ML20059P017
| ML20059P017 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1990 |
| From: | NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | |
| References | |
| NUREG-0304, NUREG-0304-V15-N02, NUREG-304, NUREG-304-V15-N2, NUDOCS 9010240391 | |
| Download: ML20059P017 (49) | |
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Vol.15, No. 2 l
Regulatory and Technical Reports (Abstract Index Journal) f I
Compilation for Second Quarter 1990 April - June U.S. Nuclear Regulatory Commission Omce of Administration i
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Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 i
A year's subscription consists of 4 issues for this publication.
Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161
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NUREG-0304 Vol.15, No. 2 Regulatory and Technical Reports (Abstract Index Journal)
Compilation for Second Quarter 1990 April - June Date Published: August 1990 Regulatory Publications Branch Division of Freedom ofInformation and Publications Services Omce of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555
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CONTENTS Preface................................................................................v Index
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Main Citations and Abstracts............................................................ 1
- Staff Reports
- Conference Proceedings t
- Contractor Reports
- International Agreement Reports '
Secondary Report N umber Index......................................................... 2 Personal A uthor index................................................................. 3 S u bject in de x......................................................................... 4 NRC Originating Organization index (Staff Reports)........................................5 NRC Originating Organization index (international Agreements)............................... 6 NRC Contract Sponsor Index (Contractor Reports).......................................... 7 Cont ractor index..................................,.................................. 8 International Organization index..........................,................................ gi Licen sed Fa cility inde x.................................................................. 10,
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N PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC's
-l Intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap- -
preciated. Please send them to:
Division of Publications Services
- Policy and Publications Management Branch i
Publishing and Translations Section Woodmont 537 U.S. Nuclear Regulatory Commission Washington, D.C. 20665 e'
q The main citations and abstracts in this compilation are' listed in NUREG number order: NUREG-XXXX, NUREG/CP XXXX, NUREG/CR XXXX, and NUREG/lA-XXXX. These precede the following indexes:
- I s
-Secondary Report Number Index Personal Author Index 1
Subject index,
1 NRC Originating Organization Index (Staff Reports)
NRC Originating Organization index (International Agreements).
NRC Contract Sponsor Index (Contractor Reports)-
Contractor Index Intemational Organization Index Licensed Facility index A detailed explanation of the entries precedes each index.
The bibliographic elements of the main citations are the following:
Staff Report NUREG-0008r MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.'
ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048. 09670:200 a
Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of -
author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession numberi 18) the microfiche address (for ir temal NRC use).
Conference Report NUREG/CP-0017: EXECUTIVE SEMIM \\R ON THE FUTURE ROLE OF RIS'K ASSESSMENT AND' RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP; J.S Argonne National Laboratory. May 1981.141 pp. 8105280299. ANL 813. 08632:070.
Where hntries are (1) report number, (2) report title, (3) report author, (4) organization that compiled the precnuings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use).
Contractor Roport i
NUREG/CR-1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER i REACTORS CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.: BENNETT, P.R.
1 Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242.
U 4
Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of authors or publisher, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if -
given), and (9) the microfiche address (for NRC Intemal use).
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n International Agreement Report
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NUREG/lA-0001: ASSESSMENT OF TRAC-PD2 USING SUPER CANNON AND HDR EXPERIMENTAL 4
4 DATA. NEUMANN, U. Kraftwerk Union.: August 1986. 223 pp. 8008270424. 376 tie:138.
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Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, 4,
(5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System
)
accession number, (8) the report number of the originating organization (if given), and (9) the microfiche e
address (for NRC internal use).
The following abbreviations are used to identify the document status of a report:
ADD
- addendum.
APP appendix -
DRFT draft
- ERR errata-N. number R
revision S - supplement u
V - volume
, Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office 4 l
(GPO) or from the National Technical Information Service, Springfield, Virginia 22161. To purchase =.
documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the following addmas:
t bvointendent of Documents U.S. Govemment Printing Office.
Post Office Box 37082-Washington, DC 20013 7082 '
You may charge any purchase to your GPO Deposit Account; Mastercerd charge card, or VISA charge card by calling the GPO on (202)275 2000 or (202)275-2171. Non U.S. customers must make pavment in '
advance either by International Postal Money Order, payable to the Superintendent of Documents, or ~
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by draft on a United States or Canadian bank, payable to the Superintendent of Documents.
NRC Reoort Codes The NUREG designation, NUREG XXXX, indicates that the document is a formal NRC staff-generated report. Contractor prepared fccmal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-L NUREG XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of
' l the work being reported.
in addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference'-
proceedings and NUREG/lA is used for international agreement reports.'
All these report codes are controlled and posigned by the staff of the Publishing and Translations Section -
of the NRC Division of Publications Services.
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Main Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG XXXX is an NRC staff-o inated report, NUREG/CP XXXX is an NRC-sponsored conference report, NUREG/CR-is an NRC contractor prepared report, and NUREG/lA XXXX is an inter-national agreement re > oft. The bibliographic information (see Preface for details) is followed by a brief abstract of tils report.
NUMEG 0030 V14 N02: LICENSED OPERATING REACTORS NUREQ4325 RIS: U.S. NUCLEAR REGULATORY COMMISSION CIATUS
SUMMARY
REPORT. Data As Of January FUNCTIONAL ORGANIZATION CHARTS. April 1,1990.
- Olc of 31,1990.(Grey Book 1) LOVELACE W.H. DMeion of Computer &
Personnel (Post 870413). April 1990. 64pp. 9005070012.
Telecommunicateons Services (Post 800205). April 1990. 564pp.
63656:347, 9005040120. 63648:104.
Functional organizaten charts for the U.S. Nuclear Regulatory THE OPERATING UNITS STATUS REPORT UCENSED Commesion offces, dMeions, and branches are presented.
OPERATING REACTORS provides data on the operation of nu.
clear uNts as timely and accurately as possibic. This infwma-NUMEG 0908 006 M06: UNITED STATES NUCLEAR REGULA-l tion is collected by the Offee of information Resources Man-TORY COMMISSION STAFF PRACTICE AND PROCEDURE 1
agement from the Headquarters staff of NRC's Office of En-DIGEST. Commission, Appeal Board And ucensing Board -
forcement (OE), from NRC's Regional Offmes, and from utihties.
DecnolontJuly 1972 Dwenby 1989.
- Office of the Genwel The three sections of the report are: monthly highlights and sta-Counsel (Post - 860701)E June 1990. 800pp. 9007120220.
54473:203.
tietes for commercial operahng uNts, and errata from previously reported deta; a compilation of detailed information on each This Revision 6 of the fifth edition of the NRC Practice and unit, provided by NRC's Regional Offices, OE Headquarters and Procedure Digest contains a digest of a number of Commission, the utilities; and an appendix for miscellaneous information such Atomic Safety and Ucensing Appeal Board, and Atomic Safety as spent fuel storage capability, reactor years of experience and and Licensing Board decisions losued during the period July 1, d
1972 to December 31, 1989, interpreting the NRC's Rules of I
non-power reactors in the U.S. It is hoped the report is helpful Practice in 10 CFR Part 2.
to all agencies and indMduals interested in = maintaining an -
awareness of the U.S. energy situation as a whole.
NUREG4430 V00 N02: UCENSED FUEL FACluTY STATUS NUREG4090 Vit N04: REPORT TO CONGRESS ON ABNOR-REPORT. Inventory Difference Data. July 1988 June 1989.(Gray Book II)-
Office of Nuclear Material Safety &
MAL OCCURRENCES. October December 1989.
- Ofice for Safeguards. Director. April 1990.18pp. 9005210145. 53861:345.
'j Analysis & Evaluston of Operatonal Data, Director, March 1990. 21pp. 9005040046. 53622:224.
NRC is committed to the periodic pubication of licensed fuel Section 200 of the Energy Reorgantzation Act of 1974 identi-faclhties inventory difference data, following agency review of l
fios an abnormal occurrence as an unscheduled incident or the information and completion of any related NRC investiga.
tions. Information in tNs report includes inventory difference event which the Nuclear Regulatory Commission determines to data for active fuel fabrication facihtees possessing more than be signifcant from the standpoint of public health and safety -
one effective kilogram of high enriched uranium, low enriched I
and requires a quarterty report of such events to be made to uranium, plutonium, or uranium 233.
Congress. This report covers the period October 1 through De-cember 31,1985. For this reporting period, there were three ab.
NUREQ4640 VII N12: TITLE LIST OF DOCUMENTS MADE normal occurrences, none invoMng a licenood nuclear power PUBLICLY AVAILABLE. DECEMBER 1 31,1989.
- DMeion of 1
plant Two of the abnormal occurrences took place under other Freedom of Information & Publications Services (Post 890205).
NRC-issued licenses. The first involved a medical diagnostic -
Apr91990. 230pp. 9005210149. 53887:2%
misadministration and the second involved a medcal therapy Th6s document is a monthly pubhcotion containing descrip-misadrninistration. The tNrd abnormal occurrence was reported Ums of infwmahon by an Agreement State (Louisiana) and involved an overexpo-Regulatwy G,no f.eceived and generated by the U.S. Nuclear -
_ (NRC). TNs information includes (1)
.v,
sure to an industrial radiographer. The report also contains in-docketed mateW assodated e cWan nuclear poww plants formation that updates a previously repor'sd abnormal occur-r a ce.
ated nt as
-t a regulatory agency. The fonowing indexes are included: Per-NUREG 0304 V16 N01: REGULATORY AND TECHNICAL RE.
smal AN, Capwate Source, Repwt Nuh and Cmsa -
PORTS (ABSTRACT INDEX JOURNAL). Compilation For First Reference to Principal Documents.
q Quarter 1990, January. March.
- DMaion of Freedom of informa.
NUREG-0640 V12 N01: TITLE UST OF DOCUMENTS MADE i
tion & Pubhcations Services (Post 890205) May 1990. 53pp.
9006000185. 54074:156.
PUBUCLY AVAILABLE. January 1 31,1990.
- DMeion of Free-This joumal includes all formal reports in the NUREG series dom of Information & Publications Services (Post 890205). April 1990 332pp. 9005210255. 53689:146.
prepared by the NRC staff and contractors; proceedings of cm-forences and workshops; as well as intomational agreement re-See NUREG-0540,V11,N12 abstract.
ports. The entries in this compilation are indexed for access by NUREQ4540 V12 N02: TITLE UST OF DOCUMENTS MADE title and abstract, secondary report number, personal author, PUBUCLY AVAILABLE. February 1 28, 1990.
- DMeion of Free-subject, NRC organ 2ation for staff and intomational agree-dom of Information & Publications Services (Post 890205). June ments, contractor, intemational organization, and Econsed facili-1990. 304pp. 9006290112. 5432(008.
- (
ty.
See NUREG-0540,V11,N12 abstract.
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Main Citations and Abstracts NURE04780 V30102: INDEXES TO NUCLEAR REGULATORY' NUREG 0837 V10 N01: NRC TLD DIRECT RADIATION MONI-COMMISSION ISSUANCES. July December 1989.
- Drvision of TORING NETWORK. Progress Report January-March 1990.
Freedom of Information & Publications Serv 6ces (Post 890205).
STRUCKMEYER,R.; MCNAMARA,N. Region 1, Ofc of the Direc.
May 1990. 87pp. 9007120183. 54486:237,
' tor. June 1990. 226pp. 9007120185. 54470:336.
Digests and indexes for issuances of the Commission, the ;
This report provides the status and results of the NRC Thor-j Atomic Safety and Licensing Appeal Panel, the Atomic Safety
' moluminescent Dossmeter (TLD) Drect Radiation Monitoring i
j and Licensing Board Panel, the Administrative Law Judges, the
- Network. It presents the radiation levels measured in the vicinity i
!=
Directors' Decisions, and the Denials of Petitions for Rulemak-of NRC licensed facilities throughout the country for the first l
ing are presented.
quarter of 1990 NUREG 0780 V30 N06: NUCLEAR REGULATORY COMMISSION NUMEG-0034 V00 N01: NRC REGULATORY AGENDA. Quarterly l
lSSUANCES FOR DECEMBER 1989. Pages 709-811,
- Division -
ReportJanuary-March 1990
- Division of Freedom of Informa-of - Freedom of information & Pubhcations Services (Post tion & Pubhcations Services (Post 890205). April 1990,145pp.
890205). April 1990.110pp. 9007120158. 54486:319, 9005210250. 53889:001, The NRC Regulatory Agenda is a compilation of all rules on i
Legalissuances of the Commission, the AtorNe Safety and Ll-
. wNch the NRC has proposed or is considering action and all
[
censing Appeal Panel, the Atomic Safety and Licensing Board l'
Panel, the Administrative Law Judges, and NRC Program Of-Petitions for rulemaking which have been received by the Com-fices are presented.
mission and are pending disposition by the Commission. The NUREG-0780 V31 N01: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR JANUARY 1990. Pages 1130.
- Drvision of NUREG-0940 V00 N04: ENFORCEMENT ACTIONS:SIGNIFICANT Freedom of information & Publications Services (Post 890205).
ACTIONS RESOLVED. Quarterly Progress Report,0ctober De-April 1990.137pp. 9007120154. 54487:064.
cember 1S89.
- Ofc of Enforcement (Post 870413). March See NUREG-0750,V30,N06 abstract.
1990. 289pp. 9005040056. 53623:190..
This compilation summarizes significant enforcement actions NUREG 0750 V31 N02: NUCLEAR REGULATORY Wic, MISSION irut have been resolved during one quarterty period (October -
ISSUANCES FOR FEBRUARY 1990. Pages 131 19b. '. Division
' Da, ember 1989) and includes copies of letters, Noticos, and of Freedom of information' & Pubhcations Services (Post Orders sent by the Nuclear Regulatory Comrression to licensees 800205). April 1990. 70pp. 9007120148. 54487:199.
with resoect to these enforcement actions. Also included are a -
See NUREG 0750,V30,N06 abstract.
number of enforcement actions that had been previously re-solved but not published in this NUREG. It is anticipated that NUREG 0750 V31 N03: NUCLEAR REGULATORY COMMISSION the information in this publication will be widely disseminated to ISSUANCES FOR MARCH 1990.Pages 197 332.
- Division of
. managers and employees engaged in activities licensed by the Freedom of information & Publications Services (Pcat 890205).
NRC, so that actions can be taken to improve safety by avoid -
May 1990.125pp. 9007120122. 54487:268.
Ing future violations similar to those described in this pubhca-See NUREG-0750,V30,N06 abstract.
tion.
NUREG-0750 V31 N04: NUCLEAR REGULATORY COMMISSION NUREG 0940 V00 N01: ENFORCEMENT ACTIONS: SIGNIFl-ISSUANCES FOR APRIL 1990.Pages 333-370.
- Division of CANT ACTIONS RESOLVED. Quarterly Progress Freedom of Information & Pubhcations Services (Post 890205).
Report, January. March 1990. M Ofc of Enforcement. ; Post June 1990. 44pp. 9007120119. 54488;047, 870413). May 1990. 403pp. 9006290085. 54320:263.
See NUREG-0750,V30,N06 abutract.
This compilation summarizes significant enforcement actions that have been resolved during one quarterty period (January -
l NUREG 0797 S24: SAFETY EVALUATION REPORT RELATED March 1990) and includes copies of letters, Notices, and Orders i
TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-sent by the Nuclear Regulatory Commission to licensees with 1RIC STATION UNITS 1 AND 2. Docket Nos. 50-445 And 50-respect to these enforcement actions. Also included are a f
446.(Texas Utihties Electric Company,et al.)
- Comanche Peak number of enforcement actions that had been previously re-Project Division (890101 900602). April 1990, 174pp.
solved but not pubhshed in this NUREG. It ta anticipated that l
9005180159. 53861:137.
the information in this publication will be widely disseminated to Supplement 24 to the Safety Evaluation Report related to the managers and employees engaged in activities licensed by the oporation of the Comanche Peak Steam Electric Station, Units NRC, so that actions can be taken to improve safety by avoid-1 and 2 (NUREG.0797), has been prepared by the Office of Nu-ing future violations similar to those described in this publica-clear Reactor Regulation of the U.S. Nuclear Regulatory Com-tion.
mission. The facihty is located in Somervell County, Texas, ap-proximately 40 miles southwest of Fort Worth, Texas. This sup.
NUREG 1021 R06: OPERATOR LICENSING EXAMINER STAND-piement reports the status of certain issues that had not been ARDS. DEAN W. Division of Licensee Performance & Quality resolved at the time of publicaten of the Safety Evaluation Evaluation (Post 870411). June 1990. 315pp. 9006290065.
Report and Supplements 1, 2, 3, 4, 6,12, 21, 22, and 23 to that 54321:339.
report. This supplement also includes the evaluations for licens.
The Operator Licensing Examiner Standards provide policy ing items resolved since Supplement 23 was issued. Supple-and guidance to NRC examiners and establish the procedures ment 5 has been cancelled. Supplements 7 through 11 were -
and practices for examining and licensing of applicants for NRC hmited to the staff evaluation of allegations investigated by the operator licenses pursuant to Part 55 of Title 10 of the Code of NRR Technical Review Team. Supplement 13 presented the Federal Regulations (10 CFR 55). They are intended to assist staff's evaluation of the Comanche Peak Response Team NRC examiners and facihty licensees to understand the examin.
(CPRT) Program Plan, which was formulated by the licensee to ing process better and to provide for equitable and consistent resolve vanous construction and design issues raised by administration of examinations to all applicants by NRC exanun-sources external to the hcensee. Supplements 14 through 20 ers. These standards are not a substitute for the operator li-i presented the staff's evaluation of the licensee's Corrective censing regulations and are subject to revision or other intemal Action Program and CPRT activities. Items identified in Supple-operator licensing program changes. As appropriate, these ments 7, B,9,10,11,13, and 15 through 20 are not included in standards will be revised periodically to accommodate corre this supplement, except to the extent that they affect the licens-ments and reflect new information and experience associated ee's Final Safety Analysis Report.
with operator licensing activities.
I l;.-
Main Citations and Abstracts 3
NUREG 1126 VII: A COMPILATION OF REPORTS OF THE AD-NUREG 1372: REGULATORY ANALYSIS FOR THE RESOLU-VISORY COMMITTEE ON REACTOR SAFEGUARDS.1989 TION OF GENERIC ISSUE C 8, " MAIN STEAM ISOLATION Annual.
VALVE LEAKAGE AND LCS FAILURE." GRAVES,C.C. Division April 1990.150pp. 9005210245. 53888:210 of Safety issue Resoluton (Post 880717). June 1990. 29pp.
1 TNs compilaton contains 54 ACRS reports submitted to the 9006290054, 54321:306 l
Commission or to the Executwe Director for Operatons during -
Generic lasue C-8 deals with staff concems about public risk t
calendar year 1989. It also includes a report to the Congress on because of the incidence of leak test failures reported for main the NRC Safety Research Program. All reports have been made steam isolaton valves (MSivs) at boiling water reactors and the available to the public through the NRC Public Document Room limitatens of the leakage control systems (LCS) for mitegabng and the U.S. Library of Congress. The reports are divided into the consequences of leakage from these valves. If the MSiv two groups: Part 1: ACRS Reports on Project Reviews, and Part leakage is greatty in excess of the allowable value in the techni-2: ACRS Reports on Generic Subjects. Part 1 contains ACRS cal specifcations, the LCS would be unavailable because of reports alphabetcod by proket name and within project name design limitations. The issue was initiated in 1983 to assess (1) by chronological order Part 2 categortzes the reports by the the causes of MSIV leakage failures, (2) the effectiveness of the mo81 appropriate generic subject area and within subject area LCS and alternative mitigabon paths, and (3) the need for addi-by chronological order, tonal regulatory acton to reduce public risk. This report pre-NUREG 1266 V04: NRC SAFETY RESEARCH IN SUPPORT OF sents the regulatory analysis for Generic issue C 8 and con-REGULATION FY 1989.
- Offee of Nuclear Regulatory Re-cludes that no new regulatory requirements are warranted.
search (Post 860720). April 1990. 61pp. 9005040092.
NUREG 1390: SAFETY EVALUATION REPORT RELATED TO 53622:113.
THE RENEWAL OF THE OPERATING LICENSE FOR THE
. TNs report, the fifth in a series of annual reports, was pre-TRIGA TRAINING AND RESEARCH REACTOR AT THE UNI-pared in response to congressional inquiries concoming how VERSITY OF ARIZONA. Docket No. 50-113. (University of Artzo-nuclear regulatory research is used. ft summartzes the accom-na)
- Division of Reactor Projects. Ill,lV.V & Special Projects plishments of the Office of Nuclear Regulatory Research dunng (Post 870411). May 1990. 74pp. 9006080200. 54074:209.
FY 1989. The goal of tNs office is to ensure that safety.related TNs Safety Evaluation Report for the application filed by the research provides the technical bases for rulemaking and for re-University of Arizona for the renewal of Operating License R-52 Lated decisions in support of NRC licensing and inspection ac-to continue operating its research reactor at an increased oper-tivities. TNs research is necessary to make certain that the reg-ating power level has been prepared by the Office of Nuclear ulatons that are imposed on iconsees provide an adequate Reac'or Regulabon of the U.S. Nuclear Regulatory Commission.
margin of safety so as to protect the health and safety of the The facility is located on the University of Arizona campus in pubhc. TNs report describes both the direct contributions to sci-
. Tucson, Arizona. The staff concludes that the reactor can con-entifc and technical knowledge with regard to nuclear safety tinue to be operated by the University of Arizona without enden-and their rogulatory apphcahons.
gering the health and safety of the pubkc.
NUREG 1333: MAINTENANCE APPROACHES AND PRACTICES NI'AEG 1391 DRFT FC: CHEMICAL' TOXICITY OF URANIUM IN SELECTED FOREIGN NUCLEAR POWER PROGRAMS AND HEXAFLUORIDE RELATED TO RADIATION DOSES. Draft OTHER U.S.
INDUSTRIES: REVIEW AND LESSONS Report For Comment. MCGUIRE S.A. Office of Nuclear Regula, j
LEARNED. DEY,M. Division of Regulatory Applications (Post tory Research (Post 860720). April 1990.16pp. 9005210238.
870413). April 1990.111pp. 9005070010. 53657:051.
53861:328.
The Commission published a Notice of Proposed Rulemaking The chemical effects from large acute exposures to uranium on Maintenance of Nuclear Power Plants on November 28, hexafluoride are compared to the effects from acute radiation
- 1988, spelling out NRC's expectations in maintenance. In pre-6ees of 25 rems to the whole body and of 300 reme to the paring the proposed rule, the NRC reviewed maintenance prac-thy,mid. The analysis concludes that an acute exposure to about tices in other countries and considered maintenance approach-es in other industries in this country. As a result of the review of 500 rug. min /m(3) of hydrogen fluoride is roughly equivalent, in terms of earty effects, to an acute whole body dose of 25 rems.
maintenance practices, it was concluded that certain practces Similarly, an intake of about 10 mg of uranium in soluble form is in the following areas have been found to contribute signifcant-roughly equivalent, in terms of earfy effects, to an acute whole ly to effective maintenance: (1) systems approach; (2) effective-body dose of 25 rems. The purpose of these analyses is to pro-ness monitoring; (3) technician quahficahons and motivation; and (4) maintenar'ce organization.
vide information for developing siting criteria for uranium ennch-ment plants based on chemical toxicity. These siting criteria are NUREG 1362: ACTION PLANS FOR MOTOR OPERATED to be sirndar, in terms of health and safety impact on workers VALVES AND CHECK VALVES SCARBROUGH,T.G. Division and members of the public, to the siting criteria in NRC regula-tions for nuclear power plants, which are based on radiation of Engineenng Technology (Post 890827) June 1990. 34pp.
doses.
9006290061, 54318:321 The proper performance of motor operated valves (MOVs)
NUREG 1393: THE INCINERATION OF LOW. LEVEL RADIOAC-and check valves is necessary for the safe operaton of a nucle-1 er power plant. Problems have been exponenced with these TlVE WASTE.A Report For The Advisory Committee On Nuclear Waste. LONG,S.W. Advisory Committee on Nuclear Wasto.
valves for many years. Currently, the U.S. Nuclear Regulatory June 1990. 75pp. 9007120183. 54469:053.
l Commession (NRC) and the nuclear industry have a number of This report is a summary of the contemporary use of inciner-
' activities under way to provide assurance that MOVs and check ation technology as a method for volume reduction of LLW It is valves will successfully perform their safety functions when Intended pnmarily to serve as an overview of the technology for needed. The Mechanical Engineering Branch (EMEB) of the Office of Nuclear Reactor Regulabon has been assigned the re-waste management professionals Involved in the use or regula-sponsibility of coordinating NRC activities and monitonng indus-tion of LLW incinerabon. It is also expected that organizations try activities regarding MOVs and check valves. To meet this re-presently considering the use of incineration as part of their ra-sponsibility, EMEB has prepared action plans to provide assur.
dioactive waste management programs will benefit by gaining a ance of the proper performance of these valves. Through the general knowledge of incinerator operating experience. Specifc action plans, the NRC statt will have an organized approach to types of incineration technologies are addressed in this report,
-resolve the concems regarding the operability of MOVs and including designation of the kinds of wastes that can be proc-check valves in a bmely manner.
essed, the magnitudes of volume reduction that are achievable in typical operation, and requirements for ash handling and off.
I
_ __a
1-4-
Main Citations and Abstrac's gas fihenng and scrubbing. A status hsbry of both U.S. and for.
NUREG-1404: LICENSEE USE OF TACTICAL EXERCISE RE-eign incinerators provides Nghhghts of a:tsvities at government, SULTS. SAWYER,C.; BROWN C. Division of Safeguards &
Industry, institutonal, and commercial nt clear power plant sites.
Transportation (Post 870413). April 1990.11pp. 9005210234.
53861:312.
NUREG 1364: EMERGENCY RESPONSE DATA SYSTEM (ERDS)
On November 10, 1988 the Nuclear Regula'ory Commission IMPLEMENTATION. JOLICOEUR,J.R. Office for Analysis &
(NRC) amended its physical secunty requirements in 10 CFR Evaluabon of Operabonal Data, Director. April 1990. 53pp.
Part 73 for fuel fac6hties possessing formula quantities of strate-9006080154. 54077;312.-
gic special nuclear material The amenoments to 10 CFR The U.S. Nuclear Regulatory Commission has begun imple.
73.46(b) require, among other things, that Icensees carry out mentabon of the Emergency Response Data System (ERDS) to performance evaluations through tactical response exercises.
upgrade its ability to acquire data from nuclear power plants in The exercises are intended to demonstrate the guard force the event of an emergency at the piant. ERDS provides a direct state of readiness and to test the effectrveness of delay mecha-real-bme transfer of data from licensee plant computers to the nisms, alarm and communication systems, response times, de-NRC Operations Center. The system has been designed to be ployment of response forces, firing skills (simulated), tactical actuated by the licensee during an emergency which has been maneuvers, etc. The purpose of tNs document is to set forth classified at an ALERT or Ngher level. The NRC porton of crtteria, acceptable to the NRC staff, wNch will enable a heens-ERDS will receive the data stream, sort and file the data. The se to use the results of an exercise to determne whether addi-j users will include the NRC Operations Center, the NRC Region.
tonal training or security improvements are needed.
I al Office of the aftseted plant, and if requested, the States NUREG 1405: INADVERTENT SHIPMENT OF A RADIOGRAPHIC.
wNch are witNn the ten mile EPZ of the site. The currently in-SOURCE FROM KOREA TO AMERSHAM stalled Ernergency Notification System will be used to supple-CORPORATION.BURLINGTON MASSACHUSETTS,
- NRC -
I rnent ERDS data. This report provides the minimum guidance No Detailed Affihation Given. May 1990.177pp. 9006110005.
for implementaton of ERDS at hcensee sites. It is intended to 54085:017 be used for planning impiettentation under the current volunta'y Amersham Corporation, Burlington, Massachusetts, a licensee program as well for providing the minimum standards for imple-of the U.S. Nuclear Regulatoey Cvie,a46 (NRC), authorized menting the proposed ERDS rule, b mandWe W dete Mum 192 and cobalt 60 source NUREG-1399: TECHNICAL SPECIFICATIONS, COMANCHE PEAK assemblies for use in radiography equipment, received a ship-STEAM ELECTRIC STATION, UNIT 1. Docket No. 50-445,Appen-ment of 14 source changers on March 8,1990, that were being retumed from meir produd dstrbdw, NG Capmahon in Seod, dix "A" to Ucense NPF 87.(Texas Utilities Electric)
- Comanche Korea. One source changer contained a small sealed source in Peak Project Division (890101900602). April 1990. 360pp.
an mehelded bcain Amersham employees rem me 9006080150. 54081:109 I
source, secured it in a hot cell, and notified NRC's Region 1.
The.Techrucal Specificabons for Comanche Peak Steam Subsequently, NRC dispatched an incident investigation Team Electric Staten, Unit 1 were prepared by the U.S. Nuclear Reg-to perform a comprehensive review of tNs incident and deter.
ulatory Commission. They set forth the limits, operating cond".
mine the potential for exposure to those who handled the tions, and other requirements apphcable to a nuclear reactor fa source changer and to members of the general public, This cility, as set forth in Section 50.36 of Title 10 of the Code of report describes the incident and the methodology used in the i
Federal Regulations Part 50, for the protecton of the health and investigation and presents the Team's findngs and conclusions.
i safety of the public.
NUREG-1410: LOSS OF VITAL AC POWER AND THE RESIDUAL NUREG 1402: CLOSEOUT OF NRC BULLETIN 88-05:NONCON-HEAT REMOVAL SYSTEM DURING MID-LOOP OPERATIONS FORMING MATERIALS SUPPLIED BY PIPING AT VOGTLE UNIT 1 ON MARCH 20,1990.
- Ofc of the Execu-SUPPLIES,1NC..AT FOLSOM NEW JERSEY,AND WEST tive Director for Operations. June 1990. 522pp. 9006290080 JERSEY MANUFACTURING COMPANY AT 54319:101.
WILLIAMSTOWN,NEW JERSEY,
- Division of Engineenng On March 20,1990, the Vogtie Electric Generating Plant Unit Technology (Post 890827). May 1990. 79pp. 9006 10010. IEB-1, located in Burko County, Georgia, about 25 miles southeast 88-005, 54083:185.
of Augusta, experienced a loss of alt safety (vital) ac power.
This report documents the activites that led to the closeout The plant was in cold shutdown w'th reactor coolant level low-of U.S. Nuclear Regulatory Commission (NRC)Bulletin 88-05, ered to "mid-loop" for various maintenance tasks. Both the which was issued on May 6,1988. The bulletin required that li-containment buildin0 personnel hatch and equipment hatch censeos submit informaton on materials supplied by Piping were open. One emorgency diesel generator and one reserve Supplies, Inc. (PSI) and West Jersey Manufacturing Company auxiliary transformw were out of service for maintenance, with (WJM), and requested that they (1) ensure that these materials the remaining reserve auxihary transformer supplying both Unit complied with the American Society of Mechanical Engineers, 1 safety buses. A truck in the low voltage switchyard backed Boiler and Pressure Vessel Code (ASME Code) and design into the support column for an offsite power feed to the reserve specifications or were suitable for their intended service or (2) auxiliary transformer which was supplying safety power The in-replace such materials. Supplements 1 and 2 were issued on sulator broke, a phaao tMround fault occurred, and the feeder June 15 and August 3,1988, respectively. In Supplement 2, an-circuit breakers for the safety buses opened. The operable other affiliated supplier, Chews Landing Metal Manufacturers, emergency diesel generator started automatically because of incorporated (CLM), was identified. The staff concluded that (1) the undervoltage condition on the safety bus, but tnpped off the analytical procedures used to qualify the nonconforming after about 1 minute. About 20 minutes later the diesel genera-parts and the analysis results provide an adequate basis for re-tor load sequencer was reset, causing the diesel generator to solving the staff's concems regarding fittings and flanges; (2) start a second time. The diesel generator started again, operat-even though the materials supplied by PSI, WJM, and CLM with ed for about 1 minuto, and tripped off. The diesel generator was falsified certified material test reports do not meet the ASME restarted in the manual emergency mode 36 minutes after the Code, their use is an acceptable attemative in accordance with loss of power. The generator remained on hne and provided Section 50.55a(a)(3)(ii) of Title 10 of the Code of Federal Regu-power to its safety tAis. Dunng the 36 minutes without safety labons; (3) activities in response to Bulietin 88-05 regarding fit-bus power, the reactor coolant system temperature rose from tings and flanges can be closed for all operating plants; and (4) about 90 F to 136'F. This report documents the results of an hcensees should evaluate the use of product forms other than incident Investigation Team sent to Vogtle by the Executive Di-fittings and flanges.
rector for Operations of the U.S. Nuclear Regulatory Commis-1
Main Citations and Abstracts 5
sion to determine what happened, identify the probable causes, NUREG/CR 2000 V00 N4: LICENSEE IVENT REPORT (LER) and make appropriate findings and conclusions.
COMPILATION;For Month Of April 199).
- Oak Ridge Natonal Laboratory. May 1990. 83pp. 90060E 3270. ORNL/NSIC 200-NUREG/CP 0100: PUBLIC WORKSHOP ON NUCLEAR POWER 54074:072.
r PLANT LICENSE RENEWAL. HUGHES,A.A.; LIGON,0.M.:
See NUREG/CR-2000,V09,N03 abstract SE MITRE April 1990. 78pp. 9005020348. MTR-NUREG/CR 2000 V09 N5: LICENSEE EVENT REPORT (LER)
On 13 October 1989, the U.S. Nuclear Regulatory Commis-COMPILATION.For Month Of May 1990.
- Oak Ridge National sion (NRC) issued an Advance Notice of Proposed Rulemaking ory. June 1990.102pp. 9007120242. ORNL/NSIC-200.
on nuclear power plant license renewal The notice presented See NUREG/CR 2000,V09,NO3 abstract the NRC's preliminary regulatory philosophy and approach for
(
developing license renewat regulations, and solicited comments NUREG/CR 2331 V09 N4:. SAFETY RESEARCH PROGRAMS on a number of technical and pohey issues. It also announced SPONSORED BY OFFICE OF NUCLEAR REGULATORY plans for a pubhc workshop to discuss the issues and to receive RESEARCH. October December 1989. WEISS,A.J. Brookhaven comments and information. Representatives from 89 organiza-National Laboratory. June 1990.152pp. 9007120223. BNL-tions attended the workshop, held on 1314 November 1989, in NUREG 51454. 54486:291, Reston, Virginia. Subsequently.12 organizatons submitted writ-This progress report dc.xtibes current activities and technicai ten comments to the NRC. This report provides a summary of progress in the programs at Brookhaven National Laboratory both workshop and written comments, sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety issue Resolution, and Division NUREG/CR 1667: RISK METHODOLOGY FOR GEOLOGIC DIS-of Systems Research of the U.S. Nuclear Regulatory Commis.
POSAL OF RADIOACTIVE WASTE. Scenario Selection Proce-sion, Office of Nuclear Regulatory Research following the reor-dure. CRANWELL,R.M.; GUZOWSKI R.V.; CAMPBELL,J.E.; et ganization in Juiy 1988. The previous reports have covered the al.
Sandia National Laboratories. April 1990. 114pp.
period October 1,1976 through September 30,1989, 9005210151. SAND 801429. 53887:090.
This report contains the descripton of a procedure for select.
NUREG/CR 3960 V06: FUEL PERFORMANCE ANNUAL ing scenarios that are potentially important to the isolation of REPORT FOR 1988. BAILEY,W.J. Battelle Memorial Institute' high level radcactive wastes in deep geologic formations. In Pacific Northwest Laboratory. WU.S. Reactor Systems Branch this report, the term scenario is used to represent a set of natu-March 1990.175pp. 9005020352. PNL 5210, 53589:223.
rally occurring o'sd/or human-induced conditions that represent This annual report, the eleventh in a series, provides a brief realistic future states of the repository, geologic systems, and description of fuel performance during 1988 in commercial nu-ground-water flow systems tnat might affect the release and clear power plants. Brief summaries of fuel design changes, fuel transport of radionuclides from the repository to humans, The surveillance programs, fuel operating experience and trends, scenario selecten procedure discussed in this report is demon-fuel problems, high bumup fuel experience, and items of gener.
strated by applying it to the analysis of a hypothetical waste dis-al significance are provided. References to additional, more de-posal site containing a bedded-satt formation as the host tailed information and related NRC evaluations are included.
medium for the repository. A final set of 12 scenarios is select-NUREG/CR 4525: CLOSEOUT OF IE BULLETIN 84 03: REFUEL.
1 ed for this site.
ING CAVffY WATER SEAL - FOLEY,W.J.; DEAN,R.S.:
HENNICK,A. PARAMETER, Inc, June 1990.88pp.9007120236.
NUREG/CR-2000 V09 N3: LICENSEE EVENT REPORT (LER)
PARAMETER IE155. 54468:336.
-l COMPILATION-For Month Of March 1990.
- Oak Ridge Nation.
Documentation is provided in this report to close IE Bulletin al Laboratory. Apol 1990.83pp.9005210152. ORNL/NSIC-200.
63886:325.
84-03 on the subject of refueling cavity water seals. The bulletin was issued on August 24, 1984, to all power reactor facilities This monthly report contains Licensee Event Report (LER) except Fort St. Vrain. Because the excluded plant is gas-cooled operational information that was processed into the LER data and graphite-moderated, the subject of the bulletin does not file of the Nuclear Safety Information Center (NSIC) during the apply to that plant. The event causing the safety concem was one month period identified on the cover of the document The failure of the pneumatic seals containing water in the refueling LERs, from which this information is derived, are submitted to cavity at Haddam Neck on August 21,1984 The cavity had the Nuclear Regulatory Commission (NRC) by nuclear power been flooded in preparation for refueling, but when the seals plant licensees in accordance with federal regulations. Proce.
failed, water level decreased from -full level just below the oper-dures for LER reporting for revisions to those events occurring.
ating floor down to the level of the reactor vessel flange. On prior to 1984 are described in NRC Regulatory Guide 1.16 and May 17,1988, another refueling cavity seal leakage (about NUREG-0161, " Instructions for Preparation of Data Entry three feet in water level) occurred at Surry 1, which was in the Sheets for Licensee Event Reports." For those events occurring middle of a refueling and maintenance outage.. Evaluation of on and after January 1,1984, LERs are being submitted in ac-
. utility responses and NRC/ Region inspection reports in accord-cordance with the revised rule contained in Title 10 Part 50.73 ance with a specific criterion shows that the bulletin is closed of the Code of Federal Regulations (10 CFR 50.73 - Licensee for 114 (97%) of the 118 affected facilities. It is concluded that Event Report System) which was published in the Federal Reg-the concems of the bulletin have been resolved through design inter (Vol. 48, No.144) on July 26,1983. NUREG-1022, "Li-and procedure reviews and corrective actions taken by licens-consee Event Report System Description of Systems and ees Follow up items are proposed for use by NRC Region ll in Guidelines for Reporting." provides supporting guidance and in-closing the bulletin for the four (4) remaining facihties with open formation on the revised LER rule. The LER summaries in this bulletin status. NRC Region li venties that the bullet _n will not i
report are arranged alphabetically by facility name and then be closed for these four TVA facilities until construction is com-chronologica!Iy by event date for each facility. Component, a ground information is supplied in the introduction system, keyword, and component vendor indexes follow the summaries. Vendors are those identified by the utility when the NUREG/CR-4550 V3R1P1: ANALYSIS OF CORE DAMAGE LER form is initiated, the keywords for the component, system.
FREQUENCY;SURRY, UNIT 1, INTERNAL EVENTS.
and general keyword indexes are assigned by the computer BERTUCIO,R.C.; JULIUS,J.A. El Services, Inc.
- Sandia Nation-using correlation tables from the Sequence Coding and Search al Laboratories. April 1990. 484pp. 9006080175. SAND 86-2084.
System.
54078:003.
6 Main Citations and Abstracts i
This document contains the accident sequence analyses of Sequoyah was calculated to be the 5.7E 5 per year, with a 95 internalty initiated events for the Surry Nuclear Station, Unit 1.
percent upper bound of 1.8E-4 and 5 percent lower bound of This is one of the five plant anafyses conducted as part of the 1.2E 5 per year. Loss-of-coolant type accidents were the largest NUREG 1150 effort by the Nuclear Regulatory Commission.
contributors to core damage frequency, accounting for approxi-NUREG 1150 documents the nok of a selected group of nuclear mately 62 percent of the total. The next most dominant type of power plants. The work performed and described here is an ex.
accidents were station blackout (loss of all AC power), which tensive reanalysis of that pubhshed in November 1986 as account for 26 percent of core dama0e frequency NUREG/CR-4550. Volume 3. It addresses comments from nu-morous reviewers and significant changes to the plant systems NUREG/CR 4550 V7 R01: ANALYSIS OF CORE DAMAGE FRE-and procedures made since the first report. The uncertainty OUENCY: ZION, UNIT 1, INTERNAL EVENTS. SATTISON.M.B.:
analysis and presentation of results a o also much improved.
HALL K.W. EL &G Idaho, Inc. (subs. of EG&G, Inc.). May 1990.
The context and detail of this report are directed toward PRA 402pp. 9006291164. EGG 2554. 54325:223.
practitioners who need to know how the work was performed This documeit contains the accident sequence analyses of and the details for use in further studies. The mean core intemally initiate 1 events for the Zion Unit 1 Nuclear Power damage frequency at Surry was calculated to be 4.0E 5 per Plant. This is oro of the five plant analyses conducted as part year, with a 05% upper bound of 1.3E-4 and 5% lower bound of the NUREG-150 crfort for the Nuclear Regulatory Commis-of 6.8E 6 per year. Station blackout type accidents (loss of all sion. The work p Wormed and described here is an extensive AC power) were the largest contributors to the core damage fre-reanalysis of the work published in October 1986 as NUREG/
quency, accountin0 or approximately 68% of the total. The CR-4550 Volume 1, it addresses comrnents from numerous re-f next type of dominant contributors were Loss of Coolant Acci-viewers and prog des significantly more detailed modelirig of dents (LOCAs). These sequences account for 15% of core most aspects of he Zion plant. The mean core damage fre-damage frequency. No other type of sequence accounts for quency et Zion was calculated to be 3.4E 4 per year, with a more than 10% of core damage frequency.
95% upper bound of 8.4E 4 per year and a 5% lower bound of 1.1E-4 per year. Note that this range ia based only on the statis-NUREG/CR-4550 V3RtP2: ANALYSIS OF CORE DAMAGE tical treatment of six specific issues. Reactor coolant pump seal FREQUENCYSURRY, UNIT 1. INTERNAL EVENTS APPENDI.
loss-of coolant accidents (LOCAs) were the largest contributors CES. BERTUCIO,R.C.; JULIUS.J.A. El Services, Inc.
- Sandia to the core damage frequency, accounting for approximately National Laboratories. April 1990. 702pp. 9006080169.
85% of the total.
SAND 66-2084. 54079:127, See NUREG/CR 4550,V03,R01,P01 abstract.
NUREG/CR 4639 V01 R1: NUCLEAR COMPUTERIZED LIBRARY NUREG/CR-4550 V5R1P1: ANALYSIS OF CORE DAMAGE FRE-FOR ASSESSING REACTOR RELIABILITY OUENCY:
SEOUOYAH, UNIT 1,1NTERNAL EVENTS.
(NUCLARR). Summary Desenption.
GERTMAN,D.L; BERTUCIO,R C.: BROWN,S.R. El Services, Inc.
- Sarda Na-GILMORE W.E.; GALYEAN,W.J.; et al. EG&G Idaho, Inc. (subs.
tional Laboratories. Apnt 1990. 422pp. 9005210262. SAND 66-of EG&G, Inc.). May 1990. 29pp. 9006080202. EGG 2458.
[
2084. 53890:118.
54074:042.
This document contains the accident sequence anafyses of The Nuclear Computerized Ubrary for Assessing Reactor Re-intemally initiated events for the Sequoyah, Unit i nuclear liability (NUCLARR) is an automated data base management power plant. This is one of the five plant analyses conducted as system for storing and processing human error probability and part of the NUREG 1150 effort by the Nuclear Regulatory Co*
hardware component failure rate data. The NUCLARR system mission (NRC). NUREG 1150 documents the nsk of a selected software resides on an IBM (or compatible) personal microcom-group of nuclear power plants. The work performed and de-puter, NUCLARR can be ucessed by the end user to fumish scribed here is an extensive reanalysis of that published in Feb-data suitable for input in human and/or hardware reliability anal-ruary 1987 as NUREG/CR 4550, Volume 5. It addresses com-ys a to support a variety of risk assessment activities. The NU-ments from numerous reviewers and significant changes to the CLARR system is documented in a five-volume series of re-plant systems and procedures made since the first report. The ports. Volume 1 of this series is the Summary Description, uncertainty analysis and presentation of results are also much which presents an overview of the data management system, improved. The mean core damage frequency at Sequoyah was including a description of data collection, data qualification, data calculated to be 5.7E 5 per year, with a 95 percent upper bound structure, and taxonomies. Programming activities, procedures i.
of 1.8E 4 and 5 percent lower bound of 1.2E 5 per year. Loss-for processing data, a user's guide, and hard copy data manual j
of coolant type accidents were the largest contributors to core are presented in Volumes 2 through 5, NUREG/CR-4639.
damage frequency, accounting for approximately 62 percent of the total. The next most dominant type of accidents were sta-NUREG/CR-4667 V07: ENVIRONMENTALLY ASSISTED CRACK-tion blackout (loss of all AC power), which account for 26 per-ING IN LIGHT WATER REACTORS. Semiannual Report, April, cent of core damage frequency.
September 1988. SHACK,W.J.; KASSNER,T.F; PARK,J.Y.; et NUREG/CR 4550 V5R1P2: ANALYSIS OF CORE DAMAGE FRE-al. Argonne National Laboratory. March 1990. 53pp.
QUENCY: SEQUOYAH. UNIT 1,1NTERNAL EVENTS APPENDI-9007120135. ANL-89/40,54470.208.
CES. BERTUCIO,R.C.; BROWN.S.R. El Services, Inc.
- Sandia This report summarizes work performed by Argonne National National Laboratones. April 1990. 686pp. 9005210142.
Laboratory on environmentally assisted cracking in light water SAND 86-2084. 53862:001.
reactors during the six months from April to September 1988.
This document contains the appendices for the accident so-The stress corrosion cracking (SCC) of Types 316NG and 304 quence analyses of internally initiated events for the Sequoyah, stainless steels (SSs) was Investigated by means of slow-strain.
Unit 1 nuclear power plant. This is one of the five plant analy-rate and fracture-mechanics crack growth rate tests in high-tem-ses conducted as part of the NUREG 1150 effort by the Nucle-perature water. The effects of load ratio and water chemistry on at Regulatory Commission (NRC). NUREG 1150 will document the crack growth behavior of Type 316NG and sensitized Type the risk of a selected group of nuclear power plants. The work 304 SS were determined in lon0-term fracture-mechanics tests.
performed and desenbod here is an extensive reanalysis of that The int;uence of organic impurtties on the SCC of sensitized published in February 1987 as NUREG/CR-4550, Volume 5. It Type 304 3S was also investigated. Fatigue tests were conduct-addresses comments from numerous reviewers and significant ed on Type 316NG SS in air and simulated boiling water reactor changes to the plant systems and procedures made since the water at 288 degrees C to assess the degree of conservatism first report. The uncertainty analysis and presentation of results in the ASME Code Section til fatigue design curves. An ongoing are a!so much improved. The mean core damage frequency at investigaton of the susceptibility of several heats of different l
i l
Main Citations and Abstracts 7
i grades of low-alloy femtic steels to transgranular SCC in slow-areas, thus effectively doubling the amount of seismic data strain-rate and fracture-mechanics tests was continued.
being received by the Ottawa data tab.
NUMEG/CR 4047 V00: ENVIRONMENTALLY ASSISTED CRACK
- NUREG/CR-4753 V04:
CANADIAN SEISMIC ING IN LIGHT WATER REACTORS.
Semiannual Report. October 1988 March 1969. KASSNER.T.F.; PARK,J.Y.;
AGREEMENT. Annual Report 1988-1989. WETMILLER,R.J.;
RUTHER.W.E.; et al. Argonne Natonal Laboratory. June 1990.
LYONS,J.A.; SHANNONW.E.; et at Canadian Commercial 53pp. 9007120144. ANL-90/4. 54468 287.
Corp. April 1990. 61pp. 9005210166. 53891:180.
This report summarizes work performed by Argonne Natonal In this time period eastem Canada experienced its largest Laboratory on environmentally assisted cracking in light water earthquake in over 50 years when a magnitude (M) 6.0 event reactors dunng the six months from October 1988 to March took place at 18:48 EST on Fnday, November 25 near 48.12 1989 The effects of load ratio on stress corrosion cracking degrees W just south of Chicoutimi, Quebec. This earthquake, (SCC) of Types 316NG,304, and CF 3M cast stainless steels which has been chnstened the Saguenay earthquake, has pro-(SSs) were investigated by fracture-mechanics crack growth-vided a wealth of new data pertinen; to earthquake engineering rate (CGR) tests in high temperature water. The influence of or*
studies in eastern North America and la the subject of many genic impurities on the SCC of Type 316NG SS was also inves-continuing studies. The bibliography gives a summary of the Sci-tigated in long-term CGR tests. Tests to determine the suscepti-entific reports on earthquake or related studies that have been bihty of 4-in. diameter Types 316NG and 304 SS pipe weld-published or submitted for publication by GD staff in this period, i
monts to SCC in simulated BWR environments have been cork NUREG/CR 4416: PR EDB: POWER REACTOR EMBRITTLE.
i ducted. The influence of carbonate at concentrations between MENT DATA BASE, VERS!ON 1.
Program Desenpton.
0.1 and 3300 ppm on the SCC behavior of sensitized Type 304 STALLMANN,F.W.; KAM.F.B.K.; TAYLOR.B.J. Oak Rdge Na-
}
SS in deoxygensted water (less than 5 ppb) was determined in tonal Laboratory. June 1990.104pp. 9006290154 ORNL/TM.
l constant extensiorw rate tensile (CERT) tests. Fatigue tests 10328, 54363:220, were conducted on Type 316NG SS in air and BWR envirork 1
monts to assess the degree of conservatism in the ASME Code Data concerning radaten embnttlement of pressure vessel l
Section 111 fatigue design curves. CGR tests to determine sus-steels in commercial power reactors have been collected from available surveillance reports. The purpose of this NRC-spon-ceptibility to SCO are being conducted on A533-Gr B low-alloy
- sorec program is to provide the technical bases for voluntary territic steelin simulated BWR environments.
consensus standards, regulatory guides, standard review plans, NUREG/CR 4491 V01: MELCOR ACCIDENT CONSEQUENCE and codes. The data can also be used for the explorebon and
}
CODE SYSTEM (MACCS) Volume ' 1: User's Guide, verification of embrittlement prediction models. The data files CHANIN,D1; SPRUNG,J.L.; RITCHIE LT.: et al. Sanda National are given in dBASE lll Plus format and can be accessed with Laboratories. February 1990. 507pp. 9005040033. SAND 86-any personal computer using the DOS operating system. Menu-1562. 54503.136, dnven software is provided for easy access to the data includ-This report describes the MACCS computer code. The pur-ing curve fitting and plotting facilities. This software has drasts-pose of this code is to simulate the impact of severe accidents cally reduced the time and effort for data processing and eval-at nuclear power plants on the surrounding environment.
uation compared to previous data bases. The current compila-MACCS has been developed for the U.S. Nuclear Regulatory tion of the Power Reactor Embrittlement Data Base (PR EDB, Commission to replace the previous CRAC2 code and it incor-version 1) contains results from surveillance capsule reports of porates many improvements in modeling flexibility in comparison 78 reactors with 381 data points from 110 different irradiated to CHAC2. The principal phenomena considered in MACCS are base materials (plates and forgings) and 161 data points from-atmospheric transport, mitigative actions based on dose projec-79 different welds. Results from heat-affected-zone materials tion, dose accumulation by a number of pathways including food are also listed. Electric Power Research institute (EPRI), reactor i
ed water ingestion, early and latent health effects, and eco-vendors, and utilities are in the process of providing back up nonic costs. The MACCS code can be used for a variety of ap-quality assurance checks of the PR.EDB and will be supple-plicaMns. These include (1) probabihstic risk assessment (PRA) menting the data base with additional data and documentation.
of nuceer power plants and other nuclear facilities, (2) sensitivi-Periodic updates of data and software will be released to au-ty studies 'o gain a better understanding of the parameters im-thorized users. Future updates will also include results from irra-portant to Ph.'
and (3) cost benefit analysis. This report is diations in materials test reactors, composed of three Omes. Volume 1, the User's Guide, de-scribes the input data requirements of the MACCS code and NUREQ/CR 5001: EFFECTS OF MANUFACTURING VARIABLES provides directions for its use as illustrated by three sample ON PERFORMANCE OF HIGH LEVEL WASTE LOW CARBON problems. Volume 2, the Model Description, descnbes the un.
STEEL CONTAINERS. FROST,R.H. Colorado School of Mines, dertying models that are implemented in the code, and Volume Golden, CO. MtfrH.T.R.; LIBY,A.L Manufacturing Sciences 3, tne Programmer's Reference Manual, describes the code's Corp. April 1990.129pp. 9005040070. 53847:242.
structure and database management.
Analytical and expenmental research was performed to detor-mine the effect of manufacturing variables on the performance NUREG/CR 4753 V03:
CANADIAN SEISMIC of cast steet overpacks. The work examines the influence of i
AGREEMENT. Annual Report-1987 1988. WETMILLER.R.J.;
casting and welding process variables on the long-term per-l LYONS.J.A.; SHANNON,W.E.; et al. Canadian Commercial Corp. April 1990. 41pp. 9005210158. 53887:048.
formance of low carbon steel overpacks in the repository envi.
During the period of this report, the contract resources were ronment. Centrifugal casting was indicated to be the most eco-spent on operation and maintenance of the Easte n Canada Te-nomical and technically favorable manufacturing approach for lemetred Network (ECTN), development of special purpose cast steel overpacks. A bottom would be welded into a hollow local network systems, servicing and maintenance of the cylinder to make the container and final closure welding to strong. motion seismograph network in eastem Canada, oper.
secure the lid would be done at the repository. Effects of alloy ation and of the Ottawa data lab and earthquake monitoring and chemistry, solidification processing, and solid state phase trans-reporting. Of special note in this period was the final completion formations on final microstructure of the cast and welded over-of the Sudbury (SLTN) and Charlevoix (CLTN) local networks pack has been examined in detail in this report. Codes and and the integration of their data processing and analysis re-standards goveming the manufacture of overpacks do not pres-quirements in the regular analysis stream for ECTN data. These ently exist. An extonsion of the ASME Boiler and pressure networks now acquire high quality digital data for detailed analy-vessel code supplemented by governrnent standards could be sis of seismic activity and source properties from these two adopted for the purpose. Standard, well established methods of non destructive evaluation are adequate for the purpose of i
^
8' Main Citations and Abstracts identifying hkely manufacturing defects. Experimer tal work fo-was developed that describes how CES can be used to provide cused on material and process combinat ons to be used in man-input to human rehability analyses (HRA) in probabilistic nsk as-ufacture of overpacks. Practical process hmits were explored sessment (PRA) studies. TNs report describes the results of and changes in microstructure due to repository thermal condi-three activitses that were performed to evaluate CES/ CREATE:
(1) A technical review was conducted by a panel of experts in tsons were investQated.
cognitive modeling, PRA and HRA (2) CES was exercised on NUREG/CR 5111: INTEGRATED RELIABILITY AND RISK ANAL.
steam generator tube rupture incidents for which data on opera.
YSIS SYSTEM (IRRAS) VERSION 2.0 USER'S GUIDE.
tm pedamance esst,,(3) A wmW wm W premoners RUSSELL,K.D.; SATTISON.M.B. EG&G Idaho, Inc. (subs. of was hew to anahze a wMed exaW, of me CREAM mem-EG&G, Inc.). RASMUSON.D.M. Probabilistic Risk Analysis odologw The resuus of an Wee evalushs Wate mat CES/
j Branch. June ' 1990. 400pp. 9007120205. EGG-2535.
CREATE is a promising approach for modeling intenbon forma-W70203 bon. Volume 1 provides a summary of the results. Volume 2 The Integrated Reliability and Risk Analysis System (tRRAS)
P6s MaHe on the three waluahs, Wng tM CES is a state of the-art, microcomputer. based probabilistic risk as.
canMer oup b me W We weMs.
sessment (PRA) model development and analysis tool to ad-dress key nuclear plant esfety issues. IRRAS is an integrated NUREG/CR-5213 V02: THE COGNITIVE ENVIRONMENT SIMU-software tool that gives the user the abihty to Create and ana*
LATION AS A TOOL FOR MODEL!NG HUMAN PERFORM-
. lyze fault trees and accident sequences using a microcomputer.
ANCE AND RELIABILITY. Main Report. WOODS,D.D. Ohio State This program provides functions that range from graphical fault Univ., Columbus, OH. POPLE,H.E. Pittsburgh, Univ. of, Pitts-
- tree construction to cut set generation and quantification. Also burgh, PA. ROTH,E.M. Westmghouse Science & Technology provided in the system rs an integrated full screen editor for use Center. June 1990. 69pp. 9007120162. 54488:091, when interfacing with remote mainframe computer systems.
See NUREG/CR-5213,V01 abstract.
l Version 1.0 of the IRRAS program was released in February of 1987. Since that bme, many user comments and tenhancements NUREG/CR 5253: PARTITION: A PROGRAM FOR DEFINING have been incorporated into the program providing a much THE SOURCE TERM / CONSEQUENCE ANALYSIS INTER.
more powerful and user.friendty system. TNs version has been
' FACE IN THE - NUREG 1150 PROBABILISTIC RISK designated IRRAS 2.0 and is the subject of this user's guide.
ASSESSMENTS. User's Guide. IMAN,R.L Sandia National Lab-Version 2.0 of IRRAS provides all of the same capabilities as oratories. ' HELTON.10. Arizona State : Univ., Tempe, AZ.
Version 1.0 and adds a relatonal data base facility for manag-JOHNSON,J.D. Science Applications-Intemational Corp. (for-ing the data, improved functionality, and improved algorithm per.
merly Science Appiscations, Inc.). May 1990.52pp.9006080142.
!ormance-SAND 88 2940,54075:318.
This document has been designed for users of the PARTI-NUREG/CR 5181: NUCLEAR PLANT AGING RESEARCH;THE TlON computer program developed by the authors at Sandia IE POWER SYSTEM. MEYER,LC.; EDSON J.L EG&G Idaho, National Laboratories for defining the interface between the Inc. (subs. of EG&G, Inc.). May 1990.102pp. 9006080179.
source term analysis and the consequence analysis. The pur-1 EGG-2545. 5407L216, pose of the PARTITION program is to form groups of source This report presents the results of a study of aging effects on terms with similar properties. One set of MACCS calculations is the Class 1E power system in nuclear power plants. The 1E performed for each of these groups. The following operations power system is the part of the plant auxiliary power system are performed in PARTITION: (1) an earfy fatality weight and a that supplies power to the safety systems. The purpose of this chronic fatality weight are defined for each source term, (2) the report is to evaluate the effects of aging caused by operation mce terms are parb%ned sto groups of source terms with within a nuclear facility and the effectiveness of maintenance, similar radiological potential on the basis of these weights and a testing, and monitoring on detecting and mitigating the effects smgle frequency-weighted source term is calculated for each -
of aging. The U.S. Nuclear Regulatory Commission's Nuclear source term group, (3) the source terms in each source term Plant Aging Research guidelines were followed in performing group are dvidedinto subgroups on the basis of evacumen the detailed study that identifies 1E power system components tirnng and a frequency-weighted source term is calculated for i
susceptible to aging, stressors, environmental factors, and fail.
each subgroup, (4) various summary plots are produced to aid ure modes. Testing, maintenance, codes and standards, and In checking the adequacy of the partitioning, and (5) an output l
regulatory issues are discussed. Degradation mechanisms, fail.
file that aerves as input to the consequence analysis is generat-ure modes and inspection, surveillance, and monitoring meth, ed. The result of the partitioning process is a subdivision of the ods are summartzed for major class 1E components. This report source terms on the basis of three dimensions: earty fatality po-also presents the results of a review of the 1E power system tential, chronic fatality potential, and evacuation timing.
operating experiences reported in Licensee Event Reports, the Nuclear Power Experience data base, the Nuclear Plant Reli-NUREG/CR-6258 V02: GEORGIA / ALABAMA REGIONAL SEIS-ability Data System, and plant maintenance records. Included in MOGRAPHIC NETWORK. Annual Report.Jufy 1986 - June 1987, this report are the alternating current system (4160 to 120 V),
LONG.LT. Georgia Institute of Technology, AUenta, GA. April the direct current system, and the vital 120 V ac Irntrument and 1990. 71pp. 9005210180. 53891:242.
control systm Data from continued operation of the setsmographic network NUREG/CR 5213 Vot: THE COGNITIVE ENVIRONMENT SIMU-were used for topical studies. A small swarm of earthquakes oc-LATION AS A TOOL FOR MODELING HUMAN PERFORM-curred in the Lake SinclairiGeorgia vicinity during the first quar-ANCE AND RELIABILITY. Executive Summary. WOODS,0.D.
ter of 1967; The events occurred in three main shocks of mag-Ohio State Urw., Columbus, OHJ POPLE,H.E. Pittsburgh, Ursv.
nitude 2 and their aftershocks. The third event had an unusually of, Pittsburgh, PA. ROTH.E.M. Westinghouse Science & Tech-rapid decay in aftershock activity, suggesting complete release nology Center. June 1990. 31pp. 9007120152. 54488:163.
of stress; Attenuation and scattering were studied on the basis The U.S. Nuclear Regulatory Comrnssion is sponsoring a pro-of theoretical evaluations of the effects of sphericalinclusions in L
gram to develop improved methods to model cognitive behavior elastic or viscoelastic media and on the basis of P wave coda.
i l
of nuclear power plant (NPP) personnel. A tool called Cognitive Resonance of the scatterer and viscosity of a solid material play Environment Simulation (CES) was developed for simulating an important role in determining scattering coefficients. Wave-l how people form intentions to act in NPP emergencies. CES lets comprising the P. wave coda show systematic increases provides an analytic tool for exploring plausible human re-and decreases in spectral peaks that may be explained by shifts sponses in emergency situations. In addition a methodology in the corner frequency caused by absorptrve attenuation and called Cognitive -Reliability ' Assessment Technique (CREATE) scattering. In locating hypocenters, the method used at Georgia w
- - ~-
lAaln Citations and Abstracts 9
Tech computes origin time Independent of location. Location function technique), and geostabstical techniques (stochastic i
and depth are then computed with a imod origin bme. When S modeling using Monte Carlo simulation and spectral analysis).
and P hmes are available, including off diagonal elements in the Advantagoc, esadvantages and applications of each technique
{
{
covariance matrix leads to locations that are more precise than those obtained with standard methods.
are presented. Propagation of uncertainties through multiple, linked models is also escussed. Applicabon of these techruques MUREQ/CR-5262: PRAMISi PROBABILITY RISK ASSESSMENT to sensWy aneWs is also presem SensMy analyses can MODEL INTEGRATION SYSTEM. User's Guide. IMAN,R.L.
be uWul to uncedeW sNdes because the ne of paraw JOHNSON,J.D.; HELTONJ.C. Sandia National Laboratories em incWed h N unmMnty anaWs can be WW W May 1990. 80pp. 9006080164. SAND 88-3093. 54077:012.
eliminating those parameters for wNch the uncertainty has a The document has been designed for users of the Probabihs-nmmal N on N pMormance vahaWsh bc Risk Assessment Model integrabon System (PRAMtS) com' puter program developed by the authors at Sandia National NUREG/CR 5397: VALUE lMPACT ANALYSIS OF REGULATORY Laboratories for easy assembly of the indevidual parts of the OPTIONS FOR RESOLUTION OF GENERIC !SSUE C-8.MSIV NUREG 1150 plant analyses into overall risk results. PRAMIS LEAKAGE AND LCS FAILURE. JAMISON J.D.; VO T.V.;
assembles the following files associated with the NUREG tt50 TABATABAIAS. Battelle Memorial Institute, Pacific Northwest Laboratory. May 1990. 53pp. 9007120218. PNL 6931, analyses in matrix format to obtain risk: the Latin hypercube sample, the results of the systems analysis, the results of the 54467:273.
accident progression analysis, the results of the source term /
TNs report descdbes the analysis conducted to establish the partitioning analysis, and the results of the consequence analy' basis for answering two remaining regulatory questions facing sie, in addition, various intermedate and conditional quantities the NRC staff regarding the resolution of Generic issue C-8,-
are calculated when requested by user.specifed input; the trac-specifically 1) What action should the NRC take concoming I
tsonal contribution to risk of individual plant damage states, acci-plants that currentty have a leakage control system (LCS)? 2) dont progression bins and source term groups are determined, What action should the NRC take concemeng plants that do not and a file containing the original Lahn hypercube sample and have an LCS? Using individual MSIV leak test data, the per-user specified dependent vanables is generated for use as input formance of a system of eight such vanes in a standard BWR to the SAS stabstical package. This report provides a tutorial configurr%n was modeled. The performance model was used that details how to use the PRAMIS program. The PRAMIS pro-along with estimates of core damage accident frequency and gram is written in ANSI standard FORTRAN 77 to make the calculated dose consequences to determine the public risk as-code as machine-independent (that is, portable) as possible*
sociated with each of the attematives. The occupabonal expo-sure imphcations of each attemabve were calculated using esti-NUREG/CR 5366: HTAS2: A THREE DIMENSIONAL TRANSIENT mates of labor hours in radiation zones that would be incurred SHIPPING CASK ANALYSIS TOOL WENDEL,M.W.; GILES,G.E.
or avoided. The costs to industry of implementing each alterna.
Oak Ridge National Laboratory. May 1990. 81pp. 9006290147, tive were estimated using standard cost formulae and NRC staff ORNL/CSD/TM-267. 54324:312.
This report desenbes the HTAS2 computer program which estimates. The costs to the NRC were estimated based on the can be used to assess the thermal behavior of shipping casks etlort incurred or avoided for reviews or other staff actions en-containing PWR type fuel assemblies. HTAS2 has two parts: a gendered by the selection of a particular attemative. The costs and risks thus calculated suggest that no regulatory action can global cask analysis and a single assembly analysis. The global be justified on the basis of risk reduction or cost savings, cask analysis is a three dimensional lumped parameter model for the entre shipping cask and its contents. The user has the NUREG/CR-5399: SURVEY OF STATE AND TRIBA'L EMERGEN-option of simulating a prefire steady state, a fire transient, a CY RESPONSE CAPABluTIES FOR RADIOLOG!OAL TRANS-postfire transient, or a final steady state in acceptable order.
PORTATION INCIDENTS.
VILARDO,F.J.;
MITTER,E.L; Details within the fuel assemblies are not resolved in this por-PALMER.J.A.; et al. Indiana Univ., Bloomington, IN. May 1990.
tion of the analysis. The single assembly analysis is two-dimen.
230pp. 9006290123. 54322-294.
sional and contains a detailed radiation and conduction model This publication is the final report of a project to survey the which can be used to represent PWR square pitched fuel as.
fifty states, the District of Columbia, Puerto Rico, and selected sembhes A simple convection modelis available. The boundary indian tribal jurisdictior's to ascertain their emergency-prepared-conditions on the basket walls surrounding the assembly can be ness planning and capabilities for responding to transportahon directty specified or may come from a previous global cask incidents invoMng radioactive materials. The survey was con-i analysis. Good companson was obtained with other computa-ducted to provide the Nuclear Regulatory Commission and tiorri resutts for both parts of the analysis. Although good other federal agencies with information conceming the current agreement was found between the single assembly results and level of emergency-response preparedness of the states and expenmental data, the global cask analysis has not yet been selected tribes and an assessment of the changes that have vahdated against expenments; this task is recommended as a occurred since 1980 (when'a similar survey was performed part of future work.
[NUREG/CR 1620]). There have been no major changes in the NUREG/CR-5393: A REVIEW OF TECHNIQUES FOR PROPA-states' emergency response planning strategies and field tac-GATING DATA AND PARAMETER UNCERTAINTIES IN HIGH.
tics. The changes noted included an increased availability of LEVEL RADIOACTIVE WASTE REPOSITORY PERFORMANCE-dedicated emergency response vehicles, wider availability of ASSESSMENT MODELS. ZIMMERMAN,D.A.; WAHL,K.K.;
specialized radiation-detection instruments, and higher propor-GUTJAHR.A.L; et al. Sandia National Laboratories. March tions of police and fire personnel with training in the handling of 1990.109pp. 9007120186. SAND 89-1432. 54470 014.
suspected radiation threats. Most Indian tribes have no capabil-Techniques for propagating data and parameter uncertainties Ity to evaluate suspected radiation threats and have no formal in high-level waste (HLW) repository performance assessment relabons with emergency response personnelin adjacent states.
models are discussed. Uncertainty analysis techniques ascribe For the nation as a whole, the incidence of suspected radiation i
Quantitabve measures of reliability to model predictions. Both 10 threats declined substantially from 1980 to 1988.
CFR Part 80 and 40 CFR Part 191 require consideraton of un-NUREG/CR 5409: NEUTRON EXPOSURE PARAMETERS FOR certainties, including uncertainties in data and parameters, in THE METALLURGICAL TEST SPECIMENS IN THE SIXTH the performance assessment of an HLW repository system.
HEAVY SECTION STEEL 1RRADIATION SERIES. MILLER,LF.;
Four categories of uncertainty analysis methods are discussed:
Monte Carlo simulation, replacement models (response surface BALDWIN,C.A.; STALLMANN,F.W.; et al. Oak Ridge National techniques), differential techniques (direct, adjoint, and Green's Laboratory. May 1990. 49pp. 9005210206. ORNL/TM 11267.
53891:314.
i
/
10 Main Citations and Abstracts The goal of the Heavy Section Steel irradation (HSSI) Pro-classified in terms of environment, context, organizational gov-(
gram Sixth irradetion Sones is to determine the effect of irre&a-ernance, organizational design, and emergent process. Initial tion on the shape and shift of the crack arrest toughness versus empirical analyses were conducted on a hmited set of hypoth-j temperature curve. Two capsules which contairW crack-arrest sees dortved from the framework. One set of hypotheese con-and Charpy V notch test specimens have bee". rradated at the corned the relationships betwoon efficiency, measured by criti-Oak Ridge Research Reactor located at the Oak Ridge National cet hours and outage rate, and safety, measured by 5 NRC Indi -
Laboratory, These capsules have bea", disassembled, intomal cators. Results of the analysis suggest that critical hours and l
dosimeters have been analyzv. ed exposure parameters are outage rates and safety, as measured in this study, are not re-presented for each irradiahon test specimen. This report de- -
lated to each other. Hypotheses were tested concoming the of-scribes the computational methodology for the least-squares facts on safety and efficiency of utility financial resources and adjustment of the dosimetry data with neutronics calculations, the legged twsy,r,a6 and correction of problems that accom-and presents exposure parameters at each test specimen loca-panies the reporting of major violations and licensee event re-tion for the fluence rate greater than 1.0 MeV, fluence rate ports. Results suggest that both financial resources and organi-greater than 0.1 MeV, and displacements por atom. The specific rational problem solving /leaming have segrecent effects on the actMty of each dosimeter at the end of irradiation is listed in the outcome variables when time is property taken into account.
Appendix.
NUREG/CR M34: BASIC CONSIDERATIONS IN PREDICTING NUREG/CR-5411: ELICITATION & USE OF EXPERT JUDGMENT ERROR PROBABILITIES IN HUMAN TASK PERFORMANCE.
IN PERFORMANCE ASSESSMENT FOR HIGH-LEVEL RADIO-FLEISHMAN,E.A.; BUFFARDI,L.C.; ALLEN,JA: et al. George ACTIVE WASTE REPOSITORIES. BONANO,E.J. Sandia Natiork Mason Univ., Fairfax, VA. April 1990. 49pp. 9005040061, CBCS al Laboratories. HORA,S.C. Hawall, Univ, of, Hilo, Hl.
RPP #90153622:022~
KEENEY,R.L; et al. Southom Califomia, Univ, of, Los Angeles H is W established that human error plays a major role in CA. May 1990. 94pp. 9006080176. SAND 891621, 54075:122.'
the mal'unctioning of complex systems. This report takes a This report presents the concept of formalizing the elicitation broad look at the study of human error and addresses the con '
and use of expert judgment in the performance assessment of is e hvoW in high. level radioactive waste (HLW) repositories in deep geologic D9 W "9
bc in addnion, a formations. The report begins with a discussion of (1) character.
review of exisung sources of human r@ dah and ap istics (advantages or disadvantages) of formalizing expert judg-proaches to human performance data base development is pre-
_{
ment, (2) examples of previous uses of expert judgment in re, sented, Altemative task taxonomies, wNch are promising for es-dioactive waste programs, (3) criteria that can assist in deciding tablishing the comparability on nuclear and non-nuc6 ear tasks, when to formalize expert bdgment, and (4) the relationship of are also identified. Based on such taxonomic schemes, various formal use of expert judgment to data collection and modeling.
data base prototypes for generalizing human error rates across The current state of the art with respect to the elicitation, use, and communication of formal expert judgment is presented. The settings are proposed report concludes with a disdussion on potential applications of NUREQ/CR-5439: HUMAN FACTORS ISSUES ASSOCIATED formal expert judgment in performance assessment of HLW re-WITH ADVANCED INSTRUMENTATION AND CONTROLS E "u "**
TECHNOLOGIES IN NUCLEAR PLANTS. CARTER,R.J.;
NUREG/CR 5436 V01: THE DEVELOPMENT AND EVALUATION UHRIG R.E. Oek Ridge National Laboratory. June 1990.208pp.
OF PROGRAMMATIC PERFORMANCE INDICATORS ASSOCI-9007120224, ORNL/TM-11319. 54467:071.
ATED WITH - MAINTENANCE AT. NUCLEAR POWER A survey ~ of advanced instrumentation and controls (l&C)
PLANTS. Main Report. - WREATHALL,J.:
FRAGOLA,J.:
t6chnc0@s and associated human factors issues in the U.S.-
APPIGNANI,P.; et al. Soence Applications international Corp.
and Canadian nuclear industnes was carried out;The purpose 1
(formerly Science Applications, Inc.). May 1990. 133pp.
of the survey was to provide background for the development of 9006080157, SAIC-90/1130. 54076:239.
regulatory policy, criteria, and guides for review of advanced This report summarizes the development and evaluation of l&C systems as well as human engineenng guidelines for evalu-q programmabc performance indicators of maintenance. These in-ating these systems. The survey found those components of dicators were solected by- (1) creating a formal framework of the U.S. nuclear industry surveyed to be quite Interested in ad-plant processes; (2) identifying features of plant behavior con-vanced l&C, but very cautious in implementing such systems in sidered importarit to safety; (3) evaluating existing indicators nuclear facilities and power plants. The trend h the facilities against these features; and (4) performing statistical analyses surveyed is to experiment cautiously when t%re is an intuitive for the selected indicators. The report recommends additional advantage or short term payoff. The mrd advanced 14C sys-testing.
tems were found in the Canadian CANDU plants, where the newest plant nas 6L, ; ay.;.ms in almost 100% of its control NUREG/CR 5436 V02: THE DEVELOPMENT AND EVALUATION systems and in over 70% of Ms plant protech syskm. The hp OF PROGRAMMATIC PERFORMANCE INDICATORS ASSOCl-pothesis that propedy " introducing dignal syskms messes ATED WITH MAINTENANCE AT NUCLEAR ' POWER safety" is supported by the Canadian experience. A number of
.j PLANTS. Appendices.
WREATHALL,J.;
FRAGOLA,J.;
safety related human factors issues were derived from the re-APPIGNANI,P.; et al. Science Applications international Corp.
suits of this survey. They include: Is an advanced l&C guideline (formerly Science Apphcations, Inc.). May 1990. 225pp, equivalent to NUREG-0700 needed? What changes will there 9006080152. SAIC 90/1130. 54076:014.
be in the role of the control room operator? The potential prob-See NUREG/CR-5436,V01 abstract.
lem of information overload needs to be addressed. How should NUREG/CR-5437: ORGANIZATION AND SAFETY IN NUCLEAR existing training technology be made applicable for advanced j'
POWER PLANTS MARCUS,A.: NICHOLS M.: BROMILEY,P,; et t&C7 How will operator acceptance and trust be accomplished 7 al. Minnesota, Univ. of, Minneapolis, MN. May 1990. 214pp.
9006080166. 54077:098.
NUREG/CR 5449:' DETERMINATION OF THE NEUTRON AND Perspectives from industry, academe, and the NRC are GAMMA FLUX DISTRIBUTION IN THE PRESSURE VESSEL brought together to develop a logical framework that linlm orga-AND CAVTTY OF A BOILILNG WATER REACTOR. ASGARIM.;
nization factors and safety in nuclear power plant performance.
WILLIAMS,M.L; KAM.F.B.K. Oak Ridge National Laboratory.
The framework focuses on intermediate outcomes which can be June 1990.115pp. 9006290142 ORNL/TM-11350. 54325:036.
predicted by organizational factors, and which are subsequently The Grand Gulf Boiling Water Reactor (BWR/6), owned and linked to safety. The intermediate outcomes are efficiency, com-operated by Mississippi Power & Light Company has been ana-pliance, quality, and innovation. The organization factors can be fyzed to determine the neutron and gamma energy spectrum
- )
. ~.
(
Main Citations and AbstrPets 11 and flux levels in regions from the reactor vessel throughout the and there were indications that other sorption processes were concrete shield wall Several two dimensional and one demon-involved.
sional transport calculabons were performed for the Grand Gulf reactor configuration. The results from these calculations were NUREG/CR 6444: OADS: A MULTIDIMENSIONAL POINT synthesized to obtain the three-dimenseonal neutron flux spectra KERNEL ANALYSIS MODULE. BROADHEAD.B.L Oak Ridge and dosimeter activities. The results from the transport calcula-Natsonal Laboratory. May 1990. 66pp. 9006080320. ORNL/
hons indicate the flux above 1 MeV peaks near the axial mid-CSD/TM-270. 54075:054.
plane and azimuthal angle between 40 degrees and 45 de-OADS is a multedimensional point kernel computer code that grees, depending on the radial locahons. The peak flux above 1 utilizes the simplified free-form input of the SCALE system as MeV incident on the vessel and at midcavity is about 1.82 x well as compatibility with ORIGEN.S produced sources, SCALE 10(9) and 1.07 x 10(8) n x cm( 2) x s( 1), respectively. The cross section hbraries, and standard compositson data sets.
vessel fluence accumulated during Cycle 2 and after 32 offec.
CADS consists of a preprocessor that takes the free-form input live full power years is about 4.41 x 10(16) and 1.84 x 10(18) n and prepares input for the widely available OAD-CGGP code, x cm( 2) x s( 1), respectivey. The peak flux above 1 MeV at the which is then automatcally executed by a driver module. This front of the concrete shield well,15.24 cm (6 in.) into the con.
report desenbes the point kernel theory bnefly, followed by nu.
crete wall, and 30.48 cm (1 ft) Into the concretc well is about merous tips on successfully applying the theory to various types 7.91 x 10(7), 7.24 x 10(6), and 6.44 x 10(5) n x cm( 2) x s( 1),
of shielding problems. The remainder of the document is devot-respectivety. The results obtained from the gamma calculahons ed to input and output desenptions of the OADS code with sev-show that the peak gamma heating at the O T location of the oralillustrative sample problems.
s s
(S p k ga a
NUREG/CR 6473: INCLUSION OF UNSTABLE DUCTILE TEAR.
at the midcavity is about 7.31 x 10(3) rad /h at full power oper-ING AND EXTRAPOLATED CRACK ARREST. TOUGHNESS ation.
DATA IN PWR VESSEL INTEGRITY ASSESSMENT, NUREG/CR 5440 A CAUSE DEFENSE APPROACH TO THE UN-DICKSON,T.L; CHEVERTON,R.D.; SHUM.D.K Oak Ridge Na-tional Laboratory. May 1990. 35pp. 9007120177. ORNL/TM-DERSTANDING AND ANALYSIS OF COMMON CAUSE FAIL.
11450, 54470:164.
URES. PAULA,H.: CAMPBELL,D.; et al. JBF Associates, Inc, Over the past several years, the Heavy-Section Steel Tech-PARRY,G. NUS Corp. March 1990. 137pp. 9005020340.
nology Program at Oak Ridge National Laboratory has per.
SAND 89-2368. 53590:038.
This report presents the results of research to develop a new formed a series of large-scale fracture-mechanics experiments.
methodology for common cause failure analysis and prevention.
Thess experiments have demonstrated that prototypical nuclear Common cause failures are defined as those caused by single reactor vessel steels can exhibit crack-arrest toughness values considerably above 220 MPa a ftnalthough urrest can be followed ^
events that can make redundant components unavailat; $. thus immediately by unstable ductile tearing. This report evaluates compromising the rehability of redundant trains of equbpment.
the influence of the crack strest toughness above 220 MPa * (m Probabiliste nok assessments (PRAs) and nuclear power plant (NPP) operations have demonstrated that CCF events are often on the integrity assessments of nuclear reactor pressure ves-major contributors to the potential risk posed at such plants, sels for pressurized-thermal shock (PTS) loading conditions, Over the years, quantative and quantitatue anatysis methods taking into account the potenbal for unstable ductile tearing fol.
lowing arrest. The influence of the high crack-arrest toughness have been developed to improve CCF analysis. However, these data and unstable ductile tearing on pre surtzed water reactor methodologies have not explicitly accounted 1or the impact of plani specific defenses, such as design features and operational vessel integrlty assessment is PTS transient dependent. It ap-and maintenance policies, in reducing the likelihood of failures pears that the potential benefit from crack-arrest events corre-at NPPs. The research documented in thir, report describes a sponding to toughness values above 240 MPa * (m for low upper-shatf weld (LUSW) material and above 370 MPa * (m for those cause-defense methodology for CCF analysis and prevention to vessels not containing LUSW motorial will ust.alty be negated help correct this deficiency. The report discusses the develop-by uns'able ductile tearing.
ment of (1) procedures for identifying the potential for CCF events at individual NPPs and (2) cause-defense matrices for NUREG/CR 5475: MODEL FEASIBILITY STUDY OF RAD (OAC-anatyzing CCF events. New concepts and more precise defin;.
TIVE PATHWAYS FROM ATMOSPHERE TO SURFACE tions are introduced to enhance CCF terminology and interpre.
WATER. SMITH,R.E.; SUMMER,R.M.: FERREIRA.V.A. Agricul-tation of historical event data.
ture, Dept. of. March 1990. 43pp. 9005020324. 53588:328.
NUREQ/CR 5443: EFFECTS OF MINERALOGY ON SORPTION A feasibility study of the atmosphere to surface water radio-OF STRONTIUM AND CESIUM ONTO CALICO HILLS TUFF.
nuchde pathways was performed for small catchments using a MEYER R.E.; ARNOLD,W.D.; CASE,F.I.: et al. Oak Ridge Na-physically based hydro-ecosystem model, Opus.. Detailed time-intensity (breakpoint) precipitation records from Arizona and tional Laboratory. April 1990. 42pp. 9005040072. ORNL 6589.
Georgia were used as input to drive the model. Tests of model 53622:071.
Sorption and desorption measurements were made of stronti-sensitivity to distnbution coefficients, Kd, for Cs 137. Cs 134, and Sr-90 illustrated different vegetation-soil-erosion-runoff um and cesium onto clinoptitolite and Calico Hills Tuff. The object was to see whether there was a correlation between pathways in response to agricultural management practices. Re-sorotion of strontium and cesium onto Cahco Hills Tuff and the suits reflected the fact that low Kd values allow a radionuclide sorption of strontium and cesium onto chnoptilolite based on the to infiltrate into the soil profile and isolate it from subsequent content of clinoptilolite in the Cahco Hills Tuff. Il sorption onto runoff and erosson. Of the radionuclides and physical settings Calico Hills Tuff is solely due to the presence of clinoptilohte, studied, only the Sr 90 with low Kd values is sufficiently mobile and long-lived to be removed from the system via percolation then the ratios of the sorption ratios on tuff to those on clinopti-below the root zone. Conversely, highly-adsorbed radionuclides lolite at similar conditions should be the weight fraction of the clinoptilolite on the tuff. Since the tuff contained about 50%
were subject to removal by adsorption to sediment particles and subsequent runoff. Comparison of different effectsve half. lives of chnoptilohte, the ratios would be expected to be about 0.5 li 1-131 demonstrated the importance of the timing of an erosion-sorption was due solely to clinoptilohte. The experimental evi-dence showed that the ratios were generally near 0.5 for both runoff storm event during or immediately after a fallout event.
cesium and strontium sorption and that looexchange processes Seasonal timing of a fallout event and crop management also were operative for both the clinoptilohte and the tuff. However, affect the fate of this short-lived radionuclide, Removal by solu-tion to surface water runoff was negligible for all nuclides stud-the ratios differed to a small extent for the different conditions, led. Model simulation results for up to 10 half-lives are corrobo-
-i 12 Wain Citations and Abstracts rated by results from long. term field studies. These results show toughness is required to provide confidence that safety margins the feasibility of modeling pathways in small catchments using do not fall below assumed levels. To assess this behavior, com-parisons of AT's defined by elastic-plastic fracture toughness and Opus.
C(v) tests have been made using data from RPV base and weld r
NUREG/CR 5440: DATA
SUMMARY
REPORT FOR FISSION metals in which irradishons were made under test reactor condi-PRODUCT -RELEASE- -TEST VI-3.
OSBORNE,M.F, LORENZ,R.A.t COLLINS.J.L; et al. Oak Ridge National Laboral,
.,3ng,'
t tory. June 1990. 70pp. 9006290135. ORNL/TM-11399.
NUREG/CR.5510: EVALUATIONS OF CORE MELT FREOUENCY 54325:151, EFFECTS DUE TO COMPONENT AGING AND MAINTE.
Test VI-3, the third in a series of high-temperature fission NANCE. VESELY,W.E.; KURTH.R.E.; SCALZO,S.M.- Science product release tests in the vertical test apparatus, was con-Applications Intemational Corp. (formerly Science Applications, ducted in flowing steam. The test specirnen was a 15.2 cm-long Inc.). June _~ 1990. 225pp. 9007120191. SAK;-89/1744.
section of a fuel rod from the BR3 reactor in Belgium, which 54486:001, had been irradiated to a burnup of 42 mwd /kg. Using an induc-A methodology is developed to incorporate aging effects into tion furnace, it was heated under simulated LWR accident con-Probabilistic Risk Analyses (PRAs). The methodology separates dit<ms to two test temperatures,20 min at 2000 K and then 20 the PRA analyses from the aging analyses, allowing available man at 2700 K. The cladding was completely oxidized dunng the.
PRAs to be efficientty used in evaluating risk effects of aging.
test, and very little melting or fuel cl adding interaction had oc-The methodology was applied to two NUREG-1150 PRAs using.
curred. Based on fission product inventories measured in the aging rate data that was developed for active Components. Vari.
fuel or calculated by ORIGEN2, analyses of test components ous surveillance and maintenance programs were evaluated to showed total releases from the fuel of 100% for Kr 85,6% for determine their effects in controlling aging. Both point evalue-Ru-106,99% for Sb 125, and 99% for both Cs 134 and Co-137.
tions and uncertainty evaluations were carried out. The results Small release fractions for many other fission products were de-of the applications showed the sensitivity of aging effects on i
tected. In addition, very small amounts of fuel material. uranium core melt frequency to the efficiency of the maintenance and and plutonium. were released. The total mass released from surveillance program in managing aging effects. The detailed the fumace to the collection system was 3.17 g,78% of which contributors to the aging effects showed relatively few compo-was collected on the filters. The results from this test were nents contributing, implying that prioritized aging management compared with previous tests in this series and with a common-programs would be most effective in controlling risk.
ly used model for fission product release, NUREG/CR-5613 V01: AOCIDENT-MANAGEMENT INFORMA-NUREG/CR-5489: BIOLOGICAL CHARACTERIZATION OF RADi*
TION NEEDS. Volume 1 Methodology Development And Appfi-ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED cation To A Pressurized Water Reactor (PWR) With A Large, URANIUM MILLING EFFLUENTS. ElDSON,A.F. Inhalation Toxi-Dry Containment. HANSON,D.J.t WARD L.W.; NELSON.W.R.; et.
cology Research Institute. June 1990. 82pp. 9007120243. LMF*
al. EG&G Idaho, Inc. (subs. of EG&G, Inc.). April 1990. 69pp.
124. 54468:105.
9005070008. EGG 2592. 53656:277, Protection of uranium mill workers from occupational exp>
In support of the U.S. Nuclear Regulatory Commission (NRC) sure to uranium through routine bioassay programs and the as-Accident Management Research Program, a methodology has
'(
sessment of accidental worker exposures are addressed. Com-been developed for identifying the plant information needs nec-pansons of chemical properties and the biological behavior of essary for personnel involved in the management of an acci-refine uranium ore (yellowcake) are made to identify important dent to diagnose that an accident is in progress, select and im '
properties that influence uranium distribution among organs.
. piement strategies to prevent or mitigate the accident, and mon-These studies will facilitate calculations of organ doses after ex-itor the effectiveness of these strategies. This report describes posures and associated health nok estimates and will identify-the methodology and presents an application of this methodolo-important bioassay procedures to improve evaluations of human gy to a pressurized water reactor (PWR) with a large dry con-exposures. Samples of airbome uranium from operating mills tainment. A risk-important severe accident sequence for a PWR and deposition mode!s were used to predict appreciable deposii is used to examine the capability of the existing measurements tion in the upper respiratory tract of workers, if respiratory pro-to supply the necessary information. The method includes an tection were not used. Laboratory analyses of commercial yel-assessment of the effects of the sequence on the measurement lowcake, and inhalation studies in rats, showed that inhalation availability including the effects.of environmental conditions.
+
of yellowcake aerosols might be considered to be inhalation of The information needs and capabilities identified using this ap-variable mixtures of ammonium diuranate and U(3)O(8). Studies proach are also intended to form the basis for more compre-of yellowcake clearance from rats after wound contamination hensive information needs assessment performed during the showed that uranium behavior in vivo could not be quantitatively analyses and development of specific strategies for use in acci-related to chemical composition. A biokinetic model of yellow-dent management prevention and mitigation.
cake inhaled by. Beagle dogs was developed. Comparison with available data from human exposures showed that organ bur.
NUREG/CR-6513 V02: ACCIDENT MANAGEMENT. INFORMA.
I Appendices. HANSON.D.J.;
dens in an exposed worker can be estimated from unnary bio.
- TION NEEDS. Volume 12 assay results and in vivo countingiif the chemical composition, WARD,LW.; NELSON,W.R.; et.al. EGSG Idaho, Inc. (subs. of or soluble fraction, of the inhaled yellowcake is known.
EG&G, Inc.). April 1990. 304pp. - 9005040140. EGO-2592.
NUREG/CR+5494: CORRELATION OF IRRADIATION-INDUCED.
NU EG/CR.5513,V01 abstract.
TRANSITION TEMPERATURE INCREASES FROM C(V) AND K(JC)/K(IC) DATA. Final Report. HISER,A. Materials Engineering NUREG/CR-5614: MODELING AND-PERFORMANCE OF THE Associates, Inc. March 1990. 224pp. 9005020331, MEA.2377.
MHTGR REACTOR CAVITY COOLING SYSTEM. CONKLIN.J.C.
53500 175.
Oak Ridge Natonal Laboratory. April 1990. 36pp. 9005210212.
Reactor pressure vesset (RPV) surveillance capsules contain ORNL/TM-11451. 53863:327.
Charpy-V (C(v)) specimens, but many do not contain fracture The Reactor Cavity Cooling System (RCCS) of the Modular
, toughness specimens; accordingly, the radiation-induced' shift High-Temperature Gas Cooled Reactor (MHTGR) proposed by (increase) in the brittle.to-ductile transition region (AT) la based the U.S. Department of Energy is designed to remove the nucle-upon the AT determined from notch ductility (C(v)) tests. Since ar afterheat passivefy in the event that neither the heat trans-the ASME K(Ic) and K(Ir) reference fracture toughness curves port system nor-the shutdown cooling circulator suosystem is are shifted by the AT from C(vl, assurance:that this AT does not available. A computer dynamic simulation for the physical and underestimate AT asnociated with the actual irradiated fracture mathematical modeling of an RCCS is desenbed here.1wo con-t
)
' Main Citations and Abstracts 13 clusions can be made from computatone performed under the NUREG/CR-6840: /ERFORMANCE TESTING OF EXTREMITY assumpton of a uniform reactor vessel temperature. First, the DOSIMETERS, STUDY 2.
HARTY,R.;
REECE W.D.;
' heat transferred across the annulus from the reactor vessel and HOOKERC.D. Battelle Memorial institute, Pacific Northwest then to ambient conditions is very dependent on the surface Laboratory. April 1990. 89pp. D005040084. PNL 7276.
emissivities of the reactor vessel and the RCCS panels. These 53646:319.
ermssivities should be perio$cally checked to ensure the safety The second of two performance tests of extremity dosimeters function of the RCCS. Second, the heat transfer from the reac-was conducted using a draft of a proposed standard for extrem-tor vessel is reduced by a maximum of 10% by the presence of ity dosimeter performance testing. The draft standard was writ-i steam at 1 atm in the reactor cav'ty annulus for an assumed ten by the Health Physics Society Standards Committee constant reactor vessel temperature of 500 K. Thus, the pres-
_ (HPSSC) Working Group on the Performance Testing of Extrem-1 once of a medium that partcipates in the transmission of radi-Ity Dosimeters. The inibal performance test study (reported in ant energy across the annulus can be expected to result in an NUREG/CR-4959) indicated that approximately 60% of the time increase in the reactor vessel temperature for the MHTGR. Fur-the processors met the performance criterion specified in the
)
ther investigation of participathg radiation media, including draft standard. Because of these results, an investigation con-small particles, in the reactor cavtti annulus is warranted, ducted to determine the sources of error during the first per-J formance test indicated that errors occurred as a result of poor NUREG/CR 6617: IMPACTS.BRC,VER0lON 2.0. Program User's procedures, equipment malfunctions, and because processors Manual. O'NEAL,B.L; LEE,C.E. San & National Laboratories, were not prepared for the tests. A second performance test re-April 1990. 244pp. 9005040021. SAND 8b3060. 53649:308.
sutted in a passing rate of approximately 70% for ring dosi-This manual describes the procedures for implementing IM-meters and 81% for wrist dosimeters. Although this is an overall PACTS-BRC Version 2.0. The 1MPACTSORC computer code improvement, the results indicated that most processors were was designed for use by the Nuclear Regulatory Commission unable to meet the performance criterion consistently for all ir-radiation categories. Variations in the results were also ob.
and industry to evaluate petitions to clatsify specific waste
- served within specific categories as a result of the irradiation streams as below regulatory concern (BRC). The code provides source. Recomrnendations for changes to the draft standard in -
a capability for calculating radiation doses to a maximal individ-ciude dviding the beta particle and mixture categories, eliminat-ual, critcal group, and the general population as a result of ing the neutron category, and changing the tolerance level for J
transportation, treatment, diso*
~.f ;.A posal actNibes the performance criterion.
involving low level rar%uve waste. Impacts are calculated for mulhpie nuclides and pathways depending on the treatment / dis.
NUREG/CR 6647: APPLICATION OF SURFACE-COMPLEXA-posal options specified by the code user. The treatment /dspos-TION MODELS FOR RADIONUCLIDE ADSORPTION. Sensitivity al options include onsite incineration, offsite incineration at mu-Analysis Of Model input Parameters. HAYES.K.F.; REDDEN.G.;
nicipal and hazardous waste facilities, and offsite disposal at ELA W.; et al Battelle Memorial Institute, Pacif'c Northwest Lab-municipal and hazardous waste landfills included within the dis-oratory. April 1990. 88pp. 9005210220. PNL-7239. 53904:280.
I posal options is the ability to calculate impacts from the sorting This report discusses activity in two areas: 1) an evaluahon of and/or recycling of metal containers and metal and glass mat
- methodologies currently used for the physical and chemical rials. Default environmental and facility parameters are devel-characterization of metal oxide and hydroxide adsorbents and 2) oped from reference treatment / disposal sites, but the user has the sensitrvity of the various surface complexaton models' ad-
' the ophon to replace default paremeters with site-specific pa-sorbent input parameters for describing adsorption. The report rameters. To facilitate use of the code, data input files are cre-describes the relative rnerits of three surface complexation models (SCM), procedures to estimate values of the modet pa-ated and/or edited using a data preprocessor with pull-down rameters from titration data and what is required of expenmen-menus and contact sensitive help screens. The code is written tal titration data sets. The ultimate goalis to determine how and 1
in FORTRAN and runs on 640K IBM-PC and compatible com-whether SCMs can successfully describe adsorption of contami-puters.
nants from disposed nuclear wastes in natural soils and sedi-NUREGICR-6623: DEVELOPMENT OF AN INFILTRATION EVAL-ment. This study's results help clarify the applicability of SCMs, particularly with respect to their sensitivity to input parameters.
UATION METHODOLOGY FOR LOW LEVEL WASTE SHAL.
A method is presented by which unique, best fit values for LOW LAND BURIAL SITES. SMYTH.J.D.; BRESLER,E.;
these parameters.may be obtained. An appendix is provided GEE.G.W.; et al. Battelle Memorial Institute, Pacific Northwest that reviews methods for determining surface area, site density, Laboratory. May 1990. 147pp. 9006290119. PNL 7356c particle size distribution and poro structure.-
An infiltration evaluation methodology (IEM) has been devel-NUREG/CR-6644: REVIEW OF GEOCHEMICAL PROCESSES oped to provide a consistent, well-formulated approach for eval-AND CODES FOR ASSESSMENT OF RADIONUCLIDE MIGRA-unting field scale infiltration and drainage at low level waste TION POTENTIAL AT COMMERCIAL LLW SITES. SERNE,T.J.:
sites. The IEM is designed to simulate significant factors and ARTHUR R.C.; KRUPKA,K.M. Battelle Memorial institute, Pacific hydrologic conditions that determine or influence moisture infil-Northwest Laboratory.. April 1990.13Spp. 9005210229. PNL-7285. 53688:074.
tration, redistribution, and drainage in engineered cover and bar' rier systems. The IEM recognizes the sources of uncertainty in information on geochemical orocesses that control contami-estimating moisture movement through engineered covers and nant solution concentrations and migration at existing LLW quantifies their influences on infiltration and drainage estimates.
waste sites is summarized. The review identifies the current status and future information needs required for the develop-The IEM is developed on the basis of regulatory requirements ment of effectrve performance assessment models for use in given in 10 CFR 61. Engineered cover design and data avail-site license applicabons. Except for some reports on LLW dis-abihty will be largely site specific. Because of the site-specific posal sites at Sheffield, IL, and West Vaney, NY, few references nature of the design and data availability, the IEM is a flexible were identified that contained adequate geochemical data nec-framework that accepts vanous numerical models, closed-form essary to model geochemical processes that affect migration.
analytical models, and uncertainty approaches. Two applications Tritium appears to be the most mobile radionuclido migrating of the IEM (a closed-form analytcal model and a series of inte.
from burial trenches at commercial LLW sites. The review iden-grated numerical models) are demonstrated for a hypothetical tified microbial-degradation induced anoxia, subsequent iron waste site.
oxide precipitation during oxidation, alkalinity controlled pH changesi and organic complexation reactions as key controls of w
-,w.
e v
-g---9--
m--
-y+
w 9
g
1 14 Main Citations and Abstracts redonuclide migraten. The quantity of experimental and field and cells, blood, specific gravity, and urine pH, blood uros nitro-data against wNch to test geochemecal codes is very limited. All gen, and blood creatinine. We concluded after reviewing two experimenta! work on radionuchde adeorption at commercial years of follow-up medical data that none of the 31 workers LLW sites relies upon the K(d(60)) concept. Studies of the ef-sustained any observable health offects from exposure to urani-facts of organics on radionuclide mobikty suggest that a Co-um. The urinary excretion data were used to develop an lm.
EDTA chelate formed in the original waste may persest indefi-proved system 6c recycling model for inhaled soluble uranium, nitely and lead to enhanced migraten. Other organic radonu-We estimated initial intakes, clearance rates, kidney burdens, chde complexes are less stable and do not significantly on-and resulting radiation doses to lungs, kidneys, and bone sur-hence mobihty under conditions anticipated at commercial LLW faces. Radiation does limits and hmits on intake, as recom-i sites.
monded by the ICRP, were not exceeded. However, the NRC derived limit of 9.6 mg was exceeded by eight of the 31 work-NUREG/CR 5563: COMPUTER PROGRAMS FOR EDDY CUR, ers. Maximum kidney countratens in exposed he ranged RENT DEFECT STUDIES. PATE.J.R.; DODO.C.V. Oak Ridge from 0.06 to 2.5 pg U/g kidney tissue. We found no toxicological.
National Laboratory. June f 990. 256pp. 9007120211. ORNL/
docts on N kW of wwws at m conwntroh r
TM-11505. 54472:307, Several computer programs to aid in the design of eddy cur-NUREG/CR 6673: BORON FLUSHING DURING A BWR ANTICl-rent tests and probes have been written. The programs, wrttten PATED TRANSIENT WITHOUT SCRAM. MIRKOVIC,D.;
in Fortran, deal in various ways with the response to defects ex.
DIAMOND,0.J. Brookhaven National Laboratory. June 1990.
Nbited by four types of probe 0: the pancake probe, the reflec-42pp. 9007120190. BNL NUREG-52237, 54468:246.
+
tion probe, the circumferental borecide probe, and the rarcum-TM W docats a SWy of an accident sequence in a forential encircling probe. Programs are included which calcu-boiling water reactor (BWR) in which there is a large reactivity late the impedance or voltage change in a coil due to a :lefect',
insertion due to the flusNng of borated water from the core.
wNch calculate and plot the defect sensitivity factor of a coil This has the potential to occur during an anticipated transient and which invert calculated or experimental readngs to obtain wit out scram (ATWS) after the injecten of borated water from the size of a defect. The theory upon which the progra ns are s
a The h ems down N based is the Burrows point defect theory, and thus the (alcula-power but if there is a rapid depressurization of the vessel (e.g.,
tons of the programs will be rnore accurate for small cefects.
N M
of N ambc depreeWra NUREG/CR-5557: RELAPS THERMAL HYDRAULIC ANALYSIS tion system), large amounts of low pressure, relatively cold, un-OF THE SNUPPS PRESSURIZED WATER REACTOR.
borated water enters the vessel causing a rapid dilution and KULLBERG.C.M. EG&G Idaho, Inc. (subs. of EG&G, Inc ). May cooling. This study was carned out to determine if the reactrvity 1990. 89pp. 9006080313. EGG-2590. 54074:283.
addition caused by tNs flusNng could lead to a power excursson Thermal-hydraulic analyses of five hypothetical accident sce-sufficient to cause catastropNc fuel damage. Calculations were narios were performed with the RELAPS cornputer code fw the carried out using the RELAPS/ MOD 2 computer code under ef-Westinghouse Standardized Nuclear Unit Power Plant System forent assumptions regarding timing and availabdity of low pres-pressunzed water reactor. TNs work was sponsored by the U.S.
sure pumps and with different reactrvtty coefficients. The results Nuclear Regulatory Commission and la being done in cajunc-showed that the fuel enthalpy rise was insufficient to cause cat-tion with future analysis work at the U.S. Nuclear Regulatory astropNc fuel damage although less severe fuel damage might Commission Technical Training Center in Chattanooga, TN.
stdl be possible due to the overheating of the fuel cladding.
These accident scenarios were chosen to assess and bench-mark the thermal hydraulic capabilities of the Technical Training NUREQ/CR 5674: DETERMINATION OF THE CHEMICAL FORM Center Standardized Nuclear Unit Power Plant System simulator OF TRITIUM IN SELF LUMINOUS SIGNS. BOWERMAN.B.S.;
to model abnormal transient conditions.
CZAJKOWSKI.C.J. Brookhaven National Laboratory. May 1990.
43pp. 9006080314. BNL-NUREG-52238. 54075:00g.
NUREG/CR 5560: AGING OF NUCLEAR PLANT RESISTANCE TEMPERATURE DETECTORS.
HASHEMIAN,H.M.;
Buddng est signs contaming trmum sehnous EgM swrces wwe demanned, and the EgM sources were tested to l
DEVERLY,0.D.; MITCHELL.D.W.; et al. Analysis & Measurement detennine the chemical fwm of Mbum in tNs sW De %-
Services Corp. June 1990. 250pp. 9007120248. 54469:127.
Uve was to quanbfy the amounts of tritiated water (T(2)O or An experimental research project was completed to identify TOH) present in the light sources. The light sources conesst of the effects of normal aging on performance of nuclear safety.
sealed glass tubes coated internally with a phosphor (zinc sul-related RTDs. The limit for initial accuracy of these RTDs was nde) and HHed with trmum gas m2)). UgM source tubes frwn established and the range of their response time was deter-mined. Representative nuclear grade RTDs were tested for cali-four exit signs were tested. Two were new signs, one was six years old, and one was tNrteen years old. In one of the new bration drift at simulated reactor conditions and for shelf-life signs, the total Mtium imentwy included two pwcent trWated drift. TNs included a number of naturally aged RTDs received water. Two signs had higher amounts of tritiated water. 4.5%
from nuclear power plants. The results of this work have shown for the other new sign, and 14.5% for the six year-old sign. In l
that penodic cahbration and response time testing performW the oldest sign, an accurate inventory of the tritium content was once every fuel cycle is a reasonable approach for manage-not available, but tritiated water accounted for 12.2% of the ment of aging of nuclear grade RTDs.
total tritium collected for counting.
NUREG/CR 5566: EVALUATION OF HEALTH EFFECTS IN SE-QUOYAH FUELS CORPORATION WORKERS FROM ACCI.
NUREG/CR 5676: SURVEY OF BORIC ACID CORROSION OF DENTAL EXPOSURE TO URANIUM HEXAFLUORIDE.
CARBON STEEL COMPONENTS IN NUCLEAR PLANTS.
FISHER,0.R.; SWINT,M.J.; KATHREN.R.L Battelle Memorial in.
CZAJKOWSKl C.J. Brookhaven National Laboratory. June 1990.
stitute. Pacific Northwest Laboratory. May 1990. 60pp.
45pp. 9007120t98,8NL NUREG 52239. 54468:203.
9006290116. PNL-7328. 54323.311.
A review of licensee responses to Generic Letter No. 88-05 Uranium urinalyses and medical laboratory results were stud-was performed by the U.S. Nuclear Regulatory Commission led to determine whether there were any health effects from (USNRC). This review encompassed 50 satisfactory responses uranium intake among a group of 31 workers exposed to urani-from the affected licensees. A series of ten (10) joint Brookha.
um hexafluoride (UF6) and hydrolysis products following the ac-von National Laboratory (BNL) and USNRC sudits were per-l cidental rupture of a 14-ton shipping cylinder in earty 1986 at formed on a selection of utdities. All of the licensees audited the Sequoyah Fuels Corporation uranium conversion facility in had program irnplementations which met the intent of the Go-Gore, Oklahoma. Physiological indicators studied to detect neric Letter. A review of the available literature and the plant kidney tissue damage included tests for urinary protein, casts audits has led to the conclusion that the requirements of the
Main Citations and Abstracts 15 Generic Letter have ossentially been met for the 10 plants au-primary coolant pumps. This test was carried out under the dited and it is recommended that resident inspectors verify that OECO LOFT Program.
a documented and irnplemented program is in effect at their own plants.
NUREG/lA-0022: TRAC-PF1/ MOD 1 POST TEST CALCULA-NUREG/CR-5579. VALUE/ IMPACT ASSESSMENT OF JET IM-TIONS OF THE OECD LOFT EXPERIMENT LP SB-3. /
PINGEMENT - LOADS AND PIPE-TO PIPE -IMPACT LLEN.E.J.; NEILL,A.P. United Kingdom Atomic Energy Authonty, DAMAGE. Revised Methods And Catena. BROWN.J.B.;
April 1990.100pp. 9006290072. AEEW R 2275. 54319:001.
BAMPTON,M.C.C.; ALZHEIMER J.M. Battelle Memonal Institute, Analysis of the small, cold leg break, OECD LOFT Experiment Pacife Northwest Laboratory. June 1990. 72pp. 9007120168.
LP SB 3 using the best-estimate computer code TRAC PF1/
PNL 7339. 54470:267.
MOD 1 is presented. Desenptons of the LOFT facility and the To account for effects that might result from a loss-of coolant LP SB-3 experiment are given and development of the TRAC-accident (LOCA). nuclear power plant designers have been re.
PF1/ MODI input model is detailed. The calculations performed quired to analyze the effects of double-ended guillotine breaks (DEGB) in high-energy piping. The U.S. Nuclear Regulatory in achieving the steady state conditions, from which the experi-ment was initiated, and the specifcation of experimental bound-Commission (NRO), through its Standard Review Plan (SRP), re-ary conditions are outlined, quires that plant designers follow certain presenbod methods and criteria in the estimahon of dynamic effects associated with NUREG/lA 0023 V01: ASSESSMENT OF TRAC-PF1/ MOD 1 the postulated rupture of piping. The work reported in this VERSION 14.3 USING SEPARATE EFFECTS CRITICAL FLOW NUREG is intended to provide the basis for NRC decisions on adopting revisions to parts of the SRP 3.6.2 entitled "Determi-AND BLOWDOWN EXPERIMENTS. Volume 1: Text And Tables.
nation of Rupture Locations and Dynamic Effects Associated SPINDLER.B.t PELLISSIER,M. France, Govt. o1. January 1990.
with the Postulated Rupture of Piping" The revisions consid-141pp. 9006110015. SETH/LEML88138. 54084:236.
ered in this wnrk evaluated updated presenptions for calculating Independent assessment of the TRAC code was conducted jet impingement forces on critical systems and the requirement at the Centre d' Etudes Nucleaires de Grenoble of the Commis-to consider pipe-whip damage to a new population of pipes. In sariate a rEnergie Atomique (France)in the frame of the ICAP.
accordanco with the procedures documented in NUREG/CR.
. This report presents the results of the assessment of TRAC-3586 entitled "A Handbook for Value-impact Assessment", this PF1/ MODI version 14.3 using entical flow steady state tests report found Indication that substantial costs and occupational (MOBY-DICK, SUPER MOBY DICK), and biowdown tests radiation exposure would result from the proposed action with-(CANNON, SUPER CANNON, VERTICAL CANON, MARVIKEN' out suostantially reducing the ned to public health and safety.
OMEGA TUSE, OMEGA BUNDLE).
i NUREG/lA 0011: TRAC PF1/ MOD 1 POST TEST CALCULA-NUREG/lA 0023 ' V02: ASSESSMENT OF TRAC-PF1/ MOD 1 TlONS OF THE OECD LOFT EXPERIMENT LP-SB 1.
VERSION 14.3 USING SEPARATE EFFECTS CRITICAL FLOW ALLEN,E.J. United Kingdom Atomic Energy Authority. April AND BLOWDOWN EXPERIMENTS. Volume 2: Figures.
1990,137pp. 9007120212. AEEW-R 2254. 54466:163.
Analysis of the small, hot leg break, OECD LOFT Experiment SPINDLER.B.t PELLISSIER.M. France, Govt. of. January 1990.
LP SB-1. using the "best-estimate" computer code TRAC PF1/
238pp. 9006110013. SETH/LEML88138. 54083:264.
MODI is presented. Descriptions of the LOFT facility and the See NUREG/lA.0023,V01 abstract.
LP SB 1 experiment are given and development of the TRAC-NUREG/lA 0030: ASSESSMENT OF RELAPS/ MOD 2 CODE PF1/ MOD 1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experi-USING LOSS OF OFFSITE POWER TRANSIENT DATA OF ment was initiated, and the specification of experimental bound-KNU #1 PLANT. CHUNG,B-D.; KIM,H-J. Korea Advanced ary conditons are outHned-Energy Research Institute. LEE,Y J. Seoul Nabonal Univ., Seoul, Republic of Korea. April 1990.100pp. 9005040134. 53620:288.
i NUREG/lA-0018: RELAPS/ MOD 2 ASSESSMENT,OECD-LOFT This report presents a code assessment study, based on a SMALL BREAK EXPERIMENT LP-SB-03. GUNTAY,S. Paul real plant transient that occurred on June 9,1961, at the KNU Scherrer Institute. April 1990.100pp. 9006060163. 54084:136.
- 1 (Korea Nuclear Unit Number 1). KNU #1 is a two-loop Wes-An analysis of the experimental results and post. test calcula-tions using RELAP5/ MOD 2 carried out for OECD LOFT small tinghouse PWR plant of 587 Mwe. The loss of offsite power break experiment LP SB-3 are presented. Experiment LP.SB-3 transient occurred at the 77.5% reactor power with 0.5%/hr was conducted on March 5,1964 in the Loss-of Fluid Test power ramp. The real plant data were collected from available (LOFT) facihty located at the Idaho National Engineering Labo-on-line plant records and computer diagnostics. The transient retory (INEL). The experiment simulated a small cold leg break, was simulated by RELAP5/ MOD 2/36.05 and the results were with concurrent loss of high pressure injection system, and cool-compared with the plant data to assess the code weaknesses down and recovery by feed and bleeo of the steam generator and strengths. Some nodalization studies were performed to secondary side and accumulator injection respectively. This contribute to developing a guideline for PWR nodalization for report documents a short post. test analysis of the experiment the transient analysis.
emphael2ing the results of additional analysis performed during NUREG/lA 0031: ICAP ASSESSMENT OF RELAPS/ MOD 2.
the course of this task. RELAPS/ MOD 2 input model and results of the post. test calculation are documented. Included in the CYCLE 36.05 AGAINST LOFT SMALL BREAK EXPERIMENT report is the results of a sensitivity analysis with show the pre-L3 7. LEE,E.J.; CHUNG,B.D.; KIM,H.J. Korea Advanced Energy dicted thermal hydraulic response to a different input model.
Research institute. April 1990.171pp. 9005040112. 53621:028.
The LOFT small break (1 in dia) experiment L3 7 has been NUREG/lA 0021: RELAP5/ MOD 2 CALCULATIONS OF OECD analyzed using the reactor thermal hydraulic analysio code LOFT TEST LP.SB 2. HALL,P.C. Central Electricity Generating RELAPS/ MOD 2, Cycle 36.05. The base calculation (Case A)
Board. April 1990. 43pp. 9005040095. GD/PE N/606.
was completed and compared with the experimental data.
53622:176.
To help in assessing the capabilities of RELAPS/ MOD 2 for Three types of sensitivity studies (Cases B, C, and D) were car-PWR Fault Analysis, the code is being used by CEGB to simu-ried out to investigate the effects of (1) break discharge coeffi-late several small LOCA and pressunzed transient experiments cient Cd. (2) pump two-phase difference multiplier, and (3) high in the LOFT experimental reactor. The present report describes pressure injection system (HPIS) capacity on major therrnal and hydraulic (T/H) parameters. A nodalizaten study (Case E) was an analysis of small LOCA test LP-SB-02, which simulated a conducted to assess the phenomena with a simplified nodaliza-1% hot leg break LOCA in a PWR, with delayed tnpping of the tion.
~
~-
i 16 Main Citations and Abstracts
- NUREG/lA-0032: ASSESSMENT OF RELAP5/ MOD 2 CYCLE effect of different nodalizaton was studied in the area of the 36.04 USING LOFT LARGE BREAK EXPERIMENT L2 5.
downcomer and core. For a sensitivity study, another calcula.
BANG,Y.S.: LEE.S.Y.; K)M,H.J. Korea Advanced Energy Re-tion was executed using an updated version of RELAP5/ MOD 2 Cycle 36,04. A split downcomer with one crossflow junction and
~ search Institute. Apru 1990,183pp. 9005040105. 53621:199. _
The LOFT L2 5 LBLOCA experiment was simulated usin0 the two core channels were found to be effective in descrbing the -
, RELAPS/ MOD 2 Cycle 36.04 code to assess its capability to -
ECC bypass and hot channel behavior. And the updated version predict the phenomena in LBLOCA. One base case calculation was found to be effectue in overcommg the code defeiency in.
- j
?
and three cases of different nodalizatons were carried out. The '
the interfacial friction and reflood quenching.
.?
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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.
N SECONDARY REPORT nut 00ER REPORT NUtoOEM ONDARY REPORT nut 0SER REPORT nut 3SER AEEW.R 2254
' NUREG/lA 0011 NL/TM-11319 NUREG/C45439
'L AEEW R 2275 NUREG!lA-0022 NL/TM 11350 NUREG/C45449 ANL 89/40.
NUREG/CR-4867 V07
% y '8 hMf$$
ANL 90/4 -
. NUREG/CR-4667 V08 g
BNL NUREG-51454 NUREG/CR 233! V00 N4 OANLitM-11451 NUREG/CR 5514 BNL NUREG42237 NUREG/C45573
' ORNWM-11505 NUREG/CR4553 BNL NUREG42238 NUREG/CR4574.
PARADETER IE155 NUREG/CR4525 PNL-5210 NUREG/CR-3950 VD6 BNL NUREG42239 NUREG/CR.5576 PNL4931 NUREG/CR 5397
!(,
CBCS RPP #901 NUREG/C45438 PNL 7239 NUREG/CR4547.
EGO-2458 NUREG/CR4639 V01 Rq PNL 7276 NUREG/C45540 EGG-2535 NUREG/CR 5111 PNL 7285 NUREG/C45548 -
EGG-2545 NUREG/CR4181 hg2 -
glg
'j EGO-2554 NUREG/CN4550 V7 Rol EGO-2592 NUREG/CR-5513 V02 PNL-7356 NUREG/C45523 -
EGO 2592 NUREG/C45513 V01 SAC 99/1744 NUREG/C45510 ~
i EGO-2599 NUREG/C45557 SAC 90rt130 NUREG/CR4436 V02 -
SAIC 90/1130 NUREG/CR-5436 V01 GD/PE N/606 NUREG/LA 0021 SAND 80-1429 NUREG/CRd667 t
i
?EB-88 005 NUREG-1402 6ANO86-1562 NUREG/CR46$1 V01 l
LMF 124 NUREG/C45430 SANO86-2084 NUREG/C44550 V3 RIP 2 i
MEA-2377 NUREG/CR-5494 SAND 86 2084 NUREG/CR-4550 V3A1P1 1
MTR 90W00013 NUREG/CP 0108 SANO88-2084 NUREG/CR 4550 V5AtP2 q
ORNL4589 -
NUREG/C45463 SANO86-2084 NUREG/CR-4550 V58191 n
ORNL/CSO/TM-267 NUREG/C45366 SAND 86-2940 NUREG/CR-5253.
]
ORNL/CSO/TM-270 NUREG/C45468 h$$
ORNL/NSC200 NUREG/C42000 V00 N3 ORNL/EC200 NUREG/CR-2000 V00 N4 SANO891821 NUREG/C45411 ORNL/HSC200 NUREG/CR 2000 V09 N5 SANO89-2368 NUREG/CR 5460 ORNL/TM-10328 NUREG/CR4816 SANO894080 i
SETH/LEML88-138.
NUREG/CR4517 ORNL/TM 11267 NUREG/CR 5409 SETM/LEML88-138 NUREG/LA-0023 V01 NUREG/lA 0023 V02 l
i q
i f
17 r
a
-_,___-----r-w--y-v--------
I l
t i
i 1
l
=
I
- .\\
i f
i i
.4 i
2 s
i t
1 i
1 i
1 L
Personal Author Index' This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by the author. If further information is needed, refer to the main cita-tion by the NUREG number.
ADAMS,J.
BERTUCIO,R.Ci NUREG/CR-4753 V03: CANADIAN SEISMIC AGREEMENT. Annual NUREG/CR-4550 V3RtP1: ANALYSIS OF. CORE DAMAGE Report 1987 1988.
FREQUENCY.SURRY, UNIT 1, INTERNAL EVENTS.
~ NUREG/OR.4753 V04. CANADAN SE!SMO AGREEMENT. Annual NUREG/CR-4550 V3 RIP 2: ANALYSIS OF CORE DAMAGE Report 1986 1989.
FREQUENCY.SURRY, UNIT 1. INTERNAL EVENTS APPENDICES.
ALLEN 1J.
NUREG/CR.4550 V5 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN-M NUREGhA-00f t: TRAC.PF1/ MOD 1 POST. TEST CALCULATIONS OF -
CY: SEQUOYAH. UNIT 1.lNTERNAL EVENTS THE OECD LOFT EXPERIMENT LP.SB-1.'
NUREG/CR 4550 V5RtP2: ANALYSIS OF CORE DAMAGE FREQUEN.
h CY: SEQUOYAH UNIT 1. INTERNAL EVENTS APPENDICES.
AL'.EN J.A.'.
=
M NUREG/CR.5438: BASO CONSOERATONS IN PREDICTING ERROR SEVERLY,0.D. :
PROBABILITIES IN HUMAN TASK PERFORMANCE.
NUREG/CR-5560 AGING OF NUCLEAR PLANT RESISTANCE TEM-PERATURE DETECTOGS.
AL2NEIMER,J.R
= NUREGICR4579: VALUE/ IMPACT ASSESSMENT OF JET IMPINGE-DONANO,E.J.
MENT LOADS AND PIPE TO. PIPE IMPACT DAMAGE. Revised Meth-NUREG/CR 5411: ELICITATION & USE OF EXPERT JUDGMENT IN
. ods Artt Crttena.
PERFORMANCE ASSESSMENT FOR HIGH-LEVEL RADCACTIVE ANDREW,KD,-
' NUREG/CR 4753 V03: CANADIAN SEISMIC AGREEMENT. Annual
' BOWERMAN,8.S.
N GG5M DETENNATON & WE CHEN N W NUR 53 V04:. CANADIAN SEISMIC AGREEMENT. Annual Report: 1988-1989.
TRITIUM IN SELF LUMINOUS SIGNS.
ANGLIN,F.M.
BRESLER,E.
NUREG/CR.4753 MXL CANADIAN SEISMIC AGREEMENT. Annual NUREG/CR-5523'-DEVELOPMENT OF AN INFILTRATION EVALUA-Report: 1907 1988.
TON METHOOOLOGY FOR LOW. LEVEL WASTE SHALLOW LAND NUREG/CR-4753. V04: CANADIAN SEISMIC AGREEMENT. Annual BURIAL SITES.
Report 1908 1989.
APPt4MANt.P.
'BRIGOS,H.C..
- NUREG/CR.5399: SURVEY OF STATE AND TRIBAL EMERGENCY RE-NUREG/CR 5436 V01: THE DEVELOPMENT AND EVALUATON OF PROGRAMMATIC *ERFORMANCE INDICATORS ASSOCIATED WITH SPONSE CAPABlWTIES FOR RADOLOGICAL TRANSPORTATION INCIDENTS.
MAINTENANCE Al NUCLEAR POWER PLANTS Mein Report-NUREG/CR-5436 V02: THE DEVELOPMENT AND EVALUATON OF BROADHEAD,B.L.' '
' PROGRAMMATIC PERFORMANCE INDLCATORS ASSOCIATED WITH NUREG/CR 5468: QADS: A MULTIDMENSONAL POINT KERNEL
.MAIMENANCE AT NUCLEAR POWER PLANTS. Appendices-ANALYSIS MODULE.
ARNOLD,W.D.
SORPTON OF R NT D ES O TO S F N E 5437: ORGANIZATION AND SAFETY IN NUCLEAR POWER PLANTS.
ARTHUR R.C.
NUREG/CR 5548: REVIEW OF GEOCHEMICAL PROCESSES AND BROWN,C.
CODES FOR ASSESSMENT OF RADIONUCLOE MiGRATON PO.
NUREG 1404: LICENSEE USE OF TACTICAL EXERCISE RESULTS.
. TENTIAL AT COMMERCIAL LLW SITES.
BROWN.J.B.
ASGAR).M. '
NUREG/CR 5579LVALUE/ IMPACT ASSESSMENT OF JET IMPINGE-
. NUREGICR-5449: DETERMINATION OF THE NEUTRON AND 4AMMA MENT LOADS AND PIPE.TCwPIPE IMPACT DAMAGE.Revtsed Moth.
Fulx DISTRIDUTION IN THE PRESSURE VESSEL AND CA 1TY OF ods And Critena.
A BOILit.HG WATER REACTOR.
SAMLWA BROWN,S.R. -
NUREG/CR4550 V5AIPt: ANALYSIS OF CORE DAMAGE FREQUEN.
HUREG/C43950 V06: FUEL PERFORMANCE ANNUAL REPORT
"$R U80-CY: SEQUOYAH. UNIT 1,!NTERNAL EVENTS.
NUREG/CR-4550 V5RtP2: ANALYSIS OF CORE DAMAGE FREQUEN-BALDWIN,C.A.
CY: SEQUOYAN, UNIT 1,1NTERNAL EVENTS APPENDICES.
NUREG/CR-5409: NEUTRON EXPOSURE PARAMETERS FOR THE BUFFARDI,L.C
^
ON T L IR 1AD ON SER '
' NUREG/CR-6438. BASIC CONSIDERATIONS IN PREDCTING ERROR
- PROBABILITIES IN HUMAN TASK PERFORMANCE.
SAMPTON M.C.C.
NUREG!CR 5579 VALUE/lMPACT ASSESSMENT OF JET IMPINGE.
BURLILE,G.
MENT f.OADS AND PIPE.TO.P1PE IMPACT DAMAGE. Revised Meth.
NUREG/CR-5436 V01: THE DEVELOPMENT AND EVALUATION OF.
oda Ard Criterio.
PROGRAMMATIC PERFORMANCE INDtCATORS ASSOCIATED WITH MAINTENANCE AT NUCLEAR POWER PLANTS. Main Report BANC,Y.S.
NUREG/CR 5436 V02:,THE DEVELOPMENT AND EVALUATON OF NVilEG/lA 0032 ASSESSLIENT OF RELAPS/ MOD 2 CYCLE 36.04 PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH USING LOf'T LARGE BREAK EXPERIMENT L2 5.
MAINTENANCE AT NUCLEAR POWER PLANTS. Appendices.
19
20
. Personal Author index.
CAMPSELL,0.
DR YSDALE,J.A.
NUREG/CR 5460 A CAUSE-DEFENSE APPROACH TO THE UNDE 4 NUREG/C44753 ' V03: CANADIAN SEISMIC AGREEMENT. Annual STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
Report 1987 1986.. CANADIAN SE!SMIC AGREEMENT. Annual HU8tEG/CR-4753 V04:
CAMP 9ELLJ.E.
Report: 1988 1989.
NUREG/CR-1667: AISK METHODOLOGY FOR GEOLOGC DISPOSAL OF RADCACTIVE WASTE. Scenario Selecten Procedure.
EDSON.J.L..
. NUREGIC45181:- NUCLEAR PLANT AGING RESEARCH.THE 1E CARTER,R)kS439 HUMAN FACTORS ISSUES ASSOCIATED WITH AD.POWER SYSTEM.
NUREG/C VANCED INSTRUMENTATON AND CONTROLS TECHNOLOGlES IN EO80NAF.
NUCLEAR PLANTS.
NUREG/CR-5489 BIOLOGICAL CHARACTERIZATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL-CASE.F.L ING EFFLUENTS.
NUREG/C45463: EFFECTS OF MINERALOGY ON SORPTON OF.
STRONTIUM AND CESIUM ONTO CAUCO HILLS TUFF.
gg n NUREG/CR 5547: APPUCATION OF SURFACE COMPLEXATION -
CHANIN,0.L MELCOR ACCIDENT CONSEQUENCE CODE MODELS FOR RADONUCLlDE ADSORPTION.Sensitwity Analysis Of 4
NUREG/C44601 V01:
t SYSTEM (MACCS) Volume 1: User's Gude.
. ModW input Parameters.
CHEVERTON,R.D.
FERREIRA,V.A.
NUREG/C45473: INCLUSION OF UNSTABLE DUCTILE TEARING AND NUREG/CR 5475: MODEL FEASIBILITY STUDY OF RADIOACTIVE EXTRAPOLATED CRACK-ARREST TOUGHNESS DATA IN PWR c PATHWAYS FROM ATMOSPHERE TO SURFACE WATER.
VESSEL INTEGRITY ASSESSMENT.
.FE8ENMAIER,J.
CHUNG,5-0.
' NUREG/CR 5399: SURVEY OF STATE AND TRIBAL EMERGENCY RE-NUREG/lA 0030: ASSESSMENT OF RELAP5/ MOD 2 COOE USING SPONSE CAPAB!LITIES FOR RADOLOGICAL TRANSPORTATION LOSS OF OFFSITE POWER TRANSIENT DATA OF KNU #1 PLANT.
INCIDENTS.
NUREG C45566: EVALUATION OF HEALTH EFFECTS IN SEQUOYAH SMAL REA EXPER NT FUELS CORPORATON WORKERS FROM ACCIDENTAL EXPOSURE TO URANIUM HEXAFLUORIDE.
COLUNS,J.L NUREG/C45480: DATA
SUMMARY
REPORT FOR FISSION PRODUCT F LDSHM AN,E.A.
RELEASE TEST VI-3.
NUREG/CR 5438: BASIC CONSIDERATIONS IN PREDICTING ERROR PROBADIUTIES IN HUMAN TASK PERFORMANCE.
CONKLIN.J.C.
NUREG/C45514: MODEUNG AND PERFORMANCE OF THE MHTGR REACTOR CAVITY COOUNG SYSTEM.
FOLEY W.J..
'7 NUREG/CR 4525: CLOSEOUT OF IE BULLETIN 84 03 REFUEUNG CRANWELL.R.M.
- CAVITY WATER SEAL.
NUREG/C41667: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL OF RADCACTIVE WASTE. Scenario Selection Procedure.
FRAGOLA.J...
THE DEVELOPMENT AND EVALUATON OF 3
NUREGIC45436 Vot:
CZAJKOWSKI,C.J PROGRAMMATIC PERrORMANCE INDICATORS ASSOCIATED WITH f
NUREG/CR 55N: DETERMINATON OF THE CHEMICAL FORM OF MAINTENANCE AT NUCLEAR POWER PLANTS Main Report.
TRITIUM IN SELF LUM! NOUS SIGNS NUREG/CR-5436 V02: THE DEVELOPMENT AND EVALUATON OF l
NUREGICR 5576: SURVEY OF BORIC ACIO CORROSION OF CARBON PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH STEEL COMPONENTS IN NUCLEAR PLANTS' MAINTENANCE AT NUCLEAR POWER PLANTS. Appendices. '
DAVIS,P.A.
FROST,R.H.
NUREG/CR-5393: A REVIEW OF TECHNIOUES FOR PROPAGATING NUREG/C45001: EFFECTS OF MANUFACTURING VARIABLES ON DATA AND PARAMETER UNCERTAINTIES IN HIGH-LEVEL RADIO.
PERFORMANCE OF HIGH. LEVEL WASTE LOW CARBON STEEL ACTIVE WASTE REPOSITORY PERFORMANCE ASSESSMEW CONTAINERS.
MODELS.
I GALYEAN,W.J.
\\
DEAN,R.S.
NUREG/CR-4639 V01 R1: NUCLEAR COMPUTERIZED UBRARY FOR
~
NUREG/CR-4525: CLOSEOUT OF IE BULLETIN 84 03 REFUEUNG ASSESSING REACTOR REUABlUTY (NUCLARR). Summary Desenp-CAVITY WATER SEAL.
tion.
DEAN.W.
NUREG 1021 R06: OPERATOR UCENSING EXAMINER STANDARDS. -
GASKINS,R.C.
NUREG/C45438: BASIC CONSIDERATONS IN PREDICTING ERROR BABMES IN HM W MNME. -
N EG-1333: MAINTENANCE APPROACHES AND PRACTICES IN SE.
LECTED FOREIGN NUCLEAR POWER PROGRAMS AND OTHER
' GEE,0 W.
NUREG/C45523: DEVELOPMENT OF AN INFILTRATON EVALUA-U.S. INDUSTRIES: REVIEW AND LESSONS LEARNED.
. TION METHODOLOGY FOR LOW-LEVEL WASTE SHALLOW LAND
'f DIAMOND.D.J.
BURIAL SITES.
NUREG/C45573: BORON FLUSHING DURING A BWR ANTICIPATED TRANSIENT WITHOUT SCRAM.
GENTILLON.C.D.
NUREG/CR-4639 V01 R1: NUCLEAR COMPUTERI2ED UBRARY FOR DICKSON,T.L.
ASSESSING REACTOR REUABILITY (NUCLARR). Summary Descrip-NUREG/CR 5473: INCLUSION OF UNSTABLE DUCTILE TEARING AND ton.
EXTRAPOLATED CRACK-ARREST TOUGHNESS DATA IN PWR VESSEL INTEGRITY ASSESSMENT.
GERTMAN,D.L NUREG/CR 4639 V01 Rt: NUCLEAR COMPUTERIZED LIBRARY FOR DfERCK8,D.R.
ASSESSING REACTOR REUABluTY (NUCLARR> Summary Descrip.
NUREG/CR4667 V00: ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS Semiannual Report. October 1988 - March ton.
GILBERT,B.G.
NUREG/CR-4639 V01 Rt: NUCLEAR COMPUTER; ZED UBRARY FOR DODD.C.V.
NUREG/CR 5553: COMPUTER PROGRAMS FOR EDDY.CURREM ASSESSING REACTOR REUABldTY (NUCLARR). Summary Descrip-l DEFECT STUDIES.
ton.
l l
l L
Personal Author index 21 GiLEs.G.E.
HUGHES,A.A-NUREG/CR-5366. HTAS2; A THREE DIMENSIONAL TRANSIENT SHIP-NU".EG/CP.0106: PUBLIC WORKSHOP ON NUCLEAR POWER PLANT PING CASK ANALYSIS TOOL LICENSE RENEWAL GILMORE,W.L IMAN R.L NUREG/CR4639 V01 R1: NUCLEAR COMPlJTERIZED LIBRARY FOR NUREG/C&5253; PARTITON: A PROGRAM FOR DEFINING THE ASSESSING REACTOR RELIABluTY (NUCLARRISummary Desenp-SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE NUREG 1150 PROBABluSTIC RISK ASSESSMENTS. User's Guide.
NUREG/CR-5262 PRAMIS: PROBABluTY RISK ASSESSMENT MODEL GRAVES,C.C.
NUREG 1372: REGULATORY ANALYSIS FOR THE RESOLUTION OF INTEGRATION SYSTEM. Users Gude. -
GENERIC ISSUE C-8,
- MAIN STEAM ISOLATON VALVE LEAKAGE JAMISON.J.D.
AND LCS FAILURE.a NUREG/C45397: VALUE lMPACT ANALYSIS OF REGULATORY OP-GROH M.R.
TONS FOR RESOLUTION OF GENERIC ISSUE C-8:MSIV LEAKAGE NUREG/CR4639 Voi R1: NUCLEAR COMPUTERIZED UBRARY FOR AND LCS FAILURE.
ASSESSING REACTOR REUABlWTY (NUCLARR) Summary Descrip.
- tion, JOHNSON,J.D.
NUREG/CR.5253: PARTITON. A PROGRAM FOR DEFINING THE GUNTAY,S.
SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE NUREG/lA-0018: RELAP5/ MOD 2 ASSESSMENT OECD' LOFT SMALL NUREG-1150 PROBABluSTIC RISK ASSESSMENTS. Users Guide BREAK EXPERIMENT LP.SB 03.
NUREG/CR-5262-PRAMIS: PROBABILITY RISK ASSESSMENT MODEL INTEGRATON SYSTEM. Users Guide.
GUTJAHRAL NUREG/C45393: A REVIEW OF TECHNIQUES FOR PROPAGATING DATA AND PARAMETER UNCERTAINTIES IN HIGH-LEVEL RADtO-NUREG-1394: EMERGENCY RESPONSE DATA SYSTEM (ERDS) IM-ACTIVE WASTE REPOSITORY PERFORMANCE ASSESSMENT PLEMENTATON.
MODELS.
GUZOWSKI,R.V.
JOW,H.N.
NUREG/C41667: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL NUREGICA-4691 V01: MELCOR ACCIDENT CONSEQUENCE CODE OF RADIOACTIVE WASTE. Scenario Selection Procedure.
SYSTEM (MACCS).Votane 1: Users Guide.
HALLK.W.
JULIVS,J.A.
NUREG/CR 4550 V7 RO1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR.4550 V3 RIP 1: ANALYSIS OF CORE DAMAGE CY: ZION. UNIT 1, INTERNAL EVENTS.
FREQUENCY:SURRY. UNIT 1,1NTERNAL EVENTS.
NUREG/CR4550 V3R1P2-ANALYSIS OF CORE DAMAGE HALLP.C' lA-0021:
FREQUENCY:SURRY, UNIT 1,1NTERNAL EVENTS APPENDICES. -
NUREG/
RELAP5/ MOD 2 CALO ^ TONS OF OECD LOFT TEST LP.SB-2.
ggy,p,g, HANSON.D.J.
NUREG/CR-5409: NEUTRON EXPOSURE PARAMETERS FOR THE NUREGICR 5513 V01: ACCIDENT MANAGEMENT INFORMATICN METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY SEC-NEEDS. Volume 1. Methodology Deveiopment And Application To A TION ETEEL IRRADIATION SERIES.
Pressunzed Water Reactor (PWR) With A Large, Dry Containment.
NUREG/CR-5513 V02: ACCIDENT MANAGEMENT INFORMATION KAM,F.B.K.
i NEEDS. Volume 2. Appendices.
NUREG/CR4816: PR-EDB: POWER REACTOR EMBRITTLEMENT DATA BASE, VERSION 1. Program Desenption.
HARTY,R-NUREG/C45449: DETERMINATION OF THE NEUTRON AND GAMMA NUREG/C45540: PERFORMANCE TESTING OF EXTREMITY FLUX DISTRIBUTON IN THE PRESSURE VESSEL AND CAVITY OF DOSIMETERS. STUDY 2.
A BOILILNG WATER REACTOR.
HASHWMIAN,H.M.
KASSNER,T.F.
NUREG/C45560: AGING OF NUCLEAR PLANT RESISTANCE TEM-NUREG/CR-4667 V07: ENVIRONMENTALLY ASSISTED CRACKING IN PERATURE DETECTORS.
UGHT WATER REACTORS. Semiannual Report, April-September 1988.
NUREG/CR4667 V08: ENVIRONMENTALLY ASSISTED CRACKING IN HAVES,K.F.
NUREG/CR 5547: APPUCATION OF SURFACE COMPLEXATON UGHT WATER REACTORS. Semiannual Report,0ctober 1988 March 1969.
MODELS FOR RADIONUCUDE ADSORPTON.Sensitmty Analysis Of KATHREN,R.L Model input Parameters.
HELTON.J.C, NUREG/CR-5566: EVALUATON OF HEALTH EFFECTS IN SEQUOYAH NUREG/C45253: PARTITION: A PROGRAM FOR DEFINING THE FUELS CORPORATION WORKERS FROM ACCIDENTAL EXPOSURE SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE TO URANIUM HEXAFLUORIDE.
NUREG 1150 PROBABIUSTIC RISK ASSESSMENTS Users Guide.
NUREGICR.5262: PRAMIS: PRCBABILITY RISK ASSESSMENT MODEL KEEMEY,R.L INTEGRATION SYSTEM. Users Guide.
NUREG/CR.5411: EUCITr. TON & USE OF EXPERT JUDGMENT IN l
PERFORMANCE ASSESSMENT FOR HIGH LEVEL RADIOACTIVE HENNICKA WASTE REPOslTOnlES.
NUREG/CR4525 CLOSEOUT OF lE BUt.LETIN 84-03:REFUEUNG CAVITY WATER SEAL KIM,H-J.
NUREG/lA-0030: ASSESSMENT OF RELAP5/ MOD 2 CODE USING HISER,A LOSS OF OFFSITE POWER TRANSIENT DATA OF KNU #1 PLANT.
NUREG/CR-5494: CORRELATION OF IRRADIATION INDUCED TRAN.
SrilON TEMPERATURE INCREASES FROM C(V) AND K(JC)/K(IC)
KIM,H.J NUREG/lA.0031: ICAP ASSESSMENT OF RELAP5/ MOD 2, CYCLE HOOKE R,C.D.
36.05 AGAINST LOFT SMALL BREAK EXPERIMENT L3-7.
NUREG/C45540: PERFORMANCE TESTING OF EXTREMITY NUREG/lA-0032: ASSESSMENT OF RELAP5/ MOD 2 CYCLE 36.04 DOSIMETERS, STUDY 2.
USING LOFT LARGE BREAK EXPERIMENT L2 5.
HORA,S.C.
KINC AID C.T.
NUREG/C45411: EUCITATION & USE OF EXPERT JUDGMENT IN NUREG/CR-5523: DEVELOPMENT OF AN INFILTRATON EVALUA-PERFORMANCE ASSESSMENT FOR HIGH-LEVEL RADIOACTIVE TON METHOOOLOGY FOR LOW LEVEL WASTE SHALLOW LAND WASTE REPOSITORIES.
BURIAL SITES.
l 1
E i
,22 Personal Author Index 3
KRUPKA,K.M.
MCOUIRE,8.A. -
. NUREG/CR-5548: REVIEW OF GEOCHEMCAL PROCESSES AND NUREG 1391 DAFT FC: CHEMICAL TOXCITY OF URANIUM HEXA-CODES FOR ASSESSMENT OF RADONUCUDE MIGRATON PO-FLUORIDE DELATED TO RADIATON DOSES. Draft Report For Com.
TENTIAL AT COMMERCIAL LLW SITES.
ment.
. KULLSERG.C.M.
MCNAMARA.N.
. NUREG/CR-5557: RELAP5 THERMAL.HYDRAULO ANALYSIS OF THE NUREG 0637 V10 N01: NRC TLD DIREG RADIATON MONITORING SNUPPS PRESSURIZED WATER REACTOR.
NETWORK.Progrees Report January March 1990.
- KURTH.R.E. -
MEYER,LC.
NUREG/C45510: EVALUATONS OF CORE MELT FREQUENCY EF*
NUREG/C45181: NUCLEAR PLANT AGING RESEARCH.THE 1E FECTS DUE TO COMPONENT AGING AND MAINTENANCE.
POWER SYSTEM.
LAMONTAGNE M.
MEYEROR.
NUREG/C447! V03: CANADIAN SEl;,MC AGREEMENT. Annual NUREG/C45513 V01: ACCOENT MANAGEMENT INFORMATON l
NUR CR.4753 V04: CANADIAN SEISMIC AGREEMENT. Annual NEEDS. Volume 1 Methodology Devempment And Application To A l
Report: 1988-1989' Proesurtred Water Reactor (PWR) With A Large. Dry Containment.
NUREG/CR-5513 V02: ACCIDENT MANAGEMENT INFORMATON LANO J.F.
NEEDS. Volume 2. Appendices.
NUREG/CR-6463: EFFECTS OF MINERALOGY ON SORPTION OF STRONTlUM AND CESIUM ONTO CAUCO HILLS TUFF.
ME "
UREG/C45463: EFFECTS OF MINERALOGY ON SORPTION OF LAPOINTE,$.P.
STRONTIUM AND CESIUM ONTO CALCO HILLS TUFF.
NUREG/CR.4753 V04: CANADIAN SEISMC AGREEMENT. Annual Report 1988 1989.
MILLER,LF.
[
NUREG/CR 5409: NEUTRON EXPOSURE PARAMETERS FOR THE
~
LECKIE.J.O.
METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY.SEC.
NUREG/CR-5547: APPLCATON OF SURFACE COMPLEXATON TON STEEL IRRADIATON SERIES.
MODELS FOR RADIONUCLOE ADSORPTON.Sensitmty Analysie Of Model input Parameters.
MIRKOVIC.D.
NUREG/C45573: BORON FLUSHING DURING A BWR ANTICIPATED LEE.C.E.
NUREG/CR5517: IMPACTS-BRC,VERSON 2.0 Program User's Manual MITCHELL.D.
LEE E.J.
NUREG/C45460 A CAUSE DEFENSE APPROACH TO THE UNDER.
NUREG/lA 0031: CAP AS1iESSMENT OF RELAP5/ MOO 2. CYCLE STANDING AND ANALYSIS OF COMMON CAUSE FA! LURES.
36.05 AGA!NST LOFT SMALL BREAK EXPERIMENT L3,7.
LEE H.K E0EME NUREG/C45460: DATA
SUMMARY
REPORT FOR FISSION PRODUCT U M558e AM & EM N MSISME E RELEASE TEST VL3.
PERATURE DETECTORS.
LEE,S.Y.
MITTER,E.L NUREG/lA-0032: ASSESSMENT OF RELAP5/ MOD 2 CYCLE 36.04 NUREG/C45399: SURVEY OF STATE AND TRIBAL EMERGENCY RE-
- USING LOFT LARGE DREAK EXPERIMENT L2 5.
SPONSE CAPABluTIES FOR RADIOLOGICAL TRANSPORTATION INCIDENTS.
. LEE Y J.
NUREG/lA 0030 ASSESSMENT OF RELAP5/ MOD 2 CODE USING MUNRO,P.S.
LOSS OF OFFSITE POWER TRANSIENT DATA OF KNU pl PLANT, NUREG/CR-4753 V03: CANADIAN SEISMIC AGREEMENT. Annual j
Report 1987 1988.
.j LIB Y,A.L.
NUREG/C44753 V04: CANADIAN ' SEISMIC AGREEMENT. Annual j
NUREG/C45001: EFFECS OF MANUFACTURING VARIABLES ON Report 1988-1989.-
PERFORMANCE OF HIGH-LEVEL WASTE LOW CARDON STEEL CONTAINERS.
MUTH T.R.
NUREG/C45001: EFFECTS OF MANUFACTURING VARIABLES ON LIGON.D.M.
PERFORMANCE OF HIGH LEVEL WASTE LOW CARBON STEEL NUREG/CP.0108. PUBUC WORKSHOP ON NUCLEAR POWER PLANT CONTAINERS.
UCENSE RENEWAL NAKAMURA,T.
LONG,L.T.
NUREG/CR-5480- DATA
SUMMARY
REPORT FOR FISSION PRODUCT NUREG/CR-5258 V02: GEORGIA / ALABAMA REGIONAL SEISMO.
RELEASE TEST VI-3.
GRAPHO NETWORK. Annual Report. July 1986 June 1987.
NEILL,A.P.
G i393: THE INCINERATON OF -LOW. LEVEL RADCACTIVE THE FT E PER ME T P-WASTE.A Report For The Adasory Committee On Nuclear Waste.
LORENZ R.A N*
~
NUREG/CR5480: DATA
SUMMARY
REPORT FOR FISSION PRODUCT WEG/C45513. m AMEW NGEM ENON RELEASE TEST VL3' NEEDS. Volume 1 - Methodology Development And Application To A 7
Pressunrod Water Reactor (PWR) With A Large. Dry Containment i
LOYELACE.WJL NUREG/C45513 V02: ACCOENT MANAGEMENT INFORMATION NUREG-0020 V14 NO2: UCENSED OPERATING REACTORS STATUS NEEDS. Volume 2. Appendices. -
SUMMARY
REPORT. Data As Of January 31,1990.(Gray Book I)
NICHOLS,M.
LYONS.J.A.
NUREG/C45437: ORGANIZATON AND SAFETY IN NUCLEAR POWER NUREG/C44753 VU3: CANADIAN SEISMC AGREEMENT. Annual PLANTS.
Report 199/.1988.
NUREG/CR-4753 V04. - CANADIAN SEISMIC AGREEMENT. Annual O'KELLEY,0.D.
Report 1988-1989.
NUREG/C45463: EFFECTS OF MINERALOGY ON SORPTON OF STRONTIUM AND CESIUM ONTO CALCO HILLS TUFF.
MARCUS,A.
NUREG/CR-5437. ORGANIZATION AND SAFETY IN NUCLEAR POWER O'NEAL,8.L
-PLANTS.
NUREG/C45517: IMPACTS RRC. VERSION 2.0. Program User's Manual
Personal Author index 23 OL90NJ.
REECE,WJ.
NUREG/CR-5437: ORGANIZATON AND SAFETY IN NUCLEAR POWER -
NUREG/CR-4639 V01 R1: NUCLEAR COMPUTERIZED LIBRARY FOR PLANTS.
ASSESSING REACTOR RELIABILITY (NUCLARR). Summary Desertp.
ORTIZ,NA i
NUREG/C41067: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL -
RITCHIE,L.T.
OF RADIOACTIVE WASTE. Scenario Selection Proce&e.
NUREG/CR-4691 V01: MELCOR ACCIDENT CONSEQUENCE CODE SYSTEM (MACCS). Volume 1: User's Guide.
gg NUREG/CR 5437: ORGANIZATION AND SAFETY IN NUCLEAR POWER ROTH.E.E PLANTS.
NUREG/CR-5213 V01: THE COGNITIVE ENVIRONMENT SIMULATON AS A TOOL FOR MODELING HUMAN PERFORMANCE AND OSSORNE,EF.
l NURE C DATA
SUMMARY
REPORT FOR FISSON PRODUCT LB TY'13 52 H
NITIVE ENVIRONMENT SIMULATON l
AS A TOOL FOR MODELING HUMAN PERFORMANCE AND PALMERJ.A.
RELIABILITY. Man Peport NUREG/CR-5399: SURVEY OF STATE AND TRIBAL EMERGENCY RE.
PONSE PABILITIES FOR RADOLOGICAL TRANSPORTATON A
EG/ 5111: INTEGRATED RELIABILITY,AND RISK ANALYSIS SYSTEM (IRRAS) VERSON 2.0 USER'S GUIDE.
PARKJ.Y.
NUREG/CR-4667 V07: ENVIRONMENTALLY ASSISTED CRACKING IN RUTHER,W.E.
LIGHT WATER REACTORS. Sermannual Report.Apr5.Seplemher 1988.
NUREG/CR 4667 V07: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR-4061 V08: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semiannual Report. April-September 1988.
LIGHT WATER REACTORS. Sermannual Report. October 1968. March.
NUREG/CR4667 V06. ENVIRONMENTALLY ASSISTED CRACKING IN :
1989.
LIGHT WATER REACTORS. Sermannual Report,0ctober 1988. March PARRY,G.
NUREG/CR-5460: A CAUSE. DEFENSE APPROACH TO THE UNDER-SATTISON.KS.
STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
NUREG/C44550 V7 R01: ANALYSIS OF CORE DAMAGE FREOVEN-CY: ZION UNIT 1. INTERNAL EVENTS.
N E /CR 5553: COMPUTER PROGRAMS FOR EDOY CURREPR YS E IR S) VERSION 2 0 USER DE DEFECT STUDIES.
SAWYER,C.
NUREGM WENSEE USE OF TACH NCISE RESM NU
/CR-5460: A CAUSE-DEFENSE APPROACH TO THE UNDEg STANDING AND ANALYSIS OF COMMON CAUSE FAILURES, SCALZO,8.E NUREG/CR-5510: EVALUATONS OF CORE MELT FREQUENCY EF.
pgttgggggg g'023 V01: ASSESSMENT OF TRAC-PFI/ MOD 1 VERSION NUREG/II-0 FECTS DUE TO COMPONENT AGING AND MAINTENANCE.
14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW.
SCARSROUGH,T.G.
NU i
23 2 S M T
PF1/ MOD 1 VERSON NUREG 1352: ACTON PLANS FOR MOTOR OPERATED VALVES AND 14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW.
DOWN EXPERIMENTS. Volume 2: Figures.
SCOTT,W.
PELTO.P.
NUREG/CR-5437: ORGANIZATON AND SAFETY IN NUCLEAR POWER NUREG/C45437: ORGANIZATON AND SAFETY IN NUCLEAR POWER PLANTS.
SERNE,T.J.
PETERSEN.K.E NUREG/C45548: REVIEW OF GEOCHEMICAL PROCESSES AND NUREG/CR-5560 AGING OF NUCLEAR PLANT RESISTANCE TEM.
CODES FOR ASSESSMENT OF RADIONUCLIDE MIGRATON PO-PERATURE DETECTORS.
TENTIAL AT COMMERCIAL LLW SITES.
PLOUFFE.E SETH,S.S.
NUREG/CR-4753 V03: CANADIAN SEISMIC AGREEMEPR. Annual -
NUREG/CP-0100: PUBLIC WORKSHOP ON NUCLEAR POWER PLANT NUR C 53 V04: CANADIAN SEISMIC AGREEMENT. Annual Report 1968-1969.
SHACK WJ.
NUREG/CR-4667 V07 ENVIRONMENTALLY ASSISTED CRACKING IN E C45213 V01: THE COGNITIVE ENVIRONMENT SIMULATON N E /CR-4 7 RON LY SS S D CRA NG N f'^"
"""*I UABILITY Ex utive NUREG/CR-5213 V02: THE ITIVE ENVIRONMENT SIMULATON AS A TOOL FOR MODELING HUMAN PERFORMANCE AND SHANNON.W.E.
RELIABILITY. Main Report NUREG/CR 4753 V03: CANADIAN SEISMIC AGREEMENT. Annual Report 1987 1968.
RASMUSON.D'5460: A CAUSE DEFENSE APPROACH TO THE UNCER.
NUREG/CR-NUREG/CR-4753 V04: CANADIAN SEISMIC AGREEMENT. Annual STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
Report 988 4 89 j
RASMUSON.D.E SHEN,Y.
4 NUREG/C45111: INTEGRATED RELIABILITY AND RISK ANALYSIS.
NUREG/C45436 V01: THE DEVELOPMENT AND EVALUATION OF SYSTEM (IRRAS) VERSION 2.0 USER'S GUIDE.
PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH MAINTENANCE AT NUCLEAR POWER PLANTS, Main Report.
REDOEN,0.
NUREG/C45436 V02: THE DEVELOPMENT AND EVALUATON OF NUREG/C45547; APPLICATION OF SURFACE COMPLEXATON PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH MODELS FOR RADIONUCUDE ADSORPTION. Sensitivity Analysis Of MAINTENANCE AT NUCLEAR POWER PLANTS. Appendices.
Model input Parameters.
REECE.W.D.
NUREG/CR-5473; INCLUSION OF UNSTABLE DUCTILE TEARING AND NUREG/CR 5540-PERFORMANCE TESTING OF EXTREMITY EXTRAPOLATED CRACK. ARREST TOUGHNESS DATA IN PWR DOSIMETERS. STUDY 2.
VESSEL INTEGRITY ASSESSMENT.
i i
24 Personal Author Index SMITH.R.E.
YO,T.V.
NUREG/C45475: MODEL FEASIBILITY STUDY OF RADIOACTIVE NUREG/CR-5397: VALUE.lMPACT ANALYSIS OF REGULATORY 09 PATHWAYS FROM ATMOSPHERE TO SURFACE WATER-TONS FOR RESOLUTON OF GENERIC ISSUE C-8.MSIV LEAKAGE AND LCS FAILURE.
SMYTH.J.D.
NUREG/C45523; DEVELOPMENT OF AN INFILTRATON EVALUA.
VON WINTERFELDT TON METHODOLOGY FOR LOW.LEVFL WASTE SHALLOW LAND HUREG/CR-5411: ELICITATION & USE OF EXPERT JUDGMENT IN BURIAL SITES.
PERFORMANCE ASSESSMENT FOR HIGH-LEVEL RADOACTIVE SOPPET,W.K.
WASTE REPOSITORIES.
NUREG/C44867 V07: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Semannual Report, April-September 1988.
W AHL.K K.
NUREG/C44667 V08. ENVIRONMENTALLY AS$1STED CPACKING IN NUREG/CR 5393: A REVIEW OF TECHNIQUES FOR PROPAGATING LIGHT WATER REACTORS. Semannual Report,0ctober 1988. March DATA AND PARAMETER UNCERTAINTIES IN HIGH. LEVEL RADO-1989.
ACTIVE WASTE REPOSITORY PERFORMANCE ASSESSMENT SPINDLER,5.
NUREG/lA 0023 Vol: ASSESSMENT OF TRAC.PF1/ MODI VERSION WARD LW.
t 14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW-NUREG/C45513 V01: ACCIDENT MANAGEMENT INFORMATON
'j SSENM T
1/ MODI VERSION NEEDS. Volume 1. Methodology Development And Application To A NU l
2 14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW.
NURE R 55 DE T M NAG NT N O M TON DOWN EXPERIMENTS Volume 2: Figures.
NEEDS. Volume 2. Apperdces.
SPRUNG,J.L NUREG/CR-4691 V01: MELCOR ACCIDENT CONSEQUENCE CODE WESSTER,C.S.
SYSTEM (MACCS). Volume 1: User's Guide.
NUREG/CR-5480: DATA
SUMMARY
REPORT FOR FISSION PROOUCT RELEASE TEST VI-3.
3, STALLMANN.F.W.
NUREG/CR-4816: PR.EDB: POWER REACTOR EMBRITTLEMENT WEISS A.J.
DATA BASE,VEASION 1. Program Desenption.
NUREG/CR-2331 V09 N4: SAFETY RESEARCH PROGRAMS SPON.
NUREG/CR-5409: NEUTRON EXPOSURE PARAMETERS FOR THE SORED BY OFFICE OF NUCLEAR REGULATORY METALLURGICAL TEST SPEC 1 MENS IN THE SIXTH HEAVY SEC.
RESEARCH. October December 1989.
TION STEEL IRRADIATION SERIES.
STRUCKMEYER,R"O N01: NRC TLD DIRECT RADIATON MONITORING NUREG-0837 VI NUREG/C45366: HTAS2: A THREE DIMENSIONAL TRANSIENT SHIP-7 NETWORK. Progress Report. January-March 1990.
PING CASK ANALYSIS TOOL SUMMER.R.M.
WETMILLER,R.J.
NUREG/C45475: MODEL FEASIBILITY STUDY OF RADIOACTIVE NUREG/C44753 V03: CANADIAN SEISMIC AGREEMENT. Annual t
PATHWAYS FROM ATMOSPHERE TO SURFACE WATER.
Report 1987 1988.
~
NUREG/CR-4753 V04: CANADIAN SEISMIC AGREEMENT. Annual SWINT,M.J.
- Report 1988-1989.
NUREG/C45568: EVALUATON OF HEALTH EFFECTS IN SEQUOYAH 7
FUELS CORPORATON WORKERS FROM ACCIDENTAL EXPOSURE WILLIAMS.M.L TO URANIUM HEXAFLUORIDE.
NUREG/CR 5449; DETERMINATON OF THE NEUTRON AND GAMMA FLUX DISTRIBUTON IN THE PRESSURE VESSEL AND CAVITY OF 1
^
^
^
NUREG/CR 5397. VALUE.lMPACT ANALYSIS OF REGULATORY OP.
TONS FOR RESOLUTION OF GENERIC ISSUE C-8.MSIV LEAKAGE WONG.C AND LCS FAILURE
- NUREd/CR-4753 V03: CANADIAN SEISMIC AGREEMENT. Annual TAYLOR,SJ.
Report 1987 1988.
NUREG/CR-4816: PREDB: POWER REACTOR EMBRITTLEMENT NUREG/CR-4753 V04: CANADIAN SEISMIC AGREEMENT. Annual DATA BASE,VERStON 1. Program Desenptiort Report 1988-1989.
THOMAS,J.T.
WOOOS,0.D.
NUREG/C44753 V03: CANADIAN SEISMIC AGREEMENT. Annual NUREG/C45213 VC1: THE COGNITIVE ENVIRONMENT SIMULATION i
Report: 1987 1988.
AS A TOOL FOR MODELING HUMAN PERFORMANCE AND NUREG/CR-4750 V04. CANADIAN SEISMIC AGREEMENT. Annual RELIABILITY.Executwo Sum < nary.
Report 1988-1989.
NUREG/CR-5213 V02: THE COGNITIVE ENVIRONMENT SIMULATION
^ ^
THURSER,J'R-5437: ORGANIZATON AND SAFETY IN NUCLEAR POWER NUREG/C BWMain RW PLANTS.
- WREATHALL,J.
TONG,Y C.
NUREG/C45436 V01: THE DEVELOPMENT AND EVALUATON OF NUREG/C45480: DATA
SUMMARY
REPORT FOR FISS!ON PRODUCT PROGRAM ATIC PERFORMANCE INDICATORS ASSOCIATED WITH RELEASE TEST VI-3.
MAINTENANCE AT NUCLEAR POWER PLANTS Main Report.
NUREG/C45436 V02: THE DEVELOPMENT AND EVALUATON OF TRAVIS J.R.
PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH NUREG/C45480: DATA
SUMMARY
REPORT FOR FISSION PRODUCT MAINTENANCE AT NUCLEAR POWER PLANTS.Appereces.
RELEASE TEST VI 3.
WU.S.
NU d 45439: HUMAN FACTORS ISSUES ASSOCIATED WITH AD-3 VANCED INSTRUMENTATON AND CONTROLS TECHNOLOGIES IN NUCLEAR PLANTS-ZIMMERMAN,0.A.
VESELY W.E NUREG/CR-5303. A REVIEW OF TECHNIQUES FOR PROPAGATING NUREG/C -5510: EVALUATONS OF CORE MELT FREQUENCY EF.
DATA AND PARAMETER UNCERTAINTIES IN HIGH-LEVEL RADIO-FCCTS DUE TO COMPONENT AGING AND MAINTENANCE.
ACTIVE WASTE REPOSITORY PERFORMANCE ASSESSMENT MODELS.
VILARDO,F.J.
NUREG/CR-5399: SURVEY OF STATE AND TRIBAL EMERGENCY RE.
/LLEN,E.J.
SPONSE CAPABILITIES FOR RADIOLOGICAL TRANSPORTATION NUREG/lA-0022: TRAC.PF1/ MODI POST. TEST CALCULATONS OF INCIDENTS.
THE OECD LOFT EXPERIMENT LP SB-3.
j l
l
..1 u
Subject index
\\
This index was developed from keywords and word strings in titles and abstracts. During this 1
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 are welcome.
i ACRS Reports Dolling Water Reester NUREG-1125 V11: A COMPILATON OF REPORTS OF THE ADVISORY NUREG-1372: REGULATORY ANALYSIS FOR THE RESOLUTION OF COMMITTEE ON REACTOR SAFEGUARDS.1989 Annuel.
GENERO ISSUE C 8, " MAIN STEAM ISOLATION VALVE LEAKAGE Abnormel Occurrence AND LCS FAILURE?
NUREG/CR-5397: VALUE lMPACT ANALYSIS OF REGULATORY OP.
NUREG-0000 V12 N04: REPORT TO CONGRESS ON ABNORMAL TIONS FOR RESOLUTION OF GENERIC ISSUE C 8:MSIV LEAKAGE
'~
OCCURRENCES. October Docornber 1989.
AND LCS FAILURE.
s Accident,i-.
3 NUREG/CR 5449: DETERMINATION OF THE NEUTRON AND GAMMA NUREG/C45513 V01: ACCIDENT MANAGEMENT INFORMATION FLUX DISTRIBUTION IN THE PRESSURE VESSEL AND CAVITY OF A BOlutNG WATER REACTOR.
NEEDS. Volume 14 Methodolo0y Development And Application To A -
NUREG/C45573: BORON FLUSHING DURING A BWR ANTICIPATED NU $R 55 NI NT A DE M
N MATION NEEDSVolume 2. Appendicos.
Bortc Acid Corroeien e
Accident Progression Bin NUREG/CR 5576. SURVEY OF BORIC ACID CORROS40N OF CARBON
<j NUREG/CR-5262-PRAMIS: PROBABluTY RISK ASSESSMENT MODEL STEEL COMPONENTS IN NUCLEAR PLANTS.
INTEGRATION SYSTEM. Users Gude.
Boron W Accident Soeuence NUREG/CR-5573: BORON FLUSHING DURING A BWR ANTICIPATED NUREG/CA-4550 V3 RIP 1: ANALYSIS '- OF.: CORE DAMAGE TRANSIENT WITHOUT SCRAM.
FREQUENCY:SURRY. UNIT 1,1NTERNAL EVENTS.
NUREG/CR-4550 V3 RIP 2: ANALYSIS OF CORE DAMAGE Carbon Stool FREQUENCY:SURRY, UNIT 1,1NTERNAL EVENTS APPENDICES.
NUREG/CR-5576: SURVEY OF BORIC ACID CORROSION OF CARBON y
NUREG/C44550 V5 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN-STEEL COMPONENTS IN NUCLEAR PLANTS, CY: SEQUOYAH, UNIT 1,1NTERNAL EVENTS. -
'j NUREGICR-4550 V5R1P2: ANALYSIS OF CORE DAMAGE FREQUEN.
CavWy Coolln0 System 1
CY: SEQUOYAH, UNIT 1, INTERNAL EVENTS APPENDICES.
NUREG/CR-5514: MODEUNG AND PERFORMANCE OF THE MHTGR Accident consequence Code System -
REACTOR CAVITY COQUNG SYSTEM.
?
NUREG/CR-4691 V01: MELCOR ACCIDENT. CONSEQUENCE CODE Cavity Water Seel j
SYSTEM (MACCS). Volume t: Users Guide. -
NUREG/C44525: CLOSEOUT OF IE BULLETIN 84 03:REFUEUNG q
Accidental Exposure NUREG/C45566: EVALUATION OF HEALTH EFFECTS IN SEQUOYAH Coelum FUELS CORPORATION WORKERS FROM ACCIDENTAL EXPOSURE NUREG/CR-5463: EFFECTS OF MINERALOGY ON SORPTION OF.
STRONTIUM AND CESIUM ONTO CAUCO HILLS TUFF.
A04ng Check Velve NUREG/CR 5560: AGING OF NUCLEAR PLANT RESISTANCE TEM-PERATURE DETECTORS.
NUREG-1352: ACTION PLANS FOR MOTOR 4PERATED VALVES AND -
Aging And Maintenance NUREG/C45510 EVALUATONS OF CORE MELT FREQUENCY EF*
Chemleel Toaletty FECTS DUE TO COMPONENT AGING AND MAINTENANCE.
- NUREG 1391 DRFT FC: CHEMICAL TOXICITY OF URANIUM HEXA-FLUORIDE RELATED TO RADIATION DOSES. Draft Report For Com-Aging Stroesor mont NU E /CR 5 81: NUCLEAR PLANT AGING RESEARCH:THE 1E Chronic Fetelity Weight -
NUREG/CR-5253: PARTITION-A PROGRAM FOR DEFINING THE l
Anticipated Trenaient Without Scram SOURCE TERM / CONSEQUENCE ANALYSl$ (NTERFACE IN THE NUREG/CR 5573: BORON FLUSHING DURING A BWR ANTICIPATED NUREG 1150 PROBABluSTIC RISK ASSESSMENTS. User's Guide.
BWR NUREG-1402: CLOSEOUT OF NRC BULLETIN 8845.NONCONFORM-NUREG 1372 REGULATORY ANALYSIS FOR THE RESOLUTION OF ING MATERIALS ' SUPPUED BY PIPING SUPPUES,1NC.,AT GENERIC ISSUE C-8, " MAIN STEAM ISOLATION VALVE LEAKAGE FOLSOM,NEW JERSEY,AND WEST JERSEY MANUFACTURING ANO LCS F AILURE."
COMPANY AT WILUAMSTOWN,NEW JERSEY.
NUREG/CR5307: VALUE lMPACT ANALYSIS OF REGULATORY Op.
NUREG/C44525: CLOSEOUT OF IE BULLETIN 84-03.REFUEUNG TONS FOR RESOLUTION OF GENERIC ISSUE C-8:MSIV LEAKAGE CAVITY WATER SEAL AND LCS FAILURE.
NUREG/CR-5449: DETERMINATON Or THE NEUTRON AND GAMMA Common Cause Fellure 1"
FLUX DISTRIBUTION IN THE PRESSURE VESSEL AND CAVITY OF NUREG/C45460: A CAUSE-DEFENSE APPROACH TO THE UNDER-A BOlulNG WATER REACTOR.
STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
NUREG/CR-5573: BORON FLUSHING DURING A BWR ANTICIPATED TRANSIENT WITHOUT SCRAM, Computer Code Seiow ReguietorY Concern NUREG/C44639 V01 R1: NUCLEAR COMPUTERIZED UBRARY FOR ASSESSING REACTOR REUABiUTY (NUCLARR). Summary Desenp-NUREG/CR-5517: IMPACTS BRC, VERSION 2.0. Program Users Manuel.
- tlon, 25
26 Subject index Computer Preyam NUREG-0940 V00 N01: ENFORCEMENT ACTONS: SIGNIFICANT AC.
NUREG/CR 5553: COMPUTER PROGRAMS FOR EDOY CURRENT TIONS RESOLVED.Quarterty Progrees Report Jonuary March 1990.
DEFECT STUDIES.
ENtremity Doelmotor Core Demoge NUREG/CR-5540: PERFORMANCE TESTING OF EXTREMITY NUREG/CR 4550 V3'11P1: ANALYSIS OF CORE DAMAGE DOSIMETERS STUDY 2.
FREQUENCY:SURRY UNrt 1,tNTERNAL EVENTS.
NUREG/CR-4550 V3h1P2: ANALYSIS OF CORE DAMAGE Plasion Product FREQUENCY:SUARY, UNIT 1,1NTERNAL FVENTS APPENDCES.
NUREG/CR-5480: DATA
SUMMARY
REPORT FOR FISSION PRODUCT NUREG/CR4550 V5H1P1: ANALYSIS OF CORE DAMAGE FREQUEN' RELEASE TEST VI-3.
CY: SEQUOYAH UNIT 1,1NTERNAL EVENTS.
NUREG/CR-4550 V5RtP2: ANALYSIS OF CORE DAMAGE FREQUEN*
Fracture CY: SEQUOYAH UNIT 1,1NTERNAL EVENTS APPENDICES NUREG/CR 4550 V7 Rot: ANALYSIS OF CORE DAMAGE FREQUEN.
NUREG/CR-5409: NEUTRON EXPOSURE PARAMETERS FOR THE METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY-SEC-CY; ZION, UNIT 1, INTERNAL EVENTS.
TION STEEL 1RRADIATION SERIES.
Core teelt NUREG/CR-5510: EVALUATIONS OF CORE WELT FREQUENCY EF.
Procture Toughnese FECTS DUE TO COMPONENT AGING AND MAINTENANCE.
NUREG/CR-5494: CORRELATION OF IRRADIATION-INDUCED TRAN-SITION TEMPERATURE INCREASES FROM C(V) AND K(JC)/K(IC)
Creek Arrest DATAFinal Report.
NUREG/CR 5400 NEUTRON EXPOSURE PARAMETERS FOR THE METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY SEC.
PuolDemage TION STEEL IARADIATION SERIES.
NUREG/CR 5480 DATA
SUMMARY
REPORT FOR FISSION PRODUCT NUREG/CR5473: INCLUSON OF UNSTA8LE DUCTILE TEARING AND RELEASE TEST VI-3.
EXTRAPOLATED CRACK ARREST TOUGHNESS DATA IN PWR VESSEL INTEGRITY ASSESSMENT, Puol Performance NUREG/CR,3950 V06: FUEL PERFORMANCE ANNUAL REPORT FOR Creek Growth 1968.
NUREG/CR4667 V07: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER REACTORS. Sementmel September 1968.
Gemme Plus NUREG/CR4067 V00: ENVIRONMENTALLY I TED CRACKING IN NUREG/CR5449: DETERMINATION OF THE NEUTRON AND GAMMA LIGHT WATER REACTORS. Semannual Report,0ctober 1988. March FLUX DISTRIBUTION IN THE PRESSURE VESSEL AND CAVITY OF 1989-A BOlulNG WATER REACTOR.
Critical Flow Generk leeue C 4 NUREG/lA4023 V01: ASSESSMENT OF TRAC-PF1/ MOO 1 VERSON NUREG/C&5307: VALUE lMPACT ANALYSIS OF REGULATORY OP.
14.3 USING SEPARATE EFFECTS CRIT: CAL FLOW AND BLOW-TiONS FOR RESOLUTON OF GENERIC ISSUE C 8:MSly LEAKAGE DOWN EXPERIMENTS Volume 1: Text And Tables NUREG/lA 0023 V02-ASSESSMENT OF TRAC PF1/ MOO 1 VERSON AND LCS FAILURE
- 14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW' Generic estety leeue C-8 DOWN EXPERIMENTS, Volume 2-Fgures.
NUREG 1372: REGULATORY ANALYSIS FOR THE RESOLUTION OF Decay Heat Removal GENERO ISSUE C 8, " MAIN STEAM ISOLATION VALVE LEAKAGE NUREG/CR5514: MODELING AND PERFORMANCE OF THE WHTGR AND LCS FAILURE."
REACTOR CAVITY COOLING SYSTEM.
OmhemioW Pres Depleted urentum
. NUREG/CR-5548: REVIEW OF GEOCHEMICAL PROCESSES AND 3
NUREG-1405: INADVERTENT SHIPMENT OF A RADIOGRAPHIC CODES FOR ASSESSMENT OF RADIONUCUDE MIGRATON PO-SOURCE FROM KOREA TO AMERSHAM TENTIAL AT COMMERCIAL LLW SITES.
-)
CORPORATION.BURLINGTON,M ASSACHUSETTS.
g D600tel System NUREG/CR 1667: RISK METHODOLOGY FOR GEOLOGIC DISPOSAL NUREG/CR-5439: HUMAN FACTORS ISSUES ASSOCIATED WITH AD-OF RADIOACTIVE WASTE. Scenerto Selection Procedure.
VANCED INSTRUMENTATION AND CONTROLS TECHNOLOGIES IN NUCLEAR PLANTS.
Health Effect NUREG/CR5566: EVALUATION OF HEALTH EFFECTS IN SEQUOYAH Does Estemete FUELS CORPORATION WORKERS FROM ACCIDENTAL EXPOSURE q
NUREG/C&5489: BIOLOGICAL CHARACTERIZATON OF RADIATION TO URANIUM HEXAFLUORIDE.
EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL.
ING EFFLUENTS.
Heat Trenefer NUREG/CR-5366: HTAS2: A THREE DIMENSIONAL TRANSIENT SHIP-Earthquake PfNG CASK ANALYSIS TOOL.
NUREG/CR5258 V02: GEORGIA / ALABAMA REGIONAL SEISMO-GRAPHIC NETWORK. Annual Report. July 1966 June 1D87.
Hith Level Radioactive Weste Repoettery NUREG/CR 5393: A REVIEW OF TECHNIQUES FOR PROPAGATING E M""*"'
DATA AND PARAMETER UNCERTAINTIES IN HIGH LEVEL RADIO-NUREG/C&5553: COMPUTER PROGRAMS FOR EDDYCURRENT ACTIVE WASTE REPOSITORY PERFORMANCE ASSESSMENT DEFECT STUDIES.
MODELS.
Embrtttlement NUREG/CR 5411: ELICITATION'& USE OF EXPERT JUDGMENT IN -
NUREG/CR-4816: PR EDB: POWER REACTOR EMORITTLEMENT PERFORMANCE ASSESSMENT FOR HIGH LEVEL RADIOACTIVE DATA BASE. VERSION 1. Program Descripton.
WASTE REPOSITORIES.
Emergency Reeponse High4 met Weste NUREG/CR-5399: SURVEY OF STATE AND TRIBAL EMERGENCY RE.
NUREG/CR5001: EFFECTS OF MNIUFACTURING VARIABLES ON SPONSE CAPABILITIES FOR RADIOLOGICAL TRANSPORTATION PERFORMANCE OF HIGH LEVEL WASTE LOW CARBON STEEL INCIDENTS.
COMAINERS. -
Emergency Reeponse Date System Hunwi Error Probability NUREG-1394: EMERGENCY RESPONSE DATA SYSTEM (ERDS) IM.
NUREG/CR 5438: BASIC CONSIDERATIONS IN PREDICTING ERROR PLEMENTATON.
PROBABILITIES IN HUMAN TASK PERFORMANCE.
Enforcement Action Human Factor NUREG-0940 V06 N04: ENFORCEMENT ACTIONS:SIGNIFICANT AC.
NUREG/CR-5213 V01: THE COGNITIVE ENVIRONMENT SIMULATION TIONS RESOLVED.Ouarterly Progress Report, October December AS A TOOL FOR MODELING HUMAN PERFORMANCE AND 1989.
. REllADIUTY.Executrve Summary.
Sut(ectindex N
NUREG/CR-6213 V02: THE COGNITIVE ENVIRONMENT SIMULATION NUREG/CR-2000 V09 h<
U $NSEE EVENT REPORT (LER)
AS A TOOL FOR MODEUNG HUMAN PERFORMANCE AND COMPILATION For Manth. * *c.1990 REUABtVTY.Mem Report NUREG/CR 2000 V00 N6: uCENSEE EVENT REPORT (LER)
NUREG/CR4439. HUMAN FACTORS ISSUES ASSOCLATED WITH AD-COMPILATION.For Mernh 01 Hey 1990.
VANCED INSTRUMENTATION AND CONTROLS TECHNOLOGIES IN NUCLLAR PLANTS.
LOCA Hwnen Reesheny NUREQ/LA4021: RELAPS/ MOO 2 CALCULATIONS OF OECD LOFT TEST LP 88 2 NUREG/CR4218 V01: THE COGNITIVE ENVIRONMENT SIMAATION AS A TOOL FOR MODELING HUMAN PERFORMANfM AND LOFT REUABILITY.Enocupve Summary.
NUREG/lA4031: ICAP ASSESSMENT OF RELAP6/ MOD 2, CYCLE NUREQ/OR42t3 V02: THE CO3NITNE ENVIRONMENT SIMULATION 36.06 AGAINST LOFT SMALL BREAK EXPERIME*17 LS 7.
AB A TOOL FOR MODEUNG HUMAN PERFORMANCE AND REUABluTY. Main Report LWR Nonen Rehabliny Date NUREG/CRae67 V07: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4639 V01 R1: NUCLEAR COMPUTERIZED UBRARY FOR UGHT WATER RE ACTORS. Sermannual Report.Apr"i _ _ 1998.
NUREGICR4067 V08: ENVIRONMENTALLY ASS'5TED 6 RACKING IN ASSESSING REACTOR REUABluTY (NUCLARR).Bumme'y Descrip-bon-UGHT WATER REACTORS. Sermannual ReportOctober 1988. March 1989.
Nunen Teek Portermance NUREG/CR4430: BASIC CONSIDERATIONS IN PREDICTING ERROR W Centrol System PROBABluTIES IN HUMAN TASK PERFORMANCE.
NUREG 1372: REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERC ISSUE C-8,
- MAIN STEAM ISOLATION VALVE LEAKAGE ICAP Program AND LCS FAILURE?
NUREG/lA 0023 V01: ASSESSMENT OF TRAC PFI/ MODI VER$10N NUREG/CR4397: VALUE tMPACT ANALYSIS OF REGULATORY OP-14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW.
TlONS FOR RESOLUTION OF GENERIC ISSUE C 8 MStV LEAKAGE DOWN EXPERIMENTS Volume tient And Tables AND LCS FAILURE.
NUREQ/lA0023 V02: ASSESSMENT OF TRAC PF1/ MOD 1 VERSION USING SEPARATE ECTS CRITICAL FLOW AND BLOW-A V30 102: INDEXES TO NUCLEAR REGULATORY COM-MISSION i&BUANCES.Jul December 1989.
I tg Ngoun8443 NUREG-0760 V30 N06. N CLEAR REGULATORY COMMISSION IS-J NUREG/CR 4525: CLOSEOUT OF IE BULLETIN E443.REFUEUNG SUANCES FOR DECEMBER 1969. Posee 709 811.
CAVITY WATER SEAL NUREG 0750 V31 N01: NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR JANUARY 1990. Pe9ee 1130 OE Peerer System NUREG4760 V31 N02: NUCLEAR REGULATORY COMMIS$10N IS-NUREG/CR4101: NUL; LEAR PLANT AGING RESEARCH.THE 1E SUANCES FOR FEBRUARY 1990. Pesos 131196.
POWER SYSTEM.
NUREG 0760 V31 NO3; NUCLEAR REGULATORY COMMISSION IS-SUANCES FOR MARCH 1990.Pe9ee 107 332.
IIAPACTSSRC NUREG 0760 V31 N04: NUCLEAR REGULATORY COMMISSION IS-NIIREG/CR 6517 IMPACTS-BRC,VER$10N 2.0 Program Users Manual.
SUANCES FOR APRIL 1990.Pepee 333470.
WRAE NUREG/CR 6111: INTEGRATLD RELLABluTY AND RISK ANALYSIS Usense Regulation SYSTEM (IRRAS) VERSION 2.0 USER'S GUIDE.
NUREG/CP4106: PUBUC WORKSHOP DN NUCLEAR POWER PLANT UCENSE RENEWAL 1880"
- NVAEG 1393: THE INCINERATION OF LOW LEVEL RADIOACTIVE Uoonee Menswel WASTE.A Report For The Advmory Committee On Nuclear Weste.
NUREG/CP 0108: PUBLIC WORKSHOP ON NUCLEAR POWER PLANT UCENSE RENEWAL
/CR 6523: DEVELOPMENT OF AN INFILTRATION EVALUA-T MET OGY FOR LOW LEVEL WASTE SHALLOW LAND 0
N02: LICENSED FUEL FACluTY STATUS REPORT. inventory Detterence Do's.Juh 1998. June 1969(Grey Book 11) instrumentotton And Control NUREG/CR4439. HUMAN FACTORS ISSUES ASSOCIA1ED WITH AD" VANC INSTR ENTATION AND CONTROLS TECHNOLOGIES IN G40 ENSED OPERATING REACTORS STATUS
SUMMARY
REPORT. Dele As Of January 31,1990.(Grey Book 1)
Integretten SyMem Woonese Event Report NUREG/CR4262: PRAMIS: PROBABluTY RISK ASSESSMENT MODEL NUREG/CR-2000 V00 N3. UCENSEE EVENT REPORT (LER)
INTEGRATION SYSTEM User s Guide.
NU G/C V
N4 N
EVENT REPORT (LER) inventory Difference Date COMPILATION;For Month Of Apre 1990 NUREG-0430 V09 No2: UCENSED FUEL FACluTY STATVS NUREG/CR 2000 V09 N5: UCENSEE EVENT REPORT (LER)
REPORT. inventory Difference Date. July 1988. June 1980.(Grey Book COMPILATION.For Month Of May 1990.
"I Ught Water Reester irredlet6on Effect NUREG/CAde67 V07: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR4409: NEUTRON EXPOSURE PARAMETERS FOR THE UGHT WATER REACTORS. Sermannual Report. April-September 1988, METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY SEC.
NUREG/CR4867 V08: ENVIRONMENTALLY ASSISTED CRACKING IN TION STEEL IRRADIATION SERIES.
LIGHT WATER REACTORS. Sermannual Report October 1988. March 1989.
Jet L _ _..; Load NUREGICR 5579: VALUE/ IMPACT ASSESSMENT OF JET IMPINGE.
Low Level Weste Desposal MENT LOADS AND PIPE TO PIPE IMPACT DAMAGE.Revoed Meth.
NUREG/CR4548. REVIEW OF GEOCHEMICAL PROCESSES AND ods And Crotone.
CODES FOR ASSESSMENT OF RADIONUCUDE MIGRATION PO-TENTIAL AT COMMERCIAL LLW SITES.
NUREG/lA 0032: ASSESSMENT OF RELAP5/ MOD 2 CYCLE 36.04 Low Level Radleesteve Weeto j
USING LOFT LARGE BREAK EXPERIMENT L2 5.
NUREG 1993: THE INCINERATION OF LOW LEVEL RADIOACTIVE WASTE.A Report For The Advteory Committee On Nuclear Weste.
NUREG/CR 2000 V00 N3. UCENSEE EVENT REPORT (LER)
Low-Level Radiosotive Weste Disposal COMPILATION For Month Of March 1900.
NUREG/CR 6617; IMPACTS BRC, VERSION 2.0. Program User s Manuel, j
f i
CD Subject index Low tmi weet.
Orseniention chart NUREQ/CR4623 DEVELOPMENT OF AN INFILTRATION EVALUA-NUREG 0326 RIS. U.S. NUCLEAR REGULATORY COMMIS$10N FUNC.
ON METHODOLOGY FOR LOW. LEVEL WASTE SHALLOW LAND TONAL ORGANIZATON CHARTS. April 1,1DDD.
BURIAL SITES.
PRA Luminescence NUREG/CR4660 V3R1P1: ANAlvSIS OF OORE DAMAGE NUREG/CR4674: DETERMINATON OF THE CHEMCAL FORM OF FREOVENCY:SURRY, UNIT 1,1NTERNAL EVENTS.
TRITIUM IN BELF LUMINOUS SIGNS.
NUREG/CR4660 V3 RIP 2; ANALYSIS OF CORE DAMAGE FREQUENCY:S71RY. UNIT 1,1NTERNAL EVENTS APPENDCES.
WELCOR NUREG/CR4660 V6R1P1: ANALYSl$ OF CORE DAMAGE FRFOUEN.
NUREG/CR4691 V01: MELOOR ACCIDENT CONSEQUENCE CODE CL SEOVOYAH. UNIT 1,1NTERNAL EVENTS.
SYSTEM (MACCS). Volume 1: User's GuKle-NUREG/CR4660 V6R1P2. ANALYSIS OF CORE DAMAGE FREQUEN-CY: SEQUOYAH. UNIT 1,1NTERNAL EVENTS APPENDCES.
NU
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^^
U EG R 614: MODEUNG AND PERFORMANCE OF THE MHTGR UNT I E L EVE REACTOR CAV!TY OOOUNG SYSTEM-NUREG/CR'4263. kARTITON A PROGRAM FOR DEFINING THE SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE gggy 4 NUREG 1160 PROBABluSTIC RISK ASSESSMENTS User's Gulde NUREG/CR4397: VALUE lMPACT ANALYSIS OF REGULATORY Op.
TIONS FOR RESOLUTON OF GENERIC ISSUE C.B.MSIV LEAKAGE pggg,,,
AND LCS F ALLURE.
NUREG/CR4P62. PRAMIS: PROBADILITi RISK ASSESSMENT MODEL Maintenance Appreach INTEGRATON SYSTEM. User's Guide.
NUREG 1333. MAINTENANCE APPROACHES AND PRACTCES IN SE.
PWR LECTED FOREIGN NUCLEAR POWER PROGRAMS AND OTHER NUREG/CR 6US INCLUSION OF UNSTABLE DUCTILE TEARING AND U.S. INDUSTRIES: REVIEW AND LESSONS LEANNED.
EXTRAPOLATED CRACK ARREST TOUGHNESS DATA IN PWR M6d Loop Operst6en VESSEL INTEGRITY ASSESSMENT, NUREG.1410. LOSS OF VITAL AC POWER AND THE RESIDUAL HEAT NUREG/CR4613 V01: ACCIDENT MANAGEMENT INFORMATON REMOVAL SYSTEM DURING MID-LOOP OPERATIONS AT VOGTLE NEEDSVolume 1. Methodology Development And Apg4ceton To A UNIT 1 ON MARCH 20,1990.
Proesurtrod Water Reactor (PWR) With A Large, Dry Containment NUREG/CR4613 V02: ACCIDENT MANAGEMENT INFORMATION M6ner 60py NEEDSvolume 2. Appereces.
NUREG/CR4463: EFFECTS OF MINERALOGY ON SORPTION OF NUREG/CR4667: RELAPS THERMAL HYDRAULIC ANALYSIS OF THE STRONilUM AND CESIUM ONTO CALCO HILLS TUFF.
$NUPPS PRESSURIZED WATER REACTOR.
Motor Operated Velve Performenee indlestor NUREG 1362: ACTION PLANS FOR MOTOR OPERATED VALVES AND NUREG/CR4436 V01: THE DEVELOPMENT AND EVALUATION OF CHECK VALVES.
PROGRAMMATIC PERFORMANCE INDICATORS ASSOCLATED WITH MAINTENANCE AT NUCLEAR POWER PLA4TS Main Report NRC Sulleth $$ 06 NUREG/CR4436 V02: THE DEVELOPMENT AND EVALUATION OF N'JREC 14N CLOSEOUT OF NRC BULLETIN 88 06:NONCONFORM-PROGRAMMATC PERFORMANCE INDICATORS ASSOCIATED WITH ING (AATERIALS SUPPUED BY PtPING SUPPLIES.fNC.,AT
^
FOLSOM.NEW JERSEY,AND WEST JERSEY MANUFACTURING NU E ZATO D SA TY i R POWER COMPANY AT WILLIAMSTOWN.NEW JERSEY.
PLANTS.
NRC Operatione Center N
N O 1 94 E RGENCY RESPONSE DATA SYSTEM (ERDS) IM.
NU G O 36 01: NRC REGULATORY AGENDA.Ouarterly Report,Jenuary March 1990.
NUCLARA NUREG/CR 4630 V01 R1: NUCLEAR COMPUTERIZED UBRARY FOR Phyeneet Securny A?3ESSING REACTOR RELIABluTY (NUCLARR) Summary Desenp.
NUREG 1404: UCENSEE USE OF TACTICAL EXEhCISE RESULTS.
l ten.
Nonconforming Malertal NUREG/CR4679: VALUEllMPACT ASSESSMENT OF JET IMPINGE-NUREG-1402: CLOSEOtif OF NRC BULLETIN BB-06:NONOONFORM-MENT LOADS AND PIPE TO PIPE IMPACT DAMAGE.Revoed Meth-ING MATERIALS SUPPLIED BY PIPING SL / PLIES.INC AT oos And Criterne.
FOLSOM.NEW JERSEY,AND WEST JERSEY MANUFACTURING COMPANY AT WILUAMSTOWN.NEW JERSEY.
g
,p Nondoetructive Testing ING MATERIALS SUPPUED BY PIPING BUPPUES.INC..AT NUREG/CR4676. SURVEY OF BORC ACID CORROSION OF CARBON FOLSOM.NEW JERSEY,AND WEST JERSEY MANUFACTURING STEEL COMPONENTS IN NUCLEAR PLANTS.
COMPANY AT WILUAMSTOWN,NEW JERSEY.
Nucteer Plant Atir.9 Plant Melntonense NUREG/CR41st: NUCLEAR PLANT AGING RESEARCH.THE 1E NUREG/CR4436 V01: THE DEVELOPMENT AND EVALUATION OF POWER SYSTEM.
PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH MAINTEIONCE AT NUCLEAR POWER PLANTS Mom Report l
OECD LOFT NUREG/CR4436 V02: THE DEVELOPMENT AND EVALUATION OF NUREG/lA 0011: TRAC PF1/ MOD 1 POST. TEST CALCULATIONS OF PROGRAMMATIC PERFORMANCE INDCATORS ASSOCIATED WITH LChLATONS OF OECD LOFT MAINTENANCE AT NUCLEAR POWER PLANTS.APpereces l
N EG A R
Point Kernel NUhGriA 0022; TRAC.PF1/ MOD 1 POST TEST CALCULATIONS OF NUREG/CR4468: QADS: A MULTIDIMENSIONAL POINT KERNEL THE OECD LOFT EXPERIMENT LP SB-3 ANALYSIS MODULE.
Offene Power Trene4ent P'A***
NUREG/lA-0030: ASSESSMENT OF RELAPS/ MOD 2 CODE USING NUREG/CR-4816: PR-EDB-. POWER REACTOR EMBRITTLEMENT LOSS OF OFFSITE POWER TRANSIENT DATA OF KNU #1 PLANT, DATA BASE.VERSON 1. Program Doocnpton.
Operetnr Laconoint Eneminor NUREG 1021 R06. OPERATOR LCENSING EXAMINER STANDARDS' O
06 ITED STATES NUCLEAR REGULATORY Organization COMMISSION STAFF PRACTICE AND PROCEDURE NUREG/CR4437: ORGAN!ZATON AND SAFETY IN NUCLEAR POWER DIGEST.Commmoon, Appeal Board And Oconsen0 Soard PLANTS.
DecmonsJuly 1972. Decorr$er 1989.
M Subject index 29 Presewe veeems Memoscove Rose ee HUREQ/CR44D9. NEUTRON EXPOSURE PARAMETERS FOR THE NUREG/CR-4691 V01: MELCOR ACCIDENT CONSEQUENCE CODE METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY SEC-SYSTEM (MACCS) Volume 1: Users Guce.
TON STEEL IRRADIATION SERIES NUREG/CR4449. DETERMINATION OF THE NEUTRON AND GAMMA ReGenetive Weste FLUX DISTRIBUTON IN THE PRESSURE VESSEL AND CAVITY OF NUREG/CR 1667: RISK METHODO:.OGY FOR GEOLOGIC DISPOSAL A SOILILNG WATER REACTOR.
OF RADOACTIVE WASTE. Scenaw Selecton Procedse.
NUREGiCR4494: CORRELATION OF IRRADIATION-INDUCED TRAN-NUREG/CR4001: EFFECTS OF MANUFACTURING VARIABLES ON SITON TEMPERATURE INCREASES FROM C(V) AND KWC)/KlC)
DATA. Final Report.
PERFORMANCE OF HIGH LEVEL WASTE LOW CARBON STEEL CONTAINERS.
Precourteed Water Roseter NUREGICR 5473. INCLUSON OF UNST ABLE DUCTILE TE ARWG AND Reeegraphic Seuree NUREG 1405: INADVERTENT SHIPMENT OF A RADIOGRAPHIC EXTRAPOLATED CRACK ARREST TOUGHNESS DATA IN PWR SOURCE FROM KOREA TO AMERSHAM N E 5
1 DEN MANAGEMENT INFORMATON CORPORATON.BURLINGTON.M ASSACHUSETTS.
NEEDS. volume 1. Methodology Development And Apscaton To A Rosetesteel Tronoportetten insident NR R 55 C DE T N
T M TON NUREG/CR4399, SURVEY OF STATE AND TRIBAL EMERGENCY RE.
NEED$ volume 2. Apoendees SPONSE CAPABILITIES FOR RADIOLOGICAL TRANSPORTATON NUREG/CR4557: RELAP5 THERMAL-HYDRAULIC ANALYSIS OF THE INCOENTS.
SNUPPS PRESSURIZED WATER REACTOR.
m gg,,,,,,
Prahamme nieli Aeoseement NUREG/CR4547; APPLICATON OF SURFACE COMPLEXATON NUREG/CR4550 # RIP 1: ANALYbl$ OF CORE EW. AGE MODELS FOR RADONUCUDE ADSORPTION.Sensitwity Analysis Of FREQUENCY:SURm UNIT 1,1NTERNAL EVENTS.
Modelinput Parameters NUREG/CR4550 V3M1P2: ANALYSIS OF CORE DAMAGE NUREG/CR4546. REVIEW OF GEOCHEMICAL PROCESSES AND FREQUENCY:SURRY. UNIT 1,1NTERNAL EVENTS APPENDICES NUREG/CR 4550 VSRIP1: ANALYSIS OF CORE DAMAGE FREQUEN-CODES FOR ASSESSMENT OF RADIONUCLIDE MiGRATON PO-CY:SEOUOYAH. UNIT 1,lNTERNAL EVENTS TENTIAL AT COMMERCIAL LLW SITES.
NUREG/CR 4550 V5RtP2: ANALYSIS OF CORE DAMAGE FREQUEN-Remoter Aceteent CY: SEOVOYAH UNIT 1,1NTERNAL EVENTS APPENDICES NUREG/CR-4550 V7 RO1: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR 4691 V01: MELCOR ACCIDENT CONSEQUENCE CODE CY: ZON, UNtf 1, INTERNAL EVENTS.
SYSTEM (MACCS). Volume 1: User's Gude.
NUREG/CR4213 V01: THE COGNITIVE ENVIRONMENT SIMULATION Reeoger Thermal Nydraults AS A TOOL FOR MODELING HUMAN PERFORMANCE AND NUREG/lA 0031: ICAP ASSGSMENT OF RELAP5/ MOD 2. CYCLE NUEY/C
{3 NH COGhiTIVE ENVIRONMENT S!MULATION 36.05 AGAINST LOFT SMALL SAEAK EXPERIMENT L3 7, B'
AS A TOOL FOR MODELING HUMAN PERFORMANCE AND Refueling PAN 1$ TION. A PROGRAM FOR DEFINING THE TY WfTER S A NU
/C 525 SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE CA NUREG 1150 PPOBABluSTC RISK ASSESSMENTS. User's Gulde.
Reguistion Preheh61 sty Risk Aseseemop NUREG 1206 V04: NRC SAFETY RESEARCH IN SUPPORT OF REGU.
NUREG/CR $262. PRAMis: PROBABILITY RISK ASSESSMENT MODEL LATION. FY 1989.
INTEGRATION SYSTEM. User's Guide.
Regulatory Agende OADe NUREG 0936 V09 N01: NRC REGULATORY AGENDA Ouarterty NUREG/CR446ti: OADS: A MULTIDIMENSIONAL POINT KERNEL ReportJanuary March 1990.
ANALYSIS MODULE.
WTelW RELAPl NOREG 0304 V16 NOI: REGULATORY AND TECHNICAL REDORTS NUREG/CR 5557: RELAP5 THERMAL-HYDRAULIC ANALYSIS OF THE (ABSTRACT INDEX JOURNAL). Comp 6taton For First Quarter l
SNUPPS PRESSURIZED WATER REACTOR.
1990 January March.
RELAP5/00002 Report To Congrees NUREG/LA 0018. RELAP5/ MOD 2 ASSESSMENT,OECD LOFT SMALL NUREG-0090 V12 N04: REPORT TO CONGRESS ON ABNORMAL BRE AK EXPERIMENT LP.SB-03.
NUREG/lA 0021: RELAPS/ MOD 2 CALCULATIONS OF OECD LOFT OCCURRENCES. October. December 1969.
TEST LP SB-2.
Ree6 duel Heat Removal NUREG/tA 0030: ASSESSMENT OF RELAP5/ MOD 2 CODE USING LOSS OF OFFSITE POWER TRANSIENT DATA OF KNU #1 PLANT.
NUREG 1410. LOSS OF VITAL AC POWER AND THE RESIDUAL HEAT ICAP ASSESSMENT OF RELAP5/ MOD 2, CYCLE REMOVAL SYSTEM DURING MID. LOOP OPERATONS AT VOGTLE NUREG/lA 0031:
36 05 AGAINST LOFT SMALL BRE AK EkPERIMENT L3 7.
UNIT 1 ON MARCH 20,1990, NUREG/lA 0032: ASSESSMENT OF RELAPS/ MOD 2 CYCLE 36.04 USING LOFT LARGE BREAK EXPERtMENT L24 Rootstence Tempereture Detector NUREG/CR4560 AGING OF NUCLEAR PLANT RESISTANCE TEM-l Redehn Does PERATURE DETECTOR $.
NUREG 1391 DRFT FC: CHEMCAL TOXICITY OF URANIUM HEXA-FLUORIDE RELATED TO RADIATION DOSES. Draft Report For Com-ni,g Anoppi, gp NUREG/CR 5111: INTEGRATED RELIABILITY AND RISK ANALYSIS i
NU
/CR-5617: IMPACTS BRC,VERSON 2.0. Program User's Manual.
SYSTEM (IRRAS) VERSION 2.0 USER'S GUCE.
Remetten Embritsement Reek Methodelegy NUREG/CR4494: CORRELATON OF 1RRADIATON-INDUCED TRAN.
NUREG/CR.1667: RISK METHODOLOGY FOR GEOLOGC Di4POSAL SITION TEMPERATURE INCREASES FROM C(V) AND KpC)/K(IC)
OF RADOACTIVE WASTE. Scenario Selection Procedure.
DATA. Final Report.
Remomen Espeouro NUREG-0936 V09 N01: NRC REGULATORY AGENDAOuartaty NUREGICR4489. BIOLOGICAL CHARACTERIZATION OF RADIATION ReportJanuayMarch 1990.
EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL.
ING EFFLUENTS-Rules Of Proottee NUREG-0386 DOS R06: UNtTED STATES NUCLEAR REGULATORY Remeeotive Pethwer COMMISSION STAFF PRACTICE AND PROCEDURE NUREG/CR4475: MODEL FEASIBILITY STUDY OF RADCACTIVE DIGEST.Comm as on, Appeal Board And Licenomg Board PATHWAYS FROM ATMOSPHERE TO SURFACE W ATER.
Dece6onaJuly 1972. December IM9.
I u
n
Sub)SCl lfMR SCALE Steel Contenner NUREG/CR4366 HTAS2. A THREE DIMENSONAL TRANSIENT SHIF-NUREG/CR4001: EFFECTS OF MANUFACTURIN3 VAR 6 ABLES ON PING CASK ANALYSIS TOOL.
PERFQRMANCE OF HIGH LEVEL WASTE LOW CARBON STEEL CONTAINERS.
N 4EGICR4437: ORGAN 12ATON AND SAFETY IN NUCLEAR POWER Stroes Corton6en Crocking PLANTS.
NUREG/CR4667 V07: ENVIRONMENTALLY ASSISTED CRACKING IN LIGHT WATER RE ACTOAS Semiannual Report.Apr#-Septemter 1968.
SoWy Evolust6en Report NUREG/CR4067 V0B. ENVIRONMENTALLY ASSISTED CRACKING IN NUREG 0797 $24: E.AFETY EVALUATION REPORT RELATED TO THE LIGHT WATER REACTORS. Semannual ReportOctober 1968. March OPERATON OF COMANCHE PEAK STEAM ELECTRIC STATON 1989.
UNITS 1 AND 2 Docket Nos. 50445 And 50446.(Texas Utstes Elec-b[kFb EVALUATON REPORT RELATED TO THE RE*
N E N E CR4463. EFFECTS OF MINERALOGY ON SORPTION OF NEWAL OF THE OPERATING LICENSE FOR THE TRIGA TRAINING STRONTIUM AND CESIUM ONTO CALICO HILLS TUFF.
AND RESEARCH REACTOR AT THE UN!YER$ TTY OF ARIZONA. Docket No. 50113. (Un#versity of Artrona)
Surteee Comptometton Model NUREG/CR4547: APPLICATON OF SURFACE COMPLEXATION U EG V04 NRC SAFETY RESE ARCH IN SUPPORT Of REGU-odel P me LATON. FY 1989.
Surtece Water Bowy Moseereh Program NUREG/CR4475: MODEL FEASIBILITY STUDY OF RADCACTIVE NUREG/CR-2331 V09 N4: SAFETY RESEARCH PROGRAMS SPON-PATHWAYS FROM ATMOSPHERE TO SURFACE WATER.
BORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH October December 1989.
TLD NUREG 0837 V10 N01: NRC TLD DIRECT RADIATON MONITORING Sommac hoek NUREG/CR4753 V03 CANADIAN SEISMIC AGREEMENT. Annual NETWORK. Progress Report January March 1990.
Report 1987 1986 TMAC pF1/ MODI NUREG/CR4753 V04' CANADIAN SEISMIC AGREEMENT. Annual NUREG/lA-0011: TRAC.PF1/ MODI POST TEST CALCULATIONS OF N'P 'l 088" THE OECD LOFT EXPERIMENT LP SB-1.
Sonom4 city NUREG/lA 0022: TRAC PF 1/ MOD 1 POST TEST CALCULATONS OF NUREG'CR 753 V03 CANADIAN SEISMIC AGREEMENT Annual E C T EXP RIME
$1 g
ES p TRAC-PF1/ MOD 1 VER$lON NU CR4753 VO4: CANADIAN SEISMIC AGREEMENT. Annua 14 3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW-i Report 1968-1989.
DOWN EXPERIMENTS Volume 1: Text And Tables NUREG/LA 0023 V02: ASSESSMENT OF TRAC.PF1/ MOD 1 VERSION Setemergraph6c Network 14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW-NUREG/CR4258 V02: GEORGIA / ALABAMA REGONAL SEISMO-DOWN EXPEFilMENTS Volume 2: Fogures.
GRAPHIC NETWORK Annual Report. July 1986. June 1987.
Tact 6 cal Response Esorcise Severe Acc6 dent NUREG 1404: LICENSEE USE OF TACTICAL EXERCISE RESULTS.
NUREG/CR4513 V01: ACCIDENT MANAGEMENT INFORMATION NEEDSVolume 1 Methodology Development And Apphcation To A Technical Specificatlon Pressurtred Water Reactor (PWR) With A Large. Dry Containment NUREG 1399.
TECHNICAL SPECthCATIONS. COMANCHE PEAK Q
NUREG/CR4513 V02. ACCOENT MANAGEMENT INFORMATION STEAM ELECTRIC ST ATION. UNIT 1. Docket No. 50445, Appendix "A" NEEDS Volume 2 Appendices.
to License NPF-87.(Texas Utilities Electnc)
Shallow Land Surnal Site Telemetred Network NUREG/CR 5523. DEVELOPMENT OF AN INFILTRATION EVALUA-NUREG/CR-4753 V03: CANADIAN SEISMIC AGREEMENT. Annual TION METHODOLOGY FOR LOW-LEVEL WASTE SHALLOW LAND Report 198719B8, BURIAL SITES.
Thermal-Hydraulic Sh6 eld 6n9 NUREG/CR4557: FtELAP5 THERMAL HYDRAULIC ANALYSIS OF THE NUREG/CR 5468. CADS: A MULTIDIMENSIONAL POINT KERNEL SNUPPS PRESSURIZED WATER REACTOR.
ANALYSIS MODULL.
Thermoluminescent Doelmeter Shipment NUREG 0837 V10 N01: NRC TLD DIRECT RADIATON MONITORING NUREG 1405: INADVERTENT SHIPMENT OF A RADOGRAPHic NETWORK. Progress Report. January-March 1990.
SOURCE FROM KOREA TO AMERSHAM CORPORATION.BURLINGTON,M ASSACHUSETT S.
Yttle llet NUREG 0540 V11 N12: TITLE LIST OF DOCUMENTS MADE PUBLICLY Shipping Caek AVAILABLE. DECEMBER 1 31 1989 NUREG/CR4366-HTAS2: A THREE DIMENSONAL TRANSIENT SHIP-NUREG 0540 V12 N01: TITLE LIST OF DOCUMENTS MADE PUBLICLY PING CASK ANALYSIS TOOL-AV AILABLE. Januay 1 31,1990.
NUREG 0540 V12 NO2; TITLE LIST OF DOCUMENTS MADE PUBLICLY RE A 0031: ICAP ASSESSMENT OF RELAPS/ MOO 2, CYCLE AVAILABLE. February 1 28,1990-36.05 AGAINST LOFT SMALL BREAK EXPERIMENT L3-7.
Trttlum Small Sreek LOCA NUREG/CR4574: DETERMINATION OF THE CHEMICAL FORM OF NUREG/lA 0018. RELAP5/ MOD 2 ASSESSMENT,OECD LOFT SMALL TRmVM IN SELF LUMINOUS SIGNS.
BREAK EXPERIMENT LP SD 03 Uranium Sorpt6on NUREG/CR4489 BOLOGICAL CHARACTER 12ATION OF RADIATON NUREG/CR 5463: EFFECTS OF MINERALOGY ON SORPTON OF EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL.
STRONTIUM AND CESIUM ONTO CAUCO HILLS TUFF.
ING EFFLUENTS.
Source Term Uranium He afluoride NUREG/CR4253. PARTITION A PROGRAM FOR DEFINING THE NUREG 1391 DRFT FC-CHEMICAL TOXICITY OF URANIUM HEXA-SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE FLUORIDE RELATED TO RADIATON DOSES. Draft Report For Com-NUREG-1150 PROBABILISTIC RISK ASSESSMENTS. User's Guide.
ment NUREG/CR4566: EVALUATION OF HEALTH EFFECTS IN SEQUOYAH Special Nuclear Meternal FUELS CORPORATON WORKERS FROM ACCIDENTAL EXPOSURE NUREG 1404. LICENSEE USE OF TACTICAL EXERCISE RESULTS TO URANIUM HEXAFLUORIDE.
n
Sub$est indes 81 veses m vaal ac power NUREG/CR4473. INCLUSON OF UNSTASLE DUCTILE TEARNG AND NUmEG4410: LOSS OF Vff AL AC power AND THE RESIDUAL FAT
)
EXTRAPOLATED CRACK-AMMEST TOUGHNESS DATA IN PWR MEMOVAL SYSTEM DURNG 20 LOOP OPERATIONS AT YOGTLE l
VESSEL NTEGRffY ASSESSMENT.
UNff 1 ON MARCH 80,1900.
l i
l
I 6'r) N FI A
d
NRC Originating Organization index (Staff Reports)
This index lists those NRC organizations that have published staff reports. The index is ar.
ranged alphabetically by ma,or NRC organizations (e.g., program offices) and then by sub-sections of these (e.g., divis ons, branches) where 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.
ADVit0RY COMMITTEE (S)
EDO.0FFICE OF NUCLEAR MATERIAL SAFETY & SAFEOUARDS ADVISORY COMMITTEE ON NUCLEAR WASTE OFF6CE OF NUCLEAR MATERIAL SAFETY NUREG 1393. THE INCINERATON OF LOW LEVEL RADIOACTIVE SAFEGUARDS. DIRECTOR WAS1E.A Repor1 For The Adwoory Comrrvitse On Nuclear Waste, NUREG 0430 V09 NO2: UCF.NSED FUEL FACluTY STATUS ACRS
- ADVISORY COMMITTEE ON REACTOR SAFEGUARDS REPORT. inventory Difference DataJuly 1988 June 1989.(Gray NUREG 1125 Vit: A COMPILATION OF REPORTS OF THE ADVISO-Book 11)
RY COMMITTEE ON REACTOR SAFEGUARDS.1989 Annust Di Ig S TA TAT ST B7 p
0FFICE OF EXECUTIVE DIRECTOR FOR OPEPAtl0NS (EDD)
OFC OF THE EXECUTIVE DIRECTOR FOR OPERATIONS U.S. NUCLEAR REOULATORY COMMISSION NUREG 1410. LOSS OF VITAL AC POWER AND THE RESOUAL OF HEAT REMOVAL SYSTEM OURING MID. LOOP OPERATONS AT NU E 3 I
U T N
R REGULATORY COMMISSION STAFF PRACTICE AND PROCEDURE REG ON C
T D RECT DIGEST.Comnussen, Appeal Board And Licensm9 Board NUREG 0837 V10 N01: NRC TLD DIRELT RADlATION MONITORING NRC 'N'O D"*
8 NETWORK Prog'ese Report Janus'y March 1990.
E F T V
OFC OF ENFORCEMENT (POST 870413)
NUREG 1405. INADVERTENT SHIPMENT OF A RADIOGRAPHIC NUREG 0940 V08 N04: ENFORCEMENT ACTONS.SIGNIFICANT AC-SOURCE FROM KOREA TO AMERSHAM TIONS RESOLVED.Quarterty Pro 9ress ReporLOctober December CORPORAT ON.DURUNGT ON,M ASSACHUSETTS.
1989-NUREG 0940 V09 N01: ENFORCEMENT ACTIONS: SIGNIFICANT AC-EDO. OFFICE Ot NUCLE AR MEGULATORY RESEARCH (POST 880406)
OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 060720)
I TIONS RESOLVED.Ouarterty Pro 9tess Report. January March 1990.
NUREG 1288 V04. NRC SAFETY RESEARCH IN SUPPORT OF REG-OFC OF PERSONNEL (POST 8704136 ULATON. FY 1989 NUREG 0325 R13. U.S. NUCLEAR REGULATORY COMMISS ON NUREG 1391 DAFT FC CHEMICAL TOXICITY OF URANIUM HEXA-FUNCTONAL ORGAN 12ATON CHARTS. April 1.1990.
FLUOROE RELATED TO RADIATION DOSES. Draft Report For EDO.0FFICE OF ADM6NtSTRAfl0N (PRE 870413 & P0sf 980306)
DIVIS N REGULATORY APPLICATONS (POST 870413)
DIVLSION OF FREEDOM OF INFORMATION & PUBUCATONS SERV.
NUREG 1333. MAINTENANCE APPROACHES AND PRAC' ICES IN ICES (POST 890205 SELECTED FOREIGN NUCLEAR POWER PROGRAMS AND NUREG 0304 V15 N01: REGULATORY AND TECHNICAL REPORTS OTHER U S. INDUSTRIES REVIEW AND LESSONS LEARNED.
(ABSTRACT INDEK JOURNAL). Compdaten For First Quarter OlVISION OF SAFETY ISSUE RESOLUTON (POST 8807jt7 1990. January March.
NUREG-1372: AEGULATORY ANALYf..lS FOR THE REwLUTION OF NUREGM40 V11 N12: TITLE LIST OF DOCUMENTS MADE PUBLIC.
GENERIC ISSUE C-8. " MAIN STEhl ISOLATON VALVE LEAKAGE LY AVAILABLE. DECEMBER 1 31,1989.
AND LCS FAILURE."
LY AV LABLE J 1 1 NURE /
11 N E RA EL ABluTY AND RISK ANALYSIS LY A A LABLE F a 1 28
~
NU R
NUREG 0750 V30102: INDEXES TO ' UCLEAR REGULATORY COM-STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
N NU NU A EQU TORY COMMISSION IS-Ah0R E(C VlS O F
ECT Il V &
P S
SUANCES FOR DECEMBER 1989. Pa0es 709-811.
NUREG 0750 V31 N01: NUCLEAR REGULATORY COMMISSION IS*
(POST 870411 l
NUREG 1390: SAFETY EVALUATON REPORT RELATED TO THE SUANCES FOR JANUARY 1990. Pa9es 1 130.
NUREG0750 V31 N02: NUCLEAR REGULATORY C")MMISSION IS-RENEWAL OF THE OPERATING LICENSE FOR THE TRIGA SUANCES FOR FEBRUARY 1990. Pa9es 131195.
TRAINING AND RESEARCH REACTOR AT THE UNIVERSITY OF NUREG 0750 V31 NO3: NUCLEAR REGULATORY COMMISSON IS-ARIZONA Docket No. 50113. (Urwersity of Artrona)
SUANCES FOR MARCH 1990.Pa9es 197 332.
COMANCHE PEAK PROJECT DIVISON NUREG 0797 S24: SAFETY EVALUATION REPORT 189010190060 NUREG 0750 V31 N04: NUCLEAR REGULATORY COMMISSON IS-SUANCES FOR APRIL 1990 Pa9es 333 37*.
THE OPERATON OF COMANCHE PEAK STEAM ELECTRIC STA.
TION UNITS 1 AND 2. Docket Nos. 50-445 And 50-448.(Texas Utdb NUREG 0938 V09 N01: NRC REGULATORY AGENDA.QuartertY tes Electre Company.et al.) SPECIFICATIONS. COMANCHE Report, January March 1990.
NUREG-1399: TECHNICAL PEAK EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL h Oc NPF-87 e sE t OFFlu FOR ANALYSIS & EVALUATON OF OPERATIONAL DATA, Ol-REG. 5 iN t
OR E T D VALVES RECTOR AND CHECK VALVES.
NUREG 0090 V12 N04: REPORT TO CONG9ESS ON ABNORMAL NUREG 1402: CLOSEOUT OF NRC BULLETIN 88-05.NONCONFORM-OCCURRENCE $ October December 1989 ING MATERIALS SUPPLIED BY PIPING SUPPUES.lNC..AT NUREG 1394: EMERGENCY RESPONSE DATA SYSTEM (EROS) IM-FOLSOM.NEW JERSEY AND WEST JERSEY MANUFACTURING PLEMENTATION.
COMPANY AT WILLIAMSTOWN,NEW JERSEY.
EDO. OFFICE OF INFORMATION RESOURCES MANAGEMENT & ARM N RE / 395 L PERFORMANCE ANNUAL REPORT (POST D611001 FOR 1988.
DIYlSON OF COMPUTER & TELECOMMUNICATONS SERVICES DIVISION OF LICENSEE PERFORMANCE & OVAUTY EVALUATION (POST 890205)
(POST 870411)
NUREG 0020 V14 NO2: UCENSED OPERATING REACTORS STATUS NUREG-1021 R06: OPERATOR LICENS!NG EXAMINER STAND-
SUMMARY
REPORT. Data As Of January 31,1990.(Gray Book 1)
ARDS.
33 y
f 6
a
. I B
1 4
.m
NRC Originating Organization index (International Agreements)
This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, brancies) where appropriate. Each cntry is followod by a NUREG number and title of the report (s), if further information is needed, refer to the main citation by NUREG number.
EDO. OFFICE OF NUCLEAR REGULATORY RESEARCH T830406)
NUREG/M-0023 V02: ASSESSMENT OF TRAC-PF1/ MOD 1 VERSION Sk
^
NU EGIA40 1 01 CULA S OF N PE M.
8o Fg es N
O A40 EL P5 D
ES ENT,OECD LOFT SMALL NUREG/M@30: ASSESSMENT OF REW6/ MOD 2 CODE USING BREAK EXPERIMENT LP SB 03.
LOSS OF OFFSITE POWER TRANSIENT DATA OF F.NU #1' NU E / A RELAP$/ MOD 2 CALCULATIONS OF OECD LOFT P
pp p,,
NUREG/iA 0022. TRAC PF1/ MODI POST. TEST CALCULATIONS OF 36.05 AGAIM : ' LOFT SMALL BREAF EXPERIMENT L3 7.
THE OECD LOFT EXPERIMENT LP-SB 3 NUREG/lA 0032: ASSESSMENT OF RELAP5/ MOD 2 CYCLE 36 04 NUREG/LA4023 VOI: ASSESSMENT OF TRAC-PF1/ MOD 1 VERSION USING LOFT LARGE BREAK EXPERIMENT L2 5.
14.3 USING SEPARATE EFFECTS CRITICAL FLOW AND BLOW.
DOWN EXPERIMENTS Volume 1. Text And Tab 6es.
35 f
n n
i j
e j
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) and then by subsections of these (e.g., divisions) where appropriate. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) pre 3ared by that organi-zation. If further information is needed, refer to the main citation by the 9UREG/CR number, EDO 0FFICE POR ANALYSIS & EVALUATION OF OPERATIONAL NUREG/CR 4867 V08. ENVIRONMENTALLY ASSISTED CRACKING OATA IN LIGHT WATER REACTORS. Semiannual Report. October 1988 OFFICE FOR ANALYSIS & EVALUATON OF OPERATONAL DATA, Dl-March 1969.
RECTOR NUREG/CP 4753 V03: CANAD6AN SEISMC AGREEMENT. Annual NUREG/CR 2000 V09 N3: LCENSEE EVENT REPORT (LER)
Report ICt1968.
COMPILATON For Month Of March 1990.
NUREG/Cd 4753 V04: CANADIAN SEISMC AGREEMENT. Annual NUREG/CR 2000 V09 N4: LCENSEE EVENT REPORT (LER)
Report 1988 1969 COMPILATION.For Month Of April 1990.
NUREG/CR 4818: PR-EDB: POWER REACTOR EMBRITTLEMENT NUREG/CR-2000 V09 N5. LICENSEE EVENT REPORT (LER)
DATA BASE.VERSON 1. Program uescription.
COMPILATION.For Month Of May 1990.
NUREG/CR4001: EFFECTS OF MANUFACTURING VARIABLES ON PERFORMANCE OF HIGH-Lf VEL WASTE LOW CARBON STEEL EPO. OPPICE OF NUCLEAR MAftRIAL SAPETY & SAPEOUAROS CONTAINERS.
DIVISON OF SAFEOUARDS & TRANSPORTATON (POST 870413)
NUREG/CR 5258 V02: GEORGIA / ALABAMA REGONAL SEISMO-NUREG/CR4388: HTAS2: A THREE-OlMENSONAL TRANSIENT GRAPHIC NETWORK. Annual R 1986 Jane 1987.
SHIPPING CASK ANALYSIS TOOL NUREG/CR4400: NEUTRON EX SUR PARAMETERS FOR THE NUREG/CR4399: SURVEY OF STATE AND TRIBAL EMERGENCY METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY SEC.
RESPONSE CAPABILITIES FOR RADIOLOGCAL TRANSPORTA-TON STEEL (RRADIATON SERIES.
TON INCIDEF'S.
NUREG/CR 5449; DETERMINATION OF THE NEUTRON AND NUREG/CR44O OADS A MULTIDIMENSIONAL POINT KERNEL GAMMA FLUX DISTRIBUTION IN THE PRESSURE VESSEL AND ANALYSIS MODULE.
CAVITY OF A BOILILNG WATER REACTOR.
DVISCN OF INDUSTRIAL & MEDICAL NUCLEAR SAFETY (POST NUREG/CR4463: EFFECTS OF MINERALOGY ON SORPTON OF B70729)
STRONTIUM AND CESIUM ONTO CALCO HILLS TUFF.
NUREG/CR4568 EVALUATON OF HEALTH EFFECTS IN SE.
NUREG/CR4473: INCLUSON OF UNSTABLE DUCTILE TEARING OUOYAH FUELS CORPORATION WORKERS FROM ACCIDENTAL AND EXTRAPOLATED CRACK ARREST TOUGHNESS DATA IN EXPOSURE TO URANIUM HEXAFLUORIDE.
PWR VESSEL INTEGRlTY ASSESSMENT.
NUREG/CR4574: DETERMINATION OF THE CHEMCAL FORM OF NUREQ/CR4494: CORRELATION OF 1RRADIATION-INDUCED TRITIUM IN SELF LUMINOUS SIGNS.
TRANSITON TEMPERATURE INCREASES FROM C(V) AND K(JC)/
DVISON OF HIGH LEVEL WASTE MANAGEMENT (POST 870413)
K(IC) DATA Frial Report NUREGICR 1667: RISK METHODOLOGY FOR GEOLOGIC DISPOS.
NUREG/CR4510; EVALUATONS OF CORE MELT FREQUENCY EF-AL OF RADCACTIVE WASTE. Scenano Selecten Procedure.
FECTS DUE TO COMPONFNT AGING AND MAINTENANCE.
NUREG/CR4393: A REVIEW OF TECHNIQUES FOR PROPAGATING NUREG/CR4523. DEVELOPMtNT OF AN INFILTRATION EVALUA-DATA AND PARAMETER UNCERTA!NTIES IN HIGH LEVEL RADIO.
TON METHODOLOGY FOR LOW LEVEL WASTE SHALLOW LAND ACT WASTE REPOSITORY PERFORMANCE ASSESSMENT NiE C
- COMPUTER PROGRAMS FOR EDDY CURRENT NUREG/CR4411: ELCITATION & USE OF EXPERT JUDGMENT IN DEFECT STUDIES.
PERFORMANCE ASSESSMENT FOR HIGH LEVEL RADCACTIVE NUREG/CR-5560: AGING OF NUCLEAR PLANT RESISTANCE TEM.
WASTE REPOSITORIES PERATURE DETECTORS.
DIVISION OF LOW. LEVEL WASTE MANAGEMENT & DECOMMISSON-NUREG/CR4579. VALUE/ IMPACT ASSESSMENT OF JET IMPINGE.
ING (POST 870413)
MENT LOADS AND PIPE.TO PIPE IMPACT DAMAGE.Rev6 sed Meth-NUR CR4517: IMPACTS-BRC.YERSION 2.0. Program User's gy,3 g'g "OF $LATORY APPLICATIONbiER!ZATONPO 00 ^ d NUREG/CR4547: APPU'4 TON OF SURFACE COMPLEXATION NUREG/CR4469: BIOLOGCAL CHAR RADb MODELS FOR RADIOiWLIDE ADSORPTON SonstMty Analysse ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANb Of Modelinput Pwwnes'OF GEOCHEMCAL PROCESSES AND UM MILLING EFFLUENTS.
NUREGICR4548: REVIEW NUREG/CR4514: MODELING AND PLRFORMANCE OF THE CODES FOR ASSESSMENT OF RADONUCLIDE MiGRATON PO-TENTIAL AT COMMERCIAL LLW SITES.
NUREG/
54 P
M C S
OF EXTREMITY DOSIMETERS, STUDY 2.
EDO OFFICE OF NUCLEAR RFOULATORY RESEARCH (POST 880406)KEG /R 9 VA PA AN Y E ULATORY OP.
NU EG C SAF TY CH RAMS PON-TlONS FOR RESOLUTION OF GENERC LSSUE C-8.MSIV LEAK-SORED BY OFFICE OF NUCLEAR REGULATORY gyi h SY T RESEARCH.Oetober-December 1989-ESEARCH (POST 880717) CORE NUREG/CR-5181: NUCLEAR PLANT AGING RESEARCH.THE 1E NUREG/CR 4550 V3 RIP 1: ANALYSIS OF DAMAGE POWER SYSTEM-FREQUENCY.SURRY. UNIT 1,1NTERNAL EVENTS.
NUREG/CR4436 VOI: THE DEVELOPMENT AND EVALUATION OF NUREG/CR-4550 V3R1P2: ANALYSIS OF CORE DAMAGE FREOVENCY:SURRY, UNIT 1,1NTERNAL EVENTS APPENDCES.
PROGRAMMATC PERFORMANCE INDICATORS ASSOCIATED NUREG/CR 4550 V5 RIP 1: ANALYSIS OF CORE DAMAGE FRE.
WITH MAINTENANCE AT NUCLEAR POWER PLANTS maw) Report OUENCY:SEQUOYAH, UNIT 1. INTERNAL EVENTS.
NUREG/CR4436 V02. THE DEVELOPMENT AND EVALUATION OF NUREG/CR 4550 V5R1P2: ANALYSIS OF CORE DAMAGE FRE-PROGRAMMATC PERFORMANCE INDICATORS ASSOCIATED OUENCY: SEQUOYAH,UNtT 1. INTERNAL EVENTS APPENDICES.
WITH MAINTENANCE AT NUCLEAR POWER PLANTS. Appendices.
NUREG/CR-4550 V7 A01: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR4475: MODEL FEASIBILITY STUDY OF RADCACTIVE CY:ZON. UNIT 1. lNTERNAL EVENTS.
PATHWAYS FROM ATMOSPHERE TO SURFACE WATER.
NUREG/CR 4639 V01 R1: NUCLEAR OOMPUTERIZED LIBRARY DIVISION OF ENGINEERING (POST 870413)
NUREG/CR 4667 V07: ENVIRONMENTALLY ASSISTED CRACKING FOR ASSESSING REACTOR RELIABILITY (NUCLARA). Summary Descripten.
IN LOHT WATER REACTORS. Sermannual Report,Apro-September NUREG/CR-4691 V01: MELCOR ACCIDENT CONSEQUENCE CODE 1988-SYSTEM (MACCSIVolurne 1: User's Guide 37 1
30 NRC Contract Sponsor Issa NUREG/CR.6111 INTEGRATED RELLASILITY AND RISK ANALYS16 NUREG/CR4613 V01: ACCIDENT MANAGEMENT INFORMATION NUREG/CR4213 ) VERSION 2 0 USER'S GUIDE.
NEEDS. Volume i C - -- _ Developmem And Aem To A SYSTEM (IRRAS VOI: THE COGNITIVE ENVrRONMENT SIMULA*
Presourtred Wawr Reactor (PWR) Wlys A Largo, Dry Contammert TION AS A TOOL FOR MODELING HUMAN PERFORMANCE AND NUMEG/CR4613 YO2: ACCIDENT MANAGEMENT INFORMATION RELIABILfTY.Execuleve NEEDS. Volume 2. Appeneces.
NUMEG/CR 6213 Vor THE NITIVE ENVIRONMENT SIMULA.
NUREG/CR4667: RELAP6 THERMAL HYDRAULC ANALYSl4 OF TION AS A TOOL FOR MODELING HUMAN PERFORMANCE AND THE SNUPPS PRESSURIZED WATER REACTOR. '
NU
^
EG PA ION: A PROGRAM FOR DEFINING THE ED1 SOURCE TERM / CONSEQUENCE ANALYST $ INTERFACE IN THE NUREG-1160 PROSASILISTIC RISK ASSESSMENTS.Unor's Guide-E00. OPPICE OF NUCLSAR flSACTOR flESULAfl0N (POST 4/98/89)
NUREG/CR4262-PRAMIS-PROSASILITY RISK ASSESSMENT DIVISION OF OPERATIONAL EYENTS ASSESSMENT (POST 870411) b FETY IN NUCLEAR NUMEG/CR-4626: CLOSEOUT OF IE SULLETW S443.REFUELNG NUR
/CR-3 A A CAVITY WATER SEAL m
NTS' SASIC 38:
CONSIDERATIONS IN PREDICTING DIVISION OF ENGINEERING TECHNOLOGY (POST 800827)
NUREG/CR ERROR PROSASILITIES IN HUMAN TASK PERFORMANCE.
NUREG/CR4476: MODEL FEASIBILITY STUDY OF RADIOACTIVE NUREG/CR4439; HUMAN FACTORS ISSUES ASSOCIATED WITH PATHWAYS FROM ATMOSPHERE TO SURFACE WATER.
ADVANCED INSTRUMENTATION AND CONTROLS TECHNOL, NUREG/CR4676: SURVEY OF SORIC ACID CORROSION OF OGIES W NUCLEAR PLANTS CARSON STEEL COMPONENTS IN NUCLEAR PLANTS.
NUREG/CR44eo A CAUSE DEFENSE APPROACH TO THE UNDER.
OlVIS40N OF ENGINEERING & SYSTEMS TECHNOLOGY (870411 STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
990826)
NUREG/CR 6480: DATA
SUMMARY
REPORT FOR FIBSION PROD.
NUREG/CR 3960 V06: FUEL PERFORMANCE ANNUAL REPORT UCT RELEASE TEST Vl.3.
FOR 1986.
l 1
I I
Contractor Index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports if further information is needed, safer to the main citation by the NUREG/CR number.
AGRICULTURE. DEPT.OF COLORADO DCHOOL OF MINES GOLDEN, CO NUREGIC45475: MODEL FEAslBILITY STUDY J RADIOACTIVE NUREG/CR-5001: EFFECTS OF MANUFACTURING VARIABLES ON PATHWAYS FROM ATMOSPHERE TO SURFACE WATER.
PERFORMANCE OF HIGH-LEVEL WASTE LOW CARBON STEEL CONTAINERS.
ANALYSIS & MEASUREMENT SERYlCES CORP.
NUREG/CR 5560- AGING OF NUCLEAR PLANT RESISTANCE TEM.
E060 IDAHO,90C- (SUBS. OF E460, INCJ PERATURE DETECTORS.
NUREG/CR4550 V7 A01: ANALYSIS OF CORE DAMAGE FREQUEN.
CY: ZION, UNIT 1, INTERNAL EVENTS.
ARGONNE NATIONAL LA80 RAT 0gy NUREG/CR 4639 V01 R1: NUCLEAR COMPUTERIZED LIBRARY COR NUREG/CR4867 V07: ENVIRONMENTALLY ASSISTED CRACKING IN ASSESSING REACTOR REllABILITY (NUCLARR).Sumnahry Descrip-NU E /
7 vo E V RONM NT Y SS D A NG NU G/CR 5111: INTEGRATED RELIABILffY AND RISK ANALYSIS HT WATER REACTOR $. Semiannual ReportOctober 1966. March NURE
-51 :
ER LA A
RESEARCH.THE 1E POWER SYSTEM.
ARI2ONA $ TATE UNIV., TEMPE. A2 NUREG/CR 5513 V01: ACCIDENT MANAGEMENT INFORMATION NUREQ/CR-5253: PARTITION. A PROGRAM FOR DEFINING THE NEEDS. Volume 1 Methodolo0y Development And Apphcetion To A SOURCE TERM / CONSEQUENCE ANALYSIS INTERFACE IN THE
^
NUREG-1150 PROBABILISTC RISK ASSESSMENTS. User's Guide.
. NU C
V DE MAN T
TION NEEDS Volume 2 Appendices.
BATTELLE MEMORIAL INSTITUTE. PACIFIC NORTHWEST NWEGM5557: N5 NENHMM WN W THE LA80RATORY SNUPPS PRESSURIZED WATER REACTOR.
NUREG/C43950 V06: FUEL PERFORMANCE ANNUAL REPORT FOR 1966-g; SERVICES, INC, NUREG/C45397: VALUE lMPACT ANALYSIS OF REGULATORY OP' NUREG/CR4550 V3R1P1: ANALYSIS OF CORE DAMAGE TIONS FOR RESOLUTION OF GENERC ISSUE C 6.MSIV LEAKAGE FREQUENCY:SURRY. UNIT 1,1NTERNAL EVENTS.
NUREG/CR4550 V3 RIP 2: ANALYSIS OF CORE DAMAGE AND LCS FAILURE.
FREQUENCY:SURRY. UNIT 1,1NTERNAL EVENTS APPENDCES.
NUREG/OR 5523: DEVELOPMENT OF AN INFILTRATION EVALUA-TION METHODOLOGY FOR LOW LEVEL WASTE SHALLOW LAND NUREG/C44550 V5 RIP 1: ANALYSIS OF CORE DAMAGE FREQUEN-CY:SEQUOYAH,UNfT 1,1NTERNAL EVENTS.
BURLAL StTES.
NUREG/CR-4550 V5R1P2: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR-5540: PERFORMANCE TESTING OF EXTREMITY CY: SEQUOYAH. UNIT 1. INTERNAL EVENTS APPENDICES.
DOSIMETERS. STUDY 2.
NUREG/CR 5547: APPLICATION OF SURFACE COMPLEXATION GEORGE MASON UNIV FAIRFAX.VA MODELS FOR RADIONUCUDE ADSORPTION Sensitety Analysis Of NUREG/C45436: BASIC CONSIDERATIONS IN PREDICTING ERROR Modelinput Parameters.
PROBABILITIES IN HUMAN TASK PERFORMANCE.
NUREG/CR-5548: REVIEW OF GEOCHEMICAL PROCESSES AND CODES FOR ASSESSMENT OF RADIONUCLIDE MIGRATION PO.
GEOR04A INSTITUTE OF TECHNOLOGY. ATLANTA. 0A TENTIAL AT COMMERCIAL LLW SITES.
NUREG/C45256 V02: GEORGIA / ALABAMA REGIONAL SEISMO.
NUREG/CR-5566: EVALUATION OF HEALTH EFFECTS IN SECOOYAH GRAPHIC NETWORK. Annual Report, July 1966. June 1967.
FUELS CORPORATION WORKERS FROM ACCIDENTAL EXPOSURE g
TO URANIUM HEXAFLUORIDE HAWAll, UNIV. OF, HILO, HI '
NUREG/CR-5579: VALUE/ IMPACT ASSESSMENT OF JET IMPlNGE-NUREG/CR 5411: ELICITATION & USE OF EXPERT JUDGMENT IN MENT LOADS AND PIPE TO PIPE IMPACT DAMAGE. Revised Meth.
PERFORMANCE ASSESSMENT FOR HIGH LEVEL RADIOACTIVE ods And Cntona.
WASTE REPOSITORIES.
DROOKHAVEN NATIONAL LADORATORY INDIANA UNIV BLOOMINGTON.IN NUREG/CR 2331 V00 N4: SAFETY RESEARCH PROGRAMS SPON-NUREG/CRd399; SURVEY OF STATE AND TRIBAL EMERGENCY RE.
SORED BY OFFICE OF NUCLEAR REGULATORY SPONSE CAPABILITIES FOR RADIOLOGICAL TRANSPORTATION RESE ARCH. October December 1969.
INCIDENTS' NUREG/CR 5573: BORON FLUSHING DURING A BWR ANTICIPATED TRANSIENT WITHOUT SCRAM.
INHALATION T0XICOLOGY RESEARCH INSTITUTE NUREG/C45574: DETERMINATION OF THE CHEMICAL FORM OF
.NUREG/CR-5469: BIOLOGCAL CHARACTERIZATICN OF RADIATION TRITIUM IN SELF-LUMINOUS SIGNS-EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM Mill-NUREG/CR-5576: SURVEY OF BORC ACID CORROSION OF CARBON ING EFFLUENTS' STEEL COMPONENTS IN NUCLEAR PLANTS.
JgF ASSOCIATES,INC, NUREG/C45460: A CAUSE DEFENSE APPROACH TO THE UNDER.
N REGI f53 V03. CANADtAN SEISMIC AGREEMENT. Annual Report: 1987 1966.
MANUFACTURING DCIENCES CORP.
NUREG/CR-4753 V04: CANADIAN SEISMIC AGREEMENT. Annual NUREG/CR 5001: EFFECTS OF MANUFACTURING VARIABLES ON Roport 1966 1969.
PERFORMANCE OF HIGH-LEVEL WASTE LOW CARBON STEEL CANADIAN COMMERCIAL CORP.
NUREG/CR4753 V03: CANADIAN SEISMC AGREEMENT. Annual MATERIALS ENetNEERING ASSOCIATES,INC.
Report 1987 1986.
NUREG/CR 5494: CORRELATION OF 1RRADIATION INDUCED TRAN.
NUREG/CR4753 V04: CANADIAN SEISMC AGREEMENT. Annual S! TION TEMPERATURE INCREASES FROM C(V) AND K(JC)/K(IC)
Report 1966-1969.
DATA. Final Report 39 i
i
40 Contractor index MINNESOT A. UNIV. 0F, MINNEAPOLIS MN NUREG/CR-5213 V02: THE COGNITIVE ENVIRONMENT SIMUL /. TON NUREG/CR 5437: ORGANIZATION AND SAFETY IN NUCLEAR POWER AS A TOOL FOR MODEUNG HUMAN PERFORMANCE AND PLANTS.
REUABluTY.Mem Report MITRE CORP.
SANDIA NATIONAL LABORATORES NUREG/CP 0100. PUBUC WORKSHOP ON NUCLEAR POWER PLANT NUREG/CR 1667: AISK METHODOLOGY FOR GEOLOGIC DISPOSAL LICENSE RENEWAL OF RADIOACTfVE WASTE. Scenano Selection Procedure.
NUREG/CR4550 V3R1P1: ANALYSIS OF CORE DAMAGE FRE URR NTT 1.1 ERN EVENTS.
~'~
U CR 5460; A CAUSE DEFENSE APPROACH TO THE UNDER.
NU STANDING AND ANALYSTS OF COMMON CAUSE FAILURES.
FREQUENCY;SURRY. UNIT 1 INTERNAL EVENTS APPENDICES.
c OAE RIDGE Natl 0NAL LABORATORY NUNEG/CR4560 V5R1P1: ANALYSIS OF CORE DAMAGE FREQUEN.
NUREG/CR4000 V09 N3; UCENSEE EVENT REPORT (LER)
CY:SEQUOYAH. UNIT 1 lNTERNAL EVENTS.
COMPILATION.Fcr Month ol March 1990.
NUREG/CR4550 V5RtP2: ANALYSIS OF CORE DAMAGE FREQUEN-NUREG/CR 2000 V00 N4: UCENSEE EVENT REPORT (LER)
CY: SEQUOYAH. UNIT 1.lNTERNAL EVENTS APPENDICES.
COMPILATION Fcr Month Ot April 1990 NUREG/CR-4891 V01: MELCOR ACCIDENT CONSEQUENCE CODE NUREG/CR-2000 V09 N5: UCENSEE EVENT REPORT (LER)
SYSTEM (MACCS). Volume 1: Users Gukte.
COMPILATON.For Month Of May 1990 NUREG/CR 6253: PARTITON; A PROGRAM FOR DEFINING THE NUREG/CR4616; PR.EDB: POWER REACTOR EMBRITTLEMENT SOURCE TERM / CONSEOVENCE ANALYSIS INTERFACE IN THE DATA BASE.VER$10N 1. Program NUREG 1160 PROBABluSTIC RISK ASSESSMENTS Users Guide.
'=
NUREG/CR 5306. HTAS2: A THREE-SiONAL TRANSIENT SHIP.
NUREG/CR 6262: PRAMIS; PROBABlUTY RISK ASSESSMENT MODEL PING CASK ANALYSIS TOOL INTEGRATION SYSTEM.Usere Guide.
NUREG/CR-6409: NEUTP.ON EXPOSURE PARAMETERS FOR THE NUREG/CR 5393: A REVIEW OF TECHNIOUES FOR PROPAGATING E
METALLURGICAL TEST SPECIMENS IN THE SIXTH HEAVY SEC*
DATA AND PARAMETER UNCERTAINTIES IN HIGH-LEVEL RADIO-fg *^8 NUR /C 6 A
ISSUES ASSOctATED WITH AD-VANCED INSTRUMENTATION AND CONTROLS TECHNOLOGIES IN NUREG/CR 6411: EllCITATION & USE OF EXPERT JUDGMENT IN NUR /CR 54 TERMINATION OF THE NEUTRON AND GAMMA PERFORMANCE ASSESSMENT FOR HIGH LEVEL RADIOACTIVE FL STR IN THE PRESSURE VESSEL AND CAVITY OF NUR G/C 54 AC SE DEFENSE APPROACH TO THE UNDER-NUREG/CR 6463. EFFECT OF' MNERALOGY ON SORPTION OF STANDING AND ANALYSIS OF COMMON CAUSE FAILURES.
STRONTIUM AND CESIUM ONTO CAUCO HILLS TUFF.
NUREG/CR 5517: IMPACTS-BRC. VERSION 2.0. Program Users Manuai.
NURE OADS. A MULTOIMENSIONAL POINT KERNEL SCIENCE APPLICATIONS INTERNATIONAL CORP. (PORedERLY NUREG/CR-6473; INCLUSION OF UNSTABLE DUCTILE TEARING AND SCIENCE APPLICATIONS, EXTRAPOLATED CRACK-ARREST TOUGHNESS DATA IN PWR NUREG/CR 5253; PARTITON; A PROGRAM FOR DEFINING THE VESSEL INTEGRITY ASSESSMENT.
SOURCE TERM! CONSEQUENCE ANALYSIS INTERFACE IN THE NUREG/CR 6490 DATA
SUMMARY
REPORT FOR FISSION PRODUCT NUREG 1150 PROBABILISTIC RISK ASSESSMENTS Users Guide.
RELEASE TEST Vl-3 NUREG/CR 6436 V01: THE DEVELOPMENT AND EVALUATION OF NUREG/CR 5614: MODELING AND PERFORMANCE OF THE MHTGR PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH REACTOR CAVITY COOUNG SYSTEM.
MAINTENANCE AT NUCLEAR POWER PLANTS Mem Report NUREG/CR 5553. COMPUTER PROGRAMS FOR EDDY CURRENT NUREG/CR 5436 V02: THE DEVELOPMENT AND EVALUATION OF DEFECT STUDIES.
PROGRAMMATIC PERFORMANCE INDICATORS ASSOCIATED WITH COLUMSUS, OH MAINTENANCE AT NUCLEAR POWER PLANTS.Appendicos.
OHIO STATE UNIV'V01: THE COGNITIVE ENVIRONMENT SIMULATON NUREG/CR 5213 NUREG/CR-6510- EVALUATONS OF CORE MELT FREQUENCY EF.
S AGM AND WWN.
AS A TOOL FOR MODELING HUMAN PERFORMANCE AND 213 S b H COG \\tTIVE ENVIRONMENT SIMULATON SOUTHERN CALIFORNIA, UN'V. OF, LOS ANGELES, CA NU E /
N EG/CRm EUTATON & USE OF NERT WDGM W AS A TOOL FOR MODEUNG HUMAN PERFORMANCE AND REUABluTY. Main RW PERFORMANCE ASSESSMENT FOR HIGH-LEVEL RADIOACTIVE
- WASTE REPOSITORIES.
PARAMETER, INC.
NUREG/CR 4525: CLOSEOUT OF IE BULLETIN 6443.REFUEUNG WESTINGHOUSE SCIENCE & TECHNOLOGY CENTER CAVITY WATER SEAL.
NUREG/CR 5213 V01: THE COGNITIVE ENVIRONMENT SIMULATON AS A TOOL FOR MODEUNG HUMAN PERFORMANCE AND PITTSSURGH, UNIV. OF PITTSSUROH. PA REUABlWTY.Executwo Summary.
NUREG/CR 5213 V01: THE COGNITIVE ENVIRONMENT SIMULATION NUREG/CR-6213 V02: THE COGNITIVE ENVIRONMENT SIMULATION AS A TOOL FOR MODEUNG HUMAN PERFORMANCE AND AS A TOOL FOR MODEUNG HUMAN PERFORMANCE AND RELIABlWTY.Executwo Summary.
REUABluTY.Mem Report.
international Organization index This index lists, ir alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed below each country and per-dorming 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.
FRANCE NUREG/ LAM ASSESSMENT OF RELAP5/ MOD 2 CYCLE 36.04 COMMISSION OF ATOMIC INERGY USING LOFT LARGE BREAK EXPERIMENT L2 5.
NUREG/LA-0023: ASSESS WENT OF TRAC PF1/ MOO 1 VERSON 14.3 USING SEPARATE EFFICTS CRITICAL FLOW AND BLOWDOWN SWftIERLAND NURE 0023 ASSESSME OF T AC.PF1/
1 YERSION 14.3 A e 6/ m ASSESWEM, OECom USING SEPARATE EFFECTS CRITICAL FLOW AND BLOWDOWN SMALL BREAK EXPFRIMENT LP-8B.03.
EXPERIMENTS. VOLUME 2: FIGURES.
gegetgp gegeGDON OREA NUCLEAR SAFETY CENTER g L l
NUREG/LAN ASSESSMENT OF RELAP5/ MOO 2 CODE USING UNITED KINGDOM ATOMIC ENERGY AUTHORITY LOSS OF OFFSITE POWER TRANS6ENT DATA OF KNU #1 NUREG/lA 0011: TRAC.PF1/ MOO 1 POST. TEST CALCULATIONS OF NU NW1: ICAP ASSESSMENT OF RELAP5/ MOD 2, CYCLE NUR G A C.P T.' TEST CALCULATIONS OF 36.05 AGA!NST LOFT SMALL BREAK EXPERIMENT L3 7.
THE OECD LOFT EXPERIMENT LP 88 3.
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Licensed Facility index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the repoM number. If further information is needed, refer to the main citation by the NUREG number S 424 AMn W Vogee Nuder Plut, Ure 1, George WM41410 S327 Geoeveh % deer hent M 1, Temosene WMGIW550 V58t1P2 S 445 Nek Senem finew 8eston, M 1, RMGW97 S24 S 200 Power m Una 1. VWgres DecW & MJREG/W550 V3R1P1 50445 Peak f.iecte Boston, Ure 1 MM41380 2 200 norry Power Baston, Urm 1, VWgrue Doctnc & MJREG/W550 V3R1P2 Tomas Ventes Enoct Power Co.
504 4 Comwee Peak Sleem Doctc Sm Ure t, WMGW97 S24 S 113 Urer of Anaone Assee@ Rosetor WMG-13eo S 327 Wesear PM pre 1, Tommesse WMGtW550 V5 RIP 1 vetor Aumorty Edson Co l
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NU G 04 2,1tiLL AND EUBis1LL Regulatory and Technical Reports (Abstract index Journal) 3 DAlt REPOR1 PUBLISH { D Compilation.for i
Second Quarter 1990 August 1990 April - June
" 'N R cRANT NvMe R
- b. AUTHORiss
- 6. TYPE OF REPORT Reference 7, PE RIOD COvi k t D isar.ww powes April - June 1990 BP FOR NG g A,NI2 AlION - N AME AND ADOR($5 fir AmC, paseas> O= epa, preste e, Aeosea, vs hurwe, servenee,y Coniaw.assa, ear medene espress. ## reat,scoes, p. eve Division of Freedom of Information and Publications Services Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 9.
O ORG AN12 AllON
- N AME AND ADDRE ES tre wec ever *1eme es esow~, et centeweee. one,apr Nec penea, ocase w Asynon, va mese, Aepvenu, coma.se.aon, Same as 8. above.
- 10. SUPPLtME NT ARY NO1($
II, ABST R ACl f200.e,sh w wwJ This journal includes all formal reports in the NUREG series prepared by the NRC staf f' and contractors; proceedings of conferences and workshops; as well as international agreement reports. The entries in this compilation are indexed for access by title and abstract, secondary report number, personal author, subject, NRC organization for staff and international agreements, contractor, international organization, and licensed facility, in t, wvR os.ui sc R w i cR5 u.
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o.....mo, u. u m i compilation Unlimited abstract index ame Unclassified a-an u Unclassified Ib. NUMBLR OF PAGE $
16 PRICL
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i THIS DOCUMENT WA8 PRINTED USING RECYCLED PAPER.
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.m Secondary Report f
Number index w
Personal Author index h
Subject index
[
k le NRC Originating Organlaation 1;;,1 Index (Staff Reports) pg h4 NRC Originating Orgenlaation Index (International Agreements) g n
i NRC Contractor 5
Sponsorinden z
Contractor index international Organisation l
-Index q
l Uoensed Facility l
.Index
-