ML20112B239

Regulatory and Technical Reports (Abstract Index Journal). Annual Compilation for 1995

ML20112B239
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Issue date:	04/30/1996
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NUREG-0304, NUREG-0304-V20-N04, NUREG-304, NUREG-304-V20-N4, NUDOCS 9605220248
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NUREG-0304 Vol. 20, No. 4 i

Regulatory and Technical Reports (Abstract Index Journal}

l Annual Compilation for 1995 U.S. Nuclear Regulatory Commission Office of Administration i

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

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 Govemment 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, joumal 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-ference 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 they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018-3308.

A year's subscription of this report consists of four quarterly issues.

4

NUREG-0304 Vol. 20, No. 4 Regulatory and Technical Reports (Abstract Index Journal) i Annual Compilation for 1995 i

i' Date Published: April 1996 J

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M. Sheehan, Project Manager i

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Regulatory Publications Branch Division of Freedom ofInformation and Publications Services Omce of Administration U.S. Nuclear Regulatory Commission

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Washington, DC 20555-0001 b

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CONTENTS i

Preface.............

v index Tab

................ 1 Main Citations and Abstracts..................

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 4

l Subject Index NRC Originating Organization index (Staff Reports).

5 NRC Originating Organization Index (International Agreements).

6 7

l NRC Contract Sponsor Index (Contractor Reports)

Contractor index......

8 International Organization index 9

10 Licensed Facility Index l

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

Technical Publications Section Publications Branch Division of Freedorn of Information and Publications Services T-6 E7 U.S. Nuclear Regulatory Commission l

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:

l Secondary Report Number Index Personal Authorindex l

Subject index NRC Originating Organization Index (Staff Reports)

NRC Originating Organization Index (International Agreements)

NRC Contract Sponsor Index (Contractor Reports)

Contractor Index l

International Organization Index l

Licensed Facility index l

l A detailed explanation of the entries precedes each index.

l The bibliographic elements of the main citations are the following:

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

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

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

1 Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND REUABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National Laboratory. May 1981.141 pp.

8105280299. ANL-81-3. 08632:070.

Where the entries are (1) report number, (2) report title, (3) report author, (4) organ:7ation 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 internal use).

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

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

Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the N RC 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).

Grant Report NUREG/GR-0013: APPLICATIONS OF A NEW MAGNETIC MONITORING TECHNIQUE TO (N SITU EVALUA-TION OF FATIQUE DAMAGE IN FERROUS COMPONENTS. JILES, D.C.; BINER, S.B.; GOVINDARAJU, M.; et al. Iowa State Univ., 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 publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Contral Sys-tem accession number, (8) the report number of the originating organization (if given), (9) the microfiche ad-dress (for NRC internal use).

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

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

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

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

ADD

- addendum APP

- appendix DRFT - draft ERR

- errata N - number R - revision S - supplement V - volume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the NationalTechnicallnformation Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a chect; 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.

vi

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

Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established codes such as 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 Division of Freedom of Information and Publications Services.

vii

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Main Citations and Abstracts 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-national agreement report. The bibliographic information (see Preface for details) is followed by a brief abstract of this report, NUREG-0020 V19: LICENSED OPERATING REACTORS STATUS tion, that have been distributed to the inspected organizations

SUMMARY

REPORT. Data As Of December 31, 1994.(Gray during the period from July through September 1995.

Book l) HARTFIELD.R.A. Office of information Resources Man-ayant (Post 890205). Apnl 1995. 350pp. 9504260255.

NUREG-0090 V17 NO3: REPORT TO CONGRESS ON ABNOR-83635:001.

MAL OCCURRENCES. July September 1994.

• Office for Analy-The Nuclear Regulatory Commission's annual summary of li-sis & Evaluation of Operational Data, Director. January 1995.

censed nuclear power reactor data is based pnmarily on the 35pp. 9504190361. 83575:183.

(

report of operating data submitted by licensees for each unit for Section 208 of the Energy Reorganization Act of 1974 identi-the month of December because that report contains data for fies an abnormal occurrence (AO) as an unscheduled incident the month of December, the year to date (in this case calendar or event that the Nuclear Regulatory Commission determines to year 1994) and cumulative data, usually from the date of com-be significant from the standpoint of public health or safety and mercial operation. The data is not independently verified, but requires a quarterly report of such events to be made to Cork various computer checks are made. The report is divided into gress. This report provides a description of those events that two sections. The first contains summary highhghts and the have been determined to be abnormal occurrences during the second contains data on each individual unit in commercial op-period of July 1 through September 30,1994. This report ad-eration Section 1 capacity and availability factors are simple dresses five abnormal occurrences (AOs) at NRC-licensed fa-anthmetic averages. Section 2 items in the cumulative column cihties. One involved a medical brachytherapy misadministration, are generally as reported by the licensee and notes as to the two involved medical teletherapy misadministrations, one in-use of weighted averages and starting dates other than com-volved a medical sodium iodide misadministration, and one in-mercial operation are provided.

volved a medical sodium iodide event. One AO report submitted by an Agreen.ent State is included. It involved the loss of man-NUREG-0040 V18 N04: LICENSEE CONTRACTOR AND agement and procedural control of a radioactive source. (Due to VENDOR INSPECTION STATUS REPORT. Quarterly publication schedule constraints, NRC was unable to include all Reg., ort, October December 1994.(White Book)

• Office of Nucle-of the AO information received from the Agreement States. Any at Reactor Regulation (Post 941001). February 1995. 200pp.

Agreement State information that was not included in this report 9503140385. 83069:031.

will be published in the next quarterly report. The report also This penodical covers the results of inspections performed by contains updates of six AOs previously reported by NRC licens-the NRC's Special inspection Branch, Vendor inspection Sec-ees and three AOs previously reported by Agreement State li-tion, that have been distnbuted to the inspected organizations censees. Two "Other Events of Interest" conceming nuclear during the penod from October through December 1994.

power reactors are also reported. One involved the fracture of a frozen pipe at Dresden Unit 1 with a consequent release of NUREG-0040 V19 Not: LICENSEE CONTRACTOR AND water, and the other involved the possible deliberate exposure VENDOR INSPECTION STATUS REPORT. Quarterfy of a contract laborer to radiation at Quad Cities Nuclear Power Report, January-March 1995.(White Book)

• Office of Nuclear Station.

Reactor Regulation (Post 941001). May 1995. 119pp.

9506020502. 84149:041.

NUREG-0090 V17 N04: REPORT TO CONGRESS ON ABNOR-This penodical covers the results of inspections performed by MAL OCCURRENCES. October-December 1994.

• Office for the NRC s Special inspection Branch, Vendor inspection Sec-Analysis & Evaluation of Operational Data. Director. May 1995.

tion, that have been distributed to the inspected organizations 39pp. 9511140069. 86174:317.

during the penod from January through March 1995.

Section 208 of the Energy Reorganization Act of 1974 identi-NUREG-0040 V19 NO2: LICENSEE CONTRACTOR AND fies an abnormal occurrence (AO) as an unscheduled incident VENDOR INSPECTION STATUS REPORT. Quarterty or event that the Nuclear Regulatory Commission determines to Report April-June 1995.(White Book)

• Office of Nuclear Reac.

be significant from the standpoint of public health or safety and for Regulation (Post 941001). August 1995. 181pp.

requires a quarterly report of such occurrences to be made to 9509070144. 85398:116.

Congress. This report provides a desenption of those incidents This periodical covers the results of inspections performed by and events that have been determined to be AOS during the the NRC's Special inspection Branch, Vendor Inspection Sec_

period of October 1 through December 31,1994. This report addresses four Aos at NRC-hcensed facilities. These occur-tion, that have been distnbuted to the inspected organizations dunng the pencd from April through June 1995.

rences involved the following: a generic concern relating to core shroud cracking in boiling water reactors; recumng incidents of NUREG-0040 V19 NO3: LICENSEE CONTRACTOR AND administenng higher doses than procedurally allowed for diag-VENDOR INSPECTION STATUS REPORT. Quarterty nostic imapg at a single facihty; one medical teletherapy mis-Report. July-September 1995.(White Book)

• Office of Nuclear administration and one medical brachytherapy misadministra-Reactor Regulation (Post 941001). December 1995. 224pp.

tion. Agreement States submitted four AO reports. These four 9601290256.86882.001.

occurrences involved the following: one major contarrination at This penodical covers the results of inspections performed by a commercial facility; two medical brachytherapy misadministra-the NRC's Special inspection Branch, Vendor inspection Sec-tions; and one medical teletherapy misadministration. The report 1

[

l 2

Main Citations and Abstracts i

also contains updates of seven AOs previously reported by the proved by the U.S. Nuclear Regulatory Commission. To assist in Agreement States. Two "Other Events of interest" are also identifying packaging, an index by Model Number and corre-being reported. These occurrences involved the operability of sponding Certificate of Compliance Number is included at the safety rehof valves at a nuclear powerplant, and an error in the front of Volumes 1 and 2. An alphabeticd listing by user name installation process of a Leksell Gamma Knife " Teletherapy is included in the back of Volume 3 of approved QA programs.

unit that resulted in an operational failure.

The reports include a listing of all useis of each package design NUREG-0090 V18 N01: REPORT TO CONGRESS ON ABNOR.

and approved QA programs pnor to the publication date.

MAL OCCURRENCES. January March 1995.

• Office for Analy-NUREG 0383 V02 R18: DIRECTORY OF CERTIFICATES OF sis & Evaluation of Operational Data, Director. July 1995.25pp, COMPLIANCE FOR RADIOACTIVE MATERIALS 9507250296. B4831:118.

PACKAGES. Certificates Of Compliance.

• Office of Nuclear Ma-Section 208 of the Energy Reorganization Act of 1974 identi-terial Safety & Safeguards. October 1995. 572pp. 9511160271.

fies an abnormal occurrence (AO) as an unscheduled incident 86231:001, or event that the Nuclear Regulatory Commission determines to See NUREG-0383.V01,R18 abstract.

be significant from the standpoint of public health or safety and requires a quarterly report of such occurrences to be made to NUREC-0383 V03 R15: DIRECTORY OF CERTIFICATES OF Congress. This report provides a desenption of those incidents COMPLIANCE FOR RADIOACTIVE MATERIALS and events that have been determmed to be AOS dunng the PACKAGES. Report Of NRC Approved Quality Assurance Pro-period of January 1 through March 31, 1995. This report ad-grams for Radioactwe Materials Packages.

• Office of Nuclear dresses one AO at an NRC-licensed facihty which involved a Material Safety & Safeguards. October 1995. 155pp.

medical brachytherapy misadministration. The report also con-9511160266. 86230.001.

tains updates of one AO previously reported by an NRC licens.

See NUREG-0383,V01 R18 abstract.

ee and three AOs previously reported by the Agreement States.

No Other Events of Inierest items are being reported.

NUREG-0390 V09 Not: TOPICAL REPORT REVIEW STATUS.

• Office of Nuclear Reactor Regulation (Post 941001). February NUREG-0304 V19 N03: REGULATORY AND TECHNICAL RE-1995. 44pp. 9503140388. 83068:272.

PORTS (ABSTRACT INDEX JOURNAL). Compilation For Third This report provides industry with procedures for submitting Quarter 1994, July-September.

• Dwision of Freedom of Informa-topical reports, guidance on how the U.S. Nuclear Regulatory i

tion & Publications Services (Post 940714). December 1994.

Commission (NRC) processes and responds to topical report 62pp. 9502080121, 82660:001.

submittals, and an accounting, with review schedules, of all topi-This journal includes all formal reports in the NUREG series cal reports currently accepted for review by the NRC. This prepared by the NRC staff and contractors, proceedings of con-report is published semiannually.

ferences and workshops, grants, and international agreement reports. The entnes in this compilation are indexed for access NUREG-0430 V14: LICENSED FUEL FACILITY STATUS by title and abstract, secondary report number, personal author, REPORT. inventory Difference Data. July 1,1993 - June 30, subject, NRC organization for staff and international agree-1994-(Gray Book 11) JOY,D.R. Office of Nuclear Material Safety ments, contractor, international organization, and licensed facili.

& Safeguards. March 1995.18pp. 9505100290. 83902:267.

ty.

NRC is committed to the periodic publication of heensed fuel facility inventory di*ference data, following agency review of the NUREG-0304 V19 N04: REGULATORY AND TECHNICAL RE-information and completion of any related NRC investigations.

PORTS (ABSTRACT INDEX JOURNAL). Annual Compilation Information in this report includes inventory difference data for For 1994.

• Dwision of Freedom of information & Publications actwe fuel fabncation facilities possessing more than one effec-Services (Post 940714). March 1995. 133pp. 9504180345.

twe kilogram of special nuclear matenal.

83566:063.

See NUREG-0304,V19,NO3 abstract.

NUREG-0498 S01: FiWAL ENVIRONMENTAL STATEMENT RE.

NUREG-0304 V20 N01: REGULATORY AND TECHN!OAL RE-LATED TO THE OPERATION OF WAITS BAR NUCLEAR PORTS (ABSTRACT INDEX JOURNAL). Compilation For First T, MS 1 AW 2. Docket Nos. 50-390 And 50-391.(Ten-Quarter 1995, January. March.

• Dwision of Freedom of Informa-msse VaHey Auhy, Associate hor for hanced b tion & Publications Services (Post 940714). July 1995. 48pp.

actors & License Renewal (ADAR) (Post 941001). April 1995.

9508160248. 85041:277 366pp. 9505180344. 83965:118.

See NUREG 0304,V19,NO3 abstract.

The Final Environmental Statement (FES) issued in 1978 rep-resents the Nuclear Regulatory Commission's (NRC's) previous NUREG-0304 V20 N02: REGULATORY AND TECHNICAL RE-environmental review related to the operation of Watts Bar Nu-PORTS (ABSTRACT INDEX JOURNAL). Compilation For clear (WBN) Plant. The purpose of this NRC review is to dis-Second Quarter 1995, April-June.

• Duision of Freedom of Infor-cuss the effects of observed changes in environment and to s

mation & Publications Services (Post 940714). September 1995.

evaluate the changes in environmental impacts that have oc-1 48pp. 9510270358. 86009:180.

curred as a result of changes in the WBN Plant design and pro-j l

See NUREG-0304,V19,NO3 abstract.

posed methods of operations since the last environmental NUREG-0325 R18: U.S. NUCLEAR REGULATORY COMMISSION review. A full scope of environmental topics has been evaluat-ORGANIZATION CHARTS AND FUNCTIONAL ed, including regional demography, land and water use, meteor.

STATEMENTS. July 23,1995.

• Ofc of Personnel (Post 870413).

ology, terrestrial and aquatic ecology, radiological and nonradio-July 1995. 67pp. 9511060194. 86094:255.

logical impacts on humans and the environment, socioeconomic Functional statements and organization charts for the U.S.

impacts, and environmental justice. The staff concluded that j

l Nuclear Regulatory Commission offices, dwisions, and branches there are no significant changes in the environmental impacts i

are presented.

since the NRC 1978 FES-OL from changes in plant design, pro-posed methods of operation, or changes in the environment.

i NUREG-0383 V01 R18: DIRECTORY OF CERTIFICATES OF The applicant's preoperational and operational monitonng pro-COMPLIANCE FOR RADIOACTIVE MATERIALS grams were reviewed and found to be appropnate for establish-j PACKAGES. Report Of NRC Approved Packages.

• Office of ing baseline conditions and ongoing assessments of environ-Nuciear Matenal Safety & Safeguards. October 1995. 619pp.

mental impacts. The staff also conducted an analysis of plant 9511160279. 86235:001.

operation with severe accident mitigation design alternatwes The purpose of this directory is to make available a conven-(SAMDAs) and concluded that none of the SAMDAs, beyond ient source of information on packagings which have been ap-the three procedural changes that the applicant committed to 4

Main Citations and Abstracts 3

implement, would be cost-beneficial for further mitigating envi-NUREG-OhD V17 N07: TITLE LIST OF DOCUMENTS MADE ronmental impacts.

PUBLICLi A 'AILABLE. July 1-31, 1995.

• Division of Freedom NUREG 0525 V02 R03: SAFEGUARDS

SUMMARY

EVENT LIST f Inf rmation & Publications Services (Post 940714). Septem-ber 1995. 278pp. 9510030198. 85664:001.

(SSEL). January 1,

1990 Through December 31, 1994 FADDEN.M.; YARDUMIAN J. Operations Branch. July 1995; See NUREG-0540,V16,N11 abstract.

140pp. 9511060198. 86103.096-NUREG-0540 V17 N08: TITLE LIST OF DOCUMENTS MADE The Safeguards Summary Event List provides brief summa-PUBLICLY AVAILABLE. August 1 31, 1995.

• Division of Free-ries of hundreds of safeguards-related events involving nuclear dom of informatson & Publications Services (Post 940714). Oc-material or facilities regulated by the U.S. Nuclear Regulatory tober 1995. 431pp. 9511060215. 86093:130.

Commission. Events are desenbed under the categories: Bomb-See NUREG-0540,V16,N11 abstract.

related, intrusion, Missing / Allegedly Stolen, Transportation-relat.

ed, Tampennc/ Vandalism, Arson, Firearms-related Radiological NUREG-0540 V17 N09: TITLE LIST OF DOCUMENTS MADE Sabotage, Not.. radiological Salotage, and Miscellaneous. Be-PUBLICLY AVAILABLE. September 1-30,1995.

Division of cause of the pu@ interest, the Miscellaneous category also in-Freedom of Information & Pubhcations Services (Post 940714).

cludes events reporteu involving source material, byproduct ma-November 1995, 346pp. 9512050194. 86412:001, terial, and natural uranium, which are exempt from safeguards See NUREG-0540,V16,N11 abstract.

"in the vent desenptions was obtained

[o$"o*f[ci NUREG 0540 V17 N10: TITLE LIST OF DOCUMENTS MADE r es.

PUBLICLY AVAILABLE. October 1-31,1995.

• Division of Free-NUREG-0540 V16 N1t ilTLE LIST OF DOCUMENTS MADE dom of information & Publications Services (Post 940714). De-PUBLICLY AVAILABLE. November 1-30, 1994

• Division of cember 1995. 344pp. 9601030180. 86689:001.

Freedom of lotormation & Pubhcations Services (Post 940714).

See NUREG-0540,V16,N11 abstract.

NUREG-0700 R01 DFC: HUMAN-SYSTEM INTERFACE DESIGN hs d n it t s a month y pub ca son containing desenp-REVIEW GUIDELINE. Draft Report For Comment. O'HARA.J.M.;

tions of inforrr! stion received and generated by the U S. Nuclear BROWN.W.S.; STUBLER W.F.; et al. Brookhaven National Lab-Regulatory Cc mmission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants oratory. February 1995. 496pp. 9503270312. 83269:001.

and other usen of radioactive materials, and (2) nondocketed NUREG-0700, Rev.1, provides humen factors engineering matenal receivt d and generated by NRC pertinent to its role as (HFE) guidance to the U.S. Nuclear Regulatory Commission staff for its: (1) review of the human system interface (HSI) j a regulatory agincy. The following indexes are included: Per-sonal Author, Corporate Source, Report Number, and Cross design submittals prepared by licensees or applicants for a li-Reference of Enclosures to Pnncipal Documents.

cense or design certification of commercial nuclear power plants, and (2) performance of HSI reviews that could be under-NUREG-0540 V16 N12: TITLE LIST OF DOCUMENTS MADE taken as part of an inspection or other type of regulatory review PUBLICLY AVAILABLE. December 1-31, 1994.

• Division of involving HSI design or incidents involving human performance.

Freedom of Information & Publications Services (Post 940714).

It consists of two major parts. Part i describes those aspects of February 1995. 319pp. 9503150148. 83100:285.

the HSI design review process that are important to the identifi-See NUREG-0540,V16,N11 abstract.

cation and resolution of human engineering discrepancies that could adversely affect plant safety. Guidance is provided that NUREG-0540 V17 N01: TITLE LIST OF DOCUMENTS MADE could be used by the staff to review an applicant's HSl design PUBLICLY AVAILABLE. January 1-31, 1995.

• Division of Free' review process. Part 1 could also be used by the staff to guide dom of information & Pubhcations Sennces (Post 940714)'

the development of an HSI design review plan, e.g., as part of March 1995. 336pp. 9504120107. 83475:001.

an inspection activity. Part 2 " Guidelines for Human Factors See NUREG 0540,V16,N11 abstract.

Engineenng Reviews," provides detailed HFE guidelines for the NUREG-0540 V17 N02: TITLE LIST OF DOCUMENTS MADE assessment of HSI design implementations.

PUBLICLY AVAILABLE. February 1 28,1995.* Division of Free-NUREG-0713 V15: OCCUPATIONAL RADIATION EXPOSURE AT dom of Information & Publications Services (Post 940714). April COMMERCIAL NUCLEAR POWER REACTORS AND OTHER 1995. 337pp. 9504250369. 83629:001.

FACILITIES 1993. Twenty-Sixth Annual Report. RADDATZ,C.T.

See NUREG-0540,V16,N11 abstract.

Division of Regulatory Applications (Post 941217).

NUREG-0540 V17 NO3: TITLE LIST OF DOCUMENTS MADE HAGEMEYER,D. Science Applications International Corp. (for-PUBLICLY AVAILABLE. March 1 31, 1995.

• Division of Free-merly Science Applications, Inc.). January 1995. 307pp.

dom of Information & Publications Services (Post 940714). May 9502080242. 82661:001.

1995. 360pp. 9506020497. 84148:024.

This report summarizes the occupationa! radiation exposure See NUREG-0540,V16,N11 abstract.

information that has been reported to the NRC's Radiation Ex.

posure Information Reporting System (REIRS) by nuclear power NUREG-0540 V17 N04: TITLE LIST OF DOCUMENTS MADE facilities and certain other categones of NRC hcensees during PUBLICLY AVAILABLE. April 1 30, 1995.

• Division of Freedom the years 1969 through 1993. The bulk of the data presented in of Information & Publications Services (Post 940714). June the report was obtained from annual radiatim exposure reports 1995 207pp. 9506280447,84468:129.

submitted in accordance with the requirements of 10 CFR See NUREG-0540,V16,N11 abstract.

20.407 and the technical specifications of nuclear power plants.

Data on workers terminating their employment at certain NRC NUREG 0540 V17 N05: TITLE LIST OF DOCUMENTS MADE licensed facilities were obtained from reports submitted pursu-PUBLICLY AVAILABLE.May 1-31, 1995.

• Division of Freedom ant to 10 CFR 20.408. The 1993 annual reports submitted by of information & Pubhcations Services (Post 940714). July 1995.

about 360 licensees indicated that approximately 189,711 Indi-viduals were monitored, 169,872 of whom were monitored by N EG 054d, N

bstract.

nuclear power facihties. They incurred an average individual NUREG-0540 V17 N06: TITLE LIST OF DOCUMENTS MADE dose of 0.16 rem (cSv) and an average measurable dose of PUBLICLY AVAILABLE. June 1-30, 1995.

• Division of Freedom about 0.31 (cSv). Termination radiation exposure reports were of Information & Pubhcations Services (Post 940714). August analyzed to reveal that about 99,749 individuals completed their 1995. 276pp. 9508160276. 85041:001.

employment with one or more of the 360 covered hcensees See NUREG-0540,V16,N11 abstract.

dunng 1993. Some 91.000 of these individuals terminated from

4 Main Citations and Abstracts power reactor facilities, and about 12,685 of them were consid.

See NUREG 0750,V40 abstract.

ered to be transient workers who received an average dose of NUREG-0750 V41 N04: NUCLEAR REGULATORY COMMISSION 0.49 rem (cSv).

ISSUANCES FOR APRll 1995.Pages 245-319.

• Division of NUREG-0725 R10: PUBLIC INFORMATION CIRCULAR FOR Freedom of information & Publications Services (Post 940714).

SHIPMENTS OF IRRADIATED REACTOR FUEL.

• Division of June 1995. 83pp. 9507120290. 84633:001, industnal & Medical Nuclear Safety (Post 870729). Apnl 1995.

See NUREG-0750,V40 abstract.

30pp. 9505180571. 83980 001.

NUREG-0750 V41 N05: NUCLEAR REGULATORY COMMISSION This circular has been prepared to provide information on the SSUANCES FOR MAY 1995.Pages 321380' Post 940714).

• Division of F ee o nfor tion & b cations Services (

t on y uclear ut tory m ssi n ( RC),

t m et J

the requirements of Public Law 96-295. The report provides a bnef desenption of NRC authonty for certain aspects of trans-See NUREG 0750'V40 abstract-porting spent fuel. It provides desenptive statistics on spent fuel NUREG-0750 V41 N06: NUCLEAR REGULATORY COMMISSION shipments regulated by the NRC from 1979 to 1994. It also hsts ISSUANCES FOR JUNE 1995.Pages 381-496.

• Division of detailed highway and railway segments used within each state Freedom of information & Pubhcations Services (Post 940714).

from January 1,1993, through December 31,1994.

August 1995.124pp. 9509080022. 85405:146.

NUREG-0750 V40: NUCLEAR REGULATORY COMMISSION See NUREG-0750,V40 abstract.

ISSUANCES. Opinions And Decisions Of The Nuclear Regula-NUREG-0750 V42101: INDEXES TO NUCLEAR REGULATORY tory Commission With Selected Orders. July-December 1994.

• COMMISSION ISSUANCES. July September 1995.

• Division of Division of Freedom of informahon & Pubhcations Services Freedom of Information & Pubhcations Services (Post 940714).

(Post 940714). July 1995. 400pp. 9508300374. 85286401.

November 1995. 21pp. 9512080146. 86450:187.

Legalissuances of the Commission, the Atomic Safety and Li-See NUREG-0750,V40,102 abstract.

censing Board Panel, the Administrative Law Judges, and NRC Program Offices are presented.

NUREG-0750 V42 N01: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JULY 1995.Pages 1-45.

• Division of Free-NUREG-0750 V40102: INDEXES TO NUCLEAR REGULATORY dom of information & Publications Services (Post 940714).

COMMISSION ISSUANCES. July-December 1994.

• Division of August 1995. 53pp. 9500080026. 85405:270.

Freedom of information & Pubhcahons Services (Post 940714).

See NUREG 0750,V40 abstract.

March 1995. 40pp. 9504120101. 83474:001.

Digests and indexes for issuances of the Commission, the NUREG-0750 V42 N02: NUCLEAR REGULATORY COMMISSION Atomic Safety and Licensing Board Panel, the Administrative ISSUANCES FOR AUGUST 1995. Pages 47-97.

• Division of Law Judges, the Directors' Decisions, and the Denials of Peti-Freedom of Information & Publications Services (Post 940714).

tions for Rulemaking are presented.

October 1995. 58pp. 9511020372. 86091:056.

NUREG 0750 V40 N05: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR NOVEMBER 1994. Pages 169 318.

• Division NUREG-0750 V42 NO3: NUCLEAR REGULATORY COMMISSION of Freedom of information & Pubhcations Services (Post ISSUANCES FOR SEPTEMBER 1995. Pages99-110.

• Division 940714). February 1995.149pp. 9503150152. 83107:001, of Freedom of Information & Publications. Services (Post See NUREG-0750,V40 abstract.

940714). November 1995.18pp. 9511270462. 86321:320.

NUREG-0750 V40 N06: NUCLEAR REGULATORY COMMISSION a s act ee ISSUANCES FOR DECEMBER 1994. Pages 319-387.

• Division NURE3-0750 V42 N04: NUCLEAR REGULATORY COMMISSION of Freedom of information & Pubhcations Services (Post ISSUANCES FOR OCTOBER 1995. Pages 111-180.

• Division 940714). February 1995. 77pp. 9503150157. 83101:242.

of Freedom of Information & Publications Services (Post See NUREG-0750,V40 abstract.

940714). December 1995.194pp. 9601180208. 86806:161.

ee 0,M abstran NUREG-0750 V41101: INDEXES TO NUCLEAR REGULATORY j

COMMISSION ISSUANCES. January March 1995.

• Division of NUREG-0837 V14 N04: NRC TLD DIRECT RADIATION MONI-Freedom of information & Publications Services (Post 940714).

TORING NETWORK. Progress Report. October-December 1994.

June 1995. 34pp. 9507060350. 84533:306.

STRUCKMEYER,R. Region 1 (Post 820201). March 1995.

See NUREG-0750,V40.102 abstract.

329pp. 9503170312. 83147.001.

NUREG-0750 V41 102: INDEXES TO NUCLEAR REGULATORY This report provides the status and results of the NRC Ther.

0"'*

""U COMMISSION ISSUANCES. January June 1995.

• Division of Network. It presents the radiation levels measured in the vicinity Freedom of information & Pubhcations Services (Post 940714).

September 1995. 65pp. 9510180192. 85849:001.

f NRC hcensed facihties throughout the country for the fourth See NUREG-0750,V40,102 abstract.

quam of M91 NUREG-0837 V15 N01: NRC TLD DIRECT RADIATION MONI-NUREG-0750 V41 N01: NUCLEAR REGULATORY COMMISSION TORING NETWORK. Progress Report. January-March 1995.

ISSUANCES FOR JANUARY 1995. Pages 169.

• Division of STRUCKMEYER,R. Region 1 (Post 820201). May 1995. 230pp.

Freedom of Information & Pubhcations Services (Post 940714).

Th s r port ovide the status and results of the NRC Ther-R -

0,V40 t

moluminescent Dosameter (TLD) Direct Radiation Monitoring NUREG-0750 V41 NO2: NUCLEAR REGULATORY COMMISSION Network. It presents the radiation levels measured in the vicinity ISSUANCES FOR FEBRUARY 1995. Pages 71178.

• Division of NRC hcensed facihties throughout the country for the first of Freedom of information & Publications Services (Post quarter of 1995.

UREG0

,V40 a' tr t'

NUMEG-0837 V15 N02: NRC TLD DIRECT RADIATION MONI-b TORING NETWORK. Progress Report.Apol-June 1995.

NUMEG-0750 V41 NO3: NUCLEAR REGULATORY COMMISSION STRUCKMEYER,R Region 1 (Post 820201). August 1995.

ISSUANCES FOR MARCH 1995.Pages 179-243.

• Division of 230pp. 9508300331. 85287:060.

Freedom of Information & Pubhcations Services (Post 940714).

This report provides the status and results of the NRC Ther-May 1995. 72pp. 9506140062. 84297:130.

moluminescent Dosameter (TLD) Direct Radiation Monitonng

Main Citations and Abstracts 5

Network. It presents the radiation levels measured in the vicinity 50-390 and 50-391, located in Rhea County Tennessee, has of NRC licensed fachties throughout the country for the second been prepared by the Office of Nuclear Reactor Regulation of quarter of 1995.

the Nuclear Regulatory Commission. The purpose of this sup-NUREG-0837 V15 NO3: NRC TLD DIRECT RADIATION MONI-piement is to update the Safety Evaluation with (1) additional in-TORING NETWORK. Progress Report. July-September 1995.

f rmation submitted by the applicant since Supplement No.17 i

STRUCKMEYER R. Region 1 (Post 820201). December 1995.

was issued, and (2) matters that the staff had under review 231pp. 9601290199. 86879:051.

when Supplement No.17 was issued.

This report provides the status and results of the NRC Ther-NUREG-0847 S19: SAFETY EVALUATION REPORT RELATED moluminescent Dosimeter (TLD) Direct Radiation Monitoring TO THE OPERATION OF WATTS BAR NUCLEAR Network, it presents the radiation levels measured in the vicinity of NRC licensed facilities throughout the country for the third PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-301.(Ten-

)

quarter of 1995.

nessee Valley Authonty) TAM,P.S. Office of Nuclear Reactor Regulation (Post 941001). November 1995.96pp.9512040337.

j NUREG-0847 S15: SAFETY EVALUATION REPORT RELATED 86385:257, TO THE OPERATION OF WATTS BAR NUCLEAR Supplement No.19 to the Safety Evaluation Report for the PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Ten-application filed by the Tennessee Valley Authority for license to i

nessee Valley Authonty) TAM,P.S. Division of Reactor Projects operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos.

1/11 (DRPE) Post 941001). June 1995.189pp. 9507120275.

50-390 and 50-391, located in Rhea County Tennessee, has 84634:001, been prepared by the Office of Nuclear Reactor Regulation of Supplement No.15 to the Safety Evaluation Report for the the Nuclear Regulatory Commission. The purpose of this sup-application filed by the Tennessee Valley Authonty for license to piement is to update the Safety Evaluation with (1) additional in-operate Watts Bar Nuclear Plant. Units 1 and 2, Docket Nos.

formation submitted by the applicant since Supplement No.18

)

50-390 and 50-391, located in Rhea County Tennesaee, has was issued, and (2) matters that the staff had under review been prepared by the Office of Nuclear Reactor Regulation of when Supplement No.18 was issued.

the Nuclear Regulatory Commission. The purpose of this sup-plement is to update the Safety Evaluation with (1) additional in-NUREG-0933 S18: A PRIORITIZATION OF GENERIC SAFETY formation submitted by the applicant since Supplement No.14 ISSUES. EMRIT,R. Division of Engineering Technology (Post was issued, and (2) matters that the staff had under review 941217). Apnl 1995. 266pp. 9505190010. 83981:001.

when Supplement No.14 was issued.

The report presents the safety prionty ranking for generic NUREG-0847 S16: SAFETY EVALUATION REPORT RELATED safety issues related to nuclear power plants. The purpose of TO THE OPERATION OF WATTS BAR NUCLEAR PLANT, these rankings is to assist in the timely and efficient allocation UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Tennessee of NRC resources for the resolution of those safety issues that Val!ey Authonty) TAM P.S. Office of Nuclear Reactor Regulation have a significant potential for reducing risk. The safoty prionty (Post 941001). September 1995. 109pp. 9510270354.

rankings are HIGH, MEDIUM, LOW, and DROP, and have been 86009:069.

assigned on the basis of risk significance estimates, the ratio of Supplement No.16 to the Safety Evaluation Report for the nsk to costs and other impacts estimated to result if resolution application filed by the Tennessee Valley Authonty for license to of the safety issues were implemented, and the consideration of operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos.

uncertainties and other quantitative or qualitative factors. To the 50-390 and 50 391, located in Rhea County Tennessee, has extent practical, estimates are quantitative.

been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of.this sup.

NUREG-0933 S19: A PRIORITIZATION OF GENERIC SAFETY plement is to update the Safety Evaluation with (1) additional in.

ISSUES. EMRIT,R. Division of Engineenng Technology (Post formation submitted by the applicant since Supplement No.15 941217). December 1995. 414pp. 96C1290268. 86878:001.

was issued, and (2) matters that the staff had under review The report presents the safety prionty rankirq for generic when Supplement No.15 was issued.

safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation NUREG-0847 S17: SAFETY EVALUATION REPORT RELATED of NRC resources for the resolution of those safety issues that TO THE OPERATION OF WATTS BAR NUCLEAR have a significant potential for reducing nsk. The safety pnonty PLANT, UNITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Ten-rankings are HIGH, MEDIUM, LOW, and DROP, and have been nessee Valley Authonty) TAM,P.S. Office of Nuclear Reactor assigned on the basis of nsk significance estimates, the ratio of Regulation (Post 941001). October 1995. 70pp. 9511020395.

risk to costs and other impacts estimated to result if resolution 86090:277.

f the safety issues were implemented, and the consideration of Supplement No.17 to the Safety Evaluation Report for the uncertainties and other quantitative or qualitative factors. To the application filed by the Tennessee Valley Authonty for license to extent practical, estimates are quantitative.

operate Watts Bar Nuclear Plant, Units 1 and 2 Docket Nos.

50-390 and 50-391, located in Rhea County Tennessee, has NUREG-0936 V13 NO3:

NRC REGULATORY been prepared by the Office of Nuclear Reactor Regulation of AGENDA. Semiannual Report, July-December 1994.

• Division of the Nuclear Regulatory Commission. The purpose of this sup-Freedom of Information & Publications Services (Post 940714).

plement is to update the Safety Evaluation with (1) additional in-February 1995. 56pp. 9503140450. 83071:106.

formation submitted by the apphcant since Supplement No.16 The NRC Regulatory Agenda is a compilation of all rules on i

was issued, and (2) matters that the staff had under review which the NRC has recently comp!sted action, or has proposed when Supplement No.16 was issued.

action, of is considenng action, and all petitions for rulemaking NUREG-0847 S18: SAFETY EVALUATION REPORT RELATED which have been received by the Commission and are pending TO THE OPERATION OF WATTS BAR NUCLEAR disposition by the Commission. The Regulatory Agenda is up-PLANT,0 NITS 1 AND 2. Docket Nos. 50-390 And 50-391.(Ten-dated and issued semiannually.

nessee Valley Authonty) TAM,P.S. Office of Nuclear Reactor Regulation (Post 941001). October 1995.174pp. 9511290117.

NUREG-0936 V14 N01:

NRC REGULATORY 86354:001.

AGENDA. Semiannual Report, January-June 1995.

• Division of i

Supplement No.18 to the Safety Evaluation Report for the Freedom of Information & Publications Services (Post 940714).

application filed by the Tennessee Valley Authonty for heense to September 1995. 56pp. 9510030203. 85663.193.

operate Watts Bar Nuclear Plant Units 1 and 2. Docket Nos.

See NUREG-0936,V13,NO3 abstract.

_m 6

Main Citations and Abstracts NUREG-0940 V13 N4 P1:

ENFORCEMENT by the NRC, so that actions can be taken to improve safety by ACTIONS:SIGNIFICANT ACTIONS RESOLVED REACTOR avoiding future violations similar to those desenbod in this publi-LICENSEES.Ouarterly Progress Report, October-December cation.

1994.

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

NUREG-0940 V14 N1 P3: ENFORCEMENT ACTIONS: SIGNIFl-360pp. 9503140428. 83071:171 This compilation summarizes significant enforcement actions CANT ACTlONS RESOLVED, MATERIAL LICENSEES (NON-that have been resolved dunng one quarterly penod (October.

MEDICAL).Ouarterly Progress Report, January-March 1995.

• December 1994) and includes copies of letters, Notces, and Ofc of Entorcement (Post 870413). May 1995. 439pp.

Orders sent by the Nuclear Regulatory Commission to reactor 9506160045. 84331:306.

i hcensees with respect to these enforcement actons. It is antci.

This compilation summarizes significant enforcement actions pated that the information in this pubhcation will be widely dis-that have been resolved during one quarterly period (January -

seminated to managers and employees engaged in activities li.

March 1995) and includes copies of letters, Notees, and Orders censed by the NRC, so that actions can be taken to improve sent by the Nuclear Regulatory Commission to Material Licens-safety by avoiding future violatons similar to those desenbed in ees (non-Medical) with respect to these enforcement actions. It this publication.

is anticipated that the informaton in this publication will be we ssemsaM to managers and Woms enga# h NUREG-0940 V13 N4 P2-ENFORCEMENT activities licensed by the NRC, so that, actions can be taken to ACTIONS:SIGNIFICANT ACTIONS hlESOLVED MEDICAL LICENSEES.Ouarterly Progress Report, October-December in 1994.

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

300pp. 9503140434. 83077:001-NUREG-0940 V14 N2 P1:

ENFORCEMENT This compilation summarizes significant enforcement actions ACTIONS:SIGNIFICANT ACTIONS RESOLVED,1NDIVIDUAL that have been resolved during one quarterly period (October -

ACTIONS.Ouarterty Progress Report, April-June 1995.

• Ofc of December 1994) and includes copies of letters, Notices, and Enforcement (Post 870413). August 1995. 226pp. 9509080019.

Orders sent by the Nuclear Regulatory Commission to medical 85404:138.

licensees with respect to these enforcement actions. It is antici-This compilation summarizes sigrufcant enforcement actions pated that the information in this publication will be widefy dis-that have been resolved dunng one quarterly period (April.

seminated to managers and employees engaged in activities h-June 1995) and includes copies of Orders sent by the Nuclear censed by the NRC, so that actions can be taken to improve Regulatory Commission to individuals with respect to these en-safety by avoiding future violations similar to those desenbed in forcement actons. It is anticipated that the information in this this publication.

pubhcaten will be widely dissemnated to managers and em-NUREG-0940 V13 N4 P3:

ENFORCEMENT ployees engaged in activities licensed by the NRC. The Com-ACTIONS:SIGNIFICANT ACTIONS RESOLVED MATERIAL Lt.

mission beheves this informaton may be useful to hcensees in CENSEES (NON-MEDICAL).Ouarterly Progress Report, October.

making employment decisions.

December 1994.

• Ofc of Enforcement (Post 870413). February NUREG-0040 V14 N2 P2:

ENFORCEMENT ACTIONS:SIGNIFICANT ACTIONS RESOLVED REACTOR Thih i

n su ma es s cant enforcement actons MSEES.Ouarterty Progress Report, April-June 1995.

Ofc that have been resolved dunng one quarterly penod (October -

i Enforcement (Post 870413). August 1995. 189pp.

December 1994) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commisson to Matenal his pila mmarizes signifcant enforcement actons Licensees (non-Medical) with respect to these enforcement ac-that have been resolved during one quarterty period (April.

tions. It is antcipated that the informaton in this pubhcation will June 1995) and includes copies of letters, Notees, and Orders be widely dissennated to managers and employees engaged in sent by the Nuclear Regulatory Commission to reactor licens-activities hcensed by the NRC, so that actions can be taken to oes with respect to these enforcement actions. It is anticipated im ve sa y a oiding future violatons similar to those de-that the information in this publication will be widely disseminat-ed to managers and awki;;; engaged in actrvities hcensed

)

NUREG-0940 V14 N1 P1: ENFORCEMENT ACTIONS: SIGNIFl-by the NRC, so that actions can be taken to improve safety by i

CANT ACTIONS RESOLVED, REACTOR LICENSEES.Ouarterly avoideng future violations similar to those described in this publi-Progress Report, January-March 1995.

• Ofc of Enforcement caton.

(Post 870413). May 1995. 360pp. 9506160009. 84333:017.

NUREG-0940 V14 N2 P3: ENFORCEMENT ACTIONS: SIGNIFl-This compilaton summarizes signifcant enforcement actions that have been resolved dunng one quarterly penod (January.

CANT ACTIONS RESOLVED MATERIAL LICENSEES.Ouarterty March 1995) and includes copies of letters, Notees, and Orders Progress Report, April-June 1995.

• Ofc of Enforcement (Post I

sent by the Nuclear Regulatory Commsson to reactor licens-870413). August 1995. 223pp. 9509 t 30141. 85432:111.

ees with respect to these enforcement actons. It is anticipated This compilation summarizes sagrufcant enforcement actions that the information in this publication will be widely dissennat.

that have been resolved during one quarterly period (Apnl -

ed to managers and employees engaged in activities Icensed June 1995) and includes copies of letters, Notices, and Orders by the NRC, so that actions can be taken to improve safety by sent by the Nuclear Regulatory Commission to material heens-avoiding future violations similar to those described in this pubh-ees with respect to these enforcement actions. It is antcipated cation.

that the information in this publicaton will be widely disseminat-ed to managers and employees engaged in activities hcensed NUREG-0940 V14 N1 P2: ENFORCEMENT ACTIONS: SIGNIFI-by the NRC, so that actions can be taken to improve safety by CANT ACTIONS RESOLVED, MEDICAL LICENSEES.Ouarterly avoiding future violatons similar to those desenbed in this puble-Progress Report. January-March 1995.

• Ofc of Enforcement cation.

(Post 870413). May 1995. J00pp. 9506140141. 84301:001.

This compilation summarizes significant enforcement actons NUREG-0980 V01 NO3:

NUOLEAR REGULATORY that have been resolved during one quarterly period (January.

LEGISLATION.103D Congress.

• Office of the General Counsel March 1995) and includes copies of letters Notees, and Orders (Post 860701). August 1995. 54?pp. 9511090181. 86131:114.

sent by the Nuclear Regulatory Commission to medical hcens-This document is a compilation of nuclear regulatory legisla-ees with respect to these enforcement actions. It is anticipated tion and other relevant material through the 103d Congress,2d j

that the information in this publication will be widely disseminat.

Session. This compilation has been prepared for use as a re-ed to managers and employees engaged in activities licensed source document, which the NRC intends to update at the end

l l

l Main Citations and Abstracts 7

of every Congress. The contents of NUREG-0980 include The plant in a manner ensuring personnel and public health and l

Atomic Energy Act of 1954, as amended; Energy Reorganiza-safety. The BWR K/A catalog is organized into six major sec.

I tion Act of 1974, as amended, Uranium Mill Taihngs Radiation tions: Organization of the Catalog; Plant Wide Genenc Knowl-l Control Act of 1978; Low-Level Radioactive Waste Policy Act; edge and Abilities; Plant Systems Grouped by Safety Functions-Nuclear Waste Pohey Act of 1982; and NRC Authorization and Emergency and Abnormal Plant Evolutions; Components; and Appropriations Acts. Other materials included are statutes and Theory.

treaties on export licensing, nuclear non-proliferation, and envi-j l

rorunental protection.

NUREG-1125 V16: A COMPILATION OF REPORTS OF THE AD.

NUREG-0980 V02 NO3:

NUCLEAR REGULATORY VISORY COMMITTEE ON REACTOR SAFEGUARDS.1994 l

LEGISLATION.103d Congress.

• Office of the General Counsel Annual.

• ACRS - Advisory Committee on Reactor Safeguards.

A ril 1995.115pp. 9505190009. 83980:082.

(Post 860701). August 1995. 469pp. 9511090195. 86130:001, P

See NUREG-0980,V01,NO3 abstract.

This compilation contains 30 ACRS reports submitted to the

• "** "' 9 NUREG-1065 R02: ACCEPTABLE STANDARD FORMAT AND alendar yea'r 1994. It also includes a report to the Congress on CONTENT FOR THE FUNDAMENTAL NUCLEAR MATERIAL the NRC Safety Research Program. All reports have been made CONTROL (FNMC) PLAN REQUIRED FOR LOW-ENRICHED available to the public through the NRC Public Document Room URANIUM FACILITIES. JOY,D.R. Division of Fuel Cycle Safety and the U.S. Library of Congress. The reports are categorized

& Safeguards (Post 930207). December 1995. 62pp.

9601290207. 86892:210.

by the most appropriate generic subject area and by chronologi-This report documents a standard format suggested by the cal order within subject area.

l NRC for use in preparing fundamental nuclear material control NUREG-1145 V11: U.S. NUCLEAR REGULATORY COMMISSION (FNMC) plans as required by the Low Enriched Uranium Refonn 1994 ANNUAL REPORT.

• Office of Administration, Director Amendments (10CFR 74.31). This report also duscribes the (Post 940714). June 1905. 300pp. 9507170098. 84671:001-necessary contents of a comprehensive plan and provides ex-ample acceptance enteria which are intended to communicate This report covers the major activities, events, decisions, and acceptable means of achieving the performance capabikties of planning that took place during Fiscal Year 1994 within the U.S.

the Reform Amendments. By using the suggested format, the li-Nuclear Regulatory Commission (NRC) or involving the NRC.

censee or applicant will minimize administrative problems asso-NUREG-1266 V09: NRC SAFETY RESEARCH IN SUPPORT OF ciated with the submittal, review and approval of the FNMC plan. Preparation of the plan in accordance with this format will REGULATION - FY 1994

• Office of Nuclear Regulatory Re-p June 1995. 100pp. 9507200206' l

assist the NRC in evaluating the plan and in standardizing the BNM

• review and licensing process. However, conformance with this l

guidance is not required by the NRC. A license applicant who This report, the tenth in a series of annual reports, was pre-l employs a format that provides a equal level of completeness pared in response to congressional inquiries conceming how j

I and detail may use their own format. This document is also in, nuclear regulatory research is used. It summarizes the accom-l tended for providing guidance to licensees when making revi-plishments of the Office of Nuclear Regulatory Research during I

sions to their FNMC plan.

FY 1994. The goal of the Office of Nuclear Regulatory Re-i search (RES) is to ensure the availability of sound technical NUREG-1100 VII: BUDGET ESTIMATES. Fiscal Years 1996-bases for timely rulemaking and related decisions in support of i

1997.

• Division of Budget & Analysis (Post 890205). February NRC regulatory / licensing / inspection activities. RES also has re-(

1995. 201pp. 9504180341. 83565:223-sponsibilities related to the resolution of generic safety issues l

This report contains the fiscal year budget justification to Con-and to the review of licensoe submittals regarding individual j

gress. The budget provides estimates for salaries and expenses plant examinations. It is the responsibility of RES to conduct the j

and for the Office of the inspector General for fiscal years 1996 NRC's rulemaking process, including the issuance of regulatory and 1997.

guides and rules that govern NRC licensed activities.

NUREG-1122 R01: KNOWLEDGE AND ABILITIES CATALOG FOR NUCLEAR POWER PLANT OPERATORS. PRESSURIZED NUREG-1272 V08 N02: OFFICE FOR ANALYSIS AND EVALUA-WATER REACTORS.

• Office of Nuclear Reactor Regulation TION OF OPERATIONAL DATA.1993 Annual Report - Nuclear (Post 941001). August 1995.500pp.9603280103.

Materials.

• Office, for Analysis & Evaluation of Operat!onal This docurnent provides the basis for the development of Data, Director. May 1995.122pp. 9511140066. 86176:001.

content-valid licensing examinations for reactor operators and The annual report of the U.S. Nuclear Regulatory Commis-senior reactor operators. The examinations developed using the sion's Office for Analysis and Evaluation of Operational Data PWR catalog will cover those topics listed under Title 10, " Code (AEOD) is devoted to the activities performed during 1993. The of Federal Regulations," Part 55. The PWR catalog contains ap-report is published in two separate parts. NUREG 1272, Vol. 8, proximately 5100 knowledge and abihty (K/A) statements for re-No.1, covers power reactors and presents an overview of the actor operators and senior reactor operators. The catalog is or-operating experience of the nuclear power industry from the ganized into six major sections: Catalog Organization; Genenc NRC perspective, including comments about the trends of some Knowledge and Abilities; Plant Systems: Emergency and Abnor-key performance measures. The report also includes the pnnch mal Plant Evolutions: Components and Theory.

pat findings and issues identified in AEOD studies over the past NUREG-1123 RO1: KNOWLEDGE AND ABILITIES CATALOG year and summarizes information from such sources as licensee FOR NUCLEAR POWER PLANT OPERATORS: BOILING event reports, diagnostic evaluations, and reports to the NRC's WATER REACTORS.

• Office of Nuclear Reactor Regulation Operations Center. NUREG-1272, Vol 8, No. 2, covers nuclear (Post 941001). August 1995. 412pp. 9509250374. 85579:001.

matenals and presents a review of the events and concems This document provides the basis for the development of dunng 1993 associated with the use of hcensed materialin non-content-vahd hcensing examinations for reactor operators and reactor applications, such as personnel overexposures and senior reactor operators. The examinations developed using the medical misadministrations. Note that the subtitle of No. 2 has BWR catalog will cover those topics hsted under Title 10, Code been changed from "Nonreactors" to " Nuclear Materials." Both of Federal Regulations, Part 55. The BWR catalog contains ap-reports also contain a discussion of the incident investigation proximately 7,000 knowledge and abihty (K/A) statements for Team program and summarize both the incident Irwestigation reactor operators and senior reactor operators. Each K/A state.

Team and Augmented Inspection Team reports. Each volume j

ment has been rated for its importance to safe operation of the contains a hst of the AEOD reports issued for 19811993.

8 Main Citations and Abstracts NUREG-1275 V11: OPERATING EXPERIENCE FEEDBACK NUREG-1323 R01: LICENSE APPLICATION REVIEW PLAN FOR REPORT - TURBINE-GENERATOR OVERSPEED PROTEC.

A GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL TION SYSTEM. Commercial Power Reactors. ORNSTEIN,H.L.

AND HIGH-LEVEL RADIOACTIVE WASTE. Draft Review Plan.

• Division of Safety Programs (Post 870413). April 1995.117pp.

Division of Waste Management (NMSS 940403). December 9505180319. 83965:001.

1995.166pp. 9601290272. 86877:168.

The report presents the results of the U.S. Nuclear Regula-The License Application Review Plan (LARP) provides guid-tory Commission's Office for Analysis and Evaluation of Oper.

ance to the NRC staff for its review of the U.S. Department of ational Data (AEOD) review of operating experience of main tur.

Energy's (DOE's) license application to construct a mined geo-bine-generator overspeed and overspeed protection systems.

logic repository for the tlisposal of spent nuclear field and other AEOD's study provides insight into the shortcomings in the high-level radioactive waste at Yucca Mountain. The LARP, Re-design, operation, maintenance, testing, and human factors as-vision 0 was issued September 1994. Revision 0 represented sociated with turbine overspeed protection systems. It includes the staff's initial efforts in developing the LARP and was com-an indepth examination of the turbine overspeed event that oc-prised of both completed and outlined individual review plans.

The LARP was and continues to be, however, a work in curred on November 9,1991, at the Salem Unit 2 Nuclear Power Plant. It 'also provides information concerning actions progress. This draft revision, designated Revision 1, represents the staff's latest efforts in the development of the LARP. Ap-taken by other utilities and the turbine manufacturers as a result pendix D provides a status of the development of the individual of the Salem overspeed event. AEOD's study reviewed operat-review plans. Future revisions to the LARP, however, are uncer-ing procedures and plant practices. It noted differences be-tain due to a changing high-level waste program and associated tween turbine manufacturer designs and recommendations for budget constraints.

1 operations, maintenance, and testing, and also.dentified signif.i-I i

cant vanations in the manner that individual plants maintain and NUREG-1350 V07: NUCLEAR REGULATORY COMMISSION IN-test their turbine overspeed protection systems.

FORMATION DIGEST.1995 Edition. STADLER,L. Division of Budget & Analysis (Post 890205). March 1995. 67pp.

NtfREG-1280 R01: STANDARD FORMAT AND CONTENT AC-9505230149. 84017:040.

CEPTANCE CRITERIA FOR THE MATERIAL CONTROL AND The Nuclear Regulatory Commission Information Digest ACCOUNTING (MC&A) REFORM AMENDMENT.

• Division of (digest) provides a summary of information about the U.S. Nu-Fuel Cycle Safety & Safeguards (Post 930207). Apnl 1995.

clear Regulatory Commission (NRC), NRC's regulatory responsi-119pp. 9505180530. 83968:001-bilities, the activities NRC licenses, and general information on I

in 1987 the NRC revised the material control and accounting domestic and worldwide nuclear energy. The digest, published requirements for NRC licensees authorized to possess and use annually, is a compilation of nuclear and NRC-related oata and a formula quantity (i.e.,5 formula kilograms or more) of strategic is designed to provide a quick reference to major facts about special nuclear material. Those revisions issued as 10CFR the agency and the industry it regulates. In general, the data j

47.51 59 require timely monitonng of in-process inventory and cover 1975 through 1993, with exceptions noted. Information on discrete items to detect anomalies potentially indicative of mate-generating capacity and average capacity factor for operating nal losses. Timely detection and enhanced loss localization ca-U.S. commercial nuclear power reactors is obtained from pabilities are beneficial to alarm resolution and also for material monthly operating reports that are submitted directly to the NRC recovery in the event of an actual loss. NUREG-1280 was by the hcensee. This information is reviewed by 0 a NRC for issued in 1987 to present entena that could be used by appli-consistency only and no independent validation and/or venfica-cants, licensees, and NRC license reviewers in the initial prepa-tion is performed.

ration and subsequent review of fundamental nuclear matenal control (FNMC) plans submitted in response to tne Reform NUREG-1363 V06: ATOMIC SAFETY AND LICENSING BOARD Amendment. This document is also intended for both licensees PANEL BIENNIAL REPORT. Fiscal Years 1993 - 1994.* Atomic and license reviewers with respect to FNMC plan revisions.

Safety & Licensing Board Panel. August 1995. 54pp.

9509070103. 85398:297.

General performance objectives, system capabilities, process monitonng, item monitoring, alarm resolution, quality assurance, in Fiscal Year 1993, the Atomic Safety and Licensing Board and accounting are addressed. This revision to NUREG 1280 is Panel ("the Panel") handled 30 proceedings. In Fiscal Year an expansion of the initial edition, which clanfios and expands 1994, the Panet handled 36 proceedings. The cases addressed upon several topics and addresses issues identified under issues in the construction, operation, and maintenance of com-Reform Amendment implementation expenence, mercial nuclear power reactors and other activities requiring a license from the Nuclear Regulatory Commission. This report NUREG-1307 R05:

REPORT ON WASTE BURIAL sets out the Paners caseload during the year and summanzes, CHARGES Escalation Of Decommissioning Waste Disposal highlights, and analyzes how the wide-ranging issues raised in Costs At Low-Level Waste Burial Facihties.

• Division of Regula-those proceedings were addressed by the Paners judges and licensing boards.

tory Apphcations (Post 941217). August 1995. 59pp.

9509070078. 85404:079.

NUREG-1415 V07 NO2: OFFICE OF THE INSPECTOR One of the requirements placed upon nuclear power reactor GENERAL. Semiannual Report To Congress, October 1,1994 -

licensees by the U.S. Nuclear Regulatory Commission (NRC) is March 31,1995. BARCHI,T.; NORTON L.: GRODIN,M.; et al.

for the licensees to periodically adjust the estimate of the cost Office of the inspector General (Post 890417). June 1995.

of decommissioning their plants, in dollars of the current year, 50pp. 9506220103. 84394:337.

as part of the process to provide reasonable assurance that The inspector General is required by the inspector General adequate funds for decommissioning will be available when Act of 1978, as amended, to prepare a " Semiannual Report" to needed. This report, which is scheduled to be revised period'-

the U.S. Congress that summanzes program activities. The 6-cally, contains the development of a formula for escalating de-month reporting penod ends March 31 and September 30. The commissioning cost estimates that is acceptable to the NRC, inspector General's report is submitted to the Chairman of the and contains values for the escalation of radioactive waste NRC not later than Apnl 30 and October 31, respectively. The burial costs, by site and by year. The licensees may use the for-Chairman comments on the Inspector Generars report and pre.

mula, the coefficients, and the burial escalation from this report pares his own, as required by the Act, and submits both reports in their escalation analyses, or they may use an escalation rate to Congress no later than November 30 and May 31, respec-at least equal to the escalation approach presented herein.

tively.

i Main Citations and Abstracts 9

NUREG 1415 V08 N01: OFFICE OF THHE INSPECTOR ence gained from license amendment apphcations to convert to GENERALSomiannual Report To Congress, April 1,1995 - Sep-these improved STS or to adopt partal improvements to exist-tember 30,1995. BARCHI,T.; WATKINS,B.; GRODIN,M.; et al.

ing technical specifications. This report is the result of extensive Office of the inspector General (Post 890417). December 1995.

pubhc technical meetings and discussions behveen the Nuclear 44pp. 9601290161. 86882:226.

Regulatory Commission (NRC) staff and various r,uclear power The inspector General Act of 1978, as amended, requires plant hcensees, Nuclear Steam Supply System (NSSS) Owners that inspectors General submit a " Semiannual Report to Con.

Groups, NSSS vendors, and the Nuclear Energy Institute (NEI).

gress" summarizing program activities. The inspector General's The improved STS were developed based on the criteria in the report is submitted to the Chairman of the NRC not later than Final Commission Pohey Statement on Technical Specifications April 30, and October 31 for each reporting period. The Chair-Improvements for Nuclear Power Reactors, dated July 22,1993.

man comments on the report and prepares the NRC's Semian-The improved STS will be used as the basis for individual nucle-nual Report to Congress as required by the Act. The Chairman ar power plant licensees to develop improved plant. specific then submits the agency's report and the OlG's report no later technical specifications. This report contains three volumes.

than November 30 and May 31, respectively.

Volume 1 contains the Specifications for all chapters and sec-tions of the improved STS. Volume 2 contains the Bases for NUREG-1423 V05: A COMPILATION OF REPORTS OF THE AD-Chapters 2.0 and 3.0, and Sections 3.1 - 3.3 of the improved VISORY COMMITTEE ON NUCLEAR WASTE. July 1993 - June STS. Volume 3 contains the Bases for Sections 3.4 3.9 of the 1995.

• Advisory Committee on Nuclear Waste. August 1995.

improved STS.

75pp. 9509130173. 85433:144.

This compilation contains 13 reports issued by the Advisory NUREG-1431 V02 Rot: STANDARD TECHNICAL SPECIFICA-Committee on Nuclear Waste (ACNW) during the sixth and sev.

TIONS WESTINGHOUSE PLANTS. Bases (Sections 2.0 3.3).

• enth years of its operation. The reports were submitted to the Office of Nuclear Reactor Regulation (Post 941001). Apnl 1995.

Chairman and Commissioners of the U.S. Nuclear Regulatory 350pp. 9506280489. 84458:001.

Commission. All reports prepared by the Committee have been See NUREG-1431,V01,R01 abstract.

1 made available to the public through the NRC Public Document NUREG-1431 V03 R01: STANDARD TECHNICAL SPECIFICA-Room and the U.S. Library of Congress.

TIONS WESTINGHOUSE PLANTS. Bases (Sections 3.4 - 3.9).

• NUREG-1430 V01 R01: STANDARD TECHNICAL SPECIFICA.

Office of Nuclear Reactor Regulation (Post 941001). April 1995.

TIONS BABCOCK AND WILCOX PLANTS. Specifications.

• 500pp. 9506280506. 84459:001.

Office of Nuclear Reactor Regulation (Post 941001). April 1995.

See NUREG-1431,V01,R01 abstract.

326pp. 9506220126. 84395:027.

This report documents the results of the combined effort of NUREG-1432 V01 Rot: STANDARD TECHNICAL SPECIFICA.

the NRC and the industry to produce improved Standard Tech.

TIONS COMBUSTION ENGINEERING PLANTS. Specifications.

nical Specifications (STS), Revision 1 for Babcock & Wilcox

• Office of Nuclear Reactor Regulation (Post 941001). April Plants. The changes reflected in Revision 1 resulted from the 1995. 442pp. 9506220142. 84397:326.

expenence gained from license amendment applications to cory This report documents the resufts of the combined effort of vert to these improved STS or to adopt partial improvements to the NRC and the industry to produce improved Standard Tech-existing technical specifications. This NUREG is the result of ex.

nical Specifications (STS), Revision 1 for Combustion Engineer-tensive public technical meetings and discussions between the ing Plants. The changes reflected in Revision 1 resulted from Nuclear Regulatory Commission (NRC) staff and various nuclear the experience gained from license amendment applications to power plant hcensees, Nuclear Steam Supply System (NSSS) convert to these improved STS or to adopt partial improve-Owners Groups, NSSS vendors, and the Nuclear Energy Insti.

ments to existing technical specifications. This report is the tute (NEI). The improved STS were developed based on the cri.

result of extensive public technical meetings and discussions teria in the Final Commission Policy Statement on Technical between the Nuclear Regulatory Commission (NRC) staff and Specifications Improvements for Nuclear Power Reactors, dated various nuclear power plant licensees, Nuclear Steam Supply j

July 22,1993. The improved STS will be used as the basis for System (NSSS) Owners Groups, NSSS vendors, and the Nucle-

)

individual nuclear power plant licensees to develop improved ar Energy institute (NEI). The improved STS were developed based on the enteria in the Final Commission Policy Statement plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters on Technical Specifications improvements for Nuclear Power and sections of the improved STS. Volume 2 contains the Reactors, dated July 22,1993. The improved STS will be used 1

as the basis for individual nuclear power plant ficensees to de-Bases for Chapters 2.0 and 3.0, and Sections 3.1 - 3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4 velop improved plant specific technical specifications. This j

report contains three volumes. Volume 1 contains the Specifica-3.9 of the improved STS.

tions for all chapters and sections of the improved STS. Volume NUREG-1430 V02 R01: STANDARD TECHNICAL SPECIFICA.

2 contains the Bases for Chapters 2.0 and 3.0, and Sections TlONS BABCOCK AND WILCOX PLANTS. Bases (Sections 2.0 3.1 - 3.3 of the improved STS. Volume 3 contains the Bases for

- 3.3).

• Office of Nuclear Roactor Regulation (Post 941001).

Sections 3.4 3.9 of the improved STS.

I April 1995. 300pp. 9506220131. 84396:001.

See NUREG 1430.V01,R01 abstract.

NUREG 1432 V02 R01: STANDARD TECHNICAL SPECIFICA.

TiONS COMBUSTION ENGINEERING PLANTS. Bases (Sec-NUREG-1430 V03 Rot: STANDARD TECHNICAL SPECIFICA-tions 2.0 3.3).

• Office of Nuclear Reactor Regulation (Post TIONS BABCOCK AND WILCOx PLANTS. Bases (Sections 3.4 941001). April 1995. 450pp. 9506220171. 84399:040.

- 3.9)

• Office of Nuclear Reactor Regulation (Post 941001)-

See NUREG-1432,V01,R01 abstract.

Apnl 1995. 350pp, 9506220136. 84396:302.

See NUREG 1430,V01,R01 abstract.

NUREG-1432 V03 RO1: STANDARD TECHNICAL SPECIFICA-TlONS COMBUSTION ENGINEERING PLANTS. Bases (Sec-l NUREG-1431 V01 RO1: STANDARD TECHNICAL SPECIFICA-tions 3.4 - 3.9).

• Office of Nuclear Reactor Regulation (Post TIONS WESTINGHOUSE PLANTS. Specifications.

• Office of 941001). April 1995. 602pp. 9506220172. 84400:126.

Nuclear Reactor Regulation (Post 941001). April 1995. 350pp.

See NUREG-1432,V01,R01 abstract.

9506280476. 84456:001.

This report documents the results of the combined effort of NUREG-1433 V01 Rot: STANDARD TECHNICAL SPECIFICA-the NRC and the industry to produce improved Standud Tech-TIONS GENERAL ELECTRIC PLANTS BWR/4. Specifications.

• nical Specifications (STS), Revision 1 for Westinghouse Plants.

Office of Nuclear Reactor Regulation (Post 941001). April 1995.

The changes reflected in Revision 1 resulted from the experi-388pp. 9506290155. 84461:001.

1

10 Main Citations and Abstracts This report documents the results of the combined effort of NUREG-1435 SO4: STATUS OF SAFETY ISSUES AT LICENSED the NRC and the industry to produce improved Standard Tech-POWER PLANTS.TMl Action Plan Requirements, Unresolved nical Specifications (STS), Revision I for General Electric Safety issues, Generic Safety issuesOther Multiplant Action BWR/4 Plants. The changes reflected in Revision 1 resulted issues.

• Office of Nuclear Reactor Regulation (Post 941001).

from the expenence gained from license amendment applica-December 1994.157pp. 9502080207. 82686:001.

tions to convert to these improved STS or to adopt partial im-As part of ongoing U.S. Nuclear Regulatory Commission provements to existing technical specifications. This report is (NRC) efforts to ensure the quality and accountability of safety the result of extensive public technical meetings and discus-issue information, a program was established whereby an sions between the Nuclear Regulatory Commission (NRC) staff annual NUREG report would be published on the status of li-and various nuclear power plant licensees, Nuclear Steam censee implementation and NRC venfication of safety issues in Supply System (NSSS) Owners Groups, NSSS vendors, and the major NRC requirements areas. This information was compiled Nuclear Energy institute (NEI). The improved STS were devel-and reported in three NUREG volumes. Volume 1, published in oped based on the enteria in the Final Cot imission Policy State.

March 1991, addressed the status of Three Mile Island (TMI) ment on Technical Specifications Improvaments for Nuclear Action Plan Requirements. Volume 2, published in May 1991, Power Reactors, dated July 22,1993. The improved STS will be addressed the status of unresolved safety issues (USIs).

used as the basis for individual nuclear power plant licensees to Volume 3, published in June 1991, addressed the implementa-develop improved plant specific technical spe :ifications. This tion and venfication status of generic safety issues (GSis). Sup-report contains three volumes. Volume 1 contains the Specifica-plement 1, published in December,1991, combined these vol-tions for all chapters and sections of the improved STS. Volume umes into a single report and provided updated information as 2 contains the Bases for Chapters 2.0 and 3.0, and Sections of September 30,1991. Supplement 2, published in December, 3.1 - 3.3 of the improved STS. Volume 3 contains the Buses for 1992, provided updated information on TMI, USI, and GSI Sections 3,4 - 3.10 of the improved STS.

issues and included status of all Other Multiplant Actions (MPAs). Supplement 3, published in December,1993, provided NUREG-1433 V02 R01: STANDARD TECHNICAL SPECFICA-updated information as of September 30, 1993. This annual TIONS GENERAL ELECTRIC PLANTS BWR/4. Bases (Secdons NUREG report provides updated information on TMI, USI, GSI 2.0 - 3.3).

• Office of Nuclear Reactor Regulation (Post and other MPAs as of September 30,1994. The data contained 941001). Apnl 1995. 338pp. 9506290159. 84462:025.

in these NUREG reports are a product of the NRC's Safety See NUREG 1433,V01,R01 abstract.

Issues Management system (SIMs) data base, which is main-NUREG-1433 V03 R01: STANDARD TECHNICAL SPECIFICA.

tained by the Program Management Staff in the Office of Nucle-TIONS GENERAL ELECTRIC PLANTS BWR/4. Bases (Sections ar Reactor Regulation and by NRC regional personnel. This 3.4 - 3.10).

Office of Nuclear Reactor Regulation (Post report is to provide a comprehensive description of the imple-rnentation and venfication status of TMI Action Plan Require-941001). Apnl 1995. 422pp. 9506290164. 84463 001.

See NUREG 1433,V01,R01 abstract.

ments USis, GSis and Other MPAs that have been resolved and li:volve implementation of an action or actions by licensees.

NUREG-1434 V01 R01: STANDARD TECHNICAL SPECIFICA-This report makes the information available to other interested TIONS GENERAL ELECTRIC PLANTS, BWR/6. Specifications.

• parties, including the public. An additional purpose of this Office of Nuclear Reactor Regulation (Post 941001). Apnl 1995.

NUREG repod is to serve as a follow-on to NUREG-0933, "A 412pp. 9506290222. 84466:001.

Priontization of Genenc Safety issues," which tracks safety This report documents the results of the combined effort of issues up until requkements are approved for imposition at h-the NRC and the industry to produce improved Standard Tech-censed plants or until the NRC issues a request for action by nical Specifications (STS), Revision 1 for General Electric licensees BWR/6 Plants. The changes reflected in Revision 1 resulted NUREG-1444 S01: SITE DECOMMCSIONING MANAGEMENT from the experierm ' gained from heense amendment applica-tions to convert ese improved STS or to adopt partial im.

PLAN. FAUVER.D.N.; WEBER M.F.; JOHNSON T.C.; et al. Divi-provements to (

og technical specifications. This report is sion of Waste Management (NMSS 940403). November 1995.

the result of exte,~ ve public technical meetings and discus.

149pp. 9512150121. 86535:001, sions between the Nuclear Regulatory Commission (NRC) staff The Nuclear Regulatory Commission (NRC) staff has identi-and various nuclear power plant licensees, Nuclear Steam fied 51 sites contaminated with radioactive matenal that require Supply System (NSSS) Owners Groups, NSSS vendors, and the spacial attention to ensure timely decommissioning. While none Nuclear Energy Institute (NEI) The improved STS were devel.

of these sites represent an immediate threat to public health oped based on the critena in the Final Commission Policy State.

and safety, they have contamination that exceeds existing NRC ment on Technical Specifications improvements for Nuclear cntena for unrestricted use. All of these sites require some Power Reactors, dated July 22,1993. The improved STS will be degree of remediation, and several involve regulatory issues used as the basis for individual nuclear power plant hcensees to that must be addressed by the Commission before they can be develop improved plant.specihc technical specifications. This released for unrestncted use and the applicable hcenses termi-report contains three volumes. Volume i contains the Specifica, nated. This report contains the NRC staff's strategy for address-tions for all chapters and sections of the improved STS. Volume ing the technical, legal, and policy issues affecting the timely 2 contains the Bases for Chapters 2.0 and 3.0, and Sections decommissioning of the 51 sites and desenbes the status of de-3.1 - 3.3 of the improved STS Volume 3 contains the Bases for commissioning activities at the sites. This is supplement number Sections 3.4 - 3.10 of the improved STS.

one to NUREG-1444, which was published in October 1993.

NUREG-1434 V02 R01: STANDARD TECHNICAL SPECIFICA-NUREG 1464: NRC ITERATIVE PERFORMANCE ASSESSMENT TIONS GENERAL ELECTRIC PLANTS, BWR/6. Bases (Sections PHASE 2.Dovelopment Of Capabilities For Review Of A Per-2.0 33).

Office of Nuclear Reactor Regulation (Post formance Assessment For A High-Level Waste Repository.

I 941001). April 1995. 356pp. 9506290228. 84464:059.

WESCOTT,R.G.; LEE,M.P. Office of Nuclear Matenal Safety &

See NUREG-1434,V01,R01 abstract.

Safeguards. MCCARTIN,T J.; et al. Office of Nuclear Regulatory Research (Post 941217). October 1995. 535pp. 9511160289.

NUREG-1434 V03 R01: STANDARD TECHNICAL SPECIFICA-86233:001.

TIONS GENERAL ELECTRIC PLANTS, BWR/6. Bases 3.4 -

In order to better review a potential license application to 3.10).

• Office of Nuclear Reactor Regulation (Post 941001).

construct and operate a geologic repository for spent nuclear April 1995. 444pp. 9506290230. 84467:049.

fuel and high-level radioactive waste (HLW), the Nuclear Regu-See NUREG 1434,V01,R01 abstract.

latory Commission staff (and its contractor) has expanded and i

i

Main Citations and Abstracts 11 improved its capabihty to conduct performance assessments.

apphcants for NRC operator hcenses pursuant to Part 55 of This report documents the demonstration of the second phase Title 10 of the Code of Federal Regulations (10 CFR Part 55).

of this capability. The demonstration made use of the scenario They are intended to assist NRC examiners and facility licens-selection procedure developed by Sandia National Laboratories ees to better understand the examination process and to to provide a set of scenarios, with corresponding probabilities, ensure the equitable and consistent administration of examina-for use in the consequence analysis of a potential HLW dispos-tions to all applicants. These standards are not a substitute for al site in unsaturated tuff. Models of release of radionuchdes the operator licensing regulations and are subject to revision or from the waste form and transport in ground water, air and by other internal operator examination heensing pokcy changes. As direct pathways provided preliminary estimates of releases to appropriate, these standards will be revised periodically to ac-the accessible environment for a 10,000 year period. The input commodate comments and reflect new information or experi-values of parameters necessary for the consequence models ence.

were sampled numerous times using Latin Hypercube Sampling from assumed probabihty distnbutions. The results from the con-NUREG-1482: GUIDELINES FOR INSERVICE TESTING AT NU-sequence models were then used to generate Complementary CLEAR POWER PLANTS. CAMPBELL,P.L. Office of Nuclear Cumulative Distnbution Functions (CCDFs) for either release to Reactor Regulation, Director (Post 870411). April 1995. 328pp.

the accessible environment or effective dose equivalents to a 9505030476. 83746:022.

target population. CCDFs were calculated for probabikstically in this report, the staff gives bcensees guidelines for develop-significant combinations (scenanos) of four disruptive events; ing and implementing programs for the inservice testing of dnlling, pluvial climate, seismicity and magmatism. Sensitivity pumps and valves at commercial nuclear power plants. The and uncertainty analyses of the calculated releases and effec-report includes U.S. Nuclear Regulatory Commission guidance tive dose equivalents were also used to determine the impor, and recommendations on inservice testing issues. The staff dis-tance of the parameters. Because of the preliminary nature of cusses the regulations, the components to be included in an in-the analysis and the lack of an adequate data base, the results service testing program, and the preparation and content of and conclusions presented in this report should be carefully in-Cold shutdown and refueling outage justifications and requests terpreted. They should not be misconstrued to represent the for relief from the American Society of Mechanical Engineers actual performance of the proposed Yucca Mountain repository Code requirements The staff also gives specific guidance on nor serve as an endorsement of the methods used.

relief acceptable to the NRC and advises hcensees in the use of this information for application at their facilities. The staff dis-NUREG-1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER cusses the revised standard technical specifications for the in-NUCLEAR POWER PLANTS. SOFFER L.; BURSON,S.B.;

service testing program requirements and gives guidance on the FERRELL,C.M.; et al Division of Systems Technology (Post process a hcensee may follow upon finding an instance of non-941217). February 1995. 40pp. 9503140449. S3071:066.

comphance with the Code.

In 1962 the U S Atomic Energy Commission published TID-14844, " Calculation of Distance Factors for Power and Test Re-NUREG-1493: PERFORMANCE-BASED CONTAINMENT LEAK-actors" which specified a release of fission products from the TEST PROGRAM. Final Report. DEY,M. Division of Regulatory core to the reactor containment for a postulated accident in-Applications (Post 941217). SKOBLAR.L.; CHEN P.; et al. S.

volving " substantial meltdown of the core". This " source term",

Cohen & Associates, Inc.

September 1995. 300pp.

the basis for the NRC's Regulatory Guides 1.3 and 1.4, has 9510200161. 85903:001.

been used to determine comphance with the NRC's reactor site The Nuclear Regulatory Commission (NRC) is implementing critena,10 CFR Part 100 and to evaluate other important plant an initiative to ehminate requirements that are marginal to safety performance requirements. Dunng the past 30 years substantial and yet impose a significant regulatory burden on licensees.

additional information on fission product releases has been de-The containment leak-testing requirements for power reactors veloped based on significant severe accident research. This have been identified as one area where performance-based re-document utihzes this research by providing, ore reakstic esti-quirements could replace the current presenptive requirements mates of the " source term" release into conta qment, in terms with only a marginal impact on safety. This technical support of timing. nuchde types, quantities and chem 6c I form, given a documeri (TSD) provides the technical bases for the NRC's severe core-melt accident. This revised " source term" is to be rulemaking to revise leak-testing requirements for nuclear power applied to the design of future hght water reactors C.WRs). Cur-reactors in 10 CFR Part 50, Appendix J. This report identifies rent LWR hcensees may voluntarily propose applications based altomatives to current containment testing requirements which upon it.

would meet the NRC's Safety Goals and achieve greater effi-ciency in the use of resources. Changes in the allowable leak NUREG 1470 V04: FINANCIAL STATEMENT FOR FISCAL YEAR rate for containment and the testing frequencies for both inte-1994.

• Office of the Controller (Post 890205). March 1995.

grated and local leak rate tests are evaluated in terms of both 84pp, 9505180277. 83967:148.

nsk and cost impacts. The feasibihty of applying statistically-The Chief Financial Officer's Act of 1990 requires the NRC based samphng techniques to local leak-rate testing, and the Chief Financial Officer to prepare and submit an annual financial use of on-hne monitonng systems to continuously monitor con-statement to the Director of the Office of Management and tainment integrity are also evaluated.

Budget (OMB). The OMB has replaced the requirement for the CFO's Annual Report with the annual financial statement. The NUREG 1493 DFC: PERFORMANCE-BASED CONTAINMENT annual financial statement was previously included in the Chief LEAK. TEST PROGRAM. Draft Report For Comment. DEY,M. Di-Financial Officer's Annual Report. This report is the fourth vision of Regulatory Applications (Post 941217). SKOBLAR,L.;

annual report for the NRC and includes an overview of the CHEN,P.; et al. S. Cohen & Associates, Inc. January 1995.

NRC, the audited pnncipal financial statements and audit re-260pp. 9503150142. 83100:001.

ports for fiscal year 1994, and supplemental financial and man-Nuclear Regulatory Commission (NRC) is implementing an ini-agement information.

tiative to ehminate requirements that are marginal to safety and NUREG-1478 Rot: NON-POWER REACTOR OPERATOR L1-yet impose a significant regulatory burden on heensees, The CENSING EXAMINER STANDARDS.

• Office of Nuclear Reac-containment leak testing requirements for power reactors have tor Regulation (Post 941001). June 1995.172pp. 9507060351.

been identified as one area where performance-based require-84533:100.

ments could replace the current prescriptive requirements with The Non-Power Reactor Operator Licensing Examiner Stand-only a marginal impact on safety. This technical support docu-ards provide pokcy and guidanco to NRC examiners and estab-ment (TSD) provides the technical bases for the NRC's rule-hsh the procedures and practices for examining hcensees and making to revise leak-testing requirements for nuclear power re-1

12 Main Citations and Abstracts tctors in 10 CFR Part 50, Appendix J. This report identifies al-TIONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS ternatives to current contairiment testing requirements which FOR VARIOUS CONTAMINANTS AND FIELD would meet the NRC's Safety Goals and achieve greater effi-CONDITIONS. Draft Report For Comment. HUFFERT,A.M. Divi-ciency in the use of resources. Changes in the allowable leak sion of Regulatory Apphcations (Post 941217).

rate for containment anj the testing frequencies for both inte-ABELOUIST,E.W.

Oak Ridge Associated Universities.

grated and local leak ra'e tests are evaluated in terms of both BROWN.W.S. Brookhaven National Laboratory. August 1995.

nsk and cost impacts. h 9 feasibihty of applying statistically-160pp. 9509200367. 85540:156.

based sampling techniques a local leak-rate testing, and the This report desenbes and quantitatively evaluates the effects use of on-line monitoring syste ns to continuously monitor con-of various factors on the detection sensitivity of commercially tainment integnty are also evalut ted.

available portable field instruments being used to conduct radio-logical surveys in support of decommissioning. NRC is currently NUREG-1505 DRFT FC: A NONPARAMETRIC STATISTICAL involved in a rulemaking effort to establish residual contamina-METHODOLOGY FOR THE DESIGN AND ANALYSIS OF tion entena for release of facilities for restricted or unrestncted FINAL STATUS DECOMMISSIONING SURVEYS. Draft Report use. In support of that rulemaking, NRC has prepared a draft Genenc Environmental impact Statement (GEIS), consistent For Comment. GOGOLAK C.V. Energy, Dept.of, Environmental with the National Environmental Policy Act. The effects of this Measurements Laboratory. HUFFERT.A.M.; POWERS,G E. Divi-new rulemaking on the overall cost of decommissioning are sion of Regulatory Applications (Post 941217). August 1995.

among the many factors considered in the GEIS. The overall i

216pp. 9509200281. 85541:080.

cost includes the costs of decontamination, waste disposal, and This report describes a nonparametric statistical methodology radiological surveys to demonstrate compliance with the appli-for the design and analysis of final status decommissioning sur-cable guidelines. An important factor affecting the costs of such veys in support of the proposed rulemaking on decommission-surveys is the minimum detectable concentrations (MDCs) of ing. The techniques desenbed are alternatives to the existing field survey instruments in relation to the residual contamination parametric statistical methodology in NRC draft NUREG/CR-guidelines. The purpose of this study was two-fold. First, the 5849, " Manual for Conducting Radiological Surveys in Support data were used to determine the validity of the theoretical of License Termination." Proposed nonparametnc statistical MDCs used in the draft GEIS. Second, the results of the study, methods for testing comphance with decommissioning cntena published herein, provide guidance to hcensees for (a) selection are provided for radionuclides which occur in natural back-and proper use of portable survey instruments and (b) under-ground and for those that do not occur in natural background.

Standing the field conditions and the extent to which the capa-The tests considered apphcable are the Wilcoxon Signed Ranks bihties of those instruments can be limited. Such instruments as test, Sign test, and Quantile test for the analysis of a single gas proportional, Geiger-Mueller, zinc sulfide, and sodium iodide data set, and the Wilcoxon Rank Sum test and a Quantile test detectors were evaluated.

for companng two independent data sets. An Elevated Meas-NUREG-1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE urement Companson is also desenbed to deal with any unusual-FOR A NUCLEAR POWER PLANT. WEINSTEIN.E.D.;

ly high observations. This report contains information on the BATES.E.F. Office for Analysis & Evaluation of Operational Data Quality Objectives process as it relates to the planning Data, Director. ADLER,M.V.; et al. Oak Ridge National Laborato-and analysis of final site surveys. The proposed process in-ry. March 1995. 78pp. 9504180482. 83570:007.

cludes methods for determining the number of samples needed Tabletop exercises are held to discuss issues related to the to obtain statistically valid compansons with decommissioning response of organizations to an emergency event. This docu-cntena and the methods for conducting the statistical tests with ment desenbes in task format the planning, conduct, and report-the resulting sample data.

ing of lessons leamed for a large interagency tabletop. A sample scenano, focus area, arid discussion questions based NUKEG-1506 DRFT FC: MEASUREMENT METHODS FOR RADI-on a simulated accident at a commercial nuclear power plant OLOGICAL SURVEYS IN SUPPORT OF NEW DECOMMIS-are provided.

SiONING CRITERIA. Draft Report For Comment. HUFFERT,A.M.

Division of Regulatory Applications (Post 941217). MILLER,K.M.

NUREG-1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MA-Energy, Dept.of, Environmental Measurements Laboratory.

TERIAL SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft August 1995. 78pp. 9509200363. 85541:001.

Report For Comment.

CAMPER,L.W.;

SCHLUETER.J.;

This report desenbes proposed methodologies for measunng HENDERSON,P.; et al. Division of Industnal & Medical Nuclear low-level radiation and radioactivity that could be used in con-Safety (Post 870729). January 1995. 178pp. 9502080136.

826631 52.

ducting surveys associated with decommissioning of kcensed NRC facilities. Guidance on survey planning within the context A Tad he ongmah composed d smn E Mar of the Data Quality Objective approach and on specific instru-Regulatory Commission and two Agreement State program statt developed the guidance contained in this report. The purpose of mentation for measurements of gross and nuchde-specific rade this report is to describe a systematic approach for effective ation and radioactivity is given. Scannin9, direct measurements' management of radiation safety programs at medical facilities.

and sampling are discussed in terms of the application to par.

This is accomplished by emphasizing the roles of institution ex.

ticular measurement locations. The basic survey meter tech-ecutive management, radiation safety committee, and radiation naques thst are commonly used at present are outlined and safety officer Vanous aspects of program management are dis-more detsited information is given on the capabilities and apph-cussed and include guidance on selecting the radiation safety cation of in situ spectrometnc techniques for providing high sen-officer, determining adequate resources for the program, the sitmty for individual photon-emitting radionuchdes. The use of use of contractual services such as consultants and service

)

vanous techniques in concert is recommended, as the different companies, the conduct of audits, the roles of authorized users measurements, taken collectively, serve as a quahty control and supervised individuals, NRC's reporting and notification re-check. The methodologies desenbed provide the means to quirements, and a general desenption of how NRC's licensing, measure residual radionuclides at concentrations corresponding inspection, and enforcement programs work. Appendices pro-to the proposed decommissioning cntena which are in the range vide detailed guidance on specific aspects of a radiation safety j

of 3 to 15 mrem per year for unrestricted release of a f acility.

program and the glossary defines terms used throughout the j

report. The guidance contained herein does not represent new NUREG-1507 DRFT FC: MINIMUM DETECTABLE CONCENTRA.

or proposed regulatory reqarements and bcensees will not be inspected against any portion of it. Addztsonally. regulatory com-

~~

. ~ _.

l l

Main Citations and Abstracts 13 pliance with all applicable regulations is not assured by 1cens-would help detect, alleviate, and correct the degraded condi-ees who adopt any portion of, or apply the principles desenbed tions of the structures and civil engineenng features.

in this report.

NUREG-1525: ASSESSMENT OF THE NRC ENFORCEMENT NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT AL.

PROGRAM. LIEBERMAN J. Ofc of Enforcement (Post 870413).

LEGATIONS REVIEW TEAM. KOKAJKO LE.; SKAY,D.M.;

COBLENTZ,L. Walnut Creek Field Ofc, R4 (Post 940404). April WANG,H-B.; et al. Office of Nuclear Reactor Regulation (Post 1995. 200pp. 9505040381. 83784:001.

941001). March 1995. 222pp. 9504180323. 83565:001.

On May 13,1994, the Nuclear Regulatory Commission's This report provides the results of the South Texas Project Al.

(NRC's) Executive Director for Operations established a Review l

legations Review Team of the U.S. Nuclear Regulatory Commis.

Team to Assess the NRC Enforcement Program. The team t

sion. This team was formed to obtain and review allegations evaluated the current system, and sohcited comments from vari-from indviduals represented by three attorneys who had con.

ous NRC Offices, other Federal agencies, members of industry, tacted Congressional staff members. The allegers were em.

and the public. This report presents the team's assessment.

ployed in various capacities at South Texas Project Electric The report summanzes current processes and suggests certain Generating Station, licensed by Houston Lighting and Power changes. It proposes: (1) increased clanty, focus, and simplicity Company, et al.; therefore, the allegations are confined to this in the enforcement program; (2) retention of four seventy levels site. The South Texas Project Allegations Review Team re-of violations, with a clear focus on safety; (3) holding enforce-viewed, referred, and dispositioned concerns related to discrimi-ment conferences only when needed, clarifying their status as natory issues (harassment and intimidation), falsification of predecisional, and making open conferences the norm; (4) a records and omission of information, and various technical streamlined civil penalty assessment process, with fewer deci-issues. The team was able to substantiate certain technical sional points and limited outcomes, and the use of discretion issues of minor safety significance or regulatory concern at the where appropriate; and (5) implementation changes to increase South Texas Project facility, but it did not find widespread dis.

efficiency. Recommendations are given in each area.

criminatory practices such as harassment and intimidation.

NUREG-1526: LESSONS LEARNED FROM EARLY IMPLEMEN.

TATION OF THE MAINTENANCE RULE AT NINE NUCLEAR NUREG-1519: SURFACE INTERACTIONS OF CESIUM AND POWER PLANTS. PETRONE,C.D.; CORREIA.R.P.; BLACK,S.C.

BORIC ACID WITH STAINLESS STEEL. GROSSMAN,N. Divi-Office of Nuclear Reactor Regulation (Post 941001). June 1995.

sion of Systems Technology (Post 941217). August 1995.

44pp. 9507060341. 84532:001.

104pp, 9509130158. 85432:009.

This report summarizes the lessons learned from the nine In this report, the effects of cesium hydroxide and boric acid pilot site visits that were performed to review early implementa-on oxidized stainless steel surfaces at high temperatures and tion of the maintenance rule using the draft NRC Maintenance f

near one atmosphere of pressure are investigated. This is the Inspection Proceaure. Licensees followed NUMARC 93-01, "In-i first expenmental investigation of this chemical system. The ex.

dustry Guidehne for Monitoring the Effectiveness of Mainte-penmental investigations were performed using a mass spec.

nance at Nuclear Power Plants." In general, the licensees were

(

trometer and a mass electrobalance. Surfaces from the different thorough in determining which structures, systems, and compo-l experiments were examined using a scanning electron micro-nents (SSCs) were within the scope of the maintenance rule at scope to idontify the presence of deposited species, and elec.

each site. The use of an expert panel was an appropriate and tron spectroscopy for chemical analysis to identify the species practical method of determining which SSCs are nsk significant.

deposited on the surface. The analysis couples vaporization, When setting goals, all heensees considered safety but many h-I deposition, and desorption of the compounds formed under consees did not consider operating expenence throughout the conditions similar to what is expected during certain nuclear re-industry. Although required to do so, licensees were not moni-actor accidents. Cesium deposits on an oxidized stainless steel toring at the system or train level the performance or condition surface at temperatures between 1000 and 1200 Kelvin and on for some systems used in standby service but not significant to stainless steel surfaces coated with boric oxide in the same risk. Most licensees had not established adequate monitonng of i

temperature ranges. The mechanism for such deposition in-structures under the rule. Licensees established reasonable i

volved the chemical reaction between cesium and chromate.

plans for doing periodic evaluations, balancing unavailability and Some revaporization in the cesium hydroxide-boric acid system reliability, and assessing the effect of taking equipment out of was observed. Under the conditions given, boric acid will react service for maintenance. However, these plans were not evalu-with cesium hydroxide to form cesium metaborate. A model is ated because they had not been fully implemented at the time proposed for this chemical reaction.

of the site visits.

l NUREG 1522: ASSESSMENT OF INSERVICE CONDITIONS OF NUREG-1527: NRC'S OBJECT ORIENTED SIMULATOR IN-SAFETY-RELATED NUCLEAR PLANT STRUCTURES.

STRUCTOR STATION. GRIFFIN,J l.; GRIFFIN,J.P. Technical ASHAR,H.; BAGCHl,G. Office of Nuclear Reactor Regulation Training Center. June 1995.107pp. 9506280427, 84469.073.

(Post 941001). June 1995.116pp. 9508230244. 85127:001.

As part of a comprehensive simulator upgrade program, the l

The report is a compilation from a number of sources of infor-s mulator computer systems associated with the Nuclear Regu-mation related to the condition of structures and civil engineer.

latory Commission's (NRC) nuclear power plant simulators were i

ing features at operating nuclear power plants in the U.S. The replaced. Because the original instructor stations for two of the most significant information came from the hands-on inspection simulators were dependent on the original computer equipment, of the six old plants (licensed prior to 1977) performed by the it was necessary to develop and implement new instructor sta-staff of the Civil Engineenng and Geosciences Branch in the Di-tior's. This report describes the Macintosh-based Instructor Sta-vision of Engineenng of the Office of Nuclear Reactor Regula-tions developed by NRC engineers for the General Electric (GE) tion. For the containment structures, most of the information re-and Babcock and Wilcox (B&W) simulators, lated to the degraded conditions came from the heensees as part of the Licensing Event Report System (10 CFR 50.73), or NUREG-1528: RECONSTITUTION OF THE MANUAL CHAPTER as part of the requirement under limiting condition of operation 2512 CONSTRUCTION INSPECTION PROGRAM FOR WATTS i

of the plant specific Technical Specifications. Most of the infor.

BAR UNIT 1. PERANICH.M.; FREDRfCKSON,P.; JAUDON.J.

l mation related to the degradation of other structures and civil Region 2 (Fost 820201). September 1995.683pp.9510310364.

engineering features was extracted from the industry survey, the 86016:001.

reported incidents, and the plant visits. The report discusses ttie This document provides information on the concepts, per-condition of the structures and civil engineenng features at op-formance, and results of the inspection Manual Chapter (MC) erating nuclear power plants and provides information that 2512 Light Water Reactor inspection Reconstitution Program for

14 Main Citations and Abstracts the construction phase of Watts Bar Nuclear Plant, Unit 1 This document includes the U.S. Nuclear Regulatory Commis-(WBNP 1). The U S Nuclear Regulatory Commission (NRC),

sion's (NRC's or Commission's) revised General Statement of Region 11, conducted the reconstitution as a follow-up to the ini-Policy and Procedure for Enforcement Actions (Enforcement taal MC 2512 inspection Program, completed in 1985. Through Policy) as it was published in the Federal Register on June 30, this initial inspection program, the NRC identified several prob-1995 (60 FR 34381). This document also includes the notice lems with the quality of construction at the facility, as well as announcing the removal of the Enforcement Policy from the weaknesses in the corrective actions taken by the licensee to Code of Federal Regulations (60 FR 34380; June 30,1995).

resolve those problems Subsequently allegations and employ-The Enforcement Policy is a general statement of policy ex-ees concerns echoed these findings, raising questions regarding plaining the NRC's policies and procedures in initiating enforce-construction quality. The NRC decided in 1994 that a real-time ment actions, and of the presiding officers and the Commission correlation or " reconstitution" was needed to venfy that the n reviewing these actions. This policy statement is applicable to final construction-related plant inspections met the requirements enforcement in matters involving the radiological health and of the MC 2512 Inspection Program. The MC 2512 Reconstitu-tion Program successfully validated completion of the WBNP 1 safety of the public, including employees' health and safety, the construction inspection program. It also inspires confidence in common defense and secunty, and the environment. This state-the effectiveness of the licensee's corrective actions in resolv-ment of general policy and procedure is published as NUREG-I i

ing construction problems and enhancing the quality of plant I

construction Successful reconstitution is integral to the overall Enforcement Policy. However, this is a pclicy statement and not reasonable assurance assessment of the readiness for WBNP 1 a regulation. The Commission may deviate from this statement to be licensed as an operating facility.

of policy and procedure as appropnate under the circumstances of a particular case.

NUREG 1530: REASSESSMENT OF NRC'S DOLLAR PER PERSON. REM CONVERSION FACTOR POLICY.

• Division of NUREG/CP-0140 V01: PROCEEDINGS OF THE TWENTY-Regulatory 9nhcations (Post 941217). December 1995. 23pp.

SECOND WATER REACTOR SAFETY INFORMATION 9601290154. 86877:334.

MEETING. Plenary Session, Advanced instrumentation & Control The U S. Nuclear Regulatory Commission (NRC) has complet-Hardware & Software, Human Factors Research, IPE & PRA.

ed a review and analysis of its dollar per person-rem conversion MONTELEONE,S. Brookhaven National Laboratory. April 1995 factor policy. As a result of this review, the NRC has decided to 403pp. 9505230106. 84016:001.

adopt a $2000 per person-rem conversion factor, subject it to This three-volume report contains papors presented at the present worth considerations, and limit its scope solely to health Twenty-Second Water Reactor Safety Information Meeting held effects. This is in contrast to the previous policy and staff prac-at the Bethesda Marnott Hotel, Bethesda, Maryland, dunng the tice of using an undiscounted $1000 per person-rem conversion week of October 24-26, 1994. The papers are pnnted in the factor that served as a surrogate for all offsite consequences order of their presentation in each session and descnbe (health and offsite property) The policy shift has been incorpo~

rated in " Regulatory Analysis Guidelines of the U.S. Nuclear rogress and results of programs in nuclear safety research Regulatory Commission, NUREG/BR-0058, Revision 2, No-conducted in this country and abroad. Foreign participation in vember 1995.

the meeting included papers presented by researchers from Fin-land, France, Italy, Japan, Russia and United Kingdom. The NUREG-1535: INGESTION OF PHOSPHORUS-32 AT MASSA.

titles of the papers and the names of the authors have been CHUSETTS INSTITUTE OF updated and rnay differ from those that appeared in the final T ECHNOLOGY, CAM BRIDG E,M ASS ACHUSETTS.lDENTIFIE D program of the meeting ON AUGUST 19,1995.* Incident Investigation Team Decem-ber 1995.129pp. 9512260311. 86626.001.

NUREG/CP-0140 V02: PROCEEDINGS OF THE TWENTY-On Monday October 16,1995, the Massachusetts Institute of SECOND WATER REACTOR SAFETY INFORMATION Technology (MIT, the hcensee) notified the U S. Nuclear Regu-MEETING. Severe Accident Research, Thermal Hydrauhc Re-latory Commission (NRC) of an incident involving ingestion of search For Advanced Passive LWRs, High-Burnup Fue! Behav-phosphorus 32 by a researcher at the MIT Center for Cancer sor. MONTELEONE.S. Brookhaven National Laboratory, April Research. The licensee informed the NRC that a researcher 1995. 430pp. 9505230109. 84018:001, had reported the incident on August 19. The licensee initially See NUREG/CP-0140,V01 abstract.

estimated the intake as 500 microcunes (19 MBq) and the dose as 4000 millirem (40 mSv) to the individual On October 12, the NUREG/CP-0140 V03: PROCEEDINGS OF THE TWENTY-licensee informed the researcher that its final intake estimate SECOND WATER REACTOR SAFETY INFORMATION was 579 microcunes (21 MBq), just under the 600 microcunes MEETING Primary Systems Integnty, Structural And Seiscnic En-(22 MBq) which would represent an overexposure. On October gineenng, Aging Research, Products And Applications.

17, the NRC established an incident investigation Team to in-MONTELEONE,S. Brookhaven National Laboratory. Apnl 1995.

vestigate the case. NRC also contracted with Lawrence Liver.

265pp. 9505230115. 84019 068 more National Laboratory and Ook Ridge Institute for Science See NUREG/CP-0140,V01 abstract.

and Education to do independent dose assessments of the unne sample data and the whole-body data. The Team conclud-NUREG/CP-0141: PROCEEDINGS OF THE 23RD DOE /NRC NU-ed that the licensee's final intake and dose estimates were in CLEAR AIR CLEANING CONFERENCE. Held In Buffalo,New accordance with accepted scientific references and NRC guid-York, July 25-28,1994. FIRST,M.W. Harvard School of Pubhc ance. However, recognizing the uncertainties involved in the Health. Boston, MA. February 1995. 800pp. 9503140367.

use of models to simulate human charactenstics, the Team de-CONF-940738 83072:114 terrnined the intake would be better charactenzed as hkely fall-ing within a range of 500 to 750 microcunes (19-28 MBq). An This report contains the papers presented at the 23rd DOE /

NRC medical consultant concluded that no symptoms or acute NRC Nuclear Air Cleaning Conference and the associated dis-effects should be observed from an intake of this level.

cussions. Major topics are: (1) nuclear air cleaning codes, (2) nuclear waste, (3) filters and filtration, (4) effluent stack monitor.

NUREG-1600: GENERAL STATEMENT OF POLICY AND PROCE.

ing, (5) gas processing (6) adsorption, (7) air treatment sys-DURE FOR NRC ENFORCEMENT ACTIONS Enforcement tems, (8) source terms and accident analysis, and (9) fuel re-Pohey

• Ofc of Enforcement (Post 870413). July 1995. 35pp.

processing.

9507250299. 84831:142.

l Main Citations and Abstracts 15 NUREG/CP-0142 V01: PROCEEDINGS OF THE 7TH INTERNA-NUREG/CP-0144 V01: A WORKSHOP ON DEVELOPING RISK TlONAL MEETING ON NUCLEAR REACTOR THERMAL HY.

ASSESSMENT METHODS FOR MEDICAL USE OF RADIOAC-DRAULICS (NURETH-7). Sessions 15. BLOCK,R.C., FEINER.F.

TIVE MATERIAL. Summary. TORTORELLt,J.P. Idaho National Amencan Nuclear Society. September 1995.

509pp.

Engineenng Laboratory. August 1993. 122pp. 9508230351 9509110196. 85399 001.

INEL-94/0111. 85118:013.

This volume includes papers presented at the 7th Internation.

A workshop was held at the Idaho National Engineenng Labo-al Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH 7) ratory, August 16-18, 1994, on the topic of nsk assessment on September 10-15, 1995 at Saratoga Spnngs, N.Y. The following medical devices that use radioactive isotopes. Its purpose was to review past efforts to develop a nsk assessment truthodolo-subjects are discussed: progress in analytical and experimental gy to evaluate these devices and to develop a program plan work on the fundamentals of nuclear thermal-hydraulics, the de-and scoping document for future development. This report pre-velopment of advanced mathematical and numerical methods, sents a summary of that workshop, related technical papers, and the application of advancements in the field in the develop-presentation matenal, and a transcnpt of the workshop. Partici-ment of novel reactor concepts. Also combined issues of ther-pants included experts in the fields of radiation oncology, med:-

mal-hydraulics and reactor / power plant safety, core neutronics cal physics, nsk assessment, humun-error analysis, and human and/or radiation.

factors. Staff from the U.S. Nuc: ear Regulatory Commission assocated w@ N mgda%n of Mcal uses d ra@-

NUREG/CP-0142 V02: PROCEEDINGS OF THE 7TH INTERNA.

TIONAL MEETING ON NUCLEAR REACTOR THERMAL-HY-ods participated in the workshop. The workshop participants DRAULICS (NURETH-7). Sessions 6-11.

BLOCK,R.C.;

concurred in NRC's intended use of nsk assessment as an im-FEINER,F. Amencan Nuclear Society. September 1995.806pp.

portant technology in the development of regulations for the 9509110202. 85402:001-medical use of radioactive matonal and encouraged the NRC to See NUREG/CP-0142,V01 abstract.

proceed rapidly with a pilot study. Specific recommendations are included in the executive summary and the bo@ of this NUREG/CP-0142 V03: PROCEEDINGS OF THE 7TH INTERNA-report.

TIONAL MEETING ON NUCLEAR REACTOR THERMAL-HY.

DRAULICS (NURETH-7) Sessions 12 16.

BLOCK,R C.;

NUREG/CP-0144 V02: A WORKSHOP ON DEVELOPING RISK FEINER,F. Amencan Nuclear Society. September 1995.916pp.

ASSESSMENT METHOD 6 FOR MEDICAL USE OF RADICAC-9509110205. 85386 001.

TIVE MATERIAL.Supportng Documents. TORTORELL1,J.P.

See NUREG/CP-0142,V01 abstract.

Idaho National Engineenng Laboratory. August 1995. 330pp.

9508230354. INEL-94/0111. 85119:001.

NUREG/CP-0142 V04: PROCEEDINGS OF THE 7TH INTERNA-See NUREG/CP-0144,V01 abstract.

TIONAL MEETING ON NUCLEAR REACTOR THERMAL-HY-NUREG/CP-0145-WORKSHOP ON DEVELOPING SAFE DRAULICS (NURETH 7). Sessions 17-24.

BLOCK,R.C.;

SOFTWARE. Held At Hotel Del Coronado, San Diego,CA, July 22-FEINER F. Amencan Nuclear Society. September 1995.845pp.

23,1992. LAWRENCE,J.D. Lawrence Livermore National Labo-9509110210. 85396.001',V01 abstract.

ratory. November 1994. 30pp. 9502080047. 82644:318.

See NUREG/CP 0142 The WorMhop on Developing Safe Software was held July 22-23, 1992, at the Hotel dei Loronado, San Diego, California.

NUREG/CP-0143: PROCEEDINGS OF THE THIRD INTERNA-The purpose of the workt. hop was to have four world experts TIONAL WORKSHOP ON THE IMPLEMENTATION OF ALARA discuss among themselvos software safety issues which are of AT NUCLEAR POWER PLANTS Held At Hauppau9e, Lon9 interes! to the U.S. Nuclear Regulatory Commission. These Island, New York. KHAN.T.A. Brookhaven National Laboratory.

ssuet concern the development of software systems for use in March 1995. 822pp. 9503270314. BNL-NUREG 52440.

nuclear power plant protection systems. The workshop com-83266.001, pf, sed four sessions. Wednesday morning, July 22, consisted of This report contains the papers presented and the discus-p,esentations from each of the four panel members. On sions that took place at the Third International Workshop on Wednesday afternoon, the panel members went through a inst ALARA implementation at Nuclear Power Plants, held in Haup-of possible software development techniques and commented pauge, Long Island, New York from May 8-11, 1994. The work-on thern. The Thursday morning, July 23, session consisted of shop brought together scientists, engineers, health physicists, an extended discussion among the panel members and the ob-regulators, managers and others who are involved with occupa-servers from the NRC. A final session on Thursday afternoon tional dose control and ALARA issues. One-hundred and seven-consisted of a discussion among the NRC observers as to what ty five persons from eleven countnes attended the workshop was leamed from the workshop.

The countnes represented were Canada, Finland, France, Ger-NUREG/CP-0146: PROCEEDINGS OF THE WORKSHOP ON many, Japan, Korea, Mexico, the Netherlands, Spain, Sweden, GATE VALVE PRESSURE LOCKING AND THERMAL BINDING.

the United Kingdom and the United States. The workshop was BROWN.E.J. Division of Safety Programs (Post 870413). July organized into twelve sessions and three panel discussions.

1995. 268pp. 9508090040. 84964:001.

The topics were as follows: Session 1, Controlling Radiation The purpose of the Workshop on Gate Valve Pressure Lock-Fields; Session 2, Panel Discussion on Recent Recommenda-ing and Thermal Dinding was to discuss pressure locking and tions on Dose Limitation; Session 3, Presentations and Panel thermal binding issues that could lead to inoperable gate valves Discussion on ALARA in New Reactors; Session 4, Pathways to in both boiling water and pressurized water reactors. The goal ALARA: Session 5 Panel Discussion on Econcmics Versus Ex-was to foster exchange of information to develop the technical cellence; Session 6 Short Presentations on ALARA Implemen-bases to understand the phenomena, identify the components tation; Session 7A, PWR and CANDU Presentations; Session that are susceptible, discuss actual events, discuss the safety 7B; BWR and Gas Cooled Presentations 1; Session 8A, PWR significance, and illustrate known corrective actions that can and CANDU Presentations; Session BB. BWR and Gas-Cooled prevent or limit the occurrence of pressure locking or thermal Presentations, Session 9 Decommissiuning of Nuclear Power binding The presentations were structured to cover U.S. Nucle-Plants; Session 10 Decontamination of Nuclear Power Plants, ar Regulatory Commission staff evaluation of operating experi-and Session 11, Robotics and Remute Handling. The workshop ence and planned regulatory activity; industry discussions of was sponsored jointly by the U.S Nuclear Regulatory Commis-specific events, including foreign experience, and efforts to de-soon and the Brookhaven Natioral Laboratory's ALARA Center.

termine causes and alleviate the affects; and valve vondor ex-

I i

l 16 Mrin Cit'.tions cnd Abstracts j

t penence and recommended corrective action. The discussions NUREG/CR-0200 V2P1R4: SCALE: A MODULAR CODE indicated that identifying valves susceptible to pressure locking SYSTEM FOR PERFORMING STANDARDlZED COMPUTER and thermal binding was a complex process involving knowl.

ANALYSES FOR LICENSING EVALUATION. Functional Modules

)

edge of components, systems, and plant operations. The cor.

F1-F8. PARKS.C.V. Oak Ridge National Laboratory. April 1995.

rective action options are vaned and straightforward.

670pp. 9506010509. ORNL/NUREG/CSD2. 84124:266.

1 See NUREG/CR-0200,V01 R04 abstract.

NUREG/CP-0147: PROCEEDINGS OF THE WORKSHOP ON THE ROLE OF NATURAL ANALOGS IN GEOLOGIC DISPOS-NUREG/CR-0200 V2P2R4: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDtZED COMPUTER AL OF HIGH LEVEL NUCLEAR WASTE. Held in San ANALYSES FOR LICENSING EVALUATION. Functional Modules Antonio, Texas, July 22 25,1991, KOVACH L.A. Division of Regu-F9-F16. PARKS C.V. Oak Ridge National Laboratory. April 1995.

latory Applications (Post 941217). MURPHY,W.M. Southwest 69pp. N1052t @NWREWCSD2. 84128:273.

Research Institute. September 1995. 131pp. 9510310349.

j ee N OW,R04 abstract.

CNWRA 93 020. 86024:001, A Workshop on the Role of Natural Analogs in Geologic Dis-NUREG/CR-0200 V3 R04: SCALE: A MODULAR CODE SYSTEM posal of High-Level Nuclear Waste was held in San Antonio, FOR PERFORMING STANDARDIZED COMPUTER ANALYSES Texas on July 22-25, 1991. The proceedings comprise seven.

FOR LICENSING EVALUATION. Miscellaneous. PARKS,C.V.

teen papers submitted by participants at the workshop. A senes Oak Ridge National Laboratory. Apnl 1995.793pp.9506010519.

{

of papers addresses the relation of natural analog studies to the ORNL/NUREG/CSD2. 84126:208.

i regulation, performance assessment, and liconsing of a geolog-See NUREG/CR-0200,V01,R04 abstract.

ic repository. Applications of reasoning by analogy are illustrat-NUREG/CR-2850 V13: DOSE COMMITMENTS DUE TO RADIO-ed in papers on the role of natural analogs in studies of earth-ACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES quakes, petroleum, and mineral exploration. A summary is pro-IN 1991, BAKER,D.A. Battelle Memorial Institute, Pacific North-vided of a recently completed, intemationally coordinated natu-ral analog study at Pogos de Caldas, Brazil. Papers also cover west Laboratory. April 1995.192pp. 9505090280. PNL-4221.

83849.001 problems and applications of natural analog studies in four tech-Population and individual radiation dose commitments have nical areas of nuclear waste management: waste form and been estimated from reported radionuchde releases from com-waste package, near-field processes and environment, far-field mercial power reactors operating during 1991. Fifty-year dose processes and environment, and volcanism, and tectonics.

Summaries of working group dekberations in these four techni-commitments for a one-year exposure from both liquid and at-mospheric releases were calculated for four population groups cal areas provide reviews and proposals for natural analog ap' (infant, child, teenager, and adult) residing between 2 and 80 plications.

km from each of 72 reactor sites. This report tabulates the re-NUREG/CP-0148: TRANSACTIONS OF THE TWENTY-THIRD sults of these calculations, showing the dose commitments for both water and airbome pathways for each age group and WATER REACTOR SAFETY INFORMATION MEETING.

organ. Also included for each of the sites is an estimate of indi-MONTELEONE,S. Brookhaven National Laboratory. September 1995. 76pp. 9510060240. 85746:182.

v dual doses that are compared with 10 CFR Part 50, Appendix I design objectives. The total collective dose commitments This report contains summaries of papers on reactor safety (from both liquid and airbome pathways) for each site ranged research to be presented at the 23rd Water Reactor Safety In-from a high of 22 person-rem to a low of 0.002 person-rem for formation Meeting at the Bethesda Marriott Hotel, Bethesda-the sites with plants in operation and producing power dunng

)

Maryland, October 23-25,1995. The summaries briefly desenbe the year. The anthmetic mean was 1.2 person-rem. The total the program and results of nuclear safety research sponsored population dose for all sites was estimated at 88 person-rem for by the Office of Nuclear Regulatory Research, U S. NRC. Sum-the 130 million people considered at risk. The individual dose maries of invited papers concerning nuclear safety issues from commitments estimated for all sites below the Appendix l U.S. government laboratories, the electnc utilities, the nuclear design objectives.

Industry, and from foreign governments and industry are also in-ciuded. The summaries have been compiled in one report to NUREG/CR-2907 V13: RADIOACTIVE MATERIALS RELEASED provide a basis for meaningful discussion and information ex.

FROM NUCLEAR POWER PLANTS. Annual Report 1992.

change during the course of the meeting and are given in the TICHLER.J.; DOTY,K.; LUCADAMO.K. Brookhaven National order of their presentation in each session.

Laboratory. August 1995. 350pp. 9509130181. BNL-NUREG-51581.85431:001.

NUREG/CR 0200 V1 R04: SCALE: A MODULAR CODE SYSTEM Releases of radioactivo materials in airborne and hquid ef.

FOR PERFORMING STANDARDIZED COMPUTER ANALYSES fluents from commercial hght water reactors during 1992 have FOR LICENSING EVALUATION Control Modu'les. PARKS,C.V.

been compiled and reported. Data on sohd waste shipments as Oak Ridge National Laboratory. April 1995.781pp.9506010500.

well as selected operating information have been included. This ORNL/NUREG/CSD2. 84122:213.

report supplements earlier annual reports issued by the former SCALE-a modular code system for Standardized Computer Atomic Energy Commission and the Nuclear Regulatory Com.

Analyses Licensing Evaluation-has been developed by Oak mission. The 1992 release data are summanzed in tabular form.

)

Ridge National Laboratory at the request of the U.S. Nuclear Data covering specific radionuchdes are summanzed.

Regulatory Commission. The SCALE system utikzes well estab-NUREG/CR-2907 V14: RADIOACTIVE MATERIALS RELEASED lished computer codes and methods within standard analysis FROM NUCLEAR POWER PLANTS. TICHLER.J.; DOTY,K.;

sequences that (1) allow an input format designed for the occa-LUCADAMO,K. Brookhaven National Laboratory. December sional user and/or novice, (2) automate the data processing 1995.351pp.9601290279. BNL-NUREG-51581. 86880:001, and coupling between modules, and (3) provide accurate and Releases of radioactive matenals in airborne and liquid ef.

reliable results. System development has been directed at prob-fluents from commercial light water reactors dunng 1993 have lem-dependent cross-section processing and analysis of cntical-been compiled and reported. Data on sohd waste shipments as ity safety, shielding, heat transfer, and depletior./ decay prob-well as selected operating information have been included. This lems. Since the initial release of SCALE in 1980, the code report supplements earlier annual reports issued by the former system has been heavily used for evaluation of nuclear fuel fa-Atomic Energy Commission and the Nuclear Regulatory Com-1 cility and package designs. This revision documents Version 4.2 mission. The 1993 release data are summanzed in tabular form.

l j

of the system.

Data covenng specific radionuchdes are summarized.

l J

l Main Citations and Abstracts 17

{

NUREG/CR 3243: COMPARISON OF ASME CODE FATIGUE NUREG/CR-4219 Vit N1: HEAVY-SECTION STEEL TECHNOL.

EVALUATION METHODS FOR NUCLEAR CLASS 1 PIPING OGY PROGRAM. Semiannual Progress Report For October WITH CLASS 2 OR 3 PIPING. RODABAUGH.E.C. E.C. Roda-1993 - March 1994. PENNELL W.E. Oak Ridge National Labora-baugh Associates, Inc. RODABAUGH,E.C. Oak Ridge National tory. November 1995. 97pp. 9512260324. ORNL/TM-9593.

Laboratory. June 1983.

59pp.

9511160312.

ORNL-86627:001.

SUB82222521. 86249:160.

The Haavy-Section Steel Technology (HSST) Program is con-The fatigue evaluation procedure used in= the ASME Boiler ducted for the U.S. Nuclear Regulatory Commission (NRC) by and Pressure Vessel Code, Section Ill, Nuclear Power Plant Oak Ridge National Laboratory (ORNL). The program focus is Components, for Class 1 piping is different from the procedure on the development and validation of technology for the as-used for Class 2 or 3 piping. The basis for each procedure is sessment of fracture-prevention margins in commercial nuclear desenbed and correlations between the two procedures are pre-reactor pressure vessels. The HSST Program is organized in sented. Conditions under which either procedure, or both, may seven tasks: (1) program management, (2) constraint effects be unconservative are noted. Potential changes in the Class 2 analytical development and validation, (3) evaluation of cladding or 3 piping procedure to explicitly cover all loadings are dis-effects, (4) ductile to cleavage fracture mode conversion, (5) cussed. However, the report is intended to be informatory and, fracture analysis metheds development and applications, (6) while the contents of the report may guide future Code material property data and test methods, and (7) integration of changes, specific recommendations are not given herein.

results into a state-of-the-art methodology. The program tasks have been structured to place emphasis on the resolution frac-NUREG/CR 3469 V08: OCCUPATIONAL DOSE REDUCTION AT tuie issues with near-term licensing significance. Resources to NUCLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY execute the research tasks are drawn from ORNL with subcon-OF SELECTED READINGS IN RADIATION PROTECTION AND tract support from universities and other research laboratones.

ALARA. SULLIVAN S G.; KHAN.T.A.; XIE,J.W. Brookhaven Na-Close contact is maintained with the sister Heavy-Section Steel tional Laboratory. May 1995. 78pp. 9506140072. BNL-NUREG-Irradiation Program at ORNL with related research programs 51708.84297:202.

both in the United States and abroad. This report provides an The ALARA Center at Brookhaven National Laboratory pub-overview of principal developments in each of the seven pro-lishes a series of bibliographies of selected readings in radiation gram tasks from October 1993 - March 1994.

protection and ALARA in a continuing effort to collect and dis.

NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND seminate information on radiation dose reduction at nuclear PIPING WELDS. Seventh Program Report March 1993 - Decem-power plants. This is volume 8 of the series. The abstracts in ber 1994. WILKOWSK!,G.M.; GHADIALI,N.; RUDLAND D.; et al.

this bibliography were selected from proceedings of technical Battelle Memorial Institute, Columbus Laboratories. April 1995.

meetings and conference journals, research reports, and 74pp. 9505180445. BMI-2173. 83967:074.

searches of the Energy Science and Technology database of This is the seventh progress report of the U.S. Nuclear Regu-the U.S. Department of Energy. The subject material of these fatory Commission's research program entitled Short Cracks in abstracts relates to the many aspects of radiation protection Piping and Piping Welds." The program objective is to verify and dose reduction, and ranges from use of robotics, to oper-and improve fracture analyses for circumferentially cracked ational health physics, to water chemistry. Material on the large-diameter nuclear piping with crack sizes typically used in design, planning, and management of nuclear power stations is leak before-break (LBB) analyses and in-service flaw evalua-included, as well as information on decommissioning and safe tions. All work in the eight technical tasks have been complet-storage efforts. Volume 8 contains 232 abstracts, an author ed. Ten topical reports are scheduled to be published. Progress index, and a subject index. The author index is specific for this only during the reporting penod, March 1993 - December 1994, volume. The subject index is cumulative and lists all abstract not covered in tho topical reports is presented in this report. De-numbers from volumes 1 to 8. The numbers in boldface indicate tails about the following efforts are covered in this report: Im-the abstracts in this volume; the numbers not in boldface repre-provements to the two computer programs NRCMPE and sent abstracts in previous volumes.

NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; Pipe mate-NUREG/CR-4219 V10 N2: HEAVY-SECTION STEEL TECHNOL-rial property database PIFRAC; Circumferentially cracked pipe OGY PROGRAM. Semiannual Progress Report For April-Sep-database CIRCUMCK.WK1; An assessment of the proposed tember 1993. PENNELL,W.E. Oak Ridge National Laboratory.

ASME Section lit design stress rule changes on pipe flaw toler-May 1995.159pp. 9506140082. ORNL/TM-9593. 84299:001, ance; and A pipe fracture experiment on a section of pipe re-The Heavy-Section Steel Technology (HSST) Program is con-moved from service degraded by microbiologically induced cor-ducted for the Nuclear Regulatory Commission by Oak Ridge rosion (MIC) which contained a girth weld crack. Progress in the National Laboratory (ORNL). The program focuses on the de-other tasks is not repeated here as it has been covered in great velopment and validation of technology for the assessment of detail in the topical reports.

fracture prevention margins in commercial nuclear reactor pres-sure vessels. The HSST Program is organized in 12 tasks: (1)

NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACK.

program management (2) fracture methodology and analysis, ING IN LIGHT WATER REACTORS.

Semiannual (3) material characterizations and properties, (4) special techni-Report, October 1993 - March 1994.

CHUNG,H.M.;

cal assistance, (5) fracture analysis computer programs, (6)

CHOPRA,0.K.; ERCK,R.A.; et al. Argonne National Laboratory.

cleavage-crack initiation, (7) cladding evaluations, (8) pressur-March 1995. 65pp. 9503280398. ANL-95/2. 83292:228.

ized-thermal-shock technology, (9) analysis methods validation, This report summarizes work performed by Argonne National (10) fracture evaluation tests, (11) warm prestressing, and (12)

Laboratory (ANL) on fatigue and environmentally assisted crack-biaxial loading effects on fracture toughness. The program tasks ing (EAC) in light water reactors during the six months from Oc-have been structured to emphasize the resolution fracture tober 1993 to March 1994. Topics that have been investigated issues with near-term licensing significance. Resources to exe-include fatigue of low-alloy steel used in piping, steam genera.

cute the research tasks are drawn from ORNL with subcontract tors, and reactor pressure vessels; EAC of wrought and cast support from universities and other research laboratories. Close austenitic stainless steels (SSs); and radiation-induced segrega-contact is maintained with the sister Heavy-Section Steel Irra-tion and irradiation-assisted stress corrosion cracking (IASCC) diation Program at ORNL and with related research programs of Type 304 SS after accumulation of high fluence. Fatigue both in the United States and abroad. This report provides an tests have been conducted on A302-Gr B low-alloy steel to overview of principal developments in each of the 12 program verify whether the current predictions of modest decreases of tasks from April - September 1993, fatigue life in simulated pressurized water reactor water are l

18 Main Citations and Abstracts vahd for high-sulfur heats that show environmentally enhanced NUREG/CR-4674 V22: PRECURSORS TC POTENTIAL SEVERE fatigue crack growth rates. Additional crack growth data were CORE DAMAGE ACCIDENTS:1994 A

STATUS obtained on fracture-mechanics specimens of austenitic SSs to REPORT. Appendix 1.

BELLES,R.J.;

CLETCHER,J.W.;

investigate threshold stress intensity factors for EAC in high-COPINGER.D.A.; et al. Oak Ridge National Laboratcry. Decem-punty oxygenated water at 289 degrees C. Microchemical ber 1995. 531pp. 9601290235. ORNL/NOAC-232. 86876:001.

changes in high-and commercial-punty Type 304 SS specimens See NUREG/CR-4674,V21 abstract.

from control-blade absorber tubes and a control-blade sheath NUREG/CR-4918 V08: CONTROL OF WATER INFILTRATION from operating boiling water reactors were studied by Auger INTO NEAR SURFACE LLW DISPOSAL UNITS. Progress electron spectroscopy and scanning electron microscopy to de-Report Of Field Experiments At A Humid Region termine whether trace impunty elements may contnbute to Site,Beltsville, Maryland. SCHULZ,R.K. California, Univ. of, Los IASCC of solution-annealed matenals.

Angeles, CA. RIDKY,R.W. Maryland, Univ. of, College Park, MD.

a anageM Rad M WM NUREG/CR-4667 V19: ENVIRONMENTALLY ASSISTED CRACK-Apnl 1995. 33pp. 9505030447. 83725:281.

ING IN LIGHT WATER REACTORS. Semiannual Report,Apnl September 1994 CHOPRA.O K.; CHUNG.H.M.; GAVENDA.D.J.';

The project objective is to assess means for controlling waste infiltration through waste disposal unit covers in humid regions.

et al Argonne National Laboratory. September 1995. 67pp-Expenmental work is being performed in large scale fysimeters 9510030169 ANL-95/25,85688:172.

(70 x45'x10') at Beltsville, MD and results of the assessment This report summanzes work performed by Argonne National are applicable to disposal of LLW, uranium mill tailings, hazard-Laboratory (ANL) on fatigue and environmentally assisted crack-ous waste, and sanitary landfills. Three concepts are under in-ing (EAC) in light water reactors from Apnl to September 1994.

vestigation: (1) resistive layer barrier, (2) conductive layer bar.

Topics that have been investigated include (a) fatigue of carbon rier, and (3) bioengineenng water management. The resistive and low-alloy steel used in piping and reactor pressure vessels, layer barrier consists of compacted earth (clay). The conductive (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and layer barner is a special case of the capillary bamer and it re-(c) arradiation assisted stress corrosion cracking (IASCC) of quires a flow layer (e g fine sandy loam) over a capillary break.

Type 304 SS. Fatigue tests have been conducted on A106-Gr B As long as unsaturated c.onditions are maintained water is con-and A533-Gr B steels in oxygenated water to determine wheth-ducted by the flow layer to below the waste. This barner is most er a slow strain rate applied during different portions of a ten.

efficient at low flow ratos and is thus best placed below a resis-sile-loading cycle are equally effective in decreasing fatigue hfe.

tive layer barrier. Such a combination of the resistive layer over Crack growth data wore obtained on fracture-mechanics speci.

the conductive layer barrier promises to be highly effective pro-mens of SSs and Alloy 600 to investigate EAC in simulated boil.

vided there is no appreciable subsidence. Bioengineenng water ing water reactor (BWR) and pressunzed water reactor environ-management is a surface cover that is designed to accommo-ments at 289 degrees C. The data were compared with predic-date subsidence. It consists of impermeable panels which en-tions from crack growth correlations developed at ANL for SSs hance run-off and limit infiltration. Vegetation is planted in narr w penings between panels to transpire water from below in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high-and commercial-punty the panels. This system has successfully dewatered two lyss-

" " "9 "t this procedure could be used Type 304 SS specimens from control-blade absorber tubes and for remedial action (. drying out,,) existing water logged disposal a control-blade sheath from operating BWRs were studied by i

sites at low cost.

Auger electron spectroscopy and scanning electron microscopy to determine whether trace impunty elements may contnbute to NUREG/CR-5229 V07: FIELD LYSIMETER INVESTIGATIONS:

lASCC of these matenals.

LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FISCAL YEAR 1994. Annual Report. MCCONNELL,J.W.;

NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE ROGERS.R.D.; JASTROW,J.D.; et al. Idaho National Engineer-CORE DAMAGE ACCIDENTS:1994 A STATUS REPORT. Main ing Laboratory. May 1995. 74pp. 9506140086. INEL-94/0278.

Report And Appendices A-H. BELLES,R.J.; CLETCHER,J.W.;

84298.232.

COPINGER.D.A.; et al. Oak Ridge National Laboratory. Decem-The Field Lysimeter investigations: Low-Level Waste Data ber 1995. 270pp. 9601290218. ORNL/NOAC-232. 86881:001.

Base Development Program, funded by the U.S. Nuclear Regu-Nine operational events that affected ten commercial hght.

latory Commission, is (a) studying the degradation effects in water reactors (LWRs) dunng 1994 that are considered to be EPICOR il organic ion exchange resins caused by radiation, (b) precursors to potential severe core damage are descnbed. All examining the adequacy of test procedures recornmended in of these events had conditional probabilities of subsequent core the Branch Technical Position on Waste Form to meet the re-damage greater than or equal to 1.0 x 10( 6). These events quirements of 10 CFR 61 using sohdified EPICOR il resins, (c) were identified by computer screening the 1994 hcensee event obtaining performance information on sohdified EPICOR-il ion-reports from commercial LWRs to identify those that could be exchange resins in a disposal environment. and (d) determining potential precursors. Candidate precursors were then selected the condition of EPICOR-Il kners. Compressive test results of and evaluated in a process similar to that used in previous as.

11 year old cement and vinyl ester styrene sohdified waste sessments. Selected events underwent engineenng evaluation forms are presented, which show effects of aging and self irra-that identified, analyzed, and documented the precursors. Other diation. Results of the ninth year of data acquisition from the events designated by the Nuclear Regulatory Commission field testing are presented and discussed. During the continuing (NRC) also underwent a similar evaluation. Finally. documented field testing, both Portland type 1-11 cement and Dow vinyl ester-precursors were submitted for review by bcensees and NRC styrene waste forms are being tested in lysimeter arrays located staff to ensure that the plant design and its response to the pre-at kgonne Nabonal LawaWad in Enm and at Oak W cursor were correctly characterized. This study is a continuation National Laboratory. The study is designed to provide continu-of earher work, which evaluated 1969-1981 and 1984-1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors.

NUREG/CR 5462: AGING STUDY OF BOILING WATER REAC-and the estimation of conditional probabihties of subsequent TOR HIGH PRESSURE INJECTION SYSTEMS. CONLEY,D.A.;

severe core damage for events. This document is bound in two EDSON.J.L.; FINEMAN.C.F. Idaho National Engineenng Labora-volumes: Volume 21 contains the main report and Appendicos tory. March 1995. 114pp. 9504120089 INEL-94/0090.

A-H, Volume 22 contains Appendix 1.

83474.051.

Main Citations and Abstracts 19 The purpose of high pressure injection systems is to maintain particular the fracture toughness properties, of typical pressure an adequate coolant level in reactor pressure vessels, so that vessel steels as they relate to light-water reactor pressure-the fuol cladding temperature does not exceed 1,200 degrees C vessel integnty. Effects of specimen size, material chemistry, (2,200 degrees F), and to permit plant shutdown during a variety product form and microstructure, irradiation fluence, flux, tem-of design basis loss-of-coolant accidents. This report presents perature and spectrum, and post. irradiation annealing are being the results of a study on aging performed for high pressure in-examined on a wide range of fracture properties. The HSSI Pro-jection systems of boiling water reactor plants in the United gram is arranged into 10 tasks: (1) program management, (2)

States. The purpose of the study was to identify and evaluate K(Ic) curve shift in high-copper welds, (3) K(la) curve shift in the effects of aging and the effectiveness of testing and mainte-high-copper welds, (4) irradiation effects on cladding, (5) K(Ic) nance in detecting and mitigating aging degradation. Guidelines and K(la) curve shifts in low upper-shelf welds, (6) irradiation ef-from the United States Nuclear Regulatory Commission's Nucie-fects in a commercial low upper shelf weld, (7) microstructural ar Plant Aging Research Program were used in performing the analysis of irradiation effects, (8) in-service aged material eval-aging study. Review and analysis of the failures reported in da-uations, (9) correlation monitor materials, and (10) special tech-tabases such as Nuclear Power Experience, Licensee Event nical assistance. This report provides an overview of the activi-Reports, and the Nuclear Plant Reliability Data System, along ties within each of these tasks from October 1991 to September with plant-specific maintenance records databases, are included 1992.

in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilis-NUREG/CR-5591 V04 N2: HEAVY-SECTION STEEL IRRADIA-tic risk assessments were reviewed to identify nsk-significant TlON PROGRAM. Semiannual Progress Report For Apnt-Sep-components in high pre %ure mjection systems. Testing, mainte-tomber 1993. CORWIN,W.R. Oak Ridge National Laboratory.

nance, specific safety issues, and codes and standards are also March 1995. 48pp. 9504180477. ORNL/TM-11568. 83569:316.

discussed.

The goal of the Heavy Section Steel Irradiation Program is to NUREG/CR-5535 V01: RELAP5/ MOD 3 CODE MANUAL. Code provide a thorough, quantitative assessment of effects of neu-Structure, System Models, And Solution Methods.

• Idaho Na.

tron irradiation on matenal behavior, and in particular the frac-tional Engineering Laboratory. August 1995.

412pp.

ture toughness properties, of typical pressure vessel steels as 9509150284. INEL 95/0174. 85495:001, they relate to light-water reactor pressure-vessel integnty. Ef-The RELAPS/ MOD 3 computer code has been developed for fects of specimen size, material chemistry, product form and mi-best-estimate simulation of light water reactor coolant systems crostructure, irradiation fluence, flux, temperature and spectrum, dunng postulated accidents. The code models the coupled be, and post-irradiation annealing are being examined on a wide havior of the reactor coolant system and the core for loss-of.

range of fracture properties. The HSSI Program is arranged into coolant accidents and operational transients such as anticipated 14 tasks: (1) program management, (2) fracture toughness transient without scram, loss of offsste power, loss of feedwater, (K(ic)) curve shift in high-copper welds, (3) crack-arrest tough-and loss of flow. A generic modeling approach is used that per.

ness (K(la)) curve shift in high copper welds, (4) irradiation ef-mits simulating a vanety of thermal-hydraulic systems. Control fects on cladding, (5) K(Ic) and K(la) curve shifts in low upper-system and secondary system components are included to shelf welds, (6) annealing effects in low upper-shelf welds, (7) permit modeling of plant controls, turbines, condensers, and irradiation effects in a commercial low upper-shelf weld (8) mi-secondary feedwater systems. RELAPS/ MOD 3 code documen-crostructural analysis of irradiation effects, (9) in-service aged tation is divided into seven volumes: Volume I presents model.

matenal evaluations, (10) correlation monitor materials, (11) ing theory and associated numencal solution schemes; Volume special technical assistance, (12) JPDR steel examination, (13) li details instructions for code application and input data prepa.

technical assistance for JCCCNRS Working Groups 3 and 12, ration; Volume lli presents the results of developmental assess-and (14) additional requirements for materials. This report pro-ment cases that demonstrate and venfy the models used in the vides an overview of the activities within each of these tasks code; Volume IV discusses in detail RELAPS models and corre-from April to September 1993.

lations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code:

NUREG/CR-5591 V05 N1: HEAVY SECTION STEEL IRRADIA.

Volume VI discusses the numencal scheme used in RELAP5; TION PROGRAM.Somiannual Progress Report For September and Volume Vil presents a collection of independent assess-1993 Through March 1994. CORWIN W.R. Oak Ridge National ment calculations.

Laboratory. April 1995. 71pp. 9505190052. ORNL/TM-11568.

NUREG/CR-5535 V02: RELAPS/ MOD 3 CODE MANUAL. User's Guide And input Requirements.

• Idaho National Eng:neenng The goal of the Heavy-Section Steel irradiation Program is to Laboratory. August 1995. 323pp. 9509150290, INEL 95/0174.

pr vide a thorough, quantitative assessment of effects of neu-85496:049 tron irradiation on material behavior, and in particular the frac-See NUREG/CR-5535,V01 abstract ture toughness properties, of typical pressure vessel steels as they relate to light water reactor pressure vessel integrity. Ef-NUREG/CR-5535 V04: RELAP5/ MOD 3 CODE MANUAL.Models fects of specimen size, material chemistry, product form and mi-And Correlations.

• Idaho National Engineenng Laboratory.

crostructure, irradiation fluence, flux, temperature and spectrum, August 1995. 430pp. 9509150303. INEL-95/0174. 85498:001, and post irradiation annealing are being examined on a wide See NUREG/CR 5535,V01 abstract.

range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness NUREG/CR-5535 V05 R1:

RELAPS/ MOD 3 CODE MANUAL. User's Guideline. FLETCHER,C.D.; GCHULTZ,R.R.

(K(Ic)) curve shift in high-copper welds, (3) crack arrest tough-Idaho National Engineering Laboratory. August 1995. 293pp.

ness (K(la)) curve shift in high copper welds, (4) irradiation ef-9509150307, INEL-95/0174. 85497:008.

fects on cladding, (5) K(Ic) and K(la) curve shifts in low upper-See NUREG/CR-5535,V01 abstract.

shelf welds, (6) annealing effects in low upper-shelf welds, (7) irradiation effects in a commercial low upper shelf weld, (8) mi-NUREG/CR-5591 V03: HEAVY-SECTION STEEL IRRADIATION crostructural analysis of irradiation effects, (9) in-service aged PROGRAM. Progress Report For October 1991 - September material evaluations (10) correlation monitor materials, (11) 1992. CORWIN,W.R. Oak Ridge National Laboratory. February special technscal assistance, (12) JPDR steel examination, (13) 1995.60pp.9503140442. ORNL/TM-11568. 83071:007.

technical assistance for JCCCNRS Working Groups 3 and 12, The pnmary goal of the Heavy-Section Steel irradiation Pro-and (14) additional requirements for materials. This report pro-gram is to provide a thorough, quantitatrve assessment of the vides an overview of the activities within each of these tasks effects of neutron irradiation on the matenal behavior, and in from September 1993 through March 1994.

20 Main Citations and Abstracts NUREG/CR-5591 V05 N2: HEAVY-SECTION STEEL IRRADIA-NUREG/CR-5758 V05: FITNESS FOR DUTY IN THE NUCLEAR TION PROGRAM. Progress Report For April 1994 Through Sep-POWER INDUSTRY. Annual Summary Of Program Performance tomber 1994. CORWIN W.R. Oak Ridge National Laboratory.

Repons CY 1994. WESTRA,C.; DURBIN,N.; FIELD,l.; et al. Bat-July 1995. 76pp. 9508160263. ORNL/TM-11568. 85042:092.

telle Seattle Research Center. August 1995.83pp.9508230332.

The goal of the Heavy-Section Steel trradiation Program is to PNL 10638. 85118.134.

provide a thorough, quantitative assessment of effects of neu-This report summarizes data from semiannual reports on fit-tron irradiation on material behavior, and in particular the frac-ness-for-duty program performance submitted to the NRC by ture toughness properties, of typical pressure vessel steels as utilities for two reporting periods: Jan.1 through June 30,1994, they relate to light water reactor pressure-vessel integnty. Ef-and Ju!y 1 through Dec. 31,1994. In 1994, licensees reported fects of specimen size, matenal chemistry, product form and mi-that they had conducted 163,247 tests for the presence of ille-crostructure, irradiation fluence, flux, temperature and spectrum, gal drugs and alcohol. Of these tests, 1,372 (.84 %) were con-and post irradiation annealing are being examined on a wide firmed pos:tive. Positive test results varied by category of test range of fracture properties. The HSSI Program is arranged into Lnd category of worker. The majority of positive test results 14 tasks: (1) program management, (2) fracture toughness (977) were obtained through pre-access testing. Of tests con-(K(Ic)) curve shift in high-copper welds, (3) crack arrest tough-ducted on workers having access to the protected area, 233 ness (K(la)) curve shift in high-copper welds, (4) irradiation ef-were positive from random testing and 122 were positive from lects on cladding, (5) K(ic) and K(la) curve shifts in low upper-for cause testing. Followup testing of workers who had previ-shelf welds, (6) annealing effects in low upper-shelf welds, (7) ously tested positive resulted in 50 positive tests. For cause irradiation effects in a commercial fow upper-shelf weld, (8) mi-testing resulted in the highest percentage of positive tests; j

crostructural analysis of irradiation effects, (9) in-service aged about 16% of for cause tests were positive. In comparison, i

material evaluations, (10) correlation monitor matenals, (11) 1.22% of pre-access tests and.28% of random tests were posi-i special technical assistance, (12) JPDR steel examination, (13) tive. Positive test rates also varied by category of worker. When technical assistance for JCCCNRS Working Groups 3 and 12, all types of tests are combined (pre-access, rendom, for-cause, and (14) additional requirements for matenals. This report pro-and followup), short-term contractor personnet had the highest vides an overview of the activities within each of these tasks positive test rate at 1.22% Licensee employees and long-term from April 1994 Through September 1994.

contractors had lower combined positive test rates (.33% and

.49% respectively). Of the substances tested, marijuana was re-NUREG/CR-5591 V06 N1: HEAVY SECTION STEEL IRRADIA' sponsible for the highest percentage of positive test results TION PROGRAM Semiannual Progress Report For October (52.79%), followed by cocaine (24.57%) and alcohol (17.93%).

1994 Through March 1995. CORWIN W.R. Oak Ridge National Laboratory. October 1995. 82pp. 9511270455. ORNL/TM.

NUREG/CR-5857: AGING OF TURBINE DRIVES FOR SAFETW 11568. 86317:001.

RELATED PUMPS IN NUCLEAR POWER PLANTS. COX,D.F.

The goal of the Heavy-Section Steel Irradiation Program is to Oak Ridge National Laboratory. June 1995. 64pp. 9507120251.

provide a thorough, quantitative assessment of effects of neu-ORNL 6713. 84634.295.

tron irradiation on matenal behavior, and in particular the frac-This study examines the relationship between aging, and cur-ture toughness properties, of typical pressure vessel steels as rent industry practices in areas of maintenance, surveillance, they relate to light-water reactor pressure-vessel integrity. Ef-and operation of steam turbine dnves for safety related pumps.

fects of specimen size, material chemistry, product form and mi.

These pumps are located in the Auxiliary Feedwater (AFW) crostructure, irradiation fluence, flux, temperature and spectrum, system for pressurized water reactor plants, and the Reactor i

and post-irradiation annealing are being examined on a wide Core Isolation Cooling and High Pressure Coolant Injection sys-

)

range of fracture properties The HSSI Program is arranged into tems for Boiling Water Reactor facilities. Findings in a recent 14 tasks: (1) program management, (2) fracture toughness study on the AFW (NUREG/CR 5404) indicate that the turbine (K(Ic)) curve shift in high-copper welds, (3) crack arrest tough-dnve is the single largest contributor to AFW system degrada-ness (K(la)) curve shift in high-copper welds, (4) irradiation ef-tion. However, examination of the data show that tne turbine fects on cladding (5) K(Ic) and K(la) curve shifts in low upper-itself is a reliable piece of equipment with a good service she!f welds, (6) annealing effects in low upper-shelf welds, (7) record. Most of the problems documented are the result of irradiation effects in a commercial low upper shelf weld, (8) mi-problems with the turbine controls and the mechanical over-crostructural analysis of irradiation effects, (9) in-service aged speed inp mechanism, which apparently stem from three major material evaluations, (10) correlation monitor materials, (11) causes.

special technical assistance, (12) JPDR steel examination, (13) technical assistance for JCCCNRS Working Groups 3 and 12, NUREG/CR 5884 V01: REVISED ANALYSIS OF DECOMMIS-and (14) additional requirements for matenals. This report pro.

SIONING FOR THE REFERENCE PRESSURIZED WATER RE-vides an overview of the activities within each of these tasks ACTOR POWER STATION Effects Of Current Regulatory And from October 1994 Through March 1995.

Other Considerations On The Financial Insurance Requirements Of The Decommissioning Rule And..

KONZEK,G.J.;

NUREG/CR-5657: AUTOCASK (AUTOMATIC GENERATION OF SMITH,R.I.; BIERSCHBACH M.C; et al. Battelle Memonal Insto 3-D CASK MODELS).A Microcomputer Based System For Ship-tute, Pacific Northwest Laboratory. November 1995. 122pp.

ping Cask Design Review Analysis.

GERHARD,M.A.;

9512260293. PNL-8742. 86610.226.

SOMMER,S.C. Lawrence Livermore National Laboratory. April With the issuance of the final Decommissioning Rule (July 27, 1995.130pp.9505180397. UCRL ID-118766. 83967:231, 1988), owners and operators of licensed nuclear power plants AUTOCASK (Automatic Generation of 3 D CASK Models) is are required to prepare, and submit to the U.S. Nuclear Regula-a microcomputer-based system of computer programs and data-tory Commission (NRC) for review, decommissioning plans and bases developed at the Lawrence Livermore National Laborato-cost estimates. The NRC staff is in need of bases documenta-ry for the structural analysis of shipping casks for radioactive tion that will assist them in assessing the adequacy of the li-material. Model specification is performed on the microcomput-censee submittals, from the viewpoint of both the planned ac-er, and the analyses are performed on the engineering worksta-tions, including occupational radiation exposure, and the proba-tion or mainframe computer, AUTOCASK is based on 80386-ble costs. The purpose of this reevaluation study is to provide 60486 compatib!e microcomputers. The system is composed of some of the needed bases documentation. This report contains a series of menus, input programs, display programs, a mesh the results of a review and reevaluation of the 1978 PNL de-generation program, and archive programs. All data is entered commissioning study of the Trojan nuclear power plant through fill-in-the-blank input screens that contain desenptive (NUREG/CR-0130), including all identifiable factors and cost data requests.

assumptions which contnbute significantly to the total cost of

Main Citations and Abstracts 21 decommissioning the nuclear power plant for the DECON, SAF-The effect of aging on the PWR Chemical and Volume Con-STOR, and ENTOMB decommissioning alternatives. These al-trol System (CVCS) has been evaluated. A detailed review of ternatives now include an initial 5-7 year period during which the NPRDS and LER databases for the 1988-1991 time period,

. time the spent fuel is stored in the spent fuel pool, prior to be-together with a review of industry and NRC experience and re-ginning major disassembly or extended safe storage of the search, indicate that age-related degradations and failures have plant. Included for information (but not presently part of the li-occurred. These failures had significant effects on plant oper-cense termination cost) is an estimate of the cost to demoksh ation, including reactivity excursions, and pressurizer level tran-the decontaminated and clean structures on the site and to re-sients. The majonty of these component failures resulted in store the site to a " green field" condition. This report also in-cludes consideration of the NRC requirement that decontamina-leakage of reactor coolant outside the containment. A repre-tion and decommissioning activities leading to termination of the sentative plant of each PWR NSSS design (W. CE, and B&S) nuclear heense be completed within 60 years of final reactor was v sited to obtain specific information on system inspection, shutdown, consideration of packaging and disposal require-smiHance, monitoring, and inspection practices. The results of these visits dicate that adequate system maintenance and in-in ments for materials whose radionuclide concentrations exceed the limsts for Class C low-level waste (i.e., Greater Than-Class spection is being performed. In some instances, the frequencies C), and reflects 1993 costs for labor, materials, transport, and of inspection were increased in response to repeated failure disposal activities. Sensitivity of the total license termination events. A parametric study was performed to assess the effect cost to the disposal costs at different tow-level radioactive of system aging on Core Damage Frequency (CDF). This study waste disposal sites, and to different depths of contaminated showed that as MOV operating failures increased the contribu-concrete surface removal within the facilities is also examined.

tion of the High Pressure injection to CDF also increased.

NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMM'S-NUREG/CR-5973 R02: CODES AND STANDARDS AND OTHER SiONING FOR THE REFERENCE PRESSURIZED WATER RE-GUIDANCE CITED IN REGULATORY DOCUMENTS.

ACTOR POWER STATION. Effects Of Current Regulatory And Other Considerations On The Financial insurance Requirements NICKOLAUS,J.R.; BOHLANDER K.L. Batteile Memorial Institute, Pacific Northwest Laboratory. August 1995.

525pp, Of The Decommissioning Rule And....

KONZEK,G.J.;

SMITH,R.I.: BIERSCHBACH,M.C; et al. Battelle Memorial Insti.

9509130146. PNL-8462. 85429:001.

tute, Pacific Northwest Laboratory. November 1995. 424pp.

As part of the U.S. Nuclear Regulatory Commission (NRC) 9512260336. PNL-8742. 86609:001.

Standard Review Plan Update and Development Program, Pa-See NUREG/CR-5884,V01 abstract.

cific Northwest Laboratory developed a listing of industry con-sensus codes and standards and other govemment and indus.

NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE AS-SESSMENT METHODOL.OGY FOR LOW-LEVEL RADIOAC-try guidance referred to in regulatory documents. In addition to TiVE WASTE DISPOSAL FACILITIES. Validation Needs.

updating previous information, Revision 2 adda more than 500 KOZAK,M.W.; OLAGUE,N E. Sandia National Laboratories. Feb.

citations. This hstang identifies the version of the code or stand-ruary 1995. 43pp. 9503170314. SAND 912801, 83151:304.

ard cited in the regulatory document, the regulatory document, in this report, concepts on how validation fits into the scheme and the latest version of the code or standard. It also provides of developing confidence in performance assessments are intro-a summary characterization of the nature of the citation. This duced. A general framework for validation and confidence build-listing was developed from electronic searches of the Code of ing in regulatory decision making is provided. It is found that tra.

Federal Regulations and the NRC's Bulletins, Information No-1 ditional vahdation studies have a very hmited role in developing taces, Circulars, Enforcement Manual, Generic Letters, Inspec-site-specific confidence in performance assessments. Indeed, tion Manual, Policy Statements, Regulatory Guides, Standard vahdation studies are shown to have a role only in the context Technical Specifications, and the Standard Review Plan that their results can narrow the scope of initial investigations (NUREG-0800).

that should be considered in a performance assessment. In ad-dition, validation needs for performance assessment of low-level NUREG/CR-5975 R01: INCENTIVE REGULATION OF INVES-waste disposal facihties are discussed, and potential approach-TOR-OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILI-es to address those needs are suggested. These areas of topi-TY REGULATORS.

MCKINNEY,M.D.;

SEELY,H.E.;

cal research are ranked in order of importance based on rel-MERRITT,C.R.; et al. Battelle Memorial Institute, Pacific North-i evance to a performance assessment and likelihood of success.

west Laboratory. April 1995. 70pp. 9505180354. PNL-8466.

NUREG/CR-5944 V02: A CHARACTERIZATION OF CHECK 83967:001.

VALVE DEGRADATION AND FAILURE EXPERIENCE IN THE The U.S. Nuclear Regulatory Commission (NRC) periodically NUCLEAR POWER INDUSTRY.1991 Failures.

surveys utilities that operate nuclear plants and state regulatory MCELHANEY.K.L Oak Ridge National Laboratory. July 1995.

commissions that regulate utihty owners of nuclear power 87pp. 9508090036. ORNL-6734. 84963:003.

plants. The NRC is interested in identifying states that have es-j Review and characterization of check valve failures taken tablished economic or performance incentive programs applica-from the Institute of Nuclear Power Operation's Nuclear Plant ble to nuclear power plants, including states with new programs, Reliabihty Data System was performed in accordance with part how the programs are being implemented, and in determining two of a three-phase project sponsored by the U.S. Nuclear the financial impact of the programs on the utihties. The NRC Regulatory Commission. The analysis was performed for check interest stems from the fact that such programs have the poten-valve failures occurnng in 1991 and is intended to update the tial to adversely affect the safety of nuclear power plants. The previous analysis performed for the time penod 1984-1990. To information in this report was obtained from interviews conduct-maintain consistency and for ease of trending, the 1991 analy-ed with each state regulatory agency that administers an incen-j sis presents the same parameters and cross-correlations in es-tive program and each utility that owns at least 10% of an af-sentially the same format as the 1984-1990 study. Additional fected nuclear power plant. The agreements, orders, and settle-data was obtained for the 1991 analysis, including information ments, that form the basis for each incentive program were re.

related to specific check valve type. This information is present-viewed as required. The interviews and supporting documenta-ed in a separate section of the report, tron form the basis for the individual state reports desenbing the NUREG/CR-5954: EFFECT ON AGING ON PWR CHEMICAL structure and financialimpact of each incentive program.

AND VOLUME CONTROL SYSTEM. GROVE E.J.; TRAVIS,R.J.

Brookhaven National Laboratory. June 1995.

196pp.

NUREG/CR-6002:

RISK-BASED MAINTENANCE 9507060335. BNL NUREG-52410. 84532:181.

~.

-~

22 Main Citations and Abstracts MODELING.Prioritization Of Maintenance importances And covered. It also examines whether current testing frequencies Quantsfication Of Maintenance Effectiveness. VESELY,W.E.;

and methods are effective in predicting such failures.

REZOS,J.T. Science Applications International Corp. (formerly Science Apphcations, Inc.).

• Brookhaven National Laboratory.

NUREG/CR-6017: FIRE MODELING OF THE HEISS DAMPF September 1995. 90pp. 9510030191. BNL-NUREG-52332.

REAKTOR CONTAINMENT, NICOLETTE,V.F. Sandia National 85672:266.

Laboratories. YANG K.T. Notre Dame, Univ. of, Notre Dame, IN.

This report desenbes approaches for priontizing the risk im.

September 1995.

115pp. 9510310360. SAND 93-0528.

portances of maintenances using a Probabilistic Risk Assess.

86018:001, ment (PRA). Approaches are then desenbed for quantifying the This report summarizes Sandia National Laboratories

• partici-reliability and nsk effects of maintenance actions Two different pation in the fire modehng activities for the German Heiss PRA importance measures, minimal cutset importances and risk Dampf Reaktor (HDR) containment building, under the sponsor.

reduction importances, are used to priontize maintenances and ship of the United States Nuclear Regulatory Commission. The the report shows that similar results are obtained if appropriate Purpose of this report is twofold: 1) to summarize Sandia's par.

criteria are used. The justifications for the particular importance ticipation in the HDR fire modeling efforts and 2) to summarize measures are also developed. The approaches which are devel, the results of the international fire modeling community involved oped for quantifying the reliability and nsk effects of mainte.

in modeling the HDR fire tests. Additional comments on the nance actions are extensions of the usual reliability models now state of fire modeling and trends in the international fire model-used in PRAs. These extended models consider degraded ing community are also included. It is noted that, although the states of the component and quantify the benefits of mainte.

trend internationally in fire modeling is toward the development nance in correcting degradations and preventing failures. The of the more complex fire field models, each type of fire model negative effects of maintenance, including maintenance down.

has something to contribute to the understanding of fires in nu.

times, are also included. These models are specific types of clear power plants.

Markov models. This report analyzes a range of postulated values of input data in order to explore the potential usefulness NUREG/CR-6046: ALERTNESS, PERFORMANCE, AND OFF-of these models. The effects of maintenance are quantified to DUTY SLEEP ON 8-HOUR AND 12-HOUR NIGHT SHIFTS IN A be large in specific cases.

SIMULATED CONTINUOUS OPERATIONS CONTROL ROOM SETTING. BAKER.T.L Institute for Circadium Physiology. April NUREG/CR 6004: PROBABILISTIC PIPE FRACTURE EVALUA-1995.157pp. 9505030466. 83745:228.

TIONS FOR LEAK-RATE-DETECTION APPLICATIONS.

This study of simulated 8-hour and 12-hour work shifts com.

RAHMAN,S.; GHADIAll,N.; PAUL,D.; et al. Battelle Memonal in-pares alertness, speed, and accuracy of subjects assigned to stitute, Columbus Laboratories. April 1995. 336pp. 9505090276.

the two shift conditions. Twenty male subjects, ages 20 to 45, BMI-2174. 83848.001, were randomly assigned to either an 8-hour or 12-hour shift pro-Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary tocol, working in teams of two in a process control simulator, in Leakage Detection Systems," was published by the U.S. Nucte-addition to the simulator monitoring task, subjects completed a ar Regulatory Commission (NRC) in May 1973, and provides 15-minute performance assossment battery and a vigor and guidance on leak detection methods and system requirements affect scale hourly. Throughout the simulated work shift, alert-for Light Water Reactors. Additionally, leak detection limits are ness was monitored by recording electroencephalogram (EEG),

specified in plant Technical Specifications and are different for electrooculogram (EOG) and electromyogram (EMG) data. At 2-Boiling Water Reactors (GWRS) and Pressunzed Water Reac.

hour intervals, subjects' sleepiness alertness level was as-tors (PWRs). These leak detection limits are also used in leak-sessed via standardized nap tests. Dunng off-shift hours, sub-before-break evaluations performed in accordance with Draft jects lived in attached, self contained apartments. Subjects Standard Review Plan, Section 3.6.3, " Leak Before Break Evat-were restricted to bed for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> sleep period in total dark-uation Procedures" where a margin of 10 on the leak detection ness one-hour after work shifts; EEG, EOG, AND EMG record-limit is used in determining the crack size considered in subse-ings were continued dunng sleep. During days off between quent fracture analyses. This study was requested by the NRC blocks of shifts, subjects left the facility, but kept detailed logs to: (1) evaluate the conditional failure probability for BWR and of sleep-wake activities. No significant differences were found PWR piping for pipes that were leaking at the allowable leak de-between subjects on 8-hour and 12-hour night shifts in perform-tection limit, and (2) evaluate the margin of 10 to determine if it ance on the simulator task, in length or quality of daytime sleep, was unnecessanly large. A probabilistic approach was undertak-or in physiological or subjective sleepiness-alertness on shift.

en to conduct fracture evaluations of circumferentially cracked Except for one subtest in the performance test battery, subjects pipes for leak rate-detection applicauons. Sixteen nuclear piping on 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts were slower, but more accurate, than those on systems in BWR and PWR plants were analyzed to evaluate 8-hour shifts. Several measures showed better alertness, mood, conditional failure probability and effects of crack-morphology and off-duty sleep on evening shifts than on night shifts.

variability on the current margins used in leak rate detection for leak before-break.

NUREG/CR-6054: ESTIMATING PRESSURIZED WATER REAC-TOR DECOMMISSIONING COSTS. A User's Manual For The NUREG/CR-6016: AGING AND SERVICE WEAR OF AIR-OPER-PWR Cost Estimating Computer Program (CECP) Software.

ATED VALVES USED IN SAFETY-RELATED SYSTEMS AT NU.

BIERSCHBACH,M.C Battelle Memorial institute, Pacific North-CLEAR POWER PLANTS. COX,D.F.; MCELHANEY,K.La west Laboratory. November 1995.174pp. 9512200287, PNL-STAUNTON R.H. Oak Ridge National Laboratory. May 1995.

8497, 86610:061, 62pp.9506140092. ORNL/TM 6789. 84299:152.

With the issuance of the Decommissioning Rule (June 27, Air-operated valves (AOVs) are used in a vanety of safety re-1988), nuclear power plant licensees are required to submit to lated applications at nuclear power plants. They are often used the U.S. Nuclear Regulato y Commission (NRC) for review, de-where rapid stroke times are required or precise control of the commissioning plans and cost estimates. This user's manual valve obturator is required. They can be designed to operate and the accompanying Cost Estimating Computer Program automatically upon loss of power, which is often desirable when (CECP) software provide a cost calculating methodology to the selecting components for response to design basis conditions.

NRC staff that will assist them in assessing the adequacy of the The purpose of this report is to examine the reported failures of licensee submittals. The CECP, designed to be used on a per.

AOVs and determine whether there are identifiable trends in the sonal computer, provides estintiates for the cost of decommis-failures related to predictable causes. This report examines the sioning PWR power stations to the point of license termination.

specific components that compnse a typical AOV, how those Such cost estimates include compuMnt, pip,ng, and equipment components fail, when they fail, and how such failures are dis-costs; packaging costs. decontamination costs; transportation

Main Citations and Abstracts 23 costs; burial costs; and manpower costs. In addition to costs, important developments: (a) two methods for estimating or the CECP also calculates burial volumes, person-hours, crew-bounding the design basis stem factor (in rising-stem valves),

hours, and exposure person-hours associated with decommis-using data from tests less severe than design basis tests; (b) a sioning.

new correlation for evaluating the opening responses of gate NUREG/CR-6074 V04: SEALED SOURCE AND DEVICE DESIGN valves and for predicting opening requirements; (c) an extrapo-SAFETY TESTING. Technical Report On The Findings Of Task lation method that uses the results of a best effort flow test to

4. Investigation Of Sealed Source for Paper Mill Digester, estimate the design basis closing requirements of a gate valve BENAC.D.J.; IDDINGS F.A. Southwest Research institute. Octo-that exhibits atypcal responses (peak force occurs before flow ber 1995. 42pp. 9512040331. 04 4448-013. 86386:086.

isolation); and (d) the extension of the original INEL closing cor-This report covers the Task 4 activities for the Sealed Source relation to include low-flow and low-pressure loads. The report and Device Safety testing program. SwRI was contracted to in-also includes a general approach, presented in step-by-step vestigate a suspected leaking radioactive source that was in-format, for determining operating margins for rising stem valves stalled in a gauge that was on a paper mill digester. The actual (gate valves and globe valves) as well as quarter-turn valves source that was leaking was not available, therefore, SwRI ex.

(ball valves and butterfly valves).

amined another sou.ce. SwRI concluded that the encapsulated NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAIL-source examined by SwRI was not leaking. However, the pres-URE BY DIRECT CONTAINMENT HEATING IN SURRY, ence of Cs-137 on the intenor and exterior of the outer encap-PILCH,M.M.; ALLEN,M.D.

Sandia National Laboratories.

sulation and handling tube suggests that contamination prob-ably occurred when the source was first manufactured, then in-SPENCER,B.W.; et al. Argonne National Laboratory. May 1995.

stalled in the handling tube.

275pp. 9506010485. SAND 93-2078. 84121:302.

This report uses the methodology and scenarios desenbed in NUREG/CR-6074 VD5: SEALED SOUNE AND DEVICE DESIGN NUREG/CR-6075 and its supplement to address the direct con-SAFETY TESTING. Technical Report On The Findings Of Task tainment heating (DCH) issue for the Surry nuclear power plant

4. Investigation Of Failed Radioactive Stainless Steel Troxler (NPP). Consistency of the initial distributions has been ensured Gauges. BENAC,D.J.; SCHICK,W.R. Southwest Research Insti-by using insights from system-level codes, specifically SCDAP/

tute. October 1995. 58pp. 9512040326. 04 4448 010.

RELAPS and CONTAIN. The most useful insights are that the 86386:129.

reactor coolant system (RCS) pressure is low at vessel breach, This report covers the Task 4 activities for the Sealed Source metallic biockages in the core region do not celt and relocate and Device Safety testing program. SwRI was contracted to in-into the lower plenum, and melting of upper plenum steelis cor-vestigate failed radioactive stainless steel troxler gauges.

related with hot leg failure. The SCDAP/RELAPE etput was SwRl's task was to determ:ne the cause of failure of the rods used as input to CONTAIN to assess the containmer.' condi-and the extent of the problem. SwRI concluded that the broken tions at vessel breach. The load evaluations for Suny sMwed rod failed in a bnttle manner due to a hard zone in the heat af-no intersections of the load distributions with the containmant fected zone.

strength distribution, and thus the DCH issue for Surry can bi NUREG/CR-6089: DETECTION OF PUMP DEGRADATION.

resolved based on containment loads alone. However, the likeli-GREENE,R.H.; CASADA.D.A. Oak Ridge National Laboratory.

hood of high RCS pressures at vessel breach was evaluated for August 1995.108pp. 9509130143 ORNL-6765. 85430:182.

Surry for a limited number of sequences. The probability of RCS This study examines the methods of detecting pump degrada, pressures greater than 1.38 MPa for all station blackout scenar-tion that are employed in domestic and overseas nuclear facili.

ios without power recovery or operator intervention was found ties. This report evaluates entena mandated by required pump to be low (0.077). This probability could have been factored into testing at U.S. nuclear plants and compares them to state-of-the containment failure probability for Surry if there had been the-art diagnostic programs and practices implemented by other intersections of the load and strength distributions.

major industnes. Since the working condition of the pump driver NUREG/CR-6112: IMPACT OF REDUCED DOSE LIMITS ON is entical to pump operability, a review of new applications of motor diagnostics is also provided that highlights developments tion Of ICRP/NCRP Dose Limit Recommendations. Final Report.

in this technology. Vibration spectral analysis is a powerful diag-MEINHOLD.C.B. Brookhaven National Laboratory. May 1995, nostic tool for the pump analyst. The routine collection and 79.9505290093. BNL/NUREG-52394. 84017:240.

analysis of spectral data is supenor tc, other technologies in its g

ability to accurately detect numerous types and causes of pump degradation. Existing ASME testing cntena do not require the a'avely, the potential impacts of reducing occupation-al dose limits below those grven in 10 CFR 20 (Revised). The evaluation of pump vibration spectra but instead overall vibra-tion amplitude. The mechanical information discernible from vi-following overall conclusions were reached: (1) Although 10 bration amplitude is limited, and several pump failures in the nu-g clear power industry were not detected by vibration amplitude structive to the continued operation of some hcensees, such as monitonng. Since spectral analysis provides pertinent informa-nuclear power, fuel fabrication, and medicine, (2) Twenty mSv tion concerning the mechanical condition of rotating machinery, yr(-1) as a limit is possible for some of these groups, but for its incorporation into ASME testing enteria may ment a relax-others it would prove difficult, (3) Firty mSv yr(-1) and age in 10s ation in the monthly-to quarterly testing schedules that seek t of mSv appear reasonable for all licensees, both in terms of the venfy pump operabihty.

lifetime nsk of cancer and severe genetic effects to the most NUREG/CR-6100: GATE VALVE AND MOTOR-OPERATOR RE-highly exposed workers, and the practicality of operation. In SEARCH FINDINGS.

STEELE,R.;

DEWALL,K.G.;

some segments of the industry, this acceptability is based on WATKINS,J.C.; et al. Idaho National Engineenng Laboratory.

the adoption of a " grandfather clause" for those people ex-September 1995.

130pp. 9510060250. INEL-94/0156.

ceeding or close to exceeding the cumulative limit at this time.

85745:001.

This report provides an update on the valve research being NUREG/CR-6114 V02: AUXILIARY ANALYSES IN SUPPORT OF sponsored by the U.S. Nuclear Regulatory Commission (NRC)

PERFORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-and conducted at the Idaho National Engineering Laboratory LEVEL WASTE FACILITY,Two-Phase Flow And Contaminant (INEL). The research addresses the need to provide assurance Transport in Unsaturated Soils With Application To Low-Level that motor operated valves can perform their intended safety Radioactive Waste Disposal BINNING,P.;

CELIA,M.A.;

function, usually to open or close against specified (design JOHNSON.J.C. Pnneeton Univ., Princeton, NJ. May 1995.

basis) flow and pressure loads. This report describes several 139pp. 9506140130. 84299:214.

i 24 Main Citations and Abstracts A numencal model of multiphase air-water flow and contami-operation and actual code operation were identified. Modifica-nant transport in the unsaturated zone is presented. The multi-tions that have been made to SAPHIRE are identified.

phase flow equatons are solved using the two-pressure, mixed NUREG/CR-6116 V10: SYSTEMS ANALYSIS PROGRAMS FOR form of the equations with a modified Picard hneanzation of the HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA-equatons and a finite element spatial approximation. A volatile PHIRE) VERSION 5.0. Data Loading Manual. VANHORN.R.L-contaminant is assumed to be transported in either phase, or in WOLFRAM LM.; FOWLER,R.D.; et al. Idaho National Engineer $

both phases simultaneously. The contaminant partitons be-tween phases with an equilibrium distribution given by Henry's ing Laboratory. April 1995.161pp. 9505180347. INEL-94/0039.

83966:120 Law or via kinetic mass transfer. The transport equations are solved using a Galccin finite element method with reduced in-The Systems Analysis Programs for the Hands on integrated tegration to lump the resultant matnces. The numerical model is Reliability Evaluations (SAPHIRE) suite of programs can be applied to published expenmental studies to examine the be-used to organize and, standardize in an electronic format infor-g havior of the air phase and associated contaminant movement examinations. The Models and Results Database (MAR-D) pro-under water infiltration. The model is also used to evaluate a hy-pothetical design for a low-level radioactive waste disposal facil-gram of the SAPHIRE suite serves as the repository for probabi-listic risk assessment and individual plant examination data and sty. The model has been developed in both one and two dimen-sions; documentaten and computer codes are available for the information. This report demonstrates by examples the common electronic and manual methods used to load these types of I

one-dimensional flow and transport model.

data. It is not a stand alone document but references docu.

NUREG/CR-6116 V06: SYSTEMS ANALYSIS PROGRAMS FOR rnants that contribute information relative to the data loading HANDS-ON INTEGRATED REL! ABILITY EVALUATIONS (SA.

process. This document provides a more detailed discussion PHIRE) VERSION 5 0. Graphical Evaluaton Module (GEM) Ref-and instructions for using SAPHIRE 5.0 only when enough infor-erence Manual RUSSELL,K.D.;

KVARFORDT,K.J.;

mation on a specific topic is not provided by another available s

HOFFMAN.C.L.; et al. Idaho Natonal Engineenng Laboratory, source.

October 1995.113pp. 9511270459. INEL-94/0039. 86344:149.

NUREG/CR-6119 V01:

MELCOR COMPUTER CODE The Systems Analysis Programs for Hands-on-integrated Reli-MANUALS. Primer And User's Guides. Version 1.8.3 September ability Evaluations (SAPHIRE) refers to a set of several micro-1994. SUMMERS,R.M.; COLE,R.K.; SMITH R.C.; et al. Sandia computer programs that were developed to create and analyze National Laboratories. March 1995. 708pp. 9505040396.

probabilistic nsk assessments (PRAs), pnmanly for nuclear SAND 93-2185. 83779:001.

power plants. The Graphical Evaluation Module (GEM) is a spe-MELCOR is a fully integrated, engineenng-level computer cial application tool designed for evaluation of operational oc-code that models the progression of severe accidents in light currences using the Accident Sequence Precursor (ASP) pro-water reactor nuclear power plants. MELCOR is being devel-gram methods. GEM provides the capability for an analyst to oped at Sandia National Laboratones for the U.S. Nuclear Reg-quickly and easily perform conditonal core damage probability ulatory Commission as a second-generation plant risk assess-(CCDP) calculations. The analyst can then use the CCDP calcu-ment tool and the successor to the Source Term Code Pack-lations to determine if the occurrence of an initiating event or a age. A broad spectrum of severe accident phenomena in both condition adversely impacts safety. It uses models and data de-boiling and pressunzed water reactors is treated in MELCOR in veloped in the SAPHIRE specially for the ASP program. GEM a unified framework. These include: thermal-hydraulic response requires more data than that normally provided in SAPHIRE end in the reactor coolant system, reactor cavity, containment, and will not perform properly with other models or data bases. This confinement buildings; core heatup, degradation, and relocation; is the first release of GEM and the developers of GEM welcome core-concrete attack; hydrogen production, transport, and com-user comrnonts and feedback that will generate ideas for im-bustion; fission product release and transport; and the impact of provements to future versions. GEM is designated as version engineered safety features on thermal-hydraulic and radionu-5.0 to track GEM codes along with the other SAPHIRE codes clide behavor. Current uses of MELCOR include estimation of as the GEM relies on the same, shared database structure.

severe accident source terms and their sensitivities and uncer.

NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR tainties in a vanety of applications.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SA.

NUREG/CR-6119 V02:

MELCOR COMPUTER CODE PHIRE) VERSION 5.0. Verification And Validation (V&V) Manual.

MANUALS Reference Manuals. Version 1.8.3 September 1994.

JONES J.L.; CALLEY,M.B.; CAPPS E.L.; et al. Idaho National SUMMERS,R.M.; COLE,R.K.; SMITH,R.C. Sandia National Lab-Engineering Laboratory. March 1995. 141pp. 9503270310.

oratones. March 1995. 885pp. 9505040402. SAND 93-2185.

INEL-94/0039. 83270:133.

83781:001.

A venfication and validation (V&V) process has been per.

See NUREG/CR-6119,V01 abstract.

formed for the System Analysis Programs for Hands-on Integrat-ed Reliability Evaluaton (SAPHIRE) Version 5.0. SAPHIRE is a NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF set of four computer programs that the Nuclear Regulatory REMOTE AFTERLOADING BRACHYTHERAPY. Human Error Commission has developed for the performance of probabilistic And Critical Tasks In Remote Afterloading Brachytherapy And nsk assessments. These programs allow an analyst to perform Approaches For improved System Performance. CALLAN,J.R.;

many of the functions necessary to create, quantify, and evalu-KELLY,R.T.; OUINN,M L.; et al. Pacific Science & Engineenng ate the risk associated with a facility or process being analyzed.

Group, Inc. May 1995. 205pp. 9506140119. 84300 001.

The programs included in this set are Integrated Reliability and Remote Afterloading Brachytherapy (RAB) is a medical proc-Risk Analysis System (IRRAS), System Analysis and Risk As-ess used in the treatment of cancer. RAB uses a computer-con-sessment (SARA), Models and Results Database (MAR-D), and trolled device to remotely insert and remove radioactive sources Fault Tree / Event Tree / Piping and instrumentation Diagram close to a target (or tumor) in the body Some RAB problems (FEP) graphical editor. The intent of this program is V&V of suc-affecting the radiaton dose to the patient have been reported cessive versions of SAPHIRE. The SAPHIRE 5.0 V&V plan is and attnbuted to human error. To determine the root cause of based on the SAPHIRE 4.0 V&V plan with revisions to incorpo-human error in the RAB system, a human factors team visited rate lessons leamed from the previous effort. The SAPHIRE 5.0 23 RAB treatment sites in the U.S. The team observed RAB vital and nonvital test procedures are based ort the test proce-treatment planning and delivery, interviewed RAB personnel, dures from SAPHIRE 4.0 with revisions to include the new SA-and performed walk throughs, dunng which staff demonstrated PHtRE 5.0. The majonty of the results from the testing was ac-the procedures and practices used in performing RAB tasks.

coptable, however, some discrepancies between expected code Factors leading to human error in the RAB system were identi-

l Main Citations and Abstracts 25 feed. The impact of those factors on the periormance of RAB shine dose, long-term inhalation dose, total food pathways was then evaluated and priontized in terms of safety signifi-dose, total ingestion pathways dose, total long-term pathways cance. Finally, the project identified and evaluated alternative dose, total latent cancer fatalities, area-dependent cost, crop approaches for resolving the safety significant problems related disposal cost, milk disposal cost, population-dependent cost, to human error, total economic cost, condemnation area, condemnation popula-NUREG/CR-6125 V02: HUMAN FACTORS EVALUATION OF tion, crop disposal area and milk disposal area.

REMOTE AFTERLOADING BRACHYTHERAPY. Function And NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS Task Analysis. CALLAN J.R.; GWYNNE,J.W.; KELLY.R.T.; et al.

OF EARLY EXPOSURE RESULTS WITH THE MACCS REAC-Pacific Science & Engmsering Group, Inc. May 1995,143pp TOR ACCIDENT CONSEQUENCE MODEL. HELTON,J.C. Anzo-9506140126. 84300:207, na State Univ., Tempe, AZ. JOHNSON J.D. GRAM, Inc.

A human factors project on the use of nuclear by-product ma-MCKAY,M.D.; et al. Los Alamos National Laboratory. January terial to treat cancer using remotely operated afterloaders was 1995.147pp. 9502080062. SAND 93-2371, 82660:063.

undertaken by the Nuclear Regulatory Commission. The pur-Uncertainty and Sensitivity analysis techniques based on pene of the project was to identify factors that contnbute to Latin hypercube sampling, partial correlation analysis and step-human error in the system for remote afterloading brachyther*

wise regression analysis are used in an investigation with the apy (RAB). This report documents the findings from the first MACCS model of the early health effects' associated with a phase of the project, which involved an extensive function and severe accident at a nuclear power station. The pnmary pur-task analysis of RAB. This analysis identified the functions and pose of this study is to provide guidance on the vanables to be tasks in RAS, made preliminary estimates of the likelihood of considered in future review work to reduce the uncertainty in human errcr in each task, and determined the skills needed to the important variables used in the calculation of reactor acci-perform each RAB task. The findings of the function and task dent consequences. The effects of 34 imprecisely known input analysis served as the foundation for the remainder of the variables on the following reactor accident consequences are project, which evaluated four major aspects of the RAB system studied. number of early fatahties, number of cases of prodro-linked to human error; human-system interfaces; procedures mal vomiting, population dose within 10 mi of the reactor, popu-and practices; training and qualifications of RAB staff; and orga-lation dose within 1000 mi of the reactor, individual early fatality nizational practices and policies. At its completion, the project probability within 1 mi of the reactor, and maximum early fatality identified and priontized areas for recommended NRC and i'b distance.

dustry attention based on all of the evaluations and analyses.

NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF NUHEG/CR-6136: UNCERTAINTY AND SENSITIVITY ANALYSIS REMOTE AFTERLOADING BRACHYTHERAPY. Supporting OF FOOD PATHWAY RESULTS WITH THE MACCS REACTOR Analyses Of Human-System Interfaces, Procedures And ACCIDENT CONSEQUENCE MODEL. HELTON,J.C. Arizona Practices, Training And Organizational Practdes And Proce.

State Univ., Tempe, AZ. JOHNSON,J.D.; ROLLSTIN.J.A.; et al.

dures. CALLAN.J.R : KELLY,R.T.; OUINN M.L.; et al. Pacific Sci.

GRAM, Inc. January 1995. 94pp. 9502080088. SAND 93-2372.

82660:210.

ence & Engineering Group, Inc. July 1995.226pp.0508090067.

84965:001.

Uncertainty and Sensitivity analysis techniques based on A human factors project on the use of nuclear by-product ma.

Latin hypercube sampling, partial correlaticn analysis and step-tenal to treat cancer using remotely operated afterloadert, was wise regression analysis are used in an investigation with the undertaken by the Nuclear Regulatory Commission. The pur-MACCS model of the food pathways associated with a sevore pose of the project was to identify factors that contribute to accident at a nuclear power station. The primary purpose of this human error in the system for remote afterloading brachyther-study is to provide guidance on the vanables to be considered apy (RAB). This report documents the findings from the second, IN Iuture review work to reduce the uncertainty in the important vanables used in the calculation of reactor accident conse-third, fourth, and fifth phases of the project, which involved de, talled analyses of four major aspects of the RAB system knked quences. The effects of 87 imprecisely known input variables on to human error: human-system interfaces; procedures and prac-the following reactor accident consequences are studied: crop tices; training practices and policies; and organizational prac-9'0**g season dose, crop long-term dose, milk growing tices and policies, respectively. Findings based on these analy, season dose, total food pathways dose, total ingestion path-ses provided factual and conceptual support for the final phase ways dose, total long-term pathways dose, area dependent of th'is project, which identified factors leading to human orror in cost, crop disposal cost, milk disposal cost, condemnation area, RAB. The impact of those factors on RAB performance was crop disposal area and milk disposal area.

then evaluated and prioritized in terms of safety significance, NUREG/CR-6141: HANDBOOK OF METHODS FOR RISK-l and alternative approaches for resolving safety significant prob-BASED ANALYSES GF TECHNICAL SPECIFICATIONS.

I lems were identified and evaluated.

SAMANTA P.K.; KIM,l.S. Brookhaven National Laboratory.

NUREG/CR-6134: UNCERTAINTY AND SENSITIVITY ANALYSIS MANKAMO,T.; et al. Avaplan Oy (Finland). December 1994.

OF CHRONIC EXPOSURE RESULTS WITH THE MACCS RE-190pp.9503010187. BNL-NUREG-52398. 82887:001.

l ACTOR ACCIDENT CONSEQUENCE MODEL. HELTON.J.C. Ar-Technical Specifications (TS) requirements for nuclear power i2ona State Univ., Tempe, AZ. JOHNSON,J.D.; ROLLSTIN J.A.;

plants define the Limiting Conditions for Operation and Surveil-et al. GRAM, Inc. January 1995. 96pp. 9502080054. SAND 93-lance Requirements to assure safety during operation, in gener.

2370. 82644:222, al, these requirements are based on deterministic analysis and Uncertainty and sensitivity analysis techniques based on Latin engineenng judgments. Expenences with plant opvation inds-hyporcube sampling. partial correlation analysis and stepwise cate that some elements of the requirements are unnecessarily regression analysis are used in an investigation with the restrictive, while a few may not be conducive to safety. Iruprov-MACCS model of the chronic exposure pathways associated ing these requirements involves many considerations and is fa-with a severe accident at a nuclear power station. The pnmary cilitated by the availability of plant-specific Probabilistic Safety purpose of this study is to provide guidance on the vanables to Assessments and development of related methods for analyses.

be considered in future review work to reduce the uncertainty in This handbook summanzes the nsk-and reliability-based meth-the important variables used in the calculation of reactor accL ods to improve TS requirements. The scope of the handbook dont consequences. The effects of 75 imprecisely known input includes reliability-and risk-based methods for evaluating al-variables on the following reactor accident consequences are lowed outage times, scheduled or preventive maintenances, studied crop growing season dose, crop long-term dose, water action statements requinng shutdown where shutdown nsk may ingestion dose, milk growing season dose, long-term ground-be substantial, surveillance test intervals, and management of I

l i

m

._.._m 26 Main Citations and Abstracts plant configurations resulting from outages of systems, or com-containment response and source terms supporting the Level 2 ponents. For each topic, the handbook summanzes analytic analysis.

methods with data needs, outlines the insights to be gained, lists additional references, and gives examples of evaluations.

NUREG/CR4144 V01: EVALUATION OF POTENTIAL SEVERE ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-NUREG/CR-6143 V01: EVALUATION OF POTENTIAL SEVERE ATIONS AT SURRY, UNIT 1. Summary Of Results. CHU,T.L; ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-PRATT,W.T. Brookhaven National Laboratory. October 1995.

ATIONS AT GRAND GULF, UNIT

1. Summary Of Results.

WHITEHEAD D.W. Sandia National Laboratories. July 1995.

118pp.9511270460. BNL-NUREG-52399. 86317:223.

This document contains a summarization of the results and 56pp. 9508230248 SAND 93-2440. 85118.215.

This document contains a summarization of the results and insights from the Level 1 accident sequence analyses of inter-insights from the Level 1 accident sequence analyses of inter.

nally initiated events, intemally initiated fire and flood events, nally initiated events, internally initiated fire and flood events, seismically initiated events, and the Level 2/3 nsk analysis of seismically initiated events, and the Level 2/3 risk analysis of internally initiated events (excluding fire and flood) for Surry, internally initiated events (excluding fire and flood) for Grand Unit 1. The analysis was confined to mid-loop operation, which Gulf, Unit 1 as it operates in the Low Power and Shutdown can occur during three plant operational states (identified as Plant Operational State 5 dunng a refueling outage. The report POSs R6 and R10 dunng a refueling outage, and POS D6 summanzes the Level 1 information contained in Volumes 2 - 5 dunng drained traintenance). The report summarizes the Level and the Level 2/3 information contained in Volume 6 of 1 information contained in Volumes 2-5 and the Level 2/3 infor.

NUREG/CR-6143.

mation contained in Volume 6 of NUREG/CR-6144.

NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL NUREG/CR-6144 V06 P1: EVALUATION OF POTENTIAL SEVERE ACCIDENTS DURING LOW POWER AND SHUT-SEVERE ACCIDENTS DURING LOW POWER AND SHUT.

DOWN OPERATIONS AT GRAND GULF, UNIT 1. Evaluation Of DOWN OPERATIONS AT SURRY. UNIT 1. Evaluation Of Severe Severe Accident Risks For Plant Operational State 5 During A Refueling Outage. Main Report And Appendices. BROWN,T.D.;

Accident Risk During Mid-Loop Operations. Main Report. JO,J.;

KMETYK,L.N.; WHITEHEAD,D.W.; et al. Sandia National Lab-LIN C.C.; NEYMOTIN.L.; et al. Brookhaven National Laboratory.

oratones. March 1995. 407pp. 9504050320. SAND 93 2440.

May 1995.

145pp.

9506220089.

BNL-NUREG-52399.

84394:192.

83378:027.

The analysis documented in this volume of the report is the This document contains the accident progression analysis of Level 2/3 analysis of the traditional internal events. Plant internally initiated events for Surry, Unit 1 as it operates in mid-damage states, which define the configuration of the plant and loop operation dunng drained maintenance or a refueling its systems at the onset of core damage for the accident sce.

outage. The report documents the methodology used during the narios developed in the Level 1 analysis, were used to define analysis, desenbes the results from the application of the meth-the interface between the Level 1 and Level 2/3 analyses. In odology, and compares the results with the results from a full the Level 2/3 analysis, the possible progressions of the acci.

power analysis performed on Surry as a part of the Nureg-1150 dent following the onset of core damage were dolineated and study.

the amount of radioactive matenal released to the environment was estimated. Based on the amount of radioactive matenal re-NUREG/CR-6144 V06 P2: EVALUATION OF POTENTIAL leased to the environment, health effects to the general public SEVERE ACCIDENTS DURING LOW POWER AND SHUT.

were estimated. In addition to the offsite consequences, a scop.

DOWN OPERATIONS AT SURRY. UNIT 1. Evaluation Of Severe ing analysis of the potential doses and dose rates within the site Accident Risk During Mid-Locp Operations. Appendices. JO,J.;

were also estimated. The final product of the analysis was the LIN C.C.; NEYMOTIN,L.; et at. Brookhaven National Laboratory.

integration of the accident frequencies with the consequences May 1995.

275pp.

9507120287. BNL-NUREG 52399.

of the accidents to form an expression for aggregate risk.

84633:084.

See NUREG/CR-6144,V06,P1 abstract.

NUREG/CR-6143 V06 P2: EVALUATION OF POTENTIAL SEVERE ACCIDENTS DURING LOW POWER AND SHUT.

NUREG/CR-6150 V01: SCDAP/RELAP5/ MOD 3.1 CODE DOWN OPERATIONS AT GRAND GULF UNIT 1. Evaluation Of MANUAL. interface Theory. ALLISON.C.M/ BERNA G.A/

Severe Accident Risks For Plant Operational State 5 Dunng A CORYELL,E.W.; et al. Idaho National En9ineen"9 Laborat0'Y!

Refueling Outage Supporting MELCOR Calculations.

June 1995. 70pp. 9507180412. EGG-2720. 84716:299.

KMETYK,L.N.; BROWN.T.D. Sandia National Laboratones.

March 1995. 431pp. 9504100148. SAND 93-2440. 83418:001.

The SCDAP/RELAP5 code has been developed for best esti-The document contains the deterministic code calculations mate transient simulation of light water reactor coolant systems performed with the MELCOR Code that were used to support dunng a severe accident. The code models the coupled behav-the development and quantification of the PRA rnodels used in ior of the reactor coolant system, the core, fission products re-the analysis of internally initiated events for Grand Gulf. Unit 1, leased dunng a severe accident transient as well as large and as it operates in the Low Power and Shutdown Plant Operation.

small break loss-of-coolant accidents, operational transients al State 5 during a refueling outage. The background for the such as anticipated transient without SCRAM, loss of offsite work documented in this report is summarized including how de.

power, loss of feedwater, and loss of flow. A gonenc modeling terministic codes are used in PRAs, why the MELCOR code is approach is used that permits as much of a particular system to used, what the capabilities and features of MELCOR are and be modeled as necessary. Control system and secondary how the code has been used by others in the past. Bnef de.

system components are included to permit modeling of plant senptions of the Grand Gulf plant and its configuration dunng controls, turbines, condensers, and secondary feedwater condi.

LP&S operation and of the MELCOR input model developed for tioning systems. This volume desenbes the organization and the Grand Gulf plant in its LP&S configuration are given. The manner of the interface between severe accident models which results of MELCOR analyses of vanous accident secuences for are resident in the SCDAP portion of the code and hydrodynam-the plant operating state (POS) $ configuration dunng refuekng ic models which are resident in the RELAP portion of the code.

(approximately Cold Shutdown as defined by Grand Gulf Tech.

A desenption of the organization and structure of SCDAP/

nical Specifications) are presented for accidents initiated at sev-RELAPS is presented Additional information is provided regard-eral different Dmes after scram and shutdown including short-ing the manner in which models in one portion of the code ened thermal hydraulic and core damage calculations done in impact other parts of the code, and models which are depend-support of the Level 1 analysis and full plant ana'yses including ent on and derive information from other subcodes.

i t

Main Citations and Abstracts 27 NUREG/CR-6150 V02: SCDAP/RELAP/ MOD 3.1 CODE NUREG/CR-6150 V05: SCDAP/RELAP5/ MOD 3.1 CODE MANUAL. Damage Progression Model Theory. ALLISON.C.M.;

MANUAL. Developmental Assessment.

ALLISON,C.M.;

BERNA G.A.; CHENG,T.C.; et al. Idaho National Engineering BERNA,G.A.; BOURDON,S M.; et al. Idaho National Engineenng Laboratory. June 1995. 190pp. 9507180418. EGG.2720.

Laboratory. June 1995. 349pp. 9507180448. EGG-2720.

84713:172.

84717:001.

The SCDAP/RELAP5 code has been developed for best esti.

The SCDAP/RELAPS code has been developed for best esti-mate transient simulation of light water reactor coolant systems mate transient simulation of light water reactor coolant systems dunng a severe accident. The code models the coupled behav-dunng a severe accident. The code models the coupled behav-ior of the reactor coolant system, the core, fission products re-ior of the reactor coolant system, the core, fission product re-leased dunng a severe accident transient as well as large and leased during a severe accident transient as well as large and small break loss of-coolant accidents, operational transients small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A genenc modeling power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary be modeled as necessary. Control system and secondary system components are included to permit modehng of plant system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater condi.

controls, turbines, condensers, and secondary feedwater condi-i tioning systems. This volume contains detailed desenptions of tioning systems. This volume contains detailed code-to-data cal-the severe accident models and correlations. It provides the culations performed using SCDAP/RELAP5/ MOD 3.1, as well as

)

user with the underlying assumptions and simplifications used to comparison calculations performed with earlier code versions.

i generate and implement the basic equations into the code, so The results of full plant calculations which include Surry, TMI-2, an intelligent assessment of the applicability and accuracy of and Browns Ferry are desenbed. The results of a nodalization the resulting calculation can be made.

study, which accounted for both axial and radial nodalization of the core, are also reported.

NUREG/CR-6150 V03: SCDAP/RELAP5/ MOD 3.1 CODE NUREG/CR-6154 V02: EXPERIMENTAL RESULTS FROM CON-MANUAL. User's Guide And input Manual. ALLISON,C.M.,

TAINMENT PIPING BELLOWS SUBJECTED TO SEVERE AC-BERNA,G.A.; CORYELL,E.W.: et al. Idaho National Engineering CIDENT CONDITIONS.Results From Bellows Tested in Cor-Laboratory. June 1995. 352pp. 9507180429. EGG-2720.

roded Conditions. LAMBERT,L.D.; PARKS,M.B. Sandia National 84715:311 s.

ctober 1995. 55pp. 9511020400. SAND 941711.

The SCbAP/RELAP5 code has been developed for best esti-86 mate transient simulation of light-water-reactor coolant systems Bellows are an integral part of the containment pressure dunng a severe accident. The code models the coupled behav-boundary in nuclear power plants. They are used at piping pen-ior of the reactor coolant system, core, fission products re-etrations to allow relative movement between piping and the leased dunng a severe accident transient as well as large and containment wall, while minimizing the load imposed on the small break loss-of-coolant accidents, operational transients piping and wall. Piping bellows are primanly used in steel con-such as anticipated transient without SCRAM, loss of offstte tainments; however, they have received limited use in some power, loss of feedwater, and loss of flow. A generic modeling concrete (reinforced and prestressed) containments. In a severe approach is used that permits as much of a particular system to accident they may be subjected to pressure and temperature be modeled as necessary. Control system and seconda y conditions that exceed the design values, along with a combina-system components are included to permit modeling of plant tion of axial and lateral deflections. A test program to determine controls, turbines, condensers, and secondary feedwater condi^

the leak-tight capacity of containment penetration bellows is tioning systems. This volume, Volume 3, provides guidelines to being conducted at Sandia National Laboratories under the code users based upon lessons learned during the developmen*

sponsorship of the U.S. Nuclear Regulatory Commission. Sever-tal assessment process. A description of problem control and al different bellows geometnes, representative of actual contain-the installation process is included. Appendix A contains the de-ment bellows, have been subjected to extreme deflections senption of the input requirements-along with pressure and temperature loads. The bellows ge-es aM loa &ng cMons am &scM abg e N NUREG/CR-6150 V04: SCDAP/RELAPS/ MOD 3.1 CODE testing apparatus and procedures. A total of nineteen bellows MANUAL.MATPRO--A Libra 7 Of Materials Properties For Light.

have been tested. Thirteen bellows were tested in,like-new Water-Reactor Accident Analysis. ALLISON,C M.; BERNA,G.A.;

condition (results reported in Volume 1), and six were tested in CHAMBEFS R ; et a!. Idaho National Engineenng Laboratory.

a corroded condition. The tests showed that bellows in "like-June 1995. 674pp. 9507180441. EGG-2720. 84714.001.

new" condition are capable of withstanding relatively large de-The SCDAP/RELAPS code has been developed for best esti-formations, up to, or near, the point of full compression or elon-mate transient simulation of light-water-reactor coolant systems gation, before developing leakage, while those in a corroded dunng a severe accident. The code models the coupled behav-condition did not perform as well, depending on the amount of ior of the reactor coolant system, core, fission products re-corrosion. The corroded bellows test program and results are leased during a severe accident transient as well as large and presented in this report.

small break loss-of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsito NUREG/CR-6159: USING MICRO SAINT TO PREDICT PER-power, loss of feedwater, and loss of flow. A genenc modeling FORMANCE IN A NUCLEAR POWER PLANT CONTROL approach is used that permits as much of a particular system to ROOM.A Test Of Validity And Feasibility. LAWLESS.M.T.:

be modeled as necessary. Control system and secondary LAUGHERY,K.R.

Micro Analysis

Design, Inc.

system components are included to permit modeling of plant PERSENSKY,J.J. Control, instrumentation & Human Factars controls, turbines, condensers, and secondary feedwater condi-Branch (Post 941217). September 1995. 38pp. 9510030209.

tioning systems. This volume, Volume 4 descnbes the matenal 85663.255.

properties correlations and computer subroutines (MATPRO)

Yhe United States Nuclear Regulatory Comrrission (NRC) re-used by SCDAP/RELAPS. Formu!ation of the matenals proper-r,uires a technical basis for regulatory actions. In the area of ties are generally semi-empincal in nature. A wide vai aty of ma-loman factors, one possible technical basis is human perform-terial property subroutines are contained in this document. This a.v.,e rnodeling technology including task network modeling document also contains descriptions of the reaction and solu-This study assessed the feasibility and validity of task network tion rate models needed to analyze a reactor accident, modeling to predict the performance of control room crews.

i l

]

28 Main Citations and Abstracts Task network models were built that matched the expenmental tearir,g and to develop reasonably simple analytical methods for 1

conditions of a study on computenzed procedures that was con-predicting when teanng will occur. Three sets of test specimens l

ducted at North Carohna State Universrty, The data from the were designed to allow individual control over and investigation j

" paper procedures" conditions were used to calibrate the task of the mechanisms believed to be important in causing failure of network models. Then, the models were manipulated to reflect the hner plate. The senes of tests investigated the effect on expected changes when computerized procedures were Lsed.

liner tearing produced by the anchorage system, the loading These models' predictions were then compared to the expen-conditions, and the transition in thickness from the liner to the mental data from the "computenzed conditions" of the North insert plate. Before testing, the specimens were analyzed using Carohna State University study. Analyses indicated that the two-and three-dimensional finite element models. Based on the models predicted some subsets of the data well, but not all im-analysis, the failure mode and corresponding load conditions placations for the use of task network modeling are discussed.

were predicted for each specimen. Test data and post-test ex-

)

NUREG/CR-6172: REVtEWING PSA.8ASED ANALYSES TO amination of test specimens show mixed agreerrrant with the

]

MODIFY TECHNICAL SPECIFICATIONS AT NUCLEAR POWER analytical predictions with regard to failure mode arid specimen j

PLANTS. SAMANTA,P.K.; MARTINEZ-GURIDI Brookhaven Na-response for most tests. Many similanties were also observed j

tional Laboratory. VESELY,W.E. Science Applications interne-een me msp nse d N Hnw in N M-scab snWcM tional Corp. (formerly Science Applications, Inc.). December concmW cetainmet W and N respmse of me kst j

1995.62pp.9601040277. BNL-NUREG-52428. 86730:170.

specimens. This work illustrates the fact that the failure mecha.

{

Changos to Technical Specifications (TSs) at nuclear power nism of a reinforced concrete containment building can be plants (NPPs) require review and approval by the United States greatly influenced by details of kner and anchorage system Nuclear Regulatory Commission (USNRC). Currently, many re-sign. Mer, d signkaW scmases N estaW of conta me ing respmse unds seem cMions.

I quests for changes to TSS use analyses that are based on a plant's probablistic safety assessment (PSA). This report pre

• NUREG/CR-6188 V02: MICROBIAL DEGRADATION OF LOW.

sents an approach to reviewing such PSA-based submittals for LEVEL RADIOACTIVE WASTE. Annual Report For FY 1994.

changes to TSs. We discuss the basic objectives of reviewing a ROGERS.R.D.; HAMILTON,M.A.; VEEH R.H.; et al. Idaho Na-PSA-based submittal to modify NPP TSs; the methodology of tional Engineenng Laboratory. August 1995.66pp.9509130162.

reviewing a TS submittal, and the diffenng roles of a PSA INEL-95/0153. 85430.287, review, a PSA Computer Code review, and a review of a TS The Nuclear Regulatory Commission stipulates in 10 CFR 61 submittal. To illustrate this approach, we discuss our review of that disposed low-level rtdioactive waste (LLW) be stabilized.

changes to allowed outage time (AOT) and surveillance test in-To provide guidance to disposal vendors and nuclear station terval (STI) in the TS for the South Texas Project Nuclear Gen-waste generators for implementing those requirements, the erating Station. Based on this expenence gained, a check-list of NRC developed Technical Position on Waste Form, Revision 1.

{

items is given for future reviewers; it can be used to venfy that That document details a specified set of recommended testing the submittal contains sufficient information, and also that the procedures and entena, including several tests for determining review has addressed the relevant issues. Finally, recommend-the biodegradation properties of waste forms. Cement has been ed steps in the review process and the expected findings of widely used to solidify LLW; however, the resulting waste forms each step are discussed.

are sometimes susceptible to failure due to the actions of waste NUREG/CR-6173: A

SUMMARY

OF THE FIRE TESTING PRO.

constituents, stress, and environment. The purpose of this re.

GRAM AT THE GERMAN HDR TEST FACILITY. NOWLEN,S.P.

search program is to develop modified microbial degradation Sandia National Laboratories. November 1995. 57pp.

test procedures that will be more appropnate than the existing j

9511290120. SAND 94-1795. 86353.298.

procedures for evaluating the effects of microbiologically influ-This report provides an everview of the fire safety expen.

enced chemical attack on cement-solidified LLW. Groups of ments performed under the sponsorship of the German govern.

microorganisms indigenous to LLW disposal sites are being em-ment in the containment building of the decommissioned pilot ployed that can metabolicelly convert organic and inorganic nuclear power plant known as HDR. This structure is a highly substrates into organic and mineral acids. Such acids aggres-complex, multicompartment, multi-level building which has been sively react with cement and can ultimately lead to structural used as the test bed for a wide range of nuclear power plant failure. Results over the past year on the application of mecha.

ooeration safety expenments. These expenments have included nisms inherent in microbially influenced degradation of cement-numerous fire tests. Test fire fuel sources have included gas based matenal are the focus of this annual report. Data-validat-burners, wood cnbs, oil pools, nozzle release oil fires, and ed evidence of the potential for microbially influenced detenora-cables in cable trays. A wide range of ventilation conditions in.

tion of cement sohdified LLW and subsequent release of radion-cluding full natural ventilation, full forced ventilation, and com.

uclides has been developed dunng this study.

bined natural and forced ventilation have been evaluated.

NUREG/CR-6 *91: SIZE AND DEFORMATION LIMITS TO MAIN-During most of the tests, the fire products mixed freely with the full containment volume. Macro-acale building circulation pat-TAIN CONST6AINT IN K(IC) AND J(C) TESTING OF BEND terns which were very sensitive to such factors as ventilation SPECIMENS. KOPPENHOEFER,K.; DODOS R.H. I!!inois, Univ of Urbana, IL. October 1995. 27pp. 9511270461. 86322:329 configuration were observed and charactenzed. Testing also in-cluded the evaluation of selective area pressunzation schemes The ASTM Standard Test Method for Plane-Strain Fracture f smoke control for emergency access and evacu-Toughness of Metallic Materials (E399-90) restncts test speci-men dimensions to insure the measurement of highly con-strained fracture toughness values (X(IC)). These requirements NUFIEG/CR-6184: SEPARATE EFFECTS TESTING AND ANALY.

insure small-sc91e yielding (SSY) conditions at fracture, and SES TO INVESTIGATE LINER TEARING OF THE 1:16-SCALE thereby the vtMity of linear elastic fracture mechanics. Recent-REINFORCED CONCRETE CONTAINMENT BUILDING.

ly, Dodds and Anderson have proposed a less restnctive size SPLETZER.B.L; LAMBERT,L.D.; BERGMAN,V.L.; et al. Sandia requirement for cleavage fracture toughness measured in terms National Laboratones. May 1995. 250pp. 9507200213.

of the Jantegral (J(C)), as give, by a, b, B 2 200 J(C) /4 SAND 92-1720. 84749:001.

The size requirement proposed by Dodds and Anderson in-The overpressurization of a 1:6-scale reinforced concrete creases the apphcability of fracture toughness expenments by containment building demonstrated that lener teanng is a plausi-expanding the range of conditions over which fracture tough-ble failure mode in such structures under severe accident condi-ness data meeting SSY conditions can be reliably measured.

tions. A combined expenmental and analytical program was de-This investigation compares the proposed size requirement with veloped to determine the important parameters that affect liner that of ASTM Standard Test Method E399 and, by companson

Main Citations and Abstracts 29 with published expenmental data for various alloys. provides NUREG/CR-6224: PARAMETRIC STUDY OF THE POTENTIAL validation of the new requirements FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GEN-ERATED DEBRIS. ZlGLER,G.; BRIDEAU,J.; RAO.D.V.; et al.

(

NUREG/CR-6192: AGING AND SERWCE WEAR OF SPRING-Science & Engineering Associates, Inc. October 1995. 465pp.

LOADED PRESSURE RELIEF VALVES USED IN SAFETY-RE-9512260318. SEA 93-554-06A1. 86607:001.

LATED SYSTEMS AT NUCLEAR POWER PLANTS.

This report documents a plant specific study for a BWR/4 STAUNTON,R.H.; COX,D.F. Oak Ridge National Laboratory.

with a Mark I containment that evaluated the potential 's LOCA Uerch 1995. 86pp. 9504180472. ORNL-6791. 83568:226.

generated debns and the probabahty of losing long term recirru-

%Sg-loaded pressure relief valves (PRVS) are used in some lation capability due ECCS pump suction strainer blockage. he WW Blated applications at nuclear power plants. In general, major elements of this study were: (1) acquisition of detailed tiey are used in systems where, dunng accidents, pressures piping layouts and installed insulation details for a derence i b nse to levels where pressure safety relief is required for BWR; (2) analysis of plant specific piping weld failure probabil-p stection of personnel, system piping, and components. This ities to estimate the LOCA frequency; (3) development of an in-sulation and other debns generation and dryweK transport repet documents a stu$ of PRV aging and considers the se-models for the reference BWR; (4) modeling of debns transport venty ad causes of sec,.ce wear and how it is oiscovered and in the suppression pool; (5) development of strainer blockage corrected in various systems, valve sizes, etc. Provided in this head loss models for estimating loss of NPSH margin; (6) esti-report are results of the examination of the recorded failures a n cm damage Weny aunWe b bss d M and identification of trends and relationships / correlations in the recirculation capability following a LOCA. Elements 2 through 5 failures when all failure related parameters are considered.

were combined into a computer code, BLOCKAGE 2.3. A point Components that compnse a typical PRV, how those compo-estimate of overall DEGB pipe break frequency (per Rx-year) of nents fail, when they fait, and the current testing frequencies 1.59E-04 was calculated for the reference plant, with a corre-and methods are also presented in detail-sponding overall ECCS loss of NPSH frequency (per Rx-year) oi 1.58E-04. The calculated point estimate of core damage fre-NUREG/CR-6214: PRODUCTi% / ND TESTING OF THE VITA-quency (per Rx-year) due to blockage related accidsnt se-MIN CC TINE GROUP AND IdE BUGLE-93 BROAD-GROUP quences for the reference BWR ranged from 4.2E-06 to 2.5E-NEUTRON / PHOTON CROS3 SECTION LIBRARIES DERIVED

05. The results of thir M show that unacceptable strainer FROM ENDF/B-VI NUCLEAR DATA. INGERSOLL.D.T.;

blockage and loss of NPSH *nargin Can occur within the first WHITE,J.E.; WRIGHT,R.O.; et al. Oak Ridge National Laborato-few minutes after ECCJ m. mss achieve maximum flows when ry. January 1995.172pp. 9502080096. ORNL-6795. 82650:146.

the ECCS strainers are ewesed to LOCA generated fibrous A new multigroup cross-section isbrary based on ENDF/B-VI debris in the presence of particulates (sludge, paint chips, con-data has been produced and tested for light water reactor crete dust). Generic or unconditional extrapolation of these ref-shielding and reactor pressure vessel dosimetry apphcations, erence plant calculated results should not be undertaken.

The broad-group library, which is designated BUGLE-93, is in-NUREG/CR-6235: ASSESSMENT OF SHORT THROUGH-WALL tended to replace the aging BUGLE-80 and SAILOR hbranes.

CIRCUMFERENTIAL CRACKS IN P! PES.Expenments And The processing methodology is consistent with ANSI /ANS Analysis, March 1990 - December 1994. BRUST,F.W.;

6.1.2, since the ENDF data were first processed into a fine-SCOTT,P.; RAHMAN.S.; et al. Battelle Memorial Institute, Co-group, pseudo-problem-independent format and then collapsed lumbus Laboratories. April 1995. 249pp. 9505040365. BMI-into the final broad-group format. The fine-group library, which is 2179. 83784:193.

designated VITAMIN-86, contains 120 nuchdes. The BUGLE-93 This topical report summanzes the work performed for the 47-neutron-group /20-gamma ray-group hbrary contains the Nuclear Regulatory Commission's (NRC) research program enti-same 120 nuclides processed as infinitely dilute and collapsed tied "Short Cracks in Piping and Piping Welds" that specifically using a weighting spectrum typical of a concrete shield. Addi-focuses on pipes with short through-wall cracks. Previous NRC tionally, BUGLE 93 contains 105 nuclides processed with reso-efforts, conducted under the Degraded Piping Program, focused nance self-shielding and weighted using spectra specific to on understanding the fracture behavior of larger cracks in piping BWR and PWR matenal compositions and reactor models. Sev-and fundamental fracture mechanics developments necessary eral dosimetry response functions and kerma factors for all 120 for this technology. This report gives details on: (i) material nuclides are also included with the library. An extensive integral property determinations (ii) pipe fracture experiments, and (iii) data testing effort was performed to quahfy the new library. In development, modification, and validation of fracture analysis general, results using the new data show significant improve-methods. The material property data required to analyze the ex-ments relative to earlier ENDF data.

penmental results are included. These data were also imple-mented into the NRC's PIRFAC database. Threo pipe experi-NUREG/CR-6220: AN ASSESSMENT OF FIRE VULNERABILITY ments with short through-wall cracks were conducted on lary3 FOR AGED ELECTRICAL RELAYS. VIGIL,R.A. Science & Engi-diameter pipe. Also, experiments wore conducted on a large-op neenng Associates, Inc. NOWLEN,S.P. Sandia National Labora-ameter uncracked pipe and a pipe with a moderate-size tories. M arch 1995. 38pp. 9504100121. SAND 94-0769.

through-wall crack. The analysis results reported here focus on 83419 ND simple predictive methods based on the J Toanng theory a well as limit-load and ASME Section XI analyses. Some o This report details testing to assess the impact of aging on these methods were improved for short-crack-length predic-the fire vulnerability of Agastat and General Electric relays. Both tions. The accuracy of the various methods was determined by aged and unaged relays were tested. Aged relays were subject-comparisons with expenmental results from this and other pro-ed to operational cycling under rated load and thermally aged grams.

for sixty days. All relays were exposed to one of bree different fire temperature profiles in the Severe Combined Enviroraents NUREG/CR-6239 V01: 6UHVE( OF STRONG MOTION EARTH-Test Chamber located at Sandia National Laboratories. The OUAKE EFFECTS ON THcRMAL POWe6 PLANTS IN CALL-ability to operate properly in the given fire environment was FORNIA WITH EMPHASIS ON P'F;NG SYSTEMS. Main Report.

I monitored. Results for the aged and unaged relays were exam-STEVENSOV LD. Stevenson & AssociMes.

• Oak Ridge Na-ined to determine the impact of agirn on the relays' abihty to tional Laboratory. November 1995. 107pp. 9512260280.

)

sustain operation under the test conditions. Overall results indi-ORNLSUB94SD4272. 86626:246.

cated that the aged relays' performance was not significantly Since 1982, there has been a major effort expended to evalu-different from that of the unaged relays.

Ste the susceptibihty of nuclear power plant equipment to failure l

i

30 Main Citations and Abstracts and significant damago dunng seismic events. This was done by AGE CASKS): A MICROCOMPUTER BASED ANALYSIS making use of data on the periormance of electncal and me-SYSTEM FOR STORAGE CASK DESIGN REVIEW. User's chanical equipment in conventional power plants and other semi-Manual To Version ib (including Program Reference).

lar industnal facilities dunng strong motion earthquakes. This CHEN,T.F.; GERHARD,M.A.; TRUMMER,D.J.; et al. Lawrence report is intended as an extension of the seismic experience Livermore National Laboratory. February 1995. 200pp.

data collection effort and a compilation of expenence data spe.

9503140405. UCRL-ID-117418. 83070:058.

cific to power plant piping and supports designed and construct.

CASKS (Computer Anafysis of Storage casks) is a micro-ed to U.S. power piping code requirements which have experi.

computer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory enced strong motion earthquakes. Eight damaging (Richter

(

)

WaluaMg sa% anahds WS on $$W sbD Magnitude 7.7 to 5.5) California earthquakes and their effects age casks. The bulk of the complete program and this user s on 8 puwer 9enerating facihties in Cal;fornia were reviewed. All manual are based upon the SCANS (Shipping Cask Analysis of these facilities were visited and evaluated. Seven fossil-System) program previously developed at LLNL. A number of fueled (dual use natural gas and oil) and one nuclear fueled enhancements and improvements were added to the original plants consisting of a total of 36 individual boiler or reactor units SCANS program to meet requirements unique to storage casks.

were investigated. Peak honzontal ground accelerations that CASKS is an easy-to-use system that calculates global re-either had been recorded on site at these facilities or were con-sponse of storage casks to impact loads, pressure loads and sidered applicable to these power plants on the basis of nearby thermal conditions. This provides reviewers with a tool for an in-recordings ranged between 0.20g and 0.51g with wrong motion dependent check on analyses submitted by licensees. CASKS durations which vaned from 3.5 to 15 seconds. Most U.S. nucle-is based on microcomputers compatible with the IBM-PC family ar power plants are designed for a safe shutdown earthquake of computers. The system is composed of a series of menus, peak ground acceleration equal to 0.20g or less with strong input programs, cask analysis programs, and output display pro-motion durations which vary from 10 to 15 seconds.

grams. All data is entered through fill-in-the-blank input screens that contain desenptive data requests.

NUREG/CR-6239 V02: SURVEY OF STRONG MOTION EARTH-OUAKE EFFECTS ON THERMAL POWER PLANTS IN Call.

NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT CONSE-FORNIA WITH EMvHASIS ON PIPING SYSTEMS. Appendices.

QUENCE UNCERTAINTY ANALYSIS. Dispersion And Deposi-STEVENSON,J.D. Stevenson & Associates.

• Oak Ridge Na-tion Uncertainty Assessment. Main Report. HARPER,F.T. Sandia National Laboratones. GOOSSENS,L.H.J.; COOKE,R.M.; et al.

tional Laboratory. November 1995. 228pp. 9512260283.

Netherlands, Govt. of. January 1995.106pp. 9503150202. EUR ORNLSUB94SD4272. 86608:102.

15855EN. 83108:230.

Volume 2 of the " Survey of Strong Motion Earthquake Effects The development of two new probabilistic accident conse-on Thermal Power Plants in California with Emmsis on Piping quence codes, MAC"S and COSYMA, was completed in 1990.

Systems" contains Appendices which detail the detail design These codes estimate the consequences from the accidental and seismic response of several power plants subjected to releases of radiological matenal from hypothesized accidents at strong motion earthquakes. The particular pMnts r.onsidered in-nuclear installations. In 1991, the U.S. Nuclear Regulatory Com-clude the Ormond Beach, Long Beach and Seal Beach, Bur-mission and the Commission of the European Communities bank, El Centro, Glendale, Humboldt Bay, Kern Valley, Pasade-began co-s>onsonng a joint uncertainty analysis of the two na and Valley power plants. Included is a typical power plant codes. The ultimate objective of this joint effort was to system-piping specification and photographs of typical power plant stically develop credible and traceable uncertainty distnbutions piping specification and photographs of typical piping and sup-for the resy tive code input vanables. Because of the magni-port installations for the plants surveyed. Detailed piping support tude and expense required to complete a full-scale conse-spacing data are also included.

quence uncertainty analysis, a trial study was performed to eva!uate the feasibility of euch a joint study by initially limiting NUREG/CR-6240: APPLICATION OF BOUNDING SPECTRA TO efforts to the dispersion and deposition code input variables. A SEISMIC DESIGN OF PIPING dASED ON THE PERFORM-formal expert judgment elicitation and evaluation process was ANCE OF ABOVE GROUND PIPING IN POWER PLANTS SUB-identified as the best technology available for developing a li-JECTED TO STRONG MOTION EARTHOUAKES.

brety of uncertainty distributions for these consequence param-STEVENSON.J.D. Steve ison & Associates.

• Oak Ridge Na-eters. This report focuses on the methods used in and results tional Laboratory. Fet ruary 1995. 104pp. 9503140368.

of tus tnal study.

ORNLSUB94SD4271. 83039:312.

NUREG CR-6244 V02: PROBABILISTIC ACCIDENT CONSE-This report extends tra potential application of Bounding OUEhCE UNCERTAINTY ANALYSIS. Dispersion And Deposi-Spectra evaluation proc;dures, developed as part of the A-46 tion Uncertainty Assessment. Appendices A And B.

Unresolved Safety issue applicable to seismic venfication of in-HAF PER.F T. Sandia National Laboratones. GOOSSENS,L.H.J.;

situ electrical and mechanical equipment, to in-situ safety relat-CCVKE.R.M.; et al. Netherlands, Govt. of. January 1995.400pp.

ed piping in nuclear power plants. The report presents a sum-9503150209. EUR 15855EN. 83109 001.

mary of earthquake experience data which define the behavior See NUREG/CR-6244,V01 abstract.

of typical U.S. power plant piping subject to strong motion NUREG/CR-6244 V03: PROBABILISTIC ACCIDENT CONSE-earthquakes. The report defines those piping system caveats which would assure the seismic adequacy of the piping systems OUENCE UNCERTAINTY ANALYSIS. Dispersion and Deposi-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 L.H.J.;

within the bounding spectra input. Based Go the observed be-

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

havior of piping in sfong motion earthquakes, be report distin-9503140347. EUR 15855EN. 83069:214.

guishes between the capabilities of the piping sy item to carry See NUREG/CR.6244,V01 abstract.

seismic loads as a function of the type of c0r'nection (i.e.,

threaded vs wolded). This report also dscusses in some detail NUPEG/CR-6251: STAINLESS STEEL SUBMERGED ARC WEl.D the basic causes and mechanisms for earthquake damages and FUSION LINE TOUGHNESS. ROSENFIELD,A.R.; HELD,r).R.;

failures to power plant piping systems.

W LKOWSK;.G M. Battelle Memonal institute, Columbus Labora-tones. April 1535. 88pp. 9505190023. BMI-2180. 83980:197.

NUREG/CR-6242: CASKS (COMPUTER ANALYSIS OF STOR-This effort evaivated the fracture toughness of austenstic steel submerged arc welc (SAW) fusion lines The incentive was to

Main Citations and Abstracts 31 explain why cracks grow into the fusion line in many pipe tests NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING conducted with cracks initially centered in SAWS. The concern ATOMIC ENERGY OF CANADA LTD CODES. JEDD,J.L.;

was that the fusion line may have a lower toughness than the SHUMWAY,R.W. Idaho National Engineenng Laboratory.

SAW. It was found that the fusion line, J(Ic) was greater than EBERT,D.D. Reactor & Plant Systems Branch (Post 941217).

the SAW toughness but much less than the base metal. Of January 1995. 88pp. 9503270305. INEL-95/0070. 83268:267.

greater importance may be that the crack growth resistance A limited number of transient scenarios were calculated using (J(D)-R) of the fusion line appeared to reach a steady-state a computer code suite and input modeling provided by the value, while the SAW had a continually increasing J(D)-R curve.

Atomic Energy of Canada Limited (AECL) for the CANDU 3 This explains why the cracks eventually turn to the fusion line in design. Emphasis was placed on a large-break loss-of-coolant the pipe expenments. A method of incorporating these results accident with delays in actuation of the two independent shut-would be to use the weld metal J-R curve up to the fusion-line down systems (shutdown rods and liquid poison injection). Al-steady-state J value. These results may be more important to though an extremely unlikely scenario, it was studied because LBB analyses than the ASME flaw evaluation procedures, since of the potential consequences that would result from a positive there is more crack growth with through-wall cracks in LBB void coefficient of reactivity. Results indicate that a few seconds analyses than for surface cracks in pipe flaw evaluations.

delay in shutdown would result in quickly reaching fuel or clad-ding melting temperatures before the emergency core cooling NUREG/CR-6256 V01: FIELD LYSIMETER INVESTIGATIONS -

system would be activated. Only small changes in the timing TEST RESULTS Low-Level Waste Data Base Program. Test Re.

and consequences of the scenano result when several param-suits For Fiscal Years 1986, 1987,1988, And 1989.

eters, of potential importance to the progression of the acci-MCCONNELL.J.W.; ROGERS.R.D.; FINDLAY,M.W.; et al. Idaho dent, are varied. Five calculations were also performed for loss-National Engineenng Laboratory. May 1995.

134pp.

of-off-site-power scenarios. These calculations assume that the 9506020457. INEL 95/0073. 84147.001, piant failed to enter the island mode, i.e., power to the main The Field Lysimeter Investigations: Low-Level Waste Data coolant pumps was not restored using on-site power generation.

Base Development Program, funded by the U S. Nuclear Regu-NUREG/CR 6259: CONSTRAINT EFFECTS ON FRACTURE INI-latory Commission (NRC), is (a) studying the degradation effects TIATION LOADS IN HSST WIDE-PLATE TESTS. DODDS A f L in EPICOR-il organic ion-exchange resins caused by radiation, Illinois, Univ. of. Urbana, IL. DODDS,R.H. Oak Ridge National (b) examining the adequacy of test procedures recommended in Laboratory. December 1994. 42pp.

9502080251.

UI-the Branch Technical Position on Waste Form to meet the re-LUENG942009. 82660:304 quirements of 10 CFR 61 using solidified EPICOR-il resins, (c)

Dunng the penod 1984 1987, researchers of the Heavy-Sec-obtaining performance information on solidified EPICOR-li ion-tion Steel Technology program at the Oak Ridge National Labo-exchange resins in a disposal environment, and (d) determining ratory performed a unique senes of fracture mechanics tests the condition of EPICOR-il liners. Results of the first 4 years of using exceptionally large, SE(T) specimens (a/W=0.2) fabricat-data acquisition from the field testing are presented and dim ed from a reactor pressure vessel material, A533B Class 1 cussed. Dunng the continuing field testing. both Portland type l-steel. This study re-examines fracture initiation loads in the ll cement and Dow vinyl ester-styrene waste forms are being wide-plate tests using two constraint assessment methodologies tested in lysimeter anays located at Argonne National Laborato" developed over the past five years: the J-O toughness locus ap-ry-East in Illinois and at Oak Ridge National Laboratory. The ex-proach and the toughness scaling approach based on a local penmental equipment is descnbed and results of waste form failure enterion for cleavage. Both upproaches demonstrate a characterization using tests recommended by the NRC's " Tech-significant loss of constraint in the elastic-plastic fields ahead of nical Position on Waste Form" are presmted. The study is de' the crack in the wide-plate specimens caused by the inherent signed to provide continuous data on n4de release and move-negative T-stress of the shallow notch SE(T) configuration.

ment, as well as environmental conditions, over a 20-year Moreover, the 25mm wide machined notch required for speci-penod.

men fabncation is shown to further reduce constraint by intro-9*

NUMG/CR-6256 V02: FIELD LYSIMETER INVESTIGATIONS these factors combined to reduce near tip stresses by 10%

TEST RESULTS. Low-Level Waste Data Base Development Pro-below those of the small-scale yieldng, SSY (T=O), fields. This gram: Test Results For Fiscal Years 1990,1991,1992, And 1993-reduction places fracture results for the wide-plate specimens MCCONNELL J.W.; ROGERS,R.D. Idaho National Engineenng within the J O toughness locus defined by fracture toughness Labo% tory. JASTROW,J.D.; et al. Argonne National Laboratory.

tests on the A5338 material and within the constraint corrected Decamber 1995.

170pp.

9601290286.

INEL-95/0073.

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

868 M 016.

The Field Lysimeter Investigations: Low-Level Waste Data NUREG/CR-6260: APPLICATION OF NUREG/CR-5999 INTERIM Base Development Program, funded by the U.S. Nuclear Regu-FATIGUE CURVES TO SELECTED NUCLEAR POWER PLANT latory Commission (NRC), is (a) studying the degradation effects COMPONENTS. WARE.A.G.; MORTON.D.K.; NITZEL M.E.

in EPICOR-il organic ion-exchange resins caused by radiation, Idaho National Engineering Laboratory. March 1995. 200pp.

(b) examining the adequacy of test procedures recommended in 9503280383. INEL-95/0045. 83293:001.

the Branch Technical Position on Waste Form to meet the re-Recent test data ino.cate that the effects of the light water quirements of 10 CFR 61 using solidified EPICOR-il resins, (c) reactor (LWR) environment could significantly reduce the fatigue obtaining performance information on solidified EPICOR-Il son-resistance of materials used in the reactor coolart pressure exchange resins in a disposal environment, and (d) determining boundary components of operating nuclear power plants. Ar-the condition of EPICOR-li liners Resutts of the second 4 years gonne National Laboratory has developed interim fatigue curves of data acquisition from the field testing are presented and dis-based on test data simulating LWR conditions, and published cussed. Dunng the continuing field testing both portland type l-them in NUREG/CR 5999. In order to assess the significance of Il cement and Dow vinyt ester-styrene waste forms are being these intenm fatigue curves, fatigue evaluations of a sample of tested in lysimeter arrays located at Argonne National-East in 11-the components in the reactor coolant pressure boundary of linois and at Oak Ridge National Laboratory. The expenmental LWRs were performed. The sample consists of components equipment is desenbed and results of waste form charactenza-from facilities designed by each of the four U.S. nuclear steam i

tion using tests recommended by the NRC's " Technical Position supply system vendors. For each facility, six locations were l

on Waste Form" are presented. The study is designed to pro-studied, including two locations on the reactor pressure vessel.

vide continuous data on nuclide release and movement, as well in addition, there are older vintage plants where components of as environmental conditions, over a 20-year penod.

the reactor coolant pressure boundary were designed to codes I

I

32 Main Citations and Abstracts that did not require an exphcit fatigue analysis of the compo-rameter was introduced. Initial results using J(M) were encour-nents. In order to assess the fatigue resistance of the older vin-aging but subsequent studies did not support the earlier results.

tage plants, an evaluation was also conducted on selected The present computational study presented in Volume 1 of this components of three of these plants. This report discusses the report investigates several forms of this parameter, how they insights gained from the application of the intenm fatigue curves are derived and the validity of these parameters for small and to components of seven operating nuclear power plants.

large amounts of crack growth. It is concluded that neither J nor NUREG/CR-6261: A

SUMMARY

OF ORNL FISSION PRODUCT J(M) (nor any single parameter) can satisfactonly capture the RELEASE TESTS WITH RECOMMENDED RELEASE RATES full range of near tip fracture states. A discussion on the range AND DIFFUSION COEFFICIENTS.

LORENZ,R.A.;

of vahdity of J(M) is given in Volume 2.

OSBORNE M.F. Oak Ridge National Laboratory. July 1995.

NUREG/CR-6264 V02: VALIDITY LIMITS IN J-RESISTANCE 83pp.9509070074. ORNL/TM 12801. 85401:146.

Fission product release data from the ORNL Hi test series CURVE DETERMINATION.A Computational Approach To Duc-tile Crack Growth Under Large-Scale Yielding Conditions.

and VI test series are summarized in this report and compared with release results from similar tests performed in France (the SHIH C.F.;

XIA,L Brown Univ.,

Providence, RI.

HEVA test series). The ORNL test results are also compared HUTCHINSON,J W.; et al. Harvard Univ., Cambridge, MA. Feb-with four in-reactor tests, the SNL ST 1 and ST-2 tests, and the ruary 1995. 50pp. 9503140430. BMI-2181. 83070:317.

INEL SFD 13 and SFD 14 bundle tests. Test atmospheres in this report, Volume 2. Mode I crack initiation and growth range from steam to hydrogen, and the temperature range is under plane strain conditions in tough metals are computed 1675 to 2700 K.

using an elastic / plastic continuum model which accounts for v id growth and coalescence ahead of the crack tip. The mate-NUREG/CR-6263 V01: HIGH INTEGRITY SOFTWARE FOR NU-rial parameters include the stress-strain properties, along with CLEAR POWER PLANTS Candidate Guidelines, Technical Basis the parameters Charactenzing the spacing and volume fraction And Research Needs Executive Summary. SETH,S.; BAILW.;

of voids in matenal elements lying in the plane of the crack. For CLEAVES,D.; et al.

MITRE Corp. June 1995. 86pp.

a given set of these parameters and a specific specimen, or 9508020314. MTR 94W0000114. 84908:262.

The work documented in this report was performed in support component, subject to a specific loading, relationships among i

of the U.S. Nuclear Regulatory Commission to examine the load, load-line displacement and crack advance can be comput-technical basis for candidate guidehnes that could be consid-ed with no restrictions on the extent of plastic deformation.

ered in reviewing and evalua'mg high integnty computer soft.

Similarly, there is no hmit on crack advance, except that it must ware used in the safety systems of nuclear power plants. The take place on the symmetry plane ahead of the initial crack.

framework for the work consisted of the following software de.

Suitably defined measures of crack tip loading intensity, such as velopment and assurance activities; requirements specification; those based on the J-integral, can also be computed, thereby l

design; coding; venfication and vahdation, including static analy, directly generating crack growth resistance curves. In this e

sis and dynamic testing; safety analysis; operation and mainte-report, the model is applied to five specimen geometnes which nance; configuration management; quahty assurance; and plan-are known to give rise to significantly different crack tip con-ning and management. Each activity (framework element) was straints and crack growth resistance behaviors. Computed re-subdivided into technical areas (framework subelements). The suits are compared with sets of experimental data for two tough report desenbes the development of approximately 200 candi.

steels for four of the specimen types. Details of the load, dis-date guidehnes that span the entire range of software hfe-cycle placement and crack growth histones are accurately repro-activities; the assessment of the technical basis for those candi.

duced, even when extensive crack growth takes place under date guidehnos; and the identification, categorization and prionti-conditions of fully plastic yieldang.

Zation of research needs for improving the technical basis. The i

report has two volumes: Volume 1. Executive Summary, in.

NUREG/CR-6265: MULTIDISCIPLINARY FRAMEWORK FOR cludes an overview of the framework and of each framework HUMAN RELIABILITY ANALYSIS WITH AN APPLICATION TO l

element, the complete set of candidate guidelines, the results of ERRORS OF COMMISSIC AND DEPENDENCIES.

the assossment of the technical basis for each candidate guide.

BARRIERE.M.T.

Brookhaven National Laboratory, kne, and a discussion of research needs that support the regu.

WREATHALL.J. John Wreathall & Co., Inc. COOPER.S.E.; et al.

latory function; Volume 2 is the main report.

Science Apphcations Intemational Corp. (formerly Science Ap-k[G 5243. 85539 094.

NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NU-CLEAR POWER PLANTS. Candidate Guidehnes. Technical Basis Since the early 1970s, human reliabihty analysis (HRA) has And Research Needs. Main Report. SETH,S.; BAIL,W.;

been considered to be an integral part of probabshstic risk as-CLEAVES D.; et al. MITRE Corp. June 1995. 430pp.

sessments (PRAs). Nuclear power plant (NPP) events, from 9508020315. MTR 940000114. 84909:001.

See NUREG/CR-6263,V01 abstract.

Three Mile Island through the mid-1980s, showed the impor-tance of human performance to NDP risk. Recent events dem-NUREG/CR-6264 V01: VALIDITY LIMITS IN J-RESISTANCE onstrate that human performance continues to be a dominant CURVE DETERMINATION.An Assessrnent Of The J(M) Param-source of nsk. In light of these observations, the current hmita-eter. SHlH,C.F.; LIU.X.H. Brown Univ., Providence, RI.

• Battelle tions of existing HRA approaches become apparent when the Memorial institute, Columbus Laboratories. February 1995.

role of humans is examined exphcitly in the context of real NPP 35pp. 9503140423. BMI-2181. 83070:282.

events. The development of new or improved HRA methodolo.

Significant advances in elastic-plastic fracture became possi-gies to more reahstically represent human performance is rec-ble witl, the introduction of Rice's path independent J-integraf ognized by the Nuclear Regulatory Commission (NRC) as a which has two physical meanings. First, the J-integral is equiva-necessary means to increase the utikty of PRAs. To accomphsh lent to the energy release rate associated with a virtual crack this objective, an Improved HRA Project, sponsored by the advance. Secondly, J can be regarded as the strength of the NRC's Office of Nuclear Regulatory Research (RES), was initi-stress and strain singulanty near a stationary crack tip. As a ated in late February,1992, at Brookhaven National Laboratory result of several expenmental studies, the J-integra! is generally (BNL) to develop an improved method for HRA that more realis-accepted as a vahd parameter to characterize a material's re-tacally assesses the human contnbution to plant risk and can be sistance to the onset of crack growth under large-scale yielding.

fully integrated with PRA. This report desenbes the research ef.

Driven by simplicity and the practical benefits that could be de-forts including the development of a multidisciphnary HRA rived from a geometry and size-independent material resistance framework, the characterization and representation of errors of I

curve for large amounts of crack growth, J(M), a modified J pa-commission, and an approach for addressing human dependen-

Main Citations and Abstracts 33 cies. The implications of the research and necessary require-been characterized. Cast stainless steel materials were ob-rnents for further development also are discussed.

tained from four cold-leg check valves, three hot-leg main shut.

NUREG/CR-4266: ANALYSIS OF BORON DILUTION IN A FOUR-off valves, and two pump volutes. The actual time-et-tempera-LOOP PWR. SUN.J.G.; SHA,W.T. Argonne Nabonal Laboratory.

ture for the materials was N13 y at N281 degrees C for the March 1995. 80pp. 9504180468. ANL-94/35. 83569:235.

hot-leg components and N264 degrees C for the cold-leg com-Thermal mixing and boron dilution in a pressunzed water re-ponents. Baseline mechanical propertees for as-cast material actor were analyzed with COMMIX codes. The reactor system were determined from tests on either recovery-annealed maten-was the four loop Zion reactor. Two boron dilution scenarios al or matwial kom the cooler region of the component. The were analyzed in the first scenano, the plant is in cold shut-Shippingport materials show modest decreases in fracture down and the reactor coolant system has just been filled after toughness and Charpy impact properties and a small increase maintenance on the steam generators. To flush the air out of in tensile strength because of relatively low service tempera-the steam gs,:erator tubes, a reactor coolant pump (RCP) is tures and fwnte content of the std The procedure and core started, with the water in the pump suction hne devoid of boron lations developed at Argonne National Laboratory for estimating and at the same temperature as the coolant in the system. In mechanical properties of cast stainless steels predict accurate the second scenario, the plant is at hot standby and the reactor or slightly lower values for Charpy-smpact energy, tensile flow coolant system has been heated to operating temperature after stress, fracture toughness J-R curve, and J(IC) of the materials.

a long outage. It is assumed that an RCP is started, with the The kinetics of thermal embrittlement and degree of embnttle-ment at saturabon were established from materials that were pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently aged further in the laboratory. The results were consistent with through the reactor core. The subsequent transient thermal the estimates. The correlations successfully predicted the me-mixing and boron dilution that would occur in the reactor system chanical properties of the Ringhals 2 reactor hot-and cross-is simulated for these two scenancs. The reactivity insertion over-leg elbows (CF-8M) after service of N15 y and the KR8 rate and the total reactivity are evaluated and a sensrtivity study reactor pump cover plate (CF-8) after N8 y of sennce is performed to assess the accuracy of the numencal modeling NUREG/CR-6276: A COMPILATION OF CURRENT REGULA-of the geometry of the reactor coolant system.

TIONS, STANDARDS, AND GUIDELINES IN REMOTE AFTER-NUREG/CR-6273: BIAXIAL LOADING ZFFECTS ON FRACTURE LOADING BRACHYTHERAPY. TORTORELLI,J.P.; SIMION G.P.;

TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL KOZLOWSKI.S.D. EG&G Idaho, Inc. February 1995. 94pp.

MCAFEE,W.J.; BASS,B.R.; BRYSON,J.W.; et al. Oak Ridge Na.

9503270298. EGG-2746. 83268:095.

tional Laboratory. March 1995. 76pp. 9503270300. ORNL/TM.

Over a dozen government and professional organizations in 12866. 83268:191.

the United States and Europe have issued regulations and guid-The preliminary phases of a program to develop and evaluate ance concerning quality management in the practice of remote fracture methodologies for the assessment of crack-tip con.

afterloading brachytherapy. Information from the publications of straint effects on fracture toughness of reactor pressure vessel these organizations was collected and collated for this report.

(RPV) steels has been completed by the Heavy-Section Steel This report provides the brachytherapy licensee access to a Technology (HSST) Program. The primary objectives of this broad field of quality management information in a single, topi-effort were to analytically and experimentally investigate the cally organized document.

effect of biaxial loading on fracture toughness, to quantify this NUREG/CR-6277 V01: HUMAN FACTORS EVALUATION OF effect through use of existing stress-based, dual-parameter, TELETHERAPY. Identification Of Problems And Alternative Ap-fracture-toughness correlations, or to propose and venfy alter-proaches. HENRIKSEN,K.; KAYE,R.D.; JONES R.; et al. Hughes nate correlations. A cruciform beam specimen with a two-di-Training, Inc. July 1995. 92pp. 9508010236, 84908:0Q9.

mensional shallow, through-thickness flaw and a special loading A series of human factors evaluations were undertaken to fixture was designed and fabricated. Tests were performed better understand the contributing factors to human error in the using biaxial loading ratios of 0:1 (uniaxial),0.6.1, and 1:1 (equi-teletherapy environment. Teletherapy is a multidisciplinary meth-biaxial). Cntical fracture toughness values were calculated for odology for treating cancerous tissue through selective expo-each test. Biaxial loading of 0.6:1 resulted in a reduction in the sure to an external beam of ionizing radiation. A team of human lower bound fracture toughness of =12% as compared to the factors specialists, assisted by a panel of radiation oncologists, uniaxial tests. The biaxial loading of 1:1 yielded two subsets of medical physicists, and radiation therapists, conducted site visits toughness values; one agreed well with the uniaxial data while to radiation oncology departments at community hospitals, uni-one was reduced by = 43% when compared to the uniaxial versity centers, and free-standing clinics. A function and task data. The results were evaluated using the J-Q theory and the analysis was initially performed to guide subsequent evaluations Dodds-Anderson (D-A) micromechanical scaling model. The D-A in the areas of system-user interfaces, procedures, training and model predicted no biaxial effect while the J-O method gave in-qualifications, and organizational policies and practices. The conclusive results. When applied to the 1:1 biaxial data, these present work focuses solely on training and qualifications of constraint methodologies failed to predict the observed reduc-personnel (e.g., training received before and during employ-tion in fracture-toughness obtained in one experiment. A strain-ment), and the potential impact of organizational factors on the based constraint methodology that considers the relationship performance of teletherapy. Organizational factors include such between applied biaxial load, the plastic zone size at the flaw topics as adequacy of staffing, performance evaluations, com-tip, and fracture-toughness was formulated and applied suc-monfy occurring errors, implementation of quality assurance pro-cessfulty to the data. Evaluation of this dual-parameter strain-grams, and organizational climate.

based model led to the conclusion that it has the capability of representing fracture behavior of RPV steels in the transition NUREG/CR-6277 V02: HUMAN FACTORS EVALUATION OF region, including the effects of out-of-plane loading on fracture-TELETHERAPY. Function And Task Analysis. KAYE,R.D.;

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

HENRIKSEN,K.; JONES,R.; et al. Hughes Training, Inc. July 1995. 250pp. 9508020293. 84910:067.

NUREG/CR-6275: MECHANICAL PROPERTIES OF THERMALLY See NUREG/CR-6277,V01 abstract.

AGED CAST STAINLESS STEELS FROM SHIPPINGPORT RE-ACTOR COMPONENTS. CHOPRA,0.K.; SHACK W.J. Argonne NUREG/CR-6277 V03: HUMAN FACTORS EVALUATION OF National Laboratory. Apnl 1995.170pp. 9506010453. ANL-94/

TELETHERAPY. Human-System Interfaces And Procedures.

37, 84121:001.

KAYE,R.D.; HENRIKSEN,K.; JONES R.; et al. Hughes Training.

Thermal embnttlement of static-cast CF-8 stainless steel Inc. July 1995. 70pp. 9508020295. 84915:001.

components from the decommissioned Shippingport reactor has See NUREG/CR-6277,V01 abstract.

I 34 Main Citations and Abstracts NUREG/CR-6277 V04: HUMAN FACTORS EVALUATION OF sion of a pressunzed water reactor severe accident. Parameters TELETHERAPY. Training And Organizational Analysis.

affecting natura! circulation in the reactor vessel and hot legs HENRIKSEN.K.; KAYE.R.D.; JONES R ; et al. Hughes Training, were identified and ranked based on their perceived impor-Inc. July 1995. 71pp. 9508020296. 84908.191.

tance. Reviews of the scaling of the 1/7-scale experiments per-See NUREG/CR-6277,V01 abstract.

formed by Westinghouse were undertaken. RELAPS/ MOD 3 cal-NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF cu!ations of two of the expenments showed generally good TELETHERAPY. Literature Review. HENRIK5EN.K.; KAYE R D.;

agreement between the calcu'ated and observed behavior.

JONES,R.; et al. Hughes Training. Inc. July 1995. 113pp.

Analyses of hydrogen behavior in the reactor vessel showed 9508020306. 84915.071.

that hydrogen stratification is not likely to occur, and that an ini-See NUREG/CR-6277,V01 abstract.

tially stratified layer of hydrogen would quickly mix with a recir-NUREG/CR-6283: FIELD SITE INVESTIGATION. EFFECT OF culating steam flow. An analysis of the upper plenum behavior MINE SEISMICITY ON GROUNDWATER HYDROLOGY.

In the Three Mile Island, Unit 2 reactor concluded that vapor OFOEGBU,G I.; HSIUNG.S-M.; CHOWDHURY,A H.; et al.

temperatures could have been significantly higher than the tem-Center for Nuclear Waste Regulatory Analyses. Apnl 1995.

Peratures seen by the control rod dnve lead screws, supporting 103pp. 9505040377. CNWRA 94-017. 83783158.

the premise that a strong natural circulation flow was likely The results of a field investigation on the groundwater hydro-present dunng the accident. SCDAP/RELAP5 calculations of a i

logic effect of mining-induced earthquakes are presented. The commercial pressunzed water reactor severe accident without e day Mine Idaho, the operator actions showed that the natural circulation flows en-investigation was conducted at the Lucky n

groundwater pressure in three fracture zones was monitored hance the likelihood of ex-vessel piping failures long before fail-over a 24 mo period. The magnitude, source location, and asso-ure of the reactor vessellower head.

ciated ground motions of mining induced seismic events were also monitored. Several seismic events of magnitude 1.0 or NUREG/CR-6287: MANAGEMENT CONCEPTS AND SAFETY larger were recorded, many of which caused a change in the APPLICATIONS FOR NUCLEAR FUEL FACILITIES. EISNER,H.;

groundwater pressure. The magnitude of groundwater-pressure SCOTTI,R.S. George Washington Univ., Washington, DC.

change vaned with the seismic-event magnitude and source dis-DELICATE W.S. The KEVRIC Company, Inc. May 1995.131pp.

tance. The data was examined using regression analysis. The 9506020467.84147:137.

statistical models obtained predicted the effects of small-magni-This report presents an overview of effectiveness of manage-tude seismic events more satisfactonly than those of larger ment control of safety. It reviews several modern management ones. The observed change 'n Oroundwater pressure due t control theones as well as the general functions of mana9e-seismic events of magnitude 3.0 or more were larger than those merit and relates them to safety issues at the corporate and at predicted using the statistical model. Based on these results, it is suggested that the effect of earthquakes on groundwater flow the process safety management (PSM) program level. Following may be better understood through mechanistic modeling The these discussions, a formal and structured technique for as-mechanical processes and matenal behavior that would need to sessing management of the safety function is suggested. Seven be incorporated in such a model are examined. They include a modern management control theories are summanzed, including description of the effect of stress change on the permeability business process reengineenng, the leaming organization, capa-and water storage capacity of a fractured rock mass; transient bihty matunty. total quahty management, quality assurance and fluid flow; and the generation and transmission of seismic control, rehabihty centered maintenance, and industnal process waves through the rock mass.

safety. Each of these thoones is examined for its pnncipal char-actenstics and imphcations for safety management. The five NUREG/CR-6284: CRITICALITY SAFETY CRITERIA FOR LI-CENSE REVIEW OF LOW-LEVEL WASTE FACILITIES.

general management functions of planning, organiz;ng, direct-HOPPER,C M ; ODEGAARDEN R.H ; PARKS,C.V.; et al. Oak

'"9 50"'tonng, and integrating, which together provide control Ridge National Laboratory. March 1995. 45pp. 9504120093, over all company operations, are discussed. Under the broad ORNL/TM 12845. 83474.165.

categones of Safety Culture. Leadership and Commitment, and This report provides recommended safety entena for NRC li.

Operating Excellence, liey corporate safety elements and their censed burial facilities. These entena have been developed with subelements are examined. The three categories under which accepted and consistent nuclear enticahty safety evaluation PSM program-level safety issues are desenbed are Technology, techniques Additionally, this report provides the bases fnr the Personnel, and Facihties.

recommended safety enteria by documenting the evaluation methods and assumptions, and by reporting the results of all NUREG/CR-6291 V01: NUCLEAR PLANT ANALYZER. Installation single-package and array calculations. These enteria were de-Manual. SNIDER,D.M.; WAGNER.K.L.; GRUSH,W.H.; et al.

veloped with care to assure consistency with data and practices Idaho National Engineenng Laboratory. January 1995. 28pp.

provided in current standards on nuclear enticality safety as well 9502150283. INEL-94/0123. 82730:001.

as conformity of the entena to apphcable NRC regulations. The This report contains the installation instructions for the Nucle-recommended safety entena are expressed in terms of surface-ar Plant Analyzer (NPA) System. The NPA System consists of density spacing cr teria, thereby greatly simphfying the cpplica-the Computer Visual System (CVS) program, the NPA hbranes, tion of hcense conditions for nuclear enticahty safety control.

the associated utshty programs. The NPA was developed at the This approach was used by an NRC licensee at the Barnwell Idaho National Engineenng Laboratory under the sponsorship of j

waste bunal facihty by limiting the specific controls to the fewest the U S. Nuclear Regulatory Commission to provide a highly l

number of parameters consistent with good nuclear safety prac-flexible graphical user interface for displaying the results of l

tice. The use of a surface-density entena can ehminate the need these analysis codes The NPA also provides the user with a

[

for numerous license amendments for vanations in package convenient means of interactively controlkng the host program contents and specifications.

through user defined pop-up menus. The NPA was designed to NUREG/CR-6285: SEVERE ACCIDENT NATURAL CIRCULATION serve primanly as an analysis tool. After a brief introduction to STUDIES AT THE INEL. BAYLESS,P.D.; BROWNSON D.A.;

the Computer Visual System and the NPA. an analyst can DOBBE.C.A ; et al. Idaho National Engineenng Laboratory. Feb.

quickly create a simple picture or set of pictures to aide in the ruary 1995. 200pp. 9M3150171. INEL-94/0016. 83108:014.

study of a particular phenomenon. These pictures can range Severe accident natural circulation flows have been investi-from simple collections of square boxes and straight knes to gated at the Idaho National Engineenng Laboratory to better un-complex representations of emergency response information derstand these flows and their potential impacts on the progres-displays.

Main Citations and Abstracts 35 NUREG/CR-6291 V02: NUCLEAR PLANT ANALYZER. Analyzer tions about the environment in which these systems will be uti-Reference Manual. SNIDER D.M.; WAGNER,K.L; GRUSH,W.H.;

lized The final chapter of Volume i deals with a framework for et al. Idaho National Engineenng Laboratory. January 1995.

standards in this field. Volume ll contains appendices dealing 71pp. 9502150287, INEL 94/0123. 82730:030.

with specific methodologies for system classification, for de-The Nuclear Plant Analyzer (NPA) system provides both a pendability evaluation, and for two software tools that can auto-highly flexible graphical user interface for displaying simulation mate otherwise very labor intensive venfication and validation data and, where applicable, a convenient means of interactively activities.

controlling the host program through user defined pop-up NUREG/CR-6293 V02: VERIFICATIONN AND VALIDATION menus. The NPA system was developed at the Idaho National Engineenng Laboratory under the sponsorship of the U.S. Nu-GUIDELINES FOR HIGH INTEGRITY SYSTEMS. Appendices A-clear Regulatory Commission (NRC). The Computer Visual D. HECHT H.; HECHT,M.; DINSMORE,G.; et al. SoHaR, Inc.

System and the Analyzer are the primary components of the March 1935. 54pp. 9504100137. 83419.304.

NPA system. This report contains the reference manual for the See NUREG/CR-6293,V01 abstract.

Analyzer. It desenbes both the NPA libranes that constitute the NUREG/CR-6297: FRACTURE EVALUATIONS OF FUSION LINE Analyzer and a set of auxiliary programs used in conjunction CRACKS IN NUCLEAR PIPE B! METALLIC WELDS. SCOTT P-with the Analyzer FRANCINI.R.; RAHMAN.S.; et al. Battelle Memonal Institute NUREG/CR-6291 V03: NUCLEAR PLANT ANALYZER. Computer Columbus Laboratones. Apnl 1995.111pp. 9505030455. BMI-Visual System Reference Manual. SNIDER.D.M.; WAGNER.K.L.;

2182. 83744:194.

GRUSH,W.H.; et al. Idaho National Engineenng Laboratory. Jan.

In both OWRs and PWRS there are many locations where uary 1995.128pp. 9502150297. INEL-94/0123. 82730.101.

carbon steel pipe or components are joined to stainless steel The Cornputer Visual System CVS) Reference Manual da.

pipe or components with a bimetallic weld. The objective of the senbes that part of the Nuclea-unt Analyzer (NPA) system research described in this report was to assess the accuracy of used to create pictures (masksJ Ihis manual is intended to current fracture analyses for the case of a crack along a carbon guide a user in creating, editing, and animating rnasks for use in steel to austenitic weld fusion line. To achieve the program ob-the NPA. The NPA was developed at the Idaho National Engi.

jective, matenal property data and data from a large-diameter neenng Laboratory under the sponsorship of the U.S Nuclear pipe fracture expenment were developed to assess current ana-Regulatory Commission to provide a highly flexible graphical lytical methods. The bimetallic welds evaluated in this program

.were bimetallic welds obtained from a cancelled Combustion user interface for displaying the results of these analysis codes.

The NPA also provides the user with a convenient means of in-Engineenng plant. The welds joined sections of the carbon steel teractively controlling the host program through user defined cold-leg piping system to stainless steel safe ends that were to pop-up menus. The NPA was designed to serve pnmanly as an be welded to stainless steel pump housings. The major conclu-analysis tool. After r. brief introduction to the Computer Visual 8'on drawn as a result of these efforts was that the fracture be-System and the P.M, an analyst can ouickly create a simple havior of the bimetallic weld evaluated in this program could be picture or set of g mures to aide in the study of a particular phe, evaluated with reasonable accuracy using the strength and nomenon. These pictures can range from simple collections of toughness properties of the carbon steel pipe matenal in con-square boxes and straight lines to complex representctions of junction with conventional elastic-plastic fracture mechanics or emergency response information displays.

limit-load analyses. This may not be generally true for all bime-NUREG/CR-6291 V04:

NUCLEAR PLANT ANALYZER. Programmer's Manual.

SNIDER D.M.;

NUREG/CR-6298: FRACTURE BEHAVIOR OF SHORT CIRCUM-WAGNER.K.L.; GRUSH,W.H.; et al. Idaho National Engineenng FERENTIALLY SURFACE-CRACKED PIPE.

Laboratory. January 1995. 202pp. 9502150309. INEL 94//0123.

KRISHNASWAMY,P.; SCOTT,P.; MOHAN.R.; et al. Battelle Me-82730.229.

morial Institute, Columbus Laboratories. November 1995.

The Nuclear Plant Analyzer (NPA) system provides both a 328pp. 9512040318. BMI-2183. 86383:001.

highly flexible graphical user interface for displaying simulation This topical report summarizes the work performed for the data and, where applicable, a convenient means of interactively Nuclear Regulatory Commission's (NRC) research program enti-controlling the host program through user-defined pop-up tied "Short Cracks in Piping and Piping Welds" that specifically menus. The NPA system was developed at the Idaho National focuses on pipes with short, circumferential surface cracks. The Engineenng Laboratory under the sponsorship of the U.S. Nu-following details are provided in this report- (i) material property clear Regulatory Commission (NRC) This manual is intended to determinations, (ii) pipe fracture experiments, (iii) development, serve as a programmers' guide for the NPA system. As such, it modification and validation of fracture analysis methods, and (iv) includes technical details regarding the design and implementa.

impact of this work on the ASME Section XI Flaw Evaluation tion of the Computer Visuals Systems (CVS) program, the Ana-Procedures. The matenal properties developed and used in the lyzer, data files used by CVS and the Analyzer, and a senes of analysis of the expenments are included in this report and have auxiliary programs that provide important services to NPA users.

been implemented into ths NRC's PIFRAC database. Six full-Scale Pipe expenments were conducted during this program.

NUREG/CR-6293 V01: VERIFICATION AND VAllDATION

• • U*

(i) limit load approaches, (ii) design entena,"and (iii) elastic-plas-GUIDELINES FOR HIGH INTEGRITY SYSTEMS. Main Report.

HECHT,H.; HECHT,M.; DINSMORE,G.; et al. SoHaR, Inc. March 8

1995. 206pp. 9504100125. 83419:107 paring the analytical predictions with expenmental data. The re-High integnty systems include all protective (safety and miti-suits, using 44 pipe experiments from this and other programs, gation) systems for nuclear power plants, and also systems for showed that the SC.TNP1 and DPZP analyses were the most which comparable reliability requirements exist in other fields, accurate in predicting maximum load. New Z-factors were de-such as in the process industnes, in air traffic control, and in veloped using these methods. These are being considered for patient monitonng and medical systems. Verification aims at de-updating the ASME Section XI criteria.

termining that each stage in the software development com-pletely and correctly implements requirements that were estab-NUREG/CR-6299: EFFECTS OF TOUGHNESS ANISOTROPY lashed in preceding phase, while validation determines that the AND COMBINED TENSION, TORSION, AND BENDING LOADS overall performance of a computer system completely and cor-ON FRACTURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

rectly meets system requirements. Volume 1 of the report re-MOHAN,R.; MARSCHALL,C.W.; KRISHNASWAMY,P.; et al. Bat-views existing classifications for high integnty systems and for telle Memorial Institute, Columbus Laboratories. April 1995.

the venfication and validation procedures, based on assump-122pp. 9506020484. 84147:268.

i 36 Main Citations and Abstracts This topical report summarizes the work on angled crack vective-dispersive, mu!tispecies, solute transport. BLT-EC ac-growth and combined loading effects performed within the Nu-counts for retardation directly by modeling the chemical proc-clear Regulatory Commissions's research program, entitled esses of complexation, sorption, dissolution-precipitation, ion-l "Short Cracks in Piping and Piping Welds". The major impetus exchange, and oxidation-reduction reactions. Herein we: (1) de-i for this work stemmed frorn the observation that initial circum

• scnbe in detail physical and chemical processes that control the i

l ferential cracks in carbon steel pipes exhibsted angular crack release and migration of radionuclides from shallow land LLW growth. This failure mode was little understood, and the effect disposal facilities; (2) formulate the mathematical models that of angled crack growth from an initially circumferential crack represent these processes; (3) outline how these models are in-raised questions of how pipes under combined loading with tor.

corporated and implemented in BLT-EC; and (4) demonstrate sional stresses would tehave. There were three major conclu-application of BLT-EC on sample problems.

sions from this work. The first was that virtually all femtsc nucle-ar pipes will have toughness anisotropy. The second was that NUREG/CR-6307:

SUMMARY

OF COMMENTS RECEIVED AT the ratio of the normalzed crack dnving force (as a function of WORKSHOP ON USE OF A SITE SPECIFIC ADVISORY angle) to the normalized toughness (also as a function of the angle of crack growth) showed that there was an equal likeli-BOARD (SSAB) TO FACILITATE PUBLIC PARTICIPATION IN hood of cracks growing at any angle between 25 and 65 de-DECOMMISSIONING CASES. CAPLIN,J.; PAGE,G.; SMITH.Dc' grees. This agreed with the scatter of crack growth angles ob.

et al. Advanced Systems Technology, Inc. June 1995.115pp.

served in pipe tests. Third, for combined loads with torsional 9507120259. 84634:190.

i stresses, an effective moment allows pure bending analyses to The Nuclear Regulatory Commission (NRC) is conducting an

(

be used up to crack initiation. Crack opening area under com-enhanced participatory rulemaking to establish radiological ente-bened loads could also be determined in this manner.

ria for the decommissioning of NRC-licensed facilities. As part NUREG/CR-6300: REFINEMENT AND EVALUATION OF CRACK.

of this rulemaking, on August 20, 1994 the NRC published a l

OPENING-AREA ANALYSES FOR CIRCUMFERENTIAL proposed rule for public comment. Paragraph 20.1406(b) of the i

THROUGH-WALL CRACKS IN PIPES..

RAHMAN,S.;

proposed rule would require that the licensee convene a Site i

BRUST,F.W.; GHADIAll.N.; et al. Battel!e Memonal institute, Specific Advisory Board (SSAB) if the licensee proposed re.

L Columbus Laboratones. Apnl 1995. 227pp. 9505030464. BMI.

lease of the site for restricted use after decommissioning. To l

2184. 83745.001.

encourage comment the NRC held a workshop on the subject i

Leak-before-break (LBE) analyses for circumferentially of SSABs on December 6, 7, and 8,1994. This report summa-cracked pipes are currently being conducted in tha nuclear in-nzes the 567 comments categonzed from the transcnpt of the dustry to justify elimiration of pipe whip restraints and jet im-workshop. The commenters at the workshop generally support-pingement shields which are present because of the expected ed public participation in decommissioning cases. Many partici-dynamic effects from pipe rupture. The application of the LSB pants favored promulgating requirements in the NRC's rules.

methodology frequently requires calculation of leak rates. Th0 Some industry participants favored relying on voluntary ex-leak rates depend on the crack-opening area of the through-changes between the public and the licensees. Many partici-wall crack in the pipe. In addition to LBB analyses which pants indicated that a SSAB or something functionally equiva-assume a hypothetical flaw size, there is also interest in the in-lent is needed in controversial decommissioning cases, but that tegnty of actual leaking cracks corresponding to current leakage some lesser undertaking can achieve meaningful public partici-detection requirements in NRC Regulatory Guide 1.45, or for as-sessing temporary repair of Class 2 and 3 pipes that have leaks pation in other cases. No analysis or response to the comments as are being evaluated in ASME Section XI. This study was re-s included in this report.

quested by the NRC to review, evaluate, and refine current ana-NUREG/CR-6308: AN OVERVIEW OF INSTABILITY AND FIN-lytical models for crack-opening-area analyses of pipes with cir" GERING DURING IMMISCIBLE FLUID FLOW IN POROUS AND cumferential through-wall cracks. Twenty-five pipe experiments were analyzed to determine the accuracy of the predictive FRACTURED MEDIA. CHEN G.; TANIGUCHi,M.; NEUMAN S.P.

models. Several practical aspects of crack-opening such as; Anzona, Univ. of, Tucson, AZ. April 1995.137pp. 9507060331.

l crack-face pressure, off< enter cracks, restraint of pressure-in-84532.043.

j duced bending, cracks in thickness transition regions, weld re.

Wetting front instability is an important phenomenon affecting l

sidual stresses, crack-morphology models, and thermal-hydrau.

fluid flow and contaminant transport in unsaturated soils and lic analysis, were also investigated.

rocks. It causes the development of fingers which travel faster than would a uniform front and thus bypass much of tne NUREG/CR-6305: BLT-EC (BREACH, LEACH, TRANSPORT me ater sa ahon aM solute concenuahon M such 6 AND EQUILIBRIUM CHEMISTRY), A FINITE ELEMENT MODEl' FOR ASSESSING THE RELEASE OF RADIONUCUDES FROM gers tend to be higher than in the surrounding medium. During LOW LEVEL WASTE DISPOSAL UNITS. Background, Theory' kabon, fingering may cause unexpectedly rapid arrival of And Model Desenption. MACKINNON.R.J.: SULLIVAN,T.M water and solute at the water table. This notwithstanding, most Brookhaven National Laboratory. SIMONSON.S.A.; et al. Mas-models of subsurface flow and transport ignore instability and sachusetts institute of Technology, Cambridge, MA. August fingering. In this report, we survey the literature to assess the 1995.155pp.9509200030. BNL.NUREG 52446. 85540.001-extent to whicn this may or may not be justified. Our overview Performance assessment models typically account for sorp.

covers expenments, theoretical studies, and computer simula-tion and dissolution-precipitation by using an empincal distribu.

tions of instability and fingenng dunng immiscible two-phase lion coefficient (K(d)) that combines the effects of all chemical flow and transport, with emphasis on infiltration into soils and reactions between solid and aqueous phases. There is an in-fractured rocks. Our desenption of instability in an ideal fracture creasing awareness that performance assessments based (Hele-Shaw cell) includes an extension of existing theory to solely on empincally based K(d) models may be incomplete, fractures and interfaces having arbitrary orientations in space.

particularly for applications involving radionuclides having sorp-Our discussion of instability in porous media includes a slight tion and solubility properties that are sensitive to variations in but important correction of existing theory for the case of an in-the in situ chemical environment. To accommodate such vari-clined interface. We conclude by outlining some potential direc-niions and to assess impact on radionuclide mobility, one must tions for future research. Among these, we single out the effect model radionuclide release, transport, and chemical processes of soil and rock heterogeneities on instability and preferential in a coupled fashion. This modeling was incorporated into the flow as menting special attention in the context of nuclear twodimensional, finite-element, computer code BLT-EC which waste storage in unsaturated media.

can predict container degradation, waste-form teaching, and ad-

Main Citations and Abstracts 37 NUREG/CR-6310: AN ANALYSIS OF POTASSIUM IODIDE (Kl) also observed to be less subject to electromagnetic interference PROPHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NUCLEAR ACCIDENT, BEHLING,H.; BEHLING K. S.

(EMI) and more capable at very low frequencies.

Cohen & Associates, Inc. AMARASOORlYA H.; et al. SCIEN-NUREG/CR-6313 V02: ROBUST, ACCURATE, AND NON-CON-TECH, Inc. February 1995. 200pp. 9503150166. 83107:157 TACTING VIBRATION MEASUREMENT A generic difficulty countered in cost-benefit analyses is the quantification of major elements that define the costs and the SYSTEMS. Supplemental Appendices Presenting Companson Measurements Of The Robust Laser interferometer And Typical benefits in commensurate units. In this study, the costs of Accelerometer Systems. GOODENOW,T.C.; SHIPMAN,R.L.;

making Kl available for public use, and the avoidance of thyroi-HOLLAND.H.M. Epoch Engineering, Inc. June 1995. 200pp.

dal health effects predicted to be realized from the availability of 9507200236. 84750:092.

that Kl (i.e., the benefits), are defined in the commensurate units of dollars.

See NUREG/CR.6313,V01 abstract.

NUREG/CR-6311: EVALUATING PREDICTION UNCERTAINTY.

NUREG/CR-6315: CANDU REACTORS, THEIR REGULATION IN MCKAY,M.D. Los Alamos National Laboratory. March 1995.

CANADA, AND THE IDENTIFICATION OF RELEVANT NRC 69pp. 9503270307 LA 12915-MS. 83270:274.

SAFETY ISSUES. CHARAK,l.; KIER.P.H. Argonne National Lab-The probability distnbution of a model prediction is presented oratory. April 1995. 48pp. 9505180578. ANL-95/5. 83980:034.

as a proper basis for evaluating the uncertainty in a model pre-Atomic Energy of Canada, Limited (AECL) and its subsidiary diction that anses from uncertainty in input values. Determina-in the Uni:ed States, are considenng submitting the CANDU 3 tion of important model inputs and subsets of inputs is made design for standard design certification under 10 CFR Part 52.

through companson of the prediction distnbution with condition-al prediction probability distnbutions. Replicated Latin hypercube CANDU reactors are pressunzed heavy water power reactors.

sampling and vanance ratios are used in estimation of the distri-They have some substantially different safety responses and butions and in construction of importance indicators. The as-safety systems than the LWRs that the commercial power reac-sumption of a linear relation between model output and inputs is tor licensing regulations of the U.S. Nuclear Regulatory Com-not necessary for the indicators to be effective. A sequential mission (NRC) have been developed to deal with. In this report, methodology which includes an independent validation step is the authors discuss the basic design characteristics of CANDU applied in two analysis applications to select subsets of input reactors, specifically of the CANDU 3 where possible, and some vanables which are the dominant causes of uncertainty in the safety-related consequences of these characteristics. The au-model predictions. Comparison with results from methods which thors also discuss the Canadian regulatory provisions, and the assume lineanty shows how those methods may fail. Finally, CANDU safety systems that have evolved to satisfy the Canadi-ggestions for treating structural uncertainty for submodels are an regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and NUREG/CR-6312: ASSESSMENT OF FIBER OPTIC PRESSURE 100, with issues for CANDU 3 reactor designs. In all, eleven SENSORS. HASHEMIAN.H.M.; BLACK,C.L.; FARMER.J.P. Anal-such regulatory issues are identified. They are: (1) the ATWS rule (Section 50.62); (2) station blackout (Section 50.63); (3) ysis & Measurement Services Corp, April 1995. 132pp 9506010456. 84121:170.

conformance with Standard Review Plan (SRP); (4) appropnate-The principle of operation of fiber optic pressure sensors and ness of the source term (Section 50.34(f) & Section 100.11);(5) the potentia! of inese sensors for use in nuclear power plants applicability of reactor coolant pressure boundary (RCPB) re-are desenbed in this report. Alec included is a review of current quirements (Section 50.55a, etc); (6) ECCS acceptance enteria research on f ber optic seriry tuchnologies, a companson of (Section 50.46(b); (7) combustible gas control (Section 50 44, i

fiber optic pressure sensors with conventional pressure sensors, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic a discussion on advantages and disadvantages of fiber optic design (Part 100); (10) environmental impacts of the fuel cycle pressure sensors, a review of failure modes of these sensors, (Section 51.51); and (11) (standards Section 50.55a)'

and results of a survey of fiber optic sensor manufacturers.

NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION NUREG/CR-6313 V01: ROBUST, ACCURATE, AND NON-CON-AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND TACTING VIBRATION MEASUREMENT SYSTEM. Summary of CONVENTIONAL SOFTWARE. MILLER,L.A.; HAYES.J.E.;

Companson Measurements Of The Robust Laser interferometer MIRSKY,S.M. Science Applications international Corp. (formerly And Typical Accelerometer Systems. GOODENOW,T.C.;

SHIPMAN.R L.; HOLLAND,H.M. Epoch Engineenng, Inc. June Science Applications, Inc.). March 1995.178pp. 9504180393.

1995. 74pp. 9507200222. 84749'285.

SAIC 95/1028. 83566:198.

Epoch Engineenng, incorporated (EEI) has completed a This report provides the project summary of the results of the series of vibration measurements companng their newly-devel-Expert System Venfication and Validation (V&V) activity that oped Robust Laser Interferometer (RLl) with accelerometer in-was jointly funded by the U.S. Nuclear Regulatory Commission strumentation systems. eel has successfully demonstrated, on and the diectric Power Research Institute to develop guidelines several pieces of commonplace machinery, that non-contact for the V&V of expert and other systems. This is the first line-of-sight measurements are practical and yield results equa' volume of an eight-volume report. The project began with a l

to or, in some cases, better than customary field implementa-survey of conventional V&V methods that covers 153 different tions of accelerometers. The demonstration included analysis techniques. Quantitative cost-benefit and an effectiveness and companson of such phenomena as nonlineanty, transverse measures were developed to permit compansons among all the sensitivity, harmonics, and signal-to-noise ratio. Fast Founer methods for three levels of stnngency of VaV: low, medium, Transformations were performed on the accelerometer and the and high (Classes 3 to 1, respectively). A survey was conducted laser system outputs to provide a companson basis. The RLI conceming V&V practices in use for export systems, finding that was demonstrated, within the limits of the task, to be a viable, they were not common, but that there was considerable activity line-of-sight, non-contract alternative to accelerometer systems.

in developing methods for knowledge bases. Selected V8V Several different kinds of machinery were instrumented and methods were applied to two existing expert systems used in compared, including a small pump, a gear driven cement mixer, nuclear power applications. Other V&V methods were investi.

A rotor kit, and two smati fans. Known machinery vibration gated in an empirical expeiiment to assess their practical utility, sources were venfied and RLI system output file formats were A method for generating validation scenarios was developed. Fi-venhed to be compatible with commercial computer programs nally, a set of guidelines recommending specific V&V methods used for vibration monitoring and trend analysis. The RLI was for 16 different system-development situations was developed.

38 Main Citations and Abstracts adREG/CR 6316 V02: GUIDELINES FOR THE VERIFICATION Idaho, and SAIC. The major conclusion was that the use of AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND tools in static knowledge base certification results in significant CONVENTIONAL SOFTWARE. Survey And Assessment Of Con-improvement in detecting all types of defects, avoiding false ventional Software Venfication And Vahdation Methods.

alarms, and completing the effort in less time. The simulated MILLER.LA.; GROUNDWATER.E.H, HAYES.J E.; et al. Science knowledge-checking tool, based on supplemental information, Apphcations international Corp. (formerly Science Apphcations, was the most effectue of the tools.

Inc.).

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

83567:017.

NUREG/CR-6316 V05: GUIDELINES FOR THE VERIFICATION By means of literature survey, a comprehensive set of meth.

AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND ods was identified for the verification and validation of conven.

CONVENTIONAL SOFTWARE Rationale And Description Of tional software. The 153 methods so identified were classified V8V Guideline Packages And Procedures. MILLER.L.A.;

according to their appropnateness of vanous phases of a devel-HAYES.J E.; MIRSKY,S.M. Science Applications International opment hfecycle - requirements, design, and implementation; Corp. (formerly Science Applications, Inc.). March 1995.103pp.

the last category was subdivided into two, static testing and dy.

9504180441. SAIC-95/1028. 83568:131.

namic testing methods. The methods were then characterized in This report is the fifth volume of a series desenbing the re-terms of eight rating factors, four conceming ease-of-use of the suits of the Expert System Venfication and Validation (V&V) methods and four concerning the methods; power to detect de-project jointly funded by the U.S.

Nuclear fects. Based on those ar,J an Effectiveness Metnc. The Effec-Regulatory. Commission and the Electnc Power Research Insti-tiveness Metric was further refined to provide three different es-tute, to formulate guidehnes for the V&V of expert and other temates for each method, depending on three classes of needed systems. This report provides the rationale for and description stringency of VaV (determined by rating 3 of a system's com-of those guidehnes. The actual guidehnes are presented in plexity and required integnty). Methods were then r&nk-ordered Volume 7, " User's Manual." Three factors determine what V&V for each of the three classes in terms of their overall cost-beno-is needed: (1) the stage of the development lifecycte; (2) wheth-fits and effectiveness. The applicabihty was then assessed of er the overall system or a speciahzed component needs to be each method for the four identified components of knowledge-tested; and (3) the stringency of V&V that is needed. A V&V based and expert systems, as well as the system as a whole.

guideline package is provided for each of the combinations of NUREG/CR-6316 V03: GUIDELINES FOR THE VERIFICATION these three vanables. The package specifies the V&V methods AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND recommended and the order in which they should be adminis-CONVENTIONAL SOFTWARE. Survey And Documentation Of tered, the assurances each method provides, the quakfications Expert System Venfication And Vahdation Methodologies.

needed by the V&V team, the performance measures that GROUNDWATER E.H; MILLER,LA.; MIRSKY,S.M. Science Ap-should be taken, and the decision entena. In addition to the phcations International Corp. (formerly Science Applications, guidehne packages, highly detailed step-by-step procedures are Inc.).

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

provided for 11 of the most important methods, to ensure that 83567:204.

they can be implemented correctly. The guidelines can apply to This report is the third volume in a series of reports desenb-conventional as well as to Al systems.

ing the results of the Expert System Venfication and Validation NUREG/CR-6316 V06: GUIDELINES FOR THE VERIFICATION (V&V) project that is jointfy funded by the U.S. Nuclear Regula.

AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND tory Commission and the Electric Power Research institute t CONVENTIONAL SOFTWARE. Validation Scenanos.

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

MIRSKY,S M.; HAYES J E'; MILLER LA. Science Apphcations The purpose of this activity was to survey and document tech-Intemational Corp (formerly Science Applications, Inc.). March niques presently in use for expert systems V&V. Via extensue telephone contacts, site visits, and through bibbographic 1995. 57pp 9504180445 SAIC 95/1028. 83567:312 g

PO searches a wide samphng of expert system V&V was accom-phshed. The major finding was that V&V of expert systems is ing the results of the Expert System Venfication and Vahdation not nearly as estabhshed or prevalent as V&V of conventional (VaV) project that is jointly funded by the U.S. Nuclear Regula.

software systems. There were few examples of V&V in the early tory Commission and the Electric Power Research Institute to stage of development. However, there is a very active research develop guidehnes for the V&V of expert and other systems.

area conceming the development of methods to assess the This achity was concerned with the development of a method-knowledge bases of expert and knowledge-based systems.

ology for selecting " validation scenarios." These are defined as

" realistic dynamic tests of software which covers only the in-NUREG/CR-6316 V04: GUIDELINES FOR THE VERIFICATION tended range of apphcations of the software and are designed AND VALIDATION EXPERT SYSTEM SOFTWARE AND CON-to sample important subsets of functions, usually for selected VENTIONAL SOFTWARE. Evaluation Of Knowledge Base Certh situations known to be challenging or problematic, to provide fication Methods. MILLER LA.; HAYES.J E.; MIRSKY,S M. Sci-assurance that the system achieves the tested functions with ence Apphcations international Corp. (formerly Science Applica-the required accuracy and performance." Such scenanos are tions, Inc.). March 1995.136pp. 9504180425. SAIC-95/1028.

used after all the V&V testing of the system is completed. Five 83568:001.

categones of validation scenanos were defined: PLANT, TEST, This report is the fourth volume in a senes of reports desenb-BASICS, CODE, and LICENSING. A sixth type, REGRESSION, ing the results of the Expert System Venfication and Vahdation is a composite of the others and refers to the practice of using (V&V) project that is jointly funded by the U.S. Nuclear Reguta-trusted scenarios to ensure that software modifications did not tory Commission and the Electric Power Research Institute to unintentionally change non-modified functions. A generalized develop guidelines for the V&V of expert and other systems.

procedure was deveioped for generating appropriate sets of val-Here are presented the results of the Knowledge Base Certifica-idation scenarios from these basic categones.

tion actuity that was concemed with developing and testing var-lous static analysis methods for assuring the quabty of knowl-NUREG/CR-6316 V07: GUIDELINES FOR THE VERIFICATION edge bases. The testing procedure used was that of a behavior.

AND VALIDAT!ON OF EXPERT SYSTEM SOFTWARE AND al experiment involving evaluation of four different V&V meth.

CONVENTIONAL SOFTWARE. User's W nual MILLER,L.A.;

ods. The study used two real nuclear expert systems: a boihng HAYES,J E.; MIRSKY,S M. Science App.ations International water reactor emergency operating procedures tracking system, Corp. (formedy Sc4ence Apphcations, Inc ) March 1995.250pp and a pressurized wa'er reactor safety assessment system. The 9504180447, SA C-95/1028, 83569:001.

twenty participants were from three nuclear utihties, the USNRC This report provides a step by-step guide. or user manual, for Technical Training Center, the Unwersity of Maryland. EG&G personnel respons,ble for the planning and execution of the von i

Main Citations and Abstracts 39 ification and validation (V&V), and also developmental testing, and Pressure Vessel Code are used to determine the adequacy of expert systems, conventional software systems, and also var-of the basket components. Special acceptance entena are pro-ious other types of artificial intelligence systems. While the posed to address the unique material charactenstics of austenit-guide was developed pnmanly for applications in the utilsty in-ic stainless steel, a matenal which is frequently used in the dustry, it applies well to all industnes. The user manual has basket assembhes.

three sections. In Section 1 the user assesses the sinngency of V&V needed for the system under consideration, identifies the NUREG/CR-6323: RELATIVE RISK ANALYSIS IN REGULATING development stage the system is en, and identifies the THE USE OF RADIATION-EMITTING MEDICAL DEVICES A component (s) of the system to be tested next. These three Preliminary Apphcation.

JONES,E.D.;

BANKS,W.W.;

pieces of information dotermine which package of V&V meth.

ALTENBACH,T.J.; et al. Lawrence Livermore National Laborato-ods, called a Guideline Package, is most appropnate for those ry. September 1995.181pp. 9510060245. UCRL ID-120051.

conditions. The V&V Guidelines Packages are provided in Sec.

85746:001.

tion 2. Each package consists of an ordered set of V&V tech.

This report desenbes a preliminary application of an analysis niques to be applied to the system, along with guides as to the approach for assessing relative nsks in the use of radiation-review / evaluation team, and the measurement entena. In Sec-emitting medical devices. Results are presented on human inits tion 3, the details of 11 of the most important (or least well ex-ated actions and failure modes that are most likely to occur in plained in the literature) methods are presented to assist the the use of the Gamma Knife, a gamma irradiation therapy user in the accurate application of these techniques.

device. This effort represents an initial step in a U.S. Nuclear Regulatory Commission (NRC) plan to evaluate the potential j

NUREG/CR-6316 V08: GUIDELINES FOR THE VERIFICATION n ana s reg a ng use of nu&ar mal 6 AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND vices. For this preliminary application of risk assessment, the CONVENTIONAL SOFTWARE. Bibliography.

MILLER.L. A.-

HAYES J E.; MIRSKY,S.M. Science Applications Internationa' t cus was to develop a basic process using existing techniques f r identifying the most likely nsk contributors and their relative Corp. (formerly Science Applications, Inc.). March 1995. 55pp.

importance. The approach taken developed relative risk rank-9504180451. SAIC-95/1028. 83568:308 This volume contains all of the techn$ cal references found in ings and profiles that incorporated the type and quality of data available and could present results in an easily understood Volumes 1-7 concerning the development of guidelines for the 8'

venfication and validation of expert systems, knowledge-based tional Laboratory for the NRC.

systems, other Al systems, object-onented systems, and con-ventional systems.

NUREG/CR-6324: OUALITY ASSURANCE FOR GAMMA NUREG/CR-6318: DATA

SUMMARY

REPORT FOR FISSION KNIVES. JONES.E.D.; BANKS.W.W.; FISCHER.LE. Lawrence PRODUCT RELEASE TEST VI-7.

OSBORNE.M F.;

Livermore National Laboratory. September 1995. 146pp.

9510060248. UCRL-ID-120056. ts5745:121.

LORENZ.R.A.; TRAVIS.J R.; et al Oak Ridge National Laborato.

ry. May 1995. 65pp. 9506220072. ORNL/TM-12937. 84394.001.

This report desenbes and summanzes the results of a quality Test VI-7 was the final test in the VI series conducted in the assurance (OA) study of the Gamma Knife, a nuclear medical vertical furnace. The fuel specimen was a 15.2-cm-long section device used for the gamma irradiation of intracranial lesions.

of a fuel rod from the Monticello boiling water reactor (BWR).

The study's focus was on the physical aspects of OA and did The fuel had expenenced a burnup of -40 Mwd /kg U. It was not address issues that are essentially medical, such as patient heated in an induction furnace for successive 20-min penods at selection or prescription of dose. A risk-based OA assessment 2000 and 2300 K in a moist air. helium atmosphere integral re-approach was used, in this report, sample programs for quality leases were 69% for 85Kr, 52% for (125)Sb, 71% for both control and assurance are included. The use of the Gamma (134)Cs and (137)Cs. and 0 04% for (154)Eu. For the non.

Knife was found to conform to existing standards and guidelines gamma-emitting species, release values for 42% for I. 4.1% for concerning radiation safety and quality control of external beam Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total therapies (shielding, safety reviews, radiation surveys, interlock mass released from the fut nace to the collection system, includ.

Systems, exposure monitonng, good medical physics practices, ing fission products, fuel, and structural matenals, was 0.89 g.

etc) and to be compliant with NRC teletherapy regulations.

with 37% being collected on the thermal gradient tubes and There are, however, current practices for the Gamma Knife not 63% downstream on filters. Posttest examination of the fuel covered by existing, formalized regulations, standards, or guide-specimen indicated that most of the cladding was completely lines. These practices have been adopted by Gamma Knife oxidized to ZrO(2), but that oxidation was not quite complete at users and continue to be developed with further experience.

the upper end The release behaviors for tho most volatile ele.

Some of these have appeared in publications or presentations ments, Kr and Cs, were in good agreement with the ORNL.

and are slowly finding their way into recommendations of pro-Booth Model fessional organizations.

NUREG/CR-6322: BUCKLING ANALYSIS OF SPENT FUEL NUREG/CR 6325: AN IMPLICIT STEADY-STATE INITIALIZATION BASKET. LEE.A.S.; BUMPAS.S E. Lawrence Livermore National PACKAGE FOR THE RELAP5 COMPUTER CODE.

Laboratory. May 1995. 60pp. 9505230342. UCRL-ID-119697.

PAULSEN.M.P.; PETERSON,C.E. Computer Simulation & Analy-84026 245.

sis, Inc. August 1995.142pp. 9509130168. 85433.001.

The basket for a spent fuel shipping cask is subjected to A direct steady-state initialization (DSSI) method has been compressive stresses that may cause global instability of the developed and implemented in the RELAPS hydrodynamic anal-basket assemblies or local buckling of the individual members.

ysis program. It provides a means for users to specify a small Adopting the common buckling design practice in which the sta-set of initial conditions which are then propagated through the bility capacity of the entire structure is based on the perform-remainder of the system. The DSSI scheme utilizes the steady-ance of the individual members of the assemblies, the typical state form of the RELAP5 balance equations for nonequilibrium spent fuel basket, which is composed of plates and tubular two-phase flow. It also employs the RELAPS component models structural members, can be idealized as an assemblage of col-and constitutive model packages for wall-to-phase and inter-umns, beam-columns and plates. This report presents the flex-phase momentum and heat exchange. A fully implicit solution of ural buckling formulas for five load cases that are common in the linearized hydrodynamic equations is implemented. An im-the basket buckling analysis: column under axial loads, column plicit coupling scheme is used to augment the standard steady-under axiat and bending loads, ptates under uniaxial loads, state heat conduction solution for steam generator use. It plates under biaxial loadings, and plato under biaxial loads and solves the primary-side tube region energy equations, heat con-lateral pressure. The acceptance entena from the ASME Boiler duction equations, wall heat flux boundary conditions, and over.

40 M:in CitSti ns cnd Ab2trccts all energy balance equations as a coupled system of equations months. A total of 371 respondents of the 589 who were sent and improves convergence. The DSSI methods for initializing questionnaires returned completed surveys, for a response rate RELAPS problems to steady. state conditions has been com-of 63% The body of the report presents the findings of the pared with the transient solution scheme using a suite of test survey including a bnef introduction to the approach used, fol-problems including; adiabatic single-phase liquid and vapor flow lowed by survey findings regarding regulations, policies and reg-through channels with and without heating and area changes; a ulatory guidance; expedence with hcensing apphcations, renew-heated two. phase test bundle representative of BWR core con-als and amendments; inspections; reporting requirements; and ditions; and a single-loop PWR model.

enforcement actions. 'ihe appendices of the report include a NUREG/CR-4327: MODELS FOR EMBRITTLEMENT RECOVERY copy of the survey as administered to licensees, a fuller de-DUE TO ANNEALING OF REACTOR PRESSURE VESSEL scription of the survey 01 sign and data collection methods, and STEELS. EASON.E.D.: WRIGHT.J E.; NELSON.E.E.: et al. Mod-detailed graphic matenal ocscribing survey responses, eling & Computer Services. May 1995. 76pp. 9505190018. MCS NUREG/CR-6331: ATMOSPHERIC RELATIVE CONCENTRA-950302,83982:001.

TIONS IN BUILDING WAKES. RAMSDELL,J.V.; SIMONEN.C.A.;

The irradiation embnttlement of the reactor pressure vessel.in SMYTH,S.B. Battelle Memorial Institute, Pacific Northwest Labo-nuclear power plants can be reduced by thermal annealing at ratory. May 1995.129pp. 9506220081. PNL-10521. 84394:066.

temperatures higher than the normal operating conditions. The This report documents the ARCON95 computer code devel-objective of this work was to analyze the pertinent data and de-oped for the U.S. Nuclear Regulatory Commission Office of Nu-velop quantitative models for estimating the recovery in 30 ft-lb clear Regulatory Research for use in control room habitability (41 J) Charpy transition temperature and Charpy and Charpy assessments. The document includes a user's guide to the upper shelf energy due to annealing. An analysis data base was code, a description of the technical basis for the code, and a developed, reviewed for completeness and accuracy, and docu-programmer's guide to the code. The ARCON95 code uses mented as part of this work, independent variables considered hourly meteorological data and recently developed methods for in the analysis included matenal chemistnes, anneahng time and estimating dispersion in the vicinity of buildings to calculate rela-temperature, arradiation time and temperature, fluence, and flux.

tive concentrations at control room air intakes that would be ex-To identify important variables and functional forms for predict-ceeded no more than five percent of the time. These concentra.

ing embnttlement recovery, pattern recognition transformation tions are calculated for averaging periods ranging from one hour analysis, and residual analysis were apphed together with cur' to 30 days in duration. Relative concentrations calculated by rent understanding of the mechanisms governing embnttlement ARCON95 are significantly lower than concentrations calculated and recovery. Models were calibrated using multivariable sur-using the currently accepted procedure when winds are less face-fitting techniques. The quality of fit was evaluated by con-than two meters per second. For higher wind speeds, sidenng both the Charpy annealing data used for fitting and a ARCON95 calculates about the same concentrations as the surrogate hardness data base. Several iterations of model cah-current procedure.

bration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness NUREG/CR-6333: BREATH VERSION 1.1 - COUPLED FLOW data produced improved correlation models for estimating AND ENERGY TRANSPORT IN POROUS MEDIA. Simulator De-Charpy upper shelf ensrgy and transition temperature after irra-senption And User Guide. STOTHOFF S.A. Southwest Research diation and annealing-Institute. July 1995. 93pp. 9509200049. CNWRA94-020.

NUREG/CR 6328: ADEOUACY OF THE 123 GROUP CROSS.

85539:00 t.

SECTION LIBRARY FOR CRITICALITY ANALYSES OF This document desenbes the BREATH computer code, in-WATER-MODERATED URANIUM SYSTEMS. PARKS,C.V.;

cluding the mathematical and numencal formulation for the sim-WRIGHT,R.O_; JORDAN W.C. Oak Ridge National Laboratory.

ulator, usage description, and sample input files with corre-August 1995.29pp.9508300310. ORNL/TM 12970. 85287:29t.

sponding output files. The BREATH computer code is designed in a recent enticahty analysis for an array of water-moderated to simulate one-dimensional flow of a liquid phase and disper-packages containing high-ennched uranium, the 123-group sive transport of the corresponding vapor species, coupled with cross-section library in the SCALE system was observed to energy transfer, in a heterogeneous porous medium. The have a nonconservative discrepancy of approximately 3 to 3.5%

BREATH code is organized into a flow equation simulator and when compared to more recently developed hbranes. A simple an energy equation simulator, which can be coupled or used in-representative system of UO(2)F(2) H(2)O was used to identi_

dependently. Both simulators use linear finite element basis fy that the problem results from a lack of resonance data for functions. A modified Picard iteration scheme is used to solve (235)U. Only a single set of self-shielded cross sections, most the nonhnear sets of equations. Heunstic algonthms are avail-hkely corresponding to a water-moderated infinite dilute system, able to control time stepping and the active solution domain.

was provided with the original data. The UO(2)F(2).H(2)O The BREATH simulator has been developed for use in auxiliary study indicates that this limitaton may cause nonconservative analyses which are a part of the Nuclear Regulatory Commis-discrepancies as high as 5.5% for some water-moderated, high-sion lterative Performance Assessment program. The simulator ennched uranium systems. Charactenstics of the systems where was developed in response to the observation from Total the discrepancy is evident are identified and discussed.

System Performance Assessments by both the Nuclear Regula-tory Commission and the U.S. Department of Energy that total-NURE3/CR-6330: RESULTS OF REGULATORY IMPACT system performance at the Yucca Mountain site in Nevada is SURVEY OF INDUSTRIAL. AND MEDICAL MATERIALS Ll-highly sensitive to the infiltration rate. Accordingly, this first ver.

CENSEES OF THE OFFICE OF NUCLEAR MATERIALS sion of the code is pnmanly intended to simulate processes im-SAFETY AND SAFEGUARDS.

LACH,D.;

MELBER,8.:

portant to infiltration and evaporation in chmatic and hydrologic BRICHOUX,J.: et al Battelle Human Affairs Research Centers-near surface environments representative of the Yucca Moun-June 1995.100pp. 9507060313. PNL 10548. 84533.007-tain site.

This report presents the findings of a regulatory impact survey of nuclear matenals heensees of the United States Nuclear NUREG/CR 6334: NEW SENSOR FOR MEASUREMENT OF Regulatory Commission (NRC). Commissioners of the NRC di-LOW AIR FLOW VELOCITY. Phase i Final Report.

rected staff to provide the Commission with first hand informa-HASHEMIAN,H.M.; HASHEMIAN,M.; RIGGSBEE,E.T. Analysis &

tion from hcensees that could be used to improve the overall Measurement Services Corp. August 1995.145pp.9508230253.

regulatory program. A self administered, mail-out survey ques-85120:001.

tionnaire was used to collect data from a sample of licensees This is the report of a six-month feasibihty study of a new who had interaction with the NRC dunng the previous 12 sensor to measure ambient air flow velocity and direction for

Main Citations and Abstracts 41 health physics applications in nuclear facilities. The information NUREG/CR-6347: MULTI-PHASE REACTIVE TRANSPORT from this sensor is to be used to determine where to place air THEORY. LICHTNER,P.C. Southwest Research Institute.

samplers to sample airborne radioactive matenal that is repre-LICHTNER,P.C. Center for Nuclear Waste Regulatory Analyses.

sentative of the air inhaled by radiation workers. A new sensor July 1995.102pp. 9508090044. CNWRA94 018. 84963.087.

was developed in this project and successfully tested in the Physicochemical processes in the near-field region of a high-AMS laboratory for measurement of low flow rates of air. The level waste repository may involve a diverse set of phenomena sensor uses a conventional thermocouple as its sensing ele.

including flow of liquid and gas, gaseous diffusion, and chemical ment and is therefore referred to as a " thermocouple flow reaction of the host rock with aqueous solutions at elevated sensor", The dynamic response of the thermocouple is meas-temperatures. This report develops some of the formalism for ured using an in-situ response time testing method. The re-desenbing simultaneous multscomponent solute and heat trans-sponse time information is then converted to a flow signal using port in a two-phase system for partially saturated porous media.

predetermined response time versus-flow correlation for the Diffusion of gaseous species is desenbed using the Dusty Gas thermocouple. The thermocouple flow sensor has the potential Model which provides for simultaneous Knudsen and Fickian to aid in determining in-door air flow patterns. This may be ac-diffusion in addition to Darcy flow. A new form of the Dusty Gas complished by using multiple thermocouples to measure air flow Model equations is denved for binary diffusion which separates velocities in several locations in the room and use the velocity the total diffusive flux into segregative and nonsegregative com-information with computational fluid dynamics or rmural network ponents. Migration of a wetting front is analy7ed using the models to establish air flow patterns.

quasi-stationary state approximation to the Richards' equation.

Heat-pipe phenomena are investigated for both gravity-and NUREG/CR-6335: FATIGUE STRAIN-LIFE BEHAVIOR OF capillary-dnven reflux of liquid water. An expression for the CARBON AND LOW ALLOY STEELS, AUSTENITIC STAIN.

burnout permeability is denved for a gravity-dnven heat. pipe. Fi-LESS STEELS, AND ALLOY 600 IN LWR ENVIRONMENTS.

nally an estimate is given for the change in porosity and perme-KEISLER.J.; CHOPRA,0.K.; SHACK,W.J. Argonne National Lab-ability due to mineral dissolution which could occur in the region oratory. August 1995. 82pp. 9508300329. ANL-95/15.

of condensate formation in a heat-pipe.

85288:221.

The existing fatigue strain vs. life (S-N) data, foreign and do-NUREG/CR-6348: THERMALLY DRIVEN MOISTURE REDlSTRI-BUTION IN PARTIALLY SATURATED POROUS MEDIA.

mestic, for carbon and low-alloy steels, austenitic stainless steels, and Alloy 600 used in the construction of nuclear power GREEN R.T.; MANTEUFEL,R.D.; MEYER,K.A.; et at Center for plant components have been compiled and categonzed accord-Nuclear Waste Regulatory Analyses. December 1995. 209pp.

9601290290. CNWRA 95-005. 86892:001.

ing to matenal, loading, and environmental conditions. Statistical Expenmental and theoretical studies have been conducted to models have been developed for estimating the effects of the develop a quantitative understar' ding of the thermohydrologic vanous service conditions on the fatigue life of these matenals.

The results of a ngorous statistical analysis have been used to phenomena induced by emplacement of high-level radioactive waste (HLW) in an unsaturated fractured porous media. A series estimate the probability of initiating a fatigue crack. Data in the of laboratory experiments was conducted in a variety of media literature were reviewed to evaluate the effects of size, geome-to study the physics of thermally dnven moisture redistnbution try, and surface finish of a component on its fatigue hfe. The through porous media. Principles of similarity theory were ap-fatigue S-N curves for components have been determined by plied to develop dimensionless parameters to be used to predict adjusting the probability distnbution curves for smooth test thermohydrologic behavior at the field and repository scales i

specimens for the effect of mean stress and applying design from information gained from the laboratory and field scales.

margins to account for the uncertainties due to component Results from the dimensional analyses indicated that dimension-size / geometry and surface finish. The significance of the effect less parameters are useful to design field-scale experiments so of environment on the current Code design curve and on the as to maximize the information gained from the experiment and j

proposed intenm design curves published in NUREG/CR-5999 to predict heat and mass transfer of larger scale porous media i

is discussed. Estimations of the probability of fatigue cracking in systems. Dimensional analysis did not prove to be useful in sample componer'ts from BWRs and PWRs are presented-scaling systems containing discrete features such as faults or j

fractures. A major conclusion of this report is that conduct of a NUREG/CR-6343: ON-LINE TESTING OF CAllBRATION OF heater test at field scale is crucial to inve,tigste and understand PROCESS INSTRUMENTATION CHANNELS IN NUCLEAR heat and mass transfer through par $ ally saturated fractured, POWER PLANTS. Phase 11 Final Report. HASHEMIAN.H.M.

porous media at the scale of a geologic repository.

Analysis & Measurement Services Corp. November 1995.

342pp. 9512260334. 86625.001.

NUREG/CR-6349: COST-BENEFIT CONSIDERATIONS IN REGU-This report presents the results of a comprehensive research LATORY ANALYSIS. MUBAYI,V.; SAILOR,V. Brookhaven Na-and development program to develop and validate a new tech-tional Laboratory. ANANDALINGAM G. Pennsylvania, Univ. of, nque for remote testing of calibration of process instrumenta-Philadelphia, PA. October 1995.148pp. 9511270458. BNL-J tion channels in nuclear power plants. This technique can be NUREG-52466. 86317:081.

used to venfy the cahbration of instrument channels (including Justoication for safety enhancements at nuclear facihties, e g.,

the field sensors) while the plant is on-hne. The method is a compulsory backfit to nuclear power plants, requires a value-simple and passive, uses the existing plant instrumentation, and impact analysis of the increase in overall public protection includes the effects of process operating conditions on the cali-versus the cost of implementation. It has been customary to bration of the instruments. The new technique is based on mon-assess the benefits in terms of radiation dose to the public storing the steady state output of process instrumentation chan-averted by the introduction of the safety enhancement. Compar-nels on a periodic or continuous basis dunng each fuel cycle to ison of such benefits with the costs of the enhancement then identify any significant dnft. If the drift is more than an allowable requires an estimate of the monetary value of averted dose value, then the channel is calibrated. Otherwise, the channel is (dollars / person rem). This report reviews available information not cahbrated. To separate process dnft from instrument dnft, on a vanety of factors that affect this valuation and assesses analytical models or a cahbrated refererme channel may be the continuing validity of the figure of $1000/ person-rem avert-used to track the process independently and determine if it is ed, which has been widely used as a guideline in performing dnfting. On-hne monitoring is performed not only dunng normal value-impact analyses. Factors that bear on this valuation in-operation but also dunng plant startup and shutdown penods to clude the health risks of radution doses, especially the higher obtain data to venty instrument cahbrations throughout their op-nsk estimates of the BEIR V committee, recent calculations of erating range.

doses and offsite costs by consequence codes for hypothesired I

I

i 42 Main Citations and Abstracts severe accidents at U.S. nuclear power plants under the the HEP results wera compared with the failure rates observed

)

NUREG 1150 program, and recent ir formation on the economic in the examinations. The ASEP pretcza was applied by PNL consequences of the Chernobyl accident in the Soviet Union operator license examiners who supplemerded the ihrmed Mo j

and estimates of risk avoidance based on the willingness-to-pay mation in the examination reports with expert jud2 ment based i

cnterion. The report analyzes these factors and presents results upon their extensive simulator examination experience. Two on the dollars /persorprem ratio arising from different assump-tests were performed to assess the bias of the ASE,P HEPs i

tions on the values of these factors.

compared with the data from the requalification examinations.

The first compared to the average of the ASEP HEP values with NUREG/CR-6351: REVIEW OF SCENARIO SELECTION AP.

the fraction of the population actual!y failed. It found the exist-PROACHES FOR PERFORMANCE ASSESSMENT OF HIGH-ence of a statistically significant factor of two bias on the aver.

LEVEL WASTE REPOSITORIES AND RELATED ISSUES.

age. The second test partitioned the critical tasks into sub-BONANO,E.J. Southwest Research Institute. BACA,R.G. Beta groups based on the ASEP HEP values and compared the sub-Corp, international. August 1995. 78pp. 9509070092. CNWRA group average ASEP HEP values with the observed subgroup 94-002. 85401:230-failure rates. It found httle or no bias for small ASEP HEP The selection of scenarios representing plausible realizations values, but a larger bias for Larger ASEP HEP Values.

of the future conditions-with associated probabilities of occur-rence-that can affect the long-term performance of a high-level NUREG/CR-6356: HYDRAULIC CHARACTERIZATION OF HY-rad:oactive waste (HLW) repository is the commonly used DROTHERMALLY ALTERED NOPAL TUFF. GREEN,R.T.;

method for treating the uncertainty in the prediction of the future MEYER JAMES K.A Southwest Research Institute. RICE,G.

states of the system. This method, conventionally referred to as George Rice & Associates. July 1995. 69pp. 9509150281.

the " scenario approach," while common is not the only method CNWRA 94-027. 85521:001.

to deal with this uncertainty; other methods, such as the envi-Understanding the mechanics of variably saturated flow in ronmental simulation approach (ESA), have also been pro-fractured-porous media is of fundamental importance to evaluat.

posed. Two of the difficulties with the scenano approach are ing the isolation performance of the proposed high-level radio-the lack of uniqueness in the definition of the term " scenario" active waste repository for the Yucca Mountain site. Developing and the lack of uniqueness in the approach to formulate scenar-that understanding must be founded on the analysis and inter.

ios, which relies considerably on subjective judgments. Conse-pretation of laboratory and field data. This report presents an quently, it is difficult to assure that a complete and unique set of analysis of the unsaturated hydraulic properties of tuff cores scenarios can be defined for use in a performance assessment.

from the Pena Blanca natural analog site in Mexico. The basic Because scenarios are key to the determination of the long-intent of the analysis was to examine possible trends and rela-term performance of the repository system, this lack of unique-tionships between the hydraulic properties and the degree of ness can present a considerable challenge when attempting to hydrothermal alteration exhibited by the tuff samples. These reconcile the set of scenanos, and their level of detal, obtained data were used in flow simulations to evaluate the significance using different approaches, particularly among proponents and of a particular conceptual (composite) model and of distinct hy.

regulators of a HLW repository, draulic properties on the rate and nature of water flow.

NUREG/CR 6354 DRF FC: PERFORMANCE TESTING OF ELEC.

NUREG/CR-6358 V01: ASSESSMENT OF UNITED STATES IN-TRONIC PERSONAL DOslMETERS. Draft Report For Comment.

DUSTRY STRUCTURAL CODES AND STANDARDS FOR AP-SWINTH,K.L.; MCDONALD,J.C.; SISK D R.; et al. Battelle Me-PLICATION TO ADVANCED NUCLEAR POWER monal Institute, Pacific Northwest Laboratory. August 1995.

REACTORS Final Report. ADAMS T.M.; STEVENSON J.D. Ste-172pp. 9508230227. PNL 10560. 85127:119' venson & Associates. October 1995. 188pp. 9510310370.

In radiation protection, incremental control of worker radiation 86032:154 exposures is important to ensure that periodic dose limits are Through out its history, the USNRC has been committed to not exceeded. Electronic personal dosimeters, EPDs, are widely the use of US industry consensus standards for the design, used for this apphcation. As their rehability has improved users have shown an interest in their use for both incremental control construction, and hcensing of commercial nuclear power facili-ties. The existing industry standards are based on the current and as the pnmary dosimeter to track the dose of record for the worker, in this application they would replace the traditional film class of light water reactors and as such may not adequately address design and construction teatures of the next generation or thermoluminescent dosimeter whose performance is thor-oughly understood. The EPD bnngs with it some of the prob-of Advanced Light Water Reactors and other types of Advanced lems of instruments which are not seen with the traditional dosi-Reactors. As part of their on going commitment to industry meters. The report contains results of a survey of users and a standards, the USNRC commissioned this study to evaluate U S, industry structural standards for apphcation to Advanced survey of vendor hterature that highhght some of the lirritations and problems of EPDs. The radiation protection comn, unity is Reactors. The initial review effort included: (1) the review and concernad that the reliabihty and accuracy of the data frorr, the study of the relevant reactor design basis documentation for EPD be comparable to traditional methods if they assume this eight Advanced Reactor Designs, (2) the review of the USNRCs additional role. The report hsts type tests, test methods and design requirements for advanced reactors, (3) the review of j

calibration methods intended to ensure the required reliabihty.

the latest revisions of the relevant industry structural standards, and (4) the identification of the need for changes to these l

NUREG/CR-6355: A LIMITED ASSESSMENT OF THE ASEP standards. The results of these studies were used to develop j

HUMAN RELIABILITY ANALYSIS PROCEDURE USING SIMU-recommended changes 'o industry consensus structural stand.

LATOR EXAMINATION RESULTS. GORE,0.F.; DUKELOW J.S ;

ards which will be used in the construction of Advanced Reac-MITTS,T.M.; et al. Battelle Memorial Institute, Pacific Northwest tors Over seventy sets of proposed standard changes were i

Laboratory. October 1995.150pp. 9511020380. PNL 10573.

recommended and the need for the development of four new

(

86000 030.

standards was identified. In addition to the recommended stand-The ASEP post-accident, post-diagnosis, nominal HRA proce-ard changes. several other sets of information were extracted dure described in NUREG/CR-4772 is assessed within the con-for use by USNRC in other programs. This information included.

text of an individual's performance of cntical tasks on the simu-(1) detailed observations on the response of structures and dis-lator portion of requalification examinations administered to nu-tribution system supports to the recent Northridge, CA (1994) clear power plant operators. The data for this study are denved and Kobe, Japan (1995) earthquakes, (2) comparison of ver-from simulator examination reports from the NRC requalification sons of certain standards cited in the standard review plan to examination cycle. The ASEP procedure was used to estimate the most current versions, and (3) companson of the seismic human error probabihty (HEP) values for the entical tasks, and and wind design basis for the subject reactor designs. Finally

F Main Citations and Abstracts 43 i

provided is a suggested plan of action for implementation of the formed by Battelle Pacific Northwest Laboratones in support of l

recommended industry standard changes.

the NRC's Standard Review Plan Update and Development Pro.

NUREG/CR-6358 V02: ASSESSMENT OF UNITED STATES IN-grarn. Significant changes to the standards, from the cited ver-l DUSTRY STRUCTURAL CODES AND STANDARDS FOR AP.

sion to the latest version, are desenbed and discussed in a tab-I filCATION TO ADVANCED NUCLEAR POWER ular format for each standard. Recommendations for updating REACTORS. Appendices. ADAMS,T.M.; STEVENSON.J D. Sto-each citation in the Standard Review Plan are presented. Tech-venson & Associates. October 1995. 150pp. 9510310376.

nical considerations and suggested changes are included for re-l 86032:008.

lated regulatory documents (i.e., Regulatory Guides and the See NUREG/CR-6358,V02 abstract.

Code of Federal Regulations) citing the standard. The results and recommendations presented in this document have not i

1 NUREG/CR-6368: EXPERIMENTAL INVESTIGATION OF SEDI-been subjected to NRC staff review.

MENTATION OF LOCA-GENERATED FIBROUS DEBRIS AND SLUDGE IN BWR SUPPRESSION POOLS. SOUTO,F.J.;

NUREG/CR-6386: COMPARISONS OF ANSI STANDARDS l

RAO,D.V. Science & Engineenng Associates, Inc. December CITED IN THE NRC STANDARD REVIEW PLAN,NUREG-0800, l

1995. 82pp. 9601290176. SEA 95-554-06A9. 86881:271.

AND RELATED DOCUMENTS.

ANKRUM,A.R.;

Several tests were conducted in a 1:2.4 scale model of a BOHLANDER,K.L.; GILBERT,E.R.; et al. Battelle Memonal Insti-l Mark I suppression pool to investigate the behavior of fibrous tute, Pacific Northwest Laboratory. November 1995. 126pp.

insulation and sludge debris under LOCA conditions.

9512260328. PNL 10769. 06626:130.

l NUKON(TM) shreds, manually cut and torn up in a leaf shred-This report provides the results of comparisons of the cited I

der, and iron oxide particles were used to simulate fibrous and and latest versions of ANSI standards cited in the NRC Stand-l sludge debris, respectively. The suppression pool model includ-ard Review Plan for the Review of Safety Analysis Reports for ed four downcomers fitted with pistons to simulate the steam-Nuclear Power Plants (NUREG 0800) and related documents.

i I

water oscillations during chugging expected dunng a LOCA. The The compansons were performed by Battelle Pacific Northwest study was conducted to provide debns settling velocity data for Laboratones in support of the NRC's Standard Review Plan the models used in the BLOCKAGE computer code, developed Update and Development Program. Significant changes to the to estimate the ECCS pump head loss due to clogging of the standards, from the cited version to the latest version, are de-strainers with LOCA generated debns. The tests showed that senbed and discussed in a tabular format for each standard.

i the debns, both fibrous and particulate, remains fully mixed Recommendations for updating each citation in the Standard l

dunng chugging; they also showed that, dunng chugging, the fi-Review Plan are presented. Technical considerations and sug-brous debns underwent fragmentation into smaller sizes, includ-gested changes are included for related regulatory documents ing individual fibers. Measured concentrations showed that f'-

(i e., Regulatory Guides and the Code of Federal Regulations) brous debns settled slower than the sludge, and that the set-citing the standard. The results and recommendations present-tling behavior of each matenal is independent of the presence ed in this document have not been subjected to NRC staff of the other matenal. Finally, these tests showed that the as' review.

j sumption of considenng uniform debns concentration dunng stra.ner calculations is reasonable. The tests did not consider NUREG/CR-6390: RADIOLOGICAL CHARACTERIZATION OF the effects of the operation of the ECCS on the transport of SPENT CONTROL ROD ASSEMBLIES.

LEPEL,E.A.;

l debns in the suppression pool.

ROBERTSON.D E.; THOMAS C.W.; et al. Battelle Memorial in-NUREG/CR-6382: COMPARISONS OF ASTM STANDARDS stitute, Pacific Northwest Laboratory. October 1995. 59pp.

95 060201 PNL 10

6. 86094 200 CITED IN THE NRC STANDARD REVIEW PLAN.NUREG 0800, i

l AND RELATED DOCUMENTS ANKRUM,A.R.;

BOHLANDER,K.L.; GILBERT E.R.; et al. Battelle Memonal Insti-ratory radionuclide measurements, as well as waste classifica-tion assessments, of BWR and PWR spent control rod assem-tute, Pacific Northwest Laboratory. October 1995. 124PP.

blies. The radionuclide invertones of these spent control rods 9511020389. PNL-10747. 86090:163.

This report provides the results of compansons of the cited were determined by three separate methodologies, including 1) direct assay techniques, 2) calculational techniques, and 3) by and latest versions of ASTM standards cited in the NRC Stand-ard Review Plan for the Review of Safety Analysis Reports for sampling and laboratory radiochemical analyses. For the BWR control rod blade (CRB) and PWR burnable poison rod assem-Nuclear Power Plants (NUREG 0800) and related documents.

bly (BPRA), (60)Co and (63)Ni, present in the stainless steel The compansons were performed by Battelle Pacific Northwest Laboratones in support of the NRC,s Standard Review Plan cladding, were the most abundant neutron activation products.

The most abundant radionuclide in the PWR rod cluster control Update and Development Program. Significant changes to the assembly (RCCA) was (108m)Ag (130 yr halfisfe) produced in stardards, from the cited version to the latest version, are de-scnbed and discussed in a tabular format for each standard.

the Ag-In-Cd alloy used as the neutron poison. This radionuclide will be the dominant contnbutor to the gamma dose rate for Recommendations for updating each citation in the Standard many hundreds of years. The results of the direct assay meth-Oeview Plan are presented. Technical considerations and sug-ods agree very well ( 10%) with the sampling / radiochemical j

gested changes are included for related regulatory documents measurements. The results of the calculational methods agreed (te., Regulatory Guides and the Code of Federal Regulations) fairly well with the empincal measurements for the BPRA, but citing the standard The results and recommendations present-often varied by a factor of 5 to 10 for the CRB and the RCCA ed in this document have not been subjected to NRC staff assemblies. If concentration averaging and encapsulation, as al-review.

lowed by 10CFR61.55, is performed, then each of the entire NVREG/CR 6385: COMPARISONS OF ANS,ASME,AWS, AND control assemblies would be classified as Class C low-level ra-NFPA STANDARDS CITED IN THE NRC STANDARD REVIEW dioactive waste PLAN,NUREG-0800 AND RELATED DOCUMENTS.

ANKRUM,A.R.; BOHLANDER,K.L.; GILBERT,E.R.; et al. Battelle NUREG/CR-6398: EVALUATION OF THE COMPUTERIZED PRO-Memonal institute, Pacific Northwest Laboratory. November CEDURES MANUAL ll (COPMA-II). CONVERSE,S.A. North 1995.130pp. 9512150115. PNL-10768. 8M34:126.

Carolina State Univ., Raleigh, NC, November 1995. 57pp.

This report provides the results of compansons of the cited 9512260314. 86627:098.

and latest versions of ANS, ASME, AWS and NFPA standards The purpose of this study was to evaluate the effects of a cited in the NRC Standard Review Plan for the Review of computerized procedure system, the Computerized Procedure Safety Analysis Reports for Nuclear Power Plants (NUREG Manual 11 (COPMA-II), on the performance and mental workload 0800) and related documents. The comparisons were per-of licensed reactor operators. To evaluate COPMA-II, eight I

l l

a

44 Main Citations and Abstracts teams of two operators were trained to operate a scaled pres-path for transmission of stress waves resulting from wire surized water reactor facility (SPWRF) with traditional paper pro-breaks. Accelerometers placed on the bearing plates at the cedures and with COPMA-II. Following training, each team oper-ends of tendons recorded high-intensity waveforms. However, ated the SPWRF under normal operating conditions with both accelerometers placed on concrete surfaces revealed substan-paper procedures and COPMA II. The teams then performed tial attenuation and reduced intensities. Locations of wire breaks one of two accident scenarios with paper procedures, but per-were determined accurately through measurement of differ-formed the remaining accident scenano with COPMA-II. Per-ences in arrival times of the signal at the two sensors. Pattern E

half as many errors dunng the accident scenarios with COPMA-formance measures and subjective estimates of mental work-recognition systems utilized in conjunction with the proposed load were recorded for each performance trial. The most impor-concept will provide a basis for an integrated and automated tant finding of the study was that the aperators comrnitted only tool for identification of wire breaks.

18 as they committed with paper procedures. However, time to NUREG/GR-0014: BALDCYPRESS TREE RING ELEMENTAL initiate a procedure was fastest for paper procedures for acci-CONCENTRATIONS AT REELFOOT LAKE TENNESSEE,FROM dent scenano tnals. For performance under normal operating AD 1795 TO AD 1820. VAN ARSOALE,R. University of Mem-conditions, there was no difference in time to initiate or to com-phis, Memphis, TN. HALL G. Rutgers, New Jersey, State Univ.

plete a procedure, or in the number of errors committed with of, Piscataway, NJ. November 1995. 27pp. 9511290103.

paper procedures and with COPMA-II There were no consistent 86353.271.

differences in the mental workload ratings operators recorded Many 200 year old baldcypress trees in Reelfoot Lake, Tenn-,

for trials with paper procedures and COPMA-II.

lived through the great New Madrid earthquakes of 1811 1812.

NUREG/CR-6420:

SELF MONITORING SURVEILLANCE This study was undertaken to determine if the elemental com-SYSTEM FOR PRESTRESSING TENDONS. TABATABAI,H.

position of baldcypress tree rings showed any systematic vana.

Construction Engineenng Laboratory, Inc. December 1995.

tion through the earthquake period of AD 1795-1820. Cores 73pp. 9601290169. 86879.282.

were collected from two Reelloot Lake baldcypress trees and Assured safety and operational reliability of post-tensioned analyzed using inductively Coupled Plasma Mass Spectrometry.

concrete components of nuclear power plants are of great im-Individual yearly nngs and five-year nng segments were an+

portance to the public, electnc utilities, and regulatory agencies.

lyzed to determene their elemental compositions. The cores Prestressing tendons provide pnncipal reinforcement for con-were analyzed for Li through U but only Ba, Ce, Cs, Cu, I, La, tainment structures. In this phase of the research effort, the fea-Mg, Mn, Nd, Rb, Sm, Sr. and ZN were found to be above the sibility of developing a passive surveillance system for identifica-detection limit. Of these only Co, I, La, Nd, Rb, and Sm showed tion of ruptures in tendon wires was evaluated and venfied. A any systematic changes within individual cores. Companson of one-tenth scale nng model of the Pa:o Verde nuclear contain-three cores taken from one tree reveal that tree-ring elemental ment structure was built inside the Structural Laboratory. Dy-concentrations and any changes through time are very different namic scaling (similitude) relationships were used to relate among the cores. When companng the elemental concentra-measured sensor responses (to intentional wire breaks) to the tions of tree nngs for the same years in the two trees neither expected prototype response. Strong and recognizable signa-elemental concentrations nor changes through time were semi-tures were detected by the accelerometers used. It is concluded lar. The elemental concentrations show no systemic changes that the unbonded prestressing tendons provide an excellent through the earthquake period of AD 1795 1820.

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.

i SECONOARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER 04-4448-010 NUREG/CR4074 V05 MCS 950302 NUREG/CR-6327 04-4448 013 NUREG/CR4074 V04 MTR 940000114 NUREG/CR-6263 V02 ANL-94/35 NUREG/CR4266 MTR 94W0000114 NUREG/CR-6263 V01 ANL 94/37 NUREG/CR-6275 ORNL4713 NUREG/CR 5857 4

ANL-95/15 NUREG/CR 6335 ORNL-6734 NUREG/CR 5944 V02 ANL-95/2 NUREG/CR-4667 V18 ORNL 6765 NUREG/CR-6089 ANL-95/25 NUREG/CR-4667 V19 ORNL4791 NUREG/CR-6192 ANL-95/5 NUREG/CR-6315 ORNL 6795 NUREG/CR-6214 BMI-2173 NUREG/CR-4599 V04 N1 ORNL/NOAC-232 NURFG/CR-4674 V21 BMi-2174 NUREG/CR-6004 ORNL/NOAC-232 NUREG/CR-4674 V22 BMI-2179 NUREG/CR-6235 ORNL/NUREG/CSD2 NUREG/CR-0200 V1 R04 BMI-2180 NUREG/CR4251 ORNL/NUREG/CSD2 NUREG/CR-0200 V2PIR4 BMI 2181 NUREG/CR-6264 V01 ORNL/NUREG/CSD2 NUREG/CR-0200 V3 R04 BMi-2181 NUREG/CR4264 V02 ORNL/NUREG/CSD2 NUREG/CR 0200 V2P2R4 BMI-2182 NUREG/CR4297 ORNL/TM-11568 NUREG/CR 5591 V03 BMI-2183 NUREG/CR4298 ORNL/TM-11568 NUREG/CR-5591 V04 N2 BMI-2184 NUREG/CR4300 ORNL/TM 11568 NUREG/CR-5591 VOS N1 BNL-NUREG-51581 NUREG/CR 2907 V13 ORNL/TM-11568 NUREG/CR-5591 VOS N2 BNL-NUREG-51581 NUREG/CR-2907 V14 ORNL/TM 11568 NUREG/CR-5591 V06 N1 BNL-NUREG-51708 NUREG/CR-3469 V08 ORNL/TM-12796 NUREG/CR4259 BNL NUREG-52332 NUREG/CR-6002 ORNL/TM-12801 NUREG/CR-6261 BNL-NUREG 52398 NUREG/CR4141 ORNL/TM-12845 NUREG/CR 6284 BNL-NUREG 52399 NUREG/CR4144 V06 P1 ORNL/TM-12866 NUREG/CR4273 BNL-NUREG 52399 NUREG/CR-6144 V06 P2 ORNL/TM-12937 NUREG/CR4318 BNL-NUREG-52399 NUREG/CR4144 V01 ORNL/TM 12970 NUREG/CR4328 BNL NUREG-52410 NUREG/CR4954 ORNL/TM4789 NUREG/CR4016 BNL NUREG-52428 NUREG/CR4172 ORNL/TM-9593 NUREG/CR 4219 V10 N2 BNL NUREG-52431 NUREG/CR4265 ORNL/TM-9593 NUREG/CR-4219 V11 N1 BNL-NUREG 52440 NUREG/CP-0143 ORNLSUB82222521 NUREG/CR-3243 BNL-NUREG 52446 NUREG/CR-6305 ORNLSUB94SD4271 NUREG/CR-6240 BNL NUREG 52466 NUREG/CR4349 ORNLSUB94SD4272 NUREG/CR-6239 V01 BNL/NUREG-52394 NUREG/CR-6112 ORNLSUB94SO4272 NUREG/CR-6239 V02 BSRC-700/95/005 NUREG/CR-5758 V05 PNL 10521 NUREG/CR-6331 CNWRA 93-020 NUREG/CP 0147 PNL 10548 NUREG/CR4330 CNWRA 94 002 NUREG/CR-6351 PNL 10560 NUREG/CR-6354 DAF FC CNWRA 94 017 NUREG/CR-6283 PNL-10573 NUREG/CR4355 q

CNWRA 94 027 NUREG/CR4356 PNL 10638 NUREG/CR-5758 VOS CNWRA 95-005 NUREG/CR-6348 PNL 10747 NUREG/CR4382 CNWR A94-018 NUREG/CR-6347 PNL 10768 NUREG/CR-6385 CNWRA94 020 NUREG/CR-6333 PNL 10769 NUREG/CR-6386 CONF-940738 NUREG/CP 0141 PNL 10806 NUREG/CR-6390 EGG-2720 NUREG/CR4150 V01 PNL-4221 NUREG/CR 2850 V13 EGG 2720 NUREG/CR4150 V02 PNL4462 NUREG/CR-5973 A02 EGG-2720 NUREG/CR-6150 V03 PNL-8466 NUREG/CR-5975 R01 EGG-2720 NUREG/CR4150 VO4 PNL-8497 NUREG/CR4054 EGG-2720 NUREG/CR4150 V05 PNL-8742 NUREG/CR4884 V01 EGG 2746 NUREG/CR4276 PNL-8742 NUREG/CR 5884 V02 EUR 15855EN NUREG/CR-6244 V03 SAIC-95/1028 NUREG/CR-6316 V01 EUR 15855EN NUREG/CR-6244 V01 SAIC-95/1028 NUREG/CR-6316 V02 EUR 15855EN NUREG/CR4244 V02 SAIC-95/1028 NUREG/CR-6316 V03 INEL 94//0123 NUREG/CR-629I V04 SAIC-95/1028 NUREG/CR-6316 V04 INEL-94/0016 NUREG/CR 6285 SAIC-95/1028 NUREG/CR4316 VOS INEL-94/0039 NUREG/CR-6116 V09 SAIC-95/1028 NUREG/CR4316 V06 INEL 94/0039 NUREG/CR-6116 VIO SAIC-95/1028 NUREG/CR4316 V07 INEL-94/0039 NUREG/CR4116 V06 SAIC-95/1028 NUREG/CR-6316 V08 INEL-94/0090 NUREG/CR-5462 SAND 91-2801 NUREG/CR 5927 V02 INEL 94/0111 NUREG/CP 0144 V01 SAND 921720 NUREG/CR-6184 INEL-94/0111 NUREG/CP-0144 V02 SAND 93-0528 NUREG/CR 6017 INEL-94/0123 NUREG/CR-6291 V01 SAND 93-2078 NUREG/CR-6109 INEL 94/0123 NUREG/CR-6291 V02 SAND 93-2185 NUREG/CR4119 V01 INEL-94/0123 NUREG/CR-6291 V03 SAND 93-2185 NUREG/CR4119 V02 INEL-94/0156 NUREG/CR 6100 SAND 93-2370 NUREG/CR-6134 INEL-94/0278 NUREG/CR-5229 V07 SAND 93-2371 NUREG/CR4135 INEL-95/0045 NUREG/CR4260 SAND 93-2372 NUREG/CR-6136 INEL-95/0070 NUREG/CR4257 SAND 93-2440 NUREG/CR4143 V06 P1 INEL-95/0073 NUHEG/CR4256 V01 SAND 93-2440 NUREG/CR4143 V06 P2 INEL 95/0073 NUREG/CR4256 V02 SAND 93-2440 NUREG/CR-6143 V01 INEL-95/0153 NUREG/CR4188 V02 SAND 94 0769 NUREG/CR 6220 INEL 95/0174 NUREG/CR-5535 V01 SAND 941453 NUREG/CR4244 W3 INEL 95/0174 NUREG/CR-5535 V02 SAND 941453 NUREG/CR-6244 V01 INEL-95/0174 NUREG/CR 5535 V04 SAND 941453 NUREG/CR 6244 V02 INEL-95/0174 NUREG/CR-5535 V05 R1 SAND 94-1711 NUREG/CR4154 V02 LA 12915-MS NUREG/CR4311 SAND 941795 NUREG/CR-6173 45

1 l

46 Secondary Report Number Index SECONDARY REPORT NUMBER REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMBER SEA 93-554MA1 NUREG/CR-6224 UCHL-ID 119697 NUREG/CR-6322 SEA 95-554MA9 NUREG/CR 6368 UCRL-ID-120051 NUREG/CR-6323 UCRL-ID 117418 NUREG/CR4242 UCRL-ID 120056 NUREG/CR4324 UCRL lO-118766 NUREG/CR-5657 UILUENG942009 NUREG/CR4259 l

l

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.

ABELOUIST.E,W.

B AC A,R.G.

NUREG-1507 DRFT FC: MINIMUM DETECTABLE CONCENTRATIONS NUREG 1464: NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE WITH TYPICAL RADIATION SURVEY INSTRUMrNTS FOR VARIOUS

2. Development Of Capatuhties For Review Of A Performance Assess-CONTAMINANTS AND FIELD CONDITIONS. Draft Report For Com-ment For A High-Level Waste Repository.

ment NUREG/CR4348. THERMALLY DRIVEN MOISTURE REDISTRIBUTION IN PARTIALLY SATURATED POROUS MEDIA.

ADAMS,T.M.

NUREG/CR-6351: REVIEW OF SCENARIO SELECTION APPROACHES l

NUREG/CR4358 VOI: ASSESSMENT OF UNITED STATES INDUSTRY FOR PERFORMANCE ASSESSMENT OF HIGH-LEVEL WASTE RE-STRUCTURAL CODES AND STANDARCS FOR APPLICATION TO POSITORIES AND RELATED ISSUES.

ADVANCED NUCLEAR POWER REACTORS Final Report.

NUREG/CR 6358 V02: ASSESSMENT OF UNITED STATES INDUSTRY BAGCHl,G.

STRUCTURAL CODES AND STANDARDS FOR APPLICATION TO NUREG 1522. ASSESSMENT OF INSERVICE CONDITIONS OF ADVANCED NUCLEAR POWER REACTORS. Appendices-SAFETY-RELATED NUCLEAR PLANT STRUCTURES.

ADLER.M.V.

RUREG 4 JIDANC FOR A LARGE TABLETOP EXERCISE FOR A NUREG/CR-6263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR POWER PLANTS Candidate Guidehnes. Technical Basis And Research I

Needs Executive Summary.

I ALLEN,K.

NUREG/CR-6263 V02: HIGH INTEGHITY SOFTWARE FOR NUCLEAR l

NUREG 1516 DAFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL POWER PLANTS Candidate Guidehnen. Technical Basis And Research j

SAFETY PROGRAMS AT MEDICAL FACILITIES Draft Report For Com.

Needs Main Report.

ment BAKER,D.A.

ALLEN,M.D.

NUREG/CR-2650 Vf 3. DOSE COMMITMENTS DUE 7" AADIOACTIVE j

NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY RELEASES FROM NUCLEAR POWER PLANT SIES IN 1M DIRECT CONTAINMENT HEATING IN SURRY.

B A K E R.D.C.

ALLISON,C.M.

NUREG/CR-6150 VOI:

SCDAP/RELAPS/ MOD 31 CODE NUREG/CR-5975 RO1: INCENTIVE REGULATION OF INVESTOR-j OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULA-MANUAL _Interiace Theory NUREG/CR-6150 V02.

SCDAP/RELAP/ MOD 3.1 CODE TORS.

1 MANUAL Damage Progression Model Theory.

NUREG/CR-6150 V03 SCDAP/RELAP5/ MOD 3.1 CODE BAKER,T.L MANUAL. User's Guide And input Manual NUREG/CR-6046. ALERTNESS. PERFORMANCE, AND OFF-DUTY NUREG/CR-6150 V04:

SCDAP/RELAPS/ MOD 3.1 CODE SLEEP ON 8-HOUR AND 12-HOUR NIGHT SHIFTS IN A SIMULATED MANUAL.MATPRO A Library Of Matenals Properties For Light-Water.

CONTINUOUS OPERATIONS CONTROL ROOM SETTING.

Reactor Accident Analysis 1

NUREG/CR-6150 V05:

SCDAP/RELAPS/ MOD 31 CODE BANKS,W.W.

MANUALDevelopmental Assessment NUREG/CR4323: RELATIVE RISK ANALYSIS IN REGULATING THE l

CATENSACH T.J.

cat NUREG/CR-6323: RELATIVE RISK ANALYSIS IN REGULATING THE NUREG/CR-6324. QUALITY ASSURANCE FOR GAMMA KNIVES.

USE OF RADIATION-EMITTING MEDICAL DEVICES.A Prehminary Ap-phcation.

BARCHI,T.

NUREG-1415 V07 NO2: OFFICE OF THE INSPECTOR NURE R 10 AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO-fg PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU' NUREG 14'15 V08 N01: OFFICE OF THHE INSPECTOR CLEAR ACCIDENT

• GENERAL. Semiannual Report To Congress,Apnl 1,1995 - September 30,19 %

ANANDALINGAM,0.

WUREG/CR4349' COST BENEFIT CONSIDERATIONS IN REGULA-BARRIERE,M.T.

i NUREG/CR-6265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN

^

b'b' RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF ANKRUM,A.R.

COMMISSION AND DEPENDENCIES.

WUREG/CR4382: COMPARISONS OF ASTM STANDARDS CITED IN THE NRC STANDARD REVIEW PLAN.NUREG-0800, AND RELATED BASS,BA DOCUMENTS.

NUREG/CR4385: COMPARISONS OF ANS.ASME,AWS, AND NFPA NUREG/CR4273: BIAXIAL LOADING EFFECTS ON FRACTURE TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL STANDARDS CITED IN THE NRC STANDARD REVIEW PLAN.NUREG 0800. AND RELATED DOCUMENTS.

BATES,E.F.

NUREG/CR4386: COMPARISONS OF ANSI STANDARDS CITED IN THE NRC STANDARD REVIEW PLAN.NUREG-0800, AND RELATED NUREG-1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A NUCLEAR POWER PLANT.

DOCUMENTS.

BAYLESS,P.D.

ASHAR,H.

NUREG 1522. ASSESSMENT OF INSERVICE CONDITIONS OF NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION SAFETY RELATED NUCLEAR PLANT STRUCTURES.

STUDIES AT THE INEL 47

48 PCrsonal Author Index BECK,S.T.

BLEY,0.C.

NUREG/CR 6116 V10- SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN HANDS-ON INTEGRATED RELIABILITY EVALUATONS (SAPHIRE)

RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF VERSION 5.0. Data Loading Manual.

COMMISSION AND DEPENDENCIES.

BEHLING.H.

BLOCK,R.C.

NUREG/CR4310: AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO-NUREG/CP-0142 VOI: PROCEEDINGS OF THE 7TH INTERNATONAL PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS CLEAR ACCIDENT.

(NURETH-7) Sessons 1-5.

NUREG/CP 0142 V02: PROCEEDINGS OF THE 7TH INTERNATIONAL NU

/CR4310: AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO-p 7) g PHYLAXi3 FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-NUREG/CP 0142 V03: PROCEEDINGS OF THE 7TH INTERNATIONAL CLEAR ACCOENT.

MEETING ON f4UCLEAR REACTOR THERMAL-HYDRAULICS BELES,RJ-(NURETH 7).Sessons 12-16.

NUREG/CP 0142 V04: PROCEEDINGS OF THE 7TH INTERNATIONAL NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS DAMAGE ACCIDENTS:1994 A STATUS REPORT. Main Report And Ap-(NURETH 7).Sessons 17-24.

pendices A-H.

NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE BOHLA NDER.K.L DAMAGE ACCIDENTS:1994 A STATUS REPORT. Appendix 1.

NUREG/CR-5973 R02: CODES AND STANDARDS AND OTHER GUID.

BENAC DJ ANCE CITED IN REGULATORY DOCUMENTS.

NUREG/CR-6074 V04. SEALED SOURCE AND DEVICE DESIGN NUREG/CR4382: COMPARISONS OF ASTM STANDARDS CITED IN SAFETY TESTING. Technical Report On The Findings Of Task 4.Inves.

THE NRC STANDARD REVIEW PLAN,NUREG4800, AND RELATED N E /C 407 V SE L RC NO DEVICE DESIGN NU G/ 4 5: COMPARISONS OF ANS,ASME.AWS, AND NFPA SAFETY TESTING. Technical Report On The Findings Of Task 4.Inves.

STANDARDS CITED IN THE NRC STANDARD REVIEW tigation Of Failed Radcactive Stainless Steel Troxler Gauges.

P

.N R G-0800 A ELA E UMENT BERGERON,K.D.

THL' NRC STANDARD F1 e EW PLAN,NUREG-0800, AND RELATED NUREG/CR4109: THE PROBABILITY OF CONTAINMENT FAILURE BY 00Ct.iMENTS.

DIRECT CONTAINMENT HEATING IN SURRY.

BONANO,EJ.

BERGMAN.V.L NUREG/CR4351. REVIEW OF SCENARIO SELECTION APPROACHES NUREG/CR-6184: SEPARATE EFFECTS TESTING AND ANALYSES TO FOR PERFORMANCE ASSESSMENT OF HIGH-LEVEL WASTE RE-INVESTIGATE LINER TEARING OF THE 1:16-SOALE REINFORCED POSITORIES AND RELATED ISSUES.

CONCRETE CONTAINMENT BUILDING.

BOURDON.S M.

BERMUDEZ,H.

NUREG/CR4150 V05:

SCDAP/RELAP5/ MOD 3.1 CODE NUREG 1516 DAFT FC: MANAGEMENT OF RAD;OACTIVE MATERIAL MANUALDevelopmental Assessrnent SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-ment.

BRAMWELL.D.

NUREG/CR-6100: GATE VALVE AND MOTOR-DPERATOR RESEARCH BERNA,G.A.

FINDINGS' NUREG/CR4150 V01:

SCDAP/RELAPS/ MOD 3.1 CODE MANUAL. Interface Theory.

BREY,R.R.

NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE NUREG/CR4256 V02: FIELD LYSIMETER INVESTIGATIONS - TEST NUREG/ 415 VE S O

/R 5/ MOD 31 CODE RESULTS. Low-Levet Waste Data Base Development Program: Test MANUALUser's Guide And input Manual.

Results For Fiscal Years 1990,1991,1992, And 1993.

NUREG/CR-6150 V04:

SCDAP/RELAP5/ MOD 3.1 CODE M

MANUALMATPRO-A Library Of Matenals Properties For Light Water-fC -6330: RESULTS OF REGULATORY IMPACT SURVEY OF NU C 4150 SCDAP/RELAPS/ MOD 31 CODE INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE MANUALDevelopmental Assessment OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

BIER 8CHBACH,M.C BRICKSTAD,B.

NUREG/CR-5884 V01: REVISED ANALYSIS OF DECOMMISSIONING NUREG/CR-6300: REFINEMENT AND EVALUATON OF CRACK-OPEN.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL STATON. Effects Of Current Regulatory And Other Considerations On CRACKS IN PIPES.

The FMencial Insurance Requirements Of The Decommissoning Rule BRIDEAU,J.

NURE5'CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING NUREG/CR-6224: PARAMETRIC STUDY OF THE POTENTIAL FOR

/

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED STATION.Ettects Of Current Regulatory And Other Consideratons On DEBRIS.

The Financial Insurance Requirements Of The Decommissioning Rule BROWN.EJ.

NU d'/CR4054: ESTIMATING PRESSURIZED WATER REACTOR DE-NUREG/CP-0146 PHOCEEDINGS OF THE WORKSHOP ON GATE COMMISSIONING COSTS. A User's Manual For The PWR Cost Esti-VALVE PRESSURE LOCKING AND THERMAL BINDING.

mating Computer Program (CECP) Software.

BROWN,T.D.

geNINNG P.

NU9EG/CR-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC.

NUREG/CR4114 V02: AUXILIARY ANALYSES IN SUPPORT OF PER-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-LEVEL GRAND GULF UNIT 1. Evaluation Of Severe Accident Risks For Plant WASTE FACILITY.Two-Phase Flow And Contaminant Transport in Un.

Operational State 5 Dunng A Refuehng Outage Main Report And Ap-saturated Soils With Application To Low-Level Radcactive Waste Dis-NU

/ R4143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT BLACK.C.L GRAND GULF, UNIT 1.Evaluaten Of Severe Accident Risks For Plant NUREG/CR4312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN-Operationai State 5 Dunng A Refueling Outage Supporting MELCOR SORS.

Calculations.

BLACK.S.C.

BROWN,W.S.

NUREG-1526: LESSONS LEARNED FROM EARLY IMPLEMENTATON NUREG 0700 RO1 DFC: HUMAN-SYSTEM INTERFACE DESIGN OF THE MAINTENANCE RULE AT NINE NUCLEAR POWER PLANTS.

REVIEW GUIDELINE. Draft Report For Comment

P:rsonil Auth:r indIx 49 NUREG-1507 DRFT FC: MINIMUM DETECTABLE CONCENTRATIONS CHAM 8ERS,R.

WITH TYPICAL RADIATION SURVEY #NSTRUMENTS FOR VARIOUS NUREGICR4150 V04:

SCDAP/RELAP5/ MOD 3.1 CODE CONTAMINANTS AND FIELD CONDITIONS. Draft Report For Com-MANUAL.MATPRO-A Library Of Matenals Properties For Leght-Water-ment.

Reactor Accident Analysis.

SROWNSON.D.A.

CHARAK,L NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION NUREG/CR4315. CANDU REACTORS. THEIR REGULATION IN STUDIES AT THE INEL CANADA, AND THE IDENTIFICATION OF RELEVANT NRC SAFETY ISSUES.

BRUST,F.W.

NUREG/CR 6235. ASSESSMENT OF SHORT THROUGH-WALL CIR-CHEN.G.

CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis, March NUREG/CR-6308. AN OVERVIEW OF INSTABILITY AND FINGERING 1990 December 1994.

DURING IMMISCIBLE FLUID FLOW IN POROUS AND FRACTURED NUREG/CR4298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-MEDIA.

TIALLY SURFACE-CRACKED PIPE.

NUREG/CR-6299: EFFECTS OF TOUGHNESS ANISOTROPY AND CHEN,P.

COMB!NED TENSION. TORSION. AND BENDING LOADS ON FRAC-NUREG-1493: PERFORMANCE-BASED CONTAINMENT LEAK. TEST TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

PROGRAM. Final Report.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK OPEN-NUREG 1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-ING AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL TEST PROGRAM. Draft Report For Commere CRACKS IN PIPES.

CHEN,T.F.

BRYSON).W.

NUREG/CR-6242; CASKS (COMPUTER ANALYSIS OF STORAGE NUREG/CR4273: BIAXIAL LOADING EFFECTS ON FRACTURE CASKS)- A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL.

STORAGE CASK DESIGN REVIEW User's Manual To Version 1b (In-SUMPAS.S.E.

'"U 9'**

NUREG/CR-8322: BUCKLING ANALYSIS OF SPENT FUEL BASKET.

CHENG,T.C.

NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE NU G 1 65: ACCIDENT SOURCE TERMS FOR LIGHT WATER NU-A Damage Wession Wel TW CLEAR POWER PLANTS' CHIN,E.

NUREG/CR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE CALLAN,J.R.

AFTERLOADING BRACHYTHERAPY. Human Error And Cntcal Tasks NUREG/CR4125 VOI: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY. Human Error And Cntcal Tasks in Remote Afterloading Brachytherapy And Approaches For improved in Remote Art ing Brachytherapy And Approaches For improved NU CR 6 2 V : HUMAN FACTORS EVALUATION OF REMOTE NU EG/CR4125 V02: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY. Function And Task Analyris.

NUREG/CR4125 V03: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY Supporting Analyses Of Human-AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human.

System lnterfaces. Procedures And Practces,Tranng And Organirahon.

al Practces And Procedures.

System Interfaces, Procedures And Practces,Tranng And Organization, al Practices And Procedures.

CHOI,Y.H.

i NUREG/CR4298: FRACTURE BEHAvlOR OF SHORT CIRCUMFEREN-CALLEY,M.5.

l NUREG/CR4116 V09: SYSTEMS ANALYSIS PROGRAMS FOR TIALLY SURFACE-CRACKED PIPE-NUREG/CR-6300: REFINEMENT AND EVALUATION OF CRACK OPEN-HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL VERSION 5.0.Venication And Validation (V&V) Manual.

CRACKS IN PIPES.

CAMP 8 ELL,P.L NUREG 1482: GUIDELINES FOR INSERVICE TESTING AT NUCLEAR CHOPRA,0.K.

NUREG/CR 4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN POWER PLANTS.

LIGHT WATER REACTORS. Semiannual Report, October 1993

• March CAMPSELL V.

1994.

NUREG-1516 DAFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL NUREG/CR-4661 V19: ENVIRONMENT ALLY ASSISTED CRACKING IN SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com.

LIGHT WATER REACTORS Semaannual Reoort.Aprd-September 1994.

NUREG/CR4275: MECHANICAL PROPERTIES OF THERMALLY AGED ment.

CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-CAMPER,LW.

PONENTS.

NUREG 1516 DRFT FC. MANAGEMENT OF RADIOACTIVE MATERIAL NUREG/CR 6335; FATIGUE STRAIN-LIFE BEHAVIOR OF CARBON SAFETY PROGRAMS AT MEDICAL FACILITIES Draft Report For Com-AND LOW-ALLOY STEELS, AUSTENITIC STAINLESS STEELS, AND ALLOY 600 IN LWR ENVIRONMENTS.

c.ient.

CAPLIN,J.

CHOWDHURY,A.H.

NUREG/CR4307:

SUMMARY

OF COMMENTS RECEIVED AT WORK.

NUREG/CR4283: FIELD SITE INVESTIGATION: EFFECT OF MINE SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO SEISMICITY ON GROUNDWATER HYDROLOGY.

FACIUTATE PUBLIC PARTICIPATION IN DECOMMISSIONING CASES.

CHU,T.L NUREG/CR 6144 VO1: EVALUATION OF POTENTIAL SEVERE ACCl-CAPPS E.L DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR4116 V09: SYSTEMS ANALYSIS PROGRAMS FOR SURRY, UNIT 1. Summary Of Results.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

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

CHUNG,H.M.

NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN CASADA.D.A-LIGHT WATER REACTORS. Semiannual Report,0ctober 1993. March NUREG/CR 6089: DETECTION OF PUMP DEGRADATION.

1994.

NUREG/CR 4667 V19: ENVIRONMENTALLY ASSISTED CRACKING IN CELIA,M.A.

UGHT WATER REACTORS. Semsannual Report April-September 1994.

NUREG/CR-6114 V02. AUXILIARY ANALYSES IN SUPPORT OF PER-FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW LEVEL CLARK,R.E.

WASTE FACILITY.Two-Phase Flow And Contaminant Transport in Un-NUREG/CR-6116 V06. SYSTEMS ANALYSIS PROGRAMS FOR saturated Sods With Appleat on To Low-Level Radioachve Waste Dis-HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE) i posal.

VERSION 5 0.Grapheal Evaluabon Module (GEM) Reference Manual.

1 I

50 Personal Author Index CLEAVES D.

NUREG/CR-5591 VOS N2: HEAVY SECTION STEEL IRRADIATION NUREG/CR4263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR PROGRAM Progress Report For Apnl 1994 Through September 1994.

POWER PLANTS Canddate Gudelines. Technical Basis And Research NUREG/CR-5591 V06 N1: HEAVY-SECTION STEEL IRRADIATION Needs Executive Summary PROGRAM. Semiannual Progress Report For October 1994 Through NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR March 1995.

POWER PLANTS Canddate Gudolines. Technical Basis And Research Needs Main Report.

CORYELL.E W.

NUREG/CR-6150 VOI:

SCDAP/RELAP5/ MOD 3.1 CODE CLETCHER,J.W.

MANUALinterface Theory.

NUREG/CR.4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR 6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE DAMAGE ACCIDENTS 1994 A STATUS REPORT Main Report And Ap-MANUAL. Damage Progression Model Theory, pendices A-H.

NUREG/CR-6150 V03:

SCDAP/RELAP5/ MOD 3.1 CODE NUREG/CR 4614 V22: PRECURSORS TO POTENTIAL SEVERE CORE MANUALUser's Gude And input Manual DAMAGE ACCIDENTS 1994 A STATUS REPORT.Appendin I.

NUREG/CR-6150 V04.

SCDAP/RELAPS/ MOD 3.1 CODE MANUALMATPRO-A Library Of Matenals Properties For Light-Water.

CouLENTZ.L Reactor Accujent Analyss.

NUREG 1525: ASSESSMENT OF THE NRC ENFORCEMENT PRO-NUREG/CR-6150 V05:

SCDAP/RELAP5/ MOD 3.1 CODE GRAM MANUALDevelopmental Assessment.

COHEN,H.

COX,D.F.

)

NUREG/CR-6263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR NUREG/CR-5857: AGING OF TURBINE DRIVES FOR SAFETY RELAT.

I POWER PLANTS Canddate Gudehnes, Technical Bases And Research ED PUMPS IN NUCLEAR POWER PLANTS Needs Executive Summa NUREG/CR-6016: AGING AND SERVICE WEAR OF AIR-OPERATED NUREG/CR-6263 V02: Hi H INTEGRITY SOFTWARE FOR NUCLEAR -

VALVES USED IN SAFETY RELATED SYSTEMS AT NUCLEAR POWER PLANTS Canddate Guidehnes, Technical Basis And Research POWER PLANTS.

l Needs Main Report.

NUREG/CR 6192: AGING AND SERVICE WEAR OF SPRING-LOADED

)

PRESSURE RELIEF VALVES USED IN SAFETY-RELATED SYSTEMS COLE.R.K.

AT NUCLEAR POWER PLANTS.

a NUREG/CR-6119 V01: MELCOR COMPUTER CODE MANUALS Pnmer And User's Gudes. Version 1.8.3 September 1994.

CUNNINGHAM,M.

NUREG/CR-6119 V02:

MELCOR COMPUTER CODE NUREG/CR-5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER MANUALS Reference Manuals.Versson 1.8.3 September 1994.

INDUSTRY. Annual Summary Of Program Performance Reports CY COLLINS,J.L NUREG/CR-6318. DATA

SUMMARY

REPORT FOR FISSION PRODUCT CYBULSKIS,P.

RELEASE TEST VI 7.

NUREG-1493: PERFORMANCE-BASED CONTAINMENT LEAK. TEST PROGRAM Final Report.

CONGER,R-NUREG 1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK-NUREG/CR-6330: RESULTS OF REGULATORY IMPACT SURVEY OF TEST PROGRAM Draft Report For Comment.

INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

DAVIS,E.C.

NUREG/CR-6256 V01: FIELD LYSIMETER INVESTIGATIONS TEST CONLEY,D.A.

RESULTS. Low-Level Waste Data Base Program. Test Results For NUREG/CR-5402: AGING STUDY OF BOILING WATER REACTOR Fiscal Years 1986,1987,1988, And 1989 HIGH PRESSURE INJECTION SYSTEMS.

DAVIS K.L CONVERSE.S.A.

NUREG/CR-6150 V01:

SCDAP/RELAPS/ MOD 3.1 CODE NUREG/CR-6398: EVALUATION OF THE COMPUTERIZED PROCE*

MANUAL. Interface Theory DURES MANUAL ll (COPMA-fl).

NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE COOKE,R.M.

MANUALDamage Progression Model T NUREG/CR-6150 V03.

SCDAP/ RELA 5/ MOD 31 CODE 1

NUREG/CR 6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE MANUAL. User's Guide And in t Manual.

UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty NUREG/CR 6150 V04:

DAP/RELAPS/ MOD 3.1 CODE Assessment. Main Report.

MANUAL.MATPRO.A Library Of Matertals Properties For Light-Water-NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE Reactor Accident Analysis.

UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty NUREG/CR-6150 V05:

SCDAP/RELAPS/ MOD 3.1 CODE Assessment. Appendices A And B.

MANUALDevelopmental Assessrnent.

NUREG/CR-6244 V03. PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS Dispersion and Deposition Uncertainty DELICATE,W.S.

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

NUREG/CR-6287: MANAGEMENT CONCEPTS AND SAFETY APPLICA-TIONS FOR NUCLEAR FUEL FACILITIES.

NUREG/CR-6265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN DEWALL,K.G.

RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF NUREG/CR-6100: GATE VALVE AND MOTOR OPERATOR RESEARCH COMMISSION AND DEPENDENCIES.

FINDINGS.

COPINGER,D.A.

DE Y,M.

NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE NUREG 1493: PERFORMANCE BASED CONTAINMENT LEAK. TEST DAMAGE ACCIDENTS 1994 A STATUS REPORT. Main Report And A -

PROGRAM Final Report.

P pendices A-H.

NUREG 1493 DFC: PERFORMANCEEASED CONTAINMENT LEAK.

NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE TEST PROGRAM. Draft Report For Comment.

DAMAGE ACCIDENTS 1994 A STATUS REPORT. Appendix 1.

DINSMORE,0.

CORREIA,R.P.

NUREG/CR 6293 V01: VERIFICATION AND VALIDATION GUIDELINES NUREG 1526: LESSONS LEARNED FROM EARLY IMPLEMENTATION FOR HIGH INTEGRITY SYSTEMS Main Report.

OF THE MAINTENANCE RULE AT NINE NUCLEAR POWER PLANTS.

NUREG/CR4293 V02: VERIFICATIONN AND VALIDATION GUIDE-LINES FOR HIGH INTEGRITY SYSTEMS Appendicos A-0.

NUREG/CR 5591 V03.

HEAVY SECTION STEEL IRRADIATION DO8SE,C.A.

PROGRAM. Progress Report For October 1991 September 1992 NUREG/CR 6285: SEVERE ACCIDENT NATURAL CIRCULATION NUREG/CR 5591 V04 N2: HEAVY,SECTION STEEL IRRADIATION STUDIES AT THE INEL PROGRAM Semiannual Progress Report For AprSSeptember 1993 NUREG/CR-5591 VOS N1; HEAVY SECTION STEEL IRRADIATION DODDS.R.H.

PROGRAM Semiannual Progress Report For September 1993 Through NUREG/CR 6191: SIZE AND DEFORMATION LIMITS TO MAINTAIN March 1994.

CONSTRAINT IN K(IC) AND J(C) TESTING OF BEND SPECIMENS.

i Personal Author Index 51 NUREG/CR-6259: CONSTRAINT EFFECTS ON FRACTURE INITIATION NUREG/CP-0142 V04 PROCEEDINGS OF THE 7TH INTERNATIONAL LOADS IN HSST WIDE-PLATE TESTS.

MEETING ON NUCLEAR REACTOR THERMAL. HYDRAULICS (NURETH-7) Sessions 17-24.

NUREG/CR4348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION FERRE LL,C.M.

IN PARTIALLY SATURATED POROUS MEDIA.

NUREG 1465: ACCIDENT SOURCE TERMS FOR LIGHT WATER NU-DOLAN.S.W.

CLEAR POWER PLANTS.

NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE FIELD 1 GE IDENTS:1994 A STATUS REPORT. Main Report And Ap-NUREG/CR 4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE INDUSTRY. Annual Summary Of Program Performance Reports CY DAMAGE ACCIDENTS.1994 A STATUS REPORT. Appendix 1.

1994-DOTY,K.

FINDLAY,M.W.

NUREG/CR-2907 V13 RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR-6256 VOI: FIELD LYSIMETER INVESTIGATIONS - TEST NUCLEAR POWER PLANTS. Annual Report 1992.

RESULTS Low-Level Waste Data Base Program. Test Results For NUREG/CR-2907 V14: RADIOACTIVE MATERIALS RELEASED FROM Fiscal Years 1986,1987,1988 And 1989.

NUCLEAR POWER PLANTS.

FINEMAN.C.F.

DUKELOW,J.S.

NUREG/CR-6355: A LIMITED ASSESSMENT OF THE ASEP HUMAN NUREG/CR-5462: AGING STUDY OF BOILING WATER REACTOR RELIABILITY ANALYSIS PROCEDURE USING SIMULATOR EXAMI-HIGH PRESSURE INJECTION SYSTEMS.

NATION RESULTS.

FIRST M.W.

DUR8IN.N.

NUREG/CP 0141: PROCEEDINGS OF THE 23RD DOE /NRC NUCLEAR NUREG/CR-5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER AIR CLEANING CONFERENCE. Held in Buffalo.New York July 25-INDUSTRY. Annual Summary Of Program Performance Reports CY 28,1994 1994.

FISCHER,LE.

EASON,E.D.

NUREG/CR-6323: RELATIVE RISK ANALYSIS IN REGULATING THE NUREG/CR.6327. MODELS FOR EMBRITTLEMENT RECOVERY DUE USE OF RADIATION-EMITTING MEDICAL DEVICES.A Preliminary Ap.

TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

plication.

NUREG/CR4324: OUALITY ASSURANCE FOR GAMMA KNIVES.

EASTERLY,J.C.

NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF REMOTE E

AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-R R 5535 VOS R 1.

RELAP5/ MOD 3 CODE MANUALUser's System Interfaces. Procedures And Practices, Training And Organization-al Practices And Procedures Guidehne.

E8ERT,D.D.

FORESTER,J.

NUREG/CR-6257; CANDU 3 TRANSIENT ANALYSIS USING ATOMIC NUREG/CR4143 V06 Pt. EVALUATION OF POTENTIAL SEVERE AC-ENERGY OF CANADA LTD CODES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1 Evaluation Of Severe Accident Risks For Plant P

pe[ce N E CR 5462: AGING STUDY OF BOILING WATER REACTOR HIGH PRESSURE INJECTION SYSTEMS.

FOWLER,R.D.

EISENBERG,N.A'NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE NUREG/CR4116 VIO: SYSTEMS ANALYSIS PROGRAMS FOR NUREG-1464:

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

2. Development Of Capabilities For Review Of A Performance assess, ment For A High-Level Waste Repository.

VER$10N 5.0. Data Loading Manual.

EISNER.H.

FOX,P.8.

NUREG/CR-6287. MANAGEMENT CONCEPTS AND SAFETY APPLICA NUREG/CR 6284: CRITICALITY SAFETY CRITERIA FOR LICENSE TIONS FOR NUCLEAR FUEL FACILITIES.

REVIEW OF LOW-LEVEL WASTE FACILITIES.

EMRIT,R.

FRANCINI,8.

NUREG4933 S18: A PRIORITIZATION OF f,ENERIC SAFETY ISSUES NUREG/CR4235: ASSESSMENT OF SHORT THROUGH WALL CIR-NUREG-0933 S19. A PRIORITt2ATION OF 3ENERIC SAFETY ISSUES.

CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis, March 1990 December 1994.

ERCK,R.A.

NUREG/CR.4667 V18. ENVIRONMENTAL.Y ASSISTED CRACKING IN NA LIGHT WATER REACTORS Semiannual Peport. October 1993 March U G/ 4297: FRACTURE EVALUATIONS OF FUSION LINE CRACKS IN NUCLEAR PIPE BIMETALLIC WELDS.

FADDEN,M.

NUREG/CR4298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-NUREG-0525 V02 R03-SAFEGUARDS

SUMMARY

EVENT LIST TIALLY SURFACE CRACKED PIPE.

(SSEL). January 1,1990 Through December 31,1994 FREDRICKSON,P.

FARMER,J.P.

NUREG-1526 RECONSTITUTION OF THE MANUAL CHAPTER 2512 NUREG/CR4312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN-CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT 1.

SORS.

FULLER,M.

FAUYER D.N*

NUREG-1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL NUREG 1444 Sot: SITE DECOMMISSIONING MANAGEMENT PLAN.

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

FEINER,F.

NUREG/CP 0142 V01: PROCEEDINGS OF THE 7TH INTERNATIONAL GALYEAN,W.J.

MEETING ON NUCLEAR REACTOR THFPMAL HYDRAULICS NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR (NURETH-7) Sessions 15 NUREG/CP 0142 V02: PROCEEDINGS OF THE 7TH NTERNATIONAL HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

MEETING ON NUCLEAR REACTOR THERb AL. HYDRAULICS VERSION 5.0.Venfication And Vahdation (V&V) Manual J

(NURETH 7) Sessions 611.

NUREG/CP-0142 V03: PROCEEDtNGS OF THE 7TH INTERNATIONAL GANT,K.S.

MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS NUREG 1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A (NURETH 7). Sessions 1216.

NUCLEAR POWER PLANT.

52 Personal Author index GAVE NDA.D.J.

PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-VI NUREG/CR-4667 V19 ENVIRONMENTALLY ASSISTED CRACKING IN HUCLEAR DATA.

LIGHT WATER REACTORS Semsannual Report.Apni-September 1994.

GREENE,R.H.

GERHARD,M.A.

NUREG/CR4089: DETECTION OF PUMP DEGRADATION.

NUREG/CR-5657: AUTOCASK (AUTOMATIC GENERATION OF 3-D CASK MODELS) A Mcrocomputer Based System For Shipping Cask GRIFFIN,J.L Design Review Analysis.

NUREG 1527: NRC'S OBJECT ORIENTED SIMULATOR INSTRUCTOR NUREG/CR-6242: CASKS (COMPUTER ANALYSIS OF STORAGE STATION.

CASKS) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR STORAGE CASK DESIGN REVIEW User's Manual To Verson 1b (In-GRIFFIN.J.P.

ciuding Program Reference).

NUREG 1527: NRC'S OBJECT-ORIENTED SIMULATOR INSTRUCTOR STATION.

GHADIAll,N.

NUREG,CR-4599 VL4 N1: SHORT CRACKS IN PIPING AND PIPING GRODIN,M.

WELDS Seventh Program Repod March 1993 December 1994 NUREG-1415 V07 NO2; OFFICE OF THE INSPECTOR NUREG/CR 6004:. PROBABILISTIC PIPE FRACTURE EVALUATIONS GENERAL. Semiannual Report To Congress, October 1,1994 March FOR LEAK-RATE DETECTION APPLICATIONS.

31,1995 NUREG/CR-6235: ASSESSMENT OF SHORT THROUGH-WALL CIR' NUREG-14'15 V08 NOI: OFFICE OF THHE INSPECTOR CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis, March GENERAL. Semiannual Report To Congress,Apol 1.1995 - September 1990. December 1994 NUREG/CR4298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-

30. M95.

TIALLY SURFACE CRACKED PIPE'OUGHNESS ANISOTROPY GROSSMAN'N-NUREG/CR-6299' EFFECTS OF T AND NUREG-1519: SURFACE INTERACTIONS OF CESIUM AND BORIC COMBINED TENSION, TORSION. AND BENDING LOADS ON FRAC-j TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

ACID WITH STAINLESS STEEL.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK OPEN-G A ALYSES FOR CIRCUMFERENTIAL THROUGH-WALL NUREG CR 3 V02: GUIDELINES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-GILBERT,E.R.

AL SOFTWARE Survey And Assessment Of Conventonal Software NUREG/CR-6382: COMPARISONS OF ASTM STANDARDS CITED IN Venfcation And Validation Methods.

THE NRC STANDARD REVIEW PLAN.NUREG-0800, AND RELATED NUREG/CR 6316 V03. GUIDELINES FOR THE VEFSICATION AND DOCUMENTS.

VALIDATION OF EXPERT SYSTEM SOFTWARE AN ' CONVENTION-NUREG/CR4385: COMPARISONS OF ANS,ASME.AWS. AND NFPA AL SOFTWARE. Survey And Documentaten Of Expert System Venfca-STANDARDS CITED IN THE NRC STANDARD REVIEW ton And Vahdation Methodologies.

PLAN.NUREG-0800. AND RELATED DOCUMENTS.

NUREG/CR4386: COMPARISONS OF ANSI STANDARDS CITED IN GROVE E.J.

THE NRC STANDARD REVIEW PLAN.NUREG-0800, AND RELATED NUREG/CR-5954: EFFECT ON AGING ON PWR CHEMICAL AND DOCUMENTS.

VOLUME CONTROL SYSTEM GOGOLAK C.V.

GRUBER,E.E.

NUREG-1505 DAFT FC: A NONPARAMETRIC STATISTICAL METHOD-NUREG/CR 4667 V19: ENVIRONMENTALLY ASSISTED CRACKING IN OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-LIGHT WATER REACTORS Semiannual Report.Apni-September 1994.

COMMISSIONING SURVEYS Draft Report For Comment.

GRUSH,W.H.

GOLDIN,D.

NUREG/CR-6291 V01: NUCLEAR PLANT ANALYZER. installation NUREG 1493: PERFORMANCE BASED CONTAINMENT LEAK TEST Manual PROGRAM. Final Report.

NUREG/CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

NUREG 1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK

• ence Manual TEST PROGRAM. Draft Report For Comment.

NUREG/CR-6291 V03: NUCLEAR PLANT ANALYZER Computer Visual S stem Reference Manual.

13 V01: ROBUST, ACCURATE, AND NON-CONTACTING Manu '

VIBRATION MEASUREMENT SYSTEM Summary of Companson Meas-urements Of The Robust Laser interferometer And Typical Accelerome-GWYNNE,J.W.

ter Systems.

NUREG/CR4313 V02: ROBUST, ACCURATE. AND NON-CONTACTING NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks VIBRATION MEASUREMENT SYSTEMS Supplemental Appendees Presenting Companson Measurements Of The Robust Laser interfer.

in Remote Aftertoading Brachytherapy And Approaches For Improved System Performance.

ometer And Typcal Accelerometer Systems-NUREG/CR4125 V02: HUMAN FACTORS EVALUATION OF REMOTE GOOSSENS LH.J.

AFTERLOADING BRACHYTHERAPY. Function And Task Analysis NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE HAGEMEYER,D.

UNCERTAINTY ANALYSIS. Disperson And Deposition Uncedainty NUREG-0713 V15. OCCUPATIONAL RADIATION EXPOSURE AT COM-NUR 2

IOBABILISTIC ACCIDENT CONSEQUENCE MERCIAL NUCLEAR POWER REACTORS AND OTHER UNCERTAINTY ANALYSIS. Dsperson And Depositen Uncertainty FACILITIES.1993. Twenty Sixth Annual Report.

Assessment.Appendees A And B-NUREG/CR.6244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE HAGGARD,D L.

UNCERTAINTY ANALYSIS. Dsperson and Deposition Uncertainty NUREG/CR 4390: RADIOLOGICAL CHARACTERIZATION OF SPENT Assessment.Appendees C,D,E.F,G.H.

CONTROL ROD ASSEMBLIES.

GORE.B.F.

HAGRMAN,0.L NUREG/CR4355: A LIMITED ASSESSMENT OF THE ASEP HUMAN NUREG/CR4150 V02:

SCDAP/RELAP/ MOD 3.1 CODE LIA IL ANALYSIS PROCEDURE USING SIMULATOR EXAMI-NUREG 41 V04 CD /R 5/ MOD 3.1 CODE MANUAL.MATPRO-A Library Of Matenals Properties For Light-Water-GREEN,R.T.

Reactor Artdent Analysis.

NUREG/CR4348. THERMALLY DRIVEN MOISTURE REDISTRIBUTION IN PARTIALLY SATURATED POROUS MEDIA HAGRMAN,D.T, NUREG/CR4356: HYDRAULIC CHARACTERIZATION OF HYOROTH.

NUREG/CR4150 Vot:

SCDAP/RELAPS/ MOD 3.1 CODE ERMALLY ALTERED NOPAL TUFF, MANUALinterface Theory.

NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE j

QREENE.N.M.

MANUAL.Camage Progresson Model T NUREG/CR4214: PRODUCTION AND TESTING OF THE VITAMIN-06 NUREG/CR-6150 V03:

SCDAP/ RELA / MOD 3.1 CODE FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

MANUALUser's Guide And input Manual

Personal Author index 53 NUREG/CR4150 V04:

SCDAP/RELAP5/ MOD 11 CODE HECHT,M.

MANUALMATPRO-A Labrary Of Matenals Properties For Light-Water-NUREG/CR4293 V01: VERIFICATION AND VALIDATION GUIDELINES Reactor Accident Analysis.

FOR HIGH INTEGRITY SYSTEMS. Main Report.

NUREG/CR-6150 V05:

SCOAP/RELAP5/ MOD 3.1 CODE NUREG/CR-6293 V02: VERIFICATIONN AND VALIDATION GUIDE-MANUAL. Developmental Assessment LINES FOR HIGH INTEGRITY SYSTEMS.Appendees A 0.

HALL,0.

HECHT,S.

NUREG/G40014; BALDCYPRESS TREE RING ELEMENTAL CONCEN-TRATIONS AT REELFOOT LAKE. TENNESSEE.FROM AD 1795 TO NUREG/CR4293 V01: VERIFICATION AND VALIDATION GUIDELINES AD 1820.

FOR HIGH INTEGRITY SYSTEMS Main Report.

NUREG/CR4293 V02: VERIFICATIONN AND VALIDATION GUIDE-HAMILTON,M.A.

LINES FOR HIGH INTEGRITY SYSTEMS. Appendices A-D.

NUREG/CR-6188 V02: MICROBIAL DEGRADATION OF LOW-LEVEL RADIOACTIVE WASTE. Annual Report For FY 1994.

HELD,P.R NUREG/CR-6251; STAINLESS STEEL SUBMERGED ARC WELD HAMPTON,N.L FUSION LINE TOUGHNESS.

NUREG/CR-6150 V04:

SCDAP/RELAP5/ MOD 31 CODE MANUAL MA AL ary Of Malenals Properties For Light-Water.

HE T j

CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-HARPER,F.T.

CIDENT CONSEOUENCE MODEL.

NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-Assessment Main Report.

. DENT CONSEQUENCE MODEL.

NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF UNCERTAINTY ANALYSIS. Dsperson And Deposition Uncertainty FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-Assensment Appendces A And B.

DENT CONSEQUENCE MODEL NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR 6244 V01: PROBABILIGTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dspersion and Depositon Uncertainty Assessment. Appendices C.D.E.F.G.H.

UNCERTAINTY ANALYSIS. Dsperson And Deposition Uncertainty Assessment. Main Report.

NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEOUENCE NUR G 0520 19: LICENSED OPERATING REACTORS STATUS SUM-

^

^

P ' " "

s ne ppe ce A B

MARY REPORT. Data As Of December 31,19941 Gray Book 1)

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE HASHEMlAN H M.

UNCERTAINTY ANALYSIS. Dsperson and Deposition Uncertainty WUREG/CR-6312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN, Assessment. Appendices C.D.E.F,G.H.

SORS NUREG/CR-6334 NEW SENSOR FOR MEASUREMENT OF LOW AIR HENDERSON,P.

FLOW VELOCITY Phase 1 Final Report.

NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL NUREG/CR-6343. ON-LINE TESTING OF CAllBRATION OF PROCESS SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-INSTRUMENTATION CHANNELS IN NUCLEAR POWER ment.

PLANTS Pha.ie !! Final Report.

HASHEMIAN M.

NUREG/CR-6277 V01: HUMAN FACTORS EVALUATION OF TELE-NUREG/CR-6334 NEW SENSOR FOR MEASUREMENT OF LOW AIR THERAPY. Identification Of Problems And Alternatsve Approaches.

FLOW VELOCITY. Phase l Final Report.

NUREG/CR4277 V02: HUMAN FACTORS EVALUATION OF HATTRUP,M.

TELETHERAPY. Function And Task Analysis.

NUREG/CR4277 V03: HUMAN FACTORS EVALUATION OF NUREG/CR 5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER INDUSTRY. Annual Summary Of Program Performance Reports CY TELETHERAPY. Human-System interfaces And Procedures.

NUREG/CR4277 V04: HUMAN FACTORS EVALUATION OF 1994-NUREG/CR4330: RESULTS OF REGULATORY IMPACT SURVEY OF TELETHERAPY. Training And Organizational Analysis.

NUREG/CR4277 V05: HUMAN FACTORS EVALUATION OF INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

TELETHERAPY. Literature Review.

HILTON LD.

NUREG R4316 V01: GUIDELINES FOR THE VERIFICATION AND

-6256 M M MWER WWWS TEST VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-S evel aste Data Base Wamlest NsuHs 6 AL SOFTWARE Fiscal Years 1986,.987,1988, And 1989.

NUREG/CR-6316' V02: GUIDELINES FOR THE VERIFICATION AND NUREG/CR4256 V02: FIELD LYSIMETER lNVESTIGATIONS. TEST VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-RESULTS Low-Level Waste Data Base Development Program Test AL SOFTWARE. Survey And Assessment Of Conventonal Software Results For Fiscal Years 1990,1991,1992. And 1993.

Venicaton And Vahdation Methods NINI' REG'/CR.4667 V19: ENVIRONMENTALLY ASSISTED CRA leUREG/CR-6316 V04: GUIDELINES FOR THE VERIFICATION AND A'O NU VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL SOFTWARE. Evaluation Of Knowledge Base Certifcation Methods.

LIGHT WATER REACTORS. Semiannual Report.Apni September 1994.

RUREG/CR4316 V05: GUIDELINES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-HODGE,5.A.

AL SOFTWARE. Rationale And Desenption Of V&V Guidehne Packages NUREG/CR4119 V01: MELCOR COMPUTER CODE MANUALS.Pnmer And Procedures And User's Guides. Version 1.8.3 September 1994.

NUREG/CR4316 V06: GUIDELINES FOR THE VERIFICATION AND NUREG/CR4119 V02:

MELCOR COMPUTER CODE VALIDATION OF EXPERT SYSTEM SOFTWARE AND COWENTION-MANUALS Reference Manuals.Verson 1.8.3 September 1994.

AL SOFTWARE.Vahdation Scenares.

MUREG/CR4316 V07: GUIDELINES FOR THE VERIFICATION AND HOFFMAN,C L.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

NUREG/CR-6116 V06: SYSTEMS ANALYSIS PROGRAMS FOR AL SOFTWARE. User's Manual.

HANDS-ON INTEGRATED RELIABluTY EVALUATIONS (SAPHIRE)

NUREG/CR-6316 V08. GUIDELINES FOR THE VERIFICATION AND VERSION 5.0.Graphcal Evaluation Module (GEM) Reference Manual.

VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AL SOFTWARE.Bibhography.

HOHORST,J.K.

NUREG/CR4150 V01.

SCDAP/RELAPS/ MOD 3.1 CODE HECHT,H.

MANUAL. interface Thoory.

NUREG/CR-6293 V01: VERIFICATION AND VALIDATION GUIDELINES NUREG/CR 6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE FOR HIGH INTEGRITY SYSTEMS Main Report.

MANUAL. Damage Progresson Model Theory.

NUREG/CR-6293 V02: VERIFICATIONN AND VALIDATION GUIDE.

NUREG/CR-6150 V03-SCDAP/RELAP5/ MOD 3.1 CODE LINES FOR HIGH INTEGRITY SYSTEMS Appendees A-D.

MANUALUser's Guede And input Manual.

v

.m 54 Personal Author index i

NUREG/CR4150 V04.

SCDAP/RELAP5/ MOD 3.1 CODE NUREG/CR4119 V02:

ME COR COMPUTER CODE

]

MANUAL.MATPRO-A Library Of Matenals Properties For Light-Water-MANUALS. Reference Manuals. Version 1.8.3 September 1994.

Reactor Accadent Analysis.

NUREG/CR4150 V05:

SCDAP/RELAP5/ MOD 3.1 CODE tDOINGS,F.A.

MANUAL. Developmental Assessment.

NUREG/CR-6074 V04: SEALED SOURCE AND DEVICE DESIGN SAFETY TESTING.Techncal Report On The Findmgs Of Task 4.inves-Naton ealed Soume W Paper W %sk NU EG/C 6313 V01: ROBUST, ACCURATE, AND NON-CONTACTING j

VIBRATION MEASUREMENT SYSTEM Summary of Companson Meas' INGERSOLL,D.T.

urements Of The Robust Laser interferometer And Typcal Accelerome-NUREG/CR-6214: PRODUCTION AND TESTING OF THE VITAMIN-06 E bb313 V02: ROBUST, ACCURATE. AND NON CONTACTING FINE GROUP AND THE BUGLE 93 BROAD. GROUP NEUTRON /

d N

P j

VfBRATION MEASUREMENT SYSTEMS Supplemental Appendees NUCLEAR DAT '

Presenting Companson Measurements Of The Robust Laser Interfer.

a ometer And Typcal Accelerometer Systems.

JASTROW,J.D.

7 NUREG/CR-5229 V07; FIELD LYSIMETER INVESTIGATIONS: LOW-HOPPER C.M.

NUREG/CR-6284. CRITICALITY SArETY CRITERIA FOR LICENSE LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR REVIEW OF LOW-LEVEL WASTE FACILITIES.

FISCAL YEAR 1994.Anrual Report.

NUREG/CR-6256 VOI: FIELD LYSIMETER INVESTIGATIONS - TEST HORA,S.C.

RESULTS. Low-Level Waste Data Base Program. Test Results For l

NUREG/CR 6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE Fiscal Years 1986,1987,1988 And 1989.

UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty NUREG/CR 6256 V02: FIELD LYSIMETER INVESTIGATIONS. TEST Assessment Main Report RESULTS Low-Level Waste Data Base Development Program: Test NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE Results For Fiscal Years 1990.1991,1992, And 1993.

UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty Assessment Appendces A And B.

JAUDON,J.

NUREG/CR-6244 V03. PROBABILISTIC ACCIDENT CONSEQUENCE NUREG-1528: RECONSTITUTION OF THE MANUAL CHAPTER 2512 UNCERTAINTY ANALYSIS. Dspersion and Deposition Uncertainty CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT 1.

Assessment.Appendees C.D.E.F,G.H HSIUNG,S-M NUREG/CR-6257; CANDU 3 TRANSIENT ANALYSIS USING ATOMIC NUREG/Cb6283. FIELD SITE INVESTIGATION: EFFECT OF MINE ENERGY OF CANADA LTD CODES.

SEISMICITY ON GROUNOWATER HYDROLOGY, HUSER D.

JO,J.

NUREG-1415 V07 NO2:

OFFICE OF THE INSPECTOR NUREG/CR-6144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC.

GENERALSemiannual Report To Congress. October 1,1994 - March CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT 31,1995.

SURRY, UNIT 1. Evaluation Of Severe Accident Risk Dunng Mid-Loop NUREG-1415 V08 N01: OFFICE OF THHE INSPECTOR Operatons Main Report.

GENERAL Semiannual Report To Congress,Apnl 1,1995 - September NUREG/CR 6144 V06 P2. EVALUATION OF POTENTIAL SEVERE AC-

)

30,1995.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1. Evaluation Of Severe Accident Risk Dunng Mid Loop HUFFERT,A.M-Operat ons Appendees.

NUREG-1505 DAFT FC: A NONPARAMETRIC STATISTICAL METHOD-OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-JOHNSEN,E.C, COMMISSIONING SURVEYS Draft Report For Comment-NUREG/CR-6150 V03-SCDAP/RELAPS/ MOD 3.1 CODE NUREG-1506 DRF' FC: MEASUREMENT METHODS FOR RADIOLOGl' MANUAL. User's Guide And input Manual.

CAL SURVEYS IN SUPPORT OF NEW DECOMMISSIONING NUREG/CR-6150 V05-SCDAP/RELAPS/ MOD 3.1 CODE MANUALDevelopmental Assessment.

Nt G

7 FT MN M DETECTABLE CONCENTRATIONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS JOHNSON,G.L.

CONTAMINANTS AND FIELD CONDITIONS Draft Report For Com-NUREG/CR 6242-CASKS (COMPUTER ANALYSIS OF STORAGE ment CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR HUGHES.K.

STORAGE CASK DESIGN REVIEW User's Manual To Version Ib (In-NUREG/CR-6330: RESULTS OF REGULATORY IMPACT SURVEY OF ciuding Program Reference)

INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE JOHN N,J.

OFFICE OF NUCLEAR MATERIALS SAFETY AND SAT MUARDS HUGHES.T.H.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR 4667 V19. ENVIRONMENTALLY ASSISTED CRACKING IN GRAND GULF, UNIT 1 Evaluaton Of Severe Accident Risks For Plant LIGHT WATER REACTORS Semiannual Report,Aptd-September 1994.

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

HUNTER,T.H.

NUREG/CR-6214. PRODUCTION AND TESTING OF THE VITAMIN-06 JOHNSON,J.C.

FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

NUREG/CR-6114 V02: AUXILIARY ANALYSES IN SUPPORT OF PER-PHOTON CROSS-SECTION LIBRAR!ES DERIVED FROM ENDF/B-Vi FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-LEVEL NUCLEAR DATA.

WASTE FACillTY,Two-Phase Flow And Contaminant Transport in Un-saturated Sods With Apphcation To Low-Level Radioactive Waste Dis-HUTCHINSON,J.W.

NUREG/CR.6264 V02: VALIDtTY LIMITS IN J-RESISTANCE CURVE p sal DETERMINATION.A Computahonal Approach To Ductde Crack Growth JOHNSON J.D.

Under Large Scale Yielding Conditons' NUREG/CR-6134 UNCERTAINTY AND SENSITIVITY ANALYSIS OF HYBERTSON,D.

CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-NUREG/CR-6263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR CIDENT CONSEQUENCE MODEL POWER PLANTS Candidate Guidehnes. Techncal Basis And Research NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF Needs Executive Summary EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CH-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR DENT CONSEOVENCE MODEL POWER PLANTS. Candidate Guidelines, Techrucal Basis And Research NUREG/CR-6136 UNCERTAINTY AND SENSITIVITY ANALYSIS OF Needs Mam Report.

FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-DENT CONSEQUENCE MODEL HYMAN,C.R.

NUREG/CR 6119 V01: MELCOR COMPUTER CODE MANUALS Pnmer JOHNSON,T,C.

And User's Guides. Verson 1.8.3 September 1994.

NUREG-1444 Set: SITE DECOMMISSIONING MANAGEMENT PLAN

.=. - --.

Personal Author index 55 JONES,E.D.

NUREG/CR-6277 V02: HUMAN FACTORS EVALUATION OF NUREG/CR-6323; RELATIVE RISK ANALYSIS IN REGULATING THE TELETHERAPY Function And Task Analysis.

USE OF RADIATION-EMITTING MEDICAL DEVICES.A Preliminary Ap-NUREG/CR-6277 V03-HUMAN FACTORS EVALUATION OF plication.

TELETHERAPY Human-System interfaces And Procedures.

NUREG/CR 6324. QUALITY ASSURANCE FOR GAMMA KNIVES NUREG/CR-6277 V04: HUMAN FACTORS EVALUATION OF TELETHERAPY.Trairung And Organaational Analysis.

JONES J.

NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL TELETHERAPY. Literature Review.

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

KEISLER J.

NUREG/CR-6335: FATIGUE STRAIN-LIFE BEHAVIOH OF CARBON JONES,J.A.

AND LOW-ALLOY STEELS, AUSTENITIC STAINLESS STEELS, AND NUREG/CR-6244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE ALLOY 600 IN LWR ENVIRONMENTS.

UNCERTAINTY ANALYSIS. Despersion And Deposition Uncertainty Assessment. Main Report KELLY,R.T.

NUREG/CR 6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE UNCERTAINTY ANALYSIS Dispersion And Deposition Uncertainty AFTERLOADING BRACHYTHERAPV. Human Error And Cntical Tasks Assessment. Appendices A And 8.

In Remote Afterloading Brachytherapy And Approaches For improved NUREG/CR 6244 V03. PROBABILISTIC ACCIDENT CONSEQUENCE System Performance.

UNCERTAINTY ANALYSIS. Dispersion and Depositon Uncertainty NUREG/CR-6125 V02: HUMAN FACTORS EVALUATION OF REMOTE Assessment. Appendices C.D.E.F,G t L AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

NUREG/CR4125 V03: HUMAN FACTORS EVALUATION OF REMOTE JONES.J.L AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Humark NUREG/CR 6116 V09-SYSTEMS ANALYSIS PROGRAMS FOR System Interfaces, Procedures And Practices. Training And Organuaton.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE) al Practices And Procedures.

VERSION 5 0.Venficaton And Vahdat on (V8V) Manual.

KHAN T.A.

JONES K.R.

NUREG/CP 0143: PROCEEDINGS OF THE THIRD INTERNATIONAL NUREG/CR-6285: SEVERE ACCIDENT NATURAL CIRCULATION WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR STUDIES AT THE INEL POWER PLANTS. Held At Hauppauge, Long Island, New York.

NUREG/CR-6291 V01: NUCLEAR PLANT ANALYZER Installaton NUREG/CR-3469 V08: OCCUPATIONAL DOSE REDUCTION AT NU.

Manual-CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-NUREG/CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer-ED READINGS IN RADIATION PROTECTION AND ALARA.

ence Manual.

NUREG/CR-6291 V03. NUCLEAR PLANT ANALYZER. Computer Visual KIER,P.H.

System Reference Manual-NUREG/CR-6315: CANDU REACTORS. THEIR REGULATION IN NUHEG/CR-6291 V04 - NUCLEAR PLANT ANALYZER. Programmer.s CANADA AND THE IDENTIFICATION OF RELEVANT NRC SAFETY Manual-ISSUES.

JONES,k M N NUREG/CR-6277 V01: HUMAN FACTORS EVALUATION OF TELE-gRE CR-6235; ASSESSMENT OF SHORT THROUGH-WALL CIR-NU E /CR 6 7 V2 FA T S EVA OF CUMFERENTIAL CRACKS IN PIPES.Expenments And Analysis, March TELETHERAPY. Function And Task Analysis 1990 - December 1991 NUREG/CR 6277 V03.

HUMAN FACTORS EVALUATION OF NUREG/CR-6298; FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-TELETHER APY. Human-System Inlenxes And Procedures.

TIALLY SURFACE-CRACKED PIPE.

NUREG/CR 6277 V04 HUMAN FACTORS EVALUATION OF KIM,1.S.

TELETHERAPY, Training And Organaatonal Analysis.

NUREG/CR-6277 V05 HUMAN FACTORS EVALUATION OF NUREG/CR 6141: HANDBOOK OF METHODS FOR RISK-BASED TELETHERAPY. Literature Review ANALYSES OF TECHNICAL SPECIFICATIONS.

JORDAN,W.C.

KINNEMAN,J.D.

NUREG/CR 6328. ADEOUACY OF THE 123 GROUP CROSS-SECTION NUREG-1444 S01: SITE DECOMMISSIONING MANAGEMENT PLAN LIBRARY FOR CRITICALITY ANALYSES OF WATER MODERATED KMETYK.LN.

URANIUM SYSTEMS.

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

JOY,0.R.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG-0430 V14.

LICENSED FUEL FACIUTY STATUS GRANO GULF UNIT 1.Evaluaton Of Severe Accident Risks For Plant REPORT Inventory Difference Data July 1,1993 - June 30,1994 (Gray Operational State 5 Dunng A Refuehng Outage. Main Report And Ap-1 Book 11) pendices.

NUREG 1065 R02-ACCEPTABLE CTANDARD FORMAT AND CON.

NUREG/CR-6143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC-TENT FOR THE FUNDAMENTAL NUCLEAR MATERIAL CONTROL CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT (FNMC) PLAN REQUIRED FOR LOW ENRICHED URANIUM FACILI.

GRAND GULF UNIT 1 Evaluation Of Severe Accident Risks For Plant TIES.

Operational State 5 Dunng A Refueling Outage. Supporting MELCOR Calculations.

KASSNER,T.F.

NUREG/CR-4667 V18-ENVIRONMENTALLY ASSISTED CRACKING IN KNUDSON.D L 1

LIGHT WATER REACTORS Semiannual Report. October 1993 - March NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY 1994 DIRECT CONTAINMENT HEATING IN SURRY.

NUREG/CR-4667 V19. ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS Semiannual Report,Apnl September 1994.

KOKAJKO.LE.

NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-GASUMOVIC,J.

TIONS REVIEW TEAM.

NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE 1

AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks KONZEK,G J.

NUREG/CR-5884 V01: REVISED ANALYSIS OF DECOMMISSIONtNG 1

in Remote Afterloading Brachytherapy And Approaches For improved FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER System Performance.

NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF REMOTE STATION EHects Of Current Regulatory And Other Considerations On AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-The Financial Insurance Requirements Of The Decommissoning Rule

]

System Interfaces. Procedures And Practices. Training And Organizaton-And....

al Practices And Procedures.

NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER CAYE.R.D.

STATION. Effects Of Current Regulatory And Other Consideratons On NUREG/CR-6277 V01: HUMAN FACTORS EVALUATION OF TELE-The Financial Insurance Requirements Of The Decommissioning Rule THERAPY Identification Of Problems And Alternative Approaches.

And-

56 Personal Author Index KDPPENHOEFER,K.

LEE,M.P.

NUREG/CR4191: SIZE AND DEFORMATION LIMITS TO MAINTA N NUREG 1464: NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE CONSTRAINT IN K(IC) AND J(C) TESTING OF BEND SPECIMENS.

2. Development Of Capabihties For Review Of A Performance Assess-ment For A High-Level Waste Repostory.

NUREG/CR 6310: AN ANALYSIS OF POTASSlUM IODIDE (KI) PRO-LEE.R.Y.

PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-NUREG-1465: ACCOENT SOURCE TERMS FOR LIGHT. WATER NU-CLEAR ACCIDENT.

CLEAR POWER PLANTS.

j I

KOVACH,LA.

LEPAGE.R.P.

NUREG/CP4147: PROCEEDINGS OF THE WORKSHOP ON THE ROLE NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE OF NATURAL ANALOGS IN GEOLOGIC DISPOSAL OF HIGH-LEVEL AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks NUCLEAR WASTE Held in San Antonio Texas. July 22-25,1991.

In Remote Afterloadng Brachytherapy And Approaches For improved System Performance.

KOZAK,M.W.

NUREG/CR-6125 V02: HUMAN FACTORS EVALUATION OF REMOTE NUREG/CR-5927 V02. EVALUATON OF A PERFORMANCE ASSESS-AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

MENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE WASTE NUREG/CR 6125 V03. HUMAN FACTORS EVALUATON OF REMOTE DISPOSAL FACILITIES Vehdation Needs-AFTERLOADING BRACHYTHERAPY.Supportng Analyses Of Human-System interfaces, Procedures And Practices, Training And Organization-al Practices And Procedures.

NUREG 6 A COMPILATION OF CURRENT REGULATIONS, STANDARDS. AND GUIDELINES IN REMOTE AFTERLOADING BRA-LEPEL,E.A.

CHYTHERAPY.

NUREG/CR-6390: RADIOLOGICAL CHARACTER 12ATION OF SPENT CONmOL ROD ASSEMES.

KRAAN.B.

NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE LICHTNER,P.C.

UNCERTAINTY ANALYSIS. Disperson And Depostion Uncertamty NUREG/CR4347; MULTI-PHASE REACTIVE TRANSPORT THEORY.

Assessment Main Report.

NUREG/CR-6244 V02 PROBABILISTIC ACCIDENT CONSEQUENCE LIE 8ERMAN.J.

UNCERTAINTY ANALYSIS Dispersion And Deposton Uncertady NUREG 1525: ASSESSMENT OF THE NRC ENFORCEMENT PRO-Assessment Appendices A And B-GRAM ^

NUREG/CR 6244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dispersion and Depostion Uncertainty UN,C.C.

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

NUREG/CR4144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1. Evaluation Of Severe Accident Risk Dunng M4tpop NUR CR 59'9 V04 N1: SHORT CRACKS IN PIPING AND PIPING STEYS E T

ROU H A L CIR-NU G/

-64 V06 : EVALUATION OF POTENTIAL SEVERE AC-NU EG/C 2 CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis. March SURRY, UNIT 1. Evaluation Of Severe Accident Risk During M4 Loop 1990 - December 1994 NUREG/CR-6298. FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN.

Operatons. App ndces.

TIALLY SURFACE-CRACKED PIPE.

UU,X.H.

NUREG/CR-6299. EFFECTS OF TOUGHNESS ANISOTROPY AND NUREG/CR4264 V01: VAllDITY UMITS IN J-RESISTANCE CURVE COMBINED TENSION, TORSION, AND BENDING LOADS ON FRAC-DETERMINATIOttAn Assessment Of The J(M) Parameter.

TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK OPEN.

GA A ALYSES FOR CIRCUMFERFNTIAL THROUGH-WALL UR CR-6261: A

SUMMARY

OF ORNL FISSION PRODUCT RE-LEASE TESTS WITH RECOMMENDED RELEASE RATES AND DIF-KVARFORDT,K.J.

FUSION COEFFICIENTS.

NUREG/CA-6116 V06. SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4318: DATA

SUMMARY

REPORT FOR FISSION PRODUCT HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

RELEASE TEST Vl-7.

VERSION 5.0 Graphical Evaluation Module (GEM) Reference Mariual.

LUCADAMO,K.

NUREG/CR-2907 V13. RADIOACTIVE MATERIALS RELEASED FROM LACH,D.

NUREG/CR-6330 RESULTS OF REGULATORY IMPACT SURVEY OF NUCLEAR POWER PLANTS. Annual Report 1992.

INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE NUREG/CR 2907 V14: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS.

OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

LUCKAS,WJ.

LAMBERT,L.D.

NUREG/CR4154 V02: EXPERIMENTAL RESULTS FROM CONTAIN-NUREG/CR4265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN MENT P1 PING BELLOWS SUBJECTED TO SEVERE ACCIDENT RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF CONDITIONS Results From Bellows Tested in Corroded Conditions.

COMMISSION AND DEPENDENCIES.

NUREG/CR4184: SEPARATE EFFECTS TESTING AND ANALYSES TO IGAT E E I THE 1:16-SCALE REINFORCED

'U E'G/CR4244 V01: PROBABILISTIC ACCIDENT CONSEOVENCE UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertaanty LAUGHERY,K.R.

Assessment. Main Report.

NUREG/CR4159: USING MICRO SAINT TO PREDICT PERFORMANCE NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of Vahdity UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty And Feasibihty.

Assessment. Appendices A And B.

NUREG/CR-6244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE LAWLESS,M.T.

UNCERTAINTY ANALYSIS. Dspersion and Deposition Uncertainty NUREG/CR4159: USING MICRO SAINT TO PREDICT PERFORMANCE Assessment. Appendices C.D.E.F,G.H.

IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of Vahdity And Feasibility.

MACFARLANE,R.E.

NUREG/CR4214: PRODUCTION AND TESTING OF THE VITAMIN-86 LAWRENCE,J.D.

FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

NUREG/CP-0145:

WORKSHOP ON DEVELOPING SAFE PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENOF/B.VI SOFTWARE. Held At Hotel Del Coronado, San Diego,CA. July 22-NUCLEAR DATA.

23,1992.

MACKINNON R.J.

LEE,A.S.

NUREG/CR4305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EOUI-NUREG/CR4322-BUCKLING ANALYSIS OF SPENT FUEL BASKET.

LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS-

Personal Author Index 57 ING THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL WASTF MCKAY M.D.

j DISPOSAL UNITS Background. Theory, And Model Descnpt on.

NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-l MADER,E.V.

DENT CONSEQUENCE MODEL NUREG/OR4327:. MODELS FOR EMBRITTLEMENT RECOVERY DUE NUREG/CR4244 VO1: PROBABILISTIC ACCIDENT CONSEQUENCE TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty r

I CANKAMO,T AssessrnenWain Repod REN6244 W2: NBABluSM AMEM CONSEQUENCE I

NUREG/CR4141: HANDBOOK OF ME* HODS FOR RISK BASED ANALYSES OF TECHNICAL SPECIFICATIONS.

UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty Assessment Appendices A And B.

i NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE MANTEUFEL,R.D.

NUREG/CR4348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION UNCERTAINTY ANALYSIS, Disperson and Depositen Uncertainty i

IN PARTIALLY SATURATED POROUS MEDIA.

Assessment. Appendices C.D.E.F.G.H.

l NUREG/CR4311: EVALUATING PREDICTION UNCERTAINTY.

MARSCHALL,C.

NUREG/CR4 RA URE B HAVIOR OF SHORT CIRCUMFEREN.

MCK NNEY M OWNED NUCLEAR POWER PLANTS BY PUBLIC JTILITY REGULA-CARSCHALL,C.W.

TORS.

l NUREG/CR4235: ASSESSMENT OF SHORT THROUGH-WALL CIR-i CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis March MCNEIL,K.A.

1990 - December 1994.

NUREG/CR-6150 V04:

SCDAP/RELAP5/ MOD 3.1 CODE NUREG/CR4299 EFFECTS OF TOUGHNESS ANISOTROPY AND MANUAL.MATPRO--A Library Of Matenals Properties For Ught-Water.

l I

COMBINED TENSION. TORSION. AND BENDING LOADS ON FRAC.

Reactor Accident Analysis.

TURE BEHAVIOR OF FERAITIC NUCLEAR PIPE.

MEINHOLD.C.B.

EARTINEZ OURfD4 NUREG/CR-6112; IMPACT OF REDUCED DOSE LIMITS ON NRC Li-l' NUREG/CH41T2: REVIEWING PSA-BASED ANALYSES TO MODIFY CENSED ACTIVITIES Malor issues in The implementaten Of ICRP/

TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

NCRP Dose Umst Recommendations Final Report.

l MASON.R.E.

MELBER,8.

NUREG/CR4150 V04:

SCDAP/RELAP5/ MOD 3.1 CODE NUREG/CR-6330; RESULTS OF REGULATORY IMPACT SURVEY OF MANUAL.MATPRO-A Library Of Matenals Properties For Ught-Water

• INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE Reactor Accident Analysis.

OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

MCAFEE,W.J.

MERRITT,L !,

NUREG/CR4273: BIAXIAL LOADING EFFECTS ON FRACTURE NUREG/CF JS RO1: INCENTIVE REGULATION OF INVESTOR-TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL-OWNED fwCLEA9 POWER PLANTS BY PUBLIC UTILITY REGULA-t MCCARTIN,T.J.

NUREG 1464 NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE MEYER K.A.

2 Development Of Capabihties For Review Of A Performance Asuss.

NUREG/CR-6348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION ment For A High-Level Waste Repository.

IN PARTIALLY SATURATED POPOJS MEDIA.

MCCOMAS,M.L M ENAMES,KA NUREG/CR 6150 V04-SCDAP/RELAP5/ MOD 3.1 CODE NUREG/CR 6356' HYDRAULIC CHARACTERIZATION OF HYDROTH-MANUALMATPRO-A Library Of Matenals Properties For Ught-Water.

ERMALLY ALTERED NOPAL TUFF.

Reactor Accident Analysis.

MICHAUD,W.F.

MCCONNELL J.W.

NUREG/CR-4667 V18: ENV"iONMENTALLY ASSISTED CRACKING IN NUREG/CR-5229 V07. FIELD LYSIMETER INVESTIGATIONS: LOW-LIGHT WATER REACTORS. Semiannual Report October 1993 - March i

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR 18 "

FISCAL YEAR 1994. Annual Report.

NUREG/CR4188 V02. MICROBIAL DEGRADATION OF LOW-LEVEL NURE'G 1506 DRFT FC: MEASUREMENT METHODS FOR RADIOLOGi-N RE 6 V01 EYDL ETE IN TIGATIONS - TEST CAL SURVEYS IN SUPPORT OF NEW OECOMMISSIONING RESULTS. Low-Level Waste Data Base Program. Test Results For CRITERIA. Draft Report For Comment.

Fiscal Years 1986.1987.1988 And 1989 NURE3/CR4256 V02: FIELD LYSIMETER INVESTIGATIONS TEST P

"I N EG R-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-Results F F aYas 90 1 1,92 d 99 '

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT l

GRAND GULF. UNIT 1. Evaluation Of Severe Accident Risi a For P' ant RCDONALD,J.C.

NUREG/CR-6354 DRF FC: PERFORMANCE TESTING OF ELECTRON-Operational State 5 Dunng A Refueling Outage Main Repori And Ap-IC PERSON'L DOSIMETERS. Draft Report For Comment.

pendices.

NUREG/CR 6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE MCDUFFIE.P.N.

UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty NUREG/CR-5884 V01: RFVtSED ANALYSIS OF DECOMMISSO%iNG Assusment. Main Report.

FOR THE REFERENCE PRESSURIZED WATER REACTOR SQWER NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSEOUENCE STATION Effects Of Current Regulatory And Other Consideratans On UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty

{

The Financsal Insurance Requirements Of The Decommissionng Rule Assessment. Appendices A And B i

NUREG/CR-6244 V03. PROBABILISTIC ACCIDENT CONSEQUENCE I

And...

NUREG/CR-5884 V02-REVtSEO ANALYSIS OF DECOMMISSIONING U"RTAINTY ANALYSIS. Dispersion and Deposition Uncertairdy FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER m wssment. Appendices C.D.E.F,G.H.

'WG/CR4316 V01: GUIDELINES FOR THE VERIFICATION AND STATION Effects Of Cunent Regulatory And Other Considerations On 4

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-The Financial Insurance Raquirements Of The Decommissanning Rule AL SOFTWARE.

And,..

NUREG/CR-6316 V02: GUIDEUNES FOR THE VERIFICATION AND i

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-MCELHANEY,K.L.

+

1 NUREG/CR 5944 V02: A CHARACTER 12ATION OF CHECK VALVE AL SOFTWARE. Survey And Assessment Of Conventional Software j

DEGRADATION AND F AILURE EXPERIENCE IN THE NUCLEAR Venficatson And Validation Methods NUREG/CR-6316 V03-GUIDELINES FOR THE VERIFICATION AND 3

POWER INDUSTRY.190 f Fa.ures NUREG/CR-6016 AGIN') AND SERVICE WEAR OF AIROPERATED VALIDATION OF EXP!'RT SYSTEM SOFTWARE AND CONVENT 60N-1 VALVES USEO IN 6AFETY RELATED SYSTEMS AT NUCLEAR AL SOFTWARE. Survey And Documentation Of Expert System Venfica-l POWER PLANTS.

ton And Validation Methodologies.

l

\\

i i

l 1

58 Personal Author index NUREG/CRE16 V04 GUIDELINES FOR THE VERIFICATION AND MONTELEONE,S.

VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL NUREG/CP-0140 V01: PROCEEDINGS OF THE TWENTY SECOND SOFTWARE Evuluaton Of Knowledge Base Certification Methods WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-NUREG/CR-6316 VOS: GUIDELINEb FOR THE VERiflCATION AND son, Advanced Instrumentation & Control Hardware & Software, VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Human Factors Research, IPE & PRA.

AL SOFTWARE.Ratonale And Desenption Of V&V Gunteline Packages NUREG/CP-0140 V02: PROCEEDINGS OF THE TWENTY SECOND WATER REACTOR SAFETY INFORMATION MEETING. Severe Acci-NU G R 316 V06: GUIDELINES FOR THE VERIFICATION AND dent Research, Thermal Hydraulic Research For Advanced Passive VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

AL SOFTWARE. Validation Scenanos.

LWRs, High-Burnup Fuel Behaver.

NUREG/CR-6316 V07. GUIDELINES FOR THE VERIFICATION AND NUREG/CP-0140 V03. PROCEEDINGS OF THE TWENTY SECOND VALIDATION OF 8NPERT SYSTEM SOFTWARE AND CONVENTION.

WATER REAC"9 OAFETY INFORMATION MEETING Primary Sys.

AL SOFTWARE %W Manual tems integnty, souctural And Seismic Engineenng, Aging Research, NUREG/CR 6316 M.' GUIDELINES FOR THE VERIFICATIGN AND Products And Applicatons.

VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CP-0148. TRANSACTIONS OF THE TWENTY THIRD WATER AL SOF1 WARE Bibiography.

REACTOR SAFETY INFORMATION MEETING.

MILLER,R L MONTGOMERY,J.

NUREG/CR-6150 V04.

SCDAP/RELAP5/ MOD 31 CODE NUREG-1516 DRFT FC-MANAGEMENT OF RADIOACTIVE MATERIAL MANUAL.MATPRO A Library Of Matenals Properties For Light Water' SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Com-Reactor Accident Ar. lysis.

ment.

MINARICK J.W.

NUhlG/CR-4074 V21: PRICURSORS TO POTENTIAL SEVERE CORE MOORE.R.A.

DAMAGE ACCIDENTS 1994 A STATUS REPORT. Main Report And Ap.

NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE pendices A-H.

AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks NUREG/CR 4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE in Remote Afterloading Brachytherapy And Approaches For Improved DAMAGE ACCIDENTS 1994 A STATUS REPORT.Appenden I.

System Performance.

NUREG/CR-6125 V03. HUMAN FACTORS EVALUATION OF REMOTE MINTON.L AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human.

NUREG-1493: PERFORMANCE-BASED CONTAINMENT LEAK-TEST F-sten, Interfaces. Procedures And Practices. Training And Organaaton-A Practices And Procedures.

NU E 493 D FORMANCE-BASED CONTAINMENT LEAK-TEST PROGRAM Draft Report for Comment.

MORtSSEAU,0.S.

MIRSKY,S.M-NUREG/CR-6277 VOI: HUMAN FACTORS EVALUATION OF TELE-NUREG/CR 6316 V01: GUIDELINES FOR THE VERIFlI%m 4 AND THERAPY. ldentificaton Of Problems And Alternative Approaches.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND L,.1 TION-NUREG/CR 6277 V02: HUMAN FACTORS EVALUATION OF AL SOFTW ARE.

TELETHERAPY Function And Task Analysis.

NUREG/CR-6316 V02. GUIDELINES FOR THE VERIFICATION AND NUREG/CR 6277 V03.

HUMAN FACTORS EVALUATION OF VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-TELETHERAPY. Human-System Interfaces And Procedures.

AL SOFTWARE. Survey And Assessment Of Conventonal Softwsm NUREG/CR 6277 V04.

HUMAN FACTORS EVALUATION OF Venficaton And Vahdation Methods.

TELETHERAPY. Training And Organizational Analysis MUREG/CR-6316 V03. GUIDELINES FOR THE VERIFICATION AND NUREG/CR-6277 V05.

HUMAN FACTORS EVALUATION OF VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-TELETHERAPY Literature Review.

AL SOFTWARE. Survey And Documentation Of Expert System Venfica.

tion And Validation Methodologies MORTON,D'K~

NUREG/CR 6316 V04: GUIDELINES FOR THE VERIFICATION AND NUREG/CR 6260. APPLICATION OF NUREG/CR-5999 INTERIM FA-VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL TIGUE CURVES TO SELECTFO NUCLEAR POWER PLANT COMPO-SOFTWARE Evaluation Of Knowledge Base Certification Methods.

NUREG/CR 6316 V05: GUIDELINES FOR THE VERIFICATION AND NENTS.

VALIDATION OF EXFERT SYSTEM SOFTWARE AND CONVENTION-AL SOFTWARE. Ratio nale And Descnpton Of V&V Guideae Packages MUBAYI,V.

And Procedures.

NUREG/CR-6144 V06 Pt: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-6316 V06 GUIDELINES FOR THE VERIFICATION AND CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT k

VALIDATION OF ELPERT SYSTEM SOFTWARE AND CONVENTION-SURRY, UNIT 1 Evaluat on Of Severe Accider; Risk Dunng Mid-Loop AL SOFTWAREAa dat on Scenanos.

Operations Main Report.

NUREG/CR-6316 V)7: GUIDELINES FOR THE VERiflCATION AND NUREG/CR-6144 V06 P2 EVALUATION OF POTENTIAL SEVERE AC.

VAllOATION OF ZXPERT SYSTEM SOFTWARE AND CONVENTION-CIDEN.! DURING L0W POWER AND SHUTDOWN OPERATIONS AT L$1DE INES FOR THE VERIFICATION AND

• ^

"""O NU

/CR-63

.8 VALIDAflON OF EXPERT SYSTEM SOFTWARE AND CONVENTION NUR 634 OST-BENEFIT CONSIDERATIONS IN REGUL A.

q; AL SOFTWARE.Bibbography' TORf ANALYSIS.

MITTS.T M.

MUCKLE RJ. A.

NUREG/CR-6355-A LIMITED ASSESSVENT OF THE ASEP HUMAN REllABILITY ANALYSIS PROCEDURE USING S!MULATOR EXAMI, NUREG/CR 6125 V01. HUMAN FACTORS EVALUATION OF REMOTE NATION RESULTS.

AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks in Remote Afterloa$ng Brachytherapy And Approaches for improved Mic h System Performance.

Nb.'

/CR-6235. ASSESSMENT OF SHORT THROUGH WALL CIR-qgqEG/CR-6125 V02 HUMAN MCTORS EV ALUATION OF REMOTE K

CUL ERENTIAL CRACKS IN PIPES Expenments And Analyus. March AFTERLOADING BRACHYTHEj. kPY Funcbon And Tar.k Analysis 1990 - Decemter 1994.

NPEG'CR 6125 V03 HUMAN FACTORS E VALUATION OF RE MOTE Af TE RLOADING BRACHYTHERAPY. Supporting Analyses Of Human.

m ms s

aeshnN N Waem JR G CR-6298 FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN

^"

TiALL V SURFACE-CRACKED PIPE NUREG/CR-6299 EFFECTS OF TOUGHNESS AN'SCTROPY AND COMB!NED TENSION. TORSIC;1 AND BENDING LOADS ON FRAC-NUREG 1517 RE PORT OF THE SOUTH TEXAS PROJE CT ALL EGA-TURE BEHAVIOR OF F ERRITIC NUCLEAR PIPE.

TlONS PE VtEW TEAM MOK,G C.

NURE G/CR-6242 CAs<S (COMPUTER ANALYSIS OF STORAGE MURPHY,W M.

CASKS) A MICROCOMPUTER BASED ANALYS!S SYSTEM FOR NURf G/CP-0147 PROCE E DINGS Or THE WOWSHOP DN THF RD '

STOR AGE CASV DESIGN REVIEW User's Manual To Wnon ib On-Or NATURAL ANALOGS IN GE Ot.OGIC OsposAL Of HiGH Leva cluderg Program Rotwence)

N ELE AR WASTE Held in San Anton.c Tewas July 22 25 1991

Personal Author index 59 I

NEILSON,R.M.

OLAGUE,N.E.

I NUREG/CR-6256 V01: FIELD LYSIMETER INVESTIGATIONS - TEST NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE ASSESS-RESULTS Low Level Waste Data Base Program. Test Results For MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE Fiscal Years 1986,1987,1988, And 1989 DISPOSAL FACILITIES. Validation Needs.

NUREG/CR-6256 V02: FIELD LYSIMETER INVESTIGATIONS TEST RESULTS Low Level Waste Data Base Development Program: Test OLSEN,C.S.

Results For Fiscal Years 1990,1991,1992, And 1993.

NUREG/CR-6150 V04:

SCDAP/RELAPS/ MOD 3.1 CODE A

A ary O Matenals Ropmes For @-Water NELSON,E[CR-6327:

E

• • # ^#'

^"" "'

NUREG MODELS FOR EMBRITTLEMENT RECOVERY DUE TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

ORNSTEIN H.L l

NUREG 1275 V11: OPERATING EXPERIENCE FEEDBACK REPORT -

NEUMAN,S.P TURBINE GENERATOR OVERSPEED PROTECTION NUREG/Cd6308: AN OVERVIEW OF INSTABILITY AND FINGERING SYSTEM Commercial oower Reactors.

DURING IMMISCIBLE FLUID FLOW IN POROUS AND FRACTURED MEDIA.

OS80RNE,M.F.

NUREG/CR-6261: A

SUMMARY

OF ORNL FISSION PRODUCT RE-NEYMOTIN,L LUSE TESTS WITH RECOMMENDED RELEASE RATES AND DIF-NUREG/CR 6144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NU E CR 318 ATA UMMARY REPORT FOR FISSION PRODUCT SURRY, UNIT 1 Evaluation Of Severe Accident Risk Dunng Mid-Loop RELEASE TEST VI Operations Main Report.

NUREG/CR-6144 V06 P2: EVALUATION OF POTENTIAL SEVERE AC.

CICENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUR G C 4285: SEVERE ACCIDENT NATURAL CIRCULATION SURRY, UNIT 1.Evaluewn Of Severe Accident Risk Dunng Mid-Loop STUDIES AT TilINEL' l

Operations. Appendices.

PAGE,G.

NICHOLSCh,W.L NUREG/CR-6307:

SUMMARY

OF COMMENTS RECElVED AT WORK-NUREG/CR-6355. A LIMITED ASSESSMENT OF THE ASEP HUMAN SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO RELIAB. lTY ANALYSIS PROCEDURE USING SIMULATOR EXAMI.

FACILITATE PUFJC PARTICIPATION IN DECOMMISSIONING NATION RESULTS.

CASES.

NICKOLAUS,J.R.

NUREG/CR-5973 R02: CODES AND STANDARDS AND OTHER GUID-PAIK.S.

ANCE CITED IN REGULATORY DOCUMENTS.

NUREG/CR4150 V02:

SCDAP/RELAP/ MOD 3.1 CODc MANUAL. Damage Progression Model Theory.

NICOLETTE,V.F.

NUREG/CR-6017: FIRE MODELING OF THE HEISS DAMPF REAKTOR PARKS.C.V.

CONTAINMENT.

NUREG/CR-0200 V1 R04: SCALE: A MODULAR CODE SYSTEM FOR i

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-NITZELM.E.

CENSING EVALUATION Control Modules.

NUREG/CR.6260: APPLICATION OF NUREG/CR 5999 INTERIM FA-NUREG/CR 0200 V2PIR4: SCALE: A MODULAR CODE SYSTEM FOR

)

TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-NENTS.

CENSING EVALUATION. Functional Modules F1 F8.

NUREG/CR-0200 V2P2R4: SCALE: A MODULAR CODE SYSTEM FOR i

NORTON,L PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-NUREG 1415 V07 NO2: OFFICE OF THE INSPECTOR CENSING EVALUATION Functional Modules F9-F16.

GENERAL. Semiannual Report To Congress, October 1,1994 - March NUREG/CR.0200 V3 R04: SCALE: A MODULAR CODE SYSTEM FOR 31,1995.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-CENSING EVALUATION. Miscellaneous.

NOVACK,S.D.

NUREG/CR4284. CRITICALITY SAFETY CRITERIA FOR LICENSE NUREG/CR-6116 V09. SYSTEMS ANALYSIS PROGRAMS FOR REVIEW OF LOW LEVEL WASTE FACILITIES.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR-6328: ADEOUACY OF THE 123-GROUP CROSS-SECTION VERSION 5 0 Venfication And Valdation (V&V) Manual.

LIBRARY FOR CRITICALITY ANALYSES OF WATER-MODERATED i

URANIUM SYSTEMS.

NUREG/CR 6173: A

SUMMARY

OF THE FIRE TESTING PROGRAM AT PARKS,M.B.

THE GERMAN HDR TEST FACILITY.

NUREG/CR4154 V02 EXPERIMENTAL RESULTS FROM CONTAIN-NUREG/CR-6220: AN ASSESSMENT OF FIRE VULNERABILITY FOR MENT PIMG BELLOWS SUBJECTED TO SEVERE ACCIDENT AGED ELECTRICAL RELAYS-CONDITIONS.Results From Bellows Tested in Corroded Conditions.

O'BRIEN.J.E.

PASLER SAVER.J.

i NUREG/CR-6285. SEVERE ACCIDENT NATURAL CIRCULATION NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE STUDIES AT THE INEL.

UNCERTAINTY ANALYSIS Dispersion And Deposition Uncertainty Assessment. Main Report.

NUREG/CF4244 V02: PROBABILISTIC ACCIDENT CMISEQUENCE N RE 4918 V08 f' WTROL OF WATER INFILTRATION INTO UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty NEAR SU3 FACE LLW DISPOSAL UNITS Progress Report Of Field Ex.

Assessment. Appendices A And B.

I periments At A Humid Region Site,Beltsville Maryland NUREG/CR4244 Yb3: PROBABILISTIC ACCIDENT CONSEQUENCE i

UNCERTAINTY ANALYSIS. Dispersion and Deposition Uncertain 4 O'HARA.J.M.

NURF G-0700 RO1 DFC: HUMAN-SYSTEM INTERFACE DESIGN Assessment. Appendices C,0,E,F,G.H.

I REVfEW GUIDELINE. Draft Report For Comment p

ODEGAARDEN.R.H.

NUREG/CR-6004: PROBABILISTIC PIPE FRACTURE EVALUATIONS j

NUREG/CR 6284 CRITICALITY SAFETY CRITERIA FOR LICENSE FOR LEAK-RATE-DETEC'lON APPLICATIONS, 4

REVIEW OF LOW-LEVEL WASTE F ACILITIES PAULSEN,M.P.

NUREG/CR-6325. AN IMPLIC*T STEADY STATE INITIALIZATION PACK.

ODETTE,G.R-I NUREG/CR4327. MODELS FOR EMBGTTLEMENT RECCVERY DUE AGE FOR THE RELAPS COMPUTER CODE.

j TO ANNE ALD tG OF REACTOR PRESSURE VESSEL STEM S NUREG/CR4382. COMPARISCNS OF ASTM STANDARDS CITED IN OFOEGBU.G.L Nt., REG /CR4; B3. FIELD SITE INVESTIGATION EFFEC' N MlNE TW NRC STANDARD REVIEW PLAN.NUREG 0800, AND RELATED SEISMICITS ON GROUNDWATER HYDROLOGY.

DOCUMENTS 3

I l

1 4

60 Person:1 Auth:r index NUREG/CR-6386: COMPARISONS CF ANSI STANDARDS FTED IN NUREGrCR-6297: FRACTURE EVALUATIONS OF FUSION LINE THE NRC STANDARD REVIEW PLAN,NUREG 0800, AND RdLATED CRACKS IN NUCLEAR PIPE BIMETALLIC WELDS.

DOCUMENTS.

NUREG/CR4298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-TIALLY SURFACE CRACKED PIPE.

PENNELL,W.E.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK OPEN.

NUREG/CR 4219 V10 N2: HEAVY-SECTION STEEL TECHNOLOGY ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH WALL PROGRAM Semiannual Progress Report For Aprd-September 1993 CRACKS IN PIPES.

NUREG/CR-4219 VII N1: HEAVY.SECTION STEEL TECHNOLOGY PROGRAM Semaannual Progress Report For October 1993 - March RAMEY SMITH,A.

1994 NUREG/CR4265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN NUREG/CR-6273 BIAX1AL LOADING EFFECTS ON FRACTURE RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF TOUGHNESS OF REACTOR PRESSURE VESSEL bTEEL-COMMISSION AND DEPENDENCIES.

PERANICH,M.

RAMSDELL,J.V.

NUREG 1528. RECONSTITUTION OF THE MANUAL CHAPTER 2512 NUREG/CR 6331: ATMOSPHERIC RELATIVE CONCENTRATIONS IN CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT 1.

BUILDING WAKES.

PERSENSKY,J J' RO1 RAO,0 V NUREG 0700 DFC: HUMAN-SYSTEM INTERFACE DESIGN NUREG/CR-6224: PARAMETRIC STUDY OF THE POTENTIAL FOR NURE /CR419 N

I I T R DICT PERFORMANCE BWR ECCS STRAlWER BLOCKAGE DUE TO LOCA GENERATED IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of Valdty BR And Feasitxhtv.

N Eg R-6368: EXPERIMENTAL INVESTIGATION OF SEDIMENTA.

TION OF LOCA GENERATED FIBROUS DEBRIS AND SLUDGE IN PETERSON,C.E.

BWR SUPPRESSION POOLS.

NUREG/CR-6325: AN IMPLICIT STEADY STATE INITIAUZATION PACK.

AGE FOR THE RELAPS COMPUTER CODE.

RE REG /C 6150 V04.

SCDAP/RELAP5/ MOD 3.1 CODE PETRONE,C.D.

MANUAL.MATPRO--A Library Of Malenals Properties For Light-Water-NUREG-1526: LESSONS LEARNED FROM EARLY IMPLEMENTATION Reactor Accident Analysis.

OF THE MAINTENANCE RULE AT NINE NUCLEAR POWER PLANTS.

REZOS.J.T.

PHILIP,J.

NUREG/CR-6002:

RISK-BASED MAINTENANCE NUREG/CR4283 FIELD SITE INVESTIGATION-EFFECT OF MINE MODELING Pnontization Of Maintenance importances And Quantifica-SEISMICITY ON GROUNDWATER HYDROLOGY.

tion Of Maintenance Effectiveness.

PILCH,M.M.

RICE,G.

NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY NUREG/CR.6348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION j

DIRECT CONTAINMENT HEATING IN SURRY.

IN PARTIALLY SATURATED POROUS MEDIA.

NUREG/CR4356: HYDRAULIC CHARACTERIZATION OF HYDROTH.

PIPER,R.K*

ERMALLY ALTERED NOPAL TUFF.

NUREG/CR4354 DAF FC: PERFORMANCE TESTING OF ELECTRON.

IC PERSONAL DOSIMETERS. Draft Report For Comment RIDGELY,J.N.

NUREG-1465: ACCIDENT SOURCE TEF!M9 FOR LIGHT WATER NU-NUREG 5 DRFT FC: A NONPARAMETRIC STATISTICAL P,iTHOD-OLOGY FOR THE DESIGN AND ANALYS!S OF FINAL bi ATUS DE-RIDKY,R.W.

COMMISSIONING SURVEYS Draft Report For Comment.

NUREG/CR-4918 V08: CONTROL OF WATER INFILTRATION INTO PRATT S.L NEAR SURFACE LLW DISPOSAL UNITS Progress Report Of Field Ex-NUREG/CR-6390: RADIOLOGICAL CHARACTER 12ATION OF SPENT penments At A Humid Region Site.BeltsvilleNaryland.

1 CONTROL ROD ASSEMBLIES RIGCSSEE.E.T.

PRATT,W.T.

NUREG/CP4334: NEW SENSOR FOR MEASUREMENT OF LOW AIR NUREG/CR 6144 V01: EVALUATION OF POTENTIAL SEVERE ACCl-FLOW VELOCITY. Phase i Final Report.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ROBERTSON.D.E SURRY, UNIT 1 Summary Of Results.

NUREG/CR-6390: RADIOLOGICAL CHARACTERIZAllON OF SPENT QUICK K.S.

CONTROL ROD ASSEMBLIES NUREG/CR4109. THE PROBABILITY OF CONTAINMENT FAILURE BY DIRECT CONTAINMENT HEATING IN SURRY.

RO NUREG R 243. COMPARISON OF ASME CODE FATIGUE EVALUA-OUINN,M.L TION METHODS FOR NUCLEAR CLASS 1 PIPING WITH CLASS 2 NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE OR 3 PIPING.

AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks R ERS O in mo e An ding Brachytherapy And Approaches For improved NU EG/CR4125 V03. HUMAN FACTORS EVALUATION OF REMOTE LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR AFTERLOADING BRT,HYTHERAPY. Supporting Analyses Of Human.

FISCAL YEAR 1994 Annual Report System lnterfaces. Pre s:rJures And Practices.Trairwng And Organization-NUREG/CR-6188 V02, MICROBIAL DEGRADATION OF LOW LEVEL al Practices And Procedures.

RADIOACTIVE WASTE. Annual Report For FY 1994 NUREG/CR 6256 V01: FIELD LYSIMETER INVESTIGATIONS TEST RADDATZ,C.T.

RESULTS Low Level Waste Data Base Program Test Results For NUREG 0713 V15: OCCUPATIONAL RADIATION EXPOSURE AT COM.

Fiscal Years 1986.1987,1988, And 1989 MERCIAL NUCLEAR POWER REACTORS AND OTHER NUREG/CR-6256 V02: FIELD LYSIMETEFI INVESTIGATIONS - TEST i

FACILITIES,1993. Twenty-Sarth Annual Report.

RESULTS. Low-Leves Waste Data Base Development Program: Test i

Results For Fiscal Years 1990.1991,1992. And 1993 RAHMAN,E NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING ROLLSTIN.J.A.

WELDF.twventh Program Report March 1993 - December 1994.

NUREC,/CR 6134. UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/yn4004: PROBABluSTIC PIPE FRACTURE EVALUATIONS CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-FOR LEAKJtATE-DETECTION APPUCATIONS.

CIDENT CONSEQUENCE MODEL.

NUREG/CR4235: ASSESSMENT OF SHORT iHROUGRWALL CIR-NUREG/CR4136 UNCERTAINTY AND SENSITIVITY ANALYSIS OF CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis. March FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-1990 December 1994.

DENT CONSEQUENCE MODEL.

l Personal Author Index 61 l

I ROMAN.W.

NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF REMOTE NUREG 1493: PERFORMANCE-BASED CONTAINMENT LEAK. TEST AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-PROGRAM. Final Report System interfaces. Procedures And Practices, Training And Organization-NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK-al Practces And Procedures.

TEST PROGRAM. Draft Report For Comment.

l SCHAEFER C.

ROSE 8-NUREG/CR-6263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR NUREG 1493: PERFORMANCE-BASED CONTAINMENT LEAK TEST POWER PLANTS. Candidate Guidelines, Techncal Basis And Research PROGRAM.Fina! Report-Needs Executive Summa <y.

NUREG 1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK

• NUREG/CR4263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR TEST PROGRAM Draft Report For Comment.

POWER PLANTS Candidate Guidehnes, Techncal Basis And Research Needs Main Report.

j ROSENFIELD.A.R.

NUREG/CR4251: STAINLESS STEEL SUBMERGED ARC WELD SCHICK,W.R.

NUREG/CR-6074 VOS: SEALED SOURCE AND DEVICE DESIGN NU E R 9 FR C FiE EVALUATIONS OF FUSION LINE SAFETY TESTING.Technmal Repoit On The Findings Of Task 4.Inves-CRACKS IN NUCLEAR PIPE BlMETALUC WELDS.

tigation Of Failed hadioactive Staeniess Steel Troxter Gauges.

ROUSSIN,R.W.

NUREG/CR4214. PRODUCTION AND TESTING OF THE VITAMIN-86 SCHLENKER,LD.

FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

NUREG/CR-6285: SEVERE ACCIDFNT NATURAL CIRCULATION l

PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B VI STUDIES AT THE INEL SCHLUETER,J.

NUREG-1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL f.UDLAND,0.

NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING SAFETY PRCGRAMS AT MEDICAL FACILITIES. Draft Report for Com-nwnt.

WELDS Seventh Program Report March 1993 December 1994.

RUSSELL,K.D.

SCHOENFELD,t.

NUREG/CR-6116 V06. SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

AFTERLOADING BRACHYTHERAPY. Human Error And Cntcal Tasks VERSION 5.0. Graphical Evaluation Module (GEM) Reference Manual in Remote Afterloading Brachytherapy And Approaches For improved i

1 System Perforr..ance.

RUSSELL.M.J.

NUREG/CR4125 UO2 HUMAN FACTORS EVALUATION OF REMOTE NUREG/CR-6100: GATE VALVE AND MOTOR-OPERATOR RESEARCH AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

l FINDINGS.

NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY Supporting Analyses Of Human-UREd C 4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN I Pra ti es A oc s

LIGHT WATER REACTORS. Semiannual Report. October 1993 - March 1994.

SCHULTZ,R.R.

NUREG/CR-4667 V19. ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR.5535 V05 R1: RELAPS/ MOD 3 CODE MANUAL. User's LIGHT WATER REACTORS. Semiannual Report.Apni-September 1994.

Guideline.

SAILOR,V.

SCHULZ,R.K.

NUREG/CR43d9. COST-BENEFIT CONSIDERATIONS IN REGULA*

NUREG/CR-4918 V08: CONTROL OF WATER INFILTRATION INTO TORY ANALYSIS NEAR SURFACE LLW DISPOSAL UNITS Progress Report Of Field Ex.

penments At A Humid Region Site.Beltsville, Maryland.

SAMANTA P.K.

NUREG/CR4141: HANDBOOK OF METHODS FOR RISK-BASED SCOTT,P.

NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING NURE C 6172:

EW G PS ASE A ALYSES TO MODIFY TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

NUREG 3 AS ES E O

T R

W L CIR-CUMFERENTIAL CRACKS IN PtPES Expenrnents And Analysis, March SANDERS R.L N

/C FR CTURE EVALUATIONS OF FUSION LINE A U e undes ersion 3 Septem 1 4 CRACKS IN NUCLEAR PIPE BlMETALLIC WELDS.

l NUREG/CR-6119 V02:

MELCOR COMPUTER CODE NUREG/CR-6298: FRACTURE BEHAVIOR OF Sif0RT CIRCUMFEREN.

l MANUALS Reference Manuals. Version 1.8.3 September 1994.

TIALLY SURFACE CRACKED PIPE.

SANECKt,J.E.

SCOTTI,R.S.

NUREG/CR 4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4287: MANAGEMENT CONCEPTS AND SAFETY APPLICA-LIGHT WATER REACTORS. Semiannual Report,0ctober 1993 - March TIONS FOR NUCLEAR FUEL FACILIT ES.

1994.

f SEELY,H.E.

SANFORD,W.E.

NUREG/CR-5229 V07: FIELD LYSIMETER INVESTIGATIONS: LOW.

NURLG/CR 5975 R01: INCENTIVE REG'RSTION OF INVESTOR-OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULA-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR TORS.

FISCAL YEAR 1994 Annual Report.

NUREG/CR4256 V02; FIELD LYSIMETER INVESTIGATIONS TEST RESULTS. Low. Level Waste Data Base Development Program: Test SERIG,D.L i

NUREG/CR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE l

l Results For Fiscal Years 1990,1991,1992. And 1993.

AFTERLOADING BRACHYTHERAPY. Human Error And Cntcal Tasks In Remote Aftertoeding Brachytherapy And Approaches For improved l

SATTISON.M.B.

1 NUREG/CR4115 V06: SYSTEMS ANALYSIS PROGRAMS FOR System Performance.

l HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR4125 V02: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

i VERSION 5 0.Graphcal Evaluation Module (GEM) Reference Manual.

NUREG/CR4125 V03: HUMAN FACTORS EVALUATION OF REMOTE RAUNDERS,W.M.

AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-r NUREG/CR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE System interfaces. Procedures And Practces.Tranng And Organization-j AFTERLOADING BRACHYTHERAPY. Human Error And Cntcal Tasks al Practces And Procedures.

in Remote Altertoading Brachytherapy And Approaches For improved NUREG/CR4277 V01: HUMAN FACTO 43 EVALUATION OF TELE-THERAPY. Identifcation Of Problems Ar d Altemative Approaches.

J System Performance.

i NUREC2/CR4125 V02. HUMAN FACTORS EVALUATION OF REMOTE NUREG/CR4277 V02: HUMAN FA CTORS EVALUATION OF j

AFTFRLOA9ING BRACHYTHERAPY. Function And Task Analysis.

TELETHERAPY. Function And Task Ana,ysis.

t l

4j

62 Personal Author index NUREG/CR 6277 V03: HUMAN FACTORS EVALUAUON OF S4MION,G.P.

TELETHERAPY Human-System Interfaces And Procedures.

NUREG/CR4276: A COMPILATION OF CURRENT REGULATIONS, NUREG/CR-6277 V04.

HUMAN FACTORS EVALUATION OF STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING BRA.

TELETHERAPY.Traiturg And Organizabonal Analysis.

CHYTHERAPY, NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF TELETHERAPY.L'terature Review.

SIMONEN,C.A.

NUREG/CR 6331: ATMOSPHrD8C RL.ATIVE CONCENTRATIONS IN SETH,S.

BUILDING WAKES.

NUREG/CR4263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR POWER PLANTS Candidate Guidelines. Technical Basis And Research SIMONSON,S.A.

Needs Executive Summary NUREG/CR4305: BLT-EC (BREACH. LEACH, TRANSPORT, AND EQUI-NUREG/CR4263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS-POWER PLANTS Candidate Guidelines, Techrucal Basis And Research ING THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL WASTE Needs Main Report DISPOSAL UNITS Background, Theory, And Model Desenphon.

SHA,W.T.

SISK,0.R.

NUREG/CR 6266: ANALYSIS OF BORON DILUTION IN A FOUR-LOOP NUREG/CR 6354 DRF FC: PERFORMANCE TESTING OF ELECTRON.

PWR.

IC PERSONAL DOSIMETERS. Draft Report For Comment.

SHACK,W.J.

SKAY,D.M[1517: REPO9T OF THE SOUTH TEXAS PROJECT ALLEGA-NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG LIGHT WATER REACTORS. Semiannual Report, October 1993 March TlONS REVIEW TEAM.

1994-NUREG/CR-4667 V19. ENVIRONMENTALLY ASSISTED CRACKING IN SKOBLAR.L

^

NU E /

275 ICA PE T F

E L A D PR aR CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-NUREG-1493 DFC: PER'FORMANCE BASED CONTAINMENT LEAK.

NU EG/CR4335; FATIGUE STRAIN-LIFE BEHAvlOR OF CARBON R6aN Repod 6 Comment AND LOW-ALLOY STEELS, AUSTENITIC STAINLESS STEELS, AND iiLATER C.O.

ALLOY 600 IN LWR ENVIRONMENTS-NUREG/CR4214: PRODUCTION AND TESTING OF THE VITAMIN-06 SHAFFER C OW AND M BWM BMAMM WRN NUREG/dR-6224. PARAMETRIC STUDY OF THE POTENTIAL FOR PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENOF/8-VI BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED NMEAR DATA.

DEBRIS.

SMITH,C.L SHIEH A.L NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

MANUALDamage Progression Model Theory.

NLRE b 1 V10 SY T ANALY I P OGRAMS FOR SHlH.C.F.

HANDS ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR-6264 V01. VALIDITY LIMITS IN J-RESISTANCE CURVE VERSION S.O. Data loading Manual.

DETERMINATION An Assessment Of The J(M) Parameter.

NUREG/CR 6264 V02: VAllDITY LIMITS IN J-RES! STANCE CURVE SMITH,0.

DETERMINATION.A Computational Approach To Ductile Crack Growth NUREG/CR-6307:

SUMMARY

OF COMMENTS RECElVED AT WORK-Under Large-Scale Yielding Conditiomt SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO FACILITATE PUBLIC PARTICIPATION IN DECOMMISSIONING SHIPMAN,R.L CASES.

NUREG/CR4313 VOI: ROBUST, ACCURATE, AND NON-CONTACTING VIBRATION MEASUREMENT SYSTEM Summary of Companson Meas.

SMITH,R.C.

urements Of The Robust Laser Interferometer And Typical Accelerome.

NUREG/CR-61?9 VO1: MELCOR COMPUTER CODE MANUALS.Pnmer ter Systems And User's Guides. Version 18.3 September 1994.

NUREG/CR-6313 V02: ROBUST, ACCURATE, AND NON CONTACTING NUREG/CR-6119 V02.

MELCOR COMPUTER CODE VIBRATION MEASUREMENT SYSTEMS Supplemental Appendices MANUALS. Reference Manuals. Version 18.3 September 1994, Presenting Companson Measurements Of The Robust Laser Interfer.

SMITH R.l ometer And Typical Accelerometer Systems.

NUREG/CR4884 V01: REVISED ANALYSIS OF DECOMMISSIONING SHIVER.A.W.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER NUREG/CR41'l4; UNCERTAINTY AND SENSITMTY ANALYSIS OF STATION Effects Of Current Regulatory And Other Considerations On CHRON!C EXPOSURE RESULTS WITH THE MACCS REACTOR AC.

The Financial Insurance Requirements Of The Decornmissioning Gule CIDENT CONSEQUENCE MODEL.

And...-

NUREG/CR4135. UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER DENT CONSEQUENOE MODEL.

STATION Effects Of Current Regulatory And Other Considerations On NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF The Financial Insurance Requirements Of The Decommissioning Rule FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-And....

DENT CONSEQUENCE MODEL SM YTH,S.B.

SHUMWAY,R.W.

NUREG/CR-6331: ATMOSPHERIC RELATIVE CONCENTRATIONS IN NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC BUILDING WAKES.

ENERGY OF CANADA LTD CODES SNIDER.D.M.

84EFKEN.LJ.

NUREG/CR4291 V01: NUCLEAR PLANT ANALYZER. installation NUREG/CR4150 V01:

SCDAP/RELAPS/ MOD 31 CODE Manual MANUALinterface Theory.

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

NUREG/CR4150 V02:

SCDAP/RELAP/ MOD 3.1 CODE ence Manul MANUAL Darnage Progression Model Theory NUREG/CR-6291 V03 NUCLEAR PLANT ANALYZER Computer Visual NUREG/CR-6150 V03.

SCDAP/RELAPS/ MOD 31 CODE System Reference Manual.

MANUALUser's Guide And input Manual.

NUREG/CR 6291 V04. NUCLEAR PLANT ANALYZER. Programmer's NUREG/CR4150 V04.

S'CDAP/RELAP5/ MOD 31 CODE Manual MANUALMATPRO-A Library Of Matenals Propertes For Light-Water-Reactor Accident Analysis.

SOFFER.L NUREG/CR4150 V05:

SCDAP/RELAPS/ MOD 31 CODE NUREG-1465. ACCIDENT SOURCE TERMS FOR LIGHT-WATER NU-MANUALDevelopmental Assessn.ent.

CLEAR POWER PLANTS.

l Personal Author Index 63 SOMMER,S.C.

NUREG/CR-6358 V02: ASSESSMENT OF UNITED STATES INDUSTRY NUREG/CR-5657: AUTOCASK (AUTOMATIC GENERATION OF 3-D STRUCTURAL CODES AND STANDARDS FOR APPUCATION TO CASK MODELS).A Mcrocomputer Based System For Shipping Cask ADVANCED NUCLEAR POWER REACTORS. Appendices.

Design Review Analysis.

I STOTHOFF S.A.

SOPPET,W.K.

NUREG/046333: BREATH VERSION 1.1 - COUPLED FLOW AND NUREG/CR-4667 V18; ENVIRONMENTALLY ASSISTED CRACKING IN ENERGY 13ANSPORT IN POROUS MEDIA. Simulator Descnption And LIGHT WATER REACTORS. Semaannual Report,0ctober 1993 March User Guide.

1994.

NUREG/CR-4667 V19: ENVIRONMENTALLY ASSISTED CRACKING IN STRUCKMEYER,r LIGHT WATER REACTORS Semiannual Report,Apnt-September 1994.

NUREG-0837 VG Nud: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report. October-December 1994.

SOUTO,F.J NUREG-0837 V15 N01. NRC TLD DIRECT RADIATION MONITORING NUREG/CR4224: PARAMETRIC STUDY OF THE POTENTIAL FOR j

R ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED NURE 0837 V NO2 N TLD D E RA ATION MONITORING NUREG/CR.6368: EXPERIMENTAL INVESTIGATION OF SEDiMENTA.

NETWORK. Progress Report.Apnl-June 1995.

NUREG-0837 V15 NO3: NRC TLD DIRECT RADIATION MONITORING TION OF LOCA-GENERATED FlBROUS DEBRIS AND SLUDGE IN NETWORK Progress Report. July-September 1995.

BWR SUPPRESSION POOLS.

SPENCER,E.W.

STUART,D.S.

NUREG/CR4109 THE PROBABILITY OF CONTAINMENT FAILURE BY NUREG/CR-6119 V01: MELCOR COMPUTER CODE MANUALS.Pnmer DIRECT CONTAINMENT HEATING IN SURRY.

And User's Guides. Version 1.8.3 September 1994.

NUREG/CR-6119 V02:

MELCOR COMPUTER CODE SPIESMAN.J.S.

MANUALS Reference Manuals. Version 1.8.3 September 1994.

NUREG/CR6382 COMPARISONS OF ASTM STANDARDS CITED IN l

THE NRC STANDARD REVIEW PLAN,NUREG-0800, AND RELATED STUART,G.

DOCUMENTS.

NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR NUREG/CR-6385: COMPARISONS OF ANS,ASME,AWS, AND NFPA POWER PLANTS Candidate Guidelines Technical Basis And R wearch STANDARDS CITED IN THE NRC STANDARD REVIEW Needs. Main Report.

PLAN.NUREG 0800, AND RELATED DOCUMENTS.

NUREG/CR-6386: COMPARISONS OF ANSI STANDARDS CITED IN STUBLER,W.F.

THE NRC STANDARD REVIEW PLAN,NUREG-0800, AND RELATED NUREG-0700 R01 DFC: HUMAN-SYSTEM INTERFACE DESIGN DOCUMENTS-REVIEW GUIDELINE. Draft Report For Corr ment.

SPLETZER.B.L SUEN.C.J, l

NUREG/CR 6184: SEPARATE EFFECTS TESTING AND ANALYSES TO NUREG/CR-6305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EQUI-INVESTIGATE LINER TEARING OF THE 1:16-SCALE REINFORCED LIBRIUM CHEMISTRY) A FINITE-ELEMENT MODEL FOR ASSESS-CONCRETE CONTAINMENT BUILDING.

ING THE RELEASE OF RADIONUCLIDES FROM LOW LEVEL WASTE DISPOSAL UNITS Background, Theory, And Model Descnption.

SPRUNG,J.L NUREG/CR-6134-UNCERTAINTY AND SENSITIVITY ANALYSIS OF

^

NU EG[C 3469 V08-OCCUPATIONAL DOSE REDUCTION AT NU-DENT CONS OUENCE M EL CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT.

NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF ED READINGS IN RADIATION PROTECTION AND ALARA.

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-DENT CONSEOUENCE MODEL NUREG/CR-6136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF SULLIVAN,T.M.

FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI.

NUREG/CR5229 V07: FIELD LYSIMETER INVESTIGATIONS: LOW-c LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR I

DENT CONSEQUENCE MODEL FISCAL YEAR 19G4. Annual Report.

STADLER,L NUREG/CR425G V02: FIELD LYSIMETER INVESTIGATiCNS - TEST NUREG-1350 V07: NUCLEAR REGULATORY COMMISSION INFORMA.

RESULTS Low-Level Waste Data Base Development Program: Test TION DIGEST.1995 Edition.

Results For Fiscal Years 1990,1991,1992. And 1993.

I NUREG/CR.6305: BLT-EC (BREACH, l.EACH. TRANSPORT, AND EQUI-(

STAMPS.D.W-LIBRIUM CHEMISTRY), A FINITE ELEMENT MODEL FOR ASSESS-NUREG/CR 6109. THE PROBABILITY OF CONTAINMENT FAILURE BY ING THE RELEASE OF RADIONUCLIDES FROM LOW LEVEL WASTE DIRECT CONTAINMENT HEATING IN SURRY-DISPOSAL UNITS Background, Theory. And Model Desenption.

STAUNTON,R.H.

SUMMERS,R.M NUREG/CR.6016. AGING AND SERVICE WEAR OF AIR-OPERATED NUREG/CR-6119 V01: MELCOR COMPUTER CODE MANUALS.Pnmer V LVES USED IN SAFETY RELATED SYSTEMS AT NUCLEAR And User's Gudes. Version 1.8.3 September 1994.

NUREG/CR-6119 V02-MELCOR COMPUTER CODE NUREG/CR4192. AGING AND SERVICE WEAR OF SPRING-LOADED MANUALS Reference Manuals. Version 1.8.3 September 1994.

PRESSURE RELIEF VALVES USED IN SAFETY RELATED SYSTEMS AT NUCLEAR POWER PLANTS.

SUN J.G.

NUREG/CR-6266. ANALYSIS OF BORON DlLUTION IN A FOUR-LOOP STEELE R.

PWR.

NUREG/CR-6100. GATE VALVE AND MOTOR OPERATOR RESEARCH f

SVEDEMAN,S.J.

NUREG/CR-6348. THERMALLY DRIVEN MOISTURE REDISTRIBUTION i

l STEVENSON,J.D.

IN PARTIALLY SATURATED POROUS MEDIA.

i NUREG/C46239 V01: SURVEY OF STRONG MOTION EARTHOUAKE EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-1 SWINTH,K.L SHASIS ON PIPING SYSTEMS Main Report NUREG/CR 6354 DAF FC: PERFORMANCE TESTING OF ELECTRON.

88EG/CR4239 V02: SURVEY OF STRONG MOTION EARTHOUAKE 1

s FECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-IC PERSONAL DOSIMETERS. Draft Report For Comment.

1 1

PHASIS ON PIPING SYSTEMS. Appendices NUREG/CR4240: APPLICATION OF BOUNDING SPECTRA TO SEIS.

TA.A.

MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF NUREG/CR4263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO POWER PLANTS. Candidate Guidelines, Technical Basis And Research Needs Executwo Summary.

STRONG MOTION EARTHQUAKES.

NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR NUREG/CR4358 VOI: ASSESSMENT OF UNITED STATES INDUSTRY 1

STRUCTURAL CODES AND GTANDARDS FOR APPLICATION TO POWER PLANTS Candidate Gudelines, Techrucal Bases And Research ADVANCED NUCLEAR POWER REACTORS Final Report.

Needs. Main Report.

i 5

f

64 Personal Author Index TABATABAl,H.

TUNG,V.X.

NUREG/CR-6420: SELF-MONITORING SURVEILLANCE SYSTEM FOR NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION PRESTRESSING TENDONS.

STUDIES AT THE INEL TADIO6.E.L ULER Y,8.

NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY NUREG/CR4263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR DIRECT CONTAINMENT HEATING IN SURRY.

POWER PLANTS. Candidate Guidelines, Technscal Basis And Research Needs Executive Summary.

TAM,P.S.

NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR NUREG-0847 S15 SAFETY EVALUATION REPORT REL.ATED TO THE POWER PLANTS. Candidate Guidehnes, Technical Basis And Research OPERATION OF WATTS BAR NUCLEAR PLANT. UNITS 1 AND Needs Main Report.

2 Docket Nos. 50 390 And 50 391.(Tennessee Valley Authonty)

NUREG-0847 S16: SAFETY EVALUATION REPORT RELATED TO THE VAN ARSDALE R.

OPERATION OF WATTS BAR NUCLEAR PLANT, UNlTS 1 AND NUREG/GR-0014: BALDCYPRESS TREE RING ELEMENTAL CONCEN-

2. Docket Nos. 50-390 And 50-391.(Tennessee Valley Authority)

TRATIONS AT REELFOOT LAKE TENNESSEE.FROM AD 1795 TO NUREG 0847 S17: SAFETY EVALUATION REPORT RELATED TO THE AD 1820.

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

VANDEN HEUVEL NUREG-0847 S18 SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND DAMAGE ACCIDENTS:1994 A STATUS REPORT.Masn Report And Ap-2 Docket Nos. 50 390 And 50 391 (Tennessee Valley Authonty) pendices A-H.

NUREG 0847 S19 SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND DAMAGE ACCIDENTS:1994 A STATUS REPORT. Appendix L

2. Docket Nos 50-390 And 50 391.(Tennessee Valley Authonty)

VANHORN,RL TANG,D.

NUREG/CR-6116 V10: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR-6293 V01: VERIFICATION AND VALIDATION GUIDELINES HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

FOR HIGH INTEGRITY SYSTEMS Main Report.

VERSION 5.0. Data Loading Manual.

NUREG/CR4293 V02: VERIFICATIONN AND VALIDATION GUIDE-LINES FOR HIGH INTEGRITY SYSTEMS. Appendices A-D.

VEEH,R.H.

NURFG/CR-6188 V02: MICROBIAL DEGRADATION OF LOW LEVEL TANIGUCHl.M-RADIOACTIVE WASTE. Annual Report For FY 1994.

NUREG/CR4308: AN OVERVIEW OF INSTABILITY AND FINGERING DURING IMMISCIBLE FLUID FLOW IN POROUS AND FRACTURED VESELY,W.E.

MEDIA.

NUREG/CR.6002; RISK-BASED MAINTENANCE MODELING Pnontization Of Maintenance importances And Quantifica-THOMAS,C.W.

tion Of Maintenance Effectiveness.

NUREG/CR4390: RADIOLOGICAL CHARACTERIZATION OF SPENT NUREG/CR-6141; HANDBOOK OF METHODS FOR RISK-BASED CONTROL ROD ASSEMBLIES.

ANALYSES OF TECHNICAL SPECIFICATIONS.

NUREG/CR-6172: REVIEWING PSA-BASED ANALYSES TO MODIFY THOMAS,W.

TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

NUREG/CR-6224: PARAMETRIC STUDY OF THE POTENTIAL FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED VIGIL,R.A.

DEBRIS.

NUREG/CR-6220. AN ASSESSMENT OF FIRE VULNERABILITY FOR AGED ELECTRICAL RELAYS.

THOMPSON,M.G.

NUREG/CR 6354 DRF FC: PERFORMANCE TESTING OF ELECTRON-WACHTELJ.A.

IC PERSONAL DOSIMETERS Draft Report For Comment.

NUREG-0700 ROI DFC: HUMAN-SYSTEM INTERFACE DESIGN REVIEW GUIDELINE. Draft Repon For Comment.

NUREG/CR4119 VOI: MELCOR COMPUTER CODE MANUALS.Pnmer WAGNER,K.L And User's GJedes. Version 1.8 3 Septernier 1994.

NUREG/CR-6291 VOI: NUCLEAR PLANT ANALYZER installation NUREG/CR4119 V02:

MELCOR COMPUTER CODE Manual.

MANUALS Reference Manuals Version 18.3 September 1994.

NUREG/CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

ence Manual.

TICHLERJ.

NUREG/CR4291 V03. NUCLEAR PLANT ANALYZER. Computer Visual NUREG/CR-2907 V13; RADIOACTIVE MATERIALS RELEASED FROM System Reference Manual.

NUCLEAR POWER PLANTS Annual Report 1992.

NUHEG/CR4291 V04: NUCLEAR PLANT ANALYZER. Programmer's NUREG/CR-2907 V14: RADIOACTIVE MATERIALS RELEASED FROM Manual.

NUCLEAR POWER PLANTS.

W ANG.H-B.

TORTORELLtd.P.

NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-NUREG/CP-0144 V01: A WORKSHOP ON DEVELOPING RISK ASSESS-TLONS REVtEW TEAM MENT METHODS FOR MEDICAL USE OF RADIOACTIVE MATERIAL. Summary.

WARE A.G.

NUREG/CP.0144 V02: A WORKSHOP ON DEVELOPING RISK ASSESS-NUREG/CR4260: APPLICATION OF NUREG/CR-5999 INTERIM FA-MENT METHODS FOR MEDICAL USE OF RADIOACTIVE TIGUE CUOVES TO SELECTED NUCLEAR POWER PLANT COMPO-MATERIALSupporti Documents NENTS.

NUREG/CR4276: A PILATION OF CURRENT REGULATIONS, STANDARDS. AND GUIDELINES IN REMOTE AFTERLOADING BRA.

WATKINS,B.

CHYTHERAPY.

NUREG 1415 V08 NO1: OFFICE OF THHE INSPECTOR GENERALSemiannual Report To Cengress.Apnt 1,1995 - September TRAVISJ.R.

30,1995.

NUREG/CR4318: DATA

SUMMARY

REPORT FOR FISSION PRODUCT RELEASE TEST VI-7.

WATKINS).C.

NUREG/CR4100. GATE VALVE AND MOTOR OPERATOR RESEARCH TRAVIS RJ.

FINDINGS.

NUREG/CR 5954: EFFECT ON AGING ON PWR CHEMICAL AND VOLUME CONTROL SYSTEM.

WEATHERBYJ.R.

NUREG/CR 6184: SEPARATE EFFECTS TESTING AND ANALYSES TO TRUMMER,DJ-INVESTIGATE LINER TEARING OF THE 116-SCALE REINFORCED p

NUREG/CR4242: CASKS (COMPUTER ANALYSIS OF STORAGE CONCRETE CONT AINMENT ButLDtNG.

CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR STORAGE CASK DESIGN REVIEW. User's Manual To Version 1b (In-WEBER.M.F.

ciuding Program Reference).

NUREG-1444 S01: SITE DECOMMISSIONING MANAGEMENT PLAN

Personal Author index 65 WESSTER,C.S.

WILSON R.

NUREG/CR4318: DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR-5758 V05. FITNESS FOR DUTY IN THE NUCLEAR POWER RELEASE TEST VI.7, INDUSTRY. Annual Summary Of Program Performance Reports CY 1994.

WEINSTEIN,E.D.

NUR G 15 4 iA FOR A LARGE TABLETOP EXERCISE FOR A EG C 4'116 V09: SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

VERSION 5.0.Venfication And Vahdation (V& Manual.

WESCOTT,R'G'. NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE NUREG/CR4116 VIO: SYSTEMS ANALY IS PROGRAMS FOR NUREG-1464

2. Development Of Capabihtes For Review Of A Performance Assess.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE) ment For A High-Level Waste Repository.

VERSION 5 0. Data Loading Manual.

WREATHALL,J.

WEST RA,C.

NUREG/CR-6265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN NUREG/CR 5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF INDUSTRY. Annual Summary Of Program Performance Reports CY COMMISSION AND DEPENDENCIES.

1994.

WRIGHT,J.E.

WHITf.J.E.

NUREG/CR-6327: MODELS FOR EMBRITTLEMENT RCCOVERY DUE NUIMG/CR4214 PRODUCTION AND TESTING C'F THE VITAMIN-B6 TO ANNEALING OF REACTOR PPESSURE VESSEL STEELS.

FUE GROUP AND THE BUGLE-93 BROAD-CROUP NEUTRON /

PVOTON CROSS-SECTION LIBRARIES DERIVER > FROM ENDF/B-VI UAEG 6214: PRODUCTION AND TESTING OF THE VITAMIN-B6 NUCLEAR DATA.

FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

WHITEHEAD.D W PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-VI NUREG/CRkl13 V01: EVALUATION OF POTENTl/L SEVERE ACCl-NUR / R43 ADEQUACY OF THE 123-GROUP CROSS-SECTION DENTS DURING LOW POWER AND SHUTDOWN UPERATIONS AT LIBRARY FOR CRITICALITY ANALYSES OF WATER-MODERATED GRAND GULF. UNIT 1 Summary Of Results, URANIUM SYSTEMS.

NUREG/CR-6143 V06 Pt: EVALUATION OF POTENTIAL SEVERE AC-CfDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT X1A L.

GRAND GULF. UNIT 1. Evaluation Of Severe Accident Risks For Plant NUREG/CR4264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE Operational State 5 Dunng A Refuehng Outage Main Report And Ap-DETERMINATION.A Osmputational Approach To Ductile Crack Growth pendees.

Under Large-Scale Yielding Conditions.

WISLIN,C.

KlE,J.W.

NUREG/CR-6307:

SUMMARY

OF COMMENTS RECElVED AT WORK, NUREG/CR-3469 V08: OCCUPATIONAL DOSE REDUCTION AT NU-SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-FACILITATE PUBLIC PARTICIPATION IN DECOMMISSIONING ED READINGS IN RADIATION PROTECTION AND ALARA.

CASES-YANG.K.T.

WICKLIFF HlCKS NUREG/CR-6017: FIRE MODELING OF THE HEISS DAMPF REAKTOR CONTAINMENT

• NUREG/CR 6256 V02: FIELD LYSIMETER INVESTIGATIONS - TEST RESULTS Low 4 eve! Waste Data Base Development Program: Test YAROUMIAN,J.

Results For Fiscal Years 1990,1991,1992, And 1993.

NUREG-0525 V02 R03. SAFEGUARDS

SUMMARY

EVENT LIST (SSEL). January 1,1990 Through December 31,1994.

NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING YOUNG.M.L WELDS Seventh Program Report March 1993 December 1994.

NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR400( PROBABILISTIC PIPE FRACTURE EVALUATIONS UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainry FOR LEAK RATE DETECTION APPLICATIONS.

Assessment. Main Report.

NUREG/CR 6235. ASSESSMENT OF SHORT THROUGH-WALL CIR-NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis. March UNCERTAINTY ANALYSIS. Dispersion And Deposition Uryertainty 1990 - December 1994 AssessmentAppendices A And B.

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONE 60UENCE NUREG/CR4251: STAINLESS STEEL SUBMERGED ARC WELD FUSION LINE TOUGHNESS.

UNCERTAINTY ANALYSIS. Dispersion and Deposition Uncertainty NUREG/CR4297: FRACTURE EVALUATIONS OF FUSION LINE AssessmenLAppendices C.D.E,F G,H.

CRACKS IN NUCLEAR PIPE BIMETALLIC WELDS.

2EIGLER,S.L NUREG/CR-6298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR TLALLY SURFACE-CRACKED P1PE-HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHlHE)

NUREG/CR4299: EFFECTS OF TOUGHNESS ANISOTROPY AND COMBINED TENSION, TORSION. AND BENDING LOADS ON FRAC.

VERSION 5.0.Venfication And Vahdation (V&V) Manual.

TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

ZlGLER,G.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK OPEN-NUREG/CR4224: PARAMETRIC STUDY OF THE POTENTIAL FOR ING AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH WALL BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED CRACKS IN PIPES.

DEBRIS.

l l

l l

l l

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

are welcome.

ACRS Report Accident Sequence NUREG-1125 V16 A COMPILATION OF REPORTS OF THE ADVISORY NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE COMMITTEE ON REACTOR SAFEGUARDS 1994 Annual DAMAGE ACCIDENTS 1994 A STATUS REPORT. Main Report And Ap-pendices A-H.

AEOD NUREG/CR-4674 V22. PRECURSORS TO POTENTIAL SEVERE CORE NUREG-1272 V06 NO2: OFFICE FOR ANALYSIS AND EVALUATION OF DAMAGE ACCIDENTS:1994 A STATUS REPORT. Appendix L OPERATIONAL DATA.1993 Annual Report - Nuclear Matenals.

Accounting ALARA NUREG-1280 R01: STANDARD FORMAT AND CONTENT ACCEPT.

NUREG/CP 0143, PROCEEDINGS OF THE THIRD INTERNATIONAL ANCE CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNT.

WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR ING (MC&A) REFORM AMENDMENT.

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

NUREG/CR-3469 V08. OCCUPATIONAL DOSE REDUCTION AT NU-Advanced Light Water Reactor CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-NUREG/CR 6358 VO1: ASSESSMENT OF UNITED STATES INDUSTRY ED READINGS IN RADIATION PROTECTION AND ALARA.

STRUCTURAL CODES AND STANDARDS FOR APPLICATION TO ADVANCED NUCLEAR POWER REACTORS Final Report.

ANS NUREG/CR-6358 V02: ASSESSMENT OF UNITED STATES INDUSTRY l

NUREG/CR-6385: COMPARISONS OF ANS.ASME.AWS, AND NFPA STRUCTURAL CODES AND STANDARDS FOR APPLICATION TO STANDARDS CITED IN THE NRC ST ANDARD REVIEW ADVANCED NUCLEAR POWER REACTORS. Appendices.

PLAN.NUREG-0800, AND RELATED DOCUMENTS.

Advanced Reactor ANSI NUREG/CR-6312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN i

NUREG/CR.6386: COMPARISONS OF ANSI STANDARDS CITED IN SORS.

THE NRC STANDARD REVIEW PLAN,NUREG 0800, AND RELATED DOCUMENTS.

Advisory Committee On Nuclear Weste NUREG-1423 V05-A COMPILATION OF REPORTS OF THE ADVISORY ASMd COMMITTEE ON NUCLEAR WASTE. July 1993 - June 1995.

NUREG/CR-6385: COMPARISONS OF ANS.ASME.AWS, AND NFPA STANDARDS CITED IN THE NRC STANDARD REVIEW Aging i

PLAN.NUREG-0800, AND RELATED DOCUMENTS NUREG/CR-5462: AGING STUDY OF BOILING WATER REACTOR i

HIGH PRESSURE INJECTION SYSTEMS LSME Code NUREG/CR-5857. AGING OF TURBINE DRIVES FOR SAFETY-RELAT.

NUREG-1482: GUIDELINES FOR INSERVICE TESTING AT NUCLEAR ED PUMPS !N NUCLEAR POWER PLANTS.

POWER PLANTS.

NUREG/CR-5944 V02. A CHARACTERIZATION OF CHECK VALVE l

NUREG/CR-3243. COMPARISON OF ASME CODE FATIGUE EVALUA-DEGRADATION AND FAILURE EXPERIENCE iN THE NUCLEAR TION METHODS FOR NUCLEAR CLASS 1 PIPING WITH CLASS 2 POWER INDUSTRY.1991 Failures.

OR 3 PIPING.

NUREG/CR-5954 EFFECT ON AGING ON PWR CHEMICAL AND VOLUME CONTROL SYSTEM.

ASTM NUREG/C46016: AGING AND SERVICE WEAR OF AIR-OPERATED 1

NUREG/CR-6382. COMPARISONS OF ASTM STANDARDS CITED IN VALVES USED IN SAFETY-RELATED SYSTEMS AT NUCLEAR THE NRC STANDARD REVIEW PLAN.NUREG-0800 AND RELATED POWER PLANTS DOCUMENTS.

NUREG/CR-6089 DETECTION OF PUMP DEGRADATION NUREG/CR-6192: AGING AND SERVICE WEAR OF SPRING-LOADED i

AUTOCASK PRESSURE RELIEF VALVES USED IN SAFETY-RELATED SYSTEMS NUREG/CR-5657: AUTOCASK (AUTOMATIC GENERATION OF 3-D AT NUCLEAR POWER PLANTS CASK MODELS).A Microcomputer Based System For Shipping Cask NUREG/CR-6220: AN ASSESSMENT OF FIRE VULNERABILITY FOR Design Review Analysis.

AGED ELECTR CAL RELAVS Abnormal Occurrence Aging Research NUREG-0090 V17 N04: REPORT TO CONGRESS ON ABNORMAL NUREG/CP 0140 V03. PROCEEDINGS OF THE TWENTY-SECOND OCCURRENCES. October-December 1994=

WATER REACTOR SAFETY INFORMATION MEETING Pnmary Sys-NUREG-0090 V18 Nol: REPORT TO CONGRESS ON ABNORMAL tems Integnty, Structural And Seismic Enginee ing. Aging Research.

OCCURRENCES. January-March 1995.

Products And Applications Accelerometer System A6r Flow Velocity NUREG/CR-6313 VO1: ROBUST, ACCURATE, AND NON-CONTACTING NUREG/CR-6334. NEW SENSOR FOR MEASUREMENT OF LOW AIR V1BRATION MEASUREMENT SYSTEM. Summary of Companson Meas-FLOW VELOCITY Phase i Final Report urements Of The Robust Laser Interferometer And Typical Accelerome-ter Systems.

Air Operated Valve NOREG/CR-6313 V02-ROBUST, ACCURATE, AND NON-CONTACTING NUREG/CR-6016 AGING AND SERVICE WEAR OF AIR-OPERATED VIBRATION MEASUREMENT SYSTEMS Supplemental Appendices VALVES USED IN SAFETY-RELATED SYSTEMS AT NUCLEAR 1

Presenting Companson Measurements Of The Robust Laser Interfer-POWER PLANTS orpeter And Typical Accelerometer Systems Airborne Effluent Acceptance Critens NUREG/CR-2907 V13 RADIOACTIVE MATERIALS RELEASED FROM NUREG-1280 R01: STANDARD FORMAT AND CONTENT ACCEPT-NUCLEAR POWER PLANTS Annual Report 1992 ANCE CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNT.

NUREG/CR-2907 V14 RADIOACTIVE, MATERIALS RELEASED FROM ING (MC&A) REFORM AMENDMENT.

NUCLEAR POWER PLANTS l

67 i

1

.I

68 Subject index Alertnese Bianial Loading NUREG/CR-6046' ALERTNESS. PERFORMANCE. AND OFF-DUTY NUREG/CR4273: BlAXIAL LOADING EFFECTS ON FRACTURE SLEEP ON 8 HOUR AND 12-HOUR NIGHT SHIFTS IN A SIMULATED TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL CONTINUOUS OPERATIONS CONTROL HOOM SETTING.

Bemetall6c Wold Allegations Rev6ew Team NUREG/CR 6297: FRACTURE EVALUATIONS OF FUSION LINE NUREG 1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-CRACKS IN NUCLEAR PIPE BIMETALLIC WELDS.

TIONS REVIEW TEAM.

Boiling Water Reactor Annealing NUREG 1123 RO1; KNOWLEDGE AND ABILITIES CATALOG FOR NU-NUREG/CR4327. MODELS FOR EMBRITTLEMENT RECOVERY DUE CLEAR POWER PLANT OPERATORS: BOILING WATER REACTORS.

TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS NUREG/CR-5462: AGING STUOY OF BOILING WATER REACTOR HIGH PRESSURE INJECTION SYSTEMS.

Annual Report NUREG/CR 6224. PARAMETRIC STUDY OF THE POTENTIAL FOR NUREG-1145 VII: U S. NUCLEAR REGULATORY COMMISSION 1994 BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED ANNUAL REPORT.

DEBRIS Atmospheric Depcsit6on Boric Acid NUREG/CR-6244 V03: PROBABILISTIC ACCIDENT CONSEOUENCE NUREG-1519: SURFACE INTERACTIONS OF CESIUM AND BORIC UNCERTAINTY ANALYSIS. Dispersion and Depositon Uncertainty ACID WITH STAINLESS STEEL Assessment Appendices C.D.E F,G.H Boron Atmospheric D6spersion NUREG/CR-6266: ANALYSIS OF BORON DILUTION IN A FOUR LOOP NUREG/CR-6244 VOI: PROBABILISTIC ACCIDENT CONSEQUENCE PWR UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty Assessment Main Report.

Bounding Spectra NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSEOUENCE NUREG/CR4240: APPLICATION OF BOUNDING SPECTRA TO SEIS-UNCERTAINTY ANALYSIS Dispersion And Depositon Uncertainty MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF Assessment. Appendices A And B.

ABOVE GROUND PtPING IN POWER PLANTS SUBJECTED TO NUREG/CR-6331; ATMOSPHERIC RELATIVE CONCENTRATIONS IN STRONG MOTION EARTHOUAKES.

I BUILDING WAKES.

Brachytherapy Atomic Safety And Licensing Board Panet NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF REMOTE NUREG-1363 V06: ATOMIC SAFFTY AND LICENSING BOARD PANEL AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks BIENNIAL REPORT, Fiscal Years 1993 - 1994-in Remote Afterloading Brachytherapy And Approaches For improved NUlh CR 6 2 VO [ HUMAN FACTORS EVALUATION OF REMOTE Austenit6c Stainless Steel NUREG/CR-6335: FATIGUE STRAIN-LIFE BEHAVIOR OF CARBON AND LOW-ALLOY STEELS, AUSTENITIC STAINLESS STEELS. AND AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

NUREG/CR-6276: A COMPILATION OF CURRENT REGULATIONS, ALLOY 600 IN LWR ENVIRONMENTS.

STANDARDS. AND GUIDELINES IN REMOTE AFTERLOADING BRA-Austenitic Steel CHYTHERAPY.

NUREG/CR-6235: ASSESSMENT OF SHORT THROUGH-WALL CIR-1 CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis. March N EG CR 2: BUCKLING ANALYSIS OF SPENT FUEL BASKET.

1990 - December 1994.

Aserege Dose Budget NUREG 0713 V15. OCCUPATIONAL RADIATION EXPOSURE AT COM.

NUREG-1100 VII: BUDGET ESTIMATES Fiscal Years 1996-1997.

MERCIAL NUCLEAR POWER REACTORS AND OTHER Building Wake FACILITIES.1993. Twenty-Sath Annual Report.

NUREG/CR-6331: ATMOSPHERIC RELATIVE CONCENTRATIONS IN

[REATN Version 1 BUILDING WAKES NUREG/CR 6333. BREATH VERSION 1.1 COUPLED FLOW AND CANDU3 ENERGY TRANSPORT IN POROUS MEDIA. Simulator Desenption And NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC ENERGY OF CANADA LTD CODES.

EUGLE-93 CANDU Reactor NUREG/CR-6214. PRODUCTION AND TESTING OF THE VITAMIN-86 FINE GROUP AND THE BUCLE-93 BROAD-GROUP NEUTRON /

NUREG/CFI-6315: CANDU REACTORS. THEIR REGULATION IN PHOTON CROSS SECTION LIBRARIES DERIVED FROM SNDF/B-VI CANADA, AND THE IDENTIFICATION OF RELEVANT NRC SAFETY NUCLEAR DATA.

ISSUES.

EWR CASKS NUREG-1123 Rol: KNOWLEDGE AND ABILITIES CATALOG FOR NU.

NUREG/CR 6242: CASKS (COMPUTER ANALYSIS OF STORAGE CLEAR POWER PLANT OPERATORS: BOILING WATER REACTORS.

CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR NUREG/CR-5462: AGING STUDY OF BOILING WATER REACTOR STORAGE CASK DESIGN REVIEW User's Manual lo Version ib (In-HIGH PRESSURE INJECTION SYSTEMS cluding Program Reference).

NUREG/CR-6224 PARAMETRIC STUDY OF THE POTENTIAL FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED COPMA-il

DEBRIS, NUREG/CR4398. EVALUATION OF THE CCMPUTERIZED PROCE-DURES MANUAL If (COPMA.It).

Baldcypress NUREG/GR-0014. BALDCYPRESS TREE RING ELEMENTAL CONCEN-Calibration TRATIONS AT REELFOOT LAKE. TENNESSEE,FROM AD 1795 TO NUREG/CR4343 ON-LINE TESTING OF CAllBRATION OF PROCESS AD 1820.

INSTRUMENTATION CHANNELS IN NUCLEAR POWER PLANTS Phase il Fost Report.

NUREG/CR-6154 V02: EXPERIMENTAL RESULTS FROM CONTAIN.

Cand6dete Guide 6tne CNT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT NUREG/CR4263 VOI: HIGH INTR _GRITY SOFTWARE FOR NUCLEAR w t DITIONS Results From Bellows Tested im CorrMed Condstsons.

POWER PLANTS.Candedate Ce d*nes, Techracal Basis And Research Needs Executwo Summary Bond Specimen NUREG/CR-6263 V02: HiGH WTEGRITY SOFTWARE FOR NUCLEAR NUREGICR-6191: SIZE AND DEFORMATION LIMITS TO MAINTAIN POWER PLANTS Candidab Guidelines, Ter:hnical Base And Reneerch CONSTRANT IN K(IC) AND J(C) TESTING OF BEND SPECIMENS.

Needs Main Report.

)

[

Subject index 69 Capillary Barr6er NUREG/CR-0200 V3 R04. SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR-4918 V08: CONTROL OF WATER INFILTRATION INTO PERFORMING Sl ANDARDIZED COMPUTER ANALYSES FOR Lt.

NEAR SURFACE LLW DISPOSAL UNITS. Progress Report Of Field Ex-CENSING EVALUATION. Miscellaneous.

penments At A Humid Region Site,Beltsville, Maryland.

Computertred Procedure Certif6 cates Of Compliance NUREG/CR-6398: EVALUATION OF THE COMPUTERIZED PROCE-NUREG 0383 V01 R18. DIRECTORY OF CERTIFICATES OF COMPLl-DURES MANUAL ll (COPMA ll).

ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC Approved Packages-Concrete NUREG-0383 V02 R18 DIRECTORY OF CERTIFICATES OF COMPL).

NUREG/CR-6188 V02: MICROBIAL DEGRADATION OF LOW-LEVEL ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of RADIOACTIVE WASTE. Annual Report For FY 1994.

Corroliance.

NUREG 0383 V03 R15. DIRECTORY OF CERTIFICATES OF COMPL1-Constraint Effect ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC NUREG/CR-6259: CONSTRAINT EFFECTS ON FRACTURE INITIATION Approved Quality Assurance Programs for Radioactive Matenals Pack-LOADS IN HSST WIDE-PLATE TESTS.

ages.

Cesium Hydroxid, Construction inspect 6on Program NUREG 1519. SURFACE INTERACTIONS OF CESIUM AND BORIC NUREG-1528: RECONSTITUTION OF THE MANUAL CHAPTER 2512 ACID WITH STAINLESS STEEL.

CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT 1.

Check Valve Containment NUREG/CR 5944 V02: A CHARACTERIZATION OF CHECK VALVE NUREG 1493: PERFORMANCE-BASED CONTAINMENT LEAK TEST DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLEAR PROGRAM. Final Report.

POWER INDUSTRY.1991 Failures.

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

Circumferential Crack NUREG 1522: ASSESSMENT OF INSERVICE CONDITIONS OF NUREG/CR-6235 ASSESSMENT OF SHORT THROUGH-WALL CIR.

SAFETY-RELATED NUCLEAR PLANT STRUCTURES.

CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis, March NUREG/CR-6017: FIRE MODELING OF THE HEISS DAMPF REAKTOR 1990. December 1994 CONTAINMENT NUREG/CR-6299: EFFECTS OF TOUGHNESS ANISOTROPY AND NUREG/CR-6154 V02: EXPERIMENTAL RESULTS FROM CONTAIN-COMBINED TENSION, TORSION. AND BENDING LOADS ON FRAC-MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE-CONDITIONS Results From Bellows Tested in Corroded Conditions.

Circumferential Surface Crack NUREG/CR-6298. FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN.

Containment Bulading TIALLY SURFACE CRACKED PIPE.

NUREG/CR 6184: SEPARATE EFFECTS TESTING AND ANALYSES TO INVESTIGATE LINER TEARING OF THE 1:16 SCALE REINFORCED Civil Penalty CONCRETE CONTAINMENT BUILDING.

NUREG 1525: ASSESSMENT OF THE NRC ENFORCEMENT PRO-GRAM.

Containment Failure NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY REG /CR-6382: COMPARISONS OF ASTM STANDARDS CITED IN THE NRC STANDARD REVIEW PLAN,NUREG-0800, AND RELATED Contaminant Transport

^

^

NU G/C 63 5. COMPARISONS OF ANS,ASME.AWS. AND NFPA FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-LEVEL STANDARDS CITED IN THE NRC STANDARD REVIEW WASTE FACILITY.Two-Phase Flow And Contaminant Transport in Un-saturated Soils With Application To Low-Level Radioactive Waste Dis.

NU C 63 6 C M R N F NS STANDARDS CITED IN THE NRC STANDARD REVIEW PLAN.NUREG 0800. AND RELATED posat DOCUMENTS.

Control Rod Code Assessment NUREG/CR-6390: RADIOLOGICAL CHARACTERIZATION OF SPENT NUREG/CR 6285. SEVERE ACCIDENT NATURAL CIRCULATION CONTROL ROD ASSEMBLIES.

STUDIES AT THE INEL Control Room Code System NUREG/CR 6046: ALERTNESS, PERFORMANCE, AND OFF-DUTY NUREG/CR 0200 V1 R04: SCALE: A MODULAR CODE SYSTEM FOR SLEEP ON 8-HOUR AND 12 HOUR NIGHT SHIFTS IN A SIMULATED PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-CONTINUOUS OPERATIONS CONTROL ROOM SETTING CENSING EVALUATION Control Modules.

NUREG/CR-6159 USING MICRO SAINT TO PREDICT PERFORMANCE NUREG/CR 0200 V2PIR4: SCALE: A MODULAR CODE SYSTEM FOR IN A NUCLEAR POWER PLANT CONTROL ROOM A Test Of Valglity PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-And Feasiblity CENSING EVALUATION Functional Modules F1 F8 NUREG/CA-6331-ATMOSPHERIC RELATIVE CONCENTRATIONS IN NUREG/CR-0200 V2P2H4. SCALE: A MODULAR CODE SYSTEM FOR BUILDING WAKES.

PERFORMING STANDARDI2ED COMPUTER ANALYSES FOR Ll-CENSING EVALUATION Functional Modules F9-F16 Convert.5on Factor NUREG/CR 0200 V3 R04: SCALE: A MODULAR CODE SYSTEM FOR NUREG-1530. REASSESSMENT OF NRC'S DOLLAR PER PERSON-PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-REM CONVERSION FACTOR POLICY.

CENSING EVALUATION. Miscellaneous C lant Cleanup System Cee And Shdarda NUREG/CR 5954. EFFECT ON AGING ON PWR CHEMICAL AND NUREG/CR 5973 R02: CODES AND STANDARDS AND OTHER GUID.

VOLUME CONTROL SYSTEM ANCE CITED IN REGULATORY DOCUMENTS.

Core Damage Computer Anatyees NUREG/CR-0200 V1 R04: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR 4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll.

DAMAGE ACCIDENTS 1994 A STATUS REPORT Main Report And Ap-CENSING EVALUATION Control Modules.

pendices A H NUREG/CR4200 V2PIR4. SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-DAMAGE ACCIDENTS 1994 A STATUS REPORT.Appendm t.

CENSING EVALUATION Functional Moduies F1-F8.

NUREG/CR 0200 V2P2R4. SCALE: A MODULAR CODE SYSTEM FOR Core Meltdown PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-NUREG-1465: ACCIDENT SOURCE TFAMS FOR LIGHT WATER NU-CENSING EVALUATION. Functional Modubs F9-F16.

CLEAR POWER PLANTS l

l l

... ~,

-~- _

70 Subject Index Corroelon Fatigue NUREG/CR-6307:

SUMMARY

OF COMMENTS RECEIVED AT WORK-NUREG/CR-4667 V18 ENVIRONMENTALLY ASSISTED CRACKING IN SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO LIGHT WATER REACTORS. Semiannual Report, October 1993 - March FACILITATE PUBLIC PARTICIPATION IN DECOMMISSIONING 1994.

CASES.

NUREG/CR 4667 V19. ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6390: RADIOLOGICAL CHARACTERIZATION OF SPENT LIGHT WATER REACTORS Semiannual Report.Apnt-September 1994 CONTROL ROD ASSEMGLIES.

Cost Benefit Degradation NUREG/CR4349: COST-BENEFIT CONSIDERATIONS IN REGULA.

NUREG-1522: ASSESSMENT OF INSERVICE CONDITIONS OF TORY ANALYSIS-SAFETY-RELATED NUCLEAR PLANT STRUCTURES.

NUREG/CR 5944 V02: A CHARACTERIZATION OF CHECK VALVE Cost Estimate NUREG/CR-6054: ESTIMATING PRESSURIZED WATER REACTOR DE-DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLEAR COMMISSIONING COSTS. A User's Manual For The PWR Cost Esti-NUREG E I f PUMP DEGRADATION, mating Computer Program (CECP) Software.

Crack Diffusion Coefficient NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK-OPEN.

NUREG/CR-6261: A

SUMMARY

OF ORNL FISSION PRODUCT RE.

ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL LEASE TESTS WITH RECOMMENDED RELEASE RATES AND DIF-l CRACKS IN PIPES.

FUSION COEFFICIENTS.

Crack Growth Dilut6on j

NUREG/CR-6264 VOI: VALIDITY LIMITS IN J-RESISTANCE CURVE NUREG/CR4268 ANALYSIS OF BORON DILUTION IN A OUR-LOOP j

DETERMINATION An Assessment Of The J(M) Parameter.

PWR.

NUREG/CR-6264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE l

DETERMINATION A Computational Approach To Ductile Crack Growth D6 rect Containment Heating i

Under Large-Scale Yielding Conditions.

NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY DIRECT CONTAINMENT HEATING IN SURRY.

NUREG/CR-6335: FATIGUE STRAIN-LIFE BEHAVIOR OF CARBON Discrimination losue AND LOW ALLOY STEELS. AUSTENITIC STAINLESS STEELS, AND w p,CG-1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-ALLOY 600 IN LWR ENVIRONMENTS-

}&S REVIEW TEAM.

Criticailty Safety Dose Commitment NUREG/CR-6328. ADEQUACY OF THE 123 GROUP CROSS-SECTION NUREG/CR 2850 V13: DOSE COMMnVQTS DUE TO RADIOACTIVE LIBRARY FOR CRITICALITY ANALYSES OF WATER MODERATED RELEASES FROM NUCLEAR POWER PLANT SITES IN 1991, URANIUM SYSTEMS.

Dosknew Cross-Sect 6on Library NUREG/CR 6354 DRF FC: PERFORMANCE TESTING OF ELECTRON-NUREG/CR-6328. ADEQUACY OF THE 123-GROUP CROSS-SECTION IC PERSONAL DOSIMETERS. Draft Report For Comment.

LIBRARY FOR CRITICALITY ANALYSES OF WATER-MODERATED URANIUM SYSTEMS.

Dosimetry NUREG/CR-6112: IMPACT OF REDUCED DOSE LIMITS ON NRC Ll-Damage Progression NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 3.1 CODE CENSED ACTIVITIES. Major issues in The implementation Of ICRP/

MANUAL. Damage Progression Model Theory.

NCRP Dose Limit Recommendatens Final Report.

Data Summary Report Dusty Gas Model NUREG/CR4318 DATA

SUMMARY

REPORT FOR FISSION PRODUCT NUREG/CR-6347: MULTI-PHASE REACTIVE TRANSPORT THEORY.

1 RELEASE TEST VI-7.

l ECCs Debris NUREG/CR-6224-PARAMETRIC STUDY OF THE POTENTIAL FOR NUREG/CR-6224: PARAMETRIC STUDY OF THE POTEN11AL FOR BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED DEBRIS.

DEBRIS NUREG/CR 6368 EXPERIMENTAL INVESTIGATION OF SEDiMENTA-ENDF/8-VI Nuclear Data TION OF LOCA GENERATED FIBROUS DEBRIS AND SLUDGE IN NUREG/CR-6214. PRODUCTION AND TESTING OF THE VITAMIN-86 i

BWR SUPPRESSION POOLS-FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

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

NUREG 1307 RO5: REPORT ON WASTE BURIAL CHARGES Escalation Of Decommissoning Waste Disposal Costs At Low-Level Waste Bunal EPICOR-Il NUk 44 S01' SITE DECOMMISSIONING MANAGEMENT PLAN NUREG/CR-5229 V07. FIELD LYSIMETER INVESTIGATIONS: LOW-NUREG 1505 DRFT FC: A NONPARAMETRIC STATISTICAL METHOD _

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE.

FISCAL YEAR 1994. Annual Report.

NUREG/CR-6256 VOI: FIELD LYSIMETER INVESTIGATIONS TEST COMMISSIONING SURVEYS Draft Repor1 For Comment NUREG-1506 DRFT FC: MEASUREMENT METHODS FOR RADIOLOGl.

RESULTSLow4evel Waste Data Base Program. Test Results For CAL SURVEYS IN SUPPORT OF NEW DECOMMISSIONING Fiscal Years 1986,1987,1988, And 1989 CRITERIA Draft Report For Comment.

NUREG-1507 DRFT FC. M!NIMUM DETECTABLE CONCENTRATIONS Earthquake WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS NUREG/CR4239 V01: SURVEY OF STRONG MOTION EARTHOUAKE CONTAMINANTS AND FIELD CONDITICNS Draft Report For Com-EFFECTS ON THERMAL POWER Pt ANT 9 IN CALIFORNIA WITH EM-ment PHASIS ON PIPING SYSTEMSMain Reoort NUREG/CR-5884 V01 REVl SED ANALYSIS OF DECOMMISSIONING NUREG/CR 6239 V02. SURV8N OF STRONG MOTION EARTHOUAKE FOR THE REFERENCE PnESSURIZED WATER REACTOR POWER EFFECTS ON THERMAL 'OWER PLANTS IW CALIFORNIA WITH EM-STATION Effects Of Current Regulatory And Other Considerations On PHASIS ON PIPING SYS TEMS Appene.:es The Fcancial insurance Re:iuirements Of Tne Decommmasoning Rule NUREG/CR-6240 APPLIC \\ TION OF POUNDING SPECTRA TO SEIS-And..

MIC DESIGN OF PIPtflG P*. SED ON THE PERFORMANCE OF NUREGrCR-5884 V02 REVISED ANALYSIS OF DECOMMISSIONING ABOVE GROUND PIPilC IN POWER PLANTS SUBJECTED TO FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER STRONG MO h0N E ARTHOUAKES.

STATION Ettects Of Cunert Hegulatory And Other Considerahons On NUREG/CR-6283 FIELD SITE INVESTIGATION. EFFECT OF MINE The Financial insurance Reqwements Of The Decomnyssoning Rule SEISMICITY ON GROUNDWATER HYDROLOGY And.

NUREG/CR-6054 ESTIMATAG PPESSURtZED WATER REACTOR DE-Embrittiement COVM?SSIONING COSTS A Wer s Manual For The PWR Cost Estb NUREG/CR-5591 V05 H2. HE AVY-SECTION STEEL HnADIATION rnaung Computer Program (CECF) Scitware PROGRAM Progress Report For Apnl 1994 T% ugh E glemter iss94

l Subject Index 71 i

NUREG/CR 6275: MECHANICAL PROPERTIES OF THERMALLY AGED NUREG 1123 R01: KNOWLEDGE AND ABILITIES CATALOG FOR NU-CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-CLEAR POWER PLANT OPERATORS: BOILING WATER REACTORS.

PONENTS.

NUREG/CR-6327: MODELS FOR EMBRITTLEMENT RECOVERY DUE Examiner Standards l

TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

NUREG-1478 R01: NON-POWER REACTOR OPERATOR LICENSING EXAMINER STANDARDS.

NUREG/CR-6310: AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO-Exercise PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-NUREG 1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A l

CLEAR ACCIDENT-NUCLEAR POWER PLANT.

Emergency Response Failure Mode NUR G 15 4 GUIDAN FOR A LARGE TABLETOP EXERCISE FOR A NUREG/CP 0145:

WORKSHOP ON DEVELOPING SAFE SOFTWARE. Held At Hotel Del Coronado, San Diego,CA, July 22-Energy Transport 23.1992.

NUREG/CR-6333. BREATH VERSION 1.1 COUPLED FLOW AND ENERGY TRANSPORT IN POROUS MEDIA Simulator Descnption And Failure Mode Analysis User Guide.

NUREG/CR-6002:

RISK B ASED MAINTENANCE MODELING.Priontization Of Maintenance importances And Quantifica-Enforcement Act6on tion Of Maintenance Effectiveness.

NUREG-0940 V13 N4 P1: ENFORCEMENT ACTIONS:SIGNIFICANT AC-TlONS RESOLVED REACTOR LICENSEES.Ouarterly Progress Fat 6gue i

Report. October December 1994 NUREG/CR-6046: ALERTNESS, PERFORMANCE, AND OFF DUTY t

NUREG.0940 V13 N4 P2: ENFORCEMENT ACTIONS SIGNIFICANT AC-SLEEP ON 8-HOUR AND 12 HOUR NIGHT SHIFTS IN A SIMULATED TIONS RESOLVED MEDICAL LICENSEES.Ouarterly Progress CONTINUOUS OPERATIONS CONTROL ROOM SETTING.

Report. October-December 1994.

NUREG-0940 V13 N4 P3: ENFORCEMENT ACTIONS.SIGNIFICANT AC-Fatigue Design Curve TIONS RESOLVED MATERIAL LICENSEES (NON-MEDICAL) Quarterly NUREG/CR-6335: FATIGUE STRAIN-LIFE BEHAVIOR OF CARBON Progress Report, October-December 1994-AND LOW-ALLOY STEELS, AUSTENITIC STAINLESS STEELS, AND NUREG-0940 V14 N1 P1: ENFORCEMENT ACTIONS. SIGNIFICANT AC-ALLOY 600 IN LWR ENVIRONMENTS.

TIONS RESOLVED, REACTOR LICENSEES.Quarterty Progress Report, January March 1995.

Fatigue Evaluation NUREG-0940 V14 N1 P2; ENFORCEMENT ACTIONS: SIGNIFICANT AC-NUREG/CR-3243: COMPARISON OF ASME CODE FATIGUE EVALUA-TIONS RESOLVED, MEDICAL LICENSEES.Ouarterly Progress TlON METHODS FOR NUCLEAR CLASS 1 PIPifM WITH CLASS 2 NVI 40 N1 3 NFORCEMENT ACTIONS: SIGNIFICANT AC-OR 3 PIPING.

TIONS

RESOLVED, MATERIAL LICENSEES (NON-Fiber Optic NURE 94 4f 1

FOR AT IFICANT AC-NUREG/CR-6312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN.

SORS.

TIONS RESOLVED,1NDIVIDUAL ACTIONS.Ouarterly Progress NUR U N2 P2: ENFORCEMENT ACTIONS.SIGNIFICANT AC-Field Lysimeter NUREG/CR-5229 V07: FIELD LYSIMETER INVESTIGATIONS: LOW-l TlONS RESOLVED. REACTOR LICENSEES.Ouarterly Progress Report. April-June 1995.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR l

NUREG-0940 V14 N2 P3: ENFORCEMENT ACTIONS: SIGNIFICANT AC.

FISCAL YEAR 1994. Annual Report

(

TIONS RESOLVED MATERIAL LICENSEES.Ouarterly Progress NUREG/CR-6256 VOI: FIELD LYSIMETER INVESTIGATIONS - TEST Report,Apr61-June 1995.

RESULTS Low-Level Waste Data Base Program. Test Results For i

NUREG-1600: GENERAL STATEMENT OF POLICY AND PROCEDURE Fiscal Years 1986,1987,1988, And 1989.

FOR NRC ENFORCEMENT ACTIONS Enforcement Policy.

NUREG/CR-6256 V02: FIELD LYSIMETER INVESTIGATIONS TEST RESULTS Low-Level Waste Data Base Development Program: Test i

Enforcement Policy Results For Fiscal Years 1990,1991,1992, And 1993, l

NUREG-1600: GENERAL STATEMENT OF POLICY AND PROCEDURE TOR NRC ENFORCEMENT ACTIONS. Enforcement Policy.

Final Environmental Statement NUREG 0498 S01: FINAL ENVIRONMENTAL STATEMENT RELATED Enforcement Program TO THE OPERATION OF WATTS BAR NUCLEAR PLANT UNITS 1 NUREG-1525: ASSESSMENT OF THE NRC ENFORCEMENT PRO.

AND 2. Docket Nos. 50 390 And 50-391.(Tennessee Valley Authonty)

GRAM.

i Financial Statement NUREG-1470 V04. FINANCIAL STATEMENT FOR FISCAL YEAR 1994.

N RE R-6260: APPLICATION OF NUREG/CR-5999 INTERIM FA-TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-Fire Modeling

NENTS, NUREG/CR-6017: FIRE MODELING OF THE HEISS DAMPF REAKTOR CONTAINMENT.

Environmental Simulation Approach NUREG/CR-6351: REVIEW OF SCENARIO SELECTION APPROACHES FOR PERFORMANCE ASSESSMENT OF HIGH-LEVEL WASTE RE-Fire Test ng ram POSITORIES AND RELATED ISSUES THE GERMAN HDR TEST F ACILITY.

Enytronmental Transport NUREG/CR-6305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EQUI-Fire Vittnerabmty LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS.

NUREG/CR-6220: AN ASSESSMENT OF FIRE VULNERABILITY FOR ING THE RELEASE OF RADIONUCUDES FROM LOW LEVEL WASTE AGED ELECTRICAL RELAYS j

DISPOSAL UNITS Background, Theory, And Model Descnption.

i Equilibrium Chemistry NUREG 1100 V11 BUDGET EST1 MATES Fiscal Years 1996 1997.

NUREG/CR-6305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EQUI-LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS.

Fission Product ING THE RELEASE OF RADIONUCLIDES FROM LOW LEVEL WASTE NUREG/CR-6261. A

SUMMARY

OF ORNL FISSION PRODUCT RE.

DISPOSAL UNITS 8ackground, Theory, And Model Desenption.

LEASE TESTS WITH RECOMMENDED RELE#SE RATES AND DIF-1 FUSION COEFFICIENTS l

Examination j

NUREG-1122 RO1: KNOWLEDGE AND ABluTIES CATALOG FOR NU-Fission Product Release CLEAR POWER PLANT OPERATORS' PRESSURIZED WATER REAC-NUREG/CR-6318 DATA

SUMMARY

REPORT FOR FISSION PRODUCT l

TORS.

RELEASE TEST VI.7.

1 i

w s

72 Subject Index Fitnese For Duty Geolcgic Repository NUREG/CR 5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER NUREG-1323 RO1: LICENSE APPUCATION REVIEW PLAN FOR A INDUSTRY. Annual Summary Of Program Performance Reports CY GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-1994.

LEVEL RADIOACT'VE WASTE. Draft Review Plan.

Fracture Graphical Evaluation Module NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING NUREG/CR-6116 V06: SYSTEMS ANALYSIS PROGRAMS FOR WELDS. Seventh Program Report March 1993 December 1994.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR4299: EFFECTS OF TOUGHNESS ANISOTROPY AND VERSION 5 0. Graphical Evaluation Module (GEM) Reference Manual.

COMBINED TENSION, TORSION. AND BENDING LOADS ON FRAC-TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

Groundwater NUREG/CR-6300: REFINEMENT AND EVALUATION OF CRACK-OPEN-NUREG/CR4283. FIELD SITE INVESTIGATION: EFFECT OF MINE ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL SEtSMICITY ON GROUNDWATER HYDROLOGY.

CRACKS IN PIPES.

Health Physic Fracture Behavior NUREG/CR 4334: NEW SENSOR FOR MEASUREMENT OF LOW AIR NUREG/CR-6298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-FLOW VELOCITY. Phase i Final Report.

TIALLY SURFACE-CRACKED PIPE.

Heat Pipe Phenomena Fracture Flow NUREG/CR4347: MULTI-PHASE REACTIVE TRANSPORT THEORY.

NUREG/CR-6308. AN OVERVIEW OF INSTABILITY AND FINGERING DURING IMMISCIBLE FWID FLOW IN POROUS AND FRACTURED Heavy-Section Steel Irradiation Program MEDIA.

NUREG/CR-5591 V03: HEAVY-SECTION STEEL IRRADIATION PROGRAM Progress Report For October 1991 September 1992.

Fracture Mechanics NUREG/CR-5591 V04 N2: HEAVY SECTION STEEL IRRADIATION NUREG/CR-5591 V05 N2; HEAVY SECTION STEEL IRRADIATION PROGRAM. Semiannual Progress Report For Apnl. September 1993.

PROGRAM. Progress Repr For Apnl 1994 Through September 1994.

NUREG/CR-5591 VOS N1: HEAVY SECTION STEEL IRRADIATION NUREG/CR-6004. PROBABILISTIC PIPE FRACTURE EVALUATIONS PROGRAM. Semiannual Progress Report For September 1993 Through FOR LEAK RATE-DETECTION APPLICATIONS.

March 1994.

NUREG/CR-6259 CONSTRAINT EFFECTS ON FRACTURE INITIATION NUREG/CR 5591 VOS N2: HEAVY SECTION STEEL IRRADIATION LOADS IN HSST WIDE-PLATE TESTS-PROGRAM Progress Report For Apnl 1994 Through September 1994.

NUREG/CR 6273: DIAXtAL LOADING EFFEC 1S ON FRACTURE NUREG/CR-5591 V06 N1: HEAVY SECTION STEEL IRRADIATION TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL.

PROGRAM. Semiannual Progress Report For October 1994 Through NUREG/CR-6298. FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-March 1995.

TIALLY SURFACE CRACKED PIPE.

Heavy-Section Steel Technology Program Fracture Toughness NUREG/CR-4219 V10 N2: HEAVY SECTION STEEL TECHNOLOGY NUREG/CR-6191: SIZE AND DEFORMATION LIMITS TO MAINTAIN PROGRAM Semiannual Progress Report For April-September 1993 CONSTRAINT IN Kt C) AND J(C) TESTING OF BEND SPECIMENS-NUREG/CR-4219 VII N1: HEAVY SECTION STEEL TECHNOLOGY l

NUREG/CR-6273: BIAXlAL LOADING EFFECTS ON FRACTURE PROGRAM. Semiannual Progress Report For October 1993 - March TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL.

1994' NUREG/CR-6275: MECHANICAL PROPERTIES OF THERMALLY AGED CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-Heiss Dampf Reaktor PONENTS.

NUREG/CR-6017: FIRE MODELING OF THE HEISS DAMPF REAKTOR CONTAINMENT.

NU FG/

308: AN OVERVIEW OF INSTABILITY AND FINGERING THE N DR ST FAC TY' DURING IMMISCIBLE FlutD FLOW IN POROUS AND FRACTURED MEDIA.

High integrity Software NUREG/CR 6263 VG1: HIGH INTEGRITY SOFTWARE FOR NUCLEAR POWER PLANTS Candidate Guidelines, Technical Basis And Research U EG/CR 283: FIELD SITE INVESTIGATION-EFFECT OF MINE SEISMICITY ON GROUNDWATER HYDROLOGY.

NUR /

V2 H INTEGRITY SOFTWARE FOR NUCLEAR Fuel Damage POWER PLANTS. Candidate Guidehnen, Technical Basis And Research NUREG/CR-6318. DATA

SUMMARY

REPORT FOR FISSION PRODUCT Needs.Mmn Report.

RELEASE TEST VI-7.

High integrity System Fundamental Nuclear Maternal Control NUREG/CR-6293 Vol. VERIFICATION AND VAllDATION GUIDELINES NUREG 1200 RO1: STANDARD FORMAT AND CONTENT ACCEPT.

FOR HIGH INTEGRITY SYSTEMS Main Report.

NUREG/CR-6293 V02: VERIFICATIONN AND VALIDATION GUIDE.

ANCE CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNT-ING (MC&A) REFORM AMENDMENT.

LINES FOR HIGH INTEGRITY SYSTEMS. Appendices A D.

Fusion Line Crack High Pressure In}ectlon System NUREG/CR-5462: AGING STUDY OF BOILING WATER REACTOR NUREG/CR-6297: FRACTURE EVALUATIONS OF FUSION LINE CRACKS IN NUCLEAR PIPE BIMETALLIC WELDS HIGH PRESSURE INJECTION SYSTEMS.

Gamma Knife High-Level Nuclear Waste NUREG/CR-6324. QUALITY ASSURANCE FOR GAMMA KNIVES NUREG/CP-0147; PROCEEDINGS OF THE WORKSHOP ON THE ROLE OF NATURAL ANALOGS IN GEOLOGIC DISPOSAL OF HIGH-LEVEL Gate Valve NUCLEAR WASTE. Held in San Antonio, Texas. July 22 25,1991.

NUREG/CP-0146 PROCEEDINGS OF THE WORKSHOP ON GATE VALVE PRESSURE LOCKING AND THERMAL BINDING.

HigtwLevel Radioactive Weste NUREG-1323 ROI: LICENSE APPLICATION REVIEW PLAN FOR A Gener6c Safety issues GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-NUREG-0933 S18. A PRIORITIZATION OF GENERIC SAFETY ISSUES.

LEVEL RADIOACTIVE WASTE. Draft Review Plan.

NUREG-0933 S19'.A PRIORITIZATION OF GENERIC SAFETY ISSUES.

NUREG 1435 SO4 STATUS OF SAFETY ISSUES AT LICENSED High-Level Weste Geologic Repoeltory POWER PLANTS.TMt Action Plan Requirements. Unresolved Safety NUREG/CR-6356: HYDRAULIC CHARACTER 12ATION OF HYDROTH-Issues.Genenc Safety issuer.Other Multipiant Action lasues.

ERMALLY ALTERED NOPAL TUFF.

Geolog6c Disposal High-Level Waste Repoaltory NUREG/CP 0147: PROCEEDINGS OF THE WORKSHOP ON THE ROLE NUREG-1464: NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE OF NATURAL ANALOGS IN GEOLOGIC DISPOSAL OF HIGH-LEVEL

2. Development Of Capatalities For Revww Of A Performance Assess-NUCLEAR WASTE. Held in San Antonio, Texas, July 22-25, 1991.

rnent For A High-Level Waste Reposstory.

l

Subject index 73 NUREG/CR4351: REVIEW OF SCENARO SELECTON APPROACHES Industrial Radlography FOR PERFORMANCE ASSESSMENT OF HIGH-LEVEL WASTE RE-NUREG-0713 V15: OCCUPATIONAL RADIATON EXPOSURE AT COM-POSITORIES AND RELATED ISSUES.

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

p NUREG/CP-0140 V01: PROCEEDINGS OF THE TWENTY-SECOND infiltration Analysis WATER REACTOR SAFETY INFORMATION MEETING Plenary Ses-NUREG/CR4333. BREATH VERSION 1.1 COUPLED FLOW AND sion, Advanced Instrumentation & Control Hardware & Software, ENERGY TRANSPORT IN POROUS MEDIA. Simulator Desenption And Human Factors Research. IPE & PRA.

User Guide.

NUREG/CP-0144 V01: A WORKSHOP ON DEVELOPING RISK ASSESS-MENT METHODS FOR MED6 CAL USE OF RADIOACTIVE Information D6 gest MATERIAL Summary NUREG-1350 V07: NUCLEAR REGULATORY COMMISSION INFORMA.

NUREG/CP 0144 V02: A WORKSHOP ON DEVELOPING RISK ASSESS-TON DIGEST.1995 Edition.

MENT METHODS FOR MEDICA L USE OF RADIOACTIVE MATERIAL Supporting Documents.

Inservice inspection NUREG/CR4125 V01: HUMAN FACTORS EVALUATON OF REMOTE NUREG/CR-6313 V01: ROBUST, ACCURATE, AND NON-CONTACTING AFTERLOADING BRACHYlHERAPY. Human Error And Cntical Tasks VIBRATION MEASUREMENT SYSTEM Summary of Companson Meas-In Remote Afterloading Brachytherapy And Approaches For improved urements Of The Robust Laser Interferometer And Typical Accelerome-System Performance.

ter Systems NUREGiCR-6125 V02: HUMAN FACTORS EVALUATION OF REMOTE NUREG/CR-6' 13 V02: ROBUST, ACCURATE, AND NON-CONTACTlWG I

3 VfBRATION MEASUREMENT SYSTEMS Supplemental Appendices NUR US I O SA N TO R D CT O ANCE esentng Canpanson Measurements O h hst Laser Inh IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of Validity i

uneter And Mcal Acceumneter @ ems.

And Feasibility NUREG/CR4265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN inservice Testing RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF NUREG-1482: GUIDELINES FOR INSERVICE TESTING AT NUCLEAR COMMISSION AND DEPENDENCIES.

NUREG/CR 6398: EVALUATION OF THE COMPUTERtZED PROCE-POWER PLANTS.

l DURES MANUAL 11 (COPMA-II) instructor Station Human Factora Evaluation NUREG 1527: NRC'S OBJECT-ORIENTED SIMULATOR INSTRUCTOR NUREG/CR-6125 V03: HUMAN FACTORS EVALUATION OF REMOTE STATION.

AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-Instrumentation System lnterfaces, Procedures And Practices,Trairung And Organizaton.

NUREG-1505 DRFT FC: A NONPARAMETRIC STATISTICAL METHOD-Nt 67 0 HU AN FACTORS EVALUATON OF TELE.

OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-oaches~

COMMISSONING SURVEYS Draft Report For Comment.

THERAPY. Identification Of Problems And Alternative AbATION NUREG/CR-6277 V02: HUMAN FACTORS EVA OF NUREG 1506 DRFT FC: MEASUREMENT METHODS FOR RADIOLOGi-TELETHERAPY. Function And Task Analysis.

CAL SURVEYS IN SUPPORT OF NEW DECOMMISSIONING NUREG/CR 4277 V03.

HUMAN FACTORS EVALUATION OF CRITERIA Draft Report For Comment.

TELETHERAPY. Human-System lnterfaces And Procedures.

NUREG 1507 DRFT FC: MINIMUM DETECTABLE CONCENTRATIONS NUREG/CR4277 V04: HUMAN FACTORS EVALUATION OF WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS TELETHERAPY. Training And Organizational Analysis.

CONTAMINANTS AND FIELD CONDITIONS. Draft Report For Com-NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF ment.

TELETHERAPY. Literature Review.

Interface Theory Human RecabilitY NUREG/CR-6150 V01:

SCDAP/RELAP5/ MOD 31 CCOE NUREG/CR-6265: MULTIDISC)PLINARY FRAMEWORK FOR HUMAN MANUALinterface Theory.

RELIABluTY ANALYSIS WITH AN APPUCATION TO ERRORS OF COMMISSION AND DEPENDENCIES Internal Event NUREG/CR-6355: A LIMITED ASSESSMENT OF THE ASEP HUMAN NUREG/CR-6143 V01: EVALUATION OF POTENTIAL SEVERE ACCI-RELIABILITY ANALYSIS PROCEDURE USING SIMULATOR EXAMI-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NATION RESULTS.

GRAND GULF. UNIT 1. Summary Of Results.

NUREG/CR4144 VO1: EVALUATION OF POTENTIAL SEVERE ACCl-ER D SHMM TERAMS AT S}mmar EG 700 A01 DFC: HUMAN-SYSTEM INTERFACE DESIGN RY,0N T Re REVIEW GUIDELINE. Draft Report For Comment.

Irradiated Reactor Fuel Human-System Interf ace NUREG-0725 R10: PUBLIC INFORMATION CIRCULAR FOR SHIP-NUREG-0700 R01 DFC: HUMAN-SYSTEM INTERFACE DESIGN MENTS OF IRRADIATED REACTOR FUEL REVIEW GUIDELINE. Draft Report For Comment.

Hydraulic Characterization J-integral NUREG/CR4356: HYDRAUUC CHARACTERIZATON OF HYDROTH.

NUREG/CR-6191: SIZE AND DEFORMATION LIMITS TO MAINTAIN ERMALLY ALTERED NOPAL TUFF.

CONSTRAINT IN K(IC) AND J(C) TESTING OF BEND SPECIMENS.

Hydrology J-R Curve NUREG/CR-4918 V08. CONTROL OF WATER INFILTRATION INTO NUREG/CR-4599 V04 Nt: SHORT CRACKS IN PIPING AND PIPING NEAR SURFACE LLW DISPOSAL UNITS Progress Report Of Field Ex.

WELDS Seventh Program Report March 1993 - December 1994.

NUREG/CR-6264 VOI: VALIDITY LIMITS IN J-RESISTANCE CURVE penments At A Humid Region Site Beltsville, Maryland.

DETERMINATION An Assessment Of The J(M) Parameter.

IRRAS NUREG/CR4264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE NUREG/CR-6116 V06. SYSTEMS ANALYSIS PROGRAMS FOR DETERMINATION A Computational Appioach To Ductile Crack Growth HANDS-ON INTEGRATED REUABILITY EVALUATIONS (SAPHIRE)

Under Large-Scale Yielding Conditions.

VERSON 5.0. Graphical Evaluation Module (GEM) Reference Manual.

NUREG/CR4297: FRACTURE EVALUATONS OF FUSION LINE NUREG/CR4116 V09-SYSTEMS ANALYSIS PROGRAMS FOR CRACKS IN NUCLEAR PIPE BIMETALUC WELC ]

HANDSON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR4300 REFINEMENT AND EVALUATON OF CRACK-OPEN-VERSON 5 0.Venfication And Vahdation (V&V) Manual ING-AREA ANALYSES FOR CIRCUMFERENTthi THPOUGH WALL NOREG/CR4116 Vio: SYSTEMS ANALYSIS PROGRAMS FOR CRACKS IN PIPES.

HANDSON INTEGRATED REUABluTY EVALUATIONS (SAPHtRE)

VERSION 5.0 Data Loading Manual.

J-Resistance NUREG/CR-6264 VOI: VALIDITY LIMITS IN J-RESISTANCE CURVE trnm6scitale \\ _

7; DETERMINATON An Assessment Of The J(M) Parameter.

r NUREG/CR4308; AN OVERVIEW OF INSTABIUTY AND FINGERING NUREG/CR4264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE DURING IMMISCIBLE FLUID FLOW IN POROUS AND FRACTURED DETERMINATON A Computational Approach To Ductde Crack Growth MEDIA.

Under Large-Scale Yielding Conditsons.

74 Subject index LWR Ucensee Event Report NUREG-1465: ACCIDENT SOURCE TERMS FOR LIGHT WATER NU-NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE CLEAR POWER PLANTS.

DAMAGE ACCIDENTS 1994 A STATUS REPORT. Main Report And Ap-NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN pendices A.H.

LIGHT WATER REACTORS. Semiannual Roport,0ctober 1993. March NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE 1994.

DAMAGE ACCIDENTS:1994 A STATUS REPORT. Appendix 1.

NUREG/CR-4667 V19: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual Report.Apnt-September 1994.

Licensing Evaluation NUREG/CR-0200 V1 R04: SCALE: A MODULAR CODE SYSTEM FOR i

Leak 4efore4reak PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-NUREG/CR4004: PROBABILISTIC PIPE FRACTURE EVALUATIONS CENSING EVALUATION. Control Modules.

FOR LEAK-RATE DETECTION APPLICATIONS-NUREG/CR-0200 V2PIR4: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4235: ASSESSMENT OF SHORT THROUGH-WALL CIR-PERFORMING STANDARDI2ED COMPUTER ANALYSES FOR Ll-CUMFERENTIAL CRACKS IN PIPES Expenments And Analysis, March CENSING EVALUATION Functional Modules Ft F8 1990. December 1994-NUREG/CR-0200 V2P2R4: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR-6251: STAINLESS STEEL SUBMERGED ARC WELD PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll.

FUSION LINE TOUGHNESS.

CENSING EVALUATION Functional Modules F9 F16.

NUREG/CR4299: EFFECTS OF TOUGHNESS ANISOTROPY AND NUREG/CR 0200 V3 R04: SCALE: A MODULAR CODE SYSTEM FOR COMBINED TENSION. TORSION, AND BENDING LOADS ON FRAC.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

CENSING EVALUATION. Miscellaneous ~

4 1

NUREG/CH-6300: REFINEMENT AND EVALUATION OF CRACK.OPEN-ING AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL L.lght Water Reactor CRACKS IN PIPES-NUREG-1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER NU-CLEAR POWER PLANTS.

NUREG/CR-4667 V18 ENVIRONMENTAL LY ASSISTED CRACKING IN UAE 149 ERFORMANCE BASED CONTAINMENT LE 4K TEST

. emiannual Report, October 1993 March PROGRAM Final Report.

F N

B^

N G/CR-4667 V19: ENVIRONMENTALLY ASSISTED CRACKING IN TES R RAM Dra t R9 mme '

LIGHT WATER REACTORS. Semiannual Report,Apni-September 1994.

Legalleeuances U"

NUREG 0750 V40:

NUCLEAR REGULATORY COMMISSION NUREG/CR-6256 V02: FIELD LYSIMETER INVESTIGATIONS - TEST ISSUANCES Opinions And Decisions Of The Nuclear Regulatory Com-RESULTS Low-Level Waste Data Base Development Program; Test mission With Selected Orders. July December 1994.

Results For Fiscal Years 1990.1991,1992, And 1993, NUREG-0750 V40102: INDEXES TO NUCLEAR REGULATORY COM.

MISSION ISSUANCES. July-December 1994.

NUREG-0750 V40 N05: NUCLEAR REGULATORY COMMISSION IS.

Uner Tearing NUREG/CR4184: SEPARATE EFFECTS TESTING AND ANALYSES TO SUANCES FOR NOVEMBER 1994. Pages 169-318.

fiUREG 0750 V40 N06: NUCLEAR REGULATORY COMMISSION IS.

INVESTIGATE LINER TEARING OF THE 1:16-SCALE REINFORCED SUANCES FOR DECEMBER 1994. Pages 319-387.

CONCRETE CONTAINMENT BUILDING.

NUREG 0750 V41 101: INDEXES TO NUCLEAR REGULATORY COM-MISSION LSSUANCES. January-March 1995.

Liquid Effluent NUREG-0750 V41 102. INDEXES TO NUCLEAR REGULATORY COM.

NUREG/CR 2907 V13: RADIOACTIVE MATERIALS RELEASED FROM MISSION ISSUANCES. January June 1995.

NUCLEAR POWER PLANTS. Annual Report 1992.

NUREG-0750 V41 N01: NUCLEAR REGULATORY COMMISSION IS.

NUREG/CR-2907 V14: RADIOACTIVE MATERIALS RELEASED FROM SUANCES FOR JANUARY 1995. Pages 1-69 NUCLEAR POWER PLANTS.

NUREG-0750 V41 NO2. NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR FEBRUARY 1995. Pages71-178.

Low Enriched Uranium NUREG 0750 V41 NO3. NUCLEAR REGULATORY COMMISSION IS-NUREG 1065 R02: ACCEPTABLE STANDARD FORMAT AND CON-SUANCES FOR MARCH 1995.Pages 179-243.

TENT FOR THE FUNDAMENTAL NUCLEAR MATERIAL CONTROL NUREG-0750 V41 N04. NUCLEAR REGULATORY COMMISSION IS-(FNMC) PLAN REQUIRED FOR LOW-ENRICHED URANIUM FACILI-SUANCES FOR APRIL 1995 Pages 245-319-TIES.

NUREG 0750 V41 N05: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR MAY 1995 Pages 321380.

Low Power NUREG 0750 V41 N06: NUCLEAR REGULATORY COMMISSION IS-NUREG/CR-6143 VO1: EVALUATION OF POTENTIAL SEVERE ACCI.

SUANCES FOR JUNE 1995 Pages 381496.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG-0750 V42101: INDEXES TO NUCLEAR REGULATORY COM-GAAND GULF. UNIT 1. Summary Of Results.

MISSION ISSUANCES July-September 1995.

NU9EG/CR-6143 V06 PI: EVALUATION OF POTENTIAL SEVERE AC.

NUREG-0750 V42 N01: NUCLEAR REGULATORY COMMISS,0N IS-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT R

aluation G Sevee Ament Risks N Mant NURE 2

02. N L EGULATORY COMMISSION IS-a al ate 5 Dunng A Refuehng Outage Main NW And Ap-SUANCES FOR AUGUST 1995. Pages 47-97.

NUREG 0750 V42 NO3 NUCLEAR REGULATORY COMMISSION IS-NU

/C 6143 V06 P2. EVALUATION OF POTENTIAL SEVERE AC-RETUL Y COMMISSION IS-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG 50 2 0 N LE GRAND GULF, UNIT 1 Evaluat:on Of Severe Accident Risks For Plant SUANCES FOR OCTOBER 1995. Pages 111 180.

Operational State 5 Dunng A Refueting Outage Supporting MELCOR Lessons Learned Calculations.

NUREG-1526. LESSONS LEARNED FROM EARLY IMPLEMENTATION NUREG/CR.6144 V01: EVALUATION OF POTENTIAL SEVERE ACCl-OF THE MAINTENANCE RULE AT NINE NUCLEAR POWER PLANTS DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1. Summary Of Results.

Ucense Application NUREG/CR-6144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-NUREG 1323 ROI: LICENSE APPLICATION REVIEW PLAN FOR A CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-SURRY, UNIT tEvaluation Of Severe Accident Risk Dunng Mid-Loop LEVEL RADIOACTIVE WASTE. Draft Review Plan.

Operations Main Report.

NUREG/CR-6144 V06 P2: EVALUATION OF POTENTIAL SEVERE AC.

Ucensed Fuel Facility Status Report CfDENTS DURING LOW POWER AND CHUTOOWN OPERATIONS AT NUREG.0430 V14:

LICENSED FUEL FACILITY STATUS SURRY, UNIT 1 Evaluation Of Severe Accident Risk Dunng Mid-Loop REPORT. inventory Difference Data. July 1,1993 June 30,1994 (Gray Operations Appendices.

Book II)

Low-Alloy Steel Licensed Operating Reactors NUREG/CR-6335 FATIGUE STRAIN LIFE BEHAVIOR OF CARBON NUREG 0020 V19 LICENSED OPERATING REACTORS STATUS SUM-AND LOW ALLOY STEELS, AUSTENITIC STAINLESS STEELS. AND MARY REPORT. Data As Of December 31,1994 (Gray Book f)

ALLOY 600 IN LWR ENVIRONMENTS

i I

Subject Index 75 Low-Level Nuclear Waste TECHNOLOGY, CAMBRIDGE, MASSACHUSETTS,lDENTIFIED ON NUREG/CR4390: RADIOLOGICAL CHARACTERl2ATION OF SPENT AUGUST 19,1995.

CONTROL ROD ASSEMBLIES.

NUREG/CR 4323: RELATIVE RISK ANALYSIS IN REGULATING THE l

Low-Level Radtoactive Weste Ica I

NUREG/CR4188 V02: MICROBIAL DEGRADATION OF LOW-LEVEL l

l RADIOACTIVE WASTE. Annual Report For FY 1994-Medical Misadministration NUREG/CR 6277 V01: HUMAN FACTORS EVALUATION OF TELE-l Low-Level Rad 6oactive Weste D6sposal NUREG/CR4114 V02: AUXILIARY ANALYSES IN SUPPORT OF PER-THERAPY. Identificaton Of Problems And Alternative Approaches.

NUREG/CR-6277 V02: HUMAN FACTORS EVALUATION OF FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-LEVEL WASTE FACILITY.Two-Phase Flow And Contaminant Transport In Un' TELETHERAPY.Functon And Task Analysis.

NUREG/CR 6277 V03: HUMAN FACTORS EVALUATION OF saturated Soils With Application To Low-Level Radoactive Waste Dis-TELETHERAPY. Human-System interfaces And Procedures.

posal.

NUREG/CR-6277 V04: HUMAN c: ACTORS EVALUATION OF Low-Level Waste Data Base TELETHERAPY. Training And Organizational Analysis.

NUREG/CR-5229 V07: FIELD LYSIMETER INVESTIGATIONS. LOW.

NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF LEVEL WASTE DATA BASE OFVELOPMENT PROGRAM FOR TELETHERAPY.Uterature Review.

F1 SCAL YEAR 1994. Annual Report.

j NUREG/CR-6256 VOI: FIELD LYSIMETER INVESTIGATIONS - TEST Medical Quality Management RESULTS Low-Level Waste Data Base Program. Test Results For NUREG/CR-6277 VOI: HUMAN FACTORS EVALUATION OF TELE-i l

Fiscal Years 1986,1987,1988, And 1989.

THERAPY. Identificalion Of Problems And Altemative Approaches.

l NUREG/CR-6256 V02; FIELD LYSIMETER INVESTIGATIONS TEST NUREG/CR-6277 V02: HUMAN FACTORS EVALUATION OF l

RESULTS. Low-Level Waste Data Base Development Program: Test TELETHERAPY. Function And Task Analysis.

Results For Fiscal Years 1990,1991,1992, And 1993.

NUREG/CR-6277 V03: HUMAN FACTORS EVALUATION OF TELETHERAPY. Human-System Interfaces And Procedures.

Low-Level Waste D6eposal NUREG/CR-6277 V04: HUMAN FACTORS EVALUATION OF NUREG/CR 4918 V08: CONTROL OF WATER INFILTRATION INTO TELETHERAPY. Training And Organizational Analysis.

NEAR SURFACE LLW DISPOSAL UNITS Progress Report Of Field Ex-NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF penments Al A Hum'd Regon Site.Bettsville, Maryland TELETHERAPY. Literature Review.

NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE ASSESS-MENT METHODOLOGY FOR LOW-LEVEL RADIOACTIVE WASTE Medical Use l

DISPOSAL FACILITIES Validaten Needs.

NUREG/CP 0144 V01: A WORKSHOP ON DEVELOPING RISK ASSESS-NUREG/CR-6305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EQUI' MENT METHODS FOR MEDICAL USE OF RADIOACTIVE LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS-MATERIAL. Summary.

ING THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL WASTE NUREG/CP-0144 V02: A WORKSHOP ON DEVELOPING RISK ASSESS-DISPOSAL UNITS. Background. Theory, And Model Desenption.

MENT METHODS FOR MEDICAL USE OF RADIOACTIVE Low-Level Waste Facil6ty MATERIALSupporting Documents.

NUREG/CR 6284: CRITICALITY SAFETY CRITERIA FOR LICENSE Microbial Degradation REVIEW OF LOW-LEVEL WASTE FACILITIES NUREG/CR-6188 V02: MICROBIAL DEGRADATION OF LOW-LEVEL MACCS RADIOACTIVE WASTE. Annual Report For FY 1994.

NUREG/CR-6134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-Mid-Loop Operat6on CIDENT CONSEOUENCE MODEL NUREG/CR-6144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-SURRY, UNIT 1 Evaluat on Of Severe Accedont Risk Dunng Mid-Loop DENT CONSEQUENCE MODEL Operatons Main Report.

NUREG/CR-6136. UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR-6144 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-SURRY. UNIT 1.Evaluaton Of Severe Accident Risk During Mid-Loop P*'#*""^PP'" ""~

MELCOR Computer Code NUREG/CR-6119 V01: MELCOR COMPUTER CODE MANUALS Pnmer Mine Seismicity NUREG/CR-6283. FIELD SITE INVESTIGATION: EFFECT OF MINE NUR /

61 0

L COMPUTER CODE SEISMICITY ON GROUNDWATER HYDROLOGY.

MANUALS. Reference Manuals.Verson 1.8.3 September 1994.

Moisture Redistribution Maintenance NUREG/CR-6348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION NUREG/CR-6002:

RISK-BASED MAINTENANCE IN PARTIALLY SATURATED POROUS MEDIA.

MODELING.Prontization Of Maintenance importances And Quantifica.

tion Of Maintenance Effectiveness.

Multiplant Action issues Maintenance Rule NUREG 1435 SO4 STATUS OF SAFETY ISSUES AT LICENSED NUREG 1526: LESSONS LEARNED FROM EARLY IMPLEMENTATION POWER PLANTS TMI Action Plan Requirements, Unresolved Safety OF THE MAINTENANCE RULE AT NINE NUCLEAR POWER PLANTS.

Issues,Genenc Safety issues.Other Multiplant Acton issues Management Plan NPRDS NUREG-1444 S01: SITE DECOMMISSION lNG MANAGEMENT PLAN NUREG/CR-5944 V02-A CHARACTERIZATION OF CHECK VALVE DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLEAR NU E 1 RE ONSTITUTION OF THE MANUAL CHAPTER 2512 CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT 1.

NURETH-7 NUREG/CP-0142 V01: PROCEEDINGS OF THE 7TH INTERNATIONAL Material Control MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS NUREG-1065 R02: ACCEPTABLE STANDARD FORMAT AND CON-(NURETH-7) Sessions 15-TENT FOR THE FUNDAMENTAL NUCLEAR MATERIAL CONTROL NUREG/CP 0142 V02. PROCEEDINGS OF THE 7TH INTERNATIONAL (FNMC) PLAN REQUIRED FOR LOW ENRICHED URANIUM FACILi-MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS TIES.

NUREG-1280 ROI: STANDARD FORMAT AND CONTENT ACCEPT.

(NURETH 7).Sessons 6-11.

NUREG/CP-0142 V03 PROCEEDINGS OF THE 7TH INTERNATIONAL ANCE CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNT.

MEETING ON NUCLEAR HEACTOR THERMAL HYDRAULICS ING (MC&A) REFORM AMENDMENT.

(NURETH-7) Sesssons 12-16.

1 Medical Device NUREG/CP-0142 V04 PROCEEDINGS OF THE 7TH INTERNATIONAL NUREG-1535: INGESTION OF PHOSPHORUS-32 AT MASSACHU-MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS I

SETTS INSTITUTE OF (NURETH-7) Sessions 17 24 i

i

76 Subject Index l

l Natural Analog NUREG-0980 V02 NO3: NUCLEAR REGULATORY LEGISLATION.103d NUREG/CP 0147: PROCEEDINGS OF THE WORKSHOP ON THE ROLE Congress.

i OF NATURAL ANALOGS IN GEOLOGIC DISPOSAL OF HIGH-LEVEL NUCLEAR WASTE. Held in San Antonio, Texas, July 22 25,1991.

Nuclear Regulatory Research NUREG 1266 V09: NRC SAFETY RESEARCH IN SUPPORT OF REGU-Natural Circulation LATION - FY 1994.

NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION STUDIES AT THC INEL Nuclear Safety NUREG-0700 RO1 DFC: HUMAN-SYSTEM INTERFACE DESIGN New Sensor REVIEW GUIDELINE. Draft Report For Comment.

NUREG/CR4334: NEW SENSOR FOR MEASUREMENT OF LOW AIR FLOW VELOCITY. Phase i Final Report.

Nuclear Safety Research NUREG/CP-0148: TRANSACTIONS OF THE TWENTY THIRD WATER NU EG/CR-6260 APPLICATION OF NUREG/CR-5999 INTERIM FA-TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-Occupational Exposure NENTS.

NUREG/CP4143: PROCEEDINGS OF THE THIRD INTERNArlONAL WORKSHOP ON THE IMPLEMENTAllON OF ALARA AT NUCLEAR "d

e o Isla ew R /C 310: AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO-NUREG/ R 9 VO PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-CLEAR ACCIDENT.

CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-ED READINGS IN RADIATION PROTECTION AND ALARA.

Nuclear Air Cleaning NUREG/CR-6112: IMPACT OF REDUCED DOSE LIMITS ON NRC Ll-NUREG/CP 0141: PROCEEDINGS OF THE 23RD DOE /NRC NUCLEAR CENSED ACTIVITIES. Maior issues in The implementaton Of ICRP/

AIR CLEANING CONFERENCE. Held in Buffalo,New York. July 25-NCRP Dose Umst Recommendations Final Report.

28J 9%

Occupational Radiation Exposure Nuclear Fuel Facility NUREG/CR 5884 VO1: REVISED ANALYSIS OF DECOMMISSIONING NUREG/CR-6287. MANAGEMENT CONCEPTS AND SAFETY APPLICA.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER TIONS FOR NUCLEAR FUEL FACILITIES.

STATION Effects Of Current Regulatory And Other Considerations On The Financial Insurance Requirements Of The Decommissioning Rule Nuclear Materials Licensees And....

NUREG/CR4330: RESULTS OF REGULATORY IMPACT SURVEY OF NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

STATION. Effects Of Current Regulatory And Other Considerations On Nuclear Plant Analyzer NUREG/CR4291 VOI: NUCLEAR PLANT ANALYZER.inetallation Manual.

Office Of The inspector General MUREG/CR 6291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer.

NUREG-1415 V07 NO2: OFFICE OF THE INSPECTOR NL EG R-6 91 V03: NUCLEAR PLANT ANALYZER. Computer Visual 995 NU C 21 04 N LEAR PLANT ANALYZER Programmer's gg-GENERAL. Semiannual Report To Congress,Apnl 1,1995 September 30,1995.

Nuclear Plant Safety System NUREG/CR-6263 VOI: HIGH INTEGRITY SOFTWARE FOR NUCLEAR On-Line Monitoring POWER PLANTS. Candidate Guidelines, Technical Basis And Research NUREG/CR4343: ON-LINE TESTING OF CAllBRATION OF PROCESS Needs Executwe Summary INSTRUMENTATION CHANNELS IN NUCLEAR POWER NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR PLANTS. Phase 11 Final Report POWER PLANTS Candidate Guidelines, Technical Basis And Research NeedsMasn Report.

Operating Experience i

NUREG 1272 V08 NO2: OFFICE FOR ANALYSIS AND EVALUATION OF Nuclear Plant Structure OPERATIONAL DATA.1993 Annual Report Nucioar Matenals.

NUREG 1522: ASSESSMENT OF INSERVICE CONDITIONS OF NUREG/CR-6089: DETECTION OF PUMP DEGRADATION.

SAFETY RELATED NUCLEAR PLANT STRUCTURES' NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-LOADED PRESSURE RELIEF VALVES USED IN SAFETY-RELATED SYSTEMS Nuclear Power Plant AT NUCLEAR POWER PLANTS.

RUREG/CR-5975 RO1: INCENTIVE REGULATION OF INVESTOR-l OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULA-Operating Experience Feedback Report l

TORS.

NUREG 1275 VII: OPERATING EXPERIENCE FEEDBACK REPORT -

l NUREG/CR.6159 USING MICRO SAINT TO PREDICT PERFORMANCE TURBINE GENERATOR OVERSPEED PROTECTION l

IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of Vahdity SYSTEM Commercial Power Reactors.

l And Feasibility.

NUREG/CR-6172: REVIEWING PSA-BASED ANALYSES TO MODIFY Operational Event TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1994 A STATUS REPORT. Main Report And Ap-Nuclear Reactor pendices A-H.

NUREG/CP 0142 V01: PROCEEDINGS OF THE TTH INTERNATIONAL NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS DAMAGE ACCIDENTS:1994 A STATUS REPORT.Appenda 1.

(NURETH-7).Sessons 1-5.

NUREG/CP-0142 V02: PROCEEDINGS OF THE 7TH INTERNATIONAL Operator Licens6ng MEETING ON NUCLEAR REACTOR THERMAL-HYDRAUllCS NUREG-1478 RO1: NON-POWER REACTOR OPERATOR LICENSING (NURETH-7). Sessions 6-11.

EXAM!NER STANDARDS.

NUREG/CP-0142 V03. PROCEEDINGS OF THE 7TH INTERNATIONAL MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS Organization Chart (NURETH-7). Sessions 1218 NUREG-0325 R18. U.S. NUCLEAR REGULATORY COMMISSION OR.

NUREG/CP 0142 V04 PROCEEDINGS OF THE 7TH INTERNATIONAL GANIZATION CHARTS AND FUNCTIONAL STATEMENTS. July 23 j

MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS 1995.

(NURETH-7). Sessions 17-24 Ove 3 peed Protection System Nuclear Regulatory Leg 6slatlon NUREG-1275 VII: OPERATING EXPERIENCE FEEDBACK REPORT -

WUREG-0980 V01 NO3. NUCLEAR REGULATORY LEGISLATION.1030 TURBINE. GENERATOR OVERSPEED PROTECTION Congress SYSTEM. Commercial Power Reactors.

1 l

l l

Subject index 77 PRA Petitions For Rulemaking NUREG/CR4116 V06. SYSTEMS ANALYSIS PROGRAMS FOR NUREG-0936 V13 NO3: NRC REGULATORY AGENDA. Semiannual HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

Report. July-December 1994 VERSION 5.0. Graphical Evaluation Module (GEM) Reference Manual.

NUREG-0936 V14 N01: NRC REGULATORY AGENDA. Semiannual Report. January June 1995.

NUREG/CR4116 VIO: SYSTEMS ANALYSIS PROGRAMS FOR Phosphorue-32 MASSACHU-HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG-1535: INGESTION OF PHOSPHORUS-32 AT VERSION 5.0 Data Loading Manual.

SETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE, MASSACHUSETTS.lDENTIFIED ON PWR AUGUST 19,1995.

NUREG 1122 R01: KNOWLEDGE AND ABILITIES CATALOG FOR NU-CLEAR POWER PLANT OPERATOR $ PRESSURIZED WATER REAC.

Pipe TORS.

NUREG/CR-4593 V04 N1: SHORT CRACKS IN PIPING AND PIPING NUREG/CR 5884 V01: REVISED ANALYSIS OF DECOMMISSIONING WELDS. Seventh Program Report March 1993. December 1994.

I FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER NUREG/CR4004: PROBABILISTIC PIPE FRACTURE EVALUATIONS l

STATICN Effects Of Current Regulatory And Other Considerations On FOR LEAK RATE DETECTION APPLICATIONS.

NUREG/CR4235: ASSESSMENT OF SHORT THROUGH WALL CIR-The Financial Insurance Requirements Of The Decommissioning Rule And....

CUMFERENTIAL CRACKS IN P: PES Expenmenu. And Analysis, March NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING 1990. December 1994.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER NUREG/CR 6251: STAINLESS STEEL SUBMERGED ARC WELD STATION Effects Of Current Regulatory And Other Considerations On NU R 629 :

A E EVALUATIONS OF FUSION LINE The Financial Insurance Requirements Of The Decommissioning Rule CRACKS IN NUCLEAR PIPE BlMETALLtc WELDS.

NUREG/CR4298: rRACTURE CEHAVIOR OF SHORT CIRCUMFEREN-NUREd/GR-5054-EFFECT ON AGING ON PWR CHEMICAL AND TIALLY SURFACE-CRACKED PIPE.

NUREG/CR-6299: EFFECTS OF TOUGHNESS ANISOTROPY AND NURE /CR 60 E Tl A PRESSURIZED WATER REACTOR DE-COMBINED TENSION, TORSION, AND BENDING LOADS ON FRAC-COMMISSIONil4G COSTS. A User's Manual For The PWR Cost Esti-TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

EGMN MNT AND NWG & NGM4 NUR G C 6 T PROBAB LIT NTAINMENT FAILURE BY DIRECT CONT AINMENT HEATING IN SURRY.

CRACKS IN PIPES.

NUREG/CR4266: ANALYSIS OF BORON DILUTION IN A FOUR-LOOP PWR.

Piping NUREG/CR-3243: COMPARISON OF ASME CODE FATIGUE EVALUA.

Packaging TION METHODS FOR NUCLEAR CLASS 1 PIPING WITH CLASS 2 NUREG-0383 V01 RI8: DIRECTORY OF CERTIFICATES OF COMPLl-ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC NtREG/

54 V02: EXPERIMENTAL RESULTS FROM CONTAIN-MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT NU 83 1'8: DIRECTORY OF CERTIFICATES OF COMPLl-NEIS-ANCE FOR RADIOACTIVE MATERIALS PACKAGES Certificates Of NU EC/ 62 AP L AI B

N S T

'ance.

MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF NURE -0383 V03 R15: DIRECTORY OF CERTIFICATES OF COMPLl-ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC STRONG MOTION EARTHOUAKES.

Approved Ouality Assurance Programs for Radioactive Matenals Pack-ages.

Piping System NUREG/CR4239 V01: SURVEY OF STRONG MOTION EARTHQUAKE R 0014: BALDCYPRESS TREE RING ELEMENTAL CONCEN-P AS ON PIP NG YST S Ma n e TRATIONS AT REELFOOT LAKE. TENNESSEE.FROM AD 1795 TO NUREG/CR-6239 V02: SURVEY OF STRONG MOTION EARTHOUAKE AD 1820.

EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-PHASIS ON PIPING SYSTEMS. Appendices.

g l

NUREG/CR4074 V04: SEALED SOURCE AND DEVICE DESIGN Porous Media l

SAFETY TESTING. Technical Report On The Findings Of Task 4.Inves' NUREG/CR-6333: BREATH VERSION 1.1 - COUPLED FLOW AND l

tigation Of Sealed Source for Paper Mill Digester.

ENERGY TRANSPORT IN POROUS MEDIA. Simulator Desenption And User Guide.

Performance NUREG/CR4348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION NUREG/CR 6046: ALERTNESS, PERFORMANCE. AND OFF-DUTY IN PARTIALLY SATURATED POROUS MEDIA.

SLEEP ON 8-HOUR AND 12 HOUR NIGHT SHIFTS IN A SIMULATED CONTINUOUS OPERATIONS CONTAOL ROOM SETTING.

Potassium lod 6de NUREG/CR4159: USING MICRO SAINT TO PREDICT PERFORMANCE NUREG/CR 6310: AN ANALYSIS OF POTASSIUM IODIDE (KI) PRO l

IN A NUCLEAR POWER PLANT CONTROL ROOM A Test Of ValdtY PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-And Feasitnkty.

CLEAR ACCIDENT.

Performance Aseeeement Pressure Locking

(

NUREG-1464: NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE NUREG/CP 0146: PROCEEDINGS OF THE WORKSHOP ON GATE

2. Development Of Capabihties For Roview Of A Performance Assess-VALVE PRESSURE LOCKING AND THERMAL BINDING i

ment For A High Level Waste Repository.

NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE ASSESS-Pressure Sensor i

MENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE WASTE NUREG/CR 6312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN-i DISPOSAL FACILITIES.Vahdation Needs.

SORS.

NUREG/CR-635h REVIEW OF SCENARIO SELECTION APPROACHES l

FOR PERFORMANCE ASSESSMENT OF HIGH-LEVEL WASTE RE.

Pressure Veseel POSITORIES AND REuTED ISSUES.

NUREG/CR-5591 V05 N2: HEAVY SECTION STEEL IRRADIATION 3

PROGRAM. Progress Report For April 1994 Through September 1994.

Performance incentive 4

NUREG/CR-5975 RO1: INCENTIVE REGULATION OF INVESTOR-Pressurized Water Reactor OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGUu-NUREG-1122 R01: KNOWLEDGE AND ABluTIES CATALOG FOR NU-CLEAR POWER PLANT OPERATORS PRESSURIZED WATER REAC.

TORS.

TORS.

Perfonnance Testing NUREG/CR-5884 V01: REVISED ANALYSIS OF DECOMMISSIONING j

NUREG/CR-6354 DRF FC: PERFORMANCE TESTING OF ELECTRON-FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER IC PERSONAL DOSIMETERS. Draft Report For Comment STATION Effects Of Current Regulatory And Other Consederations On i

a a

~ _ _ _ _ _ _ _.. _ _ _ _. _ _ _

78 Subject Index The Financial Insurance Requirements Of The Decommissiorung Rule Publ6c Utility Comm6ee60n And...

NUREG/CR-5975 ROI: INCENTIVE REGULATION OF INVESTOR.

NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULA-FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER TORS.

STATION Effects Of Current Regulatory And Other Considerations On The Financial Insurance Requirements Of The Decommissoning Rule Pump 482 6 M MNE TESM M EMM NUR /CR-5954: EFFECT ON AGING ON PWR CHEMICAL AND p

NU E /C E TIMA I PRESSURIZED WATER REACTOR DE.

NUREG/CR-5857: AGING OF TURBINE DRIVES FOR SAFET'6HELAT-COMMISSIONING COSTS. A User's Manual For The PWR Cost Est,_

ED PUMPS IN NUCLEAR POWER PLANTS mating Computer Program (CECP) Software NUREG/CR-6089 DETECTION OF PUMP DEGRADATION.

NUREG/CR-6109. THE PROBASILITY OF CONTAINMENT FAILURE BY DIRECT CONTAINMENT HEATING IN SURRY.

Quality Assurance NUREG/CR-6266: ANALYSIS OF BORON DILUTION IN A FOUR-LOOP NUREG/CR-6316 V0t GUIDELINES FOR THE VERIFICATION AND PWR.

VALIDATION OF EXPERT UYSTEM SOFTWARE AND CONNENTION-AL SCETWARE-Probab416st6c Accident Consequence NUREG/CR-6316 V02: GUIDELINES FOR THE VERIFICATION AND NUREG/CR 6244 V01: PROBABILISTIC ACCICENT CONSEQUENCE VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVE NTION-UNCERTAINTY ANALYSIS Disperson And Deposition Unccriainty AL SOFTWARE. Survey And Assessmern Of Conventonal hoftware NU G/YR 62 v

Va a V

ROBABILISTIC ACCIDENT CONSEQUENCE NUR UI ES FOR THE VERIHOAilON AND n And Deposden UncMag sse s e pe ce And B VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR-6244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE AL SOFTWARE Survey And Documentaton Of Expert System Venfica-UNCERTAINTY ANALYSIS. Disperson and Deposition Uncertainty ton And Vahdaten Methodologies.

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

NUREG/CR-6316 V04. GUIDELINES FOR THE VERIFICATION AND VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL Probabilist6c Riek Assessment SOFTWARE Evaluat.on Of Knowledge Base Certification Methods.

NUREG-1464. NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE NUREG/CR-6316 V05-GUIDELINES FOR THE VERIFICATION AND 1

2 Development Of Capatulities For Review Of A Performance Assess-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-I l

ment For A High-Level Wasle Repository.

AL SOFTWARE. Rationale And Descnption Of V&V Guidehne Packages NUREG/CR-6116 V09. SYSTEMS ANALYSIS PROGRAMS FOR And Procedures.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR 6316 V06: GUIDELINES FOR THE VERIFICATION AND l

]

NUI C 11 01 ELC 1 U ER hE ANUALS Pnmer A

AR V lida on e os OMPUTER CODE NUREG/CR-6316 V07: GUIDELINES FOR THE VERIFICATION AND NUR / 6 0

E MANUALS Reference Manuals.Verson 1.8 3 September 1994.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-l NUREG/CR-6141: HANDBOOK OF METHODS FOR RISK-BASED AL SOFTWARE. User's Manual.

1 ANALYSES OF TECHNICAL SPECIFICATIONS.

NUREG/CR-6316 V08: GUIDELINES FOR THE VERIFICATION AND i

NUREG/CR-6143 V01: EVALUATION OF POTENTIAL SEVERE ACCl-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT AL SOFTWARE.Bibhography.

' GRAND GULF, UNIT I Summary Of Results.

NUREG/CR-6324. QUALITY ASSURAr4CE FOR GAMMA KNIVES.

NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT Quahty Management GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant NUPEG/CR 6276: A COMPILATION OF CURRENT REGULATIONS, Operahonal State 5 Dunng A Refueling Outage Usin Report And Ap-STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING BRA-NU

/ R-6143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT RELAPS Computer Code GRAND GULF UNIT 1 Evaluation Of Severe Accident Risks For Plant NUREG/CR 6150 V01:

SCDAP/RELAPS/ MOD 3.1 CODE Operational State 5 Dunng A Refuehng Outage Supportmg MELCOR MANUALinterface Theory.

Calculatons-NUREG/CR-6150 V02:

SCDAP/RELAP/ MOD 31 CODE Probabilistic Safety Assessment AQamage Wesson Wel Demy NUREG/CR-6172; REVIEW lNG PSA BASED ANALYSES TO MODIFY NUREG/CR-6150 V03:

SCDAP/RELAPS/ MOD 3.1 CODE TECHN' CAL SPECIFICATIONS AT NUCLEAR POWER PLANTS' NU EG/C 61 V04 SO ELAP5/ MOD 3.1 CODE Proceed 6ng MANUALMATPRO-A Library Of Matenals Properties For Light-Water-NUREG/CP-0140 V01: PROCEEDINGS OF THE TWENTY SECOND Reactor Accident Analysis.

WATER REACTOR SAFETY INFORMATION MEETING Plenary Ses.

NUREG/CR 6150 VOS:

SCDAP/RELAPS/ MOD 31 CODE son, Advanced Instrumentation & Control Hardware & Software, MANUAL. Developmental Assessment.

Human Factors Research, IPE & PRA.

NUREG/CR-6325 AN IMPLICIT STEADY STATE INITIALIZATION PACK.

NUREG/CP-0140 V02: PROCEEDINGS OF THE TWENTY SECOND AGE FOR THE RELAP5 COMPUTER CODE.

WATER REACTOR SAFETY INFORMATION MEETING Severe Acco dent Research, Thermal Hydrauhc Research For Advanced Passive RELAPS Interf ace LWRs, Hig4Burnup Fuel Behavior.

NUREG/CR-629 f V01: NUCLEAR PLANT ANALYZER. Installation NUREG/CP 0140 V03: PROCEEDINGS OF THE TWENTY-SECOND Manual.

WATER REACTOR SAFETY INFORMATION MEETING Pnmary Sys-NUREG/CR-6291 V02 NUCLEAR PLANT ANALYZER. Analyzer Refer-tems integnty, Structural And Seismic Engineenng, Ag ng Research, ence Manual.

Products And Applicatens.

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

REG /N6291 W: MEAR MT ANAMER Wanned NURE 6

LINE TESTING OF CALIBRATION OF PROCESS anual INSTRUMENTATION CHANNELS IN NUCLEAR POWER PLANTS. Phase ll Final Report RELAP5/ MOD 3 Computer Code Wogram Performance NUREG/CR-5535 V01. RELAPS/ MOD 3 CODE MANUALCode Structure, NUREG/CR 5758 V05. FITNESS FOR DUTY IN THE NUCLEAR POWER System Models, And Soluton Methods.

INDUSTRY Annual Summary Of Program Performance Reports CY NUREG/CR-5535 V02: RELAP5/ MOD 3 CODE JAWAL User's Guide 1994 And input Requirements.

NUREG/CR-5535 V04-RELAPS/ MOD 3 CODE MANUALModels And Publ6c information Circular Correlatons.

NUREG-0725 R10- PUBLIC INFORMATION CIRCULAR FOR SHIP-NUREG/CR-5535 VOS R1: RELAPS/ MOD 3 CODE MANUALUser's MENTS OF IRRADIATED REACTOR FUEL Guidehne.

Subject index 79 Rad 6ation Dose Rad 6olog6 cal Survey NUREG/CP 0143: PROCEEDINGS OF THE THIRD INTERNATIONAL NUREG 1506 DRFT FC: MEASUREMENT METHODS FOR RADIOLOGI-WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR CAL SURVEYS IN SUPPORT OF NEW DECOMMISSIONING POWER PLANTS. Held At Hauppauge. Lonqlsland New York.

CRITERIA. Draft Report For Comment.

NUREG/CR-3469 V08: OCCUPATIONAL UUSE REDUCTION AT NU-CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-Radionuclide Characterization ED READINGS IN RADIATION PROTECTION AND ALARA.

NUREG/CR-6390: RADIOLOGICAL CHARACTERIZATION OF SPENT NUREG/CR-6112: IMPACT OF REDUCED DOSE LIMITS ON NRC LI-CONTROL ROD ASSEMBLIES.

CENSED ACTIVITIES. Mapor issues in The implementation Of ICRP/

NCRP Dose Lunt Recomrnendations. Final Report.

Radlonuclide Migration NUREG/CR-6305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EOUl-LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS-R 7 V1 OCCUPATIONAL RADIATION EXPOSURE AT COM-ING THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL WASTE MERCIAL NUCLEAR POWER REACTORS AND OTHER DISPOSAL UNITS. Background Theory, And Model Desenption-FACILITIES,1993. Twenty-Sixth Annual Report.

NUREG 1535: INGESTION OF PHOSPHORUS-32 AT MASSACHU-Reactive Transpod -

SETTS INSTITUTE OF NUREG/CR-6347: MULTI-PHASE REACTIVE TRANSPORT THEORY.

TECHNOLOGY. CAMBRIDGE, MASSACHUSETTS,lDENTIFIED ON AUGUST 19.1995.

Reactor NUREG/CR 6134-UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG-1493: PERFORMANCE-BASED CONTAINMENT LEAK TEST CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC, CIDENT CONSEQUENCE MODEL-PROGRAM Final Report.

NUREG/CR 6' 35: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG 1493 DFC: PERFORMANCE BASED CONTAINMENT LEAK-EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCT TEST PROGRAM. Draft Report For Comment.

DENT CON 3EOVENCE MODEL.

NUREG/CR4136. UNCERTAINTY AND SENSITIVITY ANALYSIS OF Reactor Accident FOOD PAlHWAY RESULTS WITH THE MACCS REACTOR ACCI.

NUREG/CR-6134. UNCERTAINTY AND SENSITIVITY ANALYSIS OF DENT CO' ASEQUENCE MODEL CHRONfC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-CIDENT CONSEQUENCE MODEL Radiationr tection NUREG/CR-6135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREGre3469 V08: OCCUPATIONAL DOSE REDUCTION AT NU-EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-CLE AP POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT-DENT CONSEOUENCE MODEL.

ED READINGS IN RADIATION PROTECTION AND ALARA.

NUREGICH-6136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-Radiation Survey DENT CONSEQUENCE MODEL.

NUREG-1507 DRFT FC: MINIMUM DETECTABLE CONCENTRATIONS NUREG/CR-6265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF CONTAMINANTS AND FIELD CONDITIONS. Draft Report For Com-COMMISSION AND DEPENDENCIES.

ment NUREG/CR-6349: COST BENEFIT CONSIDERATIONS IN REGULA-TORY ANALYSIS.

RW W h w NUREG/CR 6323: RELATIVE RISK ANALYSIS IN REGULATING THE Reactor Component USE OF RADIATION-EMITTING MEDICAL DEVICES.A Preliminary Ap-NUREG/CR-6002:

RISK-BASED MAINTENANCE DEUWnonhzahon G Maintenance impodances M Quanh NU E i CR 6324:OUALITY ASSURANCE FOR GAMMA KNIVES.

tion Of Maintenance Effectiveness.

Radioactive Material Reactor instrumentation NUREG 0383 V01 R18. DIRECTORY OF CERTIFICATES OF COMPLI.

NUREG/CP-0140 V01: PROCEEDINGS OF THE TWENTY-SECOND ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses.

Aporoved Packages.

NUAEG 0383 V02 R18: DIRECTORY OF CERTIFICATES OF COMPLl-sion, Advanced instrumentation & Control Hardware & Software, ANCE FOR RADIOACTIVE MATERIALS PACKAGESCertificates Of Human Factors Research, lPE & PRA.

Compliance.

NUREG-0383 V03 R15: DIRECTORY OF CERTIFICATES OF COMPLl-Reactor Operation ANCE FOR RADIOACTIVE MATERIALS PACKAGES Report Of NRC NUREG/CR-6172: REVIEWING PSA-BASED ANALYSES TO MODIFY TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

Approved Quahty Assurance Programs for P 'dioactive Matenais Pack.

NWEG/CP-0144 V01: A WORKSHOP ON DEVELOPING RISK ASSESS-Reactor Operator MENT METHODS FOR MEDICAL USE OF RADIOACTIVE NUREG 1122 RO1: KNOWLEDGE AND ABILITIES CATALOG FOR NU-CLEAR POWER PLANT OPERATORS Pr.ESSURIZED WATER REAC-MATERIALSummary.

NUREG/CP-0144 V02: A WORKSHOP ON DEVELOPING RISK ASSESS-TORS.

MENT METHODS FOR MEDICAL USE OF RADIOACTIVE NUREG 1123 RO1: KNOWLEDGE AND ABILITIES CATALOG FOR NU-MATERIAL.Supporbng Documents.

CLEAR POWER PLANT OPERATORS: BOILING WATER REACTORS..

NUREG/CR 2907 V13. RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS. Annual Report 1992 Reactor Pressure Vessel NUREG/CR-2907 V14. RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR-6214: PRODUCTION AND TESTING OF THE VITAMIN-86 NUCLEAR POWER PLANTS.

FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

Radioactive Material Safety Program U LEAR DATA' BIAX1AL NUREG 1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL LOADING EFFECTS ON FRACTURE NUREG/CH-6273:

SAFETY PROGRAMS AT MEDICAL FACluTIES Draft Report For Com TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL.

i ment NUREG/CR 6327: MODELS FOR EMBRITTLEMENT RECOVERY DUE TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

Radioactive Release NUREG/CR.2850 V13: DOSE COMMITMENTS DUE TO RADIOACTIVE p,,,,,,3,,,,y f

RELEASIS FROM NUCLEAR POWER PLANT SITES IN 1991' NUREG/CP-0140 VOI: PROCEEDINGS OF THE TWENTY-SECOND WATER REACTOR SAFETY INFORMATION MEETING.Pienary Ses-l Rad 6ciogical Criteria sion. Advanced instrumentahon & Control Hardware & Software, NUREG/CR-6307:

SUMMARY

OF COMMENTS RECEIVED AT WORK.

SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO Human Factors Research, IPE & PRA FACluTATE PUBUC PARTICIPATION IN DECOMMISSIONING NUREG/CP-0140 V02: PROCEEDINGS OF THE TWENTY SECOND WATER REACTOR SAFETY INFORMATION MEETING Severe Acci-CASES.

dent Research, Thermal Hydrauhc Research For Advanced Passive j

LWRs, High-Burnup Fuel Behavior.

Radiological Exposure NURrG/CP-0140 V03-PROCEEDINGS OF THE TWENTY SECOND NUREG-1530: REASSESSMENT OF NRC'S DOLLAR PER PERSON-REM CONVERSION FACTOR POLICY.

WAiER REACTOR SAFETY INFORMATION MEETING Pnmary Sys-l t

1 I

80 Subject Index i

tems integnty, Structural And Seismic Engineenng. Aging Research, Relay Products And Apphcations.

NUREG/CR4220. AN ASSESSMENT OF FIRE VULNERABILITY FOR NUREG/CR-6141: HANDBOOK OF METHODS FOR RISK BASED AGED ELECTRICAL RELAYS.

ANALYSES OF TECHNICAL SPECIFICATIONS.

NUREG/CR-6265: MULTlDISCIPLINARY FRAMEWORK FOR HUMAN Release Rate RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF NUREG/CR4261: A

SUMMARY

OF ORNL FISSION PRODUCT RE-COMMISSION AND DEPENDENCIES.

LEASE TESTS WITH RECOMMENDED RELEASE RATES AND DIF-NUREG/CR-6311: EVALUATING PREDICTION UNCERTAINTY-FUSION COEFFICIENTS.

Reactor Safety Research Rel6ef Valve NUREG/CP-0148: TRANSACTIONS OF THE TWENTY-THIRD WATER NUREG/CR-6192: AGING AND SERVICE WEAR OF SPRING-LOADED REACTOR SAFETY INFORMATION MEETING.

PRESSURE RELIEF VALVES USED IN SAFETY-RELATED SYSTEMS AT NEEAR NER MTS.

Reconstitution Program NUREG-1528: RECONSTITUTION OF THE MANUAL CHAPTER 2512 Remote Aftertoad Brachytherapy CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT 1-NUREG/CR4276: A COMPILATION OF CURRENT REGULATIONS, STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING BRA-Redoot Lake CHYTHERAPY' NUREG/GR-0014: BALDCYPRESS TREE RING ELEMENTAL CONCEN-TRATIONS AT REELFOOT LAKE. TENNESSEE,FROM AD 1795 TO Remote Aftertoading Brachytherapy AD 1820.

l NUREG/CR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE Refueling Outage AFTERLOADING BRACHYTHERAPY. Human Error And Cntical Tasks NUREG/CR4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC.

In Remote Afterloading Brachytherapy And Approaches For improved 6

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT W Q CR I HUMAN FACTORS EVALUATION OF REMOTE 2 V GRAND GULF, UNIT 1. Evaluation Of Severe Accident Risks For Plant AFTERLOADING BRACHYTHERAPY. Function And Task Analyses.

Operational State $ Dunng A Refueling Outage. Main Report And Ap-NUREG/CR4125 V03: HUMAN FACTORS EVALUATION OF REMOTE NU

/CR-6143 V06 P2 EVALUATION OF POTENTIAL SEVERE AC-AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT System intedacesh udures And Rams, Training W Ogamam al Rames And hedures.

GRAND GULF, UNIT 1 Evaluation Of Severe Accident Raska For Plant Operational State 5 Dunng A Refueling Outage Supporting MELCOR Report To Congress Calculations NUREG-0090 V17 N04. REPORT TO CONGRESS ON ABNORMAL OCCURRENCES October-December 1994.

REG @90 W8 Not REN TO CONMSS ON ABNORMAL NU 1493. PERFORMANCE-BASED CONTAINMENT LEAK TEST PROGRAM Final Report.

anuaWa@ N NUREG/CR 5975 ROI: INCENTIVE REGULATION OF INVESTOR' Reporting Requ6rement OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGULA-NUREG/CR4330: RESULTS OF REGULATORY IMPACT SURVEY OF TORS INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE NUREG/CR4315: CANDU REACTORS, THElR REGULATION IN CANADA, AND THE IDENTIFICATION OF FELEVANT NRC SAFETY OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

ISSUES-Requalification Examination NUREG/CR 6323. RELATIVE RISK ANALYSIS IN REGULATING THE NUREG/CR4355: A LIMITED ASSESSMENT OF THE ASEP HUMAN USE OF RADIATION-EMITTING MEDICAL DEVICES A Preimnary Ap-RELIABILITY ANALYSIS PROCEDURE USING SIMULATOR EXAMI-i pication NATION RESULTS.

Regulatory Agenda Rosin l

NUREG 0936 V13 NO3 NRC REGULATORY AGENDA. Semiannual NUREG/CR-5229 V07: FIELD LYSIMETER INVESTIGATIONS: LOW-l NU 093 V4 01: NRC REGULATORY AGENDA. Semiannual LEVEL WASTE DATA BASE DEVELOPMENT DROGRAM FOR Report' January' June 1995 FISCAL YEAR 1994. Annual Report.

NUREG/CR-6256 V01: FIELD LYSIMETER INVESTIGATIONS - TEST Regulatory Analysis RESULTS Low-Level Wace Data Base Program. Test Results For NUREG 1530: REASSESSMENT OF NRC'S DOLLAR PER PERSON.

Fiscal Years 1986,1987,1988, And 1989, REM CONVERSION FACTOR POLICY.

NUREG/CR-6349-COST-BENEFIT CONSIDERATIONS IN REGULA-Risk Analysis TORY ANALYSIS' NUREG/CR4311: EVALUATING PREDICTION UNCERTA!NTY.

NUREG/CR-6323: RELATIVE RISK ANALYSIS IN REGULATING THE l

Regulatory And Techn6 cal Report USE OF RADIATION-EMITTING MEDICAL DEVICES.A Preliminary Ap-i NUREG-0304 V19 NO3 REGULATORY AND TECHNICAL REPORTS plication.

l (ABSTRACT INDEX JOURNAL). Compilation For Third Quarter 1994. July September.

Risk Assessment NUHEG-0304 V19 N04 REGULATORY AND TECHNICAL REPORTS NUREG/CP-0144 VOI: A WORKSHOP ON DEVELOPING RISK ASSESS-(ABSTRACT INDEX JOURNAL) Annual Compdation For 1994 MENT METHODS FOR MEDICAL USE OF RADIOACTIVE I

NUREG-0304 V20 N01: RF.GULATORY AND TECHNICAL REPORTS MATERIALSummary.

l (ABSTRACT INDEX JOURNAL). Compilation For First Ouarter NUREG/CP-0144 V02: A WORKSHOP ON DEVELOPING R:SK ASSESS-1995. January March.

MENT METHODS FOR MEDICAL USE OF RADIOACTIVE i

NUREG-0304 V20 NO2: REGULATORY AND TECHNICAL REPORTS MATERIALSupporting Documents.

(ABSTRACT INDEX JOURNAL) Compdation For Second Quarter l

1995,Apni-June Robust Laser interferometer NUREG/CR4313 V01: ROBUST, ACCURATE, AND NON CONTACTING Regulatory Document VIBRATION MEASUREMENT SYSTEM. Summary of Companson Meas-NUREG/CR-5973 R02: COCM AND STANDARDS AND OTHER GUID.

urements Of The Robust Laser interferometer And Typcal Accelerome-ANCE CITED IN REGULATORY DOCUMENTS.

ter Systems.

NUREC-/CR-6313 V02: ROBUST, ACCURATE, AND NON-CONTACTING Regulatory impact Survey V)BR.ATION MEASUREMENT SYSTEMS. Supplemental Appendices NUREG/CR4330: RESULTS OF REGULATORY IMPACT SURVEY OF Presenting Companson Meawrements Of The Robust Laser Interfer-INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE orneter And Typcal Accelerorneter Systems.

OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS.

Rules Reinforced Concrete NUREG-0936 V13 NO3-NRC REGULATORY AGENDA. Semiannual NUREG/CR4184: SEPARATE EFFECTS TESTING AND ANALYSES TO Report, July-December 1994 INVESTIGATE LINER TEARING OF THE 1:16-SCALE REINFORCED NUREG-0936 V14 NOI: NRC REGULATORY AGENDA. Semiannual CONCRETE CONTAINMENT BUILDING.

Report, January-June 1995.

I l

Subject Index 81 j

SAPHIRE Sofety Standard NUREG/CR4116 V06: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4349: COST-BENEFIT CONSIDERATIONS IN REGULA-HANDS ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

TORY ANALYSIS.

VERSION 6.0. Graphical Evaluation Module (GEM) Reference Manual.

l NUREG/CR4116 V09-SYSTEMS ANALYSIS PROGRAMS FOR Safety Testing HANDS ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR4074 V04: SEALED SOURCE AND DEVICE DESIGN VERSION 5 0.Venfication And Vahdation (V&V) Manual SAFETY TESTING.Techncal Report Ori The Findings Of Task 4.Inves-NUREG/CR4116 VIO: SYSTEMS ANALYSIS PROGRAMS FOR tigation Of Sealed Source for Paper Mill Digester.

HANDS ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR4074 VOS: SEALED SOURCE AND DEVICE DESIGN VERSION 5.0. Data Loading Manual.

SAFETY TESTING.Tochncal Report On The Findings Of Task 4 inves-tigation Of Failed Radioactive Stainless Steel Troxler Gauges.

SCALE NUREG/CR-0200 V1 R04: SCALE: A MODULAR CODE SYSTEM FOR Safety-Related System PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-NUREG/CR4016: AGING AND SERVICE WEAR OF AIR OPERATED CENSING EVALUATION. Control Modules.

VALVES USED IN SAFETY RELATED SYSTEMS AT NUCLEAR NUREG/CR-0200 V2P1R4: SCALE: A MODULAR CODE SYSTEM FOR POWER PLANTS.

l PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Lt.

NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-LOADED PRESSURE RELIEF VALVES USED IN SAFETY-RELATED SYSTEMS CENSING EVALUATION Functional Modules F1 F8.

NUREG/CR 0200 V2P2R4: SCALE. A MODULAR CODE SYSTEM FOR AT NUCLEAR POWER PLANTS.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-Sealed Sowce CENSING EVALUATION Functonal Modules F9-F16.

NUREG/CR 0200 V3 R04: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4074 V04: SEALED SOURCE AND DEVICE DESIGN PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Lt.

"" TTY TESTING. Technical Report On The Findings Of Task 4 inves-m Of Sealed Source for Paper Mill D ester.

CENSING EVALUATION. Miscellaneous.

A/CR4074 V05: SEALED SOURC AND DEVICE DESIGN SCDAP 2TY TESTING. Technical Report On The Findings Of Task 4.inves-NUREG/CR4150 V01:

SCDAP/RELAPS/ MOD 3.1 CODE ugation Of Failed Radioactive Stainless Steel Troxler Gauges.

MANUALinterface Theory.

NUREG/CR 4150 V02:

SCDAP/RELAP/ MOD 3.1 CODE Sedier:entation NUREG/CR4368: EXPERIMENTAL INVESTIGATION OF SEDIMENTA-MANUAL. Damage Progression Model Theory.

NUREG/CR-6150 VUi SCDAP/RELAP5/ MOD 3.1 CODE TlON OF LOCA GENERATED FIBROUS DEBRIS AND SLUDGE IN MANUAL. User's Guide And input Manual.

BWR SUPPRESSION POOLS.

NUREG/CR 6150 V04

$CDAP/RELAP5/ MOD 3.1 CODE Seismic MANUALMATF'10--A Library Of Matenals Properties For Light-Water, NUREG/CR4144 V01: EVALUATION OF POTENTIAL SEVERE ACCl-Reactor Accident Analysis.

NUREG/CR4150 V05.

SCDAP/RELAP5/ MOD 3.1 CODE DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT MANUAL Developmental Assessment.

SURRY, UNIT 1. Summary Of Results.

Safeguards Summary Event List Seismic Design NUREG-0525 V02 R03: SAFEGUARDS

SUMMARY

EVENT Ll6T NUREG/CR4243: APPLICATION OF BOUNDING SPECTRA TO SEIS-(SSEL) January 1,1990 Through December 31,1994.

MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO Safety Application STRONG MOTION EARTHOUAKES.

NUREG/CR4287: MANAGEMENT CONCEPTS AND SAFETY APPLICA-TIONS FOR NUCLEAR FUEL FACILITIES.

Senettivity Analysis NUREG/CR 6134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF Safety Criterte CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-NUREG/CR4284: CRITICALITY SAFETY CRITERIA FOR LICENSE CIDENT CONSEQUENCE MODEL REVIEW OF LOW-LEVEL WASTE FACILITIES NUREG/CR-6135' UNCERTAINTY AND SENSITIVITY ANALYSIS OF EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-Safety Evaluation Repod DENT CONSEQUENCE MODEL NUREG-0847 S15: SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND FOOD PATHWAY RESULTS WITH THE MACCS fiEACTOR ACCl-2 Docket Nos. 50-390 And 50-391.(Tennessee Valley Authonty)

DENT CONSEQUENCE MODEL.

NUREG 0847 S16 SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-6311: EVALUATING PREDICTION UNCERTAINTY.

OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND Severe Accident 2 Docket Nos. 50 390 And 50-391.(Tennessee Valley Authonty)

NUREG-0847 S17: SAFETY EVALUATION REPORT RELATED TO THE NUREG-1465: ACCIDENT SOURCE TERMS FOR LIGHT-WATER NU.

OPERATION OF WATTS BAR NUCLEAR PLANT. UNITS 1 AND CLEAR POWER PLANTS.

NUREG 1519. SURFACE INTERACTIONS OF CESIUM AND BORIC 2.Docliet Nos. 50 390 And 50-391.(Tennessee Valley Authonty)

NUREGOB47 S18: SAFETY EVALUATION REPORT RELATED TO THE ACfD W(TH STAINLESS STEEL.

OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND NUREG/CP-0140 V02: PROCEEDINGS OF THE TWENTY SECOND WATER REACTOR SAFETY INFORMATION MEETING Severe Acce

2. Docket Nos. 50 390 And 50 391.(Tennessee Valley Authonty)

NUREG 0847 S19: SAFETY EVALUATION REPORT RELATED TO THE dent Research, Thermal Hydraulic Research For Advanced Passive OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND LWRs. High Burnup Fuef Behavior.

NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE BY

2. Docket Nos 50-390 And 50-391.(Tennessee Valley Authonty)

DIRECT CONTAINMENT HEATING IN SURRY.

NUREG/CR-6119 VD1: MELCOR COMPUTER CODE MANUALS Pnmer Safetyleaues And User's Guides. Verson 1.8.3 September 1994.

NUREG-1435 SO4: STATUS OF SAFETY ISSUES AT LICENSED NUREG/CR-6119 V02:

MELCOR COMPUTER CODE POWER PLANTS.TMI Action Plan Requirements. Unresolved Safety MANUALS Reference Manuals Version 18 3 September 1994.

lasues.Genenc Safety lasues.Other Multiplant Acton lasues NUREG/CR4143 VOI: EVALUATION OF POTENTIAL SEVERE ACCl-NUREG-1517; REPORT OF THE SOUTH TEXAS PROJECT ALLEGA.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TIONS REVIEW TEAM.

NUREG/CR4315: CANDU REACTORS. THEIR REGULATION IN GRAND GULF. UNIT 1 Summary Of Results.

NUREG/CR-6143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-CANADA, AND THE IDENTIFICATION OF RELEVANT NRC SAFETY CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT ISSUES.

GRAND GULF, UNIT 1.Evaluaton Of Severe Accident Risks For Plant Operatonal State 5 Dunng A Refueling Outage Main Report And Ap-Safety Management pendices NUREG/CR4287: MANAGEMENT CONCEPTS AND SAFETY APPLICA.

NUREG/CR4143 V06 P2 EVALUATION OF POTENTIAL SEVERE AC-TlONS FOR NUCLEAR FUEL FACILITIES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1.Evaluaton Of Severe Accident Risks For Plant l

Safety Reeeerch NUREG-1266 V09: NRC SAFETY RESEARCH IN SUPPORT OF REGU-Operatonal State 5 Dunng A Refuehng Outage Supporting MELCOR LATION. FY 1994-Calculatons

82 Subject index NUREG/CR4144 V01: EVALUATION OF POTENTIAL SEVERE ACCl-Software DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CP 0145:

WORKSHOP ON DEVELOPING SAFE SURRY, UNIT 1.Sumrnary Of Results.

SOFTWARE Held At Hotel Del Coronado, San Diego.CA, July 22-NUREG/CR4144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC.

23.1992.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6293 V01: VERIFICATION AND VAllDATION GUIDELINES SURRY, UNIT 1 Evaluation Of Severe Accident Risk During Mid-Loop FOR HIGH INTEGRITY SYSTEMS Main Report.

Operatens Main Report.

NUREG/CR-6293 V02. VERIFICATIONN AND VAllDATION GUIDE-NUREG/CR-6144 VUU P2. EVALUATION OF POTENTIAL SEVERE AC.

LINES FOR HIGH INTEGRITY SYSTEMS Appendices A D.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION AND SURRY, UNIT 1 Evaluaton Of Severe Accident Risk During Mid-Loop VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Operations Appendicos.

AL SOFTWARE NUREG/CR-6154 V02: EXPERIMENTAL RESULTS FROM CONTAIN.

NUREG/CR-6316 ' V02: GUIDELINES FOR THE VERIFICATION AND MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-VEk C I A I L CR ION AL SOFTWARE Survey And Assessment Of Conventional Software NU G/ R 28 STUDIES AT THE INEL Venfication And Vahdation Methods.

NUREG/CR-6316 V03 GUIDELINES FOR THE VERIFICATION AND Shiftwork VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR-6046: ALERTNESS. PERFORMANCE. AND OFF DUTY AL SOFTWARE Survey And Documentaten Of Expert System Venfica-SLEEP ON 8-HOUR AND 12-HOUR N!GHT SHIFTS IN A SIMULATED t on And Validation Methodologies.

CONTINUOUS OPERATIONS CONTROL ROOM SETTING, NUREG/CR-6316 V04. GUIDELINES FOR THE VERIFICATION AND VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL Shipment SOFTWARE. Evaluation Of Knowledge Base Certrtication Methods.

NUREG 0725 R10: PUBLIC INFORMATION CIRCULAR FOR SHIP-NUREG/CR 6316 V05. GUIDELINES FOR THE VERIFICATION AND MENTS OF 1RRADIATED REACTOR FUEL VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Cask AL SOFTWARE Rationale And Desenpton Of V8V Guidehne Packages

^"

NUREG/CR 5657: AUTOCASK (AUTOMATIC GENERATION OF 3-D NUR G C 16 V06. GUIDELINES FOR THE VERIFICATION AND CASK M DELS) A rocomputer Based System For Shipping Cask VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

NUREb/CR 6322: BUbKLING ANALYSIS OF SPENT FUEL BASKET AL SOFTWARE.Vahdaten Scenanos.

1 NUREGICR-6316 V07: GUIDELINES FOR THE VERIFICATION AND Sh6ppingport Reactor VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR-6275. MECHANICAL PROPERTIES OF THERMALLY AGED AL SOFTWARE. User's Manual.

CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-NUREG/CR 6316 V08: GUIDELINES FOR THE VERIFICATION AND PONENTS.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

AL SOFTWARE.Bibhography.

NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING Sohd Waste D6sposal WELDS Seventh Program Report March 1993 December 1994' NUREG/CR-2907 V13: RADIOACTIVE MATERIALS RELEASED FROM NtICLEAR POWER PLANTS Annual Report 1992.

EG/C 61 3 V01: EVALUATION OF POTENTIAL SEVERE ACCI-NUREG/CR 2907 V14: RADIOACTIVE MATERIALS RELEASED FROM DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUCLEAR POWER PLANTS.

1 GRAND GULF, UNIT 1 Summary Of Results.

NUREG/CR4143 V06 Pt: EVALUATION OF POTENTIAL SEVERE AC.

Sohdification CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6256 V02: FIELD LYSIMETER INVESTIGATIONS - TEST GRAND GULF, UNIT 1 Evaluation Of Severe Accident Risits For Plant RESULTS Low-Level Waste Data Base Development Program: Test s

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

Results For Fiscal Years 1990,1991,1992, And 1993.

pendices NUREG/CR-6143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC.

Source Term CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG 1465: ACCIDENT SOURCE TERMS FOR LIGHT WATER NU.

GRAND GULF. UNIT 1 Evaluaton Of Severe Accident Risks For Plant CLEAR POWER PLANTS.

}

Operat onal State 5 Dunng A Refuehng Outage Supporting MELCOR Calculations.

South Texas Project NUREG/CR-6144 V01 EVALUATION OF POTENTIAL SEVERE ACCI-NUREG 1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TIONS REVIEW TEAM.

SURRY, UNIT 1 Summary Of Results NUREG/CR-6144 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-Special Nuclear Material CiDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG-1065 R02: ACCEPTABLE STANDARD FORMAT AND CON-SURRY, UNIT 1 Evaluation Of Severe Accident Risk Dunng Mid-Loop TENT FOR THE FUNDAMENTAL NUCLEAR MATERIAL CONTROL Operations Main Report (FNMC) PLAN REQUIRED FOR LOW ENRICHED URAN 1UM FACILI-NUHEG/CR-6144 V06 P2: EVALUATION OF POTENTIAL SEVERE AC-TIES.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT SURRY, UNIT 1 Evaluation Of Severe Accident Risk Dunng Mid-Loop Spent Fuel Operatons Appendices.

NUREG 1323 RO1: LICENSE APPLICATION REVIEW PLAN FOR A GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND HIGH-A a

e an.

REG 1527: NRC'S OBJECT-ORIENTED SIMULATOR INSTRUCTOR STATION Spent Fuel Basket S6mulator Esaminat6on NUREG/CR-6322: BUCKLING ANALYSIS OF SPENT FUEL BASLET.

NUREG/CR-6355: A LIMITED ASSESSMENT OF THE ASEP HUMAN LIABIL TY ANA.YSIS PROCEDURE USING SIMULATOR EXAMi-Stalniese Stel ACID WITH STAINLESS STEEL Site Specifle Adv6sory Board NUREG/CR-6251: STAINLESS STEEL SUBMERGED ARC WELD NUREG/CR-6307:

SUMMARY

OF COMMENTS RECEIVED AT WORK.

FUSION LINE TOUGHNESS.

t SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO NUREG/CR 6275: MECHANICAL PROPERTIES OF THERMALLY AGED FACILITATE PUBLIC PARTICIPATION IN DECOMMISSIONING CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-CASES.

PONENTS.

Studge Standard NUREG/CR 6368: EXPERIMENTAL INVESTIGATION OF SEDIMENTA.

NUREG/CR-6385. COMPARISONS OF ANS ASME,AWS, AND NFPA TION OF LOCA GENERATED FIBROUS DEBRIS AND SLUDGE IN STANDARDS CITED IN THE NRC STANDARD REVIE W BWR SUPPRESSION POOLS.

PLAN NUREG-0800, AND RELATED DOCUMENTS.

Subject index 83 NUREG/CR-6386: COMPARISONS OF ANSI STANDARDS CITED IN Suppression Pool THE NRC STANDARD REVIEW PLAN,NUREG 0800, AND RELATED NUREG/CR 6368 EXPERIMENTAL INVESTIGATION OF SEDiMENTA.

DOCUMENTS.

TION OF LOCA-GENERATED FIBROUS DEBRIS AND SLUDGE IN BWR SUPPRESSION POOLS.

p NUREG/CR-5973 R02: CODES AND STANDARDS AND OTHER GUID-Surface Interact 6on ANCE CITED IN REGULATORY DOCUMENTS NUREG 1519. SURFACE INTERACTIONS OF CESIUM AND BORIC NUREG/CR 6382. COMPARISONS OF ASTM STANDARDS CITED IN ACID WITH STAINLESS STEEL THE NRC STANDARD REVIEW PLAN NUREG-0800, AND RELATED DOCUMENTS Surveillance Requ6rement NUREG/CR-6385: COMPARISONS OF ANS,ASME.AWS. AND NFPA NURFG/CR 6141: HANDBOOK OF METHODS FOR RISK BASED STANDARDS CITED IN THE NRC STANDARD REVIEW ANALYSES OF TECHNICAL SPECIFICATIONS.

PLAN,NUREG 0800, AND RELATED DOCUMENTS NUREG/CR 6386: COMPARISONS OF ANSI STANDARDS CITED IN Surve611ance System THE NRC STANDARD REVIEW PLAN NUREG-0800, AND RELATED NUREG/CR-6420: SELF MONITORING SURVEILLANCE SYSTEM FOR DOCUMENTS-PRESTRESSING TENDONS.

Standard Technical Specificat6ons NUREG-1430 V0I R01. STANDARD TECHNICAL SPECIFICATIONS I""Y NUREG/CR-6239 V01: SURVEY OF STRONG MOTION EARTHOUAKE BABCOCK AND WILCOx PLANTS Specifications NUREG-1430 V02 R0t; STANDARD TECHNICAL SPECIFICATIONS EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-PHASIS ON PIPING SYSTEMS Main Report.

BABCOCK AND WILCOX PLANTS Bases (Sectons 2 0 - 3 3) CATIONS NUREG 1430 V03 R01: STANDARD TECHNICAL SPECIFI NUREG/CR.6239 V02. SURVEY OF STRONG MOTION EARTHOUAKE BABCOCK AND WILCOX PLANTS Bases (Sectons 3 4 3 9)

EFFECTS ON THERMAL POWER PLANTS IN CAllFORNIA WITH EM-NUREG-1431 V01 ROI. STANDARD TECHNICAL SPECIFICATIONS PHASIS ON PIPING SYSTEMS Appendees.

WESTINGHOUSE PLANTS Specifcations NUREG 1431 V02 RO1: STANDARD TECHNICAL SPECIFICATIONS Survey Design WESTINGHOUSE PLANTS Bases (Sectons 2.0 3 3).

NUREG 1505 DRFT FC: A NONPARAMETRIC STATISTICAL METHOD-NUREG-1431 V03 ROI: STANDARD TECHNICAL SPECIFICATIONS OLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS DE-WESTINGHOUSE PLANTS Bases (Sectons 3 4 - 3 9)

COMMISSIONING SURVEYS Draft Report For Comment.

NUREG 1432 VOI R01: STANDARD TECHNICAL SPECIFICATIONS NUREG 1506 DRFT FC: MEASUREMENT METHODS FOR RADIOLOGi-lYA PECIFICATIONS b

O NU E 432 0 01 Ni NURIG 7

MN M ETECTABLE CONCENTRATIONS NU EG 432 0 1

PE IICA ONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS COMBUSTION ENGINEERING PLANTS Bases (Sect ons 3 4 - 3.9)

NUREG-1433 V01 ROI: STANDARD TECHNICAL SPECIFICATIONS CONTAMINANTS AND FIELD CONDITIONS. Draft Report For Com.

ment.

GENERAL ELECTRIC PLANTS BWR/4 Specifcations.

NUREG 1433 V02 A01: STANDARD TECHNICAL SPECIFICATIONS System Safety GENEFIAL ELECTRIC PLANTS BWR/4 Bases (Sectons 2 0 3 3) ONS NUREG-1433 V03 RO1: STANDARD TECHNICAL SPECIFICATI NUREG/CP-0145:

WORKSHOP ON DEVELOPING SAFE GENERAL ELECTRIC PLANTS BWR/4 Bases (Sectons 3.4 - 310)

SOFTWARE Held At Hotel Del Coronado, San Diego.CA. July 22 NUREG-1434 V01 ROI: STANDARD TECHNICAL SPECIFICATIONS 23,1992.

GENERAL ELECTRIC PLANTS. BWR/6 Specificatons NUREG 1434 V02 RO1: STANDARD TECHNICAL SPECIFICATIONS Systems integrity GENERAL ELECTRIC PLANTS. BWR/6 Bases (Sectons 2.0 - 3 3)

NUREG/CP-0140 V03: PROCEEDINGS OF THE TWENTY SECOND NUREG-1434 V03 R01: STANDARD TECHNICAL SPECIFICATIONS WATER REACTOR SAFETY INFORMATION MEETING.Pnmary Sys-GENERAL ELECTRIC PLANTS. BWR/6 Bases 3 4 - 3.10) tems Integnty, Structural And Scismic Engineenng. Aging Research, Products And Applicatons.

Steady-State Initializat6on NUREG/CR-6325 AN IMPLICIT STEADY STATE INITIALIZATION PACK-TLo AGE FOR THE RELAPS COMPUTER CODE.

NUREG-0837 V14 N04 NRC TLD DIRECT RADIATION MONITORING NETWORK.Pr ress Report. October-December 1994.

U G/C 6242. CASKS (COMPUTER ANALYSIS OF STORAGE NET ORKP ress e Janua Mar h 19 CASKS) A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR NUREG-0837 V15 NO2: NRC TLD DIRECT RADIATION MONITORING STORAGE CASK DESIGN REVIEW User's Manual To Version 1b (In-NETWORK. Progress Report.Apnt-June 1995-giuding Program Reference)

NUREG-0837 V15 NO3 NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report July-September 1995.

Strainer Blockage NUREG/CR-6224 PARAMETRIC STUDY OF THE POTENTIAL FOR TMI Action Plan BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED NUREG-1435 SO4. STATUS OF SAFETY ISSUES AT LICENSED DEBRIS.

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

Stress Corrosion Cracking NUREG/CR-4667 V18. ENVIRONMENTALLY ASSISTED CRACKING IN top l

LIGHT WATER REACTORS. Semiannual Report, October 1993 - March NUREG-1514: GUIDANCE FOR A LARGE T ABLETOP EXERCISE FOR A N

G'CR 4667 V19-ENVIRONMENTALLY ASSISTED CRACKING IN NUCLEAR POWER PLANT, LIGHT WATER REACTORS Semiannual Report.Apni-September 1994 Task Procedure NUREG/CR-6125 V03. HUf,iAN FACTORS EVALUATION OF REMOTE Stress Trias6ailty NUREG/CR-6259 CONSTRAINT EFFECTS ON FRACTURE INITIATION AFTERLOADING BRACHYTHERAPY Supporting Analyses Of Human-LOADS IN HSST WIDE PLATE TESTS.

System Interfaces. Procedures And Practces. Training And Organizat on-al Practices And Procedures Structural Analysis NUREG/CR-5657: AUTOCASK (AUTOMATIC GENERATION OF 3-D Technical Specifications CASK MODELS).A Mcrocomputer Based System For Shipping Cask NUREG/CR-6141. HANDBOOK OF METHODS FOR RISK-BASED Design Review Analysis.

ANALYSES OF TECHNICAL SPECIFICATIONS.

NUREG/CR 6172: REVIEWING PSA BASED ANALYSES TO MODIFY Structural Codes And Standards TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

NUREG/CR-6358 VOI. ASSESSMENT OF UNITED STATES INDUSTRY STRUCTURAL CODES AND STANDARDS FOR APPLICATION TO Teletherapy ADVANCED NUCLEAR POWER REACTORS Final Report.

NUREG/CR-6277 VO1: HUMAN FACTORS EVALUATION OF TELE-NUREG/CR-6358 V02: ASSESSMENT OF UNITED STATES INDUSTRY THERAPY. Identification Of Problems And Asternative Approacf es.

STRUCTURAL CODES AND STANDARDS FOR APPLICATION TO NUREG/CR 6277 V02: HUMAN FACTORS EVALUATION OF ADVANCEC' "'JCLEAR POWER REACTORS. Appendices TELETHERAPY. Function And Task Analys:s.

~

84 Subject index NUREG/CR4277 V03: HUMAN FACTORS EVALUATION OF NUREG-0837 V15 NO3. NRC TLD DIRECT RADIATION MONITORING TELETHERAPY. Human-System interfaces And Procedures NETWORK Progress Report. July-September 1995-NUREG/CR4277 V04: HUMAN FACTORS EVALUATION Or TELETHERAPY, Training And Organizational Analysis.

Title List NUREG/CR 6277 V05: HUMAN FACTORS EVALUATION OF NUREG-0540 V16 N11: TITLE UST OF DOCUMENTS MADE PUBLICLY TELETHERAPY. Literature Review.

AVAILABLE. November 1 30, 1994.

NUREG-0540 V16 N12: TITLE UST OF DOCUMENTS MADE PUBLICLY AV^

E em gE 5O o h S OF DOCUMENTS MADE PUBLICLY G/CR SELF ITORING SURVEILLANCE SYSTEM FOR AVAILABLE. January 1 31,1995.

NUREG 0540 V17 NO2: TITLE LIST OF DOCUMENTS MADE PUBLICLY Thwepy Misadministration NUREG-1535. INGESTION OF PHOSPHORUS-32 AT MASSACHU.

AVAILABLE. February 1 28,1995.

SETTS INSTITUTE OF NUREG 0540 V17 NO3: TITLE UST OF DOCUMENTS MADE PUBLICLY TE CHNOLOGY. CAMBRIDGE.M ASSACHUSETTS.lDENTIFIED ON AVAILABLE. March 1-31,1995.

AUGUST 19,1995.

NUREG 0540 V17 N04: TITLE UST OF DOCUMENTS MADE PUBUCLY AVAILABLE. April 1-30,1995.

Thermal Aging NUREG-0540 V17 NOS: TITLE UST OF DOCUMENTS MADE PUBUCLY NUREG/CR-6275: MECHANICAL PROPERTIES OF THERMALLY AGED AVAILABLE May 1-31.1995.

CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-NUREG-0540 V17 N06. TITLE UST OF DOCUMENTS MADE PUBUCLY PONENTS.

AVAILABLE. June 1 30.1995.

NUREG-0540 V17 N07; TITLE UST OF DOCUMENTS MADE PUBLICLY Thermal Bending AVAILABLE. July 1-31.1995.

NUREG/CP 0146: PROCEEDINGS OF THE WORKSHOP ON GATE NUREG 0540 V17 N08. TITLE LIST OF DOCUMENTS MADE PUBUCLY VALVE PRESSURE LOCKING AND THERMAL BINDING.

AVAILABLE. August 1 31,1995.

NUREG-0540 V17 N09: TITLE UST OF DOCUMENTS MADE PUBLICLY AVAILABLE. September 1-30.1995.

U G 23 01: SURVEY OF STRONG MOTION EARTHOUAKE RE 40 EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-4vA 8LE' to 1 1'19 5' PHASIS ON PIPING SYSTEMS Main Report.

i NUREG/CR4239 V02: SURVEY OF STRONG MOTION EAR 1rlOUAKE

" ^

"'^ ""

RG V09 N01: TOPICAL REPORT REVIEW STATUS.

S ON PIP N YSTE Ap ic Thermal-Hydraunc Transient Anahsia l

NUREG/CP 0140 V02: PROCEEDINGS OF THE TWENTY SECOND NUREG/CR-5535 V01: RELAPS/ MOD 3 CODE MANUAL. Code Structure, WATER REACTOR SAFETY INFORMATION NEETING. Severe Acci System Models. And Solution Methods.

dent Research. Thermal Hydraulic Research For Advanced Passive NUREG/CR-5535 V02: RELAPS/ MOD 3 CODE MANUALUser's Guide LWRs. Hgh-Burnup Fuel Behavior.

And input Requirements.

NUREG/CP-0142 V01: PROCEEDINGS OF THE 'TH INTERNATIONAL NUREG/CR-5535 V04. RELAP5/ MOD 3 CODE MANUALModels And MEETING ON NUCLEAR REACTOR THt3 MAL-HYDRAULICS Cerrelations.

(NURETH-7) Sessions 1-5.

NUREG/CR-5535 VOS R1: RELAPS/ MOD 3 CODE MANUALUser's NUREG/CP-0142 V02: PROCEEDINGS OF THE 71H INTERNATIONAL Guideline.

MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS NUREG/CR-6257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC (NURETH-7tSessions 6-11.

ENERGY OF CANADA LTD CODES.

NUREG/CP-0142 V03. PROCEEDINGS OF THE 7TH INTERNATIONAL MEETING ON NUCLEAR REACTOR THERMAL-HYDRAUUCS Transportation (NURETH-7) Sessions 12-16.

NUREG-0383 V01 R18. DIRECTORY OF CERTIFICATES OF COMPU-NUREG/CP 0142 V04: PROCEEDINGS OF THE 7TH INTERNATIONAL ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS Approved Packages.

(NURETH-7) Sessions 17 24-NUREG-0383 V02 R18: DIRECTORY OF CERTIFICATES OF COMPU-NUREG/CR 5535 VO1: RELAP5/ MOD 3 CODE MANUAL Code Structure.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of System Models, And Solution Methods-Compliance NUREG/CR 5535 V02: RELAP5/ MOD 3 CODE MANUALUser's Guide NUREG-0383 V03 R15: DIRECTom? OF CERTIFICATES OF COMPLi-E NUR /C V

RELAP5/ MOD 3 CODE MANUALModels And Approved Quality Assursr.co Programs ior Radeoactive Matenals Pack.

Correim NUREG/CR 5535 V05 R1: RELAPS/ MOD 3 CODE MANUAL User's ages.

Guideline.

NUREG/CR4004: PROBABILISTIC PIPE FRACTURE EVALUATIONS Tree mng FOR LEAK RATE-DETECTION APPLICATIONS.

NUREG/GR-0014: BALD'.;YPRt.'?

• REE RING ELEMENTAL CONCEN-NUREG/CR4119 V01: MELCOR COMPUTER CODE MANUALS.Pnmer TRATIONS A7 REELFOOT LAKE. TENNESSEE.FROM AD 1795 TO And User's Guides. Version 18.3 September 1994.

AD 1820.

NUREG/CR-6119 V02.

MELCOR COMPUTER CODE MANUALS Reference Manuals Version 18 3 September 1994.

Trouler Gauge NUREG/CR4074 V05: SEALED SOURCE AND DEVICE DESIGN r

Thermocoupee SAFETY TESTING Technical Report On The Findings Of Task 4.Inves-NUREG/CR 6334 NEW SENSOR FOR MEASUREMENT OF LOW AIR tigation Of Failed Radioactive Stainless Steel Troxler Gauges.

FLOW VELOCITY. Phase i Final Report.

Turbine Drive Thermohydraunc NUREG/CR-5857: AGING OF TURBINE DRIVES FOR SAFETY-RELAT.

NUREG/CR4257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC ED PUMPS IN NUCLEAR POWER PLANTS.

ENERGY OF CANADA LTD CODES Turb6ne Generator NUREG-1275 Vit: OPERATING EXPERIENCE FEEDBACK REPORT.

U

/CR 48 THERMALLY DRIVEN MOISTURE REDISTRIBUTION TURBINE-GENERATOR OVERSPEED PROTECTION IN PARTIALLY SATURATED POROUS MEDIA-SYSTEM Commercial Power Reactors.

Thermolum6nescent Dosimeter NUREG-0837 V14 N04-NRC TLD DIRECT RADIATION MONITORING Twofhase Flow NETWORK, Progress Report. October-December 1994.

NUREG/CR4114 V02: AUXIUARY ANALYSES IN SUPPORT OF PER-NUREG-0037 V1b f401: NRC TLD DIRECT RADIATION MONITORING FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW LEVEL NETWORK. Progress Report. January-March 1995, WASTE FACluTY.Two-Phase Flow And Contamenant Transport in Un-I NUREG-0837 V15 NO2: NRC TLD DIRECT RADIATION MONITORING saturated Soils With Application To Low-Level Radioactive Weste Dis-NETWORK Progress Report.Apnl June 1995.

posal.

i

f Subject index 85 Uncertaenty Analyons NUREG 0040 V19 N01: LICENSEE CONTRACTOP AND VENDOR IN-NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE SPECTION STATUS REPORT, Quarte ty Report. January-March UNCERTAINTY ANALYSIS. Dispersion And Depostten Uncertainty 1995.(White Book)

Assessment Main Report.

NUREG 0040 V19 NO2: LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR 6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE SPECTION STATUS REPORT. Quarterly Report, April-June 1995.(White UNCERTAINTY ANALYSIS. Dispersion And Depositen Uncertainty Book)

Assessment Appendices A And B-NUFIEG-0040 V19 NO3; LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR 6244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE SPECTION STATUS REPORT. Quarterly Report. July-September UNCERTAINTY ANALYSIS. Disperson and Deposition Uncertainty 1995 (White Book)

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

Y**"

Uneatursted Rock NUREG/CR 6356. HYDRAULIC CHARACTERl2ATION OF HYDROTH-NUREG/CR-6293 V01: VERIFICATION AND VALIDATION GUIDELINES ERMALLY ALTERED NOPAL TUFF.

FOR HIGH INTEGRITY SYSTEMS. Main Report.

NUREGICR-6293 V02: VERIFICATIONN AND VALIDATION GUIDE-Unsaturated Soil LINES FOR HIGH INTEGRITY SYSTEMS. Appendices A.D.

NUREG/CR 6114 V02: AUXILIARY ANALYSES IN SUPPORT OF PER.

NUREG/CR-6316 V0t: GUIDELINES FOR THE VERIFICATION AND FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-LEVEL VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-WASTE FACILITY.Two-Phase Flow And Contaminant Transport in Un-AL SOFTWARE.

satuieted Soels With Apphcation To Low-Level Radioactive Waste Dis-NUREG/CR-6316 V02: GUIDELINES FOR THE VERIFICATION AND posal.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AL SOFTWARE Survey And Assessment Of Conventonal Software Uranium Venficaton And Vahdaten Methods.

NUREG/CR-6328. ADEQUACY OF THE 123 GROUP CROSS-SECTION NUREG/CR-6316 V03: GUIDELINES FOR THE VERIFICATION AND LIBRARY FOR CRITICALITY ANALYSES OF WATER-MODERATED VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-URANIUM SYSTEMS' AL SOFTWARE. Survey And Documentation Of Expert System Venfica-ton And Vahdaten Whodologes.

VITAM 6N-56 NUREG/CR-6316 V04: GUIDELINES FOR THE VERIFICATION AND NUREG/CR-6214: PRODUCTION AND TESTING OF THE VITAMIN-B6

^

FINE GROUP AND THE BUGLE 93 BROAD-GROUP NEUTRON /

SOFTWARE Evaluation Of Knowledge Base Certificaten Methnds.

2 PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-Vt NUREG/CR-6316 V05: GUIDELINES FOR THE VERIFICATION AND NUCLEAR DATA VAllDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

Val 6datlon AL SOFTWARE.Rahonale And Desenpton Of V&V Guidehne Packages NUREG/CR-6293 VOI: VERIFICATION AND VALIDATION GUIDELINES And Procedures.

FOR HIGH INTEGRITY SYSTEMS Main Report.

NUREG/CR-6316 V06. GUIDELINES FOR THE VERIFICATION AND NUREG/CR 6293 V02: VERIFICATIONN AND VALIDATION GUIDE-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-LINES FOR HIGH INTEGRITY SYSTEMS Appendices A.D.

AL SOFTWARE.Vahdaten Scenares.

NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION AND NUREG/CR-6316 V07: GUIDELINES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AL SOFTWARE.

AL SOFTWARE. User's Manual.

NUREG/CR-6316 V02: GUIDELINES FOR THE VERIFICATION AND NUREG/CR-6316 V08: GUIDELINES FOR THE VERIFICATION AND VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-AL SOFTWARE. Survey And Assessment Of Convenhonal Software AL SOFTWARE Biblography.

Venficaton And Vahdation Methods l

NUREG/CR-6316 V03 GUIDELINES FOR THE VERIF6 CATION AND Vessel l

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION.

NUREG/CR-6260. APPLICATION OF NUREG/CR-5999 INTERIM FA.

AL SOFTWARE. Survey And Documentation Of Expert System Venfica-TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-ton And Vahdalen Methodologies NENTS.

NUREG/CR-6316 V04: GUIDELINES FOR THE VERIFICATION AND VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL V6 brat 6on Measurement System SOFTWARE Evaluaton Of Knowledge Base Certificaton Methods.

NUREG/CR 6313 V01: ROBUST. ACCURATE, ANO NON CONTACTING NUREG/CR 6316 V05: GUIDELINES FOR THE VERIFICATION AND VIBRATION MEASUREMENT SYSTEM Summary of Companson Meas.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-urements Of The Robust Laser interferometer And Typical Accelerome-i AL SOFTWARE.Ratonale And Desenpton Of V&V Guidehne Packages ter Systems.

NUREG/CR-6313 V02: ROBUST, ACCURATE, AND NON-CONTACTING NUR G CR 16 V06: GUIDELINES FOR THE VERIFICATION AND VIBRATION MEASUREMENT SYSTEMS Supplemental Appendices VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-Presenhng Compenson Measurements Of The Robust Laser Inteh AL SOFTWARE.Vahdation Scenanos.

NUREG/CR4316 V07: GUIDELINES FOR THE VERIFICATION AND ometer And Typical Accelerometer Systems.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-VI'3*II'"

AL SOFTWARE. User's Manual.

NUREG/CR4316 V08 GUIDELINES FOR THE VERIFICATION AND NUREG-1525: ASSESSMENT OF THE NRC ENFORCEMENT PRO-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-GRAM.

AL SOFTWARE.Biblography.

Validity LinWt NUREG/CR 6291 V01, NUCLEAR PLANT ANALYZER. installation NUREG/CR 6264 VOI: VALIDITY LIMITS IN J RESISTANCE CURVE Manual.

DETERMINATION.An Assessment Of The JtM) Parameter.

NUREG/CR.6291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer-NUREG/CR4264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE ence Manual.

DETERMINATION.A Computatonal Approach To Ductde Crack Growth NUREG/CR-6291 V03 NUCLEAR PLANT ANALYZER. Computer Visual Under Large-Scale Yielding Conditons.

System Reference Manual.

NUREG/CR-6291 V04 NUCLEAR PLANT ANALYZER Programmer's VaWW Analysis Manual' NUREG-1530. REASSESSMENT OF NRC'S DOLLAR PER PERSON-REM CONVERSION FACTOR POLICY.

Weste Burial NUREG-1307 R05-REPORT ON WASTE BURIAL CHARCES Escalaton yg NUREG-1482: GUIDELINES FOR INSERVICE TESTING AT NUCLEAR Of Decommissoning Waste Disposal Costs At Low-Level Waste Dunal POWER PLANTS.

Facshties Vender inspection Water Infiltration NUREG-0040 Vis N04: LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR-4918 V08-CONTROL OF WATER INFILTRATION INTO SPECTION STATUS REPORT. Quarterty Report, October-December NEAR SURFACF LLW DISPOSAL UNITS. Progress Report Of Field Ex-1994 (White Book) penments At A Humid Regon Sate.Beltsville. Maryland.

86 Subject index Wold Wire Break Detectkm NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING NUREG/CR-6420: SELF-MONITORING SURVEILLANCE SYSTEM FOR WELDS. Seventh Program Report March 1993 - December 1994 PRESTRESSING TENDONS' NUREG/CR4251: STAINLESS STEEL SUBMERGED ARC WELD FUSION LINE TOUGHNESS.

Yucca Mountain W6de-Plate NUREG-1464: NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE NUREG/CR4259: CONSTRAINT EFFECTS ON FRACTURE INITIATION

2. Development Of Capatulitws For Review Of A Performance Assess-LOADS IN HSST WIDE-PLATE TESTS.

ment For A Hgh-level Waste Repository.

r-1 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.

ADVISORY COMMITTEE (S)

EDO OFFICE OF ADMINISTRATION (PRE 870413 & POST 890205)

ADVISORY COMMITTEE ON NUCLEAR WASTE OFFICE OF ADMINISTRATION, DIRECTOR (POST 940714)

NUREG-1423 V05: A COMPILATION OF REPORTS OF THE ADVISO-NUREG-1145 VII: U S. NUCLEAR REGULATORY COMMISSION RY COMMITTEE ON NUCLEAR WASTE. July 1993 June 1995.

1994 ANNUAL REPORT.

ACRS ADVISORY COMMITTEE ON REACTOR SAFEGUARDS DIVISION OF FREEDOM OF ENFORMATION & PUBLICATIONS SERV-NUREG-1125 V16: A COMPILATION OF REPORTS OF THE ADVISO-ICES (POST 940714 RY COMMITTEE ON REACTOR SAFEGUARDS 1994 Annual.

NUREG 0304 V19 NO3: REGULATORY AND TECHNICAL REPORTS ATOMIC SAFETY BOAR S) & PANEL (S f99

^

AT l S pts r

NUREG-0304 V19 N05 REGULATORY AND TECHNICAL REPORTS NU G 63 V06 A I

SAFETY LICENSING BOARD PANEL BIENNIAL REPORT. Fiscal Years 1993 1994.

N R G 304 20 NO TO AD NIC L R PORTS OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

(ABSTRACT INDEX JOURNAL). Compilation For First Quarter REGION 1 (POST 820201) 1995 JanuarpMarch.

NUREG-0837 V14 N04. NRC TLD DIRECT RADIATION MONITORING NUREG-0304 V20 NO2: REGULATORY AND TECHNICAL REPORTS NETWORK. Progress Report. October December 1994.

(ABSTRACT INDEX JOURNAL). Compilaton For Second Quarter NUREG-0837 V15 NO1: NRC TLD DIRECT RADIATION MONITORING 1995.Apni June.

NETWORK Progress Report January March 1995.

NUREG 0540 V16 N11: TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG-0837 V15 NO2: NRC TLD DIRECT RADIATION MONITORING LY AVAILABLE. November 1 30.1994.

NETWORK. Progress Report.Aprd-June 1995.

NUREG-0540 V16 N12: TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG 0837 V15 NO3. NRC TLD DIRECT RADIATION MONITORING LY AVAILABLE. December 1 31,1994.

NETWORK. Progress Report. July-September 1995.

NUREG-0540 V17 Not: TITLE LIST OF DOCUMENTS MADE PUBLIC-REGION 2 (POST 820201)

LY AVAILABLE. January 1 31,1995.

NUREG-1528: RECONSTITUTION OF THE MANUAL CHAPTER 2512 NUREG-0540 V17 NO2: TITLE LIST OF DOCUMENTS MADE PUBLIC-CONSTRUCTION INSPECTION PROGRAM FOR WATTS BAR UNIT LY AVAILABLE. February 1-28,1995.

1.

NUREG 0540 V17 NO3: TITLE LIST OF DOCUMENTS MADE PUBLIC-WALNUT CREEK FIELD OFC. R4 (POST 940404)

LY AVAILABLE. March 1-31,1995.

NUREG-1525: ASSESSMENT OF THE NRC ENFORCEMENT PRO-NUREG-0540 V17 N04: TITLE LIST OF DOCUMENTS MADE PUBLIC-GRAM.

LY AVAILABLE. April 1-30,1995.

OFC OF ENFORCEMENT (POST 8704t3)

NUREG-0540 V17 N05: TITLE LIST OF DOCUMENTS MADE PUBLIC.

NUREG-0940 V13 N4 PI: ENFORCEMENT ACTIONS SIGNIFICANT LY AVAILABLE.May 1 31, 1995.

ACTIONS RESOLVED REACTOR LICENSEES.Ouarterly Progress NUREG-0540 V17 N06: TITLE LIST OF DOCUMENTS MADE PUBLIC-Report October December 1994 LY AVAILABLE. June 1 30,1995.

NUREG-0940 V13 N4 P2: ENFORCEMENT ACTIONS SIGNIFICANT NUREG-0540 V17 N07. TITLE LIST OF DOCUMENTS MADE PUBLIC.

ACTIONS RESOLVED MEDICAL LICENSEES.Ouarterly Progress LY AVAILABLE. July 1-31,1995.

Report, October-December 1994 NUREG-0540 V17 N08: TITLE LIST OF DOCUMENTS MADE PUBLIC-NUREG-0940 V13 N4 P3: ENFORCEMENT ACTIONS SIGNIFICANT LY AVAILABLE. August 1 31,1995.

ACTIONS RESOLVED MATERIAL LICENSEES (NON-NUREG-0540 V17 N09: TITLE LIST OF DOCUMENTS MADE PUBLIC-MEDICAL).Ouarterly Progress Report. October December 1994 LY AVAILABLE. September 1 30.1995.

NUREG-0940 V14 N1 P1: ENFORCEMENT ACTIONS: SIGNIFICANT NUREG-0540 V17 N10: TITLE LIST OF DOCUMENTS MADE PUBLIC-ACTIONS RESOLVED. REACTOR LICENSEESQuarterly Progress LY AVAILABLE. October 1 31,1995.

Report. January-March 1995.

NUREG-0750 V40-NUCLEAR REGULATORY COMMISSION NUREG 0940 V14 N1 P2: ENFORCEMENT ACTIONS SIGNIFICANT ISSUANCESOpinions And Decisions Of The Nuclear Regulatory ACTIONS RESOLVED, MEDICAL LICENSEES Ouarterly Progress Commission With Selected Orders. July-December 1994.

Report. January-March 1995.

NUREG-0750 V40102: INDEXES TO NUCLEAR REGULATORY COM-NUREG 0940 V14 N1 P3: ENFORCEMENT ACTIONS: SIGNIFICANT MISSION ISSUANCES. July-December 1994 ACTIONS

RESOLVED, MATERIAL LICENSEES (NON-NUREG-0750 V40 N05 NUCLEAR REGULATORY COMMISSION IS-MEDICAL) Quarterly Progress Report, January-March 1995.

SUANCES FOR NOVEMBER 1994. Pages 169-318 NUREG-0940 V14 N2 P1 ENFORCEMENT ACTIONS: SIGN!FICANT NUREG-0750 V40 N06. NUCLEAR REGULATORY COMMISSION IS-ACTIONS RESOLVED,1NDIVIDUAL ACTIONS Ouarterly Progress SUANCES FOR DECEMBER 1994. Pages 319-387.

Report. April-June 1995 NUREG-0750 V41101: INDEXES TO NUCLEAR REGULATORY COM.

NURtG-0040 V14 N2 P2: ENFORCEMENT ACTIONS:SIGNIFICANT MISSION ISSUANCES. January-March 1995.

ACTIONS RESOLVED, REACTOR LICENSEES Quarterly Progress NUREG-0750 V41102-. INDEXES TO NUCLEAR REGULATORY COM-Report.Apni-June 1995_

MISSION ISSUANCES. January-June 1995.

NUREG 0940 V14 N2 P3. ENFORCEMENT ACTIONS: SIGNIFICANT NUREG-0750 V41 N01: fdUCLEAR REGULATORY COMMISSION IS-ACTIONS RESOLVED MATERIAL LICENSEES Ouarterly Progress SUANCES FOR JANUARY 1995. Pages 1-69 Report Aptd-June 1995.

NUREG 0750 V41 N02: NUCLEAR REGULATORY COMMISSION IS-NUREG 1525 ASSESSMENT OF THE NRC ENFORCEMENT PRO-SUANCES FOR FEBRUARY 1995. Pages 71178.

GRAM.

NUREG-0750 V41 NO3 NUCLEAR REGULATORY COMMISSION IS-NUREG-1600: GENERAL STATEMENT OF POLICY AND PROCE-SUANCES FOR MARCH 1995 Pages 179-243.

DURE FOR NRC ENFORCEMENT ACTIONS Enforcement Policy.

NUREG-0750 V41 N04: NUCLEAR REGULATORY COMMISSION IS-OFC OF INVESTIGATIONS (POST 880201)

SUANCES FOR APRIL 1995 Pages 245-319 NUREG-1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-NUREG-0750 V41 N05: NUCLEAR REGULATORY COMMISSION IS-TlONS REVIEW TEAM SUANCES FOR MAY 1995 Pages 321-380 OFC OF PERSONNEL (POST 870413)

NUREG-0750 V41 N06. NUCLEAR REGULATORY COMMISSION IS-NUREG-0325 RIB: U.S. NUCLEAR REGULATORY COMMISSION OR-SUANCES FOR JUNE 1995 Pages 381-496 GAN12ATION CHARTS AND FUNCTIONAL STATEMENTS. July 23, NUREG.0750 V42101; lNDEXES TO NUCLEAR REGULATORY COM-1995.

MISSION ISSUANCES. July-September 1995.

87

88 NRC Originating Organization index (Staff Reports)

NUREG4750 V42 N01: NUCLEAR REGULATORY COMMISSION IS.

DIVISION OF FUEL CYCLE SAFETY & SA EGUARDS (POST 930207)

SUANCES FOR JULY 1995 Paoes 1-45.

NUREG 1065 R02: ACCEPTABLE STANDARD FORMAT AND CON-NUREG-0750 V42 N02: NUCLEAR REGULATORY COMMISSION IS-TENT FOR THE FUNDAMENTAL NUCLEAR MATERIAL CONTROL SUANCES FOR AUGUST 1995. Pages 47 97.

(FNMC) PLAN REOUIRED FOR LOW ENRICHED URANIUM FACILi-NUREG-0750 V42 NO3 NUCLEAR NEGULATORY COMMISSION IS-TIES.

SUANCES FOR SEPTEMBER 1995. Pages99-110.

NUREG-1280 ROI: STANDARD FORMAT AND CONTENT ACCEPT.

NUREG-0750 V42 N04. NUCLEAR REGULATORY COMMISSION IS-ANCE CRITERIA FOR THE MATERIAL CONTROL AND ACCOUNT.

SUANCES FOR OCTOBER 1995. Pages 111 180-ING (MC&A) REFORM AMENDMENT.

NUREG-0936 V13 NO3: NRC REGULATORY AGENDA. Semiannual OPERATIONS BRANCH Report, July-December 1994.

NUREG-0525 V02 R03: SAFEGUARDS

SUMMARY

EVENT LIST NUREG-0936 V14 NO1: NRC REGULATORY AGENDA Semiannual (SSEL). January 1,1990 Through December 31,1994.

Report, January June 1995-NUREG/CR-6125 V03: HUMAN FACTORS EVALUATON OF

^

^

b"EP *"U ^""

• EDO OFFICE OF THE CONTROLLER (PRE 020410 & POST 890205)

OFFICE OF THE CONTROLLER (POST 890205)

Of Human-System Interfaces Procedures And Practces,Trairung And NUREG-1470 V04: FINANCIAL STATEMENT FOR FISCAL YEAR Organtratonal Practces And Procedures.

9994, NUREG/CR-6277 V01: HUMAN FACTORS EVALUATON OF TELE-DIVISION OF BUDGET & ANALYSIS (POST 890205)

THERAPY. Identification Of Problems And Alternative Approaches.

NUREG 1100 V11: BUDGET ESTIMATES Fiscal Years 1996-1997.

NUREG/CR4277 V02: HUMAN FACT C EVALUATION OF NUREG-1350 V07: NUCLEAR REGULATORY COMMISSION INFOR-TELETHERAPY. Function And Task Analysis MATION DIGEST.1995 Edition.

NUREG/CH4277 V03: HUMAN FACTCHS EVALUATION OF TELE.

THERAPY. Human-System interfaces And Procedures.

EDO. 0FFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR-6277 V04: HUMAN FACTORS EVALUATION OF DATA TELETHERAPY.Tra> rung And Orgaruzational Analyses.

OFFICE FOR ANALYSIS & EVALUATON OF OPERATIONAL DATA, Dl-NUREG/CR-6277 V05: HUMAN FACTORS EVALUATON OF RECTOR TELETHERAPY.Uterature Review.

NUREG-0090 V17 NO3: REPORT TO CONGRESS ON ABNORMAL DIVISION OF WASTE MANAGEMENT (NMSS 940403)

OCCURRENCES. July-September 1994.

NUREG-1323 ROI: UCENSE APPUCATION REVIEW PLAN FOR A NUREG4090 V17 N04: REPORT TO CONGRESS ON ABNORMAL GEOLOGIC REPOSITORY FOR SPENT NUCLEAR FUEL AND OCCURRENCES October-December 1994.

HIGH-LEVEL RADIOACTIVE WASTE. Draft Rewsw Plan.

NUREG-0090 V18 N01: REPORT TO CONGRESS ON ABNORMAL NUREG 1444 S01: SITE DECOMMISSIONING MANAGEMENT PLAN.

OCCURRENCES. January March 1995.

NUREG 1272 V08 NO2: OFFICE FOR ANALYSIS AND EVALUATON U.S. NUCLEAR REGULATORY twaanseemm OF OPERATIONAL DATA 1993 Annual Report - Nuclear Matenais-OFFICE OF THE GENERAL COUNSEL (POST 860701)

NUREG-1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR NUREG-0980 VD1 NO3:

NUCLEAR REGULATORY A NUCLEAR POWER PLANT-LEGISLATON.103D Congress.

TECHNICAL TRAINING CENTER NUREG-0980 V02 NO3: NUCLEAR REGULATORY LEGISLATION.103d NUREG-1527; NRC'S OBJECT-ORIENTED SIMULATOR INSTRUC-Congress.

TOR STATION-OFFICE OF THE INSPECTOR GENERAL (POST 890417)

DIVISION OF SAFETY PROGRAMS (POST 870413)

NUREG-14 t$ V07 NO2: OFFICE OF THE INSPECTOR NUREG 1275 VII: OPERATING EXPERIENCE FEEDBACK REPORT -

GENERALSemaannual Report To Congress, October 1,1994 -

TURBINE-GENERATOR OVERSPEED PROTECTION March 31,1995 NURE PR IN OF THE WORKSHOP ON GATE

• • • M I'

~

• VALVE PRESSURE LOCKING AND THERMAL BINDING.

ber 30,1995.

EDO OFFICE INFORMATION RESOURCES MANAGEMENT & ARM NRC ET ILED F ON EN O FICE OF INFORMATION RESOURCES MANAGEMENT (POST UNCERTAINTY ANALYSIS. Dapersson and Depoestion Uncertamty 890205)

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

NUREG 0020 V19: JCENSED OPERATING REACTORS STATUS

SUMMARY

REPORT' Data As C1 December 31,1994.(Gray Book 1)

EDO WM W m EULAM RESEARM (W SN)

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 941217)

EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG-1266 V09 NRC SAFETY RESEARCH IN SUPPORT OF REG-OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS ULATION - FY 1994.

NUREG-0383 V01 R18: DIRECTORY OF CERTIFICATES OF COMPLp NUREG-1464: NRC ITERATIVE PERFORMANCE ASSESSMENT ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC PHASE 2.Deve60pment Of m; ^ ~ : For Renew Of A Performance Approved Packages.

Assessment For A High-Level Weste Repoestory.

NUREG-0383 V02 RIB: DIRECTORY OF CERTIFICATES OF COMPLk DIVISON OF ENGINEERING TECHNOLOGY (POST 941217)

ANCE FOR RADIOACTIVE MATERIALS PACKAGES.Certifcates Of NUREG 0933 S18: A PRORITIZATION OF GENERIC SAFETY Compliance.

ISSUES.

NUREG-0383 V03 R15: DIREFORY OF CERTIFICATES OF COMPLl-NUREG 0933 S19: A PRORITIZATON OF GENERIC SAFETY ANCE FOR RADOACTIV2 bATERIALS PACKAGES. Report Of NRC ISSUES.

Approved Quality Assurars:e Programs for Radioactive Matenals DIVISON OF REGULATORY APPUCATONS (POST 941217)

Packa NUREG0713 V15: OCCUPATIONAL RADIATON EXPOSURE AT NUREG 30 V14: LICENSED FUEL FACluTY STATUS COMMERCIAL NUCLEAR POWER REACTORS AND OTHER REPORT. inventory Difference Data. July 1,

1993 - June 30, FACILITIES.1993 Twenty-Sixth Annual Report.

1994 (Gray Book it)

NUREG-1307 ROS:

REPORT ON WASTE BURIAL NUROG 1464: NRC ITERATIVE PERFORMANCE ASSESSMENT CHARGES Escalation Of Decomrrusseorung Waste Dsposal Costs At PHASE 2. Development Of Capabilities For Renew Of A Performance Low-Level Waste Bunal Facilities.

Assessment For A High-Level Waste Reposit.

NUREG-1493: PERFORMANCE BASED CONTAINMENT LEAK-TEST DIVISION OF INDUSTRIAL & MEDICAL NUC AR SAFETY (POST PROGRAM.Flnal Report.

870729)

NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK.

NUREG 0725 RIO: PUBUC INFORMATION CIRCULAR FOR bHIP-TEST PROGRAM Draft Reoort For Comment.

MENTS OF IRRADIATED REACTOR FUEL NUREG-1505 DRFT FC: A NONPARAMETRIC STATISTICAL METH-NUREG-1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERI-ODOLOGY FOR THE DESIGN AND ANALYSIS OF FINAL STATUS AL SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For DECOMMISSIONING SURVEYS. Draft Report For Comment.

Comment.

NUREG-1506 DRFT FC: MEASUREMENT METFODS FOR RADIO-OPERATONS BRANCH LOGICAL SURVEYS IN SUPPORT OF NEW DECOMMISSIONING NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF CRITERIA. Draft Report For Comment.

REMOTE AFTERLOADING BRACHYTHERAPY. Human Error And NUREG-1507 DRFT FC: MINIMUM DETECTABLE CONCENTRATONS Critical Tasks in Remote Afterloading Brachytherapy And Approach-WITH TYPICAL RADIATON SURVEY INSTRUMENTS FOR VARI-es For improved System Performance.

OUS CONTAMINANTS AND FIELD CONDITIONS. Draft Report For NUREG/CR-6125 V02: HUMAN FACTORS EVALUATION OF Comment.

REMOTE AFTERLOADING BRACHYTHERAPY. Function And Task NUREG-1530 REASSESSMENT OF NRC'S DOLLAR PER PERSON-Analysis.

REM CONVERSION FACTOR POUCY, l

l l

NRC Originating Organization index (Staff Reports) 89 I

f NUREG/CP-0147: PROCEEDINGS OF THE WORKSHOP ON THE NUREG4040 V19 NO3: UCENSEE CONTRACTOR AND VENDOR IN-ROLE OF NATURAL ANALOGS IN GEOLOGIC DISPOSAL OF SPECTION STATUS REPORT. Quarterly Report, July-September HIGH-LEVEL NUCLEAR WASTE. Held in San Antonio. Texas, July 22-1995.(White Book) 25,1991.

NUREG 0390 V09 N01: TOPICAL REPORT REVIEW STATUS.

WASTE MANAGEMENT BRANCH (POST 941217)

NUREG-0847 S16: SAFETY EVALUATON REPORT RELATED TO NUREG/CR-4916 V08: CONTROL OF WATER INFILTRATON INTO THE OPERATON OF WATTS BAR NUCLEAR PLANT, UNITS 1 NEAR SURFACE LLW DISPOSAL UNITS Progress Report Of Field AND 2. Docket Nos. 50-390 And 50-391.(Tennessee Valley Authority)

Expenments At A Humid Region Site,Bettsville, Maryland.

NUREG-0847 S17: SAFETY EVALUATION REPORT RE(.ATED TO NUREG/CR4283: FIELD SITE INVESTIGATION: EFFECT OF MINE THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND

2. Docket Nos. 50-390 And 50-391.(Tennessee Valley Authonty}o TO SEISMICITY ON GROUNDWATER HYDROLOGY.

NUREG 0847 S18: SAFETY EVALUATION REPORT RELATt DMSION OF SYSTEMS TECHNOLOGY (POST 941217)

NUREG-0700 ROI DFC: HUMAN-SYSTEM INTERFACE DESIGN THE OPERATON OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND REVIFW GUIDELINE. Draft Report For Comment.

2. Docket Nos. 50-390 And 50 391.(Tennessee Valley Authonty)

NUREG-1465: ACCIDENT SOURCE TERMS FOR UGHT WATER NU-NUREG-0847 S19. SAFETY EVALUATION REPORT RELATED TO l

CLEAR POWER PLANTS.

THE OPERATON OF WATTS BAR NUCLEAR PLANT,UNITG 1 AND

2. Docket Nos. 50-390 And 50-391.(Tennessee Valley Authon_ty) FOR l

NUREG 1519: SURFACE INTERACTIONS OF CESIUM AND BORIC NUREG-1122 RO1: KNOWLEDGE AND ABILITIES CATALW ACID WITH STAINLESS STEEL CONTROL INSTRUMENTATION & HUMAN FACTORS BRANCH (POST NUCLEAR POWER PLANT OPERATORS: PRESSURIZED WATER REACTORS.

941217)_

NUREG-1123 R01: KNOWLEDGE AND ABluTIES CATALOG FOR NUREG/GR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE AFTERLOADING BRACHYTHERAPY. Human Error And NUCLEAR POWER PLANT OPERATORS: BOILING WATER REAC-Cntical Tasks in Remote Afterloading Brachytherapy And Approach-

~

STANDARD TECHNICAL SPECIFICATIONS NUI E -1430 V01 R01:

es For improved System Performance.

BABCOCK AND WILCOX PLANTS Specifications' SPECIFICATIONS NUREG/CR.6125 V02: HUMAN FACTORS EVALUATION OF NUREG 1430 V02 RO1: STANDARD TECHNICAL REleOTE AFTERLOADING BRACHYTHERAPY. Function And Task BABCOCK AND WILCOX PLANTS. Bases (Sections 2.0 - 3.3).

Anah sis.

NUREG-1430 V03 R01: STANDARD TECHNICAL SPECIFICATONS NUREG/CR4125 V03. HUMAN FACTORS EVALUATION OF BABCOCK AND WILCOX PLANTS. Bases (Sections 3 4 - 3.9).

[

REMOTE AFTERLOADING BRACHYTHERAPY. Supporting Analyses NUREG-1431 V01 RO1: STANDARD TECHNICAL SPECIFICATIONS Of Human-System interfaces. Procedures And Practices,1ranng And WESTINGHOUSE PLANTS. Specificatsons.

Orgaruzational Practices And Procedures.

NUREG-1431 V02 ROI: STANDARD TECHNICAL SPECIFICATIONS l

NUREG/CR-6159: USING MICRO SAINT TO PREDICT PERFORM-WESTINGHOUSE PLANTS. Bases (Sections 2.0 3.3) CIFICA l

ANCE IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of NUREG-1431 V03 RO1: STANDARD TECHNICAL SPE Validity And Feasitality.

WESTINGHOUSE PLANTS. Bases (Sections 3.4 - 3.9)ECIF NUREG-1432 V01 RO1: STANDARD TECHNICAL SP NUREG/CR4277 V01: HUMAN FACTORS EVALUATON OF TELE-THERAPY. Identification Of Problems And Altemative Approaches.

COMBUSTION ENGINEERING PLANTS Specifications.

NUREG/CR-6277 V02: HUMAN FACTORS EVALUATION OF NUREG-1432 V02 801: STANDARD TECHNICAL SPECIFICATONS t

TELETHERAPY. Function And Task Analysa COMBUSTION ENGINEERING PLANTS Bases (Sections 2 0 - 3.3).

NUREG/CR4277 V03: HUMAN FACTORS EVALUATION OF TELE.

NUREG-1432 V03 RO1: STANDARD TECHNICAL SPECIFICATIONS THERAPY. Human-System interfaces And Procedures.

COMBUSTION ENGINEERING PLANTS. Bases (Sections 3.4 3.9).

NUREG/CR4277 V04: HUMAN FACTORS EVALUATION OF NUREG-1433 V01 RO1: STANDARD TECHNICAL SPECIFICATIONS TELETHERAPY. Training And Organizational Analysis GENERAL ELECTRIC PLANTS BWR/4. Specifications.

NUREG 1433 V02 RO1: STANDARD TECHNICAL SPECIFICATIONS NUREG/CR-6277 V05: HUMAN FACTORS EVALUATION OF TELETHERAPY.Uterature Review GENERAL ELECTRIC PLANTS BWR/4. Bases (Sections 2.0 3.3).

NUREG-1433 V03 R01: STANDARD TECHNICAL SPECIFICATIONS PROBABluSTIC RISK ANALYSIS BRANCH (POST 941217)

NUREG/CR-6244 V01: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG 4 R :

A A I

I IC T ON UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty GENERAL ELECTRIC PLANTS. BWR/6. Specifications.

N REM 434 W2 M SMDARD TECHNM SNCMS NU V02 OBABILISTIC ACCIDENT CONSEQUENCE GENERAL ELECTRIC PLANTS, BWR/6 Bases (Sections 2.0 3.3).

UNCERTAINTY ANALYSIS. Dispersion And Deposttion Uncertainty NUREG-1434 V03 R01: STANDARD TECHNICAL SPECIFICATIONS' Assessment. Appendices A And B.

GENERAL ELECTRIC PLANTS, BWR/6 Bases 3.4 - 3.10).

j NUREG/CR4265: MULTIDISCIPUNARY FRAMEWORK FOR HUMAN NUREG 1435 SO4: STATUS OF SAFETY ISSUES AT UCENSED RELIABluTY ANALYSIS WITH AN APPUCATION TO ERRORS OF POWER PLANTS.TMI Action Plan Requerements, Unresolved Safety COMMISSION AND DEPENDENCIES.

Issues.Genenc Sa_fety issues.Other Multiplant Action issues.

REACTOR & PLANT SYSTEMS BRANCH (POST 941217)

NUREG 1478 RO1: NUN-POWER REACTOR OPERATOR UCENSING NUREG/CR 6257. CANDU 3 TRANSIENT ANALYSIS USING ATOMIC EXAMINER STANDARDS.

ENERGY OF CANADA LTD CODES.

NUREG 1517: REPORT OF THE SOUTH TEXAS PROJECT ALLEGA-INTRA-AGENCY COMMITTEES, REVIEW GROUPS, ETC.

NUl 1 2 SE ENT OF INSERVICE CONDITIONS OF INCIDENT INVESTIGATION TEAM SAFETY RELATED NUCLEAR PLANT STRUCTURES.

NUREG 1535: INGESTON OF PHOSPHORUS-32 AT MASSACHU-NUREG 1526: LESSONS LEARNED FROM EARLY IMPLEMENTA-SETTS INSTITUTE OF TON OF THE MAINTENANCE RULE AT NINE NUCLEAR POWER TECHNOLOGY, CAMBRIDGE,M ASSACHUSETTS,lDENTIFIE D ON PLANTS.

AUGUST 19,1995.

DIVISION OF REACTOR PROJECTS 1/11 (DRPE) POST 941001)

NUREG 0847 S15: SAFETY EVALUATION REPORT RELATED TO EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 000428)

THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001) 2 Docket Nos. 50-390 And 50 391fiennessee Valley Authonty)

NUREG-0040 V18 N04: UCENSEE CONTRACTOR AND VENDOR IN-ASSOCIATE DIRECTOR FOR ADVANCED REACTORS & LICENSE RE-SPECTON STATUS REPORT. Quarterly Report,0ctober December NEWAL (ADAR) (POST 1994.(White Books NUREG 0498 S01: FINAL ENVIRONMENTAL STATEMENT RELATED NUREG 0040 V19 N01: LICENSEE CONTRACTOR AND VENDOR IN.

TO THE OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 SPECTION STATUS REPORT. Quarterty Report. January-March AND 2. Docket Nos. 50-390 And 50-391TTennessee Valley Authonty) 1995.(White Book)

OFFICE OF NUCLEAR REACTOR REGULATION, DIRECTOR (POST NUREG4040 V19 NO2: UCENSEE CONTRACTOR AND VENDOR IN.

s70411)

SPECTION STATUS REPORT. Quarterty Report. April-June NUREG 1482: GUIDELINES FOR INSERVICE TESTING AT NUCLEAR 1995.(White Boolt)

POWER PLANTS.

l l

l

NRC Originating Organization index (International Agreements)

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

There were no NUREG/IA reports during this year.

l 91

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9

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d D

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 2 ajor NRC organization (e.g., program office) cnd then by subsections of these (e.g., divisind where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and utie of the report (s) prepared by that organi-zation. If further information is needed, refer to the main citation by the NUREG/CR number.

l l

EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR4259: CONSTRAINT EFFECTS ON FRACTURE INITI-DATA ATION LOADS IN HSST WIDE-PLATE TESTS.

DIVISION OF SAFETY Pf OGRAMS (POST 870413)

NUREG/CR-6331: ATMOSPHERIC RELATIVE CONCENTRATIONS IN NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE BUILDING WAKES.

CORE DAMAGE ACCIDENTS:1994 A STATUS REPORT. Main OlVISION OF ENGINEERING TECHNOLOGY (POST 941217)

Report And Appendices A-H NUREG/CR-4219 V10 N2: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE PROGRAM. Semiannual Progress Report For Aprd-September 1993.

I CORE DAMAGE ACCIDENTS:1994 A STATUS REPORT. Appendix 1.

NUREG/CR 4219 VII N1: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR4266. ANALYSIS OF BORON DILUTION IN A FOUR-PROGRAM. Semiannual Progress Report For October 1993. March i

LOOP PWR.

1994.

NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS WELDS. Seventh Program Report March 1993 December 1994.

OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING NUREG/CR4328: ADEOUACY OF THE 123-GROUP CROSS-SEC-IN LIGHT WATER REACTORS. Semiannual Report,0ctober 1993 TION LIBRARY FOR CRITICALITY ANALYSES OF WATER-MODER-March 1994.

DIVISION STR

& MEDICAL NUCLEAR SAFETY (POST l

870729) gg NUREG/CR 0200 VI R04: SCALE: A MODULAR CODE SYSTEM FOR NUAEG/CR-5462; AGING STUDY OF BOluNG WATER REACTOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-HIGH PRESSURE INJECTION SYSTEMS.

CENSING EVALUATION Control Modules-NUREG/CR-5591 V03: HEAVY SECTION STEEL IRRADIATION i

NUREG/CR 0200 V2PIR4. SCALE: A MODULAR CODE SYSTEM PROGRAM. Progress Report For October 1991. September 1992.

l FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR NUREG/CR 5591 V04 N2: HEAVY SECTION STEEL IRRADIATION LICENSING EVALUATION Functional Modules F1 F8.

PROGRAM. Semiannual Progress Report For April-September 1993.

NUREG/CR-0200 V2P2R4. SCALE: A MODULAR CODE SYSTEM NUREG/CR.5591 VOS N1: HEAVY SECTION STEEL lRRADIATION FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR PROGRAM. Semiannual Progress Report For September 1993 i

LICENSING EVALUATION. Functional Modules F9-F16.

Through March 1994 l

NUREG/CR 0200 V3 R04: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR-5591 V05' N2: HEAVY-SECTION STEEL IRRADIATION l

PERFORMING STANDARDlZED COMPUTER ANALYSES FOR Li-PROGRAM. Progress Report For April 1994 Through September CENSING EVALUATION. Miscellaneous.

ggg4 NUREG/CR-5657: AUTOCASK (AUTOMATIC GENERATION OF 3-D NUREG/CR-5591 VO6 N1: HEAVY-SECTION STEEL IRRADIATION CASK MODELS).A Microcomputer Based System For Shipping Cask PROGRAM. Semiannual Progress Report For October 1994 Through Design Review Analysis-March 1995 NUREG/CR4074 V04. SEALED SOURCE AND DEVICE DESIGN NUREG/CR-5857: AGING OF TURBINE DRIVES FOR SAFETY-RE-SAFETY TESTING. Technical Report On The Findings Of Task 4.In-LATED PUMPS IN NUCLEAR POWER PLANTS.

i vestigation Of Sealed Source for Paper Mill Digester-NUREG/CR-5944 V02. A CHARACTERIZATION OF CHECK VALVE NUREG/CR4074 V05: SEALED SOURCE AND DEVICE DESIGN DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLEAR SAFETY TESTING. Technical Report On The Findings Of Task 4.in-POWER INDUSTRY.1991 Failures.

vestigation Of Failed Radioactive Stainless Steel Troxler Gauges.

NUREG/CR-5954: EFFECT ON AGING ON PWR CHEMICAL AND NUREG/CR4276: A COMPILATION OF CURRENT REGULATIONS, VOLUME CONTROL SYSTEM.

STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING NUREG/CR-6004: PROBABILISTIC PIPE FRACTURE EVALUATIONS BRACHYTHERAPY.

FOR LEAK-RATE-DETECTION APPLICATIONS.

NUREG/CR4322: BUCKLING ANALYSIS OF SPENT FUEL BASKET.

NUREG/CR4016: AGING AND SERVICE WEAR OF AIR-OPERATED NUREG/CR-6323: RELATIVE RISK ANALYSIS IN REGULATING THE VALVES USED IN SAFETY-RELATED SYSTEMS AT NUCLEAR USE OF RADIATION-EMITTING MEDICAL DEVICES.A Preliminary POWER PLANTS.

NU G 4324:OUAllTY ASSURANCE FOR GAMMA KNIVES.

T IN ET NUREG/CR4330: RESULTS OF REGULATORY IMPACT SURVEY OF NUREG/CR-6089: DETECTION OF PUMP DEGRADATION.

t I

INDUSTRIAL AND MEDICAL MATERIALS UCENSEES OF THE NUREG/CR 6100: GATE VALVE AND MOTOR OPERATOR RE-OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS-SEARCH FINDINGS DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207)

NUREG/CR-6154 V02': EXPERIMENTAL RESULTS FROM CONTAIN.

NUREG/CR4287: MANAGEMENT CONCEPTS AND SAFETY APPLi-MENT PIPING BELLOWS SUBJECTED TO SEVERE ACCIDENT CATIONS FOR NUCLEAR FUEL FACILITIES.

CONDITIONS.Results From Bellows Tested in Corroded Conditions.

NUREG/CR4173: A

SUMMARY

OF THE FIRE TESTING PROGRAM EG/CR 2 C Ti LI Y SAF CRI E lA FOR LICENSE REVIEW OF LOW-LEVEL WASTE FACILITIES.

NUREG/

84 EPARA FE S TESTING AND ANALYSES TO INVESTIGATE LINER TEARING OF THE 1:16-SCALE REIN-EDO - OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

DIVISION OF SYSTEMS RESEARCH (880717 941217)

FORCED CONCRETE CONTAINMENT DUILDING.

NUREG/CR4141: HANDBOOK OF METHODS FOR RISK-BASED NUREG/CR-6191: SIZE AND DEFORMATION LIMITS TO MAINTAIN 1

ANALYSES OF TECHNICAL SPECIFICATIONS.

CONSTRAINT IN K(IC) AND J(C) TESTING OF BEND SPECIMENS.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 941217)

NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-NUREG/CR4116 V09: SYSTEMS ANALYSIS PROGRAMS FOR LOADED PRESSURE RELIEF VALVES USED IN SAFETY-RELATED HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

SYSTEMS AT NUCLEAR POWER PLANTS.

I VERSION 5 0 Venfication And Vahdation (V&V) Manual NUREG/CR-6214: PRODUCTION AND TESTING OF THE VITAMIN-B6 NUREG/CR4224: PARAMETRIC STUDY OF THE POTENTIAL FOR FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/B-DEBRIS.

VI NUCLEAR DATA.

i 93

94 NRC Contract Sponsor Index NUREG/CR4220: AN ASSESSMENT OF FIRE VULNERABILITY FOR NUREG/CR 5927 V02: EVALUATION OF A PERFORMANCE AS-AGED ELECTRICAL RELAYS.

SESSMENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE l

NUREG/CR4235: ASSESSMENT OF SHORT THROUGH-WALL CIR-WASTE DISPOSAL FACILITIES.Vahdaten Needs.

C1)MFERENTIAL CRACKS IN PIPES Expenments And NUREG/CR4054: ESTIMATING PRESSURtZED WATER REACTOR Analyses. March 1990 December 1994.

DECOMMISSIONING COSTS. A User's Manual For The PWR Cost NUREG/CR4239 VO1: SURVEY OF STHONG MOTION EARTH-Eshmaten0 Computer Program (CECP) Software.

QUAKE EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA NUREG/CR4112-tMPACT OF REDUCED DOSE UMITS ON NRC Ll-WITH EMPHASIS ON PtPING SYSTEMS Masn Report.

CENSED ACTIVITIES. Major issues in The implementaton Of ICRP/

1 NUREG/CR4239 V02: SURVEY OF STRONG MOTON EARTH-OUAKE EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA NCRP Dose LW Recommendehons.F6nal Report NUREG/CR4114 V02: AUXILIARY ANALYSES IN SUPPORT OF PER-A AH H

VEL NU

/

64 A I

F N

SPECTRA TO SEIS-MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF WASTE FACILITY.Two-Phase flow And Contamenent Transport in ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO Unsaturated Soils With Apphcaton To Low-Level Radcactive Waste NUR 4188 V02: MICROBIAL DEGRADATION OF LOW-LEVEL NU EG 4

K TER ANALYSIS OF STORAGE CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR NURE C I LD LY TE INV GATONS = TEST STORAGE CASK DESIGN REVIEW. User's Manual To Versson Ib RESULTS. Low-Level Waste Data Base Program. Test Results For N RE C4 S

L STEEL SUBMERGED ARC WELD Fecal Years 1986,1987,1988, And 1989.

FUSION UNE TOUGHNESS NUREG/CR4256 V02: FIELO LYSIMETER INVESTIGATIONS TEST NUREG/CR4264 V01: VALIDITY LIMITS IN J. RESISTANCE CURVE RESULTS Low-Level Waste Data Base Development Program: Test DETERMINATION.An Assessment Of The J(M) Parameter.

Results For Focal Years 1990,1991,1992, And 1993.

NUREG/CR4264 V02: VALIDtTY LIMITS IN J-RESISTANCE CURVE NUREG/CR4283: FIELD SITE INVESTIGATION: EFFECT OF MINE DETERMINATION.A Computahonal Approach To Ductile Crack SEISMICITY ON GROUNDWATER HYDAOLOGY.

Growth Under Large Scale Yielding Conditons.

NUREG/CR4305: BLT-EC (BREACH, LEACH, TRANSPORT, AND NUREG/CR4273: BIAXlAL LOADING EFFECTS ON FRACTURE EQUILIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR AS-TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL SES$ LNG THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL NUREG/CR4275: MECHANICAL PROPERTIES OF THERMALLY WASTE DISPOSAL UNITS Background Theory, And Model Desenp-AGED CAST STAINLESS STEELS FROM SHIPPINGPORT REAC.

ton.

TOR COMPONENTS.

NUREG/CR4307:

SUMMARY

OF COMMENTS RECEIVED AT WORK-NUREG/CR-6297: FRACTURE EVALUATIONS OF FUSION LINE SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB)

CRACKS IN NUCLEAR PIPE BIMETALUC WELDS.

TO FACILITATE PUBLIC PARTICIPATION IN DECOMMISSIONING NUREG/CR4298: FRACTURE BEHAVIOR OF SHORT CIRCUMFER-CASES.

ENTIALLY SURFACE-CRACKED PIPE.

NUREG/CR4333: BREATH VERSION 1.1, COUPLED FLOW AND NUREG/CR4299: EFFECTS OF TOUGHNESS ANISOTROPY AND COMBINED TENSION, TORSION, AND BENDING LOADS ON ENERGY TRANSPORT IN POROUS MEDIA.Sunulator Desenption And User Guide.

FRACTURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK-NUREG/CR-6334: NEW SENSOR FOR MEASUREMENT OF LOW AIR OPEN4NG-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-FLOW VELOCITY. Phase i Final Report.

WALL CRACKS IN PLPES NUREG/CR4347: MULTI-PHASE REACTIVE TRANSPORT THEORY, NUREG/CR4348: THERMALLY DRIVEN MOISTURE REDISTRIBU-NUREG/CR4313 V01: ROBUST, ACCURATE AND NON-CONTACT-ING VIBRATION MEASUREMENT SYSTEM. Summary of Companson TION IN PARTIALLY SATURATED POROUS MEDIA NUREG/CR 6351: REVIEW OF SCENARO SELECTION APPROACH-Measurements Of The Robust li:,ser interferometer And Typical Ac-ES FOR PERFORMANCE ASSESSMENT OF HIGH LEVEL WASTE Nt$ C 631 V ROBUST, ACCURATE, AND NON-CONTACT-NURE / R FC E RMANCE TESTING OF ELEC-ING VIBRATION MEASUREMENT SYSTEMS Supplemental Appendi-ces Presenting Compenson Measurements Of The Robust Laser in.

TRONIC PERSONAL DOSIMETERS. Draft Report For Comment NUREG/CR4356: HYDRAULIC CHARACTERIZATION OF HYDROTH-ENIT E

T RECOVERY DUE ERMALLY ALTERED NOPAL TUFF.

N EG S

TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

EN6M RADOMM CHAMCTERMON & SN NUREG/CR4335: FATIGUE STRAIN-LIFE BEHAVIOR OF CARBON CONTROL ROD ASSEMBLIES.

O'V AND LOW ALLOY STEELS, AUSTENITIC STAINLESS STEELS, N EG/ R 55 V01:

LAPS / OD CO AN ALCode Struc-AND ALLOY 600 IN LWR ENVIRONMENTS NUREG/CR4358 V01: ASSESSMENT OF UNITED STATES INDUS-N G/

535 A LA / OD3 C MANUALUser's Guide TRY STRUCTURAL CODES AND STANDARDS FOR APPLICATION NUR RELAP5/ MOD 3 CODE MANUALModels And N E 5 0 ESS T

U TE A

INDUS-TRY STRUCTURAL CODES AND STANDARDS FOR APPLICATION b35V05R1: RELAPS/ MOD 3 CODE MANUALUser's NU

/

TO ADVANCED NUCLEAR POWER REACTORS. Appendices.

NUREG/CR-6368: EXPERIMENTAL INVESTIGATION OF SEDIMEN-Guidehne.

A AG E TED FlBROUS DEBRIS AND SLUDGE NUREG/CR 2

R SK-BASED MA NTE NCE NUREG/CR4420: SELF-MONITORING SURVEILLANCE SYSTEM caton Of Maintenance Effectnreness.

FOR PRESTRESSING TENDONS.

NUREG/CR-6046: ALERTNESS, PERFORMANCE, AND OFF-DUTY DlVISION OF REGULATORY APPLICATIONS (POST 941217)

SLEEP ON 8-HOUR AND 12 HOUR NIGHT SHIFTS IN A SIMULAT-NUREG/CR-3469 VO6: OCCUPATIONAL DOSE REDUCTION AT NU-ED CONTINUOUS OPERATIONS CONTROL ROOM SETTING.

CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SE.

NUREG/CR-6109: THE PROBABILITY OF CONTAINMENT FAILURE LECTED READINGS IN RADIATION PROTECTION AND ALARA.

BY DIRECT CONTAINMENT HEATING IN SURRY.

NUREG/CR-4918 V08: CONTROL OF WATER INFILTRATION INTO NUREG/CR4116 V06: SYSTEMS ANALYSIS PROGRAMS FOR NEAR SURFACE LLW DISPOSAL UNITS, Progress Report Of Field HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

Eugenments At A Humid Regen Site.Beltsville, Maryland.

VERSION 5.0 Graphical Evaluaton Module (GEM) Reference NUREG/CR 5229 V07: FIELu LYSIMETER INVESTIGATIONS: LOW-Manual.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR-6116 V10: SYSTEMS ANALYSIS PROGRAMS FOR FISCAL YEAR 1994 Annual Report.

HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

NUREG/CR 5884 V01: REVISED ANALYSIS OF DECOMMISSIONING VERSION 5.0. Data Loading Manual.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER NUREG/CR-6119 V01:

MELCOR COMPUTER CODE STATION Effects Of Current Regulatory And Other ConsKleratons MANUALS.Pnmer And User's Guides. Version 1.8.3 Septemter On The Financial Insurance Requirements Of The Decommessoning 1994.

Rule And..

NUREG/CR-6119 V02:

MELCOR COMPUTER CODE NUREG/CR 5884 V02: REVISED ANALYSIS OF DECOMMISSIONING MANUALS. Reference Manuals.Vernon 1.8.3 September 1994.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER NUREG/CR-6125 V01: HUMAN FACTORS EVALUATION OF STATION Effects Of Current Regulatory And Other Considerations REMOTE AFTERLOADING BRACHYTHERAPY. Human Error And On The Financial insurance Requirements Of The Decommissoning Cntical Tasks in Remote Aftertoeding Brachytherapy And Approach-Rule And..

es For improved System Performance.

NRC Contract Sponsor index 95 NUREG/CR4125 V02: HUMAN FACTORS EVALUATION OF NUREG/CR-6277 V04: HUMAN FACTORS EVALUATON OF REMOTE AFTERLOADING BRACHYTHERAPY. Function And Task TELETHERAPY. Training And Organizational Anatyses.

Analysis.

NUREG/CR4277 V05: HUMAN FACTORS EVALUATION OF NUREG/CR-6125 V03: HUMAN FACTORS EVALUATON OF TELETHERAPY.uterature Review.

REMOTE AFTERLOADING BRACHYTHERAPY. Supporting Analyses NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATON Of Human-Svstem Interfaces.Procedules And Practices.Trairwn0 And STUDIES AT THE INEL Orgaruzational Practices And Procedures.

NUREG/CR4291 V01: NUCLEAR PLANT ANALYZER installation NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF Manual.

CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR NUREG/CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer-ACCIDENT CONSEQUENCE MODEL ence Manual.

NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR4291 V03: NUCLEAR PLANT ANALYZER. Computer Visual EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-System Reference Manual.

DENT CONSEQUENCE MODEL NUREG/CR4291 V04: NUCLEAR PLANT ANALYZER. Programmer's NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF Manual FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CR-6293 V01: VERIFICATON AND VAUDATION GU!DE-DENT CONSEQUENCE MODEL LINES FOR HIGH INTEGRITY SYSTEMS. Main Report NUREG/CR4143 V01: EVALUATON OF POTENTIAL SEVERE ACCl-NUREG/CR4293 V02: VERIFICATIONN AND VAUDATION GUIDE-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LIN2S FOR HIGH INTEGRITY SYSTEMS. Appendices A-D.

A ENN NUREG/

V P ALUA i

)F POTENTIAL SEVERE S O HE GENERAL PUBLIC IN THE EVENT OF A NU-ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER.

E ATONS AT GRAND GULF, UNIT 1. Evaluation Of Severe Accident NUREG/CR4311: EVALUATING PREDICTION UNCERTAINTY.

Risks For Plant Operational State 5 Dunng A Refueling Outage Main NUREG/CR4312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN-NUR CR V F 2: EVALUATION OF POTENTIAL SEVEPE NU G CR4315: CANDU REACTORS. THEIR REGULATON IN ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-CANADA, AND THE IDENTIFICATION OF RELEVANT NRC SAFETY ATONS AT GRAND GULF. UNIT 1. Evaluation Of Severe Accident ISSUES Risks For Plant Operational State 5 Dunng A Refuehng NUREG/CR4316 V01: GUIDELINES FOR THE VERIFICATION AND NU [CR VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-0 EV A 10 POTENTIAL SEVERE ACCl-DENTS DURING LOW POWER AND SHUTDOWN OPERATONS AT NURE V 2 GUIDELINES FOR THE VERIFICATON AND NUREG/

44 06 P VALL TION OF POTENTIAL SEVERE VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVEN-ACCIDENTS DURING LOW POWER AND SHUTDOWN OPER-TIONAL SOFTWARE. Survey And Assessment Of Conventional Soft-ware Venfication And Vahdation Methods.

ATIONS AT SURRY, UNIT 1. Evaluation Of Severe Accident Risk NUREG/CR4316 V03: GUIDEUNES FOR THE VERIFICATION AND Dun M 4 Loop ations. Main R NURE /CR4144 V P2: EVALUA OF POT 2NTIAL SEVERE VALIDATON OF EXPERT SYSTEM SOFTWARE AND CONVEN-ACCIDENTS DURING LOW POWER AND SHU1DOWN OPER.

TiONAL SOFTWARE. Survey And Documentahon Of Expert System ATONS AT SURRY, UNIT 1. Evaluation Of Severe Accident Risk Venfication And Vahdation Methodologies.

NUREG/CR4316 V04: GUIDELINES FOR THE VERIFICATION AND Dunng M4 Loop Operations.Appendt.es.

NUREG/CR4150 VOI:

SCDAP/RELAP5/ MOD 3.1 CODE VALIDATION EXPERT SYSTEM SOFTWARE AND CONVENTIONAL MANUAL. Interface Theory.

SOFTWARE. Evaluation Of Knowledge Base Certification Methods.

NUREG/CR4150 V02:

SCDAP/RELAP/ MOD 3.1 CODE NUREG/CR4316 V05: GUIDELINES FOR THE VERIFICATION AND MANUALDamage Progression Model Theory.

VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-NUREG/CR-6150 V03:

SCDAP/RELAP5/ MOD 3.1 CODE TIONAL SOFTWARE. Rationale And Desenption Of V&V Guidehne MANUALUser's Guide And input Manual.

Packages And Procedures.

NUREG/CR4150 V04:

SCDAP/RELAPS/ MOD 3.1 CODE NUREG/CR4316 V06: GUIDEUNES FOR THE VERIFICATION AND MANUALMATPRO-A Ubrary Of Matenals Properties For Ught-VALIDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN-Water-Reactor Accident Analysis.

TIONAL SOFTWARE.Vahdation Scenarios.

NUREG/CR4150 V05:

SCDAP/RELAPS/ MOD 3.1 CODE NUREG/CR4316 V07: GUIDEUNES FOR THE VERIFICATION AND

^

NUREG/

9 N Mi R A NT TO PREDICT PERFORM-

^

A ARE' s

ANCE IN A NUCLEAR POWER PLANT CONTROL ROOM.A Tesi Of NUREG/CR4316 V08: GUIDEUNES'FOR THE VERIFICATON AND VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVEN.

NUR CR4172 EWING PSA-BASED ANALYSES TO MODIFY TIONAL SOFTWARE.Babhography.

TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS NUREG/CR4318: DATA

SUMMARY

REPORT FOR FISSION PROD-NUREG/CR-6244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty NU G 25 AN ICIT STEADY-STATE INITIAll2ATON NUR CR424 V2 OBABILISTIC ACCIDENT CONSEQUENCE PACKAGE FOR THE RELAP5 COMPUTER CODE.

NUREG/CR4343: ON-LINE TESTING OF CALIBRATON OF PROC-UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty ESS INSTRUMENTATION CHANNELS IN NUCLEAR POWER Assessment. Appendices A And B.

NUREG/CR4244 V03: PROBABILISTIC ACCOENT CONSEQUENCE PLANTS. Phase ll Final Report.

NUREG/CR 4349: COST BENEFIT CONSIDERATIONS IN REGULA-UNCERTAINTY ANALYSIS. Dispersion and Depos. tion Uncertainty TORY ANALYSIS.

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

NUREG/CR4257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC NUREG/CR4355: A LIMITED ASSESSMENT OF THE ASEP HUMAN ENERGY OF CANADA LTD CODES.

REUABILITY ANALYSIS PROCEDURE USING SIMULATOR EXAMI-NUREG/CR4261: A

SUMMARY

OF ORNL FISSION PRODUCT RE-NATION RESULTS.

LEASE TESTS WITH RECOMMENDED RELEASE RATES AND DIF.

NUREG/CR4398: EVALUATION OF THE COMPUTERIZED PROCE-FUSON COEFFICIENTS.

DURES MANUAL ll (COPMA-ll).

NUREG/CR4263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR DIVISION OF RISK ANALYSIS & OPERATONS (840429460720 POWER PLANTS. Candidate Guidehnes. Technical Basis And Re.

NUREG/CR4308: AN OVERVIEW OF INSTABILITY AND FING'CllNG search Needs Executive Summary.

DURING IMMISCIBLE FLUID FLOW IN POROUS AND FRACL MED NUREG/CR-6263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR MEDIA.

POWER PLANTS. Candidate Guidehnen. Technical Basis And Re-DIVISION OF ENGINEERING TECHNOLOGY (PRE 860720) search Needs Man Report.

NUREG/CR-3243: COMPARISON OF ASME CODE FATIGUE EVAL NUREG/CR4265: MULTIDISCIPLINARY FRAMEWORK FOR HUMAN UATON METHODS FOR NUCLEAR CLASS 1 PIPING WITH CLASS REllABILITY ANALYSIS WITH AN APPUCATION TO ERRORS OF 2 OR 3 PIPING.

COMMISSION AND DEPENDENCIES.

NUREG/CR4277 V01: HUMAN FACTORS EVALUATION OF TELE-EDO - OFFICE OF MUCLEAR REACTOR REGULATION (POST 800428)

THERAPY. Identification Of Problems And Alternative Approaches.

OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001)

NUREG/CR4277 V02: HUMAN FACTORS EVALUATION OF NUREG/CR 2850 V13: DOSE COMMITMENTS DUE TO RADCAC-TELETHERAPY. Function And Task Analysis TIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1991.

NUREG/CR4277 V03: HUMAN FACTORS EVALUATION OF TELE-NUREG/CR 2907 V13: RADIOACTIVE MATERIALS RELEASED FROM THERAPY. Human-System Interfaces And Procedures.

NUCLEAR POWER PLANTS. Annual Report 1992.

L.

96 NRC Contract Sponsor index NUREG/CR-5758 V05: FITNESS FOR DUTY IN THE NUCLEAR NUREG/CR4385: COMPARISONS OF ANS,ASMF AWS, AND NFPA POWER INDUSTRY. Annual Summary Of Program Performance Re-STANDARDS CITED IN THE NRC STANDARD REVIEW ports CY 1994.

PLAN,NUREG 0800, AND RELATED DOCUMENTS.

NUREG/CR-5973 R02: CODES AND STANDARDS AND OTHER NUREG/CR4386: COMPARISONS OF ANSI STANDARDS CITED IN GUIDANCE CITED IN REGULATORY DOCUMENTS.

THE NRC STANDARD REVIEW PLAN,NUREG-0000, AND RELAT.

NUREG/CR4260: APPLICATION OF NUREG/CR-5999 INTERIM FA-ED DOCUMENTS.

TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COM-LICENSE HENEWAL & ENVIRONMENTAL REVIEW PROJECT DIREC-PONENTS.

TORATE (PDLR) (POST NUREG/CR4382: COMPARISONS OF ASTM STANDARDS CITED IN NUREG/CR-5975 RO1: INCENTIVE REGULATION OF INVESTOR.

THE NRC STANDARD REVIEW PLAN,NUREG4800, AND RELAT.

OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY REGO-ED DOCUMENTS.

LATORS.

i i

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.

ADVANCED SYSTEMS TECHNOLOGY, INC.

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4307:

SUMMARY

OF COMMENTS RECEIVED AT WORK.

UNCERTAINTY ANALYSIS. Dispersion and Deposition Uncertainty SHOP ON USE OF A SITE SPECIFIC ADVISORY BOARD (SSAB) TO Assessment. Appendices C.D.E.F,G H.

FACILITATE PUBLIC PARTICIPATION IN DECOMMISSIONING CASES.

ARIZONA, UNIV. OF, TUCSON, AZ NUREG/CR4308: AN OVERVIEW OF INSTABILITY AND FINGERING AMERICAN NUCLEAR SOCIETY DURING IMMISCIBLE FLUID FLOW IN POROUS AND FRACTURED NUREG/CP-0142 V01: PROCEEDINGS OF THE 7TH INTERNATIONAL MEDIA.

MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS (NURETH-7). Sessions 1-5.

AVAPLAN OY (FINLAND)

NUREG/CP-0142 V02: PROCEEDINGS OF THE 7TH INTERNATIONAL NUREG/CR4141: HANDBOOK OF METHODS FOR RISK BASED MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS ANALYSES OF TECHNICAL SPECIFICATIONS.

(NURETH-7).Sesssons 611.

NUREG/CP 0142 V03. PROCEEDINGS OF THE 7TH INTERNATIONAL BATTELLE HUMAN AFFAIRS RESEARCH CENTERS MEETING ON NUCLEAR REACTOR THERMAL-HYDRAULICS NUREG/CR4330; RESULTS OF REGULATORY IMPACT SURVEY OF (NURETH-7). Sessions 12-16 INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE NUREG/CP-0142 V04: PROCEEDINGS OF THE 7TH INTERNATIONAL OFFICE OF NUCLLAR MATERIALS SAFETY AND SAFEGUARDS.

MEETING ON NUCLEAR REACTOR THERMAL HYDRAULICS (NURETH-7). Sessions 17 24.

BATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES NUREG 1493: PERFORMANCE-BASED CONTAINMENT LEAK TEST ANALYSIS & MEASUREMENT SERVICES CORP.

PROGRAM Final Report.

NUREG/CR4312: ASSESSMENT OF FIBER OPTIC PRESSURE SEN.

NUREG-1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK.

SORS.

TEST PROGRAM. Draft Report For Comment.

NUREG/CR 6334: NEW SENSOR FOR MEASUREMENT OF LOW AIR NUREG/CR-4599 V04 N1: SHORT CRACKS IN PIPING AND PIPING FLOW VELOCITY. Phase i Final Report.

WELDS Seventh Program Report March 1993 - December 1994.

NUREG/CR4343 ON-LINE TESTING OF CALIBRATION OF PROCESS NUREG/CR-6004: PROBABILISTIC PIPE FRACTURE EVALUATIONS INSTRUMENTATION CHANNELS IN NUCLEAR POWER FOR LEAK RATE-DETECTION APPLICATIONS.

PLANTS. Phase ll Fenal Report.

NUREG/CR4235: ASSESSMENT OF SHORT THROUGH-WALL CIR-CUMFE TI CRACKS IN PIPES.Expenments And Analysis, March ARGONNE NATIONAL LABORATORY NUREG/CR-4667 V18: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6251: STAINLESS STEEL SUBMERGED ARC WELD LIGHT WATER REACTORS. Semaannual Report, October 1993 March FUSION LINE TOUGHNESS.

1994-NUREG/CR4264 V01: VALIDITY LIMITS IN J-RESISTANCE CURVE NUREG/CR-4667 V19: ENVIRONMENTALLY ASSISTED CRACKING IN DETERMINATION.An Assessment Of The J(M) Parameter.

LIGHT WATER REACTORS. Semiannual Report.Apni September 1994.

NUREG/CR 6264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE NUREG/CR4109: THE PROBABILITY OF CONTAINMENT FAILURE BY i

DETERMINATION.A Computatonal Approach To Ductile Cracli Growth DIRECT CONTAINMENT HEATING IN SURRY.

NUREG/CR4256 V02: FIELD LYSIMETER INVESTIGATIONS TEST NUR /R 2 FRA R

V UATIONS OF FUSION LINE RESULTS Low-Level Waste Data Base Development Program: Test CRACKS IN NUCLEAR PIPE BIMETALLIC WELDS.

i NUREG/CR-6298: FRACTURE BEHAVIOR OF SHORT CIRCUMFEREN-NURE /CR L St O N D UTIO IN A FOUR LOOP TIALLY SURFACE-CRACKED PIPE.

NUREG/CR4299: EFFECTS OF TOUGHNESS ANISOTROPY AND NUREG/CR4275: MECHANICAL PROPERTIES OF THERMALLY AGED COMBINED TENSION, TORSION AND BENDING LOADS ON FRAC-CAST STAINLESS STEELS FROM SHIPPINGPORT REACTOR COM-TURE BEHAVIOR OF FERRITIC NUCLEAR PIPE.

NUREG/CR4300: REFINEMENT AND EVALUATION OF CRACK-OPEN.

NU EG/

6315. CANDU REACTORS. THEIR REGULATION IN ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL CANADA, AND THE IDENTIFICATION OF RELEVANT NRC SAFETY CRACKS IN PIPES.

ISSUES.

NUREG/CR-6335: FATIGUE STRAIN-LIFE BEHAVIOR OF CARBON BATTELLE MEMORIAL INSTITUTE, P ACIFIC seORTHWEST AND LOW ALLOY STEELS. AlJSTENITIC STAINLESS STEELS AND LABORATORY ALLOY GOS.N LWR ENVIRONMENTS NUREG/CR-2850 V13: DOSE COMMITMENTS DUE TO RADIOACTIVE ARIZGNA STATE UNIV., TEMPE, AZ RELEASES FROM NUCLEAR POWER PLANT SITES IN 1991.

NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR-5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-INDUSTRY. Annual Summary Of Program Performance Reports CY CIDENT CONSEQUENCE MODEL.

1994-NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR-5884 V01: REVISED ANALYST 3 OF DECOMMISSIONING EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI.

FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER DENT CONSEQUENCE MODEL STATION. Effects Of Current Regulatory And Other Consideratx>ns On NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF The Financial Insurance Requirements Of The Decommissioning Rule FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-And....

DENT CONSEQUENCE MODEL.

NUREG/CR-5884 V02: REVISED ANALYSIS OF DECOMMISSIONING NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE FOR THE REFERENCE PRESSURIZED WATER REACTOR POWER UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty STATION. Effects Of Current Regulatory And Other Considerations On Assessment Main Report.

The Financial Insurance Requirements Of The Decommissoning Rule NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE And....

UNCERTAINTY ANALYSIS. Dispersion And Deposttion Uncertainty NUREG/CR-5973 R02: CODES AND STANDARDS AND OTHER GUlO-Assessment. Appendices A And B.

ANCE CITED IN REGULATORY DOCUMENTS.

97

\\

1 1

98 Contractor index NUREG/CR-5975 R01: INCENTIVE REGULATION OF INVESTOR-NUREG/CR4144 V06 P2: EVALUATION OF POTENTIAL SEVERE AC.

OWNED NUCLEAR POWER PLANTS BY PUBLIC UTILITY RF.GULA-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TORS.

SURRY, UNIT 1. Evaluation Of Severe Accident Risk Dunng Mid-Loop NUREG/CR4054: ESTIMATING PRESSURIZED WATER REACTOR DE-Operations. Appendices.

COMMISSIONING COSTS. A User's Manual For The PWR Cost EstF NUREG/CR4172: REVIEWING PSA-BASED ANALYSES TO MODIFY mating Computer Program (CECP) Software.

TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

NUREG/CH-6330: RESULTS OF REGULATORY IMPACT SURVEY OF NUREG/CR4256 V02: FIELD LYSIMETER INVESTIGATIONS - TEST INDUSTRIAL AND MEDICAL MATERIALS LICENSEES OF THE OFFICE OF NUCLEAR MATERIALS SAFETY AND SAFEGUARDS-RESULTS. Low-Level Waste Data Base Development Program: Test Results For Fiscal Years 1990,1991,1992, And 1993.

NUREG/CR4331: ATMOSPHERIC RELATIVE CONCENTRATIONS IN NUREG/CR 6265: MULT! DISCIPLINARY FRAMEWORK FOR HUMAN NU G/

D F FC: PERFORMANCE TESTING OF ELECTRON-M ON AND DEPEN NC ES NU E CR 55 IT AS E S N OF H SEP HUMAN NUREG/CR4305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EQUl-RELIABILITY ANALYSIS PROCEDURE USING SIMULATOR EXAMI-LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS-NATION RESULTS.

ING THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL WASTE NUREG/CR4382: COMPAR! SONS OF ASTM STANDARDS CITED IN DISPOSAL UNITS Background, Theory, And Model Desenption.

THE NRC STANDARD REVIEW PLAN NUREG-0800, AND RELATED NUREG/CR4349: COST-BENEFIT CONSIDERATIONS IN REGULA-DOCUMENTS.

TORY ANALYSIS.

NUREG/CR4385: COMPARISONS OF ANS,ASME.AWS, AND NFPA STANDARDS CITED IN THE NRC STANDARD REVIEW BROWN UNIV., PROVIDENCE, RI PLAN,NUREG 0800 AND RELATED DOCUMENTS.

NUREG/CR4264 VOI: VALIDITY UMITS IN J-RESISTANCE CURVE NUREG/CR 6386: CUMPARISONS OF ANSI STANDARDS CITED IN DETERMINATION.An Assessment Of The J(M) Parameter.

THE NRC STANDARD REVIEW PLAN,NUREG-0800, AND RELATED NUREG/CR4264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE DOCUMENTS DETERMINATION.A Computational Approach To Ductile Crack Growth NUREG/CR4390: RADIOLOGICAL CHARACTERIZATION OF SPENT Under Large-Scale Yielding Conditions.

CONTROL ROD ASSEMRLIES.

CAUFORNIA STATE UNIVERSITY AT FRESNO BATTELLE SEATTLE RESEARCH CENTER NUREG/CR4305: BLT-EC (BREACH, LEACH, TRANSPORT, AND EOUl-NUREG/CR-5758 V05: FITNESS FOR DUTY IN THE NUCLEAR POWER LIBRIUM CHEMISTRY), A FINITE-ELEMENT MODEL FOR ASSESS-INDUSTRY. Annual Summary Of Program Performance Reports CY ING THE RELEASE OF RADIONUCLIDES FROM LOW-LEVEL WASTE 1994.

OlSPOSAL UNITS Background, Theory, And Model Descnption.

BETA CORP, INTERNATIONAL NUREG/CR435t: REVIEW OF SCENAHIO SELECTION APPROACHES CAUFORNIA, UNIV. OF, LOS ANGELES, CA NUREG/CR-4918 V08: CONTROL OF WATER INFILTRATION INTO FOR PERFORMANCE ASSESSMENT OF HIGH-LEVEL WASTE RE-POSITORIES AND RELATED ISSUES.

NEAR SURFACE LLW DISPOSAL UNITS. Progress Report Of Field Ex-penments At A Humid Region Site,Beltsville, Maryland.

BROOKHAVEN NATIONAL LABORATORY CAUFORNIA, UNIV. OF SAN DIEGO, CA NUREG4700 RO1 DFC: HUMAN-SYSTEM INTERFACE DESIGN NUREG/CR4125 VOI: HUMAN FACTORS EVALUATION OF REMOTE NUR 57 R C MINI DT T BLE CONCENTRATIONS AFTERLOADING BRACHYTHERAPY. Human Error And Critical Tasks WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS in Remote Afterloadeng Brachytherapy And Approaches For improved CONTAMINANTS AND FIELD CONDITIONS. Draft Report For Con >

NU CR 2 VO : HUMAN FACTORS EVALUATION OF REMOTE NUREG/CP-0140 VOI: PROCEEDINGS OF THE TWENTY.SECOND AFTERLOADING BRACHYTHERAPY.Functon And Task Analysis.

WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-NUREG/CR4125 V03: HUMAN FACTORS EVALUATION OF REMOTE sion, Advanced Instrumentation & Control Hardware & Software, AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-Human Factors Research, IPE & PRA.

System Interfaces, Procedures And Practces, Training And Orgarwratica-NUREG/CP-0140 V02: PROCEEDINGS OF THE TWENTY SECOND al Practces And Procedures.

WATER REACTOR SAFETY INFORMATION MEETING. Severe Aces-Resea ch, T l

ic Research For Advanced Passive CAUFORNIA, UNIV. OF, SANTA SARBARA, CA NUREG/CR4327: MODELS FOR EMBRITTLEMENT RECOVERY DUE NUREG/CP 14 03: PROCEEDINGS OF THE TWENTY SECOND TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

WAldR REACTOR SAFETY INFORMATION MEETING.Pnmary Sys-tem tural And Seismo Engineering, Aging Research, CENTER FOR NUCLEAR WASTE REGULATORY ANALYSES NUREG-1464: NRC ITERATIVE PERFORMANCE ASSESSMENT PHASE NUREG/CP 0143: PROCEEDINGS OF THE THIRD INTERNATIONAL

2. Development Of Capabilities For Review Of A Performance Assess.

WORKSHOP ON THE IMPLEMENTATION OF ALARA AT NUCLEAR ment For A High-Level Waste Repository.

NUREG/CR4283: FIELD SITE INVESTIGATON: EFFECT OF MINE POWER PLANTS Held At Ha Lklsland, New York.

NUREG/CP-0148; TRANSACT S TWENTY THIRD WATER SEISMICITY ON GROUNDWATER HYDROLOGY.

REACTOR SAFETY INFORMATION MEETING NUREG/CR4347: MULTl-PHASE REACTIVE TRANSPORT THEORY.

NUREG/CR 2907 V13: RADIOACTIVE MATERIALS RELEASED FROM NUREG/CR 6348. THERMALLY DRIVEN MOISTURE REDISTRIBUTION NUCLEAR POWER PLANTS. Annual Report 1992 IN PARTIALLY SATURATED POROUS MEDIA.

NUREG/CR-2907 V14: RADIOACTIVE MATERIALS RELEASED FROM NUCLEAR POWER PLANTS.

CENTRAL RESEARCH INSTITUTE OF ELECTRIC POWER INDUSTRY NUREG/CR-3469 V08: OCCUPATIONAL DOSE REDUCTION AT NU-NUREG/CR4235: ASSESSMENT Cr SHORT THROUGH-WALL ClR.

CLEAR POWER PLANTS: ANNOTATED BIBLIOGRAPHY OF SELECT.

CUMFERENTIAL CRACKS IN PIPES.Expenments And Analysis, March ED READINGS IN RADIATION PROTECTION AND ALARA.

1990 December 1994.

NUREG/CR-5954: EFFECT ON AGING ON PWR CHEMICAL AND VOLUME CONTROL SYSTEM COMPUTER SIMULATION & ANALYSIS,INC.

NUREG/CR4002:

RISK BASED MAINTENANCE NUREG/CR4325: AN IMPLICIT STEADY-STATE INITIAll2ATION PACK-MODELING.Pnontization Of Maintenance importances And Quantifca.

AGE FOR THE RELAPS COMPUTER CODE.

tion Of Maintenance Effectiveness.

NUREG/CR4112: IMPACT OF REDUCED DOSE LIMITS ON NRC Ll-CONSTRUCTION ENGINEERING LABORATORY, INC.

CENSED ACTIVITIES. Maior issues in The implementation Of ICRP/

NUREG/CR4420: SELF-MONITORING SURVEILLANCE SYSTEM FOR NCRP Dose Ltmit Recommendations Final R PRESTRESSING TENDONS.

NUREG/CR4141: HANDBOOK OF MET S. FOR RISK BASED ANALYSES OF TECHNICAL SPECIFICATIONS E.C. RODABAUGH ASSOCIATES,INC.

NUREG/CR4144 V01: EVALUATION OF POTENTIAL SEVERE ACCl-NUREG/CR-3243: COMPARISON OF ASME CODE FATIGUE EVALUA-DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT TION METHODS FOR NUCLEAR CLASS 1 PIPING WITH CLASS 2 SURRY, UNIT 1. Summary Of Results.

OR 3 PtPtNG.

NUREG/CR4144 V06 F1: EVALUATION OF POTENTIAL SEVERE AC-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT EARTHLINGS COTTAGE SURRY, UNIT 1. Evaluation Of Severe Accident Risk Dunng Mid-Loop NUREG/CR-6354 DRF FC: PERFORMANCE TESTING OF ELECTRON-Operations. Main Report.

IC PERSONAL DOSIMETERS. Draft Report For Comment.

f

Contractor index 99 EG4G IDAHO, INC.

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEOUENCE NUREG/CR4276: A COMPILATION OF CURRENT REGULATONS, UNCERTAINTY ANALYSIS. Dsperson and Deposition Uncertainty STANDARDS, AND GUIDELINES IN REMOTE AFTERLOADING BRA-Assessment.Appereces C.D.E,F,G.H.

CHYTHERAPY.

HUGHES TRAINING, INC.

ENERGY, DEPT.0F, ENVIRONMENTAL MEASUREMENTS LABORATORY NUREG/CR4277 V01: HUMAN FACTORS EVALUATON OF TELE-NUREG 1505 DRFT FC: A NONPARAMETRIC STATISTICAL METHOD-THERAPY. Identificahon Of Problems And Alternative Approaches.

OLOGY FOR THE DESIGN AND ANALYSTS OF FINAL STATUS DE-NUREG/CR4277 V02: HUMAN FACTORS EVALUATON OF COMMISSONING SURVEYS Draft Report For Comment TELETHERAPY. Function And Task Analysis.

NUREG 1506 DAFT FC: MEASUREMENT METHODS FOR RADIOLOGI-NUREG/CR4277 V03: HUMAN FACTORS EVALUATON OF CAL SURVEYS IN SUPPORT OF NEW DECOMMISSIONING TELETHERAPY. Human-System interfaces And Procedures.

CRITERIA. Draft Report For Comment.

NUREG/CR4277 V04: HUMAN FACTORS EVALUATON OF TELETHERAPY. Training And Organizatonal Analysis.

EPOCH ENGINEERh0G, INC.

NUREG/CR4277 V05:

HL' MAN FACTORS EVALUATION OF NUREG/CR4313 V01: ROBUST, ACCURATE, AND NON-CONTACTING TELETHERAPY.Uterature Review VfBRATON MEASUREMENT SYSTEM. Summary of Companoon Meas-urements Of The Robust Laser interferonwter And Typical Accelerome-IDAHO NATIONAL ENGINEERING LABORATORY ter Systems.

NUREG/CP Ot44 V01: A WORKSHOP ON DEVELOPING RISK ASSESS-NUREG/CR-6313 V02: ROBUST, ACCURA 1E, AND NONCONTACTING MENT METHODS FOR MEDICAL USE OF RADIOACTIVE VIBRATION MEASUREMENT SYSTEM:3 Supplemental Appendices MATERIAL. Summary.

Presenting Compenson Measurements Or The Robust Laser Interfer*

NUREG/CP4144 V02: A WORKSHOP ON DEVELOPING RISK ASSESS-ometer And Typecal Accelerometer Systerrs.

MENT METHODS FOR MEDICAL USE OF RADIOACTIVE MATERIALSupporting Documents.

N REG / 434 :

MA Y DRIVEN kOISTURE REDISTRIBUTION

^

^ ^

IN PARTIALLY SATURATED POROUS MEDIA.

RE HY

^U C ARACTERIZATION OF HYDROTh' p

ppd NOP NU

/C 62 AG G U

OF BOILING WATER REACTOR HIGH PRESSURE INJECTION SYSTEMS.

GEORGE WASHINGTON UNIV., WASHINGTON, DC NUREG/CR-5535 V01: RELAP5/ MOD 3 CODE MANUALCode Structure, NUREG/CR-6287: MANAGEMENT CONCEPTS AND SAFETY APPLICA.

System Models, And Solution Methods.

TONS FOR NUCLEAR FUEL FACILITIES.

NUREG/CR4535 V02: RELAP5/ MOD 3 CODE MANUAL. User's Guide And input Requirements GERMANY. DEMOCRATIC REPUBLIC NUREG/CR-5535 V04: RELAP5/ MOD 3 CODE MANUALModels And NUREG/C94244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE Correlations.

UNCERTAINTY ANALYSIS. Dsperson and Depositen Uncertainty NUREG/CR-5535 VOS Rl: RELAP5/ MOD 3 CODE MANUALUser's Assessment. Appendices C.D.E.F,G.H.

G I

GERMANY, FEDERAL REPUBLIC OF FINDINGS.

NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4109: THE PROBABILITY OF CONTAINMENT FAILURE BY UNCERTAINTY ANALYSIS. Dsperson And Deposition Uncertainty DIRECT CONTAINMENT HEATING IN SURRY.

Assessment Main Report.

NUREG/CR-6116 V06: SYSTEMS ANALYSIS PROGRAMS FOR NUREG/CR4244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

UNCERTAINTY ANALYSIS. Dsperson And Deposition Uncertainty VERSION 5.0. Graph 6 cal Evaluation Module (GEM) Reference Manual.

AssessmentAppendices A And B.

NUREG/CR-6116 V09: SYSTEMS ANALYSIS PROGRAMS FOR HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

GRAM INC.

VERSION 5.0.Venfication And Vahdation (V&V) Manual.

NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR4116 V10- SYSTEMS ANALYSIS PROGRAMS FOR CHRONIC EXPOSURE RESULTS WITH THE MACCS REACTOR AC-HANDS-ON INTEGRATED RELIABILITY EVALUATIONS (SAPHIRE)

CIDENT CONSEQUENCE MODEL-VERSION 5.0. Data Loading Manual.

NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF NUREG/CR4150 V01:

SCDAP/RELAP5/ MOD 3.1 CODE EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-MANUALinterface Theory.

DENT CONSEQUENCE MODEL.

NUREG/CR4150 V02' SCDAP/RELAP/ MOD 3.1 CODE NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF MANUAL. Damage Progression Model Theory.

FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCI-NUREG/CR-6150 V03:

SCDAP/RELAP5/ MOD 3.1 CODE DENT CONSEQUENCE MODEL NUREG/CR-6143 V06 P1: EVALiJATION OF POTENTIAL SEVERE AC.

MANUALUser's Guide And input Manual NUREG/CR4150 V04:

SCDAP/RELAPS/ MOD 3.1 CODE CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1 Evaluation Of Severe Accident Risks For Plant MANUALMATPRO.-A Library Of Matenals Properties For Light-Water.

Reactor Accident Analysis.

Operatonal State 5 Dunng A Refueling Outage. Main Report And Ap-NUREG/CR4150 V05:

SCDAP/RELAP5/ MOD 3.1 CODE pendices.

MANUAL. Developmental Assessment NUREG/CR-6188 V02: MICROBIAL DEGRADATION OF LOW LEVEL HARRtB CORPORATION INFORMATION SYSTEMS RADIOACTIVE WASTE Annual Report For FY 1994'IGATIONS TEST NUREG/CR4293 V01: VERIFICATION AND VALIDATION GUIDELINES NUREG/CR4256 V01: FIELD LYSIMETER INVEST NU EG 3V E IF T N AN VALIDATION GUIDE-RESULTS Low-Level Waste Data Base Program. Test Results For LINES FOR HIGH INTEGUTY SYSTEMS. Appendices A-D.

NU

/ R42 V'02 LD LY E'R INVESTIGATIONS. TEST HARVARD SCHOOL OF PUBLIC HCT*ri. 80STON, MA RESULTS. Low-Level Waste Data Base Development Program: Test NUREG/CP 0141: PROCEEDINGS OF THE 23RD DOE /NRC NUCLEAR Results For Fiscal Years 1990,1991,1992, And 1993.

AIR CLEANING CONFERENCE. Held in Buffalo,New York,Juty 25 NUREG/CR4257: CANDU 3 TRANSIENT ANALYSIS USING ATOMIC 28,1994' ENERGY OF CANADA LTD CODES.

NUREG/CR4260: APPLICATION OF NUREG/CR-5999 INTERIM FA-HARVARD UNIV, CAMBRIDGE MA TIGUE CURVES TO SELECTED NUCLEAR POWER PLANT COMPO-NUREG/CR-6264 V02: VALIDITY LIMITS IN J-RESISTANCE CURVE NENTS.

DETERMINATION.A Computational Approach To Ductile Crack Growth NUREG/CR4285: SEVERE ACCIDENT NATURAL CIRCULATION Under Large-Scale Ysiding Conditons.

STUDIES AT THE INEL NUREG/CR 6291 V01: NUCLEAR PLANT ANALYZER.Installaton HAWAll. UNIV. OF, HtLO, HI Manual.

NUREG/CR4244 VO1: PROEABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4291 V02: NUCLEAR PLANT ANALYZER. Analyzer Refer-UNCERTAINTY ANALYSIS. Osperson And Depossten Urmertainty ence Manual.

Assessment. Main Report.

NUREG/CR-6291 V03: NUCLEAR PLANT ANALYZER. Computer Visual NUREG/CR4244 V02: PROBAB:LISTIC ACCIDENT CONSEQUENCE System Reference Manual.

UNCERTAINTY ANALYSIS. Deperson And Depositon Uncertainty NUREG/CR4291 V04: NUCLEAR PLANT ANALYZER. Programmer's AssessmentAppendices A And B.

Manual.

l 100 Contractor index ILLINOIS, STATE OF MODELING & COMPUTER SERVICES NUREG-1516 DRFT FC: MANAGEMENT OF RADIOACTIVE MATERIAL NUREG/CR4327: MODELS FOR EMBRITTLEMENT RECOVERY DUE SAFETY PROGRAMS AT MEDICAL FACILITIES. Draft Report For Co*

TO ANNEALING OF REACTOR PRESSURE VESSEL STEELS.

l ment.

l NETHERLANDS, GOVT.OF i

U EG/CR 1 ZE A D DEFORMATION LIMITS TO MAINTAIN

^

CONSTRAINT IN I AND J C TESTING OF BEND SPECIMENS.

" "^

I" ^ "

LOADS IN HS WIDE P TE TESTS NU G/C VO ROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty INSTITUTE FOR CIRCADIUM PHYSIOLOGY Assessment. Appendices A And B.

NUREG/CR4046: ALERTNESS, PERFORMANCE. AND OFF-DUTY NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE SLEEP ON 8-HOUR AND 12-HOUR NIGHT SHIFTS IN A SIMULATED UNCERTAINTY ANALYSIS. Dspersion and Deposition Uncertainty CONTINUOUS OPERATIONS CONTROL ROOM SETTING.

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

JOHN WREATHALL & CO.,INC.

NORTH CAROLINA STATE UNIV., RALElGH, NC NUREG/CR4265: MULTIDISCIPUNARY FRAMEWORK FOR HUMAN NUREG/CR-6398: EVALUATION OF THE COMPUTERIZED PROCE-RELIABILITY ANALYSIS WITH AN APPUCATION TO ERRORS OF COMMISSION AND DEPENDENCIES.

DURES MANUAL 81 (COPMA-II).

NOTRE DAME, UNIV. OF, NOTRE DAME, IN URE /CR RA TURE BEH I R OF SHORT CIRCUMFEREN-NU EG/ R FIRE MODELING OF THE HEISS DAMPF REAKTOR TIALLY SURFACE-CRACKED PIPE.

MUREG/CR 6300: REFINEMENT AND EVALUATION OF CRACK-OPEN-A ALYSES FOR CIRCUMFERENTIAL THROUGH-WALL 1

NUR 1507 FC M ETECTABLE CONCENTRATIONS WITH TYPICAL RADIATION SURVEY INSTRUMENTS FOR VARIOUS LAWRENCE LIVERMORE NATIONAL LABORATORY CONTAM;NANTS AND FIELD CONDITIONS. Draft Report For Com-NUREG/CP-0145:

WORKSHOP ON DEVELOPING SAFE ment SOFTWARE. Held At Hotel Del Coronado, San Dego,CA. July 22 23,1992.

OAK RIDGE NATIONAL LABORATORY WUREG/C45657: AUTOCASK (AUTOMATIC GENERATION OF 3-D NUREG-1514: GUIDANCE FOR A LARGE TABLETOP EXERCISE FOR A CASK MODELS).A Microcomputer Based System For Shipping Cask NUCLEAR POWER PLANT.

Dese Review Analysis.

NUREG/CR 0200 V1 R04: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4242: CASKS (COMPUTER ANALYSIS OF STORAGE CASKS): A MICROCOMPUTER BASED ANALYSIS SYSTEM FOR FERFORMING STANDARDIZED COMPUTER ANALYSES FOR L1-CENSING EVALUATION Controt Modules.

STORAGE CASK DESIGN REVIEW. User's Manual To Version Ib (In-NUREG/CR 0200 V2PIR4: SCALE: A MODULAR CODE SYSTEM FOR W

G b

KL N ANALYSIS OF SPENT FUEL BASKET.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-NUREG/CR4323: RELATIVE RISK ANALYSIS IN REGULATING THE CLNSING EVALUATION Functional Modules F1-F8.

USE OF RADIATION-EMITTING MEDICAL DEVICES A Prehminary Ap NUREG/CR-0200 V2P2R4: SCALE: A MODULAR CODE SYSTEM FOR plication.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Ll-NUREG/CR4324-QUALITY ASSURANCE FOR GAMMA KNIVES.

CENSING EVALUATION. Functional Modules F9-F16.

NUREG/CR-0200 V3 R04: SCALE: A MODULAR CODE SYSTEM FOR LOS ALAMOS NATIONAL LABORATORY PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-NUREG/CR4135: UNCERTAINTY AND SENSITIVITY ANALYSIS OF CENSING EVALUATION. Miscellaneous.

EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCI-DENT CONSEQUENCE MODEL.

NUREG/CR-3243: COMPARISON OF ASME CODE FATIGUE EVALUA-NUREG/CR4214: PRODUCTION AND TESTING OF THE MTAMIN-B6 TION METHODS FOR NUCLEAR CLASS 1 PIPING WITH CLASS 2 OR 3 PIPING.

FINE GROUP AND THE BUGLE-93 BROAD-GROUP NEUTRON /

PHOTON CROSS SECTION LIBRARIES DERIVED FROM ENDF/B-VI NUREG/CR-4219 V10 N2: HEAVY-SECTION STEEL TECHNOLOGY PROGRAM. Semiannual Progress Report For Apni-September 1993.

NUR / 4 V01: PROBABILISTIC ACCIDENT CONSEQUENCE 8

UNCERTAINTY ANALYSIS. Dspersson And Deposition Uncertainty Semiannual Rogress RepM 6 Odober W Ma@

gg NU G'/

2 V ROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CA-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE UNCERTAINTY ANALYSIS. Dspersion And Deposition Uncertainty DAMAGE ACCIDENTS:1994 A STATUS REPORT. Main Report And Ap-NUfYG'/

/CR$ 4 V22: PRECURSORS TO POTENTIAL SEVERE CORE 4 V03 0 B ISTIC ACCIDENT CONSEQUENCE NU UNCERTAINTY ANALYSIS. Dspersion and Deposition Uncertainty DAMAGE ACCIDENTS.1994 A STATUS REPORT. Appendix L Assessment. Appendices C,D.E,F,G,H NUREG/CR 5591 V03; HEAVY SECTION STEEL IRRADIATION NUREG/CR4311: EVALUATING PREDICTION UNCERTAINTY.

PROGRAM. Progress Report For October 1991 September 1992.

NUREG/CR-5591 V04 N2: HEAVY-SECTION STEEL IRRADIATION MARYLAND, UNIV OF, COLLEGE PARK, MD PROGRAM. Semiannual Progress Report For Apnl-September 1993.

NUREG/CR4918 V08: CONTROL OF WATER INFILTRATION INTO NUREG/CR-5591 V05 Nt: HEAVY SECTION STEEL IRRADIATION NEAR SURFACE LLW DISPOSAL UNITS. Progress Report Of Field Ex-periments At A Humed Region Site,Beltsville. Maryland.

PROGRAM. Semiannual Progress Report For September 1993 Through March 1994.

MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE, MA NUREG/CR-5591 V05 N2: HEAVY SECTION STEEL IRRADIATION NUREG/CR4305: BLT EC (BREACH, LEACH, TRANSPORT, AND EQUI-PROGRAM. Progress Report For Apnl 1994 Through September 1994 LIBRIUM CHEMISTRY), A FINITE ELEMENT MODEL FOR ASSESS-NUREG/CR 5591 V06 N1: HEAVY SECTION STEEL IRRADIATION ING THE RELEASE OF RADIONUCLlDES FROM LOW-LEVEL WASTE PROGRAM. Semiannual Progress Report For October 1994 Through DISPOSAL UNITS Background, Theory, And Model Descnptson.

March 1995 NUREG/CR-5857: AGING OF TURBINE DRIVES FOR SAFETY-RELAT-MICRO ANALYSIS & DESIGN, INC.

ED PUMPS IN NUCLEAR POWER PLANTS.

NUREG/CR-6159: USING MICRO SAINT TO PREDICT PERFORMANCE NUREG/CR-5944 V02: A CHARACTERIZATION OF CHECK VALVE IN A NUCLEAR POWER PLANT CONTROL ROOM.A Test Of Validity DEGRADATION AND FAILURE EXPERIENCE IN THE NUCLEAR And Feasibikty' POWER INDUSTRY.1991 Failures.

NUREG/CR4016: AGING AND SERVICE WEAR OF AIR-OPERATED M(TME CORP.

VALVES USED IN SAFETY RELATED SYSTEMS AT NUCLEAR NUREG/CR4263 V01: HIGH INTEGRITY SOFTWARE FOR NUCLEAR POWER PLANTS.

POWER PLANTS. Candidate Guidehnes, Technical Basis And Research NUREG/CR-6089. DETECTION OF PUMP DEGRADATION.

Neede. Executive Summary.

NUREG/CR4119 V01: MELCOR COMPUTER CODE MANUALS Pnmer NUREG/CR4263 V02: HIGH INTEGRITY SOFTWARE FOR NUCLEAR And User's Guides. Version 1.8.3 September 1994.

I POWER PLANTS. Candidate Guidehnes, Technical Basis And Research NUREG/CR4119 V02:

MELCOR COMPUTER CODE l

Needs Main Report.

MANUALS. Reference Manuals. Version 1.8.3 September 1994.

Contractor Index 101 NUREG/CR4192: AGING AND SERVICE WEAR OF SPRING-LOADED NUREG/CR4017: FIRE MODEUNG OF THE HEISS DAMPF REAKTOR PRESSURE RELIEF VALVES USED IN SAFETY RELATED SYSTEMS CONTAINMENT.

AT NUCLEAR POWER PLANTS.

NUREG/CR4109: THE PROBABILITY OF CONTAINMENT FAILURE BY NUREG!CR4214: PRODUCTION AND TESTING OF THE VITAMIN-B6 DIRECT CONTAINMENT HEATING IN SURRY.

FINE GROUP AND THE BUGLE-93 BROAD GROUP NEUTRON /

NUREG/CR4119 V01: MELCOR COMPUTER CODE MANUALS.Pnmer PHOTON CROSS-SECTION LIBRARIES DERIVED FROM ENDF/8-VI And User's Guides. Version 1.8.3 September 1994.

NUCLEAR DATA.

NUREG/CR4119 V02-MELCOR COMPUTER CODE NUREG/CR-6239 V01: SURVEY OF STRONG MOTION EARTHOUAKE MANUALS Reference Manuals.Versson 1.8.3 September 1994.

EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-NUREG/CR4134: UNCERTAINTY AND SENSITIVITY ANALYSIS OF S NG MOTION EARTHOUAKE NU E 39 02:

O C DENT S

ENCE M EL EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-NUREG/CR4135: UNCERTAINTY 'AND SENSITIVITY ANALYSIS OF NU E C 40 PL O

B DING SPECTRA TO SEIS-EARLY EXPOSURE RESULTS WITH THE MACCS REACTOR ACCl-MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF DENT CONSEQUENCE MODEL NUREG/CR4136: UNCERTAINTY AND SENSITIVITY ANALYSIS OF ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO STRONG MOTION EARTHOUAKES.

FOOD PATHWAY RESULTS WITH THE MACCS REACTOR ACCl-NUREG/CR.6256 V02: FIELD LYSIMETER INVESTIGATIONS TEST DENT CONSEOUENCE MODEL.

RESULTS. Low-Level Waste Data Base Development Program: Test NUREG/CR4143 V01: EVALUATION OF POTENTIAL SEVERE ACCl-Resutts For Fiscal Years 1990.1991,1992 And 1993.

DENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT NUREG/CR-6259: CONSTRAINT EFFECTS ON FRACTURE INITIATION GRAND GULF, UNIT 1. Summary Of Resutta.

LOADS IN HSST WIDE PLATE TESTS.

NUREG/CR4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4261: A

SUMMARY

OF ORNL FISSION PRODUCT RE-CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT LEASE TESTS WITH RECOMMENDED RELEASE RATES AND DIF-GRAND GULF. UNIT 1 Evaluation Of Severe Accident Risks For Plant FUSION COEFFICIENTS.

Operational State 5 Dunng A Refueling Outage. Main Report And Ap.

NUREG/CR-6273: BIAXIAL LOADING EFFECTS ON FRACTURE pendees.

TOUGHNESS OF REACTOR PRESSURE VESSEL STEEL NUREG/CR.6143 V06 P2: EVALUATION OF POTENTIAL SEVERE AC-NUREG/CR4284: CRITICALITY SAFETY CRITERIA FOR LICENSE CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1. Evaluation Of Severe Accdent Risks For Plant NURE R

SV MARY OR FISSION PRODUCT RELEASE TEST VI-7.

Operational State 5 Dunng A Refueling Outage. Supporting MELCOR NU EG/ 6 V02 XPER ME T ESULTS FROM CO B ARY F IT CAL AN YSES OF WATER D RA D CONDITIONS Results From Bellows Tested in Corroded Ccaditions.

PACIFIC SCIENCE & ENGINEERING GROUP,INC.

NUREG/CR-6173: A

SUMMARY

OF THE FIRE TESTING PROGRAM AT NUREG/CR4125 V01: HUMAN FACTORS EVALUATION OF REMOTE THE GERMAN HDR TEST FACILITY.

AFTERLOADING BRACHYTHERAPY. Human Error And Cntcal Tasks NUREG/CR4184: SEPARATE EFFECTS TESTING AND ANALYSES TO in Remote Afterloading Brachytherapy And Approaches For improved INVESTIGATE LINER TEARING OF THE 1:1& SCALE REINFORCED System Performance.

CONCRETE CONTAINMENT BUILDING.

NUREG/CR4125 V02: HUMAN FACTORS EVALUATION OF REMOTE NUREG/CR4220: AN ASSESSMENT OF FIRE VULNERABILITY FOR AFTERLOADING BRACHYTHERAPY. Function And Task Analysis.

AGED ELECTRICAL RELAYS.

NUREG/CR 6125 V03: HUMAN FACTORS EVALUATION OF REMOTE NUREG/CR4244 V01: PROBABILISTIC ACCIDENT CONSEQUENCE AFTERLOADING BRACHYTHERAPY. Supporting Analyses Of Human-UNCERTAINTY ANALYSIS. Disperson And Depositen Uncertainty System interfaces. Procedures And Practces Training And Organization-Assessment Main Report.

al Practices And Procedures.

NUREG/CR-6244 V02: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Dispersion And Deposition Uncertainty PENNSYLVANIA, UNIV. OF, PHILADELPHIA, PA Assessment.Appendcas A And B.

NUREG/CR4349: COST BENEFIT CONSIDERATIONS IN REGULA*

NUREG/CR4244 V03: PROBABILISTIC ACCIDENT CONSEQUENCE TORY ANALYS!S-UNCERTAINTY ANALYSIS. Dispersion and Deposition Uncertainty PLG, INC,(FORMERLY PICKG, LOWE & GARRICK,INC.)

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

NUREG/CR-6265: MULTIDiWIPLINARY FRAMEWORK FOR HUMAN RELIABILITY ANALYSIS WITH AN APPLICATION TO ERRORS OF SCIENCE & ENGINEERING ASSOCIATES,1NC NUREG/CR4220: AN ASSESSMENT OF FIRE VULNERABILITY FOR COMMISSION AND DEPENDENCIES.

AGED ELECTRICAL RELAYS.

PRINCETON UNIV, PRINCETON, NJ NUREG/CR4224: PARAMETRIC STUDY OF THE POTENTIAL FOR NUREG/CR4114 V02: AUXILIARY ANALYSES IN SUPPORT OF PER.

BWR ECCS STRAINER BLOCKAGE DUE TO LOCA GENERATED FORMANCE ASSESSMENT OF A HYPOTHETICAL LOW-LEVEL DEBRIS.

WASTE FACILITY.Two-Phase Flow And Contaminant Tran, port in Un.

NUREG/CR-6368: EXPERIMENTAL INVESTIGATION OF SEDIMENTA-saturated Sods With Application To Low-Level Radioactive Waste Dis.

TION OF LOCA GENERATED FlBROUS DEBRIS AND SLUDGE IN posat BWR SUPPRESSION POOLS.

PURDUE UNIV, WEST LAFAYETTE, IN SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY NUREG/CR4256 V02: FIELD LYSIMETER INVESTIGATIONS TEST SCIENCE APPLICATIONS, RESULTS Low-Level Wa;te Data Base Development Program: Test NUREG 0713 V15: OCCUPATIONAL RADIATION EXPOSURE AT COM-Results For Fiscal Years 1990,1991,1992, And 1993.

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

RUTGERS, NEW JERSEY, STATE UNIV. OF, PISCATAWAY, NJ NUREG/CR-4674 V21: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/GR-0014: BALDCYPRESS TREE RING ELEMENTAL CONCEN-DAMAGE ACCIDENTS:1994 A STATUS REPORT. Main Report And Ap-TRATIONS AT REELFOOT LAKE. TENNESSEE,FROM AD 1795 TO pendices A-H.

AD 1820.

NUREG/CR-4674 V22: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS:1994 A STATUS REPORT. Appendix L S. COHEN & ASSOCIATES INC WRE NURE 493 PERFOR ANCE-BASED CONTAINMENT LEAK-TEST YODEL N rmzaten Of M n n nce importances nc NUREG 1493 DFC: ERFORMANCE BASED CONTAINMENT LEAK.

tion Of Maintenance Effectiveness.

NUREG/CR-6141: HANDBOOK OF METHODS FOR RISK-BASED TEST PROGRAM. Draft Report For Comment.

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

ANALYSES OF TECHNICAL SPECIFICATIONS.

PHYLAXIS FOR THE GENERAL PUBLIC IN THE EVENT OF A NU-NUREG/CR4143 V06 P1: EVALUATION OF POTENTIAL SEVERE AC.

CLEAR ACCIDENT.

CIDENTS DURING LOW POWER AND SHUTDOWN OPERATIONS AT GRAND GULF, UNIT 1.Evaluahon Of Severe Accident Riske For Plant SANDIA NATIONAL LABORATORIES Operational State 5 Dunng A Refuehng Outage Main Report And Ap-NUREG/CR-5927 V02: EVALUATION OF A PERFORMANCE ASSESS-pendees.

l MENT METHODOLOGY FOR LOW LEVEL RADIOACTIVE WASTE NUREG/CR4172: REVIEWING PSA-BASED ANALYSES TO MODIFY DISPOSAL FACILITIES Vahdation Needs.

TECHNICAL SPECIFICATIONS AT NUCLEAR POWER PLANTS.

102 Contractor index NUREG/CR4265: MULTIDISCIPUNARY FRAMEWORK FOR HUMAN NUREG/CR-6074 V05: SEALED SOURCE AND DEVICE DESIGN REUABluTY ANALYSIS WITH AN APPLICATION TO ERRORS OF SAFETY TESTING. Technical Report On The Findings Of Tank 4.Inves.

COMMISSION AND DEPENDENCIES.

tigaton Of Failed Radioactive Stainless Steel Troxler Gauges.

NUREG/CR-6316 V01: GUIDELINES FOR THE VERIFICATION AND NUREG/CR4333: BREATH VERSION 1.1 COUPLED FLOW AND VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-ENERGY TRANSPORT IN POROUS MEDIA.Samulator Desenption And AL SOFTWARE.

User Guide.

NUREG/CR4316 V02: GUIDEUNES FOR THE VERIFICATON AND NUREG/CR4347: MULTbPHASE REACTIVE TRANSPORT THEORY.

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-NUREG/CR4348: THERMALLY DRIVEN MOISTURE REDISTRIBUTION AL SOFTWARE. Survey And Assessment Of Conventional Software N E 1

EN ELECTION APPROACHES NUR CR43 VO Ut S FOR THE VERIFICATON AND ORNE ASSESSMEM & HNEE WASTE RE-VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NU

/CR 5 Y

UL CTERl2ATION OF HYDROTH-AL SOFTWARE. Survey And Documentation Of Expert System Venfice-ERMALLY ALTERED NOPAL TUFF.

tion And Vahdation Methodologies.

NUREG/CR4316 V04: GUIDELINES FOR THE VERIFICATION AND STEVENSON & ASSOCIATES VAUDATON EXPERT SYSTEM SOFTWARE AND CONVENTIONAL NUREG/CR-6239 Vot: SURVEY OF STRONG MOTON EARTHOUAKE SOFTWARE. Evaluation Of Knowledge Base Certificaton Methods.

EFFECTS ON THERMAL POWER PLANTS IN CAUFORNIA WITH EM-NUREG/CR4316 V05: GUIDELINES FOR THE VERIFICATION AND PHASIS ON PIPING SYSTEMS. Man Report VAUDATON OF EXPERT SYSTEM SOFTWARE AND CONVENTON-NUREG/CR4239 V02: SURVEY OF STRONG MOTION EARTHOUAKE AL SOFTWARE. Rationale And Desenption Of VSV Guidelme Packages EFFECTS ON THERMAL POWER PLANTS IN CALIFORNIA WITH EM-And Procedures.

PHASIS ON PIPING SYSTEMS. Appendees.

NUREG/CR4316 V06: GUIDEUNES FOR THE VERIFICATON AND NUREG/CR4240' APPLICATION OF BOUNDING SPECTRA TO SEIS.

VALOATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-MIC DESIGN OF PIPING BASED ON THE PERFORMANCE OF AL SOFTWARE.Vahdation Scenanos.

ABOVE GROUND PIPING IN POWER PLANTS SUBJECTED TO NUREG/CR4316 V07: GUiOELINES FOR THE VERIFICATON AND STRONG MOTION EARTHOUAKES.

VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTION-NUREG/CR4358 V01: ASSESSMENT OF UNITED STATES INDUSTRY AL SOFTWARE. User's Manual.

STRUCTURAL CODES AND STANDARDS FOR APPUCATON TO NUREG/CR4316 V08: GUIDELINES FOR THE VERIFICATION AND ADVANCED NUCLEAR POWER REAC10RS. Final Report VAUDATION OF EXPERT SYSTEM SOFTWARE AND CONVENTON-NUREG/CR4358 V02: ASSESSMENT OF UNITED STATES INDUSTRY AL SOFTWARE.Beliography.

STRUCTURAL CODES AND STANDARDS FOR APPLICATION TO ADVANCED NUCLEAR POWER REACTORS.Appendees.

SCIENTECH, INC.

NUREG/CR4310- AN ANALYSIS OF POTASSIUM ODIDE (KI) PRO-SWEDEN GOVT'00: REFINEMENT AND EVALUATON OF OF PHYLAXIS FOR THE GENERAL PUBUC IN THE EVENT OF A NU-NUREG/CR43 CLEAR ACCIDENT.

ING-AREA ANALYSES FOR CIRCUMFERENTIAL THROUGH-WALL i

CRACKS IN PIPES.

SOHAR, INC.

NUREG/CR4293 V01: VERIFICATON AND VALIDATON GUIDEUNES THE KEVRIC COMPANY,INC.

NUREG/CR4287: MANAGEMENT CONCEPTS AND SAFETY APPLICA-FOR HIGH INTEGRITY SYSTEMS Man Acort TONS FOR NUCLEAR FUEL FACILITIES.

NUREG/CR4293 V02: VERIFICATIONN AND VALIDATON GUIDE.

UNES FOR HIGH INTEGRITY SYSTEMS. Appendices A-0.

UNITED KINGDOM NUREG/CR4244 V01: PROBABluSTIC ACCOENT CONSEQUENCE SOUTHWEST POWER CONSULTANTS,1NC.

UNCERTAINTY ANALYSIS. Dispersion And Depositen Uncertamty NUREG-1493: PERFORMANCE-BASED CONTAINMENT LEAK TEST Assessment.Mam Report.

PROGRAM.Fmal Report NUREG/CR4244 V02: PROBABILISTIC ACCOENT CONSEQUENCE NUREG 1493 DFC: PERFORMANCE-BASED CONTAINMENT LEAK

• UNCERTAINTY ANALYSIS. Dispersson And Depositen Uncertamty TEST PROGRAM. Draft Report For Comment.

Assessment.Appendcas A And B.

NUREG/CR4244 V03: PROBABluSTIC ACCIDENT CONSEQUENCE SOUTHWEST RESEARCH INSTITUTE NUREG/CP-0147: PROCEEDINGS OF THE WORKSHOP ON THE ROLE UNCERTAINTY ANALYSIS. Disperson and Depoortion Uncertamty AssessmentAppendees C,D,E,F.G.H.

OF NATURAL ANALOGS IN GEOLOGIC DISPOSAL OF HIGH-LEVEL NUCLEAR WASTE.He6d in San Antonio, Texas July 22-25,1991.

UNIVERSITY OF MEMPHIS, MEMPHIS TN NUREG/CR4074 V04: SEALED SOURCE AND DEVICE DESIGN NUREG/GR-0014: BALDCYPRESS TREE RING ELEMENTAL CONCEN.

SAFETY TESTING.Techncal Report On The Findmgs Of Task 4.Inves-TRATIONS AT REELFOOT LAKE TENNESSEE,FROM AD 1795 TO tigation Of Sealed Source for Paper Mdl Digester.

AD 1820.

F 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.

There were no NUREG/IA reports during this year.

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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 cnd followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

t 52405 CANDU SU AECL Technotopes, he,,

NUREG/CR4257 50-300 Watts Bar Nudear Plant, Und 1, Tennessee NUREG4647 S15 1

52405 CANDU 30 AECL Technolopes, Inc.,

NUREG/CR4315 Vessy Aumonty l

50416 Grand Gull Nucteer Staton, Urut 1, hesanopp NUREG/CR4143 V01 S 390 Wans B4 Nucteer P' ant, Unn 1, Tennessee NUREG0647 616 Power & bght Co.

Vesey Aumonty l

50-416 Grand Gull Nucimer Staton, Urut 1. necesopp NUREG/CR4143 V06 P1 50 390 Watts Bar Nuclear Plant, Unn 1, Tennessee NUREG 0647 $17 l

Power & bght Co.

Vatey Aumonly S 416 Grand GJi Nucieer Slaton, Una 1. Mnemapp NUREG/CR4143 Vos P2 50 390 Wens Bar Nucisar Plant Unn 1, Tennessee NUREG4647 S16 ST450-496 South as und 1 Houston Ughtng & NUREG 1517 50 390 Watts Nucieer Plant, Unn 1. Tennessee NUREG4647 $19 j

Power Co.

Vatev Aumorny ST450499 Soum Texas Protect, Urut 2, Houston Ughtng & NUREG1517 50-390 Watis Bar Nucteer Plant Unn 1. Tennessee NUREG 1528 Power Co.

Valey Authonty S 200 Suny Power Statog Una 1, Vrpna Eisetnc & NUREG/CR4109 50 391 Watts Bar Nucteer Plant Unt,2, Tennessee NUREG4496 S01 Power Ca Vetey Aumonly S 200 Suny Power Staton, Und 1, Vrpna Electnc 4 NUREG/CR4144 V01 S 391 Watts Bar Nucieer Plant, Unn 2. Tennessee NUREG4647 SIS Power Co.

Vetey Authorny S 200 Suny Power Staton, Unn 1, Vrgrue Electne & NUREG/CR4144 V06 P1 50 391 WatJ Bar Nudeer Plant, Und 2, Tennessee NUREG4647 $16 l

i Power Co.

Vasey Aumainy 50 200 Suny Power Staton, Und 1. Vrpne Electnc & NUREG/CR4144 V06 P2 S 391 Wens Bar Nuclear Plant, Und 2, Tennessee NUREG4647 $17 Power ca Veney Aumonly S 261 Suny Power Stahon, Urut 2, Vrpne Electnc & NUREG/CR 6100 S 391 Wans Bar Nucieer Plant und 2. Tennessee NUREG4647 S18 Power Co.

Vaney Aumorty S 390 Watta Bar Nucieer Plant und 1 Tennessee NUREG4496 601 50 391 Wans Bar Nudeer Plant, Und 2, Tennessee NUREG4647 St9 Vaney Aumoniy Vener Aumonty I

l t

l 105

950CPORM3Bs U.O. MJCLaAR REOULATORY COMh488 EON

1. REPORT NUhSER iSIn's itee, WReNM'Num.

anoi,asse 888UOGRAPHIC DATA SHEET h=r*.

" any 1

(*= instructions on w= reer=>

NUREG-0304

a. Tma e ausmu Vol. 20, No. 4

3. DATE REPORT PUBUSetD

?1 ' ^^-f and Mnical Reports l

(Abstract Inder Journal) uONTH YEAR Annual Compdation for 1995 April 1996

4. FIN OR ORANT NUMBER 8' NI8)

8. TYPE OF REPORT

7. PERIOD COVERED (inclushe Dates)

s. PERFORheNG ORGANIZATION - NAME AfC ADO.RESS (N NRC, provtes DMeton, Omos or Region. U.S. Nuclear Regulatory CornnWee6cn, and manns mesroes; w contractor, previos name and n mne aderen.)

Dmsson of Freedom of Information and Publications Services Office of Admimstration U.S. Nuclear Regulatory Commission Washmgton, DC 20555-0001

e. aPONSORING ORGANIZATION = NAME AND ADDRESS (W NRC, type "Same u abowe"; H contractor, provtes NRC DMelon, Omos or Region, U.S. Nuclear Regulatory Comminion, and memn0 address.)

Same as 8, above.

10. suFPLEMENTARY NOTES M. Sheehan, Project Manager

11. ASSTRACT (200 woros or lose)

This journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors, proceed-ings of conferences and workshops, grants, and international agreement reports. The entries in this compilation are indexed for access by title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agreements, contractor, international organization, and licensed facility.

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12. KEY WORD 8/DESCRPTORS (List words or phrases that will aeoist researchers in locating the report.)

13. AVAILABlWTY STATEMENT Unlimited

14. SECURITY CLAS$1FICATION compilation gr,i,,,,,)

abstract index Unclassified (This Report)

Unclassified

16. NUMBER OF PAGES

16. PRICE NRc FORM 335 (2-89)

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