ML20133Q055

Regulatory and Technical Reports.Compilation for Second Quarter 1985,April-June

ML20133Q055
Person / Time
Issue date:	07/31/1985
From:	NRC OFFICE OF ADMINISTRATION (ADM)
To:
References
NUREG-0304, NUREG-0304-V10-N02, NUREG-304, NUREG-304-V10-N2, NUDOCS 8508150008
Download: ML20133Q055 (88)

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NUREG-0304 Vol.10, No. 2 i

Regulatory and Technical Reports

(Abstract Index Journal) i l Compilation for i Second Quaner 1985

. April - June

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t U.S. Nuclear Regulatory Commission Offico of Administration fa %q e'

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Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's sutscription consists of 4 issues for this publication.

Single copies of this publication are available from National Technical

- Information Service, Springfield, VA 22161 I

I NUREG-0304 Vol.10, No. 2 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for Sccond Quarter 1985 April - June Date Published: July 1985 Policy and Publications Management Branch Division of Technical Information and Document Control l Office of Administration 1 U.S. Nuclear Regulatory Commiasion Wcshington, D.C. 20555 p, ..

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CONTENTS 1 i Preface..........................................................................v i

Index 4

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Main Citation and Abstracts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ~

j S taff R epo rts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,

4 Conference Proceedings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i d Contractor R eports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .  :

Contractor Report Number Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

!, Personal A uthor inde x . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 S u bject I ndex . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 NRC Originating Organization index (Staff Reports) . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . 5 NRC Contract Sponsor index (Contractor Reports) . . . . . . . . . . . . . . . .......................... 6 Contractor index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Licensed Facility index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 ,

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PREFACE l

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 Pubhcations Management Branch Publishmg and Translations Section ,

i Woodmont 501 j U.S. Nuclear Regulatory Comminaion 4

Washington, D.C. 20066 I

The main citations and abstracts in this compilation are listed in NUREG number order: NUREG XXXX,

! NUREG/CP-XXXX, and NUREG/CR XXXX. These precede the following indexes:

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l j Contractor Report Number Index Personal Author Index

! Subject Index j NRC Originating Organization index (Staff Reports) l NRC Contract Sponsor Index (Contractor Reports) l f

1 Contractor Index

• t Licensed Facility index I A detailed explanation of the entries precedes each index.

The bibliographic elements of the main citations are the following:

j Staff Report

NUREG-0508

MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

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

)

I i 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 4

System accession number, (8) the microfiche address (for internal NRC use).

Cnnference Report j NUREG/CP-0017: EXECU1.VE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND

RELIABILITY ENGINEERING IN NUCLEAR REGULATION, JANERP, J.S. Argonne National l Laboratory. May 1981.141 pp. 8106280299. ANL 813. 08632

070. <

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

Contractor Report

, NUf1EG/CR 1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR L!GHT WATER REACTORS CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.: BENNETT, P.R.

Sandia LaLoratories. May 1981.100 pp. 8107010449. SAND 80-0929. 08012:242.

l Where the entries are (1) report number, (2) report title, (3) report authors, (4) organliational unit of l at nors or publisher, (6) date report was published, (6) number of pages in the report, (7) the NRC l Document Control System accession number, (8) the report number of the originating organisation (if l given!, and (9) the microfiche address (for NRC internal use).

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

1 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 NRC-GPO Sales Office or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the NRC-GPO Sales Office send a check or money order, payable to the Superintendent of Documents, to the following address:

U.S. Nuclear Regulatory Commission ATTN: Sales Manager Washington, D.C. 20555 You may charge any purchase to your GPO Deposit Account, Master Charge card, or VIS A chargo card by calling the NRC GPO Sales Office on (300 492 9530. 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 repl'ces 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 Tran*,lations Section of the NRC Division of Technicalinformation and Document Control.

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Main Citations and Abstracts >

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The report listings in this corfoilation are ar- is an NRC contracter prepared report. The i 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 h RC-sponsored report.

- conference report, and NUREG/CR XXXX ,

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' results of certain inspectons performed poor to January 1985 NUMO-0017 R01: CALCULATION OF RELEASES OF RADIOAC- that were not included m prevous issues of NUREG 0040  !

TIVE MATERIALS IN GASEOUS AND LIQU10 EFFLUENTS  !

FROM PRESSURIZED WATER REACTORS (PW R-GALE I CODE). CHANDRASEKARAN. LEE.J V ; WILLIS,C A Divisen of NUMEG-0000 V07 NO3: REPORT TO CONGRESS ON ADNOR.

MAL OCCURRENCES July-September 1984 ' Aroo. 0wector's  ;

Systems Integraten (post 811005). Aptd 1965. 208pp Othee. Aprd 1985, Topp 8505160182. 30456 325  ;

850528036 t. 30603.161. j This report revises the ongsnal lesuances of NUf4E0 0017 Section 208 of the Energy Reorganirahon Act of 1974 edenti.

"Calcutation of Releases of Radioactive Matenale ln Gaseous fies an abnormal occurrence as an unschedufod mcklent or and Liquid Etnuents From Pressurtred Water Reactors (PWR- event which the Nuclear Regulatory Commissen octormmes to i GALE Code)* (Aptd 1976). to mcorporate more recent operateg be signifcant from the starstpoint of put he health arwt safety i

! data now ava table as well as the retutta of a number of m plant and requires a quarterty report of such events to be made to t measurement programs at operat*g pressuttred water reac* ors- Congress This report covers the period July 1 to September i The PWR GALE Code is a computented mathematical model 30, 1984. Dunng the report penod, there were four abnormal for calculating the release 9 of radioactive matenal 6n gaseous occurrences at the nuclear power plants hcensed to operate.

f and bgu6d effluents (ie. the gaseous and liquid source terms) These lnvoNed degraded isosahun vanes e emergency core t j

The U S. Nuclear Regulatory Commessen uses the PWR GALE coohng systems. degraded shutdown systems, a loss of offsite  !

l Code to determeo conformance with the requrements of Ap- and onsite AC elet.tncal power, and a refueling cavity water seat  !

pendiu i to 10 CFR Part 50- failure, respectrvely There was one abnormal occurrence at a I

fuel cycle facdity, the event mvoNed degraded matenal access NUMO 0020 VOS NO3: UCENSED OPERATING REACTORS area bemers There were four abnormal occurrences at the l i STATUS

SUMMARY

REPOM Data As Of retmary othat NRC heensees One invotved contamenated racopharma-28.1965(Gray Book 1)

• Dmsson of Dudget & Analyms Aptd j ceuticals used 6n several dagnostic admnstratens Two in-1985. 4030p 6505t00053 30269 082. VON *d th*apeutic medical misadmnstrations TNe other m. I The OPERATING UNITS STATUS REPORT . LICENSED OP- mNed segmhcant enternal esposure to loane 125 to a hosplal i ERATING REACTORS provides data ott the operaton of nucie- employee. There was one abnormal occunence reported by an 1 J at urvts as timely and accurataty as poseb6e. TNS information is AQ'oement State, the event evoNed contaminated rasophar. [

collected by the Offee of Resource Management from the maceuticals used m severet dagnothe admostraboos The Headquarters staff of NRC's Office of Inspection and Enforce- report also contams informahon updahng some previously re-ment, from NRC's Regional Othces, and from utihtees The trwee ported abnormal occurrences- i i sections of the report are mortht/ highhghts and statistcs for i t commercial operatmg unets, and errata from previously reported NUntG-0000 V07 N04: REPOR7 TO CONGnr.S's ON ADNOR  ;

j date, a completion of dotaded informahon on each unit, provid- MAL OCCURRENCES October December 1944

• AEOO. Direc- l l

ed by NRC's Regional Offices. IE Headquarters and the utihties. for e Othee May 1965. 40pp 8504180402 309e6 013 1 and an appendis for miscellaneous 6ntormabon such as spent Section 208 of the Energy Reorganization Art of 1974 kienh.

i fuel storage capabdity, reactor. years of espormnce and non, hos an abnormal occurrence as an unschedeied meulent or ,

' power reactors en the U S It is hoped the report is helpful to all event wtuch the Nuclear Regulatory Commission determenes to agencies and mdWiduals interested m mamtairung an awarenes, be signeficant from the starwipomt of pubhc health arkt seroty i of the U S. energy stuahon as a who8e. snd requres a quarterly report or such events to be made to (

j Congress TNs report covers the panod October t to December NUMEG0080 V00 N04: LICENSED OPERATING REACTORS l STATUS

SUMMARY

REPORT Data As Of March 31,1985(Graf 3 t.19M4 Durma the toport penod, there were two abnormal oc.

4 Gook l} ROSS,P A; EEEDE.M R Dmsion of Dudget A Anatyve currences at the nuclear power plants tw:ented to operate One May 1965 440pp 8506170349,30959 072 mvoNed four control rods fashng to moort durmg testing arut the See NUREG 0020,V09.NO3 abstract- other invoNed degraded upper head intachon systam accumuta. l tot isolaton valves There was one abnormal occurrence at a NUMEG 0080 V00 NOS: UCENSING OPERATING REACTORS W cycle facd4 me event evoNed buddup of urannam m a STATUS

SUMMARY

MPORT 0ata As Of Apnf 30.1965 (Gray venblation system. There was one abnormal occurrence report.  ;

I Book f) ' Dmeon of Dudget & Anafyste June 1965, 441pp mi by an Agreement State, me event invoNm1 an overosposure ,

8507060197. 31396 287 of a radographer tramee. The repod also contains informaton ,

See NUREG 0020 V09.N01 abstract. uptiabn0 some previousty reportal abnormal occurrencet Numte 0040 V00 Nott LICENSEE CONTRACTOR AND NUMEG 0304 V10 N01: REGULATORY AND TECHNICAL ,

VENDOR STATUS REPORT Ouarterty neport. January March REPORTSComplebon For first Quarter 1985 ' Dmmort of e 1965 (WNte Book) ' Dev6sion of OA, Vendor & Tochtwcal Tra n. Technical informaton & Document Control. Aptd 1985 129pp

! leig Center Programs (Post 850212) May 1985 210pp 850603006g. 30707 00 f, 8505240207.30564 195 l TNs purnal lists att formal toports m the NUREG sorte pre.

7 The penoscal covers the retalts of 6nspections performed by ,

I the NRC's Vendor Program Branch that have been estobuted pared by the NRC staff erwl contractors as woft as proceeengs to the inspected orgerusehons durmQ the penorf from January of conferences aruj workshops The entnas m trie compiahon f 198$ through March 1965. Also 6ncludef in tNs issue are the are Irwiered for access by title and abstract, contractor report i

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2 Main Citations and Abstracts number, personal author, subject, NRC organization, contractor, tions or concerns that have been utentAed to the NHC as of and hcensed facihty, March 1,1985 NUREG-0430 V05 N01: UCENSED FUEL FACILITY STATUS NUREG-0675 S30: SAFETY EVALUATION REPORT HELATED REPORT inventory Dfference Data January 1984 June TO THE OPERATION OF DIAOLO CANYON NUCLEAR 1984 (Gray Book it)

• Drector's Office, Office of Inspection and POWER PLANT, UNITS 1 AND 2 Docket Nos 50-275 And 50 Enforcement. Apot 1985.18pp. 8504290008, 30056 348 323 (Pacif c Gas And Electne Company) ' Dvmon of bcer,ang NRC is committed to the penodic pubhcation of hcensed fa- Apol 1985.137pp 8504220336 29943 202 cilities mventory difference data, following agency review of the Supplement No 30 to the Safety Evaluaton He^ ort for the mformation and completion of any related NRC mvestigat,ons apphcation by the Pacific Gas and Electnc Company (PGAE) to information m this report includes eventory difference data for operate the Diablo Canyon Nuclear Power Plant. Units 1 and 2 active fuel fabocation facihties possessing more than one effoc- (Docket Nos 50 275 and 50 323) has been prepared by the teve lulogram of high ennched uranium, low ennched uransum, Office of Nuclear Reactor Hegulation of the U S Nuclear Hegu-plutonium, or Uranium-233 latory Commissiort This suppiement reports on the staff's tech NUREG-0525 RIO: SAFEGUARDS

SUMMARY

EVENT LIST s d pi e su f (SSELI. REVISION 10.

• bcensing Pohey & Programs Dranch nt2 (Pre 850707). May 1985 59pp 8506140072. 30907;130 NUREG-0675 831: SAFETY EVALUATION REPOAT HELATED The Safeguards Summary Event List (SSEL) provides bnef TO THE OPERATION OF DIABLO CANv0N NUCLEAR summanes of several hundred safeguards related events involv- POWER PLANT, UNITS 1 AND 2 Docket Nos 50 275 Arn150-ing nuclear matenal of facihties regulated by the U S Nuclear 323 (Pacific Gas And Electnc Company)

• Dvision of Licensing Regulatory Commission (NRC) Events are descnbed under the Apnl 1905 17tpp 8505240223 30564 022 categones of bomb related, mtrusion, missing /allegedty stolen, Supplement 31 to the Safety Evaruation Heport for the apph.

transportaton, tampenng/vandaham, arson, firearms related, re- canon by Facehc Gas and Electnc Comptny for hcenses to op, diological sabotage and miscellaneous The mformation con. erste Diablo Canyon Nuclear Power hnt, Units 1 and 2 tarned 6n the event descnphone is denved pnmanly from official (Docket Nos 50 275/323) has been prepared by the Office of NAC reporting channels. Nuclear Reactor Hegulation of the U S Nuclear Hegulatory NUREG 0540 V01 N02: TITLE LIST OF DOCUMENTS MADE "# * " * * # '"

PUOLICLY AVAILABLE February 1 28.1965 ' Dmsson of Tech- related to the issuance of an operating hcense for Dablo nical Information & Document Control. Apnl 1985. 699pp Canyon Unit 2, in particular those issues identited by the NHC 8505160179 30457 296 staff en ea<her supplements, commitments made by the apple-This document to a monthly pubhcaton containing descrip- cant, and certain hcense corutes ecluded in Facility Operat.

tions of 6nformation received and generated by the U S. NRC. ing Ucense No DPR 81 for Dablo Canyon Unit 2 This mformahon ecludes (t) docheted matenal associated with NUREG-0675 831: SAFETY EVALUATION HEPORT HELATED cmhan nuclear power plants and other uses of radcactive me. TO THE OPC HATION OF DIADLO CANYON NUCLEAR tonals, and (2) nondocketed matenat recorved and Generated by POWER PLANT, UNITS 1 AND 2 Dochet Nos 50 275 And 50 NRC pertinent to its role as a regulatory agency The following 323 (Pacif< Gas And Electnc Company)

• Dvmum of Licensing mdenes are ecluded Personal Author inden, Corporate Source Apnl 1995.17 f pp 8505240223 30564 022.

Indes. Report Number inrfes, and Cross Hoference to Pnncipal Supplement 31 to the Safety EvaNabon Report for the app 4 Documents Inder cation by Pacr6c Gas and Electnc Company for hcenses to op.

erste Dablo Canyon Nuclear Power Plant. Urute 1 and 2 NUMEO 0540 V07 N03: TITLE LIST OF DOCUMENTS MADE (poci,et Nos 50 275/323) has been prepared by the Office of PUBLICLY AVAILADLE March 1 31,1985 ' Dvison of Techno cal leformation & Document Control. Aptd 1985 430cp Nuclear Reactor Regulation of the U S Nuclear Regulatory 8505210419 30522 018 Commesseort Thea supp!ement addresses a number of matters See NUAEO4540 V01,N02 abstract related to the usuance of an operating hcense for Dablo Canyon Unit 2, m particular those issues idnnt.f.ed by the NnC NUREG 0540 V07 N04: TITLE LIST OF DOCUMENTS MADE sta4 m earher supplements, comr%tments made ty thu apph.

PUBLICLY AVAILAOLE. Apol 130,1995

• Drvivon of Technical cant, and certain hcense cond tona WNfud(<f m Facer Operat-Information & Document Control June 1985 500pp ing Leonse No OPH 81 for Dablo Canyon Unit 2 8507080219 31395 143 See NUAEG 0540.V01,N02 abstra:t NURE0 0125 Hol: PUOLIC INFOnMATION CiHCULAH FOR S6til'MLNTS OF IARADIATED HrACrOH FUEL. ' Dmson of NUMEG-0606 V01 N02: UN4ESOLVED SAFETY ISSULS Safeguards June 1995 55pp 8'07080117 31394 310

SUMMARY

Data As Of May 17,1995 (Aque Dook) ' Otymon of Ih4 C'rcular has been prepared m response to numerous re.

Engineenng Techno'ogy June 1985 61pp 8507080200 4""*N f* '"b'*8t*" '*98"S"O 'outes uwt for the shipment of 31390 163 "'8'hatW mattot (speny funt sik ed t to mgulabon by the Nucto.

Provwles an overvew of the status of the prog &ss and plan, at Hegulatory Comm.osion (NHC), and to meet the requirements for resolution of the genenc tasks addfessing "UnrescNed of Mhc Law M.495 The NHC statt must approve such routos Safety lasues" as reported to Congress P'*' to ther fast use m accorrlance with the regulatory prove.

sinns of Sec mn 13 37 of 10 CP H Part 73 The er fortnatnin on.

NUMEO-06FS S28: SAFETY EVALUATION Nf POHT HEL Aff D ctwfat reflects NhC stMf kno*' edge as of June 1.1985 Spent TO THE Ort NATION OF DIAULO CANYON NUCL EAH fuel shipment toutes, pnmanly for toad transportaten t>ut a.ao POWER PLANT, UNITS t AND 2 Dochet Nns 50 275 Arwt 50- incluelmg one tail toute, are erecated on reproefur.tinns of DOT 323 (Pacific Gas And Electne Company) ' Dmtion of Licens.nQ road maps Also orkluded tre the amounts of matanal shipped Apnt 1995 6150p 8505100069 30286156 dunnu the approismate three year pennd that satequents toqu.

Supplement No 28 to the Safety Evabaton Haport for the latwms fut spent fums shipments hare treen effective in a4bt on, appbcaton by the Paohc Gas and tiectric Company (PGAE) to the Cvnmission has omsen to prov*fe mfurmahon a this doco-operate the Dablo Canyon Nuttear Power Plant, Units I and 2 ment regardmg the NHC's safety and sa'cuuards regulations for (Do(bot Nos 50 275 and to'323) has baen prepared by the spent fuel shipments as well as safeguards incidents teganbrua Othce of Nucleat Pea? tor Hegulation of the U S Nuclear NJU. spent tunt shipments (of wheth tone have teen reprettod to latory Commission This suppement reports on the status of the date) This adibtional mfr>rmation es furrusheel tiy the Commis.

staffe mvashgaton, 6nspection and evalughon of thoie aflege, sion m order to convoy to the put he a mnte complete peture of I

Main Citations and Abstracts 3 NRC regulatory practices concerning the shipment of spent fuel NUREG-0797 S11: SAFETY EVALUATION REPORT RELATED than could be obtained by the pubhcation of the Shipment TO THE OPERATION OF COMANCHE PEAK STEAM ELEC-routes and quantities alone. TRIC STATION,'JNITS 1 AND 2. Docket Nos. 50-445 And 50-448 (Texas Utikties Generating Company, et af)

• Drvision of L6 NUREG-0748 V05 NO2: OPERATING REACTORS LICENSING censq May %85 34@p. 85M00R 3W8 0R ACTIONS

SUMMARY

Data As Of February 28,1985 (Orange Supplement No.11 to the Safety Evaluation Report for the Book)

Management Support Branch. Apol 1985. 335pp. Tomas Utihties Electne Company apphcaton for a hcense to op-8505070582. 30207 295.

The Operating Reactors Licensing Actions Summary is de. erste the Comanche Peak Steam Electnc Staten. Units 1 and 2 (Ducket Nos. 50-445 and 50-446), located 6n Somervell County, signed to provide the Management of the Nuclear Regulatory Texas, has been pntly prepared by the Office of Nuclear Reac-Commission (NRC) with an overview of hcensing actions dealing tor Regulation and the Comanche Peak Techrucal Review Team with the operating power and nonpower reactors. of the U S. Nuclear Regulatory Commisseon ;NRC) and is m two NUREG-0748 V05 N03: OPERATING REACTORS LICENS,NG parts. Part 1 (Appendix O) of this supplement provides the re-ACTIONS

SUMMARY

Data As Of March 31,1985. (Oran9e suits of the TRT's evaluat>on of approximatefy 125 concerns Book)

Management Support Branch. May 1985 342pp and allegations relating specifically to quahty assurance and Q$06030192. 30688 001- quakty control (OA/OC) issues regarding construction practices See NUREG-0748,V05.N02 abstract. at the Comanche Peak facibty. Part 2 (Appendia P) contains overall summary and conclusion of the OA/OC aspects of the NUREG-0748 V05 N04: OPERATING REACTCAS LICENS'NG NRC Technical Review Team efforts as reported in Safety Eval-ACTIONS

SUMMARY

. Data As Of Apnl 301985(Orange Book)

My agement Support Branch Jane 1985. 400pp uation Report (SER) Supplements 7,8,9 and 10. Issnes raised 8507020440 31310 077. dunng recent Atomic Safet/ and Licensing Board heanngs will See NUREG 0748,V05.N02 abstract be dealt with in future supplements to the SER as needed.

NUREG-0750 V21101: INDE*ES TO NUCLEAR REGULATORY NUREG-0797 Sit: SAFETY EVALUATION REPORT RELATED COMMISSlot. ISSUANCES January March 1985 ' Division of TO THE OPERATION OF COMANCHE PEAK STEAM ELEC.

Technical vdormation & Document Control June 1985 59pp TRIC STATION. UNITS 1 AND 2. Docket Nos. 50 445 And 50-85070802(3 31380 272. 446 (Texas Utilities Generating Company, et al) ' Divison of Li-Digests and indeses for me,uances of the Commission, the consing May 1985. 349pp. 8506190054. 31018 014 Atom +c Sa'ety and L6 censing Appeal Panel, the Atomic Safety Suppioinent No 11 to the Safety Evaluaten Report for the

) and Licens4ng Doard Panel, the Administrative La* Judge, the Texas Utilities Electnc Company apphcaten for a license to op.

Directors' Decrsions, and the Denia!s of Petitions for Rulemak- e, ate the Comanche Peak Stsam Electoc Siston. Units 1 and 2 eng are presentad (Docket Nos60-445 and 50 446), located n Somenett County, Texas, has been pntry prepared by the Office of Nuclear Reac-NUREG 0750 V21 N02: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR FEDRUARY 1985 Pages 275 469

• Divison for Regulation and the Comanche Peak Technical Review Team of Technical informaton & Document Cytrol April 1985 of the U S Nuclear Regulatory Commission (NRC) and te in two 203pp 8504290238 30056 001. parts. Part 1 (Appendia O) of this supplement provnies the re.

Legalissuances of the Commission, the Atomic Safety and U. mutts of the TRT's evaluaton of approximatefy 125 concerns censing Appeal Panel, the Atomc Safety and bcensing Doard and allegations relating specifically to quahty assurance and Panet, the Administratrve Law Judge, and NRC Program Offices quality control (OA/OC) essues regarding construction practices at the Comancho Peak facihty Part 2 (Appendia P) contains NUREG-0750 V21 N03: NUCLEAR REGULATORY COMMISSION overall summary and conclus on of the OA/OC aspects of the ISSUANCES FOR MARCH 1985 Pages 471559

• Divison of NRC Temnical Review Team efforts as reported m Safety Eval-Technical Informaton & Document Control May 1985 78pp uation Report (SER) Supplements 7,8,9 and 10. Issues raised 8505280104 30606 279 riunng recent Atomic Satoty and Licensing Board hea..ngs will l See NUREG 0750,V21,N02 abstract be dealt with in future supplements to the SER as needed NUREG-0750 V21 N04: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR APRJL 1985 Pages 5611,041 Devrson of NUREG 0829 DRFT: INTEGRATED PLANT SAFETY ASSESS i MENT HEPORT, SYSTEMATIC EVALUATION PROGRAM SAN 1 Technical information & Document Control June 1985 490pp  !

ONOFHE NUCLEAR GENERATING STATION UNIT 1 Docket EG 07 V 1,N02 abstrect No 50 206 (Southern CaMornia Ed. son Company)

• Division ot t

Licensing Apnl 1985 55epp 8505'40058 30565 001.

NUREG 0797 Sto: SAFETY EVALUATlON HEPOHT HELATED The Systematic Evaluaton Program was mitiated in February TO THE OPERATION OF COMANCHE PEAK STEAM ELEC- 1977 by the U S. Nucinar Hegulatory Commisson to review the ThiC STATION UNITS t AND 2 Dor.ket Nos 50 445 And 50- devgns of older operating nuclear reactor plants to confirm and l 448 (Tesas Utilit,os Electnc Company)

• Division of Ucensing document their safety Tho review provutes (1) an assessment 1

Apne 19n5 32fipp 8506030061. 30706 008 of how those plants compare wth current hconsing safety re.

Suppinment No 10 to the Safety Evaluation Heport for the querements relating to antected issues, (2) a basis for deciding

! Tenas UtAties Electne Company apphcaton for a license to op- on how these differences should be resolvm1 in an mtegrated orate Comanche Peak Steam Electric Staten, Units I and 2 plant review, arvi (3) a documented evatuaton of plant safety (Doont Nos 50 445 and $0-446), located m Somervell County, This report documents the review of San Onofre Nuclear Gen.

Tenas, has been pntly prepared by the Office of Nuclear Heac- erating Staton, Unit 1, operated by Southern Cahtornia Edison for Hegulaton and the Comanche Peak Technical Heview Team Company The San Onofro i faciWy is one of to plants re-

) of the U S Nuclear Hegulatory Commisson This supplement y,cwmf under Phase 11 of this program This report indicates provides the results of the staff's evaluaton and fotolution of how 137 topics selected for review under Phase i of the pro-opprovimatnfy 400 techrucal concerns and allagations m the gram were addressed Equipment erw] procedural changes have rnochanical and piping area regarding constructon practicos at been idnntified as a result of the revice the Cram 4nche Peak facihty This report does not address the Wahh/Doyte allegatons regarding def<iertties 6n the pipe sup. NUREG-0444 DRFT FC: NRC INTEGHATED PROGRAM FOH port domqn process issues raised by the Walsh/Doyle allega- HESOLUTION OF UNHESOLVED SAFETY 1SSUES A-3.A-4 finns as weil as 6ssues rael durmy tecnnt Atomic Safety and AND A5 HI GAHDING STEAM GF NF H A TOR TURE Ucanvrw; Hoard hearings will be dealt with 6n future supplew IN1t GRITY Draft Hoport For Comment MUHPHY,0 Divivon of monts to the Sa%ty Evaluat on Heport as needaj Lkensing Apol 1985 f f*pp 8505310603 30666 007.

l 1

4 Main Citations and Abstracts This report presents the results of the NRC integrated pro- hcense to operate the Perry Nuclear Power Plant, Units 1 and 2 gram for the resolution of Unresolved Safety issues A-3, A-4, (Docket Nos. 50-440 and 50-441). The report has been pre-and A-5 regarding steam generator tube integnty. The report ad- pared by the OMce of Nuclear Reactor Regulation of the U.S.

dresses issues within the areas of steam generator mtegnty, Nuclear Regulatory Commisson. The facihty is located in Lake plant systems response, human factors, radiologral conse- County, Ohio, approximately 35 mdes northeast of Cleveland, quences and the response of vanous organizations to a steam Ohio. This Supplement, No. 6 addresses the remaining unre-generator tube rupture. A generic nsk assessment is provided solvaJ Atome Safety and Licensing Board contenton issues; and indicates that nsk from steam generator tutw rupture TDI diesel generator rehability in Secton 9 6.3.1, hydrogen con-events is not a significant contnbutor to total nsk at a given site, trol system design per the new hydrogen rule in Secton 6.2.7; nor to the total nsk to which the general pubhc is routinely ex- and several issues related to Emergency Plans in Secton 13.3.

posed. However, the report identifies a number of actions which the staff finds as a group would be effectrve in signihcantry re. NUREG-0910 R01 S03: NRC COMPREHENSIVE RECORDS DIS-ducing the incidence of steam generator tube degradation, the POSITION SCHEDULE.

• Drvison of Techncal information &

frequency of tube ruptures and the corresponding potential for Document Control. Apnl 1985. 22pp. 8505100062. 30286:119.

significant non-core melt radological releases, and occupahonai in comphance with statutory requirements set forth in Title 44 radiological exposures and whsch would be effective in rnitigat- U.S. Code, " Pubic Pnnbng and Documents " and in the opphca-eng the consequences of SGTR events. The actons would also ble regulabons cited in Title 41 Code of Federal Regulatons, further reduce nsk and have been designated as " staff recom. "Public Contracts and Property Management," Chapter 101, mended acbons." Final pubhcation of the report herein, follow. Subchapter B, " Archives and Records," the U.S. Nuclear Regu-ing a 90-day penod for pubhc comment, win constitute technical latory Commission has pubhshed and maintains *NRC Compre-resolubon of Unresolved Safety issues A-3 A-4, and A-5. hensrve Records Dispositen Schedule" (NUREG4910) for NUREG-0857 S08: SAFETY EVALUATION REPORT RELATED

" ' " * " * * ***'" '"* ' ***"I wards changes to the General Records Schedules as made by TO THE OPERATION OF PALO VERDE NUCLEAR GENERAT. * #' #8 " "

ING STATION. UNITS 1,2 AND 3 Docket Nos. 50-528,50-529 And 50-530(Anzona Pubhc Service Company, et al)

• Division enwal c 0 W Musca of Licensing May 1985. 37pp. 8506240081. 31177.312. NUREG-0910 R01 S03: NRC COMPREHENSIVE RECORDS DIS-Supplement No 8 to the Safety Evaluation Report for the ap. POSITION SCHEDULE.

• Division of Technical information &

phcaton filed by Anzona Pubhc Service Company, et af, for h- Document Control. Apnl 1985. 22pp. 8505100062. 30286:119.

censes to operate the Palo Verde Nuclear Generating Staten, in comphance with statutory requirements set forth in Title 44 Units 1,2 and 3 (Docket Nos. STN 50-528/529/530) located in U.S. Code, " Pubic Pnnting and Documents," and 6n the apphca-Mancopa County, Anzona, has been prepared by the Offee of ble regulations citeo 6n Title 41 Code of Federal Regulations, Nuclear Reactor Regutabon of the U S. Nuclear Regulatory " Pubic Contracts and Property Management," Chapter 101, Commisson. The purpose of this supplement is to update the Subchapter 8, " Archives and Records," the U.S. Nuclear Regu-Safety Evaluaten Report by prov@ng an evaluation of (1) addi- latory Commisson has pubhshed and maintains "NRC Compre.

tonal informaton submitted by the applicants since Supplement hensive Records Disposebon Schedule," (NUREG-0910) for No. 7 was issued and (2) matters that the staff had under records created or maintained by the NRC. Supplement 3 for.

review when Supplement No 7 was issued, specifically those wards changes to the General Records Schedules as made by issues which required resolution pnor to plant operation of Unit 1 above 5% fun pow" the National Arch #ves & Records Administration (NARA) and I

General Schedule 20 for incluson.

NUREG-0081 S06: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WOLF CREEK GENERATING NUREG-0936 V04 N01: NRC REGULATORY AGENDAOuarterty STATION, UNIT 1 Docket No. 50-482 (Kansas Gas And Electric Repat, January-March 1985.

• Divison of Rules and Records.

Company,et al)

• Divison of Licensing June 1965. 22pp. May 1985. 201pp. 8505310669. 30666.173.

8506240149 31152 311. The NRC Regulatory Agenda is a complaton of all rules on Supplement No 6 to the Svety Evaluation Report relaicd to which the NRC has proposed or is considering acton and all the operaten of the Wolf Crock Generating Staton, Unit No. I petitions for rulemaking which have been received by the Com-updates the informahon contained in the Safety Evaluaton miston and are pending dispositon by the Commission. The Report, dated Apnl 1982 and Supplements 1, 2, 3, 4, and 5. Regulatory Agenda is updated and issued each quarter. The dated August 1982, June 1983, August 1983, December 1983, Agendas for Apnl and Octotier are pubbshed in their enbrety in i and March 1985 respectively. Supplement No. 6 concludes that the Federal Register whsie a nobce of availability is pubhshed in l the facehty can be operated by the hcensee at power levels the Federal Register for the January and July Agendas.

greater than 5% without endangenng the health and safety of the pubic The Safety Evaluabon and its supplements pertains NUREG-0940 V04 N01: ENFORCEMENT ACTIONS SIGNIFICANT to the apphcaton for a hconse to operate the Wolf Creek Gen. ACTIONS RESOLVED Quarterty Progress Report. January March erating Staten Unit No.1 filed by Kansas Gas and Electre 1985.

• Enforcement Staff. Apnl 1985. 541pp. 8505100072.

Company on February 18, 1980. The Construction Permet No. 3028847L CPPR.147 was issued on May 17,1977 and a low power 5% ms compilaton summartzes shnincant enforcement actons tecense assued on March 11, 1985. The facihty is located in that have been resolved dunnq one quarterty penod (January -

Coffey County, Kansas. March 1985) and includes copies of letters, Notices, and Orders sont by the Nuclear Regulatory Commission to hcensees with NUREG-0847 So6: SAFETY EVALUATION HEPORT RELATED respect to these enforcement actons and the hcensees' re.

TO THE OPERATION OF PERRY NUCLEAR POWER sponses. It is anticipated that the informaton in this pubhcation PLANT, UNITS 1 AND 2 Docket Nos 50-440 And 50-441 (Cleve- will be widely disseminated to managers and employees en-land Electnc illuminating Company)

• Division of Licensing April gaged in activities licensed by the NRC in the interest of pro-1985.150pp. 8505010117, 30114 204. moting pubhc hoa th and safety as wen as common defense Safety Evaluaton Report, NUREG.0887, pertains to the apph- and secunty.

caton Nod by the Cleveland Electnc lHuminating Company on behalf of itself aruf as agent for the Duquestwe Light Company, NUREG-0970: PHOCEDURES FOR MEETING NRC ANTITRUST the Ohio Edison Company, the Pennsytvania Power Company, RESPONS10lLITIES. TOALSTON,A L; MESSIER.M E.;

and the Toledo Edison Company (the Central Area Power Co. LAMBE.W M; et al. Site Analysis Granch. May 1985. 29pp.

ordination Group or CAPCO), as applicants and owners, for a 8506070360. 30798 013.

l l

l

4 i

Main Citations and Abstracts 5 This report desenbes the procedures used by NRC staff to power would be restored before systems that cannot operste implement the antitrust review and enforcement prescnbed in for extended penods without AC power fail, thus resulting in Sections 105 and 186 of the Atomic Energy Act of 1954, as core damage. This report also addresses effects of different de.

J amended (the Act), as covered largely by the Commission's signs, locations, and operatonal features on the estimated fre-Rules and Regulations in 10 CFR Parts 2.101, 2.102, 2.200, quency of core damage resulting from station blackout events.

50.33a. 50.80, and 50.90. These procedures set forth the steps 4 NUREG 1033: FINAL ENVIRONMENTAL STATEMENT RELATED and entena the staff applies in the antitrust review of construc-l TO THE OPERATION OF WPPSS NUCLEAR PROJECT NO.

ton permit and operating keense apphcatons and the amend- 3. Docket No. 50-508.(Washington Pubhc Power Suppfy System)

{ ments to those apphcations that deal with changes in owner-

• Division of Licensing. May 1985. 247pp. 8505310065.

ship. In additen, the procedures describe how the staff en-forces comphance by hcensees when antitrust conditions have  ! Envronmental Statement related to the operation been appended to construction permits and operating hconses. of Washington Nuclear Project No. 3 by Washington Pubic

• NUREG-0975 V03: COMPILATION OF CONTRACT RESEARCH Power Supply System, et al (Docket No. 50-508), located in

FOR THE MATERIALS ENGINEERING BRANCH DIVISION OF Grays Harbor County, Washington, has been prepared by the i ENGINEERING TECHNOLOGY. Annual Report For FY 1984.

• Office of Nuclear Reactor Regulaten of the U S. Nuclear Regu-Division of Engineenng Technology. Apr# 1985. 400pp. latory Commission. This statement reports on the staff's review 8506040260.30711:262. of the impact of operation of the plant. Also included are com-This report presents summanes of the research work per- ments of state and federal governments, local agencies and l formed dunng Fiscal Year 1984 by laboratones and organiza- members of the pubhc on the Draft Environmental Statement i tons under contracts administered by the NRC's Matenals Engs- for this project and staff responses to these comments. The j neenng Branch, Office of Nuclear Regulatory Research. Each NRC staff has concluded, based on a weighing of environmen-

- contractor has wntten a more complete and detailed annual tal, technical and other factors, that an operahng hcense could report of their work which can be obtained by wnting to NRC; be granted.

however, we beheve it is useful to have a summary of each NUREG-1037 DRFT FC: CONTAINMENT PERFORMANCE contractor's efforts for the year combened into one volume. WORKING GROUP REPORT. Draft Report For Comment.

• DivF NUREG-0991 SO4: SAFETY EVALUATION REPORT RELATED son of Engineenng. May 1985. 322pp. 8506140588. 30935.257.

TO THE OPERATION OF LIMERICK GENERATING Containment buildings for power reactors have been studied j STATION. UNITS 1 AND 2 Docket Nos. 50-352 And 50 to estimate their leak rate as a functon of increasing internal 353 (Philadelphia Electnc Company)

• Division of Licensin9 pressure and temperature associated with severe accident se-May 1985 Sipp. 8506060596 30776 256 quences invoMng significant core damage. Potential leak paths In August 1983 the NRC btaff issued its Safety Evaluation through containment penetration assembhes (such as equip-l Report regarding the applicaten for leenses to operate the Lim- ment hatches, airlocks, purge and vent valves, and electncal j enck Generating Station, Units 1 & 2 located on a site in Mont' penetrations) have been identified and their contributions to gomery and Chester Counties, Pennsytvania. Supplement 1 was leak area for the containment are incorporated into containment 4 issued in December 1983 and addressed several outstanding leak rate and pressure temperature response as a function of issues. It also contains the comments made by the Advmory time. Because of lack of reliable expenmental data on the leak-l j Committee on Reactor Safeguards in its intenm report dated age behavior of containment penetratons and isolat on barriers October 18,1983 Supplement 2 was issued in October 1984. at pressure beyond their design conditions, an analytical ap-

) Supplement 3 was issued in October 1984 and addressed proach has been used to estimate the leakage behavior of com-

) issues that required resoluton before issuance of the operating ponents found in specific reference plants that approximately license for Unit 1. On October 26.1984 a heense (NPF.27) for charactertze the various containment types.

Unit 1 was essued which was restncted to a five percent power i level and contained condit ons which required resolution prior to NUREG-1038 S02: SAFETY EVALUATION REPORT RELATED i proceeding beyond the frve percent pcwer level. This Supple- TO THE OPERATION OF SHEARON HARRIS NUCLEAR ment 4 addresses some of those technical issues and their as. POWER PLANT, UNIT 1. Docket No. 50-400, (Carolina Power sociated hcense cond,tions which require resolution pnor to pro. And Light Company And North Carohna Eastern Municipal ceeding beyond the fue percent power level. The remaining Power Agency)

• Division of Licensing June 1985 65pp.

issues will be addressed in a later supplement to this report. 8506270137. 31258 277.

I This Supplement 4 also contains the comments made by the Supplement No. 2 to the Safety Evaluabon Report for the ap.

j Advisory Committee on Reactor Safeguards in its report dated phcation filed by Carohna Power and Light Company and North Novernber 6,1984, regarding full power operation of Limenck Carohna Eastern Municipal Power Agency for a hcense to oper-ate the Shearon Hams Nuclear Power Plant, Unit 1 (Docket No.

! Unit 1.

06, beaW 6n Wah aM Gamarn Coum M Carob NUREG 1032 DRFT FC: EVAltiATION OF STATION BLACKOUT na, s n a ce ear ac ula-ACCIDENTS AT NUCLEAR POWER PLANTS. Technical Find. tion of the U S. Nuclear Regulatory Commission. This supple.

ings Related To Unresolved Safety Issue A-44 Draft Report For ment provides more recent information regarding resolution of Comment. BARANOWSKI.P.W. Office of Nuclear RegulatO'Y some of the open rtems identified in the Safety Evalutaten Research, Director.

• Office of Nuclear Reactor Regulaton, De- Report and in Supplement No.1. It also addresses one of the rector. May 1985. 200pp. 8506250217. 31212:301. recommendations of the Advisory Committee on Reactor Safe-

" Station Blackout / which is the complete loss of alternating guards in its report on the Shearon Harns Plant, dated January current (AC) electncal power in a nuclear power plant, has been 18,1984, which was inadvertently omitted in Supplement No.1.

! designated as Unresolved Safety issue A-44. Because many l safety systems required for reactor core decay heat removal NUREG-1047 S01: SAFETY EVALUATION REPORT RELATED and containment heat removal depend on AC power, the conse. TO THE OPERATION OF NINE MILE PO6NT NUCLEAR l STATION. UNIT NO. 2. Docket No 50-410. (Niagara Mohawk

! quences of a staton blackout could be severe. TNs report doc-uments the findings of technical studies performed as part of Power Corporation)

• Division of Ucensng. June 1985. 35pp.

{

I the program to resofve this issue. The important factors ana- 850705041t 31372:268.

l lyzed include: the frequency of loss of offsite power; the proba. This report supplements the Safety Evaluation Report bahty that emergency or onsite AC power supphes would be un. (NUREG-1047, February 1985) for the application f4ed by Niag-

. available; the capabihty and reliabihty of decay heat removal ara Mohawk Power Corporaton, as apphcant and co owner, for I systems independent of AC power; and the likkhood that offsite a 16 cense to operate the Nine Mile Point Nuclear Station Unit 2 I

i

\ - - - - - . - --- --.-_-- -- .-

i 6 Main Citations and Abstracts 4

(Docket No. 50 410). It has been prepared by the Office of Nu- format, the brense appict.nt will minimize aJministrative prob-clear Reactor Regulahon of the U S. Nuclear Regulatory Com- lems associaid with the submittal, review and approval of the rrussion. The facdsty is located near Oswego, New York Subl ect FNMC plan. Preparation of the plan in accordance with this to favorable resolution of the items discussed in this report, the format wdl assist the NRC m evaluating the plan and in stand-NRC Staff conclextes that the facihty can be operated by the ap- ardizing the review and Icensing process. However, conform-plicant without erWa.1genng the health and safety of the pubic. ance with this guidance is not required by the NRC, A licensa NUREG 1061 V02: PEPORT OF THE U.S. NUCLEAR REGULA- appleant who employs a format that provides an equal level of TORY COMMISSION PIPING REVIEW COMMITTEE. Volume completeness and detad may use their own format.

2 Evaluaton Of Seismc Designs A Review Of Seisme Desgn Requirements For Nuclear Power Plant Piping

• Pipng Review NUREG-1085: FINAL ENVIRONMENTAL STATEMENT RELATED Committee. Apnl 1985. 213pp. 8505090016. 30254.043. TO THE OPERATICN OF NINE MILE POINT NUCLEAR 1,

This docurnent reports the position and recommendations of STATION, UNIT NO. 2. Docket No. 50-410.(Niagara Mohawk the NRC Pipng Review Comtruttee. Task Group on Seismic Power Corporation et al)

• Division of Licensing. May 1985.

373pp. 8505230685. 30548.00 t.

Design. The Task Group consioered overlappng conservatism in the vanous steps of seismic design, the effects of using two This Fina1 Env ronmental Statement conta ns the assessment levels of earthquake as a design enterion, and current industry of the environmental impact associated with the operaten of practicos. Issues such as dampng vatues, spectra modifcaton, the Nine tJde Point Nuclear Staten, Unit 2, pursuant to the Na.

multiple response spectra methods, nozzle and support design, tonal Environmental Pohey Act of 1969 (NEPA) and Title 10 of design margins, inelaste piping response, and the use of snub- the Code of Federal Regulatons, Part 51, as amended, of the f bers are addressed. Effects of current regulatory requirements Nuclear Regulatory Commission regulatons. This statement ex-for piping design are evaluated, and recommendations for im- amines the environment, environmental consequences and miti-mediate icensing action, changes in existing requirements, and g g research programs are presented. Additional background infor- costs.

maton and suggestens given by consultants are also present-ed NUREG-1095: EVALUATION OF RESPONSES TO IE BULLETIN

, 82-02 Degradaten Of Threaded Fasteners in Reactor Coolant j NUREG 1061 VOS: REPORT OF THE U S. NUCLEAR REGULA. Pressure Boundary Of Pressunzod Water. Reactor Plants

TORY COMMISSION PIPING REVIEW COMMITTEE. Volume ANDERSON.W.; STERNER,P. Orvision of Emergency Prepared-t

5. Summary - Piping Review Committee Conclusons and Rec- ness & Engineermg Resporve (Post 830103). May 1985. 75pp.

ommendations.

• Pipng Review Committee Apnl 1985. 55pp. 8506240221. IEB-82 02. 31177.219.

8505070580. 30209 240. IE Bulletin 82-02 was issued by t'.e NRC on June 2,1982 to This document summanzes a comprehensive review of NRC notify hcensees about incidents of severe degradaten of thread-requirements for Nuclear Pipng by the U S. NRC Piping Review ed fasteners. Responses to the Bulleto from 41 PWR hconsees Committee. Four topical areas, addressed m greater detad in included data from recent regular mspectons of reactor coolant Volumes 1 through 4 of this report, are included. (1) Stress Cor- pressure boundary components connectons of six-inch size and

rosion Cracking m Pipng of Boding Water Reactor Plants, (2) larger. Statisteal analysis is used to determine ssgnifcant fac-Evaluation of Seismic Design. (3) Evaluaten of Potential for tors related to frequency of leakage incidents in connactions, i Pipe Breaks. and (4) Evaluaton of Other Dynamic Loads and occurrence of degradation of botts and studs, and the need for

Load Combinations This volume summanzes the major issues, bolt replacement. Factors examined include the age of the

reviews the interfaces, and presents the Committee's conclu- plant, types of components, use of lubncants and sealants, and sions and recomtnendations far updating NRC requirments on differences between plants. The compded data indicate that, on these issues. This report also suggests research or other work the average,10% of the bolted connections which were irF that may be required to respond to issues not amendable to spected show evidence of leakmg and 80% of those undergo resolut on at this time. some degradation of the bolting A segrwfcant decrease in the NUREG-1061 VO2 ADO

REPORT OF THE U S. NUCLEAR REG- occunwice of bottog degradation events as the age of the i ULATORY COMMISSION PIPING REVIEW plant increases is observed. Valves appear to be less sub lect to

, COMMITTEE. Volume 2 Addendum: Summary And Evaluaton Of bottmg conosion. A group of 5 of the 41 plants accounted tw Historical Strong-Moton Earthquake Seismic Response And about one-hatt of the repurted leakage and corrosen events Damage To Aboveground Industnal Pipng.

• Piping Review The common charactenstic found for 4 of these 5 plants was i

Comtnittee. ' Stevenson & Associates. Aprd 1985. 211pp the lubricant used The use of nickel-graphite based futmcants 8505100057. 30268 005. appears to offer a significantly reduced incidence of leakage

}'

Earthquake expenence data for industrial pping has been and corrosion; while use of molybdenum disulfide-based lubre summanzed in this report. Conclusions and recommendations cants and graphite-based lubncants appears to result in a sig-for improveg the design of nuclear plant ppng are made by the nificantly mcreased mcidence of leakage and conosh j author, input from R. L. Cloud, P, Yaner, and H. Shibata has been included. The matenal in this report served as I ackgrour.d NUREG-1116: A REVIEW OF THE CURAFNT UNDERSTANDING l information for the NRC Pipng Review Committee Seismic OF THE POTENTIAL FOR CONTAINMENT FAILURE FROM Design Task Group (and their consultants) in the development IN VESSEL STEAM EXPLOSIONS.

• Steam Explosion Review Group. June 1985. 400PP, 8507030718. 31335 001, of the positons grven in NUREG 1061 Volume 2.

A group of experts was convened to review the current urk NUREG 1065 R01: ACCEPTANCE CRITERIA FOR THE LOW EN. derstanding of thb potential for containment fadure from irk RICHED URANIUM REFORM AMENDMENTS. EMEIGH,C.W.; vessel steam explosions dunng core meltdown accidents in GUNDERSEN.G E.; WITHEE.C.J. Division of Safeguards. Aprd LWRs. The Steam Explosen Review Group (SERG) was re.

I 1985. 53pp. 8504240693. 29988.183. quested to provide assessments of. (i) the conditional probabdi-

+

This report documents a standard format suggestad by the ty of containment fadure due to a steam explosion, (n) a Sandia 1 NRC for use o prepanng fundamental nuclear matenal control Natonal Laboratory (SNL) report entitled "An Uncertainty Study 4 plans as required by the Low Enrched Uransurn Reforrn Arnend- of PWR Steam Explosons," NUREG/CR 3369, (m) a SNL pro.

j rnents (portions of 10 CFR Part 74). The report also desenbes posed steam explosion research program. This report summa-the necessary contents of a comprehensive plan and provides rizes the results of the dehberations of the review group 11 elso example acceptance entena which are intended to commurw. presents the detaded response of each indevedual member to cata acceptable means of achieving the performance capabdb each of the issues. The consensus of the SERG is that the oc-i ties of the Reform Amendments. By using the suggested currence of a steam explosen of sufficient energetics wtuch

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

Main Citations and Abstracts 7 could lead to alpha-mode containment failure has a low proba- NUREG-1125 V05: A COMPILATION OF REPORTS OF THE AD-bihty. The SERG members disagreed with the methodology VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957-used in NUREG/CR-3369 for the purpose of establishing the 1984.Voluma 5,Part 2:ACRS Reports On Generic Subjects uncertainty in the probability of containment failure by a steam (HTGR - Regulatory Guides).

• ACRS Advisory Committee on explosion. A consensus was reached among SERG members Reactor Safeguards. Apnl 1985. 630pp. 8504220396.

on the need for a continuing steam explosion research program 29958:102.

which would improve our understanding of certain aspects of See NUREG-1125,V01 abstract.

steam explosion phenomenology.

NUREG-1118: ENVIRONMENTAL ASSESSMENT FOR RENEW- VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957-AL OF SPECIAL NUCLEAR MATERIAL LICENSE NO.SNM- 1984. Volume 6,Part 2:ACRS Reports On Genenc Sub lects (RPA 1107. Docket No. 70-1151. (Westinghouse Electnc Corporaton) - Appendix C).

• ACRS Advisory Committee on Reactor Safe-

• Divison of Fuel Cycle & Matenal Safety. May 1985. 140pp. guards. Apnl 1985. 567pp. 8504220402. 29960 015.

8505240050, 30566:199. See NUREG 1125,V01 abstract.

This Environmental Assessment is issued by the U.S. Nuclear Regulatory Commisson (NRC) in response to an apphcaton by NUREG 1127: RADIATION PROTECTION TRAINING AT URANI-the Westinghouse Electnc Corporaton for the renewal of Spe- UM HEXAFLUORIDE AND FUEL FABRICATION PLANTS.

cial Nuclear Matenal License No. SNM 1107 which covers the BRODSKY,A.; SOONG,A L; BELLJ. Drvison of Radiaton Pro-operations of the Columbia plant. grams & Earth Sciences (post 840429). May 1985. 33pp.

8506190048. 31017:173.

NUREG 1119: SAFETY EVALUATION REPORT RELATED TO This report provides general informaton and references THE RENEWAL OF THE OPERATING LICENSE FOR THE useful for establishing or operating radiation safety training pro.

CAVALIER TRAINING REACTOR AT THE UNIVERSITY OF grams in plants that manufacturo nuclear fuels or process urani-VIRGINTA. Docket No. 50-396.(University Of Virginia)

• Divison um compounds that are used in the manufacture of nuclear of Licensing. May 1985. 62pp. 8506060716. 30780:214. fuels. In addition to a bnet summary of the pnnciples of effec.

This Safety Evaluaton Report for the applicaten filed by the tive management of radiaten safety training, the report also University of Virginia for a renewal of operabng hcense number contains an appendix that provides a comprehensive checklist R-123 to continue to operate a tra?ung and research reactor of scientific, safety, and management topics, from which appro-(CAVAUER) has been prepared by the Office of Nuclear Reac- pnate topics may be selected in prepanng training outhnes for tor Regulaton of the U.S. Nuclest Regulatory Commission. The vanous job categones or tasks pertaining to the uranium nuclear facility is owned and operated by the Urwversity of %rginia and fuels industry. The report is designed for use by radiahon safety is located on the campus in Charlottesville, Virginia. Based on training professonals who have the expenence to utilize the its techncal review, the staff concludes that the reactor facility report to not only select the appropnate topes, but also to tailor can continue to be operated by the university without endanger

• the specific details and depth of coverage of each training ses-ing the health and safety of the pubic or endangenng the envi- sion to match both employee and management needs of a par-ronment. tcular industnal operabon.

NUREG 1125 V01: A COMPILATION OF REPORTS OF THE AD-VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957 NUREG-1128. TRIAL EVALUATIC?iS IN COMPARISON WITH THE 1983 SAFETY GOALS. RIGGS.R.; SEGE.G. Division of 1984. Volume 1,Part 1 ACRS Reports On Project Reviews (A-F).

Safety Technology. June 1985. 200pp. 8507080209. 31402.041.

ACRS - Advisory Committee on Reactor Safeguards. April 1985. 65Bpp. 8504220393. 29956:167.

This report provides retrospective compansons of selected This six-volume compilaten contains over 1000 reports pre- generic regulatory actons to the 1983 NRC safety goals, whch pared by the Advisory Committee on Reactor Safeguards from had been issued for evaluaten dunng a two-year penod. The September 1957 through December 1984. The reports are di- issues covered are those analyzed by the Office of Nuclear Re-actor Regulaton (NRR) (assisted in some cases by the Battelle vided into two groups: Part 1: ACRS Reports on Project Re-views, and Part 2: ACRS Reports on Generic Subjects. Part 1 Pacif'c Northwest Laboratory). The issues include auxiliary feed-contains ACRS reports alphabetized by project name and within water reliability, pressurized thermal shock, power-operated project name by chronological order. Part 2 categonzes the re- relief valve isolaton, asymmetnc blowdown loads on PWR pri-mary systems, pool dyname loads for BWR containments, and ports by the most appropnate genenc subject area and within steam generator tube rupture. Calculated core-melt frequencies, subject area by chronological order.

mortality risks, and cost-benefit ratos are compared with the NUREG 1125 V02: A COMPILATION OF REPORTS OF THE AD. corresponding safety-gnal quanhtatrve design objectives. Con-VISORY COMMITTEE ON REACTOR SAFEGUARDS.1957, s.derations that should inflJence interpretaten of the compari-1984. Volume 2.Part 1 ACRS Reports On Project Reviews (G-P). sons are discussed. Comments are included on whether and

• ACRS - Advisory Committee on Reactor Safeguards. Apnl how the safety goals may help in the regulatory decison proc-1985. 720pp. 8504220344. 29944:001. ess and on problems encountered.

See NUREG 1125.V01 abstract.

NUREG-1131: FINANCIAL ANALYSIS OF POTENTIAL RETF.O-NUREG-1125 V03: A COMPILATION OF REPORTS OF THE AD. SPECTIVE PREMlUM ASSESSMENTS UNDER THE PRICE-VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957 ANDERSON SYSTEM. WOOD.R.S. Office of State Programs, 1984. Volume 3,Part 1 ACRS Reports On Projo;t Reviews (0-Z). Director. Apnl 1985.17pp. 8505080348. 30218.171.

• ACRS - Advisory Committee on Reactor Safeguards. Apnl Ten representative nuclear utshties have been analyzed over 1985. 563pp. 8504220389. 29954:329. the period 1981 1983 to evaluate the effects of three levels of See NUREG 1125,V01 abstract. retrospective premiums on varcus financial indcators. This analysis continues and expands on earher analyses prepared as NUREG-1125 V04: A COMPILATION OF REPORTS OF THE AD- background for deliberabons by the U S. Congress for possible VISORY COMMITTEE ON REACTOR SAFEGUARDS,1957- extension or modifcabon of the Pnce-Anderson Act.

1984. Volume 4,Part 2:ACRS Reports On Genenc Subjects (Ac-cident Analysis - Generic items).

• ACRS Advisory Committee NUREG-1132: TECHNICAL SPECIFICATIONS FOR DIABLO on Reactor Safeguards. Apnl 1985. 627pp. 8504220406. CANYON NUCLEAR POWER PLANT, UNIT NO. 2. Docket No.

29961:225. 50-323.(Pacife Gas and Electric Company)

• Division of Licens-See NUREG 1125,V01 abstract. ing. Apnl 1985. 466pp. 8505280011. 30605:007.

8 Main Citations and Abstracts The Diablo 2 Techncal Specifcations were prepared by the as apphcants and owners, for heenses to operate the Vogtle U.S. Nuclear Regulatory Commisson to set forth the hmats, op. Electnc Generating Plant, Units 1 and 2 (Docket Nos. 50-424 erating conditions, and other requirements apphcable to a nu- and 50-425), has been prepared by the 0: fee of Nuclear Reac-clear reactor facahty as set forth in Secten 50.36 of 10 CFR tor Regulabon of the U S. Nuclear Regulatory Commission. The Part 50 for the protection of the health and safety of the pubic. facility is located in Burke County, Georgia, approximately 41.5 NUREG-1133: TECHNICAL SPECIFICATIONS FOR PALO VERDE km (26 mi) south-southeast of Augusta, and on the Savannah NUCLEAR GENERATING STATION, UNIT 1. Docket No. 50-528. River. Subject to favorable resolution of the items discussed in (Arizona Public Sennce Company)

• Division of Licensing. May s re@rt, N stan conchs mat N a$ cant can pam 1985. 515pp. 8506240646. 31151:001. N fadity wecut Wangenng N Nam aN sa% of N The Palo Verde Unit 1 Techncal Specifcatens were pre- public.

pared by the U.S. Nuclear Regulatory Commission to set forth NUREG-1140 DRFT FC: A REGULATORY ANALYSIS ON EMER.

the hmits, operating conditons, and other requirements apphca- GENCY PREPAREDNESS FOR FUEL CYCLE AND OTHER ble to a nuclear reactor facility as set forth in Secten 50.36 of RADIOACTIVE MATERIAL LICENSEES. Draft Report For Com-10 CFR Part 50 for the protecten of the health and safety of ment. MCGUIRE S.A. Div'sion of Risk Analysis & Operations the public. (post 840429). June 1985.125pp. 8507020410. 31309.033.

NUREG-1134: RADIATION PROTECTION TRAINING FOR PER- Potential accidents for 15 types of fuel cycle and other rado-SONNEL EMPLOYED IN MEDICAL FACILITIES. achve matenal heensees were analyM The most potenbany MCELROY,N.L.; BRODSKY,A. Onnsion of Radiation Programs & hazardous accident, by a large margin, was determined to be Earth Sciences (post 840429). May 1985. 61pp. 8506130363. the sudden rupture of a heated multi-ton cyhnder of UF6. Acute 30868:116 fatabbes onsite are probaW rot credible. Amte pennanent inju-This report proudes informaton useful for planning and con- Ms may be possible for many hundreds of meters, and chnically ducting radiabon safety training in medical facilities to keep ex- observable transient effects of unknown long term conse-posures as low as reasonably achievable, and to meet other quences may be possible for distances up to a few miles.

regulatory, safety and loss preventen regmrements in today's These effects would be caused by the chemmal torcity of the hospitals. A bnef discussen of the elements and basc consid- UF6. Radiation doses would not be signifcant. The most poten-erations of radiation safety training programs is followed by a tially hazardous accident due to radiation exposure was deter-short bibliography of selected references and sample lecture (or mined to be a large fire at certain facihbes handling large quan.

sesson) outlines for vanous job categones. This informaton is tities of alpha-emitting radionuclides (ie., Po 210, Pu-238, Pu=

intended for use by a professicnal who is thoroughly acquainted 239, Am-241, Cm-242, Cm-244) or radioiodines (1-125 and l-with the science and practice of radiation protecten as well as 131). However, acute fatalities or injunes to people offsite due the specife procedures and circumstances of the particular hos. to accidental releases of these materials do not seem plausible.

pital's operations. Topics can be added or subtracted, ampi.fied The only other signifcant accident was identfied as a long-term or condensed as appropriate. This document does not set forth Pulsahng cnbcahty at fuel cycle facihties handhng high-ennched specifc training program requirements for any partcular hospital uranium or plutonium. An important feature of the most senous or type of medcalinsttution or group of employees. accidents is that releases are hkely to start without pnor warn-ing. The releases would usually end within about half an hour.

.n..... n,.. 6.. m n..v.. mrvn. RELATED TO inus protection actons wouta nave to De taken quckly to be THE CONSTRUCTION PERMIT AND OPERATING LICENSE effective. There is r*ot hkely to be enough time for dose projec-FOR THE RESEARCH REACTOR AT THE UNIVERSITY OF tions, complicated decisonmaking dunng the accident, or the TEXAS. Docket No. 50-602. (University of Texas)

• Divison of partcipabon of personnel not in the immediate vicinity of the Licenssng. May 1985. 88pp. 8506240665. 31152:223.

site. The appropnate response by the facility is to immediately This Safety Evaluaton Report for the applicaton filed by the notify local fire, pohce, and other emergency personnel and give University of Texas for a constructen permit and operating li- them a bnef predetermined message recommending protective cense to construct and operate a TRIGA research reactor has actions. Emergency personnel are generally well qualified to re-been prepared by the Office of Nuclear Reactor Regulaton of spond effectively to small accidents of these types.

the U.S. Nuclear Regulatory Commission. The facility is owned and operated by the University of Texas and is located at the NUREG-1145 V01: U.S. NUCLEAR REGULATORY COMMISSION Univensty's Balcones Research Center, about 7 miles (11.6 kilo. 1984 ANNUAL REPORT.

• Office of Resource Management, Di-meters) north of the main campus in Austin, Texas. The staff rector. June 1985. 234pp. 8506260386 31246.057.

concludes that the TRIGA reactor facility can be constructed This report covers the major activities, events, decisons and and operated by the University of Texas without endangenng planning that took place dunng fiscal year 1984 within the NRC the health and safety of the pubhc. or involving the NRC.

NUREG 1136: TECHNICAL SPECIFICATIONS FOR WOLF NUREG-1147: SEISMIC SAFETY RESEARCH PROGRAM PLAN.

CREEK GENERATING STATION, UNIT 1. Docket No. 50

• Division of Engineering Technology. June 1985. 230pp.

482.(Kansas Gas And Electric Company)

• Division of Licens- 8507080215.31392:150.

ing. June 1985. 498pp. 8506270251. 31257:142. This plan desenbes the safety issues, regulatory needs, and The Wolf Creek Generating Staten Unit No.1 Technical the research necessary to address these needs. The plan also Specifcatons were prepared by the U.S. Nuclear Regulatory discusses the relatonship between current and proposed re-Commisson to set forth the limits, operating conditions, and Jarch within the NRC and research sponsored by other gov-other requirements appicable to a nuclear reactor facahty as set ernment agencies, universities, industry groups, professional so-forth in Section 50.36 of 10 CFR Part 50 for the protecten of cieties, and foreign sources.

the health and safety of the public*

NUREG/CP-0059 V01: PROCEEDINGS OF THE MITI-NRC SEIS-NUREG-1137: SAFETY EVALUATION REPORT RELATED TO MIC INFORMATION EXCHANGE MEETING VOLUME L THE OPERATION OF VOGTLE ELECTRIC GENERATING WEISS.A.J. Brookhaven Natonal Laboratory. Apnl 1985.423pp.

PLANT, UNITS 1 AND 2. Docket Nos. 50424 And 50-425.(Geor- 8506070372. BNL-NUREG-51821. 30796:001.

gia Power Company,et al)

• Division of Licensing. June 1985. The first Japan Ministry of Internatonal Trade and Industry 650pp. 8507030707. 31314:265. (MITI) - U.S. Nuclear Regulatory Commisson (NRC) Soismic In-The Safety Evaluaton Report for the apphcation filed by formabon Exchange Meeting (SIEM) was held Jufy 18-20, 1984 Georgia Power Company, Municipal Electnc Authonty of Geor- in Palo Alto, Cahfornia. The purpose of SIEM was to provide gia, Oglethorpe Power Corporaton, and City of Dalton, Georgia, technical informaton on seismic research being conducted

j i

I 4

Main Citations and Abstracts 9 under MITI and NRC sponsorships to the parteipants. The aim This monthly report contains Licensee Event Report (LER)

I was to improve understanding of the seismec research in operatonal information that was processed into the LER data file of the Nuclear Safety Informaton Center (NSIC) dunng the progress in Japan and the United States for posssble identifica-ton of areas of mutual interest which could be the basis for one month penod identified on the cover of the document. The l LERs, from whch this informaton is denved, are submitted to future cooperaton. Approximately 40 Japanese and U.S. techrw-cal speciahsts in seismc research partcipated in the meeting. the Nuclear Regulatory Commesson (NRC) by nuclear power

These proceedings represent the compilaton of the papers pre- plant Icensees en accordance with federal regulatons. Proce-

sented at the meeting. dures for LER reporting for revisons to those events occurnng pnor to 1984 are desenbed in NRC Regulatory Guide 1.16 and NUREG/CP-0062

PROCEEDINGS OF THE CONFERENCE ON NUREG-0161, Instructons for Preparation of Data Entry Sheets

! THE APPLICATION OF GEOCHEMICAL MODELS TO HIGH-for Licensee Event Reports. For those events occumng on and LEVEL NUCLEAR WASTE REPOSITORY ASSESSMENT. after January 1,1984, LERs are being submitted in accordance JACOBS,G.K.; WHATLEY,S.K. Oak Ridge Nabonal Laboratory.

May 1985.130pp. 8506130505. ORNL/TM-9585. 30892

227. with the revised rule contained in Title 10 Part 50.73 of the

, A conference on the application of geochemcal models in Code of Federal Regulatons (10 CFR 50.73 - Lcensee Event the assessment of high-level nuclear waste repositones was Report System) whch was published in the Federal Register held to discuss the current status of geochemcal code develop.

(Vol. 48, No.144) on July 26,1983. NUREG-1022, Licensee f Event Report System Desenpton of Systems and Guidelines ment, thermodyname data bases, reaction kinetes, and cou. .

pled-process models as applied to site characterizaten and per. for Reporting. provides supporting guidance and informaton on formance assessment actnnties. These proceedings include ex, the revised LER rule. The LER summanes in this report are ar-tended abstracts of the techncal presentations given at the ranged alphabetcally by facihty name and then chronologically I

conference, a discussion of the role of g 0Gerd.e modehng in by event date for each facihty. Component, system, keyword, predicting the performance of repositones, and a set of recom. and component vendor indexes follow the summanes. Vendors mendatons that identify the key developments needed in order are those identified by the utthty when the LER form is initiated; .

for geochemical models to become more apphcable for quants. the keywords for the component, system, and general keyword [

j tatrve evaluations of repositones. Detailed recommendations rei. icidexes are assigned by the computer using correlaton tables evant to the following subjects are discussed (1) improved sim. fr xn the Sequence Coding and Search System.

i '

utaten of repository performance through inclusion of additonal

{; NUREG/CR-2000 V04 N4: LICENSEE EVENT REPORT (LER) important geochemcal processes and parameters into current

. COMPILATION.For Month Of Apnl 1985.

• Oak Ridge Natonal geochemical modelg2) more careful attenten to uncertainties Laboratory. May 1985. 87pp. 8506130364. ORNL/NSIC-200.

associated with youww model calculatons, (3) assgning

30867.262

pnonties to (through sensstrvity studes and entcal evaluatons) See NUREG/CR-2000,V04,NO3 abstract.

, snd then improving and/or obtanng important thermodynamic

, data, and (4) addressing the importance of kinetcs in simulating NUREG/CR-2000 V04 N5: LICENSEE EVENT REPORT (LER) repository behavior. COMPILATION For Month Of May 1985.

• Oak Ridge Natonal NUREG/CP-0065: TRANSACTIONS OF THE 8TH INTERNATION. Laboratory. June 1985.111pp. 8507030669. ORNL/NSIC-200.

AL CONFERENCE ON STRUCTURE MECHANICS IN REAC. 31314g 55. _ _ _ _ , . ,

see nvNewwwuu,vue,rvua nostract i TOR TECHNOLOGY. Panel Sesson J-K: Status of Research in

! Structural And Mechancal Engineenng For Nuclear Power NUREG/CR-2331 V04 N3: EAFETY RESEARCH PROGRAMS Plants. BROWZIN.B.S. Drvision of Ergneer'ng Technology. SPONSORED BY OFFICE OF NUCLEAR REGULATORY

June 1985. 266pp. 8507080187. 31393 277. RESEARCH.Ouarterty Progress Report Jufy 1 -September j These transactions of the J-K/ panel sesson include prepnnts 30,1984. WEISSAJ. Brookhaven Natonal Laboratory. May 1 of papers or abstracts which are listed in Volume A, "Introduc- 1985.117pp. 8506060147. BNL-NUREG-51454. 30781

002.

i tion, General Contents, Authors' Index." Proceed.ngs of the 8th This progress report will descnbe current actrvites and techni-j international Conference on Structural Mechancs in Reactor cal progress in the programs at Brookhaven Natonal Laboratory Technology. These papers represent the body of the J-K/ panel sponsored by the Division of Accident Evaluaton, Dtvision of

session, " Status of Research in Structural and Mechanical Engi- Engineenng Technology, and Division of Risk Analysis & Oper-

! neenng for Nuclear Power Plants," sponsored by the U.S. Nu- atons of the U.S. Nuclear Regulatory Commisson Office of Nu-

) clear Regulatory Commisson. clear Regulatory Research. The projects reported are the fol-I NUREG/CR-1755 AD001: TECH 140 LOGY, SAFETY AND COSTS lowing: Hgh Temperature Reactor Research, SSC Develop-i OF DECOMMISSIONING NUCLEAR REACTORS AT MULTI. ment, Vahdaten and Applicaton, Genenc Balance of Plant Mod-PLE-REACTOR STATIONS. Effects On Decommissoning Of In. eling, Thermal-Hydraute Reactor Safety Expenments Develop-l tenrn inabihty To Dispose Of Wastes Offsite. MOORE,E.B. Bat- ment of Plant Analayrer Code Assessment and Apphcation

! telle Memonal institute, Pacifc Northwest Laboratones. April (Transient and LOCA Analyses). Thermal Reactor Code Devel-

! 1985. 41pp. 8505070571. 30209 200. opment (RAMONA-30), Calculatonal Quakty Assurance in Sup-The purpose of this addendum is to examine the impacts of port of PTS; Stress Corrosion Cracking of PWR Steam Genera-en interim inabihty to carryout offsite disposal of radioactive tor Tubing, Probabihty Based Load Comtunatens for Design of wastes and spent fuel on the decommissoning of multiple-reac- Category 1 Structures, Identificaton of Age-Related Failure for power staten. The example selected for study is a four.PWR Modes; Analysis of Human Error Data for Nuclear Power Plant

. station in whch each PWR is prepared for safe storage at two- Safety Related Events, Human Factors Aspects of Safety / Safe-

] ytar intervals, held in safe storage for 100 year intervals. BWRs guards Interactons Emergency Acton Levels, and Protective cre neglected for simplicity and in the expectaten that the re- Acton Decision Making.

suits would be semitar to those for PWRs. Only SAFSTOR is considered because DECON and ENTOMB are unsuitable by NUREG/CR-2331 V04 N4: SAFETY RESEARCH PROGRAMS SPONSORED BY OFFICE OF NUCLEAR REGULATORY definition for intenm storage of radioactive wastes and/or spent RESEARCH.Ouarterfy Progress Report, October 1 - December fuel it is assumed that all radoact:ve wastes and spent fuel are 31, 1984. WEISSAJ. Brookhaven Natonal Laboratory. May sh9 ped offsite by the end of decommissiorg 1985.139pp.8507050378. BNL-NUREG-51454. 31373:001.

NUREG/CR-2000 V04 N3: LICENSEE EVENT REPORT (LER) This progress report will desenbe current activities and techni-COMPILATION.For Month Of March 1985.

• Oak Ridge Naton- cal progress in the programs at Brookhaven National Laboratory j si Laboratory. Apnl 1985.78pp.8505070557. ORNL/NSIC 200. sponsored by the Division of Accident Evaluaton, Divison of 30210:092. Engineering Technology, and Division of Risk Analysis & Oper-l

-,-3r y .- ,93, , -.,,--.,,vvys--,-,.--9. ,--~ q ,w.,- e,--9 y-ycy.-.,, ,y-, - - -

w--y-3,-y.y.,-p. . , - - 9 ,y -y. .ww-- ---- , ,=r7my . - y r-,.

10 Main Citations and Abstracts

, ations of the U.S. Nuclear Regulatory Commission Office of Nu- and subcooling, melt temperature, and melt compositon. Also, a

clear Regulatory Research. The projects reported are the fol- a few scoping experiments were performed to explore the ef-  :

lowing: High Temperature Reactor Research, SSC/MINET De- fects of changing the nature of the coolant, and the viscosity of I velopment, Validaten and Applicabn, Thermal-Hydraulic Reac- the melt. As each of the four major parameters was vaned, tor Safety Experiments, Plant Analyzer, Code Assessment and thresholds could be located beyond which explosions were sup-Application, Code Maintenance (RAMONA-38), Calculational pressed. However, in general, the explosions could be reinstiat-Quality Assurance in Support of PTS, Stress Corrosen Cracking ed by increasing the magnitude of the tnggenng pulse. The ef-of PWR Steam Generator Tubing Probability Based Load Com- fects of increasing the ambient pressure up to 1.12 MPa were binations for Design of Category i Structures, SoiLStructure faster and finer melt fragmentation, and faster and more com-Interaction Evaluations, Identificaton of Age-Related Failure plete transfer of heat from melt to water, Moreover, tnggering j Modes; Application of HRA/PRA Results to Resolve Human became easier over the range of ambient pressure between l

)

Reliability and Human Factors Safety issues PRA Technology about 0.15 MPa and approximately 0.7 MPa. j l Transfer Program, Emergency Acton Levels, and Protective 1 Action Decisionmaking. NUREG/CR 2951: THE D9 EXPERIMENT. Heat Removal From NUREG/CR-2531 R03: INTRODUCTORY USER'S MANUAL FOR Stratified UO2 Debris. OTTINGER.C.A.; MITCHELL,G.W.;

.I S THE U.S. NUCLEAR REGULATORY COMMISSION REACTOR

, SAFETY RESEARCH DATA BANK. HARDY,H A.; LAATS.E.T. 8 SAN 4 838. 31 1 EG8G, Inc. May 1985. 95pp. 8505170007. EGG-2164. The 09 experiment investigated the coolabstity of a shallow (77 mm), strat feed urania bed in sodium. The bed was fission i Th Uni ed States Nuclear Regulatory Commission (USNRC) heated in the Annular Core Research Reactor (ACRR) at h.n cablished the NRC/ Division of Accident Evaluation (DAE) Sandia National Laboratones to simulate the effects of radioac-l Data Band Program to collect, store, and make available data tive decay heating It was the first stratified debns bed experi- ]

i from the many domestic and foreign water reactor safety re- ment to use an extended UO2 particle size distnbution (0.038 to i search programs. Local directen of the program is provided by 4.0 mm). Dryout occurred at powers ranging from 0.10 to 0.58 EG&G Idahoinc., pnme contractor for the Department of W/g, which was close to the incipient boiling power and before Energy (DOE) at the Idaho National Engineenng Laboratory channels penetrated the subcooled zone in the bed, even with

, (INEL). The NRC/DAE Data Bank Program provides a central subcoolings as low as 80 degrees centigrade. Channel penetra- I j computer storage mechanism ano access software for data that tion was observed after dryout began, but the bed became only is to be used by code development and assessment groups in moderately more coolable. All these observations agree with

. meeting the code correlation needs of the nuclear industry. The current models.

j administrative portion of the program provides data entry, h NUREG/CR-3005:

SUMMARY

OF THE NUCLEAR REGULATORY C. T C/DAE Da nk a the capabi ties f the COMMISSION'S LOFT PROGRAM RESEARCH FINDINGS.

data access software are desenbed in this document. NALEZNY,C.L EG8G, Inc. Apnl 1985. 212pp. 8507050424.

EGG-2231. 31372.001.

NUREG/CR-2663 V01: INFORMATION NEEDS FOR CHARAC- This document is a summary of the ma n research results of TERIZATION OF HIGH-LEVEL WASTE REPOSITORY SITES IN the Loss-of-Fluid Test (LOFT) Procram relative to code assess-

] S;X GECLOGiG MEDIA. Main Report.

• Ertec vestern, v Inc. ment, code development, licensing, rulemaking. Safety technolo-May 1985. 574po. 8506270457. 31261.001. gy, and reactor operations. The LOFT facility is a 50 MW(t)

Evaluaten of the geologic isolaton of radioactive materials pressunzed water reactor (PWR) system with instruments that

, from the biosphere requires an antimate knowledge of site geo- measure and provide data on the system thermal. hydraulic and logic conditons, which is gained through precharacterizaten nuclear conditions. The transient response of the LOFT system and site characterization studies. This report presents the re- to accident events is similar to large [1000 MW(e)) commercial

suits of an intensive literature review, analysis and compilaten PWRs. The main objectives of the LOFT Expenmental Program to delineate the informaton needs, applicable techniques and were to qualify the engineered safety systems used in commer-evaluaten entena for programs to adequately charactenze a site cial PWRs and to venfy the computer codes used in safety in six geologic media. These media, in order of presentation. analyses. The LOFT Program contnbuted to the improvement of

! are: granite, shale, basalt, tuff, bedded salt, and domed salt. computer codes used to predict the response of commercial Guidelines are presented to assess the effcacy (application, ef- PWRs, demonstrated the adequacy of engineered safety sys-fectiveness, and resolution) of currently used to exploratory and tems, and contnbuted to improved understanding of PWR acci-3 testing techniques for precharacterizaten of charactenzation of dent phenomena, particularty those associated with the evalua-t a site. These guidelines include the reliability, accuracy, and res- tion model in Appendix K to 10 CFR 50 (the "ECCS rule").

olution of techniques deemed acceptable, as well as cost este-mate of vanous field and laboratory techniques used to obtain NUREG/CR-3091 V04: REVIEW OF WASTE PACKAGE VERIFI-the necessary information. Guidelines presented do not assess CATION TESTS. Semiannual Report Covenng The Period Octo-the relative suitability of media. This report consists of two vol- ber 1983 - March 1984. JAIN.H.; VEAKIS,E.; SOO,P. Brookha-umes: main report and appendices. ven National Laboratory. June 1985. 29pp. 8507050398. BNL-NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARAC- NUREG-51630. 31373:314.

TERIZATION OF HIGH-LEVEL WASTE REPOSITORY SITES IN The current study is part of an ongoing task to specify tests SIX GEOLOGIC MEDIA. Appendices.

• Ertec Western, Inc . May that may be used to venfy that engineered waste package /re.

1985. 700pp. 8506270247. 31259.001. pository systems comply with NRC radionuclide containment See NUREG/CR-2663,V01 abstract. and controlled release performance objectives. Work covered in this report includes tuff packing material for use in a high level NUREG/CR-2718: STEAM EXPLOSION EXPERIMENTS WITH waste tuff repository. Ranges of repository conditions relevant SINGLE DROPS OF IRON OXIDE MELTED WITH A CO2 to its testing and other factors importart for its performance are LASER.Part II:Parametnc Studies. NELSON,LS.; DUDA.P.M. discussed.

Sandia National Laboratones. Apnl 1985.154pp. 8506140047.

SAND 82-1105. 30908:219. NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFI-1 The steam explosco experiments performed with single CATION TESTS. Semiannual Report Covering The Period Apnl

drops of mol'en iron oxide metted with a CO(2) laser, desenbed 1984 September 1984. JAIN.H.; VEAKis,E.; SOO,P. Brookha-I in Part 1 of this report, were extended here. The following mator ven National Laboratory. June 1985. 34pp. 8507050402. BNL-parameters were vaned: ambient pressure, water temperature NUREG-51630. 31372:302.

1 e

j Main Citations and Abstracts 11 i

This ongoing study is part of a task to specify tests that may data in parameter ranges not previously obtained. Companson l be used to venfy that engineered waste packages / repository of the data with current vapor generation models and wall heat systems comply with NRC radionuchde containment and con- transfer models yielded unsatisfactory results. This 6s attnbuted i

trolled releaso performance objectives. Work covered in this to the effects of nonequilibrium, quench front quality, and dis-report includes crushed tuff packing material for use in a high tance from the quench front, which ere factors not included in level waste tuff repository. A review of available tests to quanti- the current models compared. j fy packing performance is given together with recommendations

' NUREG/CR-3197 V01: REACTION DETWEEN SOME CESIUM-for future testing work.

LODINE COMPOUNDS AND THE REACTOR MATERIALS 304

, NUREG/CR-3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES STAINLESS STEEL,1NCONEL 600 & SILVER. Volume i. Cesium OF POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT Hydroxide Reactions. ELRICK,R.M.; SALLACH R.A.;

ZONE. Annual Report For 1983. HINZE,W J.; BRAILE.LW, OUELLETTE.A.L.; et el. Sandia National Laboratones. June Purdue Unrv West Lafayette, IN. KELLER,G R.; et al. Texas, 1985.156pp. 8507020369. SAND 83-0395. 31307:175.

Univ. of, El Paso, TX. Apnl 1985. 60pp. 8504220438. Laboratory scale scoping studies, using chemcal simulants, i 29946:263. are examirung physcal and chemical processes that could Recent geophyscal investigations have shown that the sers- occur between fisson products and other pnmary system mate- ,

. mcity of the New Madnd, Missouri, seismogenc regon corre- nals in a steam and hydrogen environment. The chemical sys-lates with an ancient nft complex suggesting that the anoma* tems studied were cesium hydroxide vapor reactions in steam

( lous seismcsty is the result of the locahzaten of the regional and hydrogen at 970K in a 304 Stainless steel system, at

, compressive stress pattern by basement structures. An integrat' 1120K in a 304SS system, at 1000K with iodine vapor in an alu- ,

1 ed geophyscal/ge6cgical research program is being conducted mina system and at 1000K with hydrogen iodine vapor in an

] to evaluate the nft complex hypothesis, to refine our knowledge alumina system. Major observations and conclusons are that; j of the structure and physical properties of the nft complex, and cesium in the CsOH reacts with the sihcon dioxide in the inner

to investigate the possible northern extensions of the New oxide formed on stainless steel to produce a cesium silicate; Madnd Fault zone, especially the possible northeastern connec- the availabihty of SiO(2) may therefore control the extent of re-I tion to the Anna, Oho, seismic region. Investigaton of the action of CsOH with 304SS in steam; the oxidation rate of j northeast extension has focused upon the acquisition and prep- 304SS is enhanced by the exposure to CsOH vapor; the reac-j aration of arrays of gravity and magnetic data sets. Dunng 1983, tien of CsOH with iconel 600 is slow in steam and seems to special emphasis was placed upon integration of these data react with the silica content in the oxide layer, with basement lithologc and seismcity informaton which has revealed several major hthologic/ structural features in the crust NUREG/CR-3208

TRAC.PD2 DEVELOPMENTAL ASSESSMENT.

. of the Anna area. Current seismcity in this region appears to be KNIGHT T D.; METZGER,V, Los Alamos Scient.fc Laboratory.

related to an ancient nft structure (the Fort Wayne nft) and pos- Apnl 1985. 371pp. 8504160087. LA.9700-MS. 29835:001.

sibly its contact wrth a low density pluton. Minor seismicity may This report desenbes the final results of the development as-be caused by stress concentraten associated with local base- sessment analyses conducted dunng the later stages of the j ment inhomogeneities. TRAC-PD2 development. The calculatons discussed in this NUREG/CR-3178: STRUCTURAL AND TECTONIC STUDIES IN report used the released version of TRAC-PD2 and cover sepa-NEW YORK STATE. Final Report, July 1981 - June 1982 rate-ettects Diowcown, neat transfer, and oowncomer penetra-1 i ISACHSEN,Y.W. New Yock, State Unsv. of, Albany, NY. tion tests together with mtegral tests from the Loss-of-Fluid Test )

Boston College, Chestnut Hill, MA. Apnl 1985. 84pp. and Semiscale facilities. Although these calculatons are not an j 8505100048. 30270:214. exhaustue test of the code, they demonstrate its capabilities, in-

! Subjects treated in this report include the distnbuton, trends, ciuding automate steady. state trutiahzaten and the complete exposure charactenstics, aeromagnetic signatures, and detailed transient from blowdown through refill and reflood. The results i

geometnes of fracture systems, as well as tentatue inferences show good agreement between the calculated parameters and concerning relative ages and causes of reactivation. Stress indi- the data and indcate that the code is applicable to large-break l cators are discussed, and a begintung is made at working out loss-of-coolant accident analyses.

l regional paleostress directons using the attitudes of dated NUREG/CR-3228 V03: STRUCTURAL INTEGRITY OF WATER

mafic dikes. Attempts at defining Holecene and recent crustal REACTOR PRESSURE BOUNDARY COMPONENTS. Annual

! movements using geological, geodetic, and seismological meth- Report For 1984. LOSS,F.J. Materials Engineenng Associates" ods are reviewed, as well as attempts to relate projected focal Inc. June 1985.171pp. 8506260518. MEA-2075. 31245.026.

mechanism solutons to ground geology. Finally, the distnbuten This program consists of research and engineenng relating to

of earthquakes and their relatonships to geology is reviewed. fracture, fatigue and radiaton sensituity of nuclear structural 1 NUREG/CR-3193

FORCED steels and weldments and addresses many of the key uncer.

CONVECTIVE.NONEOUILIBRIUM POST-CHF HEAT TRANS- tainties in the margin of safety in operating nuclear plants. All FER EXPERIMENT DATA AND CORRELATION COMPARISON tasks are integrated to focus on structuralintegnty of LWR pres-REPORT. GOTTULA,R.C.; CONDIE,K.G.; SUNDARUM R.K.; et sure boundary components. The approach centers on an exper-al. EG8G, Inc. Apnl 1985. 562pp. 8504160110. EGG-2245. imental characterization of nuclear grade steels and an assess-29833:001. ment of fracture and environmental cracking behavior under Forced convectue postcntical-heat-flux heat transfer experi- conditions of a nuclear environment, so investigation of irradiat-ments with water flowing upward in a vertcal tube have been ed matenals is a key element of each task. Emphasis is placed

conducted at the Idaho Natonal Engineenng Laboratory. Ther- on identifying metallurgical factors responsible for radiation em-

! modynamic nonequihbrium in the form of superheated vapor bnttlement of steels and on developing procedures for embrittle-i temperatures was measured at a maximum of three different ment relief, including guidehnes for radiation-resistant stee's. Ex-axial levels. Steady-state experiments were conducted at pres- perimental studies are supported by analytical models and in-l sures of 0.2 to 0.7 MPa mass fluxe:s of 12 to 24 kg/m(2).s, vestigations of the mechanesms responsible for the observed heat fluxes of 7.7 to 27.5 kW/m(2), and test section inlet quah- behavior. Data developed in the program will provide the basis

! tes of 38 to 64% Ouasi-steady-state (slow moving quench for recommendations for the ASME Boder and Pressure Vessel front) experiments were conducted at pressures of 0.4 to 7 Code and ASTM test methods, and revisions to NRC Guides.

MPa, mass fluxes of 12 to 70 kg/m(2).s, heat fluxes of 8 to 225 Current work is organized into three major tasks: (1) fracture kW/m(2), and test section inlet quahties of -7 to 47% The muf- mechancs investigattons, (2) environmentally-assisted crack tiple probe data and the data taken above 0.4 MPa are new growth in high temperature, primary reactor water and (3) radi-I 4

-.,--e.  % ,,,,-.--.--,--ee. - - , . - . . = - - . , ,,e.e. . - . - -e,s.----me----...-----..--.-.=-.--~-,---.-------.---.,,-c, - . , - - -

I i

12 Main Citations and Abstracts abon sensitivity and postirradiaton properties recovery. Re- are the two thermal-hydraule models used in the code. Steam search progress in these three tasks for 1984 is summarized evaporation was found to signifcantty retard deposition proc-here. esses in pools near the boiling point. The code user supplies NUREG/CR-3293 V01: TECHNOLOGY, SAFETY AND COSTS OF the values of several controlhng variables yn the code input. The SPARC output can include the decontaminabon factors (DF) of DECOMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON-FUEL CYCLE FACILITIES FOLLOWING POSTULAT-twenty different partcle size groups, an overall DF for the whole ED ACCIDENTS. Main Report. ELDER,H.K. Battelle Memorial in- particle distnbution, particle log normal distnbution parameters, stitute, Pacific Northwest Laboratones. May 1985. 327pp. and mass flow rates of partcles (wet and dry) leaving the pool.

8506140337. 30932:308. NUREG/CR-3455: A COMPARISON OF IODINE KRYPTON,AND Techncal requirements, costs and safety are conceptually XENON RETENTION EFFICIENCIES FOR VARIOUS SILVER evaluated for the post-accident cleanup and decommissoning LOADED ADSORPTION MEDIA. HUCHTON,R.L.;

of fuel cycle and non-fuel cycle facilities that have experienced TKACHYK J.W.: TAYLOR.J.T.; et al. Westinghouse Electric a significant accident. Accident cleanup is postulated to include Corp. Apnl 1985. 80pp. 8505230585. WINCO-1024. 30546:254.

1) initial decontamination of building surfaces to reduce the sub- A comparison was made among vanous sitver impregnated sequent occupatonal dose to cleanup and decommissioning adsorpton media to determine their iodine, krypton, and xenon workers and 2) management of the resulting wastes. Decom- retenton efficiencies. The program consisted of three compo-missioning is assumed to follow accident cleanup. In order to nents. First, laboratory measurements of the noble gas retenton ensure that worker doses are ALARA, despite higher radiation efficiencies of commercially available adsorpton media were exposure to workers during post-acodent operations, careful determined as a functon of relative humidsty, sample duraten, planning and rehearsal of cleanup operatens and the use of test cartndge geometry, and ambient air purge. Second, a litera-remote and semi-remote cleaning techniques are required to ture survey was performed to evaluate the iodine species reten-reduce occupancy times in high radiation areas and to minimize tion efficiencies of the selected media. Third, data associated occupational exposures during accident cleanup. The public with a media previously proposed for an emergency response safety impacts of post-accident cleanup and decommissioning air sampler were incorporated to enlarge the data base.

r are also evaluated, these are below permissable radiaton dose levels in unrestricted areas and well within the range of annual NUREG/CR-3469 V02: OCCUPATIONAL DOSE REDUCTION AT j!

radiaton doses from normal background. NUCLEAR POWER PLANTS. Annotated Bibliography Of Select-ed Readings in Radiation Protection And ALARA. BAUM.J.W.;

NUREG/CR-3293 V02: TECHNOLOGY, SAFETY AND COSTS OF WEILANCICS.C. Brookhaven National Laboratory. June 1985.

DECOMMISSIONING REFERENCE FUEL CYCLE AND NON- 150PP. 8507020380. BNL-NUREG-51708. 31306:251.

FUEL CYCLE FACtLITIES FOLLOWING POSTULATED This is the second volume of abstracts dealing with occupa-ACCIDENTS. Appendices. ELDER,H.K. Battelle Memonal insti- tional dose, dose control, dose reducten and apphcation of the tute, Pacific Northwest Laboratones. May 1985. 288pp. ALARA (as low as reasonably achievable) pnnciple at nuclear 8506170550. 30979:001. power plants. This volume contains abstracts selected from AP.

This volume contains the appendices conceming the techni- PLIED HEALTH PHYSICS ABSTRACTS AND NOTES, Volumes 1

ca! requirements, costs and safety aspects conceptually evalu- 1, No.1,1975 through Volume 5, No. 4, October 1979, and

, ated for post-accident cleanup and decommissoning of fuel frnm racant rubhentinrm knnwn ta tha amhare Anmar and m.h- ~

4 cycle and non-fuel cycle facihties that have experienced a sig- ject indexes' are included.'The ' subject in'dex in this v'olutSe nificant accident. Accident cleanup is postulated to include 1) covers abstracts in both Volumes 1 and 2. This volume contains

initial decontamination of building surfaces to reduce the sube abstract Numbers 252 through 549.

quent occupational dose to cleanup and decommissioning work-ers and 2) management of the resulting wastes. Decommission- NUREG/CR-3514 V02: THE CHEMICAL BEHAVIOR OF IODINE ing is assumed to follow accident cleanup. In order to ensure IN AOUEOUS SOLUTIONS UP TO 150 C.lf.Radiaton-Redox that worker doses are AMRA, despite higher radiaton expo. Condit#ons. TOTH LM.; DODSON K.E. Oak Ridge Natonal Lab-sure to workers during post-accident operations, careful plan- oratory. April 1985. 22pp. 8506100496. ORNL/TM-8664/V2.

ning and rehearsal of cleanup operations and the use of remote 30830:083.

and semi-remote cleaning techniques are required to reduce oc- Redox reactions that might alter the volatihty of aqueous cupancy times in high-radiation areas and to minimize occupa- iodine solutions have been examined expenmentally using ab-tonal exposures dunng cleanup. The public safety impacts of sorption spectrophotometry. Oxygen and hydrogen atmospheres post-accident cleanup and decommissioning are also evaluated; had no effect on the iodine chemistry at temperatures up to 150 i 2 these are below permissible radiaton dose levels in unrestricted degrees centigrade. However, irradiation of aqueous solutions i

areas and well within the range of annual radiation doses from with a (60)Co source,0.8 x 10(6) R/h, produced radiotysis prod-normal background. ucts that either ordized iodine ion or reduced 10(3)- in the pH

'*""

• 9"" #*" * " * " " * ' " ' *

• NUREG/CR-3317: TECHNICAL BASES AND USER'S MANUAL The amount of iodine volatihzed vaned from a few percent for 1 FOR THE PROTOTYPE OF SPARC A SUPPRESSION POOL AEROSOL REMOVAL CODE. OWCZARSKl P.C.; POSTMA,A.K.; solute concentratons of 10(-4) M to as rnuch as 10 to 19% for

,l SCHRECK,R.t. Battelle Memorial Institute, Pacific Northwest 10(-6) M Csl or Kl0(3) solutes. Silver metal has been shown to l provide an effective gettenng route for l- in solution if these ions i

Laboratories. May 1985. 68pp. 8506240650. PNL-4742. are first oxidized by OH radcats generated during the radclysis I l 31152

156. of the solutons.

The Pacife Northwest Laboratory has developed a prototype

! verson of a Suppression Pool Aerosol Removal Code (SPARC). NUREG/CR-3551: SAFETY IMPLICATIONS ASSOCIATED WITH l This code was wntten to calculate the capture of aerosol parti- IN-PLANT PRESSURIZED GAS STORAGE AND DISTRIBU-cles in the pressure suppression pool (wet well) of a boiling TION SYSTEMS IN NUCLEAR POWER PUNTS.

water reactor under hypothetical accident conditions. The code GUYMON,R.H.; CASTO,W R.; COMPERE,E.L Oak Ridge Na-incorporates five aerosol scrubbing models and two thermal-hy. tional Laboratory. May 1985.82pp.8506140622. ORNL/NOAC-1 drauhc models. The scrubbing models describe 1) steam con- 214. 30934:031.

l densation,2) soluble particle growth in a humid atmosphere,3) Storage and handling of compressed gases at nuclear power gravitational settling, 4) inertial depositen, 5) diffusonal deposi- plants were studied to identify any potential safety hazards.

< tion. Mechanical entrainment of pool houid by breaking of bub- Gases investigated were air, acetylene, carbon dioxide, chlorine, I bles at the surface was also considered. An optonal model for Halon, hydrogen, nitrogen, oxygen, propane, and sulfur hexa-equilibrium pool temperature and a model for steam evaporaten fluoride. Physical properties of the gases were reviewed as s

Main Citations and Abstracts 13 l

were appleable industnal codes and standards. Incidents involv- ments. IGSCC is caused by a combination of a sensitized mi-ing pressunzed gases in general industry and in the nuclear in- crostructure, an aggressrve environment, and tensile stress.

dustry were studsed. In this report general hazards such as mis- Control of any of these thr(e factors can ehminate IGSCC in sales from ruptures, rocketing of cylinders, pipe whipping, as- most practical situations. This program will measure and model phyxiaten, and toxicity are discussed. Even though some sen- the development of a sensitized microstructure as it pertains to ous injuries and deaths over the years have occurred in indus- welded and repair-welded SS pipe. An empincal correlaten be-tries handling and using pressurized gases, the industrial codes, tween a matenars DOS and its susceptibihty to IGSCC will be standards, practces, and procedures are very comprehensue. determined using constant extension rate tests (CERTs). The The most important safety consideraten in handkng gases is successful completon of these tasks will result in a method for the serious enforcement of these well-known and established assessing the effects of welding /repainng parameters on the methods. Recommendabons are made concerning compressed IGSCC susceptibihty of component.specife nuclear reactor gas cylinder missdes, hydrogen hne ruptures or leaks, and iden- welds /repars.

tificaton of lines and equipment.

NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PER-NUREG/CR-3558: HANDBOOK OF NUCLEAR POWER PLANT FORMANCE SIMULATION (MAPPS) MODEL DESCRIPTION SEISMIC FRAGillTIES. Seisme Safety Margins Research Pro- OF MODEL CONTENT STRUCTURE.AND SENSITIVITY TEST-pam. COVER LE.; BOHN.M.P.; CAMPBEU,R.D.; et al. Law- ING. SIEGEL,A1; BARTTER,W.D ; WOLF,J.J.; et al Oak Ridge rence Livermxe National Latoratory. June 1985. 300pp. National Laboratory. Apnl 1985. 322pp. 8504170234. ORNL/

8507080210. UCRL-53455. 31402:238. . TM-9041/V2. 29902:002.

The Seismic Safety Margins Research Program (SSMRP) is This volume of NUREG/CR-3626 presents details of the con-an NRC-funded, multryear program conducted by Lawrence tent, structure, and sensitroty testing of the Maintenance Per.

Livermore Natonal Laboratory (LLNL). Its goal is to develop a sonnel Performance Simulation (MAPPS) model that was de-complete end fully-coupled analysis procedure, including meth- scnbed in summary sn volume one of this report. The MAPPS ods and computer codes, for estimating the nsk of earthquake- model is a generalized stochastic computer simulation model induced radcactive release from a commercial nuclear power developed to simulate the performance of maintenance person-plant. As part of this program, calculations of the seismic nsk net in nuclear power plants. The MAPPS model considers work-from a typical commercial nuclear reactor were made. These place, maintenance technician, motivation, human factors, and l calculatons requwed a knowledge of the probability of farfure task onented vanables to yield predectue information about the (fragility) of safety-related components in the reactor system effects of these vanables on successful maintenace task per-that actively parberpate in the hypothesized accident scenanos. formance. All major model vanables are discussed in detail and q This report desenbes the development of the required fragility their implementabon and interactue effects are outlined. The relations and the data sources and data reducton techniques model was examined for disqualifying defects from a number of upon whch they are based. Both building and component fragi- viewpoints, including sensitmty testing. This examination led to hties are covered. The building fragilities are for the Zon Urut 1 the identifcaten of some minor recahbration efforts which were reactor, the speofic plant used for development of methodology carried out. These positue results indicate that MAPPS is ready in the program. Some of the component fragilities are site-spe- for initial and controlled appbcatons whch are in conformsty l

' cific, but most would be usable for other sites as well. with its noirnnsas NUREG/CR-3611: RADIOACTIVE MATERIAL (RAM) ACCIDENT /

INCIDENT DATA ANALYSIS PROGRAM. EMERSON.E.L; NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PER-MCCLURE.J D. Sandia Natonal Laboratories. Apnl 1985.40pp. FORMANCE SIMULATION (MAPPS) MODEL: DESCRIPTION OF MODEL CONTENT. STRUCTURE,AND SENSITIVITY TEST-8504220385. SAND 82-2156. 29946:323.

I This report descnbes the development of the Radcactue Ma, ING. SIEGEL,A.I.; BARTTER,W D.; WOLF.J.J.; et al. Oak Ridge tenals Transportation Accident / Incident Data Base (RAM.AIDB), National Laboratory. Apnl 1985. 322pp. 8504170234. ORNL/

which contains information on the occurrences of transportation TM-9041/V2. 29902.002.

. Eccidents and incidents, for radcactue matenals (RAM) that are This volume of NUREG/CR 3626 presents detatis of the con-involved in the process of transportaten, loading and unloading tent, structure, and sensitmty testing of the Maintenance Per-operations, or temporary storage. These transportabon oper. sonnel Performance Simulation (MAPPS) model that was de-i scribed in summary in volume one of this report. The MAPPS

Etions are in support of the nuclear fuel cycle for electncal energy generabons of RAM. This study analyzes in some detail model is a generafrzed stochastic computer simulaton model i

basic accident / incident stat: steal data, RAM packaging acci. developed to simulate the performance of maintenance person-

< dent response data, and the health effects associated with nel in nuclear power plants. The MAPPS model considers work.

RAM transport accidents / incidents. This report presents a sum. place, maintenance technician, motivaton, human factors, and mary of U.S. RAM transport accident / incident experience for task oriented vanables to yield predictrve information about the

! the period 1971 through December 1981. In addition, a sample effects of these vanables on successfut maintenace task per. l

! Ennual summary of accident / incident expenence is presented formance. All major model vanables are discussed in detail and for the calendar year 1981, their implementaten and interactue effects are outhned. The model was examined for disqualifying defects from a number of i

j NUREG/CR 3613 V02: EVALUATION OF WELDED AND viewpoints, including sensatmty testing. This examinaton led to i REPAIR-WELDED STAINLESS STEEL FOR LWR the identificaton of some minor recabbraton efforts which were i SERVICE. Annual Report for 1984. ATTERIDGE D.G.; carried out. These positue results indicate that MAPPS is ready l l

BRUEMMER,S.M.; PAGE.R E. Battelle Memonal Institute, Pacif* for initial and controlled applcatons whch are in conformity ic Northwest Laboratones. June 1985. 63pp. 8506270333. PNL. with its purposes.

1 4971.31262:215. .

1 Pacific Northwest Laboratory (PNL), under a program spon- NUREG/CR 3647: DESIGN AND FABRICATION OF A 1/8-sored by the Dmsion of Engineenng Technology of the U.S. Nu- SCALE STEEL CONTAINMENT MODEL REESE R.T.;

clear Regulatory Commission (NRC), is conducting a program to HORSCHEL,0.S. Sandia National Laboratones. Apnl 1985. '

determine a method for evaluating the acceptance of welded 131pp. 8504170693. SAND 84-0048. 29907.022.

Ond repair-welded stainless steel (SS) piping for hght-water re- A 1/8-scale steel model containment building was designed

) actor (LWR) service. Vahdated models, based on experimental and fabncated in support of the Containment Safety Margins i

! data, will be developed to predict the degree of sensitization Program. This program is directed to determine the margin of (DOS) and the intergranular stress corrosion cracking (IGSCC) safety of containments in severe accident conditons. It is susceptibihty in the heat affected zone (HAZ) of the SS weld- planned to internally pressunze the model to failure. In this test-l l

l 4_ _ _ ._ . _ _ _ . _ _ . . _, __ _ _ _ . . .

14 Main Citations and Abstracts ing program, failure modes of the pressure vessel and scaled tective Action Guides. Using relatively conservative assumptons penetrations will be examined in detail. The model was de- in the screening analysis, all but at most a few hundred hcenses signed according to Section ill of ASME Code for Class MC were found to have estimated doses below the Protective containment vessels with the exception that no code stamp was Action Guide levels. The few hundred identified in this initial required since no nuclear matenals would be housed within the screening should be further evaluated using realistic assump-model. All the general requirements (subsection NCA) and spe- tions and site specific information to estabhsh the need for, ap-cific requirements (subsection NE) of Section ill of the ASME propnate level and extent of, and potential effectiveness of Code were met. The majonty of the model was fabncated from emergency response planning and preparedness beyond that 3/16-in. SA516 Grade 70 steel plate in the form of a nght circu-currently required.

lar cylinder capped with a hemisphencal dome. Eleven penetra-tions and two lifting trunnions were included in the model. The NUREG/CR-3703: ASSESSMENT OF SELECTED TRAC AND cyhnder/ dome section was joinded to a 2:1 ellipsoidal base (test fixture) composed of thicker (1 1/8-in. and 1 1/2-in ) plate RELAPS CALCULATIONS FOR OCONEE-1 PRESSURIZED THERMAL SHOCK STUDY. ROHATGI,U.S ; PU,J.; SAHA P.; et matenal. The model was supported on six legs to permit access for personnel, instrumentaten, data acquisition, power, and al. Brookhaven National Laboratory. Apnl 1985. 98pp.

pressure piping. The model was fabricated in Apnl through Oc- 8505070497. BNL-NUREG-51750. 30211:005-tober 1983 by Chicago Bndge and Iron and erected at the test Several Oconee-1 overcooling transients that were computed site in Albuquerque, New Mexico, in November 1983. by LANL and INEL using the latest verseons of TRAC-PF1 and RELAPS/ MOD 1.5 codes have been reviewed by BNL Three of NUREGiCR-3651: ASSESSMENT OF THE ADEQUACY OF these transients were selectd for detailed review as they either ORNL INSTRUMENTATION IN REFLOOD TEST FACILITIES.

HARDY.J.E.; HERSKOVITZ.M.B. Oak Ridge National Laborato-had the potential of challenging the integnty of the pressure ry. Apnl 1985. 56pp. 8506070366. ORNL/TM-9067. 30798.265. vessel or highhghted the effect of code differences. These are Instrumentaten for making two-phase measurements in ex. (1) Main Steam Line Break (MSLB), (2) All Turbine Bypass penmental refill-reflood test facilities was developed by Oak Valves Stuck Open' and (3) 2-inch Small Break LOCA-Ridge Nabonal Laboratory (ORNL) through the Advanced Instru.

mentation for Reflood Studies (AIRS) program. These unique in- NUREG/CR-3721 V01: PRESSURE MEASUREMENTS IN A HY-strumentation systems were dessgned to survive the severe in- DROGEN COMBUSTION ENVIRONMENT. Hydrogen-Air Com-vessel environmental conditions that exist dunng a sirnulated bustion Test Senes 1 And 2 in The FITS Tank. ROLLER,S.F.

pressurized water reactor foss-of-coolant accident (LOCA). The Sandia National Laboratones. Apni 1985. 59pp. 8504170005.

measurements include two-phase flow velocity, void fracton, SAND 83-2621/1. 29904:307.

and film thickness and velocity, and are required for better un- Hydrogen combustion tests were performed in the Fully in-derstanding of reactor behavior dunng LOCAs. The adequacy strumented Test Site (FITS) tank under the Hydrogen Behavior (sunnvabihty and data quahty) of the instrumentaten systems in- Program performed by Sandia National Laboratones under con-stalled in four expenmental reflood test facilities is assessed. tract with the US Nuclear Regulatory Commisson. Test senes 1 Signal conditioning electronics and sensor thermocouples func- and 2 examined the effects of a number of parameters on hy-toned extremely well. For the first time, two-phas,e flow meas-drogen-air combuston: the initial temperature and pressure of wenwnm wwe mauw sin.uiu uunrig a annusaivu w. oecause the gases, the effect of added steam or carbon dioxide as di-of the harsh environment and geometncal constraints, some luents, and the percent hydrogen in air. For tests in the range of sensor failures were considered liliely; the number actually fail- 20% to 40% hydrogen in air, recorded peak pressures were ing in sennce was within expectatens. An exception to this equal to adiabatic, isochroic, complete combustion (AICC) record occurred in the Slab Core Test Facihty - Core 1. A chlo-nde-ion stress corroson problem destroyed signal cables at the values within an expenmental error of 15% This was contrary vessel seal for most sensors. This problem was corrected by to the results of tests at a number of other facihties. The pre-changing the sealant matenal at the vessel penetraton in the @ tion temperature had a strong effect on the peak pressure, subsequent facslaties. Overall, the performance of the instrumen- while pre-egniten pressure in the range examined had no effect tation was very satisfactory yielding valuable data dunng simu- on combustion pressure ratios. Calculations showed that, al-lated LOCAs in refill-reflood test facilities. though the effect of dynamic head on the peak pressure was a few percent or less, interactons of the wave preceding the NUREG/CR-3657: PREUMINARY SCREENING OF FUEL CYCLE flame front with the flame and with the vessel walls may be ap-AND BY-PRODUCT MATERIAL UCENSES FOR EMERGENCY parent in the expenmental records.

PLANNING. BENNETT D E.; RUNKLE,G E.; ALPERT,DJ.; et al.

Sandia National Laboratones. Apnl 1985.137pp. 8506060385. NUREG/CR-3746 V02: LWR PRESSURE VESSEL SURVEIL-SAND 84-0186. 30775.062.

LANCE DOstMETRY IMPROVEMENT PROGRAM. Semiannual This report summanzes work done for the U.S. Nuclear Regu- Progress Report. April 1984 September

- 19841 s latory Commisson as part of a program considenng the need LIPPINCOTT,E.P.; MCELROY,W N. Hanford Engineenng Devel- I for and appropriate level of emergency response planning at opment Laboratory. Apnl 1985.220pp.8505070562. HEDL TME fuel cycle and by-product matenal facilities. The purpose is to 84 21. 30209.020 (1) provide a base of technical information for identifying and ranking those facihties for whic's the need for emergency re- Water Reac-p sponse planning and preparedness should be further consid-ered, and (2) perform an initial screening of hcenses issued by gram (LWR-PV-SDlP) dunng FY84. The pnmary concern of this NRC. A data base containing the radionuchde possesson hmits program is to improve, test, venfy, and standardize the physics-for each hcense was developed. Dose estimates for a unit (1 dosametry-metallurgy and associated reactor and damage analy-cune) release of each of the radionuclides in the data base sis procedures and data used for predicting the integrated ef-were calculated. To account for the variability in weather, distri- fects of neutron exposure to LWR-PVs and their support struc-butions of doses were estimated for a full range of meteorologi. tures. These procedures and data are being recommended in a cal conditons. As requested by NRC, doses at the 99th per- new and updated set of ASTM standards being prepared, centile of the distnbuten were used. An initial screening analy- tested, and venfied by program participants. These standards, sis was performed for the approximately 9400t hcenses by together with parts of the US Code of Federal Regulatons ano companng the estimated 99th percentile dose for a postulated ASME codes, are needed and used for the assessment and release of a fracton of the hcensed possesson hmet to the dose control of the condition of LWR-PVs and their support structures levels suggested in the Environmental Protection Agency's Pro- during the 30. to 60-year hfetime of a nuclear power plant.

k Main Citations and Abstracts 15 NUREG/CR-3746 V03: LWR PRESSURE VESSEL SURVEIL- tained. The results suggest, as have the results of other invesh-LANCE DOSIMETRY IMPROVEMENT PROGRAM.1984 Annual gators, that the measured rates and reaction parameters may Report, October 1,1983 September 30,1984. MCELROY,W.N. not be those of any specifc reacton, but are instead the " effec-

< Hanford Engineenng Development Laboratory. Apnt 1985. bye" values of a senes of complex systems operating together.

110pp.8505070543. HEDL TME 84-31. 30208:270. However, the total quantity of hydrogen generated by this See NUREG/CR-3746,V02 abstract. mechanism is signifcantly less than can be produced from t other sources, e g., steam. zirconium.

NUREG/CR-3747: THE SELECTION AND TESTING OF ROCK FOR ARMORING URANIUM TAILINGS IMPOUNDMENTS- NUREG/CR-3804 V04: PHYSICS OF REACTOR l FOLEY M.G.; KIMBALL.C.S.; MYERS,D.A.; et at. Battelle Memo- SAFETY.Ouarterly Report. October-December 1984.

• Argonne nal institute, Pacific Northwest Laboratories. May 1985.119pp. National Laboratory. Apnl 1985. 20pp. 8504250268. ANL-84-35 8506140396. PNL-5064. 30933.275. V04. 30032.091.

Under contract to the U.S. Nuclear Regulatory Commission. This quarterty progress report summanzes wc4k done dunng Pacifc Northwest Laboratory has developed an approach for the months of October December 1984 in Argonne Nabonal selecting and teshng rock for its suitability and durabihty as Laboratory's Applied Physcs and Components Technology Divi-trmor for protecting decommissioned uranium mill tarhngs piles. sions for the Division of Reactor Safety Research in the U.S.

1 A prehminary survey of the hterature determined that existing Nuclear Regulatory Commisson. The work in the Apphed Phys-

! techniques for testing rock durability were inadequate for evalu- ics Division includes reports on reactor safety modehng and as-Eting long-term (100 years) appicatens. Suites of rock samples sessment by members of the Reactor Safety Appraisals Sec-with common hthologies and documented duratons of exposure ton. Work on reactor core thermal-hydrauhes is performed at to weathenng were then collected and submitted to three-axis ANL's Components Technology Division, emphasizing 3-dimen ultresonsc testing in an attempt to develop a more rehable test- sonal code development for LMFBR accidents under natural ing technique. We found little correlation between the duration convection conditons. An executrve summary is provided in-of weathenng and ultrasound velocity or attenuation in the rock- cluding a statement of the findings and recommendatons of the Through further study, we determined that the best screening report.

approach incorporates common geomorphologe field collecton techniques and laboratory tests. Suites of samples with known NUREG/CR-3804 V04: PHYSICS OF REACTOR durabons of exposure to weathenng can be subjected to wet SAFETY.Ouarterly Report. October-December 1984.

• Argonne abrason and wetting-drying tests to screen local rock types and Nabonal Laboratory. Apnl 1985. 20pp. 8504250268. ANL-84-35 select those with the greatest potental durabihty. Furthermore, V04. 30032-091. .

the expected decrease of rock mass with environmental This quarterly progress report summanzes work done during stresses (e.g, flood impingement and diumal wetting-drying the months of October-December 1984 in Argonne National cycles) can be estimated using this approach. Laboratory's Applied Physics and Components Technology Divi-sons for the Divison of Reactor Safety Research in the U.S.

NUREG/CR-3757: TRAN B-2:THE EFFECT OF LOW STEEL Nuclear Regulatory Commissen. The work in the Applied Phys-CONTENT ON FUEL PENETRATION IN A NON-MELTING CY- ics Division includes reports on reactor safety modehng and as-M THUR.D.A.; MAST,Pg. sessment by members of the Reactor Safety Appraisals Sec-

{NDRICAL FLOW CHAN]EL , ,

~....~.m~,-~,~s. ni- . s. , , usus . vv . . ton. Work on reactor core thermat-nyoraulcs is performe1 at SAND 84-0814. 30457.029. ANL's Components Technology Divison, emphasizing 3-dimen-l The TRAN B-Senes of expenments is being conducted at sional code development for LMFBR accidents under natural Sandia National Laboratones to investigate the charactenstics convection conditons. An executive summary is provided in-of fuel removal and freezing through the upper axial blankets of ciuding a statement of the findings and recommendatons of the

an LMFBR dunng the transiten phase of a hypothetical core report.

I disruptive accident. The second expenment in this senes. TRAN B-2, was performed in July 1983. This expenment involved the NUREG/CR-3810 V04: REACTOR SAFETY RESEARCH injection of a mixture of 95% UO(2) and 5% stainless steel into PROGRAMS Ouarterly Report. October-December 1984.

a simple thick walled steel cyhndrical flow channel. The inabal EDLER,S.K. Battelle Memorial Insttute, Pacife Northwest Lab-temperature of the steel channel was low, such that melting of oratones. May 1985. 31pp. 8506140415. PNL-5106-4. l the walls upon contact with the hot melt was not expected. Pre- 30998:188.

vious experiments under similar conditions but using pure UO(2) This document summarizes work performed by Pacific North-a melts had shown stable crust growth and fairly long penetration west Laboratory from October 1 through December 31, 1984, distances. This expenment was intended to investigate whether for the Division of Accident Evaluation and the Division of Engi-those results were also apphcable for the case of UO(2)/ steel neenng Technology, U.S. Nuclear Regulatory Commission. Re-  !

! mixtures. The results of the TRAN B-2 expenment, consisting of suits from an instrumented fuel assembly irradiation program l data from online instrumentation and post-irradiation examina- being performed at Halden, Norway, are reported. Accelerated bon, suggest that low steel content fuel / steel mixtures behave pellet-cladding interaction modehng is being conducted to pre-very similarly to pure UO(2) melts, dict the probabihty of fuel rod failure under normal operating undi . n en I at ad n al rno S are being NUREG/CR-3803: THE EFFECTS OF POST-LOCA CONDITIONS pr vided to aid in decison making regarding pipe-to. pipe im-ON A PROTECTIVE COATING (PAINT) FOR THE NUCLEAR pacts following postulated breaks in high-energy fluid system POWER INDUSTRY. LOYOLA.V.M.: WOMELSDUFF,J.E. Sandia assen es a ana al upp 4 are being pW National Laboratones. May 1985. 50pp. 8505230538. SAND 84- ed for experimental programs at the Power Burst Facihty, Idaho aWal Wenng Wa% % FaHs, MaM @h I, W have ied the oxidation of zinc in a zinc-nch coating

. perature materials property tests are being conducted to pro-I used in the rNclear power industry and have measured the yde data on severe core damage fuel behavior. Thermal-hy.  !

, rates of hydrof en generaton from these coatings due to zinc drauhc computer programs are providing best-estimate analvses j oxidaten at temperatures of up to 175 degrees ceraigrade. The for a unety of safety issues in hght-water reactors. Severe fuel results suggest that the real time rates of hydrogen generation damage tests are being conducted in the NRU Reactor, Chalk are considerab!y higher than prevously beheved. The higher River, Canada.

rates measured in this study are probably due to differences in i expenmental methodologios between this s'd previous studies. NUREG/CR-3816 V02: REACTOR SAFETY RESEARCH.Ouarterty j in this study, the measurements were real-time measurements, Report.Apnt-June 1984.

• Sandia National Laboratones. Apnl as opposed to time-averaged values which are typically ob- 1985. 211pp. 8504160554. SAND 841072. 29825:153.

1 i

16 Main Citations and Abstracts This report desenbes progress in a number of activities deal- NUREG/CR-3863: ASSESSMENT OF CLASS 1E PRESSURE ing with current safety issues relevant to both light water reac- TRANSMITTER RESPONSE WHEN SUBJECTED TO HARSH tors (LWRs) and breeder reactors. The work includes a broad ENVIRONMENT SCREENING TESTS. FURGAL.D.T.

range of expenments to sirrolate accidental conditons to pro- CRAFT,C.M.; SALAZAR,E.A. Sandia Natonal Laboratones. Apnl vide the required data base to understand important accident 1985.194pp. 8506140052. SAND 84-1264. 30907:289.

sequences and to serve as a basis for development and venfi- An expenmental investigation into the performance of Class caton of the complex computer simulation models and codes 1E electronic pressure transmitters exposed to environments used in accident analysis and licensing reviews. Such a pro- within and beyond the design basis was conducted. Emphasis gram must include the development of analytical models, veri- was placed on determirung the instruments' failure and degrada-fled by expenment, which can be used to predict reactor and ton modes in separate and simultaneous environmental expo-safety system performance under a broad vanety of abnormal sures. Five unaged ITT Barton Model 763 pressure transmitters conditons. were tested and exposed to a unique environment. The re-NUREG/CR-3820 V03: THERMAUHYDRAULIC ANALYSIS RE-sponse of the transminers showed that Mnwature was the @

SEARCH PROGRAM.Ouartert) Report. July-September 1984. mary environmental stress affecting the tested transmitters' per-THOMPSON.S.L Sandia National Laboratories. Apnl 1985. fonnance. Initial large errors that decrease with time-at-tempera-85pp. 8505160172. SAND 84-1025. 30451:130. ture were observed. The source of these errors is believ,ed to The TRAC-PF1/ MOD 1 independent assessment program at be a common mode design weakness in the transmitters cali-Sandia National Laboratones is part of a multi-faceted effort braten potentiometers. This weakness resutts from a depend-sponsored by the Nuclear Regulatory Commission to determine ency of material dielectnc properties on temperature. The modi-the ability of varcus systems codes to predict the detailed ther. fication recommended by the manufacturer, although palliative i mal /hydrautic response of LWRs during accident and off-normal in natute, did reduce this temperature-induced effect after the conditions. This program is a successor to the RELAPS/ MOD 1 first few nmMs of accMent exposa A poMntial second independent assessment prolect underway at Sandia for the common failure mode which activates slowty with time-at tem-last two years. The TRAC-PF1/ MODI code will be assessed perature was also identified. The operaten of this failure mech-against data from vanous integral and separate effects experi- anism is believed to be cata!yzed by the presence of a lubncant mental test facilibes, and the Calculated results will also be used in the production of some potentemeters. The design of compared with results from our previous RELAPS/ MODI inde- this transmitter proved to be exceptenally hard to radiation ef-pendent assessment anafyses whenever possible. fects and there appeared to be no significant synergistic effects between radiation and temperature. The observed responses of NUREG/CR-3855: CHARACTER!ZATION OF NUCLEAR REAC- the transmitters offer support for the positen of IEEE 381-1977 TOR CONTAINMENT PENETRATION - FINAL REPORT. which recommends that electronic modules aged to varying de-SHACKELFORD,M. Argonne Natonal Laboratory.

• Sandia Na- grees of advanced hfe should be tested.

tional Laboratones. Apnl 1985. 361pp. 8505060534. SAND 84-7139. 30192:018. NUREG/CR-3872: DATA ACOUISITION AND CONTROL OF THE This report summanzes the survey work conducted by Ar. HSST SERIES V IRRADIATION EXPERIMENT AT THE ORR.

, gonne National Laboratory on the design and details of major MILLER.LF.; HOBBS.R.W. Oak Ridge Natonal Laboratory. April penetrations in 48 nuclear oower clants. The survey includes all 1985.97pp.8505230574. ORNL/TM-9253. 30547:232.

containment types and matenals in current use. It also includes Documentaten relative to data acquisiten and control for details of all types of penetratens (except for electrical penetra. support of the HSST Series V Irradiation Expenment at the Oak tion assemblies and valves) and the seals and gaskets used in Edge Research (ORR) is included in this report. Part A de-them. The report provides a test matrix for testing major pene, scnbes the computer system hardware and real-time applicahon trations and for testing seals and gaskets in order to evaluate support software, and Part B desenbes the temperature control their leakage potential under severe accident conditions. methodology. Software that acquires data from analog input provides this information to the control algonthm software. Re-NUREG/CR-3862: DEVELOPMENT OF TRANSIENT INITIATING sults from the control algonthm are, in turn, uthzed by software EVENT FREQUENCIES FOR USE IN PROBABILISTIC RISK which controis digital output hardware. Time intervals of execu-ASSESSMENTS. MACKOWIAK,0.P.; GENTILLON.C.D.; tion, as well as sequencing of software modules, are controlled SMITH,K.L EG&G Idaho, Inc. (subs. of EG&G, Inc.) May 1985. through commands to the operating system. Temperature data 278pp. 8506240069. EGG-2323. 31150:002.

are recorded at one-hour intervals on computer printouts for Transient initiating event frequencies are an essential . input to documentaten and immediate analysis and one magnetc media the analysis process of a nuclear power plant probabiliste risk for permanent storage and subsequent analyses. Results from assessment. These frequencies desenbe events causing or re- processing of data files show that the average temperatures at quiring scrams. This report documents an effort to validate and the 1/4T and 3/4T positions are maintained within 2.6 degrees update from other sources a computer-based data file devel- centgrade of 288 degrees centigrade with associated standard oped by the Electnc Power Research Institute (EPRI) desenbing deviatens of less than 3 degrees centigrade. Average tempera.

such events at 52 Unit 9d States commercial nuclear power tures of the other thermocouples are maintained within 288 de-plants. Operating information from the United States Nuclear grees centigrade plus or minus 12 degrees centigrade with Regulatory Commission on 24 add; tonal plants from their date standard deviations less than 3 degrees centigrade.

of commercial operation has been combined with the EPRI data, and the entire data base has been updated to add 1980 NUREG/CR-3883: ANALYSIS OF JAPANESE-U.S. NUCLEAR through 1983 events for alt 76 plants. The validity of the EPRI POWER PLANT MAINTENANCE. BOEGEL,A.J.; CHOCKIE.A.D.;

data and data analysis methodology and the adequacy of the HUENEFELD.J.C.; et al. Battelle Memonal lastitute, Pacife EPRI transient categories are examined. New transient initiating Northwest Laboratones. June 1985.120pp. 8507080183. PNL-event frequencies are denved from the expanded data base 5160.31392:345.

using the EPRI transient categones and data display methods. This report presents the results of a project designed to com-Upper bounds for these frequencies are also provided. Additen- pare and contrast Japanese and United States nuclear power al analyses explore changes in the dominant transients, plant operating experience, preventive maintenance / surveil-changes in transient outage times and their impact on plant op- lance requirements, and organization and management prac-eration, and the effects of power level and scheduled scrams tices relating to maintenance. Findings are based on information on transient event frequencies. A more rigorous data analysis obtained on the November-December 1983 and November methodology is developed to encourage further refinement of 1984 visits to Japan by the NRC and representatives of Bat-the transient initiating event frequencies denved herein. telle's Pacific Northwest Division, and on varcus documents ob-l l

l

Main Citations and Abstracts 17 tained from the Japanese (primanly the Ministry of Intemabonal NUREG/CR 3904: A COMPARISON OF UNCERTAINTY AND Trade and Industry-MITI) dunng and subsequent to the visits. SENSITIVITY ANALYSIS TECHNIQUES FOR COMPUTER U.S. data sources included NUREG-0020 (Greybook) and plant MODELS. IMAN.R.L; HELTON.J.C. Sandia Nanonal Laborato-techncal specifications. The study shows that Japanese plants ries. May 1985.118pp. 8506190020. SAND 841461. 31019 001.

expenenced far fewer manual shutdowns, manual scrams, auto- Uncertainty analysis and sensstrvity analysis are important ele-matic scrams, and reduced loads than U.S. plants and that their ments in the development and implementation of computer mean-time-betweerHrvent (MTBE), even when adi usted for dif- models for complex processes. Typically, there are many uncer-ferences in average plant availability, was approximately 10 tainbes associated with both the development and the appica.

times greater than the U.S. MTBE. The report also points out ton of such models. Understanding of these uncertainties and significant differences in the Japanese approach to preventrve their causes is required to effectuely interpret model behavior.

maintenance, and in the Japanese regulatory approach to maim Many differer;t techniques have been proposed for performing tenance, their management and organizabonal context for rnain- uncertainty and sensitivity analyses The obt ectue of the tenance, and other socioeconomic factors that may affect the performance of maintenance. three models having large uncertainties and varying degrees of NUREG/CR-3885 V03: HIGH-TEMPERATURE GAS-COOLED RE- complexity in order to highhght some of the problem areas that ACTOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT must be addressed in actual apphcatons. The following ap-EVALUATION.Ouarterly Progress Report, July 1 - September proaches to uncartainty and sensstrvity analysis are considered.

30,1984. BALL.S.J.; CLEVELAND J.C.; HARRINGTON,R.M.; et (1) response surface methodology based on input determined at Oak R dge Nabonal Laboratory. Apnl 1985. 28pp. from a fractional factonal design (2) Latin hypercube sampling 8505230519. ORNL/TM-9267/V3. 30549.063 with and without regression analysis, and (3) differential analy-Modeling and code development work on the modular High- sis. These techniques are compared on the basis of (1) ease of Temperature Gas-Cooled Reactor (HTGR) were conbnued. The implementaton, (2) flexibility, (3) estimaton of the cumulative longer term heatup accident scenano in which cavity waft cool- distnbuten functon of the output, and (4) adaptability to differ-ing is lost was also modeled. Sensitivity studies were run for a ent methods of sensstrvity analysis. With respect to these crite-variety of parameter vanatons. Fission-product (FP) release and ria, the technique using Latin hypercube sampling and regres-transport expenments were completed for several additional sion analysis gives the best results overalt The models used in elements _ "

NUREG/CR-3889: THE MODELING OF BWR CORE MELTDOWN for researchers to make compansons of other techniques with ACCIDENTS - FOR APPLICATION IN THE MELRPLMOD2 the results in this study.

COMPUTER CODE KOH.B R.; KIM,S.H.; TALEYARKHAN,R.; et al. Oak Rdge Natonal Laboratory. May 1985. 279pp. NUREG/CR 3905 V01 R1: SEQUENCE CODING AND S~ ARCH 8505160635. 30444:230. SYSTEM FOR LICENSEE EVENT REPORTS. User's Guide.

This, report summarizes improvements and modifcatons GREEN,N.M.; MAYS,G.T.; JOHNSON.M.P.; et al. Oak R4ge Na-made in the MELRPI computer code. A major difference be' tonal Laboratory. Apol 1985.164pp. 8505070182. ORNL/NSIC-tween tNs new, updated version of the code, called 223. 30216:139' MELRPl. MOD 2. and the one reported previously, concems the ~ " " - " " P" ***

inclusion of a model for the BWR emergency core cooling sys- senbal fw sa"fety and mhability analyses, especiauy analyses of tems (ECCS). This model and its computer implementaten, the trends and pattems. The licensee event reports (LERs) that are ECCRPl subroutine, account for various emergency injecten submitted to the Nuclear Regulatory Commission (NRC) by the modes, for both intact and rubbhzed geometnes. Other changes nuclear power plant ublibes contain much of this data. The to MELRPI deal with an improved model for canister waff oxida- NRC's Offee of Analysis and Evaluaten of Operabonal Data tion, rubble bed modeling, and numencal integraton cf system (AEOD) has developed, under contract with NOAC, a system for 1 equatons. A complete documentaten of the enbre codifying the events reported in the LERs. The primary objective J MELRPl. MOD 2 code is also given, including an input guide, list of subroutines, sample input / output and program listing.

of the Sequence Coding and Search System (SCSS) is to l reduce the desenptive text of the LERs to coded sequences NUREG/CR-3900 V03: LONG-TERM PERF09MANCE OF MATE- ** " ** * '"" ***

• RIALS USED FOR HIGH LEVEL WASTE PACKACING.Quarterfy systm Wes a stmetumd famat tw maded codng of com-Report, October-December 1984. STAHL,0.; MILLER,N E. Bat- pen sysWn, aM und ems as wd as pesonM erms.

telle Memonal Institute, Columbus Laboratories. May 1985. This four volume report documents and Cscnbes SCSS in 88pp. 8506060746. 30776:173' detail. Volume 1 is a User's Guide for searching the SCSS data-Expenments for evaluating the glass-dessolubon model are base. Chapter 2 of this guide is a tutonal on retneving. display.

underway, and the procedure developed last quarter for dispers- ing, and analyzing LERs and provides hands.on expen.ence in ing RuO(2) in MCC 76-68 glass has been tested and proved to executing basic commands. Volume 2 contains all valid and ac-produce appropnate partcle concentratons. Acetic and humic ceptable codes used for searching and encoding the LER data.

acids have been chosen to test the effect of natural organic Volumes 3 and 4 provide a technical processor, new to SCSS, acids on waste glass performance. In the overpack-corrosion the informaton and methodology necessary to capture desenp-effort, potentodyname polanzation tests indicate that of the 15 tive data from the LER and to codify that data into a structured I chemmal species tested, aff but perchloate and hydrogen may format and serve as reference matenal for the more expen-effect stress-corrosion cracking behavior of carbon steel, sever.

enced technical processor, and contains informaton that is es-al synergiste effects were also indicated. In slow strasn rate studies, specimens tested in 0.0005 M FeC1(3) (a much lower sential for the more advanced user who needs to be famdiar chloride concentration than expected in groundwater) exhibited with the intreate coding techniques in order to retneve specifc details in a sequence.

signifcant cracking over the temperature range 250-315 de-grees cenbgrade. Pits were found to propagate readily, but NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH l slowfy, in 1018 carbon steel exposed to aerated basalt ground- SYSTEM FOR LICENSEE EVENT REPORTS Code Listings.

water at 90 degrees centigrade. The general-corrosen correla. GALLAHER,R.B.; GUYMON,R.H.; MAYS G T.; et al. Oak Rdge bon was changed to incorporate a finite rate of film growth. Inte-Nabonal Laboratory. Apnl 1985. 27f pp. 8505070170. ORNL/

gral expenments are being prepared to provide information on combined-effects processes that may influence the long. term NSIC-223. 30214.022.

See NUREG/CR-3905,V01 abs'ract.

performance of the waste package.

18 Main Citations and Abstracts NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH results are compared with those obtained for cesium, iodine, SYSTEM FOR LICENSEE EVENT REPORTS Code Listings. and tellunum in 26 tests of irradiated fuel and other tests using GALLAHER.R.B.; GUYMON.R.H.; MAYS.G.T.; et al. Oak Ridge tracers. In good agreement with the LWR fuel tests, the Csl be-Natonal Laboratory. Apnl 1985. 271pp. 8505070170. ORNL/

NSIC-223. 30214:022. havior in the control tests was similar to that observed for sodine See NUREG/CR-3905,V01 abstract. in the fuel tests; iodine was released pnmanly as Csl rather than highly volatile molecular iodine. Cesium (not associated with NUREG/CR 3905 V03: SEQUENCE CODING AND SEARCH Cst) behaved like CsOH in the LWR fuel tests. In both LWR fuel

$YSTEM FOR LICENSEE EVENT REPORTS. Coder's Manual. tests and the control tests, cesium hydroxide was observed to GALLAHER,R.B.; GUYMON.R.H.; MAYS.G.T.; et al. Oak Ridge react with and be retained by zirconia ceramic surfaces in the National Laboratory. Apnl 1985. 381pp. 8505070006. ORNL/ temperature range 800 to 1200 degrees centigrade, probably NSIC-223. 30213.001, forming cesium metazirconate (Cs(2)2rO(3). In one of the con-See NUREG/CR-3905,V01 abstract.

trol tests, cesium hydroxide reacted with tellurium in the gas NUREG/CR 3905 V04: SEQUENCE CODING AND SEARCH phase and was collected as CsTe. Although the results are hm.

SYSTEM FOR LICENSEE EVENT REPORTS Coder's Manual. ited at this time, the mdecated collected behavior of teelunum in GALLAHER,R.B.; GUYMON,R.H.; MAYS.G.T.; et al. Oak Ridge the LWR fuel tests has been that of a tellunde.

Na na ra . Apnl 1985. 347pp. 8505070184. ORNL/

NUREG/CR-3944: TRAN B-3 EXPERIMENTAL INVESTIGATION See NUREG/CR-3505,V01 abstract OF FUEL CRUST STABILITY ON MELTING SURFACES OF AN NUREG/CR-3906:

ANNULAR FLOW CHANNEL MCARTHUR.D A.; MAST,P.K.

URANIUM MILL TAILINGS Sandia National Laboratones. Apnl 1985. 61pp. 8505030229.

NEUTRALIZATION: CONTAMINANT COMPLEXATION AND SAND 84-1646. 30160:302.

TAILINGS LEACHING STUDY. OPITZ,B.E.; DODSON.M.E.;

SERNE,R.J. Battelle Memonal Institute, Pacife No:thwest Lab- The TRAN B senes of expenments is being conducted at oratones. May 1985. 77pp. 8506190026. PNL-5179. 31019:119. Sandia National Laboratones to investigate the charactenstics Laboratory expenments were performed to compare the ef- of fuel removal / freezing through the upper axial blankets of an fectiveness of limestone and hydrated hme for improving waste LMFBR dunng the transition phase of a hypothetical core dis-water quahty through the neutralization of acidic uranium mill ruptive accident. The third expenment in this senes, TRAN B-3, tailings hquor. The expenments were designed to assess the ef- was erformed in February 1984, and the results are reported fects of three proposed mechanisms - carbonate complexa- herein. This expenment involved the inI'ecten of molten UO(2) tion, elevated pH and colloidal particle adsorpton - on the sol. into an annular flow channel. Unlike the similar TRAN B-1 ex.

ubility of toxic contaminants found in a typical uranium mill penment, the initial steel wall temperature in B 3 was sufficiently waste solution. Of special interest were the effects of each of high that instantaneous steel melting would occur upon contact these possible mechanisms on the solution concentratens of with molten fuel. The earlier TRAN B 1 results had shown that trace metals such as Cd, Co Mo, Zn and U after neutralization. fuel crusts are mitially stable, both on the inside of a steel tube Acidic untreated solid tailings from two mill sites and taihngs as well as on tho outside of a steel rod, when no steel melting neutralized with lime were leached with a laboratory. prepared occurred. TR AN 0-3 was designed to investigate this queston ground water for savarmi prwa dWacement ve'umes. Ard/scs Of cruct eti:,tj - "--- " cppea;ts curvatu,6 wb nudaw performed on the column effluents indcate that pnor neutraliza- melting did occur, tion results in a significant reducten in the concentraton of all pH dependent consttuents in the column effluents. In contrast, NUREG/CR-3953: THE USE OF MAG-1 SPECTACLES WITH relatwely high concentratens of several trace metals and macro POSITIVE AND NEGATIVE-PRESSURE RESPIRATORS.

ions were found in effluent sofuten from the untreated tailings REED,K.A.; MOORE,T.O. Los Alamos Scientific Laboratory. May columns.

1985. 41pp. 8506060798. LA-10229-MS. 30775:198.

NUREG/CR 3913: HECTR VERSION 1.0 USER'S M ANUAL Results of testing conducted at Los Alamos National Labora-CAMP,A.L; WESTER.M.J.; CINGMAN,S E. Sandia National Lab- tory, Personnel Protection Studies Section, using MAG 1 spec-oratories. Apnl 1985. 325pp. 8504160098. SAND 84-1522. tacles in conjunction with positive- and negative-pressure full-29832:026. facepiece respirators, are reported. The purpose of the three-This report desenbes the features and use of HECTR Verson phase study was to determine if the specialty constructed strap 1.0. HECTR is a relatively fast-running, lumped-volume contain. of the MAG 1s affected the protection factors (PFs) of the res-ment analysis computer program that is most useful for per. pirators of the cyhnder hfe of selected self-contained breathing forming parametnc studies. The main purpose of HECTR is to apparatus (SCBA). The following respirators were tested with anatyre nuclear reactor accidents involving the transport and the MAG-1s: a) Phases I and it, positwe-pressure full facepiece:

combustion of hydrogen, but HECTR can also function as an Presur Pak 11 SCBA (pressure-demand) Scottoramic facepiece, expenment analysis tool and can solve a limited set of other types of containment problems. HECTR Version 1.0 has been MSA 401 Air Mask Ultravue facepiece (medium) Surv vair pres-particularly tailored to anatyze accidents in ice-condenser PWR sure-demand SCBA/sihcone futl facepiece, MSA powered air-and Mark 111 BWR containments. HECTR is designed for flexibel-punfying respirator /Ultravue facepiece (medium); b) Phase til, ity and provides for user control of many important parameters, negative-pressure full facepiece: MSA Ult avue (small, medium, particularfy those related to hydrogen combustion. Built in corre- large), MSA Ultra-twin (small, medium, large) Norton Senes lations and defauft values of key parameters are also provided. 7600 (one size only). Statistical analysis and review of the test data from Phases I and ll indicated httle, if any, vanaten with NUREG/CR-3930: OBSERVED BEHAVIOR OF and without the MAG-1s with most protecten factors greater CESIUM,10 DINE,AND TELLURIUM IN THE ORNL FISSION than 10,000. Test data also indicated little, if any, difference in PRODUCT RELEASE PROGRAM. COLLINS.J L; the cyhnder hfe with and without the MAG-Is, except the Scott OSBORNE.M.F.: LORENZ,R.A.; et al. Oak Ridge Natonal Labo- Presur-Pak 11 SCBA used with the Scottoramic facepiece. States-ratory. Apnl 1985. 73pp. 8504170679. ORNL/TM-9316. tical analysis of the quantitatrve fit test data indicated no differ.

29907:154.

ence in PFs for the negative-pressure devices for the Ultravue Two control tests were conducted to study the behavior of Cst, CsOH, and Te in the expenmental apparatus used to con- negatwe-pressure resp rator, but a significance at the 0.05 and duct fisson product release tests with highly irradiated LWR fuel 0.01 levels for the Ultra-twin and Norton full facep+ecos, respec-tively.

at ORNL in this report the control tests are desenbed, and the

Main Citations and Abstracts 19 NUREG/CR-3977

RELAPS THERMAL-HYDRAULIC ANALYSES can be closed out for 117 (92%) of the 127 current facilities on OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR H.B. the basis of specife entena Followup items for the remaining ROBINSON UNIT 2 PRESSURIZED WATER REACTOR. 10 current facihties are proposed for use by NRC/IE. Incorrect FLETCHER,C.D.; BOLANDER,M.A.; WATERMAN.M.E.; et al. weights reported for valves other than Velan swing check l EGaG, Inc. Apnl 1985. 233pp. 8506140624. EGG-2341. valves are identified as Remaining Areas of Concern. This bulle-l 30934:352. tin has served its purpose and can be closed out. A final check 4 Thermal-hydraulc analyses of fourteen hypothetical pressur- of valve weights will be made per later IE Bulletin 79-14 on seis-i ized thermal shock (PTS) scenarios for the H. B. Robinson, Unit mic analyses for as-budt safety-related piping systems.

2 pressurized water reactor were performed at the Idaho Na-tional Engineenng Laboratory (INEL) using the RELAP5 comput- NUREG/CR-4004: CLOSEOUT OF IE BULLETIN 79-25. FAIL-

~

er code. The scenanos, which were developed at Oak Ridge URES OF WESTINGHOUSE BFD RELAYS IN SAFETY-RELAT-Natonal Laboratory (ORNL), contain significant conservatsms ED SYSTEMS FOLEY,WJ.; DEAN.R.S.; HENNICK,A. Parame-i concerning equipment fadures, operator actions, or both. The ter, Inc. Apnl 1985. 44pp. 8505010088. IEB-79-25. 30112:260.

results of the thermal-hydra'.ilic analyses presented here, along Robinson 2 submitted LER 78 29 Decer9ber 19,1978 to i

j with additonal analyses of multidimensional and fracture me- report sbcking of a normally energized Wesbnghouse BFD relay.

4 chanics effects, will be utthzed by ORNL, integrator of the PTS After reviewing this problem, Westinghouse issued Service i study, to assist the U.S. Nuclear Regulatory Commission in re- Letter TS-E-412 to recommend that BFD relays in safety-related solving the pressunzed thermal shock unresolved safety issue. systems be replaced with later NBFD relays. Dunng installaten and testing of the new NBFD relays, Robnson 2 fourd some NUREG/CR-3980 V03: LIGHT WATER-REACTOR SAFETY FUEL with marginal or ur* satisfactory armature overtravel. Because of r SYSTEMS RESEARCH PROGRAMS. Quarter!y Progress this new problem, Westinghouse issued Techncal Bulletin NSD-ReportJufy-September 1984. REST.J. Argonne National Labo- TB-79-05 to recommend prompt checking of certain models of ratory. May 1985. 51pp. 8507050429. ANL-84-61. 31371:306. NBFD relays and returning those with anadequate overtravel for This progress report summanzes the Argonne National Labo- rework or replacement. IE Bulletin 79-25, with extracts of the ratory work pe-formed dunng July, August, and September 1984 on water reactor safety problems related to fuel and fuel clad- Westinghouse service letter and technical bulletin enclosed, was issued November 2,1979 to require responses and specife

! ding matenals. The research and development areas covered are Transient Fuel Response and Fission Product Release and actions by all licensees and holders of constructen permits with 1

Clad Properttes for Code Venficaton. respect to BFD and NBFD relays in safety-related systems.

Evaluation of utikty responses and NRC/IE inspection reports NUREG/CR-3987: COMPUTERIZED ANNUNCIATOR SYSTEMS. indicates that the bulletin can be closed out for 121 (94%) of RANKIN W.L; RIDEOUT.T.B.; TRIGGS,TJ.; et al. Battelle the 129 current facihties on the basis of specife enteria. Pro-Human Affairs Research Centers. June 1985. 102pp. posed followup items for the remaining 8 facihties are presented j 8507050435. PNL-5158. 31371:001. in Appendix C for use by NRC/IE. Because followup of correc-i This report presents the design philosophy and associated t ye acten is ensured, IE Bulletin 79-25 is considered closed.

l functional entena and design pnnciples for developing advanced computenzed annunciator systems for use in the control rooms NUREG/CR-4005: CLOSEOUT OF IE BULLETIN 80-12. DECAY

of nuclear power plants. The scope of the work includes ad- HEAT REMOVAL SYSTEM OPERABILITY. FOLEY,WJ.;

vanceo system recommendatens tnat couid oe incorporateo Ut- AN.H.b ; Ht

NNICK A. Parameter, Inc. June tW3, tspp.

into operating nuclear power plants. The information contained 8507030675. IE-146. 31314.098.

in this report was obtained from a review of the revelant com- On Apnl 19,1980, decay heat removal (DHR) capabhty was j puter and visual display terminal hterature, from site visits to ad- lost at Davis-Besse 1 for approximately two and one-half hours

, vanced control rooms in the nuclear power and related indus- in a refuehng mode. Typically for that mode, many systems and

[ tnes, and from a study of technical reports on computenzed components were out of service for maintenance and testing or control rooms. This report should assist the staff in develop- were deactivated to preclude inadvertent actuation. IE Bulletin ment of a regulatory position regarding the design of computer- 80-12 was issued May 8,1980 for acton by hcensees of operat-ized control room annunciator systems. The guidance in this ing pressurized water reactors (PWRs); it was issued for infor-report is consistent with that provided in NUhEG 0700. mation to nuc' ear power facihties other than operating PWRs.

The intent of the bulletin was to improve nuclear plant safety by NUREG/CR 3998 V02: LIGHT WATER-REACTOR SAFETY MA-TERIALS ENGINEERING RESEARCH PROGRAMS.Ouarterty reducing the likehhood of losing DHR capabhty in PWRs, espe-cially when some DHR components are unavailable because of Progress Report.Apnt-June 1984. SHACK,W.J. Argonne Naton-maintenance activities dunng refuehng and cold shutdown al Laboratory. Apnl 1985. 92pp. 8504220361. ANL-84-60. modes of operation. A related NRR Generic Letter was issued 29946:069.

This progress report summanzes the Argonne National Labo-June 11,1980 to hcensees of operating PWRs, requesting ratory work performed dunng Apol, May, and Jurse 1984 on amendment of technical specificatons to ensure long-term maintenance of DHR capabihty. Evaluaton of utshty responses water reactor safety problems related to out-of-core matenals.

The research and development areas covered are Environmen- and NRC/IE inspecton reports mdicates that the bulletin can be cfosed out per specirc cntena for 33 (75%) of the 44 affected i tally Assisted Cracking in Light Water Reactors, Long-Term Em.

bnttlement of Cast Duplex Stainless Steels in LWR Systems, facihties.

i and Nondestructive Evaluation and Leak Detection. NUREG/CR-4009: HUMAN RELIABILITY DATA BANK Evaluation NUREG/CR-4003: CLOSEOUT OF IE BULLETIN 79-04.INCOR- Results. COMER.M K.; DONOVAN.M,0.; GADDY,C D ; et al.

I. RECT WEIGHTS FOR SWING CHECK VALVES MANUFAC. General Physics Corp. Apnl 1985. 75pp. 8505070505. SAND 85-j TURED BY VELAN ENGINEERING CORPORATION. 7150. 30210:295.

FOLEY,W.J.; DEAN,RS.; HENNICK A. Parameter, Inc. June The U.S. Nuclear Regulatory Commisaion and Sandia Naton-t 1985. 29pp. 8507030702. IE 143. 31314:067, al Laboratories conducted a three-year research program to de-IE Bulletin 79-04 was issued March 30,1979 as a result of velop a human rehabhty data bank specifically tailored to sup-reports from three facihties that Velan Engineenng Corporaton port human reliabhty analysis segments of probabiliste risk as-had provided incorrect weights for swing check valves. The sessments for nuclear power plants. Prevous efforts of the pro-reason for concem was the possibhty that these incorrect gram include a review of existing human performance data weights had been used in ana!yses of Seisme Category 1 piping banks (NUREG/CR-2744 Vol.1) and a concept and system de-

]

, systems at a large number of plants. Evaluaton of utihty re- senpton (NUREG/CR-2744, Vol. 2). Subsequent to the system sponses and NRC/IE inspecton reports shows that the bulletin desenption, a detailed specifcation for implementing the data l

4

- - , , . - - . -- , - , x , . . - , , - , ,, -- - , - - , - - . . - - - , - ,, -,v., .

20 Main Citations and Abstracts bank was developed. An evaluation of tNs specif>cahon was NUREG/CR-4031 V02: NEUTRON SPECTRAL CHARACTERIZA-conducted and is desenbed in this report. The evaluaten con- TION FOR THE FIFTH HEAVY SECTION STEEL TECHNOLO.  !

sisted pomanly of an Operabihty Demonstration and Evaluabon GY (HSST) IRRADIATION SERIES. "Neutronics Calculatons."

of the data processing procedures and personnel required, and WILLIAMS.L; REMEC.I.; KAM,F.B. Oak Ridge National Labora-a Data Rewew Demonstration and Evaluaten involving mem- tory. May 1985. 42pp. 8505170010. ORNL/TM-9423/V2.

bers of the potential user populaton. The conclusions of tNs 30484.001, study were used to modify and improve the detailed implemen- A senes of calculations has been completed to compute do-taten specifcation. The revtsed specificaton is pubbshed as simeter activaten in the Oak Ridge Research Reactor (ORR)

NUREG/CR-4010, and rt desenbes all the necessary matenals, HSST Simulator Expenment. A companson of calculated and personnel, procedures, definitens, and data taxonomies to im- expenmental results shows that calculatons underpredict do-piement the data bank. simeter actrvities on the average of about 15%. The C/E values indicated the now famahar tendency to become lower as more NUREG/CR-4010: SPECIFICATION OF A HUMAN RELfABILITY eron is penetrated. The dosimeters in front of the simulator (and DATA BANK FOR CONDUCTING HRA SEGMENTS OF PRAS beNnd the thermal shield) typically have C/E values of 0.9-1,0, FOR NUCLEAR POWER PLANTS. COMER.M.K.; while those at the back of the smulator have values of 0.7-0.85.

DONOVAN.M D. General Physics Corp

• Sandia National Lab- The calculations also show SNfted axial distnbuten relative to oratones. Apnl 1985. 400pp. 8505060499. SAND 85-7151. the measurements: the C/E values near the bottom of the core 30190 301. are about 15% to 20% higher than those near the top. TNs is TNs document specifies the personnel, resources, pohcies, probably due to a discrepancy in the axial power distnbution and procedures for implementing and operating a human reli- computed mth VIPOR/ VENTURE. The axial distnbution of fuel abihty data bank specrfically tailored to support human rehability obtained from the correlation in VIPOR could possibly be caus-anatysis (HRA) segments of probabilishc nsk assessments ing the power sNft, although tNs speculaton has Mt been ver6 (PRAs) for nuclear power plants. The report concludes a three- fred. There is also, perhaps, a slight tendency for the C/E year research program conducted by the U.S. Nuclear Regula- values to increase along the x (the coordeste parallel to the re-tory Commission and Sandia Natonal Laboratones. Previous ef- actor face) traverse from the reactor centerhne to a point near forts of the program include a reytew of existing human per- the core boundary; however, this vanaton is much less than in formance data banks (NUREG/CR-2744, Vol 1), a concept and the axial direction. The results obtained mth different dosi-

system c,esenpbon (NUREG/CR-2744, Vol. 2), and a peer eval- meters appear generalty to be reasonably consistent, except uation study (NUREG/CR-4009). TNs report specifies the ad- that the (46)Ti C/E values seem to be consistently lower than minestrative organization of the data bank functional groups and for the other dosimeters. The systematic nature of the discrep-their proposed interection. Detailed procedures are included ancies in these calculatons mil be adjusted by the least that specify how to process submitted data, how to classify and squares procedure to produce an accurate representabon of the store the data, and how to combine similar data when appropn- flux disti,bution.

ate. Included mtNn the report is the skeleton data manual, which is a prototype, hardcopy, data manual that would be used NUREG/CR-4031 V03: NEUTRON SPECTRAL CHARACTERIZA-to disseminate data to the user population. It desenbes the data TION FOR THE FIFTH HEAVY SECTION STEEL TECHNOLO-tassmy, piccedwan iw winway data vi intvioni, ami pie- GY (HSST) IRRADIATION SERIES. "Neutrun Empusure Fararn-sents several sample data retneval problems. Definitens are eters." REMEC,1.; STALLMANN,F.W.; KAM.F.B. Oak Ridge Na-supphed for all technical and behaooral terms used in the taxo- tonal Laboratory. May 1985. 3tpp. 8505t60647. ORNL/TM-nomic structure. As its name amphes, the theleton data manual 9423/V3. 30457:18t.

embodies the data manual in structure, but is void of empencal This is the third volume of a three-volume report which de-data. scnbes the simulator expenments of the fifth series of HSST ir-radiation expenments which are sponsored by the U.S. Nuclear NUREG/CR-4015: EFFECT OF STAINLESS STEEL WELD Regulatory Commisson (NRC). The purpose of these three vol-OVERLAY CLADDING ON THE STRUCTURAL INTEGRITY OF umes is to document, in detail, the expenmental and calcula-FLAWED STEEL PLATES IN BENDING SERIES 1. tonal methodology which mil be used in determining the neu-CORWIN,W.R : ROB NSON,G C.; NANSTAD,R.K.; et al. Oak tron-exposu o parameters for the fifth and subsequent senes of

( Ridge Natonal Laboratory. Apnl 1985.103pp. 8505230524. HSST irradiaton erperiments at ORNL. The methodolgy was ORNL/TM-9390. 30549:090. also used in the fourth ser>es of HSST irradiaton expenments The HSST stainless steel cladding evaiuatens were initiated and represents the current state-of-the-art procedures devel-I to study the interacton of stainless cladding eth flaws instiated oped in the Light Water Reactor Pressure Vessel Simulation in and propagating in base metal of reactor pressure vessels. A Project which is a part of NRC's Surveillance Dosametry Im-comphcating factor in understanding the role of stainless clad- provoment Program. The neutronexposure data from the fifth ding in this setting is its toughness as a functon of radiation and subsequent senes will be documented in a loose-leaf dose and fabncaten process. The initial phase of tNs study ad- NUREG/CR report as the data become available. In this dressed tNs queston by testing the response of specimens volume, the best estimates for the values and spatial distnbu-clad with single-mre submerged-arc weld overlay in vary:ng tion of fluence rate (0) (E t.0 MeV), fluence rate (0) (E 0.1 toughness levels. The tests completed under the initial phase of Mev), and displacements per atom per second (dpa/s) are de-this study indicate that cladding of moderate Charpy toughness tr ined using LSL M2, a least squares loganthmic spectrum has only hmited capabilities to stop running cracks. This was a ad .stment procedure with input values taken from dosimetry hmited set of expenments, and the upper and lower bounds of data from Vol.1 and neutronics calculations from Vol. 2. These cracks arrest capabihties are not yet determined. The fabncaten estimates have an overall uncertainty of less than 20% relative techniques employed for this first series of tests have resulted standard deviaton. TNs volume is essential to the metallurgist in conditions that have prevented close control of the stress for defining the irradiaton strategy to meet his objective (s).

state at pop-in of the hydrogen-charged EB welds. Consequent-ty, the arrnst toughness of the stainless cladding was not close- NUREG/CR 4033: THE ROLE OF PERSONAL AIR SAMPLING IN ly bounded. General modificatens are proposed for incorpora. RADIATION SAFETY PROGRAMS AND RESULTS OF A LAB-ton in a second series of tests to provide more comprehensive ORATORY EVALUATION OF PERSONAL AIR SAMPLING conditions of testing and matenals of interest, to ehminate some EQUIPMENT. RITTER,P.D.; HUNTSMAN.B.L; NOVICK,V J; et undesirable test conditons that existed in the first senes, and to al. EGaG, Inc. May 1985. 80pp. 8505230534. EGG-2352.

provide an improved geometry for analytical interpretatens. 30549 258.

i l

i l

Main Citations and Abstracts 21 Recommended appleabons for personal air samphng in NRC noding on predcted results and on computer running time. It heensee radiaton protecting programs are presented. The rec- was found that the overall sequence of events and the impor-ommendatons are based on performance tests of currently tant trends of the transient were predicted to be nearfy the l

' available samplers, a review of research and regulatory htera- same with both the fine-node and coarse-node models. There ture, and a survey of current heensee air samphng programs. were differences in the time-dependent behavior of the cold leg i The performance tests show that personal at samplers are accumulator injecton, and the predicted PCT for the coarse-available whch can provide a reliable, convenient means for node calculation was about 75 K less than that or the fine-node

! breathing-zone samphng of workers in pracbcally any work envi- calculaton. The higher PCT of the fine-node calculabon is attnb-ronment which might be encountered in the hcensee industnes. uted pnmanly to three-dimensonal flow effects in the core. The The research literature emphasized that eshmates of an individ-i complete (steady state plus transient) coarse-node calculation uars exposure may be greatly undereshmated if based on gen- required 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of CYBER 76 computer time compared to eral area at samples, as is common practce in current hcensee 68.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the fine node calculation, yielding an overall programs, due to the unpredictable vanabihty of airbome-actrvity factor of five decrease in running time. Thus, we conclude that concentrations in the worksite. A conclusion whch may be I d

] drawn from the hterature and from expenmental results is that in j trends are of concern, the loss of accuracy resulting from use of such a coarse-node model will normally be inconsequential oath zone sa p ): the n m ns e bly e t compared to the savings in resources that are realized. Howev-

, the arbome actvity to which a worker has been exposed (MPC er, if the objective of the analyses is the investigation of the ef.

h). Research concerning the apphcabihty of ar-samphng mess.

, urements for estimating intake, uptake, and internal dose was fects of multi-dimensional flows on clad temperatures, then a i detailed model is required.

also reviewed.

NUREG/CR-4035

A HIGHWAY ACCIDENT INVOLVING RADIO- NUMEG/CR-4051: ASSESSMENT OF JOB-RELATED EDUCA-

+ PHARMACEUTICALS NEAR BROOKHAVEN. MISSISSIPPI ON TIONAL QUALIFICATIONS FOR NUCLEAR POWER PLANT DECEMBER 3,1983. MOHR,P.B.; MONT,M E.; OPERATORS. SAARI,LM.; MELBER.B D.; WHITE,A S.; et al.

SCHWARTZ.M.W. Lawrence Livermore Nabonal Laboratory. Battelle Human Affats Research Centers. April 1985. 77pp.

Apnl 1985. 52pp. 8505070560. UCRL 53587. 30210:169. 8505010277, PNL 5303. 30114.352.

A rear-end collision occurred between a passenger automo- This report identifies job-related educational quahficabons for

, bile and a luggage trailer carrying 84 packages, 76 of which the nuclear power plant hcensed operator positions of reactor contained radiopharmaceutcals, on U.S. Highway 84 near Brookhaven, Mississippi on the aft 6 moon of December 3,1983. operator (RO), sener reactor operator (SRO), and shift superv6-The purpose of this report is to document the mechanical cir, sor (SS). The extent to which college engineenng curnculum l covers jot >related academic knowledge was assessed. The ap-

! cumstances of the accident, confrm the nature and quantity of 4 radioactrve matenals involved, and assess the nature of the proach used involved systematically companng college engi-l physical environment to whch the packages were exposed and neering programs to knowledge needed on the job by having j the response of the packages. The report consists of three subject matter experts in the filed of general and nuclear engi-major sectons. The frst deals with the nature and circum- neenng cumculurrt (1) assess the Coverage of specific academ-J Steneae of tha acedaat end hading = of feet The earnnd gn,a= ic knnwtadga idenhfed by a jnh anmiy== as necessary for li-l l an accounting and descnpton of the matenals involved and the Censed operators in existing College engineenng degree pro-l consequences of their exposure. The third gives an assessment grams, cnd (2) make judgments conceming levels of formal en-l and analyses of the mechanisms of damage and the conclusions geneering education necessary for apphcation of knowledge ori j which may be drawn from the invesbgaton- the job, based on job samples from a job analysis of activities under selected normal and emergency operating sequences.

NUREG/CR-4040: OPERATIONAL DECISIONMAKING AND ACTION SELECTION UNDER PSYCHOLOGICAL STRESS IN The major conclusions of the report are: a substantial amount NUCLEAR POWER PLANTS. GERTMAN,D.L; JENKINS.JP.; (approximately 2/3) of job-related academic knowledge is cov- ,

i HANEY,LN.; et al. EGaG Idaho, Inc. (subs. of EG&G, Inc.) ered in college engineenng cumculum; college engineenng cur- J l May 1985. 68pp. 8507020109. EGG-238/ 31313.236. riculum provides considerable matenal beyond that identified as

! An extensive review of hterature on indnrdual and group per- necessary for hcensed operators; higher level operator positons l l formance and decisonmaking under psychological stress was (SS relative to SRO, SRO relative to RO) were judged as need-I 1 conducted and summarized. Specific stress-related vanables ing higher levels of educaton to perform the job.

i relevant to reactor operation were pinpointed and incorporated

) in an expenment to assess the performance of reactor opera. NUREG/CR-4064: STRUCTURAL RESPONSE OF LARGE PENE-

! tors under psychological stress. The decisionmaking perform- TRATIONS AND CLOSURES FOR CONTAINMENT VESSELS ance of 24 reactor operators under diffenng levels of workload, SUBJECTED TO LOADINGS BEYOND DESIGN BASIS.

confheting informaton, and detail of available wntten procedures KULAK R.F. Argonne National Laboratory.

• Sandia National was assessed in terms of selecting emmediate, subsequent, and Laboratones. Apnl 1985. 109pp. 8505060514. ANL 84:41.

nonapphcable actions in response to 12 emergency scenanos 30193.019.

resulting from a severe seismic event at a pressurtzed water re* This report summarizes the analyses work performed by Ar.

t actor. Specific personahty charactenstics of the operators sug- gonne Natonal Laboratory on three representative nuclear gested by the literature to be related to performance under power plant penetrations for severe accident loads beyond the stress were assessed and correlated to decisionmaking under desigq basis conditions. These include analyses of an equip-l stress- ment hatch for a steel containment, a BWR Mark 11 drywell NUMEG/CR-4044: TRAC-PF1 LOCA CALCULATIONS USING head and a bellows connection. The objectives of the analyses l

FINE NODE AND COARSE-NODE INPUT MODELS. were to identify the methodology requwed to simulate the re-DOBRANICH.D.: BUXTON.LD.; WONG,C.N. Sandia National sponse of the penetratons and determine their leakage poten-Laboratories. May 1985. 86pp. 8506190042, SAND 84 2305. tial under severe accident loads. This report provides the details 31015:118. of the analytical methodology used and the results obtained TRAC.PF1 calculatons of a 200% cold-leg break LOCA have from the analyses.

! been completed for a UH1 plant using both fine-node (with 776 i mesh cells) and coarse-node (with 320 mesh cells) input NUREG/CR-4070 V03: BlVALVE FOULING OF NUCLEAR i

! models. This study was performed to determine the effect of i

I l

_. _ _ . . = _ _ _ __ _ _ _ _ _ .___

l l

1 22 Main Citations and Abstracts

) POWER PLANT SERVICE-WATER SYSTEMS. Factors That May drodyname and hydrostate types of seals, was modeled and Intensify The Safety Consequences Of Befouhng. tested. Extrusen tests were conducted to determine if seal ma-

)

HENAGER.C.H.; DALING,P.M.; JOHNSON.K.I. Battelle Memon- tenals could withstand predicted temperatures and pressures. A al Institute, Pacific Northwest Laboratores. Apnl 1985. 61pp. taper-face seat model was tested for seal stability under condi-4 8504220375. PNL 5300. 29946.202. tons when leaking water flashes to steam across the seal face.

This report desenbes the safety and econome consequences Test information was then used as the basis for a staten black-f of bivalve fouhng in raw-water systems at nuclear power plants. out analysis. Test results indcate a potential problem with an l The report hsts events that could cause a norfral fouhng situa- elastomer matenal used for O-nngs by a pump vendor, that 1 ton to become more entcal and desenbes scenanos in whch vendor is considenng a change in matenal specifcaton. Test bivalve fouling could cause unsafe or unwanted conditons such results also indicate a need for further research on the genenc as transients and stiutdowns. Several fouhng eventf that have issue of RCP seal integnty and its possible consideration for occurred at vanous nuclear plants are bneffy reviewed, and rec- designaton as an unresolved safety issue.

ommendatons are made to aid in the detection and control of bivalve fouhng. NUREG/CR-4079: ANALYTIC STUDIES PERTAINING TO STEAM NUREG/CR-4071: EXPLORATORY TREND AND PATTERN GENERATOR TUBE RUPTURE ACCIDENTS. KASHIWA.B.A.;

ANALYSIS FOR 1981 LICENSEE EVENT REPORT DATA. MJOLSNESS.R.C. Los Alamos Scientifc Laboratory. April 1985.

! HESTER.O.V.: GENTILLON.C.D. EG&G, Inc. Apnl 1985.215pp. 88 p 8 37 10307 MS 077 1.

8505280415. EGG-2362. 30601:072.

This report presents an overvew of the 1981 Sequence generator tube rupture (SGTR) accidents leads to the conclu-Coding and Search System (SCSS) data base that contains nu- sions that (1) flashing will not occur upstream of the tube rup-clear power plant operatonal data denved from Licensee Event ture, so that the flow will be resistance hmited rather than Reports (LERs) submitted to the United States Nuclear Regula, choked, (2) there is considerable potential for discharging the tory Commission. Both overall event reporting and events relat- pnmary fluid in the form of trucron-sized droplets, particularly ed to speofic components, subsystems, systems, and person- when the fluid discharges into a vapor cavity surrounding the nel are discussed. At all of these levels of information, software tube rupture, and (3) that the surrounding of the rupture site by

, is used to generate count data for contingency tables. Contin- water rather than vapor may be a means for preventing the for-

gency table analysis is the main tool for the trend and pattern maton of rmcron-sized droplets. The p esence or absence of 1' analysis. The tables pnmanly focus on faults associated with meron-sized droplets is considered to be a key issue for the vanous components and other stems of interest across different damage assessment of SGTR acodents because they are cur-plants. The abstracts and other SCSS informaton on the LERs rently thought to be the most hkely route for radioactive iodine i accounting for unusual counts in the tables were ex1msned to to be released to the atmosphere.

gain insights from the events.

NUREG/CR-4084: DRY SPENT FUEL STORAGE TEST PLAN f NUREG/CR-4075: DESIGNING PROTECTIVE COVERS FOR FOR DESTRUCTIVE FUEL BCD EXAMINATIONS. OLSEN.C.S.

URANIUM MILL TAILINGS PILES. A Review. BEEDLOW,P.A.; EG8G. Inc. Apol 1985. 41pp. 8504220369 EGG-2367 PARKER.G.B. Battelle Memonal insbtute, Pacifc Northwest 29946:164.

l Laboratones. May 1985. 29pp. 8505280093. PNL 5323. A tesbng program using eight commercial pressunzed water

30604

247. reacte and be6ng w:!ct rc ctor cpent fuct red: ws: conducted l This report reviews design considerations for protective to investigate their long-term stabelity under a vanety of possible covers for uranium mill tashngs impoundments. The role of pro- dry storage conditions. The objective of this report is to provide I tective covers in taihngs containment systems is discussed. the Nuclear Regulatory Commisson with information to confirm Factors affecting the loreterm stabihzation of taihngs (eroson, or estabhsh dry spent fuel storage hcensing positons for long-biotic intruson, and soil moisture) are summarized. Basic ele- term, low-temperature (250 degrees centigrade) spent fuel rod ments to be considered in design of all uranium taihngs covers behavior dunng dry storage and for radoactive contamination

, are presented, and then quantitative techniques for designing that might occur with spallation of ctadding crud. Sex of the eight site-speofc covers are reviewed.

commercial fuel rods will be destructively examined. This report NUREG/CR-4076: DETERMINATION OF COMPLIANCE WITH presents the test plan for the destructive examinatens.

I CRITERIA FOR FINAL TAILINGS DISPOSAL SITE RECLAMA.

TION. BEEDLOW,P.A.; CLINEJF.; FREEMAN.H.D.; et al. Bat- NUREG/CR-4086: TENSILE PROPERTIES OF IRRADIATED NU-telle Memonal Institute, Pacfic Northwest Laboratones. June CLEAR GRADE PRESSURE VESSEL WELDS FOR THE THIRD 1985. 48pp. 8507030712. PNL-5324. 31318:250. HSST IRRADIAMN SERIES MOWANM Oak Nge Na-J This report provides methods and procedures that can be honal hatory. May W85. 23pp. 82M29. @EMW

' used to vonfy comphance with Environmental Protecton Agency 9477. 30439.288.

(EPA) engineenng standards for uranium mill tailings disposal The Heavy Section Steel Technology (HSST) Program con-sites. EPA standards for radon emissions, long. term ssolaton, ducted a senes of experiments to investigate the effect of neu-and protecten of water quahty are discussed. Taihngs isolaton tron arradiaton on the fracture toughness of nuclear pressure technologies are reviewed Informaton the licensee needs to wssd mawals. N wWs of A 508 class 2 sM we exam-provide for the regulating agency to determine comphance is med in this Therd HSST Irra presented, as is the actual comphance enteria. cated according to 'earty',diaton (pre-1972) Senes. hghtwaterThe welds reactor weldwere fabI practice (i.e., coppbr-coated electrodes). As part of this study, i NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL tensile properties were measured after irradiaton to 2 to 10 x BEHAVIOR DURING STATION BLACKOUT. KITTMER.CA; 10(22) neutrons /m(2) (E 1 MeV) at temperatures between 250 WENSEL,R.G.; RHODES.D.B.; et al. Atomic Energy of Canada. and 290 degrees centigrade. Strength properties of all four Ltd. April 1985. 93pp. 8506170667, EGG-2365. 30979:290. welds increased with exposure to irradiation. Yield strength was A testing program designed to provide fundamental informa- more sensitive to irradiaton than was ultimate strength. Tensale i ton pertaining to the behavior of reactor coolant pump (RCP) ductihty was not affected signifcantly by exposure to irradiaton.

. shaft seals dunng a postulated nuclear power plant staten I blackout has been completed. The test plan was developed by NUREG/CR-4088: METHODS FOR ESTIMATING RADIOACTIVE ,

EG&G ldaho personnel at the Idaho Natonal Engineenng Labo- AND TOXIC AIRBORNE SOURCE TERMS FOR URAN!UM l 1

ratory (INEL) and performed at the Chalk River Nuclear Labora- MILLING OPERATIONS. HARTLEYJN.; GLISSMEYER,J A.; '

tory, Ontano, Canada, under auspces of the U.S. Nuclear Regu- HILL,0.F. Battelle Memorial institute, Paofc Northwest Labora.

l latory Commisson (NRC). One seal assembly, utilizing both hy- tories. June 1985. 69pp. 8507080192. PNL-5338. 31393.099.

- - - , - - - --t-- - - - -

-e- y w wr ,ur.- w,,y w 9w ,-.,,,r-,,-- - - , --o-w, - ---,w- -~,,---e- -- - w-

Main Citations and Abstracts 23 Pacife Northwest Laboratory, under contract to the U.S. Nu- any potential performance deficiencies or confhcts between tho l clear Regulatory Commission (NRC), identified and evaluated Operations (safety) and Secunty (safeguards) organizations.

methods for estimating radcactwe and toxic partculate and This examinaton included the impacts of coordination with off-gaseous airbome releases from uranium milling operations. site emergency response personnel. Dubes, responsibilities, op-Such methods need to be standardized so that all uranium mills timal methods, and procedural actons inherent in these interac-can provide adequate data for NRC evaluaton of potenba! envi- tions were explored.

ronmental impacts and of compliance with 10 CFR 20, 40 CFR 190, and 'he Natonal Environmental Policy Act. The general NUREG/CR-4095: TEST SERIES 2: SEISMIC-FRAGILITY TESTS method for calculahng source terms is to multiply together a OF NATURALLY-AGED CLASS IE EXIDE FHC 19 BATTERY l normalized emission rate, contaminant content, emisson control CELLS. BONZON.LL; HENTE D.B. Sandia Natonal Laborato-

! factor, and processing rate for each process being evaluated. nes.Apnl 1985.190pp. 8507020389. 31307.332.

This report desenbes the sources of airborne releases (ore stor. This report, the second in a test senes of an extensue seis-age area, ore crushing and grinding, ore processing, yellowcake mic research program, covers the testing of 10-year old lead-producton, and tailings impoundment) and the calculational pro. calcium Exide FHC-19 cells from the Calvert Cliffs Nuclear cedures for estimating radioactwe and toxc source terms. Ex. Power Station operated by the Baltimore Gas and Electnc Com-ample calculatons are provided. pany. The Exide cells were tested in two configuratens using a triaxial shake table: single-cell tests, both ngidly and loosely NUREG/CR-4091: THE EFFECT OF ALTERNATIVE AGING AND mounted; and multcell (three-cell) tests, mounted in a typical ACCIDENT SIMULATIONS ON POLYMER PROPERTIES- battery rack. A total of six electncally actwe cells was used in BUSTARD LD.; CHENION.J.; CARLIN.F.; et al. Sandia National the two different cell configuratens. None of the six cells failed Laboratones. June 1985.177pp. 8507050391. SAND 84-2291. in the first stage tests dunng the actual seismic test up to the 31373:139. 1.5 g ZPAs imposed. Subsequent discharge capacity tests The influence of accident irradiation, steam, and chemcal showed, however, that only two of the cells could delwer the spray exposures on the behavior of twenty-three age-preconde- accepted standard of 80% of their rated electncal capacity for 3 toned polymer sample sets (twenty-one different matenals) has hours. When two of the same cells were exposed to the second been invesbgated. The test program vaned the following condi- stage, higher g-level tests, both cells again provided nstantane-tions: 1. Accident simulatons of irradiaten and thermodyname ous uninterrupted power. Subsequent capacity tests showed (steam and chemeal spray) condibons were performed both se- both of these cells to have capacities well below the accepted

quentially and simultaneously; 2. Accident thermodyname standard of 80% Four of the cells were disassembled for ex-(steam and chemcal spray) exposures were performed both amination and metallurgical analyses. The examination showed with and without air present during the exposures; 3. Sequential the actue matenal on the positue plates was hard and cracked accident arradiations were performed both at 28 degrees cend- and that the positwe bus bar matenal was corroded and bnttle.

I grade and 70 degrees centigrade; 4. Age precondiboning was i performed both sequentially and simultaneously; 5. Sequential NUREG/CR-4096: TEST SERIES 3. SEISMIC FRAGIUTY TESTS I aging irradiabons were performed both at 27 degrees conf' grade OF NATURALLY-AGED CLASS 1E C&D LCU-13 BATTERY and 70 degrees centigrade; and 6. Sequential aging exposures CELLS. BONZON.LL.; HENTE.D.B. Sandia Natonal Laborato-were performed using two sequences: (1) thermal followed by nes.Apnl 1985.170pp. 8507020413. SAND 84-2629. 31309:159.

, irradiation and (2) irradiaton followed by thermal. This report This report, the third in a test senes of an extensive seismic i presents both general trends apphcable to a mapnty of the research program, covers the testing of 10-year old lead-calci-j tested matenals as well as specific results for each polymer. um C&D LCU-13 cells from the North Anna Nuclear Power Sta-The data base consists of ultimate tensile properties at the ton operated by the Verginia Electnc and Power Company. The completon of the accident exposure for three XLPO and XLPE. C&D cells were tested in two configuratens using a tnaxial fue EPR and EPDM, two CS?E (HYPALON), one CPE, one shake table: single-cell tests, both ngidly and loosely mounted; VAMAC, one polydiallytphtalate, and one PPS material. Report and multicell (three-cell) tests, mounted en a typical battery rack.

bend test results at completion of the accident exposures for A total of seven electnca:ly actwe cells was used in the two dif-two TEFZEL matenals and permanent set after compresson re- ferent cell configurations. None of the seven cells failed in the i suits for three EPR, one VAMAC, one BUNA N, one SILICONE, first stage tests dunng the actual seismic test up to the 1.5 g

and one VITON matenal are also presented. ZPAS imposed. Subsequent discharge capacity tests showed NUREG/CR-4092

ORNL CHARACTERIZATION OF HEAVY-SEC- that while these cells suffered some loss of discharge capacity, t TION STEEL TECHNOLOGY PROGRAM PLATES 01,02.AND all cells could deliver the accepted standard of 80% of their

03. STELZMAN W.J.; BERGGREN,R.G.; JONES.T.N. Oak Ridge rated electncal capacity for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. When two of the same National Laboratory. April 1985.176pp. 8506060376. ORNL/ cells were exposed to the second stage, higher g-level tests, i TM-9491. 30773:101, both cells again provided instantaneous uninterrupted power.

! Charpy V-notch impact, tensile, and drop-weight data are pre. Subsequent capacity tests showed both of these cells to have sented for three 305-mm-thick (12-in.) A 533 grade B class 1 capacities well below the accepted standard of 80% Four of ,

l steel plates. The effects of specimen size and onentation were the cells were disassembled for examinaton and metallurgical l examined as well as the vanation of properties between differ. analyses. The examinaten showed that all plates and separa- '

ent plate locatons and depths. Some observatons based on tors were in very good conditen.

data obtained from an instrumented Charpy testing machine are

, NUREG/CR 4097: TEST SERIES 4: SEISMIC-FRAGIUTY TESTS also presenM 1 OF NATURALLY AGED EXIDE EMP-13 BATTERY CELLS.

4 NUREG/CR-4093: SAFETY / SAFEGUARDS INTERACTIONS BONZON LL; HENTE.D.B. Sandia National Laboratories. Aprd i DURING SAFETY-RELATED EMERGENCIES AT NUCLEAR 1985.119pp. 8507020430. SAND 84-2630. 31309:321, POWER REACTOR FACILITIES. MOUL,D.A.; PILGRIM,M.K.; This report, the fourth in a test series of an extensve seismic i SCHWElZER.R.L; et al. Brookhaven National Laboratory. June research program, covers the testing of 27-year old lead anti-1985.400pp.8507020094. BNL-NUREG-51848. 31308:138. mony Exide EMP-13 cells from the recentty decommissioned This report contains an analysis of the safety / safeguards Shippingport Atomic Power Station. The Exide cells were tested interactions that could occur dunng safety-related emergencies in two configuratens using a trixial shake table: single-cell tests, at licensed nuclear power reactors, and the extent to which ngidly mounted; and multicell (fwe-cell) tests, mounted in a typi-these interactons are addressed in existing or proposed NRC cal battery rack. A total of nine electncally actrve cells was used guidance. The safety / safeguards interaction dunng a senes of in the two different cell configurations. None of the nine cells postulated emergencies was systemabcally examined to identify failed dunng the actual seismic tests when a range of ZPAs up I

- - - _ . -_l

l l

l I

24 Main Citations and Abstracts '

2 to 1.5 g was imposed. Subsequent discharge capacity tests of thermal transients were imposed on the vessel, which was pre-five of the cells showed, however, that none of the cells could heated to 290 degrees centigrade. Two episodes of crack prop-delwer the accepted standard of 80% of their rated electncal agation and arrest occurred. The thermal transients were in-capacity for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. In fact, none of the 5 cells could delwer duced by coolant at 29 degrees centigrade to 15 degrees cen-more than 33% capacity. Two of the seismically tested cells tigrade. Pressure transients were as high as 94.4 MPa. The ex-and one untested, low capacity well were disassembled for ex- perimental objectives were attained. The inhibiting effects of amination and rretalfurgical analyses. The inspection showed warm prestressing were definitely demonstrated. Crack propa-the cells to be in poor condition. The negatue plates in the vi- gaton was nearly pure cleavage, and arrest at 30K above the cinity of the bus connections were extremely weak, the positive onset of the Charpy upper shelf was expenenced in a positive buses were corroded and bnttle, negative and positwe actwe K(1) gradient and with K(l) 300 MPa square root m. Fracture-me-

material utilizaten was extremely uneven, and corrosen prod- chanics analysis of bnttle fracture based on small-specimen ucts littered the cells. toughness measurements was reasonably accurate. Flaw eval-NUREG/CR-4101

ASSAY OF LONG-LIVED RADIONUCLIDES IN uation by procedures of the ASME Boiler and Pressure Vessel LOW-LEVEL WASTES FROM POWER REACTORS. CLINE J.E.; Code conservatively predicted vessel failure, which did not NOYCE J.R.; COE,LJ.: et al. Science Applications Intemational occur.

Corp erty ience Appications Inc.). April 1985. 615pp.

NUREG/CR-4109: TRAC-PF1 ANALYSES OF POTENTIAL PRES-The 10 CFR Part 61 waste classificaten system includes sev- SURIZED-THERMAL SHOCK TRANSIENTS AT CALVERT eral nuclides whch are difficult to assay without expensive ra, CLIFFS / UNIT 1.A Combuston Engineering PWR.

diochemcal methods. In order for waste generators to classify SPRlGGS,G.D.; KOENIG.J.E.; SMITH.R.C. Los Alamos Scientifc wastes practically, NRC Staff has recommended the use of cor- Laboratory. April 1985. 355pp. 8504220382. LA-10321.MS.

, relation iactors to scale the diffcult-to-measure nuctedes with 29953:331.

nuclides which can be measured more easily (i.e., gamma emit. Los Alamos National Laboratory part cipated in a program to ters such as (60)Co or (137)Cs). In thw study Science Applica. assess the hsk of a pressunzed thermal shock (PTS) to the re-tions intemational Corporation (SAIC) performed complete ra. actor vessel dunng a postulated overcooling transient in a pres-diochemical assays for a!! the 10 CFR Part 61 waste classifca- surized water reactor (PWR). We provided 7 0 thermal-hydraulic

.f tion nuclides on over 100 samples. These data, along with analyses of three general accident categones: steamhne almost 800 other samples in the SAIC data base, were used to breaks, runaway-feedwater transients, and small-break loss-of-4 assess the vahdaty of correlaten factors suggested for use for coolant accidents. These postulated accidents included multiple j nuclear power plant wastes. Specific genenc correlaten factors operator and equipment failures. Results were provided to Oak i are recommended with other approaches to correlate nuclides Ridge National Laboratory (ORNL) who plan to determine the for whch genenc scaling factors are not defens,ble. probability of vessel failure and accident occurrence for an NUREG/CR-4105: AN ASSESSMENT OF THERMAL GRADIENT

" "**""9 * "9 TUBE RESULTS FROM THE Hi SERIES OF FISSION PROD- ansient ReacW MaV-,is %' @AML M M h re-UCT RELEASE TESTS. NORWOOD.K.S. Oak Ridge National s s se vwy sens at Mons

,mo.

Laboratory.

, o,, May 1985. 64pp. 8505230531. ORNL/TM-9506- of the plant. If the plant was inatially at hot-Zero power (com-

)

thArmal gradient tube was used to analyze fission product pared to full power), the decay heat was much less, which vapors released from fuel heated in the HI test series. Complete made it possible for the same accident initiator to produce sig- j depositen profiles were obtained for Cs,1, Ag. and Sb. The canW lows una te atures. Howw, rouMe op-cesium profiles were complex and probably were dominated by eratw actes may Muce me consequwe of ag of hse simulated accidents if the prescnbed pressure-temperature rela- ]

Cs-S-O compounds formed by release of sulfur from furnace tionships are followed, ceramics. The iodine profiles were simple, indicating that more than 99.5% of the released iodine behaved as a s%Ie nonvoia- NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER tile species, probably Cst. Mass transfer coeffcients for this species onto platinum were estimated to be 1.9 to 5.8 cm/s. EFFICIENCIES WHEN CHALLENGED WITH THERMAL- AND Silver was probabty released in elemental form, condensed to AIR-JET-GENERATED Dl-2-ETHYLHEXYL SEBECATE.DI an aerosol, and captured by filters. Antimony was released as ETHYLHEXYL PHTHALATE.AND SODIUM CHLORIDE.

KERSCHNER,H.F.; ETTINGER.H.J.; DEFIELD.J.D.; et al. Los the element and reacted rapidly with platinum (or gold) as it der-j posited. Antimony profiles were calculated a pnori with some Alamos Scientific Laboratory. Apnl 1985. 62pp. 8504300121.

LA-9985-MS. 30070:220.

j success. A method was developed for isolating tellurium plati-num and mixed fisson products in a form suitable for neutron Respirators fitted with high-efficiency partculate (HEPA) car-activation anatysis. The platinum samples were completely dis- tridge filten are designed to remove dust, fumes, mists, and air-solved in acid (HC1/HNO(3), and the tellunum was precipitated borne particulate radonuclides. If these filters are to be reused, on selenium carrier by reduction. Finally, tellurium was loaded a Quality Assurance (OA) program must be established to onto Dowex 1X-4 ion-exchange resin for activaten and analysis. ensure that filter efficiency remains greater than 99.97 per cent.

i Tellurium recovery was 88%, and the theoretcal sensitivity was The standard method for performing OA testing is to challenge 3 ng. the filter with a thermally generated aerosol of 0.3-m-diam di i ethylhexyl phthalate (DEHP). Because of potential toxicological

, NUREG/CR-4106: PRESSURIZED-THERMAL-SHOCK TEST OF and other problems associated with the use of monodisperse

• 6-IN.-THICK PRESSURE VESSELS.PTSE 1:Investigaton Of DEHP, an investigation to study measured filter effciencies on i Warm Prestressing And Upper Shelf Arrest. BRYAN.R.H.; an HEPA respirator filter populaton, using several recommend-BASS.B.R.; BOLT.S E.; et al. Oak Ridge National Laboratory. ed replacement aerosols, has been conducted. Aerosc's com-Apnl 1985. 288pp. 8506060815. ORNL-6135. 30770:024. pared in this study were thermally generated de-2-ethylexyl se-The first pressurized-thermal-shock test of a 148-mm-thick becate (DEHS), thermally generated DEHP, air-jet-generated steel pressure vessel with a 1-m-long flaw was performed to in- DEHS, and air-jet-generated salt (NaC1). The study also fo.

vestigate fracture behavior of a vessel under condatens relevant cused on determining compatibihty for parallet use of aerosols to a flawed nuclear reactor pressure vessel dunng an overcool- generated for respirator-fit testing for use in OA filter testing.

ing accident. The objectives were to observe crack arn,st and Results indcate that a polydisperse air-jet-generated aerosol of

!i stabihty on the ductile upper shelf and effects of warm pres- DEHS can substitute for thermally DEHP as a method of provid-i tressing on crack initiation. Three coordinated pressure and ing OA testing of HEPA respirator filters and that equipment t

,.-_mv_ , , . ,__ . _ ._ _ - - -.-,. .. -s__._-..m .

Main Citations and Abstracts 25 used in the study designed for respirator quantitative-fit testing radiation exposure-rate measurements using macro-R-meters,2) can easily be modified to perform this functon. beta-gamma measurements using Geger-Mueller tubes,3) wipe sts for surface contannaton, and 4) sod anahses W (22QRa NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER and omer (23@ daugNws. Mcatons My to have (2Wa EFFICIENCIES WHEN CHALLENGED WITH THERMAL- AND concentratens that exceed standards can be identified by AIR-JET-GENERATED Dl-2 ETHYLHEXYL SEBECATE,DI gamma-radiaton exposure rate measurements. Samples of sod ETHYLHEXYL PHTHALATE,AND SOOlUM CHLORIDE. a oths matenal from locaten showing devaW exposwe rams KERSCHNER,H.F.; ETTINGER,H.J.; DEFlELD,J D.; et al. Los Alamos Scientfac Laboratory. Apnl 1985. 62pp. 8504300121, can hn be anaWed W (226)Ra to &mnnm N boundanes

• '*'*** *'""d ' ~9*"*****""*"

LA-9985-MS. 30070:220 and wipe sample analyses can b'e used to determine whether Respirators fitted with bgh-efficiency particulate (HEPA) car- wanum concenkaWs excM standads W N M or e tndge filters are designed to remove dust, fumes, mists, and air- mmaW contaminaten.

borne particulate radonuclides. If these filters are to be reused, a Quality Assurance (OA) program must be established to NUREG/CR-4124: NDE OF STAINLESS STEEL AND ON-LINE ensure that filter efficiency remains greater than 99.97 per cent. LEAK MONITORING OF LWRS. Annual Report, October 1983 -

The standard method for performing OA testing is to challenge September 1984. KUPPERMAN,0.S.; CLAYTOR,T.N.;

the filter with a thermally generated aerosol of 0.3-m-diam dF2- PRINE,D.W. Argonne Natonal Laboratory. Apnl 1985. 39pp.

ethylhexyt phthalate (DEHP). Because of potential toxicological 8505060509. ANL-85-5. 30190:287.

and other problems associated with the use of monodisperse This progress report summanzes work performed by the Ar.

DEHP, an investgation to study measured filter efficiencies on gonne Natonal Laboratory and GARD, Inc. (Drvison of Cnam-an HEPA respirator filter population, using several recommend- berlain Mfg. Corp.) as subcontractor on NDE of stainless steel ed replacement aerosols, has been conducted. Aerosols com- and orbline monitonng of LWRs dunng the twelve months from pared in this study were thermally generated dF2-ethylexyl se- October 1983 to September 1984.

becate (DEHS), thermally generated DEHP, air-let-generated DEHS, and amjet-generated salt (NaC1). The study also fo- NUREG/CR-4124: NDE OF STAINLESS STEEL AND ON-LINE cused on determening compatibility for parallel use of aerosols LEAK MONITORING OF LWRS. Annual Report, October 1983 -

September 1984. KUPPERMAN.D.S.; CLAYTOR T.N.;

generated for respirator-fit testing for use in OA filter testing.

Results indicate that a polydisperse airjet-generated aerosol of PRINE,D.W. Argonne Natonal Labora:ory. Apnl 1985. 39pp.

DEHS can substtute for thermally DEHP as a method of provid. 8505060509. ANL-85-5. 30190:287.

ing OA testng of HEPA resperator fdters and that equipment This progress report summanzes work performed by the Ar-used in the study desgned for respirator quantitative fit testing gonne Natonal Laboratory and GARD, Inc. (Dnnsion of Cham-can easily be modified to perform this functon. berlain Mfg. Corp.) as subcontractor on NDE of stainless steel and on-line monitonng of LWRs dunng the twelve months from NUREG/CR-4114: VALENCE EFFECTS ON THE SORPTION OF Octotnr 1983 to September 1984.

NUCLIDES ON ROCKS AND MINERALS.II. MEYER,R E; ARNOLD,W D.; CASE,F.I. Oak Ridge National Laboratory. May NUREG/CR-4131: INVESTIGATION OF ALTERNATIVE MEANS 1985. 53pp. 8505210576. ORNL-6137. 30521:328. TO ACCOMPLISH THE GOALS OF BIENNIAL lON CHAMBER Estimaton of the rates of migration of nuchdes from nuclear CALtBRATION. CAMERON.J.R.; DEWERD,LA: GOETSCH,S J.;

weiv revusduries requires knuwiedge of the irneraction of 61 6L W4 con &M, Us of. Madion, WL Mai 1301 30pp.

l these nuclides with the components of the geological forma- 8506030101. 30707.220.

tons in the path of the migra: ion. These interactions will be de- The research desenbed in this report was performed to inves-pendent upon the valence state and speciaton of the nuchde. tigate the feasibehty of a mailed dosimetry system as an alterna-Expenments designed to measure interaction of multivalent nu- tive method of achieving the goals of the present U.S. Nuclear clides and minerals must therefore include study of the specia- Regulatory Commission requirement that ionizaten chambers ton of the nuchdes. An electrochemical method of valence used for cahbraton of cobalt-60 teletherapy units be cahbrated state control and solvent extraction analyses of the valence every two years. Both thermoluminescent dosimeters (TLD's) states were used to study a number of reactons of interest to and a dode detector unit was used in this study. A total of 20 HLW repositones. These include the reducten of Np(V) and hospitals in the states of Ilknois, Iowa and Wisconsin participat.

Tc(Vil) by crushed basalt and other minerals. For the reduction ed in a program in which this dosimetry package was sent to of Np(V) by basalt, the expenments indicate that the sorption of each insttuten on three separate occasons. The physicist, phy.

basalt increases with pH and that most of the Np is reduced to sician or chief technologist was asked to deliver 1.50 Gray (150 Np(IV) which is very difficult to remove from the basalt even if rads) to the device, assuming the device was equivalent in rade-orygenated tracer-free solution is added to the solution. For the ation adsorption charactenstics to human tissue. A treatment l

expenments with Tc(VII), the results are considerably more field size of 10cm by 10cm was chosen and the institution was comphcated. Expenments were instated to determine the solu- requested to use their chnical source-to-surface cistance. The bikty of Tc(IV) oxides. The resufts of these expenments are accuracy of the beam locahzation as indicated by the coinci-used to assess some of the techniques and methods currently dence of the light field with the radiation field was measured as used in safety analyses of proposed HLW repositones. well. The entenon for accuracy of dose dehvery was plus minus 3.0mm. Only two hospitals dunng the course of the study had NUREG/CR-4118: MONITORING METHODS FOR DETERMINA- both a disagreement of more than 3mm between the light field TION COMPLIANCE WITH DECOMMISSIONING CLEANUP and the radiation field. It is recommended that such a ma!!ed CRITERIA AT URANIUM RECOVERY SITES. DENHAM,D.H.- pac age s c ns e as an amnadve to N RATHBUN.L.A.; BARNES.M.G.; et al. Batteile Memonal Insti een tute, Pacific Northwest Laboratories. June 1985. 31pp.

en mal ca awn of MaM c ahs usM to caWam cobam Mehrapy sources.

l 8507030713. PNL-5361. 31314:036.

l Decommissioning of a uranium processing site requires radio- NUREG/CR-4134: REPOSITORY ENVIRONMENTAL PARAM-I logical surveys to: 1) identsfy buddings, equipment, and open ETERS RELEVANT TO ASSESSING THE PERFORMANCE OF l land areas that require cleanup; 2) venfy that cleanup oper- HIGH-LEVEL WASTE PACKAGES. CLAIRBORNE,H.C.;

ations have been successful; and 3) provide a record of the ra- CROFF.A G.; GRIESS.J.C.; et al. Oak Ridge National Laborato-diological condition of the site following cleanup. This report de- ry. May 1985.130pp. 8506130358. ORNL/TM-9522. 30867.350.

scribes the instruments, measurements, quality assurance, and This document provides specifications for a model/methodol-statistical procedures that can be used to perform pre-and post- ogy and approach that could be employed in determining post-cleanup surveys. The procedures desenbed include: 1) gamma- closure repository environmental parameters relevant to high-

26 Main Citations and Abstracts level waste package performance for the Basalt Waste Isolation failure rate changes is beyond the scope of this study. The ap-1 Project (BWIP). Guidance is provided on (1) the identity of the plicatens use average ccmponents unavailability equations cur-relevant repository environmental parameters (groundwater rently employed in PRAs to calculate the nsk aging sensitivity. A characteristics, temperature, radiation, and pressure), (2) the more exact calculation is possible by using unavailability equa-models/ methodologies employed to determine the parameters, tions that include the time-dependent charactenstics of compo-and (3) the input data base for the model/ methodologies. Sup- nent failure rates; however, these time-dependent charactens-

- porting studies included are (1) an analysis of potential waste tics are not well-known. The risk aging sens.tivity measure pre-package failure modes leading to identificaten of the relevant sented here is, therefore, segregated from these time-depend-

. repository environmental parameters, (2) an evaluaten of the ent effects and addresses only the time-independent portion of l credible range of the repository environmental parameters for aging phenomena. The resutts identify the component types i the BWIP situation, and (3) a summary review of existing that show the most potential for nsk change due to aging phe-i models/ methodologies currently employed in determirung re- nomena. Future research on the time-dependent portion of pository environmental parameters relevant to waste package aging phenomena for these component types is needed to com- l performance, pletely desenbe the nsk impact due to component aging

, NUREG/CR-4139: THE MAILED SURVEY:A TECHNIQUE FOR NUREG/CR-4144: IMPORTANCE RANKING BASED ON AGING OBTAINING FEEDBACK FROM OPERATIONS PERSONNEL CONSIDERATIONS OF COMPONENTS INCLUDED IN PROB-

MCGUIRE M.V.; WALSH.M.E.; MORISSEAU,D.S.; et al. Battelle ABILISTIC RISK ASSESSMENTS. DAVIS.T.; SHAFAGHl,A.;

i Human Affairs Research Centers. May 1985. 87pp. KURTZ,R.; et al. Battelle Memonal Institute, Pacific Northwest 8505100041. PNL-5381. 30270;125- Laboratories. Apol 1985 69pp. 8504220341. PNL-5389.

This report describes a mailed survey of operatens personnel 29946:001, at a sample of commercial nuclear power plants. The survey This report presents a method for focusing additional re-was conducted for the U.S. Nuclear Regulatory Comrnission search on aging phenomena that affects nuclear power plant (NRC) as part of the Operator Feedback Project. The survey components. Specifically, the method ranks components using 4

sought to collect informabon on topics of concern to the NRC a nsk aging sensitivity measure that descnbes the change in i and to assess the feasibility of a mailed survey on an informa- risk due to changes in component failure rate. Desenbing the tion collection mecharwsm. Participants in the survey were 520 aging phenomena and the resulting berm-dapendent component i personnet at 26 nuclear power plants representing all five NRC friture rate changes is beyond the scope of this study. The ap-regions. The indmdual participants completed and retumed by plications use average components unavailabihty equations cur-

] mail a ten-page questionnaire. This contained questons about rently employed in PRAs to calculate the nsk aging sensitivity. A operatons crew practices, including work and shift schedules, more exact calculation is possible by using unavai' ability equa.

4 operations shift crew staffing, the shift technscal advisor posi- tions that include the time-dependent Charactenstics of compo-tion, respondents' own backgrounds, the questennaire, and nent failure rates; however, these time-dependent charactens-

] other informaten-collecten techniques. Results of the survey j tics are not well-known. The nsk aging sensitety measure pre- i offer some insight on operatons crew practices at the plants sented here is, therefore, segregated from these time-depend-participating in the survey. Survey results also suggest that the ent effects and addresses only the time-independent porton of mailed survey i,s an informaton-collection technique that can be aging phenomena. The results identify the component types

, used effectively to obtain feedback for the NRC from operatens that show the most potent:3! for nck change due to aging phc j personnet nomena. Future research on the time-dependent porton of 1

NUREG/CR-4140: DOMINANT ACCIDENT SEQUENCES IN aging phenomena for these component types is needed to com- ,

OCONEE 1 PRESSURIZED WATER REACTOR. DEARING.J.F.; pletely desenbe the nsk impact due to component aging.

~} HENNINGER R.J; NASSERSHARIF,0.; et cl. Los Alamos Sci-entific Laboratory. Apnl 1985.112pp. 8506240647. LA-10351- NUREG/CR-4147 THE EFFECT OF ENVIRONMENTAL STRESS MS. 31149:230. ON SYLGARD 70 SILICONE ELASTOMER. BUCKALEW,W H.;

A set of dominant accident sequences in the Oconee-1 pres. WYANT F.J. Sandia Natonal Laboratones. May 1985. 92pp.

surized water reactor was selected useng probabilistic nsk analy. 8506240313. SAND 85-0209. 31157:001.

sis methods. Because some accident scenarios were similar, a Dow Corning Sylgard 170 Silicone Elastomer has been inves- .

Subset of four accident sequences was selected to be analyzed tigated to characterize its response to accelerated thermal '

with the Transient Reactor Analysis Code (TRAC) to further our aging, radiaton exposure, and its behavior under applied com-  !

insights into similar types of accidents. The sequences selected pressive forces. Sy1 gard 170 response to accelerated thermal l

! were loss-of-feedwater, small-small bresk loss-of-coolant, loss, aging suggests the matenal properties are not parhcularty age I of-feedwater-initiated transient without scram, and interfacing dependent. Radiaten exposures, however, produce significant, systems loss-of-coolant accidents. The normal plant response monotonic changes in both elongaton and hardness with in-and the impact of equipment availability and potential operator creasing absorbed radiaton dose. Elastomer response to an ap-

, actions were also examined. Strategies were developed for op. plied compressive force was strongly dependent on environ-erator actions not covered in existing emergency operator ac- ment temperature and degree of material confinement. Vari-tions not covered in existing emergency operator guidelines and ations in temperature produced large changes in compressrve

, were tested using TRAC simulatons to evaluate their effective, forces applied to confined samples. Attempts to mitigate force J

ness in preventang core uncovery and maintaining core cooling. fluctuations by means of pressure relief paths resulted in total I loss of the applied compressive force. Thus, seat applicatons NUREG/CR-4144: IMPORTANCE RANKING BASED ON AGING employing this elastomer in Class IE equipment required to CONSIDERATIONS OF COMPONENTS INCLUDED IN PROB- function dunng or following an accident should consider the po-ABILISTIC RISK ASSESSMENTS. DAVIS,T.; SHAFAGHI,A.; tential loss of compressive force from long term aging and po-KURTZ,R.; et al. Battelle Memonal Institute, Pacific Northwest tential LOCA-temperature transient conditions.

i Laboratories. Apnl 1955. 69pp. 8504220341. PNL-5389.

t 29946:001. NUREG/CR-4149: ULTIMATE PRESSURE CAPACITY OF REIN.

j This report presents a method for focusing additional re- FORCED AND PRESTRESSED CONCRETE CONTAINMENT, j search on aging phenomena that affects nuclear power plant SHARMA,S.; WANG,Y.K.; REICH.M. Broo6 haven National Labo-2 components. Specificaffy, the method ranks components using ratory. May 1985. 95pp. 8506130467. BNL-NUREG-51857.

a nsk aging sensitivity measure that describes the change in 30901:328.

4 risk due to changes in component failure rate. Desenbing the This report presents the results of an in-depth evaluaten of 1 aging phenomena and the resulting time-dependent component current modeling techniques and analysis procedures for deter.

4

[i 4

, - . , . - , . . . . , - - - , , . - ,e+- -- - ,em> --,----,.c_, n.--, .n e.- - - . - , . , - - , .

Main Citations and Abstracts 27 mining ultimate pressure capacity of reinforced and prestressed second part discusses the uncertainties involved in precipitaten concrete containments. The material models used for desenbing scavenging rates for effluents resulting from a nuclear reactor the nonlinear material behavior of concrete and steel are re- accident. The conclusion is that mafor uncertainties are involved viewed in detail. Special attention is focused on post-cracking both as a result of the natural vanability of the atmosphenc pre-behavior of concrete which controls one of the containment fail- capitaten process and due to our incomplete understanding of ure modes, i.e., the shear failure. Vanous finite element idealiza- the underlying process. The third part involves a review of the tons utilized for containment analysis are reviewed. The effects important problems associated with modeling the interacton be-of major assumptions pef;aining to containment geometry, ba- tween the atmosphere and a forest. It gives an indication of the semat restraint, firtte element mesh, rebar locations and onen- magnitude of the problem involved in modehng dry deposition in tations, are evaluated. Finally, failure analyses of two selected such environments.

reinforced and prestressed concrete containments are per-furmed and results are compared with those presented in the NUREG/CR-4159: COMPARISON OF THE 1981 INEL DISPER.

literature. SiON DATA WITH RESULTS FROM A NUMBER OF DIFFER-ENT MODELS. LEWELLEN,W.S ; SYKES,R l; PARKER,S.F.

NUREG/CR-4155: TRAC-PF1/ MOD 1 INDEPENDENT Aeronautscal Research Associates of Pnnceton. May 1985.

ASSESSMENT-NORTHWESTERN UNIVERSITY PERFORAT.

202pp.8505310421. ARAP NO. 505. 30670 001.

ED-PLATE CCFL TESTS. DOBRANICH,0. Sandia National Lab- Results from simulatons by 12 different dispersion models oratones. Aptd 1985. 42pp. 8505060503. SAND 85-0172- are compared with observations from an extensive field expen-3C190:241. ment at the Idaho Natonal Engineenng Laboratory in July, The TRAC-PF1/ MODI independent assessment project at 1981. Compansons were made based on hourty ground-level Sandia is part of an overall effort funded by the NRC to deter- SF(6) samples, out to approximately 10 km from the 46 m re-mine the ability of vanous systems codes to predict the detailed lease tower, both during and following 7 dittarent 8-hour re-thermal / hydraulic response of LWRs dunng accident and off- leases. Compansons are also made for total mtegrated doses normal conditons. As part of this effort, calculations for some of collected out to approximately 40 km. Within the limited range the Northwestern University perforated-plate CCFL tests have appropnate for Class A models this data companson shows that been performed. Two input models were constructed to repre* neither the puff models or the transport and diffusson models sent the rectangular test channel: a two. dimensional model and agree with the data any better then the simple Gaussian plume a faster-running one-dimensional model. The results of both models. The puff and transport and diffusion mode's do show a models indicate that for h!gh water flow rates, with the water in.

slight edge in performance in companson with the total dose jected vertically above a perforated plate, TRAC overpredicts over the extended range appropnate for Class B models. The the steam flow rate necessary for complete weeping (CCFL) best model results for the hourty samples show approximately However, for flow conditions more typical of PWR transieats, 40% calculated within a factor of two when a 15 degree uncer.

TRAC provides a reasonable predicton of weeping. tainty in plume position is permitted, and it is assumed that NUREG/C3-4155: TRAC-PF1/ MOD 1 INDEPENDENT higher data samples may occur at stations between the actual ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORAT- sample sites. This is increased to 60% for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> integrat-ED PLATE CCFL TESTS. DOBRANICH.D. Sandia National Lab. ed dose and 70% for the total integrated dose. None of the oratones. Apnl 1985. 42pp. 8505060503. SAND 85-0172. models reproduce the observed patchy dose pattems. This pat.

30!90:241, chinate anpaart to ha r'nnsintent with the inharent isncertaintv The TRAC-PF1/ MODI independent assessment project at associated with time averaged plume observations.

Sandia is part of an overall effort funded by the NRC to deter-mine the ability of various systems codes to predict the detailed NUREG/CR-4160: HISTORICAL

SUMMARY

OF OCCUPATIONAL thermal / hydraulic response of LWRs dunng accident and off. RADIATION EXPOSURE EXPERIENCE IN U.S. COMMERCIAL normal conditions. As part of this effort, calculatons for some of NUCLEAR POWER PLANTS. MOELLER.M P.; STOETZEL,G.A.;

the Northwestem University perforated-plate CCFL tests have MUNSON,L.H. Battelle Memonal Institute, Pacific Northwest been performed. Two input models were constructed to repre. Laboratones. Apnl 1985. 65pp. 8505070439. PNL 5404.

sent the rectangular test channel: a two-dimensional model and 30214 293.

a faster-running one-dimensional model. The results of both This report organizes existing data on occupational radiaton models indicate that for high water flow rates, with the water in- exposure expenence for consideraton in the safety goal evalua-jected vertically above a perforated plate TRAC overpredicts tion program. The report includes a review of occupational radi-I the steam flow rate necessary for complete weeping (CCFL). ation exposures incurred by workers at commercial U.S. nuclear However, for flow conditons more typical of PWR transients, power plants. In addition, occupatonal radiation exposure infor-TRAC provides a reasonable prediction of weeping. mation is presented for work performed at commercial U.S. nu-clear power plants to meet regulatory actions and required l

NUREG/CR-4158: A COMPILATION OF INFORMATION ON UN- backfits. This informaton identifies specife operatons per-CERTAINTIES INVOLVED IN DEPOSITION MODELING. formed as part of these requirements. Where possible, actual LEWELLEN,W.S ; VARMA A.K.; SHENG,Y.P. Aeronautical Re- radiation exposure histones are provided. A bnef history of radi-search Associates of Pnnceton. Apnl 1985.79pp.8504250168. aton dose limits and a review of the biologcal and health ef-ARAP NO. 504. 30032:015 fects attnbutable to radiation exposure is included to provide a The current generation of dispersion models contains very perspective on the development of radiation protection regula-l simple parametenzations of deposition processes. The analysis tions.

here looks at the physical mechanisims governing these proc-I esses in an attempt to see if more valid parametenzations are NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-available and what level of uncertainty is involved in either LEVEL WASTE REPOSITORY, Volume 1-Basalt. BINNALL,E.P.;

these simple parameterizatens or any more advanced parame- WOLLENBERG.H.A.; BENSON.S.M.; et at. Lawrence Berkeley tenzation. The report is composed of three parts. The first, on Laboratory. May 1985. 95PP. 8506060810. UCID 20092.

d y deposition model sensitivity, provides an estimate of the un- 30769.292.

certainty existing in current estimates of the deposition velocity This report addresses entcal parameters specific to a repose-due to uncertainties in independent vanables such as meteoro- tory in basalt, using the Columbia River Basalt Group as the logical stability, partcle size, surface chemical reactivity and principal example. For the purposes of this report, a parameter canopy structure. The range of uncertainty estimated for an ap- is considered to be a physical property whose value helps de-propriate dry deposition velocity for a plume generated by a nu- termine the characteristics or behavior of a repository system.

clear power plant accident is three orders of magnitude. The Parameters which are defined as critcal are those essential to l

l

28 Main Citations and Abstracts l

evaluate and/or monitor leakage of radonuchdes from the re- egonzed and desenbed, including h gh-efficiency and moderate-pository and to evaluate the need for retrieval. The parameters effciency devces as well as other (some novel) devees useful are considered with respect to the disciplines of geomechanics, in specife situatons. Preoperational consderatons discussed geology, hydrology, and geochemistry and are rank ordered in include selecting devces, instrumentaten, and testing pro-terms of importance. The specific role of each parameter, spe- grams. Operational and maintenance consderanons related to cifc factors affecting the measurement of each parameter, and dry and wet removal processes are desenbed. Quality assur-the interrelationships between the parameters are consdered in ance documents and topics are also discussed.

detal NUREG/CR-4164: GT2F:A COMPUTER CODE FOR ESTIMAT. NUREG/CR-4177 V01: MANAGEMCNT OF SEVERE ING LIGHT WATER REACTOR FUEL ROD FAILURES. ACCIDENTS. Perspectives On Managing Severe Accidents in WILLIFORD,R E.; LANNING,0.D.; BEYER,C.E. Battelle Memon. Commercial Nuclear Power Plants. DISALVO R.; LEONARD,M.;

al Institute, Pacific Northwest Laboratones. May 1985. 277pp. MANAHAN,M.; et al. Battelle Memonal Insttute, Columbus Lab-8506060567. PNL-5354. 30771:002, oratones. May 1985. 105pp. 8506130369. BMI-2123.

This report descnbes the development, benchmarking and re- 30868:176.

suits of a computer code designed to permit companson of Accident management is examined from several related per.

BWR and PWR fuel rod failure behaviors dunng postulated re- spectives. The emphasis is on the role of the operating crew actor off-normal events such as control rod withdrawal errors. and the techncal support provded to them before, dunng, and The code is called GT2F, and was developed from the after an accdent. The relabonship among accident manage-GAPCON. THERMAL-2 code by the additon of new models for ment, risk management and emergency management is exam-calculating transient temperatures, fisson gas release, mechani- ined. The roles played by industry, regulaton, and research are cal interacton between fuel and cladding, and Zircaloy cladding reviewed. Finally, the results of viewing accident management fracture behavior. Results indicate that for the conservatively from these varcus perspectives are reflected c the artculet'on severe overpower transient scenanos assumed, a full length of issues and some proposals for their resolut,ott commercial BWR fuel rod has a fadure probabihty between 1%

and 4.5% at 27 mwd /kgM when the transient begins from high NUR[G/CR-4177 V02: MANAGEMENT OF SEVERE operating power. A full length commercial PWR fuel rod has a ACCIDENTS. Extending Plant Operating Procedures into The failure probability between 2% and 11% at 28 mwd /kgM when Severe Accident Regeme. WREATHALL,J.; LEONARD,M.;

the transient begins from low power. Failure probabhties are DISALVO.R. Battelle Memonal Insttute, Columbus Laboratones.

substantially smaller at lower bumups and for less extreme tran- May 1985. 75pp. 8506130132. BMI-2123. 30867;108.

sient conditions. This study examenes the feasibility and value/ impact of ex-NUREG/CR-4169: AN APPROACH TO TREATING RADIONU. tending emergency operahng procedures into the severe acci-CLIDE DECAY HEATING FOR USE IN THE MELCOR CODE dent regime. It reviews the types of knowledge needed to devel-SYSTEM. OSTMEYER,R.M. Sandia National Laboratones. June op such procedures and the appicabihty of exiseng regulatory 1985. 33pp. 8507050426. SANO84-1404. 31371:175. review cnteria. A method is developed and illustrated in two A new code system is being developed for use in assessment cases. This study concludes that it is feasible to develop proce-of nuclear reactor accident risks. The code system, termed dures for operators to mibgate the consequences of accdents MELCOR. will treat thermal-hvdraulic and fission prnduct tmhav- Drogressing past the onnat nf etwa damage A prolemmary ior jointly. A part of its treatment of thermal-hydrautic processes, valuehmpact assessment indcates a signifcant likehhood of the code system will evaluate decay heatng from fission prod- there being an overall net positrve betwfit of develop;ng mitiga- '

uct inventones contained within the reactor core debns and trve procedures. A phased program has been proposed. First a compartments that are defined for the reactor system and con- pilot study should develop the appication of the methods used tainment. A simple approach to treating radonucide decay in this feassbehty study and provde more precise informahon for heating is proposed for use in MELCOR. The proposed ap- a detailed value-impact assessment. Based on the results of the proach uses a table-lookup to estimate element decay powers pilot study, extension to a greater populaton of plants may be as a functon of time after reactor shutdown (start of accident). jushfied.

Decay power for each element in a compartment of the reactor i

system is found by muttsplying the mass of the element in the NUMEG/CR-4100: STATE-OF THE ART OF SOLID-STATE compartment by the element's decay-heat rate per unit mass MOTOR CONTROLLERS. JAROSS,RA; MULCAHEY,T.P.;

which is a functon of time after reactor scram. The approach KOEHL.EA Argonne Natonal Laboratory. Apnl 1985.118pp. j assumes that daughter products are transported along with the 8504180201 ANL-84102. 29921:264.

parent radionuclide dunng the accident. The valdsty of this as- The state-of.the-art of soidstate motor controllers (SSMCs) sumpton is discussed. In additon, methods for apportening the is assessed in terms of use, probabbty of Class 1E quahfcation, decay energy between the walls and the gases in a compart- l failure rate expenence, and rehabbty predictort Surveys of ment are also discussed. The proposed approach is based on commercial availabihty, nuclear and nonnuclear electric uthty SANDIA-ORIGEN calculabons for a 3412 MWt PWR and a 3578 expenence, and arch tect.engineenng use were made relative to MW1 BWR. the suitabehty of SSMCs for nuclear sennce. Reasons for the NUREG/CR-4176: EMISSION CONTROL TECHNOLOGY AND limited use of SSMCs in nuclear plants are given. Available fail- j OUALITY ASSURANCE NEEDS AT URANIUM MILLING ure rate data are meager, and are augmented by data on other i FACILITIES. Includes Supporting Methods For soldstate power electronc devices that are shown to have Testing.Operahng.And Maintaining Air Polluten Control Devees. subcomponents similar to those found in SSMCs. In addition to LUDWICK.J.D. Battelle Memonal Institute, Pacifc Northwest large nonnuclear soldstate adlustable-speed motor drives, the Laboratories. June 1985. 55pp. 8507030684. PNL-5386. rehabihty of nuclear plant inverter systems and high-voltage ,

31318:311. sohd. state DC transtmsson hne converters is assessed. Class  !

Pacife Northwest Laboratory, under contract to the U.S. Nu- 1E environmental quakfcaten expenence with nuclear plant I citar Regulatory Commisson, conducted an investigabon of converterhnverters and battery chargers in shown to be directty partculate emisson control devces for appication to process apphcable to SSMCs. No problems are expected in quahfying Cxhausts at uranium milhng facihties. The scope of this invesh. them. Actual rehabhty predictions of two typcal commercial 1 gabon included devces now in use, as well as those devees SSMCs are Diven, together with predctons of improvements I that have potential apphcation for milkng sites. This report pre- possible with use of high.quahty parts and manufacturing proce-sents the results of the study. Emisson control devees are cat- dures.

I i

l

Main Citations and Abstracts 29 NUREG/CR-4181: LEACHABILITY OF RADIONUCLIDES FROM 1986. The other two plants are currently operating, having re-CEMENT SOLIDIFIED WASTE FORMS PRODLCED AT OPER- ceived operating licenses in the mid 1970's and earty 1980's.

ATING NUCLEAR POWER REACTORS. CRONEY,S T. EG&G The major finding of this survey is that despite the fact that the Idaho, Inc. (subs. of EG&G, Inc ) . Apol 1985. 130pp. latest control-room-habitabihty systems have become large and 8505070173. EGG-2355. 30217:161. more complex than earlier systems surveyed, the latest systems Different sized samples of cement-solidified hqud wastes do not appear to be functonally supenor. The maior recommon-were collected from two nuclear power plants, a pressunzed dation of this report is to consoldate into a single NRC docu-water reactor (PWR) and a boiling water reactor (BWR), to cor- ment, based upon a comprehensive systems engineenng ap-relate radionuchde leaching from small and full sized waste proach, the pertinent cntena for control-room-habitabihty design.

forms. Diffusion-based model analysis of measured radionuchdo teach data from small samples and full size samples indicated NUREG/CR-4192: THE ANALYSIS OF DRAINAGE AND CON.

that leach data from small samples could be used to determine SOLIDATION AT TYPICAL URANIUM MILL TAILINGS SITES leachabikty indexes for full size waste form. The teachabihty in'. FAYER M J ; CONBERE W. Battelle Memortal Institute, Pacific dexes for cessum, strontium, and cobalt isotopes were deter Northwest Laboratones. May 1985. 56pp 8506190031. PNL-mined for waste samples from both nuclear plants according to 5421. 31017.324 models used in ANS 16.1. The teachabihty indexes for the PWR The computer code TRUNC was used to analyze three as-samples were 6.4 for cesum, 7.1 for strontium, and 10 4 for ects of uranium mill taihngs dewatenng- the couphng of con-cobalt. The leachabahty mdexes for BNR samples were 6 5,8 6' sohdation and fluid flow, drainage design, and cover load One-and 11.1 for cesum, stront um, and cobah, respectwey , ,

NUREG/CR-4181: LEACHABILITY OF RAD!ONUCLIDES FROM flow within a taihngs pile of either shmes or a sand /shmes mix CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT OPER- showed that drainage flux was greater for a consohdating ATING NUCLEAR POWER REACTOF S. CRONEY,S.T. EG&G system early in the smutation. After days 1,400 and 160 of the Idaho, Inc. (sebs. of EG&G, Inc.).. Apnl 1985. 130pp. simulations for tne shmes and sand /Shmes mix, respectuely, 8505070173. EGG-2355. 30217:161. however, the fluxes for the nonconsohdating systems were Different Szed samples of cement sohdified hquid wastes greater. In the sand /shmes mix, the nonconschdating system were collected from two nuclear power plants, a pressunzed had a cumulatwe flux by day 5,000 that was 93*= of that of the water reactor (PWR) and a boshng water reactor (BWR), to cor-consohdating system. At the same time, m the shmes taihngs relate radenuchde leachog from small and full sized waste forms. Diffusion-based model anafyss of measured radionuchde piles the nonconsohdating system had a cumultwe flux of only teach data from small samples and full Sze samples indicated 34% of that of the consohdating system, which mdicates that that leach data from small samples could be used to determine consohdation and fluid flow should not be decoupled for the leachabihty indexes for full sze waste form. The teachabihty m. shmes. Two-dimensional smutabons of an actual taihngs pile dexes for cesum, strontium, and cobalt isotopes were deter. drainage design showed that a sand blanket drain increased the mined for waste samples from both nuclear plants according to rate of drainage and settlement. The sand blanket drain also models used m ANS 16.1. The leachability ;.1 dexes for the PWR significantty reduced differential settlement across the pile. This samples were 6.4 for cesium, 7.1 for stron$ im, and 10.4 for indicates that the use of a sand blanket drain could enable ear.

cobalt. The teachabehty indexes for BWR sam )les were 6 % 8 6. her placement of the cover system after taihnas emplacement.

and 11.1 for cesium, strontium, and cobalt, respectuely. In simulations of covered and uncovered taihngs piles, nearly the ume quantity of water was removed from each, but drain.

NUREG/CR-4190: CALIFORNIA OFFSHORE SURVEY OF L1-CENSEES USING RADIOACTIVE MATERIAL. WONG,K.S ; age occuned much more slowty without the cover; hence, sur.

BRCWN J. Cahfomia. State of. May 1985. 22pp. 8506060007. face settlement was slower whe1 the tashrys pile was not cov-30770:318. ered.

This report is an account of offshore 'adcactrve matenal ac-tuities and was prepared to provide information about their safe NUREG/CR 4194: LOW. LEVEL NUCLEAR WASTE SHALLOW LAND BURIAL TRENCH ISOLAisON Final Report. October 1981 use m the manne environments beyond Cahtomia's junsdiction.

. September 1984 MCCRAY,J G.; NOWATZKl.E.A.;

The report supplies the essentral information called for and (a) identifies heensees with radcactue nuchde utshzation programs ARMSTaONG,G ; et al. Anzona, Urwv. of, Tucson, AZ. May (b) desenbes the hcensees' work statens, (c) identifies and/or 1985. 219pp. 8506240003. 31149 012.

desenbes radionuchde, quantities and their apphcations, and (d) This is the final report on a three year study to evaluate desenbes the radiaton safety concems and exist'ng methods trench cap designs, trench construction and trench load ng by for their resolution. Finally, three offshore sites were inspected accelerating the creation of void space to smutate waste degra-in a typical comphance manner and the findings reported En. dation in order to apply stress conditions on the trench in a rol.

closed photographs of the work statens. dunng source and ative short penod of time. Eight trenches were initialty construct-equipment use, dlustrate conditons and the 1censees' oper. ed and instrumented, four in a semi-and regon and f%r in a ations. It is concfuded from observations dunng onsite vists to more humid mountainous regon. After the first year the semi-these unusual work environments, that penodic onsite comph- and site was abandoned due to cap failures. A new trench in-ance inspections are necessary to assure radiaton protecton corporating an improved soil slab design with a wick was con-for all concemed. structed at the humid site. Conclusions from these expenments are: 1. Controlled compacton is not sufficient to mitigate long NUREG/CR 4191: SURVEY OF LICENSEE CONTROL ROOM term surface subsidence. 2. Single sheet geotextile reinforce-HABITABILITY PRACTICES. BOLAND.J F.; BROOKSHIRE,R L; DANIELSON.W.F.; et al. Argonne National Laboratory Apol ment is not adequate trench cap reinforcement. 3. Geotextile 1985. 225pp. 8505100194. ANL-85-13. 30268.213. wrapped soil slab attenuates surface subsidence and surface This document presents the results of a survey of Licensee water infdtration. 4 A steel-remforced soil.coment slab appears control-room-habitabihty practices. The survey is part of a com. to meet the requirements necessary for long term stabihty. 5. If prehensive program plan instituted in August 1983 by the NRC the crown and cap remain stable so does the trench. 6. Ahphat-to respond to ongoing questions from the Advisory Committtee ic tracers performed well and dye type of tracers poorty. 7.

on Reactor Safes,uards (ACRS). The emphasis of this survey Tracers are feasible and effectue as a trench monitonng tool, 8.

was to determine by field review the control-room habitabihty Narrow designed trenches improve trench cap stabikty. This practices at three different plants, one of which is still under report recommends a design for enhanced isolaton disposal constructen and scheduled to recetve an operating kcense m trench providing improved monitonng capabibties.

i 1

1 a

30 Main Citations and Abstracts ,

NOREG/CR-4its: OVERVIEW OF TRAC-801 (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 EG8G Inc.) Apnt 1985. 55pp. 8506060796. EGG- The results of the two samples were similar. A regresson analy-2382. 30771:279. ms was performed to estimate the contnbution of each vanable A senes of simulations were performed at Idaho Natonal En- to uncertainty in eshmated consequences. The study was first gineenng Laboratory to continue the advancement of Bothng performed with the magnitude of the source term as one of the

, Water Reactor (BWR) safety research, with the TRAC-B01 17 variables. A second analysis was performed with a fixed 3 (Vernon 12) computer code. The pnncipal motsation for per- source term. Only one sample of size 50 was generated in the

formmg these simulabons was to assess the code's capabdty to second analysis. The uncertainty /senstrwity anatyss techniques i

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) simulabons, and economic consequence model. Altematue methods for dis-

) wfuch cover a broad range of conditons dunng different types playing and desenbing the results are presented. The inssghts I of LOCA scenanos, are summanzed in this document. Selected gained from performing the analyss are revewed and maior i

compensons between calculated and measured results are pre- conclusons summarized. A comparison of tre results of this sented. Conclusons derived from those compansons are guen. study with current point eshmates of health and econome con-NUMEG/CR-4197: SAFETY GOAL SENSITIVITY STUDIES. sequences is presented.

l 8 7020 SA 4 13 J NUREG/CR-4200: BIODEGRADATION TEShNG OF SOLIDIFIED

! This study presents the results of analyses performed as part LOW-LEVEL WASTE STREAMS. PICIULO.P.L; SHEA.C E.;

! of the two-year evaluaton program for the NRC safety goals. BARLETTA.R E. Brookhaven Natonal Laboratory. May 1985.

Analyses are performed to demonstrate the senstuities of the 46pp.8506140593. BNL-NUREG-51868. 30936 219.

l 1 quanttatue design objectwo calculatons to changes in input The NRC Techncal Poston on Waste Form (TP) specifies parameters and assumptions. Results are presented whch that waste should be resstant to bodegradabon. The methods show the influence of parameter changes on the health nsk recommended in the TP for testing resistance to fungi, ASTM i

quanbtatwo design objectues and on cost-benefit calculabons. G21, and for testing re$ stance to bactena, ASTM G22. were -

I The attematwo design objectwo nsk measures are compared camed out on several types of sohdified smulated wastes, and with attematue measures of the health impacts of LWR acci. the effect of merobial actuity on the mechanical strength of the

, dents. The resutts of this study provide background informabon matenals tested was examined. The tests are beheved to be and input to be used in the NRC staff evaluation of the safety suffeient for distinguishing between matenals that are susceph-

! goals and quanbtative desgn obtectues. ble to tuodegradaten and those that are not. However, it is con-

! NUREG/CR-4tte: FRACTURE IN GLASS /HIGH LEVEL WASTE

' "" * "9 '

CANISTERS. MARTIN,D M. Iowa State Univ., Ames, lA. Apnl itself as an indcahon that the waste form will beodegrade to an 1985. 81pp. 8504170534. 29906 243 extent that the form does not meet the stabihty requirements of 10 CFR Part 61. In the case of fadure of ASTM G21 or ASTM The release rate of radonuchdes from a vitnfied waste form G22 or both, it is recommended that additonal data be supphed due to aqueous teaching by ground water will depend, among

other factors, on the waste form's surface area. Large castings by the waste generator to demonstrate the resistance of the

,,,....u...,_,ui,,~.,_...

of giass win almost certainty oe used as the waste form for high "~~~""~"""'"~~'~~**""~"'

levet nuclear wastes and such castings tend to fracture as a NUMEG/CR 4201

THERMAL STABILITY TESTING OF LOW.

result of transient and residual stresses induced by the casting LEVEL WASTE FORMS. PICIULO,P.L; CHAN,S F. Brookhaven process; such fractures increase the surface area available for 3

Natonal Laboratory. May 1985. 48pp. 8506060814. BNL.

aqueous teaching of redonuchdes from the HLW glass. The pri-NUREG 51869. 30769 247.

mary focus of this study was on achieving an understanding of the dependence of fracture surface area on glass properties The NRC Techncal Position (TP) on Waste Form specifies and processing vanables for both in-can melts and castings- that waste forms should be resistant to thermal degradabon.

j The thermal cycle tesbng procedure outhned in the TP on The maximum fracture surface area per unit volume of glass ob.

served in this study was about 7.1/cm (an equivalent sphencal Waste Form was camed out and is beheved adequate for dem-partcle 6ameter of 0 85 cm) for a water quenched in-can melt. onstrating the thermal stabdity of sohdified waste forms. The in-j

The processing parameter whch appears to most strongty cluson of control samples and the mnnetonng of sample tem-affect the extent of fracture surface area for both castings and perature are recommended additons to the test. An outhne for j reporting thermal cychng test results is given. To produce a data

in-can melts is the 6mensonless Bot modutus (thermal film co-effcient x redus/ waste form thermal conductuity). base on the apphcabihty of the thermal cychng test, the follow-Ing simulated laboratory-scale waste forms were prepared and i NUREG/CR-4199

A DEMONSTRATION UNCERTAINTY /SENSI. tested. bonc acid and sodsum sulfate evaporator bottoms, mixed j TlVITY ANALYSIS USING THE HEALTH AND ECONOMIC bed bead resins, and powdered resna each sohdified in asphalt,

CONSEQUENCE MODEL CRAC2. ALPERT,0.J.; IMAN.R.L; cement and vinyl ester-styrene.

I HELTON.J.C.; et al. Sandia Nabonal Laboratones. June 1985.

I 59pp. 8507050415. SAND 841824. 31372 210. NUREG/CR 4203: A CALCULATIONAL METHOD FOR DETER.

A demonstraton uncertain /senstuity analysis was performed MINING BIOLOGICAL DOSE RATES FROM IRRADIATED RE-for the reactor accident consequence model CRAC2 using tech. SEARCH REACTOR FUEL SCHNITZLER.B G. EG&G Idaho, niques compded as part of the NRC-sponsored MELCOR pro. Inc. (subs. of EGAG, Inc.).. Apnl 1985. 65pp 8506060811.

gram. The principal obectues l of the study were: 1) to demon- EGG-2383. 30775.237.

i strate the use of the uncertain /senstuity analysis techniques, This report desenbes a calculational method for the determo j' 2) to test the computer models that implement the techniques naten of biological dose rate from tradiated research reactor

3) to identify possbie effcutbes in performing such an analyss, fuels. The calculatonal method is implemented en a computer and 4) to explore alternatue means of analymng, displaying, and program for quick and convenient assessment of multigroup i desenbing the results. Seventeen CRAC2 input vanables gamma and beta dose rates resultng from an arbitrary (user.

thought to contnbute significantly to uncertainty in estimated supphed) rradiation history. The FUELDR program calculates consequences were selected for analyss; subective t estimates dose rates at a fixed dose point using builtan fisson product 6m-of ranges, 6stnbutons, and correlations for these vanables pulse source functons and precalculated gamma and beta j were made. Latin hypercube samphng, a modified Monte Carlo transport factors. The fixed dose pomt is located on the axial technique, was used to generate two multuanate samples of mid-plane at a distance of 91.44 cm (3 ft) from the fuef element.

I i

]

Main Citations and Abstracts 31 Transport factors are included for sixteen unique g235)U fuel A new computer code called MATADOR (Methods for the l

types in use at thirteen nonpower reactor facilibes. Ana'ysis of Transport And Depositen Of Radonuclides) has NUREG/CR-4204- LONG TERM EMBRITTLEMENT OF CAST was wnen W h Reactor Safen M WSHM ms DUPLEX STAINLESS STEELS IN LWR SYSTEMSAnnual mport contains a detailed descripion of the models used in Report. October 1983 - September 1984. CHOPRA,0 K , MATAN A companon mod pro @s a Usn Manual W i CHUNG H.M. Argonne Nabonal Laboratory. Apnl 1985. 33pg.

' the code. WADOR M interM for use in nstem nsk studes 8509140599. ANL-85-20. 30935:223. to analyze radonudde transport and depositen in reactor con-This progress report summanzes work performed by Argonne tamments. The pnncipal output of the code is inform 3h on on the

' National Laboratory dunng t% tweise months from Octobu 1983 to September 1984 on long. term embnttlement of cast timing and magmtude of radionuchde releases to the environ-duplex stainless steels used in hgtt-water reactors.

ment as a result of severely riegraded core accidents. MATA-DOR considers the :ransport of ladionuchdes through tne con-NUREG/CR-4205: 1 RAP-MELT 2 USEh'S MANUAL JORDAN,H.; tainment and their removal by natural depositen and the oper-KUHLM AN,M ft Battells Memorial Institute, Columbus Laborato- ation of engineered safety systems sxh as sprays. The code

)

nes. May 1985. 74pp. 8506190036. BMI-2124. 31017:100. requires input data on the source term from the pnmary system.

The TRAP MELT 2 cede is a development of tne prevous4 the geometry of the containraent, and the thermal-hydrauhc issued TRAP-MELT code whch simulates the transoort and conditions in the containment.

i deposition of aerosol parbcles and certain vapors in the reactor i coolant system under hyoothetcal accident conditons in a hght NUREG/CR-4211: MATADOR (METHODS FOR THE ANALYSIS water reactor. This manual contains a bnef desenption of tne OF TRANSPORT AND DEPOSITION OF RADIONUCLIDES) '

models of the processes treated in the code and of the code COOE DESCRIPTION AND USER'S MANUAL AVCl,H.I.:

organizabon. The input to the code for a sample run are pre- RAGHURAM,S.; BAYBUTT.P. Battelle Memonal Institute, Co- '

! sented and output from a run are presented as well. lumbus Laboratones. Apnl 1985. 75pp. 8505080373. BMI 2126.

NUREG/CR-4206: A SELECT REVIEW OF THE RECENT (1979- 30218:192.

1983) BEHAVIORAL RESEARCH LITERATURE ON TRAINING A new computer code called MATADOR (Methods for the SIMULATOHS. LAUGHERY,K.R. Oak Ridge Natonal Laborato. Analysis of Transport And Depositen of Radionuclides) has l

i ry. May 1985. 51pp. 8506t30489. ORNL/TM 9445. 30901:249. been developed to replace the CORRAL-2 computer code Report summanzes some selected reports of behavioral re- which was wntten for the Reactor Safety Study (WASH-1400).

search performed in years 1979-1983 on training simulator ap- This report is a User's Manual for MATADOR. A companion phcaton technology, and discusses findings related to nuclear report desenbes in detail the models used in the code. MATA-

power plant operators' simulator training. Fmdings are organized DOR is intended for use in systam risk studies to analyze rado-j as related to the design, tesbng, and use of training simulators. nuchde transport and depositen in reactor containments. Tne q

Topics include Simulator Fedehty vs. Training Effectiveness Op- pnncipal output of the code is informaton on the timing and erator Performance Measurement, Measunng Simulator Effec- magnitude of radionuchde releases to the environment as a tiveness, and Simulator Utihzation Prachces. Reviews 89 refer- result of severely degraded core accidents. MATADOR consid-

" era ihe transport of radionuciw.ies dwuugh use containmeni snd

]

NUREG/CR-4200: GASTROINTESTINAL ABSORPTION OF PLU.

their remeval by natural deposton and by engineered safety TONIUM IN MICE. RATS, AND DOGS.Apphcaten To Estabhsh. systems such as sprays. It is capabie of analyzing the behavior ing Values Of f1 For Soluble Plutonium. BHATTACHARYYA; of raionuchdes existing either as vapors or aerosols in the con-LARSEN,R.P.; OLDHAMAD.; et al. Argonne Natonal Laborato- tainment. The code requires input data on the wurce terms into

! ry. May 1985. 99pp. 8507050425. ANL.85-21. 31371:207, the containment, the geometry of the containment, and thermal-The gastrointesbnal (GI) absorption of plutonium was meas- hydraule conditons m the contamment.

I ured in mice, rats, and dogs under conditons relevant to setting I dnnkmg water standards. The fractional GI absorpton of Pu (VI) NUREG/CR-4212: IN. PLACE THERMAL ANNEALING OF NU-4 in adult mce was 2 x 10(-4) (0.02%) in fed mice and 2 x 10(-3)

CLEAR REACTOR PRESSURE VESSELS. SERVER,W.L 1 (0.02%) in fasted mice. The Gl absorpton of plutonium was in. EG8G Idaho, Inc. (subs. of EG&G, Inc.).. Apnl 1985. 250pp.

dependent of plutonium oxidaten state, administration medium, 8503070548. EGG-MS-6708. 30211:101.

j and plutonium concentration; absorpton was dependent upon Radiaten embnttlement of fernte pressure vessel steels animal species, state of animal fasting, state of Pu(IV) hydroty- changes the toughness properties. A thermal annea! cycle well I sis, and age of the animal. Fractional GI absorption values above the normal operating temperature of the vessel can re-ranged from 3 x 10(-5) (0.003%) for hydrolyzed Pu(IV) adminis- store most of the onginal properties. The Army SM 1 A test re-( tered to fed adult mee to 7 x 10(-3) (0.7%) for Pu(VI) adminis- actor vessel was wet annealed in 1J67, and wet anneahng of i

tered to fed neonatal rats. From analysis of our data, we sug* the Belgian BR-3 reactor vessel has recentty taken place. An gested values of f(1) (the fraction transferred f om gut to biood industry survey indicates that dry anneahng of a reactor vessel in humans) for use in estabhshment of oral hmits of exposure to in-place is feasible, but solvable engineering problems exist.

plutonium. For an acute exposure m the occupatorw settng, Limited toughness data available for five high copper content we propcsed one value of f(1) for fed (2 x10(-4) and one for welds were reviewed. The review suggested that significant re-a fasted (2 x 10(-3) individua's. For the environmental setbng, we covery results from annea'ing at 454 degrees centigrade (850 developed two approaches to obtaining values of f(1); suggest- degrees fahrenheit) for one week, but scatter in the data makes ed values were 6 x 10(-4) and 4 x 10(-3), respectively. Both ap- assessment of recovery ard reembnttlement response deffcult proaches took into account effects of animal age and fasting. to quantify. A thermal and structural analysis of a reactor vessel We discussed uncertainbes in proposed values, of f(1) and undergoing an annealing tieatment found no problems with the l

j made recommendatens for further research. reactor vessel itself, but did indcate a rotation at the t'ozzle NUREG/CR-4210: MATADOR.A COMPUTFR CODE FOR THE regon of the vessel which would plastcally deform the attached 1,

ANALYSIS OF RADIONUCLIDE BEHAVIOR DURING DEGRAD. pnmary piping. Further analytcal studes attempted to solve this ED CORE ACCIDENTS IN LlGHT WATER REACTORS. problem, but tt'ty were not successful. An Amercan Society for I BAYBUTT,P.; RAGHURAM,S.; AVCl,H.l. Battelle Memonal Insti- Testing and Matenals (ASTM) task group is upgrading and re.

l tute, Columbus Laboratones. Apnl 1985. 62pp. 8505080375. vising auide ASTM E 509-74 with emphasis on the materials BMI-2125. 30218.271, and surveillance aspects of anneahng. I i

I

l 32 Main Citations and Abstracts NUREG/CR-4215: TECHNICAL FACTORS AFFECTING LOW- Quaktative and quantitative informaton developed for contain-  ;

LEVEL WASTE FORM ACCEPTANCE CRITERIA. ment performance under normal operating conditons and ,

MACKENZIE.D.R.; VASLOW.F.; DOUGHERTY.D.R.; et al. design basis accidents indicate that there is room for improve-  !

Brookhaven Natonal Laboratory. May 1985.77pp.8506140405. ment. A crude estimate for overall containment unavailability for  !

BNL NUREG-51873. 30908.116. relatively stiiall leaks which violate plant technical specifications i This report provides technical support to NRC in connecten with the regulaton 10 CFR Part 61 and NRC's Techrucal Posi-is 0.3. An estimate of containment unavailability due to large l tion (TP) on waste form. Six specific areas are addressed

• leakage events is in the range of 0.001 to 0.01. These esti- l namely: the technical basis for limiting containers of radcactive mates are dependent on several assumptons (particularly on gases to atmosphenc pressure and 100 cunes; the require- event duration times) whch are documented in the report. I ments to demonstrate that a stable waste would be recogruz- NUREG/CR-4221; AN EVALUATION OF STRESS CORROSION able for 300 or 500 years; the feassbehty of achieving less than q

5% deformaton in buned wastes; the adequacy of ASTM tests CRACK GROWTH IN BWR PIPING SYSTEMS. KASSIR.M.;

G21 and G22 for testng for biodegradability; the adequacy of SHARMA.S.; REICH.M.; et al. Brookhaven Natonal Laboratory.

ASTM test 8553 for testing for thermal degradation; and the May 1985. 80pp. 8506130175. BNL-NUREG-51874. 30867:183' basis for determining if a waste is explosive or pyrophoric. The This report presents the results of a study conducted to pnncipal conclusens of the report follow. A maximum pressure evaluate the effects of stress intensity factor and environment of 1.5 atmospheres for radcactive gases is acceptable, but the on the growth behavior of integranular stress corrosion cracks radioactivity hmat should depend on the isotope, the quality of in type 304 stainless steel piping systems. Most of the detected the container and the properties of the site. Site and package cracks are known to be circumferential in shape, and initally qualitios and a wet / dry cycling test are suggested that apprecia- start at the inside surface in the heat a'fected zone near girth bly increase the probability of indicating whether a waste would welds. These cracks grow both radially in-depth and circumfer-have long-term recognizabihty. Achieving deformaton of buned entially in length and, in extreme cases, may cause leakage in waste of 5% would not be feasible using current sciidification tf a installation. The propagaton of the crack is essentially due methods with either metal or polyethylene containers. ASTM to the influence of the following simultaneous factors: (1) The l tests G21 and G22, with modifications are suitable for biodegra- action of apphed and residual stress, (2) Sensitization of the dability teshng. A modified form of ASTM B553 is adequate for base metal in the affected zone adjacent to firth weld and (3) thermal testing. Required informaton on pyrophonc and explo-sive matenals is provided. The continuous exposure of the matenal to an aggressive envi-ronment of high temper ture water containing dissolved oxygen NUREG/CR-4218: LOCA SIMULATION IN THE NATIONAL RE- and some levels of impunties. Each of these factors and their SEARCH UNIVERSAL HEACTOR PROGRAM.Postirradiation effects on the piping systems is discussed in detail in text of the Examination Results For The Third Matenals Test (MT-3) - report. Tne report also evaluates the time required for hypotheti-Second Campaign. HABERMAN J.H. Battelle Memoria! Institute, cal cracks in BWR pipes to propagate to their entical size. The Pacife Northwest Laboritones. June 1985.62pp.8506260395. pertinent times are computed and displayed graphically. Finally, PNL-5433. 31244:326. parametric study is performed in order to assess the relative in.

A senes of in-reactor expenments were conducted using full-length 32-rod pressunzed water reactor (PWA) fuel buncties an fluence and sensitivity of the vanous input parameters (residual part of the Loss-of-Coolant Accident (LOCA) Simulation Pro- struss, crack growtn iaw, osameter of pipe, irwtial size of defect, gram by Pacific Northwest Laboratory (PNL). The third matenals etc.) which have beanng on the growth behavior of the inter-granular stress corrosion cracks in type 304 stainless steel.

ra c a en s f ation ture e n nts or Cracks in large-diameter as well as in small-diameter pipes are ed in the Natonal Research Universal (NRU) Reactor, Chalk considered and analyzed.

River, Ontano Canada. The MT-3 expenment was jointly funded by the U.S. Nuclear Regulatory Commission (NRC) and the NUREG/CR-4225:

SUMMARY

OF EFFICIENCY TESTING OF Un.ted Kingdom Atome Energy Authonty (UKAFA) with the STANDARD AND HIGH-CAPACITY HIGH-EFFICIENCY PAR.

main objective of evaluaten balloorung and rupture dunng TICULATE AIR FILTERS SUBJECTED TO SIMULATED TOR.

l active two-phase cooling at elesated tempertures. All 12 test NADO DEPRESSURIZATION AND EXPLOSlVE SHOCK rods in the center of the 32-rod bundle failed with an average WAVES. SMITH,P.R.; GREGORY,W.S. Los Alamos Scientific i peak strain of 55.4%. At the request of the UKAEA, a destruc- Laboratory. Apnl 1985. 25pp. 8507020407, LA-10401-MS.

teve postirradiation examination (PIE) was perfornied on 7 of the 31309:007, 12 test rods. The results of this examinaten were presented in Pressure transients in nuclear facihty air cleaning systems can a previous report. Subsequently, and at the reqwst of UKAEA, originate from natural phenomena such as tornadoes or from PIE was performed on three additional rods along with further accident-induced explosive blast waves. This study was con-examination of one of the prevouuy exammed rods. Informaton cerned with the effactive efficiency of high-efficiency partculate obtained from the PIE incfuded cladding thickness measure- air (HEPA) filters dunng pressure surges resulting from simulat-ments, cladding metallography, and partcle size analysis of the ed tomado and explosen transients. The primary objective of fractured fuel pellets. This report desenbes the additional PIE the study was to examine filter efficiencies at pressure levels work performed and presents the results of the examinatens. below the point of structural failure. Both standard and high-ca-i NUREG/CR-4220: RELIABILITY ANALYSIS OF CONTAINMENT pacity 0.61-m by 0.61-m HEPA filters were evaluatod, as were ISOLATION SYSTEMS. PELTO,P.J.; GALLUCCt.R.H.; several 0.2-m by 0.2-m HEPA filters. For a partcular manufac-AMES,K R. Battelle Memorial Institute, Pacific Northwest Lab- turer, the material release when subjected to tomado transients oratones. June 1985. 222pp. 8506260362. PNL-5432. is the same (per unit area) for both the 0.2-m by 0.2-m and the 31245:199. 0.61-m by 0.61m filters. For tomado transients, the material was

)i This report summanzes the results of the Reliabehty Analysis on the order of micrograms per square meter. Wher' subjecting of Containment Isolaten System Project. Work was pai-,g clean HEPA filters to simulated tomado transients with aerosol j in five basic areas: design review, operating expenence review, entrained in the pressure pulse, all filters tested showed a deg-related research review, generc analysis and plant specife radation of filter effeiency. For explosive transients, the matenal analysis. Licensee Event Reports (LERs) and Integrated Leak release from preloaded high-capacity filters was as much as Rate (ILRT) Test reports provided the mator sources of contain. 340 g. When preloaded high-capacity fitters were subjected to 1 ment performance mformation used in this study. Data extracted shock waves approximately 50% of the structural limit level,1

! from the LERs were assembled into a computer data base. to 2 mg of partculate was released.

i

Main Citations and Abstracts 33 NUREG/CR-4226: NEW MADRID SEISMOTECTONIC NUREG/CR-4231: EVALUATION OF AVAILABLE DATA FOR STUDY.Actrvities Dunng Fiscal Year 1983. BUSCHBACH,T.C. PROBABILISTIC RISK ASSESSMENTS (PRA) OF FIRE St. Louis Urw., St Louis, MO. Apnl 1985.153pp.8505070552. EVENTS AT NUCLEAR POWER PLANTS. SAMANTA,P.K.;

30209:297, BOCCIO.J.L Brookhaven National Laboratory. KRASNER,LM.;

The purpose of the New Madrid Seismotectonic Study is to et al. Factory Mutual Research Corp. May 1985. 71pp.

idenbfy the earthquake mechanisms within a 20Mie radius of 8506190077, BNL-NUREG-51879. 31017:205.

New Madnd, Missouri. During 1983 there was more awareness Several crucial parameters are needed in the assessment of of the sigrAcance of current regional stress pattems and the fire nsk in nuclear power plants. Among those that need to be local concentration of stresses by basement structures and in- developed from a data base are: (1) fue frequency, (2) fire de-M.ix.swebes. The program continued to concentrate on defin- tection time, and (3) fue suppresson time. Currently, that data ing bouncanes of t proposed nft complex in the area, as well base for nuclear power plants is not large enough to develop as estabhshing the re;abonsNps of the east-west trending fault these parameters, considenng fuel locahon, fuel geometry, com-sys'. ems with the northeast-trending faults of the Wabash Valley buston properties, enclosure geometry, etc. TNs study attempts and New Madnd areas. There were 204 earthquakes located by to augment the nuclear data base by investgating the useful-the Saint Louis Urwersity meroearthquake network in 1983. In ness of other nonnuclear data bases whch contain fre incident addition, the earthquake swarm in north-central Arkansas con- loss experience of occupancy classes having somewhat similar tinued throughout the year, and 45,000 earthquakes have been physical features and fire protecten engineenng systems nor-recorded there since January,1982. TrencNng data from Late mally found in nuclear power plants. This study has found that CennzN terrace deposits along the Kentucky River Fault indeed some useful information can be gleaned from nonnucle-Systei lgest that there was post-terrace deformaton along ar sources; in partcular, detecton and suppression times. How-some < J faults. Thermal and chemical data from grouridwat- ever, other fre-nsk data needs such as fre frequency and fire ers ir 2 Mississippi Embayment appear to be useful in localiz- size would require other forms of data searches and data anaty-ing ceep faults that cut through the aquifers. Earty indcations ses that at tNs stage can only be conceptualized.

from studies of joinnng in Indiana are that the directen of major joint sets w 11 be useful in determining regonal stress drections. NUREG/CR-4237: MOBILITY OF RADIONUCLIDES IN HIGH CHLORIDE ENVIRONMENTS. SIMPSON.H.J.; HERCZEG A.L; No Ouaternary fauinng was found in the Indiana or lihnoes fault ANDEF SON.R.F.; et al. Columbia Urw., New York, NY, April studies.

1985. 77pp. 8505070484. 30210:219.

NUREG/CR-4229: EVALUATION OF CURRENT METHODOLOGY Concentratens of naturalty occumng isotopes of uranium, EMPLOYED IN PROBABlUSTIC RISK ASSESSMENT (PRA) OF thorium, radium and radon were measured in freshwaters and in FIRE EVENTS AT NUCLEAR POWER PLANTS. RUGER C.; sodium-chionde bnnes near the site of the Waste isolation Pilot BOCCIO,J.L; AZARM,M.A. Brookhaven Natonal Laboratory. Plant (WIPP) located in southeastern New Mexco. Supplemen-May 1985. 47pp. 8506190103. BNL-NUREG-51877. 31017:277. tal water chemsstry analysos (chionde, alkahrvty, P(CO2), CO(2),

The report presents a general evaluation of the current meth- Fe, Mn, H(2)S) were made to aid in interpreting the data for nat-odology used by industry for the probabihstic assessment of fire ural radionuclides. Three features of radionuchde mobihty are events in nuclear power plants. The basis for this evaluation, in evident from the results: 1) There is a shght tendency for U and which the strengths and weaknesses of the methods are identi- Ra concentrations to correlate mth the chloride content of the fied, stem from reviews of several, industry-sponsored, full- water samples. Whether this tendency resutts from complexa-scope Prot,abikste Risk Assessments (PRAs) and vanous deter- tion by Cl- ions or cation exchange competition for adsorptson i ministc/probabihste approaches used by industry to judge their sites cannot be resolved mth the available information. 2) Much compliance with or used to seek exemptons from the fre-pro- more dramatic than the correlabon with Cl- concentraton is the tecten requirements enumerated in Appendix R to 10 CFR 50. effect of the redox state of the waters on U and Ra concentra-In performing tfis evaluation of the current methodologies, tons. Chemicatty reducing groundwaters contain much lower U state-of the-art irterature on the modeling of propagabon/detec- concentratons and much hegher Ra concentratons than were tion / suppression, input parameters, and modehng uncertainties measured in oxic and suboxc samples. Calculated retardabon i

are utilized. Areas are identified where recently developed, more factors of 1 for Ra indcate that it can migrate freely in anoxic accurate and complete techniques can be implemented to brines. 3) Low chemical recovenes of Th, and to a lesser extent reduce the state-of-knowledge uncertainties that presently exist. U were observed for methods that work well with seawater

Recommendations are Cso made whch could be the basis for samples. These elements may be present in a mobile, unreac-a more suitable and complete fire-risk methodology. tive dissolved or colloidal complex with organc matter.

l NUREG/CR-4230: PROBABILITY BASED EVALUATION OF SE- NUREG/CR-4245: IN-PLANT SOURCE TERM MEASUREMENTS

( AT BRUNSWICK STEAM ELECTRIC STATION. DUCE,S.W; I LECTED FIRE PROTECTION FEATURES IN NUCLEAR POWER PLANTS. AZARM,M.A.; BOCCIO,J.L Brookhaven Na- CRONEY,S.T.; AKERS.D.W.; et al. EG&G Idaho, Inc. (subs. of tonal Laboratory. May 1985. 93pp. 8506180415 BNL-NUREG- EG&G, Inc.),. June 1985. 800pp. 8507020396. EGG 2392.

51878. 30985:281. 31311:061.

A probabiliste approach for the evaluabon of major fire pro- This report presents data obtained at Brunswick as part of tecten measures in nuclear power plants is desenbed. The the In-Plant Source Term Measurement Program in operating methods developed are applied to two representative fire areas light water reactors (LWRs). The work was conducted for the

- one similar to a cable routing room and the other typical of a Office of Nuclear Regulatory Research (RES) in support of the diesel generator room. The fire areas chosen for apphcation, Meteorology and Effluent Treatment Branch (METB) of the the fire scenarios desenbed, and the various fire-damage states Office of Nuclear Reactor Regulabon (NRR). The pnmary objec-specified in the two illustrative examples are used to evaluate trve of this program is to provide the Nuclear Regulatory Com-those fire-protecten guidehnes which deal with automatic / mission (NRC) with operational data that can be used in evalua-manual fire detection and suppression systems, rated bamcrs, tion of plant designs for liquid and gaseous radwaste treatment

! divisional separation, drainage systems, dampers, and fire rating systems. Data presented were obtained at the Brunswick Nucle-of electncal cables. Tabular results are presented, which reflect ar Generating Station, operated by Carohna Power and Light, lo-the relative ments of these systems / features in terms of condi- cated at Southport, North Carolina. in-plant measuremens were tional probabihties of achieving various room-damage states. conducted during the time penod from March 1982 to Novem-The conclusions drawn and the lessons teamed through the ber 1982. This plant is the sixth in a series of operating LWRs course of this study are discussed, and the areas that may to be studied and the first boihng water reactor (BWR) in the need further investigation are identified. senes.

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34 Main Citations and Abstracts NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAIL- bers. This study also confirmed that the malfunction of pipe URES ON TRANSIENTS AND ACCIDENTS AT A GENERAL whip restraints introduced higher thermal stresses and tended J ELECTRIC BOILING WATER REACTOR. Main Report. to reduce the overall piping rehabihty. Finally, our results indicat-BRUSKE.S.J.; BAXTER.D E.; RANSOM.C.B.; et al. EG8G Idaho, ed that supports in a flexible piping design may need to be re-Inc. (subs. of EG&G, Inc.).. May 1985. 81pp. 8506240180. EGG- evaluated and that the eliminaten of pipe supports which are 2394.31150:280. close to components should be done with care in order to mini-This report documents the evaluation of the effects of nonsa- maze the impact on the component rehabihty.

fety grade control system failures on a typical boiling water re-actor plant. The methods utilized in this evaluaton include a NUREG/CR-4264: INVESTIGATION ON HIGH. EFFICIENCY PAR-system level failure modes and effects analysis, deterministic. TICULATE AIR FILTER PLUGGING BY COMBUSTION AERO-computer analysis, a review of 3 years of recorded plant occur- SOLS. FENTON.D.L; GREGORY,W.S.; et al. Los Alamos Scien-rences, a piobability analysis and a review of applicable NRC tific Laboratory. GUNAJI.M.V. New Mexco State Univ., Las entena pertaining to control systems. This study identified three Cruces, NM. May 1985. 32pp 8507050422. LA-10436-MS.

system failures that could cause transients leading to a reactor 31375:021.

vessel overtill and of these three failures, two could also lead to Experiments were conducted to investigate high-efficiency a reactor coolant cooldown of greater than 100 degrees fahren. particulate air (HEPA) filter plugging by combustion aerosols.

heet per hour. This study concluded that the existing NRC cnte. These tests were done to obtain empincal data to improve our na, concerning control systerr:s, adequately address the poten. modeling of filter plugging phenomena using the Los Alamns tial problem areas that were identified during this evaluation. National Laboratory fire accident analysis code FIRAC. Com-Based on the results of this study,it was recommended that the mercially available 0.61-m by 0.61-m square filters were tested consequences and risk associated with overfill and overcool in a specially designed facihty to determine how airflow resist-transients be further investigated. ance vanes with increased filter loading by combustion aero-sols. Two organic fuels normally found in nuclear fuel cycle fa-NUREG/CR-4262 V02: EFFECTS OF CONTROL SYSTEM FAIL

• cihties, polystyrene (PS) and potymethytmethacrylate (PMMA),

URES ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC BOILING WATER REACTOR. Appendices. were burned under varied conditions to generate combustion BRUSKE S.J.; BAXTER.D.E.; RANSOM.C.B.; et al. EG&G ldaho, aerosols. The test facility included a combustor, a 23-m-long Inc. (subs. of EG&G, Inc.),. May 1985. 341pp. 8506240689. duct, and a specialty designed gravimetnc balance for determ-EGG-2394. 31179:036. ng the aerosol mass gain of the filters. Test results include cor-Safety implicatens of Control Systems (A-47) was approved relations of HEPA filter resistance ratios (actual resistance /ir.i-as an Unresolved Safety issue (USI) by the Nuclear Regulatory tial resistance) with aerosol mass gain. The mass gain of plugged HEPA filters was found to correlate wth the airborne Commission (NRC) in December of 1980. USl A-47 concerns the potential for transients or accidents being made more mass concentration of material in the size range greater that severe as a result of control system failures. This report de- approximately 2.0 m. Also, the fuel with a smaller soot fraction, scnbes the work performed on the effects of control system fail- PMMA, produced filter plugging at lower accumulated aerosol mass deposits on or within the filter.

ures on transients and accidents at a General Electric boihng water reactor. This work was conducted for the U.S. Nuclear NilRFmCR.427t RECOMMENDEO Hegulatory Commission, Dnnsson of Safety Technology by SAFETY, RELIABILITY OUALITY ASSURANCE AND MANAGE-EG&G Idaho, Inc. and is based on the Browns Ferry Nuclear MENT AEROSPACE TECHNIOUES WITH POSSIBLE APPLICA-Plant. This report is contained in two volumes; a main report and five appendices. The main report desenbes the study meth-TION BY THE DOE TO THE HtGH LEVEL RADIOACTIVE WASTE REPOSITORY PROGRAM. BLAND,W.M. GeeB's, Inc.

odology, the matur areas of work performed, and the results June 1985.113pp. 8507080205. 31393:164.

and conclusons. The appendices contain detailod information consisting of failure mode and effects analysis tables, a detailed Aerospace SROA and management techniques, principally those developed and used by the NASA Lyndon B. Johnson description of the computer analyses and significant transient Space Center on the manned space thght programs, have been excerpts.

assessed for possible apphcation by the DOE and the DOE-NUREG/CR-4263: RELIABILITY ANALYSIS OF STIFF VERSUS contractors to the high level radioactive waste repository pro.

FLEXIBLE PIPING FINAL PROJECT REPORT. LU.S.C.; gram that results from the implementation of the NWPA of CHOU,C.K. Lawrence Livermore National Laboratory. May 1985. 1982. Those techniques believed to have the greatest potential 78pp. 8505280086. UCRL-20410. 30604:169. for usefulness to the DOE and the DOE-contractors have been This research project is to develop a technical basis for flexi. discussed in detail and are recommended to the DOE for adop-ble piping designs which willimprove piping reliabihty and mini- ten; discussion is provided for the manner in which this transfer mize the use of pipe supports, snubbers, and pipe whip re, of technology can be implemented. Six SROA techniques and straints. This study indicated that piping design can be made two management techniques are recommended for adopton by more reliable by some reduction of nged supports and/or snub- the DOE; included w th the management techniques is a recom.

bers. This study also confirmed that the malfuncton of pipe mendation for the DOE to include a hcensing interface with the whip restraints introduced higher thermal stresses and tended NRC in the apphcation of the milestone review technique.

to reduce the overall piping reliability. Finally, our results indicat- These other techniques are recommended for study by the DOE ed that supports in a flexible piping design may need to be re, for possible adaptiorkto the DOE program.

evaluated and that the ehmination of pipe supports which are close to components should be done with care in order to mini- NUREG/CR-4276: VIBRATION AND WEAR IN STEAM GENERA-mize the impact on the component reliability' TOR TUBES FOLLOWING CHEMICAL CLEANING - SEMIAN-NUAL REPORT. ENDERLIN.W.I.; BAUGH,J.W. Battelle Memon-NUREG/CR-4263: RELIABILITY ANALYSIS OF STIFF VERSUS al Institute, Pacific Northwest Laboratones. June 1985. 38pp.

FLEXIBLE PIPING FINAL PROJECT REPORT, LU.S.C.; 8507030714. PNL-5477. 31322.273.

CHOU,C.K. Lawrence Livermore National Laboratory. May 1985. The Pacific Northwest Laboratory is studying the effects of in-78pp. 8505280086. UCRL-20410. 30604:169. creased tube / tube-support clearances in pressunzed water re-This research protect is to develop a technical basis for flexi- actor steam generators following chemical cleaning. The project ble piping designs which will improve piping rehabikty and mini- purpose is to provide NRC with cntena for evaluating licensees' mize the use of pipe supports, snubbers, and pipe whip re- specific proposals for chemical cleaning of steam generators.

straints. This study indicated that piping design can be made This report desenbes the test and data anafysis plans and pro-more reliable by some reduction of nged supports and/or snub- cedures for the flow and accelerated wear tests to be per.

I Main Citations and Abstracts ' 35 formed in a scale-model steam generator. The flow tests will es- fers, occurs. This agitated region appears to propagate down-tablish the forcing boundary conditions, using clearances repre- stream in a quasi-penodic pattern. Increased inlet hquid flow senting vanous conditions following chemical cleaning. The ac- rates, and high gas annulus flow rates tend to diminish the sig-celerated wear tests will determine the potential wear rates pos- nificance of this agitated region. Observed inverted annular flow sible, based on the vibrations characterized in the flow tests. (and subsequent downstream flow pattem) hydrodynamic be-The overall project status, including work completed to date and havior is reported, and comparisions are drawn to data generat-tasks planned for the remainder of FY85, is also documented. ed by previous expenmenters studying post-CHF flow.

NUREG/CR-4277: INVERTED ANNUAL FLOW EXPERIMENTAL NUREG/CR-4283: STUDY OF THE EFFECTS OF ELASTIC UN-STUDY. DE JARLAIS,G.; ISHil.M. Argonne National Laboratory. LOADINGS ON THE Ji-R CURVES FROM COMPACT SPECl-April 1985.115pp. 850705')406. ANL-85-31. 31338.074. MENS. SUTTON,G E4 VASSILAROS.M.G. David W. Taylor Steady-state inverted annular flow of Freon 113 in up flow N.aval Research & Development Center. June 1985. 50pp.

was established in a transparent test secton. Using a special 8506260731, 31227:118.

inlet configuration consisting of long aspect-ratio liquid nozzles An investigation was performed to evaluate the efforts of coaxially centered within a heated quartz tube, idealized invert- elastic unloadings on the J-Integral Resistance Curves of ASTM ed annular flow initial geometry (cyhndrical liquid core surround- A106 Class C steel and 3-Ni steel. Compact specimens (IT) ed by coaxial annulus of gas) could be established. Inlet liquid were tested using a multi-specimen technique, direct current po-and gas flowrates, liquid subcooling, and gas density (using vari- tential drop technique and the elastic unloading compliance ous gas species) were measured and vaned systematically. The technique with unloading ranging from 10 to 90%. The two hydrodynamic behavior of the liquid core, and the subsequent former techniques were 0% unloading procedures used to gen-downstream break-in of this core into slugs, ligaments and/or erate the reference J-R curves for compenson to the elastic un-droplets of various sizes, was observed. In generai, for low inlet loading J-R curves for the two steels. The results of the investi-Lquid velocities it was observed that after the initial formation of gations of these matenals indscate that there was no significant roll waves on the liquid core surface, an agitated region of high difference in the J-R curves that resulted from the elastic un-surface area, with attendant high momemtum and energy trans- Ioading comphance technique.

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Contractor Report Number Index This index lists, in alphabetical order, the NUREG/CR for the report and to the 10-contractor-issued 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 SECONDARY REPORT NUMSER REPORT NUMSER AECL MISC-305 NUREG/CR-4077 ORNL/NSC200 NUREG/CR-2000 V04 N4 ANL44-102 NUREG/CR-4180 ORNL/NSC200 NUREG/CR-2000 V04 N5 ANL44-35 V04 NUREG/CR-3804 V04 ORNL/NSC223 NUREG/CR-3905 V03 ANL44-41 NUREG/CR-4064 ORNL/NSC223 NUREG/C43905 V02 NUREG/CR-3998 V02 ORNL/NSC223 NUREG/CR3905 V01 R1 ANL-84-60 NUREG/CR3905 V04 ANL4441 NUREG/CR-3980 V03 ORNL/NSC223 ORNUTM-8664/V2 NUREG/CR3514 V02 ANL44-87 NUREG/CR-3855 ORNUTM-9041/V2 NUREG/CR-3626 V02 ANL45-13 NUREG/CR-4191 #

ANL45-20 NUREG/CR-4204 / 9 hU CR ANL45 21 NUREG/CR4208 OANUTM-9267/V3 NUREG/CR-3885 V03 ANL-85-31 NUREG/CR-4277 ORNUTM-9316 NUREG/CR-3930 ANL-85-5 NUREG/CR4124 ORNUTM-9390 NUREG/CR-4015 ARAP NO 504 NUREG/CR4158 ORNUTM-9423/V2 NUREG/CR4031 V02 ARAP NO. 505 NUREG/CR-4159 ORNL/TM-9423/V3 NUREG/CR4031 V03 BHARC-400/84/02 NUREG/CR4139 ORNL/TM-9445 NUREG/CR-4206 BMI-2123 NUREG/CR-4177 V02 ORNL/TM-9477 NUREG/CR4086 BMI-2123 NUREG/CR-4177 V01 ORNUTM-9491 NUREG/CR 4092 BMI-2124 NUREG/CR4205 OANL/TM-9506 NUREG/CR-4105 BMI-2125 NUREG/CR4210 ORNL/TM-9522 NUREG/CR-4134 BMI-2126 NUREG/CR-4211 ORNL/TM-9585 NUREG/CP-0062 BNL-NUREG-51454 NUREG/CR-2331 V04 N3 PNL-4742 NUREG/CR-3317 BNL-NUREG-51454 NUREG/CR-2331 V04 N4 PNL4971 NUREG/CR3613 V02 BNL-NUREG-51630 NUREG/CR3091 V04 PNL-5064 NUREG/CR-3747 BNL NUREG-51630 NUREG/CR-3091 V05 PNL 5106-4 NUREG/CR-3810 V04 BNL-NUREG-51708 NUREG/CR-3469 V02 PNL-5158 NUREG/CR-3987 BNL-NUREG-51750 NUREG/CR-3703 PNL 5160 NUREG/CR3883 BNL-NUREG-51821 NUREG/CP-0059 V01 PNL-5179 NUREG/CR3906 BNL NUREG-51848 NUREG/CR4093 PNL 5300 NUREG/CR-4070 V03 BNL-NUREG-51857 NUREG/CR4149 PNL-5303 NUREG/CR4051 BNL-NUREG-51868 NUREG/CR-4200 PNL-5323 NUREG/CR-4075 BNL-NUREG-51869 NUREG/CR4201 PNL-5324 NUREG/CR4076 BNL-NUREG-51873 NUREG/CR-4215 PNL-5338 NUREG/CR4088 BNL-NUREG-51874 NUREGICR4221 PNL-5354 NUREG/CR4168 BNL-NUREG-51877 NUREG/CR4229 PNL 5361 NUREG/CR-4118 BNL-NUREG-51878 NUREG/CR4230 PNL 5381 NUREG/CR-4139 BNL-NUREG-51879 NUREG/CR-4231 PNL-5386 NUREG/CR4176 EGG-2164 NUREG/CR 2531 R03 PNL-5389 NUREG/CR4144 EGG-2231 NUREG/CR-3005 PNL-5404 NUREG/CR-4160 EGG-2245 NUREG/CR-3193 PNL-5421 NUREG/CR4 f 92 NUREG/CR-3862 PNL-5432 NUREG/CR-4220 EGG-2323 NUREG/CR-4218 EGG-2341 NUREG/CR 3977 PNL 5433 NUREG/CR-4033 PNL 5477 NUREG/CR4276 EGG-2352 SAND 82-1105 NUREG/CR-2718 EGG-2355 NUREG/CR-4181 NUREG/CR 3611 NUREG/CR-4071 SAND 82-2156 EGG 2362 SAND 83-0395 NUREG/CR3197 V01 EGG-2365 NUREG/CR-4077 SAND 83-2621/1 NUREG/CR-3721 V01 EGG-2367 NUREG/CR-4084 EGG 2382 EGG-2383 NUREG/CR4196 NUREG/CR-4203 h [ 86 SAND 644806 NUREG/CR3803 EGG-2387 NUREG/CR-4040 SAND 84-0814 NUREG/CR-3757 EGG-2392 NUREG/CR-4245 SAND 64-1025 NUREG/CR-3820 V03 j EGG-2394 NUREG/CR-4262 V01 SAND 841072 NUREG/CR-38e V02 t

EGG-2394 NUREG/CR-4262 V02 SAND 64-1264 NU9EG/CR-3863 EGG MS4708 NUREG/CR4212 SAND 84-1404 NUREG/CR-4169 HEDL TME 84-21 NUREG/CR-3746 V02 SAND 84-1461 NUREG/CR-3904 l HEDL-TME 84-31 NUREG/CR3746 V03 SAND 841522 NUREG/CR3913 IE 143 NUREG/CR-4003 SAND 64-1646 NUREG/CR3944 IE-146 NUREG/CR-4005 SAND 64-1824 NUREG/CR4199 l IEB 79-04 NUREG/CR4003 SAND 64-1838 NUREG/CR-2951 l IEB-79-25 NUREG/CR4004 SAND 64-2291 NUREG/CR-4091 lEB-8012 NUREG/CR-4005 SAND 64-2305 NUREG/CR-4044 IEB-8242 NUREG-1095 SAND 64-2629 NUREG/CR4096

' A-10229-MS NUREG/CR 3953 SAND 64-2630 NUREG/CR4097 LA-10307-MS NUREG/CR-4079 SAND 64-7139 NUREG/CR3855 LA 10321-MS NUREG/CR4109 SAND 64-7177 NUREG/CR4064 LA-10351-MS NUREG/CR4140 SAND 85-0172 NUREG/CR-4155 LA 10401 MS NUPEG/CR-4225 SAND 85-0209 NUREG/CR.4147 LA-10436-MS NUREG/CR4264 SAND 85-0634 NUREG/CR4197 LA-9700-MS NUREG/CR 3208 SAND 85-7150 NUREG/CR4009

! LA-9985-MS NUREG/CR-4111 SAND 85-7151 NUREG/CR-4010 MEA 2075 NUREG/CR3228 V03 UCID-20092 NUREG/CR4181 V01 ORNL-8135 NUREG/CR-4106 UCRL 20410 NUREG/CR-4263 ORNL-6137 NUREG/CR4114 UCAL 53455 NUREG/CR-3558 ORNL/NOAC-214 NUREG/CR-3551 UCRL-53587 NUREG/CR-4035 ORNL/NSC200 NUREG/CR-2000 V04 N3 WINCO 1024 NUREG/CR-3455 37

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Personal Author index l

This index lists the personal authors of NRC report (s) prepared by the author. If informa-staff and contractor reports in al3habetical tion is needed, refer to the main citation by order. Each name is followec by the the NUREG number.

NUREG number and the title of the AMERS D.W. 8ARANOWSKI P.W.

NUREG/CR4245: IN-PLANT SOURCE TERM MEASUREMENTS AT NUREG 1032 DAFT FC: EVALUATION OF STATON BLACKOUT ACCI-BRUNSWICK STEAM ELECTRIC STATION. DENTS AT NUCLEAR POWER PLANTS.Techrucal Fridmgs Related To Urresolved Safety issue A44 Draft Report For Comment.

NUREG/CR-4091: THE EFFECT OF ALTERNATIVE AGING AND ACCl- BARLETTA,R.E.

DENT SIMULATIONS ON POLYMER PROPERTIES. NUREG/CR4200: BIODEGRADATON TESTING OF SOUDIFIED LOW.

ALPERT D.J.

NUREG/CR-3657: PRELIMINARY SCREENING OF FUEL CYCLE AND SARNES,M.G.

BY-PRODUCT MATERIAL LICENSES FOR EMERGENCY PLANNING NUREG/CR-4118. MONITORING METHODS FOR DETERMINATON NUREG/CR-4199: A DEMONSTRATON UNCERTAINTY / SENSITIVITY COMPUANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT

• ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE URANIUM RECOVERY SITES MODEL CRAC2.

I BARTTER,W.D.

NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PERFORMANCE NURE CR4033: THE ROLE OF PERSONAL AIR SAMPUNG IN RADI-ATION SAFETY PROGRAMS AND RESULTS OF A LABORATORY SIMULATION (MAPPS) MODEL: DFSCR:PTION OF MODEL CONTENT, STRUCTURE,AND SENSITIVITY TESTING.

EVALUATION OF PERSONAL AIR-SAMPUNG EQUIPMENT.

AWS,KI. BASS,5.R.

NUREG/CR4106: PRESSURIZED. THERMAL-SHOCK TEST OF 6-IN, NUREG/CR 3987: COMPUTERIZED ANNUNCIATOR SYSTEMS.

NUREG/CR4220 REUABluTY ANALYSIS OF CONTAINMENT ISOLA. THICK PRESSURE VESSELS PTSE.t investigation Of Warm Prestress-TON SYSTEMS. ing And Upper-Shelf Arrest.

ANDERSON,R.F. BAUGH.J W.

NUREG/CR4237: MOBlUTY OF RADIONUCLIDES IN HIGH CHLORCE NUREG/CH4276 VIBRATION AND WEAR IN STEAM GENERATOR ENVIRCNMENTS. TUBES FOLLOWING CHEMICAL CLEANING . SEMIANNUAL REPORT.

ANDERSON,W.

NUREG-1095. EVALUATION OF RESPONSES TO lE BULLETIN 82- BAUM,J.W.

02.Dogradaten Of Threaded Fasteners in Reactor Coolant Pressure NUREG/CR-3469 V02. OCCUPATONAL DOSE REDUCTION AT NU-Boundary Of Pressunzed Water. Reactor Plants. CLEAR POWER PLANTS. Annotated Behography Of Selected Read-i ings in Radiaton Protection And ALARA.

NUREG/CR 4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL BAXTER D.E.

l WASTE REPOSITORY. Volume 1 Basalt. NUREG/CH-4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES A MSTRONG,&

ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC BOlUNG WATER REACTOR Mac Report NUREG/CR-4194: LOW-LEVEL NUCLEAR WASTE SHALLOW LAND NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES BURIAL TRENCH ISOLATON Final Report,0ctober 1981. September ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC 1984 BOILING WATER REACTOR. Apperdces ARNOLD,W.D.

NUR /CR-4210- MATADOR A COMPUTER CODE FOR THE ANALY-CUDES O A M R '

SIS OF RADIONUCUDE BEHAVIOR DURING DEGRADED CORE AC.

ATTERIDGE.D.G. CIDENTS IN UGHT WATER REACTORS NUREG/CR 3613 V02: EVALUATON OF WELDED AND REPAIR. NUREG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Report for TRANSPORT AND DEPOSITION OF RADONUCLOES) CODE DE-1984. SCRIPTION AND USER'S MANUAL AVCI,H.L BECKHAM,R.

NUREG/CR-4210 MATADOR A COMPUTER CODE FOR THE ANALY. NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER EFF1-SIS OF RADIONUCUDE BEHAVIOR DURING DEGRADED CORE AC- CIENCIES WHEN CHALLENGED WITH THErlMAL. AND AIR JET- I COENTS IN UGHT WATER REACTORS. GENERATED DI 2 ETHYLHEXYL SEBECATE.DI-2-ETHYLHEXYL NUREG/CR-4211: MATADOR (METHOOS FOR THE ANALYSIS OF PHTHALATE,AND SODIUM CHLOROE.

TRANSPORT AND DEPOSITION OF RADONUCLIDES) CODE DE-BEE BE,M.R.

SCRIPTON AND USER'S MANUAL NUREG-0020 V09 N04. UCENSED OPERATING REACTORS STATUS I AZARM,M.A.

SUMMARY

REPORT.Deta As Of March 31,1985(Gray Book I)

NUREG/CR-4229- EVALUATION OF CURRENT METHOOOLOGY EM-BEEDLOW,P.A.

i PLOYED IN PROBABluSTIC RISK ASSESSMENT (PRA) OF FIRE EVENTS AT NUCLEAR POWER PLANTS NUREG/CR-4075. DESIGNING PROTECTIVE COVERS FOR URANIUM

( NUREG/CR4230: PROBABILITY-BASED E% ALUATION OF SELECTED MILL TAluNGS PILES. A Rev ew FIRE PROTECTION FEATURES IN NUCLEAR POWER PLANTS. NUREG/CR4076: DETERMINAflON OF COMPUANCE WITH CRITERIA FOR FINAL TAluNGS DISPOSAL SITE RECLAMATION SALL S.J.

NUREG/CR-3885 V03: HIGH TEMPERATURE GAS-COOLED REACTOR DELL,J.

SAFETY STUDIES FOR THE DIVISON OF ACCOENT NUREG-1127: RADIATON PROTECTION TRAINING AT URANIUM HEX. I EVALUATON Ouarterfy Progress Report, July 1. September 30,1984. AFLUORCE AND FUEL FABRICATON PLANTS.

l l 39

. . l

\

l i 40 Personal Author Index BENNETT,D.E. BRAILE,LW. l NUREG/CR-3657: PRELIM! NARY SCREENING OF FUEL CYCLE AND NUREG/CR-3174 V02. GEOPHYSICAL-GEOLOGICAL STUDIES OF i BY-PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING. POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT I ZONE Annual Report For 1983 J

NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL BRODSKY,A.

WASTE REPOSITORY. Volume 1: Basalt. NUREG 1127: RADIATON PROTECTION TRAINING AT URANIUM HEX.

SEROGREN R.G. AFLUORIDE AND FUEL FABRICATION PLANTS NUREG-1134 RADIATION PROTECTION TRAINING FOR PERSONNEL NUREG/CR-4015; EFFECT OF STAINLESS STEEL WELD OVERLAY EMPLOYED IN MEDICAL FACluTIES.

CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL PLATES IN BENDING SERIES 1. BROOKSHIRE,R.L NUREG/CR-4092- ORNL CHARACTERIZATON OF HEAVY SECTON NUREG/CR-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT-STEEL TECHNOLOGY PROGRAM PLATES 01,02.AND 03. ABluTY PRACTICES BEYER.C1 BROWN.J.

NUREG/CR4168: GT2F.A COMPUTER CODE FOR ESTIMATING NUREG/CR-4190: CAUFORNIA OFFSHORE SURVEY OF UCENSEES UGHT WATER REACTOR FUEL ROO FAILURES. USING RADIOACTIVE MATERIAL BHATTACHARYYA BROWZIN.B.S.

NUREG/CR-4208. GASTROINTESTINAL ABSORPTION OF PLUTONIUM NUREG/CP-0065 TRANSACTIONS OF THE 8TH INTERNATIONAL (N MICE, RATS, AND DOGS Appbcation To Estabbshing Values Of f t CONFERENCE ON STRUCTURE MECHANICS IN REACTOR For Soluble Plutoruum.

TECHNOLOGY. Panel Session J-K. Status of Research in Structural And Mecharucal Engineenng For Nuclear Power Plants NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL BRUEMMER,5.M.

WASTE REPOSITORY. Volume 1. Basalt-NUREG/CR-3613 V02- EVALUATION OF WELDED AND REPAIR.

BLACKMAN,H.S. WELDED STAINLESS STEEL FOR LWR SERVICE Annual Report for NUREG/CR-4040: OPERATONAL DECISIONMAKING AND ACTION SE. 1984 LECTON UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER BRUSK E,S.J.

PLANTS.

NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES BLANO,W.M ON TRANSIENTS AND ACCJDtNTS AT A GENERAL ELECTRIC NUREG/CR4271: RECOMMENDED SAFETY.RELIABluTY,0UAUTY ASSURANCE AND MANAGEMENT AEROSPACE TECHNIQUES WITH NUR CR 6 V2 FFEC OF NTROL SYSTEM FAILURES POSSIBLE APPUCATON BY THE DOE TO THE HIGH LEVEL RADIO- "# #8## C ACTIVE WASTE REPOSITORY PROGRAM. O **'

BRYAN,R.H.

SLONO.R.M'R-4197:

NUREG/C SAFETY GOAL SENSITIVITY STUDIES NUREG/CR4106: PRESSURIZED-THERMAL SHOCK TEST OF 6-IN -

THICK PRESSURE VESSELS PTSE 1 investigation Of Warm Prestress-90CCIO.J.L ing And Upper-Shelf Arrest.

NUREG/CR4229. EVALUATON OF CURRENT METHODOLOGY EM-BRYSON,J.W.

PLOYED IN FAOBABIUSTIC RISK ASSESSMENT (PRA) OF FIRE ,

EVENTS AT NUCLEAR POWER PLANTS NUREG/CR-4106: PRESSURIZED THERMAL. SHOCK TEST OF 6-IN - l NUREG/CR4230 PROBABluTY-BASED EVALUATION OF SELECTED THICK PRESSURE VESSELS PTSE t investigation Of Warm Prestress-

{

FIRE PROTECTON FEATURES IN NUCLEAR POWER PLANTS ing And Upper-Shell Arrest.

NUREG/CR4231: EVALUATION OF AVAILABLE DATA FOR PROBABI-USTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR BUCKALEW,W.H.

POWER PLANTS. NUREG/CR-4147. THE EFFECT OF ENVIRONMENTAL STRESS ON SYLGARD 70 SIUCONE ELASTOMER.

BOEGEL,A.J.

NUREG/CR-3883: ANALYSIS OF JAPANESE-U.S NUCLEAR POWER BURKE,R.P.

PLANT MAINTENANCE. NUREG/CR4197: SAFETY GOAL SENSITIVITY STUDIES.

f 90HN M.P. BURTT.J.D.

NUREG/CR 3558 HANDBOOK OF NUCLEAR POWER PLANT SEISMIC NUREG/CR 3977. RELAP5 THERMAL. HYDRAULIC ANALYSES OF FRAGluTIES. Seismic Safety Margins Research Program. PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B ROBIN-SON UNIT 2 PRESSURIZED WATER REACTOR l BOLANO,J.F.

NUREG/CR-4191: SURVEY OF UCENSEE CONTROL ROOM habit. BUSCH8ACH,T.C.

ABluTY PRACTICES. NUREG/CR4226 NEW MADRID SEISMO ECTONIC STUDY Actrnties l Dunng Fiscal Year 1983 SOLANDER,M.A.

NUREG/CR-3977: RELAP5 THERMAL-HYDRAUUC ANALYSES OF BUSTARD,LD.

PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN. NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCl-SON UNIT 2 PRESSURIZED WATER REACTOR. DENT SIMULATlONS ON POLYMER PROPERTIES DOLT,S.E. BUXTON.LD.

NUREG/CR4106: PRESSURIZED THERMAL SHOCK TEST OF 6-IN.. NUREG/CR 4044. TRAC PF1 LOCA CALCULATIONS USING FINE.

THICK PRESSURE VESSELS PTSE 1. Investigation Of Warm Prestress- NODE AND COARSE-NODE INPUT MODELS.

'"9 CAGLE,R.J.

SONZON,LL NUREG/CR-3905 V02. SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-4095: TEST SERIES 2: SEISMIC-FRAGIUTY TESTS OF FOR UCENSEE EVENT REPORTS Code bstings NATURALLY AGED CLASS 1E EXIDE FHC 19 BATTERY CELLS NUREG/CR-3905 V01 SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-4096. TEST SERIES 3: SEISMIC FRAGluTY TESTS OF FOR LICENSEE EVENT REPORTS Coder's Manual NATURALLY-AGED CLASS 1E CSD LCU-13 BATTERY CELLS NUREG/CR-3905 V04- SEQUENCE CODING AND SEAhCH SYSTEM NUREG/CR-4097. TEST SERIES 4 SEISMIC-FRAGIUTY TESTS OF FOR UCENSEE EVENT REPORTS Coder's Manual NATURALLY-AGED EXIDE EMP-13 BATTERY CELLS.

CAMERON J.R.

BOYACK,8.E. NUREG/CR4131: INVESilGATION OF ALTERNAtlVE MEANS TO AC.

NUREG/CR4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE 1 COMPUSH THE GOALS OF BIENNIAL ON CHAMBER CAUBRA-PRESSURIZED WATER REACTOR. TION i

_ _ -_ . _ - - _ - - _ _ - , _ _ - - - .- -- ._- . - _ ~ . . . - - -

J Personal Author Index 41 CAMP,A.L COE,LJ.

NUREG/CR-3913: HECTR VERSON 10 USER'S MANUAL NUREG/CR-4101: ASSAY OF LONG-LIVED RADONUCUDES IN LOW-LEVEL WASTES FROM POWER REACTORS.

CAMPOELL.R.D.

NUREG/CR-3558: HANDBOOK OF NUCLEAR POWER PLANT SEISMC COLUNS,J.L FRAGIUTIES. Seestmc Safety Margms Research Program. NUREG/CR-3930 OBSERVED BEHAVOR OF CESIUM. LODINE.AND CARUN.F. R '

NUREG/CR4091: THE EFFECT OF ALTERNATNE AGING AND ACCI.

DENT SIMULATIONS ON POLYMER PROPERTIES. COMER,M.K.

CASE,FJ. NUREG/CR4009: HUMAN RELIABluTY DATA BANK.Evaluaton Re-DES S M NER ll' NUREG/CR4010 SPECIFICATION OF A HUMAN REUABILITY DATA BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE.

CASTO,W.R. AR POWER PLANTS.

NUREG/CR-3551: SAFETY IMPUCATONS ASSOCIATED WITH IN-PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS-COMPERE,R TEMS IN NUCLEAR POWER PLANTS. NUREG/CR-3551: SAFETY IMPLCATONS ASSOCIATED WITH IN-PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS-CHAN.S.F. TEMS IN NUCLEAR POWER PLANTS.

NUREG/CR4201: THERMAL STABluTY TESTING OF LOW-LEVEL WASTE FORMS. CONSERE.W.

NUREG/CR4215: TECHNCAL FACTORS AFFECTING LOW-LEVEL NUREG/CR-4192: THE ANALYSIS OF DRAINAGE AND CONSOUDA.

WASTE FORM ACCEPTANCE CRITERIA. TION AT TYPICAL URANIUM MILL TAluNGS SITES.

CHANDRASEKARAN CONDIE.K.G.

NUREG.0017 R01: CALCULATON OF RELEASES OF RADIOACTNE NUREG/CR-3193. FORCED CONVECTNE.NONEOUlUBRIUM. POST.

MATERIALS IN GAS?)US AND UQUID EFFLUENTS FROM PRES- CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATON SURIZED WATER REACTORS (PWR-GALE CODE). COMPARISON REPORT.

CHANG.M.T. CORWIN.W.R.

NUREG/CR4221: AN EVALUATON OF STRESS CORROslON CRACK NUREG/CR4015. EFFECT OF STAINLESS STEEL WELD OVERLAY GROWTH IN BWR PIPING SYSTEMS. CMDOING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL CHEN.J.C. PLATES IN BENDING SERIES 1.

NUREG/CR-3193. FORCED CONVECTNE,NONEOUluBRIUM. POST-CFH RAN COTHAM.B.M.

R EXPERIMENT DATA AND CORRELATION NUREG/CR4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV.

OR DURING STATION BLACKOUT.

CHENION.J.

COVER,LE.

NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCI.

DENT SIMULATONS ON POLYMER PROPERTIES. NUREG/CR3558: HANDBOOK OF NUCLEAR POWER PLANT SEISMIC FRAGluTIES. Seestmc Safety Margms Research Program.

CHOCKlE.A.D.

NUPEG/CR3883 ANALYSIS OF JAPANESE-U S. NUCLEAR POWER CRAFT,C.M.

PLANT MAINTENANCE. NUREG/C43863 ASSESSMENT OF CLASS 1E PRESSURE TRANS.

MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT CHOPRA,0.K. SCREENING TESTS.

NUREG/CR4204 LONG TERM EMBRITTLEMENT OF CAST DUPLEX STAINLESS STEELS IN LWR SYSTEMS Armuel Report. October 1983 CROFF.A.G.

September 1984. NUREG/CR4134. REPOSITORY ENVIRONMENTAL PARAMETERS RELEVANT TO ASSESSING THE PERFORMANCE OF HGH. LEVEL CHOU.C.K. WAS1E PACKAGES.

NUREG/CR4263 REUABluTY ANALYSIS Or STIFF VERSUS FLEXI.

BLE PIPING FINAL PROJECT REPORT. CRONEY,S.T.

NUREG/CR4181: LEACHABluTY OF RADIONUCUDES FROM CHUNG.H.M.

CEMENT SOUDIFIED WASTE FORMS PRODUCED AT OPERATING NUREG/CR4204 LONG-TERM EMBRITTLEMENT OF CAST DUPLEX NUCLEAR POWER REACTORS.

STAINLESS STEELS IN LWR SYSTEMS Armuel Report. October 1963 -

September 1984.

NUREG/CR4245: IN. PUNT SOURCE TERM MEASUREMENTS AT BRUNSWICK STEAM ELECTRIC STATON.

CLAIR 90RNE,H.C.

DAUNO.P.M.

NUREG/CR 4134. REPOSITORY ENVIRONMENTAL PARAMETERS RELEVANT TO ASSESSING THE PERFORMANCE OF HGH-LEVEL NUREG/CR4070 V03: BlVALVE FOUUNG OF NUCLEAR POWER PLANT SERVICE. WATER SYSTEMS Factors That May intensdy The WASTE PACKAGES-Safety Consequences Of Beofoulmg CLAYTOR,7.N.

DANIELSON W.F NT I OF LWR A R 983 NUR Chi 9 VEY OF LICENSEE CONTROL ROOM HABIT.

1984.

CLEVELANO,J.C. DAVIS.C.B.

NUREG/CR3885 V03 HIGH. TEMPERATURE GAS-COOLED REACTOR NUREG/CR3977: RELAP5 THERMAL-HYDRAUUC ANALYSES OF SAFETY STUDIES FOR THE DIVISION OF ACCIDENT PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-EVALUATION Ouarterly Progress Report, July 1. September 30,1984. SON UNIT 2 PRESSURf2ED WATER REACTOR.

CUNE,J.E. DAVIS,T.

NUREG/CR 4101: ASSAY OF LONG-LfVED RADIONUCLIDES IN LOW. NUREG/CR4144 IMPORTANCE RANKING BASED ON AGING CON-LEVEL WASTES FROM POWER REACTORS SIDERATIONS OF COMPONENTS INCLUDED IN PROBABlUSTIC RISK ASSESSMENTS.

NUREG/CR-4076. DETERMINATON OF COMPUANCE WITH CRITERIA DE JARLAIS,0.

FOR FINAL TAIUNGS DISPOSAL SITE RECLAMATION. NUREG/CR4277: INVERTED ANNUAL FLOW EXPERIMENTAL STUDY.

1 i 42 Personal Author Index DEAN.R.S. DUCE,S.W.

NUREG/CR4003 CLOSEOUT OF IE BULLETIN 7944 INCORRECT NUREG/CR4245: IN PLANT SOURCE TERM MEASUREMENTS AT WEIGHTS FOR SWING CHECK VALVES MANUFACTURED BY BRUNSWICA STEAM ELECTRIC STATION

, VELAN ENGINEERING CORPORATION.

, NUREG/CR4004. CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF DUDA.P.M.

WESTINGHOUSE BFD RELAYS IN SAFETY.RELATED SYSTEMS NUREG/CR-2718. STEAM EXPLOSION EXPERIMENTS WITH SINGLE NUREG/CR4005: CLOSEOUT OF IE BULLETIN 80-12. DECAY HEAT DROPS OF IRON OxtDE MELTED WITH A CO2 LASER Part REMOVAL SYSTEM OPERABILITY. 11 Parametnc Studies.

]

j DEARING.J.F. EDLER,S.K.

NUREG/CRat40 DOMINANT ACCOENT SEQUENCES IN OCONEE 1 NUREG/CR-3810 V04 REACTOR SAFETY RESEARCH '

j PRESSURIZED WATER REACTOR. PROGRAMS Ouarterfy Report October-December 1984.

I DECK.D.L EDMONDS,D.P. I NUREG/CR4237: MOBluTY OF RADIONUCLOES IN HIGH CHLORIDE NUREG/CR-4106. PRESSURIZED THERMAL SHOCK TEST OF 6-IN- I ENVIRONMENTS. THICK PRESSURE VESSELS.PTSE 1 Inveshgation Of Warm Prestress-DEFIELD.J.D.

) NUREG/CR-4111: A COMPARATIVE STUDY OF HEPA FILTER EFFl. ELDER,H.K.

j CIENCIES WHEN CHALLENGED WITH THERMAL AND AIR 4ET. NUREG/CR-3293 V01: TECHNOLOGY. SAFETY AND COSTS OF DE.

GENERATED Dl-2-ETHYLHEXYL SEBECATE.DI 2-ETHYLHEXYL COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON-PHTHALATE,AND SODIUM CHLORIDE. FUEL CYCLE FACILITIES FOLLOWING POSTULATED

]

ACCOENTS Main Report DENHAM,0.H. NUREG/CR-3293 V02. TECHNOLOGY, SAFETY AND COSTS OF DE.

NUREG/CR-4118. MONITOR!NG METHODS FOR DETERMINATION COMMISSIONING REFERENCE FUEL CYCLE AND NON-FUEL COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT CYCLE F ACluTIES FOLLOWING POSTULATED I

URANIUM RECOVERY SITES. ACCIDENTS Appendces.

DEWERD,L.A. E LRICK,R.M.

NUREG/CR-4131: INVESTIGATION OF ALTERNAirVE MEANS TO AC. NUREG/CR-3197 V01: REACTON BETWEEN SOME CESIUM-ODINE COMPUSH THE GOALS OF BtENNIAL ION CHAMBER CALIBRA. COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS TON. STEEL,1NCONEL 600 & SILVER. Volume ICesom Hydroude Reac.

OtNGMAN.S.E.

NUREG/CR-3913. HECTR VERSION t 0 USER'S MANUAL EMElGH,C.W.

1 DISALVO,R. NUREG-106$ A01. ACCEPTANCE CRITERIA FOR THE LOW EN-RICHED URANIUM REFORM AMENDMENTS.

NUREG/CR4177 V01: MANAGEMENT OF SEVERE ACCOENTS Perspectives On Managing Severe Accident.s in Commer EMERSON,E.L coal Nuclear Power Plants. NUREG/CR-3611: HADIOACTIVE MATERIAL (RAM) ACCOENT/ INCL. ,

NUREG/CR-4177 V02: MANAGEMENT OF SEVERE DENT DATA ANALYSIS PROGRAM. '

ACCOENTS Extending Plant Operating Procedures into The Severe Accident Regime. ENOERUN,W1 1

NUREG/CR4276 VIBRATON AND WEAR IN STEAM GENCRATOR

] DOSRANICH.D. TUBES FOLLOWING CHEM: CAL CLEANING SEVIANNUAL i NUREG/CR4044 TRAC-PF 1 LOCA CALCULATIONS USING FINE- REPORT.

l NODE AND COARSE-NODE INPUT MODELS.

1 NUREG/CR4 t SS: TRAC-PF t / MOD 1 INDEPENDENT ETTINGER,H.J.

I ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORATED PLATE NUREG/CR-4111: A COMPARATivF STUDY OF HEPA FILTER EFFl.

! CCFL TESTS CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR 4ET-GENERATED Di-2 ETHYLHEXYL SEBECATE.DI-2-ETHYLHEXYL DODSON,K.E- PHTHALATE,AND SODIUM CHLORIDE.

NUREG/CR-3514 V02: THE CHEMICAL BEHAVIOR OF ODINE IN AOVEOUS SOLUTIONS UP TO 150 C.ll Radiahon-Redon Cordihons FAYER.M.J.

I NUREG/CR-4192. THE ANALYSIS OF ORAINAGE AND CONSOUDA.

! DODSON,ML TION AT TYPICAL URANIUM MILL TAluNGS SITES.

NUREG/CR 3906. URANIUM MILL TAILINGS NEUTRAUZATONCONTAM!NANT COMPLEXATON AND TAIUNGS FENTON,0.L

} LEACHING STUDY. NUREGICR-4264' INVESilGATON ON HIGH EFFICIENCY PARTICU.

I LATE AIR FILTER PLUGGING BY COMBUSTON AEROSOLS.

1 DOESOURG,J.M.

NUREG/CR-3747: THE SELECTON AND TESTING OF ROCK FOR AR. FINEMAN,C.P,

? MORING URANIUM TAIUNGS IMPOUNDMENTS NUREG/CR4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES I ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC DONOVAN.M.D. BOILING WATER REACTOR Main Report.

j NUREG/CR4009 HUMAN REUABluTY DATA BANK Evaluation Re. NUREG/CR4262 V02. EFFECTS OF CONTROL SYSTEM FAILURES 4 suits ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC i NUREG/CR4010: SPECIFICATION OF A HUMAN RELIABluTY DATA BOluNG WATER REACTOR. Appendices BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE-l AA POWER PLANTS FLETCHER,C.D.

NUREG/CR 3977 RELAPS THERMAL HYDRAUUC ANALYSES OF

! DOUGHERTY,D.R. PRESSUR12ED THERMAL SHOCK SEQUENCES FOR H D ROBIN-NUREGICR4215 TECHNICAL FACTORS AFFECTING LOW-LEVEL SON UNIT 2 PRESSURIZED WATER REACTOR.

WASTE FORM ACCEPTANCE CRITERIA.

! FOLEY,M.G.

l DOUGLAS,$.C. NUREG/CR 3747: THE SELECTON AND TESTING OF ROCK FOR AR-

+

NUREG/CR 3197 V01: REACTION BETWEEN SOME CESIUM ODINE MORING URANIUM TAIUNGS IMPOUNDMENTS.

COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS l STEELINCONEL 600 & SILVER Volume ICesium Hydronde Reac. FOLEY,W.J.

i tions, NUREG/CR4003 CLOSEOUT OF IE BULLETIN 7904 tNCORRECT WEIGHTS FOR SWING CHECK VALVES MANUFACTURED BY DRISCOLL,J.W. VELAN ENGINEERING CORPORATION NUREG/CR-4191 SURVEY OF LICENSEE CONTROL ROOM HABIT. NUREG/CR 4004 CLOSEOUT OF IE BULLETIN 79 25 FAILURES OF ABluTY PRACTICES WESTINGHOUSE Br0 RELAYS IN SAFETY RELATED SYSTEMS.

l l

4 l

l

)

1

! Personal Author index 43 I NUREG/CR4005: CLOSEOUT OF IE BULLETIN 80-12 DECAY HEAT FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA.

i REMOVAL SYSTEM OPERABILITY. TION AND EXPLOSIVE SHOCK WAVES.

• NUREGICR4264. INVESTIGATION ON HIGH EFFICIENCY PARTICU-FREEMAN,H.D. LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS.

NUREG/CR4076: DETERMINATON OF COMPLIANCE WITH CRITERIA FOR FINAL TAILINGS DISPOSAL SITE RECLAMATON. GRIESS,J.C.

NUREG/CR4134. REPOSITORY ENVIRONMENTAL PARAMETERS j NURE Ik-3863: ASSESSMENT OF CLASS 1E PRESSURE TRANS. ST PC GE 3

MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT SCREENING TESTS. GUNAJt.M.V.

NUREG,CR4264. INVESTIGATON ON HIGH-EFFICIENCY PARTICU-LATE AIR FILTER PLUGGING BY COV8USTION AEROSOLS.

. N EG/C'R-4009 HUMAN REUABILITY DATA BANK Evaluation Re-suRs. GUNDERSEN,G.E.

GALLAHEN.R.B.

NUREG-1065 RO1: ACCEPTANCE CRITERIA FOR THE LOW EN-

, NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM RICHED URANIUM REFORM AMENDMENTS.

j FOR LICENSEE EVENT REPORTS Code Ustings GUYMON,R.H.

NUREG/CR-3905 V03: SEQUENCE CODING AND SEARCH SYSTEM

'j FOR UCENSEE EVENT REPORTS Coders Manual. NUREG/CR-3551: SAFETY IMPUCATIONS ASSOCIATED WITH IN-NUREG/CR 3905 V04: SEQUENCE CODING AND SEARCH SYSTEM PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS-1 FOR UCENSEE EVENT REPORTS Coders Manual. TEMS IN NUCLEAR POWER PLANTS.

NUREG/CH-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM GALLUCCI,R.H. FOR UCENSEE EVENT REPORTS Code Listogs.

NUREG/CR-4220: REUABluTY ANALYSIS OF CONTAINMENT ISOLA. NUREG/CR-3905 V03: SEQUENCE CODING AND SEARCH SYSTEM TON SYSTEMS. FOR UCENSEE EVENT REPORTS. Coders Manual.

i NUREG/CR-3905 V04 SEQUENCE CODING AND SEARCH SYSTEM GANTI.C.S. FOR LICENSEE EVENT REPORTS Coders Manual.

NUREG/CR4231: EVALUATON OF AVAILABLE DATA FOR PROBA81-USTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR HABERMAN,J.H.

POWER PLANTS. NUREG/CR4218: LOCA SIMULATION IN THE NATONAL RESEARCH UNIVERSAL REACTOR PROGRAM Postirradiaton Exammation Re-

'""* * " " ** I' 'C0" #"Y'9"'

NU G 1091: THE EFFECT OF ALTERNATIVE AGING AND ACCI-

DENT SIMULATONS ON POLYMER PROPERTIES- HANEY,LN.

NUREG/CR4040 OPERATIONAL DECISIONMAKING AND ACTION SE.

LECTION UNDER PSYCHOLOGICAL STRESS 6N NUCLEAR POWER I GEE.G.W' NUREG/ CR4076 DETERMINATON OF COMPUANCE WITH PLANTS.

CRITERIA y FOR FINAL TAluNGS DISPOSAL SITE AFCLAMATION 1 GENTIL 1,H. HANSON,R.G.

I NUREG/CR-4077; REACTOR COOLANT PUMP SHAFT SEAL SEHAV, NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES LOR DURING STATION BLACKOUT. ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC BOILING WATER REACTOR Man Report i GENTILLON.C.D. NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES j' NUREG/CR 3862. DEVELOPMENT OF TRANSIENT INITIATING EVENT ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC FREQUENCIES FOR USE IN PROBABfUSTIC RISK ASSESSMENTS BOluNG WATER REACTOR Appendices. -

NUREG/CR4071: EAPLORATORY TREND AND PATTERN ANALYSIS FOR 1981 UCENSEE EVENT REPORT DATA. HARDY,H.A.

NUREG/CR-2531 R03. INTRODUCTORY USER S MANUAL FOR THE GERTMAN,0.l. U S. NUCLEAR REGULATORY COMMISSON REACTOR SAFETY RE-

! NUREG/CR4040- OPERATONAL DECISIONMAKING AND ACTON SE* SEARCH DATA BANK.

! LECTION UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER PLANTS. HARDY,J.E.

NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN-

-4

^

METHODS FOR ESTIMATING RADIOACTIVE AND f TOxfC AIRBORNE SOURCE TERMS FOR URANIUM MILUNG OPER- HARLAN.C.P.

ATIONS. NUREG/CR-3657: PRELIMINARY SCREENING OF FUEL CYCLE AND GOETSCH,S.J. BY DRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING NUREG/CR-4131; INVESTIGATION OF ALTERNATIVE MEANS TO AC- HARRINGTON,K.H.

, COMPUSH THE GOALS OF BIENNIAL ON CHAMBER CAUBRA. NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM TION. FOR LICENSEE EVENT REPORTSCode bstings GOOOWIN,G.M NUREG/CR-3905 V03 SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR5015: EFFECT OF STAINLESS STEEL WELD OVERLAY FO" L NSEE EN R TS sM al l NUpEG/

CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL FOR LICENSEE EVENT REPORTS Coders Manual.

, PLATES IN BENDING SERIES 1.

GOTTULA,R.C HAR M W ,M.

4 NUREG/CR 3193: FORCED CONVECTIVE.NONEOUfuBRIUM. POST. NUREG/CR 3885 V03 HIGH-TEMPERATURE GAS-COOLED REACTOR CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION SAFETY STUDIES FOR THE DIVISION OF ACCIDENT l EVALUATION Ouarterty Progress Report July 1 + Septernber 30.1984.

COMPARISON REPORT.

GRAHAM,E.D. HARTLEY,J.N.

l NUREG/CR 4191: SURVEY OF UCENSEE CONTROL ROOM HABIT, NUREG/CR4088. METHODS FOR ESTIMATING RADCACTIVE AND i ABILITY PRACTICES. TOXIC AIRBORNE SOURCE TERMS FOR URANIUM M: LUNG OPER.

I ATIONS, GREEN,N.M.

i NUREG/CR-3905 V01 R1: SEOUENCE CODING AND SEARCH HELTON.J.C.

i SYSTEM FOR UCENSEE EVENT REPORTS Users Guide. NUREG/CR.3904. A COMPARISON OF UNCERTAINTY AND SENSITIV.

ITY ANALYSIS TECHNIOUES FOR COMPUTER MODELS GREGORY,W.S. NUREG/CR.4199. A DEMONSTRATION UNCERTAINTY / SENSITIVITY

NUREG/CR-4225.

SUMMARY

OF EFFICIENCY TESTING OF STAND- ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE

{ ARD AND HIGH4APACITY HIGH-EFFICIENCY PARTICULATE AIR MODEL CRAC2.

I

---..,,.-...._------.w ww--w-- -

-w-. eg . . - - - - - - ,ep--.--,%-v ,----.------------w-- ,-1.yw, - w.-ee--, -.---w-----. ---y,-, -

44 Personal Author Index HEMAGER,C.H. JAIN H.

NUREG/CR-4070 V03: BlVALVE FOUUNG OF NUCLEAR POWER NUREG/CR-3091 V04 REVIEW OF WASTE PACKAGE VERIFICATION PLANT SERVICE WATER SYSTEMS Factors That May Intensity The TESTS Sermannual Report Covenng The Penod October 1983 March Safety Consequences Of Biotouhng. 1984.

HENN6CK,A. NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFCATION NUREG/CR4001: CLOSEOUT OF IE BULLETIN 79-04.lNCORRECT TESTS.Serniannual Report Covenng The Penod Apnl 1984 Septem.

WEIGHTS FOR SWING CHECK VALVES MANUFACTURED BY W 198L VELAN ENGINEERING CORPORATION.

NUREG/CR4004. CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF JAROSS,R.A.

WESTINGHOUSE BFD RELAYS IN SAFETY RELATED SYSTEMS NUREG/C44180- STATE OF-THE ART OF SOLID-STATE MOTOR NUREG/CR4005: CLOSEOUT OF IE BULLETIN 80-12. DECAY HEAT CONTROLLERS.

REMOVAL SYSTEM OPERABluTY.

JENKINS.J.P.

HENNINGER.R.J.

NUREG/CR4040- OPERATIONAL DECISIONMAKING AND ACTION SE-NUREG/CR4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE-1 LECTON UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER PRESSURIZED WATER REACTOR _

PLANTS.

HENTE.D.B.

JO,J.

NUREG/CR4095: TEST SERIES 2.SEISMC FRAGluTY TESTS OF NATURALLY AGED CLASS 1E EXIDE FHC 19 BATTERY CELLS NUREG/C43703: ASSESSMENT OF SELECTED TRAC AND RELAP5 NUREG/CR4096 TEST SERIES 3 SEISMC FRAGluTY TESTS OF CALCULATONS FOR OCONEE.1 PRESSURIZED THERMAL SHOCK NATURALLY-AGED CLASS 1E C&O LCU-13 BATTERY CELLS STUDY.

NUREG/CR4097. TEST SERIES 4 SEISMIC-FRAGluTY TESTS OF NATURALLY-AGED EXCE EMP 13 BATTERY CELLS. JOHNSON,J.D.

NUREG/C43657; PRELIMINARY SCREENING OF FUEL CYCLE AND HERCZEG.A.L NUREG/CR4237. MOBluTY OF RADIONUCUDES IN HIGH CHLORCE BY-PRODUCT MATERIAL LEENSES FOR EMERGENCY PLANNING.

ENVIRONMENTS. NUREG/CR-4199. A DEMONSTRATON UNCERTAINTY / SENSITIVITY ANALYSIS USING THE HEALTH AND ECONOMC CONSEQUENCE HERSKOVITZ,M.B. MODEL CRAC2.

NUREG/C43651; ASSESSMENT OF THE ADEQUACY OF ORNL IN-STRUMENTATON IN REFLOOD TEST FACluTIES. JOHNSON,KJ.

NUREG/CR4070 V03. BlVALVE FOUUNG OF NUCLEAR POWER HESTER,0.V. PLANT SERVICE-WATER SYSTEMS Factors That May intensify The NUREG/C44071: EXPLORATORY TRENO AND PATTERN ANALYSIS Safety Consequences Of B+ofoulmg.

FOR 1981 LCENSEE EVENT REPORT DATA.

JOHNSON.M.P.

HILL.O.F.

NUREG/CR4088. METHODS FOR ESTIMATING RADCACTIVE AND NUREG/C43905 V01 R1: SEQUENCE CODING AND SEARCH T AJRBORNE SOURCE TERMS FOR URANIUM MILUNG OPER. SYSTEM FOR UCENSEE EVENT REPORTS User's Gude NUREG/C43905 V02: SEQUENCE CODING AND SEARCH SYSTEM FOR LICENSEE EVENT REPORTS Code Ustngs.

HINZE.W.J. NUREG/C43905 V03. SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR.3174 V02: GEOPHYSCAL. GEOLOGICAL STUDIES OF FOR UCENSEE EVENT REPORTS Coder's Manual.

POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT NUREG/C43905 V04. SEQUENCE CODING AND SEARCH SYSTEM ZONE. Annual Report For 1983. FOR UCENSEE EVENT REPORTS Coder's Manual.

HOO88,R.W. JONES,T.N.

NUREG/CR-3872: DATA ACQUISITON AND CONTROL OF THE HSST NUREG/CR4092: ORNL CHARACTERIZATON OF HEAVY-SECTON SERIES V 1RRADIATON EXPERIMENT AT THE ORR. STEEL TECHNOLOGY PROGRAM PLATES 01,02.AND 03.

HORSCHEL.D.S.

JORDAN H.

NUREG/C43647. DESIGN AND FABRICATON OF A 1/8 SCALE STEEL CONTAINMENT MODEL NUREd/CR4205: TRAP-MELT 2 USER'S MANUAL HUCHTON,R.L KAM,F.B.

NUREG/CR-3455. A COMPARISON OF ODINE KRYPTON.AND *ENON NUREG/C44031 V02: NEUTRON SPECTRAL CHARACTERIZATION RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD. FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST) IR-SORPTION MEDtA. RADIATON SERIES. "Neutrorncs Calculahons "

NUREG/CR4031 V03. NEUTRON SPECTRAL CHARACTERIZATION HUENEFELD.J.C-FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST) IR-NUREG/CR-3883 ANALYSIS OF JAPANESE-US. NUCLEAR POWER RADIATION SERIES.

• Neutron Emposure Parameters."

PLANT MAINTENANCE.

KASHOWA,8.A.

UR G/ R 4033: THE ROLE OF PERSONAL air SAMPLING IN RAOl- NURE / 40 ANALYTIC S ES PERTAINING TO STEAM GEN-ATION SAFETY PROGRAMS AND RESULTS OF A LABORATORY ^

EVALUATION OF PERSONAL AAR SAMPUNG EOulPMENT. KASS4R,M.

IMAN,R.L NUREG/CR4221: AN EVALUATON OF STRESS CORROSION CRACK NUREG/C43904 A COMPARISON OF UNCERTAINTY AND SENSITIV. GROWTH IN BWR PIPING SYSTEMS.

ITY ANALYSIS TECHNIOUES FOR COMPUTER MODELS NUREG/CR-4199. A DEMONSTRATION UNCERTAINTY / SENSITIVITY KELLER,G.R.

ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE NUREG/C43174 V02: GEOPHYSICAL GEOLOGICAL STUDIES OF MODEL CRAC2. POSSIBLE EXTENSIONS OF THE NEW MADRIO FAULT ZONE Annual Report For 1983.

NUREGICR 3178. STRUCTURAL AND TECTONO STUDIES IN NEW KELLY,J.E.

YORK STATE. Final Report. July 1981. June 1982.

NUREG/CR-2951: THE D9 EXPERIMENT. Heat Removal From Strahfed ISHil,M. UO2 hs.

NUREG/C44277, INVERTED ANNUAL FLOW EXPERIMENTAL STUDY' KERSCHNEUF, JACO8S,0.K. NUREG/C44111: A COMPARATIVE STUDY OF HEPA FILTER EFFl.

NUREG/CP-0062. PROCEEDINGS OF THE CONFERENCE ON THE AP. CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR-JET-PUCATION OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR GENERATED Dl-2-ETHYLHEXYL SEBECATE.DI-2 ETHYLHEXYL WASTE REPOSITORY ASSESSMENT. PHTHALATE.AND SOOIUM CHLORIDE.

Personal Author index 45 KIM.S.H. LAUGHERY,K.R.

NUREG/CR-3889: THE MODEUNG OF BWR CORE MELTDOWN ACCl- NUREG/CR-4206: A SELECT REVIEW OF THE RECENT ('979-1983)

DENTS - FOR APPLICATON IN THE MELRPIMOO2 COMPUTER BEHAVIORAL RESEARCH UTERATURE ON TRAINING SIMULA-

-CODE. TORS.

KIM8ALL C.S. LEE,J.Y.

NUREG/CR-3747: THE SELECTON AND TESTING OF ROCK FOR AR- NUREG4017 R01: CALCULATION OF RELEASES OF RADIOACTIVE MORING URANIUM TAIUNGS IMPOUNDMENTS. MATERIALS IN GASEOUS AND UOUlO EFFLUENTS FROM PRES-KITTMER C.A. SNED WW MANS WWW ML NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV' LEMEUR.M.

IOR DURING STATION BLACKOUT. NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCl-KNEE.H.E. DENT SIMULATIONS ON POLYMER PROPERTIES NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PERFORMANCE LEONARD M SIMULATION (MAPPS) MODEL- DESCRIPTON OF MODEL NUREG/CR4177 V01: MANAGEMENT OF SEVERE CONTENT, STRUCTURE,AND SENSITIVITY TESTING.

ACCIDENTS. Perspectives On Managmg Severe Accidents in Commer.

KNIGHT,T.D. cel Nuclear Power Plants.

NUREG/CR 3208: TRAC-PD2 DEVELOPMENTAL ASSESSMENT, NUREG/CR-4177 V02: MANAGEMENT OF SEVERE ACCIDENTS Extending Plant Operatino Procedures into The Severe KOEHL.E.R. Accident Regirne.

NUREG/CR4180- STATE OF.THE-ART OF SOUD-STATE MOTOR CONTROLLERS. LEVERENZ,F.

NUREG/CR4144 IMPORTANCE RANKING BASED ON AGING CON-KOENIG J.E. SIDERATIONS OF COMPONENTS INCLUDED IN PROBABluSTIC NUREG/CR4109- TRAC.PF1 ANALYSES OF POTENTIAL PRESSUR- RISK ASSESSMENTS IZED THERMAL SHOCK TRANSIENTS AT CALVERT CUFFS / UNIT 1.A Combustion Engmeenng PWR. LEWELLEN,W.S.

K OH,8.R. NUREG/CR-4158. A COMPILATION OF INFORMATON ON UNCER-TAINTIES INVOLVED IN DEPOSITION MODELING NUREG/CR-3889 THE MODELING OF BWR CORE MELTDOWN ACCI- NUREG/CR-4159- COMPARISON OF THE 1981 'INEL DISPERSION j DENTS - FOR APPUCATON IN THE MELRPl. MOD 2 COMPUTER DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

CODE.

LIDIAK,E.G.

GNER1M NUREG/CR-3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES OF NUREG/CRM31: EVALUATION OF AVAILABLE DATA FOR PROBABI-LISTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT ZONE. Annual Report For 1983.

POWER PLANTS.

KUHLMAN.M.R. LIPlNSKI.R.J.

NUREG/CR-4205: TRAP-MELT 2 USER'S MANUAL NUREG/CR-2951: THE D9 EXPERIMENT. Heat Removal From Stratihed UO2 Debns.

KULAK,R.F.

! NUREG/CR4064 STRUCTURAL RESPONSE OF LARGE PENETRA. LIPPINCOTT.E.P.

TIONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED NUREG/CR-3746 V02. LWR PRESSURE VESSEL SURVEILLANCE DO-TO LOADINGS BEYOND DESIGN BASIS. SIMETRY IMPROVEMENT PROGRAM Semeannual Progress Report.Apnl 1984 September 1984 NUREG/CR-3977: RELAPS THERMAL-HYDRAUUC ANALYSES OF LORENZ,R.A.

PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B ROBIN. NUREG/CR 3930: OBSERVED BEHAVIOR OF CESIUM,10 DINE.AND SON UNIT 2 PRESSURIZED WATER REACTOR TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-NUREG/CR-4196. OVERVIEW OF TRAC-BD1 (VERSION 12) ASSESS- GRAM.

MENT STUDIES.

LOSS,F.J.

KUPPERMAN,D.5-NUREG/CR-3228 V03: STRUCTURAL INTEGRITY OF WATER HEAC-NUREG/CR-4124. NDE OF STAINLESS STEEL AND ON-UNE LEAK TOR PRESSURE BOUNDARY COMPONENTS Annual Report For MONITORING OF LWAS. Annual Report. October 1983 Sep4mber t g84.

1984.

LOYOLA,V.M.

KURTZ,R.

NUREG/CR-3803. THE EFFECTS OF POST LOCA CONDITIONS ON A NUREG/CR-4144- IMPORTANCE RANKING BASED ON AGING CON-SIDERATONS OF COMPONENTS INCLUDED IN PROBABluSTIC PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-DUSTRY, RISK ASSESSMENTS.

'A^ NU EG/CR4263. REUABILITY ANALYS;S OF STIFF VERSUS FLExi.

NU EG R-2531 R03 INTRODUCTORY USER'S MANUAL FOR THE

' BLE PIPING FINAL PROJECT REPORT.

U S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE-SEARCH DATA BANK. LUDWICK,J.D.

LAMEYAT NUREG/CR-4176. EMISSION CONTROL TECHNOLOGY AND OUALITY NUREG/CR-3889: THE MODEUNG OF BWR CORE MELTDOWN ACCI- ASSURANCE NEEDS AT URANIUM MILUNG FACIUTIESIncludes DENTS FOR APPLICATON IN THE MELRPIMOD2 COMPUTER Supporting Methods For Testog.Operatmg.And Mantainmg Air Pollu-CODE. bon Control Devices.

LAMSE,W.M. MACKENZl ,D.R.

NUREG4970: PROCEDURES FOR MEETING NRC ANTITRUST RE. NUREG/CR4215: TECHNICAL FACTORS AFFECTING LOW LEVEL SPONStBluTIES. WASTE FOF.M ACCEPTANCE CRITERIA.

LANNING,D.D. MACKOWIAK,0.P.

J NUREG/CR4168: GT2F.A COMPUTER CODE FOR ESTIMATING NUREG/CR 3862: DEVELOPMENT OF TRANSIENT INITIATING EVENT UGHT WATER REACTOR FUEL ROD FAILURES. FREQUENCIES FOR USE IN PROBABILISTIC RISK ASSESSMENTS.

LARSEN.R.P. MANAHAN,M.

NUREG/CR4208: GASTROINTESTINAL ABSORPTION OF PLUTONIUM NUREG/CR-4177 V01: MANAGEMENT OF SEVERE IN MICE. RATS, AND DOGS.Apphcata To Estabbshmg values Of ff ACCIDENTS Perspectives On Managing Severe Accidents in Commer.

For Soluble Plutoneum. cial Nuclear Power Plants.

l l

, . _ _ , _ . _ __l

46 Personal Author Index MANDLER.J.W. MCGUIRE S.A.

NUREG/CR-4245 IN-PLANT SOURCE TEAM MEASUREMENTS AT NUREG-1140 DRFT FC- A REGULATORY ANALYSIS ON EMERGENCY BRUNSW1CK STEAM ELECTRIC STATION. PREPAREDNESS FOR FUEL CYCLE AND OTHER RADCACTIVE MATERIAL LICENSEES. Draft Report For Comment.

NUREG/CR-4198: FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS- MELBER B.D.

TERS. NUREG/CR-4051: ASSESSMENT OF JOB-RELATED EDUCATONAL MARTIN,R.A. QUALIFICATONS FOR NUCLEAR POWER PLANT OPERATORS.

t NUREG/CR 4264: INVESTIGATON ON HIGH-EFFICIENCY PARTICU- ME NGS,W.J.

LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS.

NUREG/CR-4077. REACTOR COOLANT PUMP SHAFT SEAL BEHAV.

i MAST,P.K. OR DURING STATON BLACKOUT-NUREG/CR-3757: TRAN B-2 THE EFFECT OF LOW STEEL CONTENT N F EL PENETRATON IN A NON-MELTING CYUNDRICAL FLOW "L REG [CR-4015. EFFECT OF STAINLESS STEEL WFLD OVERLAY NUREG/CR-3944: TRAN B-3 EXPERIMENTAL INVESTIGATION OF CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL FUEL CRUST STABluTY ON MELTING SURFACES OF AN ANNU- PLATES IN BENDING SERIES 1.

LAR FLOW CHANNEL NUREG/CR4106: PRESSURIZED. THERMAL SHOCK TEST OF 6-IN.

THICK PRESSURE VESSELS PTSE 1. Investigation Of Warm Prestress.

MATHIEU.G.G. 6ng And Upper-Shelf Arrest.

NUREG/CR-4237: MOBluTY OF RADIONUCUDES IN HIGH CHLORIDE ENVIRONMENTS. MESSIER,M.E.

NUREG-0970- PROCEDURES FOR MEETING NRC ANTITRUST RE.

MAYS,G.T. SPONSieiUTIES NUREG/CR-3905 V01 R1: SEQUENCE CODING AND SEARCH SYSTEM FOR UCENSEE EVENT REPORTS User's Guede METCALF R.

NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4077: REACTOR CCOLANT PUMP SHAFT SEAL BEHAV.

FOR UCENSEE CVENT REPORTS. Code Listings IOR DURING STATON BLACKOUT.

NUREG/CR-3905 V03 SEQUENCE CODING AND SEARCH SYSTEM FOR UCENSEE EVENT REPORTS Coder's Manual. METZGER V.

NUREG/CR-3905 V04 SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-3208: TRAC.PD2 DEVELOPMENTAL ASSESSMENT.

4 FOR UCENSEE EVENT REPORTS Coder's Manual.

WEYER.RE.

NUR R 57: TRAN B-2:THE EFFECT OF LOW STEEL CONTENT 2

ON FUEL PENETRATON IN A NON-MELTING CYUNDRICAL FLOW CL DES ON AN Mt ERA S'I' M ER,U.

NUREG/CR 3944 TRAN 8-3 EXPERIMENTAL INVESTIGATON OF FUEL CRUST STABluTY ON MELTING SURFACES OF AN ANNU- EGG 3872: DATA AWSM M NG & NE MST SERIES V IRRADIATON EXPERIMENT AT THE ORR.

LAR FLOW CHANNEL MILLER,N.E.

MCCLURE.J.D.

i NU9EG/CR-3611: RADIOACTIVE MATERIAL (RAM) ACCIDENT / INCL. NUREG/CR-3900 V03. LONG. TERM PERFORMANCE OF MATERIALS

DENT DATA ANALYSIS PROGRAM. USED FOR HIGH-LEVEL WASTE PACKAGING Quarterty Report. October-December 1984.

MCCONNELL,R.J.

NUREG/CR4191: SURVEY OF UCENSEE CONTROL ROOM HA8IT. MITCHELL,G.W.

ABluTY PRACTICES NUREG/CR-2951: THE 09 EXPERIMENT Heat Removal From Strattfled UO2 Debns.

MCCRAY,J.G.

, NUREG/CR4194. LOW-LEVEL NUCLEAR WASTE SHALLOW LAND MJOLSNESS RC.

BURIAL TRENCH ISOLATON Final Report. October 1981. September NUREG/CR-4079: ANALYTIC STUDIES PE~1TAINING TO STEAM GEN-1984. ERATOR TUBE RUPTURE ACCOENTS.

MCCULLOCH,R.W. MOELLER,M.P.

NUREG/CR4106. PRESSURIZED. THERMAL SHOCK TEST OF 6-IN. NUREG/CR4160: HISTORICAL

SUMMARY

OF OCCUPATONAL RADI-THICK PRESSURE VESSELS PTSE.1 investigation Of Warm Prestress- ATON EXPOSURE EXPERIENCE IN U S COMMERCIAL NUCLEAR eng And Upper-Shelf Arrest POWER PLANTS MCELROY,N.L. MOHR,P.B.

NUREG.1134: RADIATON PROTECTION TRAINING FOR PERSONNEL NUREG/CR-4035. A HIGHWAY ACCOENT INVOLVING RADIOPHAR.

EMPLOYED IN MEDICAL FACIUTIES. MACEUTICALS NEAR BROOKHAVEN.MISSISSIPP1 ON DECEMBER 3,W83 MCELROY,W.N.

NUREG/CR 3746 V02: LWR PRESSURE VESSEL SURVEILLANCE DO. MONT,M.E.

SIMETRY IMPROVEMENT PROGRAM Semiannual Progress NUREG/CR4035: A HIGHWAY ACCIDENT INVOLVING RADIOPHAR.

74 V03 ESSURE VESSEL SU!!VEILLANCE DO. MALEUTICALS NLAR BROOKHAVEN.MISSISSIPP1 ON CECEMBER NUR 3,M83.

SIMETRY IMPROVEMENT PROGRAM.1964 Annual Report. October ,

1,1983 September 30,1964-MOORE.E.B. l MCEWEN,J.E. NUREG/CR 1755 ADD 01: TECHNOLOGY, SAFETY AND COSTS OF DE.

I NUREG/CR-4093: SAFETY / SAFEGUARDS INTERACTIONS DURING COMMISSONING NUCLEAR REACTORS AT MULTIPLE-REACTOR i SAFETY.RELATED EVERGENCIES AT NUCLEAR POWER REACTOR STATIONS Effects On Decomtrussioning Of intenm inability To Dispose FACluTIES. Of Wastes Offsste.

, MCGOWAN.J.J. MOORE.T.O.

NUREG/CR-4086 TENSILE PROPERTIES OF 1RRADIATED NUCLEAR NUREG/CR.3953 THE UGE OF MAG.1 SPECTACLES WITH POSITIVE-GRADE PRES $URE VESSEL WELDS FOR THE THIRD HSST IRRA. AND NEGATIVE-PRESSURE RESPIRATORS.

DIATON SERIES. MORETTI,E.S.

MCOUIRE,M.V. NUREG/CR-4206: GASTROINTESTINAL ABSORPTION OF PLUTONIUM NUREG/CR4139 THE MAILED SURVEY A TECHNIQUE FOR OBTAIN- IN MICE RATS, AND DOGS Apphcation To Estabbshing Values Of f1 ING FEEDBACK FROM OPERATONS PERSONNEL For Soluble Plutonium.

l

l l

l i

i l

Personal Author index 47 MORGENSTERN,M. NOWATZKl.E.A.

NUREG/CR-3883. ANALYSIS OF JAPANESE-U S. NUCLEAR POWER NUREG/CR-4194: LOW-LEVEL NUCLEAR WASTE SHALLOW LAND PLANT MAINTENANCE. BUHIAL TRENCH ISOLATION Final Report. October 1981 - September 1984.

MonsSSEAU.D.S.

NUREG/CR4139 THE MAILED SURVEY.A TECHNIQUE FOR OBTAIN- NOYCE.JA ING FEEDBACK FROM OPERATONS PERSONNEL. NUREG/CR4101; ASSAY 05 LONG.UVED RADONUCUDES IN LOW.

MOTES,0.G. LEVEL WASTES FROM POWER REACTORS.

NUREG/CR-3455 A COMPARISON OF IODINE. KRYPTON.AND XENON 000EN,0.M.

RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD- NUREG/CR 3977; RELAPS THERMAL-HYDRAOUC ANALYSES OF SORPTON MEDIA. PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-SON UNIT 2 PRESSUR:2ED WATER REACTOR MOUL.DA NUREG/CR-4093- SAFETY / SAFEGUARDS INTERACTONS DURING OLDHAMAO.

SAFETY-RELATED EMERGENCIES At NUCLEAR POWER REACTOR NUREG/CR4208. GASTROINTESTINAL ABSORPTON OF PLUTONIUM FACluTIES. IN MICE. RATS, AND DOGS.Applicahon To EstatAsteg values Of f t MULCAMEY,T.P.

NUREG/CR-4180: STATEOF THE-ART OF SOUD STATE MOTOR OLSEN C.S.

CONTROLLERS. NUREG/CR-4084: DRY SPENT FUEL STORAGE TEST PLAN FOR DE.

1 STRUCTIVE FUEL ROD EXAMINATIONS.

NUREG/CR-4160; HISTORICAL

SUMMARY

OF OCCUPATIONAL RADI- OLSON J.

ATON EXPOSURE EXPERIENCE IN U S. COMMERCIAL NUCLEAR NUREG/CR-3883. ANALYSIS OF JAPANESE U.S. NUCLEAR POWER POWER PLANTS. PLANT MAINTENANCE.

I I MURPHY,E opgyg,0.L NUREG4844 DRFT FC; NRC INTEGRATED PROGRAM FOR RESOLU-NUREGICR4908 URANIUM MILL TAluNGS TON OF UNRESOLVED SAFETY ISSUES A-3.A4 AND A-5 HEGARD.

ING STEAM GENERATOR TUSE lNTEGRITY Draft Report For Com- NEUTRALIZATION CONTAMINANT COMPLEXATON AND TAluNGS LEACHING STUDY.

ment.

1. YER8 0 A- OSBORNE.M.F.

NUREG/CR-3747: THE SELECTON AND TESTING OF ROCK FOR AR. NUREG/CR-3930: OBSERVED BEHAVOR OF CES1UM.ODINE.ANJ MORING URANfUM TAILINGS IMPOUNDMENTS. TELLURIUM IN THE ORNL FISSON PRODUCT RELEASE PRO-GRAM.

MALE 2NY,C.L NUREG/CR 3005:

SUMMARY

OF THE NUCLEAR REGULATORY COM' OSTMEYERAM MISSON'S LOFT PROGRAM RESEARCH FINDINGS. NUREG/CR4581k AN APPROACH TO TREATING RADIONUCLIDE DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM MANSTAO,R.K.

R 0 E E "

yg S NT A E Sy A 2951: THE 09 EXPERIMENT. Heat Removal From Stratrhed PLATES IN BENDING SERIES 1. UO2 Ms.

I NUREG/CR-4108. PRESSURIZED. THERMAL SHOCK TEST OF 8-IN.

OUELLETTE.A.L THICK PRESSURE VESSELS PTSE.1 investi9ation Of Warm Prestres>

N And uppeN Anon NUREG/CR-3197 V01

REACTION BETWEEN SOME CESIUM IODINE COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS NASSERSHARIF,0. STEEL,1NCONEL 800 A SILVER Volume ICesium Hydronde Reac-NUREG/CR-4140 DOMINANT ACCOENT SEOUENCES IN OCONEE.1 tions.

PRESSURIZED WATER REACTOR.

OWCZARSKI,P.C.

NELSON,LS. NUREG/CR-3317: TECHNICAL BASES AND USER'S MANUAL FOR NUREG/CR-2718. STEAM EXPLOSON EXPERIMENTS WITH SINGLE THE PROTOTYPE OF SPARC . A SUPPRESSON POOL AEROSOL DROPS OF IRON OXIDE MELTED WITH A CO2 LASER Part REMOVAL CODE.

II Parametnc Studes NELSONAA. NUREG/CH.40tS: EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR-3193: FORCED CONVECTIVE.NONEQUlUBRIUM. POST. CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATON PLATES IN BENDING SERIES t.

COMPARISON REPORT.

PAGEAE.

NETI,S. NUREG/CR 3613 V02' EVALUATION OF WELDED ANO REPAIR.

NUREG/CR.3193: FORCED CONVECTIVE.NONLOUlUBRIUM. POST. WELDED STAINLESS STEEL FOR LWR SERVICE Annual Report for

! CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATON 1984.

PARN ER,0.3.

NICHOLSON,PA NUREG/CR4075 DESIGNING PROTECTIVE COVERS FOR URANIUM NUREG4970 PROCEDURES FOR MEETING NRC ANTITRUST RE- MILL TAluNGS PILES. A Review SPONSIO UTIES.

PARM ER,S.F.

NORWOOO.K.S. NUREG/CR-4 t S9 COMPARISON OF THE 1981 INEL DISPERSION NUREG/CR 3930 OBSERVED BEHAVIOR OF CESIUM.ODINE.AND DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-GRAM. PELTO,P.J.

NOREG/CH.4105 AN ASSESSMENT OF THERMAL GRADIENT TUBE NUREG/CR4220 RELIA 81UTY ANALYSIS OF CONTAINMENT ISOLA.

RESULTS FROM THE HI SERIES OF FISSON PRODUCT RELEASE TlON SYSTEMS.

PtCIULO,P.L NOYlCK,V.J.

NUREG/CR4200 BIODEGRADATION TESTING OF SOUDIFIED LOW.

NUREG/CR4033: THE ROLE OF PERSONAL AIR SAMPUNG IN RADI- LEVEL WASTE STRE AMS. I ATON SAFETY PROGRAMS AND RESULTS OF A LABORATORY NUREG/CR4201: THERMAL STABluTY TESTING OF LOW-LEVEL EVALUATON OF PERSONAL AIR SAMPUNG EQUIPMENT. WASTE FORMS.

4

I i

48 Personal Author index PtLORIM.M K. REST.J.

NUREG/CR4003 SAFETY / SAFEGUARDS INTERACTIONS DURING NUREG/CR.3980 V03 LIGHT. WATER-REACTOR SAFETY FUEL SYS-SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR TEMS RESEARCH PROGRAMS. Quarterty Progress Report. July-Sep-FACIUTIES. tember 1984.

POOOWSKI.M 2.

RHOOES D.S NUREG/CR-3889. THE MODELING OF BWR CORE MELTDOWN ACCI.

D TS FOR APPLICATION IN THE MELRPt MOD 2 COMPUTER NUREd/Ck4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV.

OR DURING STATION BLACKOV,T.

POORE.W.P. RICH,8.L NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4033. THE ROLE OF PERSONAL AIR SAMPUNG IN RADi.

FOR UCENSEE EVENT REPORTS Code Ustngs= ATON SAFETY PROGRAMS AND RESULTS OF A LABORATORY NUREG/CR-3905 V03 SEQUENCE CODING AND SEARCH SYSTEM EVALUATON OF PERSONAL AIR SAMPLING EQUtPMENT, FOR LICENSEE EVENT REPORTS Coder's Manual.

NUREG/CR-3905 V04 SEQUENCE CODING AND SEARCH SYSTEM RIDEOUT.T.S.

FOR LICENSEE EVENT REPORTS Coder's Manual. NUREG/CR.3987: COMPUTERIZED ANNUNCIATOR SYSTEMS POSTMA.A.K. RtGGS R NUREG/CR.3317. TECHNICAL BASES AND USER'S MANUAL FOR THE PROTOTYPE OF SPARC . A SUPPRESSION POOL AEROSOL NUREd.1128. TRIAL EVALUATIONS IN COMPARISON WITH THE 1983 REMOVAL CODE.

SAFETY GOALS PRICE.J.C. RITTER.P.D.

NUREG/CR4131. INVESilGATON OF ALTERNATIVE MEANS TO AC. NUREG/CR4033 THE ROLE OF PERSONAL AIR SAMPLING IN RADI.

COMPUSH THE GOALS OF BtENNIAL ION CHAM 8ER CAliBRA. ATON SAFETY PROGRAMS AND RESULTS OF A LABORATORY TON EVALUATION OF PERSONAL AIR.SAMPUNG EQUIPMENT.

PRINE.D.W. RO90NSON,0.C.

NUREG/CR 4124. NDE OF STAINLESS STEEL AND ON.UNE LEAK NUREG/CR4015: EFFECT OF STA NLESS STEEL WELD OVERLAY MONITOR NG OF LWAS Annual Report,0ctober 1983 . Septemb*f CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STELL 1984 PLATES IN BENDING SERIES 1.

NUREG/CR-4106. PRESSUR12ED. THERMAL. SHOCK TEST OF 6.IN .

NdREG/CR 3703 ASSESSMENT OF SELECTED TRAC AND RELAPS THICK PRESSURE VESSELS PTSE t invest gaton Of Warm Prestress.

CALCULATIONS FOR OCONEE.1 PRESSURilED THERMAL SHOCK '"9^" # E #"

STUDY ROHATGI,U.S.

RAGNURAM.S. NUREG/CR.3703. ASSESSMENT OF SELECTED TRAC AND RELAP5 NUREG/CR-4210 MATADOR A COMPUTER CODE FOR THE ANALY. CALCULATONS FOR OCONEE.1 PRESSURIZED THERMAL SHOCK SIS OF RADONUCLIDE BEHNOR DURING DEGRADED CORE AC. STUDY.

CIDENTS IN UGHT WATER RE ACTORS NUREG/CR42 t t- MATADOR (VETHODS FOR THE ANALYSIS OF MOLLER S.F.

TRANSPORT AND DEPOSITION OF RADIONUCUDES) CODE DE. NUREG/CR.3721 VOI- PRESSURE MEASUREMENTS IN A HYDROGEN SCRIPTION AND USER S MANUAL. COM9USTON ENVIRONMENT. Hydrogen.Aa Combushon Test Seres RAM 4RE2.A.L NUREG/CR4161 V01 CRITICAL PARAMETERS FOR A HIGH-LEVEL ROSS,P,A.

WASTE REPOSITORY. Volume 1 Basart.

NUREG-0020 V09 N04 UCENSED OPERATING REACTORS STATUS RANKIN,W.L

SUMMARY

REPORT Data As Of March 31,1985 (Gray Book 1)

NUREG/CR.3987. COMDUTERIZEO ANNUNCIATOR SYSTEMS RUGE R RANSOM.C.S. NUREG/CR4229 EVALUATION OF CURRENT METHODOLOGY EM NUREG/CR4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES PLOYED IN PROBA8iUSTIC RISK ASSESSMENT (PRA) OF FIRE ON TRANS:ENTS AND ACCIDENTS AT A GENERAL ELECTRC EVENTS AT NUCLEAR POWER PLANTS.

BCluNG WATER REACTOR Mam Report NUREG/CR-4262 V02 EFFECTS OF CONTROL SYSTEM FAILURES MUNK LE,G.E.

ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRC NUREG/CR-3657. PREUMtNARY SCREENING OF FUEL CYCLE AND BOluNG W ATER REACTOR Appordces By. PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING RATHOUN,LA. gaggg,Lg, NUREG/CR4tte MONITORING METHODS FOR DETERMINATION COMPUANCE WITH DECOMVl%AONING CLEANUP CRITERIA AT NUREG/CR-4051: ASSESSMENT OF JOB RELATED EDUCATONAL URANIUM RECOVERY SITES. OVAUFICATIONS FOR NUCLEAR POWER PLANT OPERATORS RE E D,K. A. SAHA,P.

NUREGICR-3953 THE USE OF MAG.1 SPECTACLES WITH POSITIVE, NUREG/CR 3703 ASSESSMENT OF SELECTED TRAC AND RELAPS AND NEGATIVE PRESSURE RESPIRATORS. CALCULATONS FOR OCONEE.1 PRESSURIZED THERMAL SHOCK STUDY.

RE ESE,R.T.

NUREG/CR.3647. DESIGN AND FABRICATION OF A t/8 SCALE SALAZAR,E.A.

STEEL CONTAINVENT MODEL. NUREG/CR-3863 ASSESSMENT OF CLASS 1E PRESSURE TRANS.

MITTER RESPONSE WHEN SUSJECTED TO HARSH ENVIRONMENT NU Ed/CR.4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED AND PRESTRESSED CONCRETE CONTAINMENT. SALLACH.R.A NUREG/CR-4221. AN EVALUATION OF STRESS CORROSION CRACK GROWTH tN OWR PtPtNG SYSTEMS NURM/CR 3197 V01 REACTION BETWEEN SOME CESIUM ODINE COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS REMEC.L STEEL,1NCONEL 600 & SILVER Vooume ICesium Hydrorde Reac.

NUREG/CR4031 V02. NEUTRON SPECTRAL CHARACTERllATON tons FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST)IR.

R ADIATON SER!ES "Neutroruct Calculatons" SAMANTA.P.K.

NUREG/CR.4031 V03 NEUTRON SPECTRAL CHARACTER 12ATON NUREG/CR 4231: EVALUATON OF AVAILABLE DATA FOR PROBABl.

FOR THE FIFTH HEAVY SECTON ST[EL TECHNOLOGY (HSST)IR. USTIC RISK ASSFSSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR RADIATON SER!ES " Neutron Esposure Parameters POWER PLANTS

\

i Personal Author index 49 i

SCHNITZLER,s.G.

FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURl2A-NUREG/CR-4203. A CALCULATONAL METHOD FOR DETERMINING TION AND EXPLOSIVE SHOCK WAVES BIOLOGCAL DOSE RATES FROM IRRADIATED RESEARCH REAC.

j TOR FUEL SMITH,R.C.

NUREG/CR-4109 TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR-N REC /C .3317. TECHNICAL BASES AND USER S MANUAL FOR 1 st E R THE PROTOTYPE OF SPARC A SUPPRESSION POOL AEROSOL REMOVAL CODE goo,p, SCHWARTZ,M.W. NUREG/CR-3091 V04 REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR-4035: A HIGHWAY ACCIDENT INVOLVING RADIOPHAR. TESTS Sermannual Report Covenng The Penod October 1983 March M EUTICALS NEAR BROOKHAVEN.MISSISSIPPt ON DECEMBER 98 p

i TESTS Sermannual Report Covenng The Penod Apnl 1984 . Septem.

, SCHWE12ER,R.L ber 1984.

NUREG/CR 4093. SAFETY / SAFEGUARDS INTERACTONS DURING SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR SOONG,A.L FACILITIES. NUREG 1127; RADIATION PROTECTON TRAINING AT URANIUM HEX.

j ,,,,g AFLUORIDE AND FUEL FABRICATION PLANTS-l NUREG-1128 TRIAL EVALUATONS IN COMPARISON WITH THE 1983 yALETTO,M.L

SAFETY GOALS-NUREG/CR 4208. GASTROINTESTINAL ABSORPTON OF PLUTONIUM I SERNE,RJ. !N MICE. RATS, AND COGS Application To EstabitsNng Values Of II NUREG/CR 3906. URANIUM MILL TAILINGS For Soluble Plutonsum.

I NEUTRALI2ATON CONTAMINANT COMPLEXATON AND TAILINGS I

LEACHING STUDY. SPRl0GS,0.D.

SECVER,W.L.

NUREG/CR4109 TRAC-PFt ANALYSES OF POTENTIAL PRESSUR-i IZED. THERMAL SHOCK TRANSIENTS AT CALVFRT CLIFFS / UNIT i WUREG/CR4212- IN. PLACE THERMAL ANNEALING OF NUCLEAR RE- 1.A Combustion Ergneenng PWR.

l ACTOR PRESSURE VESSELS.

STAHL,0.

SHACK,WJ.

WUREG/CR 3998 V02 LIGHT-WATER REACTOR SAFETY MATERIALS NUREG/CR-3900 V03 LONG TERM PERFORMANCE OF MATERIALS ENGINEERING USED FOR HIGRLEVEL W ASTE PACKAGING Ouarterly

, RESEARCH PROGRAMS Quarterly Progress Repc4 October December 1984 i Report.ApnbJune 1984.

SHOCKELFORD,M. STALLMANN,F.W.

NUREG/CR-3855: CHARACTEHi2ATION OF NUCLEAR REACTOR NUREG/CR-4031 V03 NEUTRON SPECTRAL CHARACTERt2ATON CONTAINMENT PENETRATION FINAL REPORT. FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST)IR-RADIATON SERIES " Neutron Exposure Parameters

• SHAFAGH1,A.

NUREG/CR4144. IMPORTANCE RANKING BASED ON AGING CON. STEL2 MAN,WJ.

l SiOERATONS Or COMPONENTS INCLUDED IN PROBA8ILISTIC NUREG/CR4092 ORNL CHARACTERIZATON OF HEAVY SECTION RISK ASSESSMENTS- STEEL TECH, OLOGY PROGRAM PLATES Ot.02.AND 03 SHARMA,S.

STERNER,P.

NUREG/CR4149 ULTIMATE PRESSURE CAPACITY OF RE!NFORCED NUREG-1095. EVALUATON OF RESPONSES TO IE BULLETIN 82-NUREG 422 E ATl 0 E S CORROSION CRACK GROWTH IN B AR PtPING SYSTEMS 8 ary Of Pre ed W er React ns SHE A.C.E. STETZENSACH,K.

NUREG/CR4200- BIODEGRADATON TESTING OF SOLIDIFIED LOW. NUREG/CR-4194 LOW-LEVEL NUCLEAR WASTE SHALLOW LAND LEkEL WASTE STREAMS. BURIAL TRENCH ISOLATION Final Report.Oc.tober 198t - Septernber 1984.

SHENG,Y.P.

NUREG/CR4158 A COMPILATON OF INFORMATION ON UNCER- STITT,0.0.

TAINTIES INVOLVED IN DEPOSITON MODELING NUREG/CH-3971, RELAPS THENMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B RO8iN-UR G/ R-3626 V02. MAINTENANCE PERSONNEL PERFORMANCE SIMULATON (MAPPS) MODEL: DESCRIPTON OF MODEL STOETZEL,0.A.

CONTENT, STRUCTURE.AND SENSITIVITY TESTING.

NUREG/CR4160' HISTORICAL

SUMMARY

OF OCCUPATIONAL RADb

$4MPSON,F.3, ATION EXPOSURE EXPERIENCE IN U S COMMERCIAL NUCLEAR NUREG/CR4245 IN PLANT SOURCE TERM MEASUREMENTS AT POWER PLANTS.

BRUNSWICK STEAM ELECTRIC STAflON SfMPSON,HJ.

NUREG/CR4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR-4231. MOBILITY OF RADIONUCLIDES IN HIGH CHLORIDE ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC ENVIRONMENTS BOLLING WATER REACTOR Main Report.

NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAtLURES

^ ^ " "

" ^

N EG/CR4 t34 REPOSITORY ENVIRONMENTAL PARAMETERS DOI NG EACTOR~

RELEVANT TO ASSES $iNG THE PERFORMANCE OF HIGH LEVEL WASTE PACKAGES.

SUNOARUM.M.K.

CMITH.K.L. NUREG/CR 3191 FORCED CONVECTIVE.NONEOUILlBRIUM. POST.

N~) REG /CR-3862: DEVELOPMENT OF TRANSIENT INITIATING EVENT CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION FREQUENCIES FOR USE IN PROBABILISTIC RISK ASSESSMENTS COMPARISON REPORT.

SMITH.P.M. SUTTON,0.E.

NUREG/CR422i

SUMMARY

OF EFFICIENCY TESTING OF STAND. NUREG/CR42a3. STUDY OF THE EFF ECTS OF ELASTIC UNLOAD-ARD AND HIGH CAPACITY HIGH EFFICIENCY PARTICULATE AIR INGS ON THE #R CURVES FROM COMPACT SPECIMENS.

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

l 50 Personal Author index SWAIN,R.L WALSH,M.E. j NUREG/CR4015: EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR4139: THE MAILED SURVEY.A TECHNIOUE FOR 00TAIN-CLADDtNG ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL ING FEEDBACK FROM OPERATONS PERSONNEL PLATES IN SENDING SERIES 1.

WALTERS,W.H.

SYKES,R.L NUREG/CR4076: DETERM: NATION OF COMPUANCE WITH CRITERIA NUREG/CR-4159: COMPARISON OF THE 1981 INEL DISPERSON FOR FLNAL TAluNGS DLSPOSAL SITE RECLAMATON DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

WANG Y.K.

TALEYARKHAN,R.

NUREG/CR-4149: ULTIMATE PRESSURE CAPACITY OF REINFORCED NUREG/CR-3889. THE MODEUNG OF BWR CORE MELTDOWN ACCl-AND PRESTRESSED CONCRETE CONTAINMENT.

DENTS . FOR APPUCATON IN THE MELRPl. MOD 2 COMPUTER WATERMAN,M.E.

NUREG/CR-3977; RELAP5 THERMAL HYDRAULO ANALYSES OF TASAO.L NUREG/CR4161 VOI: CRITICAL PARAMETERS FOR A HIGH-LEVEL PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-WASTE REPOSITORY. Volume 1 Basalt. SON UNIT 2 PRESSURlIED W ATER REACTOR.

TAYLORJ.T. WESER,C.F.

i NUREG/CR-3455: A COMFARISON OF COINE, KRYPTON.AND XENON NUREG/C43885 V03. HIGH-TEMPERATURE GAS-COOLED REACTOR RETENTON EFFCIENCIES FOR VARCUS SILVER LOADED AD- SAFETY STUDIES FOR THE DIVISION OF ACCIDENT SORPTION MEDIA. EVALUATON Ouarterty Progress Report, July 1. September 30,1984.

8L WESSTER C.S.

NUREG/C'R-3820 V03: THERMAL /HYDRAUUC ANALYSIS RESEARCH NUREG/CR 3930- OBSERVED BEHAVOR OF CESIUM. LODINE.AND PROGRAM.Ouarterty ReportJuly-September 1984. TELLURIUM (N THE ORNL FISSION PRODUCT RELEASE PRO-GRAM.

THOMPSON.V.M.

NUREG/CR-4191: SURVEY OF LCENSEE CONTROL ROOM HABIT. WEILANOlCS,C.

A81UTY PRACTICES.

NUREG/C43469 V02: OCCUPATIONAL DOSE REDUCTON AT NU-THOMS.K.R. CLEAR POWER PLANTS Annotated Bibhography Of Selected Read-NUREG/CR4106. PRESSURf2ED. THERMAL-SHOCK TEST OF 6-4N.. angs in Radiate Protecta And ALARA.

THICK PRESSURE VESSELS PTSE.1. Invest gatm Of Warm Prestress.

rg And Upper.Shel' Arrest. WEISS.A.J.

NUREG/CP-0059 V01: PROCEEDINGS OF THE MITI-NRC SEISMIC IN-TMACHYKJ.W. FORMATION EXCHANGE MEETING. VOLUME I.

NUREG/CR 3455: A COMPARISON OF IODINE. KRYPTON.AND XENON NUREG/CR-2331 VO4 N3 SAFETY RESEARCH PROGRAMS SPON-RETENTON EFFCIENCIES FOR VARIOUS SILVER LOADED AD- SORED BY OFFCE OF NUCLEAR REGULATORY SORPTON MEDIA RESEARCH Ouarterty Progress Report. July 1. September 30,1984.

TOALSTON,A L NUREG/CR-2331 V04 N4: SAFETY RESEARCH PROGRAMS SPON.

SORED BY OFFCE OF NUCLEAR REGULATORY NUREG-0970: PROCEDURES FOR MEETING NRC ANTITRUST RE- RESEARCH Ouarterty Progress Report. October 1. December 31, SPONSIBluTIES. 1984 TOTH,LM. ]

NUREG/CH-3514 V02- THE CHEMICAL BEHAVOR OF IODINE IN WENSELR.G.

i AOUEO'JS SOLUTIONS UP TO 150 C.ll.RadiatmorvRedon Condites. NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV. ,

OR DURING STATON BLACKOUT. '

TRAvis,J.R.

NUREG/CR 3930 OBSERVED BEHAVOR OF CESIUM.ODINE.AND WERES,0.

TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO- NUREG/CR4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL GRAM. WASTE REPOSITORY. V sume 1 Basalt 7

TRIER.R.M. WESLEY,0.A.

NUREG/CR-4237. MOBluTY OF RADIONUCUDES IN HIGH CHLORIDE NUREG/C43558: HANDOOOK OF NUCLEAR POWER PLANT SEISMIC ENVIRONMENTS- FRAGIUTIES. Sersnuc Safety Margins Rosearch Program.

TRIGOS,T.J. *E I NUREG/CR.3907. COMPUTERIZED ANNUNCIATOR SYSTEMS' NUREd C43913. HECTR VERSION 1.0 USER'S MANUAL VARMA,A.K.

WHATLEY,5.K.

NUREG/CR4158. A COMPILATON OF INFORMATON ON UNCES TAINTIES INVOLVED IN DEPOSITON MODEUNG. NUREG/CP-0062: PROCEEDINGS OF THE CONFERENCE ON THE AP.

PUCATON OF GEOCHEMCAL MODELS TO HIGRLEVEL NUCLEAR VASLOW,F. WASTE REPOSITORY ASSESSMENT.

NUREG/CR4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL WASTE FORM ACCEPTANCE CRITERIA. WHITE.A.S.

NUREG/C44051 ASSESSMENT OF JOO-RELATED EDUCATIONAL

' ^

N CR'4 3 STUDY OF THE EFFECTS OF ELASTC UNLOAD-INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS- WHITMAN,0.D.

NUREG/CR-4106: PRESSURIZEO. THERMAL. SHOCK TEST OF 6-IN.

VEAKIS.E. THICK PRESSURE VESSELS PTSE.t investagahon Of Warm Prestress-NUREG/CR 3091 V04: REVIEW OF WASTE PACKAGE VERIFCATION mg And Upper-Shett Arrest

! TESTS.Sermannual Report Covenng The Penod October 1963 March WILUAU8

NUREG/CR4091 V05. REVIEW OF WASTE PACKAGE VERIFICATION NUREG/C44031 V02: NEUTRON SPECTRAL CHARACTER 12ATION TESTS Sermannual Report Covenng The Penod Apnf 1964 Septem.

FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST) 14 ber 1964~ RADIAflON SERIES. "Neutrorucs Calculations.

VINCENT,S.G.

WILUFORO,R.E. ,

I NUREGICR4231: EVALUATON OF AVAILAELE DATA FOR PROBA81 '

USTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR NUREG/C44168 GT2F.A COMPUTER CODE FOR ESTIMATING POWER PLANTS. LIGHT WATER REACTOR FUEL ROD F AILURES l

\

Personal Author Index 51 WILLIS C.A. WONG,K.S.

NUREG-0017 Rot: CALCULATON OF RELEASES OF RADCACTIVE NUREG/CR-4190- CAUFORNIA OFFSHORE SURVEY OF UCENSEES MATERIALS IN GASEOUS AND LOUlO EFFLUENTS FROM PRES. USING RADIOACTNE MATERIAL SURIZED WATER REACTORS (PWR-GALE CODE). ,g wetgoggj,g, NUREG 1131: FINANCIAL ANALYSIS OF POTENTIAL RETROSPECTIVE NUREG/CR 3885 V03: HIGH-TEMPERATURE GAS-COOLED REACTOR PREMlUM ASSESSMENTS UNDER THE PRICE AP.'DERSON SAFETY STUDIES FOR THE DIVISION OF ACCOENT SYSTEM.

EVALUATON Ouarterly Progress Report, July 1 - September 30,1984.

WREATHALLJ.

ME,CJ- NUREG/CR-4177 V01: MANAGEMENT OF SEVERE ACCIDENTS Perspectrwes On Managing severe Acodents in Commer NUREG 1065 RO1: ACCEPTANCE CRITERIA FOR THE LOW EN- cel Nuclear Power Plants.

, RICHED URANIUM REFORM AMENDMENTS. NUREG/CR-4177 V02: MANAGEMENT OF SEVERE ACCIDENTS Extending Plant Operahng Procedures into The Severe WOLFJJ. Accsdent Regime.

NUREG/CR-3626 V02: MAINTENANCE PERSONNEL PERFORMANCE SIMULATON (MAPPS) MODEL: DESCRIPTON OF MODEL WRIGHT,K.W.

CONTENT. STRUCTURE,AND SENSITIVITY TESTING. NUREG/CR-4101: ASSAY OF LONG-UVED RADONUCLIOES IN LOW-LEVEL WASTES FROM POWER REACTORS.

WOLLENGERG,H.A.

^^ "^" l WASTE SIT Y 18asa NU dR-4147: THE EFFECT OF ENVIRONMENTAL STRESS ON SYLGARD 70 SluCONE ELASTOMER.

WORIELSOUFF,JL

' yonggg,A, NUREG/CR-3803: THE EFFECTS OF POST-LOCA CONDITONS ON A NUREG/CR-4118 MONITORING METHOOS FOR DETERMINATON PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-i DUSTRY. COMPUANCE WITH DECOMMISSONING CLEANUP CRITERIA AT URANIUM RECOVERY SITES, I WONG,C.N. YOUNG,T.E.

NUREG/CR-4044: TRAC-PF1 LOCA CALCULATONS USING FINE. NUREG/CR-4245: IN-PLANT SOURCE TERM MEASUREMENTS AT NODE AND COARSE-NOOE INPUT MODELS. BRUNSWICK STEAM ELECTRIC STATION.

t 4

1 J

l l

l l

l i

i

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-n----.-- -.- -. _, -- -._- , , , , - - - - , ., ,,m,,_ m, ,,,- , - - , , , - -,r-_,---- ,,_ -- -- , , ,_,, , , , - . _ . _ ,,_,,_

Subject index This index was developed frorn keywords moved later when a reasonable thesaurus end 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.

AC Pourer NUREG/CR-3855 CHARACTERIZATION OF NUCLEAR REACTOR NUREG-1032 DAFT FC: EVALUATON OF STATION BLACKOUT ACCI- CONTAINMENT PENETRATON FINAL REPORT.

DENTS AT NUCLEAR POWER PLANTS. Technical FindinOs Related To NUREG/CR-3889. THE MODEUNG OF BWR CORE MELTDOWN ACCI-Unresolved Safety issue A-44 Draft Report For Comment. DENTS . FOR APPUCATION IN THE MELRPIMOD2 COMPUTER CODE.

ACMS NUREG/CR-4035: A HIGHWAY ACCIDENT INVOLVING RADIOPHAR-NUREG-1125 vot: A COMPILATON OF REPORTS OF THE ADVISORY MACEUTICALS NEAR BROOKHAVEN, MISSISSIPPI ON DECEMBER COMMITTEE ON REACTOR SAFEGUARDS.19571984. Volume 1,Part 3,1983.

NU 1 COM T PORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 19571984. Volume 2,Part 0ENT SIMULA I N LY ER PROPER ES NUREG/CR4169 AN APPROACH TO TREATING RADIONUCUDE

, NURE 1 03 LAT R PORTS OF THE ADVISORY DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM.

COMMITTEE ON REACTOR SAFEGUARDS *1957-1984. Volume 3.Part NUREG/CR-4197: SAFETY GOAL SENSITIVITY STUDIES 1:ACRS Reports On Protect Reviews (0 Z) NUREG/CR4210: MATADOR A COMPUTER CODE FOR THE ANALY-NOREG-1125 V04: A COMPILATON OF REPORTS OF THE ADVISORY SIS OF RADIONUCUDE BEHAVOR DURING DEGRADED CORE AC-COMMITTEE ON REACTOR SAFEGUARDS,19571984. Volume 4,Part COENTS IN UGHT WATER REACTORS 2.ACRS Reports On Genenc Sublects (Accdent Analysis Genenc NUREG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF ltems). TRANSPORT AND DEPOSITION OF RADIONUCUDES) CODE DE-NUREG-1125 V05: A COMPfLATON OF REPORTS OF THE ADVISORY SCRIPTION AND USER'S MANUAL COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 5,Part NUREG/CR4282 V01: EFFECTS OF CONTROL SYSTEM FAILURES 2.ACRS Reports On Genenc Sub,ects (HTGR Regulatory Gudes) ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC NUREG-1125 V06. A COMPILATON OF REPORTS OF THE ADVISORY BOILING WATER REACTOR Main Report.

COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 6,Part NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES i 2.ACRS Reports On Genenc Sublects (RPA . Appendru C). ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC J

BOluNG WATER REACTOR. Appendices NUREG/CR-3293 V01: TECHNOLOGY, SAFETY AND COSTS OF DE- Accident Analys6s COMMISSONING REFERENCE NUCLEAR FUEL CYCLE AND NON- NUREG 1125 V04 A COMPILATON OF REPORTS OF THE ADVISORY FUEL CYCLE FACIUTIES FOLLOWING POSTULATED COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 4,Part NUR CR 3293 V02 CHNOLOGY, SAFETY AND COSTS OF DE* m '

l COMMISSONING REFERENCE FUEL CYCLE AND NON-FUEL CYCLE FACIUTIES FOLLOWING POSTULATED Acc6 dent _..,..t NUR 6 NUPEG/CR_[tT7 V01: MANAGEMENT OF SEVERE CCUPATONAL DOSE REDUCTON AT NU-CLEAR POWER PLANTS. Annotated Bibliography Of Selected Read- ACCIDENTS Perspectives On Managing Severe Acceents in Commer ings in Radiation Protection And ALARA-NU C 1 V2 MANAGEMENT OF SEVERE

ASME Code ACCOENTS Extending Plant Operating Procedures into The Severe i

NUREG/CR-3847: DESIGN AND FABRICATON OF A 1/8-SCALE Accdent Regime.

! STEEL CONTAINMENT MODEL Abnormal Occurrence NUREG/CR4140. DOMINANT ACCIDENT SEQUENCES IN OCONEE 1 NUREG4090 V07 NO3: REPORT TO CONGRESS ON ABNORMAL PRESSURIZED WATER REACTOR.

OCCURRENCES - tomber 1984.

I NUREG-0090 V07 . EPORT TO CONGRESS ON ABNORMAL Acid Weste OCCURRENCES October-December 1984. NUREG/CR 3906. URANIUM MILL TAlltNGS NEUTRAlf2ATIONCONTAMINANT COMPLEXATION AND TAluNGS U 1037 DAFT FC: CONTAINMENT PERFORMANCE WORKING GROUP REPORT Draft Report For Comment Acoustic Em6eelon i NUREG-1118: A REVIEW OF THE CURRENT UNDERSTANDING OF N RA RESEARCH FOR THE POTENTIAL FOR CONTAINMENT FAILURE FROM IN VESSEL STEAM EXPLOSONS THE MATERIALS ENGINEERING BRANCH.DfVISION OF ENGINEER.

NUREG/CR-3293 V01: TECHNOLOGY, SAFETY AND COSTS OF DE. ING TECHNOLOGY. Annual Report For FY 1984 COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON- * '

j FU CYCLE CiUTIES FOLLOWING POSTULATED N G E OF STAINLESS STEEL AND ON.LINE LEAK NUREG/CR-J293 V02: TECHNOLOGY, SAFETY AND COSTS OF DE- MONITORING OF LWRS. Annual Report. October 1983 - September COMMISSIONING REFERENCE FUEL CYCLE AND NON. FUEL 1984.

CYCLE FACluTIES FOLLOWING POSTULATED ACCOENTS Appendices. Adeorption NUREG/CR-3317. TECHNICAL BASES AND USER'S MANUAL FOR NUREG/CR4237. MOBluTY OF RADIONUCLIDES IN HlGH CHLORIDE THE PROTOTYPE OF SPARC A SUPPRESSION POOL AEROSOL ENVIRONMENTS.

REMOVAL CODE.

NUREG/CR-3611: RADIOACTIVE MATERIAL (RAM) ACCOENT/ INCL. Adeorption Media DENT DATA ANALYSIS PROGRAM. NUREG/CR-3455: A COMPARISON OF OOINE. KRYPTON.AND XENON NUREG/CR-3818 V02: REACTOR SAFETY RESEARCH Ouarterty RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD-Report Apni-June 1984. SORPTON MEDIA.

53

. _ _ = . , - - - ._ _

i 1

, 54 Subject index i

Aerosol NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL 1 NUREG/CR4111: A COMPARATIVE STUDY OF HEPA FILTER EFFI- WASTE REPOSITORY. Volume 1 Basalt C1ENCIES WHEN CHALLENGED WITH THERMAL AND AIR JET-I GENERATED DI 2-ETHYLHEXYL SEBECATE.Ol-2 ETHYLHEXYL Seta Dose Rate l PHTHALATE.AND SOOlUM CHLORIDE. NUREG/CR-4203 A CALCULATIONAL METHOD FOR DETERMINING NUREG/CR4205: TRAP-MELT 2 USER'S MANUAL BIOLOGICAL DOSE RATES FROM lRRADIATED RESEARCH REAC-NUREG/CR4210: MATADOR A COMPUTER CODE FOR THE ANALY- TOR FUEL SIS OF RADIONUCUDE BEHAVIOR DURING DEGRADED CORE AC-CIDENTS IN UGHT WATER REACTORS. 560 degradation NUREG/CR4264: INVESTIGATION ON HIGH-EFFICIENCY PARTICU- NUREG/CR4200: BODEGRADATON TESTING OF SOUDIFIED LOW.

LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS. LEVEL WASTE STREAMS.

Aging savalve Foul 6ng

, NUREG/CR-2331 V04 N4: SAFETY RESEARCH PROGRAMS SPON- NUREG/CR4070 V03. BlVALVE FOULING OF NUCLEAR POWER SORED BY OFFICE OF NUCLEAR REGULATORY PLANT SERVCE-WATER SYSTEMS Factors That May Intensify The RESEARCH.Ouarterty Progress Report, October 1 - December 31. Safety Consequences Of Biotouhng 1984.

NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCI- Slowdown 4 DENT SIMULATONS ON POLYMER PROPERTIES. NUREG/CR-4196 OVERVIEW OF TRAC-BD1 (VERSION 12) ASSESS-NUREG/CR4144: IMPORTANCE RANKING BASED ON AGING CON- MENT STUDIFS SIDERATIONS OF COMPONENTS INCLUDED IN PROBABluSTIC RISK ASSESSMENTS. Bomb Threat NUREG4525 RIO: SAFEGUARDS

SUMMARY

EVENT UST Air Cleaning System (SSEL). REVISION 10' i NUREG/CR4191: SURVEY OF LICENSEE CONTROL ROOM HABIT-

! ABluTY PRACTICES. Burial NUREG/CR4225.

SUMMARY

OF EFFICIENCY TESTING OF STAND- NUREG/CR4194. LOW LEVEL NUCLEAR WASTE SHALLOW LAND

, ARD AND HIGH CAPACITY NfGH-EFFICIENCY PARTICULATE AIR BURIAL TRENCH ISOLATION Final Report' October 1981 September FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA- 1984' TON AND EXPLOSIVE SHOCK WAVES.

Burled Weste AirSa @ NUREG/CR4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL NUREG/CR-3455: A COMPAR! SON OF ODINE, KRYPTON,AND XENON WASTE FORM ACCEPTANCE CRITERIA.

f RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD-SORPTON MEDIA. By-Produd Mmal Fmy Airborne Releases NUREG/CR 3657. PRELIM 5 NARY SCREENING OF FUEL CYCLE AND NUREG/CR4088. METHODS FOR ESTIMATING RADIOACTIVE AND BY-PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING T XIC AIRBORNE SOURCE TERMS FOR URANtUM MILLING OPER. C&D LCU-13 Battery Cell

]

NUREG/CR4096: TEST SERIES 3 SEISMC FRAGluTY TESTS OF i

j Analyele NATURALLY AGED CLASS 1E C&D LCU 13 BATTERY CELLS.

NUREG-1140 DAFT FC: A REGULATORY ANALYSIS ON EMERGENCY PREPAREDNESS FOR FUEL CYCLE AND OTHER RADIOACTIVE CDA MATERIAL UCENSEES Draft Report For Comment. NUREG/CR-3944. TRAN 0-3 EXPERIMENT AL INVESTIGATION OF

, FUEL CRUST STABlUTY ON MELTING SURFACES OF AN ANNU.

' Annual Report LAR FLOW CHANNEL a NUREG 1145 V01: U S. NUCLEAR REGULATORY COMMISSON 1984 1 ANNUAL REPORT. CO6RA 1 NUREG/CR-3810 V04 REACTOR SAFETY RESEARCH Annunciator System PROGRAMS Ouarterty Report. October-December 1984.

NUREG/CR-3987: COMPUTERIZED ANNUNCIATOR SYSTEMS.

CONTING Antitrust NUREG/CR4071: EXPLORATORY TREND AND PATTERN ANALYSIS NUREG 0970: PROCEDURES FOR MEETING NRC ANTITRUST RE. FOR 1981 LICENSEE EVENT REPORT DATA.

SPONSIBluTIES.

CORRAL l

Aqueous lod 6ne Chemistry NUREG/CR4210: MATADOR A COMPUTER CODE FOR THE ANALY-I NUREG/CR-3514 V02: THE CHEMICAL BEHAVIOR OF IODtNE IN SIS OF RADIONUCLOE BEHAVIOR DURING DEGRADED CORE AC-AQUEOUS SOLUTIONS UP TO 150 C.ll. Radiation-Redox Conditions. CIDENTS IN LIGHT WATER REACTORS.

Atmospheric D6epersion CORRR 2 NUREG/CR4158- A COMPILATON OF INFORMATION ON UNCER. NUREG/CR 4211: MATADOR (VETHOOS FOR THE ANALYSIS OF 4 MA TH 1 81 INEL DISPERSON TRANSPORT AND DEPOSITION OF RADIONUCUDES) CODE DE-

< Nt /

SCRIPTON AND USER'S MANUAL DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

CRAC2 i Austenit6c Steinlose Steel NUREG/CR-36f 3 V02: EVALUAftON OF WELDED AND REPAIR, NUREG/CR-4199 t DEMONSTRATON UNCERTAINTY / SENSITIVITY l ANALYSIS USIN3 THE HEALTH AND ECONOMC CONSEQUENCE WELDED STAINLESS STEEL FOR LWR SERVICE Annual Report fo, MODEL CRAC2.

I 1984.

I CRW

eWie NUREG/CR-4134. REPOSITORY ENVIRONMENTAL PARAMETERS NUREG/CR4070 V03. BfVALVE FOUUNG OF NUCLEAR POWER PLANT SERVICE WATER SYSTEMS Factors That May intensdy The RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL Safety Consequences Of Biotouhng WASTE PACKAGES.

! seempt Cash Flow Anetye6e NUREG/CR-2663 V01: INFORMATION NEEDS FOR CHARACTERl2A. NUREG-1131. FINANCIAL ANALYSIS OF POTENTIAL RETROSPECTIVE lj TION OF HIGH LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG. PREMlUM ASSESSMENTS UNDER THE PRICE ANDERSON

> IC MEDIA Ma n Report. SYSTEM.

NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARACTER 12A-TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG. Cast Sta6ntese Steel l

) IC MEDIA Appendices NUREG/CR 3998 V02_ UGHT WATER REACTOR SAFETY MATERIALS NUREG/CR4114. VALENCE EFFECTS ON THE SORPTION OF NU- ENGINEERING RESEARCH PROGRAMS Ouartetty Progress CLCES ON ROCKS AND MiNERALSIL Report.Apnt June 1984.

___ _ _ - , _ . - _ . - _ - _ _ _ . _ _ . ~ - . _ _ _ _ . . _ . ._. -

I 5

i

' Subject index 55 NUREG/CR4124: NDE OF STAINLESS STEEL AND ON-UNE LEAK Conection Efficiency MONITORING OF LWRS. Annual Report, October 1983 September NUREG/CR-3455: A COMPARISON OF ODINE. KRYPTON ANO XENON

,I 1984. RETENTON EFFICIENCIES FOR VARIOUS SILVER LOADED AD-NUREG/CR4204: LONG-TERM EMORITTLEMENT OF CAST DUPLEX SORPTON MEDtA.

STAINLESS M EELS IN LWR SYSTEMS: Annual Report. October 1983 September 1984. Combustion NUREG/CR4231: EVALUATON OF AVAILABLE DATA FOR PROBA01-Coment LISTIC RISK ARSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR NUREG/CR4181: LEACHABluTY OF RADIONUCUDES FROM POWER PLANTS.

CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT OPERATING NUCLEAR POWER REACTORS. Compact Specimen

NUREG/CR4283

STUDY OF THE EFFECTS OF ELASTO UNLOAD-j Cesium Hydroside INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS.

1 NUREG/CR-319' V01: REACTON BETWEEN SOME CESIUM-LODINE COMPOUNDf AND THE REACTOR MATERIALS 304 STAINLESS Compilence j

STEELINCOf EL 600 & SILVER Volume I.Cesorn Hydroxide Reac. NUREG-0970: PROCEDURES FOR MEETING NRC ANTITRUST RE.

tions.

SPONSIBluTIES.

NURE3/CR-4076: DETERMINATION OF COMPUANCE WITH CRITEmA Charpy FOR FINAL TA1UNGS OtSPOSAL SITE RECtAMATION.

r NUREG/CR-4092: ORNL CHARACTER 12ATON OF HEAVY-SECTION C STEEL TECHNOLOGY PROGRAM PLATES 01,02.AN0 03.

R 51: SAFETY IMPUCATONS ASSOCIATED WITH IN-Chemical Cleoning PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS-NUREG/CR-4276: V16 RATION AND WEAR IN STEAM GENERATOR TEMS IN NUCLEAR POWER PLANTS.

j BS FOLLOWING CHEMICAL CLEANING - SEMIANNUAL C6 M i

NUREG/CR-3889. THE MODELING OF BWR CORE MELTDOWN ACCl.

I Chlortdo DENTS FOR APPUCATION IN THE MELRPIMOD2 COMPUTER NUREG/CR 237 MOBluTY OF RADIONUCUDES IN HIGH CHLORIDE NU EG CR-3913: HECTR VERSON 10 USER *$ MANUAL 1

NUREG/CR4044: TRAC-PF t LOCA CALCULATONS USING FINE-Circudet#ng New Water NODE AND COARSE NODE INPUT MODELS.

I NUREG/CR4109: TRAC PF1 ANALYSES OF POTENTIAL PRES $UR-1 NUREG/CR-4070 V03: BlVALVE FOUUNG OF NUCLEAR POWER IZED THERMAL SHOCK TRANSIENTS AT CALVERT CUFFS / UNIT PLANT SERVICE WATER SYSTEMS Factors That May intensify The Sa% Conwes O Biopg NURE 1 R C PF1/ MOD 1 (NDEPENDENT Cledding ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORATED-PLATE

] CCFL TESTS.

< NUREG/CR4015 EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR4168. GT2F.A COMPUTER CODE FOR ESilMATING I CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL LIGHT WATER REACTOR FUEL ROO FAILURES PLATES IN BENDING SERIES 1- NUREG/CR-4192: THE ANALYSIS OF DRAINAGE AND CONSOLIDA-j Cleeding Deformation TlON AT TYPICAL URANIUM MILL TAIUNGS SITES NUREG/CR 4196 OVERVIEW OF TRAC 00t (VERSION 12) ASSESS-NUREG/CR4218. LOCA SIMULATON IN THE NATONAL RESEARCH UNIWRSAL RFJCTOR PROGRAM Postirfadiation Esamination Re- NU G R f suits For The Third Matenals Test (VT 3)- Second Campaign A DEMONSTRATON UNCERTAINTY / SENSITIVITY 4 ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE

• • • "' U '

NU / 4 MATADOR A COMPUTER CODE FOR THE ANALY-NUREG/CR4861 ASSESSMENT OF CLASS tE PRESSURE TRANS- SIS OF RADIONUCLIDE BEHAVOR DURING DEGRADED CORE AC.

M ER R SE WHEN SUBJECTED TO HARSH ENVIRONMENT CIDENTS IN UGHT WATER REACTORS.

j Computer Model O'"'"I NUREG/CR.3904 A COMPARISON OF UNCERTAINTY AND SENSITIV.

j NUREG/CR4004: CLOSEOUT OF IE BULLETIN 79-25. FAILURES OF ITY ANALYSIS TECHNOVES FOR COMPUTER MODELS.

WESTINGHOUSE OFD RELAYS IN SAFETY RELATED SYSTEMS.

J Computertred System REG /CR 2331 V04 N4: SAFETY RESEARCH PROGRAMS SPON-

} SORED BY OFFICE OF NUCLEAR REGULATORY Concrete Containment

RESEARCH.Quartetty Progress Report, October i . December 31 NUREG/CR4149. ULTIMATE PRESSURE CAPACITY OF REINFORCED 1984 AND PRESTRESSED CONCRETE CONTAINMENT, l' NUREG/CR 3208. TRAC.PD2 DEVELOPMENTAL ASSESSMENT.

Containment

~

NUREG/CR 3703 ASSESSMENT OF SELECTED TRAC AND RELAPS CALCULATONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK NUREG.1116. A REVIEW OF THE CURRENT UNDERSTANDING OF STUDY. THE POTENTIAL FOR CONTA!NMENT FAILURE FROM IN VESSEL t NUREG/CR4169 AN APPRO/CH TO TREATING RADONUCUDE STFAM EXPLOSIONS.

i DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM NUREG/CR-364h DESIGN AND FADRICATON OF A 1/8-SCALE NUREG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF STEEL CONTAINMENT MODEL TRANSPORT AND DEPOSITION OF RADONUCUDES) CODE DE- NUREG/CR-3803 THE EFFECTS OF POST LOCA CONDITONS ON A SCRIPTON AND USER'S MANUAL PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-DUSTRY.

Code ueting NUREG/CR 3816 V02. REACTOR SAFETY RESEARCH Ouarterly NUREG/CH 3905 V02' SEQUENCE CODING AND SEARCH SYSTEM ReportAprlidune 1964.

FOR UCENSEE EVENT REPORTS Code Ustings. NUREG/GR-3855. CHARACTER 12ATON OF NUCLEAR REACTOR CONTAINMENT PENETRATION . FINAL REPORT.

Ceder's Manuel NUREG/CR4064' STRUCTURAL RESPONSE OF LARGE PENETRA-NUREG/CH-3905 V03. SEQUENCE CODING AND $EARCH SYSTEM TONS AND CLOSURES FOR CONTAINVENT VESSELS SUBJECTED FOR UCENSEE EVENT REPCRTS Coder's Manual, TO LOADINGS BEYOND DESIGN BASIS NUREG/CR 3905 V04- SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR4210 MATADOR A COMPUTER CODE FOR THE ANALY.

FOR UCENSEE EVENT REPORTS. Cod #s Manuel SIS OF RADIONUCLIDE BEHAVIOR DURING DEGRADED CORE AC.

CIDFNTS IN UGHT WATIR REACTOHS Coed-Leg ereek NURLG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF NUREG/CR4044, TRAC.PF1 LOCA CALCULATONS USING FINE- TRANSPGAT AND DEPOSITION OF RADIONUCLIDES) CODE DE-NODE AND COARSE NODE INPUT MODELS. SCRIPTON AND USER'S MANUAL

/

l l

56 Subject Index t

)i Containment teotet6en Crack Arroet j

NUREG/CR-4220: REUABiUTY ANALYSIS OF CONTAINMENT ISOLA- NUREG/CR-4015: EFFECT OF STAINLESS STEEL WELD OVERLAY

TION SYSTEMS.

4 CLADOING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL J Containment Perknance PLATES IN BENDING SERIES 1.

1 NUREG-1037 DAFT FC- CONTAINMENT PERFORMANCE WORKING NUREG/CR-4106. PRESSURIZED THERMAL SHOCK TEST OF 6 IN -

GROUP REPORT Draft Report For Comr%nt THICK PRESSURE VESSELS PTSE 11nvestigation Of Warm Prestress.

ing And Upper Shelf Arrest. <

Contemenent Cracking NEU R U2 ON CONTAMINANT COMPLEX TION AND AILI

" 22 " ^ " ^ ^

$SVRE NO r E R For 1984 i Contract NUREG-0978 V03: COMPILATON OF CONTRACT RESEARCH FOR Cmet stehemy THE DATERIALS FNGINEERING BRANCH.DIVISON OF ENGINEER. NUREG/CR-3944. TRAN B-3 EXPERIMENTAL INVESTIGATION OF ING TECONOLOGY. Annual Report For FY 1984 FUEL CRUST STABluTY ON MELitNG SURFACES OF AN ANNU-

{ LAR FLOW CHANNEL (

Control Rod i NUREG-0000 V07 N04. REPORT TO CONGRESS ON ABNORMAL Dete Bank OCCURRENCES October-December 1984. NUREG/CR-2531 R03 INTRODUCTORY USER S MANUAL FOR THE Control Room U.S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE.

SEARCH DATA BANK.

NUREGICR-3967: COMPUTERIZED ANNUNGATOR SYSTEMS- NUREG/CR 4009. HUMAN REUABluTY DATA BANK Evoluaten Re-NUREG/CR 4191: SURVEY OF UCENSEE CONTROL ROOM HABIT. suite.

} ABluTY FRACTICES.

i Control Systen Detehese NUREG/CR-4262 V01; EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR-390S V01 Rt: SEQUENCE COOtNG AND SEARCH i ON TRANSlENTS AND ACCOENTS AT A GENERAL ELECTRIC SVETEM FOR UCENSEE EVENT REPORTS user's Gusde 801UNG WATER REACTOR Main Report NUREG/CR 3905 V02: SEQUENCE CODING AND LEARCH SYSTEM l FOR UCENSEE EVENT REPORTS Code Ustings.

Control System Feelure NUREG/CR-3905 V03. SEQUENCE CODING AND SEARCH SYSTEM I NUREG/CR 4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES FOR UCENSEE EVENT REPORTS Coder's Monual ON TRANS;ENTS AND ACCICENTS AT A GENERAL ELECTRIC NUREG/CR-3905 V04 SEQUENCE CODING AND SEARCH SYSTEM BOfLING WATER REACTOR. Appendices FOR UCENSEE EVENT REPORTS Coder's Manuel.

4 Core Decay Heat

NUREG/CR-3757. TRAN B 2 THE EFFECT OF LOW STEEL CONTENT NUREG/CR-2951

THE 09 EXPERIMENT Heat Removal From Stratified

• ON FUEL PENETRAitON IN A NON-MLLTING CfuNDRICAL FLOW UO2 Debne CHANNEL. NUREG/CR 4169: AN APPROACH TO TREATING HADONUCUDE NUREG/CR 3889 THE MODEUNG OF BWR CORE MELTDOWN ACCI. DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM DENTS + FOR APPUCATION IN THE MELRPIMOO2 COMPUTER CODE. O*88808"'neheng i NUREG/CR-4040 OPERATONAL DECISONMAKING AND ACTON SE-Core Damage LECTON UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER NUREG-1037 DAFT FC. CONTAINMENT PERFORMANCE WORKING PLANTS I GROUP REPORT Draft Report For Comment-

.j Decommiseeening j Core Deeruptive Acchlent l NUREG/CR 1755 ADOO1 TECHNOLOGY. SAFETY AND COSTS OF DE. i

NUREG/CR-3804 V04 PHYSICS OF REACTOR SAFETY Ouarterty COMMISS60NING NUCLEAR REACTORS AT MULTIPLE. REACTOR l J

l NUR C 3 4 R B XPER MENTAL INVESTIGATON OF FUEL CRUST STABlWTY ON MELTING SURFACES OF AN ANNU-w ,, N ""

NUREG/CR-3293 V01: TECHNOLOGY. SAFETY AND COSTS OF DE.

{

j LAR FLOW CHANNEL COMMISSONING REFERENCE NUCLEAR FUEL CYCLE ANO NON- l j Cm MWt FUEL CYCLE FACluTIES FOLLOWING POSTULATED I i NUREG-ttt8 A REVIEW OF THE CURRENT UNDERSTANDING OF THE POTENTIAL FOR CONTAINMENT FAILURE FROM IN-VESSEL NUR 329 V02 CHNOLOGY, SAFE T) AND COSTS OF DE.

{ STEAM EXPLOSONS COMMISSIONING REFERENCE FUEL CYCLE AND NON FUEL j CYCLE FACluTIES FOLLOWING POSTULATED i Core Renoodeng System ACCIDENTS Appendese 2

I NUREG/CR 4277. INVERTED ANNUAL FLOW DPERIMENTAL STUDY. NUREG/CH-4118 MONITORING METHODS FOR DETERMINATON COMPLIANCE WITH DECOMMISSONING CLEANUP CRITERIA AT

! Correeien URANIUM RECOVERY SITES NUREG 1095 EVALUATON OF RESPONSES TO lE BULLETIN 82

' 02 Degradation Of Threaded Fasteners in Reactor Coolant Pressure Deposseen Modoeng BoundaryOf Pressurited Water Reactor Plants NUREG/CR.4158 A COMPILATION OF INFORMATON ON UNCER.

NUREG/CH-4134 REPOSITORY ENVlRONMENTAL PARAMETERS TAINTIES INVOLVED IN DEPOS4 TION MODEUNG.

RELEVANT TO ASSESSING THE PERFORMANCE OF HCH LEVEL 5 WASTE PACKAGES. Dewetering I

NUREG/CR 4192' THE ANALYSIS OF DRAINAGE AND CONSOUDA.

Cost TON AT TYPICAL URANIUM MILL TAlltNGS SITES.

NUREG/CR 3291 V01: TECHNOLOGY. SAFETY AND COSTS OF DE.

i COMMISSONING REFERENCE NUCLEAR FUEL CYCLE ANO NON. Desde MedioHen Detector J FUEL CYCLE FActuTIES FOLLOWING POSTULATED NUREG/CR 413f: INVESilGATON OF ALTERNATIVE MEANS TO AC.

1 ACCIDFNTS Main Report COMPUSH THE GOALS OF DIENNIAL ON CHAMBER CAUDRA.

NUREG/CR-3293 V02 TECHNOLOGY, SAFETY AND COSTS OF DC. TON COMMISSIONING REFERENCE FUEL CYCLE AND NON FUEL CYCLE FACILITIES FOLLOWING POSTULATED Deeperseen Model ACCIDENTS Appendices. NUREQ/CR 41$9 COMPARISON OF THE 1988 INf L DISPERSON

' DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS Coupled Procese Model I

NUREG/CP-0062: PROCEEDINGS OF THE CONFERENCE ON THE AP. DeePeesi j PUCATON OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR NUREG/CR 1765 ADD 01 TECHNOLOGY.SAFETV AND COSTS OF DE.

i I

WASTE REPOSITORY ASSESSMENT. COMMISSONING NUCLEAR REACTORS AT MULTIPLE REACTOR I,

I r

I l

I Subject Irtdex 57 l

1 l

STAflONS Ettects On Decommissionmg Of Intenm instality to Dspose Electren6e undule AWng i Of Wastes Ottsste. NUREG/CR-3863. ASSESSMENT OF CLASS 1E PRESSURE TRANS-i MITT;A RLSPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT I Deepessi SHe SCREENING TESTS.

NUREG/CR4C76: DETERMINATON OF COMPUANCE WITH CRITERIA J FOR FINAL TAluNGS DISPOSAL SITE RECLAMATON Embrtttlement NUREG/CR-3226 V03 STRUCTURAL INTEGRITY OF WATER REAC.

1' Deoooennon Schedule TOR PRESSURE DOUNDARY COMPONFNTS Annual Report For

? REG-0910 R01 S03. NRC COMPREHENSIVE RECORDS DISPOSI- 1984 j RON SCHEDULE- NUREG/CR 3998 V02- UGHT WATER REACTOR SAFETY MATERIALS j ENGINEERING RESEARCH PROGR AMS Ouarterfy Progress g,

hurtEG/CR-3469 V02: OCCUPATIONAL DOSE REDUCTON AT NU. NUa C L G-TERM EMBRtTTLEMENT OF CAST DUPLEX j CLEAR POWER PLANTS. Annotated Bit *ography Of Selected Read- STAINLESS STEELS IN LWR SYSTEMS Annual Report,0ctober 19n3 -

[ mgo in Radiation Protection AndOF ALARA'HE 1961 INEL DISPERSION SepteW 1964 4 NUREG/CR4159 COMPARISON T NUREG/CR-4212.'IN PLACE THERMAL ANNEAuNG OF NUCLEAR RE- '

! DATA WITH RESULTS FROM A NUMBER OF OtFFERENT MODELS ACTOR PRESSURE VESSELS ~

NUREG/CR-4160 HISTORICAL

SUMMARY

OF OCCUPATONAL RADI- l ATON EXPOSURE EXPERIENCE IN U S. COMMERCIAL PeUCLEAR Emergency Core Cee86ng System

POWER PLANTS NUREG/CR-3669 THE MODELING OF BWR CORE MELTDOWN AC41-d NUREGICR 4201 A CALCULATONAL METHOD FOR DETERMINING DENTS . FOR APPUCATON IN THE MELRPt MOD 2 COMPUTER 4 00 LOGICAL DOSE RATES FROM IRRADIATED RESEARCH REAC. CODE.

TOR FUEL Eme, - , ,re ,e,o. noes

,,,,,,, R 40 DRM K A MLATORY AWS4 ON MRGN NUREG/CR-3746 V02 LWR PRESSURE VESSEL SURVEILLANCE 00

! SIMETRY IMPROVEMENT PROGRAM Semeannual Progrese MATE RIAL LICENSEES Draft Report For Comment

! Report Apr41964. Septeerter 1984

{ NUREG/CR-3746 V03 LWR PRESSURE VESSEL SURVEILLANCE 00- g ,,,,,ney n ,,n ,pi nnen ,

i SIMETRY IMPROVEMENT PROGRAM 1964 Annual Report,0ctobe' NUREG/CR-3657, PREUMINARY SCREENING OF FUEL CYCLE AND BY. PRODUCT MATERIAF. LICENSES FOR EMERGENCY PLANNING NU E CR 4 3 NV T TION OF ALTrRNAffvE MEANS TO AC.

COMPUSH THE GOALS OF BfENNIAL ON CHAMBER CAtl0RA- EmlWen Centrol l TON

< NUREG/CR-4176 EMISSION CONTROL TECHNOLOGY AND OVAUTY ASSURANCE NEEDS At URANtUM MILUNG FACfuTIESincludet

, Downcomer Proseure Supporteg Methods For TestmgOperateg,And Mentaerung Aar Pollu-4 NUREG/CR 3703. ASSESSMENT OF SELECTED TRAC AND RELAPS hon ConW Dwices

• CALCULATONS FOR OCONEE.1 PRESSUR12ED THERMAL SHOCK

STUDY- Energy freneport l NUREGICR-4203 A CALCULATONAL METHOO FOR DETERMINING j

{ Drainage BIOLOGICAL DOSE RATES FROM IRRADIATED RESEARCH RE AC.

NOREG/CR 4t92 THE ANALYSIS OF DRAINAGE AND CONSOUDA.

TON AT TYPICAL URANIUM M;LL TAluNGS SITES TOR FUEL.

1 pry pe,,,eeson Enforcement Aetten NUREG/CR 4158 A COMPILATION OF INFORMATION ON UNCEq. NURE00940 V04 Not: ENFORCEMENT ACTIONS SIGN #iCANT AC-l TIONS RESOLVED Ouarteth Progress Report January March 1965  ;

j TAINTIES INVOLVED IN DEPOSITON MODEUNG Dry Spent Fued Sterege Engineered Sofety Feature f NURtG/CR-33 t F: TECHNICAL BASES AND USER'S MANUAL FOR i NUREG/CR4064 DRY SPENT FUEL STORAGE TEST PLAN FOR DE.

! STRUCTIVE FUEL ROD EXAMtNATONS THE PROTOTYPE OF SPARC . A SUPPRESSION POOL AEROSOL i REMOVAL CODE. i i

I ECCS NUREG/CR4277. INVERTED ANNUAL FLOW EXPERIMENTAL STUDY.

Engineered Safety System t

]

NUREGICR 4191: SURVEY OF LICENSEE CONTROL ROOM HA8lf. r j Eartiguahe ABluTV PRACTICES .

I 4 NUREG 1061 V02 REPORT OF THE U S NUCLEAR REGULATORY NUREG/CR 4210 MATADOR A COMNTER CODE FOR THE ANALY.

! COMMISSION PIPING REVIEW COMMif7E E volume 2 Evaluation Of SIS OF RADIONUCLIDE E3EHAVIOR DURING DEGRADED CORE AC. i

! Seistruc Dessne A Rowew Of Se' ems Design Requiremente For Nu. COENTS IN UGHT WATER RE ACTORS l clear Power Plant P NOREG/CR 421 t MATADOR (METHODS FOR THE ANALYSIS OF  :

TRANSPORT AND DEPO $lflON OF RADIONUCUDES) CODE DE. I

! NUREG/CR.3176 St TURAL AND TECTONIC STUDIES IN NEW t YORK STATE Feel Report July 196 t June 1942 SCRIPTION AND USER'S MANUAL l NUREG/CR4226 NEW MADRO SEISMOTECTONIC STUDY Actrwtwo Dunng Fisce' Year 1963 Engineering Curr6sueum i

' NUREG/CR dost ASSESSMtNT OF JOn RELATED EDUCAtlONAL Eddy Current QUAUFICATONS FOR NUCLEAR POWER PLANT OPEnATOns NUREG 0915 V03 COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS ENGINEERING BRANCH.DIVISON OF ENGINEER. EnWonmental Aceseement

,t ING TECHNOLOGY Annusi noport For FY 1964 NUREG-itte ENVIRONMENTAL ASSES 5 MENT FOR RENf WAL OF SPECIAL NUCLEAR MATERIAL UCENSE NO SNM.t f 0F Dochet No .

Educellenal Que4Hicetten 10-t 15 t. (Westeghouse Electnc Corporation) t NUREG/CR-4051 ASSESSMENT OF JOB RELATED EDUCATONAL QUAUFICATONS FOR NUCLEAR POWER PLANT OPERATORS Environmentes Effset ,

NUREG.09FS y03 COMPILATON OF CONTRACT RESE ARCH FOR Effluent THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER.

NUREG-Oct? R01' CALCULATON OF RELE ASES OF RADOACTIVE ING TECHNOLOGY Annusi nopor1 For FY 1964 MATERIALS IN GASEOUS AND LOUlO EFFLUENTS FROM PRES-

, SUHlHD WATER REACTORS (PWR GALE CODE) Eiride EMP 13 Settery Ces '

! NUREG/CR 4245 IN PLANT SOURCE TERM MEASUREMENTS AT NUREG/CR 409f TEST SERIES A SEISMIC FRAGluTY TESTS OF

! BRUNSWICK STEAM ELECimC STATON NAfuRALLY-AGED ExlDE EMP.13 BATTERT CELLS 1

Elaette Uniesens Espeevre -

I NUREG/CR 42s1 STUDY OF THE EFFECTS Or ELASTIC UNLOAD. NUREG t134 RADIATON PROTECTON TRAINING FOR PERSONNEL a

INGS ON THL JI R CURVES FROM COMPACT SPECIMENS E MPLOYED IN MEDICAL F ACILITIES [

1 i

5

)

.~------ - . ~ ~ - - - - - - - ,- -- =.- - - - - - - - - - - - - . -

58 Sut$ect index f NUREG/CR-4031 V03 NEUTRON SPECTRAL CHARACTER 12ATION STEELINCONEL 600 & SILVER Volume ICesium Hydronde Reac.

+

FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST) tR tions. I RADtAflON SERIES ' Neutron Exposure Parameters." NUREG/CR-3980 V03. LIGHT WATERREACTOR SAFETY FUEL SYS-NUREG/CR4147: THE EFFECT OF ENVIRONMENTAL STRESS ON TEMS RESEARCH PROGRAMS. Questerty Progress Report. July-Sep-SYLGARD 70 SfLICONE ELASTOMER tomber 1984 NUREG/CR 4105. AN ASSESSMENT OF THERMAL GRADIENT TUBE b

UREG/CR 3721 at: PRES $URE MEASUREMENTS IN A HYDROGEN COMBUSTION LNVIRONMENT. Hy@ Air Combustion Test Senes sis 1 And 2 tri The FITS Tank, NUREG/CR 4169 AN APPROACH TO TREATING RADIONUCUDE DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM FRAPCON i NUREG/CR4810 V04 REACTOR SAFETY RESEARCH F6ee6on Product Reteeee PROGRAMS Quarterty Report. October December 1984. NUREG/CR 3895 V03 HIGH-TEMPERATURE GAS-COOLED REACTOR SAFETY STUDIES FOR THE DIVISON CF ACCIDENT Failure EVALUATON Ouarter9 Progrese Report, July I . September 30.1984 NUREG-1118 A REVIEW OF THE CURRENT UNDERSTANDING OF j TLE POTENTIAL FOR CONTAINMENT FAILURE FROM IN VESSEL F6ee6on Product Treneport ,

STEAM EXPLOSONS NUREG/C44205. TRAP MELT 2 USER 1 MANUAL NUREG/C43647: DESIGN ANO F48RICATON OF A t /8-SCALE STEEL CONTAlNMENT MOQEL Fleshing NUREG/CR 4004. CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF NUREG/CR4079 ANALYTIC STUDIES PERTAINING TO STEAM GEN-WESTINGHOUSE BFD RELATS IN SAFETY RELATED SYSTEMS ERATOR TUBE HUPTURE ACCIDENTS NUREG/C441eo STATE OF THE ART OF SOUD' STATE MOTOR CONTROLLERS pie g g e.g pi,,,

pegu,, anegy,,, NUREG/CR 4015. EFFECT OF STAINLESS STEEL WELD OVERLAY NUREG/CR 4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED CLADDING ON THE STRUCTURAL 6NTEGRITY OF FLAWED STEEL AND PRESTRESSED CONCRETE CONTAINMENT, PLATES IN BENDING SER!ES 1.

Fetegue Floobie PtPeng Design NUREG/CR3228 V03 STRUCTURAL INTEGRITY OF WATER REAC. NUAEG/CR4263 RELIABILITY ANALYSIS OF SilFF VER$US FLEXI-TOR PRESSURE BOUNDARY COMPONENTS Annual Report For BW PIPING FINAL PROJECT REPORT.

1984.

Forced Convective FouM NUREG/CR3193 FORCED CONVECTIVE.NONEOuluBRIUM. POST.

NUREG/CR 3178 STRUCTURAL AND TECTONC STUDIES IN NEW CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATON YORK STATE Feel Report. July 1981 June 1982.

COMPARISON REPORT.

URE CR3174 V02 GEOPHYSICAL GEOLOGICAL STUDIES OF '" I***"**

POSSIBLE EXTENSONS OF THE NEW MADRID FAULT NUREG/CR-3881 ANALYSIS OF JAPANESE-U S NUCLEAR POWER ZONE Annual Report For 1983 PLANT MAINTENANCE.

Piem Flow Frecture NUREG/C43651. ASSESSMENT OF THE ADEQUACY OF ORNL IN. NUREG/CR 3178 STRUCTUnAL AND TECTONC STUDiE3 IN NEW STRUMENTAflON IN REFLOOD TEST FACluTIES YORK STATE Fmai Report. July 1981. June 1962.

NUREG/CR 3228 V03 STRUCTURAL INTEGRITY OF WATER REAC-Fmer TCH PRESSURE BOUNDARY COMPONENT $ Annual Report For NUREGICH ditt: A COMPARATIVE STUDY OF HEPA FILTER EFFI, 1984 CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR JET- NUREG/CR 4198 FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS.

GENERATED De 2 ETHYLHEXYL SEBECATE,De 2 ETHYLHEXYL TERS PHTHALATE.AND SODIUM CHLORCC.

Fracture Mechenice Penel Enverenmentet Statement NUREG 0975 V03 COMPILATON OF CONTRACT RESE ARCH FOR NUREG 1033 FINAL ENVIRONMENTAL STATEMENT RELATED TO THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEE4 THE OPERATON OF WPPSS NUCLEAR PROJECT NO 3 Docket No ING TECHNOLOGY Annual Report For FY 1964 NU 1 A$ RON E BENT RELATED TO NUREG/CR 4106 PRESSURilED THERMAL SHOCK TEST OF 6IN.

THE OPERATON OF NINE MILE POINT NUCLEAR STATON.UNif THCK PRESSURE VESSELS PTSE 1Investigabon Of Warm Prestress-NO. 2 Dooet No 60-410 (Niagere Mortewk Power Corporation.et el) ing And UpperW Anst Finenetel Anesyees Fregieny N') REG 1131: FINANCIAL ANALYSIS OF POTENilAL RETROSPECTIVE NUREG/CH 3558 HANDOOOK OF NUCLE AR POWER PLANT SEISMC PREMtUM ASSESSME NTS UNDER THE PRICE ANDERSON FRAGIUTIES Se+smic Setety Wergine Research Program SYSTFM Fuoi Ptre Proteeteen System NUREG t127. RADIATON PROF [CTION TRAINING At URANIUM HEX-NUREG/CR 4230 PROGABIUTY BASED EVALUATION OF SELECTED AFLUOnlDE AND FUEL FABRICATON PLANTS FIRE PROTECTON FE ATURES IN NUCLEAR POWER PLANTS NUREG/CR 3944 TRAN 0 3 EXPERIMENTAL INVESTIGATION OF FUFL CRUST STADlufY ON MELilNG SUHFACES OF AN ANNU-U G/CR 4229 EVALUATON OF CURRENT METHODOLOGY EM-

" "^N PLOYED IN PROBABlUSTIC RISM ASSESSMENT (PRA) OF FIRE Fuel Cycle EVENTS AT NUCLEAR POWER PLANTS NUREGICR 3293 V01- TECHNOLOGY. SAFETY AND COSTS OF DE.

NUMEG/CR 4211. EVALUAf 0N OF AVAILABLE DATA FOR PROBABI.

COMMISSONING REFEnrNCE NUCLEAR rutt CYCLE AND NOk USTC RISM ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR FUEL CYCLE FACIUilES FOLLOWING POSTULATED POWER PLANTS ACCIDENTS Me.n Report p,,g , NUREG/CR3291 V02. TECHNOLOGY. SAFETY AND COSTS OF DE.

NUREG/CR 3930 OBSERVE 0 DEHAVIOR OF CESIUM.ODINEAND COMMIS$lONING REFE RENCE FUEL CYCLE AND NON FUEL TELLuntVM IN THE ORNL FISSION PROOUCT oELEASE PRO. CYCL E FACluTIE S FOLLOWING POSTULATED GRAM ACCIDENTS Apperdcas Premien Product Fuoi Cycle Feesty NUREG/CR 3197 V01 REACTON DETWEEN SOME CESIUM IODINE NUREG ot25 RIO SAF EGUARDS

SUMMARY

EVENT UST COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS (SSEL) REVISION 10

l Subject index 59 l

l FuelDamage Groundwater NUREG/C43810 V04. REACTOR SAFETY RESEARCH NUREG/CR4134. REPOSITORY ENV6AONMENTAL PARAMETERS

PROGRAMS ouartery Report,0ctober-Decemtst 1984 RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL NUREG/CR-3816 VC2

REACTOR SAFETY RESEARCH Ouarterty WASTE PACKAGES j ReportApr44une 1964. NUREG/CR4198 FRACTURE IN GLASS'HIGH LEVEL WASTE CANIS-1 TERS l Fuel Failure i

NUREG/CR4168- GT2F A COMPUTER CODE FOR ESTIMATING HECTR

! LIGHT WATER REACTOR FUEL ROD r ALLURES. NUREG/C43913 HECTR VERSION 10 USER-S MANUAL J

d Fuel Memoval HfpA NUREG/CR3757. TRAN 0-2.THE EFFECT OF LOW STEEL CONTENT NUREG/C44tti. A COMPARATIVE STUDY OF HEPA FILTER EFF1-

! ON FUEL PENETRATION IN A NON-MELTING CYLINORICAL FLOW CIENCIES WHEN CHALLENGED WITH THERMAL AND Al4 JET.

} CHANNEL GENERATED De-2 ETHYLHEXYL SEBECATE.Di'2 ETHYLHEXYL PHTHALATE,AND SOOlUM CHLORICE Fulh instrumented feet Sete NUREG/CR4225.

SUMMARY

OF EFFICIENCY TESTING OF STAND-NUREG/CR-3721 V01: PRESSURE MEASUREMENTS IN A HYDROGEN ARD AND HIGHCAPACITY HtGH-EFFICIENCY PARTICULATE AIR I COM0USTION ENVIRONMENT. Hydrogen-Ar Combustion Test Senes FILTERS SUBJECTED TO SJMULATED TORNADO DEPHESSURIZA-I 1 And 2 in The FITS Tanit TION AND EXPLOStVE SHOCK WAVES.

NUREG/CH4264. INVESilGATON ON HIGH EFFICIENCY PARilCU-i CT2F LATE AIR FILTER PLUGGING BY COMBUSTION AEROGOLS NUREG/CR4t68 GT2F A COMPUTER CODE FOR ESilMAtlNG

] LIGHT WATER REACTOR FUEL ROD FAILURES UAEG/CR-3703. ASSESSMENT OF SELECTED TRAC AND RELAPS i Gemme Doee Rete CALCULATIONS FOR OCONEE t PRESSURIZED THERMA. SHOCK NUREG/C44203 A CALCULAflONAL METHOD FOR DETERMINING STUDY.

BIOLOGICAL DOSE RATES FROM IRRADIATED RESEARCH REAC.

I NUREG/CR4010 SPECIFICATION OF A HUMAN REllABILITY DATA

• Gee Buttle BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE-NUREQ/CR 27t8. STEAM EXPLOSON EXPERIMENTS WITH SINGLE AR POWER PLANTS i

DROPS OF IRON OXIDE MELTED WITH A CO2 LASER Part N it Parametnc Studies.

UAEG/CR4015 EFFECT OF STAINLESG STEEL WELD OVERLAY Gee Dietributton System CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL NUREG/CR3551 SAFETY IMPLICATONS ASSOCIATED WITH IN. PLATES IN BENDING SERIES 1 j PLANT PRESSURIZED GAS STORAGE AND DISTRfDUTION SYS- NUREG/CR4088: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL WELDS FOR THE THIRD HS$T 1RRA-TEMS IN NUCLEAR POWER PLANTS

< OlAflON SERIES

! Geotseinteetenel Ateerption NUREG/CR 4208 GASTROINTESTINAL A0 SORPTION OF PLUTONIUM IN MICE, RATS. AND DOGS Apper.abon To EstabhsNng values Of f t NUREG 1125 V05 A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUAROS,1957-1964 Volume 5,Part pg, g g %

j 2 ACRS Hoports On Genanc Subsects (HTGR Regulatory Guntes) j Geuseeen Plume Modet j NUREG/CRatS9 COMPARISON OF THE 1961 INEL DISPERSION /CR 3558 HANDOOOK OF NUCLEAR POWER PLANT SLISMIC DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS. FRAGillflES Seistruc Safety Margns Research Program J Ooeger Mue86er j NUREG/CR4 t t 0 MONif0 RING METHODS FOR DETERMINATION NfAG/C 63 ASSESSME.NT OF CLASS tE PRESSURE TRANS.

COMPLIANCE WITH DECOMMIS$10NING CLEANUP CRITERIA AT MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIROhMENT l URANIUM RECOVERY SifES SCHEENING TESTS i

l Goochemical Medotng Heet Affected Zone l NUREG/CP-00e2 PROCEEDINGS OF THE CONrERENCE ON THE AP- NUREG/CR3613 V02: EVALUATION OF WELDED AND REPAIR 4

PLICATION OF GEOCHEMICAL MODELS TO HIGH LEVEL NUCLEAR WELDED STAINtESS STEEL FOR LWR SERVICE Annual Report for j WASTE REPOSITORY ASSESSMENT. g gg4.

Geochem6etrY Heat pomoval

NUREG/C44134 REPOSif0RY ENVIRONMENTAL PARAMETERS NUREG/CR 2951

THE 09 EXPERIMENT Heat Removat From Stratdied j RELEVANT TO ASSES $1NG THE PERFORMANCE OF HIG4 LEVEL 002 Debne WASTE PACKAGES NUREG/CH 3816 V02 REACTOR SAFETY RE SE ARCH Ouarterty Games hte Form

• P 't #"84""' *4 i NUREG/CR4t98 FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS- Heat Transfer l TERS. NUREG/CR3193- FORCED CONVECTIVE.NONEOUILIRRIUM. POST-CHF HEAT TRANSFE4 EXPERfME NT DATA AND CORREL AflON {

j GrenHe COMPARISON HEPORT.

NUREG/CR 2663 V01 INFORMATON NEEDS FOR CHARACTERIZA- NUREG/CH 3208 TRAG.PO2 DEVELOPMENTAL ASSESSMENT.

TION OF HIGH LEVEL Wi STE REPOSITORY SITES IN SIX GEOLOG-i IC MEDIA Mac Report Hoevy Section Steet Technetegy j NUREG/C42663 V02 INFORMAflON NEEDS FOR CHARACTEnl2A- NUREG/CR 4015 EFFECT Or STAINtESS STEEL WELD OVf RLAY flON OF HIGH LEVEL WASTE REPOSITORY SITES lN SIX GEOLOG- CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STLEL i

IC MEDIA Appedices. PLATES IN DENDING SERIES 1.

NUREG/CR 4066- TENSILE PROPERilES OF IRRADIATED NUCLEAR l GrevHetional Sedimentation GRADE PRESSURE VESSEL WELDS FOR THE THIRD HSST 1RRA.

NUREG/CR4tS8 A COMPILAflON CF INFORMAflON ON UNCE4 OtAtlON SERIES TAINTIES INVOLVED IN DEPOSITION MODELING Hegh Efficiency Porticulate Air Filter j

Grevity NUREG/CR4225

SUMMARY

OF EFFICIENCY TESilNG OF STAND-NURE G/CR-3 t f 4 V02 GEOPHYSICAL 4EOLOGICAL STUDIES OF ARD AND HIGHCAPACITY HIGH f FFICIENCY PARflCULATE AIR I POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT FILTERS SUBJECTED TO SIMULATED IORNADO DEPHE SSunilA-ZONE Annual Report For 1963 TION AND EXPLOStVE SHOCK WAVES s

-. - - - _ - - - - - - - - - - - - - - - --,--e- - - - - - ----- -.-+-- - - - - - - , - - -

60 Subject index H6gh integrity Container Hydrogen Generation NUREGiCR-4215. TECHNICAL FACTORS AFFECTING LOW-LEVEL NUREG/CR-3803. THE EFFECTS OF POST LOCA CONDITONS ON A WASTE FORM ACCEPTANCE CRITERIA. PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-

, DUSTRY.

NUREG/CR-3703. ASSESSMENT OF SELECTED TRAC AND RELAP5 IE Bdenn 7W CALCULATIONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK NUREG/CR4003: CLOSEOOT OF IE BULLETIN 7944 INCORRECT STUDY.

WElGHTS FOR SWING CHECK VALVES MANUFACTURED BY H6gh Temperature VELAN ENGINEERING CORPORATION NUREG/CR-2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR REGULATORY IE B d een 4 25 RESEARCH.Ouarterty Progress Report October 1 December 31, NUREG/CR4004. CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF 1084 WESTINGHOUSE BFD RELAYS IN SAFETY-RELATED SYSTEMS.

NUREG/CR 3197 V01: REACTON BETWEEN SOME CESIUM-CDINE COMPOUNDS AND ThE REACTOR MATERIALS 304 STAINLESS IE Bdenn W2 -

STEELINCONEL 600 & SILVER. Volume ICessum Hydomde Reac. NUREG/CR-4005 CLOSEOUT OF IE BULLETIN 80-12 DECAY HEAT tent REMOVAL SYSTEM OPERABILITY.

NUREG/CR 3885 V03: HIGH TEMPERATUr4E GASCOOLED REACTOR SAFETY STUDIES FOR THE DIVtSION OF ACCIDENT IE BuHenn $242 EVALUATON Ouarterty Progress Report, July 1 - September 30,1964. NUREG 1095: EVALUATON OF RESPONSES TO IE BULLETIN 82-02 Degradaten Of Threaded Festeners in Reactor Coolant Pressure High-Efficiency Particulate Boundary Of Pressunted Water-Reactor Plants.

NUREG/CR-4111. A COMPARATIVE STUDY OF HEPA FILTER EFFI-CIENCtES WHEN CHALLENGED WITH THERMAL. AND AIR-JET. IOSCC GENERATED De-2-ETHYLHEXYL SEBECATE,Di 2 ETHYLHEXYL NUREG-1061 V05: REPORT OF THE U S. NUCLEAR REGULATORY PHTHALATE.AND SODIUM CHLORIDE. COMMISSION PIPING REVIEW COMMITTEE.Volurre 5 Summary -

High-Efficiency Particulate Air Filter NUREG/CR4264. INVESilGATION ON HIGH-EFFICIENCY PARTICU- ILRT LATE AIR FILTER PLUGGING BY COMBUSTON AEROSOLS- NUREG/CR4220 REUABluTY ANALYSIS OF CONTAINMENT ISOLA-H6gh-uvW WeWe TION SYSTEMS.

NUREG/CR 3900 V03: LONG TERM PERFORMANCE OF MATERIALS I '

USED FOR H EVEL WASTE PACKAGING.Quarterty

/CR 4 IMPORTANCE RANKING BASED ON AGING CON-NUR CR4134 REPOslTORY ENVIRONMENTAL PARAMETERS SIDERATONS OF COMPONENTS INCLUDED IN PROBABluSTIC RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH LEVEL RISK ASSESSMENTS.

WASTE PACKAGES. . .

NUREG/CR4tet V01. CRITICAL PARAMETERS FOR A HiGH-LEVEL .- b I WASTE REPOSITORY Volume 1 Basart. NUREG/CR4075. DESIGNING PROTECTIVE COVERS FOR URANIUM NUREG/CR-4198 FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS. MILL TAILINGS PILES. A Review TERS.

NUREG/CR-4237. MOBILITY OF RADIONUCUDES IN HIGH CHLORIDE incident ENVIRONMENTS. NUREG 0090 V07 N04. REPORT TO CONGRESS ON ABNORMAL OCCURRENCES October-December 1984.

High-Level Weste Repository NUREG/CP 0062. PROCEEDINGS OF THE CONFERENCE ON THE AP- Incident Dete Ane#yees PUCATION OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR WASTE REPOSITORY ASSESSMENT.

NUREG/CR-3611. RADCACTIVE MATERIAL (RAM) ACCIDENT / INCL-DENT DATA ANALYSIS PROGRAM Human Factors NUREG/CR-2331 V04 N3 SAFETY RESEAR PROGRAMS S N EG CR3197 V01: REACTON BETWEEN SOME CESIUM-CDINE RESEARCH Ouarterty Progress Report July 1. September 30.1964 NUREG/CR2331 V04 N4. SAFETY RESEARCH PROGRAMS SPON. STEELINCONEL 600 & SILVER Volume I.Cessum Hydronde Reac-SORED BY OFFICE OF NUCLEAR REGULATORY tions.

RESEARCH Ouarterty Progresa Report. October 1 December 31, NUREG/CR 3987. COMPUTERIZED ANNUNCIATOR SYSTEMS. NUREG 0750 V28101: INDEXES TO NUCLEAR REGULATORY COM-NUREG/CR-4093 SAFETY / SAFEGUARDS INTERACTONS DURING MISSION ISSUANCES. January-March 1985.

SAFETY RELATED EMERGENCIES AT NUCLEAR POWER REACTOR FACIUTIES InspecWon NUREG/CR 4206. A SELECT REVIEW OF THE RECENT (19791963) NUREG-0040 V09 Not: UCENSEE CONTRACTOR AND VENDOR BEHAVORAL RESEARCH LITERATURE ON TRAINING SIMULA. STATUS REPORT.Quarterty ReportJanuary March 1985 (White Book)

TORS Instrumentation Circuit Homen Performance Date NUREG/CR-3863 ASSESSMENT OF CLASS 1E PRESSURE TRANS.

NUREG/CR.4010 SPECIFICATON OF A HUMAN REUABluTY DATA MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE- SCREENtNG TESTS.

AA POWER PLANTS Human Rel6et litty inter Leek Rate Ted NUREG/CR 3626 V02: MAINTENANCE PERSONNEL PERFORMANCE NUREG/CR4220 REUABluTY ANALYSl3 OF CONTAINMENT ISOLA-TION SYSTEMS.

SIMULATION (MAPPS) MODEL- DESCRIPTON OF MODEL CONTENT. STRUCTURE,AND SENSITIVITY TESTING NUREG/CR 4009 HUMAN PEUABluTY DATA BANK Evaluation Re- 8" F d 8*fA""'**"8 suits NUREG4829 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT NUREG/CR4010. SPECIFICATION OF A HUMAN REUABluTY DATA REPORT. SYSTEMATIC EVALUATION PROGRAM SAN ONOFRE BANK FOR CONDUCTING HAA SEGMENTS OF PRAS FOR NUCLE. NUCLEAR GENERATING STATION UNIT 1. Docket No 50-206(South-AR POWER PLANTS orn Cahforrue Edison Company)

Hydrogen Combuetion interpenvier Stroes Corrocoon Cracking NUREG/CR3721 V01, PRES $URE MEASUREMENTS IN A HYDROGEN NURFG/C43613 V02: EVALUATION OF WELDED AND REPAla-COMBUSTON ENylRONMENT Hydrogen-Air Comtution Test Senes WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Report for 1 And 2 in The FITS Tank. 1964 1

1 Subject index 61 intestinel Absorpeon LM2 r,

! NUREG/CR-4208. GASTROINTESTINAL ABSORPTION OF PLUTONIUM NUREG/CR-4181: LEACHABluTY OF RADONUCUDES FROM IN MICE. RATS, AND DOGS Apphcaton To EstabhsNng Values Of f1 CEMENT SOUDiFIED WASTE FORMS PRODUCEO AT OPERATING For Soluble Plutonsurn. NUCLEAR POWER REACTORS.

Inventory Defference Data Leeching NUREG-0430 VOS N01. LICENSED FUEL FAQUTY STATUS NUREG/CR-4198: FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS.

REPORT.irwentory Dfference Data. January 1984 - June 1984 (Gray TERS.

Book 11)

Leak Inverted Annuier Flow NUREG/CR-3855: CHARACTERIZATON OF NUCLEAR REACTOR NUREG/CR-4277: INVERTED ANNUAL FLOW EXPERIMENTAL STUDY. CONTAINMENT PENETRATON - FINAL REPORT.

lodine Leak Rete NUREG/CR.3455. A COMPARISON OF CDtNE, KRYPTON,AND XENON NUREG-1037 DRFT FC. CONTAINMENT PERFORMANCE WORKING RETENTON EFFICENCIES FOR VAR!OUS SILVER LOADED AD. GROUP REPORT. Draft Report For Comment.

i SORPTON MEDA.

NUREG/CR-3514 V02- THE CHEMICAL BEHAvlOR OF ODNE IN Leekage J, AQUEOUS SOLUTIONS UP TO 150 C.ll.Radation.Redon Conditions. NUREG 1095. EVALUATON OF RESPONSES TO IE BULLETIN 82-02.Degradaten Of Threaded Fasteners in Reactor Coolant Pressure U /R ry Of Prnaunzed Water Reacta Nnts.

INVESTIGATON OF ALTERNATIVE MEANS TO AC.

l COMPUSH THE GOALS OF BIENNIAL ON CHAMBER CAUBRA- Legal M - i-5 TON. NUREG4750 V21 NO2. NUCLEAR F.EGM.ATORY COMMISSION IS-SUANCES FOR FEBRUARY 1e85. Ps s 275-469.

UE CR-2718: STEAM EyPLOSON EXPER:MENTS WITH SINGLE SUANC F M CH 1 5 4 9

, DROPS OF IRON OxtDE MELTED WITH A CO2 LASER Part NUREG-0750 V21 N04. NUCLEAR REGULATORY COMMISSON IS.

t il Parametnc Stut>es-SL'ANCES FOR APRIL 1985. Pages 581 1,041.

S FROM R A C S EAC.

2 E%ED WERADNG MAGORS SMS

SUMMARY

REPORT Data As Of Febuary 28.1985 (Gray Book t).

TOR FUEL NUREG-0020 VO9 N04: UCENSED OPERATING REACTORS STATUS bolsted Reactor Fuel

SUMMARY

REPORT.Deta As Of March 31,1985 (Gray Book I)

NUREG-0725 ROS: PUBLIC INFORMATON CIRCULAR FOR SHIP. NUREG-0020 V09 N05. UCENSING OPERATING REACTORS STATUS

SUMMARY

REPORT Data As Of April 30,1985.(Gray Book 1)

MENTS OF IRRADATED REACTOR FUEL

, irredleted Zircatoy Clodding Licensee Contractor And Vendor inepection NUREG/CR-3980 V03: UGHT WATER-REACTOR SAFETY FUEL SYS- NUREG-0040 V09 N01: UCENSEE CONTRACTOR AND VENDOR TEMS RESEARCH PROGRAMS. Quarterty Progress Report. Jury-Sep- STATUS REPORT.Quarterty Report, January-March 1985 (WNte Book) ternber 1984.

Ucenese Event Report irredletion NUREG/CR-2000 V04 N3. UCENSEE ENENT REPORT (LER) i NUREG/CR 3872- DATA ACQUlstTION AND CONTROL OF THE HSST COMPtLATON Fa Month Of March 1985.

1 SERIES V IRRADATION EXPERIMENT AT THE ORR. NUREG/CR-2000 V04 N4. UCENSEE EVENT REPORT (LER) l COMPILATON For Month Of Apnl 1985.

J-integral Ree6etance Curve NUREG/CR-2000 V04 N5. UCENSEE EVENT REPORT (LER)

NUREG/CR-4283: STUDY OF THE EFFECTS OF ELASTIC UNLOAD _ COMPILATON For Month of May 1985 INGS ON THE JI-R CURVES FROM COMPACT SPECIMENS. NUREG/CR 3905 V01 R1: SEQUENCE CODING AND SEARCH SYSTEM FOR UCENSEE EVENT REPORTS User's Guide Japan NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM 4 NUREG/CP-0059 Vot: PROCEEDINGS OF THE MITLNRC SEISMC IN. FOR UCENSEE EVENT REPORTS. Code betings FORMATION EXCHANGE MEETING. VOLUME 3. NUREG/CR-3905 V03. SEQUENCE CODING AND SEARCH SYSTEM I NUREG/CR 3883 ANALYSIS OF JAPANESE-U S. NUCLEAR POWER FOR LICENSEE EVENT REPORTS Coder's Manual.

1 PLANT MAINTENANCE. NUREG/CR-3905 V04. SEQUENCE CODING AND SEARCH SYSTEM

, FOR UCENSEE EVENT REPORTS Coder's Manual.

j Krypton NUREG/CR-4071: EXPLORATORY TREND AND PATTERN ANALYSIS i

MUREG/CR 3455: A COMPARISON OF OOINE. KRYPTON,AND XENON FOR 1981 UCENSEE EVENT REPORT DATA.

I RETENTON EFFICIENCIES FOR VARIOUS SILVER LOADED AD. NUREG/CR-4220: REUABluTY ANALYSIS OF CONTAINMENT ISOLA-SORPTON MEDIA. TION SYSTEMS.

LER Looe4Cootent Acc6 dent

(

,' NUREG/CR 2000 V04 N3. UCENSEE EVENT REPORT (LER) NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN-COMPILATION For Month Of March 19% STRUMENTATlON IN RFFLOOD TEST FACluTIES.

NUREG/CR-2000 V04 N4. UCENSEE EVENT REPORT (LER) NUREG/CR-4196: OVERVIEW OF TRAC-BD1 (VERSON 12) ASSESS-l COMPILATION For Month Of Aptd 1985 MENT STUDIES.

NUREG/CR-2000 V04 N5: UCENSEE EVENT REPORT (LER) NUREG/CR-4218 LOCA SIMULATON IN THE NATIONAL RESEARCH COMPILATION For Month Of May 1985 UNIVERSAL REACTOR PROGRAM Poststradiaten Esarnmaton Re-NUREG/CR 4220: REUABluTY ANALYSIS OF CONTAINMENT ISOLA- suits For The TNrd Materials Test (MT 3) Second Campaegn.

TlON SYSTEMS.

Looe-Of-Flu 6d Teet LOCA NUREG/CR-3005

SUMMARY

OF THE NUCLEAR REGULATORY COM-

{

4 NUREG/CR-4044 TRAC-PF t LOCA CALCUJTONS USING FINE- MISSON'S LOFT PROGRAM RESEARCH FINDINGS NODE AND COARSE NODE INPUT MODELS.

NUREG/CR-4196. OVERVIEW OF TRAC BD1 (VERSION 12) ASSESS. Low Enriched Uranium MENT STUDIES. NUREG-1065 Rol: ACCEPTANCE CRITERIA FOR THE LOW EN-NUREG/CR-4218 LOCA SIMULATON IN THE NATIONAL RESEARCH RICHED URANIUM REFORM AMENDMENTS

! UNIVERSAL REACTOR PROGRAM Postirradiation Examinetson Re. NUREG 1065 R01: ACCEPTANCE CRITERIA FOR THE LOW EN-i suits For the TNrd Matenals Test (MT 3) . Second Campaign. RICHED URANIUM REFORM AMENDMENTS.

1 j LOFT Low strateg6c Segnmcence j NUREG/CR-3005:

SUMMARY

OF THE NUCLEAR REGULATORY COM- NUREG-1065 R0f: ACCEPTANCE CRITERIA FOR THE LOW EN-

MISSON'S LOFT PROGRAM RESEARCH FINDINGS. RICHED URANIUM REFORM AMENDMENTS.

I

62 Subject Index Low-Level Weste . Motoriate Test 4 NUREG/CR-41C1: ASSAY OF LONG UVED RADIONUCUDES IN LOW- NUREG/CR-4218: LOCA SIMULATION IN THE NATONAL RESEARCH LEVEL WASTES FROM POWER REACTORS. UNIVERSAL REACTOR PROGRAM Postrradiate Exarmnation Re-

NUREG/CR-4181

LEACHABILITY OF RADIONUCLIDES FROM suits For The Thrd Matenals Test (MT-3) Second Campaign.

! CEMENT SOUDIFIED WASTE FORMS PRODUCED AT OPERATING I NUCLEAR POWER REACTORS- Mesourement NUREG/CR-4194. LOW LEVEL NUCLEAR WASTE SHALLOW LAND BURIAL TRENCH ISOLATION Final Report, October 1981 - September NUREG/CR4245. IN-PLANT SOURCE TERM MEASURLMENTS AT BRUNSWICK STEAM ELECTRIC STATON NUREG/CR4200: BCDEGRADATON TESTING OF SOLIDIFIED LOW-LEVEL WASTE STREAMS.

" d'""

NUREG/CR4201: THERMAL STAB:UTY TESTING OF LOW-LEVEL NUREG/CR-3889. THE MODELING OF BWR CORE MELTDOWN ACCI.

WASTE FORMS. DENTS - FOR APPLICATION IN THE MELRPlMOD2 COMPUTER NUREGICR 4215- TECHNICAL FACTORS AFFECTING LOW-LEVEL CODE.

) WASTE FORM ACCEPTANCE CRITERIA.

M6cro-R-Meter Low Temperature Aging NUREG/CR4118. MONITORING METHODS FOR DETERMINATION j NUREG/CR4204: LONG-TERM EMBRITTLEMENT OF CAST DUPLEX COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT STAINLESS STEELS IN LWR SYSTEMS Annual Report. October 1983 *

, URANIUM RECOVERY SITES.

September 1984.

MAO-1 Spectacle Microetructure NUREG/CR-3953: THE USE OF MAG 1 SPECTACLES WITH POSITIVE. NUREG/CR-4124: NDE OF STAINLESS STEEL AND ON UNE LEAK MO 41TORiNG OF I WRS. Annual Report,Or*ober 1983 - September 4 AND NEGATIVE-PRESSURE RESPIRATORS.

1984.

MAPPS NUREG/CR 3626 V02: MAINTENANCE PERSONNEL PERFORMANCE M6gration SIMULATION (MAPPS) MODEL: DESCRIPTON OF MODEL NUREG/CR-4114- VALENCE EFFECTS ON THE SORPTION OF NU.

CONTENT. STRUCTURE,AND SENSITIVITY TESTING. CLIOES ON ROCKS AND MINERALS.II.

MATADOR Mieelle NUREG/CR4210

MATADOR.A COMPUTER CODE FOR THE ANALY- NUREG/CH 3551: SAFETY IMPUCATIONS ASSOCIATED WITH IN.

I SIS OF RADONUCUDE BEHAVIOR DURING DEGRADED CORE AC- PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTON SYS-TEMS IN NUCLEAR POWER PLANTS.

NU 2t A ME S FOR THE ANALYSIS OF I TRANSPORT AND DEPOSITON OF RADONUCUDES) CODE DE- Modeling I

SCRIPTION AND USCA S MANUAL NUREG/CR.3626 V02. MAINTENANCE PERSONNEL PERFORMANCE MELCOR SIMULATION (MAPPS) MODEL: DESCRIPTON OF MODEL NUREG/CR-4169. AN APPROACH TO TREATING RADIONUCLiDE CONT'ENT. STRUCTURE.AND SENSITIVITY TESTING i DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM j NUREG/CR4199- A DEMONSTRATION UNCERTA:NTY/ SENSITIVITY Monitoring 1 ANALYSIS ISING THE HEALTH AND ECONOMIC CONSEQUENCE NUREG/CR4124. NDE OF STAINLESS STEEL AND ON UNE LEAK 3

MODEL CRAC2. MONITORING OF LWRS Annual Report,0ctober 1983 - September 1984

MELRM MOD 2 NUREG/CR4118. MONITORING METHODS FOR DETERMINATION NUREG/CR-3889 THE MODELING OF BWR CORE MELTDOWN ACCI- COMPUANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT l

, DENTS FOR APPUCATION IN THE MELRPt MOD 2 COMPUTER URANtUM RECOVERY SITES.

CODE.

MINET Monte-Carlo $4rnulawn NUREG/CR 2331 V04 N3 SAFETY RESEARCH PROGRAMS SPON. NUREG/CR-3626 V02. MAINTENANCE PERSONNEL PERFORMANC7 SOAED BY OFFICE OF NUCLEAR REGULATORY SIMULATION (MAPPS) MODEL: DESCRIPTON OF MODEL RESEARCH Ouarterty Progress Report. July 1 -September 30.1984 CONTENT, STRUCTURE.AND SENSITIVITY TESTING.

NUREG/CR 2331 V04 N4. SAFETY RESEARCH PROGRAMS SPON- .

, SORED BY OFFICE OF NUCLEAR REGULATORY NDE i RESEARCH.Ouarterty Progress Report, October 1, De ember 31, NUREG/CR-4124: NDE OF STAINLESS STEEL AND ON UNE LEAK )

4 1984. MONITORING OF LWRS. Annual Report. October 1983 September

' 1984 Magnetic

, NUREG/CR-3174 VC2: GEOPHYSICAL GEOLOGICAL STUDIES OF NOT

] POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT NUREG/CR.4092. ORNL CHARACTERl2ATION OF HEAVY SECTON t ZONE. Annual Report For 1983- STEEL TECHNOLOGY PROGRAM PLATES 01.02.AND 03

" "8 R CR-3626 V02. MAINTENANCE PERSONNEL PERFORMANCE NUREG/CR-3804 V04 PHYSICS OF REACTOR SAFETY.Ouarterty SIMULATON (MAPPS) MODEL: DESCRIPTION OF MODEL ReportM& December M84 CONTENT. STRUCTURE.AND SENSITIVITY TESTING.

NUREGICR 3883- ANALYSIS OF JAPANESE-U S. NUCLEAR POWER Neutron Doe 6tnetry PLANT MAINTENANCE.

j NUREG-0975 V03 COMPILATION OF CONTRACT RESEARCH FOR Mondrel Loading Test THE MATERIALS ENGINEERING BRANCH.DfVISION OF ENGINEER-

! NUREG/CR 3980 V03 LIGHT. WATER REACTOR SAFETY FUEL SYS- ING TECHNOLOGY Annual Report For FY 1984 TEMS RESEARCH PROGRAMS. Quarterfy Progress Report, July-Sep-l 1 ember 1984 Neutron irred6etion NUREG/CR4086 TENSILE PROPERTIES OF IRRADIATED NUCLEAR

^

27 RADIATION PROTECTION TRAINING AT URANIUM HEX

• ATf0N SER l AFLUORIDE AND FUEL FABRICATION PLANTS.

I Neutron Spectral Characterlietion CAUFORNIA OFFSHORE SURVEY OF UCENSEES NUREG/CR-4031 V02. NEUTRON SPECTRAL CHARACTERl2ATION EG/ R4 USING RADIOACTIVE MATERIAL. FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST)lR.

RADIATION SERIES. "Neutroruct Calculations."

Meterial License NUREG/CR 4039 V03 NEUTRON SPECTRAL CHARACTERl2ATON NUREG/CR-3657. PREUMINARY SCREENING OF FUEL CYCLE AND FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST) tR.

BY. PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING. RADIATON SERIES. " Neutron Esposure Parameters "

Subject Index 63 Neutronics Calculation Operator Training NUREG/CR-4031 VCO: NEUTRON SPECTRAL CHARACTERIZATION NUREG/CR4139 THE MAILED SURVEY A TECHNIQUE FOR OBTAIN.

FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST)lR- ING FEEDBACK FROM OPERATIONS PERSONNEL.

RADIATION SERIES *Neutronics Calculatons "

Overcooling Accident Noding NUREG/CR-4106 PRESSURIZED-THERMAL-SHOCK TEST OF 6-IN -

NUREG/CR-4044. TRAC-PF1 LOCA CALCULATIONS USING FINE- THICK PRESSURE VESSELS PTSE 1 investigation 06 Warm Prestress-NODE AND COARSE-NODE INPUT MODELS- og And Upper-Shelf Arrest Nondestructive Evaluation Overexposure NUREG/CR-3998 V02- LIGHT-WATER-REACTOR SAFETY MATERIALS ENGINEER!NG RESEARCH PROGRAMS Quarterty NUREG-0090 V07 N04. REPORT TO CONGRESS ON ABNORMAL Progress ReportApnidune 1984 OCCURRENCES October-December 1984 NUREG/CR-4124 NDE OF STAINLESS STEEL AND ON-LINE LEAK PTS MONITORING OF LWRS Annual Report. October 1983 - Septembe' 1984- NUREG/CR4109. TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR-IZED-THERMAL SHOCK TRANS!ENTS AT CALVERT CLIFFS / UNIT Nondestructive Examination 1 A Coeuston Evneenng N NUREG-097t V03. COMPILATION OF CONTRACT RESEARCH FOR PWR-GALE THE MATEllALS ENGINEER!NG BRANCH. DIVISION OF ENGINEER-ING TECHNOLOGY Annual Report For FY 1984 NUREG-0017 RO1: CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS AND LIQUID EFFLUENTS FROM PRES-Nonequilibrium SURIZED WATER REACTORS (PWR GALE CODE)

NUREG/CR-3193. FORCED CONVECTIVE.NONEOUILIBRIUM, POST-CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATICN Packaging Material COMPARISON REPORT. NUREG/CR-3091 V04 REV!EW OF WASTE PACKAGE VERIFICATION TESTS Semiannual Report Covenng The Pened October 1983 - March Nuclear Plant Aging 1984 NUREG/CR-2331 V04 PO hAFETY RESEARCH PROGRAMS SPON- NUREG/CR-3091 V05 REVIEW OF WASTE PACKAGE VERiflCATION SORED BY OF9CE OF NUCLEAR. REGULATORY TESTS. Semiannual Report Covenng The Penod Apnl 1984 - Septem-RESEARCH Ouarterty P sgress Report. July 1 -September 30.1984. ber 1984.

Occupational Radiation Panel Session ,

NUREG/CR4160- H;STORICAL SUMVARY OF OCCUPATIONAL RADI- NUREG/CP-0065. TRANSACTIONS OF THE BTH INTERNATIONAL ATION EXPOSURE EXPERIENCE IN U S. COMMERCIAL NUCLEAR CONFERENCE ON STRUCTURE MECHANICS IN REACTOR POWER PLANTS TECHNOLOGY Panel Session J-K. Status of Research in Structural And Mecharucal Engineenng For Nuclear Power Plants.

NUREG-0090 V07 N04. REPORT TO CONGRESS ON ABNORMAL Penetration OCCURRENCES October December 1984 NUREG/CR-3855. CHARACTERlZATION OF NUCLEAR REACTOR Off Normal Event CONTAINMENT PENETRATION - F:NAL REPORT.

NUREG/CR-4064: STRt.,CTURAL RESPONSC OF LARGE PENETRA-NUREG/CR-4168. GT2F:A COMPUTER CODE FOR ESTIMATING LIGHT WATER REACTOR FUEL ROD FAILURES TiONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED TO LOADINGS BEYOND DESIGN P ASIS Official R xord NU EG. "

O A01 S03 NRC COMPREHENSIVE RECORDS D!SPOSI-gE CA 033 THE ROLE OF PERSONAL AIR SAMPLING IN RADI-ATION SAFETY PROGRAMS AND RESULTS 06 A LADORATORY Offshore Radioactive Matertal EVALUATION OF PERSONAL AIR-SAMPLit.G EOUtPMENT.

NUREG/CR4190 CALIFORNIA OFFSHORE SURVEY OF LICENSEES USING RADIOACTIVE MATERIAL Personnel Monitoring NUREG/CR-4160: HtSTORICAL

SUMMARY

OF OCCUPATIONAL RADI.

Operating Experience ATION EXPOSURE EXPERIENCE IN t,S. COMMERCIAL NUCLEAR NURdG/CR 3883. ANALYSIS OF JAPANESE-U.S NUCLEAR POWER POWER PLANTS.

PLANT MAINTENANCE NUREG/CR-3905 V01 R1: SEQUENCE CODING AND SEARCH Petitions For Rulemaking SYSTEM FOR LICENSEE EVENT REPORTS User's Guide. NUREG-0936 V04 N01: NRC REGULATORY AGENDA Ovarierty NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM Report. January-March 1965.

FOR LICENSEE EVENT REPORTS Code Listings.

NUREG/CR-3905 V03: SEQUENCE CODING AND SEARCH SYSTEM Pipe Crack FOR LICENSEE EVENT REPORTS Ccder's Manual-NUREG-1061 V05: REPORT OF THE U S NUCLEAR REGULATORY NUREG/CR-3905 V04. SEOUENCE CODING AND SEARCH SYSTEM COMMISSION PIPfNG REVIEW COMMITTEE Votume 5 Summary FOR LICENSEE EVENT REPORTS Coder's Manual- Piping Review Committee Conclusions and Recommendatons.

Operating Reactors Licensing Actions Pipe Support NUREG-0748 V05 NO2: OPERATING REACTORS LICENS!NG ACTIONS

SUMMARY

. Data As Of February 28,1985 (Orange Book) NUREG/CR4263. RELIABILITY ANALYSIS OF STIFF VERSUS FLEXI-NUREG-0748 V05 NO3- OPERATING REACTORS LICENS!NG ACTIONS BLE PIPING FINAL PROJECT REPORT'

SUMMARY

. Data As Of March 31.1985. (Orange Book)

Pipe Whip Restraint NUREG-0748 V05 N04. OPERATING REACTORS LICENSING ACTIONS

SUMMARY

. Data As Of Apnl 30,1985 (Orange Book) NUREG/CR4263. RELIABILITY ANALYSIS OF STIFF VERSUS FLEXI-BLE PIPING FINAL PROJECT REPORT.

Operations Personnel NUREG/CR-4139: THE MAILED SURVEY A TECHN10UE FOR OBTAIN- Piping ING FEEDBACK FRCM OPERATIONS PERSONNEL NUREG-0975 V03- COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS ENG:NEEnlNG BRANCH, DIVISION OF ENGINEER-Operator ING TECHNOLOGY Annual Report For FY 1984.

NUREG/CR-4206 A SELECT REVIEW OF THE RECENT (1979-1983) NUREG-1061 V02: REPORT OF THE US NUCLEAR REGULATORY BEHAVIORAL RESEARCH LITERATURE ON TRAINING SIMULA- COMMISSION PIPING REVIEW COMMITTEE. Volume 2 Evaluation Of TORS. Seismc Desegns - A Review Of Se sme Design Requirements For Nu-clear Power Plant Piping Operator Feedback Project NUREG-1061 V05: REPORT OF THE U S. NUCLEAR REGULATORY NUREG/CR4139 THE MAfLED SURVEY.A TECHNIOUE FOR OBTAIN- COMMISSION PIPING REVIEW COMMITTEE. Volume 5 Summary ING FEEDBACK FROM OPERATIONS PERSONNEL. Piping Review Committee Conclus ons and Recommendatens

l 64 Subject index NUREG/CR-3613 V02: EVALUATON OF WELDED AND REPAIR- NUREG/CR-4212: lN PLACE THERMAL ANNE AUNG OF NUCLEAR RE-WELDED STAINLESS STEEL FOR LWR SERVICE Annual Report for ACTOR PRESSURE VESSELS 1984.

Pressurtzed Thermal Shock Piping Reliability NUREG/CR-2331 V04 N3. SAFETY RESEARCH PROGRAMS SPON.

NUREG-1147. SEISMIC SAFETY RESEARCH PROGRAM PLAN. SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH Ouarter y Progress Report.Ju!y 1 -September 30.1984 E 4221: AN EVALUATON OF STRESS CORROSION CRACK GROWTH IN BWR PIPING SYSTEMS.

R 8 N ER RG TOR RESEARCH Ouarterty Progress Report. October 1 December 31, Plutonium 1984-NUREG/CR-4208. GASTROINTESTINAL ABSORPTON OF PLUTONIUM NUREG/CR 3703. ASSESSMENT OF SELECTED TRAC AND RELAP3 IN MICE. RATS, AND DOGS Application To Estat* stung Values Of f1 CALCULATONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK For Scoble Plutorwurn STUDY.

NUREG/CR-3977: RELAPS THERMAL-HYDRAUllC ANALYSES OF Post Accident Cleanup PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-NUREG/CR 3293 V01: TECHNOLOGY SAFETY AND COSTS OF DE- SON UNIT 2 PRESSURIZED WATER REACTOR COMMISSONING REFERENCE NUCLEAR FUEL CYCLE AND NON- NUREG/CR-4106. PRESSURIZED-THERMAL-SHOCK TEST OF 6-IN.

FUEL CYCLE FACluTIES FOLLOWING POSTULATED THICK PRESSURE VESSELS PTSE-1 Investigaton Of Warm Prestress-ACCOENTS Main Report ing And Upper Shelf . st.

NUREG/CR-3293 V02. TECHNOLOGY. SAFETY AND COSTS OF DE- NUREG/CR-4109. TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR.

COMMISSONING REFERENCE FUEL CYCLE AND NON-FUEL IZED-THERMAL SHOC 4 TRANSIENTS AT CALVERT CLIFFS / UNIT CYCLE FACIUTIES FOLLOWING POSTULATED 1.A Combustion Engineenng PWR.

ACCIDENTS Appendices.

Preventive Maintenance N C 38 HE EFFECTS OF POST-LOCA CONDITONS ON A

"" 3 ANALYSIS OF JAPANESE-U.S NUCLEAR POWER PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-P N bA '

DUSTRY. Price-Anderson Retrospective Premium Posterstical Heat Flux NUREG-1131: FINANCIAL ANALYSIS OF POTENTIAL RETROSPECTIVE NUREG/CR-3193: FORCED CONVECTIVE.NONEQUlUBRIUM. POST. PREMlUM ASSESSMENTS UNDER THE PRICE. ANDERSON CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION SYSTEM.

Probabilistic Risk Anatysis Postirradiation Examination NUREG/CR-4140 DOMINANT ACCIDENT SEQUENCES IN OCONEE 1 NUREG/CR-3810 V04: REACTOR SAFETY RESEARCH PRESSURIZED WATER REACTOR.

PROGRAMS Quart Report. October-December 1984 NUREG/CR4218; L SIMULATION IN THE NATONAL RESEARCH Probabilist6c Risk Assessment UNIVERSAL REACTOR PROGRAM Postirradiation Examination Re- NUREG/CR-3862: DEVELOPMENT OF TRANSlENT INITIATING EVENT suits For The Third Matenals Test (MT-3) - Second Campaagn. FREQUENCIES FOR USE IN PROBABILISTIC R'SK ASSESSMENTS.

Precipitation Scaveng6ng ,

NUREG/CR-4158 A COMPILATION OF INFORMATON ON UNCER. NUREG/CR-4010- SPECIFICATION OF A HUMAN RELIABluTY DATA TAINTIES INVOLVED IN DEPOSITION MODEUNG. BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE-Pressure Boundary Component AR POWER PLANTS.

NUREG/CR-3228 V03. STRUCTURAL INTEGRITY OF WATER REAC- NUREG/CR-4144. IMPORTANCE RANKING BASED ON AGING CON-TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For SIDERATIONS OF COMPONENTS INCLUDED IN PROBABILISTIC 1984~ RISK ASSESSMENTS.

NUREG/CR-4229. EVALUATION OF CURRENT METHODOLOGY EM.

Pressure Capacity PLOYED IN PROBABluSTIC RISK ASSESSMENT (PRA) OF FIRE NUREG/CR-4149- ULTIMATE PRESSURE CAPACITY OF REINFORCED EVENTS AT NUCLEAR POWER PLANTS.

AND PRESTRESSED CONCRETE CONTAINMENT. NUREG/CR4231: EVALUATON OF AVAILABLE DATA FOR PROBABI-USTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR Pressure Transient POWER PLANTS.

NUREG/CR-4225:

SUMMARY

OF EFFICIENCY TESTING OF STAND.

ARD AND HIGH-CAPACITY HIGH-EFFICIENCY PARTICULATE AIR Project Review FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA- NUREG 1125 V01 A COMPILATION OF REPORTS OF THE ADylSORY TION AND EXPLOSIVE SHOCK WAVES. COMMITTEE ON REACTOR SAFEGUARDS.19571964. Volume 1.Part 1.ACRS Repor*t t'.i Protect Reviews (A F)

Pressure Transmitter NUREG 1125 V02- A COMPILATION OF REPORTS OF THE ADVISORY NUREG/CR-3863: ASSESSMENT OF CLASS 1E PRESSURE TRANS.

COMMITTEE ON REACTOR SAFEGUARDS.19571984. Volume 2.Part MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRONMENT 1:ACRS Reports On Project Reviews (G-P)

SCREENING TESTS.

Protectivt Cover NU G V01 COMPILATION OF CONTRACT RESEARCH FOR NUREG/CR-4075: DESIGNING PROTECTIVE COVERS FOR URANIUM THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER- TAMS MS A Am ING TECHNOLOGY Annual A For FY 1984. Psychological Stress

" ^"

SE CON AINMEN MODEL NUREG/CR-4040: OPERATIONAL DECISIONMAKING AND ACTION SE-NUREG/CR-3746 V02: LWR PRESSURE VESSEL SUHVEILLANCE DO. LECTION UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER SIMETRY IMPROVEMENT PROGRAM Semiannual Progress PLANTS.

NUR 374 V03 RE VESSEL SURVEILLANCE DO. PuM Model SIMETRY IMPROVEMENT PROGRAM 1984 Annual Roort. October NUREG/CR-4159 COMPARISON OF THE 1981 INEL DISPERSION 1.1983 September 30,1984. DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS.

NUREG/CR 4086: TENSILE PROPERTIES OF IRRADIATED NUCLEAR GRADE PRESSURE VESSEL WELDS FOR THE THIRD HSST IRRA. Qualifications DIATON SERIES. NUREG/CR4051: ASSESSMENT OF JOB-RELATED EDUCATIONAL NUREG/CR4092: ORNL CHARACTERIZATION OF HEAVY SECTION OUAUFICATONS FOR NUCLEAR POWER PLANT OPERATORS.

STEEL TECHNOLOGY PROGRAM PLATES 01.02 AND 03 NUREG/CR-4106: PRESSURIZED THERMAL SHOdK TEST OF 6-IN- Quality Assurance THICK PPESSURE VESSELS.PTSE 1 Investigation Of Warm Prestress- NUREG/CR-4271: RECOMMENDED SAFETY. RELIABILITY.OUALITY ing And Upper-Sheff Arrest. ASSURANCE AND MANAGEMENT AEROSPACE TECHNIQUES WITH

l l

Subject index 65 POSSIBLE APPLICATON BY THE DOE TO THE HIGH LEVEL RADIO- Radionuclide ACTIVE WASTE REPOSITORY PROGRAM NUREG/CR-4101. ASSAY OF LONG-LIVED RADIONUCLIDES IN LOW-RAMONA 38 LEVEL WASTES FROM POWER REACTORS NUREG/CR-4169. AN APPROACH TO TREATING RADIONUCLIDE NUREGICR-2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON- DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM SORED BY OFFICE OF NUCLEAR REGULATORY NUREG/CR-4181: LEACHABILITY OF RADIONUCLIDES FROM RESEARCH.Ouarterty Progress Report. October 1 December 31 CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT CPERATING 1984.

NUCLEAR POWER REACTORS NUREG/CR4210: MATADOR A COMPUTER CODE FOR THE ANALY-N REG /CR 3703. ASSESSMENT OF SELECTED TRAC AND RELAP5 IDE TS N G TER R ACTORS CALCULATONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK NUREG/CR4211. MATADOR (METHODS FOR THE ANALYSIS OF NUR CA-3977: RELAPS THERMAL-HYDRAULIC ANALYSES OF

" ^ "^

R PT ND US U PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-NUREG/CR-4237. MOBILITY OF RADIONUCLIDES IN HIGH CHLORIDE SON UNIT 2 PRESSURIZED WATER REACTOR.

ENVIRONMENTS RPA NUREG/CR4245 IN-PLANT SOURCE TERM MEASUAEMENTS AT NUREG-1125 V06: A COMPILATION OF REPORTS OF THE ADVISORY BRUNSWICK STEAM ELECTR:C STATION COMMITTEE ON REACTOR SAFEGUARDS.19571984. Volume 6,Part 2.ACRS Reports On Genenc Svbrects (RPA - Appenen C).

Rad opharmaceutical Package NUREG/CR-4035. A HIGHWAY ACCIDENT INVOLVING RADf0 PHAR-Radiation MACEUTICALS NEAR BROOKHAVEN MISSISSIPPI ON DECEMBER NUREG/CR4033 THE ROLE OF PERSONAL AIR SAMPLING IN RADg- 3.1983.

ATION SAFETY PROGRAMS AND RESULTS OF A LABORATORY

~

NU E 14 THE T F EN ON E TAL TRESS ON N REG /CR4245 IN-PLANT SOURCE TERM MEASUREMENTS AT SYLGARD 70 SILICONE ELASTOMER. BRUNSWICK STEAM ELECTRfC STATON Radiation Effect Reaction Kine *les NUREG/CR-3514 V02. THE CHEMICAL BEHAVIOR OF IODINE IN NUREG/CP-0062; PROCEEDINGS OF THE CONFERENCE ON THE AP-AQUEOUS SOLUTIONS UP TO 150 C il Ra$ation-Redon Conditions PLICATION OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR WASTE REPOSITORY ASSESSMENT.

Radiation Monitoring NUREG/CR4160 HISTORFCAL

SUMMARY

OF OCCUPATIONAL RADI- Reactor Coolant Pressure Boundary ATION EXPOSURE EXPERIENCE IN U S COMMERCIAL NUCLEAR NUREG-1095 EVALUATION OF RESPONSES TO IE BULLETIN 82-POWER PLANTS 02 Degradation Of Threaded Fasteners in Aeactor Coolant Pressure Radiation Protection NUREG/CR-3469 V02 OCCUPATONAL DOSE REDUCTION AT NU- Reactor Coolant Purnp CLEAR POWER PLANTS Annota'ed Bibliography Of Selected Read- NUREG/CR 4077: REACTOR COOLANT PUMD SHAri SEAL BEHAV-ings in Radiation Protection And ALARA.

IOR DURING STATION BLACKOUT.

NUREG/CR-4033. THE ROLE OF PERSONAL AIR SAMPLING IN RADI-ATON SAFETY PROGRAMS AND RESULTS OF A LABORATORY Resctor Coolant System EVALUATION OF PERSONAL AIR-SAMPLING EQUIPMENT. NUREG/CR-4205: TRAP-MELT 2 USER'S MANUAL Radiation Protection Training Reactor Operator NUREG-1127. RADIATION PROTECTION TRAINING AT URANIUM HEX- NUREG/CR 4051: ASSESSMENT OF JOB RELATED EDUCATIONAL AFLOORIDE AND FUEL FABRICATION PLANTS-OUALIFICATIONS FOR NUCLEAR POWER PLANT OPERATORS.

NUREG-1134. RADIATON PROTECTICN TRAINING FOR PERdONNEL NUREG/CA-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT-EMPLOYED IN MEDICAL FACILITIES ABILITY PRACTICES Radiation Safety Survey Reactor Safeguards NUREG/CS4190: CALIFORNIA OFFSHORE SURVEY OF LICENSEES USING RADIOACTIVE MATERIAL NUREG-1125 VOI: A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.1957-1984 Volume 1.Part Radiation Sensitivity 1 ACRS Reports On Protect Revews (A F)

NUREG-1125 V02; A COMPILATION OF REPORTS OF THE ADVISORY NURE3/CR-3228 V03 STRUCTURAL INTEGRITY OF WATER REAC.

TOR PRESSURE BOUNDARY COMPONENTS Annual Report For COMMITTEE ON REACTOR SAFEGUARDS,19571984 Volume 2.Part 1984' 1.ACRS Reports On Protect Revews (G P)

NUREG-1125 V03. A COMPILATION OF REPORTS OF THE ADVISORY Rad 6oactive Gas COMMITTEE ON REACTOR SAFEGUARDS.1957-1984 Volume 3.Part NUREG/CR-4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL 1.ACRS Reports On Protect Revews (0-Z)

WASTE FORM ACCEPTANCE CRITERIA. NUREG-1125 V04 A COMPILATION OF REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.1957-1984 Volume 4,Part I Radioactive Material 1

2 ACRS Repoes On Genenc Subrects (Accident Anafysis - Genenc NUREG/CR-3611. RADIOACTIVE MATERIAL (RAM) ACCIDENT / INCL- Items)

DENT DATA ANALYSIS PROGRAM NUREG 1125 V05: A COMPILATION OF REPORTS OF THE ADVISORY NUREG/CR 3657. PRELIMINARY SCREEN:NG OF FUEL CYCLE AND COMMITTEE ON REACTOR SAFEGUARDS.1957-1984 Volume 5.Part BY-PRODUCT MATERIAL LICENSES FOR EMERGENCY PLANNING. 2 ACRS Reports On Genenc Subl ects (HTGR Regulatory Guides)

NUREG-1125 V06: A COMPILATION OF REPORTS OF THE ADVISORY Radioacthre Waste COMMITTEE ON REACTOR SAFEGUARDS.1957-1984 Volume 6.Part NUREG-0017 RO1: CALCULATION OF RELEASES OF RADIOACTIVE 2.ACRS Reports On Genenc Subjects (APA - Appendix C).

MATERIALS IN GASEOUS AND LIQUID EFFLUENTS FROM PRES-SURIZED WATER REACTORS (PWR-GALE CODE) Reactor Safety NUREG/CR 1755 ADOO1: TECHNOLOGY. SAFETY AND COSTS OF DE- NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN-COMMISSIONING NUCLEAR REACTORS AT MULTIPLE REACTOR STRUMENTATION IN REFLOOD TEST FACILITIES STATIONS. Effects On Decommissioning Of intenm inatnlity To Dispose NUREG/CR-3804 V34: PHYSICS OF REACTOR SAFETY Ouarterly Of Wastes Offs;te. Report. October-December 1984 NUREG/CR4194. LOW-LEVEL NUCLEAR WASTE SHALLOW LAND NUREG/CR-3810 V04. REACTOR SAFETY RESEARCH BURIAL TRENCH ISOLATION Final Report. October 1981 September PROGRAMS Quarterly Report. October-December 1984 1984. NUREG/CR-3816 V02: REACTOR SAFETY RESEARCH Ouarterty NUREG/CR4271: RECOMMENDED SAFETY. RELIABILITY.OUAitTY Report.Apnt-June 1984 ASSURANCE AND MANAGEMENT AEROSPACE TECHNIQUES WITH NUREG/CR-3885 V03- HIGH TEMPERATURE GAS COOLED REACTOR POSSIBLE APPLICAT!ON BY THE DOE TO THE HIGH LEVEL RADIO- SAFETY STUDIES FOR THE DIVISION OF ACCIDENT ACTIVE WASTE REPOSITORY PROGR/M. EVALUATION.Ouarterty Progress Report, July 1 - September 30,1984

66 Subject index NUREG/CR4177 V01: MANAGEMENT OF SEVERE NUREG/CR-2663 V02: INFORMATON NEEDS FOR CHARACTERIZA-ACCIDENTS Perspectrves On Managing Severe Accidents in Commer TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-cial Nuclear Pour Plants IC MEDIA Appendices.

NUREG/CR4177 V02: MANAGEMENT OF SEVERE ACCIDENTS Extending Plant Operating Procedures into The Seywe Research Acca.ient Regime. NUREG/CP4065. TRANSACTIONS OF THE 8TH INTERNATIONAL CONFERENCE ON STRUCTURE MECHANICS IN REACTOR NU EG/C 38 7 AT ACQUISITON AND CONTROL OF THE HMT - nd M a E or N ea P Pat SERIES V !4 RADIATION EXPERIMENT AT THE ORR- NUREG/CR-3005.

SUMMARY

OF THE NUCLEAR REGULATORY COM-RW MISSION'S LOFT PROGRAM RESEARCH FINDINGS.

NUREG/CR4237. MOBILITY OF RADIONUCLIDES IN HIGH CHLORIDE popg, ENVtRONME NTS.

NUREG/CR-3953. THE USE OF MAG-1 SPECTACLES WITH POSITIVE-Redon Process AND NEGATIVE-PRESSURE RESPIRATORS NUREG/CR4? i4 VALENCE EFFECTS ON THE SORPTON OF NU-CLIDES ON ROCKS AND MINERALS II. Respirator Quality Assurance Testing NUREG/CR4111: A COMPARATIVE STUDY OF HEPA FILTER EFF1-Redos Reaction CIENCIES WHEN CHALLENGED WITH THERMAL AND AIR-JET-NUREG/CR-3514 V02. THE CHEMICAL BEHAVIOR OF ODINE IN GENERATED DI-2-ETHYLHEXYL SEBECATE.D8 2 ETHYLHEXYL AQUEOUS SOLUTIONS UP TO 150 C 11 Radiation-Redox Conditions. PHTHALATE,AND SODIUM CHLORIDE.

Refdi-Reflood Retention Efficiency NUREG/CR-3651. ASSESSMENT OF THE ADEQUACY OF ORNL IN- NUREG/CR-3455 A COMPARISON OF IODINE. KRYPTON.AND XENON STRUMENTATION IN REFLOOD TEST FACILITIES RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD-Reform Amendment NUREG-1065 R01: ACCEPTANCE CRITERIA FOR THE LOW EN- Rift NU G 1065 R A PTANCE R EI FOR THE LOW EN- NUREG/CR-3174 V02. GEOPHYSICAL GEOLOGICAL STUDIES OF RICHED URANIUM REFORM AMENDMENTS. S ENSN & M W NO NG ZONE. Annual Report For 1983 Regulatory Agenda NUREG/CR4226. NEW MADRID SEISMOTECTONIC STUDY Actmties NUREG 0936 V04 N01: NRC REGULATORY AGENDA Ouarterly Dunng Fiscal Year 1983.

ReportJanuary-March 1985.

RW Regulatory And Technical Report NUREG/CR-3747: THE SELECTION AND TESTING OF ROCK FOR AR-NUREG4304 V10 N01: REGULATORY AND TECHNICAL MO3tNG URANIUM TAILINGS IMPOUNDMENTS REPORTS Compdation For Frst Quarter 1985.

Risit Assessment Regulatory Guides NUREG/CR 2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON-NUREG 1125 V05. A COMPILATION OF REPORTS OF THE ADVISORY SORED BY OFFICE OF NUCLEAR REGULATORY COMMITTEE ON REACTOR SAFEGUARDS 19571984 Volume 5.Part RESEARCH Ouarterty Progress Report. October 1 December 31, 2 ACRS Reports On Genene Subt ects (HTGR - Regulato y Guides). 1984.

UR G/CR4159 COMPARISON OF THE 1981 INEL DISPERSION DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS N -3747 M SMCD M EM & M @ E MORING URANIUM TAILINGS IMPOUNDMENTS Reliability NUREG/CR4010: SPECIFICATION OF A HUMAN RELIABILITY DATA Rules B ANK FOR CONOUCTING HRA SEGMENTS OF PRAS FOR NUCLE- NUREG-0936 V04 N01: NRC REGULATORY AGENDA Quarterty AR POWER PLANTS. ReportJanuary March 1985 NUREG/CR-4180: STATE OF THE-ART OF SOLID-STATE MOTOR CONTROLLERS SAFSTOR NUREG/CR-4271: RECOMMENDED SAFETY, RELIABILITY.OUALITY NUREG/CR-1755 ADOO1: TECHNOLOGY,$AFETY AND COSTS OF DE-ASSURANCE AND MANAGEMENT AEROSPACE TECHNIOUES WITH COMMISSIONING NUCLEAR REACTORS AT MULTIPLE-REACTOR POSSIBLE APPLICATION BY THE DOE TO THE HIGH LEVEL RADIO- STATIONS Effects On Decommissiorung Of Intenm inabdity To Dispose ACTIVE WASTE REPOSITORY PROGRAM Of Wastes Offsite.

Reliability Anatysis SGTR NUREG/CR-4220: RELIABILITY ANALYS;S OF CONTAINMENT ISOLA- NUREG-0844 DAFT FC NRC INTEGRATED PROGRAM FOR RESOLU-TON SYSTEMS TION OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 REGARD.

ING STEAM GENERATOR 1UBE INTEGRITY. Draft Report For Com-Repository ment-NUREG/CR4114- VALENCE EFFECTS ON THE SORPTION OF NU-CLIDES ON ROCKS AND MINERALS II. SROA NUREG/CR-4134. REPOSITORY ENVIRONMENTAL PARAMETERS NUREG/CR4271: REDOMMENDED SAFETY,REUABILITY OUALITY RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL ASSURANCE AND MANAGEMENT AEROSPACE TECHNIQUES WITH NUREG/C 4 VO . CRITICAL PARAME'ERS FOR A HIGH-LEVEL POSSIBLE APPLICATION BY THE DOE TO THE HIGH LEVEL RADIO-WASTE REPOSITORY. Volume 1 Basalt.

^ ^b Repository Condition Safe Storage NUREG/CR-3091 V04. REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR-1755 AD001. TECHNOt OGY,$AFETY AND COSTS OF DE-TESTS Semiannual Report Covenng The Penod October 1983 - March COMMISSIONING NUCLEAR REACTORS AT MULTIPLE-REACTOR 99g4 STATIONS Effects On Decommiss< rung Of Intenm inabaty To Dispose NUREG/CR 3091 V05 REVIEW OF WASTE PACKAGE VERIFICATION Of Wastes Offsite.

TESTS Semiannual Report Covenng The Penod Aprd 1984 - Septem-ber 1984. Safeguards NUREG4525 RIO: SAFEGUARDS

SUMMARY

EVENT LIST Repository Site (SSEL), REVISION 10.

NUREG/CR-2663 V01: INFORMATION NEEDS FOR CHARACTERtZA. NUREG/CR-4091 SAFETY / SAFEGUARDS INTERACTIONS DURING TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG- SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR IC MEDIA Main Report. FACTUTIES.

Subject Index 67 Safety NUREG-1125 V03. A COMPILATION OF REPORTS OF THE ADVISORY NUREG/CR4093 SAFETY / SAFEGUARDS INTERACTIONS DURING COMMITTEE ON REACTOR SAFEGUARDS.19571984 Volume 3.Part SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR 1 ACRS Reprts On Prolect Rewews (0-Z)

FACluTIES NUREG-1125 V04. A COMPILATION OF REPORTS OF THE ADVISORY NUREG/CR4271: RECOMMENDED SAFETY.REl.lASILITY.OUALITY COMM11 TEE ON REACTOR SAFEGUARDS.19571984 Volume 4.Part ASSURANCE AND MANAGEMENT AEROSPACE TECHNIOUES WITH 2:ACRS Reports On Genenc Subectsl (Accident Analysis Genenc POSSIBLE APPLICATION BY THE DOE TO THE HIGH LEVEL RADIO- Items)

ACTIVE WASTE REPOSITORY PROGRAM. NUREG-1125 V05: A COMPtLATION OF REPORTS OF THE ADVISORY Safety Evaluation Report COMMITTEE ON REACTOR SAFEGUARDS.1957-1984 Volume 5.Part 2 ACRS Reports On Genenc Sublects (HTGR - Regulatory Guides)

NUREG-0675 S28. SAFETY EVALUATON REPORT RELATED TO THE NUREG-1125 V06- A COMPILAllON OF REPORTS OF THE ADVISORY OPERATON OF DIABLO CANYON NUCLEAR POWER PLANT UNITS COMMITTEE ON REACTOR SAFEGUARDS.1957-1964 Volume 6.Part 1 AND 2 Docket Nos. 50-275 And 50-323 (Pacific Gas And Electnc 2 ACRS Reports On Genenc Sub lects (RPA - Appendix C).

Company)

NUREG-1147. SEISMIC SAFETY RESEARCH PROGRAM PLAN NUREG-Od?S S30: SAFETY EVALUATON REPORT RELATED TO THE NUREG/CR-2331 V04 N3 SAFETY RESEARCH PROGRAMS SPON.

OPERATON OF DIABLO CANYON NUCLEAR POWER PLANT. UNITS SORED BY OFFICE OF NUCLEAR REGULATORY 1 AND 2. Docket Nos. 50-275 And 50-323 (Pacific Gas And Electnc RESEARCH Ouarterfy Progress Report. July 1 -September 30,1984 Company) NUREG/CR-2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON-NUREu-0675 S31: SAFETY EVALUATION REPORT RELATED TO THE SORED BY OFFICE OF NUCLEAR REGULATORY OPERATION OF DIABLO CANYON NUCLEAR POWER PLANT. UNITS RESEARCH Ouarterty Progress Report. October 1 December 31, 1 AND 2 Docket Nos. 50 275 And 50-323 (Pacific Gas And Electnc igg 4 CompanV)

NUREG.0157 S10 SAFETY EVALUATION REPORT RELATED TO THE NUREG/CR-2531 R03. INTRODUCTORY USER'S MANUAL FOR THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC U S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE-SEARCH DATA BANK.

STATION. UNITS 1 AND 2 Docket Nos 50-445 And 50446(Texas Utah.

ties Electnc Company) Safety Training NUREG-0797 S11: SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC NUREG-1127. RADIATION PROTECTION TRAINING AT URANIUM HEX.

AFLUORIDE AND FUEL FABRICATION PLANTS.

STATION. UNITS 1 AND 2. Docket Nos. 50445 And 50446 (Texas Utils ties Generating Company. et a0 Safety-Related Systems NUREG-0857 SU8 SAFETY EVALUATON REPORT RELATED TO THE OPERATION OF PALO VERDE NUCLEAR GENERATING NUREG/CR-4004 CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF STATION. UNITS 1.2 AND 3 Docket Nos 50-528.50-529 And 50- WESTINGHOUSE BFD RELAYS IN SAFETY-RELATED SYSTEMS 530 (Anzona Pubhc Service Company. et al) Salt NUREG-0881 S06. SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF WOLF CREEK GENERATING ST ATION. UNIT NUREG/CR.2663 VOI: INFORMATION NEEDS FOR CHARACTERIZA-1 Docket No 50482 (Kansas Gas And Electnc Company et al) TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-IC MEDIA Main Report NUREG4887 S06. SAFETY EVALUATION REPORT RELATFO TO THE OPERATION OF PERRY NUCLEAR POWER PLANT,UNrl3 1 AND NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARACTERIZA-

2. Docket Nos. 50440 And 50441 (Cleveland Electnc tiluminating Con TION OF HfGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-IC MEDIA. Appendices.

NUW 0991 SO4. SAFETY EVALUATION REPORT RELATED Scram TO THE mERATION OF LIMERICK GENERATING STATION UNITS 1 AND NUREG/CR-4262 VO1: EFFECTS OF CONTROL SYSTEM FAILURES Nt F 13 ET N LA T THE ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC OPLRATION OF SHEARON MARRIS NUCLEAR POWER PLANT. UNIT BOILING WATER REACTOR Main Report

1. Docket No. 50-400. (Carolina Power And Light Company And North Seal Assembly NI O SA Y EVYUAbREPORT RELATED TO THE NUREG/CR4077. REACTOR COOLANT PUMP SHAFT SEAL BEHAV-OPERATION OF NINE MILE POINT NUCLEAR STATION. UNIT NO. IOR DURING STATION BLACKOUT.

2 Docket No. 50410. (Niagara Mohaak Power ation NUREG-1119: SAFETY EV4LUATION REPORT RE TED 0 THE RE. Search System NEWAL OF THE OPERATING LICENSE FOR THE CAVALIER TRAIN- NUREG/CR-3905 V01 RI: SEQUENCE CODING AND SEARCH ING REACTOR AT THE UNIVERSITY OF VIRGINIA. Docket No. 50- SYSTEM FOR LICENSEE EVENT REPORTS User's Guide NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM NtRE 13 AFETYbALUATION REPORT RELATED TO THE CONSTRUCTION PERMIT AND OPERATING LICENSE FOR THE RE-NU EG/ 39 5 EOU NCE I ND SEARCH SYSTEM SEARCH REACTOR AT THE UNIVERSITY OF TEXAS Docket No. 50- FOR LICENSEE EVENT REPORTS Coder's Manual NUREG/CR-3905 V04 SEQUENCE COOING AND SEARCH SYSTEM NU -1 S YVALUATION REPORT RELATED TO THE OP-FOR LICENSEE EVENT REPORTS Coder's Manual ERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 Security AND 2. Docket Nos. 504?4 And 50425 (Georgia Power Company.et al) l NUREG/CR4093 SAFETY / SAFEGUARDS INTERACTIONS DURING ggg,;y go,,, SAFETY-RELATED EMERGENCIES AT NJCLEAR POWER REACTOR NUREG-1128. TRIAL EVALUATIONS IN COMPARISON WITH THE 1983 FACILITIES.

SAFETY GOALS NUREG/CR4177 V01: MANAGEMENT OF SEVERE 3 A D S Per es On Managing Severe Accidents in Commer- REG 06 VOS: REPORT OF THE U.S NUCLEAR REGULATORY COMM!SSION PIPING REVIEW COMMITTEE Volume 5 Summa <y NUREG/CR4177 V02: MANAGEMENT OF SEVERE Piping Review Committee Conclus#ons and Recommendations.

ACCIDENTS Extending Plant Operating Procedures into The Severe NUREG/CR-3558. HANDBOOK OF NUCLEAR POWER PLANT SEISMIC Accider t R me FRAGILITIES. Seismic Safety Margins Research Program.

NUREG/CR-4y97: SAFETY GOAL SENSITIVITY STUDIES' S Cm Safety Margin NUREG 1061 V02: REPORT OF THE U.S NUCLEAR REGULATORY NUREG/CR-3558. HANDBOOK OF NUCLEAR POWER PLANT SEISMIC COMMISSION PIPING REVIEW COMMITTEE. Volume 2 Evaluation Of FRAGILITIES. Seism c Safety Marg ns Research Program. Seismic Designs A Review Of Seismic Design Requirements For Nu-clear Power Plant Piping.

Safety Research NUREG-1125 VOI: A COMPILATION OF REPORTS OF THE ADVISORY Seismic Research COMMITTEE ON REACTOR SAFEGUARDS.19571984 Volume 1,Part NUREG-1147: SFISMIC SAFETY RESEARCH PROGRAM PLAN 1 ACRS Reports On Protect Reviews NUREG/CP-0059 V01: PROCEEDINGS OF THE MITI-NRC SEISMIC IN-NUREG-1125 V02: A COMPILATION F REPORTS O(AOF F)

THE ADVISORY FORMATION EXCHANGE MEETING. VOLUME I COMMITTEE ON REACTOR SAFEGUARDS.1957-1984. Volume 2,Part NUREG/CR4095 TEST SERIES 2 SEISMIC-FRAGILITY TESTS OF 1.ACRS Reports On Protect Reviews (G-P)

NATURALLY-AGED CLASS 1E EXIDE FHC-19 BATTERY CELLS.

68 Subject Index NUREG/CR4096 TEST SERIES 3 SEISMIC FRAGILITY TESTS OF Shale NATURALLY AGED CLASS 1E C&D LCU 13 BATTERY CELLS NUREG/CR-2663 V01 INFORMATION NEEDS FOR CHARACTERf2A-NUREG/CR4097: TEST SERIES 4 SEISMIC-FRAGILITY TESTS OF TlON OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-NATURALLY AGED EXiDE EMP-13 BATTERY CELLS- IC MEDIA Main Report NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARACTERIZA-UR G C 0 TEST SERIES 3 SEISMIC FRAGILITY TESTS OF IC MEDIA Appendices NATURALLY AGED CLASS 1E C&D LCU-13 BATTERY CELLS NUREG/CR-4097. TEST SERIES 4 SEISMIC-FRAGILITY TESTS OF NATURALLY AGED EXIDE EMP-13 BATTERY CELLS.

h'fRE N /CR4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED Seismicity AND PRESTRESSED CONCRETE CONTAINMENT.

NUREG/CR-3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES OF NE nual Re or 1983 UR / 4 51: ASSESSMENT OF JOB-RELATED EDUCATIONAL QUAllFICATIONS FOR NUCLEAR POWER PLANT OPERATORS Seismotectonic NUREG/CR-4226 NEW MADRID SEISMOTECTONIC STUDY.Actwitnes Shipment Routo Dunng Fiscal Year 1983 NUREG-0725 ROS PUBLIC INFORMATION CIRCULAR FOR SHIP-MENTS OF IRRADIATED REACTOR FUEL.

Seismotectonics NUREG/CR-3178. STRUCTURAL AND TECTONIC STUDIES IN NEW Silicone Elastomer YORK STATE Final Report.Juey 1981 - June 1982. NUREG/CR-4147. THE EFFECT OF ENVIRONMENTAL STRESS ON Self-Contained Breathing Apparatus NUREG/CR-3953 THE USE OF MAG-1 SPECTACLES WITH POSITIVE- $11ver Alumina AND NEGATIVE. PRESSURE RESPIRATORS. NUREG/CR 3455 A COMPARISON OF IODINE. KRYPTON.AND XENON Senior Reactor Operator RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD.

NUREG/CR4051 ASSESSMENT OF JOB-RELATED EDUCATIONAL SORPTION MEDIA.

QUAUFICATIONS FOR NUCLEAR POWER PLANT OPERATORS. 33,y,, 33:3c, c,,

Sensitiv6ty Analysis NUREG/CR-3455 A COMPARISON OF IODINE. KRYPTON.AND XENON NUREG/CR-3904. A COMPARISON OF UNCERTAINTY AND SENSITIV. RETENTION EFFICIENC!ES FOR VARIOUS SILVER LOADED AD-ITY ANALYSIS TECHNOUES FOR COMPUTER MODELS SORPTON MEDIA NUREG/CR-4199 A DEMONSTRATION UNCERTAINTY / SENSITIVITY ANALYSIS USiNG THE HEALTH AND ECONOMIC CONSEQUENCE Silver Zeolite MOCEL CRAC2. NUREG/CR-3455: A COMPARISON OF IOD;NE, KRYPTON.AND XENON RETENTION EFFICIENCIES FOR VARIOUS SILVER LOADED AD-Sensitivity Study SORPTION MEDIA.

NUREG/CR-4197: SAFETY GOAL SENSITIVITY STUDIES.

Simulation Sequence Coding NUREG/CR4091- THE EFFECT OF ALTERNATIVE AGING AND ACCI-NUREGICA-3905 V01 R1: SEQUENCE CODING AND SEARCH DENT SIMULATIONS ON POLYMER PROPERTIES.

SYSTEM FOR LICENSEE EVENT REPORTS User's Gude NUREG/CR-3905 V02. SEQUENCE CODING AND SEARCH SYSTEM Simulator NU EG/ 5 EO E I MD SEARCH SYSTEM NUREG/CR4206: A SELECT REVIEW OF THE RECENT (1979-1983)

FOH LICENSEE EVENT REPORTS Coder's Manual. BEHAVIORAL RESEARCH LITERATURE ON TRAINING SilVULA-NUREG/CR-3905 V04: SEQUENCE COC.NG AND SEARCH SYSTEM TORS.

FOR LICENSEE EVENT REPORTS Coder's Manual Site Characterization Sequence Coding And Search System NUREG/CR 2663 V01: INFORMATON NEEDS FOR CHARACTERIZA-NUREG/CR4071: EXPLORATORY TREND AND PATTERN ANALYSIS TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-FOR 1981 LICENSEE EVENT REPORT DATA. IC MEDIA Main Report.

NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARACTER 12A.

Service Wear TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG-NUREG/CR4144 IMPORTANCE RANKING BASED ON AG!NG CON- IC MEDIA Appendices.

SIDERAT60NS OF COMPONENTS INCLUDED IN PROBABILISTIC RISK ASSESSMENTS- Snubber NUREG-1061 V02- REPORT OF THE U.S. N'JCLEAR REGULATORY N G/C 4070 "" "'" "

03. BlVALVE FOULING OF NUCLEAR POWER PLANT SERVICE-WATER SYSTEMS Factors Tha' May Intensify The Safety Consequences Of Bio'ouling-c a Pow r ant Poeng NUREG/CR4263. RELIABillTY ANALYSIS OF STIFF VERSUS FLEXI-Severe Accident BLE PIPING FINAL PROJECT REPORT.

NUREG/CR-3647: DESIGN AND FABRICATION OF A 1/8-SCALE Soil AnalYsis STEEL CONTAtNMENT MODEL NUREG/CR-3855: CHARACTERIZATION OF NUCLEAR REACTOR NUREG/CR-4118- MONITORING METHODS FOR DETERMINATION CONTAINMENT PENETRATION - FINAL PEPORT COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT NUREG/CR-3930: OBSERVED BEHAVIOR OF CESIUM.ODINE.AND URAN 1UM RECOVERY SITES.

TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-GRAM SoM Response NUREG/CR4064: STRUCTURAL RESPONSE OF LARGE PENETRA. NUREG 1147: SEISMIC SAFETY RESEARCH PROGRAM PLAN.

TlONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED TO LOADINGS BEYOND DESIGN BASIS Solid-State Motor Controller NUREG/CR-4177 V01: MANAGEMENT OF SEVERE NUREG/CR4180: STATE-OF THE-ART OF SOLID-STATE MOTOR ACCIDENTS Perspectwes On Managing Severe Accidents in Commer. CONTROLLERS.

cial Nuclear Power Plants NUREG/CR4177 V02: MANAGEMENT OF SEVERE Soluble Plutonium ALCIDENTS Extending Plant Operating Procedures into The Severe NUREG/CR4208. GASTROINTESTINAL ABSORPTON OF PLUTONIUM Acesdent Regime IN MICE. RATS, AND DOGS Applicaton To Establishing Values Of f t For Soluble Plutonium.

Severe Accident Anslyses NUREG/CR-3889. THE MODELINu v ne.n CORE MELTDOWN ACCl- Sorption DENTS FOR APPLICATION IN THE MELHH MOD 2 CGCER NUREG/CR-4114. VALENCE EFFECTS ON THE SORPTION OF NU-CODE. CLIDES ON ROCKS AND MINERALS 11.

a . . -- ,

Subject index 69 Spent Fuel TECHNOLOGY. Panel Sesson J K Status of Research in Structural NUREG-0725 ROS. PUBLIC INFORMATION CIRCULAR FOR SHIP- And Mechanscal Er9neenng For Nuclear Power Plants.

MENTS OF IRRADIATED REACTOR FUEL.

NUREG/CR-1756 ADD 01: TECHNOLOGY. SAFETY AND COSTS OF DE- Super System Code COMMISSIONING NUCLEAR REACTORS AT MULTIPLE-REACTOR NUREG/CR-2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON-STATIONS Ef'ects On Decomrmssoning Of intenm inabdity To Dispose SORED BY OF FICE OF NUCLEAR REGULATORY Of Wastes OMsate. RESEARCH Ouarterty Progress Report. October 1 December 31 Spent Fuel Rod NUREG/CR4084: DRY SPENT FUEL STORAGE TEST PLAN FOR DE- Suppression Pool Aerosol STRUCTIVE FUEL ROO EXAMINATIONS-NUREG/CR-3317: TECHMCAL BASES AND USER'S MANUAL FOR Stainless Steel Piping System THE PROTOTYPE OF SPARC - A SUPPRESSION POOL AEROSOL NUREG/CR4221: AN EVALUATION OF STRESS CORROSION CRACK REMOVAL CODE.

GROWTH IN BWR PIPING SYSTEMS.

Suppression System Stainless Steel Weld NL, REG /CR4230 PROBABILITY-BASED EVALUATION OF SELECTED NUREG/CR4015: EFFECT OF STAINLESS STEEL WELD OVERLAY FIRE PROTECTION FEATURES IN NUCLEAR POWER PLANTS CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED STEEL PLATES IN BENDING SERtES 1. Survelilance Dosimetry improvement Program NUREG/CA-3746 V03: LWR PRESSURE VESSEL SURVEfLLANCE DO-Station Blackout SIMETRY IMPROVEMENT PROGRAM 1984 Annual Report. October NUREG 1032 DRFT FC EVALUATION OF STATION BLACisOUT ACCI- 1.1983 Septemt er30,1984 DENTS AT NUCLEAR POWER PLANTS Techrucal Findings Related To NUREG/CR 3746 V02: LWR PRESSURE VESSEL SURVEILLANCE DO-Unresolved Safety issue A44 Draft Report For Comment. SIMETRY IMPROVEMENT PROGRAM Sermannual Progress NUREG/CR4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV- Report.Aprd 1984 September 1984.

OR DURING STATON BLACKOUT.

Survetilance Techniques UF 6 A REVIEW OF THE CURRENT UNDERSTANDtNG OF NUREG/CR4070 V03 BIVALVE FOULING OF NUCLEAR POWER THE POTENTIAL FOR CONTAINMENT FAILURE FROM IN-VESSEL PLANT SERVICE-WATER SYSTEMS Factors That May intensify The STEAM EXPLOSONS. afeh nsequeces Befouhng.

NUREG/CR-2718. STEAM EXPLOSION EXPERIMENTS WITH SINGLE DR S OF ON OxtDE MELTED WITH A CO2 LASER Part REG /CR-4118. MONITORING METHODS FOR DETERMINATION COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT Steam Generator URANIUM RECOVERY SITES.

NUREG4844 DRFT FC: NRC INTEGRATED PROGRAM FOR RESOLU. NUREG/CR4139: THE MAILED SURVEY A TECHNIOUE FOR 00TAIN-TlON OF UNRESOLVED SAFETY ISSUES A-3.A4 AND A-5 REGARD- ING FEED 8ACK FROM OPERATIONS PERSONNEL.

ING STEAM GENERATOR TUBE INTEGRITY.Dra't Report For Com- NUREG/CR-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT.

ment ABILITY PRACTICES.

NUREG4975 V03 COMPILATION OF CONTRACT RESEARCH FOR THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER- Sy1 gard 170 ING TECHNOLOGY Annual Report For FY 1984. NUREG/CR4147: THE EFFr 'T OF ENVIRONMENTAL STRESS ON NUREG/CR4276 VtBRATION AND WEAR IN STEAM GENERATOR SYLGARD 70 SILICONE ELASTOMER.

TUBES FOLLOWtNG CHEMICAL CLEANING SEMIANNUAL REPORT. Systematic Evaluation Program Steam Generator Tube Rupture NUREG-0829 DAFT: INTEGRATED PLANT SAFETY ASSESSMENT REPORT. SYSTEMATIC EVALUATION PROGRAM SAN ONOFRE NUREG/CR-4079: ANALYTIC STUDIES PERTAINING TO STEAM GEN-ERATOR TUBE RUPTURE ACCIDENTS- NUCLEAR GENERATING STATION UNIT 1 Docket No 50-206(South-ern Cahfornia Edison Company)

Steel TRAC NUREG/CR-4283. STUDY OF THE EFFECTS OF ELASTIC UNLOAD-INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS' N EG/N3703. ASSESSMENT OF SELECTED TRAC AND RELAP5 CALCULATIONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK Steel Containment Building STUDY.

NUREG/CR-3647. DE3IGN AND FABRICATION OF A 1/8-SCALE NUREG/CR4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE 1 STEEL CONT AINMENT MODEL PRESSURIZED WATER REACTOR.

Stratmed Debris Bed TRAC PF1 NUREG/CR-2951: THE D9 EXPERIMENT. Heat Removal From Stratified NUREG/CR4044. TRAC-PFI LOCA CALCULATIONS USING FINE-UO2 Debns. NODE AND COARSE-NODE INPUT MOCELS.

Stress Corrosion Cracking I TRAC-801 (Version 12)

NUREG/CR-2331 V04 N3. SAFETY RESEARCH PROGRAMS SPON-SORED BY OFFICE OF NUCLEAR NUREG/CR4196. OVERVIEW OF TRAC-BD1 (VERSION 12) ASSESS-REGULATORY MENT STUDIES.

RESEARCH Ouarterty Progress Report July 1 -September 30.1984 NUREG/CR-3998 V02: LIGHT-WATER-REACTOR SAFETY MATERIALS TRAC-PD2 ENGINEERING RESEARCH PROGR AMS Quarterly Progress Report.Apni-June 1984 NUREG/CR-3208: TRAC-PD2 DEVELOPMENTAL ASSESSMENT.

NUREG/CR-4221; AN EVALUATION OF STRESS CORROSION CRACK TR AC-PF1 GROWTH IN BWR PlPING SYSTEMS-N ' REG /CR4109: TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR-Structural Geology i?ED-THERMAL SHOCK TRANSIENTS AT CALVERT CLIFFS / UNIT NUREG/CR-3178: STRUCTURAL AND TECTONIC STUDfES IN NEW 1 A Combusten Engineenng PWR.

YORK STATE. Final Report. July 1981 June 1982.

TRAC 4F1/ MOD 1 Structuraf Steel NUREG/CR4155: TRAC PF1/ MOD 1 INDEPENDENT NUREG/CR-3228 V03: STRUCTURAL INTEGRITY OF WATER REAC- ASSESSMENT NORTHWESTERN UNIVER$1TY PERFORATED-PLATE TOR PRESSURE BOUNDARY COMPONENTS Annual Report For CCFL TESTS.

TRAN Structure Mechanics NUREG/CR-3757. TRAN 8-2.THE EFFECT OF LOW STEEL CONTENT NUREG/CP4065: TRANSACTIONS OF THE 8TH INTERNATIONAL ON FUEL PENETRATION IN A NON-MELTING CYLINDRICAL FLOW CONFERENCE ON STRUCTURE MECHANICS IN REACTOR CHANNEL,

70 Subject index TRAN B Experiment Thermal Annealing NUREG/CR ^944. TRAN B-3 EXPERIMENTAL INVESTIGATION OF NUREG/CR-4212. IN PLACE THERMAL ANNEAUNG OF NUCLEAR RC-FUEL CRUST STABILITY ON MELTING SURFACES OF AN ANNU- ACTOR PRESSURE VESSELS.

LAR FLOW CHANNEL.

Thermal Gradient Tube TRAN B-3 NUREG/CR-4105. AN ASSESSMENT OF THERMAL GRADIENT TUBE NUREG/CR 3944: TRAN B-3 EXPERIMENTAL INVESTIGATION OF RESULTS FROM THE Hi SERIES OF FISSON PRODUCT RELEASE FUEL CRUST STABILITY ON MELTING SURFACES OF AN ANNU- TESTS.

LAR FLOW CHANNEL Thermal Stability TRAP-MELT 2 NUREG/CR-4201: THERMAL STABILITY TESTING OF LOW-LEVEL NUREG/CR-4205: TRAP-MELT 2 USER'S MANUAL WASTE FORMS.

TRUNC Thermal-Hydraulic NUREG/CR-4192 THE ANALYSIS OF DRAINAGE AND CONSOUDA- NUREG/CR-2331 V04 N4. SAFETY RESEARCH PROGRAMS SPON.

TION AT TYPICAL URANIUM MILL TAILINGS SITES. SORED BY OFFICE OF NUCLEAR REGULATORY RESEARCH Ouarterty Progress Report. October 1 December 31 Taillogs 1984.

NUREG/CR-3747: THE SELECTION AND TESTING OF ROCK FOR AR- NUREG/CR-3804 V04 PHYS 6CS OF REACTOR SAFETY Ouarterty MOR!NG URANIUM TAluNGS IMPOUNDMENTS. Report. October-December 1984.

NUREG/CR-3306 URAN 1UM MILL TAl LINGS NUREG/CR-3820 V03 THERMAL / HYDRAULIC ANALYSIS RESEARCH NEUTRALIZATION CONTAMINANT COMPLEXATION AND TAIUNGS PROGRAM Ouarterty Report.Juty-September 1984 LEACHING STUDY. NUREG/CR-4079: ANALYTIC STUDIES PERTAINING TO STEAM GEN.

NUREG/CR-4075 DESIGNING PROTECilVE COVERS FOR URANIUM ERATOR TUBE RUPTURE ACCIDENTS.

MILL TAIUdGS PILES. A Rewew. NUREG/CR-4155: TR AC-PF I / MOD 1 INDEPENDENT NUREG/ CR4192: THE ANALYSIS OF DRAINAGE AND CONSOLIDA- ASSESSMENT NORTHWESTERN UNIVERSITY PERFORATED-PLATE TION AT TYPICAL URAN!UM M!LL TA! LINGS SITES. CCFL TESTS.

Tailings isolation Thermal-Hydraul6c Analyses NUREG/CR-4076' CETERMINATION OF COMPUANCE WITH CRITERIA NUREG/CR-3977- RELAPS THERMAL-HYDRAUUC ANALYSES OF FOR FINAL TAlUNGS DISPOSAL SITE RECLAMATON PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B ROBIN-SON UNIT 2 PRESSURIZED WATER REACTOR Technical Specificstion NUREG-1132. TiCHNICAL SPECIFICATIONS FOR DIABLO CANYON Thermal-Hydraulic Reactor Safety NUCLEAR POWER PLANT, UNIT NO 2. Docket No. 50-323 (PaofC NUREG/CR-2331 V04 N3: SAFETY RESEARCH PROGRAMS SPON-Gas and Electre C6npany) SORED BY OFFICE OF NUCLEAR REGULATORY NUREG-1133 TEMICAL SPECIFICATIONS FOR PALO VERDE NU- RESEARCH Ouarterty Progress Report. July 1 September 30,1984.

CLEAR GEWRATING STATION. UNIT 1 Docket No. 50-528 (Anzona Pubuc Servor Company) Thermodynamic Database NUREG-1136. TECHNICAL SPECIFICATIONS FOR WOLF CREEK GEN- NUREG/CP-0062: PROCEEDINGS OF THE CONFERENCE ON THE AP.

ERATING STATON, UNIT 1. Docket No. 50-482(Kansas Gas And PLICATION OF GEOCHEMICAL MODELS TO HIGH-LEVEL NUCLEAR Electnc Company) WASTE REPOSITORY ASSESSMENT Tectonics Thermoluminescent Dosimeter NUREG/CR 3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES OF NUREG/CR-4131: INVESTIGATION OF ALTERNATIVE MEANS TO AC-POSSIBLE EXTENS!ONS OF THE NEW MADRID FAULT COMPLISH THE GOALS OF BIENNIAL ION CHAMBER CAUBRA-ZONE. Annual Report For 19d3. TION Teletherapy Calibration Threaded Fastener NUREG/CR-4131: INVESTIGATION OF ALTERNATIVE MEANS TO AC- NUREG-1095: EVALUATON OF RESPONSES TO IE BULLETIN 82-COMPUSH THE GOALS OF 8:ENNIAL ION CHAMBER CALIBRA. 02 Degradation Of Threaded Fasteners in Reactor Coolant Pressure TION. Boundary Of Pressunzed Water-Reactor Plants.

Temperature Title List NUREG/CR-3703: ASSESSMENT OF SELECTED TRAC AND RELAP5 NUREG-0540 V07 NO2: TITLE UST OF DOCUMENTS MADE PUBLICLY CALCULATIONS FOR OCONEE-1 PRESSURIZED THERMAL SHOCK AVAILABLE February 1 28.1985.

STUDY. NUREG-0540 V07 NO3: TITLE LIST OF DOCUMENTS MADE PUBUCLY NUREG/CR-3872. DATA ACQUISITION AND CONTROL OF THE HSST AVAILABLE March 1-31,1985.

SERIES V IRRADIATION EXPERIMENT AT THE ORR NUREG-0540 V07 N04 TITLE UST OF DOCUMENTS MADE PUBUCLY AVAILABLE. Aprd 1 30,1985.

NUREG/CR-4092. ORNL CHARACTERIZATION OF HEAVY-SECTION Training STEEL TECHNOLOGY PROGRAM PLATES 01.02.AND 03 NUREG-1134. RADtATION PROTECTION TRAINING FOR PERSONNEL EMPLOYED IN MEDICAL FACILITIES.

Testing NUREG/CR-4051: ASSESSMENT OF JOB-RELATED EDUCATIONAL NUREG/CR-3091 V04. REVIEW OF WASTE PACKAGE VERIFICATION OUALIFICATIONS FOR NUCLEAR POWER PLANT OPERATORS.

TESTS Semiannual Report Covenng The Penod October 1983 - March NUREGiCR-4206: A SELECT REVIEW OF THE RECENT (1979-1983) 1984. BEHAVIORAL RESEARCH UTERATURE ON TRAINING SIMULA-NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFICATION TORS.

TESTS Sermannual Report Covenng The Penod Aptd 1984 Septem-ber 1984. Transient NUREG/CR-3862: DEVELOPMENT OF TRANSIENT INITIATING EVENT Theft FREOUENCIES FOR USE IN PROBABluSTIC RISK ASSESSMENTS NUREG-0525 RIO: SAFEGUARDS

SUMMARY

EVENT UST NUREG/CR-4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES (SSEL). REVISION 10. ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC BOILING WATER REACTOR Main Report.

Thermal Aging NUREG/CR-4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR 4147: THE EFFECT OF ENVIRONMENTAL STRESS ON ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC SYLGARD 70 SluCONE ELASTOMER- BOluNG WATER REACTOR. Appercces.

Thermal Analysis Transient Reactor Analysis Code NUREG/CR-4064: STRUCTURAL RESPONSE OF LARGE PENETRA- NUREG/CR-3208. TRAC-PD2 DEVELOPMENTAL ASSESSMENT.

TiONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED NUREG/CR-4140: DOMINANT ACCIDENT SEQUENCES IN OCONEE-1 TO LOADINGS BEYOND DESIGN BASIS. PRESSURIZED WATER REACTOR.

Subject index 71 NUREG/CR-4155. TR AC-PF1/ MOD 1 INDEPENDENT Uncertainty ASSESSMENT NORTHWESTERN UNIVERSITY PERFORATED ? LATE NUREG/CR4199 A DEMONSTRATION UNCERTAINTY / SENSITIVITY CCFL TESTS. ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE MODEL CRAC2.

Transition-Phase NUREG/CR-3757: TRAN B-2:THE EFFECT OF LOW STEEL CONTENT Uncertainty Analysis ON FUEL PENETRAT:ON IN A NON-MELTING CYLINDRICAL FLOW NUREG/CR-3904: A COMPARISON OF UNCERTAINTY AND SENSITIV-CHANNEL ITY ANALYSIS TECHNIOUES FOR COMPUTER MODELS.

Transport Uncertainty Estimate NUREGICR-4211: MATADOR (METHODS FOR THE ANALYSIS OF NUREG/CR-4158. A COMPILATION OF INFORMATION ON UNCER-TRANSPORT AND DEPOSITION OF RADIONUCLIDES) CODE DE- TAINTIES INVOLVED IN DEPOSITON MODELING SCRIPTION AND USER'S MANUAL.

Undergromd Disposal Transportation Accident NUREG/CR 4194. LOW-LEVEL NUCLEAR WASTE SHALLOW LAND NUREG/CR4035: A HIGHWAY ACCIDENT INVOLVING RADIOPHAR. BURIAL TRENCH ISOLATION Final Report October 1981 September MACEUTICALS NEAR BROOKHAVEN, MISSISSIPPI ON DECEMBER g gg4.

3,1983.

Unresolved Safety issue A-44 UR G/ R-40 5 A H WAY A IDENT INVOLVING RADIOPHAR-032 N m MMO & SW@ BmM M MACEUTICALS NEAR BROOKHAVEN. MISSISSIPPI ON DECEMSER D ec gs aW k 3,1983. nresM Safen issue A-44 Dan W 6 Comnt.

Unresolved Safety issues P UREG/CR4194. LOW LEVEL NUCLEAR WASTE SHALLOW LAND NUREG-0606 V07 NO2: UNRESOLVED SAFETY ISSUES RIAL TRENCH ISOLATION Final Report. October 1941 - September N REG 0844 D FT FC NR I TEG ATE P OGRAM FOR RESOLU-TION OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 REGARD-Tube ING STEAM GENERATOR TUBE INTEGRITY. Draft Report For Com-NUREG/CR-4276. VIBRATON AND WEAR IN STEAM GENERATOR ment.

8S FOLLOWING CHEMICAL CLEANING SEMIANNU AL NUREG4090 V07 N04. REPORT TO CONGRESS ON ABNORMAL Tube Integrtty OCCURRENCES October-December 1984 NUREG-0844 DAFT FC: NRC INTEGRATED PROGRAM FOR RESOLU-TION OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 REGARD. Uranium ING STEAM GENERATOR TUBE INTEGRITY. Draft Report For Com- NUREG/CR 4237: MOBILITY OF RADIONUCLIDES IN HIGH CHLORIDE ment. ENVIRONMENTS.

Tuff Uranium Mill NUREG/CR 2663 V01: INFORMATON NEEDS FOR CHARACTERIZA- NUREG/CR-4075. DESIGNING PROTECTIVE COVERS FOR URANIUM TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG- MILL TAILINGS PILES. A Review.

IC MEDIA Main Report.

NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARACTERIZA- Uranium Mill Tailings TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEOLOG- NUREG/CR-3747: THE SELECTION AND TESTING OF ROCK FOR AR.

IC MEDIA.Appereces MORING URANIUM TAILINGS IMPOUNDMENTS NUREG/CR 3091 V04. REVIEW OF WASTE PACKAGE VERIFICATION NUF EG/CR-3906: URANIUM MILL TAILINGS TESTS Semiannual Report Covenng The Penod October 1983 - March NEUTRALIZATION CONTAMsNANT COMPLEXATON AND TAILINGS 1984 LEACHING STUDY.

NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR-4076. DETERMINATION OF COMPLIANCE WITH CRITERIA TESTS. Semiannual Report Covenng The Penod Apnl 1984 - Septem- FOR FINAL TAILINGS DISPOSAL SITE RECLAMATION ber 1984. NUREG/CR4192: THE ANALYSIS OF DRAINAGE AND CONSOLIDA-Turbulent Transfer NUREG/CR 4158 A COMPILATON OF INFORMATION ON UNCER- Uranium Milling TAINTIES INVOLVED IN DEPOSITION MODELING. NUREG/CR-4176 EM:SSION CONTROL TECHNOLOGY AND QUALITY ASSURANCE NEEDS AT URANIUM MILLING FACILITIESincludes NUREG/CR 4 77: INVERTED ANNUAL FLOW EXPER! MENTAL STUDY.

@ g e n abngh Maintaining Air Pob tion Control Devices.

Two-Phase Measurement NUREG/CR 3651 SES ME OF THE ADEOUACY OF ORNL IN-

~

G 08 T DS FOR ESTIMATING RADIOACTIVE AND TOXIC AIRBORNE SOURCE TERMS FOR URANIUM MILLING OPER-Type A Package ATIONS NUREG/CR4035: A HIGHWAY ACCIDENT INVOLVING RADIOPHAR-CEUTICALS NEAR BROOKHAVEN,MISSISStPPI ON DECEMBER Use Guide SYSTEM FOR LICENSEE EVENT REPORTS User's Guide USI A-47 NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES User's Manual ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC NUREG/CR-2531 R03. INTRODUCTORY USER'S MANUAL FOR THE BOILING WATER REACTOR. Appendices. U.S. NUCLEAR REGULATORY COMMISSION REACTOR SAFETY RE-SEARCH DATA BANK.

Ultrasonic Testing NUREG/CR 3913: HECTR VERSION 10 USER'S MANUAL NUREG-0975 V03: COMPILATION OF CONTRACT RESEARCH FOR NUREG/CR-4205: TRAP-MELT 2 USER'S MANUAL THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGINEER. NUREG/CR-4211: MATADOR (METHODS FOR THE ANALYSIS OF ING TECHNOLOGY Annual Report For FY 1984. TRANSPORT AND DEPOSITION OF RADIONUCLIDES) CODE DE-SCRIPTION AND USER'S MANUAL Unavailab6fity NUREG/CR4144: IMPORTANCE RANKING BASED ON AGING CON- Valence SIDERATIONS OF COMPONENTS INCLUDED IN PROBABILISTIC NUREG/CR-4114. VALENCE EFFECTS ON THE SORPTION OF NU-RISK ASSESSMENTS. CLlDES ON ROCKS AND MINERALSII.

72 Subject index Vent Vaive Flow NUREG/CR-4134 REPOSITORY ENVIRONVENTAL PARAMETERS NUREG/CR 3703: ASSESSMENT OF SELECTED TRAC AND RELAP5 RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL CALCULATIONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK WASTE PACKAGES.

STUDY.

Water Chemistry Ventilation System NUREG/CR-3998 V02: LIGHT-WATER-REACTOR SAFETY MATERIALS NUREG4090 V07 N04: REPORT TO CONGRESS ON ABNORMAL ENGINEERING RESEARCH PROGRAMS Ouarterty Progress OCCURRENCES October-December 1984. Report.Apnt June 1984.

NUREG/CR-4191: SURVEY OF LICENSEE CONTROL ROOM HABIT.

ABILITY PRACTICES Weathering NUREG/CR-4225:

SUMMARY

OF EFFICIENCY TESTING OF STAND- NUREG/CR J747: THE SELECTON AND TESTING OF ROCK FOR AR-ARD AND HIGH4APACITY HIGH-EFFICIENCY PARTICULATE AIR MORING URANIUM TAILINGS IMPOUNDMENTS FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA-TION AND EXPLOSIVE SHOCK WAVES. Weld Overlay Visual Display Terminal NUREG/CR-4124: NDE OF FTAINLESS STEEL AND ON-LINE LEAK MONITORING OF LWRS. Annual Report.Odober 1983 - Sep: ember NUREG/CR-3987: COMPUTERIZED ANNUNCIATOR SYSTEMS. 1984 Void Fraction Weldment NUREG/CR-3651: ASSESSMENT OF THE ADEOUACY OF ORNL IN-STRUMENTATION IN REFLOOD TEST FACILITIES- NUREG/CR 3228 V03. STRUCTURAL INTEGRITY OF WATER REAC.

TOR PRESSURE BOUNDARY COMPONENTS Annual F.eport For 1984 N EG/CR4200- BCDEGRADATION TESTING OF SOLIDIFIED LOW- NUREG/CR-3613 V02: EVALUATON OF WELDED AP D REPAIR-LEVEL WASTE STREAMS. STAMESS STEEL @ WR SEREE AW Repod for 1984.

Waste Disposal NUREG/CR4194: LOW-LEVEL NUCLEAR WASTE SHALLOW LAND Westinghouse BFD Relays BURIAL TRENCH ISOLATON Final Report. October 1981 - Septerr.ber NUREG/CR-4004. CLOSEOUT OF IE BULLETIN 79-25 FAILURES OF 1984. WESTINGHOUSE OFD RELAYS IN SAFETY-RELATED SYSTEMS.

Waste Form Wet Depositen NUREG/CR4181: LEACHABILITY OF RADIONUCLIDES FROM NUREG/CR-4158: A COMPILATION OF INFORMATION ON UNCER-CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT OPERATING TAINTIES INVOLVED IN DEPOSITON MODEL'NG NUCLEAR POWER REACTORS NUREG/CR-4201 THERMAL STABILITY TESTING OF LOW-LEVEL WI N RE /CR-4118 MONITORING METHODS FOR DETERMINATON NUREG/CR-4215:' TECHNICAL FACTORS AFFECTING LOW-LEVEL COMPUANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT WASTE FORM ACCEPTANCE CRITERIA _ URANIUM RECOVERY SITES.

Waste Management Xenon NUREG/CR-4101: ASSAY OF LONG-UVED RADIONUCLIDES IN LOW. NUREG/CR-3455 A COMPARISON OF IODINE. KRYPTON.AND XENON LEVEL WASTES FROM POWER REACTORS. RETENTION EFFCENCIES FOR VARIOUS SILVER LOADED AD-SORPTION MEDIA Waste Package NUREG/CR-3091 V04: REVIEW OF WASTE PACKAGE VER:FICATON Zinc-Rich Paint TESTS.Semennual Report Covenng The Penod October 1983 - Ma ch NUREG/CR-3803. THE EFFECTS OF POST-LOCA CONDITIONS ON A 1984 PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFICATON DUSTRY.

TESTS Semiannual Report Covenng The Penod Apnl 1984 Septem-ber 1984 Zircaloy NUREG/CR-3900 V03. LONG-TERM PERFOnMANCE OF MATERIALS NUREG/CR-3980 V03. LIGHT WATER-REACTOR SAFETY FUEL SYS-USED FOR HIGH-LEVEL WASTE PACKAGING Ouaderty TEMS RESEARCH PROGRAMS. Quaderiy Progress Report. July Sep-Report October-December 1984. tember 1984.

b - - - - - - - - - - - -

NRC Originating Organization Index (Staff Reports)

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

ADVISORY COMMITTEE (S) DIVISION OF EMEPGENCY PREPAREDNESS & ENGINEERING RE-ACRS - ADVtSORY COMMITTEE ON REACTOR SAFEGUARDS SPONSE (POST 830103)

NUREG-1125 V01: A COMPILATON OF REPORTS OF THE ADVISO- NUREG-1095: EVALUATON OF RESPONSES TO IE BULLETIN 82-RY COMMITTEE ON REACTOR SAFEGUARDS.1957-1984. Volume 02. Degradation Of Threaoed Fasteners in Reactor Coolant Pressure 1.Part 1 ACRS Reports On Protect Revews (A-F). Boundary Of Pressunzed Water-Reactor Plants.

NUREG-1125 V02 A COMPILATION OF REPORTS OF THE ADVISO- DIVISION OF OA. VENDOR & TECHNICAL TRAINING CENTER PRO-RY COMMITTEE ON REACTOR SAFEGUARDS,1957-1984 Volume GRAMS (POST 85021 2.Part 1:ACRS Reports On Protect Reviews (G-P) NUREG-0040 V09 N01: LICENSEE CONTRACTOR AND VENDOR NUREG-1125 V03 A COMPfLATION OF REPORTS OF THE ADVISO- STATUS REPORT.Ouarterty Report. January-March 1985 (White RY COMMITTEE ON REACTOR SAFEGUARDS,1957-1984 Volume Book)

NL EG 112 V04 AC PI T OF RE S OF THE ADVISO- OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS RY COMMITTEE ON REACTOR SAFEGUARDS.19571984 Volume U EG- 18 VI NM NTAL SE MENT FOR RENEWAL OF 4.Part 2.ACRS Reports On Genenc Sub i ects (Accident Analysis - Ge-SPECIAL NUCLEAR MATERIAL LICENSE NO SNM-1107. Docket No.

NU EG 1125 V05 A COMPILATION OF REPORTS OF THE ADVISO- DIVI ON O A G S RY COMVITTEE ON REACTOR SAFEGUARDS.1957-1984. Volume NUREG-0725 ROS: PUBLIC INFORMATION CIRCULAR FOR SHIP-5,Part 2.ACRS Reports On Genenc Sublects (HTGR - Regulatory MENTS OF IRRADIATED REACTOR FUEL Guides) NUREG-1065 RO1: ACCEPTANCE CRITERIA FOR THE LOW EN-NUREG-1125 V06 A COMPILATION OF REPORTS OF THE ADV:SO- RICHED URAN!UM REFORM AMENDMENTS RY COMMITTEE ON REACTOR SAFEGUARDS 1957-1984 Volurne LICENSING POLICY & PROGRAMS BRANCH (PRE 850707) 6.Part 2 ACRS Reports On Genenc Subl ects (RPA Appendix C) NUREG-0525 R10: SAFEGUARDS

SUMMARY

EVENT LIST EDO OFFICE OF ADMINISTRATION DIVISlON OF TECHN! CAL INFORMATION & DOCUMENT CONTROL OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

NUREG-0304 V10 NO I - REGULATORY AND TECHNICAL OFFICE OF NUCLEAR REGULATORY RESEARCH, DIRECTOR REPORTS Compilaton For First Quarter 1985. NUREG-1032 DRFT FC; EVALUATION OF STATION BLACKOUT AC-NUREG-C540 V07 NO2: TITLE LIST OF DOCUMENTS MADE PUBLIC- CIDENTS AT NUCLEAR POWER PLANTS. Technical Findings Relat-LY AVAILABLE February 1-28,1985 ed To Unresolved Safety issue A-44 Draft Report For Comment NUREG-0540 V07 NO3. TITLE LIST OF DOCUMENTS MADE PUBLIC- DIVISION OF RISK ANALYSIS & OPERATIONS (POST 840429)

LY AVAILABLE March 1-31,1985 NUREG-1140 DRFT FC: A REGULATORY ANALYSIS ON EMERGEN.

NUREG-0540 V07 N04. TITLE LIST OF DOCUMENTS MADE PUBLIC- CY PREPAREDNESS FOR FUEL CYCLE AND OTHER RADIOAC-LY AVAILABLE. Apnl 1 30.1985 TIVE MATERIAL LICENSEES Draft Report For Comment.

NUREG-0750 V21101: INDEXES TO NUCLEAR REGULATORY COM- DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST MISSION ISSUANCES. January-March 1985. 840429; NUREG-0750 V21 NO2: NUCLEAR REGULATORY COMMISSION IS- NUREG-1127. RADIATON PROTECTION TRAINING AT URANIUM SUANCES FOR FEBRUARY 1985. Pages 275-469. . 'EXAFLUORIDE AND FUEL FABRICATION PLANTS.

NUREG-0750 v21 NO3: NUCLEAR REGULATORY COMMISSION IS- NUREG-1134: RADIATION PROTECTON TRAINING FOR PERSON.

SUANCES FOR MARCH 1985. Pages 471-559 NEL EMPLOYED IN MEDICAL FACILITIES.

NUREG-0750 V21 N04. NUCLEAR REGULATORY COMMISSON IS- DIVISION OF ENGINEERING TECHNOLOGY SUANCES FOR APR'L 1985. Pages 5611.041. NUREG-0606 V07 NO2: UNRESOLVED SAFETY ISSUES NUREG-0910 RO1 S03. NRC CCMPREHENSIVE RECORDS DISPOSI.

SUMMARY

. Data As Of May 17.1985. (Aqua Book)

TION SCHEDULE. NUREG-0975 V03: COMPILATION OF CONTRACT RESEARCH FOR DIVISION OF RULES AND RECORDS THE MATERIALS ENGINEERING BRANCH. DIVISION OF ENGI-NUREG-0936 V04 N01; NRC REGULATORY AGENDA Quarterly NEERING TECHNOLOGY. Annual Report For FY 1984 Report, January-March 1985. NUREG-1147: SEISMIC SAFETY FiESEARCH PROGRAM PLAN NUREG/CP-0065: TRANSACTIONS CF THE 8TH INTERNATIONAL EDO - OFFICE OF STATE PROGRAMS CONFERENCE ON STRUCTURE MECHANICS IN REACTOR OFFICE OF STATE PROGRAMS, DIRECTOR TECHNOLOGY.Panet Session J-K: Status of Research in Structural NUREG-1131: FINANCIAL ANALYSIS OF POTENTIAL RETROSPEC- And Mechanical Engineenng For Nuclear Power Plants.

TIVE PREM:UM ASSESSMENTS UNDER THE PRICE ANDERSON INTRA-AGENCY COMMITTEES, REVIEW GROUPS, ETC.

SYSTEM.

PIPING REVIEW COMMITTEE EDO - OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL NUREG-1061 V02: REPORT OF THE U.S. NUCLEAR REGULATORY DATA COMMISSION PIPING REVIEW COMMITTEE. Volume 2. Evaluation AEOD. DIRECTOR'S OFFICE Of Seismic Designs - A Rewew Of Sersmic Degn Requirements For NUREG-0090 V07 NO3: REPORT TO CONGRESS ON ABNORMAL Nuclear Power Plant Piping.

OCCURRENCES July-September 1984_ NUREG-1061 V05: REPORT OF THE U.S. NUCLEAR REGULATORY NUREG-0090 V07 NO4. REPORT TO CONGRESS ON ABNORMAL COMMISSION PIPING REVIEW COMMITTEE. Volume 5. Summary -

OCCURRENCES October-December 1984. Ppng Rewew Cornmittee Conclusions and Recommendations.

NUREG 1061 VO2 ADO REPORT OF THE U S. NUCLEAR REGULA-OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) TORY COMMISSON PIPING REVIEW COMMITTEE. Volume 2 DIRECTOR'S OFFICE, OFFICE OF INSPECTION AND ENFORCEMENT Addendum: Summary And Evaluation Of Histoncal Strong-Motion NUREG-0430 V05 NO1: L' CENSED FUEL FACILITY STATUS Earthquake Seismic Response And Damage To Aboveground Indus.

REPORT. Inventory Difference Data. January 1984 - June 1984(Gray tnal Ppng Book 11) STEAM EXPLOS!ON REVIEW GROUP ENFORCEMENT STAFF NUREG-1116: A REVIEW OF THE CURRENT UNDERSTANDING OF NUREG-0940 V04 N01. ENFORCEMENT ACTIONS SIGNIFICANT AC- THE POTENTIAL FOR CONTAINMENT FAILURE FROM IN-VESSEL TIONS RESOLVED.Ouarterty Progress ReportJanuary-March 1985. STEAM EXPLOSIONS.

73

74 NRC Originating OrganizatlOn Index ,

EDO-RESOURCE MANAGEMENT NUCLEAR GENERATING STATION UNIT 1 Docket No. 50-OFFICE OF RESOURCE MANAGEMENT. DIRECTOR 206 (Southern Cahfornsa Edison Company)

NUREG-1145 VO1: U S. NUCLEAR REGULATORY COMMISSION NUREG-0844 DAFT FC. NRC INTEGRATED PROGRAM FOR RESO-1984 ANNUAL REPORT. LUTION OF UNRESOLVED SAFETY ISSUES A-3.A-4 AND A-5 RE-OlVISION OF BUDGET & ANALYSIS GARDING STEAM GENERATOR TUBE INTEGRITY Draft Report For NUREG-0020 V09 NO3: LICENSED OPERATING REACTORS STATUS Comment.

SUMMARY

REPORT. Data As Of Febuary 28.1985 (Gray Book 1). NUREG-0857 S08: SAFETY EVALUATON REPORT RELATED TO NUREG-0020 V09 N04 LICENSED OPERATING REACTORS STATUS THE OPERATION OF PALO VERDE NUCLEAR GENERATING

SUMMARY

REPORT Data As Of March 31.1985 (Gray Book 1) STATION. UNITS 1.2 AND 3 Docket Nos 50 528.50-529 And 50-NUREG-0020 V09 N05: LICENSING OPERATING REACTORS 530 (Anzona Pubic Servce Company. et af)

STATUS

SUMMARY

REPORT. Data As Of Apnl 30,1985 (Gray Book NUREG-0881 S06 SAFETY EVALUATION REPORT RELATED TO 1)

THE OPERATION OF WOLF CREEK GENERATING STATION. UNIT MANAGEMENT SUPPORT BRANCH 1 Docket No 50-482 (Kansas Gas And Electre Company et al)

NUREG4748 VOS NO2. OPERATING REACTORS LICENSING AC- NUREG4887 S06: SAFETY EVALUATION REPORT RELATED TO TIONS

SUMMARY

. Data As Of February 28.1985 (Orange Book) THE OPERATION OF PERRY NUCLEAR POWER PLANT. UNITS 1 NUREG4748 V05 NO3; OPERATING REACTORS LICENSING AC- AND 2. Docket Nos 50-440 And 50441 (Cleveland Electre muminat.

TIONS

SUMMARY

. Data As Of March 31,1985. (Orange Book) ng Company)

NUREG-0748 V05 N04: OPERATING REACTORS LICENSING AC- NUHEG4991 504 SAFETY EVALUATION REPORT RELATED TO TIONS

SUMMARY

. Data As Of Apnl 30,19F ~) range Book) THE OPERATION OF LIMERICK GENERATING STATION. UNITS 1 AND 2. Docket Nos. 50-352 And 50-353 (Phdadelphia Electnc Com-OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80) pany)

OFFICE OF NUCLEAR REACTOR REGULATION. DIRECTOR NUREG 1033: FINAL ENVIRONMENTAL STATEMENT RELATED TO NUREG-1032 DRFT FC: EVALVATON OF STATION BLACKOUT AC-THE OPERATION OF WPPSS NUCLEAR PROJECT NO 3 Docket CIDENTS AT NUCLEAR POWER PLANTS 3ectwcal Findings Relat- No. 50-508 (Washington Pubhc Power Supply System) ed To Unresolved Safety issue A44 Draft Report For Comment.

NUREG-1038 S02: SAFETY EVALUATION REPORT RELATED TO DtV1SION OF ENGINEERING NUREG 1037 DRFT FC: CONTAINMENT PERFORMANCE WORKING THE OPERATION OF SHEARON HARRIS NUCLEAR POWER GPOUP REPORT. Draft Report For Comment PLANT, UNIT 1. Docket No. 50400. (Carohna Power And Light Com-SITE ANALYSIS BRANCH pany And North Carohna Eastern Municipal Power Agency)

NUREG4970; PROCEDURES FOR MEETING NRC ANTITRUST RE- NUREG-1047 S01: SAFETY EVALUATION REPORT RELATED TO THE OPERATON OF NINE MILE POINT NUCLEAR STATION. UNIT DIVI I N F MS INTEGRATION (POST 811005) NUREG-  : F NAL IRO E TA STA N D TO NUREG4017 RO1: CALCULATION OF RELEASES OF RADIOACTIVE THE OPERATON OF NINE M'LE POINT NUCLEAR STATION, UNIT MATER!ALS IN GASEOUS AND LIQUID EFFLUENTS FROM PRES-SURIZED WATER REACTORS (PWR-GALE CODE) NO. 2 Docket No 50410 (Na a Mohawk Power ation.et al)

NURtEG 1119. SAFETY EVALU TION REPORT RE TO THc N E 75 28 FETY EVALUATION REPORT RELATED TO RENEWAL OF THE CDERATING LICENSE FOR THE CAVALIER THE OPERATION OF DIABLO CANYON NUCLEAR POWER TRAINING REACTCR AT THE UNIVERSITY OF VIRGINIA Docket PLANT, UNITS 1 AND 2 Docket Nos. 50-275 And 50-323.(Pacdc Gas NUREG 12 CAL ICATIONS FOR DIABLO CANYON NUR ETY EVALUATION REPORT RELATED TO A ER MT, M M 2.DocW No W23 Pade THE OPERATION OF DIABLO CANYON NUCLEAR POWER NUR - 133 H PLANT,0N TS 1 AND 2. Docket Nos. 50-275 And 50-323 (Pacdic Gas AL PECIFICATIONS FOR PALO VERDE NU-And Electnc Company) CLEAR GENERATING STATION. UNIT 1 Docket No. 50-528. (Anzo-na Pubhc Service Com )

NUREG-0675 S31: SAFETY EVALUATION REPORT RELATED TO NUREG 1135. SAFETY EV LUATION REPORT RELATED TO THE THE OPERATION OF OtABLO CANYON NUCLEAR POWER CONSTRUCTION PERMIT AND OPERATING LICENSE FOR THE PLANT. UNITS 1 AND 2 Docket Nos. 50-275 And 50-323 (Pacific Gas RESEARCH REACTOR AT THE UNIVERSITY OF TEXAS Docket And Electnc Company)

No 50-602. (Univers, of Texas)

NUREG-0797 S10 SAFETY EVALUATION REPORT RELATED TO NUREG-1136. TECHNI L SPECIFICATIONS FOR WOLF CREEK THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC GENERATING STATION UNIT 1. Docket No. 50482 (Kansas Gas STATION, UNITS 1 AND 2. Docket Nos. 50-445 And 50446(Texas And Electnc Company)

UtAties Electnc Company)

NUREG-0797 S11: SAFETY EVALUATION REPORT RELATED TO NUREG-1137. SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF VOGTLE ELECTRIC GENERATING PLANT, UNITS THE OPERATION OF COMANCHE PEAK STEAM ELECTRIC I STATION. UNITS 1 AND 2 Docket Nos. 50445 And 50448 (Texas Utihtes Generatmg Company, et al)

NUREG O829 DRFT: INTEGRATED PLANT SAFETY ASSESSMENT REPORT, SYSTEMATIC EVALUATION PROGRAM - SAN ONOFRE 1 AND 2. Docket Nos, 50424 And 50425 (Georgia Power Company.et al)

DIVISION OF SAFETY TECHNOLOGY NURFG-1128. TRIAL EVALUATIONS IN COMPARISON WITH THE 1983 SAFETY GOALS

NRC Contract Sponsor Index (Contractor Reports)

This index lists the NRC organizations that sponsor organization is followed by the sponsored the contractor reports listed in NUREG/CR number and title of the this compilation. It is arranged alphabetically report (s) prepared by that organization. If by major NRC organization (e.g., program further information is needed, refer to the office) and then by subsections of these main citation by the NUREG/CR number.

(e.g., divisions) where appropriate. The EDO - OFFICE OF ADMINISTRATION WITH POSSlBLE APPLICATION BY THE DOE TO THE HIGH LEVEL DIVISION OF RULES AND RECORDS RADIOACTIVE WASTE REPOSITORY PROGRAM NUREG/CR-4040 OPERATIONAL DECISIONMAKING AND ACTION SELECTION UNDER PSYCHOLOGICAL STRESS IN NUCLEAR OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 4/05/81)

POWER PLANTS DIVISION OF ACCIDENT EVALUATION NUREG/CR.2331 V04 N3 SAFETY RESEARCH PROGRAMS SPON EDO-OFFICF FOR ANALYSIS & EVALUATION OF OPERATIONAL SORED BY OF FICE OF NUCLEAR RE GULATOR Y DATA RESEARCH Ouarterty Progress Report. July 1 -Septemter 30.1984 AEOD. DIRECTOR S OFFICE NUREG/CR-2331 V04 N4 SAFETY RESEARCH PROGRAMS SPON NUREG/CR-2000 V04 N3 LCENSEE EVENT REPORT (LE R) SORED BY OFFICE OF NUCLEAR REGULATORY NU EG/CR V04 4 L E SE EVENT REPORT (LER) RESEARCH On% %ess Report Och 1 - kW 30 NU EG/C V04 5 LI E EVENT REPORT (LER) NU EG/CR-2511 R03. INTRODUCTORY USER'S MANUAL FOR THE COMPILATION For Month Of May 19515. U.S NUCLEAR nEGULATORY COMMIS$10N RE ACTOR SAFETY NUREG/CR-3551 SAFETY IMPUCATIONS ASSOCIATED WITH IN. RESEARCH DATA BANK PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS- NUREG/CR-2718. STEAM EXPLOSION EXPERIMENTS WITH SINGL E TEMS IN NUCLEAR POWER PLANTS DROPS OF IRON OxtDE MELTED WITH A CO2 LASER Part NUREG/CR-3905 V01 Rt- SEQUENCE CODING AND SEARCH ll Parametnc Studies SYSTEM FOR LCENSEE EVENT REPORTS User's Gude NUREG/CR 2951 THE 09 EXPER! MENT Heat Removal From Strate NUREG/CR-3905 V02 SEQUENCE CODING AND SEARCH SYSTEM fied UO2 Debns FOR LICENSEE EVENT REPORTS Code Listings NUREG/CR-3005.

SUMMARY

OF THE NUCLEAR REGUL ATORY NUREG/CR-3905 V03 SEQUENCE CODING AND SEARCH SYSTEM COMMISSION'S LOFT PROGRAM RESE ARCH FINDINGS FOR LICENSEE EVENT REPORTS Coder's Manual NUREG/CR-3197 V01. REACTION BETWEEN SOME CESIUM IODINE NUREG/CR 3905 V04 SEQUENCE CODING AND SEARCH SYSTEM COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS FOR LICENSEE EVENT REPORTS Coder's Manual STEEL.INCONEL 600 & SILVER Volume 1Cesum Hydronie Reac-NUREG/CR-4071. EXPLORATORY TREND AND PATTERN ANALY- tions SIS FOR 1981 UCENSEE EVENT REPORT DATA. NUREG/CR-3208 TRAC-PD2 DEVELOPMENTAL ASSESSMENT.

NUREG/CR-3514 V02 THE CHEMICAL BEHAVIOR OF IODINE IN OFFICE OF INSPECTION & ENFORCEMENT (POST 12/11/80) AQUEOUS SOLUTIONS UP TO 150 C ll Radiate Redon Conditions DIVISION OF EMERGENCY PREPAREDNESS & ENGlNEERING RE-NUREG/CR- 151: ASSESSMENT OF THE ADEOUACY OF ORNL IN-NUR G/CR-4003 CLOS OUT OF IE BULLETIN 79-04 INCORRECT b '#

WElGHTS FOR SWING CHECK VALVES MANUFACTURED BY 3M MSSM & SEE WC W WM VELAN ENGINEERING CORPORATION CALCULATIONS FOR OCONE E-1 PRESSURIZED THERMAL ESTING OU E B D RELAYS 1 SAFE RE TED S TEMS NUR /C 37 V01 PRESSURE MEASUREMENTS IN A HYDRO-NUREGICR-4005 CLOSEOUT OF IE BULLETIN 80-12 DECAY HEAT GEN COMBUSTION ENVIRONMENT. Hydrogm Air Combuston Test REMOVAL SYSTEM OPERABILITY. Senes 1 And 2 in The FITS Tank DIVISION OF INSPECTON PROGRAMS (POST 850212) NUREGICA-3757. TRAN B-2 THE EFFECT OF LOW STEEL CON-NUREG/CR-4190 CAUFORNIA OFFSHORE SURVEY OF LICENS. TENT ON FUEL PENETRATION IN A NON MELTING CYLINDRICAL EES USING RADICAC'lVE MATERIAL FLOW CHANNEL NUREG/CP 103- THE EFFECTS OF POST-LOCA CONDtTIONS ON OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS A PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER DIVISION OF FUEL CYCLE & MATERIAL SAFETY INDUSTRY.

NUREG/CR 4035- A HIGHWAY ACCIDENT INVOLV!NG RADIOPHAR- NUREG/CR-3804 VO4. PHYSICS OF REACTOR SAFETY Ouarterty MACEUTICALS NEAR BROOFHAVEN. MISSISSIPPI ON DECEMBER Report October-December 1984 3.1983 NUREG/CR-3810 V04 REACTOR SAFETY RESEARCH D' VISION OF WASTE MANAGEMENT PROGRAMS Quarterty Report. October Decemter 1984 NUREG/CR-3091 V04 REVIEW OF WASTE PACKAGE VER; FICA

• TION TESTS Semiannual Report Covenng The Penod October 1983 NUREG/CR-3816 V02. REACTOR SAFETY RESEARCH Ouanerty Report.Apnt-June 1984.

NbRE /C 3 1 VOS REVtEW OF WASTE PACKAGE VERIFICA. SEARCH PROGRAM Ouarterty Report. July September 1984 TION TESTS Semiannual Report Covenng The Pered Aprd 1984 -

NUREG/CR-3885 V03 HIGH-TEMPERATURE GAS-COOLED REAC-NU GC 41 ASSAY OF LONG-UVED RADIONUCLIDES IN TOR SAFETY STUDIES FOR THE DIVISION OF ACCIDENT LOW-LEVEL WASTES FROM POWER REACTORS EVALUATION Ouarterty Progress Report. July 1 September NUREG/CR-4134. REPOSITORY ENVIRONMENTAL PARAMETERS 30J 984 RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL NUREG/CR-3889 THE MODEUNG OF BWR CORE MELTDOWN AC-WASTE PACKAGES CIDENTS - FOR APPLICATION IN THE MELRPI MOD 2 COMPUTER NUREG/CR-4200 BIODEGRADATION TESTING OF SOUDIFIED CODE.

LOW LEVEL WASTE STREAMS NUREG/CR-3913 HECTR VERSION 10 USER'S MANUAL NUREG/CR-4201: THERMAL STABILITY TESTING OF LOW-LEVEL NUREG/CR-3930 OBSERVED BEHAVIOR OF CESIUM LODINE.AND WASTE FORMS TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-NUREG/CR 4215. TECHNICAL FACTORS AFFECTING LOW LEVEL GRAM WASTE FORM ACCEPTANCE CRITERIA NUREG/CR 3944 TRAN B-3 EXPERIMENTAL INVESTIGATION OF NUREG/CR-4271- RECOMMENDED SAFETY.REUABILITY.OUAUTY FUEL CRUST STABluTY ON MELTING SURFACES OF AN ANNU-ASSURANCE AND MANAGEMENT AEROSPACE TECHNIOUES LAR FLOW CHANNEL 75

76 NRC Oontract Sponsor Index NUREG/CR-3977: RELAPS THERMAL-HYDRAJUC ANALYSES OF NUREG/CR4264. INVESTIGATION ON HIGH EFFICIENCY PARTICU-PRESEURIZED THERMAL SHOCK SEQUENCES FOR H 8. ROBIN- LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS.

SON UNIT 2 PRESSURIZED WATER REACTOR DIVISION OF RADIATION PROGRAMS & EARTH SCIENCES (POST NUREG/CR-3980 V03: LIGHT WATER-REACTOR SAFETY FUEL 840429)

SYSTEMS RESEARCH PROGRAMS. Quartetty Progress ReporUuly- NUREG/CR-2663 V01: INFORMATON NEEDS FOR CHARACTERIZA-Septembe- 1984. TON OF HIGH-LEVEL WASTE REPOSITORY SITES IN SIX GEO-NUREG/CR4044. TRAC-PF1 LOCA CALCULATONS USING FINE- LOGIC MEDIA Main Report.

NODE AND COARSE-NODE INPUT MODELS. NUREG/CR-2663 V02: INFORMATION NEEDS FOR CHARACT ERIZA-NUREG/CR4079: ANALYTIC STUDIES PERTAINING TO STEAM TION OF HIGH-LEVEL WASTE REPOSITORY SITES IN Six GEO-GENERATOR TUBE RUPTURE ACCIDENTS. LOGIC MEDIA. Appendices.

NUREG/CR-4105: AN ASSESSMENT OF THERMAL GRADIENT NUREG/CR-3174 V02: GEO'HYSICAL GEOLOGICAL STUDIES OF TUBE RESULTS FROM THE HI SERIES OF FISSION PRODUCT POSSIBLE EXTENSIONS OF THE NEW MADRO FAULT RELEASE TESTS ZONE. Annual Report For 1983.

NUREG/CR4109: TRAC-PF1 ANALYSES OF POTENTIAL PRESSUR- NUREG/C43178. STRUCTURAL AND TECTONIC STUDIES N NEW IZED-THERMAL SHOCK TRANSIENTS AT CALVERT CLIFFS / UNIT YORK STATE. Final Report. July 1981 June 1982.

1 A Combuston Engneennq PWR. NUREG/CR3657: PREUMINARY SCREENING OF FUEL CYOLE AND NUREG/CRat40: wMINANT ACCOENT SEQUENCES IN OCONEE- BY-PRODUCT MATERIAL LICENSES FOR EMERGENCY PLAN.

1 PRESSURIZED WATER REACTOR. NtNG.

NUREG/CR4155: TRAC-PF1/ MODI INDEPENDENT NUREG/CR-3747: THE EELECTION AND TESTING OF ROCK FOR ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORATED- ARMORING URANIUM TA! LINGS IMPOUNDMENTS.

PLATE CCFL TESTS. NUREG/CR-3900 V03. LJNG-TERM PERFORMANCE OF MATERI-NUREG/CR-4168: GT2F:A COMPUTER CODE FOR ESTIMATING ALS USED FOR HIGH-LEVEL WASTE PACKAGING Ouarterty LIGHT WATER REACTOR FUEL ROD FAILURES. ReportOctober-December 1984.

NUREG/CR4196: OVERVIEW OF TRAC BD1 (VERSON 12) AS- NUREG/CR3906: URANIUM MILL TAlUNGS NtRE 4 1 EMELT2 USER S MANUAL

" ^ " "^ ^ " ^"

NUREG/CR4218: LOCA SIMULATON IN THE NATIONAL RE. NGS E GS Y SEARCH UNIVERSAL REACTOR PROGRAM.Postrradiaton Exarro. NUREG/CR-3953: THE USE OF MAG-1 SPECTACLES WITH POSI-TIVE- AND NEGATIVE-PRESSURE RESPIRATOPS.

nation Results For TfL TNrd Matenals Test (MT-3) - Second Cam- NUREG/CR-4033. THE ROLE OF PERSONAL AIR SAMPUNG IN RA-NU /CR-4277: INVERTED ANNUAL FLOW EXPERIMENTAL DIATION SAFETY PROGRAMS AND RESULTS OF A LABORATO-RY EVALUATION OF PERSONAL AIR-SAMPLING EOUIPMENT.

DIVt ON F RISK ANALYS:S & OPERATIONS (POST 840429) NUREG/CR-4075: DESIGNING PROTECTIVE COVERS FOR URANI-NUREG/CR-3193, FORCED CONVECTIVE,NONEOUluBRIUM POST. UM MILL TAluNGS PILES. A Revew CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION NUREG/CR-4076: DETERMINATON OF COMPLIANCE WITH CRITE-COMPAR: SON REPORT RfA FOR FINAL TAIUNGS DISPOSAL SITE RECLAMATION NUREG/CR-3455: A COMPARISON OF CDINE, KRYPTON.AND NUREG/CR4088: METHODS FOR ESTIMATING RADIOACTIVE AND XENON RETENTK'N EFFICIENCIES FOR VAROUS SILVER TOxlO AIRBORNE SOURCE TERMS FOR URANIUM MILUNG OP-LOADED ADSORPT'ON MEDIA ERATONS.

NUREG/CR-3469 V02. OCCUPATONAL DOSE REDUCTON AT NU. NUREG/CR4111: A COMPARATIVE STUDY OF HEPA FILTER EFF1-CLEAR POWER PLANTS, Anno;ated Bibliography Of Selected Road. CIENCIES WHEN CHALLENGED WITH THERMAL. AND AIR-JET-ings in Radiation Protection And ALARA. GENERATED DI-2-ETHYLHEXYL SE8ECATE.DI-2-ETHYLHEXYL NUMEG/CR 3611: RAOlOACTIVE MATERIAL (RAM) ACCIDENT /INC3 PHTHALATE.AND SODIUM CHLORIDE.

DENT DATA ANALYSIS PROGRAM. NUREG/CR-4114. VALENCE EFFECTS ON THE SORPTON OF NU.

NUREG/CR-3626 V02- MAINTENANCE PERSONNEL PERFORM- CUDES ON ROCKS AND M!NERALS il.

ANCE SIMULAT'ON (MAPP9 MODEL DESCRIPTION OF MODEL NUREG/CR4118: MONITORING METHOD 9 FOR DETERMINATION CONTENT,STRUCTUriE,AND SENSITIVITY TESTING. COMPLIANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT NUREG/CR-3862- DE6ELOPMENT OF TRANSIENT INITIATING URANIUM RECOVERY SITES.

EVENT FREQUENCIES FOR USE IN PROBABluSTIC RISK AS- NUREG/CR-4131: INVESTIGATON OF ALTERNATIVE MEANS TO SESSMENTS ACCOMPLISH THE GOALS OF BIENN!AL ON CHAMBER Call-NUREG/CR-3904. A COMPARISON OF UNCERTAINTY AND SENSI- BRATION.

TIVITY ANALYSIS TECHNIOUES FOR COMPUTER MODELS. NUREG/CR4158. A COMPILATON OF INFORMATION ON UNCER-NUREG/CR-4009: HUMAN RELIABILITY DATA BANK.Evaluaton Re- TAINTIES INVOLVED IN DEPOSITON MODEUNG.

sutts NUREG/CR-4159: COMPARISON OF THE 1981 INEL DISPERSION NU9EG/CR4010 SPECIFICATION OF A HUMAN REUABiUTY DATA DATA WITH RESULTS FROM A NUMBER OF DlFFERENT BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NU- MODELS.

CLEAR POWER PLANTS. NUREG/CR-4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL NUREG/CR4093: SAFETY / SAFEGUARDS INTERACTIONS DURING WASTE REPOSITORE Volume t Basalt.

SAFETY RELATED EMERGENCIES AT NUCLEAR POWER REAC. NUREG/CR-4176: EMISSION CONTROL TECHNOLOGY AND OUAL-TOR FACILITIES. ITY ASSURANCE NEEDS AT URANIUM MILLING NUREG/CR 4169. A'4 APPROACH TO TREATING RADIONUCUDE FACluTIES.tncludes Supporting Methods For Testing. Operating.And DECAY HEATING FOR USE IN THE MELCOR CODE SYSTEM. Maintanng Air Pollution Control Devices.

NUREGICR-4177 V01: MANAGEMENT OF SEVERE NUREG/CR-4181: LEACHABluTY OF RADIONUCUDES FROM ACCIDENTS. Perspectives On Managing Severe Acodents in Com- CEMENT SOLIDIFIED WASTE FORMS PRODUCED AT OPERAT-mercal Nuclear Power Plants. ING NUCLEAR POWER REACTORS.

NUREG/CR-4177 V02: MANAGEMENT OF SEVERE NUREG/CR-4192: THE ANALYSIS OF DRAINAGE AND CONSOLIDA-ACCIDENTS Extending Plant Operating Procedures into The Severe TION AT TYPICAL URAN!UM MtLL TAluNGS SITES.

Accident Regime NUREG/CR-4194: LOW-LEVEL NUCLEAR WASTE SHALLOW LAND NUREG/CR4197: SAFETY GOAL SENSITMTY STUDIES- BURIAL TRENCH ISOLATION Final Report. October 1981 Septem-NUREG/CR-4199 A DEMONSTRATION UNCERTAINTY / SENSITIVITY ber1984 ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE NUREG/CR-4198 FRACTURE IN GLASS /HIGH LEVEL WASTE CAN-MODEL CRAC2. ISTERS.

NUREG/CR-4206: A SELECT REVIEW OF THE RECENT (1979-1983) NUREG/CR-4203. A CALCULATIONAL METHOD FOR DETERMINING BEHAVIORAL RESEARCH UTERATURE ON TRAINING SIMULA* BIOLOGICAL DOSE RATES FROM IRRADIAltD RESEARCH RE-TORS. ACTOR FUEL NUREG/CR-4210: MATADORA COMPUTER CODE FOR THE ANAL-NUREG/C44208 GASTROINTESTINAL ABSORPTON OF PLUTONI-YSIS OF RADIONUCUDE BEHAVIOR DURING DEGRADED CORE UM IN MICE, RATS. AND DOGS Application To Estabhshing values ACCIDENTS IN LIGHT WATER REACTORS- Of f1 For Soluble Plutonium.

NUREG/CR4211: MATADOR (METHODS FOR THE ANALYSIS OF NUREG/CR-4226. NEW MADRID SEISMOTECTON!C STUDY.Actnnties TRANSPORT AND DEPOSITION OF RADIONUCLIDES) CODE DE. Dunng Fiscal Year 1983 SCRIPTON AND USER'S MANUAL NUREG/CR-4237: MOBluTY OF RADIONUCLIDES IN HIGH CHLO-NUREG/CR-4225:

SUMMARY

OF EFFICIENCY TESTING OF STAND- RIDE ENV1RONMENTS.

ARD AND HIGH-CAPACITY HIGH EFFICIENCY PARTICULATE AIR DIVISION OF ENGINEERING TECHNOLOGY FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURI. NUREG/CR-1755 ADD 01: TECHNOLOGY. SAFETY AND COSTS OF ZATION AND EXPLOSIVE SHOCK WAVES. DECOMMISSIONING NUCLEAR REACTORS AT MULTIPLE-REAC.

NRC Contract Sponsor index 77 TOR STATONS Effects On Deconessenmg Of Intenm inatulity To NUREG/CR4097. TEST SERIES 4 SEISMIC-FRAGluTY TESTS OF D spose Of Wastes Offsite NATURALLY-AGED EXOE EMP 13 BATTERY CELLS NUREG/CR-3228 V03. STRUCTURAL INTEGRITY OF WATER REAC- NUREG/CR4106. PRESSURIZED-THERMAL SHOCK TEST OF 6-IN -

TOR PRESSURE BOUNDARY COMPONENTS. Annual Report For THICK PRESSURE VESSELS PTSE 1 investgaton Of Warm Pres-1984 tressang And Upper-Shelf Arrest NUREG/CR-3293 V01. TECHNOLOGY, SAFETY AND COSTS OF DE-NUREG/CR-4124. NDE OF STAINLESS STEEL AND ON UNE LEAK COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND MONITORING OF LWRS Annual Report. October 1983 September NON-FUEL CYCLE FACluTIES FOLLOWING POSTULATED 1984 ACCIDENTS Man Report. NUREG/CR4144. IMPORTANCE RANKING BASED ON AGING CON.

NUREG/CR-3293 V02: TECHNOLOGY, SAFETY AND COSTS OF DE-SIDERATONS OF COMPONENTS INCLUDED IN PROBABILISTIC COMMISSIONING REFERENCE FUEL CYCLE AND NON-FUEL RISK ASSESSMENTS CYCLE FACluTIES FOLLOWING POSTULATED NUREG/CR-4147: THE EFFECT OF ENVIRONMENTAL STRESS ON ACCOENTS Appendicas SYLGARD 70 SIUCONE ELASTOMER.

NUREG/CR-3317. TECHNICAL BASES AND USER'S MANUAL FOR NUREG/CR-4180. STATE-OF THE-ART OF SOUD-STATE MOTOR THE PROTOTYPE OF SPARC A SUPPRESSION POOL AEROSOL CONTROLLERS.

REMOVAL CODE-NUREG/CR4204 LONG-TERM EMBRITTLEMENT OF CAST DUPLEX NUREG/CR-3558. HANDBOOK OF NUCLEAR POWER PLANT SEIS' STAINLESS STEELS IN LWR SYSTEMS Annual Report, October MIC FRAGILITIES Seistmc Safety Maroins Research Program. 1983 - Septornber 1984 NUREG/CR-3613 V02 EVALUATION OF WELDED AND REPAIR-WELDED STAINLESS STEEL FOR LWR SERVICE. Annual Report for NUREG/CR4212: IN PLACE THERMAL ANNEAUNG OF NUCLEAR REACTOR PRESSURE VESSELS NUREG/CR-3647: DESIGN AND FABRICATION OF A 1/8-SCALE G Mn M SOWa M WWWN AT BRUNSWICK STEAM ELECTRIC STATION NUR C 3746 02 LW ESSURE VESSEL SURVEILLANCE NUREG/CR-4263. RELIABluTY ANALYSIS OF STIFF VERSUS FLEXI-DOSIMETRY IMPROVEMENT PROGRAM Sermannual Progress NUR G/ R4276 V BRA I N AND WEAR IN STEAM GENEHATOR NUR C 374 VO SSURE VESSEL SURVEILLANCE TUBES FOLLOWING CHEMICAL CLEANING SEMIANNUAL

^""'

1,1 83 30, 84 NURE / 4283 STUDY OF THE EFFECTS OF ELASTIC UNLOAD-NUR .CR-3810 V04 REACTOR SAFETY RESEARCH INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS NURE /CR 385 RA A ON O UCL R REACTOR OFFICE OF NUCLEAR REACTOR REGULATION (POST 4/28/80)

CONTAINMENT PENETRATION - FINAL REPORT DIVISION OF ENGINEERING NUREG/CA-3863. ASSESSMENT OF CLASS 1E PRESSURE TRANS- NUREG/CR4149. ULTIMATE PRESSURE CAPACITY OF REIN-MITTER RESPONSE WHEN SUBJECTED TO HARSH ENVIRON- FORCED AND PRESTRESSED CONCRETE CONTAINMENT.

MENT SCREENtNG TESTS NUREG/CR4221. AN EVALUATION OF STRESS CORROSION NUREG/CR-3872: DATA ACOUISITION AND CONTROL OF THE CRACK GROWTH IN BWR PIPING SYSTEMS.

HSST SERIES V IRRADIATON EXPERFMENT AT THE ORR. NUREG/CR4229 EVALUATION OF CURRENT METHOOOLOGY EM-NUREG/CR-3998 V02: UGHT-WATERREACTOR SAF ETY MATERI. PLOYED IN PROBABlUSTIC RISK ASSESSMENT (PRA) OF FIRE ALS ENGINEERING RESEARCH PROGRAMS Quarterty Progress EVENTS AT NUCLEAR POWER PLANTS Re. port,Apni-June 1984 NUREG/CR-4230: PROBABluTY BASED EVALUATON OF SELECT-NUREG/CR4015 EFFECT OF STAINLESS STEEL WELD OVERLAY ED FIRE PROTECTON FEATURES IN NUCLEAR POWER PLANTS.

CLADDING ON THE STRUCTURAL INTEGRITY OF FLAWED NUREG/CR4231. EVALUATION OF AVAILABLE DATA FOR PROB-STEEL PLATES IN BENDING SER ES 1. ABIUSTIC RISK ASSESSMENTS (PRA) OF FIRE EVFNTS AT NU-NUREG/CR4031 V02. NEUTRON SPECTRAL CHARACTERIZATON CLEAR POWER F JNTS DIVISION OF HUMAN FACTORS SAFETY FOR THE FIFTH HEAVY SECTION STEEL IRRADIATION SERIES "Neutrorucs Calculatons . TEC.HNOLOGY (HSST)NUREG/CR-3883. ANALYSIS OF JAPANESE-U.S. NUCLEAR POWER NUREG/CR4031 V03 NEUTRON SPECTRAL CHARACTERIZATION PLANT MAINTENANCE.

NUREG/CR-3987: COMPUTERIZED ANNUNCIATOR SYSTEMS.

FOR THE FIFTH HEAVY SECTION STEEL IRRADIATON SERIES. " Neutron Erposure Parameters TECHNOL.O. GY (HSST)

NUREG/CR4051: ASSESSMENT OF JOB-RELATED EDUCATIONAL NUREG/CR-4064 STRUCTURAL RESPONSE OF LARGE Pt.NETRA. OUAUFICATIONS FOR NUCLEAR PO'VER PLANT OPERATORS TlONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECT. NUREG/CR4139 THE MAILED SURVEY.A TECHNIQUE FOR OB-ED TO LOADINGS BEYOND GESIGN BASIS TAINING FEEDBACK FROM OPERATONS PERSONNEL NUREG/CR-4077: REACTOR COOLANT PUMP SHAFT SEAL BEHAV. DMSiON OF SYSTEMS INTEGRATION (POST 811005)

OR DURING STATON BLEMOUT NUREG/CR4191: SURVEY OF UCENSEE CONTROL ROOM HABIT.

NUREG/CR4064 CRY SPENT FUEL STORAGE TEST PLAN FOR ABILITY PRACTICES.

DESTRUCTIVE FUEL POD EXAMINATIONS NUREG/CR4220 REUABILITY ANALYSIS 05 CONTAINMENT ISO-NUREG/CA-4086 TENSILE FROPERTIES OF IRRADIATED NUCLE. LATrDN SYSTEMS AR GRADE PRESSURE VESSCL WELOS FOR THE TH;RD HSST DIVISION OF SAFETY TECHNOLOGY IRRADIATON SERIES NUREG/CR4160: HISTORICAL

SUMMARY

OF OCCUPATIONAL RA-NUREG/CR4091 THE EFFECT OF ALTERNATIVE AGING AND AC. DIATION EXPOSURE EXPERIENCE IN U S. COMMERCIAL NUCLE-CIDENT SIMULATONS CN POLYMER PROPERTIES AR POWER PLANTS NUREG/CR4092: ORNL CHARACTERIZAllON OF HEAVY-SECTON NUREG/CR4262 V01: EFFECTS OF CONTROL SYSTEM FAILURES STEEL TECHNOLOGY PROGRAM PLATES 01.02.AND 03 ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC NUREG/CR4095: TEST SERIES 2 SEISMIC-FRAGIUTY TESTS OF BOluNG WATER RE ACTOR Mam Report.

NATURALLY-AGED CLASS 1E EXIDE FHC 19 BATTERY CELLS NUREG/CR4262 V02. EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR4096 TEST SEHIES 3 SEISMIC FRAGILITY TESTS OF ON TRANSIENTS AND ACCIDENTS AT A GENERAL ELECTRIC NATURALLY AGED CLASS 1E C&D LCU-13 BATTERY CELLS. BOluNG WATEM REACTOR Appes.

N

Contractor Index This index lists, in alphabetical order, the numbers and titles of their reports. If further contractors that prepared the NUREG/CR information is needed, refer to the main ci-reports listed in this compilation. Listed tation by the NUREG/CR number.

below each contractor are the NUREG/CR AERONAUTICAL RESEARCH ASSOCIATES OF PRINCETON BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NUREG/CR-4158. A CCMPILATICN OF INFCRMATlGN ON UNCER- LABORATORIES TAINTIES INVOLVED IN DEPOSITION MOCELING NUREG/CR-4159 COMPARISON OF THE 1981 INEL DISPERSION NUREG/CR-1755 ADD 01: TECHNOLOGY. SAFETY AND COSTS OF DE-DATA WITH RESULTS FROM A NUMBER OF DIFFERENT MODELS- COMMISSIONING NUCLEAR REACTORS AT MULTIPLE. REACTOR STATIONS EMects On Decommissioning Of intenm inatAty To Dispose NUREG CR 3 4 0 PHYSICS OF REACTOR SAFETY Ouartery NUREG/ R 3 3 V01: TECHNOLOGY. SAFETY AND COSTS OF DE-NLR COMMISSIONING REFERENCE NUCLEAR FUEL CYCLE AND NON-C 3 5 H CTE IZATION OF NUCLEAR REACTOR FUEL CYCLE FACluTIES FOLLCWING POSTULATED CONTAINMENT PENETRATION FINAL REPORT. ACCIDENTS Main Report.

NUREG/CR-3980 V03 LIGHT WATER-REACTOR SAFETY FUEL SYS- NUREG/CR-3293 V02 TECHNOLOGY SAFETY AND COSTS OF DE.

TEMS RESEARCH PROGRAMS. Quarter 1y Progress Report.Jufy-Sep-tember 1984 COMMISGIONtNG REFERENCE FUEL CYCLE AND NON-FUEL CYCLE F ACILITIES FOLLOWING POSTULATED NUREG/CR-3998 V02 LIGHT-WATER-REACTOR SAFETY MATERIALS ACCIDENTS Appendices ENGINEER:NG RESEARCH PROGR AMS Ouarterty Progress Report.Aont-June 1984 NUREG/CR-3317: TECHNl CAL BASES AND USER'S MANUAL FOR NUHEG/CA-4064 STRUCTURAL RESPONSE OF LARGE PENETRA- THE PROTOTYPE OF SPARC - A SUPPRESSION POOL AEROSOL REMOVAL CODE.

TIONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED NUREG/CR-3613 V02 EVALUATION OF WELDED AND REPAIR NU EG 4 24 D F STA NLES STEEL AND ON-LINE LEAK MONITORING OF LWAS Annual Report. October 1983 - Septembe' 8 NUREG/CR-3747. THE SELECTION AND TESTING OF ROCK FOR AR-NU G/CR 4 83 STATE-OF-THE ART OF SOUD-STATE MOTOR NURE CR 80 V REA TOR SAFETY RESEARCH NUREG/CR-4191. SURVEY OF LICENSEE CONTROL ROOM HABIT- PROGRAMS Ouarterfy Report. October-December 1984 ABiUTY PRACTICES NUREG/CR-3827 ANALYSIS OF JAPANESE-U S NUCLEAR POWER NUREG/CR-4204 LONG-TERM EMBRITTLEMENT OF CAST DUPLEX PLANT MAINTENANCE.

STAINLESS STEELS IN LWR SYSTEMS Annual Report. October 1983 NUREG/CR-3906 UAAN!UM MILL September 1984 TAl LINGS NEUTRALIZATION CONTAM:NANT COMPLEXATION AND TAILINGS NUREG/CR-4208 GASTROINTESTINAL ABSORPTION OF PLUTON lUM LEACHING STUDY.

IN MICE, RATS. AND DOGS Applicat:on To Estabushing Values Of f1 For Soluble Plutonium NUREG/CR-3987 COMPUTERIZED ANNUNCIATOR SYSTEMS.

NUREG 'CR-4051 ASSESSVENT OF JOB-RELATED EDUCATIONAL NUREG/CR-4277. INVERTED ANNUAL FLOW EXPERIMENTAL STUDY. OUAllFICATIONS FOR NUCLEAR POWER PLANT OPERATORS ARIZONA, UNIV. 0F. TUCSON. AZ NUREG/CA-4070 V03. BlVALVE FOULING OF NUCLEAR POWER NUREG/CR-4194 LCW-LEVEL NUCLEAR WASTE SHALLOW LAND PLANT SERVICE + TER SYSTEMS Factors That May intans?y The BUR!AL TRENCH ISCLATICN Fir al Repo& October 1981 - September Safety Consecuences M Befouling 1984 NUREG/CR-4075: DESlurJNG PROTECTIVE COVERS FOR URANIUM M!LL TAILINGS PILES A Review ATOMIC ENERGY OF CANADA, LTD. NUREG/CR 4076' DETERMINATION OF COMPUANCE WITH CRITERIA NUREG/CR-4077 REACTOR COOLANT PUMP SHAFT SEAL BEHAV- FOR FINAL TAlpNGS DISPOSAL SITE RECLAMATION.

ICR DURiNG STATION BLACKOUT. NUREG/CA 4088 METHODS FOR ESTIMATING RADIOACTIVE AND TOXIC AIRBORNE SOURCE TERMS FOR URANIUM MILLING OPER.

BATTELLE HUMAN AFFAIRS RESEARCH CENTERS A TIONS NUREG/CR-3883 ANALYSIS OF JAPANESE-U.S NUCLEAR POWER NUREG/CR4118 MONITORING METHODS FOR DETERM! NATION PLANT MAINTENANCE COMPLtANCE WITH DECOMMISSIONING CLEANUP CRITERIA AT NUREG/CR-3987. COMPUTERIZED ANNUNCIATOR SYSTEMS URANIUM RECOVERY SITES NUREG/CR-4051: ASSESSVENT OF JOB-RELATED EDUCATIONAL OUALIFICATIONS FOR NUCLEAR POWER PLANT OPERATORS NUREG/CR-4139 THE MAILED SURVEY A TECHNIQUE FOR OBTAIN-NUREG/CR-4139 THE MAILED SURVEY A TECHNtQUE FOR OBTAtN- LNG FEEDBACK FROM OPERATIONS PERSONNEL ING FEEDBACK FRCM OPERATIONS PEPSONNEL- NUREG/CR-4144 IMPORTANCE RANKING BASED ON AGING CON-SIDERATIONS OF COMPONENTS INCLUDED IN PROBABlUSTIC BATTELLE MEMORIAL INSTITUTE. COLUMBUS LABORATORIES RISK ASSESSMENTS NUREG/CR-3900 V03 LONG-TERM PERFORMANCE OF MATERIALS NUREG/CR-4160. HISTOR! CAL

SUMMARY

OF OCCUPATIONAL RADI.

I, USED FOR HIGH-LEVEL WASTE ATION EXPOSURE EXPERIENCE IN U S. COMMERCIAL NUCLEAR PACKAGlNG.Ouarterty POWER PLANTS NUR C P A CE R ANKING BASED ON AGING CON-SIDERATIONS OF COMPONENTS INCLUDED IN PROBABillSTIC L GHT WATE REACTOR FUE ROD All RES NU NUREG/CR-4176 EM!SSION CONTROL TECHNOLOGY AND OUAUTY C 4 V01. MANAGEMENT OF SEVERE ASSURANCE NEEDS AT URANIUM M! LUNG FACluTIESincludes ACCIDENTS Perspect'ves On Managing Severe Accidents in Commer. Supp rting Methods For Testing. Operating.And Maintaining Asr Pollu-tion Control Devices N 177 VO MANAGEMENT OF SEVEHE NUREG/CR 4192. THE ANALYSIS OF DRAINAGE AND CONSOLIDA-ACCIDENTS Ertending Plant Operattng Procedures into The Severe TION AT TYPICAL URANIUM M!LL TAIUNGS SITES.

Accident R ime NUREG/CR-4218. LOCA SIMULATION IN THE NATIONAL RESEARCH NUREG/CR-4h05: TRAP-MELT 2 USER'S MANUAL UNIVERSAL REACTOR PROGRAM Postvradiation Examination Re-NUREG/CR-4210. MATADOR A COMPUTER CODE FOR THE ANALY- suits For The Third Materws Test lMT-3) . Second Campaign.

sis OF RADIONUCUDE BEHAVIOR DURING DEGRADED CORE AC- NUREG/CR-4220- RELIABluTY ANALYSIS OF CONTAINMENT ISOLA-CIDENTS IN UGHT WATER REACTORS TtON SYSTEMS NUREG/CR-4211. MATADOR IMETHODS FOR THE ANALYSIS OF TRANSPORT AND DEPOSITION OF RADIONUCUDES) CODE DE. NUREG/CR 4276 VIBRATION AND WEAR IN STEAM GENERATOR SCA:PTION AND USER'S MANUAL TUBES FOLLOWING CHEMICAL CLEANING SEMlANNUAL REPORT.

79

80 Contractor index BOSTON COLLEGE, CHESTNUT HILL, MA EG&G, INC.

NUREG/CR-3178: STRUCTURAL AND TECTONIC STUDIES IN NEW NUREG/CR-2531 R03 INTRODUCTORY USER S MANUAL FOR THE YORK STATE.Fulal Report. July 1981 June 1982. U.S. NUC' EAR REGULATORY COMMISSION REACTOR SAFETY RE-SEARCH 0ATA BANI(

BROOKHAVEN NATIONAL LABORATORY NUREG/CR-3005:

SUMMARY

OF THE NUCLEAR REGULATORY COM-NUREG/CP4059 V01: PROCEEDINGS OF THE MITI-NRC SEISMIC IN- MISSION'S LOFT PROGRAM RESEARCH FINDINGS.

FORMATION EXCHANGE MEETING. VOLUME L NUREG/CR-3191 FORCED CONVECTIVE.NONEOUluBRIUM. POST-NUREG/CR-2331 V04 N3: SAFETY RESEARCH PROGRAMS SPON- CHF HEAT TRANSFER EXPERIMENT DATA AND CORRELATION SORED BY OFFICE OF NUCLEAR REGULATORY COMPARISON RE FORT.

RESEARCH.Ouarterty Progress Report. July 1 -September 30,1984. NUREG/CR-39??. RELAP5 THERMAL-HYDRAUUC ANALYSES OF NUREG/CR-2331 V04 N4: SAFETY RESEARCH PROGRAMS SPON- PRESSURIZED THERMAL SHOCK SEQUENCES FOR H B. ROBIN-SORED BY OFFICE OF NUCLEAR REGULATORY SON UNIT 2 PRESSURIZED WATER REACTOR.

RESEARCH.Ouarterty Progress Report. October 1 - December 31 NUREG/CR-4033: THE ROLE OF PERSONAL AIR SAMPUNG IN RAD 6-1984. ATON SAFE *Y PROGRAMS AND RESULTS OF A LABORATORY NUREG/CR-3091 V04. REVIEW OF WASTE PACKAGE VERIFICATION EVALUATION OF PERSONAL AIR-SAMPUNG EQUIPMENT.

TESTS. Semiannual Report Covenng The Pened October 1983 March NUREG/CR4071: EXPLORAf OHY TREND AND PATTERN ANALYSIS 1984. FOR 1981 UCENSEE EVENT REPORT D ATA.

NUREG/CR-3091 V05: REVIEW OF WASTE PACKAGE VERIFICATION NUREG/CR 4084: DRY SPENT FUEL STORAGE TEST PLAN FOR DE-TESTS Sermannual Report Covenng The Penod Apnl 1984 Septem- STRUCTIVE FUEL ROD EXAMINATONS.

ber 1984.

NUREG/CH-3469 V02: OCCUPATIONAL DOSE REDUCTON AT NU- ERTEC WESTERN,8NC.,

CLEAR POWER PLANTS. Annotated Babhography Of Selected Read- NUREG/CO-2663 V01: INFORMATON NEEDS FOR CHARACTER 12A.

ings in Radiation Protection And ALARA. TION OF HIGH-LEVEL WASTE REPOSITOHY SITES IN SIX GEOLOG-NUREG/CR-3703: nSSESSMENT OF SELECTED TRAC AND RELAP5 BC MEDIA Main Report.

CALCULATIONS FOR OCONEE 1 PRESSURIZED THERMAL SHOCK NUREG/CR-2663 V02: INFORMATON NEEDS FOR CHARACTERIZA-STUDY. TION OF HIGH-LEVEL WASTE REPOSITORY S'TES IN SIX GEOLOG-NUREG/CR4093: SAFETY / SAFEGUARDS INTERACTONS DURING IC MEDIA.Appereces.

SAFETY-RELATED EMERGENCIES AT NUCLEAR POWER REACTOR FACIUTIES. FACTORY MUTUAL RESEARCH CORP.

NUREG/CR-4149 ULTIMATE PRESSURE CAPACITY OF REINFORCED NUREG/CR-4231: EVALUATON Or AVAILABLE DATA FOR PROBABI-AND PRESTRESSED CONCRETE CONTAINMENT- LISTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR NUREG/CR4200 BODEGRADATON TESTING OF SOLIDIFIED LOW- POWER PLANTS.

LEVEL WASTE STREAMS.

NUREG/CR-4201: THERMAL STABlWTY TESTING OF LOW-LEVEL GEEB'S, INC.

WASTE FORMS NUREG/CR4271: RECOMMENDED SAFETY.REUABIUTY,OUALITY NUREG/CR-4215: TECHNICAL FACTORS AFFECTING LOW-LEVEL ASSURANCE AND MANAGEMENT AEROSPACE TECHNIQUES WITH POSSIBLE APPUCATON BY THE DOE TO THE HIGH LEVEL RADIO-NUR G/ 221: A A AT STRESS CORROSION CRACK ACTIVE WASTE PEPOSITORY PROGRAM.

GROWTH IN BWR PIPING SYSTEMS.

NUREG/CR-4229 EVALUATON OF CURRENT METHODOLOGY EM- GENERAL PHYSICS CORP PLOYED IN PROBABluSTIC RISK ASSESSMENT (PRA) OF FIRE NUREG/CR4009: HUM N RELIABluTY DATA BANK Evaluation Re-EVENTS AT NUCLEAR POWER PLANTS. suus.

NUREG/CR4230 PROBABluTY BASED EVALUATON OF SELECTED NUREG/CR4010 SPECIFICATON OF A HUMAN REUABILITY DATA FIRE PROTECTION FEATURES IN NUCLEAR POWER PLANTS NUREG/CR4231: EVALUATON OF AVAILABLE DATA FOR PROBABI- BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE-AR POWER PLANTS.

USTIC RISK ASSESSMENTS (PRA) OF FIRE EVENTS AT NUCLEAR 0 HANFORD ENGINEERING DEVELOPMENT LABORATORY CAUFORNIA, STATE OF NUREG/CR-3746 V02 LWR PRESSURE VESSEL SURVEILLANCE 00-SIMETRY IMPROVEMENT PROGRAM. Semiannual Progress NUREG/CR-4190 CAUFORNIA OFFSHORE SURVEY OF UCENSEES USING RADCACTIVE MATERIAL Report.Apnl 1984 - September 1984 NUREG/CR-3746 V03: LWR PRESSURE VESSEL SURVEILLANCE DO-COLUMBIA UNIV., NEW YORK. NY SIMETRY IMPROVEMENT PROGRAM.1984 Annual Report. October NUREG/CR-4237: MOBlUTY OF RADONUCUDES IN HIGH CHLORCE 1,1983 - September 30,1984.

ENVIRONMENTS.

lOWA STATE UNIV., AMES,IA DAVID W. TAYLOR NAVAL RESEARCH & DEVELOPMENT CENTER NUREG/CR-4198: FRACTURE IN GLASS /HIGH LEVEL WASTE CANIS-NUREG/CR-4283: STUDY OF THE EFFECTS OF ELASTIC UNLOAD. TERS.

INGS ON THE Ji-R CURVES FROM COMPACT SPECIMENS.

JBF ASSOCIATES EG&Q lDAHO, lNC. (SUBS. 0F EG&G,1NC h NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-3862 DEVELOPMENT OF TRANSIENT INITIATING EVENT FOR UCENSEE EVENT REPORTS. Code Listings.

FREQUENCIES FOR USE IN PROBABluSTIC RISK ASSESSMENTS. NUREG/CR-3905 V03: SEOUENCE CODING AND SEARCH SYSTEM NUREG/CR-4040- OPERATONAL DECISIONMAKING AND ACTION SE- FOR UCENSEE EVENT REPORTS. Coder's Manual.

LECTON UNDER PSYCHOLOGICAL STRESS IN NUCLEAR POWER NUREG/CR-3905 904: SEQUENCE CODING AND SEARCH SYSTEM PLANTS. FOR UCENSEE EVENT REPORTS. Coder's Manual.

NUREG/CR-4181: LEACHABILITY OF RADONUCUDES FROM CEMENT SOUDIF;ED WASTE FORMS PRODUCED AT OPERATING LAWRENCE BERKELEY LA8 ORATORY NUCLEAR POWER REACTORS. NUREG/CR4161 V01: CRITICAL PARAMETERS FOR A HIGH-LEVEL NUREG/CR-4196: OVERVIEW OF TAAC-BD1 (VERSION 12) ASSESS- WASTE REPOSITORY. Vo6ume 1. Basalt.

MENT STUCIES.

NUREG/CR4203: A CALCULATIONAL METHOD FOR DETERMeNiNG LAWRENCE LIVERMORE NATIONAL LABORATORY BIOLOGICAL DOSE RATES FROM 1RRADIATED RESEARCH REAC- NUREG/CR-3558: HANDBOOK OF NUCLEAR POWER PLANT SEISMIC TOR FUEL FRAGIUTIES. Snsmac Safety Mar s Research Program.

NUREG/CR-4212: IN-PLACE THERMAL ANNEAUNG OF NUCLEAR RE- NUREG/CR-4035 A HIGHWAY IDENT INVOLVING RADIOPHAR-ACTOR PRESSURE VESSELS. MACEUTICALS NEAR BROOKHAVEN.MISSISS'PPI ON DECEMBER NUREG/CR-4245: IN-PLANT SOURCE TERM MEASUREMENTS AT 3.1983.

BRUNSWICK STEAM ELECTRIC STATON. NUREG/CR4263: RELIABluTY ANALYSIS OF STIFF VERSUS FLEXI-NUREG/CR-4262 VO1: EFFECTS OF CONTROL SYSTEM FAILURES BLE PIPING FINAL PHOJECT REPORT.

ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC BOILING WATER REACTOR. Main Report. LOS ALAMOS SCIENTIFIC LABORATORY NUREG/CR4262 V02: EFFECTS OF CONTROL SYSTEM FAILURES NUREG/CR-3208: TRAC PD2 DEVELOPMENTAL ASSESSMENT.

ON TRANSIENTS AND ACCOENTS AT A GENERAL ELECTRIC NUREG/CR-3953. THE USE OF MAG 1 SPECTACLES WITH POSITIVE-BOluNG WATER REACTOR. Appereces. AND NEGATIVE-PRESSURE RESPIRATORS.

s ContraCtcr ladex 81 NUREG/CR4079 ANALYTIC STUDIES PERTAINING TO STEAM GEN- NUREG/CR-4106. PRESSURIZED THERMAL-SHOCK TEST OF 6-IN -

ERATOR TUBE RUPTURE ACCIDENTS THICK PRESSURE VESSELS PTSE t investigaton Of Warm Prestress-NUREG/CR-4109 TRAC PF1 ANALYSES OF POTENTIAL PRESSUR- ing And Upper-Shelf Anest.

IZED THERMAL SHOCK TRANSIENTS AT CALVERT CLIFFS / UNIT NUREG/CR4114. VALENCE EFFECTS ON THE SORPTION OF NU-i A Combustion Engmeenng PWR- CUDES ON ROCKS AND MINERALS 11 NUREG/CR 4111: A COMPARATIVE STUDY OF HEPA FILTER EFF1- NUREG/CR4134- REPOSITORY ENVIRONMENTAL PARAMETERS CIENCIES WHEN CHALLENGED WITH THERMAL- AND AIR, JET.

SEBECATE,DI-2-ETHYLHEXYL RELEVANT TO ASSESSING THE PERFORMANCE OF HIGH-LEVEL GENERATED DI-2-ETHYLHEXYL WASTE PACKAGES.

NURE /C 40 00 T C T SEQUENCES IN OCONEE 1

^b '

PR'SSURIZED WATER REACTOR.

NURtG/CR-4225:

SUMMARY

OF EFFICIENCY TESTING OF STAND- TORS.

ARD AND HIGH-CAPACITY HIGH-EFFICIENCY PARTICULATE AIR PARAMETER INC.

FILTERS SUBJECTED TO SIMULATED TORNADO DEPRESSURIZA.

NUREG/CR 4003: CLOSEOUT OF IE BULLETIN 7904 INCORRECT NUREG/CR 4 ES G HIGH-EFFICIENCY PARTICU. WElGHTS FOR SWING CHECK VALVES MANUFACTURED BY LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS. VELA ENG ERING ATON R

MGTERIALS ENGINEERING ASSOCIATES,INC. WFSTINGHOUSE BFD RELAYS IN SAFETY-RELATED SYSTEMS NUREG/CR-3228 V03. STRUCTURAL INTEGRITY OF WATER REAC- NUREG/CR4005 CLOSEOUT OF IE BULLETIN 8012 DECAY HEAT TOR PRESSURE BOUNDARY COMPONENTS Annual Report For REMOVAL SYSTEM OPERABluTY.

1984 PITTS8URGH, UNIV. OF PITTS8URGH, PA NEW MEXICO STATE UNIV., LAS CRUCES, NM NUREG/CR3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES OF NUREG/CR-4264. INVESTIGATION ON HIGH-EFFICIENCY PARTICU- POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT LATE AIR FILTER PLUGGING BY COMBUSTION AEROSOLS- ZONE. Annual Report For 1983 NEW YORK, STATE UNIV. OF, ALBANY, NY PURDUE UNIV., WEST LAFAYETTE lN NUREG/CR-3178 STRUCTURAL AND TECTONIC STUDIES IN NEW YORK STATE Final Report. July 1981 June 1982. NUREG/CR-3174 V02: GEOPHYSICAL-GEOLOGICAL STUDIES OF POSSIBLE EXTENSIONS OF THE NEW MADRID FAULT OACI RIDGE NATIONAL LABORATORY ZONE. Annual Report For 1983 NUREG/CP-0062: PROCEEDINGS OF THE CONFERENCE ON THE AP.

STE SO A ESSM T NU G/CR 2 8 T EXP SiON EXPERiVENTS WITH SINGLE NUREG/CR-2000 V04. N3. LICENSEE EVENT REPORT (LER) DROPS OF IRON OXIDE MELTED WITH A CO2 LASER Part COMPILATION For Month Of March 1985. Il Pararnetnc Studies NUREG/CR-2000 V04 N4: UCENSEE EVENT REPORT (LER) NUREG/CR 2951: THE D9 EXPERIMENT Heat Removal From Stratified COMPILATION For Month Of Aprd 1985 UO2 Debns.

NUREG/CR-2000 V04 N 5. UCENSEE EVENT REPORT (LER) NUREG/C43197 V01: REACTION BETWEEN SOME CESIUM IODINE COMPILATION For Month Of May 1985. COMPOUNDS AND THE REACTOR MATERIALS 304 STAINLESS NUREG/CR-3514 VC2: THE CHEMICAL BEHAVIOR OF ODINE IN STEEL,1NCONEL 600 & SILVER. Volume I Cessum Hydroode Reac-AOUEOUS SOL UTONS UP TO 150 C 11 Radiation-Redor Conditions. tions NUREG/CR-3551: SAFETY IMPUCATIONS ASSOCIATED WITH IN-PLANT PRESSURIZED GAS STORAGE AND DISTRIBUTION SYS- NUREG/CR-3611' RADIOACTIVE MATERIAL (RAM) ACCIDENT / INCL-DENT DATA ANALYSIS PROGRAM NU EG/ R 3 02 IN EN E PERSONNEL PERFORMANCE STE N A NMEN DEL SIMULATION (MAPPS) MODEL: DESCRIPTION OF MODEL NUREG/CR-3657: PRELIM! NARY SCREENING OF FUEL CYCLE AND NU EG/CR 365 S SM N THE ADEO Y OF ORNL IN- BY-PRODUCT MATERIAL UCENSES FOR EMERGENCY PLANNING STRUMENTATON IN REFLOOD TEST FACILITIES NUREG/CR-3721 V01: PRESSURE MEASUREMENTS IN A HYDROGEN NUREG/CR-3872: DATA ACQUISITION AND CONTROL OF THE HSST COMBUSTION ENVIRONMENT. Hydrogen-Aa Combustion Test Senes SERIES V IRRADIATON EXPER! MENT AT THE ORR. 1 And 2 in The FITS Tank.

NUREG/CR-3885 V03: HIGH-TEMPERATURE GAS-COOLED REACTOR NUREG/CR-3757: TRAN B-2:THE EFFECT OF LOW STEEL CONTENT SAFETY STUDIES FOR THE DIVISION OF ACCIDENT ON FUEL PENETRATION IN A NON MELTING CYLINDRICAL FLOW EVALUATON Ouarterty Progress Report. J 1 September 30,1984. CHANNEL NUREG/CR 3889 THE MODELING OF BWR E MELTDOWN ACCI. NUREG/CR-3803 THE EFFECTS OF POST-LOCA CONDITIONS ON A DENTS FOR APPUCATION IN THE MELRPIMOD2 COMPUTER PROTECTIVE COATING (PAINT) FOR THE NUCLEAR POWER IN-CODE. DUSTRY.

NUREG/CR 3905 V01 R1: SEQUENCE CODING AND SEARCH NUREG/CR-3818 V02. REACTOR SAFETY RESEARCH Ouarterty SYSTEM FOR LICENSEE EVENT REPORTS User's Guide Report.Aprd-June 1984.

NUREG/CR-3905 V02: SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-3820 V03: THERMAL /HYDRAUUC ANALYSIS RESEARCH FOR LICENSEE EVENT REPORTS. Code Listings PROGRAM.Ouarterly Report. July-September 1984 NUREG/CR 3905 V03. SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-3855: CHARACTER 12ATION OF NUCLEAR REACTOR FOR LICENSEE EVENT REPORTS Coder's Manual CONTAINMENT PENETRATION - FINAL REPORT NUREG/CR-3905 V04 SEQUENCE CODING AND SEARCH SYSTEM NUREG/CR-3863: ASSESSMENT OF CLASS 1E PRESSURE TRANS-NU EG - RVE BEH F CESIUM. LODINE.AND MITTER RESPCNSE WHEN SUBJECTED TO HARSH ENVtRONMENT TELLURIUM IN THE ORNL FISSION PRODUCT RELEASE PRO-NU E /C 3 A COMPARISON OF UNCERTAINTY AND SENSITIV-NUREG CR4015: EFFECT OF STAINLESS STEEL WELD OVERLAY ITY ANALYSIS TECHNOUES FOR COMPUTER MODELS CLADDING ON THE STRUCTURAL IN~EGRlTY OF FLAWED STEEL NUREG/CR-3913: HECTR VERSON 10 USER'S MANUAL.

PLATES IN BENDING SERIES 1 NUREG/CR-3944: TRAN B-3 EXFERIMENTAL INVESTIGATION OF NUREG/CR4031 V02: NEUTRON SPECTRAL CHARACTERIZATON FUEL CRUST STABluTY ON MELTING SURFACES OF AN ANNU-FOR THE FIFTH HEAVY SECTION STEEL TECHNOLOGY (HSST) IR. LAR FLOW CHANNEL RADIATION SERIES "Neutrorucs Calculations. NUREG/CR4009: HUMAN REUABluTY DATA BANK Evaluation Re-NUREG/CR4031 V03 NEUTRON SPECTRAL CHARACTERl2ATION suits.

FOR THE FIFTH HEAVY SECTON STEEL TECHNOLOGY (HSST) IR- NUREG/CR4010: SPECIFICATION OF A HUMAN REUABILITY DATA RADIATION SER Ei " Neutron Eronsure Parametersy BANK FOR CONDUCTING HRA SEGMENTS OF PRAS FOR NUCLE-NUREG/CR4086: TENSlLE PROPERTIES OF IRRADIATED NUCLEAR AR POWER PLANTS.

GRADE PRESSURC VESSEL WELDS FOR THE THIRD HSST IRRA. NUREG/CR4044: TRAC PF1 LOCA CALCULATIONS USING FINE-DIATION SERIES. NODE AND COARSE-NODE INPUT MODELS.

NUREG/CR4092: ORNL CHARACTERIZATION OF HEAVY SECTON NUREG/CR4064: STRUCTURAL RESPONSE OF LARGE PENETRA-STEEL TECHNOLOGY PROGRAM PLATES 01,02.AND 03 TONS AND CLOSURES FOR CONTAINMENT VESSELS SUBJECTED NUREG/CR4105: AN ASSESSMENT OF THERMAL GRADfENT TUBE TO LOADINGS BEYOND DESIGN BASIS.

RESULTS FROM THE Hi SERIES OF FISSION PRODUCT RELEASE NUREG/CR4091: THE EFFECT OF ALTERNATIVE AGING AND ACCl-TESTS. DENT SIMULATIONS ON POLYMER PROPERTIES.

82 Contractor Index NUREG/CR4095: TEST SERIES 2. SEISMIC-FRAGILITY TESTS OF ST. LOUIS UNIV., ST. LOUIS, MO NATURALLY AGED CLASS 1E EXIDE FHC 19 8ATTERY CELLS NUREG/CR-4226 NEW MADRID SEISMOTECTONIC STUDY. Activities NUREG/CR-4096. TEST SERIES 3 SEISMIC FRAGILITY TESTS OF Dunng Fiscal Year 1983 NATURALLY-AGED CLASS 1E C&D LCU 13 BATTERY CELLS NUREG/CR-4097: TEST SERIES 4 SEISMIC-FRAGtLITY TESTS OF STEVENSON & ASSOCIATES NATURALLY AGED EXfDE EMP-13 BATTERY CELLS NUREG-1061 VO2 ADO REPORT OF THE U S NUCLEAR REGULA-NUREG/CR-4147. THE EFFECT OF ENVIRONMENTAL STRESS ON TORY COMMISSION PtPING REV!EW COMMdTE E. Volume 2

^ 9 " "* ' "

NUR G/CR-415 AC- 1 MOD 1 INDEPENDENT ASSESSMENT. NORTHWESTERN UNIVERSITY PERFORATED-PLATE

1. • ** **"^ ***9' ^ Y CCFL TESTS.

NUREG/CR-4169 AN APPROACH TO TREATING RAD 60NUCLIDE TEXAS, UNIV. OF, EL PASO, TX NURE / -4197 SAFET GOAL SEN I V YSU ES NUREG/CR-3174 V02 GEOPHYSICAL-GEOLOGICAL STUDIES OF NUREG/CR-4199: A DEMONSTRATION UNCERTAINTY / SENSITIVITY POSSIBLE EXTENSIONS OF .HE NEW MADRID FAULT ANALYSIS USING THE HEALTH AND ECONOMIC CONSEQUENCE ZONE. Annual Report For 1983 MODEL CRAC2 NUREG/CR-4264I INVESTIGATION ON HIGH-EFFICIENCY PARTICU- WESTINGHOUSE ELECTRIC CORF.

NUREG/CR-3455 A COMPARISON OF IODINE. KRYPTON,AND XENON LATE A!R F!LTER PLL*GG!NG BY COMBUSTION AEROSOLS RETENTION EFFICIENCIES FOR VAROUS SILVER LOADED AD-SCIENCE APPLICATIONS INTERNATIONAL COHP.,(FORMERLY SORPTION MEDIA SCIENCE APPLICATIONS NUREG/CR-3905 V01 R1: SEQUENCE CODING AND SEARCH WISCONSIN, UNIV. OF, M ADISON, WI SYSTEM FOF1 LICENSEE EVENT REPORTS User's Guide NUREG/CR4131- INVESTIGATION OF ALTERNATIVE MEANS TO AC-NUREG/CR-4101. ASSAY OF LONG-LIVED RADIONUCLIDES IN LOW- COMPLISH THE GOALS OF BTENNIAL ION CHAMBER CAllBRA.

LEVEL WASTES FROM POWER REACTORS. TION.

l Licensed Facility index This index lists the facilities that were the Docket number and followed by the report i

subject of MRC staff or contractor reports. number. If further information is needed, The facility names are arranged in alphabet- refer to the main citation by the NUREG ical crder. They are preceded by their number.

50424 AMn W Vogne Nucies Piant. Unt 1, Georpa NUAEG 1t37 2 353 Lanench Genwa$no Staton. Unt 2. PWMpNa NUALG4991 SO4 Powa Co Decinc Co

% 425 ANm W Vogne Nuclear Plant. Unt 2, Georpa NUAEG1137 S 410 Ane We Ptwt Nurks Stanon. Unt 2, N+,ma NUAEG 1047 S01 Power Co Mohawk Poeer Corp

% 325 Brunswo Steam Elecirc Ptart, Und 1. Caohna NUAEG/CR 4245 50 410 Nw Wie Fant Nucles Stahon. Und 2. Nepa NUREG-1065 Power & Ugnt Co Vonawn Pows Corp S 324 Brunswct Eteam Enctnc Plant und 2. Caonna NUAEG/CR-4245 50 269 Oconee Nucies Staton. Umt 1. D@e Power Co NUAEG,CD 3703 Pm & Lyd Co 50 269 C:enee NucMar Stanon. Umt 1. D@e Pomw Co NUAEG CR 4140

% 317 CaNort Chtts Nucien Power PM Und t NUREG/CR 4t09 STN 9528 Pah) verde Nucks Statm Und t. Ar. zona NUAEG4657 S08 Ba!bmare Gas & Emetnc  % Serace Co 50 445 STN 50 528 Psio vede Pkten Staton, Und 1, Antona NUREG1133 Comancne Peen Steare Electnc Stanon. Und 1 50 445 Comanceo Peak Steam Docts Stanon.NUAEG4797 Urd 1 NUPEG4797 S1 510, M*c he Co

% 446 Comancne Pese Spam Enctnc Stanon, Umt 2 NUAEG4797 S10 STN %529 Ia 1 berde Nudear Stanon. Ure 2. Anzona NUAEG 0057 S06

% 446 Comancne Peak Steam Elecinc Staton, Umt 2 NUAEG4797 Sit M*c Sece Co.

% 275 Outzo Canyon Nucies Power Plant, Und 1, NUAEG4675 S28 STN %530 Palo Verde Nudes Staton. Uni 3. Anzona NJAEG4857 50a Pacmc Gas & Doctnc Co N*c Sews Co.

% 275 Outso Cayon Nucinar Poww Plant. Und 1 NUAEG4675 S30 W440 Pm kwar Pcwe Piard. Und t he NUAEG4687 W Pacec Gas & Elecine Co Doctnc mummahng C

% 275 Datze Canyon Nucwar Pteer Piant. Und 1 NUAEG4675 S31 2 44i Perfy Nucker Poww Plant. Und 2. Geweend NUREG 0887 S#

PacMc Gas & Oscinc Co Electnc murmnanng C

% 323 Datso Caryon Nuclear Power PW Und 2. 206 San Onoke Nuces Statort Und 1. Scutnem NURE G 0829 CHF T NUAEG0675 S26 Pacec Gas & Decinc Co orma b Co W323 Datso Cayon Nucies Poem Piant. Uma 2.  % 400 Sheson Hans Nucles Power Plant. Und 1 NUREG1038 502 NOREGI475 S30

% 323 Piant. Unt 2. NUAEGW75 S31 W396 or t

% 323 tem P! ant, Und 2. NUAEG 1132 p 3 ParAc Gas & EWcinc Co 70-1151 Weshnghouse wtrc Corp , PWisburgn. PA, NUAEG t I t 8 W26 t H 8 Rotw' son Piant, Und 2. Cachna Power and NUAEG!CR 3977 STN 50-482 Won Creek Generatog Staban, kansas Gas & NUAEG088I Sf6 LVs Co Eier1nc

% 352 tenero Generahng Stahon. Und t. PWiespria NUAEG 0991 SO4 STN % 482 Wow Creek Generatng StaSon. Wansas Gas & NUAEG 1136 EmcInc Co Oecinc t

83

senC FOnes $35 U S 8eUCLE.m AEGut. tom v COeIMetSION ' ak#omt mv.etM rass sw er r,0C e.s ve 4e . .f ears hk'd,'- BIBLIOGRAPHIC DATA SHEET NUREG-0304 SH th5fRUCTiohs ON Twg mEvan5E Vol. 10, No. 2 g-

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Regulatory and Technical Reports Compilation (Abstract I ex Journal) for Second Quarter 1985 April - June

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This journal lists all fonnal reportsjinkhe NUREG series prepared by the NRC staff and contractors, as well as proceedings ofvonferences and workshops. The entries in the compilation are indexed for access by thle and abstract, contractor report number, personal author, subject, NRC organization, c tractor, and licensed facility.

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