ML20247E401

Regulatory and Technical Reports (Abstract Index Journal). Annual Compilation for 1997

ML20247E401
Person / Time
Issue date:	04/30/1998
From:	NRC
To:
References
NUREG-0304, NUREG-0304-V22-N04, NUREG-304, NUREG-304-V22-N4, NUDOCS 9805180333
Download: ML20247E401 (92)

Text

NUREG-0304 Vol. 22, No. 4 Regulatory and Technical Reports (Abstract Index Journal)

Annual Compilation for 1997 U.S. Nuclear Regulatory Commission Omce of the ChiefInformation Omcer ps neuq,

) f O'A1 )i g 51 g8g980430 0304 R PDR

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 Commissico papers; and applicant and licensee docu-ments and correspondence.

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

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

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

Documents such as theses, dissertations, foreign reports and translations, and non NRC con-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.

Ooples 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.

NUREG-0304 Vol. 22, No. 4 Regulatory and Technical Reports (Abstract Index Journal)

Annual Compilation for 1997 Date Published: April 1998 L L Stevenson, Project Manager Publishing Services Branch Omce of the ChiefInformation Omcer U.S. Nuclear Regulatc7 Commission Washington, DC 20555-0001 a

E

l l

l l

l l

CONTENTS Preface..

y index Tab Main Citations and Abstracts 1

Staff Reports e

Conference Proceedings e

Contractor Reports e

Grant Reports

~

e international Agreement Reports e

Secondary Report Number index 2

Personal AuthorIndex 3

Subject index 4

NRC Originating Organization index (Staff Reports).

5 NRC Originating Organization Index (International Agreements) 6 NRC Contract Sponsor Index (Contractor Reports) 7 Contractor Index.

8 International Organization index 9

Licensed Facility Index 10 iii j

~-

PREFACE This compilation consists of bibliographic data and abstracts for the formal reg ulatory 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 ouarterly and to cumulate it annually. Your comments will be appreciated. Please send them to:

Publishing Services Branch Office of the Chief Information Officer U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NU-REG /CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indexes:

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

NRC Originating Organization Index (International Agreements)

NRC Contract Sponsor index (Contractor Reports)

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

The bibliographic elements of the main citations are the following:

Staff Report NUREG-0808: MARKll CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. ANDER-SON, 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 acces-sion number, (B) the microfiche address (for internal NRC use).

Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK AGSESSMENT AND RELIABIL-ITY ENGINEERING lN 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, j3) report author, (4) organization that compiled the proceedings, (5) date report was published, (6) number bf pages in the report, (7) the NRC Document Con-trol System accession number, (8) the report number of be originating organization, (9) the microfiche ad-dress (for NRC internal use).

v

Contractor Report

{

NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, RR. Sandia Labo-ratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242.

Where the entries are (1) report numbu, (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 Con-trol System accession number, (8) the report number of the originating organization (if given), (9) the micro-fiche address (for NRC internal use).

Grant Report NUREG/GR-0013: APPLICATIONS OF A NEW MAGNETIC MONITORING TECHNIQUE TO IN SITU EVAL-UATION 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. 60328:195.

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 Con-trol System accession number, (8) the report number of the originating organization (if given), (9) the micro-fiche address (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. 3608270424. 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 acces-sioa 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 documen', status of a report:

ADD

- addendum APP

- appendix DRFT - draft ERR

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

Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202) 512-2249 or (202) 512-2171. Non-U.S. customers must make payment in ad-vance either by international Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payabla to the Superintendent of Documents.

vi

NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated re-port. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identifi-I l

cation replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship or the work being re-ported.

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 Publications Branch of the NRC Office of Information Resources Management.

vil

.m w

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 re sort. The bibliographic information (see Preface for details) is followed by a brief abstract of t1is report.

NUREG-0040 V20 NO3: LICENSEE CONTRACTOR AND Act of 1995 (PL 104-66) reonires that AOs be reported to Con-VENDOR INSPECTION STATUS REPORT. Quarterly gress on an annual basis.1 e report includes those events that ReportJuly-September 1996(White Book)

• Office of Nuclear NRC determined to be AOs during fiscal year 1996. This report Reactor Regulation (Post 941001). January 1997. 147pp.

addresses eighteen AOs at NRC-licensed facilities. Two in-9702060133. 91659:001.

volved events at nuclear power plants, eleven involved medical This periodical covers the results of inspections performed by the NRC's Special Inspection Branch, Vendor inspection Sec-brachytherapy misadmirGtrations, and five involved radiophar-tion, that have been distributed to the inspected organizations maceutical misadministration. Eight AOs submitted by the during the period from July through September 1996.

Agreement States are included. One involved stolen radiogra-phy cameras, one involved a ruptured source, one involved re-NUREG 0040 V20 N04: LICENSEE CONTRACTOR AND lease of radioactive material while being transported, one in-VENDOR INSPECTION STATUS REPORT. Quarterly volved a lost source, two involved medical brachytherapy mis-Report, October December 1996.(White Book)

• Office of Nucle-administrations, and two involved radiopharmaceutical misad-ar Reactor Regulation (Post 941001). March 1997.154pp.

ministrations. Four updates of previously reported AOs are in-0703200250. 92191:001, ciuded in this report. Three "Other Events of Interest" events This periodical covers the results of inspections performed by are being reported and one previously reported "Other Events the NRC's Special inspection Branch, vendor inspection Sec.

of Interest" event is being updated.

t!on, that have been distributed to the inspected organizations during the penod from October - December 1996.

NUREG 0304 V21 NO3: REGULATORY AND TECHNICAL RE-PORTS (ABSTRACT INDEX JOURNAL). Compilation For Third NUREG-0040 V21 N01: LICENSEE CONTRACTOR AND Ouarter 1996, July September.

• Office of Information Resources VENDOR INSPECTION STATUS REPORT. Quarterly Management (Post 890205). February 1997.41pp.9703100239.

Report, January-March 1997.(White Book)

• Office of Nuclear 92020:309.

R: actor Regulation (Post 941001). July 1997. 69pp.

This journal includes all formal reports in the NUREG series 9707240136.93885:282.

This periodical covers the results of inspections performed by prepared by the NRC staff and contractors; proceedings of con.

the NRC's Special Inspection Branch, Vendor inspection Sec-ferences and workshops; as well as intemational agreement re-tion, that have been distributed to the inspected organizations ports. The entries in this compilation are indexed for access by during the period from January through March 1997.

title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agree-NUREG-0040 V21 N02: LICENSEE CONTRACTOR AND mente contractor, international organization, and licensed facili-VENDOR INSPECTION STATUS REPORT. Quarterty ty.

Report April-June 1997.(White Book)

• Office of Nuclear Reac-NUREG-0304 V21 N04: REGULATORY AND TECHNICAL RE.

R tion (Post 941001). November 1997. 98pp.

PORTS (ABSTRACT INDEX JOURNAL). Annual Compilation This peno' dical covers the results of Inspections performed For 1996.

• Office of Information Resources Management (Post between April 1997 and June 1997 by the NRC's Special in-890205b AM M97. 93pp. 9705010326. 92697:233.

spection Branch, Vendor inspection Section, that have been See NUREG-0304,V21,NO3 abstract.

di;tributed to the inspected organizations-NUREG-0304 V22 N01: REGULATORY AND TECHNICAL RE-NUREG-0040 V21 N03: LICENSEE CONTRACTOR AND PORTS (ABSTRACT INDEX JOURNAL). Compilation For First VENDOR INSPECTION STATUS REPORT. Quarterly Quarter 1997 January-March.

• Office of Information Resources Report. July. September 1997.(White Book)

• Office of Nuclear Management (Post 890205). June 1997. 42pp. 9709030367.

Reactor Regulation (Post 941001). November 1997. 167pp.

A0236:196.

9712110121. A1386:143.

See NUREG-0304,V21,NO3 abstract.

This penodical covers the results of inspections that were performed by the NRC's Special Inspection Branch, Vendor in-NUREG-0304 V22 N02: REGULATORY AND TECHNICAL RE-spection Section, and that were distributed to the inspected or-PORTS (ABSTRACT INDEX JOURNAL). Compilation For gInizations during the period from July through September Second Quarter 1997, April-June.

• Office of Information Re-

1997, sources Management (Post 890205). October 1997. 47pp.

9711030077. A0990:053.

NUREG-0090 V19: REPORT TO CONGRESS ON ABNORMAL See NUREG-0304,V21,NO3 abstract.

OCCURRENCES. Fiscal Year 1996.

• Office for Analysis & Eval-uation of Operational Data. Director. April 1997. 47pp.

NUREG-0325 R22: U.S. NUCLEAR REGULATORY COMMISSION 9704250153.92624:314.

ORGANIZATION CHARTS AND FUNCTIONAL Section 208 of the Energy Reorganization Act of 1974 (PL STATEMENTS. November 1997.

• NRC - No Detailed Affiliation 93-438) identifies an abnormal occurrence (AO) as an unsched-Given. November 1997. 78pp. 9801130012. A1763:204.

uled incidont or event that the Nuclear Regulatory Commission Functional statements and organization charts for the U.S.

(NRC) determines to be significant from the standpoint of public Nuclear Regulatory Commission offices, divisions, and branches heilth or safety. The Federal Reports Elimination and Sunset are presented.

1

2 Main Citations and Abstracts NUREG-0383 V01 R20: DIRECTORY OF CERTIFICATES OF Regulatory Commission (NRC). This information includes (1)

COMPLIANCE FOR RADIOACTIVE MATERIALS docketed material associated with civilian nuclear power plants PACKAGES. Report Of NRC-Approved Packages.

• Office of and other uses of radioactive materials, and (2) nondocketed Nuclear Material Safety & Safeguards. October 1997. 629pp.

material received and generated by NRC pertinent to its role as 9711060093. A1003:001.

a regulatory agency. The following indexes are includeo: Per.

The purpose of this directory is to make available a converb sonal Author, Corporate Source, Report Number, and Cross lent source of information on packagings approved by the U.S.

Reference of Enclosures to Principal Documents.

Nuclear Regulatory Commission. To assist in identifying packag-ing, an index by Model Number and corresponding Certificate of NUREG-0540 V18 N12: TITLE LIST OF DOCUMENTS MADE Compliance Number is included at the front of Volumes 1 and PUBLICLY AVAILABLE. December 1 31, 1996.

• Offee of infor-

2. An alphabetical listing by user name is included in the back mation Resources Management (Post 890205). March 1997, of Volume 3 of approved Quality Assurance programs. The re-298pp. 9704040207. 92332:001, ports include a listing of all users of each package design and See NUREG-0540,V18,N11 abstract.

approved Quality Assurance programs prior to the pubication date.

NUREG-0540 V19 N01: TITLE LIST OF DOCUMENTS MADE

^

NUREG-3383 V02 R20: DIRECTORY OF CERTIFICATES OF COMPLIANCE FOR RADIOACTIVE MATERIALS 5 970 0086.92 17 1 PACKAGES. Certificates Of Compliance.

• Office of Nuclear Ma-See NUREG 0540,V18,N11 abstract.

terial Safety & Safeguards. October 1997. 571pp. 9711060101.

A1005:001.

NUREG-0540 V19 N02: TITLE LIST OF DOCUMENTS MADE See NUREG-0383,V01,R20 abstract PilBLICLY AVAILABLE. February 1-28, 1997.

• Office of Infor-NUREG-0383 V03 R17: DIRECTORY OF CERTIFICATES.OF mation Resources Management (Post 890205). April 1997.

COMPLIANCE FOR RADIOACTIVE MATERIALS 370pp. 970t>130158. 93345:026.

PACKAGES. Report Of NRC-Approved Quality Assurance Pro-See NUREG-0540,V18,N11 abstract.

grams For Radioactive Materials Packagec.

• Office of Nuclear Material Safety & Safeguards. October 1997. 81pp.

WJ8tEG-0540 V19 N03: TITLE LIST OF DOCUMENTS MADE 9711060104. A1004:266.

PUBLICLY AVAILABLE. March 1-31, 1997.

• Office of Informa.

See NUREG-0383,V01,R20 abstrad.

tion Resources Management (Post 890205). May 1997. 382pp.

9706130181. 93344:001.

NUREG-0386 D08: UNITED STATES NUCLEAR REGULATORY See NUREG-0540,V18,N11 abstract.

COMMISSION STAFF PRACTICE AND PROCEDURE l

DIGEST. Commission, Appeal Board And Licensing Board NUREG-0540 V19 N04: TITLE LIST OF DOCUMENTS MADE Decisions. July 1972 June 1996.

• Office of the General Coun-PUBLICLY AVAILABLE. April 1-30, 1997.

• Office of Information sel (Post 860701). July 1997. 695pp. 9708080183. 94729:001.

Resources Management (Post 890205). June 1997. 405pp.

This 8th edition of the NRC Practice and Procedure Digest 9707180204. 93804:001.

contains a digest of a number of Commission, Atomic Safety See NUREG-0540,V18,N11 abstract.

and Licensing Appeal Board, and the Atomic Safety and Licens-ing Board decisions issued during the period of July 1,1972 to NUREG-0540 V19 N05: TITLE LIST OF DOCUMENTS MADE June 30,1996, interpreting the NRC's Rules.

PUBLICLY AVAILABLE.May 1-31, 1997.

• Office of Information NUREG-0390 V11: TOPICAL REPORT REVIEW STATUS.

• Office Resources Management (Post 890205). July 1997. 385pp.

N R5G 054 18,N11 abstract.

6pp 970 200111 A01 27.

This report provides industry with procedures for submitting NUREG-0540 V19 N06: TITLE LIST OF DOCUMENTS MADE topacal reports, guidance on how the U.S. Nuclear Regulatory PUBLICLY AVAILABLE. June 1 30, 1997.

• Office of Information Commission (NRC) processes and responds to topical report Resources Management 1 Post 890205). August 1997. 324pp.

submittals, and an accounting, with review schedules, of all topi-cat reports currently accepted for review by the NRC. This 9709120056. A0355:001",N11 abstract.

See NUREG-0540,V18 report is published annually.

NUREG-0525 V02 R05: SAFEGUARDS

SUMMARY

EVENT LIST NUREG-0540 V19 N07: TITLE LIST OF DOCUMENTS MADE (SSEL). January 1,1990 Through December 31,1996.

PUBLICLY AVAILABLE. July 1-31, 1997.

• Ofiice of Information FADDEN,M.A.

Operations Branch. July 1997.

o'pp.

Resources Management (Post 890205). September 1997.

9707140081. 93738:158.

419pp. 9710060476. A0622:001.

The Safeguards Summary Event List providet brief summa.

See NUREG-0540,V18.N11 abstract.

ries of hundreds of safeguards-related everas involving nuclear material or facilities regulated by the U.S. Nuclear Regulatory NUREG-0540 V19 N08: TITLE LIST OF DOCUMENTS MADE Commission. Events are descr%u under the categories: Bomb-PUBLICLY AVAILABLE. August 1-31, 1997. MORRIS.E.B. Office related, Intrusion MinM:,pAllegedly Stolen Transportation-relat.

of information Resources Management (Post 890205). October 4 T.,7&dagivandalism, Arson, Firearms-related, Radiological 1997. 353pp. 9711140040. A1107:114.

Sabotage, Non-radiological Sabotage, and Miscellaneous. Be.

See NUREG-0540,V18,N11 abstract.

i NUREG-0540 V19 N09: TITLE LIST OF DOCUMENTS MADE udes e ts d

ng e ma terial, and natural uranium, which are exempt from safeguards PUBLICLY AVAILABLE. September 1 40, 1997.

• NRC - No De-tailed Affiliation Given. November 1997. 337pp. 9712110117.

requirements. Information in the event descriptions was ob-

^

tained from official NRC sources.

NUFkEG-0540,V18,N11 abstract.

NUREG-0540 V18 N11: TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE. November 1 30, 1996.

• Office of infor.

NUREG 9540 V19 N10: TITLE LIST OF DOCUMENTS MADE mation Resources Management (Post 890205). January 1997.

PUBLICLY AVAILABLE. October 1-31, 1997.

• NRC - No De-315pp. 9702070195. 91666:001, tailed Affiliation Given. December 1997. 343pp. 9801120063.

This document is a monthly publication containing descrip-A1737:001.

tions of informaton received and generated by the U.S. Nuclear See NUREG 0540,V18,N11 abstract.

Main Citations and Abstracts 3

l NUREG-0713 V17: OCCUPATIONAL RADIATION EXPOSURE AT NUREG-0750 V44 N05: NUCLEAR REGULATORY COMMISSION COMMERICAL NUCLEAR POWER REACTORS AND OTHER ISSUANCES FOR NOVEMBER 1996. Pages 229-314.

• Offee FACILITIES.1995. Twenty-Eighth Annual Report. THOMAS,M.L of Information Resources Manacement Post 890205). January Division of Regulatory Applications (Post 941217).

1997. 95pp. 9703030170. 91G40:135.

HAGEMEYER D. Science Applications International Corp. (for-Legalissuances of the Cor, mission, the Atomic Safety and Li-merly Science /elications, Inc.). January 1997. 300pp.

censing Board Panel, the Administrative Law Ju@es, arid NRC 9702190015.918C1UO1.

Program Offces are presented.

This report summarizes the occupational exposure data that ara maintained in the U.S. Nuclear Regulatory Commission's NUREG-0750 V44 N06: NUCLEAR REGL1ATORY COMMISSION Radiaton Exposure information and Reporting System (REIRS).

ISSUANCES FOR DECEMBER 1996. Pages 315-432.

• C ffice The bulk of the information contained in the report was com-of information Resources Manager,ent (Post 890205). February piled from the 1995 annual reports submitted by the classes of 1997.123pp. 9703170243. 92175:222.

NRC licensees subject to the reporting requirements of 10 CFR See NUREG-0750,V44,N05 abstract 20.2206. Annual reports for 1995 were received from a total of NUREG-0750 V45101: INDEXES TO NUCLEAR REGULATORY 294 NRC licensees, of which 109 wers operators of nuclear COMMISSION ISSUANCES. January-March 1997.

• Office of in-power raactors in commercial operation. Compilations of the re-formation Resources Management (Post 890205). June 1997.

ports submitted by the 294 heensees indicated that 142,518 in-29pp. 9709020298. A0255:150.

dividuals were monitored,76.822 of whom received a measura-See NUREG-0750,V44,101 abstract' ble dose. The collective dose incurred by these individuals was 24,536 person-cSv (person-rem) which represents a 1% de-NUREG-0750 V45102: INDEXES TO NUCLEAR REGULATORY crease from the 1994 value. The number of workers receiving a COMMISSION ISSUANCES. January-June 1997.

• Office of in-measurable dose also decreased, resulting in the average formation Resources Management (Post 890205). September i

measurable dose of 0.32 cSv (rem) for 1995. The average 1997. 47pp. 9710100233. A0701:313.

measurable dose is defined to be the total collective dose divid-See NUREG-0750,V44,101 abstract.

ed by the number of workers roccivir.g a measurable dose. The figures have been adjusted to account for transient reactof NUREG-0750 V45 N01: NUCLEAR REGULATORY COMMISSION workers. In 1995, the annual collective dose per reactor for light ISSUANCES FOR JANUARY 1997. Pages 147.

• Office of in-vcter reactor licensees was 199 person-cSv (person-rem). This formation Resources Management (Post 890205). March 1997, is the same valuw that was reported for 1994. The annual col-53pp. 9704040210. 92330:268.

lectiv? dose per reactor for boiling water reactors was 256 See NUREG-0750,V44,N05 abstract.

person-cSv (persnn-rem) and, for pressurized water reactors it wis 170 person-cSv (person-rem). Analyses of transient worker NUREG-0750 V45 N02: NUCLEAR REGULATORY COMMISSION data indicated that 17,153 individuals completed work assign-ISSUANCES FOR FEBRUARY 1997. Pages 49-93.

• Office of r9ents at two or more licensees during the monitoring year. The information Resources Management (Post 890205). April 1997, dose distributions are adjusted each year to account for the du-52pp. 9705090049. 92827:001.

plicato reporting of transient workers by multiple licensees. In See NUREG-0750,V44,N05 abstract.

1995, the average measurable dose r:alculated from reported data was 0.26 cSv (rem). The corrected dose distribution result.

NUREG-0750 V45 NO3: NUCLEAR REGULATORV COMMISSION ed in an average measurable dose of 0.32 cSv (rem).

ISSUANCES FOR MARCH 1997.Pages 95 263.

• Office of In-formation Resources Management (Post 890205). May 1997, NUREG-0725 R12: PUBLIC INFORMATION CIRCULAR FOR 175pp. 9706180464. 93393:036.

SHIPMENTS OF IRRADIATED REACTOR FUEL

• Office of See NUoEG-0750,V44,N05 abstract.

97 030080.

8 NUREG-0750 V45 N04: NUCLEAR REGULATORY COMMISSION This circular has been prepared to provide information on the ISSUANCES FOR APRIL 1997.Pages 265-353.

• Office of Infor-shipment of irradiated reactor fuel (spent fuel) subject to rr*gula-mation Resources Management (Post 890205). June 1997, tion by the U.S. Nuclear Regulatory Commission (NRC). It pro-96pp 9

, N REG 47 N

bstract-vides a brief description of spent fuel shipment safety and safe-guards requirements of general interest, a summary of data for NUREG-0750 V45 N05: NUCLEAR REGULATORY COMMISSION 1979 1996 highway and railway shipments, and a listing, by ISSUANCES FOR MAY 1997.Pages 355-435.

• Office of Infor-St te, of recont highway and railway shipment routes. The en-mation Resources Management (Post 890205). July 1997. 88pp.

closed route information reflects specific NRC approvals that 9707230346. 93857:239.

Inve been grantd in response to requests for shipments of See NUREG-0750,V44,N05 abstract.

spent fuel. This publication does not constitute authonty for car-riers or other persons to use the routes described to ship spent NUREG-0750 V45 N06: NUCLEAR PEGULATORY COMMISSION fuel,i,:her categories of nuclear waste, or other materials.

ISSUANCES FOR JUNE 1997. Pages 437-495.

• Office of infor.

NUREG-0750 V44101: INDEXES TO NUCLEAR REGULATORY 66p 970 200220. 0141:06.

COMMISSION ISSUANCES. July-September 1996

• Office of See NUREG-0750,V44,N05 abstract.

Inforrnation Resources Management (Post 890205). January 1997. 21pp. 9701160166. 91456:300.

NUREG-0750 V46 N01: NUCLEAR REGULATORY COMMISSION Digests and indexes for issuances of the Commission, tM INUANCES FOR JULY 1997.Pages 120.

• NRC - No Detailed Atomic Safety and Licensing Board Panel, the Administrative Affetion C;wn. December 1997. 27pp. 9801130019.

Lcw Juiges, the Directors' Decisions, and the Decisions on Pe-A1762:289.

titions for Rulemaking are presented.

See NUREG-0750,V44.N05 abstract.

NUREG-0750 V44102: INDEXES TO NUCLEAR REGULATORY NUREG-0750 V46 N02: NUCLEAR REGULATORY COMMISSION COMMISSION ISSUANCES. July December 1996.

• Office of In-ISSUADOES FOR AUGUST 1997. Pages 21-48.

• NRC No f

formation Resources Management (Post 890205). April 1997.

Detailed Affiliation Given. December 1997. 34pp. 9801260109.

52pp. 9704300062. 92696:262.

A1879:108.

See NUREG.0750,V44,101 abstra'1 See NUREG-0750,V44,N05 abstract.

4 Main Citations and Abstracts NUREG-0837 V16 NO3: NRC TLD DIRECT RADIATION MONI-NUREG-0940 V15 N2 P2: ENFORCEMENT ACTIONS: SIGNIFI-TORING NETWORK. Progress Report. July-September 1996.

CANT ACTIONS RESOLVED REACTOR STRUCKMEYER,R. Region 1 (Post 820201). January 1997.

LICENSEES. Semiannual Progress ReptJuly-December 1996.

• 228pp. 9702060125. 91655:001.

Ofc nf Enforcement (Post 870413). April 1997. 400pp.

This report provides the status and results of the NRC Ther-9705210290. 93063:001.

moluminescent Dosimeter (TLD) Direct Radiation Monitoring This compilation summarizes significant enforcement actions Network. It presents the radiation levels measured in the vicinity that have been resolved during the period (July - December of NRC licensed facilities throughout the country for the third 1996) and includes copies of letters, Notices, and Orders sent quarter of 1996.

by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the NUREG 0837 V16 N04: NRC TLD DIRECT RADIATION MONI-information in this publication will be widely disseminated to TORING NETWORK. Progress Report. October December 1996.

managers and employees engaged in activities licensed by the STRUCKMEYER,R. Region 1 (Post 820201). March 1997.

NRC, so that actions can be taken to improve safety by avoid-322pp. 9704170167. 92513:001.

Ing future violations similar to those described in this publica This report provides the status and results of the NRC Ther.

tion.

moluminescont Dosimeter (TLD) Direct Radiation Monitoring NUREG-0940 V15 N2 P3: ENFORCEMENT ACTIONS: SIGNIFl.

Network. It presents the radiation levels measured in the vicinity CANT ACTIONS RESOLVED MATERIAL of NRC licensed facilities throughout the country for the fourth L.ICENSEES. Semiannual Progress Report, July-December 1996.

quartw of M Ofc of Enforcement (Post 870413). April 1997. 300pp.

NUREG 0837 V17 N01: NRC TLD DIRECT RADIATION MONI-9705,140378J2886 M This compilation summarizes significant enforcement actions TORING NETWORK. Progress Report. January-March 1997.

that have been resolved during the period (July - December STRUCKMEYER,R. Region 1 (Post 820201). May 1997. 238pp.

1996) and includec copies of letters, Notices, and Orders sent 9706160218. 93368:001' by the Nuclear Regulatory Commission to material licensees This report provides the status and results of the NRC Ther-with respect to these enforcement actions. It is anticipated that moluminescent Dosimeter (TLD) Direct Radiation Monitoring the information in this publication will be widely disseminated to Network. It presents the radiation levels measured in the vicantty managers and employees engaged in activities licensed by the of NRC licensed facilities throughout the country for the first NRC, so that actions can be taken to improve safety by avoid-quarter of 1997, ing future violations similar to those described in this publica-NUREG-0837 V17 N02: NRC TLD DIRECT RADIATION MONI-TORING NETWORK. Progress Report. Aprildune 1997.

NUREG-0940 V16 N1 P1: ENFORCEMENT ACTIONS: SIGNIFl.

STRUCKMEYER.R. Region 1 (Post 820201). September 1997.

CANT ACTIONS RESOLVED INDIVIDUAL 229pp. 9710100236. A0703:069.

ACTIONS. Semiannual Progress Report,JanuaryJune 1997.

• This report provides the status and results of the NRC Ther-Ofc of Enforcement (Post 870413). September 1997, 422pp.

moluminesetent Dosimeter (TLD) Direct Radiation Monitoring 9710070386. A0639:001.

Network. It presents the radiation levels measured in the vicinity This compilation summarizes significant enforcement actions of NRC licensed facilities throughout the country for the second that have been resolved during the period (January - June quarter of 1997, 1997) and includes copies of Orders and Notices of Violation sent by the Nuclear Regulatory Commission to individuals with NUREG-0936 V15 N02:

NRC REGULATORY respect to these enforcement actions. It is anticipated that the AGENDA. Semiannual Report. July-December 1996.

• Rules &

information in *nis publication will be widely disseminated to Directives Review Branch (Post 920323). March 1997. 58pp.

managers and employees engaged in activities licensed by the 9704080379. 92389:001.

NRC. The Commission beieves this information may be useful The NRC Regulatory Agenda is a compilation of all rules on to licenseu in making employment decisions, which the NRC Ms recency completed action, or has proposed NUREG-0940 V16 N1 P2: ENFORCEMENT ACTIONS: SIGNIFI-action, or is considering action, and all petitions for rulemaking CANT ACTIONS RESOLVED REACTOR which have been received by the Commission and are pendmg LICENSEES.Semianreal Progress ReportJanuary4une 1997.

• disposition by the Commission. The Regulatory Agenda is up-Ofc of Enforcement, Post 870413). September 1997. 430pp.

dated and issued semiannually.

9710070391. A0640:059.

s comNah sunna6s sWcant enbcenet achs NUREG-0936 Y16 N01:

NRC REGULATORY that have been resolved dunng the period (January - June AGENDASemiannual Report. January 4une 1997.

• Office of 1997) and Wes Wes of hs, Notes, and Ordws sent Administration, Director (Post 940714). August 1997. 70pp.

by the Nuclear Regulatory Commission to reactor licensees with 9709120059. A0354:287.

respect to these enforcement actions, it is anticipated that the See NUREG-0936,V15,N02 abstract.

inhtion in this publication wiH be widely disseminated to managws and employr a engaged in activiti6s licensed by the NUREG-0940 V15 N2 P1: ENFORCEMENT ACTIONS: SIGNIFl.

NRC, so that actions can be taken to improve safety by avoid-CANT ACTIONS RESOLVED INDIVIDUAL future violations similar to those desenbed in this publica.

ACTIONS. Semiannual Progress Report. July-December 1996,.

Ofc of Enforcement (Post 870413). April 1997. 407pp.

9705230149. 93068:001.

NUREG-0940 V16 N1 P3: ENFORCEMENT ACTIONS: SIGNIFI-This compilation summarizes significant enforcement actions CANT ACTIONS RESOLVED MATERIAL that have been resolved during the period (July - December LICENSEES. Semiannual Prog ee Report, January 4une 1997.

• 1996) and includes copies of Orders and Notices of Violation Ofc of Enforcement (Post 8M419. September 1997. 425pp.

sent by the Nuclear Regulatory Commission to indroduals with 9710100160. A0702:001, respect to these enforcement actions. It is anticipated that the This compilation summarizes Vo'uficant enforcement actions information in this publication will be widely disseminated to that have been resolved dunng the period (January - June managers and employees engaged in activities licensed by the 1997) and includes copies of letters, Notices, and Orders sent NRC. The Comreission believes this information may be useful by the Nuclear Regulatory Commission to material licensees to licensees in making enforcement decisions.

with respect to these enforcement actions. It is anticipated that

l I

Main Citations and Abstraf:ts 5

j the information in this publication will be widely disseminated to physical configuration, licensee practices, and licensee proce-managers and employees engaged in activities licensed by the dures. The regulations on spent fuel pools were re"iewed. Inde.

NRC, so that actions can be taken to improve safety by avoid-pendent engineering assessments on the spelt fuel pool ing future violations similar to those described in this publica-system were performed on the electncal system. instruments-tion.

tion, heat loads, and radiation. An assessrr,Jnt fin the risk of NUREG-1021 INT R08: OPERATOR LICENSING EXAMINATION loss of spent fuel whng was perfumed The eraH cmcb STANDARDS FOR POWER REACTORS.

• Office of Nuclear sions are that the typical plant may need improvements in spent Reactor Regulation (Post 941001). January 1997. 460pp.

fuel pool instrumentation, operator procedures and training, and 9703050343. 91953:001.

m W ahon w ol NUREG-1021, " Operator Licensing Examination Standards NUREG-1307 R07:

REPORT ON WASTE BURIAL for Power Rec ' tors," establishes the policies, procedures, and CHARGES. Escalation Of Decommissioning Waste Disposal practices for exmining heensees and applicants for reactor op-Costs At Low-Level Waste Burial Faciiities.

• Division of Regula-erator and senior 'eactor operator licenses at power reactor fa-tory Applications (Post 941217). November 1997. 72pp.

cilities pursuant to Title 10, Part 55, of the Code of Federal 971211012E A1387:194.

Regulations (10 CFR Part 55). The examination standards are One of the requirements placed upon nuclear power reactor intended to assist NRC examiners and facility licensees to licensees by the U.S. Nuclear Regulatory Commission (NRC) is better understand the processes associated with initial and re-for the licensees to periodically adjust the estimate of the cost qualification examinations. The standards also ensure the equi-of decommissioning their plants, in dollars of the current year, table and consistent administration of examinations for all appli-as part of the process to provide reasonable assurance that cants. The standards are for guidance purposes and are not a adequate funds for decommissioning will be available when substitute for the operator licensing regulations (i.e.,10 CFR needed. This report, which is scheduled to be revised periodi-Part 55), and they are subject to revision or other changes in cally, contains the development of a formula for escalating de-internal operator licensing policy. This interim revision permits commissioning cost estimates that is acceptable to the NRC, facility licensees to prepare their initial operator licensing exami-and contains values for the escalation of radioactive waste nations on a voluntary basis pendsg an amendment to 10 CFR bunal costs, by site and by year. The licensees may use the for-Part 55 that will require facility preparation. The NRC intends to mula, the coefficients, and the burial escalation from this report solicit comments on this revision dunng the rulemaking process in their escalation analyses, of they may use an escalation rate Cnd to issue a final Revision 8 in conjunction with the final rule.

at least equal to the escalation approach presented herein.

NUREG-1100 V13: BUDGET ESTIMATES. Fiscal Year 1998.

• Di-NUREG-1350 V09: NUCLEAR REGULATORY COMMISSION IN-vision of Budget & Analysis (Post 890205). February 1997.

FORMATION DIGEST.1997 Edition. GARVER,M. Divisica of 148pp. 9702190053, 91802:045.

Budget & Analysis (Post 890205). May 1997. 144pp.

l This report contains the fiscal year budget justification to Con-9707180165. 93805:045.

j gress. The budget provides estimates for salaries and expenses The Nuclear Regulatory Commisson Information Digest and for the Office of the Inspector General for fiscal year 1998.

(digest) provides a summary of information about the U.S. Nu-NUREG-1125 V18: A COMPILATION OF REPORTS OF THE AD.

clear Regulatory Commission (NRC), NRC's regulatory responsi-VISORY COMMITTEE ON REACTOR SAFEGUARDS.1996 bilities, NRC licensed actrvities, and general information on do-Annual.

• ACRS Advisory Committee on Reactor Safeguards.

mestic and worldwide nuclear energy. The oigest published an-April 1997.128pp. 9706250071. 93502:001.

nually, is a compilation of nuclear and NRC-related data and is This compilaton contains 47 ACRS reports submitted to the designed to provide a quick reference to major facts about the j

Commission, or to the Executive Director for Operations during agency and the industry it regulates. In general, the data cover calendar year 1996. It also includes a report to the Congress on 1975 through 1996, with exceptions noted. Information on gen-the NRC Safety Research Program. All reports have been made erating capacity and average capacity factor for operating U.S.

available to the public through the NRC Public Document commercial nuclear power readors is obtained from monthly Room, the U.S. Library of Congress, and the Internet at http://

operating reports that are submnted directly to the NRC by the www.nrc. gov /ACRSACNW. The reports are divided into two licensee. This information is reviewed by the NRC for consisten.

groups: Part 1: ACRS Reports on Project Reviews, and Part 2:

cy only and no independent validaticn and/or verification is per-ACRS Reports on Generic Subjects. Part 1 contains ACRS te.

formed.

porto by project name and by chronological order within project NUREG-1415 V10 Not: OFFICE OF THE ;NSPECTOR name. Part 2 categorizes the reports by the most appropriate GENERALSemiannual Report To Congress,Ap-il 1,1997 - Sep-generic subject area and by chronological order within subject tomber 30, 1997.

• Office of tre inspector General (Post area.

890417). November 1997. 46pp. 9801120057. A1733:228.

HUREG-1145 V13: U.S. NUCLEAR REGULATORY COMMISSION The inspector General Act of 1978, as amended, requires 1996 ANNUAL REPORT.

• Office of information Resources that inspectors General submit a " Semiannual Report to Con-Management (Post 890205). September 1997. 316pp.

gross" summarizing program activdies. The inspector Generars 9711030085. A0980:001, report is submitted to the Chairman of the NRC not later than This report covers the major activities, events, decisions, and April 30 and October 31 for eadi reporting penod. The Chair-planning that took place dunng Fiscal Year 1996 within the U.S.

man comments on the report and prepares the NRC's Semian-Nuclear Regulatory Commission (NRC) or involving the NRC.

nual Report to Congress as required by the Act. The Chairman then submits *he agency's report and the OiG's report to Con-NUREG-1275 V12: OPERATING EXPERIENCE FEEDBACK gress no lates. an November 30 and May 31, respectively.

REPORT. Assessment Of Spent Fuel Cooling. IBARRA.J.G.;

JONES.W.R.; LANIK G.F.; et al. Division of Safety Programs NUREG-1423 V07: A COMPILATION OF REPORTS OF THE AD-(Post 870413). February 1997. 48pp. 9703170237. 92131:199.

VISORY COMMITTEE ON NUCLEAR WASTE. July 1996 - June This report is an assessment of the likelihood and conse-1997.

• Advisory Committee on Nuclear Waste. August 1997.

quences of bss of spent fuol pool cooling in the nuclear power 83pp. 9709120064. A0354:151.

industry. A generic pressunzed water reactor spent fuel pool This compilation contains 1'. reports issued by the Advisory configuration is developed, and a generic boiling water reactor Committee on Nuclear Waste (ACNW) during the ninth year of spent fuel pool configuration is developed. Over twelve years of its operation. The reports wers submitted to the Chairman and operational data is reviewed and assessed. Six site visits were Commissioners of the U.S. Nuclear Regulatory Commission. All conducted to gather specific information on spent fuel pool reports prepared by the Corrmittee have been made available

6 Main Citations and Abstracts to the publec through the NRC Public Document Room, the U.S.

NUREG-1496 V01: FINAL GENERIC ENVIRONMENTAL IMPACT Library of Congress, and the internet at http://www.nrc. gov /

STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOG-ACRSACNW.

ICAL CRITERIA FOR LICENSE TERMINATION OF NRC Li-CENSED NUCLEAR FACILITIES. Main Report Final Report.

• Di-NUREG-1462 S01: FINAL SAFETY EVALUATION REPORT R5 vision of Regulatory Applications (Post 941217). July 1997.

LATED TO THE CERTIFICATION OF THE SYSTEM 80+

124pp. 9707230337. 93857:115.

DESIGN.Docnet No.52-002.(Asea Brnwn Boveri-Combustion The action being considered in this Final Generic Environ-Engineering)

• Office of Nuclear Ruactor Regulation (Post mental Impact Statement (GEIS) is an amendment to the Nucle-941001). May 1997. 32pp. 9709020345. A0234:112.

ar Regulatory Commission's (NRC) regulations in 10 CFR Part This report supplements the final safety evaluation report 20 to include radiological criteria for decommissioning of lands (FSER) for the System 80+ standard design. The FSER was and structures at nuclear facilities. Under the National Environ-Issued by the U.S. Nuclear Regulatory Commission (NRC) staff mental Policy Act (NEPA), all Federal agencies must consider as NUREG 1462 in August 1994 to document the NRC staff's the effect of their actions on the environment. To fulfill NRC's technical review of the System 80 + design. The application for responsibilities under NEPA, the Commission is preparir$ this the System 80+ design was submitted by Combustion Engi GEIS which analyzes alternative courses of action and the neering, Inc., now Asea Brown Boveri-Combustion Engineering costs and impacts associated with those attematives. In prepar-(ABB-CE) pursuant to Subpart B of 10 CFR Part 52. This sup-ing the final GEIS, the following approach was taken: (1) a list.

plement documents the NRC staff's review of the changes to ing was developed of regule. tory alternatives for establishing ra-the System 80+ design documentation since the issuance of diologicrJ critena for decommissioning; (2) for each alternative, the FSER. ABB-CE made these changes as a result of its a detailed analysis and comparison of incremental impacts, both review of the System 80+ design details. The NRC staff cork radiological and nonradiological, to workers, members of the Public, and the environment, and costs were performe4 and (3) ciudes that the changes to the System 80 + design documenta.

based on the analysis of impacts and costs, conclusions on ra-tion are acceptable, and that ABB-CE's application for design diological cnteria for decommissioning were provided. Contained certification meets the requirements of Subpart B to 10 CFR in are Ms aM concWs MaM to ahng, as Part 52 that e e apphcable and technically relevant to the an objective of decommissioning ALARA, reduction to preexist-System 80 + design.

ing background, the radiological criterion for unrestricted use, NUREG-1492: REGULATORY ANALYSIS ON CRITERIA FOR decommissioning ALARA analysis for soils and structures con.

THE RELEASE OF PATIENTS ADMINISTERED RADIOACTIVE tayg Wamina%n, restMed use and aMemaWe anaysis for special site-specific situations and groundwater cleanup.

MATERIALFinal Report. SCHNEIDER,S.; MCGUIRE,S.A. Divi-sion of Regulatory Applications (Post 941217). February 1997.

NUREG-1496 V02: FINAL GENERIC ENVIRONMENTAL IMPACT 79pp. 9703200259. 92191:156.

STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOG-This regulatory analysis was developed to respond to three ICAL CRITERIA FOR LICENSE TERMINATION OF NRC-LI-petitions for rulemaking to amend 10 CFR Parts 20 and 35 re-CENSED NUCLEAR FACILITIES. Appendices A And B. Final garding release of patients administered radioactive material.

Report.

• Division c' Regulatory Applications (Post 941217).

The petitions requested revision of these regulations to remove July 1997. 478pp. 9707230341. 93856:001.

the ambiguity that existed between the 1-mSv (0.1 rem) total ef-See NUREG-1496,V01 abstract.

factive dose equivalent (TEDE) puolic dose limit in Part 20, adopted in 1991, and the activity-based release limit in 10 CFR NUREG-14ft6 V03: FINAL GENERIC ENVIRONMENTAL IMPACT 35.75 that, in some instances, would permit release of individ-STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOG-uals in excess of the current public dose limit. Three attema.

ICAL CRITERIA FOR LICENSE TERMINATION OF NRC-LI-tives for resolution of the petitions were evaluated. Under Alter.

CENSED NUCLEAR FACILITIES. Appendices C-H. Final Report.

native 1, NRC would amend its patient release enteria in 10

• Division of Regulatory Applications (Post 941217). July 1907.

CFR 35.75 to match the annual pubhc dose limit in Part 20 of 1 651pp. 9707230343. 93854:001.

mSv (0.1 rem) TEDE. Attemative 2 would maintain the status See NUREG-1496,V01 abstract.

quo of using the activity-based release entena currently found in NUREG-1503 S01: FINAL SAFETY EVALUATION REPORT RE-10 CFR 35.75. Under Attemative 3, the NRC would revise ths LATED TO THE CERTIFICATION OF THE ADVANCED BOIL-release critena in 10 CFR 35.75 to specify a dose limit of 5 mSv ING WATER REACTOR DESIGN. Supplement No.1. Docket No.

(0.5 rem) TEDE. The evaluation demonstrates that adoption of 52-001.(General Electric Nuclear Energy)

• Office of Nuclear Attemative 1 would be considerably more expensive to the Reactor Regulation (Post 941001). May 1997. 49pp.

public compared to Altemative 2 (the status quo), primarily due 9706120299. 93337:217.

to increased health care costs associated with more patents re.

This report supplements the final safety evaluation report maining in the hospital than under the current activity-based re.

(FSER) for the U.S. Advanced Boiling Water Reactor (ABWR) quirements. The evaluation also demonstrates that adoption of str.ndard design. The FSER was issued by the U.S. Nuclear the 5-mSv (0.5-rem) dose hmst under Altemative 3 would result Regulatory Commission (NRC) staff as NUREG-1503 in July in a higher net value to the public compared to Attemative 2 1994 to document the NRC staff's review of the U.S. ABWR (the status quo). primarily due to lower health care costs and design. The U.S. ABWR design was submitted by GE Nuclear the increased psychological benefrts to patients and their famL Energy (GE) in accordance with the procedures of Subpart B to lies by permitting earlier release from the hospital. Based on Part 52 of Title 10 of the Code of Federal Regulations. This this analysis, the decision was made that adoption of the 5-mSv supplement documents the NRC staffs review of the changes to (0.5-rem) TEDE limit is consistent with the provisions in 10 CFR the U.S. ABWR design documentation since the issuance of the

)

20.1301(c), and the recommendations of the intemational Com-FSER. GE made these changes primarily as a result of first-of.

I mission on Radiological Protection that an individual be allowed a-kind-engineering (FOAKE) and as a result of the design certifL to receive annual doses up to 5 mSv (0.5 rem) TEDE under cer-cation rulemaking for the ABWR design. On the basis of its tain circumstances. Further, it no longer restncts patient release evaluation, the NRC staff concludes that the confirmatory to a specific activity, and therefore. permits release of patients issues in NUREG 1503 are resolved, that the changes to the with activities that are greater than currently allowed. The pri-ABWR design documentaten are acceptable, and that GE's ap-mary benefit is in reduced hospital stays that provide emotional plication for design certification meets the requirements of Sub-benefits to patients and their famihes, and result in lower health part B to 10 CFR Part 52 that are applicable and technically rei-care costs.

evant to the U.S. ABWR design.

Main Citations and Abstracts 7

NURE41508: FINAL ENVIRONMENTAL IMPACT STATEMENT NUREG-1532: FINAL TECHNICAL EVALUATION REPORT FOR TO CONSTRUCT AND OPERATE THE CROWNPOINT URANi-THE PROPOSED REVISED RECLAMATION PLAN FOR THE UM SOLUTION MINING PROJECT, CROWNPOINT, NEW ATLAS CORPORATION MOAB MILLSource Material License MEXICO. Docket No. 40-8968.(Hydro Resources, Inc.)

• Division No.

SUA-917. Docket No 40-3453.(Atlas Corporation) l of Waste Management (NMSS 940403). February 1997.432pp.

FLIEGEL.M.; BRUMMETT.E.; IBRAHIM,A.; et al. Division of l

9703200270. BLM NM010-93-02. 92192:001.

Waste Management (NMSS 940403). March 1997, 200pp.

This Final Environmental Impact Statement (FEIS) addresses 9704100173. 92418:016.

)

issuing a combined source and 11e(2) byproduct material li-This final Technical Evaluation Report (TER) summarizes the l

conse and minerals operating leases for Federal and Indian U.S. Nuclear Regulatory Commission staff's review of Atlas Cor-lands to Hydro Resources, Inc. (HRI). This action would author.

Poration's proposed reclamation plan for its uranium mill tailings ize the company to conduct in situ Iwh uranium mining in pile near Moab, Utah. The proposed reclamation would allow McKinley County, New Mexico. Such mining would involve drill-A as to 0) mclaim the taHings pde fu pennanem disposal aM ing wells to the ore bodies, then recirculating ground water forti-I ng enn m al care h a gomne agwy n ns cunent fied with dissolved oxygen and sodium bicarbonate to mobilize

( r inq ish e sibi e er ha ng its N C uranium minerals found in the rock. Uranium would then be re-moved from the aqueous mining solutions using ion exchange cense terminated. The NRC staff concludes that, subject to li-technology in processing plants located in three separate cense conditions identified in the TER, the proposed reclama-project areas. A central plant would provide drying and packag-tion plan rneets the requirements identified in NRC regulations, which appear primarily in 10 CFR Pad 40.

ing equipment for yellow-cake production for the entire project.

The FEIS was prepared by a joint interagency review group, in-NUREG-1536: STANDARD REVIEW PLAN FOR DRY SPENT cluding the U.S. Nuclear Regulatory Commission (NRC), the FUEL STORAGE SYSTEMS. Final Report.

• Office of Nuclear U.S. Bureau of Land Management (BLM) and the U.S. Bureau Material Safety & Safeguards. January 1997, 179pp.

of indian Affairs (BIA). This FEIS describes the staffs analyses 9703130386. 92107:104.

concerning the evaluation of: (1) the purpose of and need for The Standard Review Plan (SRP) for Dry Cask Storage Sys-the proposed action; (2) attematives to the proposed action; (3) tems provides guidance to the Nuclear Regulatory Commission the environmental resources that could be affected by the pro-staff in the Spent Fuel Project Office for performing safety re-posed action and alternatives; (4) the potential environmental views of dry cask storage systems. The SRP is intended to ensure me - aW and undonnq of me staH mWews and consequences of the proposed action and alternatives; and (5) the economic costs and benefits associated with the proposed present a basis for the review scope and requirements. Part 72, action. The evaluation is based on a comprehensive review of Subpart B generdy specifies the information needed in a h-ce se PP HRl's heense application, environmental reports, related submit-

]at and h

d s R ide 6 tals, independent information sources, and wntten and oral comments received on the Draft Environmental impact State-

.. Standard Format and Content for a Topical Safety Analysis mant. On the basis of its independent review, the staff con-Report for a Spent Fuel Dry Storage Cask" contains an outline of the specific information required by the staff. The SRP is di-cludes that the potential significant impacts of the proposed vided into 14 sections which reflect the standard application project can be mitigated, and that HRI should be issued a com-format. Regulatory requirements, staff position, industry codes bined source and ite(2) byproduct material license from NRC, and standards, acceptance criteria, and other information are and minerals operating leases from BLM and BIA.

discussed. Comments, errors or omissions, and suggestions for improvement should be sent to the Director, Spent Fuel Project NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL Office, U.S. Nuclear Regulatory Commission, DC 20555-0001.

SAFETY PROGRAMS AT MEDICAL FACILITIES. Final Report.

CAMPER,LW.; SCHLUETER,J.; WOODS,S.; et al. Division of in-NUREG-1542 V02: ACCOUNTABILITY REPORT FISCAL YEAR dustrial & Medical Nuclear Safety (Post 870729). May 1997.

1996. CONNELLY,S.R. Office of the Controller (Post 890205).

193pp. 9706120304. 93341:100.

April 1997. 92pp. 9705210298. 93064:001.

A Task Force composed of eight U.S. Nuclear Regulatory The U.S. Nuclear Regulatory Commission (NRC) is one of six Commission and two Agreement State program staff members Federal agencies participating in a pilot project to streamline fi-developed the guidance contained in this report. The purpose of nancial management reporting. The goal of this pilot is to con-this report is to describe a systematic approach for affective solidate performance.related reporting into a single accountabil-management of radiation safety programs at medical facilities.

ity report in accordance with the Government Management TNs is accomplished by emphasizing the roles of institution ex.

Reform Act (GMRA) of 1994. The NRC's second accountability ecutive managemont, radiation safety committee, and radiaton mpat consWaMs me infamah phsh mW m M safety officer. Various aspects of program management are dis-NRC s annual financial statement required by the Chief Finan-cial Of'icers Act of 1990, as amended; the chairman's annual cussed and include guidance on selecting the radiation safety officer, determining adequate resources for the program, the report to the President and the Congress, required by the Fed-use of contractual services such as consultants and service

,,al Managers' Financial Integnty Act of 1982; and the Chair-man's semiannual report to the Congress on management deci-companies, the conduct of audits, the roles of authorized users sions and final actions on Office of Inspector General (OlG) and supervised Individuals, NRC's reporting and notification re-audit recommendations, required by the Inspector General Act quirements, and a general desenption of how NRC's licensing, of 1978, as amended. This report also includes performance inspection, and enforcement programs work. Appendices pro-measures, as required by the Chief Financial Officers Act of vide detailed gu6 dance on specific aspects of a radiaton safety 1990.

program and the glossary defines terms used throughout the report. The guidance contained herein does not represent new NUREG-1545: EVALUATION CR;TERIA FOR COMMUNICA.

or proposed regulatory requirements and bcensees will not be TIONS-RELATED CORRECTIVE ADON PLANS.

• Office of inspected against any portion of it. Additionally, regulatory com.

Nuclear Reactor Regulation (Post 941001).

• Division of Sys-pliance with all apphcable regulations is not assured by licens, tems Technology (Post 941217). February 1997. 69pp.

ees who adopt any portion of, or apply the principles descnbed 97 in, Ws mpwt.

d s guidance and enteria for U.S. Nucle-ar Regulatory Commission (NRC) personnel to use in evaluating corrective action plans for nuclear power plant communications.

I I

l 1

i 8

Main Citations and Abstracts The document begins by describing the purpose, scope, and of Sealed Sources and Devices for Performing industnal Radi-applicability of the evaluation enteria. Next, it presents back-ography." This draft report, where applicable, provides a more ground information concerning the communications process, risk-informed, performance-based approach to industrial radiog-j root causes of communication errors, and development and im-raphy licensing consistent with the current regulations. This pigmentation of corrective actions. The document then defines draft NUREG Report is being distributed for comment to en-specific criteria for evaluating the effectiveness of the corrective courage public participation in its development. It represents the action plan, interview protocols, and an observation protocol re-current position of NRC staff, which is subject to change after lated to communication processes. This document is intended the review of public comments. Commen4 received will be con-only as guidance. It is not intended to have the effect of a regu-sidered in developing the final NUREG Report that represents lation, and it does not establish any binding requirements or in-the official NRC staff position. Until the final NUREG Report is terpretatione of NRC regulations.

published, this draft NUREG Report represents the best avail-NUREG-1555 DRFT: ENVIRONMENTAL STANCARD REVIEW able guidance, and may be used when preparing requests for PLAN. Standard Review Plans For Envkonmental Reviews For Ecenseg actions. Once the final NUREG Report is published, Nuclear Power Plants.

• Office of Nuclear Reactor Regulation NRC staff will use it in its review of requests for licensing ac-(Post 941001). August 1997. 650pp. 9801120289. A1734:001, tions. The draft and final NUREG Reports may differ. If your h-This document, for public review and comment, provides guid.

cense was issued or amended based on recommendations in ance to be staff in implementing provisions of 10CFR51, "Envi-the draft NUREG Report and you feel that the final guidance is ronmental Protection Regulations for Domestic Ucensing and moro advantageous to you, you may choose to request an Related Regulatory Functions," related to new site / plant appi;.

amendret.

NUREG-1556 V3 DRF FC: CONSOLIDATED GUIDANCE ABOUT ta d ev ars or n nm e

o o ttsuction Permit Applications for Nuclear Power Plants,"

MATERIALS LICENSES. Applications for Sealed Source And NUREG-0555, issued in 1978. Since then, new technical issues.

Device Evaluation And Registration. Draft Report For Comment.

such as environmental justice and severe-accident mitigation LUBINSKI,J.; BAGGETT.S.; BROADDUS D.; et al. Division of in-design altematives-and new $ censing structures-such as early dustrial & Medical Nuclear Safety (Post 870729). September site permits, combined licenses, and license renewal-have 1997,134pp. 9801120269. A1740:160.

raised the need for new regulatory guidance.

As part of its redesign of the materials licensing process, NRC is consolidating and updating numerous guidance docu-NUREG-1556 V01 CONSOLIDATED GUIDANCE ABOUT MATE-ments into a single comprehensive repository as described in RIALS UCENSES. Program-Specific Guidance About Portable NUREG-1539 and draft NUREG-1541. NUREG-1556, Vol. 3, is Gauge Licensea Final Report. VACCA,P.C.; WHITTEN,J.E.;

intended for use by applicants, registrants, and NRC staff in ap-PELCHAT.J.M.; o at Division of Industrial & Medical Nuckar plying for and evaluating applications for registration of sealed Safety (Post 81 1729). May 1997. 146pp. 9706180459.

sources and devices. The final version of this document is in-93393:214.

tended to supersede guidance provided in NUREG-1550, As part of its redesign of the materials licensing process'

" Standard Review Plan for Applications for Sealed Source and NRC is consolidating and updating numerous guidance doc

• Device Evaluations," Regulatory Guide 10.10. " Guide for the ments ir to a single comprehensive repository as desenbed in Preparation of Applications for Radiation Safety Evaluation and NUREG-1539 and draft NUREG 1541. NUREG 1556 Vol.1 is Registration of Devices Containing Byproduct Material," and the first program-specific guidance developed for the new proc-Regulatory Guide 10.11. " Guide for the Preparation of Applica-ess and will serve as a template for subsequent program specif-tions for Radiation Safety Evaluation and Registration of Sealed ic guidance. This document is intended for use by applicants, Sources Containing Byproduct Material."

licensees, and NRC staff and will also be available to Agree-ment States. This document supersedes the guidance previous-NUREG-1556 V4 DRF FC: CONSOUDATED GUIDANCE ABOUT ly fcund in draft Regulatory Guide DG-0008, " Applications for MATERIALS LICENSES. Program Specific Guidance About the Use of Sealed Sources in Portable Gauging Devices," and Fixed Gauge Licenses. Draft Report For Comment.

in NMSS Policy and Guidance Directive 2-07, " Standard Review HENDERSON,P.J.; KIRKWOOD,A.S.; LEWIS,S.H.; et al. Division Plan for Applications for Use of Sealed Sources in Portable of Industrial & Medical Nuclear Safety (Post 870729). October Gauging Devices. This final report takes a more risk-informed, 1997.190pp. 9801120274. A1739:001, performance-based approach to licensing portable gauges, and reduces the information (amount and level of detail) needed to As part of its redesign of the materials licensing process support an application to use these devices. It incorporates NRC is consolidating and updating numerous guidance docu-many suggestions sutwnstted during the comment period on ments into a single comprehensive repository as described in NUREG-1539 and draft NUREG 1541. Draft NUREG-draft NUREG 1556, Vol 1. When published, this final report 1556,Vol.4, " Consolidated Guidance about Materials Licenses:

should be used in preparing portable gauge license applications.

P am-Specific Guidance about Fhed Gauges Ucenses,

NRC staff will use this final report in reviewing these applica-W ON 199L is N M NMeck g'he &

veloped for the new process cnd is intended for use by appli-NUREG-1556 V2 DRF FC: CONSOLIDATED GUIDANCE ABOUT cants, licensees, and NRC staff, and will also be available to MATERIALS LICENSES. Program Specific Guidance About in.

Agreement States. This document combines and updates the dustrial Radiography Licenses. Draft Report For Use And Corn.

guidance found in Draft Regulatory Guide and ment. CARRICO J.B.: COLLINS,D J.; WHITE.D.; et al. Division of Valuel* ERR 17'mpact Statement FC 404-4, " Guide for the Industrial & Medical Nuclear Safety (Post 870729). August 1997.

Preparation of Applications for Licenses for the Use of Sealed 170pp. 9710150143. A0729:046.

Sources and Nonportable Gauging Devices," dated January i

l This document is ultimately intended for use by applicants, li-1985, and in NMSS Policy ana Guidance Directive, FC 85-4, consees, and NRC staff and will also be available to Agreement

" Standard Review Plan for Applications for Use of Sealed States. This guidance corresponds with the revision to 10 CFR Sources and Nonportable Gauging Devices," dated February 6, l

Part 34 published in May 1997. This document combines and 1985. This draft report takes a more risk-informed, perform-I supersedes the guidance pr3viously found in draft Regulatory ance-based approach to licensing fixed gauges, and reduces l

Guide FC 401-4, " Guide for the Preparation of Applicatms for the information (smount and level of detail) needed to support the Use of Sealed Sources and Devices for Performing industri-an application to use these devices. Note that this document is al Radiography," and in NMSS Policy and Guidance Directive strictfy for public comment and is NOT for use in preparation or FC 84-15, " Standard Review Plan for Applications for the Use review of fixed gauge licenses until it is Nblished in final form.

i Main Citations and Abstracts 9

NUREG-1556 V5 DRF FC: CONSOLIDATED GUIDANCE ABOUT ment or commercial-scale license, or for the renewal or amend-MATERIALS LICENSES Program-Specific Guidance About Self-ment of an existing license is required to provide detailed infor-

{

Shielded Irradiator Licenses. Draft Report For Comment.

mation on the facilities, equipment, and procedures used and an i

VACCA,P.C.; COLLINS.D.J.; MITCHELL.M.W.; et al. Division of environmental report that discusses the effects of proposed op-Industrial & Medical Nuclear Safety (Post 870729). October erations on the health and safety of the public and on the envi-1997.180pp. 9801120276. A1740:001.

ronment. The Standard Review Plan is prepared for the guid-As part of its redesign of the materials licensing process, ance of staff reviewers in the Office of Nuclear Material Safety NRC is consolidating and updating numerous guidance docu-and Safeguards in performing safety and environmental reviews ments into a single comprehensive repository as desenbod in of applications to develop and operate uranium in situ leach fa-NUREG 1539, " Methodology and Findings of the NRC's Materi-cilities. It provides guidance for new license applications, renew-als Licensing Process Redesign," dated April 1996, and draft als, and amendments. The principal purpose of the standard NUREG-1541, " Process and Design for Consolidating and Up-review plan is to assure the quahty and uniformity of staff re-dating Materials Licensing Guidance," dated April 1996.

views and to present a well-defined base from which to evalu-NUREG-1556, Vol. 5, " Consolidated Guidance about Materials ate changes in the scope and requirements of a review. The Licenses: Program-Specific Guidance about Self-Shielded Irra-standard review plan is written to cover a variety of site condi-diator Licenses," dated October 1997, is the fifth program-spe-tions and facility designs. Each section is written to provide a cific guidance developed for the new process and is intended description of the areas of review, review procedures, accept-for use by applicants, licensees, and NRC staff and will also be ance enteria, and evaluation of findings. However, for a given available to Agreement States. This document combines and application, the staff reviewers may select and emphasize par-updates the guidance found in Regulatory Guide 10.9, Revision ticular aspects of each standard review plan section as is ap-1, " Guide for the Preparation of Applications for Licenses for propriate for the application.

the Use of Self-Contained Dry Source-Storage Gamma Irradia-tors," dated December 1988, and in NMSS Policy and Guidance NUREG-1571: INFORMATION HANDBOOK ON INDEPENDENT Directive FC 84-16, Revision 1, " Standard Review Plan for Ap-SPENT FUEL STORAGE INSTALLATIONS. RADDATZ,M.G.;

plications for Use of Self-Contained Dry Source-Storage WATERS,M.D. Office of Nuclear Material Safety & Safeguards.

Gamma Irradiators," dated January 26,1989. This draft report December 1996.140pp. 9703030211. 91940:001.

takes a more nsk-informed, performance-based approach to li-in this information handbook, the staff of the U.S. Nuclear censing self-shielded irradiators, and reduces the information Regulatory Commission desenbes (1) background information (amount and level of detail) needed to support an application to regarding the licensing history of independent spent fuel storage use these devices. Note that this document is strictty for public installations (ISFSis), (2) a discussion of the licensing process, comment and is not for use in preparing or reviewing self-(3) a description of all currently approved or certified models of shielded irradiator licenses until it is published in final form.

dry cask storage systems (DCSSs), and (4) a description of NUREG-1562 DRFT FC: STANDARD REVIEW PLAN FOR APPLI-sites currently stonng spent fuel in an ISFSI. Storage of spent CATIONS FOR LICENSES TO DISTRIBUTE DYPRODUCT MA.

fuel at ISFSis must be in accordance with the provisions of 10 TERIAL TO PERSONS EXEMPT FROM THE REQUIREMENTS CFR PART 72. The staff has provided this handbook for infor-FOR AN NRC LICENSE.10CFR Parts 30.14,30.15, mation purposes only. The accuracy of any information herein is 30.16,30.18,30.19 & 30.20. CAMPER,LW.; RICH T.; GREENE,S.

not guaranteed. For venfication or for more details, the reader Division of Industrial & Medical Nuclear Safety (Post 870729).

should refer to the respective docket files for each DCSS and January 1997. 82pp. 9702100058. 01802:274.

ISFSI site. The information in this handbook is current as of Exemptions from the requirements for an NRC license to per.

September 1,19%.

cons who receive, possess, use, transfer, own, or acquire by-product material in exempt distribution products are provided in NUREG-1572: SAFETY EVALUATION REPORT RELATED TO THE RENEWAL OF THE OPERATING LICENSE FOR THE RE-10 CFR Part 30, " Rules of General Applicability to Domestic Li-SEARCH REACTOR AT NORTH CAROLINA STATE UNIVERSI-censing of Byproduct Material. Exempt distribution products in-ciude silicon chips, electron tubes, resins, check sources, gun-TY.

• Office of Nuclear Reactor Regulation (Post 941001). April sights, and smoke detectors and are generally distributed by 1997.120pp. 9705280265. 93126:178.

persons who have a specific license from the Commission au-This safety evaluation report (SER) summarizes the findings thortzing such distribution to persons exempt from the require-of a safety review conducted by the staff of the U.S. Nuclear ments for an NRC license. This document provides assistance Regulatory Commission (NRC), Office of Nuclear Reactor Regu-to apphcants and licensees in preparing license applications lation (NhE). The staff conducted this review in response to a and describes the methods acceptable to NRC license review-timely application filed by North Carolina State University (the li-ers in implementing the regulations and the techniques used by censee or NCSU) for a 20-year renewal of Facihty Operating L)-

the reviewers in evaluating the applications to determine if the conse R-120 to continue to operate the NCSU PULSTAR re-proposed exempt distribution activity is acceptable for licensing search reactor. The facility is locaiad in the Burlington Engineer-purposes. The guidance contained herein does not represent ing Laboratory complex on the NCSU campus in Raleigh, North new or proposed regulatory requirements, and licensees wil1 not Carolina. In its safety review, the staff considered information be inspected against any portion of it. In accordance with NRC submrtted by the licensee (including past operating history re-usa 08, the word "should', is used when discussing or referenc-corded in the licensee's annual reports to the NFC), as well as ing NRC regulations. Additionally, regulatory comphance with all inspection reports prepared by NRC Region ll personnel and first-hand observations. On the basis of this review, the staff applicable regulations is not assured by bcensees who adopt aay portion of, or apply the principles described in, this guid-concludes that NCSU can continue to operate the PULSTAR re-search reactor, in accordance with its application, without en-We.

dangering the health and safety of the public.

i NUREG-1569 DRFT: DRAFT STANDARD REVIEW PLAN FOR IN i

SITU LEACH URANIUM EXTRACTION LICENSE APPLICA-NUREG-1574: STANDARD REVIEW PLAN ON ANTITRUST TIONS.

• Dnnsion of Waste Management (NMSS 940403). Oc-REVIEWS. Final Report. LAMBE,W.M.; DAVIS,M.J. Office of Nu.

tober 1997. 250pp. 9801120284. A1738:001, clear Reactor Regulation (Post 941001). December 1997.31pp.

A Nuclear Regulatory Commission source and byproduct ma-9712230314. A1501:014.

terial hcense is required to recover uranium by in situ leach ex-The Nuclear Regulatory Cornmission is issuing this Standard I

traction techniques under the provisions of Title 10 Code of Review Plan to desenbe the procedure used to implement the Federal Regulations, Part 40 (10 CFR 40), Domestic Licensing antitrust review and enforcement process presenbed in Sections of Source Material. An apphcant for a research and develop-105 and 186 of the Atomic Energy Act of 1954, as amended.

10 Main Citations and Abstracts This SRP reflects current regulations and policy, and will be up-NUREG-1603 DRFT: INDIVIDUAL PLANT EXAMINATION dated to reflect changes in NRC regulations.

DATABASE. User's Guide. SU,T.M. Office of Nuclear Regulatory Research (Post 941217). DANZlGER,LM Office of information NUREG-1574 DRFT FC: STANDARD REVIEW PLAN ON Resources Management (Post 890205). LIN,C.C.; et al. Brook.

ANTITRUST. Draft Report For Comment.

LAMBE,W.M.;

haven National Laboratory. April 1997.150pp. 9705010320.

DAVIS,M.J. Office of Nuclear Reactor Regulation (Post 941001).

92697:069.

January 1997. 31pp. 9702190059, 91804:328.

The individual Plant Examination (IPE) database stores struc-The Nuclear Regulatory Commission is issuing this draft tured information about plant designs, core damage frequency Standard Review Plan to desenbe the procedure used to imple-(CDF) and containment performance. It records the presence or ment the antitrust review and enforcement prescribed in Sec-absence of hardware in each design, characterizes its functional tions 105 and 186 of the Atomic Energy Act of 1954, as amend-dependencies, and relates these features to the CDF and con-ed. This draft SRP reflects cunent regulations and policy, and tainmert performance. The IPE database supports detailed in-will be updated to reflect changes in NRC regulations.

quiries into these characteristics for a specific plant or class of plants. In particular, the IPE database is designed to answer NUREG-1577 DRFT FC: STANDARD REVIEW PLAN ON POWER questions that enable interested parties to compare the CDF REACTOR LICENSEE FINANCIAL QUAtJFICATIONS AND DE-and containment performance of boiling-and pressurized-water COMMISSIONING FUNDING ASSURANCE. Draft Report For reactors (BWRs and PWRs) as a function of their design fea-Comment. WOOD,R.S. Office of Nuclear Reactor Regulation tures, on the basis of information found in the IPE submittals.

(Post 941001). January 1997. 20pp. 9702190062. 91809:334.

To query the IPE database, two programs have been devel-The Nuclear Regulatory Commission is issuing this draft oped. The frst is a self-contained, user friendly, menu-driven Standard Review (SRP) to describe the process it uses to program written in Microsoft's Visual Basic language. This pro-review the financial qualifications and methods of providing de-gram answers the " basic queries" most often asked about the commissioning funding assurance required of power reactor li-IPEs, through a process of sorting records within the IPE data-consees. This draft SRP reflects current regulations and policy.

base. Queries of this type can be improvised on the spot. Other and will be updated to reflect changes in NRC regulations.

" advanced queries" that call for calculations, linking of data NUREG-1601: CHEMICAL PROCESS SAFETY AT FUEL CYCLE files, and ranking or sorting on the basis of calculation can be performed using the programming language within such person-FACILITIES. AYRES.D.A. Division of Fuel Cycle Safety & Safe-al computer data management applications as dBase, Access, guards (Post 930207). August 1997, 26pp. 9708180186.

or Paradox. This IPE database user's guide provides guidance A0077:187.

for formulating basic and advanced queries. The guidance for This NUREG provides broad guidance on chemical safety advanced queries is given in terms of Microsoft Access 2.0.

issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not ex-NUREG-1604: CIRCUMFERENTIAL CRACKING OF STEAM GEN-haustive, for addressing chemical process safety in the safo ERATOR TUBES. KARWOSKI,K.J. Office of Nuclear Reactor storage, handling, and processing of licensed nuclear material.

Regulation (Post 941001). April 1997,171pp. 9705160211.

It expounds to license holders and applicants a general philoso-93024:113.

phy of the role of chemical process safety with respect to NRC-On April 28,1995, the U.S. Nuclear Regulatory Commission licensed materials; sets forth the basic information needed to (NRC) issued Generic Letter (GL) 95-03, "Circumferential properly evaluate chemical process safety; and describes plau-Cracking of Steam Generator Tubes." GL 95-03 was issued to sible methods of identifying and evaluating chemical hazards obtain information needed to verify licensee compliance with ex.

and assessing the adequacy of the chemical safety of the pro-isting regulatory requirements regarding the integrity of steam posed equipment and facilities. Examples of equipment and generator tubes in domestic pressurized-water reactors (PWRs) methods commonly used to prevent and/or mitigate the conse-This report briefly describes the desigra and function of domestic quences of chemical incidents are discussed in this document.

steam generators and summarizes the staffs assessment of the NUREG-1602 DRFT FC: THE USE OF PRA IN RISK-INFORMED "E "***

APPLICATIONS. Draft Rept For Comment.

• Division of Systems servations related to steam generator operating experience.

Technology (Post 941217). June 1997.150pp. 9707140029.

This report is intended to be representative of significant operat-93737:167.

ing experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Oper-In August 1995, the Nuclear Regulatory Commission issued a ating experience prior to April 1995 is discussed throughout the policy statement proposing improved regulatory decisionmaking report, as necessary, for completeness.

"by increasing the use of PRA [probabilistic risk assessment /

analysis) in all regulatory matters to the extent supported by the NUREG 1606 DRFT FC: PROPOSED REGULATORY GUIDANCE I

state-of-the-art in PRA methods and data." To support the im-RELATED TO IMPLEMENTATION OF 10 CFR 50.59 pigmentation of the Commission's policy, regulatory guidance (CHANGES, TESTS, OR EXPERIMENTS). Draft Report For documents have been developed by the staff (as drafts for Comment. MCKENNA,E.M. Office of Nuclear Reactor Regula-J i

public comment) describing how PRA can be used in specific tion (Post 941001). April 1997. 61rp. 9705140368. 92887:102.

regulatory activities, many of which relate to licensee-proposed The Nuclear Regulatory Comrr6sion is issuing this draft g'M-changes to their current licensing basis (CLB). In addition, a ance document for public comt"ent tha ' describes current inter-i more general regulatory guide has been developed which de-pretations related to the proca ss by wt ch power reactor licens-scribes an overall approach to using PRA in risk-informed regu-ees may make certain plant.:hanges sithout prior NRC approv.

I lation. One key aspect of this general guidance is the attributes al. The draft guidance re'.tfirms existing regulatory practice in i

of an acceptable PRA for such regulatory activities. Detailed i

many areas; clanfies the staff's expectations in areas where in-l l

discussion is provided for a full-scope PRA (i.e., a PRA that dustry practice or position differs from the staff's and estaD-considers both internal and external events for all modes of op-lishes guidance in areas where guidance did not previously eration). In addition, discussions are provided for the use and exist.

hmitations of importance measures and sensitivity studies. Final-ty, the subject of peer review of a PRA is also discussed.

NUREG-1607: SAFETY EVALUATION REPORT RELATED TO

Main Citations and Abstracts 11 THE DEPARTMENT OF ENERGY'S PROPOSAL FOR THE IR-cation Services, Mail Stop T-6059, U.S. Nuclear Regulatory RADIATION OF LEAD TEST ASSEMBLIES CONTAINING TRIT-Commission, Washington, DC 20555-0001.

IUM-PRODUCING BURNABLE ABSORBER RODS IN COM-MERCIAL LIGHT-WATER REACTORS.

• Office of Nuclear Re-NUREG-1610: CONTROLLING THE ATOM.The Beginnings Of actor Regulation (Post 941001). May 1997. 78pp. 9706190451.

Nuclear Regulation, 1946-1962. MAZUZAN,G.T.; WALKER,S.

93409:250.

Office of the Secretary of the Commission. May 1997. 558pp.

The NRC staff has reviewed a report, submitted by DOE to 9707220273. 93827:001.

determine whether the use of a commercial light-water reactor Controlling the Atom is a study of the early history of nuclear (CLWR) to irradiate a limited number of tritium-producing burn.

regulation. It focuses on the activities of the U.S. Atomic Energy able absorber rods (TPBARs) in lead test assemblies (LTAs)

Commission (predecessor of the Nuclear Regulatory Commis-raises generic issues involving an unreviewed safety question.

sion), the agency that exercised primary responsibility for safe-The staff has prepared this safety evaluation to address the ac-guarding public health and safety from the hazards of nuclear ceptability of these LTAs in accordance with the provision of 10 power. The book reconstructs the context in which the AEC es-CFR 50.59 without NRC licensing action. As summarized in tablished its regulatory program, weighing the relationship be-Section 10 of this safety evaluation, the staff has identified tween the AEC's regulatory programs and its other major func-issues that require NRC review. The staff has also identified a tions: developing and testing of nuclear weapons and encourag-number of areas in which an individual licensee undertaking irra-ing expanded use of civilian nucler energy. A persistent theme diation of TPBAR LTAs will have to supplement the information is the AEC's effort to ensure adequate protection of public in the DOE report before the staff can determine whether the health and safety without imposing restrictive or infloxible regu-proposed irradiation is ac'eptable at a particular facility. The lations that would impede the growth of the nuclear industry, staff concludes that a licensee undertaking irradiation of TPBAR The book provides detailed accounts of key issues such as li-LTAs in a CLWR will have to submit an application for amend-censing nuclear power reactors, siting of plants, developing ment to its facility operating license before inserting the LTAs standards for radiation protection, and disposing of radioactive into the reactor.

wastes.

NUREG-1608 DRFT FC: CATEGORIZING AND TRANSPORTING NUREG-1611: AGING MANAGEMENT OF NUCLEAR POWER LOW SPECIFIC ACTIVITY MATERIALS AND SURF TAMINATED OBJECTS. Draft Rept For Comment., ACE CON-PLANT CONTAINMENTS FOR LICENSE RENEWAL. LIU,W.C/

Office of KUO,P.T.: LEE S.S. Office of Nuclear Reactor Regulation (Post Nuclear Material Safety & Safeguards. Transportation, Dept' 941001). September 1997. 63pp. 9710100155. A0703:297.

of. June 1997. 60pp. 9707140018. 93738:098-The primary purpose of this guidance is to assist shippers in In 1990, the Nuclear Management and Resources Council preparing low specific activity materials (LSA) and surface con-(NUMARC), now the Nuclear Energy Institute (NEI), submitted taminated objects (SCOs) for shipment in compliance with Fed-for NRC review, the industry reports (irs), NUMARC Report 90-eral regulations. Guidance is provided in question and answer 01 and NUMARC Report 90-10, addressing aging management issues associated with PWR containments and BWR contain-format on he classification, characterization, packaging and transportation of LSA and SCOs, including the definition of LSA ments for license renewal, respectively. In 1996, the Commis-and SCOs, the determination of distribution on of activity in LSA sion amended 10 CFR 50.55a to promulgate requirements for material or on SCO surfaces, mixing LSA and SCOs in a pack-inservice inspection of containment structures. This rule amend-age, radiation level measurements, and various other aspects of ment incorporates by reference the 1992 Edition with the 1992 transporting LSA and SCOs. There are many requirements, Addenda of Subsections IWE and IWL of the ASME Code ad-other than those addressed herein, imposed in the shipment of dressing the inservice inspection of metal containmen%/hrers LSA and SCOs. The guidance represents one or more methods and concrete containments, respectively. The purpse of this of demonstrating compliance wrth the regulatory requirements report is to reconcile the technical information arv) agree nents for LSA material and SCOs that have been found acceptable to resulting from the NUMARC IR reviews which are generally de-NRC staff; however, additional methods may also be found to scribed in NUREG 1557 and the inservice inspection require.

be acceptable with adequate justification. This document is ments of subsections IWE and IWL as promulgated in $c',3.55a being issued for public comment. As a result of the public com-for license renewal consideration. This reoort concluo6s that ments, or internal peer review and discussions, the content of Subsections IWE and IWL as endorsed in Sc50.55a are ge,'eral-the final guidance may be significantly different from that pre.

ly consistent with the technical agreements reached during ther sented in this document.

IR reviews. Specific exceptions are identified and additional evaluations and augmented inspections for renewal are recom-NUREG-1609 DRFT FC: STANDARD REVIEW PLAN FOR mended.

TRANSPORTATION PACKAGES FOR RADIOACTIVE MATERIALDraft Report For Comment.

• Office of Nuclear Ma-NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRI.

l-terial Safety & Safeguards. November 1997, 132pp.

TY DATABASE. FAIRBANKS.C.J.; LEL:,A.D.; MEDOFF,J.; et al.

9801120281, A1739.174.

Office of Nuclear Reactor Regulation (Post 941001) July 1997.

(

The Standard Review Plan for Transportation Packages for 6Spp. 9707140087, 93738:255.

Radioactive Material provides guidance for the review and ap-The U.S. Nuclear Regulatory Commission (NRC) developed proval of applications for packages used to transport radioactive the Reactor Vessel Integrity Database (RVID) following the material (other than irradiated nuclear fuel) under 10 CFR Part staff's review of licensee responses to Generic Letter (GL) 92-l

71. The Standard Review Plan is intended for use by the U.S.

01, Revision 1 (Ref.1). The database summarizes the proper.

l Nuclear Regulatory Commission staff. Its objectives are to (1) ties of the reactor pressure vessel (RPV) beltline materials for l

summarize 10 CFR Part 71 requirements for package approval, each operating commercial nuclear power plant. The RVID con-l (2) describe the procedures by which the NRC staff determines tains four tables for each plant- (1) background information I

that these requirements have been aatisfied, and (3) document table, (2) chemistry data table, (3) upper shelf energy table, and the practices developed by the staff in previous reviews of (4) pressure-temperature limits or pressurized thermal shock package applications. A separate Standard Review Plan for table. References and notes follow each table documenting the Transportation Packages for Spent Nuclear Fuel (NUREG-1617) source (s) of data and presenting supplemental information. Ad-is in preparation. Draft NUREG-1617 is scheduled to be pub-ditionally, the RVID has " sort" and " data search" capabilities.

hshed for comment in the spnng of 1998. Comments, including The user can select a desired grouping of plants and then comments regarding errors or omissions, as well as suggestions specify information categories to search and list. The design of for improvement, should be sent to the Chief, Rules Review and the RVID consolidates the industry's RPV data in a convenient Directives Branch, Division of Freedom of information and Pubh-and accessible manner. Some of the data categones contain

12 Main Citations and Abstracts data inputs of " docketed" information; other data categories NUREG/CP-0154: PROCEEDINGS OF THE CNRA/CSNI WORK-contain computed numerical values, which may or may not be SHOP ON STEAM GENERATOR TUBE INTEGRITY IN NUCLE-

" docketed". The programming logic used for calculations in the AR POWER PLANTS. DIERCKS,D.R. Argonne National Labora-RVID follows the methodology in Regulatory Guide (RG) 1.99, tory. February 1997, 625pp. 9703100272. ANL-96/14.

Revision 2 (Ref. 2). For the Palisades RPV, the data and infor-92026:001.

mation contained in the RVID, Version 1.1, are current through An intemational Workshop on Steam Generator Tube Integrity April 12,1995; the data and information for the RPVs of all in Nuclear Power Plants, sponsored by the Committee on Nu-other operating commercial nuclear power plants are current clear Regulatory Activities (CNRA) and the Committee on the through December 31,1994. The staff will update the RVID pe-Safety of Nuclear installations (CSNI) of the OECD-NEA, was riodcally to reflect the latest information available. Information held at Oak Brook (suburban Chicago), Illinois, on October 30 contained in the industry's responses to the closecut letters to November 2,1995. The USNRC Office of Nuclear Regulatory GL 92 01, Revision 1, and in the industry's responses to GL 92 Research served as host. The objective of the workshop was to 01, Revision 1, Supplement 1 (Ref. 3), are not necessarily re-provide a working forum for the exchange of information by cork flected in this version, but will appear in a future version of the tributing experts on current issues related to PWR steam gener-RVID.

ator tube integnty, One hundred persons from 15 countries at-tended the workshop, including 36 from regulatory and nuclear NUREG-1614 V01: NRC STRATEGIC PLAN. Fiscal Year 1997 -

policy agencies 28 frorr. Nsearch and development laborato-Fiscal Year 2002.

• NRC - No Detailed Affihation Given. Octo-ries,18 from nuclear vendors and consulting firms, and 18 from ber 1997. 39pp. 9712110109. A1387:155.

electrical utilities. The workshop opened with a plenary session; The U.S. Nuclear Regulatory Commission (NRC) has devel-the first part of the session covered international steam genera-oped general goals consistent with its regulatory mission for ci.

tor regulatory practices and issues, featuring speakers from reg-vilian use of byproduct, source, and special nuclear materials to ulatory bodies in Belgium, France, Japan, Spain, and the United ensure adequate protection of the public health and safety, to States. In Part 2 of the plenary session, comprehensive techni-promote the common defense and security, and to protect the cal overviews on steam generator tubing degradation, inspec-environment. This report addresses the strategies for attaining tion, and integrity were presented by authorities in these fields these goals for Fiscal Year 1997 through Fiscal Year 2002.

from the United States, France, and Belgium. Parallel working sessions on the second and third days of the workshop then NUREG-1616: FEASIBILITY OF UNDERWATER WELDING OF developed findings and recommendations in the areas of (1)

HIGHLY IRRADIATED IN. VESSEL COMPONENTS OF BOILING tubing degradation, (2) tubing inspection, (3) tubing integrity, (4)

WATER REACTORS.A Literature Review. LUND,A.L Division of preventative and corrective measures, and (5) operational as.

Engineering Technology (Post 941217). November 1997.46pp.

pects and risk analysis. On the final day of the workshop, the 9711280297. A1248:311.

working-session facilitatory presented summaries of their ses-In February 1997, the U.S. Nuclear Regulatory Commission sions to the workshop attendees.

(NRC), Office of Nuclear Regulatory Research (RES), initiated a NUREG/CP-0155: PROCEEDINGS OF THE SEMINAR ON LEAK literature review to assess the state of underwater welding tech-BEFORE BREAK IN REACTOR PIPING AND VESSELS.

nology. The objective of this literature review was to evaluate FAIDY,C. France. GILLES.PH. FRAMATOME. April 1997.

the viability of underwater welding in-vessel compononts of boil-774pp. 9704240208. 92599:001.

ing water reactor (SWR) in-vossel components, especially those The sixth in a series of intemational Leak-Before-Break (LBB) components fabricated from stainless stee6s that are subjected Seminars was held at Hotel Sofitel in Lyon, France on October to high neutron fluences. This literature review revealed a pre-9 through 11,1995. The seminar updated international policies ponderance of general information about underwater welding and supporting research on LBB. Attendees included represent-technology, as a result of the active research in this field spom atives from regulatory agencies, electric utility representatives, sored by the U.S. Navy and offshore oil and gas industry com fabricators of nuclear power plants, research organizations, and cerns. However, the literature search yielded only a limited academic institutions. The objective of the seminar was to amount of information about underwater welding of components present the current state of the art in LBB methodology devel-in low-fluence areas of BWR in-vessel environments, and no in' opment, validation, and application in an intemational forum.

formation at all conceming underwater welding experiences in With particular emphasis on industrial applications and regula-high-fluence environments. Research reported by the staff of to'y policies, the seminar provided an opportunity to compare the U.S. Department of Energy (DOE) Savannah River Site and approaches, experiences, and codifications developed by differ-researchers from the intemational fuoion reactor program docu-ent countries. The seminar was organized into four topic areas:

mented relevant experience conceming welding of stainless Status of LBB Applications, Technical issues in LBB, Methodol-steel materials in air erwironments exposed to high neutrm ogy, Complementary Requirements (Leak Detection and inspec-fluences. It also addressed problems with welding highly irradi-tion), and LBB Assessment and Margins. In addition to the ated materials, primarily helium-induced cracking in the material, formal sessions where papers were presented by participants and suggested some solutions to these problems, from France, Germany, Japan, Korea, Belgium, the Uniteo King-

)

e us a, w

anada, j

NUREG/CP-0153: PROCEEDINGS OF THE 24TH DOE /NRC NU.

the Netherlands, and the United States, informal LBB poster 1

CLEAR AIR CLEANING AND TREATMENT f

s a we avaHa e se H. As a CONFERENCE. Held in Portland, Oregon, July 15-18, 1996-a

nar, esWes of tN Ms to N WB FIRST,M.W. Harvard School of Public Health Boston, MA' "E

August 1997.1,041pp. 9709120076. CONF-960715. A0347:001.

j me This report contains the papers presented at the 24th DOE /

NRC Nuclear Air Cleaning and Treatment Conference and the NUREG/CP-0157 V01: PROCEEDINGS OF THE TWENTY.

I associated discussions. Major topics are: (1) nuclear air clean-FOURTH WATER REACTOR SAFETY INFORMATION 1

ing issues, (2) waste management, (3) instrumentation and MEETING. Plenary Session, High Burnup Fuel, Containment And I

measurement, (4) testing air and gas cleaning systems, (5)

Structural Aging. MONTELEONE S. Brookhaven National Labo-f progress and challenges in cleaning up Hanford, (6) lntemation-ratory. January 1997. 357pp. 9703120266. 92077:003.

al nuclear programs (7) standardized test methods, (8) HVAC, This three-volume report contains papers presented at the (9) decommissioning (10) computer modeling applications (11)

Twenty-Fourth Water Reactor Safety information Meeting held iodine treatment, (12) filters, and (13) codes and standards for at the Bethesda Marriott Hotel, Bethesda, Maryland, October filters and adsorbers.

21-23, 1996. The papers are printed in the order of their pres.

Main Citations and Abstracts 13 entation in each session and describe progress and results of NUREG/CP-0161: TRANSACTIONS OF THE TWENTY FIFTH programs in nuclear safety research conducted in this country WATER REACTOR SAFETY INFORMATION MEFTING.

and abroad. Foreign participation in the meeting included MONTELEONE,S. Brookhaven National Laboratory. P.d.,mber papers presented by researchers from Finland, France, Japan, 1997.125pp. 9710100229. A0705:001.

Norway, Russia and the Uruted Kingdom, The titles of the This report contains summaries of papers on reactor safety papers and the names of the authors have been updated and research to be presented at the 25th Water Reactor Safety in-may differ from those that appeared in the final program Y be formation Meeting at the Bethesda Marriott Hotel in Bethesda, meeting.

Maryland October 20-22,1997. The summaries briefly describe NUREG/CP-0157 V02: PROCEEDINGS OF THE TWS t Y-the programs and results of nuclear safety research sponsored FOURTH WATER REACTOR SAFETY INFORMArlON by the Office of Nuclear Regulatory Research, U.S. NRC. Sum-MEETING. Reactor Pressure Vessel Embrittlement And Thermal maries of invited papers concerning nuclear safety issues from Annealing. Reactor Vessel Lower Head Integrity And Evaluation U.S. government laboratories, the electric utilities, the nuclear And Projection of Steam Generator tube.... MONTELEONE,S.

industry, and from foreign governments and industry are also in-Brookhaven National Laboratory. February 1997. 444pp.

ciuded. The summanes have been compiled in one report to 9703120278. 92075:001.

provide a basis for meaningful discussion of information ex.

See NUREG/CP-0157,V01 abstract.

changed during the course of the meeting, and are given in rder of their presentation in each session.

NUMEG/CP-0157 V03: PROCEEDINGS OF THE TWENTY.

FOURTH WATER REACTOR SAFETY INFORMATION NUREG/CR-0200 R5V1P1: SCALE: A MODULAR CODE MEETING PRA And HRA. And Probabilistic Seismic Hazard As-SYSTEM FOR PERFORMING STANDARDIZED COMPUTER sessmeni And Seismic Siting Criteria. MONTELEONE,S. Brook-ANALYSES FOR LICENSING EVALUATION. Control Modules haven National Laboratory. February 1997.180pp.9703120285.

C4, C6.

• Oak Ridge National Laboratory. March 1997. 599pp.

92076:081.

9705120302. ORNLNUREGCSD2RS. 92830:001.

See NUREG/CP-0157,V01 abstract.

SCALE-a modular code system for Standardized Computer NUREG/CP-0154: PROCEEDINGS OF THE OECD/CSNI SPE.

Analyses Licensing Evaluation-has been developed by Oak CIALISTS MEETING ON BORON DILUTION REACTIVITY Ridge National Laboratory at the request of the U.S. Nuclear TRANSIENTS. Held in State College, Pennsylvania USA,0ctober Regulatory Commission. The SCALE system utilizes well-estab-18-20, 1995.

• Organization for Economic Cooperation & Devel-lished computer codes and methods within standard analysis opment.

• Pennsylvania State Univ., University Park, PA. June sequences that (1) allow an input format designed for the occa-1997.468pp.9707180208. NEA/CSNI/R(96)3. 93800:001, sional user and/or novice, (2) automated the data processing A CSNI Specialist Meeting on Boron Dilution Reactivity Tran.

and coupling between modulas, and (3) provide accurate and sients was held in State College, Pennsylvania, USA, from Oc.

reliable results. System development has been directed at prc>

tober 18-20,1995. The meeting was sponsored by the United lem-dependent cross-section processing and analysis of critical-States Nucbar Regulatory Commission (USNRC) in collabora, ity safety, shielding, heat transfer, and depletion / decay prob-tson with the Committee on the Safety of Nuclear installation lems. Since the initial release of SCALE in 1980, the code (CSNI) of the OECD Nuclear Energy Agency (NEA) and the system has been heavily used for evaluation of nuclear fuel fa-Pennsylvania State University. The objective of the meeting was cility and package designs. This revision documents Version 4.3 to bring together experts involved in the different activities relat.

of the system.

ed to boron dilution transients, to promote discussion among NUREG/CR-0200 R5V1P2: SCALE: A MODULAR CODE these experts, and to focus on the technicalissues of concem SYSTEM FOR PERFORMING STANDARDIZED COMPUTER in resolving the safety significance of such events' ANALYSES FOR LICENSING EVALUATION. Control Modules NUREG/CP-0159: PROCEEDINGS OF THE OECD/CSNI WORK-S1 - H1,

• Oak Ridge National Laboratory. March 1997. 556pp.

SHOP ON TRANSIENT THERMAL-HYDRAULIC AND NEU-9705120305. ORNLNUREGCSD2RS. 92832:001.

TRONIC CODES REQUIREMENTS. Held in See NUREG/CR-0200,R5,V1,P1 abstract.

Annapolis, Maryland, USA, November 5-8, 1996.

• Organization for Economic Cooperation & Development.

• SCIENTECH, Inc.

NUREG/CR-0200 R5V2P1: SCALE: A MODULAR CODE July 1997,842pp. 9708210006. NEA/CSNI/R(97)4. A0172:001.

SYSTEM FOR PERFORMING STANDARDIZED COMPUTER This is a report on the CSNI Workshop on Transient Thermal.

ANALYSES FOR LICENSING EVALUATION. Functional Modules Hydraulic and Neutronic Codes Requirements held at Annapolis, F1 FB.

• Oak Ridge National Laboratory. March 1997. 705pp.

Maryland, USA, November 5-8, 1996. This experts' meeting 9705120317 ORNLNUREGCSD2RS. 92843:001, consisted of 140 participants from 21 countries; 65 invited See NUREG/CR-0200,RS,V1,P1 abstract, papers were presented. The meeting was divided into five NUREG/CR-0200 RSV2P2: SCALE: A MODULAR CODE areas: (1) current and prospective plans of thermal-hydraulic SYSTEM FOR PERFORMING STANDARDIZED COMPUTER codes development; (2) current and anticipated uses of thermal-hydrauhc codes; (3) advances in modeling of thermal-hydraulic ANALYSE,S FOR LICENSING EVALUATION. Functional Modules pg, pg Oak Ridge National Laboratory. March 1997,832pp.

phenomena and associated additional experimental needs; (4) 9705120321. ORNLNU1EGCSD2RS. 92866:001.

numerical methods in multi-phase flows; and (5) programming See NUREG/CR-0200,R5,V1,P1 abstract.

language, code architectures and user interfaces. The workshop consensus identified the following important action items to be NUREG/CR-0200 RSV2P3: SCALE: A MODULAR CODE addressed by the international community in order to maintain SYSTEM FOR PERFORMING STANDARDIZED COMPUTER and improve the calculational capability preserve current code ANALYSES FOR LICENSING EVALUATION. Functional Modules expertise and institutional memory; - preserve the ability to use F16 F17,

• Oak Ridge National Laboratory. March 1997.

the existing investment in plant transient analysis codes; - main-606pp. 9705120322. ORNLNUREGCSD2RS. 92869:001.

tain essential experimental capabibties; - develop advanced See NUREG/CR-0200,RS,V1,P1 abstract.

measurement capabilities to support future code vahdation work; - integrate existing analytical capabilities so as to irnprove NUREG/CR-0200 RSV3: SCALE: A MODULAR CODE SYSTEM performance and reduce operating costs; exploit the proven FOR PERFORMING STANDARDIZED COMPUTER ANALYSES advances in code architecture, numerics, graphical user inter.

FOR LICENSING EVALUATION. Miscellaneous.

• Oak Ridge faces, and modularization in order to improve code performance National Laboratory. March 1997. 764pp. 9705120311.

and acrutibility, and - more effectively utilize user experience in ORNLNUREGCSD2RS. 92840:001.

rnodifying and improving the codes.

See NUREG/CR-0200,RS V1,P1 abstract.

)

i

14 Main Citations and Abstracts l

NUREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR execute the research tasks are drawn from ORNL with subcon.

NUCLEAR ELECTRICITY-GENERATING UNITS IN THE tract support from universities and other research laboratones.

UNITED STATES: 1997 2001. VANKUlKEN J.C.; GUZIEL,K.A.;

Close contact is maintained with the sister Heavy-Section Steel TOMPKINS,M.M.; et al. Argonne National Laboratory. Septem-Irrad;ation Program at ORNL with related research programs ber 1997. 54pp. 9710100244. ANL-AA-30. A0705:220.

both in the United States and abroad. This report provides an This report updates previous estimates of replacement energy overview of principal developments in each of the seven pro-l costs for potential short-term shutdowns of 109 U.S. nuclear gram tasks from October 1995 March 1996.

electricity units. This information was developed to assist the U.S. Nuclear Regulatory Commission (NRC) in its regulatory NUREG/CR-4409 V06: DATA BASE ON DOSE REDUCTION impact analyses, specifically those that examine the impacts of PROJECTS FOR NUCLEAR POWER PLANTS. KHAN,T.A.;

proposed regulations requiring retrofitting of or safety modifica-XIE,J.W. Brookhaven Nat>onal Laboratory. January 1997,167pp.

tions to nuclear reactors. Such actions might necessitate shut.

9702060168. BNL-NUREG-51934,91659:149.

downs of nuclear power plants while these changes are being This is the sixth volume in a series of reports that provide in-implemented. The change in energy cost represents one factor formation on dose reduction research and health physics tech-that the NRC must consider when deciding to require a particu.

nology for nuclear power plants. The information is taken from tar modification. Cost estimates were derived from probabilistic two of several databases maintained by Brookhaven Nauonal production cost simulations of pooled utility system operations.

Laboratory's ALARA Center for the U.S. Nuclear Regulatory Factors affecting replacement energy costs, such as random Commission. The research section of the report covers dose re-unit failures, maintenance and refueling requirements, and load duction projects that are in the experimental or development variations, are treated in the analysis. This report describes an phase. It includes topics such as need for cost-effective meas-abbreviated analyt6 cal approach as it was adopted to update the ures to control radiation fields, the highly effective full-system cost estimates published in NUREG/CR-4012, Vol. 3. The up-decontamination, progress in addressing the increase in radi-dates were made to extend the time frame of cost estimates ation fields upon switetu g from normal water chemistry to hy-n and to account for recent changes in utility system conditions, drogen water chemistry in BWRs, addition of depleted zine to such as change in fuel prices, construction and retirement reduce radiation fields, and cobalt free wear-resistant alloys.

schedules, and system demand projections.

The section on health physics technology discusses dose re-NUREG/CR 4219 V12 N2: HEAVY-SECTION STEEL TECHNOL, duction efforts that are in place or in the process of being im-OGY PROGRAM. Semiannual Progress Report ror April 1995 plemented at nuclear power plants. A total of 67 new or updat-Through September 1995. PENNELL,W.E. Oak Hidge National ed projects are desenbed. The appendix provides a complete Laboratory. January 1997.98pp.9702070204. ORNL/TM-9593.

listing of all the material in this area, including that from previ-91667:001.

ous reports. The matorial is available,through a fax machine The Heavy Section Stect Technology (HSST) Program is con-from our ACEFAX on-line as' ERR 17*ystem. The procedure for ducted for the Nuclear Regulatory Commission (NRC) by Oak accessing ACEFAX is also described.

Ridge National Laboratory (ORNL). The program's focus is on NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACK.

the development and validation of tochnology for the assess-ING IN LIGHT WATER REACTORS.

Semiannual ment of fracture-prevention margins in commercial nuclear reac-tor vessels. The HSST program is organized in seven tasks: (1)

Report, January 1996 June 1996. CHOPRA.O.K.; CHUNG,H.M/

program management (2) constraint effects analytical develop GAVENDA,D.J.; et al. Argonne National Laboratory. May 1997' 98pp. 9706110115. ANL-97/9. 93318:150 ment and vanlation, (3) evaluation of cladding effects, (4) duc-Thb report summarizes work Performe$1 by Ar9onne National tile-to-cleavage fracture-mode conversion, (5) fracture analysis methods development and application, (6) material property Laboratory on fatigue and environmentally assisted cracking data and test met"ds, and (7) integration of results. The pro-(EAC) in light water reactors from January 1996 to June 1996.

gram tasks have been structured to place emphasis on resolu-Topics that have been investigated include (a) fatigue of tion of fracture mechanics issues with near term licensing sig-carbon, low. alloy, and austenitic stainless steels (SSs) used in nificance. Resources to execute the research tasks are drawn reactor piping and pressure vessels, (b) irradiation-assisted from ORNL with sub-contract support from universities, and stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys other research laboratones. Close contact is maintained with 600 and 690. Fatigue tests were conducted on ferritic and aus-the sister Heavy-Section Steel Irradiation (HSSI) Program at ten, tic SSs in water that contained various concentrations of ORNL and with related research programs both in the United dissolved oxygen (DO) to determine whether a slow strain rate States and abroad. This report provides an overview of principal applied during various portions of a tensile-loading cycle are developments in each of the seven program tasks from April equally effective in decreasing fatigue life. Slow strain-rate-ten-1995 through September 1995.

sile tests were conducted in simulated boiling water reactor (BWR) water at 288 degrees C on SS specimens irradiated to a NUREG/CR-4219 V13 N1: HEAVY SECTION STEEL TECHNOL-low fluence in the Halden reactor and the results were com-OGY PROGRAM. Semiannual Progress Report For October pared with similar data from a control-blade sheath and neu-1995 - March 1996. PENNELL.W.E. Oak Ridge National Labora-tron-absorber tubes irradiated in BWRs to the same fluence tory. September 1997, 95pp. 9710100227, ORNL/TM-9593.

level. Crack-growth-rate tests were conducted on compact-ten-A0705:126.

sion specimens from several heats of Alloys 600 and 690 in air The Heavy-Section Steel Technology (HSST) Program is con-and high-punty, low-DO water.

ducted for the U.S. Nuclear Regulatory Commission (NRC) by the Oak Ridge Nabonal Laboratory (ORNL). The program focus NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACK-is on the developrnent and validation of technology for the as-ING IN LIGHT WATER REACTORS. Semiannual Report. July-sessment of fracture prevention margins in commercial nuclear December 1996. CHOPRA,0.K.; CHUNG H.M.; GAVENDA.D.J.;

reactor pressure vessels. The HSST Program is organized in et al. Argonne National Laboratory. October 1997. 108pp.

seven tasks: (1) program management, (2) constraint effects 9711030083. ANL-97/10. A0990:251.

analytical development and validation, (3) evaluation of cladding This report summarizes work performed by Argonne National effects, (4) ductile to cleavage fracture mode conversion, (5)

Laboratory on fatigue and environmentally assisted cracking fracture analysis methods development and applications, (6)

(EAC) in light water reactors from July 1996 to December 1996.

material property data and test methods, and (7) integration of Topics that have been investigated include (a) fatigue of results into a state-of-the-art methodology. The program tasks carbon, low-alloy, and austenitic stainless steels (SSs) used in have been structured to place emphasis on the resolution frac.

reactor piping and pressure vessels, (b) irradiation-assisted ture issues with near term licensing significance. Resources to stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600,

Mal'n Citations and Abstracts 15 and (d) characterization of residual stresses in welds of boiling INTO NEAR SURFACE LOW LEVEL WASTE DISPOSAL witer reactor (BWR) core shrouds by numerical models. Fatigue UNITS. Final Report On Field Experiments At A Humid Region tests were conducted on ferritic and austenitic SSs in water that Site,Beltsville, Maryland. SCHULZ,R.K. California, Univ. of, Los contained various concentrations of dissolved oxygen to deter-Angeles, CA. RIDKY,R.W. Maryland, Univ, of, College Park, MD.

mine whether a slow strain rate applied during vanous portions O'DONNELL E.

Division of Regulatory Applicahons (Post of a tensile-loading cycle are equally effective in decreasing fa.

941217). September 1997. 31pp. 9711190252. A1149:329.

tigue hfe. Slow-strairt rate-tensile tests were conducted in simu-The project objective was to assess means for controlling Lited BWR water at 288 degrees C on SS specimens irradiated waste infiltration throught waste disposal unit covers in humid to a low fluence in the Halden reactor and the results were regions. Experimental work was carried out in large scale lysi-compared with similar data from a control-blade sheath and meters (70'x45'xlO') at Beltsville, MD and results of the assess-neutron-absorber tubes irradiated in BWRs to the same fluence ment are applicable to disposal of LLW, uranium mill tai!ings, level. Crack-growth-rate tests were conducted on compact-ten-hazardous waste, and sanitary landfills. Three concepts were s!on specimens from a low carbon content heat of Alloy 600 in under investigatiott (1) resistive layer barrier, (2) conductive high-punty oxygenated water at 289 degrees C. Residual layer barrier, and (3) bioengineenng water management. The re-strisses and stress intensity factors were calculated for BWR sistive layer barrier consisted of compacted earth (clay). The core shroud welds, conductive layer barrier was a special case of the capillary bar-rier and it requires a flow layer (e g. fine sandy loam) over a NU7.EQ/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERE capillary break. As long as unsaturated conditions are main-CORE DAMAGE ACCIDENTS: 1995. A Status Report.

tained water is conducted by the ficw layer to below the waste.

BELLES,R.J.; CLETCHER,J.W.; COPINGER,0.A.; et al. Oak This barrier is most efficient at low flow rates and is thus best Ridge National Laboratory. Apnl 1997. 300pp. 9706120307.

placed below a resistive layer barrier. Such a combination of the ORNL/NOAC-232. 93343:041.

resistive layer over the conductive layer bamer promises to be Ten operational events that affocted ten commercial hght-highly effective provided there is no appreciable subsidence.

witer reactors (LWRs) during 1995 that are considered to be Bioengineering water management is a surface cover that is de-precursors to potential severe core damage are described. All signed to accommodate subsidence. it consisted of imperme-of these events had conditional probabilities of subsequent core able panels which enhance run-off and limit infiltration. Vegeta-dImage greater than or equal to 1.0 x 10(-6). These events tion was planted in narrow openings betw3en panels to tran-were identified by computer-screening the 1995 licensee event spire water from below the panels. This system has successfully reports from commercial LWRs to identify those that could be dewatered two lysimeters thus demonstrating that this proce-potential precursors. Candidate precursors were then selected dure could be used for remedial action (" drying out") exisbng waterlogged disposal sites at low cost.

and evaluated in a process similar to that used in previous ts-srssments. Selected events underwent engineering evaluation NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS:

that identified, analyzed, and docurriented the precursors. Other LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM events designated by the Nuclear Regulatory Commission FOR FISCAL YEAR 1996. Annual Report. MCCONNELLJ.W.;

(NRC) also underwent a similar evaluation. Finally, documonted ROGERS,R.D. Idaho National Engineering & Environmental precursors were submitted for review by licensees and NRC Laboratory. SANFORD,W.E.; et al. Oak Ridge Nationa: Labora-stiff to ensure that the plant design and its response to the pre-tory. August 1997. 55pp. 9709150098. INEL-94/0278.

cursor were correctly characterized. This study is a continuation A0362:267.

of earfier work, which evaluated 1969-1981 and 19841994 The Field Lysimeter investigations: Low-Level Waste Data svtnts. The report discusses the general rationale for this Base Development Program, funded by the U.S. Nuclear Regu-study, the selection and documentation of events as precursors, latory Commission, is (a) studying the degradation effects in or-and the estimation of conditional probabilities of subsequent ganic ion-exchange resins caused by radiation, (b) examining severe core damage for events.

the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE 10 CFR 61 using solidified ion-exchange resins, (c) obtaining CORE DAMAGE ACCIDENTS: 1982-83.A Status Report.

performance information on solidified ion-exchange resins in a FORESTER,J.A.; SCHRINER,H.K.; et al. Sandia Nabonal Lab-disposal environment, and (d) determining the condition of liners oratories. MINARICK,J.W. Science Applicabons international used to dispose the ion-exchange resins. During the field test ng Corp. (formerly Science Apphcations, Inc.). April 1997. 515pp.

experiments, both portland type I-11 cement and Dow vinyl ester-9703120352. SAND 97-0807, 93338:001.

styrene waste form samples were tested in lysimeter arrays lo-This study is a continuation of earlier work that evaluated cated at Argonne National Laboratory-East (ANL E) in lilinois 1969-1981 and 19841994 events affecting commercial light-and at Oak Ridge National Laborato>y (ORNL). The study was watsr reactors. One-hundred nine operational events that affect-designed to provide continuous data on nuclide release and cd $1 reactors during 1982 and 1983 and that are considered m vement, as well as environmental conditions, over an ex-t p

se n s en sM @wn aM to be precursors to potential severe core damage are de-scribed. All these events had conditional probabilities of subse-are to be exhumed. This report discusses the plans for removal, quent severe core damage greger than or equal to 1.0 x 10(-6)-

sampling, and analysis of waste form and soil cores from the These events were idenbfied by first computer screening the lysimeters. Results of partition coefficient determinabons are 1982-83 hcensee event reports from commercial light-water re-presented, as well as application of a source term computer code using those coefficients to predict the fysimeter results. A actors to select events that could be precursors to core study of radionuclides-containing colloeds associated with the damage. Candidates underwent engineenng evaluation that leachate waters removed from these lysimeters is described. An identified, analyzed, and documented the precursors. This update of upward migration of radionuclides in the sand-filled ly-report discusas the general rabonale for the study, the selec-simeter at ORNL is included.

tion and documentation of events as precursors, and the esti-mation of conditional probabilities of subsequent severe core NUREG/CR 5591 V07 N1: HEAVY-SECTION STEEL IRRADIA-damage for the events.

TION PROGRAM. Semiannual Progress Report For October 1995 Through March 1996. CORWiN,W.R. Oak Ridge National NURE3/CR 4918 V10: CONTROL OF WATER INFILTRATION Laboratory. April 1997, 63pp. 9705120292. ORNL/TM-11568.

92828:054.

16 Main Citations and Abstracts The goal of the Heavy-Seebon Steel Irradiation Program is to of 10,201 thermoluminescent phosphor elements of 40 micron provide a thorough, quantitative assessment of effects of neu-thickness, covering a 900 CM(2) area. Array substrates are 125 tron irradiation on material behavior, and in particular the frac-micron thick polyimide sheets, enabling them to easily conform ture toughness properties, of typical pressure vessel steels as to regular surface shapes, especially for survey of surfaces that they relate to light-water reactor pressure-vessel integrity. El-are inaccessible for standard survey instruments. The passive, fects of specimen size, material chemistry, product form and mi-integrating radiation detectors are sensitive to alpha and beta crostructure, irradiation fluence, flux, temperature and spectrum, radiation at contaminahon levels below release guideline limits.

and post 4rradiation annealing are being examined on a wido Required coritact times with potentially contaminated surfaces I

range of fracture properties. The HSSI Program is arranged into are under one hour to achieve detection of transuranic _ alpha 14 tasks: (1) program management, (2) fracture toughness emission at 100 dpm/100 cm(2). Positional information otained (K(Ic)) curve shift in high-copper welds, (3) crack arrest tough-from array evaluaton is useful for locating contamination zones.

ness (K(la)) curve shift in high-copper welds, (4) Irradiation ef.

Unique capabilities of this system for survey of sites, facilibes focts on cladding, (5) K(Ic) and K(la) curve shifts in low upper-and material include measurement inside pipes and other geo-shelf welds, (6) anneakng effects in low upper shelf welds, (7) metrical configurations that prevent standard surveys, and irradiation effects in a commercial low upper shelf weld, (8) mi-below-surface measurement of alpha and beta emitters in con-crostructural analysis of irradiation effects, (9) in-service aged taminated soils. These applications imply a reduction of material material evaluations, (10) correlation monitor materials, (11) that must be classified as radioactive waste by virtue of its pos-special technical assistance, (12) JPOR steel examination, (13) sibility of contamination, and cost savings in soil sampling at technical assistance for JCCCNRS Working Groups 3 and 12, contaminated sites.

and (14) additional requirements for rnaterials. This report pro-vides an overview of the actrvities within each of these tasks NUREG/CR-6042 R01: PERSPECTIVES ON REACTOR SAFETY.

from October 1995 Through March 1996.

HASKIN,F.E. New Mexico, Univ. of. Albuquerque, NM.

CAMP,A L Sandia National Laboratories. HODGE,S.A. Oak NUREG/CR-5591 V07 N2: HEAVY-SECTION STEEL IRRADIA-Ridge National Laboratory. November 1997, 714pp.

TlON PROGRAM. Semiannual Progress Report For April 9712230312. SAND 93-0971. A1499:001.

Through September 1996. CORWIN,W.R. Oak Ridge National The U.S. Nuclear Regulatory Commission (NRC) maintains a Laboratory. September 1997, 71pp. 9710070369. ORNL/TM-technical training center at Chattanooga, Tennessee to provide 11568. A0642:202.

appropriate training to both new and experienced NRC employ-The goal of the Heavy Section Steel Irradiation Program is to ees. This document describes a one-week course in reactor provide a thorr' ugh, quantitative assessment of effects of neu-safety concepts. The course consists of five modules: (1) the tron irradiation on material behavior, and in particular the frac-development of safety concepts; (2) severe accident perspec ture toughness propertes, of typical pressure vessel steels as tives; (3) accident progression in the reactor vessel; (4) contain-they relate to light water reac'.or pressure-vessel integrity. Ef' ment characteristics and design bases; and (5) source terms fects of specimen size, material chemistry, product form and mi-and offsite consequences. The course text is accompanied by crostructure, irradiation fluence, flux, temperature and spectrum, slides and videos during the actual presenta*, of the course.

and post-irradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into NUREG/CR-6074 V03: SEALED SOURCE AND DEVICE DESIGN 14 tasks: (1) program management, (2) fracture toughness SAFETY TESTING. Technical Report On The Findings Of Task (K(Ic)) curve shift in high-copper welds, (3) crack-arrest tough-

4. Investigation Of A Failed Brachytherapy Needle Applicator.

ness (K(la)) curve shift in high-copper welds, (4) irradiation ef-LUKEZlCH,S.J. Southwest Research Institute. May 1997.77pp.

fects on cladding, (5) K(Ic) and K(la) curve shifts in low upper-9706110126. 04-4448-012. 93335:225.

shelf welds, (6) annealing effects in low upper-shelf welds, (7)

As a result of an incident in which a radioactive brachyther-irradiation effects in a commercial low upper shelf weld, (8) mi-apy treatment source was temporarily unable to be retracted, an crostructural analysis of irradiation effects, (9) in-service aged analysis was performed on the needle applicator used during material evaluations, (10) correlation monitor materials, (11) the treatment. In this report, the resutts of laboratory evalua-special technical assistance, (12) JPDR steel examination, (13) tions of the physical, mechanical, and metallurgical condition of technical assistance for JCCCNRS Working Groups 3 and 12, the subject applicator and two additional applicators are pre-and (14) additional requirements for materials. This report pro-sented. A kink formed in the subject applicator during the inci-vides an overview of the activities within each of these tasks dent. The laboratory investigation focused on identifying charac-from April Through September 1996-teristics which would increase the susceptibihty of an applicator NUREG/CR-5661: RECOMMENDATIONS FOR PREPARING THE to form a kink when subjected to bending loads. The results ob-CRITICALITY SAFETY EVALUATION OF TRANSPORTATION tained during this invesbgation could not conclusively identify PACKAGES. DYER,H.R.; PARKS,C.V. Oak Ridge National Lab.

the cause of the kink. The subject applicator exhitxted no oratory. April 1997. 50pp. 9705160214. ORNL/TM-11936.

unique features which would have made it particularly suscepti-93024:284.

ble to forming a lunk. The three applicators examined represent This report provides recommendations on preparing the entp two methods of manufacturing. A number of characteristics in-cality safety section of an application for approval of a transpy-herent to the method used to rnanufacture the subject applica-tation package containing fissile material. The analytical ap.

tor which could lead to an increased susceptibihty to the forma-proach to the evaluation is emphasized rather than the perform, tion of a kink were observed. The use of an insertion device, ance standards that the package must meet. Where perform.

such as the biopsy needle used during this incident, could also ance standards are addressed, this report incorporates the re.

dramatically increase the likelihood of the formation of a kink if quirements of 10 CFR Part 71, the applicator is subjected to bending loads.

NUREG/CR-6037: MEASUREMENT OF RESIDUAL RADIOAC-NUREG/CR4153: A SIMPLIFIED MODEL OF DECONTAMINA.

TIVE SURFACE CONTAMINATION BY 2-D LASER HEATED TION BY BWR STEAM SUPPRESSION POOLS. POWERS,D.A.

TLD. JONES,S.C. Keithley instruments, Inc. June 1997.100pp.

Sandia National Laboratories. May 1997. 463pp. 9706120310.

l 9706240042. 93499-001.

93340:001.

l The feasibihty of applying and adapbng a two-dimensional An uncertainty analysis of aerosol removal by nuclear reactor laser heated thermoluminescence dosimetry system to the steam suppression pools is described. Uncertainbes considered problem of surveying for radioactive surface contamination was in the analyses include uncertainties in boundary condrtions dic-studied. The system consists of a CO(2) laser-based reader and tated by accident progression, uncertainties in bubble behavior, monohthic arrays of thin dosimeter elements. The arrays consist and uncertainties in aerosol properties. Uncertainty distribubon

l l

Main Citations and Abstracts 17 1

for decontamination factors, aerosol particle sizes, and the geo-NUREG/CR 6233 V02: STABILITY OF CRACKED PIPE UNDER metric standard deviabon of the size distnbutions are developod SEISMIC / DYNAMIC DISPLACEMENT-CONTROLLED cs functions of suppression pool depth. Results of the uncer.

STRESSES. Subtask 1.2 Final Report. KRAMER,G.; VIETH,P.;

tinty distributran are used to construct a simphfied model of de.

MARSCHALL,C.; et al. Battelle Memorial Institute, Columbus contamination by steam suppression pools.

Laboratories. June 1997. 170pp. 9707140055. BMI-2177.

93737:001.

NUREG/CR-6167: LATE-PHASE MELT PROGRESSION EXPERI-Results of displacement-controlled pipe fracture experiments, MENT MP 2.Results And Analysis.

GASSER,R.D.;

analyses, and material characterization efforts performed within GAUNTT,R.O.; BOURCIER,S.C.; et al. Sandia National Labora, the Intemational Piping integrity Research Group, IPIRG, Pro-tories. May 1997. 275pp. 9707180201. SAND 93-3931.

gram Subtask 1.2 are discussed. Effects of dynamic versus 93801:105.

quasi static and monotonic versus cyclic loading were evaluated A series cf irFpile experiments addressing the phenomenolo-for ductile tearing of two materials, A106 Grade B femtic steel gy associated with Late-Phase processes in Light Water Reac-and TP304 austenatic steel. Twelve through-wall-cracked pipe tors (LWRs) has been performed in the Annular Core Research experiments were conducted on 6-inch diameter Scheduto 120 Reactor (ACRR) at Sandia National Laboratories. The Melt Pro-pipe at 288 C (550 F). The results indicated dynamic loading at gr:ssion (MP) experiments were designed to provide informa-sssmk sVain rams rnarghah increased N bahWng ca-pa ity of austenitic steel. The ferritic steel tested was sensitive tion as part of the effort to develop and verify computer models nan c s ra ag a aseqwnW, s bakaWng ca-for the LWR core dama9e duri"9 severe accidents. The MP-2 pacity decreased at dynamic strain rates. Two parameters were experiment is the second experiment in this sen.es. The MP-2 found to affect the apparent ductile crack growth resistance experiment examine the formation and movement of ceramic during cyclic loading, load ratto (R) and incremental plastic dis-molten pools that form in the disrupted regions of a reactor placement that occurs in a cycle. Cyclic (R = 0) loading had core. The MP-2 experiment assembly consisted of three re-minimal effect on ductile teanng for both materials. However, gions: (1) a rubble bed composed of enriched UO(2) and ZrO(2) fully reversed loading decreased the load-carrying capacity and that simulated the severely disrupted regions of the reactor toughness for both materials. The incremental plastic displace-core, (2) a composite ceramic / metallic crust which represented ment can be as impor1 ant as the load ratio; however, it is harder the blockage formed by the early phase melting, relocation, and to quantify from design stress reports. Large plastic displace-rsfreezing of mostly metallic core components, and (3) an intact ments will minimize the effect of negative load ratios.

j rod stub region that remained in place below the blockage region. The test assembly was fission heated in the central NUREG/CR-6233 V03: CRACK STABILITY IN A REPRESENTA-ccvity of the ACRR at an average rate of -0.2 K/s ultimately TiVE PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC / DYNAMIC DISPLACEMENT-CONTROLLED tchieving a peak temperature in the molten pool of 3400 K.

As ACRR power levels were increased over time, the crust STRESSES. Subtask 1.3 Final Report. SCOTT,P.: OLSON,R.;

WILKOWSKI,G.M.; et al. Battelle Memorial Institute, Columbus gradually remelted and reformed, penetrating into and attacking Laboratories. June 1997. 558pp. 9707140064. BMI-2177.

the ceramic / metallic N~:kage. The metallic components of the 93734:001.

b'cckage region rWd and re6cated downward to the bottom This report presents the results from Subtask 1.3 of the Inter-of the intact rod rd egion. The ceramic pool penetrated half-nabonal Piping Integnty Research Group (IPIRG) program. The w:y into the blockage region at the end of the experiment. Pos-objective of Subtask 1.3 is to develop data to assess analysis IIxperiment examination of the assembly with the associated metho oFgies for characterizing the fracture behavior of cir-material interactions and metallurgy are discussed in detail to-cumfers aially cracked pipe in a representative piping system gether with the analyses and interpretation of the results.

under combined inertial and displacement-controlled stresses. A unique expenmental facility was designed and constructed. The NUREG/CR-6181 Rot: A PILOT APPLICATION OF RISK.IN-piping system evaluated is an expansion loop with over 30 FORMED METHOUS TO ESTABLISH INSERVICE INSPEG-meters of 16-inch diameter Schedule 100 pipe. The experimen-TlON PRIORITIES FOR NUCLEAR COMPONENTS AT SURRY tal facility is equipped with special hardware to ensure system UNIT 1 NUCLF.AR POWER STATION. VO,T.V; PHAN,H.K.;

boundary conditions could be appropriately modeled. The test GORE.B.F.: et al. Battelle Memorial Institute, Pacific Northwest matrix involved one uncracked and five cracked dynamic pipe-N'.tional Laboratory. February 1997, 74pp. 9703100220. PNNL-system experiments. The uncracked experiment was conducted 9020. 92027:244.

to evaluate piping system damping and natural frequency char-As part of the Nondestructive Evatus.on Reliabihty Program acteristics. The cracked-pipe experiments evaluated the fracture sponsorod by the U.S. Nuclear Regult,y Commission, the Pa-behavior, pipe system response, and stabahty characterishes of cific Northwest National Laboratory has developed risk-informed five different materials. All cracked-pipe experiments were con-apprcaches for inservice inspection plans of nuclear power ducted at PWR conditions. Material characterization efforts pro-platus. This method uses probabikstic risk assessment (PRA) vided tensile and fracture toughness propnties of the different results to identify and priontize the most risk-important compo pipe materials at various strain rates and temperatures. Results rants for inspection. The Surry Nucioar Power Station Unit 1 from all pipe-system experiments and material characterization was selected for pilot apphcabon of this methodology. This efforts are presented. Results of fracture mechanics analyses, l

report, which incorporates more recent plant-specific informa-dynamic finite element stress analyses, and stabikty analyses

]

are presented and compared with experimental results.

tion and improved risk informed methodology and tools, is Revi.

sion 1 of the earlier report (NUREG/CR-6181). The methodolo-NUREG/CR-6233 V04: INTERNATIONAL PIPING INTEGRITY gy discussed in the original report is no longer current and a RESEARCH PROGRAM (IPIRG) PROGRAM. Program Final prtferred methodology is presented in this Revision. This report, Report. WILKOWSKI,G.M.; SCHMIDT,R.; SCOTT,P.; et al. Bat-l NUREG/CR-6181, Rev.1, therefore supersedes the earlier telle Memorial Institute, Columbus Laboratories. June 1997.

NUREG/CR-6181 pubhshed in August 1994. The specific sys-320pp. 9707140072. BMI-2177. 93736:001.

{

tems addressed in this report are the auxiliary feedwater, the This is the final report of the internabonal Piping integrity Re-low-pressure injection, and the reactor coolant systems. The re-search Group (IPIRG) Program. The IPIRG Program was an suits provide a risk-informed ranking of components within international group program managed by the U.S. Nuclear Reg-these systems, ulatory Commission and funded by a consortium of organiza-tions from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United

18 Main Citations and Abstracts States. The program objective was to develop data needed to NUREG/CR-6361: CRITICALITY BENCHMARK GUIDE FOR verify engineering methods for assessing the integrity of circum-LIGHT WATER-REACTOR FUEL IN TRANSPORTATION AND ferentially cracked nuclea power plant piping. The primary STORAGE PACKAGES. LICHTENWALTER; BOWMAN,S.M.;

l focus was an expenmental task that investigated the behavior DEHART,M.D.; et al. Oak Ridge National Laboratory. March of circumferentially flawed piping systems subjected to high-rate 1997.358pp.9705120283. ORNL/TM-13211. 92829:001.

loadings typical of seismic events. To accomphsh these objec-This report is designed as a guide for performing cnticality tives a pipe system fabricated as an expansion loop with over benchmark calculations for light-water-reactor (LWR) fuel appli-30 rneters of 16-inch diameter pipe and five long radius elbows cations. The guide provides documentation of 180 criticality ex.

was constructed. Five dynamic, cyclic, flawed piping experi-periments with geometries, materials, and neutron interaction ments were conducted using this facility. This report: (1) pro-characteristics representative of transportation packages con-vides background information on leak-before-break and flaw taining LWR fuel or uranium oxide pellets or powder. These ex.

evaluation procedures for piping. (2) summarizes technical re-periments should benefit the U.S. Nuclear Regulatory Commis-suits of the program, (3) gives a relatively detailed assessment sion (NRC) staff and licensees in validation of computational of the results from the pipe fracture experiments and comple-methods used in LWR fuel storage and transportation concems.

mentary analyses, and (4) summarizes advances in the state-of-The experiments are classified by key parameters such as en-the-art of pipe fracture technology resulting from the IPIRG pro-richment, water / fuel volume, hydrogen-to-fissile ratio (H/X), and gram.

lattice pitch. Groups of experiments with common features such NUREG/CR-6295: REASSESSMENT OF SELECTED FACTORS as separator plates, shielding walls, and soluble borors are also AFFECTING SITING OF NUCLEAR POWER PLANTS.

identified. In addition, a sample validation using these experi-DAVIS.R E.; HANSON,A.L; MUBAYi,V.; et al. Brookhaven Na.

ments and a statistical analysis of the results are provided. Rec-tional Laboratory. February 1997.117pp. 9703170247. BNL-commendations for selecting suitable experiments and determi-NUREG-52442. 92130:197, nation of calculational bias and uncertainty are presented as Brookhaven National Laboratory has performed a series of part of this benchmark guide, prooabilistic consequence assessment calculations for nuclear NUREG/CR-6363: EFFECTS OF THERMAL AGING AND NEU.

reactor siting. This study takes into account recent insights into TRON IRRADIATION ON THE MECHANICAL PROPERTIES OF pevere accident source terms and examines consequences in a THREE-WIRE STAINLESS STEEL WELD OVERLAY CLAD-nsk based format consistent with the quantitative health objec-DING. HAGGAG,F.M.; NANSTAD,R.K. Oak Ridge National Lab-tives (OHOs) of the NRC s Safety Goal Policy. Simplified severe accident source terms developed in this study are based on the oratory. May 1997. 39pp. 9705280200. ORNL/TM-13047.

93 nsk insights of NUREG-1150 and compared to those used in The a aging of three-wire series-arc stainless steel weld earlier studies, particularly the Sandia Siting Study. The results of the present study indicate that both the quantity of radioactiv-overlay cladding at 288 degrees C for 1605 h resulted in an ap-ity released in a severe accident as well as the likelihood of a preciable decrease (16%) in the Charpy V notch (CVN) upper.

release are lower than those predicted in earlier studies. The shelf energy (USE), but the effect on the 41 J transition temper-accident risks using the simplified source terms are examined at ature shift was very sinall (3 degrees C). The combined effect a series of generic plant sites that vary in population distribu-of aging and neutron irradiation at 288 degrees C to a fluence tion, meteorological characteristics, and exclusion boundary dis _

of 5 x 10(19) neutrons /cm(2) (x 1 MeV) was a 22% reduction tances. Sensitivity calculations are performed to evaluate the ef, in the USE and a 29 degrees C shift in the 41 J transition tem-fects of emergency protective action assumptions on the nsk of Perature. The effect of thermal aging on tensile properties was prompt fatality and latont cancers fatality, and population reloca-very small. However, the combined effect of irradiation and tion. The study finds that based on the new source terms, the aging was an increase in the yield strength (6 to 34% at test prompt and latent fatality risks at all generic sites meet the temperawrss from 288 to -125 degrees C) but no apparent OHOs of the NRC's Safety Goal Policy by margins ranging from change in uttamate tensile strength or total elongation. Neutron one to more than three orders of magnitude.

irradiation reduced the initiation fracture toughness (J(Ic)) much more than did thermal aging alone. Irradiation slightly decreased NUREG/CR-6331 R01: ATMOSPHERIC RELATIVE CONCEN-the tearing modulus, but no reduction was caused by thermal TRATIONS IN BUILDING WAKES.

RAMSDELL,J.V.;

aging alone. Other results from tensile, CVN, and fracture SIMONEN.C.A. Battelle Memorial Institute, Pacific Northwest toughness specimens showed that the effects of thermal aging National Laboratory. May 1997.150pp. 9706120318. PNNL-at 288 or 343 degrees C for 20,000 h each were very small and 12521, 93339:152.

similar to those at 288 degrees C for 1605 h. The effects of This report documents the ARCON96 computer code devel-long term thermal exposure time (50,000 h and greater) at 288 oped for the U.S. Nuclear Regutatory Commission Office Of Nu-degrees C will be investigated as the specimens become avail-clear Reactor Regulation for use in control room habitability as-able in 1996 and beyond.

sessments. It includes a user's guide to the code, a description of the technical basis for the code, and a programmer's guide NUREG/CR-6370: BLOCKAGE 2.5 USER'S MANUAL RAO.D.V.;

to the code. The ARCON96 code uses hourly meteorological BRIDEAU,J.; et al. Science & Engineering Associates, Inc.

data and recently developed methods for estimating dispersion BERNAHL,W. Software Edge, Inc... December 1996. 129pp.

in the vicinity of buildings to calculate relative concentrations at 9702060212. SEA 963104010A:3. 91657:126.

control room air intakes that would be exceeded no more than The BLOCKAGE 2.5 code described in this User's Mnnual five percent of the time. These concentrations are calculated for was developed by the United States Nuclear Regulatory Com-averaging periods ranging from or; hour to 30 days in duration.

mission (NRC) as a tool to evaluate licensee compliance with ARCON96 is a revised version of nRCON95, which was devel-NRC Bulletin 96-03, " Potential Plugging of Emergency Core oped for the NRC Office of Nuclear Regulatory Research.

Cooling Suction Strainers by Debris in Boiling Water Reactors".

Changes in the code permit users to simulate releases from As such, BLOCKAGE 2.5 provides a generalized framework into area sources as well as point sources. The method of averaging which a user can input plant specrfic and insulation-specif c data concentrations for periods longer than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> has also been for performing analyses in accordance with Regulatory Guide changed. The change in averaging procedures increases rela-1.82, Rev. 2. This user's manual describes the capabilities of tive concentrations for these averaging penods. In general, the BLOCKAGE 2.5 along with a desenption of the graphics user's increase in concentrations is less than a factor of two. The in-interface provided for data entry. Each input / output dialog is de-crease is greatest for rotatively short averaging periods, for ex-scribed in detail along with special considerations related to de-ample O to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and derninishes as the duration of the aver-veloping and executing BLOCKAGE. Also, several sample prob-aging penod increases.

lems are provided such that user can easily modify them to suit

l I

Main Citations and Abstracts 19 c particular plant of interest. The models used in BLOCKAGE NUREG/CR-6372 V02: RECOMMENDATIONS FOR PROBABl.

2.5 and their validabon are presented in the accompanying LISTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCER.

NUREG/CR-6371. The BLOCKAGE models were designed to TAINTY ND USE OF EXPERTS. Appendices. BUDNITZ,R.J.;

be parametric in nature, allowing the user flexibility to examine APOSTOLAKIS,G.; BOORE.D.M.; et al. Lawrence Livermore Na-the impact of several modeling assumptions and to conduct tional Laboratory. Apnl 1997. 750pp. 9705280212. UCRL-ID-nitivity analyses. As a result, BLOCKAGE 2.5 results are 122160. 93127:001.

b,own to be very sensitive to the user provided input. It is See NUREG/CR-6372,V01 abstract.

therefore strongly recornmended that users become thoroughly familiar with BLOCKAGE models and their limitabons as de.

NUREG/CR-6379: AN IMPROVED CORRELATION PROCEDURE scribed in NUREG/CR-6224.

FOR SUBSIZE AND FULL-SIZE CHARPY IMPACT SPECIMEN DATA. SOKOLOV,M.A.; ALEXANDER.D.J. Oak Ridge Na'ional NUREG/CR-6371: BLOCKAGE 2.5 REFERENCE MANUAL Laboratory. March 1997. 302pp. 9704170020 ORNL-6888.

SHAFFER,C.J.; BRIDEAU,J.; et al. Science & Engineering Asso.

92516:001.

ciates, Inc. BERNAHL,W. Software Edge, Inc... December 1996.

To examine the potential for using subsize Charpy specimens 163pp. 9702060227. SEA 96-3104-A;4. 91654:118.

to evaluate the material properties of vessel materials for life The BLOCKAGE 2.5 code was developed by the United extension, a study was conducted on the behavior of subsize States Nuclear Regulatory Commission (NRC) as a tool to impact specimens of five different geometries. Effects of notch evaluate licensee compliance regarding the design of suction depth, angle, and radius, as well as overall specimen dimen-strainers for emergency core cooling system (ECCS) pumps in sions were determined. Correlations of the transition tempera-boiling water reactors (BWR) as required by NRC Bulletin 96-03, ture determined by the different subsize specimens as com-

" Potential Plugging of Emergency Core Cooling Suction Strain-pared to full-size specimens were evaluated. A new procedure ers by Debns in Boiling Water Reactors." Science and Engi-for transforming data from subsize specimens was developed.

neering Associates, Inc. (SEA) and Software Edge, Inc. (SE) de-NUREG/CR-6389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERI.

veloped this PC-based code. The instructions to effectively use this code to evaluate the potential of debris to sufficiently block MENTS WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT-PIPE LOCATIONS. Final Report. September 1991 - November e pump suction strainer such that a pump could lose NPSH margin was documented in a User's Manual (NRC, NUREG/CR.

1995. SCOTT,P.; OLSON,R.; MARSCHALL C.; et al Battelle Memorial Institute, Columbus Laboratories. February 1997.

6370J. The Reference Manual contains additional information 363pp. 9703170239. BMI.2187. 92129:001.

that supports the use of BLOCKAGE 2.5. It contains desenp-This report presents the results from Task 1 of the Second tions of the analytical models contained in the code, program-mer guides illustrating the structure of the code, and summaries International Piping integrity Research Group (IPIRG-2) pro-of coding verification and model validation exercises that were gram. The rationale for and objective of Task 1 was to build on performed to ensure that the analytical models were correctly the results of the first IPIRG program by evaluating: (1) the frac-coded and applicable to the evaluation of BWR pump suction ture behavior of circumferentially cracked pipe subjected to more complex load histories, such as simulated seismic load strainers. The BLOCKAGE code was developed by SEA and histories; (2) cracks at geometric discontinuities, such as elbow programmed in FORTRAN as a code that can be executed from the DOS level on a PC. A graphical users interface (GUI) was girth welds; (3) smaller circumferential surface cracks, more typ-then developed by SEA to make BLOCKAGE easier to use and ical of those considered in eservice flaw evaluations, subjected to dynamic, cyche load histories; and (4) circumferential to provide graphical output capability. The GUI was pro-through-wall-cracked pipe subjected to dynamic, cyclic load his-grammed in the C language. The bser has the option of execut-tones. As a result of these Task i efforts, it was shown that: (1)

Ing BLOCKAGE 2.5 with the GUI or from the DOS level and the the load-carrying capacity of a cracked pipe subjected to a sim-Users Manual provides instruction for both metnods of execu-ulated seismic load history is no worse than that of a cracked tion.

pipe subjected to the single-frequency excitation evaluated in NUREG/CR 6372 V01: RECOMMENDATIONS FOR PROBABl*

IPIRG-1; (2) cracks at elbow girth welds can be adequately ana-LISTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCER*

fyzed using methods previously developed for cracks in straight TAINTY AND USE OF EXPERTS. Main Report. BUDNITZ,R.J.;

pipe; and Q) analysis methods previously developed and veri-fied for large circumferential surface cracks and circumferential APOSTOLAKIS,G.; BOORE,D.M.; et al. Lawrence Livermore Na-through-wall cracks work equally well for smaller cracks, even tional Laboratory. April 1997. 277pp, 9705280207. UCAL-ID-when subjected to more complex load histones.

122160. 93137:001.

Probabilistic Seismic Hazard Analysis (PSHA) is a methodolo-NUREG/CR-6391: DETONATION CELL SIZE MEASUREMENTS gy that estimates the hkehhood that various levels of earth-IN HIGH TEMPERATURE HYDROGEN. AIR STEAM MIXTURES quake-caused ground motion will be exceeded at a given loca-AT THE BNL HIGH-TEMPERATURE COMBUSTION FACILITY.

tion in a given future time period. Due to large uncertainties in CICCARELLI.G.; GINSBERG,T.; BOCCIO,J.L; et al. Brookhaven til the geosciences data and in their modeling, multiple model National Laboratory. November 1997. 84pp. 9712230301. BNL-interpretations are often possible. This leads to disagreement NUREG 52482. A1501:253.

cmong experts, which in the past has led to disagreement on The High-Temperature Combustion Facihty (HTCF) was de-the selection of ground motion for design at a given site. The signed and constructed with the objectrve of studying detona-Senior Seismic Hazards Analysis Committee (SSHAC) reviewed tion phenomena in mixtures of hydrogen-air-steam at initially past studies, including the Lawrence Livermore National Labora-high temperatures. The central element of the HTCF is a 27-cm tory and the EPRI landmark PSHA studies of the 1980's and ex-inner diameter and 21.3-m long cylindrical test vessel capable

c. mined ways to improve on the present state-of-the-art. The of being heated to 700K xx 14K. A unique feature of the HTCF Committee's most important conclusion is that differences in is the "diaphragmless" acetylene-oxygen gas driver which is l

PSHA results are due to procedural rather than technical differ-used to inst: ate the detonation in the test gas. Cell size meas-l ences. Thus, in addition to providing a detailed documentation urements in hydrogen-air-steam mixtures have shown that in-on state-of-the-art elements of a PSHA, this report provides a creasing the initial mixture temperature, in the range of 300K to i

series of procedural recommendations. The role of experts is 650K, while maintaining the initial pressure of 0.1 MPa de-snalyzed in detail. Two entities are formally defined the Tech-creases the cell size and thus makes the mixture more detona-l nical Integrator (TI) and the Technical Facilitatory Integrator (FI) -

ble. Increasing the steam dilution increases the cell size, irre-to account for the various levels of complexity in the technical spective of initial temperature. It is also observed that the do bsues and different levels of efforts needed in a given study.

sensitizing effect of steam diminished with increased initial tem-

20 Main Citations and Abstracts perature. A one-dimensional, steady-state Zel'dovich, von Neu-plant operators shut down the plant before the planned expo-mann, Doring model, with full chemical kinetics, has beon used sure was reached. The exposure of these specimens produced to predict cell size for hydrogen-air-steam mixtures at different no significant irradiation-induced embrittlement.

initial conditions.

NUREG/CR-6400: HUMAN FACTORS ENGINEERING (HFE) IN-NUREG/CR 6393: INTEGRATED SYSTEM VALIDATION: METH' SIGHTS FOR ADVANCED REACTORS BASED UPON OPER-ODOLOGY AND REVIEW CRITERIA. O'HARA J.; STUBLER,W.;

ATING EXPERIENCE. HIGGINS.J.; NASTA.K. Brookhaven Na-HIGGINSJ.; et al. Brookhaven National Laboratory. January tional Laboratory. January 1997. 61pp. 9704100181, BNL-1997.116pp.9702060198. BNL-NUREG-52483. 91655:231.

NUREG-52485. 92416:244.

The U.S. Nuclear Regulatory Commission reviews the human The NRC Human Factors Engineering Program Review Model factors engineering (HFE) aspects of advanced nuclear power (HFE PRM, NUREG-0711) was developed to support a design plant designs. In order to support the advanced reactor design process review for advanced reactor design cerbfication under certification rewews, the HFE Program Review Model was de-10CFRS2. The HFE PRM defines ten fundamental elements of velopod. The model describes the HFE program elements that a human factors engineering program. An Operating Experience are necessary and sufficient to develop an acceptable detailed Review (OER) is one of these elements. The main purpose of design and provides the review criteria for their evaluation. One an OER is to identify potential safety issues from operating of the review elements is verification and validation. The pur-plant experience and ensure that they are addressed in a new pose of this document is to discuss the detailed methodological design. Broad-based experience reviews have typically been considerations necessary for a review of an HFE integrated performed in the past by reactor designers. For the HFE PRM, system validation. A conceptual approach, or paradigm, to inte-the intent is to have a more focussed OER that concentrates on grated system validation is presented which identifes important HFE issues or experience that would be relevant to the human-validation principles and their relationships. The validation para-system interface (HSI) design process for new advanced reac-digm was used to identify the methodological aspects of the tors. This document provides a detailed list of HFE-relevant op-validation process that are needed to meet the general para-erating experience pertinent to the HSI design process for ad-digm requirements. The methodology must support a logical vanced nuclear powor plants. This document is intended to be and defensible inference to be made from validation tests to used by NRC reviewers as part of the HFE PRM review process predicted integrated system performance under actual operating in determining the completeness of an OER performed by an conoitions. The validation paradigm is based upon four general applicant for advanced reactor design certification.

forms of validity system representation, performance represen-tation, test design and statistical conclusion validity. Validating NUREG/CR-6404: AN EXPERIMENTAL SCALE-MODEL STUDY an integrated system is based on establishing that these four OF SEISMIC RESPONSE OF AN UNDERGROUND OPENING types of validity are satisfied. Such assessments are made by IN JOINTED ROCK MASS. KANA,D.D.; FOX,0.J.; HSIUNG,S.;

reviewing the methodology used to conduct validation tests.

et al. Center for Nuclear Waste Regulatory Analyses. February Methodological factors relevant to each of the aspects of validi-1997,200pp. 9704250163. CNWRA 95-012. 92625:095.

ty are discussed.

This report describes an experimental investigation conducted NUREG/CR-6397: RADIATION SAFETY CONCERNS FOR by the Center for Nuclear Waste Regulatory Analyses (CNWRA)

PREGNANT OR BREA*T4EEDING PATIENTS.The Positions to (i) obtain a better understanding of the seismic response of Of The NCRP And The ICRP. MEINHOLD,C.B. Brookhaven Na.

an underground opening in a highly-fractured and jointed rock tional Laboratory. January 1997, 23pp. 9702070213. BNL-mass and (ii) generate a data set that can be used to evaluate NUREG-52484. 91666:316.

the capabilities (analytical methods) to calculate such response For many years, protecting the fetus has been a concern of This report describes the design and implementation of simulat-the National Council on Radiation Protection and Measurements ed seismic experiments and results for a 1/15 scale model of a (NCRP) and the International Commission on Radiological Pro-jointed rock mass with a circular tunnel in the middle. The dis-tection (ICRP). Early recommendations focused on the possibili.

cussion on the design of the scale model includes a description j

ty of a wide variety of detrimental developmental effects while of the associated similitude theory, physical design rationale, j

later recommendations focused on the potential for severe model material development, preliminary analytical evaluation, mental retardation and/or reduction in the intelligence quotient instrumentation design and calibration, and model assembly and (1.0.). The latest recommendations also note that the risk of pretest procedures. The thrust of this discussion is intended to cancer for the fetus is probably two to three times greater per provide the information necessary to understand the expenmen-Sv than in the adult. For all these reasons, the NCRP and the tal setup and to provide the background necessary to under-ICRP have provided guidance to physicians on taking all rea-stand the experimental results. The discussion on the experi-sonable steps to ascertain whether any woman requiring a radi.

mental procedures and results includes the seismic input test ological or nuclear medicine procedure is pregnant or nursing a procedures, test runs, and measured excitation and response child. The NCRP and the ICRP also advise the clinician to post.

time histories. The closure of the tunnel due to various levels of pone such procedures until after delivery or cessation of nuts.

seismic activity is presented. A threshold level of seismic input ing,if possible.

amplitude was required before significant rock mass motion oc-curred. The experiment, though designed as a two-dimensional NUREG/CR-6399: RESULTS OF CHARPY V-NOTCH tMPACT representation of a rock mass, behaved in a somewhat three-TESTING OF STRUCTURAL STEEL SPECIMENS IRRADIATED dimensional manner, which will have an effect on subsequent AT 30 DEGREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A analytical model comparison.

COMMERCIAL REACTOR CAVITY.

ISKANDEP.S.K.;

' STOLLER,R.E. Oak Ridge National Laboratory. April 1997.

NUREG/CR-6414: PIPING BENCHMARK PROBLEMS FOR THE i

51pp. 9705120288. ORNL-6886. 92828:001.

WESTINGHOUSE AP600 STANDARDIZED PLANT, BEZLEN.P.;

)

A capsule containing Charpy V-notch (CVN) and mini-tensile DEGRASSI,G.; BRAVERMAN.J.; et al. Brookhaven National specimens was irradiated at - 30 degrees C (~ 85 degrees F)

Laboratory. January 1997. 300pp. 9702250218. BNL-NUREG-in the cavity of a cciwscie; nuclear power plant to a fluence, 52487. 91871:001, of 1 x 10 (16) neutrons /cm(2) p 1 MeV). The capsule included To satisfy the need for verification of the computer programs six CVN impact specimens of archival High Flux Isotope Reac-and modeling techniques that will be used to perform the final tor A212 grade 8 ferritic steel and five CVN impact specimens piping analyses for the Westinghouse AP600 Standardized of a well-studied A36 structural steet This irradiation was part of Plant, three benchmark problems were developed. The prob.

the ongoing study of neutron-induced darnap? cffects at the low tems are representative piping systems subjected to representa-temperature and flux experienced by reasor supports. The tive dynamic loads with solutons developed using the methods

Main Citations and Abstracts 21 TYPICAL REACTOR CONTAINMENTS SUBJECTED TO being proposed for analysis for the AP600 standard design. It SEVERE ACCIDENT CONDITIONS.

KLAMERUS.E.W.;

will be required that the combined licensees demonstrate that BOHN,M.P. Sandia National Laboratories. WESLEY,0.A.; et al.

their solutions to these problems are in agreement with the EOE Engineering Consultants (formerly EOE Engineering, Inc.).

benchmark problem set.

December 1996.

125pp.

9702060245. SAND 96-2445.

91657:001.

NUREG/CR-6426 Voi: DUCTILE FRACTURE TOUGHNESS OF In SECY-90-016, the NRC proposed a safety goal of a condi-MODIFIED A 302 GRADE B PLATE MATERIALS, DATA ANALY.

tional containment failure probability (CCFP) of 0.1 and the ab SIS. MCCABE.D.E.; MANNESCHMIDT.E.; SWAIN,R.L Oak ternative acceptance enteria allowed for steel containments, Ridge National Laboratory, January 1997, 86pp. 9702190023.

which specifies that the stresses should not exceed ASME ORNL-6892, 91802:192.

Level C allowables for severe accident pressures and tempera-The objective of this work was to develop ductile fracture tures. In this work, the need for an equivalent entenon for con-toughness data in the form of J-R curves for modified A 302 crete containments was studied. Six surrogate containments grade B plate materials typical of those used in fabricating reac-were designed and analyzed in order to compare the margins for pressure vessels. A previous experimental study at Materials between design pressure, pressure resulting in exceedance of Engineering Associates, Lanham, Maryland, on one particular Level C (or yield) stress limits, and ultimate pressure. For com-heat of A 302 grade B plate showed decreasing J-R curves with paraW, ead WM Ms an Wnkal Wnal Ame a

s s

ana ss ns increased specimen thickness. This characteristic has not been to yield are comparable and display a similar margin for both observed in numerous tests made on the more recent produc-steel and concrete containments. In addition, the margin to fail-tion materials of A 533 grade B and A 508 class 2 pressure ure, although slightly higher in the steel containments, were also vessel steels it was unknown if the departure from norm for the comparable. Finally, a CCFP for code design was determined MEA material was a generic characteristic for all heats of A 302 basod on general membrane behavior and imposing an upper grade B steels or just unique to that one particular plate-bound severe accident curve developed in the DCH studies.

NUREG/CR 6426 V02: DUCTILE FRACTURE TOUGHNESS OF The resulting CCFP's were less then 0.02 (or 2%) for all the MODIFIED A 302 GRADE B PLATE MATERIALS. Data Records surrogate containments studied, showing that these contain-MCCABE,D.E.; MANNESCHMIDT,E.; SWAIN R.L Oak Ridge ment designs all achieved the NRC safety goal.

Nat>onal Laboratory. February 1997. 600pp. 9703200279.

NUREG/CR 6437: FLOW AND TRANSPORT AT THE LAS CRUCES TRENCH SITE: EXPERtMENT 11 8. VINSON,J.;

ORNL-6892. 92194:001.

The objective of this work was to develop ductile fracture HILLS.R.G.; et al. New Mexico State Univ., Las Cruces, NM.

toughness data in the form of J-R ctrves for modified A 302 WlERENGA,P.J. Arizona, Univ. of Tucson, AZ. July 1997.

grade B plate materials typical of thor, used in fabricating reac-234pp. 9708210009. A0142:001.

tor pressure vessels. A previous expenmental study at Materials Three waW How and solute transpod spenments were per-Engineering Associates (MEA) on one particular heat of A 302 formed as part of a comprehensive field trench study near Las grade B plate showed decreasing J-R curves with increased Cruces, New Mexico to test deterministic and stochastic models specimen thickness. This characteristic has not beon observed of vadose zone now aM transk Ms r@od presents Mal in numerous tests made on the more recent production materb results from the third experiment (experiment lib). Experiments als of A 533 grade B and A 508 class 2 pressure vessel steels.

lla and b were conducted on the North side of the trench, on a 11 was unknown if the departure from norm for the MEA material plot 1.22 m wide by 12 m long, perpendicular to the trench. The was a generic characteristic for all heats of A 302 grade D area was drip irrigated during two time periods with water con-steels or just unique to that one particular plate. Seven heats of taining a vanety of tracers. The water front was measured with modified A 302 grade B steel and one heat of vintage A 533 tensrometers and neutron probes. Solute fronts were deter-mined from soil solutions through suction Samplers and from grade B steel were provided to this project by the General Elec-disturbed samples. Experiment lib results show predominantly tric Company of San Jose, California. All plates were tested for downward water movement through the layered unsaturated chemical content, tensile properties, Charpy transition tempera-soil. Tritium plumes were only half as deep and half as wide as ture curves, drop-weight nibductility transrtion (NDT) tempera-the water plumes at 310 days after the start of the experiment.

ture, and J-R curves. Tensile tests were made in the three prin-Chromium, applied as Cr(VI), mneo Nmilar to tritium, but with a cipal orientations and at four temperatures, ranging from room loss of mass due to reduction of Cr(Q to Cr(lit). Chloride and temperature to 550 degrees F (288 degrees C). Charpy V-notch nitrate, initially present at high concent stions in the soil solu-transition temperature curves were obtained in longitudinal, tion, were displaced by the irrigation wa er. The extensive data transverse, and short transverse orientations. J-R curves were presented should serve well as a data lase for model testing.

rnade using four specimen sizes (1/2T, IT, 27, and 4T). The NUREG/CR-6446: FRACTURE TOUGHNESS EVALUATIONS OF fracture mechanses-based evaluation method covered three test TP304 STAINLESS STEEL PIPES.

RUDLAND,D.L; onentations and three test temperatures (180,400, and 550 de-grees F (82, 204, and 288 degrees C)]. However, the coverage BRUST,F.W.; WILKOWSKl,G.M. Battelle Memorial institute, Co-of these variables was contingent upon the amount of material lumbus Laboratones. February 1997.116pp. 9703100252. BMI-2194.92062:225.

provided. Drop-weight NDT temperature was determined for the in the IPIRG 1 program, the J-R curve calculated for a 16-T-L orientation only. None of the seven heats of roodified A 302 inch nominal diameter, Schedule 100 TP304 stainless steel grade B showed size effects of any consequence on the J-R (DP2 A8) surface-cracked pipe experiment (Experiment 1.3-3) curve behavior. Crack orientation effects were present, but none was considerably lower than the quasLstatic, monotonic J-R were severe enough to be reported as atypical. A test tempera-curve calcutated from a C(T) specimen (AB-12a). The results ture increase from 180 to 550 degrees F (82 to 288 degrees C) from several related investigations conducted to determine the produced the usual loss in J-R curve fracture toughness. Gener-cause of the observed toughness difference are: (1) Chemical ic J-R curves and mathematical curve fits to the same were analyses on sections of Pipe DP2-A8 from several surface-generated to represent each heat of material. Volume 1 deals cracked pipe and material property specimen fracture surfaces with evaluation of data and discussion of technical findings. This indicate that there are two distinct heats of material within Pipe volume (Volume 2) is a compilation of all data developed.

DP2-A8 that differ in chemical composition. (2) SEN(T) speci-men experimental results indicate that the toughness of a sur-NUREQ/CR-6433: CONTAINMENT PERFORMANCE OF PROTO-face-cracked specimen is highly dependent on the depth of the

g 22 Main Citations and Abstracts initial crack. In additon, the J-R curves from the SEN(T) speci-mens closely match the J-R curve from the surface-cracked ings to provide a forum for nuclear piping experts from around pipe experiment. (3) C(T) expenmental results suggest that the world to exchange information on the subject of pipe frac-ture technology.

there is a large difference in the quasi-static, monotonic tough.

ness between the two heats of DP2-A8, as well as a toughness NUREG/CR-6454: POOL CRITICAL ASSEMBLY PRESSURE degradation in the lower toughness heat of material (DP2-A8tt)

VESSEL FACILITY BENCHMARK. REMEC,L; KAM,F.B.K. Oak when loaded with a dyr'amic, cyclic (R = -0.3) loading history.

Ridge National Laboratory. July 1997. 52pp. 9708210013.

NUREG/CR4448 V02: EVALUATION OF NATIONAL SEISMO-ORNL/TM-13205. A0141:274.

GRAPH NETWORK CETECTION CAPABILITIES. Final Report.

The pool entical assembly (PCA) pressure vessel wall facihty MCLAUGHLIN,K.L.; BARKER T.G.; BENNETT T.J. Affiliation Not benchmark (PCA benchmark) is described and analyzed in this Assigned. October 1997. 52pp. 9711140027. A1106:251*

report. Analysis of the PCA benchmark can be used for partial This final report presents detection thresholds, detection fulfillment of the requirements for the qualification of the meth-probabilities, and location error ethpse projections for the United odology for pressure vessel neutron fluence calculations, as re-States National Seismic Network (USNSN) with and without quired by the U.S. Nuclear Regulatory Commission regulatory real-time cooperative stations in the eastern United States. Net-guide DG-1053. Section 1 of this report describes the PCA work simulation methods are used with spectral noise levels to benchmark and provides all data necessary for the benchmark simulate the processt's of excitaten, propagaton, detection, analysis. The measured quantities, to be compared with the cal-and processing of seismic phases. The USNSN alone should be culated values, are the equivalent fission fluxes. In Section 2 capable of detecting 4 or more P weves for shallow crustal the analysis of the PCA benchmark is described. Calculations earthquakes in nearly alt of the eastern and central United with the computer code DORT, based on the discrete-ordinates States at the nugnitde 3.8 level. When real-bme cooperative method, were performed for three ENDF/8-VI-based multigroup stations are incbded, the network chould be capable of detect-libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent ing 4 or more P waves from events 0.2 to 0.3 magnitude units agreement of the calculated (C) and measures (M) equivalent lower. The planned expansion of the USNSN and cooperative fission fluxos was obtained. The arithmetic average C/M for all statens should improve detection levels by an additional 0.2 to the dosimeters (total of 31) was 0.931 0.03 and 0.92 2 0.03 0.3 magnitudes units in many areas. Locaten uncertainties for for the SAILOR-95 and BUGLE-96 libraries, respectively. The the USNSN should be significant!y improved by addition of real-average C/M ratio, obtained with the BUGLE 93 library, for the time coopera9ve stations. Median error ellipses for magnitude 28 measurements was 0.93 + 0.03 (the neptunium measure-4.5 earthquakes in the eastern and central U.S. depend strongly ments in the water and air regions were overpredictad and ex-upon location but should be less than 100 square km in the cluded from the average). No systematic decrease in the C/M central U.S. and degrade to 200 square km or more off-shore ratios with increasing distance from the core was observed for and south and north of the international boundanes. Close co-any of the hbraries used, operation wth the Canadian National Network should substan-tially improve detection and location along the Canadian border.

NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE PRESSURIZED WATER REACTOR FEEDWATER NUREG/CR-0451: A SAFETY AND REGULATORY ASSESS.

MENT OF GENERIC BWR AND PWR PERMANENTLY SHUT.

NOZZLE, PIPING, AND FEEDRING CRACKING AND WALL DOWN NUCLEAR POWER PLANTS. TRAVIS,R.J.; DAVIS,R.E.;THINNING. SHAH,V.N.; WARE,A.G.; PORTER,A.M. Idaho Na-GROVE,EJ.; et al. Brookhaven National Laboratory. August tional Engineering & Environmental Laboratory. March 1997, 1997.57pp.9708080190. BNL-NUREG-52498. 94731:089.

190pp. 9704170076. INEL-96/0089. 92531:091 An evabation of the nuclear power plant regulatory basis is Review of industry efforts to manage thermal fatigue, flow-ac-celerated corrosion, and steam generator water hammer performed, as it pertains to those plants that are permanently damage to Pressurized Water Reactor (PWR) feedwater noz-shutdown (PSD) and awaiting or undergoing decommissioning.

Four spent fuel storage configurations are examined Recom, zies, piping, and feedrings is presented in this report. The review includes an evaluaten of design modifications, operating msndations are provided for those operationally based regula.

procedure changes, augmented inspection and monitoring pro-tons that could be partially or totally removed for PSD plants grams, and mitigaton, repair and replacement activities. Four without irapacting public health and safety.

specific actions were taken to perform the evaluation (a) review NUREG/CR-6452: THE SECOND INTERNATIONAL PIPING IN-of field experience to identify trends of operating events; (b)

TEGRITY RESEARCH GROUP (IPIRG.2) PROGRAM. Final review of the related technical literature; (c) visits to three PWR Report. ' HOPPER,A.; WILKOWSKI,G.M.; SCOTT,P.; et al. Bat-plants and a PWR vendor; and (d) sohcitation of information telle Memorial Institute, Columbus Laboratones. March 1997, from foreign utilities. Our assessment of field expenence indi-292pp.14704080384. BMI-2195. 92387:001.

cates the USNRC hcensees have apparently taken sufficient The IPIRG-2 program was an international group program acton to minimize the feedwater nozzle cracking caused by managed by the U.S. NRC and funded by organizations from 15 thermal fatigue, wall thinning of J-tubes and feedwater piping, nations. The emphasis of the IPIRG 2 program was the devel-and steam generator water hammer in both top-feed and pre-opment of data to venfy fracture analyses for cracked pipes and heat steam generators. A major finding of this review is that the fittings subjected to dynamic /cyche load histories typical of seis-analysis, inspection, monitonng, mitigation, and replacement mic events. The scope included: (1) the study of more complex techniques have been developed for managing thermal fatigue dynamk/ cyclic load histories, i.e., multi-frequency, variable am-and flow accelerated corrosion damage to feedwater nozzles, plitude, simulated seismic excitations, than those considered in piping, and feednngs. Adequate training and appropriate appli-the iPIRG-1 program, (2) crack sizes more typical of those cork cations of these techniques would ensure effective manage-sidered in Leak Before-Break (LBB) and in-service flaw evalua-ment of this damage. Several PWR plant operators have been tons, (3) through-wall-cracked pipe experiments which can be proactive in managing this damage.

used to validate LBB-type tracture analyses, (4) cracks in and around pipe fittmgs, such as elbows, and (5) laboratory speci-NUREG/CR-6459: FIELD STUDIES AT THE APACHE LEAP RE-men and separate effect pipe experiments to provide better in-SEARCH SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL sight into the effects of dynamic and cyclic load histones. Also MODELS. BASSETT,R.L.; NEUMAN,S.P.; WlERENGA,P.J.; et at.

undertaken were an uncertainty analysis to identify the issues Arizona, Urw. of, Tucson, AZ. August 1997. 200pp.

most important for LBB or in-service flaw evaluations, updating 9708290222. A0232:073.

computer codes and databases, the development and conduct This is a final technical report for a project of the U.S. Nucle-of a senes of round-robin analyses, and analyst's group meet-Regulatory Commission (sponsored contract at 00090-051) with The University of Arizona. The contract was an optional ex-

Main Citations and Abstracts 21 3

being proposed for analysis for the AP600 standard design. It TYPICAL REACTOR CONTAINMENTS SUBJECTED TO will be required that the combined licensees demonstrate that SEVERE ACCIDENT CONDITIONS.

KLAMERUS,E.W.;

their solutions to these prcNems are in agreement with the BOHN,M.P. Sandia Nabonal Laboratories. WESLEY,D.A.; et al.

benchmark problem set.

EQE Engineering Consultants (formerly EQE Engineenng, Inc.).

December 1996.

125pp.

9702060245. SAND 96-2445.

Nue,cG/CR-6426 V0 4: DUCTILE FRACTURE TOUGHNESS OF 91657:001, MODIFIED A 302 GAADE B PLATE MATERIALS, DATA ANALY-In SECY-90-016, the NRC proposed a safety goal of a condi-SIS. MCCABE.D.E.; MANNESCHMIDT,E.; SWAIN,RL Oak tional containment failure probability (CCFP) of 0.1 and the al-Ridge National Laboratory. January 1997. 86pp. 9702190023.

temative acceptance criteria allowed for steel containments, ORNL-6892. 91802:192.

which soecifies that the stresses should not exceed ASME The objective of this work was to develop ductile fracture Level C allowables for severe accident pressures and tempera-toughness data in the form of J-R curves for modified A 302 tures. In this work, the need for an equivalent criterion for con-grade B plate materials typical of those used in fabricating reac-crete containments was studied. Six surrogate containments tor pressure vessels. A previous experimental study at Materials were designed and analyzed in order to compare the margins Engineering Associates, Lanham, Maryland, on one particular between design pressure, pressure resulting in exceedance of heat of A 302 grade B plate showed decreasing J-R curves with Level C (or yield) stress limits, and ultimate pressure. For com-increased specimen thickness. This characteristic has not been araW, ead mWmM has an Wnbcal Wernal volume a

su a a ss hd ma@ns observed in numerous tests made on the more recent produc-to yield are comparable and display a similar margin for both tion materials of A 533 9rade B and A 508 class 2 P' essure steel and concrete containments. In addition, the margin to fail-vessel steels. It was unknown if the departure from norm for the ure, although slightly higher in the steel containments, were also MEA material was a generic characteristic for all heats of A 302 comparable. Finally, a CCFP for code design was determined grade B steels or just unique to that one particular plate.

based on general membrane behavior and imposing an upper bound severe accident curve developed in the DCH studies.

NUREG/CR-6426 V02: DUCTILE FRACTURE TOUGHNESS OF The resulting CCFP's were less then 0.02 (or 2%) for all the MODIFIED A 302 GRADE B PLATE MATERIALS. Data Records.

surrogate containments studied, showing that these contain-MCCABE.D.E.; MANNESCHMIDT,E.; SWAIN,R.L. Oak Ridge ment designs all achieved the NRC safety goal.

National Laboratory. February M97. 600pp. 9703200279.

ORNL-6892. 92194:001.

NUREG/CR-6437: FLOW AND TRANSPORT AT THE LAS The objective of this work was to develop ductile fracture CRUCES TRENCH SITE: EXPERIMENT 11 8. V!NSON J.;

toughness data in the form of J-R curves for modified A 302 HILLS,R.G.; et al. New Mexico State Univ., Las Cruces, NM.

grade B plate materials typical of those used in fabricating reac-WlERENGA,P.J. Arizona, Univ. of, Tucson, AZ. July 1997, tor pressure vessels. A previous expenmental study at Materials 234pp. 9708210009. A0142:001.

Engineering Associates (MEA) on one particular heat of A 302 Three water flow and solute transport experiments were per-grade B plate showed decreasing J-R curves with increased formed a part of a comprehensive field trench study near las Cruces New Mexico to test deterministic and stochastic models specimen thickness. This characteristic has not been observed of va se zom a

rans s mport preseds padal in numerous tests made on the rnore recent production materi-results from the tbrd experiment (experiment lib). Experiments als of A 533 grade B and A 508 class 2 pressure vessel steels.

Ila and b were conducted on the North side of the trench, on a It was unknown if the departure from norm for the MEA material plot 1.22 m wide by 12 m long, perpendicular to the trench. The was a generic characteristic for all heats of A 302 grade B area was drip irrigated during two time periods with water con-steels or just unique to that one parbcular plate. Seven heats of taining a variety of tracers. The water front was measured with modified A 302 grade B steel and one heat of vintage A 533 tensiometer and neutron probes. Solute fronts were deter-grade B steel were provided to this project by the General Elec-mined from soil solutions through sucton samplers and from inc Company of San Jose, Califomia. All plates were tested for disturbed samples. Experiment lib results show predominantly chemical content, tensile properties, Charpy transition tempera-downward water movement through the layered unsaturated ture curves, drop-weight nil-ductility transition (NDT) tempera-soil. Tritium plumes were only half as deep and half as wide as ture, and J-R curves. Tensile tests were made in the three pnn-the water plumes at 310 days after the start of the experiment.

cipal orientations and at four temperatures, ranging from room Chromium, applied as Cr(VI), moved similar to tritium, but with a temperature to 550 degrees F (288 degrees C). Charpy V-notch loss of mast due to reduction of Cr(VI) to Cr(lll). Chlorios and transition temperature curves were obtained in longitudinal, nitrate, initially present at high concentrations in the soil solu-transverse, and short transverse orientations. J-R curves were tion, were displaced by the irrigation water. The extensive data made using four specimen sizes (1/2T, IT 2T, and 4T). The presented should serve well as a data base for model testing.

fracture mechantes-based evaluation method covered three test NUREG/CR-6446: FRACTURE 10UGHNESS EVALUATIONS OF onentations and three test temperatures [180,400, and 550 de-TP304 STAINLESS STEEL PIPES.

RUDLAND.D L.;

grees F (82,204, and 288 degrees C)). However, the coverage BRUST,F.W.; WILKOWSKI,G.M. Battelle Memorial institute, Co-of these variables was contingent upon the amount of material lumbus Laboratories. February 1997.116pp. 9703100252. BMI-provided. Drop-weight NDT temperature was determined for the 2194. 92062:225.

T L orientation only. None of the seven heats of modified A 302 in the IPIRG-1 program, the J-R curve calculated for a 16-grade B showed size effects of any consequence on the J-R inch nominal diameter, Schedule 100 TP304 stainless steel curve behavior. Crack orientation effects were present, but none (DP2-A8) surface-cracked pipe experiment (Experiment 1.3-3) were severe enough to be reported as atypical. A test tempera-was considerably lower than the quasi-static, monotonic J-R ture increase from 180 to 550 degrees F (82 to 288 degrecs C) curve calculated from a C(T) specimen (A8-12a). The results produced the usual loss in J-R curve fracture toughness. Gener-from several related investigations conducted to determine the ic J-R curves and mathematical curve fits to the same were cause of the observed toughness difference are: (1) Chemical generated to represent each heat of material. Volume 1 deals anaPyses on sections of Pipe DP2-A8 from several surface-with evaluation of data and discussion of technical findings. This cracked pipe and material property specimen fracture surfaces volume (Volume 2) is a compilation of a? data developed.

indicate that there are two distinct heats of material within Pipe DP2-A8 that differ in chemical composition. (2) SEN(T) speci-NUREG/CR-6433: CONTAINMENT PERFORMNCE OF PROTO-men experimental results indicate that the toughness of a sur-face-cracked specimen is highly dependent on the depth of the

22 Main Citations and Abstracts initial crack. In addition, the J-R curves from the SEN(T) speci-ings to provide a forum for nuclear piping experts from around mens closely match the J-R curve from the surface-cracked the world to exchange information on the subject of pipe frac-pipe experiment. (3) C(T) experimental results suggest that ture technology.

there is a large difference in the quasi-static, monotonic tough-ness between the two heats of DP2-AB, as well as a toughness NUREG/CR-6454: POOL CRITICAL ASSEMBLY PRESSURE degradation in the lower toughness heat of material (DP2-A811)

VESSEL FACILITY BENCHMARK. REMEC,l.: KAM F.B.K. Oak when loaded with a dynamic, cyclic (R = -0.3) loading history.

Ridge National Laboratory. July 1997. 52po. 9708210013.

ORNL/TM-13205. A0141:274.

NUREG/CR 6448 V02: EVALUATION OF NATIONAL SEISMO-The pool entical assembty (PCA) pressure vessel wall facility GRAPH NETWORK DETECTION CAPABILITIES. Final Report.

benchmark (PCA benchmark) is described and analyzed in this MCLAUGHLIN,K.L; BARKER,T.G.; BENNETT,T.J. Affiliation Not report. Analysis of the PCA benchmark can be used for partial Assigne'1 October 1997,52pp. 9711140027, A1106:251.

fulfillment of the requirements for the qualification of the meth-This final report presents detection thresholds, detection odology for pressure vessel neutron fluence calculations, as re-probabilities, and location error ellipse projections for the United quired by the U.S. Nuclear Regulatory Commission regulatory States National Seismic Network (USNSN) with and without guide DG-1053. Section 1 of this report describes the PCA real-time cooperative stations in the eastern United States. Net-benchmark and provides all data necessary for the benchmark work simulation methods are used with spectral noise levels to analysis. The measured quantities, to be compared with the cal-simulate the processes of excitation, propagation, detectson, culated values, are the equivalent fission fluxes. In Section 2 and processing of seismic phases. The USNSN alone should be the analysis of the PCA benchmark is described. Calculations capable of detecting 4 or more P waves for shallow crustal with the computer code DORT, based on the discrete-ordinates earthquakes in nearly all of the eastern and central United method, were performed for three ENDF/B-Vl-based multigroup States at the magnitude 3.8 level. When real-time cooperative libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent stations are included, the network should be capable of detect-agreement of the calculated (C) and measures (M) equivalent ing 4 or rnore P waves from events 0.2 to 0.3 magnitude units fission fluxes was obtained. The arithmetic average C/M for all lawer. The planned expansion of the USNSN and cooperative the dosimeters (total of 31) was 0.931 0.03 and 0.921 0.03 stations should improve detection levels by an additional 0.2 to for the SAILOR-95 and BUGLE-96 libraries, respectively. The 0.3 magnitudes uruts in many areas. Localion uncertainties for average C/M ratio, obtained with the BUGLE-93 library, for the the USNSN should be significantly improved by addition of real-28 measurements was 0.93 + 0.03 (the neptunium measure-time cooperative stations. Median error ellipses for magnitude ments in the water and air rehons were overpredicted and ex-4.5 earthquakes in the eastern and central U.S. depend strongly cluded from the average). No systematic decrease in the C/M upon location but should be less than 100 square km in the ratios with increasing distance from the core was observed for central U.S. and degrade to 200 square km or more off shore any of the libranes used.

and south and north of the international boundaries. Close co-operation with the Canadian National Network should substan-NUREG/CR-6456: REVIEW OF INDUSTRY EFFORTS TO tially improve detection and location along the Canadian border.

MANAGE PRESSURIZED WATER REACTOR FEEDWATER NUREG/CR-6451: A SAFETY AND REGULATORY ASSESS-NOZZLE, PIPING, AND FEEDRING CRACKING AND WALL MENT OF GENERIC BWR AND PWR PERMANENTLY SHUT-TP" ' 9G. SHAH,V.N.; WARE,A.G.; PORTER,A.M. Idaho Na-DOWN NUCLEAR POWER PLANTS. TRAVIS.R.J.; DAVIS,R.E.;

tior Engineering & Environmental Laboratory. March 1997, GROVE.E.J.; et al. Brookhaven National Laboratory. August 190pp 9704170078. INEL-96/0089. 92531:091.

1997.57pp.9708080190. BNL NUREG-52498. 94731:089.

Review of industry efforts to manage thermal fatigue, flow-ac-celerated corrosion, and steam generator water hammer An evaluation of the nuclear power plant regulatory basis le performed, as it pertains to those plants that are permanently damage to Pressurized Water Reactor (PWR) feedwater noz-shutdown (PSD) and awaiting or undergoing decommissioning.

zies, piping, and feedrings is presented in this report. The Four spent fuel storage configurations are examined. Recom-review includes an evaluation of design modifications, operating mendations are provided for those operatier'al!y based regula, procedure changes, augmented inspection and monitoring pro-tions that could be partially or totally remov 1 for PSD plants grams, and mitigation, repair and replacement activities. Four without impacting public health and safety, specific actions were taken to perform the evaluation (a) review of field experience to identify trends of operating events; (b)

NUREG/CR-6452: THE SECOND INTERNATIONAL PIPING IN-review of the related technical literature; (c) visits to three PWR TEGRITY RESEARCH GROUP (IPIRG-2) PROGRAM. Final plants and a PWR vendor; and (d) solicitation of information Report. HOPPER.A.; WILKOWSKI,G.M.; SCOTT,P.; et al. Bat-from foreign utilities. Our assessment of field experience indi-tolle Memorial Institute, Columbus Laboratories. March 1997.

cates the USNRC licensees have apparently taken sufficient 292pp. 9704080384. BMI-2195. 92387:001.

action to minimize the feedwater nozzle cracking caused by The IPIRG-2 program was an international group program thermal fatigue, wall thinning of J-tubes and feedwater piping, managed by the U.S. NRC and funded by organizations from 15 and steam generator water hammer in both top-feed and pre-nations. The emphasis of the IPIRG-2 program was the devel-heat steam generators. A major finding of this review is that the opment of data to venfy fracture analyses for cracked pipes and analysis, inspection, monitoring, mitigation, and replacement fittings subjected to dynamic / cyclic load histories typical of seis-techniques have been developed for managing thermal fatigue mic events. The scope included: (1) the study of more complex and flow-accelerated corrosion damage to feedwater nozzles, dynamic / cyclic load histories, i.e., multi-frequency, variable am-piping, and feedrings. Adequate training and appropriate appli-plitude, simulated seismic excitations, than those considered in cations of these techniques would ensure effective manage-the IPIRG-1 program, (2) crack sizes more typical of those corb ment of this damage. Several PWR plant operators have been sidered in Leak-Before-Break (LBB) and irkservice flaw evalua-proactive in managing this damage.

tions (3) through-wall-cracked pipe experiments which can be used to validate LBB-type fracture analyses, (4) cracks in and NUREG/CR-6459: FIELD STUDIES AT THE APACHE LEAP RE-around pipe fittings, such as elbows, and (5) laboratory speci-SEARCH SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL men 'nd separate effect pipe experiments to provide better irw MODELS. BASSETT,R.L.; NEUMAN,S.P.; WlERENGA,P.J.; et I

sight, M the effects of dynamic and cyclic load histories. Also al. Artzona, Univ. of, Tucson, AZ. August 1997. 200pp.

underta. 9 were an uncertainty analysis to identify the issues 9708290222. A0232:073.

most impi ant for LBB or in-service flaw evaluations, updating This is a final technical report for a project of t!',e U.S. Nucle-computer codes and databases, the development and conduct at Regulatory Commission (sponsored contract 04 090-051) of a series of round-robin analyses, and analyst's group meet.

with The University of Arizona. The contract was an optional ex-l l

l

l 1

Main Citations and Abstracts 23 tension for the penod July 12,1994 to May 31,1995. The into a scale model of either the Zion or Surry NPP. The results project manager is Thomas J. Nicholson, Office of Nuclear Reg-from the Zion and Surry experiments were extrapolated to other ulatory Research. The objectives of this contract are to examine Westinghouse plants. This report describes tests performed with hypotheses and test alternative conceptual models conceming Combustion Engineering plant geometries (in particular, Calvert unsaturated flow and transport through fractured rock and to Cliffs-like) and the impact of codispersed water as part of the design and execute confirmatory field and laboratory experi-overall DCH issue resolution. Integral effects tests were per-ments to test these hypothesis and conceptual models at the formed with a 1/10th scale model of the Calvert Cliffs NPP Apache Leap Research Site near Superior, Arizona. Each chap-inside the Surtsey test vessel The experiments investigated the ter in this progress report summarizes research related to a effects of codispersal of water, steam, and molten core simulant specific set of objectives and can be read and interpreted as a materials on DCH loajs under prototypic accident conditions separate entity. The tasks include detection and characterize-and plant configurations. The results indicated that large tion of historical rapid flow through fractured rock and the rela-amounts of coejected water reduced the DCH load by a small tionship to perched water systems using environment isotopic amount. Large amounts of debris were dispersed from the tracers of (3)H and (14)C, fluid and rock derived (234)U/(238)U cavity to the upper dorne (via the annular gap).

measurements, and geophysical data. The water balance in a small watershed at the ALRS demonstrates the methods of ac.

NUREG/CR-6474: PRELIMINARY PHENOMENA IDENTIFICA-counting for ET, and estimating the quantity of water available TION AND RANKING TABLES (PIRT) FOR SBWR STARTUP for infiltration through fracture networks. Grain density measure.

STABILITY, ROHATGI,U.S.; CHENG,H.S.; KHAN H.J.; 6t al.

ments are now possible for core-sized samples using a newly Brookhaven National Laboratory. March 1997. 81pp.

designed gas pycnometer. The distribution and magnitude of air 9703200285. BNL-N'JREG-52504. 92196:001.

permeability rneasurements have been done in a three-dimen.

Phenomena iden4fication and Ranking Tables (PIRT) have sinnal setting and subsequent geostatistical analysis is p,esent.

been developed for a start up transient for SBWR. The informa-ed. Electronic data sets of the data presented here are avail.

tion und f~ PIRT came from RAMONA-48 and TRACG analy-able from the authors more detailed discussion and analyses ses of the transielt and from resed small scale tests. The are available in the referenced technical publications.

transient was divded into four distinct phases, namely, Sub-Cooled Core Heat up, Subcooled Chimney, Saturated Chimney, NUREG/CR-6463 R01: REVIEW GUIDELINES FOR SOFTWARE and Power Ascer:sion. The assessment enterion selected was LANGUAGES FOR USE IN NUCLEAR POWER PLANT Minimum Cntical Power Ratio. The SBWR system was divided SAFETY SYSTEMS. Final Report. HECHT,M.; DECKER,D.;

into ten componants. A total of 35 distinct phenomena among GRAFF.S.; et al. SoHaR, Inc. October 1997. 588pp.

the components were identified. The Phase I has 28 ranked 9711140044. A1105:027.

phenomena with 17 low, 6 medium, and 5 high ranking. The Guidelines for the programming and auditing of software writ-Phase 11 has 39 ranked phenomena with 18 low,13 medium ten in high level languages for safety systems are presented.

and 8 high rar. king. The Phase ill has 47 ranked phenomena The guidelines are derived from a framework of issues signifi" with 22 low,10 medium and 15 high ranking. The Phase IV has cant to software safety whicti was gathered from relevant 46 ranked phenomena with 16 low,12 medium and 18 high standards and research literature. Language-specific adapta-ranking.

tions of these guidelines are provided for the following high-level languages: Ada C/C+ +, Programmable Logic Controller NUREG/CR-6478: MOTOR-OPERATED VALVE (MOV) ACTUA-(PLC) Ladder Logic, International Electrotechnical Commission TOR MOTOR AND GEARBOX TESTING. DEWALL,K.G.;

(IEC) Standard 11313 Sequential Function Charts, Pascal, and WATKINS,J.C.; BRAMWELL.D. Idaho National Engineenng &

PL/M. Appendices to the report include a tabular summary of Erwironmental Laboratory. July 1997. 54pp. 9708210034. INEL-the guidelines and additional information on selected languages.

96/0219. A0141:217.

NUREG/CR-6464: AN EVALUATION OF METHODOLOGY FOR Researchers at the Idaho National Engineenng and Environ-SEISMIC QUALIFICATION OF EQUIPMENT, CABLE TRAYS, mental Laboratory tested the performance of electric motors AND DUCTS IN ALWR PLANTS BY USE OF EXPERIENCE and actuator gearboxes typical of the equipment installed on DATA. BANDYOPADHYAY,K; KANA,0.D.; KENNEDY,R.P.; et al.

motor-operated valves used in nuclear power plants. Using a Brookhaven National Laboratory. July 1997.

140pp.

test stand that simulates valve closure loads against flow and 9708040210. BNL-NUREG-52500, 93995:001, pressure, we tested five electric motors (four ac and one de)

Advanced Reactor Corporation (ARC) has developed a meth-and three gearboxes at conditions a motor might experience in odology for seismic qualification of equipment, cable trays, and a power plant, including such off-normal conditions as operation i

docts in Advanced Ught Water Reactor plants. A Panel (mem-at high temperature and reduced voltage. We also monitored i

bers of which acted as individuals) supported by the Office of the efficiency of the actuator gearbox. All five motors operated Nuclear Regulatory Research of the Nuclear Regulatory Com-at or above their rated starting torque during tests at normal vol-rnission has evaluated this methodology. The review approach tages aid temperatures. For all five motors, actual torque end observations are included in this report. In general, the losses cue to voltage degradation were greater than the losses Panel supports the ARC methodology with some exceptions calculated by methods typically used for predicting motor torque (nd provides recommendations for further improvements.

at degraded voltage conditions. For the de motor the actual torque losses due to elevated operating temperatures were NUREG/CR-6469: EXPERIMENTS TO INVESTIGATE DIRECT greater than the losses calculated by the typical predictrve CONTAINMENT HEATING PHENOMENA WITH SCALED methoct. The actual efficiencies of the actuator gearboxes were MODELS OF THE CALVERT CLIFFS NUCLEAR POWER generally lower than the running efficiencies published by the PLANT. BLANCHAT,T.K.; PILCH,M.M.; ALLEN,M.D. Sandia Na-manufacturer and were generally nearer the published pull-out tional Laboratories. February 1997, 195pp. 9703170250.

efficiencies. Operation of the gearbox at elevated temperature SAND 96-2280. 92130:001.

did not affect the operating efficiency.

The Surtsey Test Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments for the Nuclear NUREG/CR-6481 V01: REVIEW OF MODELS USED FOR DE-Regulatory Commission (NRC) that simulate High Pressure Melt TERMINING CONSEQUENCES OF UF(6) i Ejection (HPME) accidents in a nuclear power plant (NPP).

RELEASE. Development Of Model Evaluation Cnteria.

These experiments are designed to investigate the effects of NAIR,S.K.; CHAMBERS,D.B.; PARK,S.H.; et al.

November direct containment heating (DCH) phenomena on the contain-1997. 51pp. 9712230285. A1501:144.

ment load. In previous experiments, hsgh-temperature, chemical-The objective of this study is to examine the usefulness and ly reactive (thermatic) melt was ejected by high-pressure steam effectiveness of currently existing models that simulate the re-

I 1

24 Main Citations and Abstracts 1

l i

lease of UF(6) from UF(6)-handling facilities,9ubsequent reac-applied to concrete-filled steel structural modules. The program i

tions of UF(6) with atmospheric moisture, and the dispersion of was conducted in three phases. The objective of the first phase UF(6) and reaction products in the atmosphere. The study eval-was to identify the technical issues and the need for further i

uates screening-level and detailed public-domain models that study in order to support NRC licensing review activities. The were specifically developed for UF(6) and models that were two key findings were the need for supplementary review crite-originally developed for the treatment of danse gasea but are ria to augment the Standard Review Plan and the need for veri-applicable to UF(6) release, reaction, and dspersion. The model fied design / analysis methodology for unique types of modules, evaluation process is divided into three specific tasks: model-such as the concrete-filled steel module. In the second phase component evaluation, applicability evaluation, and user inter-of this program, Modular Construction Review Criteria were de-face and Quality Assurance and Quality Control (QA/OC) eval-veloped to provide guidance for licensing reviews. In the third uation. Within the model-component ovaluation process, a hase, an analysis effort was conducted to determine if current-model's treatments of source term, thermodynamics, and at.

Ws techniques can be used to i

mosphenc dispersion are considered anti comparisons of model ly available finite element anal predictions with observations are made Within the applicability predict the response of concrete-filled steel modules.

evaluation process, a model's applicabdity to Integrated Safety NUREG/CR-6493: DOSES TO THE HAND DURING THE ADMIN-Analysis (ISA), Emergency Response Planning (ERP), and Post-ISTRATION OF RADIOLABELED ANTIBODIES CONTAINING Accident Analysis (PAA), and to site-specific considerations are assessed. Finally, within the user interface and QA/OC evalua.

Y-90,TC-99M,1-131, AND LU-177. BARBER,D.E. Minnesota, tion process, a model's user-friendliness, presence and clarity Univ. of. Minneapolis, MN. CARSTEN,A.L.; KAURIN,0.G.L; et of documentation, ease of use, etc. ace assessed along with its al. Brookhaven National Laboratory. February 1997. 60pp.

handling of OA/OC. This document presents the complete 9703100224. BNL-NUREG-52510. 92035:235.

methodology used in the evaluation process.

Exposure of the hands of medical personnel administering re-diolabeled antibodies (RABS) was evaluated on the basis of (a)

NUREG/CR-6481 V02: REVIEW OF MODELS USED FOR DE-TERMINING CONSEQUENCES OF UF(6) RELEASE.Model observing and photo-documenting administration techniques.

Evaluation Report. NAIR,S.K.; CHAMBERS,0.B.; PARK,S.H.; et and (b) experimental data on doses to thermoluminescent dos 4 al.. November 1997. 212pp. 9712230298. A1498:139.

meters (TLDs) on fingers of phantom hands holding syringes, Three uranium hexafluoride- (UF(6)-) specific rTodels and on syringe 4, with radionuclides in the syringes in each case.

HGSYSTEM/UF(6) SAIC, and RTM-96; three denso-ges Dose rate coefficients to the skin, if in contact with the syrirago models-DEGADIS, SLAB, and the Chlorine Institute rnethodolo-wall, were 89,1.9, 3.8, and 0.41 uSv s(-1) averaged over 1 gy; and one toxic chemical model-AFTOX-are evaluated on CM(2) at 7 mg CM(-2) per 37 MBq (1 mci) for Y-90, Tc-99m, I-their capabikties to simulate the chemical reactions, thermody-131, and Lu-177, respectively. When using Y-90 the imporance namics, and atmospheric dispersion of UF(6) released from ac-of avoiding direct contact with syringes containing RABs and of cidents at nuclear fuel-cycle facihties, in support of integrated using a beta-particle shield on the synnge was indictted. In Safety Analyds, Emergency Planning, and Post-Accident Analy-using a syringe for injection, doses can best be approximated sis. The models are also evaluated for user-friendliness and for for the geometry studied by (a) wearing a finger dosirneter on quality assuranca and quality control features, to ensure the va-the middle finger, toward the outside of the hand, on the hand lidity and credibility of the resufts from the models. Model per-operating the plunger, and (b) wearing finger dosimetsrs on the formance evaluations are conducted for the three UF(6)- specif-inner (palm) side of the finger on the hand that supports the sy-ic models, using field data on releases of UF(6) and other hoavy ringe for energetic beta-particle emitters, such as Y 90 and Re-gases. Predictions from the HGSYSTEM/UF(6) and SAIC I OO' models are within an order of magnitude of the field data, but the SAIC model overpredicts beyond an order of magnitude for NUREG/CR-6497: DATA COLLECTION AND FIELD EXPERI-a few UF(6)- specific data points. The RTM-96 model provides MENTS AT THE APACHE LEAP RESEARCH SITE.May 1995 -

1996. BASSETT,R.L; NEUMAN,S.P.; WlERENGA P.J.; et al. Ari-m he F

set, er RM model severely underpredicts the observations within 200 m of zona, Univ. of, Tucson, AZ. August 1997.144pp. 9709150105.

A0063:188.

the source. Outputs of the models are most sensitive to the me-teorological parameters at large distances close to the source.

This report documents the research performed during the Specific recommendations have been made to improve the ap.

period May 1995-May 1996 for a project of the U.S. Nuclear placability and usefulness of the three models and for the choice Regulatory Commission (sponsored contract NRC-04-090-051) of a specific model to support the intended analyses. Guidance by the University of Arizona. The project manager for this re-is provided on the choice of input parameters for initial dilution, search is Thomas J. Nicholson, Office of Nuclear Regulatory building wake effects, and distance to completion of UF(8) reac-Research. The objectives of this research were to examine hy-tion with water.

potheses and test alternative conceptual models concerning un-NUREG/CR-6486: ASSESSMENT OF MODULAR CONSTRUC-saturated flow and transport through fractured rock, and to TION FOR SAFETY-RELATED STRUCTURES AT ADVANCED design and execute confirmatory fiMd and laboratory experi-NUCLEAR POWER PLANTS. BRAVERMAN.J.; MORANTE,R.;

ments to test these hypotheses and conceptual models at the HOFMAYER,C. Brookhaven National Laboratory. March 1997.

Apache Leap Research Site near Superior, Arizona. Each chap-201pp.9704170099. BNL-NUREG-52520. 92518:014.

ter in this report summarizes research related to a specific set Modular construction techniques have been successfully used of objectives and can be read and interpreted as a separate in a number of industries, both domestically and internationally, entity. Topics include: crosshole pneumate and gaseous tracer Recently, the use of structural modules has been proposed for field and modehng experiments designed to help validate the advanced nuclear power plants. The objective in utilizing modu.

applicability of continuum geostatistical and stochastic cord I

tar construction is to reduce the construction schedule, reduce cepts, theories, models, and scaling relations relevant to un-l construction costs, and improve the quality of construction. This saturated flow and transport; use of geochemistry and aquifer 4

report documents the results of a program which evaluated the testing to evaluate fracture flow and perching mechanisms; in.

I I

proposed use of modular construction for safety-related struc-vestigations of uranium isotopes to evaluate teaching selectivity; tures in advanced nuclear power plant designs. The program in-and transport and modeling of both conservative and non-cork cluded review of current modular construction technology, de-servative tracers.

velopment of licensing review criteria for modular construction, and initial vahdation of currently available analytical techniques

r Main Citations and Abstracts 25 NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICALITY WJREG/CR-6507: CRITICAL HEAT FLUX (CHF) PHENOMENON SLIDE RULE. Technical Basis.

BROADHF.AD,0.L; ON A DOWNWARD FACING CURVED SURFACE.

HOPPER,C.M.; CHILDS.R.L; et al. Oak Ridge National Labora.

CHEUNG,F.B.; HADDAD,K.H.; LIU,Y.C. Pennsylvarn State tory. April 1997. 95pp. 9705090043. ORNL/TM 13322.

Univ., University Park, PA. June 1997.171pp. 9706a0256.

92826:233.

PSU/ME-97 7321. 93422:007.

In January 1974, a limited distribution report, entitled "A Slide This report describes a theoretical and experimental s%dy of Rule for Estimating Nuclear Cnticality information," was written the boundary layer boiling and entical heat flux phenomena on a by C.M. Hopper for the Oak Ridge Y-12 Plant as a tool for downward facing curved heating surface, including both hemi-emergency response to nuclear enticality accidents. Because of spherical and toroidal surfaces. A subscale boundary layer boil-several shortcomings of the original slide rule, work began re.

ing (SBLB) test facility was developed to measure the spatial cently to update the slide rule using modem computational variation of the entical heat flux and observe the underlying tools. Volume 1 of this report describes the analyses performed mechanisms. Transient quenching and steady-state boiling ex-in support of this updated slide-rule tool and includes a sample, periments were performed in the SBLB facility under both satu-nonfunctioning version of the new shde rule. Volume 2 contains rated and subcooled conditions to obtain a complete database the functional version of the slide rule. The new slide-rule too!

on the critical heat flux. To compiernent the experimental effort, provides capabilities for the continued updating of accident in-an advanced hydrodynamic CHF model was developed from the formation during the evolution of emergency response, including conservation laws along with sound physical arguments. The victim exposure informatiort potential exposures to emergency model provides s clear physical explanation for the spatial varia-reentry personnel; estimates of future radiation fields; and fis-tion of the CHF observed in the SBLB experiments and for the sion-yield estimates.

weak dependence of the CHF data on the physical size of the vessel. Based upon the CHF model, a scaling law was estab-NUREG/CR-6505 V01: THE POTENTIAL FOR CRITICALITY FOL-lished for estimating the local critical heat flux on the outer sur.

LOWING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE face of a heated hemispherical vessel that is fully t.ubmerged in FACILITIES. Uranium Blended With Soit TORAN.LE.;

water. The scaling law, which compares favorably with all the HOPPER,C.M.; NANEY,M.T.; et al. Oak Ridge National Labora-available local CHF data obtained for various vessel sizes, can tory. June 1997. 137pp. 9707180200. ORNL/TM-13323.

be used to predict the local CHF limits on large commercial-size 93805:187.

vessels.

The purpose of this study was to evaluate whether or not fis-sile uranium in low-level-waste (LLW) facilities can be concen-NUREG/CR-6508: COMPONENT UNAVAILABILITY VERSUS IN-tr;ted b/ hydrogeochemical processes to permit nuclear critical-SERVICE TEST (IST) INTERVAL: EVALUATIONS OF COMPO-i sty. A team of experts in hydrology, geology, geochemistry, soil NENT AGING EFFECTS WITH APPLICATIONS TO CHECK chemistry, and criticality safety was formed to devolup achieva-VALVES. VESELY,W.E.; POOLE A.B. Oak Ridge National Labo-ble scenanos for hydrogeoch4.al increases in concentration ratory. July 1997. 270pp. 9707280087. ORNL-6909. 93916:086.

of special nuclear material (SNM), and to use these scenarios Methods are presented for calculating component unavailabi-to aid in evaluating the potential for nuclear enticality. The lities when Inservice Test (IST) intervals are changed and when team's approach was to perform simultaneous hydrogeochemi-component aging is explic;tly included. The methods extend I

cil and nuclear criticality studies to (1) identify some achievable usual approaches for calculating unavailability and risk effects j

scenarios for uranium migration and concentration increase at of changing IST intervals which utilize Probabilistic Risk Assess-LLW disposal facilities. (2) model groundwate Yansport and ment (PRA) methods that do not explicitly include component j

subsequent concentration increase via sorptic" / precipitation aging. Difforent IST characteristics are handled including ISTs of uranium, and (3) evaluate the potential for r,-,elear criticality which are not followed by corrective maintenances which com-

]

risutting from potential increases in uranium concentration over pletely renew or partially renew the component. ISTs which are 1

disposal limits. The analysis of SNM was restricted to (235)U in not followed by maintenance activities needed to renew the the present scope of work. The outcomo of the work indicates component are also handled. Any downtime associated with the that enticality is possible given established regulatory limits on IST, including the test downtime and the following maintenance SNM disposal. However, a review based on actual disposal downtime, is included in the unavailability evaluations. A range records of an existing site operation indicates that the potential of component aging behaviors is studied including both linear for enticality is not a concem under current burial practices, and nonlinear aging behaviors. Based upon evaluations com-pleted to date, pooled failure data on check valves show rela-NUREQ/CR-6506: EMBRITTLEMENT DATA BASE, VERSION 1.

tively small aging (e.g., less than 7% per year). Nwever, data 1

WANG.J.A. Oak Ridge National Laboratory. August 1997.

from some plant systems could be evidence for larger aging j

220pp.9709120071. ORNL/TM-13327. A0413:068.

rates occurnng in time periods less than 5 years. The methods i

Version 1 of the Embrittlement Data Base (EDB) is a compre-are utikzed in this report to carry out a range of r,ensitivity eval-hensive collection of data resutting from merging Version 2 of untions to evaluate aging effects for different possible applica-the Power Reactor Embnttlement Data Base (PR-EDB) and Ver*

tions. Based on the sensitivity evaluations, summary tables are sion 1 of the Test Reactor Embrittlement Data Base (TR EDB).

constructed showing how optimal IST interval ranges for check Fracture toughness data were also integrated into Version 1 of valves can vary relative to different aging beha Aors which might the EDB. For power reactor data, the current EDB lists 1,029 exist. The evaluations are also used to identify IST intervals for tr:mitiordtemperature shift data points (321 from plates,125 check valves which are robust to comporunt eging effects.

from forgings.115 from correlation monitor materials,246 from Generalinsights on aging effects are also extracted. These sen-welds, and 222 from heat-affected-zone (HAZ) materials) from sitivity studies and extracted results provide useful information Charpy specimens that were irradiated in 271 capsules from which can be supplemented or be updated with plant specific 101 commercial power reactors. For test reactor data, informa-information. The models and results can a'so be input to PRAs tion is available for 1,308 drfferent irradiated sets (352 from to determine associated risk implications.

plates,186 from forgings,303 from correlation monitor materi-cis. 396 from welds, and 71 from HAZs) and 268 different irradi-NUREG/CR-6511 V01: STEAM GENERATOR TUBE INTEGRITY cted plus annealed data sets (89 from plates,4 from forgings, PROGRAM. Semiannual Report, August 1995 - March 1996.

11 from correlation monitor materials, and 164 from weld mate-DIERCKS,D.R.; BAKHTIARl,S.; CHOPRA,0.K.; et al. Argonne rials). The data files of EDB are given in dBASE format and can National Laboratory. April 1997.114pp. 9705120295. ANL-96/

be accessed with any personal computer using the DOS or

17. 92828:117.

WINDOWS operating system. A utility program has been wntten This report summanzes work performed by Argonne National to investigate radiation embntilement using this data base.

Laboratory on the Steam Generator Tube Integrity Program

l 26 Main Citations and Abstracts from the inception of that program in August 1995 through NUREG/CR-6515: BLT EC (BREACH, LEACH AND TRANS-March 1996. The program is divided into five tasks, namely (1)

PORT-EQUILIBRIUM CHEMISTRY) DATA INPUT GUIDE.A Assessment of Inspection Reliability, (2) Research on ISI (in.

Computer Model For Simulating Release And Coupled Geo-service-inspection) Technology, (3) Research on Degradation chemichi Transport Of Contaminants From A Subsurface Dis-Modes and Integrity, (4) Development of Methodology and posal Facility. MACKINNON,R.J. Ecodynamics Research Asso-l Technical Requirements for Current and Emerging Regulatory ciates, Inc... SULLIVAN.T.M.; KINSEY,R.R. Brookhaven National i

lasues, and (5) Program Management. Under Task 1, progress Laboratory. May 1997. 240pp. 9706180471, BNL-NUREG-j 525t6. 93484:001, is reported on the preparation of and evaluation of Nondestruc.

tive evaluation (NDE) techniques for inspecting a mock-up The BLT-EC computer code has been developed, implement.

ed, and tested. BLT-EC is a two-dimensional finite element steam generator for round-robin testing, the development of computer code capable of simulating the time-dependent re-better ways to correlate burst pressure and leak rate with eddy lease and mache transpod of apow phase Ws h a 86 current (EC) signals, the inspection of sleeved tubes, workshop surface soil system. BLT-EC contains models to simulate the and training activities, and the evaluation of emerging NDE technolo;iy. Under Task 2, results are reported on closed-form f,$

• ct e e

s nd rad ctive pr ion solutions and finite element electromagnetic modeling of EC decay) most relevant to estimating the release and transport of probe response for various probe designs and flaw characteris-contaminants from a subsurface disposal system. Water flow is tics. Under Task 3 facihties are being designed and buitt for the provided through tabular input or auxiliary files. Container degra-production of cracked tubes under aggressive and near-prototy-dation considers localized failure due to pitting corrosion and pical conditions and for the testing of flawed and unflawed general failure due to uniform surface degradation processes.

tubes under normal operating, accident, and severe accident Waste-form pedormance considers release to be limited by one conditions. In addition, crack behavior and stability are being of four mechanisms: rinse with partitioning, diffusion, uniform modeled to provide guidance on test facility design, to develop surface degradahon, and solubility. Chemical reactions account-an improved understanding of the expected rupture behavior of ed for include complexation, sorption, dissolution-precipitation, tubes with circumferential cracks, and to predict the behavior of oxidation-reduction, and ion exchange. Radioactive production flawed and unflawed tubes urder severe accident conditions.

and decay in the waste form is simulated. Transport considers Task 4 is concemed with the cracking and failure of tubes that the processes of advection, dispersion, diffusion, chemical reac-have been repaired by sleeving, and with a review of htorature tion, radioactive production and decay, and sources (waste form on this subject.

releases). To improve the usefulness of BLT-EC, a pre-proces-sor, ECIN, which assists in the creation of chemistry input files, NUREG/CR4513 Not: NRC HIGH-LEVEL RADIOACTIVE and a post-processor, BLTPLOT, which provides a visual dis-WASTE MANAGEMENT PROGRAM ANNUAL PROGRESS play of the data have been developed. BLT EC also includes an REPORT: FISCAL YEAR 1996. SAGAR.B. Center for Nuclear extensive database of thermodynamic data that is also accessi-Waste Regulatory Analyses. January 1997.317pp.9704080389.

ble to ECIN. This document reviews the models implemented in FACA. 92385.001.

BLT-EC and serves as a guide to creating input files and apply-This annual status report for fiscal year 1996 documents ing BLT-EC.

technical work performed on ten key technical issues (KTis)

NUREG/CR4519: SCREENING REACTOR STEAM / WATER that are most important to performance of the proposed geolog.

PIPING SYSTEMS FOR WATER HAMMER. GRIFFITH,P. Mas-ic repository at Yucca Mountain. This report was prepared joint-sachusetts institute of Technology, Cambridge, MA. September ty by the staff of the Nuclear Regulatory Commission (NRC) Di-1997,52pp. 9709120067. A0354:234.

vision of Waste Management and the Center for Nuclear Waste A steam / water system possessing a certain combination of Regulatory Analyses. The programmatic aspects of restructuring thermal, hydraulic and operational states, can, in certain geome-the NRC repository program in terms of KTh % discussed and a tries, lead to a steam bubble collapse induced water hammer.

brief summary of work accomplished is provided in Chapter 1.

These states, operations, and geometries are identified. A pro.

The other ten chapters provide a comprehensive summary of cedure that can be used for identifying whether an unbuilt reac-the work in each KTI. Discussions on probability of future vol-tor system is prone to water hammer is proposed. For the most canic activity and its consequences, impacts of structural defor-common water hammer, steam bubble collapse induced water mation and seismicity, the nature of the near-field environment hammer, six conditions must be met in order for one to occur.

and its effects on container hfe and source term, flow and trans-These are: 1) the pipe must be almost horizontal; 2) the sub-port including effects of thermal loading. aspects of repository cooling must be greater than 20 degrees C; 3) the L/D must be design, estimates of system performance, and actrvities related greater than 24; 4) the velocity must be low enough so that the to the U.S. Environmental Protection Agency standard are pro.

pipe does not run full, i.e., the Froude number must be less vided.

than one; 5) there should be void nearby; 6) the pressure must be high enough so that signifcant damage occurs, that is the NUREG/CR4514: ANALYSIS OF POTENTIAL SELF-GUARAN-pressure should be above 10 atrnospheres. Recommendations TEE TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE on how to avoid this kind of water hammer in both the design BY NON-PROFIT COLLEGES, UNIVERSITIES, AND HOSPl.

and the operation of the reactor system are made.

TALS AND BY BUSINESS FIRMS THAT DO NOT ISSUE NUREG/CR4523 V01: PROBABILISTIC ACCIDENT CONSE-BONDS. BAILEY,P.; DEAN.C.; COLLIER,J.; et al. ICF, Inc. June QUENCE UNCERTAINTY ANALYSIS. Food Chain Uncertainty 1997,70pp. 9706200262. 93422:175.

Assessment. Main Report. BROWNJ. United Kingdom.

l This report describes potential financial tests which could be GOOSSENS LH.J.; KRAAN,0.C.P.; et al. Delft University of I

used by NRC as a basis for allowing certain financially strong Technology. June 1997. 78pp. 97090203t3. EUR 16771.

nonprofit licensees, and also non-bond issuing Icensees, to use A0234:001.

(

self-guarantee as a mechanism for meeting NRC financial as-The development of two new probabilistic accident conse.

surance requirements. The analysis focuses on three categories quence codes, MACCS and COSYMA, was completed in 1990.

of hcensees; colleges or unhrersities, hospitals, and commercial These codes estimate the consequence from the accidental re-j firms that do not issue bonds. The report assesses the financial leases of radiological material from hypothesized accidents at l

assurance nsk of vanous financial tests, and also estimates the nuclear installations, in 1991, the U.S. Nuclear Regulatory Com-number of licensees which could quahfy for self-guarantee mission and the Commission of the European Communities under different financial test attematives.

began cosponsoring a joint uncertainty analysis of the two

Main Citations and Abstracts 27 codes. The ultimate objective of this joint effort was to systern-See NUREG/CR-6526,V01 abstract.

(tically develop credible and traceable uncertainty distributions for the respective code. input variables. A formal expert judg-NUREG/CR4527: FINAL RESULTS OF THE XR2-1 BWR ME-ment c?icitation and evaluation process was identified as the TALLIC MELT RELOCATION EXPERIMENT. GAUNTT,R.O.

-l best !achnology available for developing a library of uncertainty Sandia National Laboratories. HUMPHRIES,L.L. Science Appli-distributions for these consequence parameters. This report fo-cations intemational Corp. (formerty Science Applications, Inc.).

cuses on the results of the study to develop distribution for vari-August 1997.177pp. 9708210414. SAND 971039. A0156:112.

ables related to the MACCS and COSYMA food chain models.

This report documents the final results of the XR2-1 boiling Both soil / plant transfer processes and radionuclides transport in water reactor (BWR) metallic melt relocation expenment, con-animals were assessed.

ducted at Sandia National Laboratories for the U.S. Nuclear NUREG/CR4523 V02: PROBABILISTIC ACCIDENT CONSE.

inves4 gate the material relocation processes and relocation OUENCE UNCERTAINTY ANALYSIS. Food Chain Uncertainty pathways in a dry BWR core following a severe nuclear reactor Assessment. Appendices.

BROWN.J.

United Kingdom.

accident such as an unrecovered station blackout accident. The GOOSSENS,LH.J.; KRAAN,8.C.P.; et al. Delft University of imposed test conditions (initial thermal state and the melt gen-Technology. June 1997. 334pp. 9709020327. EUR 16771, eration rates) simulated the conditions for the postulated acci-dent scenario and the prototypic design of the lower core test U EGJCR-6523,V01 abstract.

j section (in composition and in geometry) ensured that thermal NUREG/CR4525: SECPOP90: SECTOR POPULATION, LAND masses and physical flow barriers were modeled adequately.

FRACTION, AND ECONOMIC ESTIMATION PROGRAM.

The experiment has shown that, under dry core conditions, the HUMPHREYS,S.L.

Sandia National Laboratories.

metallic core materials that melt and drain from the upper core ROLLSTIN.J.A. GRAM, Inc. RIDGELY,J.N. Division of Systems regions can drain from the core region entirely without formation Technology (Post 941217). September 1997. 44fpp.

of robust coherent blockages in the lower core. Temporary 9710070358. SAND 93-4032. A0641:125.

blockages that suspended pools of molten metal later melted, in 1973 Mr. W. Athey of the Environmental Protection Agency allowing the metals to continue draining downward. The test fa-wrote a computer program called SECPOP which calculated cility and inst,umentation are described in detail. The test pro-population estimates. Since that time, two things have changed gression and results are presented and compared to MERIS which suggested the need for updating the original program--

code analyses.

more recent population censuses ard the widespread use of personal computers (PCs). The revised computer program uses NUREG/CR4528: ENVIRONMENTAL ASSESSMENT PRO-the 1990 and 1992 Population Census information and runs on POSED LICENSE RENEWAL OF NUCLEAR METALS,1NC.

corrent PCs as "SECPOP90". SECPOP90 consists of two parts:

CONCORD, MASSACHUSETTS. MILLER,R.L.; EASTERLY,C.E.;

site and regional. The site analysis provides population and eco-LOMBARDI D.A.; et al. Oak Ridge National Laboratory. February nomic data estimates for any location within the continental 1997,88pp. 9703100266. 92020:152.

United States. Siting analysis is relatively fast running. The re-

. This Environmental Assessment was prepared to evaluate erv gional portion assesses sito availability for different siting policy vironmental issues associated with the renewal of NRC Licens-decisions; i.e., the impact of availab!e sites given specific popu-ee Nos. SMB-179 and SUB-1452 for facilities operated by Nu-

{

lation density enteria within the continental United States. Re.

clear Metals, Inc. (NMI) k: Concord Massachusetts. License re-j l

gional analysis is slow. This report compares the SECPOP90 newal is needed to permit the continuation of NMI operations j

population estimates and the nuclear power reactor licensee-involving depleted and natural uranium.

j provided information. Although the source, and therefore, the NUREG/CR-6529: VALIDATION OF TECTONIC MODELS FOR accuracy of the licensee information is unknown, this compari-AN INTRAPLATE SEISMIC ZONE, CHARLESTON, SOUTH son suggests SECPOP90 makes reasonable estimates.

CAROLINA WITH GPS GEODETIC DATA. TALWANI.P.;

NUREG/CR4526 V01: PROBABILISTIC ACCIDENT CONSE.

KELLOGG J.N.; TRENKAMP,R. South Carolina, Univ. of, Colum-OUENCE UNCERTAINTY ANALYSIS. Uncertainty Assessment bia, SC. February 1997. 54pp. 9703100260. 92018:299.

For Deposited Material And External Doses. Main Report.

Although the average strain rate in intraplate settings is 2-3 GOOSSENS,LH.; KRAAN.B.C.; et al. Netherlands, Govt. oi.

orders of magnitude lower than at plate boundaries, there are BOARDMAN,J. AEA Technology. December 1997, 66pp.

pockets of high strain rates within intraplate regions. The results 9801260178. EUR 16772. A1879:040.

of a Global Positioning System survey near the location of cur-The development of two new probabilistic accident conse-rent seismicity (atu the inferred location of the destructive 1886 quence codes, MACCS and COSYMA, was developed in 1990.

Charleston, South Carolina earthquake) suggest that there is These codes estimate the consequence from the accidental re.

anomalous strain build-up occurring there. By reoccupying 1930 leases of radiological material from hypothesized accidents at triangulation and 1980 GPS sites with six Trimble SST dual fre-nuclear installations. In 1991, the U.S. Nuclear Regulatory Com-quency receivers, a strain rate of 0.4 x 10(-7) yr( 1) was ob-mission and the Commission of the European Communities served. At the 95% confidence level, this value is not signifi-began cosponsoring a joint uncertainty analysis of the two cant; however, at a lower level of confidence (-- 85%) it is codes. The ultimate objective of this joint effort was to system-about two orders of magnitude greater than the background of Etically develop credible and traceable uncertainty distributions 10(-9) to 10( 10) yr(.1). The direction of contraction inferred for the respective code input variables. A formal expert judg.

from the GPS survey 66 degrees i 11 degrees is in excellent i

ment elicitation and evaluation process was identified as the agreement with the direction of the maximum horizontal stress l

best technology available for developing a hbrary of uncertainty (N 60 degrees E) in the area, suggesting that the observed l

distributions for these consequence parameters. This report to-strain rate is also real.

cuses on the results of the study to develop distnbution for vari-NUREQ/CR-6530: DELIBERATE IGNITION OF HYDROGEf0A!R.

tbles related to the MACCS and COSYMA deposited material STEAM MIXTURES IN CONDENSING STEAM ENVIRON.

and Memal dose models' MENTS. BLANCHAT,T.K. Sandia National Laboratories.

]

I NUREG/CR-6526 V02: PROBABILISTIC ACCIDENT CONSE-STAMPS.D.W. Evansville Univ. of Evansville, IN. May 1997.

l OUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESS-93pp. 9706240048. SANL94-1676. 93489:266.

MENT FOR DEPOSITED MATERIAL AND EXTERNAL Large scale experiments were performed at the Surtsey Test DOSES. Appendices. GOOSSENS,LH.; KRAAN B.C.; et al.

Facility for the Nuclear Regulatory Commission to determine the Netherlands, Govt. of. BOARDMAN.J. AEA Technology. Decem-effectiveness of thermal glow igniters to bum hydrogen in a rap-i ber 1997. 403pp. 9801260185. EUR 16772. A1878:001.

idly condensing steam environment due to the presence of

I i

i 28 Main Citations and Abstracts water sprays. The experiments were designed to determine if a NUREG/CR 6534 V01: FRAPCON-3: MODIFICATIONS TO FUEL detonation or an accelerated flame could occur in a hydrogen-ROD MATERIAL PROPERTIES AND PERFORMANCE air steam mixture which was initially nonflammable due to steam MODELS FOR HIGH-BURNUP APPLICATION. LANNING D.D.;

dilution but was subsequently rendered flammable by rapid con-BEYER,C.E.; PAINTER.C.L Battelle Memorial institute, Pacific J

f densation of steam due to water sprays. The experiments were Northwest National Laboratory. December 1997. 131pp.

conducted under conditions scaled to be nearty prototypic of 9801120078. PNNL 11513. A1744:001.

those expected in Advanced Light Water Reactors (such as the This volume describes the fuel rod material and performance Combustion Engineenng (CE) System 80+), with prototypic models that were updated for the FRAPCON-3 steady-state fuel spray drop diameter, spray mass flux, steam condensation rod performance code. The property and performance models rates, hydrogen injection flow rates, and using the actual pro.

were changed to account for behavior at extended bumup posed plant igniters. The lack of any significant pressure :n.

levels up to 65 GWd/MTU. The property and performance crease during the majority of the bum and condensation events, models updated were the fission gas release, fuel thermal con-signified that localized, benign hydrogen deflagration (s) oc.

duchvity, fuel swelling, fuel relocation, radial power distnbution, curred with no significant pressure load on the Surtsey contain, solid-solid contact gap conductance, cladding corrosion and hy-ment vessel. This report describes these expenments, gives the dnding, cladding mechanical properties, and cladding axial experimen'** results, and provides interpretation of the results.

Growth. Each updated property and model was compared to well characterized data up to high bumup levels. The installation NUREG/CR-6531: EFFECTS OF RADIOACTIVE HOT PARTI.

of these properties and models in the FRAPCON-3 code along CLES ON PIG SKIN.

KAURIN,0.G.L; BAUM,J.W.;

with input instructions are provided in Volume 2 of this report CARSTEN,A.L; et al. Brookhavon National Laboratory. June and Volume 3 provides a code assessment Dased on compari-1997.315pp.9710060475. BNL-NUREG-52499. A0623:056.

son to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON 2 and The purpose of these studies was to determine the incidence GAPCON-THERMAL-2.

and severity of lesions resulting from very localized deposition of dose to skin from small(< 0.5 mm) discrete radioactive par-NUREG/CR-6535: DEVELOPMENT OF CONFORMAL RESPIRA.

ticles as produced in the work environmer,ss of nuclear reactors.

TOR MONITORING TECHNOLOGY.

SHONKA.J.J.;

Hanford mini-pigs were exposed, both on and slightly off the WEISMANN,J.J.; LOGAf'.,R.J.; et al. Affiliation Not Assigned.

skin, to localized replicate doses from 0.31 to 64 Gy (averaged April 1997. 28pp. 9705210294. 93061:317, over 1 CM(2) at 70 p m depth unless noted otherwise) using This report summarizes the results of a Small Business Inno-Sc-46, Yb-175, Tm-170, and fissioned UC(2) isotopes having vative Research Phase il project to develop a modular, surface maximum beta-particle energies from about 0.3 to 3 MeV. Ery-conforming respirator monitor to improve upon the manual thema and scabs (indicating ulceration) were scored for up to survey techniques presently used by the nuclear industry. Re-71 days post-irradiation. The responses followed normal cumu-search was performed with plastic scintillator and gas propor-lative probability distributions, and therefore, no true threshold tional modules in an effort to find the most conducive geometry could be defined. Hence,10 and 50% scab incidence rates for a surface conformal, position sensitive monitor. The respira-were deduced using probit analyses. The lowest dose which tor monitor putotype developed is a computer controlled, posi-produced 10% incidence was about 1 Gy for Yb-1'"i (0.5 MeV tion-sensitNe detection system employing 56 modular propor-maximum energy) beta particle exposures, and., Lout 3 to 9 Gy tional counters mounted in molds conforming to the inner and for other isotopes. The histnpathology of lesions was deter-outer surfaces of a commonly used respirator (Scott Model mined at several doses. Single exposures to doses as large as 801450-40). The molds are housed in separate enclosures and 1,790 Gy were also given, and results were observed for up to hinged to create a " waffle-iron" effect so that the closed moni-144 days post exposure. Severity of detriment was estimated by tor will simultaneously turvey both surfaces of the respirator, analyzing the results in terms of lesion diameter, persistence.

The proportional counter prototype was also designed to incor-and infection. Over 1,100 sites were exposed. Only two ex-porate Shonka Research Associates' previously developed posed sites became infected after doses near 500 Gy; the le.

charge-division electronics. This research provided valuable ex-sions healed quickly on treatment.

perience into pixellated position sensitive detection systems.

The technology developed can be adapted to other monitoring

'dUREG/CR-6533: CODE MANUAL FOR CONTAIN 2.0; A COM-applications where there is a need for deployment of many tra-PUTER CODE FOR NUCLEAR REACTOR CONTAINMENT ditional radiation detectors.

ANALYSIS. MURATA,K.K.: WILLIAMS,D.C. Sandia National NUREG/CR-6538: EVALUATION OF LOCA WITH DELAYED Laboratories. TILLS,J.; et al. Affiliation Not Assigned. December LOOP AND LOOP WITH DELAYED LOCA ACCIDENT SCE.

1997. 955pp. 9801120086. SAND 971735. A1731:001 NA N

SA A,

et al The CONTAIN 2.0 computer code is an integrated analysis Br khaven National Laboratory. July

1997, 231pp.

tool used for predicting the physical, chemical, and radiological

8. M8222.

conditions inside a containment building following the release of Generic Safety Issue 171 (GSI-171), Engineered Safety Fea-material from the primary system in a light-water reactor (LWR) tures (ESF) Failure from a Loss Of Offsite Power (LOOP) subse-accident. It can also predict the source term to the environment.

quent to a Loss Of Coolant Accident (LOCA), deals with an ac-The purpose of this Code Manual is to provide full documenta-cident sequence in which a LOCA is followed by a LOOP. This tion of the features and models in CONTAIN 2.0 Besides co*

issue was later broadened to include a LOOP followed by a plete desc-iptions of the models, this Code Manual provides a LOCA. Plants are designed to handle a simultaneous LOCA and complete description of the input and output from the code. The LOOP. In this report, we address the unique issues that are in.

I code includes atmospheric models for steam / air thermodynam-volved in LOCA with delayed LOOP (LOCA/ LOOP) and LOOP ics, intercell flows, condensation / evaporation on structures and with delayed LOCA (LOOP /LOCA) accident sequences, and de-l aerosols, aerosol behavior, and gas combustion it also includes termine that such sequences and the specific concerns raised models for reactor cavity phenomena such as core-concrete as part of GSI-171 are not fully addressed in Individual Plant Ex-interactions and coolant pool boiling. Heat conduction in struc-amination (IPE) submittals. The determination is based on our tures, fission product decay and transport, radioactive heating.

review of selected IPE Submittals. LOOP /LOCA accidents are and the thermal-hydraulic and fission product decontamination addressed more fully by IPEs than are LOCA/ LOOP ones.

effects of engineered safety features are also modeled. These LOCA/ LOOP accidents are analyzed further in this report by de-models allow selected design basis and severe accidents to be veloping event-tree / fault-tree models to quantify their contribu-analyzed, for both current and advanced LWR designs.

tions to core-damage frequency (CDF) in a pressurized water

Main Citations and Abstracts 29 reactor and a boiling water reactor (PWR and a BWR). Engi-This report documents measurements of basic functional cir-neering evaluation and judgements are used dunng quantifica-cuits during and up to 1 day after exposure to smoke crated by tion to estimate the unique conditions that anse in a LOCA/

burning cable insulation. Pnnted wiring boards were exposed to j

LOOP sceident. The results show that the CDF contributica of the smoke in an enclosed chamber for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For high-resist-such an accident can be a dominant contnbutur to plant risk, ance circuits, the smoke lowerod the resistance of the surface although BWRs are less vulnerable than PWRs.

of the board and caused the circuits to short dunng the expo.

sure. These circuits recovered after the smoke.sas vented. For NUREQ/CR4539: EFFECTS OF FLUORIDE AND OTHER HALO-low stane carcWs, the smoke caused thek resistance to in-GEN IONS ON THE EXTERNAL STRESS CORROSION CRACKING OF TYPE 304 AUSTENITIC STAINLESS STEEL crease shghtly. A polyurethane conformal coabng substantially reduced the effe a of smoke. A high-speed digital circuit was r

WHORLQW,K.M.; HUTTO,F.B. Affiliation Not Assigned. July 1997. 50pp. 9707240126. 93886:153.

unaffected. A second expenment on different logic chip technol-The drip procedure from ASTM C 692-95a was used to re-ogies showed that the critical shunt resistance that would cause search the effect of halogens and inhibitors on the External fa!!ure was dependent on the chip technology and that the com-Stress Corrosion Cracking (ESCC) of Type 304 stainless steel ponents used in the smoke exposures were some of the most as it applies to NRC RG 1.36 The solutions used in this re-smoke tolerant. The smoke densities in these tests were high search were prepared using pure chemical reagents to simulate enough to cause changes in high impedance (resistance) cir-the halogens and inhibitors found in insulation extraction solu-cuits during exposure, but did not affect most of the other cir.

tions. The results indicated that sodium silicate compounds that cuits. Conformal coatings and the characteristics of chip tech-were higher in sodium were more effective for preventing cho-nologies should be considerad when designing digital circuitry ride-induced ESCC in type 304 austenitic stainless steel. Petas-for nuclear power plant safety systems, which must be highly slum silicate (all-silicate inhibitor) was not as effectv; as reliable under a variety of operating and accident conditions.

sodium silicate. Limited testing with sodium hydroxido (all-NUREG/CR-6547: DOSFAC2 USER'S GUIDE. YOUNG,M.L sodium inhibt;or) indicated that it may be effective as ar' inhibi*

Sandia National Laboratories. CHANIN.D.L Technadyne Engi-tor. Fluonde, bromide, and iodide caused minimal ESO0 which neering Consultants, Inc. December 1997. 55pp 9801120074.

could be effectively inhibited by sodium silicate. The addition of SAND 97-2776. A1733:276.

I fluoride to ;he chloride / sodium silicate systems at the threshold This document is a user's guide for it's DOSFAC2 Code.

of ESCC appeared to have no synergistic enect on ESCC. The DOSFAC2 generates a file of dose-to-source conversion factors mass ratio of sodium + silicate (mg/kg) to chloride (mg/kg) at for the MACCS2 code. DOSFAC2 is a revised and updated ver-the lower end of the NRC RG 1.36 Acceptability Cuve was not sion of the DOSFAC code that was distributed with MACCS ver-sufficient to prevent ESCC using the methods of this research.

sion 1.5.11 of the MACCS code. DOSFAC did not generate NUREG/CR-6541 R02: PHENOMENA IDENTIFICATION AND ICRP 60 effective (E) dose conversion factors (DCFs) or accept RANKING TABLES FOR WESTINGHOUSE AP600 SMALL user input data. DOSFAC calculations were based on parameter BREAK LOSS-OF-COOLANT ACCIDENT, MAIN STEAM LINE values hardwired into the code. DOSFAC2 accepts user input BREAK, AND STEAM GENERATOR TUBE RUPTURE SCE.

data through a user input file and can generate ICRP 60 E NARIOS. WILSON,G.E.; FLETCHER,C.D.; DAVIS,C.B.; et al.

DCFs. The parameter values for which DOSFAC2 accepts use -

idaho National Engineering & Environmental Laboratory. June input values are: (1) the values of relative biological effective-1997. 260pp. 9709L30400. INEL-94/0061. A0235:298.

ness associated with high-LET radiations, (2) the list of organs This report documents the results of Phenomena identifica.

for which acute DCFs are to be calculated, (3) the activity tion and Ranking Table (PIRT) efforts for the Westinghouse median aerodynamic diameter, (4) the acute dose reductior, fac.

J AP600 reactor. The purpose of this PIRT is to identify important tors, and (5) the inhalation clearance class for each radionu-phenorrana so that they may be addressed in both the experi.

clide.

mental programs and the RELAPS/ MOD 3 systems analysis NUREG/CR-6557: DEVELOPMENT OF THE MAGNESCOPE AS computer code. N rtsponses of AP600 during small break AN INSTRUMENT FOR IN SITU EVALUATION OF STEEL loss-of-coolant accident, main steam line break, and steam gen-COMPONENTS OF NUCLEAR SYSTEMS. JILES,D.C.; BI,Y.;

erator tube rupture accident scenarios were evaluated by a BINER.S.B. Iowa State Univ., Ames, IA. August 1997. 64pp.

committee of thermal-frudraulic experts. Committee membership included Idaho National Engineering and Environmental Labora-9708290226 A0232:284'es continuous, cumulative microstruc-Fatigue damage caus tory staff and recognized thermal-hgraulic experts from outside tural changes in materials and the magnetic properties of steels of the laboratory. Each of the accident scenarios was subdivid-are sensitive to these microstmetural changes. This work there-ed into separate, sequential periods or phases. Within each fore focused on the relationships between fatigue damage and phase, the plant behaMor is controlled by, et most, a few ther-the measured magnetic properties of different staels under a mal-hydraulic processes. The committee identified the phenom-variety of fatigue conditions. The project also investigated the ene influencing those processes, and ranked the influences as feas bility and applicability of magnetic inspection techniques for being of f 6gh, needium, low, or insignificant importance. The pri-non-destructive evaluation of fatigue damage. From the results mary product of this effort is a series of tables, one for each of a series of fatigue tests, conducted on different steels under phase of each accident scenario, describing the thermal-hydrau-both low cycle and high-cycle fatigue conditions, the magnetic lic phenomena judged by the committee to be important, and properties, such as coercivity, remanence and Barkhausen the relative ranking of that importance. The rabonales for the effect, were found to change systematically with fatigue phenomena selected and their rankings are provided.

damage. The magnetic properties showed significant changes, NUREG/CR-6543: EFFECTS OF SMOKE ON FUNCTIONAL CIR.

especially during early stage of the fatigue and also at the end cults. TANAKA,T.J. Sandia National Laboratories. October of fatigue lifetime. An approximately linear relationship between 1997. 57pp. 9711210006. SAND 97-2544. A1181:216.

the mechanical modulus and magnetic remanence was ob-Nuclear power plants are converting to $grtal instrumentation served and was explained by a model developed in this study to and control systems; however, the effects of abnormal environ-describe the dynamic changes in the magnetic and mechanical ments such as fire and smoke on such systems ara not known.

properties. The results of this research demonstrated that map There are no standard tests for smoke, but presious smoke ex-r%c measurements are suitable for non-destructive evaluaticq posure tests at Sandia National Laboratories have shown that of fatigue damage in steels such as A5338 steel and Cr-Mo digital communications can be temporarily interrupted during a staels. These magnetic measurement techniques have been in-smoke exposure. Another concern is the long-term cotrosion of corporated into instrumentation for in-situ evaluation of steel metals exposed to the acidic gases produced by a cable fire.

structures and components.

30 Main Citations and Abstracts NUREG/CR-6558: NRC ANTITRUST LICENSING ACTIONS, analysis. The resulting updated soil hydraulic parameter distribu-1978-1996. MAYER,S.J.; SIMPSON.J.J. Oak Ridge National tons can be used to obtain an updated estimate of the proba-Laboratory. September 1997.141pp. 9710070363. ORNL/TM-bihty distribution of dose. The method is illustmted using an hy-13452. A0642:273.

pothetical example decommissioning site.

NUREG-0447. " Antitrust Review of Nuclear Power Plants,"

was published,n May 1978 and includes a compdation and dis-NUREG/CR4566: DESCRIPTION OF MULTIMEDIA ENVIRON-cussion of U.S. Nuclear Regulatory Commission (NRC) proceed-MENTAL POLLUTANT ASSESSMENT SYSTEM (MEPAS) VER-ings and activity involving the NRC's competitive review pro.

SiON 3.2 MODIFICATION FOR THE NUCLEAR REGULATORY gram through February 1978. NUREG-0447 is an update of an COMMISSION. LUCK,J.W.; STRENGE,D.L; HOOPES,B.L; et earher discussion of the NRC's entit ust review of nuclear power al. Battelle Memorial Institute, Pacific Northwest National Labo-plants, NR-AlG 001, "The US Nuclear Regulatory Commission's ratory. November 1997. 98pp. 9712110135. PNL 11676.

Atititrust Review of Nuclear Power Plants: The Conditioning of A1387:057.

Licenses," which reviewed the Commission's antitrust review The Multimedia Environmental Pollutant Assessment System functson from its inception in December 1970 through Apnl (MEPAS) is a software tool developed by Pacific Northwest Na-1978. This report summarizes the support provided to NRC staff tional Laboratory (PNNL) for the U.S. Department of Energy in updating the compilation of the NRC's antitrust licensing (DOE) to allow DOE to conduct human health risk analyses review activitms for commercial nuclear power plants that have nation-wide. This report describes modifications to the MEPAS occurred since February 1978.

to meet the requirements of the U.S. Nuclear Regulatory Com-NUREG/CR-6563: LG EXCITATION. ATTENUATION, AND missbn N staH h ther anahses of Sh DecommissMng SOURCE SPECTRAL SCALING IN CENTRAL AND EASTERN Management Plan sites. In general, these niodifications provide NORTH AMERICA. MITCHELL.B.J.; XIE.J.; BAOER,S. St. Louis the MEPAS, Version 3.2. with the capability of calculating and Univ., St. Louis, MO. October 199/ 53pp. 9712230294.

repodng annual dose / risk information. Modifications were made A1501:199 to the exposure pathway and health impact modules and the Seismic moments and corner frequencies were obtained for water and atmospheric transport modules. Several example many earthquakes in the central and eastem United States, and cases used to test the MEPAS, Version 3.2, are also presented.

for a few events in the westem United States, using the Lg The MEPAS, Version 3.2, also contains a new source-term re-phase and a recently developed inversion algorithm. Lg O lease component that includes models for estimating contami-values along paths to individual statio r were obtained together nant loss from threo different types of source zones (contams-with source parameters. For moments aetween 0.15 and 400 x nated aquifer, contaminated pond / surface impoundment, and 10(15) Nm comer frequencies vary between about 4 and 0.2 Hz contaminated vadose zone) due to decay / degradation, leach-while body wave magnitude vanes between about 3.5 and 5.8.

ing, wind suspension, water erosion, overtand flew, and/or vola-Lg O values decrease from east to wes Maximum and mini-tilization. When multiple loss routes are assumed to occur simul-mum values are 098 and 100, respectively. Lg coda O values taneously, the models account for their interaction and calculate were ubtained wnn excellent coverage in the eastem and west, an appropriate pollutant mass budget to each loss route over ern portions of the country and somewhat poorer coverage in time.

the central portion. Lg coda O is highest (700-750) in portions NbREG/CR-6581: CONSIDERATIONS IN THE APPLICATION OF of New York and Pennsylvania and lower (>200) in Califomia.

THE ELECTRONIC DOSIMETER TO DOSE OF RECORD-Lg coda O is lower (250-450) everywhere west of the Rocky Mountains than in the rest of the country (450-750). For an SWINTH,K.L. December 1997. 70pp. 9801140351. A1799:012.

earthquake of a given magnitude, Lg and its coda will propagate This report describes considerations for application of the much more efficiently, and cause damage over a wider area, in electron 6c dosimeter (ED)as a measurement device for the dose the eastem and central United States than in the western of record (primary dosimetry). EDs are widely used for second-United States.

ary dosimetry and advances in their reliability and capabilities

{

have resutted in interest in their use to meet the needs of both NUREG/CR-6565: UNCERTAINTY ANALYSES OF INFILTRA-primary and secondary dosimetry. However, the ED is an active TION AND SUBSURFACE FLOW AND TRANSPORT FOR device and more complex than the thermoluminescent and film SDMP SITES. MEYER,P.D.; ROCKHOLD,M.L; GEE G.W. Bat-dosimeters now in use ter primary dosimety The user must telle Memorial Institute, Pacific Northwest National Laboratory.

evaluate the ED in terms of reliabikty, serviceabihty and radi-September 1997 151pp.

9711030078. PNNL 11705.

ations detected its intended application (s). If an ED is selected A0990:100.

for primary dosimetry, the user must establish methods both for Traits common fo many SDMP t*es include limited data char-controlling the performance of the ED to ensure long term reli-acterizing the subsurface, the presence of long-lived radionu-ability of the measurements and for their proper use as a pri-clides neceuitating a long-term analysis (1000 years or more),

mary dosimetei. Regulatory groups may also want to develop and potential exposure through multiple pathways. As a conse-methods to ensure adequate performance of the ED for dose of quence of these traits, the uncertainty in predicted exposures record. The purpose of the report is to provide an overview of can be significant. Several tools for improving uncertainty s'ialy-considerations in the use of the ED for primary dosirnetry. Con-ses of exposure estimates through the groundwater pathway siderations include recognizing current limitations, type testing are discussed in this report. Ger. enc probabihty distributions for of EDs, testing by the user, approval performance tesSng, cali.

unsaturated and saturated zone soil hydraulic parameters are bration, and procedures to integrate the dosimeter into the presented. These distributions can be used with available dose users program.

assessment codes to estimate exposure uncertainty in screen-j ing-level and prehminary analyses where site-specific data is NUREG/CR-6586: HORIZONTAL VELOCITIES IN THE CENTRAL i

~

limited. The use of the generic distnbutions is illustrated in a AND EASTERN UNITED STATES FROM GPS SURVEYS method for the estirn& tion of net infiltration uncertainty. The DURING THE 1987-1996 INTERVAL SNAY,R.A.;

method uses a relatively simple water bud et calculation con.

STRANGE,W.E. Commerce, Dept. of National Oceanic & At-D tained in an existing multiple pathway dose assessment code. A mospheric Administration. December 1997. 37pp. 9801120071.

comparison between the distnbuten of predicted annual net in.

A1738:266.

filtration and the observed lysimeter drainage (mean and stand-The National Geodetic Survey and the Nuclear Regulatory ard error) showed an agreeable match. At many SDMP sites Commission jointly organized GPS surveys in 1987,'1990, 1993, there may be some site-specific soil hydrauhc property data and 1996 to search for crustal deformation in the United States available. A method is presented to combine the generic distri-east of longitude 108 degrees W. We have analyzed the data of butions with site-specific water retention data using a Bayesian these four surveys in combinatiori with VLBi data from the

i l

Main Citations and Abstracts 31 I

i 1979-1995 interval Horizontal velocrbes for 64 GPS and 12 teflect actual moton relative to the North Arnencan plate. We l

VLBI sites were computed relatve to a reference frame for also computed horizontal strain rates for the cells formed by a 1 which the inter 6or of North Arnerica is fixed on average. None of degree by 1 degree grid spanning the central and eastern U.S.

the velocrbes exceous 6 mm/yr in magnitude. Moreover, the de-Shearing rates are everywhere less than 60 nanoradians/yr, rived velocity at each GPS site is statistically zero at the 95%

and no sheanng rate differs statistically from zero at the 95%

confidence level except for the sites BOLTON in Ohio and confidence level except for a grid cell near BEARTOWN whose BEARTOWN in Pennsylvania. However, as statistical theory rate is 572 26 nanoradians/yr. Areal dilatatiori rates are every-l would allow 5% of the 64 GPS sites to fail our zero-velocrty hy-where less than 40 nanostrains/yr, and no dilatation rate differs l

pothesis, the veloctties for BOLTON and BEARTOWN may not statistically from zero at the 95% confidence level.

l l

i 1

l 1

l l

2 l

1 1

l

Secondary Report Number index This index lists, in alphabetical order, the performing organization-issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross-referenced to the NUREG number for the report and to the 10-digit NRC Document Control System accession number.

SECONDARY REPORT NUMBER REPORT NUMSER SECONDARY REPORT NUMBER REPORT NUMBER 00-4448412 NUREG/CR4074 V03 ORNL4892 NUREG/CR4426 V01 AEOD/E9741 NUREG/CR4456 ORNL4892 NUREG/CR4426 V02 ANL 96/14 NUREG/CP-0154 ORNL4909 NUREG/CR4508 ANL-96/17 NUREG/CR4511 V01 ORNUNOAC-232 NUREG/CR-4674 V23 ANL-97/10 NUREG/CR-4667 V23 ORNUTM-11568 NUREG/CR-5591 V07 N1 AML-97/9 NUREG/CR-4667 V22 ORNUTM 11568 NUREG/CR-5591 V07 N2 ANL AA-30 NUREG/CR-4012 V04 ORNUTM 11936 NUREG/CR-5661 BIA EIS 92 001 NUREG-1508 ORNL/TM-13047 NUREG/CR4363 BLM NM010-03-02 NUREG 1508 ORNL/TM-13205 NUREG/CR4454 3MI-2177 NUREG/CR4233 V02 ORNUTM-13211 NUREG/CR 6361 BMI-2177 NUREG/CR4233 V03 ORNUTM-13322 NUREG/CR4504 V01 BMb2177 NUREG/CR4233 V04 ORNUTM 13323 NUREG/CR4505 V01 BML2187 NUREG/CR4389 ORNL/TM-13327 NUREG/CR4506 BMI-2194 NUREG/CR4446 ORNUTM-13452 NUREG/CR4558 BML2195 NUREG/CR4452 ORNL'TM-9593 NUREG/CR-4219 V12 N2 BNL-NUREG-51/4 NUREG/CR-4409 V06 M

BNL NUREG42442 NUREG/CR4295 UA CSD2R5 CR4 5 1P1 BRL-NUREG-52482 NUREG/CR4391 ORNLNUREGCSD2RS NUREG/CR-0200 RSV1P2 BNL NUREG-52483 NUREG/CR4393 ORNLNUREGCSD2R5 NUREG/CR-0200 R5V3 BNL-NUREG42484 NUREG/CR4397 ORNiNUREGCSD2R5 NUREG/CR4200 R5V2P1 ORNLNUREGCSD2R5 NUREG/CR-0200 RSV2P2 BNL NUREG-52485 NUREG/CR4400 ORNLNUREGCSD2R5 NUREG/CR 0200 R5V2P3 BNL-NUREG 52487 NUREG/CR4414 PNL-11676 NUREG/CR4566 BNL-NUREG-52498 NUREG/CR4451 PNNL 11513 NUREG/CR4534 V01 BNL-NUREG-52499 NUREG/CR4531 PNNL 11705 NUREG/CR4565 BNL NUREG-52500 NUREG/CR4484 PNNL 12521 NUREG/CR4331 RO1 BNL-NUREG-52504 NUREG/CR4474 PNNL-9020 NUREG/CR4181 ROI BNL-NUREG-52510 NUREG/CR4493 PSU/ME-97 7321 NUREG/CR-6507 BNL-NUREG-52516 NUREG/CR4515 SAND 93-0971 NUREG/CR4042 R01 BNL NUREG-52520 NUREG/CR4486 SAND 93-3931 NUREGICR4167 BNL NUREG-52528 NUREG/CR4538 SAND 93-4032 NUREG/CR4525 CRWRA 95012 NURE3/CR4404 SAND 96 2289 NUREG/CR4469 CONF 960715 NUREG/CP-0153 SAND 96-2445 NUREG/CR4433 EUR 16771 NUREG/CR4523 V01 SAND 97 0335 NUREG/CR4523 V01 EUR 16771 NUREG/CR4523 V02 SAND 97 0335 NUREG/CR-6523 V02 EUR 16772 NUREG/CR4526 V01 SAND 97 0807 NUREG/CR-4674 V24 EUR 16772 NUREG/CR4526 V02 SAND 971039 NUREG/CR4527 FACA NUREG/CR4513 N01 SAND 971735 NUREG/CR4533 INEL-94/0061 NUREG/CR4541 R02 SAND 97-2323 NUREG/CR4526 V01 INEL-94/0278 NURFG/CR-5229 V09 SAND 97 2323 NUREG/CR4526 V02 IMEL 96/0089 NUREG/CR4456 SAND 97-2544 NUREG/CR4543 IMEL 96/0219 NUREG/CR4478 SAND 97 2776 NUREG/CR4547 NEA/CNRA/R(96)1 NUREG/CP 0154 SANL94-1676 NUREGiCR4530 NEA/CSNI/R(96!3 NUREG/CP-0158 SEA 96-3104-A;4 NUREG/CR4371 NEA/CSNI/R(97)4 NUREG/CP-0159 SEA 963104010A:3 NUREG/CR4370 ORNL4886 NUREG/CR4399 UCRL ID-122160 NUREG/CR4372 V01 ORNL4088 NUREG/CR-6379 UCRL-ID-122160 NUREG/CR4372 V02 33

l c

l 3

i

(

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

reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by the author. If further information is needed, refer to the main cita-l tion by the NUREG number.

A380TT,M.L SARSER.D.E.

NUREG/CR4523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA-UNCERTAINTY ANALYSIS. Food Cham Uncertamty Assessment.Mam TON OF RADIOLABELED ANTIBODIES CONTAINING Y-90,TC 99M,6-NurbCR4523 V02: PROBABlLISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS.

Food Chain Uncertamty BARKER,T.G.

Assessment. Appendices.

NUREG/CR4448 V02: EVALUATON OF NATONAL SEISMOGRAPH ALEXANDER,0.J.

NETWORK DETECTON CAPABILITIES. Final Report.

NUREG/CR4379: AN IMPROVED CORRELATION PROCEDURE FOR BASSETTAL SUBSIZE AND FULL-SIZE CHARPY IMPACT SPECIMEN DATA.

NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH ALLEN,K.

SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

NUREG-1516; MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY NUREG/CR4497: DATA COLCCTON AND FIELD EXPERIMENTS AT PROGRAMS AT MEDICAL FACILITIES. Final Report.

THE APACHE LEAP RESEARCH SITE.May 1995 1996.

ALLEN,M.D.

BAUM,J.W.

NUREG/CR4469: EXPERIMENTS TO INVESTIGATE DIRECT CON-NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE TON OF RADIOLABELED ANTIBODIES CONTAINING Y 90,TC-99M,l-CALVERT CLIFFS NUCLEAR POWER PLANT.

131, AND LU-177.

NU EG/CR4531: EFFECTS OF RADCACTIVE HOT PARTICLES ON NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABILISTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND BELLES,R.J.

N REG 2 0 R

'MENDATIONS FOR PROBABILISTIC D MAG NS 9 tus SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND USE OF EXPERTS. Appendices.

BENNETT,T.J.

ARCHAMBEAU,J.O NUREG/CR.6448 V02: EVALUATON OF NATIONAL SEISMOGRAPH NUREG/CR4531$ EFFECTS OF RADIOACTIVE HOT PARTICLES ON NETWORK DETECTON CAPABILITIES. Final Report.

PIG SKIN.

BERMUDEZ,H.

ARREDONDO,S.A.

NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY NUREG 1556 V01: CONSOLIDATED GUIDANCE ABOUT MATERIALS PROGRAMS AT MEDICAL FACILITIES. Final Report.

LICENSES. Pro 0 ram-Specif c Gaiance About Portable Gauge NUREG C 437C. BLOCKAGE 2.5 USER'S MANUAL AYRES,D.A.

NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL NU EG 1601: CHEMICAL PROCESS SAFETY AT FUEL CYCLE FACILi-BE M E NUREG/CR4534 V01: FRAPCON-3: MODIFICATIONS TO FUEL ROD AZARM.M.A.

MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH-NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF BURNUP APPLICATON.

GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR POWER PLANTS.

BEZLER,P.

NUREG/CR4414: PIPING BENCHMARK PROBLEME FOR THE WES-BAGGETT,8.

TINGHOUSE AP600 STANDARDIZED PLANT.

NUREG-1556 V3 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE.

l RIALS LICENSES. Applications for Sealed Source And Device Evalua-pi,Y.

tion And Registration. Draft Report For Comment NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-r STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS OF NUCLEAR SYOTEMS.

U E /CR4514: ANALYSIS OF POTENTIAL SELF GUARANTEE TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON' BINER,S.B.

l PROFIT COLLEGES. UNIVERSITIES, AND HOSPITALS AND BY BUSI-NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-NESS FIRMS THAT DO NOT ISSUE BONDS.

STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS BAKHTIARI,8.

OF NUCLEAR SYSTEMS.

NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY BLANCHAT T.K.

PROGRAM. Semiannual Report August 1995. March 1996.

NUREG/CR4469-EXPERIMENTS TO INVESTIGATE DIRECT CON.

BANDYOPADHYAY,K TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE NUREG/CR4464: AN EVALUATION OF METHODOLOGY FOR SEIS-CALVERT CLIFFS NUCLEAR POWER PLANT.

MIC QUALIFICATION OF EQUIPMENT CABLE TRAYS, AND DUCTS NUREG/CR4530: DELIBERATE IGNITION OF HYDROGEN-AIR. STEAM IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

SAOER S.

90ARDMAN.J.

NUREG/CR4563: LG EXCITATION, ATTENUATION, AND SOURGE NUREG/CR4526 V01: PROBABILISTIC ACCIDENT CONSEQUENCE SPECTRAL SCALING IN CENTHAL AND EASTERN NORTH AMER.

UNCERTAINTY ANALYSIS. Uncertainty Asse:sment For Deposited ICA.

Material And Extemal Doses. Main Report.

35 1

I l

36 Personal Author index NUREG/CR4526 V02: PROBAB!USTIC ACCOENT CONSEQUENCE NUREG/CR4446: FRACTURE TOUGHNESS EVALUATONS OF TP304 UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-STAINLESS STEEL PIPES.

POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

BUCKJ.W.

DOCCIO).L NUREG/CR4566: DESCRIPTON OF MULTIMEDIA ENVIRONMENTAL NUREG/CR4391: DETONATON CELL SIZE MEASUREMENTS IN POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSON 3.2 MODl-HIGH-TEMPERATURE HYDROGEN. AIR STEAM MIXTURES AT THE FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

BNL HIGH TEMPERATURE COMBUSTON FACILITY.

BUDNITZ,RJ.

NUREG/CRM72 V01: RECOMMENDATIONS FOR PROBABluSTIC UR G/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPI-SEISMC HAZARD ANAWSIS. MDANCE ON UNCERTAW AND j

CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCl-USE OF EXPERTS Main Report DENT CONDITIONS' NUREG/CR4372 V02: RECOMMENDATIONS FOR PROBABIUSTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND BOORE D.M.

NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABluSTIC USE OF EXPERTS.Appendees.

SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND BUEHRING,W.A.

USE OF EXPERTS. Main Report NUREG/CR4372 V02: RECOMMENDATIONS FOH PROBABluSTIC NUREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NUCLE-SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND AR ELECTRICITY GENERATING UNITS IN THE UNITED STATES:

1997-2001.

USE OF EXPERTS.Appendees.

BOUCHER,TJ.

BURGESS,M.

NUREG/CR4541 R02: PHENOMENA IDENTIF; CATION AND RANKING NUREG-1556 V3 DRF FC: CONSOLOATED GUIDANCE ABOUT MATE-TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOi.,S-OF-RIALS LICENSES.Appicatons for Scaled Source And Dev6ce Evalua-COOLANT ACCIDF.NT, MAIN STEAM UNE BREAK, AND STEAM tion And Registration. Draft Report For Comment GENERATOR TV3E RUPTURE SCENARIOS.

BURNS,R.E.

[ TOR N NURE CFkM167: LATE. PHASE MELT PROGRESSION EXPERIMENT M

EC NOL Y

}

MP-2.Results And Analysis.

BURTT,J.D.

BOWMAN,$.M NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING NUREG/CR 6361: CRITICALITY BENCHMARK GUIDE FOR UGHT.

TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSSOF.

WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE COOLANT AOCIDENT, MAIN STEAM LINE BREAK, AND STEAM PACKAGES' GENERATOR TUBE RUPTURE SCENAROS.

BRAMWELL.D.

NUREG/CR4478: MOTOR 4PERATED VALVE (MOV) ACTUATOR CAMP,A.L MOTOR AND GEARBOX TESTING.

NUREG/CR4042 RO1: PER$PLCTIVES ON REACTOR SAFETY.

BRAVERMAN,J.

CAMPBELL,V.

NUREG/CR4414: PIPING BENCHMARK PROBLEMS FOR THE WES-NUREG 1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY TINGHOUSE AP600 STANDARDIZED PLANT.

PROGRAMS AT MEDICAL FACIUTIES. Final Report NUREGICR4486: ASSESSMENT OF MODULAR CONSTRUCTION FOR SAFETY.RELATED STRUCTURES AT ADVANCED NUCLEAR CAMPER,LW.

POWER PLANTS-NUREG 1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY PROGRAMS AT MEDICAL FACluTIES, Final Report BRIDEAU,J.

NUREG-1562 DRFT FC: STANDARD REVIEW PLAN FOR APPUCA.

TIONS FOR UCENSES TO DISTRIBUTE BYPRODUCT MATERIAL TO

/

i OC RE ANUAL PERSONS EXEMPT FROM THE REQUIREMENTS FOR AN NRC BROADDUS,D.

UCENSE.10CFR Parts 30.14,30.15,30.16,30.18,30.19 & 30.20.

NUREG-1556 V3 DAF FC: CONSOUDATED GUIDANCE ABOUT MATE-CARRICO,JA.

RIALS UCENSES. Applications for Sealed Source And Device Evalua.

NUREG-1556 V2 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-tion And Registration. Draft Report For Comment RIALS LICENSES. Program Specife Guidance Ahout industrial Radog-BROADHEAD,B.L raphy Ucennes. Draft Report For Use And Comment NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICAUTY SUDE RULE.Technmal Basts.

CARSTEN,A.L NUREG/CR4505 V01: THE POTENTIAL FOR CRITICAUTY FOLLOW.

NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA.

ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE TION OF RADIOLABELED ANTIBODIES CONTAINING Y 90TC 99M,l-FACluTIES. Uranium Blended With Soil.

131, AND LU 177.

NUREG/CR 6531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON BROWNJ.

PlG SKIN.

NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Food Chein Uncertainty AssessmentMain CASTLETON,K.J.

Report NUREG/CR4506: DESCRIPTION OF MULTIMEDIA ENVIRONMENTAL NUREG/CM.,523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-UNCERTAINTY ANALYSIS.

Food Chain Uncertainty FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

Assessment.Appendcas.

CHAMBERS,D.B.

BROWN,W.

NUREG/CR4481 V01: REVIEW OF MODELS USED FOR DETERMIN-NUREG/CR4393: INTEGRATED SYSTEM VAUDATON: METHODOLO-ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model GY AND REVIEW CRITERIA.

Evaluation Criteria.

NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-BRUMMETT,E.

NUREG 1532: FINAL TECHNICAL EVALUATON REPORT FOR THE ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report PROPOSED REVISED RECLAMATION PLAN FOR THE ATLAS COR-PORATON MOAB MILLSource Material Ucense No. SUA-917. Docket CHANIN.D.L NUREG/CR4547: DOSFAC2 USER'S GUIDE.

No. 40-3453.(Atlas Corporation)

BRUST,F.W.

CHEN,0.

NUREG/CR-4667 V23; ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR 6497: DATA COLLECTON AND FIELD EXPERIMENTS AT UGHT WATER REACTORS Semiannual Report.Juty-December 1996.

THE APACHE LEAP RESEARCH SITE.May 1995 - 1996.

Personal Author Index 37 CHENG.H.S.

COOKE,R.M.

NUREG/CR4474: PREUMINARY PHENOMENA IDENTIFICATION AND NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE RANKING TABLES (PIRT) FOR SBWR STARTUP STABluTY.

UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessmentMain Report CHEUNG,F.B.

NUREG/CR4523 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A UNCERTAINTY ANALYSIS.

Food Chain Uncertainty DOWNWARD FACING CURVED SURFACE.

AssessmentAppendices.

NUREG/CR4526 Vot: PROBABILISTIC ACCIDENT CONSEQUENCE CHILDS.R.L UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited NUREG/CR4504 V01: AN U0 DATED NUCLEAR CRITICAUTY SUDE Matenal And Extemal Doses. Main Report RULE. Technical Basas.

NUREG/CR4526 V02: PROBABILISTt ACCIDENT CONSEQUENCE i

UNCERTAINTY ANALYSIS. UNCERM,4TY ASSESSMENT FOR DE-CHOPRA,0.K.

POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

NUREG/CR4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Senannual Report. January 1996 June COPINGER,0.A.

1996.

NUREG/CR-4874 V23: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR-4867 V23: ENVIRONMENTALLY ASSISTED CRACKING IN DAMAGE ACCIDENTS: 1995. A Status Report UGHT WATER REACTORS. Somsannual Report, July December 1996.

NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY COPPERSMITH,K.

l PROGRAM. Semiannual Report. August 1995 - March 1996.

NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABILISTIC 1

SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND l

CHOWDHURY,A.H.

USE OF EXPERTS Main Report NUREG/CR 6404: AN EXPERIMENTAL SCALE MODEL STUDY OF NUREG/CR-6372 V02: RECOMMENDATIONS FOR PROBABILISTIC SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINTED SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND ROCK MASS.

USE OF EXPERTS. Appendices.

CHU,T-L CORNELL,C.A.

NUREG/CR-6538. EVALUATION OF LOCA WITH DELAYED LOOP AND NUREG/CR4372 VOI: RECOMMENDATIONS FOR PROBABILISTIC LOOP WITH DELAYED LOCA ACCIDENT SCENARIOS.

SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND USE OF EXPERTS Main Report CHUNG,H.M.

NUREG/CR-6372 V02: RECOMMENDATIONS FOR PROBABILISTIC NUREG/CR.4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN SEISMIC HAZARD ANALYSIS; GUIDANCE ON UNCERTAIN 1Y AND UGHT WATER REACTORS. Semsannual Report. January 1996 - June USE OF EXPERTS. Appendices.

1996.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN CORWIN,W.R.

LIGHT WATER REACTORS. Semiannual Report. July-December 1996.

NUREG/CR-5591 V07 N1: HEAVY SECTION STEEL IRRADIATION PROGRAM.Serniannual Progress Report For October 1995 Through CICCARELL1,G.

March 1996.

NUREG/CR4391: CETONATION CELL SIZE MEASUREMENTS IN NUREG/CR-5591 V07 N2: HEAVY-SECTION STEEL IRRADIATION HIGH-TEMPERATURE HYDROGEN-AIR. STEAM MIXTURES AT THE PROGRAM. Semiannual Progress Report For April Through September BNL HIGH TEMPERATURE COMBUSTION FACluTY.

1996.

CLETCHER,J.W.

COUTTS.P.T.

NUREG/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-DAMAGE ACCIDENTS: 1995. A Status Report.

ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report CLUFF,LS.

DANZlGER.LM.

NUREG 1603 DRFT:

INDIVIDUAL PLANT EXAMINATION NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABluSTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND DATABASE. User's Guide.

USE OF EXPERTS Mam Report DASAPPA V'R4514:

NUREG/CR4372 V02: RECOMMENDATIONS FOR PROBABILISTIC NUREG/C ANALYSIS OF POTENTIAL SELF-GUARANTEE SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-USE OF EXPERTS. Appendices.

PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY BUSI.

COLLIER J.

NESS FIRMS THAT DO NOT ISSUE BONDS.

l NUREG/CR4514: ANALYSIS OF POTENTIAL SELF-GUARANTEE DAVIDSON,G.R.

TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY BUSI-SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

NESS FIRMS THAT DO NOT ISSUE BONDS-NUREG/CR4497: DATA COLLECilON AND FIELD EXPERIMENTS AT THE APACHE LEAP RESEARCH SITE.May 1995 - 1996.

COLLtNS,D.J.

NUREG 1556 V01: CONSOLIDATED GUIDANCE ABOUT MATERIALS DAVIS,C.B.

UCENSES. Program-Specific Guidance About Portable Gauge NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING Ucenses. Final Report TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF.

NUREG 1556 V2 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-COOLANT ACCIDENT, MAIN STEAM UNE BREAK, AND STEAM l

RIALS UCENSES.Prograin Specific Guidance About Industnal Radio 9-GENERATOR TUBE RUPTURE SCENARIOS.

raphy Ucenses. Draft Report For Use And Comment.

l NUREG-1556 V5 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE.

DAVi$,F.J.

f RIALS LICENSES. Program-Specific Guidance About Self-Shielded Irra.

NUREG/CR-6533: CODE MANUAL FOR CONTAIN 2.0 A COMPUTER diator Licenses. Draft Report For Comment CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

I COLTEN-BRADLEY DAVIS.MJ.

NUREG/CR4505 VO1: THE POTENTIAL FOR CRITICAUTY FOLLOW-NUREG-1574: STANDARD REVIEW PLAN ON ANTITRUST ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE REVIEWS. Final Re_ port FAC'LITIES. Uranium Blended With Soll.

NUREG-1574 DRFT FC: STANDARD REVIEW PLAN ON ANTITRUST. Draft Report For Comment.

COMPTON,E NUREG-1556 V3 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE.

DAYlS,R.E.

RIALS UCENSES. Applications for Sealed Source And Device Evalua.

NUREG/CR4295: REASSESSMENT OF SELECTED FACTORS AF.

tion And Regratration. Draft Report For Comment FECTING SITING OF NUCLEAR POWER PLANTS.

NUREG/CR-6451: A SAFETY AND REGULATORY ASSESSMENT OF CONNELLY,S.R.

GENERIC BWR AND PWR PERMANENTl.Y SHUTDOWN NUCLEAR NUREG-1542 V02: ACCOUNTABILITY REPORT FISCAL YEAR 1996.

POWER PLANTS.

38 Personal Author index -

DEAN,C.

FIRST.M.W.

NUREG/CR4514: ANALYSIS OF POTENTIAL SELF-GUARANTEE NUREG/CP 0153: PROCEEDINGS OF THE 24TH DOE /NRC NUCLEAR TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-AIR CLEANING AND TREATMENT CONFERENCE. Head h Portland, PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY BUSI-Oregon, July 15 18,1996.

NESS FIRMS THAT 00 NOT ISSUE BONDS.

FLETCHER,C.D.

l DE90RDS K NUREG/CR-6541 R02: PHENOMENA IDENTIFICATION AND RANKING l

NUREG/CR4636: DEVELOPMENT OF CONFORMAL RESPIRATOR TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF-MONITORING TECHNOLOGY.

Cf OLANT ACCIDENT, MAIN STEAM UNE BREAK, ArdD STEAM

( 2NERATOR TUBE RUPTURE SCENARIOS.

DECKER,0.

NUREG/CR4463 R01: REVIEW GUCEUNES FOR SOFTWARE LAN.

FUEW 4

A POWER W SAFEW NUREGI532: FINAL TECHNICAL EVALUATION REPORT FOR THE YSTEMS.Fina R PROPOSED REVim RECLAMATON PMN FOR THE ATLAS COR.

i DEGRASSt,G.

PORATION MOAS ELSource Material License No. SUA-917. Docket NUREG/CR4414: PIPING BENCHMARK PROBLEMS FOR THE WES.

No. 40-3453.(Atlas Corporation)

TINGHOUSE AP600 STANDARDIZED PLANT.

FORESTER,J.A.

DEHART.M.D.

NUREG/CR 4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CH4361: CRITICAUTY BENCHMARK GUIDE FOR LIGHT.

DAMAGE ACCIDENTS: 198243.A Status Report.

Y!ATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE PACKAGES.

FOX,0.J.

NUREG/CR4404: AN EXPERIMENTAL SCALE-MODEL STUDY OF RE CR4478: MOTOR OPERATED VALVE (MOV) ACTUATOR KM MOTOR AND GEARBOX TESTING.

A "'

DIERCKS,0.R.

NUREG/CR4233 V02: STABluTY OF CRACKED PIPE UNDER SEIS-NUREG/CP 0154: PROCEEDINGS OF THE CNRA/CSNI WORKSHOP MIC/ DYNAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask ON STEAM GENERATOR TUBE INTEGRITY IN NUCLEAR POWER 1.2 Final Report.

PLANTS AU9EG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT-PIPE PROGRAM.Sermannual Report, August 1995 - March 1996.

LOCATONS. Final ReportSeptember 1991 November 1995.

DINSMORE.G.

NUREG/CR4463 RO1: REVIEW GUOEUNES FOR SOFTWARE LAN.

FUHRMANN,M.

GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW.

SYSTEMS Final Report.

LEVEL WASTE DATA BASE DEVELOPMENT PROORAM FOR FISCAL YEAR 1996. Annual Report.

NUREG/CR4181 R01: A PILOT APPUCATION OF RISK-INFORMED FULLEN,M.

METHODS TO ESTABLISH INSERVICE INSPECTON PRIORITIES NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR PROGRAMS AT MEDICAL FACILITIES. Final Report.

1 NWER STATION.

GARVER M.

N REG /C'R 4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE TION IG S 7

DAMAGE ACCOENTS: 1995. A Status R NUREG/CR.4674 V24: PRECURSORS TO TENTIAL SEVERE CORE GASSER,R.D.

DAMAGE ACCIDENTS: 1982-83.A Status Report NUREG/CR4167: LATE-PHASE W.ELT PROGRESSION EXPERIMENT DONG,P.

MP-2.Results And Analysis.

NUREG/CR-4667 V23: ENVIRONMENTALLY AS: "ED CRACKING IN l

UGHT WATER REACTORS. Sermanqual Report, July-December 1996.

N G CR4167: LATE-PHASE MELT PROGRESSION EXPERIMENT EYER,H.R.

MP-2.Results And Analysis.

NUREG/CR.5661: RECOMMENDATIONS FOR PREPARING THE CRITI-NUREG/CR4527: FINAL RESULTS OF THfi XR21 BWR METALUC CAUTY SAFETY EVALUATION OF TRANSPORTATION PACKAGES.

MELT RELOCATION EXPERIMENT.

EASTERLY,C.E.

GAVENDA.D.J.

NUREG/CR4526: ENVIRONMENTAL ASSESSMENT PROPOSED U.

NUREG/CR-4867 V22: ENVIRONMENTALLY ASSISTED CRACKING IN CENSE RENEWAL OF NUCLEAR METALS.INC. CONCORD, MASSA.

LIGHT WATER REACTORS. Sem6 annual Report, January 1996 June CHUSETTS.

1996.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN ELLIOT,SJ.

LIGHT WATER REACTORS. Semiannual Report, July-December 1996.

NUREG 1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-TABASE.

GEDDIS,AR NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH 8"

E 5 V02 A05: SAFEGUARDS

SUMMARY

EVENT LIST (S$2L) January 1,1990 Through December 31,1996.

GEE.G.W.

NUREG/CR4565: UNCERTAltdTY ANALYSES OF INFILTRATION AND FAIDY,C.

SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

NUREG/CP4155: PROCEEDINGS OF THE SEMINAR ON LEAK BEFORE BREAK IN REACTOR PIPING AND VESSELS.

GELSTON,G.M.

FAIRSANKS,CJ.

NUREG/CR4568: DESCRIPTON OF MULTIMEDIA ENVIRONMENTAL i

1 NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA.

POLLUTANT ASSESSMENT SYSTEM (MEPAS) W:RSION 3.2 MODI-TABASE.

FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

FINFROCK,C.

GERLACH,L.

NUREG/CR4301: DETONATON CELL SIZE MEASUREMENTS IN NUREG/CR4391: DETONATON CELL SIZE MEASUREMENTS IN HIGH. TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE HIGH TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE BNL HIGH-TEMPERATURE COMBUSTION FACluTY.

BNL HIGH-TEMPERATURE COMBUSTION FACluTY.

)

)

i Personal Author index 39 GHADeAu,N.

GuzMAN.A.G.

NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY NUREG/CR4459-FIELD STUDIES AT THE APACHE LEAP RESEARCH RESEARCH GROUP (IPIRG-2) PROGRAM Final Report SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

REG /CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER CODE FOR "UCLEAR REACTOR CONTAINMENT ANALYSIS.

NUREG/CR4507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A DOWNWARD FACING CURVED SURFACE.

G4LLES PH.

NUREG/CP-0155: PROCEEDINGS OF THE SEMINAR ON LEAK HAGEMEYER D.

BEFORE BREAK IN REACTOR PIPING AND VESSELS.

NUREG 0713 V17: OCCUPATIONAL RADIATION EXPOSURE AT COM-MERICAL NUCLEAR POWER REACTORS AND OTHER GINGSERG,T.

FACILITIES.1995. Twenty. Eighth Annual Report NUREG/CR4391: DETONATION CELL SIZE MEASUREMENTS IN HIGH TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE HAGGAG F.M.

BNL HIGH-TEMPERATURE COMBUSTION FACluTY.

NUREG/CR4363: EFFECTS OF THERMAL AGING AND NEUTRON 1R-RADIATION ON THE MECHANICAL PROPERTIES OF THREE WIRE GOLDOERG#

l NUREG/CRd514: ANALYSIS OF POTENTIAL SELF GUARANTEE DM TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-HAMMONDS,J.S.

P ER IES, A D tOSPITALS AND BY BUSI-N T

O NUREG/CR4481 V02: REVit.W OF MODELS USED FOR DETERMIN-ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report GOOSSENS,LH.

NUREG/CR4526 V01: PROBABluSTIC ACCIDENT CONSEQUENCE HANSON,A.L UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposated NUREG/CR-6295: REASSESSMENT OF SELECTED FACTORS AF-Material And External Doses Mam Report FECTING SITING OF NUCLEAR POWER PLANTS.

NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE.

HARDIN,E.L POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

OCOSSENS.LH.J.

NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT NUREG/CR4523 V01: PROBADIUSTIC ACCIDENT CONSEQUENCE THE APACHE LEAP RESEARCH SITE.Mey 1995 1996.

UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssesernentMain Report HARPER,F.T.

NUREG/CR4523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR4523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS.

Food Chain Uncertainty UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessmentMan Assessment.Appendo.es-Report GORE 5 F.

NUREG/CR4523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR4181 RO1: A PILOT APPUCATION OF RISK-INFORMED UNCERTAINTY ANALYSIS.

Food Chain Uncertainty METHODS TO ESTABLISH INSERVICE INSPECTION PRIORITIES AssessmentAppendices.

FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR NUREG/CR4526 V01: PROBABILISTIC ACCIDENT CONSEQUENCE POWER STATION.

UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited Material And External Doses. Main Aeport GRAFF,S.

NUREG/CR4526 V02: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4463 ROI: REVIEW GUIDEUNES FOR SOFTWARE LAN.

UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY POSITED MATE'ilAL AND EXTERNAL DOSES. Appendices.'

]

SYSTEMS. Final Report.

l HASKIN,F.E.

GREEN,W.

NUREG/CR4042 R01: PERSPECTIVES ON REACTOR SAFETY.

NUREG/CR4463 R01: REVIEW GUIDEUNES FOR SOFTWARE LAN-NUREG/CR4523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE i

GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY UNCERTAINTY ANALYSIS. Food Chan Uncertainty AssessmentMain SYSTEMS Final Report Report NUREG/CR4523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE

)

^

l NURE 562 DRFT FC: STANDARD REVIEW PLAN FOR APPLICA-

• • • • • • • • " E TIONS FOR LICENSES TO DISTRIBUTE BYPRODUCT MATERIAL TO PERSONS EXEMPT FROM THE REQUIREMENTS FOR AN NRC HECHT,M.

UCENSE.100FR Parts 30.14,30.15, 30.16,30.18,30.19 & 30.20.

NUREG/CR4463 R01: REVIEW GUIDEUNES FOR SOFTWARE LAN-GRIFFITH,P.

GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY NUREG/CR-6519. SCREENING REACTOR STEAM / WATER PIPING SYSTEMS. Final Report 1

l l

SYSTEMS FOR WATER HAMMER.

p I

GRIFFITH.R.0, NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY i

l l

NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER PROGRAMS AT MEDICAL FACILITIES. Final Report l

CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

NUREG-1556 V4 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-

[

RIALS UCENSES. Program Specific Guidance About Fixed Gauge GROVE E.J.

Ucenses. Draft Report For Comment NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR HIGGINS,J.

l POWER PLANTS.

NUREG/CR4393: INTEGRATED SYSTEM VALIDATION: METHODOLO-GY AND REVIEW CRITERIA.

NUREG/CR4400: HUMAN FACTORS ENGINEERING (HFE) INSIGHTS N E 4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN

^ ^"

"^

^

^"

LIGHT WATER REACTORS. Semiannual Report January 1996 - June E'

1996.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN HI UGHT WATER REACTORS. Semiannual Report, July-December 1996.

N G CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES QUZIELK.A.

TRENCH SITE: EXPERIMENT IIB.

NUREQ/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NUCLE-AR ELECTRICITY-GENERATING UNITS IN THE UNITED STATES:

HOOGE S.A.

1997 2001.

NUREG/CR4042 RO1: PERSPECTIVES ON REACTOR SAFETY.

40 Personal Author index HOFFMAN,F.O.

ILLMAN,W.A.

NUREG/CR4481 V01: REVIEW OF MODELS USED FOR DETERMIN-NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model THE APACHE LEAP RESEARCH SITE.May 1995 1996.

Evaluation Critena.

NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN.

ISKANDER,8.K.

ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report NUREG/CR4399: RESULTS OF CHARPY V-NOTCH IMPACT TESTING OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 DE.

HOFMAYER.C.

GREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMERCIAL RE.

NUREG/CR4486: ASSESSMENT OF MODULAR CONSTRUCTION FOR ACTOR CAVITY.

SAFETY.RELAT ED STRUCTURES AT ADVANCED NUCLEAR POWER PLANTS.

JASTROW,J.D.

NUREG/CR 5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW.

HOOPES,5.L LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR4566: DESCRIPTON OF MULTIMEDIA ENVIRONMENTAL FISCAL YEAR 1996. Annual Report POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

JENSEN.J.J.

NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE HOPPERA DAMAGE ACCIDENTS: 1982-83.A Status Report NUREG/CR4389: IPIRG 2 TASK 1 PIPE SYSTEM EXPERIMENTS WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT. PIPE JILES,D.C.

LOCATIONS Flnal ReportSeptember 1991. November 1995.

NUREGICR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS RESEARCH GROUP (IPIRG 2) PROGRAM Final Report OF NUCLEAR SYSTEMS.

HOPPER C,M, JOHNSON,T.

NUREG/CR4361: CRITICAUTY BENCHMARK GUIDE FOR UGHT.

NUREG-1532: FINAL TECHNICAL EVALUATON FTPORT FOR THE WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE PROPOSED REVISED RECLAMATION PLAN FOR THE ATLAS COR-PACKAGES.

PORATION MOAB MILLSource Matenal Ucense No. SUA-917. Docket NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICAUTY SLIDE No. 40-3453.(Atlas Corporation)

RULE.Tecnnical Basis.

NUREG/CR4505 V01: THE POTENTIAL FOR CRITICALITY FOLLOW-JONES J.

ING DISPOSAL OF URAN!UM AT LCW LEVEL WASTE NUREG 1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY FACluTIES Uranium Blended With Soil.

PROGRAMS AT MEDICAL FACILITIES. Final Report HORA.S.C.

JONES.J.A.

NUREG/CR4523 V01: PROBABillSTIC ACCOENT CONSEQUENCE NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessmentMain UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessmentMain NUR CR-6523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUR CR-6523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS.

Food Chain Uncertainty UNCERTAINTY ANALYSIS.

Food Chain Uncertainty Assessment. Appendices.

AssessmentAppendices.

NUREG/CR4526 V01: PROnABluSTIC ACCIDENT CONSEQUENCE NUREG/CFv6526 V01: PROBABluSTIC ACCOENT CONSEQUENCE UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited Material And External Doses. Main Report Material And External Doses. Main Report NUREG/CR4526 V02: PROBABlUSTIC ACCIDENT CONSEQUENCE NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE.

POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

HSlUNG S.

JONES,S.C.

NUREG/CR4404: AN EXPERIMENTAL SCALE MODEL STUDY OF NUREG/CR-6037: MEASUREMENT OF RESOUAL RADIOACTIVE SUR-SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINTED FACE CONTAMINATION BY 2-D LASER HEATED TLD.

ROCK MASS.

JONES,W.R.

HUGHES,T.H.

NUREG-1275 V12:

OPERATING EXPERIENCE FEEDBACK NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN REPORT. Assessment Of Spent Fuel Cooling.

UGHT WATER REACTORS. Semiannual Report. January 1996 June 1996.

JUSTUS,P.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG-1532: FINAL TECHNICAL EVALUATION REPORT FOR THE UGHT WATER REACTORS. Semiannual Fleport, July-December 1996.

PROPOSED REVISED RECLAMATION PLAN FOR THE ATMS COR-PORATON MOAB MILLSource Material Ucense No. SUA-917. Docket HUMPHREYS,S.L No. 40-3453.(Atlas Corporation)

NUREG/CR4525: SECPOP90: SECTOR POPULATION. LAND FRAC.

TION, AND ECONOMIC ESTIMATON PROGRAM.

KAM,F.B.K.

NUREG/CR4454: POOL CRITICAL ASSEMBLY PRESSURE VESSEL HUMPHRIES,LL FACILITY BENCHMARK.

NUREG/CR 6167: LATE-PHASE MELT PROGRESSION EXPERIMENT MP-2.Results And Analysis.

KANA,0.D.

NUREG/CR4527: FINAL RESULTS OF THE XR21 BWR METALUC NUREG/CR4404: AN EXPERIMENTAL SCALE-MODEL STUDY OF j

MELT RELOCATON EXPERIMENT.

SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINTED l

ROCK MASS.

HUTTO.F.R.

NUREG/CR4464: AN EVALUATION OF METHODOLOGY FOR SEIS-NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN MIC OUALIFICATON OF EQUIPMENT CABLE TRAYS, AND DUCTS IONS ON THE EXTERNAL STRESS CORROSION CRACKING OF IN ALWR PLANTS BY USE OF EXPERIENCE DATA-TYPE 304 AUSTENITIC STAINLESS STEEL KARWOSKl,K.J.

ISARRA,J.G.

NUREG-1604: CIRCUMFERENTIAL CRACKING OF STEAM GENERA.

NUREG-1275 V12.

OPERATING EXPERIENCE FEEDBACK TOR TUBES.

l REPORT. Assessment Of Spent Fuel Cooling.

KASSNER,T.F.

ISRAHIMA NUREG/CR-4657 V22: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG-1532: FINAL TECHNICAL EVALUATION REPORT FOR THE UGHT WATER REACTORS. Semiannual Report, January 1996 June PROPOSED REVISED RECLAMATON PLAN FOR THE ATuS COR-1996.

PORATION MOAB MILLSource Material Ucense No. SUA-917. Docket NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN No. 40-3453 (Atlas Corporat6on)

UGHT WATER REACTORS. Sennannual Report. July December 1996.

l

l Personal Author Index 41 KASZA,K.E.

KUPPERMAN D.S.

NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY NUREG/CR-6511 V01: STEAM GENERATOR TUBE INTEGRITY PROGRAM.Semeannual Report, August 1995 March 1996.

PROGRAM.Sermannual Report. August 1995. March 1996.

KAURIN,0.G.L LAMBE,W.M.

NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA-NUREG-1574: STANDARD REVIEW PLAN ON ANTITRUST TION OF RADIOLABELED ANTIBODIES CONTAINING Y 90,TC-99M,1-REVIEWSFinal Report.

131, AND LU-177.

NUREG-1574 DRFT FC: STANDARD REVIEW PLAN ON NUREG/CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON ANTITRUST. Draft Report For Comment.

PlO SKIN.

LANIK.G.F.

KELLOGG,J.N.

NUREG-1275 V12: OPERATING EVEAINE FEEDBACK NUREG/CR4529: VAUDATION OF TECTONIC MODELS FOR AN IN-REPORT. Assessment Of Spent Fuel Cooling.

TRAPLATE SEISMIC ZONE, CHARLESTON.SOLTTH CAROUNA WITH GPS GEODETIC DATA.

LANNING,D.D.

NUREG/CR4534 V01: FRAPCON.3: MODIFICATIONS TO FUEL ROD KENNEDY,R.P.

MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH-NUREG/CR4464: AN EVALUATION OF METHODOLOGY FOR SEIS-BURNUP APPLICATION.

MIC QUAUFICATION OF EQUIPMENT, CABLE TRAYS AND DUCTS IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

LAYTON,M.

NUREG-192: FINAL TECHNICAL EVALUATION REPORT FOR THE KHAN.H.J.

PROPOaED REVISED RECLAMATION PLAN FOR THE ATLAS COR-NURG/CR4474: PRELIMINARY PHENOMENA IDENTIFICATION AND PORATION MOAB MILLSource Material Ucense No. SUA-917. Docket RANKING TABLES (PIRT) FOR SBWR STARTUP STABluTY.

No. 40-3453.(Atlas Corporation)

KHAN,T.A.

LEE.A.D NUREG/CR 4409 V06: DATA BASE ON DOSE REDUCTION PROJECTS NUREG 1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-FOR NUCLEAR POWER PLANTS.

TABASE.

KILINSKl.T.

S S.

NUREG/CR4452: THE SECOND INTERNATIONAL PIPtNG INTEGRITY ppO 1611: AGING MANAGEMENT OF NUCLEAR POWER PLANT RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report.

CONTAINMENTS FOR LICENSE RENEWAL KINSEY,R.R.

LEHNER,J.R.

NUREG/CR4515: BLT-EC (BREACH, LEACH, AND TRANSPORT EQUL NUREG-1603 DRFT:

INDIVIDUAL PLANT EXAMINATION l

UBRtVM CHEMISTRY) DATA INPUT GUIDE.A Computer Model For DATABASE. User's Guide.

l Simulating Release And Coupled Geochemical Transport Of Contamp nants From A Subsurface Disposai Fac61 sty.

LEWIS,C.J.

KIRKWOOD.A.S.

NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-NUREG-1556 V4 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE.

ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report.

RIALS UCENSES. Program Specific Guidance About Fixed Gauge LEWIS.SR Ucenses. Draft Report For Comment.

NUREG 1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS j

KLAMERUS,E.W.

UCENSES. Program-Specific Guldance About Portable Gauge NUREG/CR4433. CONTAINMENT PERFORMANCE OF PROTOTYPl.

Ucenses. Final Report.

CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCI-NUREG 1556 V4 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-DENT CONDITIONS.

RIALS UCENSES. Program Specific Guidance About Fixed Gauge Licensen. Draft Report For Comment.

KOCH,8.

NUREG/CR4463 R01: REVIEW GUIDELINES FOR SOFTWARE LAN.

LICHTENWALTER GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY NUREG/CR4361: CRITICALITY BENCHMARK GUIDE FOR UGHT-WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE SYSTEMS. Final Report.

PACKAGES.

KRAAN,B.C.

NUREG/CR4526 V01: PROBABILISTIC ACCIDENT CONSEQUENCE LIN,C.C.

UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited NUREG 1603 DRFT:

INDIVIDUAL PLANT EXAMINATION Matenal And Extemal Doses. Main Report.

DATABASE. User's Guide.

NUREG/CR4526 V02: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE.

LIN,D.

POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

NUREG/CR4463 R01: REVIEW GUIDEUNES FOR SOFTWARE LAN-l GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY l

KRAAN.B.C.P.

SYSTEMS. Final Report.

NUREG/CR4523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTA:NTY ANALYSIS. Food Cham Uncertainty Assessment. Main Liu,W.C.

l Report. -

NUREG 1611: AGING MANAGEMENT OF NUCLEAR POWER PLANT l

NUREG/CR4523 V02: PROBABILISTIC ACCIDENT CONSEQUENCE CONTAINMENTS FOR UCENSE RENEWAL UNCERTAINTY ANALYSIS.

Food Chain Uncertainty Assesament. Appendices.

LIU,Y.C.

NUREG/CR4507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A KRAMER,0.

DOWNWARD FACING CURVED SURFACE.

i l

NUREG/CR4233 V02-STABluTY OF CRACKED PIPE UNDER SEIS-MIC/ DYNAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask LOGAN,R.J.

1.2 Final Report.

NUREG/CR4535: DEVELOPMENT OF CONFORMAL RESPIRATOR NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE-MONITORING TECHNOLOGY.

SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report.

KRISHNASWAMY,C.

NUREG/CR-6528 ENVIRONMENTAL ASSESSMENT PROPOSED LI-NUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPl-CENSE RENEWAL OF NUCLEAR METALS.INC. CONCORD, MASSA.

CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCl-CHUSETTS.

DENT CONDITIONS.

LUSINSKI,J.

KUO,P.T.

NUREG 1556 V3 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-NUREG-1811: AGING MANAGEMENT OF NUCLEAR POWER PLANT RIALS UCENSES.Applicatx>ns for Sealed Source And Device Evalua.

CONTAINMENTS FOR UCENSE RENEWAL tion And Registration. Draft Report For Comment.

i 42 Personal Author index LUEggERS,P.R.

MCDONALD,J.P.

NUREG/CR-4867 V22 ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4566: DESCRIPTION OF MULTIMEDIA ENVIRONMENTAL l

LIGHT WATER REACTORS. Semsennual Report, January 1996. June POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-1 NUR /CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Sermannual ReportJuly-December 1996.

MCGUIRE,S.A.

NUREG 1492: REGULATORY ANALYSIS ON CRITERIA FOR THE RE-LUKENJ NUREG/CRil074 V03: SEALED SOURCE AND DEVICE DESIGN LEASE OF PATIENTS ADMINISTERED RADIOACTIVE SAFETY TESTING. Technical Report On The Findings Of Task 4.Inves.

MATERIALFinal Report tmation Of A Failed Brachytherapy Needle Applicator.

LUND.A.L.

NUREG-1606 DRFT FC: PROPOSED REGULATORY GUIDANCE RE-NUREG-1616: FEASIBILITY OF UNDERWATER WELDING OF HIGHLY LATED TO IMPLEMENTATION OF 10 CFR 50.59 (CHANGES, TESTS, IRRADIATED IN-VESSEL COMPONENTS OF BOLLING WATER OR EXPERIMENTS). Draft Report For Comment.

REACTORS.A Literature Review.

MCLAUGHLIN,K.L NUREG/CH4448 V02: EVALUATION OF NATIONAL idlSMOGRAPH NURE C 4515: BLT-EC (BREACH, LEACH. AND TRANSPORT-EOUb NETWORK DETECTION CAPABILITIESFinal Report LIBRIUM CHEMISTRY) DATA INPUT GUIDE.A Computer Model For Simulating Release And Coupled Geochemical Transport Of Contamb MED W J nante From A Sut* afece Disposal Facility.

NURFG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-MAJUMDAR,S.

TABASE.

NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY PROGRAM.Sermannual Report, August 1995 March 1996.

MEINHOLD,C.S.

NUREG/CR4397: RADIATION SAFETY CONCERNS FOR PREGNANT N

G CR43S1: DETONATION CELL SIZE MEASUREMENTS IN HIGH TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE BNL HIGH-TEMPERATURE COMBUSTION FACILITY.

MEYER.P.D.

NUREG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION AND MANNESCHMIDT,E.

NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODp SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

FIED A 302 GRADE B PLATE MATERIALS. DATA ANALYSIS.

NUREG/CR4426 V02: DUCTILE FRACTURE TOUGHNESS OF MODI.

MILLER,R.L FIED A 302 GRADE B PLATE MATERIALS. Data Records.

NUREG/CR4528; ENVIRONMENTAL ASSESSMENT PROPOSED Li-CENSE RENEWAL OF NUCLEAR METALS.INC CONCORD, MASSA.

MARSCHALL,C.

CHUSETTS.

NUREG/CR4233 V02: STABILITY OF CRACKED PIPE UNDER SEIS-MIC/ DYNAMIC OtSPLACEMENT-CONTROLLED STRESSES. Subtask MINAR6CK,J.W.

NUREG/CR 4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE Nt E C4 V03: CRACK STABILITY IN A REPRESENTATIVE DAMAGE ACCIDENTS: 1995. A Statua Report PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE NAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask 1.3 Final DAMAGE ACCIDENTS: 198243.A Status Report popo,t NUREG/CR-6233 V04: INTERNATIONAL PIPING INTEGRITY RE-SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report MITCHELL,8.J.

NUREG/CR-6389 IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS NUREG/CR4563: LG EXCITATION, ATTENUATION, AND SOURCE WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT PIPE SPECTRAL SCALING IN CENTRAL AND EASTERN NORTH AMER-LOCATIONS. Final ReportSeptember 1991 - November 1995.

ICA.

MARTINE2,G.M.

MITCHELL.D.B.

NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

DAMAGE ACCIDENTS: 198243.A Status Report MARTINEZ GURIDI NUREG/CR4538: EVALUATION OF LOCA WITH DELAYED LOOP AND MITCHELL,M.E LOOP WITH DELAYED LOCA ACCIDENT SCENARIOS.

NUREG-1556 V5 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE.

RIALS LICENSES. Program-Specific Guldance About Self-Shielded Irra-MATSON,E.R.

distor Licenses. Draft Report For Comment.

NUREG-1556 Voi: CONSOLIDATED GUlOANCE ABOUT MATERIALS LICENSES. Pro 0rarmSpecific Gu6 dance About Portable Gauge MOHAN,R.

Licenses. Final Report NUREG/CR4452 THE SECOND INTERNATIONAL PIPING INTEGRITY RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report MAYEW NUREG/CR4556: NRC ANTITRUST LICENSING ACTIONS, 1978-1996.

MONTELEONE,S.

MAZUZAN,G.T.

NUREG/CP 0157 V01: PROCEEDINGS OF THE TWENTY-FOURTH NUREG-1610 CONTROLLING THE ATOM.The Beginnings Of Nuclear WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses.

Regulation, 1946-1962.

eion High Bumup Fuel, Contamment And Structural Agog.

NUREG/CP 0157 V02 PROCEEDINGS OF THE TWENTY-FOURTH MCCASE,D E-WATER REACTOR SAFETY INFORMATION MEETING.Rowtor Pres-NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-sure Vessel Embrittlement And Thermal Annealing. Reactor Vessel

^

NU EG CR 26 2:

LE HNE OF MODb tor tub FIED A 302 GRADE B PLATE MATERIALS.Dete Records.

NUREG/h-0157 V03: PROCEEDINGS OF THE TWENTY-FOURTH 1

WATER REACTOR SAFETY INFORMATION MEETING.PRA And HRA, j

MCCART'rfY,J.F.

NUREG/CR4505 V01: THE POTENTIAL FOR CRITICALITY FOLLOW.

And Probalmstk: Seistmc Hazard Assessment And Seismic Siting Crite-ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE ria.

NUREG/CP4161: TRANSACTIONS OF THE 1WENTY-FIFTH WATER FACILITIES.Oramum B6 ended With Soll.

REACTOR SAFETY INFORMATION MEETING.

MCCONNELL,J.W.

NUREG/CR 5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW.

MONTGOMERY,J.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY FISCAL YEAR 1996. Annual Report PROGRAMS AT MEDICAL FACILITIESFinal Report.

l

_____________________--________________-___--_-________________A

l Personal Author index 43 MORANTEJL NUREG/CR-6389: IPIRG-2 TASK 1 PIPE SYSTEM EXPERIMENTS NUREG/CR446J: ASSESSMENT OF MODULAR CONSTRUCTION FOR WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT. PIPE SAFETY-RELATED STRUCTURES AT ADVANCED NUCLEAR LOCATIONS. Final Report. September 1991 November 1995.

POWER PLANTS.

NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report.

NUREG-0540 Vt9 N06: TITLE UST OF DOCUMENTS MADE PUBUCLY ORNSTEIN,HL AVAILABLE. August 1 31,1997.

NUREG 1275 V12: OPERATING EXPERIENCE FEEDBACK REPORT. Assessment Of Spent Fuel Cooling.

NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBA81USTIC PAINTER.CL SEISMIC HAZARD ANALYSIS. GUIDANCE ON UNCERTAINTY AND NUREG/CR4534 V01: FRAPCON-3: MODIFICATIONS TO FUEL ROD USE OF EXPERTS Main R NUREG/CR4372 V02: RE MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH-MENDATIONS FOR PROBABluSTIC BURNUP APPLICATION.

(

SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND USE OF EXPERTS.Appendees.

PARKJ H.

I NUREG/CR4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN E /CR4295: REASSESSMENT OF SELECTED FACTORS AF.

HT WATER REACTORS. Semiennual Report January 1996. June FECTING SITING OF NUCLEAR POWER PLANTS.

NUREG/CR 4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN MUHLHEIM,M.D.

UGHT WATER REACTORS. Semiannual Report, July-December 1996.

NUREG/CR 4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1995. A Status Rom pggg,3 y NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY MURATA,K1 PROGRAM. Semiannual Report, August 1995 March 1996.

NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

PMK,8 H MUMRELL.M.T.

ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT Fvaluation Cnteria.

THE APACHE LEAP RESEARCH SITE.May 1995 1996.

NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-l lNG CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report.

NAIR,8.K.

NUREG/CR 6481 VOI: REVIEW OF MODELS USED FOR DETERMIN.

PARKS,C.V.

ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model NUREG/CR-5661: RECOMMENDATIONS FOR PREPARING THE CRITI-Evaluation Crtteria CALITY SAFETY EVALUATION OF TRANSPORTATION PACKAGES.

NUREG/CH-6481 V02: REVIEW OF MODELS USED FOR DETERMIN-NUREG/CR4505 V01: THE POTENTIAL FOR CRITICALITY FOLLOW-ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report.

ING DISPOSAL OF URANIUM AT LCW-LEVEL WASTE FACIUTIES. Uranium Blended With So61.

NUREG/CR4505 V01: THE POTENTIAL FOR CRITICAUTY FOLLOW-PAUL.D.

ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE-FACluTIES. Uranium Blended With Soil.

SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report.

(

NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY t

NANSTAD,R.K.

RESEARCH GROUP (IPIRG 2) PROGRAM. Final Report.

NUREG/CR4363: EFFECTS OF THERMAL AGING AND NEUTRON IR-RADIATION ON THE MECHANICAL PROPERTIES OF THREE-WIRE PELCHAT,J.M.

STAINLESS STEEL WELD OVERLAY CLADDING.

NUREG 1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS NASTA,K 9'

E NUREG/CR4400: HUMAN FACTORS ENGINEERING (HFE) INSIGHTS l

FOR ADVANCED REACTORS BASED UPON OPERATING EXPERI-PELTON,M.A.

l ENCE.

14UREG/CR4566: DESCRIPTION OF MULTIMEDIA ENVIRONMENTAL POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.? MODI-N R G'1556 V2 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE.

FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

l RIALS UCENSES. Program Specific Gu6 dance About industrial Radog-PENNELL,W.E.

l raphy Licenses. Draft Report For Use And Comment.

NUREG/CR-4219 V12 N2: HEAVY SECTION STEEL TECHNOLOGY l

{

NEUMM,S.P.

PROGRAM. Semiannual Progress Report For April 1995 Through Sep-NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH NU CR 219 V13 N1: HEAVY-SECTION STEEL TECHNOLOGY NUREG/C

A C A

EL EX R ENTS AT PROGRAM. Semiannual Progress Report For October 1995 - March j

THE APACHE LEAP RESEARCH SITE.May 1995 1996.

f NOURSAKHSH,H.P PHAN,H.K.

NUREG/CR4295$ REASSESSMENT OF SELECTED FACTORS AF.

NUREG/CR-6181 ROI: A PILOT APPUCATION OF RISK-INFORMED FECTING SITING OF NUCLEAR POWER PLANTS' METHODS TO ESTABLISH INSERVICE INSPECTION PRIORITIES FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR ql?DONNELL.E.

POWER STATION.

NUREG/CR 4918 V10 CONTROL OF WATER INFILTRATK)N INTO NEAR SURFACE LOW-LEVEL WASTE DISPOSAL UNITS. Final Report PtLCH,M.M.

On Field Expenments At A Humed Region Site.Beltsville, Maryland.

NURFOCRa6469: EXPERIMENTS TO INVESTIGATE DIRECT CON-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE O'HARA,J.

CALVERT CLIFFS NUCLEAR POWER PLANT.

NUREG/CR4393: INTEGRATED SYSTEM VAUDATION: METHODOLO-GY AND REVIEW CRITERIA.

PISKURA,0.

NUREG 1556 V2 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-OLSON,R.

RIALS LICENSES. Program Specific Guidance About industnal Radiog-NUREG/CR4233 V03: CRACK STABluTY IN A REPRESENTATIVE raphy Ucenses. Draft Report For Use And Comment.

PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-NAMIC DISPLACEMENT CCi4 TROLLED STRESSES. Subtask 1.3 Final PLEUNE,T.T.

Report.

NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4233 V04: INTERNATK)NAL PIPING INTEGRITY RE.

UGHT WA1ER REACTORS. Semiannual Report. January 1996 - June SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report.

1996

44 Personal Author index POOLE,A.B.

ROHATG1,U.S.

NUREG/CR4506. COMPONENT UNAVAILABluTY VERSUS INSERV-NUREG/CR4474: PREUMINARY PHENOMENA IDENTIFICATION AND ICE TEST (IST') INTERVALEVALUATONS OF COMPONENT AGING RANKING TABLES (PIRT) FOR SBWR STARTUP STABILITY.

EFFECTS WITH APPUCATONS TO CHECK VALVES.

ROLLSTIN J.A.

PORTER,A M-NUREG/CR4525: SECPOP90: SECTOR POPULATON, LAND FRAC-NUREG/CR4456: REVIEW OF !NDUSTRY EFFORTS TO MANAGE TON, AND ECONOMIC ESTIMATION PROGRAM.

PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING, AND FEEDRING CRACKING AND WALL THINNING.

ROM,0.

I NUREG 1532: FINAL TECHNICAL EVALUATION REPORT FOR THE POWERS,0.A.

PROPOSED REVISED RECLAMATON PLAN FOR THE ATuS COR-NUREG/CR4153: A SIMPUFIED MODEL OF DECONTAMINATION BY PORATION MOAB MILL. Source Material Ucense No. SUA-917.Docatet BWR STEAM SUPPRESSION POOLS.

No. 40-3453.(Anas Corporatxm)

PRENDERGAST,K.

NUREG 1556 V2 DAF FC: CONSOLIDATED GUIDANCE ABOUT MATE-ROOD A.

RIALS UCENSES. Program Specrhc Guidance About industnal Radeog-NUREG/CR-6523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE raphy Ucenses. Draft Report For Use And Comment.

UNCERTAINTY ANALYSIS. Food Chain Uncertainty Assessment.Mam Report.

PULLANI S.V.

NUREG/CR4523 V02: PROBABIUSTIC ACCIDENT CONSEQUENCE NUREG-1275 V12:

OPERATING EXPERIENCE FEEDBACK UNCERTAINTY ANALYSIS.

Food Cham Uncertainty REPORT. Assessment Of Spent Fuel Coolmg-Assessment. Appendices.

RADCLIFFE W.H.

RUDLAND,D.L NUREG 1556 V4 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS RIALS LICENSES. Program Specific Guidance About Fixed Gauge WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT-PIPE h56 V5 FCfCO DATED GUIDANCE ABOUT MATE-LOCATIONS. Final Report. September 1991 November 1995.

NUR NUREG/CR4446: FRACTURE TOUGHNESS EVALUATONS OF TP304 RIALS LICENSES. Program-Specrhc Guidance About Self.Stueided irra-8 88 diator Ucenses. Draft Report For Comment.

WR 5 HE S COND IffTERNATONAL PIPING INTEGRITY RADDATZ,M.G.

RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report.

NUREG-1571: INFORMATION HANDBOOK ON INDEPENDENT SPENT FUEL STORAGE INSTALLATIONS.

RUTHER,W.E.

NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CPACKING IN RADONJIC,Z.R.

UGHT WATER REACTORS. Semianru;al Report, January 1996 - June NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-1996.

ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER RFi+C' ORS. Semiannual Report, July December 1996.

RMWL NUREG/CR4331 A01: ATMOSPHERIC RELATIVE CONCENTRATIONS SAGAR,8.

IN BUILDING WAKES.

NUREG/CR 6513 N01: NRC HIGH-LEVEL RADIOACTIVE WASTE MAN-AGEMENT PROGRAM ANNUAL PROGRESS REPORT: FISCAL YEAR RANDALL.K.

1996.

NUREG 1556 V3 DAF FC: CONSOLIDATED GUIDANCE ABOUT MATE.

RIALS UCENSES. Applications for Sealed Source And Device Evalua.

8A tion And Registration. Draft Report For Comment.

NUREG C 4538: EVALUATION OF LOCA WITH DELAYED LOOP AND RAO,0.V.

LOOP WITH DELAYED LOCA ACCOENT SCENARIOS.

NUREG/CR4370: BLOCKAGE 2.5 USER'S MANUAL NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL SANFORD,W.E.

NUREG/CR-5229 V00: FIELD LYSIMETER INVESTIGATIONS: LOW-RE3/CR4167: LATE-PHASE MELT PROGRESSION EXPERIMENT L YEAR 996 nua MP-2.Results And Analysis.

SA GO,P.A.

REMEC,8 NUREG 1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS NUREd/CR-6454: POOL CRITICAL ASSEMBLY PRESSURE VESSEL LICENSES Program-Specific Guidance About Portable Gauge FACIUTY BENCHMARK.

Ucense. Final Report.

RICH,T.

NUREG-1556 V3 DAF FC: CONSOUDATED GUIDANCE ABOUT MATE.

SCHAEFER,C.W.

RIALS UCENSES. Applications for Sealed Source And Device Evalua.

NUREG/CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON tion And Registration. Draft Report For Comment.

PIG SKIN.

NUREG-1562 DRFT FC: STANDARD REVIEW PLAN FOR APPUCA-TIONS FOR LICENSES TO DISTRIBUTE BYPRODUCT MATERIAL TO SCHIFF A.J.

PERSONS EXEMPT FROM THE REQUIREMENTS FOR AN NRC NUREG/CR4464: AN EVALUATION OF METHODOLOGY FOR SEIS-LICENSE.10CFR Parts 30.14.30.15,30.16,30.18,30.19 & 30.20.

MIC QUALIFICATION OF EQUIPMENTCABLE TRAYS, AND DUCTS IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

NUREG/CR-6525: SECPOP90 SECTOR POPUuTON, LAND FRAC.

SCHLUETER,J.

TON, AND ECONOMIC ESTIMATON PROGRAM.

NUREG 1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY PROGRAMS AT MEDICAL FACluTIESFmal Report.

RIDKY,R.W.

NUREG/CR-4918 V10 CONTROL OF WATER INFILTRATION INTO SCHMIDT,R j

NEAR SURFACE LOW LEVEL WASTE DISPOSAL UNITSFmal Report NUREG/CR4233 V03: CRACK STABILITY IN A REPRESENTATIVE On Field Experiments At A Humid Region Site,Bettsville, Maryland.

PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-ROCKHOLD.M.L.

NAMIC DISPuCEMENTCONTROLLED STRESSES. Subtask 1.3 Final NUREG/CR-6565: UNCERTAINTY ANALYSES OF INFILTRATION AND Report.

SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY R E-SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report.

NUREG/CR-5229 V09 FIELD LYSIMETER INVESTIGATIONS: LOW.

SCHMIDT,R.C.

LEVEL WASTE DATA B,*.SE DEVELOPMENT PROGRAM FOR NUREG/CR4167; LATE-PHASE MELT PROGRESSION EXPERIMENT FISCAL YEAR 1996. Annual Report.

MP-2.Results And Analysis.

Personal Author Index 45 SCHNEIDER,8.

SOPPET,WX NUREG-1492-REGULATORY ANALYSIS ON CRITERIA FOR THE RE.

NUREG/CR 4667 V22-ENVIRONMENTALLY ASSISTED CRACKING IN LE.ASE OF PATIENTS ADMINISTERED RADIOACTIVE UGHT WATER REACTORS. Semannual Report. January 1996 June MATERIAL.Flnal Report.

1996.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN SCHRIMER,HK UGHT WATER REACTORS. Semiannuni Report July-December 1996.

NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1962-61A Status Report SOUTOL NUREG/CR4370: BLOCKAGE 2.5 USER'S MANUAL SCHULZ,R1 NUREG/CR 4918 V10: CONTROL OF WATER INFILTRATION INTO STAMPS,0.W.

NEAR SURFACE LOW LEVEL WASTE DISPOSAL UNITS. Final Report NUREG/CR-6530: DEUBERATE IGNITION OF HYDPOGEN-AIR-STEAM On Field Experiments At A Humid Region Site,Beltsville, Maryland.

MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

SCHWARTZ,ML STEPHENS,D.M.

NUR NUREG-1556 V2 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-HE APACHE LEAP RE CH S E.Ma 1 95 1 RIALS UCENSES Program Specife Guidance About industnal Radiog-raphy Licenses. Draft Report For Use And Comment.

STOLLER,RL NUREG-1556 V5 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-NUREG/CR4399. RESULTS OF CHARPY V-NOTCH IMPACT TESTING l

RIALS LICENSES. Program-Specific Guidance About Self-Srweided Irra-OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 DE-distor Ucenses. Draft Report For Comment.

GREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMERCIAL RE.

A RWE SCOTT,P.

NUREG/CR4233 V03: CRACK STABluTY IN A REPRESENTATIVE STRAIN,R.V.

PIPtNG SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY.

NUREG/CR-4667 V22: ENVIRONMENTALLY ASS!STED CRACKING IN NAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask 1.3 Final LIGHT WATER REACTORS. Semiannual Report, January 1996 - June N8 port 1996.

NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE*

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN SEARCH PROGRAM (iPIRG) PROGRAM. Program Final Report LIGHT WATER REACTORS. Semiannual Report, July December 1996.

NUREG/CR4369: IPIRG 2 TASK 1 PIPE SYSTEM EXPERIMENTS WITH CIRCUMFERE,ITIAL CRACKS IN STRAIGHT-PIPE STRANGE,WL LOCATIONS. Final Report 3eptember 1991 November 1995.

NUREG/CR-6586: HORIZONTAL VELOCITIE3 IN THE CENTRAL AND NUREG/CR-6452: THE SECOND INTERNATIONAL PIPtNG INTEGRITY EASTERN UNITED STATES FROM GPS SOVEY3 DURING THE RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report-19871996 INTERVAL.

SHACK,W.J.

STRENGE,D.L NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-6566: DESCRIPTION OF MULTIMEDIA ENVIRONMENTAL UGHT WATER REACTORS. Sermannual Report. January 1996 - June POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-1996.

FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

NUREG/CR-4667 V23. ENVIRONMENTALLY ASSISTED CRACKING IN i

LIGHT WATER REACTORS. Sermannual Report, July-December 1996.

STROSNIDER,JA l

NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-l i

PROGRAM. Semiannual Report August 1995 - March 1996.

TABASE.

SHAFFER,CJ.

STRJCKMEYER,R.

NUREG/CH-6370 BLOCKAGE 2.5 USER'S MANUAL NUREG.0837 V16 NO3: NRC TLD DIRECT RADIATION MONITORING NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL NETWORK. Progress Report JulpSeptember 1996.

NUREG-0837 V16 N04: NRC Tw DIRECT RADIATION MONITORING SHAH,Y.N.

NETWORK. Progress Report. October-December 1996.

NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE NUREG-0837 V17 NO1: NRC TLD DIRECT RADIATION MONITORING PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING, NETWORK. Progress Report. January-March 1997.

AND FEEDRING CRACKING AND WALL THINNING.

NUREG-0837 V17 NO2: NRC TLD DIRECT RADIATION MONITORING NETWORK. Progress Report April June 1997.

SHONKA,JJ.

NU

/CR 6 5 VE OPMENT OF CONFORMAL RESPIRATOR R G CR4393: INTEGRATED SYSTEM VAUDATION: METHODOLO-GY AND REVIEW CRITERIA.

SIMOMEN,FA

)

gg g NUREG/CR4181 RO1: A PILOT APPLICATION OF RISK-INFORMED NUREG-1603 DAFT-'

INDIVIDUAL PLANT EXAMINATION METHODS TO ESTABUSH INSERVICE INSPECTION PRIORITIES I

DATABASE. User's Guide' FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR i

POWER STATION.

SUKALAC.T.R.

NUREG/CR4535: DEVELOPMENT OF CONFORMAL RESPIRATOR I

G CR4331 R01: ATMOSPHERIC RELATIVE CONCENTRATIONS IN BUILDING WAKES.

SULLIVAN,T.M.

NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW-S4MPSON,JJ LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR

{

NUREG/CR4558: NRC ANTITRUST UCENSING ACTIONS, 1978-1996-FISCAL YEAR 1996. Annual Report

-j NUREG/CR4515: BLT-EC (BREACH, LEACH, AND TRANSPORT-EOUI-SMffH,5, UBRIUM CHEMISTRY) DATA INPUT GUIDE.A L,omputer Model For NUREG-1556 V3 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-Simulating Release And Coupled Geochemical Transport Of Contarm-RIALS UCENSES. Applications for Sealed Source And Device Evalua-nants From A Subsurface Disposal Facihty, teon And Registration. Draft Report For Comment SWAIN,R L SNAY,RA NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-NUclEG/CR4586: HORIZONTAL VELOCITIES IN THE CENTRAL AND FIED A 302 GRADE B PLATE MATERIALS. DATA ANALYSIS.

EASTERN UNITED STATES FROM GPS SURVEYS DURING THE NUREG/CR4426 V02: DUCTILE FRACTURE TOUGHNESS OF MODI-19671996 INTERVAL FIED A 302 GRADE B PLATE MATERIALS. Data Records.

l SOKOLOV,MA SWINTH,K.L.

NUREG/CR4379: AN IMPROVED CORRELATION PROCEDURE FOR NUREC/CR 6581: CONSIDERATIONS IN THE APPLICATION OF THE SUBSIZE AND FULL-SIZE CHARPY IMPACT SPECIMEN DATA.

ELECTRONIC DOSIMETER TO DOSE OF RECORD.

1 i

L______________.

l l

46 Personal Author Index TADIOS,E.L VIETH,P.

NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0 A COMPUTER NUREG/CR4233 V02: STA81UTY OF CRACKED PIPE UNDER SEIS-CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

MIC/ DYNAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask j

12 Final Report NUREG/CR 4391: DETONATION CELL SIZE MEASUREMENTS IN VINSONJ.

HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE NUREG/CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES BNL HIGH TEMPERATURE COMBUSTION FACluTY.

TRENCH SITE: EXPERIMENT 118.

TALWANI.P.

YO,T.V.

AUREG/CR 6529: VAUDATION OF TECTONIC MODELS FOR AN IN' NUREG/CR-6181 RO1: A PILOT APPUCATION OF RISK-INFORMED TRAPLATE SEISMIC ZONE, CHARLESTON. SOUTH CAROLINA WITH METHODS TO ESTABUSH INSERVICE INSPECTION PRORITIES GPS GEODETIC DATA.

FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR TANAKA.T.J.

FOWER STATON.

NUREG/CR4643: EFFECTS OF SMOKE ON FUNCTIONAL CIRCUITS' WALKER,$.

TANG,J.S.

NUREG1610: CONTROLLING THE ATOM.The Bogenrungs Of Nuclear NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICAUTY SLIDE Regulation, 1946 1962.

RULE. Technical Basis.

THOMAS M.L NUREG/CR4506: EMBRITTLEMENT DATA BASE, VERSON 1.

NUREG.0713 V17: OCCUPATIONAL HADIATON EXPOSURE AT COM-MERICAL NUCLEAR POWER REACTORS AND OTHER WANG,Y.K.

FACluTIES,1995. Twenty-Eighth Annual Report NUREG/CR4414: PIPING BENCHMARK PROBLEMS FOR THE WES-TINGHOUSE AP600 STANDARDIZED PLANT.

THOMASSON.M.J.

NUREG/LR4407: DATA COLLECTION AND FIELD EXPERIMENTS AT WARE,A.G.

THE APACHE LEAP RESEARCH SITE.May 1995 1996.

NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING,

^

C'R 59: FIELD STUDIES AT THE APACHE LEAP RESEARCH SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

WASHINGTON,K.E.

NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A CCMPUTER THE APACHE LEAP RESEARCH SITE.May 1995 1996.

CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

TILLS J'G/CR 6533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER NURE WAM8'"A CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

NUREG-1571: INFORMATON HANDBOOK ON INDEPENDENT SPENT FUEL STORAGE INSTALLATIONS.

TINGLE,W.

NUREG-1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS WATKINS,JA LICENSES. Program-Specrfic Guidance About Portable Gauge NUREG/CR-6478: MOTOR OPERATED VALVE (MOV) ACTUATOR Ucenses. Final Report MOTOR AND GEARBOX TESTING.

TOMPKINS,M.M.

WATSON,G.M.

NUREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NUCLE.

NUREG-1556 V4 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-AR ELECTRICITY-GENERATING UNITS IN THE UNITED STATES:

RIALS UCENSES. Program Speerfle Guidance About Fixed Gauge 1997 D01.

Ucenses. Draft Report For Comment TDRAh V.

WEISMANN.J.J.

NURE CR4505 V01: THE POTENTIAL FOR CRITICAUTY FOLLOW.

NUREG/CR4535: DEVELOPMENT OF CONFORMAL RESPIRATOR ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE MONITORING TECHNOLOGY.

FACluTIES. Uranium Blended with So6l.

WESLEY,0.A.

TRAVIS,R.J.

NUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPI-NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCl-GENERO BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR DENT CONDITIONS.

POWER PLANTS.

WHITE D N

6 V2 N m CONSMOATED WOANCE ABOW N NUR G 4528: ENVIRONMENTAL ASSESSMENT PROPOSED U.

am S nce M lhi RM CENSE RENEWAL OF NUCLEAR METALS,tNC. CONCORD, MASSA.

CHUSETTS.

raphy U enses. Draft Report For Use And Comment TRENKAMP,R.

WHITEHEAD,D W.

NIJREG/CR4529: VAUDATION OF TECTONIC MODELS FOR AN IN-NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE I

TRAPLATE SEISMO ZONE, CHARLESTON. SOUTH CAROUNA WITH DAMAGE ACCIDENTS: 198243.A Status Report GPS GEODETIC DATA.

,gy VACCA,PA NUREG 1556 V2 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-NUREG-1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS RIALS UCENSES. Program Specific Guidance About Induatnal Radiog-UCENSES. Program Specific Guidance About Portable Gauge raphy Ucenses. Draft Report For Uw And Comment.

Liceness. Final NUREG-1556 V5. F FC: CONSOUDATED GUIDANCE ABOUT MATE-WHITTEN.J.E.

RMLS LICENSES.Progrern-Specific Guidance About Self-Srweided Irra.

NUREG-1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS dletor Licenses. Draft Report For Comment UCENSES. Program-Speerhc Guidance About Portable Gauge Ucenses. Final Report VANKUIKEN.JA NOREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NUCLE.

WHORLOW.K.M.

AR ELECTRICITY-GENERATING UNITS IN THE UNITED STATES:

NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN 1997-2001.

ONS ON THE EXTERNAL STRESS CORROSION CRACKING OF TYPE 304 AUSTENITIC STAINLESS STEEL NUREG/CR4508: COMPONEriT UNAVAILABluTY VERSUS INSERV.

WICHMAN,K.R.

ICE TEST (IST) INTERVAL: EVALUATIONS OF COMPONENT AGING NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-EFFECTS WITH APPUCATOHS TO CHECK VALVES.

TABASE.

Personal Author index 47 WlERENGA,P.J.

NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT NUREG/CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES THE APACHE LEAP RESEARCH SITE.May 1995 1996.

TRENCH SITE: EXPERIMENT IIB.

i NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH WOODS,S.

SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

l NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT NUREG-1516. MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY PROGRAMS AT MEDICAL FACILITIES. Final Report.

THE APACHE LEAP RESEARCH SITE.May 1995 1996.

WtLKOWSKI.G.M NUREG/CR4233 V02: STABILITY OF CRACKED PlPE UNDER SEIS-NUREG/CR4474: PRELIMINAR( PHENOMENA OENTIFICATION AND MIC/ DYNAMIC DISPLACEMENT CONTROLLED STRESSES Subtask RANKING TABLES (PIRT) FOR S8WR STARTUP STABILITY.

NI EG CR4 V03: CRACK STABILITY IN A REPRESENTATIVE XIE,J.

AM dis L CE N CO TR DS SE s

Fina E

A NG CENTRAL' A D RN N TH MER-Report.

ICA.

NUREG/CR 6233 V04: INTERNATIONAL PIPING INTEGRITY RE-SEARCH PROGRAM (IPIRG) PROGRAM Program F' al Report.

XIE J.W.

m NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS NUREG/CR-4409 V06: DATA BASE ON DOSE REDUCTON PROJECTS WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT. PIPE FOR NUCLEAR POWER PLANTS.

LOCATIONS Fmal ReportSeptember 1991 - November 1995.

NUREG/CR4446: FRAuuRE TOUGHNESS EVALUATONS OF TP304 YANG,J.W.

STAINLESS STEEL PIPES-NUREG/CR4538: EVALUATION OF LOCA WITH DELAYED LOOP AND NUREG/CR4452: THE SECOND INTERNATIONAL PlPING INTEGRITY RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report LOOP WITH DELAYED LOCA ACCIDENT SCENAROS.

WILLIAMS,D.C.

YOUNG,M.H.

NUREG/CR 6533: CODE MANUAL FOR CONTAIN 2 0: A COMPUTER NUREG/CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

TRENCH SITE: EXPERIMENT !IB.

WILSON,G.E.

YOUNG,M.L NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING NUREG/CR-6523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE TABLES FOR WESTINGHOUSE APOO SMALL BREAK LOSS-OF-UNCERTAINTY ANALYSIS. Food Chain Uncertamty Assesvient Mam COOLANT ACCIDENT, MAIN STE/M LINE BREAK, AND STEAM Report.

GENERATOR TUBE RUPTURE SCENARIOS.

NUREG/CR-6523 V02: PROBABILISTIC ACCIDENT Cr% SEQUENCE UNCERTAINTY

ANALYSIS, Food Chain Uncertainty WINBOW,R.T.

Assesstnent. Appendices.

NUREG/CR-6528: ENVIRONMENTAL ASSESSMENT PROPOSED U-NUREG/CR4526 V01: PROBABluSTIC ACCOEN" CONSEQUENCE CENSE RENEWAL OF NUCLEAR METALS,lNC. CONCORD, MASSA-CHUSETTS.

UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited Matenal And External Doses. Main Report.

WOLTERMAN.R NUREG/CR-6526 V02-PROBA8ILISTIC ACCIDENT CONSEQUENCE NUREG/CR4389. IPIRG-2 TASK 1 PIPE SYSTEM EXPERIMENTS UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT-PIPE SITED MATERIAL AND EMERNAL DOSES. Appendices.

LOCATIONS Final Report. September 1991. November 1995.

WOOD.R,S.

ZHANG,J.

NUREG-1577 DRFT FC: STANDARD REVIEW PLAN ON POWER REAC-NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN TOR LICENSEE FINANCIAL QUALIFICATIONS AND DECOMMIS-UGHT WATER REACTORS. Semiannual Report, July-December 1996.

SiONING FUNDING ASSURANCE. Draft Report For Comment WOODHOUSE,E.G.

NUREG/CR.6528: ENVIRONMENTAL ASSESSMENT PROPOSED LI-NUREG/CR-6459. FIELD STUDIES AT THE APACHE LEAP RESEARCH CENSE RENEWAL OF NUCLEAR METALS,1NC. CONCORD, MASSA-SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

CHUSETTS.

1 l

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

tre welcome.

10 CFR 50 NUREG/CR4497: DATA COLLECTON AND FIELD EXPERIMENTS AT NUREG-1606 DRFT FC: PROPOSED REGULATORY GUIDANCE RE-THE APACHE LEAP RESEARCH SITE.May 1995 1996.

LATED TO IMPLEMENTATION OF 10 CFR 50.59 (CHANGES, TESTS, l

OR EXPERIMENTS). Draft Report For Comment Air-Detonation NUREG/CR4391: DETONATION CELL SIZE MEASUREMENTS IN I

A 302 Grade 8 Steel Plate HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE l

NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-BNL HIGH-TEMPERATURE COMBUSTON FACILITY.

l FIED A 302 GRADE B PLATE MATERIALS, DATA ANALYSIS.

NUREG/CR4426 V02: DUCTILE FRACTURE TOUGHNESS OF MODI-Annual Report I

l FIED A 302 GRADE B PLATE MATERIALS. Data Records.

NUREG 1145 V13: U.S. NUCLEAR REGULATORY COMMISSION 1996 l

ANNUAL REPORT.

g NUREG-1125 V18: A COMPILATofd OF REPORTS OF THE ADVISORY Antitrust COMMITTEE ON REACTOR SAFEGUARDS.1996 Annual.

NUREG-1574: STANDARD REVIEW PLAN ON ANTITRUST ALARA REVIEWSFinal Report NUR 574 FC:

ARD REVIEW PLAN ON NUREG/CR-4409 V06 DAT A BASE CN DOSE REDUCTION PROJECTS FOR NUCLEAR POWER PLANTS.

NUREG/CR4558: NRC ANTITRUST LICENSING ACTONS, 1978-1996.

ALWR NUREG/CR-6464: AN EVALUATION OF METHODOLOGY FOR SEIS-M L**P MIC QUALIFICATION OF EQUIPMENTCABLE TRAYS

• AND DUCTS NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH IN ALWR PLANTS BY USE OF EXPERIENCE DATA' SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

NUREG/CR4497: DATA COLLECTON AND FIELD EXPERIMENTS AT Abnormal Occurrence THE APACHE LEAP RESEARCH SITE.May 1995 - 1996.

t NUREG-0090 V19: REPORT TO CONGRESS ON ABNORMAL l

OCCURRENCES. Fiscal Year 1996.

A Accident Scenario PROPOSED REVISED RECLAMATON PLAN FOR THE ATLAS COR-NUREG/CR4538: EVALUATON OF LOCA WITH DELAYED LOOP AND PORATON MOAB MILLSource Material License No. SUA-017. Docket LOOP WITH DELAYED LOCA ACCIDENT SCENARIOS.

No. 40 3453.(Atlas Corporation)

Accident Sequence Precursor A:-

M Diepersion NUREG/CP 4874 V23: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4331 R01: ATMOSPHERIC RELATIVE CONCENTRATIONS DAMAGE ACCIDENTS: 1995. A Status Report IN BUILDING WAKES.

l NUREG/CR4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4481 V01: REVIEW OF MODELS USED FOR DETERMIN-DAMAGE ACCIDENTS: 1982-83.A Status Report ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model Evaluation Criteria.

i

(

AccountabilHy Report NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-NUREG 1542 V02: ACCOUNTABILITY REPORT FISCAL YEAR 1996.

ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluabon Report i

Actuator Geert>ss gg,,

l NUREG/CR4478: MOTOR-OPERATED VALVE (MOV) ACTUATOR NUREG 1610: CONTROLLING THE ATOM.The Begenin9s Of Nuclear MOTOR AND GEARBOX TESTING.

R m 19M962.

Advanced Bolling Water Reactor A"*'*" HIC 8'*I'O*** 8'*d i

NUREG-1503 S01: FINAL SAFETY EVALUATON REPORT RELATED NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN l

TO THE CERTIFfCATON OF THE ADVANCED BO(LING WATER RE.

ONS ON THE EXTERNAL STRESS CORROSION CRACKING OF i

ACTOR DESIGN Supplement No.1. Docket No.52-001,(General Elec-1 TYPE 304 AUSTENITIC STAINLESS STEEL tric Nuclear Ene gy)

Advanced Nucteer Pmver Plant SLOCKAGE 2 NUREG/CR 6486: ASSESSMENT OF MODULAR CONSTRUCTION FOR NUREG/CR4370: BLOCKAGE 2.5 USER'S MANUAL NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL SAFETY RELATED STRUCTURES AT ADVANCED NUCLEAR POWER PLANTS.

BLT4C NUREG/CR4515: BLT-EC (BREACH, LEACH, AND TRANSPORT EQUI-Advisory Committee On Nucieer Waste NUREG-1423 V07: A COMPILATION OF REPORTS OF THE ADVISORY UBRIUM CHEMISTRY) DATA INPUT GUIDE.A Computer Model For COMMITTEE ON NUCLEAR WASTE. July 1996 June 1997.

Mung Release And Coupled Geochermcal Transport Of Contamh nanta From A Subourface Disposal Facility.

A9tng NUREG 1611: AGING MANAGEMENT OF NUCLEAR POWER PLANT SWR CONTAINMENTS FOR LICENSE RENEWAL NUREG 1616: FEASIBILITY OF UNDERWATER WELDING OF HIGHLY NUREG/CR4508: COMPONENT UNAVAILABILITY VERSUS INSERV.

IRRADIATED IN-VESSEL COMPONENTS OF BOILING WATER ICE TEST (IST) INTERVAL:EVALUATONS OF COMPONENT AGING REACTORS.A Literatun, Review.

EFFECTS WITH APPUCATIONS TO CHECK VALVES.

NUREG/CR4153: A SIMPLIFIED MODEL OF DECONTAMINATION BY BWR STEAM SUPPRESSION POOLS.

Air Permeability NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

POWER PLANTS.

49 l

w___-.

50 Subject index NUREG/CR4527; FINAL RESULTS OF THE XR21 BWR METALUC NUREG/CR4400 HUMAN FACTORS ENGINEERING (HFE) INSIGHTS MELT RELOCATION EXPERIMENT.

FOR ADVANCED REACTORS BASED UPON OPERATING EXPERl-ENCE.

NUREG/CR 6361: CRITICAUTY BENCHMARK GUIDE FOR UGHT-Cherpy impact WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE NUREG/CR4379: AN IMPROVED CORRELATION PROCEDURE FOR PACKAGES.

SUBSIZE AND FULL SlZE CHARPY IMPACT SPECIMEN DATA.

soning water neector Charpy V Notch NUREG-1616: FEASIBluTY OF UNDERWATER WELDING OF HIGHLY NUREG/CR4399: RESULTS OF CHARPY V-NOTCH IMPACT TESTING tRRADIATED IN-VESSEL COMPONENTS OF BOluNG WATER OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 DE-

} ^

NUREG C 53 Si MODEL OF DECONTAMINATION BY T CV

• BWR STEAM SUPPRESSON POOLS.

NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF Check Velve GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR NUREG/CR4508: COMPONENT UNAVAILABluTY VERSUS INSERV.

ICE TEST (IST) INTERVAL-EVALUATIONS OF COMPONENT AGING NU

/CR-

FINAL RESULTS OF THE XR21 BWR METALLC EFFECTS WITH APPLICATIONS TO CHECK VALVES.

MELT RELOCATON EXPERIMENT.

O Doron Dilution NUREG/CR4566: DESCRIPTON OF MULTIMEDIA ENVIRONMENTAL NUREG/CP 0158: PROCEEDINGS OF THE OECD/CSNI SPECIALISTS MEETING ON BORON DILUTON REACTIVITY TRANSIENTS. Held in POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

State College, Pennsylvania. USA. October 18-20,1995.

Chemical Process Safety Drechtherapy NUREG/CR4074 V03: SEALED SOURCE AND DEVICE DESIGN NUREG 1601: CHEMICAL PROCESS SAFETY AT FUEL CYCLE FAClu-TIES.

SAFETY TESTING. Technical Report On The Findings Of Task 4 inves.

Ingation Of A Failed Brachytherapy Needle Applicator.

ChhWM Budget Estimate NUREG-1604: CIRCUMFERENTIAL CRACKING OF STEAM GENERA-NUPEG 1100 V13: BUDGET ESTIMATES. Fiscal Year 1996.

TOR TUBES.

NUREG/CR4389. IPIRG-2 TASK 1 PIPE SYSTEM EXPERIMENTS Building Wake WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT-PIPE NUREG/CR4331 R01: ATMOSPHERIC 'dELA ilVE CONCENTRATIONS LOCATONS. Final Report. September 1991 - November 1995.

IN BUILDING WAKES.

Cladding Corros6on Byproduct Material NUREG/CR4534 V01: FRAPCON-3: MODIFICATIONS TO FUEL ROD NUREG 1562 DAFT FC: STANDARD REVIEW PLAN FOR APPUCA-MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH-TlONS FOR UCENSES TO DISTRIBUTE BYPRODUCT MATERIAL TO BURNUP APPLICATION.

PERSONS EXEMPT FROM THE REQUIREMENTS FOR AN NRC LICENSE.10CFR Parts 30.14,30.15,30.16,30.18,30.19 & 30.20.

Clodding Effect NUREG/CR-4219 V12 N2: HEAVY-SECTION STEEL TECHNOLOGY UE O

OCEEDINGS OF THE CNRA/CSNI WORKSHOP g $

ON STEAM GENERATOR TUBE INTEGRITY IN NUCLEAR POWER PLANTS.

Code Architecture NUREG/CP-0159: PROCEEDINGS OF THE OECD/CSNI WORKSHOP ON TRANSIENT THERMAL HYDRAUUC AND NEUTRON!C CODES RE /CR4533. CODE MANUAL FOR CONTAIN 2.0: A COMPUTER REQUIREMENTS. Held in Annapohs, Maryland, USA, November 54, CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

1996.

Cable Trey Code Manual NUREG/CR4464: AN EVALUATON OF METHODOLOGY FOR SEIS-NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER MIC QUALIFICATION OF EQUIPMENT, CABLE TRAYS, AND DUCTS CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

Communication Calvert C#ffe NUREG/CR4469: EXPERIMENTS TO INVESTIGATE DIRECT CON.

NUREG-1545: EVALUATION CRITERIA FOR COMMUNICATIONS-RE-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE LATED CORRECTIVE ACTON PLANS.

CALVERT CUFFS NUCLEAR POWER PLANT.

Consolidated Guldence Cell Size NUREG-1556 V01: CONSOJOATED GUIDANCE ABOUT MATERIALS NUREG/CR4391: DETONATON CELL St2E MEASUREMENTS IN LICENSES. Program Spectic Guidance About Portable Gauge HOH TEMPERATURE HYDROGEN-AIR STEAM MIXTURES AT THE Ucenses. Final Report.

BNL HIGH-TEMPERATURE COMBUSTION FACILITY.

NUREG-1556 V3 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-RIALS LICENSES. Applications for Sealed Source And Device Evalua.

Certifloatee Of CL,. -

tion And Repstratm Draft Report For Comment.

NUREG-0383 V01 R20- DIRECTORY OF CERTIFICATES OF COMPU.

NUREG-1556 v4 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC.

RIALS LICENSES. Program Specific Guidance About Fixed Gauge Approved Packages.

Ucenses. Draft Report For Comment.

NUREG-0383 V02 R20: DIRECTORY OF CERTIFICATES OF COMPLl-NUREG 1556 V5 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-ANCE FOR RADIOACTIVE MATERIALS PACKAGESCertificates Of RIALS UCENSES. Program-Specific Guidance About Self-Shielded Irra-l l

e.

distor Ucenses. Draft Report For Comrnent.

)

NURE 0383 V03 R17: DIRECTORY OF CERTIFICATES OF COMPU.

I 1

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC-Construction Permit Approved Quality Assurance Programs For Radioactive Matenals Pack.

NUREG 1555 DRFT: ENVIRONMENTAL STANDARD REVIEW ages.

PLAN. Standard Review Plans For Environmental Reviews For Nuclear NUREG-1571: INFORMATON HANDBOOK % INDEPENDENT SPENT Power Plants.

FUEL STORAGE INSTALLATIONS.

NUREG/CR-6558: NRC ANTITRUST L! CENSING ACTIONS, 1978-1996.

Certification Containment NUREG-1462 S01: FINAL SAFETY EVALUATION REPORT RELATED NUREG/CP4157 YO1: PROCEEDINGS OF THE TWENTY-FOURTH TO THE CERTIFICATION OF THE SYSTEM 80+ DESIGN Docket No.

WATER REACTOR SAFETY INFORMATION MEETING Plenary Sea-52402.(Asee Brown Bovert-Combustion Engineering) soon, High Burnup Fuel, Containment And Structural Aging

Subject index 51 NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0 A COMPUTER NUREG/CR 0200 R5V2P2: SCALE: A MODULAR CODE SYSTEM FOR CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-Containment Performance CENSING EVALUATON Functonal Modules F9 - F11.

NUREG/CR4200 R5V2P3: SCALE: A MODULAR CODE SYSTEM FOR feUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPI-CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCI-PERFORMING STANDARDl2ED COMPUTER ANALYSES FOR U-CENSING EVALUATION.Functonal Modules F16 F17.

DENT CONDITIONS.

NUREG/CR4200 R5V3: SCALE: A MODULAR CODE SYSTEM FOR Containment Structure PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-NUREG 1611: AGING MANAGEMENT OF NUCLEAR POWER PLANT CENSING EVALUATION. Miscellaneous.

CONTAINMENTS FOR UCENSE RENEWAL NUREG/CR-5661: RECOMMENDATIONS FOR PREPARING THE CRITI-CALITY SAFETY EVALUATION OF TRANSPORTATION PACKAGES.

NUREG/CR4361: CRITICAUTY BENCHMARK GUIDE FOR LIGHT.

G 4515: BLT EC (BREACH, LEACH, AND TRANSPORT-EQUI-WA UBRIUM CHEMISTRY) DATA INPUT GU!DE.A Computer Model For PAC GE '

Sunulatsng Release And Coupled Geochermcal Transport Of Conta6 Crownpoint nants From A Subsurface Deposal Facility.

NUREG 1508: FINAL ENVIRONMENTAL IMPACT STATEMENT TO Contaminakd &

CONSTRUCT AND OPERATE THE CROWNPOINT URANIUM SOLU-NUREG 1608 DRFT FC: CATEGORIZING AND TRANSPORTING LOW TION MINING PROJECT, CROWNPOINT, NEW MEXICO. Docket No.

SPECIFIC ACTIVITY MATERIALS AND SURFACE CONTAMINATED 40-8968.(Hydro Resources, Inc.)

OBJECTS. Draft Rept For Comment.

Crustal Strain Contamination Survey NUREG/CR4529: VAUDATION OF TECTONIC MODELS FOR AN IN-NUREG/CR4037: MEASUREMENT OF RESIDUAL RADIOACTIVE SUR.

TRAPLATE SEISMIC ZONE, CHARLESTON. SOUTH CAROLINA WITH FACE CONTAMINATION BY 2-D LASER HF ATED TLD.

GPS GEODETIC DATA.

Control Room DOSFAC2 feUREG/CR4393: INTEGRATED SYSTEM VALIDATON: METHODOLO.

NUREG/CR4547: DOSFAC2 USER'S GUIDE.

GY AND REVIEW CRITERIA.

Control Room Habitability NUREG/CR-6497: DATA COLLECTION AND FIELD EXPERIMENTS AT NUREG/CR4331 RO1: ATMOSPHERIC RELATIVE CONCENTRATIONS THE APACHE LEAP RESEARCH SITE.May 1995 - 1996.

IN BUILDING WAKES.

Database Coro Damage NUREG 1603 DAFT:

INDIVIDUAL PLANT EXAMINATION NUREG/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERei CORE DATABASE. User's GuKle.

DA$AAGE ACCIDENTS: 1995. A Status Report.

NUREG/CR-4674 V24 PRECURSORS TO POTENTIAL SEVERE CORE Debris Generation DAMAGE ACCIDENTS: 198243.A Status Report.

NUREG/CR4370: BLOCKAGE 2 5 USER'S MANUAL CorO Degradation NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL NUREG/CR4527: FINAL RESULTS OF THE XR21 BWR METALLIC Decommission MELY RELOCATION EXPERIMENT.

NUREG/CR4037: MEASUREMENT OF RESIDUAL RADIOACTIVE SUR-FACE CONTAMINATION BY 2-0 LASER HEATED TLD.

MUREG-'t545: EVALUATION CRITERIA FOR COMMUNICATIONS-RE-Decommissioning LATED CORRECTIVE ACTION PLANS.

NUREG 1307 R07: REPORT ON WASTE BURIAL CHARGES. Escalation Conosion Of Decommissorwng Wa;.e Deposal Costs At Low-Level Waste Bunal Facilities.

NUREG/CR4543: EFFECTS OF SMOKE ON FUNCTONAL CIRCUlTS.

NUREG 1495 V01: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-

/

667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN

^

UGHT W ATER REACTORS. Semsannual Report, January 1996 June NURE VO F E

C VIRONMENTAL IMPACT STATE-R G/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-UGHT w ATER REACTORS. Semiannual Report. July-December 1996.

RIA FOR UCENSE TERMINATON OF NRC-LICENSED NUCLEAR FACILITIES. Appendices A And B. Final Report.

Cost Estimate NUREG 1496 V03: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-RUREG-1337 R07: REPORT ON WASTE BURIAL CHARGES Escalation MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-Of DecoTunissiorung Waste Dsposal Costs At Low-Level Waste Bunal RIA FOR UCENSE TERMINATION OF NRC-LICENSED NUCLEAR Facilite' F ACILITIES. Appendices C.H. Final Report.

NUREG 1577 DRFT FC: STANDARD REVIEW PLAN ON POWER REAC-Crack Stabilty TOR LICENSEE FINANCIAL QUALIFICATIONS AND DECOMMIS.

NUREG/CR4233 V03: CRACK STABlUTY IN A REPRESENTATIVE SiONING FUNDING ASSURANCE. Draft Report For Comment.

PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-NUREG/CP-0153: PROCEEDINGS OF THE 24TH DOE /NRC NUCLEAR NAMIC DISPLACEMENT-CONTROLLED STRESSES. Subtask 1.3 Final AIR CLEANING AND TREATMENT CONFERENCE.Helo in Portland, Report Ore 9on, July 15-18,1996.

NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF Cracked Pipe GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR NUREG/CR4233 V02: STABluTY OF CRACKED PIPE UNDER SEIS-POWER PLANTS.

MIC/ DYNAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask NUREG/CR4514: ANALYSIS OF POTENTIAL SELF-GUARANTEE 1.2 Final Report, TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY BUSI-Criticanty Safety NESS FIRMS THAT DO NOT ISSUE BONDS.

NUREG/CR4200 R5V1P1: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-Decontaminatim CENSING EVALUATON. Control Modules C4, C6.

NUREG/CP 0153: PROCEEDINGS OF THE 24TH DOE /NRC NUCLEAR NUREG/CR-0200 R5V1P2: SCALE: A MODULAR CODE SYSTEM FOR AIR CLEANING AND TREATMENT CONFERENCE. Held in Portland, PERFORMING STANDAHDlZED COMPUTER ANALYSES FOR U-Oregon, July 15 18,1996.

CENS:NG EVALUATION Control Modules S1 H1.

NUREG/CR.6037: MEASUREMENT OF RESIDUAL RADIOACTIVE SUR-NUREG/CR4200 R5V2P1: SCALE: A MODULAR CODE SYSTEM FOR FACE CONTAMINATION BY 2-D LASER HEATED TLD.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-NUREG/CR4153: A SIMPUFIED MODEL OF DECONTAMINATION BY CENSING EVALUATION. Functional Modules F1 - F8.

BWR STEAM SUPPRESSION POOLS.

)

52 Subject index W Matertal EPICOM-Il NUREG/CR4526 V01: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATORS: LOW-UNCERTAINTY ANALYSIS. Uncertamty Assessment For Deposited LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR Meterial And Extemel Dooes. Main Report FISCAL YEAR 1996. Annual Report.

NUREG/CR4526 V02: PROBABluSTIC ACCOENT CONSEQUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-Earty Site Permit POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

NUREG-1555 DRFT: ENVIRONMENTAL 2:U NDARD REVIEW PLAN. Standard Review Plans For Erwironmental Reviews For Nuclear Design Cr#erts Power Plants.

NUREG/CR4433. CONTAINMENT PERFORMANCE OF PROTOTYPi-CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCl-Earthquake DENT CONDITIONS.

NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABILISTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND Detecton System USE OF EXPERTS,Maan Report.

NUREG/CR4535: DEVELOPMENT OF CONFORMAL RESPIRATOR NUREG/CR4372 V02: RECOMMENDATIONS FOR PROBABluSTIC MONITORING TECHNOLOGY.

SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND USE OF EXPERTS. Appendices.

Dec#on ThrW NUREG/CR4448 V02: EVALUATION OF NATIONAL SE!SMOGRAPH b

NETWORK DETECTON CAPABluTIES. Final Report R

Econornic NUREG/CR4525: SECPOP90. SECTOR POPULATON, LAND FRAC-Device Design NUREG/CR4074 V03: SEALED SOUPCE AND DEVICE DESIGN TION, AND ECONOMIC ESTIMATION PROGRAM.

SAFETY TESTING.Techrwcal Report On The Findings Of Task 4.Inves-E er ti9stion Of A Failed Brachytherapy Needle Applicator.

AR ELECTRICITY GENERATING UNITS IN THE UNITED STATES:

Diletetton Rate NUREG/CR4586: HORIZONTAL VELOCITIES IN TM CENTRAL AND 1997 2001, EASTERN UNITED STATES FROM GPS SURVEYS DURING THE Electronic Doe 4 meter 19871996 INTERVAL NUREG/CR4581: CONSIDERATIONS IN THE APPUCATON OF THE Direct Containment Heating ELECTRONIC DOSIMETER TO DOSE OF RECORD.

NUREG/CR4469: EXPERIMENTS TO INVESTIGATE DIRECT CON-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE Embrittlement CALVERT CUFFS NUCLEAR POWER PLANT.

NUREG/CP4157 V02: PROCEEDINGS OF THE TWENTY-FOURTH WATER REACTOR SAFETY INFORMATON MEETING. Reactor Pres-Diepiscoment4entrolled Strees sure Vessef Ernbrittlement And Thermal Annealing, Reactor Vessel NUREG/CR4233 V03: CRACK STABILITY IN A REPRESENTATIVE Lower Head Integrity And Evaluation And Projection of Steam Genera.

PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-tor tube..

NAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask 1.3 Final NUREG/CR-6506: EMBRITTLEMENT DATA BASE, VERSION 1.

NUR CR4233 V04; INTERNATIONAL PlPING INTEGRITY RE.

Embryo SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report.

NUREG/CR4397: RADIATION SAFETY CONCERNS FOR PREGNANT OR BREAST FEEDING PATIENTS.The Positions Of The NCRP And Does Aseeeement The ICRP.

NUREG/CR4586: DESCRIPTON OF MULTIMEDIA ENVIRONMENTAL POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-Emergency Planning FICATION FOR THE NUCLEAR REGULATORY COMMISSION.

NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICALITY SUDE RULE. Technical Basis.

Does Conversion NUREG/CR4547: DOSFAC2 USER'S GUIDE.

Enforcement Action NUREG-0940 V15 N2 P1: ENFORCEMENT ACTIONS: SIGNIFICANT AC-

^

R CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON Iy mb ir 1s PIG SKIN.

NUREG-0940 V15 N2 P2: ENFORCEMENT ACTIONS: SIGNIFICANT AC-TONS RESOLVED REACTOR UCENSEES.Serniannual Progress Dome Reduc #on NUREG/CR-4409 V06: DATA BASE ON DOSE REDUCTON PROJECTS N

094 NFORCEMENT ACTIONS: SIGNIFICANT AC-FOR NUCLEAR POWER PLANTS.

TIONS RESOLVED MATERIAL UCENSEES. Semiannual Progress Report. July-December 1996.

Doe 4 meter Performance NOREG-0940 V16 N1 P1: ENFORCEMENT ACTIONS: SIGNIFICANT AC-NUREG/CR4581: CONSIDERATIONS IN THE APPUCATION OF THE TONS RESOLVED INDIVIDUAL ACTONS. Semiannual Progress ELECTRONIC DOSIMETER TO DOSE OF RECORD.

Report. January-June 1997.

I NUREG-0940 V16 N1 P2 ENFORCEMENT ACTIONS: SIGNIFICANT AC.

Doelmetry NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA.

TlONS RESOLVED REACTOR UCENSEES. Semiannual Progress TON OF RADIOLABELED ANTIBODIES CONTAINING Y-90,TC 99M)-

Report. January June 1997.

NUREG4940 V16 N1 P3. ENFORCEMENT ACTIONS: SIGNIFICANT AC-131, AND LU-177.

NUREG/CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON TONS RESOLVED MATERIAL UCENSEES. Semiannual Progress

(

l PIG SKIN.

Report, January June 1997.

Duct Engineered Safety System

{

f NUREG/CR4464: AN EVALUATION OF METHODOLOGY FOR SEIS-NUREG/CR4538: EVALUATON OF LOCA WITH DELAYED LOOP AND MIC QUAUFICATION OF EQUIPMENT CABLE TRAYS, AND DUCTS LOOP WITH DELAYED LOCA ACCIDENT SCENARLOS.

IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

Environmental Asessement Ductfie Fracture NUREG/CR4528: ENVIRONMENTAL ASSESSMENT PROPOSED Li-NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-CENSE RENEWAL OF NUCLEAR METALS.INC. CONCORD, MASSA-FIED A 302 GRADE B PLATE MATERIALS, DATA ANALYSIS.

CHUSETTS.

Dynamic Load Environmental impact Statement NUREG/CR4414: PIPING BENCHMARK PROBLEMS FOR THE WES-NUREG-1496 V01: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-TINGHOUSE AP600 STANDARDIZED PLA' ?

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-

Subject Index 53 RIA FOR UCENSE TERMINATON OF NRC-LICENSED NUCLEAR Final Safety Evaeus*.lcsn Report FACluTIES Mem Report. Foal Report.

NUREG 1462 S01: FINAL SAFETY EVALUATION REPORT RELATED NUREG.1496 V02: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-TO THE CERTIFICATION OF THE SYSTEM 80+ DESIGN. Docket No.

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-52 002.(Asea Brown Boven.Combuston E

)

HIA FOi1 UCENSE TERMINATION OF NRC-UCENSED NUCLEAR NUREG-1503 S01. FINAL SAFETY EVALU TION EPORT RELATED FACluTIES. Appendices A And B. Final Report.

TO THE CERTIFICATION OF THE ADVANCED BOILING WATER RE.

NUREG-1496 V03. FINAL GENERIC ENVIRONMENTAL IMPACT STATE-ACTOR DESIGN Supplement No.1. Docket No. 52 001.(General Elec-MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-tric Nuclear Energy)

RIA FOR LICENSE TERMINATON OF NRC-LICENSED NUCLEAR FACluTIES. Appendices C-H Final Report.

Financial Amaurence NUREG/CR4514 ANALYSIS OF POTENTIAL SEtF-GUARANTEE Environmental Protection TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-I NUREG-1555 DRFT: ENVIRONMENTAL STANDARD REVIEW PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY BUSI-PLAN. Standard Rewsw Plane For Environment! Rewsws For Nuclear NESS FIRMS THAT DO NOT ISSUE BONDS.

j Power Plants.

Environmental Software NUREG-1577 DRFT FC: STANDARD REVIEW PLAN ON POWER REAC-NUREG/CR4566: DESCRIPTION OF MULTIMEDIA ENVIRONMENTAL TOR UCENSEE FINANCIAL QUAUFICATIONS AND DECOMMIS-i POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-SiONING FUNDING ASSURANCE. Draft Report For Comment.

l FICATION FOR THE NUCLEAR REGULATORY COMMISSON.

Financial Statement Examination Standard NUREG-1542 V02: ACCOUNTABluTY REPORT rtSCAL YEAR 1996.

NUREG-1021 INT ROG OPERATOR LICENSING EXAMINATON STAND-ARDS FOR FOWER REACTORS.

E/ 5661: RECOMMENDATIONS FOR PREPARING THE CRITl-Emompt Distribution License CALITY SAFETY EVALUATION OF TRANSPORTATION PACKAGES.

NUREG-1562 DRFT FC: STANDARD REVIEW PLAN FOR APPUCA-TIONS FOR UCENSES TO DISTRIBUTE BYPRODUCT MATERIAL TO NR 56 V4 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-PERSONS EXEMPT FROM THE REQUIREMENTS FOR AN NRC RIALS UCENSES. Program Speedic Guidance About Fixed Gauge UCENSE.10CFR Parts 30.14,30.15, 30.16,30.18.30.19 & 30.20, UcensesMt Report For Comment External Does pm NUREG/CR4526 VO1: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited lONS ON THE EXTERNAL STRESS CORROSION CRACKING OF TYPE 304 AUSTENITIC STAINLESS STEEL NUR R 526

R BA.1ST ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE' Food Chain POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Food Chain Uncertainty Assessment. Main EEternal Strees Corrosion Cracking NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN NUR CR4523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE lONS ON THE EXTERNAL STRESS CORROSION CRACKING OF UNCERTAINTY ANALYSIS.

Food Chan Uncertainty TYPE 304 AUSTENITIC STAINLESS STEEL AssessmenLAppendices.

Extraction Fracture Mechanica NUREG-1569 DRFT: DRAFT STANIMRD REVIEW PLAN FOR IN SITU NUREG/CP-0157 V02: PROCEEDINGS OF THE TWENTY-FOURTH LEACH URANIUM EXTRACTION LICENSE APPUCATONS-WATER REACTOR SAFETY INFORMATON MEETING. Reactor Pres-sure Vessel Embrittlement And Thermal Annealing, Reactor Vessel

• • • *""*~

NU E R4534 V01: FRAPCON-3: MODIFICATIONS TO FUEL ROD MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH-NUREG/CE4233 V02: STABluTY OF CRACKED PIPE UNDER SEIS-BURNUP APPLICATION.

MIC/ DYNAMIC DISPLACEMENT-CONTROLLED STRESSES. Subtask 1.2 Final Fleport.

Faugue NUREG/CR4233 V03: CRACK STABluTY IN A REPRESENTATIVE NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY.

STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS NAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask 1.3 Final OF NUCLEAR SYSTEMS.

p NUR CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE-Fe% Crock %

SEARCH PROGRAM (IPIRG) PROGRAM,Propam Final Report.

NUREG/CR4356: REVIEW OF INDUSTRY EFFORTS TO MANAGE NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING.

RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report.

AND FEEDRING CRACKING AND WALL THINNING.

Fracture Toughnese Foodwater Nozzle NUREG/CR-4219 V12 N2: HEAVY-SECTION STEEL TECHNOLOGY NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE PROGRAM. Semiannual Progress Report For April 1995 Through Sep.

PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING.

tamber 1995.

AND FEEDRING CRACKING AND WALL THINNING.

NUREG/CR4233 V02: STABluTY OF CRACKED PIPE UNDER SEIS-MIC/ DYNAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask Fleid Experiment 1.2 Final Report.

NUREG/CR 4918 V10: CONTROL OF WATER INFILTRATION INTO NUREG/CR4233 V03: CRACK STABluTY IN A REPRESEN) ATIVE NEAR SURFACE LOW LEVEL WASTE DISPOSAL UNITS. Final Report PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-On Field Expenments At A Humid Regon Site,Beltsville, Maryland.

NAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask 1.3 Final Field Lysimeter NUR CR4363: EFFECTS OF THERMAL AGING AND NEUTRON IR-l NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW-RADIATON ON THE MECHANICAL PROPERTIES OF THREE-WIRE LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR STAINLESS SIEEL WELD OVERLAY CLADDING.

FISCAL YEAR 1996. Annual Report.

NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT. PIPE Final Environmental Impact Statement LOCATIONS. Final Report. September 1991 - November 1995.

NUREG-1508: FINAL ENVIRONMENTAL IMPACT STATEMENT TO NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-CONSTRUCT AND OPERATE THE CROWNPOINT URANIUM SOLU-FIED A 302 GRADE B PLATE MATERIALS. DATA ANALYSIS.

TION MINING PROJECT, CROWNPOINT, NEW MEXICO Docket No.

NUREG/CR4426 V02: DUCTILE FRACTURE TOUGHNESS OF MODI-40-8968.(Hydro Resources, Inc.)

FIED A 302 GRADE B PLATE MATERIALS. Data Records.

54 Subject Index NUREG/CR4446: FRACTURE TOUGHNESS EVALUATONS OF TP304 Heat Flux STAINLESS STEEL PIPES.

NUREG/CR4507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A NUREG/CR4506: EMBRITTLEMENT DATA BASE, VERSON 1.

DOWNWARD FACING CURVED SURFACE, UREG/CR4381: CRITICALITY BENCHMARK GUIDE FOR LIGHT-NUREG/CR-0200 R5V1P1: SCALE: A MODULAR CODE SYSTEM FOR WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-PACKAGES' CENSING EVALUATION. Control Modules C4, C6.

q Fuel Cycle Facluty NUREG/CR 0200 RSV1P2-SCALE: A MODULAR CODE SYSTEM FOR j

NUREG-1601: CHEMICAL PROCESS SAFETY AT FUEL CYCLs FACill-PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-l TIES.

CENSING EVALUATON. Control Modules S1 - H1.

NUREG/CF14200 R5V2P1: SCALE: A MODULAR CODE SYSTEM FOR Fuel Rock PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-l NUREG 1275 V12:

OPERATING EXPERIENCE FEEDBACK CENSING EVALUATION. Functional Modules F1 - F8.

REPORT. Assessment Of Spent Fuel Coohn9 NUREG/CR4200 R5V2P2: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-Fuel Rod CENSING EVALUATON.Functonal Modules F9 - F11.

NUREG/CR4534 V01: FRAPCON-3. MODIFICATIONS TO FUEL ROD NUREG/CH 0200 R$V2P3: SCALE: A MODULAR CODE SYSTEM FOR MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH' PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-DURNUP APPLICATION.

CENSING EVALUATON.Functonal Modules F16 F17.

NUREG/CR4200 R5V3: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-R 4 3: EFFECTS OF SMOKE ON FUNCTIONAL CIRCUtTS.

CENSING EVALUATION. Miscellaneous.

NUREG/CR4167: LATE-PHASE MELT PROGRESSION EXPERIMENT Funding Assurance NUREG-1577 DRFT FC: STANDARD REVIEW PLAN ON POWER REAC-MP-2.Results And Analysis.

TOR LICENSEE FINANCIAL QUALIFICATIONS AND DECOMMIS-Hea SlONING FUNDING ASSURANCE. Draft Report For Comment.

UE 59 N1 H AVY ECTION STEEL IRRADIATION Geochemical Transport PROGRAM. Semiannual Progress Repori For October 1995 Through NUREG/CR4515: BLT-EC (BREACH, LEACH, AND TRANSPORT EOUI-March 1996.

LIBRIUM CHEMISTRY) DATA INPUT GUIDE.A Computer Model For NUREG/CR 5591 V07 N2: HEAVY SECTON STEEL IRRADIATION Simulating Release And Coup 6ed Geochemical Transport Of Contam6-PROGRAM. Semiannual Progress Report For April Through September nants From A Subsurface Disposal Facility.

1996.

Geodetic Data Heavy-Section Steel Technology Program NUREG/CR-6529: VALIDATION OF TECTONIC MODELS FOR AN IN-NUREG/CR-4219 V12 N2: HEAVY-SECTION STEEL TECHNOLOGY TRAPLATF SCISMIC ZONE. CHARLESTON SOUTH CAROLINA WITH PROGRAM. Semiannual Progress Report For April 1995 Through Sep-GPS GEODETIC DATA.

tember 1995.

NUREG/CR-4219 VM N1: HEAVY-SECTION STEEL TECHNOLOGY t

/ 4586: HORIZONTAL VELOCITIES IN THE CENTRAL AND 996 EASTERN UNITED STATES FROM GPS SURVEYS DURING THE 19871996 INTERVAL High Burnup Fuel NUREG/CP-0157 V01: PROCEEDINGS OF THE TWENTY-FOURTH NU G R43 2 V01: RECOMMENDATIONS FOR PROBABILISTIC WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND son, Q Burnup Fuel Containnwnt And Structural 4ng Hig nyerature NUREG CF 2 0 R MENDATONS FOR PROBABILISTIC SELSMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND NUREG/CR4391: DETONATION CELL SIZE MEASUREMENTS IN USE OF EXPERTS. Appendices' HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE BNL HIGH-TEMPERATURE COMBUSTION FACILITY.

Geostatistics NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH High-Burnup Application SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

NUREG/CR-6534 V01: FRAPCON-3: MODIFICATIONS TO FUEL ROD NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR HIGH-THE APACHE LEAP RESEARCH SITE.May 1995-1996.

BURNUP APPLICATION.

Guidance High-Level Weste NUREG 1608 DRFT FC: CATEGORIZING AND TRANSPORTING LOW NUREG/CR4513 N01: NRC HIGH-LEVEL RADIOACTIVE WASTE MAN-SPECIFIC ACTIVITY MATERIALS AND SURFACE CONTAMINATED AGEMENT PROGRAM ANNUAL PROGRESS REPORT: FISCAL YEAR OBJECTS. Draft Rept For Comment.

1996.

Gu6delines Horizontal Velocities NUREG/CR4463 RO1: REVIEW GUIDELINES FOR SOFTWARE LAN-NUREG/CR4586: HORIZONTAL VELOCITIES IN THE CENTRAL AND GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY EASTERN UNITED STATES FROM GPS SURVEYS DURING THE

)

SYSTEMS. Final Report.

1987 1996 INTERVAL HEPA Filter Human Factor i

N!IREG/CP-0153: PROCEEDINGS OF THE 24TH DOE /NRC NUCLEAR NUREG 1545: EVALUATION CRITERIA FOR COMMUNICATIONS-RE-AIR CLEANING AND TREATMENT CONFERENCE. Held in Portland, LATED CORRECTNE ACTION PLANS.

Oregon, July 15-18,1996.

Halogen lone Human Factors Engineering NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN NUREG/CR4393. INTEGRATED SYSTEM VALIDATION. METHODOLO-

^

N REG M

CTORS ENGINEERING (HFE) INSIGHTS 3 AUSTEN C STAINLE TE I

FOR ADVANCED REACTORS BASED UPON OPERATING EXPERI-Hazard Evaluation ENCE.

NU EG-1601: CHEMICAL PROCESS SAFETY AT FUEL CYCLE FACILi-Humid Region Site NUREG/CR-491f; V10: CONTROL OF WATER INFILTRATION INTO Health Phya6c NEAR SURFACE LOW-LEVEL WASTE DISPOSAL UNITS.Fmal Report NUREG/CH4547: DOSFAC2 USER'S GUIDE.

On Field Exprmments At A Hurrud Regen Site,Beltsville,Marylarwi

r Subject index 55 l

Hydrogen Combustion NOREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY i

NUREG/CR4530 DELIBERATE IGNITION OF HYDROGEN.AlR STEAM PROGRAM. Semiannual Report, August 1995 March 1996.

MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

Ineervice Test Interval HyCl-M Modelin0 NUREG/CR4508: COMPONENT UNAVAILABILITY VERSUS INSERV-NUREG/CR4505 Vol: THE POTENTIAL FOR CRITICALITY FOLLOW-ICE TEST (IST) INTERVAL: EVALUATIONS OF COMPONENT AGING ING DISPOSAL OF URANIUM AT LCW LEVEL WASTE EFFECTS WITH APPLICATIONS TO CHECK VALVES.

FACILITIES.Urardum Blended With Soil.

l Integrated System i

iPIRG2TW 1

^

^

NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS GY DREV TERlA' WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT PIPE i

LOCATIONSTmal Report. September 1991 November 1995.

Integrity Databees 18FSI NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-t NUREG-1536: STANDARD REVIEW PLAN FOR DRY SPENT FUEL TABASE.

STORAGE SYSTEMS. Final Report Interim Storage Ignetton NUREG 1571: INFORMATION HANDBOOK ON INDEPENDENT SPENT i

NUREG/CR4391: DETONATION CELL SIZE MEASUREMENTS IN FUEL STORAGE INSTALLATIONS.

l HIGH-TEMPERATURE HYDROGEN-AIR STEAM MIXTURES AT THE BNL HIGH-TEMPERATURE COMBUSTON FACILITY.

lon Exchange NUREG/CR-6530: DELIBERATE IGNITION OF HYDROGEN-AIR-STEAM NUREG/CR-5229 V09: FIELD LYSIMETER WVESTIGATONS: LOW-i MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR FISCAL YEAR 1996. Annual Report.

r impact Testing l

NUREG/CR4379: AN IMPROVED CORRELATON PROCEDURE FOR trradiated Reactor Fuel SUBSIZE AND FULL-SIZE CHARPY IMPACT SPECIMEN DATA.

NUREG-0725 R12 PUBLIC INFORMATON CIRCULAR FOR SHIP.

In Setu Evaluation NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-Irradiation STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS NUREG/CR4363: EFFECTS OF THERMAL AGING AND NEUTRON IR-OF NUCLEAR SYSTEMS.

l RADIATION ON THE MECHANICAL PROPERTIES OF THREE-WIRE in SMu unch STA!NLESS STEEL WELD OVERLAY CLADDING.

NUREG-1506: FINAL ENVIRONMENTAL IMPACT STATEMENT TO WREGERM REWS & CHM V-NM WACT TESDNG CONSTRUCT AND OPERATE THE CROWNPOINT URANIUM SOLU-OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 DE-r l

TION MINING PROJECT, CROWNPOINT, NEW MEXICO. Docket No.

GREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMERCIAL RE-40 89681 Hydro Resources. Inc.)

ACTOR CAVITY, NUREG-1569 DAFT: DRAFT STANDARD REVIEW PLAN FOR IN SITU LEACH URANIUM EXTRACTION LICENSE APPLICATIONS.

U

/CR4389: IPIRG 2 TASK 1 PIPE SYSTEM EXPERIMENTS In-Vessel Component WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT-PIPE l

NUREG 1616: FEASIBILITY OF UNDERWATER WELDING OF HIGHLY LOCATONS. Final ReportSeptember 1991 - November 1995.

l IRRADIATED IN VESSEL COMPONENTS OF BOILING WATER NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY l

REACTORS.A Literature Review.

RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report In Vossel Retention J-R Curve NUREG/CR-6507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A NUREG/CR4426 V02: DUCTILE FRACTURE TOUGHNESS OF MODI-l DOWNWARD FACING CURVED SURFACE.

FIED A 302 GRADE B Pl. ATE MATERIALS. Data Records.

NUREG/CR4446: FRACTURE TOUGHNESS EVALUATIONS OF TP304 r

i lndpendent Spent Fuel Storage installation STAINLESS STEEL PtPES.

NUREG 1571: INFORMATION HANDBOOK ON INDEPENDENT SPENT NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRITY FUEL STORAGE INSTALLATIONS.

RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report Individual Plant Esemination LOCA NUREG-1603 DRFT:

INDIVIDUAL PLANT EXAMINATION NUREG/CR4538: EVALUATION OF LOCA WITH DELAYED LOOP AND DATABASE. User's Guide.

LOOP WITH DELAYED LOCA ACCIDENT SCENARIOS.

Industriel Radiography Loop NUREG-1556 V2 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE

• NUREG/CR4538: EVALUATION OF LOCA WITH DELAYED LOOP AND RIALS LICENSES. Program Specific Guidance About industrial Ra@

LOOP WITH DELAYED LOCA ACCIDENT SCENARIOS.

rephy Licenses. Draft Report For Use And Comment.

LWR NUREG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION AND NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN SUBSURFACE FLOW AND TRANSPORT FOR SOMP SITES.

LIGHT WATER REACTORS. Semiannual Report. January 1996 - June 1996.

Information Digest NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG-1350 V00 NUCLEAR REGULATORY COMMISSON INFORMA.

LIGHT WATER REACTORS. Semiannual Report. July-December 1996.

TON DIGEST.1997 Edition.

NUREG/CR4361: CRITICALITY BENCHMARK GUIDE FOR LIGHT-WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE ingestion Pathway PACKAGES.

i I

NUREG/CR4523 Voi: PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Food Chain Uncertamty Assessment. Main Land Fraction Report.

NUREG/CR4525: SECPOP90- SECTOR POPULATION, LAND FRAC.

NUREWCR4523 V02: PROBABILISTIC ACCfDENT CONSEQUENCE TlON, AND ECONOMIC ESTIMATON PROGRAM.

UNC4 H I AINTY ANALYSIS.

Food Cham Uncertamty AseensrnentAppendees he Cruces Trench Site NUREG/CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES inservios inapoction TRENCH SITE: EXPERIMENT IIB.

NUREG/CR4181 RO1: A PILOT APPLICATON OF RISK-INFORMED METHODS TO ESTABLISH INSERVICE INSPECTION PRIORITIES Leek Rate FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR NUREG/CP 0155: PROCEEDINGS OF THE SEMINAR ON LEAK POWER STATION.

BEFORE BREAK IN REACTOR PIPING AND VESSELS.

56 Subject index Leek-severe-areek ught water Reactor NUREG/CP4155: PROCEEDINGS OF THE SEMINAR ON LEAK NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN BEFORE BREAK IN REACTOR PIPING AND VESSELS.

UGHT WATER REACTORS. Semiannual ReportJanuary 1996 - June NUREG/CR4233 V02: STABluTY OF CRACKED PtPE UNDER SEIS-

1990, MIC/ DYNAMO DISPLACEMENT CONTROLLED STRESSES. Subtask NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN 1.2 Final Report-UGHT WATER REACTORS. Sermannual Report. July-Decernber 1996.

MJREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE-NUREG/CR4361: CRITICAUTY BENCHMARK GUIDE FOR LIGHT-SEARCH PROGRAM (IPIRG) PROGRAM. Program Final Report WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE PACKAGES.

NUREG0750 V44101: INDEXES TO NUCLEAR REGULATORY COM-Lose Of Cooient AccWnt NUREG/CR4541902: PHENOMENA IDENTIFICATION AND RANKING l

NU V

DEX REGULATORY COM.

TABLES FOR./ WESTINGHOUSE AP600 SMALL BREAK LOSS-OF-MISSON ISSUANCES.

-December 1996'ATORY COMMISSON IS.COOLANT ACCIDENT, MAIN STEAM LINE BREAK, AND STEAM R REGUL NUREG-0750 V44 N05: N GENERATOR TUBE RUPTURE SCENARIOS.

SUANCES FOR NOVEMBER 1996. Pages 229-314.

NUREG4750 V44 N06: NUCLEAR RLGULATORY COMMISSION IS-Low-Lml Weste SUANCES FOR DECEMBER 1996. Paoes 315 432.

NUREG4750 V45101: INDEXES TO NUCLEAR REGUuTORY COM-NUREG/CR4229 V09-FIELD LYSIMETER INVESTIGATIONS: LOW.

MISSON ISSUANCES. January March 1997.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG-0750 V45102: INDExt S TO NUCLEAR REGULATORY COM-FISCAL YEAR 1996. Annual Report MISSION ISSUANCES. January, June 1997.

NUREG/CR4505 V01: THE POTENTIAL FOR CRITICAUTY FOLLOW-NUREG.0750 V45 N01: NUCLEAR REGULATORY COMMISSION IS-ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE SUANCES FOR JANUARY 1997. Pages 147.

FACluTIES. Uranium Blen: led With Soil.

NUREG0750 V45 NO2: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR FEBRUARY 1997. Pages 49-93.

Low-Level Weste Diepoeel NUREG.0750 V45 NO3: NUCLEAR HEGULATORY COMMISSION IS-NUREG/CR-4916 V10: CONTROL OF WATER INFILTRATION INTO SUANCES FOR MARCH 1997.Pages95-263.

NEAR SURFACE LOW-LEVEL WASTE DISPOSAL UNITS. Final Report NUREG4750 V45 N04: NUCLEAR REGULATORY COMMISSION IS-On Field Experiments At A Humid Flegion Site,Beltsville, Maryland.

SUANCES FOR APRIL 1997.Pages 265-353.

NUREG-0750 V45 N05: NUCLEAR REGULATORY COMMISSION IS-Lo,,gp,,ggig aggg,ggygegerge NRRG ATORY COMMISSION IS-SPECIFIC ACTMTY MATERIALS AND SURFACE CONTAMINATED NUREG 1606 DRFT FC: CATEGORIZING AND TRANSPORTING LOW NURE 5

SUANCES FOR JUNE 1997. P m 437 495.

NUREG4750 V46 N01: NUC R REGULATORY COMMISSION IS.

OBJECTS.Drafl Rept For Comrnent.

SUANCES FOR JULY 1997.Pages 120.

NUREG-0750 V46 NO2. NUCLEAR REGUuTORY COMMISSION IS-Low Head IntegrMy NUREG/CR4507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON A SUANCES FOR AUGUST 1997. Pages 2148.

DOWNWARD FACING CURVED SURFACE.

LG Code Q NUREG/CR4563: LG EXCITATON, ATTENUATON, AND SOURCE MEPAS SPECTRAL SCAUNG IN CENTRAL AND EASTERN NORTH AMER.

NUREG/CR4566: DESCR!PTON OF MULTIMEDIA ENVIRONMENTAL ICA..

POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-FICATION FOR THE NUCLEAR REGULATORY COMMiSSON.

Lg Excitation NUREG/CR4563: LG EXCITATION, ATTENUATION, AND SOURCE Magneocope SPECTRAL SCALING IN CENTRAL AND EASTERN NORTH AMER.

NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-ICA.

STRUMENT FOR IN SITU EVALUATON OF STEEL COMPONENTS OF NUCLEAR SYSTEMS.

License Application NUREG-1569 DRFT: DRAFT STANDARD REVIEW pun FOR IN SITU Main Steam Line Broek LEACH URANIUM EXTRACTON LICENSE APPUCATIONS.

NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF-G

. NRC ANTITRUST LICENSING ACTONS. 1976-1996-ERA OR E UP U E S NAR License Renewal NUREG 1555 DRFT: ENVIRONMENTAL STANDARD REVIEW V01: CONSOLIDATED GUIDANCE ABOUT MATERIALS N

Sta Review Plans For Env6ronmental Reviews For Nuclear LICENSES. Program Specirc Guidance About Portable Gauge UcensesFnal Report NUREG-1611: AGING MANAGEMENT OF NUCLEAR POWER PLANT NUREG-1556 V2 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-CONTAINMENTS FOR UCENSE RENEWAL NUREG/CR4528: ENVIRONMENTAL ASSESSMENT PROPOSED U-RIALS UCENSES. Program Specife Guidance About industrial Radiog-NURE6 Ucenses. Draft Report For Use And Comment.

raphv CENSE RENEWAL OF NUCLEAR METALS.INC. CONCORD, MASSA-1556 V3 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-CHUSETTS.

RIALS UCENSES. Applications for Sealed Source And Device Evalue-License Termination tion And Registration. Draft Report For Comment.

NUREG-1496 V01: FtNAL GENERIC ENVIRONMENTAL IMPACT STATE-NUREG 1556 V4 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-

)

MENT IN SUPPORT OF RULEMAK!NG ON RADIOLOGICAL CRITE.

RIALS UCENSES. Program Specific Guidance About Flxed Gauge RIA FOR UCENSE TERMINATION OF NRC-UCENSED NUCLEAR Ucenses. Draft Report For Comment FACILITIES.Mam ReportFinal Report NUREG 1556 V5 DRF FC: CONSOLIDATED GUIDANCE ABOUT MATE-NUREG 14W6 V02: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-RIALS UCENSES. Program-Specife Guidance About Self-Sivelded irra-MENT IN SUPPORT OF RULEMAKING ON P. RADIOLOGICAL CRITE-diator Ucenses. Draft Report For Comment 1

RIA FOR UCENSE TERMINATION OF NRC-UCENSED NUCLEAR FACIUTIES.Apperdces A And B. Final Report.

Medical Factilty NUREG-1496 V03: FINAL GENERIC ENVIRONMENTAL lMPACT STATE-NUREG-1516: MANAGEMENT OF RADTCTIVE MATERIAL SAFETY z

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRiTE-PROGRAMS AT MEDICAL FACIUTIES. Final Report i

I RIA FOR UCENSE TERMINATION OF NRC. LICENSED NUCLEAR l

FACIUTIES.Appendees C41. Final Report Mett Progroselon NUREG/CR4167: LATE-PHASE MELT PROGRESSION EXPERIMENT Liconese Event Report MP-2.Results And Analysis.

NUREG/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1995. A Status Report.

Metallic Melt Relocation j

NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4527: FINAL RESULTS OF THE XR2-1 BWR METALUC DAMAGE ACCIDENTS: 196243.A Status Report.

MELT RELOCATON EXPERIMENT.

l

Subject index 57 M!crostructural Change Huclear Metals, tric NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-

"UREG/CR4528: ENVIRONMENTAL ASSESSMENT PROPOSED U-STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS

'ENSE RENEWAL OF NUCLEAR METALS,1NC. CONCORD, MASSA-OF NUCLEAR SYSTEMS.

HUSETTS.

Mill Talling N clear Power Plant NUREG-1532: FINAL TECHNICAL EVALUATION REPORT FOR THE NUREG/CR4295: REASSESSMENT OF SELECTED FACTORS AF-PROPOSED REVISED RECLAMATON PLAN FOR THE ATLAS COR-WTING SITING OF NUCLEAR POWER PLANTS.

PORATION MOAB MILLSource Matenal Ucense No. SUA 917. Docket No. 40-3453 (AJas Corporation)

Nuclear Regulation Modul Eva4uation NUREG-1610- CONTROLUNG THE ATOM.The Beginrungs Of Nuclear Regulation, 1946 1962.

NUREG/CR4481 V01: REVIEW OF 3,40DELS USED FOR DETERMIN-ING CONSEQUENCES OF UF(6) RELEASE.Developrnent Of Model Nuclear Safety Research Evaluation Cntena.

NUREG/CP-0161: TRANSACTIONS OF THE TWENTY-FIFTH WATER NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-REACTOR SAFETY INFORMATION MEETING.

ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report.

Nuchar Waste Eodular Construction NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH NUREG/CRet86: ASSESSMENT OF MODULAR CONSTRUCTION FOR SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

SAFETY RELATED STRUCTURES AT ADVANCED NUCLEAR POWER PLANTS.

Nuclear Waste Management co ten Pool NUREG/CP-0153: PROCEEDINGS OF THE 24TH DOE /NRC NUCLEAR NUREG/CR4167: LATE-PHASE MELT PROGRESSION EXPERIMENT AIR CLEANING AND TREATMENT CONFERENCE. Held in Portland, Oregon, July 15-18.1996.

MP-2.Results And Analysis.

Motor-Operated Valve Occupational Radiation Exposure NUREG 0713 V17: OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR4478: MC.OR-OPERATED VALVE (MOV) ACTUATOR MOTOR AND GEARBOX TESTING.

MERICAL NUCLEAR POWER REACTORS AND OTHER FACluTIES.1995. Twenty-Eighth Annual Report.

Multi Phase Flow

• 8 NUREG/CP-0159: PROCEEDINGS OF THE OECD/CSNI WORKSHOP ON TRANSIENT THERMAL-HYDRAULIC AND NEUTRONIC CODES REG-127 2-OPERATING EXPERIENCE FEEDBACK REPORT. Assessment Of Spont Fuel CooliN REQUIREMENTS. Held in Annapolis. Maryland, USA,Novernber 54, 1996.

NUREG/CR4400 HUMAN FACTORS ENGIN'EERING (HFE) INSIGHTS FOR ADVANCED REACTORS BASED UPON OPERATING EXPERI-NRC Streiegic Plan ENCE.

NUREG 1614 V0': NRC STRATEGIC PLAN. Fiscal Year 1997 Fiscal Year 2002.

Operating License NUREG/CR4558: NRC ANTITRUST UCENSING ACTONS, 1978-1996.

Needle Applicator NUREG/CR 6074 V03: SEALED SOURCE AND DEVICE DESIGN REG NT R08: OPERATOR UCENSING EXAMINATION STAND.

SAFETY TESTING. Technical Report On The Findings Of Task 4.Inves.

tigation Of A Failed Brachytherapy Needle Applicator.

ARDS FOR POWER REACTORS' Neutron Doolmetry Organization Chart NUREG/CR4454: POOL CRITICAL ASSEMBLY PRESSURE VESSEL NUREG4325 R22: U.S. NUCLEAR REGULATORY COMMISSION OR-FACIUTY BENCHMARK.

GANIZATON CHARTS AND FUNCTONAL STATEMENTS.Novernber 1997.

Neutronic NUREG/CP 0159: PROCEEDINGS OF THE OECD/CSNI WORKSHOP PRA ON TRANSIENT THERMAL-HYDRAUUC AND NEUTRONIC CODES NUREG-1602 DRFT FC: THE USE OF PRA IN RISK-INFORMED REQUIREMENTS Held in Annapolis, Maryland. USA. November 5-8, APPLICATIONS. Draft Rept For Cornment.

I PWR Nondestructive Evaluation NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF NUREG/CR4181 RO1: A PILOT APPUCATION OF RISK-INFOR'.ED GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR METHODS TO ESTABLISH INSERVICE INSPECTON PRIOhiTIES POWER PLANTS.

FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE POWER STATION.

PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING, AND FEEDRING CRACKING AND WALL THINNING.

Nuclear Air Cleaning NUREG/CR4469: EXPERIMENTS TO INVESTIGATE DIRECT CON-NUREG/CP-0153: PROCEEDINGS OF THE 24TP DOE /NRC NUCLEAR TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE AIR CLEANING AND TREATMENT CONFE% GE. Held in Portlared, CALVERT CLIFFS NUCLEAR POWER PLANT.

Oregon, July 15 18,1996.

p Nuclear Compw NUREG-1492: REGULATORY ANALYSIS ON CRITERIA FOR THE RE-NUREG/CR4181 R01: A PILOT APPUCATION OF RISK-INFORMED LEASE OF PATIENTS ADMINISTERED RADCACTIVE METHODS TO ESTABLISH INSERVICE INSPECTION PRORITIES MATERIALFinal Report.

FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION.

Performance Assessment NUREG/CR4513 N01: NRC HIGH-LEVEL RAD'OACTIVE WASTE MAN-Nucleer Criticality AGEMENT PROGRAM ANNUAL PROGRESS REPORT: FISCAL YEAR NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICAUTY SUDE 1996.

RULE.Technscal Basis.

NUREG/CR4505 V01: THE POTENTIAL FOR CRITICAUTY FOLLOW-Performance Measure ING DISPOSAL OF URANIUM AT LOW LEVEL WASTE NUREG-1542 V02: ACCOUNTABILITY REPORT FISCAL YEAR 1996.

FACIUTIES. Uranium Blended With Soil.

Petitions For Rulemaking Nuclear Medicine NUREG 0936 V15 NO2: NRC REGULATORY AGENDA.Sem6 annual NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA-Report. July-December 1996.

TION OF RADIOLABELED ANTIDODIES CONTAINING Y-90,TC-99M,8-NUREG-0936 V16 N01: NRC REGULATORY AGENDA.Sermannual 131, AND LU 177.

Report. January-June 1997.

7.

I 58 Subject Index Phenomene identification NUREG/CR4454: POOL CRITICAL ASSEMBLY PRESSURE VESSEL NUREG/CR4474: PRELIMINARY PHE:40MENA IDENTIFICATION AND FACILITY BENCHMARl(.

RANKING TABLES (PIRT) FOR SBWR STARTUP STABILITY.

NUREG/CR4541802: PHENOMENA IDENTIFICATION AND RANKING Presourtzed Thermal Shock TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF.

NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-COOLANT ACCIDENT MAIN STEAM LINE BREAK, AND STEAM TABASE.

GENERATOR TUBE RUPTURE SCENAROS.

Pressurized Water Reactor NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF U G/CR4631: EFFECTS OF RADIOACTIVE HOT PARTICLES ON GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR PIG SKIN' POWER PLANTS.

NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE

pip, MUREG/CR4446: FRACTURE TOUGHNESS EVALUATONS OF TP304 PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING.

STAINLESS STEEL PIPES.

AND FEEORING CRACKING AND WALL THINNING.

NUREG/CR-6452: THE SECOND INTERNATIONAL PIPING INTEGRITY NUREG/CR4489 EXPERIMENTS TO INVESTIGATE DIRECT CON-RESEARCH GROUP (IPIRG-2) PROGRAM. Final Report TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE CALVERT CLIFFS NUCLEAR POWER PLANT.

NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMEtFS Primary Doolmetry WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT PIPE NUREG/CR4581: CONSIDERATIONS IN THE APPLICATON OF THE LOCATONS. Final ReportSaptember 1991 - November 1995.

ELECTRONIC DOSIMETER TO DOSE OF RECORD.

EG/CR4414 ING BE ARK R BLEMS FOR THE WES-N REG /CR4 LISTIC ACCIDENT CONSEQUENCE NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE UNCERTAINTY ANALYSIS. Food Chain Uncertamty AssessmentMam PRESSURIZED WATER REACTOR FEEDWATER NOZZLE PIPING, Report AND FEEDRING CRACKING AND WALL THINNING.

NUREG/CR4523 VU2: PROBABILISTIC ACCOENT CONSEQUENCE NUREG/CR4519: SCREENING REACTOR STEAM / WATER PIPING UNCERTAINTY ANALYSIS.

Food Cham Uncertamty SYSTEMS FOR WATER HAMMER.

Assessment.Appendees.

NUREG/CR4526 VO1: PROBABILISTIC ACC.0ENT CONSEQUENCE Piping integrity UNCERTAINTY ANALYSIS. Uncertainty Assessment For Depot.ited NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE-Material And Extemal Doses. Main Report SEARCH PROGRAM (IPIRG) PROGRAM. Program Foal Report-NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-POSITED MATERIAL AND EXTERNAL DOSES.Appendees.

E CR4181 R01: A PILOT APPLICATON OF RISK-INFORMED METHODS TO ESTABLISH INSERVICE INSPECTION PRIORITIES Probabilistic Risk Assessment FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR NUREG-1602 DRFT FC: THE IlSE OF PRA IN RISK INFORMED NU R

3 V'03: CRACK STABluTY IN A REPRESENTATIVE APPLICATIONS. Draft Rept For Comment.

NUREG/CR4508: COMPONENT UNAVAILABluTY VERSUS INSERV-PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-ICE TEST (IST) INTERVAL: EVALUATIONS OF COMPONENT AGING NAMIC DISPLACEMENT CONTROLLED STRCSSES. Subtask 1.3 Final EFFECTS WITH APPLICATIONS TO CHECK VALVES.

p,pn,t Ploetic Scintillator Probabilistic Seismic Hazard Analysis NUREG/CR4535: DEVELOPMENT OF CONFORMAL RESPIRATOR NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABluSTIC MONITORING TECHNOLOGY.

SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND USE OF EXPERTS Main Report Plate Material NUREG/CR-6372 V02: RECOMMENDATIONS FOR PROBABILISTIC NUREG/CR4426 V02-DUCTILE FRACTURE TOUGHNESS OF MODI-SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND FIED A 302 GRADE B PLATE MATERIALS.Deta Records.

USE OF EXPERTS. Appendices.

Pool Critical Assembly Program-Specific WUREG/CR4454: POOL CRITICAL ASSEMBLY PRESSURE VESSEL NUREG 1556 VO1: CONSOLIDATED GUIDANCE ABOUT MATERIALS FACILITY BENCHMARK.

LICENSES. Program-Specife Guidance About Portable Gauge Portable Gauge Lcenses. Final Report.

NUREG 1556 V01: CONSOUDATED GUIDANCE ADOUT M/.TERIALS L

Specrfc Guidance About Portable Gauge UR /CR4397: RADIATON SAFETY CONCERNS FOR PREGNANT OR BREAST-FEEDING PATIENTS.The Positons Of The NCRP And Post-Accident Analysis The ICRP.

NUREG/CR4481 V01: REVIEW OF MODELS USED FOR DETERMIN-ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model Radiation Dose Evalmtion Cntena.

NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA.

NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN-TION OF RADIOLABELED ANTIBODIES CONTAINING Y 90.TC-99M.l-ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluaton Report 131, AND LU-177.

Pewer Reactor Radiation Embrittlement NUREG/CR4506: EMBRITTLEMENT DATA BASE, VERSION 1-NUREG 1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-TABASE.

Practice And Procedure D60eet NUREG-0386 D08: UNITED STATES NUCLEAR REGULATORY COM-Radiation injury MISSON STAFF PRACTICE AND PROCEDURE DIGEST. Commission, NUREG/CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON Appeal Board And Ucensing Board Decesions. July 1972 June 1996.

PIG SKIN.

Pregnant Women WOREG/CR4397: RADIATON SAFETY CONCERNS rOR PREGNANT Radiation Protection OR BREAST-FEEDING PATIENTS.The Positions Of The NCRP And NUREG 1610 CONTROLLING THE ATOM.The Beginnogs Of Nuclear The ICRP, Regulation, 1946-1962.

NUREG/CR-4409 V06: DATA BASE ON DOSE REDUCTON PROJECTS Pressure Yessel FOR NUCLEAR POWER PLANTS.

NUREG/CR4379: AN IMPROVED CORRELATON PROCEDURE FOR NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICALITY SLIDE SUBSIZE AND FULL-S!ZE CHARPY IMPACT SPECIMEN DATA.

RULE.Techncal Basis.

Subject index 59 Radioactive Hot Particle Reactor Piping WUREG/CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON NUREG/CP 0155: PROCEEDINGS OF THE SEMINAR ON LEAK P4G SKIN.

BEFORE BREAK IN REACTOR PIPING AND VEGSELS.

RadioactNo Material NUREG-0383 V01 R20 DIRECTORY OF CERTIFICATES OF COMPLl-Reactor Preseure Vessel ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of NRC-NUREG/CR4399-RESULTS OF CHARPY V-NOTCH IMPACT TESTING OF STRUCTURAL STEEL SPECIMENS IRRADtATED AT 30 DE.

N20: DIRECTORY OF CERTIFICATES OF OOMPLI GREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMERCIAL RE-MU G ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of ACTOR CAVITY.

Comphance.

NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-HUREG4383 V03 R17: DIRECTORY OF CERTIFICATES OF COMPLf-FIED A 302 GRADE B PLATE MATERIALS DATA ANALYSIS.

ANCE FOR RAD'OACTIVE MATERIALS PACKAGES. Report Of NRC-Approved Quahty Assurance Programs Foi Radioactive Materials Pack.

Reactor Safety ages.

NUREG/CP-0157 V01: PROCEEDINGS OF THE TWENTY-FOURTH NUREG-1492: REGULATORY ANALYSIS ON CRITERIA FOR THE RE-WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses.

LEASE OF PATIENTS ADMINISTERED RADIOACTIVE MATERIALFinal Report.

sion, High Burnup Fuel, Containment And Suuctural Aging.

NUREG/CP-0157 V02: PROCEEDINGS OF THE TWENTY-FOURTH NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY PROGRAMS AT MEDICAL FACILITIES. Final Repwt.

WATER REACTOR SAFETY INFORMATION MEETING. Reactor Pres.

MUREG-1609 DRFT FC: STANDARD REVIEW PLAN FOR TRANSPOR-sure Vessel Eaue,,u,t And Thermal Annealing, Reactor Vessel TATON PACKAGES FOR RADIOACTIVE MATERIAL. Draft Report For Lower Head Integrity And Evaluaton And Projection of Steam Genera-Canment NUREG/Ck-0157 V03: PROCEEDINGS OF THE TWENTY-FOURTH Radiocarbon WATER REACTOR SAFETY INFORMATION MEETING.PRA And HRA, NUREG/CR4459: FIELD STUDIES AT THE APACHE LEAP RESEARCH And Probabiliste Seistmc Hazard Assessment And Seismic Siting Cnte-SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL MODELS.

ria.

NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT NUREG/CR-6042 RO1: PERSPECTIVES ON REACTOR SAFETY.

THE APACHE LEAP RESEARCH SITE.May 1995-1996.

NUREGiCR4295: REASSESSMENT OF SELECTED FACTORS AF-FECTING SITING OF NUCLEAR POWER PLANTS.

p NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA.

TION OF RADIOLABELED ANTIBODIES CONTAINING Y-90,TC-99M,l-Reactor Safety Research 131, AND LU 177.

NUREG/CP-0161: TRANSACTIONS OF THE TWENTY-FicTH WATER REACTOR SAFETY INFORMATION MEETING.

Radiological Crtteria HUREG 1496 V01: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-Reactor Shutdown MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE.

NUREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NUCLE-RfA FOR LICENSE TERMINATON OF NRC-LICENSED NUCLEAR AR ELECTRICITY-GENERATING UNITS IN THE UNITED STATES:

FACILITIES. Main Report. Final Report.

1997 2001.

NURF.G 1496 V02: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE.

Reactor Siting RIA FOR LICENSE TERMINATION OF NRC-LICENSED NUCLEAR NUREG/CR4525: SECPOP90: SECTOR POPULATON, LAND FRAC.

FACILITIES.Aooendices A And B Final Report.

TON, AND ECONOMIC ESTIMATON PROGRAM.

NUREG-1496 VM: FINAL GENERO ENVIRONMENTAL IMPACT STATE-MENT IN SUPPORT OF RULEMAKING ON RADIOLOG' CAL CRITE-Reactor Vessel RIA FOR LICENSE TERMINATON OF NRC-LICENSED NUCLEAR FACILITIES. Appendices C-H. Final Report NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-TABASE.

Radiological Does NUREG/CR4547: DOSTAC2 USER'S GUIDE.

Fiocternation Plan NUREG-1532: FINAL TECHNICAL EVALUATION REPORT FOR THE Radlonucade Transport PROPOSED REVISED RECLAMATION PLAN FOR THE ATLAS COR-MUREG/CR4523 V01: PROBABILISTIC ACCIDENT CONSEQUENCE PORATON MOAB MILLSource Material License No. SUA.917.Docitet UNCERTAINTY ANALYSIS. Food Chain Uncertainty Assessment. Man No. 40-3453.(Atlas Corporation)

MURWCR4523 V02:PROBABILISTIC ACCIDENT CONSEQUENCE Regulatory Agenda UNCERTAINTY ANALYSIS.

Food Chain Uncertainty NUREG-0936 V15 NO2-NRC REGULATORY AGENDA.Sermannual Assessment. Appendices.

ReportJuly-December 1996.

NUREG-0936 V16 N01: NRC REGULATORY AGENDA. Semiannual N

/CP PROCEEDINGS OF THE OECD/CSNI SPECIALISTS Repodanuarh M MEETING ON BORON DILUTON REACTIVITY TRANSIENTS. Held in State College, Pennsylvania. USA,0ctooer 18-20, 1995.

Regulatory Analyele NUREG-1492: REGULATORY ANALYSIS ON CRITERIA FOR THE RE-Reactor Accident LEASE OF PATIENTS ADMINISTERED RADCACTIVE NUREG/CR4295: REASSESSMENT OF SELECTED FACTORS AF.

MATERIALFinal Report FECTING SITING OF NUCLEAR POWER PLANTS.

NUREG/CR4527: FINAL RESULTS OF THE XR2-1 BWR METALLIC Regulatory And Technical Report MELT RELOCATON EXPERIMENT.

NUREG-0304 V21 NO3: REGULATORY AND TECHNICAL REPORTS NUREG/CR4538: EVALUATON OF LOCA WITH DELAYED LOOP AND (ABSTRACT INDEX JOURNAL). Compilation For Third Quarter LOOP WITH DELAYED LOCA ACCIDENT SCENAROS.

1996, July-September.

NUREG4304 V21 N04: REGULATORY AND TECH;41 CAL REPORTS R G/

V02 PROCEEDINGS OF THE TWENTY-FOURTH (ABSWCT iMa NW W Waton 6 m WATER REACTOR SAFETY INFORMATION MEETING. Reactor Pres-V22 N% MMM AW TECHM EMS see Vessel E.,Less And Thermal AnneabrgReactor Vessel (ABSWCT N JOURW Wlaw 6 % h er Lower Head integrtty And Evaluation And Protection of Steam Genera-NURE 22 02-REGULATORY AND TECHNICAL PEPORTS NU E C54400 HUMAN FACTORS ENGINEERING (HFE) INSIGHTS (ABSTRACT INDEX JOURNAL). Compilation For Second Quarter FOR ADVANCED REACTORS BASED UPON OPERATING EXPERI-1997,AprSJune Regulatory Assosoment Reactor Operator NUREG/CR4451: A SAFETY AND REGULATORY ASSESSMENT OF NUREG/CR4393: INTEGRATED SYSTEM VALIDATON: METHODOLO-GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR GY AND REVIEW CRITERIA POWER PLANTS.

60 Subject Index SDMP Reguietory Guidance NUREG 1606 DRFT FC: PROPOSED REGULATORY GUIDANCE RE-NUREG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION AND LATED TO IMPLEMENTATION OF 10 CFR 50.59 (CHANGES, Tr_STS, SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

OR EXPERIMENTS). Draft Report For Comment SECPOP90 Reguistory Miselon NUREG/CR4525: SECPOP90 SECTOR POPULATION, LAND FRAC.

NUREG 1614 V01: NRC STRATEGC PLANFiscal Year 1997 - Fiscal TON, AND ECONOMC ESTIMATON PROGRAM.

Year 2002.

Safeguards Summary Event List Ru___

.,t Energy Cost NUREG4525 V02 R05: SAFEGUARDS

SUMMARY

EVENT UST NUREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NUCLE *

(SSEL) January 1,1990 Through December 31,1996.

AR ELECTRICITY GENERATING UNITS IN THE UNITED STATES:

1997-2001.

Safeiy Evak ation NUREG-1636 DRFT FC: PROPOSED REGULATORY GUIDANCE RE-

^

EG 0090 9: REPORT TO CONGRESS ON ABNORMAL I E S Dr t R

~

t OCCURRENCES. Fiscal Year 1996.

Safety Evaluation Report NUREG-1572: SAFETY EVALUATION REPORT RELATED TO THE PE-CR 3 N01: NRC HIGH-LEVEL RADIOACTIVE WASTE MAN.

NEWAL OF THE OPERATING UCENSE FOR THE RESEARCH REAC.

AGEMENT PROGRAM ANNUAL PROGRESS REPORT: FISCAL YEAR TOR AT NORTH CAROLINA STATE UNIVERSITY.

N NUREG-1607: SAFETY EVALUATION REPORT RELATED TO THE DE-PARTMENT OF ENERGY'S PROPOSAL FOR THE IRRADIATION OF Respirator Monitor LEAD TEST ASSEMBLIES CONTAINING TRITIUM-PRODUCING MUREG/CR4535: DEVELOPMENT OF CONFORMAL RESPIRATOR BURNABLE ABSORBER RODS IN COMMERCIAL UGHT WATER RE-MONITORING TECHNOLOGY.

ACTORS.

Risk-informed Appilcation NUREG 1602 DRFT FC: THE USE OF PRA IN RISK-INFORMED Safety Program NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY APPUCATONS. Draft Regt For Comment.

PROGRAMS AT MEDICAL FACIUTIES. Final Report Rock Mass NUREG/CR4404: AN EXPERIMENTAL SCALE MODEL STUDY OF Safety System SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINTED NUREG/CR-6463 RO1: 3EVIEW GUIDEUNES FOR SOFTWARE LAN-ROCK MASS.

GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY SYSTEMS. Final Report NUREG 1496 V01: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-Safety BelatGd Structure MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITE-NUREG/CR4486: ASSESSMENT OF MODULAR CONSTRUCTION FOR RIA FOR UCENSE TERMINATION OF NRC LICENSED NUCLEAR SAFETY RELATED STRUCTURES AT ADVAfCED NUCLEAR FACILITIES. Main ReportFinal Report POWER PLANTS.

NUREG-1496 V02: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITF-Scaling RIA FOR UCENSE TERMINATION OF NRC UCENSED NUCLEAR NUREG/CR-6563: LG EXCITATION, ATTENUATION, AND SOURCE FACIUTIES.Apperklices A And B. Final Report SPECTRAL SCALING IN CENTRAL AND EASTERN NORTH AMER-NUREG-1496 V03: FINAL GENERIC ENVIRONMENTAL IMPACT STATE-lCA.

MENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CdlTE-RIA FOR LICENSE TERMINATION OF NRC UCENSED NUCLEAR Sealed Source FACluTIES. Appendices C-H. Final Report NUREG-1556 V3 DAF FC: CONSOUDATED GUlOANCE ABOUT MATE-RIALS LICENSES. Applications for Sealed Source And Device Evalua.

Rules tion And Registration. Draft Report For Comment NUREG4938 Vt5 NO2: NRC REGULATORY AGENDA. Semiannual NUREG/CR-6074 V03: SEALED SOURCE AND DEVICE DESi TN ReportJuly. December 1996-SAFETY TESTING. Technical Report On The Findings Of Task 4.Inv s.

NUREG-0936 V16 N01: NRC REGULATORY AGENDA. Semiannual tigation Of A Faned Brachythe apy Needle Applicator.

ReportJanuary-June 1997.

Sector Population Rules N MW NUREG/CR4525: SECPOP90 m OR POPUU, TON, LAND FRAC-01UREG4386 008: UNITED STATES NUCLEAR REGULATORY COM-TON, AND ECONOMIC ESTIMATON PROGRAM.

MISSON STAFF PRACTICE AND PROCEDURE DIGEST.Commession, Appeal Board And Ucensing Board Decisions. July 1972 - June 1996.

Seismic Event NUREG/CP 0157 V03: PROCEEDINGS OF THE TWENTY-FOURTH

$8WR WATER REACTOR SAFETY INFORMATION MEETING.PRA And HRA, NUREG/CR4474: PREUMINARY PHENOMENA IDENTIFICATION AND Pmbabilistic Seismic Hazard Assessment And Seismic Siting Cnte-RANKING TABLES (PIRT) FOR SBWR STARTUP STABluTY.

SCALE NUREG/CR-0200 RSV1P1: SCALE: A MODULAR CODE SYSTEM FOR Seismic Qualification PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-NUREG/CR4464: AN EVALUATON OF METHODOLOGY FOR SEIS.

MIC QUAUFICATION OF EQUIPMENT, CABLE TRAYS, AND DUCTS CENSING EVALUATON. Control Mcdules C4, C6.

NUREG/CR4200 R5V1P2: SCALE: A MODULAR CODE SYSTEM FOR IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LI-Seismic Response CENSING EVALUATON. Control Modules S1 H1.

NUREG/CR4200 RSV2P1; SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4404: AN EXPERIMENTAL SCALE.MODEL STUDY OF PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINTED CENSING EVALUATION. Functional Modules F1 F8.

ROCK MASS.

NUREG/CR 0200 RSV2P2-SCALE A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-Seismic Zone CENSING EVALUATION.Furetional Modules F9 - F11.

NUREG/CR4529: VALIDATION OF TLCTONIC MODELS FOR AN IN-NUREG/CR-0200 R5V2P3: SCALE: A MODULAR CODE SYSTEM FOR TRAPLATE SEISMIC ZONE, CHARLESTON, SOUTH CAROLINA WITH PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-GPS GEODETIC DATA.

CENSING EVALUATION. Functional Modules F16 - F17.

NUREG/CR-0200 RSV3: SCALE: A MODULAR CODE SYSTEM FOR Seismograph PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-NUREG/CR4448 V02: EVALUATION OF NATIONAL SEISMOGRAPH CENSING EVALUATION. M scellaneous.

NETWORK DETECTION CAPABluTIES. Final Report

I Subject Index 61 Self-Shloided irredtion NUREG/CR-0200 R5V2P2: SCALE: A MODULAR CODE SYSTEM FOR NUREG-1556 V5 DRF FC: CONSOUDATED GUIDANCE ABOUT MATE-PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-l RIALS UCENSES. Program-Specific Guidance About Self-Stwelded irra-CENSING EVALUATION Functonal Modules F9. F11.

diator Licenses. Draft Report For Comment NUREG/CR 0200 R5V2P3: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U.

Semiannual Report To Cong ese CENSING EVALUATON. Functional Modules F16 F17.

NUREG-1415 V10 N01: OFFICE OF THE INSPECTOR NUREG/CR-0200 RSV3: SCALE: A MODULAR CODE SYSTEM FOR GENERALSerrmannual Report To Congress. April 1,1997 September PERFORMING STANDARDIZED COMPUTER ANALYSES FOR Li-30,1997-CENSING EVALUATON. Miscellaneous.

Severe Accident Spent Fuel Shipment NUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPI-CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCI-NUREG-0725 R12: PUBUC INFORMATON CIRCULAR FOR SHIP-DENT CONDITIONS.

MENTS OF IRRADIATED REACTOR FUEL NUREG/CR4469: EXPERIMENTS TO INVESTIGATE DIRECT CON-Stalniese Steel ALVER C FS EAR R

NUREG/CR4363: EFFECTS OF THERMAL AGING AND NEUTRON IR-NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER RADIATION ON THE MECHANICAL PROPERTIES OF THREE-WIRE CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

STAINLESS STEEL WELD OVERLAY CLADDING.

Severe Reactor Accident Standard Review Plan NUREG/CR4167: LATE PHASE MELT PROGRESSON EXPERIMENT NUREG-1536: STANDARD REVIEW PLAN FOR DRY SPENT FUEL MP-2.Results And Analysis.

STORAGE SYSTEMS. Final Report.

NUREG-1555 DRFT: ENVIRONMENTAL STANDARD REVIEW Shearing Rate PLAN. Standard Review Plans For Environmental Reviews For Nuclear NUREG/CR4586: HORIZONTAL VELOCITIES IN THE CENTRAL AND Power Plants.

EASTERN UNITED STATES FROM GPS SURVEYS DURING THE NUREG-1562 DRFT FC: STANDARD REVIEW PLAN FOR APPUCA-19871996 INTERVAL TONS FOR LICENSES TO DISTRIBUTE BYPROCUCT MATERIAL TO PERSONS EXEMPT FROM THE REQUIREMENTS FOR AN NRC R / 44 4 PREL ARY PHENOMENA IDENTIFICATION AND RANKING TABLES (PIRT) FOR SBWR STARTUP STABluTY.

NUR G 9 DRFT DR SAdA EV OR IN SITU LEACH URANIUM EXTRACTION LICENSE APPLICATIONS.

Site Characterization NUREG 1574: STANDARD REVIEW PLAN ON ANTITRUST NUREG/CP-0157 V03: PROCEEDINGS OF THE TWENTY-FOURTH REVIEWS. Final Report.

WATER REACTOR SAFETY INFORMATION MEETING PRA And HRA, NUREG-1574 DRFT FC: STANDARD REVIEW PLAN ON And Probabihstic Sesamic Hazard Assessment And Sessmic Siting Cnte.

ANTITRUST. Draft Report For Comment ria.

NUREG 1577 DRFT FC: STANDARD REVIEW PLAN ON POWER REAC-TOR LICENSEE FINANCIAL QUAUFICATONS AND DECOMMIS-Sits Selection SiONING FUNDING ASSURANCE. Draft Report For Comment.

NUREG/CR4295: REASSESSMENT OF SELECTED FACTORS AF-NUREG-1609 DRFT FC: STANDARD REVIEW PLAN FOR TRANS"")R-FECTING SITING OF NUCLEAR POWER PLANTS.

TATON PACKAGES FOR RADIOACTIVE MATERIALDraft Report For Comment.

Slide Rule NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICALITY SUDE Startup StatWilty RULE. Technical easis.

NUREG/CR4474: PRELIMINARY PHENOMENA IDENTIFICATION AND g

RANKING TABLES (PIRT) FOR SBWR STARTUP STABluTY, NUREG/CR4541 R02 PHENOMENA IDENTIFICATION AND RANKING Station Blackout TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF.

COOLANT ACCIDENT, MAIN STEAM LINE BREAK, AND STEAM NUREG/CR4527: FWAL RESULTS OF THE XR2-1 BWR METALLIC GENERATOR TUBE RUPTURE SCENARIOS.

MELT RELOCATON EXPERIMENT

• Smoke Steam Bubble NUREG/CR4543: EFFECTS OF SMOKE ON FUNCTIONAL CIRCUITS.

NUREG/CR4519: SCREENING REACTOR STEAM / WATER PIPING SYSTEMS FOR WATER HAMMER.

Software Languages NUREG/CR4463 RO1: REVIEW GUIDEUNES FOR SOFTWARE LAN-Steam Condensation GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY NUREG/CR4530 DEUBERATE IGNITION OF HYDROGEN-AIR-STEAM SYSTEMS. Final Report.

MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

Solute Transport Steam Generator NUREG/CR4437: FLOW ANO TRANSPORT AT THE LAS CRUCES NUREG-1604: CIRCUMFERENTIAL CRACKING OF STEAM GENERA-TRENCH SITE: EXPERIMEN s 118.

TOR TUBES.

NUREG/CP-0154: PROCEEDINGS OF THE CNRA/CSNI WORKSHOP G-1508: FINAL ENVIRONMENTAL IMPACT STATEMENT TO pu CONSTRUCT AND OPERATE THE CROWNPOINT URANIUM SOLU-NUREG/CR-4409 V06: DATA BASE ON DOSE REDUCTION PROJECTS TON MINING PROJECT, CROWNPOINT, NEW MEXICO. Docket No.

FOR NUCLEAR POWER PLANTS.

4049684 Hydro Pesources, incJ NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY Spent Fuel PROGRAM. Semiannual Report, August 1995 - March 1996.

NUREG-1275 V12-OPERATING EXPERIENCE FEEDBACK NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING REPORT.Aasessment Of Spent Fuel Cooling TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF.

NUREG-1536: STANDARD REVIEW PLAN FOR DRY SPENT FUEL COOLANT ACCENT, MAIN STEAM UNE BREAK, AND STEAM STORAGE SYSTEMS. Final Report.

GENERATOR TUBE RUPTURE SCENAROS.

NUREG/CH-0200 RSVIP1: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-Steam / Water System CENSING EVALUATION Control Modules C4, C6.

NUREG/CR4519: SCREENING REACTOR STEAM / WATER PIPING NUREG/CR-0200 R5 VIP 2: SCALE: A MODULAR CODE SYSTEM FOR SYSTEMS FOR WATER HAMMER.

PERFORMING STANDARDl2ED COMPUTER ANALYSES FOR U-CENSING EVALUATON. Control Modules S1 - H1.

Steel Component NUREG/CR-0200 R5V2P1: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U.

STRUMENT FOR IN SITU EVALUATON OF STEEL COMPONENTS CENSING EVALUATION. Functional Modules F1 F8.

OF NUCLEAR SYSTEMS.

n 62 Subject Index 7

Storage Caek Therapeutic Administration NUREG-1536: STANDARD REVIEW PLAN FOR DRY SPENT FUEL NUREG-1492 REGULATORY ANALYSIS ON CRITERIA FOR THE RE-

/

STORAGE SYSTEMS. Fbal R LEASE OF PATIENTS ADMINISTERED RADIOACTIVE

/

NUREG 1571: INFORMATION 0800K ON INDEPENDENT SPENT MATERIALFinal Report.

FUEL STORAGE INSTALLATIONS.

Thermal Aging Strose Corroeion Crackin9 NUREG/CR-6363: EFFECTS OF THERMAL AGING AND NEUTRON 1R-NUREG/CP-0154: PROCEEDINGS OF THE CNRA/CSNI WORKSHOP RADIATION ON THE MECHANICAL PROPERTIES OF THREE-WIRE ON STEAM GENERATOR TUBE INTEGRITY IN NUCLEAR POWER STAINLESS STEEL WELD OVERLAY CMDDING.

PLANTS.

NUREG/CR-4667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN Thermal-Hydraul6c UGHT WATER REACTORS. Senuannual Report. January 1996 June NUREG/CP-0158: PROFEDINGS OF THE OECD/CSNI SPECIALISTS 1996.

MEETING ON BORO,; DILUTION REACTIVITY TRANSIENTS. Held in NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN State College, Pennsylvania, USA. October 18-20,1995.

UGHT WATER REACTORS. Senwannual Report, July-December 1996.

NUREG/CF-0159: PRvvEEDINGS OF THE OECD/CSNI WORKSHOP NUREG/CR4511 V01: STEAM GENERATOR TUBE INTEGRITY ON TRANSIENT THERMAL-HYDRAULIC AND NEUTRONIC CODES PROGRAM. Semiannual Report, August 1995 - March 1996.

REQUIREMENTS. Held in Annapons Maryland, USA, November 54, M6.

Structural Aging u

NUREG/CP-0157 V01: PROCEEDINGS OF THE TWENTY-FOURTH Thermoluminescent Dosimeter WATER REAC~OR SAFETY INFORMATON MEETING. Plenary Ses-NUREG-0837 V16 NO3: NRC TLD DIRECT RADIATION MONITORING sion, High Burnup Fuel, Containment And Structural Aging-NETWORK. Progress Report JugSeptember 1996.

NUREG-0837 V16 NO4: NRC TLU DIRECT RADIATION MONITORING G/

99 RESULTS OF CHARPY V-NOTCH IMPACT TESTING NUR V

iNC LD R CT IA N MONITORING OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 DE' NETWORK. Progress Report. January-March 1997.

GREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMERCIAL RE-NUREG-0837 V17 NO2: NRC TLD DIRECT RADIATION MONITORING ACTOR CAVITY.

NETWORK. Progress Report. April-June 1997, Subelze Specimen Title Ust WUREG/CR4379: AN IMPROVED CORRELATION PROCEDURE FOR NUREG4540 V18 N11: TITLE UST OF DOCUMENTS MADE PUBUCLY SUBSIZE AND FULL SIZE CHARPY IMPACT SPECIMEN DATA.

AVAILABLE. November 1 30,1996.

NUREG-0540 V16 N12: TITLE UST OF DOCUMENTS MADE PUBLICLY SubsMace N AVAILABLE. December 1 31,1996.

NL"tEG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION AND NUREG.0540 V19 N01: TITIE LIST OF DOCUMENTS MADE PUBLICLY SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES-AVAILABLE. January 1 31,1997 NUREG-0MO V19 NO2: TITLE UST OF DOCUMENTS MADE PUBUCLY Suction Strainer AVAILABLE.F 1 28,1997 AVAfLA 4 :

EC UAL

.M c 13,1997.

NUREG-0540 V19 N04: TITLE UST OF DOCUMENTS MADE PUBLICLY Supprmion Pool

^

WUREG/CR4153: A SIMPUFIED MODEL OF DECONTAMINATION BY NUR 0

Ti LICT OF DOCUMENTS MADE PUBUCLY i

BWR STEAM SUPPRESSION POOLS.

AVAILABLE.May1 31,1997.

NUREG 0540 V19 N06: TITLE UST OF DOCUMENTS MADE PUBUCLY SMace Crack NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE' NU E $4 t

T T E UST OF DOCUMENTS MADE PUBUCLY SEARCH PROGRAM ( DIRG) PROGRAM.Pr am Final Report NUREG/CR4452: THE SECOND INTERNAT NAL PIPING INTEGRITY NUR

I LE LIST OF DOCUMENTS MADE PUBLICLY RESEARCH GROUP (IPtRG-2) PROGRAM. Final RW AVAILABLE. August 1-31,1997.

NUREG 0540 V19 N09: TITLE LIST OF DOCUMENTS MADE PUBUCLY Surtsey Test FW NUREG/CR4530 DELIBERATE IGNITION OF HYDROGEN-AIR-STEAM NURE oV TL Li T OF DOCUMENTS MADE PU6UCLY MIXTURES IN CONDENSING S1EAM ENVIRONMENTS.

AVAILABLE. October 1 31,1997.

System 80+ Design NUREG-1462 S01: FINAL SAFETY EVALUATON REPORT RELATED Topical RepM NUREG 0390 V11: TOPICAL REPORT REVIEW STATUS.

TO THE CERTIFICATION OF THE SYSTEM 80+ DESIGN. Docket No.52-002.(Asea Brown BoverbCombustion Engineering)

Transportation NUREG-1608 DAFT FC: CATEGORIZING AND TRANSPORTING LOW TLD SPECIFIC ACTIVITY MATERIALS AND SURFACE CONTAMINATED NUREG-0837 V16 NO3: NRC TLD DIRECT RADIATION MONfTORING OBJECTS. Draft Rept For Comment.

NETWORK. Progress Report. July-September 1996.

NUREG-0W V16 N04: NRC TLD DIRECT RADIATION MONITORING NUR 0837 1

LD l CT I T O'N MONITORING U

1609 DR C: STANDARD REVIEW PLAN FOR TRANSPOR-TATON PACKAGES FOR RADIOACTIVE MATERIALDraft Report For NETWORK.Pr s Report. Ja

-March 1997.

NURE 7 V1 2

T RADIATION MONITORING NU

-5661: RECOMMENDATIONS FOR PREPARING THE CRITb NUREG/CR MEASUREMENT OF RESIDUAL RADCACTIVE SUR.

CAUTY SAFETY EVALUATON OF TRANSPORTATION PACKAGES.

FACE CONTAMINATION BY 2-0 LASER HEATED TLD.

TW TP304 Staineess Steel NUREG-1604: CIRCUMFERENTIAL CRACKING OF STEAM GENERA.

NURE /CR4446 RA RE TOUOHNESS EVALUATIONS OF TP304 NUR G 11 V01: STEAM GENERATOR TUBE INTEGRITY PROGRAM. Semiannual Report, August 1995 March 1996.

Technical Training Center NUREG/C94042 RO1: PERSPECTIVES ON REACTOR SAFETY, T

U EG 0154: PROCEEDINGS OF THE CNRA/CSNI WORKSHOP Tectonic Model ON STEAM GENERATOR TUBE INTEGRITY IN NbCLEAR POWER NUREG/CR4529: VAUDATION OF TECTONIC MODELS FCA AN IN-PLANTS.

TRAPLATE SEISMIC ZONE, CHARLESTON, SOUTH CAROUNA WITH GPS GEODETIC DAT/L UF6 NUREG/CR4481 V01: REVIEW OF MODELS USED FOR DETERMIN-Test Reactor ING CONSEQUENCES OF UF(6) 1ELEASE. Development Of Model NUREG/CR 6506: EMBRITTLEMENT DATA BASE, VERSION 1.

Evaluston Critena.

I Subject index 63 NUREG/CR-6481 V02: REVIEW OF MODELS USED FOR DETERMIN-NUREG4040 V20 N04: UCENSEE CONTRACTOR AND VENDOR IN-ING CONSEQUENCES OF UF(6) RELEASE.Model Evaluation Report.

SPECTION STATUS REPORT. Quarterb/ Report. October December 1996.(White Book)

Uncertainty Analysis NUREG-0040 V21 N01: UCENSEE CONTRACTOR AND VENDOR IN.

NUREG/CR4523 V01: PROBABlUSTIC ACCIDENT CONSEQUENCE SPECTION STATUS REPORT. Quarterfy Report. January-March l

UNCERTAINTY ANALYSIS. Food Chain Uncertainty Assessment. Main 1997.(White Book) i Report.

NUREG-0040 V21 NO2: LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR4523 V02-PROBABlUSTIC ACCIDENT CONSEQUENCE SPECTION STATUS REPORT. Quarterly Report, April-June 1997.(White NCER NTY NALYSIS.

Food Chan Uncertainty U

-0040 V21 NO3: UCENSEE CONTRACTOR AND VENDOR IN-NUREG/CR4526 V01: PROBABluSTIC ACCIDENT CONSEQUENCE SPECTION STATUS REPORT. Quarterty Report.Juty-September UNCERTAINTY ANALYSIS. Uncertainty Assessrnent For Deposited 1997.(Wivte Book)

Material And External Doses. Main Report.

NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE Wall Thinni UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-NUREG/C 4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE POSITED MATERIAL AND EXTERNAL DOSES.Appendaces.

PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING' AND FEEDRING CRACKING AND WALL THINNING.

Underground Disposal Waste Burial NUREG/CR4515: BLT-EC (BREACH, LEACH. AND TRANSPORT-EQUI-UBRIUM CHEMISTRY) DATA INPUT GUIDE.A Computer Model For NUREG-1307 R07: REPORT ON WASTE BURIAL CHARGES. Escalation Simulating Release And Coupled Geochenweal Transport Of Contami-Of Decommisssorung Waste Disposal Costs At Low Level Waste Bunal Facilites.

nants From A Subsurface Disposal Facility.

Water Flow Underwater Weldin9 NUREG/CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES NUREG-1616: FEASIBluTY OF UNDERWATER WELDING OF HIGHLY TRENCH SITE: EXPERIMENT 118.

IRRADIATED IN VESSEL COMPONENTS OF BOILING WATER REACTORSA uterature Review.

Water Hammer NUREG/CR4519: SCREENING REACTOR STEAM / WATER PIPING Unsaturated Zone SYSTEMS FOR WATER HAMMER.

NUREG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION AND SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

Wa NU

- 18 V10: CONTROL OF WATER INFILTRATION INTO Uranium NEAR SURFACE LOWi.EVEL WASTE DISPOSAL UNITS. Final Report NUREG-1532: FINAL TECHNICAL EVALUATION REPORT FOR THE On Field Expenments At A Humid Region Site Beltsville, Maryland.

PROPOSED REVISED RECLAMATION PLAN FOR THE ATLAS COR' PORATION MOAB MILLSource Matenal Ucense No. SUA 917. Docket Wold NUREG/CR4181 R01: s PILOT APPUCATION OF RISK-INFORMED NURE 56 FT:

F DARD REVIEW PLAN FOR IN SITU METHODS TO ESTAELISH INSERVICE INSPECTION PRIORITIES LEACH URANIUM EXTRACTION UCENSE APPLICATIONS.

FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR NUREG/CR4505 V01: THE POTENTIAL FOR CRITCAUTY FOLLOW-POWER STATION.

ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE FACluTIES.Orarvum Blended Wrth Soil.

Wold Overlay NUREG/CR-6528: ENVIRONMENTAL ASSESSMENT PROPOSED U-NUREG/CR4363: EFFECTS OF THERMAL /GING AND NEUTRON IR-CENSE RENEWAL OF NUCLEAR METALS,lNC. CONCORD, MASSA' RADIATION CN THE MECHANICAL PROPERTIES OF THREE WIRE CHUSETTS.

STAINLESS STEEL WELD OVERLAY CLADDING.

User's Guide Westinghouse AP600 NUREG/CR4414: PIPING PENCHMARK PROBLEMS FOR THE WES-NUREG/CR4547: DOSFAC2 USER'S GUIDE.

TINGHOUSE AP600 STANDARDIZED PLANT.

Vadoes Zone NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING NUREG/CR-6437: FLOW AND TRANSPORT AT THE LAS CRUCES TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-OF.

TRENCH SITE: EXPERIMENT llB.

COOLANT ACCIDENT, MAIN STEAM LINE BREAK, AND STEAM GENERATOR TUBE RUPTURE SCENARIOS.

Vendor inspection Yucca Mountain NUREG0040 V20 NO3: LICENSEE CONTRACTOR AND VENDOR IN-NUREG/CR4513 N01: NRC HIGH-LEVEL RADIOACTIVE WASTE MAN-SPECTION STATUS REPORT. Quarterty Repor' aly-September AGEMENT PROGRAM ANNUAL PROGRESS REPORT: FISCAL YEAR 1996.(White Book) 1996.

4 t

.5 s

NRC Originating Organization Index (Staff Reports) l This index lists those NRC organizations that have published staff reports. The index is ar-l 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.

GMISORY COMMITTEE (S)

ADVISORY COMITTEE ON NUCLEAR WASTE EDO OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM (POST 861109)

NUREG-1423 V07: A COMPILATION OF REPORTS OF THE ADVISO-OFFICE OF INFORMATION RESOURCES MANAGEMENT (POST RY COMMITTEE ON NUCLEAR WASTE. July 1996 - June 1997.

890205)

ACRS ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUREG 1125 V18: A COMPILATON OF REPORTS OF THE ADVISO-NUREG 0304 V21 NO3: REGULATORY AND TECHNICAL REPORTS RY COMMITTEE ON REACTOR SAFEGUARDS.1996 Annual.

(ABSTRACT INDEX JOURNAL). Cornpilation For TNrd Quarter 1996, July-September.

OFFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)

NUREG4304 V21 N04: REGULATORY AND TECHNICAL REPORTS REGION 1 (POST 820201)

(ABSTRACT INDEX JOURNAL). Annual Compilation For 1996.

NUREG-0837 V16 NO3: NRC TLD DIRECT RADIATON MONITORING NUREG-0304 V22 NOI: REGULATORY AND TECHNICAL REPORTS NETWORK. Pro 9ress Report. July-September 1996.

(ABSTRACT INDEX JOURNAL). Compilation For First Quarter NUREG-0837 V16 N04: NRC TLD DIRECT RADIATON MONITORING 1997,J

-Mard NUREG-0304 V22 NO2: REGULATORY AND TECHNICA'. REPORTS NURE 1 N 1:N TLD E

TION MONITORING NUR 4837 1 2N T C

D AflON MONITORING LY AVAILABLE.Novernber 1 30,1996.

OFC OF N C E T(

T8 1 0 W8 M W W W EMA M MM NUREG-0940 V15 N2 P1: ENFORCEMENT ACTONS: SIGNIFICANT kY ^V^

ACTIONS RESOLVED INDIVIDUAL ACTONS.Semennual Progress NUREG-0 0V E

O DOCUMENTS MADE PUBUC-LY AVAILABLE. January 1 31,1997.

NUR 0940 P NFORCEMENT ACTONS: SIGNIFICANT

[Y V

B F 28 ACTONS RESOLVED REACTOR UCENSEES. Semiannual Progress Rep 1, July-December 1996.

NUREG4540 V19 NO3: TITLE UST OF DOCUMENTS MADE PUBUC-NURE40940 V15 N2 P3: ENFORCEMENT ACTONS: SIGNIFICANT LY AVAILABLE March 1-31.1997.

ACTONS RESOLVED MATERIAL UCENSEES. Semiannual Progress NUREG-440 V19 N04: TITLE UST OF DOCUMENTS MADE PUBLIC-LY AVAILABLE. April 1-30,1997.

NUR P1 NFORCEMENT ACTONS: SIGNIFICANT LY AVA BLE M y 31 199 ACTONS RESOLVED INDIVIDUAL ACTIONS. Semiannual Progress Report, January 4une 1997.

NUREG 0540 V16 N06: TITLE UST OF DOCUMENTS MADE PUBLIC-NUREG4940 V16 N1 P2: ENFORCEMENT ACTIONS: SIGNIFICANT LY AVAILABLE.JJne 1 3),1997.

ACTIONS RESOLVED REACTOR UCENSEES. Semiannual Progress NUREG 0540 V19 N07: TITLE UST OF DOCUMENTS MADE PUBUC-Report, January-June 1997-LY AVAILABLE July 1 31,1997.

NUREG-0940 V16 N1 P3: ENFORCEMENT ACTIONS: SIGNIFICANT NUREG4540 V1J N08 TITLE UST OF DOCUMENTS MADE PUBUC-ACTIONS RESOLVED MATERIAL UCENSEES.Sermannual Progress LY AVAILAB' E. August 1 31,1997.

Report. January 4une 1997-NUREG 0750 V44101: INDEXES TO NUCLEAR REGULATORY COM-MISSION ISSUANCES. July September 1996.

ECD OFFICE OF ADMINISTRATION (PRE 870413 & POST 890205)

RULES & DIRECTIVES REVIEW BRANCH (POST 920323)

C NUREG-0936 V15 NO2-NRC REGULATORY AGENDA. Semiannual NUREG-0750 V44 N05: NUCLEAR REGULATORY COMMISSION IS.

SUANCES FOR NOVEMBER 1996. Pages 229-314.

OFFICE DM i TON DIRECTOR (POST 940714)

NUREG 0750 V44 N06: NUCLEAFi REGULATORY COMMISSION IS-NUREG 0936 V16 N01: NRC REGULATORY AGENDA. Semiannual ReporLJanuary4une 1997.

NUREG-0 V 5101 N XE NUCL R REGULATORY COM-MISSION ISSUANCES. January-March 1997.

EC3 OFFICE OF THE CONTROLLER (PRE 820418 & POST 890205)

NUREG4750 V45102: INDEXES TO NUCLEAR REGULATORY COM-OFFICE OF THE CONTROLLER (POST 690205)

MISSON ISSUANCES.Januarydune 1997.

NUREGo542 V02 ACCOUNTABluTY REPORT FISCAL YEAR 1996.

NUREG4750 V45 Not: NUCLEAR REGULATORY COMMISSION IS-DMSiON OF BUDGET & ANALYSIS (POST 890205)

SUANCES FOR JANUARY 1997. Pages 1-47.

NUREG-1100 V13: BUDGET ESTIMATES. Fiscal Year 1998.

NUREG-0750 V45 NO2: NUCLEAR REGULATORY COMMISSION IS-NUREG-1350 V09. NUCLEAR REGULATORY COMMISSION INFOR.

SUANCES FOR FEBRUARY 1997. Pages 49-93.

MATON DIGEST.1997 Edition-NUREG-0750 V45 NO3; NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR MARCH 1997.Pages95-263.

EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG-0750 V45 N04: NUCLEAR REGULATORY COMMISSION IS-DATA SUANCES FOR APRfL 1997.Pages 265-353.

OFFICE FOR ANALYSIS & EVALUATON OF OPERATIONAL DATA, DI-NUREG-0750 V45 N05: NUCLEAR REGULATORY COMMISSION IS-RECTOR SUANCES FOR MAY 1997.Pages 355-435.

NUREG-0090 V19-REPORT TO CONGRESS ON ABNORMAL NUREG-0750 V45 N06: NUCLEAR REGULATORY COMMISSION ;S-OCCURRENCES. Fiscal Year 1996.

SUANCES FOR JUNE 1997. Pages 437 495.

DIVIS!ON OF SAFETY PROGRAMS (POST 870413)

NUREG 1145 V13: U.S. NUCLEAR REGULATORY COMMISSION NUREG-1275 V12: OPERATING EXPERIENCE FEEDBACK 1996 ANNUAL REPORT.

REPORT. Assessment Of Spent Fuel Cooling.

NUREG-1603 DRFT:

INDIVIDUAL PLANT EXAMINATON DATABASE. User's Guide.

65

66 NRC Originating Organization index (Staff Reports)

EDO OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NRC - NO DETAILED AFFILIATION GIVEN OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS NUREG-0325 R22: U.S. NUCLEAR REGULATORY COMMISSON OR.

NUREG-0383 V01 R20- DIRECTORY OF CERTIFICATES OF COMPU-GANIZATION CHARTS AND FUNCTIONAL ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of STATEMENTS. November 1997.

NUREG4540 V19 N09: TITLE LIST OF DOCUMENTS MADE PUBLIC-NRC-Approved Packages.

NUREG 0383 V02 R20: DIRECTORY OF CERTIFICATES OF COMPLI.

LY AVAILABLE. September 1-30,1997.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Certificates Of NUREG-0540 V19 N10: TITLE UST OF DOCUMENTS MADE PUBLIC-LY AVAILABLE. October 1 31,1997.

Comphance.

NUREG-0750 V46 N01: NUCLEAR REGULATORY COMMISSION IS-NUREG-0383 V03 R17: DIRECTORY OF CERTIFICATES OF COMPLI.

ANCE FOR RADIOACTIVE MATERIALS PACKAGES. Report Of SUANCES FOR JULY 1997.Pages 120.

NUREG4750 V46 NO2: NUCLEAR REGULATORY COMMISSION IS-NRC Approved Quality Assurance Programs For Radioactive Materi.

SUANCES FOR AUGUST 1997. Pages 21-48.

als Packa a NilP' G-1614 V01: NRC STRATEGIC PLAN. Fiscal Year 1997 - Fiscal NUREG-072 R12: PUBLIC INFORMATION CIRCULAR FOR SHIP-

.r2002.

MENTS OF IRRADIATED REACTOR FUEL NUREG 1536: STANDARD REVIEW PLAN FOR DRY SPENT FUEL EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

STORAGE SYSTEMS. Fird Nport.

OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 941217)

NUREG 1571: INFORMA" A HANDBOOK ON INDEPENDENT NUREG-1603 DRFT:

INDIVIDUAL PLANT EXAMINATION SPENT FUEL STORAGE INSTALLATIONS-DATABASE. User's Guide.

NUREG 1608 DAFT FC: CATEGORIZING AND TRANSPORTING LOW DIVISION OF ENGINEERING TECHNOLOGY (POST 941217)

SPECIFIC ACTIVITY MATERIAL.S AND SURFACE CONTAMINATED NUREG 1616: FEASIBluTY OF UNDERWATER WELDING OF OBJECTS. Draft Rept For Comment.

HIGHLY IRRADIATED IN-VESSEL COMPONENTS OF BOILING NUREG-1609 DRFT FC: STANDARD REVIEW PLAN FOR TRANS-WATER REACTORS.A Uterature Review.

PORTATION PACKAGES FOR RADIOACTIVE MATERIALDraft DIVISION OF REGULATORY APPLICATIONS (POST 941217)

NUREG-0713 V17; OCCUPATIONAL RAD!ATION EXPOSURE AT Report For Comment.

DIVISON OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST COMMERICAL NUCLEAR POWER REACTORS AND OTHER 870729)

IAdTIES.1995. Twenty-Eighth Annual Report.

NUREG 1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY NURE,.M30/

A07:

REPORT ON WASTE BURIAL PROGRAMS AT MEDICAL FACluTIES. Final Report.

CHARGES Escalation Of DC....Ang Waste Disposal Costs At NUREG 1556 V01: CONSOUDATED GUIDANCE ABOUT MATERIALS Low Level Waste Bunal Facihties.

UCENSES. Program-Specific Guidance About Portable Gauge NUREG-1492: REGULATORY ANALYSIS ON CRITERIA FOR THE RE-LEASE OF PATIENTS ADMINISTERED RADIOACTIVE Licenses. Final Report NUREG 1556 V2 DAF FC: CONSOLIDATED GUIDANCE ABOUT MA-M ATERIALFinal Report.

TERIALS UCENSES. Program Specife Guidance About Industnal Ra.

NUREG 1496 V01: FINAL GENERIC ENVIRONMENTAL IMPACT STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL diography Licenses. Draft Report For Use And Comrnunt.

NUREG 1556 V3 DRF FC: CONSOUDATED GUIDANCE ABOUT MA-CRITERIA FOR LICENSE TERMINATON OF NRC-LICENSED NU-TERIALS UCENSES. Applications for Sealed Source And Device CLEAR FACILITIES. Main Repor1. Final Report.

NUREG 1496 V02: FINAL GENERIC ENVIRONMENTAL IMPACT Evaluation And Re0 stration. Oraft Report For Comment i

NUREG-1556 V4 DRF FC: CONSOUDATED GUIDANCE ABOUT MA.

STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION OF NRC-UCENSED NU.

TERIALS UCENSES. Program Spectre Guidance About Fixed Gauge CLEAR FACluTIES. Appendices A And B. Final Report.

Ucenses. Draft Report For Comment.

NUREG 1556 V5 DRF FC: CONSOUDATED GUIDANCE ABOUT MA.

NUREG 1496 V03: FINAL GENERIC ENVIRONMENTAL IMPACT STATEMENT IN SUPPORT OF RULEMAKING ON RADIOLOGICAL TERIALS UCENSES Program Specifc Guidance About Self-Shielded CRITERIA FOR LICENSE TERMINATON OF NRC-UCENSED NU-frradiator Ucenses. Draft Report For Comment.

NUREG-1562 DRF1 FC: STANDARD REVIEW PLAN FOR APPUCA.

CLEAR FACILITIES.Ap ndices C-H. Final Report.

NUREG/CR-4918 V10-TROL OF WATER INFILTRATION INTO TIONS FOR LICENSES TO DISTRIBUTE BYPRODUCT MATERIAL NEAR SURFACE LOW LEVEL WASTE DISPOSAL UNITSFrial TO PERSONS EXEMPT FROM THE REQUIREV WTS FOR AN NRC LICENSE.10CFF3 Parts 3014.30.15,30.16,30.1u,30.19 & 30.20.

Report On Field Exper6ments At A Humid Region DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207)

Site Beltsville, Maryland.

NUREG-1601: CHEMICAL PROCESS SAFETY AT FUEL CYCLE FA.

DIVISION OF SYSTEMS TECHNOLOGY (POST 941217)

NUREG 1545: EVALUATION CRITERIA FOR COMMUNICATIONS-RE-CluTIES LATED CORRECTIVE ACTION PLANS.

OPERATIONS BRANCH NUREG 1602 DRFT FC: THE USE OF PRA IN RISK-INFORMED NUREC4525 V02 ROS: SAFEGUARDS

SUMMARY

EVENT LIST APPLICATIONS. Draft Rept For Comment.

OfVfSION OF WASTE'1990 Through December 31,1996' (SSEL). January 1 NUREG/CR 6391: DETONATION CELL SIZE MEASUREMENTS IN MANAGEMENT (NMSS 940403)

HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT NUREG 1506: FINAL ENVIRONMENTAL IMPACT STATEMENT TO THE BNL HIGH TEMPERATURE COMBUSTION FACILITY.

CONSTRUCT AND OPERATE THE CROWNPOINT URANIUM SO.

NUREG/CR 6525: SECPOP90: SECTOR POPULATION, LAND FRAC-LUTON MINING PROJECT, CROWNPOINT, NEW MEXICO. Docket TON, AND ECONOMIC ESTIMATION PROGRAM.

No. 40-8968.(Hydro Resources, Inc.)

NUREG 1532: FINAL TECHNICAL EVALUATON REPOHT FOR THE EDO OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428)

PROPOSED REVISED RECLAMATION PLAN FOR THE ATLAS OFFICE OF NUCLEAR REACTOR REGULATION (POST 941001)

CORPORATON MOAB MILLSource Matenal Ucense No. SUA-NUREG-0040 V20 NO3: UCENSEE CONTRACTOR AND VENDOR IN-917. Docket No. 40-3453 (Atlas Corporation)

SPECTION STATUS REPORT, Quarterty Report July September NUREG-1569 DRFT: DRAFT STANDARD REVIEW PLAN FOR IN SITU 1996.(White Book)

LEACH URAN!VM EXTRACTON UCENSE APPUCATIONS.

NUREG4040 V20 N04: UCENSEE CONTRACTOR AND VENDOR IN-PERFORMANCE ASSESSMENT & HYDROLOGY BRANCH (NMSS SPECTION STATUS REPORT. Quarterty Report, October December 940403) 1996.(White Book)

NUREG/CR-6505 V01: THE POTENTIAL FOR CRITICALITY FOLLOW-NUREG-0040 V21 N01: UCENSEE CONTRACTOR AND VENDOR IN-ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE SPECTION STATUS REPORT. Quarterly ReportJanuary March FACluTIES. Uranium Blended With Soil.

1997.(White Book)

NUREG-0040 V21 NO2: LICENSEE CONTRACTOR AND VENDOR IN-U.S. NUCLEAR REGULATORY COMMISSION SPECTION STATUS REPORT. Quarte <ty Report April-June OFFICE OF THE GENERAL COUNSEL (POST 860701) 1997.(White Book)

NUREG 0386 008: UN'TED STATES NUCLEAR REGULATORY COM-NUREGK)040 V21 NO3: UCENSEE CONTRACTOR AND VENDOR IN-MISSION STAFF PRACTICE AND PROCEDURE SPECTON STATUS REPORT. Quarterly Report. July-September OlGEST.Comnnaion, Appeal Board And Ucensing Board 1gg7,(wd,te Book)

Decisions. July 1972 June 1996.

NUREG-0J90 V11: TOPICAL REPORT REVIEW STATUS.

OFFICE OF THE INSPECTOR GENERAL (POST 890417)

NUREG-1021 INT R08: OPERATOR UCENSING EXAMINATON NUREG 1415 V10 N01: OFFICE OF THE INSPECTOR STANDARDS FOR POWER REACTORS.

GENERALSemiannual Report To Congress, April 1,1997 - Septem.

NUREG 1462 S01: FINAL SAFETY EVALUATION REPORT RELATED bei 30,1997.

TO THE CERTIFICATION OF THE SYSTEM 80+ DESIGN. Docket OFFICE OF THE SECRETARY OF THE COMMISSION No. 52 002.(Asea Brown Boveri-Combustion Engineeri )

NUREG-1610- CONTROLLING THE ATOM.The Beginnings Of Nuclear NUREG 1503 S01: FINAL SAFETY EVALUATION RE T RELATED f

Regulation, 1946 1962.

TO THE CERTIFICATION OF THE ADVANCED BOiUNG WATER f

i

NRC Originating Organization Index (Staff Reports) 67 REACTOR DESIGN. Supplement No.1. Docket No. 52 001.(General NUREG-1604: CIRCUMFERENTIAL CRACKING OF STEAM GENERA-Electne Nuclear Energy)

TOR TUBES.

NUREG-1545: EVALUATION CRITERIA FOR COMMUNICATIONS.RE-LATED CORRECTIVE ACTION PLANS.

NUREG 1606 DRFT FC: PROPOSED REGULATORY GUIDANCE RE-NUREG-1555 DRFT: ENVIRONMENTAL STANDARD REVIEW LATED TO IMPLEMENTATION OF 10 CFR 50.59 (CHANGES' PLAN Standard Rawew Plans For Environmental Rewsws For Nucle-TESTS, OR EXPERIMENTS). Draft Report For Comment.

NUREG-1607: SAFETY EVALUATION REPORT RELATED TO THE NURE 15 7 ETY EVALUATION REPORT RELATED TO THE DEPARTMENT OF ENERGY'S PROPOSAL FOR THE IRRADIATION RENEWAL OF THE OPERATING UCENSE FOR THE RESEARCH OF LEAD TEST ASSEMBLIES CONTAINING TRITIUM-PRODUCING REACTOR AT NORTH CAROLINA STATE UNIVERSITY.

BURNABLE ABSORBER RODS IN COMMERCIAL LIGHT-WATER NUREG-1574: STANDARD REVIEW PLAN ON ANTITRUST REACTORS.

REVIEWS.Rnal Report NUREG-1611: AGING MANAGEMENT OF NUCLEAR POWER PLANT NUREG-1574 DRFT FC: STANDARD REVIEW PLAN ON ANTITRUST. Draft Report For Cornment.

CONTAINMENTS FOR LICENSE RENEWAL.

NUREG-1577 DRFT FC: STANDARD REVIEW PLAN ON POWER RE' NUREG-1612: STATUS REPORT: REACTOR VESSEL INTEGRITY DA-ACTOR LICENSEE FINANCIAL QUALIFICATIONS AND DECOMMIS-TABASE' SiONING FUNDING ASSURANCE. Draft Report For Comment.

6 l

l 1

a 5

l l

1

N.

/

)'N NRC Originating Organization Index (international Agreements)

This index lists those NRC organizations that have published inte national agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-I fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each l

entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

There were no NUREG/lA reports published this year.

O 69

7

NRC Contract Sponsor index (Contractor Reports)

This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program office) l and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-l tion is followed by the NUREG/CR number and title of the report (s) arepared by that organi-zation. If further information is needed, refer to the main citation by tie NUREG/CR number.

I EDO

• OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DIVISION OF REGULATORY APPLICATIONS (870413 941217)

DATA NUREG/CR-6037: MEASUREMENT OF RESIDUAL RADIOACTIVE OFFICE FOR ANALYSIS & EVALUATON OF OPERATIONAL DATA, Ol-SURFACE CONTAMINATION BY 2.D LASER HEATED TLD.

RECTOR DIVISON OF ENGINEERING TECHNOLOGY (POST 941217)

NUREG/CR4042 R01: PERSPECTIVES ON REACTOR SAFETY.

NUREG/CR-4219 V12 N2: HEAVY-SECTION STEEL TECHNOLOGY DIVISION OF SAFETY PROGRAMS (POST 870413)

PROGRAM, Semiannual Progress Report For April 1995 Through NUREG/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERE September 1995.

CORE DAMAGE ACCIDENTS: 1995. A Status Heport.

NUREG/CR 4219 V13 N1: HEAVY SECTION STEEL TECHNOLOGY NUREG/CR-4674 V24: PRECURSORS TO POTENT 1,*.L st:vD5:

PROGRAM. Semiannual Progress Report For October 1995 March CORE DAMAGE ACCIDENTS: 198243.A Status Report.

1996.

NUREG/CR4456: REVIEW OF INDUSTRY EFFORTS TO MANAGE NUREG/CR-4687 V22: ENVIRONMENTALLY ASSISTED CRACKING PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING, AND FEEDRING CRACKING AND WALL THINNING.

IN LIGHT WATER REACTORS. Senuannual Report, January 1996 -

June 1996.

NUREG/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING F

N AT RI A ETY SA EGU enuannual d,

eceh NUREG/CR-0200 R5V1P1: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR NUREG/CR-5591 V07 N1: HEAVY-SECTION STEEL IRRADIATION LICENSING EVALUATlON. Control Modules C4, C6.

PROGRAM.Senuannual Progress Report For October 1995 Through NUREG/CR 0200 R5V1P2: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR NUR G/

91 V07 N2: HEAVY-SECTION STEEL IRRADIATION LICENSING EVALUATON. Control Modules S1 H1.

PROGRAM.Senuannual Progress Report For April Through Septern-NUREG/CR-0200 RSV2P1: SCALE: A MODULAR CODE SYSTEM FOR PERFORMING STANDARD 12ED COMPUTER ANALYSES FOR NU E / R4181 RO1: A PILOT APPLICATON OF RISK-INFORMED METHODS TO ESTABLISH INSERVICE INSPECTION PRIORITIES NUREG/

2 RSV2 SC A O N

E SYSTEM FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR LICENSING EVALUATON. Functional Modules F9 - F11.

NU EG/ R 3 d2: STABILITY OF CRACKED PIPE UNDER SEIS-NUREG/CR4200 R5V2P3: SCALE: A MODULAR CODE SYSTEM MIC/ DYNAMIC DISPLACEMENT CONTROLLED FOR PERFORMING STANDARDIZED COMPUTER ANALYSES FOR STRESSES Subtask 1.2 Falai Report LICENSING EVALUATON Functional Modules F16 F17 NUREG/CR4233 V03: CRACK STABILITY IN A REPRESENTATIVE NUREG/CR4200 R5V3; SCALE: A MODULAR CODE SY' STEM FOR PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /DY-PERFORMING STANDAROtZED COMPUTER ANALYSES FOR Li-NAMIC DISPLACEMENT-CONTROLLED STRESSES. Subtask 1.3 CENSING EVALUATION' Miscellaneous Final Report NUREG/CR 5661: RECOMMENDATIONS FOR POEPARING THE NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY RE-RITICALITY SAFETY EVALUATON OF 1,4ANSPORTATON 8E C P RAM i RG PR RAM mF R WR NUREG/CR4361: CRITICALITY BENCHMARK GUIDE FOR LIGHT.

IRRADIATION ON THE MECHANICAL PROPERTIES OF THREE-WATER REACTOR FUEL IN TRANSPORTATION AND STORAGE WIRE STAINLESS STEEL WELD OVERLAY CLADDING.

PACKAGES'NDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST NUREG/CR4370: BLOCKAGE 2.5 USER'S MANUAL DIVI OF l U

/

hjED REFER E

FOR BABILISTIC NUREG/CR4074 V03: SEALED SOURCE AND DEVICE DESIGN SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND SAFETY TESTING. Technical Report On The Findings Of Task 4 in.

USE OF EXPERTS. Main Report.

vestigation Of A Failed Brachytherapy Needle Applicator NUREG/CR4372 V02: RECOMMENDATIONS FOR PROBABILISTIC NUREG/CR4528: ENVIRONMENTAL ASSESSMENT PR'OPOSED Ll-SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND EA EWAL OF NUCLEAR METALS,1NC. CONCORD, MAS.

NI EG CR43 9 AN D CORRELATION PROCEDURE FOR DIVISION OF FUEL CYCLE SAFETY & SAFEGUARDS (POST 930207)

SUBSIZE AND FULL-SIZE CHARPY IMPACT SPECIMEN DATA.

NUREG/CR4481 V01: REVIEW OF MODELS USED FOR DETERMIN.

NUREG/CR4389: IPIRG-2 TASK 1 PIPE SYSTEM EXPERIMENTS ING CONSEQUENCES OF UF(6) RELEASE. Development Of Model WITH CIRCUMFERLNTIAL CRACKS IN STRAIGHT-PIPE Evaluabon Crtteria.

LOCATIONS. Final Report. September 1991 November 1995.

NUREG/CR4481 V02: REVIEW OF MODELS USED FOR DETERMIN.

NUREG/CR4399: RESULTS OF CHARPY V-NOTCH IMPACT TEST.

ING CONSEQUENCES OF UF(6) RELEASE.Modd Evaluation ING OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 Report DEGREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMER-OlVISION OF WASTE MANAGEMENT (NMSS 940403)

CIAL REACTOR CAVITY.

NUREG/CR4505 YO1: THE POTENTIAL FOR CRITICALITY FOLLOW.

NUREG/CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE FIED A 302 GRADE B PLATE MATERIALS DATA ANALYSIS.

FACILITIES.Uraruum Blended With Soil.

NUREG/CR4426 V02 DUCTILE FRACTURE TOUGHNESS OF MODI-NUREG/CR4513 NOI: NRC HIGH-LEVEL RADIOACTIVE WASTE FIED A 302 GRADE B PLATE MATERIALS. Data Records.

MANAGEMENT PROGRAM ANNUAL PROGRESS REPORT:

NUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPl.

FISCAL YEAR 1996.

CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCT-DENT CONDITONS.

EDO OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

NUREG/CR4448. FRACTURE TOUGHNESS EVALUATONS OF OFFICE OF NUCLEAR REGULATORY RESEARCH (860720-941217)

TP304 STAINLESS STEEL PIPES MUREG/CR4530- DEUBERATE IGNITION OF HYDROGEN AIR.

NU71EG/CR4448 V02: EVALUATON OF NATIONAL SEISMOGRAPH STEAM MIXTURES IN CONDENS NG STEAM ENVIRONMENTS.

NETWORK DETECTION CAPAS!LITIES Final Report.

71

72 NRC Contract Sponsor index NUREG/CR4452: THE SECOND INTERNATIONAL PIPING INTEGRI-Model For Sanulating Release And Coupled Geochemical TranspM TY RESEARCH GROUP (IPIRG-2) PROGRAM Fined Report Of Contarmnants From A Subsurface Disposal Facility.

NUREG/CR4454. POOL CRITICAL ASSEMBLY PRESSURE VESSEL NUREG/CR4531: EFFECTS OF RADIOACTIVE HOT PARTICLES ON FACluTY BENCHMARK.

PIG SKIN.

NUREG/CR4464: AN EVALUATON OF METHODOLOGY FOR SEIS-NUREG/CR 6535: DEVELOPMENT OF CONFORMAL RESPIRATOR MIC QUAUFICATON OF EQUIPMENT. CABLE TRAYS, AND DUCTS MONITORING TECHNOLOGY.

IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

NUREG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION NUREG/CR4478: MOTOROPERATED VALVE (MOV) ACTUATOR AND SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

NUREG/CR4566: DESCRIPTON OF MULTIMEDIA ENVIRONMEN-MOTOR AND GEARBOX TESTING.

NUREG/CR4486: ASSESSMENT OF MODULAR CONSTRUCTION TAL POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 FOR SAFETY-RELATED STRUCTURES AT ADVANCED NUCLEAR MODIFICATION FOR THE NUCLEAR REGULATORY COMMISSION.

NUREG/CR4581: CONSIDERATIONS IN THE APPUCATION OF THE POWER PLANTS NUREG/CR4506. EMBRITTLEMENT DATA BASE, VERSION 1.

ELECTRONIC DOSIMETER TO DOSE OF RECORD.

0I 6

M$ C ST 41 NUREG/CR4508: COMPONENT UNAVAILABluTY VERSUS INSERV-

-6 AS E

MM ICE TEST (IST) INTERVAL: EVALUATIONS OF COMPONENT NUR V1 ST M ENERA OR BE EGRITY P R s

A PROGRAM.Sernannual Report, August 1995 March 1996.

NUREG/CR4295: REASSESSMENT OF SELECTED FACTORS AF-NUREG/CR4529: VAUDATION OF TECTONIC MODELS FOR AN IN' FECTING SITING OF NUCLEAR POWER PLANTS TRAPLATE SEISMIC ZONE, CHARLESTON, SOUTH CAROLINA NUREG/CR4391: DETONATION CELL SIZE MEASUREMENTS IN WITH GPS GEODETIC DATA-HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT NUREG/CR4538: EVALUATON OF LOCA WITH DELAYED LOOP THE BNL HIGH-TEMPERATURE CCMBUSTION FACluTY.

AND LOOP WITH DELAYED LOCA ACCIDENT SCENARIOS.

NUREG/CR4463 RO1: REVIEW GUIDEUNES FOR SOFTWARE LAN-NUREG/CR4539: EFFECTS OF FLUORIDE AND OTHER HALOGEN GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY IONS ON THE EXTERNAL STRESS CORROSION CRACKING OF SYSTEMS. Final Report.

WPE 304 AUSTENITIC STAINLESS STEEL NUREG/CR4469: EXPERIMENTS TO INVESTIGATE DIRECT CON-NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF INSTRUMENT FOR IN SITU EVALUATION OF STEEL COMPO-THE CALVERT CLIFFS NUCLEAR POWER PLANT.

NENTS OF NUCLEAR SYSTEMS.

NUREG/CR4474: PRELIMINARY PHENOMENA IDENTIFICATION NUREG/CR4563: LG EXCITATON, ATTENUATON, AND SOURCE AND RANKING TABLES (PIRT) FOR SBWR STARTUP STABluTY.

SPECTRAL SCALING IN CENTRAL AND EASTERN NORTH AMER-NUREG/CR4507: CRITICAL HEAT FLUX (CHF) PHENOMENON ON ICA.

A DOWNWARD FACING CURVED SURFACE.

NUREG/CR4586. HORIZONTAL VELOCITIES IN THE CENTRAL AND NUREG/CR4519: SCREENING REACTOR STEAM / WATER PIPING EASTERN UNITED STATES FROM GPS SURVEYS DURING THE SYSTEMS FOR WATER HAMMER.

19871996 INTERVAL NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE DIVISION OF REGULATORY APPLICATIONS (POST 941217)

UNCERTAINTY ANALYSIS. Food Chain Uncertainty NUREG/CR-4012 V04: REPLACEMENT ENERGY COSTS FOR NU-Assessment. Main Report.

CLEAR ELECTRICITY-GENERATING UNITS IN THE UNITED NUREG/CR4523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE STATES: 1997-2001.

UNCERTAINTY ANALYSIS.

Food Chain Uncertainty NUREG/CR-4409 V06: DA A BASE ON DOSE REDUCTON Assessment Apperdces.

PROJECTS FOR NUCLEAH POWER PLANTS.

NUREG/CR 4525: SECPOP90 SECTOR POPULATION, LAND FRAC.

NUREG/CR-4918 V10- CONTROL OF WATER INFILTRATION INTO TION, AND ECONOMIC ESTlHATION PROGRAM.

NEAR SURFACE LOW-LEVEL WASTE DISPOSAL UNITS. Final NUREG/CR4526 V01: PROBABILISTIC ACCIDENT CONSEQUENCE Report On Field Expenments At A Humed Regson UNCERTAINTY ANALYSIS, Uncertainty Assessment For Deposited S4e.Beltsville. Maryland.

Matenal And External Doses. Main Report.

NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW-NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR FlSCAL YEAn 1996. Annual Report.

DEPOSITED MATERIAL AND EXTERNAL DOSES. Appendices.

NUREG/CR4397: RADIATON SAFETY CONCERNS FOR PREG.

NUREG/CR4527: FINAL RESULTS OF THE XR21 BWR METALUC NANT OR BREAST-FEEDING PATIENTSThe Positions Of The MELT RELOCATION EXPERIMENT.

NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0: A COMPUTER NCRP And The ICRP.

NUREG/CR4404: AN EXPERIMENTAL SCALE-MODEL STUDY OF CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

NUREG/CR4534 V01: FRAPCON 3: MODIFICATIONS TO FUEL ROD SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINT.

MATERIAL PROPERTIES AND PERFORMANCE MODELS FOR ED ROCK MASS.

AP T

,8 R NUREG CR443 : FL RANSPORT AT THE LAS CRUCES H

p9 E MENA IDENTIFICATION AND RANK-NUREG/CR4451: A SAFETY AND' REGULATORY ASSESSMENT OF ING TABLES FOR WESTINGHOUSE AP600 SMALL BREAK LOSS-GENERIC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR OF COOLANT ACCIDENT, MAIN STEAM LINE BREAK, AND STEAM GENERATOR TUBE RUPTURE SCENAROS.

POWER PLANTS' FIELD STUDIES AT THE APACHE LEAP RE-NUREG/CR-6543: EFFECTS OF SMQKE ON FUNCTIONAL CIR-NUREG/CR4459:

SITE IN SUPPORT OF ALTERNATIVE CONCEPTUAL NU EG/CR4547: DOSFAC2 USER'S GUIDE.

NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINIS-EDO - OFFICE OF NUCLEAR REACTOR REGULATION (POST 800428)

TRATON OF RADIOLABELED ANTIBODIES CONTAINING Y-90,TL,-

OFFICE OF NUCLEAR REACTOR REGULATION (r 1ST 941001) 99M,1-131, AND LU-177.

NUREG/CR4331 R01: ATMOSPHERIC RELMNE CONCENTRA-NUREG/CR4497: DATA COLLECTON AND FIELD EXPERIMENTS TIONS IN BUILDING WAKES.

AT THE APACHE LEAP RESEARCH SITE.May 1995 -1996.

NUREG/CR4393: INTEGRATED SYSTEM VALIDATION: METHODOL.

NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITICAUTY SLIDE OGY AND REVIEW CRITERIA.

RULE.Technscal Basis.

NUREG/CR4400: HUMAN FACTORS ENGINEERING (HFE) IN-NUREG/CR4514: ANALYSIS OF POTENTIAL SELF GUARANTEE SIGHTS FOR ADVANCED REACTORS BASED UPON OPERATING TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-EXPERIENCE.

PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY NUREG/CR4414: PIPING BENCHMARK PROBLEMS FOR THE WES-BUSINESS FIRMS THAT 00 NOT ISSUE BONOS.

TINGHOUSE AP600 STANDARDIZED PLANT, NUREG/CR4515 BLT-EC (BREACH, LEACH, AND TRANSPORT.

NUREG/CR4558: NRC ANTITRUST LICENSING ACTONS,1978 EQUILIBRIUM CHEMISTRY) DATA INPUT GUIDE.A Computer 1996.

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 l

titles of their reports. If further information is needed, refer to the main citation by the l

NUREG/CR number.

AEA TECHNOLOGY BROOKHAVEN NATIONAL LABORATORY NUREG/CR4526 V01: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG 1603 DRFT; INDIVIDUAL PLANT EXAMINATION UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposted Matenal And External Doses Main Report.

DATABASE. User's Guide.

NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CP-0157 V01: PROCEEDINGS OF THE TWENTY-FOURTH UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-WATER REACTOR SAFETY INFORMATION MEETING. Plenary Ses-POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

saon, High Burnup Fuel, Containrnent Anti Structural A ing.

Q NUREG/CP0157 V02: PROCEEDINGS OF THE TWENTY-FOURTH ARGONNE NATIONAL LABORATORY WATER REACTOR SAFETY INFORMATION MEETING. Reactor Pres.

NUREG/CP.0154: PROCEEDINGS OF THE CNRA/CSNI WORKSHOP sure Vessel Embnttlement And Thermal Annealing, Reactor Vessel ON STEAM GENERATOR TUBE INTEGRITY IN NUCLEAR POWER Lower Head interty And Evaluation And Projection of Steam Genera-PLANTS.

tor tube....

NUREG/CR-4012 V04. REPLACEMENT ENERGY COSTS FOR NUCLE-NUREG/CP-0157 V03: PROCEEDINGS OF THE TWENTY #OURTH AR ELECTR! CITY-GENERATING UNITS IN THE UNITED STATES:

WATER REACTOR SAFETY INFORMATON MEETINGPRA And HRA, Nt G H 667 V22: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Serniannual Report January 1996 - June NUREG/CP-0161: TRANSACTIONS OF THE TWENTY #lFTH WATER 1

G/CR-4667 V23: ENVIRONMENTALLY ASSISTED CRACKING IN REACTOR SAFETY INFORMATON MEETING.

LIGHT WATER REACTORS. Semiannual Report July-December 1996.

NUREG/CR-4409 V06: DATA BASE ON DOSE REDUCTION PROJECTS NOREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW.

FOR NUCLEAR POWER PLANTS.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW.

FISCAL YEAR 1996. Annual Report.

LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUREG/CR 4511 V01: STEAM GENERATOR TUBE INTEGRITY FISCAL YEAR 1996. Annual Report.

PROGRAM. Semiannual Report, August 1995 - March 1996.

NUREG/CR-6295: REASSESSMENT OF SELECTED FACTORS AF.

FECTING SITING OF NUCLEAR POWER PLANTS.

NU

/

37 F A

TRANSPORT AT THE LAS CRUCES EG/CR4390 DETONATON M SCE MEASMENTS IN HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE NL G R

1 AT THE APACHE LEAP RESEARCH BNL HIGH TEMPERATURE COMBUSTION FACILITY.

SITE IN SUPPORT OF ALTERNATIVE CONCEP"JAL MODELS.

NUREG/CR4393. INTEGRATED SYSTEM VALIDATION: METHODOLO-NUREG/CR4497: DATA COLLECTION AND FIELD EXPERIMENTS AT GY AND REVIEW CRITERIA.

THE APACHE LEAP RESEARCH SITE.May 1995 1996.

NUREG/CR4397: RADIATION SAFETY CONCERNS FOR PREGNANT CATTELLE MEMORIAL INSTITUTE, COLUMBUS LABORATORIES OR BREAST-FEEDING PATIENTS.The Positions Of The NCRP And NURE'3/CR.4657 V23 ENVIRONMENTALLY ASSISTED CRACKING IN The ICRP.

NUREG/CR4400- HUMAN FACTORS ENGINEERING (HFE) INSIGHTS NURE /

V 2-l'L O

AC PI E DE R AWANCED MANS BASED N MRAM EW%

C DYNAMIC DISPLACEMENT-CONTROLLED SIM9SES. Subtask NUREG CR4414. PIPING BENCHMARK PROBLEMS FOR THE WES-NUREG/CR4 V03 CRACK STABILITY IN A REPRESENTATIVE TINGHOUSE AP600 STANDARDIZED PLANT.

PIPING SYSTEM UNDER COMBINED INERTIAL AND SEISMIC /Dy.

NUREG/CR-6451: A SAFETY AND REGULATORY ASSESSMENT OF NAMIC DISPLACEMENT CONTROLLED STRESSES. Subtask 1.3 Final GENERC BWR AND PWR PERMANENTLY SHUTDOWN NUCLEAR Report.

POWER PLANTS.

NUREG/CR4233 V04: INTERNATIONAL PIPING INTEGRITY R E-NUREG/CR-6464: AN EVALUATON OF METHODOLOGY FOR SELS-SEARCH PROGRAM OPtRG) PROGRAM. Program Final Report.

MIC QUALIFICATION OF EOulPMENTCABLE TRAYS, AND DUCTS NUREG/CR4389: IPIRG-2 TASK 1 - PIPE SYSTEM EXPERIMENTS IN ALWR PLANTS BY USE OF EXPERIENCE DATA.

WITH CIRCUMFERENTIAL CRACKS IN STRAIGHT PLPE

~

NUREG/CR4474: PRELIMINARY PHENOMENA IDENTIFICATION AND NUR /CR 644 R

ESE 10 S OF TP304 NUREG/CR4486: ASSESSMENT OF MODULAR CONSTRUCTION FOR Nt R / 44.

ND INTERNATIONAL PIPING INTEGRITY SAFETY-RELATED STRUCTURES AT ADVANCED NUCLEAR RESEARCH GROUP (IPIRG-2) PROGRAM Final Report.

WER P S

EATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NATIONAL TON OF RADIOLABELED ANTIBODIES CONTAINING Y-90,TO-99M,l-LABORATORY 131, AND LU-177.

NUREG/CR4181 R01: A PILOT APPLICATION OF RISK-INFORMED METHODS TO ESTABLISH INSERVICE INSPECTON PRIORITIES NUREG/CR4515: BLT-EC (BREACH, LEACH, AND TRANSPORT.EOUI.

LLBRtVM CHEMISTRY) DATA INPUT GUOE.A Computer Model For FOR NUCLEAR COMPONENTS AT SURRY UNIT 1 NUCLEAR POWER STATION.

Simulating Release And Coupled Geochenweal Transport Of Contam6-NUREG/CR4331 ROI: ATMOSPHERIC RELATIVE CONCENTRATIONS nants From A Subsurface Disposal Facility.

IN BUILDialG WAKES.

NUREG/CR453 t: EFFECTS OF RADIOACTIVE HOT PARTICLES ON NUREG/CR4534 VOI: FRAPCON-3: MODIFICATIONS TO FUEL ROD pig gggy MATERIAL OPER S AND PERFORMANCE MODELS FOR HIGH-NUREG/CR4538: EVALUATION OF LOCA WITH DELAYED LOOP AND.

LOOP WITH DELAYED LOCA ACCIDENT SCENAROS.

NUREG/CR4565: UNCERTAINTY ANALYSES OF INFILTRATION AND SUBSURFACE FLOW AND TRANSPORT FOR SDMP SITES.

CALIFORNIA, UNIV. OF, LOS ANGELES, CA NUREG/CR4566: DESCRIPTON OF MULTIMEDIA ENVIRONMENTAL NUREG/CR-4918 VIO: CONTROL OF WATER INFILTRATION INTO POLLUTANT ASSESSMENT SYSTEM (MEPAS) VERSION 3.2 MODI-FICATION FOR THE NUCLEAR REGULATORY COMMISSON.

NEAR SURFACE LOW-LEVEL WASTE DISPOSAL UNITS. Final Report On Field Expenments At A Humid Region Site Beltsville, Maryland.

73

74 Contractor index CENTER FOR NUCLEAR WASTE REGULATORY ANALYSES NUREG/CR4523 V02: PROBABlUSTIC ACCIDENT CONSEQUENCE NUREG/CR4404: AN EXPERIMENTAL SCALE MODEL STUDY OF UNCERTAINTY ANALYSIS.

Food Chain Uncertanty SEISMIC RESPONSE OF AN UNDERGROUND OPENING IN JOINTED AssessrnentAppendees.

ROCK MASS.

NUREG/CR4541 R02: PHENOMENA IDENTIFICATION AND RANKING NUREG/CR45t3 N01: NRC HIGH-LEVEL RADIOACTIVE WASTE MAN-TABLES FOR WESTINGHOUSE AP600 SMALL DREAK LOSS-OF-AGEMENT PROGRAM ANNUAL PROGRESS REPORT: FISCAL YEAR COOLANT ACCIDENT, MAIN STEAM UNE BREAK, AND STEAM 1996.

GENERATOR TUBE RUPTURE SCENARIOS.

COMMERCE. DEPT. OF, NATIONAL OCEANIC & ATMOSPHERIC ILLINOIS, STATE OF NUREG-1516: MANAGEMENT OF RADIOACTIVE MATERIAL SAFETY NU 458 HORIZONTAL VELOCITIES IN THE CENTRAL AND PROGRAMS AT MEDICAL FACluTIESfinal Report EASTERN UNITED STATES FROM GPS SURVEYS DURING THE 19871996 INTERVAL lOWA STATE UNIV., AMES,IA NUREG/CR4557: DEVELOPMENT OF THE MAGNESCOPE AS AN IN-DELFT UNIVERSITY OF TECHNOLOGY NUREG/CR4523 V01: PROBABlUSTIC ACCIDENT CONSEQUENCE STRUMENT FOR IN SITU EVALUATION OF STEEL COMPONENTS UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessrnentMain OF NUCLEAR SYSTEMS.

NUR CR-6523 V02 PROBABluSTIC ACCIDENT CONSEQUENCE KEITHLEY INSTRUMENTS, INC.

UNCERTAINTY ANALYSIS.

Food Chain Uncertainty NUREG/CR4037: MEASUREMENT OF RESIDUAL RADIOACTIVE SUR-AssessrnentAppendces.

FACE CONTAMINATION BY 2 D LASER HEATED TLD.

ECODYNAMICS RESEARCH ASSOCIATES,INC, LAWRENCE LIVERMORE NATIONAL LA6 ORATORY NUREG/CR4515: BLT EC (BREACH, LEACH, AND TRANSPORT EQUl-NUREG/CR4372 V01: RECOMMENDATIONS FOR PROBABluSTIC LIBRIUM CHEMISTRY) DATA INPUT GUIDE.A Cornputer Model For SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND Simulating Release And Coupled Geochemical Transport Of Conu USE OF EXPERTS. Main Report nants From A Subsurface Disposal FaciMy.

NUREG/CR4372 V02: RECOMMENDATIONS FOR PROBA81USTIC SEISMIC HAZARD ANALYSIS: GUIDANCE ON UNCERTAINTY AND EQE ENGINEERING CONSULTANTS (FORMERLY EQE ENGINEERING, USE OF EXPERTS.Appendees INC.)

NUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPI-CAL REACT NTAINMENTS SUBJECTED TO SEVERE ACCI-UREG/

E ECTS O IOACTIVE HOT PARTICLES ON PIG SKIN.

EVANSVILLE. UNIV. OF, EVANSVILLE, IN NUREG/CR4530 DEUBERATE IGNITION OF HYDROGEN-AIR STEAM MARYLAND, UNIV. OF, COLLEGE PARK. MD MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

NUREG/CR-4918 V10: CONTROL OF WATER INFILTRATION INTO NEAR SURFACE LOW LEVEL WASTE DISPOSAL UNITS, Final Report NUREG C 0155: PROCEEDINGS OF THE SEMINAR ON LEAK BEFORE BREAK IN REACTOR PIPING AND VESSELS.

MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAM 6 RIDGE, MA NUREG/CR4519: SCREENING REACTOR STEAM / WATER PIPING FRANCE SYSTEMS FOR WATER HAMMER.

NUREG/CP-0155: PROCEEDINGS OF THE SEMINAR ON LEAK BEFORE BREAK IN REACTOR PIPING AND VESSELS.

MINNESOTA, UNIV, OF, MINNEAPOLIS, MN NUREG/CR4493: DOSES TO THE HAND DURING THE ADMINISTRA-GRAM, INC.

TlON OF RADIOLABELED ANTIDODIES CONTAINING Y 90,TC-99Ml-NUREG/CR4525: SECPOP90 SECTOR POPULATION, LAND FRAC.

TION, AND ECONOMIC ESTIMATION PROGRAM.

131 AND LU-177.

HARVARD SCHOOL OF PU8UC HEALTH, BOSTON, MA NETHERLANDS,GOYT.OF NUREG/CP4153: PROCEEDINGS OF THE 24TH DOE /NRC NUCLEAR NUREG/CR4526 VO1: PROBA81USTIC ACCIDENT CONSEQUENCE I

AIR CLEANING AND TREATMENT CONFERENCE. Held in Portland, UNCERTAINTY ANALYSIS. Uncertainty Assessment For Deposited Oregon, July 15-18,1996.

Material And Extemal Doses. Main Report.

NUREG/CR4526 V02-PROBABluSTIC ACCIDENT CONSEOVENCE RTM WNS. WEMN ASSEWW M DE-NU EG 523 PRO 8ABluSTIC ACCIDENT CONSEQUENCE ED WER M EMM MSWes.

UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessmentMain NEW MEXICO STATE UNIV., LAS CRUCES, NM NUR CR4523 V02: PROBABlUSTIC ACCIDENT CONSEQUENCE NUREG/CR4437: FLOW AND TRANSPORT AT THE LAS CRUCES UNCERTAINTY ANALYSIS.

Food Chain Uncertainty TRENCH SITE: EXPERIMENT 118.

AssessmentAppendices.

NUREG/CR4526 V01: PROBABILISTIC OCCIDENT CONSEQUENCE NEW MEXICO, UNIV. OF, ALBUOUEROUE, NM UNCERTAINTY ANALYSIS. Uncertainty Assessment For Depoested NUREG/CR-6042 RO1: PERSPECTIVES ON REACTOR SAFETY.

Matenal And External Doses. Main Report NUREG/CR4526 V02: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS UNCERTAINTY ASSESSMENT FOR DE.

UNCERTAINTY ANALYSIS.Focd Chain Uncertainty Assesstnent. Main POSITED MATERIAL AND EXTERNAL DOSES. Appendices.

Report NUREG/CR 6523 V02: PROBABluSTIC ACCIDENT CONSEQUENCE ICF, INC.

UNCERTAINTY ANALYSIS.

Food Chain Uncertainty NUREG/CR4514: ANALYSIS OF POTENTIAL SELF-GUARANTEE AssessmentAppendices.

TESTS FOR DEMONSTRATING FINANCIAL ASSURANCE BY NON-PROFIT COLLEGES, UNIVERSITIES, AND HOSPITALS AND BY BUSI-NORTH CAROUNA, STATE OF NESS FIRMS THAT 00 NOT ISSUE BONDS.

NUREG 1556 V01: CONSOLIDATED GUIDANCE ABOUT MATERIALS UCENSESRogra@c M nce M Pm Gaup IDAHO NATIONAL ENGINEERING & ENYlRONMENTAL LA80RATORY Licenses. Final Report NUREG/CR 5229 V09-F' ELD LYSIMETER INVESTIGATIONS: LOW-LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR NUCLEAR POWER ENGINEERING COHP.

NUREG/CR439t: DETONATION CELL SIZE MEASUREMENTS IN NUR /C 4456 EV DUSTRY EFFORTS TO MANAGE HIGH-TEMPERATURE HYDROGEN-AIR-STEAM MIXTURES AT THE PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING, BNL HIGH-TEMPERATURE COMBUSTION FACIUTY.

AND FEEDRING CRACKING AND WALL THINNING.

NUREG/CR4478: MOTOR-OPERATED VALVE (MOV) ACTUATOR OAK RIDGE NATIONAL LABORATORY MOTOR AND GEARBOX TESTING.

NUREG/CR4523 V01: PROBABluSTIC ACCIDENT CONSEQUENCE NUREG/CR-0200 R5 VIP 1: SCALE: A MODULAR CODE SYSTEM FOR UNCERTAINTY ANALYSIS. Food Chain Uncertainty AssessmentMain PERFORNilNG STANDARDIZED COMPUTER ANALYSES FOR U-Report CENSING EVALUATION. Control Modules C4, C6.

-~

Contractor Index 75 NUREG/CR4200 R$V1P2 SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4042 Rot: PERSPECTIVES ON REACTOR SAFETY.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-CENSING EVALUATON. Control Modules S1 H1.

NUREG/CR4153; A SIMPLIFIED MODEL OF DECONTAMINATION BY BWR STEAM SUPPRESSION POOLS.

NUREG/CR 0200 RSV2P1: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4167: LATE-PHASE MELT PROGRESSION EXPERIMENT PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-CENSING EVALUATION. Functional Modules F1 F8.

MP-2.Results And Analysis.

NUREG/CR 0200 R5V2P2-SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4433 CONTAINMENT PERFORMANCE OF PROTOTYPl.

l PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-CAL REACTOR CONTAINMENTS SUBJECTED TO GEVERE ACCl-CENSING EVALUATION. Functional Modules F9 - F11.

DENT CONDITONS.

NUREG/CR4200 R5V2P3: SCALE: A MODULAR CODE SYSTEM FOR NUREG/CR4469-EXPERIMENTS TO INVESTIGATE DIRECT CON.

PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-TAINMENT HEATING PHENOMENA WITH SCALED MODELS OF THE CMVERT CUFFS NUCLEAR POWER PN NU 02 R5 d

SYSTEM FOR NUREG/CR4523 V01: PROBASILISTIC ACCIDENT CONSEQUENCE PERFORMING STANDARDIZED COMPUTER ANALYSES FOR U-UNCERTAINTY ANALYSIS. Food Chain Uncertainty Assessment. Main bYCTON STEEL TECHNOLOGY NUR R4523 V02: PROBA81USTIC ACCOENT CONSEQUENCE NL E/

19 V12 N2-PROGRAM.Sermannual Progress Report For April 1995 Through Sep.

UNCERTAINTY ANALYSIS.

Food Chain Uncertanty tomber 1995.

AssessmentAppendees-NUREG/CR-4219 V13 N1: HEAVYSECTION STEEL TECHNOLOGY NUREG/CR4525: SECPOP90 SECTOR POPULATION, LAND FRAC-PROGRAM. Semiannual Progress Report For October 1995 March TON, AND ECONOMIC ESTIMATON PROGRAM.

1996.

NUREG/CR4526 V01: PROBA81USTIC ACCIDENT CONSEQUENCE NUREG/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1995. A Status Report.

UNCERTAINTY ANALYSIS. Uncertanty Assessment For Deposited Material And Extemal Doses. Main Report.

NUREG/CR-5229 V09: FIELD LYSIMETER INVESTIGATIONS: LOW-NUREG/CR4526 V02: PROBABILISTIC ACCIDENT CONSEQUENCE LEVEL WASTE DATA BASE DEVELOPMENT PROGRAM FOR

~

UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-NUl C

1 07 H VY-SECTON STEEL IRRADlATON PROGRAM.Sermannual Progress Report For October 1995 Ttrough ME R X R E NUR /CR 5591 V07 N2-HEAVY SEC~ON STEEL IRRADIATON NUREG/CR4530- DELIBERATE IGNITION OF HYDROGEN-AIR-STEAM PROGRAM.Sermannual Progress Report For April Through September MIXTURES IN CONDENSING STEAM ENVIRONMENTS.

1996 NUREG/CR4533: CODE MANUAL FOR CONTAIN 2.0 A COMPUTER NUREG/CR-5661: RECOMMENDATIONS FOR PREPARING THE CRITI.

CODE FOR NUCLEAR REACTOR CONTAINMENT ANALYSIS.

CAUTY SAFETY EVALUATON OF TRANSPORTATION PACKAGES.

NUREG/CR4543: EFFECTS OF SMOKE ON FUNCTONAL CIRCUITS.

NUREG/CR 6042 ROI: PERSPECTIVES ON REACTOR SAFETY.

NUREG/CR4547: DOSFAC2 USER'S GUIDE.

NUREG/CR4361: CRITICALITY BENCHMARK GUIDE FOR UGHT-WATER-REACTOR FUEL IN TRANSPORTATION AND STORAGE SARGENT & LUNDY,INC.

PACKAGES.

NUREG/CR4363: EFFECTS OF THERMAL AGING AND NEUTRON IR-NUREG/CR4433: CONTAINMENT PERFORMANCE OF PROTOTYPI-RADIATION ON THE MECHANICAL PROPERTIES OF THREE-WIRE CAL REACTOR CONTAINMENTS SUBJECTED TO SEVERE ACCl-STAINLESS STEEL WELD OVERLAY CLADDING.

DENT CONDITONS.

NUREG/CR4379: AN IMPROVED CORRELATON PROCEDURE FOR SCIENCE & ENGINEERING ASSOCIATES,INC, I

43 SLS F V

C M AC T STING NUREG/CR4370: BLOCKAGE 2.5 USER'S MANUAL OF STRUCTURAL STEEL SPECIMENS IRRADIATED AT 30 DE.

NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL GREES C TO 1 X 10(16) NEUTRONS / CM(2) IN A COMMERCIAL RE-SCIENCE APPUCATIONS INTERNATIONAL CORP. (FORMERLY NUR G CR4426 V01: DUCTILE FRACTURE TOUGHNESS OF MODI-SCIENCE APPLICATIONS, FIED A 302 GRADE 8 PLATE MATERIALS. DATA ANALYSIS.

NUREG-0713 V17: OCCUPATIONAL RADIATION EXPOSURE AT COM-NUREG/CR4426 V02: DUCTILE FRACTURE TOUGHNESS OF MODI-MERICAL NUCLEAR POWER REACTORS AND OTHER FIED A 302 GRADE B PLATE MATERIALS. Data Records.

FACluTIES.1995. Twenty-Eighth Annual Report.

NUREG/CR4454. POOL CRITICAL ASSEMBLY PRESSURE VESSEL NUREG/CR-4674 V23: PRECURSORS TO POTENTIAL SEVERE CORE FACILfTY BENCHMARK.

NUREG/CR4504 V01: AN UPDATED NUCLEAR CRITCAUTY SLIDE DAMAGE ACCIDENTS: 1995. A Status Report.

RULE. Technical Basis.

NUREG/CR-4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4505 V01: THE POTENTIAL FOR CRITICAUTY FOLLOW-DAMAGE ACCIDENTS: 198243.A Status Report.

NUREG/CR-6167: LATE PHASE MELT PROGRESSON EXPERIMENT ING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE MP-2.Results And Analysis.

NU GC 8

M NT TA VERSION 1.

NUREG/CR4527: FINAL RESULTS OF THE XR21 BWR METALLIC NUREG/CR4508: COMPONENT UNAVAl 8 Y VERSUS INSERV-MELT RELOCATON EXPERIMENT.

ICE TEST (IST) INTERVAL: EVALUATIONS OF COMPONENT AGING SCIENTECH, INC.

N CR4528 E I A

S El PROPOSED U.

NUREG/CP-0159-PROCEEDINGS OF THE OECD/CSNI WORKSHOP CENSE RENEWAL OF NUCLEAR METALS,1NC. CONCORD, MASSA.

ON TRANSIENT THERMAL-HYDRAUUC AND NEUTRONC CODES CHUSETTS.

REQUIREMENTS. Held in Annapolis. Maryland, USA. November 5-8, NUREG/CR4558: NRC ANTITRUST UCENSING ACTONS. 1978 1996.

1996.

ORGANIZATION FOR ECONOMIC COOPERATION & DEVELOPMENT SOFTWARE EDGE, INC, NUREG/CP-0158: PROCEEDINGS OF THE OECD/CSNI SPECIAUSTS MEETING ON BORON DILUTON REACTIVITY TRANSIENTS. Held in NUREG/CR4370 BLOCKAGE 2.5 USER'S MANUAL NUREG/CR4371: BLOCKAGE 2.5 REFERENCE MANUAL Sttta Colleos, arma. USA,0ctober 18-20.1995.

NUREG/CP4159: PR EDWGS OF THE OECD/CSN1 WORKSHOP SOHAR, INC, ON TRANSIENT THERMAL-HYDRAULC AND NEUTRONIC CODES NUREG/CR4463 RO1: REVIEW GUIDEUNES FOR SOFTWARE LAN-REQUIREMENTS. Held in Annapolis. Maryland. USA, November 54-GUAGES FOR USE IN NUCLEAR POWER PLANT SAFETY 1996.

SYSTEMS. Final Report.

PENNSYLVANIA STATE UNIV, UNIVERSITY PARK. PA NOREG/CP-0158: PROCEEDINGS OF THE OECD/CSNI SPECIAUSTS SOUTH CAROUNA, UNIV. OF COLUMBIA, SC MEETING ON_ BORON DILUTION REACTIVITY TRANSIENTS. Held inNUREG/CR4529: VALIDATON OF TECTONC MODELS FOR AN IN-TRAPLATE SEISMIC ZONE, CHARLESTON, SOUTH CAROUNA WITH f

G/

7:MIE EA

(

NOMENON ON A GPS GEODETIC DATA.

DOWNWARD FACING CURVED SURFACE.

SOUTHWEST RESEARCH INSTITUTE SANDIA NATIONAL LABORATORIES NUREG/CR4074 V03: SEALED SOURCE AND DEVICE DESIGN NUREG/CR4674 V24: PRECURSORS TO POTENTIAL SEVERE CORE SAFETY TESTING.Techncal Report On The Fs'idings Of Task 4.Inves-DAMAGE ACCIDENTS: 198243.A Status Report.

tgation Of A Failed Brachytherapy Needle Appleator.

76 Contractor index ST. LOUIS UNIV., St. LOUIS, MO UNITED KINGDOM NUREG/CR4523 Vot: PROBABILISTIC ACCIDENT CONSEQUENCE NUREG/CR4563: LG EXCITATION, ATTENUATION, AND SOURCE UNCERTAINTY ANALYSIS. Food Cham Urcertamty AssessmentMam SPECTRAL SCALING IN CENTRAL AND EASTERN NORTH AMER.

NUR CR4523 V02-PROBABILISTIC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS.

Food Cham Uncertainty TECHNADYNE ENGINEERING CONSULTANTS, INC.

NUREG/CR4547: DOSFAC2 USER'S GUIDE.

NUR G/

52 PROBABILISTC ACCIDENT CONSEQUENCE UNCERTAINTY ANALYSIS. Uncertamty Assessment For Depostled TRANSPORTATION, DEPT. OF Material And External Doses. Main Report.

NUREG-1608 DAFT FC: CATEGORIZING AND TRANTORTING LOW NUREG/CR4526 V02: PROBABILIST6C ACCIDENT CONSEQUENCE SPECIFIC ACTIVITY MATERIALS AND SURFACE CONTAMINATED UNCERTAINTY ANALYSIS. UNCERTAINTY ASSESSMENT FOR DE-OBJECTS. Draft Rept For Comment POSITED MATERIAL AND EXTERNAL DOSES.Appendees.

international Organization Index This index !!sts, in alphabetical order, the countries and performing organizations that pre-l 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/lA reports published this year.

77

l l

tg a so 10 l

1

Licensed Facility Index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is needed, refer to the main citation by the NUREG number.

lectnc Demon, General Electnc Co., f S01 NuRE = 4541 R02

g;p gg,,gg, gg_ g;gg

=

7pcg

• w.

(

443453 Anas Corp., Denver, CO, NUREG-1532 54317 CaNert cms Nudeer Power Plant, Und 1, NUREG/CR4469 S 280 Surry Power Stanon, Und 1, Vrgne Electnc & NUREG/CR4181 R01 Bamrnere G4 & Elecinc P w Ca 50 318 CaNert cms Nudeer Power Plant, Und 2.

NUREG/CR4469 52 @ 2 System 80+ Standanized Nudeer Power Plant NUREG 1462 S01 Bamrnore Gas & Electne Des., Combuscan Engnee 79

I NRC FORM 336 U.s. NUCLEAR REGULATORY COMMiss4ON

1. REPORT NUMBER Q4%

(Assigned ty NRC, Add Vol., Supp., Rev.,

78 BIBLIOGRAPHIC DATA SHEET rsee muuucems an te mase)

NUREG-0304 2.TrrLE AND SUBTITLE Vol. 22, No. 4 Regulatory and Technical Reports (Abstract Index Journal) 3.

DATE REPORT PUBLISHED Annual Compilation for 1997 Mom l

YEAR l

April 1998

4. FIN OR GRANT NUMBER

5. AUTHOR (S)

6. TYPE OF REPORT

7. PERICO COVERED (inclusse Deses)

January - December 1997

8. PERFORMING ORGANIZATION NAME AND ADDRESS (rNRc, powde omem otree or asom u.s. NxmarRegumkry conmason, omt n=Ang senes, remtech povce nome omt merkng somes )

Publishing Services Branch Office of the Chief Information Officer U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (rNRc, type seme es above; acontactr. povee NRc Owam care or Regon, u s NucAser Rogue try comm,ssm a

amt medme amens) s Sama as 8, above.

Io, SUPPLEMENTARY NOTES L L Stevenson, Project Manager

11. ABSTRACT (:100 wmis or aess)

This journalincludes all formal reports in the NUREG series prepared by the NRC staff and contractors; proceedings of conferences and workshops; as well as intemational 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.

12. KEY WORDS/DESCRIPTORS (Lier motis er pireaes ret ws# essist manchers a face 6ag me <epcrf) 13 AvAM8W SWEMENT compilation unlimited abstract index H SECURITYCLASSWICATION rins Page) unclassified trnus Repero unc'assified

15. NUMBER OF PAGES

16. PRICE NRc FORM 335 (2 8g)

Printed on recycled paper Federal Recycling Program

WL E

P A A 2aS o S=08 EN S R A

H acu<E e0s L

I R

O N

U T

E Y

G N

F FF T GI OlQ OUT R

LE A

PL

,N AD R B D T S NU O

$8cvmtKoge AS C

T TI R A E N E

2 Y T z=E2' 2g U S 0 C E SS 5

S

,E 5O 5

M EM 300 XI

)

S 1

IS a.e8= <si s4x O

N msEg _E@x W2TU1 AWPS2 SFS 0 HN N5 zmo ou sNs oEcE=8 m

m I

PR0 N6OC5 sug m3ue rs a

GER 1 T7 O5 O NI4 N UR4 RM8 E 6 G

zmoou sms o$eEn8 m

a sug s tciog5n8e5 w 1

1 A

N I

D A

C C

zmo o0cEUSm I

A m.agm8 cug 2

5 0

1 5

C 5

V 5

1 oogeuo;Cex ssEa=o8 a=g5=8 sus PO S

P TIF E

AR R

GS M

ET T

AC aogmE eUEx NL N

D A S

sug FS E

G EM 4

SA 7

l P l A

ID