ML20011D815

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Third Quarter 1989,July-September

ML20011D815
Person / Time
Issue date:	11/30/1989
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V14-N03, NUREG-304, NUREG-304-V14-N3, NUDOCS 9001020017
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Regulatory and Technical Reports (Abstract Index Journal)

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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.

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

NUREG--0304 Vol.14, No. 3 Regulatory and Technical Reports (Abstract Index Journal)

Compiiation for Third Quarter 1989 July - September Date Published: November 1989 Regulaton Publications Branch Division of Freedom ofInformation and Publications Services Omcc of Administration I

U.S. Nuclear Regulatory Commission Washington, DC 20555 y u v9 Q6)

1 CONTENTS Prefa'ce....................................................

v.

Index Tab Main Citations and Abstracts.................................................... 1

• Staff Reports

• Conference Proceedings

• Contractor Reports

• International Agreement Reports Secondary Report N umber index........................................................ 2 Personal Author Index.............................................................. 3 S u bject i n de x..................................................................... 4 NRC Originating Organization index (Staff Reports)............................

.5 NRC Originating Organization index (International Agreements)...........................

6-NRC Contract Sponsor index (Contractor Reports)...........................

.............-7 Cont ra cto r Inde x ;.......................................................

..... 8 International Organization index...................................................... 9 -

Licensed Facility index...................................................

....... 10 O

iii 6

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le 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 Publications Services i

Policy and Publications Management Branch Publishing and Translations Section Woodmont 537 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR XXXX, and NUREG/lA-XXXX. These precede the following indbxes:

Secondary Report Number index Personal Author Index Subject Index NRC Originating Organization Index (Staff Reports)

NRC Originating Organization Index (Intemational Agreements)

NRC Contract Sponsor index (Contractor Reports)

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

The bibliographic elements of the main citations are the following:

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

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

Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of 1

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-813. 08632:070.

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

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

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

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

l l

V

-,v.--

international Agreement Report I

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

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

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

ADD

- addendum APP appendix DRFT - draft ERR

- errata N

number R - revision S - supplement V - volume Availability of NRC 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 (2021275-2171. Non-U.S. customers must make payment in advance either by intemational 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 Heport Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor prepared formal NRC reports carry the report code NUREG/CR XXXX. This type of identification replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported.

In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/lA is used for international agreement reports.

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

l vi

(

r Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG XXXX is cn NRC staff-oriainated report, NUREG/CP-XXXX is an NRC-sponsored conference report, NUREG/CR XXXM is an NRC contractor 9repared report, and NUREG/lA XXXX is an inter-national agreement re aort, The bibliograp1ic information (see Preface for details) is followed by a brict abstract of tais report.

NUREG 0020 V13 N06: LICENSED OPERATING REACTORS involved a steam generator tube rupture at McGuire Unit 1.

STATUS

SUMMARY

REPORT. Data As Of May 31,1989.(Gray There were three abnormal occurrences under other NRC-Book 1) LOVELACE,W.H. Division of Computer & Telecommuni-issued ficenses. Two involved medical therapy misadministra-cations Services (Post 890205). July 1989. 542pp. 8907250316.

tions and one involved a medcal diagnostic misadministration.

50610.272.

There were no abnormal occurrences reported by the Agree-THE OPERATING UNITS STATUS REPORT LICENSED ment States. The report also contains information updating i

OPERATING REACTORS provides data on the operation of nu-some previously reported abnormal occurrences.

clear units as timely and accurately as possible. This informa.

NUREG-0304 V14 N01: REGULATORY AND TECHNICAL RE-tion is collected by the Office of Information Resources Man-PORTS (ABSTRACT INDEX JOURNAL). Compilation For First agement from the Headquarters staff of NRCs Office of En-Quarter 1989, January-March.

• Division of Freedom of Informa-forcement (OE), from NRC's Regional Offices, and from utihties.

ton & Pubhcations Services (Post 890205). July 1989. 56pp-The three sectons of the report are: monthly higNights and sta-8908150066. 50888:328.

tistics for commercial operating units, and errata from previously This joumal includes all formal reports in the NUREG series reported data; a compilation of detailed informaton on each prepared by the NRC staff and contractors; proceedings of con.

unit, provided by NRC's Regional Offices, OE Headquarters and forences and workshops; as well as intemational agreement re-the utilities; and an appendix for miscel:eneous information such ports. The entries in this compilation are indexed for access by as spent fuel storage capability, reactor years of experience and title and abstract, secondary report number, personal euthor, non-power reactors in the U.S. It is hoped the report is helpful subject, NRC organization for staff and intemational agree-

?

to all agencies and individuals interested in maintaining an monts, contractor, intemational organization, and heensed f acih-awareness of the U.S. energy situation as a whole.

NUREG 0020 V13 N07: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of June 30,1989.(Gray NUREG 0304 V14 N02: REGULATORY AND TECHNICAL RE.

PORTS (ABSTRACT INDEX JOURNAL). Compilation For Book 1) LOVELACE,W.H. Division of Computer & Telecommuni.

Socorxi Quarter 1989, April-June.

• Division of Freedom of Infor.

cations Services - (Post 890205). August 1989. 551pp.

mation a Publications Services (Post 890205). August 1989.

8908210225. 50988:3 f 3.

See NUREG-0020,V13,N06 abstract.

54pp. 8900120088. 51170:221.

See NUREG 0304,V14,N01 abstract.

NUREG 0020 V13 N08: LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT. Data As Of July 31,1989.(Gray NUREG-0325 R12: U.S. NUCLEAR REGULATORY COMMISSION Book 1) LOVELACE,W.H. Division of Computer & Telecommuni.

FUNCTIONAL ORGANIZATION CHARTS. July 1.1989.

• Ofc of cations Services (Post 890205). August 1989. 541pp.

Personnel (Post 870413). July 1989. 62pp. 8908150087.

8909180268. 51212:037, 50888:265.

See NUREG-0020,V13,N06 abstract.

Functinnal organizaten chads for the U.S. Nuclear Regulatory NUREG 0040 V13 N02: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT, Quarterly NUREG-0525 R15: SAFEGUARDS

SUMMARY

EVENT LIST Report,AprilJune 1989.(White Book)

• Division of Reactor in-(SSEL).

• Division of Safeguards & Transportation (Post spection & Safeguards (Post 870411). August 1989. 54pp.

870413). July 1989. 260pp. 8908070458. 50810:001, 8909260186. 51293.027, The Safeguards Summary Event List provides brief summa-This periodical covers the results of inspections performed by ries of hundreds of safeguards-related events involving nuclear the NRC's Vendor inspection Branch that have been distributed material or facilities regulated by the U.S. Nuclear Regulatory to the inspected organization during the period from April 1989 Commission. Because of public interest, also included are through June 1989, events reported invoMng byproduct material which is exempt fr m safeguards requirements. Events are described under the NUREG-0090 V12 N01: REPORT TO CONGRESS ON ABNOR-cawges: boe Maw, intmson, nyssing/aWy sh,

\\.

MAL OCCURRENCES. January. March 1989.

• Office for Anaty-transportaten-related, tampering / vandalism, arson, firearms-re.

i a s & Evaluation of Operational Data, Director. August 1989' lated, radiological sabota0e, nonradiological sabotage, alcohol-29pp. 8909260159. 51274:351.

and drug related, and miscellaneous. Information in the event Section 208 of the Energy Reorganization Act of 1974 identi-desenptions were obtained from official NRC reports.

fies an abnormal occurrence as an unscheduled incident or event which'the Nuclear Regulatory Commission determines to NUREG-0540 V11 N05: TITLE LIST OF DOCUMENTS MADE be significant from the standpoint of pubhc health and safety PUBLICLY AVAILABLE.May 1 31, 1989.

• Division of Froedom and requires a quarterly report of such events to be made to of information & Publications Services (Post 890205). July 1989.

Congress. This report covers the penod January f, to March 31, 393pp. 8907280328. 50664:322.

1989. For this reporting period, there were two abnormal occur-This document is a monthly pubhcaton containing descrip.

rences at nuclear power plants licensed to operate. The first tiora M information received and generated by the U.S. Nuclear had generic implicatens and involved a plug failuce resulting in Regulatory Commission (NRC). This information includes (1) a steam generator tube leak at North Anna Unit 1. The becond docketed material associated with ct/ilian nuclear p -wer plants 1

1 1

_.m.

2 Mtin Citations Cnd Abstr: cts and other uses of radioactive materials, and (2) nondocketed Directors' Decisions and the Denials of Petitions for Rulemak-matenal received and generated by NRC pertinent to its role as ing are presented.

a regulatory agency. The following indexes are included: Per-sonal Author Corporate Source, Report Number, and Cross NUREG-0750 V29 N05: NUCLEAR REGULATORY COMMISSION Reference to Principal Documents.

ISSUANCES FOR MAY 1989.Pages 395-463.

• Division of

"'0 NUREG-0540 V11 N06: TITLE LIST OF DOCUMENTS MADE pp7N300 g

2 0 17.

PUBLICLY AVAILABLE. June 1 30, 1989.

• Division of Freedom of Information & Publications Services (Post 890205). August Legalissuances of the Commission, the Atomic Safety and Li-censing Appeal Panel, the Atomic Safety and Licensing Board hanel, the Administrahve Law Judge, and NRC program offices eN E 0540 1 N05 abstract.

NUREG 0683 S03: PROGRAMMATIC ENVIRONMENTAL IMPACT STATEMENT RELATED TO DECONTAMINATION AND DIS-NUREG-0750 V29 N06: NUCLEAR REGULATORY COMMISSION POSAL OF RADIOACTIVE WASTES RESULTING FROM ISSUANCES FOR JUNE 198tPages 465-558.

• Division of MARCH 28,1979 ACCIDENT THREE MILE ISLAND NUCLEAR Freedom of Information & Pub 4 cations Services (Post 890205).

STAllON, UNIT 2. Final Supplement Dea!ing With..

• Division August 1989.103pp. 8910100333. 51405:304.

of Reactor Projects 1/11 (Post 870411). August 1989. 600pp.

See NUREG 0750,V29,N05 abstract.

8909180269. 51213.218.

In accordance with the National Environmental Policy Act, the NUREG-0800 02.4.2 R3: STANDARD REVIEW PLAN FOR THE Commission's emplementing regulations, and its April 27,1981, REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR Statement of Pohey, the " Programmatic Environmental impact POWER PLANTS. LWR Edition. Revision 3 To SRP Section Statement Related to Decontamination and Disposal of Radio-2.4.2, " Floods."

• Office of Nuclear Reactor Regulation, Direc-active Wastes Resulting from March 28,1979 Accident Three tor (Post 870411). August 1989. Opp. 8909010070. 51055:024.

Mile Island Nuclear Station, Unit 2," NL' REG-0683 (PElS) has The section of the safety analysis report addressed by this been supplemented. This supplement provides an environmen-section of the standard review plan (SRP) identifies historical tal evaluation of the licensee's proposal to complete the current flooding at the proposed site or in the region of the site,11 sum-cleanup effort and place the facility into monitored storage, and marizes and identifies the ?dividual types of flood-producing a number of alternative courses of action from the end of the phenomena, and combinations of flood-producing phenomena cutrent defueling effort to the beginning of decommissioning or considered in establishing the flood design bases for safety re-refurbishing. The NRC staff has concluded that the licensee's lated plant features. It also covers the potential effects of local proposal to place the facility in monitored storage will not signifi-intense precipitation. The flood history and potential for flooding cantly affect the qualit/ of the human environment. Further, any are reviewed. Factors affecting potential runoff (such as urban.

.mpacts associated with this action are outweighed by its bene-ization, forest fire, or change in agriculture use), erosion, and fits. The benefit of this action is the ultimate elimination of the sediment deposition are considered in the review.

small but continuing nsk associated with the conditions of the facility resulting from the March 28,1979 accident.

NUREG-0800 02.4.3 R3: STANDARD REVIEW PLAN FOR THE NUREG-0713 V08: OCCUPATIONAL RADIATION EXPOSURE AT REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR COMMERCIAL NUCLEAR POWER REACTORS AND OTHER POWER PLANTS, LWR Edition. Revision 3 To SRP Section FACILITIES FOR 1988. Nineteenth Annual Report.

2.4.3, " Probable Maximum Flood (PMF) On Streams And BROOKS,B.G. Division of Regulatory Applications (Post Rivers."

• Office of Nuclear Reactor Regulation, Director (Post 870413). HAGEMEYER,D. Science Applications International 870411). August 1989. 2f pp. 8910060348. 51405:238.

Corp. (formerly Science Applications, Inc.). August 1989.250pp.

The section of the safety analysis report addressed by this 8908140124. 50888:208 section of the Standard Review Plan addresses the hydrome-This reps rt summanzes the occupational radiation exposure teorological design basis developed to determine the extent of information that has been reported to the NRC's Radiation Ex.

any flood protection required for those structures, systems, and posure information Reporting System (REIRS) by nuclear power components necessary to ensure the cSpability to shut down facilities and certain other categories of NRC heensees during the reactor and maintain it in a safe shutdown condition. The the years 1969 through 1986. The bulk of the data presented in areas of review include the probable maximum precipitation the report was obtained from annual radiation exposure reports (PMP) potential and precipitation losses over the applicable submitted in accordance with the requirements of 10 CFR drainage area, tha runoff through nver channels and reservoirs, 20.407. Data on workers terminating their employment at certain the estimate of the discharge rate trace (hydrograph) of the NRC licensed facihties were obtained from reports submitted PMF at the plant site, the determination of PMF water level con-pursuant to 10 CFR 20.408. The 1986 annual reports submitted ditions at the site, and the evaluation of coincident wind-gener.

by 500 licensees indicated that approximately 225.000 individ-ated wave conditions that could occur with the PMF. The analy-uals were monitored. 207.000 of whom were monitored by nu-ses involve modehng of physical rainfall and runoff processes to clear power facihties. They incurred an average individual dose estimate the upper level of possible flood conditions adjacent to of 0.20 rem (cSv) and an average measurable dose of 0.40 rem and on the site.

(cSv). Termination radiation exposure reports were analyzed to NUREG-080013.1.2 R3: STANDARD REVIEW PLAN FOR THE reveal that about 77,600 individuals completed their employ-rnent with one or more of the 500 covered hcensees during REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR 1985. Some 75,500 of these individuals terminated from power POWER PLANTS. LWR Edition. Revision 3 To SRP Section reactor facilities, and about 6,700 of them were considered to 13.1.213.1.3, " Operating Organization."

• Office of Nuclear Re-be transient workers who received an average dose of 0.75 rem actor Regulation, Director (Post 870411). August 1989.11pp.

(cS4 8909010111, 51055:035.

The revision of SRP Section 13.1.2-13.1.3 (Revision 3), "Op-NUREG-0750 V29101: !NDEXES TO NUCLEAR REGULATORY erating Organization," added reference to Regulatory Guide COMMISSION ISSUANCES. January-March 1989.

• Division of 1.114. " Guidance to Operators at the Controls and to Senior Freedom of information & Publications Services (Post 890205).

Operators in the Control Room of a Nuclear Power Plant."

July 1989. 47pp. 8908300143. 51075:256.

Changes were made to delete the description of certain require-Digest and indexes for issuances of the Commission, the ments with respect to licensed operators and senior licensed Atomic Safety and Licensing Appeal Panel, the Atomic Safety operators that are now desenbed in 10 CFR 50.54(m), and to and Licensing Board Panel, the Administrative Law Judge, the include reference to the Commission Pokey Statement on Engi-

Main Citations and Abstracts 3

neering Expertise on Shift in place of a statement on the Shift This document is a Compilation of nuclear regulatory legista-Techrucal Advisor, tion and other relevant matenal through the 100th Congress, 2nd Session. This compilation has been prepared for use as a NUREG-0837 V09 N01: NRC TLD DIRECT RADIATION MONI.

resource document, which the NRC Intends to update at the TORING NETWORK. Progress Report. January March 1989.

end of wey Ngress. W com M MM-MM NW STRUorMEYER.R.; MCNAMARA,N. Region 1 Ofc of the Direc-The Atomic Energy Act of 1954, as amended; Energy Reorgani.

tor. June 489. 229pp. 6907250267. 50609.272.

zation Act of 1974, as amended; Uranium Mill Tailings Radiation This report provides the status and results of the NRC Ther, Control Act of 1978; Low-Level Radioactive Waste Polcy Act; moluminescent Dosimeter (TLD) Direct Radiation Monitonng Nuclear Waste Policy Act of 1982; and NRC Authorization and Network, it presents the radiation levels measured in the vicinity Appropriahons Acts. Other materials included are statutes and

- of NRC licensed facility sites throughout the country for the first treaties on export heensing, nuclear non-proliferation, and envi-quarter of 1989.

tonmental protection.

NUREO-0837 V09 N02: NRC T!.D DIRECT RADIATION MONI.

TORING NETWORK. Progress Report. April-June 1989.

NUREG-0991 809: SAFETY EVALUATION REPORT RELATED STRUCKMEYER.R.; MCNAMARA.N. Region 1, Ofc of the Direc-TO THE OPERATION OF LIMERICK GENERATING tor. September 1989. 227pp, 8910100303. 51409:135.

STATION UNITS 1 AND 2. Docket Nos. 50-352 And 50-This report provides the status and results of the NRC Ther-353.(Philadelphia Electric Company)

Division of Reactor moluminscent Dosimeter (TLD) Direct Radiation Monitonng Net-Projects 1/II (Post 870411). August 1989. 46pp. 8909120142.

l work. It presents the radiation levels measured in the vicinity of 51145:122 NRC licensed facility sites throughout the country for the in August 1983 the staff of the Nuclear Regulatory Commis-second quarter of 1989.

sion issued its Safety Evaluation Report (NUREG 0991) regard-ing the application of the Philadelphia Electric Company (the li-NUREG-0936 V08 NO2: NRC REGULATORY AGENDA. Quarterly consee) for the licenses to operate the Limenck Generating Report, April June 1989.

• Division of Freedom of Information &

Station, Units 1 and 2, located on a site in Montgomery and Publications Services (Post 890205). Juiy 1989. 129pp.

Chester Count.es, Pennsytvania. Supplement 1 was issued in 8908150121. 50888:090.

December 1983. Supplement 2 was issued in October 1984.

The NRC Regulatory Agenda is e compilation of all rules on which the NRC has proposed or is considenng action and all Supplement 3 was issued in October 1984. Supplement 4 was issued in May 1985. Supplement 5 was issued in July 1985.

petitions for rulemaking which have been received by the Com, Supplement 6 was issued in August 1985 and Supplement 7 mission and are pending disposition by the Commission. The was issued in April 1989. Supplement 7 addresses the major Regulatory A0enda is updated and issued each quarter.

design difterences between Units 1 and 2, the resolution of all NUREG-0940 V06 NO2: ENFORCEMENT ACTIONS: SIGNIFl*

issues that remained open when the Unit 1 full-power license CANT ACTIONS RESOLVED.Ouarterly Progress Report, April-was issued, the staff's assessment regarding the application by June 1989.

• Ofc of Enforcemem (Post 870413). September the licensee to operate Unit 2 and issues that require resolution 1989. 386pp. 8910100283. 51406:272-before issuance of an operating license for Unit 2. Supplements This compilation summarizes significant enforcement actions 8 and 9 address further issues that require resolution prior to that have been resolved during one quarterly penod (April.

issuance of an operating license.

June 1989) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to licensees with NUREG 1150 V01: SEVERE ACCIDENT RISKS: AN ASSESS-respect to these enforcement acbons. Also included are a MENT FOR FIVE U.S. NUCLEAR POWER PLANTS. Summary number of enforcement actions that had been previously re-Report.Second Draft For Peer Rev;ew.

• Division of Systems solved but not published in this NUREG. It is anticipated that Research (Post 880717). June 1989. 301pp. 8907250331.

the information in this publication will be widely disseminated to 50608:331.

managers and employees engaged in activities licensed by the This document discusses the nsks from severe accidents in NRC, so that actions can be taken to improve safety by avoid-five commercial nuclear power plants. Information is presented ing future violabons similar to those desenbed in this publica-on the frequencies of core damage accidents from intemally ini-tion.

tiated accidents (and from extemally initiated accidents for two plants), containment performance under severe accident loads.

NUREG-0974 SUPP: FINAL ENVIRONMENTAL S1 ATEMENT RE-releases of radioactive material and offsite consequences, and 4

LATED TO THE OPERATION OF LIMERICK GENERATING nsk (the product of accident frequencies and Consequences).

STATION, UNITS 1 AND 2. Docket Nos. 50-352 And 50-This report is a second draft for peer review, modified to ac-353 (Philadephia Electric Company)

Division of Reactor count for comments on a February 1987 draft from the public Projects.1/11 (Post 870411). August 1989. 52pp. 8909120120.

and three formal peer reviews of that draft. Following a peer 51170:005.

review of this version, a final report will be issued. Volume 1 of in April 1984 the staff of the Nuclear Rogulatory Commission this report provides summaries of the risk analysis results for issued its Final Environmental Statement (NUREG4974) related the five studied plants, perspectives on these results, and a dis-to the operation of Limenck Generating Station Units 1 and 2 cussion of the role of these risk analyses in the NRC staff's (Docket Nos. 50-352 and 50-353), located on the Scht/ylkill Rrver, near Pottstown, in Limerick Township, Montgomery and severe accident regulatory program.

Chester Counties, Pennsylvania. The NRC has prepared this NUREG 1150 V02: SEVERE ACCIDENT RISKS: AN ASSESS-supplement to NUREG 0974 to present its evaluation of the al-MENT FOR FIVE U.S.

NUCLEAR POWER temative of facility operation with the installation of further PLANTS. Appendices.Second Draft For Peer Review.

• Division severe accident mitigation design features. The NRC staff has of Systems Research (Post 880717). June 1989 310pp.

discovered no substantial changes in the proposed action as 8907250337. 50608:021 previously evaluated in the Final Environmental Statement that This document discusses the risks from severe accidents in are relevant to environmental concerns nor significant new cir-five commercial nuclear power plants. Information is presented cumstances or information relevant to envircnmental concerns on the frequencies of core damage accidents from intemally ini-an bear ng on the licensing of Limerick Generating Station, tiated accidents (and from extemally initiated accidents for two plants), containment performance under severe accident loads, NUREG-0980 R04: NUCLEAR REGULATORY LEGISLATION.

• releases of radioactive matenal and offsite consequences, and Office of the General Counsel (Post 860701). August 1989 nsk (the product of accident frequencies and consequences) 200pp. 8910100280. 51406-045.

This report is a second draft for peer review, modified to ac-h i

1

4' Main Citations and Abstracts count for comments on a February 1987 draft from the public NUREG 1229: REGULATORY ANALYSIS FOR RESOLUTION OF and three formal peer reviews of that tiraft. Following a peer USl A-17. Systems Interactions in Nuclear Power Plants.

rev6ew of this version, a final report will be issued. Volume 2 of THATCHER.D.F. Division of Safety issue Resolution (Post this report provides more detailed discussion of the methods 880717). August 1989. 26pp. 8908210132. 50986:193.

used in the nsk analyses, additional discussion on specific tech-This report presents a summary of the regulatory analysis recal issues important in the analyses, and responses to com.

conducted by the NRC staff to evaluate the value and impact of ments received on the February 1987 draft.

potential attematives for the resolution of Unresolved Safety issue (USI) A 17 " Systems Interactions in Nuclear Power NUREG 1217: EVALUATION OF SAFETY IMPLICATIONS OF Plants." The NRC staff's proposed resolution offered in this CONTROL SYSTEMS IN LWR NUCLEAR POWER report is based on this analysis. The staff's technical finding re-PLANTS. Technical Findings Related To USl A-47. Final Report.

garding systems interactions can be found in NUREG 1174. Ad-SZUKIEWICZ,A.J. Division of Safety issue Resolution (Post verse systems interactions (Asis) involve subtle and often very 880717). June 1989. 64pp. 8907250271. 50610:141, complicated plant-specific dependencies between components This report sumrnartzes the work performed by the Nuclear and systems, possibly compounded by inducing erroneous Regulatory Commission (NRC) staff and its contractors, Idaho human intervention. The staff has identified actions to be taken National Engineering Laboratory, Oak Ridge National Laborato.

by licensees and the NRC to resolve USl A 17; the staff has ry, and Pacific Northwest Laboratory, leading to the resolution also made the judgment that these actions, together with other of Unresolved Safety issue (USI) A-47, " Safety Imphcations of ongoing activities, would reduce the risk from adverse systems Control Systems." The technical findings and conclusions pre-Interactions. As discussed further in this report, the staff judg-sented in this document are based on the technical work com-ment that the actions are sufficient is not based on the asser.

plated by the contractors. The principal documents that contain tion that all systems interactions have been identified, but rather the technical findings and conclusions of the contractors who that the A 17 actions, plus other activities by the licensees and worked on USl A-47 are summanzed in Appendix B. An in-depth staff, will ider'tify precursors to potentially risk significant interac-evaluation was performed on non-safety-related control systems tions so that action Can be taken if deemed necessary.

(See Section 1) that are typically used during normal plant oper*

atson on four nuclear steam supply system plants: a General NUREG-1233: REGULATORY ANALYSIS FOR USl A-40, "SEIS.

MIC DESIGN CRITERIA." Final Report. SHAUKAT,S.K.;

Electric Company boiling-water reactor, a Westinghouse 3 loop CHOKSHi,N.C. DiAsion of Safety issue Resolution (Post pressurized-water reactor (PWR), a Babcock & Wilcox Co.

880717). September 1989. 27pp. 8910100308. 51409:001.

(B&W) once-through steam generator PWR, and a Combustion Engineering PWR design. A study was also conducted to deter-This report consists of a regulatory analysis for Unresolved trine the generic applicability of the results to the class of Safety issue A-40, " Seismic Design Cntena." The regulatory plants represented by the specific plants analyzed. Generic con-analysis discusses the impact of the proposed changes in the Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and 3.7.3 clusions were then developed. Steam generator and reactor (NUREG-0800).

vessel overfill events and reactor vessel overcooling events were idontified as major classes of events having the potential NUREG-1267: TECHNICAL RESOLUTION OF GENERIC SAFETY to be more severe than previously analyzed. Specific subtasks ISSUE A-29. Nuclear Power Plant Design For Reduction Of Vul-of this issue were to study these events to determine the need nerability To industrial Sabotage. SERKl2 A.W. Division of for preventive and/or mitigating design measures. This report Safety Issue Resolution (Post 880717). September 1989.35pp.

describes the technical studies performed by the laboratones, 8910100321. 51409:029.

the NRC staff assessment of the results, the generic applicabil.

This report summarizes key technical findings related to ity of the evaluations, and the tJchnical findings resulting from Safety issue A 09," Nuclear Power Plant Design for the Reduc-these studies. This final report contains the staff's responses to, tion of Vulnerability to Industnal Sabotage." The findings in this and resolution of, the public comments that were solicited and report deal with (1) an historical review of reported sabotage-received before September 16.1988 in response to the draft re, related events at nuclear facilities, (2) NRC physical security re-ports issued for public comments on May 27,1988.

quirements. (3) industry measures to prevent / mitigate sabotage, (4) design and procedural approaches that could be used to NUREG-1218: REGULATORY ANALYSIS FOR RESOLUTION Op deter sabotage, (5) current NRC and industry initiatives aimed at USI A 47. Safety implications Of Control Systems in LWR Nucle-personnel screening and selection, and (6) design consider-ar Power Plants. Final Report. S2VKIEWICZ,A.J. Division of ations applicable to Advanced Light Water Reactors (ALWRs).

Safety issue Resolution (Post 880717). July 1989. 49pp.

The results reveal that insider sabotage at U.S. operating nucle-8908080281. 50824:307, er plants has not been a significant problem in the United This report presents a summary of the regulatory analysis States to date and that there are no singular design modifica-conducted by the NRC staff to evaluate the value impact of al-tions or procedures that by themselves would completely elimi-ternatives for the resolution of Unresolved Safety Issue (USI) A.

nate or mitigate the threat of insider sabotage. Rather, it will 47, " Safety implications of Control Systems." The NRC staff take a combination of systematic and focused improvements in proposed resolution is based on these analyses and the techni-the three areas of reliable personnel, effective design features, cal findings and conclusions presented in NUREG-1217. The and plant procedures developed to provide a strategy to deal staff has concluded that certain actions should be taken to im-with prevention of insider sabotage and to De able to initigate prove safety in light water reactor (LWR) plants. The actions ahse ahs.

recommended that certain plants upgrade their control systems NUREG 1272 V03 N01: AEOD OFFICE FOR ANALYSIS AND to preclude reactor vessel / steam generator overfill events and EVALUATION OF OPERATIONAL DATA 1988 ANNUAL

)

to prevent cteam generator dryout, modify their technical speci.

PEPORT. Power Reactors.

• Office for Analysis & Evaluation of fication to penodically venfy operability of these systems, and Operational Data, Director. June 1989. 221pp. 8907250282.

modify selected emergency procedures to ensure plant safe 50680:263.

shutdown following a small-break loss-of. coolant accident. This The annual report of the U.S. Nuclear Regulatory Commis-report was issued as a draft for.public comment on May 27, sion's Office for Analysis and Evaluation of Operational Data 1988. As a result of the pubhc comments recewed, this report (AEOD) is devoted to the activities performed dunng 1988. The was revised. The NRC staff's responses to and resolution of the report is published in two separate parts. NUREG 1272, Vol. 3, public comments are included as Appendix C to the final report, No. I covers Power Reactors and presents an overview of the NUREG-1217.

operating expenence of the nuclear power industry, including

M^ln Citations and Abstracts 5

comments about the trends of some key performance meas-NUREG 1337 R01: STANDARD REVIEW PLAN FOR THE ures. The report also includes the pnncipal findings and issues REVIEW OF FINANCIAL ASRURANCE MECHANISMS FOR identified in AEOD studies over the past year and summartzes DECOMMISSIONING UNDER 10 CFR PARTS 30, 40, 70 AND information from Licensee Event Reports, Dagnostic Evalua-

72.

• Dvision of Low Level Waste Management & Decommis-tions, and reports to the NRC's Operations Center. NUREG-tioning (Post 870413). August 1989. 43pp. 8908290283.

1272 Vol. 3, No. 2 covers Nonreactors and presents a review 51048:109.

of the nonreactor events and misadministrations that were re-Standard Review Plan (SRP) for the Review of Financial As-ported in 1988 and a brief synopsis of AEOD studies published surance Mechanisms for Decommissioning Under 10 CFR Parts in 1988. Each volume contains a list of the AEOD reports 30,40,70 and 72,is prepared for the guidance of Nuclear Reg-Issued for 1980 1988.

ulatory Commission staff reviewers in portrming reviews of ap-plications from matenal heensees affecM) by the decommis-NUREG-1272 V03 NO2: AEOD OFFICE FOR ANALYSIS AND sioning regulations established June 27,1988 (53 FR 24018).

EVALUATION OF OPERATIONAL DATA 1988 ANNUAL The pnncipa! purpose of the SRP is to assure the quality and REPORT.Nonreactors.

• Office for Analysis & Evaluation of uniformity of staff reviews and to present a base from which to Operational Data, Drector. June 1989. 88pp. 8907250276.

evaluate the financial assurance aspects of the apphcations.

50610.208.

NUREG-1337, Rev.1, identifies who performs the review, the See NUREG-1272,V03,N01 abstract.

matters that are reviewed, the basis of the review, how the NUREG 1275 V05 ADD: OPERATING EXPERIENCE FEEDBACK review is performed, and the conclusions that are sought.

REPORT PROGRESS IN SCRAM REDUCTION. Commercial Power Reactors. BELLL. Office for Analysis & Evaluation of NUREG 1347: NRC STAFF SITE CHARACTER 1ZATION ANALY-Operational Data, Drector, August 1989. 21pp. 8909200229.

SIS OF THE DEPARTMENT OF ENERGY'S SITE CHARAC-51274:262.

TERIZATION PLAN, YUCCA MOUNTAIN SITE, NEVADA.

• Di-The U.S. Nuclear Regulatory Commission's (NRC) Office for vision of High-Level Waste Management (Post 870413). August Analysis and Evaluation of Operational Data (AEOD) evaluated 1989. 220pp. 8909260173. 51277:245.

U.S. Light Water Reactors (LWR) unplanned reactor scram ex-This Site Characten2ation Analysis (SCA) documents the penence in light of ongoing industry scram reduction programs NRC staffs concerns resulting from its review of the U.S. De-in NUREG-1275 Vol. 5. The purpose of this work which cov' partment of Energy's (DOE's) Site Characterization Plan (SCP) ered the years 1984 through 1987 was to provide feedback to for the Yucca Mountain site in southern Nevada, which is the industry, the NRC staff, and the public regarding the trends in candidate site selected for characterization as the nation's first unplanned scrams at U.S. commercial power reactors. A pri-geologic repository for high-level rc'hactive waste. DOE's SCP mary objective of AEOD's analysis was to determine the maior explains how DOE plans to obtain the information necessary to sources of unplanned scrams for the most recent data and to determine the suitability of the Yucca Mountain site for a reposi-determine whether the scram reduction programs supported by to'Y. NRC's specific objections related to the SCP, and major various nuclear steam supply system (NSSS) owners groups comments and recommendations on the vanous parts of DOE's were addressing the proper areas for future scram reduction.

program, are presented in SCA Section 2, Dractor's Comments This addendum updates that work through March 1989.

and Recommendations. Section 3 contains summanes of the NUREG-1335: INDIVIDUAL PLANT EXAMINATION; SUBMITTAL NRC staff's concerns fnr each specific program, and Section 4 GUIDANCE. Final Report.

• Office of Nuclear Regulatory Re.

contains NRC staff point papers which set forth in greater detail search, Director (Post 860720).

• Office of Nuclear Reactor particular staff concerns regarding DOE's program. Appendix A Regulation, Director (Post 870411). August 1989. 99pp.

presents NRC staff evaluations of those NRC staff Consultation 8909110348. 51143:328.

Draft SCP concems that NRC considers resolved on the basis Based on the Pohey Statement on Severe Reactor Accidents of the SCP. This SCA fulfills NRC's responsibihties with respect Regarding Future Designs and Existing F. ants, the performance to DOE's SCP as specified by the Nuclear Waste Policy Act of a plant examination is requested from the hcensee of each (NWPA) and 10 CFR 60.18.

nuclear power plant. The plant examination looks for severe ac-cident vulnerabihties and cost effective cafety improvements NUREG 1356: STATE COST SHARING OF TRAINING.A Task that would reduce or eliminate any discovered vulnerability. This Force Report.

MONTGOMERY,,t.M.;

FLATER,0.A.;

document delineates the guidance for reporting the results of a HUGHES D.R.; et al. State, Local & Indian Tribe Programs.

plant examination-August 1989.117pp. 8909110360. 51143:211, NUREG-1336 R01: STANDARD FORMAT AND CONTENT GUIDE "U*E 8

"O FOR FINANCIAL ASSURANCE MECHANISMS REQUIRED NRC Training Program for States., This report re'sponded to a FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,70 Commission's request for study of NRC's long-standing practice AND 72.

• Division cf Low. Level Waste Management & Decom-of paying the travel and per diem of state personnel who attend missioning (Post 870413). August 1989.143pp. 8908290286.

NRC sponsored training. In May 1988, the Chairman endorsed 51048:155.

the report in most respects but asked for further study of a cost Standard Format and Content Guide for Financial Assurance sharing of travel and per diem costs. As a result, the Director of Mecharusms Required for Decommissioning Under 10 CFR GPA's State, Local and Indian Tribe Programs estabhshed a Parts 30,40,70 and 72, NUREG 1338, Rev 1, discusses the in-Task Force compnsed of representatives from the Conference formation to be provided in a hcense apphcation, and estab-of Radiation Control Program Directors, Inc., the Agreement hshes a uniform format for presenting the information required States and the NRC to look at ways that the States can share to meet the decommissioning licensing requirements. The use the costs of NRC training, particularly travel and per diem. At of the Standard Format and Content Guide will (1) help ensure that the heense application contains the information required by the request of the Director, GPA, the Task Force also looked at the regulations, (2) aid the applicant in ensunng that the infor.

related Cost and quantity issues associated with the NRC train-mation is complete, (3) help persons reading the Standard ing program for State personnel. This report includes a discus-Format to locate information, and (4) cor' tribute to shortening sion of NRC and State perspectives on the issue of sharing the time required for the review process. The Standard Format travel and per diem costs, a discussion of options, and recom-and Content Guide ensures that the information required to per-mendations for likely cost savings and quality of training im-form the review is provided and in a useable format.

provement.

\\

6 Main Citations and Abstracts

(

NUREG-1866: REVISED SEVERE ACCIDENT RESEARCH PRO.

NUREG 1375 V01: SAFETY EVALUATION STATUS REPORT GRAM PLAN. Fiscal Year 1990 1992.

• Division of Systems FOR THE PROTOTYPE LICENSE APPLICATION SAFETY Research (Post 880717). August 1984. 36pp. 8908310061.

ANALYSTS REPORT. Earth Mounded Concrete Bunker.

• Divi-51075:140.

sion of Low Level Waste Management & Decommissioning The revised Severe Accident Research Program has been (Post 870413) July 1989. 93pp. 8908210185. 50986:220.

prepared by the Office of Nuclear Regulatory Research to sup-The U.S. Nuclear Regulatory Commission (NRC) staff and port the tasks and objectives discussed in the staff's "Integra, consultants reviewed a Prototype License Application Safety bon Plan for Closure of Severe Accident issues " SECY 88147, Analysis Report (PLASAR) submitted by the U.S. Department of The revised SARP addresses both the near term research di, Energy (DOE) for the Earth Mounted Concrete Bunker (EMCB) rected at providing a technical basis upon which decisions on afternative medcd of low-level radioactive waste disposal. The important containment performance issues can be made, and NRC reviewers relied extensively on the Standard Review Plans the long-term research needed to confirm and refine our unde *-

(SRPs), Rev.1 (NUREG-1200) to evaluate the acceptability of standing of severe accidents.

the information provided in the EMCB PLASAR. Certain review areas in the PLASAR were selected for development of evalua-NUREG 1368: DRAFT PREAPPLICATION SAFETY EVALUATION ti n report input by the NRC staff, to provide examples of safety REPORT FOR THE POWER REACTOR INHERENTLY SAFE assessments that are necessary as part of a licensing review.

MODULE LIQUID METAL REACTOR. LANDRY,R.R.t KING,T.L; th so I ni he eview ea ay ou WILSON.J.N. Division of Regulatory Applications (Post 870413).

September 1989. 2pp. 8909290216. 51576:135.

review status, the NRC staff report is labeled a " Safety Evalue.

This draft safety evatuation report (SER) presents the prelimi-ton Status Report" (SESR). Appendix A provides the NRC review comments and questions on the informatson submitted in nary results of a preapplication design review for the standard Power Reactor inherently Safe Module (PRISM) liquid metal re-the PLASAR. The WRC review concentrated on the design and operations-related portions of the EMCB PLASAR.

actor (Project 674). The PRISM conceptual design was submit-ted by the U.S. Department of Energy (DOE) in accordance with NUREG-1376: TECHNICAL SPECIFICATIONS FOR LIMERICK the U.S. Nuclear Regulatory Commission (NRC) " Statement of GENERATING STATION, UNIT 2. Docket No. 50-353.(Philadel-Policy for the Regulation of Advanced Nuclear Power Plants" phia Electric Company)

• Division of Reactor Projects 1/11 (51 FR 24643), which provides for earty Commission review and (Post 870411). August 1989. 559pp. 8909120128. 51197:017, interacton. The standard PRISM plant consists of nine identical The Limehek Unit 2, Technical Specifications were prepared reactor modules, each with a thermal output of 425 MWt, cou-by the U.S. Nuclear Regulatory Commission to set the hmits, pied with three steam turbine gene #: ais to produce a total operating conditions, and other requirements apphcable to a nu-plant electrical output of 1245 MW. (he reactors are liquid clear reactor facility as set forth in Section 50.36 of 10 CFR sodium cooled and utihre metallic typ fuel. The design includes Part 50 for the protection of the health and safety of the public.

passive reactor shutdown and decay heat removal features. The NUREG-1377: NRC RESEARCH PROGRAM ON PLANT AGING:

staff and its contractors at the Brookhaven National Laboratory LISTING AND ABSTRACTS OF REPORTS ISSUED THROUGH have reviewed this design with emphasis on those unique provi-FEBRUARY 1,1989. KONDIC.N.N. Division of Engineering sions In the design that accomplish the key safety functions of (Post 870413). August 1989. 73pp. 8908310069. 51075:068.

reactor shutdown, decay-heat-removal, and containment of re-The U.S. Nuclear Regulatory Commission is conducting the dioactive material Final guidance on the acceptability of the Nuclear Plant Aging Research (NPAR) Program. This is a com-PRISM standard design is contingent on receipt and evaluation prehensive hardware-oriented program focused on understand-of additional information requested from DOE pertaining to the Ing the aging mechanisms of components and systems in nucle-adequacy of non-conventional containment designs.

ar plants. The NPAR Program also focuses on methods for sim-ulating and monitoring the aging-related degradation of these NUREG-1371: TECHNICAL SPECIFICATIONS FOR LIMERICK components and systems. This document contains a hsting and GENERATING STATION, UNIT 2. Docket No. 50-353.(Philadel-index of reports generated in the NPAR Program that were phia Electnc Company)

• Division of Reactor Projects 1/11 issued through February 1,1989, and abstracts of those re-(Post 870411). July 1989. 541pp. 8908080315. 50834:086.

ports. Each abstract describes the elements of the research The Limerick, Unit 2, Technical Specifications were prepared covered in the report and outlines the significant results. For the by the U.S. Nuclear Regulatory Commission to set the limits, convenience of the user, the reports are indexed by personal

- operat ng conditions, and other requirements applicable to a nu.

author, corporate author, and subject.

clear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public.

NUREG/CR-2000 V08 N6: LICENSEE EVENT REPORT (LER)

COMPILATION.For Month Of June 1989.

• Oak Ridge National NUREG-1373: TECHNICAL POSITION ON POSTCLOSURE Laboratory. July 1989.100pp. 8908140143. ORNL/NSIC 200.

SEALS, BARRIERS AND DRAINAGE SYSTEM IN AN UNSATU-50887d 00.

RATED MEDIUM. GUPTA,0.C.; BUCKLEY,J.T. Division of High-This monthly report contains Licensee Event Report (LER)

J Level Waste Management (Post 870413). Au9ust 1989. 35pp' operational information that was processed into the LER data i

8909180310. 51211:085.

file of the Nuclear Safety Information Center (NSIC) during the The purpose of this technical position is to provide guidance one month period identified on the cover of the document. The with respect to the current DOE sealing and drainage concepts LERs, from which this information is derived, are submitted to for a geologic repository in an unsaturated medium. Section 2.0 the Nuclear Regulatory Commission (NRC) by nuclear power i

of the technical position provides a listing of the 10 CFR Part 60 plant licensees in accordance with federal regulations. Proce-regulations which are applicable to the design, testing, selection dures for LER reporting for revisions to those events occumng of materials and placement of the postclosure seals, barriers prior to 1984 are desenbed in NRC Regulatory Guide 1.16 and NUREG-0161, " Instructions for Preparation of Data Entry End drainage system. Staff position statements and the corre.

Sheets for Licensee Event Reports." For those events occurring sponding discussions are presented in Sections 3.0 and 4.0, re-on and after January 1,1984, LERs are being submitted in ac-spectivety. Technical positions are organized according to the cordance with the revised rule contained in Title 10 Part 50.73 following topics: (1) design consideration, (2) site charactenza-of the Code of Federal Regulations (10 CFR 50.73 - Licensee tion considerations, (3) performance confirmation consider-Event Report System) which was published in the Federal Reg-ations, and (4) performance analysis considerations.

inter (Vol. 48, No.144) on July 26,1983. NUREG-1022, "Li-m

Main Citations cnd Abstracts 7

consee Event Report System - Desenption of Systems and tween these features and earthquakes. (8) Vertical crustal Guidehnes for Reporting," provides supporting guidance and in-movements along the coast of Maine were studied.

formation on the revised LER rule. The LER sumrnaries in this report are arranged ulphabetically by facility name and then NUREG/CR 4219 V06 N1: HEAVY SECTION STEEL TECHNOL-chronologically by event date for each facihty. Component, OGY PROGRAM. Semiannual Progress Report For October aystem, keyword, and component vendor indexes follow the 1988 March 1989. CORWIN.W.R. Oak Ridge National Labora-summanes. Vendors are those identified by ?e utility when the tory. September 1989. 77pp. 8909290245. ORNL/TM 9593.

LER form is initiated, the keywords for the componsnt, system, 51333:065.

and general keyword indexes are assigned by computer using The Heavy Section Steel Technology (HSST) Program studies correlation tables from the Sequence Coding and Search concern all areas of the technology of materials fabricated into System.

thick-section, primary-coolant containment systems of light-water cooled nuclear power reactors. The focus is on the be-NUREG/CR 2000 V06 N7: LICENSEE EVENT REPORT (LER) havior and structural integrity of steel reactor pressure vessels COMPILATION.For Month Of July 1989.

• Oak Ridge National (RPVs) containing cracklike flaws. During this period, analytical Laboratory. August 1989.100pp. 8909180334. ORNL/NSIC-efforts included examining the influence of high crack arrest i

200. 51215:098.

toughness on RPV integnty and an increased emphasis on eval-See NUREG/CR.2000,V08,N06 abstract.

uating large international structural experiments. Two areas of NRC topical support were continued: (1) the evaluation of NUREG/CR-2000 V08 N8: LICENSEE EVENT REPORT (LER) mechanisms for enhanced low temperature, low-flux irradiation COMPILATION.For Month Of August 1989.

• Oak Ridge Nation-embrittlement that may affect the integrity of RPV supports, and al Laboratory. September 1989.156pp. 8910100312, ORNL/

(2) an overall assessment of low upper shelf (LUS) welds in NSIC 200. 51410:005.

RPVs with special emphasis on reevaluating ductile teanng cri-See NUREG/CR-2000,V08,N06 abstract.

teria. The first four stut> panel crack arrest tests were per-NUREG/CR-2331 V08 N4: SAFETY RESEARCH PROGRAMS formed. Posttest material characterization was performed for SPONSORED BY OFFICE OF NUCLEAR REGULATORY clad-plate and wide plate Senes 2 test materials. Statishcal RESEARCH. Progress Report, October December

1988, analyses were performed on the data from the Fifth HSST Irra-WEISS A.J. Brookhaven Natenal Laboratory. July 1989.164pp.

diation Series on the study of K(ic) shifts. Analysis of the irradi-8908150135. BNL-NUREG-51454. 50887:286.

ated fracture-toughness testing was completed for the Seventh This progress report describes current activities and technical HSST trradiation Series on cladding. Detailed planning was progress in the programs at Brookhaven National Laboratory begun for the next pressurized-thermal-shock expenment, sponsored by the Division of Regulatory Applications, Division PTSE-4, to examine the extent of ductile tearing and its interac-of Engineenng, Division of Safety issue Resolution, and Dmsion tion with cleavage fracture in an LUS weld metal.

of Systems Research of the U.S. Nuclear Regulatory Commis-NUREG/CR-4234 V02: AGING AND SERVICE WEAR OF ELEC-sion, Offce of Nuclear Regulatory Research following the reor-TRIC MOTOR-OPERATED VALVES USED IN ENGINEERED ganization in July 1988. The previous reports have covered the SAFETY-FEATURE SYSTEMS OF NUCLEAR POWER period October 1,1976 through September 30,1988.

PLANTS. Aging Assessments And Monitonng Method Evalua.

NUREG/CR 2331 V09 N1: SAFETY RESEARCH PROGRAMS tions. HAYNES,H.D. Oak Ridge National' Laboratory. August SPONSORED BY OFFICE OF NUCLEAR REGULATORY 1989. 212pp. 8909120103. ORNL 6170. 51170:301, RESEARCH. Progress Report. January-March 1989. WEISS A.J.

In recent years MOVs have received considerable attention Brookhaven National Laboratory. August 1989. 153pp.

by the Nuclear Regulatory Commission and the nuclear power 8908290274. BNL-NUREG-51454. 51048:298.

industry and were identified as a component for study by the This progress report desenbes current activities and technical Nuclear Plant Aging Research (NPAR) Program. In support of progress in the programs at Brookhaven National Laboratory the NPAR Program, a comprehensive Phase ll aging assess-sponsored by the Division of Regulatory Applications, Division ment on MOVs was performed by the Oak Ridge National Labo.

of Engineenng, Dmsson of Safety issue Resolution, and Division ratory (ORNL), and the results of this study are presented in of Systems Research of the U.S. Nuclear Regulatory Commis-this report. An evaluation of commercially avadable MOV moni-sion, Office of Nuclear Regulatory Research following the reor-toring methods was carried out, as well as an assessment of ganization in July 1988. The previous reports have covered the other potentially useful techniques. These assessments led to penod October 1,1976 through December 31,1988.

the identification of an effective, nonintrusive, and remote tech-nique, namely, motor current signature analysis (MCSA). The NUREG/CR 3252: NEW ENGLAND SEISMOTECTONIC STUDY capabilities of monitonng methods (especially f>CSA) for detect-ACTIVITIES DURING FISCAL YEAR 1980. BAROSH,P.J.;

ing changes ir operating conditions and MOV degradation were SMITH,P,V. Boston Col;ege, Weston, MA. September 1989, investigated in controlled laboratory tests at ORNL, en situ MOV 241pp. 89101C0298. 51408:115-tests at a neig1bonng nuclear power plant, and the Gate Valve The Fiscal Year 1980 New England Seismotectonic Program Flow interrupticn Blowdown test in Huntsville, Alabama.

consisted of thirty separate but coordinated studies covenng New England and adiacent Canada, which are desenbed in this NUREG/CR-4530 V03: U.SJFRENCH JOINT RESEARCH PRO-report. Those studies can be grouped into eight separate cate-GRAM REGARDING THE BEHAVIOR OF POLYMER BASE gones: (1) Regional and local earthquake activity were studied MATERIALS SU3JECTED TO BETA RADIATION.Votume 3:

and seismic hazard estimates made, an earthquake catalogue Phase-2b Expaaded Test Results.

BUCKALEW,W.H.;

was developed for the penod 1534-1977, and a seismometer CHENION.J.; CARLIN,F.; et al. Sandia National Laboratories.

array was installed in the Moodus, Connecticut seismic area. (2)

July 1989. 71pp. 8907280307. SAND 86-0366. 50667.087.

The causes of seismicity along tne northern Fall Zone from This document is the fmal account of a multiyear joint NRC/

New Jersey to Connecticut were investigated. (3) The level of CEA cooprative research program to investigate the relative activity of faults in the Champlain Hudson area were investigat-effectiveness of beta and gamma irradiations to produce ed. (4) Geologic mapping of fault systems was conducted in the damage in polymer base electncal insulation ar.d jacket maten.

Adirondack Mountains. (5) Geophysical studies were camed out als. Dunng the piogram, jointly executed by research laborato-in southern New England. (6) Geologic mapping was accom-nes in the U.S. Wandia National Laboratones) and France plished at selecteo locations to study Mesozoic and younger (Campagnie ORis Mdustrie), a number of matenal and radiation tectonics. (7) Bnttle fractures in bedrock were mapped within parameters were investigatea Results obtained were reason-areas of seismicity to determine if there is a relationship be-ably independent of the radiation parameters and most material

~

.-.. - - - =

l 8

Main Citations Cnd Abstracts

. parameters investigated and would suggest tw material NUREG/CR-4650 V6R1P1: ANALYSIS OF CORE DAMAGE FRE-damage resulting from electron beam and garamu ray irradia-OUENCY: GRAND GULF, UNIT 1 INTERNAL EVENTS.

tions can be correlated, within hmits, on the bees o! absorbed DROUIN,M.T.; LACHANCE.J.L; SHAPIRO B.J.; et al Science

.radiaton oose.

Apphcations internabonal Corp. (formerly Science Applications, Inc.). September 1989. 600pp. 8910060349. SAND 86-2084.

NUREG/CR-4550 V4 RIP 1: ANALYSIS OF COHE DAMAGE FRE-51404:049.

QUENCY: PEACH BOTTOM, UNIT 2, INTERNAL EVENTS.

This document contains the accident sequence analysis of in-KOLACZKOWSKI A. Science Apphcations Interna'ional Corp.

ternally initiated events for the Grand Gulf Unit 1 Nuclear (formerly Science Apphcahons, Inc.)

CRAMOND,W.R.;

Power Plant. This is one of the five plant analyses conducted as SYPE T,T. et al. Sandia National Laboratries. August 1989.

part of the NUREG-1150 effort for the Nuclear Regulatory Com-535pp. 8909260025. SAND 86-2084. 51276%8.

mission. The work performed and desenbed here is an exten.

This document contains the accident sequence analysis of in-sive reanalysis of that published in April 1987 as NUREG/CR-tornelly inibated events for the Peach Bottom, Unit 2 Nuclear 4550, Volume 6. It addresses comments from numerous review-Power Plant. This is one of the five plant enalyses conducted as ers and agnificant changes to the plant systems and proce-part of the NUREG 1150 effort for the Nuclear Regulatory Com-dures made since the first report. The uncertainty analysis and mission. The work performed and described here is an exten-Fese on of resuus am also @ hpW aM mh sive reanalysis of that published in October 1986 as NUREG/

able effort was expended on an improved analysis of loss of CR 4550, Volume 4. It addresser cemments from numerous re-offsite power. The content and detail of this report are directed viewers and significant changes to the plant systems and proce' toward Probabihstic Risk Assessment practitioners who need to know how the work was done and the details for use in further dures made since the first report. The uncertainty analysis and studies. The me:P core damage frequency is 4.0E-6 with 5%

presentabon of rasults are also much improved, and consider-able effort was expended on an improved analysis of loss of and 95% uncertainty bounds of 1.7E 7 and 1.2E-5, respectively.

Station blackout type accidents (loss of all AC power) dominate offsite power. The content and detail of this report are directed the overall results, contributing about 97% of the core damage toward PRA practitioners who need to know how the work was frequency. Anticipated transient without scram accidents con-

- done and the details for use in further studies. The mean core tributed another 3%. The numencal results are driven by loss of damage frequency is 4.5E 6 with 5% and 95% uncertainty offsite power, failure of the diesel generators, failure of the bounds of 3.5E 7 and 1.3E 5, respectively. Station blackout type steam-driven reactor core isolation cookng system, and accidents (loss of sh AC power) contributed about 46% of the common cause failure of the battenes, core damage frequency with Anticipated Transient Without Scrt.m (A1WS) occidents contributing another 42%. The numer.

NUREG/CR 4550 V6MtP2: ANALYSIS OF CORE DAMAGE FRE-ict.1 resutts are dfrven by loss of offsite power, transients with QUENCY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPEN-the power conversion system initially available, operator errors' DICES. DROUIN,M.T.; LACHANCE,J.L; SHAPIRO,B.J.; et al.

and mechanical failure to scram. Extemal events were also ana-Sc ence Applications Intemational Corp. (formerly Science Ap-ly24d using the intemal event fault tree and event tree models plications, Inc.). September 1989. 800pp. 8910060347, SAND 86-2084. 51401:272.

as a basis, and are reported separately in Part 3 of NUREG/

CR 4550, Volume 4, Revision 1.

This document contains the appendices for the accident se-Quence analysis of internally initiated events for the Grand Gulf NUREG/CR-4550 V4 RIP 2: ANALYSIS OF CORE DAMAGE FRE-Unit 1, Nuclear Power Plant. This is one of the five plant analy-QUENCY: PEACH BOTTOM, UNIT 2, INTERNAL EVENTS AP-ses conducted as part of the NUREG-1150 effort for the Nucle-PENDICES. KOLACZKOWSKI.A. Science Applications interna-ar Regulatory Commission. The work performed and desenbed here is an extensive reanalysis of that pudished in April 1987 tional Corp.

(formerly Science Applications, Inc.).

as NUREG/CR-4550, Volume 6. It addresses comments from CRAMOND W.R.; SYPE,T.T,; et al. Sandia National Laborato-numerous reviewers and significant changes to the plant sys-ries. August 1989. 752pp. 8909260256. SAND 86-2084.

$1307:306-tems and procedures made since the first report. The uncertain-ty analysis and presentation of results are also much improved, This document contains the appendices for the accident se-and considerable effort was expended on an improved analysis quence analysis of internally initiated events for the Peach of loss of offsite power. The content and detail of this report is Bottom Unit 2 Nuclear Power Plant. This is one of the five plant directed toward Probabilistic Risk Assessment practitioners who analyses conducted as part of the NUREG 1150 effort for the need to know how the work was done and the details for use in Nuclear Regulatory Commission. The work performed and de.

further studies. The mean core damage frequency is 4.0E-6 with scribed here is an extensive reanalysis of that pubhshed in Oc-5% and Of.% uncertainty bounds of 1.7E-7 and 1.2E 5, respec-tober 1986 as NUREG/CR 4550, Volume 4. It addresses com-tively. Sta. ion blackout type accidents (loss of all AC power) ments from numerous reviewers and significant changes to the dominate the overall results contributing about 97% of the core plant systems and procedures made since the first report. The damage frequency. Anticipated transient without scram acci-uncertainty analysis and presentation of results are also much dents contnbuted another 3%. The numerical results are driven improved, and considerable effort was expended on an im-by loss of offsite power, failure of the diesel generators, failure proved analysis uf loss of offsite power. The content and detail of the steam-driven reactor core isolation cooling system, and of this report is directed toward PRA practitioners who need to common cause failure of the batteries.

Know how the work was done and the details for use in further NUREG/CR 4639 V5P2R2: NUCLEAR COMPUTERIZED Li-studies. The mean core damage frequency is 4.5E-6 with 5%

BRARY FOR ASSESSING REACTOR RELIABILITY and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively.

Station blackout type accidents (loss of all AC power) contribut-(NUCLAAR). Data Manual.Part 2: Human Error Probability (HEP) ed about 46% of the core damage frequency with Anticipated Estimates. GERTMAN.D.1,; GILBERT B.G.; GILMORE,W.E.; et al. EG&G Idaho, Inc. (subs. of EG&G, Inc.). July 1989. 964pp.

Transient Without Scram (ATWS) accidents contnbuting another 8908070421. EGG-2458. 50810:270.

42%. The numencal results are dnven by loss of offsite power, This volume of a five-volume senes summarizes those data f

1 transients with the power onversion system initially available, currently resident in the first release of the Nuclear Computer-I operator errors, and mechanical failure to scram. Extemal ized Library for Assessing Reactor Rehabihty (NUCLAP.R) data events were also analyzed using the internal event fault tree base. The raw human error probabihty (HEP) and hardware and event tree models as a basis, and are reported separately component failure data (HCFD) contained herein are accompa-in Part 3 of NUREG/CR-4550, Volume 4, Revision 1.

nied by a glossary of terms and the HEP and hardware taxono-l

.--,-.l.-.

';,,'l,,,i...-".

,,;'-l,-

^,,,,

'l, Miln Citations and Abstracts 9

mies used to structura the data. Instructons are presented on NUREG/CR-4674 V08: PRECURSORS TO POTENTIAL SEVERE how the user may navigate through the NUCLARR data man.

CORE DAMAGE ACCIDENTS:1967 A

STATUS aDement system to find anchor values to assist in solving nsk-REPORT. Appendixes B C, And D.

MINARICK,J.W.;

related oroblems. Volume 5: Data Manual will be updated on a HARRIS J.D.; CLETCHER,J.W.; et al. Oak Ridge Nabonal Labo-periodic bania so that nsk analysts Whout access to a comput.

ratory. July 1989. 744pp. 8908070438. ORNL/NOAC-232.

er may have access to tne lats NCLARR data. Those users 50806:023.

wishing to learn more regarditg W omputer based interactive See NUREG/CR-4674,V07 abstract.

search and report-generatior. capabilities of the NUCLARR NUREG/CR-4706 V03: PROGRESS IN EVALUATION OF RADIO-system are referred to tha other volumes in the NUREG/CR-NUCLlDE GEOCHEMICAL INFORMATION DEVELOPED BY 4639 series, e.g., Volume 1: Summary Descripton or Volume 4:

DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE User's Guide.

June 1989.

PROJECTS. Report For October 1987 MEYER,R.E.; ARNOLD,W.D.; O'KELLEY,G.D.; et al. Oak Ridge NUREG/CR-4639 V5P3R2: NUCLEAR COMPUTERIZED Ll-BRARY FOR ASSESSING REACTOR RELIABILITY Natonal Laboratory. August 1989. 63pp. 8909180318. ORNL/

(NUCLARR). Data ManualPart 3: Hardware Component Failure n

ation the s being developed by projects within the De-Data (HCFD). GERTMAN,0.l.: GILBERT,B.G.; GILMORE,W.E.;

partment of Energy (DOE) pertinent to the potenbal geochemi-et al. EG&G Idaho, Inc. (subs. of EG&G. Inc.). July 1989.

cal behavior of redonuclides at candidate sites for a high-level 1,530pp. 8008080266. EGG-2458. 50779:258.

radioactive weste repository is being evaluated by Oak Ridge See NUREG/CR-4639,V05,P02,R02 abstract-Natonal Laboratory (ORNL) for the Nuclear Regulatory Com-NUREG/CR-4639 V5P4R2: NUCLEAR COMPUTERIZED Li' th opo ig le er w ea ucc ounta BRARY FOR ASSESSING REACTOR RELIABILITY Nevada. The pnncipal emphasis was on column studees of mi-(NUCLARR). Data Manual.Part 4: Summary Aggregations.

grabon of uranium and technetium in water from well J-13 at the GERTMAN,D.L; GILBERT,B.G.; GILMORE,W.E.; et al. EG&G Yucca Mountain site. The effects of flow rate and temperature Idaho, Inc. (subs. of EG&G, Inc.). July 1989. 470pp.

on uranium migration were studied. Sorpton ratos calculated 8908080322. EGG-2458. 50832:004.

from the elution peaks became larger as the flow rate de-See NUREG/CR-4639,V05,P02,R02 abstract.

creased and as the temperature increased. These observations support the conclusion that the sorption of uranium is kinetically NUREG/CR-4667 V06: ENYlRONMENTALLY ASSISTED CRACK-hindered. Batch sorption ratio experiments as a funcbon of time ING IN LIGHT WATER REACTORS.

Semiannual confimied that the reaction was slow because 20 to 30 days Report. October 1987 - March 1988.

SHACK,W.J.;

elapsed before sorption ratios reached steady state values. A KASSNER,T.F.; MAlYA,P.S.; et at Argonne National Laboratory, preliminary column experiment was completed under conditens August 1989. 66pp. 8908310080. ANL-89/10. 51075:001, simulating unsaturated flow in tuff for transport of Sr(2+),

This report summarizes work performed by Argonne National Cs(+), and TcO4., The significance of these experiments with Laboratory on environmentally assisted cracking in light water respect to data obtained by DOE investigators for evaluation of reactors dunng the six months from October 1987 through the suitability of the Yucca Mountain site is discussed.

March 1988. The stress corrosion cracking (SCC) of Tvpes 304, 318NG,347, and CF3M cast stainless steels (SSs) was investi.

NUREG/CR 4744 V02 N2: LONG TERM EMBRITTLEMENT OF Gated by means of slow-strain-rate, crevice-bent beam, and CAST DUPLEX STAINLESS STEELS IN LWR fracture-mechanics crack-growth-rate tests in high-temperature SYSTEMS. Semiannual Report, April-September 1987.

water. The effects of temperature and water chemistry on the CHOPRA,0.K.; CHUNG,H.M. Argonne Nabonal Laboratory, crack growth behavior of Type 316NG and sensitized Type 304 August 1989. 57pp. 8909120073. ANL-89/6. 51169:213.

SS were determined in long-term fracture-mechanics tests. The This progress report summarizes work performed by Argonne influence of dissolved copper and organic impuribes on the SCC National Laboratory on long-term embrittlement of cast duplex of sensitized Type 304 SS was also invesbgated. Fatigue tests stainless steels in LWR systems during the six months from are being conducted on Type 316NG SS in air at room temper.

April to September 1987, Microstructural studies were conduct-ature to nrovide baseline data for companson with results that ed to investigate the kinetics of spinodal decompositen and G-precipitation in CF4 and CF 8M grades of cast will be obtained in high temperature water. The investigation of phase and y,l. The results indicate that the presence of Mo in stainless stee the susceptit:llity of several heats of different grades of low-alloy femtsc steels to transgranular SCC in slow strain-rate tests CF-8M steel accelerates spinodat decomposition as well as G-was continued' phase and y, precipitation. Examination of the long-term-eged CF-8M steels also revealed a "spinodal-like" decomposition of NUREG/CR-4674 V07: PRECURSORS TO POTENTIAL SEVERE the austenite caused by segregation of Fe and Ni in the matrix.

CORE DAMAGE ACCIDENTS: 1987 A STATUS REPORT. Main Preliminary results indicate that local regions of austenite are Report And Appendix A.

MINARICK,J.W.; HARRIS,J.D.;

significantly hardened by the decomposition. Charpy-impact, CLETCHER.J.W.; et al. Oak Ridge National Laboratory. July tensile, and J-R curve data are presented for several heats of 1989.136pp. 8908070433. ORNL/NOAC-232. 50805:001, cast stainless steels aged at temperatures between 320 and 450 degrees C for times up to 10,000 h. The results indicate Forty'eight operational events ranking 1.0 x 10(-6) or higher at cmcmahs d ca@n and nege M N SW aM N and 16 lesser events, reported in 3490 licensee event reports (LERs) and revisions and occumn9 at commercial light. water re-ling low-temperature embnttlement. The existing correlations for actors dunng 1987, are considered to be precursors to potential estimahng the extent and kinetics of embrittlement do not accu-severe core damage. These are desenbed along with associat-rately represent the properties of different grades and composi-ed significance estimates, categortzation, and subsequent analy-tions of cast stainless steel after thermal aging.

ses. This study is a continuation of earlier work, which evaluat-ed the 19691981 and 19841986 events. The report discusses NUREG/CR-4918 V03: CONTROL OF WA1ER INFILTRATION (1) the general rationale for this study, (2) the selection and INTO NEAR SURFACE LLW DISPOSAL UNITS. Progress documentaton of events as precursors, (3) the estimation and Report. SCHUL 2,R.K.

Califomia, Univ. of, Berkeley, CA.

use of conditional probabilities of subsequent severe core RIDKY,R.W.

Maryland. Univ. of, College Park, MD.

damage to rank precursor events, and (4) the plant models O'DONNELL,E. NRC No Detailed Affiliation Given. August used in the evaluat on process.

1989. 30pp. 8908250012. 51018:154.

l

10 Main Cit tions cnd Abstracts The project objectives is to assess means for controlling NUREG/CR 4977 V01: SHAG TEST SERIES. Seismic Research water infiltration through waste disposal unit covers in hunid re-On An Aged United States Gate Valve And On A Piping System gions. Experirrental work is being performed in large scale tysi-in The Decommissioned Heissdampfreaktor (HDR):Sammary.

meters (7'xd5'z10') at Beltsviile, MD. anG resutts of the asseas-STEELE,R.; ARENDTS,1G. EG&G Idaho, Inc. (subs. of EG&G, ment are opphcable to disposal of LLW, uranium mill tailings, Inc.).

• Idaho National Engineering Laboratory. August 1989.

hazardous waste, and sanitary landfills. Three concepts are 69pp. 6909120202. EGG-2505, 51171:292.

under investigation: (1) resistive layer t' amer, (2) conductive The Idaho National Engineenng Laboratory (INEL) participat-layer bamer, and bioengineering water management. The resis-ed in an internationally sponsored seismic research program tive layer bamer consists of compacted earth (clay). The con-conducted at a decommissioned experimental reactor facility, ductive layer bamer is a special case of the capillary barrier and the Heissdamptreaktor (HDR), located in the Federal Republic it requires a flow layer (e g. fine sandy loam) over a capillary of Germany (FRG). The research program included the study of break. As long as unsaturated conditions are maintained water the effects of excitation, produced during a simuiated seismic is conducted by the flow layer to below the waste. This barrier event, on (a) the operability and integrity of a naturally aged 8-la most efficient at low flow rates and is thus best placed below in. motor operated gate valve installed in the Versuchskreislauf a resistive layer bamer, Such a combination of the resistive (VKL), an existing piping system ln the HDR, (b) the dynamic re-layer over the conductive layer bamer promises to be highly of-epons6 of the VKL and the operabihty of snubbers, and (c) the fective provided there is no appreciable subs.dence. Bioengin-dynamic responses of various piping Support systems installed eering water management is a surface cover thet is designed to on the VKL. The INEL work, sponsored by the U.S. Nuclear accommodate subsidence. It consists of impermeable panels Regulatory Commission (USNRC), contnbutes to earthquake in-which enhance run off and limit infiltration. Vegetation is planted vestigations being conducted by the Kemforschungszentrum in narrow openings between panels to transpire water from Karlsruhe (KfK) and is part of the general HDR Safety Program below the panels. This system has successfully dewatered two performed in behalf of the FRG, Federal Ministry for Research fysimeters thus demonstrating that this procedure could be used and Technology. This report presents the results of the KfK.

for remedial action (" drying out") existing water-logged dmpos-designated SHAG (Shakergebaude) test series; the:e are the al sites at low cost.

first in situ experiments involving an actual nuclear power plant and a full scale piping system under simulated seismic loading, NUREG/CR-4949: SOURCE TERM CALCULATIONS FOR AS-Volume l presents a summary of the tests and results, and SESSING RADIATION DOSE TO EQUIPMENT. DENNING,R.S.;

Volume 11 contains appendices that present details and specif-FREEMAN {ELLY; CYBULSKIS,P.; et al. Battelle Memorial in-ics of the tests and results of Volume 1.

stitute, Columbus Laboratones. July 1989. 64pp. 8909200246.

BMi-2153. 51274:286.

NUREG/CR-4977 V02: SHAG TEST SERIES. Seismic Research This study examines resofts of analyses performed with the On An Aged Uruted States Gate Valve And On A Piping System Source Term Code Package to develop updated source terms in The Decommissioned Heissdampfreaktor (HDR): Appendices.

using NUREG 0956 methods. The updated source terms are to STEELE.R.; ARENDTS,J.G. EG&G Idaho, Inc. (subs. of EG&G, be used to assess the adequacy of current regulatory source Inc.).

• Idaho National Engmeenng Laboratory. August 1989.

terms used as the basis for equipment qualification. Time de-139pp. 8909120157. EGG-2505. 51171:153.

pendent locational distributions of radionuclides within a con-See NUREG/CR-4977,V01 abstract.

tainment following a severe accident have been developed. The Surry reactor has been selected in this study as representative NUREG/CR 5004: RESOLUTION OF RECURRING LOSS of PWR containment designs. Similarly, the Peach Bottom reac-ALARMS. SMITH,B.W. Battelle Memorial Institute, Pacific North-tor has been used to examine radionuclide distributions in boil-west Laboratory. HEBBLE T.L.; EHINGER,M.H.; et al. Oak

= ing water reactors. The time dependent inventory of each key Ridge National Laboratory. August 1989. 84pp. 8909120062, radionuclide is provided in terms of its activity in curies. The PNL-6240. 51169:279.

data are to be used by Sandia National Laboratories to perform Patterns in material control and accounting data may indicate shielding analyses to estimate radiation dose to equipment in a loss of nuclear material. Inadvertent MC&A system errors, each containment design. See NUREG/CR-5175, " Beta and biases, or changes in the rate of holdup accumulation are nor-Gamma Dose Calculations for PWR and BWR Containments."

mally the cause of pattems in MC&A data. This report describes a program for resolving pattems of safeguards significance in NUREG/CR-4967: NUCLEAR PLANT AGING RESEARCH ON MC&A data and provides some examples of the activities. Res.

HIGH PRESSURE INJECTION SYSTEMS. MEYER,LC EG&G olution of pattems in process monitonng data used for MC&A idaho, Inc. (subs. of EG&G, Inc.).

• Idaho National EngineerinQ focuses on systematic characteristics of the process and may Laboratory. August 1989. 152pp. 8909120121. EGG-2514.

involve large quantities of data. Resolution consists of localizing 51145:179-the potential causes by detailed anatyses and tests, investigat-d This report represents the results of a review of lght water ing potential causes, and taking corrective actions to prevent re-1 reactor High Pressure injection System (HPIS) operating experi-currence.

ences reported in the Nuclear Power Experience Data Base, Li-consee Event Reports (LER)s, Nuclear Plant Reliabihty Data NUREG/CR-5085: PROGRESS IN DEVELOPMENT OF A METH-

{

System, and plant records. The purpose is to evaluate the po-ODOLOGY FOR GEOCHEMICAL SENSITIVITY ANALYSIS tential significance of aging as a contributor to degradation of FOR PERFORMANCE ASSESSMENT. Parametric Calculations, the High Pressure injection System. Tables are presented that Preliminary Databases, And Computer Codo Evaluation.

show the percentage of events for HPIS classified by cause, SIEGEL,M.D.; LEIGH,C.d Sandia National Laboratories. August component, and subcomponents for PWRs. A representative 1989.113pp. 8909110389. SAND 85-1644. 51145:004.

Babcock and Wilcox plant was selected for detailed study. The The purpose of the Geochemical Sensitivity Analysis Project U.S. Nuclear Regulatory Commission's Nuclear Plant Aging Re-is to develop a methodology to ioentify physicochemical and hy-search guidelines were followed in performing the detailed study drogeological conditions wherein use of a simple retardation

(

that identifies materials susceptible to aging, stressors, environ-factor will lead to underestimation or overly conservative esti-

' mental factors, and failure modes for the HP'S. In addition to mation of the cumulatrve radionuclide discharge over the

~ the engineering evaluation, the components that contnbuted to 10,000-year regulatcry period. This report describes activities j

system unavailabihty were determined and the aging contribu-frorn the initiation of the Geochemical Sensitivity Analysis tion to HPIS unavailability was evaluated. The unavailability as-Project in Apnl 1984 to September 30,1985. During this first sessment utilized an existing probabilistic risk assessment phase of the project, compilation of necessary geochemical (PRA), the linear aging model, and generic failure data.

data was started, radionuchde transport codes were evaluated

)

o

1 Main Citations and Abstracts 11 and parametric methods to a6sess the significance of matrix dif-ging criteria for steam generators. This report is Volume 1 of fusion, colloidal transport and reaction kinebcs on integrated re-two volumes documenting the results of analyses performed to dionuchde discharge from high-level waste repositones were de-evaluate and compare candidate samphng/ inspection schemes veloped. Although available data are inadequate to support final for in-service inspection of steam generator tubes. The estimat-conclusions for site specific conditions, the method of sensitivity ed EC inspection rehability determined from round robin exami-analysis being developed in this project can be used to guide nation of the Surry generator was used to guide the selection of i

future data collection.

input parameters for these analyses. Analytical evaluations of l

candidate sampling / inspection schemes based on the ranges of N',,AEG/CR 5152: COMPARISON AND REGULATORY IMPACT pr bability of detection and probability of exceeding the EC OF NOA 1 AND NOA-2 WITH N45.2 OA STANDARDS.

plugging hmit, as estimated from the round robin data, are de-SCANGA.B.: STOKLEY,J. Science Applicatonw international scribed. Monte Carlo simulation analyses designed to supple-j Corp. (formerty Science Apphcotions, Inc.). July 1989. 52pp.

ment the analytical results and test key assumptions are pre.

8908070426. SAIC 88/3015. 50808:047.

sented. Finally, conclusions and recommendations for further This report compares the 1983 edition of ANSI /ASME NOA 1 study are presented, j

and the 1986 editon of NOA.2 standards with the ANSI N45.2 senes of OA standards. NOA 1 (1983) has been endorsed by NUREG/CR 5174: A REFERENCE MANUAL FOR THE EVENT NRC by Revision 3 of Regulatory Guide 1.28. NOA.2 has not PROGRESSION ANALYSIS CODE (EVNTRE).

yet been formally endorsed by the NRC. Where differences be-GRIESMEYER,J.M. Sandia National Laboratories. SMITH,L.N.

tween the two sets of standards exist, the authors have as-Science Applcations International Corp (formerly Science Ap-i sessed the impact of these differences on the regulatory proc-plications, Inc.). September 1989. 220pp. 8909290240.

ess. The analysis considers both NRC Regulatory Guide en.

SAND 88-1607. 51332:027.

dorsements and current NRC emphasis on the performance This document is a reference guide for the Event Progression based inspection process. The reviewers find that the NOA Analysis (EVNTRE) code developed at Sandia National Labora-standards are more focused on the technical aspects of prod-tories. EVNTRE is designed to process the largo accident pro-ucts and processes to achieve quality as compared to N45.2 grossion event trees and associated files used in probabilistic standards emphasis on program verification activities. In con-nsk analyses for nuclear power plants. However, the general trast to NOA 1, NOA 2 has incorporated many lessons leamed nature of EVNTRE makes it apphcable to a wide variety of anal-dunng plant construction experience and the operations phase yses that involve the invesbgation of a progression of events of nuclear facihties. The summary of the impact on regulatons which lead to a large number of sets of conditions or scenarios, that is presented In this report provides a comparison for utilk The EVNTRE code efficiently processes large, complex event ties, vendors, and NRC inspectors of existing standards.

tree branch points in several different ways, to classify path-NUREGICR 5155: THE THERMAL INSTABILITY OF CESIUM ways or outcomes into user-specified groupings, and to sample IODIDE. ELRICK,R.M.; MERRILL R.M.; OUELLETTE,A.L Sandia input distributions of probabilities and parameters.

National Laboratories. August 1989, 110pp. 8908300116.

NUREG/CR 5175: BETA AND GAMMA DOSE CALCULATIONS SAND 881187,51053:214.

FOR PWR AND BWR CONTAINMENTS. KING,0.B. Sandia Na-F ecent data indicate cesium iodide can react with stainless steer in a high temperature steam and hydrogen environment. A tional Laboratories. July 1989. 225pp. 8908150075. SAND 88-j quan%tative model proposed for this reachon suggests that the 1605. 50889:024.

decomposition of Cal depends strongly on the surface area of Analyses of gamma and beta dose in selected regions in steel, the reaction time and the concentrations of Cal vapor and PWR and BWR containment buildings have been performed for hydrogen. A two-level factorial matnx of ten experiments was a range of fission product releases from selected severe acci-designed to exploru the stahstical significance of the effect of dents. The objective of this study was to determine the radiation these four variables on the instabihty of Cst Each test was di.

dose that safety related equipment could experience during the vided into time-quarters that simulated parttCular aspects of a selected severe accident sequences. The resulting dose calcu-degraded core accident. Steel oxidation dunng the first quarter lations demonstrate the extent to which design basis accident conditioned the system for Cst interact on wita oxidized surfaces qualified equipment would also be qualified for the Sevote accb dunng the second quarter -similar to the initia' release of radion.

dent envircnments. Surry was chosen as the representative uclides. Here, Col behavior was consistent with predictions of PWR plant while Peach Bottom was selected to represent the physical model, in the third quaiter edditional hydrogen was BWRs. Battelle Columbus Laboratory performed the source added-corresponding to Zircaloy oxidation, then cut off during term release analyses. The AB epsilon scenario (an intermedi-

't the fourth quarter-similar to cessation of Zircaloy oxidahon. in-ate to large LOCA with failure to recover onsite or offsite electri-stability behavior during the third and fourth quarters, dissimilar cal power) was selected as the base case Surry accident, and to model predichons, was attributed to a fundar1 ental change in the AE scenario (a large break LOCA with one initiating event the surface oxide wtsch was reduced during the addition of hy, and a combination of failures in two emergency coohng sys-drogen and reoxidized in the fourth quarter: Empincal model tems) was selected as the base case Peach Bottom accident.

Radionuclide release was bounded for both scenarios by includ-equations are presented for Csl behavior dunng the three quar-ters.

Ing spray operation and arrested sequences as variations of the base scenanos. Sandia National Laboratones used the source NUREG/CR-5161 V01: EVALUATION OF SAM 0 LING PLANS terms to calculate dose to selected containment regions. Sce-FOR IN-SERVICE INSPECTION OF STEAM GENERATOR narios with sprays operahonal resulted in a total dose compara-TUBES.Modehng Of Eddy-Current Reliabihty Data, Analytical ble to that (2.20 x 10(8) rads) used in current equipment quahfi.

Evaluations, And initial Monte Carlo Simulat ons BOWEN W.M.:

cation testing (for design basis accidents or Regulatry Guide HEASLER P.G.; WHITE,R.B. Battelle Memorial Institute. Pacific 1.89). The base case scenarios resulted in some calculated Northwest Laboratory. April '989.130pp. 8907250300. PNL-doses roughly an order of magnitude above the current 2.20 x 6462. 50607.251.

10(B) rad equipment qualification test regicn The Steam Generator Group Project (SGGP) was a mulb-task effort using the retired-from-service Surry 2A pressunzed water NUREG/CR 5225 ADD 01: AN OVERVIEW OF BWR MARK l reactor steam generator as a test bed to investigate the reliabil-CONTAINMENT VENTING RISK IMPLICATIONS An Evaluation sty and effectiveness o' in-service nondestructve eddy current Of Potential Mark l Containment Improvements. WAGNER,K.C.;

(EC) inspection equipment and procedures. The information thai DALLMAN R.J.; GALYEAN.W.J. EG&G Idaho, Inc. (subs. of was developed provided the technical basis for recommenda-EG&G Inc.). July 1989. 12 Bop. 8908140139. EGG 2548.

tions for improved in-service inspection methocs and tube plug.

50880:046.

i it Mein Citations and Abstracts 7th nout supp6emeds contaenment "enting rek evaluabons consequence projectons. The model was developed to attow pew N for the Mara l Containment Performance improve-considerston of the dominant aspects of source term, transport, mt 4 rV4) Program. Quantitatrve evaluatons using simplified dose, and consequences. The model graphew were designed

%nwnment eart trees for staton blackoet sequences were for use on the COMPAQ Portable til microcomputer used by the perforrnd *o idei4 potental nok redecton offered by con-NRG response personnel wto respond to the emergency. How.

ta, ament WW en improved automatic depressunzabon ever, the model can be run on any DOS system. The model re-s)ttom with a ded1stwd power source, and an aeditonal supply sutts can be displayed as text or in graphcal form. The develop-of water to either tic coa.tainment sprays or the vessel with a ment of the modelis ongoing. Subsequent versions are planned dedcated power source. The rak calculations were based on to: (1) show the provous cases to be stored (2) include fuel the Draft NUREG 1150 results for Peach Bottom with selected cycle faciEss and containment monitor reading; in the source enhancements. Several sensitMty ciudes were performed to in-term methoes, (3) include terrain and bulkiing wake in the trans-vestigste phenc nenolog6 cal, operatonal, and equipment per.

port and diffuson methods and (4) ethmate exposure rates, formance uncertaintes. Quahtettve risk evalueuons were provid-od for loss of lonD. term containment heat removal and antci-NUREG/CR 5363: THE RISK MANAGEMENT IMPLfCATIONS OF pated transients without scram for the same set of improvo.

NUREG 1150 METHODS AND RESULTS. CAMP,A.L -

monts. A limited discussion to provided on the generic applca-MALONEY,4.J.: SYPE.T.T. Sande National Laboratones. Seh bility of these results to othe7 plants with Mark t conta6nments.

tember 1989.17Bpp.8909290255 SANDBB 3100. $1332:247.

This report tzacnbet the potental uses of NUREG 1160 and NURfC/CR4tse: VISCOPLASTIC STRESS. STRAIN CHARAC.

similar Probabihste Risk Assessments (PRAs) in NRC and in-TERtZATION OF A533 GRADE B CLASS 1 STEEL dustry rmk management programs. NUREG 1150 uses state-of-GIOVANOLA.J.H.; KLOOP,R.W. Oak Ridge National Laboratory, the-art PRA techniques to estimate the risk from frve nuclear September 1989. 68pp. 6909260217. ORNLSUB87SA1931-power plants. The methods and results produced in NUREG-

$1292:320, 1150 provide a framework within whch current nok manage-The Heavy Socton Steel Technology (HSST) Program is in-ment strateges can be evaluated, and future nok management vestigetsng the role of nonlinear rate-dependent effects in the programs can be developed and assessed. While the develop-Interpretaten of crack run arrest events in ducti6e matenals ment of plant-specific risk management strategies is beyond the through thL 1 velopment and apphcation of viscoplaste dynam-scope of this document, examples of the use of NUREG 1160 ic fracture analysis techruques. The report describes studies framework for 6dentifying and evaluabng risk management op-that determined the viscoplashc matenal responsa of a heat of tons are presented. All phases of risk management from pre.

A533 grade B class 1 pressure vessel steel over a range of venton of initiabng events through reducten of off-site conse-strain rates and temperatures. Hopi'enson torsion bar experi-quences are discussed, with particular attention given to the monts were conducted at temperatures of.60, 20, and 150 de-early phases of ecce6nts.

grees C and at engineenny shear strain rates of 400,150,and 3000 s(1). The ressts of these expenments are presented in NUREG/CR 5273 V01:

SCDAP/RELAPS/ MOD 2 CODE the form of stress-strain cc ves, stress-time curves, strain rate.

MANUAL.RELAP5 Code Structure, System Models, And Solo time curves, and strain rate-strain curves. For all strain rates in, uon Methods. ALLISON.C.M; BERNA,G.A.: CHENG T.C.; et al.

vestigated in this study, the A533 grada 9 class i steel has a EG&G 6daho, Inc. (subs. of EG&u, Inc.). September 1989, significant temperature sonsttivity; for all temperatures, the ma.

40Bpp. 8909260136. EGG 2555. 51275 020.

tenal also exhibits a small, yet clearly measurable, strain rate The SCDAP/RELAPS code has been developed for best-esti-sensitivity in the range frem 400 to 3000 s( 1) engineenng shear mate transient s:mulation of hght water reactor cCWant systems strain rate. The torsonal data compiled in this study are shown dunng a severe accident. The code models the coupled behay-to be in good agreem6nt with previously measured NA strain.

ior of the reactor coolant system, the core, and the fission prod-rate tensile data over the same temperature range.

ucts and aerosols in the system during a severe accident tran-NUMEG/CR n231: COBALT 60 SIMUtnON OF LOCA RADI.

setnt as well as large and small break loss-of coolant accidents, operatonal transients such as anticipated transicat without ATION EFFECTS. BUCKALEW W.H. Seadic Natonal Laborato-SCRAM, loss of offsite power, loss of feedwater, and loss of ries. July 1989. 37pp. B907250326. SAND 661054. 50607:001, fio.t A generic modeling approach is uLed that permits na much The consequences of simulahng nuclear reactor loss-of-cool-of a particular wtem to be modeled as necessary. Control ant accident (LOCA) radiation effects with Cobalt-60 gamma ray system and secondary system components are included to triadiators have been lnvesbgeted. Based on radiaton Induced permit modehng of plant controls, turbines, condensers, and damage in polymer base matenais, it was demonstrated that secondary feedwater conditoning systemL The modehng theory e!ectron/ photon induced radiaton damage could be related on and associated numerical schemes are documented in Volumes the basis of r.orage abso@d radiabon dose. This result was 1 and 2 to acquaint the user with the modeling base and mus used to aahmate the relative eMectrveness of the mixed beta /

aid in effectrve use of the code. Volume 3 contains detailed in-gamma LOCA and Cobalt-60 radiation environments to damage struebons fut code spplication and input data properation, in ad-both barv and lacketed polymer base eNtrical insulation mate-ditson, Volume 3 contains user guidehnes that have evolved r als. From the resutts obtained, it is concluded that present over the past several years from apphcahon of the RELAP5 and simutaton techniques are a conservative tnethod for simulating SCDAP codes at the Idaho National Engineenng Laboratory, at LOCA radiation effects and that the practices have probably other natonal laboratones, and by users throughout the world, substantially overstressed both bare and jacketed matonals dur:ng qualifical;on testing.

NUREG/CR 5273 V02:

SCDAP/RELAPS/ MOD 2 CODE MANUALSCDAP Code Structure, Models, And Soluton Meth.

NOREQ/CR 524h RASCAL VERSION 1.3 USER'S GUIDE.

A7 HEY,G.F.; SJOREEN.A.L.; MCKENNA.T.J. Oak Ridge Naton-ods. ALLISON.C.M.; BERNA G.A.; CHENG,T.C.; et al. EG&G Idaho, Inc. (subs. of EG&G, Inc.). September 1989. 333pp.

al Laboratory. Sootember 1989.175pp. 8910100289. ORNL/

TM 10955. 5140h300 8909260121. EGG 2555. 51279.248.

See NUREG/CR.5273,V01 abstract.

A new dost: vesessment system,' RASCAL, has been devel-opea for the U.S. Nuclear Regulatory Commissic 1 (NRC) for NUREG/CR 5273 V03:

SCDAP/RELAPS/ MOD 2 CODE i

uso dunng response to emergencies. The model is designed to MANUALUser's Guide And input Requirements. ALLISON C.M.;

provida a rough com;anson with EPA Protective Acton Guid-BERNA.G.A.; CHENG,T.C.; et al. EG&G Idaho, Inc. (subs. of l

ance and thresholds for acute health offects. RASCAL wai be EG&G inc.). September 1989. 503pp. 8909260107. EGG 2555.

used by the NRC personnel who repo, to the site of a nuclear

$1278.105.

eccedent to conduct an independent evaluato'. of dose and c>ee NUREG/CR 5273,V01 abstract.

'f i

1 Main Citations and Abstracts 13 i

NUMEQ/CR 62th CLOSEOUT OF NRC BULLETIN 87 01: THIN-3 facihty in Moi, Belgium, are described. Compansons of trans-j NING OF PIPE W ALLS IN NUCLEAR POWER PLANTS.

Port calculatons using the method and many measurements in-1 FOLEY,W.J.; DEAN.R.S.; HENNICK A.

PARAMETER, Inc.

volving nckel, indium, and aluminum dosimaters indcate agree-August 1909. 56pp. 8908230236. PARAMETERNRC17P.

ment generatty to within 5% tf effects of inaccuracies m the do i

509B4:268.

simeter cross sectons are minimtred and proper onentaton of

]

Documentation is provided in this report for the closeout of the coordinate system used in the synthesis procedure is ob-NRC Bulletin 87 01 regarding monitoring the thickness of pipe served. These conobsons bode well for the success of this and component walls in hig4 energy single-phase and two*

method in solving neutron transport proolems involving the use phase carbon steel piping systems of nLclear power plants.

of PLSAs in light water reactors to reduce core leakage in pres-Both safety related and non-safety related systems are included sunted thermal shock programs. A second report describing the in the monttonng program. The bulletin required five (5) actons expenmental details of the measurements will serve as compan-by bconsees and holders of constructon permits. Complete ton documentaton to this one and will be furnished by the Stu-cloteout of the bulletin for 119 affected facilmes and conclu-discentrum voor Kernenergie/ Centre d' Etude de rEnergie Nu-sons are based on Gerenc Letter (GL) 89 08. This GL requires c% aire, M, BNgium wntten assurance that the beensing basis t,Ttinues to be met by high-energy carbon steel piping system subject to erosion /

NUREG/CR 6339: DATA

SUMMARY

REPORT FOR FISSION corrosion. Background informaton is provided in the introduc.

PRODUCT RELEASE TEST VI 1, OSBORNE.M.F.;

n and Appen@x A COLLINS.J.L; LORENZ,R.A.; et al. Oak Ridge Natonal Labora-NUMEG/CR 6290: CLOSEOUT OF IE BULLETIN 79-28.POSSI-tory. June 1989. 68pp. 6907250320. ORNL/TM-11104.

DLE MALFUNCTION OF NAMCO MODEL EA180 LIMIT 50607:039.

SWITCHES AT ELEVATED TEMPERATURES. FOLFY,W.J.;

The first in a senes of high-temperature fission product re-DEAN,R.S.; HENNICK.A. PARAMETER, Inc. August 1989.23pp-lease tests in a new vertcal test apparatus was conducted in 8903120095. PARAMETER IE181, 51170:278.

flowing steam. The tost specimen was a 15.2 cm-long secton Documentation is provided in that report for the closeout of IE of a fuel rod from the Oconee 1 PWR, it had been irradiated to Bulletin 7b28 tegarding the possible malfuncton of NAMCO a burnup of ~42 mwd /kg. Using an induction f urnace,it was Model EA180 timit switches en safety related systems at elevat-heated under simulated LWR accident conditons 20 min at ed temperatures. Closeout is based on the implementaton and 2000 K and 20 min at 2300 K. In a hot cell-mounted test appa-venficaton of four required actions by holders of operating b-retus. Posttest inspection showed severe oxidation but only conses or construction permits for nuclear power fac6tities. Eval-minimal fragmentaton of the fuel specimen; cladding melting untion of utihty responses and NRC/ Region inspection reports was apparent only Dear the top end. Based on fission product irtcates that the bulletin is closed for all of the 119 facihties to inventories measured in the fuel and/or calculated by ORIGEN, which it was issued for acton. It is concluded from the results that although the potential problem was widespread, supporting analyses of test components showed total releases from the documentation shows that the bulletin concern has been re.

fuel of 47% for (85)Kr 33% for (125)Sb,37% for (129)l,64%

solved. Background informaton is supphed in the introduction for (110m)Ag, and 63% for (137)Cs. Large fractions (36% and and Appendix A.

30%, respectivety) of the released (110m)Ag and (125)SD were retaiM in me fumace above M fud % test and posnest NUREG/CR 5322: DETECTION AND CHARACTERl2ATION OF analysis of the fuel specimen indcated a (134)Cs release of INDICATIONS IN SEGMENTS OF REACTOR PRESSURE VES-65%, which is very good agreement with the (137)Cs value. The SELS. COOK,K.V.; CUNNINGHAM,R.A.; MCCLUNG,R.W. Oak first phase (2000 K) of this test was conducted at the same Ridge Natonal Laboratory. August 1989. 33pp. 8908250014.

conditions as test HL2, However, the releases of Cs, I, and Kr ORNL/TM-11072. 41018:121, were much less than those in Hi-2. Variations in fuel character-Studies have been conducted to estimate 9ew density in seg, istes (H.B. Robinson in test Hi 2 vs Oconee in test VI 1) are be-

[

ments cut from hght water reactor 1 LWR) pressure ve=sels as hoved to be major factor.iin this difference, and the large axial l

part of the ORNL's HSST Program. Objectives were to +aluate crack that occurred in Hl.2 probably was a significant factor, LWR segments for flaws with uttrasonic and hqu'd puaetrant techniques and to compare the results with current assumptions NUREG/CR-5340: DATA

SUMMARY

REPORT FOR FISSION related to probabiliste risk assessment. Bott. objectives were PRODUCT RELEASE TEST VI-2.

OSBORNE,M.F.;

successfully completed. One significant indicaton was detected COLLINS.J.L; LORENZ,R.A.; et al. Oak Ridge National Labora-h a Hope Creek boiling water reactor seam weld by ultrasonic techniques. This lndication (with a through-wall dimension of 6 tory. September 1989. 69pp. 8910100307. ORNL/TM 11105.

mm) was detected in only 3 m of weldment and offers extreme-

$1409:065.

ly hmited data when compared to the extent of welding even in The second in a senes of high-temperature fission product re-a single pressure vessel. The Pilgrim pressunzed water reactor lease tests in a new vertcal test apparatus was conducted. The

=

segments contained relatively httle weldment, thus, we hmited 15.2 cm-long test specimen had been irradiated to a burnup of our ultrasonc examinatons to the cladding and subcladding re-M2 mwd /kg. Denng an induction funwce and col 6ection appara-gions. Only one indcaton of note was detected ultrasonically (11 tus mounted in a het cell, it was heated under simulated LWR was below recordable size). Fluorescent hauid penetrant inspec-acc6 dent conditons for 60 min at 2300 K in flowing steam. Post-tion of the Cladding surfaces for both LWR segments detected test inspection showed severe oxidation of the cladding but only no signifcant indcatons (for a total of 6.8 m(2) of cladding sur-minimal fragmentation of the Zirceloy cladding. Based on fission L

face). The detecten of the discontinuities in the arbitrarity se-product inventories in the tuel, total releases of 30.7% or lected sections imphes that the Marshall report estimates (and (85)Kr. 8.93% for (110m)Ag 68.2% for (125)Sb, and 63.4% for others) are nonconservative for such small flaws. However.

(137)Cs were measured. Smaller release fractions for many since the flaw in the Hope Creek weld is subsurface, it would other fission products and fuel material uranium and plutonium have httle influence on calculated failure probabikty.

. were determined also. Total mass release from the fumace to NUREG/CR-5338: ANALYSIS OF THE VENUS-3 EXPERIMENTS.

the collecton system was 1.08 g Compenson of the results MAERKERAE. Oak Ridge National Laboratory. August 1989.

from this test showod reasonable agreement with the results 74pp.8909120082. ORNL/TM 11106. $1170:143.

from two earlier testa at similar conditions. The release rate co-The results of applying a hybnd superposit on-synthesis calcu-efficients from the six tests in 'he H1 senes and tests VI-1 and latonal method to a mockup of a three dimensional geometry VI 2 fait significantly below a wide'y accepted standard used in involving a panial length shield assembly (PLSA) at the VENUS-LWR safety analysts.

m

14 Main Citations and Abstracts NUREQ/CR 5348: MAN MACHINE INTERFACE ISSUES IN NU-computer code BLT (Breach, Leach, and Transport), a modifca-CLEAR POWER PLANTS. Report On A Workshop Held On Jan-ton of FEMWATER, has been selected to prer,ct the processes vary 1012,1989. DEBOR.J.; SWEZEY,R. Science Appications of container degradation (Breach), r:ontaminant release from the intomational Corp. (formerly Science Appucatons, Inc.). July waste form (Leach), and contaminant migraton (Transport). In 1989. 46pp. 8908150110. SAIC-89/1114. 508BB 219.

conjuncton, these two codes have the Capabihty to account for The U S. Nuclear Regulatory Commission Offee of Nuclear the effects of disposal geometry, unsaturated / saturated water Regutatory Research held a workshop on nuclear power plant flow, container degradaton, waste form leaching, and mig'aton man. machine entericce issues in Bethesda, Maryland, January of contaminants released within a single disposal trench. In ad-10 to 12,1989. The objective of the workshop was to address dit on to the input requirements. this report presents the funda-potential ettects of Introducing advanced technology nuclear mental equations and relationships used to model the four dif-power plant systems on plant operators and system interface forent processes previously discussed. Further, the appendees design. The workshop was attended by approximately 50 lead-provide a representative sample of data required by the differ-ing experts representing a broad range of applicable disciplines.

ent models.

The workshop resulted in proposed man machine interface en-tens and guidelines, as wull as ventscation and validation experi-NUREQ/CR 63BB: STEEL IMPURITY ELEMENT EFFECTS ON monts.

POSTIRRADIATION PROPERT!ES RECOVERY BY NUREQ/CR 6363: A STUDY OF THE USE OF CROSSLINKED ANNEALING. Final Report. HAWTHORNE J.R. Matenals Engi-HIGH-DENSITY POLYETHYLENE FOR LOW LEVEL RADIOAC.

neenng Associates, Inc. August 1989.185pp. 8908300054.

TlVE WASTE CONTAINERS. BOO,P.; ANDERSON.C.I.;

ME A 2354. 51053.029.

CLINTON,J.H. Brookhaven Natonat Laboratory. June 1989.

The influence of copper content and phosphorus content on 141pp.8907250307, BNL NUREG 52196. 50607:110.

notch ductility recovery by 399 degrees C postirradiaton heat A comprehensive research program has been completed to treatment was explored for A 533-B and A 302 B pressure evaluate the effects of chemcal and gamma trradiation environ-vessel steels. Charpy.V (C(v)) Specimens for the investigaton ments on the mechancel properties of crosslinked high density were obtained from ten plates produced from four (4.way split) polyethylene. The studies. included uniaxial creep tests and laboratory melts. The plates were heat treated to reproduce the crack initiaton and crack propagation studies in statically-microstructure of 150-mm and thicker A 302 B plates at t,ie stressed U bend samples. From the results obtained, standard quarter thickness location. Specimens were irradiated at 288 testing protocols were recommended for quantifying the various degrees C (550 degrees F) to a fluence of 2.5 x 10(19) n/cm(2) failure modes whch could be present in this matenal during in the UBR test reactor. Other specimens were thermally condh servce as a low level waste container.

tioned at 288 darees C and at 288 degrees C followed by 399 NUREQ/CR 6367: COMPARISON OF STRONGLY HEAT DRIVEN degrees C (550 degrees F) to a fluence of 42.6 x 10(19) n/cmt2)

FLOW CODES FOR UNSATUPATED MEDIA.

sence of 6rradiat on. A 0.30% copper content was found to have UPDEGRAFF,C.D. Gram, Inc. ' Sandia National Laboratones.

a sigNficant influence on the magnitude of residual embnttle.

August 1989.127pp. 8908230226. SAND 88-7145. 50984:141, ment after a 168-hour anneal. In contrast, phosphorus contents Under the sponsorship of the U.S. Nuclear Reguistory Com.

In the range of 0.002 to 0.025% were not found to have an mission, Sandia Natonal Laboratones (SNL) is develop ng a effect oa notch ductahty recovery for either a high or a low performance assessment methodology for the analysis n long-cooper content. Essentially full recovery in 41 J transiten tem-r term disposal of high-level radioactive waste (HLW) in usatu-perature was observed for the high phosphorus, low copper rated welded tuff. As part of this effort, SNL evaluated existing content steel versions. A detnmental influence of a 0.68%

strongly heat driven flow computer codes for simulating ground-nickel content on recovery behavior is shown by the data for water flow in unsaturated media. The three codes tested, the 0.28% copper content plates but not the 0.16% copper NORIA, PETROS, and TOUGH, were compared against a suite content plates.

of problems for which analytical and numencal solutions or ex.

=

penmental results exist. The problems were selected to test the NUREG/CA-6390: ROCK MASS MODIFICATION AROUND A NU.

abilities of the codes to simulate situatons ranging from simple, CLEAR WASTE REPOSITORY IN WELDED TUFF. MACK,M G.;

uncoupled processes, such as two-phase flow or heet transfer, BRANDSHAUG.T,; BRADY,B.H. Itasca Consulting Group, Inc.

to fully coupled processes, such as vaportzation caused by high August 1989.112pp. 8909180264. 51211:285, tsmperatures. In general, all three codes were found to be diffe-This report presents the results of numencal analyses to esti-cult to use because of (1) built-in time stepping critona, (2) the mate the extent of rock mass modifcation resulting from the treatment of boundary conditions, and (3) handling of evapora-presence of a High Level Wasto (HLW) repository. Changes in tion /condensaton problems. A drawback of the study was that rock mass conditions considered are rock stresses and nt de-I adequate problems related to expected repository conditions formations resulting from disposal room excavation and from were not available in the literature. Nevertheless, the results of the heat generated by nuclear waste. Rock properties and site this study suggest the need for thorough investigations of the conditions were taken from the Site Charactenzation Plan Con-impact of heat on the flow field in the veinity of an unsaturated ceptual Design Report (MacDougall et al.,1987) for the poten-HLW repository. Recommendations are to develop a new flow tial repository site at Yucca Mountain, Nevada. Analyses were code combining the best features of these three codes and conducted using boundary element and distinct element meth-eliminating the worst ones-ods. Room-scale models and repository scale models were in-NUREG/CR 5387: LOW LEVEL WASTE SHALLOW LAND DIS-vestigated up to 500 years after waste emplacement. The ther.

POSAL SOURCE TERM MODEL: DATA INPUT GUIDES.

moelaste boundary element analysis indcates that a modified SULLIVAN.T.M SUEN,C.J. Brookhaven National Laboratory.

zone of rock develops around the disposal rooms for both verto Juty 1989.

325pp.

8907280333.

BNL-NUREG-52206.

cat and horizontal waste emplacement. This zone is estimated 50666:001, to extend a distance of roughty two room diameters from the This report provides an trput guide for the computational room surface. The repository scale analysis using the thermoe-models developed to predict the rate of radionuchde rettese lastic boundary element code, and the distinct element code, in-from shallow land disposal of low-leve. waste. Release o' con-deates a modified zone of rock starting approximately 100 m taminants depends on four processes: water flow, container above and below the repository honzon, with a thickness of ap-degradation, waste from leaching, and contaminant transport.

proximately 200 m above and 150 m below the repository.

The computer code FEMWATER has been selected to predct These estimates reflect uncertainties and simphtying assump-the movement of water in an unsaturated porous media. The tons introduced in the analysis.

j

Main Citations and Abstracts 15 NUREG/CR 5335 V09: MULT1 LOOP INTEGRAL SYSTEM TEST models is developed through an "lterative" approach of model (MIST) FINAL

REPORT, inter Group Comparsons.

companson to rock mass response. This program approach is GLOUDEMANS.JA Babcock & Wilcox Co. July 1989. 297pp.

opposed to one 6n which point" tesbng is conducted "a prori" 8909110309. EPRl/NP-6480. 51144:067.

at determined locations in an attempt to statistically specify The Muttiloop Integral System Test (MIST) is part of a multi-given rock properties.

phase prog'am started in 1983 to address small break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wiicox NUREQ/CR-5401: BOND STRENGTH OF CEMENT BOREHOLE designed plants. MIST is sponsored by the U.S. Nuclear Rego-PLUGS IN SALT. AKGUN.H.; DAEMEN.J.J.K. Anzona. Univ. of, latory Commission, the Babcock & Wilcox Owners Group, the Tucson, AZ. July 1989. 246pp. 8908070429. 50805:137.

Electrc Power Research institute, and Eabcock and Wilcox.

This report presents results of a senes of push out tests per-The unique features of the Babcock and Wilcox design, specife-formed on cement plugs emplaced in satt cores. Two types of Caily the hot leg U bends and steam generators, prevented the cement have been tested in satt cores drilled from blocks ob-use of existing integral system data or existing integral facilities toined from a mine in a bedded satt formaton. Expenments on to address the thermal hydraulic SBLOCA Questons. MIST and rock (satt) bndges provide a relerence basis. Test durations two other supporting facilities were specifically designed and vary from minutes to 2 days. A rather broad range of results constructed for this program, and an existing facility-the Once was obtained, depending pnmarily on cement type. Averago lhrough Integ'al System (OTIS)-was also used. Data from interface shear strengths, assuming uniform shear stress distri-MIS 1 and the other facilities will be used to benchmark the ade-butons, range from about 2 to 12 MPa. Peak shear stresses at quacy of system codes, such as RELAP5 and TRAC, for pro-failure, calculated assuming an elastic exponential shear stress dicting abnormal plant transients. The individual tests are de-distributon along the 6nterface, are up to more than four times scribed in detail in Volumes 2 through 8, Volume it, and are higher. Considerable differences are observed within each test summarized ln Volume 1, Inter-group compensons are ad-group (standard deviations commonly are 20% and more). Dis-dressed herein. These compansons are grouped as follows:

solutoning along the interface has been observed in a signife-mapping versus SBLOCA transients, SBLOCA, pump effects, cant number of samples. It is postatated that this dissolutioning and the effects of noncondensible gases. Appendix A provides may have been enhanced by dispersion of clay inclusons within an index and description of the microfiched plots for each test.

the salt. Visible dissolutoning appears to be related to reduced which are enclosed with the corresponding Volumes 2 through bond strengths. Eighteen samples have been submitted to

8. The discharge ortfice charactonstics are tabulated in Appen-short-term (up to 34 days) drying and rewetting (for 8 to 28 dix B.

days) after they had been pushed to fallute (slip). Depending NUREQ/C45596: FINANCIAL IMPACT OF IMPLEMENTING upon the reconditioning history, retesteo strengths range from DRAFT ANSI STANDARD N13.30, PERFORMANCE CRITERIA 18 to 105% of the initial strength, with an average value of FOR RADIOBIOASSAY, TRAUB R.J.; MACLELLAN,J.A. Battelle about 60%.

D NUREG/CR 5402:

CRUSHED SALT CONSOLIDATION.

7pp 890 0 0'. PNL 6924 0887 1 In order to establish standards of bioas' ay performance upon U DA Anzma, W d, Msm, E M s

1989.121pp. 8907280316. 50666:326.

which a uniform natonal program of performance testing might be based, the Health Physics Society StandarJs Committee The objective of this study is to investigate the time depend-ent cmsolidation of confined crushed satt under axial loading.

(HPSSC) formed Working Group 2.5 to prepare the draft ANS.I The crushed salt is geometrically charactenrod in terms of its Standard N13.30, " Performance Critena for Radobioassay..

Because the U.S. Nuclear Regulatory Commssion (NRC) is grain size distributon and by means of particle shape daterms-nations. Crushed salt of a selected grain tire distnbution is em-considinng whether to require that all bionssay service labora-tones meet the critoria of the draft Standard, the NRC staff re, placed in a steel cylinder, and an axial load is applied to the salt. The deformation is monitored. The axial load is increased quested that the Pactfic Northwest Laboratory (PNL) estimate in steps. Results are presented as void ratio vs. time curves, for I

the costs that may be incurred in implementing the draft Stand-ard. Two types of laboratones were involved an the cost study:

the varous applied stresses. Air dned samples have a bone contet of 0.2N moistened samples have a brine content rang-service laboratones and a performance testing laboratory. Cost estimates based on responses to questonnaires sent to seven ing from 1.75 to 8.8%. The general consolidaten trend can be facilities performing radiobioassays vaned in relation to the represented as a linear function of the loganthm of time, but extent of each facility's radiobionssay program, and their per.

with noticeable temporary acceleratons. Consolidation time cal-cepton of their readiness for accreditation implementation of Culations, predicted from the results of fitted Curve parameters, ac reditaten was estimated to cost each facility between a few admittedly extrapolated far beyond the measurement range, in-dicate that reconsolidahon of the matenals as emplaced and thousand dollars and over one quarter million dollars, depending on the facility type. Likewise, annual costs ranged from a few tested here wouVi require many centunes. It deserves emphasis hundred dollars to $170,000 For all but one facility annual that these expenments were initiated on an extremely loose crushed salt fill. Consolidation rate increases with bnne content.

costs were estimated to be less than $25,000, in addition to these costs, start up costs for the testing laboratory were esti.

The dependency upon applied stress remains ambiguous. The mated to be approximately $322,000, obtained void reduction rates are significantly lower than pub-lished results determined from isostatic consolidation. Given the NUREG/CR 5400: BASIS FOR IN SITU GEOMECHANICAL small number of tests completed, all results and conctusions TESTING AT THE YUCCA MOUNTAIN SITE. BOARD,M. Itasca must be considered preliminary. Tests in configurations that Consulting Group, Inc. July 1989. 115pp. 8907280389.

may be more representative for expected in-situ consolidation 50664 207, conditons are recommended.

This report presents an enalysis of the in-situ geomechanical testing needs for the Yucca Mountain site, and discusses a pos-NUREG/CR 5415: ENGINEERING DESIGN FOR THE NRC OP-Sible testing program whrch addresses these needs. The testing ERATIONS CENTER.

PIERCE,W.R.;

BAVIER R.N.;

needs are denved from 10CFR60 regulations, as well as simple BORLAND R.S.; et al. Science Applications intematonal Corp.

thermomechanical canister and room scale studies The testing (formerly Science Applicatons, Inc.). August 1989. 84pp.

approach suggested here is based heavily on demonstraton 890918028b. SAIC-89/1478. 51211:121.

and monitonng of full scale rock mass response. This is t.ccom-A design study of the NRC Operatons Center was initiated to plished through instrumented excavaton of exploration dnfting anticipate the movement of the Center from its present Iccaton throughout the repository honzon, as well as thermal loading of en Bethesda, Maryland to the White Flint Two building in Rock-fulbsCale repository excavations. ConfiJence in pred4Clive ville, Maryland. Based on a review of documentation on the Op-1

/

93 Main Citations and Abstracts oraton Center functons, interviews wrth selected personnel, and "hmit" values grven in the Nevada Nuclear Waste Storage and observaton of an emerDency exercise, the design team de-Investigaton (NNWST) Project Site Charactenraton Plan Con-veloped requirements for a new Center. The requirements basis ceptual Design Report. The ground reacton curve concept ts in-consisted of commurucatons diagrams documenting the infor-troduced to study the potential liner loading resutting from ther-maton f6ow witNn the Operatons Center, an ad a'ancy matrix, mally induced borehole closure. Analytcat solutons for varous and resutts from the interviews. Froni ihs basis a new concep-nng loadings were combined to develop solutons for appropri-tual layout.as developed for the Center, consistent with the ate liner loading configuratons. Results are presented in terms constraints imposed by the proposed new building. Work sta-of dimensionless bending stress versus flexibilitv rato.

tons were defined for the staff, both for day to day operations and for emergency conditons. Recomreendations are included NUREQ/CR 5428: VARIATION OF HEAT LOADING FOR A RE-ooncoming the use of supporting equipment currently in the POSITORY AT YUCCA MOUNTAIN. BRANDSHAUG T. fiasca Center as well as equipment planned for future use.

Consulting Gro>p, Inc. September 1989. 67pp. 8909200193 51293.082.

NUREQ/CR4425: EVALUATION OF ALLOWED OUTAGE TIMES (AOTS) FROM A RISK AND RELIABILITY STANDPOINT.

This report presents the results of numerical analyses te de-VESELY W.E. Science Applicatons Intemational Corp. (formerty termine the range in container pitch (i.e., the spacing petween vertcally emplaced containers), doposal room extracton rato, Scence Applicatons, inca

• Brookhaven National Laboratory, and waste stand-off distance that will satisfy design entena ed Aug e 1989. 74pp. 8909180260. DNL NUREG-52213.

$1210008.

pressed for a repository at Yucca Mountain. Effects are investe-This report describes the basic risks which are associated Dated for a range in thermal properties of the rock represented by the " saturated" and " dry" condit ons expressed in Chapter 2 with allowed outage times (AOTs), defines strategies for select-of the SCPCDR. A number of heat transfee analysos were per-ing the nsks to be quantified, and describes how the nsks can be quantifed. The report furthermore desenbes Crnena consid-formed for a time penod of 50 years after inrtial waste emplac-mont. Within this penod, temperatures have peaked in the vicini-eratons in determining the acceptability of calculated AOT noks, ty of the waste containers. Th6 anaiyses thcluded three dimen-and discusses the merits of relative risk criteria versus absolute sonal heat transfer models that account for the explcit interAt-nak critena. The detailed evaluatons which are involved in cal-culating AOT nsks, including uncertainty consideratons are also ton of single waste containers emplaced in a repository panel.

Vertical and honzontal waste emplacement concepts of com-discussed. The report also describes the proper ways that risks from multiple AOTs should be considered so that nsks are prop-mingled SF and DHLW were investigated. The analyses indcate that the configuration of container boreholes and extracton erty accurnulated from proposer! multiple AOT changes, but are ratio, as well as the stand-off distance to waste proposed h the not double-counted. Generally, everage ACT nsks which inctode SCPCDR, Chapter 4, could result in the development of tem-the frequency of occurrence of the AOT need to be accumulat-peratures that exceed design goals currently expressed in the ed but single downtime nsks don't since they apply to individual SCP and the SCPCOR.

AOTs.

NUREG/CR-5426: EXAMINATION OF THE USF. OF CONTINUUM NUREG/CR 5431: A REVIEW OF GEOSCIENCE CHARACTERIS.

VERSUS DISCONTINUUM MODELS FOR DESIGN AND PER.

TICS AND DISPOSAL EXPERIENCE AT THE COMMERCIAL FORMANCE ASSESSMENT FOR THE YUCCA MOUNTAIN LOW LEVEL RADIOACTIVE WASTE DISPOSAL FACILITY SITE. BOARD.M. Itasca Consulting Group, Inc. August 1989.

NEAR WEST VALLEY, NEW YORK. SMOOT,J.L. Battelle Me-80pp. 8909120079. 51170.060.

morial Institute, Pacifc Northwest Laboratory August 1989.

This report examines the use of continuum versus discontin-81pp. 8909180271, PNL-6970. 51211:205, uum-based numencal raodels for geomechanical design and The West Valley Commercial Low Level Radoactrve Waste performance assessment studies at the Nucca Mountain site.

disposal site is located about 48 km south of Buffalo, New York.

Welded tuff is a hard, heavity-fractured rock. The Topopah Operation of the site began in 1961 by Nuclear Fuels Service Spnngs formaton appears, from examination of vertcal bore, and was terminated in 1975. The disposal trenches at the site holes, to have a preferred vertcal #nt onentaton. Vanous con, are excavated about 5 m into glacial till that has a thickness of statutive models for representing bnted rock behavior in contin-about 28 m. About 65,000 m(3) of the waste containing approxi-uum-based models s's reviewed, as well as the discontinuum mately 710,000 C6 were disposed at the site dunng the oper-method. It is concluded that the rock mass is amenable to a atonal period. Ground-water movement through the till to pre-continuum plasteity formutation whch takes into account the dominantly downward as indicated by measurements and nu-anisotropy introduced by a dominant nt set. Numerous fuutt merical simulaton of hydraulic head. Radionuclidos do not structures cross the site. The stability of these structures under appear to have migrated more than 3 m either laterally or verti.

thermal seismic loading and their effects on drift stability, seal Cally from the waste dispor at trenches. Numerical simulatons of performance and fluid transport has not yet been 6nvestigated.

(3)H. (90)Sr, and (14)C mystion are able to reproduce the ob.

A true discontinuum model allowing dyname stress analysis is served concentration in the till beneath selected trerches. Un.

needed in this Case certainty remains with respect to the Continuity and hotorogene-Itw of the hydrostratigraphc units and the spatial distributon of NUREG/CR 5427: ANALYSIS OF EMPLACEMENT BOREHOLE hwiraute conductrvity and effective porosity. More work is ROCK AND LINER BEHAVIOR FOR A REPOSITORY AT c.beded to better dehne the waste inventory and any long-term YUCCA MOUNTAIN. LORIG,L.J.; DASGUPTA.B. Itasca Consutt-changes that might be expected. Erosion poses a potential 6ng Group, Inc. September 1989. 139pp. 8909260226.

threat to the long-term integnty of the disposal area.

51292:181.

That report presents the results of studies aimed at assessing NUREG/lA 0014: ANALYSIS OF THE THETIS BOILDOWN EX-the quasi-stats behavior of both the rock surrounding an em-PERIMENTS USING RELAPS/ MOD 2.

CROXFORD,M.G.;

placement borehole and the lining within an emplacement bore-HALL,P.C. United KMgdom. Govt. of. July 1989. 51pp.

hole for a nuclear waste repository in tuff. Two-dimensional 8908080281. GD/PE Nr576. 50830:067.

thermomechanical analyses of conditions similar to those repre-To test the ability of RELAPS/ MOD 2 to model two phase mix-sentative of the hortrontal emplacement option were performed ture level and fuel rod heat transfer when the core has become i

using a distinct element code. Three different behavior models partially uncovered, post test Calculations have been camed out 1

(equivalent continuum, wedge, and parallel nt) were used to of a series of boildown tests in the AEEW THETIS out.of pite f

investigate the state of deformation at 0 and 100 years follow-test faciltty. This report desenbes the Companson betwoon the ing waste emplacement. Three different rock strength assump-code calculations and the test data. Excellent agreement is ob.

tions were studied corresponding to " design," " recommended" tained with mixture level boildown rates in tests at pressures of

Main Citations and Abstracts 17 40 bar and 20 bar. However at pressures be.Jw 10 bars the number of vanables avadable from the measurements dunng boildown rato ts considerably overpred.cted. A general tendency the experiment. The break mass flows were generally underpre-for RELAP5/ MOD 2 to overpredet voed fracton below the two-deted at the same time as the depressurizaton rate was over-1 phase mixture level is observed, whch is traced to defects an predcted.

the interphane drag modets within the code. The heat up of ex-posed rods above the two phase mixture level is satisfactonty NUREG/lA 0017: ASSESSMENT OF TRAC PF1/ MOD 1 AGAINST calculated by the code. The resutta support the use uf A LOSS OF GRID TRANSIENT IN RINGHALS 4 POWER RELAP5/ MOD 2 for analysis of high pressure core boildown PLANT. SJOBERG.A.

ALMBERGER.J.: SANDERVAG.O.

1 events in PWRs.

Sweden, Govt. of. July 1989. 55pp. 8938080299. STUDS-VIKNP87/10. 50830.244.

NUMEG/lA 0015: ASSESSMENT OF INTERPHASE DRAG COR-A loss of gr6d transient in a three loop Westinghouse PWR RELATIONS IN THE RELAP/5 MOD 2 AND TRAC PF1/MODt has been simulated with the frozen version of TRAC PF1/

CODES. ARDRON.K.H.; CLARE A.J. United Kingdom, Govt. of.

MOD) computer code. he resuus tweal N capaWy of the July 1989. 31pp. 8908080290. GD/PE N/557. 50830:120.

code to quahtatively predet the different pertinent phenomena An assessment is carned out of the interphase drag correls-tions used in modelling vertical two-phase flows in the ad-and the data comparison was quite encouraginD. Accurate pre-venced thermal hydraulic codes RELAPS/ MOD 2 and TRAC-detions of the system response required careful determinston PF1/ MOD 1. The assessment ts performed by using code of the boundary conditions simulating the turbine govemot models to calculate void tracton in fully developed steam-water valves and steam dump valves behavor. An explcit modehng of flows, and companno results with predctions of standard corte-the steam generetor internals was also found to be important lations and test data. The study is restncted to the bubbly and for N resuus. H was also rweeled that N pressuds system slug flow regimes (void fractons below 0.75). The numercal re-including spray and heaters and their operaton should be mod-suits given in this paper allow a rapid estimate to be made of eled in some detall for proper response.

void traction errors hkely to anse in a particular code apphcaton NUMEG!lA-0020: ASSESSMENT STUDY OF RELAP5/ MOD 2, due to defceencies in interphase drag modelhng.

CYCLE 36.04 BASED ON SPRAY START UP TEST FOR NUREG/lA 0018: ASSESSMENT OF RELAPS/ MOD 2, CYCLE DOEL 4. MOEYAERT,P.: STUBBE.E. Belgium, Govt. of. July 36.04 AGAINST FIX Il GUILLOTINE BREAK EXPERIMENT NO.

1989. 421pp. 8908080303. 50830.303.

5061. ERIKSSON.J. Sweden, Govt. of. July 1989. 85pp.

This report presents a code assessment study for the 8908080294. STUDSVIKNP86109. 50830.154.

HELAPS/ MOD 2 code using a pressunzer spray start-up test for The FIX Il guillotine break experiment No. 5061 has been the DOEL-4 power plant. The pressunzer plays a vital role in analyzed using the RELAPS/ MOD 2 code. The code version determining the plant behavior dunng transients and hence cor-used. Cycle 36.04, is a frozen version of the cod 6. Four differ.

rect simulaton of the pressurtzer 6s essential to the simulation ont calculatons were carried out to study the sensittvity of initial of plant behavior duhng off normal Conditons. This full scale coolant mass, junction operations and break discharge hne no.

study addresses pressunzer thermal hydrauhes, surge hne ther-dalization. The differences between the calculations and the ex-mal hydrauhes, structural heat transfer, asymmetric loop behav-periment have been quantified over intervals in real time for a ior, and single phase pump behavior, i

i 1

1 m

s

4 m-

Secondary Report Number Index This index lists, in alphabetical order, the performing organization-issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross-referenced to the NUREG number for the report and to the 10-digit NRC Document Control System accession number, SECONDARY REPORT NUM94.It MEPORT NUMBEM DECONDARY MdPORT NUI4DEM REPORT NUGAGEM ANL 89/10 NUREG/CR4667 V06 OANL/NSIC 200 NUPEG/CR-2000 V0B N7 ANL 89/6 NUREG/CR4744 V02 N2 ORNL/NSC200 NURfG/CR 2000 VO8 N8 BAW-2021 NUREGICR4395 V09 ORNL/TM-10147 -

NUREG/CR-4708 V03 h#Mjf fC

'y BMt-2153 NUREG/CR4949 BNL NUREG 51454 NUREG/C42331 V00 N4 ORNL/TM-11104 NUREG/ R'4339 BNL NUREG41454 NUREG/CR-2331 V09 N1 ORNL/TM-11105 NUREG/ R4340 i

BNL-NUREG42196 NUREG/C45363 ORNL/TM 11106 NUREG/ R4338 BNL NUREG42208 NUREG/C45387 ORNL/TM-DA93 NUREG/CR4219 V06 N1 BNL NUREG 52213 NUREG/CR4425 ORNLSUB87SA1931 NURE G/CR4226 LOG-2458 NUREG/C44639 V5P2R2 PARAMET ER IE181 NURE G/CR-5290 EGG 2458 NUREG/GR4639 V5P3R2 PARAMETERNRC178 NURE G/CR4287 EGG-2458 NUREG/CR-4639 V5P4R2 PNL 6240 NUREG/OR 5004 h

6 V01 E GG-2505 NUREG/CR-4977 V02 EGG 2505 NUREG/C44977 V01 PNL4970 NURE G/C45431 EGG 2514 NURE G/CR-4967 SA60 88/3015 NUREG/CR 5152 EGG-2548 NUREG/CR4225 ADD 01 SAIC-89/1114 NUREG/C45348 EGG-2555 NUREG/CR 5273 V03 SAIC 89/1478 NUREG/CR *2415 EGG 2555 NUREG/CR4273 V02 SANDe51644 NUREG/CR 5085 EGG-2555 NUREG/CR427J VD1 SAND 66 0366 NURE G/CR 4530 V03 EPRl/NP 6480 NUREG/CR4395 V09 SAND 86-2064 NUREG/CR 4550 V4R1P1 GD/PE.N/557 NUREG/lA 0015 SAND 86 2084 NUREG/CR-4550 V4 RIP 2 GD/PE N/576 NUREG/LA 0014 SAND 66-2084 NURE G/CR 4550 V6R1P2 L B 78 28 NUREG/CR4290 SAND 66-2084 NUREGICR-4550 V6 RIP 1 jfy h

f$

B 87-001 NUREGICR4287 IPSN 1/88 NUREG/C44530 V03 SANDBB 1605 NUREG/C45175 MEA 2354 NUREG/CR 5388 SANDB81607 NUREG/CR4174 ORNL-6170 NUREG/CR-4234 V02 sANoog.3100 NURE G/CR-5263 ORNL/NOAC 232 NUREG/CR-4674 V07 SANDBB 7145 NUREG/C45367 ORNL/NOAC 232 NUREG/CR 4674 V08 STUDSVIKNP86109 NUREG/lA 0016 ORNL/NSIC 200 NUREG/CR $000 V0B N6 STUDSVIKNP87/10 NUREG/lA@17 r

19 k

i l

I ms

l -

Personal Author index This index licts the personal authors of NRC staff, contractor, and international agreement l

reports in alphabetical order. Each name is followed by the NUREG number and the title of I

the report (s) prepared by the author, if further information is needed, refer to the main cita-l tion by the NUREG number, i

L NUREG/CR 5273 V03. SCDAP/RELAP5/ MOD 2 CODE MANUALusers l~

AKQUN.H.

NUREG/CR 5401: BOND STRENGTH OF CEMENT BOREHOLE PLUGS Gude Arad input Requirements.

}

IN SALT, o

ALL190N,C.M.

NUREG/CR-4674 V07: PRECURSORS TO POTENTIAL SEVERE CORE NUREGICR-6273 V01:

SCDAP/RELAP5/ MOD 2 CODE DAMAGE ACCIDENTS: 1967 A STATUS REPORT. Main Report And

-wr "h.U/[

j MANUALRELAP5 Code Structure, System Models, And Solution Meth-A v A.

X ods NU R-4674 V08: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR 6273 V02: SCDAP/RELAP5/ MOD 2 CODE MANUALSCDAP DAMAGE ACCIDENTS 1967 A ST ATUS REPORT.Appendmes B.C. And 44'

. M"-

Code Structure, Mocols, And Soluten Methods.

D.

NUREG/CH-5273 V03: SCDAP/RELAPS/ MOD 2 CODE MANUALUsers Gude And input Requirements.

90ARD M.

HUREG/CR-5400: BASIS FOR IN. SITU GEOMECHANICAL TESTING AT ALM 9ERGER.J.

THE YUCCA MOUNTAIN SITE NUREG/lA4017: ASSESSMENT OF TRAC PFl/ MOD 1 AGAINST A NUREG/CR 5426. EXAMINATION OF THE USE OF CONTINUUM LOSS OF GRID TRANSIENT IN RINGHALS 4 POWER PLANT-VERSUS DISCONTINUUM MODELS FOR DESIGN AND PERFORM-ANCE ASSESSMENT FOR THE YUCCA MOUNTAIN SITE.

ANDERSON.CJ.

NUREG/CR-5363. A STUDY OF THE USE OF CROSSLINKED HIGH-SORLAND,R.S.

DENSITY POLYETHYLENE FOR LOW LEVEL RADIOACTIVE WASTE NUREGICR-5415: ENGINEERING DESIGN FOR THE NRC OPER-CONTAINERS.

ATONS CENTER.

ARDRON.K.H.

BOWEN,W.M NUREG/lA4015: ASSESSMENT OF INTERPHASE DRAG CORRELA-NUREG/C55161 V01: EVALUATION OF SAMPLING PLANS FOR IN-TIONS IN THE RELAP/5 MOD 2 AND TRAC-PF1/ MOD 1 CODES.

SERVICE INSPECTION OF STEAM GENERATOR TUBES Modeling Of Eddy Current Relatutity Data, Analytical Evaluatens And Instel Monte g

Anggoyg,J g, Carlo Simulatons.

NUREG/CR 4077 V01: SHAG TEST SERIES.Seamic Research On An Aged Urvied States Gate Velve And On A Piping System in The De-DRADY,BH NUR 77 VO S7AG ESIS Emic Research On An NUREG/CR 5390: ROCK MAS $ MODIFICATION AROUND A NUCLEAR WASTE F2POSITOPY IN WELDED TUFF.

A9ed United States Gate Valve And On A Pipeng System in The De-commesoned HeesdamptreeMor (HDR) Appendices.

BRANDSHAUG,T.

NUREG/CR.5390: ROCK MASS MODIFICATON AROUND A NUCLEAR ARNOLD.W.D.

WASTE REPOSITORY IN WELDED TUFF.

I NbnEGICR-4708 VO*: PROGRESS IN EVALUATION OF RADIONU-MEWh28: umAhm W.M WM M A M%

CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-TORY AT YUCCA MOUNTAIN.

LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For October 1967 June '989-DROOKS.S.G.

NUREG 0713 V08: OCCUPATIONAL rat 4ATION EXPOSURE AT COM-ATHEY G.y MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES NUREG/CR 5247: RASCAL VERSION 1.3 USER'S GUIDE, FOR 1986. Nineteenth Annual Report I

AUSTIN.P.N.

BUCKALEW.W.H.

NUREG/CR 4674 V07: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-4530 V03: U.S./ FRENCH JOINT RESEARCH PROGRAM DAMAGE ACCIDENTS: 1987 A STATUS REPORT. Main Report And REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-Appenda A.

JECTED TO BETA RADIATIONVolume 3: Phase-2b Expanded Test NUREG/CR-4674 V08: PRECURSORS TO POTENTIAL SEVERE CORE MAGE ACCIDENTS:1987 A STATUS REPORT.Appendmes B,C, And NUR CR-5231: COBALT-60 SIMULATION OF LOCA RADIATON EF.

FECTS.

BAROSH.P.J.

90CKLE Y,J.T.

(

NUREG/CR-3252: NEW ENGLAND SEISMOTECTONIC STUDY ACTIVI.

NUREG 1973 TECHNICAL POSITON ON POSTCLOSURE SEALS.

TIES DURING FISCAL YEAR 1980.

BARRIERS AND DRAINAGE SYSTEM IN AN UNSATURATED BAVIER.R.N.

MEDIUM.

I NUREGICR-5415: ENGINEERING DESIGN FOR Thf NRC OPER-CAMP,A.L.

ATONS CENTER NUREG/CR 5263 THE RISK MANAGEMENT IMPLICATIONS OF BELL.L.

NUREG 1150 METHODS AND RESULTS.

NUREG-1275 Yo5 ADDt OPERATING EXPERIENCE FEEDBACK REPORT PROGRESS IN SCRAM REDUCTON Commercel Power CARLIN,F.

NUREG/CR-o30 V03 U SJFRENCH JOINT RESEARCH PROGRAM Reactors.

REGARDING THE BEHAvlOR OF POLYMER BASE MATER:ALS SUB-DERNA,0 A.

JECTED TO BET A RADIATONVolume 3: Phase 2b Eimued Test NUREG/CR-5273 V01:

SCDAP/RE'APS/ MOD 2 CODE Results-MANUALRELAPS Code Structure. System Models, And Soluton Meth-CASEI.l.

ods NUREG/CH 5273 V02. SCDAP/RELAP5/ MOD 2 CODE MANUAL 3CDAP NUREG/CR 4708 V03. PROGRESS IN EVALUATION OF RAIAONU-Code Structure, Modets, And Soluton Methods.

CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-21

i

\\

J Personal Author Irdex LEVEL NUCLE AR WASTE FIEPOS! TORY SITE PROJECTS Report For DAE MENJ.J.K.

October 1987 June 1989.

NUREG/CR 5401. BOND STREN3TH OF CEMENT BOREHOLE PLUGS CHENQ T.C.

IN SALT.

NUREG/CR-5402: CRUSHED SALT CONSOLIDATON.

NUREG/CR 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE MANUAL.RELAPS Code Structura System Modets. And Solution Meth.

DALLMAN.RJ.

ods NUREG/CR-5225 ADD Ot AN OVERVIEW OF BWR MARK 1 CONTAIN-NUREG/CR 5273 V02. SCDAP/RELAPS/ MOD 2 CODE MANUAL SCDAP MENT VENTING RISK IMPLICATIONS An Evaluahon Of Potental Code Firucture. Monet And Soluteon Methods Mark-l Comamment Improvements NUREG/CR 5273 V01 SCDAP/RELAP5/ MOD 2 CODE MANUALUser's Guide And input Requuements DANIELS.L CHENIOd).

NUREG/CR-4550 V4 RIP 1; ANALYSIS OF CORE DAMAGE FREQUEN-CY: PE ACH BOTTOM. UNIT 2. INTERNAL EVENTS.

NUREGICR4530 V03. U S./ FRENCH JOtNT FtESE ARCH PROGRAM NUREG/CR-4550 V4 RIP 2. ANALYSIS OF CORE DAMAGE FREQUEN.

REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB.

CY: PEACH BOTTOM. UNIT 2, INTERNAL EVENTS APPENDICES.

JECTED TO DETA RADIATON Volume 3. Phase 2b Espanded Test Resulti,.

DASGUPTA.B.

CHOKSHl.N.C.

NUREG/CR-5427. ANALYSIS OF EMoLACEMENT BOREHOLE ROCK AND LINER BEHAVOR FOR A REPOSITORY AT YUCCA MOUN-NUREG.t233 REGULATORY ANALYSIS FOR USl A40 " SEISMIC T AIN.

DESIGN CRITERIA." Final Report DE AN,R.S.

CHOPR A.O.K.

NUREG/CR4744 V02 N2. LONG TERM EMBRITTLEMENT OF CAST NUREG/CR.5287. CLOSEOUT OF NRC BULLETIN 87 01;THINNtNG OF pipe WALLS IN NUCLE AR POWER PLANTS DUPLEX STAINLESS STEELS IN LWR SYSTEMS Senuannual NUREG/CR-5290: CLOSEOUT OF IE BULLETIN 79-28 POSSIBLE MAL.

Report.Aprt$eptember 1987.

FUNCTION OF NAMCO MODEL EA100 LIMIT SWITCHES AT ELE-VATED TEMPERATURES.

NUREG/CR 4744 V02 N2. LONG-TERM EMDRITTLEMENT OF CAST DEBORJ.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Sermanqual NUREG/CR.5348. MAN-MACHINE INTERFACE ISSUES IN NUCLEAR Report. April. September 1967.

POWER PLANTS Report On A Workshop Hdd On January 1012.

CLARE,AJ.

1989.

NUREG/lA.0015. ASSESSMENT OF INTERPHASE DRAG CORRELA-DENNING,R.S TIONS IN THE RELAP/5 MOD 2 AND TRAC PF1/ MODI CODES.

NUREG/CR4949. SOURCE TERM CALCULATIONS FOR ASSESSING RADIATION DOSE TO EQUIPMENT, NUREG/CR 4674 V07. PRECURSORS TO POTENTIAL SEVERE CORE DROUIN.M.T.

DAMAGE ACCIDENTS 1987 A STATUS REPORT. Main Report And NUREG/CR4550 V6R1P1: ANALYSIS OF CORE DAMAGE FREQUEN-Appenda A.

CY: GRAND GULF. UNIT 1 INTERNAL EVENTS NUREG/CR4674 V08. PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR 45b0 V6 RIP 2: ANALYSIS OF CORE DAMAGE FREQUEN-DAMAGE ACCIDENTS 1987 A STATUS REPORT.Appendmes B.C And CY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

D.

EHINGER.M.H.

CLINTON,J.H.

NUREG/CR-5004. RESOLUTION OF MCURRING LOSS ALARMS.

NUREG/CR 5363. A STUDY OF THE USE OF CROSSLINKED HIGH-DENSITY POLYETHYLENE FOR LOW LEVEL RADIOACTIVE WASTE ELRICK.R.M.

3 CONTAINERS NUREG/CR 51b5: THE THERMAL INSTABILiit OF CESillM IOOIDE, COLLtNR,J.L ERIKSSON J.

NUREG/CR 5339 DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/lA-0016. ASSESSMENT OF RELAP5/ MOD 2, CYCLE 36.04 RELEASE TEST VM.

AGAINST FlX-Il GUILLOTINE BREAK EXPERIMENT NO. 5061.

NUREG/CR-5340. DATA

SUMMARY

RuORT FO9 FISSION PRODUCT RELEASE TEST VI-2.

FLATER,0.A.

NUREG-1356 STATE COST SHARING OF TRAINING.A Task Force COOK.K.V.

Report NUREG/CR-5322: DETECTION AND CHARACTERIZATON OF INDICA.

TIONS IN SEGMENTS OF REACTOR PRESSURE VESSELS.

FOLEY,WJ.

NUREG/CR 5287: CLOSEOUT OF NRC BULLETIN 87 01. THINNING OF CORWIN.W.R.

PIPE WALLS IN NUCLE AR POWER PLANTS.

NUREG/CR 4219 V06 N1: HEAVY SECTION STEEL TECHNOLOGY NUREG/C9 5290: CLOSEOUT OF IE BULLETIN 79 28.POSSIBLE MAL.

PROGRAM Sermannual Progress Report For October 1988 March JUNCTION OF NAMCO MODEL EA180 LIMIT SWITCHES AT ELE-1989.

VATED TEMPERATURES.

CRAMOND,W.R.

FREEMAN-KELLY NUREG/CR 4550 V4R1Pt ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG/CR4949. SOURCE TERM CALCULATONS FOR ASSESSING CY: PE ACH DOTTOM. UNIT 2, INTERNAL EVENTS RADIATON DOSE TO EQUIPMENT.

NUREGICR4550 V4R1P2: ANALYSIS OF CORE DAMAGE FREQUEN.

CY: PEACH BOTTOM UNIT 2, INTERNAL EVENTS APPENDICES OA NUREG R 639 V5P2R2: NUCLEAR COMPUTEnt?ED LIBRARY FOR CROXFORD,M.G.

ASSESSING REACTOR RELIABILITY (NUCLAAR). Data Manual.Part 2:

NUDEG/lA-0014. ANALYSIS OF THE THETIS BOILDOWN EXPERI-Human Error Probabihty (HEP) Estimates MENTS USING RELAP5/ MOD 2.

NUREG/CR-4639 V5P3R2: NUCLEAR COMPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 3:

CUNNINGHAM.R.A.

Hardware Component Failure Data (HCFD)

NUREG/CR 5322: DETECTION AND CHARACTER 12ATION OF INDICA-NUREG/CR 4639 V5P4R2. NUCLEAR COMPUTERIZED LIBRARY FOR TONS IN SEGMENTS OF REACTOR PRESSURE VESSELS ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 4.

CURTIS.LA.

Summary Aggregations NUREG/CR 5525 ADD 01: AN OVERVIEW OF BWR MARK-1 CONTAIN-NUREG/CR4949 SOURCE TERM CALCULATIONS FOR ASSESSING MENT VENTING RISK IMPLICATONS An Eveluation Of Potential RADIATON DOSE TO EQUIPMENT.

Mark-l Containment improvements.

CYBULSKis,P.

GAUSSENS.G.

NUREG/CR4949: SOURCE TERM CALCULATONS FOR ASSES $1NG NUREG/CR4530 V03: U S / FRENCH JOINT RESEARCH PROGRAM RADIATION DOSE TO EQUIPMENT.

REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-1

Personal Author index 23 JECTED TO BETA RADiATIONVo6ame 3. Phase-2b Erpended Test TURE SYSTE4tS OF NUCLEAR POWER PLANTS. Aging Assessments Results.

And Monstonng Method Evaluatons GERTMAN.D.L HEASLER.P.G.

NUREG/OL 4639 V5P2R2: NUCLEAR COMPUTERIZED LIBRAPY FOR NUREG/CR 5161 V01: EVALUATION OF SAMPLIN3 PLANS FOR IN-ASSESSING REACTOR RELIABILITY (NUCLAAR) Data ManuLPart 2 SF.RVICE INSPECTION OF STEAM GENERATOR TUBES Modehng Of Human Error Probab4 sty (HEP) Estimates Eddy C trent Rehability Data. Anaytical Evaluatens. And initial Monte NUREG/0R-4639 V5P3R2 NUCLEAR COMPUTERIZED LIBRARY FOR Carlo Semulatons.

ASSESSING REACTOR RELIABILITY (NUCLARR) Data Manual.Part 3 Hardware Component Failure Dat3 (HCFD)

HEBOLE.T.L.

NUREGICR 4639 V5P4R2: NUCL( t.R COMPUTERIZED LIBRARY FOR NUREG/CR 5004. RESOLUTION OF RECURRING LOSS ALARMS.

ASSESSING REACTOR RELIAB!LITY (NUCLARA) Data Manual.Part 4.

Summary Aggregatons.

HENNICK.A.

NUREG/CR4287; CLOSEOUT OF NRC BULLETIN 87<Dt. THINNING OF GILDERT.S.O.

PIPE WALLS IN NUCLE AR POWER PLANTS NUREG/CR-4639 V5P2R2. NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5290. CLOSEOUT OF IE BULLETIN 79 28.POSSIBLE MAL.

ASSESSING REACTOR RELIABILITY (NUCLAAR) Data ManualPa'12:

FUNCTION OF NAMCO MODEL EA180 LIMIT SWITCHES AT ELE.

Human Error Probabihty (HEP) Estimates VATED TEMPERATURES.

NUREG/CR 4639 VbP3P.2. NUCLEAR CChPUTERIZED LIBRARY FOR ASSESSING REACTOR RELIABILITY (NUCLARR). Data Manual.Part 3 HUGHES.D.R.

Hardware Component Fa: lure Date (HCFD)

NUREG 1350: STATE COST SHARING OF TRAINING.A Task Force NUREG/CR-4639 V5P4R2: NUCLEAR COMPUTERIZED LIBRARY FOR Report ASSESSING REACTOR RELIABILITY (NUCLAP.R) Data Manual.Part 4:

Summary Aggregatons JOHNSEN,G.W.

NUREG/CR 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE GILMORE,W.E.

MANUALRELAP5 Code Structure, System Models, And Solution Meth.

NUREG/CR-4639 V5P2R2 NUCLEAR COMPUTERIZED LIBRARY FOR ods ASSESSING REACTOR REttABILITY (NUCLAAR). Data ManualPart 2:

NUREG/CR 5273 V02: SCDAP/RELAP5/ MOD 2 CODE MANUALSCDAP Hcman Error ProbaNIQ (PE LEnumates Code Structure. Models. And Solution Methods.

P NUREG/CR 4639 V5PNG: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-5273 V03: SCDAP/RELAP5/ MOD 2 CODE MANUALUser's ASSES $1NG REACTOR RELIABILITY (NUCLARR) Data Manual.Part 3.

Guide And input Requirements.

Hardware Component Failure Data (HCFD)

NUREG/CR-a639 V5P4R2. NUCLEAA COMPUTERIZED LIBRARY FOR KASSNE R,T.F.

ASSESSING REACTOR RELIABILITY (NUCLARR).3ata Manual Part 4.

NUREG/CR4S67 V06: ENVIRONMENTALLY ASSISTED CRACKING IN Summary Agg egatons.

LIGHT WATER REACTORS. 3ermannual ReportOctober 1987 - March OIOVANOLAJ.H.

NUREG/CR-5226. VISCOPLASTIC STRESS-STRAIN CHARACTER 12A-KING,0.8.

TlON OF A533 GRADE B CLASS 1 GTEEL NUREG/CR 5175; BETA AND GAMMA DOSE CALCULATIONS FOR

" ^ "

"^

OLOUDEM ANS J.R.

NUREG/CR-5395 V09. MULTILOOP INTEGRAL SYSTEM TEST KING T.L (MIST) flNAL REPORT. inter Group Compansons NUREG 1368: DRAFT PREAPPLICATION SAFETY EVALUATION REN FOR THE WER REACER WHEWW WE WW ORIESMEYERJ.M NUREG/CR-61N A REFERENCE MANUAL FOR THE EVENT PRO-LIQUID METAL REACTOR.

GRESSION ANALYSIS CODE (EVNTRE)-

KISER.D.M.

QUPTA.D.C.

NUREG/CR 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE NUREG 1373. TECHNICAL POSITION ON POSTCLOSURE SEALS, MANUALRELAP6 Code Structure, System Models. And Solution Meth.

BAR ERS AND DRAINAGE SYSTEM IN AN UNSA1VRATED NU G/CR-5273 V02: GCDAP/RELAP5M92 CODE MANUALSCDAP Code Structure. Models. And Soluton Metnuds.

HAGEMEYER D.

NUR3G/CR-5273 V03: SCDAP/RELAP5/ MOD 2 CODE MANUALUser's t

NUREG-0713 V08. OCCUPATIONAL RADIATION EXPOSURE AT COM.

Guide And input Requirements.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES KLOOP'R W FOR 1986 Nineteem5 Annual Report NUREG/CR-5226: VISCOPLASTIC STRESS-STRAIN CHARACTERIZA.

HAGRMAN.D.L TION OF A533 GRADE B CLASS 1 STEEL NUREG/CR-5273 V01.

SCDAP/RELAPS/ MOD 2 CODE KO C2K0W I,

MANUALRELAP5 Code Structure, System Models. And Solution Meth-MAN ALSCDAP NI R R 5 4 P N

IS RE DAMAGE FREQUEN-S eM u

thods NUREG/CR4273 V03. SCDAP/RELAP5/ MOD 2' CODE MANUALUser's CY; PEACH BOTTOM. UNIT 2. INTERNAL EVENTS APPENDICES.

Guide And input Requnements.

KONDIC.KR H ALL.P.C.

NUREG 1377: NRC RESEARCH PROGRAM ON PLANT AGING: LIST.

NUREG/tA-0014: ANALYSIS OF THE THETIS BOILDOWN EXPER6 ING AND ABSTRACTS OF REPORTS ISSUED THROUGH FEBRU-MENTS USING RELAP5/ MOD 2 ARY 1,1989.

HARRISJ.D.

LACHANCE J.L NUREG/CR.4674 V07; PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-4550 V6R1P1; ANALYSIS OF CORE DAMAGE FREQUEN-j DAMAGE ACCIDENTS 1987 A STATUS REPORT Main Report And CY: GRAND CULF. UNIT 1 INTERNAL EVENTS.

App.erdx A.

NUREG/CR 4550 V6 RIP 2: ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG/CR-4674 V08 PRECURSORS TO POTENTIAL SEVERE CORE CY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

AMAGE ACCIDENTS.1987 A ST ATUS REPORT.Appendines B.C, And NUREG/CR 5415: ENGINEERING DESIGN FOR THE NRC OPER.

HAWTHORNEJ.R.

ATIONS CENTER.

NUREG/CR-5388 STEEL IMPURITY ELEMENT EFFECTS ON POSTlR.

RADIATION PROPERTIES RECOVF.RY BY ANNEALING Final Report LA O Eg HAYNESA.D.

CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-NUREG/CR 4234 V02: AGING AND SERVICE WEAR OF E'LECTRIC LEVEL NUCLEAR WASTE REf*OSITORY SITE PROJECTS. Report For MOTOR OPFRATED VALVES USED IN ENGINEERED SAFETY FEA-October 1987 Jure 1989.

24 Personal Author index LANDRY,R.R.

MERRILL,R.M.

NUREG-1368: DRAFT PREAPPLtCATON SAFETY EVALLATON WREG/CR4155: THE THERMAL INSTA3tLt'.Y C# CESI'JM ODCI.

REPORT FOR THE POWER REACTOR INHEREUTLY SAFE MODULE LOUID MET AL RE ACTOR.

MEYER,L.C.

NUREG/CR 4967; NUCLEAR PLANT AGING RESEARCH ON HIGn

~

N G/CR-4530 V03. U.S1 FRENCH JOINT RESEARCH PROGRAM REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-ggygg,gg JECTED TO BET A RADIATONVolume 3. Phase 2b Expanded Test HUREG/CR4708 V03. PROGRESS IN EVALUATON OF RADONU-Results.

CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LE TUTOUR.P.

LEVEL NUCLEAR WASTE AEPOSITORY SITE PROJECTS Report For NUREG/CR.4530 V03 U.S / FRENCH JOINT RESEARCH PROGRAM October 1987. June 1989.

REGARDING THE BEHAVIOR OF POLYMER BASE IfATERIALS SUB.

C#

E D TO BETA RADIATON Volume 3: Phase-2b Expanded Test G/CR4273 V01:

SCDAP/RELAP5/ MOD 2 '

CODE MANUALRELAP5 Code Structure. System Models. And Solution Meth-LilGH.C.D.

Ods NUREGICR4085: PROGRESS IN DEVELOPMENT OF A METHOOOLO.

NUR* G/CR4273 V02. SCDAP/RELAP5/ MOD 2 CODE MANUALSCDA*

c GY FOR GEOCHEMICAL SENSITIVITY ANALYSIS FOR PERFORM.

Code Structure. Models. And Solution Methods.

ANCE ASSESSMENT. Parametric Calculations. Prelmnary Databass, NUREG/CR 5273 V03: SCDAP/RELAPS/ MOD 2 CODE MANUALUser's aim Computer Code Evaluatiort Guide And inrJt Requrements.

LOREN2,R.A.

MILLER.S.

NUREG/CR4339. DATA

SUMMARY

REPORT FOR FISSION PdODUCT NUREG/CR4550 YSRIP1: ANALYSIS OF CORE DAMAGE FREOVEN-RELEASE TEST Vbi.

CY: GRAND GULF, UNIT 1 INTERNAL EVENTS.

NUREG/CR.5340. DATA

SUMMARY

REPORT FOR FISSON PRODUCT NUREG/CR4550 V6 RIP 2 ANALYSIS OF CORE DAMAGE FREQUEN.

RELEASE TEST Vb2-CY; GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDeCES.

UE l'CR4427 ANALYSIS OF EtJIPLACEMENT BOREHOLE ROCK MINARICUW.

AND LINER BEHAVOR FOR A REPOSITORY AT YUCCA MOUN-NUREG/CR4674 V07: PRECURSORS TO POTENTIAL SLVERE CORE TAIN^

DAMAGE ACCIDENTS: 1967 A STATUS REPORT. Main Report And Appendix A.

LOVELACE.W.H.

NUREG/CR4674 '.'08: PRECURSORS TO POTENTIAL SEVERE CORE NUREG.On20 V13 N06: LICENSED OPERATING REACTORS STATUS DAMAGE ACCIDENTS 1987 A STATUS REPORT.Appendmes B,C, And SUMMAR) REPORT.Deta As Of May 311989 (Gray Book 1)

D.

NUREG4020 V13 N07: LICENSED OPENATING REACTORS STATUS

SUMMARY

REPORT. Data As Of June 30.1989 (Gray Book 11 MOSLEY,M.H.

NUREG 0020 V13 NO6: LICENSED OPERATING REACTORS STATUS NUREG 1356. STATE COST SHARING OF TRAINING.A Task Force

SUMMARY

REPORT. Data As Of July 31,1989 (Gray Book 1)

Report.

LUDENAU,J.0-MOEYAtRT,P.

MEG-1358: STATE COST SHARING OF TRAINING.A Task Force NUREG/lA 0020 ASSESSMENT STUDY OF RELAP5/ MOD 2, CYCLE Report 36.04 DASED ON SPRAY START UP TEST FOR DOEL4.

MACK.M.G NUREG/CR4390- ROCK MASS MODIFICATON AROUND A NUCLEAR MONTGOMERY,J.M.

WASTE REPOSITORY IN WELDED TUFF.

NUREG-1356: STATE COST SHARING OF TRAININGA Task Force Report MACLELLAN,J.A.

NUREG/CR4396: FINANCIAL IMPACT OF IMPLEMENTING DRAFT O'DONNELL.E.

ANSI STANDARD N13 30, PERFORMANCE CRITERIA FOR RADIO.

NUREG/CR4918 '/03. CONTROL OF WATER If71LTRATION INTO BIOASSAY.

NEAR SURFACE ILW D.SPOSAL UNITS.Frogress Report MAERKER.R.E.

O'KELLEY.G.D.

NUREG/CR.5338: ANALYS!S OF THE VENUS.3 EXPERIMENTS.

NUMG/CR4708 V03: PROGRESS IN EVALUATich OF RADIONU-CLIDE GEOO4EMICAL INFORMATION DEVELOPED BY DCE HIGH-N Ed CR4667 V06. ENVIRONMENTALLY ASSISTED CRACKING IN gg' LIGHT WATER REACTORS. Semiannual ReportOctober 1987. March 1988-OS50RNE,W.F.

MALONEY,K J.

NUREG/CR4339. DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR.4550 V4 RIP 1: ANALYSIS OF CORE DAMAGE FREOUEN.

RELEASE TEST Vb1.

CY' PEACH BOTTOM. UNIT 2. INTERNAL EVENTS NUREG/CR4340: DATA St}MMARY REPORT FOR FtSSCN PRODUCT NUREG/CR4550 V4 RIP 2: ANALYSIS OF CORE DAMAGE FREQUEN.

RELEASE TEST VI 2.

CY: PF.ACH BOTTOM UteT 2. INTERNAL EVENTS APPENDICES.

NUREGrCR.5263. TH8 RISK MANAGEMENT IMPLICATIOWS OF OUELLETTE.A.L NUREG.1150 METHODS AND RESULTS.

CREG/CR4155: THE THERV AL INST ABILITY C" CESIUM 'Ot.XDE.

MCCLUNG.R.W.

OUYANG.S.

NUREG/CR4322: DETECTION AND CHARACTER 12ATON OF INDICA.

NUREG/CR44@ C4JSHED SALT CONSOLIDATON.

TIONS IN SEGMENTS OF REACTOR PRESSURE VESSELS.

PARK.J.Y.

MCKENNA TJ.

NUREG/CR.5247; RASCAL VERSON 1,3 ! PER'S GUIDE.

NUREG/CR 4667 V06: ENVIRONMENTALLY ASSISTEC CRACKPiG IN LIGHT WATER REACTORS. Smannual ReportOctober 1987. Mar:h MCNAMARA N.

1988.

NUREG 0837 V09 N01: NRC TLD DIRECT RADIATON MONITORING b N C' EhR iTON MONITORING NUR G/C 4315: ENGINEERING DESIGN FOR THE NRC OPER.

NURE 3

D NETWORK Progress Report April. June 1989 N,0NS CENTER.

WERGES,P.J.

RAGLIN.K.A.

NUREG.1356. STATE COST SHARING OF TRAINING.A Task Force NUREG 1356 STATE COST SHARING OF TRAINING.A Task Force Roport F epert

Personal Author index 25 RANSOM.V.N.

SJOREEN.A.L.

NUREG/CR-5273 V01:

SCDAP/RELAPS/ MOD 2 CODE NUREG/CR.5247. RASCAL VIDSION 1.3 USER'S GUIDE.

MANUALRELAP5 Code Structure. System Modeis, And Soluton Meth-ods SMITH.S.W.

NUREGICR 5273 V02. SCDAP/RELAP5/ MOD 2 CODE MA NUALSCDAP NUREG/CR.5UO4: RESOLUTION OF RECURRlNG LCSS ALARMS Code Structors. Models. And Soluton Methods NUREG/CR4273 V03. SCDAP/RELAP5/ MOD 2 00DE MANUALUser's SMITH.Lal.

Gede And tnput Regerements.

NUREG LR-5174: A REFERENCE MANUAL FOR THE EVENT PRO.

GRESflON ANALYS!$ CODE (EVNTRE)

NUREG/CR.4918 V03: CONTROL OF WATER INNLTRATION INTO SMITH.P.V.

NEAR SURFACE LLW DISPOSAL UNITS Progress Report NUFJG/CR 3252. NEW ENGLAND 3EISM0 TECTONIC STUDY ACTM-RIEMKE.R.A.

NUREG/CR 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE SMOOT.J.L.

. MANUALRELAP5 Code Structere. System Models. And Soluton Meth.

NUREG/CR.5431: A REVIEW OF GEOSCIENCE CHARACTERISTICS ods NURE'3/CR 5273 V02: SCDAP/RELAP5/ MOD 2 CODE MANUALSCDAP AND DISPOSAL EXPERIENCE AT THE COMMERCIAL LOW LEVEL Code Structure. Models. And Soluton Methreds RADIOACTIVE WASTE DISPOSAL FAriLITY NEAR WEST VALLEY, NUREG/CR 5273 V03 SCDAP/RELAP5/ MOD 2 CODE MANUALUser's NEW YORK.

Guide And input Requirements.

goo,p, NUREG/CR-5363. A STUDY OF THE USE OF CROSSLINKED HIGH-RUTHER.W.E.4667 V06: ENVIRONMENTALLY ASSISTEO CRACKING IN DENSITY PCiYETHYLENE FOR LOW LEVEL RADIOACTIVE WASTE NUREG/CR LIGHT WATER REACTORS.'Somiannual ReportOctober 1987 March CONTAINERS.

1988.

STEELE.R.

SANDERVAQO.

NUREG/CR 4977 VO1: SHAG TEST SERIES Soismic Research On An NUREG/lA 0017: ASSESSMENT OF TRAC PF1/ MODI AGAINST A Aged Urvted States Gate Valve And On A Piping System in The Do.

LOSS-OF GRID TRANSIENT IN RINGHALS 4 POWER PLANT.

commessoned Heisadampfreaktor (HDR) Summary.

NUREGICR 4977 V02: SHAG TEST SERIES Seismic Research On An SCANGA.S.

Aged United States Gate Valve And On A Pong System in The De-NUREG/CR 5152-COMPARISON AND REGULATORY IMPACT OF commissoned Heitadampfreaktor (HDR). Appendices.

NOA 1 AND NQA-2 WITH N45.2 OA STANDARDS.

STOKLE Y.J.

. SCHUL 2 R.K.

NUREG/CR-5152: COMPARISON AND REGULATORY IMPACT OF NUREG/CR-4918 V03. CONTROL OF WATER INFILTRATION INTO NOA4 AND NOA 2 WITH N45.2 0A ST ANDARDS.

NEAR SURFACE LLW DISPOSAL UNITS. Progress Report.

STRUCKMEYER.R.

BERKt2.A.W.

NUREG-0837 V09 N01: NRC TLD DIRECT RADIATION MONITORING NUREG 1267: TECHNICAL RESOLUTION OF GENERIC SAFETY ISSUE NETWORK Progress Report Janugry March 1989.

A-29 Nuc6 ear Power Pant Design For Reducton Of Vulnerabel,ty To in-NUREG-0837 V09 NO2: NRC TLD DIRECT RADIATION MONITORING dustnal Sabotage.

NETWORK. Progress Report April-June 1989.

SHACK.WJ.

STUSSE.E.

NUREG/CR-4867 V06. ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/tA 0020: ASSESSMENT STUDY OF RELAPS/ MOD 2, CYCLE LIGHT W ATER REACTORS. Semiannual ReportOctober 1987. March 36.04 BASED ON SPRAY START UP TEST FOR DOEL 4.

4 1988.

SUEN.C.J.

SHAPIRO.SJ, NUREG/CR-5387: LOW LEVEL WASTE SHALLOW LAND DISPOSAL NUREG/CR4550 V6 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN' SOURCE TERM MODEL: DATA INPUT GUIDES.

(

CY: GRAND GULF UNIT 1 INTERN AL EVENTS.

{!

NUREG/CR 4550 V6R1P2: ANALYSIS OF CORE DAMAGE FREQUEN-BULLIVAN.T.M.

CY; GRAND GULF UNIT 1 INTERNAL EVENTS APPENDICES-NUREG/CR-5387; LOW-LEVEL WASTE SHALLOW LAND DISPOSAL l

SHAUKAT.S.K.

SOURCE TERM MODEL DATA INPUT GUIDES.

j NUREG 1233: REGULATORY ANALYSIS FOR USt A40. "SELSMIC gwgggy,g, DESIGN CHf7LRIA." Fcd Report NUREG/CR-5348. MAN-MACHINE INTERFACE ISSUES IN NUCLEAR SNIEHAS.

POWER PLANTS. Report On A Workshop Held On January 1012.

i NUREG/CR-5273 V01:

SCDAP/RE2.AP5/ MOD 2 CODE 188E MANUALREAAP5 Code Structure. Sysic;n mom And Solution Meth.

SYPE.T.T.

b G/CR $273 V02. SCDAP/RELAP5/ MODI' CODE MANUALSCDAP NUREQ/CR 4550 V4R1P1: ANALYSIS OF CORE DAMAGE FREQUEN-NU CY: PEACH BOTTOM UNIT 2. INTERNAL EVENTS.

NR CR 5 7 OAP LA MOD CODE kANUALUser's NUREG/CR-4550 V4R1P2: ANALYSIS OF CORE DAMAGE FREOVEN-Guide And input Reautrements' CY: PE ACH BOTTOM. UNIT 2. INTERNAL EVENTS APPENEWCES.

NUREG/CR-5263. THE RISK MANAGEMENT lMPLICATIONS OF LIEFKEN.L J.

NUREG 1150 METHODS AND RESULTS.

NJREG/CR 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE MANUALREiAP5 Code Structure. System Models. And Selon Meth.

8I NU E 2 : EVALUATION OF SAFETY IMPLICATIONS OF CONTROL NUREG/CR-5273 V02-SCDAP/RELAP5/ MOD 2 CODE MM 1ALSCDAp SYSTEMS IN LWR NUCLEAR POWER PLANTSTechrscal Fendings Coq Structure. ModMs. And Soluten Methont Related To USI A-47. Final Report i

NUREG/CR 5273 M SCCAP/CELAPS ' MOD 2 CODE WJ@ JALUser's NUREG 1218 REGULATORY ANALYSIS FOR RESOLUTION OF USl A.

Guide And input Reqmments.

47. Safety Imphcatons Of Control Systems in LWR Nuclear Power Plants Fnal Report S4EGEL.M.D.

NUREG/CR 5085. PROGRE.SS IN DEVERPMEW OF A METHOCOLO-THATCHER.D.F.

GY FOR GEOCHEMICAL SENAVITf ANALYSIS FOR PERFORM-NUREG-1229. REGULATORY ANALYSIS FOR RESOLUTION Of USl A.

ANCE ASSESSMENT.Parametne Csicvigtrobs. Prehrrunary Databases,

17. Systems Interactons in Nuclear Power Plants.

And computer Code Evaluatiort pp fuOSERO.t NUREG/CR-5273 V31:

SCDAP/RELAPS/ MOD 2 CODE NUREG/lA 0017: ASSESSMENT OF TRAC-PF1/ MOD 1 AG.NST A MANUALRELAPS Code Structure. System Models. And Soluton Meth.

LOSS OF GRID TRANSIENT IN RINGHALS 4 POWER PLANL ods.

N Potional Author index NUREG/CR1473 V02. SCOAP/RELAP5/ MOD 2 CODE MANUALSCDAP WESSTER.C.S.

Code Struchare. Models. And Solubon Methods.

NUREG/CR-5339. DATA SUM *AARY REPORT FOR FIS$60N PRODUCT NUREG/CR4273 V03. SCDAP/RELAPS/ MOD 2 CODE MANUALUner's RELEASE TEST VI1.

Guide And input Regwements.

NUREG/CR4340. DATA SUMMARN REPORT FOR FISSION PRODUCT TRAUS,RJ.

RELEASE TEST VI-2.

NUREG/CR4396: FINANCIAL IMPACT OF IMPLEMENTING DRAFT ggggg,gg, ANSI STANDARD N13.30, PERFORMANCE CRITERIA FOR RADIO-

BIOASSAY, NUREG/CR-2331 V06 N4: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORf TRA VISJ.R.

RESEARCH. Progress Report. October December 1988.

NUREG/CR4339. DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR-2331 V09 N1: SAFETY RESEARCH PROGRAMS SPON-RELEASE TEST VI 1.

SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR4340 DATA

SUMMARY

REPORT FOR FISSION PRODUCT RESEARCH. Progress ReportJanuary, March 1989.

RELEASE TEST VI-2.

WHEELER,T.A.

UPDEGRAFF.C.D NUREG/CR-4550 V4 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR43ts:C MPARISON OF STRONGLY HEAT. DRIVEN FLOW CY: PEACH BOTTOM. UNIT 2. INTERNAL EVENTS CODES FOR UNSATURATED MEDIA NUREG/CR-4550 V4 RIP 2. AHALYSIS OF CORE DAMAGE FREQUEN-VESELY,W.E.

CY: PEACH BOTTOM. UNIT 2. INTERNAL EVENTS APPENDICES.

NUREG/CR-4550 V6R1P1: AllALYSIS OF CORE DAMAGE FREQUEN.

NUREG/CR 5425. EVALUATION OF ALLOWED OUTAGE TIMES CY: GRAND GULF, UNIT 1 INTERNAL EVENTS.

(AOTS) FROM A RISK AND RELIABILfTY ST ANDPOINT.

NUREG/CR 4550 V6 RIP 2: ANALYSIS OF CORE DAMAGE FREQUEN-CACHTERJ.W.

CY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

NUREGICR4004. RESOLUTION OF RECURRING LOSS ALARMS.

WHITE.R.S.

CAGNER.K.C.

NUREG/CR4161 V01: EVALUATION OF SAMPLING PLANS FOR IN-NUREG/CR4225 ADO 01: AN OVERVIEW OF BWR MARK.1 CONT AIN-SERVICE INSPECTION OF STEAM GENERATOR TUBES Modelen0 Of MENT VENTING RtSK IMPLICATIONS.An Evaluabon Of Potentel Eddy-Current Reletety Data. Anatyucal Evaluations, And treal Monte Mas-! Containment improvements.

Car'v Sem:.;!ations.

WAGNER,RJ.

WILSON J.N.

NUREG/CR427'4 V01:

SCDAP/RELAP5/ MOD 2 CODE NUREG 1368. DRAFT PREAPPLICATION SAFETY EVALUATION D NUREG/CR4273 V02: SCDAP/RELAP5/ MOD 2 MANUALRELAPt', Code Structure, System Models, And Solution Meth.

REPORT FOR THE POWER REACTOR INHERENTLY SAFE MODULE ods LIQUID METAL REACTOR.

Code Structure, Models. And Solution Methods.

VAMASHITA.T.

NUREG/CR4273 V03, SCDAP/RELAP5/ MOD 2 CODE MANUALUser's NUREG/CR4339 DATA

SUMMARY

REPORT FOR FIS$lON PRODUCT Guide And input Requirements.

RELEASE TEST Vl-1.

l 1

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

A 633 3 steel Alerm Pesolution NUREG/CR 5368: STEEL IMPURITY ELEMENT EFFECTS ON POSilR-NUREG/CR-5004:RESOLUTON OF RECURRING LCSS ALARMS RADIATION PFOPERTIES RECOVFRY BY ANNEALING Final Report.

A683 Orade 3 Steet NUREG/CR-5425: EVALUATON OF ALLOWED OUTAGE TIMES NUREG/CR 5226: VISCOPLASTIC STRESS STRAIN CHARACTERIZA-(AOTS) FROM A RISK AND RELIABILttY STANDPOINT.

TON OF A533 GRADE B CLASS 1 STEEL-Anneeling AEOD' NUREG/CR-5388. STEEL IMPURITY ELEMENT EFFECTS ON POSTiR.

NUREG 1272 V03 NOI: AEOO OFFICE FOR ANALYSIS AND EVALUA*

RADIATON PROPERTIES RECOVERY BY ANNEALING Final Report.

TON OF OPERATIONAL DATA 190 ANNUAL REPORT. Power Reac-tors BWR NUREG 1272 V03 NO2: AEOD OFFICE FOR ANALYSIS AND EVALUA-NUREG/CR-5225 ADD 01: AN OVERVIEW OF BWR MARK 1 CONTAIN-TION OF OPERATIONAL DATA 1998 ANNUAL REPORT.Nonreactors.

MENT VENTING RISK IMPLICATONS.An Evaluaton Of Potential Mark 4 Containment Improvements.

NUREG/CR-5396 FINANCIAL IMPACT OF IMPLEMENTING DRAFT Beta Dose Calculat60n ANSI STANDARD N13 30, PERFORMANCE CRITERIA FOR RADIO-NUREG/CR-5175: BETA AND GAMMA DOSE CALCULATIONS FOR BOASSA Y.

PWR AND BWR CONTAINMENTS.

Abnormal Occurrence Beta Radtat6on NUREG-009C V12 Not: REPORT TO CONGRESS ON ABNORMAL NUREG/CR4530 V03. U.81 FRENCH JOINT TEARCH PROGRAM OCCURRF.NCES. January-March 1989, REGARDING THE BEHAVOR OF POLYMER tv.SE MATERIALS SUB-JECTED TO BETA RADIATON Volme 3. Phow2t> Expanded Test Resuus.

NUREG 0304 Vid N01: REGULATORY AND TECHNICAL REPORTS (ABSTRACT INDEX JOURNAL). Compilation For First Quarter 96owdown Experiment G RE }P / MOD 2

^^ '

N EG 304 4 NO2: REGULATORY AND TECHNICAL REPORTS ME (ABSTRACT INDEX JOURNAL). Cornpilation For Second Quarter 1989,Apre June.

Soi44ng Weser Reactor NUREGICR-5225 ADD 01:. AN OVERVIEW OF BWR MARK 1 CONTAIN-MENT VENTING RISK IMPLICATIONS An Evaluation Of Potential NU R

ENGINEERING DESIGN FOR THE NRC OPER.

ATONS CENTER.

Mark 4 Cosa.mnent improvements Borehole

[

Accident Progress 6on NUREG/CR.5174: A REFERENCE MANUAL FOR THE EVENT PRO-NUREG/CR 5427: ANALYSIS OF EMPLACEMENT BOREHOLE ROCK GRESSON ANALYSIS CODE (EVNTRE)

AND LINER BEHAVIOR FOR A REPOSITORY AT YUCCA MOUN-TAIN.

Acc6 dent Sequ9nce NUREG/CR 4550 V4R1P1: ANALYSIS OF CORE DAMAGE FREQUEN.

Coment Dorehole Plug CY: PE ACH BOTTOM. UNIT 2. INTERNAL EVENTS.

NUREG/CR-5401: BOND STRENGTH OF CEMENT BOREHOLE PLUGS NUREGICR4550 V4R1P2: ANALYSIS OF CORE DAMAGE FREQUEN.

IN SALT.

CY: PE ACH BOTTOM. UNIT f', INTERNAL EVENTS APPENDICES.

NUREGICR-4550 V6 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN.

Ceanum lod 6de CY: GRAND GULF, UNIT 1 INTERNAL EVENTS.

NUREG/CR 5155: THE THERMAL INSTABILITY OF CESIUM ODIDE.

NUREG/CR 4550 VORIP2. ANALYSIS OF CORE DAMAGE FREQUEN-CY; GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

Closeout NUREG/CR4674 V07 PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-5287: CLOSEOUT OF NRC BULLETIN 87 01: THINNING OF DAMAGE ACCOENTS: 1987 A STATUS REPORT. Main Report And PiF.I WALLP (1 NUCLEAR POWER PLANTS.

Appenom A.

NUREG/CR-5290: CLOSEOUT OF IE BULLETIN 79-2e POSSIBLE MAL-NUREG/CR4674 V08; PRECURSORS TO POTENTIAL SEVERE CORE FUNCTION OF NAMCO MODEL EA180 LIMIT SWITCHES AT ELE-DAMAGE ACCIDENTS 1987 A STATUS REPORT.Appendues B.C, And V ATED TEMPERATURES.

Cobalt-60 Aged Gate Valve NUREG/CR-5231: COBALT 60 SIMULATION OF LOCA RADIATON EF-NUREGICR4977 V01: SHAG TEST SERIES Seismic Research On An FECTS.

T Aged Uruted States Gate Valve And On A Piping System in The De-commissioned Heisadamptreaktor (HDR) Summary.

Computer Code NUREGICR-4977 V02. SHAG TEST SERIES. Seismic Research On An NUREG/CR 4639 V5P2R2. NUCLEAR COMPUTERIZED LIBRARY FOR Aged United States Gate valve And On A Piping System in The De-ASSESSING REACTOR RELIABILITY (NUCLAAR). Data Manual.Part 2:

I comnvassoned Heisadampfreaktor (HDR)- Appendices.

Human Error Probability (HEP) Estimates.

NUREG/CR-4639 V5P3R2: NUCLEAR COMPUTERIZED LIBRARY FOR Aging Assessment ASSESSING REACTOR RELIABILITY (NUCLAnR). Data ManualPart 3.

NUREG/CR4234 V02. AGING AND SERVICE WEAR OF ELECTRIC Hardware Component Failuie Data (HCFD)

MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY.FEA.

NUREG/CR-5273 V01:

SCDAP/RELAP5/ MOD 2 CODE j

TURE SYSTEMS OF NUCLEAR POWER PLANTS. Aging Assessments MANUAL.RELAP5 Code Structure, System Models. And Solution Meth.

And Monatonng Method Evaluabons-ods.

27 k

N-buhject index NUREG/CR-5273 V03: SCDAP/RELAP5/ MOD 2 CODE MANUALuser's Discontwuum Mosel

. Guide And enput Ree,arements NUREG/CR-6426: EAAMINATION OF THE USE OF CONTINUUM VERSUS DISCONTINUUM MODELS FOR DESIGN AND PERFORM-Cenerete tunke' ANCE ASSESSMENT FOR THE YUCCA MOUNT AIN SITE.

NUREG 1375 V01: SAFETY EVALUATON STATUS REPo1T FOR THE PROTOTYPE LICENSE APPLICATON SAFETY ANALYSIS Does Ateeeement REPORT.Ee th Mounded Concrete Bunker.

NUREG/CR-5247: RASCAL VERSION 1.3 USER'S GUOE.

Container Dronnese System NUREG/CR 5363 A STUDY OF THE USE OF CROSSLINKED HIGH-NUREG 1373: TECHNICAL POSITION ON POSTCLOSt.%E SEALS, DENSITY POLYETHYLENE FOR LOW LEVEL RADOACTIVE WASTE BARRIERS AND DRAINAGE SYSTEM IN AN UNSATURATED CONTAINERS.

MEDIUM.

Centeenment Earthouske NUREG-1365: REVISED SEVERE ACC:lDENT RESEARCH PROGRAM NUREG/CR-3252: NEW ENGLAND SEISMDTECTONIC STUDY ACTIVI.

TIES DURING FISCAL YEAR 1980.

NUR /C SU RM CALCULATIONS FOR ASSES $1NG RADi4 TION DOSL TO EQUIPMENT.

E86y Current NU G/CR 61 AD AMMA DOSE CALCULATIONS FOR NUREG/CR 5161 VOI: EVALUATON OF SAMPLit4G PLANS FOR IN-SERVICE INSPECTON OF STEAM GENERATOR TUBES ModehnJ Of Conwel System Eddy Current Rehetwhty Dete Analytcal Evoluetens. And instel Monte NUREG 1218: REGULATORY ANALYSIS FOR RESOLUTION OF USI A.

Carlo Simulatone.

47.Setety Impicotene Of Control Systems in LWR Nucieer Power E

6ement Ptents Final Report NUREG/CR 6400. BASIS FOR IN SITU GEOMECHANICAL TESTING AT NUREG/CR 4744 V02 N2: LONG TERM EMBRITTLEMENT OF CAST THE YUCCA MOUNTAIN SITE.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Somennual Report, April. September 1967.

Coolant System NUREG/CR-6273 V01:

SCDAP/RELAP5/ MOD 2 CODE Emergency Plan MAUUALRELAP5 Code Structure, System Models, And Solution Meth-NUREG/CR 6415 ENGINEERING DESIGN FOR THE NRC OPER-ods.

ATONS CENTER.

NUREGICR 5273 V02: SCDAP/RELAPS/ MOD 2 CODE MANUALSCDAP Code Structure, Models. And Soluten Methods Enforcement Act6on NUREG/CR-6273 V03. SCDAP/RELAP6/ MOD 2 CODE MANUALUser's NUREG/CR 3252: NEW ENGLAND SEISMOTECTONIC STUDY ACTIVI-Guide And input Regurements.

TIES DURING HSCAL YEAR 1980.

Core Damage Environmental impact Statemei.t NUREG 1150 V01: SEVERE ACCIDENT RISKS: AN ASSESSMENT FOR NUREG 0683 SOA PROGRAMMATIC ENVIRONMENTAL IMPACT FfvE U.S. NUCLEAR POWER PLANTS. Summary Report.Second Drott STATEMENT RELATED TO DECONTAMINATON AND DISPOSAL OF For Peer Review.

RADIOACTIVE WA9TES RESULTING FROM MARCH 28,1979 ACCl-NUREG 1150 V02: SEVERE ACCOENT RISKS: AN ASSESSMENT FOR DENT THREE MILE ISLAND NUCLEAR STATON, UNIT 2. Final Sup.

FNE U.S. NUCLEAR POWER PLANTS. Appendices.Second Drott For piement Deshng With...

Peer Review.

NUREG/CR-4660 V4R1P1: ANALYSIS OF CORE DAMAGE FREOVEN-Equipment Queltf6cetion CY: Pi ACH BOTTOM, UNIT 2, INTERWAL EVENTS.

NUREGICR-4949: SOURCE TEAM CALCULATONS FOR ASSESSitd NUREG/CR-4550 V4 RIP 2; ANALYSIS OF CORE DAMAGE FREOVEN-RADIATON DOSE TO EQUIPMENT.

CY: PEACH BOTTOM. UNIT 2, INTERNAL EVENTS APPENDICES.

NUREG/CR 4550 VmtP1: ANALYSIS OF CORE DAMAGE FREQUEN-CY: GRAND GULF, UNIT 1 INTERNAL EVENTS.

NUREG-0525 R15: SAFEGUARDS

SUMMARY

EVENT LIST (SSEU.

NUREGICR-4560 VOR1P2; ANALYSIS OF CORE DAMAGE FREQUEN-CY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

Event Progression Analvels Code NUREG/CR-4674 V07: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR 5174: A REFERENCE MANUAL FOR THE EVENT PRO.

DAMAGE AOCOENTS: 1987 A STATUS REPORT. Main Report And GRESSION ANALYSIS CODE (EVNTRE).

N CR 674 V08: PRECURSORS TO POTENTIAL SEVERE CORE Exp6eretory Shaft DAMAGE ACCOENTS.1987 A STATUS REPORT. Appendixes B,C, And NUREG/CR-5400 BASIS FOR IN SITU GEOMECHANICAL TESTING AT D-THE YUCCA MOUNTAIN SITE.

C mehed ten Final Environmental statement NUREG/CR-5402: CRUSHED SALT CONSOL'DATON.

NUREG-0974 SUPP: FINAL ENVIRONMENTAL STATEMENT RELATED Deeny-Heat Romovel TO THE OPERATON OF LIMERICK GENERATING STATON, UNITS NUREG 1368: DRAFT PREAPPL1 CATION SAFETY EVALUATION 1 AND 2. Docket Not,60-352 And 60-353,(PNiadophia Electne Compe-REPORT FOR THE POWER REACTOR INHERENTLY SAFE MODULE "N

LIQUID METAL REACTOR.

p g

Decommientoning NUREG 1336 R01. STANDARD FORMAT AND CONTENT GUIDE FOR NUREG 1336 RL1: STANDARD FORMAT AND CONTENT GUIDE FOR FINANCIAL ASSURANCE MECHANISMS REWIRED FOR DECOM-FINANCIAL ASSURANCE MECHANISMS REQUIRED FOR DECOM.

MISSONING UNDER 10 CFR PARTS 30,40,70 AND 72.

MISSIONING UNDER 10 CFR PARTS 30,40,70 AND 72.

NUREG-1337 RO1: STANDARD REVIEW PLAN FOR THE REVIEW OF NUREG 1337 R01: STANDARD REVIEW PLAN FOR THE REVIEW OF FINANCIAL ASSURANCE MECHANISMS FOR DECOMMISSONING FINANCIAL ASSURANCE MECHANISMS FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,70 AND 72.

UNDER 10 CFR PARTS 30,40,70 AND 12.

Decentaminetten NUREG/CR 53G6. FINANCIAL IMPACT OF IMPLEMENTING DRAFT NUREG 0683 S03: PROGRAMMATIC ENVIRONMENTAL IMPACT ANSI STANDAAD N13.30, PERFORMANCE CRITERIA FOR RADIO.

STATEMENT RELATED TO DECONTAMINATON AND DISPOSAL OF BIOASSAY.

RAD 60 ACTIVE WASTES RESULTING FROM MARCH 28,1979 ACCl-DENT THREE MILE ISLAND NUCLEAR STATION, UNIT 2. Final Sup.

Floelon Prodset piement Deshr.g With.

NUREG/CR-5339 DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST V)1.

Depresed Core Aeoident NUREG/CR-5340- DATA

SUMMARY

REPORT FOR FISSON PRODUCT NUREG/CR-5155: THE THERMAL INSinalltJTY OF CES!UM ODIDE.

RELEASE TEST V)-2.

=

_ __ ]

Subject index 29 Fracture Wochen6ce Neph-Level meteactive Weste Desposal NOREG/CR4219 V06 Nt: HEAVY SLCTION STEEL TECHNOLOGY NUREG/CR4367:OOMPANISON OF STRONGLY HEAT DRIVEN FLOW PROGRAM Semiennual Progress Report For October 1988. March CODES FOR UtsSATURATED MEDA 1989 NUREG/CR4226. VtSCOPLASTIC STRESS-STRA'N CHARACTERIZA-High Level Weste Repeettory TICN OF A533 GRADE B CLASS 1 STEEL NUREG/CR-4708 V03: PROGRESS IN EVALUATION OF RADIONU-CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-Fracture Toughneet LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report For NUREG/CR 4744 V02 N2: LONG-TERM EMBRITTLEMENT OF CAST October 1987. June 1989.

DUPLEX STA,htESS STEELS IN LWR SYSTEMS Semiannual NUREG/CR4085: PROGRESS IN DEVELOPMENT OF A METHODOLO.

R*PorLApril-September 1967.

GY FOR GEOCHEMICAL SENSITIVITY ANALYSIS FOR PERFORM-ANCE ASSESSMENT.Parametnc Calculatons. Preimnary Databases.

O R4339. DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST VL-1 H6storteel Hoseng NUREGICR 5340. DATA

SUMMARY

REPORT FOR FtSSION PFtODUCT NUREG 0800 02.4.2 R3. STANDARD FtEVIEW PLAN FOR THE REVIEW FtELEASE TEST VI-2 OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER i

Gemme Does Ce6culet6on PLANTS LWR E$ ton.Reveen 3 To SRP Secten 2.4.2, "F6 pods "

NUREG/CR 5175: BETA AND GAMMA DOSE CALCULATIONS FOR Human Error PreablNIy PWR AND BWR CONTAINMENTS NUREG/CR4636r V5P2R2: NUCLEAR COMPUTERIZED LIBRARY FOR Genorte Setety leave A 2, ASSESSING REACTOR RELIABILITY (NUCLAAR) Data Manual.Part 2:

NUREG-1267; TECHNICAL RESOLUTION OF GENERIC SAFE FY ISSUE NUR" R463 U

A PUTERIZED LIBRARY FOR A 29 Nuclear Power Plant Desen For Reducten Of Vulneratulity To in-dustnal Sabotape.

ASSESS'NG PEACTOR RELLABILITY (NUCLARR).Deta Manual. Pert 3:

Hardware womponent Failure Data (HCFD)

Geoct.emical Sonettivity NUREGICR-4639 V5P4R2: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR 6085. PROGRESS IN DEVELOPMENT OF A METHODOLO-ASSESSING REACTOR RELIABILITY (NUCLARR).Dete Lianual.Part 4:

GY FOR GEOCHEMIC/.L SENSITIVITY ANALYSIS FOR PERFORM.

Summ A99'opetens.

ANCE ASSESSMLNT F arametrc Calculatons. Prehmrary Databases, "i'NYR[G5bb R3. STANDARD REVIEW PLAN FOR THE R And Computer Code Evaluaton.

W RepWory OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG 1373. TECHN< CAL POSfTION ON POSTCLOSURE SEALS.

PLANTS LWR E$ ton Roween 3 To SRP Secten 2 4.3, ' Probable BARRIERS AND DRAINAGE SYSTEM IN AN UNSATURATED Mammum Flood (PMF) On Streams And Revers."

MEDtVM.

ICAP Ptnerem Orain Stae D etributlen NUREG/iA 0017 ASSESSMENT OF TRAC.PF1/ MOD 1 AGAINST A NUREG/CR4402: CRUSHED SALT CONSOLIDATION' LOSS OF GRID TRANSIENT IN RINGHALS 4 POWER PJNT, NUREG/lA 0020: ASSESSMENT STUDY OF RELAPS/ MOD 2, CYCLE Oroundwater 36.04 DASED ON SPRAY START UP TEST FOR DOEL 4.

NUREG/CR 5431: A REVIEW OF GEOSCIENCE CHARACTERISTICS AND DISPOSAL EXPERIENCE AT THE CCMMERCIAL IOW-LEVEL II RADIOACTIVE WASTE DISPOSAL FACILITY NEAR WEST VALLEY' NUREG/CR4290. CLOSEOUT OF IE BULLETIN 79 2f'POSSIBLE MAL.

NEW YORK.

FUNCTION OF NAMCO MODEL EA180 LIMIT SwifCHES AT ELE-VATED TEMPERATURES.

Outloot6ne Dreek NUREG/lA 0016 ASSESSMENT OF RELAP5/ MOD 2, CYCLE 36 04

'"4' * **'"***'#'"

AGAINST FIX il GUILLOTINE BREAK EXPERIMENT NO. 5061.

NUREG/CR 5161 V01: EVALUATION OF SAMPLING PLANS FOR IN-SERVICE INSPECTION OF STEAM GENERATOR TUBES Modehng Of Heat Loading Eddy Current Rehabihty Data. Analytcal Evaluatons. And initial Monte NUREG/CR4420. VARIATION OF HEAT LOADING FOR A REPOSI.

Carlo Semulatons.

TORY AT YUCCA MOUNTAIN Host Ortwen Flow NUREG/CR4400: BASIS FOR IN-SITU GEOMECHANICAL TESTING AT NUREG/CR4367: COMPAdlSCN OF STRONGLY HEAT DRIVEN FLOW THE YUCCA MOUNTAIN SITE.

CODES FOR UNSATURATED MEDIA.

Inden Heavy Section Steelischnolo9y (HSST) Program NUREG-0304 V14 NC1: REGULATORY AND TECHNICAL REPORTS NUREG/CR4322. DETECTION AND CHARACTERl2ATION OF M)CA-(ABSTRACT INDEX JOURNAL). Compilaton For First Quarter TIONS IN SEGMENTS OF REACTOR PRESSURE VESSELS.

1989. January March.

l NUREG-0304 V14 N02: REGULATORY AND TECHNICAL REPORTS Heavy Section Steel Technology (ABSTRACT INDEX JOURNALL Compilaton For Second Quarter NUREG/CR-4219 V06 Nt: HEAVY SECTION STEEL TECHNOLOGY 1989, April June.

PROGRAM Semiannual Progress Report For October 1988. March 1989 Indtv6 dual Plant Enemonet6on NUREG 1335:

INDIVIDUAL PLANT EkAMINATION. SUBMITTAL Holosdamptreektor GUIDANCE 7enal Report.

(

NUREG/CR 4977 VD1: SHAG TEST SERIES Seisme Research On An A ed Uruted States Gate Valve And On A Pipsng System in The D,.

Induction Furnace 9

connssoned Hotssdamptreaktor (HDR) Summary, NUREG/CR 5339-DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR 4977 V02: SHAG TEST SERIES Seisme Research On An RELE ASE TEST VI-1.

A9ed United States Gate Valve And On A Piping System in The De-NUREGICR4340: DATA

SUMMARY

REPORT FOR FISSION PRODUCT commissoned Heesdamptreaktor (HDR).Appendees RELE ASE TEST VI-2.

l H69h Protoure injection System industrial Sebote9*

NUREG/CR-4967: NUCLEAFt PLANT AGING RESEARCH CN HIGH NUREG 1267: TECHNICAL RESOLUTION Or GENERIC SAFETY ISSUE PRESSURE INJECTION SYSTEMS.

A 29 Nuclear Power Plant Design For Reducten Of Vulnerabihty To in.

H6ph-Denelty Polyethylene NUREG/CR4363 A STUDY OF THE USE OF CROSSLINKED HIGH-Interphase Oreg l

DENSITY POLYETHYLENE FOR LOW LEVEL RADIOACTIVE WASTE NUREG/LA 0015 ASSESSMENT OF INTERPHASE DRAG CORRELA-CONTAINERS TIONS IN THE RELAP/5 MOD 2 AND TRAC.PF1/ MODI CODES lk

i 30 Subject index trementor NUREC-1276 Vo$ ADO. OPERATING EXPERIENCE FEEDBACK NUREG/CR-6231: COBALT-60 SIMULATON OF LOCA RAD 6ATON EF.

FCPDRT. PROGRESS IN SCRAM REDUCTON Commercial Power 7 ECTS.

Reactors NUREG/CR4867 V06: ENVIRONMENTALLY ASSISTED CRACKING IN UREG/CR 6426. [KAMINATON OF THE USE OF CONTINUUM 988 VERSUS DISCONTINUUM MODELS FOR DESIGN AWD PERFORM-NUREG/CR4674 V07: PRECURSORS TO POTENTIAL SEVERE CORE ANCE ASSESSMENT FOR THE YUCCA MOUNTAIN SITE-DAMAGE ACCIDENTS: 1987 A STATUS REPORT Main Report And M

NU R 674 V08-PRECURRORS TO POTENTIAL SEVERE CORE NUREG/CR-2000 V08 N6. LICENSEE EVENT REPORT (LER)

DAMAGE ACCOENTS.1987 A STATUS REPORT.Appendmes B.C. And COMPILATION For Month Of June 1989 D

NUREG/CR-4744 V02 N2. LONG-TERM EMBRITTLEMENT OF CAST MPILA IONFor Month Of J bENSEE 89 NUREG/CR-2000 V0B NB Li EVENT REPORT (LER)

DUPLEX STAINLESS STEELS IN LWR SYSTEMS. Semiannual R

COMPILATION.For Month Of August 1989.

gg.

u-tog g

LOCA TONS IN 3EGMENTS OF REACTOR PRESSURE VESSELS.

NURE G/CR-5231: COBALT 60 SIMULATON OF LOCA RADIATION EF-FECTS.

Limit Switch NUREG/CR 5290. CLOSEOUT OF IE BULLETIN 79 26 FOShtBLE MAL-LWR FUNCTON OF NAMCO MODEL EA180 LIMIT SWITCHES AT ELE.

NUREG 1218. REGULATORY ANALY.4lS FOR RESOLUTION OF USI A-VATED TEMPERATURES.

47.Safet/ Irn> cations Of Control Systems in LWR Nuclear Power Piants. F6nal Report Liquid Metal Reactor NUREG 1276 V05 ADD. OPERATING EXPERIENCE FEEDBACK NUREG-1368. DRAFT PREAPPLICATON SAFETY EVALUATON REPORT PROGRESS IN SCRAM REDUCTION Commercial Power REPORT FOR THE POWER REACTOR INHERENTLY SAFE MODULE NUF CR-4667 V00: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS Somiannual Report October 1987 - March Loeo-Of Coolant Acc6 dent

[L

[f[roup N

/CR 4674 V07: PRECURSORS TO POTENTIAL SEVERE CORE ST) r on DAMAGE ACCIDENTS.1987 A STATUS REPORT.Mam Report And nda A LowOf Gr6d NU i u/CR4674 V08. PRECURSORS TO POTENTIAL SEVERE CORE NUREG/lA 0017: ASSESSMENT OF TRAC PF1/ MODI AGAINST A DAMAGE ACCIDENTS.1987 A STATUS REPORT.Appendmes B.C And LOSS OF-CRID TRANSIENT IN RINGHALS 4 POWER PLANT.

D NUREG/CR4744 Vn2 N2: LONG-TERM EMBRITTLEMENT OF CACT Low Level Resoective Weele DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Sermannual NUREG/CR-5363: A STUDY OF THE USE OF CROSSLINKED HIGH-NTYCT N AND CHARACTERIZATION OF INDICA-DENSITY POLYETHYLENE FOR LOW LEVEL RADIOACTIVE WASTE NU 3

TIONS IN SEGMENTS OF REACTOR PRESSURE VESSELS CONTAINERS.

T E YUCC O'

AIN T'

L***L'"I NOU"

• • • I' OPO'*I NUREG-1375 V01: SAFETY EVALUATON STATUS REPORT FOR THE Legalleeuencee PROTOTYPE LICENSE APPLICATION SAFETY ANALYSIS NUREG 0750 V29 tot: INDEXES TO NUCLEAR REGULATORY COM.

REPORT. Earth-Mounded Concrete Bunker.

MISSION ISSUANCES. January March 1989 NUREG/CR 5431: A REVIEW OF GEOSCIENCE CHARACTERISTICS NUREG 0750 V29 N05: NUCLEAR REGULATORY COMMIS$10N IS-AND DISPOSAL EXPERIENCE AT THE COMMERCIAL LOW LEVEL SUANCES FOR MAY 1989 Pag-e8 395463 RADIOACTIVE WASTE DISPOSAL FACILITY NEAR WEST VALLEY.

NUREG 0750 V29 N06. NUCLLAR REGULATORY COMMISSON IS-NEW YORK.

SUANCES FOR JUNE 1989.Paget 465 558.

Low Level Weste Ucensed Operat6ng Reactor

• NUREG/CR-5387. LOW LEVEL WASTE SHALLOW LAND DISPOSAL NUREG4020 Via N06 UCENSED OPERATING REACTORS STATUS SOURCE TERM MODEL DAT A INPUT GUIDES.

SUMMARY

REPORT. Data As Of Ma 311989 (Grav Book 1)

NUR.iG 0020 V13 N07. LICENSED hi#lATING REACTORS STATUS MIST k O S STATUS N REG 20 V1 N E

N ST F A in r roup Co r

a

SUMMARY

REPORT. Data As Of July 31.1989(Giay Book 1)

Man Mechine interface leeue Ucensed Operator NUREG/CR-5348 MAN-MACHINE INTERFACE ISSUEb N NUCLEAR NUREG 0800131.2 R3. STANDARD REVIEW PLAN FOR THE REVIEW POWER PLANTS Report On A Wodshop Held On January 1012.

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER 4 89 PLANTS. LWR EdmortRevision 3 To SRP Section 131.2131.3, "Oper, sting Organization "

Mark-l Containment Licensee Event Report NUREGICR-5225 ADD 01: AN OVERVIEW OF BWR MARK 1 CONTAIN.

NUREG/CR 2000 V08 N6-LICENSEE EVENT REPORT (LER)

MENT VENTING RISK IMPLICATIONS An Evaluation Of Potential COMPILATION For Month Of June 1989 Mark l Containment Improvements.

NUREG/CR 2000 V08 N7. LICENSEE EVENT REPORT (LER)

COMPILATION For Month Of Jutv 1989 Mater 6el Control And Accounting NUREG/CR-2000 V06 NB LICENSEE EVENT REPORT (LER)

NUREG/CR-5004: RESOLUTION OF HECURRING LOSS ALARMS.

COMPILATION For Month Of August 1989.

g, Licensing Application NUREG/CR 4234 V02: AGING AND SERVICE WEAR OF ELECTRIC NUREG 1336 RO1: STANDARD FORMAT AND CONTENT GUIDE FOR MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY FEA.

FINANCIAL ASSURANCE MECHANISMS REQUIRED FOR DECOM-TURE SYSTEMS OF NUCLEAR POWER PLANTS Aging Assessments MISSIONING UNDER 10 CFR PARTS 30. 40. 70 AND 72.

And Monitonng Method Evaluations.

Ught Water Reactor MM'Itrop Integral System NUREG 1217. EVALUATION OF SAFETY IMPUCATONS OF CONTROL NUREG/CR-539'. V09 MULTILOOP INTEGR AL SYSTEM TEST SYSTEMS IN LWR NUCLEAR POWER PLANTS.Techrecal Fin @ngs (MIST) FINA RE PORT. inter Group Compansons.

Related To USI A-47 Fr.at Report NUREG 1218 REGULATORY ANALYSIS FOR RESOLUTION OF U$l A.

NRC Bulletin 87 01 47.Salety implications Of Control Systems in LWR Nuclear Power NUREG/CR-5287: CLOS. OUT OF NRC BULLETIN 87 01. THINNING OF Plants Final Report PIPE WALLS IN NUCLLAR POWER PLANTS r

Subject index 31 NUCLARn entsu NUREG/CR4639 V5P2R2. NUCLEAR COMPUTERIZED LIBRARY FOR NUREG 1368 DRAri PREAPPLICATION SAFETY EVALUATION ASSESSING REACTOR RELIABluTY (NUCLAAR1 Data Manual.Part 2 Human Error Probabiltty (HEP) Estrmates REPORT FOR THE POWER REACTOR INHERENTLY SAFE MODULE NUREG/CR-4633 V5P3R2: NUCLEAR COVPUTERIZED LfBRARY FOR LtOU1D MLT AL RE ACTOR.

ASSESSING REACTOR RELIABILITY (NUCLARA) Data Moiual.Part 3:

Performance Aeoesement Hardware Component Faauro Data (HCFD)

NUREG/GR-4639 V5P4R2 NUCLEAR COMPUTERIZED ttBRARY FOR NUREG/CR4085. PROGRESS IN DEVELOPMENT OF A METHODOLO-ASSESSING FtEACTOR RELIABILITY (NUCLARR) Data ManualPart 4:

GY FOR GEOCHEMICAL SENSITIVITY ANALYSIS FOR PERFORM-Summary ADgregatons.

ANCE ASSESSMENT.Parametnc Celculatons, Preliminary Databases, And Computer Code Evatustert NUMEG 1160 Pettt6on For Rulemaking NUF EG/CR4263: THE RISK MANAGEMENT IMPLICATIONS OF NUREG 1150 METHODS AND RESULTS-NUREG 0936 V08 N02: NRC REGULATORY AGENDA Ouarterly Report.Apru June 1989 Neutron Transport NUREG/CR4338. ANALYSIS OF THE VENUS 3 EXPERIMENTS.

P

• " gp ggg g gg ggg Nucteer Plant Aging PIPE WALLS IN NUCLEAR POWER PLANTS.

NUREG/CR 4967. NUCLEAR PLANT AGING RESEARCH ON HIGH PRESSURE INJECTION SYSTEMS.

Piping Syelent NUREG/CR-4977 V01: SHAG TEST SERIES Seisme Research On Ari Nucteer Plant Ag6ng Roeoarch NUREG-1377: NRC RESEARCH PROGRAM ON PLANT AGtNG. LIST.

Aged United States Gate Valve And On A Pipeng System in The De-commissoned Herssdamptreaktor tHDR) Summary.

ING AND ABSTRACTS OF REPORTS ISSUED THROUGH FEBRU-NUREG/CR 4977 V02. SHAG TEST SERIES Se'smic Research On An ARY 1,1969.

Aged United States Gate valve And on A Piping System in The De.

Nuclear Regulatory Legnelat60n NUREG 0980 RD4: NUCLE AR REGULATORY LEGISLATION.

Polymer Seee Material Nuclear Weste NUREG/CR.4530 V03. U.S / FRENCH JCINT RESEARCH PROGRAM NUREG/CR 5390. ROCK MASS MODIFICATION AROUND A NUCLEAR REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB.

WASTE FiEPOSITORY IN WJLDED TUFF-JECTED TO BETA RADIATION. volume 3. Phase-2b Experded h 1 NUREG/CR4427: ANALYSIS OF EMPLACEMENT BOREHOLE ROCK Results.

AND LINER BEHAVIOR FOR A REPOSITORY AT YUCCA MOUN-Postclosure Seal NI G/CR4428 VARIATION OF HEAT LOADING FOR A REPOSi.

NUREG 1373: TECHNICAL POSITION ON POSTCLOSURE SEALS, TORY AT YUCCA MOUNTAIN.

BARRIERS AND DRAINAGE SYSTEM IN AN UNSATURATED MEDIUM.

Occupat6onal Radiation NUREG-0713 V08 OCCUPATIONAL RADIATION EXPOSURE AT COM-P*P M '"

Mr 9CIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES NUREG 0800 02 4 2 R3. STANDARD REVIEW PLAN FOR T6'E REVIEW

. 41986 Nineteenth Annual Report OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Oparating Exper6ence PLANTS. LWR EditnrtRevision 3 To SRP Secton 2 4.2, " Floods?

NUREG-1272 V03 N01: AEOD OFFICE FOR ANALYSIS AND EVALUA.

Preneure vessel TION OF OPERATIONAL DATA 1988 ANNUAL REPORT. Power Reac.

NUREG/CR 4219 V06 N1: HEAVY SECTION STEEL TECHNOLOGY tors NUREG-1272 V03 NO2. AEOD OFFICE FOR ANALYSIS AND EVALUA-PROGRAM. Semiannual Progress Repot For October 1988 Mt.rch toco.

TION OF OPERATIONAL DATA 1988 ANNUAL REPORT.Nonreactors.

NUREG/CR4226: VISCOPLASTIC STdWTRAIN CHARACTER 12A-NUREG-1275 V05 ADD. OPERATING EXPERIENCE FEEDBACK TION OF A533 GRADE B CLASS 1 W ;L.

REPORT - PROGRESS IN SCRAM REDUCTION Commercial Power NUREG/CR-S388. STEEL IMPURITY M 2ENT EFFECTS ON POSTIR-Reactors.

RADIATION PROPERTIES RECOVERY BY ANNEALING.Fenal Report Operational Event NUREG/CR 4674 V07: PRECURSORS TO POTENTIAL SEVERE CORE Proesurtrer Sprey NUREG/lA-0020: ASSESSMENT STUDY OF RELAP5/ MOD 2, CYCLE g

DAMAGE ACCIDENTS: 1987 A STATUS REPORT Main Report And 36.04 BASED ON SPRAY START UP TEST FOR DOEL 4.

Apperxtin A.

NUREG/CR-4674 V08. PRECURSORS TO POTENTIAL SEVERE CORE Probabilletic Riek Analysis DAMAGE ACCIDENTS 1987 A STATUS REPORT.Appendexes B,C, And NUREG/CR 6174: A REFERENCE MANUAL FOR THE EVENT PRO-D.

I GRESSION ANALYSIS CODE (EVNTREL Operations Center NUREG/CR 5415: ENGINEERING DESIGN FOR THE NRC OPER.

Probabileetic R6sk Assosoment ATIONS CENTER.

NUREG 1150 V01: SEVERE ACCIDENT RISKS: AN ASSESSMENT FOR FIVE U S. NUCLEAR POWER PLANTS. Summary Report.Second Draft NURE 032

2. U S. NUCLEAR REGULATORY COMMISSION FUNC.

NU 50 4R P NALY IS E D MAGE FREQUEN-TIONAL ORGANIZATION CHARTSJuly 1.1989^

NUREG/CR-4550 V4R1Pt ANALYSIS OF CORE DAMAGE FREQUEN-PRA CY: PEACH BOTTOM, UNIT 2, INTERNAL EVENTS APPENDICES.

NUREG-1150 V02; SEVERE ACCIDENT RISKS. AN ASSESSMENT FOR NUREG/CR 4550 V6 RIP 1: ANALYSIS OF CORE D/ MAGE FREQUEN-F U $ NUCLEAR POWER PLANTS. Appendices Second Draft For C

,RA D ULF U 1

ER AL V ygp NURLG/CR-4550 V4 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN-CY: GRAND GULF, UNIT 1 INTERNA L EVENTS APPENDICES.

CY: PEACH BOTTOM. UNIT 2. INTERNAL EVENTS NUREG/CR-5263. THE RISK MANAGEMENT lMPLICATIONS OF HUREG/CR4550 V4R1P2. ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG 1150 METHODS AND RESULTS CY: PEACH BOTTOM. UNIT 2. INTERNAL EVENTS APPENDICES NUREG/CR 5425: EVALUATION OF ALLOWED OUTAGE TIMES NUREGiCR 4550 V6 HIP 1: ANALYSIS OF CORE DAMAGE FREQUEN.

(AOTS) FROM A RISK AND RELIABILITY STANDPOINT.

CY: ORAND GULF. UNIT 1 INTERNAL EVENTS.

NUREG/CR-4550 V6 RIP 2 ANALYSIS OF CORE DAMAGE FREQUEN.

Probable Maximum Ficod CY: GRAND GULF, UNIT 1 INTEHNAL EVENTS APPENDICES.

NUREG 0800 02 4.2 R3: STANDARD REVIEW PLAN FOR THE REVIEW NUREG/CR 5263. THE RISK MANAGE MENT IMPLICATIONS OF OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG4150 METHODS AND RESULTS.

PLANTS. LWR Editon.Re.sion 3 To SAP Secton 2 4 2, " Floods "

NUREG/CR-5425: EV ALUATION OF ALLOWED OiJTAGE TIMES NUREG 0800 02 4.3 R3 STANDARD REVIEW PLAN FOR THE REVIEW (AOfS) FROM A RtSK AND REllABILITY ST ANDPOINT.

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER

l l

32 Subject index PLANTS. LWR EditortRewson 3 To SRP Secton 2 4.3. 'Probatne Regulatory Analyons Manwnum Flood (PMF) On Stretms And Rwo a "

NUREG-1229 REGULATORY ANALYSIS FOR RESOLUTON OF USI A-

17. Systems internetons in Nuclear Power Plants p

NUREG/CR-5401: BOND STREN3TH OF CEMENT BOREHOu! PLUGS Regulatory And Techrocal Report IN SALT.

NUREG 0304 V14 Not: REGULATORY AND TECHNICAL REPORTS Quality Aeourance Standard I h.J)nua y 19 a ch NUREG/CR4152. COMPARISON AND REGULATORY IMPACT OF y.JREG4304 V14 NO2. REGULATORY AND TECHNICAL REPORTS NOA.1 AND NQA 2 WITH N45 2 0A STANDARDS.

(ABSTRACT INDEX JOURNAL). Compilaton For Second Quaner

1. 8'AP'N RADCAL NUREG/CR 5247: RASCAL YERSON 1.3 USER'S GUIDE.

Report To Congea NUREG 0000 VO N01: REPORT TO CONGRESS ON ADNORMAL RELAPS/ MOD 2 OCCURRENCES January March 1989 NUREG/lA 0016. ASSESSMENT OF RELAP5/ MOD 2, CYCLE 36.04 AGAINST FlK Il GUILLOTINE BREAK EXPERIMENT NO $061.

g9'*"Y NUREG/lA 0020. ASSESSMENT STUDY OF RELAPS/ MOD 2. CYCLE NUREG 1347. NRC STAFF SITE CHARACTERl2ATION ANALYSIS OF 3304 BASED ON SPRA4 GT ART UP TEST FOR DOEL 4' THE DEPARTMENT OF ENERGY'S SITE CHARACTER 12ATON PLAN. YUCCA MOUNT AIN SITE. NEVADA.

RgtApgMoog NUREG/CR4402: CRUSHED SALT CONSOLIDATON NUREG/tA.0014: ANALYSIS OF THE THETIS BOtLDOWN EXPERI.

NUREG/CR4427: ANALYSIS OF EMPLACEMENT BOREHOLE ROCK MENTS USING RELAP5/ MOD 2 AND LINER BEHAVIOR FOR A REPOSITORY AT YUCCA MOUN-NUREG/lA 0016. ASSESSMENT OF INTERPHASE DRAG CORRELA.

TA N TIONS Its THE RELAP/5 MOD 2 AND TRAC-PF1/ MOD 1 CODES.

p p

TORY AT YUCCA MOUNTAIN.

Rad 6et6en NUREG/CR4231: COBALT 40 SIMULATION OF LOCA RADIATION EF-R6ek Menegement FECTS.

NUREG/CR4263. THE RISK MANAGEMENT IMPLICATIONS OF Redtetton Does NUREG 1150 METHODS AND RESULTS.

NUREGICR-4530 V03: U.SJFRENCH JOINT RESEARCH PROGRAM REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB-Reek Me**

JECTED TO BETA RADIATION Volume 3. Phase 2b Expanded Test NUREG/CR 5390; ROCK MASS MODIFICATION AROUND A NUCLEAR WASTE REPOSITORY IN WELDED TUFF.

Results NUREG/CR-4949. SOURCE TERM CALCULATONS FOR ASSES $1NG RADIATON DOSE TO EQUIPMENT.

Rook Seit NUREG/CR4175. BETA AND GAMMA DOSE CALCULATONS FOR NUREGICR-5401: BOND STRENGTH OF CEMENT BOREHOLE PLUGS PWR AND BWR CONT AINMENTS.

IN SALT.

Redist6on Embritt6ement Rune NUREGICA 5388. STEEL IMPURITY ELEMENT EFFECTS ON POSTIR-NUREG-0936 Vos N02. NRC REGULATORY AGENDA. Quarterly RADIATION PROPERTIES RECOVERY BY ANNEALINGFinal Report Report.Aprildune 1989.

Redtat6on Monitortne Network SCDAP/RELAPl NUREG4837 V09 N01: NRC TLD DIRECT RANTON MONITORING NUREG/CR 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE NE'TWORK Progress Report. JanuarpMarch 1989.

MANUALRELAPS Code Structure. S, stem Models. And Soluton Meth.

NUREG4837 VOD NO2: NRC TLD D. RECT RADIATON MONITORING ods.

NETWORK.Progrsas Report. April 4une 1989.

NUREG/CR 5273 V02: SCDAP/RELAP5/ MOD 2 CODE MANUALSCDAP Code Structure. Models. And Soluton Methods.

Red 6ation Trenaport NUREG/CR-5273 V03. SCDAP/RELAP5/ MOD 2 CODE MANUALUser's NUREO'CR 5247: RASCAL VERSION 13 USER'S GUIDE.

Ouide And input Requirements.

Redloactive Weste SHA0 Test NUREG 0683 S03. PROGRAMMAtlC ENVIRONMENTAL IMPACT NUREG/CR 4977 Vot: SHAQ TEST SERIES. Seismic Research On An ST ATEMENT RELATED TO DECONTAMINATON AND DISPOSAL OF Aged United States Gate Valve And On A Pipmg System in The De-RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 ACCb DENT THREE MILE ISLAND NUCLEAR STATION, UNIT 2 final Sup-NUYd7CI4Y77 S

ES S ame Research On An A ed Urvted States Gate Valve And On A Piping System in The De-piement Dealing WA" D

commissoned Heissdamptreaktor (HDR): Appendices.

Redtobleesey NUREG/CR 5396: FINANCIAL IMPACT l)F IMPLEMENTING DRAFT 88 S STANDARD N13.30. PERFORMANCE CRITERIA FOR RADtO-02.4.3 R3: STANDARD REVIEW PLAN FOR THE REVIEW T SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER AANTS. LWR Editon.Revison 3 To SRP Secten 2.4.3. "Prouble Re 16cnuclide NUREG/CR 4708 V03: PROGRESS IN EVALUATON OF RADONU.

Maximum Flood (PMF) On Streams And Rivers.

CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-L N

EA W TE REPOSITORY SITE PROJECTS. Report For R 1267; TECHNICAL RESOLUTON OF GENERIC SAFETY ISSUE A 29. Nuclear Power Plant Design For Fieducton Of Vulneratniity To in-NUREG/CR4155: THE THERMAL INSTABILITY OF CESIUM ODOE.

dustnal Sabotage.

Med6onuclide 06echerge NUREGICR4086: PROGRESS IN DEVELOPMENT OF A METHODOLO-Safe 9uerde Summary Event Laat GY FOR GEOCHEMICAL SENSITIVITY ANALYSIS FOR PERFORM-NUREG 0525 R15: SAFEGUARDS

SUMMARY

EVENT LIST (SSEL).

ANCE ASSESSMENT.Parametrte Calculatons. Preliminary Databases.

Sately Evaluation j

And Computer Code Evaluaton.

NUREG 1375 V01: SAFETY EVALUATON STATUS REPORT FOR THE

/

Reactor Preeeuro Vessel PROTOTYPE LICENSE APPLICATON SAFETY ANALYSIS NUREG/CR-5322: DETECTION AND CHARACTERlZATON OF INDICA.

REPORT. Earth Mounded Concrete Bunker.

TONS IN SEGMENTS OF REACTOR PRESSURE VESSELS.

(

Setety Evolustion Report R*9ulatory Agende NUREG4991 SO9. SAFETY EVALUATION REPORT RELATED TO THE NUREG 0936 V08 NO2. NRC REGULATORY AGENDA.Ouarterly OPERATON OF LIMERICK GENERATING STATION. UNITS 1 AND ReporLApril4une 1989.

2. Docket Nos. 50 352 And 50-353.(Ptuladelphia E6ectre Company)

(

Subject inder 33 NUREG 13SB-DRAFT PREAPPLICATION SAFETY EVALUATON Standard Review Plan REPORT FOR THE POWER REAClOR INHERENTLY SAFE MODULE NUREG 0800 02 4 2 R3 STANDARD REVIEW PLAN FOR THE REVIEW LOUID MET AL REACTOR.

OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS LWR E$ ton.Renson 3 To SRP Secten 2 4.2.

• Floods "

Sotely heaeerch Program NUREG/CR 2331 VOS N4 SAFETY RESEARCH PROGRAMS SPON' NUREG-0800 02.4.3 R3: STANDnRD REVIEW PLAN FOR THE RFVtEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER SORED BY OFFICE OF HUCLEAR REGULATORY PLANTS. LWR Eston.Revson 3 To SRP Secten 2 4.3, "Probab6e N RE 31 09 4 ET A H OGRAMS SPON-himum FlDod (PMF) On Strearis And Rwers "

SORED BY OFFICE OF NUCLEAR REGULATORY NUREG 080013.1.2 R3: ST ANDARD REVIEW PLAN FOR THE REVIEW RESEARCH. Progress Report. January March 1989 OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS. LWR Eston Revison 3 To SRP Secten 13.1.2131.3. "Oper-Sentiary Landfill atmo Orgenmoton."

NUREG/GR 4316 V03: CONTROL OF WATER INFILTRATON INTO NUREG-1337 Ro1: STANDARD REVIEW PLAN FOR THE REVIEW OF NEAR SURFACE LLW DISPOSAL UNITS Progress Report FINANCIAL ASSURANCE MECHANISMS FOR DECOMMISSIONING UNDER 10 CFR PARTS 30,40,70 AND 72 NUREG 1275 V05 ADD. OPERATING EXPERIENCE FEEDBACK State Coet REPOR1 PROGRESS IN SCRAM REDUCTON Commercel Power NUREG 1356. STATE COST SHARING OF TRAINING.A Task Force

Reactors, Report Setem6c Deenen NUREG 1233 REGULATORY ANALYSIS FOR USI A 40. " SEISMIC SteHon 96eck N DESIGN CRITERIA." Finel Report.

NUREG/CR-5395 V09: MULTILOOP INTEGRAL SYSTEM TEST (MIST) FINAL REPORT. Inter Group Compensons.

Solemic Hazard NUREG/CR 3252. NEW ENGLAND SEISMOTECTONIC STUDY ACTIVI.

Steam Generator Tube TIES DURING FISCAL YEAR 1980.

NUREG/CR 5161 V01: EVALUATION OF SAMPLING PLANS FOR IN-SERVICE INSPECTION OF STEAM GENERATOR TUBES Mode 16n0 Of a

eta. AnaWal beens, And W %

G 77 VO1: SHAG TEST SERIES.Seemic Research On An Apsd Outed States Gate Valve And On A Pipmg System in The De-NUR /

-4 77 67A E S ES Sesmic Research On An 8"#8 "P'I "'*'

Aged Urmed States Gate Velve And On A Pipmg System in The De.

NUREG/CR 5004: RESOLUTION OF RECURRING LOSS ALARMS.

commsssoned Heissdampfreaktor (HDR) Appeneces.

Street Corros6on Cracking Setemotectonic NUREG/CR-4667 V06: ENVIRONMENTALLY ASSISTED CRACKING IN i

NUREGICR.3252.14EW ENGLAND SEISMOTECTONIC STUDY ACTIVI-LIGHT WATER REACTORS. Semiennual ReportOctober 1967. March TIES DURING FISCAL YE AR 1980.

1988 Severe Accident System Interact 60n NUREG 1335:

INDIVIDUAL PLANT EXAMINATION. SUBMITTAL NUREG 1229. REGULATORY ANALYSIS FOR RESOLUTION OF USl A-NUR

$ 3 01:

SCDAP/RELAP5/ MOD 2 CODE MANUALRELAP5 Code Structure. System Models. And Soluten Meth-THE18 NUNC/CR.5273 V02: SCDAP/RELAP5/ MOD 2 CODE MANUALSCDAP NUREG/ A 4 AN LYSIS OF THE THETIS BOILDOWN EXPERI.

Code Structure. Models. And Soluton Methods.

NUREG/CR 5273 V03: SCDAP/RELAP5/ MOD 2 CODE MANUALUser's Guse And input Requirements ~

E NUREG 0837 V09 N01: NRC TLD DIRECT RADtATION MONITORING Severe Acc6 dent Research Program NETWORK. Progress Report January March 1989.

NUREG 1365. REVISED SEVERE ACCIDENT RESEARCH PROGRAM NUREG 0837 V09 NO2-NRC TLD DIRECT RADIATION MONITORING PLAN. Fiscal Year 1990 1992.

NETWORK. Progress Report Apr# June 1989.

Severo Acc6 dent R6ek TRAC-PFf ' MOD 1 NUREG 1150 V01: SEVERE ACCIDENT RISKS: AN ASSESSMENT FOR NUREG/lA 0015: ASSESSMENT OF INTERPHASE DRAG CORRELA-FIVE U.S NUCLEAR POWER PLANTS Summary ReportSecond DraN For Peor Review.

TiONS IN THE RELAP/5 MOD 2 AND TRAC-PF1/ MOD 1 CODES.

NUREG 1150 V02: SEVERE ACCIDENT RISKS AN ASSESSMENT FOR TRAC-P01/ MODI i

FIVE U S. NUCLEAR POWER PLANTS.Appenscos Second Draft For N'***'

NUREG/iA 0017: ASSESSMENT OF TRAC PF1/ MOD 1 AGAINST A LOSS-OF GRID TRANSIENT IN RINGHALS 4 POWER PLANT.

Shallow Land D6sposal NUREGICR-5387; LOW-LEVEL WASTE SHALLOW LAND DISPOSAL Technical Specmoetion SOURCE TERM MODEL: DATA INPUT GUIDES.

NUREG 1371: TECHNICAL SPECIFICATIONS FOR LIMERICK GENER-ATING STATION, UNIT 2. Docket No. 50353.(PhliedelpNa Electric Shift Crew Company)

NUREG-08001312 R3: STANDARD REVIEW PLAN FOR THE REVIEW NUREG 1376: TECHNICAL SPECIFICATIONS FOR LIMERICK GENER-OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER ATING STATION. UNIT 2. Docket No. 50353.(PhiladelpNe Electric PLANTS. LWR Editen. Revision 310 SRP Secten 13.1.213.1.3. "Oper-Company) eting Orgaruston."

NUREG/CR-5425: EVA'.UATON OF ALLOWED OUTAGE TIMES f

Site Charactertaation

(^

}

^

NUREG-1347: NRC STAFF SITE CHARACTER 12ATION ANALYSIS OF Thermoluminescent Doeineter THE DEPARTMENT OF ENERGY'S SITE CHARACTERtZATON PLAN. YUCCA MOUNTAfN SITE. NEVADA.

NUREG 0837 V09 N01: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report January-March 1989.

Site Characterisation Plan NUREG-0837 V09 NO2: NRC TLD DIRECT RADIATON MONITORING NUREGICR 5428 VARIATION OF HEAT LOADING FOR A REPOSI-NETWORK. Progress Report April-Jone 1989.

TORY AT YUCCA MOUNTAIN Steintees Steel NUREG-0540 V11 N05: TITLE LIST OF DOCUMENTS MADE PUBLICLY NUREG/CR 4744 V02 N2. LONG TERM EMBRITTLEMENT OF CAST AVAILABLE May 131,1989.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS.Som: annual NUREG-0540 V11 N06: TTTLE LIST OF DOCUMENTS MADE PUBLICLY RepartApre-September 1987.

AVAILABLE. June 1 30.1989.

34 Subject index Topopen Sprines ventine not NUREG/CR 6426. EXAMINATON OF THE USE OF CONTINUUM NUREG/CR-5225 ADD 01: AN OVERVIEW OF BWR MARK 4 CONTAIN-VERSUS DISCONTINUUM MODELS FOR DESIGN AND PERFORM-MENT VENTING RISK IMPLICATIONS An Evaluation Of Potential ANCE ASSESSMENT FOR THE YUCCA MOUNTAIN SITE.

Ma k4 Containment Irnprovernents, Tro6n6ng y copi gi, NUREG-1356 STATE COST SHARING OF TRAINING.A Task Force NUREG/CR-5226. VISCOPLASTIC STRESS-STRAIN CHARACTERIZA.

N'Pl TION OF A533 GRADE B CLASS 1 STEEL.

USI A 17 NUREG 1229. REGULATORY ANALYSIS FOR RESOLUTION OF USl A-

• • • O*P'**I 17.Systern. Interactions in Nuclear Power Plants.

NUREG/CR-4918 V03 CONTROL OF WATER INFILTRATION INTO NEAR SURFACE LLW DISPOSAL UNITS. Progress Report Ull A 40 NUREG 1233. FiEGULATORY ANALYSIS FOR USl A-40,

• SEISMIC Water Chemtetty DESIGN CRITERIA " Final Report NUREG/CR 4667 V06. ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS Sernsannual ReportOctober 1967 - March USI A 47 1960.

NUREG 1217: EVALUATION OF SAFETY fMPLICATIONS OF CONTROL SYSTEMS IN LWR NUCLEAR POWER PLANTS.Techrucal Findings Water Infittret6on Related To OSI A-47. Final Rom NUREG/CR-4918 V03. CONTROL OF WATER INFILTRATON INTO NUREG 1218; REGULATORY ANALYSIS FOR RESOLUT60N OF USl A-NEAR SURF ACE LLW DISPOSAL UNITS. Progress Report

47. Safety Irnpications Of Control Systems in LWR Nuclear Power Plants. Fer:al Report W M TuH Unsetureted Contaminent NUREG/CR $393: ROCK MASS MODIFICATION AROUND A NUCLEAR NUREG/CR-S387. LOW LEVEL WASTE SHALLOW LAND DISPOSAL WASTE REPOSITORY IN WELDED TUFF.

SOURCE TERM MODEL: DATA INPUT GUIDES.

Yucca Mounte6n uneaturated Med6e NUPEG 1347; NRC STAFF SITE CHARACTERIZATION ANALYSIS OF NUREG/CH 5367: COMPARISON OF STRONGLY HEAT DRIVEN FLOW THE DEPARTMENT OF ENERGY'S SITE CHARACTERIZATION CODES FOR UNSATURATED MEDIA.

PLAN. YUCCA MOUNTAIN SITE NEVADA.

NUREG/CR 5426: EXAMINATION OF THE USE OF CONTINUUM Unnawated Medium VERSUS DISCONTINUUM MODELS FOR DESIGN AND PERFORM-NUREG-1373: TECHNICAL POSITION ON POSTCLOSURE SEALS.

ANCE ASSESSMENT FOR THE YUCCA MOUNTAIN SITE.

BARRIERS AND DRAINAGE SYSTEM IN AN UNSATURATED NUREG/CR 5427: ANALYSIS Or EMPLACEMENT DOREHOLE ROCK MEDIUM.

AND LINER BEHAVIOR FOR A REPOSITORY AT YUCCA MOUN-VENUS-3 TAIN.

NUREG/CR-5338. ANALYSl$ OF THE VENUS 3 EXPERIMENTS.

NUREG/CR4428; VARIATION OF HEAT LOADING FOR A REPOSI-TORY AT YUCCA MOUNTAIN.

Vendor inspection NUREG0040 V13 NO2: LICENSEE CONTRACTOR AND VENDOR IN-Yucca Mountain Site B SPECTION ST ATUS REPORT. Quartetty Report.Apnt June NUREG/CR-5400 BASIS FOR IN-SITU GEOMECHANICAL TESTING AT Book)

THE YUCCA MOUNTAIN Sl1E.

I

J i

NRC Originating Organization Index (Staff Reports)

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

ranged alphabetically by ma,or NRC organizations (e.g., program offices) and then by sub-sections of these (e.g., divis ons, branches) where appropnate. Each entry is followed by a NUREG number and title of the report (s), if further information is needed, refer to the main citation by NUREG number.

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

DIVISION OF HiGH-LEVEL WASTE MANAGEMENT (POST 870413)

FtEGON 1, Of C OF THE DIRECTOR NUREG 1347. NRC STAFF SITE CHARACTEnt2ATION ANALYSIS OF NUREG 0837 V09 Not: NRC TLD DIRECT RADIATION MONITORING THE DEPARTMENT OF ENERGY'S SITE CHARACTER 12ATON NETWORK Progress Report Janua'y-March 1989 NUREG 0837 V06 NO2. NRC TLD DIRECT RADIATION MONITORING PLAN YUCCA MOUNT AIN SITE. NEVADA.

NUREG 1373. TECHNICAL POSITION ON POSTCLOSURE SEALS.

T 1

BARRIERS AND DRAINAGE SYSTEM IN AN UNSATURATED OFC T

NUREG 0940 VOS NO2. NFORCEMENh ACTONS. SIGNIFICANT AC.

MEDIUM TIONS RE SOLVED Ouarterty Progress Report.ApniJune 1989.

DivlSON OF LOW LEVEL WASTE MANAGEMENT & DECOMMISSION.

OFC OF PERSONNEL (POST 8704138 ING (POST 070413)

NUREG4326 R12: U S. NUCLEAR REGULATORY COMMISSION NUREG 1336 R01: STANDARD FORMAT AND CONTENT GUIDE FUNCTIONAL ORGANIZATON CHARTSJuly 1,1989.

FOR FINANCIAL ASSURANCE MECHANISMS REQUIRED FOR DE.

~

COMMISSONING UNDER 10 CFR PARTS 30,40. 70 AND 72.

EDO. OFFICE OF ADMINISTRATION (PRE 870413 & POST $90206)

DIVISON OF FFIELDOM OF INFORMAllON & PUDLICATIONS SERV.

NUREG 1337 Rot: ST ANDARD REVIEW PLAN FOR THE REVIEW OF FINANCIAL ASSURANCE MECHANISMS FOR DECOMMISSIONING NU EG03 V1 01: REGULATORY AND TECHNICAL REPORTS UNDER 10 CFR PARTS 30,40,70 AND 72.

(ABSTRACT INDEX JOURNAL) Compilation For Frat Quarter NUREG 1376 V01: SAFETY EVALUATON STATUS REPORT FOR 1989,Janua March THE PROTOTYPE LICENSE APPLICATION SAFETY ANALYSIS NUREG 0304 14 NO2; REGULATORY AND TECHNICAL REPORTS REPORT Earth-Mounded Concrete Bur *er.

(A INDEX JOURNAL) Compilation For Second Quarter U.S. NUCLEAR REGULATO'lY COMMISSION NUREG.0540 Vit N05. TITLE LIST OF DOCUMENTS MADE PUBLIC-OFFICE OF THE GENERAL COUNSEL (POST 860701)

LY AVAILABLE.May 1 31 1989.

NUREG 0980 R04: NUCLEAR FIEGULATORY LEGISLATION.

NUREG 0540 Vit N66. TITLE LIST OF DOCUMENTS MADE PUBLIC.

NRC NO DETAILED AFFILIATION 'AIVEN LY AVAILABLE.Jene 130.198g NUREG/CR 4918 V03 CONTROL OF WATER INFILTRATON INTC6 i

NUREG 0760 V29101: INUEXES TO NUCLEAR REGULATORY COM-NEAR SURF ACE LLW DISPOSAL UNITS. Progress Report MISSON ISSUANCESJanuery March 1989 NUREG 0750 V29 N05. NUCLEAR REGULATORY COMMISSION IS.

EDO. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 320405)

SUANCE S FOR MAY 1989 Pages 395-463.

OFFICE OF NUCLEAR HEGULATORY RESEARCH, DIRECTOR (POST NMEGV750 V29 F.06: NUCLLAR REGULATORY COMMISSON IS-860720)

SUANCES FOR JUNE 1989 Pages 465 658 NUREG 1335' INDIVOUAL PLANT EXAMINATION SUDMITTAL NUREG 0936 VOS NO2; NRC REGULATORY AGENDA. Quarterly GUIDANCEftnal Report Report.AprilJune 1989-DIVISON OF ENGINEERING (POST 870413)

EDO OFFICE FOR ANALYSIS & EVALUATION OF OPER ATIONAL NUREG 1377; NRC RESEARCH PROGRAM ON PLANT AGING LIST-ING AND ABSTRACTS OF REPORTS ISSUED THROUGH FEDRU-OF FIC OR ANALYSIS & EVALUATON OF OPERATIONAL DATA, DI-D 0

GULATORY APPLICATONS (POST 870413)

NUREG 0090 V1P N01: REPORT YO CONGRESS ON ABNORMAL NUREG-0713 V08: OCCUPATIONAL RADIATON EXPOSURE AT OCCURRENCES. January March 1989 COMMERCIAL NUCLEAR POWER REACTORS AND OTHER FA-NUREG-1272 V03 N01: AEOD OFFICE FOR ANALYSIS AND EVAL.

CILITIES FOR 1986. Nineteenth Annual Report UATION OF OPERATIONAL DATA 1988 ANNUAL REPORT.Powy NUREG-1368: DRAFT PREAPPLICATION SAFETY EVALUATON Reactors REPORT FOR THE POWER REACTOR INHERENTLY SAFE NUREG 1272 V03 NO2 AEOD OFFICE FOR ANALYSIS AND EVAL-MODULE LOUC METAL REACTOR UATON OF OPERATONAL DATA 1988 ANNUAL DIVISION OF SAFETY ISSUE RESOLUTON (POST 680717)

REPORT.Nonreactors NUREG-1217; EVALUATION OF SAFETY IMPLICATONS OF CON-NUREG-1276 VDS ADO. OPERATING EXPERIENCE FEEDBACK TROL bYSTEMS IN LWR NUCLEAR POWER PLANTS.Techrwcal REPORT. PROGRESS IN SCRAM REDUCTON Commercial Power Findings Related to USI A-47. Fenal Report Reactors NUREG 1218 REGULATORY ANALYSIS FOR RESOLUTON OF USI A

.a knplicatons Of ContrW Systens in N har %

OFFICE OF GOVERNMENTAL & PUBLIC AFFAIRS (POST 870413)

ST ATE, LOCAL & INDIAN TRIBE PROGRAMS Plants. Final Report NUREG 1356 STATE COST SHARING OF TRAINING.A Task Force NUREG-1229 REGULATORY ANALYSl$ FOR RESOLUTON OF USI Report A 17. Systems interactions in Nuclear Power Plants.

NUREG 1233. REGULATORY ANALYSIS FOR USl A 40, " SEISMIC EDO. OFFICE OF INFORMATION RESOURCES MANAGTMENT & ARM DESIGN CRITERIA Final Report (8704?3-800204)

NUREG-1267, TECHNICAL RESOLUTON OF GENERIC SAFETY DIVISION OF COMPUTER & TELECOMMUNICATONS SERVICES ISSUE A-29 Nuclear Power Plant Design For Reduction Of Vulner.

(POST 890205)

NOREG-0020 Vt3 N06. LICENSED OPERATING REACTORS STATUS atulity To industrial Sabotage DIVtSON OF SYSTEMS RESEARCH (POST 880717)

CUMMARY REPORT. Data As Of May 31,1989 (Gray Book i)

NUREG-1150 Vot: SEVERF. ACCIDENT RISKS. AN ASSESSMENT NUREG 0020 V13 N07 LICENSED OPERATING REACTORS STATUS 50R FIVE U.S NUCLEAR POWER PLANT S Summary

SUMMARY

REPORT. Data As Of June 30.1989 (Gray Book 1)

NUREG4020 V13 NOB-LICENSED OPERATING REACTORS ST ATUS Report Second Dran For Pew Rowew

SUMMARY

REPORT. Data As Of Juty 31,1989 (Gray Boon 1)

NUREGd150 V02 SEVERE ACCIDENT RISKS AN ASSESSMENT FOR FIVE U S. NUCLEAR POWER PLANT 3 Apperdces Second EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS Draft For Peer Rewow OlVISON OF SUEQUARDS & TRANSPORT ATION (POST 670413)

NUREG 1365: REVISED SEVERE ACCIDENT RESEARCH PROGRAM NUREG-0525 R15. SAFEGUARDS

SUMMARY

EVENT LIST ($$EL).

PLAN Fiscal Year 1990 1992.

35 i

36 NRC Originating Organiz:ti n index (Staff R: ports)

ACCIDENT THREE MILE ISLAND NUCLt M STATION, UNIT 2.Fmal EDO. OFFICE OF NUCLEAR RE ACTOR REGULATION (PO$T 4/24/80)T OFFICE OF NUCLEAR REACTOR REGULATION. DIRECTOR (POS Sappiement Dealing Wittt.

870411)

NUREG 0974 $UPP; FINAL ENVIRONMENTAL STATEMENT RELAT.

NUREG 0800 02 4.2 R3: STANDARD REVIEW PLAN FOR THE ED TO THE OPERATON OF LIMERICK GENERATING STATION, REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR UNfTS 1 AND 2 Docket Nos. t,0-352 And 50 353.(PNiedepNa Elec-POWER PLANTS. LWR E$uon.Recsson 3 to SRP Section 2 4.2.

tnc Company)

"rloods "

NUREG-0991 SO9. SAFETY EVALUATION REPORT RELATED TO NOREG-0800 02 4 3 R3. STANDARD REVIEW PLAN FOR THE THE OPERATION OF LIMERICK GENERAtlNG STATION. UNITS 1 REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR AND 2. Docket Nos. 60352 And 50353.(PNiasespNa Electne Com.

POWER PLANTS. LWR E$ ton nevisen 3 To SRP Secton 2 4.3, pany)

" Probable Maximum riood (PMF) On Streams And Rivers."

NUREG 1371: TECHNICAL SPECIFICATIONS FOR LIMERICK GEN.

NUREG 0800 13.1.2 R3. STANDARD DEVIEW PLAN FOR THE ERATING STATION, UNIT 2. Docket No. 60 353.(PNidedelpNa Electne REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR Company)

POWER PLANTS. LWR E$ ton. Revision 3 To SRP Soeten 13.1.2-NUREG 1376-TECHNICAL $PECIFICATIONS FOR LIMERICK GEN-1313 "Operetng Organtration."

ERATING STATION, UNIT 2. Docket No. 60 353.(PNiedelpNa Electne NUREG-1335: INDIVIDUAL PLANT EXAMINATION SUBMITT AL Company)

GUIDANCE Final Report.

DIVISION OF REACTOR INSPECTION & SAFEGUARDS (POST DIVISION OF REACTOR PROJECTS.1/11(POST 870411) 870411)

NUREGMB3 $03 PROGRAMMATIC ENVIRONMENTAL IMPACT NUREG.0040 V13 N02: LICENSEE CONTRACTOR AND VENDOR IN.

STATEMENT RELATED TO DECONTAMINATION AND DISPOSAL SPECTION STATUS REPORT. Quanerty Report,AprtWune OF RADIOACTIVE WASTES RESULTING FROM MARCH 28,1979 1989 (WNte Book)

R m

i 4

=

NRC Originating Organization Index (International Agreements)

This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-ontry is followed by a NUREG number and title of the report (s), if further,opriate. Each fices) and then by subsections of these (e.g., civisions, branches) where appt information is needed, refer to the main citation by NUREG number.

EDO. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST $20406) f 4UREQ/tA-0016: ASSESSMENT OF RELAP$/ MOD 2. CYCLE 3604 OFFICE OF NUCLEAR REGULATORY RESE ARCH, DIRECTOR (POST AGAINST FIX Il GUILLOTINE BREAK EXPERIMENT NO. 5061.

860720)

NUREG/lA-0017: ASSESSMENT OF TRAC-PF1/ MOD 1 AGAINST A NUREG/tA@te: ANALYSIS OF THE THETIS BOILDOWN EXPERI-LOSS.0F. GRID TRANSIENT IN RINGHALS 4 POWER PLANT.

MENTS USING RELAPS/ MOD 2.

NUREG/tA-0020, ASSESSMENT STUDY OF RELAP5/ MOD 2 CYCLE T

N RELAP/5M 2 RAC

/ MOD CODES.

t 37 L________:____--_______________-_________

M m..mi--

--- i ii-i n

i i--i l

NRC Contract Sponsor Index (Contractor Reports)

This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program office)

End then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) prepared by that organi-zation. If further information is needed, refer to the main citation by the NUREG/CR number.

EDO OFFICE FOR ANALYSl$ & EVALUATION OF OPERATIONAL NUREG/CR 4219 V06 N1 HEAVY SECTON STEEL TECHNOLOGY DATA PROGRAM Semiannual Progress Report For Octoter 1980 - March OFFICE FOR ANALYSIS & EVALUATON OF OPERATIONAL DAT A. DL 1989 RECTOR NUREG/CR 4234 V02 AGING AND SERVICE WEAR OF E LECTRIC NUREG /CR2000 V00 N6 LICENSEE EVENT REPORT (LER)

MOTOR OPERATED V ALVES USED IN ENGINE ERED SAFETY-COMPILATION For Month 04 June 1989 FEATURE SYSTEMS OF NUCLEAR POWER PLANTS Aging At NUREG/CR-20U V0B N7 UCENSEE EVENT REPORT (LER) sessments And Monitonng Method Evaluations COMPILp rlON F or Month Of July 1989 NUREG/CR-4530 V03 U S / FRENCH JOINT RESE AxCH PROGRAM NUREG/CR-2000 V06 N8 LICENSEE EVENT REPORT (LER)

REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS DIVIS ON EAI AL ASSE E (POST 870413) i t

NUREG/u 5247 RASCAL VER$lON 1.3 USER'S OUIDE NUREG/CR-4667 V06 ENVIRONMENT ALLY ASS STED CRACAiNG NUREG/CR-6415 ENGINEERING DESIGN FOR THE NRC OPER' ATIONS CENTER IN LIGHT WATER REACTORS Semiannual Heport October 1987 -

"*'C" ' "8 DIVISION OF SAFETY PROGRAMS (POST 870413)

NUREG /CR 4744 V02 N2 LONG-TERM EMBRITTLEMENT OF CAST NUREGICR-4674 V07 PRECURSORS TO POTENTIAL SEVERE DUPLEX ST AINLESS STEELS IN LWR SYSTEMS $emiannual CORE DAMAGE ACCIDENTS 1987 A STATUS REPORT Main Report And Appendia A Report.Apol-Septernber 1967 NUREG/CR 4674 V0B PRECURSORS TO POTENTIAL SEVERE NUREG/CR 4018 V03 CONTROL OF WATER INFILTRATION INTO CORE DAMAGE ACCIDENTS 1987 A ST ATUS REPORT Appendmes NEAR SURFACE LLW DISPSSAL UNITS Progress Report B'C. And D NUREG/CR 4949 SOURCE TERM CALCULATIONS FOR ASSESS-ING RADIAllON DOSE TO EQUIPMENT EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NURE G/CR-4967 NUCLEAR PLANT AGING RESE ARCH ON HIGH OF FICE OF NUCLEAR MATERIAL SAFETY g

PRESSURE INJECTION SYSTEMS bAFEGUARDS. DIRECTOR NUREG/CR 4977 V01 SHAG TEST SERIES Seismic Research On An NUHEGICA-5004 RESOLUTON OF RECURRING LOSS ALARMS Aged United States Gate Valve And On A Pong Svalem in The De-i DIVISION OF HIGH LEVEL WASTE MANAGEMENT (POST B704:3) commissoned Heisodempfreaktor (HDR) Summary l

NUREG/CR-4708 V03 PROGRESS IN EVAli $ TION OF RADONU.

NUREG/CR 4977 V02 SHAG TEST SERIES Seismic Research On An CLIDE GEOCHEMICAL INrORM ATION u / ELOPED BY DOE Aged Unsted States Gate Vaive And Or A Pong System in The De.

HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE commissoned He#3sdamptreaktor (HDR) Appendices PROJECTS Report For October 1987 - June 1989 NUREG/CR 6161 V01 EVALUATION OF SAMPLING PLANS FOR IN NUREG/CR-5065 PROGRESS IN DEVELOPMENT OF A METHOD.

SERVICE INSPECTION OF STEAM GENERATOR TUDES Modeling OLOGY FOR GEOCHEMICAL SENSITIVITY ANALYSl$ FOR PER.

Of Eody Current Rehability Data. Analytical Evaluatens. And initial FORMANCE ASSESSMENT Parametnc Calculatons. Preliminary Da Monte Carlo Simulatons tabases And Computer Coos Eva:uaten NUREG/CR 5175 BE'TA AND GAMMA DOSE CALCULATIONS FOR NUREG/CR 5390 ROCK MASS MODIFICATON AROUND A NUCLE.

PWR AND BWR CONTAINMENTS AR WASTE REPOSITORY IN WELDED TUFF NUREG /CR-5226 VISCOPLASTIC STRESS-STRAIN CHARACTER NUREG/CR $400 BASIS FOR IN SITU GEOMECHANiCAL TEST NG 12ATON OF A533 GRADE B CLASS 1 STEEL AT THE YUCCA MOUNT AIN SITE NUREG /CR-523 : COBALT 60 SIMULATION OF LOCA RADIATION NUREG /CR-S426 EXAMINATON OF THE USE OF CONTINJUM EFFECTS VERSUS DISCONTINUUM MODELS FOR DESIGN AND PERFORM NUREG/CR5322 DETECTION AND CHARACTER 17ATION OF INDI-ANCE ASSESSMENT FOR THE YUCCA MOUNT AIN SITE CATONS IN SEGMENTS OF REACTOR PRESSURE VESSELS NUREG/CR-5427 ANALYSIS OF EMPLACEMENT BOREHOLE ROCg NUREG/CR 5338 ANALYSIS OF THE VENUS-3 EXPERIMENTS AND LINER BEHAVIOR FOR A REPOSITORY AT YUCCA MOUN NUREG/CR 5363 A STUDY OF TdE USE OF CROSSLINKED HIGH-TAIN DENSITY POLYETHYLE NE FOR LOW LEVEL RADIOACTIVE NUREG/CR 5428 VARIATION OF HEAT LOADING FOR A REPOSn.

WASTE CONTAINERS TORY AT YUCCA MOUNTAIN NUREG /CR-5367 COMPAP. SON OF STRONGL Y HE AT DRIVEN DIVISON OF LOW LEVEL W ASTE MANAGEMENT & DECOMMISSON.

FLOW CODES FOR UNSATURATED MEDIA ING (POST 870413)

NUREG /C45387 LOW-LEVEL W ASTE SHALLOW LAND DISPOSAL NOREG/CR 5431 A REVIEW OF GEOSCIENCE CHARACTERISTICS SOURCE TERM MODEL DAT A INPUT GUIDES AND DISPOSAL EXPERIENCE AT THE COMMERCLAL LOW-LEVEL NUREG /CR-5388 STEEL IMPURfTY ELEMENT EFFECTS ON POS-RADIOACTIVE WASTE DISPOSAL r ACILITY NEAR WEST V ALLEY.

TIRR ADi ATON PROPERTIES RECOVERY BY ANNEALING Fina!

NEW YORK Report NURE G /CR-5401 DONC STRENGTH OF CE ME NT BORE HOLE EDO OFFICE OF NUCLEAR REOUL ATORY RESEARCH (POST 820405)

PLUGS IN SALT OFFICE OF NUCLEAR REGULATORY RESE ARCH. DIRECTOR (POST NUREG'CR 5402 CRUSHED SALT CONSOLIDATION 860720)

DIVISION OF REGULATORY APPLICATIONS (POST B70413)

NUREG/CR 2331 V08 N4 SAFETY RESEARCH PROGRAMS SPON NUREG'CR 5396 FINANCIAL IMPACT OF IMPLEMENTING DRAFT SORED BY OFFICE OF NUCLEAR REGULATORY ANSr ST ANDARD N13 30. PERFORMANCE CRfTLRIA FOR RADIO.

RESE ARCH Progress Report,0ctober December 1988 BIOASSAT NUREG/CR 2331 V09 N1 SAFETY RESEARCH PROGRAMS SPON-DivlSION Or SAFETY ISSUE RESOLUTION (POST 880717)

SORED BY OrFICE OF NUCLEAR REGULATORY NUREG/CR-5225 ADD 01 AN OVERVIEW OF BWA M ARKa CON RESE ARCH Progress Report January-March 1989 T AINMENT VENTING RISK IMPLICATIONS An Evaluaton Of Poten-NUREG/CR 5004 RESOLUTON OF RECURRING LOSS ALARMS lus' MaC Containment improvements DIVISION OF ENGINEERING IPOST 870413)

DiviStON OF SYSTEMb RESE ARCH (POST 88071')

NUREG'CR 3252 NEW ENGL AND SEiSMOTEC'ONIC STUD 4 AC-NURE G 'CA -4550 v 4 A i ro ANA ' SIS Or CORE D/ MAGE FRE Tiv: TIES DURING riSCAL YE AR 1980 OUEND PE ACH gQTTQy gq7 p lNT[HN4g [y[N?$

39

40-NRC C2ntract Sponsor index NUREG/CR4550 V4R1P2: ANALYSIS OF CORE DAMAGE FRE-NUREG/C45273 V02:

SCDAP/RELAP5' MOD 2 CODE OUENCY; PEACH DOTTOM, UNIT 2, INTERNAL EVENTS APPEN-MANUALCCDAP Code Structure, Models. And Soluton Methods NUREG/CR-5273 V03 SCDAP/RELAP5/ MOD 2 CODE DICES.

MANUALUwr's Guide And input Regarements.

- NUREG/CR 4550 V6R1P1: J WLYSIS OF CORE DAMAGE FRE.

NUREG/CR 5339. DATA SUiWARY REPORT FOR FISSION PROD-QUINCY: GRAND GULF, UNil 1 INTERNAL EVENTS.

,NUREG/CR 4550 V6 RIP 2. ANALYSl0 OF CORE DAMAGE FRE-NU G 40 A SOMMARY REPORT FOR FISSION PROD-r QUENCY; GAAND GULF, UNIT 1 INTERNAL EVENTS APPENDL-UCT RELEASE TEST Vl-2.

4 CES.

NUREG/CR 5348. MAN-MACHINE INTERFACE ISSUES IN NUCLEAR NUREG/CR4639 V5P2R2 NUCLEAR COMPUTERIZED UBRARY POWER PLANTS. Report On A Workshop Held On January 10-12, FOR ASSESSING REACTOR RELIABluTY (NUCLAAR)Osta

198g, M<cual.Part 2: Human Enor Probabdity (HEP) Eshmates.

NUREG/CR-5395 V09-MULTILOOP INTEGRAL SYSTEM TEST NUREG/CR-4639 V5P382: NUCLEAR COMPUTERIZED UBRARY (MIST) FINAL REPORT. Ints Group Compansons.

FOR ASSESSING REACTOR REUADiLITY (NUCLARR). Data NUREGICR-5425: EVALUATION OF ALLOWED OUTAGE TIMES Manual.Part 3: Hardy are Cunporant Failure Data (HCFD).

(AOTS) FROM A RISK AND RELIABIUTY ST AdDPOINT.

NUREG/CR-4639 V5P4R2: NUr', LEAR COMPUTERIZED UBRARY g gR Rygg(

4/g/ )

A E

OF IC

-FOR ASSESSING REAOTCH REUABluTY (NUCLARR). Data NURE /CR 15T TH TABILITY OF CESIUM IODIDE.

PlPE LLS N RPOWER S

NUREG/C45174: A REFERFNCE M*NUAL FOR THE EVENT PRO-NUREG/CR-5290: CLOSEOUT OF BE BULLETIN 79-28.POSSIBLE CRESSION ANALYSl$ t'C.H (EVNTREL MALFUNCTION OF NAMCO MODEL EA180 UMIT SWITCHES AT NUREG/CR5263: THE RJK MANAGEMENT IMPLICATIONS OF ELEVATED TEMPERATURES.

NUREG-1150 METHODS AND REbulTS.

DIVISION OF UCENSEE PERFORMANCE & OVAUTY EVALUATION NUREG/CFt 5273 V01:

SCDAP/RELAP5/ MOD 2 CODE (POST 870411)

MANUALRESPS Code Structure, System Models, And Soluton NUREG/CR-5152: COMPARISON AND REGULATORY IMPACT OF Methods.

NOA 1 AND NOA-2 WITH N45.2 OA STANDARDS.

R a

E

=

]

?

4 1

1

Contractor Index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation.' Listed below each contractor. <e the NUREG/CR numbers and titles of their reports, If further information is needed, refer to the rnain citation by the NUREG/CR number.

ARGONNE NATIONAL LADORATORY NUREG/CR4967: NUCLEAR PLANT AGING RESEARCH ON HIGH NUREG/CR4667 V08: ENVIRONMENTALLY ASSISTED CRACKING IN PRESSURE INJECTION SYSTEMS.

. LIGHT WATER REACTORS. Semannual ReportOctober 1987 March NUREG/CR-4077 V01: SHAG TEST SERIES.Seismc Research On An 198tt i

Aged United States Gate Valve And On A Pipmg System in The De.

NUREG/CR 4744 V02 N2: LONG TERM EMBRITTLEMENT OF CAST comrmssioned Heisadampireaktor (HDR) Summary.

DUPLEX STAINLESS STEELS IN LWR SYSTEMS Semannual NUREG/CR-4977 V02: SHAG TEST SERIES.Seismc Research On An Report,Apni-September 1987, Aged United States Gate Vstve And On A Piping System in The De.

comrmssioned Heisadampfreaktor (HDR). Appendices.

ARIZONA, UNIV. OF, TUCSON, AZ NUREG/CR-5225 ADD 01: AN OVERVIEW OF BWR MARK.i CONTAIN-NUREG/CR-5401: BOND STRENGTH OF CEMENT BOREHOLE PLUGS MENT VENTING RISK IMPLICATIONS.An Evaluation Of Potential IN SALT.

Mark.1 Contamment improvements.

NUREG/CR-5402: CRUSHED SALT CONSOLIDATION.

NUREG/CR 5273

- V01:

SCDAP/RELAP5/ MOD 2 CODE

^

SAM,0CK & WILCOX CO.

INT EGRAL SYSTEM TEST oo, NUREG/CR-5395 V09: MULTILOOP (MIST) FINAL REPORT. Inter Group Compensons.

NUREG/CR-5273 V02: SCDAP/RELAP5/ MOD 2 CODE MANUALSCDAP Code Structure Models. And Solution Methods.

BATTELLE MEMORIAL INSTITUTE, COLUM8US LABORATORIES

^

G rnen NUREG/CR 4949: SOURCE TERM CALCULATIONS FOR ASSESSING RADIATION DOSE TO EQUIPMENT.

GRAM, INC, I

^

BATTELLE MEMORIAL INSTITUTE. PACIFIC NORTHWEST LADORATORY DES O U T RA M

NUREG/CR-5004: RESOLUTION OF RECURRING LOSS ALARMS.

IDAHO NATIONAL ENQlNEERING LABORATORY NUREG/CR-5181 VO1: EVALUATION OF SAMPLING PLANS FOR IN-NUREG/CR 4967: NUCLEAR PLANT AGING RESEARCH ON HIGH SERVICE INSPECTION OF STEAM GENERATOR TUBES Modehng Of PRESSURE INJECTION SYSTEMS.

Eddy Current Rehatility Data. Analytical Evaluatsons. And initial Monte Carlo Samulations.

NUREG/CR-4977 V01: SHAG TEST SERIES.Seisrme Rosearch On An NUREG/CR 5398: FINANCIAL lMPACT OF IMPLEVENTING DRAFT Aged United States Gate Valve And On A Pipmg System in The De-ANSI STANDARD N13.30, PERFORMANCE CRITERIA FOR RADIO-comrrssioned Heisadampfreaktor (HOR). Summary.

BIOASSAY.

NUREG/CR 4977 V02: SHAG TEST SERIEE.Senmic Research On An NUREG/CR 5431: A REVIEW OF GEOSCIENCE CHARACTERtSTICS Aged Urvted States Gale Valve And On A Piping System in The Os-AND DISPOSAL EXPERIENCE AT THE COMMERCIAL LOW LEVEL commissioned Heisadampfreaktor (HDR): Appendices RADIOACTIVE WAGTE DISPOSAL FACILITY NEAR WEST VALLEY.

ITASCA CONSULTING GROUP,INC.

NEW YORK.

NUREG/CR-5390. ROCK MASS MODIFICATION AROUND A NUOLEAR BOSTON COLLEGE, WESTON, MA WASTE REPOSITORY IN WELDED TUFF.

NUREG/CR-3252: NEW ENGLAND SEISMOTECTONIC STUDY ACTIVI-NUREG/CR-5400: BAtlS FOP. IN-SITU GEOMECHANICAL TESTING AT TIES DURING FISCAL YEAR 1980.

THE YUCCA MOUN1 A.N SITE.

NUREG/C.15426: EXAMINATION OF THE USE OF CONTINUUM BROOKHAVEN NATIONAL LABORATORY VERSUS DISCONTINUUM MODELS FOR DESIGN AND PfRFORM-NUREG/CR-2331 V08 N4: SAFETY RESEARCH PROGN11S SPUN-ANCE ASSESSMENT FOR THE YUCCA MOUNTAIN SITE.

NUREG/CR-542h ANALYSIS OF EMPLACEMENT BOREHOLE ROCl'.

SORED BY. OFFICE OF NUCLEAR REGULATORY AND UNER BEHAVIOR FOR A REPOSITOP.Y AT YUCCA MOUN-RESEARCH Progress Report. October-December 1988' GRAMS SPON-TAIN NUREGICR-2331 V09 Nt: SAFETY RESEARCH PRO NUREG/CR-5428: VARIATION OF HEAT LOADING FOR A REPOSI-SORED BY OFFICE OF NUCLEAR REGULATORY TORY AT YUCCA MOUNTAIN.

RESFARCH Dr gross Report. January-March 1989.

NUREG/CR-5363: A STUDY OF THE USE OF CRC %SLINKED HIGH-JAPAN ATOMIC ENERGY RESEARCH INSTITb7E DENSITY POLYETHYLENE FOR LOW LEVEL RADIOACTIVE WASTE CONTAINERS.

NUREG/CR-5339. DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST VI 1.

NUREG/CR-5387. LOWU4 L WASTE SHALLOW LAND DISPOSAL SOURCE TERM MOCG t ATA INPUT GUICES.

MARYLAND, UNIV. OF, CULEGE PARK, MD NUREGICR 5425: EVAa,ATIOf4 OF ALLOWED OUTAGE TIMES NUREG/OR-4918 V03: CONTROL OF WATER INFILTRATION INTO (AOTSi FROM A RISK AND RELIABILITY STANDPOINT.

NEAR SURFACE LLW D7POSAL UNITS. Progress Report.

CALIFORNIA, UNIV. OF BERKE.EY. CA MATERIALS ENGINEERING ASSOCIATES,INC.

NUREG/CR-4918 V03: CONTROL OF WATER INFILTRATION INTO NUREG/CR-5368: STEEL IMPUR!TY ELEMENT EFFECTS ON POSTIA.

NEAR SURFACE LLW Dt% SAL UNITS. Progress Repors.

RADICION PROPERTIES RECOVERY BY ANNEALING Foal Report.

EG40 'r1AHO, INC, (SUBS. OF EGAG. INC.)

OAK RlDGE NATIO-4AL LABORATORY NUREG/CR 4639 V5P2P2: NUCLEAR COMPUTERIZED LIBRARY FOR NUREG/CR-2000 V08 N6: LICENSEE EVENT REPOR7 (LER)

ASSESSING REACTOH RELIABILITY (NUCLAAR). Data Manual.Part ?:

COMPILATION:For Month Of June 1989 Human Enor ProbatAity (HEP) Estimates.

NUREG/CR 2000 V08 N7: UCENSEE EVENT REPORT (LER)

NUREG/CR-4639 V5P3R2. NUCLEAR COMPUTERIZED UBR*RY FOR COMPILATION For Month Of July 1989 ASSESSING REACTOR RELIABlUTY (NUCLARR). Data Manu.J Part 3:

NUREG/CR 2000 V0B N8. UCENSEE EVENT REPORT (LER)

Hacdware Component Failure Data (HCFDA COFILATION:For Month Of August 1989.

NUREG/CR 4639 V5P4R2: NUCLEAR COMPUT5RIZED UBRARY FOR NUREG/CR 4219 V06 N1: HEAVY-SECTION STEEL TECHNOLOGY ASSESSING fiEACTOR RELIABILITY gNUCLARR). Data Manual.Part 4:

PROGRAM.Senuannual Progress Report For October 1989 March Summary Aggregatsuns.

1989.

i 41

42 C ntract:r index NUREG/CR-4234 V02: AGING AND SERVICE WEAR OF ELECTRIG NUREG/CR4550 V6 RIP 2: ANALYSIS OF CORE DAMAGE FREQUEN-MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY FEA-CY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

TURE SYSTEMS OF NUCLEAR POWER PLANTS Aging Assessments NUREG/CR 50BS: PROGRESS IN DEVELOPMENT OF A METHODOLO-And Monitoring Method Evaluations GY FOR GEOCHEMICAL SENSITIVITY ANALYSl$ FOR PERFORM.

NUREG/CR4574 V07. PRECURSORS TO POTENTIAL SEVERE CORE ANCE ASSESSMENT.Parametnc Calculations, Prelmnary Databases DAMAGE ACCIDENTS: 1967 A STATUS REPORT. Main Report And And Computer Code Evaluation.

Appenda A.

NUREG/CR 5155: THE THERMAL INSTABILITY OF CESIUM IODIDE.

NUREG/CR4674 V08: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR 5174: A REFERENCE MANUAL FOR THE EVENT PRO-DAMAGE ACCIDENTS 1987 A STATUS REPORT.Appendues B.C, And GRESSION ANALYSIS CODE (EVNTRE)

NUREG/CR 4708 V03. PROGRESS IN EVALUATION OF RADIONU.

R AND B A NMENTS CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH*

NUREG/CR-5231: COBALT 60 SIMULATION OF LOCA RADIATION EF-LEVEL NUCLEAR WASTE REPOSITORY SITE PFsOJECTS Report For FECTS.

^

NU C 5 S L TION OF RECURRING LOSS ALARMS

^'" "^"^

NUREG/C 56:

M AR N STRONGLY HEAT DRIVEN FLOW T

OF A533 G A 1 EEL NUREG/CR-5247: RASCAL VERSION 1.3 USER'S GUIDE.

CODES FOR UNSATURATED MEDIA.

NUREG/CR-5322: DETECTICH AND CHARACTERIZATION OF INDICA-SCIENCE APPLICATIONS INTERN ATIONAL CORP,(FORMERLY TlONS IN SEGMENTS OF REACTOR ORESSURE VESSELS.

NUREG/CR-5338: ANALYSIS OF ThE VENUS-3 EXPERIMENTS SCIENCE APPLICATIONS, NUREG/CR 5339: DATA

SUMMARY

AEPCRT FOR FISSION PRODUCT NUREG-0713 V08. OCCUPATIONAL RADIATION EXPOSURI AT COM-RELE ASE TEST VI 1-MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES NUREG/CR-5340. DAT A

SUMMARY

REPORT FOR FISSION PRODUCT FOR 1986.Nneteenth Annual Report RELEASE TEST VI NUREG/CR 4550 V4 RIP 1: ANALYSIS OF CORE OAMAGE FREQUEN-CY: PEACH BOTTOM. UNIT 2, INTERNAL EVENTS NUREG/CR 4550 V4R1P2 ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR i: CLOSEOUT OF NRC BULLETIN 87 01;TH NNING OF CY: PEACH BOTTOM. UNIT 2, INTERNAL EVENTS APPENDICES.

NUREG/CR4550 V6 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN-NU EG CR UT I

L IN 79-28.POSSIBLE MAL-F CiN NM MODEL EAleo LIMIT SWITCHES AT ELE-NUR /CR-45 0 V6'R1P2. A ALYS S C

DAMAGE FREQUEN-CY: GRAND GULF, UNIT 1 INTERNAL EVENTS APPENDICES.

SANDIA NATIONAL LABORATORIES NUREG/CR-5152: COMPARISON AND REGULATORY IMPACT OF NUREG/CR-4530 V03: U.S1 FRENCH JOINT RESEARCH PROGRAM NOA 1 AND NOA 2 WITH N45 2 CA STANDARDS REGARDING THE BEHAVIOR OF POLYMER BASE MATERIALS SUB.

NUREG/CR 5174: A REFERENCE MANUAL FOR THE EVENT PRO-JE D TO BETA RADIATION Volume 3: Phase 2b Expanded Test NUREG/ 53 M CHI E NTE FACE ISSUES IN NUCLEAR NUHEG/CR4550 V4 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN.

POWER PLANTEReport On A Workshop Held On January 1012, CY: PEACH BOTTOM. UNIT 2, INTERNAL EVENTS.

1989.

NUREG/CR4550 V4 RIP 2: ANALYSIS OF CORE DAMAGE FREQUEN.

NUREG/CR-54i$: ENGINEERING DESIGN FOR THE NRC OPER-CY: PEACH BOTTOM. UNIT 2, INTERNAL EVENTS APPENDICES.

ATIONS CENTER.

HUREG/CH 4550 V6 RIP 1: ANALYSIS OF CORE DAMAGE FREOVEN-NUREG/CH 5425: EVALUATION OF ALLOWED OUTAGE TIMES CY; GRAND GULF, UNIT 1 INTEMAL EVENTS.

(AOTS) FROM A RISK AND REllABILITY STANDPOINT.

'l 4

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International Organization Index Th's index lists, in alphabetical order, the' countries and performing organizations that pre-i pared tha NUREG/lA reports listed in this compilation. Listed below each country and per-forming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number.

DELGIUM NUREG/lA 0017: ASSESSMENT OF TRAC.PF1/ MODI AGAINST A N

IA@20: ASSESSMENT STUDY OF RELAPS/ MOD 2, CYCLE 36.04 BASED ON SPRAY START.UP TEST FOR DOEL.4 UN K

SWEDEN NUREG/lA-0014: ANALYSIS OF THE THETIS BOILDOWN EXPERI-NU

/A 16 A SSMENT O RE P5/ MOD 2, CYCLE 36.04 NURE /lA 5A ME OF INTERPHASE DRAG CORRELA.

AGAINST FIX-II GUILLOT!NE BREAK EXPERIMENT NO. 5061 TIONS IN THE

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Licensed Facility index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

1

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$416 Grand GJf Nuclear Sunon, Urut 1. Essesppi HUREG/CR4550 V6AIP1 S 353

' Lnenck Generatrg Statm Urut 2, PMadelptua NURE41371 Power & Ugle Co Electnc Co.

4 416 Grand Gdt Nuclear Statiort Urut 1, hisstopp NUREG/CR4550 V6 rip 2 S 353 Lrnero Generstm0 Staten, Unit 2 P%delphe NURE41376 Pceer & Ught Co.

Elecinc k M352 -

Umarr* Generatng Stanon, Unit 1. PMadelpha NUREG474 $UPP l

50-2U Peach Bottom Atome Power $tston, brut 2.

NUREG/CR4500 V#a.'1 G352 atog Staton, Urut 1, PWadelptua NURE40991 SOD S2D Peach Power Staten, Unit 2, NUREG/CR.4550 V4 RIP 2 a'aa.n S m 2.

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11, ADST R AC1 troo noros or un a This journal-includes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceedings of conferences and workshops; as well as international agreement reports. The entries in this comoilation are indexed for access by title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agreements, contractor.. international organization, and licensed facility.

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PUBLICATIONS SVCS-TPS PDR-NUREG A H NGTON DC 20555 Secondary Report Number index Personal Author index Subject index E

NRC< Originating Organization Index (Staff Reports)

Eaj NRC Originating Organization index (International Agreements)

NRC Contractor SponsorIndex Contractor Index L international Organization g

Index En

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