHER ACTOR COOLANT LOOPS & BABCOCK AND WM PWR DEVELOPMENT AND EVALUATION PLANTS. Volume 1: Summary Report.

NUREG/CR4516: INTERNATIONAL SAFEGUARDS AT FACluTIES NUREG/CR-4315 V01: EVALUATION OF NUCLEAR FACluTY DE.

EMPLOYING SPENT FUEL ROD CONSOUDATION.

NUREG/CR-4569: A REvtEW OF THE SEVERE ACCtDENT RISK RE.

COMMISSONING PROJECTS. Summary Status Report Three Mise letand Urut 2 Reactor Coolant System & Syelems Decontammeton.

DUCTON PROGRAM (SARRP) CONTAINMENT EVENT TREES NUREG/CR4315 V02: EVALUATION OF NUCLEAR FACluTY DE.

DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST 840429L COMMISSIONING PROJECTS. Summary Status Report Three Mile NUREG/GR-2675 V05: RELEVANCE OF 80 TIC PATHWAYS TO THE island urut 2 Reactor BuAsno Decontammaton.

LONG-TERM REGULATON OF NUCLEAR WASTE DISPOSAL (Est.

NUREG/CR43t5 V03: EVALUATION OF NUCLEAR FACluTY DE.

COMMISSIONING PROJECTS. Summary Stense Report Three Mise meton Of Radiaton Dooo To Men Resuleng From Bootc Trenaport The BOPORT/MAxit Sonwere Packaget tenend Urut 2 Reactor Deluelmg & F NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ILThe Sandle NUREG/CR4315 V09: EVALUATION OF NUCLEAR FACluTY DE-Weste-loolation Flow And Transport Model For Fractured Mede Re-COMMISSIONING PROJECTS Summary Status Report Three Mise lease 484.

Island Urut 2 Radioechve Weste And Laundry Shipments.

NUREG/CH-3441: RADONE.A COMPUTER CODE FOR SIMULATING NUREG/CR43t6: EVALUATION OF NUCLEAR FACluTY DECOM-FAST. TRANSIENT ONE-DIMENSIONAL HYDRODYNAMIC CONDI-MISSIONING PROJECTS STATUS REPORT. HUM 80LDT BAY TIONS AND TWCLLAYER RADONUCLIDE CONCENTRATIONS IN-POWER PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

CLUDING THE EFFECT OF BED-DEPOSITION IN CONTROLLED NUREG/CR4336: TELLURIUM BEHAVOR IN CONTAINMENT AlVERS AND TIDAL ESTUARIES.

UNDER UGHT WATER REACTOR ACCIDENT CONDITIONS.

NUREG/CR.3472 V02: SURFACE PROPERTIES AND PERFORM-NUREG/CR4349: LSL.M2:A COMPUTER PROGRAM FOR LEAST.

ANCE PREDICTON OF ALTERNATIVE WASTE FORMS. Final SOUARES LOGARITHMIC ADJUSTMENT OF NEUTRON SPECTRA.

NOREG/CR4449: A PWR HYBRID COMPUTER MODEL FOR AS.

NU CR-4309: VALENCE EFFECTS ON SOLUBluTY AND SESSING THE SAFETY IMPUCATIONS OF OONTROL SYSTEMS.

NUREG/CR4484: STATUS OF ACTIVlilES FOR INSPECTING WELD NU

/CR4 T

PE ORMANCE OF MATERI-OVERLAfD PIPE JOINTS.

ALS USED FOR HIGH-LEVEL WASTE PACKAGING. Annual NUREG/CR4489: HISTORICAL

SUMMARY

OF THE HEAVY-SEC-Raport, Year Four Apnl 1985. March 1986 TION STEEL TECHNOLOGY PROGRAM AND SOME RELATED AC.

NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOSE TIVITIES IN UGHT WATER REACTOR PRESSURE VESSEL REDUCTON RESEARCH PROJECTS.

SAFETY RESEARCH.

NUREG/CR-4511: ASSESSMENT OF THE ADEQUACY OF THE CAU.

NUREG/CR4490 V01: UGHT. WATER-REACTOR SAFETY MATERI.

BRATIONS PERFORMED BY COMMERCIAL Call 8 RATION SERV.

ALS ENGINEERING RESEARCH PROGRAMSOuarterty Progrees ICES FOR IONf/ING RADIAflON SURVEY INSTRUMENTS.

Report.Jenuary-March 1985.

NUREGICR4520: PREDICTIVE GEOCHEMICAL MODEUNG OF CON.

NUREG/CR-4503 V01: LONG TERM EMBRITTLEMENT OF CAST.

TAINMENT CONCENTRATIONS IN LABORATORY OOLUMNS AND IN PLUMES MtGRATING FROM URANIUM MILL TAIUNGS WASTE DUPLEX STAINLESS STEELS IN LWR SYS Annual Report, October 1984 September 1985.

IMPOUNOMENTS. Final Report NUREG/CR-4580 STONY BROOK SElSMIC NETWORK ON LONG NUREG/CR4519 TECHNOLOGY, SAFETY AND COSTS OF DECOM.

ISLAND,NEW YORK Feel Report (September 19T9 March 1985).

MISSIONING NUCLEAR FUEL CYCLE FACluTIES CLASSIFICA.

TEMPERATURE EFFECTS ON THE SOLUBluTY TON OF DECOMMISSIONING WASTE.

NUREG/CR4582:

AND SPECIATON OF SELECTED ACTINIDES NUREG/CR-4526: FINITE ELEMENT ANALYSIS OF THE 2240 MW NUREG/CR4807: THE EFFECTS OF ENVIRONMENT AND GAMMA HTOR PCRV.

1RRADIATION ON THE MECHANICAL PROPERTIES OF HIGH DEN-NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WELDED TYPE SITY POLYETHYLENE.

304 STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION NUREG/CR4632: NEW MADRID SEISMOTECTONIC PROGRAM Final AND UMIT LOAD ANALYSIS NUREG/CR4539: INVESTIGATON OF TEARING INSTA01UTY PHE.

NU CR4641: EXPERIMENTAL ASSESSMENT OF BOREHOLE NU

/

543 FIRS RE LTS FROM ELECTION-PHOTON NU E /CR S

IN PERIMENTAL AS.

DAMAGE EQUlVALENCE STUDIES ON A GENERIC ETHYLENE.

SESSMLNT OF BOREHOLE PLUG PERFORMANCE. Annual P

L E UBB.

NU C

OlviYtOF EhFI ECHNOLOGY CA8LE FAILURE WITH MATERIALS DEGRADATON.

NUREG/CR 3064 Vot: COMPUTATONAL METHODOLOGY FOR NUREG/CR4572 NRC LEAK BEFORE-BREAK (LG8 NRC) ANALYS4S OAK RIDGE RESEARCH REACTOR (OnR) AND BULK SHIELDING METHOO FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED REACTOR (BSR)Crose-Section Generation And Validation, Volume PIPES UNDER AXIAL PLUS BENDING LOADS. Topical Report g

NUREG/CR-4579 APPUCATON OF THE NEY CURVE AND MULTI-NUREGICR 3064 V02. COMPUTATONAL METHODOLOGY FOR SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF OAK RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING ALLOY STEEL REACTOR (DSR)The VICTOR't input Proceemng Code For The NUREG/CH 4583 V01: DEVELOPMENT AND VAUDATION OF A Bokt Venture System. Volume IL REAL. TIME SAFE UT SYS1EM FOR THE INSPECTION OF UGHT NUREG/CH-3228 V04 STRUCTURAL INTEGRITY OF WATER REAC.

WATER COMPONENTS Some. Annual Report For Apnl 1984 Septem.

TOR PRESSURE BOUNDARY COMPONENTS Annual Report for ber 1984 19n5 NUREG/CR 3453 ELECTRONIC ISOLATERS USED IN SAFETY SYS.

NUREG/CR4$88 V01: SOIL-STRUCTURE INTERACTON Vol t Influ-TI MS OF U S NUCLE AR POWFR PL ANTS ence of Leyenng NUREG/CR 3587: IDENTIFICAflON AND EVALUATON OF FACluTA.

NUREG/CR4588 V02: SOIL. STRUCTURE INTERACTON Vol finflu-ence Of trit Off-TON TECHNOUES FOR DECOMMISSONING LIGHT WATER POWtR REACTORS NUREG/CR4588 V03 SOfL. STRUCTURE INTERACTON Vol S influ-ence Of Ground Water.

NRC Contract Sponsor Index 65 NURE3/CR-4802: UNIOUENESS OF BOiUNG WATER REACTOR NUREG/CR4488: VENTING OF NONCONDENSIBLE GAS FROM PRIMARY WATER CHEMISTRY. Fool Report October 1985. March THE UPPER HEAD OF A B4W REACTOR VESSEL US;NG HOT 1988.

LEG U-BEND VALVES.

NUREG/CR-4831: ELECTROMAGNETIC ACOUSTC TRANSDUCER DIVISION OF PRESSURIZED WATER REACTOR LICENSING 8 (EMAT) DEFECT CHARACTER 12ATION OF NUCLEAR REACTOR (POST 851125)

PtPtNG WELDS. Phase 1 Final Report,0ctober 1985 - March 1988.

NUREG/CR 4404: ANALYSIS OF ALLOWED OUTAGE TIMES AT NUREG/CR4834: DEVELOPMENT OF A REAL TIME RESIDUE BYRON GENERATING STATION.

NUMBER PROCESSOR FOR SAFT INSPECTION. Phase 11 Final NUREG/CR4803: APPRAISING ATMOSPHERIC TRANSPORT AND r": ' ' ' ~ - - tgg4.Apr81988' DIFFUSION MODELS FOR EMERGENCY RESPONSE FACluTIES.

a DIVISION OF SAFETY REVIEW & OVERSIGHT (POST 851125)

EDCHIESOURCE MA8844EMElff NUREG/CR4048: A METHODOLOGY FOR ALLOCATING RELIABIL.

OFFICE OF RESOURCE MANAGEMENT, DIRECTOR ITY AND RISK.

NUREG/CR-4827: GENERC COST ESTlWATES. Abstracts From Ge.

NUREG/CR4281: ASSESSMENT OF SYSTEM INTERACTION EXPE-nonc Studes For Use in Propenng Regulatory impact Analysee RIENCE IN NUCLEAR POWER PLANTS.

DIVISION OF BUDGET & ANALYSIS NUREG/CR4374 V03: A REV.EW OF THE OCONEE-3 PRO 9A01US.

NUREG/CR-2850 V04: POPULATION DOSE COMMITMENT DUE TO TIC RISK AMESSMENT CONTAINMENT RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES PERFORMANCE. RADIOLOGICAL SOURCE TERMS AND RISK ES-IN 1982.

TIMATES.

NUREG/CR4483: REACTOR PRESSURE VESSEL FAILURE PROBA.

l OFFICE OF NUCLEAR REACTOR REOULATION (POST 4/30/80)

BluTY FOLLOWING THROUGH-WALL CRACKS DUE TO PRES.

OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST SURIZED THERMAL SHOCK EVENTS.

851125)

NUREG/CH4557: A REVIEW OF ISSUES RELATED TO IMPROVING NUREG/CR4142: A REVIEW OF THE MILLSTONE 3 PROBA01USTC NUCLEAR POWER PLANT DIESEL GENERATOR REUABluTY.

SAFETY STUDY.

NunEG/CR 4588: A HANDBOOK FOR QUICK ESTIMATES.A Method NUREG/CR4207: FAULT TREE APPLICATION TO THE STUDY OF For Developog Quick Approxrnate Estrnates Of Costa For Genenc SYSTEMS INTERACTIONS AT IN0 TAN POINT 3 Actions For Nuclear Power Pter.ts.

DIVISION OF PRESSURIZED WATER REACTOR UCENSING. A NUREG/CR4589: REVIEW OF SELECTED AREAS OF YANKEE NU(POST 851125)

ROWE PROBABlUSTC SAFETY STUDY.

REG /CR-4481: TORNADO CUMATOLOGY OF THE CONTIGUOUS NUREG/CR 4594: ESTIMATED SAFETY SIGNIFICANCE OF GENER.

UNITED STATES.

IC SAFETYISSUE81.

l

\\

r--m---

_---__________..,__-m_,,,_,

v

Contractor Index This index lists, in alphabetical order, the numbers and titles of their reports. If further contractors that prepared the NUREG/CR information is needed, refer to the main ci-reports listed in this compilation. Listed tation by the NUREG/CR number.

below each contractor are the NUREG/CR ANALYSIS & TECHNOLOGY,INC.

NUREG/CR-2850 V04: POPULATON DOSE COMMITMENT DUE TO NUREG/CR4207: FAULT TREE APPUCATON TO THE STUDY OF RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES SYSTEMS INTERACTIONS AT INDIAN POINT 3.

IN 1982.

NUREG/CR-3282 V01. COBRA-NCA THERMAL-HYDRAUUC CODE APOSTOLAKIS, G.

FOR TRANGlENT ANALYSIS OF NUCLEAR REACTOR NUREG/CR4142: A REVIEW OF THE MILLSTONE 3 PROBABIUSTIC COMPONENTS. Volume 1. Equations And Constitutive Models.

SAFETY STUDY.

NUREG/CR-3282 V02: COBRA-NC.A THERMAL HYDRAUUC CODE FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR NO M TECHNMY CMP.

COMPONENTS. Volume 2: COBRA-NC Numerical Solution Methods.

NUREG/CR-4142: A FEVIEW OF THE MILLSTONE 3 PROBABluSTIC NUREG/CR-3262 V07: COBRA.NC A THERMAL-HYORAUUC CODE SAFETY STUDY.

FOR TRANSIENT ANALYSIS OF NUCLEAR REACTOR Ananms NATIONAL LA80RATORY COMPONENT. Volume 7:Assosoment Manual for Containment Applica-NUREG/CR-4124 V02: NDE OF STAINLESS STEEL AND ON-UNE tions.

NUREG/CR-3820 S01: INTRUDER DOSE PATHWAY ANALYSIS FOR LEAK MONITORING OF LWRS. Annual Report October 1984 - Sep-THE ONSITE DISPOSAL OF RADIOACTIVE WASTES.The ONSITE/

tomber 1985.

NUREG/CR-4453 VC2: LIGHT WATER-REACTOR FUEL SAFETY SYS.

MAX 11 Computer Progrant TEMS RESEARCH PROGRAMS. Quarterty Progress Report.Aprildune NJREG/CR-3882: A METHOD TO CHARACTERIZE LOCAL METEOR-1985.

OLOGY AT NUCLEAR FACiUTIES FOR APPUCATON TO EMER-NUREG/CR4490 V01: LIGHT-WATER-REACTOR SAFETY MATERIALS GENCY RESPONSE NEEDS.

t ENGINEERING RESEARCH PROGRAMS:Quarterty Progress NUREG/CR4330 V01: REVIEW OF UGHT WATER REACTOR REGO-Report. January. March 1985.

LATORY REQUIREMENTS. Volume 1:ldentrfication Of Regulatory Re-NUREG/CR4503 V01: LONG TERM EMBRITTLEMENT OF CAST.

quirements That May Have importance To Risk DUPLEX STAINLESS STEELS IN LWR SYS. Annual Report,0ctober NUREG/CR4330 V02: REVIEW OF UGHT WATER REACTOR REGU-1984 September 1985.

LATORY REQUIREMENTS ASSESSMENT OF SELECTED REGULA-NUREG/CR-4505: A SCOPING STUDY OF THE POTENTIAL EFFEC-TORY REQUIREMENTS THAT MAY HAVE MARGINAL IMPORTANCE TIVENESS OF AN OPERATIONAL SAFETY REUA8luTY PROGRAM TO DISK. Reactor Containment Leekage Rates. Main Steam isolation IN ADDRESSING GENERIC SAFETY PROBLEMS.

Valve Leakage.

NUREG/CR4508 AN OPERATIONAL SAFETY REUABluTY PROGRAM NUREG/CR-4412: AN ASSESSMENT OF SAFETY MARGINS IN ZlRCA-APPROACH WITH RECOMMENDATIONS FOR FURTHER DEVELOP-LOY OXIDATON AND EMBRITTLEMENT CRITERIA FOR ECCS AC-MENT AND EVALUATION.

CEPTANCE.

NUREG/CR-4511: ASSESSMENT OF THE ADEQUACY OF THE CAU.

NUREG/CR-4461: TORNADO CUMATOLOGY OF THE CONTIGUOUS BRATIONS PERFORMED BY COMMERCIAL Call 8 RATION SERV-UNITED STATES.

ICES FOR IONIZING RADIATION SURVEY INSTRUMENTS.

NUREG/CR4483: RLACTOR PRESSURE VESSEL FAILURE PROBA-NUREG/CR 4568: A HAN0000K FOR QUICK ESTIMATES.A Method BluTY FOLLOWING THROUGH WALL CRACKS DUE TO PRESSUR-For Developing Oulen Apprommate Estimates Of Costs For Generte Ac-IZED THERMAL Sh0CK EVENTS tions For Nuclear Power Plards.

NUREG/CR4484: STATUS OF ACTIVITIES FOR INSPECTING WELD NUREG/CR4584: REDUCED PRESSURE AND FLUID TO FLUID SCAL-OVERLAID PtPE JOINTS.

ING LAWS FOR TWO PHASE FLOW LOOP-NUREG/CR4518: INTERNATIONAL SAFEGUARDS AT FACILITIES EM-NUREG/CR-4621: FLOW VISUAUZATION EXPERIMENT ON HOT-LEG PLOYING SPENT FUEL ROD CONSOLIDATON.

U-8END TWO-PHASE NATURAL ORCULATON PHENOMENA.

NUREG/CR4519: TECHNOLOGY, SAFETY AND COSTS OF DECOM-MISSIONING NUCLEAR FUEL CYCLE FACluTIES CLASSIFICATION ARIZONA, UMV. OF, N AZ

""L NU E E GEOCHEHICAL MODEUNG OF CON.

WALL DA GE N SAL RO S TAINMENT CONCENTRATIONS IN LABORATORY COLUMNS AND IN NUREG/CR4642: ROCK MASS SEALING - EXPERIMENTAL ASSESS-PLUMES MIGRATING FRJM USANIUM MILL TA:UNGS WASTE IM-MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1984. May 1985' POUNDMENTS. Final Repurt.

NUREG/CR-4583 V31: DEVELOPMENT AND VAUDATON OF A REAL-ASPEN SYSTEMS, INC.

TIME SAFE UT SYSTEM FOR THE INSPECTION OF UGHT WATER NUREG4388 004 R01: UNITED STATES NUCLEAR REGULATORY COMPONENTS. Semi-Annual F.eport For Apnl 19844eptember 1984.

COMMISSION STAFF PRACTICE AND PROCEDURE DGEST JULY SROOKHAVEN NATIONAL LASORATORY 1.1972. SEPTEMBER 30,1985.

NUREG/CR-2331 V05 N3: SAFETY RESEARCH PROGRAMS SPON-BATTELLE MEMORIAL INSTITUTE. COLUMOUS LA80RATORIES SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CP4077: PROCEEDINGS OF THE SEMINAR ON LEAK.

RESEARCH Ouarterty Progress Report. July-September 1985.

BEFORE-BREAK: INTERNATIONAL POUCtES AND SUPPORTING NUREG/CR-3705: IMPROVED MODEUNG AND NUMERICS TO SOLVE RESEARCH.

TWO-DIMENSIONAL ELUPTIC FLUID FLOW AND HEAT TRANSFER NUREG/CR4379 V04: LONG-TERM PERFORMANCE OF MATERIALS PROBLEMS.

USED FOR HIGH-LEVEL WASTE PACKAGING Arweal Report, Year NUREG/CR-3957: REUAB UTY ASSESSMENT AND PROBABILITY Four. April 1985 March 1988.

BASED DESIGN OF REINFORCED CONCRETE CONTAINMENTS NUREG/u14572-NRC LEAK-BEFORE-BREAK (LBB.NRC) ANALYSIS AND SHEAR WALLS.

Report METHOO FOR CIRCUMFERENTIALLY THROUGH-WALL CRACKED NUREG/CR-4048: A MET LOGY FOR ALLOCATING REUA81UTY PIPES UNDER AXIAL PLUS BENDING LOADS. Topical Report AND RISK.

NUREG/CR4207: FAULT TREE APPUCATION TO THE STUDY OF SATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST SYSTEMS INTERACTONS AT INDIAN POINT 3.

LASOMATORIES NUREG/CR4319: NUCRAC. A CODE FOR THE ESTIMATON OF AD-NUREG/CR-2875 VOS: RELEVANCE OF BOTIC PATHWAYS TO THE VERSARY-ACTON CONSEQUENCES IN THE NUCLEAR POWER LONG TERM REGULATION OF NUCLEAR WASTE DISPOSAL (Esti-FUEL CYCLE.

mation Of Radacon Dose To Man Resulting From Beotic TransportThe NUREG/CR-4374 V03: A REVIEW OF THE OCONEE-3 PROBA81USTIC BOPORT/ MAX 11 Software Packa0e).

RISK ASSESSMENT CONTAINMENT 67

68 Contractor index PERFORMANCE.RADIOLOGK/L SOURGE TERMS AND RISK ESTI-LAWIENCE BERKELEY LASORATORY MATES.

NUREG/CR-45m2 iMPERATURE EFFECTS ON ThE SOLUBluTY NUREG/CR-4404: ANALYSIS OF ALLOWED OUTAGE TIMES AT ann muATION OF SELECTED ACTINCES.

BYRON GENERATING STATION.

NUREG/CR-4409: DATA BASE ON NUCLEAR POWER PLANT DOCE LAWRENCE LIVERMORE NATIONAL LA50RATORY REDUCTION RESEARCH PROJECTS.

NW1EG/CR-4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC NUREG/CR4493: AN EXPERIMENTAL AND ANALYTICAL INVESTIGA-SAFETY STUDY.

TION OF QUENCHING OF SUPERHEATED DEBRIS BEDS UNDER NUREG/CR-4290 VOI: PROBABluTY OF PIPE FAILURE IN THE REAC.

TOP-REFLOOD CONDITIONS Final Report.

TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR NUREG/CR-4549: DETERMINATION OF APPENDIX K CONSERV-PLANTS. Volume 1: Summary Report.

ATISMS FOR LARGE BREAK LOCA IN WESTINGHOUSE PWR NUREG/CR-4009: EFFECTS OF EARTH 0UAKES ON UNDERGROUND USING TRAC-PD2/ MOD 1 CODE.

FACIUT'ES.Utcreture Rev6ew And Dw==n.

NUREG/CR 4557: A REVIEW OF ISSUES RELATED TO IMPROVING NUCLEAR POWER PLANT DIESEL GENERATOR REUABluTY.

LEHIGH UNIV., PETHLEHEM, PA NUREG/CR-4588 V01: SOIL-STRUCTURE INTERACTION.Vol 1;lnflu-NUREG/OR-43E3: ASSESSMENT OF POST-CRITICAL HEAT FLUX ence Of Layenng.

MODELS WITH LEHIGH NONEQUluBRIUM DATA.

NUREG/CR-4588 V02-SOIL STHUCTURE INTERACTION.Vol 2:Influ-ence Of Uft-Off.

LOS ALAMOS SCIENTIFIC LABORATORY NUREG/CR4588 V03: SOIL-STRUCTURE INTERACTION Vol 3.Influ-NUREG/CR-3965: AN INVESTIGATION OF THE STRENGTH OF H440 ence Of Ground Water.

GRAPHITE WHEN SUBJECTED TO COMBINED PRIMARY AND SEC-NUREG/CR-4589 REVIEW OF SELECTED AREAS OF YANKEE ROWE ONDARY STRESS.

PROBABlUSTIC SAFETY STUDY.

NUREG/CR-4497: NRCPAGE APPLICATIONS MANUAL NUREG/CR4594: ESTIMATED SAFETY SIGNIFICANCE OF GENEHIC NUREG/CR-4526: FINITE ELEMENT ANALYSIS OF THE 2240 MW SAFETYISSUE6t.

HTGR PCRV NUREG/CR4601. TECHNICAL CONSIDERATIONO AFFECTING PREP

NUREG/CR45' 1: FIRAC USER'S MANUAL:A COMPUTER CODE TO 6

ARATION OF IO4 EXCHANGE RESINS FOR DISPOSAL SIMULATE FIRE ACCOENTS IN NUCLEAR FACluTIES.

NUREG/CR4607: THE EFFECTS OF ENVIRONMENT AND GAMMA IR-NUREG/CR4595: ENHANCEMENT TO THE LAFM COMPUTER CODE.

RADIATION ON THE MECHANICAL PHOPERTIES OF Ht3H DENSITY NUREG/CR-4615 V01: MODEUNG STUDY OF SOLUTE TRANSPORT POLYETHYLENE-IN THE UNSATURATED ZONE. Information And Date Sets.

NUREG/CR4619: STRESS CORROSON CRACKING TESTS ON HIGH-LEVEL-WASTE CONTAINER MATERIALS IN SIMULATED TUFF RE-naamm arHUSETTS INSTITUTE OF TECHNOLOGY, CAMERIDGE, MA POSITORY ENVIRONMENTS.

NUREG/CR-4561: DRYOUT FRONT MODEUNG FOR PWR THERMAL COLORADO STAT'. UNIV., FT. COLLINS, CO NUREn/CR-4620: METHODOLOGCS FOR EVALUATING LONG-TERM MATERIALS ENGINEERING ASSOCIATES,INC.

STABluZATION DEEIGNS OF URANIUM MILL TAIUNGS IMPOUND NUREG/CP4067 V01: PROCEEDINGS OF THE SECOND IAEA SPE.

MENTS.

CIAUSTS' MEETING ON SUBCRITICAL CRACK GROWTH.Seessons 1 And ll. Held At sendai, Japan.Mey 15-17,1985.

COMMERCE DEPT. OF, NATIONAL OCEANIC & ATMOSPHERIC NUREG/CP4067 V02 PROCEEDINGS OF THE SECOND IAEA SPE.

ADMNIISTRATION CIAUSTS MEETING ON SUBCRITICAL CRACK GROWTH, Sessions ill NUREG/CR4603: APPRAISING ATMOSPHERIC TRANSPORT AND

& IV Held At Sendal, Japan.Mey 15-17,1985.

DIFFUSION MODELS FOR EMERGENCY RESPONSE FACluTIES.

NUREG/CR-3228 V04: STRUCTURAL INTEGRITY OF WATER REAC-DAVID W. TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER NUREG/CR-4538 V01: FRACTURE ANALYSIS OF WELDED TYPE 304 MATHTECH, INC.

STAINLESS STEEL PIPE: J-R CURVE CHARACTERIZATION AND NUREG/CR4627: GENERIC COST ESTIMATES. Abstracts From Generic NU E /CR 59i TIGATION OF TEARING INSTABluTY PHE-NOMENA IN ASTM A106 STEEL MINNESOTA, UNIV. OF, MINNEAPOLIS, MN NUREG/CR4579: APPUCATON OF THE KEY CURVE AND MULTI-NUREG/CR-3150: SEISMICITY AND TECTONIC RELATIONSHIPS FOR SPECIMEN TECHNIOUES TO DYNAMIC J-R CURVE TESTING OF UPPER GREAT LAKES PRECAMBRIAN SHIELD PROVINCE FINAL ALLOY SitEL REPORT July 1961 -December 1962.

DAVlS, P.

NEW YORK, STATE OF NUREG/CR-4142: A REVIEW OF THE MILLSTONE 3 PROBABIUSTIC NUREG-1188: THE AUBURN STEEL COMPANY RADOACTIVE CON-SAFETY STUDY.

TAMINATION INCIDENT.

EG&G IDAHO. INC. (SUSS. OF EG4G, INC.)

NEW YORK, STATE UNIV. OF STONY OROOK. NY NUREG/CR-3453: ELECTRONIC ISOLATERS USED IN SAFET( SYS-NUREG/CR-4580 STONY BROOK SEISMC NETWORK ON LONG ISLAND NEW YORK. Final Report (September 1979. March 1965).

NU E /

S RU A YSIS FOR SMALL BREAK LOSSOF COOLANT ACCIDENTS IN A RESAR-3S PLANT.

OAK RfDGE NATIONAL LA90RATORY NUREG/CR-4454: RELAP5/ MOD 2 CODE ASSESSMENT AT THE NUREG/CR-2000 V05 N3: UCENSEE EVENT REPORT (LER)

IDAHO NATONAL ENGINEERING LABORATORY.

COMPILATON.For Month Of Merch 1986.

NUREG/CR-4488: VENTING OF NONCONDENSIBLE GAS FROM THE NUREG/CR-2000 V05 N4: UCENSEE EVENT REPORT (LER)

UPPER HEAD OF A B&W REACTOR VESSEL USING HOT LEG U-COMPILATON For Month Of April 1986.

BEND VALVES NUREG/CR-2000 V05 N5: UCENSEE EVENT REPORT (LER)

NUREG/CR-4498: FIELD TESTING OF WASTE FORMS CONTAINING EPICOR-li ION EXCHANGE RESINS USING LYSIMETERS.

NU EG

-3064 1 COM T

L METHOOOLOGY FOR OAK RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-FLORIDA, UNIV. OF, GAINESVILLE, FL NUREG/CR 3472 V02-SURFACE PROPERTIES AND PERFORMANCE TOR (BSR)Crose-Section Generation And Valide6on, Volume 1.

NUREG/CR-3064 V02: COMPUTATONAL METHODOLOGY FOR OAK PREDICTON OF ALTERNATIVE WASTE FORMS Final Report RIDGE RESEARCH REACTOR (ORR) AND BULK SHIELDING REAC-FOX CONSULTING ENGINEERS & GEOLOGISTS TOR (BSR):The VICTORR input Process 6ng Code For The Bold Ven-NUREG/CR-4602-UNIOUENESS OF BOfuNG WATER REACTOR PRI.

ture System. Volume 11.

MARY WATER CHEMISTRY Final Report. October 1985 - March 1986.

NUREG/CR-3441: RADONE:A COMPUTER CODE FOR SIMULATING FAST.TRANS,ENT ONE-DIMENSONAL HYDRODYNAMIC CONDI-IMPELL CORP.

TONS AND TWO LAYER RADIONUCUDE CONCENTRATONS IN-NUREG/CR4207; FAULT TREE APPLICATON TO THE STUDY OF CLUDING THE EFFECT OF BED-DEPOSITON IN CONTROLLED SYSTEMS INTERACTONS AT INDIAN POINT 3.

RIVERS AND TIDAL ESTUARIES.

NUREG,CR-3572 V02: DETERMINATION OF METABOUC DATA AP.

JACK R. SENJAMIN & ASSOCIATES,INC.

PROPRIATE FOR HLW DOSIMETRY.II.Gestromtestmel Absorpeon.

NUHEG/CR4142: A REVIEW OF THE MILLSTONE 3 PROBABluSTIC NUREG/CR-3770: PREUMINARY DEVELOPMENT OF AN INTEGRAT-SAFETY STUDY.

ED APPROACH TO THE EVALUATON OF PRESSURIZED THERMAL

Contractor Index 69 SHOCK AS APPUED TO THE O(X) NEE UNIT 1 NUCLEAR POWER NUREG/CR4463: HUMAN FACTORS IN ANNUNCIATOMALARM PLANT.W/TWO OVERSIZE DRAWINGS.

SYSTEMS.ANNUNOATOR EXPERIMENT PLAN L NUREG/CR-4047: AN ASSESSMENT OF THE SAFETY IMPUCATIONS NUREG/CR-4467: RELATIVE IMPORTANCE OF INDIVIDUAL CLE-OF CONTROL AT THE OCONEE 1 NUCLEAR PLANT FINAL MENTS TO REACTOR ACODENT CONSEQUENCES ASSUMh/1 REPORT.

EQU AL RELEASE FRACTIONS.

NUREG/CR-4236 V03: PROGRESS IN EVALUATION OF RADONU-NUREG/CR4507: HECTR VERSON 1.5 USER'S MANUAL CUDE GEOCHEMISTRY INFORMATION DEVELOPED BY DOE HIGH-NUREG/CR4543: FIRST RESULTS FROM ELECTION-PHOTON LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For DAMAGE EQUIVALENCE STUDIES ON A GENERO ETHYLENE-PRO-Apni-June 1985.

PYLENE RUBBER.

NUREG/CR-4261: ASSESSMENT OF SYSTEM INTERACTON EXPERI-NUREG/CR-4546. CORRELATION OF ELECTRICAL REACTOR CA8LE ENCE IN NUCLEAR POWER PLANTS.

FAILURE WITH MATERIALS DEGRADATION.

NUREG/CR-4265 V01: AN ASSESSMENT OF THE SAFETY IMPUCA.

TIONS OF CONTROL AT THE CALVERT CUFFS 1 NUCLEAR PLANT.

SCPENCE & ENGINEERING ASSOCIATES,INC.

NUREG/CR 430.: VALENCE EFFECTS ON SOLUBluTY AND NUREG/CR-4627: GENERIC COST ESTIMATES. Abstracts From Gener6c NU G CR-4 NA SAFETY ANALYSIS REPORT FOR VERTICAL FISSON PRODUCT RELEASE APPARATUS IN HOT SCIENCE APPUCATIONS INTERNATIONAL CORP. (PORMERLY CELL B, BUILDING 4501.

SCIENCE APPUCATIONS, NUREG/CR-4338: TELLURIUM BEHAVIOR IN CONTAINMENT UNDER NUREG/CR-4319: NUCRAC. A CODE FOR THE ESTIMATION OF AD.

LIGHT WATER REACTOR ACCIDENT CONDITIONS.

VERSARY ACTION CONSEQUENCES IN THE NUCLEAR POWER NUREG/CR 4349: LSL-M2:A COMPUTER PROGRAM FOR LEAST.

FUEL CYCLE.

SOUARES LOGARITHMO AIMUSTMENT OF NEUTRON SPECTRA.

NUREG/CR4402 V03: HIGH-TEMPERATUR5 GASCOOLED REACTOR SIGMA RESEARCH, NIC.

SAFETY STUDIES FOR THE DIVISON OF ACCOENT NUREG/CR-4631: ELECTROMAGNETO ACOUSTIC TRANSDUCER EVALUATON.Ouart one Report. July 1. September 10.1985-(EMAT) DEFECT CHARACTERIZATON OF NUCLEAR REACTOR j

NUREG/CR-4413: L OF TROL AIR AT BROWNS FERRY UNIT ONE. ACCIDENT SEQUENCE ANALYSIS-PIPING WELDS. Phase i Feel Report,0ctober 1985. March 1986.

NUREG/CR-4634: DEVELOPMFNT OF A REAL. TIME RESIDUE l

NUPEG/CH-4449: A PWR HYBRID COMPUTER MODEL FOR ASSESS-i ING THE SAFETY IMPUCATIONS OF CONTROL SYSTEMS.

NUMBER PROCESSOR FOR SAFT INSPECTONPhase 11 Final NUREG/CR4489: HISTORICAL

SUMMARY

OF THE HEAVY SECTION ReporLSeptember M. @ 1966.

STEEL TECHNOLOGY PROGRAM AND SOME RELATED ACTIVITIES IN UGHT-WATER REACTOR PRESSURE VESSEL SAFETY RE-ST. LOUIS UNIV, ST. LOUIS, MO NUREG/CR-4632: NEW MADRID SEISMOTECTONIC PROGRAM. Final SEARCH.

NUREG/CR-4620- METHODOLOGIES FOR EVALUATING LONG TERM R'PU'L STABlu2ATION DESIGNS OF URANIUM MILL TAluNGS IMPOUND-TLG ENGINEERING, INC.

MENTS-NUREG/CR-3567: IDENTIFICATION AND EVALUATION OF FACluTA.

PARAMETER, NIC.

TION TECHNIOUES FOR DECOMMISSIONING UGHT WATER NUREG/CR-3960: CLOSEOUT OF IE BULLETIN 6001.Operatnlity Of POWER REACTORS.

U.S. NAVAL ACADEMY, ANNAPOUS, MD NU E CR 3 2 SEO E

IN 80-a es Wes-tinghouse Type W-2 Spring Retum To Neutral Control Switches.

NUREG/CR-4579: APPUCATION OF THE KEY CURVE AND MULil-SPECIMEN TECHNIQUES TO DYNAMIC J-R CURVE TESTING OF PUROUE UllIV WEST LAFAYETTE. IN ALLOY STEEL NUREG/CR-3701: REMIX.A COMPUTER PROGRAM FOR TEMPERA.

INTERRUPT N TURAL i UR CR 3 EVALUATION OF NUCLEAR FACluTY DECOM-NUREG/CR-3702: BUOYANCY EFFECTS IN OVERCOOUNG TRAN.

MISSIONING PRWECTS. Summary Status Report Three Mde Island S

L LATED FOR THE NRC PRESSURIZED THERMAL U

h R

C 5

DECOM-MISSIONING PROJECTS. Summary Status Report Three Mde Island S. COHEN & ASSOCIATES, INC.

Urut 2 Reactor Budeng Decontamination.

NUREG/CR-4627: GENERO COST ESTIMATES. Abstracts From Genenc NUREG/CR-4315 V03: EVALUATION OF NUCLEAR FACluTY DECOM-Sludes For Use in Propenng Regulatory impact Analyses.

MISSIONINf3 PRGlECTS. Summary Status Report Three Mde Island Unit 2 Reactor Defueling & N : EVALUATION O[1 NUCLEAR FACluTY DECO G

SANOIA NATIONAL LASORATORIES NUREG/CR4315 V09:

NUREG/CR-3162: DATA INPUT GUIDE FOR SWIFT ll.The Sanda MISSONING PROJECTS. Summary Status Report Three MWe leland Weste-Isolanon Flow And Transport Model For Fractured Mede Re.

Urut 2 Radoactrve Weste And Laundry Stupments.

lease 4 84_

NUREG/CR4316 EVALUATON OF NUCLEAR FACluTY DECOMMIS-NUREG/CR-3970- TRAC PF1/ MOD 1 INDEPENDENT ASSESSMENT:

SONING PRCUECTS STATUS REPORT. HUMBOLDT BAY POWER LOBIINTERMEDIATE BREAK TEST B-R1M.

PLANT UNIT 3 SAFSTOR DECOMMISSIONING.

NUREGICR4027:

TRAC.PF1/ MOD 1 INDEPENDENT ASSESSMENT Condensation in Strabfied Cocurrent Flow WISCONSIN, UNIV. OF, MAOISON, WI NUREG/CR4241: CHEMICAL ASPECTS OF CESIUM ODIDE INTERAC.

NUREG/CR-4569: A REVIEW OF THE SEVERE ACCCENT RISK RE.

TION IN STEAM WITH 304 STAINLESS STEEL AND INCONEL400.

DUCTON PROGRAM (SARAP) CONTAINMENT EVENT TREES.

'4-,m.,,

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I I

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i

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i I

l l

l

Licensed Facility Index This index lists the facilities that were the Docket number and followed by the report subject of NRC staff or contractor reports.

number. If further information is needed, The facility names are arranged in alphabet-refer to the main citation by the NUREG ical order. They are preceded by their number.

54424 Ainn W. Vogne Nudeer Plant, Urut 1, Georps MJREG1137 $02 54206 Inden Pomt Staton, Und 3, Power Aumorty of NUREG/CR4207 Power Ca Siste of New York S425 Ahm W. Vogne Nudeer Plert, Une 2. George NUREG1137 S02 444027 KerrascGee Nudeer Corp., Okiehoms Oly OK, NUREG1179 V02 Power Ca 444027 KerviscGee Nudeer Corp., Ohlehome Oly, OK, NUREG1198 S412 Beever Vasey Power Staton, Urul 2 Ducksene NURE41057 $01 50 423 Mdletone Nudeer Power Stenon, Una 3.

NUREG1152 Lgt Ca Normonal Nudeer Co 54250 Browns Ferry Nudeer Power Staton Una 1, NUREG/CR4413 54423 heestone Nucher Power Und 3, NUREG/CR4142 Tennessee Vessy Authout Northeast Phraser Energy Co ST4S454 Byron Stemon Urut 1, Commonneeuh Edson Ca NIWG/CR4404 S 260 Oconee Nudeer Stenon, und 1. Duha Power Ca NUREG/CR-3770 ST4S455 Byron Stenon, Und 2. Commanusenh Edmun C4 NUhE3/CR4404 54200 Oconee Nudeer Saeon, und 1 Duke Power Ca NUREG/CR4047 S413 Cataste P6 miser Stenon, Urd 1, DAe Poeur NUREG4954 $06 54287 Oconee Nudeer Staton, Und 3, DAe Power Ca NUREG/CR4374 V03 Ca ST4S$20 Pelo Verde Nudeer Stenort und 1, Artrone NUREG457 $10 544t3 Catsobe Nudeer Stemon, Und 1, Duke PAer NUPEG1191 Pneec Serwce Ca Ca ST454529 Pelo Verde Nudeer Staton, Unt 2, Artone NUREG4057 $10 54414 Catsubs Pbdear Seeman, und 2 Duke Power NUREG0054 S06 Pthec Seruce Ca Ca ST4S$29 Pelo Verde Nudeer Stolbn, Und 2, Artons NUREG1181 54414 Catsube Nudeer Staton, Une 2, DAe Power NURE41191 Pkes Sernce Ca Ca ST4%$30 Pelo Verde Nudeer Stason, unt 3, Artzona NUREG4857 510 54445 Comerume Peak Steam Electic Stenort Uldt 1 NURE4047 513 Pteac Sernce Ca fenas Utsees Elecir S 443 Sestroc* Nudeer Stenon, Und 1, Putic Serwce NUREG4096 SO4 S 446 Comanch Peak Steam Decmc Staton, Und 2, NUREG4797 $13 Co. of New Taas Utetes Decir S 444 Sestrook Nudeer Une 2, Pteac Sarwce NUREG4096 SO4 S 346 Domessee ibdear Power Stenort Une 1, NUREG1177 Ca of New Hampshr Toledo Edson Ca S 400 Sheeran Hems Nudeer Power Plant, Une 1, NUREG1038 S03 54346 Dee8 eses Nudeer Paser Sistort Unit 1, NUREG 1201 Carotne Power & L(t C Toledo Edson Ca

$TM4498 South Texas Proled, Und 1, Houston Ugheng & NUREG4781 54275 Detto Ceyri peudeer Power Plant, Urd 1,

>JREG475 $33 Power Ca Pause Gas 4 Doctic Co STN-5C400 South Temas Protect, Und 2, Houelon Ughtng & NUREG4781 54323 Datso Canyon Nudeer Power Plert, Une 2.

NUREGJ75 $33 Power Co Petiac Gas 4 Oseme Co S 320 Three hide leland Nudeer Staton, Urut 2, NtBEG/CR4315 V01 ST4S447 GESSAR-238 General Flectnc Ca NUREG4979 S05 Metoposten Edson Ca 54354 Hope Creek Nudeer Staton, Urut 1, Pietc NUREGt048 SOS 54320 Three hees leiend Nudeer Staton, Unit 2, NUREG/CR4315 V02 Sernte Elecetc & Gas Co betrocotteri Edson Ca 54354 Hope Creek Nudest Season, Une 1 Pteac NUREG1193 S 120 Thee hele leiend Nudeer Staton, Unit 2, NUREG/CR4315 V03 Sernce Decec 4 Gas Co Metroposten Edson Ca

$4133 Hunen*1 Boy Power Plant, Urd 3, Pecdc Gas & NUREG',106 S 220 Tivee 6tes leiend Nudeer Staten, Und 2.

NUREG/CR4315 V09 Oscme Ca Woropostan Edson Co 54133 Hurnbddt Boy Power Plant, Urut 3, Pacdc Gas & NUREGSR 4316 56 29 Ver*seAows Nutteer Power Stabon, Yar$se NUREG/CR4589 Decmc Ca 4anic Dactic Co.

71 I

IIRC PC8tEB M W3 NUCL51 AGOULATORY COMastessOss i utPOR T NuMetu sAuW py TlOC, add Fef No, et eneys hb'E BIBUOGRAPHIC DATA SHEET

/1 NUREG-0304, Vol.f, No. 2 us rRUCricp O 1,..Rev Ru 3 Tif LE AND Lt 3 LI Avt 0 LANE Regulator d Technical Reports (Abstract Index Journal)

Compilation Second Quarter 1986 April-June l

,,AR mon,,,

S AUTe*Onlls

[D A fl REPOR T ISSutD MON vtAR Ju 1986 7 Pt 9ORMING ORGAMilA1804 NAMt AND MA G ADOR E$$ flagsym g, Ceej 4 PROJECT /

Env0RE umii NUMStR Division of Technical Inf ation and Document Control I

Office of Administration U.S. Nuclear Regulatory Com ion Washington, DC 20555 10 SPONSORING ORGANil Af SON NAME AND MAILING ADOREna,/

I, cases TYPE OP REPOR F l

Quarterly Same as 7, above.

,,,,,,,o o c o,, R, o,,,,,,,,.,,,,,

April - June 1986

\\/

13 SUPPL E ME N T AR Y NOf t s

.3 A.

f aci,m This journal includes all formal reports in the t G series prepared by the NRC staff and contractors, as well as proceedings of confe es and workshops. The entries in tha compilation are indexed for access by title nd stract, contractor report number, p;rsonal author, subject, NRC organization, co ract and licensed facility.

i. oOcu. NT A A L........ ORoi.o..cR,P r om.

,.A.,Ag,3,,

abstract index f

Unlimited it SECURif y CL ASSiplCAflON

< r. n, n o.N,,l a,opiN No.oei.I s Unelassified i F4. wff Unclassified t 7 NLAMGER OF PAQ($

it PR.cs

UNITE 3 STATES C'UCLEA3 EECULATCLY COMMISSION poornes oesse emo WASHINGTON, D C. 20666 9,, c,

2 OFFICIAL 8USINESS m

PENALTY FOR PRIVATE USE. 6300 O

1 Main Citations Q

and Abstracts 1 1AN1AC19Q19T f.

12u b bbu (B B ~i i NM-V 0F<TI BR-PDR NUREG POLICY & PUB G

z

,o W-bo1 0C 20555 g

yAsHILGTON Contractor Report Number Index i

Personal Author Index

n m

l O

1 C

5 1

-4o" 4

Subject Index

>2 O

Nm

=n FI2 NRC Originating L.5 c

l Organization index 2g

-* x 3m

=3

n NRC Contractor O

SponsorIndex j

{

5do2 7

Contractor Index i

Licensed Facility index E

.a l

- - - - - - _ _ _.

ML20205G571

Text

SUMMARY

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