ML20087G135

Regulatory and Technical Reports (Abstract Index Journal). Compilation for First Quarter 1995,January-March

ML20087G135
Person / Time
Issue date:	07/31/1995
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V20-N01, NUREG-304, NUREG-304-V20-N1, NUDOCS 9508160248
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NUREG-0304 Vol. 20, No.1 Regulatory and Technical Reports (Abstract Index Journal) l Compilation for First Quarter 1995 January - March U.S. Nuclear Regulatory Commission Office of Administration f

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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1.

The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2.

The Superintendent of Documents U.S. Government Printing Office, P. O. Box 37082, Washington, DC 20402-9328 3.

The National Technical Information Service, Springfield, VA. 22161-0002 l

Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

i Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda: NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu-ments and correspondence.

The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, international agreement reports, grantee reports, and NRC booklets and bro-chures, Also available are regulatory guides, NRC regulations in the Code of Federal Regula-tions, and Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service include NUREG-series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions, Federal Register notices, Federal and State legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC con-forence proceedings are available for purchase from the organization sponsoring the publica-tion cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library Two White Flint North,11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if thesy are American National l

Standards, from the American National Standards institute,1430 Broadway, New York, NY l

10018-3308,

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l

J NUREG-0304 Vol. 20, No.1 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for First Quarter 1995 January - March

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Date Published: July 1995 Regulatory Publications Branch Division of Freedom ofInformation and Publications Services Omce of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 f - s,,

s.....

I CONTENTS Preface.........

v Index Tab Main Citations and Abstracts.....

1 e Staff Reports e Conference Proceedings e Contractor Reports e Grant Reports e International Agreement Reports Secondary Report Number index.

2 Personal Author index................

3 Subject Index......

4 NRC Originating Organization index (Staff Reports) 5 6

NRC Originating Organization index (International Agreements).

NRC Contract Sponsor index (Contractor Reports) 7 Contractor index.......

8 International Organization index 9

Licensed Facility index.......

. 10 i

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PREFACE.

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This compilation consists of bibliographic data and abttracts for the formal regulatory and technical reports -

' Issued by the U.S. Nuclear Regulatory Commission (Ni1C) Staff and its contractors. it is NRC's intention to r

publish this compilation quarterly and to cumulate it annut fly. Your comments will be appreciated. Please send i

the m to:

4 s

Technical Publications Section Publications Branch Division of Freedom ofinformation and Publications Services T-6 E7.

U.S. Nuclear Regulatory Commission '

Washington, D.C. 20555-0001 j

The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA XXXX. These precede the following indexes:

Secondary Report Number index f

Personal Authorindex Subject index

-i NRC Originating Oroanization index (Staff Reports) l NRC Originating Cvqanization index (Intemational Agreements)

NRC Contract Sponsv;Index (Contractor Reports) i Contractor index l

International Organization index l

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

The bibliographic elements of the main citations are the following:

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

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

j Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date

- i report was published, (6) romber 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 RELIABluTY ENGINEERING IN NUCLEAR REGULATION, JANERP, J.S. Argonne National Laboratory. May 1981,141 pp.

I 8105280299. ant,-81-3. 08632:070.

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

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~ Contractor Report 1

NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR UGHT WATER REAC-TORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, RR. Sandia Laboratories.

j Mrj 1981,100 pp. 8107010449. SAND 80-0929. 08912:242.

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

publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control Sys-

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- tem accession number, (8) the report number of the originating organization (if given), (9) the microfiche ad-i dress (for NRC intamal use).

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Grant Report j

NUREG/GR-0013: APPUCATIONS OF A NEW MAGNETIC MONITORING TECHNIQUE TO IN SITU EVALUA-TION OF FATIQUE DAMAGE IN FERROUS COMPONENTS. JILES, D.C.; BINER, S.B.; GOVINDARAJU, M.; et al. Iowa Sta*a i.iniv., Ames, IA. June 1994. 41 pp. 9407250286. 80328:195.

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

publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control Sys-tem accession number, (8) the report number of the originating organization (if given), (9) the microfiche ad-dress (for NRC Internal use).

Intemational Agreement Report NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA.

l NEUMANN, U. Kraftweek Union. August 1986. 223 pp. 8608270424. 37659:138.

Where the entries are(1) report number, (2) report title, (3) repnrt 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 rep 0rt 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 i

ERR

- errata N - number R - revision S - supplement V - volume Availability of NRC Publicat one, I

Copies of NRC staff and contractor reports may be purchased either from the Govemment Printing Office (GPO) or from the National Technicallnformation 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) 512-2249 or (202) 512-2171. 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.

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NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report.

Contractor-propared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established codes such as ORNUNUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship or the work being reported.

in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference pro-ceedings NUREG/GR is used for NRC grant reports, and NUREG/lA is used for international agreement reports.

All these report codes are controlled and assigned by the staff of the Technical Publications Section of the NRC l

DMslon of Freedom of Information and Publications Services.

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Main Citations and Abstracts I

The report listings in this compilation are arranged by report number, where NUREG-XXXX is an NRC staff-onginated report, NUREG/CP-XXXX is an NRC-sponsored conference report,

, NUREG/CR-XXXX is an NRC contractor-prepared report, and NUREG/lA-XXXX is an inter-nItional agreement reaort. The bibliographic information (see Preface for details) is followed by a brief abstract of 11is report.

NOREG-0040 V18 N04: LICENSEE CONTRACTOR AND NUREG-0304 V19 N04: REGULATORY AND TECHNICAL RE-VENDOR INSPECTION STATUS REPORT. Quarterly PORTS (ABSTRACT INDEX JOURNAL). Annual Compilaton Report. October. December 1994. (White Book)

• Office of No.

For 1994.

• Division of Freedom of Information A Pubications clear Reactor Regulatton (Post 941001). February 1995.200pp.

Services (Post 940714). March 1995. 133pp. 9504180345.

9503140385. 83069:031.

83566.063.

+

This perodcal c:nrers the FSsults of inspections performed by See NUREG-0304,V19,NO3 abstract.

the NRC's Special inspecton Branch, Vendor inspection Sec-NUREG-0390 V09 N01: TOPICAL REPORT REVIEW STATUS.

• ton, that have been distnbuted to the inspected organizatons Office of Nuclear Reactor Regulaton (Post 941001). February dunng the penod from October through December 1994.

1995. 44pp. 9503140388. 83068-272.

NUREG-0090 V17 N03: REPORT TO CONGRESS ON ABNOR-This report provides industry with procedures for submitting MAL OCCURRENCES. July-September 1994.

• Offee for Anaty-topcal ps, @ance on W N M Mar Wam l

Commisson (NRC) processes and responds to topical report sis A Evaluston of Operatonal Data, Director. January 1995-su als, a an accountng, e ew scWes, of an W 35pp. 9504190361 83575:183-cal reports currently acceptea for review by the NRC. This Section 208 of the Energy Reorganization Act of 1974 identi-report is published semiannually.

fies an abnormal occurrence (AO) as an unscheduled incident or event that the Nuclear Regulatory Commission determines to NUREG-0430 V14: LICENSED FUEL FACILITY STATUS be segrufcant from the standpoint of public health or safety and REPORT. Inventory Difference Data. July 1,1993 - June 30, requires a quarterly report of such events to be made to Con.

1994. JOY,D.R. Office of Nuclear Matenal Safety & Safeguards.

(

grras. This report provides a desenption of those events that March 1995.18pp. 9505100290. 83902:267.

hove been determined to be abnormal occurrences dunng the NRC is committed to the penode publication of licensed fuel penod of July 1 through September 30,1994. This report ad.

facility inventory difference data, following agency review of the dresses five abnormal occurrences (AOs) at NRC-licensed fa.

informaton and completion of any related NRC investigations.

cilities. One involved a medical brachytherapy misadministration, Informaton ir, this report includes inventory difference data for two involved medical teletherapy misadministratons, one in-active fuel fabncation facilities possessing more thar* one effec-volved a medical codium iodido misadministraton, and one in.

tive kilogram of special nuclear material.

volved a medical sodium iodide event. One AO report submitted NUREG-0540 V16 N11: TITLE LIST OF DOCUMENTS MADE by an Agreement State is included. It involved the loss of man-PUBLICLY AVAILABLE. November 1 30, 1994.

• Division of cgement and procedural control of a radioactive source. (Due to Freedom of informaton & Publications Services (Post 940714).

pubication schedule constraints, NRC was unable to include all January 1995. 307pp. 9502080222. 82662:001, of the AO information received from the Agreement States. Any This document is a monthly publication containing desenp-Agreement State information that was not included in this report tions of information received and generated by the U.S. Nuclear will be published in the next quarterly report. The report also Regulatory Commission (NRC). This information includes (1) contains updates of six AOs prevously reported by NRC licens-docketed matenal associated with civilian nuclear power plants ees and three AOs previously reported by Agreement State li-and other uses of radioactive materials, and (2) nondocketed consees. Two "Other Events of interest" conceming nuclear matenal received and generated by NRC pertinent to its role as power reactors are also reported. One involved the fracture of a a regulatory agency. The following indexes are included: Per-frozen pipe at Dresden Unit I with a consequent release of sonal Author, Corporate Source, Report Number, and Cross witer, and the other involved the possible deliberate exposure Reference of Enclosures to Principal Documents.

of a contract laborer to radiation at Quad Cities Nuclear Power

- St ton.

NUREG-0540 V16 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. December 1 31, 1994.

• Division of NUREG-0304 V19 NO3: REGULATORY AND TECHNICAL RE.

Freedom of information & Publications Services (Post 940714).

PORTS (ABSTRACT INDEX JOURNAL). Compilation For Third February 1995. 319pp. 9503150148. 83100:285.

t Quarter 1994 July-September.

• Division of Freedom of informa.

See NUREG-0540,V16,N11 abstract.

i ton & Publications Sennees (Post 940714). December 1994.

NUREG-C540 V17 Not: TITLE LIST OF DOCUMENTS MADE 62pp. 9502000121. 82660.001.

PUBLICLY AVAILABLE. January 1-31, 1995.

• Division of Free.

This joumal includes all formal reports in the NUREG senes dom of Information & Publicatons Services (Post 940714).

prepared by the NRC staff and contractors, proceedings of con-March 1995. 336pp. 9504120107. 83475:001, firences and workshops, grarts, and intomational agreement See NUREG-0540,V16,N11 abstract.

riports. The entnes m this compilaton are indexed for access by title and abstract Secondary report number, personal author.

NUREG-0700 R01 DFC: HUMAN-SYSTEM INTERFACE DESIGN subect, NRC organizaten for staff and internatoral agree-REVIEW GUIDELINE. Draft Report For Comment. O'HARA.J M.;

l ments, contractos, international organizaton, and Icense.d facili-BROWN.W.S.; STUBLER,W.F.; et af. Brookhaven Natiord Lab-ty.

oratory. February 1995. 496pp 9503270312. 83269:001.

1

I 2

Main Citations and Abstracts NUREG-0700, Rev.1, provioes human factors engineenng NUREG-0750 V41 N01: NUCLEAR REGULATORY COMMISSION i

(HFE) guidance to the U.S. Nuclear Regulatory Commission ISSUANCES FOR JANUARY 1995. Pages 1-69.

• Division of staff for rts: (1) review of the human system interface (HSI)

Freedom of information & Pubhcations Services (Post 940714).

design submsttals prepared by licensees or applicants for a li-March 1995. 75pp. 9503280211. 83302:178.

cense or design certification of commercial nuclear power See NUREG-0750,V40,N05 abstract.

J plants, and (2) performance of HSI reviews that could be under-taken as part of an inspection or other type of regulatory review NUREG-0837 V14 N04: NRC TLD DIRECT RADIATION MONI.

involving HSI design or incidents involving human performance.

TORING NETWORK. Progress Report. October-December 1994.

i it consists of two major parts. Part i descnbes those aspects of STRUCKMEYER.R. Region 1 (Post 820201). March 1995.

J the HSI design review process that are important to the edentifi-329pp. 9503170312. 83147:001.

j cation and resolution of human engineenng discrepancies that This report provides the status and results of the NRC Ther-could adversely affect plant safety. Guidance is providod that moluminescent Dosimeter (TLD) Direct Radiation Monitoring could be used by the staff to review an applicant's HSl design Network. It presents the radiation levels measured in the vicinity review process. Part I could also be used by the staff to guide of NRC licensed facilities throughout the country for the fourth i

the development of an HSI design review plan, e g., as part of quarter of 1994.

an inspection activity. Part 2 "Guidehnes for Human Factors NUREG-0936 V13 N03:

NRC REGULATORY Engineenng Reviews, provides detailed HFE guidelines for the assessment of HSI design implementations.

AGENDA. Semiannual Report July-December 1994.

• Division of i

Freedom of Information & Publications Services (Post 940714).

NUREG-0713 V15: OCCUPATIONAL RADIATION EXPOSURE AT February 1995. 56pp. 9503140450. 83071:106.

l COMMERCIAL NUCLEAR POWER REACTORS AND OTHER The NRC Regulatory Agenda is a compilation of all rules on FACILITIES,1993 Twenty. Sixth Annual Report RADDATZ C T.

which the NRC has recently completed action, or has proposed Division of Regulatory Applications (Post 941217) action, or is considenng action, and all petitions for rulemaking HAGEMEYER,D. Science Applications International Corp. (for, which have been received by the Commission and are pending merly Science Applications. Inc ).

January 1995. 307pp.

disposition by the Commission. The Regulatory Agenda is up-9502080242. 82661:001.

dated and issued semiannually.

This report summanzes the occupational radiation exposure NUREG-0940 V13 N4 P1:

ENFORCEMENT information that has been repo,ted to the NRC's Radiation Ex-ACTIONS SIGNIFICANT ACTIONS RESOLVED REACTOR posure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees dunng LICENS.EES.Ouarterly Progress Report October-December 3994.

Ofc of Enforcement (Post 870413). February 1995.

the years 1969 through 1993. The bulk of the data presented in 360pp. 9503140428. 83071:171.

the report was obtained frorn annual radiation exposure reports This compilation summanzes significant enforcement actions submitted in accordance with the requirements of 10 CFR that have been resolved dunng one quarterly penod (October -

20 407 and the technical specifications of nuclear power plants December 1994) and includes copies of letters Notices, and Data on workers terminating their employment at certain NRC Orders sent by the Nuclear Regulatory Commission to reactor licensed facilities were obtained from reports submitted pursu-licensees with respect to these enforcement actions. It is antici-ant to 10 CFR 20.408 The 1993 annual reports submitted by pated that the information in this publication will be widely dis-about 360 licensees indicated that approximately 189,711 indi-seminated to managers and employees engaged in activities li-viduals were monitored, 169,872 of whom were monitored by censed by the NRC so that actions can be taken to improve nuclear power facilities. They incurred an average individual safety by avoiding future violations similar to those desenbed in dose of 016 rem (cSv) and an average measurable dose of this pubhcation.

about 0 31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 99,749 individuals completed their NUREG-0940 V13 N4 P2:

ENFORCEMENT employment with one or more of the 360 covered licensees ACTIONS SIGNIFICANT ACTIONS RESOLVED MEDICAL dunng 1993. Some 91,000 of these individuals terminated from LICENSEES.Ouarterly Progress Report, October-December power reactor facihties, and about 12.685 of them were consid 1994.

• Ofc of Enforcement (Post 870413). February 1995.

ered to be transient workers who received an average dose of 300pp. 9503140434. 83077:001.

0 49 rem (cSv).

This compilation summanzes significant enforcement actions that have been resolved dunng one quarterly penod (October -

NUREG-0750 V40102: INDEXES TO NUCLEAR REGULATORY December 1994) and includes copies of letters, Notices, and COMMISSION ISSUANCES. July-December 1994

• Division of Orders sent by the Nuclear Regulatory Commission to medical Freedom of Information & Publications Services (Post 940714) licensees with respect to these enforcement actions. It is antici-March 1995 40pp. 9504120101. 83474 001, pated that the information in this pubhcation will be widely dis-Digests and indexes for issuances of the Commission, the seminated to managers and employees engaged in activities li-Atomic Safety and Licensing Board Panel, the Administrative consed by the NRC, so that actions can be taken to improve Law Judges, the Directors' Decisions, and the Denials of Peti-safety by avoiding future violations similar to those desenbod in tions for Rulemaking are presented this publication.

NUREG-0750 V40 N05: NUCLEAR REGULATORY COMMISSION NUREG-0940 V13 N4 P3:

ENFORCEMENT ISSUANCES FOR NOVEMBER 1994. Pages 169-318

• Division ACTIONS SIGNIFICANT ACTIONS RESOLVED MATERIAL Lt-of Freedom of information & Pubhcations Services (Post CENSEES (NON-MEDICAL) Quarterly Progress Report. October-940714) February 1995 149pp 9503150152. 83107:001.

December 1994.

• Ofc of Enforcement (Post 870413). February Legalissuances of the Commission, the Atomic Safety and Li-1995. 360pp. 9503140438. 83075:001.

consing Board Panel, the Administrative Law Judges, and NRC This compilation summanzes sigruficant enforcement actions Program Offices are presented.

that have been resolved dunng one quarterly penod (October -

December 1994) and includes copies of letters, Notices, and NUREG-0750 V40 N06: NUCLEAR REGULATORY COMMISSION Orders sent by the Nuclear Regulatory Commission to Matenal ISSUANCES FOR DECEMBER 1994. Pages 319 387.

• Division Licensees (non-Medical) with respect to these enforcement ac-of Freedom of information & Pubhcations Services (Post tions It is anticipated that the information in this pubhcation will 940714). Fetaruary 1995 77pp. 9503150157. 83101242.

be widely disseminated to managers and employees engaged in See NUREG 0750,V40,N05 abstract.

actrvities bcensed by the NRC, so that actions can be taken to

1 Main Citations and Abstracts 3

improve safety by avoiding future violations similar to those de-rent LWR heensees may voluntarily propose applications based scribed in this publication.

upon it.

NUREG-1100 VII: BUDGET ESTIMATES Fiscal Years 1996 NUREG 1493 DFC: PERFORMANCE-BASED CONTAINMENT 1997.

• Dwision of Budget & Analysis (Post 890205). February 1995. 201pp. 9504180341,83565:223.

LEAK-TEST PROGRAM. Draft Report For Comment. DEY,M. Di-This report contains the fiscal year budget justifcation to Con-vision of Regulatory Applications (Post 941217). SKOBLAR.L.;

gress. The budget provides estimates for salaries and expenses CHEN.P.; et al. S. Cohen & Associates, Inc. January 1995.

and for the Office of the inspector General for fiscal years 1996 260pp. 9503150142. 83100:001.

and 1997.

Nuclear Regulatory Commission (NRC) is implementing an ini-NUREG-1435 SO4: STATUS OF SAFLTY ISSUES AT LICENSED 1

" * * ~

POWER PLANTS.TMl Action Plan Requirements, Unresolved Safety issues.Genenc Safety issues.Other Multiplant Action issues.

• Office of Naclear Reactor Regulation (Post 941001)'

December 1994.157pp. 9502080207. 82686.001.

ments could replace the current prescriptive requirements with As part of ongoing U.S. Nuclear Regulatory Commission nly a marginal impact on safety. This technical support docu-(NRC) efforts to ensure the quality and accountability of safety ment (TSD) provides the technical bases for the NRC's rule-issue information, a program was established whereby an ng se leaWshng m@ments b ndar pows re-annual NUREG report would be published on the status of li-actors in 10 CFR Part 50, Appendix J. This report ident:fies a!-

censee implementatron and NRC venfication of safety issues in tematwes to cunent containment testing requirements which major NRC requirements areas. This information was compiled would meet the NRC's Safety Goals and achieve greater effi-and reported in three NUREG volumes. Volume 1, published in ciency in the use of resources. Changes in the allowable leak March 1991, addrassed the status of Three Mile Island (TMI) rate for containment and the testing frequencies for both inte-Action Plan Requirements. Volume 2, pubhshed in May 1991, grated and local leak rate tests are evaluated in terms of both addressed the status of unresolved safety issues (USIs).

nsk and cost impacts. The feasibility of applying statistically-Volume 3, published in June 1991, addressed the implementa-based sampling techniques to local leak-rate testing, and the tion and venfication status of genenc safety issues (GSis). Sup.

Use of on-hne monitoring systems to continuously monitor con-piement 1, pubbshed in December,1991, combined these vol_

tainment integnty are also evaluated.

umes into a single report and provided updated information as of September 30,1991. Supplement 2, published in December NUREG-1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE 1992, provided updated information on TMI, USl, and GSj FOR A NUCLEAR POWER PLANT. WEINSTEIN,E.D.:

issues and included status of all Other Multiplant Actions DATES.E.F. Offee for Analysis & Evaluation of Operational (MPAs). Supplement 3, pubbshed in December,1993, provided Data. Director. ADLER.M.V.; et al. Oak Ridge National Laborato-updated information as of September 30, 1993. This annual ry. March 1995. 78pp. 9504180482. 83570 007, NUREG report provides updated information on TMI, USl, GSI Tabletop exercises are held to discuss issues related to the and other MPAs as of September 30,1994 The data contained response of organizations to an emergency event. This docu-in these NUREG reports are a product of the NRC's Safety ment desenbes in task format the planning, conduct, and report-Issues Management system (SIMs) data base, which is main-ing of lessons learned for a large interagency tabletop. A tained by the Program Management Staff in the Office of Nucle-sample scenario, focus area, and discussion questions based at Reactor Regulation and by NRC regional personnel. This on a simulated accident at a commercial nuclear power plant report is to provide a comprehensrve desenption of the imple-are provided.

mentation and venfication status of TMI Action Plan Require-ments. USis, GSis and Other MPAs that have been resolved NUREG-1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MA-and involve implementation of an action or act ons by Icensees.

TERIAL SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft This soport makes the information available to other interested Report For Comment. CAMPER.LW.; SCHLUETER.J.;

parties, including the pubhc. An additional purpose of this HENDERSON.P.; et al Division of Industrial & Medical Nuclear NUREG report is to serve as a follow on to NUREG-0933. "A Safety (Post 870729). January 1995, 178pp. 9502080136.

Pnontization of Generc Safety issues," which tracks safety 82663:152.

issues up until requirements are approved for imposition at li-A Task Force originally composed of seven U.S. Nuclear censed plants or until the NRC issues a request for action by Regulatory Commission and two Agreement State program staff licensees.

developed the guidance contained in this report. The purpose of NUREG-1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER this report is to desenbe a systematic approach for effective NUCLEAR POWER PLANTS. SOFFER.L.; BURSON.S B.;

management of radiation safety programs at medcal facilities.

FERRELL.C M.; et al. Division of Systems Technology (Post This is accomphshed by emphasizing the roles of institution ex-941217). February 1995. 40pp. 9503140449. 83071:066.

ecutive management, radiation safety committee, and radiation in 1962 the U S. Atomic Energy Commission published TID.

safety offcer. Various aspects of program management are dis-14844, " Calculation of Distance Factors for Power and Test Re.

cussed and include guidance on selecting the radiation safety actors" which specified a release of fission products from the oMicer, determining adequate resources for the program, the core to the reactor containment for a postulated accident in-use of contractual services such as consuhants and service volving " substantial meltdown of the core". This " source term",

companies, the conduct of audits, the roles of authorized users the basis for the NRC's Regulatory Guides 1.3 and 1.4, has and supervised individuals, NRC's reporting and notification re-been used to determine comphance wrth the NRC's reactor site quirements, and a general description of how NRC's licensing, entena,10 CFR Part 100 and to evaluate other important plant inspection, and enforcement programs work. Appendices pro-performance requirements. Dunng the past 30 years substant:al vide detailed guidance on specifc aspects of a radiation safety additional information on fission product releases has been de-program and the glossary definos terms used throughout the veloped based on significant severe accident research. This report. The guidance contained herein does not represent new document utilizes this research by providsng more realiste esti-or proposed regulatory requirements and licensees will not be j

mates of the "cource term" release into containment, in terms inspected against any portion of it. Additionally, regulatory com-j of timing, nuchde types, quanttses and chemical form, given a phance with all applicable regulations is not assured by heens-i I

severe core. melt accident. This revised " source term" is to be ees who adopt any portion of, or apply the principles described applied to the design of future light water reactors (LWRs). Cur.

6n this report.

1 i

4 Main Citations and Abstracts NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT AL-The purpose of the workshop was to have four world experts LEGATIONS REVIEW TEAM. KOKAJKO,L E.; SKAY,0.M.;

discuss among themselves software safety issues whch are of WANG H-B.; et al. Offce of Nuclear Reactor Regulation (Post interest to the U.S. Nudear Regulatory Commission. These 941001). March 1995. 222pp. 9504180323. 83565:001, issues concern the development of software systems for use in This report provides the results of the South Texas Project Al-nuclear power plant protection systems. The workshop com-legations Review Team of the U.S. Nuclear Regulatory Commis-prised four sessions. Wednesday morning, July 22, consisted of sion. This team was formed to obtain and review allegations presentations from each of the four panel members. On from individuals represented by three attorneys who had con-Wednesday afternoon, the panel members went through a list tacted Congressional staff members. The allegers were em-of possible software development techniques and commented ployed in vanous capacities at South Texas Project Electnc on thern. The Thursday morning, July 23, session consisted of Generating Station, hcensed by Houston Lighting and Power an extended discussion among the panel members and the ob-Company, et al.; therefore, the allegations are confined to this servers from the NRC. A final session on Thursday afternoon site. The South Texas Protect Allegations Review Team re-consisted of a discussion among the NRC observers as to what viewed, referred, and dispositionod concerns related to discrimi-was teamed from the workshop.

natory issues (harassment and intimidation), falsification of records and omission of information, and vanous technical NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACK.

issues The team was able to substantiate certain technical ING IN LIGHT WATER REACTORS.

Semiannual issues of minor safety signifcance or regulatory concern at the Report, October 1993 - March 1994.

CHUNG H.M.;

South Texas Project facdity, but it did not find widespread dis-CHOPRA,0.K.; ERCK,R.A.; et al. Argonne National Laboratory.

cnminatory practices such as harassment and intimidation.

March 1995. 65pp. 9503280398. ANL-95/2,83292:228.

This report summarizes work performed by Argonne National NUREG/CP-0141: PROCEEDINGS OF THE 23hD DOE /NRC NU.

Laboratory (ANL) on fatigue and environmentally assisted crack-CLEAR AIR CLEANING CONFERENCE. Held in Buffalo,New Ynrk. July 25 28,1994. FIRST,M W. Harvard School of Pubhc ing (EAC) in light water reactors during the six months from Oc-tober 1993 to March 1994. Topics that have been investigated Health, Boston, MA. February 1995. 800pp. 9503140367, include fatigue of low-alloy steel used in piping, steam genera-CONF.940736. 83072:114.

This report contains the papers presented at the 23rd DOE /

tors, and reactor pressure vessels; EAC of wrought and cast austenste stainless steels (SSs); and radiation-induced segrega-NRC Nuclear Air Cleaning Conference and the associated dis.

tion and irradiaton-assisted stress corrosion cracking (IASCC) cubsons Major topes are: (1) nuclear air cleaning codes, (2) of Type 304 SS after accumulation of high fluence. Fatigue nuclear waste, (3) filters and filtration, (4) effluent stack monitor.

tests have been conducted on A302-Gr B low-alloy steel to ing, (5) gas processing, (6) adsorption, (7) air treatment sys.

venty whether the current predctions of modest decreases of tems, (8) source turms and accident analysis, and (9) fuel re.

fatigue life in simulated pressurized water reactor water are processing.

NUREG/CP-0143: PROCEEDINGS OF THE THIRD INTERNA.

9 "" 9

• *'

• 9 TIONAL WORKSHOP ON THE IMPLEMENTATION OF ALARA obtained on fracture-mechanics specimens of austenitic SSs to 1

AT NUCLEAR POWER PLANTS. Held At Hauppauge, Long investigate threshold stress intensity factors for EAC in high-Island, New York, KHAN,T.A. Brookhaven National Laboratory punty oxygenated water at 289 degrees C. Microchemical March 1995. 822pp. 9503270314. BNL-NUREG 52440.

changes in high-and commercial-punty Type 304 SS specimens om cpoWa@ abse Ws and a conWWa@ sMam T is e contains the papers presented and the discus-aW Wg waW mans e mum W @

sions that took place at the Third international Workshop on electron spectroscopy and scanning electron meroscopy to de-ALARA Implementation at Nuclear Power Plants, held in Haup-termine whether trace impurity elements may contribute to pauge, Long Island, New York from May 8-11, 1994. The work-of soluhamaled maWals.

shop brought together scientists, engineers, health physicists, regulators, rnanagers and others who are involved with occupa-NUREG/CR-5462: AGING STUDY OF BOILING WATER REAC-tional dose control and ALARA issues. One-hundred and seven-TOR HIGH PRESSURE INJECTION SYSTEMS. CONLEY,D.A.;

ty five persons from eleven countnes attended the workshop.

EDSON J.L.: FINEMAN C.F. Idaho National Engineering Labora-The countnes represented were: Canada, Finland, France, Ger',

tory. March 1995. 114pp. 9504120089. INEL-94/0090.

many, Japar), Korea, Mexico, the Netherlands, Spain, Sweden 83474:051.

the United Kingdom and the United States. The workshop was The purpose of high pressure injection systems is to maintain organized into twelve sessions and three panel discussions.

an adequate coolant level in reactor pressure vessels, so that The topes were as follows: Session 1, Controlling Radiation the fuel cladding temperature does not exceed 1,200 degrees C Fields; Sesson 2 Panet Discussion on Recent Recommenda-(2,200 degrees F), and to permit plant shutdown during a variety

{

tions on Dose Limitation; Session 3, Presentations and Panel of design basis loss-of-coolant acerjents. This report presents 1

Discussion on ALARA in New Reactors; Session 4, Pathways t the results of a study on aging performed for high pressure in-ALARA; Sesson 5, Panel Discussion on Economics versus Ex-jection systems of boiling water reactor plants in the United cellence; Session 6. Short Presentations on ALARA implemem States. The purpose of the study was to identify and evaluate taton; Session 7A, PWR and CANDU Presentations; Session the effects of aging and the effectrveness of testing and mainte-78, BWR and Gas-Cooled Presentations 1; Session BA, PWR nance in detecting and mitigating aging degradation. Guidelines and CANDU Presentations; Session BB, BWR and Gas-Cooled from the United States Nuclear Regulatory Commission's Nucle-Presentatons, Sesson 9 Decommissioning of Nucioar Power ar Plant Aging Research Program were used in performing the Plants, Session 10 Decontaminaton of Nuclear Power Plants' aging study. Review and analysis of the failures reported in da-and Session it, Robotes and Remote Handling. The workshop tabases such as Nuclear Power Expenence, Licensee Event was sponsored jointty by the U.S. Nuclear Regulatory Commis-Reports, and the Nuclear Plant Reliability Data System, along sion and the Brookhaven Natonal Laboratory s ALARA Center.

with plant-specife maentenance records databases, are included NUREG/CP-0145: WORKSHOP ON DEVELOPING SAFE in this report to provide the informaton required to identify aging SOFTWARE. Held At Hotel Del Coronado San Diego.CA, July 22-stressors, failure modes, and failure causes. Several probabilis-23,1992. LAWRENCE J.D. Lawrence Livermore Natonal Labo-te nsk assessments were reviewed to identify nsk-significant ratory. November 1994. 30pp. 9502080047, 82644:318.

components in high pressure injection systems. Testing, mainte-1 The Workshop on Developing Safe Software was held Juty nance, specifc safety issues, and codes and standards are also 22-23, 1992, at the Hotel del Coronado, San Diego, Calitornia.

discussed.

)

i

1 Main Citations and Abstracts 5

NUREG/CR-5591 V03: HEAVY-SECTION STEEL IRRADIATION NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR PROGRAM. Progress Report For October 1991 September HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-1992. CORWIN,W.R. Oak Rioge National Laboratory. February PHIRE) VERSION 5.0.Venfication And Validation (V&V) Manual.

1995.60pp.9503140442. ORNL/TM-11568. 83071:007.

JONES.J.L.; CALLEY M.B.; CAPPS,E.L.; et al. Idaho National The pnmary goal of the Heavy-Section Steel irradiation Pro-Engineenng Laboratory. March 1995. 141pp. 9503270310.

gram is to provide a thorough, quantitative assessment of the INEL-94/0039. 83270:133.

effects of neutron irradiation on the material behavior, and in A venfication and validation (V&V) process has been per-particular the fracture toughness properties, of typical pressure formed for the System Analysis Programs for Hands-on Integrat-vessel steels as they relate to hght-water reactor pressure.

ed Reliabihty Evaluation (SAPHIRE) Version 5.0. SAPHIRE is a vessel integnty. Effects of specimen size, matenal chemistry, set of four computer programs that the Nuclear Regulatory product form and microstructure, irradiation fluence, flax, tem.

Commission has developed for the performance of probabilistic perature and spectrum, and post-irradiation annealing are being risk assessments. These programs allow an ana'vst to perform examined on a wide rt.nge of fracture properties. The HSSI Pro-many of the functions necessary to create, quantify, and evalu-gram is arranged into 10 tasks: (1) program management, (2) ate the risk associated with a facility or process being analyzed.

K(Ic) curve shift in high-copper welds, (3) K(la) curve shift in The programs included in this set are Integrated Rehabihty and high-copper welds, (4) irradiation effects on cladding, (5) K(ic)

Risk Analysis System (IRRAS), System Analysis and Risk As-and K(la) curve shifts in low upper-shelf welds, (6) irrad.ation ef-sessment (SARA), Models and Results Database (MAR-D), and fects in a commercial low upper-shelf weld. (7) microstructural Fault Tree / Event Tree / Piping and Instrumentation Diagram analysis of irradiaton effects, (8) in-service aged matenal eval-(FEP) graphical editor. The intent of this program is V&V of suc-untions, (9) correlat on monitor matenals, and (10) special tech-cessive versions of SAPHIRE. The SAPHIRE 5.0 VaV plan is necal assistance This report pro 9 des an overview of the activi-asM on me ME 4.0 V&V plan with revisions to incorpo-ties within each of these tasks from October 1991 to September rate lessons leamed from the previous effort. The SAPHIRE 5.0 1992' vital and nonvital test procedures are based on the test proce-dures from SAPHIRE 4.0 with revisions to include the new SA-NUREG/CR-5591 V04 N2: HEAVY-SECTION STEEL IRRADIA-PHIRE 5.0. The majonty of the results from the testing was ac-TION PROGRAM Semiannual Progress Report For Apni-Sep-ceptable; however, some discrepancies between expected code tember 1993. CORWIN.W.R. Oak Ridge National Laboratory.

operation and actual code operation were identified. Modifica-tons mat han Men map 2 SAM are Mede March 1995. 48pp. 9504180477. ORNL/TM 11568. 83569.316.

The goal of the Heavy Section Steel irradiation Program is to NUREG/CR-6134: UNCERTAINTY AND SENSITIVITY ANALYSIS provide a thorough, quantitative assessment of effects of neu-OF CHRONIC EXPOSURE RESULTS WITH THE MACCS RE-tron irradiation on matonal behavior, and in particular the frac-ACTOR ACCIDENT CONSEQUENCE MODEL. HELTON,J.C. Ar-ture toughness properties, of typical pressure vessel steels as izona State Univ., Tempe, AZ. JOHNSON J.D.; ROLLSTIN J.A.:

they relate to light-water reactor pressure-vessel integnty. Ef-et al. GRAM, Inc. January 1995. 96pp. 9502080054. SAND 93-fects of specimen size, matenal chemistry, product form and mi-2370. 82644:222.

crostructure, irradiation fluence, flux, temperature and spectrum, Uncertainty and sensitivity analysis techniques based on Latin and post-irradiation annealing are being examined on a wide hypercube sampling, partial correlation analysis and stepwise range of fracture properties. The HSSI Program is arranged into regression analysis are used in an investigation with the 14 tasks: (1) program management, (2) fracture toughness MACCS model of the chronic exposure pathways associated (K(Ic)) curve shift in high-copper welds, (3) crack-arrest tough.

with a severe accident at a nuclear power station. The primary ness (K(la)) curve shift in high-copper welds, (4) irradiation of.

purpose of this study is to provide guidance on the vanables to fects on cladding. (5) K(Ic) and K(la) curve shifts in low upper.

be considered in future review work to reduce the uncertainty in shelf welds, (6) annealing effects in low upper-shelf welds, (7) the important vanables used in the calculation of reactor acci-irradiation effects in a commercial low upper shelf weld, (8) m,.

dent consequences. The effects of 75 imprecisely known input crostructural analysis of irradiation effects, (9) in-service aged vanables on the following reactor accident consequences are matenal evaluations, (10) correlation monitor matenals (11) studied: crop growing season dose, crop long-term dose, water special technical assistance, (12) JPDR steel examinaton, (13) ingestson dose, milk growing season dose, long-term ground-technical assistance for JCCCNRS Working Groups 3 and 12, shine dose, long term inhalation dose, total food patt. ways and (14) additonal requirements for matenals. This report pro-dose, total ingeston pathways dose, total long-term pathways vides an overview of the ectivities within each of these tasks dose, total latent cancer fatakties, area dependent cost, crop from Apol to September 1993.

disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation popula-NUREG/CR 5927 V02: EVALUATION OF A PERFORMANCE AS.

tion, crop disposal area and milk disposal area.

SESSMENT METHODOLOGY FOR LOW-LEVEL RADIOAC-NUREG/CR $135: UNCERTAINTY AND SENSITIVITY ANALYSIS TiVE WASTE DISPOSAL FACILITIES Vahdation Needs.

OF EARLi EXPOSURE RESULTS WITH THE MACCS REAC-KOZAK.M W.; OLAGUE N E. Sandia National Laboratones. Feb-TOR ACCIDENT CONSEQUENCE MODEL. HELTON,J.C. Arizo-ruary 1995 43pp. 9503170314. SAND 912801,83151:304.

na State Univ, Tempe, AZ. JOHNSON,J.D. GRAM, Inc.

In this report, concepts on how vahdation fits into the scheme MCKAY M D.; et al. Los Alamos National Laboratory. January of developing confidence in performance assessments are intro-1095. *47pp. 9502080062. SAND 93-2371. 82660:063.

duced. A general framework for validation and confidence build-Uncertainty and Sensitivity analysis techniques based on eng in regulatory decision making is provided. It is found that tra-Latin hypercube samphng, partial correlation analysis and step-ditional vahdation Fludies have a very hmited role in developing wise regresson analysis are used in an investigation with the site-specific Confidence in performance assessments. Indeed.

MACCS model of the earfy hearth effects associated with a vahdation studies are shown to have a role only in the context severe accident at a nuclear power station. The pnmary pur-that their results can narrow the scope of initial investigations pose of this study is to provide guidance on the vanables to be that should be considered in a performance assessment. In ad-considered in future review work to reduce the uncertainty in i

dition, validation needs for performance assessment of low-level the important variables used in the calculation of reactor acci-waste disposal facilities are discussed, and potential approach-dont consequences. The effects of 34 imprecisely known input es to address those needs are suggested. These areas of topi-vanables on the following reactor accident consequences are cal research aie ranked in order of importance based on rel-studied: number of early fatalities, number of cases of prodro-1 evance to a performance assessment and hkelihood of success mal vomiting, population dose within 10 mi of the reactor, popu-

6 Main Citations and Abstracts lation dose within 1000 mi of the reactor, individual early fatality were also estimated. The final product of the analysis was the probability within 1 mi of the reactor, and maximum early fatality integration of the accident frequencies with the consequences distance.

of the accidents to form an expression for aggregate nsk.

NUREG/CR-6136: UNCERTAINTY AND SENSITIVITY ANALYSIS NUREG/CR-6143 V06 P2: EVALUATION OF POTENTIAL 4

OF FOOD PATHWAY RESULTS WITH THE MACCS REACTOR SEVERE ACCIDENTS DURING LOW POWER AND SHUT-ACCIDENT CONSEQUENCE MODEL. HELTON,J.C. Arizona DOWN OPERATIONS AT GRAND GULF, UNIT 1. Evaluation Of State Univ., Tempe, AZ. JOHNSON J.D.; ROLLSTIN.J.A.; et al.

Severe Accident Risks For Plant Operational State 5 Dunng A GR, nc. January 1995. 94pp. 9502080088. SAND 93-2372.

Refueling Outage. Supporting MELCOR Calculations.

KMETYK,LN.; BROWN,T.D. Sandia National. Laboratories.

Uncertain' ty and Sensitivity analysis techniques based on March 1995. 431pp. 9504100148. SAND 93-2440. 83418:001.

Latin hypercube sampling, partial correlation analysis and step-The document contains the deterministic code calculations wise regression analysis are used in an investigation with the performed with the MELCOR Code that were used to support MACCS model of the food pathways associated with a severe the development and quantification of the PRA models used in accident at a nuclear power station. The pnmary purpose of this the analysis of intemally initiated events for Grand Gulf, Unit 1, study is to provide guidance on the vanables to be considered as it operates in the Low Power and Shutdown Plant Operation-in future review work to reduce the uncertainty in the important at State 5 during a refueling outage. The background for the vanables used in the t'alculation of reactor accident conse-work documented in this report is summarized including how de-quences. The effects of b7 imprecisely known input vanables on terministic codes are used in PRAs, why Qe MELCOR code is the following reactor accident consequences are studied crop used, what the capabilities and features of MELCOR are and growing season dose, crop long-term dose, milk growing how the code has been used by others in the past. Onef de-season dow total food pathways dose, total ingestion path-scnptions of the Grand Gutf plant and its configuration during ways d w Mtal long-term pathways dose, area dependent LP&S operation and of the MELCOR input model developed for cost, crx ausal cost, milk disposal cost, condemnation area' the Grand Gulf plant in its LP&S configuration are given. The crop dr area and milk disposal area.

results of MELCOR ana!yses of vanous accident sequences for NUREG/Cu t41: HANDBOOK OF METHODS FOR RISK.

the plant operating state (POS) 5 configuration dunng refuelir:g BASED ANALYSES OF TECHNICAL SPECIFICATIONS.

(approximately Cold Shutdown as defined by Grand Gulf Tech-SAM AN T A,P.K.;

KIM,1.S Brookhaven National Laboratory.

nical Specifications) are presented for accidents initiated at sev-MANKAMO,T.; et al. Avaplan Oy (Finland). December 1994.

eral different times after scram and shutdown including short-190pp.9503010187. BNL-NUREG-52398. 82887.001, ened thermal hydraulic and core damage calculations done in Technical Specifications (TS) requirements for nuclear power support of the Level 1 analysis and full plant analyses including plants define the Limiting Conditions for Operation and Surveil.

containment response and source terms supporting the Level 2 lance Requirements to assure safety dunng operation. In gener.

analysis al, these requirements are based on deterministic analysis and NUREG/CR 6192: AGING AND SERVICE WEAR OF SPRING-engineenng judgments. Expenences with plant operation ind,.

LOADED PRESSURE RELIEF VALVES USED IN SAFETY-RE.

cate that some elements of the requirements are unnecessanly LATED SYSTEMS AT NUCLEAR POWER PLANTS.

restrictive, while a few may not be conducive to safety. Improv.

STAUNTON,R.H.; COX,D.F. Oak Ridge National Laboratory, ing these requirements involves many considerations and is fa.

March 1995. 86pp. 9504180472. ORNL-6791. 83568:226.

cilitated by the availability of plant specific Probabilistic Safety Spnng-loaded pressure relief valves (PRVS) are used in some Assessments and development of related methods for analyses.

safety-related apphcations at nuclear power plants. In general, This handbook summarizes the risk-and reliability-based meth, they are used in systems where, during accidents, pressures ods to improve TS requirements. The scope of the handbook may nse to levels where pressure safety relief is required for includes reliability-and risk-based methods for evaluating al.

protection of personnel, system piping, and components. This Iowed outage times, scheduled or preventive maintenances, report documents a study of PRV aging and considers the se-action statements requinng shutdown where shutdown nsk may venty and causes of service wear and how it is discovered and be substantial, surveillance test intervals, and management of corrected in various systems, valve sizes, etc. Provided in this plant configurations resulting from outages of systems, or com-report are results of the examination of the recorded failures ponents. For each topic, the handbook summanzes analytic and identification of trends and relationships / correlations in the methods with data needs, outlines the insights to be gained, failures when all failure-related parameters are considered.

lists additional references, and gives examples of evaluations.

Components that comprise a typical PRV, how those compo-1 NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL nents fail, when they fail, and the current testing frequencies SEVERE ACCIDENTS DUR!NG LOW POWER AND SHUT

• and methods are also presented in detail.

DOWN OPERATIONS AT GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant Operational State 5 Dunng A NUREG/CR-6214: PRODUCTION AND TESTING OF THE VITA-Refueling Outage Mala Report And Appendices. BROWN T.D.;

MIN B6 FINE GROUP AND THE BUGLE-93 BROAD GROUP KMETYK,L.N.; WHITEHEAD.D ; et al. Sandia National Laborato-NEUTRON / PHOTON CROSS-SECTION LIBRARIES DERIVED ries. March 1995. 407pp. 9504050320. SAND 93-2440.

FROM ENDF/B VI NUCLEAR DATA. INGERSOLL.D.T.;

83378.027.

WHITE J E.; WRIGHT,R O.; et al. Oak Ridge National Laborato-The analysis documented in this volume of the report is the ry. January 1995.172pp. 9502080096. ORNL-6795. 82650:146.

Level 2/3 analysis of the traditional internal events. Plant A new multigroup cross-section library based on ENDF/B-VI damage states, which define the configuration of the plant and data has been produced and tested for light water reactor Its systems at the onset of core damage for the accident sce-shielding and reactor pressure vessel dosimetry applications.

nanos developed in the Level 1 analysis, were used to define The broad-group library, which is designated BUGLE-93, is in-l the interface between the Level 1 and Level 2/3 analyses. In tended to replace the aging BUGLE 80 and SAILOR libraries.

the Level 2/3 analysis, the possible progressions of the acci-The processing methodology is consistent with ANSI /ANS dent following the onset of core damage were delineated and 61.2, since the ENDF data were first processed into a fine-j the amount of radioactive matenal released to the environment group, pseudo-problem-independent format and then collapsed was estimated. Based on the amount of radioactive matenal re-into the final broad-group format. The fine-group library, which is leased to the environment, health effects to the general public designated VITAMIN-B6, contains 120 nuclides. The BUGLE 93 were estimated. In addition to the offsite consequences, a scop.

47-neutron-group / 20-gamma-r ay-group library contains the ing anaiysis of the potential doses and dose rates within the site same 120 nuclides processed as infinitely dilute and collapsed j

Main Citations and Abstracts 7

using a weighting spectrum typical of a concrete shield. Addi-dependent check on analyses submitted by licensees. CASKS tionally, BUGLE-93 contains 105 nuchdes processed with reso-is based on microcomputers compatible with the IBM PC family rtnce self. shielding and weighted using spectra specific to of computers. The system is composed of a series of menus, BWR and PWR material compositions and reactor models. Sev.

input programs, cask analysis programs, and output display pro-eral dosimetry response functions and kerma factors for all 120 grams. All data is entered through fill-in-the-blank input screens nuclides are also included with the library. An extensive integral that contain descriptrve data requests.

data testing effort was performed to qualify the new hbrary. In general, results using the new data show significant improve.

NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT COdSE-ments relative to earlier ENDF data.

QUENCE UNCERTAINTY ANALYSIS. Despersion And Deposi-NUREG/CR-6220: AN ASSESSMENT OF FIRE VULNERABILITY tion Uncertainty Assessment. Main Report. HARPER,F.T. Sandia National Laboratories. GOOSSENS.L.H.J.; COOKE,R.M.; et al.

FOR AGED ELECTRICAL RELAYS. VIGIL.R.A. Science & Engi-neering Associates, Inc. NOWLEN S.P. Sandia National Labora-Netherlands, Govt. of. January 1995.106pp. 9503150202. EUR 15855EN. 83108:230.

tories. March 1995. 38pp. 9504100121, SAND 94-0769.

834 4 069.

The development of two new probabilistic accident conse-Thes report detasis testing to assess the impact of aging on quence codes, MACCS and COSYMA, was completed in 1990.

These codes estimate the consequences from the accidental the fire vulnerability of Agastat and General Electnc relays. Both aged and unaged relays were tested. Agod relays were subject.

releases of radiological material from hypothesized accidents at j

ed to operational cycling under rated load and thermally aged nuclear installations. In 1991, the U.S. Nuclear Regulatory Com-for sixty days. All relays were exposed to one of three different mission and the Commission of the European Communities fire temperature profiles in the Severe Combined Environments began co-sponsonng a joint uncertainty analysis of the two Test Chamber located at Sandia National Laboratones. The codes. The ultimate objective of this joint effort was to system-l ability to operate property in the given fire environment was atically develop creu.ble and traceable uncertainty distributions I

monitored. Results for the aged and unaged relays were exam-for the respective code input variables. Because of the magni-tude and expense required to complete a full-scale conse-intd to determine the impact of aging on the relays' ability to sustain operation under the test conditions. Overall results and'-

quence uncertainty analysis, a trial study was performed to ceted that the aged relays' performance was not significantly evaluate the feasibihty of such a joint study by initially hmating I

different from that of the unaged relays.

efforts to the dispersion and deposition code input variables. A formal export judgment elicitation and evaluation process was NUREG/CR4240: APPLICATION OF BOUNDING SPECTRA TO identified as the best technology available for developing a li-SEISMIC DESIGN OF PIPING BASED ON THE PERFORM.

brary of uncertainty distnbutions for these consequence param-ANCE OF ABOVE GROUND PIPING IN POWER PLANTS SUB, eters. This report focuses on the methods used in and results JECTED TO STRONG MOTION EARTHOUAKES.

of this trial study.

STEVENSON,J D. Stevenson & Associates.

• Oak Ridge Na-tional Laboratory. February 1995. 104pp. 9503140368.

NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSE-ORNLSUB94SD4271. 83069.312.

QUENCE UNCERTAINTY ANALYSIS. Dispersion And Deposi-This report extends the potential apphcation of Bounding tron Uncertainty Assessment. Appendices A And B.

Spectra evaluation procedures, developed as part of the A-46 HARPER,F.T. Sandia National Laboratories. GDOSSENS,L.H.J.I Unresolved Safety issue applicable to seismic venfication of in.

COOKE R.M.; et al. Netherlands, Govt. of. January 1995.400pp.

}

situ electncal and mechanical equipment, to in-situ safety relat.

9503150209. EUR 15855EN. 83109:001.

ed piping in nuclear power plants. The report presents a sum.

See NUREG/CR-6244,V01 abstract.

mary of earthquake experience data which define the behavior of typical U.S, power plant piping subject to strong motion NUREG/CR-6244 V03: PROBABILISTIC ACCIDENT CONSE-eirthquakes. The report defines those piping system caveats OUENCE UNCERTAINTY ANALYSIS. Dispersion and Deposi-which would assure the seismic adequacy of the piping systems tion Uncertainty Assessment. Appendices C.D.E.F,G,H.

which meet those caveats and whose seismic demand are HARPER.F.T. Sandia National Laboratories. GOOSSENS,LH.J.;

within the bounding spectra input. Based on the observed be.

COOKE.R.M.; et al. Netherlands, Govt. of. January 1995.98pp.

havior of piping in strong motion earthquakes, the report distin-9503140347. EUR 15855EN. 83069.214.

guishes between the capabihtles of the piping system to carry See NUREG/CR-6244,V01 abstract.

seismic loads as a function of the type of connection (i.e.'

NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING threaded vs welded). This report also discusses in some detail the basic causes and mechanisms for earthquake damages and ATOMIC ENERGY OF CANADA LTD CODES. JEDD,J.L.;

failures to power plant piping systems' SHUMWAY,R.W. Idaho Nationd Engineering Laboratory.

EBERT,D.D. Reactor & Plant Symms Branch (Post 941217).

NUREG/CR-6242: CASKS (COMPUTER ANALYSIS OF STOR-January 1995. 88pp. 9503270305. INEL-95/0070. 83268:267.

AGE CASKS): A MICROCOMPUTER BASED ANALYSIS A limited number of transient scenarios were calculated using SYSTEM FOR STORAGE CASK DESIGN REVIEW. User's a computer code suite and input modehng provided by the Minual To Version ib (Including Program Reference).

Atomic Energy of Canada Limited (AECL) for the CANDU 3 CHEN,T.F.; GERHARD,M.A.; TRUMMER,D.J.; et al. Lawrence design. Emphasis was placed on a large-break loss-of-coolant Livermore National Laboratory. February 1995. 200pp.

accident with delays in actuation of the two independent shut-9503140405. UCRL-ID-117418. 83070:058.

down systems (shutdown rods and liquid poison injection). Al-CASKS (Computer Analysis of Storage casks) is a micro-though an extremely unlikely scenario, it was studied because computer. based system of computer programs and databases of the potential consequences that would result from a positive developed at the Lawrence Livermore National Laboratory void coefficient of reactivity. Results indicate that a few seconds (LLNL) for evaluating safety analysis reports on spent-fuel stor-delay in shutdown would result in Quickly reaching fuel or clad-tge casks. The bulk of the complete program and this user's ding melting temperatures before the emergency core cooling m"nual are based upon the SCANS (Shipping Cask ANa!ysis system would be activated. Only small changes in the timing System) program previously developed at LLNL A number of and consequences of the scenano result when several param-Enhancements and improverrents were added to the original eters, of potential importance to the progression of the acci.

SCANS program to meet requirements unique to storage casks.

dent, are varied. Five calculations were also performed for loss-CASKS is an easy-to-use system that calculates global re-of off-site-power scenarios. These calculations assume that the i

sponse of storage casks to impact loads, pressure loads and plant failed to enter the island mode, i.e., power to the main th2rmal conditions. This provides reviewers with a tool for an in-Coolant pumps was not restored using on-site power generation.

8 Main Citations and Abstracts NUREG/CR-6259: CONSTRAINT EFFECTS ON FRACTURE INi-aging but subsequent studies did not support the earber results.

TIATION LOADS iN HSST WIDE PLATE TESTS. DODDS.R H.

The present computational study presented in Volume 1 of this lihnois, Unrv. of, Urbana, IL. DODDS.R.H. Oak Ridge National report investigates several forms of this parameter, how they Laboratory. December 1994. 42pp.

9502080251, UI-are derrved and the validity of these parameters for small and LUENG942009. 82660:304.

large amounts of crack growth. It is concluded that neither J nor Dunng the penod 1984 1987, researchers of the Heavy Sec-J(M) (nor any single parameter) can satisfactorily capture the tion Steel Technology program at the Oak Ridge National Labo-full range of near-hp fracture states. A discussion on the range ratory performed a unique series of fracture mechanics tests of validity of J(M) is given in Volume 2.

using exceptionally large, SE(T) specimens (a/W-0.2) fabncat-ed from a reactor pressure vessel matenal, A533B Class 1 NUREG/CR-6264 V02: VAllDITY LIMITS IN J-RESISTANCE

.teet This study re-examines fracture initiation loads in the CURVE DETERMINATION.A Computational Approach To Duc-wide-plate tests using two constraint assessment methodologies tile Crack Growth Under Large-Scale Yielding Condrtions.

developed over the past tive years: the J-O toughness locus ap.

SHlH,C.F.;

XIA L Brown Univ.,

Providence, RI.

proach and the toughness scaling approach based on a local HUTCHINSON.J.W.; et al. Harvard Univ., Cambridge, MA. Feb-fadure entenon for cleavage. Both approaches demonstrate a ruary 1995. 50pp. 9503140430. BMI-2181. 83070:317.

significant loss of constraint in the elastic-plastic fields ahead of in this report, Volume 2, Mode I crack initiation and growth the crack in the wide-plate specimens caused by the inherent under plane strain conditions in tough metals are computed negative T-stress of the shallow notch SE(T) configuration.

using an elastic / plastic continuum model which accounts for Moreover, the 25mm wide machined notch required for speci-void growth and coalescence ahead of the crack tip. The mate-men fabncation is shown to further reduce constraint by intro-nal parameters include the stress-strain properties, along with ducing a traction free surface very near the crack tip Both of the parameters charactenzing the spacing and volume fraction these factors combined to reduce near.tip stresses by 10%

of voids in matenal elements lying in the plane of the crack. For below those of the small-scale yielding. SSY (T=O), fields. This a given set of these parameters and a specific specimen, or reduction places fracture results for the wide-plate specimens component, subject to a specific loading, relationships among within the J O toughness locus defined by fracture toughness load, load-line displacement and crack advance can be comput-tests on the A5338 matenal and within the constraint corrected ed with no restnctions on the extent of plastic deformation.

J(c) values defined by the toughness scaling methodology.

Similarly, there is no limit on crack advance, except that it must take place on the symmetry plane ahead of the initial crack.

NUREG/CR-6260: APPLICATION OF NUREG/CR-5999 INTERIM Sultably defined measures of crack tip loading intensity, such as FATIGUE CURVES TO SELECTED NUCLEAR POWER PLANT those based on the J-integral, can also be computed, thereby COMPONENTS WARE A.G; MORTON,D.K.; tw r2EL,M E.

directly generating crack growth resistance curves. In this Idaho National Engineenng Laboratory. March 1995. 200pp.

report, the model is applied to five specimen geometries which 9503280383 INEL-95/0045. 83293.001.

Recent test data indicate that the effects of the light water are known to give nse to significantly different crack tip con-reactor (LWR) environment could significantly reduce the fatigue straints and crack growth resistance behaviors. Computed re-suits are compared with sets of expenmental data for two tough resistance of matena!s used in the reactor coolant pressure boundary components of operating nuclear power plants. Ar.

steels for four of the specimen types. Details of the load, dis-gonne National Laboratory has developed intenm fatigue curves placement and crack growth histories are accurately repro.

based on test data simulating LWR conditions, and published duced, even when extensive crack growth takes place under them in NUREG/CR-5999. In order to assess the significance of conditions of fully plastic yielding.

these interim fatigue Curves, fatigue evaluations of a sample of NUREG/CR-6266: ANALYSIS OF BORON DILUTION IN A FOUR.

the components in tr.e reactor cociant pressure boundary of LOOP PWR. SUN,J.G ; SHA W.T. Argonne National Laboratory.

LWRs were performed The sample consists of components March 1995. 80pp. 9504180468. ANL-94/35. 83569-235.

from facilities designed by each of the four U.S nuclear steam Thermal mixing and boron dilution in a pressunzed water re-supply system vendors For each facility, six locations wer actor were analyzed with COMMIX codes. The reactor system studied, including two locations on the reactor pressure vessel.

In addition, there are older vintage plants where components of were analyzed. In the first scenano, the plant is in cold shut-the reactor coolant pressure boundary were designed to coJes down and the reactor coolant system has just been filled after that did not require an explicit fatigue analysis of the compo-maintenance on the steam 9eneratort. To flush the air out of nents in order to assess the fatigue resistance of the older vin-i the steam generator tubes, a reactor coolant pump (RCP) is tage plants, an evaluation was also conducted on selected started, with the water in the pump suction line devoid of boron components of three of these plants. This report discusses the and at the same temperature as the coolant in the system. In insights gained from the application of the intenm fatigue curves the second scenario, the plant is at hot standby and the reactor to components of seven operating nuclear power plants.

coolant system has been heated to operating temperature after NUREQ/CR 6264 V01: VALIDITY LIMITS IN J-RESISTANCE a long outage. It is assumed that an RCP is started, with the CURVE DETERMINATION.An Assessment Of The J(M) Param-pump suction hne filled with cold unborated water, forcing a slug eter SHIH.C F.; LIU,X H. Brown Univ., Providence, RI.

• Battelis of diluted coolant down the downcomer and subsequently Memonal institute, Columbus Laboratones February 1995.

through the reactor core. The subsequent transient thermal 35pp. 9503140423 BMI-2181. 83070 282.

mixing and boron dilution that would occur in the reactor system Significant advances in elastic-plastic fracture became poss8-is simulated for these two scenanos. The reactivity insertion ble with the introduction of Race's path independent J-integral rate and the total reactivity are evaluated and a sensitivity study which has two physical meanings First, the J-integral is equiva-is performed to assess the accuracy of the numencal modeling lent to the energy release rate associated with a virtual crack of the geometry of the reactor coolant system.

advance. Secondly, J can be regarded as the strength of the stress and strain singutanty near a stationary crack tip. As a NUREG/CR-6273: BIAXIAL LOADING EFFECTS ON FRACTURE result of severa. < xpenmental studies, the J-integral is generally TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL.

accepted as a vahd parameter to charactenze a matenars re-MCAFEE,W.J.; BASS,8=R.; BRYSON J.W.; et al. Oak Ridge Na-sistance to the onset of crack growth under large scale yielding tional Laboratory. March 1995. 76pp. 9503270300. ORNL/TM-Driven by semplicity end the practical benefits that could be de-12866. 83268:191.

nved from a geometry and size-independent matonal resistance The preliminary phases of a program to develop and evaluate curve for large amounts of crack growth, J(M), a modified J pa.

fracture methodologies for the assessment of crack tip con-rameter was introduced initial results using J(M) were encour-straint effects on fracture toughness of reactor pressure vessel

Main Citations and Abstracts 9

- (RPV) steels has boen completed by the Heavy-Section Steel NUREG/CR 6285: SEVERE ACCIDENT NATURAL CIRCULATION Technology (HSST) Program. The pomary objectives of this STUDIES AT THE INEL. BAYLESS P.D.; BROWNSON D.A.;

effort were to analytscally and experimentally investigate the OOBBE,C.A.; et al. Idaho National Engineenng Laboratory. Feb-effect of biaxial loading on fracture toughness, to quantify this ruary 1995. 200pp. 9503150171. INEL 94/0016. 83108:014.

effect through use of existing stress-based, dual-parameter, Severe accident natural circulation flows have been investi-fracture-toughness correlations, or to propose and venty after-gated at the Idaho National Engineenng Laboratory to better un.

nate correlations. A cruciform beam specimen with a two-di-derstand these flows and their poterit al impacts on the progres-mensional shallow, through-thickness flaw and a special loading sion of a pressurized water reactor s> mre accident. Parameters fixture was designed and fabnceted, Tests were pedormed affecting natural circulation in the reactor vessel and hot legs using biaxial loading ratios of 0:1 (uniaxial),0.6.1, and 1:1 (equi.

were identified and ranked based on their perceived impor-biaxial). Cntical fracture toughness values were calculated for tance. Reviews of the scahng of the 1/7-scale expenments per-each test. Biarial loading of 0 6.1 resulted in a reduction in the formed by Westinghouse were undertaken. RELAPS/ MOD 3 cal lower bound fracture toughness of = 12% as compared to the culations of two of the expenments showed generally good uniaxial tests. The biaxial loading of 1:1 yielded two subsets of agreement between the calculated and observed behavior, toughness values; one agreed well with the uniaxial data while Analyses of hydrogen behavior in the reactor vessel showed one was reduced by = 43% when compared to the uniaxial that hydrogen stratification is not likely to occur, and that an ini-data. The results were evaluated using the J O theory and the tsally stratified layer of hydrogen would quickly mix with a recir-Dodds Anderson (D-A) micromechanical scaling model. The D-A culating steam flow. An analysis of the upper plenum behavior N

Na, W 2 mas mW M w model predicted no biarial effect while the J-O method gave in-tamperatures could have been significantly higher than the tem-conclusive results. When applied to the 1:1 biaxial data, these Puratures seen by the control rod drive lead screws, supporting constraint methodologies failed to predict the observed reduc-the premise that a strong natural circulation flow was likely tion in fracture-toughness obtained in one expenment. A strain-present during the accident. SCDAP/RELAPS calculations of a based constraint methodology that considers the relationship commercial pressunzed water reactor severe accident without between applied biarial load, the plastic zone size at the flaw operator actions showed that the natural circulation flows en-tip, and fracture toughness was formulated and applied suc-hance the likelihood of ex-vessel piping failures long before fail-cessfully to the data. Evaluation of this dual. parameter strain-ute of the reactor vessel lower head.

based model led to the conclusion that it has the capabihty of representing fracture behavior of RPV steels in the transition NUREG/CR 6291 V01: NUCLEAR P!. ANT ANALYZER. installation region, including the effects of out of-plane loading on fracture-Manual. SNIDER D.M.; WAGNER.K.L,; GRUSH.W.H.; et al.

toughness. This report is designated as HSST Report No.150.

Idaho National Engineenng Laboratory. January 1995. 28pp.

9502150283. INEL-94/0123. 82730:001.

NUREG/CR-6276: A COMPILATION OF CURRENT REGULA-This report contains the installation instructions for the Nucle-TIONS, STANDARDS, AND GUIDELINES IN REMOTE AFTER-at Plant Analyzer (NPA) System. The NPA System consists of LOADING BRACHYTHERAPY, TORTORELU,J P.; SIMION.G.P.;

the Computer Visu# System (CVS) program, the NPA kbranes, KOZLOWSKI,S D. EG&G Idaho, Inc. February 1995. 94pp.

the associated utikty programs. The NPA was developed at the 9503270298. EGG 2746. 83268 095.

Idaho National Engineenng Laboratory under the sponsorship of Over a dozen government and professional organizations in the U.S. Nuclear Regulatory Commission to provide a highly the United States and Europe have issued regulations and guid-flexible graphical user interface for displaying the results of ance concerning quahty management in the practice of remote these analysis codes. The NPA also provides the user with a afterloading brachytherapy. Information from the publications of Convenient means of interactively controlling the host program these organizations was collected and collated for this report.

through user-defined pop up menus. The NPA was designed to This report provides the brachytherapy bcensee access to a serve pnmanly as an anatysis tool. After a brief introduction to broad feld of quality management information in a single, topi.

the Computer Visual System and the NPA, an analyst can cally organized document.

quickly create a simple picture or set of pictures to aide in the study of a particular phenomenon. These pictures can range NUREG/CR-6284: CRITICALITY SAFETY CRITERIA FOR Lt.

from simple collections of square boxes and straight knes to CENSE REVIEW OF LOW-LEVEL WASTE FACILITIES.

complex representations of emergency response information HOPPER.C.M.; ODEGAARDEN R.H.; PARKS.C.V.; et al. Oak displays.

Ridge National Laboratory. March 1995. 45pp. 9504120093.

ORNL/T M.12845. 83474.165.

NUREG/CR-6291 V02: NUCLEAR PLANT ANALYZER. Analyzer Reference Manual. SNIDER,0.M.; WAGNER,K.L.; GRUSH,W.H.;

Th's report provides recommended safety enteria for NRC h-censed bunal facihtes. These critena have been developed with et al. Idaho National Engineenng Laboratory. January 1995.

71pp. 9502150287. INEL-94/0123. 82730:030.

accepted and consistent nuclear enticakty safety evaluation techniques. Additionally, this report provides the bases for the The Nuclear Plant Analyzer (NPA) system provides both a highly flexible graphical user interface for displaying simulation recommended safety entena by documenting the evaluation data and, where applicable, a convenient means of interactively methods and assumptions, and by reporting the results of all controlkng the host program 1;1 rough user defined pop-up single-package and array calculations. These enteria were de-menus. The NPA system was developed at the Idaho National veloped with care to assere consistency with data and practices Engineering Laboratory under the sponsorship of the U.S. Nu-provided in current standards on nuclear ent calty safety as well clear Regulatory Commission (NRC). The Computer Visual as conformity of the entena to apphcable NRC regulations. The System and the Analyzer are the pnmary components of the recommended safety entena are expressed in terms of surface' NPA system. This report contains the reference manual for the density spacing entena, thereby greatly simphfying the applica-Analyzer. It describes both the NPA libraries that constitute the tion of heense conditions for nuclear enticahty safety control.

Analyzer and o set of auxiliary programs used in conjunction This approach was used by an NRC heensee at the Bamwell with the Analyzer.

waste bunal facihty by limiting the specific controls to the fewest number of parameters consistent with good nuclear safety prac-NUREG/CR-6291 V03: NUCLEAR PLANT ANALYZER. Computer tice. The use of a surface-density cntena can ekminate the need Visual System Reference Manual. SNIDER,D.M.; WAGNER,K.L.;

for numerous bcense amendments for vanations in package GRUSH.W.H.; et al. Idaho National Engineenng Laboratory. Jan-contents and specifications uary 1995.128pp. 9502150297. INEL-94/0123. 82730:101.

10 Main Citations and Abstracts The Computer Visual S stem (CVS) Reference Manual de-suggesSons for treating structural uncertainty for submodels are scnbes that part of the Nuc! ear Plant Analyzer (NPA) system presented.

useo to create pictures (masks) This manual is intended to guide a user in creating. editing and animating masks for use in NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION the NPA. The NPA was developed at the Idaho National Engi.

AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND neenng Laboratory under the sponsorship of the U S. Nuclear CONVENTIONAL. SOFTWARE. MILLER,LA.: HAYES J.E.;

Regulatory Commission to provide a highly flexible graphical MIRSKY,S M. Science Applications Intemational Corp. (formerly user interface for displaying the results of these analysis codes.

Science Applications, Inc.). March 1995.178pp. 9504180393.

The NPA also provides the user with a convenient means of in.

SAIC-95/1028. 83566:198.

teractively controlling the host program through user-defined This report provides the project summary of the results of the pop-up menus The NPA was designed to serve pnmanly as an Expert System Venfication and Validation (V&V) activity that analysis tool After a bnef introduction to the Computer Visual was jointly funded by the U.S. Nuclear Regulatory Commission System and the NPA, an analyst can quickly create a simple and the Electnc Power Research Institute to develop guidelines picture or set of pictures to aide in the study of a particular phe.

for the V&V of expert and other systems. This is the first nomenort These pictures can range from simple collections of volume of an eight volume report. The project began with a square boxes and straight lines to complex representations of survey of conventional V&V methods that covers 153 different emergency response information displays.

techniques. Quantitative cost benefit and an effectiveness measures were developed to permit comparisons among all the NUREG/CR-6291 V04:

NUCLEAR PLANT methods for three levels of stnngency of V&V: low, medium, ANALYZER Programmer's Manual SNIDER.D M.;

and high (Classes 3 to 1, respectively). A survey was conducted WAGNER.K.L, GRUSH.W.H.; et al Idaho National Engineenng concerning V&V practices in use for expert systems, finding that Laboratory January 1995. 202pp. 9502150309. INEL 94//0123 they were not common, but that there was considerable activity 82730 229 in developing methods for knowledge bases. Selected V&V The Nuclear Plant Analyzer (NPA) system provides both a methods were applied to two existing expert systems used in highly flexible graphical user interface for displaying simulation nuclear power apphcations. Other V&V methods were investi-data and. where applicable, a convenient means of interactively gated in an empincal experiment to assess their practical utility.

controlling the host program through user-defined pop.up A method for generating validation scenanos was developed. Fi-menus. The NPA system was developed at the Idaho National nally, a set of guidelines recommending specific V&V methods Engineenng Laboratory under the sponsorship of the U.S. Nu-for 16 different system-development situations was developed.

clear Regulatory Commission (NRC) This manual is intended to serve as a programmers' guide for the NPA system. As such, it NUREG/CR-6316 V02: GUIDELINES FOR THE VERIFICATION includes technical details regarding the design and imp!ementa-AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND tion of the Computer Visuals Systems (CVS) program, the Ana-CONVENTIONAL SOFTWARE. Survey And Assessment Of Con-lyrer, data files used by CVS and the Analyzer, and a senes of ventional Software Venfication And Validation Methods.

auxiliary programs that provide important services to NPA users.

MILLER.L.A.; GROUNDWATER.E.H, HAYES,J E.: et al. Science Applications international Corp. (formerly Science Applications, NUREG/CR-6293 V02: VERIFICATIONN AND VAllDATION Inc ).

March 1995. 190pp. 9504180403 SAIC-95/1028.

GUIDELINES FOR HIGH INTEGRITY SYSTEMS Appendices A-83567.017.

D. HECHT,H.; HECHT,M ; DINSMORE.G.; et al. SoHaR, Inc.

By means of literature survey, a comprehensive set of meth.

March 1995. 54pp. 9504100137,83419 304 ods was identified for the venfication and validation of conven.

See NUREG/CR-6293,V01 abstract-tional software. The 153 methods so identified were classified NUREG/CR-6310: AN ANALYSIS OF POTASSIUM IODIDE (KI) according to their appropnateness of vanous phases of a devel-PROPHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT opment lifecycle - requirements, design, and implementation; OF A NUCLEAR ACCIDENT. BEHLING.H ; BEHLING,K. S.

the last category was subdivided into two, static testing and dy-1 Cohen & Associates, Inc. AMARASOORlYA.H ; et at. SCIEN namic testing methods. The methods were then characterized in TECH, Inc February 1995. 200pp 9503150166. 83107.157.

terms of eight rating factors, four conceming ease-of-use of the A genenc difficulty countered in ecst-benefit analyses is the methods and four conceming the methods; power to detect de-quantification of major elements that define the costs and the fects. Based on these and an Effectiveness Metnc. The Effec-benefits in commensurate units. In this study, the costs of tiveness Metnc was further refined to provide three different es-making Kt available for public use. and the avoidance of thyro,_

timates for each method, depending on three classes of needed dal health effects predicted to be realized from the availability of stnngency of V&V (determined by ratings of a system's com-that Kl (i e, the benefits). are defined in the Commensurate plexity and required integnty) Methods were then rank-ordered units of dollars _

for each of the three classes in tezms of their overall cost-bene-fits and effectiveness The applicability was then assessed of NUREG/CR 6311: EVALUATING PREDICTION UNCERTAINTY.

each method for the four identified components of knowledge-MCKAY.M D Los Alamos National Laboratory. March 1995 based and expert systems, as well as the system as a whole.

69pp. 9503270307. LA 12915 MS. 83270 274.

The probability distnbution of a model prediction is presented NUREG/CR-6316 V03: GUIDELINES FOR THE VERIFICATION as a proper basis for evaluating the uncertainty in a model pre-AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND diction that anses from uncertainty in input values Determina.

CONVENTIONAL SOFTWARE. Survey And Documentation Of tion of important model inputs and subsets of inputs is made Expert System Venfication And Validation Methodologies.

through companson of the prediction distnbution with condition-GROUNDWATER,E.H. MILLER.L.A.; MIRSKY,S.M. Science Ap-al prediction probability distnbutions Replicated Latin hypercube plications International Corp. (formerly Science Applications, sampling and vanance rattos are used in estimation of the distn-Inc ).

March 1995. 112pp. 9504180421. SAIC-95/1028.

butions and in construction of importance indicators. The as-83567.204.

sumption of a linear relation t>etween model output and inputs is This report is the third volume in a senes of reports descnb-not necessary for the indicators to be effective. A sequential ing the results of the Expert System Venfication and Validation methodology which includes an independent validation step is (V&V) project that is jointly funded by the U S. Nuclear Regula-applied in two analysis applications to select subsets of input tory Commission and the Electnc Power Research Institute to vanables which are the dominant causes of uncertainty in the develop guidelines for the V&V of expert and other systems.

model predictions. Companson with results from methods which The purpose of this activity was to survey and document tech-awume lineanty shows how those methods may fail Finally, niques presently in use for expert systems V&V. Via extensive

1 Main Citations and Abstracts 11 telephone contacts, 6ste visits, and through bibliographic NUREG/CR-6316 V06: GUIDELINES FOR THE VERIFICATION searches a wide samphng of expert system V&V was accom-AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND plished. The major finding was that VaV of expert systems is CONVENTIONAL SOFTWARE.Validaten Scenarios.

not nearly as established or prevalent as V&V of conventional MIRSKY,S.M.; HAYES.J.E.; MILLER LA. Science Applicatons software systems. There were few examples of V&V in the early international Corp. (former1y Science Applications, Inc.). March stage of development. However, there is a very active research 1995. 57pp. 9504180445. SAIC-95/1028. 83567:312.

l area concerning the development of methods to assess the This report is the sixth volume in a series of reports describ-

)

knowledge bases of expert and knowledge-based systems, ing the results of the Expert System Verification and Validation NUREG/CR-6316 V04: GUIDELINES FOR THE VERIFICATION (V&V) project that is jointly funded by the U.S. Nuclear Regula.

tory Commission and the Electric Power Research Institute to AND VALIDATION EXPERT SYSTEM SOFTWARE AND CON-develop guidelines for the V&V of expert and other systems.

VENTIONAL SOFTWARE Evaluaton Of Knowledge Base Certi-This activity was concemed with the development of a method-ficaton Methods. MILLER,LA.; HAYES.J.E.; M:RSKY,S.M. Sci-ology for selecting " validation scenarios." These are defined as ence Apphcations Intemational Corp. (formerly Science Applica-

" realistic dynamic tests of software which covers only the in-tions, Inc.). March 1995.136pp. 9504180425. SAIC-95/1028.

83568 00t tended range of applicatons of the software and are oesigned to temple important subsets of functions, usually for selected This report is the fourth volume in a senes of reports desenb-situatons known to be challenging or problematic, to provide ing the results of the Expert System Venfication and Validation assurance that the system achieves the tested functions with (V&V) project that is jointty funded by the U.S. Nuclear Regula-the required accuracy and performance." Such scenarios are tory Commission and the Electnc Power Research institute to used after all the V&V testing of the system is completed. Five i

develop guidelines for the V&V of expert and other systems.

categones of validation scenarios were defined: PLANT TEST, l

Here are presented the results of the Knowledge Base Certifica-BASICS, CODE, and LICENSING. A sixth type, REGRESSION, tion actrvity that was concerned with developing and testing var-ious static analysis methods for assuring the quality of knowl-is a composite of the others and refers to the practice of using trusted scenarios to ensure that software modifications did not edge bases. The testing procedure used was that of a behavior

• unintentionally change non-modified functions. A generalized al experiment involving evaluaton of four different V&V meth-procedure was developed for generating appropnate sets of val-ods. The study used two real nuclear expert systems: a boiling idaten scenanos from these basic categories.

water reactor emergency operating procedures tracking system, and a pressunzed water reactor safety assessment system. The NUREG/CR-6316 V07: GUIDELINES FOR THE VERIFICATION twenty participants were from three nuclear utshties, the USNRC AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND Technmal Training Center, the University of Maryland, EG&G CONVENTIONAL SOFTWARE. User's Manual. MILLER LA.;

Idaho, and SAIC. The major conclusion was that the use of HAYES,J E.; MIRSKY,S.M. Science Applications international tools in static knowledge base certification results in significant Corp. (formerly Science Apphcations, Inc.). March 1995.250pp.

improvement in detecting all types of defects, avoiding false 9504180447, SAIC-95/1028. 83569:001, alanns, and completing the effort in less time. The simulated This report provides a step-by step guide, or user manual, for knowledge-checking tool, based on supplemental informaton, personnel responsible for the planning and execution of the ver-was the most effective of the tools ification and validation (V&V), and also developmental testing, of expert systems, conventional software systems, and also var-b!UFEG/CR 6316 V05: GUIDELINES FOR THE VERIFICATION ious other types of artificial intelhgence systems. While the AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND guide was developed pnmarily for applications in the utility in-CONVENTIONAL SOFTWARE. Rationale And Description Of dustry, it apphes well to all industries. The user manual has V8V Guideline Packages And Procedures. MILLER L.A.;

three sections. In Section 1 the user assesses the stringency of HAYED.J E.; MIRSKY,S M. Science Apphcations Intematonal V&V needed for the system under consideration, identifies the wr @rmerly Science Apphcations, Inc.). March 1995.103pp development stage the system is in, and identifies the 9504180o41. SAIC-95/1028. 83568.131.

component (s) of the system to be tested next. These three This report is the fifth volume of a series describing the re-pieces of information determine which package of V8V meth-sults of the Expert System Venficaton and Validation (VaV) ods, called a Guidehne Package, is most apptopriate for those project jointly funded by the U.S.

Nuclear conditions. The V&V Guidehnes Packages are provided in Sec-RegulatoryCommisson and the Electric Power Research Insth tion 2. Each package consists of an ordered set of V&V tech-tute, to formulate guidelines for the V&V of expert and other niques to be apphed to the system, along with guides as to the systems. This report provides the rationale for and desenption review / evaluation team, and the measurement criteria. In Sec-of those guidelines. The actual guidehnes are presented in tion 3, the details of 11 of the most important (or least.well ex-Volume 7, " User's Manual." Three factors determine what V&V plained in the literature) methods are presented to assist the is needed. (1) the stage of the development hfecycle; (2) wheth-user in the accurate application of these techniques.

er the overall system or a speciahzed component needs to be tested. and (3) the stringency of V&V that is needed. A V&V NUREG/CR-6316 V08: GUIDELINES FOR THE VERIFICATION guidehne package is provided for each of the combinations of AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND these three variables. The package specifies the V&V methods CONVENTIONAL SOFTWARE. Bibliography.

MILLER L.A.;

recommended and the order in which they should be adminis.

HAYES J.E.; MIRSKY,S.M. Science Apphcations Intematonal tered, the assurances each method provides, the qualifications Corp. (formerly Science Applications, Inc.). March 1995. 55pp.

needed by the V8V team, the performance measures that 9504180451 SAIC-95/1028. 83568:308.

should be taken, and the decision entena. In addition to the This volume contains all of the technical references found in Duidehne packages, highly detailed step-by-step procedures are Volumes 17 concerning the development of guidelines for the provided for 11 of the most important methods, to ensure that verification and vahdation of expert systems, knowledge-based they can be implemented correctly. The guidehnes can apply to systems, other Al systems, object oriented systems, and con-convent onal as well as to Al systems.

ventional systems.

1

1 i

l I

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i 1

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i i

i i

4 m

i a

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i i

0 6

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n,--,,--

n--

- - - -n-nn,-

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.

l i

SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUM8ER C4NL 94/35 NUREG/CR4266 ORNL/TM-11568 NUREG/CR-5591 V03 ANL 95/2 NUREG/CR-4667 V18 ORNL/TM-11568 NUREG/CR 5591 V04 N2 a

BMb2181 NUREG/CR4264 V01 OANL/TM 12796 NUREG/CR4259 BML2181 NUREG/CR4264 V02

$ h h lll6

/b '

BNL NUREG-52398 NUREG/CR4141 ORNLSUB94SD4271 NUREGICR4240 BNL NUREG-52440 NUREG/CP4143 SAIC-95/1028 NUREG/CR4316 V01 '

CONF-940738 NUREG/CP 0141 SAIC-95/1028 NUREG/CR4316 V02 EGG 2746 NUREG/CR4276 SAIC-95/1028 NUREG/CR4316 V03 i

EUR 15855EN NUREG/CR4244 V03 SAKE 95/1028 NUREG/CR-6316 V04 f

EUR 15855EN NUREG/CR-6244 V01 SAIC-95/1028 NUPEG/CR4316 VOS EUR 15855EN NUREG/CR-6244 V02 ff l$

hhh'hhj$llh INEL-94//0123 NUREG/CR 6291 V04 SAIC 95/1028 NUREG/CR-6316 V08 INEL 94/0016 NUREG/CR4285 SAND 912801 NUREG/CR-5927 V02 HEL 94/0039 NUREG/CR4116 V09 SAND 93 2370 NUREG/CR-6134 IREL 94/0000 NUREG/CR4462 SAND 93-2371 NUREG/CR-6135

]

IREL 94/0123 NUREG/CR 6291 V01 SAND 93 2372 NUREG/CR4139 IREL 94/0123 NUREG/CR4291 V02 SAND 93 2440 NUREG/CR 6143 V06 P1 INEL 94/0123 NUREG/CR4291 V03 SAND 93 2440 NUREG/CR4143 V06 P2 kgh$ll hE INEL 95/0045 NUREG/CR 6260 V03 INEL-95/0070 NUREG/CR-6257 SAND 941453 NUREG/CR4244 V01 Le 129t5-MS NUREG/CR-6311 SAND 941453 NUMEG/CR4244 V02 ORNL 6791 NUREGICR-6192 UCRL ID 117418 NUREG/CR 6242 ORNL 6795 NUREG/CR-6214 UILUENG942009 NUREG/CR4259 I

h l

l 13 j

1

4

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u

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+-aer

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e d

4 4

i il

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

ADLER,M.V.

CALLEY,M.B.

NUREG 1514 GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A NUREG/CR4116 V09: SYSTEMS ANALYSIS PROGRAMS FOR NUCLEAR POWER PLANT.

HANDS ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

VERSION 5.0.Venficaton And Vahdaten (V&V) Manual.

ALLEN,K.

NUREG-1516 DAFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL CAMP 9 ELL,V.

SAFETY PROGRAMS AT MEDICAL FACILITIES Draft Report For Com.

NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL ment.

SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-

nent AMARASOORIYA.H.

NUREG/CR4310. AN ANALYSIS OF POTASSIUM ODIDE (KI) PRO.

CAMPER.LW.

PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU.

NUREG-1516 DRFT FC: MANAGEMENT OF RADOACTIVE MATERIAL CLEAR ACCOENT.

SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-ment.

BASS.B.R.

NUREG/CR4273 BtAXIAL LOADING EFFECTS ON FRACTURE CAPPS.E.L.

TOUGHNESS OF RE ACTOR PRESSURE VESSEL STEEL NUREG/CR4116 V09: SYSTEMS ANALYSIS PROGRAMS FOR HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

CATES.Ef, VERSION 5.0.Venficaton And Valklaton (V&V) Manual.

.i NUREG-1514 GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A C E" 'G 1493 DFC:

NUCLEAR POWER PLANT.

U PERFORMANCE-BASED CONTAINMENT LEAK.

CAYLESS,P.D.

TEST PROGRAM Draft Report For Comment.

NUREG/CR4285. SEVERE ACCOENT NATURAL CIRCULATION

~

SlUDIES AT THE INEL NU EG/CR 6242: CASKS (COMPUTER ANALYSIS OF STORAGE DEHLING,H.

CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR j

NUREG/CR 6310: AN ANALYSIS OF POTASSIUM ODIDE (KI) PRO.

STORAGE CASK DESIGN REVIEW. User's Manual To Vernon 1b (In-PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-ciudmg Program Reference).

CLEAR ACCIDENT.

CHOPRA,0.K.

DEHLING,K.

NUREG/CR-4867 V18: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4310 AN ANALYSIS OF POTASSIUM ODIDE (KI) PRO-LIGHT WATER REACTORS. Semiannual Report. October 1993 - March PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-CLEAR ACCIDENT.

CHUNG.H.M.

BERMUDE2,H NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL LIGHT WATER REACTORS. Semiannual Report. October 1993. March SAFETY PROGRAMS AT MEDICAL FACILITIES Draft Report For Com-ment CONLEY,0.A.

NUREG/CR4462: AGING STUDY OF DOILING WATER REACTOR HIGH PRESSURE INJECTION SYSTEMS.

NU

/ R4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT COOKE,R.M.

GRAND GULF. UNIT 1 Evaluahon Of Severe Accdent Risks For Plant NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE Operational State 5 Dunng A Refueling Outage Main Report And Ap.

UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty pereces Assessment Mam Report.

NUREG/CR-6143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4244 V02. PROBABILISTIC ACCIDENT CONSEQUENCE CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty GRAND GULF. UNIT 1.Evaluaten Of Severe Accident Riska For Plant Assessment.Appereces A And B.

Operatonal State 5 Dunng A Refueino Outage Supportng MELCOR NUREG/CR 6244 V03. PROBABILISTIC ACCOENT CONSEQUENCE Calculatons.

UNCERTAINTY ANALYSIS. Dispersion and Depositen Uncertainty AssessmentAppendices C.D.E,F,G,H.

NUREG4700 R01 DFC: HUMAN-SYSTEM INTERFACE DESIGN CORWIN W.R.

REVIEW GUIDELINE Draft Report For Comment NUREG/CR4591 V03: HEAVY-SECTION STEEL IRRADIATION PROGRAM. Progress Report For October 1991 September 1992.

BROWNSON.D.A.

NUREG/CR-5591 V04 N2: HEAVY-SECTION STEEL IRRADIATION NUREG/CR-6285' SEVERE ACCOENT NATURAL CIRCULATION PROGRAM. Semiannual Progress Report For April-September 1993.

STUDIES AT THE INEL COX.DJ, BR YSON.J.W.

NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-LOADED NUREG/CR4273 BIAXIAL LOADING EFFECTS ON FRACTURE PRESSURE RELtEF VALVES USED IN SAFETY RELATED SYSTEMS TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL AT NUCLEAR POWER PLANTS.

BURSON.S.B.

CY80LSKIS,P.

NUREG-1465: ACCOENT SOURCE TERMS FOR LIGHT. WATER NO-NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-CLEAR POWER PLANTS TEST PROGRAM Draft Report For Comment.

15

[

16 Personal Author index DEY,M.

GREENE,N.M.

NUREG 1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-NUREG/CR 6214: PRODUCTION AND TESTING OF THE VITAMIN-B6 TEST PROGRAM. Draft Report For Comment FINE GROUP AND THE BUGLE-93 BROADGROUP NEUTRON /

PHOTON CROSS SECTION LIBRARIES DERIVED FROM ENDF/B-VI NUREG/C'E NUCLEAR DATA' R4293 V02. VERIFICATIONN AND VALIDATION GUIDE-LINES FOR HIGH INTEGRfTY SYSTEMS.Appendaces A-D.

GROUNOWATER.E.H I

NUREG/CR,6316 V02: GUIDELINES FOR THE VERIFICATION AND DOSSE,C.A.

VALOATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR-6265: SEVERE ACCIDENT NATURAL CIRCULATION AL SOFTWARE. Survey And Assessment of Conventonal Software STUDIES AT THE INEL Venfcation And Vahdation Methods.

I DOODE,R.H.

NUREG/CR4316 V03: GUIDELINES FOR THE VERIFICATION AND NUREG/CR4259 CONSTRAINT EFFECTS ON FRACTURE INITIATION VALIDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-LOADS IN HSST WIDE-PLATE TESTS.

AL SOFTWARE. Survey And Documentaten Of Expert System Venfca-tion And Vahdaten Methodologies.

NUREG/CR 6257; CANDU 3 TRANSIENT ANALYSIS USING ATOMIC GRUSH,W.H.

ENERGY OF CANADA LTD CODES NUREG/CR4291 V01: NUCLEAR PLANT ANALYZER Installaton -

Manual EG CR 5462: AGING STUDY OF BOILING WATER REACTOR HIGH PRESSURE INJECTION SYSTEMS NUREG/CR-6291 V03: NUCLEAR PLANT ANALYZER. Computer Visual i

System Reference Manual.

ERCK R A'CR 4667 VtB ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6291 V04: NUCLEAR PLANT ANALYZER. Programmer's NUREG/

Manual LIGHT WATER REACTORS. Semiannual Report October 1993 March 1884 HAGEMEYER.D.

NUREG 07t3 V15: OCCUPATIONAL RADIATION EXPOSURE AT COM-FERRELL,C.M.

NUREG 1465: ACCOENT SOURCE TERMS FOR LIGHT WATER NU.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES 1993. Twenty Sixth Annual Report.

CLEAR POWER PLANTS FINEMAN.C.F.

HARPER,F.T.

NUREG/CR 5462. AGING STUDY OF BOILING WATER REACTOR NUREG/CR4244 V01: PROBABILtSTIC ACCOENT CONSEQUENCE HIGH PRESSURE INJECTION SYSTEMS, UNCERTAINTY ANALYSIS Disperson And Depositen Uncertainty Assessment. Main Report.

. FIRST,M.W.

NUREG/CR.6244 V02. PROBABILISTO ACCIDENT CONSEQUENCE I

NUREG/CP 0141: PROCEEDINGS OF THE 23RD DOE /NRC NUCLEAR UNCERTAINTY ANALYSIS. Dsperson And Depositen Uncertainty AIR CLEANING CONFERENCE. Held in Buffalo,New York July 25 Assessment.Appendees A And B.

28.1994-NUREGICR4244 V03. PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dspersion and Deposition Uncertainty FORESTER,J AssessnwntAppendices C.D.E.F G,H.

NUREG/Cd6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT HAYES.J.E.

GRAND GULF UNIT 1.Evaluaton Of Severe Accident Risks For Plant NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION AND Operatonal State 5 Dunn0 A Refuehng Outage. Main Report And Ap-VALOATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-pendees.

AL SOFTWARE.

NUREG/CR 6316 V02: GUOELINES FOR THE VERIFICATION AND FOX,P.5 VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR-6264. CRITICALITY SAFETY CRITERIA FOR LICENSE AL SOFTWARE. Survey And Assessment Of Conventional Software j

REVIEW OF LOW LEVEL WASTE FACILITIES Ventcaton And Vuhdaten Methods.

i FULLER.M.

NUREG/CR4316 V04: GUIDELINES FOR THE VERIFICATION AND

]

NUREG 1516 DAFT FC; MANAGEMENT OF RADIOACTIVE MATERIAL VALCATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-SOFTWARE.Evaluaton Of Knowledge Base Certifcaten Methods.

NUREG/CR 6316 V05: GUIDEUNES FOR THE VERIFICATION AND rnent.

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-GALVEAN W.J.

AL SOFTWARE.Ratonale And Desenpton Of V&V Guideline Packages NUREG/CR-6116 V09-SYSTEMS ANALYSIS PROGRAMS FOR And Procedures.

i i

HANDS-ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

NUREG/CR 6316 V06: GUIDEUNES FOR THE VERIFICATION AND VERSION 5.0 Venfcation And Vahdation (V8V) Manual.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON.

l AL SOFTWARE.Vahdaten Scenanos.

GANTES*

NUREG/CR4316 V07: GUOELINES FOR THE VERIFICATION AND NUREG-1514. GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A VAllDATON OF EXPEFsT SYSTEM SOFTWARE AND CONVENTON-NUCLEAH POWER PLANT.

AL SOFTWARE. User's Manual NUREG/CR 6316 V08. GUOEUNES FOR THE VERIFICATION AND GERHARD,M.A VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR$242: CASKS (COMPUTER ANALYSIS OF STORAGE AL SOFTWARE.Bbleography.

CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR STORAGE CASK DESIGN REVIEW User's Manual To Vernon 1b pn-HECHT H ciuding Program Reference).

NUREGICR-6293 V02: VERIFICATIONN AND VALIDATON GUIDE-LINES FOR HIGH INTEGRITY SYSTEMS Appendees A-D-GOLDIN.D.

NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK.

HC7.

TEST PROGRAM. Draft Report For Comment.

g c/CR4293 V02. VERIFICATIONN AND VALCATON GUIDE-GOOSSENS.LH.J.

LINES FOR HIGH INTEGRITY SYSTEMS Appendees A-D.

NUREG/CR4244 VO1: PROBABluSTIC ACCOENT CONSEQUENCE UNCERTAINTY ANALYSIS Dsperson And Depositon Uncertainty HECHT.S.

NUREG/CR 6293 V02: VERIFICATIONN AND VALIDATION GUIDE.

Assessment. Main Report.

NUREG/CR4244 V02: PROBABILISTIC ACCOENT CONSEQUENCE UNES FOR HIGH INTEGRITY SYSTEMS.Appendees A D.

UNCERTAINTY ANALYSIS. Dsperson And Depositen Uncertainty HELTON.J.C.

Assessment.Appendees A And B.

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4134. UNCERTAINTY AND SENSITIVITY ANALYSIS OF UNCERTAINTY ANALYSIS Disperson and Depositen Uncertainty CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC.

Assessment.Appendcas C,0.E.F.G.H.

CtDENT CONSEOUENCE MODEL

i h

?

Personal Author index 17 I

I i

NUREG/CR4135; UNCERTAINTY AND SENSITIVITY ANALYSIS OF

. JONES J.A.

I EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CR4244 V01: PROBABluSTIC ACCOENT CONSEQUENCE l

DENT CONSEQUENCE MODEL UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty.

NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF Assessment Main Report FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CR4244 V02: PROBABILISTIC ACCOENT CONSEQUENCE j

DENT CONSEQUENCE MODEL UNCERTAINTY ANALYSIS. Disperson And Deposition Uncertainty

)

NUREG/CR 6244 V01: PROBABILISTIC ACCOENT CONSEQUENCE 4

Assessment Appendices A And B.

UNCERTAINTY ANALYSIS. Dsperson And Depositen Uncertainty NUREG/CR-6244 V03. PROBABluSTIC ACCIDENT CONSEQUENCE 1

Assessment Main Report NUREG/CR4244 V02: PROBABlWSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dsperson and Deposition Uncertainty Assessment.Appendees C.D.E F,G.H.

UNCERTAINTY ANALYSIS. Deperson And Depositen Uncertainty j

Assessment. Appendices A And B.

JONES,J.L NUREG/CR-6244 V03: PROBABlWSTIC ACCIDENT CONSEQUENCE NUREG/CR-6116 V09. SYSTEMS ANALYSIS PROGRAMS FOR UNCERTAINTY ANALYSIS. Dsperson and Deposition Uncertainty HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

AssessmentAppendees C.D.E,F,G.H.

VERSION 5.0.Venfcation And Vahdaten (V&V) Manual.

HENDERSONA JONES,K.R.

NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION SAFETY PROGRAMS AT MEDICAL FACluTIES Draft Report For Com-STUDIES AT THE INEL.

ment.

NUREG/CR4291 V01: NUCLEAR PLANT ANALYZER installaton U EG Fk4284: CRITICAUTY SAFETY CRITERIA FOR UCENSE af REVIEW OF LOW LEVEL WASTE FACluTIES.

NUREG/CR4291 V03: NUCLEAR PLANT ANALYZER. Computer Visual l

System Reference Manual.

UR CR-6244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR4291 m MEM m mmWraWe and UNCERTAINTY ANALYSIS Dsperson And Depositon Uncertainty Assessment. Main Report JOY,D.R NUREG/CR.6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG-0430 V14:

UCENSED FUEL FACILITY STATUS UNCERTAINTY Ar4ALYSIS. Osperson And Depositen Uncertainty Assessment.Appendees A And B.

REPORT. Inventory Ofference Data. July 1,1993 - June 30,1994.

NUREG/CR4244 V03: PROBABlUSTIC ACCIDENT CONSEQUENCE KASSNER,T.F.

UNCERTAINTY ANALYSIS. Dsperson and Deposition Uncertainty Assessment.Appendees C.D.E.F,G.H.

NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual Report,0ctober 1993 March i

HUNTER,T.H.

199d' RUREG/CR-6214 PRODUCTION AND TESTING OF THE VITAMIN-B6 FINE GROUP AND THE BUGLE.93 BROAD. GROUP NEUTRON /

KHAN,T.A.

PHOTON CROSS-SECTION UBRARIES DERIVED FROM ENDF/B.VI NUREG/CP-0143. PROCEEDINGS OF THE THIRD INTERNATIONAL NUCLEAR DATA.

WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR POWER PLANTS. Held At Hauppauge, Long Island. New York.

HUTCHINSON,J.W.

RUREG/CR-6264 V02. VAUDITY UMITS IN J-RESISTANCE CURVE DETERMINATION A Computatonal Approach To Ductile Crack Growth N R G/CR 6141: HANDBOOK OF METHODS FOR RISK BASED Under Large Scale Yielding Conditions.

ANALYSES OF TECHNICAL SPECIFICATIONS.

INGERSOLL,0.T.

KMETYK,LN.

MUREG/CR4214. PRODUCTION AND TESTING OF THE VITAMIN-B6 NUREG/CR4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PHOTON CROSS.SECTION UBRARIES DERIVED FROM ENDF/B-VI GRAND GULF, UNIT 1.Evaluaton of Severe Accident Risks For Plant NUCLEAR DATA.

Operatonal State 5 Dunng A Refueling Outage.Mam Report And Ap-pendices JEDD.J.L NUREG/CR-6143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC.

OlVREG/CR4257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ENERGY OF CANADA LTD CODES.

GRAND GULF,0 NIT 1. Evaluation Of Severe Accident Risks For Plant f

Operatonal State 5 Dunng A Refueling Outage. Supporting MELCOR JOHNSON,G.L Calculations.

NUREG/CR4242: CASKS (COMPUTER ANALYSIS OF STORAGE CASKS) A MICROCOMPUTER BASED ANALYSl$ SYSTEM FOR KOKAJKO,LE.

STORAGE CASK DESIGN REVIEW. User's Manual To Verson 1b (In.

NUREG 1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-ciuding Program Reference).

TIONS REVIEW TEAM.

JOHNSON,J.

KOTSCH,J.

MUREG/CR-6143 V06 Pt. EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4310: AN ANALYSIS OF POTASSIUM IODIDE (Kl) PRO-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-t GRAND GULF. UNIT 1. Evaluation Of Severe Accident Riska For Plant CLEAR ACCIDENT.

I Operatonal State 5 Dunng A Refuehng Outage Main Report And Ap-pendices.

KOZAK,M.W.

i NUREG/CR 5927 V02: EVALUATION OF A PERFORMANCE ASSESS-JOHNSON.J.D.

MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE NUREG/CR4134. UNCERTAINTY AND SENSITIVITY ANALYSIS OF DISPOSAL FACluTIES.Vahdaten Needs.

CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC.

CfDENT CONSEQUENCE MODEL KO2LOWSKI S.D.

RUREG/CR4135. UNCERTAINTY AND SENSITIVITY ANALYSTS OF NUREG/CR4276; A COMPILATION OF CURRENT REGULATIONS.

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-STANDARDS, AND GUIDEUNES IN REMOTE AFTERLOADING BRA.

DENT CONSEQUENCE MODEL CHYTHERAPY.

MUREG/CR 6136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl.

KRAAN.B.

j DENT CONSEQUENCE MODEL NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Deperson And Deposition Uncertainty JONESA AssessmentMain Report.

NUREG-1516 DAFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL NUREG/CR4244 V02: PROBABluSTIC ACCIDENT CONSEQUENCE SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-UNCERTAINTY ANALYSIS. Deperson And Depoorhon Uncertainty ment Assessment.Appendcas A And B.

10 Personal Author index i

. NUREG/CR4244 V03; PROBABILtSTIC ACCIDENT CONSEQUENCE AL SOFTWARE. Survey And Documentaten Of Expert System Vsnfce-UNCERTAINTY ANALYSIS. Deperson and Depositen Uncertainty ton And Vahdaten Methodolo0ies.

Assessment Appendees C.D.E F,G,H.

NUREG/CR4316 V04: GUCEUNES FOR THE VERIFICATION AND VAUDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL "E"OE M SOFTWARE.Evaluaten Of Knowledge Base Certifcation Methods.

L NUREG/CP.0145.

WORKSHOP ON DEVELOPING SAFE NUREG/CR4316 V05: GUIDEUNES FOR THE VERIFICATION AND SOFTWARE. Held At Hotel Del Coronado. San Diego,CA. July 22-VAUDATsuN OF EXPERT SYSTEM SOFTWARE AND CONVENTON-23,W2' AL SOFTWARE.Ratonale And Desenption Of V&V Guidehne Packages And Procedures.

LEE.R.V.

NUREG 1465: ACCIDENT SOURCE TERMS FOR LIGHT WATER NU-NUREG/CH4316 V06: GUIDELINES FOR THE VERIFICATON AND VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-CLEAR POWER PLANTS.

AL SOFTWARE.Validetion Scenenos.

UU,X.H-NUREG/CR4316 V07: GUIDEUNES FOR THE VERIFICATION AND NUREG/CR4264 V01: VAUDITY UMITS IN J-RESISTANCE CURVE VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON.

DETERMINATION.An Assessment Of The J(M) Parameter.

AL SOFTWARE. User's Manual.

NUREG/CR4316 V06: GUCEUNES FOR THE VERIFICATION AND U E'G/CR4244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE A

R B i

UNCERTAINTY ANALYSIS. Deperson And Depositen Uncertainty Assessment Main Report-MINTON,L.

NUREG/CR4244 V02. PROBABluSTIC ACCOENT CONSEQUENCE NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-UNCERTAINTY ANALYSIS. Daperson And Deposition Uncertainty TEST PROGRAM. Draft Report For Comment.

Assessment Appendices A And B.

NUREG/CRM44 V03. PROBABILISTO ACCIDENT CONSEQUENCE 1

e A d.E G NU E R4316 V01: GUOELINES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-MACFARLANE,R.E.

AL SOFTWARE.

NUREG/CR4214. PRODUCTON AND TESTING OF THE VITAMIN.B6 NUREG/CR-6316 V02: GUCEUNES FOR THE VERIFICATION AND FINE GROUP AND THE BUGLE-93 BROAD GROUP NEUTRON /

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-PHOTON CROSS-SECTION UBRARIES DERIVED FROM ENOF/B VI AL SOFTWARE. Survey And Assessment Of Conventonal Software NUCLEAR DATA.

Vertficaton And Vahdaton Methods.

NUREG/CR4316 V03: GUOELINES FOR THE VERIFICATON AND I

MANNAMO,T.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR 6141: HANDBOOK OF METHODS FOR RISK-BASED AL SOFTWARE. Survey And Documentation Of Expert System Venfica-ANALYSES OF TECHNICAL SPECIFICATONS-ton And Vahdation Methodologses.

NUREG/CR4316 V04: GUIDELINES FOR THE VERIFICATION AND

]

MCAFFE W.J VAUDATON EXPERT SYSTEM SOFTWARE AND CONVENTIONAL NUREd/Cfi4273. BIAXtAL LOADING EFFECTS ON FRACTURE SOFTWARE.Evaluaten Of Knowledge Base Certification Methods.

TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL' NUREG/CR4316 V05: GUIDELINES FOR THE VERIFICATION AND i

McKAY,M.D.

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-i NUREG/CR4135. UNCERTAINTY AND SENSITIVITY ANALYSIS OF AL SOFTWARE.Ratonale And Desenpton Of V&V Guideline Packages 1

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl.

And Procedures.

DENT CONSEQUENCE MODEL NUREG/CR4316 V06: GUIDELINES FOR THE VERIFICATION AND NUREG/CR4244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-

)

UNCERTAINTY ANALYSIS Dsperson And Depositon Uncertainty AL SOFTWARE.Validaten Scenanos Assessment Main Report NUREG/CR-6316 V07: GUIDEUNES FOR THE VERIFICATON AND i

NUREG/CR4244 V02: PROBABluSTC ACCOENT CONSEQUENCE VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

UNCERTAINTY ANALYSIS. Deperson And Depositon Uncertainty AL SOFTWARE. User's Maitual.

Assessment Appendices A And B.

NUREG/CR4316 V06: GUIDEUNES FOR THE VERIFICATION AND NUREG/CR4244 V03: PROBASIUSTIC ACCOENT CONSEQUENCE VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-i UNCERTAINTY ANALYSIS. Daperson and Deposition Uncertainty AL SOFTWARE Biblography.

Assessment Appendices C.DE.F.G H 1

NUREG/CR4311; EVALUATING PR$ DICTION UNCERTAINTY.

M OK,G A MtCHAUD,WI.

NUREG/CR4242: CASKS (COMPUTER ANALYSIS OF STORAGE NUREG/CR-4667 Vie: ENVIRONMENTALLY ASSISTED CRACKING IN CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR LIGHT WATER REACTORS, Samaannual Report. October 1993 March STORAGE CASK DESIGN REVIEW User's Manual To Vernon 1b (In-1994.

ciuding Program Reference).

MILLER.L.A.

MONTGOMERY,J.

NUREG/CR-6143 V06 Pt: EVALUATION OF POTENTIAL SEVERE AC-NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SAFETY PROGRAMS AT MEDOAL FACILITIES. Draft Report For Com-GRAND GULF, UNIT 1 Evaluation Of Severe Accedent Reaks For Plant ment.

Operatonal State 5 Dunng A Refuehng Outage. Main Report And Ap-j pendices MORTON,0.K.

i NUREG/CR4244 V01: PROBABluSTIC ACCOENT CONSEQUENCE NUREG/CR4260: APPUCATON OF NUREG/CR-5999 INTERIM FA-UNCERTAINTY ANALYSIS. Deperson And Deposition Uncertainty TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-4 NUR 6 4 ROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Deperson And Depositen uncertainty MURPHY,D.D.

NUREG 1517. REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-NUR C 42 P

B ISTIC ACCIDENT CONSEQUENCE TIONS REVIEW TEAM.

UNCERTAINTY ANALYSIS. Daperson and Depositen Uncertamty

1. MM NUR CR I L FOR THE VERIFICATION AND NUREG/CR4260; APPUCATION OF NUREG/CR 5999 INTERIM FA-VAltDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-AL SOFTWARE.

i NUREG/CR4316 V02: GUOEUNES FOR THE VERIFOATON AND NENTS.

1 VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-NOVACK,8.D.

AL SOFTWARE. Survey And Assessment Of Conventonal Software Venticaten And Vahdation Methods.

NUREG/CR.6116 V09. SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4316 V03: GUIDEUNES FOR THE VERIFICATION AND HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

VAllDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTON-VERSION 5.0 Venfication And Vahdaten (V8V) Manual.

P;rsonal Author index 19 NOWLEN,S.P.

S %NECKl,J.E NUREG/CR4220: AN ASSESSMENT OF FIRE VULNERABluTY FOR NUREG/CR-4067 V18: ENVIRONMENTALLY ASSISTED CRACKING IN j

AGED ELECTRCAL RELAYS.

UGHT WATER REACTORS. Semiannual Report October 1993. March O'9AIENJI.

I#'

NUREG/CR4206: SEVERE ACCIDENT NATURAL CIRCULATION SCHLENKER,LD.

- STUDIES AT THE INEL NUREG/CR-6285. SEVERE ACCIDENT NATURAL CIRCUl ATION O'HARA).M.

STUDIES AT THE INEL.

NUREG-0700 ROI DFC: HUMAN-SYSTEM INTERFACE DESIGN REVIEW GUIDEUNE. Draft Report For Comment.

UAEG 1 6 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL ODEGAARDENAM.

SAFETY PROGRAMS AT MEDICAL FACluTIES. Draft Report For Com-NUREG/CR4284: CRITICAUTY SAFETY CRITERIA FOR UCENSE ment.

REVIEW OF LOW LEVEL WASTE FACluTIES.

SHA.W.T.

OLAGUE,NA.

NUREG/CR4266: ANALYSIS OF BORON DILUTION IN A FOUR LOOP NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE ASSESS.

pwg' MENT METHODOLOGY FOR LOW-LFVEL RAN>ACTr/E WASTE DISPOSAL FACluT!ES.Vahdaten Needs.

SHACK,W.J.

NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN PAFFORD.D.J'6285:

LIGHT WATER REACTORS. Semiannual Report. October 1993 - March NUREG/CR SEVERE ACCIDENT NATUHAL CIRCULATION STUDIES AT THE INEL 1994-PARKS.C.V.

SHIH,C.F.

NUREG/CR4284: CRITICAUTY SAFETY CRITERIA FOR LICENSE NUREG/CR4264 V01: VALIDITY UMITS IN J-RESISTANCE CURVE REVIEW OF LOW LEVEL WASTE FACluTIES.

DETERMINATION.An Assessment Of The J(M) Parameter.

r NUREG/CR4264 V02: VAUDITY LIMITS IN J-RESISTANCE CURVE PASLER-SAUER,J.

DETERMINATION.A Computatonal Approach To Ductile Crack Growth NUREG/CR4244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE Under Large-Scale Yielding Conditions.

UNCERTAINTY ANALYSIS. Disperson And Deposition Uncertainty i

Assessment Main Report SHIVER,A.W.

NUREG/CR4244 V02: PROBABluSTIC ACCIDENT CONSEOUENCE NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC.

Assessment. Appendices A And B-CIDENT CONSEQUENCE MODEL NUREG/CR4244 V03. PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF UNCERTAINTY ANALYSIS. 06spersion and Depositen Uncertainty EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-Assessment. Appendices C.D.E.F,G.H.

DENT CONSEQUENCE MODEL NUREG/CR-6136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NU E / R 6273: BIAXIAL LOADING EFFECTS ON FRACTURE FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL DENT CONSEQUENCE MODEL PERSENSKYJ J.

SHUMWAYAW.

NUREG-0700 R01 DFC: HUMAN SYSTEM INTERFACE DESIGN NUREG/CR.6257; CANDU 3 TRANSIENT ANALYSIS USING ATOMIC REVIEW GUIDELINE. Draft Report For Comment.

ENERGY OF CANADA LTD CODES.

CADDATZ,C.T.

SIMlON.G.P.

NUREG-0713 V15: OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR4276: A COMPILATION OF CURRENT REGULATIONS, MERCIAL NUCLEAR POWER REACTORS AND OTHER STANDARDS, AND GUiOELINES IN REMOTE AFTERLOADING BRA-FACluTlES.1993. Twenty Smth Annual Report.

CHYTHERAPY.

RfDGELY,J.N.

SKAY,D.M.

NUREG 1465: ACCIDENT SOURCE TERMS FOR UGHT-WATER NU-NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-CLEAR POWER PLANTS.

TIONS REVIEW TEAM, ROLLS TIN.J.A.

NUREG/CR 6134. UNCERTAINTY AND SENSITIVITY ANALYSIS Op SKD8LAR,L CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-NUREG 1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-CtDENT CONSEQUENCE MODEL.

TEST PROGRAM. Draft Report For Comment.

NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF F D AT AY RESULT WITH THE MACCS REACTOR ACCl-SLATER C.O FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEtJTRON/

ROMAN.W.

PHOTON CROSS-SECTION UBRARIES DERIVED FROM ENDF/B Vi NUREG.1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK.

NUCLEAR DATA.

TEST PAOGRAM Draft Report For Comment.

SMITH.C.L ROSE.S.

NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR NUREG-1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK' HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

TEST PROGRAM Draft Report For Comment-VERSION 5 0.Venficaton And Vahdaten (V&V) Manual ROUSSIN.R.W.

SNIDER,D.M NUREG/CR4214: PRODUCTION AND TESTING OF THE VITAMIN-B6 NUREG/Cb6291 V01: NUCLEAR PLANT ANALYZER.installaton FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

PHOTON CROSS-SECTION UBRARIES DERIVED FROM ENDF/B V1 NURE CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

NUCLEAR DATA.

Mand KUTHER,W.E.

NUREG/CR4291 V03. NUCLEAR PLANT ANALYZER. Computer Visual NUREG/CR-4667 V18. ENVIRONMENTALLY AS$1STED CRACKING IN System Reference Manual.

UGHT WATER REACTORS Semiannual Report. October 1993 - March NUREG/CR-6291 V04: NUCLEAR PLANT ANALYZER. Programmer's 1994.

Manual.

SAMANTA,P.K.

SOFFER.L NUREG/CR-6141: HANDBOOK OF METHODS FOR RISK BASED NUREG 1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER NU.

ANALYSES OF TECHNICAL SPECl"lCATIONS.

CLEAR POWER PLANTS.

20 Personal Author Index SOPPET,W.K.

WAGNER,K.L NUREG/CR-4667 V18 ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6291 V01: NUCLEAR PLANT ANALYZER.installat on LIGHT WATER REACTORS. S4nuannual Report. October 1993 March Manual.

1994 NUREG/CR-6291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

ence Manual SPRUNG,JL NUREG/CR-6291 V03: NUCLEAR PLANT ANALYZER. Computer Visual NUREG/CR 6134 UNCERTAINTY AND SENSITIVITY ANALYSIS OF System Reference Manual.

CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-NUREG/CR-6291 V04: NUCLEAR PLANT ANALYZER Programmer's CIDE NT CONSEQUENCE MODE L Manual.

NUREG/CR 6135 UNCERTAINTY AND SENSITIVITY ANALYSIS OF EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-WANG,H-B.

DE NT CONSEQUENCE MODEL NUREG-1517. REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-NUREG/CR 6136 UNCERTAINTY AND SENSITIVITY ANALYSIS OF TIONS REVIEW TEAM.

FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-DENT CONSEOUENCE MODEL.

W ARE,A.G.

j NUREG/CR-6260: APPUCATION OF NUREG/CR-5999 INTERIM FA.

N JREG/C

92. AGING AND SERVICE WEAR OF SPRING LOADED E

S' PRESSURE REUEF VALVES USED IN SAFETY-RELATED SYSTEMS AT NUCLEAR POWER PLANTS WEINSTEIN,E.D.

^

STEVENSON.J.D NUCLEAR POWER PLANT.

NUREG/CR4240 APPLICATION OF DOUNDING SPECTRA TO SEIS-MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF WHITE,J E.

ABOVE GROUND PtPING IN POWER PLANTS SUBJECTED TO NUREG/CR-6214 PRODUCTION AND TESTING OF THE VITAMIN-86 STRONG MOTION EARTHOUAKES FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-VI STRUCKMEYER,R.

NUCLEAR DATA.

j NUREG.0837 v14 N04 NRC TLD DIRECT RADIATION MONITORING NETWORK Progress Report. October-December 1994 WHITEHEAD.D.

NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-STUBLER,W.F.

NUREG 0700 ROI DFC: HUMAN SYSTEM INTERF ACE DESIGN CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT REVIEW GUIDEUNE Draft Report For Comment GRAND GULF, UNIT 1. Evaluation Of Severe Accident Riska For Plant Operational State 5 Dunng A Refuehng Outage Main Report And Ap-SUN.J.G.

pendices NUREG/CR-6266. ANALYSIS OF BORON DILUT:ON IN A FOUR-LOOP PWR WOLFRAM L.M.

NUREG/CR-6116 V09-SYSTEMS ANALYSIS PROGRAMS FOR TANG.D.

HANDS-ON INTEGRATED RELIABIUTY EVALUATIONS (SAPHIRE)

WUREG/CR 6293 V02 VERIFICATIONN AND VAUDATION GUIDE-VERSION 5 0.Venfication And Vahdation (V&V) Manual.

LINES FOR HIGH INTEGRITY SYST EMS Appendices A D WRIGHT.R.O.

TORTORELLI,J.P-NUREG/CR-6214 PRODUCTION AND TESTING OF THE VITAMIN-86 NUREG/CR 6276 A COMPILATION OF CURRENT REGULATIONS.

FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING BRA' PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-VI CHYTHERAPY-NUCLEAR DATA.

TRUMMER,0.J.

ygg,g NUREG/CR 6242. CASKS (COMPUTER ANALYSIS OF STORAGE NUREG/CR-6264 V02 VAUDITY LIMITS IN J-RESISTANCE CURVE CASKS) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR DETERMINATION.A Computat:onal Approach To Ductile Crack Growth STORAGE CASK DESIGN REVIEW User's Manual To Version 1b (In-Under Large-Scale Yielding Conditions clutiing orogram Reference)

YOUNG,M.L O^

NU G CR-6265. SEVERE ACCIDENT NATURAL CIRCULATION UNCERTAINTY ANALYSIS Dispersion And Deposmon Uncertainty STUD ES AT THE INEL.

Assessment Main Report.

NUREG/CR-6244 V02: PROBABluSTIC ACCIDENT CONSEOUENCE VESELY,W.E.

UNCERTAINTY ANALYSIS Dispersion And Deposition Uncertainty NUREG/CR 6141. HANDBOOK OF METHODS FOR RISK BASED Assessment. Appendices A And B.

ANALYSES OF TECHNICAL SPECIFICATIONS NUREG/CR 6244 V03 PROBABILISTIC ACCIDENT CONSEQUENCE tflGILAA.

UNCERTAINTY ANALYSIS. Dispersion and Deposmon Uncertainty NUREG/CR 6220: AN ASSESSMENT OF FIRE VULNERADIUTY FOR Assessmer 1. Appendices C.D.E.F,G.H.

AGED ELECTRICAL RELAYS ZEIGLER.S.L CACHTEL.J.A.

NUREG/CR 6116 V09-SYSTEMS ANALYSIS PROGRAMS FOR NUREG-0700 ROI DFC HUMAN-SYSTEM INTERFACE DESIGN HANDS-ON INTEGRATED REUABluTY EVALUATIONS (SAPHIRE)

REVIEW GUIDELINE Draft Report For Comment VERSION 5 0 Venfication And Validation (V&V) Manual.

l.

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

ALARA CANDU3 NUREG/CP 0143: PROCEEDINGS OF THE THIRD INTERNATIONAL NUREG/CR4257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR ENERGY OF CANADA LTD CODES.

POWER PLANTS Held At Hauppauge, Long Island, New York.

l CASKS NUREG/CR-6242: CASKS (COMPUTER ANALYSIS OF STORAGE REG /CR 5462: AGING STUDY OF BOILING WATER REACTOR HIGH PRESSURE INJECTION SYSTEMS CASKS). A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR NUREG/CR4192 AGING AND SERVICE ' EAR OF SPRING-LOADED STORAGE CASK DESIGN REVIEW. User's Manual To verson 1b (In-W PRESSURE REUEF VALVES USED IN SAFETY-RELATED SYSTEMS ciuding Program Reference).

AT NUCLEAR POWER PLANTS NUREG/CR 6220. AN ASSESSMENT OF FIRE VULNERABILITY FOR Code Assessment AGED ELECTRICAL RELAYS.

NUREG/CR4285. SEVERE ACCIDENT NATURAL CIRCULATION i

STUDIES AT THE INEL.

Allegat6one Review Team i

NUREG 1517. REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-Constraint Effect t

TIONS REVIEW TEAM-NUREG/CR4259: CONSTRAINT EFFECTS ON FRACTURE INITIATION LOADS IN HSST WIDE. PLATE TESTS.

Atmoepheric Deposition NUREG/CR4244 V03: PROBABluSTIC ACCIDENT CONSEQUENCE Containment P

NUREG 1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK-asess D,E TEST PROGRAM Draft Report For Comment.

Atmospheric D6spersion NUREG/CR-6244 VOI: PROBABILISTIC ACCIDENT CONSEQUENCE Core Meltdown UNCERTAINTY ANALYSIS. Dispersson And Depositen Uncertainty NUREG-1465: ACCIDENT SOURrX TERMS FOR UGHT-WATER NU-Assessment Main Report.

CLEAR POWER PLANTS.

NUREG/CR-6244 V02: PROBABluSTIC ACCIDENT CONSEOVENCE UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty Corrosion Fatigue i

Assessment. Appendices A And B.

NUREG/CR 4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS. Semiannual Report. October 1993 March bg Dm g '

E NUREG 0713 V15: OCCUPATIONAL RADIATION EXPOSURE AT COM-MERCIAL NUCLEAR POWER REACTORS AND OTHER Crack Growth FACluTIES,1993 Twenty-Sath Annual Report.

NUREG/CR-6264 V01: VALIDITY LIMITS IN J RESISTANCE CURVE EUGLE43 DETERMINATION.An Assessment Of The J(M) Parameter.

NUREG/CR-6214; PRODUCTION AND TESTING OF THE VITAMIN-86 NUREG/CR4264 V02: VAUDITY UMITS IN J-RESISTANCE CURVE FINE GROUP AND THE BUGLE-93 BROAD GROUP NEUTRONf DETERMINATION.A Computational Approach To Ductile Crack Growth PHOTON CROSS-SECTION UBRARIES DERIVED FROM ENDF/B-VI Under Large Scale Yielding Condit ons.

NUCLEAR DATA.

BWR NUREG/CR4266: ANALYSIS OF BORON DILUTION IN A FOUR-LOOP NUREG/CR-5462: AGING STUDY OF BOILING WATER REACTOR PWR.

HtGH PRESSURE INJECTION SYSTEMS.

Discrimination lesue NURE /CR 273-BIAXIAL LOADING EFFECTS ON FRACTURE T

ElEW EAM.

TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL EW Nuclear Date Souing Water Reactor NUREG/CR-5462. AGING STUDY OF BOluNG WATER REACTOR NUREG/CR-6214: PRODUCTION AND TESTING OF THE VITAM:N-06 HIGH PRESSURE INJECTION SYSTEMS.

FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-VI Boron NUCLEAR DATA.

NUREG/CR4266: ANALYSTS OF BORON DILUTION IN A FOUR LOOP PWR Earthquake NUREG/CR-6240: APPUCATION OF BOUNDING SPECTRA TO SEIS-Bounding spectra MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF

. NUREG/CR 6240: APPUCATION OF BOUNDING SPECTRA TO SEIS-ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF STRONG MOTION EARTHOUAKES' ABOVE GROUND PtPING IN POWER PLANTS SUBJECTED TO STRONG MOTION EARTHOUAKES.

Emergency Planning Brachytherapy NUREG/CR 6310: AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO-NUREG/CR-6276' A COMPILATION OF CURRENT REGULATIONS.

PHYLAXIS FOR THE GENERAL PUBUC IN THE EVENT OF A NU.

STANDARDS, AND GUIDEUNES IN REMOTE AFTERLOADING BRA.

CLEAR ACCIDENT.

Emergency Response

[

Budget NUREG 1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A NUREG-1100 V11: BUDGET ESTIMATES Fracal Years 1996-1997.

NUCLEAR POWER PLANT.

l 21

22 Subject Index Enforcement Action J-Reeletence NUREG-0940 V13 N4 P1: ENFORCEMENT ACTIONS SIGNIFICANT AC-NUREG/CR4264 V01: VALIDITY LIMITS IN J-RESISTANCE CURVE T10NS RESOLVED REACTOR UCENSEESQuarterly Progress DETERMINATION An Assessment Of The J(M) Parameter.

Report, October December 1994.

NUREG/CR4264 V02: VALIDITY LIMITS IN JRESISTANCE CURVE NUREG-0940 V13 N4 P2: ENFORCEMENT ACTIONS.StGNIFICANT AC-DETERMINATON.A Computational Approach To Ductile Crack Growth -

TIONS RESOLVED MEDICAL LICENSEES. Quarterly. Progress Under Large-Scale Yielding Conditions.

Report. October-December 1994 NUREG 0940 V13 N4 P3' ENFORCEMENT ACTIONS.SIGNIFICANT AC-LWR TONS RESOLVED MATERIAL LICENSEES (NON MEDICAL).Ouarterly NUREG 1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER NU-Progress Report, October-December 1994.

CLEAR POWER PLANTS.

NUREG/CR-4667 VIB: ENVIRONMENTALLY ASSISTED CRACKING IN

~

NU R4260. APPLICATICN OF NUREG/CR4999 INTERIM FA-9' TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-NENTS.

Leak-Test Program NUREG 1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK-TEST PROGRAM.Dra't Report For Comment.

NUR G 1514' GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A NUCLEAR POWER PLANT, Legallesuancee NUREG-0750 V40102: INDEXES TO NUCLEAR REGULATORY COM-Failure Mode M

iSSUAN J ly 1 4 NUREG/CP-01d5:

WORKSHOP ON DEVELOPING SAFE NU oV SOFTWARE Held At Hotel Del Coronado, San Deego CAJuly 22 SUANCES FOR NOVEMBER 1994. Pages 169-318.

23J992.

NUREG 0750 V40 N06: NUCLEAR REGULATORY COMMISSION IS-F6re Vulnerability SUANCES FOR DECEMBER 1994. Pages 319-387.

I NUREGICR4220 AN ASSESSMENT OF FIRE VULNERABILITY FOR NUREG-0750 V41 N01: NUCLEAR REGULATORY COMMISSION IS-

)

AGED ELECTRICAL RELAYS.

SUANCES FOR JANUARY 1995. Pages 169.

Flacal Year Licensed Fuel Factitty Status Report l

NUREG-1100 Vit; DVDGET ESilMATES.Fascal Years 1996-1997.

NUREG-0430 V14:

LICENSED FUEL FACILITY STATUS REPORT. Inventory Difference Data. July 1.1993 June 30,1994.

Fracture Mechan 6ce NUREG/CR4259. CONSTRAINT EFFECTS ON FRACTURE INITIATON Ught Water Reactor LOADS IN HSST WIDE PLATE TESTS.

NUREG-1465: ACCIDENT SOURCE TERMS FOR LIGHT WATER NU-NUREG/CR4273 BIAXIAL LOADING EFFECTS ON FRACTURE CLEAR POWER PLANTS.

TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual RepotOctober 1993 - March i

Fracture Toughness 9994.

j NUREG/CR4273 BIAXtAL LOADING EFFECTS ON FRACTURE i

TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL Low Power NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-l Gener6c Safety lasun CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG 1435 SO4. STATUS OF SAFETY ISSUES AT LICENSED GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant POWER PLANTS.TMI Action Plan Requirements, Unresolved Safety Operational State 5 Dunng A Refuehng Outage. Main Report And Ap-lasues,Genere Safety issues.Other Multiplant Action issues.

pendcet NUREG/CR4143 V06 P2: EVALUATON OF POTENTIAL SEVERE AC-Heavy-Section Steel irredtation Program CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-5591 V03. HEAVY-SECTION STEEL IRRADIATION PROGRAM Progress Report For October 1991 September 1992.

GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant NUREG/CR4591 V04 N2: HEAVY SECTION STEEL IRRADIATION Operational State 5 Dunng A Refuehng Outage Supporting MELCOR j

PROGRAM. Semiannual Progress Report For April-September 1993.

Calculationt High integrity System Low-Level Weste D6aposal NUREG/CR-6293 V01: VERIFICATION AND VALIDATION GUIDELINES NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE ASSESS-FOR HIGH INTEGRITY SYSTEMS Main Report MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE NUREG/CR-6293 V02. VERIFICATIONN AND VAllDATION GUIDE-DISPOSAL FACILITIES.Vahdation Needs, LINES FOR HIGH INTEGRITY SYSTEMS Appendices A.D.

Low-Level Weste Facility H6gh Pressure injection System NUREG/CR 6284: CRITICALITY SAFETY CRITERIA FOR LICENSE NUREG/CR4462: AGING STUDY OF BOILING WATER REACTOR REVIEW OF LOW-LEVEL WASTE FACILITIES.

HOH PRESSURE INECTON SYSTEMS.

MACCS Human facto' NUREG/CR-6134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG-0700 ROI DFC: HUMAN-SYSTEM INTERFACE DESIGN CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-REVIEW GUIDELINE Draft Report For Comment.

CIDENT CONSEQUENCE MODEL NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF Human-System interface EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-NUREG-0700 ROI DFC: HUMAN-SYSTEM INTERFACE DESIGN REVIEW GUIDELINE Draft Report For Comment.

NURE /

N RT nth AND SENSITIVITY ANALYSIS OF IRRAS FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CR4116 V09-SYSTEMS ANALYSIS PROGRAMS FOR DENT CONSEQUENCE MODEL HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

M n Ac on use VERSON 5.0.Venicat on And Validation (VSV) Manual STATUS OF SAFETY ISSUES AT ~ UCENSED industrial Radiography POWER PLANTS.TMl Acton Plan Requirements. Unresolved Safety NUREG4713 V15. OCCUPATONAL RADtATION EXPOSURE AT COM-Issues Genere Safety lasues.Other Multiplant Action lasues.

MERCIAL NUCLEAR POWER REACTORS AND OTHER FACILITIES 1993. Twenty-Sixth Annual Report Natural Circulat6on NUREG/CR 6285: SEVERE ACCIDENT NATURAL CIRCULATION J-R Curve STUDIES AT THE INEL NUREG/CR 6264 V01: VALIDITY LIMITS IN J RESISTANCE CURVE DETERMINATON An Assessment Of The J(M) Parameter, Nor21e NUREG/CR4264 V02. VALIDITY LIMITS IN J-REStSTANCE CURVE NUREG/CR4260 APPLICATION OF NUREG/CR-5999 INTERIM FA-DETERMINATION A Computational Approach To Ductile Crack Growth TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-Under Large-Scale Yielding Conditions.

NENTS

Subject incex 23 Nuclear Accident GRAND GULF,0 NIT 1 Evaluaton Of Severe Accident Risks For Plant NUREG/CR4310 AN ANALYSIS OF POTASSIUM ODIDE (KI) PRO-Operatonal State 5 Dunng A Refuehng Outage Supporting MELCOR PHYLAXIS FOR THE GENERAL PUBUC IN THE EVENT OF A NU-Calculatons.

CLEAR ACCIDENT.

Quality Assurance Nuclear Air Cleaning NUREG/CR4316 V01: GUIDELINES FOR THE VERIFICATON AND NUREG/CP 0141: PROCEEDINGS OF THE 23RD DOE /NRC NUCL EAR VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AIR CLEANING CONFERENCE. Held in Buffalo,New YorkJufy 25-AL SOFTWARE.

28,1994.

NUREG/CR4316 V02: GUIDELINES FOR THE VERIFICATION AND VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AL S Ehey And Assessant O Convenpal ham UR / 42 V01: NUCLEAR PLANT ANALYZER.Instatlaton Venficaten And Validaten Methods.

NUREG/CR4316 V03: GUIDELINES FOR THE VERIFICATON AND NURE CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-NLRTG 91 V03. NUCLEAR PLANT ANALYZER Computer Visual AL SOFTWARE. Survey And Documentaten Of Expert System venfica-System Reference Manual.

ton And Validaten Methodologies-NUREG/CR-6291 V04. NUCLEAR PLANT ANALYZER. Programmer's NUREG/CR-6316 V04: GUIDELINES FOR THE VERIFICATION AND Manual.

VAUDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL SOFTWARE.Evaluaton Of Knowledge Base Certificaton Methods.

Nuclear Safety NUREG/CR-6316 VOS: GUIDELINES FOR THE VERIFICATION AND NUREG4700 RO1 DFC: HUMAN SYSTEM INTERFACE DESIGN VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-REVIEW GUIDELINE Draft Report For Comrnent.

AL SOFTWARE. Rationale And Descnpton Of V8V Guidehne Packages And Procedures.

Occupational Exposure NUREG/CR-6316 V06: GUIDELINES FOR THE VERIFICATION AND NUREG/CP 0143 PROCEEDINGS OF THE THIRD INTERNATIONAL VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR AL SOFTWARE Validaten Scenares.

POWER PLANTS Held At Hauppauge, Long Island, New York NUREG/CR-6316 V07. GUOELINES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-

^L R / 6 2 GING AND SERVICE WEAR OF SPRING-LOADED PUREG/

V G IDE INES FOR THE VERIFICATION AND PRESSURE REUEF VALVES USED IN SAFETY RELATED SYSTEMS VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AT NUCLEAR POWER PLANTS.

AL SOFTWARE Biblography.

PWR Q

y ae NU G/CR4266. ANALYSIS OF BORON DILUTION IN A FOUR-LOOP 26 A COMPILATION OF CURRENT REGULATIONS, STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING BRA-Performance Assessment CHYTHERAPY.

NUREG/C9-5927 V02: EVALUATION OF A PERFORMANCE ASSESS-MENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE WASTE RELAP51nterface DISPOSAL FACluTIES Vahdaten Needs.

NUREG/CR 6291 V01: NUCLEAR PLANT ANALYZER. installation Manual.

Petitione For Rulemaking NUREG/CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

NUREG 0936 V13 NO3: NRC REGULATORY AGENDA. Semiannual ence Manual.

ReportJuly-December 1994.

NUREG/CR-6291 V03 NUCLEAR PLANT ANALYZER. Computer Visual System Reference Manual.

Tipin9 NUREG/CR4291 V04: NUCLEAR PLANT ANALYZER. Programmer's NUREG/CR 6240- APPUCATION OF BOUNDING SPECTRA TO SEIS-Manual.

MIC OESIGN OF PIPING 6ASED ON THE PERFORMANCE OF ABOVE GROUND PIPINO IN POWER PLANTS SUBJECTED TO Radiation Dose

' STRONG MOTION EARTMUAKES NUREG/CP 0143 PROCEEDINGS OF THE THIRD INTERNATIONAL WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR V G

10. AN AUALYSIS OF POTASSIUM IODIDE (KI) PRO-PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-Radiation Exposure CLEAR ACCOENT.

NUREG-0713 V15: OCCUPATIONAL RADIATON EXPOSURE AT COM-MERCIAL NUCLEAR POWER REACTORS AND OTHER UR CR 6 ANALYSIS OF BORON DILUTION IN A FOUR-LOOP NURE 134 U CE I TY A D S SITIVITY ANALYSIS OF PWR CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-COENT CONSEQUENCE MODEL Probabilistic Accident Consequence NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR-6244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty DENT CONSEQUENCE MODEL NU G/C 62 4 ROBABlUSTIC ACCOENT CONSEQUENCE NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-UNCERTAINTY ANALYSIS Dispersion And Depositen Uncertainty DENT CONSEQUENCE MODEL.

Assessment Appendices A And B.

NUREG/CR4244 V03 PROBABlUSTIC ACCOENT CONSEQUENCE N RE 15 6 RFT F MAN EMENT OF RADIOACTIVE MATERIAL s

A e C D,E G SAFETY PROGRAMS AT MEDICAL FACluTIES. Draft Report For Com-Probabilist6c Risk Assessment ment.

NUREG/CR-6116 V09. SYSTEMS ANALYSIS PROGRAMS FOR HANDSON INTEGRATED RELIABlUTY EVALUATIONS (SAPHIRE)

Reactor VERSION 5 0 Venticaton And Vahdaten (V&V) Manual.

NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-NUREG/CR-6141: HANOBOOK OF METHODS FOR RISK-BASED TEST PROGRAM. Draft Report For Comment.

ANALYSES OF TECHNICAL SPECIFICATIONS.

NUREG/CR4143 V06 P1: EVALUATON OF POTENTIAL SEVERE AC-Reactor Accident CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS A1 NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF GRAND GULF. UNIT 1 Evaluaton Of Severe Accident Risks For Plant EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-Operational State $ Dunng A Refueling Outage Man Report And Ap.

DENT CONSEQUENCE MODEL pendices NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR4143 V06 P2. EVALUATION OF POTENTIAL SEVERE AC-FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT DENT CONSEQUENCE MODEL

24 Subject index.

Reactor Preesure Vee.ei Seism 6c Dewen

. NUREG/CR 6214: PRODUCTION AND TESTING OF The VITAMIN B6 NUREG/CR4240 APPLICATION OF BOUNDING SPECTRA TO SEIS-FINE GROUP AND THE BUGLE 93 BROADGROUP NEUTRON /.

MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF PHOTON CROSS SECTON LIBRARIES DERIVED FROM ENDF/B-VI ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO NUCLEAR DATA.

STRONG MOTION EARTHOUAKES.

NUREG/CR 6273. BIAXIAL LOADING EFFECTS ON FRACTURE TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL Sensitivity Analysis i

NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF f

NUR G

-6141: HANDBOOK OF METHOOS FOR RISK-BASED O S OUE EM L

ANALYSES OF TECHNICAL SPECIFICATIONS.

NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR 6311: EVALUATING PREDICTON UNCERTAINTY.

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-Reactor sce6 dent DENT CONSEQUENCE MODEL NUREG/CR 6134; UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR-6136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC.

COENT CONSEQUENCE MODEL' DENT CONSEQUENCE MODEL i

NUREG/CR4311: EVALUATING PREDICTION UNCERTAINTY-Refue66ng Outage NUREG/CR4143 V06 P1: EVALUATON OF POTENTIAL SEVERE AC-Severe Accident CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG 1465: ACCIDENT SOURCE TERMS FOR UGHT-WATER NU-l GRAND GULF, UNIT 1 Evaluaton Of Severe Accident Risks For Plant CLEAR POWER PLANTS.

Operatonal State 5 Dunng A Refuehng Outage Main Report And Ap.

NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-perdices.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-6143 V06 P2-EVALUATION OF POTENTIAL SEVERE AC.

GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Operatonal State 5 Dunng A Refuehng Outage. Main Report And Ap-GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant pendices.

Operatonal State 5 Dunng A Refuehng OutageSupporting MELCOR NUREG/CR4143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC-Calculatons.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Evaluation Of Severe Accident Riska For Plant Regulatory Agende Operational State 5 Dunng A Refuehng Outage. Supporting MELCOR NUREG-0936 V13 NO3: NRC REGULATORY AGENDA Serniannual Calculatons.

ReportJufy-Decernber 1994.

NUREG/CR-6285: SEVERE ACCIDENT NATURAL CIRCULATION Regulatory And Technical Report NUREG 0304 V19 NO3: REGULATORY AND TECHNICAL REPORTS Shutdown Operation 1

(ABSTRACT INDEX JOURNAL). Compilation For Third Quarter NUREG/CR4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-1994 July-September.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG 0304 V19 N04. REGULATORY AND TECHNICAL REPORTS GRAND GULF, UNIT 1.Evaluaton Of Severe Accident Resks For Plant ~

(ABSTRACT INDEX JOURNAL) Annual Compilation For 1994.

Operational State 5 Dunng A Refuehng Outage. Main Report And Ap-pendices.

Reley NUREG/CR4143 V06 P2: EVALUAT!ON OF POTENTIAL SEVERE AC-j NUREG/CR-6220: AN ASSESSMENT OF FIRE VULNERABILITY FOR CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT i

AGED ELECTRICAL RELAYS.

GRAND GULF, UNIT 1.Evaluaton Of Severe Accident Risks For Plant Re46ef Valve Operatonal State 5 Dunng A Refuehng Outage. Supporting MELCOR NUREG/CR 6192: AGING AND SERVICE WEAR OF SPRING-LOADED Calculatons.

PRESSURE RELIEF VALVES USED IN SAFETY RELATED SYSTEMS AT NUCLEAR POWER PLANTS NUR G/CP 0145:

WORKSHOP ON DEVELOPING SAFE Remote Aftertoeding SOFTWARE. Held At Hotel Del Coronado, San Diego,CAJuly 22-NUREG/CR4276. A COMPILATION OF CURRENT REGULATIONS, 23,1992.

i STANDARDS, AND GUOEUNES IN REMO E AFTERLOADING BRA.

NUREG/CR4293 V01: VERIFICATION AND VAUDATION GUIDELINES CHYTHERAPY.

FOR HIGH INTEGRITY SYSTEMS. Main Report NUREG/CR4293 V02: VERIFICATIONN AND VAUDATION GUIDE.

Riek Ano6yone LINES FOR HIGH INTEGRITY SYSTEMS Appendices A-D.

NUREG/CR4311: EVALUATING PREDICTON UNCERTAINTY.

NUREG/CR4316 V01: GUIDELINES FOR THE VERIFICATION AND VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Rulee AL SOFTWARE.

NUREG-0936 V13 NO3. NRC REGULATORY AGENDA Semiannual NUREG/CR4316 V02: GUIDELINES FOR THE VERIFICATION AND ReportJuly-December 1994.

VALCATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-AL SOFTWARE. Survey And Assessrnent Of Conventonal Software IAN Venfication And Validation Methods.

NUREG/CR4116 V09: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4316 V03: GUIDELINES FOR THE VERIFICATION AND HANDS ON INTEGRATED REUA0luTY EVALUATIONS (SAPHIRE)

VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-VERSION 5.0.Venfication And Vahdaten (V8V) Manual-AL SOFTWARE. Survey And Documentaten Of Expert Systern Venfica.

ton And Vahdation Methodologies.

Safety Criterte NUREG/CR-6284. CRITICAUTY SAFETY CRITER;A FOR LICENSE VA DATO X

TS M

AR A NVENTIONAL REVIEW OF LOW LEVEL WASTE FACIUTIES.

SOFTWARE.Evaluaton Of Knowledge Base Certificaten Methods.

gegegygegy, NUREG/CR-6316 V05: GUIDELINES FOR THE VERIFICATION AND NUREG-15th REPORT OF THE SOUTH TEXAS PROJECT ALLEGA.

VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

TlONS REVIEW TEAM.

AL SOFTWARE.Ratonale And Desenpton Of V&V Guidehne Packages And Procedures.

Safety leeuce NUREG/CR4316 V06: GUIDEUNES FOR THE VERIFICATION AND NUREG 1435 SO4: STATUS OF SAFETY ISSUES AT LICENSED VALIDATION OF EXPER1 SYSTEM SOFTWARE AND CONVENTION.

POWER PLANTS.TMI Acton Plan Requirements. Unresolved Safety AL SOFTWARE. Validation Scenanos.

Iseues,Genenc Safety issues.Other Multiplant Action issues.

NUREG/CR4316 V07: GUIDELINES FOR THE VER:FICATON AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Safety Holsted System AL SOFTWARE. User's Manual.

NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-LOADED NUREG/CR4316 V08 GUIDELINES FOR THE VERIFICATION AND PRESSURE RELICF VALVES USED IN SAFETYAELATED SYSTEMS VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AT NUCLEAR POWER PLANTS.

AL SOFTWARE. Bibliography.

Subject index 25 Source Term PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/8-VI NUREG-1465 ACCIDENT SOURCE TERMS FOR LIGHT WATER NU-NUCLEAR DATA _

CLEAR POWER PLANTS Vahdation UR G 1 EPORT OF THE SOUTH TEXAS PROJECT ALLEGA-TONS REVIEW TEAM F

GHI TEGR Y SY EM Ma e

NUREG/CR-6293 V02: VERIFICATIONN AND VALIDATION GUIDE-Storege Cask LINES FOR HIGH INTEGRITY SYSTEMS Appendices A-D.

NUHEG/CR 6242. CASKS (COMPUTER ANALYSIS OF STORAGE NUREG/CR-6316 V01. GUIDELINES FOR THE VERIFICATION AND CASKSt A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-STORAGE CASK DESIGN REVIEW User's Manual To version 1b (In.

AL SOFTWARE.

ciudmg Program Reference)

NUREG/CR-6316 V02: GUIDELINES FOR THE VERIFICATION AND VALOATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Stress Corrosion Cracking AL SOFTWARE. Survey and Assessment Of Conventonal Software NUREG/CR-4667 V18-ENVIRONMENTALLY ASSISTED CRACKING IN Verrfication And Vahdation Methods.

j LIGHT WATER RE ACTORS. Semiannual Reportfctober 1993 - March NUREG/CR-6316 V03: GUIDELINES FOR THE VERIFICATION AND 1994.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Stress Trianlaitty AL SOFTWARE survey And Documentaten Of Expert System Venica-NUREG/CR 6259 CONSTRAINT EFFECTS ON FRACTURE INITIATION ton And Vandaten Methodologies i

LOADS IN HSST WOE-PLATE TESTS-NUREG/CR4316 V04. GUIDELINES FOR THE VERIFICATION AND i

VALIDATION EXPERT SYSTJM SOFTWARE AND CONVENTIONAL Survelitance Requirement SOFTWARE Evaluaton Of Knowledge Base Certifcation Methods.

NUREG/CR-6141. HANDBOOK OF METHODS FOR RISK-BASED NUREG/CR4316 V05: GUIDELINES FOR THE VERIFICATION AND ANALYSES OF TECHNICAL SPECIFICATIONS.

VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-AL SOFTWARE Ratonale And Desenptron Of V8V Guidehne Packages System Safety And Procedures.

NUREG/CP.0145 WORKSHOP ON DEVELOPING SAFE NUREG/CR-6316 J06 GUIDELINES FOR THE VERIFICATON AND SOFTWARE. Held At Hotel Del Coronado, San Diego.CA. July 22 VALIDATION OF EANRT SYSTEM SOFTWARE AND CONVENTON-23.1992 AL SOFTWARE Vahdaen Scenaros.

TLD NUREG/CR-6316 V07: Gt IDELINES FOR THE VER!FICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG 0837 V14 N04 NRC TLD DIRECT RADIATION MONITORING AL SOFTWARE User's Manual.

NETWORK Progress Report October-December 1994 NUREG/CR4316 V06. GUIDELINES FOR THE VERIFICATION AND TMI Act6on Plan VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG 1435 SO4 STATUS OF SAFETY ISSUES AT LICE NSED AL SOFTWARE Bibbography.

POWER PLANTS TMI Acton Plan Requwements. Unresolved Safety lasues.Genenc Safety issues.Other Multiplant Action issues.

Vahdity Limit gg, g

g Tabletop DETERMINATION.An Assessment Of The J(M) Parameter.

NUREG 5514. GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A NUREG/CR 6264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE NUCl. EAR POWER PLANT.

DETERMINATION A Computatonal Approach To Ductde Crack Growth Technical Specifications NUREG/CR-6141: HANDBOOK OF METHODS FOR RISK BASE D Vendor inspection ANALYSES OF TECHNICAL SPECIFICATIONS-NUREG-0040 VIB M*

lCENSEE CONTP ACTOR AND VENDOR IN-Thermohydraul6c S

ON $

c RL Quadwty Repod.Octobe - Decemba NUREG/CR4257. CANDU 3 TRANSIENT ANALYSIS US!NG ATOMIC

• ** E h ENERGY OF CANADA LTD CODES VeMicatu Thermoluminescent Dosimeter NUREG/CR-6293 V01: VERIFICATON AND VALIDATION GUIDELINES NUREG-0837 V14 N04 NRC TLD DIRECT RADIATION MONITORING FOR HIGH INTEGRITY SYSTEMS Main Report.

NETWORK Progress Report 0 tober December 1994.

NUREG/CR4293 V02 VERIFICATIONN AND VAllDATION GUIDE-Title Ust LINES FOR HIGH INTEGRITY SYSTEMS Append ces A-D.

NUREG/CR4316 VO1: GUIDELINES FOR THE VERIFICATON AND NUREGJ A0 V16 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AVAILABLE. November 1 30, 1994 AL SOFTWARE.

NUREG4540 V16 N12 TITLE LIST OF DOCUMENTS MADE PUBLICLY NUREG/CR4316 V02: GUIDELINES FOR THE VERIFICATON AND NURE 4 V N T L LS OF DOCUMENTS MADE PUBLICLY VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AVAILABLE January 1 31,1995 AL Ehey And Assessnmnt O Conventonal Sohare Venficaton And Validaten Methods.

Topical Report NUREG/CR4316 V03' GUIDELINES FOR THE VERIFICATION AND NUREG4390 V09 Not: TOPfCAL REPORT REVIEW STATUS.

VALIDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

Trans6ent Analysis AL SOFTWARE Survey And Documentaten Of Expert System Venfica-t on And Vahdaten Methodologies.

NUREG/CR 6257. CANDU 3 TRANS!ENT ANALYSIS USING ATOMIC NUREG/CR-6316 VD4. GUIDELINES FOR THE VERIFICATION AND ENERGY OF CANADA LTD CODES.

VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL Uncertainty Analysis SOFTWARE Evaluaton Of Knowledge Base Certification Methods.

NUREG/CR-6244 VO1: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-6316 V05 GUIDELINES FOR THE VERIFICATION AND UNCERTAINTY ANALYSIS. Disperson And Deposrton Uncertainty VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Assessment Main Deport AL SOFTWARE.Ratonale And Desenpton Of V8V Guedehne Packages NUREG/CR 3244 V02. PROBABILISTIC ACCIDENT CONSEOUENCE NU EG/ R 16 V06. GUIDELINES FOR THE VERIFICATON AND And Npos%n Uncedainty n

s s e A And a VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

NUREG/CR-6244 V03. PRO"'ABILISTIC ACCIDENT CONSEOUENCE AL SOFTWARE Vahdaten Scenanos.

I UNCERTAINTY ANALYSIS Disperseon and Deposton Uncertainty NUREG/CR4316 V07: GUIDELINES FOR THE VERIFICATION AND Assessment Appendces C.D.E.F.G.H VALIDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AL SOFTWARE. User's Manuel j

VITAMIN 86 NUREG/CR-6316 V08 GUIDELINES FOR THE VERIFICATON AND l

NURE G/CR 6214 PRODUCTON AND TESTING OF THE VITAMIN-86 VALIDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-l FINE GROUP AND lHE BUGLE.93 BROAD-GROUP NEUTRON /

AL SOFTWARE Bibhography.

l

26 Subject Indax Vessel NUREG/CR-6291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

NUHEG/CR4260 APPUCATION OF NUREG/CR-5999 INTERIM FA-kgREG CR4291 V03 NUCLEAR PLANT ANALYZER. Computer Visual TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-NENTS NUdstem Reference Manual. NUCLEAR PLANT ANALYZER Programmer's S

a EG/CR-6291 V04:

Visual Display NUREG/CR-6291 V01: NUCLEAR PLANT ANALYZER 1nstal:ation WLddlate Manual.

NUREG/CR 6259: CONSTRAINT EFFECTS ON FRACTURE INIT'ATION LOADS IN HSST WIDE-PLATE TESTS.

~ w w'

e)

NRC Originating Organization Index (Staff Reports)

This index lists those NRC organizations that have published staff reports. The index is ar-ranged alphabetically by major NRC organizations (e.g., program offices) and then by sub-sections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

NUREG-1514. GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR REGION 1 (POST 820201)

A NUCLEAR POWER PLANT.

NUREG 0837 V14 N04 NRC TLD DIRECT RADIATION MONITORING NETWORK ProQ'ess Report October-December 1994.

EDO - OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS OFC OF ENFORCEMENT (POST 870413)

OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG 0940 V13 N4 P1: ENFORCEMENT ACTIONS SIGNIFICANT NUREG-0430 V14.

LICENSED FUEL FACILITY STATUS ACTIONS RESOLVED REACTOR LICENSEES.Quarterty Progress REPORT Inventory Dfference Data. July 1.1993 June 30,1994.

Report. October-December 1994 DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST NUREG-0940 Vt3 N4 P2-ENFORCEMENT ACTIONS SIGNIFICANT 870729)

ACTIONS RESOLVED MEDICAL LICENSEES Quarterty Progress NUREG 1516 DRFT FC. MANAGEMENT OF RADIOACTIVE MATERI-Report, October December 1994 AL SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For NUREG-0940 V13 N4 P3 ENFORCEMENT ACTIONS SIGNIFiCANT Comment ACTIONS RESOLVED MATERIAL LICENSEES (NON-MEDICAL) Quarterbr Prooress Report October December 1994 U.S. NUCLEAR REGULATORY COMMISSION OFC OF INVESTIGATIONS (>OST 880201)

NRC NO DETAILED AFFILIATION GIVEN NUREG 1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA.

NUREG/CR-6244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE TIONS REVIEW TEAM.

UNCERTAINTY ANALYSIS. Dapersion and Deposition Uncertavnty Assessment Appendices C.D.E.F,G,H.

EDO OFFICE OF ADMINISTRATION (PRE DIVISION Of FREEDOM OF INFORMATION & PUBLICATIONS 870413 & POST 890205)oERV-EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

ICES (POST 940714 DIVISION OF REGULATORY APPLICATIONS (POST 941217)

NUREG 0304 V19 NO3 REGULATORY AND TECHNICAL REPORTS NUREG 0713 V15: COCUPATIONAL RADIATION EXPOSURE AT (ABSTRACT INDEX JOURNAL) Compilation For Third Quarter COMMERCIAL NUCLEAR POWER REACTORS AND OTHER 1994.Juiv September.

FACILITIES,1993 Twenty-Sixth Annual Report.

NUREG 0304 V19 N04 REGULATORY AND TECHNICAL REPORTS NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK.

(ABSTRACT INDEX JOURNAL). Annual Compilation For 1994 TEST PROGRAM Draft Report For Comment NUREG4540 V16 N11 TITLE LIST OF DOCUMENTS MADE PUBLIC.

DIVISION OF SYSTEMS TECHNOLOGY (POST 941217)

LY AVAILABLE November 1 30,1994.

NUREG-0700 R01 DFC: HUMAN SYSTEM INTERFACE DESIGN NUREG-0540 V16 N12. TITLE LIST OF DOCUMIENTS MADE PUBLIC.

REVIEW GUIDELINE. Draft Report For Comment.

LY AVAILABLE December 1-31,1994 NUREG 1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER NU-NUREG-0540 V17 N01: TITLE LIST OF DOCUMENTS MADE PUBLIC.

CLEAR POWER PLANTS.

LY AVAILABLE January 1 31.1995 PROBABILISTIC RISK ANALYSIS BRANCH (POST 941217)

NUREG4750 V40102' INDEXES TO NUCLEAR REGULATORY COM.

NUREG/CR-6244 V01: PROBAslLISTIC ACCIDENT CONSEQUENCE MISSION ISSUANCES Juiv-December 1994 UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty NUREG-0750 V40 N05. NUCLEAR REGULATORY COMMISSION IS.

Assessment. Main Report.

SUANCES FOR NOVEMBER 1994 Pages 169 318 NUREG/CR-6244 V02 PROBABILISTIC ACCIDENT CONSEQUENCE NUREG-0750 V40 N06 NUCLEAR REGULATORY COMMISSION IS-UNCERTAINTY ANALYSIS. Dspersson And Deposition Uncertainty SUANCES FOR DECEMBER 1994 Pa0es 319-367.

Assessment. Appendices A And B.

NUREG-0750 V41 N01: NUCLEAR REGULATORY COMMISSION iS-REACTOR & PLANT SYSTEMS BRANCH (POST 941217)

SUANCES FOR JANUARY 1995 Pages 1-69 NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC NUREG4936 V13 NO3. NRC REGULATORY AGENDA Semiannual ENERGY OF CANADA LTD CODES.

EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428)

EDO - OFFICE OF THE CONTROLLER (PRE 820418 & POST 890205)

OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001)

DIVISION OF BUDGET & ANALYSIS (POST 890205)

NUREG-0040 V18 N04 LICENSEE CONTRACTOR AND VENDOR IN-NUREG-1100 V11. BUDGET ESTIMATES Fiscal Years 1996 1997.

SPECTION STATUS REPORT. Quarterly Report. October - December 1994 (White Book)

EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG-0390 V09 N01: TOPICAL REPORT REVIEW STATUS DATA NUREG-1435 SO4-STATUS OF SAFETY ISSUES AT LICENSED OFFICE FOR ANALYSIS 4 EVALUATION OF OPERATIONAL DATA, DI-POWER PLANTS TMI Acton Plan Requirements. Unresolved Safety RECTOR issues.Genenc Safety lasues.Other Multiplant Action Issues.

NUREG-0090 V17 NO3 REPORT TO CONGRESS ON ABNOnMAL NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-OCCURRENCES July September 1994 TIONS REVIEW TEAM.

r 27

NRC Originating Organization index (International Agreements)

This iridex lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, brancies) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

There were no NUREG/lA reports put*shed during thie querier f

a 29

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) cnd 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-ntion. If further information is needed, refer to the main citation by the NUREG/CR number.

EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF DATA EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-DIVISION OF SAFETY PROGRAMS (POST 870413)

DENT CONSEQUENCE MODEL.

NUREG/CR4266. ANALYSIS OF BORON DILUTION IN A FOUR-NUREG/CR-6136 UNCERTAINTY AND SENSITIVITY ANALYSIS OF LOOP PWR-FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS DENT CONSEOUENCE MODEL'UATION OF POTENTIAL SEVERE NUREG/CR-6143 V06 P1: EVAL DIVISION OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-A A

A

. UNIT 1. Evaluation Of Severe Accident NU G/ R4276. A COMPILATION OF CURRENT REGULATIONS s

ant Opwatonal SWe 5 Dunng A Refuehng OutageMain STANDARDS. AND GUOELINES IN REMOTE AFTERLOADING BRACHYTHERAPY Report And Appendices.

43 ON & WNM WE DIVISION OF WASTE MANAGEMENT (NMS$ 940403)

NUREG/CR 6284 CRITICALITY SAFETY CRITERIA FOR LICENSE ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-REVIEW OF LOW-LEVEL WASTE FACILITIES ATIONS AT GRAND GULF, UNIT 1.Evalualen Of Severe Accident Risks For Plant Operabonal State 5 Dunng A Refuehng EDO = OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

Outage Supporting MELCOR Calculatons.

DIVISION OF SYSTEMS RESEARCH (860717-941217)

NUREG/CR-6244 VO1: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR 4141: HANDBOOK OF METHODS FOR RISK BASED UNCERTAINTY ANALYSIS. Dispersion And Deposrnon Uncertainty ANALYSES OF TECHNICAL SPECIFICATIONS.

Assessment Main Report.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 941217)

NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-6116 V09 SYSTEMS ANALYSIS PROGRAMS FOR UNCERTAINTY ANALYSIS. Dispersion And Deposton Uncertainty HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

Assessment. Appendices A And B.

VER$lON 5 0 Venfication And Vahdaten (V&V) Manual.

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-6259. CONSTRAINT EFFECTS ON FRACTURE INITI-UNCERTAINTY ANALYSIS. Disperson and Depositen Uncertainty ATiON LOADS IN HSST WIDE-PLATE TESTS Assessment Appendices C.D.E.F.G,H DIVISION OF ENGINEERING TECHNOLOGY (POST 941217)

NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC NUREG/CR 4667 V18. ENVIRONMENTALLY ASS!STED CRACKING ENERGY OF CANADA LTD CODES IN LIGHT WATER REACTORS. Semiannual Report,0ctober 1993 -

NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION March 1994.

STUDIES AT THE INEL.

NUREG/CR 5462; AGING STUDY OF BOILING WATER REACTOR NUREG/CR-6291 V01: NUCLEAR PLANT ANALYZER insta!tation HIGH PRESSURE INJECTION SYSTEMS Manual.

NUREG/CR-5591 V03. HEAVY-SECTION STEEL IRRADIA TION NUREG/CR-6291 V02: NUCLEAR PLANT ANALYZER.Anaiyzer Refer.

PROGRAM Progress Report For October 1991 September 1992 ence Manual.

NUREG/CR-5591 V04 N2. HEAVY SECTION STEEL IRRADIATION NUREG/CR4291 V03: NUCLEAR PLANT ANALYZER Computer Visual PROGRAM Semiannual Progress Report For Apni-September 1993 System Reference Manual.

NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-NUREG/CR-6291 V04. NUCLEAR PLANT ANALYZER. Programmer's LOADED PRESSURE RELIEF VALVES USED IN SAFETY-RELATED Manual SYSTEMS AT NUCLEAR POWER PLANTS NUREG/CR-6293 V02: VERIFICATIONN AND VALIDATION GUIDE-NJREG/CR 6214. PRODUCTON AND TESTING OF THE VITAMIN-B6 LINES FOR HIGH INTEGRITY SYSTEMS Appendices A-D.

FINE GROUP AND THE BUGLE 93 BROAD GROUP NEUTRON /

NUREG/CR 6310: AN ANALYSIS OF POTA$$1UM IODIDE (Kl) PRO-PHOTON CROSS-SECTON LIBRARIES DERIVED FROM ENDF/B.

PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU.

VI NUCLEAR DATA.

CLEAR ACCIDENT.

NUREG/CR 6220: AN ASSESSMENT OF FIRE VULNERABILITY FOR NUREG/CR4311: EVALUATING PREDICTION UNCERTAINTY.

AGED ELECTRICAL RELAYS NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION AND NUREG/CR-6240. APPLICATION OF BOUNDING SPECTRA TO SEIS-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF TIONAL SOFTWARE.

ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO NUREG/CR-6316 V02: GUIDELINES FOR THE VERIFICATON AND STRONG MOTION EARTHOUAKES VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-NUREG/CR4242: CASKS (COVPUTER ANALYSIS OF STORAGE TIONAL SOFTWARE. Survey And Assessment Of Conventonal Soft-CASKS). A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR were Venfication And Validation Methods.

STORAGE CASK DESIGN REVIEW. Users Manual To Verson ib NUREG/CR4316 V03: GUIDELINES FOR THE VERIFICATON AND

('ncluding Program Reference)

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-NUREG/CR4264 V01: VALIDITY LIMITS IN J-RESISTANCE CURVE TIONAL SOFTWARE. Survey And Documentation Of Expert System DETERMINATION An Assessment Of The J(M) Parameter.

Venficat on And Vahdation Methodologies.

NUREG/CR4264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE NUREG/CR4316 V04: GUCELINES FOR THE VERIFICATION AND DETERMINAtlON A Computational Approach To Ductile Crack VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTONAL Growth Under Large Scale Yieldmg Conditions SOFTWARE Evaluation Of Knowledge Base Certificaton Methods.

NUREG/CR 6273. BIAXtAL LOADING EFFECTS ON FRACTURE NUREG/CR4316 V05: GUCELINES FOR THE VERIFICATION AND TOUGHNESS OF RE ACTOR PRESSURE VESSEL STEEL.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-DIVISON OF REGULATORY APPLICATIONS (POST 941217)

TIONAL SOFTWARE.Ratonale And Desenption Of V&V Guidehne NUREG/CR-5927 V02: EVALUATON OF A PERFORMANCE AS-Packages And Procedures.

SESSMENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE NUREG/CR4316 V06: GUIDELINES FOR THE VERIFICATION AND WASTE DISPOSAL FACILITIES Vahdaton Needs VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-OfVISION OF $YSTEMS TECHNOLOGY (POST 941217)

TIONAL SOFTWARE Vahdation S:enanos.

NUREG/CR-6134 UNCERTAINTY AND SENSITIVITY ANALYS:S OF NUREG/CR-6316 V07. GUIDELINES FOR THE VERIFICATON AND CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN.

ACCIDENT CONSEQUENCE MODEL.

TONAL SOFTWARE User's Manual.

31

32 NRC Contract Sponsor index NUREG/CR4316 V06 GUIDELINES FOR THE VERiriCATION AND EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428)

VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001)

NUREG/CR4260: APPLICATION OF NUREG/CR-5999 INTERIM FA-TiONAL SOFTWARE.Behography.

TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COM-PONENTS.

..a,.-

Contractor Index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation, Listed below each contractor are the NUREG/CR numbers and titles of their reports, if further information is needed, refer to the main citation by the NUREG/CR number.

ORcONNE NATIONAL LABORATORY NUREG/CR 6244 V02'- PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-4667 V18' ENVIRONMENTALLY ASSISTED CRACKING IN UNCERTAINTY ANALYSIS Dspersion And Depositen Uncertaintv LIGHT WATER REACTORS Semiannual Report. October 1993 - March Assessment.Appendeces A And B.

1994.

NUREG/CR4266. ANALYSIS OF BORON DILUTON IN A FOUR-LOOP GRAM, INC.

PWR.

NUREG/CR 6134 UNCERTAINTY AND SENSITIVITY ANALYSIS OF CD6W E M M M MM BUM &

ARIZONA STATE UNIV., TEMPE, A2 NUREG/CH 6134. UNCERTAINTY AND SENSITIVITY ANALYSIS OF NU EG CR 35 UN RT AND SENSITIVITY ANALYSIS OF CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-COENT CONSEQUENCE MODEL NUREG/CR 6135 UNCERTAINTY AND SENSITIVITY ANALYSIS OF DENT CONSEQUENCE MODEL.

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI.

NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF DENT CONSEQUENCE MODEL FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CR-6136. UNCERTAINTY AND SENSITIVITY ANALYSIS OF DENT CONSEQUENCE MODEL FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI.

NUREG/CR4143 V06 P1: EVALUATON OF POTENTIAL SEVERE AC.

DENT CONSEQUENCE MODEL COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NUREG/CR-6244 VC1. PROBABILISTIC ACCOENT CONSEQUENCE GRAND GULF, UNIT 1.Evaluaten Of Severe Accident Risks For Plant UNCERTAINTY ANALYSIS Otspersion And Depositen Uncertainty Operatonal State 5 Dunng A Refueling OutageMain Report And Ap-Assessment Main Report pendices.

NUREG/CH-6244 V02: PROBABILISTIC ACCIDENT CONSEOUENCE UNCERTAINTY ANALYSIS. Osperson And DepoMon Uncertainty HARRIS CORPORATION INFORMATION SYSTEMS Assessment Appendices A And B.

NUREG/CR4293 V02: VERIFICATIONN AND VALtDATION GUIDE-NUREG/CR4244 V03. PROBABILISTIC ACCIDENT CONSEQUENCE LINES FOR HIGH INTEGRITY SYSTEMS.Appendees A-D.

UNCERTAINTY ANALYSIS Dispersion and Deposition Uncertainty Assessment Appendices C.D.Ef,0,H HARVARD SCHOOL OF PUBLIC HEALTH, BOSTON, MA NUREG/CP-0141: PROCEEDINGS OF THE 23RD DOE /NRC NUCLEAR AVAPLAN OY (FINLAND)

AIR CLEANING CONFERENCE. Held in Buffalo,New York. July 25-NUREG/CR4141: HANDBOOK OF METHODS FOR RISK. BASED 28,M ANALYSES OF TECHNICAL SPECIFICATONS.

BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES HARVARD UNIV, CAMBRIDGE, MA NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK.

NUREG/CH-6264 V02: VALIDITY LIMITS IN J RESISTANCE CURVE TEST PROGRAM Draft Report For Comment.

DETERMINATION.A Computahonal Approach To Ductste Crack Growth NUREG/CR-6264 V01: VAUDITY LIMITS IN J-RESISTANCE CURVE Under Lage-Scale Yelding Conditons.

DETERMINATION An Assessment Of The JIM) Parameter NUREG/CR4264 V02. VAUDITY LIMITS IN J-RESISTANCE CURVE HAWAll, UNIV. 0F, HILO, HI DETERMINATION A Computatonal Approach To Ductile Crack Growth NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE Under Large Scale Yekkng Conditions.

UNCERTAINTY ANALYSIS. Daperson And Deposition Uncertainty Assessment. Main Report.

N EG 0 R1 DFC HUMAN SYSTEM INTERFACE DESIGN UNCERTAINTY ANALYSIS. Dispersson And Depositon Uncertainty RFVIEW GUIDELINE Draft Report For Comment.

NUREG/CP 0143 PROCEEDINGS OF THE THIRD INTERNATONAL AssessmentAppendices A And &

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR POWER PLANTS Held At Hauppauge, L island, New York UNCESTAINTY ANALYSIS. Disperson and Deposition Uncertainty NUREG/CR 6141; HANDBOOK OF ME S FOR RISK-BASED Assessment.Apo-hces C.D.E.F,G.H.

ANALYSES OF TECHNICAL SPECIFICATIONS.

IDAHO NATIONAL ENGINEL31NG LABORATORY BROWN UNTY, PROVIDENCE, RI NUREG/CR-5462: AGING MUDY OF BOILING WATER REACTOR NUREG/CR4264 V01: VALIDITY tiMITS IN JRESISTANCE CURVE HIGH PRESSURE INJECTON SYSTEMS.

DETERMINATION An Assessment Of The J(M) Parameter.

NUREG/CR4116 V09. SYSTL*VS ANALYSIS PROGRAMS FOR NUREG/CR4264 V02: VAllDITY LIMi1S IN J-RESISTANCE CURVE HANDS-ON INTEGRATED RELaBILITY EVALUATONS (SAPHIRE)

DETERMINATION A Computational Approach To Ductile Crack Growth VERSION 5 0 Verifcation And Valc&,:. (V&V) Manual Under Large-Scale Vie 6 ding Conditons.

NUREG/CR4257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC ENERGY OF CANADA LTD CODES.

^

NUR G/CR 7 A COMPILATON OF CURRENT REGULATONS STANDARDS. AND GUIDELINES IN REMOTE AFTERLOADING BRA $

E S CHYTHERAPY.

NUREG/CR4285: SEVERE ACCCENT NATURAL CIRCULATON GERMANY, DEMOCRATIC REPUBLIC STUDIES AT THE INEL NUREG/CR 6244 V03 PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4291 V01: NUCLE *9 PLANT ANALYZER.installaton es D.E NURE CR-6291 V07_ NUCLEAR PLANT ANALYZER. Analyzer Refer-ence Manual.

GERMANY, FEDERAL REPUBLIC OF NUREG/CR4291 V03: NUCLEAR PLANT ANALYZER. Computer Visual NUREG/CR4244 V01 PROBABILISTIC ACCIDENT CONSEQUENCE System Reference Manual UNCERTAINTY ANALYSIS Disperson And Depositen Uncertainty NUREG/CR4291 V04 NUCLEAR PLANT ANALYZER. Programmer's Assessment Main Repoit Manual.

l l

33 ii

l 1

34 Contractor Index fLLINOIS, STATE OF NUREG/CR 6135: UNOERTAINTY AND SENSITIVITY ANALYSIS OF NUREG-1516 DHFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-SAFETY PROGRAMS AT MEDICAL FACluTIES Draft Report For Com-DENT CONSEQUENCE MODEL.

ment NUREG/CR6136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF FOOD PATHWAY RESULTS WITH THE MACCS FIEACTOR ACCI-lLLINOIS, UNIV. OF, URSANA, IL NUREG/CR4259. CONSTRAINT EFFECTS ON FRACTURE INiTIATON DENT CONSEQUENCE MODEL.

LOADS IN HSST WIDE. PLATE TESTS.

NUREG/CR-6143 V06 P1: EVALUATON OF POTENTIAL SEVERE AC-COENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT LAWRENCE LIVERMORE NATIONAL LA80RATORy GRAND GULF,0 NIT 1.Evaluaton Of Severe Accident Risks For Plant NUREG/CP4145 WORKSHOP ON DEVELOPING SAFE Operatonal State 5 Dunng A Refuelmg Outage Masn Report And Ap-SOFTWARE. Held At Hotel Del Coronado, San Dego.CA luly 22 pendices 23.1992 NUREG/CR4143 V06 P2: EVALUATON OF POTENTIAL SEVERE AC.

NUREG/CR4242: CASKS (COMPUTER ANALYSIS OF STORAGE COENTS DUR:NG LOW POWER AND SHUTDOWN OPERATONS AT CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR GRAND GULF, UNIT 1 Evaluation Of Severe Accident Risks for Plant STORAGE CASK DESIGN REVIEW tiser's Manual To Versson Ib (In-Operational State 5 Dunng A Refuehng Outage. Supporting MELCOR ctudmg Program Refererce).

Calculations.

LOS ALAMOS NATIONAL LABORATORY

-622& M ASSNEM & ME WNW M NUREG/CR 6135. UNCERTAINTY AND SENSITIVITY ANALYSIS OF AGED ELECTRICAL RELAYS.

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-NUREG/CR4244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE DENT CONSEQUENCE MODEL.

UNCERTAINTY ANALYSIS. Dsperson And Depos! ton Uncertanty NUREG/CR-6214. PRODUCTION AND TESTING OF THE VITAMIN-86 Assessment. Main Report.

FINE GROUP AND THE BUGLE-93 BROADGROUP NEUTRON /

NUREG/CR-6244 V02. PROBABILISTIC ACCOENT CONSEQUENCE PHOTON CROSS-SECTON UBRARIES DERIVED FROM ENDF/B-VI UNCERTAINTY ANALYS!S. Dspersion And Depositen Uncertamty NUCLE AR DATA.

Assessment.Appendces A And B.

NUREG/CR 4244 VOI: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4244 V03: PROBABILtSTIC ACCIDENT CONSEQUENCE UNCERTAINTY AN6 LYSIS. Osperson And Depos4 tion Uncertainty UNCERTAINTY ANALYSIS. Dsperson and Depositen Uncertainty NU GfCR 62 VO PROBABIUSTIC ACCOENT CONSEQUENCE UNCERTAINTY ANALYSIS Dsperson And Depositen UncertamtY SCIENCE S ENGINEERING ASSOCIATES,lNC.

NUF[G C 2

V B B USTIC ACCOENT CONSEQUENCE NUREG/G4220: AN ASSESSMENT & ME WEMBW M UNCERTAINTY ANALYSIS Dsperson and Deposition Uncertainty AGED ELECTRICAL RELAYS.

NU GC 3 LA N P ICTION UNCERTAINTY, SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY SCIENCE APPLICATIONS, NETHERLANDS, GOVT. OF NUREG-0713 V15 OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE MERCIAL NUCLEAR POWER REACTORS AND OTHER UNCERTAINTY ANALYSIS. Osperson And Deposton Uncertamty FACluTIES 1993 Twenty-Sixth Annual Report.

Assessment Main Fleport NUREG/CR 6141: HANDBOOK OF METHODS FOR RISK. BASED NUREG/CR-6244 V02. PROBABluSTIC ACCIDENT CONSEQUENCE ANALYSES OF TECHNICAL SPECIFICATIONS.

UNCERTA!NTY ANALYSIS. Dsperson And Depositen Uncertainty NUREG/CR4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC.

NUF$/CR 424 V3 OB B USTIC ACCOENT CONSEQUENCE COENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT VNCERTAINTY ANALYSIS. Osperson and Depositen Uncertanty GRAND GULF, UNIT 1 Evaluation Of Severe Accident Riske For Plant Assessment Appendices C.D.E F.G,H.

Operatonal State 5 Dunng A Refueling OutageMain Report And Ap-pendees.

OAK RIDGE NATIONAL LABORATORY NUREG/CR4316 V01: GUOELINES FOR THE VERIFICATION AND NUREG 1514. GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A VALOATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUCLEAR POWER PLANT.

AL SOFTWARE.

NUREG/CR 5591 V03. HEAVY SECTION STEEL IRRADIATON NUREG/CR6316 V02: GUOEUNES FOR THE VERIFICATON AND PROGRAM Progress Report For October 1991. September 1992.

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR4591 V04 N2: HEAVY SECTON STEEL IRRADIATION AL SOFTWARE. Survey And Assessment of Conventional Software M AMS'E WE OF S GL DED Venication And Vahdation Methods.

NU E / 6 PRESSURE REUEF VALVES USED IN SAFETY-RELATLD SYSTEMS NUREG/CR 63t6 V03: GUCELINES FOR THE VERIFICATION AND AT NUCLEAR POWER PL ANTS VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR4214 PRODUCTION AND TESTING OF THE VITAMIN 06 AL SOFTWARE Survey And Documentation Of Expert System Ventca-FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

ton And Validation Methodologies.

PHOTON CROSS.SECTION UBRARIES DERIVED FROM ENDF/B-VI NUREG/CR4316 V04. GUOELINES FOR THE VERIFICATION AND NUCLEAR DATA.

VAUDATON EXPERT SYSTEM SOFTWARE AND CONVENTONAL NUREG/CR4240 APPUCATON OF BOUNDING SPECTRA TO SEIS-SOFTWARE Evaluation Of Knowledge Base Certifcaton Methods MO DESIGN OF PIPING BASED ON THE PERFORMANCE OF NUREG/CR6316 V05: GUOELINES FOR THE VERIFICATION AND ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-N

/C 59 O ST T

ECTS ON FRACTURE INITIATON p

y NUREG/CR6316 V06 GUIDEUNES FOR THE VERIFICATION AND N E C 62 I

LOAD EFFECTS ON FRACTURE TOUGHNESS OF REACTOR PRESSURE VESSFL STEEL VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR4284-CRITICAUTY SAFETY CRITERIA FOR UCENSE AL SOFTWARE.Vahdaten Scenares.

REVIEW OF LOW LEVEL WASTE FACluTIES.

NUREG/CR6316 V07: GUIDEUNES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-S. COHEN & ASSOCIATES. INC-AL SOFTWARE User's Manuat NUREG 1493 DFC-PERFORMANCE-BASED CONTAINMENT LEAK-NUREG/CR4316 V08 GUCEUNES FOR THE VERIFICATON AND TEST PROGRAM Draft Report For Comrnent VALOATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR 6310: AN ANALYSIS OF POTASSIUM ODOE (KI) PRO-AL SOFTWARE B%ography.

PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-CLEAR ACCIDENT-SCIENTECH, INC.

SANDIA NATIONAL LASORATORIES NUREG/CR6310. AN ANACYSIS OF POTASSIUM ODOE (Kl) PRO-NUREG/CR 5927 V02. EVALUATON OF A PERFORMANCE ASSESS, PHYLAXIS FOR THE GENERAL PUBUC IN THE EVENT OF A NU-MENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE WASTE CLEAR ACCIDENT.

DISPOSAL FACIUTIES Vahdaten Needs NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF SOH AR, 8NC.

CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-NUREG/CR-6293 V02. VERIFICATIONN AND VAUDATION GUOE-CIDENT CONSEQUENCE MODEL UNES FOR HIGH INTEGRITY SYSTEMS Append ces A-D.

Contractor Index 35 SOUTHWEST POWER CONSULTANTS,INC.

UNITED KINGDOM NUREG-1493 DFC. PERFORMANCE BASED CONTAINMENT LEAK.

NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE TEST PROGRAM Draft Report For Comment.

UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty Assessment. Main Report STEVENSON & ASSOCIATES NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-6240 APPLICATION OF BOUNDtNG SPECTRA TO SELS.

As s Appe es A And 8 MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO UNCERTAINTY ANALYSIS. Disperson and Depositon Uncertainty STRONG MOTION EARTHQUAKES.

Assessment. Appendices C.D.E.F G.H.

a

international Organization index This index lists, in alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed below each country and per-forming, organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number.

e o

I 37

I!!I

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 Gnd followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

$2@5 CANDU 3U AECL Technologes. Inc.,

NUREG/CR4257 STN B498 South Texas Propct. Und 1. Hae Lghtmg a NUREG 1517 50 416 Grand Gun Nudear Stahon. Und 1. Wesespp NUREG/CR4143 V06 P1 Power Co.

Power & Ls0hl Co STNM499 Scuth Texas Propet. Und 2, Houston Lightmg 4 NUREG-1517 E416 Grand Gulf Nudear S'abon. Urat 1. Massapp NUREG/CR4143 V06 P2 Power Co.

Power & Lgit Co, 39

NRC FORM 336 U.S. NUCLEAR REGULATORY COAAASSION

1. REPORT NUMBER (2-49)

Assypnad try NRC, Add Vol.,

NRCM 1102, dDD a Rev., and Addendum Num-3201.3202 BIBLIOGRAPHIC DATA SHEET

• * * - " *"Y 3 ca.e wwtructions on in. r. vers.)

NUREG-0304 V 1. 20, No.1

a. TnLe AND suorirte

3. DA1E REPORT PUBLISHED Regulatory and Tbchnical Reports (Abstract Index Journal)

MONTH YEAR Compilation for First Quarter 1995 July 1995 January - March

e. nN OR oRANT NUueER 6.AUTHORtS)

8. TYPE OF REPORT Reference

7. PEROD COVERED (inclusive Dates)

January-March 1995

8. PERFORMtv0 ORGANIZATION - NAME AND ADDHESS (If NHC, provide Division Orfee or Rege, U.S. Nuclear Regulatory Commission, and mamna addr..s; is contr.ctor, prova nam..nd mamna addr.ss.)

Division of Freedom of Information and Publications Services Office of Administrafion U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 S. LPONSOHING ORGANIZATON - NAME AND ADDRESS (if NRC, type *$ame as above"; if Contractor, provkle NRC Divlsion. Offce or Roge, U.S. Nuclear A ogulatory Commission, and mailing address.)

Same as 8, above.

10. $UWLEMENT AHY NOTES

11. ABSTRACT (200 words or less)

'Ihis journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceed.

ings of conferences and workshops; as well as international agreement reports. The entries in this compilation are indexed for access by title and abstract, secondary report num'ocr, personal author, subject, NRC organization for staff and international agreements, contractor, international organization, and licensed facility.

12. KEY WORDS/DESCRIPTORS (List words or phrases that wiH assist researchers in locatin0 the report.)

13. AVAILABILITY STATEMENT Unlimited Ia. SECURITY CLASSIRCATION compilation (Th P'8')

abstract index Unclassified (This Report)

Unclassified

15. NUMBER OF PAGES

16. PHiCE j NRe FORM 335 (2-89)

Printed on recycled paper Federal Recycling Program

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