Text

Regulatory and Technical Reports.Annual Compilation for 1985
	ML20154G747
Person / Time
Issue date:	02/28/1986
From:	; NRC OFFICE OF ADMINISTRATION (ADM)
To:	;
References
	NUREG-0304, NUREG-0304-V10-N04, NUREG-304, NUREG-304-V10-N4, NUDOCS 8603100092
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i s Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013 7082 A year's subscription consists of 4 issues for this publication. Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 l

NUREG-0304 Vol.10, No. 4 Regulatory and Technical Reports (Abstract Index Journal? Annual Compilation for 1985 Date Published: February 1986 Pclicy and Publications Management Branch

              - Divi:lon of Technical Information and Document Control Office of Administration U.S. Nuclear Regulatory Commission Wcshington, D.C. 20555
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CONTENTS Preface . . . . . .. . .. . . . . ... ... ... v index Tab Main Citation and Abstracts . . . .. .. 1 Staff Reports. . .. Conference Proceedings . . . . . . . . .. . .. . Contractor Reports . . . . . . . . . . . . . . . . .. .. .. ... . Contractor Report Number index . . .. .. . .. . . .... ... ... .2 Personal Author index . .. . .... ... . . . .. .. . . . 3 Subject index .... .. .............. . .. . . . ..... .. . . . 4 NRC Originating Organization index (Staff Reports) . . ... . . . . 5 NRC Contract Sponsor index (Contractor Reports) . . . .. .6 Contractor index . . . . . . . . . ... . .. . . . . .7 Licensed Facility index . . . .. .. . . . . . . . . . .. . .. ..8 iii

I PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. Please send them to: Division of Technical Information and Document Control Policy and Publications Management Branch Publishing and Translations Section Woodmont 501 U.S. Nuclear Regulatory Commission Washington, D.C. 20565

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

Contractor Report Number index Personal Author Index Subject Index

                - NRC Originating Organization Index (Staff Reports)

NRC Contract Sponsor index (Contractor Reports) 4 Contractor Index Licensed Facility index *

;,     A detailed explanation of the entries precedes each index.

4

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

Staff Report 5 NUREG-0508: MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

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

Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the microfiche address (for internal NRC use). Conference Report i NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND RELIABILITY ENGINEERING IN NUCLEAR REGULATION JANERP, J.S. Argonne National-Laboratory. May 1981.141 pp. 8105280299. ANL-81-3. 08632:070. , Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled the proceedings, (5) date report was published, (6) number of pages in the report, (7) the NRC Docu-ment Centrol System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use). i Contractor Report NUREG/CR 1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R. Sandia Laboratories. May 1981.160 pp. 8107010449. SAND 80-0929. 08912:242. Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizitional 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 infomal use). v

The following abbreviations are usm1 to identify the document status of a report: I ADD - addendum APP - appendix DRFT - draft ERR - errata N - ' number R - revision S - supplement V - volume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the following address: Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washingtnn, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275-2171. Non U.S. customers must make payment in advance either by international Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents. NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported. In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings. All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Technical Information and Document Control.' I l l l i vi

Main Citations and Abstracts The report listings in this compilation are ar- is an NRC contractor-prepared report, The ranged by report number, where NUREG- bibliographic information (see Preface for XXXX is an NRC staff-originated report, details) is followed by a brief abstract of this NUREG/CP-XXXX is an NRC-sponsored report. conference report, and NUREG/CR-XXXX NUREG-0017 R01: CALCULATION OF RELEASES OF RADIOAC- See NUREG.0020,V08,N12 abstract. TlVE MATERIALS'IN GASEOUS AND LIQUID EFFLUENTS FROM PRESSURIZED WATER REACTORS (PWR-GALE NUREG-0020 V09 N05: LICENSED OPERATING REACTORS CODE). CHANDRASEKARAN; LEE,J.Y.; WILLIS.C.A. Division of STATUS

SUMMARY

REPORT. Data As Of Aprd 30,1985 (Gray Systems Integration (811005-851124). Aprd 1985. 208pp. Book l)

Divison of Budget & Analysis. June 1985. 441pp.

8505280361.30603:161, 8507080197. 31396:267. This report revises the original issuances of NUREG-0017, See NUREG-0020,V08,N12 abstract.

               " Calculation of Releases of Radioactive Matenals in Gaseous and Uguid Effluents From Pressunzed Water Reactors (PWR-                            NUREG-0020 V09 N06: LICENSED OPERATING REACTORS GALE Code)" (Apnl 1976), to encorporate more recent operating                          STATUS

SUMMARY

REPORT. Data As Of May 31,1985 (Gray data now available as well as the resufts of a number of in-plant Book 1) ROSS,P.A.; BEEBE.M R. Dwison of Budget & Analysis. measurement programs at operating pressunzed water reactors. July 1985. 437pp. 8508190629. 32261.001. The PWR GALE Code is a computenzed mathematical model See NUREG-0020,V08,N12 abstract. for calculating the releases of radcactrve material in gascous and liquid effluents (i.e., the gaseous and liquid source terms)- The U.S. Nuclear Regulatory Commission uses the PWR-GALE NUREG-0020 V09 N07: LICENSED OPERATING REACTOHS Code to determine conformance with the requirements of Ap* STATUS

SUMMARY

REPORT. Data As Of June 30,1985 (Gray pendix i to 10 CFR Part 50. Book l) ROSS.P.A.; BEEBE M R. Dwison of Budget & Analysis. August 1985. 426pp. 8509130022. 32620.001. NUREG-0020 V08 N12: LICENSED OPERATING REACTORS See NUREG.0020,V08,N12 abstract. STATUS

SUMMARY

REPORT. Data As Of November 30,1984 (Gray Book !)

Dwison of Budget & Analyss. February NUREG-0020 V09 N08: LICENSED OPERATING REACTORS 1985. 427pp. 8502210264. 29047:289. STATUS

SUMMARY

REPORT. Data As Of July 31,1985 (Gray The OPERATING UNITS STATUS REPORT - LICENSED OP- Book f)

Division of Budget & Analysis. September 1985.

ERATING REACTORS provides data on the operation of nucle- 440pp. 8510070175. 32902 037, ar units as timely and accurately as possble. This informaton is See NUREG-0020,V08,N12 abstract. collected by the Office of Resource Management from the Headquarters staff of NRC's Of+ ice of Inspecten and Enforce- NUREG-0020 V09 N09: LICENSED OPERATING REACTORS ment, from NRC's riegional Offices, and from utshties. The three STATUS

SUMMARY

REPORT. Data As Of August sectons of the report are; monthly highhghts and statistics for 31,1985 (Gray Book 1) POSS P.A.; BEEBE.M R. Divison of commercial operating units, and errata from previously reported Budget & Analyss. October 1985. 471pp. 8511210285, data; a compdaten of detaded information on each unit, provid- 33556:216. ed by NRC's Regonal Offices, IE Headquarters and the utikties; See NUREG-0020,V08,N12 abstract. and an appendix for mascellaneous information such as spent fuel storage capabihty, reactor-years of expenence and non-NUREG-0020 V09 N10: LICENSED OPERATING REACTORS power reactors in the U.S It is hoped the report is helpful to all STATUS

SUMMARY

REPORT Data As Of September agencies and individuals interested in maintainsng an awareness 30,1985.(Gray Book 1)

Division of Budget & Anafysis. Novem-of the U.S. energy stuation as a whole. ber 1985. 455pp. 8512190244. 34015 345 NUREG 0020 V09 N01: LICENSED OPERATING REACTORS See NUREG-0020,V08,N12 abstract.

STATUS

SUMMARY

REPORT. Data As Of December 31,1984.(Gray Book I)

Divison of Budget & Analysis. February NUREG-0020 V09 N11: LICENSED OPERATING REACTORS 1985. 396pp. 8503220010. 29487.265. STATUS

SUMMARY

REPORT. Data As Of October See NUREG-0020,V08,N12 abstract. 31,19851 Gray Book I) ROSS.P.A.; BEEBE.M R Divison of NUREG-0020 V09 NO2: LICENSED OPERATING REACTORS Analyss. December 1985. 200pp. 8601070520. STATUS

SUMMARY

REPORT. Data As Of January f31 31,1985.(Gray Book I)

Division of Budget & Analysis. March See NUREG-0020,V08.N12 abstract.

1985. 421pp. 8504090008 29740:311. See NUREG-0020.V08,Nf 2 abstract. NUREG 0040 V08 N04: LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT, Quarterly NUREG 0020 V09 NO3: LICENSED OPERATING REACTORS Report, October-December 1984. (Whito Book)

Divison of OA, STATUS

SUMMARY

REPORT. Data As Of Febuary Safeguards & Insp Programs (830103-850212). January 1985. 28,1985.(Gray Book l)

Division of Budget & Analyss. Apnl 238pp. 8502210401. 20056.014, 1985. 403pp. 8505100053, 30263 082-See NUREG4020,V08,N12 abstract- This penodical covers tho results of iropectons pe formed by the NRC's VeMor Program Branch that have been distnbuted NUREG-0020 V09 N04: LICENSED OPERATING REACTORS to the inspected organizations donng the pered from October STATUS

SUMMARY

REPORT Data As Of March 31,1985(Gray 1984 through D9cember 1984. Also ricluded in this issue are Book }l ROSS P.A.; BEEBE M.R. Drvison of Budget & Analys s. the results of ce tain inspectsons pedormed poor to October May 1985. 440pp. 8506170349. 30959.072, 1984 that were not included in previous assues of NUREG-0040. 1 4

         -n        -  -e- . ,- . . , en,,+ ,-v, -.---mm-y,--vw.n--.-       -w--w- ..-mn-,---,w--               w, , , , - - , - . - - , - , , , , , , , - , - , - - - ----.-,c-.- --.,,,.n-~ . - - - - - - ,

J 2 Main Citations and Abstracts NUREG-0040 V09 N01: LICENSEE CONTRACTOR AND involved four control rods failing to insert dunng testing and the ! VENDOR INSPECTION STATUS REPORT. Quarterfy other involved degraded upper head injecten system accumula-Report. January-March 1985.(White Book)

Dvision of OA, tor isolation valves. There was one abnormal occurrence at a

Vendor & Techncal Training Center Programs (Post 850212). fuel cyclo facihty; the event invoNed buildup of uranium in a M2y 1985. 219pp. 8506030069. 30707: 001. ventdaten system. There was one abnormal occurrence report-This penodical covers the results of inspect ons performed by ed by an Agreement State; the event invoNed an overexposure the NRC's Vendor Program Branch that have been distnbuted of a radiographer trainee. The report also contains information to the inspected organizations dunng the pered from January updating some prevously reported abnormal occurrences.

1985 through March 1985. Also included in this issue are the results of certain inspections performed pnor to January 1985 NUREG-0090 V04 N01: REPORT TO CONGRESS ON ABNOR- + tfut were not included in previous issues of NUREG-0040. MAL OCCURRENCES.Janurary-March 1985.

AEOD, Director's NUREG-0040 V09 N02: LICENSEE . CONTRACTOR AND Offee. August 1985. 46pp. 8509060189. 32505.273.

VENDOR INSPECTION - STATUS REPORT. Quarterty Secten 208 of the Energy Reorganizaten Act of 1974 idenu. Report. April-June 1985. (White Book)

Division of OA, Vendor - fies an abnormal occurrence as an unscheduled incident or

    & Techncal Training Center Programs (Post 850212). August            event whch the Nuclear Regu*atory Comrnission determines to 1985. 245pp. 8509060260. 32503:078.                                 be significant from the standpoint of pubhc health and safety This penodcal covers the results of inspections performed by     and requires a quarterly report of such events to be made to the NRC's Vendor Program Branch that have been distnbuted            Congress. This report covers the pered January 1 to March 31, to the inspected organizations dunng the penod from Apnl 1985         1985. Dunng the report pered, tisere was one abnormal occur-l through June 1985. Also included m this issue are the results of     rence at the nuclear power plants hcensed to operate; the event

- certain inspections performed prior to Apnl 1985 that were not invoNed a premature entcality dunng reactor startup. There included in prevous 'ssues of NUREG-0040. were three abnormal occurrences at the other NRC hcensees.

NUREG-0040 V09 NO3: LICENSEE CONTRACTOR AND Two events invo ved diagnoste medical misadministrations and VENDOR INSPECTION STATUS REPORT. Quarterty the other event involved unlawful possession of radioachve ma-Report. July 1985-September 1985.(Whste Book)

Division of tenal. There were four abnormal occurrences reported by an OA, Vendor & Techncal Training Center Programs (Post Agreement State (Texas). Three events invoNed radiaten o<er.

850212). December 1985.137pp. 8601070495. 34210:127, exposures; the other event involved a well logging source whch Thss penodical covers the results of inspections performed by was apparently stolen, but later was recovered. The report also the NRC's Vendor Program Branch that have been distnbuted contains information updating some prevously reported abnor-to the inspected orgaruzatens dunng the period from July 1985 mal occurrences. l through September 1985. Also included in this issue are the re-suits of certain inspectons performed pnor to July 1985 that NUREG-0090 V09 N02: REPORT TO CONGRESS ON ABNOR-were not included in previous issues of NUREG-0040. MAL OCCURRENCES Apni-June 1985. AEOD, Director's Office. November 1985. 54pp. 8512190002. 33966 012. NUREG-0000 V07 NO3: REPORT TO CONGRESS ON ABNOR. Secten 208 of the Energy Reorganizaten Act of 1974 idents. MAL OCCURRENCES. July-September 1984.

AEOD, Director's Office. Apnl 1985. 70pp. 8505160182. 30456:325. fies an abnormal occurrence as an unscheduled incident or Secten 208 of the Energy Reorganization Act of 1974 idenb- event whch the Nuclear Regulatory Commissen determines to fies an abnormal occurrence as an unscheduled incident or be signifcant from the standpoint of pubic health and safety event whch the Nuclear Regulatory Commissen determines to and requires a quarterly report of such events to be made to be sigruficant from the standpoint of pubhc health and safety Congress. This report covers the penod Aprd 1 to June 30, l

and requires a quarterfy report of such events to be made to 1985. Dunng the report penod. there were three abnormal oc-Congress. This report covers the penod July 1 to September currences at the nuclear power plants hcensed to operate. l These events involved, respectively, (1) inoperab!e safety inlec-30, 1984. Dunng the report penod, there were four abnormal i occurrences at the nuclear power plants hcensed to operate. tion pumps, (2) significant deficiencies in reactor operator train-These involved degraded isolaten valves in emergency core ing and matenal fake statements, and (3) loss of main and, cochng systems, degraded shutdown systems, a loss of offsite auxihary feedwater systerm There were four abnormal occur-tnd onsite AC electncal power, and a refuehng cavity water seal rences at other NRC heensees. Three events involved diagnos. friture, respectively. There was one abnormal occurrence at a tc or thorapeutic medical misadmnustratens; the other involved fuel cycle facihty; the event involved degraded matenal access a breakdown in management controls. There was one abnnrmal stea barners. There were four abnormal occurrences at the occurrence reported by an Agreement State; the event involved other NRC hcensees. One involved contaminated radopharma- overexposures of a radiographer and an assistant radiographer. ceutcals used in several diagnostic administratens. Two in- The report also contains information updating some prevously volved therapeutic medical rnisadministratens. The other in- reported abnormal occurrences. volved signifcant intemal exposure to lodine-125 to a hosptal i employee. There was one abnormal occurrence reported by an NUREG 0304 V09 N04: REGULATORY AND TECHFICAL i Agreement State; the ever.t involved contam6nated ridophar- REPORTS. Annual Compdaten For 1984.

Division of Techncal

' maceuticals used in several diagnoste administrabons. The Informaton & Document Control. January 1985. 501pp. report also contair.s infortnation updating some prevously re- 8502210096. 29049 001, ported abnormal occurrences. This loumal includes all formai reports in the NUREG senes prepared by the N9C ra'1 and connactuis, as well as poceed-1 NUREG-0000 V07 N04: REPORT TO CONGRESS ON ABNOR. ings of conferences and workshops. The entnes in the compila-MAL OCCURRENCES. October-December 1984.

AEOD, Direc.

tor's Office. May 1985. 40pp. 8506180402. 30986.013. tion are indexod for access by title and abstract, contracter l report number, personal auths, sub iec*, NRC organization, con-

Secton 208 of the Errgy Reorganization Act of 1974 identi-fies an abnormal ocgurrence as an unscheouled incident or tractor, and hcensed facshty.

) event whch the Nuclear Regulatory Commission determines to be signifcant from the standpoint of pubhc health and safety NUREG-0304 V10 N01: REGULATORY AND TECHNICAL I REPORTS Compdaten For First Quarter 1985. January-March.

l and requires a quarterfy report of such events to be made to Dev.sson of Tectwca' Intvmation & Document Controt. Apnl

Cong ess. This report covers the pered October 1 to December 3 31,1984. Dunng the report penod, there were two abnormal oc. 1985. f 29pp. 8505240207. 30564:195.

currences at the nuclear power plants hcensed to operate. One See NUREG-0304,V09.N04 abstract. l 1 i

Main Citations and Abstracts 3

    . NUREG-0304 V10 NO2: REGULATORY AND TECHNICAL                                                                  September 30,1983 interpreting the NRC's Rules of Practice in
         ' REPORTS. Compilation For Second Quarter 1985,Apnl-June.

10 CFR Part 2. This edston replaces earber editions and supple-Division of Techncal informaton & Document Control. July ments and includes appropnate changes reflect'ng the amend-

,           1985. 88pp. 8508150008. 32198:129.

See NUREG-0304,V09,N04 abstract. ment to the Rules of Practice effectwo September 30,1983. i NUREG-0304 V10 NO3: REGULATORY AND TECHNICAL NUREG-0420 SO9: SAFETY' EVALUATION REPORT RELATED REPORTS. Compilation For Third Quarter 1985, July - Septem- TO THE OPERATION OF SHOREHAM NUCLEAR POWER j ber.

Dvison of Technical Information & Document Control. STATION. UNIT 1. Docket No 50-322. (Long Island Lighting October 1985. 72pp. 8511260053. 33639:193. Company)
Office of Nuclear Reactor Regulaten. Drector See NUREG4304,V09,N04 abstract. (post 851125). December 1985. 171pp. 8601070479.

' NUREG-0325 R07: U.S. NUCLEAR REGULATORY COMMISSION FUNCTIONAL ORGANIZATION CHARTS.

Office of Resource Supplement 9 (SSER 9) to the Safety Evaluaten Report on Management. Director. January 1985. 56pp. 8501180528. Long Island Lighting Company,a application for a hcense to op.

28471:117, erate the Shoreham Nuclear Power Station, Unit 1, located in Furtenal organizaten charts for the NRC Commission Of- Suffolk County, New York, has been prepared by the Office of

,         fices Dvisens, and Branches are presented.                                                                Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commesson. This supplement addresses several items that NUREG-0383 V01 ROS: DIRECTORY OF CERTIFICATES OF                                                              have been reviewed by the staff since the prevous supplement COMPLIANCE               FOR           RADIOACTIVE         MATERIALS                                     was issued' PACKAGESSummary Report Of NRC Approved Packages.

Divison of Fuel Cycle & Material Safety. October 1985.477pp.

6511110217. 33408.043. NUREG-0430 V05 N01: LICENSED FUEL FACILITY STATUS This directory contains a Summary Report of NRC Approved REPORT. inventory Dfferonce Data January 1984 - June Packages (Volume 1), Certficates of Compliance (Volume 2), 1984.(Gray Book 11)

Drector's Offee, Office of Inspecten and and a Summary Report of NRC Approved Quakty Assurance Enforcement. Apnl 1985.18pp. 8504290008. 30056:348.

': Programs for Radcactwe Matenal Packages (Volume 3). The NRC is committed to the penodic pubhcaten of hcensed fa-purpose of this directory is to make available a convenient ciWs imenton difference data, following agency review of the source of information on packagings whch have been approved information and completon of any related NRC investgatons. by the U.S. Nuclear Regulatory Commission. To assist in identi- Informaton an this report includes inventory difference data for fying packaging, an index by Model Number and corresponding actue fuel fabncaten facilibes possessmg more than one effec-i Certificate of Comphance number is included at the back of twe kilogram of high enrched urar sum, low ennched uran,um, each volume of the directory. The Summary Report includes a plutonsum, or Uransum-233. hstng of all users of each package design pnor to the publica-l tion date of the directory Shipments of radcactwe matenal uts- NUREG-0430 V05 N02: LICENSED FUEL FACILITY STATUS hzmg these packagings must be in accordance with the provi- REPORT. inventory Dfference Data July 1984 . December sons of 49 CFR 173.471 and 10 CFR Part 71, as applicable. In 1984.(Gray Book 1)

Drector's Offee, Offee of inspecten and a

satisfying the requirements of Secten 71.12, et is the responsi. Enforcement. October 1985.17pp. 8511220456. 33601:030. bility of the hcensees to insure them that they have a copy of See NUREG-0430,V05,N01 abstract. the current approval and conduct their transportaten activities 1 in accordance with an NRC approved quahty assurance pro- NUREG-0525 R10: SAFEGUARDS

SUMMARY

EVENT LIST gram. Copies of the current approval may be obtained from the (SSEL). REVISION 10.

Lscensing Pohey & Programs Branch U S. Nuclear Regulatory Commisson Puble Document Room (Pre 850707). May 1985. 59pp. 8506140072. 30907;130.

!                                                                                                                     The Safeguards Summary Event List (SSEL) provnies bnet fi!es (see Docket No. listed on each certificate) at 1717 H i        Street, Washington, DC 20555. Note that the general heense of                                            summanes of several hundred safeguards-related events mvo!v.

10 CFR 71.12 does not authonze the receipt, possessen, use ing nuclear matenal of facihtes regulated by the U.S. Nuclear or transfer of byproduct source, or special nuclear matenal, Regulatory Commisseon (NRC). Events are desenbed under the such authonzaten must be obtained pursuant to 10 CFR Parts categones of bomb-related, entruson, misseng/ allegedly stolen,

30 to 36, 40, 50, or 70. transportation, tampenng/vandahsm, arson, firearms-related, ra-NUREG-0383 V02 ROS: DIRECTORY OF CERTIFICATES OF dologeal sabotage and miscellaneous. The informaton con-COMPLIANCE FOR RADIOACTIVE MATERIALS tained in the event desenptons is denved pnmanly from official PACKAGES.Certficates of Compliance.

Dwison of Fuel Cycle NRC reportng channels.

M enal Safety. October 1985. 633pp. 8511110338. NUREG-0540 V06 N11: TITLE LIST OF DOCUMENTS MADE See NUREG-0383,V01,R08 abstract PUBLICLY AVAILABLE. November 1 30,1984.

Cuisco of

' Techncal Information & Document Control. January 1995. NUREG-0383 V03 R05: DIRECTORY OF CERTIFICATES OF $69pp. 8502060305. 28754:136. COMPLIANCE FOR RADIOACTIVE MATERIALS This document is a monthly pubication containing desenp-PACKAGES. Summary Report Of NRC Approved Quality Assur- tions of mformaton recewed and generated by the U S. NRC.

once Programs For Radcactue Matenal Packages.

Duison of Fuel CnJa & Material Safety. Octot>er 1985. 128pp. This informaton includes (1) docketed matenal assocated with cuihan nuclear power plants and other uses of radcactue ma-8511070490. 33383:243.

, See NUREG-0383,V01,908 abstract- tenals, and (2) nondocketed tratenal recewed and generated by ' NRC pertinent to its role as a regulatory agency. The following rdUREG-0386 D03: UNITED STATES NUCLEAR REGULATORY indexes are included: Parsonal Author index, Corporate Svurca COMMISSION STAFF PRACTICE AND PROCEDURE Inder, Report Number index, and Cross Reference to Pnnetpal DIGEST. JULY 1972 SEPTEMBER 1983.

Office of the Execu. Occumer'ts inder.

tue Legal Director.

Aspen Systems, Inc. July 1985. 800pp.

I 8506210006. 32303:334. NUREG-0540 V06 N12: TITLE LIST OF DOCUMENTS MADE This editon of the NRC Staff Practice and Procedure Digest PUBLICLY AVAILABLE. December 1 31,1984.

Dwison of contains a digest of a number of Commessen, Atomic Safety - Technscal informaton & Document Control. February 1985.

i and Licensing Appeal Board, and Atome Safety and Lcensing 614PP. 8503200131, 29469 001. { Board decisions issued dunng the period from Jufy 1,1972 to See NUREG4540,V06.N11 abstract. k I i i 4

      . . .       --r         --m,,,-.._                    mm       _ _._..__--...-- ,,------,_ ,- -.-- - ,_,-- -,

4 4 Main Citations and Abstracts NUREG-0540 V07 N01: TITLE LIST OF DOCUMENTS MADE Provides an overview of the status of the progress and pians a PUBUCLY AVAILABLEJanuary 1 31,1985.

Dmsson of Technt- for resoluton of the genene tasks addressing " Unresolved cal information & Document Control. March 1985. 665pp. Safety issues" as reported to Congress.

NUREG-0606 V07 NO2: UNRESOLVED SAFETY ISSUES f AdG 545 V06*N11 abstract.

SUMMARY

. Data As Of May 17,1385. (Aqua Book)

Descn of NUREG 0540 V07 N02: TITLE LIST OF DOCUMENTS MADE Safety Technology (800428-851124). June 1985. 61pp.

PUBLICLY AVAILABLE. February 1 28,1985.

Oms on of Tech- 8507080200. 31390:169

] nical Information & Document Control. Apnl 1985. 699pp. See NUREG-0606.V07 N01 abstract. l RN 054 ,VO 'N11 aostract' NUREG-0606 V07 NO3: UNRESOLVED SAFETY ISSUES e

SUMMARY

. Data As Of August 16.1985. (Aqua Book)

Dms on j

l NUREG 0540 V07 NO3: TITLE UST OF DOCUMENTS MADE of Safety Technology (800428 851124) August 1985. 55pp. l PUBLICLY AVAILABLE. March.131.1985.

Desen of Techni- 8509760507. 11631.363.

cal informaton & Document Control Apnl 1985. 430pp See NUREG 0606.V07.N01 abstract i NUREG-0675 S28: SAFETY EVALUATION REPORT RELATED N RdG 54d'V06.N11 abstract TO THE OPERATION OF DIABLO CANYCN NUCLEAR j. 1 NUREG-0540 V07 N04: TITLE LIST OF DOCUMENTS MADE POWER PLANT. UNITS 1 AND 2 Docket Nus. 50-275 And 50-PUBUCLY AVAluBLE. Apnl 1 30. 1985.* Desion of Technical 323 (Pacif4c Gas And Electnc Company)

Des'on of Licensing l (800428 851124) Apnl 1985. 635pp. 8505100069 30286.156.

informaton & Document Control. June 1985. 484pp. 8507080219. 31395.143. Supplement No. 28 to the Sa'ety Eva'uaten Report for the i' See NUALG-0540 V06.N11 abstract. apphcat.on by the Pacif4c Gas and Electnc Company (PG&E) to ' NUREG-0540 V07 N05: TITLE UST OF DOCUVENTS MADE " ** PUBUCLY AVAILABLE.MaY 1985

Dmson of Technical O*fice of Nuciear Reactor Regulation of the U.S Nuclear Regu-Informaton & Document 1'31' Co ntrolJuly 1985. 489pp- latory Commission. This supplement reports on the s'atus of tre 8507250203. 31790.011. sta'f s invest gaton, inspecten and eva!uaton of those aMega-

. See NUREG-0540.V06 N11 abstract. tions or concerns that hare been identified to tre NRC as of March 1,1985 NUREG 0540 V07 N06: TITLE UST OF DOCUMENTS MADE PUBLICLY AVAILABLE. June 1-30. 1985.

Dmsion of Technical NUREG-0675 S29: SAFETY EVALUATION REPORT RELATED Information & Document Control July 1985- 577pp. TO THE CPERATION OF DIABLO CANYON NUCLEAR !

8508150439 32218 236- POWER PLANT. UNITS 1 AND 2 Docket Nos. 50 275 And 50-See NUREG-0540.V06.N11 abstract. 323 (Pactic Gas And Electnc Company)

Dms on of Ucensing (800428 851124), March 1985. 64pp 8503280011. 29548 075 i NUREG-0540 V07 N07: TITLE UST OF DOCUMENTS MADE Supplement No. 29 to the Safety Eva'uaten Report for Pacific j PUBUCLY AVAILABLE. July 1 31, 1985.
Dvsion of Technicat '

1 informaton & Document Control August 1985 630pp. Gas and Electnc Company's apphcaton for bcenses to operate ' Diablo Canyon Nuclear Power P! ant. Units 1 and 2 IDocket Nos. 8509190f 03. 32670122. 50 275 and 50-323) has been orepared by the Office of Nacle-See NUREG 0540.V06.N11 abstract. ar Aeactor Regutaten of the U.S Nuc: ear Regulatory Commis-l NUREG-0540 V07 N08: TITLE LIST OF DOCUVENTS MADE sion. Thrs supplement presents the sta'f eva uation of the h-PUBUCLY AVA!LABLE August 1 31, 1985.

Dmston of Techn6- cersce's Internal Rewew Program for Diablo Unit 2 appbcabaty ca! Informaton & Decument Control September 1985. 656pp. and resoluten of concerns that had been rased du ring tre "

8510070182. 32903:117. Otablo Unit I design verificaten by tne independent Design Ver-See NUREG 0540.V06.N11 abstract ification Program, the licensee's internal Technical Program and the NRC sta*f-NUIEG-0540 V07 N09: TITLE LIST OF DOCUMENTS MADE PUBUCLY AVAILABLE. September 1 30,1985

Dms:on of NUREG 0675 S30: SAFETY EVALUATION REPORT RELATED

 ,                     Technical informaton & Document Control. October 1995.                     TO THE OPERATION OF D:ABLO CANYON NUCLEAR 406pp. 8511220014. 33604 248                                               POWER PLANT. UNITS 1 AND 2 Docket Nos 50 275 And 50 See NUREG-0540.V06.N11 abstract.                                       323 (Pacific Gas And Electnc Company?

Dmsten of Licensing (800428 851124h Apnl 1985 137pp. 8504220336 2%43 202 NUREG-0540 V07 N10: TITLE. LIST OF DOCUVENTS MADE Supplement No. 30 to the Sa'ety Evaluaton Report for the PUBUCLY AVAILABLE. October 1 31.1985.* Dvsion of Technt- appkcaton by the Pacific Gas and Electnc Company (PG&E) to

!                      cai informaton & Document Control December 1985. 452pp                     operate the Diablo Canyon Nuclear Pceer Plant. Units 1 and 2

8512270359 34089.109. (Docket Nos 50 275 and 50-323) has teen prepared by the i See NUREG-0540,V06 N11 abstract. Off.ce of Nuclear Reactor Regulaton of the U S Nuclear Regu-j NUREG-0544 R02: A HANDBOOK OF ACRONYMS AND INITIA- latory Commession. This supplement reports on the staff's tech-nicar rev,ew ano evaivation of the des gn and analys4 of p,p.ng i USMS.

Dmsson of Technical Informaten & Document Control January 1985.131pp. 8502150692. 28960:175 systems and pipe suppcrts for Unit 2.

This Handbook records in alphabetcal order abbrevatons NUREG 0675 S11 SAFETY EVALUATION REPORT RELATED (scronyms, initialisms, and other condensed forms) that have TO THE OPERATION OF DIAi$LO CANVON N'JCLEAh ! been used in the nuclear industry, both foreign and domestic. POWER PLANT, UNITS 1 AND 2 Docket Nos. 50-275 And 50- ! The present vefume is an attempt by the ed. tonal staff of the 323(Pacifc Gas And Eiectnc CompanP

Dvson of Ocensng Dmscn c' Tech,wcar inforr at on and Document Con:rol to com- (800428 851124). Apnl 19851/1pp 850524L223 %564.022 pile rn one updated putAcation the atbrewatons used in NRC Supplemem 31 to the Safety Evaluaton Report for tne apph-staff and consultant reports. This issue, although not all inclu- caton by Pacific Gas and Electnc Company for hcenses to op-srve, is to be treated as a compvaton c f available information. DiaNo Canyon Nuclear Power Plant. Uni's 1 and 2 [

j (Docliet Nos. 50 275/323) has been p'epared by the Office of , NUEEG 0606 V07 N01: UNRESOLVED SAFETY ISSUES Nuclear Reactor Regulaton of the U.S. Nscles.r Regulatory

SUMMARY

. Data As Of February 15 1985(Aqua Booy

De Commasion. This supplement addresses a number of matters son of Safety Technology (80C428 851124). Varch 196155pp related to the issuance of an operatng hcense for Diatlo I 8504040429.29627:321. r h

i

1-

Main Citations and Abstracts 5 I

Canyon Unit 2, in partcular those issues identified by the NRC 1,11 rems (cSv).

The most recent year for whch most of the l staff in earlier supplements, commitments made by the appli- termination data are available for analysis.

cant, and certain heense conditons included in Facility Operat-ing License No. OPR-81 for Diablo Canyon Unit 2. NUREG-0725 R05: PUBLIC INFORMATION CIRCULAR FOR j NUREG-0675 $32: SAFETY EVALUATION REPORT RELATED SH!PMENTS OF IRRADIATED REACTOR FUEL

Dmson of Safeguards. June 1985. 55pp. 8507080177. 31394 330.

TO THE OPERATION OF DIABLO CANYON NUCLEAR This circular has been prepared in response to numerous re-i POWER PLANT, UNITS 1 AND 2. Docket Nos.50-275 And 50-quests for information regarding routes used for the shipment of i 323.(Pacife Gas and Electnc Company)

Desion of Licensng irradiated reactor (spent) fuel subject to regulation by the Nucle-

              ' (800428-851124). August 1985. 50pp. 8508210398. 32337.324.               at Regulatory Commission (NRC), and to meet the requirements Supplement 32 to the Safety Evaluaton Report for the appli-of Public Law 96-295. The NRC staff must approve such routes t                  cation by Pacife Gas and Electnc Company for heenses to op-erate Diablo Canyon Nuclear Power Plant, Units 1 and 2                  poor to their first use in accordance with the regulatory provF (Docket Nos. 50-275/323) has been prepared by the Office of             sions of Secten 73.37 of 10 CFR Part 73. The informatinn in-cluded reflects NRC staff knowledge as of June 1.1985. Spent Nuctear Reactor Regulaton of the U.S. Nuclear Regulatory               fuel shipment routes, pnmanly for road transportation, but also 4                .Commisson. This supplement provides the staff evaluation of
'                                                                                        including one ral route, are indicated on reproduct>ons of DOT those matters that require an appropnate resolution pnor to full-                                                                            ,

power operation of Unit 2 and upda'es previous supplements to road maps. Also included are the amounts of matenal shipped , the Safety Evaluaton Report, dunng the approximate three year penod that safeguards regu-lations for spent fuel shipments have been effective. In addition, NUREG-0713 V05: OCCUPATIONAL RADIATION EXPOSURE AT the Commission has chosen to provide information in this docu-COMMERCIAL NUCLEAR POWER REACTORS - 1983 ment regarding the NRC's safety and safeguards regulations for ANNUAL REPORT. BROOKS,B.G. Drvison of Radiation Pro- spent fuel shipments as well as safeguards incidents regarding grams & Earth Sciences (post 840429). March 1985.128pp. spent fuel shipments (of whch none have been reported to . 8504050287. 29674.001. date). This additonal information is furnished by the Commes-This report summanzes the occupational radiaton exposure son in order to convey to the public a more complete picture of ' information that has been reported to the U.S N.R.C. by com- NRC regulatory practices concerrung the shipment of spent fuel mercial nuclear power reactors dunng the years 1969 through than could be obtained by the publication of th0 shipment 1983. The bulk of the data presented in the report was obtained routes and quantities alone. from annual radiatiori exposure reports submitted in accordance i with the requirements of 10 CFR 20.407 and leense technical NUREG-0748 V04 N12: OPERATING REACTORS LICENSING f specifications. Data on workers terminating their employment at ACTIONS

SUMMARY

. Data As Of December 31,1984.(Orange 4 nuclear power facilities was obtained from reports submitted Book)

Management Support Branch. January 1985. 320pp.

pursuant to 10 CFR 20.408. The annual reports submitted by 8502210263. 29056.355. 4 the 76 nuclear power plants that had completed at least one full The Operating Reactors Ucensing Actons Summary is de-

 .              year of operation as of December 31,1983, indicated that the            signed to provide the Management of the Nuclear Regulatory l                number of personnel monitored dunng 1983 was 136,700 pe,.               Commassen (NRC) with an overview of Icensing actons dealing i                sons and the annual collective dose incurred by these individ-          with the operating power and nonpower reactors.                      3
 ;              uals was 56,500 man-rems (man-cSv). The average annual dose for each worker that received a measurable dose was               NUREG-0748 V05 N01: OPERATING REACTORS LICENSING 0.66 rems (cSv), and the average collective dose per reactor            ACTIONS

SUMMARY

. Data As Of January 31.1985 (Orange was 753 man-rems (man-cSv). The termination reports revealed BW

Management Support Branch. March 1985. 332pp.

{ 8504030070. 29598.326 that some 56,500 individuals completed their employment with See NUREG-0748,V04,N12 abstract. , one or more reactor facilities dunng 1982.* Approximately 4,500 of these workers could be considered transients and they re-cerved an average dose of 1.11 rems (cSv). 'The most recent NUREG-0748 V05 N02: OPERATING REACTORS LICENSING year for whch most of the terminaten data are available for ACTIONS

SUMMARY

. Data As Of February 28,1985 (Orange analysis. Book) Management Support Branch. Apna 1985. 335pp. 8505070582. 30207:295. i NUREG-0714 V04-05: OCCUPATIONAL RADIATION See NUREG-0748,V04,N12 abstract. l EXPOSURE. Fifteenth And Sixteenth Annual Reports,1982 And

,                1983 BROOKS.B.G.; MCDONALD.S.; RICHARDSON,E. Divisen                  NUREG-0744 V05 NO3: OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data As Of March 31,1985. (Orange 985 3p8 2 12. 33603 3 Book) Management Support Branch. May 1985. 342pp. This report summanzes the occupational radiation exposure 8500030192. 30688 001. informaton that has been reported to the NRC by certain cate. See NUREG-0748,V04,N12 abstract. gones of NRC licensees dunng the years 1973 through 1983. j The bulk of the data presented in the report was obtained from NUREG4744 VOS N04: OPERAT!NG REACTORS LICENSING annual radiaton exposure repcrts submitted in accordance with ACTIONS

SUMMARY

. Data As Of Apnl 30,1985(Orange Book) the requirements of 10 CFR 20.407. Data on workers terminat- Management Support Branch. June 1985. 344pp. 8507020440. 31310:071, ing their employment at certain NRC Icensed facilities were ob* See NUREG-0748,VO4,N12 abstract. tained from reports submitted pursuant to 10 CFR 20.408. The

 ,              annual reports submitted by nearty 500 licensees indicated that        NUREG-0744 V05 N05: OPERATING REACTORS LICENSING                      -

approrimatery 154 000 individuals were morntored in 1982 and ACTIONS

SUMMARY

. Data As Of May 31,1985. (Orange Book) { about 173,000 individuals were monitored in 1983. They in-

Management Support Branch. July 1985.357pp.8507250167.

cuned average annual doses of 0.37 rem (cSv) and 0.35 rem 01792 028. (cSy respectrvely. Terrrunaten radiaton exposure reports re- See NUREG-0748,V04.N12 abstract. l! quired to be submitted pursuant to 10 CFR 20.408 were ana- i e lyzed te revea: that about 59,000 individuals completed their NUREG-0744 V05 N06: OPERATING REAC7 ORS LICENSING

employment with one or more of the 500 covered licensees ACTIONS

SUMMARY

. Data As Of June 30,1985. (Orange Book)

,               dunng 1982*. Some 56,000 of these endnnduals terminated from                 Management Support Branch. August 1985. 358pp.

I power reactor facilities, wt'en about 4,500 of them were consid- 8508210039. 32302:336.

;               ered to be transient workers who received an average dose of                See NUREG-0748,V04,N12 abstract.

i i

     .---n.,   - . - _ - -., , - - -, -                -.r _ n _ _   ,,_,-n--,,.                                                          ,.,.,-,,,vn_-_~-

l 6 Main Citations and Abstracts NUREG-0748 V05 N07: OPERATING REACTORS LICENSING NUREG-0750 V21 NO2: NUCLEAR REGULATORY COMMISSION ACTIONS

SUMMARY

. Data As Of July 31.1985.(Orange Book)

ISSUANCES FOR FEBRUARY 1985. Pages 275-469.
Division Management Support Branch. September 1985. 362pp. of Technical Information & Document Control Apnl 1985.

8510030438. 32846.287. 203pp. 8504290238. 30056 001. See NUREG 0748,V04.N12 abstract. See NUREG-0750,V20.N04 abstract. I NUREG-0748 V05 N08: OPERATING REACTORS LICENSING NUREG-0750 V21 N03: NUCLEAR REGULATORY COMMISSION ACTIONS

SUMMARY

. Data As Of August 31,1985. (Orange ISSUANCES FOR MARCH 1985. Pages 471559.

Division of Book).* Management Support Branch. October 1985. 365pp. Technical Informaton & Document Control May 1985. 78pp.

8511070084. 33377:161. 8505280104.30606 279. See NUREG-0748,V04 N12 abstract- See NUREG-0750.V20.N04 abstract. NUREG-0748 V05 N09: OPERATING REACTORS LICENSING NUREG-0754 V21 N04: NUCLEAR REGULATORY COMMlSSION ACTIMS

SUMMARY

. Data As Of September 30,1985 (Orange ISSUANCES FOR APRIL 1985. Pages 5611.041.

Oivision of Technical information & Document Control. June 1985.490pp.

8511 20465 3360 046 85 See NUREG-0748,V04.N12 abstract. Se N REG 075 v20.N04 abstract. NUREG-0748 V05 NIO: OPERATING REACTORS UCENSING . NUREG 0750 V21 N05: NUCLEAR REGULATORY COMM;SSION ACTIONS

SUMMARY

. Data As Of October 31.1985 (Orange Book)

Management Support Branch. December 1985.225pp. ISSUANCES FOR MAY 1985. Pages 1.043-1,567.
Division of Technical in8crmation & Document Control. July 1985. 524pp.

8512270422. 34073.081. See NUREG-0748.V04.N12 abstract. 8508260300. 32370.304 NUREG-0750 V20101: INDEXES TO NUCLEAR REGULATORY COMM:SSION ISSUANCES FOR JULY SEPTEMBER 1984.

NUREG 0750 V21 N06: NUCLEAR REGULATORY COMMiSSICN Division of Technical information & Document Control January ISSUANCES FOR JUNE 1995. Pages 1,569-1,786.
Division of 1985. 78;p. 8503270298. 29540:261. Technical Infortratcn & Document Control August 1985.

Digests and indexes for issuances of the Commiss.on, t*e 2150p. 8509230739 32702.248. Atornic Safety and Licensing Appeal Panel, the Atomic Safety See NUREG-0750 V20.N04 abstract. and Ucensing Board Panel, the Administra$ve La* Judge. the NUREG-0750 V22101: INDEXES TO NUCLEAR REGULATORY Director's Decisions. and the Deniais of Petitions for Rulemak. COVMISSION ISSUANCES. July-September 1985.

Commis-ing are presented. sioners December 1985. 69pp. 8601090315. 34239 016.

NUREG-0750 V20102: INDEXES TO NUCLEAR REGULATORY See NUREG-0750,V20.101 abstract. COMM:SSION ISSUANCES FOR JULY-DECEVBER 1984.

De v'sion of Technical Informaton & Document Control. March '

NUREG 0750 V22 N01: NUCLEAR REGULATORY COMMISSION 1985.105pp. 8504040416 29618193. ISSUANCES FOR JULY 1985. Pages 1176.

Division of Tech-See NUREG-0750,V20.101 abstract. nical information & Document Control. August 1985. 184pp.

8509300527. 32787:308. NUREG-0750 V20 N04: NUCLEAR REGULATORY COMMISSION See NUREG-0750,V20 N04 abstract. (SSUANCES FOR OCTOBER 1984. Pages 1,0551.435.

Divi-s;on of Technical Informaton & Document Control. Janua y NUREG-0750 V22 N02: NUCLEAR REGULATORY COMVISSION 1985. 388pp. 8502060353. 28748 001. Legal issuances of tne ISSUANCES FOR AUGUST 1985. Pages 177 457.
Division of Commission, the Atomic Safety and Ucensing Appea! Panel, the Technicat Informat on & Document Control August 1985 Atomic Sa'ety and Ucensing Board Panel, tne Administratme 286pp. 8511010269. 33305 048.

Law Judge and NRC Program Offices. See NUREG-0750,V20.N04 abstract. NUREG-0750 V20 N05: NUCLEAR REGULATORY COMMISSION NUREG 0750 V22 NO3: NUCLEAR REGULATORY COVViSSION ISSUANCES FOR NOVEVBER 1984. Pages 1.4371.572.

D'- ISSUANCES FOR SEPTEMBER 1985. Pages 459-649
Devi-vision of Technical information & Document Control. January sion of Technical information & Document Controt November 1085.143pp. 8502190320. 29012 226. 1985. 201p0. 8512100721. 33830 086 See NUREG-0750.V20.N04 abstract. See NUREG-0750 V20.N04 abstract.

NUREG-0750 V20 N06: NUCLEAR REGULATORY CCMMiSSION NUREG-0750 V22 N04: N'UCLEAR REGULATORY COMMISSION ISSUANCES FOR DECEMBER 1984. Pages 1,573-1.7C6

Dive ISSUANCES FOR OCTOBER 1985 Pages 651769
Divison sion of Techrucal Informaton & Document Control. February of Techrucal informaten & Document Centrol December 1985 1985.134pp. 8503110591, 29307.239 125pp. 8601070494. 34190.094.

See NUREG-0750.V20.N04 abstract. See NUREG 0750.V20.N04 abstract. NUREG-0750 V21101: INDEXES TO NUCLEAR REGULATORY* NUREG-0787 S10: SAFETY EVALUATICN AEPORT RELATED COMMISSION ISSUANCES FOR JANUARY MARCH 1985 70 THE CPCHATION OF WATERFORD STEAM ELECTRIC Division of Technical information & Document Contro'. Jure 1985. 59pp. 8507080203. 31380 272. S7ATION. UNIT 3 Docket No. 50-382. (Louis >ana Power And See NUREG 0750,V20.101 abstract. Ugnt Company)

Dnnsson of Licensing (800428 851124). March 198 3 85032702R 29M2a NUREG 0750 V21 l02: INDEXES TO NUCLEAR REGULATORY Suppament m to the Sa'eh Evaluaben Report 6 the appb COMVISS:ON ISSUANCES FOR JANUARY-JUNE 1985
D.vi- caton Ned by Lousiana Power & M Company for a license sicn of Technical informatun & Document Control. September to opera *e the Waterford Steam Electne Station, Unit 3 (Cocket 1985.108pp. 8510030126. 32855.009. No. 50-382), located in St. Charfes Parish, Louisiana, has been See NUREG 0750,V20.101 abstract. ,,,,,,ed by the Off,ce of Nuclear Reactor Regu'abon of the U S. Nuclear Regu'atory Comm.asics. The purpose of this suo-NUCEG 0750 V21 N01: NUCLEAR REGULATORY CCMMISSION piement is to update the Safety Eveiaation Report ey providing ISSUANCES FCR JANUARY 1985. ges 1273,
Crvis.on of Technical Information & Document Control. March 1985.281pp the staff's evaluat.on of informaton submitted by the hcensee s4nce the Safety Evaluation Report and its nine suppfements 8504030299 29624.081.

See NUREG 0750,V20 N04 abstract. aere issued.

Main Citations and Abstracts 7 NUREG-0797 S07: SAFETY EVALUATION REPORT RELATED approximately 400 technical Concerns and a' legations in the TO THE OPERATION OF COMANCHE PEAK STEAM ELEC- mechanical and piping area regarding constructen practices at TRIC S1 ATION. UNITS 1 AND 2 Docket Nos. 50-445 And 50- the Comanche Peak facihty. This report does not address the 446.(Texas UtAties Generat ng Company et aa

Division cf Li- Walsh/Doyle allegatons regard.ng defic.encies in the pipe sup-censing (800428-851124). January 1985.100pp. 8502150150. port design process. Issues raised by the Walsh/Doyle ailega-28958.071. tions as well as issues ra: sed dunng recent Atomic Safety and Supplement No. 7 to the Safety Eva'uaten Report for the Licensing Scard heanngs wdl be dealt with in future supple-Texas Utihties Generating Company appbcation for a hcense to ments to the Safety Evaluation Report as needed coerate the Comanche Peak Steam Electnc Station located in Somervell County. Texas has been jointly prepared by the NUREG 0797 S11: SAFETY EVALUATION REPORT RELATED Office of Nuclear Reactor Regulation and the Technical Rewew TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-Team of the U S. Nuclear Regulatory Commission. This Supple. TRIC STATION.UN TS 1 AND 2 Docket Nos. 50-445 And 50-ment provides the results of the staff's evaluation and resolution 446 (Texas Utahties Generating Company, et af)
Division of U-of approximately 80 technical concems and aflegations in the censing (800428-851124). Vay 1985 349pp 8506190054.

areas of Efectncal/ Instrumentation and Test Program regarding 31018.014 construction practices at the Comanche Peak facihty. Supplement No.11 to the Sa'ety Evaluaten Report for the NUREG-0797 S08: SAFETY EVALUATION REPORT RELATED as s cm meany apphcaton for a hcense to op e he anc a eam ecmc Statc% Unds 1 and 2 TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-TRIC STATION UN!TS 1 AND 2. Docket Nos. 50-445 And 50-

                                                                                                                       ""           ' #8   '" "                       0""#Y' 446 (Texas UtAties Generating Company, et al)

Dvsion vf Le ma s. s 'ee Ii pre are y me ce f Nuctear Read censing (800428-851124). February 1985 196pp.8503110247. 9* * "" * ** * "'##' **** * *
293 W M of the U.S. Nuclear Regulatory Commissacn (NRC) and is in two Supplement No. 8 to the Safety Evaluaton Report for the *" " 0 * * '*'

Texas Utihtees Generating Company apphcaton for a heense to # 0"# '*

operate the Comanche Peak Steam Electnc Station located in a a a gs caw m waW assurance and Somervell County. Texas has been lointry prepared by the D "" # #"##"

a anc a facW Pad 2 %ndu 9 conta,ns Office of Nuclear Reactor Regulaten and the Technical Revaw Team of the U.S. Nuclear Regulatory Commissen. This Supple- man su m a and conclusen of me OA/OC aspects of the ment provides the results of the staff s evaluaton and resolu* ion

as e in ak of approximately 80 technical concerns and allegatons relating to civil / structural and miscellaneous issues regarding construc*

h* a .[ i be dealt with in future supp: aments to the SER as needed. ton and plant readiness testing practices at the Comanche Peam facility. NUREG-0797 S12: SAFETY EVALUATION REPORT RELATED NUREG-0797 SO9: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-TO THE OPERATION OF CCMANCHE PEAK STEAM ELEC- TRIC STATION. UNITS 1 AND 2 Docket Nos 50-445 And 50-TRIC STATION. UNITS 1 AND 2 Docket Nos 50 445 And 50 446. (Texas Utihtses Generating Company)

Comanche Peak 446 (Texas Utihties Generating Company, et af)
Diviseen of U. Project (Technical Review Team) Octot>er 1985 283pp.

censing (800428-851124) March 1985. 170pp. 8504090015. 8511040056 33338.200. 29730 116. Supplement No.12 to the Sa'ety Evaluaton Report related to Suppfement 9 to the Safety Evaluation Report for the Texas the operation of the Comanche Peak Steam Electnc Station. Utahties Electnc Company's apphcation for a heense to operate Units 1 and 2 (NUREG-0797) has been prepared by the Office Comanche Peak Steam Electnc Station. Units 1 and 2 located of Nuclear Reactor Regulaton of the U.S Nuclear Regulatory in Somervell County Texas, has been prepared jointly by the Commisson. rhe facihty is located in Somervell County, Texas. Office of Reactor Regulaton and the Comanche Peak Techne. approximately 40 miles southwest of Fort Worth. Texas This cal Review Team of the U S Nuclear Regulatory Commission supplement reports the status of certain issues that had not This supprement addresses Texas Utaties analyses in support been resoNed at the time of pubhcat on of the Safety Evafuation of its request to amend the Comanche Peak Final Sa'ety Analy- Report and Supplements 1, 2, 3. 4 and 6 to that report This sis Report to eliminate the commitment that coatings inside the supplement a!so I:sts the new issuct that have been identif)ed reactor Containment Budding be quali f ed for Units 1 and 2. In since Supplement 6 was issued and .ncludes the evaluatens for addition, this supplement provides the results of the staff's eval, hcensing items resoNed in this inter m pered Supplement 5 has uaton and resolution of 82 technical concerns and attegations not been issued. Supplements 7. 8 9.10 and 11 were hmeted to in the coating area for Unit 1. Because of the tavorable resolu- the staff evaluations of allegateora investigated by the NRC tion of the items discussed in this report. the staff concludes for Technscal Review Team, and items identified therein are not in-the issues considered hereas, that there is reasonable assur, ctuded in this suppinment ance that the facihty can be operated without endangenng the health and safety of the pubhc. NUREG-0798 S05: SAFETY EVAL *JATION REPORT RELATED TO THE OPERATION OF ENRICJ FERMI ATOM.O POWER NUREG-0797 S10: SAFETY EVALUATION REPORT RELATED PLANT, UNIT NO. 2 Docket No. 50-341. (Detroit Ed. son Compa-TO THE OPERATION OF COMANCHE PEAK STEAM ELEC- ny)

Duson of bcensing (800428 851124) March 1985 TRIC STATION, UNITS 1 AND 2 Docket Nos. 50-445 And 50- 200pp 8504040427. 29628 243 446 (Texas Utihties Electnc Company, et al)
Divisen of Lb Supplement No 5 to the Safety Evaluaten Peport tSER) re-censing (800428 851124). Apnf 1985. 326pp 8506030061. late j to the operation of the Fermi-2 facihtv. provides the N 1C 30706 008. sta'f's evaluation of add; tonal informaten submitted by the ap-Supplement No.10 to the Safety Evaluaton Report for the pl' cant regarong the outstaad,r g review issues ident.fied in Eup.

Texas Utihties Electnc Company applicaten for a heense to op- piement No 4 to the SER cated Eeptember '984 Tms suppie. erate Comanche Peax Steam Electnc StaSon, Un.ts 1 and 2 ment contains the staff's conclusion that there are no outstand. (Docket Nos. 50-445 and 50-446). located in Somervell County, ing issues which must be resolved pnor to issuance of a low-Texas, has been jointly prepared by the Office of Nuclear Reac- power operating hcense (i e, less than five percent of full rated for Regulation and the Comanche Peak Technical Review Team power) for the FermF2 facky. Supplement No 5 to the SER of the U S. Nuclear Regulatory Cornmission. This supplement also summar'zes the conditions which are placed in the Ferme2 provides the results of the staff's evaluation and resolution of operating hcense The FermF2 facil't, is located on Lake Ene in _ _ - . _ _ _ _ _ _ . . _ _ _ _ . _ __, , _- _ _ _ _ ,-- - .m_. - _ _ - - - _

8 Main Citations and Abstracts Monroe County, almost 8 mdes east-northeast of Monroe, NUREG-0824 S01: INTEGRATED PLANT SAFETY ASSESSMENT Michigan. REPORT, SYSTEMATIC EVALUATION PROGRAM MILLSTONE NUCLEAR POWER STATION. UNIT 1. Docket No. 50-245 4 NUREG-0798 S06: SAFETY EVALUATION REPORT RELATED (Northeast Nuclear Energy Company)

Office of Nuclear Reac- )

TO THE OPERATION OF FERYl-2 Docket No 50-341 (Detroit tor Regulation, Director (post 851125). November 1985. 58pp Edison Company)

Diwsion of Licens.ng (800428 851124). July 8512120114. 33874 202.

1985. 56pp. 8508210034 32302 259. The Nuclear Regulatory Commission (NRC) has pubkshed its Supplerrent No. 6 to the Safety Evaluation Report (SER) re- Supplement No 1 to the Final Integrated Plant Safety Assess-lated to operation of the Fermr2 facihty addresses items perts- ment Report (IPSAR) (NUREG-0824) under the scope of the nent to the issuance of the fufi power license for Fermi-2. The Systematic Evaluation Program (SEP), for Northeast Nuclear Fermi-2 facihty is located on Lake Erie in Monroe County. Energy Company's Millstone 1 Plant located in Waterford. New almost 8 miles east northeast of Monroe, Michigan. London County, Connecticut. The SEP was initiated by the NRC to review the design of older operating nuclear power plants to NUREG 0800 06.2.2 R4: STANDARD REVIEW PLAN FOR THE reconfirm and document their safety. This report documents the REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR review completed under the SEP for those issues that reauired POWER PLANTS LWR Ed, tion Revision 4 To Section 6 2 2, refined engineenng evaluations a+ter the Final IPSAR for Mill-

 "Conta:nment Heat Removal System."

Office of Nuclear Re* stone 1 Plant was issued The review has provided for (1) an actor Regulation Director (pre 851125). October 1985. 11pp assessment of the signif.cance of differences between current 8512110128. 33845-301. technical pos+tions on selected safety issues and those that ex-Revision 4 to SRP Section 6 2 2 incorporates guidehnes de- ,sted when the Millstone 1 Plant was hcensed. (2) a basis for ve'oped from the technical resolution of UnresoNed Sa'ety deciding on how these differences should be resofved in an in-Issue (USI) A-43, " Containment Eme'gency Sump Perform- tegrated pfant review, and (3) a documented evatuation of plant ance." safety when the supplement to the Final IPSAR and the Sa'ety Evaluation Report for converting the hcense from a provisional NUREG-080013.5.2 R1: STANDARD REVIEW PLAN FOR THE to a fuit term hcense have been issued The Final IPSAR and its REV:EW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR supplements wm form part of the bases for considenng the con-PCWER PLANTS LWR Ed: tion. Revision 1 To Section 13.5 2. version of the hcense.

  "Operat:ng And Maintenance Procedures." and Revision 0 of Appendix A to Section 13 5 2, "Rev ew-

Off.ce of Nuclear NUREG 0829 DRFT: INTEGRATED PLANT SAFETY ASSESS-Reactor Regu!at.on. Director (pre 851125). Jufy 1985. 290p MENT REPORT, SYSTEMATIC EVALUATION PROGRAM - SAN 8508150055.32195.180. ONOFRE NUCLEAR GENERATING STATION UNIT 1 Docket Revision No.1 to Section 13 5 2 and Revision 0 to Append:x No. 50 206(Southern Cahfornia Ed son Company)
Division of A of Section 13.5 2 of the Standard Review Plan incorporates Licensing (800428-851124). Apnl 1985. 558pp. 8505240058 changes that have been developed since the orig nal issuance 30565 001.

in July 1981. This revision incorporates guidehnes of Task The Systema'c Evaluation Program was initiated in February Action Plan Iterns i C.1 and I C.9 of NUREG 0660 as clanf ed in 1977 by the U d Nuclear Regulatory Commission to review the Supplement 1 of NUREG 0737. Appendix A to SAP Section o,3,gns of older operating nuclear reactor plants to confirm and 13.5 2 was formerly NUREG 0899. document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety re-NUCEG-0800 18.2 R0: STANDARD REV:EW PLAN FOR THE quirements relating to setected issues. (2) a basis for deciding REVIEW CF SAFETY ANALYSIS REPORTS FOR NUCLEAR on how these d fferences should be resolved in an integrated POWER PLANTS LWR Ed tion Revision 0 To SRP Sect;on 18 2* sant review, and (3) a documented evaluation of plant safety.

   " Safety Parameter Display System (SPDS)/'

Office of Nuclear This report documents the review of San Onofre Nuclear Gen-Reactor Regulation, Director (pre 851125) January 1985.11pp erating Statron Unit 1, operated by Southern Cahfornia Edison 8502150068.28962 310. Company The San Onofre 1 facihty is one of 10 plants re-This rev'sion incorporates tne gutdehne of Task Action Plan viewed under Phase 11 of this program. This report indicates item 1.D 2 of NUREG-0660 as clanf ed in Supplement 1 to

                                          .                             how 137 topics seiected for review under Phase 1 of the pro-NUREG-0737," Safety Parameter Display Systern?                       gram were aodressed. Eampment and procedural changes have been identified as a result of the review NUCEG-080018.2A1 R0: STANDARD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR                     NUREG-0837 V04 N03: NRC TLD DIRECT RADIATION MONI-POWER PLANTS LWR Ed. tion Revision 0 To Appendrx A To                TORING NETWORK Progress Report, July Septerrbor 1984.

SRP Section 18 2. " Human Factors Review Guidehnes For The JANG.Ja KRAMARIC.M ; CLHEN.L Aegu 1, C %ce c' Direc-Safety Parameter Display System (SPOS)"

O'fice of Nuctear tor January 1985 146pp 8502040043. 28727.001.

Reactor Regulation, Director (pre 851125). Janua*y19e5 46po Tiis report provides the status and results of the NRC Ther-8502149040. 28943 317. moluminescent Dosimeter (TLD) Direct Radiat<on Monitonng TNs revrsion incorpoia:es the gudehne of Tau Ac ion P!a" Network It presents the radiation levels meast. red in the .icinit, item 1.D.2 of NUREO-0660 as clanfied in Supplement 1 of of NRC bcensed f acihty sites throughout the ca antry for the third NUREG-0737. Appendix A to SRP Section 18.2 was formerly quarter of 1984 draft NUREG-0835. " Human Fa-tors Acceptance Cr.tena for the Sofety Pararreter Display System," Draft Report issm.J fo' NUREG-0637 V04 NC4: NAC TLD DIRECT RADIATION MONI. TORING REPOHT. Progress Report. October December 1984. Com nsnt. JANG.J ; KRAMARIC.M ; COHEN L Reg >on 1 Office of Droc-NUREG-0800 ROS: STANDA RD REVIEW PLAN FOR THE tor. July 1985. 316pp. 8508010750. 31924148. REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR Ths report provides the status snd results of the NRC The'. POWER PLANTS LWR Edition.Reveon 5 To SRP Table Of moluminescent Dosimeter (TLD) Direct Radiation Monitonng Contents.

Office of Nuclear Reactor Regulation Director (pre- Network. It preser fs *he radiation lerefs measured in the vicinity 851125). January 1985. 24pp. 8502140198. 28955140 of NRC hcenard facihty sites throughout the coentry for the Revision 5 to SRP Table of Contents. fourth quarter of 1984.

_ . ._. _. . . _ . _ _ _ _ _ _ _ _ _ _ _ _ . . .m . , i Main Citations and Abstracts 9 NUREG-0837 V05 N01: NRC TLD DIRECT RADIATION MONI- 3 (January 1985) issued by the Office of Nuclear Reactor Regu. TORING NETWORK. Progress Report. January-March 1985. laton of the U S. Nuclear Regulatory Commission with respect j .1 JANG,J.; KRAMARIC,M.; COHEN,L. Regon 1, Offee of Direc- to the apphcaton filed by the Tennessee Valley Authonty, as ' tor. July 1985.152pp. 8508020373. 31961:073. apphcant and owner, for hcenses to operate the Watts Bar Nu-This report provides the status and results of the NRC Ther- clear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). moluminescent Dosimeter (TLD) Direct Radiation Monitonng The facility is located in Rhea County, Tennessee, near the Network. It presents the radiation levels measured in the vicinity Watts Bar Dam on the Tennessee Arver. This supplement pro-of NRC licensed facihty sites throughout the country for the first vides vecent informabon regarding resolution of some of the quarter of 1985 open and confirmatory items and hcense conditions identified in

'                    NUREG 0837 V05 NO2: NRC TLD DIRECT RADIATION MONI.                                       the Safety Evaluation Repert.

TORING NETWORK. Progress Report, Apni-June 1985. JANG.J. KRAMARIC,M.: COHEN.L Region 1. Office of Direc- NUREG 0853 SO4: SAFETY EVALUATION REPORT RELATED tor. September 1985. 225pp. 8510020232, 32828.338. TO THE OPERATION OF CLINTON POWER STATION, UNIT

 ;                          This report provides the status and results of the NRC Ther-                      1 Docket No. 50-461.(liknots Power Company.et al)

Dnnsson of moluminescent Dosimeter (TLD) Direct Radiabon Monitonng Ucensing (800428 851124). February 1985. 70pp. 8503210295.

1 Network. It presents the radiation levels measured in the vienty 29480:158 c*" ' : sites throughout the country for the Supplement No. 4 to the Safety Evaluaten Report on the ap-ane l98 phcation filed by lihnoes Power Company, Soyland power Coop- t i erative, Inc., and Western lihnois Power Cooperative., Inc., as 1 NUREG-0844 DRFT FC: NRC INTEGRATED PROGRAM FOR appheants and owners, for a hcense to operate ire Chnton

!                        RESOLUTION OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 REGARDING STEAM GENERATOR TUBE                                              Power Stabon, Unit No.1, has been prepared by the Office of l

INTEGRITY. Draft Report For Comment. MURPHY,E. Divison nf Nuclear Reactor Regulation of the U S. Nuclear Regulatory Commission The facihty is located in Harp Township, Dww.tt Licensing (800428-851124). Apnl 1985.166pp. 8505310%3. County, Ilhnois. This supplement reports the status of stems that d 30666.007. have been resolved by the staff since Supplement No. 3 was This report presents the results of the NRC integrated pro- issued gram for the resoluton of Unresolved Safety issues A-3 A-4, { and A-5 regarding steam generator tube integnty. The report ad-

  ;                                                                                                     NUREG-0856 DRFT FC: REASSESSMENT OF THE TECHNICAL dresses issues within the areas of steam generator integnty,                         BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For
  )                     plant systems response, human factors, radological conse-                           Comment). SILBERBERG.M ; PASEDAG.W F.; RYDER.C P.; et t

quences and the response of vanous organizations to a steam al Accident Source Term Program Office. July 1985. 265pp. I generator tube rupture. A genenc nsk assessment is provided 8508190634.32262 085 and andcates that nsk from steam generator tube rupture NUREG-0958 describes the NRC staff and contractor efforts events is not a significant contnbutor to total nsk at a given site. to reassess and update the agency's analytcal procedures for i nor to the total nsk to whch the general pubhc is routnely ex- estimating accident source terms for nuclear power plants. The

;                       posed. However, the report identibes a number of achons which effort included development of a new source term analytical              '

l the staff finds as a group would be effective in significantly re- procedure - a set of computer codes - that is rntended to re-ducing the incidence of steam generator tube degradabon, the l i ' j place the methodology of the Reactor Safety Study (WASH-frequency of tube ruptures and the corresponding potental for 1400) and to be used in reassessing the use of TID 14844 as-

}                       significant non-core melt radclogical releases. and occupahonal                     surnptons (10 CFR 100) NUREG 0956 describes the develop-radclogical exposures and whch would be effective in mitigat-                        ment of these codes, the demonstrabon of the codes to calcu-ing the consequences of SGTR events. The actons would also                          late source terms for specife cases, the peer revev of this              l
!                       further reduce nsk and have been designated as " staff recom-                       work, some perspectives on the everall impact of new source mended actons." Final pubhcaten of the report herein, follow-                       terms on plant nsks, the plans for related research protects, and j                       sng a 90-day penod for pubhc comment, will consttute technical the concbseons and recommendatens resutting from the effort.
,                      resolution of Unresolved Safety issues A-3, A-4, and A 5.

NUREG-0847 S03: SAFETY EVALUATION REPORT RELATED NUREG-0857 SOS: SAFET) EVALUATION REPORT RELATED TO THE OPERATION OF WATTS BAR NUCLEAR TO THE OPERATION U PALO VERDE NUCLEAR GENERAT. PLANT, UNITS 1 AND 2. Docket Nos. 50 390 And 50-391.(Ten- ING STATION. UNITS 1.2 AND 3 Docket Nos 50-528,50-529 nessee Valley Authonty)

Divison of Lcensing (800428 And 50-530.(Anzona Pubhc Service Company, et a0
Devison 851124). January 1985. 77pp. 3502060553. 28745:199. of Licensing (800428-85112 t) May 1985. 37pp. 8506240081.

This report supplements the Safety Evaluation Report, 3 M 77 312. ', NUREG-0847 (June 1982), Supplement No. 1 (September Supplement No. 8 to the Safety Evaluabon Report for the ap-1982), am! Suppiement No. 2 (January 1984) essued by the phcaton filed by Anzona Pubhc Service Company. et af, for li. Offee of Nuclear Reactor Regulaton of the U.S. Nuclear Regu- censes to operate the Palo Verde Nuclear Generating Stabon, l latory Commission with respect to the appicaten fJed by the Urwts 1. 2 and 3 (Docket Nos STN 50-528/529/530) located in , Tennessee Valley Authonty, as apphcant and owner, for 14 Mancopa County, Anzona, has been prepared by the Office of L censes to operate the Watts Bar Nuclear Plant. Units 1 and 2 Nuclear Reactor Regulation of the U.S. Nuclear Regulatory ' (Docket Nos. 50-390 and 50-391). The facihty is located in Commesson. The purpose of this supplement is to update the Rhea County, Tennessee, near the Watts Bar Dam on the Ton, Safety Evaluaten Report by providing an evaluaton of (1) addi-nestes Arver. This supplement provides recent informiten re- tionalinformation submitted by the apphcants since Supplement , g,arding resolution of some of the open confirmatory items and No 7 was issued and (2) matters that tne staff had under . facense conditons identified in the Safety Evaluation Report. . review when Supplement No. 7 was issued spacs'acalry those ;

assues which required resuluton onor to plant operation of Unit i NUREG-0447 SO4: SAFETY EVALUATION REPORT RELATED 1 above 5% full power.

' TO THE OPERATION OF WATTS BAR NUCLEAR PLANT,0 NITS t AND 2. Docket Ncs. 50-390 And 50-391. (Ten- NUREG-0857 SO9: GAFETY EVALUATION R8EPORT RELATED nessee Valley Authonty) ' Onnsion of Ucensing (800428 TO THE OPERATION OF PALO VERDE NUCLEAR GENERAT. l 051124). March 1985. 45pp. 8504050283. 29673:18J. ING STATION. UNITS 1.2 AND 3 Docket Nos. 50 528.50-529 , ) This report supplements the Safety Evaluaten Report. And 50-530.(Anzona Pubic Servce Company)

Division of

. NUREG-0847 (June 1982). Supplement No. 1 (September Pressunzed Water Reactor Licensing B (post 851125). Decem-1982), Supplement No. 2 (January 1984), and Supplement No. ber 1985. 63pp. 8601070472. 34187.319. t s l

l

 '10         Main Citations and Abstracts Supplement No. 9 to the Safety Evaluaton Report for the ap-     NUREG-0481 S06: SAFETY EVALUATION REPORT RELATED phcahon filed by Anzona Pubic Sennce Company, et al, for h-          TO THE OPERATION OF WOLF CREEK GENERATING csnses to operate the Palo Verde Nuclear Generating Station,         STATION, UNIT 1. Docket No. 50-482.(Kansas Gas And Electnc Units,1,2 and 3 (Docket Nos. STN 50-528/529/530), located            Company.et at)

Divisen of Licensing (800428-851124). June in Mancopa County, Anzona has been prepared by the Office of 1985. 22pp. 8506240149. 31152.311.

Nuclear Reactor Regulations of the Nuclear Regulatory Com- Supplement No. 6 to the Safety Evaluaton Report related to misson. The purpose of this supplement is to update the Safety the operation of the Wolf Creek Generating Station. Unit No.1 Evaluation Report by providing an evaluation of (1) additonal in- updates the mformation contained in the Safety Evaluaten formation submitted by the applicants since Supplement No. 8 Report, dated Aprd 1982 and Supplements 1, 2, 3, 4, and 5, was issued and (2) matters that the staff had under review dated August 1982, June 1983, August 1983, December 1983, ' when Supplement No. 8 was issued, specifically those issues and March 1985 respectively. Supplement No. 6 concludes that which required resolution pnor to Unit 2 fuel loading and testing the facahty can be operated by the hcensee at power levels up in 5% of full power. greater than 5% without endangenng the health and safety of the pubhc. The Safety Evaluaton and its supplements pertains NUMEG-0069 R01: USI A-43 REGULATORY ANALYSIS. to the apphcation for a hcense to operate the Wolf Creek Gen-SERKlZ,A.W. Dnnsson of Safety Technology (800428-851124). erating Station, Unst No.1 fded by Kansas Gas and Electric October 1985.130pp. 8512100503. 33833.001. Company on February 18, 1980. The Constructon Permit No. This report consists of (1) the regulatory analysis for Unre. CPPR 147 was issued on May 17,1977 and a low power 5% solved Safety issue (USI) A-43," Containment Emergency Sump hcense issued on March 11, 1985. The facdity is located m P$rformance"; (2) the proposed resolution; (3) a summary of Coffey County, Kansas. pubhc comments received and action taken; (4) the Committee NUREG-0845104: U S. NUCLEAR REGULATORY COMMISSION to Review Genenc Requirements (CRGR) minutes related to this USl; and (6) appendices that summarize assumphons, cal- POLICY AND PLANNING GUIDANCE 1985.

Commissioners. ,

Februa 1985.25p 8 0005. 94950 cu me , cons ence analyses, and cost estimates are to set forth the regulatory approach of the Nuclear Regula-tory Commission and to provide the supporting pnnciples to that NUREG-0471 V04 N01:

SUMMARY

INFORMATION approach, to state the major pohcies and planning object ves of REPORT. Data As Of June 30,1985 (Brown Book)

Desion of the Commissiort and to provide a common basis for the devel-Budget & Analysis. October 1985 Sipp. 8511010465. opment of programs, for the estabhshment of prionties, and for 33333 094. the aHocaten of resources.

Provides summary data conceming NRC and its licensees for general use by the Chairman, other Commissioners and Com- NUREG-0487 S05: SAFETY EVALUATION REPORT RELATED misson staff offices, the Executive Drector for Operations, and TO THE OPERATION OF THE PERRY NUCLEAR POWER

    ' the Office Directors-                                                PLANT, UNITS 1 AND 2 Docket Nos. 50-440 And 50-441 (Cleve-NUREG-Oe76 Soe: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF BYRON STATION, UNITS 1 AND 8h2 8            12      F           19     2           50         '

29220:135

2. Docket Nos. 50-454 And 50-455, (Commonwea th Edison Supplement No. S to the Safety Evaluaten Report (NUREG-Cornpany)

Dmson of Licensing (600428-851124). February 0887) pertains to the apphcabon fded by the Cleveland Electnc 1985. 52pp. 8503050075. 29243 290. Pfuminating Company, the Ohio Edison Company, the Pennsyl-Supplement No. 6 to the Safety Evaluaten related to Com" vania Power Company, and the Toledo Edisen Company (the monwealth Edison Company's apphcaten for hcenses to oper- Central Area Power Coordinaton Group or CAPCO). as apph-ate the Byron Staten. Units 1 and 2, located m Rockvale Town- cants and owners, for a heense to operate the Perry Nuclear ship. Ogle County, Ilknois. has been prepared by the Office of Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441)

Nuclear Reactor Regulaton of the U.S. Nuclear Regulatory The report has been prepared by the Office of Nuclear Reactor Commesson. This supplement provides recent information re- Regulation of the U.S. Nuclear Regulatory Commisson. The fa- > g;.rding resolution of the heense conditions identified in the cihty is located in Lake County, Ohio, approximately 35 mdes l SER. Because of the favorable resoluton of the items dis- northeast of Cleveland. Ohc. This supplement reports the ! cussed in this report, the staff concludes that the Byron Staten, status of certain issues that had not been resolved at the time i Unrt 1 can be operated by the hcensee at power levels greater of pubhcation of the Safety Evaluation Report and Supplement than 5% without endangenng the health and safety of the Nos.1 through 4. pubhc. NUREG 0047 S06: SAFETY EVALUATION REPORT RELATED NUf4EG-0481 S06: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF PERRY NUCLEAR POWER TO THE OPERA'lON OF WOLF CREEK GENERATING PLANT. UNITS 1 AND 2. Docket Nos. 50-440 And 50-441.(Cleve-STATION. UNIT 1.Dmket No. 50-482(Kansas Gas And Electnc land Electnc liluminatng Company)

Dmson of LicenSng Company,et al)
Deson of Lsce<ising (800428-851124). March (800428 851124). Apnl 1985.150pp. 8505010117. 30114 2C4 1985. 240pp. 8504030055. 29599.298. Safety Evaluaton Report, NUREG-0887, perta.ns to the apph-Supplement No. 5 to the Safety Evaluation Report related to caton fded by the Cleveland Electnc fl.aminat ng Company on the operation nf the Wcif Creek Generating Station, Unit No.1 behalf of itself and as agent for the Duquesne Light Company, ;

codates the informa, son contained m the Safety Evaluation the Ohc Edison Corrpany, the Pennsyfvania Powei Company, t Report, dated Apnl 1982 and Supplements 1,2,3 and 4, dated and the Toledo Edison Company (the Central Area Power Co-August 1982 June 1983. August 1983, and December 1983, re- ordination Group or CAPCO), as applicants and cwners, for a r spectively Supplement No. 5 also addresses open issues and bcense to operate the Perry Nuclear Power Plant. Units 1 and 2 items concerning the issuance of a five percent low power h- (Docket Nos. 50-440 and W441) The report has been pre-conse. The Safety Evaluaten and its supplements pertains to pared by the Office of Nuclear Reactor Rego!aton of tne US. the apphcation for a hcense to operater the Wolf Creek Generat- Nuclear Regulatory Commesson. The facihty is located m Lake og Station, Unit No.1 fded by Kansas Gas and Electnc Comps- County, Ohio, approximately 35 mdes northeast of Chvetand, ny on February 18, 1980. The Construction Permit No. CPPR- Ohio. This Suppieenent, No. 6 addresses the remaining unre-147 was issued on May 17, 1977. The facihty is located in solved Atomic Safety and Licensing Board contention essues; Coffey County, Kansas, TDI diesel generator rehabihty m Section 9 6.3.1; h/ d ogen con.

Maln Citations and Abstracts 11 trol system design per the new hydrogen rute in Secton 6.2.7, similar events have been reported since IE Bulletin 79-12 was and several issues related to Emergency Plans in Secton 13 3. issued NUREG-0887 S07: SAFETY EVALUATION REPORT RELATED NUREG-0910 R01 S01: NRC COMPREHENSIVE RECORDS DIS-TO THE OPERATION OF PERRY NUCLEAR POWER PLANT POSITION SCHEDULE

D vision of Technical Informaton &

UNITS 1 AND 2. Docket Nos. 50-440 And 50-441. (Cleveland Document Control. January 1985 18pp. 8502060493. Electnc illuminating Company)

Office of Nuclear Reactor Reg- 28749.329 ulation. Director (post 851125). November 1985. 243pp. In compliance with statutory requirements set forth in Title 44 8512 t 20154. 33875:280.

This Supplement No. 7 reports the status of certain issues U S. Code, "Pubhc Pnnting and Documents," and in the apphca-that had not been resolved at the time of pubhcaton of the ble regolations etted in Title 41 Code of Federal Regulations, Safeiy Evaluaton and Supplement Nos. I through 6 to inat "Public Contracts and Property Management." Chapter 101,

     + port. The Perry Nuclear Power Plant facility is located in Lake                     Subchapter B. " Archives and Records " the U.S. Nuclear Regu-County, Ohc. approximately 35 miles northeast of Cleveland,                           latory Commission submitted to the General Services Adminis.

Oho. This report relates to the applicaton for licenses in oper. traton Natonal Archrves and Records Serwces, and to the ate the Perry Nuclear Power Plant Units 1 and 2 tiled by the Comptrofler General a schedule (commonly referred to as a dis-Cleveland Electnc illuminating Company on behalf of itself and positen or retenten schedule) proposing the appropnate dura. as agent for the Duquesne Light Company, the Ohc Edison tion of retenten and the final disposition for records created or maintained by the NRC. Company, the Pennsylvania Power Company and the Toledo Edison Company as apphcants and owners. NUREG-0910 R01 S02: NRC COMPREHENSIVE RECORDS DIS-NUREG-0896 S03: SAFETY EVALUATION REPORT RELATED POSITION SCHEDULE.

Divison of Technical information &

TO THE OPERATION OF SEABROOK STATION, UNITS 1 AND Document Controt February 1985. 24pp 8503010298.

2. Docket Nos. 50-443 And 50-444 (Pubhc Service Company of 29188 287.

New Hampshire,et al)

Divisen of Licensing (800428-851124). See NUREG-0910.R01 S01 abstract July 1985. 04pp. 8508090565. 32112.094.

Supplement No. 3 to the Safety Evaluaten Report for the ap- NUREG 0910 R01 S03: NRC COMPREHENSIVE RECORDS DLS-phcaton filed by Public Service Company of New Hampshire, et POSITION SCHEDULE.

Drvision of Technical information &

Document Control Apol 1985. 22pp. 8505100062. 30286:119 al. for licenses to operate the Seabrook Staten. Units 1 and 2 In compliance with statutory requirements set forth in Tit:e 44 located in Rockingham County, New Hampshire, has been pre- U S. Code, "Public Pnnting and Documents " and in the apphca-pared by the Office of Nuclear Reactor Regulaton of the U S. ble regulations cited in Title 41 Code of Federal Regulatons. Nuclear Regulation Commiss#on. This supplement provides in-

                                                                                          "Public Contracts and Property Management," Chapter 101, format on to update the status of the NRC rewew of the apphca-Subchapter 8, " Arch ves and Records," the U S. Nuclear Regu-tiQn.

latory Commission has published and maintains "NRC Compre. NUREG-0897 R01: CONTAINMENT EMERGENCY SUMP hensive Records Disposition Schedule." (NUREG-0910) for PERCORMANCE.(Technical Findings Related To Unresolved records created or maintained by the NRC. Supplement 3 for-Safety issues). SERKlZ,A.W. Division of Safety Technology wards changes to the General Records Schedules as made by (800428-851124). October 1985. 240pp 8512110209, the National Archives & Records Administration (NARA) and 33847 235. Gewal ScWe 20 for incluson. This report summanzes key technical findings related to Unre' solved Safey issue (USI) A-43, Containment Emergency Sump NUREG 0910 R01 SO4: NRC COMPREHENSfVE RECORDS DIS-Performance. Although this issue was formulated considenng POSITION SCHEDULE.

Diws on of Technical Information &

Document Control. October 1985. 9pp. 8511040062. 33337:091, pressunzed water reactor (PWR) sumps, the genenc safety in comphance with statutory requirements set fortn in Title 44 questions apply to both boiling water reactors (BWRs) and PWRs. Hence, both BWRs and PWRs are considered in this U S. Code, "Pubhc Pnnting and Documents," and in applicable report. The technical findings in this report prowde information regulaton cited in 36 CFR Subchapter XII, National Archives on post LOCA recirculation performance. These findings have and Records Administrabor (NARA) The U S. Nuclear Regula-been denved from extensive expenmental studies, genenc plant tory Commission has published and maintains "NRC Compre-studies, and assessments of sumps and purrps used for long- hensive Records Disposition Schedule," (NUREG-0910 Rev 1) term cochng. The results of hydrauhc tests have shown that the for records created or maintained by the NRC. Supplement 4 potential for air ingeston is less severe than previously hypoth* forwards schedules to the General Records Schedules and ap-proved NRC schedulas as made by the Natonal Archives and esized. The effects of debns blockage on NPSH margin must be dealt with on a plant-scecific basis. These findings have Records Administration for incluson. been used to develop revisens to Regulatory Guide 182 and NUREG-0933 S02: A PRIORITIZATION OF GENERIC SAFETY Standard Rewew P!an Section 6.2 2. (NUREG-0800). ISSUES. EMRIT.R : MINNERS.W; VANDER MOLEN,H.; et al. NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12 SHORT- Division of Safety Technology (800428 851124) January 1985 288pp. 8502210083. 29053 003-PERIOD SCRAMS AT BCiLINSWATER REACTORS CEBEVEC,CJ; HOLLAND,R A. Emergency Preparedness The eport presents the pronty rarkings for genenc safety Branch. March 1985,2tpp 8504050280. IEB-7912. 23673144' issues related to nuclear power plants The purpose c' these IE Circular 77-07 was issued on Apnl 14,1977 because of the ranhngs is to assist in the timdy and Oficient allocation of NRC occurrence of short penod scram events at Dresden Unit 2 on resources for the resoluton of those safety issi.es thac teve a recember 28,1976 and at Monticello on February 23. 1977, significant potential for reducing nsk The safory gnont) rankngs The cirular ady-sed BWR plants to rewse their control rod with- are HrGH, MEDIUM, LOW, arv* DROP and have been assigned drawsl equences and operating procedurus to reduce the likel" on the basis of r'sk significance estimates, the ratio of nsli to hood of futre short perud scrams. However similar events costs and other ingacts e:,tima'ad to sesult si tesolut.ons of the safety issues were imphmented. and the con 9deraton of un-continued December 14, to 1978; occur.etThese Brcwnsincludui Ferry Unit even's at Oyster 1 on January 18 Creek o", certainties and other quantitatue or quahtatrve factors. To the 1979 and at Hatch Unit 1 en January 31,1979 As a result of entent practical, estimates are quantitatue these events, 'E Bulletin 7912 was issuod on Vay 31, 1979 NUREG-0933 S03: A PRIORITIZATION OF GENERIC SAFETY This bulletin required a wntton response from licensees of GE- ISSUES. EMRIT,R ; MINNERS.W; VANDERMOLEN.H.; et al designed BWRs regarding specific actions 1.sted in the bulletin Divmon of Safety Techcology (800428-851124). July 1985 Al; of the hcensees responded in a satisfactory manner. No 269pp. 8508150024 32194 252 l

   . -.                -            . . . -       -         ...        .-. -             __           . _ .   . - - _ _ - . -                - _ ~ --

l 12 . Main Citations and Abstracts I See NUREG-0933,S02 abstract. NUREG-0940 V04 NO3: ENFORCEMENT ACTIONS SIGNIFICANT

                                                                                 #CTIONS RESOLVED.Ouarterfy Progress Report. July-Septem-
]       NUREG 0936 V03 N04: NRC REGULATORY AGENDA.Ouarterly

Enforcement Staff. November 1985. 243pp.

ber 1985. Report,0ctober December 1984.

Dnnsson of Rules and 8512120125. 33875.023.

Records. February 1985. 207pp. 8502250788. 29094 043. This compilation summanzes signifcant enforcement actons , The NRC Regulatory Agenda is a compilaton of all rules on that have been resolved during one quarterty pered (Jufy - Sep-whch the NRC has proposed or is considenng acten and all tember 1985) and includes cop +es of lesters, Notices, and petitons for rulemaktng which have been received by the Com- Orders sent by the Nuclear Regulatory Commission to hcensees l mission and are pendng dispositen by the Commisson. The with respect to these enforcement actons and the hcensees' l Regulatory Agenda is updated and issued each quarter. The responses. It is anticipated that the informabon in this pubica-1 Agendas for Apnl and October are pubhshed in their entirety in tion will be widely disseminated to managers and employees the Federal Register while a notice of availability is published in engaged in activities hcensed by the NRC sn the interest of pro-j the Federal Register for the January and July Agendas- moting pubhc health and safety as well as common defense aM seny I NUREG 0936 V04 N01: NRC REGULATORY AGENDA Quarterly Rzport, January-March 1985.

Divison of Rules and Records. NUREG-0946: AN EVALUATION OF RADIONUCLIDE CONCEN.

Mzy 1985. 201pp 8505310669. 30666.173. TRATIONS IN HIGH-LEVEL RADIOACTIVE WASTES.

;             See NUREG-0936,V03,N04 abstract.                                    FEHRINGER,D.J. Divison of Waste Management. October                 ;

1985. 34pp. 8510310448. 33278.231. NUREG-0936 V04 N02: NRC REGULATORY AGENDAOuarterfy This repoit descnbes a possible approach for development of Report.Apnt-June 1985.

Divison of Rules and Records. July a numencal definiton of ine term high-level radioactue waste."

1985. 216pp. 8508090725. 32101.305. Five wastes are 6dentified which are recognized as being high-See NUREG-0936,V03,N04 abstract. level wastes under current, non-numercal definitens. The Con-sbtuents of Nse wastes are examined and the nest hazardous NUXEG 0936 V04 NO3: NRC REGULATORY AGENDA.Ouarterty component radionucides ve identfied. This report suggests Rrport. July September 1985.

Divison of Rufes and Records. that other wastes with similar concentratens of these radionu-l' October 1985. 210pp. 8511070469. 33384:292. chdes could a'so be defined as high-level wastes.

See NUREG-0936,V03,N04 abstract. NUREG-0956 DRFT FC: REASSESSMENT OF THE TECHNICAL

!       NUREG-0940 V03 N04: ENFORCEMENT ACTIONS SIGNIFICANT                        BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For ACTIONS RESOLVED.Ouarteny Progress Report, October De-                 Comment). SILBERBERG.M,; MITCHELL.J A: MEYER,R.O.; et cember 1984.

Enforcement Staff. February 1985. 402pp- al. Accident Source Term Program Office. Jufy 1985. 265pp.

i 8502250240. 29093.001. 8508190634. 32262.085. I This compilaton summanzes significant enforcement actons NUREG-0956 desenbes the NRC staff and contractor efforts j that have been resolved dunng one quarterty period (October - to reassess and update the agency's analytcal procedures for December 1984) and includes copies of letters, Notees, and .eshmating accident source terms for nuclear power plants. The l Orders sent by the Nuclear Regulatory Commission to icensees effort included development of a new source term ana9tical 7 with respect to these enforcement actons and the hcensees' procedure - a set of computer codes - that is intended to re-responses. It is anterpated that the informahon in this pubhca' place the methodology of the Reactor Safety Study (WASH. 4 ton will be widely disseminated to managers and employees 1400) and to be used in reassessing the use of TID 14844 as-

.          engaged in acturbes hcensed by the NRC,in the interest of pro-          sumptons (10 CFR 100). NUREG-0956 descnbes the develop-
!          moting pubic health and safety as well as common defense and            ment of these codes, the demonstration of the codes to calcu-J           securdy.                                                                fate source terms for specific cases, the peer review of this work, some perspectues on the overalt impact of new source NUREG-0940 V04 N01: ENFORCEMENT ACTIONS SIGNIFICANT                        terms on plant nsks, ne plans for related research projects, and ACTIONS RESOLVED.Ouarterfy Progress Report. January-March               the conclusom and ecommendatom resumng kom N eM 1985.

Enforcement Staff. Apnl 1985. 541pp. 8505100072.

30288.071. NUREG-0970: PROCEDLT".S FOR MEETING NRC ANTITRUST This compHation summanzes significant enforcement actons RESPONSIBILITIES. TOALSTON.A.L.; MESSIER M E.; i that have been resolved dunng one quarterty pered (January . LAMBE,W.M.; et al. Site Analysis Branch. May 1985. 29pp. March 1985) and includes copies of letters, Notices, and Orders 8506070360. 30798.013. j sent by the Nuclear Regulatory Commission to hcensees wth This report descr,bes the procedures used by NRC staff to 2 respect to these enforcement actsons and the hcensees' re- implement the anbtrust review and enforcement presenbed in I- sponses. It is antcipatoo that the informaton in this pubhcation Sections 105 and 186 of the Atome Erwrgy Act of 1954, as f will be widely dissemenated to managers and employees en- amended (the Act), as coverad largely by the Commisson's gaged in actuities Icensed by the NRC in the interest of pro- Rules and Regutabons in 10 CFR Parts 2.101, 2.102, 2 290, l moting pubhc heafth and safety as well as common defense 50.33a. 50.80, and 50 90. These procedures set forth the steps and secunty. and critena the statt apphes an the entitrust review of construc-i ten per' nit and operating hcense apphcahons and the amend-l NUREG-0940 V04 NO2: ENFORCEMENT ACTIONS.SIGNIFICANT monts to those appicabons that deal with changes in owner-j ACTIONS RESOLVED.Ouarterty Progress Report.Apnt- shrp. In additen, the procedures desenbe how the staff en-l June,1985.

Enforcement Staff. July 1985.341pp.8508190616 forces comphance by hcensees when antitrust condiuons have 32260.001. been appended to constructen permits and operating Icensos. .

I This compnauon summanzes significant enforcement actions ' that have been resolved dunng one quarterfy penod (Apnl - NUREG-0975 V03: COMPILATION OF CONTRACT RESEARCH - June 1995) and includes copies of letters, Notices, and Orders FOR THE MATERIALS ENGINEERINf3 BRANCH. DIVISION OF l ENGINEERING TECHNOLOGY. Annual Report For FY 1984.

l l sent by the Nuclear Regulatory Commission to Icensees wita

!           respect to these enforcement actens and the Icensees' re-                Divis on of Engineenng Technology. Apnl 1985. 400pp.              ,

sponses. It is anucipated that tne informaton in this pubhcaton 8506040260.30711.262. will be widely disseminated to managers and employees en- This report presents summanes of the research work per-l formed dunng Fiscal Year 1984 by laboratones and organiza-i gaged in actuities hcensed by the NRC in the interest of pro- bons under contracts administered by the NRC's Matenals Engs-i motng pubhc health and safety as well as common defense I snd secunty. neenng Branch, Offce of Nuclear Regulatory Research. Each l t

Main Citations and Abstracts 13 contractor has wntten a more complete and detaded ar.nual latory Commission. The facihty is located in West Fekcsana report of their work which can be obtained by wr: ting to N3C; Pansh, near St. FrancisvMe, Lou siana This supplement reports however, we beheve it as useful to have a summary of each contractor's efforts for the year combined into one <olume the status of certa:n items that had not been resofved at tr'e time of publication of the Safety Evaruat:en Report and Supple- - NUREG-0979 S03: SAFETY EVALUATION REPOAT RELATED ment No t TO THE FINAL DESIGN APPROVAL OF THE GESSAR il NUREG-0989 S03: SAFETY EVALUATION REPORT AELATED BWR/6 NUCLEAR ISLAND DESIGN. Docnet No. 50-447. (Gen . TO THE OPERATION OF RIVER BEND STATION Docket No eral Electnc Company)

Divtsion of Licensing (800428-851124)

January 1985. 33pp. 8502110615 28860 291 50 458 (Gulf States Utates Company. Cajun E!ectnc Power Co-Supptement 3 to the Safety Evaluat:en Report (SER) for the operatve)

Division of Licensing (800428 851124) August 1985. 30$pp. 8509100337. 32529.148 applicaton fded by General Electnc Company for the final destgn approval for the GE BWR/6 nuclear is'and design has Supp,ement No. 3 to the Safety Evaluat on Report for the ap-been prepared by the Office o' Nuclear Reactor Regulation of obcation tied by Guif States Utaties Company as apphcant and the Nuclear Aegulatory Comrnission. This report suppiements for itself and Ca,un Electnc Power Cooperat ve. as owners, for a the GESSAR II SER (NUREG 0979), issued in Apnl 1983, sum- hcense to operate River Bend Station has been prepared by the manrng the resu!!s of the staff's safety review of the GESSAR O*fice of Nuclear Reactor Regulat4cn of the U S. Nuclear Aegu.

11 BWR/6 nuclear island des gn. Subject to favorable resoluton latory Commission. The facihty is located in West Fenciana of the <! ems discussed in this supp'ement. the staff concludes Pansh, near St Francisvale. Louis.ana The supptement reports that the GESSAR 11 design sat.sfactonly adcresses the severe- the status of ccrta.n :tems that had not teen reserved at the acc: dent concerns descrted in dra't NUAEG 1070 t;me of pub 6 cation of the Safety Eva!uat on Repo't and Supple-ment Nos 1 and 2 NUREG 0979 SO4: SAFETY EVALUATION REPCRT RELATED TO FINAL DES:GN APPROVAL CF THE GESSAR 11 BWR/6 NUREG-0989 SO4: SAFETY EVALUATiCN REPCRT AELATED NUCLEAR ISLAND DEStGN Docket No. 50-447.(Generaf E ec. TO THE OPEAATICN OF R VER BEND STATION Docket No. tnc Company)

Division of Ucens1ng (800428-851124) July 50 458 (Gulf States Utdit es Company Capn E: ectr,c Power Co.

1985. 91pp. 6508020356. 31961226. operatve)

Division of Ucensing (800428 851124) September Suppiertent 4 to the Safety Evatat:on Recert (SEA) for tne 1985. 39pp 8509260506 32758 043.

appbcat:on fded by General E!ectnc Company for the final Suppeent No 4 to tre Sa'ery Evaluat.on Aeport for the ap-des.gn approval for the GE BWR/6 nucrear island design pbcat.on f< led by Gu f States Utd tes Company as apphcant and (GESSAR 11) has been prepared ty the Off;ce of Nuclear Reac. for itself and Ca;un E'ectrte Power Coope'ative, as owners. for a for Regulation of the Nuclear Regulatory Commissicn Th s hcense to operate R ver Bend Staton has teen prepared by tre report supp'ements the GESSAR 11 SER (NUREG-0979) issued Off ce of the Nuclear Reactor Aegu'abon of the U S Nuc! ear in Apnl 1983 summanzcg the resutts of the sta+f s safety rev'en RegWatory Commission. The facety is located en West Fehciana of the GESSAR 11 BWR/6 nuciear istand design; Supplement 1 Parish, nea' St. Franc sche. Louis <ana issued in Jufy 1983; Supp!ement 2. issued in Novemter 1984-and Supp:ement 3. issued in Janua y 1985. Subject to favoratie NUREG 0989 S05: SAFETY EVALUATION AEPCAT AELATED resolution of the items d scussed in this supplement the staff TO THE CPEAATION OF river BEND STAT >CN Docket No. concludes that the GESSAR 11 design sat!sfactonly accresses 50-a58 (Gulf States UtAt.es Company. Cajun Electnc Power Co-the severe accident concerns descrted in the Commiss,on's operatNe)

Ctt,ce of Nuclear Reactor RegWat on, D, rector (post Pobey Statement on Severe Reactor Accidents Aegarding 851125) November 1985 5800 8512100373 33933 224 Future Des gns and Exist ng Ptants Supplerrent No. 5 to the Sa'ety Evafuaton Report for tae ap-phcaton taed cy Gulf States Utaties Company as apphcant and NUREG-0981 R01: NAC/FEVA CPEAATIONAL PESPCNSE for 'tse'f and Ca,un Electric Po*er Cooperat,ve. as ceners for a PROCEDURES FOR RESPONSE TO A COMVEACIAL NUCLE. hcense to coerate A ver Bend Station has been prepared by tne AR AEACTOR ACCIDENT.
Director's C'f.ce. C'fice of Inspec. O'fice of the Nuc ear Reactor Aequiation of the U S Nue:ea-tron and Enf0rcement.
Federal Emergency Management Aegulatory Comm.ssion The facAty is located in West Febc>ana Agency. FeOruary 1985. 34pp. 8503060141 FEMA 51. Pansh. near St_ Franc!sel!e. Louts ana 2926128L Procedures have been developed by the U.S Nuclear Regu- NUREG 0991 SO4: SAFETY EVALUATION AEPCAT AELATED latory Commiss.on (NRC) and tne Federaf Emergency Manage. TO THE OPERAT!CN CF UVERfCK GENERATING ment Agency (FEMA) which prowde tre response teams of both STATION UNITS 1 AND 2 Docket Nos ~,0 352 And 50-agencies
tn the steps to be tamen in respond ng to an emer. 333 (png,3,,phia Electnc Cortpany)
Dews on of Licens.ng gency at a commercial nuclear oc*er plant The emphas.s of (800428-851124) May 1985 5100 8506060596 30776 256.

these procedures is ma:nfy on the inter' ace between NRC and in August 1983 the NAC Sta'f issued its Sa'ety Eva'uaton FEMA at their respectve Headqua ters and Regional O'frcos Aeport regard ng the appbcation for hcer ses to operate the Lim-and at the vanous s.tes at *hich such an emergency could enck Generat ng Staton. Units 1 & 2 rocated on a s.te :n Mont-occur. Detaled procedures are presented that cover for Doth gorrery and Chester Countes. Penrisylvania Supp'ement 1 *as ogencees, hoMcaten schemes and manner of actvat.on, orga- issued in Dece ate' 1983 aad add'essed several o/ stand.ng nizations at Headqua*ters and the s.te, interface procedures. co. issues It also contains the comrrents made by the Adesory ordinaton of cnsite and of' site coerators, the roie of the Senior Committee on Reactor Sa'eguards in its intenm report dated FEMA oftc al, and the cooperative ef' orts of each agency's October 18, 1983 Supplement 2 *as issued in Octccer 1994 pubhc informaton sta'f Supp>ement 3 *as issued in October 1984 and accressed issues that reouired resoluton before issuance of the operateg NUREG-0989 S02: SAFETY EVALUATION AEPOAT AELATED hcense for Unit 1. On October 26,1994 a hcense (NPF 27) for TO THE OPERAT6CN OF A VER BEND STATION Docket No. Unit 1 was is ued which was restncted to a f:ve percent power 50 458 (Gulf States utst.es Correany. Cajun Electnc Power Co- level and conta red cond't ons which required rescuttor, unor to operat ve)

Division of Licensing (800428-851124). August proceed!ng beyond the five percent power level Th>5 Sum e.

1985. 247pp. 8508210406 32338 015 Supplement No 2 to the Safety Evatation Report for the ap. ment 4 addresses some of those technical issues and therr as. sociated kcense cond tions which reqare resolut on prior to pro. phcation fJed by Gulf States Ut+tes Corapany as appocant and fvr 4tself and Cajun Electnc Power Cooperatve, as owners, for a ceeding beyond the five percent power levet The terra ning issues will be addressed in a later supp!ement to this report-bcense to operate River Bend Staton has been prepared by the Thc load.ng at afl cperat;rg plants, in Leu of requinng qua'.fw 1984. This Supp:ement 5 to the SER addresses further issues caten to the current cr,tena that a e appted to rew plants that require resolution pnor to proceed rg beyond the five per-cent level. NUREG-1031 501: SAFETY EVALUATION REPCRT RELATED TO THE CPERATION OF MtLLSTONE NUCLEAR PCWER NUREG 0991 S06: SAFETY EVALUATION REPCRT RELATED STATICN.UN!T 3 Docket No. 50 423. (No'treast Nue! ear TO THE OPERATICN CF LIVER:CK GENERAT.NG Energy Company)

~~Divis on of Ocers.rg (tt00428-851124L STATICN UNITS 1 AND 2. Docket Ncs 50 352 And 50- Ma'ch 1985. 7500 8504090018 29753 001.~~
~~353 (Philade phia Elecir.c Corrpany)~~
Div:s:en of Licens ng The Sa'ety Eva'uation Report prov' des the resuits of tre NAC (800428-851124) August 1985.19pp. 85C8210037. 32302 317. StaM rev'e* of Northeast Noctear Ene gy Compary's apphcat.cn In August 1983 the sta'f of the Nuclear RegJatory Commis- for a beerse to cperate the M,l: store Nuclear Power Plaat. Un:t sien issued 4ts Safety Evatat.on Report (NUREG 0991) regard, No. 3 The facmty is located in Water *cid Townsh'p. New ing tre appbcation of tre Philade:phia Eiectnc Compary (tre 1,. London, Conrect. cut This Suppiemert No 1 updates tre .nfor-censee) for hcenses to operate the Limer,ck Gererat.rg Station. mat,on conta.ned in ',he Sa'ety Eva4at on Aeport. dated July Units 1 and 2, located on a s te in Mor!gomery and Chester 1984 Th.s s pp;ement s'so addresses tae ACAS Report issued Count:es. Pennsylvania Suppiement 1 *as issued in December Septemoer 10,1984 1983, Sopplement 2 was issued in Octeter 1984. Supp;ement 3 was issued Octcber 1984. Supplemert 4 was issued in Vay NUREG 1031 S02: SAFETY EVALUATICN RELATED TO THE 1985. Suppiement 5 was issued in July 1985 and Suppfement E CPERATICN CF M LLSTONE NUCLEAR PCWER issued in August 1985 This suppiement 6 addresses fu rther STATION. UNIT 3 Docket No SO 423 (Northeast Nweiear Erergy issues, pnncipa4y the status cf of*s te emergency p anneg. trat Corrpacy)

Division of Licens.rg (800428 851124) September recu;re resolution pner to proceed.rg beyond the f,ve percent 1985 10000 8510010227. 32836 195 power level. The Safety Evatation Report issued in August 1984 provided the results of the NRC staff rev ew of Ncrtreast Noctear Energy NUREG-1021 R01: CPERATOR LICENS+NG EXA%NER STAND. Company's apphcat,on for a hcense to operate the M.ustone Nu-ARDS.

Division of Human Factors Sa'ety (800428-851124) c! ear Power Stat cn, Unit No 3 Swpoiement No.1 to that Februa*y 1985.195pp. 8503200229 29473 058. report. issued in March 1985 updated the inforrration conta ned The Operator Licens:ng Examiner Standards provide pokey 'n tre Safety Evatuation Report and addressed the ACRS and gwdance to NRC examiners and estachsh the procedures Peport issued on September 10,1984 This Report. Supprerrent and practices for examining and hcensing of Tit:e 10 of t*e No. 2 updates the eformaten conta,ned in the Sa'ety Evalua-CCCE CF FEDERAL REGULATIONS (10 CFR 55t They are in. t.on Report and Suppterrent No 1 and aodresses poor unre-tended to ass st NRC examrers and 'ac+ty bcensees to under. so'ved items. The facar, is :ocated .n Water *crd Tcwa s%p New stand the exammation process cetter and to provide for eccita-cle and centstent admin strat on of exammations to a!! apph. London. Conneticut cants by N9C eamme s These star dards eri not a tutsttute NUREG 1011 S03: SAFETY EV ALU ATION REDCRT RELATED for the coeratur heens.ng requistions and are subject to rarsion TO THE OPERAT'CN OF VLLSTONE NUCLEAR POAER or other internal operator examination ucenstng pobcy changes STATICN.UNF NC 3 Ccdet du. 50-423 (Ncrtaeas Nuc! ear As apprognate, these standards *di be rev: sed penodicany to Energy Corrpany)

O'f.ce of Nuclear Reactor 81 egutaton. De accommodate comments and ref:ect new information or expen- rector (post 851125) Noember '985 83pp 8512030475 enca. 33724 29 EVENT REPCRT The Safety EvaLation Report issued in August IC64 provided NUREG 1072 S02: ' LtCENSEE the resa:ts of the NRC staff review of Ncrtheast Nuc' ear Energy SYSTEM Eva:uat.on Of Frst Year Results And Recommenca. Corrpany's appLcat on for a 4 cense to operate the Matstone Ne tons For irrprovements. HE900N F J. AEOD, Director s O*f-ce .
Septemter 1985. 84pp. 8509230665. 32701299 clear Power Stat.on, Unit No 3 Suppiement No 1 to that This :eport desentes an evaNabon of an industrpw de report, issued in March 1985 updated the inbrrnatan conta.ned sarrple of Licensee Event Reports (LERs) that was conducted in the Safety Eva uation Report and addressed the ACRS to determine whether or not these LERs were prepared in ac- Report issued on Septerrter 10. 1984 Supplement No. 2 L
Main Citations and Abstracts 15 issued in September 1985 updated the information contained in the Safety Evaluation Report and Supplement No I and ad- NUREG 1037 DRFT FC: CONTAINVENT PERFORMANCE WORKING GROUP REPORT Draft Report For Comment.
Divi-dressed prior unresolved items This Supplement No 3, pro- sion of Engineenng (pre-851125). May 1985 322pp vides more recent information regarding resolution or updating 8506140588. 30935 257.
of some of the open and confirmatory items and license conds-Conta ninent ouildings for power reactors have been studied tions identified in the Safety Evaluaten Report. The facihty is to-cated in Waterford Township, New Condon County, Connects- to estimate their leak rate as a funct on of increasing internal cut. pressure and temperature associated with severe accident se. ouences involving s.gnificant core damage Potential leak paths NUREG 1031 SO4: SAFETY EVALUATION RELATED TO THE througn conta.nment penetraten assembhes (such as equip. OPERATION OF MILLSTONE NUCLEAR ment hatches, airlocks. purge and vent va!ves, and electncal POWER STATION _ UNIT 3 Docket No. 50 423. (Northeast Nuclear penetratons) have been identified and their contrtutions to Energy Company)
Office of Nuclear Reactor Regulation De leak area for the containment are incorporated into containment rector (post 851125r November 1985. 232pp 8512260100 leak rate and pressure temperature response as a functon of 34063 073- time Because of lack of rehable emperimental data on the feak.
The Safety Evaluaton Reoort issued in August 1984 provided age behaver of conta nment penetrations and isolaten bamers the results of the NAC staff rcyten of Northeast Nuclear Energy at pressure beyond their des.gn cond tions, an anarytical ai> Company's apphcation for a hc6nse to operate the Milfstone Nu- proach has been used to estimate the leakage behaver of com-crear hwer Station, Unit No 1 Suoptement Na.1 to that ponents found n specif.c reference plants that approximately report, iss ad in March 1985 updated the informatori contained charactenze the vanous containment types in the SaV/ $ valuation Report and addressed the ACRS Report issued on September 10,1984 Supplement Nos 2 and NUREG-1038 S02, SAFETY EVALUATION REPORT RELATED 3 dated September 1985 and November 1985, respectively, up- TO THE CPERATION OF SHEARON HARRIS NUCLEAR dated the inforrnation contained in the Sataty Evaluation Report POWER PLANT, UNIT 1. Docket No. 50-400. (Carohna Power and Supplement No.1 and addressed pnor unresolved items And Ught Company And North Carchna Eastern Munic pal Th s Supplement No. 4, addresses the items concerning the is- Power Agency)
~~Dmsson of Licensing (800428 851124) June 1985 65pp. 8506270137. 31258 277.~~
suance of a low power hcense (5*,) The facshty is located in Waterford Township, New London County, Connecticut- Supp!ement No 2 to the Sa'ety Evaivaton Report for the ap-phcaten fJed by Carchna Power and Ught Company and North NUREG 1032 DRFT FC: EVALUATION OF STATION BLACKOUT
* # # #NW *"" "
ACCIDENTS AT NUCLEAR POWER PLANTS Technical Find-ings Re:ated To Unresofved Safety issue A-44 Draft Report For aW m Wau aN Nam Mu M Ca6 Comment. BARANOWSKI.P W. Office of Nuclear Regulatory na. as been maw h N N of Mar nan NgWa-Research, Director.
Office of Nuclear Reactor Regu!ation, Di- ton of the U S Nctear Watory Gmon %s sWe-rector (pre-851125) May 1985. 200pp 8506250217. 31212 301. ment provides m re recent information regarding resolution of
"$taton 8:ackout," which is the Complete loss of alternatng some of the open items identified in tr'e Safety Evalutaten current (AC) electncal power in a nuclear power plant, has been Re ort and in Supplement No 1. It also addresses one of the designated as Unresolved Svery issue A-44. Because many recommendatons of the Advisory Committee on Reactor Safe-safety systems required for reactcr ccre decay heat removal guards in its report on the Shearon Hams Plant, dated January and contairsment heat removal depend un AC power, the conse- 16,1984, which was inadvertentty omitted in Suppiement No 1.
quences of a station blackout could be severe This report doc-uments the findings of technical stud:es per'ormed as part of NUREG 1046: DISPOSAL CF HIGH-LEVEL RADtOACTIVE the program to resolve this issue. The imponant factors ana- WASTES IN UNSATURATED ZONE. TECHNICAL CONSIDER-tyred include. the frequency of Icss of offsite power; the proba- ATIONS AND RESPONSE TO COMVENTS HACK 8ARTH.CJ; bihty that emergency or onsite AC power supphes would be un- NtCHOLSON.T J ; EVANS.D D Divison of Radiaton Programs avadable, the capabihty and rehacihty of decay heat removal & Earth Sciences (post 840429) October 1985 49pp systems independent of AC power; and the Okl! hood that offsite 8510290415 33259 295 power would be restored before systems that cannct operate On Ju'y 22,1985, the U S Nuclear Regulatory Commission for extended penods w'thout AC powe, f&l, thus resulting in (NRC) promulga.ted amendments to 10 CFR Part 60 concerning core damage. Thrs report also addresses effects of different de- disposal of high-level radoactrve waste (HLW) in geologic re-signs, locations, and operatonal features on the estimated fre- ostones in the unsaturated zone (50 FR 29641). This report quency of core damage resulting from station blackout events. pnncipal technical issues consid-ered by the NRC staff dunng the development of these amend-NUREG 1033: FINAL ENVIRONMENTAL STATEMENT RELATED ments it expands or revises certain technical discussions ongo TO THE OPERATION OF WPPSS NUCLEAR PROJFCT NO nally presented in draft NUREG 1048 (February 1984) based on 3 Docket No 50-508 (Washington Pubhc Power Supply System) pubhc comment letters and a1 increastng understanding of the
Drve on of Licensing (800428-851124) May 1980 247pp physical, geochemical, and hydrogeolog'c procssses operative 8575310C35. 30671.001. in unsaturated geologic rnedia The following issues related to the Final Environmental Statement related to tre operation d'sposal of HLW witNn the unsaturated zone are discussed hy-of Washington Nuclear Proicct No. 3 by Washington Pubhc drogeologic properties and conditons, heat dissipaton and tom-Power fupp9 ivstem, et .nl (Locket No. 50 508), located in perature, geocnemistry, retnevauhty, potential for exhumation of Grays Hartor Ct anty, Washington, has been prepared by the the radcactive waste by natural causas and by human mtrusion, Office of Nuclear Reactor Regulation of the U S. Nuclear Regu. the effects of future chmatic changes on the level of the region-latory Commisson. This statement reports on the staff's review al water table, and transport of radionuchdes in the gaseous of the impact of operaten of tne plant Also included are com- state The changes to 10 CFR Part 60 st. definitens, siting crito-ments of Sia's and 'ederal governtrents, local agencies and na, and demgn cntena for the geomgic repostory operations members of the pubhc on the Draft Environmental Statement area are discussed Other entena esamened by tne NRC staff for tNs project and statt responses to tnese comments The but which were not changed in the rule are the minimum 300-NRC staff has concluded, based on a weighing of environmen. meter depth for waste emplacement, hmitations on esploratory tal, technical and other factors, that an operating license could boreholes, backfill requirements. waste package design cntena.
~~be granted and provisons for ventdation~~
~~- - _ . - -_ _ _ . -- - - - - . - - - _ - - - _ . - - ~ - _-~~
o i 1 i 16 Main Citations and Abstracts NUREG 1047: SAFETY EVALUATION REPORT RELATED TO Supplement No. 2 to the Safety Evaluation Report on the ap- ' THE OPERATION OF N!NE M!LE POINT NUCLEAR phcaten fded by Pubhc Senoce Electnc and Gas Company as STATION UNIT NO. 2 Docket No. 50 410 (N,agara Moha*k appl. cant for itself and At: antic City Electnc Company, as Power Corporation.et al)
Divis on of Licensing (800428- owners, for a hcense to operate Hope Creek Generating Station 851124). February 1985. 652pp. 8502210335 29054 001. has been prepared by the Office of Nuclear Reactor Regulation The Stety Evaluation Report for the apphcahon fded by the of the U S Nuclear Regulatory Commrssion. The facihty is locat-N.agara Mohawk Power Corporation, as applicant and co- ed in Lower Alicways Creek Townsh,p in Sa:em County, New owner, for a hcense to operate the Nine Mae Point Nuclear Sta- Jersey. This supplement reports the status of certain items that tion, Unit No. 2 (Docket No. 50 410), has been prepared by the had not been resolved at tne t me of pubkcation of the Safety Office of Nucl ear Reactor Regulat on of the U S. Nuclear Regu- Evaluaten Report.
latory Commisson. The facility is located near Oswego, New York. Subject to favorable resolut2on of the items discussed in NUREG 1048 S03: SAFETY EVALUATION REPORT RELATED this report, the NRC staff concludes that the facdity can be CD. TO THE OPERATION OF HOPE CREEK GENERATING erated by the appbcant w>tnout endangenng the heaftn and STATION Docket No. 50-354 (Pubhc Senoce Electnc And Gas Company,At' antic City Electnc Company)
~~Division of Licensing sa'ety of the pubhc.~~
(600428 851124). October 1985 100pp. 8511110437, NUEEG 1047 S01: SAFETY EVALUATION REPORT RELATED 33416 273 TO THE OPERATION CF NINE M;LE POlNT NUCLEAR Supprement No 3 to the Safety Evaluation Report on the ap-STATION. UNIT NO. 2 Docket No 50 410. (Niagara Mohawk phcaton fded by Pubhc Senece Electnc and Gas Company as Power Corporat:on.et a))
Dms on of Licensing (800428 applicant for itself and Attantic City Electnc Company, as 851124). June 1985. 35pp 8507050411. 31372 268. owners. for a hcense to operate Hope Creek Generating Staton This report supplements the Safety Evaluation Report has been prepared by the Oftce of Nuclear Reactor Regulation (NUREG-1047, February 1985) for the apphcation fded by N,ag- of the U S Nuc' ear Regulatory Commission. The f achty is locat-ara Mohawk Power Corporation, as appbcant and co-owner, for ed in Lower Alloways Creek Township in Salem County, New a hcense to operate the Nine Mde Point Nuclear Station Unit 2 Jersey This supplement reports the status of certain items that (Docket No 50-410). It has been prepared by the Office of Nu- had not been resolved at the time of pubhcaton of the Safety clear Reactor Regulaten of the U.S. Nuclear Regulatory Com. Evaluaten Report and Supplements 1 and 2.
mission. The fac,hty is located near Oswego, New York Subject to favorabie resolution of the eterns d scussed in this report. the NUREG 1048 SO4: SAFETY EVALUATION REPORT RELATED NRC Sta't Concludes that the facMfy can be operated by the ap- TO THE OPERATION OF HOPE CREEK GENERATING pbcant without endangenng the hea'th and safety of the pubhc. STATION Docket No 50-354 (Pubhc Senace Electnc And Gas Company.Atlant,c C ty Electt:c Company)
Omco of Nuc! ear NUREG 1047 S02: SAFETY EVALUATION REPORT RELATED Reactor Regutation. Director (post 851125) December 1985 TO THE OPERATION OF NINE MILE POINT NUCLEAR $3pp 8601070492 34197.329 STATION. UNIT NO 2 Docket No 50-410. (Niagara Mona*k Supplement No 4 to the Safety Eva'uation Report on the ap-Power Corporaten)

Omce of Nuclear Reactor Regulaton, Di. phcation fded by Pubhc Senoce Electnc and Gas Company on rector (post 851125), November 1985. 97pp 8512190253 'ts own behalf as co-owner and as agent for the other co-340f t.023. owner, the Atlantic City E!ectnc Company, for a hcense to oper-Thrs report supplements the Safety Evaluation Report ate Hope Creek Generat.ng Station has been prepared by the (NUREG 1047. February 1985) for the appkcation fded t'y N ag. Office of Nuclear Reactor Regulation of the U S Nuclear Regu-era Mohawk Power Corporation, as apphcant and co owrer, for a kcense to operate the N<ne Mde Point Nuclear Station Unit 2 Jatory Commission. The facility is located in Lower Alloways Creek Township in Satem County, New Jersey This supplement (Docket No 50 410). It has been prepared by the Office of Nu.
reports the status of certain items that had not been resolved at clear Reactor Regulation of the U S Nuclear Regulatory Com. the time of pubhcation of the Safety Evaluation Report. misson. The facihty is located near Oswego, New Ycrir Suppie. ment 1 to the Safety Eva!uation Report was poveshed in June NUREG-1057: SAFETY EVALUATION REPORT RELATED TO 1985 and contained the report from the Advis>y Cohmttee on THE OPERATION OF BEAVER VALLEY POWER Reactor Safeguards as well as ther resoluton to a number of STATION. UNIT 2 Docket No. 50 412 (Duquesne Light outstanding issues from the Safety Evaluation Report. Subsect Company et al)
~~Divisinn of Licensing (800428-851124). Octo-to favorable resolution of the issues d'$ cussed in this report, the ber 1985. 639pp. 8510290574. 33260 001.~~
NRC sta'f concludes that the f acihty can be operated by the ap- The Sa'ety Evaluation Report for the apphcation filed by Du-phcant without endangenng the health and safety of the pubhc. Quesne Light Company, et al for a bcense to operate the Beaver Vaney Power Station Unit 2 in Beaver County, Pennsyi. NUREG 1048 $01: SAFETY EVALUATION REPORT RELATED vania. has been prepared by the Office of Nuclear Reactor Reg-TO THE OPERATION OF HOPE CREEK GENERATING ufaton of the U S, Nuclear Regulatory Commission. Subject to STATION Dockst No. 50 354 (Pathc Service Electne And Gas favorable resolution of the items discussed in the Safety Evalua-Company.At: antic Oty Enctne Company)
Du sion of Licensing tion Report, the sta'f concludes that the plant can be operated (800428-851124r March 1985 77pp. 8503270534 29542 275 Suoplement No.1 to the Safety Evaluation Report on the ap- by the Duquesne Light Company without er dangenng the health phcaten fded by Pubiec Servce Electnc and Gas Company as and ssfety of the pubhc.
tppricant 'or itself and Atlantic Oty Electnc Compary, as NUREG 1061 V02: REPORT OF THE U S NUCLEAR REGULA-owners, for a keense to operate Hope Creek Generating Staten TORY COMMISSION P! PING REVIEW COMMITTEE. Volume hcs oeen prepa's1 by the Othce of Nuclear Reactor Regulat on 2 Evaluaton Of Seismic Designs A Review Of Seismic Design of the U S. Nuclear Regulatory Commission. The facility is locat- Requirements For Nuclear Power Plant Pcing
Piping Review ed in Lower Alloways Creek Township in Satem County, New Committee. Apnl 1985. 213pp. 8505090016. 30254 043 Jersey. This supplement reports the status of certain items that Th.s document repo'ts tn1 posnion and recommendations of h d not been resolved at the time of pubhcation of the Safety the NRC Piping Review Committee, Task Group on Seismic Evaluat on Report. Design. The Task Group considered overlapping conservatism NUIEG-1048 S02: SAFETY EVALUATION REPORT RELATED in the vanous steps of seismic design, the effects of using two TO THE OPERATION OF HOPE CREEK GENERATING tevets of eartnquake as a design cntenon, and current industry STATION Docket No. 50-354 (Pubhc Service Electnc And Gas practices issues such as damping values, spec'ra modification, Company. Atlantic Ory Electnc Company)

Division of Licensing multiple response spectra methods, norffe and support design, (800428-851124). August 1985. 92pp. 8508190624 32302143 design margins, snelastic piping response, and the use of snub-l l
L
Main Citations and Abstracts 17 bers are addressed. Effects of current regulatory requirements Future Designs and Existing Plants." It provides an overview of for piping design are evaluated, and recommendations for im-mediate beensing action, changes in existing requirements, and comments recewed from the pubhc and the Advisory Committee on Reactor Safeguards and the staff response to these. In adds research programs are presented Additenal background infor-mation and suggeshons given by consultants are also present- tion to the Pohcy Statement, the report discusses how the poh. d caes of this statement relate to other NRC programs, including the Severe Accident Research Program: the implementation of NUREG-1061 V02 ADD: REPORT OF THE U.S. NUCLEAR REG. safety measures resulting from lessons learned in the accident ULATORY COMMISSION PlPING REVIEW at Three Mi'e Island, safety goal development. the resoluton of COMMITTEE. Volume 2 Addendum Summary And Evatuaton Of Unresolved Safety issues and other Generc Safety issues, and Historical Strong. Motion Earthquake Seismic Response And possible revisions of rules or regulatory requirements resulting Damage To Aboveground Industnal Piping.
Piping Review from the Severe Accident Source Term Program. Also dis-Committee.

Stevenson & Associates. Apnl 1985. 211pp. cussed are the main features of a genenc decisen strategy for 8505100057. 30268.005. resolving Regulatory Questions and Techncal issues relating to Earthquake expenence data for industnal piping has been severe accidents, the development and regulatory use of new summanzed in this report. Conclusions and recommendations safety informaton, the treatment of uncertainty in severe acci-for improving the design of nuclear plant piping are made by the dent decison making; and the development and implementaten author, input from R. L Cloud, P. Yaner, and H. Shibata has of a Systems Reliability Program for both existing and future been included. The matenal in this report served as background plants to ensure that the realized level of safety is commensu-informaton for the NRC Piping Review Committee Seismic rate with the safety anafyses used in regulatory decisons.
Desgn Task Group (and their consultants) in the development of the positions given in NUREG 061 Volume 2, NUREG-1073: FINAL ENVIRONMENTAL STATEMENT RELATED NUREG 1061 V05: REPORT OF THE U.S. NUCLEAR REGULA- TO THE OPERATION OF RIVER DEND STATION Docket No. TORY COMMISSION PIPING c1EVIEW COMMITTEE. Volume 50-458 (Gu!f States Utilities And Cajun Electnc Power Coopera.
~~5. Summary Piping Review Committee Conclusens and Rec. tive)~~
Divison of Licensing (800428 651124). January 1985 ommendations.

Piping Review Committee. Apnl 1985. 55pp. 140pp. 8501240053. 28559.174.
8505070580. 30209 240. This Final Environmental Statement contains the second as-This document summanzes a comprehensive review of NRC sessment of the environmentalimpact associated with the oper. requwements for Nuclear Piping by the U S. NRC Piping Review aten of River Bend Station, pursuant to the Natonal Envson-Committee. Four topical areas, addressed in greater detail in mental Policy Act of 1969 (NEPA) and Title 10 of the Code of Volumes 1 through 4 of this report, are included (1) Stress Cor. Federal Regulat ons, Part 51, as amended, of the Nuclear Reg. rosion Cracking in Piping of Boiling Water Reactor Plants, (2) ufatory Commission regulatons. This statement examines the Evaluabon of Seismc Design. (3) Evaluaten of Potential for environment. envronmental consequences and mitgating ac-Pipe Breaks. and (4) Evaluaten of Other Dyname Loads and tons. and environmental and econome benefits and costs. Load Combinations. This volume summanzes the major issues, reviews the interfaces, and presents the Committee's conclu. NUREG-1079 DRFT FC: ESTIMATES OF EARLY CONTAINMENT sons and recommendations for updating NRC requirments on FROM CORE MELT ACCIDENTS Draft Report for Comment
these issues. This report also suggests research or other work Othee of Nuclear Reactor Regulat on, Director (post 851125) that may be required to respond to issues not amendable to December 1985. 255pp. 8601070488. 34184 024.
resoluton at this time. The thermat-hydraulic processes and conum debns-matenal NUREG-1065 Rot: ACCEPTANCE CRITERIA FOR THE LOW EN-
' "" #" "" "O'"*""' * #" "' **' ' ** P #" "' *#' ""
RICHED URANIUM REFORM AMENDMENTS E MEIGH.C.W - *
GUNDERSEN.G.E.; WITHEE.CJ. Onnsion of Safeguards. Aprii " " #8 1985. 53pp. 8504240693 29988.183 * # # *# **N"#" **
This report documents a standard format suggested by the been estimated for the sa vanous LWR containment types used NRC for use in prepanng fundamental nuclear matenal control w n n a s unvnanes Mmng me analyses am plans as required by the Low Ennched Uranium Reform Amend- presented and an interpretatinn of the results provided ments (portions of 10 CFR Part 74) The report also riesenbes the necessary contents of a comprehensive plan and provides NUREG-1080 V02: LONG-RANGE REdtARCH PLAN FY 1986-example acceptance entena whch are intended to communi- FY 1990.
~~Office of Nuclear Regutatory Research, Director.~~
August 1985.157pp. 850913004 7. 32621067. cate acceptable means of achieving the performance capabili-ties of the Reform Amendments. By using the suggested The Long. Range Research Plan (LRRP) was prepared by the format, the hcense applcant i wil! minimize administrative prob- Office of Nuclear Regulatory Research (RES) to s4sist the NRC lems associated with the submittal, review and approval of the in coerdinahng its long. range research planning with the short-FNMC plan. Preparation of the plan in accordance with this range budget cycks. The LRAP lays out programmate ap-format w"! nest the NRC :n eva uatng the plan and in stand' proa@ M for research to Np resuvo regulatory is ues. The ardizing the review and Icensing process. However, conform- plan will he updated annually. ance with this guidance is not regurred by the NRC. A hcense vphcint who employs a format that provides an equal level of NUREG 1085: FINAL ENVIRONMENTAL STATEMENT RELATED comnli teness and d. tail may use their own format- TO THE OPERATION OF NINE MILE POINT NUCLEAR STATION. UNIT NO 2 Docket No 50 410 (Niaga a Mnhawk NUREG-10/0: NRC POLICY ON FUTURE REACTOR Power Corporation.et al)
~~Dwison of Licensing (800428 031 GPS Decisons On 3evere Accident issues In Nuclear 851124) May 1985. 373pp 8505230685. 30548 001.~~
~~Power Plant Regulat on.~~
Office of Nelaar Reactor Regulatior This Final Environmental Statement contains the assessmem Director (pre-851125). July 1985. 147pp. 8508150036. of the environmental impact associated with the operation of 32197.342. the Nine Mile Point Nuclear Station. Unit 2. pursuant to the Na-On Apnf 13, 1983, tr a UE. Nuclear Regu atory Commnsion tional Environmental Pohcy Act of 1969 (NEPA) and Tit!o 10 of issued for pubhc comment a " Proposed Pohey Statement on the Code of Federal Pegulabons. Part St. as amended, of the Swere Accidents and Related Views on Nuclear Reactor Regu- Nuclear Regulatory Commission regulations This statement en-11 tion" (48 SR 16014) This report presents and discusses the amines the environment, environmental consequences and miti-Commission's final version of that pohcy statement now entitled, gating actions, and environmental and economic benefits and "Polcy Statement on Severe Reactor Accidents Regarding costs.
_ ._ _ . _ _ . __ ~_ __.m _ _ _ - . ._ 18 Main Citations and Abstracts cants and graphite based tubreants appears to result in a sig-NUREG-1087: FINAL ENVIRONMENTAL STATEMENT RELATED TO THE OPERATION OF VOGTLE ELECTRIC GENERATING nificantly increased incidence of leakage and corrosion. PLANT, UNITS 1 AND 2. Docket Nos. 50-424 And 50 425.(Geor-gia Power Company)
~~Dwison of Licensing (800428-851124). NUREG-1093: SAFETY EVALUATION REPORT RELATED TO I~~
March 1985. 461pp. 8504090235. 29748.035. THE RENEWAL OF THE OPERATING LICENSE FOR THE This Final Envronmental Statement contains an assessment TRIGA TRAINING AND RESEARCH REACTOR AT THE UNI . of the environmental impact associated with the operation of VERSITY OF UTAH. Docket No. 50-407. (University of Utah)
~~ths Vogtle Electnc Generating Plant, Units 1 and 2 pursuant to Dwison of Licensing (800428-851124). March 1985. 70pp.~~
the National Envronmental Policy Act of 1969 (NEPA) and Title 8504090013J 29754:045. 10 of the Code of Federal Regulations, Part 51 (10 CFR 51), as This Safety Evaluaton Report for the apphcation filed by the amended, of the Nuclear Regulatory Commission regulations. Unwersity of Utah (UU) for a renewal of Operabng License R-This statement examines the envronmental impacts, environ- 126 to continue to operate a training and research reactor facih- t I mental consequences and mitigating actions, and envronmental ty has been prepared by the Office of Nuclear Reactor Regula-economic benefits and costs assooated with station oper. tion of the U.S. Nuclear Regulatory Commission. The facility is owned and opera:ed by the University of Utah and is located on its campus in Salt Lake City. Salt Lake County, Utah. The staff NUMEG-1089: TECHNICAL SPECIFICA'lONS FOR FERMJ- concludes that this training reactor facility can continue to be
~~2. Docket No. 50-341. (Detrost Edsson Company)~~
Dwision of U-censing (800428-851124). March 1985. 501pp. 8504050281. operated by UU without endangenng the health and safety of 29672.001, the public.
The Fermi-2 Facihty Technical Specifications were prepared by the U.S. Nuclear Regulatory Commission to set forth the NUREC 1098: SAFETY EVALUATION REPORT RELATED TO limits, operating conditions, rnd other requrements apphcable THE RENEWAL OF OPERATING LICENSE FOR THE RE-to a nuclear reactor facdity as set forth in Secten 50.36 of 10 SEARCH REACTOR AT MANHATTAN COLLEGE. Docket No. CFR Part 50 for the protection of the health and safety of the 50-199. (Manhattan Col!ege)
~~Dwision of Licensing (800428-851124). February 1985. 58pp. 8503130281. 29360 064.~~
pubhc. This Safety Evaluation Report for the appicaten filed by Man-NUREG 1094: FINAL ENVIRONMENTAL STATEMENT RELATED hattan College (MC) for a renewal of Operating License R.94 to TO THE OPERATION OF BEAVER VALLEY POWER contmue to operate the MC 0.1 W open pool trairwng reactor STATION, UNIT 2. Docket No. 50-412. (Duquesne Light Compa, has been prepared by the Office of Nuclear Reactor Regulabon ny)
Dwison of Licensing (800428-851124). September 1985. of the U.S. Nuclear Regulatory Commission. The facihty is 300pp. 8509300559. 32792:017. owned and operated by MC and is located two blocks away The Final Envronmental Statement related to the operation from the MC main campus in the Ruerdale area of New York of Beaver Valley Power Station Unit 2 by Duquesne Light Com. City, New York. The staff concludes that the reactor facihty can pany, et af (Docket No. 50-412), located itt Beaver County, conhnue to be operated by MC without endangenng the health Pennsytvania, has been prepared by the Offee of Nuclear Re- and safety of the pubhc.
actor Regulation of the U.S. Nuclear Regulatory Commisson. This statement reports on the staff's review of the impact of op-eration of the plant. Also included are comments of state and NUREG-1100 V01: FY 1986 BUDGET ESTIMATES.
~~Duison of Budget & Analysis. January 1985. 87pp. 8502190023.~~
federal governments, local agencies and members of the pubhc 29013:211, on the Draft Environmental Staternent for this protect and staff This report contains the fiscal year budget justifcatons to responses to these comments. The NRC staff has concluded, based on weighing of environmental, technical and other fac- Congress. The budget estimates for salanes and expenses for fiscal year 1986-87 provide for obhgations of $429.000,000 to
~~. tors, that an operating Icense could be granted.~~
be Nnded n total by a new appropnaten. NUREG-1096: EVALUATION OF RESPONSES TO IE BULLETIN 8242 Degradation Of Threaded Fasteners in Reactor Coolaat NUREG-1103: CONTAMINATED MEXICAN STEELimportaton Of Pressure - Boundary Of Pressunzed Water-Reactor Plants. Steel into The United States That Had Been inadvertently Con-ANDERSON.W.; STERNER,P Drvmon of Emergency Prepared- taminated With Cobatt.60 As A Result Of Scrameng Of A Tele-ness & Engineenng Response (Post 830103). May 1985. 75pp, therapy Unit.
~~Safeguards & Matenals Program Branch. Janu-8506240221. IEB-82 02,31177:219. ary 1985. 84pp. 8502110623. 28905:023.~~
IE Bulletin 82-02 was issued by the NRC on June 2,1982 to Dus W docunets me crcumstances contnbuting to the notify Icensees about incidents of severe degradation of thread- inadetent meltng of Co-60 contaminated scrap metal in two ed fasteners. Rssponses to the Bulletin from 41 PWR Icensees Mexican steel foundnes and me subsequent 4stnbuton of con-included data from recent regular inspectons of reactor coolant faminated steel products into the United States. The report pressure boundary components connections of six inch size and covers the tracing of the source to its on0'n, response actons larger. Statistcal analysis rs used to determine signifcant fac- to recover radcactue steel in the Uruted States, and return of lors rsiated to frequency of leakage incidents in connectons, the contaminated matenals to Meuco. Informaton outsede of occurrence of degradation of bolts and studs, and the need for ttus scope is recounted as necessary, e.g, details of the inci-bolt replacement. Factors examined include the age of the dent on the Mexcan side of the border. The incident resulted in plant, types of components, use of tubrcants and searants, and very significant exposure to citizens of the United States, differences between plants. The compded data indicate that, on the average 10% of the bolted connections whch were in- NUREG-1104: TECHNICAL SPECIFICATIONS FOR WOLF spected show evidence of leaking and 80% of those undergo CREEK GENERATING STATION, UNIT 1. Docket No. 50-some degradation of the botting. A signifcant decrease in the 482.(Kansas Gas And Electric Company)
~~Dwision of Licensing occurrence of boating degradation events as the age of the (800428 851124). March 1985. 500pp. 8504030425. 29601:354.~~
plant increases is obseqed. Valves appear to be less subject to The Wott Creek Generahng Staton, Unit 1 Techncal Specife botting corrosen. A group of 5 of the 41 plants accounted for cabons were prepared by the U.S. Nuclear Regulatory Commis-about one-half of the reported leakage and corrosion events. son to set forth the hmits, operating conditions, and other re-The common charactenste found for 4 of these 5 plants was quirements apphcable to a nuclear reactor facdity as set forth in the lubncant used. The use of ncke6 graphite based lubrcants Sechon 50.36 of to CFR Part 50 for the protecton of the health appears to offer a sigrufcantly reduced incidence of leakage and safety of the pubic. End corroson; whde use of mofybdenum disulfide-based fubn-
I r l Main Citations and Abstracts 19 ' NUREG-1105: REVIEW AND EVALUATION OF THE NUCLEAR 50.36 of 10 CFR Part 50 for the protecten of the hearth and REGULATORY COMMISSION SAFETY RESEARCH PRO- safety of the public. GRAM FOR FISCAL YEARS 1986 AND 1987.
ACRS - Adviso-1 ry Committee on Reactor Safeguards. February 1985. 59pp. NUREG-1115: CATEGORIZATION OF REACTOR SAFETY j 8503010055. 29186:001. ISSUES FROM A RISK PERSPECTIVE.

Dmson of Risk Anal-1 Pubhc Law 95-209 includes a requirement that the Advisory ysis & Operatons (post . 840429). March 1985. 167pp.
Committee on Reactor Safeguards submit an annual report ta 8504030427. 29598.072. , Congress on the safety research program of the Nuclear Regu- This report presents the results of an effort to identify and latory Commisson. This report presents the results of the ACRS rank reactor safety and nsk issues identified from past Probabi-
; review and evaluation of the NRC safety research program for Fiscal Years 1986 and 1987. The report contains a number of hste Risk Assessments (PRAs) and other safety analyses. Be-comrnents and recommendations. cause of the vaned scope of these analyses, the lest of issues may be incomplete. Nevertheless, those studies compnsed or.
~~NUREG-1108: -TECHNICAL SPECIFICATIONS FOR CATAWBA dered analyses to whatever their respectue depths; hence, they~~
~~NUCLEAR STATION. UNIT 1. Docket No. 50-413.(Duke Power warranted scrutiny for whatever insights they could reveal with j Company)~~
Dvision of Licensing (800428-851124). January respect to issue importance. The top ranked issues in terms of 1985. 525pp. 8502060481. 28746:001, ther contnbution to the uncertainty in nsk are desenbed in 4
The Catawba Nuclear Staten, Unit 1 Technscal Specifcat ons some detail. All of these nsk issues are compared to the "ge-were prepared by the U.S. Nuclear Regulatory Commisson to nene safety issues" for completeness and omession. i set forth tne limits, operat.ng conditons, and other requirements J appleable to a nuclear reactor facshty as set forth in Secten NUREG 1116: A REVIEW OF THE CURRENT UNDERSTANDING
! 50.36 of to CFR Part 50 for the protecten of the health and OF THE POTENTIAL FOR CONTAINMENT FAILURE FROM safety of the pubic. IN-VESSEL STEAM EXPLOSIONS.
~~Steam Explosen Review NUREG-1100: RADIOACTIVITY TRANSPORT FOLLOWING Group. June 1985. 521PP. 8507030716. 31335:001.~~
STEAM GENERATOR TUDE RUPTURE. HOPENFELD,J. Divi. A group of experts was convened to review the current un-sion of Accident Evaluaton. March 1985. 45pp. 8504040001. derstanding of the potential for containment failure from in. 29630:269. vessel steam explosons dunng core meltdown accidents in t A review of the capabihties of the CITADEL computer code LWRs. The Steam Explosen Review Group (SERG) was re-as well as plant expenence to project radioactmty releases fol- quested to provide assesstnents of: (i) the conditional probab h-i lowing steam generator tube rupture in PWR's shows that cer- ty of containment failure due to a steam explosen, (ii) a Sandia l j tain expenmental data is needed for rehable offsite dose predic- National Laboratory (SNL) report entitled "An Uncertainty Study tons. This artcle defines five parameters whch are the key for of PWR Steam Explosons," NUREG/CR 3369, (iii) a SNL pro-such predictions and discusses the functional dependence of posed steam explosion research program. This report summa-these parameters on varcus operabonal vanables. A joent Wes- nzes the results of the dehberations of the review group. It also i tinghouse. Electric Power Research Institute, and the Nuclear presents the detailed response of each indmdual member to i Regulatory Commission program a.med at obtaining the five pa* i rameters empincally is desenbed. Present status of the CITA* each of the issues. The consensus of the SERG is that the oc-DEL code as also reviewed. currence of a steam explosen of suffeient energetics which { could lead to alpha-mode containment failure has a low proba-NUREG 1110: COMPARISON OF LICENSING ACTIVITIES FOR blity. The SERG members disagreed with the methodology i OPERATING PLANTS DESIGNED BY BABCOCK & WILCOX. used in NUREG/CR 3369 for the purpose of estabhshing the
! THOMA.J.O. Division of Licensing (800428-851124)c January uncerta.nty in the probabhty of containment failure by a steam j 1985. 29pp. 8502070590. 28807:175. exploseon. A consensus was reached among SERG members i This report provides a companson of a number of licensing on the need for a continuing steam explosion research program actmties for the operating Babcock & Wilenx (B&W) plants with which would improve our understanding of certain aspects of
~~emphasis on Rancho Seco. The factors selected were a com- steam explosion phenomenology.~~
panson of staff resources expended in FY84, actve hcensing acton reviews, implementsten of NUREG-0737 modifcations. NUREG 1117: TECHNICAL SPECIFICATIONS FOR WATERFORD exemptions to regulations, SALP reports, enforcement actons, STEAM ELECTRIC STATION UNIT 3 Docket No. 50-382.(Lou. 4 and Licensee Event Reports (LERs). The eight Icensed operat- esiana Power And Light Company)
~~Offee of Nuclear Reactor~~
) ing plants examined are as follows: Arkansas Nuclear One Unit Regulabon, Drector (pre-851125). March 1985. 504pp. 4 1 (ANO 1). Crystal River Unit 3. Davis Besse, Oconee Units 1 8504030439.29600 178- l j 2, and 3, Ran-ho Seco, and Three Mile Island Unit 1 (TMI 1)- The Waterford. Unit 3 Technical Specifcahons were prepared 4 NUREG-1112: EfNIRONMENTAL ASSESSMENT FOR RENEW. by the U.S. Nuclear Regulatory Commissen to set forth the 1 AL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM. hmits, operating condibons, and other requirements apphcable 368.(UNC Naval Products Dvisco Of UNC Resources.Inc)
De- to a nuclear reactor facility as set forth sn Secbon 50.38 of 10 vision of Fuel Cycle a Matenal Safety. January 1985. 74pp. CFR Part 50 for the prctecten of the health and safety of the 8502120056. 28671:266. pubic.
i This Environmental Assessment is issued by the U.S. Nuclear Regulatory Commission (NRC) in response to an apphcation by NUREG-1118: ENV:RONMENTAL ASSESSMENT FOR RENEW. UNC Naval Products, Dvison of UNC Resources, Inc., for the l AL OF SPECIAL NUCLEAR MATERIAL LICENSE NO. SNM. >
~~renewal of Special Nuclear Material (SNM) License No. SNM- 1107. Docket No 70-1151. (Westinghouse Electnc Corporaten) 368 for the operaton of the existing fuet fabncaten facility.~~
~~Dvision of Fuel Cycle & Material Safety. May 1985.140pp.~~
i NUREG-1113: TECHNCAL SPECIFICATIONS FOR BYRON STA. 8505240050. 30566.199. TION UNITS 1 AND 2. Docket Nos. 50-454 And 50-455. (Com. This Environmental Assessment is issued by the U S. Nuclear momwealth Edison Company)
Dmson of Licensing (800428 Regulatory Commession (NRC) in response to an appication by 851124). February 1985. 510pp. 8503110132. 29326.001. the Westinghouse Electre Corporaten for the renewal of Spe-
! The Byron Staten Unit 1 and Urvt 2 Technical Specifcations cial Nuclear Matenal License No. SNM 1101 which covers the I were prepared by the U.S. Nuclear Regulatory Commission to operabons of the Columba plant. I set forth the hmits, operating condatens, and other requirements appicable to a nuclear reactor facility as set forth in Section NUREG-1119: SAFETY EVALUATION REPORT RELATED TO { i l l
20 Main Citations and Abstracts THE RENEWAL OF THE OPERATING LICENSE FOR THE NUREG-1125 V05: A COMPILATION OF REPORTS OF THE AD-CAVALIER TRAINING REACTOR AT THE UNIVERSITY OF VISORY COMMITTEE ON REACTOR SAFEGUARDS 1957-VIRGINIA. Docket No. 50-396 (University Of Virginia)
Divison 1984. volume S.Part 2:ACRS Reports On Genenc Subjects of Licensing (800428-851124). May 1985. 62pp. 8506060716. (HTGR Regulatory Guidest

ACRS - Advisory Committee on ,
30780 214. Reactor Safeguards. Apnl 1985. 630pp 8504220396. l This Safety Evaluation Report for the application filed by the 29958.102. University of Virginia for a renewal of oparating hcense number See NUREG 1125,V01 abstract. R 123 to continue to operate a training and research reactor j (CAVALIER) has been prepared by the Office of Nuclear Reac- NUREG 1125 V06: A COMPILATION OF REFORTS OF THE AD- i tor Regulation of the U.S. Nuclear Regulatory Commission. The VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957-facihty is owned and operated by the University of Virginia and 1984. Volume 6,Part 2:ACRS Reports On Genenc Subjects (RPA is located on the campus in Charlottesville, Virginia. Based on - Appendix C).
ACRS - Advisory Committee on Resctor Safe-its technical review, the staff concludes that the reactor facihty guards. Apol 1985. 567pp. 8504220402. 29960 015 can continue to be operated by the university without endanger. See NUREG-1125,V01 abstract.
ing the health and safety of the public or endangenng the envi-ronment. NUREG 1126: TECHNICAL SPECIFICATIONS FOR SHOREHAM NUREG-1122: KNOWLEDGES AND ABILITIES CATALOG FOR NUCLEAR POWER STATION. UNIT NO.1. Docket No. 50-NUCLEAR POWER PLANT OPERATORS Pressunzed Water 322.(Long ! stand Lighting Company)
Divison of Licensing Reactors.

Divison of Human Factors Safety (800428-851124) (800428-851124). July 1985. 489pp. 8507250207. 31784 001, July 1985. 400pp. 8508090488. 32120:352. The Shoreham, Unit 1, Test.hnical Specifications were pre.
This document catalogs roughly 5300 knowledges and abili- pared by the U S. Nuclear Regulatory Commission to set forth ties of reactor operators and senior reactor operators. It results the limits, operating conditons, and other requirements applica-from a reanylsis of a much larger job-tasis analysis data base ble to a nuclear reactor facihty as set forth in Secten 50.36 of compiled by the Institute of Nuclear Power Operatons (INPO). 10 CFR Part 50 for the protection of the health and safety of Xnowledges and abihties are cataloged for 45 mator power the pubhc. plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functons (e g., reacuvity control and NUREG-1127: RADIATION PROTECTION TRAINING AT URANt-reactor coolant system inventory control). With appropnate sam- UM HEXAFLUORIDE AND FUEL FABRICATION PLANTS. phrig from this catalog, operator licensing examinations having BRODSKY,A : SOONG,A L.; BELL.J. Division of Radiaton Pro-content vahdity can be developed. A structural sampling proce- grams & Earth Sciences (post 840429). May 1985. 33pp. dure for this catalog is under development by the Nuclear Regu- 8506190048. 31017:173. latory Commission (NR(') and will be pubhshed as a companion This report provides ganeral information and references document. " Examiners Handbook for Developing Operator La- useful for estabhshing or operating radiaton safety training pro-consing Examinations (NUREG 1121). The examinations devel-oped by using the catalog and handbook will cover those topics 9 um compounds that are used in the manufacture of nuclear hsted under Title 10, Code of Federal Regulations, Part 55. fuels in addition to a bnef summary of the pnnciples of effec-NUREG-1125 vot: A COMPILATION OF REPORTS OF THE AD- tive management of radiation safety training. the report also VISORY COMMITTEE ON REACTOR SAFEGUARDS.1957- contains an appendix that provides a comprehensive checkhst 1084. Volume 1 Part 1:ACRS Reports On Project Reviews (A F) of scientific, safety. and management topics, from which appro.
ACRS - Advisory Committee on Reactor Safeguards. Apnl pnate topics may be seiected in prepanng training outhnes for 1985. 658pp. 8504220393. 29956:167. various job categones or tasks pertaining to the uranium nuclear This six volume compilation contains over 1000 reports pre- fuels industry The report is designed for use by radiation safety pared by the Advisory Committee on Reactor Safeguards from training professionals who have the expenence to utihre the September 1957 through December 15 A The reports are di- re ort to not only select the appropnate topics, but also to tailor vided into two groups: Part 1: ACRS Reports on Project Re- the specific detaits and depth of coverage of each training ses-mews, and Part 2: ACRS Reports on Genenc Subjects. Part 1 sion to match both employee and management needs of a par.
contains ACRS reports a!phabetized by project name and within ticular ,ndustnal i operation, project name by chronological order. Part 2 categonzes the re. ports by the most appropnate genenc subject area and within NUREG-1128: TRIAL EVALUATIONS IN COMPARISON WITH subject area by chronological order. THE 1983 SAFETY GOALS RIGGS R.; SEGE,G. Divison of NUREG-1125 V02: A COMPILATION OF REPORTS OF THE AD* Safety Technology - (800428-851124). June 1985. 200pp. VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957- 8507080209 31402.041. 1984, Volume 2,Part 1;ACRS Reports On Project Reviews (G-P). This report provides retrospective compansons of selectec*
ACRS Advisory Committee on Reactor Safeguards. Apnl genenc regulatory actions to the 1983 NRC safety Coafs, which 1985. 720pp. 85042203/4. 29944.001. 5ad been issued for evaluaton dunng a two-year penod. The See NUREG-1125,V01 abstract. issues covered are those anatyzed by the Office of Nuclear Re-NUREG-1125 V03: A COMPILATICN OF REPORTS OF THE AD- acor Regulation (NRR) (sssisted in some cases by the Battelle VISORY COMMITTEE ON REACTOR SAFEGUARDS.1957- Pacific Northwest Labororyt The issues include auxikary feed-1984. Volume 3,Part 1:ACRS Reports On Project Revews (0 Z). water rehabihty, pr%unzed thermal shock, power operated

ACRS - Advisory Committee on Reactor Safeguards. Apnl relief valve isolation, asymmetric blowdown loads on PWR pn-1985. 563pp. 8504220389. 29954 329. mary systems, pool dynamic loads for BWR containments, and See NUREG-1125,V01 abstract. steam generator tube rupture. Calculated core-melt frequencies.
mortahty nsks, and cost benefit rabos are compared with the NUREG 1125 V04: A COMPILATION OF REPORTS OF THE AD-corresponding safety-goal quantitatrve design objectives. Con-VISORY COMMITTEE ON REACTOR SAFEGUARDS 1957-siderations that should influence interpretation of the compan- ! 1984. Volume 4,Part 2:ACRS Reports On Genenc Subects l (Ac. I cadent Analysis - Genenc items).
ACRS - Advisory Committee sons are discussed. Comments are included on whether and i on Reactor Safeguards. Apnl 1985. 627pp. 8504220406. how the safety goals may help in the regulatory decision proc-29961 225. ess and on problems encountered.
l j See NUREG-1125,V01 abstract. l i 1
Main Citations and Abstracts 21 NUREG 1130: ENVIRONMENTAL ASSESSVENT FOR RENEW. cense to construct and operate a TRIGA research reactor has AL AND CONSOUDATION OF MATERIALS LICENSE NOS been prepared by the Office of Nuclear Reactor Regulation of SNM-362.SMS-405,08-00566-05, 08 0056610.AND 08 00C66- the U S. Nudear Regulatory Commiss.on The facdtv is owned
~~12.~~
~~Dmsion of Fuel Cycle & Matenal Safety, March 1985 and operated by the University of Temas and is located at the 45pp. 8504080548. 297'3.038.~~
Univensty's Ba! cones Research Center, about 7 mdes (116 kdo-This Environmental Assessment is issued by the U.S. Nuclear meters) north of the ma;n campus in Austin. Teras. The staff Reguiatory Commission (NRC) in response to an apphcatron by concludes that the TRIGA reactor facihty can to constructed the U S. Department of Commerce, National Bureau of Stand-ards, for the renewal and conschdat:en of five Matenats Ls and operated by the Univers.ty of Teras witnout endangenng the hea'th and safety of the pubhc. nses for radiological activities at the National Bureau of NUREG 1136: TECHNICAL SPECIFICATIONS FOR WOLF CREEK GENERATING STA TION, UNIT 1 DoJet No. 50-NUREG-1131: FINANCIAL ANALYSIS OF POTENTIAL RETRO- 482 (Kansas Gas And E'ectnc Company)
~~Division of Licensing SPECTIVE PREMtVM ASSESSMENTS UNDER THE PRICE- (800428-851124) June 1985 198pp. 8506270251 31257.142.~~
ANDERSON SYSTEM. WOOD.R S. Cffice of State Programs. The Wolf Creek Generating Stat:on. Un t No 1 Techn. cal Director. Apnt 1985.17pp 8505080348. 33218:171. Ten representatrve nuclear utattes have been ana'yred over Specificatrons were prepared by the U S NuC' ear Regu!atory the pened 1981 1983 to eva'uate the effects of three lesels of Commission to ,et forth the hrmts, operat ng cond.t ons, and retrospective premiums on vanous financel in@cators This other requirements apphCab!e to a nuclear reactor facarp as set analysis continues and espands on earber ana>yses prepared as forth in Section 50 36 of 10 CFR Part 50 for the protect,on of the hea!th and sa'ety of the pubhc. background for dehberations by the U S. Congress for possible extension or rnod.fication of the Pnce Anderson Act NUREG 1137: SAFETY EVALUATION AEPCRT RELATED TO NUREG-1132: TECHNICAL SPECIFICATIONS FOR CfABLO THE CPERATION CF VOGTLE ELECTRrC GENERATING CANYON NUCLEAR POWER PLANT. UNIT NO 2.Dochet No PLANT.UN!TS 1 AND 2 Docnet Nos 50-424 And 50-425 (Geor-50 323 (Pacifrc Gas and Electr.c CompanyJ
Dmsson of Licens- gia Power Company et a!)

Dvston of Licens+ng (600428-ing (800428 851124). Apnl 1985. 466pp 8505280011 851124) Ju ne 1985. 747pp 8507030707. 31314 265 30605 007. The Sa'ety EvaNation Report for the appucat>on faed by The D+ab!o 2 Technical Specificawns were prepared by the Georg'a Po*er Company, Munopal Etectric Authorty of Geor-U S. Nuclear Regulatory Cor mission to set forth the I.mits, cp- g a. Og'ethorpe Po*er Corporat.on. and Cty of Da; ton. Georg a.
erating cond tions. and other requirements apphcable to a nu- as acphcants and owners, for hcenses to cperate the Vogt'e clear reactor facibty as set forth in Sect;on 50 36 of to CFR E'ectric Gererating Pfant. Unrts 1 and 2 (Docket Nes 50 424 Part 50 for tne protection of the health and safety of the pubhe- and 50-425), has been prepared by the'O%ce of Nuciear Reac-tor Regulat.on cf the U S. Nuc' ear Regu'atory Commission The NUREG 1133: TECHNCAL SPECIFICATIONS FOR PALO VERDE faobty is loca +1.n Burke County, Georg.a approomatefy 415 NUCLEAR GENERATING STATfCN, UNIT 1 Docket No 50 528 km (26 mi) souto southeast of Augusta. and on the Sa.annah (Anzora Pubbe Service Company)
Division of Lonseg R.ver Subject to favorade resoLtion of the items d scussed in (800428 851124). May 1985. 515pp. 8506240646. 3115t.001. this report. t*e staff concludes that the apphcant can operate The Palo verde Unit 1 Technical Spec:f: cations were pre. the facil.ty *4hout endangering tne hea.th and sa'ety cf the pared by the U S. Nuclear Regulatory Comm;ssvon to set forth pubhc.
the hmits operating cond.t.ons, and other requ.rements apphca-ble to a nuclear reactor fac+ty as set forth in Sect 1on 50 36 of NUREG 1137 Sot: SAFETY EVALUATION REPORT RELATED. 10 CFR Part 50 for the protection of the hearth and sa'e'y of TO THE OPERATION OF VOGTLE ELECTRIC GENERATING the public PLANT. UNITS 1 AND 2 DOCMET Nos 50 424 And 50-425 (Georg.a Power Company et a')
Dvs on of Licenvng NUREG 1134: RADIATION PACTECTION TRAIN:NG FOR PER. (800428 851124) Octoter 1985. 56pp 8511210589 SONNEL EMPLOYED IN MEDICAL FACILITIES 33586 265 MCELRDY.N L.; BRODSKY.A. Dms on of Rad.ation Programs & in Ju re 1985. the sta'f of the Nuclear Regulatory Comm ss:en Earth Sciences (post 840429) May 1985 61pp 8506130363 issued .ts Safety Eva6ation Reoort (NUREG 1137) regard ng 30868.116. the appbcatron of Georg a Power Company. Yu nictpa! Eiectnc This report provides information use'ul for planning and con. Author:ty of Georg a. Og'ethorpe Power Corporation. and City of ducting rad.aSon safety tra'n:ng in medical facAt es to keep en. Datton, Georg a. for a hcense to operate the Vogte Electnc posures as low as reasonably ach,evab:e, and to meet other Generating Plant. Units 1 and 2 (Dochet Nos 50 424 and 50-regulatory, safety and loss prevention requirements in today's 425) The faobty is located in Burke CounY Georg a. approns hospita's. A bnef d.scussion of the elements s9d basic consid. mately 26 mdes south-southeast of Augusta. Georg'a ard on erasons of rad at on safety tra n.ng programs is foHo*ed by a the Savannah R,ver. This f.rst suppleraent to NUREG 1137 pro-short D bhograpny of sefected references and sampie lecture (or V' des recent information regarding resolut.on of some of the session) outnnes for vanous job categones This rfor at on rs cpen sad confirmatcry items that re a.ned unresLed at the intended for use by a professional who is thoroughty acquainted tee the Sa'ety EvaNat t- Report was issued and prov. des the wth the science and pract.ce of radtation protection as well as Adwsory Comm.ttee on Reacter Sa'eguards letter dated August the specif,c procedures and circumstances of the part:cu!ar hos- 13.1985.
pitat's operations Top cs can be added or subtracted, amphf ed or condensed as appropnate This document does not set forth NUREG 1138: SAFETY EVALUATION REPORT RELATED TO specific tra.ning program requirements for any particular hospital THE RENEWAL OF THE OPERATING LICENSE FOR THE of type of medicalinstitution or group of employees TRAIN,NG AND RESEARCH REACTOR AT THE U%ERSITY OF MICHIGAN Docket No 50-2 (University of M chegan)
~~O'vi.~~
NUREG 1135: SAFETY EVALUATION REPORT RELATED TO sion of Licensing (800428-851124) July 1985 73pp THE CONSTRUCTION PERMIT AND OPERATsNG LICENSE 85080t0299 31928 095 FOR THE RESEARCH REACTOR AT THE UNIVERSITY OF This Safety Evaluation Report for the apphcat.on fded by the TEXAS Docket No. 50-602. (Universsty of Texas)
Devision of University of M.chigan (UM) for a renewal of the Ford Nuclear Licensing (800428-851124). May 1985 88pp 8506240665 Reactor operating hcense R-28 to cont.nue to cperate a tra.ning 31152 223 and researen reactor facihty has teen prepared by the Omce of This Safe 4 Evaluation Report for the applicatton fded by the Nuclear Reactor Regu'ation of the U S Nuclear Regufatory University of Tenas for a construction permit and operating h. Commiss.on The facilty >$ owned and operated by the Universo
22 Main Citations and Abstracts The River Bend Staton Technical Specfications were pree ty of Michigan and is located at the North Campus of the Uni-versity in Ann Arbor, Michigan. The staff concludes that this pared by the U S. Nuc! ear Regulatory Commisson to set forth training reactor facihty can continue to be operated by UM with- the limits, operating cond.tions and other requirements apphca-ble to a nuclear reactor facihty as set forth in Secton 50.36 of out endangenng the health and safety of the public. 10 CFR Part 50 for the protection of the health and safety of NUREG 1139: SAFETY EVALUATION REPORT RELATED TO the public. THE RENEWAL OF THE OPERATING LICENSE FOR THE TRAIN!NG & RESEARCH REACTOR AT THE UNIVERSITY OF NUREG 1143: SAFETY EVALUATION REPORT PELATED TO LOWELL. Docket No. 50-223.(Unwersity of Lowell)
Division of THE FULL TERM OPERATING UCENSE FOR MILLSTONE M L N ber 1985 84pp NUCLEAR POWER STATION UNIT NO.' 1. Docket No. 50-5120 668. 33734.089 26 peau Mear hergy @pand

Nsm d Ws-This Safety Eva!uatork Report for the apphcation filed by the 9 University of Lowell for renewal of operatng license number R.
125 to confine to operate the open pool type tra,ning and re. [338f'0 The Saf 2'ety Evaluaton Report for the full-term ope Search reactor has been preparea by the Office of Nuclear Re- conse appbcaten hied by Northeast Nuclear Energy Company actor Regulation of the U.S. Nuclear Reactor Regulation of the for Millstone Nuclear Power Station, Unit No.1 has been pre-U.S. Nuclear Regulatory Commission. The facihty is owned and pared by the Office of Nuclear Reactor Regulaton of the U S. operated by the Unwersity of Lowell and is located on the uni- Nuc! ear Regulatory Commission. the facihty is located in New versey campus in Lowell, Massachussetts. The staff concludes London County, Waterford, Connecticut. The sta'f concludes that the open pool type reactor facihty can continue to be oper. that the facihty can continue to be operated without endanger-ated by the Unwersity of Lowell without endangenng the heatth ing the health and safety of the pubhc. and safety of the pi.bbc. NUREG-1140 DRFT FC: A REGULATORY ANALYSIS ON EMER- NUREG 1144: NUCLEAR PLANT AGING RESEARCH (NPAA) PROGRAM PLAN. MORRIS.B.M; VORA.J P. Divis on of Eng-GENCY PREPAAEDNESS FOR FUEL CYCLE AND OTHER RADIOACTIVE MATERIAL LICENSEES Draft Report For Com- neenng Technology. July 1985. 48pp. 8508210443. 32337.277, ment. MCGUIRE.S A. Division of R sk Ana'vs.s & Operatons The nuclear plant aging research desenbed in this plan is in-(post 840429). June 1985.125pp. 8507020410. 31309.033. tended to resolve issues related to the aging and service wear Potental accidents for 15 types of fuel cycle and other rad,o- of equipment and systems at commercial reactor facihties and actve matenal licensees were analyzed The most potentially - their poss.ble impact on piant safety. Errphasis has been hazardous accident, by a large margin, was determined to be placed on identfication and charactenzation of the mechanisms the sudden rupture of a heated multi ton cyhnder of UF6. Acute of matenal and component degradation dunng sevice and eval-fatahties offsite are probabiy not cred.ble. Acute permarient inju- unton of methods of inspection, surveillance, cond, tion monitor-ries may be possible for many hundreds of meters, and chnically ing and maintenance as means of mitigatng such effects. Spe-observable transient effects of unknown long term conse- cifically, the goals of the program are as follows: (1) To identify Quences may be possible for distances up to a few miles. and charactenze aging and service wear effects which, f un-These e*fects would t e caused by the chemical toxicity of the checked, could cause degradaton of structures, components, UF6. Radiaton doses would not be signif. cant. The most poten- and systems and thereby impair plant safety, (2) To identfy tally hazardous accident due to radiaton expcsure was deter- methods of inspection, surveillance and monitonng or of evalu-mined to DS a large fire at certain facihties handhng large quan- ating residual hfe of structures, components. and systems, tites of atpha-emettng radionuclides (i.e., Po-210 Pu-238, Pu- which will assure t mely detecton of significant aging effects 239, Am-241, Cm 242, Cm-244) or radioodines (1125 and l- pnor to loss of safety functen, and (3) To evaluate the effec-131) However, acute fatal,tes or injunes to people offsite due tiveness of storage, ma.ntenance, repair and replacement prac-to accidental releases cf these matenals do not seem plausible. tices in mitgating the f ate and extent of degradaten caused by The only other significant accident was identfied as a long-term agtng and service wear. pulsatng enticabty at fuel cycle facilites handhng high ennched uranium or plutonium. An important feature of the most senous NUREG 1145 V01: U S. NUCLEAR REGULATORY COMMISSION accidents is that releases are likely to start without pnor warn- 1984 ANNUAL REPORT,
~~C+fice of Resource Management, Di-rector. June 1985. 234pp. 8506260386. 31246 057, sog The releases would usually end within about half an hour.~~
Thus protectcn actons would have to be taken quickly to be This report covers the maior activities, events, decisons and effectve. There is not hkety to be enough time for dose projec- planning that took place durtng fiscal year 1984 within the NRC tons, comphcated decisonmaking dur ng the accident, or the or involving the NRC. partcipation of personnel not in the immedrate v.cinity of the ste. The appropnate response by the facilty :s to immediately NUREG 1147: SEISMIC SAFETY RESEARCH PROGRAM PLAN.
~~Division of Engineenng Technology. June 1985. 195pp.~~
notty local fire, pohce, and other emergency personnel and give them a b set predetermined message recommending protectve 8507080215. 31392:150. This plan describes the safety issues, regulatory needs, and actons. Emergency personnel are generally well qual:fied to re-the research necessary to address these needs. The plan afso spond effectvely to small acccents of these types. discusses the relationship between current and proposed re-NUREG-1141: TECHNICAL SPECIFICATIONS FOR FERMI-2 search within the NRC and research sponsored by other gov-FACluTY. Docket No. 50-341 (Detroit Edison Company)
~~Divi- ernment agences, wiversites, industry groups, professonal so-son of Licensing (800428-851124). Juty 1985. 490pp. ceties, aN foregn soven 8508070371,32058.001.~~
The Fermi-2 facility Technical Specificabons were prepared NUREG 1148: NUCLEAR POWER PLANT FIRE PROTECTION by the U.S. Nuclear Regulatsy Commiss on to set forth the RESEARCH PROGRAM. DATTA.A. Division of Engineenng hmits, operating conditons, and other requirements appbcable Tecnnology July 1985. 32pp. 8508080059. 32072 263. to a nuclear reactor facility as set forth in Secten 50 36 of to A program plan for nuclear power plant fire pro:ecton re-CFR Part 50 for the protection of the health and safety of the search has been presented in this report. The pnncipal ob l ec-public. tive of the program is to create a data base that would reduce the uncertaintes in fire probabilistic nsk assessment of plants. A NUREG-1142: TECHNICAL SPECIFICATIONS FOR RIVER BEND three pronged approacn of charactenzaten of potential fires, STATION. Docket No. 50-458. (Gulf States Utiktes Company) determinaten of the ensuing environment, and determination of BENEDICT,R. Divesen of Licensing (800428 851124). August
~~' . failure thresholds of safety-related equipment in that environ-30,1985. 541pp. 8509180507. 32663 337.~~
i
Main Citatloas and Abstracts 23 ment is doscribed. The techniques are to be applied to estimat. clear Regulatory Research. It is one of four plans desenbmg the ing the fire safety margin available in a control room. ongoing research in the corresponding areas of MEBR actmty, NUREG-1149: TECHN: CAL SPECIFICATIONS FOR LIMERICK which are being published simu!!aneously in four volumes as GENERATING STATION, UNIT 1. Docket No. 50 352. (Dhiladel- fo4ws: WI.1 Reactor Vesseis. Vol. 2 Steam Generators, Vol. 3 phia Electnc Company) MARTIN.R E. Offee of Nuclear Reactor Piping and Vol 4 Non-Destructwe Examination. These plans Regulation, Director (pre-851125). June 1985. 500pp have been updated and are more detailed expansions of those 8508270346. 32380.272. 0"9'nally pubbshed as part of the Lorq Range Research Plan The Limenck Generateg Station, Unit No.1 Technical Speci- for the Off ce of Nuclear Regulatory Research in NUREG 1080 fications were prepared by the U.S. Nuclear Regu!atory Com- VOI' I' mission to set forth the hmits, operating Cond:trons and other re. quirements applicable to a nuclear reactor facity as set forth in NUREG-1155 V02: RESEARCH PROGRAM PLAN Steam Genera-Section 50.36 of 10 CFR Part 50 for the protection of the health tors. MUSCARA.J , SERPAN.C Z. Drvision of Engineenng Tech-nology. Jufy 1985.18pp. 8508150425. 32220 093. and safety of the public. This report desenbes the NRC's research program related to NUREG-1151: TECHNICAL SPECIFICATIONS FOR DIABLO steam generators. Mainly it discusses the program for evalua-CANYON NUCLEAR POWER PLANT UNITS 1 AND 2 Docket tion of a removed-from-service degraded steam generator. Also Nos. 50-275 And 50-323 (Pacific Gas And E!ectnc Company)
discussed are projects to evaluate the vibration and wear that Drvision of Licensing (800428-851124). August 1985. 465pp. could result from chemical cleaning and NDE tasks for inservice 850910052t 32535.001. inspection of steam generators.
The Diabfo Canyon 1 and 2 Technical Specifications were prepared by the U.S. Nuclear Regulatory Commission to set NUREG-1155 V03: RESEARCH PAOGRAM PLAN PiPin9 forth the limits. operatrng conditions, and other requirements ap- VAGINS.M.; STROSN! DER.J. Dmsion of Engineonng Technolo-phcable to a nuclear reactor facihty as set forth in Sect on 50 36 gy. July 198516pp 8508160080 32230 056 of 10 CFR Pad 50 for the protection of the hea!!h and safety of This document presents a plan for research in Piping to be the public. performed by the Materiais Engineenng Branch. MEBR, Division of Engineenng Technoiogy, (DET), Off,ce of Nuclear Regulatory NUREG-1153: INSPECTION REPORT OF UNAUTHCRf2ED POS- Research. !! rs one of four plans describing the ongoing re-SESSION AND USE OF UNSEALED AMERICIUM-241 AND search in the corresponding areas cf MEBR actrvity, which are SUBSEQUENT CONFISCATION. J.C. Haynes beirg pubhshed simultaneously in four volumes as follows Vol. Company. Newark,Ohic. CANIANO,R J Reg >on 3, Office of Di- 1 Reactor Vessels. Vol 2 Steam Generators, Vol. 3 Piping. and rector. November 1985.100pp. 8512190248. 34n10:162. Vol. 4 Non-Destructive Examinat.on These plans have been up. This U.S Nuclear Regulatory Commission report documents dated and are more detarted expansions of those onginally pub-the circumstances surrounding the March 26,1985, confiscation lished as part of the Long Research Plan for the Office of Nu-and subsequent decontamination actwibes related to the use of clear Regu!atory Research in NUREG 1080 Vol.1. unauthonzed quantrties of amencium-241 at the John C. Haynes Company (licensee) of Newark Ohio. It focuses on the penod NUREG-1155 V04: RESEARCH PROGRAM PLAN Non-Destrue. from earfy Februaqr to July 26,1985. The incident started when twe Examination. MUSCARA.J. Dvsson of Engineenr,g Technol-NRC Region 111 recewed informatron that John C. Haynes pos- ogy. Jufy 1985. 27pp. 8508160074. 32230 073. sessed unauthorzed quantt4es of amencium-241 and was con. This report desenbes the NRC research program in non-de-ductng unauthonzed activities (diamond arradiation) By July 26, structrve evaluation. Projects are desenbed for the development 1985, the decontaminaton actmees at the licensee's laboratory and evaluation of techniques for penod.c inservice inspection of were concluded The licensee's actions with diamond irrad+ation reactor components and for the continuous onhne monitonng of resulted in contaminaton in restncted and unrestncted areas of reactors The areas of study described are ultrasonic, eddy cur. the facility. The confiscation and decontaminaton actmties re, rent testing and acoustic emiss+on. quired the corrbined efforts of NRC, Federal Bureau of Investi-gation. U S. Department of Energy, Oak Ridge Associated Uni- NUREG-1157: ENVIRONMENTAL ASSESSMENT FOR RENEW. versities, the State of Ohio. and the U S. Environmental Protec. AL OF SOURCE MATERIAL LICENSE NO. SUB-1010 Docket tion Agency. The report desenbes the f7ctual information and No. 40-8027. (Sequoyah Fuels Corporation)
D esion of Fuel significant findings associated with the confiscaton and decon. Cycle & Matenal Safety. August 1985. 406pp. 8509060254 tam. nation actmties 32504 010'nse to an application for renewal of Source Maten in respo NUREG-1154: LOSS OF MAIN AND AUXILIARY FEEDWATER LZense SUB 1010 for the Sequoyah Fuels Corporation facihty.
~~EVENT AT THE DAVIS-BESSE PLANT ON JUNE 9,1985.~~
the NRC staff prepared this Environmental Assessment. The Office of the Executwe Director for Operations. Jufy 1985. Environmental Assessment includes discussions of the need for 103pp. 8508150428. 32220.150. the proposed renewal action, aNrnatives to the action, and the On June 9,1985, Toledo Edison Company's Davis Besse Nu- ~ environmental impacts of the proposed action and alternatwes clear Power Plant, located in Ottawa County, Ohio, expenenced a part al loss of feedwater while the plant was operat ng at 90*. NUREG-1161: TECHNICAL SPECIFICATIONS FCR MILLSTONE power. Following a reactor inp, a loss of all feedwater occurred. NUCLEAR POWER STATION UNIT 3 Docket No. 50-The event involved a number of equipment rnalfuntions and ex. 423 (Northeast Nuclear Energy Company)

Dmsion of Pressur.
tenswe operator actions, including operator actions outside the tred Water Reactor Liceneng - A (post 851125). NSvember control room. Several operator errors also occurred dunng the 1985. 488pp. 8512180379 33954 015. event. This report documents tr,e findings of an NRC Team sent The Millstone Nuclear Power Stabon. Unit No 3 Technical to Davis-Besse by the NRC Executwe Director for Operations in Specifications were prepared by the U.S Nuclear Regulatory conformance with the staff proposed incident Investigation Pro. Commission to set forth the limits, operating conditions, and gram. other requirements apphcable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of NUREG-1155 V01: RESEARCH PROGRAM PLAN Reactor Ves- the health and safety of the public. sels. VAGINS.M. Dmsion of Engineerng Technology Jufy 1985. 41pp. 8508150435. 32220:110. NUREG 1164: INFORMATION ON THE CONFINEMENT CAPA. This document presents a plan for research in Reactor Ves- BILITY OF THE FACILITY DISPOSAL AREA AT WEST sels to be performed by the Materials Engineenng Branch, VALLEY,NEW YORK.
~~Office of Nuclear Regulatory Research, MEBR, Dmsson of Engineenng Technology, (DET), Office of Nu- Director. December 1985 62pp 8601070531. 34155 001.~~
24 Main Citations and Abstracts This report summanzes the previous NRC research studies. The Palo Verde Unit 2 Technical Specifications were pre-NRC licensee source term data and recent DOE site investiga- pared by the U.S Nuclear Regulatory Commission to set forth tions tht deal with assessment of the radioctive waste inventory the hmits, operating conditions, and other requirements applica-and confinement capability of tne Facihty Disposal Area (FDA) ble to a nuclear reactor facility as set forth in Section 50 36 of at West Varley, New York. The radioactive waste inventory for 10 CFR Part 50 for the protection of the health and safety of the FDA has a totat rd:cactivity of about 135.000 cunes (Ci) and the public. is compnsed of H-3 (9.500 Ci), Co-60 (64.000 Ci), SR-90/Y-90 (24,300 Ci). Cs 137/Ba 137m (24,400 Ci). and Pu 241 (13.300 NUREG/CP 0058 V01: PROCEEDINGS OF THE TW ELFTH C4 These wastes are buned in the Lavery Tdl. a giacial hil unit WATER REACTOR SAFETY RESEARCH INFORMATION comprised of a clayey sitt with very low hydraulic conductWy MEET!NG. SZAWLEWICZ.S A. Office of Nuclear Regulatory Re-propertres Recent studies of a tnbutyfphcshate-kerosene plume search, Director. January 1985. 434pp. 8502040194. 28714.188 moving through the shallow ground-water flow system in the The papers published in this seu volume report were present-FDA andicate a need to better assess the fracture flow compo^ ed at the Tweifth Wa'er Reactor Safety Research Information nents of this system particularly the weathered and fractured Meetirig held at the Na!'onal Bureau of Standards, Gaithersburg, Lavery Till unit. The ana!ysis of the deeper ground-water flow Maryland dunng the ween of October 22 26, 1984. The papers system studied by the USGS and NYSGS staffs ind: cates rela' desenbe progress and results of programs in nuclear safety re-tively long pathways and travel t;mes to the access,ble environ-ment. Mass wasting, endemic, to the glacial-filled valley, contnb-utes to the active slumping in the ravines surround 3 g the FDA ed by researchers from seven European countnes, Japan, and and also need attention. Canada. Volume 1 presents information on Plenary Session - I, NUREG-1165: ESRP 7,1.1 "ENVIRONVENTAL IMPACTS 0" Integral System Tests, separate Effects, International Programs POSTULATED ACCIDENTS INVOLVING RELEASES OF RA- in Thermal Hydraubcs, and Calculation of Appendix K Conserv. DIOACTIVE MATERIALS TO GROUNOW ATER ' atisms. WESCOTT,R G Office of Nuclear Reactor Regulation, Director (post 651125). November 1985. 13pp. 8512120111. ESRP NUREG/CP 0058 V02: PROCEEDINGS OF THE TW ELFTH 7.1.1. 33875.266- WATER REACTOR SAFETY RESEARCH INFORMATION Environmental Standard Review P,an (ESRP) 7.1.1 provides MEETING. SZAWLEWIC2.S A Office of Nuclear Regulatory Re-guidance to the staff for preparation of envircomental assess- search, Director. January 1985. 459pp. 8502060596. 28753.039 ments of "Radiolog cal Impacts - Releases to Groundwater " an The papers pubbshed in this six volume report were present. input to the staff's environmentai statement wrisch addresses ed at the Twelfth Water Reactor Safety Research information the groundwater pathway consequences from postulated reac- Meeting held at the National Bureau of Standards, Ga.thersburg, tor core-melt accidents. The ESRP lists the type of information aryland dunng the week of October 22 26, 1984. The papers which should be collected, references that may be useful, and g g 9 provides a procedure for uniform staff review of applicant ana!y. search conducted in this country and abrcad Foreign participa-ses. The ESRP is appbcable to both Construction Permit and tion in the meeting included twenty-six different papers present-Operateng License Stage reviews. ed by researchers from seven European countnes, Japan, and NUREG 1167: TPDWR2: THERMAL POWER DETERMl NATION Canada. Volume 2 presents information on Pressunzed Thermal FOR WESTINGHOUSE REACTORS. VERSION 2. User's Guide.
~~Shock, Code Assessment and improvement, 2D/3D Research Division of Emergency Preparedness & Engineenng Response Program, and the Nuclear Plant Analyzer Program.~~
(Post 830103). December 1985 158pp. 8601070496. 04198 236- NUR EG/CP-0058 V03: PROCEEDINGS OF THE TW5:!LFTH TPOWA2 is a computer program which was developed to de- WATER REACTOR SAFETY RESEARCH INFORMATION termine the amount of thermal power generated by any Wes* MEETING.
~~Office of Nuclear Regulatory Research, Director.~~
tinghouse nuclear power plant. From system cond.tions. January 1985. 731pp 8502060432. 28750 001. TPDWR2 calculates entha! pies of water and steam and the The papers pubbshed in this sin volume report were present-power transferred to or from vanous components in the reactor ed at the Twelfth Water Reactor Safety Research information coolant system and to or from the chemical and volume control Meeting held at the National Bureau of Standards. Gaithersburg. system. From these results and assuming that the reactor core Maryland dunng the week of October 22-26, 1984. The papers is operaing at constant power and is at thermal equil.bnum. desenbe progress and results of programs in nuclear safety re-TPDWR2 calcula'es the 'hermal power generated by the reactor search conducted in this country and abroad Foreign participa-core. TPDWR2 runs on the IBM PC and XT computers when tion in the meeting included twenty sir different papers present-IBM Personal Computer DOS, Version 2 00 or 2.10. and IBM ed by researchers from seven European countnes, Japan. and Personal Computer Basic, Version D2 00 or D2.10. are stored Canada Volume 3 presents information on Contatnment Sys-on the same diskette with TPOWR2 tems Research. Fuel Systems Research Accident Source Term NUREG-1172: TECHNICAL SPECIFICATIONS FOR RIVER BEND Assessment, and Japanese Industry Safety Researt.h. STATION. Docket No. 50 458 (Gulf States Utaties Company) BENEDICT,R. Office of Nuclear Reactor Regulation, Director NUREG/CP-0058 V04: PROCEED NGS OF THE TWELFTH (post 851125). November 1985 539pp 8512180358. WATER REACTOR SAFETY RESEARCH INFCRMATION 33955.143 MEETING. SZAWLEWICZ,5 A. Office of Nuclear Regulatory Re-The River Bend Station Technical Specifications were pre- search Director. January 1985. 388pp. 8502060428. 28752 012. pared by the U.S. Nuclear Regulatory Commission to set forth The papers pubbshed in this six volume repor* were present-the limits, operating conditions and other requirements apphca- ed at the Twelfth Water Reactor Safety Research information ble to a nuclear reactor facihty as set forth in Sect,on 50.36 of Meeting held at the Nationat Bureau of Standards, Ga.thersburg. 10 CFR Part 50 for the protection of the heaith and safety of Maryland dunng the week of October 22-26. 1984. The papers the public. desenbe progress and results of programs in nuclear safety re-search conducted in this country and abroad. Foreign participa-NUREG-1173: TECHNICAL SPECIFICATIONS FOR PALO VERDE tion in the meeting included twenty-six different papers present-NUCLEAR GENERATING STATION, UNIT 2. Docket No 50 529.(Anzona Pubhc Service Company)
Division of Pressunred ed by researchers from seven European countnes, Japan, and Water Reactor Licensing B (post 851125) December 1985. Canada. Volume 4 presents information on Matenals Engineer-513pp. 8601070482, 34191.289. ing Research
Main Citations and Abstracts 25 NUREG/CP-0058 V05: PROCEEDINGS OF THE TWELFTH pnonties to (through sensitivity studies and cnScal evaluations) WATER REACTOR SAFETY RESEARCH INFORMATION and then improving and/or obtaining important thermodynamic MEETING. SZAWLEWICZ,S.A. Office of Nuclear Reguratory Re-search, Director. January 1985. 470pp 8502060359. 28755:345. data, and (4) addressing the importance of kinetics in simulating repository behavior.
. The papers published in this six volume report were presented at the Twelfth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards. Ga thersburg. NUREG/CP-0063: PROCEEDINGS OF THE 1984 STATISTICAL Maryland dunng the week of October 22 26, 1984. The papers SYMPOSIUM ON NATIONAL ENERGY ISSUES KiNNISON.R.:
DOCTOR,P. Battelle Memonal Institute, Pacific Northwest Lab-describe progress and results of programs in nuclear safety re-oratones. July 1985. 244pp. 8507230193. 31754 241. search conducted in this country and abroad. Foreign participa-tion in the meeting included twenty six different papers present. The 1984 Statistical Symposium on Nat.onal Energy issues ed by researchers from seven European countnes, Japan, and was the tenth in a ser es of annual symposia bringing togetner Canada. Volume 5 presents information on Mechanical Engs- statisticians and other interested parties who are actively en-neenng. Structural Engineenng. Seismic Research, Process gaged in the pursuit of solving the nation's ener9y problems, Inn Control, instrumentation and Control Program, and Equipment taw the symposium was sponsored by U S Department of Quahfication and Nuclear Plant Aging. Energy (DOE) and named the DOE Statistcal Syrrposrum. The symposium is organized by a steenng committee made up of NUREG/CP-0058 V06: PROCEEDINGS OF THE TWELFTH representatives from the nat.onallaboratones The 1984 sympo-WATER REACTOR SAFETY RESEARCH INFORMATION sium was hosted by Pacific Northwest Laboratory, and it was or-MEETING. SZAWLEWICZ,S.A. Office of Nuclear Regulatory Re- ganized around four special topical sessions. (1) Assessnq and search, Director. January 1985. 515pp. 8502060357. 28757.094 Assunng High Reliab.hty. (2) Spatial Statistical. (3) Quant fica #on The papers published in this six volume report were present. of informed Opinion, and (4) Health Effects of Energy Technob ed at the Twelfth Water Reactor Safety Research Information ogies These were chosen research and data analys;s Several Meeting held at the Nat onal Bureau of Standards, Garthersburg, contnbuted papers were aiso presented Maryfand dunng the week of October 22 26, 1984. The papers desente progress and results of programs in nuclear sa'ety re- NUREG/CP-0065: TRANSACTIONS OF THE 8TH INTERNATION-Search conducted in this country and abroad. Foreign participa-AL CONFERENCE ON STRUCTURE MECHANICS (N REAC-tror, in the meeting included twenty srx different papers present. TOR TECHNOLOGY Panel Session J-K- Status of Research in ed by researchers from seven European countnes. Japan, and Structural And Mechanical Engtneenng For Nuclear Power Canada. Votume 6 presents information on Plenary Session II, Human Factors and Safeguards Research, Heafth Effects and Plants. BROWZIN.B S. Dmsion of Eng neenng Technology. June 1995. 266pp 8507080187. 31393 277, Radiation Prctection, Risk Ana'ysis, and EPRI Safety Research These transactions of the J-K/ panel session include preponts NUREG/CP-0059 V01: PROCEEDINGS OF THE MITI NRC SEIS. of papers or abstracts which are listed in Volume A. "!ntroduc. MIC INFORMATION EXCHANGE MEETING VOLUME l. tion. General Contents, Authors' Index," Proceed ngs of the 8'n WEtSS.A.J. Brookhaven National Laboratory. Apnl 1985 423pp international Conference on Structural Mechanics in Reactor 8506070372. BNL NUREG-51821. 30796 001. Technology. These papers represent the body of tne J-K/ panel The first Japan Ministry of International Trade and industry session, " Status of Research in Structural and Mechanical Ergo (MITI) U S. Nuclear Regulatory Commiss.on (NAC) Seismic in. neenng for Nuclear Power Plants," sponsored by the U S Nu-formation Exchange Meeting (SiEM) was held July 18 20, 1984 clear Regulatory Comrmssion in Palo Alto, Cahfornia The purpose of SIEM was to provide technical information on seismic research being conducted NUREG/CP-0066: PROCEEDINGS OF AN INTERNATIONAL under MITI and NRC sponsorships to the participants The aim WORKSHOP ON HISTORIC DOSE EXPERIENCE AND DOSE was to improve understanding of the seismic research in REDUCTICN (ALARA) AT NUCLEAR POWER PLANTS MAY progress in Japan and the United States for possible identifica. 29-JUNE 1,1984 HORANJR, B AUM.J W , DIONNE.B J t>on of areas of mutual interest which could be the basis for Brookhaven National Laboratory September 1985 279pp future cooperation. Approximatefy 40 Japanese and U S. techni- 8510040389 BNL-NUREG-51901 32856 242 cat specialists in seismic research partopated in the meeting Dose reduction data and experience from 28 foreign and to These proceed ngs represent the compilation of the papers pre- U S nuclear power plants was examined to determine causes sented at the meeting. for the wide variations in occupational dose from country to NUREG/CP 0062: PROCEEDINGS OF THE CONFERENCE ON cow ajor t pics discussed were. steam generator and re. THE APPUCATION OF GEOCHEMICAL MODELS TO H'GH- fuehng maintenance problems. utility and suppher ALARA pro-LEVEL NUCLEAR WASTE REPOSITORY ASSESSMENT. grams. effectiveness of dose-reduction modifications; att,tudes JACOBS.G K.; WHATLEY,S K. Oak Ridge National Laboratory and training. current and future dose-reduction research While May 1985.130pp 8506130505. ORNL/TM 9585. 30892 227. many parameters contnbute to o fferences of occupational A conference on the application of geochemical models in doses between plants from different nations, it is clear that the assessment of high-level nuclear waste repositones was most U S plants have higher collective dose eouivaient per re-held to discuss the current status of geochemical code develop. actor per magawatt-year than most erher countnes, even for ment, thermodynamic data bases, reaction kinetics, and cou- plants of similar site and age World*de. Fennesh and Swedrsh pied process models as apphed to site charactentation and per, plants, both PWR and DWR. have acheeved the lowest values formance assessment activities. These proceedings include e,. Major factors which contr bure to low doses include 1) minime tended abstracts of the technical presentations given at the ration of Coba:t in primary system components esposed to conference, a discussion of the role of geochemical modekng in water, 2) careful plant des +gn, layout and component segrega. predicting the performance of repositones, and a set of recom. tion and snielding. 3} plant standard >2ation, 4) selection of com. mendations that identify the key developments needed in order ponents and systems for increased rehabihty, 5) management for geochemical rnodels to become more applicab!e for quante interest and commitment, 6) minimum number of workers and tative evaluations of repositones. Detailed recommendations rel. indepth worker training. 7) careful control of primary systar, evant to the follomng subjects are discussed (1) improved sim. Oxygen and pH. 8) good pnmary system water punty to minimize ulation of repository performance through inclusion of additional corrosion product formation, 9) use of special tools and robot-smportant geochemical processes and parameters into current ics,10) decontamination and passivation of pnmary systems geochemical models, (2) more careful attention to uncertainties and components, and 1 t) extent of backfitting and mandated in-associated with geochemical model calculations, (3) assigning spections.
26 Main Citations and Abstracts suits would be similar to those for PWRs. Only SAFSTOR is NUREG/CP-0070: PROCEEDJNGS OF THE WORKSHOP ON considered because DECON and ENTOMB are unsuitable by SEISV4 AND DYNAMIC FRAGILITY OF NUCLEAR POWER definition for intenm storage of radioactive wastes and/or spent PLANT COMPONENTS. HOFMAYER.C.H.; BANDYOPADHYAY Brockhaven National Laboratory. August 1985. 278pp. fuel. It is assumed that all radioactive wastes and spent fuel are shipped offsite by the end of decommissiong. 8511180634. BNL-NUREG-51924. 33507.085. The Workshop on Seismic and Dynamic Fragility of Nuclear NUREGICR-2000 V03N12: LICENSEE EVENT REPORT (LER) Power Plant Components was held at Brookhaven National COMPILATION For Month Of December 1984.
~~Oak Ridge Na-Laboratory (BNL) on June 5-7, 1985. The purpose of the work- tional Laboratory. Janua#Y 1985. 49pp. 8501280397. ORNL/~~
shop was to provide a forum for exchangrng concepts, informa- NS C-200 28 72 297 tion and expenences on the fragibty of electncal, control and m;chanical equipment used in nuclear power plants when sub- o erational information that was processed into the LER data jected to seismic and other dynamic environments. The work- g g shop was divided into six sessions which included discussions one month penod identified on the cover of the document. The on definition, uses and importance of component fragility; pa- LERs. from which this information is denved, are submitted to rameters affecting component fragility; categonzing equipment the Nuclear Regulatory Commission (NRC) tf nuclear power and existing test results; methodology and apphcation of fragibty lant licensees in accordance with federal regulatons Proce-data to equipment assembhes; equipment requinng future fragth- dures for LER report:ng for those events (and revisions to thvse fy testing; and, use of fragihty data in PRA and Seismic Margin events) occurnng pnor to 1984 are described in NRC Regula-studies. The proceedings represent the compdation of the g g g g papers presented at the workshop. d WW Em RWL 6 Nw events occurnng on and after January 1,1984, LERS are being NUREG/CP-0071: TRANSACTIONS OF THE THIRTEENTH WATER REACTOR SAFETY RESEARCH INFORMATICN submitted in accordance w th the revised rule contained in Title MEETING. WEISS.A.J. Office of Nuclear Regulatory Research. 10 Part 50 73 of the Code of Federal Regufations (10 CFR Director. October 1985. 276pp 8510170230. 33046.076. 50.73 1447 on July 26.1983. NUREG 1022. Licensee Event This report contains summanes of papers on reactor safety Report System Desenption of Systems and Guidehnes for Re-research to be presented at the 13th Water Sa'ety Research in- porting. provides supporting guidance and informahon on the re-vised LER rute. The LER summanos in this report are arranged formaton Meeting held at the National Bureau or Standards in alphabetically by facihty name and then Chronofogicaffy by event Gathersburg. Maryland. October 22-25, 1985. The summanes date for each f acility Component, system, keyword, system and briefly descnte the programs and resuits of nuclear safety re-search sconsored by the Office of Nuclear Regulatory Re- general keyword indenes are assigned by the computer using searen. USNRC. Summanes of invited papers are also included, correlation tables from the Sequence Coding and Search which cover the highlights of reactor safety research conducted System. by the electnc utihties through the Electnc Pcwer Research In-stitute, the nuclear industry, and the research of government CCMPILATION For Month Of Janua'Y 1985.
~~Oak Red 9e Na-and industry in Europe and Japan. The summanes have been tena Laboratory February 1985. 59pp. 8503150302. ORNL/~~
compiled in one report to provide a basis for meaningful d.scus- NS ) 2 2 sion and information exchange dunng the course of the meet- p ' eng and are given in the order of their preser'tation in each ses-son NUREG/CR 2000 V04 N2: LICENSEE EVENT REPORT (LER) COMPILATION For Month Of February 1985.
~~Oak Ridge Na-NUREG/CR 1677 V02: P! PING BENCHMARK tronal Laboratory February 1985 7Cpp 8504030412. ORNL/~~
PROBLEMS. VOLUME 11 DYNAMIC ANALYSIS INCEPENDENT NSIC-200. 29604 359 SUPPORT MOTION RESPONSE SPECTRUM METHOD See NUREG/CR-2000,V03.N12 abstract. BEZLER.P.; SUBUDHl.M.; HARTZMAN.M Brookhaven Nat,onal Laboratory. August 1985. 401pp. 8509160031. BNL-NUREG- NUREG/CR 2000 V04 N3: UCENSEE EVENT REPORT (LER) COMPILATION For Month Of Maren 1985.
~~Cak Ridge Nation-51267.32625:321.~~
Four benchrnark prob! ems and solutions were developed for af Laboratory Apnl 1985.78pp 8505070557. ORNL/NSiC-200 ventying the adequacy of computer programs used for the dy- 30210.092. namic analysis and design of elastic piping systems by the inde- See NUREG/CR 2000.V03.N12 abstract pendent support motion, response spectrum method The dy- NUREG/CR 2000 V04 N4: LICENSEE EVENT REPORT (LER) namic loading is represented by distinct sets of support excita. COMPILATtCN For Month Of Apnl 1985.
~~Oak R.dge National tion spectra assumed to be induced by non-uniform excitation in Laboratory May 1985 87pp 8506130364. ORNL/ NSIC-200.~~
three spattal directions. Complete input descnptions for each 30867 262. problem are provided and the solutions include predicted natu- See NUREG/CR 2000,V03.N12 abstract. ral frequencies, participation factors, nodat displacements and e!ement forces for independent support excitation and also for NUREG/CR 2000 V04 N5: LICENSEE EVENT REPORT (LER) uniform envelope spectrum excitation. Solutons to the associat' COMPILATION For Month Of May 1985.
Oak Ridge National j ed anchor point pseudo-static displacements are not included- Laboratory. June 1985.111pp 8507030669 ORNL/ NSiC-200 i 3t 314155 NUREG/CR 1755 ADD 01: TECHNOLOGY. SAFETY AND COSTS See NUREG/CR-2000,V03.N12 abstract.
CF DECOMMISS6ONING NUCLEAR REACTORS AT MULTI-PLE REACTOR STATIONS Effects On Decommissioning Of in' NUREG/CR 2000 V04 NG: LICENSEE EVENT REPORT (LER) tenm inabihty To Dispose Of Wastes Offsite. MOORE.E B Bat- COMPILATION For Month Of June 1985
~~Oak Ridge National telle Memonal institute. Pacific Northwest Laboratones. April Laboratory. Juff 1985 133pp. 8508150071. ORNL/NSIC 200.~~
1985. 41pp. 8505070571. 30209 200. 32196 336. The purpose of this addendum is to examine tho irrpacts of See NUREG/CR 2000,V03.N12 abstract an intenm inabihty to carryout offsite disposal of radeoactive wastes and spent fuel on the decommissioning of mutt;ple-reac. NUREG/CR 2000 V04 N7: LICENSEE EVENT REPORT (LER) COMPILATION For Month Of July 1985
Oak Ridge National for power station. The example so'ected for study is a four PWR Laboratory August 1985. 129pp 8509060195 ORNL/NSIC-station in which each PWR is prepared for safe storago at two-200 3250514f.
year intervals, held in safe storage for 100 year intervals BWRs See NUREG/CR 2000,V03.N12 abstract are neglected for simpilCity and in the espectaton that the re-
___. _ - _ _- . _ __ __. ~- . _ _ _ ._- _ Main Citations and Abstracts 27 4 NUREG/CR-2000 V04 N8: LICENSEE EVENT REPORT (LER) This report is part of an ongoing effort to review the natonal COMPILATION.For Month Of August 1985.
Oak Ridge Nat on-al Laboratory. September 1985.147pp. 8510030309. ORNL/ high level waste package program. The contnbutions of individ-NSIC-200, 32848.162. ual waste package components to cor'lainment and controlled <
See NUREG/CR-2000,V03.N12 abstract. release of radonuclides after emplacement in salt, basalt, tuff and granite repositones are evaluated. The U S. crystalline NUREG/CR-2000 V04 N9: UCENSEE EVENT REPORT (LER) (granite) reposdwy program is reviewed and relevant foreign COMPILATION For Month Of September 1985.
Oak Ridge data are outhned. The use of crushed salt, bentonite and reo-National Laboratory. October 1985. 118pp. 8511110422. hte contaming packmg matenals is discussed. Temperatures ORNL/NSIC-200. 33417:093. and gammayrradiation are shown to be important environmental See NUREG/CR-2000.V03,N12 abstract parameters in assessing waste package performance.
~~NUREG/CR-2000 V04N10: LICENSEE EVENT REPORT (LER) NUREG/CR 2482 V07; REVIEW OF DOE WASTE PACKAGE j( COMPILATION For Month Of October 1985.~~
Oak Ridge Na- PROGRAM. Subtask 1.1 National Waste Package Program tional Laboratory. November 1985.123pp. 8512100733. ORNL/ Apnl 1984. September 1984. SOO.P. Brookhaven Nahonal Labo-4 NSIC-200. 33832:146. ratory. March 1985. 88pp. 8504030408- BNL-NUREG-51494.
3 See NUREG/CR-2000,V03,N12 abstract. 29604:266. NUREG/CR-2000 V04N11: LIC8ENSEE EVENT REPORT (LER) e pesent eM is pad of an pgoing task to ww the na-COMPILATION.For Month Of November 1985.
Oak Ridge Na- nal waste package Mod H mcWes waluatons of tional Laboratory. December 1985.117pp. 8601070500. ORNL- refere ce waste form, container, and packing matenal compo-NSIC-200. 34155:131 nents with respect to determming how they may contnbute to See NUREG/CR-2500,V03,N12 abstract. the containment and controlled release of radionuchdes after waste packages have been emplaced in salt, basaft, tuff, and NUREG/CR-2331 V04 N2: SAFETY RESEARCH PROGRAMS granite repositones. In the current Biannual Report a review was SPONSORED BY OFFICE OF NUCLEAR REGULATORY camed out to determine the ability of spent fuel cladding to pro.
I RESEARCH. Quarterly Progress Report,Apnl 1 June 30,1984- vide additional radonuclide containment caoabAty should the WEISS.A.J. Brookhaven National Laboratory. February 1985. container /overpack system fail prematurely. T is' prog ess e rt wl d n rre a iv tes and techns, NUREG/CR 2482 V08: REVIEW OF DOE WASTE PACKAGE l cal progress in the programs at Brookhaven National Laboratory PROGRAM Semiannual Report Covent$g The Penod October sponsored by the Division of Accident Evaluaton, Divisen of 1984 - March 1985 DAVIS.M S.; BREWSTER.C.; GAUSE.E.; et j Engineenng Technology, and Devision of Risk Analysis & Oper- at Brookhaven National Laboratory. December 1985. 300pp. 1 atrons of the U.S. Nuclear Regulatory Commission, Office of Nu- 8601070538. BNL-NUREG-51494. 34181:2a9, ,
! clear Regulatory Research. The projects reported are the foi- A large number of technical reports on waste package com-lowing: High Temperature Reactor Research, SSC Develop-I ponent pertemance were reviewed over the last year in support ment, Vahdaten and Apphcation, CRBR Balance of Plant Mod- of the NRC's review of the Department of Energy's (DOE's) En-4 shng ThermalHydraulic Reactor Safety Expenments, Develop- vironmental Assessment reports. The intent was to assess iri ; ment of Plant Analyzer, Code Assessment and Application some detail the quantity and quality of the DOE data and their (Transient and LOCA Ana r/ses), Thermal Reactor Code Devel-
' relevance to the high-level waste repository site selection proc. " opment (RAMONA 38), Calculational Ouakty Assurance in Sup- ess A representatue selecton of the rewews es presented fa j port of PTS: Stress Corrr'sion Cracking of PWR Steam Genera- the saft, basaft and tuff repository projects. Areas for future re-
~~search have been outhned.~~
tor Tubing Probabshty Based Load Combinations for Design of Category 1 Structures, Mechanical Piping Benchmark Problems, Identification of Age Related Failure Modes; Analysis of Human NUREG/CR-2482 V09: REVIEW OF DOE WASTE PACKAGE } Error Data for Nuclear Power Plant Safety Related Events, PROGRAM. Semiannual Report Covenng The Pered Apol 1985-Human Factors Aspects of Safety / Safeguards interactons, September 1985 SULLIVAN.T.; JAIN,H.; ABRAHAM.T,; et at Emergency Action Levels, a9d Protective Action Decison Brookhaven National Laboratory. December 1985. 98pp. Miking. 8601070503. BNL NUREG 51494. 34183.180. i Detarted evaluatons continued on DOE reports and papers NUREG/CR 2331 V04 N3: SAFETY RESEARCH PROGRAMS concerned witn the evaluaten of waste package component be. ! SPONSORED BY OFFICE OF NUCLEAR REGULATORY havior. The entent was to estimate the quantity and relevance of 1 RESEARCH.Ouarterty Progress Report, July * , . September data being generated for bamer system performance analysis J 30,1984. WEISS.A.J. Brooithaven National Laboratory. May in additon, several review studies have been completed to 1985.117pp. 8506060147. BNL-NUREG 51454. 30781:002. evaluate progress in the DOE waste package program. These ! See NUREG/CR 2331,V04,N02 abstract. enclude work on the selecten of a glass composition for West i NilREG/CR 2331 V04 N4: SAFETY RESEARCH PROGRAMS Val lwy,. New York, high level waste, a descnption of the system SPONSORED BY OFFICE OF NUCLEAR REGULATORY at West Valley for utnfying the waste, and reviews of papers in-l- RESEARCH.Ouarterfy Progress Report, October 1 December cluded m the Defense High-level Waste Leaching Program and 31, 1984. WEISS.A.J. Brookhaven National Laboratory. May the recent Tucson, Anzona, " Waste Management '85" Confer-ence, 1985.139pp.8507050378. BNL NUREG-51454. 31373 001. N-See NUREG/CR 2331,V04,N02 abstract. NUREG/CR 2531 R03: INTRODUCTORY USER'S MANUAL FOR ) NUREG/CR 2331 V05 N1: SAFETY RESEARCH PROGRAMS THE U.S. NUCLEAR REGULATORY COMMISSION REACTOR
) SPONSORED BY OFFICE OF NUCLEAR REGULATORY SAFETY RESEARCH DATA BANK. HARDY,H.A.; LAATS.E.7, RESEARCH Ouarterly Progress Report. January 1 March EG&G, Inc. May 1985. 95pp. 8505170007. EGG 2164. ; 31,1985. WEISS.A.J. Brookhaven National Laboratory. August 1985.250pp.8512120133. BNL NUREG-51454. 33874:133' 30483.079'ed The Unit States Nuclear Regulatory Commission See NUREG/CR 2331,V04,N02 abstract. has estabhshed the NRC/ Division of Accident Evaluation (DAE)
Data Bank Program to collect, store, and make available data NUREG/CR-2482 V06: REVIEW OF DOE WASTE PACKAGE from the rnany domestic and foreign water reactor safety re-PROGRAM Subtask 1,1 - Nat onal Waste Package ProgramOctober 1983 - March 1984. SOO,P. Brookhaven Na- search programs. Local direction of the program is provided by EG8G Idaho.Inc., pnme contractor for the Department of tional Laboratory. March 1985. 53pp. 8504030414. BNL. ! NUREG-51494. 29605.065, Ener0y (DOE) at the Idaho National Engineenng Laboratory (INEL). The NRC/DAE Data Bank Program provides a central i I I ,i
28 Main Citations and Abstracts computer storage mechanism and access software for data that na tive human factors methodology approach may be u;ed in is to be used by code develcpment and assessment groups in sg scific future cases in which the methods identified in the ini-meeting the code correlation needs of the nuclear industry. The tia. report (NUREG/CR-2800) may not adequately assess the administrative po-tion of the program provides data entry, docu. proper impact for resolution of new safety issues. The altema-mentaton, training. and advisory services to users and the tive methdology included in this suppfement is entitled Method-USNRC. The NRC/DAE Data Bark and the capabihties of the ology for Est, mating the Public Risk Reduction Affected by data access software are desenbed in this document. Human Factors improvement. The priontization section of this report is ent.tfed Pnontrzation of the U.S. Nuclear Regulatory NUREG/CR 2663 V01: INFORMATION NEEDS FOR CHARAC. Commission Human Factors Program Plan. TERIZATION CF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOGIC MEDIA Main Report.
Ertec Western. Inc. NUREG/CR 2815 V01 R1: PROBABfLISTIC SAFETY ANALYSIS May 1985. 574pp. 8506270457. 31261.001. PROCEDURES GUIDE Sections 17 And Append.ces.
Evaluation of the geologic iso!ation of radioactive materials BARI.R A.; BUSLIK,A.J ; CHO.N Z; et al. Brookhaven National from the biosphere requires an int. mate kno*Iedge of site geo- Laboratory August 1985. 203pp. 8509110037. BNL-NUREG-logic conditions, whicn is gained through precharactenzation 51559. 32561.010 and site charactenzation studres. This report presents the re- A procedures guide for the performance of probabilistic safety suits of an intensive laterature rev,ew. analys.s and compJation assessment has been prepared for intenm use in the Nuclear to dehneate the information needs, apphcable techniques and Regulatory Commission programs. It will be revised as com-evaluation entena for programs to adequately charactenze a s te , , g in six geolog'c media These media, in order of presentation. The probabihst.c sas ety assessment studes performed are in-are granite, shate, basaft. tuff, bedded salt, and domed satt tended to produce probabikstic predictive models that can be Guidelines are presented to assess the efficacy (apphcation. ef- used and extended by the utshties and by NRC to sharpen the fectiveness and resolutioni of currently used to exploratory and focus of inquines into a range of issues affecting reactor safety testing techniques for precharacterizamn or charactenzation of ggg gg9 g a site. These guidehnes include the rehabihty. accuracy. and res- probability (per year) of core damage resulting from accident ini-olutica of techniques deemed acceptable, as well as cost est,. t;ators mtemal to the plant (i.e , intnnsic to plant operation) and mate of vanous field and laboratory techniques used to obtain from loss of off.stte electnc power, The scope includes human the necessary information. Gu<dehnes presented do not assess '8I'abihty analyss, a determination of the importance of vanous the relative suitabihty of media This report consists of two vof- core damage acciderit sequences, and an cap;ic:t treatment and umes main report and appendices. display of uncertainties for key accident seq ences. The second NUREG/CR-2663 V02: INFORVATION NEEDS FOR CHARAC- volume dea's with the treatment of the so caUed external events TER12ATION OF HIGH LEVEL WASTE REPOSITCRY SITES IN including seismic disturbances fires, floods, etc. Ultimatory, the SIX GEOLOGIC MEDIA. Appendices.
Ertec Western, Inc... Ma/ guide will be augmented to include the piant specific analysis of 1985. 700pp 8506270247. 31259 001. in. plant processes (ie , containment perforrnance). This guide See NUREG/CR-2663.V01 abstract, prov, des the structure of a probabilistic safett study to be per-formed, and mdicates what products of the study are valuable NUREG/CR 2718: STEAM EXPLOSION EXPERIMENTS WITH for regulatory decisron making For interr.al events, methodology SINGLE DROPS OF IRON OXIDE MELTED WITH A CO2 is treated in the guide only to the extent necessary to indicate LASER Part 11 Parametnc Stud.es. NELSON.L S ; DUDA.P M_
Sandia National Laboratones Apnl 1963 154pp. 8506140047. the range of methods which is acceptable, ample reference is grven to a:ternative methodologies which may be utibzed in the SAND 82.f105 30908 2t9 performance of the study For external events, more expbcit The steaan explosion expenments performed with s.ngte drops of molten iron oxice melted with a CO(2) laser, descnted guidance is given. in Part I of this report. were extended here The followirg ma,or NUREG/CR 2815 V02 R1: PROBABILISTIC SAFETY ANALYSIS pwameters were varied ambient pressure, water temperature PROCEDURES GutDE Sections 812. MCCANN.M.; REED.J W ' and subcoohng melt temperature, and melt composition Also, RUGER.C.; et al Brockhaven National Laboratory August 1985 a few scopng expenments were performed to erpfore the ef- 369pp 8503110050. BNL-NUREG 51559 32560 001 fects of changmg the nature of the coofant, and the viscosty of See NUREG/CR-2815,V01.R1 abstract the melt. As each of the four marr parameters was vaned, thresholds could be located beyond which emplosions were sup- NUREG/CR 2850 V03: POPULATION DOSE COMMITMENTS pressed. However, in general, the explosions could be reinitiat- DU'. TO RADIOACTIVE RELEASES FROM NUCLEAR POWER ed trf increasing the magnitude of the triggering pulse. The ef. PLANT SITES IN 1981. BAKER.D A.; PELOCUIN,R A Battelle fects of increasing the ambient pressure up to 1.12 MPa were Memonal Institute, Pacift Northwest Laboratones. January faster and finer melt fragmentation, and faster and more com- 1985 128pp. 8502060603. PNL 4221. 28745 070 plete transfer of heat from melt to water. Moreover, tnggenng Perulation radtation dose commitments have been estimated became easier over the range of ambient pressure between from reported radionuclide releases from commercial power re-about 01S MPa and approrimately 0 7 MPa actors operating dunng 1981. Fifty year dose commitments from a one year exposure were calcufated from both hquid and at. NUREG/CR-2800 S03: GUIDELINES FOR NUCLEAR POWER mosphertC releases for four population groups (infant, chid, l PLANT SAFETY ISSUE PRIORITIZATION INFORMATICN DE-ANDREWS.W B ; BICKFORD.W E., teen-ager and aduft) residing between 2 and 80 km from each VELOPMENT. site This report tabulates the resu'f s of these calculatsons, COUNTS.C A ; et al. Battelte Memonal Institute. Pacific North. west Laborarones. September 1985, 143pp 8510030435. snowing the dose commitments for both hquid and airborne pathways for each aje group and organ. Also included for each 32847.283 This supplemental report is the fourth in a senes that docu- site is a histogram showing the fraction of the total population ment and use methods developed by the Pacific Northwest Lab. withn 2 to 80 km a ound each site receiving vanous average dose commitments from the airborne pathways The total dose oratory to calculate, for pnontization purposes, the nsk, dose and cost impacts of implementing resolutions to reactor safety commitment from both hquid and airborne pathways from 48 issues. The initial repcrt in this senes was pubbshed by An- sites ranged from a high of 20 person rem to a low of 0 008 drews et al in 1983 as NUREG/CR 2800. This supplement con- person-rom with an arithmetic mean of 3 person-rem. The total population dose for all sites was estimated at 160 person-rem l sists of two parts desenbing separate research efforts. (1) an al-ternative human factors methodougy approach and (2) a pnon- for the 98 milhon people considered at nsk. The average individ- l tization of the NRC's Human Factors Program Plan. The alter- ual dose commitment from all pathways on a sito basis ranged ( l l
i Main Citations and Abstracts 29 from a low of 1 x 10(-5) mrem to a high of 0.05 mrem. No at-tempt was made in this study to determine the maximum dose and deterministic analysis, a data marna approach could be commitment received by any one individual from the radionu- measured and recorded, and those which aro required for cer-clides released at any of the sites, tain types of events involving thermal hydraulics and neutronics as illustrative of events requinng in-depth analysis. Also includ-NUREG/CR-2951: THE D9 EXPERIMENT. Heat Removal From ed in the study was a review of INPO's Nuclear Plant Reliability Strahfied UO2 Debos. OTTINGER.C.A.; MITCHELL.G W.; Data System; NASA's Problem Reporting and Correctve Action LIPINSKI,R.J.; et al. Sandia Natonal Laboratories. June 1985, (PRACA) program. Electncite de France's KIT system, an auto-74pp. 8507050431, SAND 84-1838. 31371:101. matic computer based reactor parameter monitonng and record-The D9 expenment investigated the coolabary of a shallow ing system; and the regulatory relationship between the FAA (77 mm), stratified urania bed in sodiurn. The bed was fission and the commercial airline industry. heated in the Annular Core Research Reactor (ACRR) at Sandia National Laboratones to simulate the effects of radioac. NUREG/CR 3091 V04: REVIEW OF WASTE PACKAGE VERIFl-tive decay heating. It was the first stratified debris bed expen-CATION TESTS Semiannuar Report Covenng The Penod Octo-ment to use an extended UO2 pa.ticle size distnbution (0.038 to ber 1983 - March 1984 JAIN.H ; VEAKIS.E.; SOO.P. Brookha-4 0 mm). Dryout occurred at powers ranging from 0.10 to 0 58 W/g. which was close to the incrpient boiling power and before ven National Laboratory. June 1985. 29pp. 8507050398. BNL-Channels penetrated the subcooled zone in the bed, even with NUREG-51630. 31373 314. subcoolings as low as 80 degrees centigrade. Channel penetra' The current study is part of an ongoing task to specify tests ton was observed after dr)out began, but the bed became or;ly that may be used to venty that engineered waste package /re-moderately more coolable. All these observations agree with pository systems comply with NRC radionuchde containment current models. and controffed release performance oblectives Work covered in th.s report includes tuff packing matenal for use in a high level NUREG/CR-3005:
~~SUMMARY~~
OF THE NUCLEAR REGULATORY waste tuff repository Ranges of repository cond tions relevant COMMISSION'S LOFT PROGRAM RESEARCH FINDINGS to its testing and other factors important for its performance are NALEZNY,C.L EG&G. Inc. Apnl 1985.199pp. 8507050424. discussed EGG-2231. 31372.001. This document is a summary of the main research results of the Loss-of Fluic Test (LOFT) Program relative to code assess- NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIF1-ment, code dwiopment, licensing, rutemaking, safety technolo- CATION TESTS Semiannual Report Covering The Penod Apnl gy and reactor operat.ons. TV LOrT facaty is a 50 MW(t) 9984 - September 1984 JAIN.H ; VEAKIS.E.; SOO.P, Drookha-pressunzed water reactor (PWR) system with instrurrents that ven National Laboratory. June 1985 34pp 8507050402. DNL. measure and provide data on the system thermal-hydraulic and NUREG-51630 31372 302. nuclear cond: tens. The transient response of the LOFT system This ongoing study is part of a task to specify tests that may to accident events is similar to large [1000 MW(e)) commercal be used to venfy that engineered waste packages / repository PWRs. The main objectives of the LOFT Expenmental Program systems comply with NRC radonuchde containment and con-were to qualify the engineered safety systems used in commer. troHed release performance objectives Work covered in this cial PWRs and to venfy the computer codes used in safety report includes crushed tuff packing matenal for use in a hcgh analyses. The LOFT Program contnbuted to the irrorovement of fevel waste tuff repository. A review of avadable tests to quanto computer codes used to predict the response of commercial fy packing performance is g:ven together with recommendations PWRs demonstrated the adequacy of engineered safety sys. for future testing work. tems, and contnbuted to improved understanding of PWR acci-dent phenomena. particularty those associated with the evalua. NUREG/CR 3091 V06: REVIEW OF WASTE PACKAGE VERIFi-tion model in Appendm K to 10 CFR 50 (the "ECCS rule") CATION TESTS Semiannual Report Covenng The Penod Octo-NUREG/CR-3019: RECOMMENDED WELDED CRITERIA FOR ry u 98S pp 822 3 B UF E USE IN THE FABRICATION OF SHIPPING CONTAINERS FOR O RADIOACTIVE MATERIALS. MONROE.R E.; WOO,H H.; 32346 154' SEARS.R.G. Lawrence Livermore National Laboratory. March e poyntial of WAPPA, a second generation waste package 1985.16pp. 8504040007. UCRL 53044. 29619 236. system code, to meet the needs of the regulatory community Welding and related operations are evaluated to assess the are anatyred The anatysis includes an indepth review of centrols required to prevent weld related fadure of sh!pp#ng con- WAPPA's individual process models and a review of WAPPA's tainers used for transportaten of radioactive matenats The operation it is concluded that the code is of limited use to the report includes (1) recommended entena for controlling welding NRC in the present form Recommendations for future improve-as apphed to shipping containers, and (2) a discussion of mods ment, usage, and implementation of the code are given. This ficatens of the recommended industry Codes as apphed to report also descobes the results of a testing program undertak-shipping containers. en to determine the chemical environment that will be present NUREG/CR 3026: FEASIBILITY STUDY ON THE ACOUISTION near a high-level waste package emplaced in a basalt reposs-OF LICENSEE EVENT DATA. K ATO,W.Y; HALL,R E ; tory 6 this purpose, low cart on 1020 steel (a current BWIP TEICHMANN.T.; et al. Brookhaven National Laboratory. Febru- reference container matenal). synthetic basartic groundwater ary 1985. 267pp. 8503080498. BNL-NUREG 51609. 29285134. and a mixture of bentonite and basalt were exposed in an auto-Brookhaven Nat2cnal Laboratory's Department of Nuclear clave, to espected conditions some penod after repository seal. Energy (DNE) has performed a study of the Licenseo Event ing (150 degrees centigrade. approvimately 10 4 MPa) Param. Report (LER) system. The obect;vel of the study was to assess eters measured include changes in gas pressure with time and the feasibility of modifying the LER reporting system as pro. gas compositen, vanahon in dissolved orygon (DO). pH and posed by NRC-AEOD, and/or developing an alternative plan certain ionic concentrations of water in the packing matenal that would in addition collect informat on about significant across an smposed thermal gradient, mineralogic alteration of ever.ts amenable to statistical analysis, such as multH:ase, the basaft/ bentonite mixture, and carbon steel corroton behav-multevanate analysis. The study indicated that the LERs conste cr. A second test.ng program was also initiated to check the tute reports from a large vanety of events which have in most likehhood of stress corrosion cracking of austenitic stainless cases many different plant parameters, both measured and cur- steels and Incoloy 825 which are being considered for use as rently not measured, to charactente the event. In order to do-termine event-specific plant parameters required for statistical waste container matenals in the tuff repository program
)
30 Main Citations and Abstracts NUREG/CR 3193: FORCED NUREG/CR 3145 V03: GEOPHYSICAL INVESTIGATIONS OF CONVECTIVE,NONEOUILIBRIUM. POST CHF HEAT TRANS-THE WESTERN OHIO-INDIANA REGION - ANNUAL September 1983, Volume 3). FER EXPERIMENT DATA AND CORRELATION COMPARISON REPORT.(October 1982 - REPORT. GOTTULA,R C.; CONDJE,K.G ; SUNDARUM.R K.; et POLLACK,H.N.; CHRISTENSEN,D.; WELC,J. Michigan, Univ. of. Ann Arbor, Mi, May 1985. 51pp. 8507230043. 31753.299. al. EG&G, Inc. Apnl 1985. 562pp. 8504160110. EGG-2245 Earthquake activity in the Western Oho - Indana regon has 23833.00 f. Forced convective postcntical-heat flux heat transfer expen-been monitored with a precison seismograph network consist. ments with water flowing upward in a vertical tube have been ing of nine stations located in west-central Ohio and four sta. conducted at the Idaho National Engineenng Laboratory. Thor-tions sited in Indiana. Twelve local and near regional eartn. modynamic nonequilibnum in the form of superheated vapor quakes have been recorded and located dunng this report temperatures was measured at a maximum of three different penod, ranging in magnitude from 0.3 to 4.0 m(big). An event axial levels. Steady-state expenments were conducted at pres-which occurred on January 14,1984, in Toledo, Ohio, and two sures of 0.2 to 0.7 MPa, mass fluxes of 12 to 24 kg/m(2)e, events on July 28 and August 29,1984, near Terre Haute, Indi, heat fluxes of 7.7 to 27.5 kW/m(2), and test secton intet qcali-ana, were fell Only minor damage was reported from these ties of 38 to 64% Ouase-steady state (slow moving quench ever.ts. Of the twelve events, four occurred in the center of the front) expenments were conducted at pressures of 0.4 to 7 Oho array, three occurred near the city of Toledo, Ohio, four MPa, mass fluxes of 12 to 70 kg/m(2) s, heat fluxes of 8 to 225 occurred in Indiana (including one on the Indiana-lffinois border), kW/m(2), and test section inlet qualites of 7 to 47%. The rnul-and one was located near Chicago, Illinois. Teleseism.c P-wave tiple probe data and the data taken above 0.4 MPa are new rssiduals have been updated and evaluated by back projection data in parameter ranges not previously obtained. Compar son to varcus depths in the lower crust. The residuals are found to dau e cunent vapor generaton Mb aM way Nat correspond roughly to magnetic anomalies in the lower cruit of transfer models yielded unsatisfactory results. This is attntuted Ohio. It is thought that these magnetic anomahes may represent to the effects of nonequilibnum, quench front quahty, and dis-the remains of an ancient nft zone ur perhaps they are the sig- e n w am fa@s not EW in nature of the Grenville Front complex which may cross through the current models compared. this area. NUREG/CRd197 V01: REACTION BETWEEN SOME CESIUM. NUREG/CR-3174 V02: GEOPHYSICAL GEOLOGICAL STUDIES ICDtNE COMPOUNDS AND THE REACTOR MATERIALS 304 OF POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT STAINLESS STEEL,1NCONEL 600 & SILVER, Volume ICesium ZONE. Annual Report For 1983. HINZE,W J ; BRAILE LW- Hydroxide Reactons. ELRICK,R M.; SALLACH.R.AJ Purdue Univ., West Lafayette, IN KELLER.G R.; et al. Texas. OUELLETTE.A L; et al. Sandia Natienal Laboratones. June Univ. of. El Paso, TX. Apnl 1985. 60pp. 8504220438 1985.156pp 8507020369. SAND 83-0395. 31307:175. 29946'263. Laboratory scale scoping studies, using chemical simulants. Recent geophysical investigatons have shown that the seis- are examining physical and chemical processes that could occur between fisson products and other pnmary system mate-micity of the New Madnd, Missoun, seismogenic regon corre-lates with an ancient nft complex suggesting that the anoma- nals in a steam and hydrogen environment The chemical sys-fous seismscity is the result of the localizaton of the regonal tems studed were cesium hydroxtde vapor reactons in steam compressive stress pattern by basement structures. An integrat- and hydrogen at 970K in a 304 Stainless steel system, at 1120K in a 304SS system, at 1000K with edine vapor en an alu-ed geophysscal/gelogical research progra:n is being conducted to evaluate the nft complex hypothesis, to refine our knowledge mina system and at 1000K with hydrogen odine vapor in an of the structure and physical properties of the nft complex, and alumina system. Ma;or observations and conclusions are that: to investigate the possible northern extensions of the New cesium in the CsOH reacts with the secon derede en the inner oxide formed on stainless steel to produce a cessum secate, Madnd Fault zone, especially the possibfe northeastern connec-the availabAty of SiO(2) may therefore control the extent of re-tion to the Anna, Ohio, seismic region. InvesbgaDon of the action of CsOH wth 304SS in steam; the oxidabon rate of northeast extensson has focused upon the acquisiten and prep- 304SS is enhanced by the exposure to CsOH vapor; the reac-aration of arrays of gravity and magnetic data sets. Dunng 1983, special emphasis was placed upon integration of these data tion of CsOH wtth Iconel 600 is slow 6n steam and seems to react eth the SAca content in the oxide layer. with basement lithologic and seismicity information which has revealed several major lithologic / structural features in the crust NUREG/CR 3208: TRAC PD2 DEVELOPMENTAL ASSESSMENT of the Anna area. Current seismicity in this region appears to be KNIGHT,T.D; METZGER V Los Afamos Scientific Laboratory related to an ancient nft structure (the Fort Wayne nft) and pos- A nl 1985 371pp. 8504160087, LA-9700-MS 29835 001 sibly its contact with a low density pluton. Minor seismicity may This report descnbes the final results of the development as-be caused by stress concentraton associated with local base- sessment analyses conducted dunng the later stages of the ment inhomogeneities TRAC-PD2 development. The calculabons discussed in this report used the released version of TRAC.PD2 and cover sepa-NUREO/CR 3178: STRUCTURAL AND TECTONIC STUDIES IN rato effects blowdown, heat transfer, and downcomer penetra-NEW YORK STATE. Final Report.Juty 1981 - June 1982.* ton tests together with integral tests from the Loss-of Fluid Test ISACHSEN,Y.W. New York, State Univ. of, Albany, NY and Semiscale facilibes Although these calculatons are not an Boston College, Chestnut Hill, MA. Apnl 1985 84pp. exhaustive test of the code, they demonstrate its Capabeties. in-8505100048. 30270 214 cio ng automatic steadtstate inibalizabon and the compInte i SubIects treated in this report include the d'stnbuton, trends- transient from blowdown through refill and reflood The results ' exposure charactenstics, aeromagnetic signatures, and detailed show good agreement between the calculated parameters and geometnes of fracture systems, as well as tentative inferences the data and indicate that thu code is applicable to large-break concerning relative ages and causes of reactr<ation Stress indi- loss-of-coolant accident anafyses. cators are discussed, and a beginning is made at working out regional paleostress directons using the attitudes of dated NUREO/CR 3228 V03: STRUCTURAL INTEGRITY OF WATER mnf,c dAes. Attempts at defining Holecene and recent crustal REACTOR PRESSURE BOUNDARY COMPONENTS Annual movements using geological, geodebc, and seismological meth. Report For 1984. LOSS.F.J Mater als Enginoenng Associates, ods are reviewed, as well as attempts to relate projected focal Inc. June 1985 171pp 85062fM18 MEA-2075. 312 tt 026 mechanism solutions to ground geo,ogy. Finalty, the distnbut;on This program consists of research and engineenng refahng to fracture, fatigue and radiation sensitiety of nuclear structuraf of earthquakes and their relatonships to geology is reviewed
Main Citations and Abstracts 31 steels and weldments and addresses many of the key uncer-tainties in the margin of safety in operating nuclear plants. All NUREG/CR-3293 V02: TECHNOLOGY. SAFETY AND COSTS OF DECOMMISSION!NG REFERENCE FUEL CYCLE AND NON. tasks are integrated to focus on structuralintegnty of LWR pres-FUEL CYCLE FACILITIES FOLLOWING POSTULATED sure boundary components. The approach centers on an exper. ACCIDENTS. Appendices. ELDER,H K. Battelle Memonal Insh-imental charactenzation of nuclear grade steels and an assess- tute, Pacific Normwest Laboratones. May 1985. 288pp. ment of fracture and environmental cracking behaver under 8506170550.30979 001. conditions of a nuclear environment, so investigation of irradiat- This volume contains the appendices concerning the techni-ed matenals is a key element of each task. Emphasis is placed cal requirements, costs and safety aspects conceptually evalu-on identifying metallurgical factors responsible for radiation em- ated for post-accident cicanup and decommissioning of fuel bnttlement of steels and on developing procedures for embnttle. Cycle and non-fuel cycle facdities that have exper'enced a ag-ment relief, including guidelines for radiaton-resistant steels. Ex- " cant accient. Accident cleanup is postulated to include 1) penmental studies are supported by anatyhcal models and in- initial @contaminaten of building surfaces to reduce the subse-vestigatons of the mechanisms responsable for the observed quent occupational dose to cleanup and decommissoning work-behaver. Data developed 6n the program will provide the basis # # " 0 *#
for recommendations for the ASME Boiler and Pressure Vessel ing is assumed to follow accident cleanup In order to ensure Code and ASTM test methods, and revisions to NRC Guides that worker doses are ALARA. despite higher radiation expo.
Current work is organized into three major tasks. (1) fracture sure to workers dunng post-accident operations, careful plan-mechanics investigatons, (2) environmentally-assisted crack ning and rehearsal of cleanup operations and the use of remote growth in high temperature, pnmary reactor water and (3) radi- and semi-remote cleaning techniques are required to reduce oc-ation sensitivity and postirradiauon properties recovery. Re- cupancy times in high-radiation areas and to minimize occupa-search progress in these three tasks for 1984 is summanted tional exposures dunng cleanup The public safety impacts of here, post accident cleanup and decommissoning are also evaluated, these are below permissible radiaton dose levels in unrestncted areas and well within the range of annual radiaton doses from NUREG/CR 323h CONTROL OF EXPLOSIVE MIXTURES IN normal background PWR WASTE GAS SYSTEMS. RANDOLPH P D.; ISAACSON.L; AYERS.A L; et al. EGAG. Inc. January 30, 19d5. 122pp NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT 8502130454. EGG-2251. 28914 224. SEQUENCE INFORMAT'ON. CATHEY,N G ; KR ANTZ.E. A ; A study has been performed to evaluate problems associated POLOSKI.J P; et al. EG&G ldaho, Inc. (subs of EG8G, Inc.) with the existence of flammable or explosrve gas mixtures in August 1985 ?9pp. 8509130115 EGG 2259. 32605 2b4. Pressunted Water Reactor waste gas systems. Information on Infonnaten L *cenmg me donmant acceent sequees existing waste gas systems, waste gas concentratons, and gas m ee ed probaMsk n% assessments WA) is monitonng instrumentaten obtained from six operating nuclear cataloged in this re;at, which is published as a part of the Ac. power plants is summanted. A comparatrve nsk evaluahon has cident Sequence Evaluation Program (ASEP) The purpose of been performed for several genene types and configuratons of this report is to provide users of PRA informaton a single refer. PWR waste gas systems. Waste gas systems in the plants vis- ence document. The cataloged resutts include plant operation sted are included and categonted as part of the nsk evaluation. information, core. melt and sequence frequencies, and a de-Existing data on the effset of initial pressure on flammabdity senption of each dominant accident sequence. The report pro. vides a consistent set of insights on the factors that dnve the limits, as well as recently reported data on flammabihty and de-tonabihty of hydrogen / air mixture has been collected and sum-dominant accident sequences. ASEP has reconstructed the PRA fault tree models at the system or train level of detail and manzed. A survey of commercially avadable instruments for requantified the sequence hkehhoods to provide the consistent monstonng hydrogen and oxygen concentratons has been per- insights. This work provides the informaton for the other ASEP formed and the results tabulated. A senes of observations, con- activities on accident hkehnood assessment for the operabng clusions and recommendations are given. and near term operating plants NUREG/CR-3293 V01: TECHNOLOGY, SAFETY AND COSTS OF NUREG/CR 33th TECHNICAL BASES AND USER'S MANUAL DECOMMISSIONING REFERENCE NUCLEAR FUEL CYCLE FOR THE PROTOTYPE OF SPARC - A SUPPRESSION POOL AND NON-FUEL CYCLE FACILITCS FOLLOWING POSTULAT. AEROSOL REMOVAL CODE. OWCZARSKI,P.C.; POSTMA,A K.; ED ACCIDENTS Vrn Report. ELDER H K. Battelle Memonal In_ SCHRECK.R.l. Battelle Memorial Institute, Pacific Northwest stitute, Pacific Northwest Laboratones. May 1985. 327pp- Laboratones. May 1985 68pp. 8506240650. PNL-4 742. 8506140337, 30932.308. 31152.156 Technical requirements. costs and safety are conceptuaffy a* Mwest Lawam hat WW a pdow evaluated for the post-accident cleanup and decommissioning version cf 8 SUPpresson Pool Aerosol Removal Code (SPARC) of fuel cycle and non-fuel cycle facdities that have expenenced ' * " "' "' * " ' a significant accident. Accident cleanup is poslulated to include clos in the pressure suppression pool (wet well) of a bothng
1) initial decontamination of building surfaces to reduce the sub- water reactor under hypothetical accident conditions The code sequent occupabonal dose to cleanup and decommissionin9 incorporates five aerosol scrubt':ng models and two thermal hy-workers and 2) management of the resulting wastes. Decom- draube models. The scrubbing models describe 1) steam con-missioning is assJmed to follow accident c'eanup. In order to densation,2) soluble partscle growth in a humid atmosphere. 3) gravitatonal setthng, 4) inertial deposition. 5) diffusenal deposi-cnsure that worker doses are ALARA, despite higher radiaton ton. Mechanical entrainment of pool hquid by breaking of bub-Exposure to workers dunng post accident opera 00ns, careful bles at the surface was also considered. An optional model for planning and rehearsal of cleanup operahons and the use of equdibnum pool temperature and a model for steam evaporation remote and semi-remote cleaning techniques are required to are the two thermal hydrauhc models used in the code Steam reduce occupancy times in high-radiaton areas and to minimite evaporaton was found to significantry retard deposition proc.
occupational exposures dunng accident cleanup. The pubhc esses in pools near the borbng point The code user supplies safety impacts of post accsdent cleanup and decommissoning the values of several controlhng vanables in the code input. The tre also evaluated, these are below permissible radiation dose SPARC output can include the decontam#nat on factors (DF) of kvels in unrestncted areas and well within the range of annual twenty different particle site groups,'an overall DF for the whole rcdiation doses from normal background. part:cle distrbution, particle log normal distnbuton parameters, and mass flow rates of parbcles (wet and dry) leaving the pool
32 Main Citations and Abstracts NUREG/CR 3319: LWR PRESSURE VESSEL SURVEILLANCE losses, the quahty of various typos of labor and equipment nec-DOSIMETRY IMPROVEMENT PROGRAM LWR Power Reactor essary to complete the decontamination. dose to radiation work. Surveillance Physics-Dosimetry Data Base Compendium. ers, the costs for surveying and monitonng activities, and the MCELROY,W.N. Hanford Engineenng Development Laboratory. disposal costs associated with radiological waste gerierated August 1985.533pp.8509110278. HEDL TME-85-3. 32563.020. dunng cleanup The program and data base are demonstrated This NRC physics-dosimetry compendium (Sections 1.0 with a decontamination analysis of a hypothetical site through 4.0) is a collation of information and data developed from available research and commercial fight water reactor NUREG/CR-3426 V01: THERMAL AND FLUID MIXlNG IN 1/2 SCALE TEST FACILITY, Facihty And Test Design Report vessel survedlance program (RVSP) documents and related sur. DOLAN,F.X.; VALENZUELA.J A Creare. Inc. September 1985 veillance capsule reports. The Section 4 0 data represents the results of the HEDL least-squares FERRET SAND fl Code re. 130pp.8510020227. EPRI NP 3802. 32839 028. This report descnbes the test facility and program designed to evaluation of exposure units and values for 47 PWR and BWR rneasure fluid mmng and heat transfer in a 1/2-scale model uf surveillance capsules. Using a consistent set of aux *ary data and dosimetry adjusted reactor physics results, the revised the cold-leg downcomer and lower plenum of a pressunzed ; water reactor under conditions of interest to the issues of pres- i fluence values for E > 1 MeV averaged 25*. higher than the sunzed thermal shock Several cold-leg assemhies are modeled onginally reported values. The range of fluence vafues (new/ and the downeomer arrangement can be a4ered to match old) was from a low of 0 80 to a hegn of 2.38. These HEDL-de. nved FERRET-SAND 11 exposure parameter va!ue are being vendor specific configurations. The facihty can be operated to used for NRC supported HEDL and other PWR anu 3WR trend model flow rates based on Froude number of the s'ilected flow curve data development and testing stud es, which support Re. in the cold leg and with steady or transient inlet boundary cond-tions. Entensive instrumentation rs provided to measure flow vision 2 of Regulatory Guide 199 These trend curves are used rates, temperatures and pressure at the facility boundaries and by the utilities and by the NRC to account for neutron radiation for detailed measurements of temperatures, velocity and heat damage in settng pressure / temperature limits, in analysing trac. tures, and in predicting neutron-induced changes in reactor PV transfer data in the cold-leg and downcomer rnodels The test steel fracture toughness and embrittlement dunng the vessel's data are monitored and recorded by a computer data acquis#- service hfe. The status of the development and apphcation of tion system that is also used for post test reduction and plot-new advancements in LWR reactor surveillance programs is dis- ting The planned test matnx includes 75 tests with vanations in cussed, such as cavity physics-dosimetry for improving the reir cold-leg and downcomer geometnes, loop and HPI flow rates. abihty of current and end-of-life (EOL) predictions on the metal- cold-leg Froude number and loop to HPl density difference. lurgical condtions of pressure vessels and their support struc- Test results will be reported in a senes of Quick-Look Reports
#** NUREG/CR-3426 V02: THERMAL AND FLUID MIXING IN 1/2-NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON SCALE TEST FACILITY Data Report. VALENZUELA.J A ;
THE RATES OF HYDROGEN GENERATION FROM GALVA- DOLAN.FX Creare, Inc. September 1985 208pp.8510020217. NIZED STEEL CORROSION AT POST-LOCA CONDITIONS EPRI NP-3802 32838 069 LOYOLA.V.M WOMELSDUFF.J E. Sanda National Laborato- This report presents data frorn an expenmental study of f!ud nes January 1985 40pp. 8502130366 SAND 831326 mining in a 1/2 scate model of the cold leg, downcomer, lower 28920 328 plenum, pump simulator. and loop seal typical of a Westing-The rates of hydrogen generation are measured for the corro- house Pressunzod Water Reactor. The tests were transient son of galvanized steel in three efferent light water cooled re* cooldown tests in that they simulated an estreme condition of actor (LWR) water chemistnes. The results were obtained over Small-Break Loss-of-Coolant Accident (SDLOCA) dunng which a temperature range of 100 degrees to 175 degrees centsgrade cold High Pressure injection (HPt) fluid as inlected into stagnant, and indicate that in a boshng water reactor (BWR) water chemis- bot onmary fluid with complete loss of natural circulation in the try, the reaction is f aster than in those of two pressunzed wate' loop. Estensive temperature, velocity, and beat transfer coeffi. reactors (PWR's) A mechanism is proposed which would ex- c,ent data are presented at two cold leg Froude numbers 0 052 plain the observed results without requinng that the chemical and 0 076 The 1/2-scale data are compared with earlier data adetwes come in direct centact with the corrodible unoxidzed from a 1/5 scale.geometncally similar facihty to assess scahng metal. Such a mechanism is required because electron micro- pnnciples probe anatysis suggests that no chemical addtives have ef-fused into the protective ZnO layer which forms on the unomi- NUREG/CR-3430 V02: NUCLEAR POWER PLANT OPERATJNG dized metal. Arrhenius parameters are calculated for the three EXPERIENCE - 1982 Annual Report SILVER E.G Oak Ridge chemistnes, but some questions are raised about whether those National Laboratory January 1985 393pp- 8502150078. parameters are associated with a d ffusion process of wrth the 29004 010. This report is the ninth in a senes of reports issued annually actual hydrogen producing reaction. that summantes the operating espenence of nuclear powa NUREG/CR 3413: OFF SITE CONSEQUENCES OF RADIOLOGi- * ' " ' * * * ' ' "" * " '" *"G" CAL ACCIDENTS METHODS. COSTS AND SCHEDULES FOR "" " '"' 9" '
~~C C""*"C * * * *'~~
DECONTAMINATION TAWIL,J'J : BOLD.F.C,; HARRER.B J.; et nt Amance, aN occupaWat radaten oposum fu af. Battelle Memonal Institute Pacific Northwest Laboratones each plant are presented and discussed, and summary high. lights are given. The report includes 1982 data from 72 plants Au9ust 1985. This report 379PP 8509110274 documents PNL-4790. a data base and a computer32562 pr 001' ogram 24 Miling-watnr-reactor plants. 47 pressunzed water-reactor for conducting a decontamination analysis of a large, radiologi- p! ants, and I high temperature Gas. cooled reactor plant. cally contaminated area The data base, which was compiled largely through interviews with knowledgeable persons both in NUREG/CR 3442: RADTWO A COMPUTER CODE FOR SIMU-the pubhc and pnvate sectors, consists of the costs, phys 6 cal LATING FAST TRANSIENT, TWO-DIMENSIONAL,TWO LAYER inputs, rates and contaminant removal efficiencies of a large RADIONUCLlDE CONCEN TRATION CONDITIONS IN number of decontaminaten procedures. The computer program LAKE S.RE SE RVOIRS, RIVERS,E ST U ARIE S. AND COASTAL utilizes this data base along with information specific to the con- REGIONS ERASLAN,A H ; DIAMENT H. Oak Ridge National taminated site to provide detailed information that includes the Laboratory. July 1985 444pp 8509180502. ORNL/TM 8869 least costfy method for effectively decontaminating each sur- 32666 013 face at the site, vanous types of property losses associated with RADTWO is a computer codo for predicting the trans><?nt. the contaminaton, the time at which each subarea within the two d mensional trar, sport of radonuclides in receiving water site should be decontaminated to minimite these property bodes The modet formulation consHjers two coupled, depth-
Main Citations and Abstracts 33 averaged transport equat;ons for the water layer and tne bottom measurement tests and human factors pnnciples This report sediment layer. The coupling conditions incorporate bottom descnbes the development of TAPS and presents its first dem-deposition and resuspension effects. The computer code uses a onstration. It begins with characteristics of skilled human per-discrete-element method which offers variable size gnd cells. formance and proceeds to postulate a cognitive model to for-accurate shorehne representation, and numencal accuracy. A ma;fy desenbe these charactenstics A rnethod is denwd for sample apphcation is provided for the problem of a hypothetical linking SKAA charactenstes to measurement tests The entire accidental release of radionuchdes to the coastal environment. process is then automated in the form of a task ana'ysis com-Results are presented as contours of constant radionuchde con-puter program. The development of the program is deta' led and centratron in the water layer and the bottom sediment layer at vanous times dunng the model simulation penod. a user guide w'th annotated code listings and support:ng test in-formation is provided. NUREG/CR 3444 V02: THE IMPACT OF LWR DECONTAMiNA-TlONS ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCI. NUREG/CR 3485: PRA REVIEW MANUAL EL-BASSIONLA ; ATED OCCUPATIONAL EXPOSURE. D AVIS.M.S ; CHO.N Z.; HANAN.N ; et al Brookhaven National Laboratory PICIULO.P.L; BOWERMAN.B S.; et al. Brookhaven National September 1985. 2pp 8509230653 BNL-NUREG.51710 Laboratory. July 1985 102pp. 8507250157. BNL NUREG- 32702 019 51699. 31794.123. This PRA Review Manual descr,bes the approach for review-This report descr bes work conducted by BNL on the degra- ing a Level 1 PRA, te, one which carnes the accident analysis dation of simulator chemical decontamination wastes by com. to the point of determinatron of core damage frequency. but en-bustion and acid digestion. Both acid digestion and combustron cludes questions of conta<nment inlegnty (but does include con-are capable of effecting 90*. destruction cf the matenals stud- tainment failure induced core damage) and of offtte conse. ed. as measured by the conversion of carbon compounds in quences. The manual *di be revised as comments are received, the waste to carbon dioxide. Work on the direct sohd.fication of and as experience is ga'ned from its use The procedure in-simulated decontamination wastes in cement and vinyl ester- voNes three parts The first (Phase 1) is concerned with the styrene is reported also. Laboratory Scale waste forms were formal aspes-ts of the PRA Phase 1 sunseys its apparent com-prepared using these binders. However, process control pro- p;eteness, scrutabihty, and determines to what entent the PRA grams and fulf sca!e sohdrficat on studies are necessary to con- can usefully be fer*ner examined it a!$o identifies sahent and firm the acceptability of the wastes-distanctive features. of the study, methods. and reported resu'ts NUREG/CR 3455; A COMPARISON OF IODINEMRYPTON AND The second part (Phase 2) reviews the anatyses in a correre-XENON RETENTION EFFICIENCIES FOR VARIOUS SILVER hensive and thorough but quaktative way. which is des gned to LOADED ADSORPTION MEDIA. HUCHTON,R L.; focus cn unusual or unsupported features. and to lay the TKACHYK.J W.; TAYLOR.J T : et al. Westinghouse Electnc groundwork for further, more detailert stud es The final stage Corp. Apnl 1985. 80pp. 8505230585 WINCO-1024 30546 254 (Phase 3) addresses detads of issues and concerns ra sed in A companson was made among var.ous sdver impregnated the earber phases, and involves detaded quant,tative enamna-adsorption med a to determine their iodine, krypton, and renon tion of selected areas to ensure the overall vahd,ty. The trst retent.on ef%cas The p*cy ennostad nf three compo. part of this manual, deakng w.th " internal" event PRAs. hand,es nents. First, laboratory measurements of the noble gas retent:en these phases sequent.a'ly as a whole in the second part. whicn efficiencies of commercially available adsorption med a were treate, "enternal'* events, the phases are vient,f ed w4trin each determined as a function of relative humidity, sample duration, event section, while Chapter 9 g'ves a sequent.af surr mary of test cartndge geometry, and arrbient air purge Second. a htera. the end resutts for each event ture survey was performed to evaluate the iodine species reten-tion efficiencies of the selected media. Third, data associated NUREG/CR 3488 V03; IDAHO FIELD EXPERIMENT 1981 Volurre with a med,a previously proposed for an emergency response 3 Companson Of Traiectones Concentration Patterns And YE-air sampler were incorporated to enlarge the data base SODIF Model Calculations START.G E.; CATE J H; SAGENDORF.J F : et al. Commerce. Dept of. Natt Oceano-NUREG/CR 3469 V02: OCCUPATIONAL DOSE REDUCTION AT NUCLEAR POWER PLANTS. Annotated 04bhography Of Select- graphic & Atmosphenc Administration. February 1985 750p ed Readir*gs in Radiaton Protection And ALARA B AUM.J W ; 8503120064 29340 116 WEILANDICS C. Brochhaven National Laboratory. June 1985 The 1981 Idaho Field Espenment was conducted in south. 150PP. 8507020380 BNL NUREG 51708. 31306 251. eastern Idaho over the Upper Snake Rever Plain. N ne test day This is the second volume of abstracts deakng with occupa. case studies were measured between July 15 and 30,1981 tsonal dose, dose control, dose reduct on and apphcation of the Eight hour releases of SF(6) gaseous tracer were made from 46 ALARA (as low as reasonably achievable) pnnctpfe at nuclear m above the ground Tracer was sampred houriy. for 12 se-power prants This volume conta,ns abstracts selected from AP. quantial hours at about 100 locatons within an area 24 km PLIED HEALTH PHYSICS ABSTRACTS AND NOTES, Volumes square Also, a single total integrated sampie of about 30 hours 1, No.1,1975 through Volunw 5, No 4. October 1979. and duration was collected at approximately 100 s,tes witnen an area from recent pubhcat.ons known to the authors Author and sub' 48 by 72 km (usng 6 km spacings) Extensive tower profJes of ject indenes are included The subject ender in this volume meteorology at the release point were coliected RAWIN-covers abstracts in both Volumes 1 and 2 This volume conta.ns SONDES. RADALS and PIBALS were collected at 3 to 5 sites abstract Numbers 252 through 549 Honzental, low altitude winds were monitored utng the INEL MESONET. SF(6) tracer plumes were marked with co-located NUREG/C43481 V02: NUCLEAR POWER PLANT PERSONNEL od fog releases and behourly sequentA launches of tetroon QUALIFICATIONS AND TRAINING TAPS - The Task Analysis parrs Aonai LIDAR observatons of the oil fog plume and air. Profihng System. JORGENSEN.C C. Oak Rdge National Labo- borne samples of SF(6) were collected High altifade senal pho-rctory. July 1985 24fipp, 8508090705. ORNL/TM 9308/V2 tographs of daytime piumes were also collected The Idaho 32104 128 held Espenment is reported in three volumes Volume 3 con. This report discusses an automated task analyss profiling tains descriptions of the nine intensve measurement days. Gen-system (TAPS) desgred to provide a knkir'g tool between the eral meteurolorycal conditions are described, traiectones and behaviors of nuclear power plant operators en performing their their relatonsheps to analyses of gaseous tracer data are dis-t:sks and the measurereent tools necessary to evaluate their in- Cussed. and overviews of test day cases are presented Calcu. plant performance. TAPS assists in the identificat<on of the fations utng the ARLF RD MESODIF model are included and re-ently level skill, knowledge, abikty and att,tude (SKAA) requsre- lated to the gaseous tracer data FinaHy, a summary and hst of ments for the vanous tasks and rapidfy associates them with recommendations are presented 1 - . - . , . - -., , _ _ _ _ _ _~- - .- m -__ -__,...__m . . . . _ . _ _ _ . - - , - - - . . _ _
~~- - - - , y--.___, _ v. - - - - , - ,.~~
34 Main Citations and Abstracts NUREG/CR-3498: TWO DIMENSIONAL MODELING OF INTRA- NUREG/CR 3519: HUMAN ERROR PROBABILITY ESTIMATION USING LICENSEE EVENT REPORTS VOSKA.K J ; SUBASSEMBLY HEAT TRANSFER AND BUOYANCY IN- O' BRIE 1J N Brockhaven Nationai Laboratory. February 1985 DUCED FLOW REDISTRIBUTION IN LMFBRS. KHATIB- 116pp 8502220414. BNL NUREG 517f 7. 29062.f 63. RAHBAR, CAZZOLI.E.G. Brookhaven National Laboratory. Jan-179pp. 8501210102. BNL NUREG 51713 The obi ective of this report is to present a method for using uary 1985. field data from nuclear power plants to estimate human error 28496.154 Phenomenological models and nomencal techniques for pre- probabihties (HEPst These HEPs are then used in probabihstic diction of coolant flow and temperature fields dunnq forced, nsk actwities. This n(thod of estimating HEPs is one of four mued, and free convecten regimes of operation in LMFBR sub- being pursued in NRC-soonsored research The other three are assembhes are addressed. It is shown that simphfied integral (1) structured empert judgment. (2) analysis of training simulator data. and (3) performant.e modehng The type of field data ana-solutions provide an excellent approach to assessing the impor-tance of the intra-subassembly buoyancy induced flow redistn- fyred en thrs report as from Licensee Event Reports (LERs) which are anatyred using a method specifically developed for bution, and the transverse thermal conduction and mixing et-that purpose However, any type of field data or human errors fects on the assembly wide peak coolant temperatures Further- ' could be anafyrted using this method with minor adjustments. more, a more detarled steady state and transient parabohc two- This report assesses the practicahty. acceptabihty, and useful dimensonal porous body model resulting #n the TWIST comput. ness of estimat.ng HEPs from LERe and comprehensively pre-er code is developed. Comparison of calculated results and out-sents the method of use of-pde sodium and water test data indicate generafly good agreement in cross-assembly temperature profdes However, NUREG/CR 3537: EXPEDIENT METHODS OF RESPIRATORY the impact of fuel pin distoroon and bowing caused by large PROTECTION lit SUBYlCRON PARTICLE TESTS AND SUM-transverse power gradients on transverse distributions are found MARY OF QUALITY FACTORS PRICE.J M ; COOPER,0 W ; to be significant. YEE.C S : et al. Sand'a National Laboratories September 1985. 95pp 8509260257. SAND 83 7450. 32757.118. NUREG/CR-3514 V02: THE CHEM: CAL BEHAvtOR OF f00lNE The erficacy of readily ava table mater,ats, such as cotton fab-IN AQUEOUS SOLUTIONS UP TO 150 C.li Radiation Redox ncs. towehng a surgical mast and a singte use respirator, for Conditions TOTH.L M.; DODSON.K E. Oak Ridge Natenal Lab- providing emergency respiratory protection was evaluated by oratory Apnf 1985. 22pp. 8506100496 ORNL/TM-8664/V2 determing the filtration efficiency as a function of aerosol parts-30830.089. cle size over the size range of 0 001 to 5 0 mm and as a func-Redox reactions that rnight after the volatility of aqueous ton of fdtraton face velocity Filtration face velocity was set at iodine solutions have been examined espenmentally using ab. 15. 5 0. and 15 0 cm/s. This report describes the equipment sorption spectropnotometry. Oxygen and hydrogen atmospheres and procedures used to obtain efficiency measurements for par-had no effect on the cdine chemistry at temperatures up to 150 ticles 0 5 mm in diameter and smalier, and summanzes the re-degrees centigrade However, irradiation of aqueous solutions suits of all three phases of this research. Particles with diame. with a (60)Co source. 0 8 x 10(8) R/h. produced radiolysis prod- ters from 0.10 to 0 50 mm proved to be the most difficult sizes ucts that either oudized ed.no son or reduced 10(3). in the pH of particles to remcve Particles smaller than 010 mm were re. range 6-9 and generated $sgnificant amounts of volatde iodine moved due to diffuson while particles larger than 0 50 mm were The amount of iodine volatihred vaned from a few percent for removed due to inertia and gravitational setthng Deposition of solute concentrations of 10(-4) M to as much as 10 to 19*. for me smanest particles was favored by the use of low face veloco-10(.6) M Cat or K!0(3) solutes SJver metal has been shown to ties A fracticnal efficiency curve was determined for each ma-provide an effectue gettenng route for I. in solution af these ons Wal at ead Woc 4 W cynpanson Vabes of M quaMy are first oxidized by OH rad'cafs generated dunng the radiofys's a , nevaW@psn M wm cakuW of the solutions ~ Ouahty factors were less for wet matenats than for dry, less at high velocities rather than low, and best for the singte-uso respe-NUREG/CR 3516: A SURVEY OF THE USES OF RADIOACTIVE rator mask, next best for the surgical mask and often third best MATERIALS IN LOUISI AN A'S CFFSHORE WATERS for the towehng BENNETT J J.; HOOK.S E4 PALAZZO.R J ; of al. Louisiana, State of February 1985. 37pp. 8503130140 29360122. NUREG/CR 3551: SAFETY IMPLICATIONS ASSOCIATED WITH As a result of a contract agreement with the U S Nuclear IN PLANT PRESSURIZED GAS STOAAGE AND DISTRIOU. Regulatory Commission, the State of Louisiana, and in particu- TION SYSTEMS IN NUCLEAR POWER PLANTS lar, the Louisiana Nuclear Energy Duisson (LNED), conducted a GUYMON R H ; CASTO.W R : COMPERE.E L. Oak Ridge Na-survey of the use of radioactive matenals in Louis:ana's "off- tional Laboratory May 1985 82pp 8506f 40622 ORNL/NOAC-shore waters." Offshore waters are here defined as "that area 214. 30934.031. of land and water on and above the United States' Outer Conti. Storage and handhog of compressed gases at nuclear power nental Shelf " The obiectives of the survey were fourfold 1) plants were studied to identify any potential safety hazards. adentification of those hcensees using radioactive matenais off. Gases investigated were air, acetylene. carbon dionde, chlorine, shore Louisiana, 2) Identification of work locations where radio- Halon, hydrogen, nitrogen, osygen, propane, and sulfur hena-actue matenals are being used,3) a descnption of the types of fluonde Physical properties of the gases were reviewed as work performed, and 4) performance of at least three site visits were appbcable industnal codes and standards Incidents involv. to offshore locations where radcactue matenals are being ing pressunted gases in general industry arvi in the nuctear in-used. By telephone survey. LNED attempted to contact those dustry were studied in this report general hazards such as m s-licensees thought to be using radioactive matenals offshore. Of sales from ruptures, rocketing of cyhnders, pipe wh<pping, as-the 69 hcensees reached by telephone, 43. or 61*., indicated phyniation, and toxicity are discussed Even though some sen-ous injunes and deaths over the years have occurred in indus-they have current offshore activities. The results of the tele- tnes handkng and using pressunted gases, the industnal codes. phone survey, conducted in May June 1983 are presented in standards, practices, and procedures are very comprehensive detail in this report To meet objectue four of the survey, three The most important safety consideration in handbng gases is visits were made to offshore ogs and platforms, two involving industnal radiography and one envofving welllogging Also in- the sorcus enforcement of these well known and estabbshed methods Recommendations are made concerning compressed cluded in this report are summanes of these visits and a de- gas cyhnder rnissdes. hydrogen hne ruptures or leab s. and den-acnption of previous work done by LNED concerning radiation tification of knes and equipment safety on " fay barges " i
~~. _ ~ -~~
Main Citations and Abstracts 35 NUREG/CR 3558: HANDBOOK OF NUCLEAR POWER PLANT actor (LWR) service Vahdated models, based on expenmental SEISMIC FRAG 1LITIES Se.smic Safety Margins Research Pro- data, will be developed to predict the degree of sensitization
~~gram.. COVER.L E., BOHN.M P.; CAMPBELL.R D ; et al. Law-~~
' (DOS) and the intergranular stress corrosen cracking (IGSCC) rence Livermore Nat onal Laboratory. June 1985 321pp. susceptibikty in the heat affected zone (HAZ) of the SS weld-8507080210. UCRL 53455. 31402.238. ments IGSCC is caused by a combination of a sensitized mi-The Seismic Salery Margins Research Program (SSMAP) is crostructure, an aggressive environment, and tensile stress an NRC-funded, multyear program conducted by Lawrence Control of any of these three factors can ehminate IGSCC in Livermora National Laboratory (LLNL). Its goal is to develop a I most practical situatens. This program will measure and model complete and fully-coupled analyms procedn, including meth* the development of a sens.tized microstructure as it pertains to ods and computer codes, for estimating the risk of earthquake- welded and repair welded SS pipe An empincal correlation be-induced radcactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic nsk tween a matenal's DOS and its susceptibety to IGSCC will be from a typical commercial nuclear reactor were made These determined using constant extension rate tests (CERTs). The successful completion of these tasks wal result in a method for ! calculations required a knowledge of tne probabikty of failure (fragibty) of safety-related components en the reactor system assessing the effects of welding /reparnng parameters on the that cctively participate in the hypothesized accident scenanos' IGSCC susceptibihty of component. specific nuclear reactor welds / repairs. This report describes the development of the required frag.I,ty relations and the data sources and data reduction techniques upon which they are based. Bcth building and component fragi- NUREG/CR 3613 V03 N1: EVALUATION OF WELDED AND REPAIR. WELDED STA'.NLESS STEEL FOR LWR lit.es are covered The building fragilities are for the Zion Unit i reactor, the specific plant used for development of methodology SERVICE. Semiannual Report For October 1984 Through March in the program. Some of the component fragMties are site-spe- 1985 ATTERIDGE.D G; CHARLOT.L A ; BRUEMMER,S M ; et cific, but most would be usable for other sites as well- al. Battene Memonal Institute. Pacific Northwest Laboratones September 1985 59pp 8510040360 PNL 4941 32856184 NUREG/CR 3609: EVALUATION OF NEUTRON DOSIMETRY Pacific Northwest Laboratory, under the soonsorship of the TECHNIQUES FOR WELL LOGGING OPERATIONS Divisen of Engineenng Technology of the U S Nuc! car Regula-CUMM:NGS.F M ; HAGGARD,0.L.; ENDRES.G W. Battelle Me- tory Commission. is conductsng a program to determine a monal Institute. Pacific Northwest Laboratones. July 1985. Sipp method for evaluating welded and repair-welded stainless steel 8500010304 PNL-4942. 31928170 tSS) piping for hght water reactor service Vabdated models, Neutron dose and energy spectral measurements from (241) based on experimental data. are being developed to pred.ct mi. AmBe and a 14 MeV neutron generator were performed at a crostructural development (e g , the degree of sensit.zation) and well-logging laboratory The measurement technique included the stress-corrosion cracking (SCC) resistance in the heat-af-the tissue equrvaient proportonal counter, multisphere, two fected zone of the SS weldments Stress-corrosen cracking is types of remmeters and five types of personnel neutron dos'- caused by a combination of a suscept ble microstructure, an al meters Several source configuratons were used to attempt to gressive environment, and tensile stress Control of any of relate data to field situatons. The resutts of the measurements these three factors can ehminate SCC in most practical situa. indicated that the thermoluminescent albedo dosimeter was the tons This program will measure and model the development of most appropr ate personnel neutron dosimeter, and that the most appropnate cahbraton source would be the source nor- a susceptible microstructure as it pertains to welded and repair. welded SS pepe Empincal correlatons between matenal micros-rnaHy employed in the feld with the cakbration source being tructure and SCC wol be determined using constant entension used in the unmoderated conhguration. rate tests. The successful completion of these tasks will result NUREG/CR 3611: RADIOACTIVE MATERIAL (RAM) ACCIDENT / in a method for assess.ng the effects of welding / repairing pa-INCIDENT DATA ANALYSIS PROGRAM. EMERSON.E L ; rameters on the SCC resstance of component specific nuclear MCCLURE,J D. Sandia Natonal Laboratones Apnl 1985 40pp reactof welds / repairs The present report desenbes the 8504220385 SAND 82 2156 29946 323 progress of these studies dunng the first haff of the 1985 fiscal This report desenbes the development of the Radioactwo Ma- year tenafs Transportaten Accdent/ Incident Data Base (RAM-AIDB), which contains information on the occurrences of transportaten NUREG/CR 3626 V0a MAINTENANCE PERSONNE L PER, accidents and incidents, for radioactive matertals (RAM) that are FORMANCE SIMULnTION tMAPPS) MODEL. DESCRIPTION involved in the process of transportation, loading and unloading OF MODEL CONTENT, STRUCTURE.AND SENSITIVITY TEST-operations, or temporary storage These transportaten oper- ING SIEGEL.A t; BARTTER.W D; WOLF.JJ; et al Oak Ridge atens are in support of the nuclear fuel cycle for electncal National Laboratory Apnl 1985 322pp 8504170234 ORNL/ energy generations of RAM This study anatyres in some detail TM-9041/V2. 29902 002 basic accident / incident statistical data, RAM packaging acci- This volume of NUREG/CR 3626 presents details of the con-dent response data, and the heatth effects associated with tent, structure, and sensitwity testing of the Maintenance Pet. RAM transport accidents / incidents This report presents a sum- sonnel Performance Simulaton (MAPPS) model that was de-mary of U S. RAM trentport accdent/incdent empenence for scobed in summary in volume one of this report The MAPPS the pered 1971 through December 1981. In addition, a sample model is a generabred stochastic computer simulaton model annual summary of accident / incident espenence is presented developed to simulate the performance of marntenance person. for the calendar year 1981. nel en nuclear power plants The MAPPS model consdors work-NUREG/CR-3613 V02: EVALUATION OF WELDED AND place, maintenance technician, motivation human factors and REPAIR WELDED STAINLESS STEEL FOR LWR task onented vanables to yield predictive information about the SERVICE Annual Repcrt for 1984 ATTERIDGE,D G ; effects of these vanabics on successfut maintenace task per-BRUEMMER.S M ; PAGE,a E. Battelle Memonal Institute, Pac,t. formance All major model variables are d.scussed in detail and ic Northwest Laboratones Jur a 1985 63pp 8506270333 PNL. their implementation and interactive effects are outhned Tho 4971. 31262 215. model was erarnined for disquahtying defects from a number of Pacific Northwest Laboratory (PNL). under a program spon- viewpoints, including sens,tivity terting This esamination led to sored by the Division of Engineenng Technology of the O S. Nu- the identification of some rninor recahbration efforts which were clear Reg slatory Commission (NAC; is conducting a program to carned out. These positive results indicate that MAPPS is ready determine a method for evaluating tre acceptance of welded for initial and controlled apphcations which are in conformity and repair weided sta,nless steel (SS) piping for hght water re- with its purposes
36 Main Citations and Abstracts NUREG/CR-3633 V01 S1: TRAC BD1/ MODI.AN ADVANCED ed. Section 3 provides the detailed guidance to users for inter-BEST ESTIMATE COMPUTER PROGRAM FOR BO! LING act.ng with MAPPS via a terminal. WATER REACTOR TRANSIENT ANALYSlS TAYLOR.D D ; i SHUMWAY,R W; SINGER.G L ; et al EG&G Idaho. Inc. (subs NUREG/CR 3636: HYDROGEN STEAM JET-FLAME FACILITY of EGAG, inc). September 1985.122pp 8510040411. EGG- AND EXPERIMENTS SHEPHERD.J E Sand.a National Labora-2294. 32855 116. tones. Jufy 1985. 138pp 8508010764. SAND 84-0060. The TRAC-BD1/ MODI computer program provides a best es- 31925:104. timate anatysis capab:hty for the analysis of the full range of As part of NRC sponsored research on light-water reactor postulated accidents in Boehng Water Reactor (BWR) systems safety, the high-temperature combustion of steam-hydrogen gets l and related expenmental facihties The program es described in in an air atmosphere is being investigated at Sandia This re- i four volumes. Volume 1. Code Desenption, Volume 2. User's search is oriented at understanding the genonc issues involved Gu'de; Vofume 3. Code Structure and Programming Information. in accident-generated lets and the specific problems of using and Volume 4, Developmental Assessment. Volume 1 desenbes del. berate flanng from high-point vents to ehminate hydrogen the thermal-hydrauhc models, numencal mehtods, and compo- from the primary system in this report we give some back-nent models available Volume 2 descobes the input and output ground on d,ffusion flame combustion, descnbe the expenmen-of the TRAC BD1/ MOD 1 code and provides guidelines for use tal facility constructed at Sandia to study high. temperature, of the code modeling of BWR systems. Volume 3 #s dessgned steam-hydrogen jets and descuss our results for the programmer or model developer who needs to imple-ment or modify the TRAC-BD1/ MODI program, Volume 4 dis- NUREG/CR 3646: TR AC-PF1 INDEPENDENT ASSESSMENT. cusses the results of the development assessment calculations KNIGHT,T D, BOOKER C P.; BOYACK.B E.; et al Los Alamos performed with TRAC BD1/MODt. Scientific Laboratory. October 1985. 229pp 8512050451. LA. 10548-MS 33770 246 NUREG/CR 3633 V04: TRAC B01/ MOD 1'AN ADVANCED BEST The Trans,ent Reactor Analysis Code (TRAC) provides an ad-ESTIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSIENT ANALYSIS Volume 4 Developmental vanced, best estimate analysis capabety for pressunted water Assessment SHUMWAY,R.W. EGAG Idaho. Inc (subs of reactors and for many thermal-hydrauhc test facihties The most EGAG, inc ). September 1985 101pp. 8510040407. EGG 2294 recent pubhcly released version of TRAC is TRAC PF1 This 32858 314. code verson includes a full two fluid modeleng capability in both This volume of the TRAC-BD1/ MODI manual discusses the the three-dimensonal vessel component and the one-dimon-results of developmental assessment calculations performed sional components. We have improved the numerict" methods ma'nfy with a prehm nary verson of TRAC-BD1/ MODI (V21). en the one-dimensional components to provide a more stable and some selected cases performed with the off.cial verson of soluton and to permit the code to run faster The Los Alamos TRAC-BD1/ MOD 1 (V22), which differed from the preliminary report, " TRAC PF1 An Advanced Best Estimate Computer Pro-version due to few small correctons. Twenty one test cases gram for Pressunrod Water Reactor Analysis," LA-9944 MS have been performed, ranging from simple s.ng'e effect flow (NUREG/CR 3567). provides a detailed descnption of the code. tests up to a full BWR/6 system calculaten. The four groups of This report documents the Los Alamos results of the second tests are separate effects hydraubc tests, steady state heat assessment phase, independent assessment for TRAC PF1, We transfer tests. transient heat transfer tests, and integral system analyzed separate-effects tests in the Semiscate facility to in-effects tests. The separate effects test cases were each vestigate natural-circulation and reflus coohng We ana!yred m-chosen to exercise a specific hydraubc or heat transfer model m tegral tests from the Semiscafe and the Loss-offluid Test facili-the code, while the integral system effects tests were chosen to ties to emptore the small and mterrnudiate-break loss of coolant
' exercise the code as a whole. The TR AC code verson initia4y accident (LOCA) capabd cy and the non-LOCA capabibty We used was TA021 for all cases. Code errors were evident en also anatyred the loss of-feedwater transient en the Crystal some of the heat transfer runs. Arver plant. The results show reasonably good agreement with the data, but indicate that improvements are required for the NUREG/CR 3634: MAINTENANCE PERSONNEL PERFORM. critical-flow model and the interphatic condensaton model ANCE SIMULATION (MAPPS) MODEL: tJSER'S MANUAL.
KOPSTEIN F F : WOLF.J J Cak Ridge National Laboratory September 1985 115pp. 8512270273 OR NL/ TM-9545. NUREG/CR 3647: DESIGN AND FABRICATION OF A 1/8-SCALE STEEL CONTAINMENT MODEL REESE R T , 34079 172 HORSCHELD S Sandia Nat onal Laboratones Apnl 1985 This document serves as the user s manual for the MAinte. 13tpp 8504170693 SAND 84 0048 29907 022 nance Personnel Performance Simulation (MAPPS) model A 1/8-scale steel model containment buil4ng was designed MAPPS is a genera'ited, stochastic computer simulation model which simulates the performance of maintenance personnel in and fabricated m support of the Containment Safety Margins nuclear power plants The model considers workplace, mainte- Program This program is directed to determine the margin of nance technician, mot vational, human factors, and tasbonent- safety of containments in severe accident conditions it is ed variables to yield predictive information about the effects of planned to enternally pressunto the model to failure in this test-these vanables on successful maintenance task performance eng program, failure modes of the pressure vessel and scated As such, MAPPS is beheved to represent a fundamental prob- penetrations will be enam ned m detail The modni was de-Ebilistic nsk assessment (PRA) analytic tool Moreover, it serves signed according to Section lit of ASME Code for Class MC the needs of nuclear power plant maintenance managment for containment vessels with the enception that no code stamp was maintenance operations analysis, and the needs of architectural required sence no nuclear matenals would be housed within the i model. All the general requirements (subsection NCA) and spe-and engineenng firms for maintenance system design evalua-t on This manual deals with the procedural aspects of operating cific requirements (subsection NE) of Secton lit of the ASME the MAPPS computer program The first secMn of the present Code were met The ma son'y of the model was fabncated from report desenbes the use of the MAPPS model e inm the nuclear 3/164n. SA516 Grade 70 steel plate in the form of a nght Circu-power plant contest. This section also presents the pnncipal po- lar cylinder capped with a hemisphencal dome Eleven penetra-tential uses of the model and the major types of informaton tons and two htting trunnions were encluded m the model The that can be helpful to its vanous types of users m making dect- cyhnder/ dome secten was joinded to a 2.f ellipsoidal base sions. Secton 2 provides an overview of MAPPS utilitat on and (test forture) composed of thicker (1 1/8 m and 1 1/2.an) plate esplains the process of planning and executing simualtions material The rrohl was supported on ses kgs to permit access runs Data enput requirements are outhned, and the varous data for personnel, matrumentation, data acquisiten, power, and types are explanced Also, the model s data outputs are illustrat- pressura piping The model was fabricated in Apr1 through Oc-
~~. - - - - - -- = - - - - - - - - - - - - - - - - -~~
Main Citations and Abstracts 37 tober 1983 by Chicago Dndge and iron and erected at the test A method of assessing the quantitative uncertainties in the site in Albuquerque. New Moxico, in November 1983 determinaten of PWR powers and coolant flows caused by NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF measurement uncertarnties is provided. The method dehnes the ORNL INSTRUMENTATION IN REFLOOD TEST FACILITIES. parameters entering into the calculation. the types and sources HARDY,J E4 HERSKOVITZ,M D. Oak Ridge National Laborato- of measurement errors wtuch must be considered, together with ry Apnl 1985. 56pp. 8506070366. ORNL/ TM-9067. 30798.265 sources of quantitative data for the uncertainties. A mathemato instrumentation for making two-phase measurements in ex- cat model is developed which combines the measurement un-penmental refill reftood test facilities was developed by Oak certainties in a rigorous statistical manner to give the overall un-Ridge National Laboratory (ORNL) through the Advanced Instru- cedainty in the desired p' ameter together with a sample calcu-mentation for Reflood Studies (AIRS) program. These unique in- laton. strumentation systems were designed to survive the severe in-vessel environmental conditions that exist dunng a simulated NUREG/CR 3660 V01: PROBABILITY OF PIPE FAILURE IN THE pressunzed water reactor loss-of coolant accident (LOCA) The REACTOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS Volume t. Summary Report HOLMAN.G S ; CHOU C K measurements include two phase flow velocity. void fracton-and film thickness and velocity, and are required for bettor un- Lawrence Livermore Natonal Laboratory Jufy 1985 103pp derstanding of reactor behavior dunng LOCAs. The adequacy 8508010773. UCID 19988. 31923 247 (survivabihty and data quahty) of the instrumentation systems in- As part of its reevaluation of the double-ended guillotine stalled in four expenmental reflood test facihties is assessed- break (CEGB) of reactor coolant loop piping as a design basis Signal cond>tioning electronics and sensor thermocouples func. avent for nuclear power plants, the U S Nuclear Regulatory tioned extremely well For the first time, two-phase flow meas- Jommisson (NRC) contracted with the Lawrence Livermore Na-urements were made in-core dunng a simulated LOCA. Because tional Laboratory (LLNL) to estimate the probability of occur-of the harsh environment and geometncal constraints, some rence of a DEGB. and to assess the effect that earthoaakes have on DEGB probabihty This report describes a probabihstic sensor failures were considered hkely; the number actually fail" ang in service was within espectations. An eucepton to this evaiuation of reactor coolant loop piping in PWR plants hawng record occurred in the Slab Core Test Facihty -- Core 1. A chlo- nuclear steam supply systems desegried by Westinghouse Two ride 4on stress corroson problem destroyed segnal cables at the causes of pipe break were considered p'pe fracture due to the vessel seal for most sensors This problem was Corrected by growth of cracks at WWded joints (" direct" DEGO), and pte chang:ng the sealant matenal at the vessel penetraton in the rupture indirectfy caused by failure of component supports due subsequent faciht>es. Overall, the performance of the instrumen- to an earthquake (" indirect" DEGB) The prot,abihty of direct DEGB was estimated using a prcbabihstic fracture mechanics taten was very satisfactory yeding valuable data dunng simu-model. The probabihty of end-rect DEGB was estimated by esti-
!ated LOCAs an refill reflood test facihties mating support frathty and then convolving frag 4 sty and seismic NUREG/CR 365h PREllMiNARY SCREENtNG OF FUEL CYCLE hatard The results of this study indicate that the probabihty of AND BY PRODUCT MATERIAL LICENSES FOR EMERGENCY a DEGB from either cause is very low for reactor coolant loop PLANNING. BENNETT.D Ea RUNKLE.G Ea ALPERT,D J : et af. Piping in these plants. and that NRC should therefore consider Sandia Natonal Laboratones Apnl 1985.137pp. 8506060385 chminating DEGB as a des 4gn basis event in favor of more real.
SAND 84 Ot88 30775 062. istic cntena This report summarizes work done for the U S Nuclear Regu-latory Commission as part of a program considenng the need NUREG/CR 3660 V03: PROBASILITY OF PIPE FAILURE IN THE for and appropnate level of emergency response planning at REACTOR COOLANT LOOP OF WESTINGHOUSE PWR fuel cycle and by-product matenal facihtres The purpose is to PLANTS Volume 3 Guillotine Dreak Indirectfy Induced By Earth-(1) provide a base of technical information for identifying and quakes RAVINDRA.M K ; CAMPDELL,R D. KENNEDY.R P; et rank:ng those facilities for which the need for emergency re- al Lawrence Livermore Natonal Laboratory February 1985
~~' sponse pla9twng and preparedness shouH be further consid. 199pp 8503130008 UCID-199M 29360183.~~
ered, and (2) perform an initial screening of heenses issued by The requirements to design the nuclear power plants for the NRC. A data base containing tf9e radionuchde possession hmits effocts of an ensta,taneous doubio ended guillotine break j for each license was developed. Dose estimates for a unit (t (DEGB) of the reactor coolant loop (RCL) piping have iM to ex-cune) release of each of the radeonuclides in the data base cessive design costs. interference of normal plant operation and were calculated. To account for tho vanabihty in weather, distn. maintenance, and unnecessary radiation esposure of plant butons of doses were estimated for a full range of meteorolog,. mainMnance personnel This report describes an aspect of the cal cond. tons. As requested by NRC, doses at the 99th per- f 51C/ Lawrence ( Nermoro National Laboratory sponsored re. centile of the distnbut on were used An initial screening analy, search program aimed at demonstrat;ng that the probab bty of sis was performed for the approsimately 9400 + licenses by CEGB in RCL piping of nuclear power plants #s acceptably sma'l companng the estimated 99th percentile doso for a postulated and the requirements in design for the DEGB effects te g , pro-release of a fraction of the hcensed possesson hmet to the dose vision of pipe wNp restraints) may be removed This study esti-4 levels suggested in the Environmental Protection Agency's Pro, mated the probabihty of indirect DEGB in RCL piping as conse. tective Acton Gusdes Using relatively conservative assumptions quence of seismic. induced structurat failures within the contain-in the screening analysis, all but at rnost a few hundred hconses ment of %stinghouse supphed pressun&d water reactor nucle-were found to have estimated doses below the Protective er power plants in the United States The medan probabAty of Action Guide levels The few hundred identAnd in this initial 'ndirect DEGB was estimated to be about 3s1006) per year with screening should be further evaluated using reahstic assump, a 10*4 to 90% subjectrve probabikty range approximately for tons and site specific information to estabhsh the need for, ap. t 1007) per ya to 4s1005) per year propnate level and estent of, and potential effectiveness of emergency response planning and preparedness beyond that NUREG/CH 3660 V04: PROBADILITY OF PIPE FAILURE IN THE currently required REACTOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS Volume 4 Pipe Failure Induced By Crack Growth in NUREG/CR 3659: A MATHEMATICAL MODEL FOR ASSESSING West Coast Plants CHINN.D J ; HOLMAN.O S ; LO T.Y ; el al THE UNCERTAINTIES OF INSTRUMENTATION MEASURE. Lawrence Livermore National Laboratory July 1985. 5Ppp MENTS FOR POWER AND FLOW OF PWR REACTORS 8508020207. UCID 19988 V04 31962143 HESSON.G M ; CUFF,W C ; STEVLNS.D I Battelle Memonal The U S Nuclear Regulatory Commission contracted with the Inst.tute Pacific Northwest Laboratones February 1985. depp. Lawrence trvermore National laboratory to conduct a study to 8503220007, PNL473 29487 217 determine of the probabihty of occurrence of a doubic ended f
i 38 Main Citations and Abstracts guillotine break in pnmary coolant piping warrants the current NUREG/CR-3688 V01: GENERATING HUMAN RELIABILITY ES- j design requirements that safeguard against the effects of such TIMATES USING EXPERT JUDGMENT. Volume 1. Main Report. a break. Tr:ss report assesses the reactor coolant-loop piping COMER.M.K.; SEAVER.D A ; STILLWELL,W G ; et al General system of west coast Westinghouse plants. The results indicate Physics Corp. January 1985. 6tpp. 8502210260. SAND 84-7115. thit directly induced DEGB 6s an unlikety event in the west 29028 272. coast Westinghouse plants. For the Trolan plant, leak is far The U S Nuclear Regulatory Commission is conducting a re-more likely than a direct DEGB. Further, earthquakes have very search program to determine the practicality, acceptability, and little effect on the probabihties of leak and direct DEGB. At the usefulness of several different methods for obtaining human re-liability data and estimates that can be used in nuclear power Deblo Canyon plant, the increase in postulated seismic levels plant probabilistic nsk assessments (PRA) One method, envesto due to reevaluation of the site to account for the Hosgn Fault gated as part of this overall research program, uses expert judg-his caused directly Induced DEGB fadure probability to be de. ment to generate human error probability (HEP) estimates and pendent on earthquake occurrences. The resutting direct DEGB associaW unwam he N pW esch m M failure probability is stdl much lower than the indirect Dt GB fail- document evaluated two techniques for using expert judgment. ure probability for Diablo Cat' yon. pa, red compansons and direct numencal estimation. Volume 1 of this report provides a bnef overview of the background of the NU~.EG/CR-3663 V01: PROBABILITY OF PIPE FAILURE IN THE project, the p ocedure for using psychological scaling tech-REACTOR COOLANT LOOPS OF COMBUSTION ENGINEER- niques t generate HEP estimates and conclusions from evalua-ING PWR PLANTS. Volume 1: Summary Report. HOLMAN.G S; tion of the techniques Volume 2 provides detailed procedures LO T.Y.; CHOU.C.K. Lawrence Livermore National Laboratory. for umng the techniques, deta' led desenptions of the anatyses Jinuary 1985. 81pp. 85021200 % UCAL 53500 V01. performed to evaluate the techniques, and HEP estimates gen. 28972.057- erated as part of this proiect. The results of the evaluation indi. As part of its reevaluaton of the double-ended guillotine cafe that technu ues umng emport Judgment should be given break (DEGB) as a design requirement for reactor coolant strong consdera on for use in developing HEP estimates in piping, the U.S. Nuclear Regulatory Commisson (NRC) con- addit on. HEP et. . nates for 35 taske related to boiling water re-tracted with the Lawrence Uvermore Natonal Laboratory (LLNL) actors (BWRs) m te obtained as part of the evaluation. These to estimate the probability of occurrence of a DEGB, and to HEP estimates are a!so included 6n the report. assess the effect that earthquakes have on DEGB probability. This report desenbes a probabilistic evafuaton of reactor cool. NUREG/CR 3688 V02: GENERATING HUMAN RELIABILITY ES-TiMATES USING EXPERT JUDGMENT. Volume 2. Appendices. ant loop piping in PWR plants having nuclear steam supply sys. COMER.M K.; SEAVER.D A : STILLWELL.W G ; et at General tzms desgned by Combustior. Engineenng Two causes of pipe Physics Corp January 1985. 176pp 8502210262. SAND 84 break were contdered: pipe fracture due to the growth of 7 8 cracks at welded joints ( dirxt" DEGB). and pipe rupture indo gEG/CR '3688,V01 abstract. rectly caused by failure of component supports due ta an earth-quake (" indirect
DEGB) The probability of direct DEGB was NUREG/CR 3703: ASSESSMENT OF SELECTED TRAC AND estimated us ng a probabihstic fracture rnechanics model The RELAPS CALCULATIONS FOR OCONEE 1 PRESSURIZED probability of indirect DEGB was estimated by estimat:ng sup- THERMAL SHOCK STUDY. ROHATGl.U S.; PU,J , SAHA.P.; et port fragility and then convolving fragility with seismic hazard al. Brookhaven National Laboratory. Apnl 1985 98pp.
The results of this study indicate that the probability of a DEGB 8505070497. BNL NUREG 51750. 3021 LOO 5 from either cause is very low for reactor coolant loop piping in Several Oconee-1 overcooling transeents that were computed thase plants, and that NRC should therefore consider eliminat- by LANL and INEL u$ng the latest versons of TRAC PF1 and ing DEGB as a design basis en favor N more reakstic cntena RELAPS/ MODI 5 codes have been reviewed by BNL Three of these transents were seloctd for detailed review as they either NUREG/CR 3663 V03: PROBABILITY OF PIPE FAILURE IN THE had the potential of challenging the 6ntegnty of the pressure REACTOR COOLANT LOOPS OF COMBUSTION ENGINEER- vessel or highlighted the effect of code d.fferences. These are ING PWR PLANTS, Volume 3 Double Ended Guittotine Break in- (1) Main Steam Line Dreak (MSLB), (2) All Turbine Bypass directly Induced By Earthquakes. RAVINDRA.M K ; Valves Stuck Open, and (3) 2-Inch Small Break LOCA. CAMPBELL.R D,; KENNEDY,R.P; et al. Lawrence Livermore Natonal Laboratory January 1985. 118pp. 8501280412. NUREG/CR 3706: TRAC ANALYSES OF SEVERE OVERCOOL-e ING TRANSIENTS FOR THE OCONEE 1 PWR. IRELAND.J R 28573 001. Los Alamos Scietdic Laboratory. August 1985, 26tpp. The requirements to de$gn the nuclear power plants for the effects of an enstantaneous double-ended guillotine break 8m0026. LA 10055M 32625 059 This report descnbes the resufts of several Transient Reactor (DEGB) of the reactor coolant loop tRCL) piping have led to en- Anatysis Code (TRAC) PF1 calculations of overcooling tran-esssive desgn costs, interference of normal plant operation and tents in a BaW & Mcon lowereMoop, pressunted water maintenance personnel. This report describes an aspect of the reactor (Oconee 1). The purpose of this study is to provide de-NRC/ Lawrence Livermore National Laboratory sponsored re- tailed input on thermal-hydrauhc data to Oak Ridge National starch program aimed at demonstrating that the probability o' Laboratory for pressunted thermal-shock analyses The tran-DEGB in RCL piping of nuclear power plants is acceptably small sient calculations performed were plant specific in that details of c.nd the requirements to demgn for the DEGB ef'ects (e g , pro ~ the pnmary system, the secondary system, and the plantante-vison of pipe whip restraints) may be removed. This study est'- grated control system of Oconee 1 were encluded in the TRAC rnated the probability of 6ndirect DEGB in RCL piping as a con
input modet The results of the calculations indicate that the tur-sequence of seismac-ir'duced structural failures within the con- bine bypass valve f ailure transont was the most severe in t4rms tonment of Combuston Engineenng supplied pressunted wate' of resulting in relatrvely cold liquid temperatures in the down-reactor rur,. lear power plants en the United States The median comer region of the vessel The power-operated relief valve probability of indirect DEGB was estimated to be in the range of sogg or. coolant accident transient was the leist severe in terms 10(-6) per year of older plants, and less than 10( 8) per year for of downcomer hquid temperatures because of vent valve fluid modern plants; u$ng very conservatrve assuroptions, the 90% mining and near satura'ed conditions in the pnmary system. It is subjective probabslity value (confidence) of P DEGB was found recommended that future calculations con $ der a widor range of to be less than 5x10(-5) per year for older plants and less than operator actons to cover the spectra of overcoohng transient 3:1007) per year for modern plants. sequences more completely
Main Citations and Abstracts 39 NUREG/CR 3709: METHODS OF MINIMlZING GROUND-WATER to the results of tests at a number of other facihties. The pre. CONTAMINATION FROM IN SITU LEACH URANIUM ignit.on temperature had a strong effect on the peak pressure, MINING Final Report. DEUTSCH,W.1: MARTIN.W J : while pre 4gnition pressure in the range examined had no effect EARY,L.E.; et al. Battelle Memonal Institute Pacific Northwest on combustion pressure ratios. Calculations showed that, al-Laboratones. March 1985. 158pp. 8504040003. PNL.5319. 29629 071 though the effect of dynamic head on the peak pressure was a This is the final report of a research project dealing with few percent or less, interactions of the wave preceding the methods of minim 12ing ground-water contamination from in situ flame front with the flame and with the vessel walls may be ap-parent in the expenmental records teach uranium mining. Field work and laboratory expenments were conducted to identify excursion indicators for monitonn9
, purposes dunng mining, and to evaluate effective aquifer resto. NUREG/CR-3723: STRESS INTENSITY-FACTOR INFLUENCE ration techniques following mining Many of the solution con- COEFFICIENTS FOR SURFACE FLAWS IN PRESSURE VES-stituents were found to be too reactive with the aquifer sed SELS. BALL.D G.; BASS B R ; BRYSON.J W ; et al Oak Ridge ments to reliably indicate excursion of leaching solution from National Laboratory. February 1985. 53pp. 8503290280 ORNL/
CSD/TM 216 29563 295, the ore zone, however, in many cases, the concentratons of chlonde and sulfate and the total d,ssolved solids level of the In the fracture-mechanics analysis of reactor pressure ves- ) solution were found to be good excursion indicators Aquifer sels, stressantensty-factor influence coefficients are used in restoraton by ground-water sweeping consumed ground water conjunction with superpositen techniques to reduce the cost of and was not effectve for the redou-sensitive contaminants often calculatmg stresuntensty factors. The present study uses a present 6n the ore tone. Surface treatment methods were effec- finite. element code, together with a virtual crack extensson tech-tive in lowenng the amount of water used, but also had the po- nique, to obtain influence coefficients for semeelkptical surface tential for creating conditions in the aquifer under which the flaws in a cy!>nder, and particular emphasis was placed on redox sensitive contaminaats would be mobile. In situ restora, mesh convergence Cess than 1% error was sought in the re. tion by chemscal reduction, in which a reducing agent is added suits from any one mesh construction parameter). Companson to the soluton recircu!ated through the ore zone dunng restora. of the coeffoents wth those obtu,ed by other investigators ton, has the capability of restonng the ore tone sediment as shows good agreement. Furthermore, stressantensity factors well as the ground water. This method could lead to a stable obta.ned by superpositen for a severe thermal-transient loading chemical cond: tion in the aquifer similar to cond,tions before condition agree w than 1*v cf the values calculated by a d, rect mining, finite-element method. Influence coefficients were calculated for NUREG/CR-3710: LABORATORY STUDIES OF A BREACHED Wee specdic amiah onented semdptical surface flaws. The NUCLEAR WASTE REPOSITORY IN BASALT. SEIT2.M G.; I'rst was a 2-maong innesudace Han in a nuclear macts pres. BOWERS,D L; GERDING T.J: et al Argonne National Labora- sure vessel with depth to wall-thickness ratos between 0 2 and tory. August 1985.150pp. 8508263304 ANL 8416. 32368 071. 0 9. The cecond was an inner-surface flaw in the reactor vessel Espenments are desenbed that combine backfill, radioactwo with a surface length to-depth ratto of 6 and with depth-to wall-waste, and basalt rock in a single flowing gioundwater stream in thickness ratos between 0 05 and 0 2. The third was a 1 m-long a manner analogous to a hydraulic breach of a waste repos,. flaw on the outer surface of a test vessel with depth-to-wall-tory. The expenments were used to study chemical interactions thickness ratios between 0.1 and 0.9 For the reactor vessel, that would occur if repository components were breached by separate coefficients were calculated for the cladding on the flowing water. The result of most significance to issues of re. inner surface and for the base-matenal regon This allows for pository performance was that uranium, neptunium, and pluton,. an accurate accounting of the effect of thermal stresses in the um were found to move more rapidly through repos, tory compo. cladding on the stressantensity factor for surface flaws that nents that were a!!ered to represent aging than through fresh extend through the cladding into the base matenal. matenals. In contrast, cesium moved slower through aftered re. pository matenals, as had been deduced from prevous work NUREG/CR-3736: FIELD AND THEORETICAL INVESilGATIONS using batch adsorption tests. Two other parameters studied es. OF FRACT'JRED CRYSTALLINE ROCK NEAR penmentally, the metal alfoy used in the apparatus and an sontr. ORACLE ARIZONA. JONES.J W ; SIMPSON E S ; l ing radiation field imposed on the expenmental apparatus, had NEUMAN.S P.; et al. Antona, Unry of. Tucson, A2. August little or no measurable effect on radtoactive element transport 1985 115pp. 8508290526 32410 254. , by flowing water. Inasmuch as ti,e alteration of the repos tory A combination of geophyucal and hydrautic test:ng has been matena!s aging in an actual repository, we conclude that conducted in granite near Oracle, Antona The purpose of the changes with age will detnmentally affect the ability of a repose work is to determine relationships, if any, among (1) fracture , tory to isolate uranrum, neptunium, and plutonium Because distribution, (2) geophysical properties, and (3) hydraulic proper. { these elements have long lived radoactwo isotopes in nuclear ties of fractured rock of low hydrauhc conductivity To date, waste, the degradation with time is a ma or issue regarding the eight vertical bonngs spaced 20 to 50 feet apart, ranging from performance of a nuclear waste repository in basalt. 250 to 300 feet in depth, have been dnlled The data obtained NUREG/CR 3721 V01: PRESSURE MEASUREMENTS IN A HY. from nwtron, gamma, acoustic velocity, electncal-resistivity, and DROGEN COMBUSTION ENVIRONMENT. Hydrogen-Air Com- acoustic televiewer logs, with the results of over 100 ser'gle-bustion Test Senes 1 And 2 in The FITS Tank ROLLER.S F hole. straddle packer intection tests make possibio a detailed Sandia National Laboratones. Apnl 1985 59pp 8504170005 desenption of the fracture system. Geophysscal logs readily SAND 83-2621/1. 29904 307. detect fractures and are sensitwo to subtle lithologic vanations Ffydrogen combuston tests were performed in the Fulty in, of the granito Onentaton and distnbution of indwidual fractures strumented Test Site (FITS) tank under the Hydrogen Behaver were determined from the interpretation of the acoustic-tele. Program perforrned by Sandia National Laboratones under con. viewer data, and from the analysis of core obtained from one tract wrth the US Nuclear Regulatory Commission Test senes 1 borehole Fracture dansties over the 13 foot long straddle. and 2 examined the effects of a number of parameters on hy. packer test intervals did not correlate with measured hydraulic i drogen air combuston: the initial temperature and pressure of conductivity measurements A strong correlaton between the the gases, the effect of added steam or carbon dioxide as di- noutron log response and measured hydraulic conductivity does luents, and the percent hydrogen in air. For tests in the rarige of exist, it was used to supplant conductivity measurements The 20% to 40% hydrogen in air, recorded peak pressures were geostatistical technique of knging provufnd a three-dimensional equal to adiabs'ac, isochroic, complete combustion (AICC) map of hydraulic conductwity that can be compare : with sub-values within an expenmental enor of 85% This was contrary surface interpretafsons of the geophysical logs i
40 Main Citations and Abstracts NUREG/CR 3738: ENVIRONMENTAL EFFECTS OF THE URANI- ies, (5) crack arrest technology. (6) irradiation effects studies, UM FUEL CYCLE.A Review Of Data For Technetium. TilL.J E. (7) cladding evaluatens. (8) intermediate vessel tests and anal-Radiological Assessments Corp. SHOR,R.W.; HOFFMAN,F.O. yms. (9) thermal-shock technology, and (10) pressunted ther. Oak Ridge National Laboratory February 1985. 135pp. mal-shock technology. 8503120452. ORNL/TM-9150. 29340:293 NUREG/CR-3746 V02: LWR PRESSURE VESSEL SURVEIL-Sources of (99)Tc releases to the environment are reviewed LANCE DOS! METRY IMPROVEMENT PROGRAM.Semaannual
~~,r the uranium fuel cycle considenng the recycle of spent ura. Report.Apl September 1984.~~
Progress 1984 - nium fuel and no fuel recychng Without recycling the only LIPPINCOTT,E.P4 MCELROY,W N. Hanford Enginoenng Devel-source of (99)Tc release is an extremely small amount associat. ed with airborne emissions from the processing of high-level opment Laboratory. Apnl 1985.220pp.8505070562. HEDL-TME wastes. With recycling. (99)Tc releases are associated with the 84 21. 30209.020. This report desenbes progress made in the Light Water Reac-operation of reprocessing facihties, UF(6) converson plants. tor Pressure Vesset Surveillance Dosimetry improvement Pro-uranium ennchment plants, fuel fabncation facihties, and low. and high-level waste processing and storage facihties. Among gram (LWR-PV SDlP) d.enng FY84. The pnmary concern of this these, the most prominent (99)Tc releases are from the liquid program is to improve test. venty. and standardize the physics-dosimetry-metalluqy and associated reactor and damage anaty-effluents of uranium ennchment facihties. An extenssve review of sis procedures and data used for predicting the integrated ef-data estmate parameters for predicting the environmental be. haver and fee of (99)Te indicates a reduced radological signifi- fects of neutron exposure to LWR.PVs and their support struc-cance for the ingestion of milk and meat. More important path. tures. These procedures and data are being recommended in a ways of exposure to (99)Tc will probably be associated with new and updated set of ASTM standards being prepared. dnnking water and the consumptenn of aquatic organisms, tested, and ventied by program participants. These standards, garden vegetables, and eggs. For each parameter reviewed in together with parts of the US Code of Federal Regulatons and this study, a range of values is recommended for radiological ASME codes, are needed and used for the assessment and assessment calculations. Where obvious discrepancies exist be- control of the condition of LWR PVs and their support structures tween these range and the default values hsted in USNRC Reg- dunng the 30- to 60-year lifetime of a nuclear power plant. NUREG/CR-3746 V03: LWR PRESSURE VESSEL SURVEIL-efau t va s is m d LANCE DOSIMETRY IMPROVEMENT PROGRAM.1984 Annual NUREG/CR 3741 V02: EVALUATION OF POWER REACTOR Report, October 1,1983 September 30.1984. MCELROY.W N. FUEL ROD ANALYSIS CAPABILITIES Phase 2 Topecal Hanford Engineenng Development Laboratory. Apr l 1985 Report Volume 2 Code Evaluation COLEMAN.O.R. Control Data 110pp 8505070543 HEDL TME 84 31. 30208 270. Corp. November 1985.120pp. 8512190246 34010 267. See NUREG/CR-3746.V02 abstract. FRAPCON-2 (V1MS) was apphed to generate fuel perform-ance predictions for setected rods of a recently eva!uated NUREG/CR 3747: THE SEl ECTION AND TESTING CF RCCK power reactor data sample Rod design, operatonal, and per. FOR ARMORING URANIUM TAltlNGS IMPOUNDMENTS FOLEY.M G.; KIMBALL,C S ; MYERS,0.A ; et al. Battelle Memo. formance data was obtained from the EPR) Fuel Performance tral Institute Pacific Northwest Laboratones. May 1985 119pp. Data Base. The data was systematically processed to generate code oput garameters. After initial debugging. FRAPCON 2 was 8506140396. PNL 5064 30933 275. Under contract to the U S Nuclear Regulatory Commission, apphed for bnechmark studies to qualify revised cladding creep. Pccific Northwest Laboratory has developed an approach for down, growth, and fisson gas release models Benchmark ro. suits indicated improved gas release and clad deformaton mod. selecting and testing rock for ris suitabikty and durabil<ty as el.ng. but increased computer requvements and mechanical armor for protecting decommissioned uranium mill taslings pJes conservatism dunng hard PCI Compansons between measured A preliminary survey of the hterature determined that existing and calculated fuel and cladding corroson are presented and techn*1'ses for testing rock durability were inadequate for evalu-interpreted relative to code error magnitudes, distributons. and ating long term (> 100 years) apphcatons Suites of rock sam-trenda versus rod design and operating parameters The toruits pies *sth common litholog:es and documented duratons of eu. indicate that FRAPCON 2 (V1MS) is an irnproved code having posure to weathenng were then collected arid submitted to adequate capabihty for best estimate analysis of moderate duty three-axis ultrasonic testing in an attempt to rievelop a more re-fuel rod performance, up to average burnups of at least 40 liable testeg technique We found little correlation between the duraton of weathenng and ultrasoiand velocity or attenuation in GW )/MTU. The Pain model lirnstations involve lack of certain physical effects which can gerera!!y be compensated for when the rock. Through further studf we determined that the best sett.ng up input or interpreting results, namely, densehcaton screening approach encorporates common geomorphologic field effect on cecreaseg pellet relocation, sweihng accommodation collection techniques and laboratory tests. Suites of samples with known duratons of esposure to weathenng can be subsect-by fuel porosity, cladding creepout effect on PCI stress re:ax. aten, and coolant chemistry and petiet impunty effects on clad ed to wet abramon and wetting-drying tests to screen local rock types and select those with the greatest potential durabahty Fur. ID and OD corros on. cmase of rnass wdh emnron. NUREG/CR 3744 V02 HEAVY SECTION STEEL TECHNOLOGY "M g. WP a al
~~m t PROGRAM SEMIANNUAL PROGRESS REPOR1 FOR april- drying cy les) can be eemated umng this approach SEPTEMBER 1984. PUGH.C E. Oak Ridge National Laboratory.~~
January 1985 244pp 8502210332. ORNL/TM-9154/V2. NUREG/CR 3752: EFFECTS OF HYDROLOGIC VARIABLES ON 29052.001, ROCK RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUND-The Heavy-Section Steel Technology (HSST) Program is an VENTS WALTERS.W H ; SKAGGS.R L Battelle Memonal Insts-engineenng research activity conducted by the Oak Ridge Na- tute. Pacific Northwest Laboratones January 1985 55pp tional Laboratory for the Nuclear Regufatory Comm.sson. The 8501280379 PNL-5069. 28572 072. program compnces studies related to all areas of the technolo- Pacific Northwest Laboratory es studying the metigaton of ero-gy of matenals fabncated into thick secton pnmary. coolant con. sion of earthen radon suppresson covers for uranium tailings tainment systems of hght-water-cooled nuclear power reactors impoundments. Because the covers will require croson protec. The investigaton focuses on the behaver and structural integn- tion for upwards of 1000 years, rock nprap (armonog) has been ty of steel pressure vessels containing cracklike flaws Current proposed as the pnmary protection method This study investi-work as organ 2ed into ten tasks (1) program management, (2) gates the sensitivity of riprap design procedures to entreme fracture-methodology and analysis, (3) matenal charactentaton flood events that can generate h*gh flow velocities and shear and properties, (4) environmentatty assisted crack grusth stud- stresses. The study uses two decommissioned taihngs sites
Main Citations and Abstracts 41 i (Grand Junction and Slick Rock. Colorado) as case studes to workstation and computer / simulator is an effectwo configura-eva!uate the sensitivity of design rock size with respect to vana- tion The imphcotons of these results for display evaluation and bles such as flood discharge, side slope, specific gravity, safety design are discussed. factor, and channel roughness. The results indicate that design rock size can vary significantfy for d:fferent design procedures. NUREG/CR 3772: RELAPS ASSESSMENT.SEMISCALE SMALL Other significant results indicate that embankment side slopes BREAK TESTS S-UT 1,S-UT 2, S-UT 6,S-UT 7 AND S-UT 8. of about 4H;1V are optimum for rock nprap and that the use of PETERSON.A.C. Sanda National Laboratones. February 1985. rock matenal with specific gravities less than 2 50 may prove 230pp. 8503040547. SAND 84 0884. 29198 001. too costly. The RELAPS independent assessment protect is part of an NUREG/CR-3757: TRAN B-2:THE EFFECT OF LOW STEEL overalt effort to evaluate the capabshty of vanous system codes CONTENT ON FUEL PENETRATION IN A NON-MELTING CY. to calculate the detailed thermal / hydraulic response of LWRs LINDRICAL FLOW CHANNEL MCARTHUR.D.A ; MAST,P.K. dunng accident and off-normal constions. The RELAPS com. Sandia Nat onal Laboratories. Apnl 1985. 72pp. 8505160176. puter code is being assessed against test data from vanous in-SAND 84-0814. 30457:029. tegral and separate effects test facilities As part of the assess-The TRAN B-Senes of expenments is being conducted at ment effort, several small break tests with and without upper Sandia National Laboratones to investigate the charactenstics head injection ( UHl) of emergency core coolant (ECC), per-of fuel removal and freezing through the upper axial blankets of formed m the Semiscale Mod-2A facihty, have been analyzed an LMFBR dunng the transition phase of a hypothetical core disruptrve acesdent. The second exper: ment in this senes. TRAN The results shcw that RELAP5/ MOD 1 is capable of calculating B-2. was performed in July 1983 This expenment involved the some aspects of the important phenomena dunng small breaks Injection of a mixture of 959. UO(2) and 5*. stainless steelinto both with and without UHl. The times for the system to depres-a simple thick walled steel cylindncal flow channel The initial sunze to the UHf and/or loop accumulator flow irutiabon were temperature of the steel channel was low, such that melt;ng of calculated satisfactonly. The correct trends of the effects of the walls upon contact with the hot melt was not expected. Pre- break size and of UHI on the system pressure response were vious expenments under similar conditions t,ut using pure UO(2) also calculated The ingection rate from the UHI and loop accu-melts had shown stable crust growth and fairfy long penetration mulators was not afways calculated correctly: the flows cycled estances. This expenment was mtended to investigate whether on and off tecause large flow surges caused the accumulator those results were also applicable for the case of UO(2)/ steel pressures to temp ranty decrease below the system pressure mixtures. The results of the TRAN B-2 expenment, consisting of This cycling of the flow had a significant effect on the system data from online instrumentation and post erradiaton examina- response dunng UHf accumulator flew When the upper head tion, suggest that low steel content fuel / steel mixtures behave was hound-filled from UHI flow, a core hquid level depression very semifarty to pur e UO(2) melts, was calculated, but not measured, that resulted in a dryout of NUREG/CR-3764: BWR-LTAS: A DOILING WATER REACTOR "
LONG-TERM ACCIDENT SIMULATION CODE

HARRINGTOftRM FULLER,L C. Oak Ridge Na!ional Labora$ * "
tory. February 1985. 167pp. 8503120456. OANL/TM 9163. * " "" 29338 179 kanseents, con n'g *to a low W Ind in N vessel and The BWR LTAS code was developed by the SASA program la% time core heatup Kghe late-bme cm tempeatures wee at Oak Ridge Natonal Laboratory for the detailed study of spe- cabulated than measured both with and without UHl. cific accident sequences at Browns Ferry Unit One station blackout, small break LOCA outside pnmary containment, loss NUREG/CR 3774 V02: ALTERNATIVE METHODS FOR DISPOS. of decay heat removal, loss of vessel water iniection, and antics- AL OF LOW LEVEL RACIOACTIVE WASTES Task 2A Technical pated transient without scram. The pomary use of the code has Requirements For Belowground Vault Disposal Of Low Levet been to estimate the effects of operator actions on the timing Radoactiva Waste WARRiNER.J B ; DENNETT.R 0 Army-and course of events dunng the part of the sequence lea 4ng Dee of Amy Wne Waman Exemum 9 awn. Och up to but not including severe fuel damage. This report docu- ber 1985 93pp 8511110452 33412 224 ments the basis of the methods used to simulate the response The stud / reported herein contains the results of Task 2a of reactor vessel, pnmary coolant system, pnmary containment. (Technical Requirements for Belowfound Vault Disposal of and other reactor systems, the output from a sample use of the low-level Radoactive Waste) of a fountask study ertit;ed "Cn-code is presented tena for Evaluating Engineered Facihties " The ovmall objec-
~~** ' Y '* *" ' ' ""~~
NUREG/CR 3767: INTERACTIVE SIMULATOR EVALUATION * * " FOR CRT GENERATED DISPLAYS. 'O* #
~~BLACKMAN.H S -~~
GILMORE.W E. EGaG. Inc. January 1985. 43pp. 8502210188 methnds are available to potential license apphcants The be-EGG 2308. 29035 284 towgwd vaW esp sat ayernatwo is % f samal nwhods The United States Nuclear Regulatory Commesson (USNRC) that may bc porosed for d sposal of low level radoactive is sponsonng an on-gomg research program to develop meth- ***I" N " M N # N O"N "" # # *
ods of assessing vanous types of computer-generated esplays to a near-surface dsposal alternative in which the wastes would currarmy being implemented in nuclear power plant control be esposed of in vaults constructed below ground in encava-rooms. The purr.,ose of this report is to present an interactivt tions and covered with so,t The esper ence and knowledge simulation technique for the evaluation of computer. generated 93'Nd with this method are descnbed and updated in this esplays. The independent vanables for thes esponment were report A genonc descnption of the featurens aruf components transient type (six levels), and esplay type including the levels and operation of a belowground vauft d sposal facihty is proved.
of star + control panel, bar e control panel, meter & contros ed Features and components that Could enhance tne long term panel, pressure temperature map + control pancl. and cor: trol perfor .ance are descnbe 1. 6nclueng site condtions for which panel only. The dependent measures were deviations of param. they would be apphcable The apphcabihty of esisting cntena de-eter values cornpnsing the safety functions at nsk, percent of veloped for near surface dsposal (10 CFR Part 61 Subpart D) time these parameters were out of tolerence from onset of the to the belowground ve Jit esposal method, as assessed in Task transient, and accuracy of the operator path in transient mitiga. t, are reassessed herein. With few exceptions, these entena t on. The results todcate that an interactive simula%n method were found to be applicable in the reassessment. These conclu-can be used to evaluate varcus d splay types, and that the sions dffer slightly from the Task I findngs )
42 Main Citations and Abstracts NUREG/CR-3774 V03: ALTERNATIVE METHODS FOR DISPOS- is one of several methods that may be proposed for disposal of AL OF LOW LEVEL RADIOACTIVE WASTES. Task 2B-Technical low-level radioactive waste. In this report, the term shaft dispos-Requirements For Aboveground Vault Disposal Of Law Level al refers to a near-surface disposal alternative in which the Radioactwe Waste. BENNETT.R D.; WARRlNER JB. Army, wastes would be disposed of in shafts or boreholes augered. Dept. of, Army Engineer Waterways Espenment Station. Octo- bored or sunk by any other conventional method. The expen- l ber 1985. 89pp. 8511110459. 33412:135. ence and knowledge gained with this method are descnbed and The study reported herein contains the results of Task 2b updated in this report. This includes expenence in the use of (Technical Requirements for Aboveground Vault Disposal of shafts for storage in the U S. and Canada and research with the Low-Level Radioactive Waste) of a four task study entitled "Cn- method both here and abroad. A genenc desenption of the fea-tzna for Evaluating Engineered Facihties." The overati objective tures and components that could enhance the long term per-of this study is to ensure that the critena needed to evaluate formance are descnbed, including site conditions for which they fNe alternative low-level radioactwe waste (LLW) disposal meth- would be appbcable. The apphcabihty of existing cntena devel-ods are availabe to poential license applicants. The above- oped for near surface disposal ( 10 CFR Part 61 Subpart D) to ground vault disposal alternative is one of several methods that the shaft disposal method, as assessed in Task 1, are reas-may be proposed for disposal of low level radioactrve waste. In sessed herein. Without esception, these entena were found to this report, the term aboveground vault refers to an engineered be apphcable in the reassessment. These conclusions differ structure with roof, walls and floor enclosing the disposal space. slightly from the Task i findings The limited exponence and knowledge gained with this method are desenbed and updated in this report. The short term expen- NUREG/CR 3791: CLOSEOUT OF IE BULLETIN 7949 FAILURE ence does not conclusively demonstrate tr.e capability of this OF GE TYPE AK 2 CIRCUIT BREAKERS IN SAFETY RELATED method to satisfy the Part 61 Performance Objectwes. A gener- SYSTEMS. DEAN.R S ; FOLEY,W J.; MILLS.W R ; et al. Param. ic descrption of the features and components and operation of eter, Inc. January 1985 47pp 8501280735 IEB 79 09 an aboveground vault disposal facility is provided Features and 28629 130. components that could enhance the long-term performance are Twelve fadures of General Electnc Type AK 2 safety-related desenbed. The apt.!icabihty of existing cntena developed for circuit breakers reported in 1975,1978 and 1979 are desenbed near-surf ace disposal (10 CFR Pe t 61 M,tspa,1 C) to W.e Lbo.6- ,ggg7g g g g g,g ground vault disposal method, as a.sessed in Task 1, are reas- issued Apnl 17,1979 to require responses and specihc actions sossed herein. With few exceptiom, these entena were found t by all licensees and holdets of construction permits The fail-be apphcable in the reassessment. These concluseons differ , shght!y from the Task 1 find,ngs. g NUIEG/CR-3774 V04: ALTERNATIVE METHOD FOR DISPOSAL faulty adjustment of that linkage mechanism it was concluded OF LOW LEVEL RADIOACTIVE WASTE. Task 2C. Technical Re- that the twelve failures resulted from inadequate preventive quirements For Earth Mounded Concrete Bunker Disposal Of maintenance Because many occurrences of the same kind Low Level Radioactive Waste, MILLER.W O.; BENNETT,R.O happened after 1979, a significant number of later NRC docu-Army, Dept. of Army Engineer Waterways Expenment Station. ments whech are included in Appendia A were issued The bulle-October 1985 96pp. 8511110415. 33417 218- tin has been closed out for 101 of the 129 current facilities The study reported herein contains the results of Task 2c which reported either that they had no Type AK 2 breakers in (Technical Requirements for Earth Mounded Concrete Bunke' safety-related systems or none with undervoltage inp devices Disposal of Low-Level Radioactrve Waste) of a four-task study Proposed followup items for the remaining 28 current facihties entitled "Cntena for Evaluating Engineered Facihties " The over- are presented in Appendia C. Because followup is based on the til ob iectue of this study is to ensure that the cntena needed to requirements of later Bulletins 83 04 and 83 08 Bulletin 79-09 evaluate five afternatve low level radioactue waste (LLW) dis- is considered closed. posal methods are availabio potential license apphcants. The ezrth mounded concrete bunker disposal alternaitve is one of NUREG/CR 3794: CLOSEOUT OF IE BULLETIN 80-25 0PERAT-several methods that may be proposed for disposal of low-level ING PROBLEMS WITH TARGET ROCK SAFETY RELIEF radioactive waste, The name of this afternatrve is desenpt:ye of VALVES AT BWHS FOLEY,W.J ; HENNICK.A. Parameter, Inc. the disposal method used in France at the Centre de la January 1985 44pp. 8501280090 PARAMETER IE 13 Manche Exponence gained with this method at the Centre is pg$ 743 $ 7, descnbod, including unit operations and features and compo- Dunng the thri.e-month penod beginning July 25,1980, five nents. Some img.rovements to the French systern are recom- events occurred involving two types of malfunctions of Target mended here n, including the use o previous backfrit around mo- Rock safety rehof valves at Boston Edison Company's Pilgnm mohths and entend,ng the hmits of a low permeabihty surface Nuclear Power Station Unit 1. The first three events were layer. The appbcability of existing critena developed for near. caused by direct failures of the vatves, the last two events were surface disposal (10 CFR Part 61 Subpart D) to the earth caused by nr.rogen supply system overpressure which led to mounded concrete bunker disposat method, as assessed in valve failure IE Information Notice 80-40 was issued November Task 1, are reassessed herein. With minor Qualifications, these 1,1980 to call attention to the two rttrogen overpressure entena were found to be apphcable in the reassessment. These events As a result of all five events, IE Dulietin 80 25 was conclusions differ shghtfy from the Task 1 findings. issued December 19, 1980 for action to alt 30 BWR facihties with operating licenses or near term operating liconses, and for NUREG/CR-3774 V05: ALTERNATIVE METHODS FOR DISPOS. information only to 24 facihties then under constructon Actions AL OF LOW LEVEL RADIOACTIVE WASTE. Task 2E Technical worn to be taken with respect to (1) all Target Rock two stage. Requirements For Shaft Disposal Of Low Level Radioactive Waste OENNETT,R D. Arrny, Dept. of, Army Engineer Water- pdot. operated safety rebet valves (SRVe!, (2) any make or model of SRV wNch fails to function as designed, encepting for ways Esportment Staton. October 1985 105pp. 8511110400 pressure setpoint requirements and (3) SRV nitrogon/ air supply 334ty03f, The study reported herein contains the results of Task 2e systems. Upon evaluation of utikty responses and NRC inspec. tion reports. the bulletin has beer closed out for eight of the 30 (Technical Requirements for Shaft Disposal of low-Level Radio- facilfics to *Nch tne bulletin was issued for action For use by actue Waste) of a four-ta@ study entitled "Cotena for Evaluat. NRCHE, fullowup items for 22 current facihtees with open t utle-ing Engineered Facihties" The overall obtec3ve of this study is tin status are proposed in Apperds C Remaining areas of con to ensure that the entena needed to evaluate ftve attemerve corn and continuing actions deakng with them arn desenbod low-level radioactive waste (LLW) disposal methods are avail- The bulletin has served its purpose by result:ng in identification able potentral license appbcants. The shaft disposal alternat,vo
Main Citations and Abstracts 43 of the need for Correctrve actons at all of the 27 Current operat. Laboratory's Applied Physics and Components Technology Divh ing facihties to which the bullet n was issued for action. sons for the Division of Reactor Safety Research in the U S. NUREG/CR 3802: RELAPS ASSESSMENT:OUANTITATIVE KEY Nuclear Regulatory Commission. The work in the Applied Phys-PARAMETERS AND RUN TIME STATISTICS. KMETYK,L.N ; ics Dwison mcludes mports on reactor safety modemg and as-BUXTON.L.D.; THOMPSON S L. Sandia National Laboratones. usmM W men d N %N Sa% Wam h February 1985. 30pp. 8503210472. SAND 841013. 29479 072. tion. Work on reactor core thermal-hydrauhcs is performed at The advanced best-estimate systems codes currently being ANL's Components Technology Division, emphasizing 3-dimen-developed for the NRC are designed to provide reahstic, rather sional code development for LMFOR accidents under natural than conservative, predictions of LWR plant behavior dunng a convection conditions An executive summary is proved J iri-vanety of accidents and transients. The RELAPS independent etuding a statement of the findings and recommendat ons of the code assessment protect at Sandia Naticnal Laboratones is part PO'l of an overati code assessment effort funded by the NRC to arrrve at a qualified judgment of the accuracy with which these NUREG/CR 3805 V02: ENGINEERING CHARACTERl2ATION OF codes can predict the detailed thermal / hydraulic response of GROUND MOTION Task II. Effects Of Ground Motiort Charac-LWRs dunnq accident and off-normal conditions. The heart of tenstics On Structural Response Considenng Locahred Structur. the assessment project involves entensive comparison of code at Nnnhneantes And Sod-Structure interaction Effects results with test data RELAP5/ MOD 1 has been assessed at KENNEDY,R P.; KINCAID.R H ; SHORT,S A Structural Mechan-Sandia against a vanety of test data from both integral and sep- ics Associates. March 1985.158pp. 6504030436. 29604105. arate effects test facilities. All these anafyses have been docu- g g, mented in detail in individual topical reports and in an overall the engineenng charactentation of earthquake ground moton summary and conclusions report. In this paper, we tabulate the for nuclear power plant design. Task I of the study, which is quantitative key parameters for those scenanos which involve presented in NUREG/CR 3805, Vol 1, developed a basis for leanage of coolant from the pnmary system to containment, as selecting dessgn response spectra taking into account the char-well as the run time statistics for att the analyses we have per, acteristics of free-field ground motion found to be significant in formed. causing structural damage Task it incorporates add,tional con. huFEG/CR4603. TH2 EFFECTS OF PO3T40CA COfsDIT; Ohs siderations of effects of spatial vanations of ground mottons and x,,,.steucture inaraction on lounoation mot.om aaa strea. ON A PROTECTIVE COATING (PAINT) FOR THE NUCLEAR response. The results of Task 11 are presented in four parts (1) POWER INDUSTRY. LOYOLA V M : WOMELSDUFF.J E. Sandia effects of ground motion charactenstics on structural response National Laboratones. May 1985. 50pp 8505230538 SAND 84-of a typical PWR reactor budding with locahied honlinearities 0806. 30549 014. and sod structure interaction effects, (2) empincal data on spa-We have studied the oxidation of zinc in a rinc nch coatin9 taal vanatens of earthquake ground motion; (3) sod structure used in the nuclear power industry and have measured the rates cf hydrogen generaten from these coatings due to anc enteraction effects on structural response; and (4) summary of evidation at temperatures of up to 175 degrees centigfade. The conclusons and recommendations based on Tasks I and il studies This report presents the results of the first part of Task results suggest that the real time rates of hydrogon generation 11 The results of the other parts will be presented in NUREG/ are considerably higher than previously believed The higher CR 3805, Vols 3 5 rates rneasured in this study are probably due to differences in empenmental mothodologies between this and prevous studies. in this study, the measurements were real-time measurements. NUREQ/CR 3810 V03: REACTOR SAFETY HLbEARCH PROGRAMS Ouarterty Report.Jufy-September 1964 as opposed to time-averaged values which are typically ob- EDLER.S K, BatteHe Memonal Institute. Pacific Northwest Lab-tanned. The results suggest, as have the results of other invests- oratones. February 1985 35pp 8503130105 PNL 5106 3 gators, that the measured rates and reaction parameters may 29360-151. not be those of any specific reaction, but are instead the " effec- This document summantes work performed by Pacific North. tive" values of a seres of comples systems operating together. west Laboratory from July 1 through September 30,1964. for However, the total quantity of hydrogen generated by this the Division of Accident Evaluation and the Div+on of Enge-mechanism is significant!y less than can be produced from neenng Techno6ogy, U S Nuclear Regulatory Commission. Re-other sources, e g., steam; Urconium suits from an instrumented fuel assembty irradiateor* program NUREG/CR 3804 V03: PHYSICS OF REACTOR being performed at Hafden, Norway, are reported Accelerated SAFETY.Ouarterly Report. July September 1984
Argonne Na- peHetclaoing interaction modeling is being conducted to pre-tionat Laboratory. January 1985. 23pp 6502210261. ANL 84 35 dict the probability of fuel rod failure under normat operating V03. 20057.325. c netons Espenmental data and analytical models are being This quarterly progress report summanzes work done dunng P'0V'ded to ard in decision making regarding pipe to pipe im.
the months of July September 1984 in Argonne National Lab- pacts fonowing postulated breaks in high-energy flunt system oratory's Apphed Physics and Components Technology Deve- piping Fuel assembbes and analytical support are being proved-tions for the Divison of Reactor Safety Research m the U S ed for empenmental programs at the Powot Burst Facihty, Idaho Nuclear Regulatory Commission. The work in the Applied Phys- Natenal Engineenng Laboratory, Idaho Falls Idaho. High tem-ics Orvision includes reports on reactor safety modehnq and as- porature matenals property tests are being conducted to pro-sessment by members of the Reactor Safety Appraisa!$ Soc- V'de data on sovere core damage fuel behavior T hermal by-tion Work on reactor core thermal hydrauhes is performed at draubc models are being developed to provide better digital ANL's Components Technology Division, emphasizing 3-dimen. codes to compute the behavior of fuH scafo reactor systems sional code development for LMFOR accidents under natural under postulated accident cordtions Severe fuel damago acci-convection conditions. An esecutive summary is provided in- dent tests are being conductnd in the NRU Reactor, Chark cluding a statement of the findings and recommendations of the River, Canada report. NUREG/CR 3810 V04: REACTOR SAFETY RESEARCH NUREG/CR 3804 V04: PHYSICS OF REACTOR PROGR AMS Quarterty Report, October December 1984 SAFETY.Ouarterty Report. October December 1984 ' Argonne EDLER.S K Dattelle Memonal !nstitute. Paafic Northwest t ab. National laboratory. Apol 1985 20pp 8504250268 ANL'84-35 oratones May 1985 3fpp 8506140415 PNL 5106 4 VO4. 30032 091. 30908 t88 This quarterfy progress rnport summanies work donc dunny This document summanies work performed by Pacific North. the months of October December 1984 in Argonne National west Laboratory from October 1 through December 31, 1984,
44 Main Citations and Abstracts for the Division of Accident Evaivation and the Division of Enge censing decisions are made. This report descnbes progress in a neenng Technology, U.S. Nuclear Regulatory Commission. Re- number of activities dealing with current safety issues relevant suits from an instrumented fuel assembly arradiation program to both light water and breeder reactors. The work includes a being performed at Halden Norway, are reported. Accelerated broad range of expenments to simulate accdontal conditions to pellet cladding interaction modehng is being conducted to pre- provide the required data base to understand important accb dict the probability of fuel rod failure under normal operating dont sequences and to serve as a basis for development and I conctions. Expenmental data and anarytical models are being venfication of the compteu computer simulation models and provded to a d in decision making regarding pipe to-pipe im- codes used in accident analysis and hcensing reviews. Such a picts following postulated breaks in high-energy fluid system program must include the development of anafytical moders, piping. Fuel assemblies and anaWeat support are being provid- venfred by esponment, which can be used to predict reactor ed for expenmental programs at the Power Durst Facihty, Idaho and safety system performance under a broad vanety of abnor. National Engineenng Laboratory, Idaho Faus, Idaho. High-tem- mal conditions. perature matena!s property tests are being conducted to pro-vide data on severe core damage fuel behavior. Thermal hy- NUREG/CR 3818 VOC REACTOR SAFETY RESEARCHOuarterly draubc computer programs are provdinq best-estimate analyses Report. October. December 1984.
~~Sandia National Laborato-for a vanety of safety issues in hght. water reactors Severe fuel nos September 1985 242pp. 8510040365. SAND 84 1072.~~
damage tests are being conducted in the NRU Reactor, Chafk 32855 238 Arver, Canada. Sandia National Laboratones is conducting, under the sans t , nome cal msearch Mated to NUREG/CR 3816 V01: REACTOR SAFETY RESEARCH Ouarterly ' ' * * * * ' " " ' E**'" ****#" Report. January March 1984
Sand:a National Laboratones includes expenments to simulate the phenomenology of the ac-Janua'Y 1985 160pp 8501300074. SAND 84-1072 28680 095- cident conditions and the development of analytical models.
The overall objectve of this report is to provide NRC with a venfied by openment, which can be used to piedict reactor comprehensive data base essential to (1) defining key safety and safety systems pe<formance and behavior under abnormal isaues, (2) understanding risk significant accident sequences, conditions The objective of this work is to provde NRC reque (3) developing and ventying models used in safety assess- site data bases and analytical methods to (t) dent ry and def.ne ments, and (4) assunng the public that Ocnet reacter systne safsty issues, G) t,r.d>stanj the p ogvssion e r nsospiricant will not be hcensed and placed 6n commercial service in the accident sequences, and (3) conduct safety astiessments The United States without appropnate consideration being given to collective NRC-sponsored effort at Sandia National Laboratones their effects on health and safety. This report descnt'es is directed at enhancing the technology base supporting bcons-progress in a number of activities dealing with current safety ing decis<ons. issues relevant to both light water reactors (LWRs) and breeder reactors The work includes a broad range of expenments to NUREG/CR 3817: DEVELOPME NT,USE AND CONTROL OF simutate accdental conditions to provide the required data base MAINTENANCE PROCEDURES IN NUCLEAR POWER f to understand important accident sequences and to serve as a PLANTS Problems And Recommendations. basis for development and venfication of the complex computer MORGENSTERN,M.; BARNES V E ; HADFORD.L.R ; et al. Dat. simulation models and codes used in accident analysis and h- te,te Human Affairs Research Centers. January 1985 120pp censing reviews Such a program must include the development 8501280403 PNL 5121. 28572.136 of analytical models. venfied by expenment, which can be used This report descnbes the results of activities conducted to to predct reactor and safety system pertormance under a broad assess and document the need for guidance of regulatory in-vanety of abnormal conditions- volvement by the Nuclear Regulatory Commission (NRC) en the d'v00**"' upgrading, use and control of maintenance proce. NUREG/CR 3818 V02; REACTOR SAFETY RESEARCH Ouarterly dures in U S nuclear power plants. Reported are the findings of Report.Apni-June 1984
Sandia National Laboratones Apol the following four actmties. (1) a survey of current maintenance 1985 211pp. 8504160554 SAND 841072 29825153 Procewre pracket in seven U S nuclear power plants, (2) a This report desenbes progress e a number of activities deal. Y'ow and anal ysis of plant administrative and maintenance ing with current safcty issues relevant to both light water reac. procedures. (3) a survey of maintenance procedure practices in tors (LWRs) and breeder reactors The work includes a broad sndustnet that share some characteristlCs with the nuclear in.
range of empenments to simulate accidental condtjons to pro, dustry, and (4) a review of the research periaining to job per-vide the required data base to understand important accident formance a ds and a boet anahsis of their applicabihty to main. sequences and to serve as a basis for development and vento tenance in nucWar power plants Bamj on thew findings, sov. cation of the comples computer simulation models and codes eral recommendatons for NRC action to upgrade maintenance < uced in accident analysis and licensing reviews Such a pro. procedure programs are offered gram must include the development of analytical models, vers fsed by empenment. *Nch can be used to predict reactor and NUREG/CR 3819: SURVEY OF AGED POWER PLANT FACILf-safety system performance under a broad vanety of abnormal TIES ROSE.J Aa OEWALL,K G ; STEELE R , et al EG4G conditions Idaho, inc. (subs of EG40, inc ) July 1985. Sipp fl507250162. NUREG/CR 3818 V03: REACTOR SAFETY RESEARCH Ouarterfy EGO 2317. 31786 06t l Report. July September 1984
~~Sandia Natonal Laboratones This report presents the results of the survey of Agod Nucicar Power Plant F acht,es conducted for the USNRC Office of No.~~
July 1985.190pp 8507250114 SAND 641072 31787.147. cicar Regulatory Research The results of this report recom. Sanda National Laboratones es conductino under USNRC s mend methods to heip formulate comprehensive research pro- ' sponsorship, phenomenological research related to the sa'ety of commercial nuclear power reactors The overalt ob iectives of gram theit wilt systematically afontify aging and service wear ef-this work is to provide NRC a comprehenstvo data baso essen- fects which are likely to affect plant safety. The survey centered Ital to (1) de6ning key safety issues, (2) understandng nsk sig- on safety related plant systems with regard to component fa5 nif icant acc dent sequences. (3) developing and vonfring ures from nperating histones The age related fadure information models used m safety assessments, and (4) assunng the pubhc gainered from the plant histones was analyted for reoccurnng that power rcactor systems wdl not be hconsed and placed in failure patterns Emphams was on 6dont;fication of specific commercial sery ce in the United States without appropriate eqwpment with high failure rates and of failure mechanism rela. consideraton boing given to their effects on health and safety tionships The data would not support spoofic equipment wiento ficaton, it ed empty a drect relationship between failure and fail-Together with other programs, the Sanda effort is directed at ute mechanism 70*% of the failures reportnd worn due to four assuring the soundness of the technology base upon which k-
Main Citations and Abstracts 45 failure mechanisms in addition there appeared to be a strong correlation between cause of failure and the system in which NUREG/CR 3830 V02: AEROSOL RELEASE AND TRANSPORT PROGRAM Semiannual Progress Report For Apnl 1984 Sep-the component operates. This is verified by detailed study of tember 1984 ADAMS.A E ; TOBIAS.M L Oak Ridge National several plant systems and corroborated by personnel inter. Laboratory January 1985. 50pp. 8502210193 29035 231 views. This survey endicates identification and elimination of This report summantes progress for the Aerosol Release and system level cause of component fadure is a viable approach to Transport Program sponsored by the Nuclear Regulatory Com-prevent and m;tigate major reported aging effects mission. Office of Nuclear Regutatory Research. Divison of Ac. NUREG/CR-3820 V03: THERMAL / HYDRAULIC ANALYSIS RE- cadent Evaluation, for the penod April 1984-September 1984 SEARCH PROGRAM.Ouarterly Report. July-September 1984 Topics discussed include (t) the experimental program in the THOMPSON.S L Sandia National Laboratones Apnl 1985. Fuel Aerosol Simulant Test Facihty, (2) NSPP espenments in-05pp 8505140172. SAND 841025. 30451.130. volving an aerosol of kmestone amgate concrete in a steam The TRAC-FF1/ MODI independent assessment program at air atmospWe @ revisons in the NSPP empenmental program. Sand:a National Laboratones es part of a multi-faceted effort W eenments mahng to N Nmo %drauhc conddont FA sponsored by the Nuclear Regulatory Commission to determine aerosobmcisMe interaction test pians. (6) aerosol code impie. the abihty of vanous systems codes to predict the detaded ther. mentaton acWes, m improvements in data procesong proce-mal /hydrauhc response of LWRs dunng accident and off-normal dures for NSPP esperiments, and (8) a study comparing in-conditions. This program is a successor to the RELAP5/ MOD 1 vesse and enetw cascade impactor aerosol sie measve. independent assessment proiect underway at Sandia for the ** * * '" Ih* # last two years The TRAC PFt/ MOD 1 code wdl be assessed against data from vanots integr61 and separate effects empen- NUREG/CR 3831: THE IN. PLANT RELIABILITY DATA BASE mental test facihties, and the calculated results w4 also be FOR NUCLEAR PLANT CCVPONENTS intenm Report D4esel compared with results from our previous RELAP5/ MOD 1 inde- Generators.Datter es. Chargers And Inverters x AHL W K -, pendent assessment analyses whenever possible BORKOWSMR J Oak fi dge National Laboratory Februa y 1985.10900 8502250363 ORNL/TM 9216 29094 248 NUREG/CR 3825 V03-4: ACOUSTIC EVISSION/ FLAW RELA. The objectrve of the in Plant Rebsaihty Data CPRD) program TIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR is to devesco a comprocenwe, component 4pecific rekatA4 FRE55LAE VE55El.S Cuarterly Heport. Apnl 1984 September cata taas for procaciuntic nsk assessment and for other statisti-1981 Volumes 3 and 4 HUTTON,P H : KURTZ.R J. Batteile Me. cal analyses relevant to con pcroat robabihty evar u atons This monal Inst;tute, Pacific Northwest Laboratones March 1985 cb;ectrve is being attained through a cooperat:ve eaort with sev-22pp 8504090010. PNL 5125 29754 018 eral utikties which have provided access to maintenance fdes Technical progress toward continuous acoustic emission and pert.nent population information This odot data base in-rnonstonng of nuclear reactor pressure boundanes for fla* de- cfudes (1) a corrponent populanen hst (for each piant) or SMect-tecton is desenbed for the penod Apni September 1984 A draft ed efectromechanical and mechanical egwpment te Q , purep. report of Z81 vessel test results was completed Growth of varves. etc ). and (2) comprehensive compor'ent fa ture and (nachined flaws was detected by AE dunng both 65 degree C repa.r histones based on corrective ma.ntenance act ons on and 285 degree C testing AE data was generany preportional these components This document is the product CI a pJot to crack growth A key resutt was clear detection of a naturat study that was undertamen to demonstrate the methodology and Crack in a fabncation weld by AE Crack growth rates estimated feas,bei% of applying IPRDS techniques to develop and ana ge from AE data compared well with measured crack growth rates the rehab hty characteristics of key electncas components in fue in service hydro test monitonng gave mixed results Impending nuclear power plants These electrical components include fadure cond:t.ons are readdy detectable Hewever, with low d4esel generators. battenes, battery charges and inverters The overpressure (1.15 m operating pressure). flaws as deep as 70*. sources used to develop the data base and produce the com-through-wall did not produce significant AE With higher over. ponent fadure rates and mean repair times were the p' ant pressure (1.4 s operating pressure), flaws produced identifying equ pment bsts, plant drawings ma,ntenance work requestt AE An engineenng prototype AE monitor system has been Final Safety Anafysis Reports FSAHs). and interviews w'tn completed for use in operational monitonng at Watts Bar Unit 1 plant personnel The data spanned approssmatefy 33 reactor-reactor A modahed approach to Crack growth AE signal sdentife- years of commerciat operation cat:on is producing about 95*. correct determ. nations on re-corded waveforms from the ZB-1 vessel test A report on re. NUREG/CR.383h MULTIPLE-SEQUENTIAL FA#LUAE suits from AE monitonog hot functional testing at Watts Bar Un,, MODEL Evaluahon Of And Procedures For Human Error De-1 has been pubbshed pendency SAMANTA.P K , O D AIEN J N . MOARISON H W Brookhaven National Laboratory May 1985. flapp NUREG/CR 3829: AN EVALUAftON OF THE STABILITY TESTS 8512270358 DNL NUREG 51786 34030 203 RECOMMENDED IN THE BRANCH TECHN CAL POSITION ON This report provides an evaiuat,on of the practicahty. accept. WASTE FORMS AND CONTAINER MATERtALS BOWERMAN 8 S ; SWYLER.KJ ; DOUGHERTY.O R ; et al. abihty, and usefulness of uSng the Muftiple Sequent at Failur e Brookhaven National Laboratory March 1985 (MSF; model ong.nany descnbed in NUHEG/CH 2111, 1981 167pp. The MSF model is descobed. discussed. and proceores for its 8503280017. ONL.NUREG 51784 29548 t42. use provided The model was found to be practical, acceptable. The Technical Position on Was'e Form and Container Maten- and useful as a PHA tool for asscS9ng the dependence due to cle (TP) provides guidance to generators of low-level rad.oactive waste for meeting the regulations under 10 CFR Part 61 gov- human interact,ons with components en systems employing re-dundant components orning the disposal of these wastes Testing methods are rec. ommended in the TP for assessing matenal properties relevant to long term performance in shanow land bunal. These tests NUREG/CR 3851 V03: PHOGAE SS IN EVALUATION OF HADIO. were reviewed with respect to their apphcation to specific mate. NUCLIDE GEOCHEMICAL INFO 4VATION OEVELOPED BY DOE HIGH LEVEL NUCLEAR WASTE HESPOSITO4v' SITE nats cement, bitumen, vinyt ester styrene. and polyethylene in PROJEC TS Deport For Apr i Junn 1984 kELMEAS A D; some cases. the apphcabMy of the tests was found to be inad- ARNOLD,W 0. MEYEH R E ; et al Oak Hidge National Labora-equale, and modifications to the esisting tests or afternatrve tory January 1985 43pp 8501280387 OONL/TM 9191/V3 rnethods were recornmended An esperimental evaluation of 28572 255 one of the recommended biodegradation tests (the Bartha. Goochemical information rnlevant to the retention of ra4onu. Pramer methodl was also carned out Cond-tions under which chdes by candidate high level waste repositones bang devel-this test should be conducted are recommendod oped by Department of Energy (DOE) prolmts is being evaluah
46 Main Citations and Abstracts ed by Oak Ridge National Laboratory (ORNL) for the Nuclear of the entena wdl ensure the structural integnty of shipping con-Regulatory Commission (NRC). Dunng this report penod, the tainers at levels consistent with the radioactive materiars being project has evaluated radionuchde sorption and solubehty values transported. (* NOTE. Apphes to all metals used in shipping con-apphcable to the candidate repository site in the Columbia River tainers construction except cast irons ) basalts at the Hanford Reservation. The removal of technetium from pertechnetate-traced groundwater by McCoy Canyon NUREG/CR-3855: CHARACTERIZATION OF NUCLEAR REAC. bisa! under anoxic rMov conditions (air excluded) at 27 de- TOR CONTAINVENT PENETRATION - FINAL REPORT. SHACKELFORD.M. Argonne Nabonal Laboratory
Sanda Na- j grees centigrade ca4 found to be sensitive to the groundwater tional Laboratones. Apol 1985. 38tpp. 8505060534. SAND 84-composition. Sorption of uranium from groundwater by McCoy l Canyon basalt under omic redom cond:trons at 60 degrees centi- 7139.30192.018.
This report summanzes the survey work conducted by Ar-grade showed low sorption ratios (1.8 to 2.4 L/kg) similar to those previously obtained at 27 degrees centigrade. The aver- gonne National Laboratory on the dosgn and details of malor , age sorption ratio for strontium in groundwater onto McCoy penetrations in 48 nuclear power pfants. The survey includes aff l Canyon basalt under omic redox conditions at 27 degrees centi- containment types and matenals in current use. It afso includes details of afl types of penetrations (cucept for electncai penetra. grade was 225 L/kg. Column chromatographic expenments with neptunom in groundwater to measure retardation factors at tion assembhes and valves) and the seals and gaskets used in tsmperatures from 25 to 80 degrees centigrade gave calculated them The report provides a test matrix for testing mator pene-sorptron rat:0 values that were in good agreement with the trations and for testing seals and gaskets in order to evafuate values previously obtained in batch contact tests their leakage potentiat under severe accident cond:trons, NUREG/CR 3851 V04: EVALUATION OF RADIONUCLIDE GEO- NUREG/CR 3862: DEVELOPMENT OF TRANSIENT INITIATING CHEMICAL INFORMATION DEVELOPED BY DOE H!GH- EVENT FREQUENCIES FOR USE IN PROBABILISTIC RISK LEVEL NUCLEAR WASTE REPOSITORY SITE ASSESSMENTS MACKOWIAK.D P.; GENTILLON C D - ' PROJECTS Annual Progress Report For October 1983-Septem' SMITH.K L EGaG Idaho, Inc. (subs of EG&G. Inc ) May 1985' ber 1984. KELMERS A D.; KESSLERJH4 SEELEY,F G : et al. 278pp. 8506240069 EGG-2323 31150 002 ansient inmahng went hquencies are an essential mput to 5092 79 O L/TM 9191 260 66 the anatysis process of a nuclear power plant probabihstic nsk Geochemical information relevant to the retention cf radonu- assessNnt Rese Wqwncies descnbe events causing or re-chdes by canddate high-level nuclear waste geologic repositor- quinng scrams This repod docunwnts an eMod to vaf i date and 4s being charactenzed by Department of Energy (DOE) update frorr other sources a computer based data file devel-projects is being evaluated by Oak Ridge National Laboratory oped by the Electrc Power Research Institute (EPRI) describing (ORNL) for the Nuclear Regulatory Commission (NRC) Empha- $d evens at M Wed Wes comNcial nucMar power sis has been given to the espenmental evaluat.on of key racon- p anh Operamg informat.on from the United States Nuclear uchdes relevant to the Hanford Ste being charactented by the Regulatory Commission on 24 additiona! plants from their date Basait Waste Isolation Protect (BWIP) In work by the BWIP, hy. of commercial operat.on has been corr:bined with the EPRI dra2:ne was added to groundwater to simulate the reducing data, and the entire data base has been updated to add 1980 redor condition espected in the repository Such laboratory through 1983 events for all 76 plantt The vahdity of the EPRI methodo8ogy r"ay not adequatee y model in situ repository geo- data and data analysis methodology and the adequacy of the chemical conditions. We have beem employing anoxic redox EPRI transient categones are esamined New transient initiating condtons to allow the basaft to estabbsh the effect:ve redor event frequencies are derived from the espanded data base condtion in batch contact sorption esper ments Sorpton of using the EPRI transient categones and data display methods. Np(V) or TC(Vil) by basalt from synthetic groundwater under Upper bounds for these frequencies are a'so provided Addition-anovic redoit condtjons may involve chemisorption reduction re. al analyses espiore enaNos in the dominart transients. actions on the basatt surface Our sorption rat:0 for neptunium changes in transient outage Smes and their impact on plant op-under osic redor cond,tions does not compare favorabty *.th eration, and the effects nf power level and scheduled scrams the value pubbshed by the BWIP The pubbshed sciub:hty of on transient event frequencies A more ngorous data analysis technetium under the reducing redos condtions espected by methodology es developed to encourage further refinement of BWIP at the repository probably is based on caiculations involv- the transient initiating event frequencies derived herein ing inadequate thermodynamic data Under omic rodos condi. tions, our uranium sorption ratio was much lower than vafues re. NUREG/CR 3863: ASSESSMENT OF CLASS 1E PRESSURE ported by the BWIP A mineralogical and chemical charactenza- TRANSMITTER RESPONSE WHEN SUBJECTED TO HARSH tion was completed for the three basalt sampres used in our ENVIRONVENT SCRE ENING TESTS FURGAL,0.T , work. Significant d frerences were seen in both the quantity and CRAFT.C M . SALAZAR.E A Sandia National Laboratones. Apna composition of the mesostasis A potential defoency in the in- 1985 194pp 8506140052 SANO841264 30907 283 formation pubbshed by the BWIP is the absence Cf hthological An expenmental investigation into the performance of Class information as well as mineralogical and cr emical character:2a. 1E electronic pressure transmitters esposed to environments tion for the basait samples Our geochemicat modehng work w< thin and beyond the design basis was conducted Emphasis suggested that code to-code evaluation for geochemical calcu- was placed on determining the instruments
fa, lure and degrada-lations may be less important than a deta, led evafuation of the tion modes in separate and simultaneous environmental espo-data bases' sures. Five unaged ITT Barton Model 763 pressure transmitters were tested and esposed to a unique environment. The re-NUREG/CR 3854: FABRICATION CRITERIA FOR SHIPPING sponse of the transmitters showed that temperature was the pn-CONTAINERS FiSCHER L E : LAl.W Lawrence Lrvermore Na- mary environmental stress affecting the tested transmitters' per.
~~tional Laboratory March 1985 17pp 8504040417. UCRL. formance initiallarge errors that decrease with time.at tempura-53544. 29619 007. Cntena are #dentified for controlhng the fabncaton of metal~~
ture were observed The source of these errors is behewed to be a comrnon mode design weakness in the transmitters' cah-components of shipping containers used for transporting rado- bration potentiometers This weakness results from a depend-active matenals. The entena have been selected from the ency of mater al detectnc properties on temperature The modo AEME Code and are based on the levet of radioactive materits fication recommended t y the manufacturer, a'though padiative being transported and the nuclear safety function of the con- in nature did reduce this temperature. induced offect after the tainer's components Cr tena are ident;fied for fabricat.on proc- first few minutes of accident esposure A potent.af second esses which are related to rnatenals controls. forming heat Common failure mode which activates slowly with time at tem-treatment, esamination and acceptance testing Implementation
Main Citations and Abstracts 47 perature was also identified. The operation of this failure mech- are recorded at one-hour intervals on computer pnntouts for anism is believed to be catafyzed by the presence of a lubricant documentation and immediate analysis and one magnetic media used in the production of some potentcmeters The design of for permanent storage and subsequent analyses Results from this transmitter proved to be exceptionally hard to radiation ef- processing of data files show that the average temperatures at fects and there appeared to be no significant synergistic effects the 1/4T and 3/4T positions are maintained within 2 6 degrees between radiation and temperature. The observed responses of centig ade of 288 degrees centigrade with associated standard the transmitters offer support for the pos4 tion of IEEE 38t 1977 deviations of less than 3 degrees centigrade. Average tempera. which recommends that electronic modules aged to varying de- tures of the Cther thermocouples are maintained within 288 de-grees of advanced life should be tested-grees centigrade plus or minus 12 degrees centigrade with NUREG/CR-386S EVALUATION OF THE RADIOACTIVE INVEN- standard deviations less than 3 degrees centigrade TORY IN AND ESTIMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN EIGHT TRENCHES AT THE SHEF. NUREG/CR 3876: PROBABluTY BASED LOAD COMBINATION FIELD LOW LEVEL WASTE BURIAL SITE. MACKENZIE.D.R ; CRITERIA FOR DESIGN OF CONCRETE CONTAINMENT SMALLEY.J F.; KEMPF,C.R.; et al. Brookhaven National Labora- STRUCTURES. HW A NG.H.; K AG AMi,5 ; nEICRM ; et at tory. January 1985 196pp. 8503040014 BNL NUREG.51792 Brookhaven National Laboratory. August 1985 99pp. 29197.113. 8509110003 BNL-NUREG.51795. 32561213 An inventory has been compiled of the isotopes of half-hfe This report desenbes a research effort for the development of
>5 years buned in eight of the trenches at the Sheff. eld rado, the probabikty-based load combination cntena for design of active shipment records (RSRs). Pertinent information from concrete contain ent structures The proposed critena are in a some 3200 fuel cycle RSRs and 1700 non-fuel cycle RSRs has load and resistance factor design (LRFD) format in order to been stored in a computenzed data base and used to develop test the performance cbjectives of the proposed criteria, four the inventory. Results of the compilation are in disagreement representative structures are selected using a Latin hypercube with the two previous estimates for H-3 and C-14 In particular, sarnpling technique. Next, the rehabil.ty ana'ysis method devel-non fuel cycle H-3 inventory for the eight trenches of the oped by Brookhaven Natonaf Laboratory ss employed to assess present study is approximately a factor of 2 higher than either the reliability of these representative containments Further-prevous estimates of total site inventory. Modeling of release more, an ob l ective function is de6ned and a minimization tech-processes has been camed out in order to obtain estimates of nique is developed to find the optimum load factors The load isotopic release rates from waste packages to the trenches factors for accident pressure due to the design basis accident This modehng is highly speculative, but beheved to be state of. and safe shutdown earthquake are derived for three target limit the art. It required information not only on amounts of the differ- state probabilities Other load factors are also discussed on the ent isotopes, but also on the waste forms and containers hold. basis of pror empenence with probabelty based design entena ing them. Such information was generally not given on the for ordinary building construction The proposed load combina-RSRs and had to be cbtained by contact with the ger erators tons are based on the best ava:lable data to date pertarning to Estimated numencal release rate data are given for each trench loads and resistances. If in the future the data base changes.
for H 3, C 14 Cs-137, Sr 90, and Co-60.1129 is eupected to the developed methodology can readJy be utAred to update the have been totally released within a year of container breaching load factors resulting from these changes by corrosion. Most of the Pu, in the form of oinde, will probably not be released at a signif, cant rate. NUREG/CR 3883: ANALYSIS OF JAPANESE U S NUCLEAR HUREG/CR-3868: TRAC-PD2 INDEPENDENT ASSESSMENT- HUENEFELD.J C ; et al. BatteHe Memonal Inst >tute. Pacifrc KNIGHT,T.D. Los Atamos Scient,fic Laboratory March 1985 440pp. 8503220009 LA 10166-MS 29488 307. Northwest Laboratones June 1985 120pp 8507080183 PNL-This report documents the Los Alamos results of the second 5160 :" 392 345. assessment phase, independent assessment, for TRAC PD2' This report presents the results of a preject designed to com-We documented the results of the developmental assessment are and contrast Japanese and United States nuclear power for TRAC-PD2 in an earlier report. This report desenbes calcula- plant operating emperie ce, preventive maintenance / surveil-tions run with the released versions of TRAC-PD2. We analyzed lance requirements, and organeration and management prac-separate-effects tests to investigate the entical-flow calculation, teces relating to maentenance Findings are based on information the emergency core cooling (ECC) bypass behaver, and the re- Wained on the November-Occomber 1983 and November
,9g g g g g flood-tracking capability We analyzed integral tests to esplore the gravity dnven reflood behacor, and the small-break LOCA telle a Pacific Northwest Division, and on vanous documents ob.
behavior. The results show good agreement between the calcu-lated parameters and the data for those tests related to large- Trade and industry-MITI) dunno and subsequent to the visits break LOCA in general, the compensons to small-break LOCA U.S data sources included NUREG 0020 (Greybook) and plant tests indicated that the code can be useful but that some model technical specifications The study shows that Japanese plants emprovements are required. expenenced far fewer manual shutdowns, manual scrams, auto-mabc scrams, and reduced loads than U S plants and that their NUREG/CR 3872: DATA ACQUISITION AND CONTROL OF THE mean time between event (MTBE), even when adl usted for dit. HSST SERIES V IRRADIATION EXPER MENT AT THE ORR ferences in average plant availabihty, was approvimately to MILLER.L.F.; HOBBS.R W. Oak Ridge National Laboratory Apnl t+mes greater than the U.S MTDE. The report also points out 1985. 97pp 8505230574. ORNL/TM-9253 30547 232. significant differences in the Japanese approach to preventive Documentation relative to data acquisition and control for maintenance, and in the Japanese regulatory approach to main-support of the HSST Senes V trradiation Espenment at the Oak tenance, their management and organtiational contest for main. Ridge Research (ORR) is included in this report Part A de- tenance, and other socioeconomic factors that may artect the scobes the computer system hardware and real time application performance of maintenance support software, and Part B desenbes the temperature control methodology. Software that acquires data from analog input NUREG/CR 3885 V03: HIGH TEVPERATURE GAS-COOLED RE. provides this information to the coritrol algonthm software. Re. ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT suits from the control algorithm are, in turn, utilized by software EVALUATION Ouarterly Progress Report, July t September which controls digital output hardware Time intervals of execu- 30,1984. DALL.S J . CLEVELAND.J C , HARRINGTON.R M ; et tion, as well as sequencing of software modules, are controlled al Oak R,dge N ttonal Laboratory Apnl 1985 28pp through commands to the operating system Temperature data 8505230519 ORNL/f M 9287/V3 30549 063 1
48 Main Citations and Abstracts Modehng and code development work on the modular High- focused on dissolution /reprecipitation kinetics An expenment is Temperature Gas-Cooled Reactor (HTGR) were continued The planned to veri ^f this model. A procedure was developed to dis-longer term heatup accident scenano in which cavity wall cool- perse RuO(2) in MCC 76 68 glass Potentiodynamic polanzation ing is lost was also modeled. Sensitivity studies were run for a tests were performed to determine the effects of single chemi-vanety of parameter vanatens. Fiss on-product (FP) release and cal species in groundwater on the cracking and pitting suscepts transport expenments were completed for several ad4tional bety of carbon steel Slow strain rate tests show that carbon siements steel is especially susceptible to stress-corroson cracking in NU".EG/CR 3885 V04: HIGH TEMPERATURE GAS COOLED RE. aqueous FeCl(3) at low strain rates The strength of commercial l ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT hgh-punty iron was found not to be affected by hydrogen; how- ' EVALUATION Ouarterty Progress Report October 1-December ever, ductility was somewhat reduced The description of 31,1984 BALL,S J.; CLEVELAND.J.C.: HARRINGTON,R M.; et groundwater rad olysis was further refined dunng this quarter. al Oak Rdge National Laboratory. August 1985. 24pp Integral expenments are being prepared to provde information 8508210424. ORNL/TM-9267/V4. 32334 209 on combined effects processes that may influence the long. Modehng and code development work on the modular High" ] term performance of the waste package, , < Temperature Gas-Cooled Reactor (HTGR) continued with the development and testing of a thermal model of the upper reflec- NUREG/CR 3900 V03: LONG TERM PERFORMANCE OF MATE-tor. The longer term heatup accident scenano in which cavity RIALS USED FOR HIGH-LEVEL WASTE PACKAGING Ouarterly wall cochng is lost was also modeled Sensitmty sto -s were Report. October-December 1984 STAHL.D ; MILLER.N E Bat-run for vanatons in soil conductivity and decay heat g, stahon telle Memonal Institute. Columbus Laboratones May 1985 rate Fisseen. product (FP) release and transport expenments 88pp. 8506060746. 30776.173 were completed and initiated for several additional elements Espenments for evaluating the glass-dissolutron model are ogr s was made in estachshing an FP reestnbution capabil- , ing RuO(2) en MCC 76-68 g! ass has been tested and proved to NUREG/CR-3887: HUMAN FACTORS REVIEW FOR SEVERE produce appropnate particle concentratons. Acetic and humic ACCICENT SEQUENCE ANALYSIS KROIS P A: HAAS.P M ; acids have been chosen to test the effect of natural organic MANN!NG.J J ; et al Oak Ridge National Laboratory October acids on waste g! ass performance. In the overpack corrosion 1985.237pp 8512120160. ORNL/TM-9266 33871.139 effort, potentodynam.c polantation tests indicate that of tre 15 This report descobes a human factors research project per- chemical species tested. all but perchloate and hydrogen may formed to- (1) Support the Severe Accdent Sequence Analysis efeect stress-corrosion cracking behaver of carbon steel, sever. (SASA) program and (2) develop a descnptive model of cpera- al synergistic effects were also indicated. In slow strain rate tot response in accident managment The first goal was accom-phshed by working with SASA anatysts on the Browns Ferry 0N M %M 4 M W Unit One anticipated transient without scram (ATWS) accdent chlorde concentration than espected in groundwater) euhrbited sequence to systematically assess entical operator actions and S'9ficant cracking over the temperature range 250-315 de-thereby demonstrate contributions to SASA analyses from grees centigrado. Pits were found to propagate readdy, but human factors data and methods The second goal was accom- slowly, in 1018 carbon steel esposed to aerated basatt ground-plished by developong a model catted the Function Onented Ac- water at 90 degrees centigrade The general-corrosen correla-cadent Management (FOAM' model, which provides both a con- tion was changed to incorporate a finite rate of film growth Inte-ceptual structure linking off-normal safety functions with poten- gral espenments are being prepared to provde informaton on tial unconventonal emergency responses and a method for de- combined offects processes that may influence the long-term veloping technical gu dance for those responses based on oper- performance of the waste package ations engineenng, and human factors data and expertise. The fuur components compnsing the model are descnbed and their NUREG/CR 3900 V04: LONG TERM PERFORMANCE OF MATE-use is shown through a table-top demonstration. RIALS USED FCR HIGH LEVEL WA3TE PACMAGING Annual RepoaApril 1984 Apnl 1985 STAHL.D ; M.LLER.N E. Battelle NUREG/CR 3889: THE MODELING OF BWR CORE MELTDOWN Memonal Inst @te. Columbus Laboratones. July 1985 235pp ACC1 CENTS FOR APPLICATION IN THE MELPol MOD 2 COMPUTER CODE. KOH.B R ; KIM.S H ; TALEVARKHAN R ; et 850815004 7. BMI-2127. 32197.107. al. Oak Ridge National Laboratory May 1985 279pp Waste form esponmentaten has focusmj on boros hcate 8505160635 30444 230 glass, using the reference composition MCC 76M An esport-This report summanzes improvements and modf<ations ment investigated the influence of continuous contact between made in the MELRPt computer code A maior ofference be- the glass specimen and the teachate on the results of corrosen tween this new, updated verson of the code. CCed studes it was found that precipitates formed dunng coohng can MELRPl MOD 2, and the one reported prevously, concerns toe affect the results. Other esperiments evaluated the influence of incluson of a model for the BWR emergency core cooling sys- crystatittation on glass waste form performance and the influ-tems (ECCS). This model and its computer emplementation, the ence of organic acd on the wasto form and ra4onuchde mobili. ECCRPI subroutine, account for vanous emergency iniection ty in groundwater Modols were used to analyte glass dissolu-modes, for both intact and rubbhted geometnes Other change
tion, inclueng the reprecipitation of dssolved glass specees, to V LAPI deal with an improved model for canister wall orda- The effect of groundwater species on the electrochemistry of ton, rubble bed modekng, and numencal integration of system steels is being analyted to evaluate susceptabihty to pitting and equations. A complete documentation of the entire stress-corroson cracking Species dentified as potential crack.
MELRPI MOD 2 code is also given, includng an input guide, hst ing agents are tung investigated by slow strain rate espen-of subroutines, sample input / output and program hsting ments Hydrogen embnttlement studres of steel showed an. NUIER/CR 3900 V02: LONG TERM PERFORMANCE OF MATE- nealed cast s'cet to be more sensitive to embrittlement Reahn-RIALS USED FOR HIGH LEVEL WASTE PACKAGING Ouarterly tic general and pitting corrosion modals are being developed. Report.Jufy September 1964. STAHL.D ; MILLER,N E. Battelle based on known pnnciples of mass transport and radofytic pro-Memonal Inst,tute, Columbus Laboratones January 1985 duction Mechanical and water chemistry related stresses which 117pp. 8502130361, 28914 098. influence mechanical dogradat6on were evaluated Groundwater. Dunng this teriotting penod. it was found that glass-*ater rasolysis and water <hemistry studes are continuing as part of contact dunng the nonisothermal penorfs of leech testing maf the integrated system performance task influence the test resuffs Modehng of waste'urm degradaten I k
1 l Main Citations and Abstracts 49 NUREG/CR-3901: DOLUMENTATION AND USER S GUIDE GS2 that are both computer-readable and computer searchable. This
& GS3. VARIABLY SATURATED FLOW AND MASS TRANS- system provides a structured format for detaried codmg of com-PORT MODELS DAVIS.LAa SEGOLG. Water. Waste & Land, ponent, system. and un:t effects as weil as personnel errors.
Inc. June 1985. 305pp. 8507250143. VW,L/TM 17912. This four volume report documents and desenbes SCSS n 31789 002-This report presents documentaten and user's manual for datail Volume 1 is a User's Guide for searching the SCSS data. programs GS2 (two dimensional version) and GS3 (three dtmen- base. Chapter 2 of this guide is a tutonal on retrieving. d splay-ing. and analyzing LERs and provides hands on empenence en sional version). Mathematical equations and physical pnnciples utshzed to develop the code are presented in Section 2. The nu- execut ng basic commands Volume 2 contains all vahd and ac-mencal approach used (GareArn Finite Elements) is presented ceptable codes used for searching and encoding the LER data in Section 3. Section 4 presents an overview of how prcblems Volumes 3 and 4 provide a technical processor, new to SCSS. should be set up to properly use the code while detaded input the informaton and methodology necessary to capture desenp. tive data from the LER and to codsfy that data into a structured instructions are presented in Secton 5. Output produced by the code is discussed in Section 6. Three example probtems, in- format and serve as reference matenal for the more emperi-cluding sample input data sets and output data, are presented enced technicat processor, and conta.ns information that is es-in Section 7. Program information is provided in Section 8 A sential for the more advanced user who needs to be famihar hsting of important program vanabies along with comp!ete pro- with the intricate coding techniques in order to retneve specif>c detads in a sequence gram listings are presented in the Appendices This report was prepared as part of a project in which NRC staff was presented NUREG/CR 3905 V02: SEQUENCE CODING AND SEARCH a training course on how to property use this computer program SYSTEM FOR LICENSEE EVENT REPORTS Code Listegs Programs GS2 and GS3 can be utibzed to raaiyze flcw and GALLAHER R BJ GUYMON R H ; MAYS.G T , et al Oak Roge mass transport in unsaturated, partially satura ed. or fully satu. Natonal Laboratory Apnl 1985 271pp 8505070170 CANL/ rated flow regions. It is anticipated that the NRC *di use the NSIC 223 30214 022 model for checlung informaten provided by a hcensee, fcr eval. See NUREG/CR-3905,V01 abstract uating alternative sites and des.gns for waste disposal and for companng their results with results from other methods of solu. NUREG/CR 3905 V03: SEQUENCE CODING AND SEARCH tion. SYSTEM FOR LICENSEE EVENT REPORTS Coder s Vanual NUREG/CR 3904: A COMPARISON CF UNCERTAINTY AND GALLAHER.R B ; GUYMON R H. MAYS G 7 ; et at Oak Rdge National Laboratory April 1985. 38 tpp 8505070006 ORNL/ SENSITIVITY ANALYSIS TECHNIOUES FOR COMPUTER NS C-223 30213 001 MOCELS. IMAN.H L.; HELTON.J C. Sandia National Laborato-hes. May 1985.110pp. 8506190020. SANC841461. 31019 001. See NUREG/CR.3905.v01 atstract Uncertainty ana'ysis and sensitivity analys s are ireportant ele. NUREG/CR 3905 V04: SEQUENCE COD NG AND SEARCH ments in the development and implementat;on of computer SYSTEM FOR LICENSEE EVENT REPORTS Coder's Manual models for complex processes Typically, there are many uncer. GALLAHER R B ; GUYMON R H ; MAYS G T et at Oak Rdge tainties associated with both the development and the apphca. National Laboratory Apnl 1985 347pp 8505070t84 CANL/ tion of such models Understanding of these uncertaint.es and NSC223 30212 001. their causes is required to effectively interpret model behavior See NUREG/CR-3905.v01 abstract Many different techniques have been proposed for performing uncerta.nty and sensitivity analyses The objective of the NUREG/CR 3906: URAN:UM MILL TAlLINGS present study is to compare several widely used techniques on NEUTRAll2ATION CONTAMiN ANT COMPLEXATION AND three rnodels having large uncertaint.es and varying degrees of TAILINGS LEACHING STUOY OPITZ.B E . DOOSON M E . complexity in order to highhght some of the problem areas that SERNE.R J Datteile Memonal Inst,tute. Paci6c Northwest Lab. must be addressed in actual appkcations The fotlowing ap. orator:es May 198$. 77pp 8506190026 PNL.5179 31010119 proaches to uncerta,nty and sensitivity ana!ysis are considered Laboratory evpenments were performed to compare the of. (1) response surface methodology based on input determined fectiveness of hmestone and hydrated hme for improving waste from a fractional factonal design. (2) Latin hypercube samphng water quahty through the neutral + ration of acdc uranium mdl with and without regression analysis, and (3) d.fferent al ana!y. taihngs hquor. The espenments were designed to assess the ef-srs These techruques are compared on the basis of (1) ease of fects of tnree proposed rnachanisms carbonate compiena. implementation, (2) f!esibibty, (3) estimation of the cumulative tion elevated pH and colloidal particle adsorption on the sol-distribution function of the output. and (4) adaptabihty to differ, ubikty of toxic contaminants found in a typecal uranium mal ent methods of sensitivi anasysis Wrth respect to these cnte. waste solution Of special interest were the etiects of each of na, the technique using Latin hypercube samphng and regres, these possible rnechanisms on the solubon concentrabons of Sion ana'ysis gives the best results overall. The models used in trace metafs such as Cd. Co. Mo. Zn and U after neutrakrahon the compansons are well documented, thus making it poss>ble Acidic untreated sohd taihngs from two mill s.tes and ta>hngs for researchers to make comparisons of other techniques with neutrahted with hme were leached with a faboratory prepared the results in this study. ground witer for several pore despiacement volumes Anatyses performed on the column efffuents indicate that pnor neutrahla NUREG/CR 3905 V01 R1: SEQUENCE CODING AND SEARCH tion results en a significant reduction in the concentration of all SYSTEM FOR LICENSEE EVENT REPORTS User s Guido pH dependent constituents in the column effluents in contrast. GREEN N M ; MA VS.G T.; JOHNSON M P ; et al Oak Ridge Na* tional Laboratory. Apnl 1985 164pp 8505070182. ORNL/NSIC- relatively high concentrations of several trace metais and macro 4 223. 30216 139 ,ons were found in effluent solution from the untreated faihngs columns Operating empenence data from nuclear power plants are es-sential for safety and rehability analyses, especially analyses of NUREG/CR 3911 V02: EVALUATION OF WELDED AND trends and patterns The hcensee ever t reports (LERs) that are REPAIR-WELDED STAINLESS STEEL FOR LWR submitted to the Nuclear Regulatory Commission (NRC) by the SERVICE Ouarterly Report. Apol June 1984 A TTERf DGE.D G ; nuclear power pfant utihties contain much of this data The BRUEMVER.S M , PAGE.R E Batteile Memonal Institute. Pacif. NRC's Office of Anatysis and Evaluabon of Operabonal Data ic North
cst Laboratones February 1985 42pp 8503220003 (AEOD) has developed under contract with NOAC, a system for PNL 5181. 2948 7. t 76 codfying the events reported in the LERs The primary objective The Division of Engineering Technology. U S Nuclear Reguta-of the Sequence Coding and Search System (SCSS) is to tory Comm4sion. is sponsonna a program at Pacific Northwest reduce the descr!ptrve test of the LERs to coded sequences Laboratory to e>aluate welded and repair weldet stainless steel
~~'! (~~
50 Main Citations and Abstracts piping for bght water reactor service. Stainless stoels often boundanes to detect and evaluate growing flaws. Several areas become sensitized, or less resistant to stress corrosion cracking of technical concern are addressed. Results support the feasi-(SCC), after undergoing heating and cookng cycles such as bikty of effective continuous monitonng. those encountered in welding. The weld heat affected zone is often the site of crack initiation. This program will therefore NUREG/CR 3919: TR AC-PF1/ MOD 1 INDEPENDENT measure and model the development of a sensitized microstruc- ASSESSMENT.NEPTUNUS PRESSURIZER TEST YO5. ture and its resultant resistance to SCC in welded and repair- PETERSOKA C. Sandia National Laboratones. Februay 1985. we'ded stainless steel pipe. The result will be a method to 60cp 8500040543. SAND 841534. 29218 281. assess the effects of welding vanables on the SCC susceptabili- T'1e TF.AC er dependent assessment protect is part of an ty of component specific nuclear reactor / repairs- overall effort to determine the capabihty of vanous system codes to preda.:t the detailed thermal /hydrauhc response of bght NUREG/CR-3912: MARCH.HECTR ANALYSIS OF SELECTED ACCIDENTS IN AN ICE CCNDENSER CONTAINMENT. water reactors dunng accident and off. normal conditions The CAMP,A.L.; BEHR,V.L.; HASKIN,F E. Sandia National Latorato- TRAC compu'er code is being assessed against test data from rics. January 1985. 214pp. 8503050511. SAND 83-0501. vanous integral and separate effects test facilities. As part of 29264 001- this assessment effort, a separate effects component test per. The MARCH and HECTR computer cedea are used in this formed in the NEPTUNUS pressunzer test facility for Thermal study to examine hydrogen prod % tion, transport, and combus- Power Engineerng at Delft University of Technology was ana-tion ln an ice-conde ser containment for a mmber of hypoth. lyzed with TRAC-PF1/ MOD 1. The test simulated insurges, com. esized severe acdcan's. Both degradedcoro and core-melt- bined with spray flow, and Jutsurges from a pressunzer, and down accidents are t:eated. The sensitivfy of the conta:nment was selected for code assessment because the capability of the pressure-temperature response is assessed for a number of computer codea used in safety analyses to calculate the correct factors, including the hydrogen and steam source-term assump- g tions, ignitcn and propagation hmits, combustion completeness, flame speed. spray operation, and recirculation fan operation. and fluid temperatures were calculated dunng insurges with The highest contarnment pressures occur for those cases spray flow than were measured in the test. A contnbuting factor where the igniters are assumed to fad, the recsrcufation fans or to the calculation of high pressures and fluid temperaUres ap-containment sprays are assumed to fail, or very large steam pears to be that the interfacial heat transfer from superheated and hydrogen releases accompanying vessel breach are pre-vapor to subcooled liquid was too low. The calculatenal results dicted. for the base analysis and some modeling studies are discussed NUREG/CR 3913: HECTR VERSION 1.0 USER'S MANUAL. A TRAC PF1/ MOD 1 input hsting of the base case model is also CAMP,A L: WESTER.M.J.; DINGMAN,S E. Sandia Nat onal Lab. provided or: tones. Apnl 1985. 325pp. 8504160098. SAND 841522. 29832.026 NUREG/CR 3922 V01: SURVEY AND EVALUATION OF SYSTEM This report desenbes the features and use of HECTR Verson INTERACTION EVENTS AND SOURCES Main Report And Ap-1.0. HECTR is a relatively fast-running lumped-votume contain- pendices A And B. MURPHY,G A. Oak Ridge Natenal Laborato-ment ana!ysis computer program that is most usefLJ for per. ry. CASADA.M L.; JOHNSON.M P.; et al. JBF Associates. Janu-forming parametnc stud.es. The main purpose of HECTR is to ary 1985.114pp. 8502120062. ORNL/NOAC-224. 28872.137. analyze nuclear reacter a:.cidents invotving the transport and This report desenbes the first phase of an NRC-sponsored combustion of hydrogen, but HECTR can also function as an expenment analysis tool and can solve a limited set of other project that identified and evaluated system interaction (SI) events that have occurred at commercial nuclear power plants hpes of contafnment problems. HECTR Version 1.0 has been sn the Un red States. The project included. an assessment of partcularly ta:lored to anatyre accidents in ice-condenser PWR and Mark 111 BWR containments. HECTR is designed for flexibil- nuciear pvwer plant operating expersence data sources; the de-ity and provides for user control of many important parameters, velopmen; of search methods and event selection cntena for ptrtcularly those related to hydrogen combuston. Bunt-in cerrt identifying Si events; review of possible SI events; and final Ictions and default values of key pararaeters are also provided evaluation and categonzation of events. The report outlines each of these steps and presents the results of the protect. The NUREG/CR-3914: PUMP AND VALVE OUALIFICATION REVIEW results include 235 events identified as adverse system Pterac. GUIDE. MILLEP.B E. Brookhaven Nat onal Laboratory. October tions and 2? categones into which those events were assigned 1985.58pp.8611050305. BNL NUREG 51807. 33339.122. The categnnes represent groups of similar ever.:s and include This report provides NRC reviewers with guidance, assist-ance, and examples relating to the information and procedures areas such as: adverse interactions between normal or offsite 13 be socluded in an apphcant's pump and vatve operability as- power and emergency power systems; degradation of safety surance program, and to the scope and depth of the review. systems by vapor or gas intruson; degradation of safety-refated Discussed are the applicable components, concerns, method- equipment by fire protection systems; and flooding of safety-re-oiogies, documentaton, eva:uation proceuures, and examples of lated equipment through plant drain systems After evaluating dagn and operabihty issues ror pump and vatve assembhes. each category (and the events contained in them), the empha-These items should be of concern and included in an appli- ses on the potential for continued prob! ems in these areas cmt's quahfication program. should be examined; and current system interacton analyses methods should be studied to determine their effectiveness for NUREG/CR-3915: ACOUSTIC EMISSION RESULTS OBTAINED identifying system interaction events. (Phase Il of this project. FROM TESTING THE 28-f INTERMEDIATE SCALE PRES- " Evaluation of System Interaction Methods," will assess the ef-SURE VESSEL HUTTON.P.H.; KURTZ.R.J.; PAPPAS.R.A.; et fectiveness of current methods using the events identified in ti. Battelte Memonal Insttute, Pacific Northwest Laboratones. this report). l September 1985. 240pp. 8509300231 PNL 5184. 32811:317. Acoustic emission (AE) monitonng of flaw growth in an inter-mediate scale vessel dunng cyche loading at 65 degmes cents. NUREG/CR-3922 V02: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES Appendices C And D. grade and 288 degrees centgrade is desenbed in this reoort. MURPHY,G.A. Oak Ridge National Laboratory. CASACA,M L.; The report deals with background, methodology, and results. JOHNSON.M.P.; et al. JBF Associates. January 1985. 265pp. The work discussed is of major significance in a program sup-8502120069. ORNL/NOAC-224. 28871:002. ported by NRC to develop and demonstrate apphcation of AE See NUREG/CR 3922,V0f abstract. monitonng for continuous surveillance of reactor pressure l l
Main Citations and Abstracts 51 NUREG/CR-3930: OBSERVED BEHAVIOR OF in this paper we attempt to evaluate the overall code perform-CESIUM. LODINE.AND TELLURIUM IN THE ORNL FISSION ance by companng results from many different calculations, and PRODUCT RELEASE PROGRAM COLLINS.J L.; to offer other users some guidehnes based on our experience to OSBORNE.M.F.; LORENZ,R. A.; et al. Oak Ridge National Labo- date. All results show that good pnmary S de steady state initial ratory Apnl 1985. 73pp. 8504170679. ORNL/TM 9316 and/or operating conditions are readily obta.ned, given ade-Two corktrol tests were conducted to study the behaver of
0" Csi, CsOH, and Te in the expenmental apparatus used to con-ene' rat seco a sdes duct fisson product release tests with highly irradiated LWR fuel at ORNL in this report the control tests are described, and the resutts are compared with those obtained for cesium, iodine, NUREG/CR 3937: STEAM GENERATOR TUBE RUPTURE IODINE TRANSPCRT MECHANISMS Task 1.Expenmental and telfunum in 26 tests of irradsted fuel and other tests using tracers. In good agreement with the LWR fuel tests, the Csl be- Studies GIESEKU A.: FLANIGAN.LJ.; COLLIER,R P.; et al.
Battelle Memonal Institute, Columbus Laboratones. October havior in the control tests was similar to that observed for iodine 1985. 36pp. 8511110327 BMI 2114. 33412.317-in the fuel tests; iodine was released pnmanly as Cs! rather than highly volatile molecular iodine. Cesium (not associated with lodine release from a nuclear power plant dunng steam gen-Csi) behaved like CsOH in the LWR fuel tests in both LWR fuel wator tube mpture accidents is expected to be strongly de-tests and the control tests, cestum hydroxide was observed to pendent on the drop sizes formed as high pressure pnmary react with and be retained by zirconia ceramic surfaces en the system water is f: ashed and atomized as it passes through the temperature range 800 to 1200 degrees centigrade, probably rupture opening This study was based on the need for informa-forming cesium metazirconate (Cs(2)ZrC(3). In one of the con. tion on drop sizes formed under such conditions Expenmants trol tests, cessum hydroxide reacted with tellunum in the gas to measure the fraction of water flashed and the drop sizes phase and was collected as CsTe. Although the results are lim. formed were performed at typical operating pressures and tem-ited at this time, the indicated collected behavior of teeiunum in peratures with the actual tube diameters and lengths neady to the LWR fuel tests has been that of a tellunde. scate. The mass median drop sizes measured were in the range fr m about 20 to 60 micrometers for both open ended and slit NUREG/CR-3935: THERMAL HYDRAULIC ANALYSES OF OVERCOOLIN3 SEQUENCES FOR THE HB ROBINSON ruptum geometnes. No sigmficant effect on drop size of pnmary UNIT 2 PRESSURIZED THERMAL SHOCK STUDY. system pressure level was noted over the range from 1100 to FLETCHER.C.D.; DAVIS.C.B.; OGDEN.D M. EG&G Idaho, Inc. 2100 psig subs of EG&G, lnc.). July 1985. 291pp. 8507250171. EGG-NUREG/CR 3943: THE BWR PLAN ANALYZER. WULFF,W.; Oak Ridge National Laboratory (ORNL), as a part of the Nu- CHENG.H S.; LEKACH S V.; et al. Brookhaven National Labora-clear Regulatory Commission's (NRC's) pressunzed thermal % uary M85. 345pp. 8503120461. BNL-NUREG 51812. shock (PTS) integration study for the resoluton of Unresolved 2933M 3t Safety issue A49, identified overcooling sequences of interest This final report desenbes the modeling. the software and the to the H.B. Robinson PTS study. For each sequence, reactor hardware of the plant analyzer. The report also presents the vessel down-comer fluid pressure and temperature histones first developmental assessment and contains a user guide for were required for the two-hour period following the initiating the plant analyzer. A large number of transients have been sim-event. Analyses previously performed at the Idaho National En. ulated The simulation encompasses the neutron kinet es, the gineenng Laboratory (INEL) fully investigated a limited number thermal conduction in fuel structures and the hydrauhes of non-of the sequences using a detailed RELAPS model of the H B. equilibnum, nonhomogeneous two-phase flow in the nuclear Robinson, Unit 2 (HBR.2) plant. However, a full investigation of steam supply system, steam line dynamics, turbines, condens-all sequences using the detailed model was not economically ers, feedwater trains and tho suppression pool, as well as the practical. New methods were required to generate results for control and plant protection systems. All simulations can be car-the remaining sequences. Pressure and temperature histones ned out at speeds up to 10 times faster than real. time p*ocess for these remanng sequences were generated at the INEL speeds. The technology presented here has been developed through a process combining: (a) ;,artial-length ca!:ulations primanly for cost-effectue safety analyses, but is also invaluable using the detailed RELAP5 model, (b) full-length ca:culations for plant monitonng. failure diagnosis and computer aided mits-using a simplified RELAPS model, and (c) hand caiculations gation of accidents. This report documents both the methods used in this process and the results. The sequences investigated contain sigmficant NUREG/CR-3944: TRAN B-3 EXPERIMENTAL INVESTIGATION conservatisms concerning equipment failures, operator actions. OF FUEL CRUST STABILITY ON MELTING SURFACES OF AN or both. Consequentfy, care should be taken in applying the re- ANNULAR FLOW CHANNEL. MCAR THUR.D. A.; MAST,P.K. suits presented herein without an understanding of the conserv- Sandia National Laboratones. Apnl 1985. 61pp. 8505030229 visms and assumptions. The results of the thermal-hydraulic SANDE 4-1646. 30160.302. analyses presented here, along with additional analyses of mul- The TRAN B senes of expenments as being conducted at tdmensional and fracture mechanics effects, will be utilized by Sandia National Laboratones to invesbgate the characteristics ORNL to assist the NRC in resolving the PTS unresolved safety of fuel removal / freezing through the upper svial blankets of an issue- LMFBR dunng the transibon phase of a hypothetica! core dis-NUREG/CR-3936: RELAPS ASSESSMENT. CONCLUSIONS AND ruptue accident. The third expenment in this senes TRAN B-3, USER GUIDELINES. KMETYK,LN. Sandia National Laborato- was performed in February 1984, and the results are reported nes. January 1985. 182pp. 8502010096. SAND 84 1122. herein. This expenmer't involved the injection of molten UO(2) 28700.001. into an annular flow channel. Unlike the similar TRAN B 1 ex. The RELAPS independent assessment project at Sandia Na. penment, the initial steel wall temperature in B-3 was sufficiently tional Laboratones is part of an overall effort funded by the hgh that instantaneous steel melting would occur upon contact NRC to determine the ability of vanous systems codes to pre- with molten fuel. The earher TRAN 0-1 results had shown that dict the detailed thermal / hydraulic response of LWRs dunng ac- fuel crusts are initially stable, both on the inside of a steel tube cident and off-ncrmal condebons. The RELAP5/ MODI code has as well as on the outside of a steel rod, when no steel melting been assessed at Sandia against a variety of test data from occurred. TRAN B-3 was designed to invest gate this question both integral and separate effects test facilities. All these anaty- of crust stability on surfaces of opposite curvature when surface ses have been documented in detail in individual topical reports; melting did occur.
52 Main Citations and Abstracts NUREG/CR-3945: FATIGUE CRACK GROWTH RATES OF LOW- rejection of tube support signals. We adapted our new IBM CARBON AND STAINLESS PIPING STEELS IN PWR ENVI- System 9000 computer to take and process the targer amounts RONMENT. CULLEN W.H. Matenals Engineenng Associates, of data required by additional vanables, such as copper coating Inc. February 1985. 65pp. 8502150048. MEA 2055. 29958.202. and intergranular attack. We also completed construction of the Fatigue crack growth rates of A 106 Gr. C and A 516 Gr. 70 hand. wired versions of the 8- and 16-coil arrays and the multi-carbon steels, and A 351-CF8A stainless steelin PWR environ- plex,ng crrcuitry and computer codes to handlei the data. ments have been determined over a load ratio range (R) of 0 2 to 0.85, a temperature range of 93 degrees cent grade to 338 NUREG/CR-3949 V02: EDDY-CURRENT INSPECTION FOR degrees centigrade, and a test frequency range of 17 mHz to 1 STEAM GENERATOR TUBING PROGRAM. Annual Progress H2 using sinusoidal waveforms. In addition, growth rates have Report For Penod Ending December' 31,1984. DODD.C.V.; been determined for vanous onentations of the crack plane with DEEDS W E.; SMITH.J H.; et al. Oak Ridge National Laboratory. respect to the product form. Crack growth rates in 288 degrees August 1985. 16pp. 8509110016. ORNL/TM-9339/V2 cent, grade air environments have been measur.3d in order to 32564 288. provide a reference basehne. These results define the magni. Eddy current inspecten is the most suitable method for rapid tude of and major influences on the environmentally-assisted f a- boreside evaluaten of steam generator tubing. However, small tigue cracn growth rates for these piping steels. and are sup- flaws can be masked by the effects of harmless vanables. such ported by fractographic observatens of the fatigue fracture sur- as tube supports. To identify the entical properties accurately face. and rehably in the presence of extraneous signals caused by vanat:ons of unimportant properties, sufficient information is NUREG/CR-3948: EXPERIMENTAL RESULTS OF THE OPER-ATIONAL TRANSIENT (OPTRAN) TESTS 11 AND 1-2 IN THE needed to identify harmful vanatens and reject harmless ones. POWER BURST FAC!LITY. MCCARDELL R K.; PLOGER,S A : For this reason we have been developing instrumentaten capa-MCCORMICK R D.; et al EG&G Idaho, Inc. (subs. of EGaG, ble of measunng both the amplitude and phase of the eddy-cur-Inc.). September 1985. 78pp. 8510030433, EGG 2297. rent signal at several different frequencies, as well as computer 32911:221. equipment capable of processing the data quickly and rehably. Operatonal transients occur occasionalty in light water reac- Our probes and test conditions are also Computer optimized. tors when minor malfunctions of cedarn system components The most recent probe design embodies an array of small flat affect the reactor core. This report preseO the results of the " pancake" coils and improves the detecton of small flaws and operational transient Test OPTRAN 11 ani., OPTRAN 1-2, in- the rejection of tube support signals We have also expenmen-cluding a companson of the data with posttest calculatons and tally venfied the acr'uracy of our computer programs for calcu-the postirradiaton examir ation results. The OPTRAN11 tests lat,ng the signals produced by defects in tubing and are adapt-s;mulated operatonal transients with reactor scram. Four pro- ing our new IBM System 9000 computer to take and process gressrvety higher and broader power transients at a constant the larger amounts of data requrred by additonal vanables, such coolant flow rate were performed The first transient simulated a as copper coating and intergranular attack. BWR 5 turbine tnp without steam bypass, with fuel rods operat-ing near BWR-6 core average rod powers. The second transient NUREG/CR-3950 V01: FUEL PERFORMANCE ANNUAL simulated a generator load rejecten without steam bypass, with REPORT FOR 1983. BAILEY,W J.; DUNENFELD M S. Battene fuel rods operating near core average powers. The last two Memonal Institute, Pacific Northwest Laboratones March 1985. transients were performed at higher core average peak rod 123pp. 8503280020. PNL 5210,29547.139. powers than safety analyses predict to be ecssible in commer* This annual report, the sixth in a senes, provides a bnef de-cial reactors to define failure threshold margins. Test OPTRAN scnpten of fuel performance dunng 1983 in commercial nuclear 12 was performed to evaluate the probability and extent of fuel power plants. Onef summanes of fuel design changes, fuel sur-rod damage for the most severe BWR anticipated transient veillance programs, fuel operating expenence, fuel problems, without scram (ATWS) that results in boshng transiten, a main high-burning fuel expenence, and items of general significance steam kne isolation vatve closure transient without scram. Two are provided. References to additional, more detailed informa-sets of two fuel rods were tested. In each set, an unirradiated tion and related NRC evauations are included. fuel rod was used to heat the coolant to typical BWR conditions for each prevous!y irradiated fuel rod. Following an extensive NUREG/CR-3952: SEOUOYAH EQUIPMENT HATCH SEAL fuel conditening penod of operation, a single power transient LEAKAGE. GREIMANN,L.; FANOUS.F,; BLUHM.D. Ames Labo-was performed that simulated the power history and coolant ratory, Energy & Mineral Resources Research Institute. July cond tions calculated for a main steam line isolation valve clo- 1985. 49pp. 8508090579. IS-4862 32105 200. sure ATWS- Nuclear containments which will not leak significantry, that is, NUREG/CR-3949 V01: EDDY-CURRENT INSPECTION FOR beyond technical specificatons. dunng a design accident may STEAM GENERATOR TUBING PROGRAM. Semiannual leak severely dunng a severe accident when the pressure in. Progress Report For Perod Ending June 30,1984. DODD C.V.; creases beyond the design level. Smaff leaks which are visual-DEEDS W.E.; SMITH J H.; et al. Oak Ridge National Laboratory. ized as occurnng at local details may occur before complete January 1985 14pp. 8502210202. ORNL/TM-9339/V1. vesset failure. Buckhng of the hatch door, large deformatens 29028 333. and ovakng of the hatch sleeve are potential causes of mis-Eddy-current inspecten is the most suitable method for rapid match at the seahng surface which can result in a leakage path. boreside evaluation of steam generator tubing However, sma'l As a typical example of steel containments the Sequoyah equ:p-flaws can be masked by the effects of harmless vanables, such ment hatch was selected. If penetratons effects are neglected, as tube supports. To identify the entical properties accurately gross yielding of the 1/2-inch shell plate near the spnnghne of and reliably in the presence of extraneous signals caused by the Gequoyah containment will occur at an internal pressure of vanations of unimportant properties, sufficient information is between 50 and 60 pst The results of a finite element analysis needed to identify harmful vanatens and to reject harmless showed that a maximum of 0.9 inch of 85 to 90 psi, far above ones. For this reason we are developing instrumentaten capa- gross yielding of shell. Although buckhng increased the relative ble of measunng both the amplitude and phase of the eddy-cur- seal motions, they remained sufficiently smaff to prevent leak-rent signal at several different frequencies and computer equip- age. The Sequoyah equipment hatch should not feak before ment capable of processing that data quickly and rehably. Our strains of several percent develop in the 1/2-inch containment probes and test conditions are also computer opterrized. The shell plate near the spnnghne, which occurs between 50 and 60 most recent probe design embodies an array of small flat " pan- pst in the unhkely event of hatch buckhng, postbuckkng defor-Cake" Coils and improves the detection of small flaws and the mations would not introduce leakage.
i Main Citations and Abstracts 53 NUREG/CR-3953: THE USE OF MAG-1 SPECTACLES WITH partially saturated conditons. If the magnitude of partially satu-POSITIVE- AND NEGATIVE-PRESSURE RESPIRATORS. rated settlement is considered significant, then the time over REED,K.A.; MOORE,T.O. Los Alamos Scientifc Laboratory. May which it occurs will most hkely be the deciding factor in deter-1985. 41pp. 8506060798. LA 10229-MS. 30775:198. d Results of testing conducted at Los Alamos Natonal Labora- mining when to place the cover on the tanhngs pile.
' tory, Personnel Protection Studies Section, using MAG-1 spec. NUREG/CR-3977: RELAPS THERMAL-HYDRAULIC ANALYSES tacles in conjunction with positwe- and negatwe-pressure full- OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B.
facepiece respirators, are reported. The purpose of the three. ROBINSON UNIT 2 PRESSURIZED WATER REACTOR. phase study was to determine if the specially constructed strap FLETCHER.C.D.: BOLANDER.M.A.; WATERMAN,M E.; et al. of the MAG 1s affected the protecten factors (PFs) of the res- EGaG, Inc. Apnl 1985. 233pp. 8506140624. EGG-2341. pirators or the cylinder life of selected se'f contained breathing 30034.352. apparatus (SCBA). The following respirators were tested witn Thermal. hydraulic analyses of fourteen hypothetical pressur-the MAG-1s: a) Phases I and 11, positwe-pressure full facepiece: ized thermal shock (PTS) scenarios for the H. B. Robinson, Unit Presur-Pak 11 SCBA (pressure-demand) Scottoramic facepiece, 2 pressurized water reactor were performed at the fdaho Na-4 MSA 401 Air Mask Ultravue facepiece (medium), Survivait pres- tional Engineenng Laboratory (INEL) using the RELAPS comput-sure-demand SCBA/sihcone full facepiece, MSA powered air. er code. The scenares, which were developed at Oak Ridge punfying respirator /Ultravue facepiece (medium); b) Phase lit, Natenal Laboratory (ORNL), contain significant conservatisms j negatwe-pressure tual facepiece: MSA Ultravue (small, medium,
' concerning equipment failures, operator actions, or both. The large), MSA Ultra-twin (small, medium, large), Norton Senes results of the thermal-hydrauhc analyses presented here, along 7600 (one size only). Statistical analysis and review of the test with additonal analyses of multidimensional and fracture me-data from Phases I and ll indicated httle, if any, variation with chanics effects, will be utihred by ORNL, integrator of the PTS
( and without the MAG 1s with most protection factors greater study,' to assist the U.S. Nuclear Regulatory Commisson in re-than 10,000. Test data also indicated httle, af any, difference in solving the pressanzed thermal shock unresolved safety issue. the cyhnder hfe with and without the MAG-Is, except the Scott Presur Pak il SCBA used with the Scottoramic facepiece. States- NUREG/CR-3978: TENSILE PROPERTIES OF IRRADIATED NU-tical analysis of the quantitatue fit test data indicated no differ, CLEAR GRADE PRESSURE VESSEL PLATE AND WELDS ence in PFs for the negatwe pressure devices for the Ultravue FOR THE FOURTH HSST lRRADIATION SERIES. negatwe-pressure respirator, but a significance at the 0.05 and MCGOWAN.J.J. Oak Ridge National Laboratory. January 1985. 0.01 levels for the Ultra. twin and Norton full facepieces, respec- 25pp.8503110098. ORNL/TM-9516. 29328 206. twely. The Heavy Section Steel Technology (HSST) program office
! conducted a senes of expenments to determine the effect of . NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPER
~~ATURE RESPONSES TO SELECTED HYDROGEN BURNS IN neutron irradiaton on the fracture toughness of nuclear pres- ,~~
sure vessel materials. One plate (HSST plate 02) and four AN ICE CONDENSER CONTAINMENT. DANDINI,V.J.; MCCULLOCH.W.H. Sandia National Laboratones. February welds of A533 grade B class I steel were examined. The welds 1985.135pp. 8503280013. SAND 841704. 29547:001. were made by current (about 1979) practice. As part of this The temperature response of three generic surface models in , study, tensile properties were measured after irradiation to 2 x each of three locations in an ice condenser containment build- 10{23) neutrons /m(2) (>1 MeV) at 288 degrees C. The 1 ing were calculated assuming a hydrogen deflagration event strength of all four welds incieased with irradiation. Yield j and using the HECTR code. The intent of using the three gener- strength was about 10% more sensitwe to irradiaton than was e surfaces was to conservatwely represent surfaces of vanous uitmate strength. Tensile ductihty was not affected segnificantry types of safety equipment. Analyses were performed for four by irradiation. accident sequence types with variations. The general observa-NUREG/CR-3980 V02: LIGHT WATER-REACTOR SAFETY FUEL
! tions drawn from these analyses are that (1) higher equipment SYSTEMS RESEARCH PROGRAMS. Quarterty Progress surface temperatures than calculated for S{2)D type arrested Report.Apnl. June 1984. REST,J. Argonne National Laboratory.
sequences were calculated for other sequence types of compa- February 1985. 36pp. 8503290274. ANL-84-61 V02, 29563.354. rable core melt frequency, and (2) surface temperatures greater This progress report summantes the Argonne Nabonal Labo-
~~. than quahfication temperatures were calculated to occur for ratory work performed dunng Apni, May, and June 1984 on some sequence types.~~
water reactor safety problems related to fuel and fuel cladding NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS matenals. The research and development areas covered are PILES: A COMPARISON OF ANALYSIS TECHNIQUES. Transient Fuel Response and Fission Product Release and Clad i FAYER.M.J. MCKEON T.J. Battelle Memonal Institute Pacific Properties for Code Venfication. Northwest Laboratones. December 1984. 9'pp. 8503040052. PNL-5222. 29198:274. NUREG/CR-3980 V03: LIGHT WATER-REACTOR SAFETY FUEL Two empincal methods of settlement anafysis (Terraghfs SYSTEMS RESEARCH PROGRAMS. Quarterly Progress Report. July September 1984. REST,J. Argonne National Labo-theory and a simphfied version of the Fredlund-Morgenstern ratory. May 1985. 51pp. 8507050429. ANL-84 61. 31371:306. two-stress-state approach) were compared to the computer , code TRUNC, a modified version of the TRUS9 code for van- This progress report summanzes the Argonne National Labo- ! j ably saturated flow in deformable porous media. The three ratory work performed dunng July, August, and September 1984 on wafer reactor safety problems related to fuel and fuel clad-i methods were used to predict settlement of a 12.2-m-deep pile ding matenals. The research and development areas covered { of tashngs shmes with a drain at the bottom. The simpler, empin-are Transient Fuel Response and Fission Product Release and cal methods of settlement analysis were just as effective as Clad Properties for Code Venfication. i TRUNC in predicting total settlement. For saturated taihngs, pre-i
- dictrorts of total settlement by Terzaghts theory and TRUNC NUREG/CR 3980 V04: LIGHT-WATER-REACTOR SAFETY FUEL wero in close agreement (1.69 and 1.73 m, respectively). For SYSTEMS RESEARCH PROGRAMS. Quarterly Progress partially saturated tashngs, the simphfied stress-state approach Report. October-December 1984. CHUNG.H M.; REST J. Ar-and 'TRUNC pre &ted similar total settlements (0.52 and 0.51 gonne National Laboratory. September 1985. 63pp.
m, respectively). Terzaghrs theory, as apphed, overestimated 8510040332. ANL 84 61. 32856.118. the time of settlement under saturated conditions (170 days This progress report summantes the Argonne National Labo. , versus 140 days predicted by TRUNC) because it did not ac. ratory work performed during October, November, and Decem-count for gravitational gradients. No empincal or analytical her 1984 on water reactor safety problems related to fuel and means were available to predict the time of settlement under fuel cladding matenals. The research and development areas 4
b 54 Main Citations and Abstracts covered are Transient Fuel Response and Fission Product Re- NUREG/CR-3989: TIME- AND VOLUME AVERAGED CONSER-lease and Clad Properties for Code Venficaten. VATION EQUATIONS FOR MULTIPHASE FLOW.Part One: System Without Intemal Solid Structures. SHA,W.T.: NUREG/CR-3981: BIOACCUMULATION OF P 32 IN BLUEGlLL CHAO,B.T.; SOO,S.L Argonne National Laboratory. February
~~. AND CATFISH. KAHN.B.; TURGEON,K.S.: MARTINI,0.K.: et al. 1985.127pp. 8503110328 ANL-84-66. 29328.001.~~
Georgia institute of Technology. Atlanta, GA. February 1985. A set of ngorously denved conservation equatens of mass. 170pp. 8502210154. 29051:001, momentum, and energy for multiphase systems without internal Bluegill and catfish maintained in flow-through tanks were fed solid structures via time-volume averaging of point, instantane. P-32 at two feeding levels. Fish were analyzed in tnplicate for P. ous conservation equaboris is presented. These equatons are 32 and phosphorus at intervals of 1 8 days. Additional aguana differential-integral equations in which the area integrals ac-count for the interfacial transport of mass, momentum, and expenments were performed to determine the effects of other factors and to observe P-32 uptake from water by unfed fish (en. energy. The equations from volume averaging contain averages i of the product of the dependent vanables which must be ex. ciuding fish with blocked esophagus). The bluegal showed a pressed in terms of the products of their averages. In nonturbu- } weight gain of 0.2 %/d, a phosphorus turnover constant in lent flows, this is achseved by expressing the " point" vanables muscle of 0.43 %/d. and a BF(r)/BF rato of 0.081 at the higher as the sum of its intnnsic volume average and a spatial devi-i feeding rate, and 0.05 %/d,0.34 %/d, and 0.064 at the lower ation. In turbulent flows for which further time averaging is re-feeding rate. Hence, respective P 32 BF(r) values are 6.000 and quired, the " point" vanable is then decomposed into a low-fre-4,000 at a phosphorus BF of 70,000. The BF(r) values for cat- quency component and a high-frequency component. Time fish were approximately twice as high. The aquanum experi- averaging following volums averaging preserves the identity of ments suggest that the higher factors are due to a much higher the dynamic phases. Under certain samphfying conditons, the d, phosphorus intake, higher water temperature, higher retenton proposed set of ngorously denved Conservaton equations, re. from pellets than from worms, and possible higher retenten by duces closely to vanous forms that are currently ' accepted' for catfish than bluegill under the same conditons. two-phase flow analysis. This set of conservation equatens Smes as a refsence point for modding muniphase flow and NUREG/CR-3984: BIOLOGICAL CHARACTERf2ATION OF RADi. pr sides theoretical guidance and physica! insight that may be ATION EXPOSURE AND DOSE ESTIMATES FOR INHALED useful to develop correlations for quantifying interfacial transport URANIUM MILLING EFFLUENTS. Annual Progress Report.Apnl mass, momemtum, and enwgy. 1983 March 1984. FIDSON.A.F. Inhalation Toxicology Re-1 search Institute January 1985. 2fpp. 8501230614. LMF-11. NUREG/CR-3990: CHARCOAL PERFORMANCE UNDER ACCI-28536:228. DENT CONDITIONS IN LIGHT WATER REACTORS. The problems addressed are the protection of uranium mill DEITZ,V.R. Navy, Dept. of, Naval Research Lab. March 1985. workers from occupational exposure to uranium through routine 190pp. 8504040424. 5528. 29628.058. j boassay programs and the assessment of accedental worker Nuclear grade carbons were systematically degraded by ex- ! exposures. Chemical properties of refined uranium ore (yellow- posure to enfdtered outdoor air with decrease in radcactive e cake) and uranium distnbution patterns among organs are com- methyl iodide trapping. Local meteorological conditions of high pared. These studies will facilitate calculatons of organ doses humidity combined with atmospheric pollutants in the test vican6-for potenbal exposures and will identify important bioassay pro- ty contnbuted jointly to the degradaten. When service carbons cedures. Results of studies in rats to investigate retention of were exposed to radiabon levels of 10(7) to 10(9) rads, the yellowcake in a wound showed that retention of less soir.o;e iodine isotope exchange capacity was regenerated. The adsorp-yellowcake from the body was significantly more prolonged than tive properties were only shghtly improved. It was possible to re-of more solubie yellowcake. However, retention could not be generate the iodine isotope-exchange efficiencies by reacton with airborn chemical reducing agents such as hydrazine for
! quanbtatively related to the chemical composition or in vitra dis. carbons removed from nuclear power operations. The depth solution behavior of the implanted powder. Studies of Bragte 4 profde in methyt iodide 131 penetration changed from simple dogs following nose-only inhalation of aerosols of commercial exponentional through new carbons to a non-linear profile for ,
yellowcake were continued. Histological observabons showed weathered and sennce aged carbons. The behaver is attnbuted kidney damage that appeared 4 to 8 days after exposure to the to the chromatographic distnbuten of the contaminants that ac-mure soluble yellowcake with repair occumng by 64 days after cumu! ate en the bed. The removal of radcactive iodine depends exposure. The concentraton of uranium in kidney was 8-17 mg on a minimum of 3 distinguishable processes: adsorption on the . U/g kidney at 4-8 days after exposure. No evidence of kidney activated carbon, iodine isotope exchange with impregnated damage was observed in dogs exposed to the less soluble yel- iodine-127, and chemical combinaton with impregnated tertiary lowcake form. amines when present. When a carbon is new, all 3 mechanisms are at peak performance. After the carbon is placed in service, NUREG/CR 3987: COMPUTER! ZED ANNUNCIATOR SYSTEMS. the 3 mechanisms degrade at different rates; the adsorption RANKIN,W.L; RIDEOUT,T.B.: TRIGGS T.J.; et al. Battelle process degrades faster than the others. Human Affairs Research Centers. June 1985. 102pp. 8507050435. PNL-5158. 31371:001. NUREG/CR 3991: FAILURE MODES AND EFFECTS ANALYSIS l This report presents the design philosophy and associated (FMEA) OF THE ICS/NNI ELECTRIC POWER DISTRIBUTION functional entena and design pnnciples for developing advanced CIRCUITRY AT THE OCONEE-1 NUCLEAR PLANT.
~~Oak '~~
computerized annunciator systems for use in the control rooms Ridge National Laboratory. MCBRIDE.A.F.; MAYO.C.W.; et al. of nuclear power plants. The scope of the work includes ad. Science Apphcahons international Corp. (formerly Science Ap-vanced system recommendations that could be incorporated plicatons, Inc.). October 1985. 96pp. 8512270363. ORNL/TM. into operating nuclear power plants. The information contained 9383. 34078:185. in this report was obtained from a review of the revelant com- The effects of nonnuclear instrumentaten (NNI) and integrat-puter and visual display terminal literature, from site visits to ad- ed control system (ICS) electnc power supply failures have ' vanced control rooms in the nuclear power and related indus- been analyzed for the Oconee Unit 1 nuclear plant. The instru-tnes, and from a study of technical reports on computerized ment and control system power distnbution circuits were ana-control rooms. This report should assist the staff in develop- lyzed to define a comprehensive set of 19 single-point failure ment of a regulatory position regarding the design of computer- modes. For each power supply failure, the fanied and operahng ired control room annunciator systems. The guidance in this control system signal inputs were propagated through the par- ) bally energized control system circuits as well as the energized report it consistent with that provided in NUREG-0700. i i
~~~ _ . - _ ._ -. - - - . - - .. -- - - .~~
l i Main Citations and Abstracts 55 I g,nd deenergized output control devices to evaluate the initial at Pacife Northwest Laboratory pnmanly to determine the sus-plant response. In addition, the effects of the power supply fail- ceptibility of irradiated pressunzed-water reactor Zircaloy-4 clad-ures on the pnncipal control room parameter displays were ding to failures caused by pellet-cladding mechanical interacton combined with the initial plant response to the automate control (PCM1). A secondary objective was to acquire kinetic data (e g., 4 circuits to evaluate possible control room operator responses. ndge growth or relaxation rates) that might be helpful in the in. Plant responses to the defined power supply failures are de- terpretation of in-reactor performance results and/or the model-scnbed in detail The automatic responses of the plant to the ing of PCMI. No cladding failures attnbutable to PCMI occurred instrument and control system power supply failures were not durir,g the six tests. This report describes the testing methods, found to be severe. Possible operator responses to spunous testing apparatus, fuel rod diametral strain measunng devce, control room displays generally did not result in signifcant tran- and test matrix. Test results are presented and discussed. ssents. Improved automate transfer of control system input cir-cuits to operable power supplies, automatic Inp of feedwater NUREG/CR-4003: CLOSEOUT OF IE BULLETIN 79-04:lNCOR-pumps on loss of certain power supply branch circuits, and sup. RECT WEIGHTS FOR SWING CHECK VALVES MANUFAC-pression of spunous alarms have been identified as possible TURED BY VELAN ENGINEERING CORPORATION.
; ways to further hmst the effects of transients. FOLEY,WJ.; DEAN.R.S.; HENNICK,A. Parameter, Inc. June NUREG/CR-3992: COLLECTION AND EVALUATION OF COM- 1985. 29pp. 8507030702. IE 143. 31314.067.
PLETE AND PARTIAL LOSSES OF OFF-SITE POWER AT NU. IE Bulletin 79-04 was issued March 30, 1979 as a result of CLEAR POWER PLANTS. BATTLE,R.E. Oak Ridge National reports from three facihties that Velan Engineenng Corporation Laboratory. February- 1985. 63pp. 8502280542. ORNL/TM- had provided incorrect weights for swing check valves. The 9384, 29169:202 reason for concern was the possibility that these incorrect Events involvirig loss of off-site power that ha<e occurred at weights had been used in anafyses of Seismic Category 1 piping nuclear power plants through 1983 are desenbed and catego- systems at a large number of plants. Evaluation of utility re-l nzed as complete or partial fosses. The events were identified sponses and NRC/IE inspect:on reports shows that the bullet n' as plant-centered or gnd-related failures. In addition, the causes can be closed out for 117 (92%) of the 127 current facilities on of the failures were classified as weather, human error, design the basis of specife entena. Followup stems for the remaining error, or hardware failure. The plant-centered failures were usu- 10 current facilities are proposed for use by NRC/IE. Incorrect ally of shorter duraton than the weather related grid failures, weights reported for valves other than Velan swing check i For this reason, the weather related events were reviewed in valves are identified as Remaining Areas of Concem. This bulle-detait Design features that may be important factors affecting tin has served its purpose and can be closed out. A final check off-site power system reliability were tabulated for most of the of valve weights will be made per later IE Bulletin 79-14 on seis-operating nuclear power plants. The tabulated information was mic analyses for as-built safety-related peping systems. [ provided to NRC for a statistcal analysis to determine the im-l portance of these design features for losses of off-site power. NUREG/CR-4004: CLOSEOUT OF !E BULLETIN 79-25 FAIL-
~~. The frequency of losses of off-site power versus durabon were URES OF WESTINGHOUSE BFD RELAYS IN SAFETY RELAT.~~
estmated for three time penods. The frequency of loss of off- ED SYSTEMS. FOLEY,W.J.; DEAN,R.S.; HENNICK,A. Parame-site power was estimated to be 0.09/ reactor year based on in- ter, Inc. Apnl 1985. 44pp. 8505010088. IEB 79-25. 30112:260. dustry-wide data for the years 1959 through 1983. Robinson 2 submitted LER 78-29 December 19,1978 to re ort stching of a normally energized Westinghouse BFD relay. J NUREG/CR-3998 V02: LIGHT WATER-REACTOR SAFETY MA- After reviewing this problem West nghouse issued Service TERii,LS ENGINEERING RESEARCH PROGRAMS.Ouarterfy Letter TS-E-412 to recommend that BFD relays in safety related )l
^
Progra ss Report,Apni-June 1984. SHACK,W.J.. Argonne Nation- systems be replaced with later NBFD relays. Dunng installation at Laboratory. Apnl 1985. 92pp. 8504220361. ANL-84-60. and testng of the new NBFD relays, Robinson 2 found some 29946:069. with marginal or unsatisfactory armature overtravel. Because of This progress report summan::es the Argonne Nabonal Labo- this new problem, Westinghouse issued Techncal Bulletin NSD-f ratory work performed dunng Apni, May, and June 1984 on TB 79-05 to recommend prorrpt checking of certain models of 7 water reactor safety problems related to out-of-core matenals. NBFD relays and returning those with snadequate overtravel for The research and development areas covered are Environmen- rework or replacement. IE Buitetin 79 25, with extracts of the taffy Assisted Cracking in Ught Water Reactors, Long-Term Em- Westoghouse servce letter and technical bul!etn enclosed, bottlement of Cast Duplex Stainless Steels in LWR Systems, was issued November 2,1979 to require responses and specific and Nondestructive Evaluation and Leak Detection. actions by all Iscensees and holdors of construction permits with NUREG/CR 3998 V03: LIGHT WATER-REACTOR SAFETY MA, respect to BFD and NBFD relays in safety-related systems. TERIALS ENGINEERING RESEARCH PROGRAMS.Ouarterty Evaluation of utility responses a .. NRC/IE inspecton reports Progress Report, October-December 1984. SHACK,W.J. Ar- indcates that the bulletin can be closed out for 121 (94?.) of gonne National Laboratory. October 1985. 75pp. 8512120141. the 129 cunent facilities on the basis of specific cnteria. Pro-ANL 84-60 33873.039. posed foil wup stems for the remaining 8 facilities are preser-ted This progress report summanzes the Argonne National Labo- 'n Appendix C for use by NRC/IE. Because followup of correc-ratory work performed dunng October November and Decem- tive action is ensured, IE Bulletin 79-25 is considered closed. ber 1984 on water reactor safety problems related to out-of-core materials. The research and development areas covered NUREG/CR-4005: CLOSEOUT OF IE BULLETIN 8012. DECAY are Environmentally Assisted Cracking in Light Water Reactors HEAT REMOVAL SYSTEM OPERABILITY. FOLEY,W.J.; and Long-Term Embnttlement of Cast Duplex Stainless Steels DEAN R.S.; HENN!CK.A. Parameter, Inc. June 1985. 59pp. in LWR Systems. 8507030675. lE-146. 32106:149. On Apnl 19,1980, decay heat removal (DHR) capability was NUREG/CR-3999: ELECTRICALLY HEATED EX REACTOR lost at Davis-Besse 1 for approximately two and one half hours PELLETCLADDING INTERACTION (PCI) SIMULATIONS UTI- in a refueling mode. Typically for that rnode, many systems ar.d LIZING 1RRADIATED ZlRCALOY CLADDING. BARNER,J.O.; components were out of survice for maintenance and testing or i FITZSIMMONS,D. Battelle Memonal insttute, Pacific Northwest were deactvated to preclude inadvertent actuation. IE Bulletin 4 Laboratories. February 1985.104pp. 8503120445. PNL-5245. 80-12 was issued May 8,1980 for action by licensees of operat- , 29340:190. ing pressurized water reactori (PWRs); it was issued for infor-1 In a program sponsored by the Fuel Systems Research mahon to nuclear pnwer facilities other than operating PWRs. j Branch of the U.S. Nuclear Regulatory Commission, a series of The intent of the bulletin was to improve nuclear plant safety by six electnCally heated fuel rod simulaf on tests were Conducted ' reducing the likelihood of losing DHR capability in PWRs, espe.
56 Main Citations and Abstracts cially when some DHR components are unavaiiable because of of the data processing procedures and personnel required, and marntenance activities dunng refuehng and cold shutdown a Data Review Demonstraton and Evaluation involving mem-modes of operatic, t A related NRR Genenc Letter was issued bers of the potential user population. The conclusons of this June 11, 1980 to Icensees of operating PWRs. requesting study were used to modify and improve the detailed implemen-amendment of technical specifications to ensure long-term tation specification. The revised spuufication is pubhshed as maintenance of DHR capabihty. Evaluaten of utility responses NUREG/CR-4010, and it desenbes all the necessary matenals,' and NRC/IE inspection reports indcatas that the bulletin can be personnel, procedures, definitions, and data taxonomies to im-closed out per specific cntena for 33 (75*.) of the 44 affected piement the data bank. facilities. NUREG/CR-4010: SPECIFICATION OF A HUMAN RELIABILITY NUREG/CR-4006: CLOSEOUT OF tE BULLETIN 81-01: SURVEIL. LANCE OF MECHANICAL SNUBBERS. FOLEY,W.J.; DATA BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLEAR POWER PLANTS. COMER.M.K.; DEAN R.S; HENNICK,A. Parameter, Inc. August 1585. 02pp. DONOVAN M D. General Physics Corp.
~~Sandia Natonal Lab-8508260307. IE-145. 32368 219.~~
In the penod from August 1974 to May 1980, failures of me- oratones. Apr i 1985. 400pp. 8505060499. SAND 85 7151. chanical snubbers were desenbed in event reports issued for 30190 331. nine facilities and in a NRC/IE study of the DOE Fast Flux Test This document specifies the personnel, resources, policies, Facility. In most failures, the snubbers were frozen and would and procedures for implementing and operating a human reli-not permit free piping motions dunng thermal transients. In ability data bank specifically tailored to support human reliability some cases, the failed snubbers no longer provided seismic analysis (HRA) segments of probabikstic nsk assessments shock restraint. Because of concern about the reported failures (PRAs) for nuclear power plants. The report concludes a three-of mechanical snubbers, standard technical specification rev'- year research program conducted by the U.S. Nuclear Regula-sions for snubber surveillance were issued by NRC/DL on No- tory Commission and Sandia National Laboratones. Previous ef-vember 20, 1980. IE Bulletin 81-01 was issued January 27 forts of the program include a review of existing human per. 1981 to require examinaton and testing of mechanical snubbers formance data banks (NUREG/CR 2744, Vol.1), a concept and in safety-related systems at Icensed facilities and at selected system desenption (NUREG/CR-2744, Vol. 2), and a peer eval-facilities under construction. Evaluation of utikty responses and uation study (NUREG/CR-4009). This report specifies the ad-NRC/IE inspection reports indicates that the bulletin can be ministraWe organization of the data bank functional groups and closed out per specific entena for 73 (95%) of the 77 facilities their proposed interaction. Detailed procedures are included to which it was issued for action. Followup stems are proposef that specify how to process submitted data, how to classify and for use by NRC/IE to ensure sat sfactory completion of correc- store the data, and how to combine similar data when appropri-tive acten at the remaining four (4) facil t:es. ate. Included with.n the report is the skeleton data manual, NUREG/CR-4008: GENERAL EXTRAPOLATION MODEL FOR which is a prototype, hardcopy, data manual that would be used AN iMPORTANT CHEMICAL DOSE-RATE EFFECT. to disseminate data to the user population. It describes the data GILLEN.K.T.: CLOUGH,RL Sandia National Laboratones. Janu- taxonomy, procedures for retrieving data of interest, and pro-ary 1985. 51pp. 8503010341. SAND 84-1948. 29186:186. sents several sample data retneval problems. Definitions are in order to extrapolate material accelerated aging data, meth- supplied for all technical and behavoral terms used in the taxo-odologies must be developed based on sufficient understanding nomic structure. As its name implies, the skeleton data manual of the processes leading to matenal degradation. One of the most important mechanisms leading to chemical dose-rate ef- data. fects in polymers involves the breakdown of intermediate hydro, peroxide species A general model for this mechanism is de- NUREG/CR-4015: EFFECT OF STAINLESS STEEL WELD nved based on the undertying chemical steps. The results fead OVERLAY CLADDING ON THE STRUCTURAL INTEGRITY OF to a general formahsm for understanding dose rate and sequen- FLAWED STEEL PLATES IN BENDING SERIES 1 taal aging effects when hydroperoxide breakdown is important. CORWIN.W.R.; ROB'NSON,G C.; NANSTAD R K.; et af. Oal We apply the model to combined radiation / temperature aging Ridge National Laboratory Apnl 1985 103pp 8505230524-data for a PVC matenal and show that this data is consistent ORNL/TM-9390. 30549 090. with the model and that model extrapolations are in excellent The HSST sta.nless steel cladding evaluatons were initiated agreement with 12-year real-time aging resu!ts from an actual nuclear plant. This model and other techniques discussed in this to study the interac' ion of stainfess cfadding with flaws enetiated report can aid in the selection of appropnate accelerated aging in and propagating in base metal of reactor pressure ves'els. A methods and can also be used to compare and select matenals complicating factor in understanding the role of stainless clad-for use in safety-related components. This will result in in- ding in this setting is its toughness as a function of radiation creased assurance that equipment qualification procedures are dose and fabncation process. The initial phase of this study ad-adequate. dressed this question by testing the response of specimens clad with single wire subraerged-arc weld overlay in varying NUREG/CR-4009: HUMAN RELIABILITY DATA BANK Evaluation toughness levels. The tests completed under the initial phase of Results. . COMER.M K; DONOVAN M D.; GADDY,C.D ; et at this study indicate that cladding of moderate Charpy toughness General Physics Corp. April 1985. 75pp 8505070505. SAND 85 has only limited capabilities to stop running cracks. This was a 7150. 30210 295. limited set of expenments, and the upper and lower bounds of The U.S. Nuclear Regulatory Commission and Sandra Nation. al Laboratones conducted a three-year research program to de. cracks arrest capabilities are not yet de' ermined. The fabncaton techniques employed for this first tones of tests have resulted velop a human reliability data bank specificalfy tailored to sup-in conditions that have prevented close control of the stress port human rehability anafysis segments of probabilistic nsk as-sessments for nuclear power plants. Previous efforts of the pro- state at pop in of the hydrogen-charged EB welds Consequent-gram include a review of existing human performance data fy, the arrest toughness of the stainless cladding was not close-banks (NUREG/CR-2744, Vol.1) and a cencept and system de- ly bounded. General modifications are proposed for incorpora-senption (NUREG/CR-2744, Vol 2) Subsequent to the system tion in a second senes of tests to provide more comprehensive desenption, a detailed specification for implementing the data conditions of testing and matenals of interest, to eliminate some bank was developed. An evaluaton of this specification was ur. desirable test conditions that existed in the first senes, and to conducted and is desenbed in this report. The evaluation con- provide an improved geometry for analytical interpretations sisted primanly of a, Operability Demonstration and Evaluation
Main Citations and Abstracts 57 NUREG/CR 4016 V01: APPLICATION OF SLIM-MAUD.A TEST This report presents the status of Pacific Northwest Laborato. OF AN INTERACTIVE COMPUTER-BASED METHOD FOR OR. ry's program through the end of FY 83 assessing the perform. GANIZING EXPERT ASSESSMENT OF HUVAN PERFORM. ance of synthetic liners used in uranium ta; lings ponds Synthet-
'ANCE AND RELIABILITY. Volume I Main Report. ROSA.E.A; ic hner fadure mechanisms, impoundment design, installation, HUMPHREYS.P.Ca SPETTELL.C M.: et al. Bro % haven National and inspection techniques are presented frc a information col.
Laboratory. September 1985. 54pp. 8512270232. BNL-NUREG-51828. 34084.323. tected from consultants. rMi operators, and the synthette liner The U.S. Nuclear Regulatory Commission (NRC) has been industry. Progress is reported on laboratory tests on accelerated conducting a mutli-year research program to investrgate differ- aging of liners, physical properties of aged matena!s, and non. ent methods for us;ng expert judgments to estimate human destructive examination of seams. error probabihties (HEPs) in nuclear power plants. One of the methods investigated, denved from mu!ti attr,bute utility theory- NUREG/CR 4030: RADIONUCLIDE MIGRATION IN GROUND is the Success Likehhood Index Methodology implemented WATER (Final Report) FRUCHTER.J.S ; 'COWAN.C E through Multi-Attibute Utikty Decomposition (SLIM MAUD). This ROBERTSON.D E , et al. Battolie Memonal Inst,tute. Pacif$c report describes a systematic test application of the SLIM- Northwest Laboratones. March 1985.56pp.8504030419 PNL-MAUD methodology The test application is evaluated on the 5299. 29605~ 123' basis of three enteria: practicality, acceptability, and usefulness. For the past several years, data on radionuchde migration in Volume I of this report presents an overview of SLIM MAUD. ground water at a low level disposal sita were colected Most of desenbes the procedures followed in the test apphcation, and the radionuchdes were removed in the disposal basin and provides a summary of the results cbtained. Volume 11 cons 4sts trench by estner prec pitation or adsorption mechanisms. How-of technical appendices to support in detail the rnatenals con- ever, three radionuchdes (60)Co. (106)Ru. and (125)Sb showed tained in Volume I, and the user's package of exphcit proce. somewhat greater than espected mobihty The elements of dures to be followed in implement;ng SLIM MAUD. The results these three isotopes were found to be in either anionic or non-obtained in the test apphcation provide support for the apphca- ionic charge-forms. Comp' emes with both natural and r9an-made tion of SUM MAUD to a wide vanety of applications requinng organics were impbcated in the increased mobihty, particulany in estimates of human errors. the case of (60)Co Charactenzation studies of the organic frac-NUREG/CR-4020: HMSA COMPUTER PROGRAM FOR tion were performed Ruthenium 103. (60)Co. and (125)Se were TRANSIENT THREE DIMENSIONAL MIXING GASES. found to be associated w:th the higher molecutar weights great-TRAVIS.J.R. Los Alamos Scientfic Laboratory. Fetiruary 1985. er than 1000. Studies were also performed that proved the hy. 52pp. 8503210474. LA 10267 MS. 29479.114. pothesis that the adsorption behavior of (235)No on soils of the i ' A numencal technique has been developed for calculabng the site is dominated by adsorptron on iron hydroxide Fina'ly, geo-futt three-c rienseat time-dependeat equa$cns of mobon w,9 chemical modeling of 'he chemical and charge form data mult;ple species transport. The method is a modif ed form of the showed the ground water to be in equebnum with several solids imphcit Continuous-fluid Eulenan (ICE) technique to solve the that could be important in controtiing the concentrat.ons of trace governing equations for low Mach number flows where pressure elements and radionuclides waves and local vanatrons in compression and expansion are not significant. Large density vanations. due to thermal and spe, NUREG/CR 4031 V02: NEUTRON SPECTRAL CHARACTER'ZA-cies concentration gradients, are accounted for wrthout the re. TION FOR THE FIFTH HEAVY SECTION STEEL TECHNOLO-stnctions of the classical Boussinesq approximaton Example GY (HSST) IRRADIATION SERIES. "Neutronics Catculatons " calculations of the EPRl/HEDL standard problems verify the WILLIAMS.L; REMEC.l; KAM.F B Oak Ridge National Labora-feasibility of using this firute-d;fference technique for analyzing tory. May 1985. 42pp. 8505170010 ORNL/TM.9423/V2. hydagen transport and mmng within LWR containments 30484 001. NUREG/CR-4022: PRESSURIZED THERMAL SHOCK EVALUA- A senes of calculations has been completed to corepute do-T!ON OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER simeter actvation in the Oak R dge Research Reactor (ORR) PLANT. BALL,0 G.; CHEVERTON R D.; FLANAGAN,G.F : et al. HSST Simulator Expenment. A companson of ca:culated and Oak Ridge National Laboratory. September 1985. 702pp expenmental results shows that calculabons underpredict do-851227C369 ORNL/TM-9408. 34074 038. 5' meter activit:es on the average of aoout 15% The C/E valves An evaluation of the nsk of pressunzed thermal shock (PTS) indicated the now farnifiar tendency to become lower as more resulting in a through the-wall crack in a reactor pressure vessel iron is penetrated The dosameters in front of the simulator (and was performed for the Calvert Chffs Unit I nuclear power plant. behind the thermal shield) typicaffy have C/E values of 0.910. The information presented in this report covers one of three while those at the back of the simulator have values of 0 7 0 85 plant specific stud <es performed for NRC. The other two studies, The calculations also show shifted axial distnbution relative to for Oconee Unit 1 and H. B. Robinson Unit 2. are documented the measurements the C/E values near the bottom of the core in NUREG/CR 3770 and NUREG/CR-4183, respectively. The are about 15% to 20% higher than those r' ear the top This is specific objectives of the Calvert Chtfs study were (1) to further probably due to a descrepancy in the ax al power d stnbuton refine the methodology for evaluating the nsk of PTS, (2) to pro- computed with VIPOR/ VENTURE. The asial distr:bution of fuel vide a best estmate of the frequency of a through-the-wall obtained from the correlation in VIPOR could possibly be caus-crack for the Calvert Chffs Unit 1 vessel. (3) to determine the ing the power shift, altnough this speculation has not been ven-dominant PTS sequences for the unit. and (4) to evaluate the effectiveness of potential corrective measures. The examination r,ed There is also, perhaps, a shgnt tendency for the C/E values to increase along the x (the coordinate parallel to the re-of tens of thousands of transients indicated that PTS was not an important core melt initiator for Calvert Chtfs Unit 1. The actor face) traverse from the reactor centerline to a point near dominant risk sequences were determined to be small-break the core boundary; however, this vanation is much less than sn LOCAs at low core decay-heat cond.tions which led to total the axial direction. *he results obtarned with different dosi-soop flow stagnation. meters appear generally to be reasonably cons stent, except that the (46)Ti C/E values seem to be consistently lower than NUREG/CR-4023: FIELO PERFORMANCE ASSESSMENT OF for the other dosimeters The systematic nature of the discreo-SYNTHETIC LINERS FOR URANIUM TAILINGS POND.A ancies in these calculations will be adjusted by the least Status Report. MITCHELL.D H.: SPANNER,G E. Batteue Memo-squares procedure to produce an accurate representation of the nal Institute. Pacific Northwest Laboratones. January 1985 flux distnbution. 78pp. 8502040788. PNL-5005. 28716 0,96
58 Main Citations and Abstracts major sections The first deals witi the nature and circL.m-NUREG/CR-4031 V03: NEUTRON SPECTRAL CHARACTERIZA- stances of tne accident and findings of fact. The second gives TION FOR THE FIFTH HEAVY SECTION STEEL TECHNOLO- an accounting and desenption of the matenafs involved and the GY (HSST) IRRADIATION SERIES. " Neutron Exposure Param-eters." REMEC,l.; STALLMANN F.W.; KAM,F.B. Oak Ridge Na- consequences of their exposure. The third gives an assessment t.onal Laboratory. May 1985. 31pp. 8505160647. ORNL/TM- and analysis of the mechanisms of damage and the conclusions which may be drawn from the investigation. 9423/V3. 30457:181. This is the third volume of a three volume report which de-scobes the simulator expenments of the fifth senes of HSST ir- NUREG/CR-4037: DATA
~~SUMMARY~~
REPORT FOR FISSION RELEASE TEST HI-5. OSBORNE.M F.; radiation expenments which are sponsored by the U S. Nuciear PRODUCT COLLINS J L ; LORENZ,R.A.; et al. Oak Ridge National Labora-RegJatory Commission (NRC). The purpose of these three vol- 75pp 85080907t2. ORNL/TM-9437. umes is to document, in detail, the experimental and calcula- tory. July 1985. tonal methodology which wiH be used in determining the neu- 32104 055 tron-exposure parameters for the fifth and subsequent senes of The f.fth in a senes of high-temperature fission product re-HSST irradiaton expenments at ORNL. The methodolgy was lease tests was conducted for 20 min at 1700 degrees centi-also used in the fourth series of HSST irradiation expenments grade in flowing steam. The test specimen, a 15 2-cm-long sec-and represents the current state-of the-art procedures devel- tion of a fuel rod which had been irradiated to a burnup of 38,3 oped in the Light Water Reactor Pressure Vessel Simulation mwd /kg was heated in an induction fumace under simulated Project which is a part of NRC's Surveillance Dosimetry im- LWR accident condibons in a hot cell. Fosttest inspecton provement Prog'am. The neutron-exposure data from the fifth showed severe oxidation and fragmentation of the fuel speci-and subsequent senes will be documented in a loose-leaf men, but no cladding melting was apparent. Analyses of test NUREG/CR report as the data become available. In this components showed total releases from the fuel of 19 9*. for volume, the best estimates for the values and spatial distribu- (85)Kr. 22 4*. for (129)l.18 0*. for (110m)Ag. and 20.3*. for bon of fiuence rate (0) (E > 1.0 MeV) fluence rate (0) (E > 0.1 (137)Cs. A smaller fracten of the (125)Sb (0 326*.) was re-MeV), and displacements per atom per second (dpa/s) are de- leased from the fuel, and 99*. of the (110m)Ag and (125)Sb termined using LSL M2, a least squares logarithmic spectrum p g g adjustment procedure w th input values taken from dosimetry men indicated a (134)Cs release of 24.5*., which is reasonably data from Vol. I and neutronics calculations from Vol 2. These estimates have an overall uncertainty of less than 20 . relative less than half those in test Hi-2, where more oxidation and a standard deviation. This volume is essential to the metaflurg st for defining the irradiabon strategy to meet his objective (s). NUREG/CR-4033: THE ROLE OF PERSONAL AIR SAMPLING IN RADIATION SAFETY PROGRAMS AND RESULTS OF A LAB- NUREG/CR 4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE CRAC2 COMPUTER CODE. KOCHER,0Ca ORATORY EVALUATION CF PERSONAL AIR SAMPLING WARD.R,C ; KILLOUGH.G G ; et al. Oak Ridge National Labora-EQUIPMENT. RITTER.P.D; HUNTSMAN.B L; NOVICK,V.Ja et al. EG&G, Inc. May 1985. 80pp. 8505230534. EGG 2352. tory May 1985. 247pp. 8507250201. ORNL 6114 31793 025. Ths report presents a study of the sensitmty of reactor acci-30549 256. Recommended apphcations for personal air samplirg in NRC dent consequences predicted by the CRAC2 computer code to bcensee radiabon protecting programs are presented The rec- uncertainties in selected models and parameters used in the ommendations are based on perfo.mance tests of currently code The sources of uncertainty that were investigated include ava41able samplers, a review of research and regulatory h' era- (1) the model for plume n$e, (2) the model for wet deposition, ture, and a survey of current hcensee air samphng programs. (3) the meteorological bin-samphng procedure for selecting The performance tests show that personal air sample s are weather sequences involving rain, (4) the dose conversion fac-available which can provide a rehable, convenient means for tors for inhalation as they are affected by uncertainties in the breathing-zone samphng of workers in practically any work envi- physical and chemical form of the released radionuchdes, (5) ronment which might be encountered in the licensee industnes the weathenng half time for external ground-surface exposure, The research hterature emphasized that estimates of an individ- and (6) the transfer coefficients for terrestnal foodchain path-ual's exposure may be greatly underestimated if based on gen- ways. The most important sources of uncertainty in our analy-eral area air samples, as is common practice in current hcensee ses were the choice of wet deposition model, the dose conver-programs, due to the unpredictable vanability of airborne-actmty sson factors for inhalation, and the weathering haff-time for concentratons in the worksite. A conclusion which may be ground surface exposure The choice of plume-nse model, the drawn from the hterature and from expenmental results is that in use of an alternative bin-samphng procedure, and uncertainties mcst situations, personal air samphng (or more generally, truo in terrestnat foodchain pathways usually had ins,gnificant effects breathing-rene sampling) is the only means to rehably estimate on CRAC2 prediction. the airborne activity to which a worker has been exposed (VPC h). Research concerning the appbcabity of air-samphng meas- NUREG/CR 4039: GAMMA-RAY CHARACTERl2ATION OF THE urements for est mat ng intake, uptake, and intemal dose was TWO-YEAR IRRADIATION EXPERIMENT PERFORMED AT THE POOLSIDE FACILITY, MAERKER R E. Oak Ridge National also reviewed-Laboratory. January 1985.21pp.8502220412 ORNL/TM-9440, NUREG/CR-4035: A HIGHWAY ACCIDENT INVOLVING RADIO- 29073 230 PHARMACEUTICALS NEAR BROOKHAVEN.MlSS:SSIPPI ON MONT,M E.; Average gamma-ray group fluence rates are calculated for 3,1983. MOHR.P 8; DECEMBER each of the three exposures in the two-year metallurgical bhnd SCHWARTZ M W. Lawrence Livermore National Laboratori test experiment at the ORR-Poolside Facility in Oak Ridge, thus Apnl 1985. 52pp. 8505070560. UCRL 53587, 30210:169. completing the charactenzation of the radiation held for this ex-A rear end colhsion occurred between a passenger automo-penment, which is intended to serve as an internatzonal metalur-bile and a luggage tra.fer carrying 84 packages, 76 of which gical benchmark. Heating rates in the steel derived from these contained radiopharmaceuticals, on U.S Highway 84 near calculations varied from about 0 23 watts / gram in the simulated Brookhaven, Mississippi on the afternoon of December 3,1983. surveillance capsule to 1.4 milbwatts/ gram at the three-quarters l The purpose of this report is to document the mechanical cir- depth location in the simulated pressure vessel capsule, with I cumstances of the accident, confirm the nature and quantary of secondanos ansing from non-fission reactions in the core and l radioactive matenals involved, and assess the nature of the ex-core steel contnbuting between seventy seven and ninety-physical environment to which the packages were exposed and three percent of the totaf Contributions from photofission to fis-l the response of the packages. The report consists of three l l l
l Main Citations and Abstracts 59 sion foil activities are estimated to be less than five percent of followed and concentraton vs. time telationships can be deter-those prevously calculated ansing from neutron-induced fissen. mined anywhere in the system. Some examples are sncluded. NUREG/CR-4040: OPERATIONAL DECISIONMAKING AND ACTION SELECTION UNDER PSYCHOLOGICAL STRESS IN NUREG/CR-4043: DATA
~~SUMMARY~~
REPORT FOR FISSION NUCLEAR POWER PLANTS. GERTMAN.D.I.; JENKINS.J.P ; PRODUCT RELEASE TEST Hi-6. OSBORNE.M F.; HANEY,L.N ; et al. EG&G Idaho. Inc. (subs. of EG&G, Inc ) May COLLINS,J.L; LORENZ,R.A.; et at Oak Rdge Natonal Labora-1985. 68pp. 8507020109. EGG-2387. 31313:236. tory. October 1985. 65pp. 8511220329. ORNL/TM 9443. 33602.318. An extenswe review of hterature on individual and group per-formance and decisionmaking under psychological stress was The sixth in a senes of high-temperature fission product re-conducted and summanzed. Specific stress related vanables lease tests was Conducted for 1 minute at 1950 degrees centi-relevant to reactor operaton were pinpointed and incorporated grade in a steam-helium atmosphere. The 15 2-cm-long test in an expenment to assess the performance of reactor opera- specimen was a section of fuel rod which was irrad.ated to 40.3 tors under psychological stress. The decisonmaking perform- mwd /kg in the Monticello BWR. Posttest ana:yses showed total ance of 24 reactor operators under diffenng feve:s of workload. releases of 29 6*. for (85)Kr (includes 2*. that was released to confhcting mformation, and detail of available wntten procedures the plenum dunng irradiation), 33.1*. for (137)Cs 24.7*. for was assessed in terms of selecting immediate, subsequent, and (129)l,6 0** for (110m)Ag, and 0 06*. for (125)Sb. A stainless nonapphcable actons in response to 12 emergency scenarcs steel thermal gradient tube was used to examine the retenten resulting from a severe seismic event at a pressunzed water re- of fission product cessum by stainless steel Gainma scans actor. Specific personality charactenstics of the operators su9-showed that a significant fracton of the released cesium was gested by the hterature to be related to performance under released at temperatures >600 degrees centigrade where the stress were assessed and correlated to decisionmaking under surface had been oxidized Cesium that was released as Csl stress appeared unaffected by its contact with the sta:nless steel. A companson was made of Cs. I, and Kr release rate coeff<cients NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. obtained in the HI and HT test senes with NUREG-0772, "Tec4 LARSON J.R. EG&G. inc. January 1985. 75pp. 8502210404. nical Bases for Estimating Fisson Product Behavior Dunng EGG-2354. 29057.349. LWR Accidents,', values. This handbook provides simple procedures for calcu;ating the behav cr of hght water reactors dunng a vanety of incidents. It NUREG/CR-4044: TRAC-PF1 LOCA CALCULATIONS USING provides an additional tool for assessment of ongoing and pos- FINE NODE AND COARSE NODE INPUT MODELS tincident behavior. The handbook consists of a main body de- DOBRANICH D; BUXTON,L D.; WONG,C.N. Sar.daa Natonal scnoing genenc procedures, an appendix providing specific Laboratones. May 1985. 86pp. 8506190042. SAND 84 2305 design data for a kmated number of plants for appkcaten with 31015:116. the procedures, and an appendix listing existing and planned TRAC PF1 calculatons of a 200*. cold. leg break LOCA have BWR and PWR plants by containment types and thermal-hy- been completed for a UHI plant using both fine-node (with 776 drauhc parameters. The procedures are currently hmited to mesh cells) and coarse-node (with 320 mesh cells) input break flow rate, decay heat power and entegrated power, steam models. This study was performed to determine the effect of grneration from decay heat, mass balance, shutdown margin, noding on predicted results and on computer running time, it natural c rculation, noncondensable gas generation, dose esti- was found that the overall sequence of events and the impor-mates, and DNB evaluaten void formation in the upper head' tant trends of the transtent were predicted to be nearly the and torus heatup. same with both the fine-node and coarse. node models. There were differences an the time-dependent behaver of the cold 4eg NUREG/CR-4041 RO1: SYSTEM ANALYSIS HANDBOOK. accumulator injection, and the predicted PCT for the coarse-LARSON.J.R. EG&G inc. November 1985. 97pp. 8512050433 node calculation was about 75 K less than that or the fine-node EGG-2354. 33776:154 calculation. The higher PCT of the fine-node calculation is attnb-See NUREG/CR-4041 aostract. uted pnmanly to three-dimensionat flow effects in tr core The complete (steady state plus transrent) coarse. node calculation NUREG/CR-4042: A 3-DIMENSIONAL COMPUTER MODEL TO requred 13 5 hours of CYBEM 76 computer time compared to SIMULATE FLUID FLOW AND CONTAINMENT TRANSPORT 68.3 hours for the fine node calculation, yielding an overall THROUGH A ROCK FRACTURE SYSTEM. HUANG.C.; factor of five decrease in running time Tf us, we conclude that EVANS.D D. Anzona, Univ. of Tucson, AZ. January 1985. 116pp. 8502210181. 29051:171. for any large break LOCA analyses 6n which only the overall A 3-dimensional fracture generating scheme is presented trends are of concern, the loss of accuracy resulting from use of such a coarse-node model will normally be inconsequential which can be used to simulate water flow and containment compared to the savings in resources that are reahzed Howev-(solute) transport through fracture system of a rock. It es pres-entfy hmited to water saturated conditions, zero permeabihty for er, if the objective of the analyses Is the anvestigation of the ef-the rock matnx, and steady state water flow, but allows for tran- fects of muttedimensenal flows on clad temperatures then a detailed model is required. sient solute transport. The scheme creates finrte planar plates of uniform thickness which represent fractures in 3-dimensional NUREG/CR-4045: LITERATURE REVIEW ON AEROSOL SAM-space. A given fracture (plate) has tne followirg cesenptors: center location, onentation, shape, areal extent and aperture. PLINO DEVICES FOR RESPIRATORY FIELD STUDIES SUTCLIFFE.C.R Los Alamos Scientific Laboratory. February Esch parameter can be desenbed by an appropnate probability 1985. 58pp. 8502250844. LA 9977 MS. 29095 216. distnbuton. Individual fractures are generated to form an as. As part of the first phase of a Respirator Field Performance semblage of a certain fracture dens:ty. All fracture intersections project for the Occupational Safety and Heaith Administraten/ and boundary / fracture intersections are determined and dea- Nuclear Regulatory Commission, s entical review of the htera-dend fractures are ehmanated. Flow through the fracture assem. ture avadable on respirator protection studies was completed bt!.ge is considered faminar and desenbed by Poiseuille's law Little information was available on expenmental conditions, and The pnnciple of mass conservation at each intersection is used when the informaton was available, each study was different in to develop the global matnx equation, which is solved subject to how the aerosol measurements were made and in which param-specified boundary conditons to yield the head and flow distn- eters were controlled. Under these cond4tions, it as difficult to bution at each intersection. Solute transport is considered to be compare results obtained from different investigators. The htera-advective between intersectons with complete mixing at each ture was also surveyed for characteristics desirable in an aero-intersecton, Solutos added to the flow system can be exphertly sol-samphng erlet in order to representatively sample respirable
60 Main Citations and Abstracts particles. Available ambient aerosol samplers were entically re- The major conclusens of the report are: a substantial amount viewed for their performance charactenstics. Recommendatens (approximately 2/3) of job-related academic knowledge is cov-are made to 6 void the pitfa!!s present in many respirator field ered in college engineenng curnculum; college engineenng cur-studies and to he'p standard:ze these stud es. neulum provides considerable matenal beyond that #dentified as necessary for heensed operators; higher level operator positions NUREG/CR-4046: DETERM NING CRITICAL FLOW VALVE (SS relative to SRO. SRO relative to RO) were judged as need-CHARACTERISTICS USING EXTRAPOLATION TECHNIOUES. ing higher levels of education to perform the icb. JARRELL.D.B. EG&G, Inc. March 1985. 28pp. 8504030456. EGG-2357. 29605 220. NUREG/CR 4055: THE D10 EXPERIMENT.COOLABILITY OF This report presents the methodology and documentaten of UO2 DEBRtS IN SODIUM WITH DOWNWARD HEAT REMOV-the cahbration of the Loss of-Fluid Test (LOFT) power-operated AL. MITCHELL,G W; OTTINGER C.A : MEISTER.H. Sandia Na-relief and safety relief valve (PORV + SRV) for the L9-3 antici- tional Laboratones. January 1985. 84pp. 8503050012. SAND 84-pated transient without scram (ATWS) expenment A multiposi- 1144. 29246 001. tion globe va've was cabbrated to produce scaled high-pressure The LMFBR Debns Coolabuty Program at Sandia Natenal flow rates vsmg a low pressure cahbration facility and a simple Laboratones investigates the coolabihty of particle beds wh :h RELAPS cntical flow model to extrapolate the cabbraton data to may form following a severe accident involving core disassem-expected operating pressures. It was demonstrated that an ac- bly in a nuclear reactor. The D senes expenments utilize fisson curate high pressure, multiphase flow cabbration can be per- heating of fully enriched UO(2) particles submerged in sodium formed without the necessity of actual high-pressure test:ng. to reakstically simut i decay heating The D10 expenment is Th;s technique, when apphed to large pressunzed water reactor the first in the senes study the effects of bottom cochng of (LPWR) sa'ety and rehet valves, represents a potentially large the debns which coulo ,,e provided in an actual accident conde-savings in the Capacity Qualtfacation procedure of full-scale pres- tion by structural materials onto which the debris might settle. sure reduction valves- Additionally, the D10 expenment was designed to achieve man-mum temperatures in the debns s;proaching the melting point NUREG/CR-4050: A REVIEW OF THE SHOREHAM NUCLEAR of UO(2). The experiment was successfully operated for over 50 POWER STATION PROBABILISTIC RISK hours and investigated downward neat removal in a packed bed ASSESSMENT. internal Events And Core Damage Frequency at specific powers of 0.16 to 0 50 W/g Dryout in the debns was ILBERG.D.; SHlU.K.; HANAN.N ; et al. Brookhaven National achieved at powers from 0.42 to 0.58 W/g Channels were in. Laboratory. November 1985, 330pp 8512190024. BNL/ duced in the bed and channeled bed dryout was achieved at NUREG 51836. 33966 065. powers of 106 to 1.77 W/g Maximum temperatures in excess A review of the Probabihstic Risk Assessment of the Shore. of 2500 degrees centigrade were attained ham Nuclear Power Staton was conducted with the broad cb, lect:ve cf evaluat ng its nsks in relation to those identified in the NUREG/CR 4056: PARTICULATE AND GAS RELEASE FROM Reactor Safety Study (WASH-1400) The scope of the review LIGHT. WATER-REACTOR (LWR) FUEL RODS STORED IN was hmited to the " front end" part, a e., to the eva!uaton of the INERT AND DRY AIR ATMOSFMERES. OLSEN.C S. EG&G. frequences cf states in which core damage may occur. Futher. Inc. January 1985. 24pp 8501210003 EGG-2359. 28497 013 more, the review considered only internally generated acc'- A testing program using eight commercial pressurized water dents, consistent with the scope of the PRA. The review includ- reactor (PWR) and boihng water reactor (BWR) spent fust rods ed an assessment of the assumptions and methods used in the was conducted to investigate their long-term stability under a Shoreham stuc*y. It also encompassed a re-evaluaten of the vanety of possible dry =torage cond'tions. The objectae of this main results w4nin the scope and general methodolog> cal project is to provide the Nuclear Regulatory Commission (NRC) framework of the Shoreham PRA, including both quahtatrve and with information to conf +rm or estabbsh heensing positons for quantitative analyses of accident initiators. data bases, and ac- dry spent fuel rod storage with regard to long-term, low-temper-cident sequences which result in initiation of core damage. Spe- ature (<250 degrees centigrade), spent fuel rod behuor dunng cific compansons are given between the Shnreham study, the dry storage and radcactive contamination ansing from spall-results of the present review, and the WASH 1400 BWR, for the ation of claddng crud. The results of the analyses of the crud, core damage frequency. The effect of modehng uncertarnties fuel particu' ate, and gas release from these eight fuel rods is was considered by a bmited senst:vity study so as to show how presented, which includes weight change measurements, de-the results would change if other assumptions were made. This layed neutron measurements, and isotopic analysis of smears review provides an independent?y assessed point value estimate used to assess the particulate release Gas analyses of the fuel of core damage frequency and desenbes the maior contnbutors, rod capsule environments were made to determine the fisson by fronthne systems and by accident sequences- gas release, and flow tests were performed to determine the extent of filter blockage from particle entrapment. NUREG/CR-4051: ASSESSMENT CF JOB RELATED EDUCA-TIONAL QUALIFICATIONS FOR NUCLEAR POWER PLANT NUREG/CR-4057: RADIOLOGICAL ASSESSMENT OF THE OPERATORS. SAARI L M ; MELBER.B D ; WHITE A.S ; et al. TOWN OF EDGEMONT. JACKSON.P O.; THOMAS.V.W; Battelle Human Attairs Research Centers. Apnl 1985. 77pp YOUNG J A. Battelle Memonal Institute, Pacific Northwest Lab-8505010277. PNL-5303. 30114.352. oratores. January 1985. 183pp. 8502040625. PNL-5320 This report identifies job-re!ated educatenal quahfications for 28727:151. the nuclear power plant hcensed operator positions of reactor This document is the final report for radclogical surveys con-operator (RO), sener reactor operator (SRO), and shift supervi- ducted in the community of Edgemont, South Dakota for the sor (SS). The extent to which college eng'neenng cumculum purpose of locating residual radcactive matenals from the urani-covers job-related academic knowledge was assessed. The ap- um processing industry. It contains a discussion of the histoncal proach used involved systematica!!y companng college engi- justificaton for the surveys, and a summary of activities dunng neenng programs to knowledge needed on the job by having the survey, from September 1980 through November 1984. The subject matter experts in the filed of general and nuclear engi- survey protocols are presented and discussed The results of neenng cumculum: (1) assess the coverage of specific academ- several studies of relevance to the surveys are also included. ic knowledge identified by a job anasysis as necessary for li. The results of the survey are presented in tabular form. censed operators in existing college engineenng degree pro-grams, and (2) make judgments concerning levels of formal en- NUREG/CR-4060: THE DC-1 AND DC-2 DEBRIS COOLABILITY gineenng educaten necessary for apphcation of knowledge on AND MELT DYNAMICS EXPERIMENTS. HITCHCOCK.J T.; the job, based on job samples from a job analysis of activities KELLY,J E. Sandia National Laboratones. July 1985. 163pp. under selected normat and emergency operating sequences. 8508090648 SAND 84-1367. 32101:142.
Main Citations and Abstracts 61 The DC expenment senes investigates the heatup and melt of NUREG/CR 4064: STRUCTURAL RESPONSE OF LARGE PENE. dry reactor core debns through nuclear heating of actual reactor TRATIONS AND CLOSURES FOR CONTAINMENT VESSELS matenals in order to obtain the thermal properties of dry debns, SUBJECTED TO LOADINGS BEYOND DESIGN BASIS. the nature of the transition from a debns bed to a molten pool, KULAK.R.F. Argonne National Laboratory.
~~Sanda Natonal and the thermal and kinetic behavior of molten pools. The pur. Laboratones. Apnl 1985. 109pp. 8505060514. ANL-84-41.~~
pose is to develop a data base in support of model develop. 30193.019. ment. The work is jointly sponsored by the USNRC, the PNC This report summanzes the analyses work performed by Ar-(Japaq, and EURATOM. This report provides a desenption of gonne Natonal Laboratory on three representative nuclear the two experiments in the DC senes and documents the con, power plant penetrations for severe accident loads beyond the figuration and the data. These tests invesagated dry debns beds design basa conditions. These include anafyses of an equip-(2 kg) composed of pure UO(2) and mixed UO(2) and stainless ment hatch for a steel containment. a BWR-Mark ll drywell steet Heat transfer charactenstics were studied at several head and a bellows connect on. The objectwes of the analyses steady state cond;tions below melt. The beds were then taken were to. identify the methodology required to simulate the re. into melt to observe the growth of a mo!!en pool en the UO(2) sponse of the penetratons and determine their leakage poten. bed and the agglomeration and migration of steel in a compos. tial under severe accident loads. This report provides the details ite bed. The peak measured temperature in the UO(2) bed was of the analytical methodology used and the results obtained above 3000 degrees centigrade. Approximately 50*. of the from the analyses. urania formed a molten pool, in the mixed UO(2) and steel bed. the peak measured temperature was 2600 degrees centigrade. NUREG/CR-4067:
~~SUMMARY~~
OF BARRIER DEGRADAT!ON With about 90*. of the steel molten, matenal migration occurred EVENTS AND SMALL ACCIDENTS IN U S COMMERCIAL NU-resulting in a significant increase in the gross bed thermal con- CLEAR POWER PLANTS. SAILOR,V.L; COLBERT.J J. Brook. ductwity. haven National Laboratory. March 1985. 66pp 8504020086. BNL-NUREG 51842. 29585135. NUREG/CR-4061: LEACHATE PLUME MIGRATION DOWNGRA- e expenece of M commscial nuclear power plants with DIENT FROM URANIUM TAILINGS DISPOSAL IN MiNE respect t small accidents and events involving the breach of STOPES. NELSON.R.W.; MCKEON.T.J.; CON 8ERE,W. Battette any i van us bames t radoactwe rnatenal release is + Memonal Institute, Pac fic Northwest Laboratones. Februa0/
* * " '#" 8"*8
9 " U' "#

1985. 82pp. 8504050284. PNL 5318. 29673 228. report is intended to provide background information for the A method previousfy developed at Pacific Northwest Labora- NRC staff eva!uation of the proposed NRC safety goals includ-tory has been simphfied and extended to better evaluate the en- ed are events that resulted in the breach of one or more bar-virontMntal consequences of below-water table disposal of ura- riers (fuel cladding fa!!ures, pnmary coolant leakage, compro-mise of containment integnty), or in unintentional release of ra-neum rr.ill tailings in mine stopes. The method desenbed smes analytical expressions for the velocity potential and examines ?oactwe matenals Also included are miscellaneous small aces-tents or failures not resutting in radioactive releases, but which numencally the convectue transport of tailings hquor and teach- had special safety implications. The 1979 TMI 2 accident is not ate through the aquifer and into a water supply well located inc!uded. The report does not attempt to evafuate the signifi-downgradient from the mine stope. The overall dependence of cance of the events as potential precursors of more severe ac-the teachate plume size and shape an the hydrologic param. cidents (such evaluations are the subject of vher stud +est eters and the tailings disposal geometry are presented in graph- Rough statistics are presented on the frequency of events de-ical form for use in prehminary assessments. The graphical re- fined above for the penod. 1974 1982. It is noted that none of suits are also used to set up worst-case scenanos for return of the events resulted in fatalities or injunes attnbutable to radio-the leachate constituents to the biosphere via the pumped logical causes water supply well. The interactwe computer models developed to evaluate such worst-case conditions are presented, d:s- NUREG/CR 4068:
~~SUMMARY~~
OF HISTORICAL EXPERIENCE cussed, and used to evaluate four typical situations. WITH RELEASES OF RADIOACTIVE MATERIALS FROH COM-MERCIAL NUCLEAR POWER PLANTS IN THE UNITED NUREG/CR-4062: EXTENCED STORAGE OF LOW-LEVEL RA- STATES. SAILOR.V L; COLBERT.J.J Brockhaven National DIOACT.VE WASTES Potential Problem Areas SISKIND.B.; Laboratory. March 1985 73pp. 8503280025. BNL NUREG. DOUGHERTY,D.R.; MACKENZIE,D.R. Brookhaven National 51843. 29548 007. Laboratory. December 1985.149pp.8601070490. BNL-NUREG. This report presents a summary of the histoncal empenence 51841. 04189:141. concerning releases of radioactwe matenals from U.S commer. If a state or state compact does not have adequate disposal cial nuclear power plants. The matenal was compiled specifical-capacity for low-level radioactwo waste (LLRW), then extended o pmvide background information for the Nuclear Regulatory storage of certarn LLRW may be necessary. Extended storage Commission (NRC) Staff Evaluation of the proposed NRC of LLRW is considered in order to determine for the Nuclear a% &als W Ws of avadaW m on rahactw unis-Regulatory Commission areas of concern and actions recom- sions are identifet reviewed and summanted The annual 50-mended to resolve these concems. The focus is on the prcpor- year population radiation dose commitments for the annular re-ties and performance of the waste form and waste container, gions between 2 and 80 km surrounding each plant resul ting Storage afternatives are considered in order to charactenze the from the radioactwe emrssions are summanzed for the period, likely storage environments for these wastes. The areas of con- 1975 1981. These doses are compared with the annual pr'pula-tion dose commitments from natural background radiation for cern are grouped into two categones: 1. Performance of the the same areas, and Eth the proposed NRC Societal Safety wtste form and/or container dunng storage, e g radiolytic gas Goal. The question of endependent venficatun of hcensee data generation, radiation enhanced degradat on of polymenc maten- on emissions is examined. als, and corrosion. 2. Effects of extended storage on the prop- . erties of the waste form and/or container that are important NUREG/CR-4069: ANALYSES OF SOILS FROM AN AREA AD-after storage (e g , radiation-induced embnttlement of high-den- JACENT TO THE' LOW LEVEL RADIOACTIVE WASTE DIS-sity polyethytene and the weakening of steel containers result. POSAL SITE AT SHEFFIELD,lLLINOIS. PICIULO.P L.: ing from corrosion A discussion is given of additional informa. SHEA.C E.; BARLETTA R E. Brookhaven Natonal Laboratory tion and actions required to address these concerns. March 1985. 54pp. 8503200236. BNL NUREG-51844. 29470 255.
62 Main Citations and Abstracts Soit samples and field resistwity data were collected from an accourmng for unusual counts in the tables were examined to uea adiacent to the Sheffield site. Specimens of Peona Loess. gain ensights from the events. Roxana Silt, Radnor Titi, sand from the Toulon member, Hulick Till, and shale from the Pennsylvania system were collected and NUREG/CR 4072: THE ESTIMATION OF ATMOSPHERIC DIS-analyzed. Resisittvities of the soils are all greater than 2500 PERSION AT NUCLEAR POWER PLANTS UTILIZING REAL ohm cm, indicating an environment which can be moderately TIME ANEMOMETER STATISTICS. Lt.W.W.; MERONEY,R.N. corroswe to steet. Measurements of soil pH range from 6 2 to Colorado State Univ., Ft. Collins, CO. January 1985. 236pp. 8.6. Determination of the tett acidity of the so:Is indicates an 8502010665. 28703.001. alkaline environment. The moisture content of the soils are rcp- Dispersion and turbulence measurements were conducted in . resentative of a wet site. The ion content of the soils show high a sir:ulated atmosphenc boundary layer. Field expenments and l l levels of calcium consistent with the calcareous nature of the ,,nd tunnel results for the behavior of lateral plume dispersion soils. Both the extractable and exchangeable co.,centrations of are compared to three semi-empencal expressions based on calcium, magnesium, potassium, and sodium in tne soils are re- Taylor's diffusion theory. Agreement between the field data and ported. The content of the following solubre anions is also laboratory measurements supports using wind tunnel results to gwen: carbonate, bicarbonate, sulfate, sulfide, and chlonde. simulate atmosphenc transport phenomena. Eulenan space-time correlations with streamwise separatnns were measured for all NUREG/CR-4070 V02: BlVALVE FCULING OF NUCLEAR three velocity components in the simulated boundary layer Re-POWER PLANT SERVICE. WATER SYSTEMS. Volume 2. Current suits were compared to prevous measurements which were Status Of Biofouling Surveillance And Control Techniques. DALING P.M.; JOHNSON.K.l Battelle Memor;al Institute. Pacific performed under different flow configurations. A universal shape Northwest Laboratones. March 1985. 68pp. 8503270015. PNL. of the Eulenan space-time correlaton seems to exist when pre-5300. 29541:211, sented in a normalized time coordinate. Turbulence measure. This report desenbes the current status of techniques for de- ments of fixed-point Eulenan velocity stat stics were employed tection and control of cooknga 3ter system fouhng by byalve to estimate the Lagrangian velocity stat!stics through the Bald-mollusks at nuclear power plants. The effectiveness of these win and Johnson approach. The approach was modified to ac-techniques is evaluated on the basis of informaton gathered count for the uniform shear stress effect in a homogenous tur-from a literature review and in interviews with nuclear power bulent flow field. The estimated Lagrangian integral time scale plant personnel. Bofouling detection techneques examined in agrees with estimates inferred from dispersion measurements this report include regular maintenance, in service inspection. ethin only a 20% error. Such agreement supports the method-and testing. Generally, these methods have been inadequate for ology of using real time anemometer statistics to predict the at-detecting biolouhng Recommendatens for improving biofouhng mosphenc turbulent dispersion near a nuclear reactor site. detection capabihties are presented Biofouhng prevention (or control) methods that are examined in this report includo intake NUREG/CR-4073: RESULTS OF THE SEMISCALE MOD-28 screen systems, thermal treatment, prevenhve maintenance. STEAM GENERATOR TUBE RUPTURE TEST SERIES. chemical treatment alternatives, and antifoulant coatings. Rec- LOOMIS.G G. EG&G, Inc. January 1985. 75pp. 8502060492. ommendat ons for improving befouhng control methods at oper- EGG 2363. 28745.278. ating nuclear power plants are presented. Additional techniques A senes of expenments was conducted in a scaled model of that could be implemented at future power plants or that require a pressunzed water reactor (Semiscale Mod-2B) to investigate further research are also desenbed. steam generator tube rupture system signature response and recovery 8echniques The tube rupture was assumed to occur NUREG/CR-4070 V03: BlVALVE FOULING OF NUCLEAR dunng normal full power operahon [15 6 MPa (2262 psia) POWER PLANT SERVICE WATER SYSTEMS Factors That May Intensity The Safety Consequences Of Biofouhng. System pressure; 37 K (67 degrees fahrenheit) core differential HENAGER.C.H.; DALING.P.M; JOHNSON.K.t. Bat'elle Memon- temperature). From the expenmental resufts, the charactenst'c at insbtute Pacific Northwest Laboratones. Apnl 1985. 61pp. system signature responsu for a wide range of number of 8504220375. PNL 5300. 29946:202. tubes ruptured and rupture locations have been examined. In This report desenbes the svety and economic consequences add 1 tion, recovery techniques requinng operator act ons were of bwafve foahng in raw water systems at nuclear power plants. examined. These recovery techniques included the use of pres. The report hste events that could cause a normal fouhng situa- sunzer auxikary spray and internal heaters, steam generator tion to become snore enacal and descnbes scenanos in which feed and steam, pomary feed and bleed, and safety injection. bwalve fouhng could cause unsafe or unwanted condit ons such The effectiveness of using these techniques for pnmary system as transients and shutdowns. Several fouhng events that have pressure and subcoohng control is discussed occurred at vanous nuclear plants are t; netty revewed, and rec-ommendations are made to a.d in the detection and control of NUREG/CR 4074: THE PERFORMANCE OF DEFECTED SPENT bivalve fouhng. LWR FUEL RODS IN INERT GAS AND DRY AIR STORAGE ATMOSPHERES OLSEN C S EG&G. Inc. January 1985.35pp NUREG/CR-4071: EXPLORATORY TREND AND PATTERN 0502220293. EGG-2364. 29073:105~ ANALYSIS FOR 1981 LICENSEE EVENT REPORT DATA. A testing program using eight commerc # pressunzed water HESTER.O.V.: GENTILLON,C.D. EG&G, Inc. Apol 1985. 21Spp. reactor and boiling water reactor spent fuel rods was conducted 8505280415. EGG-2362. 30601:072. to invest gate their long-term stabihty under a vanety of possible This report presents an overview of the 1981 Sequence dry storage conditions The objective of this project was to pro-Coding and Search System (SCSS) data base that contam nu, vide the Nucitar Regulatory Commission with informaton to clear power plant operational data denved from Licensee Event confirm or estabhsh dry spent fuel storage hcensing posibons Reports (LERs) submitted to the United States Nuclear Regula-for long term, towdemperature (< 250 degrees centigrade) tory Commission. Both overall event reporting and events relat. spent fuel rod behavior dunng dry storage and radioactive con-ed to specific components, subsystems, systems, and person-nel are discussed. At all of these levels of information, software tarunation ansing from spattation of cladding crud. The results ss used to generate count data for contingency tables. Conbn- of a nonriestructive examinaten of eight fuel rods, which includ-ed color closed-circuit television visual examinations, Color pho-gency table analysis is the main tool for the trend and pattern analysis. The tables pnmanly focus on faults associated with tography, dimercional measurements, and neutron radiography, vanous components and other items of tnterest across different are presented plants. The abstracts and other SCSS information on the LERs
Main Citations and Abstracts 63 NUREG/CR-4075: DESIGNING PROTECTIVE COVERS FOR NUREG/CR-4080: DETERMINATION OF THE AVAILABILITY OF URAN!UM MILL TAILINGS PILES. A Review. BEEDLOW,P A.; CORE EXIT THERMOCOUPLES DURING SEVERE ACCIDENT
PARKER,G.B. Battelle Memonal Institute. Pacific Northwest SITUATIONS. EDSON,J L EGaG Idaho, Inc. (subs. of EGaG, t Laboratones. May 1985. 29pp. 8505280093. PNL-5323. Inc.). September 1985. 47pp. 8510030430. EGG-2366.
l 30604:247. 32848 082. I This report reviews design consderations for protective This report presents the findings and recommendations of the covers for uranium mill tailings impoundments. The role of pro. Nuclear Power Plant instrumentaten Evaluation (NPPIE) pro-l tective covers in tailings containment systems is discussed. gram concerning signal validation methods to determine the on-Factors affecting the long-term stabikzation of tailings (erosion, line availability of core exit thermocouples dunng accident situa. biotic intrusion, and sod moisture) are summarized. Basic ele. tions. Methods of selecting appropnate signal validaten tech-ments to be considered in design of all uranium tarkngs covers niques are discussed and sources of error identified. This report are presented, and then quant;tative techniques for designing shows that through the use of these techniques the esistence sde-specific covers are reviewed. of high-temperature-caused errors may be detected as they occur. Specific recommendatens for applicaten of selected NUREG/CR-4076: CETERMINATION OF COMPLIANCE WITH signal validation techniques to core exit thermocouples and CRITERIA FOR FINAL TAILINGS DISPOSAL SITE RECLAMA. other measvement systems are made. TION. BEEDLOW,P.A.; CLINE,J.F.; FREEMAN,H D.; et al. Bat-title Memonal Institute Pacife Northwest Laboratones. June NUREG/CR-4081: ASSORPTION OF GASEOUS LODINE BY 1985. 48pp. 8507030712. PNL 5324. 31318.250. WATER DROPLETS. ALBERT,M.F. Oak Ridge National Labora-This report provides methods and procedures that can be ton. August 1985. 212pp. 8509110032. ORNL/TM-9488. used to venfy comphance with Environmental Protection Agency 32558 154 (EPA) engineenng standards for uranium mill tailings disposal sites. EPA standarda for radon emissons, long-te m isolation. and protecten of water quality are discussed. TaAngs isolation TNWmap4% m e mW e ME technologies are reviewed. Informaton the licensee needs to ciudes the odine hydrolysis reactons. The parameters of the provide for the regulating agency to determine compliance is model are spray drop size, initial concentraton of the gas and presented, as is the actual compliance entena. liquid phases, temperature, pressure buttered or unbuffered s ray solution, spray flow rate, containment diameter and drop NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL fall height. The results of the model were studied under many BEHAVIOR DURING STATION BLACKOUT. KITTMER.C.A.; values of these parameters. Plots of concentraton of sodine WENSEL,R.G.; RHODES,0.B.: et al. Atomic Energy of Canada, species in the drop versus time have been produced by varying Ltd. Apnl 1985. 93pp. 8506170667. EGG 2365. 30979.290. the initial gas phase concentration of molecular iodine over the A testing program designed to provide fundamental informa- rango of J x 10 W mesMw to 1 x M W) rnolesMa, a ten pertaining to the behaver of reactor coolant pump (RCP) buttered pH of 7 or 9, and a drop size of 1000 mcrons Results shaft seals dunng a postulated nuclear pcwor plant station e a c&npa to s avadaW kn W Gn-blackout has been completed. The test plan was developed by tainment Systems Expenments at Pacife Northwest Laboratory. EG8G Idaho personnel at the Idaho National Eng neenng Labo- The difference between the model predictons and the expen-rttory (INEL) and performed at the Chalk River Nuclear Labora- a a rany W m 2 m W W cWst age to7, Ontano, Canada, under auspices of the U S. Nuclear Re9u- * ## * "* litory Commisson (NRC). One seal assembly, utilizing boti hy- fy existing spray models. At begh concentrations of gaseous mo-drodynamic and hydrostate types of seals, was modeled and 1ecular edine, the new spray model is considered to be less ac-tested. Extrusion tests were conducted to determine if seat ma- curate than the previous models. At low concentratons, the 1:nals could withstand predicted temperatures and pressures. A new model predicts results that are closer to the expenmental tJper face seaf model was tested for seal stabihty under conds- data. Inclusen of the odine hydrolysis reactions is shown to be tions when leakino water flashes to steam across the seal faca important for determining the removal of molecular indine frnm Test informaten was then used as the basis for a station black- as pheu by water sprays for most conditions. out analysis. Test results indicate a potent 6al problem with an elastomer matenal used for 0-nngs by a pump vendor; that NL' REG /CR-4082 V01: DEGRADED PIPING PROGRAM - PHASE vendor is considenng a change in matenal specificaton. Test il Semiannual Report March 1984 - September 1984 WILKOWSKI.G.M.; AHMAD J.; BARNES C R : et af Battelle Me. results a'so indicate a need for further research on tne genenc issue of RCP seat integnty and its possible consideration for morial Institute, Columbus Laboratones. January 1985. 118pp. 8501280617. BMI-2120. 28574.202. designation as an unresolved safety issue. The objective of the Degraded Piping Program Phase 11is to develcp simple engineenng analyses to assess the fracture be-NUREG/CR-4079. ANALYTIC STUDIES PERTAINING TO STEAM GENERATOR TUBE RUPTURE ACCIDENTS. KASHlWA,B.A.; haver of nuclear peping. Such anatyses must give reahste ests-MJOLSNESS.R.C. Los Alamos Scientifc Laboratory. Apnl 1985- mates of actual fracture events. Hence this is an intensety inte-88pp. 8506060372. LA 10307.MS. 30773 271. grated program involving laboratory material property evaluaten. A study of thermal-hydraulc phenomena of possible steam analytical developments, and full-scale pipe fracture expen. generator tube rupture (SGTR) accidents leads to the conclu- ments to venfy the simple engineenng analyses. Both advance fracture mechan 4cs analyses (i e, J/T), and hmit-load analyses sens that (1) flashing will not occur upstream of the tube rup-ture, so that the flow will be resistance limited rather than will be assessed. Thss is a 3 year program which began in March,1984. Consequently, this first semaannual report de-choked, (2) there is considerable potential for discharging the scnbes work in progress rather than completed efforts. pnmary fluid en the form of micron-sized droplets, partcularty when the fluid discharges into a vapor cavity surrounding the NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE tube rupture, and (3) that the surrounding of the rupture site by ll. Semiannual Report. October 1984 - March 1985. witer rather than vapor may be a means for preventing the for. WILKOWSKI G M ; AHMAD.J.; BARNES,C A ; et al. Battelle Me. mation of micron-sized droplets. The presence or absence of monal Institute, Columbus Laboratones July 1985. 367pp meron-sized droplets rs considered to be a key issue for the 8508090632. BMI 2120. 32100.136. damage assessment of SGTR accidents because they are cur. The efforts in this report are broken into six work packages rently thought to be the most likely route for radioactive sodine related to pipe-fracture research efforts and six work packages to be released to the atmosphere. that are supporting research efforts. The pipe. fracture efforts in-
i 64 Main Citations and Abstracts volve only circumferential crack onentations. Twenty.six pipe ex. includes atmosphere models for steam / air thermodynamics, in-i tercell flows, condensation / evaporation on Sturctures and aero-j penments have been conducted to date, with all but two at 500F (288C). Approximately 35 additional pipe expenments sols, aerosol behavior, hydrogen burning, sodium / atmosphere f from past programs were also analyzed. Anatysis efforts include chemistry, sodium-spray fires, and sodium-pool fires. It also in-j hmit load and elastc-plastic fracture mechanics analyss. Elas- cfudes models for reactor cavity phenomena such as core / con-tc-plaste fracture-mechanics analytical efforts concentrate on crete interactons, coolant-pool boiling, and sodium / concrete J-integral estimaton schemes that can predict loads and dis- ir:teractions. Heat conducton in structures, fisson-product placements (predictive J-estimation schemes), rather thart those decay and transport, radioactive heating, and the thermal-hy-that can only be used to calculate the toughness (n-factor anal- drauhc and fisson-product decontaminaton aspects of engi-yms). Finite-element analyses are conducted in selected cases. neered safety features are also modeled. i Supporting research efforts involve geome.ry effects on J-R i curves, notch acuity effects, predicting J.R curves with large NUREG/CR 4086: TENSILE PROPERTIES OF IRRADIATED NU. , 4 amounts of crack growth from small specimens, development of CLEAR GRADE PRESSURE VESSEL WELDS FOR THE THIRD l HSST IRRADIATION SERIES. MCGOWAN.J.J. Oak Ridge Na-I a large comphant pipe test system, evaluabon of cracks in i welds, and procurement of cracked pipe removed from service. tional Laboratory. May 1985. 23pp. 8505160629 ORNL/TM-9477. 30439 288. ,' NUREG/CR-4083: ANALYSES OF SOILS FROM THE LOW. The Heavy Secten Steel Technology (HSST) Program con-j LEVEL RADIOACTIVE WASTE DISPOSAL SITES AT ducted a senes of expenments to investigate the effect of neu-BARNWELL,SC AND RICHLAND,WA. PICIULO.P.L; SHEA.C.E.; tron irradation on the fracture toughness of nuclear pressure BARLETTA.R.E. Brookhaven National Laboratory. March 1985. vessel matenals. Four welds of A 508 class 2 steel were exam < ] - 62pp.8503290285. BNL NUREG-51846. 29564.029 ined in this Third HSST Irradiaton Senes. The welds were fabri-To evaluate the performance of a buned waste form or waste cated according to "earty" (pre-1972) hghtwater reactor weld container, consideraton must include the interacton of the practice (i.e., copper coated electrodes). As part of this study,
; package with the bunal enveronment. This report presents the tensile properties were measured after arradiation to 2 to 10 m i results of physical and chemical measurements of soils from 10(22) neutrons /m(2) (E > 1 MeV) at temperatures between ,
i two currently operating commercial radioactue waste disposal 250 and 290 degrees centigrade. Strength properties of all four sites; one at Barnwell, SC, and the other near Richland, WA. welds increased with exposure to irradiation. Yield strength was Soil samples beheved to be representatwe of the soil that will more sensstive to irradiation than was ultimate strengtn. Tenssie i contact the buned waste forms were collected and anatyzed. ductility was not affected significantly by exposure to irradiation. Resistuity data given for soils from both sites endicate meidly
]
corrosiva environments. The soil acidity measurements show NUREG/CR-4087: MEASUREMENTS OF URANIUM MILL TAIL-the Bamwell site to have acide soil, whereas, tne Rechland site INGS CONSOLIDATION CHARACTERISTICS. FAYER,M.J. Bat- ' has soils ranging from acidc to near neutral in pH. The moisture telle Memonal Institute, Pacific Northwest Laboratones. Febru-content and the 6on content of the soils from each site are pre- ary 1985. 44pp. 85030f 0322. PNL 5339,29186.059. sented. The extractable ion content of the soils is given for the Expenments were conducted on uranium mdl taihngs from the i following ions: calcium, magnesium, potassium, sodium, carbon- taihngs pde in Grand Juncton, Colorado, to determine their con-ate, bicarbonate, sulfate, sulfide, and chlonde. Additionally, the sohdaten charactensbes. Three matenals (sand, sand /shmes exchangeable cabons were measured for the soils from the two mix, shmes) were loaded under saturated conditons to deter-r sites. mine their saturated consohdaten behavior. Dunng a separate expenment, samples of the shmes matenal were kept under a NUREG/CR-4084: DRY SPENT FUEL STORAGE TEST PLAN constant load whde the pore pressure was increased to deter. FOR DESTRUCTIVE FUEL ROD EXAMINATIONS. OLSEN.C.S. mine the partially saturated consohdation behavior. Results of EG8G, Inc. Apnl 1985. 41pp. 8504220369. EGG 2367. the saturated tests compared well with pubhshed data. Sand j 29946:164. consohdated the least, whde shmes consohdated the rnost. As ,i
. A testing program using eig'it commercial pressurized water each matenal consohdated, the measured hydrauhc conductNity L reactor and boshng water reactor spent fuel rods was conducted decreased in a knear fashon with respect to the void ratio. Par. ' < to investigate their long term stabdity under a vanety of possible baHy saturated expenments with the shmes indicated that there i dry storage conditons. The objectue of this report is to provide was httle consohdaten as the pore pressure was increased pro-I the Nuclear Regulatory Commsssion with informahon to confirm gressively above 7 kPa. The small amount of consohdation that i or establish dry spent fuel storage licensing positens for long, d'd occur was only a fraction of the amount of saturated con-l term, low-temperature R250 degrees centigrade) spent fuel sohdaton. Prehminary measurements between pore pressures rod behavior dunng dry storage and for radioactive contamina, ,
4 ton that might occur with spallation of cladding crud. Six of the of 0 and 7 kPa indcated that measurable consohdation could , i occur in this range of pore pressure, but only if there was no I eight commercial fuel rods wdl be destructuely examined. This load. report presents the test plan for the destructue examinatons. NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Com- NUREG/CR-4088: METHODS FOR ESTIMATING RADIOACTIVE puter Code for Severe Reactor Accident Containment Analysis. AND Toxic A!RBORNE SOURCE TERMS FOR URAN!UM BERGERON K.D.: CLAUSER,M.J.; HARRISON.B.D.; et al. MILLING OPERATIONS. HARTLEY,J.N; GLISSMEYER.J.A.; Sandia National Laboratones. July 1985. 354pp. 8508090656. HILL.O.F. Battelle Memonal Institute, Pacife Northwest Labora- 1 SAND 841204. 32099:142. tones. June 1985. 69pp. 8507080192. PNL 5338. 31393 099. i i The CONTAN 1.0 computer code is an integrated analysis Pacife Northwest Laboratory, under contract to the U.S. Nu- i t tool for the physical, chemcal, and radclogical conditions inside clear Regulatory Commisson (NRC), identfied and evaluated j a containment budding following the release of radcactNe ma- methods for eshmating radcactive and toxc partculate and i tenal from the pnmary system in a severe reactor accident. It gaseous airborne releases from uranium milhng operations. can also predct the source term to the environment. The pur- Such melhoris need to be standardized so that all uransum mdls pose of this User's Manual is to provide a basic understanding can provide adequate data for NRC evaluaten of potential envi-of the features and models in CONTAIN 1.0 so that users can ronmental impacts and of comphance with 10 CFR 20,40 CFR I prepare reasonable input and understand the output and its sig- 190. and the Natonal Environmental Polcy Act. The general nifcance for particular apphcatons. Besides input instructions, method for calculating source terms is to multiply together a j i the User's Manual also contains brief desenptons of the base normah2ed etnisson rate, contaminant content, emisson control features of the models. Both fight-water reactors and hquid- factor, and processing rate for each process being evaluated. j This report desenbes the sources of airborne releases (ore stor- )
metal reactors can be modeled with CONTAIN 1.0, The code

i I,
~~~fT*-w Y p immpag e emy .&=,.^s'fMe--pmm._ _re-m.-e-'-'m2--*'-9hm rMM---- .P w-* Wh+ ----TWnm--t-M1*mV1.= -. S- fm -#*-? *--r --S-.--T . . .-tm"'"~~
l Main Citations and Abstracts 65 age area, ore crushing and gnndmg, ore processing, yellowcake suits for threc EPR, one VAMAC, one BUNA N, one SILICONE, produchon, and taihngs impoundment) and the calculational pro- and one VITON matenal are also presented. cedures for estimating radcactue and toxc source terms. Ex-Emple calculations are provided. NUREG/CR 4092: ORNL CHARACTERl2ATION OF HEAVY SEC-TION STEEL TECHNOLOGY PROGRAM PLATES 01,02.AND i ' NUREG/CR 4089: EVALUATION OF FIELD. TESTED FUGITIVE 03. STELZMAN,W.J.; BERGGREN,R G.; JONES,T.N. Oak Ridge DUST CONTROL TECHNIOUES FOR URANIUM MILL TAIL. Natonal Laboratory. Apnl 1985.176pp. 8506060376. ORNL/
~~.INGS PILES. ELMORE,MA: HARTLEY,J.N. Battelle Memonal TM-9491. 30773:101.~~
[. Instituto. Pacife Northwest Laboratones. January 1985. 68pp. Charpy V-notch impact, tensile, and drop-weight data are pre-i 8502150694. PNL-5340 28960:305. sented for three 305-mm-thsck (12-in.) A 533 grade B class 1 l Seventeen chemical stabihzers, rated as the most promising steel plates. The effects of specimen size and orientaton were
~~i. of those tested in earher laboratory studies, were apphed to test examined as well as the variaten of properties between differ- i i~~
plots on a uranium mill tailings pile at the Amencan Nuclear ent plate locatons and depths. Some observatons based on Corporaten-Gas Hills Project mill site in central Wyoming. The data obtained from an instrumented Charpy testing machine are ' } durability of these matenals when exposed to actual site conds. also presented.
; tions was evaluated over time. In additon, eight commercially available windscreens were field tested. Test panels of the eight NUREG/CR-4093: SAFETY / SAFEGUARDS INTERACTIONS
, matenals were constructed at the Wyoming site to compare DURING SAFET%RELATED EMERGENCIES AT NUCLEAR ] their relative ressstance to weathenng. A second test was con- POWER REACTOR FACILITIES. MOUL.D.A.; PILGRIM,M KJ
ducted near Pacife Northwest Laboratory to evsluate the effec- SCHWElZER.RL; et al. Brookhaven National Laboratory. June tiveness of the windscreens at reducing wind velocity. Results 1985.229pp.8507020094. BNL-NUREG-51848. 31308.138.
of the field tests on the chemcal stabihrers and windscreens This report contains an analyss of the safety / safeguards r are presented in this report, along with effectiveness.versus- interactons that could occur dunng safety-related emergencies j cost informaton. Direct companson of these two dust control at licensed nuclear power reactors, and the extent to which methods is diffcutt due to the dependence of each on site-spe- these interactions are addressed m exishng or prcposed NRC '
, cific factors. However, simphfied model case studies were de- guidance. The safety / safeguards interaction dunng a seres of i veloped to assess the cost of chemmal stabihzaten versus postutated amergencies was systematically examined to identify I windscreen systems for a hypothetcal, inactNe tailings pile. any potent al performance defciencies or conficts between the Operations (safety) and Secunty (safeguards) organizatons.
NUREG/CR-4090: EVALUATION OF NUCLEAR FACILITY DE- This examinahon included the impacts of coordinaton with off. COMMISSIONING PROJECTS. Annual Summary Report sim emergecy msoow persenei. Dubes, respons bahties, op-1 Fiscal Year 1984. MILLER.R.L; BAUMANN.BL; DOERGE,D.H.- timal methods, and procedural actions inherent in these interac-j United Nuclear Corp. (subs. of UNC Resources, Inc.). January tons wem explM J 1985.124pp. 8502010524. 28702:207, This document summanzes work performed dunng the 1984 NUREG/CR 4094: FIELD EXPERIMENT DETERMINATIONS OF i fiscal year for the Nuclear Regulatory Commissen's Evaluaton DISTRIBUTION COEFFICtENTS OF ACTIN 1DE ELEMENTS IN i SULFATE LAKE ENVIRONMENTS. SIMPSON.H.J.; TRIER,R.M.; I
~~). of Nuclear Facihty Decommissoning Projects program. This HERCZEG,A.L; et al. Columbia Un#v., New York, NY. January q~~
report doscobes actual work performed dunng the reportog 1985. 72pp. 8501280372. 28572.001. j penod and work planned for the future, included as appendices The concentratons of a number of radcasotopes of some ele-to this report are drafts of the current difs from the TMI-2 re' covery efforts and Shippingport Atome Power Station decom- ments (Pu, U, Th, Pa. Ac, Ra, Bi, Po, Pb, Cs, Sr, and K) were i measured in a group of lakes that are dominated by SO(4)(2-) l missonin9-ion in their anonc compositon. Only Pu and the Th show pos- ! NUREG/CR-4091: THE EFFECT OF ALTERNATIVE AGING AND
* ""*"#** "
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~~ACCIDENT SIMULATIONS ON POLYMER PROPERTIES se aa I~~
BUSTARD.LD: CHENION.J.; CARLIN F; et al. Sandia Natenaj la ace waws W kn Lau"' >> l - Laboratones. June 1985.177PP. 85d7050391. SAND 84-2291' am at saMaW e msW b cam haw ap proximately the same radionuchde content as seawater. The i f uence of accident tradiaten, steam, and chemical aWs (anoxd sh a maM mcmase M h aM Ra. spray exposures on the behavior of twenty-three age-precond" This mdicates that these elements may be coupled with the
~~; tioned polymer sample sets (twenty-one different matenais) has redox cycle of Fe and Mn wnich under oxygenated conditions~~
~~been investgated. The test program vaned the following cond" effectively sequester Pu, Th. and Ra as Fe(.Mn) oxyhydroxides.'~~
; tions: 1. Accident simufations of irradiat on and thermodynamic Another possibehty for the enhanced (226)Ra in the deep water i (steam and chemical spray) conditions were performed both se- is by coprecipitaticn with CACO (3) in the surface water and sub-sequent dissoluten in the deep water thereby releasmg radium {
quentially and simultaneously; 2. Accident thermodynamic l (steam and chemical spray) exposures were performed both into the water. L [- with and without air present dunng the exposures; 3. Sequenbal NUREG/CR-4095: TEST SERIES 2. SEISMIC-FRAGILITY TESTS t accident irradiatons were performed both at 28 degrees cents- OF NATURALLY. AGED CLASS 1E EXiDE FHC 19 BATTERY
. grade and 70 degrees centigrade: 4. Age preconditioning was CELLS. BONZON.LL; HENTE.D B. Sandia Natonal Laborato-performed both sequentially and simultaneously; 5. SequentmJ nes.Apnl 1985.166pp. 8507020389. 31307;332.
~~aging irradiations were performed both at 27 degrees centigrade This report, the second in a test senes of an extensrve seis-~~
; and 70 degrees centigrade; and 6. Sequential aging exposures mic research program, covers the testing of 10-year old lead-were performed using two sequences: (1) thermal followed by calcium Exide FHC-19 cells from the Calvert Chtfs Nuclear i
~~irradsation and (2) irradiation followed by thormal. This report Power Staten operated by the Ba!t more Gas and Electnc Com- i~~
! presents both general trends apphcable to a masonty of the pany. The Exide cells were tested in two configuratens usmg a tzsted materiats as well as specife results for each polymer. tnaxial shake table: single. cell tests, both rigidfy and loosely l The data base consists of ultimate tensile properties at the mounted; and multcell (three-cell) tests, mounted in a typical i completion of the accident exposure for three XLPO and XLPE, battery rack. A total of six electncally active cells was used in five EPR and EPDM, two CSPE (HYPALON), one CPE, one the two different cell configurations. None of the six cells failed l VAMAC, one polydiallyiphtalate, and one PPS matenal. Report m the fast stage tests dunng the actual seismic test up to the
~~bend test results at completion of the accident exposures for 1.5 g ZPAs imposed. Subsequent discharge capacity tests l~~
[ two TEFZEL matenals and permanent set after compression re- showed, however, that only two of the cells could deliver the l 4-I f n - , . -?' _ , - - _ ,-- - -_---,------,,--n---------- .-c-n - .-,n,.-,v--
a l. Main Citations and Abstracts 66 accepted standard of 80% of their rated electncal capacity for 3 tively newer instrumental methods. Direct-spectrometre and iso-hours. When two of the same cells were exposed to the second topcally excited x-ray fluorescence instrumental analysis meth-stage, higher g-level tests, both Cells again provided instantane- ods were evaluated. Because of a redirection in funding, the ous uninterrupted power. Subsequent capacity tests showed evaluation was not completed in terms of identifying instrumen-both of tnese cells to have capacities well below the accepted tal interferences and fielo testing of the chosen methods. How-standard of 80% Four of the cells were disassembled for ex- ever, in hght of readily available technology, a preferred method smination and metallurgcal analyses. The examinaten showed for sampling and analysis of yellowcake from uranium mill ex-
; the active material on the positive plates was hard and cracked hausts is proposed. This method would sample the exhaust and that the positive bus bar material was corroded and bnttle, stacks continuously using a continuous, automatic, isokinetic stack sampler with depositen of the exhaust gas partcutates NUREG/CR-4096: TEST SERIES 3. SEISMIC FRAGILITY TESTS onto filter paper. The deposited partculates would then be ana-OF NATURALLY-AGED CLASS 1E C&D LCU-13 BATTERY ly7ed by x-ray fluorescence using (57)Co as an excitation CELLS. BONZON.LL; HENTE.D.B. Sandia National Laborato- source. It is also recommended that a paper-tape sampler that i '
ries. Apnl 1985.170pp. 8507020413. SAND 84-2629. 31309:159. houses an isotopc excitaten source and detector be interfaced This report, the third in a test senes of an extensive seismk to a continuous stack sampler. This system would require eval-i ' rcsearch program, covers the testing of 10-year old lead-calci- uat.on and field testing aher development. um C&D LCU-13 cells from the North Anna Nuclear Power Sta-tion operated by the Virginia Electre and Power Company. The NUREG/CR-4101: ASSAY OF LONG-LIVED RADIONUCLIDES IN 4 C&D cells.were tested in twa configuratens using a tnaxial LOW-LEVEL WASTES FRCM POWER REACTORS. CLINE.JE.; i shake table: single-cell tests, both ngidly and loosely mounted- NOYCE,J R.; COE,LJ.; et al. Science Applications internatonal and multicell (three-cell) tests, mounted in a typecal battery rack. Corp. (formerfy Science Appleations, Inc.). Apnl 1985. 615pp.
~~j. A total of seven electncally active cells was used in the two dif. 8505100034.30271.001.~~
; ferent cell configurations. None of the seven cells failed in the The 10 CFR Part 61 waste classificaton system includes sev-4 first stage tests dunng the actual seismc test up to the 1.5 g eral nuclides which are diffcult to assay without expensive ra-1 ZPAS imposed. Subsequent discharge capacity tests showed o,ochemcal methods. In order for waste generators to classify that while these cells suffered some loss of discharge capacity, wastes practcally, NRC Staff has recommended the use of cor-all cells could deliver the accepted standard of 80% of their relaton factors to scale the diffcult to-measure nuchdes with '
rated electncal capacity for 3 hours. When two of the same nuchdes whch can be measured more easily (i.e., gamma emit-4 cells were exposed to the second stage, higher g-level tests- ters such as (60)Co or (137)Cs). In this study, Science Appica- . both cells again provided instantaneous uninterrupted power. tons international Corporation (SAIC) performed complete ra-
! Subsequent capacity tests showed both of these cells to have diochemeal assays for all the 10 CFR Part 61 waste classifica-l capacities well below the accepted standard of 80% Four of ten nuchdes on over 100 samples. These data, along with the cells were disassembled for examination and metal!urgical almost 800 other samples in the SAIC data base, were used to analyses. The examination showed that all plates and separa- assrss the vahdity of correlation factors suggested for use for tors were in very good condaten. nuclear power plant wastes. Specife genere correlaten factors are recomniended with other approaches to correlate nuchdes ! NUREG/CR-4097: TEST SERIES 4: SEISMIC-FRAG'LITY TESTS for whch genere scaling factors are not defensible.
OF NATURALLY AGED EXIDE EMP-13 BATTERY CELLS. BONZON.LL; HENTE.D.B. Sandia National Laboratories. Apnl NUREG/CR-4103: USES OF HUMAN RELIABILITY ANALYSIS 1985.119pp. 8507020430. SAND 84-2630. 31309 321. PROBABILISTIC RISK ASSESSMENT RESULTS TO RESOLVE This report, the fourth in a test senes of an extensive seismc PERSONNEL PERFORMANCE ISSUES THAT COULD AFFECT research program, covers the testing of 27 year old lead antg SAFETY. O'BRIEN,JN; SPETTELL.C.M. Brookhaven National
! mony Exide EMP 13 cells from the recently decommessened Laboratory. October 1985.100pp. 8601020811. BNL-NUREG-i Shippingport Atomic Power Staten. The Exide cells were tested 51849. 34118:120.
! in two configurations using a (nxial shake table: single-cell tests, This report is the first in a series which documents research ngidfy mounted; and multeell (five-cell) tests, mounted in a typg aimed at improving the usefulness of Probabiliste Risk Assess-j cal battery rack. A total of nine electrically active cells was used ment (PRA) results in addressing human nsk issues. This first
in the two different cell configurations. None of the nine ceffs report desenbes the results of an assessment of how well cur-I failed dunng the actual seismc tests when a range of ZPAs up rently available PRA data addresses human nsk issues of cur.
to 1.5 g was imposed. Subsequent discharge capacity tests of rent concern to NRC. Findings indcate that PRA data could be I five of the cells showed, however, that none of the cells could far more useful in addressing human nsk issues with modifca- ! deliver the accepted standard of 80% of their rated electncal ton of the development process and documentaten structure of i capacity for 3 hours. In fact, rene of the 5 ceffs could dehver PRAs. In addition, information from non-PRA sources could be
~~' more than 33% capacity. Two of the seismically tested cells integrated with PRA data to address many other issues.~~
j and one untested, low capacity well were disassembled for ex-j amination and metallurgical analyses. The inspection showed NUREG/CR-4104: MAINTENANCE PERSONNEL PERFORM-the cells to be in poor conditen. The negative plates in the vi- ANCE SIMULATION (MAPPS) MODEL Field Evaluation /Valida-e cinity of the bus connections were extremely weak, the positive ton. SIEGEL,A.I.; WOLF,JJ.; BARTTER,W.D.; et at Oak Ridge buses were corroded and bnttle, negative and positive active National Laboratory. August 1985. 86pp. 8512120145. ORNL/ material utilization was extremely uneven, and corros'on prod- TM-9503. 33870:175. ucts littered the cells. This report discusses the results of efforts focused upon the evaluaten of the practicality, acceptabit ty, usefulness and vahd-l NUREG/CR-4100: EVALUATION OF ' INSTRUMENTAL METH-ODS FOR THE MEASUREMENT OF YELLOWCAKE EMIS-ity of the Maintenance Personnel Performance Simulaten ,' SiONS. LEPEL,E.A.; THOMAS V.W. Battelle Memonal institute, (MAPPS) model. Subsequent to the completion of model devef-opment efforts, MAPPS was subjected to a number of evalua- ' Pacife Northwest Laboratones. February 1985. 35pp. 8503200121, PNL-5350. 29470:306. tons that included tab;e-top analyses, case study analyses, l An evaluaten of current. sampling and analysis methods used analyses involving the correlaton of MAPPS output to the con- ' for morntonng yellowcake emissons from uranium mill exhausts sensus of subsect matter experts, and analyses involving the was performed by Pacrfc Northwest Laboratory. The represent- correlaten of MAPPS output to directly observed empincal data I atives of once per quarter sampling was felt to be questonable. gathered in the field. Results of these evaluaten erforts are re-A more represerMtive sample would be obtained by a continu- ported in this NUREG/CR. Overall results indcate that no j unduly burdensome practicality issues were identified. IN adde i cus samphng system. The analysis could be performed by rela-t i I l
Main Citations and Abstracts 67 tion, identified user groups ( the U.S. Nuclear Regulatory Com- modif,ed versions of Page's ' cot ~td power-one procedures. We mssion, nuclear power plant maintenance supervisors, and ar- used simulatcd data from a MJJ in a conversion /fabncation chatect and engineenng firms) rated the model as having rela- process that tool into account process vanations, matenals tively high acceptability and found the model to be very useful holdup, and measurement uncertainties. Compansons were for a number of specific problems relating to user group inter- made over a 60-day accounting penod under different loss sce-ests. Predictive validity was also shown to be in good agree- nanos. Some important findings include. (1) No single procedure ment with both the consensus of subject matter experts and is best for all diversion scenanos. (2) Power-one procedures are with the empincal data observed in the field. The positwo re- best for protracted losses that occur early in the accounting suits of the accomplished evaluation efforts indicate that the pened and Page's test is best for late loss occurrence. (3) If modelis ready for widespread apphcation. holdup process vanations are not included in the Inventory Dif. NUREG/CR-4105: AN ASSESSMENT OF THERMAL GRADIENT ference model but are present in the process, then assuming 1 TUBE RESULTS FROM THE HI SERIES OF FISSION PROD- steady-state conditions, flase-alarm probabil. ties can double. UCT RELEASE TESTS. NORWOOD,K.S Oak Ridge National Laboratory. May 1985. 64pp. 8505230531. ORNL/TM-9506. NUREG/CR-4108: DEVELOPMENT OF MC&A ALARM RESOLU-30549:192. TION PROCEDURES. SMITH,8 W. Battelle Memonal Institute, A tnermal gradient tube was used to analyze fission product Pacific Northwest Laboratones October 1985. 53pp vapors released from fuel heated in the HI test senes. Complete 8510250531. PNL 5154. 33225.154. deposition profiles were obtained for Cs, I, Ag and Sb. The The NRC has proposed reform of the matenal control and ac-cesium profiles were complex and probably were dominated by counting (MC&A) requirements for facshties authonzed to pos. Cs-S-O compounds tormed by release of sulfur from furnace sess and use formula quantities of strategic special nuclear ma. ceramics. The iodine profiles were simple, indicating that more tenal (SSNM). The purpose of the reform is to strengthen than 99.5*. of the released iodine behaved as a single nonvota* MC&A capabibtles by requinng more timely detection of possible tile species, probably Cst Mass transfer coefficients for this SSNM losses and by providing for more rapid and conclusive species onto platinum were estimated to be 1.9 to 5.8 cm/s- resolution of discrepancres This report provides guidanco for l Siver was probably released in elemental form, condensed to developing *a set of procedures to resolve alarms from the pro-an aerosol, and captured by filters. Antimony was released as posed near-real time loss detection system. An alarm resolution the element and reacted rapidly with platinum (or gold) as it de- program distinguishes between a system error and an actual posited. Antimony profiles were calculated a pnon with some loss of nuclear matenal. An alarm resolution program consists success. A method was developed for isolating tellunum plati- of procedures to ider'tify causes of alarms, entena for accepting num and mixed fission products in a form suitable for neutron resolution of alarms,'and a program to monitor resoluton i effec-activation analysis. The platinum samples were completely dis- tweness Development of alarm resolution procedures consists solved in acid (HC1/HNO(3), and the tellunum was precipitated of identify'ng potential causes of afarms, ordenng the general on selenium camer by reduction. Finally, tellunum was loaded eiements of resolution, and determining entena for accepting onto Dowex 1X-4 ion-exchange resin for actNaton and analysis. resolution. A monitonng program is performed to ensure con-Teilunum recovery was 88%, and the theoretical sensitrvity was sistent an1 acceptable apphcat,on of the procedures. 3 ng. NUREG/CR 4109: TRAC-PF1 ANALYSES OF POTENTIAL PREb-NUREG/CR-4106: PRESSURIZED-THERMAL-SHOCK TEST OF SURIZED-THERMAL SHOCK TRANSIENTS AT CALVERT 3-IN. THICK PRESSURE VESSELS PTSE 1. Investigation Of CLIFFS / UNIT 1. A Combustion Engineenng PWR Warm Prestressing And Upper Shelf Arrest. BRYAN.R H.; SPRIGGS.G D.; KOENIG.J E.; SMITH,R C. Los Alamos Scientific BASS.B.R.; BOLT,S.E.; et af. Oak Ridge National Laboratory Laboratom Apnl 1985. 355pp. 8504220382. LA-10321-MS. Apnl 1985. 288pp. 8506060815. ORNL-6135. 30770.024. 29953.331. The first pressunzed-thermal-shock test of a 148-mm-thick Los Alamos National Laboratory participated en a program to steel pressure vessel with a 1 m-long flaw was performed to in- assess the nsk of a pressunzed thermal shock (PTS) to the re-vestigate fracture behavior of a vessel under conditions relevant actor vessel dunng a postulated overcochng transient in a pres-to a flawed nuclear reactor pressure vessel during an overcool- sunzed water reactor (PWR) We provided the thermal-hydrauhc ing accident. The objectues were to observe crack arrest and analyses of three general accident categones- steamhne stabiity l on the ductile upper sheif and effects of warm pres- breaks, runaway-feedwater transients, and small-break loss-of-tressing on crack initiation. Three coordinated pressure and coolant accidents. These postulated accidents included multiple thermal transients were imposed on the vessel, which was pre- operator and equipment failures Results were provided to Oak heated to 290 degrees centigrade. Two episodes of crack prop- Ridge Nat onal Laboratory (ORNL) who plan to determine the agation and arrest occurred. The thermal transients were in- probabihty of vessel failure and accident occurrence for an duced by coolant at -29 degrees centigrade to 15 degrees cen- overall assessment of PTS nsk. Our study was performed for a tigrade. Pressure transients were as high as 94.4 MPa. The ex- Combustion Engineenng PWR, CaNert Cistfs/ Unit 1 using the penmental obl ectues were attained The inhibiting effects of Transient Reactor Analysis Code (TRAC.PF1). We found the re-warm prestressing were definitely demonstrated. Crack p'opa, Sults of the analyses to be very sensstNe to the initial conditions gation was nearly pure cleavage, and arrest at 30K above the of the plant. If the plant was initially at hot zero power (com-onset of the Charpy upper shelf was expenenced in a positwe pared to full power), the decay heat was mech less, which X(I) gradient and with K(I) 300 MPa square root m. Fracture me- made it possible for the same accident initiator to produce sig-chanics analysis of bnttle fracture based on small-specimen nificantly lower downcomer temperatures. However, routine op-toughness measurements was reasonably accurate. Flaw eval- erator actions may reduce the consequences of any of these uation by procedures of the ASME Boiler and Pressure Vessel simulated accidents <f the prescnbed pressure-temperature rela. Code conservatuety predicted vessel failure, which did not tionships are followed cccur. NUREG/CR-4110: REPOSITORY SITE DATA REPORT FOR UN-NUREG/CR-4107: SEQUENTIAL TEST PROCEDURES FOR DE- SATURATED TUFF, YUCCA MOUNTAIN, NEVADA- TIEN.P.L; TECTING PROTRACTED MATERIALS LOSSES. SIEGEL.M.D.; UPDEGRAFF,C D ; et al. Sandia National Labora-GOLDMAN.A.S. Los Alamos Scientific Laboratory. July 1985. tones. November 1985. 400pp 8512100278. SAND 84-2668. 51pp. 8508150033. LA-10319-MS. 32196 285. 33835 001, Sequential tests are required for detecting protracted (tnckle) The U S. Department of Energy is currently considenng the losses of strategic special nuclear matenals from a single mate- thick sequences of unsaturated, fractured tuff at Yucca Moun-rials control unit (MCU). We compared apphcable tests including tain, on the southwestern boundary of the Nevada Test Site, as
68 Main Citations and Abstracts a possible candidate host rock for a nuclear-waste repository. additional information that would be useful in assessing cable Yucca Mountain is in one of the most and areas in the United system performance. States. The site is within the south-central part of the Great Basin section of the Basin and Range physiographic province NUREG/CR-4112 V02: INVESTIGATION OF CABLE AND CABLE and is located near a number of silicic calderas of Tertiary age. SYSTEM FIRE TEST PARAMETERS. Task D Ferestop Test Although locakzed zones of seismic actvity are common Method.
~~Underwnters Laboratory, Inc. January 1985, 7tpp.~~
throughout the province, and faults are present at Yucca Moun. 8502t30568. US 75-2. 28922:351. ton, the site itself is basically aseismic. No data are available An experimental investigation was conducted to provide data on the compositen of ground water in the unsaturated zone at concoming the effects that changes in pressure daterential, fire , Yucca Mountain. It has been suggested that the composition is exposure and sample construction have on firestop perform. ance when exposed to a standard fire test. Fifty-one fire test l bounded by the compositions of water from wells USW-H-3, UE25p-1, J-13, and snow or rain. There are relatively few data expenments were conducted using pressure differentials be-available from Yucca Mountain on the moisture content and tween -12 to +120 Pa, different sample constructions and two saturation, hydraulic conductivity, and charactenstic curves of fire exposure conditions. Findings were that small changes in ! the unsaturated zone. The available literature on therroome. pressure different al did not have a significant effect on firestop chanical properties of tuff does not always distinguish between matenals that dd not have cracks or through openings to allow data from the saturated zone and data from the unsaturated passage of gases dunng fire exposure. If the matenals allowed zone. Geochemical, hydrologic, and thermortechanical data passage for gases through cracks or holes, such as those left available on the unsaturated tuffs of Yucca Mountain are tabu- open after pulling a cable, changing the pressure different al af-lited in this report. Where the data are very sparse, they have fected the firestop performance. Also, it was demonstrated that been supplemented by data from the saturated zone or from changing the size of the opening, size, location and type of the areas other than Yucca Mountain, penetrating items installed through the opening; and seventy of NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER
**" "" # * * * '
E' EFFICIENCIES WHEN CHALLENGED WITH THERMAL. AND NUREG/CR-4114: VALENCE EFFECTS ON THE SORPTION OF AIR-JET-GENERATED DI-2-ETHYLHEXYL SEBECATE.DI NUCLlDES ON ROCKS AND MINERALS.II. MEYER.R E.;
ETHYLHEXYL PHTHALATE AND SODIUM CHLORIDE. ARNOLD,W.D.; CASE.F.f. Oak Rdge National Laboratory. May KERSCHNER,H.F.; ETTINGER.H.J.; DEFIELD.J.D.; et al. Los 1985. 53pp. 8505210576 ORNL-6137. 30521:378 Alamos Scientfic Laboratory. Apnl 1985. 62pp. 8504300121. Estimahon of the rates of migraten of nuclides from nuclear LA-9985-MS. 30070 220. waste repositones requires knowledge of the interaction of Resperators fitted with high. efficiency particulate (HEPA) car- these nucides with the components of the geological forma-tndge filters are designed to remove dust, fumes, mists, and air- bons in the path of the migraton. These interactions will be de-bome particulate radionuchdes. If these filters are to be reused, pendent upon the valence state and speciation of the nucide. a Quality Assurance (OA) program must be established to Expenments designed to measure interaction of multivalent nu-ensure that filter efficiency remains greater than 99.97 per cent. clides and minerats must therefore include stady of the specia-The standard method for performirv) OA teshng is to challenge tion of the nucides. An electrochemical method of valence the filter with a thermally generated aerosol of 0.3-m-diam da state control and solvent ntraction analyses of the valence ethylhexyf phthalate (DEHP). Because of potential toxicological states were used to study a number of reactions of interest to and other problems associated with the use of monodisperse HLW repositones. These include the reducten of Np(V) and DEHP, an investigation to study measured filter efficiencies on Tc(Vil) by crushed basalt and other minerals. For the reduction an HEPA respirator filt0r population, using several recommend- of Np(V) by basalt, the expenments indicate that the sorphon of ed replacement aerosols, has been conducted Aerosols com- basalt increases with pH and that most of the Np is reduced to pared in this study were thermally generated di-2 ethytexyl se- Np(IV) which is very difficult to remove from the basalt even af becate (DEHS), thermally generated DEHP, air-jet-generated oxygenated tracer free soluton is added to the solution. For the DEHS, and air-jet-generated salt (NaC1). The study also fo- expenments with Tc(VII), the results are considerably more cused on determining compabbility for parallel use of aerosols complicated. Expenments were inibated to determine the solu-generated for respirator-fit testing for use in OA filter testin9- bility of Tc(IV) or des. The results of these expenments are Results indicate that a polydisperse air-jet-generated aermol of used to assess some of the techniques and metnods currently DEHS can substitute for thermalry DEHP as a method of provd- used in safety analyses of proposed HLW repositones. ing OA testing of HEPA respirator filters and that equipment used in the study designed for respirator quanbtative-fit tesbng NUREG/CR-4115: INTERNATIONAL STANDARD PROBLEM 13 can easil y be modified to perform this function. (LOFT EXPERIMENT L2-5) Final Companson Report.
~~' nc. January 1985. 229pp. 8502210258.~~
NUREG/CR-4112 V01: INVESTIGATION OF CABLE AND CABLE G 69 2 58 SYSTEM FIRE TEST PARAMETERS. Task A lEEE Flame Test. LOFT Expenment L2-5 was designated Intemabonal Standard Underwnters Laboratory, inc January 30, 1985. 105pp. Problem 13 by the Organization for Economic Cooperabon and ome mpans ns en measumnents fan Ex. f a e' est tr nst te' of Electncal and Electronics penment L2-5 were made with calculabons from 11 internationat Engineers (IEEE) Standard 383 was invesbgated. The investiga- participants using frve different computer codes. LOFT Experi-bon was to develop possible modificatons in test equipment ment L2-5 simulated a double ended guillotine cold leg rupture End test procedure that would increase the repeatabildy of re- of a pnmary coolant loop of a large pressunzed water reactor, suits and provide additional information useful in assessing coupled with a loss of offsite pwer. Cable system performance in response to a real fire. Several fire expenments were conducted varying different test parameters. NUREG/CR-4116: NUFEGO-NP.A DIGITAL COMPUTER CODE The expenmental data were analyzed and modifications of both FOR THE LINEAR STABILITY ANALYSIS OF BOILING WATER trst equipment and test procedure were developed to increase NUCLEAR REACTORS. PE NG.S.J.; PODOWSKI.M Z,; repeatability. These modifications we'e: An enclosure for the LAHEY,R.T. Rensselaer Polytechnic Institute Troy. NY. August sample, defining cable damage; cable fastening and the cable 1985. 437pp. 8509060217. 32502.001. tray to be used, establishing tolerances for exhaust of the en- The phenomena of nuclear-coupled density-wave oscillatons closure; starting temperature of the ambient air cable sample; are of considerable importance in boiling water nuclear reactor locabon of the bumer and the flow rates of fuel and air into the (8WR) stabihty analysis. A state-of-the-art knear frequency bumer. Suggested also, was to report the maximum flame domain digital computer code, NUFREO-NP, has been devel-height versus time and the rate of heat released versus time as oped for either forced or natural circulation BWR stability anafy. i
Main Citations and Abstracts 69 sis. The NUFREO-NP code can be excited by many external tives of this program are discussed as well as the test matnx, perturbations, including system pressure perturbation. It is test facilities, and test procedures. based on one dimensional dnft-flux thermal hydraulics, and allows for subcooled botting, arbitrary nonuiuform axial and NUREG/CR-4120: MATHEMATICAL MODELING OF ULTIMATE radial power shapes, distnbuted local losses (e g., spacers), HEAT SINK COOLING PONDS. POLICASTRO,A.J ; point or multi-dimensional neutron kinetics, and detailed fuel WASTAG.M; DUNN.W E.; et al. Argonne National Laboratorf. element dynamics. It has been compared with both out-of-core March 1985. 276pp. 8504050372. ANL/ES-143 29671:054. and in-core data, and good agreement has been found. A general treatment of ultim0 te heat sank (UHS) ccohng pond thermal performance is proposed through the apphcaten of a NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR three-dimensioaal gnd model. Validation of the model has been SURFACE MINES OF SOUTHWESTERN INDIANA. AULT.C.H.; shown through compansons of predict ons with data from a field HARPER,D.; SMITH C Fi.; et al. Indiana Geological Survey. Jan- and laboratory pond The advantage of the model hos an its abil-uary 1985. 39pp. 8502070553. 28795:316. ity to determine the detailed character of the flow field whether '
~ This project was directed towards the charactenzation of: (1) it be one, two, or three dimensional. Existing rnodels require a the known large faults in southern Indiana, ie., the Georgetown pron knowledge of the character and dimensionahty of the flow Fault in Floyd County and the newly named Crandall Fault in field in such ponds. Applicaten of the model to a prototype Harnson County; and (2) the small scale fractures endemic to UHS pond revealed that the balance of physical mechanisms in-southwestern Indiana. The Georgetown and Crandall Faults are vulved in the thermal hydraulics of these ponds is quite different normal faults that have a maximum verhcal displacement of than for ponds used in normal cooling The smali, heavily-about 65 feet. They are post-Valmeyeran and pre-Pleistocene in loaded, irregularly-shaped nature of the UHS pond should, in age and are probably the result of hinge-line deformaton be- many cases, lead to a vertically mixed pond with only a one-tween the subsiding lilinois Basin and the Cincinnati Arch. In dimensional (longitudinal) vanaten in pond temperature.
Contrast, abundant small-scale faults and joints are related to regional compressive lithosphenc stress or to sedimentologic NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND processes that operated penecontemporaneously with deposi. FLOW RATE ON FATIGUE CRACK GROWTH RATES IN LWR' tion of the rocks in which they are found Structures related to ENVIRONMENTS. CULLEN.W.H. Matenals Engineenng Associ-regional stress include sma!!-scale thrust faufts with displace- ates, Inc, KEMPPAINEN,M.; HANN!NEN.H ; et al. Finland. Govt. ments of a few inches to a few feet and joints that are wide. of. February 1985. 49pp. 8503040021. MEA 2053 29190.223. spread in mines and outcrops in rocks of Mississippian and Fatigue crack growth rate tests, at a load ratio of 0 2, have Pennsytvanian age. The jointing and most of the small-scale been conducted on steels of low, med.um and high sulfur con-thrust faulting indicate that southern Indiana is affected by the tents (0.004%,0 013% and 0 025%) in PWR water at both low Midcontinent Stress Province in the northern part of the study and high flow rates. Crack growth rates show no dependence area and by another stress field in the southern part. An east. on flow rate, but are strongfy dependent on sulfur content, with west boundary can be defined between the two stress fields. a large proporton of environmental assistance for the highest sulfur contents. Tests of low and high sulfur content steels at a NUREG/CR-4118: MONITORING METHODS FOR DETERMINA- load ratio of 0 7 show relatively httle environmental assistance TION COMPLIANCE WITH DECOMMISSIONING CLEANUP in either case. The fractography of these specimens shows the CR"ERIA AT URANIUM RECOVERY SITES. DENHAM.D H.; usual bnttle appearance for environmentally-assisted fatigue RATHBUN,LA.; BARNES.M G ; et al. Battelle Memonal Insti- crack growth. In addition, the opposing fracture surfaces match tute. Pacific Northwest Laboratones. June 1985. 31pp. perfectly, indicating that httle or no dissoluton of the metal 8507030713. PNL-5361,31314:036. matnx has occurred, and there is very little plastic flow associat-Decommissioning of a uranium processing site requires radio- ed with the fatigue cracking process. The X-ray photoelectron logical surveys to: 1) identify buildings, equipment, and open emission examinaton of the fracture surface oxides shows that land areas that require cleanup; 2) venfy that cleanup oper- FeS and FeS(2) coexist in the oxide layer, suggesting that the atons have been successfuf; and 3) provide a record of the ra- conditions within the crack enclave involved near. neutral pH diological condition of the site following cleanup This report de- and catf. odic potentials. scnbes the instruments, measurements, quahty assurance, and statistical procedures that can be used to perform pre and post. NUREG/CR 4122: A FORTRAN 77 PROGRAM AND USER'S cleanup surveys. The procedures descnbed includo: 1) gamma. GUIDE FOR THE CALCULATION OF PARTIAL CORRELATION radiaton exposure-rate measurements using micro-R meters,2) AND STANDARDlZED REGRESSION COEFFICIENTS. beta-gamma measurements using Geiger Mueller tubes 3) wipe IMAN.R.L.; SHORTENCARlER, JOHNSON.J D. Sandia Natonal tests for surface contamination, and 4) soil analyses for (226)Ra Laboratories. August 1985. 54pp. 8508210437. SAND 85 0044. and other (238)U daughters. Locations hkely to have (226)Ra 32333.270. concentrations that exceed standards can be identified by This document is for users of a computer program developed gamma-radiation exposure rate measurements. Samples of soil by the authors at Sandia Natonal Laboratones The computer or other matenal from location showing elevated exposure rates program is designed to be used in conjunction with sensitivity can then be analyzed for (226)Ra to determine the boundanes analyses of complex computer models. In particular, this pro-of areas that exceed standards. Beta-gamma measurements gram is most useful in analynng input-output relationships when and wipe sample analyses can be used to determine whether the input has been selected using the Latin hypercube sampling uranium concentrations exceed standards for either fixed or re- program developed at Sandia (Iman and Shortencaner,1984) movable contamination. The present computer program calculates the partial correlation coefficients and/or the standardized regression coefficients NUREG/CR-4119: INTEGRITY OF CONTAINMENT PENETRA- from the multivanate input to, and output from, a computer TlONS UNDER SEVERE ACCIDENT CONDITIONS FY84 model. These coefficients can be calculated on either the ongi-ANNUAL REPORT. SUBRAMANIAN.C. Sandia National Labora- nal observations or on the ranks of the onginal observations. tones August 1985. 37pp. 8508210429. SAND 85-0016. The coefficients provide alternative measures of the relative 32334:167. contnbution (importance) of each of the vanous inputs to the This document is an annual report for FY84 on the NRC- observed output vanations. Relatonships between the coeffi-funded program titled "Integnty of Containment Penetrations cients and differences in their interpretations are identified. If Under Severe Accident Conditions." The purpose of this pro- the computer-model output has an associated time or spatial gram is to evaluate the behavior of seals and gaskets and history then the computer program will generate a graph of the major fixed and operable penetrations. The scope and objec- coefficients over time or space for each input-vanable, output
~~. .- . - - _-. -- - - _ _ - ~ - - . - -- -_ --~~
70 Main Citations and Abstracts venable combinaten of interest, thus indcating the importance A complete technical basis for implementation of the 3-D fast of each input over time or space. The computer program is numencs in TRACB04 is presented. The 3-D numercs is a gen. cra4zaten of the predictor / corrector method previously devel-
~~. user-fnendly and wntten in FORTRAN 77 to facihtate portabihty.~~
~~oped for the 1 D components in TRACB. NUREG/CR-4123: SEISM!C FRAGlLITY OF REINFORCED CON-~~
! CRETE STRUCTURES AND COMPONENTS FOR APPLICA. NUREG/CR 4127 V02: BWR FULL INTEGRAL SIMULATION TION TO NUCLEAR FACILITIES GERGELY,P Lawrence Liver. PROGRAM TRAC-BWR MODEL 7 TEST (FIST) ; Na na ory. March 1985.107pp. 8503280022. DEVELOPMENT. Volume 2.Models. CHU.K.H.; ANDERSEN.J G ;
~~CHEUNG,Y.K.; et at General Electnc Co. November 1985,~~
. The fadure and fragility analyses of reinforced concrete struc- pp. W2m6L M W98L MM286 tures and elements in nuclear reactor facihties within the Seis- TRAC-BWR (Transient Reactor Analysis Code) is a computer me Safety Margins Research Program (SSMRP) at the Law- code for best estimate analysis of the thermal hydraulc condi- . rence Livermore National Laboratory are evaluated. Uncertain-
' ties in material modehng, behavior of low shear walls, and sees-tions to a Boihng Water Reactor system, in this report, the de-velopment of new models and the implementation of the bal-mic nsk assessment for nonhnear response receive special at-ance of plant models leading to the creacton of the TRACB04 i tention. Problems with ductihty-based spectral deamphfcaten version of the code, is desenbed. The new models include an and predicton of the stiffness of reenforced concrete walls at improved model for boron transport whch accounts for none low stress levels are examined. It is recommended to use rela- form mixing and stratification, and a model for the interfactal trvsly low damping values in connection with ductihty-based re- heat transfer at two phase levels. The balance of plant models ! sponse reductens. The study of static nonhnear force-deflection curves is advocated for better nonlinear dynamic response pre. (turbine, containment and heat exchanger) developed at INEL I dictions. Several details of the seismic nsk analysis of the Zion were evafuated, adapted, and implemented into TRAC 804. j plant are also evaluated. NUREG/CR-4127 V03: BWR FULL INTEGRAL SIMULATION NUREG/CR-4124: NDE OF STAINLESS STEEL AND ON-LINE TEST (FIST) PROGRAM TRAC-BWR MODEL j LEAK MONITORING OF LWRS. Annual Report, October 1983 - DEVELOPMENT. Volume 3. Development Assessment For Plant Ses,tember 1984. KUPPERMAN,D.S.; CLAYTOR,T.N.; Appicaten. CHEUNG,Y.K.; ANDERSEN.1G.; CHU.K.H.; et al. PRINE.D.W. Argonne National Laboratory. Apnl 1985. 39pp. Generat Electre Co. November 1985. 90pp. 8512050464. EPRI 8505060509. ANL 85-5. 30190:287. NP-3987. 33769196. ] l This progress report summanzes work performed by the Ar. The TRAC 804 computer code has twen developed under the gonne Natonal Laboratory and GARD, Inc. (Division of Chank model development tasks in the FIST Program. This report de-t>stiain Mfg. Corp.) as subcontractor on NDE of stainless steel scnbes two developmental assessment calculations performed s;nd on-line monitoring of LWRs dunng the twelve months from on BWR plants with TRACB04. A BWR/2 Design Basis Acci-October 1983 to September 1984- dent (DBA) including the containment response and a BWR/4 DBA with Low Pressure Coolant Injecten (LPCI) water injected NUREG/CR-4125 V01: GUIDELINES AND WORKBOOK FOR AS-SESSMENT OF ORGANIZATION AND ADMINISTRATION OF into the lower plenum were calculated and results of these UTILITIES SEEKING OPERATING LICENSE FOR A NUCLEAR cases were documented. These cases serve to test some of POWER PLANT. Volume 1:Guidehnes For Utility Organization the new features of the TRAC 804 (air field. containment model, i 1 And Administration Plan. THURBER.J.A.; OLSON,J ; " water packing" fries and faster numercs in the three dimen. OSBORN,R.N.; et al. Battelle Human Affairs Research Centers. sonal vessel component) and to demonstrate that the code has August 1985. 41pp. 8508230324. PNL 5374. 32358.304. been assembled properly. They also provide best estimate l This report is a partial response to the requirements of item LOCA results for the two plant types. [ I.B.1.1 of the "NRC Acton Plan Developed as a Result of the
~~. TMI-2 Accident" NUREG-0660, and is designed to serve as a NUREG/CR-4128: BWR FULL INTEGRAL SIMULATION TEST 4~~
bisis for replacing the earher NUREG-0731, "Guidehnes for (FIST) PHASE 11 TEST RESULTS AND TRAC-BWR MODEL OUALIFICATION. SUTHERLAND.W A.; ALAMG;R.M.; Utahty Management Structure and Techncal Resources." The Guidehnes are intended to provide guidance to the user in pre. FINDLA(,J.A.; et al. General Electnc Co. October 1985.200pp. j penng a wntten plan for a proposed nuclear organization and 8511220149. EPRI NP-3988. 33604:051. administraton. The purpose of the Workbook is to guide the A full height BWR system simulator built under the Full inte. NRC reviewer through a systematic revew and assessment of a gral Simulaton Test (FIST) program is used to investigate i I proposed organizat on and administraten. It is the NRC's inten- system responses for vanous transients. Tne test program con-4 ' ton to incorporate these Guidehnes and Workbook into a future sists of two test phases. This report provides a summary, dis-1 revision of the Standard Review Plan (SRP), NUREG-0800. cussion, highhghts, and conclusons of the FIST Phase 11 tests. However, at this time the report is being pubhshed so that the The Phase I tests are reported in NUREG/CR 3711. EPRI NP. 7' mitenal may be used on a voluntary basis by industry to sys- 3602, GEAP-30496. Eight matnx tests were conducted 6n the tzmatically prepara or evaluate their organizaten or administra. FIST Phase 1. These tests investigated the large break, small ton plans. Use of the report by the NRC would not occur unt:1 break and steamhne break LOCA's. as well as natural circula-s.fter it has been incorporated in the SRP. tion and power transients. There are nine tests in Phase ll of the FIST program. They include the following LOCA tests-NUREG/CR-4125 V02: GUIDELINES AND WORKBOOK FOR AS- BWR/6 W kne break, BWR/6 iMennediate size recirculaton i' SESSMENT OF ORGANIZATION AND ADMINISTRATION OF break, and a BWR/4 large break. Steady state natural circula-UTILITIES SEEKING OPERATING LICENSE FOR A NUCLEAR ton Msts with feedwater makeup performed at high and low POWER PLANT. Volume 2. Workbook For Assessment Of Orga- pressure, and at high pressure with HPCS makeup, are includ. nizaton And Management. THURBER J A.; OLSON J.; ed. Simulaton of a transient without rod insertson, and with con. 4 OSBORN.R N.; et al. Battelle Human Affairs Research Centers. tr lied depressunz2 ton, was performed Also included is a simu-J August 1985. 92pp. 8508260008. PNL 5374. 32363 272. lation of the Peach Bottom turbine tnp test. The final two tests I See NUREG/CR-4125,V01 abstract. simulated a failure to maintain water level dunng a postulated 1 l NUREG/CR-4127 V01: BWR FULL INTEGRAL SIMULATION accident. A FIST program obl ective is to assess the TRAC code TEST (FIST) PROGRAM TRAC-BWR MODEL by compansons with test data. Two post tes' predictons made DEVELOPMENT. Volume 1.Numencal Methods. HECK,C.L; with TRACB04 are compared with Phase il test daia in this , i ANDFRSEN.J.G. General Electnc Co. November 1985. 66pp. report. These are for the BWR/6 LPCI kne break LOCA, and j i l 8512050366. EPRI NP 3987,33769.032. the Peach Bottom turbine tnp test simulation. p r
Main Citations and Abstracts 71 NUREG/CR-4130: ICEDF:A CODE FOR AEROSOL PARTICLE This document provides specifications for a model/methodol. ( CAPTURE IN ICE COMP /.R TMENTS. OWCZARSKI.P.C.; ogy and approach that could be employed in determining post-SCHRECK,R.I.; WINEGARDNER,W. Battelle Memonal Institute, closure repository environmental parameters relevant to high. Pacific Northwest Laboratories. September 1985. 93pp. level waste package performance for the Basalt Waste Isolation 8510040377. PNL-5379. 32857:295. Project (BWIP) Guidance is provided on (1) the identity of the This report desenbes the technical bases and use of comput- relevant repository environmental parameters (groundwater er code ICEDF. ICDEF was developed to serve as a tool for charactenstics, temperature, radiation, and pressure), (2) the calculating particle retenten in pressunzed water reactor (PWR) models/mothodologies employed to determine the parameters, ice compartments dunng severe accidents. This report also and (3) the input data base for the model/ methodologies. Sup-serves ss a complete user's guide for the most recent stand- porting studies included are (t) an analysis of potential waste alone version of ICDEF. A complete code desenpton, code op- package failure modes leading to identification of the relevant erating mstructions, code listing, examples of the use of ICEDF, repository environmental parameters. (2) an evaluat!on of the and a summary of a parameter sensstrvity study support the use credible range of the repository environmental parameters for of code ICEDF. the BWIP situation, and (3) a summary review of existing modals/ methodologies currently employed in determining re-NUREG/CR-4131: INVESTIGATION OF ALTERNATIVE MEANS pository environmental parameters relevant to waste package TO ACCOMPLISH THE GOALS OF BIENNIAL ION CHAMBER performance. Call 8 RATION. CAMERON,J.R.; DEWERD.L.A.; GOETSCH.S.J.; et al. Wisconsin. Urw. of, Madison, WI. May 1985. 38pp. NUREG/CR-4136: SMOKE.A Data Reduction Package For Analy. 8506030101. 30707:220. . sis Of Combustion Expenments. RATZEL.A C.; KEMPKA.S N ; The research desenbod in this report was performed to inves- SHEPHERD.J E.; et al Sandia Natonal Laboratones. Septem-tigate the feasibility of a mailed dosimetry system as an afterna- ber 1985.131pp 8511190556 SAND 83-2657. 33533.235. tive method of achieving the goals of the present U S. Nuclear A suite of computer codes, collectively referred to as SMOKE, Regulatory Commission requirement that enization chambers has been developed to expedite the processing and analysis of used for ca!.bration of ccbalt-60 teletherapy units be cahbrated data obtained from transient combustion tests. Instrumentation every two years. Both thermoluminescent dosameters (TLD's) signals which can be processed using SMOKE include pressure and a diode detector unit was used in this study. A total of 20 sensors, gas and wall thermocouples. and different types of ca. hospitals in the states of Illinois. Iowa and Wisconsin participat- lonmetry such as Sandia-developed thin-film gauges, capace-
, ed in a program sn which this dosametry package was sent to tance-type slug calonmeters, and commercial Gardon-type heat ; each institution on three separate occasons. The physicist, phy- flux gauges. This package has been used to anatyze data from sician or chief technologist was asked to deliver 1.50 Gray (150 combustion expenments conducted at the Sandia National Lab-rads) to the device, assuming the device was equivalent in radi- oratones FITS facility and in hydrogen dewar at the Nevada aton adsorption charactenstics to human tissue. A treatment Test Site In this report, we discuss the theory and models used field size of 10cm by 10cm was chosen and the institution was in the computer codes compnsing SMOKE. Details of the data requested to use their clinical source-to-surface distance. The files. signal preparation, and processing procedures for execut-accuracy of the beam localizaton as indicated by the coenci- ing SMOKE are provided. Sample data files and representative dence of the fight field with the radiation field was measured as results using SMOKE are included.
weit. The entenon for accuracy of dose delivery was plus minus 3.0mm. Only two hospitals dunng tne course of the study had NUREG/CR 4137: PRETEST PREDICTIONS FOR THE RE-both a disagreement of more than 3mm between the lignt field SPONSE OF A 1.8 SCALE STEEL LWR CONTAINMENT and the radiation field. It is recommended that such a mailed BUILDING MODEL TO STATIC OVERPRESSUR12ATION dosimetry package be considered as an afternative to the CLAUSS.D.B. Sandia National Laboratones. Juty 1985. 53pp present NRC requirement for biennial cakbraton of sonization 8508120552. SAND 85-0175. 32147.107. chambers used to calibrate cobatt-60 teletherapy sources. The analyses used to predict the behaver of a 1.8-scale model of a steel LWR containment building to static overpres-NUREG/CR 4133: NUCLEAR POWER SAFETY REPORTING sunzation are desenbed and results are presented Finite strain, SYSTEM IMPLEMENTATION AND OPERATIONAL SPECIFICA- targe displacement, and nonlinear matenal properties were ac-TlONS. NEWTON.R.D.; IMS.J R.; FINt AYSON.F.C. Aerospace counted for using finite element methods. Three-dimensional Corp. November 1985.117pp 8512050321. ATR-85(5818)2. models were needed to analyze the penotrat:ons, which includ-33776.037. ed operable equipment hatches, personnellock representations. This report is the last in a senes invest gating the feasibility of and a constrained pipe. It was concluded that the scale model adapting a voluntary, anonymous, non-punitive, third-party man- would fail due to leakage caused by large deformations of the aged reporting concept in a U.S. commercial nuclear endustry/ equipment hatch sleeves. regulatory environment. Such a system is intended for use in identifying and quantifying, in an uninhibited manner, the factors NUREG/CR-4138: DATA ANALYSES FOR NEVADA TEST SITE that contnbute to the occurrence of significant safety incidents (NTS) PREMIXED COMBUSTION TESTS RATZELA C. Sandia which ehcit eitner positrve or negative responses from humans National Laboratones July 1985.179pp. 8507250130. SAND 85-in U.S. nuclear power plants. This report specifies the elements 0135. 31794.341. cf a Nuclear Power Safety Reporting System (NPSRS), along This report provides results from an in-depth analysis of with operating procedures and forms to be used for accepting, twenty.cne premixed large-scale combustion expenments spon-integrating and processing reports submitted to the system. sored by the NRC and EPRI and conducted by EG&G at the Also included is a taxonomy for collating and stonng reports re. Nevada Test Site (NTS) These expenments were performed in ceived from a vanety of sources addressing myriad safety-relat. a 2048 m(3) sphencal vessel (hydrogen dewar) with mixtures of ed topics. A companion NUREG/CR-4132 presents the results hidrogen, steam, and air ignated by glow plugs or heated resist-of a limited evaluation of the technical specif cations contained ance coils. Hydrogen concentrations ranged from 5 to 13*. (by in this report. volume) and steam concentrations from 4 to 40% Several tests also incorporated spray systems and/or fans which enhanced NUREG/CR-4134: REPOSITORY ENVIRONMENTAL PARAM- the combustion rate and significantly altered the postcombus-ETERS RELEVANT TO ASSESSING THE PERFORMANCE Or tion gas coohng Data providod by EPRI from instrumentation H!GH-LEVEL WASTE PACKAGES. CLAIRBORNE,H.C.; designed to charactenze the thermal environment in the dewar CROFF,A.G; GRlESS.J C; et at Oak Ridge Natonal Laborato- dunng and following combustion have been evaluated The data ry. May 1985.130pp. 8506130358. ORNL/TM-9522. 30867:350. reductmo package SMOKE has been used to process data from
72 Main Citations and Abstracts thin-film gauges, commercial heat flux gauges. capacitance ca- ANSI test sectior'. Dynamic flow tests were pedormed over the lonmeters, gas and wall thermocouples, and pressure sensors. range of a design basis accident. Leak integnty testing was also Local measurements of the heat transfer are provided from the performed and extended into severe accident conditions. Anaty-catonmetry, and global averages are inferred from the pressure. sis of the test results produced a technical basis to assess in-Instrumentation " goodness for each test is assessed based on dustry purge and vent valve closing toroue extrapolation meth-the raw data and on comparisons of local and global results. odology and quantified the influence of worst case piping geom-Graphical and tabular results are provided for each test, and etry on valve torque response. It was also determined that trends observed from the results are reported some valve designs will leak in single isolation when exposed to design basis and severe accident environments. NUREG/CR-4139: THE MAILED SURVEY:A TECHNIOUE FOR OBTAINING FEEDBACK FROM OPERATIONS PERSONNEL. NUREG/CR-4143: REVIEW AND EVALUATION OF THE MILL-MCGUIRE.M.V.; WALSH M E4 MORISSEAU D.S ; et si. Batteile STONE UNIT 3 PROBABILISTIC SAFETY STUDY. Containment Human Affairs Research Centers. May 1985. 87pp. Failure Modes. Radiological Source-Terms And Offsite Conse-8505100041. PNL 5381. 30270:125. quences. KHATIB-rah 8AR, PRATT,W.; LUDEWIG.H ; et al. This report descnbes a mailed survey of operations personnel Brookhaven National Laboratory. September 1985. 75pp. at a sample of commercial nuclear power plants. The survey 8510020257. BNL-NUREG-51907. 32829.198. was condur ted for the U S. Nuclear Regulatory Commission A technical review and eva8uaton of the Millstone Unit 3 (NRC) as .. art of the Operator Feedback Project. The survey Probabilistic Safety Study has been performed it was deter- , sought to Collect informaton on topics of concern to the NRC mined that: (1) long. term damage indices (latent fatal. ties, I and to assess the feasibility of a mailed survey on an informa' person-rem, etc.) are dominated by late failure of the contain-tion collect.on mechanism. Participants in the survey were 520 ment. (2) short-term damage indices (earty fatalitres, etc.) are personnel at 26 nuclear power plants representing all five NRC dominated by bypass sequences for internally initiated events, regiors. The individual participants completed and retumed by while severe seismic sequences can also contnbute significantly mail a ten-page questionnaire. This contained questions about to earty damage indices These overall estimates of severe ac-operations crew practices, including work and shift schedules. cident nsk are extremely low compared with other societal operations shift crew staffing. the shift technecal arfv sor poss- sources of nsk. Furthermore, the nsks for Millstone-3 are com-tion, respondents' own backgrounds, the questionnaire, and parable to nsks from other nuclear plants at high populaton other information-collection techniques. Results of the survey sites. Seismically induced accidents dominate the sever acci-offer some insight on operations crew practices at the plants dont nsks at Millstone-3 Potential mitigative features were partacipating in the survey. Survey results also suggest that the shown not to be cost-effective for internal events. Value-impact mailed survey is an information-collection technique that can be analyses for seismic events showed that a manually actuated used effectrvely to obtain feedback for the NRC from operatens containment spray system might be cost-effective. personnel. NUREG/CR-4140: DOM;NANT ACCIDENT SEQUENCES IN NUREG/CR-4144: IMPORTANCE RANKING BASED ON AGING OCONEE 1 PRESSURIZED WATER REACTOR. DEARING,J F.; CONSIDERATIONS OF COMPONENTS INCLUCED IN PROB-ABILISTIC RISK ASSESSMENTS DAVIS,T C.; SHAFAGHl.A.; HENNINGER,R.J.; NASSERSHARIF.B ; et al. Los Alamos Sci-KURTH.R.; et al. Battello Memonal institute. Pacific Northwest entific Laboratory. Apnf 1985.112pp. 8506240647. LA 1035t- Laboratones. Apnl 1985. 69pp. 8504220341. PNL 5389. MS. 31149 230. A set of dominant accident sequences in the Oconee-1 prog, 29946.001. This report presents a method for focusing additional re-sunzed water reactor was selected using probabilistic risk anafy-sis methods. Because some accident scenanos were similar, a search on aging phenomena that affects nuclear power plant subset of four accident sequences was selected to be analyzed components. Specificany, the method ranks components using with the Transient Reactor Analysis Code (TRAC) to further our a nsk aging sensitmty measure that descr:bes the change in risk due to changes in component failure rate. Desenbing the insights into similar types of accidents. The sequences selected aging phenomena and the resulting time-dependent component were loss-of feedwater, small-small break loss-of-coolant, loss-failure rate changes is beyond the scope of this study, The ap-of-feedwater-enitiated transient without scram, and interfacing placattons use average components unavailability equations cur-systems loss-of-coolant accidents The normal plant response renUy employed in PRAs to calculate the nsk aging sensitivity. A and the impact of equipment availability and potential operator more exact calcdaten is possible by using unavailability equa-actons were also examined. Strategies were developed for op- tions that include the time-dependent charactenstics of compo-erator actons not covered in existing emergency operator ac. nent faifure rates; however, these time-dependent characteris-tions not covered in existing emergency operator guidelines and tics are not well-known. The risk aging sensitivity measure pre-were tested using TRAC simulations to evafuate their effective- sented here is. therefore, segregated from these time-depend-ness in preventing core uncovery and rnanntaining core cooling. ent effects and addresses only the time-independent portion of NUREG/CR-4141: CONTAINMENT PURGE AND VENT VALVE aging phenomena The results identify the component types TEST PROGRAM FINAL REPORT. STEELE,R.; WATKINS.J C- that show the most potential for nsk change due to aging phe-EGaG Idaho, Inc. (subs. of EG&G, Inc ) September 1985.79pp. nomena. Future research on the time-dependent portion of 8511910188. EGG-2374. 33304:164. aging phenomena for these component types is needed to com-This report presents the results of the containment purge and pletely descnbe the nsk empact due to component aging vent valve test program, conducted under the sponsorship of the United States Nuclear Regulatory Commission. Office of Nu- NUREG/CR-4145: EARTHOUAKE RECURRENCE INTERVALS clear Regulatory Research. The test program investigated valve AT NUCLEAR POWER PLANTS: ANALYSIS AND RANKING. functionahty and leak integnty. Three nuclear designed butterfly HILEMAN J A,; KNOPOFF,L.; MANN.N R.; et al. Earth Technol-valves typical of those used in domestic nuclear power plant ogy Co. March 1985.142pp. 8503200110. 29471.001. containment purge and vent applications were tested. For a Five methods for estimating earthquake recurrence were companson of response, two valves of the same size with dif- ranked The methods represent those used, or proposed, in nu-fenng internal designs were tested. For extrapolaton inseghts, a clear power plant studies through 1982 and include Log Linear larger s. zed valve was also tested. The valve expenments were Poisson, Extreme Value SemoMarkov, Bayesian, and Uniform performed with varcus piping configurations and vatve disc ori- Hazard Method. Ranking focused on recurrence estimates for entations to the flow to simulate various installation options in earthquake sources, excluding attenuation and site response. flGld applicatens. As a standard for companng the effects of the Scores were asssgned to each method for 12 cntena such as installaton optons, testing was a!so performed in a standard accuracy, use of geologic data, and subjective judgment Cnte.
Main Citations and Abstracts 73 ria scores were weighted by their importance and summed Dif- tations, are evaluated. Finally, fa, lure analyses of two selected ferent sconng and weighting schemes were used to identify any reinforced and prestressed concrete containments are per. sensitmties. To and in sconng statistical cntena, methods were formed and results are compared with those presented in the tested on synthetic earthquake catalogs with known statistics, hterature, and natural catalogs were tested against theoretical magnitude distnbutions. The uniform Hazard Method scored high because, NUREG/CR-4150: EPICOR-il RESIN DEGRADATION RESULTS in pnncipal, expert judgement draws upon all seismologic knowl. FROM FIRST RESIN SAMPLES OF PF 8 AND PF-20. edge. The Bayesian Method scored low uecause data require- MCCONNELL.J.W.; SANDERS R D EG&G Idaho, Inc. (subs. of ments are severe for practical cases. The other methods were EG&G, Inc.). July 1985. 52pp. 8508150017-. EGG 2376. intermediate. These observations seem insensitive to scorer, 32190:267. sconng approach, or weighting scheme. The semeMarkov The 28 March 1979 accident at Three Male Island Unit 2 re-Method scores were sensit ve to the weighting scheme. leased approximately 560.000 gallons of contaminated water to NUREG/CR-4146: SIMULATION OF AN EPRI-NEVADA TEST the Auxihary and Fuel Handkng Buildings, The water was decon-SITE (NTS) HYDROGEN BURN TEST AT THE CENTRAL RE. ama usmg a ninea ah n system called EPICOR il de-CEIVER TEST FACILITY, DANDINI,V.J.; ARAGON J.J. Sandia ve ciped by Epicor,inc. The Low-Level Waste Data Base Devel-n s ner s gaton Rosct funded by Natonal Laboratones. October 1985. 77pp. 8510280470. SAND 85-0205. 33228 002 the U.S Nuclear Regulatory Commission, is studying the cheme in order to augment results obtained from the hydrogen burn Ia cal cems d N synmeN en exchange msins tests performed by the Electnc Power Research Institute (EPRI) 'n sewal pr us. w is %ng dow by at the Nevada Test Site (NTS), a senes of tests was conducted a aN Na@nal Engin*nng LapaW at the Sandia National Laboratones Central Receiver Test Facil-ity (CRTF). The CRTF tests simutated a 13 volume-percent burn phng f a n exchange resns from EPICOR-il prefilters PF 8 and from the EPRI-NTS series. Dunng the tests, the responses of n ma amne sa kun teus p several specimens of nuclear power fant safety-related equip- nun a . In ins t ne egra-ment were monitored when subjected to the simulated hydro- radaMn h + gen burn. The specimens were pressure transmitters, solenoid mm u abo are cWaw e valves, and single and multconductor electnc cables. All were 9* #" " 9 # " " # nuclear servce quahfied. The simulation was conducted with e am and without steam in the vicinity of the test specimens Pnor to exposure, metalkc specimens were preheateo to temperature NUREG/CR-4151: INTEGRATION OF EMERGENCY ACTION corresponding to the precombuston environment in the EPRI- LEVELS WITH COMBUSTION ENGINEERING EMERGENCY NTS test vessel. OPERATING PROCEDURES.9y Use Of Combuston Engineer-ing Owners Group Emergency Operating Procedure Technical NUREG/CR-4147: THE EFFCCT OF ENVIRONVENTAL STRESS Guidelines. FALETTI.D W ; JAMISON.J D. Battelle Memonal in. ON SYLGARD 70 SILICONE ELASTOMER. BUCKALEW.W H.; statute, Pacific Northwest Laboratones September 1985.137pp. WYANT.F J. Sandia Natonal Laboratones. May 1985. 92pp. 8509250160. PNL 5392. 32753 295. 8506240313. SAND 85-0209, 31157.001. Pacife Northwest Laboratory, under contract to the U S Nu-Dow Corning Sylgard 170 Sihcone Elastomer has been inves. clear Regulatory Commisson developed a method for integrat-bgated to charactanze its response to accelerated thermal ing Emergency Action Levels (EALs) with plant-specific Emer. ag 7 radiation exposure, and its behavior under apphed com- gency Operating Procedures (EOPs) using the Combuston En-pressive forces. Sylgard 170 response to accelerated thermat 9'neenng Owner's Group Emergency Operatn.g Procedure aging suggests the matenal properties are not particularly age Technical Guidehnes (CEOG EOPTGs) Using the Combust:on dependent. Radiaton exposures, however, produce significant, Engineenng Owner's Group Technical Guidehnes document, a monotonic changes in both elongation and hardness with in- set of emergency class definitens and entena were developed creasing absorbed radiaton dose. Elastomer response to an ap. based on the status of the three main f ssen product barriers phed compressive force was strong!y dependent on environ, (fuel cladding, pnmary coolant system and containment) The ment temperature and degree of matenal confinement. Van. EOPTGs were then annotated to point out where, in a symp-ations in temperature produced large changes in compressive tom /functon based EOP patterned after the EOPTG. the in-forces apphed to confined samples. Attempts to mitigate force ferred plant condition is such that a specife EAL may have fluctuatons by means of pressure rehef paths resulted in total been exceeded. After the EOPTGs have been annotated, the loss of the apphed compressive force Thus, seal apphcations proposed method was tested by applying it to the reactor acce employing this elastomer in Class 1E equipment required to dent sequences that were shown in the reactor safety study to function dunng or fohowing an accident should consider the po- d" mate accident nsk to determine if an EAL set linked to the tential loss of compressive force from long-term aging and po- EOP annotations would produce timely and accurate classifica-tential LOCA-temperature transient conditions. ton of the nsk-dominant sequences Add, tonal annotations and additions to the EOPTGs were developed and the revised anno. NUREG/CR 4149: ULTIMATE PRESSURE CAPACITY OF REIN- tations were shown to produce timely and accurate event clas-FORCED AND PRESTRESSED CONCRETE CONTAINMENT. safications for all the accident sequences SHARMA.S.; WANG,Y.K.; REICH.M. Brookhaven Natonal Labo-ratory. May 1985. 95pp. 8506130467. BNL NUREG-51857. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION 30901:328. KATO.W Y.; WEINSTOCK.E.V.; CAREW.J F ; et at. Brookhaven This report presents the results of an in-depth evaluaton of Nabonal Laboratory February 1985 327pp 8502260121 BNL. Current modehng techniques and analysis procedures .or deter- NUREG 51858. 29109 001. mining ultimate pressure capacity of reinforced and prestressed Brookhaven National Laboratory has conducted a study on concrete containments. The matenat models used for descnbing the need and feassbehty of an independent organization to inves-the nonhnear matenal behaver of concrete and steel ato re- tigate significant safety events for the Office for Analysis and viewed en detail. Special attention is focused on post-cracking Evaluation of Operatonal Data. USNRC. This is being carned behavior of concrete which controls one of the containment fail- out in response to a Congressional request to the NRC for such ure modes, i e , the shear failure. Varcus finite element ideahza- a study. The study consists of three parts the need for an inde-tons utih2ed for containment analys#s are reviewed. The effects pendent organization to investigate significant safety events, al-of major assumptons pertaining to containment geometry, ba- ternative organizations to conduct investigations, and legislative semat restraint, finite element mesh, rebar locatens and onen- requirements The determination of need was investigated by
74 Main Citations and Abstracts reviewing current NRC investigation practices, companng avia- and accidental stressors are determined, and their effect on tion and nuclear industry practices, and interviewing a spectrum promoting aging degradation is assessed. Failure modes, mech-of representatives from the nuclear industry, the regulatory anisms, and their effect on promoting agog degradation is as-agency, and the public sector. The advantages and disadvan- sessed. Failure modes, mechanisms, and causes have been re-tages of aiternatue independent organizations were studied, viewed trum operating expenences and existing data banks. namely, an Office of Nuclear Safety headed by a director report- The study has also included consideration for the seismic corre-ing to the Executwe Director for Operations (EDO) of NRC; an laton of age-degraded motor components. The aforementioned Office of Nuclear Safety headed by a director reporting to the reviews and assessments were assimilated to charactenze the NRC Commissioners; a multi-member NTSB-type Nuclear de-electric, rotational, and mechanical hazards on motor per-Safety Board independent of the NRC. The costs associated formance and operational readiness. The functonal indicators with operating a Nuclear Safety Board were also included in the which can be monitored to assess motor component detenora. study. The legislatrve requirements, both new authonty and ten due to aging or other accidental stressors are identified changes to the existing NRC legislatwe authonty, were studied. Conforming with the NPAR strategy as outlined in the program These legislatue requirements were based upon the Edwards- plan, the study also includes a preliminary discussion of current Udall Bill H.R. 6390 introduced in the 96th Congress and study standards and guides, maintenance programs, and research ac-of the NRC Organizaten Act. twities pertaining to nuclear power plant safety-related electnc motors. NUREG/CR-4153: APPLICATIONS OF FOREIGN PROBABILIS-YlC SAFETY ASSESSMENT EXPERIENCE TO THE U S. NU- NUREG/CR-4157: A SCIENTIFIC CRITIQUE OF AVAILAOLE CLEAR REGULATORY PROCESS. ANDREWS,W.B. Battelle MODELS FOR REAL-TIME SIMULATIONS OF DISPERSION. Memonal Institute, Pacific Northwest Laboratones. February LEWELLEN W.S.; SYKES,R 1. Aeronautical Research Assocs-1985.163pp. 8503130128. PNL 5388. 29359 207. ates of Pnnceton. March 1985.180pp. 8503200t26. ARAP 472. This report is a summary of applications of probabilistic safety 29468.19 t . assessment (PRA) in the United States and foreign countnes. It This report provides an evaluation of soveral available dispor. is intended to stimulate discusson on the applicability of foreign soon models to determine their suitability for providing the capa-expenence to the United States, provide information on foreign bility for estimating the effects of accidental discharges of radio-safety technology development and focus the United States actwe matenal at nuclear power plants. A entique of the as-goals for future participation in the actwities of the Committee sumptions involved and a review of existing venfication studies for the Safety of Nuclear Installations (CSNI), Pnnciple Working are made for models ranging from the Geussian plume with Group 5 (PWG5) Results indicate that the United States leads straight line winds to models which attempt a complete solution the surveyed countnes in the completon and application of of the pnmitive equations of motort it is demonstrated that al-comprehensive PSAs of public safety impacts from nuclear though even the simple models are capable of providing rea-power plants. European expenence has focused on the use of sonably accurate predictions under ideal conditons, there are reliability analyses in support of design and operational dech reasons to expect relatwely severe limits on plume predictability sions. It is recommended that use of probabilistic analyses be when certain emission conditons are combined with certain me-expanded in the United States for engineenng applicatens teorological conditions. The usefulness of a real-time dispersion based on the success in European countnes. model is thus likely to be dependent on a complementary esti-mate of the vanability expected about the mean dispersion for NUREG/CR-4155: TR AC-PF t / MOD 1 INDEPENDENT the conditions existing at that time. This report is one of a set of ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORAT. three dealing with real-time disperson models. The other two ED-PLATE CCFL TESTS. DOBRANICH.D. Sandia National Lab. deal with the uncertainties invo.ved in the depositon module of oratories. Apnf 1985. 42pp. 8505060503. SAND 85-0172 drspersion models and the results of testing some of the disper. 30190:241, sion models reviewed in this report by comparing them with the The TRAC-PF1/ MODI independent assessment protect at data collected at the Idaho National Engineenng Laboratory in Sandia is part of an overall effort funded by the NRC to deter. July,1981 dunng an NRC sponsored field test. mine the ability of vanous sys! ems codes to predict the detailed thermal / hydraulic response of LWRs dunng accident and off- NUREG/CR-4158: A COMPILATION OF INFORMATION ON UN-normal conditions. As part of this effort, calculatons for some of CERTAINTIES INVOLVED IN DEPOSITION MODELING the Northwestem Unuersity perforated-plate CCFL tests have LEWELLEN W S.; VARMA.A K ; SHENG.Y.P. Aeronautical Re-been performed. Two input models were constructed to repre- search Associates of Pnnceton. Apnl 1985.79pp.8504250168 sent the rectangular test channel: a two-dimensional model and ARAP NO. 504.30032.015. a faster running one-dimensonal model The results of both The current generaton of disperson models contains very models indicate that for high water flow rates, with the water in- simple parametenzatons of deposition processes. The analysis lected vertically above a perforated plate TRAC overpredicts here lonks at the physical mechanisims goveming these proc-the steam ibw rate necessary for completo weeping (CCFL)- esses in an attempt to see if more valid parametenzations are However, for flow conditions more typical of PWR transients, available and what level of uncertainty is involved in either TRAC provides a reasonabie pied.ct.on of weeping. these simple parametenzat.ons or any more advanced parame-tenzaton. The report is composed of three parts. The first, on NUREG/CR-4156: OPERATING EXPERIENCE AND AGING-SEIS. MIC ASSESSMENT OF ELECTRIC MOTORS. SUBUDHl.M ; dry deposition model sensitivity, provides an estimate of the un. BURNS.E.L; TAYLOR.J. Brookhaven Natonal Laborato y June certainty existing in current estimates of the deposition velocity due to uncertainties in independent vanables such as meteoro. 1985.150pp.8511180647. BNL-NUREG-51861. 33506 297. This report provides an aging assessment of electnc motors logical stability. particle size, surface chemical reactivity and and was conducted under the auspices of the NRC Nuclear canopy structure. The range of uncertainty estimated for an ap-Plant Aging Research Program (NPAR). The objectwes of this propnate dry deposition velocity for a plume generated by a nu-clear power plant accident is three orders of magnitude. The program are to ident;fy concerns related to the aging and serv. ice wear of equipment operat:ng in nuclear power plants, to second part discusses the uncertainties involved in precipitaton assess their possible impact on plant safety, to identify effectue scavenging rates for effluents resulting from a nuclear reactor inspection, survedlance and monitonng methods and to recom- accident The conclussen is that maior uncertainties are involved rr.end suitable maintenance practices for m.tigating aging relat- both as a result of the natural vanability of the atmosphenc pre-ed concerns and diminish the rate of degradaten due to aging cipitaten process and due to our incomplete understanding of f I and service wear. Motor design and matenals of constructen the undertying process The third part involves a review of the are reviewed to edentify age. sensitive components. Operational important problems associated with modeling the interaction be-l t
l Main Citations and Abstracts 75 l tween the atmosphere and a forest. It gives an indication of the NUREG/CR-4164: DATA REPORT FOR THE TPFL TEE /CRITI-magnitude of the problem involved in modeling dry depositen in CAL FLOW EXPERIMENTS. ANDERSONJL.: . OWCA,W.A. { such environments. EG8G Idaho, Inc. (subs. of EG&G, Inc.). November 1985. t NUREG/CR-4159: COMPARISON OF THE 1981 INEL DISPER- 118pp. 8512270266. EGG-2377. 34081:188. i SiON DATA WITH RESULTS FROM A NUMBER OF DIFFER- A senes of expenments have been performed investigating ENT MODELS. LEWELLEN,W.S.; SYKES,R.I.; PARKER,S.F. the phenomena of hquid entrainment and vapor pull through at Aeronautcal Research Associates of Pnneeton. May 1985. a tee pnction between a honzontal pipe and a small branchhne. l 202pp. 8505310421. ARAP NO. 505. 30670:001. These experiments were performed under conditions of strats-Results from simulations by 12 different disperson models fied steam-water flow at 3.4. 4.4, and 6.2 MPa in the 28 4cm are compared with observations from an extensive field expen- diameter mainhne, and entical flow through a nozzle installed in ment at the Idaho National Engineenng Laboratory in July, the branchhne. Two orientations of the branchline were invests. 1981. Compansons were made based on hourly ground-level gated: horizontal and vertical downflow. This report documents SF(6) samples, out to approximately 10 km from the 46 m re- the expe e ner tal program, presents the data obtained, and dis-lease tower,' both dunng and following 7 different 8-hour re- cusses condatens for predicting the levels at which the onset 4 leases. Compansons are also made for total integrated doses of vapor pull-through and hquid entrainment occur and correla-l collected out to approximately 40 km. Within the hmited range tions for predicting the flow quahty into the branchhne l appropnate for Class A rnodels this data companson shows that 4 neither the puff models or the transport and diffusion models NUREG/CR-4166: ANALYSIS OF FLECHT SEASET 163 ROD agree with the data any better then the simple Gaussian plume BLOCKED BUNDLE DATA USING COBRA.TF. PA1K,C.Y.; models. The pff and transport and diffusion models do show a HOCHREITEH.LE. Westinghouse Electnc Corp. KELLY,J.M.; et slight edge in performance in companson with the total dose al. Battelle Memonal Institute, Pacific Northwest Laboratones. over the extended range appropnate for Class B models. The October 1985.692pp.8511010008. EPRI NP-4111, 33288:150. Flow blockage and spacer gnd heat transfer models for rod
; best model results for the hourly samples show approximatefy j bundle arrays have been developed for a two-phase flow situa.
40% calculated within a factor of two when a 15 degree uncer-tainty in plume positen is permitted, and it is assumed that ton charactenstic of a PWR reflood. These models have been y' higher data samples may occur at statens between the actual incorporated into COBRA-TF. which is a three-dimensional, three-field, two-fluid mechaniste two-phase flow subchannel sample sites. This is increased to 60% for the 12 hour integrat-ed dose and 70% for the total integrated dose. None of the computer code. Companson of the predcted flow blockage heat transfer m large rod bundle arrays with test data endcates that l models reproduce the cbserved patchy dose pattems. This pat-j the blockage and gnd heat transfer models used with the chiness appears to be consistent with the inherent uncertainty COBRA-TF code agree quite well with the measured data. Bias j associated with time averaged plume observabons. plots of the predicted and measured temperatures nses from
~~} NUREG/CR-4160: HISTORICAL~~
~~SUMMARY~~
OF OCCUPATIONAL different tests indicate that, in general, the computer code cal- { RADIATION EXPOSURE EXPERIENCE IN U S. COMMERCIAL culations tend to underpredict the heat transfer improvement NUCLEAR POWER PLANTS. MOELLER,M P.; STOETZEL.G.A.; observed to have been caused by gnds and blockage in the ex-1 MUNSON.L.H. Battelle Memonal Institute Pacific Northwest penments. The pnncipal reason for heat transfer improvement i Laboratories. Apnl 1985. 6Spp. 8505070439. PNL-5404. due to blockage and gnds is the breakup of the entrained hquid i 302 t 4.293. droplets in the supurheated steam flow above the quench front.
! This report organizes existing data on occupatonal radiation The breakup of these entrained drops results an a populaton of f exposure experience for consideration in the safety goal evaiua- much smaller drops, which are more easily evaporated in the ; tion program. The report includes a review of occupatonal radi. superheated vapor. The enhanced heat transfer observed in j ation exposures incurred by workers at commercial U.S. nuclear and downstream of blockages aind gnds is abo attnbute to in- , power plants. In additon, occupatonal radiat.on exposure infor. creased turbulence caused by the droplets in the steam flow.
j mation is presented for work performed at commercial U.S. nu. The resulting computer models and methods of modehng both I clear power plants to meet regulatory actions and required gnds and blockages, whch are desenbed in this report, are be-backfits. This information identifies specife operations per. heved to be apphcable to PWR safety analysis. Appication of j formed as part of these requirements. Where possible, actual such models is expected to signifcantly reduce or ehminate the radiation exposure histones are provided. A bnef history of radi. calculated peak clad temperature penalty due to flow blockage j ation dose hmits and a review of the biological and health er. for a hypothetical PWR LOCA, using the Appendix K cntena fects attnbutable to radiation exposure is included to provide a perspective on the development of radiation protecten regula- NUREG/CR 4167: FLECHT SEASET PROGRAM Final tons Report NRC/EPRI Westinghouse Report Number 16.
'. HOCHREITER.L.E. Westinghouse Electrc Corp. November ! NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HlGH- 1985.200pp.85t2300016. EPRI NP-4112. 34092:177.
LEVEL WASTE REPOSITORY. Volume 1. Basalt BINNALLE.P.;. This report presents the highlights and main findings of the 4 WOLLENBERG.H.A.; BENSON.S M; et al. Lawrence Berkeley USNRC, EPRI, and Westinghouse cooperative FLECHT j Laboratory. May 1985. 95PP. 8506060810. UCID-20092. SEASET program. The repert indicates areas in whch results of 30769:292. the program can contnbute to revising the current heensing re. This report addresses entrcal parameters specific to a reposi- quirements for Loss of Cool.nt (LOCA) safety analysis for i tory in basalt, using the Columbia River Basalt Group as the PWRs Also identified are severil technical areas in which the pnncipal example. For the purposes of this teport, a paiametar new FLECHT SEASET data and analysis can lead to improved is considered to be a physical property whose value helps de- safety analysis,modehng, and thereby to predicted PWR re- ; l termine the charactenstics or behaver of a repository system. sponse for postulated accident scenarcs. Signifcant progress , [ Parameters whch are defined as entical are those essential to has been made in the modehng areas of nonequihbnum dis- ; j evaluate and/or monitor leakage of radionuclides from the re- persed two-phase flow dunng reflood. Improved models and un. j pository and to evaluate the need for retneval. The parameters derstanding of this rod bundle coohng regime are summanzed in i are considered with respect to the disciphnes of geomechanics, this report. Another important result of the FLECHT SEASET geclogy, hydrology, and geochemistry and are rank ordered in program anses from the natural circulaton test senes, which in- , terms of importance. The specific role of each parameter, spe- vestigated significant single. phase, two-phase, and reflux con-
! crfic factors affecting the measurement of each parameter, and densaton coohng modes of scaled PWR undcrsmall-break j the interrelationships between the parameters are consedered in LOCA conditons. The tests and subsequent analysis constitute detail, one of few complete sets of data for these coohng modes in j
~~{ 1 i i~~
~~- , - _ . . - . _ - . . - . _ , _ . _ . --m.-,.m. .r__c__.r, . . _ . .- _ , . , , - . _ , - - - _ _ _ , _ _ , , _ . _ _ ~ , , _ , _ . . - , . , _ -~~
76 Main Citations and Abstracts which full-height, multitube steam generators with sufficient in- result through development of a prototype real. time RNS proc-strumentation were used to examine pnmary-to-secondary heat essor, The hardware is expected to be made up of 100 nsec bit transfer in the generators. It is believed that the natural circula- slice microprocessor components and large RAM storage units. tion test data will be estremely useful to benchmark the im- Based on the performance estimates of the Phase l effort, this proved post TMI small-break LOCA computer codes. new image processing system has the potential to acquire and focus the equivalent of the 145 A-scans per second, which NUREG/CR-4168: GT2F.A COMPUTER CODE FOR ESTIMAT. ING LIGHT WATER REACTOR FUEL ROD FAILURES. translates into more than 1000 cube inches per man. inspection WILLIFORD.R.Ea LANNING.D.D ; BEYER.C.E. Battelle Memon- rate for typical pressure vessel specimens. al Institute, Pacific Northwest Laboratones. May 1985. 277pp. 8506060567. PNL-5354, 30771:002. NUREG/CR-4172: A USER'S GUIDE FOR MERGE. FREEMAN. This report desenbes the development, benchmarking and re- KELLY; JUNG.R G. Batteile Memonal Institute, Columbus Lab-suits of a computer code designed to permit comparison of oratones. . March 1985. 41pp. 8504040006 BMI-2121. BWR and PWR fuel rod failure behaviors dunng postulated re- 29629 233. actor off-normal events such as control rod withdrawal errors. The MERGE code acts as the interface between the MARCH-The code is called GT2F and was developed from the 2 code, which is used to determine overall accident progres. GAPCON-THERMAL-2 code by the addition of new models for sion, and the TRAP-MELT code, which is used to evaluate reac-calculating transient temperatures, fisaon gas release, mechans- tor coolant system fission product transport and deposition. cal interacton between fuel and cladding, and Ziredluy cladding MERGE uses MARCH 4:siculated core exit flows ar,d tempara-
*scture behavor. Results indicate that for the conservatwely tures to perform a detailed gas-to-structures heat transfer analy-severe overpower transient scenanos assumed, a full length s S for the control volumes in the flow path through the reactor commercial BWR fuel rod has a failure probability between 1t' coolant system and converts these results into a form required and 4.5*e at 27 mwd /kgM when the transient begins from high as input to TRAP-MELT. MERGE can treat up to nine control operating power. A full length commercial PWR fuel rod has a volumes, containing up to fwe structures each. Required inputs fDlure probability between 2*. and 11*. at 28 mwd /kgM when include desenptons of the control volumes and their slow con-the transient begins from low power. Failure probabilities are nectons, as well as initial conditions substantially smaller at lower burnup and for less extreme tran-sient conditions. NUREG/CR 4173: CORSOR USER'S MANUAL. KUHLMAN M.R.;
NUREG/CR-4169: AN APPRCACH TO TREATING RADIONU. LEHMICKE.D J. Battelle Memorial institute, Columbus Laborato-CLIDE DECAY HEATING FOR USE IN THE MELCOR CODE nes. MEYER.R O. NRC - No Detailed Affiliaton Given. March SYSTEM. OSTMEYER.R.M. Sandia National Laboratones. June 1985. 58pp 8504040423. BMI-2122. 29618 312. 1985. 33pp. 8507050426. SAND 841404, 31371:175. The CORSOR code simulates the release of fission products A new code system is being developed for use in assessment and structural matenals from a reactor core dunng the in-vessel of nuclear reactor accident nsks. The code system, termed penod of a severe accident in a light water reactor. The code is MELCOR, will treat thermabhydraule and fission product behav- a simple, empincally based treatment of release and does not ior jointly. A part of its treatment of thermal-hydrauhc processes, treat detailed mechanisms for release from fugh temperature the code system will evaluate decay heating from fissson prod- fuel. The first-order release rate coefficients for the species uct inventones contained wit hin the reactor core debns and considered are presented, the input requirements of the code compartments that are defined for the reactor system and con- are desenbed, and an example input and output stream is sup-tiinment. A simple approach to treating radonuchde decay phed an an appendix. heating is proposed for use in MELCOR. The proposed ap-proach uses a table-lookup to estimate element decay powers NUREG/CR-4174: ROCK MASS SEALING - EXPERIMENTAL AS-ts a functon of time after reactor shutdown (start of accident). SESSMENT OF BOREHOLE PLUG PERFORMANCE. Annual Decay power for each element in a compartment of the reactor Report. June 1983 May 1984. DAEMEN.J Jf; GREER.W Ba system is found by multiplying the mass of the element in the ADISOMA.G S ; et af. Anzona, Unsv. of, Tucson, AZ. March compartment by the element's decay-heat rate per unit mass 1985. 384pp. 8504090456. 29739:193. which is a function of time after reactor scram. The approach This report descnbes expenmental borehole plugging per-Essumes that daughter proriucts are transported along with the formanca assessments performed, started, or planned dunng parent radionuclide during the accident. The validity of this as, sumpton is discussed. In additon, methods for apportioning the June 1983 - May 1984. Results are given from field fluw tests decay energy between the walls and the gases in a compart. on three coment plugs installed in vertical boreholes in basalt ment are also discussed. The proposed approach is based on and on one nearly honzontal cement plug The honzontal plug SANDIA ORIGEN calculations for a 3412 MWt PWR and a 3578 and one vertical plug seal very well. Hydrauhc conductivity of MWt BWR. two vertical field plugs has been re!atively high. Remedial act.on is descnbed. Laboratory simulatons of dyname loading of NUREG/CR-4170: AN ULTRA-HIGH SPEED RESIDUE PROCES- cement plugs show no detnmental effects. Drying of Cement SOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCE- plug, especially for months, at elevated temperatures, increases MENT. POLKY,J N ; MILLER.D.D. Sigma Research. Inc. March the hydraule conductivity of the plugs severely, and reduces 1985. 42pp. 8504030453. 29605:178. their bond strength along the interface. Microscopic inspection, The Phase-I feassbehty study of residue number system (ANS) strength and flow tests identify the dnihnganduccd damaged image processing for SAFT inspection has successfully deter-mined that an advanced inspecton system may be budt using a ZON 'n basalt. While such a damaged zone exists, and has typi-correlation-reconstruction SAFT algonthm. implemented with cal features (e g. fracture density, size, location, onentation), it RNS techniques and off the-shelf electronic components. is thin and not likely to be a preferential flowpath. Engineenng Images are reconstructed in a number theoretic transform charactenstics tests on bentonite plugs, chemical analysis and domain with simple pointwise multiphcation of the A. scan data sweihng tests. Expenmental designs are given for the study of volume by a custom point spread functon (PSF), all in a highly s:re and of thermal effects on plug performance. Preliminary re. parallel computational architecture. These methods also allow suts are presented Results are included from ongoing cement image enhancement to be easily performed for improved flaw push out tests and swelling measurements. visuahzation, and with neghgrble speed reduction. It has been determined that high resolution three dimensonal flaw images NUREG/CR-4176: EMISSION CONTROL TECHNOLOGY AND may be generated and that a commercially viable product coulo
Main Citations and Abstracts 77 OUALITY ASSURANCE NEEDS AT URANIUM MILLING from this evaluation that many of the LERs failed to meet all of FACILITIES Includes Supporting Methods For the requirements. The report presents the methodology that Testing. Operating And Maintaining Air Pollution Control Devices. was used to evaluate the LERs, the conclusions reached con-LUDWICK,JD. Battelle Memonal Institute, Pacific Northwest cerrsing problem areas in the reports, and suggestens as to Laboratones. June 1985 55pp. 8507030684. PNL-5386. how the overall quahty and completeness of reports can be im. 31318.311. proved. In addition, plant trecific informaton is provided that Pacific Northwest Laboratory, under contract to the U S Nu. will perrnt an assessment of each lecensee's perfonnanco clear Regulatory Commission, conducted an investigation of particulate emission control devices for apphcation to process NUREG/CR-4180: STATE OF THE-ART OF SOLID-STATE exhausts at uranium millino facilities. The scope of this investi. MOTOR CONTROLLERS. JAROSS,R A; MULCAHEY,T.P.; l gation included devices now in use, as well as those devices KOEHL.E.R. Argenne National Laboratory. Apol 1985. I f 8pp. ! that have potential apphcation for milhng sites. This report pre. 8504180201. ANL-84102. 29921.264. Sents the results of the study. Emission control devices are cat. The state-of-the-art of sohd-state motor controllers (SSMCs) (gonzed and desenbed, including high-efficiency and moderate. is assessed in terms of use, probabihty of Class 1E quahfication, cfficiency devices as well as other (some novel) devices useful failure rate expenence, and rehabihty prediction. Surveys of in specific situations. Preoperational consideratens discussed commercial availabihty, nuclear and nonnuclear electnc utshty include selecting devices. instrumentation, and testing pro- expenence, and architect-engineenng use were made relative to grams. Operational and maintenance considerations related to the suitabikty of SSMCs for nuclear service. Reasons for the dry and wet removal processes are desenbed. Quakty assur- hmited use of SSMCs in nuclear plants are gwen Available fad-Ence documents and topics are also discussed ure rate data are meager, and are augmented by data on other sohd-state power electronic devices that are st'own to have NUREG/CR-4177 V01: MANAGEMENT OF SEVERE subcomponents simdar to those found in SSMCs. In addition to ACCIDENTS. Perspectives On Managing Severe Accidents in large nmnuclear sokd-state adjustable-speed motor dnves, the Commercial Nuclear Power Plants. DISALVO.R.; LEONARD.M ; rehabihty of nuclear plant inverter systems and high-voltage MANAHAN,M.; et al. Battelle Memonal Institute, Columbus Lab- sohd-state DC transmission hne converters is assessed Class oratories. May 1985. 105pp. 8506130369 BMI-2123. 1 evir nnwntal quahkaton expenence wdh near pant 30868E6- convertechnverters and battery chargers in shown to be directly Accident management is examined from several related per- apphcable to SSMCs No problems are expected in quahfying spectives. The emphasis is on the role >f the operating crew them. Actual rehabihty predictions of two typical commercial SSMCs are grven, together with predictions of improvements and the technical support provided to tnem before, dunng, and possible with use of high-quahty parts and manufactunng proce-after an accident. The relationship among accident manage- dures. ment, nsk management and emergency management is exam-ined. The roles played by industry, regulaton. and research are NUREG/CR-4181: LEACHABILITY OF RADIONUCLIDES FROM reviewed. Finally, the results of viewing accident management CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT OPER. from these vanous perspectives are ref ected in the articulation ATING NUCLEAR POWER REACTORS. CRONEY,S.T. EG&G of issues and some proposals for their rescltten. Idaho, Inc. (subs of EG&G, Inc ) Apn! 1985 130pp. 8505070173. EGG-2355. 30217.161. NUREG/CR-4177 V02: MANAGEMENT OF SEVERE Different sized samples of coment soldfied hquid wastes ACCIDENTS Extending Plant Operating Procedures into The were collected from two nuclear power plants, a pressunted Severe Acedent Regeme WR EATHALL.J ; LEONARD.M ; water reactor (PWR) and a boshng water reactor (BWR), to cor-DISALVO.R Battelle Memonal Institute Columbus Laboratones. relate radionucl.de teaching from small and full sized waste May 1985. 75pp. 8506130132. BMI-2123. 30667:108 arms. Diffusion-based model analyses of measured radionuchde This study examines the feasibihty and valuehmpact of ex. leach data from small samples and full size samples indicated t nding emergency operating proccdures into the severo acci. that leach data from small samples could be used to determene dent reyme. It reviews the types of knowledge needed to devel- leachabehty indexes for full size waste form. The leachabikty in-op such procedures and the apphcabihty of existing regulatory deres for cesium, strontium, and cobalt isotopes were deter-review cntena. A method is developed and ilfustrated in two mined for waste samples from both nuclear plants according to cases This study concludes that it is feasible to develop proce- models used in ANS 16.1. The leachabihty indexes for the PWR dures for cperators to mitigate the consequences of accidents samples were 6 4 for cesium, 7.1 for strontium, and 10.4 for progressing past the onset of core damage. A prehminary cobalt. The leachabihty indexes for BWR samples were 6 5,8 6 valuehmpact assessment indicates a significant hkekhood of ' '
there being an overall net positive benefit of developing mitiga-tive procedures. A phased program has been proposed. First a NUREG/CR-4182: VERIFICATION OF SOIL STRUCTURE INTER-ACTION METHODS. MILLERC A ; COSTANTINO.C J ; PHILIP-pdot study should develop the apphcaton of the rnethods used in this feasibehty study and provide more precise information for PACOPOULO, et al. Brookhaven National Laboratory. July 1985.182pp. 8508090676, BNL NUREG-51893 32102161.
e detailed value-impact assessment. Based on the results of the so,l. structure interacton (SSI) methods currently used by in-pdot study, extension to a greater population of plants may be dustry to evaluate the seismic response of nuclear power plant justified. facihties are revvewed with the aim of evaluating those areas of uncertainty which still exist in the analytic approaches. The pn-NUREG/CR-4178 DRFT: AN EVALUATION OF SELFCTED Li- mary methodologies used by vanous agencies generally can be CENSEE EVENT REPORTS PREPARED PURSUANT TO 10 grouped into three areas, namely, lumped parametermeethods, CFR 50 73 Draft Report. ANDERSON.B S ; MILLER.C F.; finite element methods of combined soil / structure systems, and VALENTINE.B.M. EG&G Idaho, Inc. (subs. of EGaG, Inc ). substructunng methods of analysis. Each of these are dis-March 1985. 28pp. 8512120149. EGG-2381. 33873 011. This cussed in the report. In general, it was found that lumped pa. report desenbes an evaluation of an industry-wide sample of Li- rameter approaches yield reasonable results providerf that the censee Event Reports (LERs) that was performed to determine ote is relatively uniform and the seismic inputs are low enough whether or not these LERs were prepared in accordance with such that nonhnear effects are unimportant. The finite element the requirements set forth in 10 CFR 50.73. It was determined results are reasonable provided that extreme care is taken in determining mesh geometry, boundary conditions. 3D effects,
78 Main Citations and Abstracts etc. Similar conclusions can be applied to the structuring ap- in a typical compliance manner and the findings reported. En-proaches. Closed puotographs of the work stations, dunng source and equipment use, illustrate conditions and the licenseos' oper. NUREG/CR-4185: AN ASSESSMENT OF DOSIMETRY DATA ations. It is concluded from observations dunng onsite visits to FOR ACCIDENTAL RADIONUCLIDE RELEASES FROM NU- these unusual work environments, that penodic onsite comple-CLEAR REACTORS. RUNKLE.G E.: OSTMEYER R H. Sandia ance inspections are necessary to assure radiatson protection li!ational Laboratories. September 1985. 6tpp. 85f 00f 0219. f r all concerned. SAND 85-0283. 32833:329. This report reviews dosimetry models for estimatng the ab- NUREG/CR 4191: SURVEY OF LICENSEE CONTROL ROOM sorbed dose from external and internal exposure to radionu- HABITABILITY PRACTICES. BOLAND.J F.; BROOKSH:RE,R L.; chdes. Important modehng parameters and assumptons are de- DANIELSON,W.F.; et al. Argonne National Laboratory Apnl scribed. Recommendations for the dosimetry data to be used in 1985. 225pp. 8505100194. ANL-85-13. 30268 213. the MELCOR health and economic consequence model are This document presents the results of a survey of Licensee made. For estimating the dose from cloudshine and ground- control-room-habitability practices. The survey is part of a com-shine, the models for extemal exposure developed by Kocher prehensive program plan instituted in August 1983 by the NRC ue recommended. The ICRP-30 models and metabolic param- to respond to ongoing questions from the Advisory Committtee sters are recommended for estimating the dose from radenu- on Reactor Safeguards (ACRS). The emphasis of this survey chdes deposited internally via inhalaton and ingestion. Dose was to determine by field rsview the control-room habitability conversion factors calculated with these rnodels for a vanety of practices at three different plants, one of which is still '.ander rtdionuclides, clearance classes, particle sizes and integration construction and scheduled to receive an operating hcense in penods were obtained from Oak Ridge National Laboratory for 1986. The other two plants are currently operating having re-use in the MELCOR health and economic consequence model. cerved operating hcenses in the mid-1970's and earty 1980's. Sources and magnitude of uncertainty in dose factors were The major finding of this survey is that despite the fact that the svaluated. Recommendations are made for assessing the un- latest control-room-habitabihty systerns have become large and certainty in estimated consequences due to uncertainty in dose more complex than earlier systems surveyed, the latest systems converson factors. do not appear to be functionally supenor. The major recommen-NUREG/CR-4189: TRAC-PF1/ MODI INDEPENDENT dation of this report is to consohdate into a single NRC docu-ASSESSMENT.Semiscale MOD-2A Feedwater Line Break (S- ment, based upon a comprehensive systems engineenng ap-proach, the pertinent cntena for control-room-habitability design. SF-3) And Steam-Line Break (S-SF.5) Tests. DOBRANICH.D. Sandia Natonal Laboratones. November 1985. 144pp. NUREG/CR-4192: THE ANALYSIS OF DRAINAGE AND CON-8512270342. SAND 85-0576. 34078.347. The TRAC-PF1/ MODI independent assessment protect at SOLIDATION AT TYPlCAL URANIUM MILL TAfLINGS SITES. FAYER,M J.; CONBERE,W. Battelle Memonal Institute, Pacific Sandia is part of an overait effort funded by the NRC to deter. Northwest Laboratones. May 1985. 56pp. 8506190031. PNL. mine the ability of vanous systems codes to predict the detailed thermal / hydraulic response of LWRs dunng accident and off. 542 f. 31017.324. normal conditions. As part of this effort, calculations for Semis. The computer code TRUNC was used to analyze three as-cafe Mod-2A test S-SF-3, a feedwater.hne break test, and S SF- pects of uranium mill taihngs dewatenng the couphng of Con-soFdation and fluid flow, drainage design, and cover load One-5, a steam-hne break test, were performed with TRAC-PF1/ MODt. Most aspects of both the S-Sf-3 and S-SF-5 steady. dimensenal simulations of the effects of consohdation on fluid state calculations were found to be in good agreement with flow within a taihngs pile of either shmes or a sand /shmes mix data. However, the need for a better steam separator model showod that drainage flux was greater for a conschdating was adentified from the S-SF-3 calculaton. Overall, the quahta. system earty in the simulatson. After days 1.400 and 160 of the 8'mulations for tt'e shmes and sand /shmes mix, respectively, f.ve behavior of both transients was calculated reasonably well; however, the fluxes for the nonconsolidating systems were however, there were some disc *cpancies in the prediction of the greater. In the sand /shmes mir, the nonconsohdating system quantitative behavior. The results for the S-SF 3 transient calcu. had a cumulative flur by day 5,000 thol was 93% of that of the liten indicate that the pnmary to secondary heat transfer deg. consohdating system. At the same time, in the shmes taihngs radation began too earfy. This was possibly due to overpredic. piles the nonconschdating system had a cumultsve flux of only tion entrainment in the steam generator boiler, leading to an in. 34% of that of the consohdahng system, whsch #ndscates that correct calculation of the secondary-side fluid distnbutson dunng consohdation and fluid flow shou!d not be decoupled for the the steady state. However, there was insufficie'1t data to venfy shmes. Two-dimensional simulatons of an actual taihngs pile this. Results for the S-SF-5 transient calculation indicate that drainage design showed that a sand blanket drain increased the the pnmary-side fluid temperature response to a steam-hne rate of drainage and settlement. The sand blanket drain also break was in reasonable agreement with data but the pressure S'gnificantly reduced differential settlement across the pile. This response did not coincide with the data. Uncertainties in the indicates that the use of a sand blanket drain could enable ear. data are sufficient to prevent us from determining the ejract her placement of the cover system after taihngs emplacement. cause of this discrepancy, but there is indirect evidence that the In simulations of covered and uncovered taihngs piles, nearty calculated rate of phase change in the pressunzer was incor- the same quantity of water was removed from each, but drain-rect. age occurred much more slowly without the cover; hence, sur-NUREG/CR-4190: CALIFORNIA OFFSHORE SURVEY OF Ll- face settlement was slower when the taihngs pile was not cov-CENSEES USING RADIOACTIVE MATERIAL WONG,KS.; cred. BROWN.J. Cahfornia, State of. May 1985. 22pp. 8506060807. 30770.316. NUREG/CR-4194: LOW LEVEL NUCLEAR WASTE SHALLOW This report is an account of offshore radcactNe material ac- LAND BURIAL TRENCH ISOLATION Final Report. October 1981 tvities and was prepared to provide information about their safe September 1984. MCCRAY,J G ; NOWATZKl,E.A.; use in the manne environments beyond Cakfornia's junsdicton- ARMSTRONG.G,; et al. Anzona, Univ gf. Tucson, AZ. May The report supphes the essential information called for and (a) 1985. 219pp. 8506240003 31149 012. identifies hcensees with radioactive nuchde utshzaton programs, This is the final report on a three year study to evaluate (b) describes the hcensees' work stations, (c) identifies and/or trench cap designs, trench construction and trench loading by desenbes radionuchde, quantities and their apphcations, and (d) accelerating the creation of void space to simulate waste degra-desenbes the radiation safety concerns ;nd existing methods dation in order to apply stress conditions on the trench in a rel-for their resolution. Finally, three offshore sites were inspected stive short penod of time. Eight trenches were initially Construct. I
l Main Citations and Abstracts 79 ed and instrumented, four in a semi and region and four in a of glass will almost certainly be used as the waste form for high more humid mountainous region. After the first year the semi- level nuclear wastes and such castings tend to fracture as a
; trio site was abandoned due to cap failures. A new trench in- ' result of transient and residual stresses induced by the casting corporating an improved soil slab design with a wick was con- process; such fractures increase the surface area avaliable for i structed at the humid site. Conclusions from these expenments sre: 1. Controlled compaction is not suffcient to mitigate long aqueous leaching of radonuclides from the HLW glass. The pri-mary focus of this study was on achieving an understanding of term surface subsidence. 2. Single sheet geotextile reinforce- the dependence of fracture surface area on glass properties ment is not adequate trench cap reinforcement. 3. Geotextle 4 wrapped soil slab attenuates surface subsidence and surface and processing vanables for both in-can melts and castings.
~~The maximum fracture surface area per unit volume of glass ob. water infiltration. 4. A steel-reinforced soil-cement slab appears served in this study was about 7.1/cm (an equivalent sphencat~~
~~; to meet the requirements necessary for long term stability. 5. If particle diameter of 0.85 cm) for a water quenched in-can melt.~~
the crown and cap remain stable so does the trench. 6. Ahphat. ic tracers performed well and dye type of tracers poorty. 7. The processing parameter which appears to most strongty affect the extent of fracture surface area for both castings and Tracers are feasible and effectue as a trench monitonng tool. 8. Narrow designed trenches improve trench cap stabihty. This in-can melts is the dimensionless Bot modutus (thermal film co-efficient x radius / waste form thermal conductuity). i report recommends a design for enhanced isolation disposal trench providing improved monitonng capabshties. NUREG/CR 4199: A DEMONSTRATION UNCERTAINTY /SENSI. 1-4
' NUREG/CR-4195: OVERVIEW OF TRAC-PD2 ASSESSMENT TlVITY ANALYSIS USING THE HEALTH AND ECONOMIC CALCULATIONS. WATERMAN.M E. EG8G Idaho, Inc. (subs. of CONSEQUENCE MODEL CRAC2. ALPERT,DJ.; IMAN.R L; EG&G Inc.). November 1985. 70pp. 8512270237. EGG-2380. HELTON,J.C.; et at Sandia National Laboratones. June 1985. -
59pp. 8507050415. SAND 841824. 31372.210. su rnary of Transient Reactor Analysis Code Version A densstraten unmain/sensitwW anaWs was perfounM PD2(TRAC-PD2) calculations performed at the Idaho Natonal for the resctor accident consequence model CRAC2 using tech-Engineenng Laboratory (INEL) is presented in this report as part niques compiled as part of the NRC-sponsored MELCOR pro-of the U.S. Nuclear Regulatory Commission's (NRCs) overall as- gram The pnncipal objectwes of the study were: 1) to demon-sessment program of TRAC-PD2. The calculated and measured strate the use of the uncertain /sensmvrty anatysis techniques, 4 parameters summanzed in this report are break mass flow rate, 2) to test the computer models that implement the techniques pnmary coolant system pressure, reactor core flow rates, and 3) to identify possible difficulties in performing such an analysis, l fuel rod cladding temperatures. The data were obtained from and 4) to explore alternatwe means of analyzing, displaying. and
seven tests that were performed at two test facilities. The tests desenb.ng the results. Seventeen CRAC2 input vanables were conducted to study the various aspects of cold leg break thought to contnbute significantly to uncerta:nty in estimated
~~; transients, including the effects of large and small breaks, and consequences were selected for anatysis; subjective estimates i~~
core reflood phenomena. User expenence gained from the vari- of ranges, distnbubons, and correlatons for these vanables ous calculations is also summanzed. were made. Latin hypercube sampling, a modified Monte Carlo technique, was used to generate two multuar'ste samples of NUREG/CR-4196: OVERVIEW OF TRAC-BD1 (VERSION 12) AS- size 50 from the distnbutions assigned to the 17 input vanables. SESSMENT STUDIES. KULLBERG C.M. EG&G Idaho, Inc. A total of 100 CRAC2 runs,50 for each sample, was performed. (subs. of EG&G, Inc.). Apnl 1985. 55pp. 8506060796. EGG-2382. 30771:279. The results of the two samples were similar. A regression analy-sis was performed to estimate the contnbution of each vanable A senes of simulatons were performed at idaho Natonal En-to uncertainty in estimated consequences. The study was first gineering Laboratory to continue the advancement of Boiling performed with the magnitude of the source term as one of the Water Reactor (BWR) safety research, with the TRAC-BD1 17 vanables. A second analysis was performed with a fixed (Version 12) computer code. The pnncipal motwaton for per-source term. Only one sample of size 50 was generated in the forming these simulatons was to assess the code's capabikty to
' second analyrJs. The uncertainty /sensitnnty analysis techniques calculate Loss-of-Coolant Accident (LOCA) related phenomena- compiled for MELCOR appear well suited for use with a health The results of a number of TRAC-BD1 (Verson 12) simulatons, and economic consequence model Altemative methods for dis-which cover a broad range of conditons dunng different tyec' of LOCA scenanos, are summanzed in this document. Selem playing and desenbing the results are presented. The insights gained from performing the analysis are reviewed and major compansons between calculated and measured results are pre- conclusions summarized. A companson of the r6sults of this sented. Conclusions derived from those compansons are gwen- study with current point estimates of health and economic con-NUREG/CR-4197: SAFETY GOAL SENSITIVITY STUDIES. sequences is presented.
BURKE,R.P.; BLOND,R.M. Sandia Natonal Laboratones. June 15785. 50pp. 8507020415. SAND 85-G634. 31313.301. NUREG/CR-4200: BIODEGRADATION TESTING OF SOLIDIFIED This study presents the results of analyses performed as part LOW-LEVEL WASTE STREAMS. PICIULO.P.L: SHEA,C.E.; 4 of the two-year evaluabon prugram for the NRC safety goals. BARLETTA R E. Brookhaven National Laboratory. May 1985. Analyses are performed to demonstrate the sensituibes of the 46pp.8506163593. BNL-NUREG-51868. 30936:219. quanbtatue design objectue ca'culations to changes in input The NRC Technical Position on Waste Form (TP) specines j parameters and assumptions. Results are presented which that waste should be resistant to biodegradation. The methods show the influence of parameter changes on the health risk recommended in the TP for testing resistance to fungi, ASTM quantitatue design objectives and on cost-benefit calculations. G21, and for tesbng resistance to bactena, ASTM G22, were , The attematue design objectue nsk measures are compared camed out on several types of sohdified simulated wastes, and with alternatue measures of the health impacts of LWR acci- the effect of microbial activity on the mechanical strength of the
~~' dents. The results of this study provide background informabon materials tested was examined. The tests are beheved to be j~~
~~and input to be used in the NRC staff evaluation of the safety sufficient for distinguishing between matenals that are suscepti-~~
. goals and quanbtatue design objectwes. ble to biodegradation and those that are not. However, it is con-1 cluded that failure of these tests should not be regarded of NUREG/CR-4198

FRACTURE IN GLASS /HIGH LEVEL WASTE itself as an indicaton that the waste form will bodegrade to an CANISTERS. MARTIN.D M. Iowa State Univ., Ames, IA. Apnt extent that the form does not meet the stabihty requirements of 1985. 81pp. 8504170534. 29906.243, 10 CFR Part 61, in the case of failure of ASTM G21 or ASTM The release rate of radenuchdes from a vitnfied waste form G22 or both, it is recommended that additional data be supphed due to aqueous leaching by ground water will depend, among by the waste generator to demonstrate the resistance of the other factors, on the waste form's surface area. Large castings waste form to microbial degradaten.
v_m-rm - - - - . , _ ,--m. --m,,,,-..-n-.-.---...----+-,-,-.--.,-
80 Main Citations and Abstracts NUREG/CR-4201: THERMAL STABILITY TESTING OF LOW- NUREG/CR-4208: GASTROINTESTINAL ABSORPTION OF PLU-i- LEVEL WASTE FORMS. PICIULO,P.L: CHAN,S F. Brookhaven TONIUM IN M:CE. RATS, AND DOGS.Appleation To Establish. ' - National Laboratory. May 1985. 48pp. 8506060814. BNL- ing Values Of f t For Soluble Plutonium. BHATTACHARYYA; NUREG-51869. 30769:247, LARSEN.R.P.; OLDHAM R.D.; et al. Argonne National Laborato-4 The NRC Technical Positen (TP) on Waste Form specifies ry. May 1985. 99pp. 8507050425. ANL-85 21. 31371:207. that waste forms should be resistant to thermal dogradation. The gastrointestinal (GI) absorption of plutonium was meas-The thermal cycle testing procedure outhned in the TP on ured in mice, rats, and dogs under cond. tons relevant to setting Waste Form was camed out and is believed adequate for dem- dnnking water standards. The fractonal Gi absorption of Pu (VI) onstrating the thermal stability of solidified waste forms. The in- n adult mice was 2 x 10(-4) (0.02%) in fed mice and 2 x 10(.3) clusion of control samples and the monitonng of sample tem- (0.02%) in fasted mice. The GI absorption of plutonium was in-perature are recommended additens to the test. An outline for dependent of plutonium oxidation state, administration medium, reporting thermal cycling test results is guen. To produce a data and plutonium concentration; absorption was dependent upon J base on the applicability of the thermal cycling test, the follow- animal species, state of animal fasting, state of Pu(IV) hydroly-ing simulated laboratory-scale waste forms were prepared and sis, and age of the animal. Fractional GI absorption values tested; bonc acid and sodium sulfate evaporator bottoms, mixed ranged from 3 x 10(-5) (0.003**) for hydrolyrod Pu(IV) adminis-b3d bead resans, and powdered resins each solidified in asphalt, tered to fed adult mice to 7 x 10(-3) (0.7**) for Pu(VI) adminis-j cement and vinyt ester-styrene. tered to fed neonatal rats. From analysis of our data, we sug-NUREG/CR-4203: A CALCULATIONAL METHOD FOR DETER, gested values of f(1) (the fracion transferred from gut tc blood in humans) for use in estabhshment of oral hmits of exposure to MINING BIOLOGICAL DOSE RATES FROM IRRADIATED RE. plutonium. For an acute ermsure in the occupatenal setting, SEARCH REACTOR FUEL SCHNITZLER.B.G. EG&G Idaho, Inc. (subs. of EG&G, Inc.). Apnl 1985. 65pp. 8506060811. EGG. we proposed one value of f(1) for fed (2 x10(4) and one for ; i fasted (2 x 10(-3) individuals. For the environmental setting, we 2383. 30775:237. developed two approaches to obtaining values of f(1), suggest-I This report desenbes a calculatonal method for the determ,. ed values were 6 x 10(-4) and 4 x 10(-3), respectuely. Both ap-nation of biological dose rate from irradiated research reactor fuels..The ca'culational method is implemented in a computer proaches took into account effects of animal age and fasting. program for quick and convenient assessment of multsgroup We discussed uncertainties in proposed values of f(1) and gamma and beta dose rates resulting from an arbitrary (user- made recommendatens for further research. supplied) irradiation history. The FUELDR program calculates
' dose rates at a fixed dose point using built-in fission product im- NUREG/CR 4209: COMPARISON OF ANALYTICAL PREDIC-pulse source functions and precalculated gamma and beta TIONS AND EXPERIMENTAL RESULTS FOR A 1:8-SCALE transport factors. The fixed dose point is located on the axial STEEL CONTAINMENT MODEL PRESSURIZED TO FAILURE.
mid-plane at a distance of 91.44 cm (3 ft) from the fuel element. Ct.AUSS,0.B. Sandia Natonal Laboratories. September 1985. l 68pp. 851206000S. SAND 85-0679. 33789-045 Transport factors are included for sixteen unique (235)U fuel types in use at thirteen nonpower reactor facibtles. Predictions for the response of a 1.8-scale model of a steel nuclear containment building to overpressurizabon are com. j NUREG/CR-4204: LONG TERM EMBRITTLEMENT OF CAST pared to expenmental results. Finite element analyses were DUPLEX STAINLESS STEELS IN LWR SYSTEMS: Annual used to predict the model's response. Strains, displacements. Report. October 1983 - September 1984. CHOPRA,0.K.; and leak rate measurments were made at 21 different pressure 1' CHUNG.H.M. Argonne National Laboratory. Apnl 1985. 33pp- levels. Compansons of the pressure histones for strain and dis-8506140599. ANL-85-20. 30935.223- placement at a point, and the spatial variaton of strain and dis-This progress report summanzes work performed by Argonne placement are made. In addition, compansons of a more global i National Laboratory dunng the twelve months from October nature, such as the capacity of the model and the failure mode, j 1983 to September 1984 on long-term embnttlement of cast are discussed. An evaluation of the predictive capabilities and
~~duplex stainless steels used in light-water reactors. the failure cnteria is made in the light of these comapnsons.~~
NUREG/CR-4205: TRAP-MELT 2 USER'S MANUAL JORDAN.H; NUREG/CR-4210: MATADOR:A COMPUTER CODE FOR THE l KUHLMAN.M.R. Battelle Memorial Institute, Columbus Laborato- ANALYSIS OF RADIONUCLIDE BEHAVIOR DURING DEGRAD. nes. May 1985. 74pp. 8506190035. BMI 2124. 31017:100. ED CORE ACCIDENTS IN LIGHT WATER REACTORS. The TRAP-MELT 2 code is a development of the previously BAYBUTT.P.: RAGHURAM.S.: AVCI,H.l. Battelle Memonal Insh-issued TRAP-MELT code which simulates the transport and tute, Columbus Laboratones. Apnl 1985. 62pp. 8505080375. depositen of aerosol particles and certain vapors in the reactor BMI-2125. 30218.271 j coolant system under hypothetical accident conditions in a hght water reactor. This manual contains a bnef desenption of the models of the processes treated in the code and of the code Analysis of Transport And Depositon Of Radonuchdes) has ' been developed to replace the CORRAL computer code which 4 organization. The input to the code for a sample run are pre, was wntten for the Reactor Safety Study (WASH 1400). This 5 sented and output from a run are presented as well. report contains a detailed desenption of the models used in MATADOR. A companion report provides a User's Manual for 1 NUREG/CR-4206: A SELECT REVIEW OF=THE RECENT (1979 the code. MATADOR is intended for use in system nsk studies 1983) BEHAVIORAL RESEARCH LITERATURE ON TRAINING SIMULATORS. LAUGHERY,K.R. Oak Ridge Natonal Laborato- to analyze radionuchde transport and deposition in reactor con-ry. May 1985. 51pp. 8506130489. ORNL/TM.9445. 30901:249. tainments. The pnncipal output of the code is information en the Report summanzes some selee,tod reports of behavioral re- timing and magnitude of radionuchde releases to the environ-I search performed in years 1979-1983 on training simulator ap- ment as a result of severely degraded core accidents. MATA. , phcation techtclogy, and discusses findings related to nuclear DOR considers the transport of radonuchdes through the con-power plant operators' simulator training. Findings are organized tainment and their removal by natural depositen and the oper-as related to the design, testing, and use of training simulators. aten of engineered safety systems such as sprays. The code Topics include Simulator Fidehty vs. Training Effectiveness, Op- requires input data on the source term from the pnmary system, j erator Performance Measurement, Measunng Simulator Effec- the geometry of the containment, and the thermalhydrauhc
~~tiveness, and Simulator Utahzation Practices. Reviews 89 refer- conditions in the containment.~~
~~I ences. i l i~~
~~..-- _..,- --- _ _--- ___ . ,_,_. - ., _ -..____ - - - ... _., .- - . ~ . - - - - - _ - - . . - . .~~
Main Citations and Abstracts 81 NUREG/CR-4211: MATADOR (METHODS FOR THE ANALYSIS Analysis of the radiological health effects of nuclear power OF TRANSPORT AND DEPOSLTION OF RADIONUCL1 DES) plant accidents requires models for predicting early health ef-CODE DESCRIPTION AND USER'S MANUAL AVCl,H.I.; fects, cancers and benign thyroid nodules, and genetic effects. RAGHURAM,S.; BAYBUTTP. Battelle Memorial Institute, Co- Since the pubhcation of the Reactor Safety Study, additionaf in-lumbus Laboratories. Apnl 1985. 75pp. 8505080373. BMI-2126 formation on radiological health effects has become available. 30218:192. This report summanzes the effort of a program designed to pro-l A new computer code called MATADOR (Methods for the vide revised health effects models for nuclear power plant acci-l Analysis of Transport And Deposition of Radionuchdes) has dent consequence modelkng The new models for early effects been developed to replace the CORRAL-2 computer code address four causes of mortality and nine categones of morbedi-l' which was wntten for the Reactor Safety Study (WASH-1400). ty. The models for earfy effects are based upon two parametet t This report is a User's Manual for MATADOR. A companion Weibell functions. They permit evaluation of the influence of report describes in detail the models used in the code MATA- dose protraction and address the issues of vanation in radiosen-DOR is intended for use in system nsk studies to anafyze radio- sitwity among the population. The piecewise-knear dose re-nuchde transport and deposition in rr: actor containments. The sponse models used in the Reactor Safety Study to predict can-pnncipal output of the cooe is infor' nation on the timing and cers and thyroid nodules have been rep'3ced by linear and magnitude of radionuchd, releases to the environment as a linear-quadratic models. The new models reflect tne most re-result of severely degraded core accidents. MATADOR consid- cently reported results of the follow up of the survivors of the ers the transport of radionuchdes through the containment and bombings at Hiroshima and Nagasaki and permit analysis of their removal by natural deposition and by engineered safety both morbidity and mortality. The new models for genetic ef-systems such as sprays. It is capable of analyzing the behavior fects allow prediction of genetic nsks in each of the first fNe of radionuclides existing either as vapors or aerosols in the con- generations after an accident and include information on the rel-tainment The code requires input data on the source terms into ative sever ty of vanous classes of genetic effects. The uncer-the containment, the geometry of the containment, and thermal- tainty in modelhng rad.ological health nsks is addressed by pro-hydrauhc conditions in the containment. viding central, upper and lower estimates of nsks. An approach NUREG/CR-4212: IN-PLACE THERMAL ANNEALING OF NU- is outlined for summanzing the health conseouences of nuclear CLEAR REACTOR PRESSURE VESSELS. SERVER.W L. p wer plant accidents. EG&G Idaho, Inc. (subs. of EG&G, Inc ). Apnl 1985. 250pp-8505070548. EGG MS-6708. 30211:101. NUREG/CR-4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL WASTE FORM ACCEPTANCE CRITERIA. Radiation embnttlement of femtic pressure vessel steels MACKENZIE.D R ; VASLOW,F.; DOUGHERTY,D R ; et al. changes the toughness properties. A thermal anneal cycle well Brookhaven National Laboratory. May 1985.77pp.8506140405 above the normal operating temperaturt of the vessel can re* BNL NUREG-St873. 30908116. store most of the onginal properties. l'he Army SM-1 A test re- This report provides technical support to NRC in connection actor vessel was wet annealed in 1967, and wet annealing of with the regulation 10 CFR Part 61 and NRC's Technical Post. the Belgian BR-3 reactor vessel has recently taken place. An tion (TP) on waste form Six specAc areas are addressed, industry survey indicates that dry anneahng of a reactor vessel namely: the technical basis for hmating containers of radioactwe in-place is feasible, but solvable engineenng problems exist-gases to atmosphenc pressure and 100 cunes; the require. Limited toughness data available for fue high Copper content ments to demonstrate that a stable waste would be recogniz-welds were reviewed. The review suggested that significant re-able for 300 or 500 years; the feasibility of achieving less than covery results from annealing at 454 degrees centigrade (850 5% deformation in buned wastes; the adequacy of ASIM tests degrees fahrenheit) for one week, but scatter in the data makes G21 and G22 for testing for b;odegradabihty; the adequacy of assessment of recovery and reembrittlement response difficult to quant fy. A thermal and structural analysis of a reactor vessel ASTM test B553 for testing for inermal degradation; and the basis for determining if a waste is explos ve of pyrophonc. The undergoing an anneahng treatment found no problems with the reactor vessel itself, but did indicate a rotaGon at the nozzle pnncipal conclusions of the report follow. A maximum pressure of 1.5 atmospheres for radioactive gases is acceptable, but the region of the vessel which would plastically deform the attached pnmary piping. Further analytical studies attempted to solve this radioactrvity hmit should depend on the isotope, the quality of the container and the properties of the site. Site and package probtera, but they were not successful. An Amencan Society for quahties and a wet / dry cychng test are suggested that apprecia-Testing and Matenals (ASTM) task group is upgrading and re- bly increase the probabihty of indicating whether a waste would vising guide ASTM E 509-74 with emphasis on the matenals have long-term recognizabihty. Achieving deformation of buned and survedlance aspects of anneahn9 waste of <5% would not be feasible using current sohdification NUREG/CR 4213: SETS REFERENCE MANUAL WORRELL.R B. methods with eitner metal or polyethylene containers. ASTM Sandia National Laboratones. Jury 1985. 250pp. 8508090642. tests G21 and G22, with modifications are suitable for biodegra. SAND 83-2675. 32097:137. dabikty testing A modified form of ASTM B553 is adequate for The set Equation Transformation System (SETS) is used to thermal testing Required information on pyrophonc and explo-achseve the symbohc manipulation of Boolean equations. Sym. save matenals is provided. bolic manipulation involves changing equations from their ongs-nal forms into more useful forms - particularly by applying Bool- NUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR ean identities. The SETS program is an interpreter which reads. POWER PLANT VALVE FAILURE-RATE VARIABILITY--SOME interprets, and executes SETS user programs. The user wntes a PRELIMINARY RESULTS. BECKMAN.R.J ; MARTZ,H F. Los SETS user program specifying tha processing to be achieved Alamos Scientific Laboratory July 1985.70pp.8508090719 LA-and submits it, along with the required data for execution by 10396-MS. 32105 076. SETS. Because of the general nature of SETS, io., the capabil- Valve failure data from the in-Plant Rehability Data System sty to manipulate Boolean equations regardless of their ongin, (IPROS) are statistically analyzed using the Fadure Rate Anaty-the program has been used for many different kinds of analysis ses Code (FRAC). Data from the five failure modes, four of which are time related and the other demand rela'ed, are ana. NUREG/CR 4214: HEALTH EFFECTS MODEL FOR NUCLEAR fyzed to determine which of the factors--operating system, valve POWER Pi ANT ACCIDENT CONSEQUENCE ANALYSIS Part size, valve type, operating type, and operating mode- most Untroduction, Integration & Summary.Part ll. Scientific Basis For affect valve failure rates. A separate analysis is given for each Health Effects Models. EVANS.J.S ; MOELLER.D.W ; of two plants, a pressunzed water reactor (PWR) and a boshng COOPER.D.W.; et al. Harvard Univ., Cambridge, MA. August water reactor (BWR). For both plants and each failure mode, 1985. 357pp. 8509190140. SAND 85-7185. 32681:109. multipkcative adjustments for the mean are obtainnd for catego-
82 Main Citations and Abstracts nes, such as nuclear or containment systems, of the varcus related research review, generic analysis and plant specific factors. These multipliers indicate whether a particular category analysis. Licensee Event Reports (LERs) and Integrated Leak of a factor has a correspond.ng failure rate that is less than the Rate (ILRT) Test reports provided the major sources of contain-tverage failure rate (a multiplier less than one) or greater than ment performance information used in this study. Data extracted average (a multiplier greater than.one). Based on the multiphca- from the LERs were assembled into a computer data base, twe adjustments, ball valves are shown to be the most reliable Oualitative and quantitatue information developed for cuntain- i valves for the PWR plant. Globe and gate valves have the high. ment performance under normal operating conditions and est failure rates for this plant. The average failure rate for the design basis accidents indicate that there is room for improve-BWR plant is found to be hatt that of the PWR plant for thr?e of ment. A crude estimate for overall containment unavailabikty for the fue failure modes studied. In addition to the multipliers, relativefy small leaks which violate plant technical specifications point estimates and confidence intervals on the failure rates are is 0.3. An estimate of containment unavailabihty due to large given for selected valve factor combinations. These estimates leakage events is in the range of 0.001 to 0.01. These esti-End intervals are compared with several other estimates. mates are dependent on several assumptions (particularty on event duration times) which are documented in the report. NUREG/CR-4k.8: LOCA SIMULATION IN THE NATIONAL RE-SEARCH UNIVERSAL REACTOR PROGRAM.Postirradiation NUREG/CR 4221: AN EVALUATION OF STRESS CORROSION Examination Results For The Third Matenals Test (MT-3)
~~CRACK GROWTH IN BWR PIPING SYSTEMS. KASSIR.M.;~~
Second Campaign. HABERMAN.J.H. Battelle Memonal Institute. SHARMA,S.; REICH M ; et al. Brookhaven Natonal Labora*,ry. Pacific Northwest Laboratones. June 1985. 62pp. 8506260395. May ITS. 80pp. 8506130175. BNL NUREG-51874. 30867:183. PNL-5433. 31244:326. This report presents the results of a study conducted to A senes of in-reactor expenments were conducted using full- evaluate the effects of stress intonssty factor and environment length 32 rod pressunzed water reactor (PWR) fuel bundles as on the growth behavior of integranular stress corrosen cracks ptrt of the Loss-of-Coolant Accident (LOCA) Simulaton Pro- in type 304 stainless steel piping systems. Most of the detected gram by Pacific Northwest Laboratory (PNL). The third matenals cracks are known to be circumferential in shape, and initially test (MT-3) was the sixth experimer't in a series of thermal-hy* start at the inside surface in the heat affected zone near girth draulic and matenals deformation / rupture expenments conduct- welds These cracks grow both radially in-depth and circumfer-ed in the National Research Unwersal (NRU) Reactor, Chalk entially in length aqd, in extreme cases, may cause leakage in R ver, Ontano. Canada. The MT-3 expenment was jointly funded the installation. The propagation of the crack is essentially due by the U.S. Nuclear Regulatory Commission (NRC) and the to the influence of the following $4multaneous factors: (1) The United Kingdom Atomic Energy Authonty (UKAEA) with the acton of applied and residual stress. (2) Sensitization of the main objectue of evaluation ballooning and rupture dunn9 base metal in the affected zone adjacent to firth weld and (3) tctrve two-phase cooling at elevated temportures. All 12 test The continuous exposure of the matenal to an aggressive envs rods in the center of the 32-rod bundle failed with an average ronment of high temperature water containing dissolved oxygen peak strain of 55.4% At the request of the UKAEA, a destruc* and some levels of impunties. Each of these factors and their twe postirradiaton examination (PIE) was performed on 7 of the effects on the piping systems is discussed in detail in text of the 12 test rods. The results of this examination were presented in report. The report also evaluates the time required for hypothets a previous report. Subsequently, and at the request of UKAEA, cal cracks in BWR pipes to propagate to their cntical size. The PIE was performed on three additional rods along w:th further pertinent times are computed and displayed graphically. Finally, examination of one of the previously examined rods. Information parametric study is performed in order to assess the relative in-obtained from the PIE included cladding thickness measure- fluence and sensitwity of the vanous input parameters (residual ments, cladding metallography, and particle size analysis of the stress, crack growth law, d4ameter of pipe, enetial size of defect, fractured fuel pellets. This report desenbes the additional PIE etc) which have beanng on the growth behavior of the inter-work performed and presents the results of the examinations granular stress corrosion cracks in type 304 stainioss steel. Cracks in large-diameter as well as in small-diameter pipes are NU7.EG/CR-4219 V01: HEAVY SECTION STEEL TECHNOLOGY considered and analyzed. PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTO-BER 1984 - MARCH 1985. PUGH.C E Oak Ridge National Lab- NUREG/CR-4225:
~~SUMMARY~~
OF EFFICIENCY TESTING OF oratory. July 1985. 216pp. 8507250168. ORNL/TM 9593/V1. STANDARD AND HIGH-CAPACITY HIGH-EFFICIENCY PAR. 31783:146- T6CULATE AIR FILTERS SUBJECTED TO SIMULATED TOR-The Heavy-Section Steel Technology (HSST) Program is an NADO DEPRESSURIZATION AND EXPLOSIVE SHOCK engineering research actwity conducted by the Oak Ridge Na- WAVES. SMITH.P.R.; GREGORY,W.S. Los Alamos Scientific tional Laboratory for the Nuclear Regulatory Commission. The Laboratory. Apnl 1985. 25pp. 8507020407. LA 10401 MS. program compnees studies related to all areas of the technolo- 31309 007. gy of matenals fabncated into thick section pnmary-coolant con- Pressure transients in nuclear facslity air efeaning systems can trinment systems of light-water-cooled nuclear power reactors. onginate from natural phenomena such as tornadoes or from The investigation focuses on the behavior and structural integri- accident-induced explosive blast waves. This study was con-ty of steel pressure vessels containing cracklike flaws. Current cerned with the effectwo efficiency of high-efficiency particulate work is organized into ten tasks: (1) program management, (2) air (HEPA) filters dunng pressure surges resulting from simulat-fracture-methodology and analysis, (3) matenal charactenzation ed tornado and explosion transients. The pnmary obectue l of and properties, (4) environmentally assisted crack growth stud- the study was to examine filter efficiencies at pressure levels les, (5) crack arrest technology (6) irradiation effects studies, below the point of structural failure. Both standard and high-ca-(7) cladding evaluations, (8) intermediate vessel tests and anal- pacity 0 61 m by 0.61 m HEPA filters were evaluated, as were ysis, (9) thermal-shock technology, and (10) pressunzed ther- several 0.2-m by 0.2 m HEPA fdters. For a particular manufac-mal-shock technology. turer, the matenal release when subjected to tornado transients is the samo (per unit area) ror both the 0 2-m by 0.2-m and the NUREG/CR-4220: RELIABILITY ANALYSIS OF CONTAINMENT SYSTEMS. PELTO.P.J.; GALLUCCl,R H.; O 61 m by 0.61m filters. For tornado transients, the matenal was ISOLATION AMES.K.R. Battelle Memonal Institute, Pacific Northwest Lab. on the order of micrograms per square meter. When subjecting oratones. June 1985. 222pp. 8506260362. PNL 5432. clean HEPA filters to simulated tornado transients with aerosol entrained in the pressure pulse, all filters tested showed a deg-3 f 245:199. This report summarizes the results of the Reliability Analysis radation of filter efficiency. For explosse transients, the matenal of Containment isolation System Project. Work was performed release from preloaded high-capacity filters was as much as in fue basic areas: design review, operating expenence review, 340 g When preloaded high-capacity filters were subjected to
i 1 1 Main Citations and Abstracts 83 shock waves approximately 50*'. of the structural hmit level, I fied, stem from reviews of several, industry sponsored, full-to 2 mg of particulate was released scope Probabilistic Risk Assessments (PRAs) and vanous deter-NUREG/CR-4226: NEW MADRID SEISMOTECTONIC nwnistic/probabihstic approaches used by industry to judge their STUDY.Actuities Dunng Fiscal Year 1983. BUSCHBACH,T.C. canpliance with or used to seek exemptions from the fire-pro-St. Louis Univ., St. Louis, MO Apnl 1985.153pp. 8505070552. tection requirements enumerated in Appendix R to 10 CFR 50. 30209:297. In pufaming this evaluation of the current methodologies, The purpose of the New Madnd Seismotectonic Study is to state-oMhe-art hterature on the modehng of propagation /detec-identify the earthquake mechanisms within a 200-mile radius of ton / suppression, input parameters, and modeling uncertainties New Madrid, Missouri. Dunng 1983 there was moro awareness am utihred. Areas are adentified where recently-developed, more of the significance of current regional stress patterns and the accurate and complete techniques can be implemented tc local concentraton of stresses by basement structures and in- reduce the state-of-knowledge uncertainties that presently exist. homogeneities. The program continued to concentrate on dehn- Recommendatens are also made which could be the basis for ing boundaries of a proposed nft complex in the area, as well a more suitable and complete fire-nsk methodology. Es establishing the relationships of the east west trending fault systems with the northeast trending faults of the Wabash Valley NUREG/CR-4230: PROBABILITY BASED EVALUATION OF SE-and New Madnd areas. There were 204 earthquakes located by LECTED FIRE PROTECTION FEATURES IN NUCLEAR the Saint Lc sis University microearthquake network in 1983. In POWER PLANTS. AZARM.M,A.; BOCCIO.J L. Brookhaven Na-addition, the earthquake swarm in north-central Arkansas con- tional Laboratory. May 1985. 93pp. 8506180415. BNL-NUREG-tinued throughout the year, and 45,000 earthquakes have been recorded there since January,1982. Trenching data from Late g Cenozoic terrace deposits along the Kentucky River Fault tection measures in nuclear power plants is desenbed. The System suggest that there was post-terrace deformation along methods developed are apphed to two representatue fire areas some of the faults. Thermal and chemical data from groundwat-9 P' g ers in tae Mississippi Embayment appear to be useful in locahz- diesel generator room. The fire areas chosen for application, ing deep faults that cut through the aquifers. Earty indications the fire scenanos desenbod, and tr.e vanous fire-damage states from studies of jointing in Indiana are that the direction of major specified in the two illustrative examples are used to evaluate joint sets will be useful in determining regional stress directions. those fire-protection guidelines which deal with automatic / No Quaternary faulting was found an the Indiana or lihnois fault manual fire detection and suppresson systems, rated barners, studies. divisional separation, drainage systems, dampers, and fire rating of electrical cables. Tabular results are presented, which reflect NUREG/CR-4227: HUMAN ENGINEERING GUIDELINES FOR the relative ments of these systems / features in terms of conds-THE EVALUATION AND ASSESSMENT OF VIDEO DISPLAY tional probabihties of achieving vanous room-damage states. UNITS. GILMORE.W E. EG&G idaho, Inc. (subs: of EG&G, Inc ) July 1985. 535pp 8508150085. EGG-2388. 32193.075. The conclusions drawn and the lessons learned through the This report provides the Nuclear Regulatory Commissen with course of this study are discussed, and the areas that may need further investigation are identified. a single source that documents known guidelines for conducting formal Human Factors evaluations of Visual Display Units NUREG/CR-4231: EVALUATION OF AVAILADLE DATA FOR (VDUs). The handbook is a " cookbook of acceptance guide- PROBABILISTIC RISK ASSESSMENTS (PRA) OF FIRE lines for the reviewer faced with the task of evaluating VDUs al- EVENTS AT NUCLEAR POWER PLANTS. SAMANTA P K.; ready designed or planned for service in the control room. The BOCCIO.JL. Brookhaven National Laboratory. KRASNER.L.M.; creas addressed are visual displays, controls, control / display in- et al. Factory Mutual Research Corp. May 1985. 71pn tegraton, and workplace layout. Guidehnes relevant to each of 8506190077, BNL-NUREG 51879. 31017.205, those areas are presented. The existence of supporting re- Several crucial parameters are needed in the assessment of search is also indicated for each guidehne. A comment section fire nsk in nuclear power plants. Among those that need to be and Method for Assessment section are provided for each set developed from a data base are: (1) fire frequency, (2) fire de-of guidelines. tection time, and (3) fire suppression time. Currently, that data NUREG/CR-4228: REVIEW OF THE VOGTLE UNITS 1 AND 2 base for nuclear power plants is not large enough to develop AUXILIARY FEEDWATER SYSTEM RELIABILITY ANALYSIS. these parameters, considenng fuel location, fuel geometry, com-FRESCO A.; YOUNGBLOOD,R.; PAPAZOGLOU,1 A. Brookha- bustion properties, enclosure geometry, etc. This study attempts ven National Laboratory. October 1985. 95pp. 8511120038. to augment the nuclear data base by investigating the useful-BNL NUREG-51876. 33440:228. ness of other nonnuclear data bases which contain fire incident This report presents the results of the review of the Auxihary loss expenence of occupancy classes having somewhat similar Feedwater System rehabihty analysis for the Vogtle Elactnc physical features and fire protecten engineenng systems nor-Generating Plant (VEGP) Units 1 and 2. The objective of this ma!!y found in nuclear power plants. This study has found that report is to etttimate the probability that the Auxihary Feedwater indeed some useful information can be gleaned from nonnucle-System will fail to perform its miss on for each of three different ar sowces; in part cular, detection and suppression times. How-initiators: (1) loss of main feedwater with offsite power avai ' able, ever, other fero-nsk data needs such as fire frequency and fire (2) loss of offsite power, (3) loss of all ac power except vital size would require other forms of data searches and data analy. instrumentaton and control 125-V dc/120-V ac power. The ses that at this stage can only be conceptualized. scope, methodology, and farfure data are presenbed by NUREG 0611 Appendix IIL The results are compared wrth NUREG/CR-4232: THE RESPONSE OF VENTILATION those obtained in NUREG-0611 for other Westinghouse plants. DAMPERS TO LARGE AIRFLOW PULSES. GREGORY,W.S; SMITH.P R. Los Alamos Scientific Laboratory. July 1985. 71pp. NUREG/CR-4229: EVALUATION OF CURRENT METHODOLOGY 8507250121. LA 10413-MS 31794 051. EMPLOYED IN PROBABILtSTIC RISK ASSESSMENT (PRA) OF The results of an expenmental program to evaluate the re-FIRE EVENTS AT NUCLEAR POWER PLANTS. RUGER,C.; sponse of ventdation system dampers to simulated tornado BOCCIO.J.L; AZARM,M A. Brookhaven National Laboratory. transients are reported. Relevant data, such as damper re-May 1985. 47pp. 8506190103. BNL-NUREG-51877. 31017:277, sponse time, flow rate and pressure drop, and flow / pressure vs The report presents a general evaluation of the current meth- blado angle, were obtained, and the response of one tornado odology used by industry for the probabilistic assessment of fire protective damper to simulated tornado transients was evaluat. events in nuclear power plants. The basis for this evaluation, in ed. Empincal relationships that will allow the data to be integrat-which the strengths and weaknesses of the methods are identi- ed into flow dynamics codes were developed Theso flow dy.
84 Main Citations and Abstracts namics codes can be used t'y safety analysts to predict the re- identifying the Sohd phase in radonuclide exponments and high-sponse of nuclear fac;hty ventilation systems to tornado dcpres- bghts the weaknesses of the actinide thermodynamic data sunzations. bases used in geochemical modeling calculations. An evaluaton was made of the information developed by DOE on the natve NUREG/CR 4233: DISTRIBUTION OF CORBICULA FLUMINEA copper deposits of Michigan as a natural analog for the possi-AT NUCLEAR FACILITIES. COUNTS.C L Delaware, Univ. of, ble emplacement of copper canisters in repository in basalt. Lewes, DE. November 1985. 86po. 8512270228. 34084.171. e similanty in bulk chemistry of the basalts, rehed upon heavi-A review of the zoogeographic records for the exotic Asian ly by DOE in their analysis, cannot be used to unequivocally clam, CORBICULA FLUMINEA (Muller,1774), reveals its pres- conclude that similar geochemical controls, particularfy controls ence in 27 states where nuclear powered electnc generating on the geochemical canditons, exist within the basalt / water plants are either operating or under constructon. Mneteen plant systems at Mich gm anJ the Hanfo'd Site. Thus, the DOE anal-sites reported infestations of varying seventy in facihties or ysis is insufficient to conclude, with reasonable assurance, that source waterbodies immediately adjacent to the facility by popu- copper wdl be stacle at the Hanford Site, lations of C. FLUMINEA. Thirteen plant sites are focated within the zoogeographic hmits of C. FLUMINEA but have a low nsk of NUREG/CR 4236 V02: PROGRESS IN EVALUATION OF RADIO-infestation due to either salt water cochng systems or locations NUCLIDE GEOCHEMICAL INFORMATION DEVELOPED BY a great distance from known populations. Eighteen plant sites DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE are located wholly outside of the known zoogeographic range of PROJECTS Report for January. March 1985_ KELMERS.A.D.; C. FLUMINEA. Thirty plant sites are located in close proximity SEELEY,F.G.; ARNOLD.W D ; et at. Oak Ridge Natonal Labora-to known populatons of C. FLUMINEA and therefore should tory. December 1985. 52pp. 8601070466. ORNL/TM 9614. mainta'n surveillance of the source water body and within plant 34155 078. water systems for possible infestations by these bivalves- Geochemical informaton relevant to the retention of radonu. chdes by the Hanford Site (in basalt) and the Yucca Mountain NUREG/CR-4234 V0f: AGING AND SERVICE WEAR OF ELEC- S'te (in tuff). candidate high. level nuclear waste geologic repose. TRIC MOTOR-OPERATED VALVES USED IN ENGINEERED tones being developed ty U.S. Department of Energy (DOE) SAFETY FEATURE SYSTEMS OF NUCLEAR POWER projects, is being evaluated by Oak Rdge National Laboratory PLANTS. GREENSTREET,W.; MURPHY,G A; (ORNL) for the U.S. Nuclear Regulatory Commission (NRC) Our EISSENBERG,0.M. Oak Ridge National Laboratory. July 1985. evaluation of the sorption of technetium by basalt / groundwater f 2'pp. 8507250t49. ORNL-6170/VL 31808.069. This is the first in a senes of three reports on electric motor. systems was essentially completed this quarter and the results summanzed; we conclude that the expenmental methodology operated valves (MOVs) to be produced under the U S. Nuclear and results reported by the DOE for the Hanford Site have not Regufatory Cornmisson's Nuclear Plant Agsng Research pro. conclusively estabbshed tha. significant retardation of techneti-gram. This program addresses the evaluaten and identification um migraton may be provided by phases present in the basalts of practical and cost-effectue methods for detecting, monitor. of the Hanford Site. We have shown that sodium boltwoodite is ing, and assessing the seventy of time-dependen' dogradation the saturating uranium solid phase in two basaft/ groundwater (aging and ser< ice wear) of MOVs in nuclear plants. These systems. Because thermodynamic data are not available for methods are to provide capabilities for establishing degradation sodium boltwoodite, calculated solubahties for uranium are erro-trends pror to failure and developing guidance for effectwe neous in these systems. Results of radionuedde solubehty/speco maintenance. This report examines fadure modes and causes aton calculabons, published by the DOE for the Yucca Moun-resulting from aging and service wear, manufacturer-recom. tain site, were evaluated this quarter urider our geochemical mended maintenance and surveillance practicns, and measura. ble parameters (including functonal indicators) for use in as. mcdehr,g task. We express concerns relatwo to the inherent i:mitatens of such calculations. Samples of Yucca Mountain tuff sessing operational readiness, estabbshing degradation trends, and J-13 well water were received for use in our planned radio-and detecting incipient failure. The results presented are based nuchde sorption /solubdiff expenments. These Yucca Mountain on information denved from operating expenence records, nu. matenals will be used to evaluate radionuchde sorption and ap-clear industry reports, manufacturer-supplied information, and pamnt concentrafen hmit values pubbshed by the Nevada Nu-input from architect engineer firms and plant operators. NUREG/CH 4236 V01: PROGRESS IN EVALUATION OF RADIO-NUCUDE GEOCHEMICAL INFORMATION DEVELOPED BY NUREG/CR-4237: MOBILITY OF RADIONUCLIDES IN HIGH CHLORIDE ENVIRCNMENTS. SIMPSON.HJ.; HERCZEG A.L.; DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE ANDERSON,R.F.; et al. Columbia Univ , New York, NY, Apnl PROJECTS REPORT FOR OCTOBER-DECEMDER 1984. KELMERS.A.D.; SEELEY,F /;; ARNOLD.W D.; et al. Oak Ridge 1985. 77pp. 8505070484. 30210 219. Natonal Laboratory. October 1985 59pp. 8511180625 ORNL/ Concentratons of naturally occurnng isotopes of uranium, thonum, radium and radon were measured in freshwaters and in TM-9614. 33504:334. Geochemical information relevant to the retention of radionu- sodium-chionde bnnes near the site of the Waste isolation Pdot chdes by candidate high-level nuclear waste geologic repositor- Plant (WIPP) located in southeastern New Mexico. Supplemen-ies being charactenzed by Department of Energy (DOE) tal water chemistry analyses (chionde, afkahnity, P(CO2), CO(2), projects is being evaluated by Oak Ridge National Laboratory Fe, Mn. H(2)S) were made to aid in interpreting the data for nat-(ORNL) for the Nuclear Regulatory Commmsen (NRC). Dunng ural radionuchdes, Three features of radionuchde mobility are this report penod, emphasis was given to evaluaton of pub- evident from the results: 1) There is a shght tendency for U and lished sorpton and solubdity information for key radenuchdes - Ra concentratons to correlate with the chlonde content of the which is relevant to the Hanford Site in the Columbia River ba- water samples. Whether this tendency resufts from complexa-salts. The removal of neptunium from soluton by basaft/ ground- tion by Cl- ens or cation exchange competiton for adsorption water systems under anoxic redox conditions at 60 degrees sites cannot be resolved with the available information. 2) Much more dramatic than the correlation tydh Cl- concentraton is the centigrade proved to be sensitive to the basalt particle size and the test contact time. It was not possible to estabhsh if the nep- effect of the redox state of the waters on U and Ra concentra-tunium removal from solution was due to sorption or precipita- tions. Chemically reducing groundwaters contain much lower U tion processes. In studies of uranium solubehty, sodium boltwoo- concentrations and much higher Ra concentrations than were dite was shown to be the U(VI)- containing phase that precips- measured in omic and suboxic samples. Calculated retardation lates from synthetic groundwater at 60 degrees centigrade. The factors of 1 for Ra indicate that it can migrate freely in anoxic precipitation of sodium boltwoodite, rather than schoepite which bnnes 3) Low chemical recovenes of Th and to a lesser extent is pred:Cted by geochemical modeling, shows the importance of U were observed for methods that work well with seawater - - ~
Main Citations and Abstracts 85 samples. These elements may be present in a mobile, unreac- policy on overtime: 1) hmit personnel to 112 hours of work in a
tive dissolved or colloidal complex with organic matter. 14-day penod.192 hours in 28 days, and 2.260 hours in one NUREG/CR-4239

ANALYSIS OF THE ABILITY OF CURRENT year; exceeding these limits would require plant manager ap-
! MEALTH PHYSICS INSTRUMENTS TO PREDICT DOSE IN EX- proval,2) add a requirement that licensees obtain approval from i POSED INDIVIDUALS. ARMANTROUT,G.A. Lawrence Liver- NRC if plant personnel are expected to exceed 72 hours of more Nat:onal Laboratory. July 1985. 315pp. 6507250126. work in a 7-day penod,132 hours in 14 days, 228 hours in 28 UCID-20398. 31788.001. days, and 2,300 hours in one year, and 3) make the pohcy a i in this study, theoretical calculations of effective dose to body requiremont, rather than a nonbinding recommendation. tissue using Monte Carlo simulation techniques have been per, Second, it is recommended tSat hcensees be required to obtain formed for teoth gamma ray and beta ray irradiation. Similar cal. NRC approval to adopt a routine 12-hour / day shift schedule. culations for neutron irradia* ion by other workers have also Third, it is recommended that NRC actd several nonbinding rec-been reviewed. Evaluations were made of the performance of a ommondations concerning routine 8-hour / day schedules. Final-series of the more common health physics instruments. In this ly, because additional data can strengthen the basis for future evaluaton, representatue instruments for both gamma-ray and NRC pokey on overtime, five methods are suggested for collect-beta-ray survey work were evaluated using a senes of calibrated ing data on overtime and its effects. radiatiori sources. These instrument evaluations were then com-pared against similar evaluations in the literature, and an eval- NUREG/CR-4249: PRESSURE VESSEL FRACTURE STUDIES uation of basc instrument response by type was performed. In PENETRATING TO THE PWR THERMAL SHOCK addition, data on calculated health effects was used to evaluate ISSUE EXPERIMENTS TSE-5 TSE-5A AND TSE 6 the ability of these instruments to predict health effects. Key re- CHEVERTON,R.D.; BAl ' .D G.; BOLT,S E.; et al. Oak Ridge NaI sults for the gamma-ray, beta-ray, and neutron survey meters tional Laboratory July 1985. 255pp. 8507250152. ORNL 6163 are gwen. 31791:140. Thermal-shock expenments TSE 5, TSE-6 were conducted for NUREG/CR 4240 V01: PHYSICS OF REACTOR the purpose of investigating the behavicr of surface flaws under SAFETY.Ouarterly ReporLJanuary-March 19,
Argonne Na- pressunzed-water-reactor (PWR) overcoohng accident conds-tional Laboratory. July 1985. 27pp. 85086.0/79 ANL-85-23 tions. These expenments wer, the fifth, sixth, and seventh in a VC1, 31923:219. ,
senes of thermal-snock expenments conducted with large steel This quarterly progress report summanzes work done dunng cyhnders (A508, class 2 chemistry; 991 mm OD x 76- and 152-the months of January-March 1985 in Argonne National Labora- ...a, wall x 12 m length) as a part of the Heavy-Section Steel tory's Applied Physics and Components Technology Divisions Technology (HSST) Program for this purpose. For each of these for the Dwision of Reactor Safety Research in the U S. Nuclear expenments the initial flaw was on the inner surface and ex-Regulatory Commission. The work in the Apphed Physics Divi- tended the full length of the cylinder The thermal shock was sion includes reports on reactor safety modeling and assess-ment by members of the Reactor Safety Appraisals Section. apphed to the inner surface only, and this was accomphsbed by Work on reactor core thermal-hydrauhes is performed at ANL's effectively dunking the test cyhnder, initially at 93 degrees cente. grade, into a large volume of liquid nitrogen. Results of the ex-Components Tect'nology D.vison, emphasizing 3-dirnenssonal penments have confirmed that (1) knear-elastic fracture me-code development for LMFBR accidents under natural convec-chanics (LEFM) is vahd for thermal-shock loading. (2) crack tion conditions. An executive summary is provided including a arrest will take place in accordance with recently developed statment of the fndings and recommendations of the report-crack-arrest concepts, (3) the crack-arrest toughness values for NUREG/CR-4245: IN-PLANT SOURCE TERM MEASUREMENTS nsing and fa:hng K(I) fields are the same, (4) warm prestressing AT BPUNSWICK STEAM ELECTRIC STATION. DUCE S.W: 's effective in preventing crack init.ation, (5) thermal shock CRONEY,S.T.; AKERS,D.W.; et al. EG&G Idaho, Inc. (subs. of alone cannot dnve a flaw all the way through the wall, (6) dy-EG&G, Inc.). June 1985. 895pp. 8507020396. EGG-2392. namic effects for PWR vessel thermaLshock loading conditions 31311:061, are neghgible (7) in the absence of cladding and under severe This report presents data obtained at Brunsw>ck as part of thermal-shock loading conditions finite-length flaws will extend the IrdPlant Source Term Measurement Program in operating on the surface to become very long. and (8) there can be very hght water reactors (LWRs). The work was conducted for the large scatter in small-specimen fracture-toughness data Office of Nuclear Regulatory Research (RES) in support of the Meteorology and Effluent Treatment Branch (METB) of the NUREG/CR-4250: VEHICLE BARRIERS EMPHASIS ON NATU-Office of Nuclear Reactor Regulation (NRR). The pnmary oblec- RAL FEATURES. ADAMS.K,G; ROSCOE.B.J. Sandia National tive of this program is to provide the Nuclear Regulatory Com- Laboratones. September 1985 110pp. 8509300238. SAND 85-mission (NRC) with operational data that can be used en evalua- 0935. 32815 292. tion of plant designs for hquid and gaseous radwaste treatment The recent increase in the use of car and truck bombs by ter-systems. Data presented were obtained at the Brunswick Nucle- ronst organaations has led NRC to evaluate the adequacy of ar Generating Station, operated by Carohna Power and Light, lo- 16censee security against such threats. As part of this evaluation, cated at Southport, North Carohna. In-plant rneasuremens were one of the factors is the effectiveness of terrarn and vegetation conducted dunng the time penod from March 1982 to Novem- 'n providing barners against the vehicle entry. The effectiveness ber 1982. This plant is the sixth in a senes of operating LWRs of natural features is presented in two cor' texts. First, certain to be studied and the first boiling water reactor (BWR) in the natural features are presented in addeton to the discussion of Senes~ natural features, this report provides a discussion of methods to slow vehicles Also included is an overview cf man-made barner NUREG/CR 4248: RECOMMENDATIONS FOR NRC POLICY ON systems, with particular attention to ditches. SHIFT SCHEDULING AND OVERTIME AT NUCLEAR POWER PLANTS. LEWIS.P.M. Gattelle Memonal Institute, Pacific North- NUREG/CR 4251 V01: MITIGATIVE TECHNIOUES FOR west Laboratones. July 1985.150pp. 8508090710. PNL 5435. GROUND-WATER CONTAMINATION ASSOCIATED WITH 32102.343. SEVERE NUCLEAR ACCIDENTS Volume 1. Analysis Of Genenc This report contains the Pacific Northwest Laboratory's Site Conditions. OBERLANDER.P L; SKAGGS,A L; (PNL's) recommendations to the U S. Nuclear Regulatory Com- SHAFER J M Battelle Memonal Institute, Pacific Northwest Lab-mission (NRC) for an NRC policy on shift scheduling and hours oratones. August 1985. 321pp. 8509110279. PNL 5461. of work (including overtime) for control room operators and 32559 004. Other safety-related personnel in nuclear pcwer plants. First. it Pacific Nortl.*est Laboratory evaluated the feasibihty of using is recommended that NRC make three additions to its present grnund. water containment mitigation techniques to control rade.
o 86 Main Citations and Abstracts onuchde migraton following a severe commercial nuclear power NUREG/CR-4254: OCCUPATIONAL DOSE REDUCTION AND reactor accident. The two types of severe commercial reactor ALARA AT NUCLEAR POWER PLANTS. Study On HegrrDose tecidents envestigated are 1) containment basemat penetration Jobs.Radwaste Handkng,And ALARA incentives. DIONNE,B.J.; of core melt debns, which slowty cools and leaches radionu- BAUM,J W. Brookhaven National Laboratory. July 1985.104pp. clides to the subsurface environment; and 2) containment base- 85080>0697, BNL-NUREG-51888. 32105:192. mat penetration of sump water without full penetration of the The purpose of this report is to provide the NRC and the nu-core mass. Six genenc hydrogeologic site classifications were clear industry with information and data which will be useful for developed from an evaluation of reported data pertaining to the occupational dose reduction at nuclear power plants. The objec-hydrogeologic proporties of all existng and proposed commer- tives of this effort were to: 1. identify the repetitive high-dose cial reactor sites. One-dimensional radionuchde transport analy- fobs, related collective dose ranges and apphcable dose reduc-ses were conducted on each of the indnndual reactor sites to tion techniques,2. investigate and recommend improvements in determine the genenc charactonstics of a radionuclide dis- the selection of high reliability and low maintenance equipment charge to an accessible environment. Ground. water contain- to assure that collective doses received during equipment repair ment mitigaton techniques that may be suitable for severe is considered, 3. recommend improved radioactive waste han-power plant accidents, depending on specific site and accident dhng procedure and equipment which could reduce collective conditions, were identitied and evaluated. Feasible mitigative dose equivalent, and 4. examine current ALARA incentives and techniques and associated constraints on feasibihty were deter- recommend new positive steps which will provide additional mined for each of the six hydrogeologre site classifications. dose-reduction incentives. Ten nuclear sites were visited by two Three case studies were conducted at power plant sites located Brookhaven health physicists to collect the needed dose-reduc-tiong the Texas Gulf Coast and the Ohio River. Mitigative strat- tion data and information. This report summanzes the findings egres were evaluated for their impact on containment transport. and recommendations on the above objectives. Results show that the techniques evatuated significantly in-creased ground-water travel times and reduced contaminant rni. NUREG/CR-4255 V01: AEROSAL RELEASE AND TRANSPORT gration rates. PROGRAM SEMIANNUAL PROGRESS REPORT FOR OCTO-BER 1984 - MARCH 1985. ADAMS.R E.; TOBIAS,M.L. Oak NUREG/CR-4251 V02: MITIGATIVE TECHNGUES FOR Ridge National Laboratory. August 1985. 56pp 8509110010. GROUND-WATER CONTAMINATION ASSOCIATED WITH ORNL/TM-9632/VI. 32565 009. SEVERE NUCLEAR ACCIDENTS. Volume 2. Case Study Analy- This report summanzes progress for the Aerosol Release and sis Of Hydrologic Charactenzation And Mitigative Schemes. Transport Program sponsored by the Nuclear Regulatory Com-OGERLANDER.P.L; SKAGGS.R.L.; SHAFER.J.M. Battelle Me- mission Office of Nuclear Regulatory Research. Division of Ac-monal Institute, Pacific Northwest Laboratones. August 1985 cident Evaluation, for the penod October 1984. March 1985. 301pp. 8509100515. PNL-5461, 32538 094. Topics discussed include (1) steam only expenments in the See NUREG/CR-4215,V01 abstract. NSPP facility; (2) tests in small vessels to study thermal output, mass generation rates, and other operating features of plasma NUREG/CR-4252: INDEPENDENT ASSESSMENT OF TRAC- torch aerosol generators in support of the development of the PD2/ MOD 1 CODE WITH BCL ECC BYPASS TESTS. Aerosol Moisture interaction Tests (AMIT) facihty and to support SLOVIK G.C.: SAHA,P. Brookhaven National Laboratory. August the LWR Aerosol Containment Expenments (LACE) program at 1985. 78pp. 8509180076. 32668:213. Hanford, (3) analysis of data from plasma torch aerosol genera. This report presents the TRAC-PD2/ MOD 1 indepe-dent as- tor tests; (4) analysis of steam behavior in the NSPP vessel in sessment calculations performed at Brookhaven National Labo- aerosol expenments and in steam only tests, and (5) a study of ratory (BNL) using the Emergency Core Cooling (ECC) bypass the feasibihty of experiments for shape factor measurements. expenments cormcted in a 2/15-scale PWR vessel at Battelle Columbus Laboratones (BCL). Both steady state expenments NUREG/CR-4256: MEASUREMENT OF RESPONSE TIME AND with vanous ECC water subcoohngs and transient tests with hot DETECTION OF DEGRADATION IN PRESSURE SENSOR / wall effects were simulated. Besides the base cases, several SENSING LINE SYSTEMS. BUCHANAN.M E.; MILLER.L.F.; sensetivity calculations were performed to study the effects of THIE J.A.; et al, Oak Ridge National Laboratory. September nodalization, particularly the relative locations of the hot leg 1985 107pp.8512270230. ORNL/TM-9574. 34080 258. penetrataors in the downcomer, in addition, calculations were A team evaluated several methods for remote measurement performed to determine the effect of shg.' ecruses in the rr- of the response time and dotaction of degradation (blockage or verse core steam flow and the assctiated form losses due to air in lines) of pressure sensor /scnsing hne syrtems typical of the hot leg penetrat.ons. Code corrections as received from the nuclear power plants A method was developed for obtaining code developers at Los Alamos National Laboratory (LANL) the response t;me of force-balance pressure transmit' ors by were also incorporated into this study. bnetly interrupting the power supply to the transm *.ler. The data thus generated are then analyzed in conjunction with a model to NUREG/CR-4253: REVIEW OF TRAC CALCULATIONS FOR predict transmitter response to an actual pressure porturbat on. CALVERT CLIFFS PTS STUDY. JO,J H.; ROHATGl,U S. Brook. The research team also evaluated a pressure perturbation haven National Laboratory. Apnl 1985.114pp. 8511010391- method for determining the asymptotic delay time of a pressure-BNL-NUREG-51887. 33300:231. sensing hne and found that this method yields accurate results Six selected transient calculations out of thirteen performed for essentialy unblocked sensing hnes. However, these pressure by LANL using the TRAC-PF1 code for the USNRC PTS study perturbation tests are not recommended for use in nuclear of the Calvert Chffs Nuclear Power Plant have been reviewed in power piants because they to difficult to emplement on-line. A depth at BNL Simple hand calculations based on the mass and third method for remote measurement applied noise analysis energy balances have been performed to predict the tempera- methods that yieldod accurate estimates of asymptotic delay ture and pressure of the reactor system, and the results have times for blockage or air in sensing knes. Even though noise been compared with those of TRAC. Companson was also analysis methods worked well in the laboratory, it is recom-made between the TRAC and RETRAN calculations for two of mended that further evaluation be performed in operating nucle-these transients, which were performed by ENSA. In general, ar plants. the results calculated by TPAC appear to be reasonable based on the comp 3nson with RE'RAN and hand calculations.
l l Main Citations and Abstracts 87 NUREG/CR-425h INSPECTION. SURVEILLANCE.AND MONi- NUREG/CR 4260: TORAC USER'S MANUALA Computer Code TORING OF ELECTR! CAL EQUIPMENT INSIDE CCNTAIN- For Analyzing Tornado-Induced Flow And Matenal Transport in MENT OF NUCLEAR POWER PLANTS-WITH APPLICATIONS Nuclear Facilities. ANDRAE,R W.; TANG.P.K.; MARTIN,R A.; et TO ELECTRICAL CABLES. AHMED.S.; CARFAGNO,S.P. al. Los Alamos Scientific Laboratory. July 1985. 145pp. ARVIN/CALSPAN Advanced Technology Csnter. August 1985. 8507250118. LA 10435 M. 3t 787.001. 101pp. 8509100530. 32536.234. Inis manual desenbes the TORAC computer code, which can The general concepts of equipment condition monitonng as model tornado-induced flows, pressures, and matenal transport cpplicable to the detection of age.related detenoration of within structures. Future versions of this code will have im-safety-related equipment are desenbed. The goal is to detect proved analysis capabilitics. In addition, it is part of a family of detenoration in the incipient stage, pnor to in-service failure and computer codes that is designed to provide improved methods pnor to the point at which equipment can no longer be expect, of safety analysis for the nuclear industry. TORAC is directed ed to perform its function when exposed to design basis acci- toward the analysis of facihty ventilation systems, including dent conditions. The application of condition monitonng is dis- 'nterconnected rooms and corndors. TORAC is an improved cussed specificalfy for electncal cables. The goal of cable con- version of the TVENT code. In TORAC, beowers can be turned dition monitonng is to determine the degree of cable degrada- on and off and dampers can be controlled with an arbitrary time tion and to predict the remaining useful hfe. In situ nondestruc- nction. The matenal transport capabikty is very basic and in-tive testing and destructive laboratory testing are discussed. in- cludes convection, depletion, entrainment, and fdtration of mate-tenm recommendations are given for the implementation of a nal he W specifications for the code and vanety of sample condition monitonng program. problems are provided. NUREG/CR-4258: AN APPROACH TO TEAM SKILLS TRAINING NUREG/CR 4262 V01: EFFECTS OF CONTROL SYSTEM FAIL-OF NUCLEAR POWER PLANT CONTROL ROOM CREWS- URES ON TRANSIENTS AND ACCIDENTS AT A GENERAL DAVIS L.T.; GADDY,C.D.; TURNEY,J A General Physics Corp. ELECTRIC BOILING WATER REACTOR Main Report. July 1985. 83pp. 8508200093. GP-R 123022. 32278.020- BRL)bKE,S.J ; BAXTER.D E.; RANSOM.C B ; et al EG&G Idaho, Inc. (subs. of EG&G, inci May 1985. 81pp 8506240180. EGG-An investigation of current team skills training practices and 2394. 31150:280. rssearch was conducted by General Physics Corporation for the This report documents the evaluation of the effects of nonsa-Oifice of Nuclear Reactor Regulation. The methodology used in-cluded a review of relevant team skills training hterature and a fety grade control system fadures on a typical boshng water re-actor plant. The methods utilized in this evaluation include a workshop to collect inputs from team training practitioners and system level failure modes and effects analysis, deterministic rzsearchers from the public and pnvate sectors. The workshop computer analysis, a review of 3 years of recorded plant occut-was attended by representatives from nuclear utshty training or- rences, a probabihty analysis and a review of apphcable NRC ganizations, the commercial airhne industry, federal agencies, entena pertaining to control systems. This study identified three and defense training and research commands. The hterature re-system failures that could cause transients leading to a reactor views and workshop results provided the input for a suggested vessel overfdl and of these three failures, two could also lead to approach to team skdis training that can be integrated into en- a reactor Coolant cooldown of greater than 100 degrees fahren-ating training programs for control room operating crews. The heit per hour. This study concluded that the existing NRC cnte-rpproach includes five phases. (1) team skills objectives deves. na, conceming control systems, adequately address the poten-opment, (2) basic team skills training, (3) team task training, (4) tial problem areas that were identified dunng this evaluation. team sMis evaluaticn, and (5) team training program evaluation. Based on the results of this study, it was recommended that the Supporting background information and a user.onented descnp. consequences and nsk associated with overfill and overcool tion of the approach to team skills training are provided. transients be further investigated NUREG/CR-4259; TAILINGS NEUTRAllZATION AND OTHER NUREG/CR 4262 V02: EFFECTS OF CONTROL SYSTEM FAIL-ALTERNATIVES FOR IMMOBILf2lNG TOxlC MATER!ALS IN URES ON TRANSIENTS AND ACCIDENTS AT A GENERAL TAILINGS. Final Report. CFITZ.B E.: SHERWOOD.D R ; ELECTRIC BOILING WATER REACTOR. Appendices. DODSON,M E.; et at Battelle Memonal Institute, Pacific North. BRUSKE.S J; BAXTER.D E.; RANSOM.C B ; et at EGAG Idaho, wsst Laboratones. September 1985.131pp. 8510040347. PNL. Inc. (subs. of EG&G. Inc ) May 1985. 34tpp. 8506240689 5467. 32857:159. EGG-2394. 31179 036 This final document, in a senes of six, summanzes research Safety implications of Control Systems (A 47) was approved completed since the beginning of the protect. Three subtasks as an Unresolved Safety issue (USI) by the Nuclear Regulatory tre included. Subtask A - Neutrahzation Methods Selection; Commission (NRC) in December of 1980. USI A.47 concerns Subtask 8 Laboratory Analysis, and Subtask C Field Testing the potential for transients or accidents being made more Subtask A reviews treatment processes from other industnes to severe as a result of control system failures. This report de. , evaluate if current waste technology from other fields is apphca- scnbes the work performed on the effects of control system fad. ble to the uranium industry. This task also ident<fies several rea- wes on transients and accidents at a General Electnc boikng gents that were tested for their effectiveness in treating acidic water reactor. This work was conducted for the U S. Nuclear tishngs and taihngs solution in order to immobihze the contami- Regulatory Commission. Division of Safety Technology by rttnts associated with the acid waste Subtask B dascribes the * * * * * 11borato'Y batch and column treatment studies performed on Plant This report is contained in two volumes, a rnain report sohd waste taihngs and tarbngs solution over the course of the and five appendices. The main report desenbes the study meth-pro}ect. The evaluation of several reagents identified in Subtask odology, the major areas of work performed, and the results A was based on three entona- 1) treated effluent water quality- and conclusions The appendices contain dotaded information
~~2) neutralized sludge handhng and hydrauhc properties, and 3) consisting of failure mode and effects analysis tables, e detaded reagent costs and acid neutrakring efficiency. Subtask C pre~~
desenption of the computer analyses and signi'ica ' * *nsient encerpts sents a field demonstration plan that will evaluate the effecti-venss, costs and benefits of neutralizing acidic uranium mile tad-ings solution to reduce the potential teaching of toxic trace NUREG/CR-4263: RELIABILITY ANALYSIS OF SilFF VERSUS FLEXIBLE PIPING FINAL PRO JEC1 REPORL LU,S C ;
metals, radionuchdes and macro ions from a taihngs impound- CHOU.C K. Lawrence Livermore National Labor tory May 1985. rnent. 78pp 8505280080. UCRL 20410. 30604169
88 Main Citations and Abstracts This research project is to develop a technical basis for flexi- NUREG/CR-4267: VESSEL INTEGRITY SIMULATION (VISA) ble piping designs which will improve piping reliability and mini- CODE SENSITIVITY STUDY. SIMONEN.E P.; JOHNSON.K.I.; maze the use of pipe supports, snubbers, and pipe whip re- SIMONEN,F.A. Battelle Memonal Institute, Pacific Northwest straints. This study indicated that piping design can be made Laboratones. December 1985. 49pp. 8601070522. PNL-5469. more reliable by some reduction of ngid supports and/or snub- 34186.043. bers. This study also conismed that the malfunction of pipe in a study conducted for the Nuclear Regulatory Commissen whip restraints introduced higher thermal stresses and tended by Pacific Northwest Laboratory, the sensitivity of through-wall to reduce the overall piping rehabihty. Finalty, our results indicat- crack probabihty to input distnbutions was stud.ed. Flaw growth charactenstics were evaluated for three pressunzed water reac-ed that supports in a flexible piping design may need to be re-evaluated and that the elimination of pipe supports which are tor plants (Oconee 1. Calvert Chtts 1, and a hypothetical plant similar to H. B. Robinson 2). Three postulated pressunzed ther-close to components should be done with care in order to min,- mal shock (PTS) transients were considered for each plant. This maze the impact on the component rehability. report desenbos the results of matenal and flaw distnbution as. sumptons on calculated conditional failure probabilities for the NUREG/CR 4264: INVESTIGATION ON HIGH EFFICIENCY PAR. pred'cted sensitivites are evaluated and are related to require-TICULATE AIR FILTER PLUGGING BY COMBUSTION AERO-SOLS. FENTON.D L.; GREGORY,W S.; et al. Los Alamos Scien. ments for dehning snput distnbutions for probabilistic failure pre-tific Laboratory GUNAJi,M.V. New Mexico State Univ., Las dictons. es M. May 1985. 32pp. 8507050422. LA-10436 MS. NUREG/CR-4268: RATIO METHODS FOR COST-EFFECTIVE FIELD SAMPLING OF COMMERCIAL RADIOACTIVE LOW-Expenments were conducted to investigate high-efficiency LEVEL WASTES. EBERHARDT,LL.; SIMMONS M A; particulate air (HEPA) filter plugging by combustion aerosols. THOMAS.J.M. Battelle Memonal institute, Pacific Northwest These tests were done to obtain empincal data to improve nur Laboratones. July 1985. 81pp. 8508090570. PNL 5156 modeling of filter plugging phenomena using the Los Alamos 32106.069. National Laboratory fire accident analysis code FIRAC. Com- An investigaton of cost-effective methods for samphng at mercially available 0.61 m by 0.61 m square filters were tested commercial radioactive low-level waste sites has been one goal in a specially designed facihty to determine how airflow resist- of this protect. To that end, double samphng was investigated, ance vanes with increased f. Iter loading by combustion aero- and we found that the method appears useful when est. mating sols. Two organic fuels normally found in nuclear fuel cycle fa- total radionuclide inventory in waste site environs. The methods cilities, potystyrene (PS) and polymethylmethacrylate (PMMA)- are explained, decision critena for cost effectiveness presented, were burned under vaned conditons to generate combuston and a worked example based on field data is provided. The sta-aerosols. The test facihty included a combustor, a 23-m-lon9 tistical basis for the conclusion that dout"e samphng appears to duct, and a specialty designed gravimetnc balance for determ- be robust and cost effective is in separate sections Field tests ing the aerosof mass gain of the fdters Test results include cor- and add,tional estimates of " field instrument" errors are needed relatons of HEPA filter resistance ratios (actual resistance /ini- to substantiate the findings. tial resistance) with aercsol mass gain. The mass gain of plugged HEPA filters was found to correlate with the airborne NUREG/CR 427t: RECOMMENDED mass concentration of material in the size range greater that SAFETY, RELIABILITY OUALITY ASSURANCE AND MANAGE-approximately 2.0 m. Also, the fuel with a smaller soot fracton, MENT AEROSPACE TECHNIOUES WITH POSSIBLE APPLICA-PMMA, produced fdter plugging at lower accumulated aerosol TION BY THE DOE TO THE HIGH LEVEL RADIOACTIVE WASTE REPOSITORY PROGRAM. BLAND,W M. Gee 9'n, Inc. mass deposits on or within the filter, June 1985.113pp. 8507080205,31393.164. Aerospace SROA and management techniques, pnncipally NUREG/CR-4266: STANDARD BETA-PARTICLE AND MONOEN. those developed and used by the NASA Lyndon D. Johnson ERGETIC ELECTRON SOURCES FOR THE CAllBRATION OF PROTECTION INSTRUMENTATION. Space Center on the manned space thght programs, have been GETA-RADIATION EHALICH,Ma PRUITT,J S.; SOARES.C G ; et al. Commerce, assessed for possible apphcation by the DOE and the DOE-Dept. of, National Bureau of Standards August 1985 86pp. contractors to the high level radioactive waste repository pro-gram that results from the implementation of the NWPA of 8509060202. N8 SIR 85-3169. 32505.057. 1982. Those techniques believed to have the greatest potential in a protect funded jointly by the National Bureau of Stand. for usefulness to the DOE and the DOE-contractors have been ards (NBS) and the Nuclear Regulatory Commission (NRC). d'scussed in detail and are recommended to the DOE for adop-NBS has developed a cahbration facihty for beta-particle instru- tion, discussion is provided for the manner in which this transfer ments and sources used in radiation-protection dosimetry. The of technology can be implemented Six SROA techniques and facihty consists of beta.part#cle and nearly monoenergetic elec- two management techniques are recommended for adoption by tron beams charactenred in terms of absorbed-dose rates to the DOE; included with the management techniques is a recom-plastic and in terms of beta-particle spectra. A second phase of mendation for the DOE to include a kconsing interface with the the project was concemed with estabbshing secondary cahbra- NRC in the apphcation of the milestone review technique. tion laboratones for radiation-protection instruments. This final These other techniques are recommended for study by the DOE report includes a detailed discussion of (1) the determination of for possible adapton to the DOE program. absorbed-dose rates to plastic for each Deta-particle and nearly monoenergetic electron beam, dose rate dependence on alti- NUREG/CR-4272: RESPONSE TREE tude above sea level, and an estimate of the overall uncertain. EVALUATION EXPERIMENTAL ASSESSMENT OF AN EXPERT tres in dose-rate measurements, (2) beta particto and nearly SYSTEM FOR NUCLEAR REACTOR OPERATORS monoenergetic electron spectra and their dependence on NELSON.W R ; BLACKMAN.H S. EG&G idaho, Inc. (subs. of source configuraten; and (3) degree of achievable uniformity of EG&G. Inc ) September 1985. 08pp. 8510040397. EGG 2397. beam cross sections. Included also is a revew of the results of 32859 203. a first attempt to predict enstrument response to reahstic beta- The United States Nuclear Regulatory Commission (USNRC) particle environments from therir response to rnonoenergetic sponsored a project performed by EGaG Idaho, Inc., at the electrons and knowledge of the approximate beta particle spec- Idaho National Engineenng Laboratory (INEL) to evaluate differ-tra. Attached to the report are proposed guidekneo for estabhsh- ent display concepts for use in nuclear reactor control rooms. ing secondary cabbration laboratones for radiation protection in- Included in this project was the evaluation of the response treo
~~.truments. computer-based decison aid and its associated displays The~~
Main Citations and Abstracts 89 response tree evaluation task was designed to (a) assess the sible, based on the vibrations characten2ed in the flow tests. ment of the response tree decision aid and (b) develop a tech- The overall protect status, including work completed to date and nical basis for recommendations, guidelines, and entena for the tasks planed for the remainder of FY85, is also documented. ! design and evaluaten of computenred decision aids for use in I reactor control rooms. Two major expenments have been con. NUREG/CR 4271: INVERTED ANNUAL FLOW EXPERIMENTAL l ducted to evaluate the response tree system. This report em. STUDY. DE JARLAIS.G; ISHil.M. Argonne National Laboratory phasizes the conduct and results of the second expenment. An Apnl 1985.115pp. 8507050406. ANL 85-31. 31338 074 enhanced version of the response tree system, known as the Steady state inverted annular flow of Freon 113 in up flow automated response tree system, was used in a controlled ex. was estabhshed in a transparent test section. Using a special penment using trained reactor operators as test subjects. This inlet configuraton consisting of long aspect-ratio hquid nonles report discusses the automated response tree system, the coaxially centered within a heated quart 2 tube, ideGred invert-design of the evaluation expenment, and the quantitative results ed annular flow initial geometry (cylindncal hquid core surround-of the expenments. The results of the expenments are com. ed by coamal annulus of gas) could be estabbshed. Inlet bquid pared tc the results of the previous expenments to provide an and gas flowrates, hquid subcochng, and gas density (using van-integrated perspective of the response tree eva!uation project. ous gas species) were measured and vaned systematically The In addition, a subjective assessment of the results addresses hydrodynamic behavtor of the hquid core, and the subsequent the implicatons for the use of advanced " intelligent" decision downstream break-in of this core into slugs, ligaments and/or aids in the reactor control room droplets of vanous sizes, was observed. In general, for low inlet u NUREG/CR-4274: ANALYSIS AND TESTS ON SMALL SCALE c s it was se that a r the instat fematon of SHEAR WALLS FY 82 FINAL REPORT. ENDEBROCK.E.G '- r it waves n the hquid core surface, an ag)tated region of high DOVE.R CJ DUNWOODY,W E. Los Alamos Scientific Laborato surface area. with attendant high momemtum and energy trans-ry. September 1985 59pp. 8512270234. LA 10443-MS- fas, occurs. This agitated region appears to propagato down. 34084 109 stream in a quasopenodic pattern ins eased inlet liquid flow The Phase-f expenmental program was completed dunng FY r s. a gas annulus rams W to diminish the sig. 1982. This report summarizes the results of (1) quasistatic aWMh (monotonic and load-cycling) tests, (2) sinusoidal vibration tests' ""*# and (3) simulated earthquake tests conducted on small scafe, #" ** ## ' 0"' twinforced-concrete shear walls. Model construction, test meth-
9 ods, instrumentation, and expenmental results are presented in this report. Experimental results are interpreted to investigate NUREG/CR-4278: TRAC PF1/ MOD 1 DEVELOPMENT ASSESS-MENT. SAHOTA,M S; ADDESSIO.F L Los Alamos Scientific the effects of high-load levels (which produce cracking and fail- Laboratory. August 1985, 350pp. 8511010462. LA 10445-MS.
ure of the whils) on stiffness, damping, and on deformation and 33302 268. acceleraton transmissibihty. The noninear analysis method that has been developed as part of this program has been used to The Transient Reactor Analysis Code (TRAC) is being devel. aid in the interpretation of these expenmental results. oped at Los Alamos National Lavoratory to provide advanced best-estimate predictons of postulated accidents in light water NUREG/CR 4275: HEAVY SECTION STEEL TECHNOLOGY reactors The TRAC-PF1/ MODI program provides this capabil-PROGRAM FNE YEAR PLAN FY 1984 1988.
Oak Ridge sty for pressunzed water reactors and for many thermal hydrau-National Laboratory, August 1985.160pp. 8509110024. ORNL/ lic expenmental facilities. The code features either a one or TM 9654. 32558 001. three-dimensional treatment of the pressur vessel and its assa-The second in an annual senes of five-year program plan ciated internals, a two-phase, two-fluid, nonequihbnum hydro-documents is presented for the Heavy Secten Steel Technolo- dynamics model with a noncondensable gas field, flow-regime-gy program. The program is carned out by the Oak Ridge Na- dependent constituSve equation treatment. optional reflood-tonal t ar. oratory for the Matenals Engineenng Branch Division tracking capability for both bottom flood and falhng-film quench of Ens,ineonng Technology. Office of Nuclear Regulatory Re- fronts; and consistent treatment of entire accident sequences search of the U S. Nuclear Regulatory Commission. The pro- from nomal operat ng conditions through severe trasients A gram is aimed at advancing the understanding and vahdaten of new numencal algonthm is used in the one-dimensional hydro-matenals and structures behavior as they relate to hght water dynamscs that permits this portion of the fluid dynamics to vio-reactor pressure vessel integnty. The program has nine techne late the matenal Courant condition. This technique permits large cal tasks and a management function. A background statement time steps and, hence, reduced running time for slow transients end a plan-of-action is given for each. The nine technical tasks This report presents the results of initial developmental assess-address fracture methodology and anatysis, matenals character- ment calculations performed with TRAC-PF1/ MOD 1 before its ization, crack growth, crack arrest, irradiation effects, cladding pubhc release. The assessment set consists of sin integral ef-evaluations, intermediate vessel testing. thermal-shock testing. 1ects calculations in the Loss of Fluid test and Semiscale facih.
and pressunzed thormal-shock expenments. bes. Computer run times required to predict each test also are reported. NUREG/CR-4276: VIBRATION AND WEAR IN STEAM GENERA. TOR TUBES FOLLOWING CHEMICAL CLEANING - SEMIAN- NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERI. NUAL REPORT. ENDERLIN Wl; BAUGH.J.W. Battelle Memon-ENCE AND ASSISTANCE OF A SHIFT TECHNICAL ADVISOR at institute. Pacific Northwnst Laboratones June 1985. 38pp. (STA) ON CREW PERFORMANCE IN CONTROL ROOM SIMU-8507030714. PNL 5477,31322 273. LATORS. BEAF E.A N ; DONOVAN,M Da LASSITER.D L; et al. The Pacific Northwest Laboratory is studying the effects of in- General Physics Corp September 1985 202pp. 851000318 creased tube / tut;e-support uearances in pressunted water re- ORNL/TM.9660. 32861049 actor steam generators folicw'ng chemical Clean ng The protect This report descobes the second expenment using a training purpose is to provide NRC with cntena for evaluating licensees' simulator to evaluate effects of expenence level of Senior Re-specific proposals for chemical cleaning of steam generators actor Operators (SRO) in the supervisor's role, and presence of This report descnbes the test and data analysis plans and pro- a Shift Technica Advisor (STA) on performance of nuclear cedures for the flow and accelerated wear tests to be per. power plant control room operators / crews The emperiment was formed in a scalo modol steam generator. The flow tests will es- conducted in a pressunzod water reactor (PWR) plant-refer-tablish the forcing boundary conditions, using clearances repre-enced simulator Data was collected on 20 three man crews of senting varcus conditions following chemical cleaning The ac- hcensed operators performing four sequences Performance celerated wear tests will deterrmne the potential wear rates pos- measures were denved from task analyses of the sequences
90 Main Citations and Abstracts One set of measures focused on task performance; the second in order not to preludice future investigations, all three damage set measured control of system parameters. Instructors' ratings parameters F > 10 MeV, F > 0.1 MeV and dpa will be listed in End performance scores of trainees were compared to scores this report. This NUREG will contain the t'eutron exposure ga-of operators / crews, to validate the performance measures. rameters for the 12 metallurgical specimen capsules which System parameters and control manipulations were recorded by compnse the Fifth HSST Irradiation Series. In order to mske the simulator's computer. Communications and selected venfi- available the data in a timely manner and not to delay the anal-cations were recorded on checkhsts and videotapes. Question- ysis of capsulcs irradiated early in the senes, it was decided to naires recorded biographical information and self-reported work- make this NUREG a looseleaf document. The exposure values loads. No significant differences in overall performance were will be distnbuted as they become available. found attnbutable to expenence of supervisors, nor to presence of a STA. Results were similar to results of an earlier expen- NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING ment performed with boiling water reactor (BWR) crews, These IN LIGHT WATER REACTORS. Annual Report,0ctober 1983 . results are also reported. Reported workloads of supervisors as- September 1984. SHACK,W.J.; KASSNER,T.F.; MAlYA P.S.; et l sisted by STA/s were t gnificantly lower than workloads report- al. Argonne National Laboratory. August 1985. 149pp. j ed by those unassisted. 8508210430. ANL 85-33. 32338 265. This progress report summanzes work performed by the Ar-NUREG/CR-4281: AN EMPIRfCAL ANALYSIS OF SELECTED gonne National Laboratory and a subcontractor, E F Rybicki, NUCLEAR POWER PLANT MAINTENANCE FACTORS AND Inc., on environmentally asssted cracking in light water reactors PLANT SAFFTY. OLSON.J; OSBORN,R N.; THURBER,J.A.; et dunng the twelve months from October 1983 through Septem-al. Battelle Human Affairs Research Centers. July 1985. 62pp_ ber 1984. 8508150065. PNL-5487. 32196 224. This report contains a statistical analysis of the relatonship g g between selected aspects of nuclear power plant maintenance GINIA AND EASTERN TENNESSEE EARTHOUAKES (1978-programs and safety related perf^rmance. The report identifies 1984). DILLINGER.G A ; TEAGUE,A.G ; MUNSEY,J.W.; et al-a large number of maintenance resources which can be expect. Virginia Polytechnic Institute & State Unrv., Blacksburg. VA. ed to influence maintenance performance and subsequent plant June 1985. 91pp. 8508010243. 31928 001. safety performance. The resources for which data were readily Focal mechanisms are presented for 11 carthquakes from the available were related statistically to two sets of performance in. Giles County, Virginia, seismic zone and its vicinity and for 12 dicators: maintenance intermediate safety indicators, and final eathquakes from the Central Virginial seismic zone. These safety performance indicators. The resuits show that the admin. earthquakes (0<M<4) were monitored by local networks be-istrative structure of the plant maintenance program is a signifi. tween January 1978 and October 1984. In Giles County, the cant predictor of performance on both sets of mdicators. data base consists of 43 P wave polarities and 50 SB to P am-plitude ratios (SV/P) that yielded w single event focal mechan-NUREG/CR 4283: STUDY OF THE EFFECTS OF ELASTIC ims LOADINGS ON THE JI.R CURVES FROM COMPACT SPECI UN'.(SEFM's) and five composite event focal mechanisms MENS. SUTTON G E.; VASSILAROS.M.G David W. Taylor (CFM's) In Central Virginia, 79 P-wave polanties and 51 SV/P Naval Research & Development Center. June 1985. 50pp. ratios are used to determine 11 SEFM's and four CFM s. A 8506260731.31227:118 computer program FOCMEC was used to determine the focal An investigation was performed to evaluate the efforts of mechanism solutions. The results for the Giles County seismic elastic unloadings on the J. Integral Resistance Curves of ASTM zone show mainly strike-slip mechanisms on steeply dipping (73 A106 Class C steel and 3 Ni steel. Compact specimens (IT) degrees plus minus 16 degrees) NNE (nght lateral motion) and were tested using a multi-r.pecimen technique, direct current po- ESE (left lateral motion) trending nodal planes However, some tentral drop technique and the elastic unloading compliance (4/11) counterclockwise, The P aves in Central Virginia are gen-technique with unloading ranging from 10 to 90% The tw erally northeast trending for sha; low earthquakes (>8 km) and former techn> ques were 0% unloading procedures used to gen- northwest trending for deeper ones (<8 km), in Giles County, erate the reference J R curves for companson to the elastic un- where the seismic activity is occurnng beneath the Appalachian toading J-R curves for the two steels The results of the investi- dewlement, faulting and inferred stress onentations are more gations of these matenats indicate that there was no significant uniform than in Central Virginia. sorr,e 200 km away, where the difference in the J-R curves that resulted from the elastic un- seismicity is occumng near and above the decollonent, loading compliance technique. NUP' G/CR 4290 V02: PROB ABILITY OF PIPE FAILURE IN THE NUREG/CR-4284: NEUTRON EXPOSURE PARAMETERS FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIA. REACTOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR PLANTS Volume 2 Guillotine Break Indirectly induced By TION SERIES. STALLMANN.F.W ; KAM.F 8; DALDWIN.C A. Earthquakes. RAVINDRA,M K.; CAMPBELL,R D , KIPP T R ; et Oak Ridge National Laboratory August 30, 1985. 39pp. & Lawrence Livermore National Laboratory. July 1985.146pp 8509110035. ORNL/TM-9664. 32564 337. The Nuclear Regulatory Commission's (NRC's) Heavy Section 8508010757. UCRL 53644. 31924 00f. Steel Technology (HSST) Program is concerned with the inves- The requirements to design nuclear power plants for the ef-fects of an instantaneous double ended guillotine break (DEGB) tigation of cracklike flaws in reactor pressure vessel steels, in of the reactor coolant loop (RCL) piping have fed to excessive the fifth irradiation senes, capsules containing a vanety of met. abrgical test specimens were tradiated to fluences in the design costs, interference with normal plant operation and main-range of 1.10(19) to 3,10(19) nuetrons/cm(2) (E > 1.0 MeV). In tonance, and unnecessary radiation esposure of plant mainte-order to correlate radiation embnttlement to damage fluences, nance personnel. This report describes an aspect of the NRC/ accurate determination of the neutron fluence spectra at the Lawrence Livermore National Laboratory sponsored research cntical location of the test specimen is needed The part of the program aimed at demonstrating that the probability of DLGB in neutron spectrum which is responsible for the radiation damage RCL Piping of nuclear power plants is acceptably small ar.J the is charactenzed as " damage exposure parameter." Fluences for requirements to design for the DEGB effects (o g , provision of energ es greater than 1 g MeV (F > 1.0 MeV) is the most widely pipe whrp restraints) may be removed This study estimatos the used parameter; however, current thinking favors displacements probability within the containment of Babcock & Wifcon supplied per atom (dpa) in iron as better related to the physical rnecha- pressunzod water reactor nuclear power plants in the United nism of radiation damage. Fluences for energies greater than States The medium probability of indirect DEGB was estimated to range between 610( 11) and 1x10(.7) per fear. Using very O f MeV (F> O f MeV) are also considered since neutrons in conservative assumptions, the 90*w subjective probabetity value the 0.1 to 10 MeV range are likely to contribute to the damage.
Main Citations and Abstracts 91 (confidence) of P(DEGB) was found to be less than 1x10( 5) per year, derstood. The recommendatons of vanous radiaton advisory groups are a'so summarized and a cntical review of pubbshed NUREG/CR-4291: CONCLUSION AND
~~SUMMARY~~
REPORT ON expenmental data applicable to entremity dosimetry is present-PHYSICAL BENCHMARKING OF PIPING SYSTEMS. ed. These recommendatons and data are used 'o identify the BEZLER,P.; 5UBUDHl.M.; SHTEYNGART,S.; et al. Brookhaven radiosensitive tissues in the extremities and rara the relative National Laboratory. September 1985. 105pp. 8509300512. risk to each organ or tissue. The results of labo,atory and field BNL-NUREG 51897. 32792:322. measurements of extremity dose to personnel engaged in work Physical benchmark evaluations were used to assess the ac- typical of different classes of heensees are presentod. The h-curacy and adequacy of the anafysis methods and assumptions consee classes include fuel fabocation facilities, pressunzed used in typical piping quakfication evaluations. Physical bg.nch- water reactors, boshng water reactra, medical radioisotopes op-mark evaluatens have been completed for six systems involving erations, and well-logging operations. Finally, recommendations both laboratory and in situ tested piping. In each evaluaten and guidelines for several specific situatons are provided to elastic finite element methods were used to predict the time his_ help assist the practicing health physicist with extremity dosime. tory response of a system for which physical test results were try. avanablo. In the analytical simulations the measured support ex~ citations and the measured damping properties were used as NUREG/CR 4298: DESIGN AND INSTALLATION OF COMPUTER input and the acceleration and displacement response of piping SYSTEMS TO MEET THE REOUIREMENTS OF 10 CFR 73.55 intenor points were predicted as output. The linear anafysis LEWIS J R.; BYERS.K.R.; FLUCKlGER.J D.; et al Battolle MeI methods were found to provide reasonable estimates of system monal Institute, Pacific Northwest Laboratones. July 1985' response. For a near linear system and using conservative esti- 120pp. 8507250138. PNL-5490. 31794 221 mates for system damping, a good correlation of response p O'Y g 9 traces and acceptable estimates of response peaks can be ex- InstaHaton of computer managed systems that can help nuclear pected. Using realistic estimates of uniform system damping. power plant licensees to meet the physical sesunty require-large underestimates of peak response components were ob- ments of 10 CFR 73 55 (for access control, alarm monitonng, served and deviations of 100*. or greater should be expected and alarm recording) Two ob;ectives were to study the power plant secunty functions that could be aided by a ccmputer-man-NUREG/CR 4292: A COMPARATIVE ANALYSIS OF CONSTITU-TIVE RELATIONS IN TRAC PFL AND RELAPS/ MOD 1. aged physical secunty system and to evaluate the safety and ROHATGI,U S ; JO.J.H.; SLOVIK,G C. Brookhaven National Lab- secunty considerations of such a system. A further obloctive oratory. June 1985.104pp. 8601070499. BNL NUREG 51898.. was to develop guidance on system design, selecten, and in-34183.279- ' stallation. The design guidance includes safety and seconty re-The purpose of this document is to desenbe the basic ther- querements, design alternatives, computer secunty, workspace mal hydraulic models and correlations that were used in the design and user interface design. Guidance is also provided on TRAC-PF1 (Verson 7.0) and RELAP5/ MOD 1/ CYCLE 14 codes. wnting a system specification for procurement, bid review proce. dures, and site preparation. Concerted efforts have also been made to assess the models desenbod 6n the code manuals and to compare them with their FORTRAN verssons in the code. Some discrepancies between NUREG/CR-4300 V01: ACOUSTIC EMISSION / FLAW RELATION. the documentation and the code, some errors in the models. SHIP FOR INSERVICE MONITORING OF NUCLEAR PRES. and vanety of constraints on the models were found and have SURE VESStas Progress Report. October March 1985. been reported here. Comments based on 8NL expenence with HUTTON,P.H ; KURTZ.R.J Battelle Memonal Institute, Pacific Northwest Laboratones. August 1985. 27pp 8508210022. PNL. TRAC-PF1 and RELAPS/ MOD 1 assessment have been made. 5511. 32302.232. The text contains many FORTRAN vanables in order to help Technscal progress in developing continuous acoustic emis-the readers who might be interested in modifying these codos. sion (AE) monitonng of nuclear reactor pressure boundanes for A table companng the constitutive relatonships in these two codes is also presented. flaw detection is discussed in this report The penod covered as October 1,1984, to Apol 1,1985. Topics include final anas yss NUREG/CR 4294: LEAK RATE ANALYSIS OF THE WESTING- of 281 vessel test data, preparaton for continuous AE monitor. HOUSE REACTOR COOLANT PUMP. BOARDMAN.T.: ing of Watts Bar Unit I reactor dunng operation, AE signal pat-JEANMOUGIN.N.; LOFARO,R.; et al. Rockwell international torn recognition development, and deve!opment of an ASTM Corp. July 1985. 66pa 8508020424. 85 ETEC-DRF 1?1. standard for application of continuous AE monitonno to pres-31961;008. sure boundanos. An independent anafysis was performed to determine seal inkage rates for the Westinghouse Reactor Coolant Pump NUREG/CR 4303: HIGH-LEVEL Y \STE PRECLOSURE SYS-(RCP) dunng a postulated station blackout resulting from losa af TEMS SAFETY ANALYSIS. Phase 1. Final Report HARRIS.P.A.; ac electne power Tho anafys:s confirmed Westinghouse cak.u- LIGON D M : STAMATELATOS.M ; et at G A Technologtos, ir c) lations on RCP seats performance for the three conditions in. General Atomic Co. September 1985. 330pp 8509300521. vsstigated. (1) all three seals function, (2) No 1 seal fails open SAND 85 7192. 32810 001, while Nos 2 and 3 seals function, and (3) all three seals fail The major effort for this project has been on the gathenng, open. organizing. and assembling of inforrnation pertinent to the safety assessment of a nuclear waste repository dunng preclosure op-NUREG/CR-429h EXTREMITY MONITORING Considerations For erations Specific issues a Ltressed in this report are.1. De-Use.Do$ meter Placement And Evaluaton, REECE,W D.; tailed analyss of a conceptual basalt repository desegn in order HARTY,R.; BRACKEN 8USH,L; et al Battelle Memonal Institute, to identify potential initiating event / accident scenanos capable P:cific Northwest Laboratones December 1985. 110pp of causing radiological and/or nonradiological consequences 2. 8601070502. PNL-5509. 34155 250. Evaluaton of radrological and nonradiological consequences rel-Vanous aspects of extremity dosimetry aro presented in that event to a nuclear repository and recommerutation of an ap-p;per to help the licensee decide when to use extremity dosim. proach for quantitative evaluation of these consequences 3. atry, what method of dosmetry to use, and how to interpret the Comparative evaluation of severalimportance ranking measures results of the dommetry system The current regulations as they that had been used in the nuclear industry in order to select a apply to extremity dosimetry and the Nuclear Regulatory Com- measure to best meet the neods of the program 4. Develop-missiorfs interpretation of these regulations are reviewed The ment of event and fault tree models for those initiating events history of the current Code of Federal Regulations is examined which have passed the preliminary screening process b Com-so that the reasoning behsnd the present regulations may be un. pelation of specific data such as initiating event frequencies,
92 Main Citations and Abstracts component / system failure rates and repair times, personnel system developed by EPB to monitor the regional seismicity. injury, and basic information necessary for more detailed rado. There are detailed desenptions of the logic used to select the logical consequence evaluations at a ,ater time. 6. Selection of vanous physical componerts of the system, the hardware and a set of accident scenanos to be quantified in the next study software that constituto the recordi1g system, and the data flow phase to demonstrate the apphcabihty of the proposed method- from detection to archive. In addition to the engineenng detads, ology that will identify and quantitatively prioritize structures, there is a discussion of the scientific results from the analysis of components, systems, and operations which are important to the regional seismicity dunng the reporting penod. Particular safety dunng the preclosure phase of a HLW rep. emphasis has been placod on the Miramichi. New Brunswick , earthquakes of January 1982. l NUREG/CR 4304: PRESSURE VESSEL FRACTURE STUDIES PERTAINING TO THE PWR THERMAL-SHOCK NUREG/CR 4318 V01: REACTOR SAFETY RESEARCH ISSUE Expenment TSE 7. CHEVERTON,R D.; BALL.D G.; PROGRAMS Ouarterly Report. January-March 1985. EDLER,S K. BOLT,S E.; et al, Oak Ridge National Laboratory. September Battelle Memonal Institute, Pacific Northwest Laboratones 1985.152pp. 8510040402. GSCA-46. 32859 050. August 1985. 27pp. 8509100359. PNL 5516-1. 32528 024. l Thermal-shock expenment TSE 7 was conducteo for the pur- This document summanzes work performed by Pacific North-pose of investigating the behaver of surface flaws under pres- west Laboratory from January 1 through March 31,1985, for the sunzed water-reactor (PWR) overcookng-accident condtions. Division of Accident Evaluation and the Division of Engineenng This expenment was the eighth in a senes of thermaf-shock ex- Technology, U S. Nuclear Regulatory Commission. PNL is oper-penments conducted for this purpose with large steel cyhnders ated for the U S. Department of Energy by Battelle Memonal in. (A508, class 2 chemistry; 991 mm outside diameter x 152-mm stitute under Contract DE AC06-76ALO 1830. Results from an wait x 1.2 m length) as a part of the Heavy-Sect n Steel Tech- MW d nology (HSST) Program. The initial flaw for TSE 7 was a shal- at Halden, Norway, are reported Expenmental data and analyti-low, semielhptical, inner surface, axially onented, sharp crack lo- cal models are being provded to aid in decison making regard-cated at mdiength of the test cyhnder The thermal shoca was 9 9 applied to the inner surf ace only, and this was acccmplished by energy fluid system piping Fuel assembhos and analytical sup-effectively dunking the test cyhnder, initidify at equivalent 93 de- port are being provided for expenmental programs at the Power grees centigrade, into a large volume of liquid nitrogen The Burst Facility, Idaho National Engineenng Laboratory, Idaho specific purpose of TSE 7 was to determine whether, in agree- Falls, Idaho. High temperature matenals property tests are ment with analysis, a short and shallow surface flaw, in the ab- being conducted to provde data on severe core damage fuel sence of cladding, would extend on the surface to effectivety behavor. Thermal-hyttrauhc computer programs are providing become a very long flaw as a result of severe thermal-shock best+ stimate analyses for a vanety of safety essues in hght-loading. Dunng the expenment, there were three major initiaton- water reactors. Severe fuel damage tests are being conducted arrest events. The first event consisted of some radaal propaga, in the NhU Reactor, Chalk River, Canada tion and very extensive surface extension, with many bifurca. 1.ons taking place. The second and third events consisted pn- NUREG/CR 4318 V02: REACTOR SAFETY RESEARCH manly of radial propagation. A fourth initiation event was pre. PROGRAMS Ouarterfy Report.Apnt-June 1985 EDLER,S K. Bat-vented by warm prestressing These results were in good telle Memonal Institute, Pacific Northwest Laboratones. October agreement witn predictions' 1985 25pp. 8511210266 PNL 5516-2. 33558 070. This document summanzes work performed by Pacific North. NUREG/CR 4305: COMMENTS ON THE LEAK BEFORE-BREAK CONCEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS. west Laboratory (PNL) from Apnl 1 through June 30.1985, for RODABAUGH.E.C. E C. Rodabaugh Associates. Inc.
Oak the Divison of Accident Evaluaton and the Divison of Engs-Rdge National Laboratory. September 1985 58pp. neenng Technology, U S. Nuclear Regulatory Commiswon. PNL 85100303t2. 32854 257. is operated for the U S. Department of Energy by Battelle Me.
Leak-before-break enfads the concept that, with a high monal Institute under Contract DE AC06 76ALO 1830 Results degree of probabihty, failure of the pressure bounday of piping from an instrumonted fuel assembly irradiation program being systems will be signaled by a detectable leak which will provide performed at Halden, Norway, are reported H,gh-temperature ample time to shut down and repair that leak. The status of the matenals property tests are being conducted to provide data on leak before-break concept is discussed in this report, including severe core damage fuel behavior Thermal. hydraulic computer a review of industnal and nuclear power plant exponence with programs are providng best-estimate analyses for a vanety of respect to leak before break, fracture mechanics and potential safety issues in hght water reactors Severe fuel damage tests ehminaton of postulated pipe breaks in nuclear power plant are being conducted in the National Research Universal (NRU) Reactor, Chaik River, Canada piping design NUREG/CR 4314; BRIEF SURVEY AND COMPARISON OF NUREG/CR 4321: FULL SCALE MEASUREMENTS OF SMOME COMMON CAUSE FAILURE ANALYSIS. WALLER.R A Los TRANSPORT AND DEPOSITION IN VENTILATION SYSTEM Alamos Scientific Laboratory August 1985. 28pp. 8509130416 DUCTWORK. MARTIN.R A : FENTON.D L Los Alamos Scientif-ratory. July 1985. 39pp. 8511010254 LA.10478 MS s pape p e ont a bnet survey of methods and a hst of hLa g references for analyzing common cause (mnde) failure Implicit This study is part of an effort to obtain expenmental data in models, exphcit modehng techniques and computer aids are in- support of the fire accident analysis computer code FIRAC' ciuded in the discussion. It is suggested that although current which was developed at the Los Alarnos National Laboratory trends are emphasizing development of emphcit models, a reaks- FIRAC can predict the transient movement of aerosohted or tic assessment of data availability will force continued use of im' gaseous matenal throughout the Comples ventilation systems of phcit or hybnd models in the immediate future. nuclear fuel cycle facihtles We conducted a prehminary set of CANADIAN SEISMIC full-scale matenal depletion /moefication eipenments to help NUREG/CR 4317 V01: Report Covenng 1979 1985 assess the accuracy of the code's aerosol depletion modet AGREEMENT. Technical Such tests were portormed under reakstic condbons using real HAYMAN.R B ; B ASHAM.P.W ; WETTMILLER.R.J.; et al Canada, Govt. of. JJy 1985 71pp 850809058132106 001. combustion products in full-stied ducts at typical air flow rates This Final Report provides a comprehensive summary of the To produce a combuskon aerosof, we burned both pofystyrene activit<cs of the Eanh Physics Branch fEPB)in eastern Canada. and polymethyl methacry! ate, the most and least smoky fuels i particularly with respect to the Eastern Canadian Telemetered typically found in fuel cycle plants, under vaned ventilation l (oxygen-lean and oxygen-nch) condtions Aerosol mass deposi-l Network (ECTN) The report descnbes the seismographic
Main Citations and Abstracts 93 t on. Siza, and concentration measurements were performed. maximum of the anafysis indicate that for some conditions the We found that as much as 25% of polystyrene smoke mass benefit in uung the 2-m flaw is substantial, but it decreases with and as little as 2% of the polymethly methacrytate generated at the entrance to a 15.2-m duct is deposited on the duct walls. increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the We also compared our expenmental results with theoretical limit on crack-arrest toughness. equations currently used in FIRAC. NUREG/CR-4322 V01: CORPORATE DATA NETWORK (CDN) NUREG/CR 4326 V01: EFFECTS OF CONTROL SYSTEM FAIL. DATA REQUIREMENTS TASK.Vol 1: Enterpnse Model.
URES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP Touche Ross & Co. November 1985 400pp. 8601070529. WESTINGHOUSE PRESSURIZED WATER REACTOR Main 34190.222. Report. BRUSKE.S J; DAVIS.C.B.; OGDEN.D M.; et al. EG&G The NRC has initiated a multeyear program to centralize its Idaho, Inc. (subs. of EGaG. Inc.). August 1985. 162pp.
information processing in a Corporate Data Network (CDN). The 8509100428. EGG-2405. 32530 093. information processing environment will include shared data- This report documents the evaluation of the effects of nonsa-bases, telecommunications, office automation tools, and state- fety grade control system failures on a typical 3-loop Westing. of-the art software. Touche Ross and Company was contracted house pressun2ed water reactor plant. The methods utilized for to perform a general data requirements analysis for shared da- this evaluation include a system level failure modes and effects tabases and to develop a preliminary plan for emplementation of analysis, deterministic computer anafysis (utilizing a plant model the CDN concept. The ENTERPR'SE MODEL (Vol.1) provided that includes the nuclear steam supply system, balance of plant the NRC with agency-wide enformation requirements in the form systems and control systems), a review of plant occurrences, a of data entities and organizational demand patterns as the basis probabehty analysis and a review of applicable Nuclear Regula-for clustenng the entities into logical groups. The DATA DIC- tory Commission (NRC) cntena pertaining to control systems. TIONARY (Vol. 2) provided the NRC with defan,tions and exam
~~This study identified two system failures that could cause tran.~~
pie attributes and properties for each entity. The DATA MODEL sients leading to a steam generator overfill and two system fail-(VOL. 3) defined logical databases and entity relationships ures that could lead to a reactor coolant coofdown of greater within and between data'ases. The PRELIMINARY STRATE. than 100 degrees fahrenheit per hour. It a!so identified two GIC DATA PLAN (Vol. 4) pnontized the development of data- system failures that could lead to an overpressunzation of low bases and included a workplan and approach for implementa- temperatures and two steam generator tube rupture events that tion of the shared database component of the Corporate Data could be further aggravated by additional system failures This Network, study concluded that the existing NRC critena, concerning con-trol systems, adequately addressed the potential problem areas NUREG/CR-4322 V02: CORPORATE DATA NETWORK (CDN) that were identified dunng this evafuation. Based on the results DATA REQUIREMENTS TASK.Vol 2: Data Entity Dictionary . of this study, it is recommended that the consequences and risk Touche Ross & Co. November 1985. 100pp 8601070524 associated with overfill and overcool transients be further inves-34198 031. gated it is afso recommended that the probabilities associated See NUREG/CR 4322,V01 abstract. *'th the low temperature overpressunration and the steam gen-erator tube rupture sequences be evaluated by the NRC 6taff. NUREG/CR 4322 V03: CORPORATE DATA NETWORK (CON) The results of this study will be factored together with other DATA REQUIAEMENTS TASK.Vol 3: Data Model
~~Touche Hoss & Co. November 1985.100pp. 8601070528. 34188143. studies being performed on the effects of control system fail.~~
See NUREG/CR-4322,V01 abstract- ures to estabhsh a position for msolution of Unresofved Safety Issue A 47 (Safety imphcations of Control Systems). NUREG/CR 4322 V04: CORPORATE DATA NETWORK (CDN) DATA REQUIREMENTS TASK.Vol 4 Prehminary Strategic Data NUREG/CR-4326 V02: EFFECTS OF CONTROL SYSTEM Fall-URES ON TRANSIENTS AND ACCiOENTS AT A 3-LOOP 8 010705t7 4188 3 WESTINGHOUSE PRESSURIZED WATER See NUREG/CR 4322,V01 abstract ~ REACTOR Appendices ORUSKE,S J ; DAVIS,C B , OGDEN.D M ; et al. EG&G Idaho, Inc (subs of EGAG. Inc) NUREG/CR 4325: A PARAMETRIC STUDY OF PWR PRESSURE October 1985. 233pp. 8511040050. EGG-2405 33337 327. VESSEL INTEGRITY DURING OVERCOOLING This report documents the evaluation of the effects of nonsa. ACCIDENTS,CONSIDERING BOTH 2-0 AND 3 D FLAWS. fety grade control system failures on a typical 3 foop Westing. CHEVERTON,R D.; BALL.D G. Oak Ridge National Laboratory house pressunted water reaClor plant The rnethods utilized for September 1985. 40pp 8510040371. ORNL/ TM-9682. 32860 335, this evaluation include a system level failure modes and offects analysis, deterministic computer analysis (utihting a plani model A continuing anafysis of the pressunted water reactor pres- that includes the nuclear steam supply system, balance of plant sunted thermal-shock problem indicates that the previously ac- systems and control systems), a review of plant occurrences, a cepted degree of conservatism in the fracture-mechanics model probabikty analysis and a review of applicable Nuclear Regula-needs to be more closely evaluated and, if excessive, reduced tory Commission (NRC) cntena putaining to control systems. One feature that was beheved to be conservative was the use This study identified two system failures that could cause tran-of two-dunensional as opposed to finite-length flaws. The sients leading to a steam generator overfall and two system fail-degree of conservatism could not be adequately investigated ures that could lead to a reactor coolant cooldown of greater because of computational hmitations and a lack of knowledge than 100 degrees fahrenheit per hour It also identified two regarding flaw behavior; however, that situation has changed to system fartures that could lead to an overpressuntation at low the entont that some cases involving finite-length flaws can be teroperatures and two steam generator tube rupture events that studied. A flaw of particular interest is one that is located in an could be further aggravated by additional system failures This axial wetd of a plate type vessel For those vessels that suffer study concluded that the existng NRC cntena, concerning con-relativefy high radiation damage in the welds, the length of the trol systems, and adequately address the potential problem flzw will be no greater than the length of the weld, and recent areas that were identified dunng this evaluation Dased on the calculations indicate that a deep ' law of that length (as equiva- results of this study, it is recommended that the consequences tent to 2rn) is not effectively infinitely long. contrary to previous thinking The benefit to be denved from consideration of the 2- and nsk associated wrth overfill and overcool transients be fur-ther investigateu 11 is also recommended that the probabihties m flaw and also a semielbpt<al flaw with a length-to depth ratio associated with the low temperature overpressuntation and the of 6/1 was investigated by anatyring several postulated tran- steam generator tube rupture sequences be evaluated by the sients in doing so the sensitivity of the benefit to a specified NRC staff The results of this study will be factored together
94 Main Citations and Abstracts with other studies being performed on the effects of control sion. We conclude that the zone may cor'tinue to expenence system failures to estabhsh a position for resolution of Unre- small earth movements, but catastrophic quakes similar to solved Safety issue A-47 (Safety Imphcatens of Control Sys- those at New Madnd in 1811 12 are unlikely. tems). NUREG/CR-4334: AN APPROACH TO THE QUANTIFICATION NUREG/CR-4329: RELIABILITY EVALUATION OF CONTAIN- OF SEISMIC MARGINS IN NUCLEAR POWER PLANTS. MENTS INCLUDING SOIL-STRUCTURE INTERACTION. BUDNITZ,R.J.; AMICO,P.J ; CORNELL.C.A.; et al. Lawrence PIRES.J.: HWANG,H ; REICH.M. Brookhaven National Laborato- Livermore National Laboratory. August 1985 307pp. ry. December 1985.128pp. 8601070439. BNL-NUREG-51906- 8508260301. UCID-20444. 32369 001. 34182.290- This report is the second report of the Expert Panel on Ouan-Soil-structure interaction effects on the reliability assessment tification of Seismic Margins. The Paners first report was enti-of containment structures are examined. The probabihty-based tied, "NRC Seismic Design Margins Program Plan." The objec-method for reliabihty evaluation of nuclear structures developed tive of this report is to discuss progress to date in studying the at Brookhaven National Laboratory is extended to include soil' issue of quantification of seismic margins in nuclear power structure interaction effects in this method, rehabihty of struc- plants. In particular, it deals with progress towards the estabhsh-tures is expressed in terms of hmit state probabihtees. Futher- ment of review guidelines that would be usefut in studying how more, rardom vibration theory is utshred to calculate hmit state much seismic margrn exists. The guidehnos themselves will be probabilities under random seismic loads Earthquake ground the subject of the next Panel report. The work presented in this motion is modeled by a segment of a zero-mean, stationary, fil' report is the result of a detailed study of seismic Probabikstic tered Gaussian white noise random process, represented by its Risk Assesssments, histoncal earthquake performance of the power spectrum. All possible seismic hazards at a site, repre- nuclear and non-nuclear facihties, and test data, augmented by sented by a hazard curve, are also included in the analysis. The the individual emperaence and emportise of the Panel merrbers. soil-foundation system is represented by a ngid surface founda- The major development discussed in this report is the HCLPF ton on an elastic halfspace. Random and other uncertair t:es in concept, which demonstrates margin by showing that there is a the strength properties of the structure, in the stiffness and i'* High Confdence of a Low Probability of Failure for a given tornal damping of the sos!, are also included in the anafysis Fi- earthquake size. nalty, a reakstic reimorced concrete containment is analyzed to demonstrate the application of the method. For this contain. NUREG/CR-4335: POTENTIAL BENEFITS OBTAtNED BY RE-ment, the soil-structure interaction effects on (1) hmet state OUIRING SAFTEY-GRADE COLD SHUTDOWN SYSTEMS probabihties, (2) structural fragihty curves, (3) floor response GALLUP,0.R.; KUNSMAN.D M ; BOHN.M P. Sandia National sg.ectra with probabikstic content, and (4) correlation coeffi- Laboratones. November 1985.127pp 8512030665 SAND 85-cients for total acceleration response at specif>ed structural lo- 1339. 33734:170. cations, are examined in detail. This work investigates 2 stems of concern to pressunzed water reactois: the potential benefits that may be obtained by NUREG/CR-4331: SIMPLIFIED SEISMIC PROBABILISTIC RISK requinng safety-grade cold shutdown systems, and the best ASSESSMENT Procedures And Limitations. SHIEH.LC ; suction valve arrangements for residual heat removal systems JOHNSON J.J.; WELLS.J E.; et al Lawrence Livermore National The approach taken is to perform a case study of a reference Laboratory. August 1985. 198pp. 8508220290. UCID-20468 power plant and to extrapolate the results obtained to other 32345.320. PWRs. Our analysis concfudes that the core melt frequency at At the request nf the U.S. Nuclear Regulatory Commission, the reference plant can be reduced by an estimated 7.1E 6 por the Lawrence Livermore National Laboratory has developed a year by requinng safety grade cold shutdown systems Most of simphfied seismic probabilistic risk assessment (PRA) methodot- this reduction can be attnbuted to making the systems Saismic ogy. The purpose of this methodology is to reduce the costs Category L For the accident scenanos inveshgated in this work, while adequately performing seismic probabihstic nsk assess. adding redundancy to the cold shutdown systems does not sig-ments of nuclear power plants. This report summanzes the do. nificantly reduce tho Core melt frequency Further, the abehty of velopment of the simphfied methodology and explains quede. PWRs to reach cold shutdown using the aunihary feedwater knes for applying the procedures. The development effort is part system and atmosphenc dump valves must be investigated on a of the scope of work of the Seismic Safety Margins Research plant-by plant basis. The ADVs at some plants may be sized too Program (SSMRP). small to allow the plant to reach cold shutdown. Finally, we be-NUREG/CR 4333: STE. GENEVIEVE FAULT ZONE. MISSOURI lieve that the "best" RHR suction valve arrangement is to have AND ILLINOIS. NELSON.W.J ; LUMM.D K. lilinois, State of. July a single suction kne without prim ary system overpressure hter-1985.104pp. 8508090574. 32103 255. locks on the valves. The Ste. Genevieve Fault Zone is a major structural feature which stnkes NW-SE for about 190 km on the NE flank of the NUREG/CR 4339: A REVIEW OF RECENT RESEARCH ON THE Ozark Dome. There is up to 900 m of vertical displacement on SEISMOTECTONICS OF THE SOUTHEASTERN SEABOARD high angte normal and reverse faults in the fault tone At both AND AN EVALUATION OF HYPOTHESES ON THE COUNCE ends the Ste. Genevieve Fault Zone dies out into a monochne. OF THE 1886 CHARLESTON. SOUTH CAROUNA CARTH-Two penods of faulting occurred. The first was in late Middle OUAKE. DEWEY,J W. Intenor, Dept. of, Geological Survey Devonian time and the second from latest Mississippian through August 1985. 45pp. 8508290532 32410197. earty Pennsylvanian time, with possible minor post-Pennsylva. In spite of extensive research on the source of the 1866 nian movement. No evidence was found to support the hypoth- Charleston, S C. Earthquake, there is not yet a ccnsensus esis that the Ste. Genevieve Fault Zone is part of a north- among earth scientists on the charactenstics of the fault that westward extension of the late Precambnan-early Cambnan produced the earthquake or on the hkehhood of future large Reelfoot Rift The magnetic and gravity anomalies cited in sup- earthquakes at other locations of the Southeastern Seaboard port of the "St. Louis arm" of the Reelfoot Rift possibly reflect This report reviews the evidence from recent research on thren deep crustal features underlying and older than the volcanic ter- categones nf hypothesis: (A) hypotheses on the specific geolog-rain of the St. Francois Mogntains (12 to 1.5 bilhon years old) ic structures that might cause largo ear'hquakes in the South. In regard to neotectonics no dispiacements of Quaternary sedi- eastern Seaboard (D) hypotheses on the seismotectonic sonos ments have been detected, but small earthquakes occur from in which large earthquakes might occur; and (C) hypotheses on time to time along the Ste. Genevieve Fault Zone. Many faults temporal vanations of seismicity in the Southeastern Seaboard. in the zone appear capable of shpping under the current stress Hypotheses that are representative of each caingory are sum-manzed, and evidence for and against each hypothesis is given. regime of east. northeast to west-southwest her tontal compres-
Main Citations and Abstracts 95 af such evidence is available. When data are interpreted in the tonal Skills Standard describing the skills / knowledge that the ways that currently seem to be the rnost straightforward, the hy. indwidual should possess in order to have achieved mastery. potheses that are supported by one kind of evidence are usual. The instructional skills obloctwes are organized according to the fy opposed by another kind of evidence. Reaching a consensus essentral elements of the Systems Approach to Training and on the cause of the Charleston earthquake, and on the likeli- are cross-referenced to three categories of instructional person- , hood of such an earthquake occurring at other locatons of the nei. developers of instruction, instructors, and instructional man-Southeastern Seaboard, will therefore probably require the rec- agers / supervisors. Use of the instructonal skills obectives is l oncihation of what currently appear to be contrary pieces of evi- demonstrated for reviewing instructional staff training and quali-dence. fication programs, developing cntenon-tests, and reviewing the NUREG/CR 4340 V01: REACTOR SAFETY RESEARCH SEMI. perf rmance and work products of sndividual staff members. l
~~ANNUAL REPORT. January - June 1985.~~
Sandia Natonal Lab-oratones. October 1985. 344pp. 8512270343 SAND 85-1606 NUREG/CR-4345: INVESTIGATION OF THE STABILITY OF LWR l SPENT FUEL RODS DELOW 250 C. OLSEN,C.S. EG&G Idaho, 34082 M Sandia National Laboratones is c>nducting, under the Inc. (subs. of EG&G, Inc.). September 1985. 153pp.
~~8511010394. EGG-2/09. 33335.071,~~
USNRC's sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research A testing program was conducted to investigate the long-term includes expenments to simulate the phenomenologu of the ac- stability of commercial pressurized. water-reactor (PWR) and cadent conditons and the development of anarytical models* boiling-water-reactor (BWR) spent fuel rods under a vanety of venfied by expenment, which can be used to predict reactor possible dry storage conditons. The obl ectue of this project is and safety systems performance and behavior under abnormal to provide the Nuclear Regulatory Commission (NRC) with an conditons. The objective of !his work is to provide NRC reque- expenmental bass to evaluate the results of short term, high-site data bases and anatytical methods to (1) identify and define temperature tests and to estabhsh a licensing positen for long-safety issues (2) understand the progression of nsk significant term, low-temperature (less than 250 degrees centigrade) spent accident sequences, and (3) conduct safety assessments. The fuel rod dry storage. A total of nine fuel rods (five BWR and four collective NRC-sponsored effort at Sandia National Laboratones PWR) were nondestructively examined after enterim heating pen.
is directed at enhancing the technology base supporting licens- ods, and seven of those fuel rods were destructively examined ing decis ons. to determine the degradaten of the fuel rods dunng the long-term, low-temperature dry fuel storage. The results of the non. NUREG/CR 4342: UNCEr4TAINTY AND SENSITIVITY ANALYSIS destructive examinatens of all nine fuel rods and the destruc. OF A MODEL FOR MULTICOMPONENT AEROSOL DYNAM. tive examination of one fuet rod were reported prevously. This ICS. HELTON,J C.; IMAN.R L; JOHNSON.J D ; et al. Sandia Na- report presents detailed results of the destructive examinations tional Laboratones September 1985. 59pp. 8511070252. of six of the remaining fuel rods, with bnef descnptions of the SAND 84 1307, 33300197. examinations of all the test fuel rods as they pertain to the fuel An uncertainty and sensitrvity analysis of the MAEROS model rod behaver dunng dry spent fuel storage conditions. for multicomponent aerosol dynamics is presented. Analysis tzchniques based on Latin hypercube samphng and regression NUREG/CR-4346: AEROSOL RELEASE EXPERIMENTS IN THE analysis are used to study the behaver of a two componont FUEL AEROSOL SIMULANT TEST FACILITY.UNDERSODIUM i aerosol in a nuclear power plant containment for a transient ac. EXPERIMENTS. PETRYKOWSKI,J.; LONGEST,A W ; cadent with loss of AC power (i e., a TML8' accident). Conditen- ROCHELLE J M : et al. Oak Ridge Natonal Laboratory. Septem-al on assumed ranges and distnbutons for selectod independ- ber 1985. 56pp. 8511070242. ORNL/TM-9479 33383.186. ent vanables (e.g , initial distnbutions and mass loadings for The release of UO(2) aerusols from pools of sodium was each component, termperature, pressure, shape factors), esti- studied in a senes of ten expenments in the Fuel Aerosol Simu-mates are made for distnbutens of model predictions and for tant Test (FAST) facihty at Oak Ridge National Laboratory The the endependent vanables which influence these predictions. expenments were designed to provide a mechanistic basis 'or The analysis indicated that, for the situaton under consider- evalating the radiological source term associated with a postu-ation, vanables related to agglomeration (e g , dynamic shape lated, hypothetical core disruptive accident (HCDA) in a liquid fictor, matenal density, agglomeraten shape factor, and turbu- metal fast breeder reactor (LMFDR) Aerosol was generated by lence dissipation rate) tended to dominate the observed vanabel. capacitor discharge vaporitation of UO(2) pellets which were sty. For companson, an analysis based on differential techniques submerged in a sodium pool under an argon cover gas. Meas-is also given. Further, a study of the effects on MAEROS pre- urementa of the pool and cover gas pressure were used to dictons due to the number of particle size classes and the party study the transport of aerosol contained by vapor bubbles within cle size class boundanes is presented This analysis was per, the pool Cover gas samples were filtered to determine the formed as part of a project to develop a new system of comput- quantity of aercsol released from the pool. Trace amounts of er codes (ie., the MELCOR code system) for use in nsk assess. UO(2) aerosol (<0.3% of the total pellet mass) were detected ments for nuclear power plants. In the cover gas samples suggesting that the bulk of aerosol was trapped within bubbles confined by the pool. The report NUREG/CR 4344: INSTRUCTIONAL SKILLS EVALUATION IN contains (1) a desenption of the expenments (2) data records NUCLEAR INDUSTRY TRAINING. MAZOUR,T.J; DALL,F M. for the pool pressure, the cover gas pressure, and the UO(2) Analysrs & Technology, Inc. November 1985. 136pp. aerosol concentration in the cover gas and, (3) an analysis of 8512300185. 34094.052. the expenmental findings using simplihed models of bubble be-This report provides informaton to nuclear power plant train- havor. ing managers and their staffs concerning the job performance requirements of instructonal personnel to implement perform. NUREG/CR 4347: EMERGENCY DIESEL GENERATOR OPER-Ence-based training programs (also referred to as the Systems ATING EXPERIENCE,19811983 BATTLE.RE. Oak Ridge Na. Approach to Training) The information presented in this report tional Laboratory. December 1985. 80pp. 8601070491. ORNL/ kB a compilation of bnformation and lessons leamed in the nucle. TM.9739 34198155, tr power industry and in other industnes using performance. The purpose of this report is to update the operating expen-b sed training programs The job performanco requerements in ance of emergency diesel generators in nuclear power plants. this report are presented as instructional skills objectives. The Previously, similar data for 1976 through 1980 were reported in process L, sed to develop the instructional skills objectives is de- NUREG/CR 2989. "Rehabikty of Emergency AC Power Systems scnbed. Each ob l octive includes an Instructional Skills State- at Nuclear Power Plants " The two data sets are used to show ment describing the behavior that is expected and an Instruc. trends of diesel generator performance, and responses by nu-
96 Main Citations and Abstracts clear plant hcenses to a nuclear regulatory questonnaire is in- This course employs a combination of lecture matenal and cluded for additional data and comparison with the Licensee practical problem solving in order to develop competence and Event Report (LER) data collected for this report. The LER da- understanding of the pnnciples and techniques of event tree tabase was used to collect diesel generator failures, and the da- and fault tree analysis The role of these techniques in the over-t; bases for diesel generator successes were from nuclear plant all context of PRA is descnbed. The emphasis of this course as licensees' responses to NRC questionnaires. Estimates of on the basic, traditional methods of event treo and fault tree diesel generator failure on demand were calculated from the analysis diesel generators test data, from data reported in response to an NRC questionnaire (genenc letter 8415), from diesel gener, NUREG/CR 4350 V05: PROBABILISTIC RISK ASSESSMENT COURSE DOCUMENTATION Volume 5 Systems Reliabikty ator performance dunng comp'ete and partial losses of off-site And Analysis Techniques, Session D Quantification. power, and from diesel generator performance for safety injec-LOFGREN.E.V. Science Applications international Corp. (for-tion actuation signafs. merly Science Apphcahons, Inc)
~~Sandia National Laborato.~~
~~NUREG/CR 4350 V01: PROBABILISTIC RISK ASSESSMENT nes. August 1985. 215pp. 8511190554. SAND 851495/5 COURSE DOCUMENTATION Volume 1 - PRA Fundamentals~~
~~33533.029 Sandia National Laboratones. BREEDING.R J.; LEAHY,T.J ; et This course focuses on the probabikstic quant,ficaton of acci.~~
al. Energy, Inc. August 1985 425pp 8512270366. SAND 85' dont sequences and the link between accident and conse-1495. 34083.101. quences Previous sessions in this senes have focused on the The full range of PRA topics are presented, with special em- quantification of system reliability (Session A) and the develop-phasis on systems analysis and PRA appkcations. Systems ment of event trees and fault trees (Session B/C) This course analysis topics include system modeling such as fault tree and takes tho viewpoint that the event tree sequences or combina-event tree construction, failure rate data, and human rehabikty. tions of system failures and success are avadable, and that the The discussen of PRA apphcations is centered on past and Boolean equations for system fault trees have been developed present PRA-based programs such as WASH-1400 and the In- and are available. tznm Rehabihty Evaluation Program, as well as on some of the potential future apphcations of PRA. The relationship of PRA to NUREG/CR-4350 V06: PROBABILISTIC RISK ASSESSMENT genenc safety issues such as station blackout and Anticipated COURSE DOCUMENTATION Volume 6 Data Development. Transient Without Scram (ATWS) is also discussed in r,ddition LEVERENZ,F.L.; COX.D C. Battelle Memonal Institute, Colum-to system modehng the major PRA tasks of accident process bus Laboratones
~~Sandia Natonal Laboratones August 1985.~~
analysis, and consequence analysis are presented An explana- 165pp. 8512120164. SAND 851495/6. 33876.163. tion of the results of these activities and the techniques by This course desenbes the process used to develop quantita-which these results are donved forms the basis for a discussion tive values for events other than human errors in fault trees, of these topics An additional topic presented in this course is event trees, or other system models. The events included are the topic of PRA management, organsration. and evaluat!on. (1) initiating events,(2) unavadabdity due to hardware fadure,(3) This discussion explains the relatcnship of sound managment, unavadabshty due to maintenance activity, and (4) unavadabihty proper organization, and thorough evaluaton to the perform- due to test activities. The course matenal covers the most ance of credible nsk assessment. common rehabdity models fcr these events and explains how to evaluate them with uncertaint:es, determine their parameters, NUREG/CR-4350 V02: PROBADlLISTIC RISK ASSESSMENT and obtain the raw data for this process. Both Bayesian and COURSE DOCUMENTATION Volume 2-Probabihty And Statss-classical approaches to the data development process are pre-tics For PRA Applications. IMAN.R.L; PRAIRIE.R R. Sandia Na-tional Laboratones September 1985. 285pp. 8510040381. sented SAND 851495. 32858 029. NUREG/CR-4350 V07: PROBABluSTIC RISK ASSESSMENT This course is intended to provide the necessary probabibstic COURSE DOCUMENTATION. Volume 7 Environmental Trans-and statistical skdis to perform a PRA. Fundamental background port And Consequence Analysis RITCHIE,L.T. Sandia Nat onal informaton is reviewed, but the pnncipal purpose is to address Laboratones. August 1985. 400pp. 8511070092. SAND 851495/ specific techniques used in PRAs and to illustrate them with ap- 7, 33376.120. phcations. Specific examples and problems are presented for Consequence models have been designed to assess health most of the topics. and economic nska from potential accidents at nuclear power NUREG/CR 4350 V03: PROBABILISTIC RISK ASSESSMENT plants. These models have been apphed in an ever increasing Systems Rehabihty vane'y of problems with ever increatog demands to improve COURSE DOCUMENTATION Volume 3 And Analysis Techniques. Session A Rehabihty. LOFGREN.E.V. modehng capabihtees and provide reaksm. This course dis-cusses the environmental transport of postulated radiological re-Science Apphcations International Corp (formerly Science Ap. leasen and the elements and purpose of accident consequence phcatons, Inc )
~~Sandia National Laboratones. August 1985.~~
~~210pp 8511190550. SAND 85-1495/3. 3J341225. evaLetion. This course focuses on the quantitative eshmation of rehabd* NUREG/CR 4351:~~
~~SUMMARY~~
REPORT FOR LOFT ANTICIPAT-sty at the systems level. Vanous methods are reviewed, but the ED TRANSIENT EXPERIMENT SERIES L6-8. NALE2NY,C L.; structure provided by the f ault tree method is used as the basis CHEN,T. EG&G Idaho, Inc. (subs. of EG&G. Inc). November fu system rehabihty estimates The pnnciples of fault tree analy- 1985.109pp. 8511250106. EGG-2406. 33619 346. sis are bnefty reviewed Contnbutors to system unrehabehty and Espenment Senes L6-8 was conducted in the Loss.of Fluid unavadabihty are reviewed, models are given for quantitative Test (LOFT) facihty between August 26 and August 31,1982. evaluaton, and the requirements for both genenc and plant- TNs mpenment senes simulated six individual transients that specific data are discussed. Also covered are issues of quantify- have a high probability of occurrence dunng the lifetime of a ing component faults that relate to the systems contest in which commercial pressunted water reactor (PWR) The transients the components are embedded. All rehabikty terms are care'ulty simulated for the espenment consisted of two control rod with-defined. drawafs (L6 801 and B-2), three small break recovery methods NUREG/CR-4350 V04: PROBABILISTIC RISK ASSESSMENT (L6-8C-1 C-2 and C 3), and one natural circulaton cooldown COURSE DOCUMENTATION Volume 4 - System Rehabikty And with low decay heat (L6-8D). This report presents the esperi-Analysis Techniques.Sessons B/C Event Troos/Fau!t Trees. ' ment results and compares the expenmental data with postex-Sandia National Laboratones. HAASL.D Institute of Systems penment calculabons made uung the RELAPS/ MODI computer Sciences. YOUNGJ Energy, Inc October 1985. 143pp code. The system response donng the two rod withdrawal es. 8510290410. SAND 851495. 33259151. penments (L6 881 and L6 80-2) was successful, dunng L6 . - _ . _
Main Citations and Abstracts 97 1, negative reactivity feedback was dominated by moderate initial Pu content. The initial Pu concentration on trabecular feedback, while dunng L6-88-2 negative reactivity feedback was bone surfaces in red marrow was calculated to be about 4.5 dominated by Doppler feedback. The companson between L6- times greater than on compact bone surfaces. About one half of 8C-1 and L6-8C-2 demonstrated that s'though the pressunzer initial skeletal Pu was ehminated in 1100 days mainly from ca+ spray is as effectivo as the power operated refref valve (PORV) cellous structures in red marrow imphcations for some changes in depressunzing the pnmary system, it is also more sensit ve to in Internatoni Commisson on Radiological Protection metauche operator control. The results of L6-8C-3 indicate that while and dosimetnc models for Pu are noted. pump current (or power), is sensittve to void fraction, or density, j in the intact loop, it is not an accurate measure of system in. NUREG/CR-4357: THE FEASIBILITY OF DETECTING THE i ventory. The void fracton in the intact loop was about 50'. of IMPORT OF UNAUTHORIZED RADIOACTIVE MATERIALS the void fraction in the entire pnmary system. Pnmary system INTO THE UNITED STATES BEE.R W.; GORDON.J ; voiding, which occurred as intended dunng L6 80, did not sig. KWAN.0 ; et al. Aerospace Corp. September 1985. 235pp l nificantty affect natural circulation. The data obta.ned from 8510040578.32862 213 these six simulations will be valuable for qualifying computer This report explores the feasibility of establishing and operat. codes used to calculate anticipated transients an commercial ing a radiation monitonng system at U S borders to operate in PWRs. conjunction witn normal U S Customs Service inspection proce-NUREG/CR-4352: SUGGESTED STATE REQUIREMENTS AND E "#* * # # "'# nation in imported matenais The study defined potential inad-CRITERIA FOR A LOW-LEVEL RADIOACTIVE WASTE DIS-POSAL SITE REGULATORY PROGRAM. RATLIFF,R A ; vertent contamination threats of radioactive matenals in industn. DORNSIFE.B ; AUTRY,V.; et al. Conference of Radiation Con- al and commercial usage and evplored how such matenals trol Program Directors, Inc. August 1985. 50pp 8509060210- might accidentally enter new product manufactunng processes 32503 323. Radiation monitoring equipment necessary to detect such con-Desenption of cntena and procedure for a state to follow in tamination was examined, and a technical system description the development of a program to regufate a LLW disposal site. and capital, operating. and ma!ntenance costs were developed This would include identifying those portions of the NRC regula- Hypothetical scenanos were developed to estimate a range of tions that should be matters cf compatibikty, ident;fying the van- costs for recall, cleanup, and disposal of contaminted products ous expertise and discipt;nes that wIf be necessary to effective- well after distnbution to consumers Estimates were rnade of ly regulate a disposal site,6dentifying the resources necessary added health risk to consumers associated with exposure to contaminant ra6,ation if the contaminated products go undetect-for conducting a confirmatory monitoring program, and providing suggestions in other areas which, based on expenences, would ed An economic analysis was performed to provide a common result in a more effective regulatory program basis for their companson Because such aspects as frequency, level of contamination, charactenstics of product distribution, NUREG/CR-4354; A STUDY OF SEISMICITY AND TECTONICS and health effects cannot be assessed using absolute bounding IN NEW ENGLAND Final Report. E8EL.J E. Boston College. cond,tions, no direct compansons of courses of action can be Chestnut H,ll, MA August 1985. 100pp. 8509100335 made. Rather, the study provides decison mahers with vanous 32528.181, d'mensions of the problems, thus providing an improved basis The operation by Weston Observatory of a seismic network in for related decison making New England from 1974 to 1995 is desenbed, and the results of the seismic monitonng are summanzed. The network coverage g of Weston Observatory increased from two operating stations in EQUIPMENT OUALIFICATION ISSUES GILLEN K T , 1974 to 38 stations in 1979 and was stabalized at 30 stations in CLOUGH.R L; DHOOGE.N J Sandia National Laboratories the early 1980's. The network was used to find the locations September 1985 38pp 8510040354 S AND851557. tnd magnitudes of all earthquaka activity detected during the 32861 012. study penod Most earthqupkes from 1974 to 1985 were found This paper reviews the density profshng technique, a new, in-to occur en the same places as those which have been docu- expensive and versatile analytical method which can yield ex-mented histoncally, atthough the activity appears to be random tremely useful information on beterogenesties in polymers. The both in space and time. Studies of aftershocks and detailed I'Ch"'que makes use of a density gradent column to measure monitonng in selected areas did not show any strong correla- the density of a senes of successively-cut shces across a tions between the earthquake locations and mapped geologic sample Since the density of very thin shces can easily be ob-structures. It is concluded that the relationship among earth- tained, density profdes across very small cross-sectons quakes, tetonic or structural zones and faults exposed on the (< 1mm) are readdy avaslable. A maior appheation of the tech-surface are not well understood. The causes of the earthquake naque involves oxidation studies of polymers, since oxidation re-tctivity in the northeast are not clearfy established with the seis. actions usually lead to substantial increases in polymer density mac data which was gathered and analyzed. O'Huson hmited ouviation effects, which lead to heterogeneous-ly oxidized matenals, are of ten piesent in polymer aging studies NUZEO/CR 4355 V01: 238 PU(IV) IN MONKEYS Overview Of in air. Since these effects are responsible for the commonly-ob-Metaboksm. DURBIN,P W4 JEUNaN ; SCHMIOT,C T. Lawrence served physical dose-rate effects in radiation aging environ. Berkeley Laboratory Sep%nber 1985. 96pp 8510020240. LBL- ments and for non Arrhenius behavior in thermal aging environ. 20022.32839 157. ments, the availabihty of simple oxidation profiling techniques is Complete balance studies were performed using 21 adult and a tremendous aid in vahdating tne aging simulation aspects of four adolescent Macaque monkeys bhree species, both sexes) equipment qualificaton procedures This paper gives examples to def,ne distnbuten and retention of (238)Pu(IV) citrate from 2 of the utshty of density profihng for studying oxygen diffusion-hm-hr to 1100 d after parenteral insection Expenmental methods ited degradaten in both raddson and thermal aging environ-are descobed in detail. Initial distnbution (6 adults, 7 5 plus ments and in discovenng/ understanding chemical dose-rate ef-minus 0.5 d) and retenton (6 adults, 711 plus minus 310 d) of fects in high energy radiation environments Pu were, respectively, as follows: skeleton and teeth 28 and 14% ID, hver 60 and it'. ID, other soft tissues 6 and 0 9 % ID NUREG/CR 4360 V01: CALCULATIONAL METHODS FOR ANAL-cnd excretion 5 and 74% ID Initial Pu content of ovary and YSIS OF POSTULATED UF6 RELEASES. WILLIAMS.W R Oak teshs was 0.005 and 0 06% 10, rospectively, and both declined Ridge National Laboratory Septemt er 1985. 135pp eth T1/2 equivalent to 1 yr. Liver Pu was cleared mainly by ex- 8512120100 ORNL/E NG/TM.31 33873 215 creton to feces, but also by recirculation, with an average T1/2 Calculabonal methods and computer progacms for the anary.
~~= 180 d. By 1100 d, most soft tissues lost 50 to 90*. of their sa of source terms for postulated releases of UF(6) are pre.~~
98 Main Citations and Abstracts sented. Required thermophysical properties of UF(6), HF, and This report is a summary of progress in the Surry Steam Gen-H(2)O are desenbed in detail. UF(6) reacts with moisture in the erator Group Project for 1984. Information is presented on the ambient environment to form HF and H(2)O. The coesstence of anatysis of two baseline eddy current inspections of the genera. HF and H(2)O significantly alters their pure component proper- tor. Round robin series of tests using standard in-service in-ties, and HF vapor pofymenzes. A release rate model of UF(6) spection techniques are desenbed along with some preliminary is presented that considers the transient conditions inside con- results. Observations are reported of degradation found on ta;nment and the flashing. multiphase flow of UF(6) along the tubing specimens removed from the generator, and on support r$ lease pathway. Transient compartment models for simulating plates charactented in-situ. Residual stresses r,easured on a UF(6) release inside rooms ventilated by forced- and induced- tubing specimen are reported Two steam generator repair dem. draft systems are also desenbed. The basic compartment model onstratens are descnbed, one for antivibration bar replacement, mass and energy balances are supported by simple heat trans- and one on tube repair methods. Chemical analyses are shown fer, ventilation system, and deposition models. A model that can for sludge samples removed from above the tube sheet simulate either a closed compartment of a steadt state ventila-tion system is also discussed. Listings of all main programs (in. NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL ciuding two plotting routines) and subroutines are included. Ex. PURPOSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUA-tmple problems illustrate the analysis of postulated releases TION OF NUCLEAR REACTOR VESSELS AND PIPING COM-PONENTS. GANAPATHY.S; SCHMULT.0 ; WU,W S; et al. using the desenbed programs. Michigan, Univ of, Ann Arbor. MI. August 1985. 126pp. N'JREG/CR 4360 V02: CALCULATIONAL METHODS FOR ANAL- 8509100506 32536:106 YSIS OF POSTULATED UF6 RELEASES. WILLIAMS.W R Oak This report desenbes the design details of a special purpose Ridge National Laboratory September 1985. 214pp system for real-time nondestructive evaluaten of reactor ves-85t2120102. ORNL/ENG/TM 31. 33873 350 seis and piping components. The system consists of several Calculational methods and computer programs for the analy- components and the report presents the results of the research sis of source terms for postulated releases of UF(6) are pre- aimed at the design of each component and recommendations sented. Required thermophysical properties of UF(6), HF, and based on the results. One major component of the NDE system, namely the real-time SAFT processor was designed H(2)O are desenbed in detail. UF(6) reacts with moisture in the with sufficient details to enable the fabncation of a prototype by ambient environment to form HF and H(2)O. The coexistence of HF and H(2)O signifcantly alters their pure component proper- GARD inc. under a subcontract from The University of Michigan t:es, and HF vapor polymenzes. A release rate model of UF(6) and the report includes their results and conclusions. is presented that considers the transient conditions inside con-tainment and the flashing multiphase flow of UF(6) along the NU"tEG/CR 4367: ORViRT.PC A 2 0 FJNITE ELEMENT FRAC. release pathway. Transient compartment models fro simulating TURE ANALYSIS PROGRAM FOR A MICROCOMPUTER. BRYSON.J W. Oak Ridge National Laboratory. October 1985. UF(6) releases inside rooms ventilated by forced and induced-60pp. 8512270362. ORNL-6208 34078 271. draft systems are also descr. bed. The basic compartment model ORVIRT PC (Oak Ridge virtual crach extenson Personal mass and energy balaaces are supported by simple heat trans. Computer) is a 2 0 finite element fracture analysis program for fer, ventilation system, and deposition models. A model that can an IBM PC/AT microcomputer. The code is based to a large simulate either a closed compartment or a steady-state ventila, entent on the techniques i *ad in the ORMGEN ADINA-ORVIRT tion system is also discussed. Listings of all main programs (in. fracture anafysis syste.a m/lRT PC is a stand-alone program ciuding two plotting routines) and subroutines are included. Es. capable of performing 2-D linear elastic stress and fracture-me-ample problems illustrate the analysis of postulated releases chanics anafyses Thermal loadings may be anafyred in add, tion using the desenbed programs. to mechanical loadings Crack face tractinng rney also be cork NUREG/CR 4361: S'EAM GENERATOR GROUP sidered. Eight-noded isoparametnc elements which comtane PROJECT. Annual Report - 1983 CLARK,R A.; LEWIS.M ; both performance and ease of modelling are employed in the MUSCARA.J. Battelle Memonal institute, Pacific Northwest Lab- program. Special crack tip elements which allow for an inverse oratones. September 1985 125pp. 8511050309 PNL 5017. square root vanaten in the near.tip stress and strain fields are used at the crack tip. Detailed user instructions are provided 33339:181. The Stearn Generator Group Project (SGGP) is an NRG pro- which desenbe both preparation of input data and program op. gram jointly funded by consortia from France, Italy and Japan, eration. Sample problems are presented which demonstrate and the Electnc Power Research institute. The SGGP uti!ites a good agreement with known solutions. steam generator removed from service at a nuclear plant as a vehicle for research on a vanety of safety and reliability issues NUREG/CR 4368: NDE OF STAINLESS STEEL AND ON LINE LEAK MONITORING OF LWAS. Semiannual Report. October This report is an annual summary of progress on the program 1984 - March 1985 KUPPERMAN.D S ; CLAYTOR.T N ; for 1983. Information is presented on the task of removing 969 MATHIESON.T.; et al Argonne Natonal Laboratory October plugs by dnlling them out of the tubes in the steam generator to 1985 38pp 851 f 250006. ANL 85-46. 33620 2:5. permit access for eddy current nondestructive testing probes A Two pipe-to-endcap weldments with overlays were examined descnption is given of two post service baseline eddy current an. Results conclude that it is entremely difficult to inspect pipes spections, one by Zetec and one by intercontrole, of about with overlays because of ur predictable beam distortion due to 3000 tubes Photographic documentation is provided of damage the overlay and the absence of effective reference pipes The observatons on the secondary side of the unit Corrosion use of 1-Mh2 longitudinal angle beam probes rather than thear. coupon results are reported from the channel head decontami- wave probes may facilitate inspecten of such pipes Fout 60-nations descnbed in the 1982 annual report. Plans for further mm-thick cast stainless steel plates with microstructures ranging eddy current testing are discussed, including a plan to conduct from equiased to pnmariy t columnar grains have been esamined a round robin senes of tests in which up to eight independent, with ultrasonic waves 11 was found that the longitudinal velocity commercial inspecton teams would all inspect the same set of of sound and the fato of longitudinal to shear velocity as a selected tubes in the steam generator. function of position can be used to characterite the crystallo-GENERATOR GROUP graphic testure. It was also found that the beam skewing that NUREG/CR 4362: STEAM occurs in columnar (but not equiased) structures is strong PROJECT. Annual Report . 1984. CLARK.R A ; LEWIS.M ; MUSCARA.J. Battelle Memonal Institute, Pacific Northwest Lab- enough to that measurements of probe separation at mansmum oralones. September 1985, 109pp 8511210279 PNL 5417 received signal intnntry for 45 degrees shear wave pitch catch 33557.321. transducers can be correlated with microstructure Leaks from a i i
Main Citations and Abstracts 99 24n. ball valve and a flange were studied and compared with and the hquid phase velocity No momemtum equation is re-leaks from ntergranular stress corrosson cracks (IGSCCs) and quired for SIT SG because the detailed pressure distnbution in fatigue cracks. The dependence of acoustic signal on flow rate the vesselis not important for the blowdown process. Based on and frequency for the valve and the flange was comparable to the compansons between the code predictions and the data ob-that of fatigue cracks (thermat and mechanica!) and different tained from the expenments conducted in Battelle4rankfurt and from that of IGSCCs. GE, the best flux model constants for vanous flow regimes are NUREG/CR-4373: COMPENDIUM OF COST EFFECTIVENESS selected SIT SG has been used to predict the carryover, fall EVALUATIONS OF MODIFICATIONS FOR DOSE REDUCTION back, and heat transfer for the MIT steam generator blowdown AT NUCLEAR POWER PLANTS. BAUM,) W.; MATTHEWS.G R expenments The results are encouraging it is found that the Brookhaven National Laboratory December 1985. 142pp nwasured dryout kont is much higher than the calculated mis-8601070478. BNL NUREG-51915. 34187.005. ture levet if the effective heat transfer are is determined from This report summanzes available information on cost effec- the mixture level, the pnmary-to-secondary heat teransfer will be tiveness of engineenng modifications of potential value in dose substantia!!y underpredicted From the result of the liquwt hold reduction at nuclear power plants Data was gathered from sev- up study we would espect to find two misture levels, one in the eral U S. utilities. published literature, equipment and service b ttom of the steam generator and one above the top tube sup-suppliers, and recent technical meetings. Five simphf.ed econo- port plate, provided that flooding occurs at all metnc models were employed to evaluate data and arrive at a value for cost effectiveness empressed in either (a) dollars / rem; NUREG/CR 4377: EVALUATIONS AND UTILIZATIONS OF RISK or (b) total dollar savings calculated using a nominal value of IMPORTANCES. VESELY,W E.; DAVIS.T C Battelle Memonal
~~$1.000/ rem. Models employed were. a basic rrodel with no institute. Columbus Laboratones August 1985 13tpp.~~
consideration given to the time value of money; two models in 8509100355 BMI-2129 32528 049 which discounting was used to evafuate costs and savings in Thes report presents approaches for utilizing Probabilistic R,sk terms of present values; and two models in which income taxes Analys s (PRA) to determine risk importances PRAs can be and revenue requirements were considered Results from differ- used lo identify the importances of nsk contnbutors or proposed ent models vaned by as much as a factor of 10. and were gen- changes to designs or operations The obl octivo of this report is eralty lowest for the basic model and highest for the before tan to serve as a handbook and guide in evaluating and applying revenue requirements model. Results for 151 evaluations em- nsk importances. The utthration of both quahtative nsk impor. ploying d.fferent assumptions concerning number of plants per tances and quantitative nsk importances is descnbed in this sete and outage impacts were tabulated in order of decreasing report. Quaktatrve nsk emportances are based on the logic cost effectiveness Twenty five evaluations were identified as models in the PRA while quant;tative nsk importances are exceptionally cost effective since both costs and dose were based on the quantitative results of the PRA Both types of im. saved. Forty evaluations indicated highly cost-effective changes portances are among the most robust and meanengfut informa-tion a PRA can provide based on costs of less than $1,000/ rem saved using results of the present. worth fleudel that included discounting of future dose savings. NUREG/CR 4379 V01: LONG TERM PERFORMANCE OF MATE. RIALS USED FOR HIGH LEVEL WASTE PACKAGING Ferst NUREG/CR 4375: THEORY. DESIGN.AND OPERATION OF Ouarterly Report Year Four Apni June 1985 STAHL.D ; LIQUlO METAL FAST BREEDER REACTORS. INCLUDING MILLER N E. Battelle Memonal Institute. Columbus Laborato-OPERATIONAL HE ALTH PHYSICS, ADAMS.S R. EG&G Idaho, nes September 1985.111pp. 85t0020228. 32838 277. Inc. (subs of EG8G, Inc ). October 1985. 258pp 8512270240 High level waste glass studies are berg concluded and ef. EGG-2415. 34079 355 forts are being directed toward studying spent fuel performance A comprehensive evaluation was conducted of the radiation The effects of devitnf cation on glass leach rates are being in-protection practices and programs at prototype LMFBAs wdh vestigated, and sihca dissolution was studaed to provide data for long operational expenence installations evaluated were the the glass dissolution model Preliminary data support this model fast Flux Test Facihty (FFTF) Richland, Washington: Expen. A leach test using organic acids was conducted and inaching mental Breeder Reactor 11 (EBR II) Idahn Falls, Idaho, Prototype tronds were observed Real and simulated spent fuels are being Fast Reactor (PFR) Dounreay. Scotland. Phenir, Marcoule, incorporated in integral tests using simulated grourx1 water in a France, and Kompakte Natnumgekuhite kernreak Toraniange prototypic repository environment The reactions of groundwater (KNK 11), Kartsruhe, Federal Repubhc of Germany. The evalua. species with steels are being analyzed to evaluate susceptabi. tion included external and internal exposure control, respratory hty to pitting and stress-corrosion cracking Potential cracking protection procedures, radiation survellance practices, radioac- agents are being investigated by slow strain rate expenments tive waste management, and engineenng controls for confining General and pitting corrosion modets were further developed, radiation contamination The theory, design, and operating en, based on known pnnciples of mass transport and radiohtic pro-ponence at LMFORs iS desenbod Aspects of LMFOR health duction A simplif ed groundwater radiolysis model, developed physics different from the LWR exponence in the United States for use with the corrosion models, was compared with other tre identifed. Suggestions are made for modifications to the mechanisms for species concentration prod.ctions NRC Standard Revew Plan based on the differences' NUREG/CR 4382: CONCENTRATIONS OF URANfUM AND THO. NUREG/CR 4378: HEAT TRANSFER. CARRYOVER AND FALL RIUM ISOTOPES IN URANIUM MILLERS
AND MINERS' TIS-BACK IN PWR STEAM GENERATORS DURING TRANS:ENTS SUES. WRENN.M E.; SINGH.N P ; PASCHOA.A S ; et af. Utah, LtAO.LH.; PARLOS A , GRIFFITH.P. Massachusetts Institute of Univ of, Saft Lake City, UT. September 1985. 58pp Technology, Cambndge, MA September 1935. 212pp 85t0250529 3319t.286.
6510020307. EPRI NP 4298. 32828124 The alpha. emitting isotopes of uranium and thorium were de-The concern over Pressunrod Thermal Shock (PfS), along termined en the lungs of 14 former uranium miners and in soft with many other concerns, indicates the need for accurate fictues and bones of three miners and two millors These ra. knowledge of the steam generator behavior dunng the blow- dionuchdos were also delermined in soft tissues and bones of down of the steam generator secondary side To fulfill this need seven normal controls The average concentrations in pCa/hg a computer program, SIT SG (Simulator of Transient in Steam wet weight in 17 former miners' lungs are as follows U 230, 75. Generator) is deveioped This is a one dimensional best ests. U 234, 80. Th 230. 79 Concentrations of each nuclide rarupd mate code with the assumption that the vapor and liquid phases from 2 to 325 pCi/kg The average ratio of U 238/U 234 was are in thermal equibbnum but not homogeneous The dnft flus 0 92, ranging fiom 0 64 to 1.06 The mean ratio of Th-230/U model is used to descnbe the relationship between the vapor 234 indicetes that the fato of elimination of uranium and thonun
100 Main Citations and Abstracts from lungs is the same in former uranium miners. The concen- NUREG/CR 438h EFFECTS OF CONTROL SYSTEM FAILURES trations of U-234 and U-238 were highest in lung, however, the ON TRANSIENTS ACCIDENTS AND CORE-MELT FREQUEN-concentration of Th 230 in bones was either higher than or CIES AT A GENERAL ELECTRIC PRESSURIZED WATER RE-comparable to its concentraton in lung The concentraton ACTOR. BICKFORD,W E.; TABATABAl,A S. Battelle Memonal rctos of Th-230/U 234 in bone of uranium miners and millers Institute, Pacife Northwest Laboratones December 1985.87pp. measured in our laboratory have been compared w.th results 8512270082. PNL 5545. 34078 099. predicted by ICRP-30 metabolic models. These results indicate Pacific Northwest Laboratory (PNL) performed probabilistic that the ICRP metabohc models for thonum and uranium were nsk analyses to estimate core-melt frequency and pubhc nsk as-only marginalfy successful in predicting the ratio of Th 230/U. sociated with control system failures in a General Electnc boil-234 in bones, and that effective release rate of uranium from ing water reactor, PNL also conducted valuehmpact analyses of skeleton may be more rapid than predicted by the ICRP model, proposed modifications of these control systems to prevent these failures These analyses were based on failure modes NUREG/CR 4383: HIGH PRESSURE INJECTION OF MELT and effects analyses prevousty performed by the Idaho Nat on-FROM A REACTOR PRESSURE VESSEL - THE DISCHARGE at Engineenng Laboratory (INEL) The control system failure , PHASE. PILCH,M ; TARBELL,W W. Sandia Natonal Laborato- modes identafied by INEL and analyzed by PNL fall into three j nes September 1985. 4tpp 85f1070237. SAND 85-00f 2. ' main scenanos 1) failures that initiate feedwater overfall and 33380 262 also defeat the high level feedwater trip,2) a fadure of the con. Recent probabilistic nsk-assessment studies identified poten- densate booster pump that results in increased flow to the tial accident sequences in which reactor vessel failure occurs vess (ovwfill), and 3) an inadvertent actuation of the low pres-while the pnmary system is at elevated pressure. The phenome- sure coviant injection system (LPCI) that also produces an ex-nology of the discharge phase is rev,ewed here We propose an ces.svo enoldown (overcool) For each of these modes, two faii-emproved model for hole ablation following vessel fanfure, and ure sequences were postulated The results of PNL's probabilis-we compare the model with expenment data Gas blowthrough tc analysis of failure progresson to core damage and value/ is ident.fied as a mechanism that allows steam to escape impact analyses of possible resolutions to prevent the occur-th'ough the vessel breach before melt ejection is complete. Gas rence of these failures are presented in this report. bluwthrough leads to pneumatic atomizaton of the remaining melt before significant depressuntation of the pnmary system NUREG/CR 4388: AEROSOL BEHAVIOR MODELING (TASK 3)- occurs. SUPPORT SERVICES FOR RESEARCH AND EVALUATION OF NUREG/CR 4385: EFFECTS OF CONTROL SYSTEM FAILURES SEVERE ACC! DENT PHENOMENA AND MITIGATION. IN TRANSIENTS. ACCIDENTS, AND CORE MELT FREQUEN- JORDAN.H . KOGAN V. Battelle Memonal Institute. Columbus Labcratones. October 1985 78pp 8511110431. BV12130. CIES AT A WESTINCHOUSE PRESSURIZED WATER REAC. TOR BiCKFORD,W E.; TABATABAl.A S. Batteile Memonal in. 334 t 7 0 f 6 stitute. Pacif;c Northwest Laboratories November 1985. 95pp This repnr1 covers emploratory research done on a number of aerosol topics relevant to nuclear safety dunng the penod 5/1/ 8512050467. PNL 5543 33769 098 84 4/30/85. Much of this research rnquered the modification Pacific Northwest Laboratory (PNL) performed a probablistic nsk assessment to develop estimates of core-mett frequency and development of the OUICKM code to accomodate steam and public nsk due to control system failures in a Westinghouse condensation / evaporation at wait and particle surfaces in a more mechanistic fashon than previousfy possible. The pnnce-pressunzed water reactor. Valuehmpact analysis of proposed systems modifications to prevent the control system failures pal finding of this effort is that steam condensation on aerosol were also performed. Four control system failure modes were particles is a very sensitive function of the numencal model and anatyred 1) overfill, 2) overcoof, 3) overpressure, and a) steam that all present approaches are probably inadequate for its generator tube rupture. For each mode, two failure sequences proper assessmen'. Further work es needed in this area before were postulated These analyses were based on the results of prod!ctons uf aerosol growth by steam condensaton can be failure modes and effects analyses prevously performed at bel'eved Another aerosof topic approached with the auf of the Idaho Natonal Engineenno Laboratory and conducted in sup. QUICKM code was that of the coagglomeraton of aerosol parti-port of the U S. Nuclear Regu!atory Commission's program for cles of different types, as might occur in LWR meltdown se-Unresolved Safety Issue A-47 Safety implications of Control quences when the aerosol generated by Core / Concrete interac-tion mingles with the existing radioactive aerosol of the contain-Systems ment. This study showed that a " multiple component" ap. NUREG/CR 4388: EFFECTS OF CONTROL SYSTEM FAILURES proach, such as that of OUICKM and CONTAIN can yield mark-ON TRANSIENTS. ACCIDENTS. AND CORE-MELT FREQUEN' edly different predictions to those of a " sing's component" ap-CIES AT A BABCOCK AND WiLCOX PRESSURIZED WATER proach as used in the NAUA code it also revealed additonal REACTO3 BiCAFORD.W E.; TABATABAl A S Battelle Memon- sensitivities of the multiple component approach to enput data al Institum, Pacific Northwest Laboratories December 1905 that are generany not well known, Finally, the possible effect on 62pp 851270329 PNL 5544 34084 259 aerosol behavior of the decay of fission products associated Pacific Northwest Laboratory (PNL) performed probabilistic with the aerosol particles in containment was investigated This nsk analyses aimed at developing estimates of core-melt fro
limited investigation revealed that decay can potentially affect quency and public risk associated with control system failures in acrosol behavior under some circumstances.
a Babcock and Wilcos pressunted water reactor, and value/ impact analyses of proposed systems modificatons These NUREG/CR 4392: MEASURES OF SAFEGUARDS AlbK EM-Enalyses were based on the results of failure modes and effects PLOYING PRA (MOSREP) A Methodology For Estimating Risk snalyses previously performed at the Oak Ridge National Labo- Impacts Of Safeguards Measures HORION.W ; LOUNER.P.; ratory (ORNL) The control system failure modes that were K A AIMIAN.S ; et al. Brookhaven Natonal Laboratory. October ident,fied by ORNL and analyzed by PNL fall into three train 198'i 257pp 8511190505 DNL NUREG $1926. 33542134 scenanos.1) overfill of the steam generators progressing to this report presents a method called Measures of Sabotage spillover into the steam lines, 2) ICS hand power failure pro- Risk Employing PRA (MOSREP) which was developed to sys-gressing to steam generator dryout, and 3) ICS automat c power tematically evaluate the desirability of sabotage vulnerability re-failure progress.ng to steam generator failure For each of these duct >on measures The method has been specifically designed modes, failure sequences were postulated. The results of PNL's to be a starting point for resolving Generic issue A 29 probabilistic analysis of failure progresson to core damage and MOSREP has been designed to provide a technical basis for value/ impact analyses of poss ble resolutions to prevent the oc- regulatory actons envolving design changes, damage control currence of these failures are presented en this report. measures, physical protection vandalism, and tampenng at nu-
Main Citations and Abstracts 101 clear power plants. It is the intent of MOSREP to provide sup- dominate all others in terms of cost impacts. The apphcable port to the NRC staff in determing effective regulatory strategies paragraph numbers from Draft E2 of the Appendix J revision and evaluating trade offs associated with acttvities which impact and the nature of the change follows: lil A(4) & lil.A(6) Test both safeguards and safety. Pressure & Ter, ting at Reduced Pressure No Longer Allowed. NUREG/CR-4395: CORRELATION OF CV AND K:C/KJC TRAN. 111 A(7)(b)(i) Acceptance Cntena 10 La Acceptable "As Found" SITION TEMPERATURE INCREASES DUE TO IRRADIATION Leakage; til A(8)(a) Retesting Following Failure of ' As Found" HISER.A. Matonals Engineenng Associates, Inc. November Type A Test Corrective Action Plan, and ill A(8)(b)(n) Opt.on to 1985. 94pp. 8512270355. NEA-2086. 34082 024. Do More Frequent Type B & C Testing Rather Than More Type Reactor pressure vessel (RPV) surveillance capsules contain A Penarty Tests The best estimate is that the proposed Appen-Charpy V (C(v)) specimens, but rnany do not contain fracture d's J would result in cost savings ranging from about $100 me toughness specimens; accordingly, the radiatiorunduced shift lion to $160 milton, and increase routine occupatonal exposure (increase) in the bnttle-to-ductile transation regon (tnangleT) is on the order of 10.000 person tem These estimates capture based upon the tnangfeT determined from notch ductshty (C(v) the total impact to industry and the NRC over the assumed op-tests. Since the ASME K(Ic) and KIIR) reference fracture tough- erating hfe of all en sting and planned future power reactors All ness curves are shifted by the tnangfoi from C(v), assurance dollar impacts projected to occur in future years have been that this triangleT does not underestimate tnangreT associated present worthed at discount rates ranging from 5*. to 10*. with the actual arradiated fracture toughness is required to pro-vide confidence that safety margins do not fall below assumed NUREG/CR 4399: POSSIBLE OPTIONS FOR REDUCING OCCU. levels. To assess this behavior, compansons of inanglers de- PA TIONAL DOSE FROM THE TMI2 8tSEMENT fined by elastic-plastic fracture toughness and C(v) tests have MUNSON,LF,; HARTY,R Battelle Memonal Inst tute. Pacific been made using data from RPV base and weld metals in which Northwest Laboratories November 1985 125pp 8512090545 irradiations were made pgL,gg$7. 33g yg 93$ companson of tnangleT,unders at vanoustest inderreactor conditions. levels gives indica- As well' The March 28,1970 accident at Three Mae Island Unit 2 fated tions of C(v), K(Jc) and K(84 curve shape change due to irra- the basement to a depth of several feet with highly contaminat. g g g diation. Lastly, comparisons between measured tnangfeT(C(v)) and values ueing vanous correlations or models are also made tion efforts are underway Dose rates range from approximatery 40 to more than 1100 R/hr. Identified sources includo a struc-NUREG/CR 4316: SIMMER POSTPROCESSOR MANUAL. ture of hollow concrete blocks that is estimated to contain be-PARKERI. Los Alamos Scientific Laboratory. November 1985 tween 11,000 and 19,000 cunes of comum 137 and other quan-14000 8512050431, LA 10549 M. 33770 010 The T6P postprocessor analyzes SIMMER-il TAPE 6 and t.f od sources that contain between 570 and 1800 additonal cunes Decontamination methods and approaches available for TAPE 36 data The processor calculates new vanabtes, inte- cleanup are discussed grates vanables over reg.ons of the mesh, compares values of a vanable of a given probicm at selected times, compares NUREG/CR 4400: THE IMPACT OF MECHANICAL.AND MAIN-values of a variable at selected times between problems for pa. TENANCE-INDUCED FAILURES OF MAIN REACTOR COOL-rameter studies, calculates denvatives of a vanable with respect ANT PUMP SEALS ON PLANT SAFETY AZARM.M A , to time or another venable, traces a vanable at a requested lo- BOCCIO.J L Brookhaven Natenal Laboratory MITRA,5 Impell cation :n the mesh over time The data are presented to the Corp. December 1985. 103pp 86010/0464 UNL NUREG-user graphically, using two-drmenseonal graphs and three-demen- $1928 34186 093 This document presents an tmestigation of sional perspective or contour plots. Interactive graphic tech. the safety impact resulting from mechanical- and maintenance-niques may be used with perspective plots induced reactor coolant pump (HCP) seal failures in nuclear
"* " #8 " ^d""" ' P""# " '##" ' ' # ^
NUREG/CR-439h IN-PLANT SOURCE TERM MEASUREMENTS AT PRAIRIE ISLAND NUCLEAR GENERATING STATION ing nuclear power plants in the U S from severat a<adabte MANDLER.JW; STALKER A C ; CRONEY,S T.; et al EG&G s wen wu p rn e annual ke@wy of pump wal Idaho, Inc. (subs. of EG8G, Inc ) Captomber 1985 565pp fa wo in a numar powe va was eshmated based on the 8510040592. EGG 2420. 32863172. c neept of hatard rate and dependoacy evaluation. The conds-This report presents data obtained at Praine Island as a part wat maWy of vam sm of W rah wn seM raden was paluaW W sa% gact of D wal fah, en of the in-Plant Source Term Measurement Program in operating light water reactors (LWRs) the work was conducted for the U" * * " ' Office of Nuclear Regulatory Research (RES) in support of the ##N * " Meteorology and Effluent Treatment Branch (METO) of the the n rmal makeup capacity and the impact of piant safety was Office of Nuclear Reactor Regulation (NRR) The pnmary objec-teve of this program is to provide the Nuclear Regulatory Com. al make up capacity, formal PRA methodologies were ap. messon (NAC) with operational data that can be used in evalua-tion of plant designs for liquid and gaseous radwaste troatment systems. Data presented were obtained at the Prasne Island Nu^ NUREG/CR-4401: DEHAVIOR OF CONTROL RODS DURING clear Generating Staton, operated by Northern States Power' CORE DEGRADATION PRESSURIZATION OF SILVER-located near Red Wing. Minnesota in-plant measurements were INDIUM CADM;UM CONTROL RODS POWERS D A Sandia conducted dunng the time penod from October 1980 to August National Laboratones November 1985 187pp 8512270269 1981. This plant is the fifth in a senes of operating LWRs to be SAND 85-0469 34n81001. studied. Activity data for the liquvf binary systems Ag-Cd Ag In, and In-Cd are correlated in terms of the Wilson equation. These cor. NUREG/CR 4398: COST ANALYSIS OF REVISIONS TO to CFR relations are used to construct a model of the tornary system PART 50, APPENDIX J, LEAK TESTS FOR PRIMARY AND Ag In-Cd Spect.uscupic data for the vapor spncies Aq(g), SECONDARY CONTAINVENTS OF LIGHT. WATER COOLED Ag(2)(g), Ag(3)(g), Ag > (g), Int g), in(2)(g). In + (gh Cd(g). NUCLEAR POWER PLANTS SCI ACCA.F ; NE LSON,W ; Cd(2)fgh Cd e (g), Agin(g), and Cdtnig) are reviewed and are SIMPKINS,8 ; et al Science & Engenconng Anociates, Inc used to def<ne thermodynamic funct one for these specins for September 1985. 92pp. 8510030131. 32849 332 temperatures between 298 and 1500 K Vapor pressures for the This report addresses the differences betwo,m the costing liquid pnaw pure einments, liquid benary alloys, and the liquid and proposed Appendix J and identifies eleven substantive ternary alloy are calculated uvng the Wilson r*quation model areas where quantiftable impacts will 14cly result. The analyvs and uvng the assumption that the condensed phase is an ideal indicated that there are four areas of Change which tend to mixture An ateotrope is predicted for the Ag in system Predic-
102 Main Citations and Abstracts tions are made of the vaponzation of alloys of 80 percent Ag. Expenmental data are reported on: Local condensation heat 15 percent in, and 5 percent Cd used as control matenals in transfer coefficients and local interfacial shear stresses for some pressun2ed water reactors. countercurrent stratified flow of steam and cold water at atmos-phenc pressure in a flat plate geometry at an inclit.aton of 4 NUREG/CR-4402 V01: HIGH TEMPERATURE GAS-COOLED RE- ees. m aM M %es kn @zontal Data am ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT cmW in Ws d N gas aM W ReynWs news. EVALUATION Ouarterfy Progress Report. January 1 March 31,1985. BALL.S J.; CLEVELAND.J C.; HARRINGTON.R M ; et NUREG/CR 4415: COUNTER CURRENT STEAM / WATER FLOW c1 Oak Ridge National Laboratory. October 1985 15pp. ABOVE A PERFORATED PLATE VERTICAL INJECTION OF 8512270361. ORNL/TM-9798/V1. 34078.329 WATER. DiLBER.I. Northwestern Univ., Evanston. IL. October l Modehng, code deve:opment, and analyses of the modular 1985. 62pp. 8511120087. 33441.283. High Temperature Gas-Cooled Reactor (HTGR) continued with Expenmental data are presented on steam / water counter-cut-work on the side-by-side design. Fission-product relesse and rent flow limiting phenomena. Weep points and total dumping transport expenments were completed. A desenption and as- points are determined for low and high water injection rates sessment report on Oak Ridge National Laboratory HTGR above a perforated plate. safety codes was issued. NUREG/CR 4416: STABILITY OF STEAM-WATER COUNTER. NUREG/CR-4403:
~~SUMMARY~~
OF THE WASTE MANAGCMENT CURRENT STRATIFIED FLOW. LEE.S C. Northwestern Univ., PROGRAMS AT URANIUM RECOVERY FACILITIES AS THEY Evanston, IL October 1985. 242pp. 8511120083. 33441:045. RELATE TO THE 40 CFR PART 192 STANDARDS. GILLEN,0.; The stabili./ of countercurrent flow of steam and water in in. BALDWIN.J S.; CAMPBELL.A.W.; et al. Oak Ridge National Lab- chned rectangular ducts is investigated Two flow instabihtees oratory. November 1985.270pp.8512270087. ORNL/TM-9797. which limit the normal condensation process in countercurrent 34076 001- stratifed flow have been identified expenmentally: flooding and This study evaluates the degree to which surface impound- condensation induced water hammer. Analyses of both conde-ments at licensed facihties compfy with significant changes in tions are performed on a basis of flow stabihty and heat transfer NRC requirements initiated by enactment of EPA's final environ. consideration. mental standards for uranium recovery facilities (40 CFR Part 192) Impoundment liner requirements, groundwater protection NUREG/CR-4417: LOCAL PROPERTIES OF COUNTERCUR-stIndards, groundwater monitonng and corrective action pro- RENT STRATIFIED STEAM. WATER FLOW. KIM.H.J. North-western Univ., Evanston, IL October 1985.227pp.8511120029 grams, and site closure standards are the most significant regu. litory modificatons. The comphance status of 30 conventional 33440.001. mills and 31 in-situ menes is determined from a review of Nucle- A study of condensation in countercurrent stratified flow of steam and subcooled water has been carned out in rectangular ar Regulatory Commission and agreement state docket files through November,1983. Results of this review show that a channel / flat geornetry over a wide range of inclinator angles majonty of conventional uranium mill tailings management sys. and aspect ratios. Local condensation rates and pressure gradi-tems are proposed expansions to existing impoundments, as ents were measured from which local heat transfer coefficients well as with respect to some aspects of groundwater monitonng and interfacial shear stresses were determined and correlated End comphance programs. Furthermore, the status of conven. in terms of Reynolds and Freid1 numbers. tional mall site closure plans as highly uncertain at this time. Al- NUREG/CR 4422: A REVIEW Or THE MODELS AND MECHA-though surface impoundments at in-situ uranium mines also are NISMS FOR ENVIRONMENTALLY ASSISTED CRACK deficient with respect to groundwater monitonng programs, they GROWTH OF PRESSURE VESSEL AND PIPING STEELS IN nera o I with other changes in requirements imposed PWR ENVIRONMENTS CULLE N.W ; G ABETTA.G ; HANNINEN.H Matonals Engineenng Associates. tnc. December NU .EG/CR 4406: AN ANALYSIS OF LOW LEVEL 1985.116pp 8601070455 MEA 2078. 34183 063-WASTES Review of Hazardous Waste Regulations And identifi- The crack-tip micromechanisms and the computational cation of Radioactive V med Wastes. Final Report. models for environmentally-assisted cracking in pressure vessel BOWERMAN.B S ; KEMPF.C.R.; MACKENZIE.D.R ; et al. Brook- and piping steels in high-temperature. Iow-oxygen (PWR). reac-haven National Laboratory. December 1985. 172pp. for grade water are descnbed and evaluated in this report. The 8601070500, BNL NUREG-51933. 34189 286 micromectanishc models are discussed in some detail, with Regulations promulgated by the U.S. Environmental Protec- anodic dissolution and hydrogen assistance being the pnme ton Agency governing the disposal of hazardous wastes were candidates for the successful explanation of the observed phe-reviewed A survey was carned out to estabbsh a data base on nemona The anodic dissolubon model offers far better quantifi-the nature and composition of low-level radioactive wastes cation of the environmentally assisted crack growth rats, but tends to overpredict the rates for a large number of constions (LLW) in crder to determine whether some LLW could also be considered hatardous as defined in 40 CFR Part 261. Complet. The hydrogen assistance models quahtatrvely could account for ed questionnaires were returned by 97 of the 238 teactor and a wide range of effects, but quant.fication of the model is virtu-nonreactor facshties contac'ed. The waste volumes reported by ally noneristent. A vanety of calculational models are en vanous these respondents corresponded to approximately 29*. of all stages of devebpment; all of them are far from use as a predic. LLW disposed of en 1984. The analysis of the survey results in- tive tool Crack tip strain rate modefs have received the most at. dicated that three broad categones of LLW may be radioactive tention, and the approach to their use has been to partition the mored wastes. They include: waste containing organic hquids, envvonmentally-assisted gmth rates into a mechanically-dnven disposed of by all types of generators; wastes containsng lead component with the envirocmental enhancement Superposed. metal, i e., discarded shielding or lead containers; and wastes The environment component i then correlated with a calculat-containeng chromates, ie., twclear power plant process wastes ed crachtip strain rate where Chromates are used as corroseon inhibitors. Certain NUREG/CR 4424: DROPLET SIZES. DYNAMICS AND DEPOSI-
" '*
~~LOPES,J C ;~~
TION IN VERilCAL ANNULAR FLOW. ntial med aste a DUKLER.A E. Houston. Univ. of. Houston. TX. October 1985 NUREG/CR 4414: DIRECT-CONTACT CONDENSATION OF 318pp. 8511110335 03411074 STEAM ON COLD WATER IN STRATIFIED COUNTERCUR- lodne release from a nuclear power plant during steam gen-RENT FLOW. BANKOFF,S G ; KIM.H.J. Northwestorn Univ , Ev. erator tube rupture accidents is expected to be prongfy de-Enston, lL. October 1985. 8 tpp. 8511120063. 33440 323 pendent on the drop sites formed as high pressure onmary
Main Citations and Abstracts 103 system water is flashed and atomized as it passes through the NUREG/CR 4432: COMPARISON OF DYNAMIC CHARACTERIS-rupture opening. This study was based on the need for informa. TICS OF FUKUSHIMA NUCLEAR POWER PLANT CONTAIN-ton on drop sizes formed under such conditions Expenments MENT BUILDING DETERMINED FROM TESTS AND EARTH-to measure the fraction cf water flashed and the drop sizes OUAKES. SRINIVASAN,M G ; KOT,C A.; HSIEH.B.J. Argonne formrd were performed at typical operating pressures and tem- National Laboratory. October 1985. 29pp. C511190570. ANL perr.tures with the actual tube diameters and lengths nearly to 67. 33534 004 scale. The mass median drop sizes measured were in the range . Modal parameters determined from response measured in dy-from about 20 to 60 micrometers for both open-ended and sht namec tests and from anafytical models for simulating the tests rupture geometnes No significant effect on drop size of pnmary and two subsequent earthquakes expenenced by the contain-system pressure level was noted over the range from t100 to ment Nilding of Unit 1 of the Fukushima Power Staton com-2100 psig. plex in Japan are compared for it a purpose of evaluating the effectmeness of the dynamic tests en earthquake response pre-NURFG/CR-4428: OVERVIEW OF TRAC-BD1/ MODI ASSESS- dicton. The tests are found to have led to the correct identifica-MENT STUDIES. CHARBONEAU.B L. EGaG Idaho, Inc. (subs tion of a fundamental frequency. The lack of agreement be-of EG&G, Inc.). November 1985. 53pp. 8512270264. EGG- tween test- and earthquake determined modeshapes and damp-2422. 34079 299. This report summanzes a senes of computer ing is attributable more to the shortcomings of the simulaton simulations sponsored by the United States Nuclear Regulatory m dels than to daemnces 6n actual beWr. Commisson'(USNRC) performed at the Idajo National Engineer-ing Laboratory (INEL) to continue the advancoment of boilen9 NUREG/CR-4435: ORGANIC COMPLEXANT-ENHANCED MO-BILITY OF TOXIC ELEMENTS IN LOW-LEVEL WASTES Annual water reactor (BWR) safety research. The simulations were per- Report. July 1984 - June 1985 SWANSON.J L. Battelle Memorb formed to evaluate the analysis capabilities of the Transient Re-al Institut A Pacific Northwest Laboratones December 1985 actor Analysis Code BWR verson (TRAC BD1/ MOD 1) to calcu- 68pp. 86]IO70474. PNL 4965-8. 34186 299 late operaton.sl transients, including anticipated transients with- This report contains the results of the second year's efforts of out scram (ATWS) and loss-of-coolant accidents (LOCAs). The a project whose objectwe is to determine how and to eat assessment simulations were performed for a broad range of extent organic complemants affoct the mobility of toxic elements scenanos, to encompass as many different phenomena as pos- in subsurface groundwater at commercial low level waste dis-sable Compansons are made between the measured and calcu- posal sites The complexants EDTA and picolinate, both of fated data. Conclusons are made with respect tc the calculated which are used en reactor decontaminaton operations, were system pressure response, thermal response, and break flow studied most extensmefy. Hydrous ferne oxide, Fe(2)O(3) H(2)O, response, as well as the capabalities to model containment and and kaolinite clay were the soil Components most used. Three natural circulation conditons. Recommendatons are made with toxic elements wers studied, Ne, Am. and Cd Ni and Am have respect to user guidehnes. radioactwe isotopas that are commonly present in commercial low level wastes, and Cd is an example of a nonradcactive NUREG/CR 4429: TRAC-BD1/ MOD 1 USERS GUIDELINE. toxic element that might also be in such wastes. A wide dworse. HANSON,R G. Idaho National Engineenng Laboratory Novem- ty of effects of organic compiezant*, on toxic elements sorption ber 1985. 77pp. 8512120107. ECG-2423. 33873114 *** * ' * **** b* * "Y "*" Code assessment s'udies and specific code applications have at some pH values, but others are not, important reactons are provided insight into the effective use of the TRAC-BWR senes slow in some systems but rapid in others There are two sepa-of codes. This document reports the expenence gained from rate reactions in which slow kinetics have been observed in some systems; one is the slow dissociation of a preformed the studies and serves to assist the user in the effective apple- comp!er and the other is the slow desorption by complesant so-cation of the TRAC-BDt/ MODI computer code. This document lutens of a previously sorted uncomplexed element. stresses the user's perspective relatme to appropriate use of the TRAC-BD1/ MOD 1 code and is considered an adjunct to NUREG/CR 4437: EXPLORATORY STUDIES OF ELEMENT other documentation provided with the code. INTERACTIONS AND COMPOSITION DEPENDENCIES IN RA-DIATION SENSITIVITY DEVELOPMENT, HAWTHORNE.J R NUREG/CR-4430: CURRENT METHODOLOGIES FOR ASSESS. Matonals Engineenny Associates. Inc. November 1985, 94pp ING THE POTENTIAL FOR EARTHOUARKE INDUCED LIOUE. 8512190250. MEA 2113. 34010 068 FACTION IN SOILS. KOESTER J P.; FRANKLIN.A.G. Army, This 6nvest gation mth laboratory melts of pressure vessel Dept. of, Army Engineer Waterways Expenment Station Octo. steels (A 533 8 or A 302-8 base) probes suspect interactions ber 1985. 69pp. 8511120176. 33439 087. between copper impurities and manganese, molybdenum, chro. The geotechnical engineenng hterature reflects continuing meum and nickel alloying as influencing elevated termperature, evolution of methnds for eva'uation of hauefaction potential. and radiation sensettvety development The investigation also quah-several significant advances have been achieved in the past fios the influence of phosphorus Content on radiaton sens'tivity few years, notabry in the areas of in situ testing and the use of as a function of copper content und explores suspect cor:tnbu-data from past occurrences of hauefacton, strain based ap- tons of n an arsenc Radiaton resistance es judged on the proaches, the steady state concept, and non hnear, effective asis of a n an a 88 de stress analysis. In the hght of new knowledge and the reezam- grees centigrade arradiaton to 2 5 x 10(19) n/cm(2), E > 1 ination of old data, hquef action occurrence is no longer behoved MeV The findings derNnstrate clearty that impnrtant composs-hon interachons esist in radiaton sensitivity development, to be restncted to relativery clean, uniform, loosoly-deposited, saturated sands, and a great deal of research emphasis has HUREG/CR 4439: TRAINING REVIEW CRITERIA AND PROCE-thus been given, or is proposed, to understanding the dynamic DURES.
~~Analysis & Technology, Inc June 1985. 125pp berhavior of saturated gravelly sods and fine grained soils with 8512310038. 34104 181.~~
some plasticity This report discusses conditions under which This report provides a set of training review cntena and pro-the potential for earthquake-induced hquefaction should be eval. cedures which constitute a systematic means of implementing usted, and desenbes procedures and cntena that are currently two moniotonng functions identified in the " Commission Pohcy apphed to assess the hquefaction potential of soils ranging in Statement on Training & Quahrications of Nuclear Power Plant gradaton from gravels to clays. Emphasis is given to several of Personnel" (March 20,1984, 50 FR 11147) These functons the more recent field, laboratory, and theoretical approaches are 1. Continuing evaluation of industrywide train!ng & quahfica-tion program effectmeness, and 2. monitonng plant and industry
104 Main Citations and Abstracts trends and events involving personnel errors. The procedures An evaluation of standby diesel generator performance at nu-are organared around the frve essential elements of perform- clear power plants between 1980 and 1983. All diesel generator ance - based training articulated in the Policy Statement. The vendors except Transamenca Delaval were evaluated. Material package was designed for use by NRC personnel engaged in reviewed included failure data, inspection reports and previous the review of performance-based trainsng programs in nuciear studies by others Charts and tables of data including manufac-power plants it has been published in a modified version in furer versus site location. Conclusion #s that diesel generator 2 NUREG/CR format to enable large - scale production and distn- performance and reliabihty is reasonably good, when TDI expen-bution for information purposes ence is factored out in addition, total loss of offsite power NUREG/CR-4440: A REVIEW OF EMERGENCY DIESEL GENER. events are decreasing, thus increased inspection activity at ATOR PERFORMANCE AT NUCLEAR POWER PLANTS. dieso! generator manufacturers is not required. SUBUDHl.M ; HIGGINS.J C. Brookhaven National Laboratory November 1985, 35pp. 8512050369 33768 355.
Contractor Report Number Index 1 j This index lists, in alphabetical order, the NUREG/CR for the report and to the 10-I contractorissued report codes for the NRC digit NRC Document Control System acces- ; contractor reports in this compilation. Each sion number. ' contractor code is cross referenced to the SECONDARY REPORT NUMSER REPORT NUMSER SECOte0ARY REPORT NUM9ER REPORT NUMSER 5528 NUAEG/CR 3990 BNL NUREG-5t844 NUAEG/CR 4069 83-E TEC-DAF.171 NUAEG/CR4294 BNL-NUREG 51846 NUAEG/CR4083 AECL MtSC *05 NUAEG/CR4077 BNL-NUAEG-51848 NUREG/CR-4093 ANL-84 102 NUAEG/CR-4180 BNL NUREG-51849 P4UAEGICR 4103 ANL 8416 NUAE G/CR-3710 DNL-NUREG-51857 NUREG/CR 4149 ANL-84-35 V03 NUREG/CA-3004 V03 BNL NUPEG-51858 NUREGe CR4152 ANL 84-35 V04 NUAEG/CR 3804 V04 BNL-NUREG 51861 NUREQ/CR4156 ANL 8441 NUAEG/CR 4064 BNL NUREG 51868 NUREG/CR 4200 ANL 84 80 NUAEG/CR-3998 V02 BNL-NURE G-S t B69 NUREG/CR 4201 ANL6441 NUAEG/CR-3998 V03 BNL-NUAEG 51073 NUAEG/CR 4215 ANL-84-61 NUAEG/CR-3980 V03 BNL-NUREG-51874 NUREG/CR-4221 ANL-84-61 NUAEG/CR-3980 V04 BNL NUREG-51876 NUAEG/CR 4228 QNL-84 41 V02 NUAEG/CR 3980 V02 BNL-NUAEG 51877 NUAEG/CR 4229 QNL 84 66 NUAEG/CR 3999 BNL NUREG-51878 NUREG/CR 4230 ANL 84-6/ NUREG/CR-3855 BNL-NUREG-518 79 NUREG/CR 4231 ANL 8513 NUAEG/CR-4191 BNL NUAEG-51887 NUAEG/CR 4253 ANL-85-20 NUREG/CR 420 4 BNL-NUREG 51888 NUAEG/CR-4254 j ANL-85 21 NUREG/CH-4208 BNL-NUREG 51893 NUAEG/CR4182 ANL-85 23 V01 NUREG/CR 4240 V01 BNL NUREG 51897 NUREG/CR4291 ANL-85 31 'i NUAEG/CR-4277 BNL NUAEG-51898 NUREGICR4292 DNL 85 33 NUREG/CR-4287 BNL-NUREG-51901 NUREG/CP-0068 ANL8546 NUREG/CR4368 BNL-NUREG-51906 NUREG/CH-4329 AE-85 5 NUREG/CR4 t24 BNL NUREG-St907 NUREG/CR 4f 43 ANL 85 67 NUREG/CR-4432 BNL-NUREG 51915 NUREG/CR4373 Ah/ES 143 NUREG/CR 4120 BNL NUREG-51924 NUREG/CP-0070 ARAP 472 NUREG/CR 4157 BNL-NUAE G-51926 NUHEG/CR 4392 AR AP NO. 504 NUArG/CH 4158 BNL-NUREG 51928 NVAEG/CR 4400 ARAP NO 505 NUREG/CR 4159 BNL NUAE G-51933 NURE G/CR4406 i ATR 85(5818)2 NUREG/CR 4133 BNL/NUAEG 51836 NUREG/CR 4050 4 BHARC400/84/02 NUAEG/CR4139 CREARE IN 384 NUAEG/CR 3426 V01 BMI 2114 NUAEG/CR 3937 CREAVE IN-384 NUREG/CH-3426 V02 BML2120 NUAEG/CR4082 V01 CGG 2164 NUREG/CR 2531 A03 BUL2120 NUREG/CR4082 V02 EGG 223 f NUREG/CR 3005 BML2121 NUREG/CR4172 EGG-2245 NUAEG/CH 3191 y BMI-2122 NUREG/CR 4173 EGO-2251 NUREG/CR-3237 EML2123 NUREG/CR-4177 V02 EGG 2259 NUAE G/GR-3301 BMI-2123 NUREG/CR 4177 V01 EGG-2294 NUAEG/CR-3633 V04 BMI 2124 NUREG/CR 4205 EGG 2294 NUREG/CR 3633 V01 St BMt 2t25 NUREG/CR 4210 EGG 2297 NUHEG/CA 3948 BML2126 NUAEG/CR4211 EGG 1308 NUAEG/CR1767
~~.l BMI 2127 NUAEG/CR-3900 V04 E GG-2317 NunEG/CH 3819 I~~
~~BML2129 NUAEG/CR 4377 EGG-2323 NUREG/CR 3862 BML2130 NUAEG/CR 4388 EGO 2335 NUAE G/CR-3935~~
' BWNUAEG 51267 NUAEG/CR 1677 V02 EGG-2341 NUAEG/CR 39 77 BNL-NUAE G-S t 454 NUREG/CR-2331 V04 N2 EGG 2352 NUR E G /CH-4033 BNL NUREG-51454 NUREG/CR 233t V04 N3 E GG-2154 NUREG/CR4041 BNL NUAEG 51454 NUAEG/CR 2331 V04 N4 EGG 2354 NUHEG/CR4041 ROI BNL-NUREG 51454 NUHEG/CR-2331 VOS N1 EGG 2355 NUAE G/CR-4 ta t BNL NUAEG-St494 NUAEG/CR-2482 VOT E GG-2357 NUAEG/CR 4046 BYt NUREG-51494 NUAEG/CR 2482 V06 EGG 2159 NUREG/CR 4056 BNL NUnEG 51494 NUREG/CR 2482 V09 EGG 2362 NURE G /CR-4071 BNL NUREG-St494 NURE G/CR 2482 V06 EGG 2363 NUAEG/CH 4073 4
BNL NUAEG 51559 NUAEG/CR 2815 V01 Rt EGG 2364 NUREG/CH 4074 BNL-NUAEG Sf 559 NUAEG/CR 2815 V02 Rf EGG 2365 NUHE G/CR-4077 BNL NURE451609 NUREG/CR 3026 EGG 2366 NURE G/CR 4080
BNL NUAEG-51630 NUAEG/CR 3091 V04 EGG 2367 NUHLG/CH 4084 1 BNL-NUREG 51630 NUAEG/CR 3091 V05 EGG 2369 NURE G/CR-4 t t 5 i BNL-NUAEG-51610 NUAEG/CR 1091 V06 EGG 2374 NUREG/CR4 t 41 BNL NUAEG-51899 NUAEG/CR 3444 V02 EGG 2376 NURE G/CR4150 BNL NUAEG 51708 NUREG/CR-3469 V02 EGG 2377 NUAEG/CR-4164 i BNL NUAEG-51110 NUAEG/CR 3485 EGG 2380 NURE G/CR4195 1 BNL NUAEG St113 NUAEG/CR 3498 EGG 2381 NUAEG/CH 4178 DAFT 1
BNL NUAEG 51717 NUREG/CR 3519 E G42382 NUAE G/CR 4196 BNL NUAEG St 750 NUREG/CR-3703 EGG 2381 NUAEG/CR 4203 BNL-NUREG 51784 NUREG/CR 3829 EGG 2317 NUREG/CH 4040 BNL NUAEG-51786 NUAEG/CR-3837 E GG1388 NUnE G/CR 4227 1 BNL-NUAEG 51792 NUAEG/CR 3865 EGO 2392 NUHf G/CR4245 1 BNL.NUREG 51795 NURFG/CH 3878 E GG1%4 NURE G/CH4262 V01 UNL NUAEG 51807 NUAEG/CR-3914 EGG 2394 NURE G/CA4262 V02 BNL NUREG 51812 NUAE G/CR-3943 EGG 2397 NUHE G/CH 4272 BNL NUREG 51821 NUAEGICP 0059 V01 EGO 2405 NURFG/CR4326 V0f DNL-NUAE G-51828 NUREG/CR 4016 V01 E GG 2405 NUHE G/CH4126 V02 BNL NUREG 5f 841 NURE G/CA4062 EGG 2406 NUAE G/CH 4351 BNL-NUAEG 51842 NUREGICR 4067 EGG 2409 NUHE G/CH 4345 BNL-NUREG-51843 NURE G/CR4068 EGG 2415 NURE G/CH-4375 105 1
106 Contractor Report Number index SECONOARY REPORT NUMSER REPORT NUMBER $ECONDARY REPORT NUMSER REPORT NUMBER EGG 2420 NUREG/CR4397 ORNUNSC200 NUREG/CR-2000 V04 N9 EGO-2422 NUREG/CR 4428 ORNUNSC200 NUREG/CR-2000 V04N10 EGG 2423 NUREG/C44429 ORNUNSC223 NUAEG/CR-3905 V03 EGG-MS-6708 NUAEG/CR-4212 ORNUNSC223 NUAEG/CA-3905 V02 EPRI NP 3802 NVAEG/CR 3426 V02 ORNUNSC223 NUREG/C43905 VO1 R1 EPRI NP-3802 NUREG/CR 3426 V01 ORNL/NSC223 NUAEG/CR-3905 V04 EPAl NP-3987 NUREG/CR4127 V01 ORNL/TM-8664/V2 NUREG/CR-3514 V02 EPRI NP-3987 NUREG/CR4127 V03 ORNUTM4869 NUREG/CR-3442 EPRI NP-3987 NUAEG/CR4127 V02 ORNUTM-9041/V2 . NUREG/CR-3626 V02 EPRI NP 3988 NUAEG/CR-4128 ORNL/TM 9067 NUREG/C43651 EPRI NP-4111 NUREGICR4166 ORNUTM-9150 NUREGIC43738 EPRINP4112 NUAEG/CA-4167 ORNUTM-9154/V2 NUAEG/CR-3744 V02 EPRI NP 4298 NUREC/CR 4376 ORNUTM-9163 NUREG/CR 3764 ESRP 7.1.1 NUAEG 1165 ORNUTM-9191 NUREG/C43851 V04 FEMA 51 NUREG 0981 RO1 ORNL/TM 9191/V3 NUREG/CR-3851 V03 GEAP-30875 NUREG/CR-4127 V01 ORNUTM 9216 NUREG/C43831 GEAP-30875 NUREG/CR-4127 V03 OANUTM-9253 NUREG/C43872 i NUREG/C43887 GEAP 30875 NUREG/CR-4127 V02 ORNUTM-9266 GEAP-30876 NUREG/CR4128 ORNUTM-9267/V3 NUREG/C43885 V03 GP-R 123022 NUAEG/CR 4258 ORNL/TM-9267/V4 NUREGIC43885 W GSCA-46 NUAEG/C44304 ORNUTM-9308/V2 NUREG/CR 3481 V02 HEDL TME 84 21 NUREG/CR-3746 V02 ORNL/TM 9316 NUREG/CR-3930 HEDL-TME 84-31 NUAEG/CR-3746 V03 ORNL/TM-9339/V1 NUREG/C43949 V01 NEDL-TME-85-3 NUREG/CR-3319 ORNUTM-9339/V2 NUREG/CR-3949 V02 IE-143 NUREG/C44003 ORNUTM-9383 NUAEG/C43991 i ORNUTM 9384 NUREG/CR 3992 lE 145 - NUREG/CR-4006
IE 146 NUREG/CR 4005 ORNUTM 9390 NUAEG/CR-4015 t IEB 79-04 NUREG/CR4003 ORNL/TM 9408 NUREG/CR-4022 lE8 79-09 NUREG/CR-3791 ORNL/TM-9423/V2 NUAEG/CR-40'll VC2 IEB-79-12 ~ NUREG 0905 ORNUTM-9423/V3 NUAEG/CR4031 V03 t
' J IEB-79-25 NUREG/CR-4004 ORNUTM 9437 - NUREG/CA4037 IEB4012 NUAEG/CR-4005 ORNUTM-9440 NUREG/CR4039 ' IE8-80-25 NUREG/CR-3794 OANUTM-9443 NUAEG/CR-4043 IEB4101 NUAEG/C44006 ORNUTM-9445 NUREG/CR4206 ' NUREG-1095 OANUTM 9477 NUREG/CR4086 IEB-82-02 NUREG/CR-4346 8
IS 4862 NUREG/CR4952 ORNUTM-9479 NUREG/C43706 ORNUTM-9488 NUREG/CR-4081 i LA 10055-MS NUREG/C44092 I LS10166-MS NUREG/C43866 ORNUTM 9491 NUREG/CR-3953 OANL/TM-9503 NUREG/C44104 LS10229-MS NUAEG/CR 4105
, , LA 10267 MS NUREG/CR-4020 ORNL/TM 9506 ' NUREG/CR4079 ORNUTM 9516 NUREG/CR-3978 LA 10307 MS NUREG/CR4134 r LA-10319-MS NUREG/CR-4107 ORNUTM-9522 NUREG/CA-4109 OANL/TM-9545 NUREG/CR-3634 LA10321-MS NUAEG/CR4256 LA10351 MS NUREG/CR4140 ORNL/TM-9574 NUAEG/CR-4217 ORNUTM 9585 NUREG/CP-0062 LA10396-MS NUREG/CR4219 V01 LA 10401 MS NUREG/CR-4225 ORNUTM-9593/V1 NUAEG/CR-4232 ORNL/TM 9614 NUAEG/CR4236 V01 LS10413-MS NUAEG/C44236 V02 LA10435 M NUREG/CR-4260 ' OANUTM 9614 NUAEG/C44264 OANL/TM-9612/V1 NUAEGICR 4255 VOt LS10436-MS NUREG/CR-4275 ' LS10443-MS NUAEG/CR-4274 ORNLITM 9654 NUREG/CR4278 OANL/TM-9660 NUREG/CR4280 LA10445-MS NUREG/CR4284 LA 10474 MS NUAEG/CR-4314 ORNL/TM 9664 NUREG/CR4321 ORNUTM 9682 IUREG/C44325 LA10478-MS NVAEG/CR4347 - LA-10548-MS NUREG/CR-3646 ORNUTM-9739 NUAEG/C44396 ORNUTM 9797 NUREG/CR4403 LA 10549-M NUAEG/CR4402 V01 LA 9700-MS NUAEG/C43208 ORNL/TM 9798/V1 ' NUREG/CA-40 5 PARAMETER IE 13 NUAEG/CR 3794 LA 9977 MS NUREG/CR 2850 W3 LA9985-MS NUAEG/CR-4111 PNL4221 NUAEG/CA4355 V01 PNL4742 NUAEG/C43317
, LBL 20022 NUAEG/CR 3413 LMF 11 NUREG/CR 3984 PNL-4790 NOREG/C43613 V03 Nt ' NUAEGICR4121 PNL-4941 MEA-2053 NUREG/C43609 MEA-2055 NUREG/CR-3945 PNL4942 NUAEG/CR 3228 V03 PNL 4965-8 NUREG/CR4435 MEA-2075 NUREG/CR-3613 V02 MEA-2078 NUAEG/CR-4422 PNL 4971 NUREG/CR4437 PNL4973 NUREG/CR-3659 ME A 2113 NUREG/CR-4023 (' , PNL 5005 R8SlH 85-3169 NUREG/CR4266 NUREGICR4361 REA-2086 NUAEG/C44395 PNL 5017 NUAEGICR4038 PNL 5064 NUHEG/C43747 l ORNL4114 PNL-5069 NUREG/C43752 ORNL4135 NUREG/CR4106 NUREG/CR-3810 V03 NUAEG/CA 4114 PNL 5106 3 ORNL4137 PNL 51064 NUREG/CR-3810 V04 OANL4163 NUREG/CR-4249 NUAEG/CR4234 V01 PNL 512' NUREG/CR-3817 i OANL4170/V1 PNL 5123 NUAEG/CR 3825 V03-4
~~ORNL4177 NUREG/CR-4304 NUREG/CR4367 PNL 5154 NUREG/CR4108~~
! ORNL4208 PNL 5156 NUAEGIC44268 OP.NL NSC200 NUREG/CR 2000 VG4N11 t NUREG/CR-3723 PNL 5158 NUREG/C43987 i ORNUCSD/TM 216 PNL-5160 NUREG/CR-3883 ORNUENG/TM-31 NUREG/CR4360 V01 NUAEG/CR-3906 NUREG/CR4360 V02 PNL-5179 ORNL/ENG/TM-31 PNL 5181 NUREG/C43911 V02 I ORNL/NOAC 214 NUREG/CR-3551 , NUREG/CR-3922 V01 PNL 5184 NUREGiCR 3915 ORNUNOAC-224 PNL 5210 NUREG/C43950 V01 ORNUNOAC 224 NUREG/CR-3922 V02 NUREG/CR 3972
. i :' ORNL/hSIC-200 NUPEG/CR-2000 V03N12 NUREG/CR 2000 V04 N1 PNL 5222 PNL 5245 NUREG/C43999 i ORNUNSC200 NVAEG/CR-2000 V04 N2 PNL 5299 NUREGICR4030
/ ORNUNSC200 PNL 5300 NUREG/CR4070 V02 09NUNSC200 NUAEG/CR-2000 V04 N3 NUREG/C44070 V03 NUREG/CR-2000 V04 N4 PNL 5300 OANUNSC200 PNL 5303 NUREG/CH4051 NUAEG/C42000 V04 N5 i: OANUNSC200 PNL 5318 NUREG/C44061 ORNUNSC200 NUREG/CR 2000 V04 N6 NUREG/CA 2000 Voi N7 PNL 5319 NUREG/CR-3709 ORNL/NSC200 PNL-5320 NUAEG/CR4057 ORNUNSC200 NUREG/CR-2000 V04 N8 l l i
~~r 1~~
~~l Contractor Report Number index 107~~
'(
SECONDARY REPORT NUMSER PNL-5323 NUREG/CR 4075 REPORT NUMBER SECONDARY REPORT NUMSER SAND 84-1461 REPORT NUNSER NUREG/CR-3904 - PNL 5324 NUREG/CR 4076 SAND 841522 NUREG/CR-3913 L PNL 5338 NUREG/CR-4088 SAND 841534 NUREG/CR-3919 PNL-5339 NUREG/CR4087 SAND 841646 NUREG/CR 3944 PNL 5340 NURtEG/CR-4089 SAND 841704 NUREG/CR 3954 PNL 5350 NUREG/CR-4100 SAND 841824 NUREG/CR-4199 ! PNL-5354 NUREG/CR-4168 SAND 841838 NUREG/CR-2951 PNL-5361 NUREG/CR-4118 SAND 84-1348 NUREG/CR-4008 , PNL 5374 NUREG/CR-4125 V01 SAND 84 2291 NUREG/CR-4091 I PNL-5374 NUREG/CR4125 V02 SAND 84 2305 NUREG/CR-4044 ! PNL-5379 NUREG/CR4130 SAND 84-2629 NUREGICR-4096 i ' PNL-5381 NUREG/CR-4139 SAND 84-2630 NUREG/CR-4097 PNL 5386 NUREG/CR-4176 SAND 84-2668 NUREG/CR-4110 *
' SAND 84 7115 NUREG/CR-3688 V01 PNL-5388 NUREG/CR 4153 PNL 5389 NUREG/CR-4144 SAND 84 7115 NUREG/CR 3688 V02 PNL-5392 NUREG/CR-4151 ' SAND 84 7139 NUREGICR-3855 , PNL-5404 NUREG/CR-4160 SAND 84-7177 NUREGICH-4064 i
~~PNL-5417 NUREG/CR-4362 SAND 85-0012 NUREG/CR 4383 5 /~~
PNL 5421 NUREG/CR4192 3 PNL-5432 NUREG/CR-4220 PNL-5433 NUREG/CR 4218 SAND 850135 NUREG/CR4138 SAND 85-0172 NUREG/CR-4155 PNL-5435 ' NUREG/CR-4248 SAND 85 0175 i PNL-5461 NUREG/CR-4251 V02 M IREG/CR-4137 SAND 85-0205 NUREG/CR-4146 PNL 5461 NUREG/CR4251 V01 SAND 85-0209 PNL 5467 HUREG/CR-4147
'NUREG/CR-4259 SANDe5-0283 NUREG/CR-4185 PNL-5469 NUREG/CR-4267 SAND 85 0469 NUREG/CR-4401 PNL 5477. NUREG/CR-4276 SAND 85-0576 NUREG/CR 4189 PNL-5487 NUREG/CR-4281 SAND 85-0634 NUREG/CR-4197 PNL 5490 NUREG/CR-4296 SAND 85-0679 NUREG/CR-4200 PNL 5509 NUREG/CR-4297 SAND 85-0935 NUREG/CR4250 I PNL 5511 NUREG/CR-4300 V01 SAND 85-1339 NUREG/CR4335 PNL-55161 ' / NUREG/CR4318 V01 SAND 851495 NUREG/CR-4350 V02 ' ~
PNL 5516 2 NUREG/CR 4318 V02 SAND 851495 NUREG/CR-4350 V04 PNL 5543 NUREG/CR-4385 SAND 851495 ' NUREG/CR-4350 V01 PNL 5544 NUREG/CR-4386 SAND 851495/3 NUREG/CR-4350 V03 PP$L 5545 NUREG/CR-4387 SAND 851495/5 NUREG/CR 4350 V05 , PPO 5557 NUREG/CR 4399 SAND 851495/6 NUREG/CR-4350 V06 SAND 82-1105 NUREG/CR.2718 SAND 85-1495/7 NUREG/CR-4350 V07 SAND 82 2156 NUREG/CR-3611 SAND 85-1557 NUREG/CR-4358 SAND 83-0395 NUREG/CR 3197 V01 SAND 851606 NUREG/CR-4340 V01 SAND 83-0501 NUREG/CR-3912 SAND 85-7150 NUREG/CR4009 l SAND 83-1326 NUREG/CR-3361 SAND 85-7151 NUREG/CR 4010 SAND 83-2621/1 NUREG/CR-3721 V01 SAND 85 7185 NUREG/CR-4214 SAND 83 2657 NUREG/CR 4136 ' SAND 85 7192 NUREG/CR-4303 SAND 83 2675 NUREG/CR4213 UCID 19988 NUREGICR 3660 V03 l SAND 83-7450 . NUREG/CR-3537 UCID 19988 NUREG/CR-3660 V01 SAND 84-0048 NUREG/CR-3647 UCIO 19988 V04 NUREG/CR-3660 V04 1 SAND 84 0060 NUREG/CR 3638 UCID+20092 NUREG/CR 4161 V01
' SAND 844186 NUREG/CR-3657 UCID-20164 NUREG/CR-4123 SAND 84-0806 NUREG/CR-3803 UCID-20398 NUREG/CR 4239 SAND 84 0814 NUREG/CR-3757 UCrD-20444 NUREG/CR 4334 SAND 84-0884 NUREG/CR-3772 UCfD-20468 NUREG/CR 4331 SAND 841013 NUREG/CR-3802 UCRL 20410 NUREG/CH 4263 SAND 841025 NUREG/CR-3820 V03 UCRL 53044 NUREGICR-3019 SAND 841072 NUREG/CR 3816 V01 UCRL 53455 NUREG/CR-3558 SAND 84-1072 NUREG/CR-3816 VO2 UCRL 53500 V01 NUREG/CR 3663 V01 SAND 841072 NUREG/CR-3816 V03 UCRL 53544 NUREG/CR-3854 SAND 84-1072 NUREG/CR-3816 V04 UCAL 53587 NUREG/CR-4035 SAND 84-1122 NUREG/CR-3936 UCRL-53644 NUREG/CR-4290 V02 SAND 84-1144 NUREG/CR-4055 US 751 NUREG/CR-4112 V01 SAND 841204 NUREG/CR4085 US 752 NUREG/CR-4112 V02 SAND 84-1264 NURE6/CR-3863 WCAP 10375 NUREG/CR4166 SAND 84-1307 NUREG/CR-4342 WCAP 10926 NUREG/CR 4167 SAND 84-1367 NUREG/CR-4060 WINCO 1024 . NUREG/CR-3455 SAND 641404 NUREG/CR-4169 WWL/TM-1791-2 ' NUREGICR 3901
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~~! i 4~~
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I l l
l l Personal Author index This index lists the personal authors of NRC report (s) prepared by the author, if informa-staff and contractor reports in alphabetical tion is needed, refer-to the main citation by order. Each name is followed by the the NUREG number. NUREG number and the title of the ABEL,K.H. ALPERT.DJ, NUREG/CA-4030: RADIONUCUDE MIGRATION IN GROUND WATER.(Final Report) NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY PROTECTIONill SUBMICRON PARTICLE TESTS AND
~~SUMMARY~~
ABRAHAM,T. OF QUAUTY FACTORS. NUREG/CR 2482 V09. REVIEW OF DOE WASTE PACKAGE NUREG/CR 3657: PRELIMINARY SCREENING OF FUEL CYCLE AND PROGRAM Semiannual Report Covenng The Penod Apnl 1985-Sep- gy. PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING tomber 1985- NUREG/CR4199. A DEMONSTRATON UNCERTAINTY / SENSITIVITY ANALYSIS USlNG THE HEALTH AND ECONOMIC CONSEQUENCE MODEL CRAC2 NUREG 488 V03: OAHO FIELD EXPERIMENT 1981. Volume NUREG/CR-4214' HEALTH EFFECTS MODEL FOR NUCLEAR POWER 3 panso Trajectones.Concentraton Patterns And MESODIF l O $ ^^ f Summa P W Ba s or H ADAMS.K.G. NUREG/CR4250: VEHICLE BARRIERS EMPHASIS ON NATURAL FEA- ALVARE2,J.L TURES NUREG/CR4033. THE ROLE OF PERSONAL AIR SAMPLING IN RADI-AD AMS.R.E. ATON SAFETY PROGRAMS AND RESULTS OF A LABORATORY EVALUATION OF PERSONAL AIR-SAMPUNG EQUIPMENT. RUREG/CR 3830 V02: AEROSOL RELEASE AND TRANSPORT PROGRAM. Semiannual Progress Report For Apnl 1984-September AMES.K.R. 1984 . NUREG/CR 3987: COMPUTERIZED ANNUNCIATOR SYSTEMS NUREG/CR4255 V01: AEROSAL RELEASE AND TRANSPORT PRO- NUREGICR4220: REUABluTY ANALYSIS OF CONTAINMENT ISOLA. GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 TION SYSTEMS. MARCH 1985.. ADAMS,S.R. AMICO.PJ. WUREG/CR-4375: THEORY.OESIGN.AND OPERATON OF LIQUlO NUREG/CA-4334. AN APPROACH TO THE QUANTIFICATON OF SEIS-MIC MARGINS IN NUCLEAR POWER PLANTS. METAL FAST BREEDER REACTORS. INCLUDING OPERATONAL HE ALTH PHYSICS.. ANAVIM.E. CDDES$10,F.L. NUREG/CR-4050: A REVIEW OF THE SHOREHAM NUCLEAR POWER WUREG/CR4278 TRAC-PF1/ MOD 1 DEVELOPMENT ASSESSMENT. STATION PROBABILISTIC RISK ASSESSMENT. internal Events And Core Damage Frequency. AmSOMA G.S. NUREG/CM174; ROCK MASS SEAUNG EXPERIMENTAL ASSESS. ANDERSENJ G MENT OF BOREHOLE PLUG PERFORMANCE. Annual ReportJune NUREG/CR 127 VO1: BWR FULL INTEGRAL SIMULATION TEST (FIST) PROGRAM TRAC-BWR MODEL DEVELOPMENT Volume 1.Nu. mencal Methods AHMADJ. NUREG/CR4127 V02. BWR FULL INTEGRAL G 74ULATION TEST NUREG/CR4082 V01: DEGRADED PIPING PROGRAM PHASE (FIST) PROGRAM TRAC-BWR MODEL DEVELOPMENT. Volume NR /C 08 2 EGA EC N P OGRAM - PHASE NURE R4127 V03- BWR FULL 6NTEGRAL SIMULATON TEST
~~1. Semiannual Report. October 1984 March 1985.~~
(FIST) PROGRAM TRAC-BWR MODEL DEVELOPMENT. volume 3 De-AHMED S. velopment Assessment For Plant Apphcation. t 4 NUREG/CR4257: INSPECTION. SURVEILLANCE,AND MONITORING ANDERSON,8.S. OF ELECTRICAL EQUIPMENT INSIDE CONTAINMENT OF NUCLEAR POWER PLANTS-WITH APPLICATIONS TO ELECTRICAL CABLES NUREG/CR4178 DRFT: AN EVALUATON OF SELECTED UCENSEE EVENT REPORTS PREPARED PURSUANT TO 10 CFR 50 73 Draft AKERS D.W. NUREG/CR-4245: IN. PLANT SOURCE TERM MEASUREMENTS AT ANDERSON,J.L NUREG/C - 9 IN P S U CE AM MEASUREMENTS AT ~ FL ER MENTS PRAIRIE ISLAND NUCLEAR GENERATING STATION. ANDERSON.R.F. NURE /CR4174: ROCK MASS SEAUNG EXPERIMENTAL ASSESS- NUREG/CR4094- FIELD EXPERIMENT DETERMINATIONS OF DISTRI. MENT OF BOREHOLE PLUG PERFOAMANCE. Annual ReportJune O" FF CIEN S OF ACTINOF ELEMENTS IN SULFATE 1983. May 1984' EN NUREG/CR 4237: MOOluTY OF RADIONUCUDES IN HIGH CHLOROE ALAMGIR.M. ENVIRONMENTS. NUREG/CR 4128: BWR FULL INTEGRAL SIMULATION TEST (FIST) A E PHASE fl TEST RESULTS AND TRAC-BWR MODEL QUAUFICAT;ON. R ALBA.C. EVmTO OF RGMG M E Em n
02. Degradation Of Threaded Fasteners in Reactor Coolant Pressure NUREG/CR-4091: THE EFFECT OF ALTERNATIVE AGING AND ACCI. Boundary Of Pressu12ed Water. Reactor Plants DENT SIMULATIONS ON POLYMER PROPERTIES.
ANDRAE,R.W. ALBERT,M.F. NUREG/CR4260: TORAC USER'S MANUAL A Computer Code For Ana-NUREG/CR4081: ABSORPTION OF GASEOUS ODINE BY WATER lyzing Tornado induced Flow And Matenal Transport in Nuclear Facels. OROPLETS ties. 109
110 Personal Author Index ANOREWS,W.8. NUREG/CR4400: THE IMPACT OF MECH' ANICAL AND MAINTE-NUREG/CR-2800 S03. GUCEUNES FOR NUCLEAR POWER PLANT NANCE-INDUCED FAILURES OF MAIN REACTOR COOLANT PUMP SAFETY ISSUE PRIORITIZATON INFORMATION DEVELOPMENT. SEALS ON PLANT SAFETY. 4 W a8G/CR-4153: APPLICATIONS OF FOREIGN PROBABluSTIC SAFETY ASSESSMENT EXPERIENCE TO THE U S. NUCLEAR REG- SADALAMENTE.R. ULATORY PROCESS NUREG/CR-3817: DEVELOPMENT.USE AND CONTROL OF MAINTE-NANCE PROCEDURES IN NUCLEAR POWER PLANTS Problems And NU 4148: SIMULATION OF AN EPRI-NEVADA TEST SITE 6NTS) HYDROGEN BURN TEST AT THE CENTRAL RECEIVER TEST BAILEY,W.J. FACILITY. NUREG/CR 3959 V01; FUEL PERFORMANCT ANNUAL REPOHT FOR 83' A~.MANTROUT G A. NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LE/EL BAKER.D.A. NUREGICR-2850 V03. POPULATION DOSE COMMITMENTS DUE TO NURE /C AN YS S$ k BILITY OF CURRENT HEALTH CTIVE RELEASES FROM NUCLEAR POWER PLANT SITES PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INOlVID-UALS. BALDWW.C.A. ARMSTRONG'G NUREG/CR-4284. NEUTRON EXPOSURE PARAMETERS FOR TFE NUREG/CR 4594 LOW-LEVEL NUCLEAR W4TE SHALLOW LAND FIFTH HEAVY SECTON STEEL TECHNOLOGY IRRADIATON BURIAL TRENCH ISOLATON Final ReportCctober 1981. September SERIES. 1984 SALDWIN.J.S. AINOLD.W.D. NUREG/CR4403:
~~SUMMARY~~
OF THE WASTE MANAGEMENT PRO-NUREG/CR-3851 V03. PROGRESS IN E /ALUATON OF RADf0NU. GRAMS AT URAN!UM RECOVERY FACIUTIES AS THEY RELATE TO CLOE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH. THE 40 CFR FART 192 STANDARDS. LEVEL NUCLEAR WASTE RESPOS! TORY SITE PROJECTS. Report For Aprd-June 1984. BALL.D.G. NUREG/CR-3851 V04. EVALUATON OF RADIONUCUDE GEOCHEMI. NUREG/CR 3723 STRESS-INTENSITY FACTOR INFLUENCE COEFFl. CAL INFORMATION DEVELOPED RY COE HIGH-LEVEL NUCLEAR CIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS.
. WASTE REPOSITORY SITE PROJdCTS Annual Progress Report For NUREG/CR4022: PRESSURIZED THERMAL SHOCK EVALUATION OF October 1983-September 1984 THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT.
J NUREG'CR-4114. VALENCE EFFECTS ON THE SORPTION OF NU. NUREG/CR4249 PRESSURE VESSEL FRACTURE STUDIES PENE- , CUDES ON ROCKS AND MINERALS 11. TRATING TO THE PWR THERMAL-SHOCK ISSUE. EXPERIMENTS NUREG/CR4236 V01: PROGRESS IN EVALUATION OF RADIONU- TSE 5.TSE-5A AND TSE-6 CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH. NUREG/CR4304' PRESSURE VESSEL FRACTURE STUDIES PER. LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS REPORT TAINING TO THE PWR THERMAL-SHOCK ISSUE Expenment TSE-7. FOR OCTOBER-DECEPBER 1984 NUREG/CR-4325- A PARAMETRIC STUDY OF PWR PRESSURE NUREG/CR-4236 V02: PROGRESS IN EVALUATON OF RADONU- VESSEL INTEGRITY DURING OVERCOOUNG CLOE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH- ACCOENTS.CONSOERING BOTH 2-D AND 3-D FLAWS. - LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report for t January-March 1985= ggLL,p y, NUREG/CR4344 INSTRUCTIONAL SKILLS EVALUATON IN NUCLE-ATTERIDGE.D.G. AR INDUSTRY TRAINING. NUREG/CR-3613 V02: EVALUATON OF WELDED AND REPA R-WELDED STAINLESS STEEL FOR LWR SERVICE. Ant'ual Report For B A LL,$.J. 1984. NUREG/CR.3685 V03. HIGH TEMPERATURE GAS-COOLED REACTOR NUREG/CR-3613 V03 N1; EVALUATION OF WELDED AND REPAIR. WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATON Ouarteny Progress Report July 1-September 30.1984 For October 1984 Through March 1985 NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR NUREG/CR-3911 V02- EVALUATON OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Ouarterty SAFETY STUDIES FOR THE DIVISION OF ACCIDENT EVALUATION Ouarterly Progress Report. October 1 Decernber Report,Apni-June 1984. 31.1984. j AULT,C.H. NUREG/CR 4402 V01: HIGH TEMPERATURE GAS-COOLED REACTOR NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE SAFETY STUDIES FOR THE DIVISION OF ACC! DENT MINES OF SOUTHWESTERN INDIANA-EVALUATON Ouartetty Progress Report January 1 March 31,1985. AUSTIN.P A. .BANDYOPADHYAY NUREG/CR4022; PRESSURIZED THERMAL SHOCK EVALUATION OF NUREG/CP-0070; PROCEED!NGS OF THE WORKSHOP ON SEISMIC i AND DYNAMIC FRAGILITY OF NUCLEAR POWER PLANT COMPO-THE CAi. VERT CUFFS UNIT 1 NUCLEAR POWER PLANT. AUTRY,V. NUREG/CR-4352: SUGGESTED STATE REQUIREMENTS AND CRITE' BANKOFF.S G. RIA FOR A LOW-LEVEL RADCACTIVE WASTE DISPOSAL SITE NUREG/CR4414: DIRECT-CONTACT CONDENSATION OF STEAM ON REGULATORY PROGRAM. COLD WATER IN STRATIFIED COUNTERCURRENT FLOW. 0A N EG/CR4210 MATADCR A COMPUTER CODE FOR THE ANALY- N RE /CR-3660 V03. PROBABluTY OF PIPE FAILURE IN THE REAC-N ^ TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTSVolume ENTS N L GH ATER REA ORS 3 Guillotine Break indirectly induced By Earthquakes. NUREG/CR-4211: MATADOR (METHODS FOR THE ANALYSIS OF NUREG/CR-3663 V03: PROBABILITY OF PIPE FAILURE IN THE REAC. TRANSPORT AND DEPOSITION OF RADONUCUDES) CODE DE- TOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR SCRIPTION AND USER'S MANUAL PLANTS. Volume 3. Double Ended Guillotine Break Indirectly Induced By Earthquakes. AYERS.A.L NUREG/CR-3237; CONTROL OF EXPLOSIVE MIXTURES IN PWR BARANOWSKI,P.W. WASTE GAS SYSTEMS. NUREG 1032 DAFT FC: EVALUATION OF STATION BLACKOUT ACCI-DENTS AT NUCLEAR POWER PLANTS. Technical Findings Related To AZARM M.A. Urvesotved Safety Issue A44 Draft Repor1 For Comrnent. NUREG/CR4229: EVALUATON OF CURRENT METHODOLOGY EM-d PLOYED IN PROBABluSTIC RIS< ASSESSMENT (PRA) OF FIRE 8ARI,R A-EVENTS AT NUCLEAR POWER PLANTS. NUREG/CR 2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE. NUREG/CR4230 PROBABluTY BASED EVALUATON OF SELECTED DURES GUOE. Sections 17 And Appendices. FIRE PROTECTION FEATURES IN NUCLEAR POWER PLANTS. l
Personal Author index III BARLETTA R.E. NUREG/CR43'3: COMPENDIUM OF COST-EFFECTIVENESS EVALUA. NUREG/CR 3829: AN EVALUATION OF THE STABluTY TESTS REC. TIONS OF MODIFICATIONS FOR DOSE REDUCTION AT NUCLEAR OMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE POWER PLANTS. FORMS AND CONTAINER VATERIALS. NUREG/CR-3865: EVALUAT ON OF THE RAD.OACTIVE INVENTORY BAUMANN.B.L IN.AND f STIMATION OF ISOTOPIC RELEASE FROM.THE WASTE IN NUREG/CR-4090 EVALUATON OF NUCLEAR FACluTY DECOMMIS. EIGHT TMNCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL SiONING PROJECTS. Annual Sumanary Repc t . Fescal Year 1984. SITE. NUREG/CR-4069: ANALfSES OF SOfLS FROM AN AREA ADJACENT BAXTER,D.E. TO THE LOW LEVIL RADCACTIVE WASTE DISPOSAL SITE AT SHEFFIELD.lLUNOIS- NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR4083: ANALYSES OF SOILS FROM THE LOW-LEVEL RA. ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC DOILING WATER REACTOR Main Report DIOACTIVE WASTE DISPOSAL SITES AT SARNWELL.SC AND NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES NUR 2 BCDEGRADATION TESTING OF SOUDIFIED LOW- ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC LEVEL WASTE S1 REAMS. ^ ^ ^E **' BARNER,J.O. BAY BUTT,P. NUREG/CR-3999: ELECTRICALLY HEATED EX-REACTOR PELLET. NUREG/CR-4210: MATADOR A COMPUTER CODE FOR THE ANALY. CLADDING INTERACTON (PCI) SIMULATONS UTILIZING IRRADIAT. SIS OF RADf0NUCLIDE BEHAVOR DURING DEGRADED CORE AC. ED ZlRCALOY CLADDING. CIDENTS IN UGHT WATER REACTORS. BARNES,C.R. - NUREG/CR-4211: MATADOR (METHODS FOR THE ANALYSIS OF TRANSPORT AND DEPOSITION OF RADIONUCUDES) CODE DE. NUREG/CR-4082 V01: DEGRADED PIPING PROGRAM - PHASE SCalPTION AND USER'S MANUAL. Il Sermannual Report. March 1984 September 1984. NUREG/CR-4082 V02: DEGRADED PIPING PROGRAM . PHASE BEAL,S. II.Sermannual Report, October 1984. March 1985-NUREG/CR4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART BARNES.M.G. R J. STS M NW AW SENW NUREG/CR-4118: MONITORING METHODS FOR DETERMINATION CONTAINMENTS OF LIGHT WATER-COOLED NUCLEAR POWER COMPUANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT PLANTS. URANIUM RECOVERY SITES. BEARE,A.N. BARNES,V.E. NUREG/CR-4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND NUPEG/CR-3817: DEVELOPMENT.USE AND CONTROL OF MAINTE. ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW NANCE PROCEDURES IN NUCLEAR POWER PLANTS Problems And PERFORMANCE ;N CONTROL ROOM SIMULATORS. BECKHAM,R. BARRETT,R. NUREG/CR4111: A COMPARATIVE STUDY OF HEPA FILTER EFF1-NUREG/CR4143: REVIEW AND EVALUATON OF THE MILLSTONE. CtENCIES WHEN CHALLENGED WITH THERMAL. AND AIRVET. UNIT 3 PROBABIUSTIC SAFETY STUDY. Containment Failure GENERATED DI 2-ETHYLHEXYL SEBECATE.DI 2-ETHYLHEXYL Modes. Radiological Source Terms And Offste Consequences. PHTHALATE,AND SODIUM CHLORIDE. BARTTER,W.D. BECKMAN.R.J. NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PERFORMANCE NUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER StMULATION (MAPPS) MODEL: DESCRIPTION OF MODEL PLANT VALVE FAILURE. RATE VARIABluTY-SOME PREUMINARY CONTENT. STRUCTURE,AND SENSITIVITY TESTING RESULTS' NUREG/CR-4104: MAINTENANCE PERSONNEL PERFORMANCE SIM-ULATION (MAPPS) MODEL Feld Evaluation / validation. BEE.R.W. BASHAM.P.W. NUREG/CR-4357; THE FE.*SIBUTY OF OETECTING THE IMPOPT OF NUREG/CR-4317 V01: CANADIAN SEISMIC AGREEMENT. Technical UNAUTHORIZED RADICAC1tVE MATERIALS INTO THE UNITED Report Covenng 19?9-198y STATES. GASS.B.R. BE E BE.M.R. NUREG/CR-3723 STRESS-INTENSITY-FACTOR INFLUENCE COEFFI- NUREG-0020 V09 N04. LICENSED OPERATING REACTORS STATUS CIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS
~~SUMMARY~~
REPORT. Data As Of March 31.1985 (Gray Book I) NUREG/CR4106. PRESSbRIZED-THERMAL. SHOCK TEST OF 6-IN. NUREG 0020 V09 N06 LICENSED OPERATING REACTORS STATUS THICK PRESSURE VESSELS.PTSE-1: Investigation Of Warm Prestress-
~~SUMMARY~~
~~REPORT. Data As Of May 31.1985 (Gray Book l) ing And Upper-Shelf Arrest. NUREG.0020 V09 N07. UCENSED OPERATING REACTORS STATUS~~
~~SUMMARY~~
~~REPORT. Data As Of June 30.1985 (Gray Book I) BATTLE,R.E-NUREG-0020 V09 T9. LICENSED OPERATING REACTORS STATUS NUREG/CR-3991: FAILURE MO"E5 AND EFFECTS ANALYSIS (FMEA)~~
~~SUMMARY~~
~~REPOHT. Data As Of August 31.1985.(Gray Book 1) OF THE ICS/NNI ELECTRIC FOWER DISTRIBUTION CIRCUITRY AT NUREG-0020 V09 N11: UCENSED OPERATING REACTORS STATUS THE OCONEE-1 NUCLEAR PLANT.~~
~~SUMMARY~~
REPORT. Data As Of October 31,1985 (Gray Book 1) NUREG/CR-3992: COLLECTION AND EVALUATION OF COMPLETE AND PARTIAL LOSSES OF OFF-SITE POWER AT NUCLEAR POWER BEEDLOW.P.A. 4347; EMERGENCY DIESEL GENERATOR OPERATING " NU EXPER:ENCE.1981-1983. ML f fP E R NUREG/CR-407tx DETERMINATON OF COMPUANCE WITH CRITERIA BAUGH.J.W. FOR FINAL TAILINGS DISPOSAL SITE RECLAMATION. NUREG/CR-4276. VIBRATION AND WEAR IN STEAM GENERATOR BS FOLLOWING CHEMICAL CLEANING . SEMIANNUAL BE R . G/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCI.' DENTS IN AN ICE CONDENSER CONTAINMENT. BAUM J.W. NUREG/CP 0066: PROCEEDINGS OF AN INTERNATIONAL WORK. BELL.J. SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION ' NUREG-1127; RADIATION PROTECTION TRAINING AT URANIUM HEX-(ALARA) AT NUCLEAR POWER PLANTS,MAY 29-JUNE 1.1984 AFLUORIDE AND FUEL FABRICATION PLANTS NUREG/CR-3469 V02: OCCUPATIONAL DOSE REDUCTION AT NU-CLEAR POWER PLANTS. Annotated Bibhography Of Selected Read. BENEDICT,R. ings in Radiation Protection And ALARA NUREG 1142: TECHNICAL SPECIFICATONS FOR RIVER BEND NUREG/CR-4254: OCCUPATIONAL DOSE REDUCTON AND ALARA STATION Docket No. 50-458 (Gulf States Utihties Company) AT NUCLEAR POWER PLANTS Study On High Dose Jobs.Radwaste NUREG 1172. TECHNICAL SPECIFICATIONS FOR RIVER BEND STA-Handkng.And ALARA Incentives. TION. Docket No. 50458 (Gulf States Utihtes Company) 4
112 Personal Author Index SENNETT.D.E. NUREG/CR4272: RESPONSE TREE EVALUATION EXPERIMENTAL NUREG/CR-3657: PRELIMINARY SCREENING OF FUEL CYCLE AND ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR BY-PRODUCT MATERIAL LICENSES FOR EMERGENCY PLANNING. OPERATORS. SENNETT,J.J- BLAND,W.M. NUREG/CR-3516: A SURVEY OF THE USES OF RADIOACTIVE MATE- NUREG/CR-4271: RECOMMENDED SAFETY. RELIABILITY.OUALITY RIALS IN LOUISIANA S OFFSHORE WATERS. ASSURANCE AND MANAGEMENT AEROSPACE TECHNIQUEF WITH POSSIBLE APPLICATION BY THE DOE TO THE HIGH LEVEL RADIO-
^
NURE CR 3774 V02: ALTERNATIVE METHODS FOR DISPOSAL OF r LOW LEVEL RADIOACTIVE WASTES. Task 2A. Technical Requrements BLENCOE.J.G. NUREG/CR-4236 V02: NU EG C 7 VO A V . IS L OF E MHEW, PROGRESS WWWINDM&ED EVALUATION BY M OF & RADIONU-LOW LEVEL RADIOACTIVE WASTES. Task 28: Technical Regterements LEVEL NUCLFilt P'h1E REPOSITORY SITE PROJECTS. Report for For Aboveground Vault D,tposal Of Low Level Rseactwe Waste. NUREG/CR-3774 V04: ALTERNATIVE METHOD FOR CISPOSAL OF > :% . ah1985. LOW LEVEL RADIOACTIVE WASTE. Task 2C: Technical Reg.w" BLOND R.M For Earth Mounded Concrete Bunker Disposal Of Low Levet naf.oac-NUREG/CR-4197: SAFETY GOAL SENSITIVITY STUDIES. tive Waste. NUREG/CR-3774 V05: ALTERNATIVE METHODS FOR DISPOSAL OF BLUHM D LOW LEVEL RADCACTIVE WASTE. Task 2E. Technical Requrements NUREG/CR 3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE. For Shaft Disposal Of Low Level Radioactue Waste. SOAROMAN,T. BENSON,$.M NUREG/CR4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE RUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL REACTOR COOLANT PUMP. WASTE REPOSITORY. Volume 1:Basaft. 80CCIO,J.L.
~~; BERGERON.K.D.~~
NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer NUREG/CR4229 EVALUATON OF CURRENT METHODOLOGY EM-Code for Severe Reactor Accident Containment Analysis. PLOYED IN PROBABILISTIC RISK ASSESSMENT (PRA) OF FIRE EVENTS AT NUCLEAR POWER PLANTS. SERGGREN.R.G. NUREG/CR4230: PROBABILITY-BASED EVALUATION OF SELECTED NUREG/CR-4015. EFFECT OF STAINLESS STEEL WELD OVERLAY FIRE PROTECTION FEATURES IN NUCLEAR POWER PLANTS. CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL NUREG/CR4231: EVALUATION OF AVAILABLE DATA FOR PROBABI-PLATES IN BENDING SERIES 1. LISTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR NUREG/CR-4092. ORNL CHARACTERIZATION OF HEAVY-SECTION POWER PLANTS. STEEL TECHNOLOGY PROGRAM PLATES 01.02.AND 03. NUREG/CR4400: THE IMPACT OF MECHANICAL.AND MAINTE-NANCE-INDUCED FAILURES OF MAIN REACTOR COOLANT PUMP BEYER.C.E. SEALS ON PLANT SAFETY. NUREG/CR-4168: GT2FA COMPUTER ' CODE FOR ESTIMATING LIGHT WATER REACTOR FUEL ROD FAILURES- BOEGEL,A.J. NUREG/CR 3883. ANALYSIS OF JAPANESE-U.S. NUCLEAR POWER E lCR 1677 V02. PIPING BENCHMARK' PROBLEMS. VOLUME il DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION R E- BOhN.M.P. . SPONSE SPECTRUM METHOD. NUREG/CR-3558: HANDBOOK OF NUCLEAR POWER PLANT SEISMIC NUREG/CR-4291: CONCLUSION AND
~~SUMMARY~~
REPORT ON PHYSI- FRAGluTIES. Seismic Safety Margins Research Program. CAL BENCHMARKING OF PIPING SYSTEMS. NUREG/CR4335: POTENTIAL BENEFITS OdTAINED BY REQUIRING SAFTEY-GRADE COLD Sh:JTDOWN SYSTEMS. BHATTACHARYYA
~~' NUREG/CR-4208: GASTROINTESTINAL ABSORPTION OF PLUTONIUM BOLAND,J.F.~~
IN MICEJtATS, AND DOGS. Application To Establisheng Values Of f1 NUREG/CR4191: SURVEY OF LICENSEE CONTROL ROOM HABIT-For Soluble Plutonium. ABILITY PRACTICES. SICKFORD.W.E* M.A. WUREG/CR-2800 S03: GUIDELINES FOR NUCLEAR POWER PLANT BOLANDER'R-3977: NUREG/C RELAP5 THERMAL-HYDRAUUC ANALYSES OF i SAFETY ISSUE PRIORITIZATION INFORMATON DEVELOPMENT. PRESSURIZED THERMAL SHOCK SEQUENCES FOR N B. ROBIN-RUREG/CR-4385. EFFECTS OF CONTROL SYSTEM FAILURES IN SON UNIT 2 PRESSURIZED WATER REACTOR. TRANSlENTS. ACCIDENTS, AND CORE-MELT FREQUENCIES AT A WESTINGHOUSE PRESSURIZED WATER REACTOR. C MUREGICR4386: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS.ACCOENTS. AND CORE-MELT FREQUENCIES AT A
' BOLD NURE F'G/CR - : OFF-SITE 3413 CONSEQUENCES OF RADIOLOGICAL ACCOENTS METHODS, COSTS AND SCHEDULES FOR DECON-BABCOCK AND WILCOX PRESSURIZED WATER REACTOR. TAMINATION. .NUREG/CR4387; EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS.ACCOENTS AND CORE MELT FREQUENCIES AT A BOLT S.E j GENERAL ELECTRIC PRESSURIZED WATER REACTOR. NUREG/CR-4106: PRESSURIZED THERMAL SHOCK TEST OF 6-IN.-
THICK PRESSURE VESSELS PTSE-1 investigation Of Warm Prestress-StHL,N.K. . ing And Upper Shelf Arrest. NUREG/CR4397: IN PLANT SOURCE TERM MEASUREMENTS. AT NUREG/CR-4249: PRESSURE VESSEL FRACTURE STUDIES PENE-PRAIRIE ISLAND NUCLEAR GENERATING STATON. TRATING TO THE PWR THERMAL SHOCK ISSUE EXPERIMENTS CLLINGER,G.A. TSE-5.TSE 5A AND TSE 6. NUREG/CR 4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA NUREG/CR 4304: PRESSURE VESSEL FRACTURE STUDIES PER-TAINING TO THE PWR THERMAL-SHOCK ISSUE Expenment TSE-7. AND EASTERN TENNESSEE EARTHOUAKES (1978-1984). SINNALL,EP. BONZON,LL NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL NUREG/CR4095: TEST SERIES 2. SEISMIC-FRAGILITY TESTS OF WASTE REPOSITORY. Volume 1; Basalt. NATURALLY AGED CLASS 1E EXOE FHC 19 BATTERY CELLS. NUREG/CR4096: TEST SERIES 3. SEISMIC FRAGILITY TESTS OF BLACKMAN.H.S. NATURALLY AGED CLASS 1E C&D LCU-13 BATTERY CELLS. NUREG/CR 376/: INTERACTIVE SIMULATOR EVALUATION FOR CRT- NUREG/CR4097: TEST SERIES 4 SEISMIC FRAGILITY TESTS OF GENERATED DISPLAYS NATURALLY. AGED EXOE EMP.13 BATTERY CELLS. MUREG/CR4040- OPERATONAL DECISIONMAKING AND ACTION SE-LECTION UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER 800KER,C.P. PLANTS. NUREG/CR-3646 TRAC.PF1 INDEPENDENT ASSESSMENT. l L
Personal Author index 113 80RKOWSKI,R.J. BRUEMMER.S.M. NUREG/CR-3831: THE IN-PLANT RELIA 8luTY DATA BASE FOR NU-CLEAR PLANT COMPONENTS intenrq Report - Diesel NUREG/CR-3613 V02: EVALUATION OF WELDED AND REPAIR-Generators.Battenes, Chargers And inverters. WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Report For 1984. NUREG/CR-3613 V03 N1: EVALUATON OF WELDED AND REPAlR-NURE [CR-3887: HUMAN FACTORS REVIEW FOR SEVERE ACCl- fo, '"'"""' DENT SEQUENCE ANALYSIS, ago, hfT hM 98 80WERMAN,8.S. NUREG/CR-3911 V02. EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE.Ouarterty l NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATIONS ReportAprNune 1984. I ON SOUDIFICATION, WASTE DISPOSAL AND ASSOCIATED OCCU-PATONAL EXPOSURE BRUSKE*S'J' NUREG/CR-3829: AN EVALUATON OF THE STA8tuTY TESTS REC, NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES OMMENDED IN THE BRANCH TECHNICAL POSITON ON WASTE ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC FORMS AND CONTAINER MATERIALS BOIUNG WATER REACTOR Main Report NUREG/CR4406: AN ANALYSIS OF LOW-LEVEL WASTES Review of NUREG/CR-4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES Hazardous Wasle Regulations And identification of Radioactive Mmed Wastes. Final Report ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC 801UNG WATER REACTOR. Appendices. BOWERS.D.L NUREG/CR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES
~~. NUREG/CR-3710 LABORATORY STUDIES OF A BREACHED NUCLE- ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR Main Report.~~
AR WASTE REPOSITORY lN BASALT. NUREG/CR4326 V02: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE NUREG C 3646: TRAC.PF1' INDEPENDENT ASSESSMENT. ^ RUREG/CR-4140- DOMINANT ACCIDENT SEQUENCES IN OCONEE 1 BRUST F. PRESSURIZED WATER REACTOR. 80ZARTH,D. NUREG/CR4082 V02- DEGRADED PIPING PROGRAM PHASE II.Serrmannual Report, October 1984 - March 1985. NUREG/CR4022: PRESSURf2ED THERMAL SHOCK EVALUATION OF THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT, SRYAN R.H. ( BRACKEN 8USH,L. NUREG/CR-4106: PRESSURIZED-THERMAL-SHOCK TEST OF 6-IN- [ THICK PRrSSURE VESSELS.PTSE 1. Investigation Of Warrn Prestress-NUREG/CR4297: EXTREMITY - MONITORINGConsiderahons For ing And Mer-Shelf Arrest. Use. Dosimeter Placement.And Evaluahnn. BRAILE LW. NUREG/CR-3723. STRESS-INTENSITY. FACTOR INFLUENCE COEFFl. MUREG/CR-3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES OF CIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. POSSIBLE EXTENSONS OF THE NEW MADRO FAULT NUREG/CR-4106: PRESSURFZED THERMAL. SHOCK TEST OF 6-IN, ZONE. Annual Report For 1983. THICK PRESSURE VESSELS PTSE 1:Inveshgahon Of Warm Prestresn-BREEDfMG,RA ing And Upper. Shelf Arrest NUREG/CR-4350 V01: PROBA8ILISTIC RISK ASSESSMENT COURSE NUREG/CR-4367: ORVIRT.PC:A 2 D FINITE ELEMENT FRACTURE DOCUMENTATION. Volume 1 PRA Fundarnentals. ANALYSIS PROGRAM FOR A MICROCOMPUTER. BREWSTER.C- BUCHANAN,M.E. NUREG/CR 2482 V08:. REVIEW OF DOE WASTE PACKAGE NUREG/CR4256: MEASUREMENT OF RESPONSE TIME AND DETEC-PROGRAM.Serruannual Report Covenng The Penod October 1984 - TON OF DEGRADATION IN PRESSURE SENSOR / SENSING UNE March 1985. SYSTEMS. NUREG/CR-2482 V09: REVIEW OF DOE WASTE PACKAGE PROGRAM. Semiannual Report Covering The Pened Apnt-Septernber CUCKALEW W.H. M85^ NUREG/CR-4147: THE EFFECT OF ENVIRONMENTAL STRESS ON SYLGARD 70 SiUCONE ELASTOVER. BRODSKY,A. NUREG 1127: RADIATION PROTECTION TRAINING AT URAN!UM hex, BUDNITZ,R.J. AFLUORIDE AND FUEL FABRIC % TION PLANTS. NUREG/CR-4334. AN APPROACH TO THE QUANTIFICATION OF SEIS-NUREG 1134: RADIATION PROTECTON TRAINING FOR PERSONNEL ' MIC MARGINS IN NUCLEAR POWER PLANTS. EMPLOYED IN MEDICAL FACluTIES. BURKE.R.P. SROEK,D. NUREG/CR4197. SAFETY GOAL SENSITIVITY STUDIES. NUREG/CR4082 V01: DECRADED PIPING PROGRAM - PHASE . 11 Semsannual ReportMarch 1984 - Septembe 1984, BURNS,E.L MUREG/CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE ll.Serniannual Report. October 1984 - March 1985. NUREG/CR4156: OPtERATING EXPERIENCE AND AGING SEISMIC ASSESSMENT OF ELECTRIC MOTORS. BROOKS.S.G. BURTT,J.D. NUREG-0713 V05: OCCUPATONAL RADIATON EXPOSURE AT COM-MERCIAL NUCLEAR POWER REACTORS - 1983 ANNUAL REPORT. NUREG/CR-3977: RELAP5 THERMAL-HYORAUUC ANALYSES OF NUREG-0714 V04-05: OCCUPATONAL RADIATION PRESSURIZED THERMAL SHOCK SEQUENCES FOR H 8. ROBIN-EXPOSURE Fifteenth And Sateenth Annual Reports.1982 And 1983. SON UNIT 2 PRESSURIZED WATER REACTOR. BROOKSHIRE,R.L NUREG/CR 4115: INTERNATIONAL STANDARD PROBLEM 13 (LOFT EXPERIMENT L2-5). Final Companson Report. NUREG/CR4191: SURVEY OF LICENSEE CONTROL ROOM HABIT-A8luTY PRACTICES. BUSCH8ACH,T.C. BROWNJ. NUREG/CR-4226: NEW MADRID SEISMOTECTONIC STUDY.Actnnhos Dunng Fiscal Year 1983-NUREG/CR-4190: CAUFORNIA OFFSHORE SURVEY OF L.CENSEES USING RADIOACTIVE MATERIAL BUSUK,A.J. BROWZIN,8.S - NUREG/CR-2815 Vol R1: PROBABluSTIC SAFETY ANALYSIS PROCE-DURES GUOE. Sections 1-7 And Appendices. NUREG/CP-0065: TRANSACTIONS OF THE 8TH INTERNATIONAL CONFERENCE ON STRUCTURE MECHANICS IN REACTOR BUSTARD.LD. TECHNOLOGY. Panel Session J-K: Status of Research in Structural And Mecharucal Engineenng For Nuclear Powe Plants. NUREG/CH-4091: THE EFFECT OF ALTERNATIVE AGING AND ACCl-DENT SIMULATIONS ON POLYMER PROPERTIES.
114 Persoral Author Index BUXTON.LD. CASTO.W.R. NUREG/CR-3802 RELAP5' ASSESSMENT.OUANTITATIVE KEY PA- NUREG/CR 3551: SAFETY IMPUCATONS ASSOCIATED WITH IN-RAMETERS AND RUN TIME STATISTICS. PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS-NUREG/CR4044. TRAC-PF1 LOCA CALCULATIONS US6NG FINE- TEMS IN NUCLEAR POWER PLAN TS. NODE AND COARSE-NODE INPUT MODELS. CATE J.H. 8YERS,K.R. NUREG/CR-3488 V03: IDAHO FIELD EXPERIMENT 1981.Volurne NUREG/CR 4298. DESIGN AND INSTALLATION OF COMPUTER SYS- 3 Cornpanson Of Trasectones. Concentration Pa' terns And MESODIF TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55. Model Calculations. CATHEY,N.G. CADDINGTON.P'46; NUREG/CR-36 TRAC-PF1 INDEPENDENT ASSESSMENT. NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE-QUENCE INFORMATION. CAGLE.R.J. NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM CAZZOLI.E.G. D SEARCH SYSTEM NUREG/CRM8: NDWSONAL NEN & WA-SUBAS-NU EG 5V . EQU NCE IN SEMBLY HEAT TRANSFER AND BUOYANCY. INDUCED FLOW RE. FOR UCENSEE EVENT REPORTS Coder's Manual . DISTRIBUTON IN LMFBAS. NUREG/CR-3905 V04. SEQUENCE CODING AND SEARCH SYSTEM FOR UCENSEE EVdNT REPORTS Coder's Manual. CERBONE,R.J. CAMERON,J.R. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATION NUREG/CR 4131. INVESTIGATION OF ALTERNATIVE MEANS TO AC- " PUSH THE GOALS OF B3ENNIAL ION CHAMBER CAllBRA- NUF /CR 4201: THERMAL STABluTY TESTING OF LOW-LEVEL WASTE FORMS. . NUREG/CR-4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL CAMP,A.L. WASTE FORM ACCEPTANCE CRITERIA. NUREG/CR 3912 MARCH-HECTR At3ALYSIS OF SELECTED ACCl-DENTS IN AN ICE. CONDENSER CONTAINMENT. CHANDRASEKARAN NUREG/CR-3913. HECTR VERS 10N 1.0 USER'S MANUAL NUREG-0017 RO1: CALCULATON OF RELEASES OF RADIOACTIVE CAMP 8 ELL,A.W. MATERIALS IN GASEOUS AND UQUID EFFLUENTS FROM PRES-SURIZED WATER REACTORS (PWR-GALE CODE). NUREG/CR-4403:
~~SUMMARY~~
OF THE WASTE MANAGEMENT PHO. GRAMS AT URANIUM RECOVERY FACluTIES AS THEY RELATE TO CHANG M.T. THE 40 CFR PART 192 S" ANDARDS GROWTH IN BWR PIPING SYSTEMS CAMP 8 ELL R.D. RUREG/CR-3558: HANDBOOK OF NUCLEAR POWER PLANT SEISMIC CHANG T.Y. FRAGIUTIES. Seismic Safety Margins Research Prograrn. NUREG-1030 DFT FC: SEISMIC QUAUFICATION OF EQUIPMENT IN NUREG/CR 3660 V03. PROBABlWTY OF PIPE FAILURE IN THE REAC. OPERATING NUCLEAR POWER PLANTS. Unresolved Safety issue A-TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume 46 Draft Report For Comment
~~'3 Guillotine Rreak Indirectly Induced Earthquakes.~~
NUREG/CR-36.3 V03: PROBABlWTY PIPE FAILURE IN THE REAC-CHAO.B.T. TOR COOLANT LOOPS OF COMBUSTION ENGINEERING DWR NUREG/CR-3989: TIME AND VOLUME-AVERAGED CONSERVATION - PLANTS. Volume 3 Double Ended Guillotine Break Indirectfy Induced By EQUATONS FOR MULTIPHASE FLOW.Part One Systern Without in- 'I Earthquakes. ternal Solid Structures. RUREG/CR 4290 V02 PROBABILITY OF PIPE FAILURE IN THE REAC. TOR COOLANT LOOPS OF BABCOCK AND "/lLCOX PWR CHAR AU . PLANTS. Volume 2. Guillotine Break Indirectfy induced By "arthquakes. CANIANO,R.J. STUDIES. NUREG-1153: INSPECTON REPORT OF UNAUTHORIZED POSSES- CHARLOT,L.A. SiON Af 40 USE OF UNSEALED AMERICIUM-241 AND SUBSEQUENT NUREG/CR 3613 V03 N1: EVALUATON OF WELDED AND REPAIR-CONFISCATION. J.C. Haynes Company. Newark. Ohio. WELDED STAINLESS STEEL FOR LWR SERVICE.*emiannual Report For October 1984 Through March 1985. CAREW.J.F. NUREG/CR-4152: AN INDEPENDENT SAFETY ORGANIZATON NUREG/CR-3193: FORCED CONVECTIVE.NONEOUILIBRIUM, POST. CARFAGNO.S.P. RUREG/CR 4257: INSPECTON. SURVEILLANCE.AND MONITORING CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION OF ELECTRICAL EQUIPMENT INSIDE CONTAINMENT OF NUCLEAR COMPARISON REPORT. POWER PLANTS-WITH APPUCATIONS TO ELECTRICAL CABLES. NUREG/CR-4331: SiMPUFIED SEISMIC PROBABluSTIC RISK ASSESSMENT.P ocedures And Lmtations. CARHART,R.A. CHEN.T. HUREG/CR-4120: MATHEMATICAL MODEUNG OF ULTIMATE HEAT SINK COOUNG PONDS. NUREG/CR-4351:
~~SUMMARY~~
REPORT FOR LOFT ANTICIPATED TRANSIENT EXPERIMENT SERIES L6-8. CARUN.F. CHENG.H.S. 19UREG/CR-4091: THE EFFECT OF ALTERNATIVE AGlNG AND ACCI-NUREG/CR-3943: THE BWR PLAN ANALYZER. DENT SIMULATIONS ON POLYMER PROPERTIES. CHENION,J. CASADA.M.L NUREG/CR-4091: THE EFFECT OF ALTERNATIVE AGING AND ACCl-NUREG/CR-3922 V01: SURVEY AND EVALUATION OF SYSTEM INTERACTON EVENTS AND SOURCES Main Report And Appendices DENT SIMULATO1S ON POLYMER PROPERTIES. A And B. CHEUNG,Y.K. NUREG/CR-3922 V02: SURVEY AND EVALUATON OF SYSTEM NUREG/CR-4127 V0J BWR FULL INTEGRAL SIMULATION TEST INTERACTON EVENTS AND SOURCES. Appendices C And D. (FIST) PROGRAM TRAC-BWR MODEL DEVELOPMENT. Volume CASE,F.L 2 Models NUREG/CR-4114. VALENCE EFFECTS ON THE SORPTON OF NU- NUREG/GH-4127 V03: BWR FULL INTEGRAL SIMULATION TEST CUDES ON ROCKS AND MINERALS.ll (FIST) PROGRAN TRAC-BWR MODEL DEVELOPMENT. Volume 3.De-velopment Assessment For Plant Application. CHEVERTON,R.D. NUREGICR-4352: SUGGESTED STATE REQUIREMENTS AND CRITE-RIA FOR A LOW LEVEL RADIOACTIVE WASTE DISPOSAL SITE NOREG/CR 3723: STRESS-INTENSITY-FACTOR INFLUENCE COEFFl-REGULATORY PROGRAM. CIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS.
Personal Author Index 115 NUREG/CR4022- PRESSURIZED THERMAL SHOCK EVALUATION OF CLAYTOR,T.N. THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. NUREG/CR4124: NDE OF STAINLESS STEEL AND ON-LINE LEAK NUREG/CR-4249: PRESSURE VESSEL FRACTURE STUDIES PENE- MONITORING OF LWRS. Annual Report. October 1983 September TRATING TO THE PWR THERMAL-SHOCK ISSUE. EXPERIMENTS 1984.
~~.TSE 5,TSE-SA AND TSE-6.~~
NUREG/CR-4368. NDE OF STAINLESS STEEL AND ON-LINE LEAK NUREG/CR-4304: PRESSURE VESSEL FRACTURE STUDIES PER. MONITORING OF LWRS: Semaannual ReportOctober 1984 March TAINING TO THE PWR THERMAL SHOCK ISSUE.Expenment TSE-7. 1985. NUREG/CR-4325; A PARAMETRIC STUDY OF PWR PRESSURE VESSEL INTEGRITY DURING OVERCOOUNG CLEVELAND J.C. ACCOENTS.CONSOERING BOTH 2-D AND 3-D FLAWS. NUREG/CR-3885 V03: HIGH TEMPERATURE GAS-COOLED REACTOR CH4NN,0.J. SAFETY STUDIES FOR THE DIVISION OF - ACCOENT l EVALUATION Ouarterfy Progress Report. Jufy 1-September 30.1984. i^ NURFG/CR-3660 V04: PROBABluTY OF PIPE FAILURE 6N THE REAC- NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS-COOLED REACTOR l TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume SAFETY STUDIES FOR THE DIVISION OF ACCOENT
~~4. Pipe Farfure induced By Crack Growth in West Coast Plants. EVALUATON Quarterly Progress Report, October t December 31.1984.~~
CHO,N Z - NUREG/CR4402 V01. HIGH-TEMPERATURE GAS-COOLED REACTOR NUREG/CR-2815 V01 R1: PROBABILISTIC SAFETY ANALYSIS PROCE. SAFETY STUOJES FOR THE DIVISION OF ACCOENT DURES GUOE. Sections 1-7 And Appendees. EVALUATION Ouarterty Progress Report, January 1 March 31,1985. NUREG/CR-3485: PRA REVIEW MANUAL. CUFF,W.C.
~~; CHOCKIE,A.D.~~
NUREG/CR-3659: A MATHEMATICAL MODEL FOR ASSESSING NE t NUREG/CR-3883: ANALYSIS OF JAPANESE-U.S. NUCLEAR POWER UNCERTAINTIES OF INSTRUMENTATION MEASUREMENTS FOR j PLANT MAINTENANCE. POWER AND FLOW OF PWR REACTORS. CHOPRA,OL CUNE.J.E. NUREG/CR-4204: LONG-TERM EMBRITTLEMENT OF CAST DUPLEX NUREG/CR-4101: ASSAY OF LONG-UVED RADIONUCLOES IN LOW-STAINLESS STEELS IN LWR SYSTEMS Annual Report,0ctober 1983 - LEVEL WASTES FROM POWER REACTORS September 1984. CUNE.J.F. CHOU.C1 NUREG/CR 4076: DETERMINATION OF COMPLfANCE WITH CRITERIA NUREG/CR-3660 V01:. PROBABILITY OF PIPE FAILURE IN THE REAC- FOR FINAL TAluNGS DISPOSAL SITE RECLAMATION TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 1 Summary Report CLOUGH.R.L NUREG/CR-3663 V01: PROBABluTY OF PIPE FAILURE IN THE REAC- NUREG/CR4008. GENERAL EXTRAPOLATION MOOEL FOR AN IM-TOR COOLANT LOOPS OF COMBUSTON ENGINEERING PWR PORTANT CHEMICAL DOSE. RATE EFFECT. PLANTS. Volume 1: Summary Report NUREG/CR-4358: APPUCATONS OF DENSITY PROFlUNG TO EOUlP. NUREG/CR4263: REUABluTY ANALYSIS OF STIFF VERSUS FLEXI- MENT OUALIFICATON ISSUES. BLE PIPING FINAL PROJECT REPORT. COE.L.J. CHRISTENSEN,0. NUREG/CR4101: ASSAY OF LONC-UVED RADIONUCLOES IN LOW- !' NUREG/CR-3145 V03. GEOPHYSICAL INVESTIGO'lONS OF THE LEVEL WASTES FROM POWER REACTORS. WESTERN OHIO-INDIANA REGION - ANNUAL REPORT.(October O 1982 September 1983, Volume 3). URE'G 0837 V04 NO3 NRC TLD DIRECT RADIATION MONITORING CHU K.H NETWORK.r8 rogress Report. July September 1984. NUREG/CR-4127 V02: BWR FULL INTEGRAL SIMULATION TEST NUREG 0837 V04 N04: NRC TLD DIRECT RADIATON MONITORING (FIST) PROGRAM TRAC-BWR MODEL DEVELOPMENT. Volume NURE 05 NO : CT b A ION MONITORING NUR -4127 V03: BWR FULL INTEGRAL SIMULATION TEST NE ' (FIST) PROGRAM TRAC-BWR MODEL DEVELOPMENT. Volume 3.De-NUpEG 083 LDhAE AA ATION MONITORING velopment Assessment For Plant Appication. NETWORK. Progress Repo t, Apni,une 1985. COMEN,S. CHUNG,H.M.
'l NUREG/CR-4398: COST ANALYSIS OF REVISIONS TO to CFR PART NUREG/CR-3980 V04: LIGHT-WATER-REACTOR SAFETY FUEL SYS- 50. APPENDIX J LEAK TESTS FOR PRIMARY AND SECONDARY TEMS RESEARCH PROGRAMS. Quarterty Progress Report,0ctobe'-
December 1984- CONTAINMENTS OF LIGHT WATER-COOLED NUCLEAR POWER PLANTS. NUREG/CR-4204: LONG TERM EMBRITTLEMENT.OF CAST DUPLEX STAINLEGS STEELS IN LWR SYSTEMS. Annual Report, October 1983 - COL 8ERT,J.J-i Septemtu 1984. NUREG/CR4067:
~~SUMMARY~~
OF BARRIER DEGRADATON EVENTS 1 CLAIRSORNE.H.C. AND SMALL ACCOENTS IN U S. COMMERCIAL NUCLEAR POWER PLANTS. NUREG/CR-4134: REPOSITORY ENVIRONME'fTAL PARAMETERS NUREG/CR-4068:
~~SUMMARY~~
OF HISTORICAL EXPERIENCE WITH RE-RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL LEASES OF RADIOACTIVE MATERIALS FROM COMMERCIAL NU. WASTE PACKAGES. CLEAR POWER PLANTS IN THE UNITED STATES. CLARK.R.A. COLEMAN,D.R. NUREG/CR4361: STEAM GENERATOR GROUP PROJECT, Annual NUREG/CR-3741 V02: EVALUATION OF POWER REACTOR FUEL Report 1983. ROD ANALYSIS CAPABILITIES. Phase 2 Topcal ReportVolume
~~! NUREG/CR4362: STEAM GENERATOR GROUP PROJECT. Annual 2. Code Evaluation.~~
~~Report 1984. COLUER,R.P.~~
~~. CLAUSER,MJ.~~
NUREG/CR-3937: STEAM GENERATOR TUBE RUPTURE IODINE NUREG/CR4085: USERS MANUAL FOR CONTAIN 10.A Cornputer TRANSPORT MECHANISMS. Task 1.Espenmental Studies. Code for Severe Reactor Accident Containment Analyssa.
! COLUNS,J.L I CLAUSS.D.S. NUREG/CR-3930: OBSERVED BEHAVIOR OF CESIUM. LODINE.AND NUREG/CR-4137: PRETEST PREDICTIONS FOR THE RESPONSE OF TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-A 1.8-SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO GRAM STATIC OVERPRESSURl2ATION. NUREG/CR4037: DATA
~~SUMMARY~~
REPORT FOR FISSION PRODUCT NUREG'CR-4209: COMPAR! SON OF ANALYTICAL PREDICTONS AND RELEASE TEST Hi-5 EXPERIMENTAL RESULTS FOR A 18-SCALE STEEL CONTAINMENT NUREG/CR4043: DATA
~~SUMMARY~~
~~REPORT FOR FISSION PRODUCT MODEL PRESSURIZED TO FAILURE. RELEASE TEST H14 I~~
, , , ..~., -- - - , - . . - . . . , ..w.~-.,r,,,,-.- . - - . .-_,g.-_ - . _ _ , n.n-- .,-.~,n.,-- , , . ._-n. - - -
116 Personal Author Index COLMAR.R. CROFF,A.G. NUREG-0933 S02: A PRIORITIZATON OF GENERIC SAFETY ISSUES. NUREG/CR-4134: REPOSITORY ENVIRONMENTAL PARAMETERS NUREG4933 S03: A PRIORITIZATON OF GENERIC SAFETY ISSUES. RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL WASTE PACKAGES. COMER.M.K. NUREG/CR-3688 V01: GENERATING HUMAN REUABluTY EST1- CRONEY,S.T. MATES USING EXPERT JUDGMENT. Volume 1. Main Report NUREG/CR4181: LEACHABluTY OF RADIONUCUDES FROM NUREG/CR-3688 V02: GENERATING HUMAN REUABluTY ESTI- CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT OPERATING MATES USING EXPERT JUDGMENT. Volume 2.Appendees. NUCLEAR POWER REACTORS. WUREG/CR 4009. HUMAN REUABluTY DATA BANK.Evaluabon Re- NURE3/CR4245; IN-PLANT SOURCE. TERM MEASUREMENTS AT suits. BRUNSWICK STEAM ELECTRIC STATION. MUREG/CR4010. SPECIFICATION OF A HUMAN RELIABILITY DATA . NUREG/CR4397: IN. PLANT SOURCE TERM MEASUREMENTS AT BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE- PRAIRIE ISLAND NUCLEAR GENERATING STATON. AR POWER PLANTS. CULLEN W. COMPERE.E.L NUREG/CR-4422: A REVIEW OF THE MODELS AND MECHANISMS MUREG/CR 3551: SAFETY IMPLICATIONS ASSOCIATED WITH IN- FOR ENVIRONMENTALLY-ASSISTED CRACK GROWTH OF PRES-PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS- SURE VESSEL AND PIPING STECLS IN PWR ENVIRONMENTS. TEMS IN NUCLEAR POWER PLANTS. CMEN,WR CONSERE.W NUREG/CR-3945: FATIGUE CRACK GROWTH RATES OF LOW-RUREG/Cd-4061: LEACHATE PLUME MIGRATON DOWNGRADIENT CARBON AND STAINLESS PIPING STEELS IN PWR ENVIRONMt NT. FROM URANIUM TAtuNGS DISPOSAL IN MINE STOPES NUREG/CR-4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW WUREG/CR4192: THE ANALYSIS OF DRAINAGE AND CONSOLfDA. RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRON-TION AT TYPICAL URANIUM MILL TAluNGS SITES. MENTS. CONDIE.K.G. CUMMINGS F.M. RUREG/CR-3193: FORCED CONVECTIVE.NONEOUluBRiUM. POST- NUREG/CR-3609: EVALUATION OF NEUTRON DOSIMETRY TECH-CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION NIOUES FOR WELL-LOGGING OPERATONS COMPARISON REPORT. CURRIE,J.W. COOPE R.D.W. NUREG/CR 3537. EXPEDIENT METHODS 'OF RESPIRATORY NUREG/CR-3413: OFF-SITE CONSEOUENCES OF RADIOLOGICAL ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON-PROTECTION.llt SUBMICRON PARTICLE TESTS AND
~~SUMMARY~~
OF QUALITY FACTORS. TAMINATON. RUREG/CR4214; HEALTH EFFECTS MODEL FOR NUCLEAR POWER CUTSHALL,N.H. PLANT ACCIDENT CONSEQUENCE ANALYSIS Part I sntroduction, Integration & Summary Part 11 Scientife Basis For Health NUREG/CR-3851 V04. EVALUATON OF RADIONUCUDE GEOCHEUl-CAL nNFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR Effects Models. WASTE REPOSITORY SITE PROJECTS Annual Progress Report For ' CORNELL C A. October 1983-September 1984. NUREG/CR-4334. AN APPROACH TO THE QUANTIFICATION OF SEIS-DAEMEN,J.J. MIC MARGINS IN NUCLEAR POWER PLANTS. NUREG/CR4174. ROCK MASS SEAUNG EXPERIMENTAL ASSESS. CORNWELL.B.C. MENT OF BOREHOLE PLUG PERFORMANCE Annual Report. June WUREG/CR-3819: SURVEY OF AGED POWER PLANT FACluTIES. 1983 May 1984. CORWIN,W.R. DAKE,LS. MUREG/CR-4015: EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR 3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL TESTING THE Z81 INTERMEDIATE SCALE PRESSURE VESSEL PLATES IN BENDING SERIES 1. DAUNG.P.M. COSTANTINO.C.J. NUREG/CR4070 V02: BlVALVE FOUUNG OF NUCLEAR POWER NUREG'CR-4182: VERIFICATION OF SO'L STRUCTURE INTERACTION PLANT SERVICE-WATER SYSTEMS Volume 2. Current Status Of Bio-METHODS. fouhng Survedlance And Control Techruques. NUREG/CR-4070 V03. BlVALVE FOULING OF NUCLEAR POWER CSTHAM,5.M. . PLANT SERVICE WATER SYSTEMS. Factors That May intensify The NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV- Safety Consequences Of Biofouling. TOR DURING STATION BLACKOUT. DALY.8.J. COUNTS,C.A. NUREG/CR-4022: PRESSURIZED THERMAL SHOCK EVALUATON OF NUREGICR-2800 S03 GUIDELINES FOR NUCLEAR POWER PLANT THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT. DANDINI,V.J. COUNT &C.L WUREG/CR-4233: OtSTRIBUTION OF CORBICULA FLUMtNEA AT NU, NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE . RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CON-CLEAR FACIUTIES' OENSER CONTAINMENT. NUREG/CR-4146: SIMULATON OF AN EPRI NEVADA TEST SITE COVFR.LE. (NTS) HYDROGEN BURN TEST AT THE CENTRAL RECEIVER TEST NUREG/CR 3558. HANDBOOK OF NUCLEAR POWER PLANT SEISMIC FACILITY. FRAGluTIES. Seismc Safety Margins Research Prograrrt DANIELSON,W.F. MAN.C.E. NUREC/CR-4030: RADIONUCLIDE ' MIGRATON IN GROUND NUREG/CR4191: SURVEY OF UCENSEE CONTROL ROOM HABIT. WATER (Final Report) ABILITY PRACTICES. COX,D.C. DATTA.A. NUREG 1148; NUCLEAR POWER PLANT FIRE PROTECTION RE. WUREG/CR-4350 V06: PROBABluSTIC RISt' ASSESSMENT COURSE DOCUMENTATON Volume 6 Data Development SEARCH PROGRAM CRAFT,C.M. D A VIS,C.8. NUREG/CR-3863; ASSESSMENT OF CLASS 1E PRESSURE TRANS. NUREG/CR-3935: THERMAL.HYORAUUC ANALYSES OF OVERCOOL. MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT ING SEQUENCES FOR THE H B. ROBINSON UNIT 2 PRESSURIZED SCREENING TESTS. THERMAL SHOCK STUDY.
i Personal Author Index 117 NUREGiCR-3977: RELAPS THERMAL-HYDRAULIC ANALYSES OF DEITZ,V.R. PRESSURIZED THERMAL SHOCK SEQUENCES FOR H.B. ROBIN-i SON UNIT 2 PRESSURIZED WATER REACTOR. NUREG/CR-3990. CHARCOAL PERFORMANCE UNDER ACCIDENT CONDITIONS IN LIGHT WATER REACTORS. NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES '. ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE DENHAM.D.H. PRESSURIZED WATER REACTOR Main Report. WUREG/CR-4326 V02 EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR4118 MONITORING METHODS FOR DETERMINATION ON TRANSIENTS AND ACCfDENTS AT A 3-LOOP WESTINGHOUSE COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT PRESSURIZED WATER REACTOR Appendees. URANf0M RECOVERY SITES. DAVIS.LA. " "
. NUREG/CR-3901: DOCUMENTATON AND USER'S GUIDE.GS2 & GS3 NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR. . VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS. POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU-CLEAR REACTOR VESSELS AND PIPING COMPONENTS.
DAVIS LT. NUREG/CR-4258. AN APPROACH TO TEAM SKILLS TRAINING OF NU. DENNING,R.S. CLEAR POWER PLANT CONTROL ROOM CREWS. NUREG/CR 3937: STEAM GENERATOR TUDE RUPTURE LODINE Davis,M.S. TRANSPORT MECHANISMS. Task 1 Expenmental Studes. WUREG/CR-2482 V08: REVIEW OF DOE WASTE PACKAGE DEUTSCH,WJ. PROGRAM Senwannual Report Covenng The Penod October 1984 . NUREG/CR-3709: METHODS OF MINIMlZING GROUND-WATER CON-March 1985. TAMINATON FROM IN SITU LEACH URANIUM MINING Final Report. NUREG/CR-2482 V09. REVIEW OF DOE WASTE PACKAGE PROGRAM. Semiannual Report Coveang The Penod April-September DEWALL,K.G. 1985. NUREG/CR-3444.V02: THE IMPACT OF LWR DECONTAMINATONS NUREG/CR-3819 SURVEY OF AGED POWER PLANT FACluTIES. ON SOUDIFICATON, WASTE DISPOSAL AND ASSOCIATED OCCU- DEWERD,LA. PATONAL EXPOSURE. NUREG/CR-4131: INVESilGATON OF ALTERNATIVE MEANS TO AC-l DAVIS.R.E. " WUREG/CR-3329: AN EVALUATION OF (HE STABIUTY TESTS REC-N i OMMENDED ly THE BRANCH TECHNICAL POSITION ON WASTE DEWEY,J.W. FORMS AND CONTAINER MATERIALS. NUREG/C' 339- A REVIEW OF RECENT RESEARCH ON THE SEIS-DAVIS.T.C. MOTECTONICS OF THE SOUTHEASTERN SEA 80ARO AND AN NUREG/CR4144: IMPORTANCE RANKING BASED CN AGING CON- EVALUATION OF HYPOTHESES ON THE SOURCE OF THE 1886 SOERATONS OF COMPONENTS INCLUDED IN PROBA88USTIC CHA9LESTON. SOUTH CAROUNA EARTHOUAKE. RISK ASSESSMENTS. PUREG/CR-4377: EVALUATIONS AND UTIUZATIONS OF RISK IM- DHOOGE,NJ. PORTANCES. NUREG/CR4358. APPUCATIONS OF DENSITY PROFluNG TO EQUIP. DAWSON,J.Fl MENT QUAUFICATION ISSUES. NUREG/CR 3915. ACOUSTIC EMISS:ON RESULTS OBTAINED FROM DIAMENT,H. TESTING THE ZB-1 INTERMEDIATE SCALE PRESSURE VESSEL NUREG/CR-3442: RADTWO A COMPUTER CODE FOR SIMULATING 4 DE JARLAIS,G. FAST-TRANSIENT, TWO-DIMENSIONAL,TWO-LAYER RADIONU-CUDE CONCENTRATION CONDITIONS IN NUREG/CR4277: INVERTED ANNUAL FLOW EXPERIMENTAL STUDY. LAKES, RESERVOIRS. RIVERS. ESTUARIES.AND COASTAL REGIONS. DEAN,R.S. DICK C.E NUREG/CR-3791. CLOSEOUT OF IE BULLETIN 79-09 FAILURE OF GE NUREG/CR-4266; STANDARD BETA-PARTICLE AND MONOENERGE-NU / -4 C FI ET N 4 INCORRECT TIC ELECTRON SOURCES FOR THE CAUBRATON OF BETA.RADI-WEIGHTS FOR ' SWING CHECK VALVES MANUFACTURED BY ATON PROTECTION INSTRUMENTATION. UELAN ENGINEERING CORPORATON-NUREG/CR-4004: CLOSEOUT OF lE BULLETIN 79-25 FAILURES OF DICKSON.C.R* WESTINGHOUSE BFD RELAYS IN SAFETY RELATED SYSTEMS. NUREG/CR 3488 V03. IDAHO FIELD EXPERIMENT 1981 Volume NUREG/CR4005: CLOSEOUT OF IE BHLLETIN 8012 DECAY HEAT 3.Companson Of Tralectores, Concentration Patterns And MESODIF REMOV LL SYSTEM OPERABluTY. Model Calculations NUREG/CR4006: CLOSEOUT OF IE BULLETIN 8141. SURVEILLANCE OF MECHANICAL SNUBBERS. DIL8ER,1. DEARING.J.F. NUREG/CR4415: COUNTER CURRENT STEAM / WATER FLOW RUREG/CR4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE-1 ABOVE A PERFORATED PLATE VERTICAL INJECTON OF WATER PRESSURIZED WATER REACTOR. DINGMAN,S.E.
~~- NUREG/CR-3913: HECTR VERSION 10 USER'S MANUAL NUREG-0905: CLOSEOUT OF IE BULLETIN 79-12.SHORT.PEROD DIONNE.8J.~~
SCRAMS AT BOILING-WATER REACTORS-NUREG/CP4066; PROCEEDINGS OF AN INTERNATIONAL WORK. SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION NU G/CR4237: MO8luTY OF RADIONUCUDES IN HIGH CHLORIDE ENVIRONMENTS. N R G/CR 4254 UP T NAL SE REDUCT kN ALARA AT NUCLEAR POWER PLANTS Study On High-Dose Jobs Radwaste 5 DEEDS W L Handhng.And ALARA Incentues. NUREG/CR-3949 V01: EDDY-CURRENT INSPECTION FOR STEAM RA TUBING PROGRAM. Serniannual Progress Report For D A ,R NUREG/CR-3949 V02: EDDY. CURRENT INSPECTON FOR STEAM ACCIDENTS Perspectwes On Managing Severe Accidents in Commer.. GENERATOR TUBING PROGRAM. Annual Progress Report For Penod cial Nuclear Power Plants. Ending December 31,1984. NUREG/CR-4177 V02- MANAGEMENT OF SEVERE ACCIDENTS. Extending Plant Operating Procedures into The Severe DEFIELD.J.D. Accident Regime. 1 NUREG/CR4111: A COMARATIVE STUDY OF HEPA FILTER EFFI-l CIENCIES WHEN CHALLENGED WITH THERMAL- AND AIR-JET- DOORANICH.D. GENERATED DI-2-ETHYLHEXYL SEBECATE,DI-2-ETHYLHEXYL PHTHALATE,AND SODIUM CHLORIDE. NUREG/CR4044. TRAC-PF 1 LOCA CALCULATONS USING FINE-NODE AND COARSE-NODE INPUT MODELS. 1 l l l i
i 118 Personal Author index NUREG/CR-4155: TRAC PF1/MCOI INDEPENDENT DUCE,S.W. ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORATED-PLATE NUREG/CR 4245: IN. PLANT SOURCE TERM MEASUREMENTS AT CCFL TESTS. BRUNSWICK STEAM ELECTRIC STATON NUREG/CR-4189: TPAC PF1/ MOD 1 INDEPENDENT ASSESSMENT.Semiscale MOD-2A Feedwater-Lee Break (S-SF-3) And DUDA,P.M. Steam-tme Break (S-SF 5) Tests. NUREG/CR-2718. STEAM EXPLOSION EXPERIMENTS WITH SINGtE DROPS OF IRON OXIDE MELTED WITH A CO2 LASER Past
~~" "** 8#~~
UR CP-0063. PROCEEDINGS OF THE 1984 STATISTICAL SYMPO-S!UM ON NATONAL ENERGY ISSUES, DUKLER A.E. NUREG/CH-4424: DROPLET SIZES. DYNAMICS AND DEPOSITON IN DODD.C.V. VERTICAL ANNULAR FLOW. NUREG/CR-3949 V01: EDOY-CURRENT INSPECTON FOR STEAM GENERATOR TUBING PROGRAM. Semiarmual Progress Report For NUR /R 49 02. EDDY-CURRENT INSPECTION FOR STEAM NUREG/ 3950 V01: FUEL PERFORMANCE ANNUAL REPORT FOR GENERATOR TUBING PROGRAM. Annual Progress Report For Period 1983 Ereng December 31,1984. DODSON.K.E. NUREG/CH-3981: BIOACCUMULATON OF P 32 IN BLUEGILL AND l NUREG/CR-3514 V02: THE CHEMICAL BEHAVOR OF ODINE IN CATFISH. AQUEOUS SOLUTIONS UP TO 150 C.ll. Radiation-Redox Coretions. DUNN,W.E. DODSON,M.E. NUREG/CR-4120: MATHEMATICAL MODEUNG OF ULTIMATE HEAT NUREG/CR-3906: URANIUM MILL TAILINGS SINK COOUNG PONDS. NEUTRAUZATONCONTAMINANT COMPLEXATION AND TAILINGS LEACHING STUDY- DUNNING,D.E. NUREG/CR4259: TAluNGS NEUTRAUZATON AND OTHER ALTER. NUREG/CA-4038. SENSITIVITY AND UNCERTAINTY STUDIES OF THE NATIVES FOR IMMOBILIZING TOXIC MATERIALS IN TAIUNGS Final CRAC2 COMPUTER CODE. Report W ,W.E. DOERGE.D.H. NUREG/CR4274. Af4ALYSIS AND TESTS ON SMALL-SCALE SHEAR NUhEG/CR4090: EVALUATION OF NUCLEAR FACluTY DECOMMIS- WALLS FY-82 FINAL REPORT. SiONING PROJECTS. Annual Summary Report Fiscal Year 1984. DOE
~~SURG,J.M. DURSIN.P.W.~~
NUREG/CR4355 V01: 238 PU(IV) IN MONKEYS. Overview Of Metabo. NUREG/CR 3747: THE SELECTION AND TESTING OF ROCK FOR AR-lism. MORING URANIUM TAluNGS IMPOUNDMENTS. DOLAN.F.X. EARY,LE. - NUREG/CR-3426 V01: THERMAL AND FLUID MIX 1NG IN 1/2-SCALE NUREG/CR-3709 METHODS OF MINIMl2lNG GROUND-WATER CON-TEST FACILITY, Facdity And Test Design Report TAMINATON FROM IN SITU LEACH URANIUM MINING Final Report NUREG/CR 3426 V02- THERVAL AND FLUlu MIXING IN 1/2-SCALE TEST FACluTY. Data Report. _ EASLEY,P. NUREG/CR-4143. REVIEW AND EVALUATON OF THE MILLSTONE UNIT 3 PROBABluSTIC SAFETY STUDY. Containment Failure G RdOO9: HUMAN REUABIUTY DATA BANK Evaluation Re. Modes. Radiological Sourco Terms And Offsite Consequences. NUREG/CR4010- SPECIFICATION OF A HUMAN REUABluTY DATA EBEL J.E BAN OR CON ING HRA SEGMENTS OF PRAS FOR NUCLE- NUREG/CR-4354; A STUDY OF SEISMICITY AND TECTONICS IN NEW ENGLAND Final Report. i NUREG/CR4280 THE' EFFECTS OF SUPERVISOR EXPERIENCE AND ASSISTANCE OF A SHlFT TECHNICAL ADVISOR (STA) ON CREW EBERHARDT,LL PERFORMANCE IN CONTROL ROOM SIMULATORS. NUREG/CR-4268: RATIO METHODS FOR COST-EFFECTIVE FIELD SAMPUNG OF COMMERCIAL RADIOACTIVE LOW-LEVEL WASTES. DORNSIFE.B. NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE. EDLER.S.K. RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE NUREG/CR-3810 V03. REACTOR SAFETY RESEARCH REGULATORY PROGRAM. PROGRAMS Quarterty Report. July September 1984 DOUGHERTY,D.R. NUREG/CR 3810 V04: REACTOR SAFETY RESEARCH PROGRAMS Quarterly Report October-December 1984. NUREG/CR-3829: AN EVALUATON OF THE STABluTY TESTS REC. REACTOR SAFETY RESEARCH NUREG/CR-4318 V01: OMMENDED IN THE BRANCH TECHNICAL POSITON ON WASTE PROGRAMS Quarterty Report. January-March 1985. FORMS AND CONTAINER MATERIALS. REACTOR SAFETY RESEARCil NUREG/CR-4062: EXTENDED STORAGE OF LOW LEVEL RADIOAC- NUREG/CR-4318 V02: PROGRAMS Quarterfy Report.Apni-June 1985. TIVE WASTES Potential Problem Areas. NUREG/CR-4215: TECHNICAL FACTORS AFFECTING LOW LEVEL WASTE FORM ACCEPTANCE CRITERIA. EDMONDS D.P. NUREG/CR-4106. PRESSURIZED-THERMAL-SHOCK TEST OF 6 IN.- LS PTSE 1: Investigation Of Warm Prestress-
/ 3197 V01: REACTON BETWEEN SOME CESIUM-CDINE g COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS STEEL,1NCONEL 600 & SILVER. Volume f:Cessum Hydroxide Reac- EDSON.J.L t'ons. NUREG/CR4080 DETERMINATION OF THE AVAILABluTY OF CORE EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATIONS.
DOVE.R.C. NUREG/CR4274: ANALYSIS AND TESTS ON SMALL-SCALE SHEAR EHRUCH,M. WALLS FY 82 FINAL REPORT. NUREG/CR4266: STANDARD BETA. PARTICLE AND MONOENERGE-TIC ELECTRON SOURCES FOR THE CALIBRATION OF BETA-RADf. DRAKE J.S. ATON PROTECTON INSTRUMENTATON. NUREG/CR-3723: STRESSINTENSITY-FACTOR INFLUENCE COEFFl. CIENTS FOR SURFACE FLAWS IN PRESSURE VESSELS. EIDSON.A.F. NUREG/CR-3984: BIOLOGICAL CHARACTERIZATON OF RADIATON DRISCOLL.J.W. NUREG/CR-4191: SURVEY OF LICENSEE CONTFIOL ROOM HABIT- EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL. ING EFFLUENTS Annual Progress Report,Apnl 1983 March 1984. ABluTY PRACTICES l l
i l
~~t. Personal Author Index 119 i~~
I EISSENDERG,0.M. ' EVANS.D.D. NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC NUREG-1046: DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY-FEA- UNSATURATED ZONE: TECHNICAL CONSIDERATIONS AND RE-TURE SYSTEM 3 OF NUCLEAR POWER PLANTS. SPONSE TO COMMENTS. EL-SASSIONI.A. NUREG/CR-4042 A 3-DIMENSIONAL COMPUTER MODEL TO SIMU-MUREG/CR-2815 V01 R1: PROBA81USTIC SAFETY ANALYSIS PROCE-LATE F6Ul0 FLOW AND CONTAINMENT TRANSPORT THROUGH A DURES GUOE Sections 17 And Appendices. ROCK FRACTURE SYSTEM. NUREG/CR-3485:PRA REVIEW MANUAL EVANS J.S. EL-SHINAWY,R.M. NUREG/CR-4214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER NUREG/CR-3981: BiOACCUMULATON OF P 32 IN BLUEGILL AND PLANT ACCOENT CONSEQUENCE ANALY SIS Part CATFISH' tintroducta, Integration & Summary Part II.Sciantifc Bas.s For Health Effects Models. ELDER.H.K. NUREG/CR 3293 V01: TECHNOLOGY, SAFETY AND COSTS OF DE- FALETTI,D.W. COMMISSONING REFERENCE NUCLEAR FUEL CYCLE AND NON- NUREG/CR-4101: INTEGRATION OF EMERGENCY ACTION LEVELS FUEL CYCLE FACILITIES FOLLOWING POSTULATED WITH COMBUSTION ENGINEERING EMERGENCY OPERATING ACCOENTS Main RW PROCEDURES By Use Of Combustion Engineenng Owners Group NUREG/CR-3293 V02: TFCM80 LOGY,SATETY Ai40 COSTS OF DF. Emergency Operating Procedure Techrucal Guidelines. COMMISSONING REFERENCE FUEL CYCLE AND NON-FUEL CYCLE FACILITIES FOLLOWING POSTULATED FAlOUS,F. ACCIDENTS. Appendices. NURFG/CR-3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE. ELLINGWOOD,5- FAYER,M.J. NUREG/CR-3876: PROBA8ILITY BASED LOAD COMBINATION CRITE. NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PILES RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. A COMPARISON OF ANALYSIS TECHNIQUES. ELMORE,M.R. NUREG/CR4087: MEASUREMENTS OF URANIUM MILL TAIUNGS CONSOLOATON CHARACTERISTICS. WURCG/CR-4089- EVALUATION OF FIELD TESTED FUGITIVE DUST NUREG/CR4192: THE ANALYSIS OF DRAINAGE AND CONSOLOA-CONTROL TECHNIQUES FOR URANIUM MILL TAILINGS PILES. TION AT TYPICAL URANIUM MILL TAIUNGS SITES. ELOCK.R.M. FEHRINGER.DJ. RUREG/CR-3197 VO1: REACTION BETWEEN SOME CESlUM.ODINE COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS NUREG 0946. AN EVALUATON OF RADIONUCLIDE CONCEN1RA. STEEL,1NCONEL 600 8 SILVER. Volume I. Cesium Hydroxide Reac- TONS IN HIGH-LEVEL RADtOACTIVE WASTES. tions. FENTON,0.L NUREG/CR-4264. INVESilGATION ON HIGH EFFICIENCY PARTICU-UAEd CR 4022: PRESSURIZED THERMAL SHOCK EVALUATION OF NU E / R 4 21 FULL THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. L MEA UR ME OF SM K TRANS-PORT AND DEPOSITION IN VENTILFON SYSTEM DUCTWORK. EMBREY,D.E. RUREG/CR-4016 VOI: APPLICATION OF SUM-MAGO A TEST OF AN INTERACTIVE COMPUTER-BASED METHOD FOR ORGANtZtNG NUREG/CR.4128. BWR FULL INTEGRAL SIMULATON TEST (FIST) EXPERT ASSESSMENT OF HUVAN PERFORMANCE AND PHASE il TEST RESULTS AND TRAC-BWR MODEL QUAUFICATION. REUABluTY. Volume i Main Report. FINEMAN.C.P. EMElGH.C.W. NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAILU3ES RUREG-1065 RO1: ACCEPTANCE CRITERIA FOR THE LOW EN- ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC RICHED URANIUM REFORM AMENDMENTS _ 80tLING WATER REACTOR Main Report NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES EMERSON.E.L ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC NUREG/CR-3611: RADOACTIVE MATERIAL (RAM) ACCOENT/ INCL- BOluNG WATER REACTOR. Appendices. DENT DATA ANALYSIS PROGRAM. FINLAYSON.F.C. EMRIT R. NUREG/CR-4133. NUCLEAR POWER SAFETY REPORTING SYSTEM WUREG-0933 S02: A PRIORITIZATION OF GENERIC SAFETY ISSUES. IMPLEMENTATON AND OPERATIONAL SPECIFICATONS. NUREG4933 S03. A PRORITIZATION OF GENERIC SAFETY ISSUES. FISCHER,LE. N REG *1 274: ANALYSIS AND TESTS ON SMALL SCALE SHEAR ERS WALLS FY-82 FINAL REPORT. FITZSIMMONS.D. W REG R 276: Vl8RATON AND WEAR IN STEAM GENERATOR NUREG/CR-3999- ELECTRICALLY HEATED EX-REACTOR PELLET. TUBES FOLLOWING CHEMICAL CLEANtNG - SEMIANNUAL CLADDING INTERACTION (PCI) SIMULATIONS UTIUZING IRRADIAT. REPORT. ED ZnRCALOY CLADDING. ENDRES,G.W. FLANAGAN,G.F. WUREG/CR-3609: EVALUATION OF NEUTRON DOSIMETRY TECH- NUREG/CR-4022. PRESSURIZED THERMAL SHOCK EVALUATON OF NIQUES FOR WELL LOGGING OPERATIONS. THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. ERASLAN,A.H. FLAN 1GAN.LJ. NUREG/CR-3442: RADTWO:A COMPUTER CODE FOR SIMULATING NUREG'CR-3937: STEAM GENERATOR TU8E RUPTURE LODINE FAST. TRANSIENT. TWO-DIMENSIONAL,TWO-LAYER RADIONU. TRANSPORT MECHANISMS. Task 1 Eirpenmental Studies. CUDE CONCENTRATION CONDITIONS IN LAKES, RESERVOIRS, RIVERS. ESTUARIES.AND COASTAL REGIONS. FLETCHER.C.D. NUREG/CR-3935: THERMAL-HYDRAULIC ANALYSES OF OVERCDOL-ETTINGER,H.J. ING SEQUENCES FOR THE H 8. ROBINSON UNIT 2 PRESSWl2ED NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER EFFl- THERMAL SHOCK STUDY. CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR 4ET- NUREG/CR-3977: RELAP3 THERMAL-HYDRAUUC ANALYSES OF GENERATED DI-2. ETHYLHEXYL SE8ECATE.DI-2-ETHYLHEXYL PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B ROBIN-PHTHALATE.AND SODIUM CHLOROE. SON UNIT 2 PRESSURIZED WATER REACTOR.
120 Personsi Author index FLUCKlGER J.D. GALLUCCI.R.H. NUREG/CH4298: DESIGN AND INSTALLATION OF COMPUTER SYS- NUREG/CR-2800 S03: GulDEUNES FOR NUCLEAR POWER PLANT TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73.55. SAFETY ISSUE PRIORITIZATON INFORMATION DEVELOPMENT. NUREGICR-4220: REUABluTY ANALYSIS OF CONTAINMENT ISOLA-FOLEY M.G. TION SYSTEMS. NUREG/CR-3747: THE SELECTION AND TESTING OF ROCK FOR AR. MORING URANIUM TAILINGS IMPOUNDMENTS. GALLUP.D.R. NUREG/CR43J5: POTENTIAL BENEFITS OBTAINED BY REQUIRING FOLEY,W.J. SAFTEY.GP \DE COLD SHUTDOWN SYSTEMS. NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79 09. FAILURE OF GE TYPE AK 2 CIRCUti BREAKERS IN SAFETY-RELATED SYSTEMS GANAPATHY,S. NUREG/CR 3794. CLOSEOUT OF IE BULLETIN 80-25: OPERATING NUREG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-PROBLEM 9 WITH TARGET ROCK SAFETY-RELIEF VALVES AT POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU-Nt REG /CR4003: CLOSEOUT OF IE BULLETIN 70-04. INCORRECT WEIGHTS FOR SWING CHECK VALVES MANUFACTURED BY GANT1.C S. VELAN ENGINEERING CORPORATION. NUHEG/CR4231: EVALUATION OF AVAILABLE DATA FOR PROBABI-NUREG/CR4004: CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF LISTIC RISK ASSESSMENTS (PRA) OF FIHE EVENTS AT NUCLEAR WESTINGHOUSE BFD RELAYS IN SAFETY.RELATED SYSTEMS. POWER PLANTS' NUREG/CR4005: CLOSEOUT OF IE BULLETIN 80-12. DECAY HEAT REMOVAL SYSTEM OPERA 81UTY. GAUSE E. NUREG/CR4006: CLOSEOUT OF IE BULLETIN 8141. SURVEILLANCE NUREG/CR-2482 V08. REVIEW OF DOE WASTE PACKAGE OF MECHANICAL SNUBBERS- PROGRAM Semiannual Report Covenng The Penod October 1984 - March 1985. FRAGOLA J. NUREG/CR-2815 V01 RI: PROBA81USTIC SAFETY ANALYSIS PROCE* GAUSSENS.G. NUREG/CR-4091: THE EFrECT OF ALTERNATIVE AGING AND ACCl-NU EG/C 3 2d h81 LIT T THE ACQUISTION OF U-DEM SWWW ON EMR NERE CENSEE EVENT DATA. GAVIN P. FRANKLIN.A.G. NUREG/CR-4120: MATHEMATICAL MODEUNG OF ULTIMATE HEAT ' RUREG/CR-4430: CURRENT METHODOLOGIES FOR ASSESSING SINK COOUNG PONDS. THE POTENTIAL FOR EARTHOUAR (E-INDUCED UOUEFACTON IN SOfLS. GE T.G.W. NUREG/CR-4076. DETERMINATON OF COMPUANCE WITH CRITERIA FREEMAN,H.D. FOR FINAL TAILINGS DISPOSAL SITE RECLAMATON. NUREGICR-4076: DETERMINATION OF COMPUANCE WITH CRITERIA FOR FINAL TAILINGS DISPOSAL SITE RECLAMATON. GE NTIL1,H. NUREG/CR4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV-FREEMAN-KELLY OR DURING STATION BLACKOUT. WUREG/CR-4172: A USER'S GUIDE FUR MERGE. GENTILLON.C.D. FRESCO A. NUREG/CR-3862: DEVELOPMENT OF TRANSlENT INITfATING EVENT NUREG/CR4228. REVIEW OF THE VOGTLE UNITS 1 AND 2 AUXILIA. FREQUENCIES FOR USE fN PROBABluSTIC RISK ASSESSMENTS RY FEEDWATER SYSTEM REUABluTY ANALYSIS. NUREG/CR4071: EXPLORATORY TREND AND PATTERN ANALYSIS FRUCHTER.J.S. FOR 1981 UCENSEE EVENT REPORT DATA.' WUREG/CR-4030: RADIONUCUDE MiGRATON IN GROUND GERDING T.J. WATER (Final Report) NUREG/CR-3710- LABORATORY STUDIES OF A BREACHED NUCLE-FUENKAJORN,K. AR WASTE REPOSITORY IN BASALT. NUREG/CR-4174 ROCK MASS SEAUNG EXPERIMENTAL ASSESS- GERGELY,P. MENT OF BOREHOLE PLUG PERFORMANCE Annual Report. June NUREG/CR4123: SEISMIC FRAGluTY OF REINFORCED CONCRETE 1983 May 1984. STRUCTURES AND COMPONENTS FOR APPUCATION TO NUCLE-FULLER,LC. AR FACILITIES. NUREG/CR-3764: BWR-LTAS: A BOfLING WATER REACTOR LONG- GERTMAN,D.L TERM ACCIDENT SIMUL % TON CODE-NUREG/CR-4040: OPERATIONAL DECISIONMAKING AND ACTION SE-
~~' FURGAL.D.T. .~~
LECTION UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER NUREG/CR-3863. ASSESSMENT OF CLASS 1E PRESSURE TRANS- PLANTS. MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT GHERSON.P. SCREENING TESTS. NUREG/CR-4022: PRESSURIZED THERMAL SHOCK EVALUATON OF THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT, GA8ETTA G. NUREG/CR-4422: A REVIEW OF THE MODELS AND MECHANISMS GlESEKE.J.A. FOR ENVIRONMENTALLY-ASSISTED CRACK GROWTH OF PRES- NUREG/CR 3937: STEAM GENERATOR TUBE RUPTURE ODINE SURE VESSEL AND plPING STEELS IN PWR ENVIRONMENTS. TRANSPORT MECHANISMS. Tasit 1.Empenmental Studies. GADDY C.D. GILLEN,D. NUREG/CR-3688 V01: GENERATING HUMAN REUABILITY ESTI. NUREG/C94403-
~~SUMMARY~~
OF THE WASTE MANAGEMENT PRO-MATES USING EXPERT JUDGMENT. Volume 1 Man Report. GRAMS AT URANIUM RECOVERY FACIUTIES AS THEY RELATE TO NUREG/CR-3688 V02: GENERATING HUMAN REUA81UTY ESTl. THE 40 CFR PART 192 STANDARDS, i MATES USING EXPERT JUDGMENT. Volume 2 Appendices. NUREG/CR4009: HUMAN REUA81UTY DATA BANK Evaluaton Re- GILLEN K.T. suits NUREG/CR4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU- NUREG/CR-4008: GENERAL EXTRAPOLATION MODEL FOR AN IM. PORTANT CHEMICAL DOSE-RATE EFFECT. CLEAR POWER PLANT CONTROL ROOM CREWS. NUREG/CR4358' APPUCATIONS OF DENSITY PROFluNG TO EQUIP-GALLAHER R.B. MENT OUAUFICATION ISSUES. WUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM GILMORE,W.E. FOR UCENSEE EVENT REPORTSCode bstmqs NUREG/CR-3905 V03: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-FOR UCENSEE EVENT REPORTS Coder's Manual. GENERATED DISPLAYS. NUREG/CR 3905 V04: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4227: HUMAN ENGINEERING GUIDEUNES FOR THE FOR UCENSEE EVENT REPORTS Coder's Manual EVALUATION AND ASSESSMENT OF VIDEO DISPLAY UNITS.
Personal Author Index 121 GIRVIN,D.C. GRIESS.J.C.
~~. NUREG/CR-4030: RADIONUCUDE MlGRATION IN GROUND _ NUREG/CR41?4- REPOSITORY ENVIRONMENTAL PARAMETERS WATER (Final Report)~~
RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL GLISSMEYER,J.A. WASTE PACKAGES. NUREG/CR4088: METHOOS FOR ESTIMATING PADIOACTIVE AND GRIFFtTH,P, TOXIC AIRBORNE SOURCE TERMS FOR URANIUM MILLING OPER-ATONS. NUFiEG/CR-4376: HEAT TRANSFER. CARRYOVER AND FALL BACK IN PWR STEAM GENERATORS DURING TRANSIENTS. GODFREY,D. GRONEMYER.L FlUREG/CR4398: COST ANALYSIS OF REVISIONS TO to CFR PART 50 APPENOlX J. LEAK TESTS FOR PRIMARY AND SECONDARY NUREG/CR4352: SUGGESTED STATE REQUIREMENTS AND CRITE. CONTAINMENTS OF UGHT.WATERCOOLED NUCLEAR POWER HlA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE PLANTS. REGULATORY PROGRAM. GUNAJt,M.V. GOETSCH S.J.
- NUREG/CR4131; INVESTIGATION OF ALTERNATIVE MEANS TO AC- NUREG/CR4264. INVESTIGATION ON HIGH-EFFICIENCY PARTICU-LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS.
COMPUSH THE GOALS OF BIENNIAL ION CHAMBER CAUBRA TON. GUNDERSEN,G.E. GOLDIN D. NUREG 1065 Rot: ACCEPTANCE CRITERIA FOR THE LOW EN-RICHED URANIUM REFORM AMENDMENTS. NUREG/CR-4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART
~~50. APPENDIX J, LEAK TESTS FOR PRIMARY AND SECONDARY GUPPY,J.G.~~
CONTAINMENTS OF UGHT WAeER-COOLED NUCLEAR POWER NUREG/CR-4152:AN INDEPENDENT SAFETY ORGANIZATION. PLANTS. GOLDMAN,A.S. GUYMON.R.H. NUREG/CR-3551: SAFETY IMPUCATIONS ASSOCIATED WITH IN-NUREG/CR4107: SEQUENTIAL TEST PROCEDURES FOR DETECT. ING PROTRACTED MATERIALS LOSSES. PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS. TEMS IN NUCLEAR POWER PLANTS. GOODWIN,G.M. NUREG/CR 3905 V02- SEQUENCE CODING AND SEARCH SYSTEM . FOR UCENSEE EVENT REPORTS Code Listings. NUREG/CR-4015: EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR 3905 V03. SEQUENCE CODING AND SEARCH SYSTEM CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL FOR UCENSEE EVENT REPORTS. Coder's Manual. PLATES IN BENDING SERIES t. NUREG/CR-3905 V04. SEQUENCE CODING AND SEARCH SYSTEM GORDON J. FOR LichNME EVENT HEPORTS. Coder's Manual. PilREG."A4357: THE FEASIBluTY OF DETECTING THE IMPORT OF GUZOWSKI.R.V. UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED NUREGICR 4110: REPOSITORY SITE DATA REPORT FOR UNSATU-STATES. RATED TUFF, YUCCA MOUNTAIN. NEVADA. GOTTULA.R.C. HAAS,P.M. P!UREG/CR-3193: FORCED CONVECTIVE.NONEOUluBRIUM. POST- NUREG/CR 3887: HUMAN FACTORS REVIEW FOR SEVERE ACCl-CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION DENT SEQUENCE ANALYSIS. COMPARISON REPORT. HAASL,D. GRAHAM,E.D. NUREG/CR4350 V04: PROBABIUSTIC RISK ASSESSMENT COURSE
. NUREG/CR4191: SURVEY OF UCENSEE CONTROL ROOM HABIT. DOCUMENTATON Volume 4 - Systern Rehatahty And Analyssa ABluTY PRACTICES. Techniques.Sesssons B/C. Event Trees / Fault Trees.
GRAY,LH. HABERMA%J H. NUREG/CR4280. THE EFFECTS OF SUPERVISOR EXPERIENCE AND NUREG/CR4218; LOCA SIMULATION IN THE NATIONAL RESEARCH ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW UNIVERSAL REACTOR PROGRAM Postirradiation Exarnination Re-PERFORMANCE IN CONTROL ROOM SIMULATORS. suits For The Third Matenals Test (MT-3) . Second Campaign. GREEN.N.M. . HACNBARTH,C.J. NUREG/CR-3905 V01 R1: SEQUENCE CODING AND SEARCH NUREG-1046: DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN SYSTEM FOR UCENSEE EVENT REPORTS User's Guide. UNSATURATED ZONE: TECHNICAL CONSIDERATIONS AND RE-GhEENSTREET,W. SPONSE TO COMMENTS. NUREG/CR-4234 V01: AGING AND SERVICE WEAR OF ELECTRIC HAGGARD,D.L MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY FEA-TURE SYSTEMS OF NUCLEAR POWER PLANTS- NUREG/CR 3609- EVALUATION OF NEUTRON DOSIMETRY TECH-NIQUES FOR WELL-LOGGING OPERATIONS. GREER,W.B. HALL,R.E. NUREG/CR-4174: ROCK MASS SEAUNG . E)PERIMENTAL ASSESS-MENT OF BOREHOLE PLUG PERFORMANCE Annual Report. June NUREG/CR-2815 V01 R1: PROBABluSTIC SAFETY ANALYSIS PROCE. DURES GU!DE. Sections 17 And Appendices 1983. May 1984. NUREG/CR-3026. FEASIBluTY STUDY ON THE ACOUISTION OF U. GREGORY,W.S. CENSEE EVENT DATA. NUREGi;A-4152: AN INDEPENDENT SAFETY ORGANIZATION. NUREG/CR-4225:
~~SUMMARY~~
OF EFFICIENCY TESTING OF STAND. ARD AND HIGH CAPACITY HIGH-EFFICIENCY PARTICULATE AIR HALL.W.J. FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA. . NUREG/CR-4334: AN APPROACH TO THE QUANTIFICATION OF SEIS. TION AND EXPLOStVE SHOCK WAVES. MIC MARGINS IN NUCLEAR POWER PLANTS. NUREG/CR4232; THE RESPONSE OF VENTILATON DAyp{Q$ TQ LARGE AIRFLOW PULSES. HAMMOND R.A. NUREG/CR4260: TORAC USER'S MANUALA Computer Code For Ana-lymng Tornado-Induced Flow And Matenal Transport in Nuclear Facili- NUREG/CR-3981: BIOACCUMULATION OF P-32 IN BLUEGlLL AND CATFISH. ties. NUREG/CR4264: INVESTIGATON ON HIGH-EFFICIENCY PARTICU. HANAN,N. LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS. NUREG/CR-3485. PRA REVIEW MANUAL. GREIMANN,L NUREG/CR4050 A REVIEW OF THE SHOREHAM NUCLEAR POWER j - STATION PROBABluSTIC RISK ASSESSMENT. internal Events And NUREG/CR 3952: SEQUOYAH EQUIPMENT HATCH SEAL LEAKAGE. Core Damage Frequency
122 Personal Author Index HANEY,LN. HARTZMAN.M. NUREG/CR-4040: OPERATONAL DECISIONMAKING AND ACTION SE- NUREG/CR-1677 V02: PIPING BENCHMARK PROBLEMS. VOLUME ll LECTION UNDER PSYCHOL @lCAL STRESS IN NUCLEAR POWER DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION RE-PLANTS. SPONSE SPECTRUM METHOD. HANNINEN.H- HASKIN.F.E. NUREG/CR4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW NUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCl-RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRON- DENTS IN AN ICE CONDENSER CONTAINMENT, MENTS. NUREG/CR 4422: A REVIEW OF THE MODELS AND MECHANISalS HAWTHORNE,J'R* FOR ENVIRONMENTALLY ASSISTED CRACK GROWTH OF PRES- NUREG/CR4437: EXPLORATORY STUDIES OF ELEMENT INTERAC. SURE VESSEL AND PIPING STEELS IN PWR ENVIRONMENTS. TONS AND COMPOSITION DEPENDENCIES IN RADIATION SENSI-TIVITY DEVELOPMENT. HANSON.R.G. MUREG/CR4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES HAYMAN.R.8. ON TRANS!ENTS AND ACCIDENTS AT A GENERAL ELECTRIC NUREG/CR-4317 V01: CANADIAN SEISMC AGREEMENT.Tects,ical BOfUNG WATER REACTOR Man Report. MUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES Report Covenng 1979 1985. ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC HEASERUN,S.W. BOfDNG WATER REACTOR. A es NUREG/CR-4429: TRAC.BD1/M 10SERS GUIDEUNE. - NUREG/CR-2800 S03: GUIDEUNES FOR NUCLEAR POWER PLANT SAFETY ISSUE PRIORITIZATION INFORMATION DEVELOPMENT. HARDY,H.A. NUREG/CR-2531 R03. INTRODUCTORY USER'S MANUAL FOR THE HEATON,H.T U.S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE- NUREG/Cd4266: STANDARD BETA-PARTICLE AND MONOENERGE-SEARCH DATA BANK. TIC ELECTRON SOURCES FOR THE CALIBRATION OF BETA RADt. ATION PROTECTION INSTRUMENTATION HAROY,J.E. NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN- HE8 DON.F.J. STRUMENTATION IN REFLOOD TEST FACluTIES. NUREG-1022 S02: LICENSEE EVENT REPORT SYSTEM Evaluation Of HARLAN,C.P. Frst Year Results And Recommendations For improvements. NUREG/CR-3657: PRELIMINARY SCREENING OF FUEL CYCLE AND HECK,C.L BY-PRODUCT MATERIAL LICENSES FOR EMERGENCY PLANNING NUREG/CR-4127 V01: BWR FULL INTEGRAL SIMULATION TEST UREd R-4117: FAULTING AND JOINTING IN AND NEAR SURF ACE y , MINES OF SOUTHWESTERN INDIANA. HENAC. HARRER.B.J. NUREG/CR-3904: A COMPARISON OF UNCERTAINTY AND SENSITIV. WUREG/CR.3413. OFF. SITE CONSEQUENCES OF RADIOLOGICAL ITY ANALYSIS TECHNIQUES FOR COMPUTER MODELS. ACCIDENTS METHOM, COSTS AND SCHEDULES FOR DECON- NUREG/CR 4199. A DEMONSTRATION UNCERTAINTY / SENSITIVITY TAM; NATION- ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE MODEL CRAC2. HARRINGTON K.M NUREG/CR-4342- UNCERTAINTY AND SEN$lTIVITY AP.ALYSIS OF A NUREG/CR-3905 V02' SEQUENCE CODING AND SEARCH SYSTEM MODEL FOR MULTICOMPONENT AEROSOL DYNAMICS. FOR UCENSEE EVENT PEPORTS Code Listings. NUREGICA 3905 V03: SEQUENCE CODING AND SEARCH SYSTEM HEN AGER,C.H. FOR UCENSEE EVENT REPORTS Coder's Manual. NUREG/CR-3905 V04. SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4070 V03: BlVALVE FOUUNG OF NUCLEAR POWER FOR UCENSEE EVENT REPORTS. Coder's Manual. PLANT SERVICE-WATER SYSTEMSFactors That May intenssty The HARRINGTON.R.M. NUREG/CR-3764 BWR-LTAS. A BOluNG WATER REACTOR LONG- HENNICK A. NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79-09 FAILURE OF GE NU /CR 3885 VO H GH-TEMP RATURE GAS-COOLED REACTOR TYPE AK-2 CIRCUlf BREAKERS IN SAFETY-RELATED SYSTEMS SAFETY STUDIES FOR THE DIVISON OF ACCIDENT NUREG/CR-3794. CLOSEOUT OF IE BULLETIN 80-25 OPERATING PROBLEMS WITH TARGET ROCK SAFETY REUEF VALVES AT NURE 88 V04 i H. E P 0 GE D OR BWRS. SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NUREG/CR4003. CLOSEOUT OF IE BULLETIN 79-04 tNCORRECT EVALUATION Ouarterty Progress Report. October 1 December WElGHTS FOft SWING CHECK VALVES MANUFACTURED BY 3t.1984 VELAN ENGINEERING CORPORATION. NUREG/CR4402 V01: HIGH-TEMPERATURE GAS COOLED REACTOR NUREG/CR-4004: CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF SAFETY STUDIES FOR THE DIVISION OF ACCIDENT WESTINGHOUSE BFD RELAYS IN SAFETY-RELATED SYSTEMS. EVALUATION Ouarterly Progress Report. January 1. March 31,1985. NUREG/CR-4005: CLOSEOUT OF IE BULLETIN 8012 DECAY HEAT HARRIS,P.A. REMOVAL SYSTEM OPERABluTY. NUREG/CR4006: CLOSEOUT OF IE BULLETIN 81-01. SURVEILLANCE NUREG/CR4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS CF MECHANICAL SNUBBERS. SAFETY ANALYSIS Phase 1. Fmal Report HENNINGER,RJ. HARRISON.B.D-NUREGICR-4085: USERS MANUAL FOR CONTAIN 1.0 A Computer NUREG/CR-4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE 1 Code for Severe Reactor Acendent Containment Analysis. PRESSURIZED WATER REACTOR. HARTLEY,J.N. HENSLEY W T NUREG/CR4088: METHODS FOR ESTIMATING RADIOACTIVE AND NUREG/CN4022: PRESSURIZED THERMAL SHOCK EVALUATON OF TOXIC AIRBORNE SOURCE TERMS FOR URANIUM MILLING OPER- THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. ATONS. NUREG/CR4089- EVALUATION OF FIELD TESTED FUGITIVE DUST HNE CONTROL TECHNIOUES FOR URANIUM MILL TAIUNGS PILES. p p NATURALLY AGED CLASS 1E ExtDE FHC-19 BATTERY CELLS HARTY,R. NUREG/CR-4096: TEST SERIES 3 SEISMIC FRAGluTY TESTS OF NUREG/CR4297. EXTRE MITY MONITORING Considerations For NATURALLY-AGED CLASS 1E C&D LCU 13 BATTERY CELLS. Use.Dossmeter Placement.And Evaluation. NUAEG/CR4097: TEST SERIES 4 SEISMIC-FRAGluTY TESTS OF NUREG/CR 4399: POSSIBLE OPTIONS FOR REDUCING OCCUPA-NATURALLY AGED EXIDE EMP-13 BATTERY CELLS. TIONAL DOSE FROM THE TVL2 BASEMENT.
Personal Author index 123 HERCZEG,A.L NOREG/CR-3663 V01: PROBABlUTY OF PIPE FAILURE IN THE REAC. NUREG/CR4094. FIELD EXPERIMENT DETERMINATONS OF DISTRI-TOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR BUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE PLANTS. Volume 1. Summary Report. LAKE ENVIRONMENTS. NUREG/CR-4237: MOBILITY OF RADIONUCUDES IN HIGH CHLORIDE HOOK,5.L i ENVIRONMENTS. i NUREG/CR-3518: A SURVEY OF THE USES OF RADIOACTIVE MATE-HERSKOVITZ,M.5. RIALS IN LOUISIANA'S OFFSHORE WATERS. NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN- HOPENFELD.J. STRUT 8ENTATION IN REFLOOD TEST FACluTIES. NUREG-1108: RADIOACTIVITY TRANSPORT FOLLOWING STEAM HESSON,G.M. GENERATOR TUBE RUPTURE. NUREG/CR3659- A MATHEMATICAL MODEL FOR ASSESSING THE HORAN.J.R.
' UNCERTAINTIES OF INSTRUMENTATON MEASUREMENTS FOR POWER AND FLOW OF PWR REACTORS. NUREG/CP-0068: PROCEEDINGS OF AN INTERNATIONAL WORK-SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTON (ALARA) AT NUCLEAR POWER PLANTS.MAY 29-JUNE 1.1984.
NUREG/CR-407L EXPLORATORY TREND AND PA' TERN ANALYSIS HORSCHEL.D.S. FOR 19810CENSEE EVENT REPORT DATA. NUREG/CR-3647: DESaGN AND FABRICATON OF A 1/8-SCALE STEEL CONTAINMENT MODEL NUREG/CR-4038 SENSITIVITY AND UNCF RTAINTY STUDIES OF THE HORTON.W. CRAC2 COMPUTER CODE. NUREG/CR-4392: MEASURES OF SAFEGUARDS RISK EMPLOYING HIGGINS.J.C. PRA (MOSREP) A Methodology For Estimating Risk impacts Of Safe-NUREG/CR-4440 A REVIEW OF EMER3ENCY DIESEL GENERATOR Quards Measures. PERFORMANCE AT NUCLEAR POWER PLANTS. HOSKER.R.P. HILEMAN,J.A. NUREG/CR4038:
~~^ SENSITIVITY AND UNCERTAINTY STUDIES OF THE NUREG/CR-4145: EARTHOUAKE RECURRENCE INTERVALS AT NU.~~
~~CLEAR POWER PLANTS: ANALYSIS AND RANKING.~~
HSIEH.B.J. HILL,0.F NUREG/CR4432: COMPARISON OF DYNAMIC CHARACTERISTICS NUREGICR-4088: METHODS FOR ESTIMATING RADIOACTIVE AND OF FUKUSHIMA NUCLEAR POWER PLANT CONfAINMENT BUILD-TOXIC AIRBORNE SOURCE TERMS FOR URANIUM MILUNG OPER- ING DETERMINED FROM TESTS AND EARTHOUAKES. ATONS. HUANG.C. NUREG/CR4042. A 3-DIMENSIONAL COMPUTER MODEL TO SIMU-HINKLE.N.E'R-4403: NUREG/C
~~SUMMARY~~
OF THE WASTE MANAGEMENT PRO- LATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A GRAMS AT URANIUM RECOVERY FACluTIES AS THEY RELATE TO ROCK FRACTURE SYSTEM. THE 40 CFR PART 192 STANDARDS. HUCHTON.R.L HINZE.W.J. NUREG/CR-3455: A COMPARISON OF IODINE. KRYPTON.AND XENON NUREG/CR3174 V02: GEOPHYSICAL. GEOLOGICAL STUDIES OF RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD-POSSIBLE EXTENSIONS OF THE NEW MADRIO FAULT SORPTION MEDtA. ZONE Annual Report For 19c3. HUENEFELD.J.C. HISER.A. NUREG/C43883: ANALYSIS OF JAPANESE-U.S. NUCLEAR POWER NUREG/CR4395: CORRELATON OF CV AND KIC/KJC TRANS4 TON PLANT MAINTENANCE. TEMPERATURE INCREASES DUE TO IRRADIATION. HUMPHREYS.P. HITCHCOCK J.T. NUREG/CR4022. PRESSURIZED THERMAL SHOCK EVALUATION OF NUREG/CR4060: THE DC 1 AND DC-2 DEBRIS COOLA81UTY AND THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. MELT DYNAMICS ELPERIMENTS' HUMPHREYS.P.C. HOSSS.R.W. NUREG/CR4016 V01: APPUCATION OF SUM-MAUD A TEST OF AN NUREG/CR.3872: DATA ACQUISITION AND CONTROL OF THE HSST INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING SERIES V IRRADIATON EXPERIMENT AT THE ORR. EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND REUA8fuTY. Volume f. Main Report. HOCHREITER.LE. NUREG/CR 4166: ANALYSIS OF FLECHT-SEASET 163-ROD BLOCKED HUNTSMAN.R.L . CONDLE DATA USING COBRA.TF. NUREG/CR4167. FLECHT SEASET PROGRAM Final Report.NRC/EPRI NUREG/CR 4033: THE ROLE OF PERSONAL AIR SAMPUNG IN RADI. Westmghouse Report Number 16. ATION SAFETY PROGRAMS AND RESULTS OF A LABORATORY EVALUATION OF PERSONAL AIR. SAMPLING EQUIPMENT. HOFFMAN.F.O. HUTTON.P.H. NUREG/CF,-3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUCL CYCLE.A Renew Of Data For Technetom. NUREG/CR3825 V034: ACOUSTIC EMISSION / FLAW RELATIONSHIP FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE HOFMAYER,C.H. VESSELS Quarterty Report, Aprd 1984 September 1984. Volumes 3 and 4. NUREG/CP-0070; PROCEEDINGS OF THE WORKSHOP ON SEISMIC CaND DYNAMIC FRAGIUTY OF NUCLEAR POWER PLANT COMPO- NUREG/CR3915: ACOUSTIC EMISSION RESULTS OBTAINED FROM NENTS. TESTING THE Z8-1 INTERMEDIATE SCALE PRESSURE VESSEL NUREG/CR 4300 V01: ACOUSTIC EMISSION / FLAW RELATIONSHIP HOLLAND.R.A. FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE VESSELS Drogress Report. October-March 1985. NUREG-09M: CLOSEOUT OF IE BULLETIN 7912:SHORT-PERIOD SCRAus AT BOillNG-WATER REAR. TORS. HWANG.H.
~~' HOLMAN.G.S. NUREG/CR3876: PROBABiUTY BASED LOAD COMBINATION CRITE.~~
NUREG/CA-3660 V01: PROBA81UTY OF PIPE FAILURE IN THE REAC- RfA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS Volume NUREG/CR4329- REUABILITY EVALUATON OF CONTAINMENTS IN-t: Summary Report. CLUDING SOIL. STRUCTURE INTERACTION. NUREG/CR-3660 V04: PROBA81LITY OF PIPE FAILURE IN THE REAC. H W A NG,W.S. TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 4 Pye Fadure induced By Crack Growth in West Coast Plants. NUREG/CR4128- 8WR FULL INTEGRAL SIMULATION TEST (FIST) PHASE 11 TEST RESULTS AND TRAC-8WR MODEL CUAUFICATION
i
'124 Personal Author Index ILSERG,0. . . NUREG/CR-2482 V09: REVIEW OF DOE WASTE PACKAGE NUREG/CR-2815 V01 R1: PROBABrUSTIC SAFETY ANALYSIS PROCE- PROGRAM.Semannual Report Covenng The Per'od Aprd 1985-Sep-DURES GUIDE. Sections 17 And Appendices. tomber 1985.
NUREG/CR4050: A REVIEW OF THE SHOREHAM NUCLEAR POWER NUREG/CR-3091 V04: REVIEW OF WASTE PACKAGE VERIFICATON STATON PROBABILISTIC RISK ASSESSMENT. internal Events And TESTS Semiannual Report Covermg The Peod October 1983. March Core Damage Frequency 1984. NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFICATON i N RE /CR-3904: A COMPARISON OF UNCERTAINTY AND SENSITIV. 198 iTY ANALYSIS TECHNIOUES FOR COMPUTER MODELS. l- NUREG/CR.4122: A FORTRAN 77 PROGRAM AND USER'S GUOE JAMtSON.J.D. FOR THE CALCULATON OF PARTIAL CORRELATION AND STAND- NUREG/CR-4151: INTEGRATION OF EMERGENCY ACTON LEVELS ARDlZED REGRESSION COEFFICIENTS. WITH COMBUSTION ENGINEERING EMERGENCY OPERATING NUREG/CR411nr A DEMONSTRATON UNCERTAINTY /SENSITMTY PROCEDURESBy Use Of Combustion Engineenne Owners Group A YS U NG THE. HEALTH AND ECONOMIC CONSEQUENCE Emergency Opeiating Procedure Technical Guidelines. NUREG/CR-4342: UNCERTAINTY AND SENSITMTY ANALYSIS OF A JANG J NT COURSE NUREG-0837 V04 NO3: NRC TLD DIRECT RADIATION MONITORING NU E / 50 V02 PROBA IUS I ASS S DOCUMENTATON. Volume 2-Probability And Statistics For PRA Apph. NURE 083 N C R R TON MONITORING cations. RIPORT. Progress Report. October-December 1984. NUREG-0837 VOS N01: NRC TLD DIRECT RADIATION MONITORING gyg,j,g, NETWORK Progress Report, January. March 1985. j NUREG/CR-4133: NUCLEAR FOWER SAFETY REPORTING SYSTEM NUREG-0837 V05 NO2: NRC TLD DIRECT RADIATION MONITORING IMPLEMF.NTATION AND OPERATIONAL SPECIFICATIONS. NETWORKProgress Repor1, Aprd June 1985. IRELAND,J.R.
~~.JANKOWSKl.M.W.~~
NUREG/CR-3706: TRAC ANALYSES OF SEVERE OVERCOOUNG NUREG-0856 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES TRANSIENTS FOR THE OCONEE.1 PWR. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment). ISAACSON,L. JAROSS R.A. NUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR WASTE GAS SYSTEMS. NUREG/CR4180: STATE-OF.THE-ART OF SOUD. STATE MOTOR v CONTROLLERS. ISACHSEN,Y,W.
, NUREG/CR-3178: STRUCTURAL AND TECTONIC STUDIES IN NEW JARRELL.D.B. ' YORK STATE. Final Report. July 1981. June 1982. NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARAC.
TERISTICS USING EXTRAPOLATION TECHNIQUES. ISHil.M. NUREG/CR4277: INVERTED ANNUAL FLOW EXPERIMENTAL STUDY. JEANMOUO8N,N. NUREG/CR-4294. LEAK RATE ANALYSIS OF THE WESTINGHOUSE REANR COON N NU F 249: PRESSURE VESSEL FRACTURE STUDIES PENE-i TRATING TO THE PWR THERMAL-SHOCK ISSUE. EXPERIMENTS JENKINS,J.P. TSE-5.TSE-5A AND TSE4 NUREG/CR4040: OPERATIONAL DECISIONMAKING AND ACTION SE. NUREG/CR-4304; PRESSURE VESSEL FRACTURE STUDIES PER- LECTION UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER TAINING TO THE PWR THERMAL-SHOCK ISSUE.Expenment TSE-7. PLANTS.
~~~ JENNE,E.A.~~
UR G/CR4022: PRESSURIZED THERVAL SHOCK EVALUATION OF NUREG/CR4030. RADIONUCUDE MIGRATION IN GROUND THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. WATER.(Final Report) JACKSON.D.H. JEUNG,N. 1 NUREG/CR4261: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR NUREG/CR-4355 V01: 238 PU(IV) IN MONKEYS.Overvew Of Metabo-POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. lism. I JACKSON.P.O. NUREG/CR4057; RADOLOGICAL ASSESSMENT OF THE TOWN OF JO.J. NUREG/CR-3703: ASSESSMENT OF SELLCTED TRAC AND RELAPS
EDGEMONT. CALCULATIONS FOR OCONEE.1 PRESSURIZED THERMAL SHOCK JACDes.G.K. STUDY.
~~NUREG/CP-0062 PROCEEDINGS OF THE CONFERENCL JN THE AP. - JO,J.H.~~
~~! PUCATON OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR NUREG/CR4022: PRESSURIZED THERMAL SHOCK EVALUATION OF~~
~~WASTE REPOSITORY ASSESSMENT, THE CALVERT CUFFS UNIT 1 NUCLEAR PCWER PLANT.~~
NUREG/CR-3851 V03: PROGRESS IN EVALUATON OF RADONU- NUREG/CR4253. REVIEW OF TRAC CALCULATONS FOR CALVERT CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- CUFFS PTS STUDY. LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS Report For Apni June 1964. NUREG/CR4292: A COMPARATlVE ANALYSIS OF CONSTITUTIVE RE-
~~, LATIONS IN TRAC.PFL AND RELAP5/ MOD 1.~~
NUREG/CR3851 V04: EVALUATON OF RADIONUCUDE GEOCHEMI-CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR JOHNSON.A.C. WASTE REPOSITORY SITE PROJECTS Annual Progrers Report For October 1983-September 1984 NUREG/CR4288: FOCAL MECHANISM ANALYSES FOR VIRGINIA AND EASTERN TENNESSEE EARTHOUAKES (19781984).
~~, NUREG/CR-4236 V01: PROGRESS IN EVALUATION OF RADONU-CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- JOHNSON,J.D.~~
LEVEL NUCLEAR WASTE FIEPOSITORY SITE PROJECTS AEPORT NUREG/CR-3857: PREUMINARY SCREENING OF FUEL CYCLE AND FOR OCTC8E4 DECEMBER 1984. BY-PRODUCT MATERIAL LICENSES FOR EMERGENCY PLANNING. NUREG/CR4236 V02: PROGRESS IN EVALUATION OF RADIONU- NUREG/CR4122' A FORTRAN 77 PROGRAM AND USER'S GUIDE CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH- FOR THE CALCULATION OF PARTIAL CORRELATION AND STAND-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTSReport for 4 ARDIZED REGRESSION COEFFICIENTS.
! January-March 1985. NUREG/CR4199: A DEMONSTRATION UNCERTAINTY / SENSITIVITY i 1 JAIN,H. ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE '
~~MODEL CRAC2.~~
~~! NUREG/CR.2482 V08. REVIEW OF DOE WASTE PACKAGE NUREG/CR4342: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A i~~
~~PROGRAM Semiannual Report Covenng The Penod October 1984 - MODEL FOR MULTICOMPONENT AEROSOL DYNAMICS March 1985 I i~~
~~. _ - - . _ _ . . .=__ - _-~~
Personal Author index 125 l JOHNSON.J.J. KASSIR,M, NUREG/CR-4331: SIMPLIFIED SElSMIC PROBABluSTIC RISK NUREG/CR-4221: AN EVALUATION OF STRESS CORROSION CRACK i ASSESSMENT. Procedures And Lrrwtatens. l GROWTH IN BWR PIPING SYSTEMS.
~~JOHNSON.K.L~~
' KASSNER.T.F. ' NUREG/CR-4070 V02: BlVALVE FOUUNG OF NUCLEAR POWER NUREG/CR4287; ENVIRONMENTALLY ASSISTED ' CRACKING IN PLANT SERVICE-WATER SYSTEMS. Volume 2. Current Status Of Bio-I fou4ng Sunreillance And Control Techniques. UGHT WATER REACTORS. Annual Report,0ctober 1983 - September t984. NUREG/CR4070 V03: BlVALVE FOUUNG OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS Factors That May intensify The N ATO,W.Y. Safety Consequences Of Befouhng NUREG/CR 3026; FEASIBluTY STUDY ON THE ACQUISTON OF U-NUREG/CR-4267: VESSEL INTEGRITY SIMULATION (VISA) CODE CENSEE EVENT DATA. SENSITivlTY STUDY. NUREG/CR4152: AN INDEPENDENT SAFETY ORGANIZATION. JOHNSON.M.P. ' KE LLER.G.R-
, NUREG/CR-3905 V01 R1: SEQUENCE CODING AND SEARCH NUREG/CR-3174 V02: GEOPHYSICAL GEOLOGICAL STUDIES OF NUI R3 5 2 OE E N N S CH SYSTEM POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT FOR UCENSEE EVENT REPORTS. Code Ustings. ZONE. Annual Report For 1983.
NUREG/CR-390$ V03: SEQUENCE CODING AND SEARCH SYSTEM KELLY,J E. ~ NU EG 3 . EQU NCE G SEARCH SYSTEM NUREG/CR 2951: THE D9 EXPERIMENT. Heat Removal From Stratified FOR UCENSEE EVENT REPORTS Coder's Manual 002 Debns. PUREG/CR-3922 Vot: SURVEY AND EVALUATION OF SYSTEM NUREG/CR-4060: THE DC 1 AND DC-2 DEBRIS COOLABluTY AND INTERACTION EVENTS AND SOURCES Main Report And Appendices MELT DYNAMICS EXPERIMENTS. A And B NUREG/CR 3922 V02: SURVEY AND EVALUATION OF SYSTEM KELLY,J.M. INTERACTON EVENTS AND SOURCES. Appendices C And D. NUREG/CR-4166: ANALYSIS OF FLECHT-SEASET163-ROD BLOCKED BUNDLE DATA USING COBRA.TF. JONES.J.W. [ MUREG/CR-3736: FIELD AND THEORETICAL INVESTIGATONS OF KELLY,P. FRACTURED CRYSTALUNE ROCK NEAR ORACLE, ARIZONA. NUREG/CR4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR-JONES,T.N. POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU-CLEAR REACTOR VESSELS AND PIPING COMPONENTS. NUREG/CR4092: ORNL CHARACTERIZATON OF HEAVY SECTION STEEL TECHNOLOGY PROGRAM PLATES 01,02.AND 03. KELMERS.A.D. JORDAN.H. NUREG/CR 3851 V03. PROGRESS IN EVALUATON OF RADIONU-CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH. CR 8 AE d BEH V RM L NG (TASK 3) - SUP. o, [ j^9 A SG S Npon PORT SERVICES FOR RESEARCH AND EVALUATION OF SEVERE ACCOENT PHENOMENA AND MITIGATON NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI-CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR JORGENSEN.C.C- WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For NUREG/CR-3481 V02 NUCLEAR POWER PLANT PERSONNEL OUAU- October 1983-September 1984 FICATIONS AND TRAINING: TAPS - The Task Analyses Profihng NUREG/CR 4238 V01: PROGRESS IN EVALUATON OF RADIONU. Y" CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-JUNG,R.G. LEVEL NUCLEAR WAS1E REPOstTORY SITE PROJECTS REPORT e FOR OCTOBER-DECEMBER 1984. WUREG/CR4172: A USER'S GUIDE FOR MERGE. NUREG/CR-4236 V02: PROGRESS IN EVALUATION OF RADIONU-KAGAMt.S. CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report for NUREG/CR 3878: PROBABluTY BASED LOAD COMBINATION CRITE- January-March 1985. RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. K E MPF,C.R. KAHL,W.K. WUREG/CR 3831: THE IN. PLANT REUAaluTY DATA BASE FOR NU- NUREG/CR '865: EVALUATION OF THE RADIOACTIVE INVENTORY IN.AND ESilMATION OF ISOTOPIC RELEASE FROM,THE WASTE IN
~~, CLEAR PLANT COMPONENTS intenm Report Desel 1~~
Generators,Battenes. Chargers And Inverters. ElGHT TRENCHES AT THE SHEFFIELD LOW LEVEL WASTE BURIAL SITE. NUREG/CR4406. AN ANALYSIS OF LOW-LEVEL WASTES Review of WUR G/CR-3981: BICACCUMULATON OF P 32 IN BLUEGILL AND a ns An scaton of Ra@acWe Wed CATFISH. , *sfa p KEMPKA.S.N. NU' REG /CR4031 V02: NEUTRON SPECTRAL CHARACTER 12ATIONNUREG/CR4136: SMOKE.A Data Reduction Package For Analysis Of Combustion Expenments. FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST) IR-RADIATION SERIES. "Neutronics Calculations." KEMPPAINEN M NUREG/CR4031 V03: NEUTRON SP2CTRAL CHARACTERIZATON FOR THE FIFTH HEAVY SECTON STEEL TECHNOL,0GY (HSST)lR- NURUi/ CHI 4121. ErFECTS OF SULFUR CHEM 6STRY AND FLOW RADIATON SERIES. " Neutron Exposure Parameters RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRON-MUREG/CR4284. NEUTRON EXPOSURE PARAMETERS FOR THE MENTk
~~, FIFTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATION KENNE DY,R.P.~~
SERIES-NUREG/CR.3660 V03: PROBABILITY OF PIPE FAILURE IN THE REAC. KAO.C- TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS Volume NUREG/CR 3876: PROBABlUTY BASED LOAD COMBINATION CRITE- 3 Guillotine Break Indirectly induced By Earthquakes. RfA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. NUREG/CR-3663 V03: PROBABluTY OF PIPE FAILURE IN THE REAC-TOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR KARIMIAN.S. PLANTS, Volume 3. Double Ended Guillotine Break Indrectfy induced By WUREG/CR4392. MEASURES OF SAFEGUARDS RISK EMPLOYlNG Earthquakes. PRA (MOSREP) A Methodology For Estimating Risk Impacts Of Safe- NUREG/CR-3805 V02. ENGINEERING CHARACTERIZATION OF guards Measures. GROUND MOTION Task 11: Effects Of Ground Motion Charactenstics nASHlWA,5.A. On Structural Response Considenng Locahzed Structural Nonhneantes And Soil-Structure interaction Ettects NUREG/CR4079: ANALYTIC STUDIES PERTAINING TO STEAM GEN. NUREG/CR4334. AN APPROACH TO THE QUANTIFICATION OF SEIS-ERATOR TUBE RUPTURE ACCIDENTS. MIC MARGINS IN NUCLEAR POWER PLANTS. l
126 Personal Author Index KEIUN,T.W. KOCHER.D.C. NUREG/CR-4256: MEASUREMENT OF RESPONSE TIME AND DETEC- NUREG/CR-4038. SENSITIVITY AND UNCERTAINTY STUDIES OF THE TION OF DEGRADATON IN PRESSURE SENSOR / SENSING LINE CRAC2 COMPUTER CODE. SYSTEMS. KOEHL,E.R. N E /CR 111: A COMPARATIVE STUDY OF HEPA FILTER EFF1- NTRO R S. CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR-JET-GENERATED Dl-24THYLHEXYL SEBECATE.DI-2-ETHYLHEXYL KOENIG,J.E. PHTHALATE.AND SODIUM CHLORfDE. NUREG/CR-4022. PRESSURIZED THERMAL SHOCK EVALUATON OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. KESSLER J.H NUREG/CR-4109. TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR-f4UREG/CRI3851 V04. EVALUATION OF RADIONUCUDE GEOCHEMI- IZED-THERMAL SHOCK TRANSIENTS AT CALVERT CUFFS / UNIT CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR LA Cornewton Evneenng PWR. WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For October 1983-September 1984. KOESTER,J.P. NUREG/CR-4430: CURRENT METHODOLOGIES FOR ASSESSING KEYS W.S. THE POTENTIAL FOR EARTHOUARKE INDUCED UOVEFACTION IN NUREG/CR-3736: FIELD AND THEORETICAL INVESTOATONS OF SOfLS. FRACTURED CRYSTALLINE ROCK NEAR ORACLE. ARIZONA. KOGAN,V, KHATIB-RAHSAR NUREG/CR-3498: TWO-DIMENSIONAL MODEUNG OF INTRA SUBAS- NUREG/CR-4388- AEROSOL BEHAVOR MODELING (TASK 3) - SUP-SEMBLY HEAT TRANSFER AND BUOYANCY INDUCED FLOW RE. PORT SERVICES FOR RESEARCH AND EVALUATON OF SEVERE DISTRIBUTON IN tMFBRS. ACCIDENT PHENOMENA AND MITlGATlON. NUREG/CR4143: REVIEW AND EVALUATON OF THE MILLSTONE KOH,8.R. UNIT 3 PROBABlUSiaC SAFETY STUDY Containment Fadure Modes. Radiological Source Terms And Offsste Consequences. NUREG/CR-3889 THE MODELING OF BWR CORE MELTDOWN ACCl-DENTS FOR APPUCATON IN THE MELRPIMOD2 COMPUTER KILLOUGH.G.G. CODE. NUREG/CR 4038: SENSITIVITY AND UNCERTAINTY STUDIES OF THE CRAC2 COMPUTER CODE. KOMRT.J.M. NUREG/CR-4166: ANALYSIS OF FLECHT SEASET 163-ROD BLOCKED - KIM,H.J'G/CH4414. DIRECT CONTACT CONDENSATION OF STEAM ON BUNDLE DATA USING COBRA TF. i NUHE COLD WATER IN STRATIFIED COUNTERCURRENT FLOW. KOLLAR,F NUREG/CR-4317 V01: CANADIAN SEISMIC AGREEMENT.Techrucal STRAT ED S EAM ATER FLOW Report Covenng 1973-1985. KilS.H. KOPSTEIN,F.F. NUREG/CR 3889: THE MODEUNG OF BWR CORE MELTDOWN ACU, NUREG/CR-3634: MAINTENANCE PERSONNEL PERFORMANCE SIM-DENTS - FOR APPUCATION IN THE MELRPl. MOD 2 COMPUTER CODE. ULATION (MAPPS) MODEL USER'S MANUAL NUREG/CR-4104. MAINTENANCE PERSONNEL PERFORMANCE SIM-KIM8ALL.C S. ULATON (MAPPS) MODEL Feld Evaluation /Vahdation. NUREG/CR 3747: THE SELECTION AND TESTING OF ROCK FOR AR- < MOR!NG URANIUM TAluNGS IMPOUNDMENTS. KOT C.A. NUREG/CR4432: COMPARISON OF DYNAMIC CHARACTERISTICS KINCAID,R.H. OF FUKUSHIMA NUCLEAR POWER PLANT CONTAINMENT BUILD-
~~, MUREG/CR-3805 V02: ENGINEERING CHARACTERIZATION OF ING DETERMlNED FROM TESTS AND EARTHOUAKES.~~
GROUND MOTON Task II: Effects Of Ground Motion Charactenstics On Structural Response Cons <senng Localized Structural Nonhneanties KOUSAR1,8. And Sod-Structure interaction Effects. _NUREG/CR-4174. ROCK MASS SEAUNG EXPERIMENTAL' ASSESS-MENT OF BOREHOLE PLUG PERFORMANCE Annual Report. June i KINNISON.R. 1983 May 1984' I NUREG/CP-0063. PROCEEDINGS OF THE 1984 ST .nSTICAL SYMPO- a SIUM ON NATIONAL ENERGY ISSUES. KRAMARIC,M. NUREG-0837 V04 NO3: NRC TLD DIRECT RADIATON MONITORING KIPP,T.R. NETWORK. Progress Report. July-September 1984. NUREG/CR4290 V02- PROBABiUTY OF PIPE FAILURE IN THE REAC- NUREG 0837 V04 N04: NRC TLD DIRECT RADIATON MONITORING TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR PLANTS. Volume 2.Gudlotine Break Indirectty induced By Earthquakes. NURE 3 05 N C D ADI TION MON!TORING NETWORK. Progress Report. January-March 1385. KITTMER,C.A. NUREG-0837 VOS NO2 NRC TLD DIRECT RADIATION MONITORING NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV. NETWORK Progress Report Aprd June 1985. OR DURING STATION BLACKOUT. KRAME R,0. KMETYK,L.N. NUREG/CR-3802: RELAPS ASSESSMENT.OUANTITATIVE KEY PA. NUREG/CR-4082 V01: DEGRADED PIPING PROGRAM - PHASE 11 Semiannual Report. March 1984 - September 1984. RAMETERS AND RUN TIME STATISTICS NUREG/CR 3936 RELAPS ASSESSMENT. CONCLUSIONS AND USER NUREG/CR-4062 V02: DEGRADED PIPING PROGRAM - PHASE ll Semiannual Report, October 1984 March 1985 GUIDEUNES. K NEE.H.E. K RANTZ,E.A. NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PERFORMANCE NUREG/CR-3301: CATALOG OF PRA DOMINANT ACCIDENT SE. SIMULATION (MAPPS) MODEL: DESCRIPTION OF MODEL QUENCE INFORMATON. CONTENT. STRUCTURE.AND SENSITIVITY TESTING. KRASNER.LM. NUREGICR4231: EVALUATON OF AVAILABLE DATA FOR PROBABI-KNIGHT NUREG/CR- T.D' 3208: TRAC-PO2 DEVELOPMENTAL ASSESSMENT.USTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR NUREG/CR 3646: THAC PF1 INDEPENDENT ASSESSMENT. POWER PLANTS. NUREG/CR-3866: TRAC PD2 INDEPENDENT ASSESSMENT. KROIS,P.A. KNOPOFF,L. NUREG/CH-4145: EARTHOUAAE RECURRENCE INTERVALS AT NU- NUREG/CR-3887: HUMAN FACTORS REVIEW FOR SEVERE ACCI. CLEAR POW'CR PLANTS: ANALYSIS AND RANKING. DENT SEQUENCE ANALYSIS.
Personal Author Index 127 K UJ.Y. LARSEN,R.P. l NUREG/CR4038. SENSITIVITY AND UNCERTAINTY STUDIES OF THE CRAC2 COMPUTER CODE NUREG/CR4208: GASTROINTESTINAL ABSORPTION OF PLUTONIUM IN MICE. RATS, AND DOGS Applicaton To EstablisNn0 Values Of it [ KUHLMAN M.R. OO "D"'""' i NUREG/CR-4173: CORSOR USER'S MANUAL l WUREG/CR4205: TRAP-MELT 2 USER'S MANUAL R KULAK,R.F. LARSONJ[Chl4041: EG SYSTEM ANALYSIS HAND 8OOK. NUREG/CR4041 ROI: SYSTEM ANALYSIS HANDBOOK. NUREG/CR-4064 STRUCTURAL RESPONSE OF LARGE PENETRA-I TIONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED LASSITER.D.L TO LOADINGS BEYOND DESIGN BASIS. NUREG/CR4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND i KULLDERG.C.M. ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW ~
~~. PERFORMANCE IN CONTROL ROOM SIMULATOR $.~~
WUREG/CR-3977: RELAPS THERMAL-HYDRAUUC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN- LAUGHERY,K.R. SON UNIT 2 PRESSURIZED WATER REACTOR NUREG/CR-4196: OVERVIEW OF TRAC-801 (VERSION 12) ASSESS- NUREG/CR4206. A SELECT REVIEW OF THE RECENT (1979-1983) MENT STUDIES- BEHAVIORAL RESEARCH UTERATURE ON TRAINING SIMULA-TORS. KUNSMAN,0.M. NUREG/CR 4335: POTENTIAL BENEFITS OBTAINED BY REOUIRING LEAHY'T.J-SAFTEY-GRADE COLD SHUTDOWN SYSTEMS. NUREG/CR4350 V01: PROBABruSTIC RISK ASSESSMENT COURSE DOCUMENTATON. Volume 1. PRA Fundamentals. KUPPERMAN D.S. WUREG/CR 4124: NDE OF STAINLESS STEEL AND ON-UNE LEAK LEEJ.Y. MONITORING OF LWRS. Annual Report. October 1983 . September NUREG-0017 R01: CALCULATION OF RELEASES OF RADIOACTIVE 1984 MATERIALS IN GASEOUS AND UQUO EFFLUENTS FROM PRES-- NUREG/CR 4368: NDE OF STAINLESS STEEL AND ON-UNE LEAK SURIZED WATER REACTORS (PWR-GALE CODE). MONITORING OF LWRS: Semiannual Report. October 1984 . March 1985. LE E,S.C. CURTH.R. NUREG/CR-4416. STA8iUTY OF STEAM. WATER COUNTERCURRENT STRATIFIED FLOW. NUREG/CR 4144. IMPORTANCE RANKING BASED ON AGING CON-SIDERATIONS OF COMPONENTS INCLUDED IN PROBABluSTIC LE E.S.Y. RSit MSESSMENTS. NUREG/CR-3851 V03: PROGRESS IN EVALUATION OF RADONU-KURTZ,RJ CLlDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH. NUREG/CR-3825 V03-4: ACOUSTIC EMISSION / FLAW RELATONSHIP LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS Report FOR IN-SERV!CE MONITORING OF NUCLEAR PRESSURE VESSELS.Ouarterly Report Apnl 1984 . September 1984. Volumes 3 NIRE R 51 V04: EVALUATON OF RADIONUCUDE GEOCHEMI-CAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR NURE /C43915: ACOUSTIC EMISSION RESULTS O8TAINED FROM WASTE REPOSITORY SITE PROJECTS Annual Progress Report For TESTING THE Z8-1 INTERMEDIATE SCALE PRESSURE VESSEL October 1983-September 1984. NUREG/CR4300 V01; ACOUSTIC EMISSION / FLAW RELATONSHIP FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE LEHMICKE,D.J. VESSELS Progress Report October-March 1985. NUREG/CR4173: CORSOR USER'S MANUAL M W A N.Q. LEIGH.C.D. NUREG/CR-4357: THE FEASl81UTY OF DETECTING THE IMPORT OF NUREG/CR-4342: UNCERTAINTY AND SENSITIVITY ANALYSIS OF A UNAUTHORIZED RADOACTIVE MATERIALS INTO THE UNITED MODEL FOR MULTICOMPONENT AEROSOL DYNAMICS. STATES. LENACH.S.V. LAATS.E.T. NUREG/CR 3943: THE BWR PLAN ANALYZER RUREG/CR-2531 R03: INTRODUCTORY USER'S Mat.UAL FOR THE U.S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE. LEMEUR,M. SEAhCH DATA BANK. NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCf. EY E DENT SIMULATIONS ON POLYMER PROPERTIES. MUREG/CR-3889: THE MODEUNG OF BWR CONE MELTDOWN ACCI- LEONARD,M. DENTS FOR APPUCATION IN THE MELRPl. MOD 2 COMPUTER NUREG/CR4177 V01: MANAGEMENT OF SEVERE NU EC/CR4116. NUFEGO-NP:A OtGITAL COMPUTER CODE FOR ACCIDENTS Perspectrves On Managing Severe Accdents in Commer THE UNEAR STABluTY ANALYSIS OF BOfLING WATER NUCLEAR REAMORS. NU GC 17 VO : MANAGEMENT OF SEVERE ACCIDENTS Extending r> tant Operating Procedures into The Severe LAl,W. Accident Regime NUREG/CR-3854: FABRICATON CRITERIA FOR SHIPPING CONTAIN-LAMSE,W.M. NUREG/CR4100' EVALUATON OF INSTRUMENTAL ME. HODS FOR THE MEASUREMENT OF YELLOWCAKE EMISSIONS. NUREG-0970: PROCEDURES FOR MEETING NRC ANTITRUST RE-SPONSIBluTIES. LEVERENZ,F.L. LAMONICA,LO. NUREG/CR-4144: IMPORTANCE RANKING BASED ON AGING CON-NUREG/CR-4022: PRESSURIZED THERMAL SHOCK EVALUATION OF SIDERATIONS OF COMPONENTS INCLUDED IN PROBABluSTIC RISK ASSESSMENTS. THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. NUREG/CR4350 V06. PROBABILISTIC RISK ASSESSMENT COURSE LANDOW.M. DOCUMENTATON. Volume 6. Data Development NUREG/CR-4082 V01: DEGRADED PIPING PROGRAM . PHASE LEWELLEN.W S NR CR DEGR D NN ROGRAM . PHASE NUREG/CR4157: A SCIENTIFIC CRITIOUE OF AVAILABLE MODELS itsermannual Report. October 1984. March 1985, FOR REAL. TIME SIMULATIONS OF DISPERSION. NUREG/CR4158: A COMPILATION OF INFORMATON ON UNCER. LANNING.D.D. TAINTIES INVOLVED IN DEPOSITION MODELING NUREG/CR4168: GT2F. A COMPUTER CODE COR ESTIMATING NUREG/CR-4159: COMPARISON OF THE 1981 INEL OtSPERSION LIGHT WATER REACTOR FUEL ROD FAILURES. DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.
128 Personal Author Index LEWIS J.R. LOPES.J.C. NUREG/CR-4298: DESIGN AND INSTALLATION OF COMPUTER SYS- NUREG/C44424. DROPLET SIZES DYNAMICS AND DEPOSITON IN TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55. VERTICAL ANNULAR FLOW. LEWIS.M. . LORENZ,R.A.
- NUREG/CR-4361: STEAM GENERATOR GROUP PROJECT Annual NUREG/C43930: OBSERVED BEHAVIOR OF CESIUM.ODINE,AND Report - 1983. TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-NUREG/CR-4362: STEAM GENERATOR GROUP PROJECT. Annual GRAM.
~~Report 1984. NUREG/CR-4037: DATA~~
~~SUMMARY~~
~~REPORT FOR FISSION PRODUCT RELEASE TEST HI-5. LEWIS.P.M. NUREG/CR-4043: DATA~~
~~SUMMARY~~
REPORT FOR FISSION PRODUCT NUREG/CR-4248: RECOMMENDATIONS FOR NRC POUCY ON SHIFT RELEASE TEST Hit SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS. LORET,L.S. LEYLAK.J. NUREG/CR4397. IN-PLANT SOURCE TERM MEASUREMENTS AT NUREG/CR-4120:' MATHEMATICAL MODEUNG OF ULTIMATE HEAT PRAIRIE ISLAND NUCLEAR GENERATING STATON. SINK COOLING PCNDS. LOSS,M < U,0.W NUREG/CR 3228 V03: STRUCTURAL INTEGRITY OF WATEH REAC. NUREG/CR4072: THE ESTIMATON OF ATMOSPHERIC DISPERSION TOR PRESSURE BOUNDARY COMPONENTS Annual Report For AT NUCLEAR POWER PLANTS UTILIZING REAL TIME ANEMOME, 1984. TER STATISTICS. LOYOLA,V.M. WAO.LH. NUREG/C43361: THE EFFECT OF WATER CHEMISTRY ON THE NUREG/CR 4376 HEAT TRANSFER. CARRYOVER AND FALL BACK IN RATES OF HYDROGEN GENERATON FROM GALVANIZED STEEL PWR STEAM GENERATORS DURING TRANSIENTS. CORROSION AT POST-LOCA CONDITIONS. NUREG/CR-3803: THE E'FFECTS OF POST LOCA CONDITIONS ON A UDIAK.E.G. NUREG/CR-3174 V02: GEOPHYSICAL. GEOLOGICAL STUDIES OF PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-DUSTRY, POSSISLE EXTENSIONS OF THE NEW MADRID FAULT ZONE. Annual Report For 1983. UGON D.M. NUREG/C44263: REUABluTY ANALYSIS OF STIFF VERSUS FLEXI-NUREG/C44303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS' BLE PIPING FINAL PROJECT REPORT. SAFETY ANALYSIS. Phase 1 Final Report. LUCKAS W.J. UPINSKI R.J. NUREG/C43026: FEASIBTUTY STUDY ON THE ACQUISTION OF U. , NUREGICR-2951: THE 09 EXPER: MENT. Heat Rernoval From Stratified CENSEE EVENT DATA. 002 Debns. LUDEWlG,H. WPPINCOTT,E.P. NUREGICR-4143. REVIEW AND EVALUATON OF THE MILLSTONE NUREG/CR3746 V02: LWR PRESSURE VESSEL SURVEILLANCE DO- UNIT 3 PROBABILISTIC SAFETY STUDY.Contamment Fadure SIMETRY IMPROVEMENT PROGRAM Serniannual Progress Modes. Radiological Source-Terms And Offsete Consequences. - Report.Apnl 1984. September 1984.
! LLCYD R.D. NUREG/CR-4176 EMISSION CONTROL TECHNOLOGY AND QUAUTY NUREG/CR-4382: CONCENTRATONS OF URANIUM AND THORIUM ASSURANCE NEEDS AT URANIUM MILLING FACluTIESIncludes l lSOTOPES IN URANIUM MILLERS
~~AND M:NERS' TISSUES- Supporting Methods For Teshng, Operating.And Martainin0 Air Pollu-tion Control Devices.~~
NUREG/C43660 V04. PROBABILITY OF PIPE FAILURE IN THE REAC- LUMM D.K. TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS.Volurne NUREG/C44333. STE GENEVfEVE FAULT ZONE. MISSOURI AND IL-4 Pipe Fadure induced By Crack Growth in West Coast Plants. LINOIS. NUREG/C43663 V01: PROBABILITY OF PIPE FAILURE IN THE REAC. TOR COOLANT LOOPS OF COMBUSTON ENGINEERING PWR LYONS,J.A. PLANTS. Volume 1. Summary Report. NUREG/CR4317 VOI: CANADIAN SEISMIC AGREEMENT.Techrscal Report Covenng 1979-1985. p, NUREG/C44392 MEASURES OF SAFEGUARDS RISK EMPLOYING MACKENZIE D.R. PRA (MOSREP).A Methodology For Eshmahng Risk impacts Of Safe- NUREG/CR 3865. EVALUATION OF THE RADIOACTIVE INVENTORY guards Measures- IN.AND ESTIMATON OF ISOTOPIC RELEASE FROM THE WASTE IN ROA hlGHT TRENCHES AT THE SHEFFIELD LOW-LEVE PUREG/CR-4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE NUREG/CR4062: EXTENDED STORAGE OF LOW-LEVEL RADICAC-REACTOR COOLANT PUMP. TlVE WASTES Potenhal Problem Areas. I NUREG/CR4215: TECHNICAL FACTORS AFFECTING LOW LEVEL
*ASTE OR A TA C CR TER LOFGREN NUREGICR E.V'2815 V01 R1: PROBABILISTIC SAFETY ANALYSIS REG / PROCE-Hazardous Waste Regulabons And Identification of Radoactive Maed NU E /C 3 3 PR ABUS K ASSESSMENT COURSE Wastes. Final Report-DOCUMENTATION. Volume 3 . Systems Rehabelity And Analyses MACKOWIAK,0.P.
NUREG 3 $ C L'ISTIC RISK ASSESSMENT COURSE DOCUMENTATONVolume 5 - 9/ stems Rehabihty And Analysis NUREG/C43862: DEVELOPMENT OF TRANSIENT INITIATING EVENT FREQUENCIES FOR USE IN PROBABluSTIC RISK ASSESSMENTS. Techniques. Session D - QuantAcation. ' MADDEN.E.G. LONGEST,A.W. ' NUREG/CR4104. MAINTENANCE PERSONNEL PERFORMANCE SIM-NUREG/CR4346: AEROSOL RELEASE EXPERIMENTS IN THE FUEL ULATON (MAPPS) MODEL Field Evaluation / Validation. j AEROSOL SIMULANT TEST FACluTY:UNDERSODIUM EXPERI. MENTS. MAERKER.R.E. LOOM 88,G.G. NUREG/CR4039: GAMMA RAY CHARACTERIZATON OF THE TWO. NUREG/CR-4073: RESULTS OF THE SEMISCALE MOD.28 STEAM YEAR 1RRADIATION EXPERIMENT PERFORMED AT THE PCOLSIDE GENERATOR TUBE RUPTURE TEST SERIES. FACluTY.
~~--, , ~ ~ , _ _ _ _ _ _ _ . _ _ . _ _ - _ - , - . _ _ . - - _ _ _ , _ , _ . - - _ _ _ . _ _ _ _ _~~
Personal Author Index 129 MAlYA P.S. MATTHEWS,0.R. NUREG/CR4287; . ENVIRONMENTALLY ASSISTED CRACKING IN UGHT WATER REACTORS. Annual Report,0ctober 1983 September NUREG/CR-4373: COMPENDIUM OF COST. EFFECTIVENESS EVALUA. 1984. TIONS OF MODIFICATONS FOR OOSE REDUCTION AT NUCLEAR l POWER PLANTS. ! MALLEN,A.N. M AX EY,W. j NUREG/CR-3943: THE BWR PLAN ANALYZER. NUREG/CR 4082 V01: DEGRADED PIPING PROGRAM - PHASE MANAHAN'M* ll Semiannual Report. March 1984 September 1984 NUREG/CR4177 VO1: MANAGEMENT OF SEVERE NUREG/CA 4082 V02: DEGRADED PIPING PROGRAM - PHASE ACCIDENTS Perspectives On Managing Severe Accidents in Comrner- ll Semiannual Report, October 1984 March 1985. cial Nuclear Power Plants. MAYO.C.W. MANDLER J W NUREG/CR 3991: FAILURE MODES AND EFFECTS ANALYSIS (FMEA) NUREG/CR 5245: IN. PLANT SOURCE TERM MEASUREMENTS AT OF THE ICS/NNI ELECTRIC POWER DISTRIBUTION CIRCUlTRY AT BRUNSWICK STEAM ELECTRIC STATON THE OCONEE-1 NUCLEAR PLANT. NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT PRAIRIE ISLAND NUCLEAR GENERATING STATION. NU /CR 3905 V01 R1: SEQUENCE CODING AND SEARCH MANN.N.R. SYSTEM FOR LICENSEE EVENT REPORTS. User's Guide. NUREG/CR4145: EARTHOUAKE RECURRENCE INTERVALS AT NU. NUREG/CR 3905 V02: SEOUENCE CODING AND SEARCH SYSTEM CLEAR POWER Pt ANTS. ANALYSIS AND RANKING. FOR UCENSEE EVENT REPORTS Code Ustings MANNING J.J- NUREG/CR 3905 V03: SEQUENCE CODING AND SEARCH SYSTEM FOR UCENSEE EVENT REPORTS Coder's Manual. NUREG/CR-3887: HUMAN FACTORS HEVIEW FOR SEVERE ACCI. NUREG/CR-3905 V04: SEQUENCE CODING AND SEARCH SYSTEM DENT SEQUENCE ANALYSIS FOR UCENSEE EVENT REPORTS Coder's Manual MARCH-LEUSA.J. MAZOUR.T.J. NUREG/CR 4256: MEASUREMENT OF RESPONSE TIME AND DETEC- NUREG/CR4344: INSTRUCTIONAL SKILLS EVALUATION IN NUCLE. TION OF DEGRADATION IN PRESSURE SENSOR / SENSING UNE AR INDUSTRY TRAINING. SYSTEMS. MCARTHUR D.A-MARSCHALL,C.W. NUREG/CR 3757; TRAN B 2;THE EFFECT OF LOW STEEL CONTENT NUREG/CR-4082 V01: DEGRADED PlPING PROGRAM . PHASE ll Semiannual Report, March 1984 September 1984- ON FUEL PENETRATON IN A NON-MELTING CYUNDRICAL FLOW CHANNEL NUREG/CR4082 V02: DEGRADED PIPING PROGRAM . PHASE ll Semaannual Report, October 1984 March 1985- NUREG/CR-3944 TRAN 0-3 EXPERIMENTAL INVESTIGATON OF FUEL CRUST STABluTY ON MFI TING SURFACES OF AN ANNU. CARTIN.D.M. LAR FLOW CHANNEL NUREG/CR-4198: FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS. TERS. M29lDE.A.F. NL. REG /CR-3991: FAILURE MODES AND EFFECTS ANALYSIS (FMEA) MARTIN,R.A. JF 1HE ICS/NNI ELECTRIC POWER DISTRIBUTION CIRCUlTRY AT NUREG/CR 4260: TORAC USER'S MANUALA Computer Code For Ana. THE OCONEE 1 NUCLEAR PLANT. lyzing Tornado induced Flow And Matenal Transport in Nuclear Feois- NUREG/CR4022: PRESSURIZED THERMAL SHOCK EVALUATION OF ties. THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. NUREG/CR-4264: INVESTIGATION ON HIGH-EFFICIENCY PARTICU-LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS. MCBRIDE,K.C. NUREG/CR4321: FULL SCALE MEASUREMENTS OF SMOKE TRANS. NUREG/CR4298. DESIGN AND INSTALLATION OF COMPUTER SYS-PORT AND DEPOSITION IN VENTILATON SYSTEM DUCTWORK. TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55. MARTIN,R.E. MCCANN.M. NUREG-1149: TECHNICAL SPECIFICATONS FOR UMERICK GENER- NUREG/CR-2815 V02 R1; PROBABluSTIC SAFETY ANALYSIS PROCE-ATING STATON, UNIT 1. Docket No. 50-352. (Phaladetptua Electne DURES GUIDE. Sections 8-12. MCCANN.M.W. MARTIN W.J. NUREG/CR 3485: PRA REVIEW MANUAL NUREG/CR-3709 METHODS OF minim 121NG GROUND. WATER CON. TAMINATON FROM IN SITU LEACH URANfUM MINING Final Report MCCARDELL.R.K. MARTIN 1.D.K. NUREG/CR-3948. EXPERIMENTAL RESULTS OF THE OPERATIONAL TRANSIENT (OOTRAN) TESTS 11 AND 12 IN THE POWER BURST NUREG/CR-3981: BIOACCUMULATION OF P-32 IN BLUEGILL AND FACIUTY. CATFISH. MARTZ,H.F. MCCLUNG.R.W. WUREG/CR-4217: A STATISTICAL ANALYSIS OF NUCLEAR POWER NUREG/CR 394v V01- EDDY CURRENT INSPECTON FOR STEAM PLANT VALVE FAILURE-RATE VARIABlWTY-SOME PREUMINARY GENERATOR TUBING PROGRAM. Serrmannual Progress Report For RESULTS- Penod Fnding Ane 30,1984 NUREG/CR-3949 V02: EDDY-CURRENT INSPECTON FOR STEAM MAST,P.K. CENERATOR TUBING PROGRA M. Annual Progress Report For Penod NUREG/CR 3757: TRAN 0-2:THE EFFECT OF LOW STEEL CONTENT Ending December 31,1984 ON FUEL PENETRATION IN A NON-MELTING CYUNDRICAL FLOW MCCLURE.J.D. NURE / 394a: TRAN B-3 EXPERIMENTAL INVESTIGATON OF NUREG/CR-3611: RADIOACTIVE MATERIAL (RAM) ACCIDENT / INCL-FUEL CRUST STABIUTY ON MELTING SURFACES OF AN ANNU- DENT DATA ANALYSIS PROGRAM. LAR FLOW CHANNEL MCCONNELL J.W. MATHEWS.P NUREG/CR-4150 EPICOR-il RESlN DEGRADATION RESULTS FROM WUREG4933 S02: A PRIORITilATON OF GENERIC SAFETY ISSUES. FiRST RESIN SAMPLES OF PF-8 AND PF-20. MATHIESON,T. MCCONNELL,R.J. RUREG/CR-4368. NDE OF STAINLESS STEEL AND ON-LINE LEAK NUREG/CR-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT. MONITORING OF LWRS: Semeannual Report October 1984 . Maren ABILITY PRACTICES MCCORMICK,R.D. MATHIEU,G.G. NUREG/CR 3948 EXPERIMENTAL RESULTS OF THE OPERATONAL NUREG/CR4237: MOBluTY OF RADIONUCUDES IN HIGH CHLORIDE TRANSIENT (OPTRAN) TESTS 11 AND 1-2 IN THE POWER BURST ENVIRONMENTS. FACluTY.
130 Personal Author Index MCCRAY,J.G. ME NSING.R.W. i NUREG/CR-4194. LOW LEVEL NUCLEAR WASTE SHALLCW LAND NUREG/CR-3660 V04: PROBABILITY OF PIPE FAILURE IN THE REAC-BURfAL TRENCH ISOLATION Fmal Report. October 1981 September TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS. Volume 1984. 4 Pipe Failure Induced By Cracil Growth in West Coast Plan's. MCCULLAGH.C.M. MERKLE,J.G. WUREG/CR-3237: CONTROL OF EXPLOSIVE MIXTURES IN PWR NUREG/CR 4015: EFFECT OF STAINLESS STEEL WELD OVERLAY WASTE GAS SYSTEMS. CLADDIN3 ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL PLATES IN BENDING SERIES 1. MCCULLOCH,R.W* NUREG/CR-4106: PRESSURIZED THERMAL SHOCK TEST OF 6-IN.- WUREG/CR-4106: PRESSURIZED THERMAL-SHOCK TEST OF 6-IN' THICK PRESSURE VESSELS.PTSE.1.Investigaton Of Warm Prestress-THICK PRESSURE VESSELS PTSE 1.Investigaton Of Warm Prestress- ing And Upper-Shelf Arrest. Mg And Upper Shell Arrest MERONEY,R.N. MCCULLOCH,W.H. NUHEG/CR4072: THE ESTIMATON OF ATMOSPHERIC OISPERSION NUREG/CR-3954: HECTR ANALYSIS OF EQUIPMENT TEMPERATURE AT NUCLEAR POWER PLANTS UTILIZING REAL TIME ANEMOME-RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CON- TER STATISTICS. DENSER CONTAINMENT. MESSIER,M.E. MCDONALD.S. WUREG-0714 V04 05: OCCUPATIONAL RADIATION NUREG-0970- PROCEDURES FOR MEETING NRC ANTITRUST RE-EXPOSURE. Fifteenth And Sateenth Annual Reports.1982 And 1983. SPONSIBluTIES. MCELROY,N.L METCALF,R-WUREG 1134 RADIATON PROTECTION TRAINING FOR PERSONNEL NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV-EMptOYED IN MEDICAL FACluTIES. OR DURING STATON BLACKOUT. MCELROY,W.N. METZGER,V, WUREG/CR 3319: LWR PRESSURE VESSEL SURVEILLANCE DOSIME- NUREG/CR-3208. TRAC-PD2 DEVELOPMENTAL ASSESSMENT. TRY IMPROVEMENT PROGRAM. LWR Power Reactor Survetilance Physses-Dosametry Data Base Compendium. MEYER.R.E. NUREG/CR-3746 V02: LWR PRESSURE VESSEL SURVEILLANN DO- NUREG/CR 3851 V03. PROGRESS IN EVALUATION OF RADIONU-SIMETRY IMPROVEMENT PROGRAM. Semiannual F.wress CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-Report.Apnl 1984. September 1984. LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTS Report NUREG/CR-3746 V03: LWR PRESSURE VESSEL SURVEILLANCE DO- For Apn4-June 1984. SIMETRY IMPROVEMENT PROGRAM.1984 Annual Report. October NUREG/CR 3851 V04. EVALUATION OF RADIONUCUDE GEOCHEMI-1,1983 - September 30,1984. CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Annual Progress Report For McEWEN.J.E. October 1983-September 1984. NUREG/CR4093: SAFETY / SAFEGUARDS INTERACTONS DURING NUREG/CR-4114. VALENCE EFFECTS ON THE SORPTON OF NU-
; SAFETY-RELATED EMERGENC.ES AT NUCLEAR POWER REACTOR CUDES ON ROCKS AND MINERALS.fi FAC1UTIES. NUREG/CR4236 V01: PROGRESS IN EVALUATION OF RADIONU.
CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS REPORT N G/ -3978: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL PLATE AND WELDS FOR THE NUREG 42 6 V E S IN EVALUATON OF RADIONU-CLIDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-NUR G/CR40 E ROPE ES OF IRRADIATED NUCLEAR LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report for GRADE PRESSURE VESSEL WELDS FOR THE THIRD HSST IRRA. January-March 1985. DIATION SERIES. MCGUtRE,M.V. MEYER R.O. NUREG 0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES NUREG/CR-4139: THE MAILED SURVEY A TECHNIOUE FOR 00TAIN. FOR ESTIMATING SOURCE TERMS (Draft Report For Comment). ING FEED 8ACK FROM OPERATONS PERSONNEL NUREG/CR-4173 CORSOR USER'S MANUAL MCGUIRE.R.K. NUREG/CR-4145: EARTHOUAKE RECURLENCE INTERVALS AT NU- MILIAN L CLEAR POWER PLANTS. ANALYSIS AND RANKING. NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMINATONS ON SOUDIFICATON. WASTE DISPOSAL AND ASSOCIATED OCCU-MCGUIRE,S.A. PATIONAL EXPOSURE. NUREG 1140 DAFT FC: A REGULATORY ANALYSIS ON EMERGENCY PREPAREDNESS FOR FUEL CYCLE AND OTHER RADOACTIVE MILLER,5.E. MATERIAL UCENSEES Draft Report For Comment NUREG/CR 3914: PUMP AND VALVE QUAUFICATON REVIEW GUIDE-MCNEON,T.J. MILLER,C.A. NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAILINGS PlLES: NUREG/CR-4182: VERIFICATION OF SOIL STRUCTURE INTERACTION A COMPARISON OF ANALYSIS TECHNOUES. METHOCS. NUREG/CR4061: LEACHATE PLUME MIGRATION DOWNGRADIENT FROM URANIUM TAluNGS DISPOSAL IN MINE STOPES. MfLLER,C.F. NUREG/CR4178 DRFT: AN EVALUATION OF SELECTFD UCENSE'E 3 "" N G[CR-3646: TRAC-PF1 INDEPENDENT A",SESSMENT. ' MEISTER.H. MEER.D'D' NUREG/CR-4055: THE D10 EXPERIMENT.COOLABILITY OF 002 NUREG/CR-4170 AN ULTRA-HIGH SPEED RESIDUE PROCESSOR DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. MELBER,5.D. MILLER,L F. NUREG/CR-4051: ASSESSMENT OF JOB-RELATED EDUCATONAL NUREG/CR-3872: DATA ACOUISITION AND CONTROL OF THE HSST
~~QUAU:~lCATONS FOR NUCLEAR POWER PLANT OPERATORS.~~
SERIES V IRRADIATON EXPERIMENT AT THE ORR. ? MENGS,W.J. NUREG/CR-4256: MEASUREMENT OF RESPONSE TIME AND DETEC-NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV- TON OF DEGRADATION IN PRESSURE SENSOR / SENSING LINE OR DURING STATON BLACTOUT. SYSTEMS. E l 1
Personal Author Index 131 MILLER.N.E. MOHR.P.B. NUREG/CR-3900 V02: LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR4035: A HIGHWAY ACCIDLNT INVOLVING RADIOPHAR-USED FOR HIGH-LEVEL WASTE PACKAGlNG Ouarterty Report. July- MACEUTICALS NEAR BROOKHAVEN.MISS SSIPPI ON DECEMBER September t984- 3.1983. NUREG/CR-3900 V03: LONG-TERM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING Ouarterfy MONROE.R.E. NUR V4 LN NR 30G EONED WQ NM @ USE W RM PERFORMANCE OF MATERIALS USED FOR HIGH-LEVEL WASTE PACKAGING Annual Report.Apnl THE FAB CATION OF SHIPPING CONTAINERS FOR RADIOACTIVE 1984 - Apnl 1985 ' NUREG/CR4379 V01: LONG-TERM PERFORVANCE OF MATERIALS S FOR HI LEVEL ASTE PACKAGING First Ouarterly MO . pE /CR4035: A HIGHWAY ACCIDENT INVOLVING RADIOPHAR. MACEUTICALS NEAR BROOKHAVEN.M SSISSIPPI ON DECEMBER MILLER.R.L 3.1983 NUREG/CR4090: EVALUATION OF NUCLEAR FACILITY DECOMMIS-SiONING PROJECTS. Annual Summary Report . Fiscal Year 1984 MOOR E.E.B. NUREG/CR-1755 ADD 01: TECHNOLOGY. SAFETY AND COSTS OF DE. MILLER.W.O. COMMISSIONING NUCLEAR REACTORS AT MULTIPLE-REACTOR NUREG/CR3774 V04: ALTERNATIVE METHOD FOR DISPOSAL OF STATIONS Effects On Decomm.ssioning Of Intenm inatality To Dispose LOW LEVEL RADIOACTIVE WASTE. Task 2C. Technical Requirements Of Wastes Offsite For Earth Mounded Concrete Burker Disposal Of Low Level Radioac. trve Waste. MOORE,T.O. MILL 3.W.R NUREG/CR-3953 THE USE OF MAG 1 SPECTACLES WITH POSITIVE-NUREG/CR 3791: CLOSEOUT OF IE BULLETIN 79 09 FAILURE OF GE AND NEGATIVE-PRESSURE RESP'RATORS TYPE AK.2 CIRCUIT BREAKERS IN SAFETY-RELATED SYSTEMS. MORETTI.E.S. MILSTEAD.W. NUREG/CR4208. GASTROINTESTINAL ABSORPTION OF PLUTONfUM NUREG-0933 SO2: A PRIORITIZATION OF GENERIC SAFETY ISSUES IN MICE. RATS. AND DOGS Application To Establishing VWes Of f t NUREG-0933 S03: A PRIORITIZAT!ON OF GENERIC SAFETY ISSUES For Soluble Plutonium. MINARICK.J.W. MORGENSTERN.M. NUREG/CR-4022. PRESSURIZED THERMAL SHOCK EVALUATION OF THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. NUREG/CR-3817: DEVELOPMENT.USE AND CONTROL OF MAINTE-NANCE PROCEDURES IN NUCLEAR POWER PLANTS Problems And MINNERS.W. Recommendations. NUREG-0933 S02: A PRIORITIZATION OF GENERIC SAFETY ISSUES NUREG/CR-3883. ANALYSIS OF JAPANESE-U S NUCLEAR POWER NUREG-0933 S03. A PRIORITIZATION OF GENERIC 6Al'ETY ISSUES PLANT MA4NTENANCE. MITCHELL.D.H. MORISSEAU,D.S. NUREG/CR-4023 FIELD PERFORMANCE ASSESSMENT OF SYN- NUREG/CR4139 THE MAILED SURVEY A TECHNIQUE FOR ODTAIN-THETIC LINERS FOR URAN!UM TAILINGS POND A Status Report. ING FEEDBACK FROM OPERATIONS PERSONNEL MITCHELLG.W. MORRIS,B M. NUREG/CR-2951: THE D9 EXPERIMENT. Heat Removal From Stratified NUREG 1144. NUCLEAR PLANT AGlNG RESEARCH (NPAR) PRO-NU EG CR 055. THE D10 EXPER6 MENT COOLABILITY OF UO2 GRAM PLAN DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL-MORRISON.H.W. MITCHELL,J.A. NUREG/CR-3837 MULTIPLE SEQUENflAL FAILURE NUREG-0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES MODEL Evaluat.on Of And Procedures For Human Error Dependency FOR ESTIMATING SOURCE TERMS (Draft Report For Comment). MOTES.B.G. MITRA.S. NUREG/CR3455. A COMPARISON OF IODINE. KRYPTON.AND XENON NUREG/CR4400: THE IMPACT OF MECHANICAL-AND MAINTE- RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD-NANCE-INDUCED FAILURES OF MAIN REACTOR COOLANT PUMP SORPTION MEDIA. SEALS ON PLANT SAFETY. MOUL.D.A. MlZNER.A.A' NURE CR- 3981: BIOACCUMULATION OF P-32 IN BLUEGILL AND NUREG/CR 4093. SAFETY / SAFEGUARDS INTERACTIONS DURING SAFETY.RELATED EMERGENCIES AT NUCLEAR POWER REACTOR F ACILITIE S. MJOLSNESS.R.C. NUREG/CR 4079. ANALYTIC STUDIES PERTAIN!NC TO STEAM GEN. MUHLHEIM.M.D. ERATOR TUBE RUPTURE ACCIDENTS. NUREG/CR-3922 VO1: SURVEY AND EVALUATION OF SYSTEM INTERACTION EVENTS AND SOURCES Main Report And Appendices MOAYERI.N. A And B NUREG/CR 4365. DESIGN AND DEVELOPMENT OF A SPECIAL PUR- NUREG/CR3922 V02. SURVEY AND EVALUATION OF SYSTE M POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU. INTERACTION EVENTS AND SOURCES Apperdces C And D CLEAR REACTOR VESSELS AND PtPING COMPONENTS MULCAHEY.T.P. NUR G/C d214: HEALTH EFFECTS MODEL FOR NUCLEAR POWER PLANT ACCIDENT ANALYSl$ Part NTRO A' CONSEQUENCE lintroduction. Integration & Summary Par 1 li. Scientific Basis For Health MUNRO,P.S. Effects Models NUREG/CR-4317 V01. CANADlAN SEISMIC AGREEMENT.Technscal MOERER.M.P. Report Covenng 1979 1985. NUREG/CR4160: HISTORICAL
~~SUMMARY~~
OF OCCUPATIDNAL RADI-E POSURE EXPERIENCE IN U S. COMMERCIAL NUCLEAR M S W AND EASTERN TENNESSEE EARTHOUAkES (1978-1984). MOHR.C.M. NUREG/CR3633 V01 S1: TRAC-BD1/ MOO 1 AN ADVANCED BEST ES. MUNSON.LF. TIMATE COMPUTER PROGRAM FOR BOfLING WATER REACTOR NUREG/CR 4399 POSSIBLE OPTIONS FOR REDUCING OCCUPA. TRANSIENT ANALYSIS. TIONAL DOSE FROM THE TMI-2 BASEMENT.
~~-w =_~~
~~r 132 Personal Author Index~~
~~} i MUNSON,LH. NELSON W.J ,~~
~~NUREG/CR4333 STE. GENEVIEVE FAULT ZONE. MISSOURI AND IL. is NUREG/CR4160. HISTORICAL~~
~~SUMMARY~~
OF OCCUPATIONAL RADI-ATION EXPOSURE EXPERIENCE IN U S. COMMERCIAL NUCLEAR LINOIS. - POWER PLANTS. E NELSON.W.R. - E MURATA.M.K. NUREG/CR4272- RESPONSE TREE EVALUATION EXPERIMENTAL NUREG/CR4085: USERS MANUAL FOR CONTAIN 10 A Compute' ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR Code for Severe Reactor Accident Containment Analysis OPERATORS. MURPHY,E. NETI,S NUREG 0844 DRFT FC: NRC INTEGRATED PROGRAM FOR RESOLU- A NUREG/CR-3193. FORCED CONVECTIVE.NONEOUILIBRIUM.PGsT-TION OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 REGARD- CHF HEAT TRANSFER EXPEHiMENT DATA AND CORRELATION ING STEAM GENERATOR TUBE INTEGRITY. Draft Report For Com- COMPARISON REPORT. c ment NEUM AN.S.P. i MURPHY,G.A. NUREG/CR-3736. FIELD AND THEORETICAL INVESTIGATIONS OF NUREG/CR-3922 V01: SURVEY AND EVALUATION OF SYSTEM FRACTURED CRYSTALLINE ROCK NEAR 09ACLE. ARIZONA. INTERACTION EVENTS AND SOURCES Main Report And Appendices '; A And B NEWTON,R.D. NUREG/CR-3922 V02: SURVEY AND EVALUATION OF SYSTEM NUREG/CR-4133. NUCLEAR POWER SAFETY REPORTING SYSTEM INTERACTION E' VENTS AND SOURCES Appendccs C And D f IMPLEMENTATION AND OPERATIONAL SPECIFICATIONS.
~~NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC -~~
~~MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY-FEA- NIC LS.F A. - TURE SYSTEMS OF NUCLEAR POWER PLANTS.~~
~~' LIGHT WATER REACTORS. Annual Report. October 1983 - Septembw MUSCARA.J.~~
1984. NUREG 1155 V02: RESE ARCH PROGRAM PLAN Steam Generators. NUREG-1155 V04. RESEARCH PROGRAM PLAN Non-Destructnre Ex- NICHOLSON,P.R. " aminaten. NUREG/CR4361: STEAM GENERA'OR GROUP PROJECT. Annual NUREG 0970. PROCEDURES FOR MEETING NRC ANTITRUST RE-Report - 1983 SPONSIBILITIES NUREG/CR4362. STEAM GENERATOR GROUP PROJECT. Annual Report - 1984. NICHOLSON T.J. r NUREG-1046: DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN
~~^ ZONE TECHNICAL CONSIDERATIONS AND RE-NUREG CR-3747: THE SELECTION AND TESTING OF ROCK FOR AR- TO -~~
MORING URANIUM TAILINGS IMPOUNDMENTS CO @ ,L T NAKAGAKIM NUREG/CR 3444 V02 THE IMPACT OF LWR DECONTAM; NATIONS NUREG/CR4082 V01: DEGRADED PIPING PROGRAM PHASE ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU- ', ll Semiannual Report. March 1984 - September 1984 PATIONAL EXPOSURE. NUREG/CR4082 V02: DEGRADED P6 PING PROGRAM PHASE ll Semiannual Report. October 1984. March 1985. WNM NALEZNY,C.L NUREG/CR-3930- OBSERVED BEHAVIOR OF CESIUM. LODINE.AND NUREG/CR-3005.
~~SUMMARY~~
~~OF THE NUCLEAR REGULATORY COM-TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-y-- MISSION'S LOFT PROGRAM RESEARCH FINDINGS GRAM y' NUREG/CR4351.~~
~~SUMMARY~~
~~REPORT FOR LOFT ANTIC! PATED #CREG/CR4037. DATA~~
~~SUMMARY~~
~~REPORT FOR FISSION PRODUCT TRANSIENT EXPERIMENT SERIES L6-8. RELEASE TES f Hf-5. l NUREG/CH-4043 DATA~~
~~SUMMARY~~
REPORT FOR FISSION PRODUCT ' NANSTAD,R.K. RELEASE TEST HI-6 . NUREG/CR-4015. EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR 4105. AN ASSESSMENT OF THERMAL GRADtENT TUBE
-a
- CLADDING ON THE STRUCTURAL INTEGRITY CF FLAWED STEEL RESULTS FROM THE Hi SERIES OF FISSION PRODUCT RELEASE - U TESTS. 8 PLATES IN BENDING SERIES 1 NUREG/CR4106 PRESSURIZED-THERMAL-SHOCK TEST OF 6-IN.. j THICK PRESSURE VESSELS PTSL1 investigatior Of Warm Prestress NOURBAKHSH,H.P. 1 - ing And Upper-Shelf Arrest NUREG/CR4022: PRESSURilED THERMAL SHOCK EVALUATION OF mir NUHEG/CR4249 PRESSURE VESSEL FRACTURE STUDIES PENE- THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. - - TRATING TO THE PWR THERMAL SHOCK ISSUE EXPERIMENTS TSE-5.TSE.5A AND TSE-6 NOVICK,V.J. " E NUREG/CR4304. PRESSURE VESSEL FRACTURE STUDIES PER. NUREG/CR4033 THE ROLE OF PERSONAL AIR SAMPLING IN RAOl-c TAINING TO THE PWR THERMAL. SHOCK ISSUE Expenment TSE 7. 5 ATION SAFETY PROGRAMS AND RESULTS OF A LABORATORY EVALUATION OF PERSONAL AIR. SAMPLING EQUIPMENT. NASSERSHARIF,0. - NUREG/CR4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE.1 NOWATZKl,E.A. g g PRESSURIZED WATER REACTOR' NUREG/CR4194- LOW-LEVEL NUCLEAR WASTE SHALLOW LAND BURIAL TRENCH ISOLATION Final Report. October 1981. September 5 NELSON.L.S. NUREG/CR 2718 STEAY EXPLOSION EXPERIMENTS WITH SINGLE 1964-g DROPS OF IRON OxlDE MELTED WITH A CO2 LASER Part II Parametnc Studies. NOYCE,J.R. NUREG/CR 001 ASSAY OF LONG-LIVED RADIONUCLIDES IN LOW-LEVEL WASTES FROM POWER REACTORS. ' NELSON.R.A. NUREG/CR-3193. FORCED CONVECTIVE.NONEOUILIBRIUM. POST- NUKARf,N.H. ; E CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION NUREG/CR-3488 V03. IDAHO FIELD EXPERIMENT 1981 Volume i E g COMPARISON REPORT. 3 Companson Of Tralectones.Concentrabon Patterns And MESODIF NELSON,R.W. Model Calculat ons - NUREG/CR4061. LEACHATE PLUME MIGRATION DOWNGRADIENT O'BRIEN.J. m._ FROM URANIUM TAIUNGS DISPOSAL IN MiNE STOPES. NUREG/CR 2815 V01 Rt- PROGABILISTIC SAFETY ANALYSIS PROCE-NELSON,W. DURES GUIDE Sections 17 Md Appendices.
~~NUREG/CR-3485 PR A REVIEW MANUAL.~~
~~i F NUREG/CR4398. COST ANALYSIS OF REVISIONS TO 10 CF9 PART NUREG/CR4392 MEASURES OF SAFEGUARDS RISK EMPLOYING~~
~~50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECWDARY PRA (MOSREP) A Methodology For Estimahng R:sk impacts Of Safe.~~
E CONTAINMENTS OF LIGHT-WATERCOOLED NUCLEAF POWER PLANTS guards Measures
= Personal Author Index 133 O'BRIEN,J.N. NUREG/CR4281: AN EMPIRICAL ANALYSIS OF SELECTED NUCLEAR NUREG/CR-3519. HUMAN ERROR PROBABILITY ESTIMATION USING POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. , LICENSEE EVENT REPORTS. L NUREG/CR-3837: MULTIPLE-SEOUENTIAL FAILURE OSBORNE,M.F. MODELivaluation Of And Procedures For Human Error Dependency NUREG/CR-3930. OBSERVED DEHAVIOR OF CESluM. LODINE.AND NUREG/CR4103. USES OF HUMAN RELIABluTY ANALYSIS PROB- TELLURIUM IN THE ORNL FISSION PRODUCT REl. EASE PRO-ABILISTIC RISK ASSESSMENT RESULTS TO RESOLVE PERSON- GRAM NEL PERFORMANCE ISSUES THAT COULD AFFECT SAFETY. NUREG/CR4037: DATA
~~SUMMARY~~
~~REPORT FOR FISSION PRODUCT RELEASE TEST Hi-5. OBERLANDER,P.L NUREG/CR-4043 DATA~~
~~SUMMARY~~
REPORT FOR FISSION PRODUCT NUREG/CR4251 V01: MITIGATIVE TECHNIQUES FOR GROUND- RELEASE TEST Hb6. WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR ACCIDENTS. Volume t Analysis Of Genenc Site Conditions OSTM EYER.R.H. i NUREG/CR4251 V02. MITIGATIVE TECHNIQUES FOR GROUND- NUREG/CR4185 AN ASSESSMENT GF DOSIMETRY DATA FOR AC-1 WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR CIDENTAL RADIONUCLIDE RELEASES FROM NUCLEAR REAC-ACCIDENTS Volunie 2. Case Study Analysis Of Hydrologic Character- TORS. ization And Mitgatrwe Schemes. OSTMEYER.R.M. s OGDEN.D.M. NUREG/CR4169: AN APPROACH TO TREATING RADIONUCLIDE Nt lREG/CA-3935. THERMAL-HYC,RAULIC ANALYSES OF OVERCOOL-DECAY HE ATING FOR USE IN THE MELCOR CODE SYSTEM ING SEQUENCES FOR THE H B RODINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUUY. OTTINGER.C.A. NUREG/CA-3977: RELAP5 THERMAL-HYDRAUUC ANALYSES OF NUREG/CR-2951- THE 09 DPER1 MENT. Heat Removal From Stratified PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B RO3lN- UO2 Debns. SON UNIT 2 PRESSURIZED WATEA REACTOR. NUR EG /CR-4055. THE D10 EXPERIMENT COOLA8lLITY OF 002 NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURFS DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL 7 ON TRANSIENTS AND ArCIDENTS AT A 3-LOOP WESTINGHOUSE i PRESSURIZED W ATER REACTOR Maa Report. OUELLETTE.A1. NUREG/CR4326 V02. EFFECTS OF CONTROL SYSTEM FAILURES NUREG.CP 397 V01 REACTION BETWEEN SGVE CESIUM-LODINE ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE COMPOUNDS AND 'HE R" ACTOR MATERIALS 304 STAINLESS PRESSURIZED WATER REACTOR Appendices STEEL.INCONEL 600 & SILVER * %ne I Cesium Hurom Reac-C LDH A M,R.D. _ NUREG/CR4208 GASTROINTESTINAL ABSOADM*: GF FLuiONIUM OWC A.W.A.
;N MICE. RATS. AND DOGS Appunon To Establishing Values Of f1 NUREG/CR4164 DATA REPORT FOR THE TPFL TE Ee CRIT. CAL For Soluble Plutonium. FLOW EXPER MENTS OLSE N,C.S. - OWCZARSKI.P.C.
NUREG/CR.4056 PARTICULATE AND GAS RE6 EASE FROM LIGHT. NUREG/CR-3317. TECHNICAL BASES AND USER S MANUAL FOR WATER-REACTOR (LWR) FUEL RODS STORED IN INERT AND DRY THE PROTOTYPE OF SPARC - A SUPPRESSION POOL AE3OSOL AIR ATMOSPHERES REMOVAL CODE. e ~ NUREG/CR-4074 THE PERFORMANCE OF DEFECTED SPENT LWR NUREG/CR-4130 ICEDF A CODE FOR AFROSOL PARTICLE CAP. FUEL RODS IN INERT GAS AND DRY AIR STORAGE ATMOS- TURE IN ICE COMPARTMENTS. PHERES F NURE 3/CR4084. DAY SPENT FUEL STORAGE TEST PLAN FOR DE. OWINGS T.D. STRUCTIVE FUEL ROO EXAM! NATIONS NUREG/CR-4015. EFFECT OF STAINLESS STEEL WELD OWHLAY a NUREG/CR-4345. INVESTIGATION OF THE STABruTY OF LWR CLADDING ON THE STRUC NRAL INTEGR'TY OF FLAWED STEEL g SPENT FUEL RODS DELOW 250 C- PLATES IN BENDING SERtES 1. r- OLSON.J. PAGE.RE. r NL REG /CR-3883 ANALYSIS OF JAPANESE-U S NUCLEAR POWER NUREG/CR-3613 V02. EVALUATION OF WELDED AND REPAlR. NU C 25 V I UIDEUNES AND WORKBOOK FOR ASSESS- fg D MAMSS Sm @ N WG. Annual Re# 6 MENT OF ORGANIZATION AND ADMINISTRATICN OF UTIUTIES NUREG/CR-3613 V03 N1: EVALUATION OF WELDED AND REPAlR-SEEKING OPERATING UCENSE FOR A NUCLEAR POWER WELDED STAINLESS S1 EEL FOh LWR St RVICE semiannual Report PLANT. Volume f. Guidelines For Utety Orgsnization And Admnstrahon For October 1984 Through March 1985 Pian- . NUREG/CR4125 VC2. GUIDELINES AND WORKBOOK FOR ASSESS-NUREG/CR-J911 V02. EVALUATION OF WTLDED At40 REFAIR-WELDED STAINLESS STEEL FOR LWR SERVICE Ouarterty MENT OF ORGANIZATION AND ADMIN!STRATION OF UTluTIES Report. April-June 1984 i SEEKING OPERATING LICENSE FOR A NUCLEAR POWER [- PLANT Volume 2 Workbook For Assessment Of Organization And Man- P A IK.C.Y. agemer.t NUREG/CR4166 ANALYSIS OF FLECHT-SEASET 163 ROD Bl.OCKED NUREG/CRJ281. AN EMPIRlCAL ANALYSIS OF SELECTED NUCLEAR BUNDLE DATA USING COBRA-TF. POWER PLANT MAINTENANCE FACTOHS AND PLANT SAFETY PALAZ20.R.J. OPITZ.B E. == NUREG/CR-3906- tsRANIUM M:LL TAluNGS NUREG/CR-3516 A SURVEY OF THE USES OF RADIOACTIVE MATE-Rf ALS IN LOUWANA'S OFFSHORE W ATERS . WEUTRALIZAT:ON CONTAMINANT COMPLEXATION AND TAILINGS F LEACHING STUDY PAPASPYROPOULOS NUREG/CR 4259 TAILINGS NEUTRAUZATION AND OTHER ALTER- NUREG/CR 4082 V01 - DEGRADED PIP:NG PROGRAM PHASC NATIVES FOR IMYOBILIZING TOxlC MATERIALS IN TAILINGS Final 11 Semiannual Report. March 1984 September 1984 - Report. NUREG/CR 4082 VO? DEGRADED PIPING PROGRAM PHASE OS80RN.R.N. NUREG/CR4125 V01 GUIDELINES AND WORKBOOK FOR ASSESS- PAPAZOGt OU,LA. MENT OF ORGANIZATION AND ADVtN STRATION OF UTILITIES NUREG/CH-2815 V01 Rt. PROBABluSTIC SAFETY ANALYSIS PROCE-SEEKING OPERATING UCENSE FOR A NUCLEAR POWER DURES GUIDE Sections 17 And Appendices PLANT. volume 1 Guidelines For Utility Organization And Admns*ahon NUREG/CR-3485 PHA REVIEW MANUAL Plan. NUHEG/CR 4228 REV EW OF THE VOGTLE UNITS 1 AND 2 AUX 1UA. NUREG/CR-4125 V02. GUIDEUNES AND WORKBOOK FOR ASSESS- RY FEEDWATER SYSTEM REUABlui/ ANALYSIS MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES SEEKING OPERATING LICENSE FOR A NUCLEAR POWER PAPPA S.R A. PLANT. Volume 2 Workbook For Assessment Of Organization And Man- NUREG/CA 3915 ACCUSTIC EMISSION RESULTS OBTA NLD FROM agement. TESTING THE 201 INTERMEDIATE SOALE PhESSURE VESSEL M E. M
134 Personal Author Index PARK.J.Y. NUREG/GR4200 Bf0 DEGRADATION TESTING OF SOLIDIFIED LOW-NUREG/CR4287: ENVIRONMENTALLY ASSISTED CRACKING IN LEVEL WASTE STREAMS. LIGHT WATER REACTORS. Anttal Report. October 1983 - September NUREG/CR-4201: THERMAL STABldTY TESTING OF LOW-LEVEL 1984. WASTE FORMS. NUREG/CR-4406: AN ANALYSIS OF LOW LEVEL WASTES Revew of PARKER.F. Hazardous Waste Regulations And identifcation of Radioactive Mixed NUREG/CR-4396: SIMMER POSTPROCESSOR MANUAL Wastes. Final Report. PARKER.G.B. PILCH.M. NUREG/CR-4075: DESIGNING PROTECTIVE COVERS FOR URANIUM NUREG/CR-4383. HIGH PRESSURE INJECTION OF MELT.FROM A MILL TAILINGS PILES. A Review. REACTOR PRESSURE VESSEL THE DISCHARGE PHASE. PARKER.S.F. PILGRIM'M K' NUREG/CR-4159: COMPARISON OF THE 1981 INEL DISPERSON NUREG/CR-4093: SAFETY / SAFEGUARDS INTERACTONS DURING DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS' SAFETY.RELATED EMERGENCIES AT NUCLEAR POWER REACTOR PARLOS.A. FACILITIES. NUREG/CR4376: HEAT TRANSFER. CARRYOVER AND FALL BACK IN PWR STEAM GENERATORS DURING TRANSIENTS. PIN.F G. I PASCHOA,AA GRAMS AT URANIUM RECOVERY FACIUTIES AS THEY RELATE TO NUREG/CR4382. CONCENTRATONS OF URANIUM AND THORIUM THE 40 CFR PART 192 STANDARDS ISOTOPES IN URANIUM MILLERS
~~AND MINERS' TISSUES.~~
FIRES,J. PASEDAG.W.F. NUREG/CR-4329: RELIABluTY EVALUATON OF CONTAINMENTS IN-NUREG-0856 DRFT FC: REA5SESSMENT OF THE TECHNICAL BASES CLUDING SOIL-STRUCTURE INTERACTION. FOR ESTIMATING SOURCE TERMS. (Draft Report For Cornrnent). PITTMAN.J. PASUPATHl.V. NUREG4933 S02: A PRIORITIZATION OF GENERIC SAFETY ISSUES. NUREG/CR4082 V01:' DEGRADED PIPING PROGRAM - PHASE NUREG 0933 S03: A PRIORITIZATION OF GENERIC SAFETY ISSUES. 11 Serniannual Report. March 1984 Septernber 1984. Nur1EG/CR-4082 V02: DEGRADED PIPING PROGRAM - PHASE PLOGER.S.A. fl.Semannual Report, October 1984 - March 1985. NUREG/CR 3948: EXPERIMENTAL RESULTS OF THE OPERATIONAL TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST PEA 80DY.C.A. FACILITY. NUREG-0856 DAFT FC: REASSESSMENT OF THE TECHN! CAL BASES FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment)- POOOWSKI M.Z. PELOOUfN.R.A. NUREG/CR-3889 THE MODEUNG OF BWR CORE MELTDOWN ACCI. WUREG/CR-2850 V03. POPULATON DOSE COMMITMENTS DUE TO DENTS . FOR APPUCATION IN THE MELRPIMOD2 COMPUTER RA ACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES NU EG CR-4116. NUFEGO-NP A OlGITAL COMPUTER CODE FOR THE UNEAR STABluTY ANALYSIS OF BOlUNG WATER NUCLEAR PELTO.P.J. REACTORS. NUREG/CR-4220: REUABILITY ANALYSIS OF CONTAINMENT ISOLA-TlON SYSTEMS, POLICASTRO.A.J. NUREG/CR-4120: MATHEMATICAL MODEUNG OF ULTIMATE HEAT PE NG.S.J. SlNK COOUNG PONDS. NUREG/CR4116: NUFEGO.NP A DIGITAL COMPUTER CODE FOR THE UNEAR STABILITY ANALYSIS OF BOluNG WATER NUCLEAR POLKY,J.N. . REACTORS. NUREG/CR4170: AN ULTRA-HIGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTION SYSTEM IMAGE ENHANCEMENT. PESCATORE.C. NUREG/CR.2482 V08: REVIEW OF DOE WASTE PACKAGE POLLACK H.N. < PROGRAM. Semiannual Report Covenng The Penod October 1984 - NUREG/CR-31;5 V03. GEOPHYSICAL INVESTIGATONS OF THE March 1985- WESTERN OHIO-INDIANA REGION . ANNUAL REPORT (October PETERSON.A.C. RUREG/CR-3772: RELAPS ASSESSMENT:SEMISCALE SMALL BREAK potoggs,J.p,
~~^ ^~~
~~W C 3919 RA 1/ 1 INDEPENDENT MT ENCE N~ ASSESSMENT.NEPTUNUS PRESSURIZER TEST YOS. POORE.W.P.~~
PETRYKOWSKI.J. NUREG/CR.3905 V02

SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4346:' AEROSOL RELEASE EXPERIMENTS IN THE FUEL .FOR UCENSEE EVENT REPORTS. Code Ustings.
AEROSOL SIMULANT TEST FACluTY.UNDERSODIUM EXPERI- NUREG/CR-3905 V03: SEQUENCE CODING AND SEARCH SYSTEM MENTS' FOR UCENSEE EVENT REPORTS Coder's Manual. NUREG/CR.3905 V04. SEQUENCE CODING AND SEARCH SYSTEM PHluPPACCPOULO FOR UCENSEE EVENT REPORTS. Coder's Manual. NUREG/CR4182: VERIFICATION OF SOIL STRUCTURE INTERACTON ODS POPELAR,C.
~~, PHILLIPS.L.D. NUREG/CR-4082 V01: DEGRADED PIPING PROGRAM - PHASE 11 Semannual ReportMarch 1984 . Septernber 1984.~~
NUREG/CR4022: PRESSURIZED THERMAL SHOCK EVALUATON CF THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANL NUREG/CR4082 V02: DEGRADED PIPING PROGRAM PHASE il Serrmannual Report October 1984. March 1985. PICIULO.P.L. NUREG/CR 3444 V02: THE IMPACT OF LWR DECONTAMINATIONS POSTMA.A.K. ON SOUDIFICATION. WASTE DISPOSAL AND ASSOCIATED OCCU- NUREG/CR.3317: TECHNICAL BASES AND USER'S MANUAL FOR ' THE PROTOTYPE OF SPARC - A SUPPRESSION POOL AEROSOL PATONAL EXPOSURE. NUREG/CR-4069- ANALYSES OF SOfLS FROM AN AREA ADJACENT REMOVAL CODE. TO THE LOW LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT SHfFFIELD,lLUNOIS. POWERS.D.A. NUREG/CR4083: ANALYSES OF SOILS FROM THE LOW LEVEL RA- NUREG/CR4401: BEHAVIOR OF CONTROL RODS DURING CORE i OlOACTIVE WASTE DISPOSAL SITES AT BARNWELLSC AND DEGRADATON: PRESSURIZATION OF SILVER-INDIUM.CADMlUM RICHLAND.WA, CONTROL RODS.
Personal Author index 135 POWERS.T.B. RANSOM.C.B. NUREG/CR-2800 S03. GUIDEUNES FOR NUCLEAR POWER PLANT NUREG/CR4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES SAFETY lSSUE PRIORITIZATION INFORMATION DEVELOPMENT- ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC PRAIRIE PLR BOILING WATER REACTOR Main Report. NUREG/CR4350 V02: PROBABILISTIC RISK ASSESSMENT COURSE NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES DOCUMENTATON Volume 2-Probatxlity And Statistes Fo* PAA Apph- ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC cahons. BOLLING WATER REACTOR. Appendices. NUREG/CR4326 VOI: EFFECTS OF CONTROL SYSTEM FAILURES PRATT,W. ON TR ANSIENTS AND ACCOENTS AT A 3-LOOP WESTINGHOUSE NUREG/CR-4143: REVIEW AND EVALUATION OF THE MILLSTONE PRESSURIZED WATER REACTOR Main Report. UNIT 3 PROBABluSTIC SAFETY STUDY. Containment Fadure NUREG/CR4326 V02: EFFECTS OF CONTROL SYSTEM FAILURES Modes.Radrologmal Source Terms And Offsite Consequences. ON TRANSIENTS AND ACCOENTS AT A 3 LOOP WESTINGHOUSE e PRESSURIZED WATER REACTOR Appendices. I PREVOST,J. NUREG/CR4294: LEAK RATE ANALYSIS OF THE WESTINGHOUSE RAO,K.S. ,' REACTOR COOLANT PUMP- NUREG/CR4038. SENSITIVITY AND UNCERTAINTY STUDIES OF THE PRICE.D.S. CRAC2 COMPUTER CODE. NUREG/CR 35te: A SURVEY OF THE USES OF RADCACTIVE MATE
RATH8UN.LA RIALS IN LOUISlANA'S OFFSHORE WATERS' NUREG/CR 4118. MONITORING METHODS FOR DETERMINATION PRICEJ.C. COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT NUREG/CR-4131: INVESTIGATION OF ALTERNATIVE MEANS TO AC. URANfUM RECOVL, Y SITES COMPLISH THE GOALS OF BIENNIAL lON CHAMBER CAllBRA-TON. RATLIFF,R.A.
NUREG/CR4352. SUGGESTED STATE REQUIREMENTS AND CRITE. PRICE.J.M. RfA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE NUREG/CR-3537: EXPEDIENT METHODS OF RESPIRATORY REGULATORY PROGRAM. PROTECTONill. SUBMICRON PARTICLE TESTS AND
~~SUMMARY~~
OF QUALITY FACTORS. RATZEL.A.C. PRINE.D.W NUREG/CR-4136: SMOKE.A Data Reduction Packag+ For Analyses Of NUREG/CR4124: NDE OF STAINLESS STEEL AND ON-LINE LEAK Combustion Erpenments. MONITORING OF LWAS. Annual Report. October 1983 September NUREG/CR-4138 DATA ANALYSES FOR NEVADA TEST SITE (NTS) 99g4 PREMIXED COMBUSTION TESTS. NUREG/CR4368: NDE OF STAINLESS STEEL AND ON-LINE LEAK MONITOR!NG OF LWRs: Semiannual Report. October 1984 March RAVINDRA.M.K. 1985. NUREG/CR-3660 V03 PROBABillTY OF PIPE FAILURE IN THE REAC-TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTSVotume PRUITT.J.S. 3 Gudiohne Break Indvectly induced By Earthquakes. NUREG/CR4266: STANDARD BETA-PARTICLE AND MONOENERGE- NUREG/CA 3663 V03. PHOBABILITY OF PIPE FAILURE IN THE REAC. TIC ELECTRON SOURCES FOR THE CAllBRATION OF BETA-RADI- TOR COOLANT LOOPS OF COMBUSTON ENGINEERING PWR ATION PROTECTION INSTRUMENTATON. PLANTS. Volume 3 Double Ended Guillotme Break Indireif" Induced By Earthquakes.
~~; PU.J.~~
~~NUREG/CR4290 V02: PROBABILITY OF PIPE FAILURE IN THE REAC.~~
- NUREG/CR-3703: ASSESSMENT OF SELECTED TRAC AND RELAPS CALCULATONS FOR OCONEE-1 PRESSURIZED THERMAL SHOCK TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR PLANTS Volume 2 Gudiotine Break Indirectfy Induced By Earthquakes STUDY.
PUGH.C.E. REECE W.D. NUREG/CR 3744 V02: HEAVY SECTION STEEL TECHNOLOGY PRO. NUREG/CRJ297: EXTREMITY MONITORING Considershons For GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL SEPTEMBER Use. Dos 4 meter Placement.And Evaluat ort 1984 NUREG/CR4219 V01: HEAVY SECTON STEEL TECHNOLOGY PRO REED.A.W. GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 NUREG/CR-4136: SMOKE.A Data Reduction Package For Analysis Of MARCH 1985. Combustion Expenments. C]ADFORD.LR. REED,J.W. NUREG/CR-3817; DEVELOPMENT.USE AND CONTROL OF MAINTE-NUREG/CR-2815 V02 R1: PROBABILISTIC SAFETY ANALYSIS PROCE. NANCE PROCEDURES IN NtOLEAR POWER PLANTS Problems And DURES GUIDE. Sections 8-12. Recomrnendations. NUREG/CR-3485. PRA REVIEW MANUAL RAGAN.G.E. NUREG/CR4334. AN APPROACH TO THE QUANTIFICATION OF SEIS-NUREG/CR-4256: MEASUREMENT OF RESPONSE TIME AND DETEC. MIC MARGINS IN NUCLEAR POWER PLANTS. TION OF DEGRADATION IN PRESSURE SENSOR / SENSING LINE R E E D.K.A. SYSTEMS. NUREG/CR-3953 THE USE OF MAG-1 SPECTACLES WITH POSITIVE-RAGNURAM,S. AND NEGATIVE-PRESSURE RESPIRATORS. NUREG/CR 4210: MATADOR A COMPUTER CODE FOR THE ANALY. SIS OF RADIONUCLIDE BEHAVOR DURING DEGRADED CORE AC. RE ESE.R.T. CIDENTS IN LIGHT WATER REACTORS NUREG/CR-3647. DESIGN AND FABRICATION OF A 1/8-SCALE i NUREG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF STEEL CONTAINMENT MODEL TRANSPORT AND DEPOSITION OF RADONUCLIDES) CODE DE. SCRIPTION AND USER'S MANUAL RElCH.M. i NUREG/CR.3876 PROBABILITY BASED LOAD COMBINATON CRiTE-RAMIRE2,A.L RfA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL NUREG/CR 4149. ULTIMATE PRESSURE CAPACITY OF REINFORCED WASTE REPOSITORY. Volume 1 Basalt- AND PRESTRESSED CONCRETE CONTAINMENT. RANDOLPH.P.D. NUREG/CR4182. VERIFICATION OF SOIL STRUCTURE INTERACT!CN NUREG/CR 323': CONTROL OF EXPLOSIVE MIXTURES IN PWR METHODS. WASTE GAS SYSTEMS. NUREG/CR4221 AN EVALUATION OF STRESS CORROSION CRACK GROWTH IN BWR PlPING SYSTEMS-RANKIN.W.L. NUREG/CR4329 RELIABluTY EVALUATON OF CONTAINMENTS IN-l NUREG/CR-3987: COMPUTERIZED ANNUNCfATOR SYSTEMS. CLUDING SOIL. STRUCTURE INTERACTION 1 a
~~, , , _ . _.___.w._r_, _m._ + . _ . . _ , _ _ _ _ _ , , _ , ,_y._,,_.,,,,-,m_-. _.,_.m ,,.r- _ y_, , , . . . - -~~
~
136 Personal Author Index REMEC.L RODA8AUGH,E.C. NUREG/CR 403t V02i NEUTRON SPECTRAL CHARACTERIZATION NUREG/GR-4305: COMMENTS ON THE LEAK-BEFORE BREAK CON-FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST) lR- CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS. RADIATION SERIES. Neutronics Calculatinns." NUREG/CR4031 V03: NEUTRON SPECTRAL CHARACTERIZATION ROHATGI,U.S. FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST) lR. NUREG/CR-3703. ASSESSMENT OF SELECTED TRAC AND RELAPS RADIATON SERIES. " Neutron Exposure Parameters." CALCULATIONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK STUDY. REST.J. NUREG/CR-4253: REVIEW OF TRAC CALCULATIONS FOR CALVERT NUREG/CR-3980 V02: UGHT-WATER-REACTOR SAFETY FUEL SYS- CUFFS PTS STUOY. TEMS RESEARCH PROGRAMS. Quarterty Progress Report.ApniJune NUREG/CR 4292: A COMPARATIVE ANALYSIS OF CONSTITUTIVE RE. 1984. LATIONS IN TRAC-PFL AND RELAPS/MODt. ! leUREG/CR-3980 V03: LIGHT. WATER. REACTOR SAFETY FUEL SYS-TEMS RESEARCH PROGRAMS. Quarterty Progress Report. July Sep.. ROLLER.S.F. tomber 1984. NUREG/CR-3721 V01: PRESSURF MEASUREMENTS IN A HYDROGEN NUREG/CR-3980 V04: LIGHT-WATER-REACTOR SAFETY FUEL SYS- COMBUSTION ENVIRONMENT. Hydrogen-At Combuston Test Senes ' TEMS RESEARCH PROGRAMS. Quarterty Progress Report. October- 1 And 2 in The FITS Tank. December 1984. p REXROTH.P.E NUREG/CR-3922 V01: SURVEY AND EVALUATION OF SYSTEM NUREG/CR4085: USERS MANUAL FOR CONTAIN 10 A Computer INTERACTION EVENTS AND SOURCES. Main Report And Appendices Code for Severe Reactor Accident Containment Analyses. A And B NUREG/CR 3922 V02: SURVEY AND EVALUATON OF SYSTEM CHODES.D.B. INTERACTION EVENTS AND SOURCES Appendices C And D. NUREG/CR4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV-OR DURING STATION BLACKOUT. ROSA.E.A. NUREG/CR4016 V01: APPUCATON OF SLIM.MAUD A TEST OF AN QANI.L , INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING
~~. NUREG-0933 SO2: A PRIORITIZATON OF GENERIC SAFETY ISSUES. EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND 2~~
NUREG4933 S03. A PRIORITIZATION OF GENERIC SAFETY ISSUES. REUABILITY. Volume I.Maen Report. RICH.B.L . ROSCOE.5.J. NVREG/CR 4033 THE ROLE OF PERSONAL AIR SAMPLING IN RADI- NUREG/CR-4250: VEHICLE BARRIERS. EMPHASIS ON NATURAL FEA-ATION SAFETY PROGRAMS AND RESULTS OF A LABORATORY TURES. EVALUATON OF PERSONAL AIR SAMPLING EQUIPMENT. ROSE,J.A. OCHARDSON.E. NUREG/CR-3819 SURVEY OF AGED POWER PLANT FACILITIES. NUREG-0714 V04-05: OCCUPATIONAL RADIATION EXPOSURE Fe enth And Sateenth Annual Reports,1982 And 1983. ROSS.P.A. NUREG 0020 V09 NO4. UCENSED OPERATING REACTORS STATUS
~~SUMMARY~~
REPORT. Data As Of March 31.1985(Gray Book 1) ' RIDEOUT,T.B. RUREG/CR-3987; COMPUTERIZED ANNUNCIATOR SYSTEMS. NUREG-0020 V09 N06: UCENSED OPERATING REACTORS STATUS
~~SUMMARY~~
REPORT Data As Of May 31.1985 (Gray Book 1) RIGGS.R. NUREG-0020 V09 N07: LICENSED OPERATING REACTORS STATUS NUREG U)3 SO2: A PRORITIZATION OF GENERIC SAFETY ISSUES
~~SUMMARY~~
REPORT Data As Of June 30.1985 (Gray Book 11 NUREG 0933 S03. A PRORIT12ATION OF GENERIC SAFETY ISSUES. NUREG-0020 V09 N09: UCENSED OPERATING RfACTORS STATUS
~~! MUREG 1128: TRIAL EVALUATIONS IN COMPARISON WITH THE 1983~~
~~SUMMARY~~
REPORT. NUREG-0020 Data As Of OPERATING V09 N11: LICENSED August 31.1985 (Gray BookS1) TATUS REACTORS SAFETY GOALS. RIORDAN.B. NUREG/CR-4398. COST ANALYSIS OF REVISIONS TO to CFR PART RUGER.C.
- 50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY NUREG/CR-2815 V02 R1: PROBABILISTIC SAFETY ANALYSIS PROCE-CONT 4tNMENTS OF LIGHT WATER-COOLED NUCLEAP POWER DURES GUIDE. Sections 812.
PLANTS. NUREG/CR4229 EVALUATION OF CURRENT METHODOLOGY EM-PLOYED IN PROBABIUSTIC RISK ASSESSMENT (PRA) OF FIRE RITCHIE.LT. EVENTS AT NUCLEAR POWER PUNTS. < NUREG/CR4350 V07; PROBABfUSTIC RISK ASSESSMENT COURSE DOCUMENTATON Volume 7 - Envronmental Transport And Conse- RUNKLE.G.E-quence Analysis. NUREG/CR-3657. PRELIM:NAAY SCREENING OF FUEL CYCLE AND BY-PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING QTTER.P.O. NUREG/CR-4185. AN ASSESSMENT OF DOSIMETRY DATA FOR AC. NUREG/CR4033 THE ROLE OF PERSONAL AIR SAMPLING IN RADI- CIDENTAL RADIONUCLIDE RELEASES FROM NUCLEAR REAC-i ATON SAFETY PHOGRAMS AND RESULTS OF A LABORATORY TORS. r EVALUATION OF PERSONAL AIR-SAMPLING EQUIPMENT. I RUTHER.W.E. ROBERSON.P.L NUREG/CR4287. ENVIRONMENTALLY ASSISTED CRACKING IN
NUREG/CR-4297

EXTREMITY MONITORING Considerations For LIGHT WATER REACTORS. Annual Report,0ctober 1983 September Use. Dosimeter Placement.And Evaluaton. 1984.
ROSERTSON.D.E. R YDE R.C.P. NUREG/CR4030: RADONUCUDE M1GRATON IN GROUND NUREG.0856 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES WATER (Frial Report) FOR ESTIMATING SOURCE TERMS (Draft Report 4r Comment). ROSINSON,G.C. SAARI.LM. ! NUREG/CR4015. EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR4051: ASSESSMENT OF JOB RELATEu EDUCATIONAL i CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL QUAUFICATIONS FOR NUCLEAR POWER PLANT OPERATORS PLATES IN BENDING SERIES 1. i NUREG/CR4106. PRESSURIZED-THERMAL-SHOCK TEST OF 6 IN.- SACCOMANNO.G.
~~THICK PRESSURE VESSELS PTSE 1.Investigaten Of Warm Prestress- NUREG/CR-4382: CONCENTRATIONS OF URANIUM AND THORIUM -~~
ing And Upper Shelf Arrest ISOTOPES IN URAN!UM MILLERS' AND MINERS' TISSUES. COCHELLE.J.M. S AGE NDORF.J.F. NUREG/CR4346. AEROSOL RELEASE EXPERlMENTS IN THE FUEL NUREG/CR 3488 V03 IDAHO FIELD EXPERIMENT 1981 Volume AEROSOL SIMULANT TEST FACIUTY UNDERSODIUM EXPERl- 3 Companson Of Trajectones. Concentration Patterns And MESODIF MENTS. Model Calculatons
~~--__r--- ,-, . _ , , , . _ _ . . . _ . - . . . _ . _ _ - . . _ , _ _ _ . . , _ . . _,_-_,,.. - _ _ _ _ , . _ - = . _ _ - - _~~
I Personal Author Index 137 SAHA.P. SCHWElZER,R.L. WUREG/C43026; FEASIBluTY STUDY ON THE ACO;.llSTON OF Ll-NUREG/C44093. SAFETY / SAFEGUARDS INTERACTIONS DURING CENSEE EVENT DATA. SAFETY RELATED EMERGENCIES AT NUCLEAR POWER REACTOR NUREG/CR 3703. ASSESSMENT OF SELECTED TRAC AND RELAP5 FACILITIES. CALCULATIONS FOR OCONEE.1 PRESSURIZED THERMAL SHOCK STUDY. SCIACCA.F. ' NUREG/CR-4252; INDEPENDENT ASSESSMENT OF TRAC.PD2/ MOD 1 CODE WITH BCL ECC BYPASS TESTS NUFIEG/CR-4398. COST ANALYSIS OF REVISIONS TO 10 CFR PART l go,AoFO;DtX J. LEAK TESTS FOR PRIMisRY AND SECONDARY SAHOTA M.S. CONTAINMENTS OF LIGHT. WATER COOLED NUCLEAR POWER PLANTS' NUREG/CR4278: TRAC-PF1/ MOD 1 DEVELOPMENT ASSESSMENT. SAROR,V L SCIACCA F.W. NUREG/CR-4067:
~~SUMMARY~~
OF BARPIER DEGRADATION EVENTS NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0.A Computer AND SMALL ACCIDENTS IN U.S. COMMERCIAL NUCLEAR POWER Code tw Severe Reacta Accident Conta.nment Anahsis. 3 4068:
~~SUMMARY~~
OF HISTORICAL EXPERIENCE WITH RE- SC W .R LEASES OF RADIOACTIVE MATERIALS FROM COMMERCIAL NU- NUREG/CR-4082 V01: DEGRADED PIPING PROGRAM PHASE CLEAR POWER PLANTS IN THE UNITED STATES. NRG 8 2 DEGR ED N OGRAM
~~PHASE SALAZAR.E.A. ILSemiannual Report October 1984 March 1985.~~
~~4 NUREG/CR-3863; ASSESSMENT OF CLASS 1E PRESSURE TRANS-ER ES S WHEN SUBJECTED TO HARSH ENVIRONMENT SEARS R~~
~
THE FABRICATION OF SHIPPING CONTAINERS FOR RADIOACTIVE SALLACH,H.A. MATERIALS. NUREG/CR-3197 V01: REACTION BETWEEN SOME CESlUM LODINE COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS SEAVER.D.A. STEEL,1NCONEL 600 & SILVER. Volume Ices.am HydrozKle Reac. NURFG/CR3688 V01: GENERATING HUMAN REUABIUTY ESil-bons. MATES USING EXPERT JUDGMENT. Volume t Ma#n Report. NUREG/CR-3688 V02; GENERATING HUMAN REUABluTY EST). SAMANTA,P.K. MATES USING EXPERT JUDGMENT Volume 2 Apperxisces. WUREG/CR 2815 V01 Rt. PROBABiUSTIC SAFETY ANALYSIS PROCE. DURES GUIDE. Sections 17 And Appendices. SEELEY,F.G. NUREG/CR-3026: FEASIBILITY STUDY ON THE ACQUISTION OF Ll. NUREG/CR-3851 V04. EVALUATION OF RADIONUCLIDE GEOCHEMI-CENSEE EVENT DATA. CAL INFORMATION DEVELOPED BY DOE HIGH LEVEL NUCLEAR RUREG/C43837: MULTIPLE SEQUENTIAL FAILURE WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For MODELEvaluation Of And Procedures For Human Error Dependency. October 1983-September 1984. NUREG/CR-4231: EVALUATION OF AVAILABLE DATA FOR PROBABI- NUREG/CR-4236 V01: PROGRESS IN EVALUATION OF RADIONU.
~~, LISTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR CLIOE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-POWER PLANTS.~~
LEVEL NUCLEAR WASTE REPOslTORY SITE PROJECTS REPORT a SANDERS,R.D. FOR OCTOBER-DECEMBER 1984. NUREG/C44150: EPIC 04tl RESIN DEGRADATION RESULTS FROM NUREG/CR-a236 V02: PROGRESS IN EVALUATON OF RADIONU-j FIRST RES4N SAMPLES OF PF 8 AND PF-20. CLlDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS. Report for
~~, SA WYER.W.D. January-March 1985.~~
l RUREG/CR 4174 ROCK MASS SEALING EXPERIMENTAL ASSESS- SEGEG l MENT OF BOREHOLE PLUG PERFORMANCE Annual Report. June i 1983 May 1984 NUREG-0933 502: A PRIORITIZATION OF GENERIC SAFETY ISSUES. NUREG 09M S03. A PRIORifl2ATION OF GENERIC SAFETY ISSUES. l SCHELLING.F.J. NUREG 1128: TRIAL EVALUATIONS IN COMPARISON WITH THE 1983 NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0 A Computer SAFETY GOALS. Code for Severe Reactor Accxtent Containment Analysis.
~~. SEGOL G.~~
SCHMIDT,C.T. NUREG/CR-3901: DOCUMENTATION AND USER'S GUIDE GS2 & GS3 NUREG/CR-4355 V01: 238 PU(IV) IN MONKEYS Overview Of Metabo- . VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS. SEITZ,M.G. SCHMULT,8. NUREG/CR-3710: LABORATORY STUDIES OF A BREACHED NUCLE-NUREG/C44365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR. AR WASTE REPOSITORY IN BASALT. POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU-SE CLEAR REACTOR VESSELS AND PIPING COMPONENTS. NUR G C44022: PRESSURIZED THERMAL SHOCK EVALUATION OF SCHNITZLER,8.G. THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. NUREG/C44203: A CALCULATIONAL METHOO FOR DETERMINING L GL DOSE RATES FROM ARRADIATED RESEARCH REAC- SE N R G/C 4085: USERS MANUAL FOR CONTAIN 10 A Cornputer Code for Severe Reactor Accident Containment Analyses. SCHRECK.R.L NUREG/C43317: TECHNICAL BASES AND USER'S MANUAL FOR SERKlZ.A.W. THE PROTOTYPE OF SPARC A SUPPRESSION POOL AEROSOL NUREG-0869 Rot: USl A-43 REGULATORY ANALYSIS. REMOVAL CODE. NUREG 0897 A01: CONTAINMENT EMERGENCY SUMP MUREG/C44130: ICEDF:A CODE FOR AEROSOL PARTICLE CAP. PERFORMANCE (Techrucal Find.ngs Related To Unresolved Safety TURE IN ICE COMPARTMENTS. Issues). . SCHWARTZ.M.W . SERNE R.J. NUREG/CR4035 A HIGHWAY ACCIDENT INVOLVING RADIOPHA4 NUREG/CR 3709: METHODS OF MINIM 121NG GROWD-WATER CON-MACEUTICALS NEAR BROOKHAVEN. MISSISSIPPI ON DECEMBER TAMINATlON FROM IN SITU LEACH URANfUM MINING Final Report 3,1983. NUREG/CR 3906- URANIUM MILL TAILINGS NEUTRALIZATIONCONTAMINANT COMPLEXATION AND TAIUNGS SCHWARTZ.R.8. LEACHING STUDY. *' NUREG/C44266: STANDARD BETA-PARTICLE AND MONOENERGE- NUREG/CR4259 TAILINGS NEUTRAUZATION AND OTHER ALTER. TIC ELECTRON SOURCES FOR THE CAUBRATION OF BETA-RADI- NATIVES FOR IMMOBlUZING TOXIC MATERIALS IN TAluNGS Final ATION PROTECTION INSTRUMENTATION. Report
, . , . - - . - - - e _-.- .-- . - _ . . + - , , , ., - . . - - - - , , , - , . - . _ , _ - . , . , , . . . - - . , .w_-m--=--------r
138 Personal Author Index SERPAN C.Z. SHERWOOD.D.R.- feUREG-1tSS V02: RESEARCH PROGRAM PLAN. Steam Generators. NUREG/CR-4259: TAILINGS NEUTRAUZATION AND OTHER ALTER- ! NATIVES FOR IMMOBluZlNG TOXIC MATERIALS iN TAILINGS Final SERVER,W.L. Report. NUREG/CR4212:IN-PLACE THERMAL ANNEAUNG OF NUCLEAR RE-ACTOR PRESSURE VESSELS, SHIEH.LC. NUREG/OR4331: WPLIFIED SEISMIC PROBABluSTIC RISK SetA W.T. ASSESSMENT.Proredures And Lmtatens. 1%REGiCR-3969: TIME. AND VOLUME WEAAGEle CONSEAVATiOh ! EQUATIONS FOR MULTIPHASE FLOW Part One System Without in- SHINOZUKA.M. temal Sohd Structures. NUREG/CR-3876: PROBABluTY BASED LOAD COMBINATION CRITE- l RIA FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES. 1 SHACK W.J. NUREG/CR4334: AN APPROACH TO THE QUANTIFICATION OF SEIS-NUREG/CR-3998 V02: UGHT WATER REACTOR SAFETY MATERIALS MIC MARGINS IN NUCLEAR POWER PLANTS. ENGINEERING RESEARCH PROGRAMS Quarte ty Progress , Report.Apni-June 1984. - SHIRE.P.R. NUREG/CR-3998 V03: UGHT-WATER REACTOR SAFETY MATERIALS NUREG/CR-4085: USERS MANUAL FOR CONTAIN 1.0.A Computer ENGINEERING RESEARCH PROGRAMS Quarterfy Progress Code for Severe Reactor Accident Containment Analysis. Report.Octot,eoDecember 1984. NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN SHiU K. UGHT WATER REACTORS. Annual Report, October 1983 - September NUREG/CR-4050- A REVIEW OF THE SHOREHAM NUCLEAR POWER 1984. STATION PROBABIUSTIC RISK ASSESSMENT.traemal Events And SHACKELFORD,M. ***" NUREG/CR 3855: CHARACTERIZATION OF NUCLEAR REACTOR SHgy,g g, CONTAINMENT PENETRATION FINAL REPORT- NUREG/CR-2815 V02 R1: PROBABluSTIC SAFETY ANALYSIS PROCE-DURES GUIDE. Sections 8-12. NURE /CR-4144: IMPORTANCE RANKING BASED ON AGING CON-SiOERATIONS OF COMPONENTS INCLUDED IN PROBABluSTIC SHOR,R.W. RISK ASSESSMENTS- NUREG/CR-3738. ENVIRONMENTAL EFFECTS OF THE URANIUM SHAFER,J.M. RUREG/CR4251 V01: MITIGATIVE TECHNIQUES FOR GROUND- SHogy,3,A, WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR NUREG/CH-3805 V02: ENGINEERING CHARACTERIZATION OF ACCIDENTS Volume 1. Analysis Of Genenc Site Conditions GROUND MOTION. Task 11: Effects Of Ground Motion Charactenstics NUREG/CR4251 V02: MITIGATIVE TECHNIOUES FOR GROUND- On Structural Response Consadonng Locabred Structural Nonhneanties WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR And Sod-Structure interar ton Effects. ACCIDENTS. Volume 2 Case Study Analysis Of Hydrologee Character. Iration And Mitigative Schemes. SHORTENCARIER NUREG/CR-4122: A FORTRAN 77 PROGRAM AND USER'S GUIDE OR H ^ ^ ^" NU G CR-3948. EXPERIMENTAL RESULTS OF THE OPERATIONAL gD ED R FF l S' TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST
~~. FACILITY. SHTEYNGART,5-NUREG/CR4291: CONCLUSION AND~~
~~SUMMARY~~
REPORT ON PHYSI-SHAPIRO B.J CAL BENCHMARKING OF PIPING SYSTEMS. NUREG/Cd3301: CATALOG OF PRA DOMINANT ACCIDENT SE. QUENCE INFORMATION. SHUMWAY,R W. NUREG/CR-3633 V01 S1: TRAC BD1/ MOD 1.AN ADVANCED BEST ES. SHARMAS. TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR NUREG/CR4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED TRANSIENT ANALYSIS. AND PRESTRESSED CONCRETE CONTAINMENT NUREG/CR-3633 V04: TRAC-BD1/ MOD 1 AN ADVANCED BEST ESil-RUREG/CR-4221: AN EVALUATION OF STRESS CORROSION CRACK MATE COMPUTER PROGRAM FOR BOlUNG WATER REACTOR GROWTH IN BWA PIPING SYSTEMS ~ TRANSIENT ANALYSIS. Volume 4 Developmental Assessment. SHAUG,J.C. SIEGEL.A.L
~~. NUREG/CR-4127 V02. BWR FULL INTEGRAL SIMULATION TEST NUREG/CR 3626 V02: MAINTENANCE PERSONNEL PERFORMANCE 4~~
(FIST) PROGRAM TRAC BWR MODEL DEVELOPMENT. Volume 2 Models. SIMULATION (MAPPS) MODEL: DESCR!PTION OF MODEL CONTENT, STRUCTURE.AND SENSITIVITY TFATING. NUREG/CR-4127 V03 BWR FULL INTEGRAL SIMULATION TEST NUREG/CR4104. MAINTENANCE PERMrdt PERFORMANCE SIM-l- (FIST) PROGRAM TRAC BWR MODEL DEVELOPMENT. Volume 3 De- ULATION (MAPPS) MODEL Field Evar,ation/Vahdaten. velopment Assessment For Plant App 4 cation. l, SHEA.C.E. . SIEGEL.M.D. NUREG/CR4069. ANALYSES OF SOILS FROM AN AREA ADJACENT NUREG/CR-4110: REPOSITORY SITE DATA REPORT FOR UNSATU-RATED TUFF, YUCCA MOUNTAIN. NEVADA. TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT SHEFFIELD.tLUNOIS. SILBERPERG,M. NUREG/CR 4083, ANALYSES OF SOILS FROM THE LOW LEVEL RA. DICACTIVE WASTE DISPOSAL SITES AT BARNWELLSC ANU NUAEM856 DAFT FC. REASSESSMENT OF THE TECHNICAL BASES RICHLAND,WA. FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment). RUREG/CR4200 BIODEGRADATION TESTING OF SOUOlFIED LOW. SILVER E.0-LEVEL WASTE STREAMS. NUREG/CR-3430 V02: NUCLEAR POWER PLANT OPERATING EXPE-SHENG,YI, RIENCE - 1982 Annualfteport. I i NUREG/CH-4158. A COMPILATION OF INFORMATION ON UNCER-TAINTIES INVOLVED IN DEPOSITION MODEUNG. SIMMONS M.A. NUREG/CR-4268 RATIO METHODS FOR COST. EFFECTIVE FIELD f SHEPHERD,J.E- SAMPLING OF COMMERCIAL RADIOACTIVE LOW 4EVEL WAS1ES l NUREG/CR-3638. HYDROGEN-STEAM JET FLAME FACIUTY AND EL PERIMENTS. SIMONEN.E.P. NUREG/CR-4136: SMOKE A Data Reducten Package For Analysts Of NUREG/CR-4267: VESSEL INTEGRITY SIMULA00N (VISA) CODE Combustion Espenments. SENSITIVITY STUDY. l I l L ..
i Personal Author index 139 SIMONEN.F.A. WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For NUREG/CR-4267: VESSEL INTEGRITY SIMULATION (VISA) CODE October 1983 September 19C4. SENSITIVITY STUDY. NUREG/CR-4134: REPOSITORY ENVIRONMENTAL PARAMETERS SIMPKINS.B. RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL WASTE PACKAGES. NUREG/CR4398: COST ANALYSIS OF REVISIONS TO 10 CFR PART
~~50. APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY SMITH.J.H.~~
CONTAINMENTS OF LIGHT-WATER-COOLED N'JCLEAR POWER NUREG/CR-3949 V01: EDDY-CURRENT INSPECTION FOR STEAM PLANTS. GENERATOR TUBING PROGRAM. Semiarrsal Progre?a Report For SIMPSONJ.S. Penod Ending June 30.1984. NUREG/CR-3736: FIELD AND THEORETICAL INVESTIGATIONS OF NUREG/CR-3949 V02: EDDY CURRENT INSPECTION FOR STEAM GENERATOR TUBING PROGRAM. Annual Progress Report For Pered FRACTURED CRYSTALUNE h0CK NEAR ORACLE.ARI2ONA. Ending December 31.1984 SIMPSONAE SMITH,K L NUREG/CR-4245: T RCE AM MEASUREMENTS AT NUREG/CR-3862: DEVELOPMENT OF TRANSIENT INITIATING EVENT FREQUENCIES FOR USE IN PROBABluSTIC RISK ASSESSMENTS. SIMPSON.H.J.
~~] NUREG/CR 4094: FIELD EXPERIMENT DETERMINATIONS OF DISTRI- SMITH.P.D.~~
d BUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE NUREG/CR4331: SIMPLIFIED SEISMIC PROBABluSTIC RISK l LAKE ENVIRONMENTS. ASSESSMENT. Procedures And Umstator+s. 1 NUREG/CR 4237: MOBILITY OF RADIONUCUDES IN HIGH CHLORIDE NURE [CR4225.
~~SUMMARY~~
OF EFFICIENCY TESTING OF STAND-SINGER,G.L ARD AND HIGH-CAPACITY HIGH-EFFICIENCY PARTICULATE AIR NUREG/CR-3633 V01 S1: TRAC BD1/ MOD 1:AN ADVANCED BEST ES- FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA. TIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TION AND EXPLOSIVE SHOCK WAVES. TRANSIENT ANALYSIS. NUREG/CR-4232; THE RESPONSE OF VENTILATION DAMPERS TO LARGE AIRFLOW PULSES SeNGH.N.P. NUREG/CR4382: CONCENTRATIONS OF URAN!UM AND THORIUM SMITH.R.C. ISOTOPES IN URANfuM MILLERS
~~AND MINERS' TISSUES. NUREG/CR4022; PRESSURIZED THERMAL SHOCK EVALUATION OF THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT.~~
~~SlSKINO.B.~~
~~! NUREG/CR-4109 TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR.~~
NUREG/CR-3879 AN EVALUATION OF THE STABILITY TESTS REC. 12ED THERMAL SHOCK TRANSIENTS AT CALVERT CLIFFS / UNIT OMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE 1.A Combusuon Engineenng PWR FORMS AND CONTAINER MATERIALS NUREG/CR-4062 EXTENDED STORAGE OF LOW-LEVEL RADIOAC- SCARES.C.G. T1VE WASTES Potenha1 Problem Areas. NUREG/CR-4406. AN ANALYSIS OF LOW 4EVEL WASTES.Rewew of NUREG/CR-4266: STANDARD BETA. PARTICLE AND MONOENERGE.
; Hazardous Waste Regulabons And identihcahonpf Radioactwo Mixed TIC ELECTRON SOURCES FOR THE CAUBRATON OF BETA-RADI-Wastes. Final Report- ATION PROTECTION INSTRUMENTATION.
l NU Ed -3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK NUREG/CR4125 V01: GUIDELINES AND WORKBOOK FOR ASSESS-RIPRAP DESIGN FOR URANIUM TAlUNGS IMPOUNDMENTS MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES NUREG/CR-4251 V01: MITIGATIVE TECHNIQUES FOR GROUND. SEEKING OPERATING LICENSF FOR A NUCLEAR POWER WATER CONTAMINATON ASSOCIATED WITH SEVERE NUCLEAR PLANT. Volume 1.Guidehne. r:or Utihty Orgamzahon And Adrmmstraton 4 NVI E -42 2 MT TV EH IOU GROUND. NU EG/CR4125 V02: GUIDEUNES AND WORKBOOK FOR ASSESS.
; WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR MENT OF ORGANIZATION AM ADM:NISTRATION OF UTluTIES ACCIDENTS. Volume 2: Case Study Analysis Of Hydrologc Character. SEEKING OPERATING LICENSE FOR A NUCLEAR POWER traton And Mibgatwo ScQ. PLANT. Volume 2 Workbook For Assessment Of Organizanon And Man.
agement.

SKORPIK.J.R. NUREG/CR4281; AN EMPtRICAL ANALYSIS OF SELECTED NUCLEAR NUREG/CR-3915

ACOUSTIC EMISSION RESULTS OBTAINED FROM POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY.
TESTING THE 281 INTERMEDIATE SCALE PRESSURE VESSEL SOO.P. l SLOVIK.G.C. NUREG/CR-2482 V06: REVIEW OF DOE WASTE PACKAGE NUREG/CR4252; lNDEPENDENT ASSESSMENT OF TRAC PD2/ MODI PROGRAM Subtask 1.1 Natonal Waste Package Program October CODE WITH BCL ECC BYPASS TESTS. 1983 - March 1984 NUREG/CR4292: A COMPARATIVE ANALYSIS OF CONSTITUTIVE RE* NUREG/CR-24f 2 V07: REVIEW OF DOE WASTE- PACKAGE LATIONS IN TRAC-PFL AND RELAPS/ MOD 1. PROGRAM Subtask 1.1. Natonal Waste Package Program Aprd 1984 SMALLEY,J.F. September 1984 NUREG/CR-3865: EVALUATION OF THE RADCACTIVE INVENTORY NUREG/CR 2482 V08: REVIEW OF DOE WASTE PACKAGE PROGRAM.Sermannual Report Covenng The Penod October 1984 IN.AND ESTIMATION OF ISOTOPIC RELEASE FROM.THE WASTE IN March 1985 ElGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL NUREG/CR-2482 V09- REVIEW OF DOE WASTE PACKAGE-SITE. PROGRAM Semiannual Report Covenng The Pered Aprd-September 1985 i EG CR4108: DEVELOPMENT OF MC&A ALARM RESOLUTION NUREG/CR-3091 V04: REVIEW OF WASTE PACKAGE VERIFICATION PROCEDURES. E r nnui e d nn e t ber W83. March fg8 SMITH.C.R. NUREG/CR-3091 V05 REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR-4117: FAULTING AND JOINTING IN AND NEAR SURFACE TESTS Sermannual Report Covenng The Penod Apnl 1984 . Septem-MINES OF SOUTHWESTERN INDIANA. b*' 94 SMtTH.F.J. R M M RW & MM PWM smW TESTS Sermannual Report Covenng The Pered October 1984. March NUREG/UR 3851 V03: PROGRESS IN EVALUATION OF RAtilONU. 1985 CUDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE RESPOSITORY SITE PROJECTSReport SOO.S.L. For Apnt-June 1984. NUREG/CR 3989. TIME AND VOLUME-AVERAGED CONSERVATON NUREG/CR-3851 V04: EVALUATION OF RADIONUCUDE GEOCHEMI- EQUATIONS FOR MULTIPHASE FLOW Part One System Without in. CAL INFORMATON DEVELOPED BY DOE HIGH4EVEL NUCLEAR ternal Sohd Structures. e.e.. . - . ._r e
4 140. Personal Author index i SOONG.A.L STETZEN8ACH K. NUREG-1127 RADIATION PROTECTION TRAINING AT URANIUM HEX-
. NUREG/CR4194: LOW-LEVEL NUCLEAR WASTE SHALLOW LAND AFLUORIDE AND FUEL FABRICATION PLANTS. BURIAL THENCH ISOLATION Final Report. October 1981 - September U8#'
SPALETTO,M.L NUREG/CR4208 GASTROINTESTINAL ABSORPTION OF PLUTONIUM STEVENS,DL. IN M RA , DOGS Apphcation To Estabksling Values Of it NUREG/CR-3659. A MATHEMATICAL MODEL FOR ASSES $1NG THE UNCERTAINTIES OF INSTRUMENTATION MEASUREMENTS FOR SPANNER,G.E. POWER AND FLOW OF PWR REACTORS. NUREG/CR-4023: FIELD PERFORMANCE ASSESSMENT OF SYN-THETIC LINERS FOR URANIUM TAILINGS POND A Status Report. STILLWELL.W.G. NUREG/CR-3688 V01. GENERATING HUMAN RELIABILITY ESTi-SPETTELL.C.M- MATES USING EXPERT JUDGMENT. Volume 1 Main Report. NUREG/CR-4016 V01: APPLICATION OF SLIM MAUD:A TEST OF AN NUREG/CR-3688 V02: GENERATING HUMAN RELIABILITY ESil-INTERACTIVE COMPUTER-BASED METHOD FOR ORGANIZING MATES USING EXPERT JUDGMENT. Volume 2 Appendices. EXPERT ASSESSMENT OF HUMAN PERFORMANCE AND RELIABILITY. Volume I. Main Re STITT 8.D NUFkEG/CR-3977; RELAP5 THERM 4l. HYDRAULIC ANALYSES OF AB L IC S SESSMENT ESUL S O E LVE PER PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-NEL FERFORMANCE ISSUES THAT COULD AFFECT SAFETY. SON UNIT 2 PRESSURIZED WATER REACTOR. SPRIGGS,G.D. - NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES WUREG/CR4022: PRESSURIZED THERMAL SHOCK EVALUATION OF ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. PRESSURIZED WATER REACTOR. Main Report. NOREG/CR-4109: TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR- NUREG/CR4326 V02: EFFECTS OF CONTROL SYSTEM FAILURES IZED THERMAL SHOCK TRANS!f NTS AT CALVERT CLIFFS / UNIT ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE 1.A Combustion Engineenng PWR. PRESSURIZED WATER REACTOR.Appendeces. l,. SRINIVASAN.M.G- STOETZEL G.A NUREG/CR-4432: COMPARISON OF DYNAMIC CHARACTERISTICS NUREG/CR 160: HISTORICAL
~~SUMMARY~~
OF OCCUPATIONAL RADi-OF FUKUSHIMA NUCLEAR POWER PLANT CONTAINMENT BUILLt ATION EXPOSURE EXPERIENCE IN U.S. COMMERCIAL NUCLEAR ING DETERMINED FROM TESTS AND EARTHOUAKES. POWER PLANTS. l STAHL.D. NURESC,13900 V02: LONG-TERM PERFORMANCE OF MATERIALS STROM8 ERG.H.M. ' USED FOR HIGH-LEVEL WASTE PACKAG;NG Ouarterfy Report. July- NUREG/CR4262 VOI: EFFECTS OF CONTROL SYSTEM FAILURES September 1984. ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC a NUREG/CR-3900 V03. LONG-TERM PERFORMANCE OF MATERIALS BOLLING WATER REACTOR Masn Report. USED FOR HIGH-LEVEL WASTE PACKAGtNG Ouarterly NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES Report. October. December 1984 ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC RUREG/CR 3900 V04: LONG-TERM PERFORMANCE OF MATERIALS BOILING WATER REACTOR Appendices. USED FOR HIGH-LEVEL WA9TE PACKAGING Annual Report.Apnl NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3. LOOP WESTINGHOUSE REG'/ 3 9 V01: LONG-TERM PERFORMANCE OF MATERIALS PRESSURIZED WATER REACTOR Main Report. USED FOR HIGH-LEVEL WASTE PACKAGING First Quarterty NUREG/CR-4326 V02: EFFECTS OF CONTROL SYSTEM FAILURES Report, Year Four Aprd June 1985. ON TRANSIENTS AND ACC; DENTS AT A 3-LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR Appendices. STALKER.A.C. NUREG/CR-4397: IN-PLANT SOURCE TERM MEASUREMENTS AT STROSNIDER,J. PRAIRIE ISLAND NUCLEAR GENERATING STATION NUREG-1155 V03 RESEARCH PROGRAM PLAN Piping. NUREG/CR-4031 V03: NEUTRON SPECTRAL CHARACTER 12ATION SUBRAMANIAN.C. FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST)lR" NUREG/CR4f t9- INTEGRITY OF CONTAINMENT PENETRATIONS UNDER SEVERE ACCIDENT CONDITIONS FY84 ANNUAL REPORT. NLRE 2 UR E S l$E NA ERS FOR THE FI H HEAVY SECTION STEEL TECHNOLOGY 1RRADIATION SUSUDHI,M. NUREG/CR 1677 V02: PIPING BENCHMARK PROBLEMS. VOLUME 11 STAMATELATOS,M. DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTION RE-NUREG/CR4303. HIGH-LEVEL WASTE PRECLOSURE SYSTEMS SPONSE SPECTRUM METHOD. SAFETY ANALYSIS Phase 1, Final Report. NUREG/CR4156 OPERATING EXPERIENCE AND AGING. SEISMIC ASSESSMENT OF ELECTRIC MOTORS. START,GI. NUREG/CR-4291: CONCLUSION AND
~~SUMMARY~~
REPORT ON PHYSI-NUREG/CR-3488 V03. IDAHO FIELD EXPERIMENT 1981. Volume CAL BENCHMARKING OF PIPING SYSTEMS. 3 Compenson Of Traiectones, Concentration Patterns And MESODIF NUREG/CR-4440: A REVIEW OF EMERGENCY DIESEL GENERATOR Model Calculabons. PERFORMANCE AT NUCLEAR POWER PLANTS. SUES,R.H. NURE CR4403.
~~SUMMARY~~
OF THE WASTE MANAGEMENT PRO, NUREG/CR-4290 V02: PROBABILITY OF PfPE FAILURE IN THE REAC-GRAMS AT URANIUM RECOVERY FACILITIES AS THEY RELATE TO TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR THE 40 CFR PART 192 STANDARDS. PLANTS Volume 2 Gudlotine Break Indirectly induced By Earthquakes. I STEELE,R. SULLIVAN,T, NUREG/CR.3819 SURVEY OF AGED POWER PLANT FACILITIES NUREG/CR-4141: CONTAINMENT PURGE AND VENT VALVE TEST NUREG/CR.2482 V08, REVIEW OF DOE WASTE PACKAGE PROGRAM FINAL REPORT. PROGRAM Senwannual Report Covenng The Penod October 1984 arch 1985. STELZMAN,W.J NUREG/CR.2482 V09- REVIEW OF DOE WASTE PACKAGE RUREG/CR4092: ORNL CHARACTERIZATION OF HEAVY SECTION PROGRAM Sern, annual Report Covenng The Penod Aprd 1985-Sep. STEEL TECHNOLOGY PROGRAM PLATES 01.02.AND 03 tomber 1985. STERNER.P. SULLIVAN,W.H. 3 NUREG 1095 EVALUATION OF RESPONSES TO IE BULLETIN 82
~~- 02. Degradation Of Threaded Fasteners in F er.ctor Coolant Pressure NUREG/CR.3301: CATALOG OF PRA DOMINANT ACCIDENT SE-~~
! Boundary Of Pressurtred Water. Reactor Plants OUENCE INFORMATlON. l
I Personal Author index 141 SUNDARUM R.K. I TAYLOR D.D. NUREG/CR3193. FORCED CONVECTIVE.NONEOUILl8RfUM. POST. NUREG/CR 3633 V01 S1: TRAC 801/MODt AN ADVANCED BEST ES-CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATON TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR COMPARISON REPORT. i TRANSIENT ANALYSIS. SUTCLIFFE,C.R. TAYLOR,J. WUREG/CR 4045: UTERATURE REVIEW ON AEROSOL SAMPUNG NUREG/CR-4156: OPERATING EXPFRIENCE AND AGING-SEISMIC DEVICES FOR RESPIRATORY FIELD STUDIES ASSESSMENT OF ELECTRIC MOTORS. SUTHERLAND,W.A. TAYLOR,J.H. NUREG/CR4128: BWR FULL INTEGRAL SIMULATION TEST (FIST) NUREG/CR 3026: FEAS181UTY STUDY CN THE ACQUISTION OF U-PHASE il TEST RESULTS AND TRAC BWR MODEL QUALIFICATION ' CENSEE EVENT DATA. SUTTON.G.E NUREG/CR4152: AN INDEPENDENT SAFETY ORGANIZATION. NUREG/CR4283. STUDY OF THE EFFECTS OF ELASTIC UNLOAD- TAYLOR,J.T. INGS ON THE JI.R CURVES FROM COMPACT SPECIMENS. NUREG/CR-3455: A COMPAR: SON OF ODINE, KRYPTON.AND XENON SCAIN,R.L RETENTION EFFICIENCIES FOR VAdlOUS SILVER LOADED AD-SORPT ON MEDIA. WUREG/CR 4015: EFFECT OF STAINLESS STEEL WELD OVERLAY CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL TEAGUE,A.G. PLATES IN BENDING SERIES 1. NUREG/CR4288. FOCAL MECHANISM ANALYSES FOR VIRGINIA AND EASTERN TENNESSEE EARTHOUAKES (1978-19NL WUREG/CR4435: ORGANIC COMPLEXANT-ENHANCED MOBlUTY OF TEICHMANN,T. TOXIC ELEMENTS IN LOW-LEVEL WASTES Annual ReportJuly 1984 NUREG/CR-2815 V01 R1: PROBABlUSTIC SAFETY ANALYSIS PROCE.
- June 1985, DURES GUIDE. Sections 17 And Appendices @VYLER,K.J. NUREG/CR-2815 V02 R1: PROBA81USTIC SAFETY ANALYSIS PROCE.
(WRES GUIDE. Sections 8-12. NUREG/CR-3829: AN EVALUATION OF THE STABIOTY TESTS REC- NUREG/CR 3026. FEASl91LITY STUDY ON THE ACQUISTON OF U. OMMENDED IN THE BRANCH TECHNICAL POSITION ON WASTE CENSEE EVENT DATA. FORMS AND CONTAINER MATERIALS. NUREG/CR-3485: PRA REVIEW MANUAL SYKES,R.I. THATCHER,0. NUREG/CR-4157: A SCIENTIFIC CRITICsE OF AVAILABLE MODELS NUREG 0933 SO2 A PRIORITIZATION OF GENERIC SAFETY ISSUES. d 4 FOR REAL TIME SIMULATIONS OF DISPERSON NUREG/CR4159- COMPARISON OF THE 1981 INEL DISPERSION THEOFANOUS T.O a DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. NUREG/CR4022 PRESSURIZED THERMAL SHOCK EVALUATION OF THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. 52AWLEWIC2,5.A-THIE.J.A. WOREG/CP 0058 V01: PROCEEDINGS OF THE TWELFTH WATER RE-1 ACTOR SAFETY RESEARCH INFORMATON MEETING. NUREG/CR-425& MEASUREMENT OF RESPONSE TIME AND DETEC-DlUREG/CP 0058 V02: PROCIEDINGS OF THE TWELFTH WATER RE. TION OF DEGRADATON IN PRESSURE SENSOR / SENSING UNE ACTOR SAFETY RESEARCH INFORMATON MEETING. SYSTEMS' NUREG/CP 0058 V04: PROCEEDINGS OF THE TWELFTH WATER RE-THOMA J.O.
~~! ACTOR SAFETY RESEARCH INFGRMATON MEETING.~~
NUREG/CP 0058 VOS. PROCEEDINGS OF THE TWELFTH WATER RE' NUREG 1810: COMPARISON OF LICENSING ACTIVITIES FOR OPER. ACTOR SAFETY RESEARCH INFORMATON MEETING ATING PLANTS DESIGNED BY 8ABCOCK & WILCOX I. I WUREG/CP4058 V06: PROCEEDINGS OF THE TWELFTH WAftiR RE. THOMAS,J.M. ACTOR SAFETY RESEARCH INFORMATION MEETING NUREG/JR4268- RATIO METHODS FOR COST-EFFECTIVE FIELD TA8ATA8Al.A S. A OWEROAL RADOAWE WMEm WASTES WUREG/CR4385: EFFECTS OF CONTROL SYSTEM FAILURES IN THOMAS,V.W. 4 TRANSIENTS. ACCIDENTS, AND CORE MELT FREQUENCIES AT A WESTINGHOUSE PRESSURIZED WATER REACTOR NUREG/CR4057. RAD 60 LOGICAL ASSESSMENT OF THE TOWN OF " q EDGEMONT. NUREG/CR4386. FFFECTS OF CONTROL SYSTEM FAILURES ON TRANSlENTS. ACCIDENTS, AND CORE-MELT FREQUENCIES AT A NUREG/CR4100: EVALUATION OF INSTRUMENTAL METHODS FOR 8ABCOCK AND WILCOX PRESSURIZED WATER REACTOR. THE MEASUREMENT OF YELLOWCAKE EMISSIONS. NUREG/CR4387: EFFECTS OF CONTROL SYSTEM FAILURES ON THOMPSON.S.L TRANS1ENTS. ACCIDENTS AND CORE-MELT FREQUENCIES AT A GENERAL ELECTRIC PRESSURIZED WATER REACTOR. NUREG/CR 3802: RELAPS ASSESSMENTOUANTITATIVE KEY PA-RAMETERS AND RUN TIME STATISTICS. TALEVARKHAN,R. NUREG/CR-3820 V03. THERMAL / HYDRAULIC ANALYSIS RESEARCH
~~. PROGRAM Ouarterty Rept.ntJuly-September 1984.~~
NUREG/CR.3889. THE MODEUNG OF BWR CORE MELTDOWN ACCl-DENTS . FOR APPLICATION IN THE MELRPI MOD 2 COMPUTER THOMPSON.V.N. CODE. NUREG/CR-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT-TANG,P.K. ABluTY PRACTICES. 1 NUREG/CR-4260 TORAC USER'S MANUALA Computer Code For Ana- THOMS,K.R. lymng Tornado-induced Flow And Matenal Transport in Nuclear FacA-ties. NUREG/CR4106. PRES $URl2ED-THERMAL-SHOCK TFST OF 6-IN. THICK PRESSURE VESSELS PTSE 1.Inveshgabon Of Warm Prestress. ing And Upper-Sheff Arrest TAR 8 ELL,W.W. NUREG/CR4383. HIGH PRESSURE INJECTON OF MELT FROM A THORNGREN.LG. REACTOR PRESSURE VESSEL THE DISCHARGE PHASE. NUREG/CR-3488 V03- 10AHO FIELD EXPERIMENT 1981 Volume 3 Compenson Of Trasectones.Concentrahon Patterns And MESODIF TASAO.L Model Calculabons NUREG/CR4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL WASTE REPOSITORY. Volume 1 Basaft. THURBER,J.A. l TACILJJ, NUREG/CR 4125 V01: GUIDEUNES AND WORKBOOK FOR ASSESS-MENT OF ORGANIZATON AND ADMINISTRATON OF UTILITIES NUREG/CR 3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL ACCIDENTS METHOPS, COSTS AND SCHEDULES FOR DECON. SEEKING OPERATING UCENSE FOR A NUCLEAR POWER PLANT. Volume 1 Guideines For ut hty Organizabon Arid Admirnstrahon
~~; TAMiNATON. Plan }~~
~~I i~~
~~.~ __-~_ _ _ . - - _. _~~
142 Personal Author Index NUREG/CR-4125 V02: GUIDEUNES AND WORKBOOK FOR ASSESS- TURAEY,J.R. MENT OF ORGANIZATION AND ADMINISTRATION OF UTILITIES NUREG/CR-4258. AN APPROACH TO TEAM SKILLS TRAINING OF NU-SEEKING OPERATING LICENSE FOR A NUCLEAR POWER CLEAR POWER PLANT CONTROL ROOM CREWS. PLANT. Volume 2 Workbook For Assessment Of Organization And Man-agement. UNIONE,A. RUREG/CR-4281: AN EMPIRlCAL ANALYSIS OF SELECTED NUCLEAR NUREG/CR-2815 V0t R1: PROBABluSTIC SAFETY ANALYSIS PROCE-POWER PLANT MAINTENANCE FACTORS AND PLANT SAFETY. DURES GUIDE. Sections 17 And Appendices. I NURFG/CR-2815 V02 R1: PROBABILISTIC SAFETY ANALYSIS PROCE-NU EG/CR4110- REPOSITORY dlTE DATA REPORT FOR UNSATU-RATED TUFF, YUCCA MOUNTAIN. NEVADA. UPDEGRAFF,C.D. NUREG/CR4110: REPOSITORY SITE DATA REPORT FOR UNSATU-CA M M W N REG /CR 3738: ENVIRONMENTAL EFFECTS OF THE URANIUM FUEL CYCLE A Review Of Data For Technetium. URIBE.R. TKACHYK,J.W. NUREG/CR 3981: BIOACCUMULATION OF P 32 IN BLUEGILL AND NUREG/CR-3455. A COMPARISON OF ODINE, KRYPTON.AND XENON CATFISH. I RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD. YADEN J
~~SORPTON MEDIA. N f~~
TCALSTON A.L RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE NUREG-0970: PROCEDURES FOR MEETING NRC ANTITRUST RE- REGULATORY PROGRAM. i l j SPONSIBlUTIES. ! TOBIAS.M.L NUREG 1155 V01: RESEARCH PROGRAM PLAN Reactor Vessels. NUREG/CR 3830 V02: AEROSOL RELEASE AND TRANSPORT NUREG 1155 V03 RESEARCH PROGRAM PLAN Piping l PROGRAM. Semiannual Progress Report For Apql 1984-September VALENTINE,0.M l 1984. NUREG/CR4255 V01: AEROSAL RELEASE AND TRANSPORT PRO. 'NUREG/CR4178 CAFT: AN EVALUATION OF SELECTED UCENSEE GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 - EVENT REPORTS PREPARED PURSUANT TO 10 CFR 50 73 Draft MARCH 1985. Report. TORRONEN.K- VALENZUELA.J.A. NUREG/CR4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW NUREGICR-3426 V01: THERMAL AND FLUID MIXING IN 1/2-SCALE RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRON
~~TEST FACluTY. Facility And Test Design Report. .~~
MENTS. NUREG/CR.3426 V02: THERMAL AND FLUID MIXING IN 1/2-SCALE j * ~ TOSTE A.P. l NUREG/CR4030. RADIONUCUDE MIGRATION IN GROUND VANDEGRIFT,G.F. I WATER (Final Report) NUREG/CR-3710 LABORATORY STUDIES OF A BREACHED NUCLE-AR WASTE REPOSITORY IN BASALT. TSTH LM, I NUREG/CR 3514 V02: THE CHEMICAL BEHAVIOR OF IODINE IN VANDER MOLEN H ! AQUEOUS SOLUTIONS UP TO 150 C ti. Radiation-Redox Conditions. NUREG-0933 S02: A PRIORITIZATION OF GENERIC SAFETY ISSUES. TCAVIS,J.R. VANDERMOLEN H. EdUREG/CR-3930: OBSERVED BEHAVIOR OF CESIUM. LODINE.AND NUREG-0933 S03. A PRORITl2ATON OF GENERIC SAFETY ISSUES. I TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO. GRAM VARMA A.K. NUREG/CR4020 HMSA COMPUTER PROGRAM FOR TRANSIENT,THREE-DIMENSIONAL MixlNG GASES. NUREG/CR4 t S8: A COMPILATION OF INFORMATION ON UNCER. NUREG/CR4037. DATA
~~SUMMARY~~
~~REPORT FOR FISSON PRODUCT TAINTIES INVOLVED IN DEPOSITION MODELING RELEASE TEST Hi-5. VASLOW,F. G6UREG/CR-4043: DATA~~
~~SUMMARY~~
~~REPORT FOR FiSSON PRODUCT RELEASE TEST HI-6. NUREG/CR4215. TECHNICAL FACTORS AFFECTING LOW-LEVEL WASTE FORM ACCEPTANCE CRITERIA.~~
~~TREBILCOCK,W.~~
NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0 A Cornputer V ASS:LAROS M.G. Code for Severe Reactor Accident Containment Andysis. NUREG/CR4283 STUDY OF THE EFFECTS OF ELAGTIC UNLOAD-INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS. j TRIER.R.d.
~~GdUREG/CR4094 FIELD EXPERIMENT DETERMINATIONS OF DISTRI- VEAKIS.E.~~
l BUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE NUREG/CA-2482 V08- REVIEW OF DOE WASTE PACKAGE LAKE ENVIRONMENTS PROGRAM Sermannual Report Covereg The Penod October 1984 - NUREG/CR4237: MOBILITY OF RADONUCUDES IN HIGH CHLORIDE ya,ch 1985. ENVIRONMENTS. NUREG/CR 3091 V04. REVIEW OF WASTE PACKAGE VERIFICATON fgg4
**'*"""*' "9 C ~ *"
i TCIGGS T.J. NUREG/CR.3987: COMPUTERIZED ANNUNCIATOR SYSTEMS. NUREG/CR-3091 V05- REVIEW OF WASTE PACKAGE VERIFICATION l TESTS Semsannual Report Covereg The Penod Apnl 1984 Septern-TSANG F.Y. ber 1964 NUREG/CR-3237; CONTROL OF EXPLOSIVE MtXTURES IN PWR WASTE GAS SYSTEMS- VESELY,W. NUREG/CR-2815 V01 R1: PROBABIUSTIC SAFETY ANALYSIS PROCE. TURGEON.K.S. DURES GUIDE.Sectiorm 17 And Appendices. l NUREG/CR-398 t: BIOACCUMULATION OF P-32 IN BLUEGILL AND 8"' VESELY,W.E. NUHEG/CH4377; EVALUATIONS AND UTIL12ATIONS OF RISK IM. TURNER J.H. NUREG/CR-3922 V0t: SURVEY AND EVALUATION OF SYSTEM PORTANCES INTERACTION EVENTS AND SOURCES Ma n Report And Appendices VICKROY,S.C. A And B. NUREG/CR-3922 V02: SURVEY AND EVALUATION OF SYSTEM NUREG/CR4296. DESIGN AND INSTALLATION OF COMPUTER SYS-INTERACTON EVENTS AND SOURCES Appendices C And D. TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55 i l l l
l Personal Author index 143 VIERZ8A.E. WEBER,C.F. NUREG/CR-4357: THE FEASIBILITY OF DETECTING THE IMPORT OF NUREG/CR-3885 V03: HIGH-TEMPERATURE GASCOOLED REACTOR UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED SAFETY STUDIES FOR THE DIVISlN OF ACCIDENT STATES-EVALUATON Ouarterly Progress Report, July 1 Sept 1mber 30.1984. NUREG/CR-3885 V04: HIGH-TEMPERATURE GAS COOLED REACTOR NU / 4231: EVALUATON OF AVAILABLE DATA FOR PROBABI- SAFETY STUDIES FOR THE DIVISION OF ACCIDENT LISTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR EVALUATON. Quarterly Progress Report, October 1-December S. NU E C'R4402 V01: HIGH-TEMPERATURE GAS-COOLED REACTOR VORA.J.P. SAFETY STUDIES FOR THE DIVISON OF ACCOENT NUREG 1144; NUCLEAR PLANT AGING RESEARCH (NPAR) PRO. EVALUATON Ouarterly Progress Report, January 1 March 31,1385. GRAM PLAN. WEBSTER C.S. VOSKA.KA NUREG/CR-3519 HUMAN ERROR PROBABILITY ESTIMATION USING NUREG/CR-3930: OBSERVED BEHAVIOR OF CESIUMlODINE.AND UCENSEE EVENT REPORTS- TELLURIUM IN THE ORNL FtSSON PRODUCT RELEASE PRO-GRAM. CAH1,K K NUREG/CR4037: DATA
~~SUMMARY~~
~~REPORT FOR FISSION PRODUCT~~
^
~~RA ED UFF Y NTAl ,NEVAD NURE CR4 3 T'A~~
~~SUMMARY~~
REPORT FOR FISSION PRODUCT RELEASE TEST HI-8. WALLER,R.A. NUREG/CR4314: BRIEF SURVEY AND COMPARISON OF COMMON WEILANDICS,C. CAUSE FAILURE ANALYSIS. NUREG/CR-3469 V02: OCCUPATONAL DOSE REDUCTION AT NU. CLEAR POWER PLANTS Annotated Bibhography Of Selected Read-URE'G/CR4357: THE FEASl81UTY OF DETECTING THE IMPORT OF UNAUTHORIZED RADIOACTIVE MATERIALS INTO THE UNITED WEINSTOCK E.V. STATES-NUREG/CR4152 AN INDEPENDENT SAFETY ORGANIZATON WALSH.M.E. NUREG/CR-4139. THE MAILED SURVEY.A TECHNIQUE FOR 00TAIN- WEISS.A.J. ING FEEDBACK FROM OPERATIONS PERSONNEL NUREG/CP-0059 V01: PROCEEDINGS OF THE MITi.-NRC SEISMIC IN. FORMATION EXCHANGE MEETING. VOLUME L CALTERS,W.H. NUREG/CP-0071: TRANSACTIONS OF THE THIRTEENTH WATER RE. NUREG/CR-3752: EFFECTS OF HYDROLOGIC VARIABLES ON ROCK ACTOR SAFETY RESEARCH INFORMATION MEETING. RIPRAP DESIGN FOR URANIUM TAILINGS IMPOUNDMENTS. NUREG/CR4076: DETERMINATION OF COMPLIANCE WITH CRITERIA NUREG/CR 2331 V04 N2: SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY FOR FINAL TAILINGS DISPOSAL SITE RECLAMATON. RESEARCH Ouarterly Progress Report.Apnl 1-June 30,1984. WANG,Y.K. NUREG/CR-2331 V04 N3: SAFETY RESEARCH PROGRAMS SPON-NUREG/CR-4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED SORED BY OFFICE OF NUCLEAR REGULATORY AND PRESTRESSED CONCRETE CONTAINMENT RESEARCH Ouarterty Progress Report. July 1.Septerr:ber 30,1984. NUREG/CR 4291: CONCLUSON AND
~~SUMMARY~~
REPORT ON PHYSt. NUREG/CR-2331 V04 N4: SAFETY HESEARCH PROGRAMS SPON-CAL DENCHMARKING OF PIPING SYSTEMS SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH.Ouarterty Prr gress Report. October 1. December 3t, WARD.R.C. 1984. NUREG/CFau38: SENSITMTY AND UNCERTAINTY STUDIES OF THE CRAC2 COMPUTER CODE- NUREG/CR2331 V05 N1: SAFETY RESEARCH PROGRAMS SPON. SORED BY OFFICE OF NUCLEAR REGULATORY WARRlNER,J.S. RESEARCH Ouarterty Progress Report. January 1-March 31,1985. NUREG/CR-3774 V02: ALTERNATIVE METHODS FOR DISPOSAL OF LOW LEVEL RADIOACTIVE WASTES. Task 2A.Techrucal Requaements For Bel NUREG/CR 3145 V03. GEOPHYSICAL INVESTIGATIONS OF THE NUREG/C ound 774 Vault V03. ALT Dshsal Of Low Level NAflVE METHODS Radosctive FOR DISPOSAL OFWaste. WESTERN OHIOLIND8ANA REGION . ANNUAL REPORT.(October LOW LEVEL RADIOACTIVE WASTES. Task 28.Techrucal Requirements 1982. September 1983, Volume 3). For Aboveground Vault Dsposal Of Low Level Radioactrve Waste. CASCOM,RL NUREG/CR 4331: SIMPUFIED SEISMIC PROBABILISTIC RISK NUREG/CR-3516. A SURVEY OF THE USES OF RADIOACTIVE MATE. ASSESSMENT Procedures And Lmtatens. RIALS IN LOUISlANA S OFFSHORE WATERS. WENSEL.R.G. NURE / R4120: MATHEMATICAL MODEUNG OF ULTIMATE HEAT NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV-SINK COOLING PONDS. OR DURING STATION BLACKOUT. WATERMAN.M.E. WERES,0. NUREG/CR3977. RELAP5 THERMAL.HYDRAUUC ANALYSES OF NUREG/CRat61 V01: CRITICAL PARAMETERS FOR A HIGH,.LVEL PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B ROBIN. WASTE REPOSITORY. Volume t Dasalt. SON UNIT 2 PRESSURllED WATER RE ACTOR NUREG/CR4195: OVERVIEW OF TRAC-PD2 ASSESSMENT CALCULA. WESCOTT,R.G. TONS NUREG 1165 ESAP 7.1.1 " ENVIRONMENTAL IMPACTS OF POSTU. NUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES LATED ACCOENTS INVOLVING RELEASES OF RADIOAC11VE MA-ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE TERIALS TO GROUNDWATER." PRESSURIZED WATER REACTOR Maen Re. port. NUREG/CR 4326 V02: EFFECTS OF CONTROL SYSTEM FAILURES WE SLEY,D.A. o ON TRANSIENTS AND ACCIDENTS AT A 3 LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR Apperdces NUREG/CR3558 HANDBOOK OF NUCLEAR POWER PLANT SEISMIC FRAGILITIES. Seistmc Safety Margms Research Program. WATKINS,J.C. WESTER,M.J. NURFG/CR414 t: CONTAINMENT PURGE AND VENT VALVE TFST PROGRAM FINAL REPORT. NUREG/CR3913. HECTR VERSION 10 USER S MANUAL WEAKLEY,S.A. WETTMILLER.R.J. NUREG/CR2800 S03 GUOEUNES FOR NUCLEAR POWER PLANT CANADIAN SEISMIC AGREEMENT Techrwcal SAFETY ISSUE PRORITl2ATON INFORMATlON DEVELOPMENT. NUREG/CR4317 Report Covenng 1979 V01'.t965
~
144 Personal Author index WHATLEY,S K. EVALUATION Ouarterty Progress Report, October 1-December NUF.2G/CP.0062: PROCEEDINGS OF THE CONFERENCE ON THE AP. 31,1984. PUCATION OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR NUREG/CR-4402 V01: HIGH. TEMPERATURE GAS-COOLED REACTOR WASTE REPOSITORY ASSESSMENT. SAFETY STUDIES FOR THE DIVISION OF ACCIDENT NOREG/CR-3851 V04: EVALUATION OF RADIONUCLIDE GEOCHEMI- EVALUATION Ouarterly Progress Report, January 1 - March 31,1985. CAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Annual Progress Report For WILSON,M.D. October 1983-September 1984 NUREG/CR-398 t BIOACCUMULATION OF P-32 IN DLUEGILL AND NUREG/CR-4236 V01: PROGRESS IN EVALUATION OF RADONU. CATFISH. CLOE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS REPORT WINEGARDNER,W. FOR OCTOBER-DECEMBER 1984 NUREG/CR-4130; ICEDF.A CODE FOR AEROSOL PARTICLE CAP. NUREG/CR4236 V02: PROGRESS IN EVALUATION OF RADIONU- TURE IN ICE COMPARTMENTS. CLIDE GEOCHEMICAL INFORMATION DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report for WITHEE,CJ. i January March 1985. NUREG 1065 R01: ACCEPTANCE CRITERIA FOR THE LOW EN-RICHED URANfuM REFORM AMENDMENTS. WHEELER,W.A. NUREG/CR-3817: DEVELOPMENT,USE AND CONTROL OF MAINTE- WOLF,J.J. NANCE PROCEDURES IN NUCLEAR POWER PLANTS Problems And NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PERFORMANCE Recommendates. SIMULATION (MAPPS) MODEL: DESCRIPTION OF MODEL CONTENT, STRUCTURE.AND SENSITIVITY TESTING WHITE,A S. NUREG/CR-3634 MAINTENANCE PERSONNEL PERFORMANCE SIM-NUREG/CR4051: ASSESSMENT OF JOB-RELATED EDUCATIONAL ULATON (MAPPS) MODEL: USER'S MANUAL OUAUFICATIONS FOR NUCLEAR POWER PLANT OPERATORS- NUREG/CR4104: MAINTENANCE PERSONNEL PERFORMANCE SIM-ULATON (MAPPS) MODEL Field Evaluation /Vahdatson. ' WHITE.J.D. NUREG/CR4022: PRESSURf2ED THERMAL SHOCK EVALUATION OF WOLLEN8 ERG,H.A. THE CALVERT CLIFFS UNIT 1 NUCLEAR POWER PLANT. NUREG/CR4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL WASTE REPOSITORY. Volume 1 Basalt. i WHITE J.R. NUREG/CR-3646: TRAC PF1 INDEPENDENT ASSESSMENT. WOMELSDUFF,J.E. WHITMAN,G.D. NUREG/CR-3361: THE EFFECT OF WATER CHEMISTRY ON THE NUREG/CR4106: PRESSURIZEDTHERMAL SHOCK TEST OF 6-IN.. RATES OF HYDROGEN GENERATON FROM GALVANIZED STEEL THICK PRESSURE VESSELS PTSE.1:fnvesbgata Of Warm Prestress. CORROSION AT POST-LOCA CONDITONS. ing And Upper-Shelf Arrest. NUREG/CR-3803. THE EFFECTS OF POST.LOCA CONDITIONS ON A PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-WIDRIG,R.D. DUSTRY. NUREG/CR-4125 V01: GUIDEUNES AND WORKBOOK FOR ASSESS. MENT OF ORGANIZATON AND ADMINISTRATION OF UTILITIES WONG,C.N. SEEKING OPERATING UCENSE FOR A NUCLEAR POWER NUREG/CR 4044. TRAC.PF1 LOCA CALCULATONS USING FINE. PLANT. Volume 1.Guidehnes For Utility Organization And Admirustraten NODE AND COARSE NODE INPUT MODELS. Plan.
~~# NUREG/CR-4125 V02: GUOEUNES AND WORKBOOK FOR ASSESS. WONG.K.S.~~
MENT OF ORGANIZATION AND ADMINISTRATION OF UTIUTIES NUREG/CR4190 CAUFORNIA OFFSHORE SURVEY OF LICENSEES SEEKING OPERATING UCENSE FOR A NUCLEAR POWER USING RADIOACTIVE MATERIAL i PLANT. Volume 2 Workbook For Assessment Of Orgaruration And Man. WOO.H.H. agement NUREG/CR.3019 RECCMMENDED WELDED CRITERIA FOR USE IN WILKOWSKI,G.M. THE FABRICATION OF SHIPPING CONTAINERS FOR RADIOACTIVE NUREG/CR4082 V01: DEGRADED PIPING PROGRAM . PHASE MATERIALS, il Semiannual Report, March 1984 - Septemtar 1984
~~! NUREG/CR4082 V02; DEGRADED PIPING PROGRAM PHASE WOOD.R.S.~~
it.Sermannual Report. October 1984 March 1985. NUREG 113t: FINANCIAL ANALYSIS OF POTENTIAL RETROSPECTIVE PREMlUM ASSESSMENTS UNDER THE PRICE ANDERSON R.LUAMS D.C. SYSTEM' NUREG/CR4085 USERS MANUAL FOR CONTAIN 1.0 A Computer Code for Severe Reactor Accident Containment Analysss. WORRELL,R.g. NUREG/CR4213: SETS REFERENCE MANUAL CILLIAMS.L NUREG/CR 4031 V02: NEUTRON SPECTRAL CHARACTERl2ATION WREATHALL,J. FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST) lR. NUREG/CR 41F7 V01: MANAGEMENT OF SEVERE RADtATON SERIES. "Neutronics Caledations." ACCOENTS Perspectives On Managing Severe Accidents in Commer. cial Nuclear Power Plants ERE NU E R4 V01 CAL T ONAL METHODS FOR ANALYSIS gcC NS wng Plant at Proc es n o The e Accident Regime NUREG/CR4360 V02: CALCULATIONAL METHODS FOR ANALYS3S OF POSTULATED UF6 RELEASES. WRENN.M.E. WILLIFORD,R.E. NUREG/CR4382: CONCENTRATIONS OF URANIUM AND TFORIUM ISOTOPES IN URANIUM MILLEAS' AND MINERS
~~TISSUES.~~
! NUREG/CR4168. GT2F A COMPUTER CODE FOR ESTIMATING LIGHT WATER REACTOR FUEL ROD FAILURES. WILLIS C.A. NUREG/CR-4346. AEROSOL RELEASE EXPERIMENTS IN THE FUEL NUREG 0017 R0f: CALCULATION OF RELEASES OF RADOACTIVE AEROSOL SIMULANT TEST FACluTY.UNDERSODtUM EXPERI. MENTS, MATERIALS IN GASEOUS AND LloutD EFFLUENTS FROM PRES. SURIZED WATER REACTORS (PWR GALE CODE). WILSON.J.H. NUREG/CR 4101: ASSAY OF LONG UVED RADIONUCLIDES IN LOW. NUREG/CH 3886 V03: HIGH. TEMPERATURE GAS-COOLED REACTOR LEVEL WASTES FROM POWER REACTORS SAFETY STUDIES FOR THE DIVISION OF ACCIDENT WRIGHT,M.A. EVALUATION Ouarterfy Progress Report. July 1-September 30,1984 NUREG/CR 3885 V04: HIGH-TEMPERATURE GAS. COOLED REACTOR NUREG/CR 4117: FAULTING AND JOINTING IN AND NEAR SURFACE SAFETY STUDIES FOR THE DIVISION OF ACCIDENT MINES OF SOUTHWESTERN INDIANA. l l E
r Personal Author index 145 wu,0.S. YOUNG,J.A. NUMG/CR-4365: DESIGN AND DEVELOPMENT OF A SPECIAL PUR- NUREG/CR-4057; RADIOLOGICAL ASSESSMENT OF THE TOWN OF POSC SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATION OF NU" EDGEMONT, CLEAR REACTOR VESSELS AND PIPING COMPONENTS. NUREG/CR-4118: MONITORING METHODS FOR DETERMINATON wULFF,W. COMPLIANCE WITH DFCOMMISSIONING CLEANUP CRITERIA AT NUREG/CR-3943 THE BWR PLAN ANALYZER. ' URANIUM RECOVERY SITES. CYANT,F.J. YOUNG,T.E. NUREG/CR-4147; THE EFFECT OF ENVIRONMENTAL STRESS ON SYLGARD 70 SILICONE ELASTOMER. NUREG/CH 4245: IN-PLANT SOURCE TERM MEASUREMENTS AT BRUNSWICK STEAM ELECTRIC STATION. j YWNDOOSTA NUREG/CR4397: IN-PLANT SOURCE TERM MEASUREMENTS AT NUREG/CR 4174: ROCK MASS SEALING - EXPER MENTAL ASSESS- PRAIRIE ISLAND NUCLEAR GENERATING STATION. MENT OF BOREHOLE PLUG PERFORMANCE. Annual Report, June 1983 May 1984. YOU M OOD,BA VEE,C.S. NUREG/CR-2815 V02 R1: PROBABILISTIC SAFETY AN ALYSIS PROCE. DURES GUIDE.Sectens 8-12. i . NUREG/CR 3537: EXPEDIENT METHODS OF RESPIRATORY l ' PROTECTIONill. SUBMICRON PARTICLE TESTS AND
~~SUMMARY~~
~~~ YOUNGSLOOO.M.~~
OF OUALITY FACTORS-NUREG/CR-4228

REVIEW OF THE VOGTLE UNITS 1 AND 2 AUXILIA.
y , RY FEEDWATER SYSTEM RELIABILITY ANALYSIS. NUREG/CR-4350 V01: PROBABILISTIC RISK ASSESSMENT COURSE 4 DOCUMENTATION. Volume 1 PRA Fundarnentals YOUNGSLOOO,M'W* s NUREG/CR-4350 V04: PROBABILISTIC RISK ASSESSMENT COURSE NUREG/CR-2815 V01 R1: PROBABILiSTIC SAFETY ANALYSIS PROCE. j DOCUMENTATON. Volume 4 - System Reliatuisty And Analysis DURES GUIDE. Sections 17 And Appareces. Tectwnques,Sessons B/C Event Trees / Fault Trees. NUREG/CR-3485: PRA REVIEW MANUAL l 4 l-l t i i 1 } ) l f i 4 i 1 i 4 1 i
i l l l l t l l l l l l l l l l l l l
~~-s-- w----- . - - ,.y -, -.-. , .. ,. ,_ _ _. _ _ _ _ __ _~~
Subject Index i This index was developed from keywords moved later when a reasonable thesaurus and word strings in titles and abstracts. has been developed through experience. During this development period, there will Suggestions for improvements are wel-be some redundancy, which will be re- come.
. 10 CFR Part 80 ATWS l NUREG-104d: DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN NUREG/CR-3633 V01 S1; TRAC BDt/ MOD 1 AN ADVANCED BEST ES- ; UNSATURATED ZONE: TECHNICAL CONSOERATONS AND RE- TIMATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR
( SPONSE TO COMMENTS. TRANSIENT ANALYSIS. A302-8 Steet NUREG/CR 3704. 8WR-LTAS: A BOluNG WATER REACTOR LONG-TERM ACCOENT SIMULATON CODE. NUREG/CR-4437: EXPLORATORY STUDIES OF ELEMENT INTERAC-NUREG/CR4046: DETERMlNING CRITICAL FLOW VALVE CHARAC-TIONS AND COMPOSITON DEPENDENCIES IN RADIATON SENSI-TlVITY DEVELOPMENT. TERISTICS USING EXTRAPOLATON TECHNIOUES Abnormal Occurrence N RE / R-4437: EXPLORATORY STUDIES OF ELEMENT INTERAC- NUREG-0090 V07 NO3. REPORT TO CONGRESS ON ABNORMAL TIONS AND COMPOSITION DEPENDENCIES IN FIADIATION SENSI' NU 00 NO TlVITY DEVELOPMENT. E TT CONGRESS ON ABNORMAL OCCURRENCES October-December 1984.- AC Power NUREG 0090 V08 N01: REPORT TO CONGRESS ON ABNORMAL NUREG 1032 DAFT FC: EVALUATION OF STATION BLACKOUT ACCl- OCCURRENCES.Janurary-March 1985. DENTS AT NUCLEAR POWER PLANTS.Techrwcal Findings Related To NUREG-0090 Vos NO2: REPORT TO CONGRESS ON %8 NORMAL Unresolved Safety issue A44. Draft Report For Comment OCCURRENCES Aptd June 1985. ACMS Aboveground Vault WUREG-1125 V01: A COMPILATION OF REPORTS OF THE ADV.SORY NUREG/CR 3774 V03. ALTERNATIVE METHODS FOR DISPOSAL OF CCMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 1,Part LOW LEVEL RADIOACTIVE WASTES. Task 28.Techrucal Requirements 1 ACRS Reports On Protect Reviews (A-F) For Aboveground Vault Disposal Of Low Level Raeoective Waste. NUREG-It25 V02: A COMPtLATION OF REPORTS OF THE A3VISORY COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 2.Part Abetract 1 ACRS Reports On Protect Reviews (G P) NUREG4304 V09 N04. REGULATORY AND TECHNICAL NUREG 1125 V03: A COMPILATION OF REPORTS OF THE ADVISORY REPORTS Annual Compelahon For 1984. COMMITTEE ON REACTOR SAFEGUARDS 19571984 Volume 3.Part NUREG-0304 V10 NO2: REGULATORY AND TECHNICAL 1 ACRS Reports On Protect Reviews (0 Z) REPORTS Compdation For Secund Quarter 1985,Aprd June. NUREG 1125 V04: A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 4.Part Access Control 2 ACAS Reports On Genenc Subrects (Accident Analysis . Genenc NUREG/CR4298. DESIGN AND INSTALLATION OF COMPUTER SYS-Items). TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55. NUREG 1125 V05: A COMPILATIO*4 OF REPORTS OF THE ADVISORY NUREG/CR4392: MEASUPES OF SAFEGUARDS RISM EMPLOYING COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 5,Part PRA (MOSREP) A Methodology For Estamating Resh Impacts Of Safe-NUREG 1 06 LA OF E S T$E A IORY COMMITTEE ON REACTOR SAFEGUARDS.19571984 Volume 6,Part Accident 2;ACRS Reports On Genenc Sub l ects (RPA . Appendu C) NUREG-0958 DRFT FC: REASSESSMENT OF THE TECHNICAL BASES ALARA FOR ESTIMATING SOURCE TERMS. (Draft Report For Comment) NUREG/CP-0066 PROCEEDINGS OF AN INTERNATIONAL WORK- NUREG-0981 R01: NRC/ FEMA OPERATONAL RESPONSE PROCE-DURES FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTION ACCOENT, N R G/ 32 3 V01 TE HN LOGY ETY T OF DE. COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON' GR E R Draft R F t FUEL CYCLE FACluTIES FOLLOWING POSTULATED NUREG-1116: A REVIEW OF THE CURRENT UNDERSTANDING OF ACCIDENTS Main Report THE POTENTIAL FOR CONTAINMENT FAILURE FROM IN VESSEL NUPEG/CR-3293 V02: TECHNOLOGY, SAFETY AND COSTS OF DE' STEAM EXPLOSIONS NUREG/CR 3293 V01: TECHNOLOGY, SAFETY AND COSTS OF DE-COMMISSIONING REFERENCE FUEL CYCLE AND NON FUEL CYCLE FACIUTIES FOLLOWING POSTULATED COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON. FUEL CYCLE FACluTIES FOLLOWING POSTULATED NUR / 34 CCUPATONAL DOSE REDUCTION AT NU- ACCOENTS Main Report. CLEAR POWER PLANTS. Annotated Bibliography Of Selected Road. NUREG/CR-3293 V02. TECHNOLOGY, SAFETY AND COSTS OF DE-COMMISSIONING REFERENCE FUEL CYCLE AND NON-FUEL NUN R V ION N LEAR FACluTY DECOMMIS. CYCLE FACluTIES FOLLOWING POSTULATED NU E /CR 4 P ION bS d i N b ALARA NUR G 330 OG OF PRA DOMINANT ACCIDENT SE-A NUCLEAR ER NTS. Study On High-Dose Jobs,Radwaste OUE E OMT N NURE /b4373. COMPENDIUM OF COST-EFFECTIVENESS EVALUA. THE PROTOTYPE OF SPARC A SUPPRESSION POOL AEROSOL TONS OF MOOlFICATIONS FOR DOSE REDUCTON AT NUCLEAR REMOVAL CODE. POWER PLANTS. NUREG/CR-3413 OFF SITE CONSEQUENCES OF RADIOLOGICAL ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON-AS88E Code TAMINATION NUREG/CR 3019 RECOMMENDED WELDED CRITERIA FOR USE IN NUREG/CR 3485, PRA REVIEW MANUAL THE FABRICATION OF SHIPPING CONTAINERS FOR RADOACTIVE NUREG/CR 3611: RADIOACTIVE MATERIAL (RAM) ACCIDENT / INCL-MATERIALS. DENT DATA ANALYSIS PROGRAM NUREG/CR 3647: DESIGN AND FABRICATON OF A 1/8 SCALE NUREG/CR-3816 V02: REACTOR SAFETY RESEARCH Ouarterty STEEL CONTAINMENT MODEL Report.Apnt June 1964 147
148 Subject index NUREG/CR-3816 V03: REACTOR SAFETY RESE ARCH ouarterly NUREG/CR4383 HIGH PRESSURE INJECTON OF MELT FROM A ReportJuly September 1984 REACTOR PRESSURE VESSEL . THE DISCHARGE PHASE. NUREG'CR-3855. CHARACTER 12ATION OF NUCLEAR REACTOR NUREG/CH4385 EFFECTS OF CONTROL SYSTEM FA! LURES IN CONTAINMENT PENETRATION . FINAL REPORT. TRANSIENTS. ACCIDENTS. AND CORE. MELT FREQUENCIES AT A NUREG/CR-3889 THE MODELING OF BWR CORE MELTDOWN ACCl- WESTINGHOUSE PRESSURIZED WATER REACTOR DENTS . FOR APPLICATION IN THE MELRPI MOD 2 COMPUTER NUREG/CR4386 EFFECTS OF CONTROL SYSTEM FAILURES ON CODE. TRANSIENTS. ACCIDENTS AND CORE-MELT FREOUENCIES AT A RUREG/CR-3912: MARCH-HECTR ANALYSIS OF SELECTED ACCl- BADCOCK AND WILCOX PRESSURllED WATER REACTOR DENTS IN AN ICE-CONDENSER CONTAINMENT. NUREG/CR-4387. EFFECTS OF CONTROL SYSTEM FAILURES ON NUREG/CR3943 THE BWR PLAN ANALYZER TRANSIENTS. ACCIDENTS AND CORE MELT FREQUENCIES AT A NUREG/CR3954 HECTR ANALYSIS OF EQUIPMENT TEMPERATURE GENERAL ELECTRIC PRESSURIZED WATER REACTOR RESPONSES TO SELECTED HYDROGEN BURNS IN AN ICE CON- NUREG/CR 4388, AEROSOL BEHAv OR MODELING (TASK 31 SUP. R "V R E ^" ^ EVALUATION OF SEVERE NURE /CR 90 R DAL PERFORMANCE UNDER ACCIDENT DE TP O E DM O NUREG/CR4399- POSSIBLE OPTIONS FOR REDUCING OCCUPA. kl R G 4 03 H HWA A DEN NVOLVING RADIOPHAR- TlONAL DOSE FROM THE TMI-2 BASEMENT. MACEUTICALS NEAR BROOKHAVEN, MISSISSIPPI ON DECEMBER NU E CR4055. THE Oto EXPERIMENT COOLA81LITY OF 002 A**'d*"' A"*'Y NUREG-1125 V04 A COMPILATION OF REPORTS OF THE ADVISORY DEBRIS 'N SODIUM WITH DOWNWARD HEAT REMOVAL COMMITTEE ON REACTOR SAFEGUARDS,1957-1984 Volume 4.Part NUREG/CR4067;
~~SUMMARY~~
OF 8ARRIER DEGRADATION EVENTS 2 ACAS Reports On Genenc Subrects (Accident Anarysis Generic AND SMALL ACCIDENTS IN U S. COMMERCIAL NUCLEAR POWER PLANTS Itemst NUREG/CR4080 DETERMINATION OF THE AVAILABILITY OF CORE NUREG/CR-3887 HUMAN FACTORS REVIEW FOR SEVERE ACCl-EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATIONS DENT SEQUENCE ANALYSIS MUREG/CR4085. USERS MANUAL FOR CONTAIN 10 A Computer NUREG/CR42td HEALTH EFFECTS MODEL FOR NUCLEAR POWER Code for Severe Reactor Accident Containment Analysis PLANT ACCOENT CONSEQUENCE ANALYSIS Part NUREG/CR4091: THE EFFECT OF ALTERNATIVE A'MNG AND ACCl- lintrnduction. Integration & Summary Part 11 Scientific Bases For Hea th DENT SIMULATIONS ON POLYMER PROPERTIES. Effects Models. NUREG/CR-4119 INTEGRITY OF CONTAINMENT PENETRATIONS UNDER SEVERE ACCIDENT CONDITONS FY84 ANNUAL REPORT, Accident Management NUREG/CR4146 SIMULATION OF AN EPRI-NEVADA TEST SITE NUREG/CR4177 V01: MANAGEMENT OF SEVERE (NTS) HYDROGEN BURN TEST AT THE CENTRAL RECEIVER TEST ACCIDENTS Perspectsves On Managing Severe Accidents in Commer. FACILITY. cial Nuclear Power Plants NUREG/CR4 tG9 AN APPROACH TO TREATING RADIONUCLIDE NUREG/CR4177 V02 MANAGEMENT OF SEVERE DE' CAY HE ATING FOR USE IN THE MELCOR CODE SYSTEM ACCIDENTS Estending Plant Operating Procedures into The Severe NURFG/CR 4172: A USER'S GUIDE FOR MERGE. Accident Regime NUREG/CR4f t 5: AN ASSESSMENT OF DOSIMETRY DATA FOR AC-CIDENTAL PADIONUCLIDE RELEASES FROM NUCLEAR REAC- Accident Mitigation
~~' * ^ "~~
NUREG/CR4197: SAFETY GOAL SENSITIVITY STUDIES UNIT 3 PROdABILISTIC SAFETY STUDY Containtnent Falure MUREG/CR4210 MATADOR A COMPUTER CODE FOR THE ANALY-Modes Radiolog. cal SourceJerms And OHsste Consmuences SIS OF RADIONUCUDE DEMAVIOR DURING DEGRADED CORE AC. CIDENTS IN LIGHT WA1ER RE ACTORS Accident Sequence RUREG/CR-4211: MATADOR (METHODS FOR THE ANALYSIS OF NUREG/CR 4140- DOMINANT ACCIDENT SEQUENCES IN OCONEE-1 TRANSPORT AND DEPOSITION OF RADIONUCLIDES) CODE DE- PRESSURIZED W ATER HEACTOR SCRIPTON AND USER'S MANUAL NUREG/CR 4214. HEALTH EFFECTS MODEL FOR NUCLEAR POWER Acid Digestion PLANT ACCIDENT CONSEQUENCE ANALYSIS Part NUREG/CR3444 V02. THE IMPACT OF LWR DECONTAMINATIONS I fntroduction. Integration & Summary Part il Scientific Bas #s For Health ON SOLIDIFICATION. WASTE DISF0 SAL AND ASSOCIATED OCCU-PATIONAL EXPOSURE. NURE C 5\ V01: MITIGATIVE TECHNOUES FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR Acid Waste NUREG/CR-3906 URANIUM MILL TAILINGS NUR 42 V2 MIT TVE C U F 1 GROUND- NEUTRAll2ATIONCONTAMlNANT COMPLEXATION AND TAILINGS WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR LEACHING STUDY. ACCOENTS Volume 2 Case Study Analysis Of Hydrologic Character-ltaton And M.ty;ative Schemes Acoustic Emission NUREG/CR4262 V01. EFFECTS OF CONTROL SYSTEM FAILURES NUREG-0915 V03 COMPILATION OF CONTRACT RESE AhCH FOR ON TR ANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC THE MATERIALS ENGlNE ERING BRANCH. DIVISION OF ENGINEER BOluNG WATER REACTOR Main Report. NUREG/CR4262 V02. EFFECTS OF CONTROL SYSTEM FAILURES NL 1 5 04 ESEA H RAM N Non-Destructive En-ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC NUREG/ 3825 V03 4 ACOUSTIC EMISSION / FLAW RELATONSHIP NUR CR 4 7 PF MO EVELOFMENT ASSESSMENT FOR IN-SERVICE MONITORING OF NUCLEAR PRESSURE NUREG/CR4294 LEAK RATE ANALYSIS OF THE WESTINGHOUSE VESSELS Ouarterly Report, Apne 1934 September 1984 Volumes 3 RE ACTOR COOLANT PUMP NUREG/CR4304 PRESSURE VESSEL FRACTURE STUDIES PER. and4. TAINING TO THE PWR THERMAL SHOCK ISSUE Expenma.nt TSE.7 NUREG/CR 3915- ACOUSTIC EMISSON RESULTS OOTAINED FROM NUREG/CH-4321 FULL SCALE MEASUREMENTS OF SMOKE TRANS. TESTING THE 20-1 INTERMEDIATE SCALE PRESSURE VESSEL PORT AND DEPOSITION IN VENTILATON SYSTEM DUCTWORK. NUREG/CR 4300 V0f ACOUSTIC EMISSION / FLAW RELATIONSHIP NUREG/CR-4325. A PARAMETRIC STUDY OF PWR PRESSURE FOR IN SERVICE MONITORING OF NUCLEAR PRESSURE VESSEL INTEGRITY DURING OVERCOOLING VESSELS Progress Report. October March 1965 ACCIDENTS.CONSIDERiNG E40TH 2-D AND 3 D FLAWS WUREG/CR4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES Acoustic Leak Detection ON TRANSIENTS AND ACCIDENTS AT A 3 LOOP WESTINGHOUSE NUREG/CR 4124 NDE OF STAINLESS STEEL AND ON LINE LEAK PRESSURilED WATER REACTOR Men Report MONITORING OF LWOS Annual Report October 1983 September NUREG/CR4326 V02. EFFECTS OF CONTROL SYSTEM FAILURES 1984 ON TRANSIENTS AND ACCOENTS AT A 3 LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR Appendiceg Acronyms And initialisms RUREG/CR4342 UNCERTAINTY AND SENSITIVITY ANALYSIS OF A NUREG4544 R02 A HAND 000K OF ACHONYMS AND INITIALISMS MOCEL FOR MULTICOMPONENT AEROSOL DYNAMICS MUREG/CR 4360 V01 CALCULATIONAL METHOD 3 FOR ANALYSIS AClimde Element OF POSTULATED UF 6 REL EASES NUREG/CR 4094 FIELD EXPERIMENT DETERMINATONS OF DISTRL NUREG/CR4360 V02. CALCULATIONAL METHOOS FOR ANALYSIS BUTION COf FFICIENTS OF ACT6NiDE ELEMENTS IN SULFATE OF POSTULATED UF6 RELEASES. LAKE ENVIRONMENTS
Subject index 149 Administrat6ori NUREG/CR4008: GENERAL EXTRAPOLATION MODEL FOR AN iM-NUREG/CR-4125 V01: GUIDEUNES AND WORKBOOK FOR ASSESS. PORTANT CHEMICAL DOSE-RATE EFFECT. MENT OF ORGANIZATION AND ADMINISTRATON OF UTluTIES NUREG/CR-409t; THE EFFECT OF ALTERNATIVE AGING AND ACCI. SEEKING OPrRATING UCENSE FOR A NUCLEAR POWER DENT SIMULATIONS ON POLYMER PROPERTIES. PLANT. Volume 1 Guidelines For Utility Orgarszation And Adrmnestration Plan. NUREG/CR4144: IMPORTANCE RANKING BASED ON AGING CON-NUREG/CR-4125 V02- GUIDEUNES AND WORKBOOK FOR ASSESS-SIDERATONS OF COMPCNENTS INCLUDED IN PROBABILIS10 RISK ASSESSMENTS. MENT OF ORGAN!ZATION AND ADMINISTRATON OF UTluTIES NUREG/CR-4156: OPERATING EXPERIENCE AND AGING-SEISMIC SEEKING OPERATING UCENSE FOR A NUCLEAR POWER ASSESSMENT OF ELECTRIC MOTORS. PLANT. Volume 2. Workbook For A*sessment Of Organtration And Man-agement. NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC
~~.m __~~
MOTOR-OPERATED VALVES USED IN ENGINEERED SAFETY-FEA-TURE SYSTEM $ OF NUCLEAR POWER PLANTS. IdURE5CR-3455. A COMPARISON OF ODINE. KRYPTON.AND XENON Air Cleaning System R ET NE ICIENCIES FOR VARIOUS SILVER LOADED AD-NUREG/CR-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT. RUREG/CR4094 FIELD EXPERIMENT DETERMINATIONS OF OtSTRI. ABILITY PRACTICES. BUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE NUREG/CR 4225:
~~SUMMARY~~
OF EFFICIENCY TESTING OF STAND-LAKE ENVIRONMENTS. ARD AND HIGH-CAPACITY HIGH-EFFICIENCY PARTCULATE AIR NUREG/CR-4237: MOBIUTY OF RADIONUCUDES IN HIGH CHLORIDE FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURt2A-ENVIRONMENTS. TION AND EXPLOSIVE SHOCK WAVES. Adverse Systems interact 6ori Air Sampler NUREG/CR-3922 V01: SURVEY AND EVALUATION OF SYSTEM NUREG/CR-3455. A COMPARISON OF ODINE. KRYPTON.AND XENON INTERACTON EVENTS AND SOURCES Main Report And Appendices RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD. NUR / R 3922 V02: SURVEY AND EVALUATON OF SYSTEM INTERACTION EVENTS AND SOURCES Appendices C And D. Airtiorne Reeeeee 4 ,,,,,g NUREG/CR4088. METHODS FOR ESilMAfiNG RADOACTIVE AND NUREG/CR 3537: EXPED'ENT METHODS OF RESPIRATORY TOXIC AIRBORNE SOURCE TERMS FOR URANIUM MfluNG OPER-PROTECTIONill. SUBMICRON PARTICLE TESTS AND
~~SUMMARY~~
ATONS. OF OVAUTY FACTORS. NUREG/CR 3830 V02: AEROSOL RELEASE AND TRANSPORT Alarm Resolution PROGRAM Semiannual Progress Report For Apnl 1984-September NUREG/CR-4108 DEVELOPMENT OF MC&A ALARM RESOLUTION 1984_ PROCEDURES. NUREG/CR-3984: BIOLOGICAL CHARACTER 12ATON OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL. All0Y Ste*88 ING EFFLUENTS Annual Progess Report.Apr41983 4 March 1984 NUREG/CR4437. EXPLORATORY STUDIES OF ELEMENT INTERAC-NUREG/CR4045: LITERATUNE REVIEW ON AEROSOL-SAMPUNG TIONS AND COMPOSITION DEPENDENCIES IN RADIATION SENSI-DEVICES FOR RESPIRATORY FIELD STUDIES TlVITY DEVELOPMENT. NUREG/CR-4085. USERS MANUAL FOR CONTAIN 1.0 A Computer Code for Severe Reactor Accident Containment Analyses. Alternating Current NUREG/CR 4111: A COMPARATIVE STUDY OF HEPA FILTER EFFi-CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR JET. NUREG/CR4294 LEAK RATE ANALYSIS OF THE WESTINGHOUSE GENERATED Dl-2-ETHYLHEXYL SEBECATE.DI-2 ETHYLHEXYL REACTOR COOLANT PUMP' NU E / 30 CDFA R AEROSOL PARTICLE CAP. U EG C V TERNATIVF METHODS FOR DISPOSAL OF NUREG/CR 4205: TRAP-VELT2 USER'S MANUAL LOW LEVEL RADIOACTIVE WASTES Task 2A.Techrucal Requwements NUREG/CR-4210 MATADOR A COMPUTER CODE FOR THE ANALY. For Belowground Vault Dsposal Of Low Level Radhoactrve Waste. SIS OF RADIONUCUCE BEHAVOR DURING DEGRADED CORE AC. NUREG/CR 3774 V03 ALTERNATIVE METHODS FOR DISPOSAL OF COENTS IN UGHT WATER REACTORS. LOW LEVEL RADOACTIVE WASTES Tasl6 28. Technical Requwements NUREG/CR 4255 V01' AEROSAL RELEAS". AND TRANSPORT PRO- For Aboveground Vault Onposal Of Low Level Radsoactive Waste. GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 - NUREG/CR 3774 V04: ALTERNATIVE METHOO FOR DISPOSAL OF MARCH 1985. LOW LEVEL RADIOACTIVE WASTE. Task 2C: Technical Requwements RUREG/CR4264: INVESTIGATON ON HIGH EFFICIENCY PARTICU- For Earth Mounded Concrete Bunlier Disposal Of Low Leve4 Radioac. LATE AIR FILTER PLUGGING OY COMBUSTON AEROSOLS tive Waste NUREG/CR-4342: UNCERTAINTY AND SENSlflVlf Y ANALYSIS OF A NUREG/CR 3774 V05 ALTF9 NATIVE METHODS FOR DISPOSAL OF 94UR / -44 ASE R N THE FUEL LOW LEVEL RADOACTNei WASTE. Task 2E:Techrscal Requwements AEROSOL SIMULANT TEST FACluTY.ONDERSODIUM EXPFRI- For Shaft Dsposal Of Low Leve6 Radioactive Waste. MENTS Amtplent Radletlen Level NUREG/CR4388: AEROSOL BEHAVOR MODELING (TASK 3) SUP-PORT SERVICES FOR RESEARCH AND EVALUATON OF SEVERE NUREG4837 V04 NO3. NRC TLD DIRECT RADIATON MONITORING ACCIDENT PHENOMENA AND MITIGATION. NETWORK Progress Report. Jury September 1984 NUREG4837 V04 N04. NRC TLD DIRECT RADIATON MONITORING Agende REPORT. Progress Report, October December 1984. NUREG-0936 V03 N04: NRC REGULATORY AGENDA.Ouarterty Report. October-December 1984 Americium 241 NUREG 1153: INSPECTION REPORT OF UNAUTHORIZED POSSES-Atene SiON AND USE OF UNSEALED AMERICIUM 241 AND SUBSEQUENT NUREG 1144. NUCLEAR PLANT AGING RESEARCH (NPAR) PRO. CONFISCATION JC. Haynes Company. Newark. Ohio. GRAM PLAN NUREG4,R-233l V04 N2. SAFETY RESEARCH PROGRAMS SPON. Anney,6, SORED BY OFFICE OF NUCLEAR REGULATORY NUREG 1140 DAFT FC: A REGULATORY ANALYSIS ON E.MERGtNCY NURE /CR 23 SNET E A CH PRd MS SPON- PREPAREDNFSS FOR FUEL CYCLE AND OTHER RADIOACTIVE
- SORED BY OfflCE OF NUCLEAR REGULATORY MATERIAL UCENSEES Draft Report for Comment RESEARCH Ouarterty Progress Report, October 1. December 31, NUREG/CR4360 V01. CALCULATONAL METHODS FOR ANALYSIS OF POSTULATED UF8 RELEASES NUREG/CR 3710 LABORATORY STUDIES OF A BREACHED NUCLE, NUREG/CR 4360 V02: CALCULATONAL METHODS FOR ANalV$lS AR WASTE REPOSITORY IN BASALT OF POSTUt ATED UF8 RELEASES NUREG/CR-3819- SURVEY OF AGED POWER PLANT F AClutlFS NUREG/CH 3W8 V03. UGHT WATER-REACTOR SAFETY MATERIALS Annual Report ENG'NEERING RESEARCH PROGRAMS Ouartedy Progress NUREG 1945 V01: U S. NUCLEAR REGULATORY COMMISSION 1984 Report Ortober-December 1984. ANNUAL REPORT.
150 Subject index Annuncletor System BWR-LTAS NUREG/CR-3987. COMPUTERIZED ANNUNCIATOR SYSTEMS. NUREG/CR-3764. BWR-LTAS A BOILING WATER REACTOR LONG-TERM ACCOENT SIMULATION CODE. NUREG/CR-4422: A REVIEW OF THE MODELS AND MECHANISMS potence Study FOR ENVIRONMENTALLY-ASSISTED CRACK GROWTH OF PRES ~ NUREG/CR 4355 V01. 238 PU(IV) IN MONKEYS Overwew Of Metabo-SURE VESSEL AND PIPING STEELS IN PWR ENVIRONMENTS. Imm. SeMer DepedeHon Event i H CR-4133: NUCLEAR POWER SAFETY REPORTING SYSTEM NUREG/CR 4067.
~~SUMMARY~~
OF BARRIER DEGRADATION EVENTS r ilWPLEMENTATON AND OPERATONAL SPECIFICATONS. ANr) SMALL ACCIDENTS IN U S. COMMERCIAL NUCLEAR POWER Anticipated Trenaient W6thout Scram PLANTS. NUREG/CR4633 V01 Sf: TRAC-BDt/ MOD 1.AN ADVANCED BEST ES-seean TIMATE COMPUTER PROGRAM FOR BOILING WATER REACTOR TRANSlENT ANALYSIS. NUREG/CR.2663 V01: INFORMATON NEEDS FOR CHARACTER 12A-NUREG/CR-4046: DETERMINING CRITICAL FLOW VALVE CHARAC- TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-TERISTICS USING EXTRAPOLATON TECHNIOUES. IC MEDIA Main Report. NUREG/CR-2663 V02: INFORMATON NEEDS'FOR CHARACTER 12A-Ant #truet TON OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-NUREG-0970: PROCEDURES FOR MEETING NRC ANTITRUST RE- IC MEDIA Appendees. SPONSIBILITIES NUREG/CR-3710 LABORATORY STUDIES OF A BREACHED NUCLE. AR WASTE REPOSITORY IN BASALT. Aqueous lod 6ne Chem 6etry NUREG/CR-3851 V04. EVALUATION OF RADIONUCLIDE GEOCHEMI-NUREG/CR-35t4 V02: THE CHEMICAL BEHAVIOR OF ODINE IN CAL INFORMATON DEVELOPED BY DOE HIGH. LEVEL NUCLEAR AQUEOUS SOLUTONS OP TO 150 C 11 Radaten-Redon Condtsons' WASTE REPOSITORY SITE PROJECTS. Annual Progress Report For October 1983-September 1984 Agulfer Restoration NUREG/CR-4114 VALENCE EFFECTS ON THE SORPTON OF NU-NUREG/CR 3709 METHODS OF MINIMl21NG GROUND-WATER CON- SA NER TAMINATON FROM IN SITU LEACH URANIUM MINING Final Report CpgOE,S ON R Art 6ficial inteil6gence WASTE REPOSITORY. Volume 1 Basaft. NUREG/CR-3481 V02: NUCLEAR POWER PLANT PERSONNEL QUAll. NUREG/CR-4236 V01: PROGRESS IN EVALUATION OF RADtONU-FICATIONS AND TRA NiNG TAPS - The Task Analyse Profiling CLIDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-System. LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS REPORT NUREG/CA-4272- RESPONSE TREE EVALUATON EXPERIMENTAL FOR OCTOBER DECEMBER 1984. ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR NUREG/CR-4303: HIGH-LEVEL WASTE PRECLOSURE SYSTEMS SAFETY ANALYSIS Phase 1. Final Report. OPERATORS. Aenette Clem Bayeeson Method NUREG/CR-4233 DISTRIBUTON OF CORBICULA FLUMINEA AT NU- NUREG/CR4145. EARTHOUAKE RECURRENCE INTERVALS AT NU. CLEAR FACILITIES. CLEAR POWER PLANTS: ANALYSIS AND RANKING. Atmosphere Solow-Weter Table PdVREG/CR4074 THE PERFORMANCE OF DEFECTED SPENT LWR NUREG/CR 4061. LEACHATE PLUME MIGRATION DOWNGRADIENT FUEL RODS IN INEAT GAS AND DRY AIR STORAGE ATMOS- FROM URANIUM TAILINGS DISPOSAL IN MINE STOPES. PHERES. Belowground Vault
~~# II# #~~
N R SCIENTIFIC CRITIQUE OF AVAILABLE MODELS LOW LEVEL RADIOACTIVE WASTES Tasti 2A Technical Requirements FOR REAL TIME StMULATONS OF DISPERSON For Belowground Vault Dsposal Of Low Level Radoective Waste NUREGICR 4158 A COMPILATON OF INFORMATION ON UNCER. TAINTIES INVOLVED IN DEPOSITION MODELING. senc6nerti Probleme
~~@dVREG/CR-4159: COMPARISON OF THE 1981 INEL DISPERSION NUREG/CR 1677 V02. PIPING BENCHMARK PROBLEMS. VOLUME 11 DATA W'TH RESULTS FROM A NUMBER OF DiFFERENT MODELS.~~
DYNAMIC ANALYSIS INDEPENDENT SUPPORT MOTON RE. Atomisetton SPONSE SPECTRUM METHOD. NUREGICR-3937: STEAM GENERATOR TUBE RUPTURE ODINE TRANSPORT MECHANISMS. Ted 1 Espenmental Studes Sentonite NUREG/CR-4383: HIGH PRESSURE INJECTON OF MELT FROM A NUREG/CA 3710' LABORATORY STUD:ES OF A BREACHED NUCLE-REACTOR PRESSURE VESSEL . THE DISCHARGE PHASE. AR WASTE REPOSITORY IN BASALT. Austentt6c Steiniees Steel Bete Does Mete NUREG/CR-3613 V02. EVALUATION OF WELDED AND REPAIR- NUREG/CR-4203. A CALCULATONAL METHOD FOR DETERMINING WELDED STAINLESS STEEL FOR LWR SERVICE Annual Report For BOLOGICAL DOSE RATES FROM IRRADIATED RESEARCH REAC. NU G/CR 3613 V03 N1: EVALUATON OF WELDED AND REPAlR. WELDED STAINLESS STEEL FOR LWR SERVICE. Semiannual Report Sete Mediation For October 1984 Through March 1985- NUREG/CR 4266 STANDARD BETA-PARTICLE AND MONOENERGE.
TIC ELECTRON SOURCES FOR THE CAllBRATON OF BETA-RADI-ATON PROTECTON INSTRUMENTATION NU GC 28R EW OF THE VOGTLE UNITS t AND 2 AUXILIA-RY FEEDWATER SYSTEM RELIABILITY ANALYSIS. g,,,y 3,gy p Velve NUREG/CR-4401. BEHAVOR OF CONTROL RODS DURING CORE NUREG/CR 4145. EARTHOUAKE RECURRENCE INTERVALS AT NU. DEGRADATION PRESSURIZATON OF SILVER INDIUM.CADMluu CLEAR POWER PLANT *: ANALYSIS AND RANKING. CONTROL RODS.
gWip 860eccumulatnon NUREG/CH-3851 V04. EVnLUATON OF RADONUCLIDE GEOCHEMI- NUREG/CR-3981. BIOACCOMULATION OF P 32 IN BLUEGILL AND CAL INFORMATON DEVELOPED Bf DOE HtGHLEVEL NUCLEAR CATFISH WASTE REPOSITORY SITE PROJECTS Annual Progress Report For October 1983 September 1984 peoseeay NUREG/C44134 REPOSITORY ENVIRONMENTAL PARAMETERS NUREG/CR 3964 BIOLOGICAL CHARACTERl2ATION OF RADIATION RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL EXPOSURE AND DOSE ESilMATES FOR INHALED URANIUM MILL. WASTE PACKAGES. ING EFFLUENTS Annual Progress Report.Apnl 1983. March 1984.
Subject index 151 36odegradation By-Product Motorial Factitty NUREG/CR4200 BIODEGRADATION TESTING OF SOLIDIFIED LOW- NUREG/CR3657: PREUMINARY SCREENING OF FLTL CYCLE AND LEVEL WASTE STREAMS. BY PRODUCT MATERIAL LICENSES FOR EMERGENCY PLANNING. Biofouling Bypese Test NUREG/CR4070 V02: BlVALVE FOUUNG OF NUCLEAR POWER NUREG/CR4252. INDEPENDENT ASSESSMENT OF TRAC-PO2/ MOOT PLANT SERVICE WATER SYSTEMS. Volume 2. Current Status Of Bio. CODE WITH BCL ECC BYPASS TESTS. j fouhng Surveillance And Centrol Techtwques. i Shake Fouling C&O LCO 13 Settery Cell l NUREG/CR4070 V03: BlVALVE FOULING OF NUCLEAR PC?ER NUREG/CR-4096. TEST SERIES 3. SEISMIC FRAGILITY TESTS OF PLANT SERVICE WATER SYSTEMS Factors That May Intensafy The NATURALLY AGED CLASS 1E C&D LCU-13 BATTERY CELLS. Safety Consequences Of Biofouhng. CDA NUREG/CR 3944: TRAN B-3 EXPERIMENTAL INVESilGATON OF CR-4196: OVERVIEW OF TRAC-801 (VERSON 12) ASSESS-MENT STUDIES. FLO CH NE l Sody Waves CITADEL RUREG/CR4354: A STUDY OF SEISMICITY AND TECTONICS IN NEW NUREG-1108 RADIOACTIVITY TRANSPORT FOLLOWING STEAM ENGLAND. Final Report. GENERATOR TUBE RUPTURE. Semb Threat COSRA ZUREG-0525 RfD SAFEGUARDS
~~SUMMARY~~
EVENT UST NUREG/C43810 V03: REACTOR SAFITY RESEARCH ' (SSEL).REVISON 10. PROGRAMS Quarterly Report. July-September 19f 4 NUREG/CR 3810 V04. REACTOR SN ETY RESEARCH Boolean Algetra . PROGRAMS Quarterty Report. October Decembrt 1984 i NUREG/CR4213 SETS REFERENCE MANUAL NUREG/CR4318 V01: REACTOR FAFETY RESEARCH PROGRAMS Quarterly Report. January-March 1985. NUREG/CR-4174 ROCK MASS SEAUNG E)PERIMENTAL ASSESS- COSRA.TF MENT OF BOREHOLE PLUG P8hrORMANCE Annual Report, June NUREG/CR4166. ANALYSIS OF FLECHT SEASET 163-ROD BLOCKED 1983. May 1984. BUNDLE DATA USING COBRA-TF. Doroselicate Glase CONTAIN Y NUREG/CR3900 V04. LONG TERM PERFORMANCE OF MATERIALS NUREG/CR4085: USERS MANUAL FOR CONTAIN 1.0 A Computer USED FOR HIGH-LEVEL WASTE PACKAGING Arinual Report.Aprd Code for Sewte Reactor Accxtent Containment Analyss. 1984. Apnl 1985 NUREG/CR4340 V01: REACTOR SAFETY RESEARCH SEMIANNUAL REPORT. January a June 1985. I Steals Flow Rete , NUREG/CR-4041: SYSTEM ANALYSIS HANDBOOK. CONTING 1 NUREG/CR4071: EXPLORATORY TREND AND PATTERN ANALYSIS RUREG/CR3791: CLOSEOUT OF lE BULLETIN 79-09 FAILURE OF GE TYPE AK 2 CIRCUIT BREAKERS IN SAFETY RELATED SYSTEMS- CORCON 1 Su m Beha
NUREG 0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS (Draft Report For Commenti NUREG/CR 3830 V02: AEROSOL RELEASE AND TRANSPORT PROGRAM Semeanrw Progress Report For Apnl 1984-September NUREG/CR-4340 VOI: REACTOR SAFETY RESEARCH SEblANNUAL 1984 REPORT.aanuary . June 1W66 CORRAL E"d9'I NUREG-1100 V01: FY 1986 BUDGET ESTIMATES' NUREG/CR4210: MATADOR A COMPUTER CODE FOR THE ANALY-
} ' SIS OF RADIONUCLIDE BEHAVOR DURING DEGRADED CORE AC-Buoyant Plume CIDENTS IN LIGHT WATER REACTORS. RUREG/CR3426 V01 THERMAL AND FLUID MixlNG IN 1/2 SCALE CORRAL 2 i NU /CR 4 VO HE MAL A LU MIXING IN 1/2-$CALE NUREG/CR42 t t: MATADOR (METHODS FOR THE ANALYSIS OF TEST FACluTY. Data Report. TRANSPORT AND DEPOSITION OF RADIONUCUDES) CODE DE. ' SCRIPTION AND USER'S MANUAL. Burtal RUREGICR 4194 LOW LEVEL NUCLEAR WASTE SHALLOW LAND CORSOR
' BURIAL TRENCH ISOLATON Final Report. October 1981 September NiIREG-0956 DAFT FC: REASSESSMENT OF THE TECHNICAL BASES 1984 s J ESTIMATING SOURCE TERMS (Draft Report For Comment)
, NUREG/CR4173 CORSOR USER'S MANUAL i Burial Environment i CRAC2 NUREG/CR4083. ANALY5ES OF SOfLS FROM THE LOW LEVEL RA. DIOACTIVE WASTE DISPOSAL SITES AT BARNWELL SC AND NUREG/CR4038 SENSITIVi!Y AND UNCERTAINTY STUDIES OF THE R rCHLAND.WA. CRAC2 COMPUTER CODE j NUREG/CR-4 t 99 A DEMONSTRAff0N UNCERTAINTY / SENSITIVITY Burial Site ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQULNCE I f00 REG /CR 3865: EVALUATION OF THE RADCACTIVE INVENTORY MODEL CHAC2, l IN.AND ESTIMATON OF ISOTOPIC RELEASE FROM.THE WASTE IN EIGHT TRENCHES AT THE SHEFFIELD LOW-LEVEL WASTE BURIAL CRT D6eplay SITE. NUREG/CR-3767. INTERACTIVE SIMULATOR EVALUATON FOR CAT. GENERATED DSPLAYS Buried Weste NUREGICR4227. HUMAN ENGINEERING GUIDELINES FOR THE NUREG/CR-4063 ANALYSES OF SOIL 3 FROM THE LOW LEVEL RA. EVALUATION AND ASSESSMENT OF VIDEO DISPLAY UNITS DIOACTIVE WASTE DISPOSAL SITES AT BARNWELL SC AND RICHLAND,WA. CAW RUREG/CR4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL NUREG/CR-4070 V02. BIVALVE FOUUNG OF NUCLEAR POWER WASTE FORM ACCEPTANCE CRITERIA. PLANT SERVICE WATER SYSTEMS Volume 2 Cunent Status Of Bio-foubtag Surveeltance And Control facteruques Burner Location NUREG/CR4070 V03. BlVALVE FOULING OF NUCLEAR POWER HUREG/CR4112 V01: INVESilGATON OF CABLE AND CABLE PLANT SERVICE-WATER SYSTEMS Factors That May intenefy The SYSTEM FIRE TEST PARAMETERS Tash A fEEE Flame Test. Safety Consequences Of Biotouhng I _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ , _ _ _ _ _ - _ _ _ _ ~ ~ _
152 Subject index Cable NUREG/CR-3980 V02. LIGHT-WATER-REACTOR SAFETY FUEL SYS-NUREG/CR-4112 V01: INVESTIGATION OF CAF E AND CABLE TEMS RESEARCH PROGRAMS Quarterly Progress Report.Aprd-June SYSTEM FIRE TEST PARAMETERS Task A IEEE Flame Test 1984 NUREG/CR-4112 V02: INVESTIGATION OF CABLE AND CABLE NUREG/CR-4015. EFFECT OF STAINLESS STEEL WELD OVERLAY SYSTEM FIRE TEST PARAMETERS Task B Firestop Test Method. CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL PLATES IN BENDING SERIES 1 0/W218. m SMAM W M WN MSNH R CR-4266. STANDARD BETA-PARTICLE AND MONOENERGE- swamn Dawon h TIC ELECTRON SOURCES FOR THE CAUBRATION OF BETA RADI. suits For The Third Matenals Test (MT 31 Second Campa gn. ATION PROTECTION INSTRUMENTATION- NUREG/CR-4219 V01. HEAVY-SECTION STEEL TECHNOLOGY PRO-GRAM SEMIANNUAL PROGRESS REPOFIT FOR OCTOBER 1984 - Canada NUR C 4317 V01 ADIAN SEISMIC AGREEMENT. Technical A CH 95 p
~~. FIVE-YEAR PLAN FY 1984-1968.~~
Capacitor Discharge vaporization Esperiment NUREG/CR-4346. AEROSOL RELEASE EXPERrMENTS IN THE FUEL Class IE Pressure Transmitter AEROSOL SfMULANT TEST FACILITY UNDERSODIUM EXPERI. NUREG/CR-3863 ASSESSMENT OF CLASS 1E PRESSURE TRANS. MENTS MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT SCREENING TESTS Cash Flow Analyste NUREG 1131: FINANCIAL ANALYSIS OF POTENTIAL RETROSPECTIVE Cleanup PREMlUM ASSESSMENTS UNDER THE PRICE-ANDERSON NUREG/CR 4399 POSSIBLE OPTIONS FOR REDUCING OCCUPA. SYSTEM. TlONAL DOSE FROM THE TMI-2 BASEMENT. Cast Stainless Steel Closecut NUREG/CR-3998 V02. LIGHT-WATER-REACTOR SAFETY MATERIALS NUREG4905- CLOSEOUT OF IE BULLETIN 7912 SHORT-PERIOD ENGINEERING RESEARCH PROGRAMS Quarterty Progress SCRAMS AT BOILING. WATER REACTORS NUREG/CR-3791: CLOSEOUT OF IE BULLETIN 79 09 FAILURE OF GE NUR Cl 12 P E OF STAINLESS STEEL AND ON UNE LEAK TYPE AK 2 CIRCUlf BREAnERS IN SAFETY RELATED SYSTEMS MONITORING OF LWRS. Annual Report. October 1983 September NUREG/CR-3794 CLOSEOUT OF IE BULLETIN 80 25 OPERATING 9934 PROBLEMS WITH TARGET ROCK SAFETY RELIEF VALVES AT NUREG/CR-4204 LONG TERM EMBRITTLEMENT OF CAST DUPLEX S AIN S TEELS IN LWR SYSTEMS Annual Report.0ctottar 1983 - NU EG/CR-4004 CLOSEOUT OF IE BULLETIN 79 25 FA! LURES OF WESTINGHOUSE BFD RELAYS IN SAFETY-RELATED SYSTEMS Cement NUREG/CR-4006 CLOSEOUT OF IE BULLETIN 8t 01 SURVEfLLANCE NUREG/CR-4181: LEACHABiUTY OF RADIONUCUDES FROM OF MECHANICAL SNUBBERS. CEVENT SOLIDIFIED WASTE FFRMS PRODUCED AT OPERATING Cobalt 40 Teletherapy incident NUCLEAR POWER REACTORS. NUREG-1103 CONTAMINATED MEXICAN STEEilmportaton Of Steel Ces6um Hydroside into The United Slales That Had Been inachertently Contarr4nated NUREG/CR 3197 V01. REACTION BETWEEN SOME CESIUM-LODINE With Cobal. 60 As A Result Of Scrapping Of A Telethorapy Unit COMPOUNDS AND THE REACTOR MATERulS 304 STAINLESS STEELINCONEL 600 & SILVER. Volume ICessum Hydroxide Reac. Code tons. NUREG-0956 DRFT FC. REASEESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS (Draft Report For Comment) CharpY NUREG/CR 233t V04 N2 SAFETY RESEARCH PROGRAMS SPON-i 'JREG/CR 4092 ORNL CHARACTERl2ATION OF HEAVY.SECTION SORED BY OFFICE OF NUCLEAR REGULATORY FEEL TECHNOLOGY PROGRAM PLATES 01.02.AND 03 RESEARCH Ouarterfy Progress Report.Apnl 1 June 30,1984 NUREG/CR 2331 V04 N4 SAFETY RESEARCH PROGRAMS SPCN. SORED BY OF FICE OF NUCLEAR REGULATORY N E /C 435. ORGANIC COMPLEXANT-ENHANCED MOBfLITY OF RESEARCH Ouarterty Progress Report. October 1 Decerrcer 31, TOx!C ELEMENTS IN LOW-LEVEL WASTES Annual ReportJdy 1984
~~-June 85 NUREG/CR-2331 V05 N1. SAFETY RESEARCH PROGP nMS SPON.~~
SORED BY OFFICE OF NUCLEAR REGULATORY Chem 6 cal Cleaning RESE ARCH Ouarterly Progress Report. January 1-March 31,1985 NUREG 1155 V02. RESEARCH PROGA AM PLAN Stsam Generators NUREG/CR 4276 V10HATION AND WEAR IN STEAM GENERATOR NUREG/CR 3091 V06 REVIEW OF WASTE PACKAGE VERIFICATION SEMIANNUAL TESTS Semiannual Report Covenng The Pered October 1984 March TUBES FOLLOWING CHEMICAL CLEANING REPORT. t985 NUREG/CR-3208 TRAC PD2 DEVELOPMENTAL ASSESSMENT. Chemical Dose Rate NUREG/CR-3319 LWR PRESSURE VESSEL %URVEILLANCE DOS 4ME-NUREG/CR 400e GENER AL EXTRAPOLATION MODEL FOR AN IM- TRY IMPROVEMENT PROGRAM LWR Pr ser Peactor Survedlance PORTANT CHE ACAL DOSE RATE EFFECT. pnysics-Dosametry Data Base Cornpendium NUREG/CR 3498 TWO-OtMENSIONAL MODE LING OF INTRA-SUBAS-b NURE / R 89 VALUATION OF FIELD-TESTED FUGlitVE DUST D STR BU NI LMFORS CONTROL TEEHNIOUES FOR URANIUM MILL T AiUNGS PILES NUREG/CR 3646 TRAC-PF1 INDEPENDENT ASSESSMENT. NUREG/CR 37a3 ASSESSMENT OF SELECTED TRAC AND RELAPS CALCULATIONS FOR OCONEE 1 PRESSuni/ED THERMAL SHOCK LG/CR 4237 MOBluTY OF RADIONUCUDES IN HIGH CHLORIDE ENVIRONMENTS NURE /CR 374t V02 EVALUATION OF POWER REACTOR FUEL ROD ANALYSIS CAPABluTIES Phase J Topical Report Volume Circulating Raw Water 2 Code Evaluation NUREG/CR-4070 V02. BtVALVE FOULING OF NUCLEAR POWER NUREG/CR 3M6 TRAC.PD2 INDEPENDE'!T ASSESSMENT PLANT SERVICE WATER SYSTEMS Votume 2 Current Status Of Bio. NUREG/CR-3901. DOCUMENTATON AND USER S GUIDE GS2 & GS3 fouhng Survemance And Control Techniques
- VARIABLY SATURATED F LOW AND MASS TRANSPOHT MODELS NLREG/GR 4070 V03 BlVALVE FOUUNG OF NUCLEAR POWER NUREG/CR 3972 SETTLEMENT OF URANIUM MtLL TA UNGS PILES PLANT SERVICE WATER SYSTEMS Factors That May intenwfy The A COMPARISON OF ANALYSIS TECHNIOUES Safefy Consequences Of Biofouhng NUREGICR 4130 ICEDF A CODE FOR AEROSOL PARTICLE CAP-Cladding TURE IN ICE COMPARTMENTS NUREG/CR 3744 V02 HE AVY SECTION STEEL TECHNOLOGY PRO- NUPEG/CR 4169 AN APPROACH TO TREATING RADIONUCUDE GRAM SEMIANNUAL PROGRESS REPORT FOR APHIL SEPTEMBER DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM 1984 NUREG/CR-4172 A USER S GUIDE FOR MERGE
Subject index 153 NUREG/CR-4189: TRAC PFt/ MOD 1 INDEPENDENT Compliance ASSESSMENT.Semiscale MOD 2A Feedwaterene Break (S-SF 3) And NUREG4970 PROCEDURES FOR MEETING NRC ANTITRUST RE. Stsam-Line Break (S SF-5) Tests SPONSIBILITIES. NUREG/CR-4195. OVERVIEW ')F TRAC PD2 ASSESSMENT CALCULA- NUREG/CR4076 DETERMINATION OF COMPLIANCE WITH CRITERIA TlONS FOR FINAL TAILINGS DISPOSAL SITE RECLAMATION NUREG/CR421 f: MATADOR (METHODS FOR THE ANALYSIS OF TRANSPORT AND DEPOSITION OF RADIONUCLIDES) CODE DE. Component Degradation SCRIPTION AND USER'S MANUAL. NUREG 1144 NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-NUREG/CR4252. INDEPENDENT ASSESSMENT OF TRAC PD2/ MODI GRAM PLAN CODE WITH BCL ECC BYPASS TESTS NUREG/CR4253. REVIEW OF TRAC CALCULATIONS FOR CALVERT Compressed Gas CLtFFS PTS STUDY. NUREG/CR 3551: SAFETY IMPLICATIONS ASSOCIATED WITH IN. l NUREG/CR4267: VESSEL INTEGRITY SIMULATION (VISA) CODE PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTON SYS-NU E 9 COMPARATIVE ANALYSIS OF CONSTITUTIVE RE. LATIONS IN TRAC PFL AND RELAP5/ MOD 1- Computer Code NUREG/CR4367: ORVIRT.PC.A 2-0 FINITE ELEMENT FRACTURE l ANALYSIS PROGRAM FOR A MICROCOMPUTER NUREG tt08- RADIOACTIVITY TRANSPORT FOLLOWING STEAM NUREG/CR4388. AEROSOL BEHAVIOR MODELING (TASK 3) SUP- GENERATOR TUBE RUPTURE PORT SERVICES FOR RESEARCH AND EVALUATION OF SEVERE NUREG/CR-3442 RADTWO A COMPUTER CODE FOP SIMULATING ACCIDENT PHENOMENA AND MITIGATION FAST TRANSIENT. IWO-DIMENSIONAL TWO-LAYER RADIONU.
~~^ "~~
SAF Y STUD E FO THE D IS N ACCtDENi KES.RESERVO RS ERS UARIES.A D C A TAL REGIONS EVALUATION Ouarterty Progress Report, January 1. March 31,1985 ' T M TE O PUTER R RAM F R BO G W T A A ACT Code Listing TRANSIENT ANALYSIS NUREGICR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR 3633 V04. TRAC-801/ MOD 1.AN ADVANCED BEST ESTi. FOR LICENSEE EVENT REPORTS Code listings. MATE COMPUTER PROGRAM FOR BOILING WATER REACTOR
, TRANSIENT ANALYSIS Volume 4 Developmental Assessment Coder's Manual NUREG/CR.3764. BWRITAS A BOILING WATER REACTOR LONG.
NUREG/CR-3905 V03. SEQUENCE CODING AND SEAACH SYSTEM TERM ACCIDENT SIMULATION CODE FOR LICENSEE EVENT REPORTS Corfer's Manual NUREG/CR-3772 RELAPS ASSESSMENT SEMISCALE SMALL BREAK NUREG/CR-3905 V04. SEQUENCE CODING AND SEARCH SYSTEM TESTS S UT-1.S-UT 2 S-UT-6.S-UT 7 AND S-UT-8 FOR LICENSEE EVENT REPORTS Coder's Manual NUREG4R-3802 RELAPS ASSESSMENT:OUANTITATIVE KEY PA. RAMETERS AND RUN TIME STATISTICS. Cold Leg NUREG/CR-3810 V03 REACTOR SAFETY RESEARCH NUREG/CR4115 INTERNATIONAL STANDARD PROOLEM 13 (LOFT PROGRAMS Ouartetty Report.Jufy September 1984 EXPERIMENT L2-5). Final Companson Report NUREG/CR-3889 THE MODELING OF BWR CORE MELTDOWN ACCl-Cold-Leg Break DENTS < FOR APPLICATION IN THE MELRPI MOD 2 COMPUTER CODE. NUREG/CR4044. TRAC-PF1 LOCA CALCULATIONS USING FINE- NUREG/CR 3912: MARCH-HECTR ANALYSIS OF SELECTED ACCI-NODE AND COARSE-NODE INPUT MODELS. DENTS IN AN ICE-CONDENSER CONTAINMENT. Collection Efficiency NUREG/CR-3913 HECTR VERSION 1.0 USER S MANUAL NUREG/CR-3919 TR AC.PF1/ MOD 1 INDEPENDENT NUREG/CR-3455 A COMPARISON OF ICDtNE. KRYPTON.AND XENON ASSESSMENT NEPTUNUS PRESSURtZER TEST YO5 RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD- NUREG/CR 3936 RELAP5 ASSESSMENT CONCLUSIONS AND USER SORPTION MEDIA GUIDELINES Combustion NUREG/CR4038 SENSITIVITY AND UNCERTAINTY STUDIES OF THE CRAC2 COMPUTER CODE. NUREG/CR 3444 V02. THE IMPACT OF LWR DECONTAMINATiONS NUREG/CR4044. TRAC-PF I LOCA CALCULATIONS USING FINE-ON SOLIDIFICATION. WASTE DISPOSAL AND ASSOCIATED f OCU- NODE AND COARSE NODE INPUT MODELS PATIONAL EXPOSURE. NUREG/CR 4085 USERS MANUAL FOR CONTAIN 1.0 A Computer NUREG/CR4136 SMOKE A Data Reduction Package Fe* \nalysis Of Code for Severe Reactor Accident Containment Analysis Combustion Expenments NUREG/CR4109 TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR-NUREG/CR423t: EVALUATION OF AVAILABLE DATA FOR PRO 8ABI- 12ED-THERMAL SHOCK TRANSIENTS AT CALVERT CLIFFS / UNIT LISTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR f A Combustion Engineennq PWR POWER PLANTS NUREG/CR4116 NUFEGO-NP.A DIGITAL COMPUTER CODE FOR Commitment THE LINEAR STABILITY ANALYSIS OF BOILING WATER NUCLEAR REACTORS WUREG/CR 2850 V03 POPULATION DOSE COMMITMENTS DUE TO NUREG/CR4127 V01. BWR FULL INTEGRAL SIMULATION TEST RADtOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES (FIST) PROGRAM TRAC BWR MODEL DEVELOPMENT Volume 1 Nu-IN 1981. mencal Methods. Common Cause Fa61ure NUREG/CR-4136 SMOKE A Data Reduction Package For Ana!yses Of Combuston Espriments WUREG/CR4314 BRIEF SURVEY AND COMPARISON OF COMMON Nt' REG /CR 4155 TR AC-PF t / MODI CAUSE FAILURE ANALYSIS. INDEPENDENT ASSESSMENT NORTHWESTERN UNIVERSITY PERFORATED-PLAiE Compact $pecimen CCFL TESTS. WULEG/C'14281 STUDY OF THE EFFECTS OF ELASTIC UNLOAD- NUREG/CRa t68 GT2F A COMPUTER CODE FOR ESilMA TING INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS LIGHT WATER REACTOR FUFL ROD FAILURES NUREG/CR 4185 AN ASSESSMENT OF DOSMETRY DATA FOR AC. Comparison Of Licensing ActMttee CIDE NTAL RADIONUCLIDE RELEASES FROM NUCLEAR REAC. NUREG 1110 COMPARISON OF LICENSING ACTIVITIES FOR OPER. TORS ATWG PLANTS DESIGNED BY BABCOCK & WILCOX. NUREG/CR-4192 THE ANALYSIS OF DRAINAGE AND CONSOLIDA-IlON AT TYPICAL URANIUM MILL TAILINGS SITES Comporison Studies NUREG/CR4196 OVERVIEW OF TRAC-BDt (VERSION 12) ASSESS-NUREG/CR4090 EVALUATION OF NUCLEAR FA 'lJTY DECOMMIS. MENT STUDIES STONtNG PROJECTS Annual Summary Repurt . Fiscal Year 1984 NUREG/CR4199 A DEMONSTHATlON UNCERTAINTY /SENSitivlTY ANALYSIS USING hie HEALTH AND ECONOMIC CONSEQUENCE Compdatson Of Rules MODEL CHAC2. NUREG-0936 V04 NO2. NRC REGULATORY AGENDA Ouarterty NUREG/CR-4210 MATADOR A COMPUTER CODE FOR THE ANALY-Repo. rt.Aprd June 1985 SIS OF RADIONUCLIDE BEHAVIOR DURING DEGRADED CORE AC-NUREG4936 V04 NO3 NRC REGULATORY AGENDA Quarterty CIDENTS IN LIGHT WATER REACTORS ReportJufy September 1985 NUREG/CR 4260 TORAC USER'S MANUAL A Computer Code For Ana NUREG 0936 V03 N04 NRC REGULATORY AGEND A Quarterly tynng Tornado Indsced Flow And Matenal Transport in NWear Face Report,0ctober December 1984 ties
154 Subject index NUREG/CR-4318 V01: REACTOR SAFETY RESEARCH Confiscation PROGRAMS Ouarterty Report. January-March 1985. NUREG 1151 INSPECTION REPORT OF UNAUTHORIZED POSSES-NUREG/CR 4321: FULL SCALE MEASUREMENTS OF SMOKE TRANS- SiON AND USE OF UNEEALED AMERICIUM 241 AND SUBSEQUENT PORT AND DEPOSITION IN VENTILATION SYSTEM DUCTWORK. CONFISCATION J C. Haynes Company. Newark,Otvo. NUREG/CR-4340 vot; REACTOR SAFETY RESEARCH SEMlANNUAL REPORT January June 1985. Congress NdREG/CR 4360 V0t; CALCULATIONAL METHODS FOR ANALYSIS NUREG 0090 V08 N01: REPORT TO CONGRESS ON ADNORMAL OF POSTULATED UF6 RELEASES OCCURRENCES Janurary March 1985 NUREG/CR 4360 V02 CALCULATIONAL METHODS FOR ANALYSIS OF POSTULATED UF6 RELEASES NU G/ R t EVIEW AND EVALUATION OF THE MILLSTONE A BD SER U DEL N UNIT 3 PROBABfLISTIC SAFETY STUDY. Containment Failure Computer Display Modes. Radiological Source-Terms And Ottsite Consequences. NUREG/CR 3767: INTERACTIVE SIMULATOR EVALUATION FOR CRT-GENERATED DISPLAYS. Consequence Model NUREG/CR-4350 V07. PROBABILISTIC RISK ASSESSMENT COURSE Computer Model DOCUMENTATION Volume 7 Environmental Transport And Conse-NUREG/CR-3904 A COMPARISON OF UNCERTAINTY AND SENSITIV* quence Analyses ITY ANALYSIS TECHNIQUES FOR COMP' ITER MODELS NUREG/CR 4042: A 3-DIMENSONAL COMPUTER MODEL TO S.MU" Consondation Characteristics LATE FLUID FLOW AND CONTAINMENT TRANSPORT THROUGH A NUREG/CR-4087: MEASUREMENTS OF URANIUM MILL TAILINGS
^ " "^# 8 NUR G/ 45 Of T GATIVE TECHNIQUES FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR Container ACCIDENTS Volume 1 Analyses Of Genere Site Conditions- NUREG/CR 3829 AN EVALUATION OF THE STABILITY TESTS REC-OMMENDED IN THE RAANCH TECHNICAL POSITON ON WASTE CON ^' ^ ^t UR / 42 7 HU AN ENGINEERING GUIDELINES FOR THE FOR ,
NUp( y02 TER PERFORMANCE OF MATERIALS EVALUATION AND ASSESSMENT OF VIDEO DISPLAY UNITS UMD FOR HIGH-LEVEL WASTE PALKAGING Ouar1erty Report. July-Septemtyer 1984. Computer Program NUREG/CR 3900 V04 LONG TERM PERFORMANCE OF MATERIALS NUREG t tS7; TPOWR2: THERMAL POWER DETERMINATON FOR USED FOR HIGH-LEVEL WASTE PACKAGING Annual Report.Aptd WESTINGHOUSE REACTORS VERSION 2 0ser's Guide 1984 Aptd 1985 NUAEG/CR 3413. OFF-SITE CONSEQUENCES OF RADOL OGICAL NUREG/CR-UN V01. LONG. TERM PERFORMANCE OF MATERIALS ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON. TAMINATION USED FOR HIGH-LEVEL WASTE PACKAGING Fwst Ouarter9 NUREG/CR-4020 HMSA COMPUTER PROGRAM FOR Report. Year Four Apr& June 1985, TRANSIENT,THREE DIMENSIONAL MIXING GASES NUREG/CR 4122: A FORTRAN 77 PROGRAM AND USER'S GUOE Containment FOR THE CA.CULATION OF PARTIAL CORRELATION AND STAND- NUREG-0800 06 2 2 R4 STANDARD REVIEW PLAN FOH THE REVIEW ARDIZED REGRESSION COEFFICENTS OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER NUREG/CR42F8 FOCAL MECHANISM ANALYSES FOR VIRGINIA PLANTS LWR Edition Revision 4 To %ection 6 2 2 " Containment Heat AND EASTERN TENNESSEE E ARTHOUAkES 0978-1984) Removal System " NUREG/CR4376 HEAT TRANSFER. CARRYOVER AND FALL BACK IN NUREG-0891 R01. CONTAINMENT EMERGENCY SUMP PWR STEAM GENERATORS DURING TRANS!ENTS PERFORMANCE (Technical Findings Related to UnresoNed Safety luues) NUR / 429 DESIGN AND INSTALLATION OF COMPUTER SYS' R EF 0R T F F TEMS TO MEET THE REQUIREMENTS OF to CFR 73 55 NUREG-1079 DAF T FC: ESilMATES OF EARLY CONTAINMENT FROM CORE MELT ACCIDENTS Draft Report for Comment Com uter System NUREG 1116 A REVIEW OF THE CURRENT UNDERSTANDING OF NUREG/CR 4298' DESIGN AND INSTALL ATON OF COMPUTER r"- THE POTENTIAL FOR CONTAINMENT FAILURE FROM IN VESSEL TEMS TO MEET THE REQUIREVEN1S OF 10 CFR 73 55 STE AM EXPLOSIONS NUREG/CR 3647; DESIGN AND F ABRICATION OF A t/8 SCALE Computertzed 8ystem STEEL CONTAINMENT MODEL. NUREG/CR-3307: COMPUTERIZED ANNUNCIATOR SYSTEMS NUREG/CR 3803 THE EFF ECTS OF POST LOCA CONDITIONS ON A Concentration PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-NUREG/CR 4382' CONCENTRATIONS OF URANIUM AND THORIUM DUSTRY NUREG/CR-3616 V02 REACTOR SAF ETV RESE ARCH Ouarterty ISOTOPES IN URANIUM MtLLERS' AND MINERS' TISSUES Report.Apr0 June 1964 Concrete NURLG/CR 3655 CHARACTE nl2ATION OF NUCLEAR REACTOR NUREG/CR 3476 PROBABILITY BASED LOAD COMBINAT!ON CRITE- CONTAINMENT DE NF"aTION FINAL REPORT. RIA FOR DESIGN OF CONGRETE CONTAINMENT STRUCTURES NUREG/CR 3876 PP2 wiLITY HASED LOAD COMBINAflON CRITE. Rf A FOR DESIGN OF SONCRETE CONTAINVENT STRUCTURES Concrete Bunke' NUREG/CR 3952; SEQUOY AH EQUIPMENT HATCH SE AL LE AKAGE. NUREG/C43774 V04. ALTERNATIVE METHOD FOR DISPOSAL OF NUREG!CH 3954 HECTH ANALYSIS OF COUIPME NT TEMPERATURE LOW LEVEL RADIOACTIVE WASTE. Task 2C Technical Requuements RESPONSES TO SELECTED HYDROGEN DURNS aN AN ICE CON-For Earth Mounded Concrete Duriker Disposal Of Low Level Radioac- DENSER CONTAINMENT. tive Waste NUREG/CR 4020 HMSA COMPUTE R PROGRAM FOR TRANSIENT,THREE-DIMENSONAL MIXING GASES Concrete Containment NUREG/CR 4042 A 3 DMENSIONAL COMPUTER MODEL TO SIMO. NUREG/CR-4149' ULTIMATE PRESSURE CAPACITY OF REINFORCED LATE FLUID FLOW AND CONTAINMENT TRANSPOHT THROUGH A AND PRESTRESSED CONCRETE CONTAINMENT. ROCK FRACTURE SYSTEM NUREG/CR-4064 STRUCTURAL RESPONSE OF LARGE PENETRA-TIONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED F G/ 4414 DIRECT-CONTACT CONDENSATON OF STEAM ON TO LOADINGS BEYOND DESIGN HASIS cot 0 W ATER IN STRATIFIED COUNTE RCURRENT FL OW NUREG/CR 4081 ABSORPTON OF GASFOUS LODINE DY WATER NUREG/CH-4416. STABILITY OF STEAM WATER COUNTERCURRENT DROPT E TS STRATIFIED FLOW NUREG/CR 4119 INTEGRITY OF CONTAINMENT PENETRATONS NUREG/CR 4417. LOCAL PROPE RTIE S OF COUNTERCI RRENT UNDER SEVERE ACCIDf NT CONDITIONS FY84 ANNUAL REPORT STRATIFIED STEAM WATER FLOW NUREG/CR 4137 PRETEST PRI DICTIONS FOR THE RE SPONSE OF Cone Penetrat)on test A 18 SCALE STEEL LWR CONTAINVENT BUILDING MODEL TO NUREG/CR 4430. CURRENT METHODOLOGIES FOR ASSESSING ST ATIC OVE RPRE SSURl/ ATION THE POTENTIAL FOR EARTHOUARKE-INDUCED LIQUEFACTION IN NUREG/CR 4t41 CONTAINVENT PURGE AND VENT VALVE TEST PROGnAM FINAL REPORT. SOILS.
l Subject index 155 i I NUREG/CR4143 REVtEW AND EVALUATON OF THE MILLSTONE Control system Fe61ure UNIT 3 PROBABluST C SAFETY STUDY.Contaminent Fadure Modes.Radological Source Terms And ottaite consequences NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR4t$1. INTEGRATION OF EMERGENCY ACTION LEVELS ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC BOfLING WATER REACTOR Appendices. WITri COMBUSTON ENGINEER 6NG EMERGENCY OPERATING PROCEDURES By Use Of Combustion Engmeenng Owners Group NUREG/CR-4326 V02. EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE NUR G CE4 C PA I N LY EDICTIONS AND
^
~~EXPERIMENTAL RESULTS FOR A 16-SCALE STEEL CONTAINMENT Cooient Bottaway MODEL PRESSURIZED TO FAILORE.~~
l' NUREG/CR-3810 V03: REACTOR SAFETY RESEARCH NUREG/CR 4210. MATADOR A COMPUTER CODE FOR THE ANALY. PROGRAMS. Quarterly Report. July September 1984 SIS OF RADtONUCUDE BEHAVOR DURING DEGRADED CORE AC. CIDENTS IN LIGHT WATER REACTORS. i Coolant SO46ewey And Demoge NUREG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF NUREG/CR4318 J V01: REACTOR SAFETY RESEARCH TRANSPORT AND DEPOSITON OF RADIONUCLICES) CODE DE-i SCRIPTON AND USER'S MANUAL PROGRAMS Quarterly Report, January March 1985, NUREG/CR4220- REUABluTY ANALYSIS OF CONTAINMENT ISOLA
Coo 46ng Pond Modeling NU EG/C 4 9 REUABluTY EVALUATON OF CONTAINMENTS IN. NU EG 4120 MATHEMATICAL MODEUNG OF ULTIMATE HEAT l CLUDING SOIL STRUCTURE INTERACTION '
~~j NUREG/CR4198: COST ANALYSIS OF REVISIONS TO to CFR PART~~
; $0, APPENDIX J. LEAK TESTS FOR PRIMARY AND SECONDARY Corb6cule Flumenee CONTAINMENTS OF UGHT WATER COOLED NUCLEAR POWER NUREG/CR4233 OISTRIBUTtON OF CORBICULA FLUMINEA AT NU.
?l , PLANTS CLEAR FACIUTIES. NUREG/CR-4432: 'OMPARISON OF DYNAMIC CHARACTERISTICS Core l OF FUKUSHiMA vuCLEAR POWER PLANT CONTAINMENT ButLO. t ING DETERMINED FROM TESTS AND EARTHOUAKES' NUREG/CR-3485; PRA REVIEW MANUAL NUREG/CR-37$T: TRAN 8 2 THE EFFECT OF LOW STEEL CONTENT i Contaminent NUREG/CR3906- ON FUEL PENETRATION IN A NON-MELTING CYUNDRICAL FLOW 'i URANIUM MILL TAILINGS CHANNEL NEUTRAUZATONCONTAM:NANT COMPLEXATION AND TAluNGS NUREG/CR3889 THE MODEUNG OF BWR CORE MELTOOWN ACCT-4 LEACHING STUDY. OENTS . FOR APPUCATON IN THE MELRPIMOO2 COMPUTER i CODE. Cantemeneted Steel Products j huREG 1103. CONTAMINATED MEXICAN STEELimportaten Of Steet NUREG/CR-3948. EXPERIMENTAL RESULTS OF THE OPERATIONAL into The United States That Had Been inadvertently Contammated TRANSIENT (OPTRAN) TESTS 11 AND 12 IN THE POWER BURST 4~ FACluTY. Wit % Cobalt 60 As A Result Of bcrappng Of A Tesetherapy Unit. NUREG/CR4080 DETERMINATION OF THE AVAILABluTY OF CORE Contamenenon EXIT THERMOCOUPLES DURING SEVERE ACCIDENT SITUATONS. NUREG/C43709. METHOOS OF MINIM 12tNG GROUND-WATER CON. Core Cooling t TAMINATION FROM IN SITU LEACH URANIUM MINtNG Feel Report. NUAEG/CR4060; THE DC-9 AND DC2 DEBRIS COOLABluTY AND 1 NUREG/CR4251 V01: MITIGATIVE TECHNIOUES FOR GROUND-1 MELT DYNAMICS EXPERIMENTS. WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR i ACCIDENTS Volume 1 Analysis Of Genene Site Cordtione. Core Demage j NUREG/CR4251 V02: MITIGATIVE TECHNIQUES FOR GROUND-WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR NUREG 1037 DRFT FC' CONTAINMENT PERFORMANCE WORKING GROUP REPORT Draft Report For Comment. 4 ACCCENTSVolume 2 Case Study Analyses Of Hydrologic Character-
) ' laaton And Mitigative Sr. homes NUREG/CR4050: A REVIEW OF THE SHOREHAM NUCLEAR POWER ) STATON PROBA81USTIC RISK ASSESSMENT. internal Events And Core Damage Frequency NUREG-0975 V03- COMPILATION OF CONTRACT RESEARCH FOR Core Degredetton THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER ) ING TECHNOLOGY. Annual Report For FY 1984. NUREG/CR 41T3. CORSOR USERS MANUAL I NUREG/CR4401: BEHAVOR OF CONTROL ROOS DURING CORE Control Med DEGRADATON- PRESSUR12AfiON OF SILVER INDIUM CAOMlUM
i. CONTROL RODS' NUREG-0090 V07 N04 REPORT TO CONGRESS ON ABNORMAL l OCCURRENCES October-Occomber 1964 j Core D6erupttve Accident NOREG/CR4401: BEHAvlOR OF CONTROL ROOS DURING CORE DEGRADATON PRESSURIZATION OF SILVER-INDIUM-CADMfUM NUREG/CR4240 V01: PHYSICS OF REACTOR SAFETYouarterty CONTROL RODS, Report. January March 1985
. NUREG/CR3804 V03 PHYSICS OF REACTOR SAFETY.Ouarterty Report July September 1964 REG 398h COMPUTERIZED ANNUNCIATOR SYSTEMS NUREG/CR3804 V04 PHYSICS OF REACTOR SAFETV.Ouarterty NU E R 4 91 SURVEY OF UCENSEE CONTROL ROOM HABIT- ' %hC R EXPERIMENTAL INVESTIGATON OF NUREG/CR4280: THE EFFECTS OF SUPERVISOR EXPERIENCE AND FUEL CRUST STABluTY ON MELTING SURFACES OF AN ANNU.
ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW LAR FLOW CHANNEL PERFORMANCE IN CONTROL ROOM SIMULATORS. j Core Mott Control System i NUREG 1079 DAFT FC' ESTIMATES OF EARLY CONTAINMENT FROM NUREG/CR4262 Vot: EFFECTS OF CONTROL SYSTEM FAILURES CORE MELT ACCIDENTS. Dr.n Heport for Comment i ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC NUREGflt6 A REVIEW OF THE CURRENT UNDERSTANDING OF BOIUNG W ATER REACTOR Mac Report.
NUREGmR-4326 V01: EFFECTS OF CONTROL SYSTEM FAILURES THE POTENTIAL FOR CONTAtNMENT FAlLURE FROM IN VESSEL STEAM EXPLOSIONS ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE NUREG/C439t2- MARCH HECTR ANALYSIS OF $[LECTED ACCl-
) PRESSURt2ED WATER RE $CTOR Main Report - FFNTS IN AN ICE CONDENSER CONTAINMENT. ' NUREG/CR4365 EFFECTS OF CONTROL SYSTEM FAILURES IN NUREG/CR4130: ICEDF A CODE FOR AEROSOL PARTK;LE CAP.
TRANSIENTS. ACCIDENTS. AND CORE MELT FREQUENCIES AT A TURE IN ICE COMPARTMENTS WESTINGHOUSE PRESSURIZED WATER REACTOR. NUREG/CR 42$1 V01. MITIGATIVE TECHNiOUES FOR GROUNTb NUHEG/CR4386: EFFECTS OF CONTROL SYSTEM FAILURES ON l TRANSIENTS. ACCIDENTS. AND CORE MELT FREQUENCIES AT A WATER CONTAMINATION ASSOCIATED WITH SEVERE NUCLEAR i ACCIDENTS Volume 1 Analysis Of Genanc Site Conditions BABCOCK AND WfLCOX PRES $URIZED WATER REACTOR NUREG/CR4367. EFFECTS OF CONTHOL SYSTEM FAILURES ON NUREG/CR 4251 V02- MITIGA flVE TECHNIQUES FOR GROUNCA I WATER CONTAM6 NATION ASSOCIATED WITH SEVERE NUCLEAR TRANSIENTS. ACCIDENTS AND CORE MELT FREQUENCIES AT A ACCOENTS Volur,e 2 Case Study Analysse Of Hydrologc Character. { GENERAL ELECTRIC PRES $URIZED WATER REACTOR aration And M>tigetsve Schemet. T
156 Subject index NUREG/CR4385: EFFECTS OF CONTROL SYSTEM FAILURES IN Crack TRANSIENTS. ACCIDENTS, AND CORE-MELT FREQUENCIES AT A NUREG/CR-3228 V03. STRUCTURAL INTEGRITY OF WATER REAC. WESTINGHOUSE PRESSURIZED WATER REACTOR - TOR FRESSURE BOUNDARY COMPONENTS Annual Report For NUREG/CH-4386: EFFECTS OF CONTROL SYSTEM FAILURES ON 1984. TRANSIENTS. ACCIDENTS. AND CORE MELT FREQUENCIES AT A NUREG/CR-3660 V04: PROBAB"JTY OF P PE FAILURE IN THE REAC. BABCOCK AND WILCOx PRESSURIZFD WATER REACTOR- TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTSVolume NOREG/CR4387: EFFECTS OF CONTROL SYSTEM FAILURES ON 4 Poe Failure Induced By Crack Growth in West Coast Plants. TRANSIENTS. ACCIDENTS AND CORE MELT FREQUFNCIES AT A NUREG/CR4082 V01: DEGRADED PIPI: 3 PROGRAM . PHASE H Senwannual Npo@a@ 1984. Sepumpr 1984. NR /CR 44 HE AN C AND MAINTE- " G' "^ CT STEEL TECHNOLOGY PROGRAM NANCE INDUCED FAILURES OF MAIN REACTOR COOLANT PUMP pryE- RP N Y 4 SEALS ON PLANT SAFETY' NUREG/CR4305: COMMENTS ON THE LEAK BEFORE-BREAK CON-CEPT FOR NUCLEAR POWER PLANT PIPING SYSTEMS Core Heflooding SyWom NUREG/CR4277; INVERTED ANNUAL FLOW EXPERIMENTAL STUDY' CM M l Corporate Data postwork NUREG/CR4015. EFFECT OF STAINLESS STEEL WELD OVERLAY i l NUREG/CR4322 V01: CORPORATE DATA NETWORK (CDN) DATA CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL REQUIREMENTS TASK Voi t Entarpnse Modes PLATES IN BENDING SERfES 1. NUREG/CR-4322 V02: CORPOf1ATL DATA NETWORK (CDN) DATA NUREG/CR-4106. PRESSURIZED THERMAL SHOCK TEST OF 6-IN. REQUIREMENTS TASK Vol 2' Data Entity Dictionary THICK PRESSURE VESSELS PTSE l. investigation Of Warm Prestress-NUREG/CR4327 VPJ: CORPORATE DATA NETWORK (CDN) DATA ing And Upper Shett Arrest. REOutREMENTS TASK Vol 3- Data Modet NUREG/CR 4322 V04 CORPORATE DATA NETWORK (CDN) DATA Crack Growtfi REQUIREMENTS TASK.Vol 4. Preliminary Strategic Data Plan. NUREG/CR 3744 V02; HEAVY SECTON STEEL TECHNOLOGY PRO. GRAM SEMIANNUAL PROGRESS REPORT FOR APRIL SEPTEMBER Cem Maintenance NU EG/CR 377; EVALUATIONS AND UTlu2ATIONS OF RISK iM- NUREG/CR 3998 V03 LGHT WATER REACTOR SAFETY VATERIALS ENGINEERING RESEARCH PROGRAMS ouarteily Progress ReportOctoter-December 1984. Corroedon NUREG-1095: EVALUATION OF RESPONSES TO IE BULLETIN 82- NUREGICR4219 V01: HEAVY SECTON STEEL TECHNOLOGY PRO-02.Degradaten Of Threaded Fasteners in Reactor Coolant Pressure GRAM SEMIANNUAL PROGRESS REPORT FOR OCTOBER 1984 - Boundary Of Pressunred Water-Reactor Piants MARCH 1985 NUREG 1155 V02. RESE ARCH PROGHAM PLAN Steam Generators NUREG/CR4287: ENVIRONMENTALLY ASSISTED CRACKING IN NUREG/CR 4134. REPOSITORY E NVIRONMENTAL PARAMETERS LIGHT WAttR REACTORS Annual Report. October 1983. September RELEVANT TO ASSESSING THE PERFOnMANCE OF HIGH LEVEL 1984 WASTE PACKAGES, Crew Performance Corrosion Ase6sted Fougue NUREG/CR4280 THE EFFECTS OF SUPERVISOR EXPERIENCE AND
^
~~y y yy[CHH O~~
~~^ 0a or Vessets' p~~
ASSISTANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CREW PERFORMANCE IN CONTROL ROOM SIMULATORS. Corrosion Fatigue Crew Tre6n6ng WUREG/CR 4422: A REVIEW OF THE MODELS AND MECHANISMS NUREG/CR4258' AN APPROACH TO TEAM SKILLS TRAINING OF NU-FOR ENVIRONMENTALLY ASSISTED CRACK GROWTH OF PRES. CLEAR POWER PLANT CCNTROL ROOM CREWS. SURE VESSEL AND PIPING STEELS IN PWR ENVIRONMENTS. Coet Crttwal Flow NURFG/CR 3293 V01: TECHNOLOGY. SAFETY AND COSTS OF DE- NUREG/CR-3866. TRAC-PD21NDEPENDENT ASSESSMENT. COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON. Cruet Stalwitty FUEL CYCLE FACluTIES FOLLOWING POSTULATED ACCIDENTS Masn Report NUREG/CR-3944 TRAN 0-3 EXPERIMENTAL INVESTIGATION OF
~~' NUREG/CR-3293 V02. TECHNOLOGY. SAFETY AND COSTS OF DE- FUEL CRUST STADIUTY ON MELTING SURFACES OF AN ANNU.~~
COMMISSIONING REFERENCE FUEL CYCLE AND NON-FUEL LAR FLOW CHANNEL. CYCLE FACIUTIES FOLLOWING POSTULATED D10 Esperiment ACCIDENTS Appendices. NUREG/CR409u EVALUATION OF NUCLEAR FACILITY DECOMMIS- NUREG/CR4055 THE Oto EXPERIMENT COOLABILITY OF UO2 SiONING PROJECTS Annual Summary Report Fiscal Year 1964 DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL NUREG/CR-4268 RATIO METHODS FOR COST-EFFECTIVE FIELD SAMPLING OF COMMERCIAL RADIOACTIVE LOW-LEVEL WASTES DEGil NUREG/CR4373' COMPENDIUM OF COSI EFFECTIVENESS EV ALVA- NUREG/CR-3660 V03. PROBABILITY OF PIPE FAILURE IN THE REAC. TIONS OF MODIFICATIONS FOR DOSE REDUCTION AT NUCLEAR TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS Volume
~~' NU E / R 3 8 COST ANALYSIS OF REVISIONS 70 to CFR PART NUREO/ R 3 P B OF P P Al RE IN THE REAC.~~
50. APPENDIX J. LEAK TESTS FOR PRIMARY AND. SECONDARY TOR COOLANT LOOPS OF COMBUSTION ENGINEERING PWR CONTAINMENTS OF UGHT WATER COOLED NUCLEAR POWER PLANTSVolume t.Summarypeport PLANTS NUREG/CR4290 V02. PROBABILITY OF PIPE FAILURE IN THE REAC.
NUREG/CR4398 COST ANALYSIS OF REVISCNS TO to CFR PART TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR 50.APPENDtX J. LEAK 1ESTS FOR PRIMARY AND SECONDARY PLANTSVolume 2 Guellotine Breah Indirectty induced By Earthquakes. CONTAINMENTS OF UGHT WATER COOLED NUCLEAR POWER PLANTS- Damage Control NUREG/CR4392. MEASURES OF SAFEGUARDS RISK EMPLUv4NG PRA (MOSREP) A Methodology For Estimating Rah impacts Of Safe. R G/ 15 COUNTER CURRENT STEAM / WATER FLOW guards Measures ABOVE A PERFORATED PLATE VERTICAL INJECTION OF WATER. Damage Definet6on Countercurrent Flow NUREG/CR 4t t2 V01. INVESTIGATION OF CADLE AND CADLE NUREG/CR4414 DIRECT <,0NTACT CONOENSAflON OF STEAM ON SYSTEM FIRE TEST PARAMETERS Task A lEEE Flame Test. COLD WATER IN STRATIFIED COUNil RCURRENT FLOW NUREG/CR-4417. LOCAL PROPLFlflES OF COUNT ERCURRENT Damage MmgeHon STRATIFIED STEAM WATER FLOW NUREG 1155 V03 RESEARCH PROGRAM PLAN Piping C r Fi n z Modet Damper Response Time NUREG/CP-0062: PROCEEDINGS OF THE CONFERENCE ON THE AP-PLICATION OF GEOCHEMICAL MODELS TO HIGH LEVEL NUCLEAR NUREG/CR4232- THE RESPONSE OF VENTILATION DAMPERS TO WASTE REPOSITORY ASSESSMENT. LARGE A!RFLOW PULSES
Subject index 157 Data NUREG/CR 3293 V01: TECHNOLOGY, SAFETY AND COSTS OF DE. NUREGICR 4182. VERIFICATION OF SOIL STRUCTURE INTERACTION COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON. METHODS. FUEL CYCLE FACILITIES FOLLOWING POSTULATED Data Analysis ACCIDENTS Main Report. NUREG/CR-4138: DATA ANALYSES FOR NEVADA TEST SITE (NTS) NUREG/CR-3293 V02: TECHNOLOGY. SAFETY AND COSTS OF DE-PREMIXED COMBUSTION TESTS- COMMISSIONING REFERENCE FUEL CYCLE AND NON FUEL CYCLE FACILITIES FOLLOWING POSTULATED Data Bank ACCIDENTS Appendices. NUREG/CR-2531 R03. INTRODUCTORY USER'S MANUAL FOR THE NUREG/CR 4118- MONITORING METHODS FOR DETERMINATION U S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE. COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT SEARCH DATA BANK URANIUM RECOVERY SITES NUREG/CR 4009 Pi) MAN RELIABILITY DATA BANK Evaluation Re- NUREG/CA-4090. EVALUATION OF NUCLEAR FACILITY DECOMMIS. suits $10NING PROJECTS. Annual Summary Report Fiscal Year 1984. Data Base Decontamination NUREG 1148 NUCLEAR POWER PLANT FIRE PROTECTION RE. NUREG 1153: INSPECTION REPORT OF UNAUTHORIZED POSSES-SEARCH PROGRAM SiON AND USE OF UNSEALED AMERICIUM 241 AND SUBSEQUENT NUREG/CR 3319 LWR PRESSURE VESSEL SURVEILLANCE DOSIME- CONFISCATION. J C Haynes Company. Newark. Ohio. TRY IMPROVEMENT PROGRAM LWR Power Reactor Surveillance NUREG/CR 3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL Physics-Dosemetry Data Base Compendium ACCIDENTS METHODS, COSTS AND SCHEDULES FOR DECON-NUREG/CR-3413: OFF-SITE CONSEQUENCES OF RADIOLOGICAL T AMINATION ACCIDENTS METHODS, CCSIS AND SCHEDULES FOR DECON-TAMINATION NUREG/CR-3444 V02: THE IMPACT OF LWR DECONTAMiNATIONS NUREG/CR 3905 V01 Rt: SEQUENCE CODING AND SEARCH ON SOLIDIFICATION WASTE DISPOSAL AND ASSOCIATED OCCU-PATIONAL EXPOSURE. SYSTEM FOR UCENSEE EVENT REPORTS User % Guide NUREG/CR-3905 V02. SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-4399. POSSIBLE OPTIONS 'OR REDUCING OCCUPA-FOR LICENSEE EVENT REPORTS Code Listings TIONAL DOSE FROM THE TMI-2 BASLMENT. NUREG/CR-3905 V03. SEQUENCE CODING AND SEARCH SYSTEM Defect NU EG/ 95 . EQU E N'O N SEARCH SYSTEM NUREG 1144 NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-FOR LICENSEE EVENT REPORTS Coder's Manual. #" Data Development Degradation NUREG/CR-4350 V06. PROBABILISTIC RISK ASSESSMENT COURSE NUREG 1155 V02: RESEARCH PROGRAM PLAN Steam Generators DOCUMENTATION Volume 6 Data Development NUREG/CR 4067.
~~SUMMARY~~
OF BARRIER DEGRADATION EVENTS Data Entity Dictionary AND SMALL ACUDENTS IN U.S. COMMERCIAL NUCLEAR POWER PLANTS NUREG/CR-4322 V02: CORPORATE DATA NETWORK (CDN) DATA NUREG/CR-4156. OPERATING EXPERIENCE AND AGING-SEISMIC REQUIREMENTS TASK.Vol 2: Data Entity Dictsonary ASSESSMENT OF ELECTRIC MOTORS Dato Model NUREG/CR-4256. MEASUREMENT OF RESPONSE TIME AND DETEC-RUREG/CR-4322 V03 CORPORATE DATA NETWORK (CON) DATA TION OF DEGRADATION IN PRESSURE SENSOR / SENSING UNE SYSTE MS. REQUIREMENTS TASK.Vol 3. Data Modet. Degradt1 Core UE /C 4 22 V01. CORPORATE DATA NETWORK (CDN) DATA DENTS IN A E CON ENS R C REOurREMENTS TASK Vol 1- Enterpnse Modal N ENT RUREG/CR-4322 V02. CORPORATE UATA NETWORK (CDN) DATA Domineralisation System NURE / 4 22 V C TE A ORK (CDN) DATA NUREG/CR 4150: EPICOR il RESIN DEGRADATION RESULTS FROM REQUIREMENTS TASK Vol 3 Data Model FIRST RESIN SAMPLES OF PF 8 AND PF-20 NUREG/CH 4322 V04 CORPORATE DATA NETWORK (CDN) DATA REQUIREMENTS TASK.Vol 4 Prelemenary Strateg#c Data Plan. " ' g 4358. APPUCATIONS OF DENSITY PROFILING TO EQUIP. Debris Bed MENT OVAUFICATION ISSUES NUREG/CH-4060- THE DC 1 AND DC-2 DEBRIS COOLABILITY AND MELT DYNAMtCS EXPER!MENTS, DensHy Wave Debris CooHng NUREG/CR-4116- NUFEGO NP.A DIGITAL COMPUTER CODE FOR THE UNEAR STABluTY ANALYSIS OF 801UNG WATER NUCLEAR NUREG/CR 4055 THE D10 EXPERIMENT COOLA8tuTY OF 002 REACTORS DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL. Deposition Decay Heat NOHEG/CR 2951: THE D9 EXPERIMENT Heat Removal From Stratified NUREG/CR-3384 BIOLOGICAL CHARACTERilATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOR INHALED URANIUM MILL-RU EG C 041: SYSTEM ANALYSIS HAND 000k * ' NUREG/CR-4055. THE D10 EXPERIMENT COOLABluTY OF UO2 Deposition Modeling Nl f E 4 9 AN PR CH TO F TP A ONUCUDE NUREG/CR 4tS8 A COMPILATION OF INFOHMATION ON UNCER. DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM TAINTIES INVOLVED IN DEPOS4flON MODELING Doctelon Aid DepotlHon Velooty M sREG/CR-4272. RESPONSE TREE EVALUATION EXPERIMENTAL NUREG/Ch 4157. A SCIENTIFIC CRifiOUE OF AVAILADLE MODELS ASSEFSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTCn FOH REAL TIME SIMULATIONS OF DISPERSION Design Accident Decisionmak6pg NUREG/CR 3M2. SEQUOYAH EQUIPMENT MATCH SEAL LEAKAGE. NOREG/CR 4040 OPERATIONAL DECISIONMAkiNG AND ACTION SE. LECTION UNDER PSYCHOLOG> CAL STRESS IN NUCLEAR PuwER Design Basis Event PLANTS, N IREG/CR 36/iG V01 PROHABILITY OF PIPE FAILURE IN THE RE AC-Decommissioning TOR COOLANT LOOPS OF WESTINGHOUSE PWR PLANTS Volume i Summary Report WUREG/CR 1755 ADD 01: TECHNOLOGr. SAFETY AND COSTS OF DE-COMMISSIONING NUCLEAR REACTORS AT MULTIPLE RCACTOR Dewatering STAflONS Effects On Decommiswoneng of intenm inatnisty To Dispose NUREG/CR 4192 THE AP ALYSIS OF DRAINAGE AND CONSOUDA-Of Wastes Offsate TION AT TYPICAL URANIUM MILL TAILINGS SITES
158 Subject index Diesel Generator NUREG 0910 ROI S02: NRC COMPREHENSIVE RECORDS DISPOSI-NUREG/C43831: THE IN-PLANT RELIABILITY DATA BASE FOR NU- TION SCHEDULE. CLEAR PLANT COMPONENTS Intenm Report - Diesel NUREG 0910 R01 S03 NRC COMPREHENSIVE RECORDS DISPOSI-Generators. Batteries, Chargers And inverters TON SCHEDULE. NUREG/CR4347. EMERGENCY DIESEL GENERATOR OPERATING NURFG-0910 R01 SO4- NRC COMPREHENSIVE RECORDS DISPOSI-EXPERIENCE 1981-1983 TlON SCHEDULE. NUREG/CR 4440: A REVIEW OF EMERGENCY DIESEL GENERATOR PERFORMANCE AT NUCLEAR POWER PLANTS. Distribution Coefficient NUREG/CR4094. FIELD EXPERIMENT DETERMINATONS OF DISTRI. Differentist integral Conservation Equation BUTION COEFFICIENTS OF ACTINIDE ELEMENTS IN SULFATE NUREG/CR-3989. TIME AND VOLUME-AVERAGED CONSERVATION LAKE ENVIRONMENTS. EQUATIONS FOR MULTIPHASE FLOW Part One System Without in-tornal Sohd Structures. Dose NUREG/CP4066. PROCEEDINGS OF AN INTERNATIONAL WORK-Diffusion SHOP ON HISTORIC DOSE EXPERIENCE AND DOSE REDUCTON NUREG/CR-4072: THE ESTIMATON OF ATMOSPHERIC DISPERSION (ALARA) AT NUCLEAR POWER PLANTS MAY 29-JUNE 1,1984 AT NUCLEAR POWER PLANTS UTill2iNG REAL TIME ANEMOME- NUREG/CR 2850 V03. POPULATION DOSE COMMITMENTS DUE TO TER STATISTICS. RADIOACTIVE RELEASES FROM NUCLEAR POWER PLANT SITES IN 1981 NUREG/CR3469 V02: OCCUPATIONAL DOSE REDUCTION AT NU-NU /CR 3638. HYDROGEN-STEAM JET FLAME FACILITY AND EX- EAR WR MTS. AnnotaW B@aography G Sewed Rea$ PERIMENTS' ings in Radiation Protection And ALARA D6 gests And Indones NUREG/C43984 BIOLOGICAL CHARACTER 12ATION OF RADIATION WUREG 0750 V20101: INDEXES TO NUCLEAR REGULATORY COM. EXPOSURE AND DOSE ESilMATES FOR INHALED URANIUM MtLL-ING EFFLUENTS Annual Progress Report.Apnf 1983 March 1984 MISSON ISSUANCES FOR JULY-SEPTEMBER 1984 NUREG-0750 V20102: INDEXES TO NUCLEAR REGULATORY COM. NUREG/CR 4068
~~SUMMARY~~
OF HISTORICAL EXPERIENCE WITH RE-MfSS10N ISSUANCES FOR JULY-DECEMBER 1984 LEASES OF RADIOACTIVE MATERIALS FROM COMMERCIAL NU-NUREG 0750 V22 601: INDEXES TO NUCLEAR REGULATORY COM- CLEAR POWER PLANTS IN THE UNITED STATES MISSION !SSUANCES. July-September 1985 NUREG/CR4159 COMPARISON OF THE 1981 INEL DISPERSION DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS Diode Radiation Detector NUREG/CR 4160: HISTORICAL
~~SUMMARY~~
OF OCCUPATIONAL RADI-NUREG/CR4131. INVESTIGATON OF ALTERNATIVE MEANS TO AC- ATION EXPOSURE EXPERIENCE IN U S. COMMERCIAL NUCLEAR l COMPLISH THE GOALS OF BIENNIAL ON CHAM 8ER CALIBRO POWER PLANTS.
~~TON. NUREG/CR4203. A CALCULATONAL METHOD FOR DETERMINING~~
' BIOLOG3 CAL DOSE RATES FROM IRRADIATED RESEARCH REAG. Directory Of Certificate Of Compl6ance TOR FUEL. NUREG 0383 V01 R08. DIRECTORY OF CERTIFICATES OF COMPLI- NUREG/CR4239 ANALYSIS OF THE ADILITY OF CURRENT HEALTH ANCE FOR RADIOACTIVE MATERIALS PACKAGES Summary Report PHYSICS INSTRUMENTS TO PREDICT DOSE IN EXPOSED INDfVID-1 Of NRC Approved Packages UALS NUREG 0383 V02 R08. DiHECTORY OF CERTIFICATES OF COMPLi* NUREG/CR4266 STANDARD BETA PARTICLE AND MONOENERGE-ANCE FOR RADOACTIVE MATERIALS PACKAGESCertificates of Tic ELECTRON SOURCES FOR THE CAllBRATION OF BETA-RADI. Comphance ATON PROTECTION INSTRUMENTAflON NUREG-0383 V03 ROS- DIRECTORY OF CERTIFICATES OF COMPLi- NUREG/CH 4373 COMPENDIUM OF COST EFFECTIVENESS EVALUA-ANCE FOR RADIOACTIVE MATERIALS PACKAGES Summary Report TiONS OF MODIFICATIONS FOR DOSE REDUCTION AT NUCLEAR Of NRC Approved Quakty Assurance Programs For Radioactive Maten- POWER PLANTS, alPackages. I Dose Reduction I D6screte Element NUREG/CR4254 OCCUPATIONAL DOSE REDUCTION AND ALARA NUREG/CR-3442: RADTWO A COMPUTER CODE FOR SIMt" ATING AT NUCLEAR POWER PLANTS Study On High-Dose Jobs.Radwaste FAST. TRANSIENT. TWO-DIMENSIONAL,TWO-LAYER RADONU* Handlin0 And ALARA Incentives i CLIDE CONCENTRATION CONDITIONS IN LAKES. RESERVOIRS. RIVERS. ESTUARIES.AND COASTAL REGIONS Desimeter NUREG/CR4239 ANALYSIS OF THE ABILITY OF CURRENT HEALTH CR 1 7. A SCIENTIFIC CRITIQUE OF AVAILABLE MODELS h[3 F' NU EG d A OF TH 1 INEL DISPERSION U PI A be ' l DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS Dosometry NUREG/CR 3319 LWR PRESSURE VESSEL SURVEILLANCE DOSiME-NR 1048 DISPOSAL OF HIGH-LEVEL RADIOACllVE W/ 3TES IN TRY IMPROVEMENT PROGRAM LWR Power Reactor Surveillance UNSATURATED 20NE: TECHNICAL CONSOERATONS AND RE- Ph secs-Oowmetry Data Base Comperusium NUR G/CR3746 V02 LWH PRESSURE VESSEL SURVEILLANCE DO-NUR 64 IN ORMAT N ON THE CONFINEMENT CAPABILITY OF SIMETRY IMPROVEMENT PROGRAM Semiannual Progress THE FACillTY DISPOSAL AREA AT WEST VALLEY.NEW YORK Report >pnl 1984 September 1984 NUREG/CR-1755 ADOO1; TECHNOLOGY,SM ETY AND COSTS OF DE. NUREG/C43746 V03 LWH PRESSURE VESSEL SURVEILLANCE DO-COMMISSIONING NUCLEAR REACTORS AT MULTIPLE-REACTOR S: METRY IMPROVEMENT PROGRAM 1984 Annual Report. October STATIONS Effects On Decommesseorwng Of interim inatahry to Dispose 1,1983 September 30,1984 NUREG/CR4131: INVESTIGATION OF ALTERNATIVE MEANS TO AC-Nt EG/ 406 LEACHATE PLUME MIGRATION DOWNGRADIENT COMPLlSH THE GOALS OF BIENNIAL ION CHAV0ER CALIDRA-FROM URAN'UM TAILINGS DISPOSAL IN MINE STOPES NUREG/CR4351 SvGGESTED STATE REOuiHEMENTS AND CRITF. TON NUREG/CR4t85 AN ASSESSMENT OF DOSIMETRY DATA FOR AC-RfA FOR A LOW LEVEL RADIOACTIVE WASTE DISPOSAL SITE l REGULATORY PROGRAM, CIDENTAL RADIONUCLIDE RELEASES FROM NUCLEAR REAC-l TORS Disposal Site NUREG/CR4069 ANALYSES OF SOILS FROM AN AREA ADJACENT Double Ended Guillotine 8teek TO THE LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE AT NUREG/CR 3660 V03 PROBABILITY OF PIPE FArLURE IN THE RE AC-SHEFFIELD itLINOIS TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTSvolume NUREG/CR 4076 DETERMINATON OF COMPLIANCE WfTH CRITERIA 3 Gudiotine Break Irwierectly Inrfuced By Eartrwavahes l FOR FINAL TAILINGS DISPOSAL SITE RECLAMATION NUREG/CR 3M3 V01 PROBAHILtTV OF PIPE F AILURE IN THE REAC-I TOR COOLANT LOOPS OF COMHUSTION ENGINEERING PWR i Disposition Schedule PLANTS Volume 1 Summary Report l NURFG4910 H01 S0t; NRC COMPREHENSrVE RECORDS DISPOSI- NUREG/CR 3M3 V03 FROOABILITY OF PIPE F ALLURE IN THE REAC-TION SCHEDULE TOR COOLANT LOOPS OF COM8USTON ENGINEERING PWR l l I
f, I t Subject index 159 PLANTS. Volume 3. Double Ended Gudlotine Break Indrectty induced By Earthquehe Eanhquakes. NUREG/CR-4290 VC2: PROBABILITY OF PtPE FAILURE IN THE REAC- NUREG 1061 V02. REPORT OF THE U S. NUCLEAR REGULATORY TOR COOLANT LOOPS OF BABCOCK AND WILCOX PWR COMMISSION PIPING REVIEW COMMJTTEE. Volume 2 Evaluation Of Sestmc Desagne - A Review Of Seisme Desgn Recurements For Nu-PLANTS. volume 2Gudlotine Break Indrectfy Induced By Earthquakes. clear Power Plant Pong Downcomer Pressure NUREG-1061 V02 ADD REPORT OF THE U S. NUCLEAR REGULA-NUREG/CR-3703: ASSESSMENT OF SELECTED TRAC AND RELAPS TORY COMMISSON PIPtNG REVIEW COMMITTEE Volume 2 CALCULATONS FOR OCONEE 1 PRESSURt2ED THERMAL SHOCK Addendum Sumrnary And Evaluahon Of Histoncal StrongMoton Earth-STUDY. quake Seesme Response And Damage To Aboveground Irwstnas Piping NUREG/CP4XI70- PROCEEDtNGS OF THE WORKSHOR ON SEISMIC
^
UE CR 05 TH D10 EXPER MENTCOOLABIUTY OF UO2 T DEBRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL NUREG/CR-3178: STRUCTURAL AND TECTONIC STUDES IN NEW YORK STATE Final Report. July 1981. June 1982. . WREG/CR-4192 THE ANALYSIS OF DRAINAGE AND CONSOUDA- NUREG/CR-3660 V03 PROBABruTY OF PIPE FAILURE IN THE REAC.. ' TON AT TYPtCAL URANfUM MILL TAluNGS SITES' TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTSVoaume 3 Guirlotine Break fndrectfy induced By Earthquakes DrMe NUAEG/CR 3805 V02- ENGINEERING CHARACTERIZATION OF RUREG/CR4258: AN APPROACH TO TEAM SKILLS TRAINING OF NU. GROUND MOTON Task & Enocts Of Ground Moten Charactenstics CLEAR POWER PLANT CONTROL ROOM CREWS. On Simchaaf Response Consideng LocaAzed StructLaal Nonlineanties And Sod Structure Interacten Enocts. i Drop-54:e Distreutson NUREG/CR 3876 PROBABluTY BASED LOAD COMBINATON CRITE- I RUREG/CR4424: DROPLET SIZES. DYNAMICS AND DEPOSITION IN Rf A FOR DESIGN OF CONCRETE CONTAINMENT STRUCTURES VERTICAL ANNULAR FLOW. NUREG/CR4117 FAULTING AND JOINTING IN AND NEAR SURFACE M:NES OF SOUTHWESTERN INDIANA. Dry Air Storage
~~'JUREG/CR4 t 45: EARTHOUAKE RECURRENCE INTERVALS AT NU.~~
~~NUREG/CR4074: THE PERFORMANCE OF DEFECTED SPENT LWR CLEAR POWER PLANTS. ANALYSIS AND RANKING.~~
~~] FUEL RODS IN INERT GAS AND DRY AIR STORAGE ATMOS- NUREG/CR4226. NEW MADRID SEISMOTECTONC STUDY Actmties 1~~
~~PHERES' Dunng Fiscal Year 1983~~
+
Dry Depoeltlon NUREG/CR4290 V02: PROBABluTY OF P!PE FAILURE IN THE REAC-l TOR COOLANT- LOOPS OF BABCOCK AND WILCOx PWR NUREG/CR-4157: A SCIENTIFC CRITOUE OF AVAILABLE MODELS PLANTS Volume 2 Gudiotine Break Indrectly induced By Earthquakeit FOR REAL TIME S:MULATONS OF OtSPERSICN. NUREG/CR4329. REUABsUTY EVALUATON OF CONTAJNMENTS IN-NUREG/CR-4158. A COMPILATON OF INFORMATON ON UNCER- CLUDNG SOIL-STRUCTURE INTERACTION 1 TAINT:ES INVOLVED IN DEPOSITON MODEUNG NUREJ/CR4334 AN APPROACH TO THE QUANTIFICATION OF SEIS-Dry Spent Fuel Storage MtC MARG 4NS IN NUCLEAR POWER PLANTS l NUREG/CR4339 A REVIEW OF RECENT RESEARCH ON THE SEIS-NUREG/CR4084: DAY SPENT FUEL STORAGE TEST PLAN FOR DE- MOTECTONOS OF THE SOUTHEASTERN SEABOARD AND AN { STRUCTIVE FUEL ROD EXAMINATONS. EVALUATON OF HYPOTHESES ON THE SOURCE OF THE 1886
~~} , Dry Storage CHARLESTON. SOUTH CAROUNA EARTHOUAKE.~~
~~WREG/CR4345: INVESTIGATION OF THE STABluTY OF LWR NUREG/CR4354 A STUDY OF SEISMICITY AND TECTONCS IN NEW j~~
~~; ENGLAND Final Report SPENT FUEL RODS BELOW 250 C.~~
NUREG/CR4430: CURRENT METHOOOLOGIES FOR ASSESSJNG ! Dynernic Analysis THE POTENTIAL FOR EARTHOUARnE INDUCED LOUEFACTION IN ! So,t$ NUREG/CR 1677 V02 PIPtNG BENCHMARM PROBLEMS. VOLUME D ~ DYNAMC ANALYS4S INDEPENDENT SUPPORT VOTON RE- NUREG/CR4432: COMPARISON OF DYNAMO CHARACTERISTICS i SPONSE SPECTRUM METHOO- OF FUKUSHIMA NUCLEAR POWER PLANT CONTAINMENT ButLD- ~ 1 ING DETERMINED FROM TESTS AND EARTHOUAKES. Dynam6c Soil Test 1 4 Econometric Model NUREG/CR4430- CURRENT METHODOLOG:ES FOR ASSESS 4NG THE POTENTIAL FOR EARTHOUARKE4NDUCED LOUEFACTON IN NUREG/CR4373 COMPENDf0M OF COST EFFECTIVENESS EVALUA-j SOILS- TONS OF MODIFICATONS FOR DOSE REDUCTION AT NUCLEAR POWER PLANTS. I E MUREG/CR4375: THEORY, DES 4GN.AND OPERATON OF UOutD Eddy Current
]
METAL FAST BREEDER REACTORS.lNCLUDNG OPERATONAL NUREG-09/5 V03. COMPtLATON OF CONTRACT RESEARCH FOR HEALTH PHYSICS ~ THE MATERIALS ENGINEERrNG BRANCH.DtVISON OF ENGINEER-ING TECHNOLOGY Annual Report For FY 1984 ECCS NUREG-1155 V02- RESEARCH PHOGRAM PLAN Steam Generators NUREG/CR-4277; INVERTED ANNUAL FLOW EXPERIMENTAL STUOY. E%55 V04 RESEARCH PROGRAM RANDesimche Eo amination EOCMEC NUREG/CR-3949 V01: EDOYCURRENT INSPECTION FOR STEAM
,I NUREG/CR4288: FOCAL MECHANISM ANALYSES FOR VIRGINTA GENERATOR TUBtNG PROGRAM Semiannual Progress Report For AND EASTERN TENNESSEE EARTHOUAKES (1978-1964) June g hndi V2 EDDYCURRENT INSPECTON FOR STEAM EPIC 04-il GENERATOR TUBING PROGRAM Annual Progress Report For Penod NUREG/CR-4150- EPICOR-il RESAN DEGRADATON RESULTS FROM Ending December 38,1964.
FIRST REStN SAMPLES OF PF 8 AND PF 20. Educational Ouellficat6on Egp NUREG/CR-4051: ASSESSMENT OF JOO-RELATED EDLCATIONAL i NUREG/CR-4130: ICEDF:A CODE FOR AEROSOL PARTICLE CAP- OVAUFICATIONS FOR NUCLEAR POWER PLANT OPERATORS. t TURE IN ICE COMPARTMENTS. Earth Mounded Coricrete Sunker D6spoest . NUREG/CR-3805 V02. ENGINEERING CHARACTERl2ATON OF NUREG/CR-3774 V04. ALTERNATIVE METHOD FOR DSPOSAL OF GROUND MOTION Task it. Effects Of Ground Motion Charactenstics LOW LEVEL RADCACTIVE WASTE. Task 2C.Technscal Requrements On Structural Response Consadonne Locanted Structural Nonhneante For Earth Mounded Concrete Bunker Dsposar Of Low Level Radoac. And Sod-Structure interacton Effecta , trve Waste. I
; Earthen Redon Suppression Cover NUREG4017 ROI: CALCULATON OF RELEASES OF RADOACTIVE WREG/CR-3752. EFFECTS OF HYDROLOGIC VARIABLES ON ROCK MATERIALS IN GASEOUS AND UQUID EFFlutNTS FROM PRES-RIPRAP DESIGN FOR UAANtUM TAIUNGS IMPOUNOMENTS. SUAllED WATER REACTORS (PWR GALE CODE)
J
' 160 Subject Inder NUREG/CR4245; IN-PLANT SOURCE TERM MEASUREMENTS AT Emergency Sump Performance BRUNSW1CK STEAM ELECTRC STATON NUREG-0897 R0t: CONTAINVENT EMERGENCY SUMP RUREG/CR4397; IN-PLANT SOURCE TERM MEASUREMENTS AT PERFORMANCE (Techrncal Fru$ngs Related To Unresolved Safety
, PRAIRIE ISLAND NUCLEAR GENERATING STATION. Issues) Elastic Unicodeng Em#esson Control NUREG/CR4283 STUDY OF THE EFFECTS OF ELASTC UNLOAD- NUREG/CRat76 EMISSION CONTROL TECHNOLOGY AND QUAUTY INGS ON THE #R CURVES FROM COMPACT SPECIMENS. ASSURANCE NEEDS AT URANIUM MiLUNG FACruTIESinc6 des Elastic-Plastte Fracture Mechanece Supportng Methods For Testmg. Operating And Mainta, rung Aa Pollu-NUREG 1155 V03; RESEARCH PROGRAM PLAN P tion Contros Devces NUREG/CR-4082 V02. DEGRADED PtPING P OuRAM - PHASE 11 Sertsannual Report October 1984 - March 1985. Ermastons 7 NUREG/CR 4100 EVALUATON OF INSTRUMENTAL METHOOS FOR 8 a G 4156 OPERATING EXPERIENCE AND AGING-SEISMIC ASSESSMENT OF ELECTRC MOTORS Energetsc Core Deeruptive Acc6 dent i NUREG/CR4346 AEROSOL RELEASE EXPERIMENTS IN THE FUEL Electric Motor-Operated Vh AEROSOL SIMULANT TEST F ACtuTY UNDERSOOtVM E XPE RL NUREG/CR4234 V01: AGING AND SERVICE WEAR OF ELECTRIC MENTS. MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY-FEA. TURE SYSTEMS OF NUCLEAR POWER PLANTS Energy Transport Doctrical Systent Failure NUREG/CR4203 A CALCULATONAL METHOO FOR DETERM:NhG NUREG/CR-3991: FAILURE MODES AND EFFECTS ANALYSIS (FMEA) BIOLOGICAL OOSE RATES FROM IRRA01ATED RESEARCH RE AC. OF THE CS/NNI ELECTRIC POWER DISTRIBUTON CIRCUlTRY AT TOR FUEL. THE OCONEE 1 NUCLEAR PLANT. Enforcement Actions Doctronic Module Aging NUREG4940 V03 N04 EN80RCEMENT ACTCNS S GN FICANT AC
, NUREG/CR-3863: ASSESSMENT OF CLASS 1E PRESSURE TRANS. TONS RESOLVED Ouarter'y Progress Report.0ctoeer~0ecemcor MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONVENT $984 SCREENING TESTS. NUREG 0940 V04 Not ENFOACEVENT ACTONS S6GN1FICANT AC-TIONS RESOLVED Quarterty Pr ess ReportJanuary-Marcn 1985
~~W NO2 EWC W ACW SMOT &~~
3228 V03. STRUCTURAL IPTEGRiTY OF WATER READ V ' ^ TOR PRESSURE BOUNDARY COMPONENTS Annual Report For pg 3 (pO T T10NS ANT AC-TtONS RESOLVED ouarterty Progress Report.uy-September 1985 NUREG/CR-3998 V02. UGHT-WATER-REACTOR SAFETY MATERIALS ENGINEERING RESEARCH PROGRAMS Quartery Progress Eywred DepoW i NUREG/CR-3774 V02 ALTL'RNATNE METHOOS FOR OiSPOSAL 0F NUR UGHT WATER REACTOR SAFETY MATERIALS LOW LEVEL RADOACTIVE WASTES Tap 2A TecFwucal Requaements ENGINEERING RESEARCH PROGRAMS Quarterfy Progress For Bedo* ground Vault Dsposal Of Low Leve8 Radioactrve Waste pepo,t. October-Oecember 1984 NUREG/CR-J774 V03 ALTERNATNE METHOOS FOR QSPOSAL OF NUREG/CR-4204 LONG-TERM EMBRITTLEVENT OF CAST DUPLEX LOW LEVEL RADICACTNE W ASTES Task 2B Technical Requirements 4 STAINLESS STEFLS IN LWR SYSTEMS Annual Report. October 1983 - For Aboveground Vaust Deposal Of Low Levet Radoactare Waste Septerrcer 1984 NUREG/CR-4212. IN PLACE THERMAL ANNEAUNG OF NUCLFAR RE. NUREG/CR 3774 V04 ALTERNATNE METHOD FOR DSPOSAL OF LOW LEVEL R ADCACTIVE W ASTE Tap 2CcTechrucal Regerements ACTCR PRESSURE VFSSELS i NUREG/CR4395. CORRELATON OF CV AND KlC/KJC TRANSITON For Eartn Mounded Concrete Burmer Deposal Of Low Lover Resoec. ? TEMPERATURE INCAEASES DUE TO IRRADATON tive Waste l NUREG/CR 3774 VM ALTERNAftVE METHOOS FOR D SPOSAL OF LOW WEL RADOAW WASUash 2ETechnical Woments A C 3 C.PD2 INCEPENDENT ASSESSMENT FC' Sha4 Dsposal Of Low Level Radoactive Waste. ! NUREG/CR-3881 THE MODEUNG OF BWR CORE MELIDOWN ACCl-l TS - FOR APPUCATON IN THE MELRPIMOD2 COMPUTER Enganeered Safety Feature f NUREG/CR 3317. TECHNICAL BASES ANO USER S MANUAL f OR
'NUREG/CR4252: INCEPENDENT ASSESSMENT OF TRAC PD2/ MOO 1 THE PROTOTYPE OF SPARC A SUPPRESSON POOL AEROSOL CODE WITH BCL ECC BYPASS TESTS REMOVAL CODE.
Emergency Deeeel Generator NUREG/CR4130 CEDF A CODE FOR AEROSOL PARTICLE CAP-NUREG/CR4347: EMERGENCY DESEL GENERATOR OPERATING TURE IN CE COMPARTMENTS i EXPER:ENCE.1981 1983. NUREG/CH-4234 V01 AGING AND SERVCE WEAR Or ELECTRC MOTOR OPERATED VAL %ES USED IN ENG.NEERED SAFETY FEA-1 Emergency Opereung Procedure TURE SYSTEMS OF NUCLEAR POWER PLANTS NUREG-080013 5 2 At STANDAAD REVIEW PLAN FOR THE REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER Engeneered Safety System ! PLANTS LWR Edition Revison 1 To Secten 13 5 2. " Operating And NUREG/CR4191: SURVEY OF UCENSEE CONTROL ROOM HABIT. j Maintenance Procedures." and Reveon 0 of Appendu A to Secton ABIUTY PRACTICES l ' 13 5 2,
~~Review - -" NUREG/CR 4210 MATADOR A COMPUTER CODE FOR THE ANALY- ,~~
I SIS OF RADONOCUDE BEHAVOR DUR:NG OEGRADED CORE AC- ! Emergency Preparednme CIDENTS IN UGHT WATER REACTORS j NUREG 1140 DAFT FC. A REGULATORY ANALYS43 ON EMERGENCY NUREG/CR 42t t. MATADOR (METHODS FOR THE ANALYSIS OF PREPAAEDNESS FOR FUEL CYCLE AND OTHER RADOACTIVE TRANSPORT AND DEPOSJTON OF RADONUCUDES) COOE DE-MATERIAL LOENSEES DraN Report For Comment SCRIPTON AND USER S MANUAL ' NUREG/CR-4151. INTEGRATON 08 EMERGENCY ACTON LEVELS , WITH COMBUSTON ENGINEERING EMERGENCY OPERATING E CmM j PROCEDURES By Use Of Combuston Engmeereg Owners Group NUREG/C44051 ASSESSMENT OF JOBRELATED EDUCATONAL ! Emergency Operating Procedure Technical Guidelina OUAUFCATONS FOR NUCLEAR POWER PLANT OPERATORS I Esnergency Responee EntPN** Mod i NOREG4981 RO1: NRC/ FEMA OPERATONAL RESPONSE #ROCE- NUREG/GR 4322 V01. CORPORATE DATA NETWORN (CON) DATA i DURES FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR REQUtRE MENTS TASK Voi t. Eriterpnse Modes j ACCIDENT. Entradewnent l Emergency Reeponse Planrung NURE G/CR4424 DROPLET SilES.DYNAMCS AND DEPOSITON IN , NUREG/CR-3657: PREUMINARY SCREENING OF FUEL CvCLE AND i VERTICAL ANNULAR f LOW ! BY-PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING I i I l i__. . , _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ , , , . _ , . . _ _ ._ _--_, ___
Subject index 161 Environment Empert System NUAEG/CR3738 ENVIRONMENTAL EFFECTS OF THE UAANIUM NUREG/CR 4272 RESPONSE TREE EVALUATON EXPERMENTAL FUEL CYCLE A Review Of Data For Technebum ASSESSMENT OF AN EXPEAT SYSTEM FOR NUCLEAR REACTOR NUAEG/CR-4138. DATA ANALYSES FOR NEVADA TEST SITE (NTS) OPERATORS PAEMIXED COMSUSTON TESTS. Emplosive Gas RG 1 NV RON tENTA6 ASSESSMENT FOR RENEWAL OF ROL OF EXPLOSNE MIXTUAES IN PWR STE G SYS ' SPECIAL NUCLEAR MATERIAL UCENSE NO. SNM-1107 Docket No. 70-1151. (Westirghouse Electne Corporator9 Esposure NUAEG-1130. ENbAONMENTAL ASSESSMENT FOR RENEWAL AND NUREG 0713 V05. OCCUPATONAL RADIATON EXPOSUAE AT COM. NL E IR L SS NT R A O I SOURCE MATERIAL UCENSE NO. SUB-1010 Docket No 40 8027. E L ED MFD L ACILIT ES NUREG/CR3444 V02 THE IMPACT OF LWR DECONTAM'NATONS MU 112 ENN ME AL ASSESSMENT FOR RENEWAL CF WASTE DISPOSAL AND ASSOCIATED OCCU-l SPECIAL NUCLE AR MATER:AL LICENSE NO SNM-368(UNC Naval i Products Omsson Of UNC Resources,Inc) NUHEG CR 3 4 i OGICAL CHAAACTER12ATION OF RADIATION EXPOSURE AND DOSE ESTIMATES FOA INHALED URANNM MILL. Environmental Con =equences ING EFFLUENTS Annual Progress Report Aprd 1983 March 1984 NUREGICR-4061; LEACHATE PLUME M.GAATON DOWRGAADIENT NUAEG/CR-3990' CHARCOAL PEAFORMANCE UNDER ACUOE NT FROM URANfUM TAILINGS OtSPOSAL IN M:NE STODES CONDITIONS IN UGHT W ATER AE ACTCAS NUREG/CR 4031 V03 NEUTRON SPECTRAL CHAAACTER12ATION Env6ronmental Effect FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST)1R-NUREG4975 V03: COMPILATION OF CONTRACT RESEAACH FOR RADIATON SERES " Neutron Emposure Parameters " THE MATERIALS ENGtNEERING BRANCH.DIVISJON OF ENGINEER _ NUREG/CR 414T THE EFFECT OF ENV;AONMENTAL STAESS ON ENG TECHNOLOGY. Annual Aeoort For FY 1964 SYtGARD 70 SILICONE ELASTOME A NUREG/CR3945 FATGUE CRACK GROWTH RATES OF LOW-CARBON AND STAINLESS PIPING STEELS IN PWA ENVtAONMENT Estended Storage NUREG/CR 4062 EXTENDED STORAGE OF LOW-LEVEL AADCAC. Enctronmental Impact TNE W ASTES Potential Prctiem Areas NUAEG-1165 ESRP 71.1
ENVIRONMENTAL IMPACTS OF PCSTU-LATED ACCOENTS INVOLVING RELEASES OF RADCACTNE MA. Entremety Dosimetry TERIALS TO GROUNOW ATER " PUAEG/CR 4297 EXTREMITY MONITOntNG Cons.decahons For Use Dos meter PlacementAnd Evasuaton Environmental Transport NUREG/CR4350 V07. PROBA81USTIC RISK ASSESSMENT COURSE FAST DOCUMENTATION Volume 7 Environmentas Transport And Conne- NUnEG/CR3830 V02 AEROSOL AELEASE AND TRANSPORT quence Analyses PROGA AM Semiantmas Progress Report For Apr4 1964-September 1984 Equation Transformate System NUR2G/CA 4346 AEROSOt. RELEASE ESPE AiMENTS IN THE FUEL NUREG/CR4213 SETS REFERENCE MANUAL AERCSOL S'MULANT TEST FACJUTY UNDERSODruu E sPE Ai-Equipment MENTS NUAEG/CR.4156 OPERATING EXPER'ENCE AND AGING SEISuiC ASSESSMENT OF ELECTRO MOTOAS FEMA NUAEG 0991 R0t NRC/ FEMA ODERAflONAL AESPONSE PROCE.
Equipment Hatch DUACS FOR RESPONSE TO A COMMERCIAL NUCLEAR REACTOR NUREG/CR4137. PRETEST PAEDICTONS FOR THE RESPONSE OF ACC CEN T A 18 SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO STATIC OVERPRESSURl2ATION FERRET SAND H NUAEG/CA 3319 LWR PRESSUAE VESSEL SUAVEILLANCE DOS ME-Equipment Ouahf*ation TRY IMPROVEMENT PROGRAM LWR Power Reactor Surve.liance NUAEG/CR-3914 PUMP AND VALVE OUALIFICATON PEVIEW GU;DE Phys.cs Des metry Data Bas a Compendium NUAEGICR-4358. APPLICATIONS OF DENSITY PAOFluNG TO EQUIP. MENT OUAUFICATON ISSUES FIRAC NUREG/CR 4121 FULL SCALE MEASUREMENTS OF SMORE TRANS-Equipment Survivability PORT AND DEPOSITON IN VENTILATON SYSTEM DUCTWOAK. NU AEG/CR-4146 SaMULATION OF AN EPRI NEVADA TEST SJTL (NTS) HYDAOGEN BUAN TEST AT THE CENTRAL RECENER TEST FtST FACluTY. NUREG/CR 4127 V01 BWR FULL INTEGRAL SIMUL A TON TEST (FIST) PROGRAM T1AC BWR MODEL DEVELOPMt NT Volume 1 N# Econt Tme mencal Methods WUAEG/CA-4050 A REVIEW OF THE SHOREHAM NUCLEAR POWER STATON PAOBA01USTIC A'SM ASSESSMENiintomal Events And FITS Core Damage Fr - NUAEG/CR 3721 VG1 PAE SSUHE MEASUAEMENTS IN A HYDAOGF N NUREG/CR 4350 V 4 P 08A8:USTIC RISK ASSESSMENT COUASE COMBUSTION ENVIRONMENT Hydrogen At Combuston Test Seres DOCUMENTATON volume 4 System Rehatzkty And AnaPyses 1 And 2 M The FITS Tank. Tecnewwes Sessions B/C - EveM Trees / Fast Trees NUAEG/CR4377. EVALUATIONS AND UTru2ATIONS OF RISK IM' FLECHT PORTANCES. NUREG/CR.4ter FLECHT SE ASET PROGRAM Final Report NRC/EP41 Esamination Weshnghw.e Report Number 16 NUREG-1122. l'NOWLEDGES AND A8tuTIES CATALOG FOR NUCLE. " $ AR POWER PLANT OPEAATORS Pressunzed Water Reactors 4EG/CA 4164 ANALYSIS OF FLE CHT SE ASET 163 AOD BLOCMED Esaminer Standard BUNDLE DATA USING COBRA-TF. WUAEG-1021 Rot; OPERATOR UCENSiNG EXAMINER STANDARD $ p py Eshaust Flow NUAEG/CR 4292 A COMPARATIVE ANALYSIS OF CCNSTITUffVE RE. RUREG/CR4tt2 vot: INVEShGATON OF CABLE AND CABLE LATONS IN TRAC PF L AND RELAPS/MODt SYSTEM FiAE TEST PARAMETERS Test AIEEE Flame Test FORTRAN 77 EsWe EMP 13 Battery Cett NUREG/CR 4122 A FORTRAN 77 PROGRAM AND USE A S Guirt NUAEG/CR 4097. TEST SERtES 4 SEiSMCFRAGiu?Y TESTS OF FOH THE CALCULATION OF PAAtlAL CORHELATION AND ST AND-NATUAALLY-AGED EXCE EMP 13 8ATTERV CELLS A ADi/ED REGHE SSON COf F FICIF N TS
162 Subject Index FRAC NUREG/CR-4387: EFFECTS OF CONTROL SYSTEM F AILURES ON NUREG/CR-4217. A STATISTCAL ANALYSIS OF NUCLEAR POWER TRANS!ENTS ACCIDENTS AND CORE-MELT FREQUENCICS AT A PLANT VALVE FAILURE-RATE VARIA81UTY-SCUE PREUMINARY GENERAL ELECTRIC PRESSURt2ED WATER REACTOR RESULTS. NUREG/CR4440 A REV!EW OF EMERGENCY DIESEL GENFRATOR PERFORMANCE AT NUCLEAR POWER PLANTS NUREG/CR-3810 V03 REACTOR SAFETY RESEARCH Fadure Ariaiysse PROGRAMS Ouarter eport, July temeer 1984 NUREG/CR-4149 ULTIMATE PRESSURE CAPACITY OF RElNFORCED NUREG/C&3810 : REA R SAFETY RESEARCH W PR M W SSED @ % R M W M 6 T PROGR AMS Ouart Report,0ctot3er-Decemtser 1984. NUREG/CR4318 1- REACTOR SAFETY RESEARCH Fadure Mecharusm NUREG/CR-3819 SURVEY OF AGED POWER PLANT F ACIUTIES BlU E / 43 C O FETY RESEARCH PROGRAMS.Ouarterty Report.Apre-June 1985 , PCAPCON-2 NUREG/CR-4217 A STATISTCAL ANALYSIS OF NUCLE AR POWER NUREG/CR-3741 V02 EVALUATON OF POWER REACTOR FUFL PLANT VALVE FAILURE RATE VARIABiUTY-SOME PREUMAARY ROO ANALYSIS CAPABluTIES Phase 2 Topical Repod Volume RESULTS
~~2. Code Evaluabon.~~
FaHure Rates Fat r6catiort Crtteria NUREG/CR 4228 REvtEW OF THE VOGTLE UNITS 1 AND 2 AuxiuA. NUREG/CR 3854 FABRCATON CRITERIA FOR SHIPPtNG CONTAIN. RY FEEDWATER SYSTEM REUABluTY ANALYSIS ERS Fast Breeder Reactor NUREG/CR 4375 THEORY DESIGN AND OPE RATON Or tiOUiD REG-t t t6. A REVIEW OF THE CURRENT UNDERSTANDING OF METAL FAST OREEDER RE ACTORS.!NCLUDiNG OPE R ATONAL THE POTENTIAL FOR CONTA NMENT FAILURE FAOM IN VESSEL " """*bC3 STEAM EXPLOSIONS WUREG 1 NUCLEAR PLANT AGlNG RESEARCH (NPAR) PRO' FM Fu TM FMy NUREG/CR 43 75 THEORY.DE S+GN. AND OPERATON OF UOUiD NUREG/CP-0070 PRUCEEDINGS OF THE WORASHOP ON SEISMC METAL FAST BREECER RE ACTORS. INCLUDING OPERATONAL AND DYNAMO FRAG!UTY OF NUCLEAR POWER PLANT COMPO. HE ALTH PHYSCS NENTS NUREG/CR 3647. DESIGN AND FA8ACATON OF A 118 SCALE Faa h r STEEL CONTAINMENT MODEL NUREG/CR-3660 V04 PROSAB1UTY OF PtPE FAILURE IN THE REAC- NUREGrCR 41 t2 V01 INV E ST!GATON OF CABLE AND CABLE TOR COOLANT LOOPS CF WESTINGHOUSE PWR PLANTSVolume SYSTEM F:RE TFST PAR AMETERS Testi A IEEE Flame Test 4 Pipe Fadure Induced By Crack Groeh in West Coast Plants NUREG/CR3663 V03: PRC0A8tuTY OF PIPE FAILURE IN THE REAC- Fatigue TOR COOLANT LOOPS OF COMBUSTON ENGINEERING PWR NUREG/CR 3226 V03 STRUCTURAL INTE GRITY OF W ATER REAC. PLANTS. Volume 3 Doutse Ended Gudictme Break inerectfy induced By TOR PRESSUHE BOUNDARY COMPONENTS Anrmat Report F or Earthquakes 1984 NUREG/CR 3831 THE IN-PLANT REUABILITY DATA BASE FOR NU. CLEAR PLANT CCMPONENTS Interm Report Diesef Fatigue Crack Growth Generators 8attenes, Chargers And inverters NUREG/CH 3945 F ATiGUE CHACE GROWrTH R A TE S OF LOW-NUREG/CR-3943 THE BWR PLAN ANALY2ER CARBON AND STA!NLESS PIP 1NG STEELS IN PWR E NVIRCNMENT NUREGiCR-3952 SEQUOYAH E'OUIPMENT HATCH ECAL LEAKAGE NUREG/CR4121 EFFECTS OF SULFUR CHE M'STRY AND FLOW NUREG/CR-3991: FA4 LURE MOCES AND EFFECTS ANALYS.S (FMEA) RATE ON FAT GUE CRACK GROWTH RATES IN LWR ENVIRON-OF THE ICS/NNI ELECTRC POWER DISTRIBUTON CIRCUITRf AT ENTS THE OCONEE-1 NUCLEAR PLANT NUREG/CR-4004 CLOSEOUT OF IE BULLET:N 79-25 FAILURES Or I'# WESTINGHOUSE BFD RELAYS IN SAFETY RELATED SYSTEMS NUREG/CR 3178 STRUCTURAL AND TECTOPcC STUDIES IN NEW NUREG/CR4123- SEISMIC FRAGIUTY OF REINFORCED CCNCRETE YORK STATE Fnal Report.My 1981. June 1962 STRUCTURES AND COMPONENTS FOR APPUCATION TO NUCLE. NUREG/CR-4117 F AULTING AND JOINTING IN AND NE AR SURF ACE AR FACiUTIES NUREG/CR 4tS6. OPERAT6NG EXPERIENCE AND AGlNG-SEISMC MINES OF SOUTHWESTE RN 'NDIANA ASSESSMENT OF ELECTR C MOTORS NUREG/CR-4339 A REV'E Ar OF RECENT RESE ARCH ON THE SEli NUREG/CR-4180- STATE OF THE ART CF SOUD STATE MOTOR MOTECTONCS CF THE SOUTHE ASTERN SEABOARD AND AN CONTROLLERS EVALUATION 08 HYPOTHESES ON THE SOURCE OF THE 16e6 NUREG/CR 4209. COMPARISON OF ANALYTICAL PAEDICTONS AND CHARLESTON SOUTH CAROUNA E ARTHOUA> E EXPER. MENTAL RESULTS FOR A 18-SCALE STEEL CONTAINMENT MOCEL PRESSUR LED TO F AtVRE Fauft free NUREG/CR-4217. A STAftsitCAL ANALYSIS OF NUCLEAR POWER NUREG/CR 3301 CATALOG OF PRA DOM:NANT ACCOENT SE' PLANT VALVE FAILURE RATE VARIABILITY -SOME PREUM NARY OUE NCE INFORMATION RESULTS NURE G/CR 4213 SE TS AEFERENCE MANUAL NUREG/CR-4234 V01: AGING AND SERVCE WEAR OF ELECTAC NUDEG/CR 4350 V04 PROBABILISTC RISM ASSESSMENT COURSE MOTOR OPERATED VALVES USED IN ENGINEERED SAFETY-FEA" DOCUME NT A T ON VoNrne 4 System RekataMy And AnaWe TURE SYSTEMS OF NUCLEAR POWER PLANTS Tecnniques Sem 0/C E vent Trewf suit Trees NUREGICR430$ COMYENTS ON THE LEAM-BEFORE BAE AM CON-CEPT FOR NUCLEAR POWER Pt>NT P'P!NG SYSTEMS Fault Zone NUREG/CR-4314 BR;EF SURVEY AND COMPARISON OF COMMON P4UREG/CR 3 t 74 V02 GE OPHYSICAL GEOLOGsCAL S TUDIE S OF POSSIBL E E X TE NSONS OF THE NE W MADRID FAULT NURE / 326 V ECTS OF CONTROL SYSTEM FAILURES **E A 1 F ON TRANSlENTS AND ACCIDENTS AT A 3tOOP WESTINGHOUSE NURE G, 3 t F AUG ZOM Mmm M U PAESSURilED WATER RE ACTOR Mac R UNOS NUREG/CR4326 V02. EFFECTS CF CON OL SYSTEM FARURES ON TRANSIENTS AND ACCIDENTS AT A 3-LOOP WESTINGHOUSE I"" I'"' PAESSURilED WATER REACTOR Appeneces NUREG/CR 3744 V02 NE AVY M CTON S* EEL TECHNOt OGY PRO, NUREG/CR-4383 HIGH PRESSURE 6NJECTON OF met T FROM A GRAM SE MIANNUAL PROGRESS RE PORT FOR APRt-M PTE UHE R REACTOR PAESSURE VESSEL . THE 0 SCHARGE PHASE NUREG/CR-4385 EFFECTS OF CONTROL SYSTEM FAILURES IN 1984 TRANSIENTS. ACCIDENTS. AND CORE. MELT FREQUENCIES AT A NUREG/CR 4219 V01 HE AVY SE CTION STE E L TECHNOt OGY PRO WES'INGHOUSE PRESSURllED WATER REACTOR GRAM SEMIANNUAL PROGRESS RE PORT F OR OCTORER 1944 - NUREG/CR4396 EFFECTS OF CONTROL SYSTEM FAILURES ON MARCH 1985 TRANSIENTS. ACCIDENTS. AND CORE-MELT FREQUENCIES AT A NUREG/CR42/$ HE AVY SECTON STEEL TECHNOLOGY rROGRAM BABCOCK AND WtLCOX PRESSURilED WATER FIEACTOR FIVE YE AR PLAN FY 1964 1969
Subject index 163 Film Flow Fission NUREGICR-3651. ASSESSMENT OF THE ADEQUACY OF ORNL IN NUREG/CA-3930 OBSERVED BEHAVOR OF CE S:UM IODtNE. AND STRUMENTATON IN REFLOOO TEST F ACILITIES TELLUR UM IN THE ORNL FISSION PRODUCT RELE ASE PRO. Film Rupture I NUAEG/CR4422 A REV:EW OF THE MOCELS AND VECHANl SUS Fissson Gas Release FOR ENVIAONVENTALLY-ASSISTED CRACK GROWTH CF PRES-NUREG/CR 4056 PARTICULATE AND GAS RE LEASE FRCU LGHT. SUAE VESSEL AND P1P1NG STEELS IN PWR ENVIRONMENTS W ATER AC ACTOR RWR) FUEL RODS STOHED IN INERT AND ORY p AIR ATMOSPHERES NUAEG/CH4111 A COMPARA'IVE STUDY OF HEPA FILTER EFF1-F4ssion Product CIENCIES WHEN CHALLENGED W1TH THERMAL- AND A R-JET. l GENERATED DI 2 ETHYLHEXYL SEBECATE.DI-2 ETHYLHEXYL NUREG/CR-3t97 V01. HE ACTON BETWEEN SOME CESTUM ODiNE PHTHALATE.AND SODIUM CHLOACE- COMPOUNDS AND THE RE ACTOR MATERIALS 304 STArNLESS STEEL.INCONEL 600 & SILVER Volume I Cowm Hydronde Reac-Filter Flow Test t*"S NUMEG/CR4056 PARTCULATE AND GAS RELEASE FACM LIGHT. NUAEG/CR 3980 V03 LtGHT WATEA RE ACTOR SAFETY FUEL SYS-WATER AEACTOR (LWR) FUEL ROOS STORED IN INERT AND ORY TEMS RESEARCH PAOGRAMS Quarter +y Progess ReportJuh Sep-air ATMOSPHEAES '*** N NUAEG/CA 3980 V04 LtGHT W ATER REACTOH SArETV FUEL SYS-Filtration TEMS AESE ARCH PROGRAMS Oi arterty Progess Deport. October. NUA E G/ CA-3537 EXPEDtEN T METHOD 3 OF RE SPlR ATORY Decemrer 1994 PAQTECTON iil SUBU CRON PARTICLE TESTS AND
~~SUMMARY~~
~~NUREG/CA 4037 DATA~~
~~SUMMARY~~
AEPOAT FOR FISSON PRODUCT OF QUALIT'Y F ACTORS RELEASE TEST Hi S Final Environmental Statement NUREG/CR 4081 AB50APTON OF GASEOUS ODiNE BY W ATER DROPT ETS NUAEG-1033 FINAL ENV;RONMENTAL STATEMENT RELATED TO NUREG/CA 4085 USERS MANUAL FOH CONTA2N 9 0 A Cornouter THE OPEAATON CF WPPSS NUCLEAR PAOJECT NO 3 Docket No Code for Severe Aeactor Accdant Conta.nment Anaivvs 50&8 (WasNneon Pubhc Pe*er Suppy System NUAEG/CR 4105 AN ASSESSMr.NT OF THF RMAL GRADeENT TUBE NUHEG-1073 FAAL ENVtRONVENTAL STATEMENT RELATED TO RESULTS FACM THE Ht SE AiES OF FiSSON PRODUCT RELE ASE THE OPERATON OF AIVER BENO STATON Docket No 50456(Gurt TESTS States utstes And Ca un Electnc Pewar Conperatrve) NUAE G-1085 FINAL ENVRONME NT AL STATEMENT RELATED TO NUREGrCA 4169 AN APPAOACH TO TRE AfiNG AADONUCLOE DECAY HEAftNG FOR USE SN THE VELCOA CODE SYSTEM THE OPERATON OF N'NE MtLE FC NT NUCLEAR STATON UNff NUAEG/CA-4172 A USE A S GUCE FOH MERGE NO 2 Docket No 50410 Wa' ara uvaat Power Corcora'en et a:) NUAEG 10e7: FsNAL ENURO4 MENTAL STATEMENT RELATED TO Fission Product Modet THE OPERATON OF VOGTLE ELECT A C GE NER ATING NUREG/CH 39eo V02 LGHT W ATER RE ACTOR SAFE TY FUf L SYS-PLANT. UNITS 1 AND 2 Docket hos 50424 And 50425 (Georgia TEMS RESEARCM PROGRAMS Quartery Progress Report Apris4sie i Power CompaebAL NUREG-1094 F ENV:AONMENTAL STATEMENT RELATED TO 19sa THE OPEAATCN OF BEAVER VALLEY POWER STATION UNIT Fission Product Release 2 Docnet No 504 t2 (Duquesne Lft Company) NUREG/CR-3885 V03 HIGH TEMPERATUAE GA% COOLED 4E ACTOH Financial Analysse SAFETV STUDtE S FOR THE DIVISON OF ACCIDE NT NUREG-1131. FINANCIAL ANALYSIS OF POTENTIAL AETROSPECTIVE EV ALUATICN Ouarterty Pr ess Aeport. Jufy 1 Septemte 30 1944 PREurVM ASSESSVENTS UNDER THE PRCE ANDERSON NUHEG/CA 4402 V01 HG EMPEHATURE GASWOLED RE ACTOn SYSTEM SAFETY G TUDtE S FOR THE DIVISON OF ACCIDE N T EVALUATON Quartetty Prcqress Aaport, January 1 March 31.1965 Finite Element Analysis NUAEG/CA4367 OHV'R T PC A 2 O FINITE ELEMENT FRACTURE b884" Ef U'"M" ANALYSIS PAOGRAM FOR A M CROCCMPUTER NUREG/CH 4130 ICEDF A CODE FOH AE ROSOL PAhflCLE CAP-TUAE IN CE COMPAR TMENTS F# nite Element Method NUREG/CA-4182 VERFICATON OF SOfL STAUCTUAE INTEAACTION Fieston Product Transport METHODS NUREG/CR 420$ THAP ME LT2 USE A S MANUAL Fire Acc6 dent Analysis Cornputer Code Five-Year Research Plan NUAEG/ CR-4 321 FULL SCALE MEASUREMENTS OF SMOxE TAANS- NUAEG 1080 V02 LONG AANGE RE SEAACH PL AN FY fp6 FY 1HO POAT AND DEPOS1 TION IN VENTILATON SYSTEM DUCTWOAM. Flame Test Fire tsposure NL AEG/CH 4t t2 Mt INVE STIGA TION OF CABLE AND CABLE NUAEG/CA4112 V02 INVESTIGATON OF CABLF AND CABLE SYSTEM FIRE TEST PARAME TERS Tasa A #EE E Flame Test SYSTEM FIAE TEST PARAMETEAS Tasa B Firesten Test Method Flashmg F6re Protection Research NUHE G/CA1917 STE AM GENERATOH TUDE RUPTURE OO4NE NUAEGtf44 NUCLEAR POWER PLAN T FIAE PRO TECTON AE- fnANspOAT MEEHANISMS Yask t Er rwnentad $turtse SEAACH PROGRAM NUHEG/CR 4019 ANALYTC STUDIES PERTA4PdG TO Sif AM GEN F6re Protection System EHATCH TUBE AUPTUHE ACCCf NTS NVAEG/CA-4230 PROBABILf7V-GASED EVALUATON OF SFLECTE D Flaw FIRE PAOTECTON FEATURES IN NUCLE AR POWER PLANTS NUAEG/CH 3721 STHE SSINTENs Ty rACfDA INFLUF NCE CCE F FI Cif NYS F04 SUHF ACE FL AWS IN PHE SSUHE VF SSF LS U EG/CR-4229 EVALUATON OF CUAHENT METHODOLOGY E M "U' *# ## M C N IE *** PLOYED IN PROBAB!LISTIC AISK ASSESSMENT (PRA) OF FIHE #M SEMIANNUAL PROGAESS REPOHf FOR APnit SEPTEM0E n
,4 NU / A 421 EVALU C A LABLE DATA FOA PACHABI, NUHf G/CA 1925 V014 ACOU 9fC EM!SSON FL AW HELATONSHiP LISTIC AtSK ASSESSMENTS (PH A# OF FIRE EVENTS AT NUCLEAA FOR IN SERVICE MONITORING OF NUCLEAH PRESsunE POWER PLANTS WES Oua% HW Apd N - See W was 3 and 4 pm,op NUHE G/CA 191S ACOUSTC E MISSON Af SULT5 00i ALNE D FHOU NUAEG/Cn-4 t f 2 V02 INVEs tlGA TICN OF CABLE AND CABLE TF STING THE IB t INTEnMf DATF SCALF PHf %SUHF VF S$f L SYSTEM FIAE TEST PARAMETERS Tash B Feestop Test Method NUDEG/CA 4219 V0f HE AVV SECf K)N Sif EL T E CHNOL OG Y PHO GAAM St M ANNUAL PROGHF SS HE PO4 f FOH OCTOUE H 19n4 -
Pleh MARCH 1945 NUAEG/CR 3931: BOACCUMULATION OF P 32 IN BLUEGILL AND NVHE G/CH-4275 HF AVV SF CTON STEI L TELHNOLOGY PROGHAM CATFISH FIVE YE AR PLAN FY Ia4419H6
164 Subject Index NUREG/CR-4284. NEUTRON EXPOSURE PARAMETERS FOR THE Fracture Mechanece FicTH HEAVY SECTION STEEL TECHNOLOGY IRRADIATION NUREG 0975 V03 COMPILATION OF CONTRACT RESEARCH FOR SER:ES- THE MATERIALS ENGINEEHiNG BRANCH. DIVISION OF ENG!NEER - NUREG/CR 4304- PRESSURE VESSEL FRACTUhE STUDIES PE R- lNG TECHNOLOGY Annual Report For F Y 1984 TAIN NG TO THE PWR THERMAL-SHOCK ISSUE Empenmer't TSE 1 NUREG.t 155 V01 RESE ARCH PROGR AM PLAN Reactor vesWe NUREG/CR-4325. A PARAMETRIC STUDY OF PWR PRESSURE NUR E G /CR-3723 STRESSINTE NSITY F ACTCA INF LUENCE COEF Fb VESSEL INTEGRITY DURING OVERCOOLING CIENTS FOR SURFACE F LAWS IN PRESSURE VESSELS ACCIDENTS.CONSOERING BOTH 2-0 AND 3-0 FLAWS NUREG/CR J744 V02 HE AVY-SECTION STEEL TECHNOLOGY PHO-NUREG/CR-4367- ORVIRT.PC A 2 O F# NITE ELEMENT FRACTURE GRAM Si MlANNUAL PROGAESS RE PORT FOR APAll SE PTEMUEH ANALYSIS PROGRAM FOR A M!CROCOMPUTER 1984 NUREG/CR 4022. PHESSUHi2ED THERVAL SHOCK EVALUATON OF Flaw Detect 6on THE CALVERT Clif FS UNtT 1 NUCLE AR POWER PLANT NUHt.G/CR-4300 V01: ACOUSTIC EMtSSON/ FLAW RELATONSHip NUREG/CR 4082 V01. DEGRADED PlPtNG PHOGRAM PHASE FOR IN-SE RVICE MONITORING OF NUCLEAR PRESSURE si Semannual Aeport. March 1984 Secter%er 1984 VESSELS Progress Report. October March t 995 NUHEG/CR 4082 V02 DE GR ADE D P PtNG PROGRAM PHASE il Sermannual Report. Octetier 1984 March 1985 Flawed Steel Ptate NUREG/CR 4106 PHE SSUHilE D- T HE R M AL-SHOCK TEST OF 8- a N - NUREG/CR-4015 EFFECT OF STAINLESS STEEL WELD OVERLAY THICK PRESSURE VESSELS PTSE 1 invesogaton Of Warm Prestress-CLADO:NG ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL rg And Upper Sheff Arrest PLATES JN BENDING SERtES 1. NUREG/CR 4219 V01- HE AVY-SECTION STE EL TECHNOLOGY PHO-GAAM SEMIANNUAL PROGnESS REPORT FOR OCTODER 1994 - FDuble Ptpeng Design yancy 9ggg NUREG/CR-4263 RELIABILITY ANALYSIS OF STIFF VERSUS FLEXI- NUAEG/CR 424) PHESSUAE %ESSEL FRACTURE STUO45 FENE-BLE PIPING FINAL PROJECT REPORT. TR A TING TO THE PWR THE AMAL SHOCR ISSUE E APFH UENTS TSE 5.TSE-5A AND TSE 6 Flow Blockage Heat Transfer NUAEG<CR 4275 HE AVY-SECTON STEEL TECHNOLOGY PROGRAM MUREG/CA 4166 ANALYS,$ OF FLECHT SEASET 163 ROD BLOCKED FIVE YEAR PLAN FY 19A41989 BUNDLE DATA USING COBRA TF NUAEG/CR 4304 PRESSURE VE SSE L FRACTURE STUDiE S PE R NUREG/CR-4t67 FLECHT SEASET PROGRAM Final Report NRC,EPRI f AANG TO THE PWR THF AMAL $HOCm ISSUE Erpenment TSF 1 Westinghouse Report Nur@er 16 NUREG'CR 4305 CCVVENTS ON THE LE Am DE FOHE BREAK CON-CEPT FOR NUCLE AR POWER PLANT PANG SYST[MS Flow Rate NUREG/CR 4325 A PAHAVETHO STUDY OF PWH PRESSUHf NUREG/CR 4t21 EFFECTS OF SULFUR CHEM STRY AND FLOW VE SSE L INTE GRITY DURiNG OV E RCootlNG RATE ON F AT:GUE CRACK GROWTH RATES IN LWR ENV:AON' ACCIDE NTS CONSOERrNG DOTH 2 O AND 3 D FLAWS MENTS. ' **'" "8 ** Flow Redistnbution NUAEG/CR 3979 TE NSILE PROPERT'ES OF IDHAOlATE D NUCL[ AA NUREG/CA-3436 TWO-OlMt NSiONAL MODELtNG OF INTR A-SUB AS- W E AW MM W M M WM SEMBLY HEAT TRANSFER AND BUOY ANCY4NDUCED FLOW RE- FOUATH HSST IHR ADIATON $EH:FS DISTRIBUTON IN LMFB AS NUREG/CR 4395 COOHELATON OF CV ANO siCesJC THAWTON poc , y,enny, TEMPERATUHE INCHE ASES DUE TO IARADIAf rON NUREG/CA-3145 V03 GEOPHYSICAL INVESTIGATIONS OF THE Y WESTERN OHtO INDIANA REGON ANNUAL REPOR T (Octteer t EG/CR 3Me HAND 000M OF NUCL EAH POWt A PLANT SE iSU'C 1992 September 1993 Volume 31 NUREG/CA 4286 FOCAL MECHANISM ANALYSES FCA VIRGAA S"
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7 , am L O CONCHE f f AND EASTERN TENNESSEE EARTHOUAnES (19781994) STRUCTunES AND COMPCM NTS FOR APPt sCATON TO NUCLE . Force-8atance Pressure Tranermtters AR F ACILIT!ES NUREG/CR-4256 MEASUAEMENT OF RESPONSE T:ME AND DETEC-
~~N OF DEGRADATsCN IN PRESSURE SENSOR /SENS.NG LiNE gU $ 99 TWO DIME NSIONAL MODE t tNG OF ;Nf H A SUH AS.~~
SEMBLY HE AT TRANSFER AND OUOYAP4CY 6NOVEED FLOW HE PErced Convective DISTR'0UTON IN LVFORS NUAEG/CR 3193 FOACED CONVECTIVE NONEOljauBRtM. POST-N CF EA TRAN R EXPER; MENT DATA AND COnHELATION gg gg , y g g
~
TE ST F ACillTY Faudy And iest DesNn Report Fordga Espenence NUHf G/CR 3426 v02 fHF AMAL AND f LUtO MmNG IN 1/2 SCAE TE ST F ACILITY Da'a RepM NUREG/CA-3883 ANALYSIS OF JAPANESE U S NUCLE AR POWE R PLANT MAJNTENANCE Fractography NbHE G 1927 RADIA TON PHOTECTION TRA4NG Af UHANtM HE t. AF LlKW OE AND F UI L F AHR'C AT ON Pt AN TS NURE G/CR4945 FATIGUE CRACM GAOWTH RATES OF LOW NUF:G/CR 3944 TRAN A 3 E XPE H.ME NT AL INVE STsGAf 80N Os CAABON AND STA>NLESS P@NG FTEELS IN PWR ENV@ONMENT FUEL ( HUST St A8'tsf y ON MF L YtNG SUHF ACs S OF A N A P,Nu NURE G /CR -4121 EFFEC'S CF SULFUR CHEMisfHY AND F LOW RATE ON FATIGUE CRACx GROWTH RATES IN LWR ENV@ON LAR FL .h CHANNE L MENTS Fuel Aerosof Semulant Test Fracture NUHf G/CH 3810 V02 Al f*O%OL fiE LE A%E A P4D T RAP 4SPOH T NUAEG/CR 3178 9iHUCTURAL AND TECTONtC STUDIES IN NE
PHOGRAM Seannual Proyess Heport For Aprs IM4 Septeenter YOAK STATE Final Aeport.JuPy 1941. Jun e 1992 1M4 NUHtG/CR 4346 AE AOSOL HE L F ASt, E sPE R ME NTS IN THF f ut t NUAEG/CR 3229 V03 $THUCTURAL INTEGHITY OF WATER RE AC-TOR PRES $URE BOUNCARY COMPONENTS Annual Report For AlROSOL S Mut AN T TEST F ACILITY UNUf HSODiUM E RPL Al-1984 MLNTS NbAEG/CR-4042 A 3 DiMEN510NAL COMPUTER MODE L TO hiMu Fuel Cladating LATE FlutD h.OW AND CONTAAMENT TRANSPORT THHOUGH A ROCK FRACTURE SYSff M NUHFG/CHmo V04 LONG fE HM Pf Hf 04MANCE OF MAf(R:A(S NUHEG/CR 4t98 FRACTUNE IN GLASS /HIGH LEVEL WASTE CAN S USE D F OR HtGH t E VI L W A$f E PACK AG Ped Ant %et Hegwwt Apert TERS '984 A 'il P 148S NUHf G/CH 4 tit IN fl GH A f ON OF f Mt HGl pry ACTIGN Lt vf L S Fracture Analytse WI T H COMBUSf04 (NGNffH 41 F Mi f 4GE NCY OPE H A fiPM NUR EG / CR-4367 OAVtRT PC A 2 D FiNif E ELt MENT F R AC TUDf- PROC E DUHf S 0y Use Of Lomt.uenn E nynnerwy Owners Group ANALYS:S PROGRAM FOR A M CROCOMPUTE A E mergany Og crat<vJI 'r9mfore f evhnerat GmM.rwe
Subject index 165 i Fuel Cycle GS2 NUREG/CR3293 V01: TECHNOLOGY. SAFETY AND COSTS OF DE- NUAEG/CA-3901. DOCUMENTATON AND USEA S GUOE GS2 & GS3 COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON- - VARIABLY SATURATED FLOW AND MASS TRANSPORT MODELS FUEL CYCLE FACTUTIES FOLLOWING POSTULATED ACCOENTS Maen Report GS3 NUAEG/CR 3293 VC2. TECHNOLOGY, SAFETY AND COSTS OF DE. NUREG/CR-3901 DOCUMENTATON AND USER S GUOE GS2 4 GS3 , COMM:SSONING REFERENCE FUEL CYCLE AND NON FUEL . VAAtABLY SATURATED FLOW AND MASS TRANSPORT MODELS I CYCLE FACfUTIES FOLLOWING POSTULATED ACCIDENTS Appendices. GITF l NUAEG/CA 4232: THE RESPONSE OF VENTILATON DAMPE AS TO NUREG/CR 4168 GT2F A COMPUTE R CODE FOR ESTIMA TING LARGE AjRFLOW PULSES UGHT WATE R RE ACTOR FUEL ROD FA%UHES Fuel Cycle Facility Garvanized Steel Corrosion I NUREG4525 A10: SAFEGUARDS
~~SUMMARY~~
E%ENT UST NUAEG/CA 3361 THE EFFECT OF WATER CHEMtSTRY ON THE (SSEL),REVIS'ON 10. RATES OF HVDHOGEN GENERATON FROM GALVANilED STEEL COAROSON AT POST LOCA CONDITONS Fues Damage NUAEG/CR4810 V04 AEACTOR SAFETY RESEARCH Gamma Dose Rate PAOGR AMS Ouarteriy Repnrt. October December 1984 NUMEG/CR-4203 A CALCULATONAL METHOD FOR DETE RM N NG NUAEG/CR-38 t 6 VC2. REACTOR SAFETY AE SE AACH Ouarterfy DIOLOGCAL DOSE RATES FROM IRAADIATED RESE ARCH AE AC-Aeoort Apr4-June 19A4 TOR FUEL NUAEG/CR 4037 DATA
~~SUMMARY~~
REPORT FOA FISS;ON PACDUCT RELEASE TEST Hi S Gamma-Ray NURE G /CR-4039 GAMMA RAf CHAR ACTE Ai!ATON OF THE TWO-Fuel Failure TEAH IRAADtATON EXPER#ENT PE AFORMED AT THE POOLSOE NUREG/CA.4168. GT2F A COMPUTER CODE FOR E STIM Af fNG F ACILITY UGHT WATER AEACTOR FUEL AOD FAILURES Gas Anatyrer UA G/CR4112 vot: INVESTIGATON OF CABLE AND CABLE ST GA SYST SYSTEM FIRE TEST PARAMETERS Tap A fEEE Flame Test Gas Blowthrough A H 4383 HGH NS%nE MN & W t T m A UA C 50 V0t FUEL PE AFCAM WCE ENNUL AEPCAT FCA McACTOR PHI SSURE VESSEL THE DSCmRGE PHASE 39g Gas But>tHe Fuet Removat NOREG/CA 27t8 STEAM E tPLOSON E uPt AMENTS WITH 5.NGLE NUREG/CA 3757. TRAN G-2 THE EFFECT CF LOW STEEL CONTENT COOPS OF IHON OuCE MELTED WITH A CO2 LASE R Part ON FLEL PENETRATON IN A NON MELTING CYUNDRICAL FLOW u Par m S W CHANNEL Gao Distrtbuhon System FW Rod NUAEG/CR3 741 V02 EVALUATON OF POWER REACTOR FUEL NUAEGrcn 3551 SAFETY #PUCATONS ASSOCIATED WITH IN. ROD ANALYSIS CAPA0luTIES Phase 2 Topsal Report votume PLANT PRESSUPilED GAS STORAGE AND DiST AsuisON SvS. TEMS IN NUCLEAR POWER Pt ANTS 2 M bauw NUAEG/CR3S48 EXPER MENTAL RESULTS OF THE OPEnATONAL
T N NT (OPTRAN) TESTS t f AND t 2 IN THE POWE A BUAST y gG y488 V03 IDAHO F iE LD f aPEn#E NT 1941 Volume 3 Compenson Of Traectoriee s Contectraton Pa' terne And Mt SOOF NVAEGICAI4037 DATA
~~SUMMARY~~
AEPOAT FOR FISSION PRODUCT M de# Casculatzes AELEASE TEST Hi-5 NUAEG/CH.4056 FARTICULATE AND GAS RELEASE FROM UGHT. O** '" ' TER A ACTCA WR} FUEL RODS STOHED IN INERT AND DAY gg C TEsm ARWWTON & wmM IN M'CE.AATS AND DOGS Aswaten To E statash.ng Venuee Of f t Fuel Rod Behavtor For Sos ube Plutonium NUAEG/CH-4345 INVE SilGA TION OF THE STA0iUTY OF LWR **"
l
$ PENT FUEL ROCS BELOW 250 C. gg g pg Fuel Rod Claddmg DATA WITH AF SULTS F ROM A NUMHf R OF DIF F ERENT MODE LS NUAEG/CH 39W ELECTRICALLY HEATED EX AE ACTOR PE L L F. T. Ge+ger#wner CLADDING INTERACTON (PC4 SiMULATONS UTiuliNG IAAAD'AT, NUREG/CR 41te MON:Tos0NG ME THOOS 7054 Of Tt HM NATION ED l'ACALOY CLADDlNG COMPUANCE WITH DECOMMtSSON<NG CLE ANUP CHITE HrA AT Fugettve Duet UHAN1UM RECOVERf S4TE S t NUAEG/CR4089 EV ALUATON OF FIELD-TESTED FUGiTIVT DUST CONTROL TECHNIQUES FOR URANfUM Mit TAIUNGS PILES G***M *d AM NURiG/CR 4192 Mt ASUHES OF SAFIGUAHDS DISK E M5'LOy NG Fukuehema Data PH A (MOSOf P) A Methodoeugy For E stimahng Ret impa<.to Of N4 NUAEG/CR 4182 VEAiFtCATION OF Soil ST AUCTURE INTE RACTON guardo Meawree UETHOOS Generic Safety issue Full Integral Semulation Test NUMEG 093J $03 A PAQFOTl/A TION OF GE NE H:G SAF ETY ISSUIS NUAEG/CR4t27 V01 BWA FULL INTEGAAL S MULATON TEST G*n* rte Safety leaves (FIST) PROGAAM TRAC-aWA MODEL DEVELOPME NT Volume i Nu.
mermal Methorte NUAl G 09n 502 A PHCHITilA TON Or GF NE AIC SAF E TY ISSUF S NUHlG/CR 410J USE S OF HUMAN RE UAmtit Y ANALYSIS PHou-Fury lnetrumented Test Site AmUSTIC HISW ASSE SSMf NT OF SULTS TO HE sot VI P(HSON. NUAEG/CA 3721 V01 PRESSUR( ME ASUHE MENTS IN A HYDHOGEN NE L Pi HF OHMANCE ISSUE S THA T Cout 0 AF i t CT SAF E T V COMBUSTON ENVIRONMENT. Hyrtrogener Coretmston Test Ser==e 1 And 2 In The FITS Tar
Genehe EHede Nunt G/CH 4214 Hf AL TH f F F(CTS MOOF L FOH NUCLE AR POAf R GE Type AK 2 Circuit PL AN T ACCOENT CON 5f GUf NCE- ANALYSIS Past NUHEG/CA 3791 CLOSEOUT OF IE DULLETIN 7909 FAituRE OF GE f introdurte intayrewn 4 Sumenary Pa t H SoerMm Ane F or Heann TYPE AW 2 CIHCUIT BREAPE45 IN SAFE TY DELAit D SYSTEMS Ef'ene Metre i
~~= . . , , . . _ - - - _ _ _ , - . - - . _ . - , , _ - - . . . . - - - - - - - - , - _ _ - - . _ - --~~
166 Subject index Geocherrocal Condition NUAE G/CH 41M F RACTURE IN GLASS /HOH LEVEL WASTE CANIS-NUREG/CR 4236 V02 FROGRESS IN EVALUAT'ON OF RADtCNU- TERS. CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HOH. NUREG/CR 4251 V01. MITIGATIVE TECHNOUE S FO4 GROUND-LEVEL NUCLEAR WASTE REPOSITORY SITE PROJECTS Report for WATER CONTAMINATION ASSOCIATED WITH SEVEHE NUCLEAH January March 1985 ACCOf NTS bolume t Analyses Of Genere Sete Corotone NLnEG/CR 4251 V02 MlflGA TIVE T ECHNiQUE S FOH GHOUND-Geochemical Mode 16ng WATER CONF AMINATION ASSOClATED WITH SEVEHE NUCLT AR WUREG/CP 0062. PROCEEDINGS OF THE CONFEAENCE ON THE AP- ACCOENTS Volume 2 Case Study Anaryses Of Hydrologic CParacter-PUCATON OF GEOCHEMICAL MODELS TO HIGH LEVEL NUCLEAR iration And Mitgative Schemes WASTE AEPOSITORY ASSESSMENT NUHEG,CH 4403 SUMMAHY OF THE WASTE MANAGE MENT PRO-NUAEG/CR 3851 V04 EVALUATON OF RADONUCLIDE GEOCHEMI- GRAMS AT ORANIUM RtCOVERY FACluTIES AS THEY RELATE TO CAL 6NFOPMATON DEVELOPED BY DOE HIGHLEVEL NUCLEAR THE 40 CF A PART 192 STANDARDS WASTE REPOSITORY SITE PROJECTS Annual Progress Report for October 1983-September 1984 Guido NUREG 1021 A01 OPERATOO UCE NSiNG E N AM Nf A ST ANDAADS Geochemestry NUqE GtCR 3914 PUMP AND VALVE OVAUFICAfiON 'LVit W GUOE PeUAEG.1164 INFOAMATON ON THE CONFINEMENT CAPA8fLITY Or THE FACILITY OfSPOSAL AREA AT WEST VALL EY NEW YC9m Guadonnes NUREG/CR4110 REPOSif0RV SITE DATA REPORT FOA UNSATO- NURE G/CH 2n00 S03 GUOEUNES FOH NUCLEAR POWER PLANT RATED TUFF. YUCCA MOUNTAIN NEVADA SAFETV ISSUE PH:OHITi2ATON INF 0AV Af K3N DE VE t OPMf NT WUAEGiCR4134 AE POSITORY E NViHCNUE NTAL PARAMETERS NURE G/CH 4125 V01 GOOELINE S AND WOHafREM F OR ASSI SS-RELEVANT TO ASSESSING THE PE AFORMANLE OF HIGH LEVEL ME NT OF CHGAN!2ATON AND ADMtNISTRATON Or UT Uf:ES WASTE PACMAGES SEDuNG OP( R A TING LICE NM FOH A NUCLEAR POWER PLANT Voeume 1 Gualehrws For Utey Orgarviation And A.fmanretraton NUREG-1164 INFOAMATON ON THE CONFlNEMENT CAPA01UTY OF NUHEG/CH 4129 V02 GuCEUNE S AND WO4wPOOM FOR A5SISS-THE FACiUTY DISPOSAL AREA AT WEST VALLEY.NE W YOR" ME NT OF CHGAN;lATON AND ACM@STD4f 0N OF UTIUT<f S SEEM:NG CPE R ATING l ICE NSE FOH A NUCLEAA POWEH H CR 1 6 L AND THEORE TICAL INVESilGATONS OF TVace2W e Assesswd G OrgaNaton And Mam FAACTUAEO CAYSTALUNE ROCm NEAR OHACLE.AH:lONA NL) H 4tSt IN TE GR A TION OF emf RGf NGY ACTON LEVELS Glacial Hydrogeo6ogy WITH COMRUS T ON E%NE E HiNG EVEHGtNCY OPE H ATING NUREG-I tta tNFOAMAllGN ON THE COM iNEMENT CAPABiUTV OF PROCEDUnE S Br Use Of Comtut oe. f aeneering Oeners Group THE FACluTY DISPOSAL AREA AT WEST VALLE Y.NEW YORg E mgancy CWating procasure Te&n= as G 16ne, NUHEGeCR 4221 HUMAN i NGINE E HiNG GOOF LINE S f 00 fHE Gaass Weste Form EVALUAflON AND ASSE SSMENT OF VOf 0 DISPLAY UNITS NUAEG/CR4196 FRACTUHE IN GLASS /HiGH LEVEL WASTE CANIS-TERS HECm NUnEG/CH 3912 MAnCH NICTR ANA6YS'S OF SELECfED ACCO Grenste Of NTS 'N AN ICE CONCf %f A CONT A:NU(NT NOAEGICA 243 V01 INFOAMATON NEEDS FOR CHAAACTE AJA NUH(GeCH 3313 He CiR VE H$sOto 10 USf A S MANUAL TON OF HIGH LEVEL W ASTE REPOSITORY SsTES IN S!x GEOLOG- NUHE G/CH 3%4 HE GTR ANALYSIS Of f OUiPMtNT T[ MPt RA TUHE IC UE DIA Main Reciert HESPONSE S TO Fst L ECil D HvDHOGE N HU ANS IN AN ICE CON NUAEG/CH-26M V02 INFOAMATON NEE DS FOR CHAHACTE R!2A DE NSF H CONTA:NME NT TION OF H2GH LEVEL WASTE REPOSITOAY SITES IN Six GEOLOu
^
N RE GrCH 4111 A COMPARATIVE STUDY OF HEPA FILTER E F Fi. Gravitat6cnal Sedarnentation Ci( NCiES WHE N CHALLINGFD WifH THE AMAL. AND A;H JE T. OU AEG/CR4158 A CCUPSLATON OF INf 04MATON ON UNCEn GE NE R AT E D OtPETHVLHERvL SEBf CATE 012 ( THYLHf sVL TA4NTIES INVOLVED IN CEPOSITON MOOEllNG PHTHAL ATE AND SCOiUM CHL OHOE NUHfG/CHal25 SUMMAHf OF lifiCrE NCY TE ST&T OF STANO-GrQtty APD AND HCH CAP ACaf Y HCH I FFICif NCY PARTICUL ATE AiH RUREG/CR 3174 V02 GEOPHYSICAL GEOLOGICAL S TUD.E S OF FITTERS SUtutCTID TO SiUVL Aff D TORNADO ETPHf ssonilA POSSd8LE EXTE NSiONS OF THE NE W MADHiD FAULT TON AND E APLO9VF SHOC88 WAvFS 7CNE Anroad Report For 19e3 NUREG/CH 4/64 INVE 5 f CA f CN ON HIGH f F FICit NCY PAHf kJU-LATE AIR FILTEH PLUGGeNG BY COMBUSTON Af HOSOLS Ground Motion NVAEG/CR 3eCS V02 ENG NE E AING CHAD ACTE Ar2Af 0N OF HMS GROUND MOTON Test it Ef'ects Of Ground Mvtion CPteracteristw;s NUAEG/CH 4020 HMSA COMPUflR PHOGRAM FOH On Structural Aeepon$e Conwienng Locahied St'ucual Nr*neant.eg TRANSrE NT.THHf E D'M[NSONAL M10NG GASES And So45trucVe Interaction Ertacts Groundwater NUHf G/CH 3/01 ASSESSMf NT OF Si tf Cff D T A AC AND Hit APg NUAEG 1046 DISPOSAL OF HIGH LEVEL nAD60 ACTIVE WASTE S IN CALCULATOP45 FOH OGONt i? t PHf SSuni/E D THE HMAt sHocu UNSATUAATED ZCNE TECHNICAL CONSOfRATONS AND AE S T UDY. SPONSE TO COMMENTS NUREG-1164 INFOAMATON ON THE CONriNE MENT CAPADIUTY OF HPt det THC F ACluTV DISPOSAL AnE A AT WEST VALLFY NFW YOHr NUHEG/EH 3426 V0' f Hf DM AL AND F L UO M:E,4G 1741/2 $ CAL E NVAEG-f t65 ESHP 71 t ' ENylHONVE NT AL IMPACTS OF POSTU TF ST F Aca lf Y F acov Antt Test Dawfi neport LATED ACCOENTS INVOLV+4G AELE AST S OF RADOACTIVE MA. NUHI G/CH 3426 Vt2 f Hf hMAL ANO F L UO MulW3 IN t /2 5 CAL E TE AtAt S TO GROUNOW ATE R " if ST F ACIUTY Data Harport NUAEG/CH 3109 UE THOOS OF M'Niu IING GAOUND W A TE R COP 4 TAM + NATION FRCM IN SITU LE ACH UHAN!UM M%NG Firsal Aepo.rt HMA NVAEG/CH 145t VC4 EVALUATON CF AADONUCUDE GE OCHLMf- NUREG/CH 4010 SPf CIFICATON OF A HUMAN Hf LIAAltiTY DATA CAL tNFOAMATON DEVFLOPED Rf DOE HIGH LEVEL NUCLE AH D ANN F OH CONDUCilNG HD A St Gut NT S Of I'HAS F OH NUCL E - WASTE REPOSif 0AY SITE PROJf CT9 Annual Pmgress Report For AH POWE H PLANTS October 1961 Septemtear 1994 NUHf G/CA 4030 NADONUCUDt MORATON IN GROUND HSST W A TE R IFeal Reporg NUAf G/CH W9 TI NSit E PHOPEnfif 9 Of if4HADIATI O NUCtf AH NUAEG/CA 4t to HIPDS4 TONY SifE DAT A Hf PORT F00 UNSATU- GHAtt PHFSSUHE vf W L et. A f f AND wtLDS FOH THE AAYED TUFF, YUCCA MOUNTA4N NF VADA F OUHfH HS$f IHHADI Af CN sf Hit S NUREG/CA 4t34 Af POSif 0Rf E NTHONME N f AL PAHAMEffHS NUHiG/CH 4014 tFitGT OF s f A.Pu f SS Sf E ( L WE t 0 Ovf HL A Y AELEVANT TO ASSESS.NG fME PEHFORMANCE OF HIGH Lf ViL CL AfXANG ON THE S THUC TUH AL IN f f GFtif Y OF F L AWt D Sig i L WASTE PACKAGES Pt A f t S IN Bf NDf 4G St H:( % 1
Subject index 167 NUREG/CR4086 TENStLE PROFERTIES OF IRRADIATED NUCLEAR NUREG/CR 4376 HEA! THANSFEH,CARRYOVE R AND F ALL BACK 6N GRACE PRESSURE VESSEL WELDS FOR THE THIRD HSST 1RRA- PWR STE AM GENERATORS DURING TRANSE NTS OtATON SERtES. NUREG/CR4414 DIRECT < CONTACT CONDENSATON C5' STE AM ON NUREG/CR 4284 NEUTRON EXPOSURE PARAMETERS FOR THE COLD W ATER IN STR AflFIE D COUNTERCURRE N T FLOW FIFTH htAVY SECTON S7 EEL TECHNOLOGY IRRADIATON NUREG/CR 4417 LOCAL PROPf Rf tf S OF COUNTERCURHE N T SERIES STR ATIRF.D STEAM W ATER F LOW HTGR Heat Affected Zone NUREG.t125 V05 A CCMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SA'EGUARDS,19571984 Volume 5 Part NUREG /CR-3911 V02 EVALUATON OF WELDED AND REPA;R. 2 ACRS Reports On Genenc Sul ett (HTGR - Repatory Gates) WELDE D ST A6NL E SS STEEL FOR LWR SE HVtCE Ousterry Report'Apr4 June t 944 NUREG/CR 3885 VG4 HIGH-TEM"ERATURE GAS 400 LED REACTOR SAFETY STUDES FOR THE DIVISCN OF ACCIDEN T Heavy Secten Steet Techno#ogy EV ALU ATION Qwarteth Progre .s Report, Octot>er 1 -December M84 NUREG/CR 3978 TE NSILE PROPERTIES OF IRRADIATED NUCLEAR GR ADE PfiE SSURE VE SSE L PLATE AND W E LDS FOH THE Handttock FOURTH HS$7 IRRADIAfiON SER:ES NUREG-c544 R02 A HAND 000ey>F ACRONYMS AND iPWTIALISMS NURE G/CH 4 H5 EFFECT OF STA.Ntf SS STE E L WELD OVERLAY . NUREG/CR-3558 HANDOOOn OF NUCLEAR POWER PLANT SEISMIC CLADDING ON THE STRUCTURAL rNTEGHITY OF FLAWED STEEL FR AGIUT ES Se smic Sascey Magns Reseesch Program. PLATES IN OF NDiNG SER'E S 1 NUREG/CR 4C41 SYSf44 ANALYSIS HAND 800M NURE G /CR .4084 TE N4LE PROPERilE S OF IRR ADIATID MJCL E AR NUREG/CR4041 R01 SYSTEM ANALYSIS HAND 800K GRADE PRESSUHE VESSEL WELDS FOR THF TH'RD HSST IHR4-Harsh Environment OfATON SERIE S NUREG/CR-3663 ASSESSMENT OF CLASS TE PRESSURE TRANS- NUREG/CR 4284 NE UTHON E *POSURE P AR AME TI HS F OR THE MITTER RESPONSE WHFN SUBJECTED TO HARSH ENVIRONMENT FIFTH HEAVY SE CTON STE E L TECHNOLOGY IHRADIATO N SCREENING TESTS StRTS I High Confksence Of Low Protpasselsty Of Padure Hatch Seas NUREG/CR 3952 SEOVOY AH EOutPMENT HATCH SE AL LE AM AGE NUREG/CR4D4 AN APPROACH TO THE QUANfl> CAf ON OF SEIS. U,C VARG.NS IN NUCLFAR POWL R Pt. ANTS Hasardous Weste Regulation r NUREG/CR-4406 AN ANALYSIS OF LOW LEVEL WASTES Review of High Efflciency Particulate Air Filter l Hazardous Waste Regulations And kier tAation of RMoactive und NUA EG, CR-4225 $UMMARY OF EFFICrE NCY testing OF ST AND. Wastes Final Report ARD AND HIGH CAPACif Y HIGH EF FICIE NCY PARTICut, ATE air Health Fit TERS SUBJE CTED TO S4 MUTATE D TORNADO DEPHESSURflA-TON AND E aPLOStVE SHOCM W AVE S NUAEG/CR 3384 BiOLOG3 CAL CHARACTERllATON OF RADIATON DPOSURE AND DOSE ESTIMATES FOR INHALED URAv04 M'tt. Hegh integrity Contaener ING EFFLUENTS Annual Progrese ReportApre t9R3. Mych t944 NypCG/CR 4215 TE CHNCAL F AC TORS AF F E Ct h) LOW LE VEL H mhEH e M W ASTE FORM ACCEPTANCE CRf TEH4A WUREGICR 4214 HE ALTH EFFECTS MODEL FOR NUCLEAR POWER H4gh Pressure Electw>n PLANT ACCJDENT CONSE QUE NCE ANALYS;S Part 1 tratroducton.ir'tegration & Summary Part H Soorth R4sre For Hea'in NUREG/CR 4M3 HIGH Pf4E SSUf4 iNdCTON OF Mf L T F ROM A ENects Modeis Rf ACTOR PRE SSDHE VT W4 . fHF DKCHanG3 FmasE Health Phye6ce N N "" M '** NUREG/CR4239 ANALYSIS OF THE A8fuTV OF CURRENT HEALTH WWG 3426 W2 NWAL AND WD WW W 1/2 SCALE PHYStCS INSTRUMENTS TO PREDtCT DOSE IN DPOSE D INDIVO. TEM FAOW Data Repod UALS NUHE G/CR 3703 ASSE SSME NT OF SElf Cit D TRAC AND HE t APS NUREG/CR4375 THEORY. DESIGN AND OPrRATON OF LouiD CALCUL ATONS FOR OCONEE t PHI 550HilED THE AMAL 6HOCM METAL FAST BAEEDER RE ACTCHS INQ UD NG OPE RATONAL STUDY HEALTH PHY$1CS Hegh Temperature Heat Affected Zone NUHt GNR 2311 V04 N4 SAF ETY RE SE ARCH PROGR AMS WON NUREG/CR3611 V02 EVAL.UATON OF WE LDt 0 AND EE P A,R. SOHED fM OF F ICE Or NUCL E AR HEGut A f oH Y WELDED STAiNLLSS STEEL FCH LWR SERVfCE Awa Hegvt for RF SE ARCH Ouar*er+y Pwese Hoport Or. toter 1 Decemtier 31, 1964 1964 NUHf G/CR 319? V0t RE ACitON (iF TWFf N SOVE CESIUM IOC)NF Heat Balance CCMPOUNDS AND THE HE ACTOR MATIH'ALS 304 Sf AINt E SS NUREG-1107 TPOWR2 THERMAL POWER CE f EnM NA TON FOR S f f EUNCONE L 600 A Silver Volume ICesium Hyrtresoe Heec, WESilNGHOUSE REACTOR $ VERSON2 User e Gwe tons NUHE G/CR 3Ae5 VO) HIGH TI MPEHATURE GAS COOL ED RF ACf OF4 PtREG # 06 22 R4 STANDARD HEriW PLAN FOR THE HEVitW OF SAFETY AN ALYS.S REPORTS FOR NUCLEAR POAER
~~" D *" "# M ' W"*' 3" PLANTS LWR E$ tion Rev+ ort 4 70 Secten 6 2 2. ' Cor44merwnt Hea' M6gh fernperature Combustion NU E UO2 D.w.~~
R 51 HE D9 DPER: MENT Heat Removal From StratAed pfp f3 # # I U## # ^WI* #T U ' NUREG/CR M16 V02 REACTOR SAFETY HE SE ARCH Ouarterly High Temperature Graphete Reactor ReponMJune N NURFG/CR 2Mt VOS Nt RAF E f Y RE SE / HCH Pfuy, HAM 9 SPON Heat Transfe, SOHID BY Of FICE 09 NUCL F AR 4t gut Af 0fM NUREG/CR-3193 FORCED CONyt Cf N E NONF Outt inn,UM POSr. Hest AHCH Owterly Prgress Hepart January 1 Marcri 31,1p5 CHF HEAT TRANSFER DPER MENT DATA AND CORHE LATON CGMPARISON RFPOHf I "'""8 F "'* NUHFG/CR 3206 TRAC PD2 DFVFLOFMFNTAL ASSFSSMrNt NUHf G/CH 39% V91 f Uti Pt HFOHMANCE ANNOAt HE P0H f f OR NUREG/CR 3499 TWO DIMENSONAL MOfA LING 08 IN f H A SUf3 AS 1983 SEMBLY HE AT TRANSF f R AND BUOYANCY INDUCED FLOW HE. DISTR:8UTON IN LMF RRS High-Dose Jot > NUHEG/CR "633 V04 THAGBDt/ MOOT AN ADVANCE D BEST F Sii. NunE G/CH 4254 Dat # A f ONAL Dost nt DUCTON AND at AnA MATE COMPUTER PROGR AM FOR BOill4G W A TE R Af ACf 0H Af NUCLE AH FT)WEH PLANTS Study ()n H,gh D%e Ms.Hadweee THAN$if NT ANALYSIS Volume 4 Developmental Asseurecr*t Hareng Arvj ALAHA irw.arttives
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168 Subject index Higrt-Effic6ency Particulate NUnEG1CR 3688 V02 GE NE H A fiNG HUMAN F4E LI A81U f f E Sil-R$U AE G/CR-4111 A COMPARAT!VE STUDY OF HEPA FILTER EFF1- MATES US NG EXPERT JUDGV(NT Volume 2 Apperstices C:ENCIES WHEN CHALLENGED WITH THE HMAL AND AsR JET- NUHE Ge CH-4016 v01 APPUCATON OF Stim MAUD A TEST OF AN GENE R ATE D 002 ETHYLHEXYL SE8ECATE.D12 ETHYLHExvL IN TE R ACIIVE COVPUTE A DASED UE f HOD FOH OHGAN@NG PHTHALATE AND SOOiUU CHLOHiDE E R PE RT ASSE S5ME NT OF HUMAN PE R F ORMANCE. AND PSUREGiCR.4264 INbESTIGATON ON HGH EFFIC:ENCY PAHTICU nE LIABitlTY VNume t Ma.n Heport LATE air FILTER PLUGWNG SY COV8USrCN AEROSOLS NUHEG/CH 41M NUCLE AR POWE R SAF ETY 84f PORTING SYSTE V iMPtEMENf ATiON AND OPE AAiONAL SPFC;FICAfCNS HigPt-Level Waste NUREG/CH 4;p0 THE E F F E CTS Of SUPE HviSOH E WER>ENCE AND NUREG-0946 AN EVALUATION OF RADONUCLICE CONCENTHA- AS$$f ANCE OF A SHJT TE CHNCAL ACWSOH (STA) ON CHEW TONS IN HIGH LEVEL nADiOACTi%E W ASTES PE HF ORMANCE IN CONTHOL HOOM Si%)L AYOCS NUREG-1046 OsSPOSAL OF HPGH. LEVEL RADOACTIVE W ASTES IN UNSATUR ATED 2ONE TECHNICAL CONSOEHAtlONS AND AE' Human Factors SPONSE 'IO COMMENTS NuHEG esco 18 2 A0 STANOAAD Af view PLAN FOH THE HE Vif W NL,AEG/CP OO62 PHOCEEDINGS OF THE CONFERENCE ON THE AP. OF SAFETV ANALYSIS REPOHTS FOH NUCLEAH POWEH PUCATON OF GEOCHEWCAL MODELS TO HIGH LENEL NUCLEAR PLANY$ LWH E 4 tion 84em.on 0 Yo SHP Sante 18 2. ' Sa'e9 Par am-W ASTE REPOSaTORY ASSESSVENT S NUREG/CRJ851 V03 PROGHESS IN EVALUATON OF HADONU-CUDE GEOCHEU' CAL INFOAVATON CEVEiOPED BY DOE HGH-gy $3fi%4T2 ..AF Eiv HE SE AnCH PHOGH AUS SPON-SOHED Dy OF F ire OF NUCLEAR HfGULATOHY LENEL NUCLEAR W ASTE AESPOSITORY SITE PROJECTS neport nE SE AHCH Ouyte-N Proces Heport AprJ 1. June 30144 CR 2m W M W W M SI ANH PHWHM M NUAE CH St V EVALUAflON OF RADtONUCLOE GEOCHE Vo OHV CAL INFORMATON DEVELOFED BY OOE HGH-LEVEL NUCLE AR W ASTE REPOSITORf SITE PROJECTS Annual Prowess Aaport F ' p g , ,p QA[h $ Wm MH1m va N4 SApEfy HE M AHCH V6H A VS VON SCHFD Bf 08 FiC E Of NUCLEAR Hf Gut ATOHy NU C 3 L P G TE AM PE AFOAVANCE OF MATERIALS E AM:HOua% NnuRem NWr i Ne%er 3t USED FOR HtGH LEVEL W ASTE PACh AGING OuarterN Heport. Jury.
*"I O'CH 2311 V95 N1 %AF E TY HE SI ARCH Pf*CGHAus SPON NUf O CR J V03 LONG fERV PE AFCHMANCE OF MATEn:ALS SO*4 E 0 HY OF F *CE OF NUCLE AR pf. Gut.A f DH y USED FOR HGH LE%EL WASTE PACR AG.NG Ouanecy #if SE A HeH Ouarterly Puyest Accort Ja wary 1 Mae 3 9 iW$
p p g g, % ,,%,,g4 NUHEG Cn 34n1 ve? Nutt t AR POWf R PLANT Pt HSONNE L QUAll NUHEG/CHDJO V04 LONG f ERM PE AFOAUANCE CF MATERIALS The Tau Ana'yte Pro 4ng USED FOR HiGH LEVEL WASTE PACKAGiP G Annu at Aerort Aprit I'CATONS AND inA4NG TAPS 9** t t#4 AordtMS NUAf G CH 1M4 VA.NTE NANCE Pf ASONNf L PE RF OAV ANCE S.V NUAE GICA-4134 AE POSiTOA V E NV:RONVENTAL PAAAVEffng RELEVANT TO ASSESSING THE PERFOAVANCF CF HGH LEVEL UL A 70N (U APPS) UOE f ( U%fA1UANUAL NUHE G CR M17 UULTIPtf %( QUE NilAL p Alt VHF WASTF PACKAGES WTEL Evesaw cA Amt r, mores im Hean Frme Detm,ety NURE G rCH-4161 v01 CAIT6 CAL PAH AVETEMS FOR A HGH LE%EL Nb"f 0'CH 3M t HOVAN F AC TCHS HLV-{ h 7 04 M g( H[ Aa j. WASTE AEPOSTCAr bovne t aawt NVAEG/CA 4tM FRACTUHE !N GLASS >HGH LENE L W Asit CAN S- CINI SEQUINCf ANAL W 4 Nuht G cn m r COMPuf f Helt D AP.NLNO A f 04 sysf f u9 TENS NUM G CH 4016 s0t APPt tr, A f ,GN 08 M 1V V Aun A T E S T OF AN NUHEG/CR-4236 V01 PAOGHE SS IN EV ALUATCN OF AACONU. iN T E H Aci tvt. CCVPUTE H c A%,0 VE f HoO .nH nHGANeNG CUCE GEOCHEWCAL INFCAVATON CEVE LOPE D By DOE HiGH. LEVEL NUCLE AR WASTE REPOSTCHY SITE PHOJE Cf 3 AE PGAT EnPf47 A%E %VE NT C# HUM AN PE RF OHM ant E AND FnA nrTOAF A CECEV0E R 1M4 AFUAH4 f f Y VNr"# I Ma'n H*Put NUREG/CR 42M VC2 PHOGHE M 'N E V ALUA f CN C# H AC'CNU NUH[ GeCH 4N1 4AF E M / %AF ( MAHD1 ME HAC f TONS DUH N3
% A8 E f y DE LAf( D E VE HGI Nrl 5 Af NUCt f AH Pf %t H W ACf 04 CUCE GEOCHE M' CAL **F OAV A f CN CE V E LOP 1 O o f DOE H'Go LEVEL NUCLEAR W ASTE AE POSif 0AY SITE PHOJECTS Reput for F ACUCE S Jarwary Mcch t es NUHE G/LH 4'01 USM Of HLVAN Nf UAfillTY ANALf%9 PHDH NUHLG CR 4/37 MOD UT f CF RAD ONUCUOES iN HGH CHLCHOE AUU4f C HiSE A%5f S5MI NT HE SUL f S TO Af 50L VE 6 E HSOPe E NyinONVE NTS NF t Pi ASOHMA'# I 6%Mcf $ THA f CDut D Al FI Cf % AF f fy NUNEG/CH 4301 NGH LEVE L W AS TE PDECLCSUHE SYSf f uS NtpfG'LH 4t04 UAMI N AV [ P( HSOf4 Nil PI HIOHMANCI SU SAFE TY ANALYSIS Pmasa 1. Feel Apt Ut AflON (VAPPN UFCf 1 7 et i vaivatevawf aw PilAEG/CH-4119 VOi LOPdG-f E HM FI 645 CHM AP+Cf 08 UAfEH ALS NuntGfH47.4 A %fLtLf Hfvt W 07 f HE Htti NT 09?1 taalt USED FGH HGHLEVEL W ASTE PACP AG NG First Ouwwy f4 HA CCHAL HE r.f AHtH UtlHAfbHE ON T H A 4 NCe EV JL A Heport. Year Four AprNue s tiGS TODS P#uME GrCH 4217 HLVAN ( Ny Nt I n.NQ quG L,NLg f04 THE HigNSpeed Semulatron y v At UAfiON AND AW UVF Nf CF Vit f n D %Pt A v uppfi NWEG/CR H43 f HE B AA PLAN ANALYl! A NLHi *( H 4/am fu aVUF NLA f eCN4 f on NHC Pet g f UN Snp i M A ' Nur t i A H PC W f H Pt A N T 4 H6gN f emperature NyH,HI GC4{HJUNG 412 HEANDWNSIOwl Hfutf F.Vff v Am Af auN f wt H Ve NT At NUAEG/CH 4037 DATA SUVVAhY HLPCHf FCA 8 S%CN PHOfA>CT 4Q %Vf NT OF AN I sFf Hf Systi V F OH Nutt i AH Hi AC fnfi ACLF ASE TEST He $ g,,g47ggg HegNTemperature Gae-Ct.oied Mentor gom,n p,,,,,,,nc , o,,,
NUAEG/CH 345 Va4 HGH T E MPE HAfunf GA5 COOT f D Pt ACf 0H %M uH Mto Vf uf 4, A fiCN GF A tetNAN HfliAn utv DAT A SAFETV S TUD'E S FGH THE DmsON OF Accitt NT mA W f m 9 w PCs r> p m ptu % tf E VALUA TON Quarter 4 Progress Hatort. Cedee t (M erat +r g gg n g m 31 1 p4 NUHEG CR 4402 V01 H.GH fE VPt HAf uHF GAS COOLt D AF AC TOH g,,,n g,3,,,n,, SAFETV SfLDfS pOn THE OtW9CP4 OF A % Di N f g g, q q g yg 9, ggy,q g ggr;g g,ggg3,M If f I 57 VA f(We US NG EV ALUAf CP4 Quarter'y Progtess He CWrt J4^4ary t - UNCh 3 t t MS g rg qsg g g g p4y gq pr jnig NUlit rp( H VM VQ/ U A N fl P A Nl [ l'f HSi ANT l F( 48 ( liv Aper:( 4 S W # ICD ##" " #I "*"#I " " " ' " ' NA c 4341 H6H PHF S WHE IN J G T04 OF VF L ? F HCM A t rME Nt MHWoHf Av M Newrv fMf NG REACTCH PHES5UDE vt $5Et. . f ME DISCHARGE PHASF NUHf G/r.H a we HLVAN Hr t iAmt if y DAfA t, A NN ( ,g,4Nin He siin Human tngineenne NU"' G'CH 41' O %II L W 'C A fif /4 OF A HUM AN Nf LIA01lf t D A T A NUAEG%R 422/ HI N A Pd (fe ra P4f f A 6P G GtJ.Dilif 4E $ F OH THE OAN" f OH (LNht C f W's HH A Sf Guf P4 f 3 or On A3 5 oH Pain t g EV A4.UA TCPd AND AMES;VFter Gr wtrO De,PLAv Ur e ts AH POWt H Pt ANTS Human Error Human M*'iabehty Analysse reuntrPCn m + Hovah f HHOA PHOHAH!Uf f f SfiVA f FON usWG NUHf fa'f H 4 t TP t1%f % OF HUS,e4N Nf t lana lty AP4A(y%s Psit;H i rrt Ngs f 9 VP NT H(P6H fi N'#L 4 LH l'A9 V01 GENC H A T NG HUM A P4 HEUAH.Uf f ( $fi ArbU$f C HAN A%i %%Vf fdf' Hl hot f % T O HF %t)t W P( H%N VA!! S U%Paq EyPE Rf JULg3Vf N f ten f Veri HaHort Nt L Of HF OHVAP# t t%5Uf 5 THA f LOUT O Al f 4 C T SA8 4 f V
Subject index IC9 Human Re46absty Estimate IE Bunetm 80 25 NUREG/CR-3686 V0t: GENERATING HUMAN REUABtUTY ESTL NUREG/CR 3794 CLOSEOUT OF IE BULLEilN flo 2S Of't R ATING MATES USWG DPERT JUDGME NT Volume 1 Main Report PROBLEMS WifH TAHGEf ROCK SAF L TY HE UE F vat %ES AT NUREG/CR.3688 V02: GENERAflNG HUMAN HELIAD UTY ESTO gw n g. MATES USING EXPERT JUDGMENT.Voivme 2 Appereces IE Bulletm 8141 Hydrogo" PdUHEG/CR 4006 CLOSE OUT OF IE BULLE fin t t 01 SURVElLLANCE NUREG/CR 3312 MARCRHECTR ANALYSIS OF SELECTED ACCI- OF MECHANICAL SNUBBf HS DENTS IN AN ICE-CONDENSE R CONTAWMENT NUHEG/CR 4121 EFrECTS OF SUtFUR CHEM:STRY AND FLOW IE Sufletin $242 RATE ON FATIGUE CRACA GROWTH RATES IN LWR ENVIRON- NURE G- 1095 EVALUAfiON OF RESPONSES TO IE BULLEilN 82 MENTS- 02 0egradition Of finaded Fasteners in Hractor Coolant Presswo Hydrogen Abeorption 6 undary O %ssuW WaM Reactor Rants NUREG/CH 4422 A REVIEW OF THE MODELS AND MECHANISMS IEEE 343 Flame feet FOR ENVIRONMENTALLY. ASSISTED CRACK GnOWTH OF PHES-SURE VESSEL AND PIPWG STEELS IN PWR ENVIRONVENTS NUnEG/CR 4112 V01 INVESilGATION OF CABL E AND CABLE SYSTEM FIRE TEST PARAMETEHS f ash A ItEE Flame Test Hydrogen Burn IGSCC NUREG/CR-3254 HECTR ANALYSIS OF EOUfPMEN r TEMPERATURE RESPONSES TO SELECTED HYDROGEN BURNS W AN ICE CON. NUHEG 1061 VOS REPOAf OF THE U S NUCLE AR REGULATOHf DFNSER CONTAWMENT COMMISSION PIPWG F4EVIEW COMUlff t E Voeume 5 Summary NUREG/CR 4146 SiMULATCN OF AN EPRINEVADA TEST SITE P'P'ng Rewe Cominee Cmsons and Remmmendatene ( (NTS) HYDROGEN BURN TEST At THE CENTRAL RECETVER TEST gg7 l FAC W NURf G.CH 42/0 REUA3;LITY ANALYSTS OF CONT A:NMt NT ISOLA ( Hydrogen Combustion TION SVSTIMS NUREG/CR3721 V01 PAESSURE MEASUREMENTS IN A HYDHOGEN COMB TON ENV RONMENT Hy Jrogen-Ast Comt)ustion fest Sereg N RfO/CH3831 THE l 81A f HEllABiUTY DA A f+Ast FOH NU CLEAR PL AN T COMPON[ NTS inteem Report Deses Hydrogen Generation Generators Hatteres Chargers And m.mters NUf4G/CH-3361. THE EFFECT OF WATER CHf U $f RY ON THE RATES OF HvDAOGEN GENERAflON FRCM GALVANflED STIEL lee Condeneer CORROSION AT POST.LOCA CONDefiONS NL"f G' CR-3112 U AAO4 HE CTR ANAL v$ 5 OF SELFCTED ACCl-NUREG/CR 3803 THE EFFECTS CF POST LOCA CONDitlOr.S ON A DE NTS IN AN ICE CONOf NSE n CONT A:NME NT PROTECTIVE COAtlNG (PAWT) FOR THE NUCLE AR POWE R IN Nb54t G/CR 3954 HE CTH ANALYSIS OF f Oti10MI Nf it Vrt AATUHF DUSTny HtSsOPest S TO SFLECTED HYDHOG( N BUHNS (Pd AN ICE COPS. CE NSE A CONI AWWI NT Hydrogen Transport NtJHf G/CH 413n ICE L4 A CODE FOH A&ROSOL PARfict F CAP NUAE 3 /CR-4020 HMSA COMPUTE R PROGR A M IOR TbHE IN ICE COMPARTUE NTS TRANS;ENT.THPEE DIMEPeSONAL MniWG GASES litmose Hyttogen Onygen Flammatweity NUHE G/CH 4 m SfE Gt Nf V:()L F AUL7 206E MMOLHI AND #L. NUREG/CR 3237 CONTROL OF E sPLOSfVE MitTU54ES IN PWH UNOis W ASTE GAS SYSTEMS Immurwty l Hydrogeolog#c Sete Charactertsation NURE G /CH 4133 NUCLT AH POWE R SAFf TY HFmHTM SYSTE M f NVAEG/CH 42S t V01 MITKiAfivE TECHNIQUE 4 FOR GROUNO- iMpt(M(NT Af SON AND Ort n A f)ONAL Ssf CMCA flON9 WATER COPafAMWATON A$sOCIATED WifH st VtHE NUCLEAR l ACCIDENTS Volume 1 AnaWe Of Genenc Me Corotoes importance Renteg l NUAEG/CR 4251 V02 MifiGAf tVE TICHNIQUE S FOR G540UND- Nang G cn 4 44 iMpOHf ANtt n AN,M D ALE D ON AGWG CON WATER CONTAM NATON ASSOCIATED W f H SEst nE NUCLE AH s2DtnAfiONS os Cr.,MuGNt Nf 3 Wu bbt D N phon 4wu<;f c ACCICD4TS Vo ume 2 Case Stafy Anawo Of Hrwosupc Charact*. s Hism AssFSsMENTS iration And Mitigative Scr' emes importateon Hydro 609Y NUHl.G tR A M1 THE 8 i A%HfUf Y Os Df flCf WG THE IMNHf CF NUR E G 'CA 1134 FtEL D AND THECHEilCAL 'NVt SfiGAf 0NS OF tAAuf HOH:ll D R AD OACfivt M A f f RIAL S rN to f Hf UNiftD FRACTURED CRfStALUNE AOCR NE AR CHACLE.AHi/ONA. gf Af gS ICEDF NunEG49'A DRrf FC RE ASSE SsME NT cf THE f t CHN' CAL i Ast S , 493 g my 94 ,gg,, g y,yg ggy, gg , gg ug,q,L, FOR ESt.MAhNG SOURCf TE AMS fDraft Fiapnr' F or Commeno NUAEG/GR e t30 Kill 4 A CODE FGH AE HO50L PAH fiCL E CAP " ## N 9 ^ "'"* funE IN #CE COMPARf MrNT5 in $du temg IE Buttetm f9-04 NU"I O'C" W UII"OO9 O' U h V 1 M GROUND d'AflH W 4 NUPEG/CR 4003 CLOLIOUf Or IF 90t Lf fin f 4 04 4 COH4t cf f AMhA f104 f RCU iN 9f U LF ACH UuAN.UU M NM ivai n, cort WEIGHTS FCR SWWG CHf C8r V ALVE S MANUF ACTUHt O fit '"' *
VELAf t t NGWE ERM COAPDHATION g *[ , , r IN imt Hf WM m p p p m W G Bulletm f9-09 ClfAA PLAPO CCVPONt N f 9 'r*r'm H+ ,rt tweet NUREG/CH 3f91 CLOS (OUf OF ff HULL t TIN 79 01 F Att UHf OF Gy NonwaWe Da nmewe. Umtgers Anit in,cetere f YPE AK 2 CWCUlf OnEAzFRS W sap tfy HILAf tO SrSitUS ,
IE Bunetm f912 NUHF G4H 4410 Cos*Hr N T Mr f HODot f u S pre A% um NUAEG 0905 CLOSEOUT OF IE Gut L L f W 1* 12 SHOH f Pt Hoo THf POff Nf>AL F OH f AHh*AAna r iNLots D LKm[ f AC tK.W W SCHAMS At EtOruNG WAf t n ht ACf0HS SOlt S O Sullete f9-26 lac 8 dent NuntG/CD 4004 CLOSE 007 Or it put t F. fin 11 X F ArtU4f 4 OF NuHe G orm vor No4 ntvonf fo ceNon,sq ON anNonUAg WEShNGHOUSE BF D Rf L AVS W SA$ t f y Ar LAfIO S f STt US OCCUHHf PdCl 3 Or um tw mt e tM4 f t Dunetm 8012 incident Data Ana#yese NUPf G/CH 4004 CLO% TOUT OF Ir DULL Ef14 flo t2 0t CAY MfAf NU4t G/CH Mt t H ArpOAC f 4f MA f g HiAL th AM, M wg y f nen D(MOVAL SYSf t M OPiHA01Uf Y Of NT DATA ANAt VN10 Hor. HAM
170 Subject Index incone8600 NUREG 115$ V02 RESE AACH PROGRAM PLAN Steam Gene!< store NUREG/CH 3197 V01 AEACTON BETWEEN SOME CEStUM ODNF NURE G/CH N49 V02 EDOY-CUHHENT INSPECTION FOH Sif AM COMPOUNOS AND THE AEACTOR MATERIALS 304 ST AINL E SS GENE RATOH TUUING PROGF4AM Anrual Progrote ficport For Perni STEELJNCONEL 600 & SUER volume ICeesum Hydrou1e Heac- Endng Decenter 31.1984 tions NUHEG,CH 4170 AN ULTRA HIGH SPEED HESOUE P640CESSOH FOH SAFT INSPECTON SV57E M ' MAGE f NHANCE Mt NY Independent Assessment NUREG/CH 4234 V01 AGING AND St HVICf_ WI An OF ILECTRO
% REG /CH 3919 1 R AC PF t / U001 INDEPE NDf.NT MOTOH OPERATED VALVtS USf 3 IN ENGiNtEHE D SAF E TV F E A-ASSESSMENT NEPTUNUS PAFSSUTER TEST YOS TUHE SYSTEMS OF NUCl[AH POWf H Pt_ ANTS P11HEG/CH 4252. 4NDEPENDENT ASMhMENT OF TRAC P02/ MOOT NyngG,cH 4440 A HEwt W Of t ME HGF Ncv Dit.SE L GENIHATOH COCE WITH BCL ECC BYPASS TESTS PERFOHMANCE AT NUCLE AR POWE R PLANTS Independent Safety Or0snisation ,,,,,,,,,,y,,,,,,,
NUREG/CA 4tS2 AN INDEPENDENT SAFETY OAGANilATON NUAE G /CH 4344 INSTRUCTONAL SAILLS EV ALUATON IN NUCLf4 Indones AH INOUSTRY TRAINING NUREGON4 V09 N04 AEGUL A TOHY AND TECHNCAL AFPOATS Annual Complaton For 1984 instrurnentation NUHE G 0304 VIO NO2 AEGULA TOHY AND TECHNCAL NUHEG/CH 4239 ANALYSIS OF THr An Lify OF cur He NT HE ALTH AFPOATM Cornp.dation For Sacormt Quarter t#5 Apria-Jurie PHYSICS INSf HUVf Nf 3 TO PRE DCT DOSE .N E XPO%f D INOlVID NUREG 0r50 v10101 INDE RES TO NUCLEAA Ht GULATOAV COM UALS M'SSION ISSU ANCES FOA JUL Y SEPTE UUE A 1%4 NUHEG-07SO V20102 INDERES TO NUCLE AR RE GutATOHY COM- Instrumentanon Circuit M'SSON ISSUANCES FOR JULY DECEU0E R tM4 NUAE G 'CH 386.1 A%f $5MrNt OF CLA%S tE Pnt ssUnr THANS NUHEG4750 V21 #01: INDERES TO NUCLE AR HEGULATOAV COM W:7ff A Rf SPONSE WHEN SUfDECTE D TO HARSH I NWHONMI NT MoSON ISSUANCES FO4 JANUAAY M ANCH 1985 SCHE EN<NG TE Sf 5 NUREG0/50 V21 tot INDEuS TO NUCtE AA REGUL ATOAV COM MISSION ISSUANCES FOA JANUARY MARCH 196$ Integral Systeme NUH GeC P 0064 VO t PHOCE EDiNGS Or iHE fWf LF TH WAig H pE Mdlana ACTOR SAP E TY AE SE ARCr4 Ne opM A f 0N MF E TING NUHEG/CR 3t4$ V01 GE OPHY SICAL INVESf 0AflONS OF THE WESTERN OHOINDI ANA HEGON ANNUAL HE POHf (Oocte' tritegrated Control System 1962. Sectemter 1983 Volume 3) NUDIG/CH 3 1 F AltOHE MODE S AND E F F f CTS ANALys S (F MI Al inert Cae Os THE ICS/NNI E LICf MC POWE A DISTHPUflON CiHCuliny A f WUREG/C A-4074 THE PERSOAMANCE OF DEFECTED SPENT LWR THE OCOPd E t fiUCLE AA Pt ANT FUEL ACOS IN INERT GAS AND DAY AIR STOHAGE AfMOS PHERES lategrated Leet Rate feet NUHf G'CH 4220 RELIAiOLif V ANALYSIS OF CONT A NME NT ISOLA. Inftuence Coeffic6ent ION Sysff MS NVAEGs CR 3723 STAESS-tNf f hSif v1 ACTOH INFLLENCI COE F Fi-CtENTS FOA SURF ACE FLAWS IN PAESSURE bESSE LS Integrated Plant Safety Asseeement NUAtG O424 50t INTIGHAftD Pt ANT SAfEfy A%M %Ut P4f Informatson Engmeenng HI POR T SYSf f M A TC I V ALU A flON PHOGHAu Mit L STONF Nu-NUREGeCH-4122 Vot COAPOA A TE DA T A NE fwCA n (CDN) DA T A CLE AR POWf H ST ATON UNtf 1 Dos set No S0 244 teuwineast Ph AFOtJtArMENTS TAsm Vna 1 Feverpnse Martes ci,,, g n,,gy Corvany) NVAEG/CH 4322 V02 COHPCHATE DATA NE f*Ons (CDN) DAT A NUHt G on20 DHFT INflGHAflu PL ANT SAFITV ASS / 5SME NT Ni E /C 4 4 22 V0J CN IfE A NOAM ICONt OAT A HEPOHf Systf MATC E VALUAf 04 P8*)GHAM I" " * I ' "' "O SAN ONtVHf
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AFOU1Hf MENTS TASM Vnd 1 Oate Model NVAE G/CH-4322 V04 COHPOAATE DAT A NEtwo84n (CDN) DAT A "'" C8'd*'me E ttienn CompenW AEQUlHEMINTS TASK Vol 4 Pre 8+menery Stratege Data Pran g,g 9, informed Opmeon NUntG/CH 17to L AliGHAfCev 5YUDit S OF A UHf ACHf D NiJCLI-WURE GICP 00:0 PAOCEEOiNGS OF THE 1944 StAfiSTCAL SvMPO AH WASfE HlPOSITOHY tN RASAL T 5tuM ON NAf 0N AL ENE AGv ISSUE S Interfectal frenefer Mtegraf Mfrerad Anatyese NijntG/CH May iIVt. . ANO vf 4 UMF A yi H AG( 0 CONSI HV A flON NUnf G'C.H h44 BOLOOCAL CHAn ACTf nitAfCN OF AADIAf ON IOUAf ONS FOH Uut f PHAst #iOvv Pat One 4,com W trw n E*POSUME At#O DOSE ESTiMAf f S FOR INHALt D UMAPUUU M't L- tornal Sotwf Structures IP#G EP FLUE NTS Armual Progrese Report Apr41M 3 Mach 1944 Intergrenuser Strees Corroeien Cre<6mg Mhelation Nyn(0/Cn.1gt) yop t y Ag y A hopg Of Wg (L( O Apen (qq pa,n NUME G/CR h44 BOLOGCAL CHAAACfERt2Af 80Pd OF AA0iA f 0P4 WI L t4 0 37 A:NLI SS S f t t L P 084 LWet St H ywy Antwei H,p,,1 p o, ErPOsunt ANO Dost E ST:MAfE S FCH IP4 Half 0 URAN 10M MILL- , g4 ING EFFLUENTS ANwel Progrese Report Armo 1M1 Mach iM4 NUHgGe(H 1613 y01 N1 f y At UAfiON or Wr t pt U ANO Hg p4m WELD 4 D ST A;Nt f SS StiI L F OH LWR SI HvK t %ernav%el Haport ~ ' For O(f her 1M4 Through Marth 1945 a 1 VO Rf %f AtK H PHOGA AM PL AN Stearn Generatore NUHEG 115$ V04 HE SE AHCH PFt(14 HAM PLAN Non De *ructwo is g PdUHF G/CT4 41$ ) APPilCA f CNS Of F OH( IGP4 l'n&H A ntt lSilC Mgpection SAf I f f A%f.%Uf Nf 1. IPt H'E Nt I f O f HI U S P4ULLi AH Hf G NUPEG 0041 V04 404 LICF N'd E Wef HACfC4t AP40 VENOOH IN Ut Af OH f 8'HOLf SS PT CTON S T A f US DE POR f O ete'y naPort Ortoter Dwemter S, g4 gwn,,, go,g Mternational Standard Problem 13 NUHE G 4040 VO4 N0t LICE NVf CONF nACf06 ANO Vf NDOH IN NUHf Gd H 4 9 f 5 INf t HNAflOPdAL Sf ANDAHD FHOHL F M 93 (LO8 f
$1T C TON STATUS OfPOHf O,atar'y Heport.Jamery Wee n E *Pt H ME N T L2 % F emi Corycen H*twel 1Mi(WMe hontl Inteounal Absorption NUnf G-0040 Vov Nn2 LICE NS4 E CONfHACTOH AND VtNUOH IN SPE CT ON StAfUS HEPOHf Quartar'y Hermet Apr4 June tMi NiSI 4/LH ang oArif Ho<Nf f 9 t NAt. Anv>HPf ua of Pg ytoepoM (ww.te g6ry iN UK F $4 A f 3 AP40 [f)49 Aps,1w e%n to I statMevy venpg os ii Nun t, G- 1144 Nf)CLFAH PLANT AGrNG Df'J ANCH (NPAIN PHO F r/ Wutm fMutorwrn GR AM PL AN Nunf 4 tig1 INV1GTON Af POHf OF UNAuf Hf)$U/f D PO%f S Inventory SiON AND USE OF UN'J ALE D AMEHC8';M 24 f ANU SUBSI QUs NT NuHr o/CH ws i v AtuAfioN OF THf NArnoAcfivl in,r n f ort y GOPtf t%AflON J C Hayne Corvag Newet ONo IN ANO I 9f MA flON Of VJf DPC Hf L t A*,f P sH MA f Hf W A9 tg IN
Subject index 171 EIGHT TRENCHES AT THE SHEFFIELD LOW LEVEL W ASTE BUAIAL ECHT TREPdCHES Af THE SHEFFIELD LOW LEVEL W ASTE BURIAL SITE. S6TE Inventory Difference Data Issues NUREG44'.to VOS No t . UCENSED FUEL fAC)UTY STATVS REPOHT inventory Deererte Data January 1984 June 1964 (Grav NUREG/CA 4382 CONCENfHAflOP45 OF UHANiUU AND THOfouV Book 10 6507 OPES IN URANIUM VILLE AS AND %NE nS' TISSUES invened Annuter % J-entegral Reseetance Curve NUPEG/CH-42ff INVERTED ANNUAL FLOW EXPEn;VENTAl STUDY NUREG/CA 4243 STUCY OF THE EFFECTS OF ELASTC UNLOAD' INGS ON THE Ji H CURv'ES Ff40M COMPACT SPI CIVt NS ledme NUAEG/CH 3455 A COMPAAISON OF ODINE kAYPTON AND XEPdON gre earmg Modulus AETENTON EFFLCIENCIES F08 VARCUS S:LyER LOADED AO- NUREG/CH 4082 v01 DEGHADf D PIPING PHOGHAM PHASE goppyON VEDtA H Semannual RepWtuarch 1M4 Septemtaer t964 NUREGICH 3514 VC2 THE CHEUCAL BEHAv0A OF ociNE IN NUHt G 'CR 40n2 v02 DE GRADt D PtFWG PHOGHAM PHASE AQUEOUS SOLUTONS Us TO t50 C H Haaanon Asen conet.one n Senwannual Heport. Octu,or 1984. March tMi NUAE G / C A-3990- CHARCOAL PE AFORM ANCE UNDE.R ACCIDE NT CONOfCNS IN LIGHf-w ATER AF ACTOH9 Japan NUHEG/CA.4001 ABSORPTON OF GASEOUS ODiNE BY W AfEn NURf G/CP 00S9 V01 PnOCf EDNGS OF THE MiflNHC St1540 W DROPLETS FonM AfiON E RCHANGF Ut E 7WG VOLUME I ton Eachange Resana NUftEG/CR 3AM ANALYSIS Or JAPANt >E U S NVCLE An POWtn PLANf WAWf E NANCE PdVAEG/CH 4t50 EPCOA H AE9N DEGRADATON PESULTS FROM FIRST RESIN SAVPLE S OF PF a AND PF 20 Jott Performance Aad i NUHtGrCH Mt f Of %ELOf'UE Nf u$E AND CONTHOt Or MAIN?E. l HE / "
~~INVE SfiGAflON OF AL TEf4 NATIVE ME ANS TO AC-COUPUSH THE GOALS OF 8:tNNiAL ON CHAUSf R CAL l0AA N $~~
l TON RNE le NU8if G C4 4 3 7$ THE O84 f DE NGN AND OPtRA70N OF L OUiD UAE Cn-2f te STEAM EMPLOSON EnPER UENTS WITH &NGLE " Al DAOPS OF lRON OulCE MELTED wifH A CO2 LASER Part HE AUH Pur5d i H Parametnc Sturhee grypgen frfeested Fuel NUREGrCH 34SS A COMPAHf 50N OF IOpeNF a AfPf 04 ANO 3(agoN PeUAEG CT25 Aci PURUC iP4FOHUATON CinCULAR FOR Ship. HE f t NfiON IFFIC ENCit S FOH V Ar4003 54vt H LOADf D AD-MFNTS OF inn ADIATED RF ACTOR FUE L SORPflON Mt DI A NUHEG CN 4203 A CALCULATONAL UFTHOO FOH DEfERM NWG 90tOGicAL DOSE RAfIS FROM IHHADIATED RE SEAACH HE AC. N"E TOH FUEL NunFG/CM 4 s t$ TH(OHY Dt UGN AND OPEHAfCN Of tIOVO lereested Zircanoy Cladeng Mt.YAt FAST 9Hf f DE 4 Ht AOfOHS *(:LUDWG OPE RA f OttAL HF AL TH PHYSaCS NVAf G/CA TM0 V03 UGHT WATE A AE ACTOH SAFEfv FUEL Sv5 TEVS At SF ARCH PROGRAUS Ouerter'y Progrese ReprwtAs, Sep- LER temtwr tu4 NUHF G tfs22 S02 KiNM f f vt NT Hf MN1f BrSit U Eva#oshon Of l NURf G/CF4 JMO V04 UGHf w ATE A AE ACTOR SAf ETY FUf L Sv$- 8 ret veer huts AM Hav omensatawe f or nrrerove*ents I f tv5 AESE AACH Pf40GAAUS Quer'er'y Progreet Report Ortder. NUHf G<CH 2ono vo3Nt2 Licf Nst t i vt pe r nfPOHf (t g R; l Deceneettp4 U)Mfat A f ON f or urm Of D* **hr iM4 Brtseat6on PeUREOtCH 2000 VO4 N1 ULINMI I Vt Ni AI M M Y (t i fN COMP 4 AfIOP4 For Meth Of Ja%e, tM% NUAFG lt$9 V01 AE%EARCH PROGAAU PL AN Amartor Vessels f(UML QiCH 2rJtX) VM N/ L il [.N%( g (V(NT 84g pOH f (d H) NUMEG/Cn 3472 DAf A ACQtJISifiO4 AND CONTROL 08 f HE HSS T cOvPq AtON Fy urm of t.4,ns, nag M A ES v 6AnADtATOma E stt H Vf Nf At fHf Onn Nyngm/co pyry g4 N3 g(g N5g g gyght pg ggn y gg g ggg NLHIG'CH We 7t N9tt P8+GPt Hfil s OF IMHADIA ff D NUCll AR COMPit AfiON For %mm Of March mag GAACE PHE%SVht VI SM L PtAfE AND WELC1 F04 fME NUHt Gf Cn ye Vo4 N4 Lw g Nst E g yg NT na gu f g g n; i FOUnfM HSST inn AD AtON Sf s4 i S COVP4 A f 0*d f or %w'm Of Apr* 1p'l NUHf G CA 3w) (LICfHCALLY Nf A f t0 E 8 Hf ACTOH PE L L E f' CLADOWG WTf AACTON (PCo S-vULATONS UfrutP#G iHnADiA T' NUHfG T 4 2000 VW PM tlCE'6fF l yt Nf HfPOHf 4 t H) COUW Afu1N For M mth Of May 19mg
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172 Subject index LOCA NUREG 0?50 v20 NOS NUctE An et GULAfOHv CCVMSSiON 63-NUAEG/CAs#633 V01 St. TAAC-BC1,VOO1 AN ADVANCEO BEST ES- SUANCE S f OH NOvf VdEH 1+n4 Pape t 41' ' U2 fiMATE COUPUTER PAOGRAM FOR DOsUNG WAfE A RE ACTOH NuntG 0750 v20 Nos NUCLE An Htuut ATOH, (cVV4scNis f AANSEENT AN ALYSiS SUANCIS f OH OFCI Upt R i#4 Pajas t 57 L t 706 NUHEG/CH 3646 TRAC PF1 INDEPENDENT ASSE55 VENT NUHEG 0/to v21 Not NUCLE AH Ht GUL A TOH r COVv 9 SON IS NUAEG/CR J802. RELAPS ASSESSVE NT OU AN fit A fiv E KEY PA- suANct S FOH JANUAAv tpg Papg 1 JM AAVEfE AS AND AVN fiVE STAfaSTCS NUntG 0 40 Vlt N02 NUCLE AH HL GULATOHY COVV SSON l$ NU AE G/ C A-4044 TRAC PF1 LOCA CALCULATONS US.NG FING SUANrf S FOH f f HHUAHV tMS Pays 2?S 4M NODE ANO COARSE-NCOE INPUT VODELS NUHI G o r50 v21 M3 NUC1E AH Ht GUL A f 0H f CGVV:SSION IS NUAEG/CA 4127 V01 BWA F ULL IN T EGH AL SiVULATON TE ST SUANCI S FOA V A ACH 1*A815 Pe.rs ef t 549 (Fr$f) PAOGR AV TRAC BWA VOCEL CEVELOPVENT voru me 1 N# NUHEG 0750 v21 N04 NUCL E A4 Bf GJL ATOHV COUV SS ON IS-mentaf Vefh015 $U ANCE S FCA APIM 1945 Pape Set 104f NUHEG/CR416 7 FLECHT SE ASET PAOGA AV Feel Report NRC/E PH: NUHE 0 0 ?$/) v21 N05 NUCLIAH ALGutAfCHY COUV SSON G Westeghousa Aerort Numt)ef 16 5U ANCE S FOH VAV t ei Pe ys 104 3 t *+7 NUHEG/CA-419e OVEHv:EW OF TRAC 801 tie P8.4N 12) ASSE SS, NUHLG c150 v21 Me NUCLt An Nt GUL A70HY COVu.ss.ON is VENT STUO4 5 $UANCE S FOH JUNE 19ei Pays t SM I TAA NUAEGJCA 4218 LOCA SiVULATON IN THE NATON AL AE SE AACH NUHIG 0fM v22 Not NU(LEAH DEGULAf0Hf CCUV%GN I4 UNrvf ASAL AEACTOA PROGA AV Postirca44 tion E namwiat,on Ha- SUANetS son JUtv twi Pape t t m ou te For the TPvd Veter$s fest (UT.3) . Second Camra>gn NUHE G 07% v22 Nd2 NUCI E AM Hf GUL A TOH f COUU*S'PON is SU ANLE 9 8 00 AUGUST IW9 Paj+s 1714S' LOFT NUHE G 0tso v12 4n NUCLE AR Ht GUL Af 0Hy cCVV%ON IS NUAEG/CA 3009 SuyVAAv OF THE NUCLE AA AEGULATCAV COM %U ANCf % 7OH St PTEVDE H IMS Papt 48< 3 M A V:5SION S LOF T PAOGR AV AE SE AACH F.N0iNGS NUHtG 0750 v!2 N64 MM E AH nt uutA f oH y ccVV WON is SUANCES FOH OCfC0f R 1MS Pape 611 TM LOFT L24 NVAEG/CA 4t tS INTERNATON AL STANDAAD FHCOLEV 13 (LOFT Licensed Operating Aeectore EnPE A uENT L2 Si Frat Compar son Report NUHt G 0020 VOS N'? L Kl ME D Os4 H AfiNG Af AC f CH% S T A f u$ SUVVAR Y Hf POH T Ca'a As Of Nwe*t'e' M 19A4 Ferar Pme il j Leech feet NUHf G 000 vn9 Mt LK t NM O OPE H AfiP6G Nf AC1CHS &T A f ui NVAEG/CH-4P9 901 LONGTEny Pf arCAV ANCE OF V4f tniALS si VVAH, nf POHf osa A Of pece-tw it tua pa, n#. 9 i USED FCA HiGH lev E L W A$fE PACM AGING Fest Quanecy NUHE G OGN vo9 *('2 UCf %E O opt H A f feG HF Auf C44% $ f A f US A=purtMee Fuut Apre de t 783 SUWAH r Hf PORf Da'a As Cf Ja%4q ll 14A$ edsay Pma i NUHE G OGG vo9 M3 LCI NSIO Ort.H A hM) Hf ACf 0H5 Sf AfUS LeechetHhty SUVV AHY HE POHf Oca At Of 8 et%4ry 24 9 was mas, nona y NUMEGIC A 4t91 LE ACHAH LITY OF A A0f0NUCUCE S FACV NLnEG 0020 V&a Noe UCIN%ED OPlo Af 4 4 HE At f uo) $fAfv5 CEMENT 50uCJ (o W AStE FOAV9 PAGOUCED Af OrtHAf tNG St VV A4 f 4f POH f Dra A s O' Vs"a 3' t *S pm, Bma 4 4 NUCLE AA POWL A f4E ACfCAS NU8+E C 1020 VJ9 Pe05 UCE NSI O OPE H A fi*ci Ot AC f o+44 Sf A f uS
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NUqEG.tolf CRF f FC CONT A NVf 4f Pf h8 04VANCE WOHe deG CcVsm AfsnNFo,V y m Oepe.eme- tp4 4 a GROUP AlPOHf Ors t Mevort 8 or Comener't Nuns Orth Do v04 Nt La i Nsf I t vf Nf Hi a ut f niH) UiVPy A h0N P nr W m Of Januey 1M4 Lees Seforel'eae P40sq uf(H 2ere pje N2 Lei g P4st t (ytNT H( M jH f g[ Hj
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[40spwjag Of P' ester *1 Ws'*r ha*** P'aNS P#UH8ei' 'f D 7 M) v'i4 Na I w I P448 I f blNT Nf P( >81' (i f Hi Pdyll(G/CH 4/7A GGVPA AN)4 OF AP4AL T f k4Al PflI C'CIC8 4 ANO 4 ( UVon A f 0Pe o r # MWPs Of Apw t w$ F #Ptf4tVt P4f AL HF %Ut f % FQH A t o SGALE S ff LL U/P if A#eVf.P4 f PebHf GM H rF)f) V14 4? ( d. l Mf f t vF fef DIPUHf til HP VOOEL PAE $%84'/tD TO F AiLVHi- (.nVrm A hs ;*# p re Uwh of f /, I wat PeUHf O tH is N #1 pr;4 44 (nJ P43f t f y( P(f O( J T )H f p[pp I Legal leeuentes (OVP4 A fit;P4 7 or Umm Of A, pet t seg NuHt G c7',0 vio M4 Ni>Ct E AH Hf GUI A f OH F CGVV%QN G NUHf Q,f H TOO V14 Ne t K 4 M(l i ,1 N I DiPUHf (( l Hj SUANCIS FOH OCIDO(H (9e4 F49ee t 0% t 4 t$ COV89 A hf)P4 i'9' % C'h Of Womt*r t '8HS J i i I
~~~ __ ____._ ___ _._ _ _ _ _ _ _ _ _ . _~~
Subject Index 173 NUAEG/CA 2000 V04N10 LICE NSE E EYENT AEPORY (LE $ Lost Of Offsete Power CCMPILATION For Month Of Octaer 99 tis NUAEG/CA JMJ COLLFCTiON AND f V AtU AflON OF COUPL E Y f NUREG< CA 2000 V04N11. UCt.N$ti EVENT REPORT REA) ANg p ARil AL L OSSE S OF OF F Sd TE POW E H A f NUCLI A4 POW E H COMPILAh0N For Month Of Noverrter 19A$ PLANTS l NUAEG/CA 3026 FEAS$UTY SillOY ON THE ACQUISTION OF U-CENSEE EbENT DATA Lege.og coosang Acced,ng NUREG/CA 3519 Huu4N ERAOA PAGGA81UfY EStiMAtlON USING , LICENSEE EVENT REPOHi9 NUNI GJCH 3426 V01 THf HU AL ANO # LU'O MC#G IN t> 2 SCALE Tf St F ACluTY Da'a Report l NUREG/CH 390$ V01 At SCOUENCE COOING AND SEAACH SYSTEM FOR LICENSEE E%ENT HEPOnf S Usee e Guwte NUntG CA M3) vat St THAC RD1'uOOt AN ADb ANCiO fit St & S-NVAEG/CH 3J05 V02 S&QUENCE CODNG AND SE AHCH SVSf EV f VATE COMPUTER PHOGnau FOn no: LING W Af tH HE ActOH FOR UCENSEE E\ENT AEPOHTS Co1e Listcqs yngqg,gqy gg,g yg;g NUAEG'CA 3905 V03 SEQt;ENCE CODNG AND SE ARC
~~SYSTEV NtntorCR X44 THAC PF t #NCIPENDf NY ASsf $$ MENT FOR LtCENSEE EVE NT AtPOATS CO1er e Marwal ( qgo,CA 3A$t ASST %%Uf NT OF THE ADEQUACY 05 CHNi IP+~~
NUAEG CH 390$ v04 5EOutNCE CODNG AND SEAACH SYSTEV 5f HLVENT AT80N IN HIFLOOD it $f F ACluf fE 5 FOA UCENSEE EVENT REPOAfS Ca1ers Ma%e' pm,49.CH Und BWA L t As A DoiN
~~A TE H HL ACTOH Wed~~
; NVAFGe CR 4071 E u PLCA A TORY THE NO AND PATTf HN ANALY St$ f t PV ACCIDE NT S.uut AfiON CODF rOn 1941 LICENSFE EVFNT REPOAf DAT A WM CA W HfL M M W NY QUANilf Aff%E sEV PA NUAEG'CA dire CAFT AN EV ALUAtlCN OF SLLf Ctf D LtCENStE n4uf TENS ANO nun f ME sf Afistics ENENT REPOATS PREPARED PURSUANT TO 10 Cf H $0 73 ora't Nunr G,cn 3%e f n AC PO2 iNDE PE NDf NT AW SWt Ni Apport ggngg,CH me Rf L APS ASSE 55ME NY C04uuseONS AND US(R PcVHEG CA 4229 AEUA98tify ANALY S S OF CONT A.NVLNt ISOLA gggg gqg g icN SYSTEV9 NuHf G LH 4127 v0t 8WH Fut t eNffGHAL S Vvt A f pON f f f.f p DST) enOGn4M TRAC 0W A UCC(L Df V[ LOPU(NT Vowe t Nu e%,(si Men c=1s N GC vf MICLE BAAn E AS(VFHAS S ON NATUMAL F E A NUNIO'CH 419e OV(Hvfw OF THAG HDt (Vf HS ON th A%f 5%
DEI VENTSTUDES NLHtGTH 4214 LOCA S MULA f CN IN THf PeAf >ONAL ne st Anr H L4uefecteort M"$^l NE ACDN P"N ' M "0*h" ** **"" I '8 *'* *
~~N *'~~
NvAEG7CA 4410 CUAArNT VE THOOOL OG E 9 FOR AS%E SS NG '#' 0 'Hf a""CR 44X' NTHE I* V8w*
~~I * 'V I M ' 5unt r'a7 4. p THF POTENT! AL FCA f AAf HQUA4*E *NOUCED trQUE FAC f'ON 'N Ad W W LnAN' Cat AND U A Pv tt -~~
~~SOtLS NANCE INDUCI O F Antnr S OF WA<4 nt ACf 0H COOT ANY PUUP SE ALS ON PLANT $AFE f y Lau6d Entre +nenent~~
' NUAEGeCN 4t64 DATA AEPOAf FOR THE ?PFL TE EicniflC Ag LOHOllues fell FLOW DPT A VEN f S N@FG4 A 3005 SuuvAn< D* f Hf Notti AA niGut A70HV (CU LMund Metag UT&ON 3 LOF f PnoGn Au Hf 5t AHO4 F NO!NQ%
i NLH[GiCH 4049 Cf iE HV4N'] CH'f C At F LOW b Ag g( (HAn A NuREG/CA 4P5 f"FOAv DE S GN ANO OMR A TION OF L tGUlO f f HISTCS USNG EITH APtX A f'ON f f(. bNQUt % i UEfAL FAST GHEECt A RE ACTOn$ iNCLUDM+G OPf H Af CNAL NURt'UCH 42t9 THAC PF1 A40010EvitO* VENT A%t %UrNT Low Entsched uten6 san l Ligued Pethways NWG WS Hot ACCE PT A4ct CHl fl Hi A FOR THf LOW (N NVAEGitel E $AP 111 ' t Nvl40NVrNf AL !VPACf1 OF POS f u- ACHI D UHAPe!OU Hf FCHU Auf NDMINf % LATED ACC!DE NTS INv0LV'NQ HEtt A5f1 CF A ADtCAt'ilVt UA NLHL G 1CA9 40t ACCf PT ANCf L84tif 9 A 7 084 f HF L OW IN TEntALS TO GnOUNDWAf t A - mcHE O UR4%M H[ $ OAV AVt NDV( N T& Load Coentunation Low Stratepe Segnetwente NUAtG<CA m t V04 N2 SAFEtv nESE ArecH P40GnAM$ $pON Nonp G tong e40, ArigptANLE Cmftn.4 p On f H( gOW gy ' SGREO BV CF FICE OF NUCLFAA HE GUL A TOHV *KHE D UnAny Hl.tOHu AVtNCut.Nf s AE %E ARCH Quarte $ Prgrete popor9 Ape 4 I kne M 1444 P#UAEG,CH #331 VrA N t SAf E'r Ht5F A8% H $'840GH AUS 9 PUN Lawlevel Wette 50Ht0 av C8 F sCE CF NuCLEAA HIGutAfony RESEAACH QWor'y Pqrett 83eport JervFy i Vettft 31144S NontG<n 1714 Vo2 AL f tDNA f vt uttHOO1 POn (MPOS AL Os- ' LOW L f Vf L HAD50ACilVE W Ag?[$ f eo pA for Wes Heggenes, E0ed Pe<for for8.*wy m fveetD.*pouiOfLrneLev=4Ha& ettwo W nte
*,:,r4p u<tn iris v0)
At fi HP4 Afi%f Ut f HUO15 084 ().SPO4AL OF NVAE G/CA 19?6 PROG AWUtv DA$t 0 t CAD coua NA TON tHefE- LOW L t yl L HAOiOACf rvt W A%f t 1 f ed pn ipnr ad Hwee<ene, RA FOH DESrN OF CONCritiE CO4fA NVE Nt sinucf UHf % e o, Acmepno Ves, o,you, Of t rm i ever He t.o , twe Wu's Leed Motto NUHf gel 84 U/4 v74 At tf its eA flyt VL f p(>O p ost (pssM) SAL OF LOA L g y g g H Agig Ac t r,y A A gf y r ag pc f,, pu g, p,,p,,n,,,,, NUMEG/CA 4t21 EFF(CfS OF SUtFUA CHE V4f AV AND FLOW FiA'E ON F AfiGut CHACM GnOWTH PAf tS iP4 LWA ENv HON pjgg ye g %,,,e WeD elOfM W N h VFNTS %Wm Pdus4f G/CA 4422 A AE V f W OF THf' UOct LS AND ME CHAN-W% NUNUMet 3/14 VOS At fIHNA f f)f Ut f
JOS f OH Dt%#T)%AL C# y FOR ENvpONVf Nf atty Anst!0 08tACs GHOWfH OF Pnt4 tow Lt vp t nAcoActrvt W A 51F f e.a 21 fe< hrwei 84e pernenet Fw thon ve W nte Sunt vissEL ANO Pw.O mets wWn iwnONorNiS NoH,.v ce"n y ,seintti,e m ( v LevetHa y U A nONwO, so,H, StA,n,fy f, ,f , n, c
{ LOtsi enledeCW GMVf NUf D IN THf flH ANf'64 f f f HNC AL PO% f d)$ (pg W A%f t PsVAE G/CH 44 t4 Dint ct conf Ac t U340t N$A f roe 4 Op gf t Au ON f OHV1 999,3 1 n AND 3,,4 CO4f g ygg g4 A fNfggit g, UA gf F H A,t %, p3nggc,g, g,9 7gg, COLD WAfE A IN SfHAf# 4 O COUNf tHCUHA8 NT ILOW lN AND t $ f VA f a(#4 OF 6%O f Ds9C H$ ([ A%( ) HOy f Hf W Agf g sN Leeg-Menge Meteetch Plan f CHf S M THE W 8 F 4i O (OW W L W A W tuum AL I NvHEG tomo V02 LONG H ANGE DE tt A ACH Pt AN 8 Y t+M 8' Y t M to,, 0, ,c ,,we,
~~,hHWiWWDW f a~~
A s ti s Pe,- P..e- A,eo WU mtNLWM NUMEG/CA 4142 UNCF 4f A.P 4 f f AP40 9F PesifiVif f ANA L v5ii OF A N C 'fM N M*AL W 9 W W 1 IN M M ADF A CMINI VOCEL FOH U AfiCOVPGNINf At HOVA CVNAVO M 'Hf L OW (I W L MADOAOH W AW MMAL W Af
~~%HF f F t i O it t 40%~~
~~Lese Of feedwet** N' J"F G t" d~~
A '6%A ' O' ' O'F i t fVf 0 4 AUCN' t' 4 0f 9 'N t OW Nuf4f G fh#) V?4 P,02 At Posif 70 CO*eNHi %9 ON AH'v?fiU AL t f 'e f t

A *4f l 4

H' N N )W' H Hf AC f' >H s OCCUHntNCES Asse4 Ane 1995 4 '"I ' ' 4 " *

8 * >" d hl S#8 O' U" A U A ICN W %"I I4 I H' M -
F fH%f Nf 4441AV8*t f 9 Of PF e ANO PF 14 Legg Of Ween And Ausehery Peedwe'e, Nohl m tn 4tet i s A H Alia lf f of NArnOhiitt:(t g s Hr M P40HfG 9154 LOgg GF MA## AND AuotrAHy # f t OM fin i vt Nf r W 4f wim 4 0 mf f it*W Nunn t o At (M HAf se 4 Af THE DAyt1 Of $5f PLANT ON JONT 9 IW'iS NT ti AH POWI H HF At 70844
174 Subject index NUAf G/CA 4t94 LOW LEVEL NUCLEAR W ASTE SHALLOW LAND WERGE BUAIAL TAf NCH ISOLATION F.nal ReportOctober 1981 Septer4*r NUHEG OW DHFT FC Af AS$f 5SMENT OF THE flJ+NCAL DASIS 19A4 FOR EST,UAflNG SOuHCE TFRUS (Ora tr HeDort For Commeet) NUAEG/CA 4200 9:COfGAAOATcN TESTING OF SOLOrito LOW NUHEG/C84 4112 4 Us(H $ GUOE F OH VEHGE LIbf L W ASTE STDE AVS NUREG/CA 4201 THERVAL STABiLifY TESilNG OF LOW LE%EL ug$oosp NVAE G'CR J-sM vc) (CANO F tt l D t a rt n VE NT 19a t yneurae NUREG A S TECHN CAL FACTOAS AF FECT.NG LOW LEVEL 3 Comrere G frarm hectrabon Pam AN Vt w W ASTE FOAM ACCEPTANrE CAiTE AiA M Carcutshore NUAEG/CH 4268 RATO UtfMOO5 FOR COST tFFtCtrvE FIELD SAVPUNG OF COVVF ACIAL A ADCACTI%E LOW LEVf L W ASff S WNET NVAEG CH 4352 SUGGf 5TED STATE AtOU:HE VENTS AND CHif t. P4URE G/Cn 213 t v04 N2 $Ap[fY A[$( ARCH PHOGR AVS SPOP4 AIA FOA A LOW LE%EL AADICACTivt W ASTE DISPOSAL SifE AEGUL ATCAV PROGRAM SOnto By OF F CE OF NUCitAn HI GULA f 0H f NUAEG/CA 4406 AN ANALv51$ OF LOW LEvtL W ASTES Revee of HE SE AHCH Ouarter+v Pro,pese Repwt Aerd t June 30 ted Hazardous Waste Regdat,one And ktentrfication of AasicactNe Veed NUDE G/CH 2331 VO4 N3 SAF E IV RE M A Ai H P840GH AUS %PCh Waves f eel Report SONED BV 07 F ICE OF NUCLEAR A( GUL Af 0Hf l NUf1EG CA 4435 CHGANC COVPLEX ANT f NHANCEO MC04 if Y OF At SE AHCH Quador'y Prograss Hetat July 1 Scotamtw M tM4 1 TOUC ELEVENTS IN LOW LEVEL W ASTES Amsar Aepar t.Jutv 1M4 NUNIG CH 2311 vc4 N4 SAPEYV HI 54 AHCH PHOOH AVS 580N-June 1965 50HED By OF F Cf OF NUCLfAR HtGULATOHy H( L( ARCH (berterfy Pmgrege f4eget Ot t ote f De( end et 31, Loefemperature Aging ,g4 NURFG CH 4204 LONG-f t AV EVashtTLEVENT C# CA$f OLPtf a NUHra cn 2331 VG$ Nt SA8(fv A(5f ANCH FNOQNAVS SDth STAINLESS Sf EELS iN LW81 Sv$f E US Anrwa Aeoort Cktorer t M3 - SOnto Dr CF F (t OF NuctfAH ntGutAf0Hv September 1984 pg sg Arics on.n.,,, p,(y,,,, g;,pc,, j,,y,,, , y,,t n 3, ,9a g Lumped Parameter Method " NUPEG CA 4182 bE A F CATON Cf $04, STAUCTUAE WitRACTON NUDIG 'CR 4De vn2 Pf 40GHI S9 IN E V ALU ATON OF H A DioNu VETHOOS CLCE GrOCHE M: cat IN8 OHV AiON L:t%f t 08 t 0 Hy OOF HtGH LEvtl NUG E AH W ASf E Nt 80970H y Sif t Pnost C f 5 A.g. ort 9,v MAtROS NUA(G/CA 4342 UNC(HtA:Nfy ANO $tN$fMTV Ans Alv$43 08 A January March t ang UODEL FOA VULitCCVP% TNT At AOSCL CvNAUCS WINf E O M AG.I Spectacte NUHf G CH 1a59 v0) PHOGHf $9 iN (w At UATON Of A A0040-NUAf GrCA 335) THE USE OF WAG t srf Cf ACLIS WifH PO9fivE- Ct Ct Gf OCHF V> CAL Np rHU Af CN Df VI L opt 0 er DOf mGH-ANC N(GAfr%( PAE SSUAf H( $P A A f CA S trgg( a,9ctgAq waggg gay spO9f 084f $if f PHO)f C f1 Heport f os Apr* June '9M4 I #I # N EGeCA V26 VC2 MAhtt NAP *CE Pt R$0NNF L Pt HFOAM ANCE CAL iN7 088VAf ore Of bitOPE D by Det Hanu tivt t NUCLI AH SMULATCN IVAPPS) MOctL Ot %CHiPf C Pd 08 VOOfL I # **' V"' N*# CONTE NT StAUCfUAr AND M NN fMf y ff 5f:NG bl " 2 ' ' ' I V AL U AI'CN OI NAUCNU U A IC P VAPP4 f t U haS W P Ct Cf Gt (A Hf WC AL WrOAV Af o'*4 De vt (07f 0 ev C00 tuGH NUHEG CA-4104 UAhfl NANLE Pt 83SONNI L Ptf4FO6V ANCE 8 M ULATcN (U APPS) uoC(L F i,1 E vabav vatafaboa U dl NUGIAN *AST N " D ' MI N'"JICI4'iIPO"I FOH OCf 0f*t n M Cf Vet 84 iva4 CAACH Nun (G ova D:47 FC At A%FSSUF Nf G8 THE ff CHNEAt B AM S MODI FOA (57 VAf %G SOU84Cf it AVS IDean Decut Fen (nrep NUHt G <rH h02 MtAP% A94 %Vf Nf GUANfif A fi v t at, PA NUntGiCR 3912 U AHLH Ht Cf n ANAL v 44 G8 Si tt t t t u A(~ CI HAvt f F Hi ANO livN f rVF ST A fif C4 CENTS h AN ICf CONC 1NM R CONT AAVf NT NuntuitH4isi fHAC Pti<uoot hotN NUsNF A%f MUf Psf Semes a>e UDO J A $ eeces'e# Lerw Drean 14 M 3) AM MARCH 2 p,,,,, gn, g,,,, g y q g ,,,, NU84G/CA diff A U%( A S GWCE FCA VtnGt gg g g g g, , g g 7pg g,pg, gg gn, gg MAfADOR SIM % NU84E G/CA 4210 U A f ADON A CCMpu tt A CUCf f 04 f Hf ANAty TH W VIASUH89 08 MAP (4U AHn1 HTm i vet ov 4 a f f1*L. WAfr s A r PHA IVMh' P1A V'"* *Ny F or i et*avy H.es $ rape, to Oe %,. NUniG CH 4291 Maf Af084 Vt THCL9 FOH THE ANAlvv4 f # T AAPe%PQ4f ANO LS PO%f CPd GF F4 ADCe#)CLCt 4
%) COCf f f - 7*A"4 V'aoums SCH PfCN AP 40 LSE A 1 UAPc)AL Wagnetie NoHs GTn it r4 va2 r.n n y w Atta niO % At n f Hort g Or WC&A NU81FO/CA 4904 D(vlLOPVFNT OF W 4A ALAAV 4[501ViiGP4 M1%F4 8 IattN90P49 or fMF NI W V A! H8'O F Aut ?
PfVJCf DUAE S AteE Arcuel Hoport F re 1481J MtLCOM Weentenance NyntG/CM 4149 AN A8+00ACH TO f pF A f NO R ACiC PC AUCI PmHf Q LH W6 vp2 UAhflNAPp [ t'( H N 'hM t Pt NH AV Appg Cf CAf pt AtWG # Cn uv iN for urg r4s (Of f gyg f f y Sym4909 ,ygppg, p f, t pg y n.p r og e, pg, t heUHF Q/CA 4t9% AN A%E MUf N f Of Of A VE T H f D A f A # CH A4 WPd f NT S f hw f op$ AND M PW f M t y M S t to ) CCf NT AL M AD'ONVGL OT At Lf AM 4 8 84GV NurLI AM Hi AG Nupt on n wie UA NTr NAPp s Pt H9 WNr L Pt Hf OHV APM t 5 V NU8 1/CA 4 t99 A Of VOPe%f A A f O'd fY N"" ' " ' H " " =~"'H" ^N"">""""'*A*" A,m , $1, UsNG I, c M AtIH A,o V'#'I
-G84f m.AhiT / M,NN w TM,,a P6APatf PHMf fMOf 4 'Pd NULL f AH POWI H Pt ANI% Profoeme Arut nor nmmerwla%s NU" M " " I' V'll I "L I M O' i Al ' A 'l l '"'
M(OROO M N I l I '# 4a%" OI A'*1 P" 8"t9'** I'd H # a" I"'* I'l e t**'*V P4Un(0/CA 4140 V0t Al ACf GH S Ai( f y Pf M ANCH ',( y APsNUAL DiPr)f4f Jeruary . Jve ip$ bW G IN D"I A'8AI f M I # J Af AM M U % N'" II A" UWH Pt ANf M AL(fl P( Atet i WtLMPtWOO3 4
84 4 H d OPf N A f h4 i

N H 8 N8 F AND AGWOMAVC NONIG 4FIlan? fM MGM L WO QF lihH Ihf4 UI L 708 /WN Af Ll AMF UNI Pa f h8 Ill CIHC M( J T f 09 C(Nf4 F OH 2 PptXA f ON IN TP4 Vil f 98'l MOO 2 COV8N fl H NIi HI41 H 4191 Afe f 489H4 At APsAt y ta% Of y (( C fl p Ni N [( A4 CQQ(. WIWl H PL ANI VA!Pif 6 4 ANI I F Alf D843 APeh PL A8 4 f $4f t t f
l Subject index 175 l l Weintenance Pmonnes Weefde.n NUREG/CA 4104 MAWfENANCE PERSONNEL PEAFOAMANCE SiM NUHf 3/CH 3rW THf' UCJE LWG OF BWH COAC MELTDOWN ACCI ULAf CN (VAPPS) MODEL Fiesd Evapuaten/Vandato't C4 N f S FOH APPUCAtlON IN fHE Ut L apt MOOJ COMPUTER CODE Marfunetton Nus4t o,CH 4 t 72 A U9E R S GU'Of FOn ut RGE NUAEG ttS4 LOSS OF M A N AND AJUUAAv FEEDW ATER EVENT P4Unf G/CH 41A4 AEHO%OL lif HAVIGH UOOf LiedG (f ASet 3) . SUP. At fP*E CAVIS BESSE PLANT ON JUNE 91985 POHf MI AVICE S FOH Rf SF AHCH AND EV ALUATON OF StVIDE l l ACCIDt P4 f PHE NOME NA ANO u.f 4 A f ev L Mandrel Leading Test NUMEG/CR 3440 VC2 L6GMt.W AfER RE ACTOR S AFEh FLEL $4 5 Wetasturg. cal emnd f eet TEMS RESEAACH PAOGAAMS Ovadedy Progress Report.Asms Jurv Nur4EG CR 4039 GAVVA A Af CHAHAGff Hf/AflON OF THE TWO. 19f44 V[ AR IHR AQf Af ops f BPEH VENT rt HFOHVE D Af fHE POOtLD( NUHEG/CR 3960 V0) LIGHT WAff A-AE ACTOR SAFEtv FUEL SvS. F ACrLif y fEMS RE SE AACH PROGR AU$ Ouarter+y Proryogo Hepoti.Jul'y S + tomt>er 1984 Wetecto6cgy ManufadWe NUHEG/CA 3484 V03 ICAHO f if L D E Wl R!M[ NT t$41 Volume 3 Compar<non Of tra, < soe e corwenwahon reneerse An.s ut sOOrF NUAEG 1127 AADtATCP4 PnOTECTON TR A4NG At UAANiUM HE R. MWief Calcudahone AFLVOACE ANO FUEL FABAiCAfCN PLANTS Wethodotogy For Estimating Mieh impacts Wartne Environment NuntG/LHdn1 UfASOHf9 08 5 A8- t GUAHOS H:5M L UPLOYtNG NUREG/CA 4t90 CAuf 0ANA OFFSHOnE SundY OF LKENSEES pnAivogregp34 w easnsyy For t enmat.ng oea p Impe, te C# Sais. i USAG RADC ACfort MATEHtAL guer:te y,seutee 1 Watenal Control And Accoununt Weveen incident NunEG/CA 4108 CEvf LCPVENT OF VCAA ALM 4U AE SOtufCN PROCLOVAES NuntG.t tol CONF AV NAf tD Vt oCAme 6 fit t leveeton Of Stees into f be Unted Sia'es that Ha f Meen una.fvedee4try Ceterrwnated W th Cobe-t Ad As A Hetuft Of 'xf appsag Of A f eietherapy Ufwt NUAEG/CA 3 elf PAEUMAAAV SCAttN NG OF #Uf L CYCLE AND y ,ggg, eV-PACOUCT MATER!AL UCENSE S FOR E Vf 8%E NCY PLANNM) %W G 414 De M[4Mwvu Watertal Transport COMPuANet MfH Of COUV-SMO4NG CL f ANUP CH4f t ntA Af NUAIlG/CA 4?M TOnAC USER S U ANUAL A Comg% fee Cose For Ana UN A%V N M AY SiIl S rg f orWWLced Flew And V4'eetal t'ensport #n N#iese Fem gg,,, ,,, g i NU54f G CA 447 CW'H f PC A 2O FiNiftiLtutNT FnACTUHF , l Waternate Contred Unst ANAL 4 S'S PMMH AW FOH A MOiOLOMPUtt R NUHE W CA 4t07 SE Out NiiAL f E S T PnocF DU AE S 9 0n CE T E C T. ""'""*'" ING P40f AAcitD WAf t H'ALS LCS$r1
; hunt G'fH 4124 ND6 OF 37 AM t 45 5fl( L AP40 (W UNt (t Ay, Waterials troperty feet %Deo f 08444 OP LWHS Annu al Netv1 CAh4mr Wii . September PtVREG/C'4 4314 VQt A[ ACIC414 SAFFfy b( %( A A(H IM4 i PROGA AV4 Quadeey Aes 99 Ja% vy Ma%h t esl$
Wigration r Wetettale Test NUHt'G/CR 4W A AD'OMJCi rCT 4044 flON IN G8900NO
~~! NUAtG/CR 4218 LOCA SMui A70N eN f"f NAf 0NAL Rt $f AACH~~
A f184 #8 car neport) i LNr,t ns AL pat ACTCf4 PAv 4AV Poetwees+cn Esa*wahon Ae. NUHfG'CA41'4 VAlf PK F f f f f Cf1 ON fHf %Q8tPfod (# Pv) outtt For ihe f Puri V4'e*'ais Iest (Vf-h . tet.orwt i Campagn CUDt S ON stOr a it ANo u Ng n AL 5 It Wathematical Modet W*e's l
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176 Subject Inden Mitegahort Tochtwque NUHf G/CH M4i THE DW4 Pt AN ANAt Yli H NVHEGeCH 4251 V01 MifiGA tlNE f t CHNIOUE S FOH GHOUND' NUHf G CH 4100 ( V ALUA TON Of tNSTRUVf NT AL M[ f HOOi F OH WATER CONTAM NAflON ASSOCIAfED WifH SENEHE NUCLE AR THE VE ASUNIVE N f C+ V[t LOWCAaf (%%SOfe4 ACCOf NTS Volume 1 Ana'ys's Of Geanne S.te Corwe NUHE G CH 4118 Ui N f 4 )H AG VffHODS FOH Dt f f HVAA7CN NUNEG CR 425t V02 Mi flGA flVE fECHNCUtS 7 084 GHOUNO- COMPL r AWf v4 ' + i( OVU%ON NG CLi ANUP CHe f f RA AY WAff A CONTA%NATON ASSOCIATE D W'fH SE%tHE NUCLE AR UH AN4 v ' i m "Y Nff S ACCCCNTS Volume 2 Case Study Anawee Of Hydrooec Character- NU54 0 4 ' a i AG?eG AND SE HveCi WIAH OF fit Cf HC eAhort And U.tyetme Seemere 90f04 C ' .1 V AL bl 5 USE D iPd E Pe<ihf f.HE D SAF t f y F t A. fLHL 5 e , % Lf NtCt[ AH POWt 4 Pt ANTS 1 MotHiltY NuHL G/(H 4r a f a f HE Mif y UOP41f CH 60 Consata= af.ons F or NL'H(G 1164 IPeFOHMATON SN THE CONONIUENT CAPA4UIY OF Use Dr+rw F% emont AM L ,etuaton THE F ACILifV OtSPOSAL Ant' A Af Wf r % ALL EY Nf W YOHK NpHtg<cn 4 m yot Acet;gic gy woqrLA4 nt t AfoNsn;p NUHEGrCH 4435 OHO ANiG COMPLE R AN f LNHANCE D MOH>lifY OF FOH W M HVG W TOHM OF NM AH PHI W Hf TOluC ELEMENTS IN LOW LLbEL W ASTLS Arvwm Repoet Jur y ten 4 g gg g g p,,, ng my ug ,, a June f MS Peuf tt G,LH 4 M/ %UGG8 Sfl D % f A f f DI OUthf Ut N T S AND CHif f - R6A FOH A LOWlfbfL H ADiOACfist W AS TE DiVO%AL S'f f 1 COMPARISON OF OVNA%C CHARACf;R+Sf CS M OM A N" ' ""*"" U AEG NLRI G Ln 4 4 A 1f UDV O $I 4 C1ty A D tt CtOhK'S M Nt A OF FUaVSHMA NUCLI AFI POWER PLANY CONYAAV(NT flV!LD I*A## ING DE?! AV NEO FROM f t ST$ AND E AMTHOUA nES l Modet Devesogwnent M8"O*ae'9*he Il*4 tron Sourcee NUNI O'CH 42*4 St AND AHO Dt i A pan f r i t ANO w 3esef Na Hc,a . NUHk G/CA 4121 V01 BWA FULL INTI GA At AVUL A fiON fist IC llI Cf HON SM f 4 FUH f Hf C AL tpH4 f ON Op hp r A n Agt ,! (Fisf) PHOGAAM TRAC 0W84 UOCf L ptg((Cry (Ny ycue g Ny metcel Methorts A f( N PHOif C f C*e 6Ns f Huvi N f A f(4 Modes uncertainty Monte Certo Samus4 tion NUHE G /CR etsf A SCf NT FIC CHtt Ott C* Av44LAftt t MOO (LS NUHI G c't M/e vn2 U Ahtf PdA*er( Pt HNONNf L Pf HF OfsW AN([ FOR DE AbTME S VULA'ONS OF 045Pt ASON S Mut 4f oN quAw%t WCf L Cf 'K H Pf CN OF Mnet L CChtt NT Sf Hittt;v4I AND 5t Nsif tsif v f f if %G Mode #ng NVREG CA 2447 vec nE v'E W OF Ort WAstt par m AGE Moodue PROGRAW S4tatii 19 $44ferial Weste Parmajo 8%ogram Or. tater NU68[ G.( H 41$4 A SfUOV Of M AVOf y ANO fICf r4( s N Nf W t441 M4r* 19A4 ( ngl. ANO Fa.44 Heps1 i NORE G'CH MI6 V02 V4bNf t NANCf S'f HSONNF L Pf RFOHMAPeff S MULAfON (V APPS) MOOfL Cf SCH PTON OF WX 4 L Multiphase Flow SyCom CONTINT STAUCti;nF AND M Nmf rytty it SthG NLHf G 'CH 1)4 4 f Mi AND VUt LVt Avi H AGE D WNsf Hv A f ?ON NUA E Gt CR46 }4 U A A f r NAN (( Pf RgGNNE ( P(HP O*4M A*sCE 4V g gUA f ten.g p on vUt f rema g s t oW r.,1 One g y ,t,, 4,%ot in j UL A f ON (U AP95) MOOf L U%F F4 % U ANU AL pe, w gwg NVAE G 'L4- F4 9 V02 f. v &L L A f k'W OF POWIR A(AG'DH FUTL AOO ANALYSIS CAPA0 uf tS P54to 2 Tomad Heport voeume yon g p,, p y 2Cortefusbahoa Nung G,( n yn g 909 g gpy g i:Hr4g e,f IN%Pf C f(N f rH StlAU
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.. n.uS s.e,.~, H,. 4 t 9 Pr4 P4HL L Umf C f N ADI A f e C4 NM i i VI Pdf DA f A NUHf,sn(A1f H, yfr ,.,,H.,,-<>,- n . ... ofs W;tefr#nsq N1 JHF (A (Al f S Paf)i NHG f tO O f 4 L f H A(N A f s04 Wi telf OH agj N M HyWen bef NT fWO88W Pem ett itact,rt JaN4e Westa 17 9 E89"' #8 M A' N UA I A N8A A "8 9 ' U" hI W AU A II % I % 4 (NI'H NUHEG 1144 t#At(AH PtAP4f A#de' A Hf M AHLH (NFAH) PHO I Hf V 80 U IiMMUSI/h II SI')
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NUHF OiLH WS ACOuif C $ U WO4 Nf %Un f 4 OH f A AI D f 4M f Hl (Wf Aff Mf Afblif f APsAt V91 Of fust Wd e A A!l 4 Nt)( ( f AH f($ fit #4 fM& /0 t INflHVf DIATL SGAL( Pf 48 %UHf VI%f L HI AG N)f ti i r
Subject inden 177 Naturel Carcutetion Nonnuclear instrumentehon NuntG'CH 4ter FLE CHT St ASEi PROGHAM Finet Avrort NHCit pne Westanghouse Heport Numt>er 16 NUHI GrCH M t 7 AeLuHE MODE 1 AND F F FiCf S ANALYSIS (P Us A) OF Tbt IC5 NNi iLiCfH6C IUWL H DI$fn UUf CN CIHCUsiHr Af Naturae Convection Core CooNng f Hf OLONE E t NtEt E AH PL AN f l NUAEG/CH 3804 V03 PHYSICS OF HEACTCH SAF E TY Quader'y Notch 0vetehty HeretJuv Septemt3er 1984 NUHf G/CH 4M COHHE tAfON OF CV AND mIC/80C f AANSif CN NVAEG/CH 3804 Vo4 PHY$iCS OF AEACTOR SAF t f v 0ue"e'N f t Mot HATU'it INCHE Ast in Dut TO iHH AD'ATON AeportOcteer Decen'twr 19A4 NuHa usH 4417 t uptosi4f 0Hy Sf tOit S OF tiivt NT INfl HAC NUHf G/CH-4240 VOL PHYSaCS OF HEACfOH SAf f f Y Quaderly iIONS AND CCMf051 ION DE Pt NDE PpJt $ IN HAD! A f OPd M NSA ReptetJarwary March 198% f rvlTY OfitLOPVt.NT leeutrensation Nuc6eet Ptent Ag6pg NUHtG/CH 42W TAILINGS NIVIRAlllATON ANO OTHER Atit R NAfivtS FOA IUUCB'UltNG TOtic UAff RIALS IN f AIUNG% 7est NUNI G'CH 2J1t V04 N1 SAftf? H( St AH1'H PHOGHAMS %8Pft- i gggig g gy py y ,cg, pp gg;t t e n geg ggg A f ggy ' j Heport et $t AncH Oue,teriy Prryese Hegnt Juy i screemt.,30 tan 4 I Neutron Ooesmetry y4, ,, NU53r0 0975 V03 COMPetAfCN Or CONT 4ACT F4 St AMCH FrH THE V4f t AIAL 5 ENCoNEEN M) 844NCH DeVi$OPg OF (NGiNf (f4 MA4gg G. LH Wg t VM PtN ,9
$5 iP4 h At UM ON OF H ADOW N 51 S 0 Rt 50 4 A Pt AN nee, tcw vnws W L M AH WASR Hf W ifOfH *i4 NWf S Hgort NUnf G/CR 3609 EVAtUA fON OF Ni U f $40N 005 Vi f n y f(CH- E 0' #FdJW* I'"4 NOUtS FOn WtLL LOGGrNG CrtnAficNg kww Carhon Neutron Embettflement NUlt(O CH Ma0 CHAHCCAL rt Hf Od4MW t UNDt H ARtD( N f NUnfG<CA 4417 titPLOnAf 0Af StuOrt s OF ELiut Nt iNit nac COWees pe UC.Hf W Af f H HE ACros6 TONS ANO COV509af ON Of et NCE NCit S IN H AC*ATON $4 NV OCA P f tVffY DtvtLOrvtNT NUntMNm Pry w wnN m em te m m Neutton irrediation f Ht CA Mi H T 014 F S t##f 1 MJCL T Art POWI et Pt ANf ;
NORE G/CH 40e4 f tN$ tit PnOPERf ttS OF IHM AD4Af tQ NUCL E AH g,9,gM 9 9 ! C A t %%nt visstL WILOS 700 THE TH Ho esst mn A N bH m tm Hw m m u n m W m m m fega%s NT IOPf AAN) fitf 5 t t AND t 2 IN ?>4 POWE H Husest l
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P4UntGiCn 4011 V02 NEUF nON SriCf AAL CHAH ACfl Hi/ A f ON I "" " FOO THE FIF f H HE AVf SI Cf CN Sfit L f t CNNOL OQ f lH5%fi m i A AD ATON H A:r g ' Neutroew s Cau vetwee " Nutit GTH M4a f n8't H Vf Nf At Ht EULis OF THE OPf H A roNat NUntG/tH 40)t vo) Nf'UfAON 65't C T H AL CH AAACit Rf / A fiCN f HAN4 f Nf (OPf n AN) It %f S 14 ANO t 2 IN THF NJWL H DuHgf FOH THE Fif fH HEAVY 56 CTON SiEi L f t CHNOLOGf (H%5f t m P Aciuf v RAOtATON St R tS Neutron t aposure Peronisters " pg i i PteutronMe Calcul446en NdiO'CN 1**% V04 HOH II VI't H AIDHI GA% M)OlI O HI ACf OH j NUDIG/CR 401t V02 NE Uf RON M'I CM it CH A A AC f t Hi/ A TON 8Affff 9 7 Uf *t S FOH THE D'VISON Of AM IDe rt f ;. 70H TP't F FTH Hi Avv SECf 04 Sf t EL f /LHNOLOGY Pi%5f)'N 1V4tu4f0NQuenwy Prayne H pint or h+.e t O , eme., , I MAOIADON St AitS "Neutrorse Cascuidhoes " 3 U 984
) Nevede Nuclear Weele Storege investigation Omiot N 1 OsuntG CH 4De V01 PHOGrit S% iN E V AL VA f 04 C# A AOtOPeu. M#4f G Na.4 DH8't FC Hf A%8 Mut NT OF f He f t.HavCAL HAst 4 i CUCE GtOCHE UtcAL #NFOHMAf CN ttvI LOs=to ef twy H.nn 8 OH IifiVA f W4 h00H( F f f HMS Dre*t H pet P rw (nmnet)
Ltvt L NUCI T A4 W AS f t At POSIfORY Saf f PHOat Cf M H( #084 f , FOM OCfCBE R Ottt MPtR 1944 OnYin? PC 1
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New heedr6d ANat rv$ ettor;H AW f 04 A Mg Hor 1 Assyf tri j NUHLG/CH 4311 STt GtNtriVt F AULf /ONE VVOJtu ANO st. ' phoeg OccupenenetDeee Pr#is o4n 4 m t m &ful OPfioNt scos Hp(n 0 No O(cura i
!' Noepng YK'#4AL ()U'.E P ftOM f Hi fMi# BA%IMtPdf Pc;pitr)iCn 4044 fnACPFt iOCA CALCULAf ONS (J5#3 fINt-PeODE AND COAH5E PeODE iPePUf UOOttg Oecupenenes Medtemm 1 NLHf 4 0PI) VM UrLUPA f 0N AL I4 ACn A f oP4 I urynuter At COAs peondeettuctive Evefustson ut prJ At Pee W L i An sCW F H Hi AC f osig - t en i AP#v J At fit er o f I NUAEG 9t$% V94 N( $t AACH PHOGHAM PL AP4 PMA Det%A teve is Peuht O'LH 4 HlO HHlf 08M Al $UVVAHF Cf (al OPA f M>s AL F4 ads-emneten AfloN t PrMUtti ( wt n l Nr t $4 y g Lt y yttu,iAt Nutgg Art L NUHt G'CH WM V0J LIGHT W A f t A HE ACf 0H 4Af t f y W A f t tt1AL S 80Wt H Pt ANf 9 I t NGINE t to PF) MtSt A54 H PrV)r,H AM4 Quadw*y Pg ett j i A prvt Apr4 Jurg f 944 Oecurrence 1 NuhtG CH 4124 PeOt OF Sf AvM r $$ 4fi t L AP40 C+# L WT lt As P40pt u orni V0 f Peo4 HipOHf f 0 (OPv 6Hi M 04 AMP #>Hual UQfeifQtW) OF LWrit Aarwei Heprirt OttrAer tem) Sootemener (x tuhHf *e I 4 (A k4.r Oe-en time t was ,
1944 i NUnt G/CH 41st DC %iGN ANO (?t VTt OPuf Pdf Or A %Pt tiAt Pun O't Normed tvent POV MAF f Sv%f f M FGH NGNff4 StHUCf rvt t vat UA Tg/4 OF Pej- NUsig G,Us 4184 4f# A ( Ow'U f f H TOM P 080 i % f M4 f.P,0 CL E AM HT ACf 0H vtut LS AND PIPtNo Cr wthe NTS ( CHf
A fl H Hi M f OH P ut t **Jo F A:n ont 5 Nondestruettve I semenehon Othc648 Mecord NUnt G 09 f t V01 COMPft Af 04 OF CONF MACT Nf %I ANO4 f 081 Nooiq wto Hot %I P44G U At0Ht HE P#*,1vt tes LOHug Disemi
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178 Subject index Oftehore Med6eactive Watertal Organw Compsesant NUHtG/CH-41 AJ CAurOHNi4 CFFSHOHE Sun %LY OF LICE NST t$ NunEG'CH 4435 OHGANic COMPLt e ANT I NHANCE D MOH:lif Y Of USNG RADsOACTIvt M4iEHiAL TOmic i1E Mt NT% W LOW LibtL W ASilS Arinual Repythy 1M4 June 1965 l' NUREGcCH 3145 V03 GEOPHY$iCAL INvistIGAf0NS OF THE Organ 4satsen WE SitRN OHIO-INDtANA 9tGiON ANNUAL Ht POH f (Octotwr NUHtuiCHet16 V0t GUOf lwr m AND WOfteilOOn FOH A%f SS 1987 September 1983, Volume 3) MF NT OF 08*0ANI/AfiON ANO APM4%f HATM)N OF U fit ifis's SI i n&& Ort F4 A f WG L R I N5f fOH A NLCLEAH POWIH Operating Iaperience PLANT vcWine t Gu=1.wwe F or Ut*ty Orpnisation Ar=1 A.imnwstration NUREG'CH J430 v02 NUCLE AH POWtH PLANf OrtnAf WG t ipr- Pian HitNCE 1M1 Antbal Hepyt NuHF w CH 4'29 902 GUOf L At1 AND WOHmHO(4 f OH A%i 54 NUHtG'CH Jeal ANALYS4S Or JAPANE5E U S NUCLE AH POWEH VENT Of OHGAPc/Af 0N ANO Af M 411H Af t'W OF Ufitif ff S PLANf MAtNff NANCE Min #3 opt H Af 60 L K I NSI 80H A NUCLtAH POWIH NUF4EG/CH MOS V01 Hf Stout NCE COO 4) AND M A8404 Pt ANT vo+ume J Warstwo F or Assewnent Of Orgervseton Aruf Ms+ i SvSit V Fat LectNSE E EVINf HE POHf S user e Gwe opment NUHEG,CH 3 m VC2 stout NCE COOdeG AND $t ARCH 8Y$7EM FOn LICEN%f f (VE Nf HiN'Nf S Cate Lastme Orgenisat+on Chart feUHaG/CH 3w6 v&3 StOvtNCE COOING AND St AHCH St1ff M Pc#i1G uili H0t U S NUCLt AH Hf Gut Af 084Y COUVtSSION f UNC FOR LICE N%f F E% TNT Rt POHf S Cater e Me<wel f10N AL OHG App / Af ON ( HAH f $ COUnf G/CF4 390% v04 StoutNCE COOING ANU $t AHCH Sv$f t M FOH LICEN$t t t% f N T HI POsif S cator e Marmai Overtoo8+ng NUREG/CH 41Se OPE H A f WO f es't H t NCE ANG AGWG MISMC Nt et G 4 A m% fHf HVAL Hv0HAUt M: ANAL Tst 5 (N 0%f HmOL A%I $5MFNT Or (Lt Cinc MOfoH9 WO u GUI Pe i s t on 79f' H H HOH NsOP4 UNt f a PHI % uni /t D NUHE G CH 4134 V0t. AG.NG AND St HviCf WL AR OF f LICf hic fHe ngAt stey n stoo< MOTOH opt H Af t0 V ALvF S Ust 0 iN t NGWt L F*tD SAFt.fi f f A NuHs G1 H 4a>2 PHf %un /I D THf HV AL 58*
u f v Ai UAf m0N OF TUHE SvSf tMS OF NUCit AH POWin PLANTS THE C ALvt H f CL

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%UMMARY Date As Of )v,1919A$ (Orery Hme) UGOs t PHf $%Unt/f D fO P A.LUfat NUHf G0744 v0% Pm41 Ot'lHAf WG Hf ACTOS 4% Ler.t N%Aq ACfiONS SUUV AHf De's At Of Aoquet It tMi (OreNo Ibwei Oveetene j
NUf 46 0 0144 voi Nto 001HAth) Hi ACf 0H3 (M t N%NG ACflONS POJ8'l ad fl 4244 Hn COUV6 Mi Af'rW9 7(W4 NHC POL K:V O*e % Hit' f SUMMAf4f Dein As Of Ottor.e 31.1945 (Orange isoral SCH4 DUt 4) ANO 0)B Hfiut At Pc A;t t 484 mWe n Pt ANf 3 Opetetsenal Proteuuree On ygen DHuesen mVHEG OM 9 Haf NHC/F I u A opt H A f CNAL H[ %PONV PH(N O PeoHf G T H 4 M4 APPt If. A f(W4 U (4 %f Y PtV 4 it IPF 4 f O I OVP j DUHf 9 F(#1 Rt $PON$t iO A COUVE HQAL PcJCL t An Hf AcfOH ur Nf OUAus iCA f 4W % f 9 AGCIDE N f P-we,e Meowfuese Operatione personnes P4UHr G 4 H itas voi Ot t )fH v %g A t INyt %flGAf K W4 (V fH( PdOHf G/CH 4 t 31 THf M Aft f D SOHviY A f tf DNtQi,1 PDH Orf f AW WF %f f 8*N OH4 deDiANA Nf4N)4 ANNUAL HF ty dt f f(). bte, WG F tiOfl ACp FRGM OPtHA fiONg P( H3(ANg L tMJ , %erforater i M1 V@en M 5 Opeestor P' M M;nt GtCH 4706 A %f L tCf HFV E W C# f HE fit 08 NT 091$ 1Mb Pei nns GeCH 4 t rg f H$ nH e (A v.4 AM) Os f HAI r W f* tuyno HT H Av10f1 AL HEMANH Lif 8 HA TUHE ON 188 A N P40 S.MVt A MI I A' IA%I I'"I I UI 4 DI A' Il#'% 4 (UI"Y3 UCI H 4 I 'h AL TOH1 "I Al IH IHYb"A j Operator Error PHHftOt i NuntG t tle 1094 Of MAW AND AusruAHy (((pW Af g y g;ytNf POHIG4 d 4 )m%9 40) PHfhnfT9 W (yALUAfgW (y H 4[w pp)

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ABUSTIC H4m A%5f 55Ut NT HISULTS TO H* 4t hf Pif4%ON Nunf G'CH 4 344 INSTHUCT@AL %ai([S f y ALUAf TON W Pofit .

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OF WESTINGHOUSE PW4 PLANf 5 Volurae CCUMITTEE ON AEACTOA SAFEGUAACS 19571964 Votume tyart 1 ACAS Aeports On Pro **ct Aev e*e ( AJ) 3 Gstotme Break trotectry rouced p By Eartewanes NUAEG-it25 4C2 A COMRLATON CF REPORTS OF THE A0vtSOAY RELAPS CCvutTTEE CN RE ACTOR SAFEGUARCS IM71984 Volume 2 Part NUREG'CH 3703 ASSESSUENT OF SELECTED TAAC AND RELAPS 1 ACAS Reporte On Proiect Revwee (G Pl CALCULATONS FOA OCONFE 1 PHESSUHJED THEHMAi SHOCm STUDY Protective Cover NUnEGICH 3772 RELAPS ASSESSVENT SE M SCALE SMALL OHE AR NUAEG<CA 4075 CE$iGNiNG PROTECirvE CObE AS FCA UA AN4LM TE STS S UT 1 S UT 2. S UT 6 s UT 7 ANO s ut a V LL TArUNGS PILES A A*we* NunE G<cH 3802 AtLAPS ASSESSVE NT OUANfit4fiVE *Ev PA Prototype Feet Reettof R AVE f f RS AND AUN T'VF ST Af tsflCS NU AE G/C443 ?S THEO4v CES;GN AND OP(AATCN OF L OUID NUHE G CH 3916 PELAPS A%ESSMENT CONCLU$CNS AND USER GLTDf DNES j VETAL FAST BAEEOEA AE ACf0AS :NCLLOWG CPE AAT CNAL NUAE G CH D77 RFtAPS THIHUAL Hvcn A.JUC AP4 ALy$ES OF HE Alf'H PHYSICS

^ PR(MUnif[Q THE RMAL SHOCS STOUE NCE S FOH H B AOUW Protracted Matenete Loseet SON UNif 2 PAE%5UA'lE D W4 f f A HF ACf 0H NUAEG/CA 4048 CL YE F1u tpPa} CHifiCAL F L OW V ALVE CHAH AC-NU A E G' CA-4 t 0 7 SEQUENftAL TEST PAOCEDUAE S FOA CE fECf-ING PROYAACTED uATEniALS LOSSES TE R'$fiCS U%NG E NTRAPOLAf CN TECHVOUTS Psychotegical Strsee NUAEGeCA 4340 OPE A ATCNAL DE CIS ONV Am$eG AND ACilON SF AEG'CA MM AELAPS A%ES$Vf NT CONCLULONS AND USER G LECTCN UNCER PSYCHOLOGICAL STAESS IN NUCLEAR POWE A PLAN f S g[E 4 2 A Co gycy 3 ynAc p,uPAHAfivE g ma ngAN ALY%S gpg ygg, OF CON 3fifufrvE nr. Public Safety NUAEG'CR 4tS1 APPLIC ATONS OF POAE GN l'ACB A B'USfic U" SA8ETY ASSE% VENT ESPE A ENCE 70 THE U S NUCL( AA AEG N CH 22 HE MUAW D Ht AV L SHOCm I v A C4 OF UuTCAf P40 CESS THF CAL'vE AT CUFFS UNif 1 NVCLEAH POWf 4 Pt AP47 Pwff Model APA NUAEG/CA 419 COMPAAiSCN OF THE tMt INEL DiSFCA&ON NunEG 112% VCe A CCUs"L AfCN OF Hf ecnf S cs THE AC/ViSOHV DATA WifH AESafS F ACV A NOV0f R OF 04st nEP47 voOELg COVuif FEE ON for ACTCH SAFEGuAHOS19571944 voiume e Pa t 2 ACRS Aeoorts On Gerwe< %twe<.te (HP A Ac5 ros C) Pump NUnFG <CA Mid PUUD ANO V ALVE Ch AltflCAfCN pf y-[W G(rC( II8d84ho9 PsVAEG'CR 421e AEv f f OF THE v04fLE UN;f s t ANO 2 Aus:UA. PF.4EG 0417 VOS Not N8C TLD O'HtCf nAOrA f CN uoNifCH'NG Rf Ff EDW4f ER Sv5fEW AEUAO'Ufy AN ALvSis ' i f *0A* Pmem A* Port Jasa4 Mann i mS NVHf G/LH T484 (4CtOWLAL CH AH AU f t~ HQ A f C4 CF H ads A f TON Purg$ [ EPO$U$3E AND (f)$( ( $f M A f[ $ f Q$4 INHAi(Q UH APfuy y ((, ! PdLAf GdCA diet CONT AAVEN' PUAGE ANO VE NT V ALVE TE ST WG ( FF L UF N f S Am af 5%ym He PHOG84 AV 7 *a AL HE POH f NUHf d'C84 *M 3 THF HOLI O' 8'l Hyet Apr41 'UN AL A+88M1 Vevh 1444 SAMHUNil W H AD'- AflCN RAFI TV PHOGH AM% ANO Hi %UL f S Of: A L AF0H Af DHv OUIC st M Ev At U A TOP 4 OF Pt H%CNAL A>H % AVPt %Q ( Q nry( NT NURF G1Cf4 dies AI Rn%O( DEHACCH UCC[UNG (f A9g h . SUP. NUHy @( A 4CM GAVyA H A r (HAH ACf( HgAf tGN OF THg TWO POHf Sf 4WCES 804 AE %E AACH AND t v4(V Af'ON OF SE vEHf vi AR spH AD'AfCN EsPt n Uf P4f Pt HFt'nVI O At f HE FvjoLSpr ACOC(NT PHENOVf NA A PdQ UtTGA704 F AOUTy PeUHI G ift 4147 f ME I f r E C f OF E NV fK+ Vf NT AL Sf HE % ON Owaah anone Sett,Ano 70 Slant t LA$fcut n tvunt GsLA 405 t A%f % VIN' CF sOH HE L A f r0 t our A firN AL QUAUNCAflCNS FOH P4UCLE AH POWt A PLANT 08'E nA f 0H% Meestion Effect Nt;fitG'l'H 35t4 VU2 iHi CHF V C AL H(HAyCH OF 10(~t Np IN Owe #ify Aeseronce AQUI OUS SQL Uf CNS UP fO t 50 C ree n eawm H.+. coroe ,,,, NUnf G 0940 v04 902 ENFOHCEVENT ACTONS SGN F' CANT AC f 0N% Af SolvtD readeev Pr+rees A mrt Ac<4 hr t 4a5 Medastiea Esposure NU84Eq/CR 4271 NE UGVME NLl O SAf t f r Af uAtouf v 0VAlif y NUHt oicH 40w t vAUjA TOP 4 OF NULLT AM ( AQUf y D(rchy.s A%$UA ANCE Af40 VANAGF Vf_NT A(HOSP Af f f( f,HPesQUt 5 Wif H NOP4 PF) PhOjt L t 5 Am,as e,urwary Hepwt . P k et fear i94 4 POS's GLf APPUC A 704 0f THF DOF f O THf H6H Lt viL MADiG-AOflVE W A%fE HEPO%fOHy PHoonav Meenhen Monitoneg NUF4F 4 04'67 VM NO2 P4HG f t D DiHi C f D ADiA fiON MUPp f OH Pp) Owenth shon Nr f WOHu Pmpm H. pat At <4 > p. t m NUHtG/CR 4M VM PAGOA&USfiC H %# A%F MVf 4f COU$4%E NUPf 4/t H 4 9eo H4f 0HCAL %uYV AH f 04 (w ruP A f CN AL H ADI DCX.UVt N f A f CPd volurce 9 Sveterne HenatW, Arvi Aea ye s A t'04 i aPO%UHf f art H t Pe.I (Pi V 4 LOMUt HC1AL NU( t EMt fe W m et Sete.r>n 0 - Quar + A 4 hon POWS n Pt ANf S CA0fwo Madeshon Protection P4vnt 'l/CA 1442 AADf A0 A CCUPufl 4 CM C FGH % QUL Af WG NUHI G'CP t>rA9 PH'Xt f DAM OF ape W f f HN A f fnN AL WOHm l F A1f f PAP 4%tE P,7 TWO D ut NtGPeAL.f WO L AvF H H A D50Peil % HOP DN HR f 084>G tlO5F 5 8P1 H f tv7 ANO LOM Hf ()ut f 0P4 ! CUlf' UEt 8 f M# A fiGN CONru fiO8ei tN # Al AH A) A f PcKt f an sv>Wt H Pi ant S MA,79 p NF t e gg4 U#i % f4E SIN (Otni f"vlH4 f SfUAA ( S AND U?A%f AL h(4CN3 NUNIO'LH 34#4 V92 O'LUF A f K>P4AL Os F4 Hi l M f 0N A t Pel Llf AH POWI H h ANf 4 Arma!*.f H pegreg Ay Of $eseg twj Hee j CAWONA38 op ie eis.t. awn 8%,er te Arvj Al AH A NUngq/CH 2}11 V04 N2 sap (TV H( %f ARCH Pf 80GH A V4 %fV)Pd NbHl WL.H 4Q O fHf 740t f OF Pf H%( A AL A.H %AVPt seeQ iP4 H ADt %QHt 0 Of Of FiCE OF NUCt F AH 8'f Gift A f 0Hf AfiGN hAF L f Y PHOT'eHAUS AND Hf Mit f S Of A 6 A Hf 41 A f 0f t y 14E 5( AnCH Quener*y Pnpegg Hoppt Apr41 Jurg 10 t944 f V At U A f 4JN OF Pf HV;4 AL AiH RAUPt;Ppn (QM Vt Nf 184 Subject index NUREG'CR 4254. OCCUPATONAL DOSE REDUCTON AND ALARA Radioactive Waste AT NUCLEAR POWER PLANTS Study On HighDose Jees Radweste NUREG 0017 H01 CALCULATION OF HELE ASES OF RADIOACfivE Handling And ALARA Incentwes MATER:ALS IN GASEOUS AND LiOUiD UFLUENTS FROM PRES-NUREG/CR 4373. COMPENDIUM OF COST EFFECTIVENESS EVALUA- SURilED W ATER RE ACTORS (PWH. GALE CODE) TONS OF MOOtFICATONS FOR DOSE REDUCTION AT NUCLEAR NUHE G 0946 AN EV ALUAflON OF RAD 80NUCLIDE CONCE NTHA-POWER PLANT 1 TONS IN HsGH LEVEL R ADtCACfivE WAstts ,' NvREG 1164 INFOHMAflON ON THE CONFINEMENT CAPAstLliv Or Radiation Protection traenin9 THE FACILITY D:SPOSAL AREA Af WEST VALLE Y.NEW TCHE NUREG-It27 RAD!Af!ON PaOTECTION TRAINING AT URANIUM HEX- NUREG/CH 1755 ADD 01 f f CHNOLOGY. SAFETY ANO COSTS OF DE-AFLUORCE AND FUEL FABRFCATON PLANTS COMMtSSIONiNG NUCLE AR RE ACTOHS AT MuttleLE HE ACTOH NUREG.1134 RADIATON PROTECTION TRAINING FOR PERSONNEL STAfiONS Effects On Decommissiorwng Of interim Inst. ,ty To Depone EMPLOYED IN MEOCAL F ACILITlES of Wastes Ortsste NUAEG/CH 34t3 OFF Sif E CONSE QUE NCE S OF R ADIOt OGOAL Rad 6ation Safety Program ACCCENTS METHODS, COSTS AND SCHEDULES FOH DECON NUREG 0940 V04 NO2 ENFORCEMENT ACTIONS StGN1FICANT AC' T M NATON TONS RESOLVED Ouaderty Progress Report.Apr4Vune.19tl5 NUHFG/CR 37t0 LABOHAfORV STUDIES OF A DHE ACHED NUCLE + AA WASTE REPOSifony IN BASALT Radianon Safety Survey NUREG/CR 3774 V02 ALTEHNAflVE METHODS FOH D6POSAL OF NUREG/CR-4190 CALIFORN A OFFSHORE SURVEY OF LICE NSEES LOW (EDEL RADICACTfvE W ASTES fask 2A f ednecal Requvements U$iNG RADIOACTIVE MATER:AL FM HWW VM Dim # 0 W W Raoact've Wave NU'4EG/CH 3774 V03 Atif HNATIVE METHOOS FOH D5POSAs OF * $ ** *' "N"#'#'" ' N E CR 8 V03 STRUCTURAL INTEGRITY OF WATER REAC- For Aboveyound Vault Dmposal Of Lew Levet Radeactne Wete TOR PRESSURE BOUNCARY COMPONENTS Annual Report F o' NUAEG:CH 3774 V04 Atit HN A tlVE Uf. f HOD F OH CrSPOSAL Os LOW LEvf L AADOACf fvt W ASTE f ad 2C forhrwai Rmremer ts NUREG/CR 4437 EXPLORATORY STUD!ES OF ELEMENT INTEnAC- ' ' "" L U* t**' "# flONS AND COMPOSITON DEPENDENCIES IN HADIATON SEN% TMTV DEVELOPVENT' NUREG CR 3714 VOS At f f RN Af fit ME THOOS FOH D6POSAL OF fiadtstion StatHhty LOW LEVEL HADCACfrSE W ASTE f asa 2E fechncal Requnements I NUREG/CA 3829 AN EVALUATON OF THE STAR'Liff TEsis REC- NUrY E G H LOW VIL Li HW $E SHALLOW LAND i OVVENDED IN THE BRANCH TECHNICAL POSif 0N ON W ASTE BURi AL f nE NCH ASOLAf!ON F eal Report Otteer 1981 Septernt.or 4 FORMS AND CONTAINER MATER'ALS % NUnf G/CA 427t RE COUVE NOf D SAF F f y nf tIABILITV QUAlif y Radioactive Emiss6on NUREG/CR-4068 $UVMARY OF HiSTOACAL EXPEn;ENCE WiTH RE. MSun ANCf' AND MANAGEUF NT Af HOSPACE ff CHNOUE9 W1TH LEASES OF HADCACTIVE MATERtALS FROM COVWRCIAL NU PO%BJ AFWTON M THf M TO THE HM M HAM ACitVE W A$fE Hf POSITOHY PROGR AM CLEAR POWER PLANTS IN THE UNiff D STATES Ree* oetement Trensport Radeoective Oas NUREG/CR 4215 TECHNCAL F ACTORS AFFECTING LOW LEVf L NUHE G/CH If f 0 L AnonATOHV STUDFS OF A fiAE ACHf D NtKit AR W ASil Rf PO5170Hy IN DASAL T W ASTE FORV ACCEPTANCE CRifEn'A Ead6cactive loome Radiographer NUREG/CR 3910 CHAHCOAL PE H8 0HVANCE UNDER ACCfDE NT NUHEG 0940 404 407 (NFOHL f Mf NT ACTONS SGN!T CANT AC. TONS HESOLVED Quarter *v Pety.se Heport Apro June 1ims CONDtflONS IN LIGHT.W A Tt A RE ACTORS Radioactive footopes Radeoseotope NU64 E G/ CR.3Aa 5 EVALUAfCN OF THE RADtOAC*iVE #NVENTORy NUf4EGICH 4094 FIELO DPf H utNf DE!IHU NATOS4 0F Ds4fh8-IN AND ESf!MATU OF ISOICPO RE LF ASE S HCM.fHE W A$fE rN UUf 0N COUt st f Nt S OF AC f >N Dt ( L t MI N f S its Sut F A f E EIGHf TRENCHES Af THE SHEFFiELD LOW LEVEL W ASTE BUHtAL LA* f ENV'HONME Nf1 Siff Radiological Assessment NUHF G/CR 3138 E NVtRONyt NT AL ffffCf5 07 fHf t;nANou Red 6cacttve heaterial NUnEG4383 V01 Ro# OtnECTOHY 06 Cf Rf FCATES OF COVPti- FUt L cvCLE A Hev+w 7e Dea For 7were%m ANCE FOH HAD OACTIVE MAff.HiALS PACK AGES Summary Heport F4Uhl O'CH 447 HADIOLOGsCAL A%f t %MENf 05 THE TL;WN OF j 08 NRC Approved Patnage. EDGEMONf. NUnEG 038) V02 R04 DmtCf04Y OF Cf nf feCATES OF CGVPL8- Radionucl6de ANCE FOR RADCACitVE MAfENIALS PACw AGE S Certha'es of NUs4E G 0944 A P4 f v ALUA flON OF DACICNUCLIDI CONCENTHA-CorrphaNe NUREG can) V03 Rc5 DtnECf0Hf OF Cf Rf CCAfES OF COMPLt TONS IN HiGH (f vt L RADiOACflVF W Aif f 9 ANCE FCH RADICACTlyE MAf[ sbALS PAC # AGE 5 Sumanary Hepr>rt NUHf G 1W IN8 ONUAfiON CN THf CON 8 NE V8 NT CAPAltetstY OF fHf: F Acetitv D6POsAL AHr A Af Wt %f VAg(f f Nf W YOHK Of NRC Approved Qua4ty Assurance f% grams For Ra+oactve Mate,, NUHIGftPS E 5HP 1 f 1 ' ENyiHONVf Nf AL IUPACT4 OF PO9 f u-ge pop Pet / nf G/CH-}019 DECOMVf NCF.D WELOFO CHif tf0A FOH USE IN L Af f D ACClCE N f 3 INv0L v'P+G HE LE ASE S GF HADtOACflVE MA. iI 4 At 9 TO GHOUNOW A f f H ' THE F ABA.CATON OF SHIPPING C047A'Nt H4 FOH HAD'OActivt UA ff 80Alg H M G/CH 2MO vol POPut AflON DO%E CCMVif MI Nf 3 DUE TO NUHf G/CR 3916 A SUHVE v OF tHE USF S OF HADCACfirE MAfE- He t CACrivt Hf LE AM S iHOU NUCLt AH POnt R PL ANT nif f S IN ' fM t HIALS IN LOUISIANA 9 OF f SHOHf W Af f HS NUHfG/CH 1egt VO) PHOGHT % qN FV ALU A fiON cf MADIONU NUPf o/CA 1611 RADtOACilVI! MATE RIAL (HAV) ACCIDE N T /1NC4 Ct IDE GI 05. Hf MicAL its OHM A f x ;N Dt. Vi t Os f D ny 00: HmH. DrNf DAf A ANAL v%S PHOGnAv ifVEL Ni#;LE AN W A$f f H(%POSifDstf Si1E PHWI Cf S etepnet NUnfG/CH 3657 PHELIMINAHf SCAf f NM OF FUTL CvCt f AND 9 f PRODUCf M A f f FO AL LICf NSF 9 FOH f Vf HGF NCY PL AN4NG F re Apris June t #4 NURf G/CH 38154 F Ah80CAfiON CHitt Hr A f OH SHmNG CONT A44 NHHf G/CH Mt V94 iV ALUAflON OF HADIONUCLICE GI DCHIMi-Eng CAL INi OHM A fiON Di VI LOP ( D H f Opf H%H Lt yf ( Ntyg i Ay NUHf G/CH 43$7 f Hf F f AS@lif Y OF DE f f Cf M THE IMDOHf OF W A s f E nE PGst f 0H f 9 tg PnO;t Cf 3 Am,.i rw9.n H prn, go, UNAJfHQH1f D HADCACTIVE M4ff H ALS sNfD THE UNiffD Or tut.e t u t %ptamt, t w $ FATES NUhl G/CH 4010 N ADIONUCLIOf M6H A fiON IN GHOUP#D W A fI H (t snat Heps at) flad60ective Re6eese ht;Hi rpCH 4 t g t AwAy Op g DNo g tyg g Haptr,paigg epg g ,p4 ; gg NUHEG/CH Disa ygj pqpyg Af 0f4 DQgt; COyyttyf Nf g OUE in ggggg4ggggyngypqgggggyggg P4UHF G/CH 4t&# AN apt' HOAR H 70 f H1 AT M H AbiOP40Clit t RADCActrvE HELE ASES FR/M NUCLI AH POWTH Pt ANT Siffi l iN 199 t Of CAf Hf Af M f OH U%f IN f Hf Ut t Myt Cf)(t Sr37g M l f t Subject index 185 NUREG/CR 4181. LEACHABILITY OF RADIONUCLIDES FROM NUREG/CH 3M3 V03 PHOOAFPuf v Or PtPE FAILURE IN THE RE AC-CEMENT SOLO 6FIED WASTE FORMS PROJUCED Af OPERAflNG TOR COOLANT LOOPS OF COMBOSilON ENGINEER!NG PWR NUCLEAR POWER REACTORS PLANTS Voaume 3 Doutae Ered Gudiosne Bream irstirecty Irwfuced By NUREG/CR-4185 AN ASSESSMENT OF DOSJMETRY DATA FOR AC- E arthquakes COENTAL RADONUCLOE RELEASES FROM NUCLEAR REAC- NUREG/CH-4290 V02 PHOBA8ttifY OF PIPE F A! LURE 6N THE REAC. ^ NURE /CR-4210 MATADOR A COMPUTER CODE FOR THE ANALY. sis OF RADIONUCUDE BEHAVOR OURiNG DEGRADED CORE AC-CIDENTS IN L GHT W ATE R RE ACTOAS Reactor Coolant Pressure Boedary NUREG/CR 4211. MATADOR (METHODS FOR ThE ANALYSIS OF NUREG 1095 EVALUATON OF nESPONSES TO IE BULLEfiN 82-A PT AND $R AN L da n eaM Faste 8n ReadW Co$dM Nsture NUREG/Cfb4236 Vot- PROGRESS'IN EVALUATON OF RADONU. Boundary O ReWM WaWReador heeft CUDE GEOCHEM' CAL INFORMATON DEVELOPED BY DOE NfGH, R or MM W OR TOBER E EV R 194 URf GICR-4077 REACTOR COOLANY PUMP SHAFT SE AL BEHAV-NUAEG/CR-4236 V07 PROGRESS IN EVALUATON OF RADIONU- OR DURJeG STAfiCN BLACKOUT CUDE GEOCHEMICAL INFORMATON DEVELOPED BY DOE HIGH-LEVEL NUCLEAR WASTE REPOSITORY bifE PROJECTS Report for Reactor Coolant Pump Seal Ja%ary March 1985 NUREG/CR 4400 THE iMPACf OF VECHANICAL AND M AiN TI . NUREG/CR 4237 MOO!UTY OF RADONUCUDES IN H8GH CHLORCE NANCE INDUCED FA'LUHE S or MAIN HEACTON COOL ANT PUMP ENVIRONMENTS SEALS ON PLANT SAFETY NUREG/CR 4245 IN PLANT SOURCE TERM MEASUREMENTS At BRUNSWICK STEAM ELECTRC STATON Reactor Coodant Systern NUREG/CR 4251 V0t MtT GA TIV E TECHN'OUES FOR GROUND- nun E G.1167 TPCWH2 THf RM AL POWER Of if RM:NATON 7OA W ATER CONTAMNATON ASSOC'ATED WifH SEVERE NUCLEAH WE Sf:NGHOUSE RF ACTCHS3f RSON2 User e Gude ACCCENTS Volume 1 Anaps Of Gewe S<te Conditions NUHEG-CR 4205 TRAP MELf 2 USE A S MANUAL NUREG/CR 4251 V02 Mif GAfiVE TECHN OUE S FOH GROLNO WATER CONTAMJ4ATON ASSOCIATED WifH SEVERE NUCLEAA Reactor Core Thermal Hydraunc ACCOENTS Volume 2 Case Stufy Ar*aryses Of Hydrocyc s Characto" Gat'on And M tqative Schemes NUHEG/CR 4240 V31 PHY S CS OF REACTOR SAF E f t Quarter +v Report,Jafwery March tp5 NUREG/CR 4382 CONCENTRATONS OF URANrUM AND THORIUM fSOTOPES IN URANrUM MILLE AS' AND M NER$' TISSUES p, ecto, o,.,gn NURE G/ CR-43F IN PLANT SOURCE TE AM MEASUREMENTS AT PRAiPrE ISLAND NUCLE AR GENERAf!NG STATON NUHEG 1070 NRC Policy ON FUTURE REACTOR DESGN3 Deowrs On Severe Accmunt lowes in Nwear Po ee Radenuchde Transport P' ant Repnon NUPE G/ CR-3442 RADTWO A COMPUTER COD ( FOR S>MULAt-NG , d F AST TR ANSI [NT, TWO-O' MENS 4CNALTWO L AYF A RADONU- Reactor Operator CUCE CONCE NTRA TON CONDef0NS IN NUREG/CR 40$1 AS$f SSME NT Or JOU At L A TF 0 FDUCA TONAL LAaES RESERVO!RS ArVERS ESTUAR'ES.AND COASTAL REGIONS OUAUFICA TONS FOR NUCL F AR POWE R Pt ANT opt A Af 0HS ' NUHEG/CR 419 9 $URVE Y OF UCE NSE F CONTROL HOOM HABif-Radiopharrnaceutical Pact age AUll if Y PH ACilCf S NUREG/CR 4035 A HGHWAY ACCCENT INNOLviNG RADIOPHAR MACEUTICALS NEAR BROOttHAVEN VSSiSSIPP1 ON DECEMOER Reactor Operators Ucensing 1 ; 3,1983 NUHFG/CR 4280 THE EFF ECTS Or SUPERvtSOR (1lPf foe Pett AND Cadtum Treatment A%3f ANCE OF A SniFT TECHNCAL ADVisOH (St A) ON CHFW PERFOHM ANCE IN CONTHOL ROOM S MULATOHS NUREG/CR 4259 TAILINGS NEutRAU2AfiON AND OfHER ALTE R j NAfrVES FCR IMMO8fullNG toxic MATERIALS IN 7AJUNGS Finaf Reettor Pressure Boundary Report l 1 NUnE G/CR 382S V03 4 ACOUSTC E MtSSiON ' FLAW Hi L A flON9'iP FOR IN SF HWCE MONITOH+NG Of NUCLI AR PHE MUDE NUAE /CR 4245 IN PLANT SOURCE TERM MEASUAE MENTS Af P4U EG/C $ P A 10 L E RFOUCflON AND ALAHA nun H4m W A Sic ( %m4f t AW HELAWH+ Af NUCLEAR POWER PLANTS Study On t*gh Dose Aes RMees's FOH INSFRvCE MONif 0H NG Or NUCL E AR PHISsunE san,p,ng AfvgALAR4inc.nc,,. VE %ELS Progress Hefnt M*er 4 arch W NUAEG'CH 4W1 14 PLANT SOURCE TERM MEASUntME NTS At PRA-R:E ISLAND NUCLE AR GENERAfiP44 ST ATON - "' ' # ' 8 " g Reectson lunetice (NTS) HYDHOGEN flunN fist Af THE CENinAL Rf.CEI)LR f(St NUREG'CP 0062 PROCEEDINGS OF fME CONF (RENCF ON fME AP- F ACIUf Y PUCAflON OF GEOCHE WCAL MODELS TO HiGH LEVE L NUCLE AH WA$fE REPOSifCRY ASSESSM(NT, Reactor Safeguarde Meector NUHFG t12% V01 A COMP'LAfiCN OF Af POHf 9 Of THf ACVMHY COMMif fE E ON DFACTOR SAf f GUAADS 195119st4 VoNrne 1 Part NUREG/CR M41 THE RWR PL AN ANALY/FR t ACHS Repor's On Petyect Pwwes (A 6') NURE GtCR N54 HECTR ANALYS.S OF EOU'PME NT TEMPERATUAf p4yng G ,123 V02 A COMPILATON 0+ AFPOHfS OF THf ADVKORY RESPON5f S TO SELECTED HrDHOGEN DUDNS IN AN ICE COP 4 ConytyEE ON Hf ACf 04 SAf f GUAHDS 'W 1984 V9ume 2 Part DE NSE R CON TAiNMENT-1 ACHS Reports On Prowt Rev=es (G P) NUnf G t t2% V03 A COMPtt AiIOP4 Or nr POHf 3 Or THE ADVtsOny P G/ 4 t PROHAsittfY OF P PE F AllUHf IN THE HI AC- CrNwf ftf ON Hf ACTOH SAF(Guanos tW tys4 Vwne 3 Part iOR COOuNT <OOn OF WofwCU$i PWn Pu~TS v--e ' ,,,= y g l,T ;; = pfamn,5c,,,,,, ,On, 1 ggn 73 01 PHOOAulutY Or PIPE F AitunE IN THE Af AC. COUWTf Et ON hl ACf 0R SAf t GUAHOS 195119M4 Volume 4 Part TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS Volurne 2 ACHS Rapsts On (m; S4 Mst (Amtent Amarvs s . Genenc 3 Gmitotne greek trvierer-fty treed Ry f edNtuanes "*'"*) NVAf G/CH yW) Vr;4 PHr4A04 f y OF Pt81 F A4UNE IN THE Hf AC. NUDE G t12% VCS A COMPILAflOf4 OF HE POHf S Of THf MOVMHY 700 COOLANT LOOPS Or WtSfiPMOUsE PWR PLANTS Voeume O 'M*if f f ON H( AC FOH SA8 i 40 ADDS M631 PM 4 v+sne 5 Part 4 Ppe Fe. lure trhed Hy Crar k Gro**h in WeV Coast P'a'ts I AbHS U"P1 U" b*08" SOI* IHIGII ' N"'P AaNry Gwtest NUF EG/CH FA) VQt PHOOAHillf Y OF PWI F A4UHE IN IHE NI AC- NUHE G t129 V99 A UNFfl AIiUN OF HI PUNIS OF IHI AUVA'JHY i TOH COOtANT LOOPS Or CONBUSf 0N (NGiNCIRNO F%H COMW fit E GN HF ACf 0H RAf t OU AHDS 1991 tv64 votume m Part ptANTS Volume t Sumeaary negort 2 ACHS 4, ports On Genenc S4ev ts (HPA . Agg+ewtis C) - - _ . ---_--.___.---__._--c _ _ - - . - - - _ - - - , - _ . - - _ _ . - - - - - , - . - , . - - . - _ - - _ - - - - - _ _ - - , _ . . _ , - , . - - -- 186 Subject index Reactor safety Reference Manoas , f NUREG/CP-0058 V01: PROCEEDINGS OF THE TWELFTH W ATER RE- NUREG/CR 4213. SETS REFERENCE MANUAL ACTOR SAFETY RESEARCH INFORMATION MEETING NUREG/CP-0071. TRANSACTONS OF THE THIRTEENTH WATER RE- Refdi-Reflood ACTOR SAFETY RESEARCH INFORMATON MEETING NUREG/CR-3651 ASSESSMENT OF THE ADEOUACY OF ORNL 8N-NUREG/CR.2331 V04 N2. SAFETY RESEARCH PROGRAMS SPON- STRUMENTATION IN REFLOOO TEST F ActLTTIES. SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH Ouarted Prtwyess Report Apr41 June 30.1964 Reflood NUREG/CR-3361: THL EFTECT OF WATER CHEMrSTRY ON THE NUREG/CR 4166 ANALYSIS OF FLECHT SEASET 163 ROO BLOCKED HATES OF HYDROGEN GENERATION FROM GALVANIZED STEEL DUNDLE DATA USING COBRA.TF CORROSION AT POST LOCA CONDIT ONS NUREG/CR 4167. F LECHT SE ASET PROGRAM Final Report NRC/EPRI NUREG/CR-3485. PR A REVIEW MANUAL Westmghouse Report Number 16 NUREGICR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN-STRUMENTATON IN REFLOOD TEST F ACIUTIES Reform Amendment NOREG./CR-3804 V03 PHYSICS OF REACTOR SAF ETY Ouart*'4 NUREG 1065 RO1 ACCEPTANCE CRITERIA FOR THE LOW E N-ReportMSeptember 1964 RICHED UR ANfuM REFORM AMENOMENTS NUREG/CR 3804 V04 PHYSICS OF REACTOR SAFETY ouarter y NUREG 1065 R01 ACCEPTANCE CRifERIA FOR THE LOW E N-RICHED URAN UM REFORM AMENDMENTS NURWC 3 REACTOR SAFETY RESEARCH '8"*" RE / R 38 04 RE C SAF TV RESEARCH NUAEG 1070 NRC POLICY ON FUTURE RE. ACTOR PAOGRAMS Ouarterty Report. October December 1984 DE SIGNS Oecri.one On Serere Accuser't lasues in Nuc+ ear Power NUREG/CR-3816 V0 f . REACTOR SAFETY RESEARCH Ouarte,9 NUR 6 EACTOR SAFETY RESEARCH Ouarter'y NU EG C 4 DEVELOPMENT.USE AND CONTROL Of MAeNTE. NANCE PROCEDURES IN NUCLEAR POWER PLANTS Protnems And l Report Apni-June 1984 J NUREG/Cn3816 V03 REACTOR SAFETY RESEARCH Ouartorg Rocommandaims ReportJuly~e,eptomter 19R4 NUREG/CR-3816 V04 REACTOR SAFETY RESEARCH Oua torey Regulatory Agenda Report Octeter-December 1984 NURE G-0936 V04 N01 NRC REGULATORY AGE NDA Ouarterty OdVREG/GR-3885 V03. HGH-TEMPERATUAE GASCOOLED REACTOR Report. January March 1945 SAFETY STU0iES FOR THE DIVISION OF ACCID'N T Regulatory Analysse EVALU ATON QuarterN Popess Aeport. July 1-Septemter 30 1964 NUREG/CR-3885 V04 HGH TEMPERATURE GAS COOLED REACTOR NURE G-06e9 R01 USI A 43 REGtJLATOR f ANALY$19 SAFETY STUDIES FOR THE Ovv1SiON OF ACCIDENT Regulatory And Technecal Report EVALUATON Ouarter'y Progress Report. October 1-Decembe, 3 t. t 984 NUREG 0304 V09 N04 REGULATORY AND TECHNICAL RE PORTS Annual Compelafon For 1984 NUREG/CR 4143 REVEW AND EVALUATON OF THE M'LLSTONE TECHNICAL NUREG 0304 V10 Not RE GULATORY ANO UNIT 3 PROSA8!USTIC SAFETY STUDY Corta.nment Faho RE POR TS Completon For Fvot Quarter 19M January March Mortes Ra@ologcal Source Terms Art! Of's.te Consequences NUHE G/ CR-4117 V01. MANAGEMENT OF SEVERE NUREG 0304 Vio NO2 REGULATOHY AND TE CHNICAL ACCIDENTS Perspectives On Manageg Sevare Accxtents in Comme . RE PORTS Comptaban For Serorws Quarter 1985.Ap+ June cial Nuclear pnwer Ptents NUREG4304 Vt3 NO3 REGUL A TORY AND TE CHNiCAL NUR E G/ CH-4177 V02 MANAGEMENT OF SEVERE REPORTS Co@ ate For TPwd Quarter 1965. July Septemter ACCIDENTS Estendog Plant Operating Procedures acto The Severe Regutatory Approacti Acewsent Reeme NUREG/CR 4240 V01 PHYStCS OF PEACTOR SAF ETY Ouartery NUREG 0845104 U $ NUCLE AR REGULATORY COMMISSION POUCY ReportJanuary-March 1985 AND PLANNWG GutOANCE 1995 NUREG/CR43 t e V0t REACTOR SAFETY RESEARCH PAOGRAMS Quarter'y Report. January March 19e Regulatory Gu6 des NURE G/CRa318 v07 HEACTOR SAF ETY RESEARCH NUREG 1125 V05 A CCMPILATON OF Hf PORTS OF IHE ADV:SORf PROGRAMS Ouaderfy Report Apr$ June 1995. COMMiff EE ON REACTOR SAFEGUARDS 195719e4 Volume 5.Part 2 ACAS Reports On Genanc Sut sects (HfGR Regu6atory GeieH Reactor Scram NUREG/CR 3948 EXPER MENTAL RESULTS CF THE OPERAflONAL Regulatory lesues TRANS ENT (OPTRAN) TESTS 11 AND 1-2 IN THE POWER BURST NUREG/CR 4103 USES OF HOP AN RELiABiUTY ANALYSIS PROH-F ACiU TY. As UST C RiSM ASSESSMENT RESULTS TO RE SOLvl PERSON. NEL PERFORMANCE ISSUES THAT COULD AFFECT SAFE TV Reactor Trip NUREG I t S4 LOSS OF MA;N AND AUCUARY FEEDWATER EVENT Reenforced-Concrete Shear Watt AT THE DAVIS-DESSE PLANT ON JUNE 9.1995 NURIG/CR 4274 ANALYSIS AND TESTS ON SMALL SCALE SHE AR W ALLS FY 82 FINAL Rf POR T Reactor Vestof WUREG.1155 Vol. RE SE ARCH PROGR AM Pt AN Reactor VegWs p ,g,gg, NUREG/CR4365 CES GN ANO CEVELOFVE NT OF A SPECIAL PUR- NURf G/CR 4tS9 COMPARISON 08 THE 1941 INFL DISP ( RSION POSE SAFT SYSTEM FOR NONDESTRUCTIVE EVALUATON OF NU- DATA WITH RESulfS FROM A NUMOER OF D4 f f Rf NT MODELS CLEAR REACTOR VESSELS AND PPtNG COMPONE NT S NURt G/CR 4%0 Vot CALCULAfiONAL ME THOOS F OR ANALYSIS OF POSTUL A TE D UF6 RE L i ASE S NUREG/CR 43e0 V02 CALCULAflONAL ME THOOS FOR ANALYSIS NU / 672 i ACOutSITION ANO CONTROL OF THE HSST WATE D W6 RME S SERIES V IRRACIATON EXPEHiMENT Af THE ORA. P'8i4tHi'tY Redof NUREG 1144 NUCtEAR PLANT AGING Rt EEARCH (NPAR) PRO-Ni> REG /CR 4237 MO9'UTY OF RADOFHJCUCES IN HiGH CHLOR 0E ENVIRONMENTS [g"#G V04 M WRCH PROGRAW PUN hhm h Re . ,,m ese a~a-NURIG/CR 30/fi FF A&Ft1Utv STUOY ON THE ACOutSflON OF Lt NUREG 'CR di14 VALFNCE EFFECTS ON THE SORPTON OF NU. CrNstr EVf Nf DAf A CUDEG ON ROCF$ AND MiNEnALS 11 NURE G/CR 38 31 THE IN PL ANT REUARittfY oaf A BASE FOR NV. CLE AR Pt ANT COMPONI Nf 5 Inteern Report Dienne Redos Reect>on NUREG/CR 1514 V02 THE CHEurCAL BEHAVOR OF IOOtNE IN Generatore Battee+s. Chemere Arut inverters NUNI G/CR 387e PHOHAa:Ur y UASt D LOAD CouniNAfiON CHitt-AQUEOUS SOLUTIONS UP TO 150 C al Radabr>rt Redos Con 4tions RIA FOR Of SIGN OF COPERf TE CONTAINMt NT SiRUCTURf S Reeffoot R#ft NUfif G/CR-4010 SOf CIFICAflON OF A HUMAN RF UAustif Y DATA NUREG/CR 4333 STE. GENEV' EVE F AULT ZONE MISSOURI AND IL- DANM FOR CONI UCIWG HR A St GUE NTS OF PRAS FOR NUCLE - LINOiS AR POWER PLANf S Subject index 187 NUREG/CR-4153: APPLICATONS OF FOREGN PROBABILISTC Research SAFETY ASSESSMENT EXPERIENCE TO THE U S NUCLEAR REG-l ULATORY PROCESS. NUREG/CP 0065: TRANSACTONS OF THE 87H INTERNATIONAL NUREG/CR4180: STATE-OF THE-ART OF SOUD' STATE MOTOR CONFERENCE ON STRUCTURE MECHANICS IN REACTOR l CONTROLLERS TECHNOLOGY. Panel Sessaan IK; Status of Research in Structural [ NUREG/CR-4271; RECOMMENDED SAFETY.REUABILITY.OUALITY And Mecharwcal Engmoenng For Nuclear Power Plants l i-ASSURANCE AND MANAGEMENT AEROSPACE TECHNOUES WITH NUREG/CA-3005.

~~SUMMARY~~
OF THE NUCLEAR REGULATDRY COM-POSSIBLE APPLCATION BY THE DOE TO THE HGH LEVEL RADIO- MISSON'S LOFT PROGRAM RESEARCH FINDINGS' ! ACTIVE WASTE REPOSITORY PROGRAM NUREG C Research Uttilaat6cn Report 47 EMERGENCY DESEL GENERATOR OPERATING NUREG 1000 V02 LONG-RANGE RESEARCH PLAN FY 198&FY 1990. NUREG/CR-4353 V03. PROBA81USTIC RISK ASSESSMENT COURSE _ DOCUMENT IONVolume 3 Systems Rehabihty And Ana %s NUR 373. COMP Di OF COST-EFFECTIVENESS EVALUA- GRAMG. PLAN, 144- NUCLEAR PLANT AGING RESEARCH (NPAR) PRO-TONS OF MODFCATONS FOR DOSE REDUCTON AT NUCLEAR 1 POWER PLANTS. Resadual Strese Measurement I ReNetHhty Analye6e NUREG/CR-4287: ENVIRONMENTALLY ASSISTED CRACKING IN i LIGHT WATER REACTORS A'w.ual Report. October 1983 + September i NUREG/CR-4220t REUABIUTY ANALYS.* OF CONTAINVENT ISOLA - 1984. TON SYSTEMS. NUREG/CR-4228. REVIEW OF THE VOGTLE UNITS 1 AND 2 AUXIUA- Rooklue Number System RY FEEDWATER SYSTEM REUA81UTY ANALYSS.
, NUREGNR-4170 AN ULTRA-HGH SPEED RESIDUE PROCESSOR FOR SAFT INSPECTON SYSTEM IMAGE ENHANCEMENT, NUREG/CR-4350 V06. PROBA81USTC RISK ASSESSMENT COURSE DOCUMENTATON Volume 6 - Data Cevelopment. R m Processor Repeer-Wehsed Sta6n6ese Steel .
NUREG/CR 4170- AN ULTRA-HIGH SPEED RESOUE PROCESSOR , FOR SAFT INSPECTON SYSTEM IMAGE ENHANCEMENT. NUREG/CR-3613 V03 N1: EVALUATON OF WELDED AND REPAIR. WELDED STA NLESS STEEL FOR LWR SERVICE Semiannual Report Resistance Factor For October 1984 Through March 1985. NUREG/CR-3876: PROGA8fuTY 8MSED LOAD COM9tNATON CRITE-Repeatabietty RIA FOR DESGN OF CONCRE (E CONTAJNMENT STRUCTURES. , NUREG/CR-4112 V01; INVESTIGATION CF CABLE AND CABLE Reepirator SYSTEM FIRE TEST PARAMETERS Task AJEEE Flame Test. NUREG/CR-3537: E)PEDfENT METHODS OF RESPLRATORY Report To Congrees FROTECTIONilt SUBMCRON PARTICLE TESTS AND
~~SUMMARY~~
NUREG-0090 V07 NO3: REPORT TO CONGRESS ON ABNORMAL ' OF QUAUTY FACTORS NU NUREG/CR-3953 THE USE OF MAG-1 SPECTACLES WITH POSITIVE-4' E T CONGRESS ON ABNORMAL AND NEGATIVE-PRESSURE RESPsRATORS. NU N NUREG/CR4045. UTERATURE REVIEW ON AEROSOL SAMPUNG R POR CONGRESS ON ABNORMAL DEVCES FOR RESPIRATORY FIELD STUDIES. NU $0 EPORT CONGRESS ON ABNORMAL Reeparator Questy Assurance Test 6ng 5 OCCURRENCES Aprd June 1985. NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER EFF4 Reporting System CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR 4ET-
. GENERATED D-2. ETHYLHEXYL SE8ECATE.DI-2 ETHYLHEXYL NUREG/CR-4133. NUCLEAR POWER SAFETY REPORTING SYSTEM PHTHALATE,AND SODtUM CHLOROE, IMPLEMENTATON AND OPERATONAL SPECIFICATONS-4 RepoeNory Reeponee Tree NUREG/CR-3710- LABORATORY STUDES OF A BREACHED NUCLE- NUREG/CR4272 RESPONSE TREE EVALUATON EXPER!Mc'NTAL AR WASTE REPOSITORY IN SASALT. ASSESSMENT OF AN EXPERT SYSTEM FOR NUCLEAR REACTOR NUREG/CR4110: REPOSITORY STE DATA REPORT FOR UNSATU- OPERATORS RATED TUFF. YUCCA MOUNTAIN NEVADA.
R h w E m ne CL S N U NERA SII. NUREG/CR-3455 COMPARsSON OF ODINE. KRYPTON.AND XENON NUREG/CR4134 REPOSFORY ENVIRONMENTAL PARAMETERS RETENTON EFFCIENCIES FOR VAROUS SLVER LOADED AD-RELEVANT TO ASSESSNG THE PERFORMANCE OF HGH-LEVEL SORPTION MEDIA WASTE PACKAGES. NUREG/CR4161 VOI: CRITCAL PARAMETERS FOR A HtGH-LEVEL Rg Weves WASTE AEPOSTORY. Volume 1 Basart NUREG/CR4236 V01: PROGRESS IN EVALUATON OF RADIONU- NUREG/CR-4354- A STUDY OF SEISMICITY AND TECTONCS IN NEW ENGLAND Fanal Report CLOE GEOCHEMCAL INFORMATON DEVELOPED BY DOE HIGH-i LEVEL NUCLEAR WASTE REPOSITORY STE PROJECTSREPORT R ft FOR OCTOBER-DECEMBER 1984 NUREG/CR4236 V02: PROGRESS IN EVALUATON OF RADONU- NUREG/CR.3174 V02. GEOPHYSICAL GEOLOGCAL STUDIES OF CUDE GEOCHEMCAL INFORVATON DEVELOPED BY DOE HIGH- POSSiBLE EXTENSONS OF THE NEW MADRID FAULT ZONE Annual Report For 1983. LEVEL NUCLEAR WASTE REPOSTORY STE PROJECTSReport for January March 1985. NUREG/CR4226. NEW MADRO SEISMOTECTONIC STUDY Actuties Dunng Fiscal Year 1983 Repository Condstion g4p,,, R gi er 3 37 THE SEL ION D T OF ROCK FOR AR. p NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFICATON T ST Sermannual Report Covenng The Penod April 1984 - Septem. Report. October-December 1994 Repoe6 tory Site i NUREG/CR4022: PRESSURIZED THERMAL SHOCK EVALUATION OF NUREG/CR-2663 V01: INFORMATON NEEDS FOR CHARACTERIZA. THE CALVERT CUFFS UNIT 1 NUCLEAR POWER PLANT. TON OF HIGH-LEVEL WASTE REPOSTORY SITES IN SIX GEOLOG. IC MEDIA Masn Report NUREG/CR 40% A REVIEW OF THE SHOREHAM NUCLEAR POWER NUREG/CR-2663 V02: INFORMATON NEEDS FOR CHARACTERIZA- STATON PROBABlUSTIC RtSK ASSESSMENTInternal Events And Core Damage Frequency TON OF HIGH-LEVEL WASTE REPOSITORY SITES IN SX GEOLOG-IC MEDIA.Apperdcas. NUREG/CR-4377. EVALUATONS AND UTILIZATONS OF RISK IM-PORTANCES r 4
- .--rr - - - - . - , . - - - nm.-g, ,1-- , - - - . - .n . ,e-- -.,--,r. - - . . ,-y --m,-v---,-,-,,-
188 Subject index NUPFG/CR-4385: EFFECTS OF CONTROL SYSTEM FAILURES IN SGTR TRANSIENTS.ACCCENTS. AND COHE-MELT FREQUENCIES AT A NUAEG4844 OAFT FC- NAC tNTEGRATED PROGRAM FOA AESOLU-WESTINGHOUSE PAESSURIZED W ATER REACTOR. TON OF UNAESOLVED SAFETY ISSUES A 3 A4 AND A 5 REGARD-NUAEG/CA-43S8: EFFECTS OF CONTROL SYSTEM FAILUAES ON ING STEAM GENERATOR TUBE INTEGRITY Ovan Report For Com-TRANS ENTS.ACCCENTS AND COAE MELT FREQUENC ES AT A rnent. i BABCOCK AND WILCOX PRESSUAIZED WATER REACTOR NUAEG/CR4387; EFFECTS OF CONTROL SYSTEM FAILURES ON SIMMER TRANSIENTS.ACCCENTS AND CORE MELT FAEQUENCES AT A NUAEGICR4396. S.MMER POSTPAOCESSOR MANUAL j GENERAL ELECTA 1C PRESSURIZED WATER REACTOR. A6ek Arudye6e NUREG/CR4336 SIMVER POSTPROCESSCR MANUAL i NUREG 1115: CATEGORIZATION ' OF REACTOR SAFETY ISSUES I FROM A RISK PERSPECTIVE SIMOUAKE l NUREG/CR4342: UNCERTA!NTY AND SENSITIVITY ANALYSIS OF A NUREG/CR4182 VERIFICATON OF Soil STAUCTURE iNTERACTON I MODEL FOR MULTICOUPONENT AEROSOL DYNAMICS- METHOOS-Q6ek Asseeement SIT SG NUREG/CA-2331 V04 N4; SAFETY RESEARCH PROGAAMS SPON- NVAEG/CR-4376 HEAT TRANSFER CARRYOVER AND FALL BACK IN SOAED BY OFFICE OF NJCLEAR AEGULATORY PWR STEAM GENERATORS DUR'NG TAANSENTS RESEARCHOuarterfy Progress Report. Octooer 1 - December 31 1984 Suu.uAUo NUAEGICR4133. NUCLEAA POWER SAFETY REPOAtlNG SYSTEM NUREG/CR4016 V01. APPUCATCN OF Slim MAUD A TEST CF AN IMPLEMENTATON AND OPERATONAL SPECIFICATONS INTERACTIVE CCMPUTE A.B ASED VETHOO FOR ORGAN @NG NUREG/CA 4392- MEASURES OF SAFEGUAACS RISK EMPLOYING EXPEAT ASSE SSUENT OF HUMAN PERFOAVANCE AND PRA (VOSAEP) A Methodology For Esbmat.ng R.sk Impacts Of Sa'e- REUA81UTY voiume e Main Repert. guarcs Measwes-ION Rock Armor NUREGICR4136 SMOwE A Data Reduct.on Papage For Anaps Of NUREG/CR-3747: THE SELECTON AND TESTING OF AOCK FOR AR. Combustion Espenments MORING URAN 10M TAAUNGS IMPOUNCUENTS, SOLA#TS Rock Fracture System NUREG/CR4022 PAESSUR' ZED THERMAL SHOCK EVALUATON OF
' NUAEG/CR4042: A 3-01MENSONAL CCVPUTER MOCEL TO SIVU, THE CALVERT CUFFS UNIT 1 NUCLE AR POWER PLANT LATE FLUC FLOW AND CONTAaNMENT TRANSPCAT THAOUGH A ROCK FAACTURE SYSTEM. 3,3,e Rock Mass Seahng NUREG4956 OAFT FC. AEASSESSVENT OF THE TECHNtCAL BASES FOR ESTIMATING SOURCE TERVS (Dre't Report For Commer't)
~~NUREG/CA4174. ROCK MASS SEAUNG EXPER! MENTAL ASSESS. MENT CF 90AEHOLE PLUG PEAFOAVANCEAnnuat AeportJune~~
'# NUHEG/CR 3767. INTERACTIVE SrMULATOR EVALUAtCN FOR CRT-Qock Riprop GENEAATED DISPLAYS NUAEG/CR-3752 EFFECTS OF HYCROLOGIC VAA1 ABLES ON ROCK SPOS RiPAAP CESIGN FOR URANIUM TAluNGS iMPOUN0MENTS NUREG-080018 2 RO STANDARO REWEW PLAN FOA THE AEV'EW Queos OF SAFETY ANALYSIS REPOATS FCR NUCLEAR POW ER NUREG-0936 V04 N01: NRC REGULATORY AGENCA Quartery PLANTS LWR Editon Rews.on 0 To SRP Secten 18 2. "Sa'e'y Param-ReportJanuary March 1985. eter Dsplay System (SPOS)"
NUREG-oe00 18 2A1 A0- STANDARD AEVEW PLAN FOR THE Rules Of Practice REVIEW CF SAFETY ANAlv515 PEPCATS FOR NUCLEAR POWER NUREG4386 003 UNITED STATES NUCLEAR REGULATORY COM- PLANTS LWR Estion Reveson 0 To Appenen A To SAP Sect.on 18 2. MiSSCN STAFF PRACTICE AND PROCEDURE OiGESTJULY 1972 - " Human Factors Review Gsdehaos For The Safety Paremeter Dspray SEPTEMBER 1981 System (SpcS)- SAFSTOR SROA NUAEG/CR-1755 AD001: TECHNOLOGY. SAFETY AND COSTS OF CE* NUREG/CR4271. PECCMVENCEO SAFETY.AEUAB:UTY.OU AUT Y CCVM;SSCNING NUCLEAR AEACTORS AT MULTaPLE-AEACTOR ASSUAANCE AND MANAGEVENT AEROSPACE TECHNCUES wtTH STATONS Effects On Decomrmss orwng Of intenm snatalery to Ospose POSSIBLE APPUCATON BY THE DOE TO THE HeGH LEVEL RAOC. Of Wastes 0"s.te. ACTfvE WASTE REPOSITORY PAOGRAV SAFT - SSeem NUREG/CA4365 CESIGN AND CEVELOPVENT OF A SPECAL PUR' NUAEG/CR4107: SEQUENTIAL TEST PROCEDUAES FOR DETECT-POSE SAFT SYSTEM FOA NONDESTRUCTIVE EVALUATON OF NU- ING PAOTRACTED MATERIALS LOSSES CLEAA REACTOR VESSELS AND PIPING COMPONENTS. NUREG/CR.4108 CEVELOPMENT OF MC&A ALAAM RESOLUTCN PROCEDURES-NUAEG/CR4170: AN ULTRA-HIGH SPEED RESCUE PROCESSOR Sabotage FOR SAFT INSPECTION SYSTEM 19 AGE ENHANCEMENT, NUREG/CA-4392 MEASURES OF SAFEGUAROS AtSK EMPLOWNG PPA (VOSAEP) A Methodology For Estimateg R>s= lrrpacts Of Safe-guan2s Measm REGICR-3764. BWR-LTAS. A BOtDNG WATER REACTOR LONG, TERM ACCCENT SIVULATON COCE. S.,,5,or,9, NUREG/CR-t755 ADOO1. TECHNOLOGY SAFETY AND COSTS OF DE. NO EG/CR416h FLECHT SEASET PROGRAU Fral AeportNRC/EPRI CCMMISSIONING NUCLEAR REACTORS AT MULTIPLE AEACTOR STATIONS Effects On Decommiss.orwng Of intenm inatWity To Oscose Westmgnouse Report Number 16 Of Wastes Offsste SEFOR Safeguardo NUAEG/CR4375 THEOAY.DEStGN AND OPERATON OF UQUC NOAEG4525 Ato SAF EGU ARDS
~~SUMMARY~~
EVENT UST METAL FAST BREEDER AEACTORS.lNCLUDING OPERATIONAL (SSEL). REVISION to HEALTH PHYSICS NUAEG/CR-38t 7 DEVELOPMENT.USE AND CONTROL OF MA6NTE-NANCE PROCEDURES IN NUCLEAR POWER PLANTS Problems Arvj SETS NUAEG/CR4213: SETS REFERENCE MANUAL Recommendanons
Subject Index 189 NUREG/CR-4093. SAFETY / SAFEGUARDS INTERACTIONS DUR;NG NUREG-0847 SO4 SAFETY EVALUATON REPORT RELATED.TO THE SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER AEACTO4 OPERATON OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND FACIUTIES 2 Docket Nos 50-390 And 50-391. (Tennessee VaHey AuthontW NUREG/CA-4392: MEASURES OF SAFEGUARDS RISK EMPLOYING NUREG4853 SO4 SAFETY EVALUATION REPORT RELATED TO THE PRA (MOSREP) A Methodology For Estimateg Rrsk impacts Of Sa'e- OPERATON OF CUNTON POWER STATON UNIT 1 Docket No. 50-guards Measures. 461 (ulinors Power Company et af) 3,,g NUREG-0857 S08 SAFETY EVALUATON REPORT AELATED TO THE NUHEG/CR-4093 SAFETY /SAFEGL'ARDS INTERACTONS DUR>NG OPERATON OF PALO VERDE NUCLEAR GENERATING SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR 2 AW 3 Docket Nos 50-528M29 W S F ACILIT'E S 530 (Antona Mc SeMe Company. et al WUREG/CR-4271; RECOMMENDED SAFETY.REUABluTY.OUALITY NUREG4857 SO9 SAFETY EVALUATION REPORT RELATED TO THE ASSURANCE AND MANAGEMENT AEAOSPACE TECHN!OUES WITH OPERATON OF PALO VERDE NUCLEAR GENERAftNG POSS:BLE APPLOATON BY THE DOE TO THE HGH LEVEL RADO- STATON, UNITS .1.2 AND 3 Docket hos %528.50-529 And 50-ACTIVE W ASTE REPOSITORY PROGRAM 530 (Ar:rona PutAc Service Companyl NUREG/CR4281 AN EMPIRtCAL ANALYS;S OF SELECTED NUCLEAR NUAEG-0876 Sc6 SAFETY EVALUATON AEPORT RELATED TO THE POWER PLANT MA:NTENANCE FACTOAS AND PLANT SAFETY OPERATION OF BYRON STATION UNITS 1 AND 2 Docket Nos 50-NUREG>CR-4388 AEROSOL BEHAVIOR MODELING (TASK 3) . Sup. 454 And 50455 (Commonweartn Edrson Compang PORT SERVICES FOR RESEARCH AND EVALVATON OF SEVERE NUAEG4881 SOS SAFETY EVALUATON AEPORT RELATED TO THE ACC: DENT PHENOMENA AND MITIGATCN OPERATON OF WOLF CREEK GE NER ATING STATON UNIT 1 Docket No 50-482 (Kansas Gas And Electnc Company.et ad Safety Assurance NUAEG-0881 S06 SAFETY EVALUATON REPORT RELATED TO THE NUREG/CR4377.. EVALUATONS AND UTILIZATONS OF R!SK IM- OPERATON CF WOLF CREEK GENERATaNG STATON UNIT PORTANCES- 1 Dochet No 50-482 4 Kansas Gas And Electre Company et ar) Safety Evaluation Aeport NUREG4887 305 SAFETY EVALL ATON AEPORT RELATED TO THE CPERATION CF THE PERRY NUCLE AR POWER PLANT. UNITS 1 NUREG 0420 SO9. SAFETY EVALUATON REPORT RELATED TO THE AND 2 Docket Nos 50-440 And 50 441 (Cleve.and Esctre iuuminanng OPERATON OF SHOREHAM NUCLEAR PCWER STATCN UNIT CompanW NUR GC 75 S2 F TY vLA b i ELATED TO THE D m THE 3'#* CPERATON OF DIABLO CANYON NUCLEAR POWER PLANT. UNITS c and Electnc enating Com 1 AND 2 Docket Nos 50-275 And 50-323(Paafe Gas And Electnc NL NU $ 5 528: SAFETY EVALUATON REPCRT RELATED TO THE 4687 507 SAFETY EVALUATION REPORT RELATED TO THE CPERATCN OF DIABLO CANYON NUCLEAR POWER PLANT. UNITS OPERATON OF PERRY NUCLEAR POWER PLANT UNITS 1 AND 1 AND 2.Docnet Nos 50-275 And 50-323 (Pacfc Gas And Electne 2 Docket Ms % 440 W M 441 (Nand Esctne muminamg Company) Com ) NUAEb75 S29 SAFETY EVALUATION AEPORT RELATED TO THE NUAEG-0896 503 SAFETY EVALUATION REPORT RELATED TO THE CPERATON OF DIABLO CANYCN NUCLEAR POWER PLANT. UNITS CPERATiON OF SEABAOOK STATON UNITS 1 AND 2 Oochet Nos 50-443 And 50-444 (PutAc Senoce Company of New HampsNro et a ) 1 AND 2 Docket Nos 50-275 And 50-323 (Pac +c Gas And Electnc NUAEG 0940 V04 NO2 ENFOACEMENT ACTONS S GN$ CANT AC-NL $ 5 S30; SAFETY EVALUATON REPORT RELATED TO THE TIONS RESOLVED Ovaneny Progress Report.Apr$ June.1985 CPERATON OF DIABLO CANYON NUCLEAR POWER PLANT. UNITS NUA G-0979 SO3 SAFETY EVALUATON REPORT RELATED TO THE 1 AND 2 Docket Nes 50-275 And 50-323 (Paofc Gas And Ehectnc FINAL CESGN APPACVAL OF THE GESSAR 11 BAR/6 NUCLEAR Compan i ISLAND CESGN Docket No 50447 (General Electnc Company) NUAEG4d75 S31: SAFETY EVALUATON REPORT PELATED TO THE NVAEG4979 SO4 SAFETY EVALUATON REPORT AELATED TO CPERATON OF OfABLO CANYCN NUCLEAR POWER PLANT. UNITS FINAL CESON APPROVAL OF THE GESSAR li BWR/6 NUCLEAR 1SLAND DES GN Docmet No 50 447 (General Eectre Company) 1 AND 2 Docket Nos 50-275 And 50-323 (Paofc Gas And Electnc Co NUREG4969 SO2 SAFETY EVALUATON REPORT RELATED TO THE NUREd 7 SI& SAFETY EVALUATON REPORT AELATED TO THE OPERATON OF AiVER BEND STATON Docket No 50-458 (Gutt CPERATON 08 WATERFOAD STEAM ELECTR6C STATON UNIT States UtStes Company Capn Electnc Power Cocceratwo) 3 Docket No 50-382 flours.ana Power And is ) NUAEG 09e9 S03 SAFETY EVALUATON REPORT AELATED TO THE NUREG-0797 SOB SAFLTY EVALVATON AEP et RTCombed A TO THE OPERATON CF AlVER BEND STA10N Docnet No 50-458 (Gurf OPERATON OF CCMANCHE PEAK STEAM ELECTRC States Ut>htes Company Capn Electnc Power Cooperative) STATON. UNITS 1 AND 2 Docket Nos 50445 And 50-446 (Texas Ut* NUPEG-0969 SO4 SAFETY EVALUATON REPORT RELATED TO THE tes Generabrv Compaey et at) OPE AATION OF A;VER BEND STATON Docket No 50 458 (Gu ef NUREG-0797 Sv)a. SAFETY EVALUATON REPORT AELATED TO THE States UtAtes Company C4,un Electnc Power Cceperstwe) CPERATON OF COMANCHE PEAK STEAM ELECTRfC NUREG-0989 SOS SAFETY EVALUATON REPORT AELATED TO THE STATON UNITS 1 ANO 2 Docket Nos 50-445 And 50-446 (Tesas Ute OPERATION OF RAER BEND STATON Docket No 50-458 tGud t,es Generatm Company et all States Othtes Company Capn Electre Power Cooperative) NUREG-0797 S69 SAFETY EVALUATON REPORT AELATED TO THE NUREG 0991 504 SAFETY EVALUATON REPORT RELATED TO THE OPERATION OF CCMANCHE PEAK STEAM ELECTRIC OPERATON OF UMERICK GENERATING STATON UNTS 1 AND STATON. UNITS 1 AND 2 Occhet Nos 50-445 And 50-446 (Tesas Utg 2 Docnet Nos 50-352 And %353 :PNadeiprua E6ectnc Company) tms Geeeratng Company. et af) NUREG-0991 SOS SAFETY EVALUATON REPORY RELATED TO THE NUREG 0797 Sfo. SAFETY EvALUATON REPOAT RELATED TO THE OPERATON OF L:MERICK GENERATING $TATION UNITS 1 AND CPERATION OF CCMANCHE PEAK STEAM ELECTRO 2 Cbchet Nos 50-352 And 50 35; (PNadeepNa Electrc Company) STATION. UNITS 1 AND 2 Docket Nos 50445 And 50-446(Tewas Uth NURE G-0991 506 SAFETY EVALUATON REPORT DELATED TO THE tres Electne Cornpany. et ar) OPERATION OF UMERCK GENERATING STATON UNITS 1 AND NUREG-0797 St t. SAFETY EVALUATION REPORT AELATED TO THE 2 Docket Nos %352 And 50 353 (PNadelpha Electnc Companvi CPERATION OF COMANCHE PEAK STEAM ELECTAC NUREG 1031 S01 SAFETY EVALUAT60N REPORT RELATED TO THE STATON. UNITS 1 AND 2 Doctet Nos 50445 And 50446(Texas Ute OPE A ATON OF M"LLSTONE NUCLEAR POWER STATON UNIT ties Generatco Comcany et a:) 3 Docket No 50 423 Wertheast N4 ear Energy Compend PdUAEG-0797 Sf2 SAFEfY EVALUATON REPORT RELATED TO THE NUAEG-1031 SO2 SAFETY EVALUATON RELATED TO THE OPER. OPERATON OF COMANCHE PEAK STEAM ELECTRIC ATON OF M!LLSTONE NUCLEAR POWER STATON UNIT 3 Dor. net STATON. UNITS 1 AND 2 Docket Nos 50445 And 50446 (Temas No S0-423 (Nortreast Nuclear Energy Companyl Ut3ktwts Generateg Com .) NUREG t031 504 SAFETY EVALUATION RELATED TO THE OPER. NUREG-0798 SOS SAFE Y VALUATON REPORT RELATED TO THE ATON OF M LLSTONE NUCLEAR POWER STATON. UNIT 3 Docket CPERATION OF ENRICO FERMI ATOM:C POWER PLANT, UNIT NO No %423 (Northeat Nuclear Energy Company) 2 Docket No %341 (Detroit Edison Comea NUPEG-1038 S02 SAFETY EVALUATON AEPORT RELATED TO THE NUAEG-0798 SO6 SAFETY EVALUATON 4E T AELATED TO THE OPERATON OF SHEARON HARRIS NUCLEAR POWER PLANT. UNIT OPERATICN OF FERMb2 Dochet No %341 (Detroet Edison Compa- 1 Doctet No %400 (Carosena Power And Leght Company And North ny) Carohna Eastern Mun.cipai Power Agency) NUAEG-C847 S03 SAFETY EVALUATION AEPORT AELATED TO THE NUREG-1047 SAFETY EVALUATON REPORT RELATED TO THE OP. OPERATON OF WATTS BAR NUCLEAR PLANT,UNTS 1 AND ERATON OF NINE M!LE POINT NUCLEAR STATON UNT NO 2 Docket Nos 50-390 And 50 391 (Tennessee Vaiwy AuthontyJ 2 Docket No %4 to (Niagara Moha*h Power Corporaton.et ai)
190 Subject index NUREG 1047 Sot: SAFETY EVALUATON REPORT RELATED TO THE Safety Matenals OPERAT ON OF NtNE MILE PO;NT NUCLEAR bTATON. UNIT NO NUREG/CR-3998 V03 LIGHT W ATER-REACTOR SAFETY MATERIALS 2 Docket No 50-410 (N agara Mot ank Power Corporaton et s') ENGNEERtNG RESEARCH PROGR AVS Quarterfy Progress NUREG-1047 SO2. SAFETY EVALUATION REPORT RELATED TO THE Report October December 1984 OPERATON OF NiNE MILE POINT NUCLEAR STATON UNfT NO 2 Docket No. 50-410 (Neagara Mona*k Power Corporaton) Safety Parameter Display System NUREG-1048 S01. SAFETY E% ALUATON REPORT RELATED TO THE NUREG 060018 2 R0 STANDARD REVIEW PLAN FOR THE REV:EW OPERATION OF HOPE CREEK GENERATING STATON Docket No. OF SAFETY ANALYSIS REPORTS FOR NUCLE AR POWER 50 354 (Pubic Servce Electre And Gas Corepany Atiante City Elecinc PLANTS LWR Ed> ton Revision 0 To SRP Section 18 2 "Sarety Param-WU 1 8 S02. SAFETY EVALUATON REPORT RELATED TO THE N G 1'82 1 R STANDARD REY:EW PLAN FOA THE OPERATON OF HOPE CREEK GENERATING STATION Docket No REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER 50-354 (Pubhc Senece Electnc And Gas Company.Atlante Oty E'ectre PLANTS LWR Editon Rewson 0 To Appenos A To SAP Secton 18 2. NU 1 8 S03 SAFETY EVALUATON REPORT RELATED TO THE " Human Factors Reve Guidehnes For The Safety Parameter Disp'ay CPERATON OF HOPE CREEK GENERATING STATON Docket No { 50-354 (Pub:c Serwce Electre And Gas Company.Atlante C4y Electre } g g , GENERATED DISPLAYS NU 1 8 SO4. SAFETY EVALUATON REPORT RELATED TO THE Safet Reh Valve OPERATION OF HOPE CREEK GENERATtNG STATON Docket No 50-354 (Pubic Serwce Electre And Gas Company.Atlante City Electnc - PROBLEYS WITH TARGET ROCK SAF E Tv-REUEF VALVES AT NU 1 7: SAFETY EVALUATON REPORT RELATED TO THE OP- BWRS ERATON OF BEAVER VALLEY POWER STATON. UNIT 2 Docket No NUREG/CR4046 DETERM;N:NG CRITICAL FLOW VALVE CHARAC-50 412 (Du@esne Lght Corr pany et al TERtSTICS USING EXTHAPOLATON TECHNOUE5 NUREG4096 SAFETY EVALUATON REPORT RELATED TO THE RE. NEW AL OF THE OPERATING UCENSE FOR THE TRIGA TRA NiNG Safety Research AND RESEARCH REACTOR AT THE UNIVERSITY OF UTAH Docket NUREG-1105 REV'EW AND EVALUATON OF THE NUCLEAR REGU-No 50-407. (Urwversty of Utah) LATORY COMM;SSON SAFETY RESEARCH PROGRAM FOR NUREG 1098 SAFETY EVALUATON REPORT RELATED TO THE RE- FISCAL YEARS 1986 AND 1997 NEWAL OF OPERATING UCENSE FOR THE RESEARCH REACTOR NUREG-1125 V01 A COMPILATION OF REPORTS OF THE ACVISORY AT MANHATTAN COLLEGE Docket No 50-199 (Manharan College) COMVITTEE CN REACTOR SAFEGUARDS.19571984 Vo8ume 1,Part NUREG 1119 SAFETY EVALUATON REPORT RELATED TO THE HE- 1 ACRS Aeports On Pro,cct Reviews (A F) NEWAL OF THE OPERATING UCENSE FCR THE CAVALIER TRAtN- NUREG-1125 V02 A COMPLATION OF REPORTS OF THE ADV.SORY ING REACTOR AT THE UNIVERSITY OF VIRGINIA Docket No 50- COMMITTEE ON REACTOR SAFEGUARDS 195719A4 Volume 2.Part 396 1 ACAS Reports On Proiect Reviews (G-P) NUREw(Unrversty Of Vrginia)EVALUATON REPORT RELATED TO THE it35. SAFETY NUREG-1125 V03 A COMPILATON OF REPORTS OF THE ADVISORY CONSTRUCTON PERMIT AND OPERATING UCENSE FOR THE RE- CCMM;TTEE ON AEACTOR SAFEGUARDS 19571984 Voaume 3 Part SEARCH REACTOR AT THE UN VER$1TY OF TEXAS Dochet No 50- 1 ACRS Reports On Protect beviews 101) 602 (Unrverssty of Texas) NUREG-1125 V04 A COMPILATON OF REPORTS OF THE ADVISORY NUREG-1137. SAFETY EVALUATON REPORT RELATED TO THE OP- COMViTTEE ON REACTOR SAFEGUARDS.19571964 vosume 4 Part ERATCN OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 2 ACRS Reports On Genere Subr e cts (Acceent Analyses Generc AND 2 Docket Nos 50424 And 50-425 IGeorga Power Company et a ) items) NUREG-1138. SAFETY EVALUATON REPORT RELATED TO THE RE- ggREG 1125 V05. A COMPILATON OF REPORTS OF THE ADVISORY NEWAL OF THE OPERATING UCENSE FOR THE TRAIN:NG AND CCMMITTEE ON REACTOR SAFEGUARDS 19571964 Vo8ume 5 Pa1 RESEARCH REACTOR AT THE UNrVERSITY OF M;CHfGAN Docket 2 ACRS Reports On Genenc Subtects (HTGR Regulatory Guidest F4UREG-1125 V06 A CCMPtLATION OF REPCRTS O' THE ADV!SORY NU EG1 i E b7lON REPORT RELATED TO THE COMMlTTEE ON REACTOR SArEGUARDS,19571984 Volume e.Part FULL TERM OPERATING LICENSE FOR M:LLSTONE NUCLEAR ## POWER STATON. UNIT NO 1 Docket No 50-245(Northeast Nuclea' NUREG 1 S CS T E ARCH PR LAN Energy Company) NUREG/CP 0071 TRAP.SACTONS OF THE THIRTEENTH WATER AE-Safety Fuel Systems Research ACTOR SAFETY RESEARCH INFOAMATON MEETING NUREG/CR-2331 V04 N2 SAFETY RESEARCH PROGRAMS SpON-NUREG/CR3980 V02 UGHT-WATERREACTOR SAFETY FUEL SYS- ^ MS RESEARCH PROGRAMS Quarterty Progress Report.Apr& June D nEGULATOR Y g RCH O R 3ms NUREG/CR-2331 V04 N3 SAFETY RESE ARCH PROGRAMS SPON-Safety Goals SORED BY OF FICE OF NUCLEAR REGUL ATOR Y NUREG 1128- TRIAL EVALUATONS IN COMPARISON WITH THE 1983 RESE ARCH Ouar+erty Progress Aeport. July 1 September 30.1964 SAFETY GOALS NUREG rCR 2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON-NUREG/CR-4067.
~~SUMMARY~~
OF BAAAiER DEGRADATON EVENTS SORED BY OFFICE ' OF NUCLEAR RE GULATOR r AND SMALL ACCOENTS IN U S. COMMERCIAL NUCLE AR POWER RESEARCH Ouarter'y Progress Report. Octoter 1 Decemtwr 31 PLANTS 1964 NUREG/CR4068
~~SUMMARY~~
OF HiSTOAiCAL EXPER:ENCE WITH RE- NUREG/CR2331 VOS N1 SAF ETY RESEARCH FROGRAMS SPON LEASES OF RADCACTIVE MATERIALS FROM COUVERCIAL NU- SORED BY OFFICE OF NUCLEAR REGUL A TORY CLEAR POWER PLANTS IN THE UNITED STATES RESEARCH Ouartody Progress ReportJanuary 1 March 31,1985 NUREG/CR4 t T7 V01. MANAGEMENT OF SEVERE NUREG/CR.2531 R03 iNTHOOUCTORY USER S MANUAL FOR THE ACCOENTS Perspectrves On Managing Severe Acc4ents in Commer U S NUCLE AR AEGULATORY COMM SSON REACTOR SAFETV RE-ciel Nuclear Power Plants SE ARCH DATA BANK NUREGIC44177 V02 MANAGEMENT OF SEVERE NUREG/CR 3816 V03 REACTOR SAFETY RESE ARCH Ouano f ACCOENTS Estereng Piant Operahng Procedures into The Severe Report. July-September 1984 Acceent Regime %gg g,CH-4 318 V01 HEACTOR SAFETV RESEARCH NUREG/CR4f 37 SAFETY GOAL SENStilVITY STUDIES PROGRAMS Ouanerty ReportJanuary March 1985 NUREG/CR-4';40 V01 REACTOR SAFETY RESE ARCH SE MtANNUAL g gpg NUREG 0090 V08 NO2: REPORT TO CONGRESS ON ABNORMAL RE A a 1965. g g OCCURRENCES Apnt-June 1985 STUDIES Safety Isave A-47 NUREG/CR4326 V01. EFFECTS OF CONTROL SYSTEM FAILURES Souty Trammg ON TRANSIENTS AND ACCIDENTS Ar A 3 LOOP WESTINGHOUSE NUREG-1129 RADIATON PROTECTION TnAiNiNG AT URANfUM HEL PRESSURIZED W ATER REACTOR Maen Report AF LUOROE AND FUEL FABRfCATON PLANTS Safety Margm Safety-Related Equipment NUREG/CR-3558 HANDBOOK OF NUCLEAR POWER PLANT SEISM:C NUREG-1148 NUCLEAR POWER PLANT FIRE PROTECTON RE-FRAG 1UT;ES. Seistnac Safety Margins Research Program SE ARCH PROGRAM
Subject Index 191 Safety-Related Personnel Setsm6c History NUREG/CR4248. AECOMMENDATONS FOR NRC POLICY ON SHIFT NUAEGICR4430: CURRENT METHOOOLOGIES FOR ASSESSING SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS THE POTENTIAL FOR EARTHOUARnE INDUCED LOVEFACTON IN Safety-Related Systems NUREG/CR-3791: CLOSEOUT'OF IE BULLETtN 79-09 FArLURE OF GE Seismic Margan TYPE AK-2 CIRCUIT BREAKERS IN SAFETY-AELATED SYSTEMS. NUAEG/CP0070 PROCEEDINGS OF THE WORKSHOP ON SEISMC NVAEG/CR-4004 CLOSEOUT OF #E BULLETIN 79-25 FAILURES OF AND OYNAMIC FRAGluTY OF NUCLEAR POWER PLANT COMPO-WESTINGHOUSE BFD RELAYS IN SAFETY RELATED SYSTEMS. NENTS I Satt NUREG/CR4334 AN APPROACH TO THE QUANTIFICATON OF SEIS. MC MARGINS IN NUCLEAR POWER PLANTS NUREG/CR-2663 V01: INFOAMATION NEEDS FOR CHARACTER!2A-TON OF H,GH LEVEL W ASTE REPOSITORY SITES IN Six GEOLOG- Seisme Quai bcation IC MEDIA Main Report. NUAEG/CR 2663 V02 INFOAMATON NEEDS FOR CHARACTERr2A. Seesm6C Reflection TON OF HIGH-LEVEL WASTE REPOSITOAY SITES IN Six GEOLOG. NVAEG/CR4339 A REVIEW OF RECENT RESEARCH ON THE SEIS-IC MEDIA.Appereces. MOTECTONICS OF THE SOUTHEASTERN SEABOARD AND AN EVALUATION OF HYPOTHESES ON THE SOURCE OF THE 1686 Saturated Fractured Rocks CHARLESTON, SOUTH CAROUNA E ARTHOUAkE. NUAEG/CR 3736. FIELD AND THEORETCAL INVESTIGATIONS OF FRACTUAED CRYSTALUNE ROCK NEAR ORACLE, ARIZONA Semsc Research NUAEG-1147. SEISMC SAFETY RESEARCH PROGRAM PLAN , Scenario NUREG/CPOO59 V01. PROCEEDINGS OF THE MITI NRC SEISMC IN-NUREG/CR4360 V01. CALCULATONAL METHODS FOR ANALYSIS FOAMATON EXCHANGE MEETING VOLUME l OF PCSTULATED UF6 RELEASES. NUAEG/CR4095. TEST SERIES 2 SEISMCFRAGIUTY TESTS OF 4
NUAEG/CR-4360 V02. CALCULATONAL METHOOS FOR ANALYSIS NATURALLY AGED CLASS 1E ExOE FHC-19 BATTERY CELLS OF POSTULATED UF6 RELEASES NUREG/CR.4096. TEST SERIES 3 SEISMO FRAGiUTY TESTS OF 4
NATURALLY AGED CLASS tE C&O LCU t3 BATTERY CELLS i Scram NUREG/CR.4097. TEST SERIES 4 SEISMCFRAGsUTY TESTS OF , j NUAEG/CR4262 V01. EFFECTS OF CONTROL SYSTEM FAILURES NATURALLY-AGED ExtDE EMP 13 BATTERY CELLS. ON TRANS:ENTS AND ACCCENTS AT A GENERAL ELECTRC BOIUNG WATER REACTOR Mac Aeport Seesmic moh Aseessment l j NUAEG/CR-4123 SEISMC FRAGILITY OF REINFORCED CONCRETE Sesi STRUCTURES AND COMPONENTS FOR APPLCATON TO NUCLE- i NUREG/CR-4174 ROCM MASb SEAUNG EXPEAIMENTAL ASSESS. AR FACIUTIES ; MENT OF BOREHOLE PLUG PERFORMANCE Annual ReportJune NUAEG/CR4331. SIMPUFIED SEISMC PROBAB1USTO AtSK , i 1983 May 1984. ASSESSMENT Procerkres And bmitations i l Seal Assembly SeismecengMy Test i NUREG/CR-4077. REACTCR COOLANT PUMP SHAFT SEAL BEHAV- NUREG/CR-4096 TEST SERIES 3 SEISMIC FRAGIUTY TESTS OF 4 OR DURING STATON BLACKOUT' NATURALLY-AGED CLASS tE C&D LCU-13 BATTERY CELLS NUREG/CR-4097. TEST SER!ES 4 SEISMCFRAGIUTY TESTS OF l Search System NATURALLY AGED ExtDE EMP 13 BATTERY CELLS. NUREG/CR-3905 V01 Rt: SEQUENCE CODING AND SEARCr4 g,gg
*. SYSTEM FOR LICENSEE EVENT REPORTS User's Guide NUREGICR-3905 V02' SEQUENCE COOING AND SEARCH SYSTEM NUREG/CR-3145 V03 GEOPHYSCAL INVESTIGATONS OF THE d
~~FOR LCENSEE EVENT REPORTS Code Ustmqs. WESTERN OHO INDIANA REGION ANNUAL REPORT (October NUREG/CR 3905 V03 SEQUENCE COOING AND SEARCH SYSTEM~~
~~NUR /C 2 GE YS CAL.GEOLOGCAL STU0iES CF~~
! NOR G E E G SEARCH SYSTEM POSSIBLE EXTENSONS OF THE NEW MADRIO FAULT FOR UCENSEE EVENT REPORTS Coders Manual' NUR G/ 88 MECHANSM ANALYSES FOR VIRGINtA Secumy AND EASTERN TENNESSEE EARTHOUAMES (1978-1984)
NUREGICR-4093: SAFETY /SAFEGUAROS INTERACTIONS DURING NUREG/CR-4317 V01: CANADIAN SEISMIC AGREEMENT.Techrmcal SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR FACluTIES NUR C R OF RECENT RESEARCH ON THE SEIS. j MOTECTONCS OF THE SOUTHEASTERN SEABOARO AND AN i Security System EVALUATON OF HYPOTHESES ON THE SOURCE OF THE 1886 NUREG/CR-4298 DESIGN AND INSTALLATON OF COMPUTER SYS- CHARLESTON SOWH CAROWA EARTHOUAM i TEMS TO MEET THE REQUIREMENTS OF 10 CFR 73 55, G 34 A ONCS W W , ENGLAND Frial Report Seesmic Dee69n 5,,, ,,,,,,n , NUREG-1061 V02 ADD REPORT OF THE U S NUCLEAR DEGULA. ' TORY COMMISSON PtPING REVIEW COMMITTEE volume 2 NUREG/CR 4317 V01. CANADIAN SEISMIC AGREEMENT Techrucai Addendum Summary And Evaluation Of Histoncal Strong Moton Earth- Repor1 %nng 19794985 $ Quake SeesemC Response And Damage To Aboveground industnal Seiemotectonico l j, Pong NUREG 1061 V05: REPORT OF THE U S NUCLEAR REGULATORY NUREG/CR 3178 STRUCTURAL AND TECTONC STUOtES IN NEW VORK STATE Feel ReportMy 1981 - June 1M2 1 COMVISSICN PtP!NG REVIEW COMMITTEE Volume 5 Summary Pong Revew Commsttee Conclussons and Recommendabons NURE G/CR-4226 NEW MADRO SEISMOTECTONC STUDY Actnritses 'I Dunng Fecal Year 1983 NUREG/CR-3558 HANDBOOK OF NVCLEAR POWER PLANT SEISMC
! FRAGIUTIES. Seismic Safety Magns Research Program. Sett-Conta6ned Broething Apparetus Seismic Deeegn CmMe NUREG/CA-3953 THE USE OF MAG-1 SPECTACLES WITH POSITIVE-AND NEGAflVE-PRESSURE RESPIRATORS.
~~NUREG 1061 V02 REPORT OF THE U.S. NUCLEAR REGULATORY~~
, COMMISSON PIPING REVIEW COMMITTEE. volume 2 Eve;uat.on Of Semesca6e i ! Seestruc Dessgns A Revew Of Seismic Design Requrements For Nu- NUAEG/CR 3646 TRAC PF1 INDEPENDENT ASSESSMENT. 6 clear Power Plant Ppng. NUREGICR4278 TRAC.PF1/ MOO 1 DEVELOPMENT ASSESSMENT. I Setemic Hazard Servuecale MOO 2A Foodwater-Une Greek NUREG/CR-3660 V01 PROBABluTY OF PIPE FA! LURE IN THE REAC- NUREG/CR 4189 TRAC-PF1/ MOO 1 INDEPENDENT TOR COOLANT LOOP OF WESTINGHOUSE PWR PLANTS. Volume ASSESSMENT Sermscale MOD'2A Feedwater-Une Break (S SF 3) And !
~~3 Guillotine Break Indrectly Induced By Earthquakes. Steer~~
~~Lee Break (S-SF 5) Tests 1~~
i i 192 Subject Index Semiocale Mod-29 NUREG/CR-3647- DESGN AND FABRICATON OF A 1/8-SCALE 2 NUREG/CR4073: RESULTS OF THE SEMISCALE MOD 20 STEAM STEEL CONTAINMENT MODEL GENERATOR TUBE RUPTURE TEST SERIES. NUREG/CR-3764 BWR-LTAS. A BOILING WATER REACTOR LONG-TERM ACCIDENT SMULATON CODE. f Semiscale Small Sreak Test NUREG/CR.3855 CHARACTER;2ATON OF NUCLEAR REACTOR
NUREG/CR 3772

RELAPS ASSESSMENT.SEMISCALE SMALL BREAK CONTAINMENT PENETRATION . FINAL REPORT.
TESTS S UT-1.S UT-2. S-UT-6.S UT 7 AND S-UT-8 NURLG/CR-3897. HUMAN FACTORS REVIEW FOR SEVERE ACO-l DENT SEOUENCE ANALYSIS. { Sen6er Reactor Operator NUREG/CR 3889 THE MODEUNG OF BWR CORE MELTDOWN ACO- ! NUREG/CR4051: ASSESSMENT OF JOB-RELATED EDUCATIONAL DENTS FOR APPtiCATON IN THE MELRPlMOD2 CCMPUTER QUAUFICATONS FOR NUCLEAR POWER PLANT OPERATORS. CODE. NUREG/CR-3912 MARCH-HECTR ANALYSIS OF SELECTED ACO. I Seneruvity Analyses DENTS IN AN #CE CONDENSER CONTAINMENT NUREG/CR-3904: A COMPARISON OF UNCERTAINTY AND EENSITIV- NUREG/CA-3930 OBSERVED BEHAVOR OF CESIUM.ODtNE AND ! ITY ANALYSIS TECHNIQUES FOR COMPUTER MODELS TELLURIUM IN THE ORNL FISSON PRODUCT RELEASE PROS l NUREG/CR4199:- A DEMONSTRATON UNCERTAINTY / SENSITIVITY GR ANALYS U NG THE HEALTH AND ECONOMIC CONSEQUENCE gggEGICR-3952. SEQUOYAH EQUIPMENT HATCH SE AL t LAWAGE NUREG/CR4055 THE DIO EXPER MENT.COOLABiUTY OF 002 l SensetMty Study - DEPRIS IN SODIUM WITH DOWNWARD HEAT REMOVAL NUREG/CR-4197; SAFETY GOAL SENSTIVITY S1830tES NUREGICR4064 STRUCTURAL RESPONSE OF LARGE PENEIR.A-TtONS AND CLOSURES FOR CONT AiNMENT VESSELS SUBJECTED i Seno6tiaation TO LOADINGS BEYOND DESGN BASIS. NUREG/CR4287: ENVIRONMENTALLY AS$1STEC CRACKING IN NUREG/CR-4080' DETERViNATON OF THE AVAILABiUTY OF CORE LIGHT WATER REACTORS. Annual Report.0ctocer 1983 - September ExlT THERMOCOUPLES DUR:NG SEVERE ACCIDENT SITUATONS NUREG/CR4005 USERS MANUAL FOR CONTAIN t o A Computer 1984. Code for Severe Reactor Accident Conta nment Analysis , Sequence Coeng NUREG/CR4119 INTEGRITY OF CONT AiNMEN . PENETRAftONS NUREG/CR-3905 V01 R1: SEQUENCE COOING AND SEARCH UNDER SEVERE ACC1 CENT CONDITONS F)84 ANNUAL REPORT. SYSTEM FOA UCENSEE EVENT REPORTS User's Gwde NUREG/CR4130' ICEDF A CODE FOR AEROSOL PART CLE CAP-NUREG/CR-3905 V02: SEQUENCE COOING AND SEARCH SYSTEM TURE IN ICE COMPARTMENTS FOR UCENSEE EVENT REPORTS Code Lishngs- NUREG/CR4137: PRETEST PREDeCTONS FOR THE RESPONSE OF NUREG/CR-3905 V03. SEQUENCE CODING AND SEARCH SYSTEM A 18 SCALE STEEL LWR CONTAINMENT BUILDING MODEL TO FOR LICENSEE EVENT REPORTS Coder's Manual. STATIC OVERPRESSURIZATON NUREG/CR-3905 V04. SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4173 CORSOR USER S MANUAL FOR LtCENSEE EVENT REPORTS Coder's Manuai NUREG/CR4177 V01. MANAGEMENT OF SEVERE ACCDENTS Perspectives On Mana+ng Severe Acceents in Commer-Sequence Coeng And Search System Cd NUREG/CR4071: EXPLORATORY TREND AND PATTERN ANALYSS ggRE C 41 7 MANAGEMENT OF SENERE FOR 1981 UCENSEE EVENT REPORT DATA ACCIDENTS Enterdng Plant Operating Procedures into The See Accdent Regeme Sequential Fastures . NUREG/CR-4209 COMPAR: SON OF ANALYTCAL PREDCTONS AND NUREG/CR 3837: MULTIPLE SEQUENTIAL ' FAILURE MODELEvaluation Of And Procedures For Human Error Oependency EXPER2VENTAL RESULTS FOR A 18 SCALE STEEL CONT AINVENT MOGEL PRESSUnt2ED TO F AILURE. Service Water NUREG/CR4388 AEROSOL BEHAVIOR MODEUNG (TASK 3) . SUP-NUREG 1144- NUCLEAR PLANT 4GING RESEARCH (NPAR) PRO. POFT SERVICES FOR RESEARCH AND EVALUATON OF SEVERE GRAM PLAN ACODENT PHENOMENA AND MITtGATON Service Water System Severe Fuel Damage NUREG/CR4318 V01 REACTOR SAFETY RESEARCH NUREG/CR4233: DISTR:BUTON OF CORBICULA FLUMiNEA AT NU-PROGRAMS Quar 1erh ReportJanuary March 1985 CLEAR FAOUTIES Shaft Dspotel Service Weer NUREG It44. NUCLEAR PLANT AGING RESEARCH (NPAR) PRO- NUREG/CR-3774 V05 ALTERNATIVE METHOOS FOR DISPOSAL OF NUREG C 19 SURVEY OF AGED POWER PLANT FACUTIES. For Shatt Dsposal Of Low Level Radioacttwo Waste NUREG/CR4144 IMPORTANCE RANAING BASED ON AGING CON-SDERATONS OF COMPONENTS INCLUDED IN PROBAB USTIC Sn.i, 2W W WMM WS M GARACp2A NU 4 4 VO AGING AND SERVICE WEAR OF ELECTRIC OF HGH LEm WASTE REWORY STES W $x GEOLOG-MOTOR 4PERATED VALVES USED IN ENGINEERED SAFETYJEA-TURE SYSTEMS OF NUCLEAR POWER PLANTS NU EG/C - 3V INFORMATON NEEDS FOR CHARACTERilA. TON OF HGH LEVEL WASTE REPOSTORY STES IN SIX GEOLOG-
, C MEDIA Apper@ces NUREG/CR4070 V02- BlVALVE FOUUNG OF NUCLEAR POWER PLANT SERVICE WATER SYSTEMSVolume 2 Current Status Of Bio-UREG/CR4414. DRECTCONTACT CONDENSATON OF STEAu ON NU CR-40 V0 B A VE OU N OF NUCLEAR POWER COLD WATER IN STRATIFIED COUNTERCURRENT FLOW PLANT SERVICE WATER SYSTEMS Factors That May Intensdy The Safety Consewences Of B.ofous.ng 33 ,,, ,,,,,,
NUREG/CR 4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED Setnemeng AND PRESTRESSED CONCRETE CONTAfNMENT. NUREG/CR-3972: SETTLEMENT OF URANIUM MILL TAluNGS PILES. A COMPARISON OF ANALYSS TECHNIQUES NUREG/CR-4274 ANALYSIS AND TESTS ON SMALL SCALE SHEAR Severe Acc6 dent WALLS FY-82 FINAL REPORT. NUREG 1070: NRC POUCY ON FUTURE REACTOR DESGNSOecisions On Severe Acesdent issues in Nuclear Power Shan Scheduhng Plant Regulaton. NUREG/CR-4244. RECOMYENDATIONS FOR NRC POUCY CN SHIFT NUREG 1079 DAFT FC. ESTIMATES OF EARLY CONTAINMENT FROM SCHEDUUNG AND OVERTIME AT NUCLEAR POWER PLANTS CORE MELT ACODENTS Draft Report for Comment NUREG-1080 V02. LONG-RANGE RESEARCH PLAN FY 1986 FY 1990 Shsft Supervisor NOREG/CR-2331 VOS N1: SAFETY RESEARCH PROGRAMS SPON. NUREG/CR 4051. ASSESSMENT OF JOB RELATED EDUCATONAL SCRED BY OFFCE OF NUCLEAR REGULATORY RESEARCH Ouarterty Progress ReportJanuary 1-March 31.1985 OUAUFOATONS FOR NUCLEAR POWER PLANT OPERATORS I w
Subject index 193 Shift Technical Adytsor Sod Response NUAEGICR4290- THE EFFECTS OF SUPERVISOR EXPER!ENCE AND NUAEG 1147 SEISMIC SAFETY RESE ARCH PROGRAM PLAN ASS: STANCE OF A SHIFT TECHNICAL ADVISOR (STA) ON CAEW PERFORMANCE IN CONTROL AOOM S4MULATORS Sod-Structure interaction i Shipment Route NUREGICR 4182 VERIFICATON OF SOfL STRUCTURE INTERACTION NUREG-0725 A05. PUBLIC INFOAMATION CIRCULAR FOR SHIP- METHODS l NUREG/CR4329 RELIABluTY EVALUATION OF CONTAINMENTS IN-MENTS OF 1RRADIATED REACTOR FUEL CLUDlNG SOIL-STRUCTURE INTERACTION Shipping Container WUREGICR-3019- RECOMVENDED WELDED CRITERIA FOR USE IN Solenoid Vatve THE FABACATON OF SHIPPING CONTA:NERS FOR RADIOACTIVE NUREG/CR4146 SiUULATON OF AN EPRI-NEVADA TEST S3TE MATERIALS (NTS) HYDROGEN BUAN TEST AT THE CENTRAL RECEIVER TEST NUREG/CR-3854 FABRfCATON CRITERIA FOR SHIPPING CONTAM F ACILITY. ERS Solid-State Motor Controller Short-Pertod Scram NUREG/CR4180 STATEOF THE. ART OF SOL O-ST AT E VOTOR NUREG 0905. CLOSEOUT OF IE BULLETIN 79-12 SHORT PEROD CONTROLLERS SCRAMS AT BOfLING-W ATER REACTCAS Solidification I' g'R /C 44: THE EFFECT OF ENVIRONMENTAL STRESS ON NUREG/CR 3444 V02 THE IMPACT OF LWR DECONTAMsNATONS SYLGARD 70 StUCONE ELASTCMER ON SOLOtFICATON W ASTE OtSPOSAL AND ASSOCIATED OCCU-PATIONAL EXPOSURE SHver Alumina So6utpe Plutoneum NUREG/CR-3455. A CCMPARiSON OF ODiNE. KRYPTON AND *ENON RETENTON EFFCENCIES FOR VAR'OUS SILVER LOADED AD. NUREG/CR 4006 GASTROINTESTINAL ABSORPTON OF P UTONIUM SORPTION MEDIA. IN MSCE. RATS. AND DOGS Applicaton To Estat>hsrung value Of it For Soluble Plutomum. Sdver Seca Get NUAEG/CR 3455 A COMPARISON OF ODINE. KRYPTON.AND XLNON Solute Transport RETENTON EFFCENC.ES FOR VARIOUS SIL%ER LOADED AD- NUREG/CR 4042 A 3-DIMENSIONAL COMPUTER MODEL TO SMU-SOAPTON MEDIA LATE FLUC FLOW AND CONTAthMENT TRANSPORT THROUGH A Sdver Zeohte NUREG/CR-3455 A COMPARISON OF OctNE. KRYPTON AND XENON so,ption AETENTON EFFCENCIES FOR VAROUS SILVER LOADED AD' NUAEG/CR 4114 VALENCE EFFECTS ON THE SORPTON OF NU-SORPTION MEDIA. CLOES ON AOCMS AND M NERALS 11 Sdver-indium-Cadm6um NUREG/CR4401: BEHAVOR OF CONTAOL AODS DUAiNG CORE Source Term DEGRADATON - PRESSUAIZATtON OF SILVER-INDfUM-CADMIUM NUREG /CR4143 AEviEW AND EVALUATON OF THE MILLSTONE CONTROL RODS. UNIT 3 PRCBABiUSTIC SAFETY STUDY Conta4nment Failure Modes.Aadiological Source-Terms And Orts te Consequences Simulation NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCs. Space-Time Correlation DENT SiMULATONS ON POLYMER PROPERTIES NUAEG/CR4072 THE EST!MATON OF ATMOSPHERO OfSPERSON AT NUCLEAR POWER PLANTS UTIL.12:NG RE AL TIME ANEMOME-TER STATISTICS. Simulato'.CR NUREG 4206 A SELECT REVIEW OF THE RECENT (1979-1983) BEHAVIOAAL PESEARCH UTERATURE ON TRAtNtNG S.MULA- Spacer Gnd Heat Transfer TCAS- NUAEG/CR4166 ANALYSIS OF FLECHT-SEASET 163.AOD BLOCKED Sete Cha actertration BUNDtE DATA USING COBRA.Tr NUAEG/CR-2663 V01 INFORMATON NEEDS FOR CHAAACTERf2A- g TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN Six GEOLOG-IC MEDIA Ma,n Report NUAEGICR4030. RADONUCLOE MGRATION IN GROUND NUAEG/CA 2M3 VC2 INFOAMATON NEEDS FOR CHARACTER 12A. W ATER (Final Report) TON OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-UA 3609 EVALUATON OF NEUTRON DOSMETRY TECH-Smoke Transport NiOUES FCR WEL' ' OGGING OPERATIONS WUREG/CR4321. FULL SCALE MEASbHEMENTS OF SMOP E TRANS-PORT AND DEPOSTON IN VENTILAT ON SYSTEM DUCTWORK Spent Fuel NVAEG 0725 R05 PUBLIC INFORMATON CIRCULAR FOR Ship-Snubber MENTS OF IRRADtATED REACTOR FUEL NUAEG-1061 VC2 REPORT OF THE L S. NUCLEAR AEGULATORY NUREG/CR 1755 AD001 TECHNOLOGY. SAFETY AND COSTS OF DE-COMMISSION P!PfNG AEVIEW COMYlTTEE Volume 2 Evaluation Of COMM'SSIONING NUCLEAR REACTORS AT MULTIPLE REACTOR Sesmic Lesegns A Revew Of Sesme Des.gn Requirements For No- STATONS Effects On Decomrmss.oning Of ir'tenm inataty To Dispose clear Power Plant P> pint Of Wastes Oesite NUREGICR4263 REUAmiUTY ANALYul OF STIFF VERSUS FLEXI-BLE PPtNG FINAL PROJECT REPORT Spent Fuel Rod 3,g NUREGICR4074 THE PERFOAMANCE OF DEFECTED SPENT LWR WUAEG/CR4069 ANALYSES OF SOILS 5 AOM AN AREA ADJACENT FUEL ROOS IN INERT GAS AND DAY AIR STORAGE ATMOS-HE L W LE L AADICACTIVE W4STE OtSPOSAL STE AT NUREG/CR 4084 DRV SPENT FUEL STORAGE TEST PLAN FOR DE-NUREG/CR-4083: ANALYSES OF SOILS F2 OM THE LOW LEVEL RA- STAUCTIVE FUEL ROD EXAM >NATONS DOACTIVE WASTE DISPOSAL SITES MT BARNWELL.SC AND NUREG/CR4345 INVE STIGATION OF THE STABILITY OF LWR RtCHLAND,WA s SPENT FUEL RODS BELOW 250 C. Sod Analys#s Spen 0 Fuel RUREG/CR4tt8 MONITORING METHODS} FOR DETERM'NATON NUAEG/CR 4379 V01. LONG TERM PERFOAMANCE OF MATERIALS COMPUANCE WITH DECOMMtSSIONING (tEANUP CA4TERIA At USED FOR HIGH LEVEL WASTE PACK AGING Fwst Quarterey UAANiUM RECOVERY SITES ReportYear Four Apr& June 1985 g 4 o
~~- . - - - _~ --- - - - -..-.- -- .-_- -- - - - _ _ - _ -~~
~~194 Subject index~~
. Stabsitty Analyede NUREG/CR 2118. STEAM ExPLOSON EXPER4MENTS WITH SINGLE i NUREG/CR4116. NUFEGO-NP.A DIGITAL COMPUTER COOE FOR DROPS OF IRON OxlOE MELTED WITH A CO2 LASER Part 3 THE UNEAR STABILITY ANALYSIS OF BOfLING WATER NUCLEAR tt Parametnc Staties.
~~, REACTORS.~~
i Steam Generat6on Stalsielty Test } NUREG/CR-4041 SYSTEM ANALYSS HANDBOOK NUREG/CR-3829: AN EVALUATON OF THE STABILITY TESTS REC-l OMMENDED IN THE BRANCH TECHNICAL POSITON ON WASTE Steam Generator
! FORMS AND CONTAINER MATERIALS. NUREG4844 DRET FC NRC INTEGRATED PROGRAM FOR RESOLU-TON OF UNRESOLVED SAFE;Y ISSUES A 3.A4 AND A-5 HEGARD-Staineses Steel NUREG/CR-3911 V02- EVALUATON OF WELDED AND REPAlR- ING STEAM GENERATOR TUBE INTEGRITY Draft Report For Com-l j WELDED STAINLESS STEEL - FOR LW R SERVICE Ouarterty NU 4975 V03. COMPtLATON OF CONTRACT RESEARCH FOR I NUAW 3 ATIGUE CRACK GROWTH RATES OF LOW, THE MATERIALS ENGINEERING BRANCH.D'VISON OF ENGINEER-I CARBCN AND STAINLESS P'PtNG STEELS IN PWR ENVIRONMENT. ING TECHNOLOGY Armual Report For FY 1984 NUREG/CR4060: THE DC f AND 0C 2 OEBRIS COOLABILITY AND NUREG-1108 RADOACTIVITY TRANSPORT FOLLOWING STEAM i MELT DYNAMICS EXPER:MENTS GENERATOR TUBE RUPTURE.
i - NUREG-1155 V02 RESEARCH PROGRAM PLAN Steam Generators { Steen6ees Steel Pepeng System NUREG/CR 2331 V05 Nt: SAFETY RESEARCH PROGRAMS SPON-NUREG/CR-4221: AN EVALUATON OF STRESS CORROSON CRACK SORED BY' OFFICE OF. NUCLEAR RE GULATORY ,
~~GROWTH IN BWR PtPING SYSTEMS RESEARCH Quarterfy Progress Report. January 1-Maren 31.1985 l I~~
Steineses Steel Weed NUREG/CR 3937. STEAM GENERATOR TUBE RUPTURE ODINE TRANSPORT MECHANISMS Task 1 Esperweental Stumes l NUREG/CR4015: EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG'CR-3949 V01. EDOY.CURdENT INSPECTON FOR STEAM
CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL GENERATOR TUBING PROGRAM Semsannual P ogress Report For

PLATES IN BENDING SERIES 1 Pered Endng June 30.1984 Stansserd Penetration Test NUREG/CR.3949 V02 EDDY-CURRENT INSPECTON FOR STEAM WUREG/CR4430

CURRENT METHODOLOGIES FOR ASSESSING GENERATOR TUBING PROGRAM Annual Progress Report For Pered ;
T E POTENTIAL FOR EARTHOUARKE-sNDUCED LOUEFACTON :N E l kUR G C 40 RE ULTS OF THE SEM) SCALE MOO 28 STEAM GENERATOR TUBE RUPTURE TEST SERIES Standard Review Plan . NUREG/CR 4276 VtBRATION AND WEAR IN STEAM GENERATOR WUREG-0800 06.2.2 R4. STANDARD REV1EW PLAN FOR THE REVIEW TUBES FOLLOWiNG CHEMICAL CLEANING . SE M1 ANNUAL i OF SAFETY ANALYSS REPORTS FOR NUCLEAR POWER REPORT. PLANTS LWR Eston.Revson 4 To Secten 6.2 2, "Contamment Heat NUREG/CR4361 STEAM GENERATOR GROUP PROJECT Annual i a ! Removal System Report . t983 j NUREG-0800 18.2 RO- STANDARD REVIEW PLAN FOR THE REVIEW NUREG/CR-4376 HEAT TRANSFER. CARRYOVER AND FALL BACK If4 ( r i OF SAFETY ANALYSIS REPORTS FOR NUCLEAR FOWER PWR STEAM GENERATORS OURING TRANSIENTS 4 PLANTS. LWR Eeten Revison 0 To SAP Secton 18 2. " Safety Param- < eter D spiav System (SPDS)" Steein Generefor Tut e Rupture ] NUREG-0600 f 8 2A1 RO: STANDARD REVIEW PLAN FOR THE NUREG/CR 4079 ANALYTO STUDES PERTAINING TO STEAM GEN-i REVIEW OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER ERATOR TUBE RUPTURE ACCIDENTS t I PLANTS LWR Eston.Revoon 0 To Appenen A To SAP Secten 18 2. '
~~"Hurnan Factors Renew Guidehnes For The Safety Parameter Dsplay SteenLine Greek i System (SPDSP -~~
NUREG/CR4189 TRAC PF1/ MOO 1 INDEPENDE NT NUREG-0800 ROS- STANDARD REVIEW PLAN FOR THE REVIEW OF ASSESSMENT Semiscale MOO'2A Feedwater Lee Breat (S-SF 3) And SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS LWR StearwLee Break (S-SF-5) Tests ,
~~Esten Revison S To SRP TatWe Of Contents.~~
[ Stearn Water Countercurrent Flow g G/ 2 V01' INVESTIGATON OF CABLE AND CABLE ' STRAT DF
. SYSTEM FIRE TEST PARAMETERS Task A IEEE Flame Test UR 52: SUGGESTED STATE REQUtREMENTS AND CRITE, REG /CH4283. STUDY OF THE EFFECTS OF ELASTC UNLOAD-RIA FOR A LOW-LEVEL RADIOACTIVE WASTE DISPOSAL SITE INGS ON THE #R CURVES FROM COMPACT SPECIMENS REGULATORY PROGAAM.
Steei containment l Station Boeckout . NUREG/CR 3647. CESIGN AND FABRICATON OF A 1/8-SCALE 4 NUREG-1032 DAFT FC. EVALUATON OF STATON BLACxOUT ACCI. STEEL CONTAINMENT MODEL DENTS AT NUCLEAR POWER PLANTS.Tectncal Feengs Related To NUREG/CR-4209 COMPARISON OF ANALYTICAL PREDeCTONS AND l Unresobed Sa'ety issue A44 DraN Report For Comment EXPERIMENTAL RESULTS FOR A 18 SCALE STEEL CONTAlNMENT I NUREG/CR-3764 BWR-LTAS. A BOILING WATER REACTOR LONG- MODEL PRESSURi2ED TO F AILURE. TERM ACCIDENT SIMULATON CODE. , NUREG/CR 3992 COLLECTON AND EVALUATON OF COMPLETE Strategic information Plann6ng , l AND PARTIAL LOSSES OF OFF-SITE POWER AT NUCLEAR POWER NUREG/CR4322 V01 CORPORATE DATA NETWORK (CON) DATA PLANTS REQUIREMENTS TASK Voi t Enterpnse Model NUREG/CR-4077. REACTOR COOLANT PUUP SHAFT SEAL BEHAV- NUREG/CR4322 V02; CORPORATE DATA NETWORM (CDN) DAT A i j OR DURfNG STATON BLACMOUT RFOUIREMENTS TASE Vol ? Data Entdy Octw> nary NOREG/CR4347 EMERGENCY D:ESEL GENERATOR OPERATING NUREG/CR4322 V03 CORPORATE DATA NETWORM (CDN) DATA EXPERIENCE.1981+1983- REOuiREMENTS TASM Vol 3 Data Model
~~NUREG/CR-4322 V04 CORPORATE DATA NETWORK (CDN) DATA REQUIREMENTS T ASK Voi 4 Prohrmnary Strategc Data Pfart~~
/60I)63. PROCEEDNGS OF THE 1984 STATISTOAL SYMPO-( SIUM ON NAitONAL ENERGY ISSUES. Strateg6c Special Nuc4eer Meternal l gg g,,,,,,,, NUREG/CR.4107 SEQUENTIAL TEST PROCEDUAES FOR DETECT.
NUREG/CR-4333 STE. GENEVIEVE FAULT 2ONE.M1SSOURI AND IL. ING PROTRACTED MATERIALS LOSSES gg. NUREG/CR 4 ton DEVELOPMENT OF MC&A ALARM RESOLUTON PROCEDURE 5. Seeam Espeoe6on NUREG-1116. A REVIEW OF THE CURRENT UNDERSTANDING OF Strettfied Det no Sed THE POTENTIAL FOR ODNTAINUENT FAILURE FROM .N-VESSEL NUREG/CR 2951. THE D9 EXPERIMENT Heat Remnval From Stratified STEAM EXPLOSIONS. UO2 Debns b v-m---< ,,,_,.e _.,,-- ,o- y-.,,._,- m , ,. , _ - _ _ , . _.m y_,.-m_., -,__y-m.--y , , _ _ , , -
~~. ~ _ _ _ . _ _ _ _ . . _ . ._ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ ._~~
Subject index 195 Stretified Steam Weter Sulfur NUREG/CR4417: LOCAL PROPERTIES OF COUNTERCURRENT NUREG/CR4121: EFFECTS OF SULFUR CHEMISTRY AND FLOW f' STRATIFIED STEAM-WATER FLOW. 5 RATE ON FATIGUE CRACK GROWTH RATES IN LWR ENVIRON-MENTS. Strettf6ed Two Phase Floar i ' NUREG/CR-4416: STA81UTY OF STEAM-WATER COUNTERCURRENT Summary Informatson Report STRATIFIED FLOW. NUREG-0871 V04 N01:
~~SUMMARY~~
INFORMATON REPORT Data As Of June 30,1985. (Brown Book) I NUREG/CR4387. ORVIRT PCA 2 0 FINITEELEMENT FRACTURE Sump Design i ANALYS15 PROGRAM FOR A MICROCOMPUTER NUREG4800 08 2 2 R4- STANDARD REVIEW PLAN FOR THE REVIEW
, Strees Corroeson Cracking OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER l NUREG-1155 V03; RESEARCH PROGRAM PLAN PLANTS LWR Eaton Revison 4 To Secten 6 2 2. "Contaentnent Heat
~~NUREG/CR2331 V04 N2; SAFETY RESEARCH PR RAMS SPON- Removal System "~~
~~i SORED BY OFFICE OF NUCLEAR REGULATORY Super System~~
~~, RESEARCH Ouarterty Progress Report.Apnl 1 June 30.1984 i i~~
NUREG/CR2331 V04 N3. SAFETY RESEARCH PROGRAMS SPON. NUREG/CR 2331 V04 N2, SAFETY RESEARCH PROGRAMS SPON-f SORED BY OFFICE OF NUCLEAR SDRED BY OFFICE OF NUCLEAR REGULATORY REGULATORY , RESEARCH Ouarterfy Progress Report. September 30.1984 RESE ARCH Ouarterty Progress Recort.Aprd t, June 30,1984 NUREG/CR 2331 VOS N1; SAFETY RES ARCH PROGRAMS SPON-SORED Super System Code i BY OFFICE OF NUCLEAR REGULATORY , RESEARCH Quarterty Progress Report. January 1. March 31,1985- NUREG/CR2331 V04 N4 SAFETY RESEARCH PROGRAMS $PON-p NUREGeCR-3911 V02. EVALUATON OF WELDED AND REPAIR- SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH Quarterty Progress Report. October 1 December 31, WELCEO STAINLES! STEEL FOR LWR SERVCE.Ouarterty ggg4
; Report.Apni-June 1984 j NUREG/CR-3998 V02- UGHT WATER-REACTOR SAFETY MATER;ALS ; ENGINEERING Superpoeation Techruque RESEARCH PROGRAMS Ouartery P' ogress NUREG/CR 3723 STRESS-INTENS TY FACTOR INFLUENCE COEFF6- I DiUR C 3 UGHT WATER-AEACTOR SAFETY MATERIALS " ^ "
ENGINEER:NG RESEARCH PROGRAMS.Quarterfy Progress a.;; ,__ p,,g g,,,,og b i NURE CR4 N L TON OF STRESS CORROSON CRACK NUREG/CR-3317. TECHNICAL BASES AND USER'S MANUAL FOR GROWTH IN r1WR PIPING SYSTEMS THE PROTOTYPE OF SPARC . A SUPPRESSON POOL AEROSOL atava' c=< ' St,e. ,,s.d
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M T ET 0 A' REIG[C 4 PROBA81UTY-BASED EVALUATION OF SELECTED j FIRE PROTECTON FEATURES IN NUCLEAR POWER PLANTS i Strees Corrosson Cracking
! NUREG/CR4422; A REVIEW OF THE MODELS ANO MECHANISUS "" ' ] FOR ENVIRONMENTALLY ASSISTED CRACX GROWTH OF PRES- NUREGICR3319 LWR PRESSURE VESSEL SURVEILLANCE DOS ME-1 SURE VESSEL AND PIPING STEELS IN PWR ENVIRONMENTS TRY IMPROVEMENT PROGRAM LWR Power Reactor Suntedlance l Pntses-Dosenetry Data Base Compendium i Strece-State VarietWe NUR$G/CR4070 V02 BIVALVE FOUUNG OF NUCLEAR POWER NUREG/CR4087; MEASUREMENTS OF URANIUM MILL TATUNGS PLANT SERVICE-WATER SYSTEMS volume 2 Current Statue Of Bo-CONSOUCA TON CHARACTERISTICS}}

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